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Attachment 3, WCAP-17441-NP, Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation
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Attachment 3 WCAP-17441-NP Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation

Westinghouse Non-Proprietary Class 3 WCAP-17441-NP October 20)11 Revision 0 Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17441-NP Revision 0 Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation F. A. Alpan*

Radiation Engineering and Analysis B. A. Rosier*

Aging Management and License Renewal Services October 2011 Reviewers: A. E. Freed*

Aging Management and License Renewal Services B. W. Amiri*

Radiation Engineering and Analysis Approved: M. G. Semmler*, Acting Manager Aging Management and License Renewal Services

Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

Westinghouse Non-Proprietary Class 3 RECORD OF REVISION Revision 0: Original Issue WCAP-17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 iii TABLE OF CONTENTS L IS T O F TA B L ES ....................................................................................................................................... iv L IS T OF F IG U RE S ..................................................................................................................................... vi E X E C U TIVE SU M M A RY ......................................................................................................................... vii I IN T R OD U C T IO N ........................................................................................................................ 1-1 2 CALCULATED NEUTRON FLUENCE ..................................................................................... 2-1 2.1 IN TR O DU C T ION ........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS ...................................................................... 2-1 2.3 CALCULATIONAL UNCERTAINTIES ........................................................................ 2-3 3 FRACTURE TOUGHNESS PROPERTIES ................................................................................. 3-1 4 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ................ 4-1 4.1 OVERALL APPROACH .................................................... 4-1 4.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE D E VE L O P M E N T ............................................................................................................ 4-1 4.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS ........................................... 4-5 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .......................................... 5-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ................... 6-1 7 RE F E RE N C E S ............................................................................................................................. 7-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (Ki) ................................ A-I WCAP- 17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 2-1 Pressure Vessel M aterial Locations .................................................................................. 2-4 Table 2-2 Seabrook Unit I Calculated Neutron Fluence Projections at the Reactor Vessel Clad/Base M etal Interface at 22, 28, 36, and 55 EFPY ..................................................................... 2-5 Table 2-3 Seabrook Unit I Calculated Neutron Fluence at the Reactor Vessel Clad/Base Metal Interface for Cycles I through 14 and Future Projections ............................................... 2-6 Table 2-4 C alculational U ncertainties .............................................................................................. 2-6 Table 3-1 Summary of the Best-Estimate Cu and Ni Weight Percents and Initial RTNDT Values for the Seabrook Unit I Reactor Vessel M aterials ................................................................. 3-2 Table 3-2 Summary of the Seabrook Unit I Reactor Vessel Material Chemistry Factor Values per R egulatory G uide 1.99, R evision 2 .................................................................................. 3-4 Table 5-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Seabrook Unit I Reactor Vessel Materials at 36 EFPY ....................................... 5-2 Table 5-2 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Seabrook Unit I Reactor Vessel Materials at 55 EFPY ............................................. 5-3 Table 5-3 Adjusted Reference Temperature Evaluation for the Seabrook Unit I Reactor Vessel Beltline Materials through 36 EFPY at the 1/4T Location .............................................. 5-4 Table 5-4 Adjusted Reference Temperature Evaluation for the Seabrook Unit I Reactor Vessel Beltline Materials through 36 EFPY at the 3/4T Location .............................................. 5-5 Table 5-5 Adjusted Reference Temperature Evaluation for the Seabrook Unit I Reactor Vessel Beltline Materials through 55 EFPY at the 1/4T Location .............................................. 5-6 Table 5-6 Adjusted Reference Temperature Evaluation for the Seabrook Unit I Reactor Vessel Beltline Materials through 55 EFPY at the 3/4T Location .............................................. 5-8 Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Seabrook Unit I Heatup and Cooldown Curves at 36 and 55 EFPY ........................................................ 5-10 Table 6-1 Seabrook Unit 1 36 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ Ki, w/o Flange Notch, and w/o Margins for Instrum entation E rrors) .................................................................................................... 6-5 Table 6-2 Seabrook Unit 1 36 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/o Flange Notch, and w/o Margins for Instrum entation E rrors) ................................................................................................... 6-.7 Table 6-3 Seabrook Unit 1 55 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1,, w/o Flange Notch, and w/o Margins for Instrum entation E rrors) .................................................................................................. 6-10 Table 6-4 Seabrook Unit 1 55 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1,, w/o Flange Notch, and w/o Margins for Instrum entation E rrors) .................................................................................................. 6-12 WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 V Table A-I K1, Values for Seabrook Unit 1 at 36 and 55 EFPY 100°F/hr Heatup Curves (w/o Flange Requirements and w/o Margins for Instrument Errors) .................................................. A-2 Table A-2 KI, Values for Seabrook Unit I at 36 and 55 EFPY 100°F/hr Cooldown Curves (w/o Flange Requirements and w/o Margins for Instrument Errors) ...................................... A-3 WCAP- 17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 vi LIST OF FIGURES Figure 2-1 Seabrook Unit 1 r-0 Reactor Geometry at the Core Mid-plane - 12.5' Neutron Pad C onfi guration ................................................................................................................... 2-7 Figure 2-2 Seabrook Unit I r-0 Reactor Geometry at the Core Mid-plane - 20.00 Neutron Pad C o nfig uratio n ................................................................................................................... 2-8 Figure 2-3 Seabrook Unit I r-0 Reactor Geometry at the Core Mid-plane - 22.50 Neutron Pad C o nfi gu ratio n ................................................................................................................... 2 --9 Figure 2-4 Seabrook Unit I r-z Reactor Geom etry .......................................................................... 2-10 Figure 6-1 Seabrook Unit I Reactor Coolant System Heatup Limitations (Heatup Rates of 80 and 100 0F/hr) Applicable for 36 EFPY (without Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G M ethodology (w/ K ij ...................................................................................................... 6-3 Figure 6-2 Seabrook Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, -80, and -100°F/hr) Applicable for 36 EFPY (without Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G M ethodology (w/K,,) ............................................................ 6-4 Figure 6-3 Seabrook Unit I Reactor Coolant System Heatup Limitations (Heatup Rates of 80 and I00°F/hr) Applicable for 55 EFPY (without Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G M ethodology (w/ K i,) ...................................................................................................... 6-8 Figure 6-4 Seabrook Unit I Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, -80, and -100°F/hr) Applicable for 55 EFPY (without Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G M ethodology (w/Klc) ............................................................ 6-9 WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 vii EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure-temperature (P-T) limit curves for normal operation of the Seabrook Unit I reactor vessel. The heatup and cooldown P-T limit curves were generated using the limiting Adjusted Reference Temperature (ART) values for Seabrook Unit 1. The limiting ART values were those of Lower Shell Plate RI 808-1 (without surveillance data) at both 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations. The P-T limit curves were generated using the K1, methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code, Section XI, Appendix G.

The P-T limit curves were generated for 36 and 55 Effective Full Power Years (EFPY) using heatup rates of 80 and 100°F/hr, and cooldown rates of 0, -20, -40, -60, -80 and -100 0 F/hr. The curves were developed without flange requirements and without margins for instrumentation errors. They can be found in Figures 6-1 through 6-4.

Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 36 and 55 EFPY.

October 2011 WCAP- 17441 -NP WCAP-17441I-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION Heatup and cooldown P-T 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 (RTNDT( U)). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The U.S. Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 [Ref. 1]. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values (RTNDT(U) + ARTNoT + margins for uncertainties) at the 1/4T 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 P-T limit curves documented in this report were generated using the most limiting ART values and the NRC-approved methodology documented in WCAP-14040-A, Revision 4

[Ref. 2]. Specifically, the K1, methodology of the 1998 through the 2000 Addenda Edition of ASME Code, Section XI, Appendix G [Ref. 3] was used.

The calculated ART values for 36 and 55 EFPY are documented in Tables 5-3 through 5-6 of this report.

The fluence projections used in calculation of the ART values are provided in Section 2 of this report.

The purpose of this report is to present the calculations and the development of the Seabrook Unit I heatup and cooldown P-T limit curves for 36 and 55 EFPY. This report documents the calculated ART values and the development of the P-T limit curves for normal operation. The P-T limit curves herein were generated without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [Ref. 4] have been eliminated from the P-T limit curves per the technical justification provided in WCAP-17444 [Ref. 5].

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-1 2 CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

A discrete ordinates SN transport analysis was performed for the Seabrook Unit I reactor to determine the neutron radiation environment within the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant- and fuel-cycle-specific basis. An evaluation of the most recent dosimetry sensor set from Capsule V, withdrawn at the end of the tenth plant operating cycle, is provided in Reference 6. The dosimetry analysis documented in Reference 6 showed that the +/-20% (1])

acceptance criteria specified in Regulatory Guide 1.190 [Ref. 7] is met. The results of this analysis are consistent with those of Reference 6 within the uncertainty of the methodology. Therefore, the acceptance criterion continues to be met. The validated calculations form the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 EFPY.

All of the calculations described in this section were based on nuclear cross-section data derived from ENDF/B-VI.3 and made use of the latest available calculational tools. Furthennore, the neutron transport evaluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 7]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-14040-A, Revision 4 [Ref. 2].

2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the Seabrook Unit 1 reactor vessel, a series of fuel-cycle-specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:

q'(r, 0, z) = (p(r,0)

go(r, z)

(p(r) where co(r.O,z) is the synthesized three-dimensional neutron flux distribution, CO(r,O) is the transport solution in r,0 geometry, 9o(r, z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and yp(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 Seabrook Unit 1.

For the Seabrook Unit 1 transport calculations, the r,0 models depicted in Figures 2-1 through 2-3 were utilized since, with the exception of the neutron pads, the reactor is octant symmetric. These rO models include the core, the reactor internals, the neutron pads - including explicit representations of an octant not containing surveillance capsules and octants with surveillance capsules at 20.00 and 22.50 - the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. 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 WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 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 proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the r,0 calculations was set at a value of 0.001.

The r,z model used for the Seabrook Unit I calculations is shown in Figure 2-4 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 approximately six feet below to five feet above the active fuel. 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 r,0 calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

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

The data utilized for the core power. distributions in plant-specific transport analyses included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions. This information was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle-averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code, version 3.2 [Ref. 8], and the BUGLE-96 cross-section library [Ref. 9]. 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 Ps Legendre expansion and angular discretization was modeled with an S1 6 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.

In Table 2-1, locations of the lower shell to lower vessel head circumferential weld, lower shell longitudinal welds, lower shell plates, intermediate shell to lower shell circumferential weld, intermediate shell longitudinal welds, intermediate shell plates, upper shell to intermediate shell circumferential weld, upper shell longitudinal welds, upper shell plates and outlet/inlet nozzle to upper shell welds are given.

The axial position of each material is indexed to z =0.0 cm, which corresponds to the mid-plane of the active fuel stack.

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-3 Selected results from the neutron transport analyses are provided in Tables 2-2 and 2-3. In Table 2-2, calculated fast neutron (E > 1.0 MeV) fluence for reactor vessel materials, on the pressure vessel clad/base metal interface, is provided at future projections to 22, 28, 36 and 55 EFPY. Cycle-specific calculations were performed for Cycles I to 14, where a core thermal power of 3411 MWt was used in Cycles 1-10, 3587 MWt was used in Cycle 11 and 3648 MWt was used in Cycles 12-14. The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycles 11, 12, and 13 were representative of future plant operation. In Table 2-3, calculated fast neutron (E > 1.0 MeV) fluence on the pressure vessel clad/base metal interface is provided for Cycles I through 14 and future projections, at various azimuthal locations.

2.3 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Seabrook Unit I reactor pressure vessel materials 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 benchmnark 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 Seabrook 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 Seabrook Unit I analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Seabrook Unit I 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.

Table 2-4 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 2. The net WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-4 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 given in AppendixA of Reference 6 support these uncertainty assessments for Seabrook Unit 1.

Table 2-1 Pressure Vessel Material Locations r-0 Neutron Pad Material Axial Location azimh Configuration (cm) Location used in Exposure (0) Calculations L ower shell to lower vessel head -310.59 0 to 360 12.5' neutron pad circumferential weld Lower shell longitudinal weld 1 90 22.50 neutron pad Lower shell longitudinal weld 2 -310.59 to -36.27 210 12.50 neutron pad Lower shell longitudinal weld 3 330 12.50 neutron pad Lower shell plate 1 90 to 210 12.50 neutron pad Lower shell plate 2 -310.59 to -36.27 210 to 330 12.50 neutron pad Lower shell plate 3 330 to 90 12.50 neutron pad Intermediate shell to lower shell -36.27 0 to 360 12.5' neutron pad circumferential weld Intermediate shell longitudinal weld 1 0 12.5' neutron pad Intermediate shell longitudinal weld 2 -36.27 to 238.05 120 20.0' neutron pad Intermediate shell longitudinal weld 3 240 22.50 neutron pad Intermediate shell plate 1 0 to 120 12.5' neutron pad Intermediate shell plate 2 -36.27 to 238.05 120 to 240 12.50 neutron pad Intermediate shell plate 3 240 to 360 12.50 neutron pad Upper shell to intermediate shell 238.05 0 to 360 12.50neutron pad circumferential weld Upper shell longitudinal weld 1 42 12.50 neutron pad Upper shell longitudinal weld 2 238.05 to 489.83 160 12.50 neutron pad Upper shell longitudinal weld 3 278 20.00 neutron pad Upper shell plate 1 42 to 160 12.50 neutron pad Upper shell plate 2 238.05 to 489.83 160 to 278 12.50 neutron pad Upper shell plate 3 278 to 42 12.50 neutron pad Outlet Nozzle to Upper Shell Weld"a) 1 22 12.50 neutron pad Outlet Nozzle Outlet Nozzle to Upper Shell to Upper Shell Weld Weldýa)a)23 158 12.50 neutron pad 202 12.5° neutron pad Outlet Nozzle to Upper Shell Weld a) 4 338 12.5' neutron pad Inlet Nozzle to Upper Shell Weld(")1 67 22.50 neutron pad Inlet Nozzle to Upper Shell Weld'a) 2 113 12.50 neutron pad Inlet Nozzle to Upper Shell Weld(') 3 271.35 247 22.50 neutron pad Inlet Nozzle to Upper Shell Weld(') 4 293 12.50 neutron pad Note for Table 2-1:

(a) Lowest extent WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-5 Table 2-2 Seabrook Unit 1 Calculated Neutron Fluence Projections at the Reactor Vessel Clad/Base Metal Interface at 22, 28, 36, and 55 EFPY Fluence(a)

Reactor Vessel Material (n/cm 2, E > 1.0 MeV) 22 EFPY 28 EFPY 36 EFPY 55 EFPY Outlet Nozzle to Upper Shell Welds (Lowest Extent) 2.29 x 1016 2.91 X 1016 3.74 x 1016 5.70 x 1016 1, 2, 3, and 4 Inlet Nozzle to Upper Shell Welds (Lowest Extent) I and 3 3.11 x 1016 3.95 x 1016 5.07 x 1016 7.72 x 1016 Inlet Nozzle to Upper Shell Welds (Lowest Extent) 2 and 4 4.12 x 1016 5.23 x 1016 6.70 x 1016 1.02 x 1017 Upper Shell Plates 3.00 x 1017 3.79 x 1017 4.84 x 10"' 7.35 x 1017 Intermediate Shell Plates 1.23 x 1019 1.56 x 1019 2.00 x 10'9 3.05 x 1019 Lower Shell Plates 1.23 x 1019 1.56 x 1019 2.00 x 10' 9 3.05 x 1019 Upper Shell Longitudinal Weld 1:101-122 A(b" 2.79 x 1017 3.50 x 1017 4.45 x 1017 6.70 x 1017 Upper Shell Longitudinal Weld 2:101-122 B1b) 2.88 x 1017 3.65 x 1017 4.68 x 1017 7.12 x 1017 Upper Shell Longitudinal Weld 3:101-122 C"b) 1.99 x 1017 2.53 x 1017 3.26 x 10I 4.97 x 1017 Upper Shell to Intermediate Shell Circumferential Weld 3.00 x 1017 3.79 x 1017 4.84 x 10i" 7.35 x 1017 Intermediate Shell Longitudinal Weld 1:101-124 A(b) 7.06 x 1018 9.00 x 1018 1.16 x 1019 1.78 x 10'9 Intermediate Shell Longitudinal Weld 2:101-124 B1b3 7.14 x 10"8 9.04 x 1018 1.16 x 10"9 1.76 x 1019 Intermediate Shell Longitudinal Weld 3:101-124 Cfbl 6.44 x 1018 8.14 x 1018 1.04 x 10'9 1.58 x 10'9 Intermediate Shell to Lower Shell Circumferential Weld 1.23 x 1019 1.56 x 10'9 1.99 x 10'9 3.03 x 1019 Lower Shell Longitudinal Weld 1:101-142 A"b) 7.05 x 10"8 8.99 x 1018 1.16 x 10"9 1.77 x 1019 Lower Shell Longitudinal Welds 2&3: 101-142 B&C"b' 1.19 x 10'9 1.51 x 1019 1.93 x 1019 2.93 x 1019 Lower Shell to Lower Vessel Head Circumferential Weld 1.02 x 10"5 1.28 x 101" 1.63 x 1015 2.46 x 1015 Notes for Table 2-2:

(a) Extended beltline materials are currently interpreted to be the reactor vessel materials that will be exposed to a neutron fluence greater than or equal to I x 1017 n/cm 2 (E > 1.0 MeV). Only the materials that are projected to experience a fluence value of at least I x 10"7 n/cm 2 (E > 1.0 MeV) will be included in the subsequent evaluations contained within this report.

(b) The fluence value at each individual weld location (A, B, C) was reported here for documentation purposes; however, the maximum fluence value across all three longitudinal welds in each reactor vessel shell will be used as the bounding fluence value in the subsequent calculations contained within this report.

WCAP- 17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-6 Table 2-3 Seabrook Unit 1 Calculated Neutron Fluence at the Reactor Vessel Clad/Base Metal Interface for Cycles 1 through 14 and Future Projections Fluence (n/cm 2, E > 1.0 MeV) Maximum Azimuthal Cycle Cumulative Cycle Fluence Location of ED TeFTim 00 150 300 450 (n/cm 2, E > Maximum (EFPY) (EFPY) 1.0 MeV) Fluence (0) 7 1 0.91 0.91 4.04 x 1017 6.16 x 1017 7.25 x 101 7.92 x 1017 7.92 x 1017 45 2 0.87 1.78 7.58 x 1017 1.15 x 108 1.27 x 10" 1.35 x 10" 1.36 x 1018 22 3 1.21 2.99 8 1.21 x 1018 1.80x 118 2.04x 10" 2.20x 1018 2.20x 1018 45 4 8 1.21 4.20 1.63 x 10" 2.38 x 101 2.63 x 108 2.78 x 101 2.78 x 10" 45 5 1.37 5.57 1.99 x 10" 2.94 x 10"8 3.31 x 1018 3.46 x 1018 3.46 x 10" 45 6 8 1.49 7.06 2.44 x 10 3.59 x 1018 4.04 x 1018 4.17 x 1018 4.18 x 1018 23 7 1.41 8.47 2.83 x 1018 4.17 x 10"8 4.81 x 1018 4.95 x 10" 4.95 x 1018 28 8 1.20 9.67 3.17 x 10" 4.66 x 10" 5.40 x 1018 5.57 x 10" 5.57 x 108 45 9 1.34 11.01 3.55 x 1018 5.24 x 10s 6.10 x 1018 6.31 x 1018 6.31 x 1018 45 10 1.40 12.41 3.95 x 1018 5.83 x 1018 6.82 x 1018 7.07 x 1018 7.07 x 1018 45 11 1.40 13.81 4.36 x 108 6.44 x 1018 7.53 x 1018 7.80 x 101I 7.80 x 1018 45 12 1.34 15.15 4.80 x 1018 7.10 x 1018 8.27 x 1018 8.53 x 1018 8.53 x 1018 45 13 1.39 16.54 5.27 x 10 7.77 x 101" 8.98 x 1018 9.22 x 1018 9.25 x 1018 28 14 1.31 17.85 5.71 x 1018 8.40 x 018 9.74 x 1018 9.99 x 10t" 1.00 x 1019 28

... ... 22.00 7.07 x 1018 1.04 x 1019 1.19 x I019 1.22 x 1019 1.23 x 10 _9 -_-_ -

-.-.- 28.00 9.03 x 1018 1.33 x 1019 1.51 x 1019 1.54 x 1019 1.57.x 1019 - - -

9

- - .36.00 1.16 x 1019 1.71 x 1019 1.94 x 10' 1.96 x 1019 2.01 x 1019 ---

-.-.- 42.00 1.36 x 1019 1.99 x 1019 2.26 x 1019 2.28 x 10'9 2.35 x 1019 - - -

-.-.- 48.00 1.56 x 1019 2.28 x 1019 2.58 x 10'9 2.59 x 1019 2.68 x 109 - - -

... 55.00 1.79 x 1019 2.62 x 1019 2.95 x 1019 2.96 x 1019 3.07 x 1019 - - -

... 60.00 1.95 x 1019 2.86 x 10'9 3.22 x 1019 3.23 x 1019 3.35 x 1019 - - -

Table 2-4 Calculational Uncertainties Description Uncertainty Capsule Vessel IR PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-7 Figure 2-1 Seabrook Unit 1 r-0 Reactor Geometry at the Core Mid-plane - 12.50 Neutron Pad Configuration Seabrook Unit 1 - 12.5 Degree Neutron Pad R,T Model Meshes: 183R, 998

-core - Sim. Swe

-M3 M Douuw -M ftwm Yawne

- Q~ W~our~ily Wr R

w The stainless steel regions include the core baffle, core barrel, thermal shield, and vessel clad.

4-t

'0.0 SR WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-8 Figure 2-2 Seabrook Unit 1 r-0 Reactor Geometry at the Core Mid-plane - 20.00 Neutron Pad Configuration Seabrook Unit 1 - 20.0 Degree Neutron Pad R,T Model Meshes: 183R, 998

-Cu.. M SI*. SW M D..v~u M Cubur SIf

- R.'u C.1,7Ai J Cur.tuN 0

The stainless steel regions include the core baffle, core barrel, neutron pad, surveillance capsule holder, and vessel clad The carbon steel regions include the surveillancecapsule specimens and pressure vessel.

1'4 r'-

60 U'

[cm]

WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-9 Figure 2-3 Seabrook Unit 1 r-0 Reactor Geometry at the Core Mid-plane - 22.50 Neutron Pad Configuration Seobrook Unit 1 - 22.5 Degree Neutron Pad R,T Model Meshes: 183R, 998

-Core - taiien se. "pas

- mm M Carbon SIW

-Reactor Cm~it Ai ~hRoO 0

The stainless steel regions include the core baffle, core barrel, neutron pad, surveillance capsule holder, and vessel clad The carbon steel regions include the surveillance capsule specimens andpressure vessel.

as-0 5.2 0.0 68.6 137.2 205.8 274.4 34ý3. R

[cm]

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 2-10 Figure 2-4 Seabrook Unit 1 r-z Reactor Geometry Seabrook Unit 1 - R,Z Model Meshes: 153R,188Z The stainlesssteel regions include the core baffle, formers, core barrel, neutronpad,pressure vessel clad, and upper core plate B, which is located below outlet plenum B.

Seobrook Unit 1 - R,Z Model Meshes: 153R,188Z

- W*hWr -4W tow *.I

- T. fo

%WA -.

C. ft. A humC W-6up C- AW

- ý C. %%g LMA

'b.0 68.6 137.2 205.8 274.4 343.0 R

[cm]

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The fracture toughness properties of the ferritic materials in the Seabrook Unit I reactor vessel are presented in Table 3-1.

The Regulatory Guide 1.99, Revision 2, methodology used to develop the heatup and cooldown P-T limit curves documented in this report is the same as that documented in WCAP-14040-A, Revision 4 [Ref. 2].

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 Regulatory Guide along with the best-estimate copper and nickel weight percents, which are presented in Table 3-1. Position 2.1 uses the surveillance capsule data from all capsules withdrawn and tested to date. Table 3-2 summarizes the Position 1.1 and 2.1 CFs determined for the Seabrook Unit I reactor vessel materials.

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of the Best-Estimate Cu and Ni Weight Percents and Initial RTNDT Values for the Seabrook Unit 1 Reactor Vessel Materials Fracture Chemical Composition(a) Toughness Reactor Vessel Material(a" Property(a)

Cu Ni Initial RTNDT(b)

(wt. %) (wt. %) (*F)

Inlet Nozzle RI804- 1c) 0.10 0.85 0 Inlet Nozzle R1804-2(c) 0.09 0.89 -20 Inlet Nozzle RI 804-3(c) 0.08 0.88 -20 Inlet Nozzle R1804-4(c) 0.10 0.83 -20 Upper Shell (US) Plate R1807-1l) 0.08 0.60 30 US Plate RI18 0 7 -2(d) 0.09 0.61 30 US Plate R1807-3(d) 0.06 0.67 10 Intermediate Shell (IS) Plate R1806-1 0.045 0.61 40 IS Plate R1806-2 0.06 0.64 0 IS Plate RI 806-3 0.075 0.63 10 Lower Shell (LS) Plate RI 808-1 0.06 0.58 40 LS Plate R1808-2 0.06 0.58 10 LS Plate R1808-3 0.07 0.59 40 Surveillance Plate R1808-3 Inlet Nozzle to Upper Shell Welds(e) 0.35 1.0 -56(9 US Longitudinal (Long) Welds 101-122A,B,C (Heat # 8 6 9 9 8)(d0 US to IS Circumferential (Circ) Weld 103-121 0.045 0.06 -56(g)

(Heat # 9 0 12 8 )da IS Long Welds 101-124A,B,C (Heat # 4P6052)(0 0.047 0.049 -60 IS to LS Circ Weld 101-171 (Heat # 4P6052)7 0.047 0.049 -60 LS Long Welds 101-142A,B,C (Heat # 4P6052)jf) 0.047 0.049 -60 Seabrook Unit I Surveillance Weld 0.047 0.049 (Heat # 4 P 6 0 5 2)(h) 0.047_ 0.049 -_-_-

(See next page for notes)

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 3-3 Notes for Table 3-1 :

(a) All values are taken from WCAP-15745 [Ref. 10], unless otherwise noted.

(b) The initial RTNDT values are based on measured data, unless otherwise noted.

(c) Copper and initial RTNDT values for the Inlet Nozzles were taken from Table 5.3-3 of the Seabrook Unit 1 UFSAR [Ref. 11]. Note that these values are consistent with those documented in the Certified Material Test Report (CMTR). Nickel values for the Inlet Nozzles were obtained from the CMTR.

(d) Chemistry and Initial RTNDT values are consistent with the values documented in the Seabrook Unit I License Renewal Application (LRA) [Ref. 12].

(e) The Inlet Nozzle to Upper Shell Welds are comprised of Linde 0091 flux type welds. Generic values for Cu, Ni, and initial RTNDT are conservatively used for the Inlet Nozzle to Upper Shell Welds. The generic Cu and Ni values are taken from Regulatory Guide 1.99, Revision 2 [Ref. 1].

(f) The beltline weld seams were fabricated with weld wire Heat # 4P6052, Flux Type 0091, Flux Lot #

0145. The copper and nickel weight percents were taken from CE Reports NPSD-1039, Revision 2

[Ref. 13] and NPSD-1 119, Revision I [Ref. 14].

(g) Generic initial RTNDT value taken from 10 CFR 50.61 [Ref. 15] for Flux Type Linde 0091 welds.

(h) Values consistent with those documented in WCAP-16526-NP [Ref. 6] and the Seabrook Unit I LRA

[Ref. 12].

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 3-4 Table 3-2 Summary of the Seabrook Unit 1 Reactor Vessel Material Chemistry Factor Values per Regulatory Guide 1.99, Revision 2 Chemistry Factor (OF)

Reactor Vessel Material Position 1.1 Position 2.1 Inlet Nozzle R1804-1 67 - - -

Inlet Nozzle RI 804-2 58 ---

Inlet Nozzle RI 804-3 51 ---

Inlet Nozzle RI 804-4 67 - - -

US Plate RI807-1 51 - - -

US Plate R1807-2 58 -- -

US Plate R1807-3 37 ...

IS Plate R1806-1 28.5 ---

IS Plate RI1806-2 37 - - -

IS Plate RI1806-3 47.5 -- -

LS Plate R1808-1 37 - - -

LS Plate R1808-2 37 - - -

LS Plate R1808-3 44 45.0(a)

Inlet Nozzle to Upper Shell Welds 272 ---

US Long Welds 101-122A,B,C 38.7 - - -

US to IS Circ Weld 103-121 31.3 ---

IS Long Welds 101 -124A,B,C 30.7 30.0(a, IS to LS Circ Weld 101-171 30.7 30.0(a' LS Long Welds 101-142A,B,C 30.7 30.0(a)

Seabrook Unit I Surveillance Weld 30.7 ---

(Heat # 4P6052)

Note for Table 3-2:

(a) Per Appendix D of WCAP-16526-NP [Ref. 6], the Seabrook Unit 1 surveillance plate and weld data is deemed credible.

WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 4-1 4 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 4.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kj,, 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 Xl, Appendix G of the ASME Code [Ref. 3]. The KI, curve is given by the following equation:

Kc= 33 + 20.734

e[0 02 (T- RTNT)] (1)

where, Ki, (ksi'Jin.) = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This KI, 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 Class 1, SA-508-1, SA-508-2, and 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 < Kic (2)

where, Kim = stress intensity factor caused by membrane (pressure) stress Kit = stress intensity factor caused by the thermal gradients K1. reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNOT 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-1 7441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 4-1 Westinghouse Non-Proprietary Class 3 4-2 For membrane tension, the corresponding K, for the postulated defect is:

Kim mx (pRi/t) (3) where, Mm for an inside axial surface flaw is given by:

M = 1.85 for -/t <2, Mm = 0.92617 for.2 _< ft _<3.464, Mm = 3.21 for 1 > 3.464 and, Mm for an outside axial surface flaw is given by:

Mm = 1.77 for f-4 < 2, Mm = 0.893 U for 2*4 t -3.464, Mm = 3.09 for ,ft > 3.464 Similarly, Mm for an inside or an outside circumferential surface flaw is given by:

Mm = 0.89 for ft- <2, Mmn = 0.443ft for 2< It < 3.464, Mmn = 1.53 for it- > 3.464 Where:

p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in).

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

KIb = Mb

Maximum Stress, where Mb is two-thirds of Mm (4)

The maximum K, produced by radial thermal gradient for the postulated axial or circumferential inside surface defect of G-2120 is:

K], = 0.953x10- 3 x CR x t 25 (5) where CR is the cooldown rate in 'F/hr., or for a postulated axial or circumferential outside surface defect Kit = 0.753x10- 3 x HU x t25 (6) where HU is the heatup rate in °F/hr.

October 2011 WCAP-1 7441 -NP WCAP-17441I-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 4-3 The through-wall temperature difference associated with the maximum thermal K, can be determined from ASME Code, Section XI, Appendix G, Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code, Section XI, Appendix G, Fig. G-2214-2 for the mnaximum thermal K1 .

(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 axial or circumferential inside surface defect using the relationship:

Ki, = (1.0359Co + 0.6322C, + 0.4753C2+ 0.3855C 3)* (7) or similarly, Kit during heatup for a 1/4A-thickness outside axial or circumferential surface defect using the relationship:

Ki, = (1.043C0o+0.630C + 0.481C2 + 0.401C3)* * (8) where the coefficients C0, C 1, C 2 and C 3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

or(x) = Co+ C1(x/ a) + C2(x / a) 2 + C3(x / a)3 (9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in).

Note that Equations 3, 7, and 8 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 2] Section 2.6 (equations 2.6.2-4 and 2.6.3-1).

At any time during the heatup or cooldown transient, K1 c is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code, Section XI, paragraph G-2120), the appropriate value for RTNDT, and the reference fracture toughness curve (Equation 1). The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K11, 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 1/4T flaw of Appendix G to Section Xl of 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 vessel wall WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 4-4 because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable pressure-temperature curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.

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) across the vessel wall developed during cooldown results in a higher value of K1c at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in Kj, exceeds KI,, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the I/4T location and, therefore, allowable pressures could be lower 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 1/4T 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 inside 1/4T flaw during heatup is lower than the K1, for the flaw 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 third 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 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 least of the three values taken from the curves under consideration. The use of WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 4-5 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 HEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [Ref. 4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RTNDT by at least 120'F for nonnal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure. However, WCAP-17444 [Ref. 5] provides the technical basis for elimination of the flange requirements for the development of P-T limit curves for Seabrook Unit 1. Hence, the Seabrook Unit I heatup and cooldown limit curves will be generated without flange requirements.

WCAP-17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-1 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 (10)

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 Code [Ref. 16]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided 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:

ARTNDT = CF

f(0.28-o.lolog f) (11)

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.

f*depth xi = fsurface

e (-0.24x) (12) 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 11 to calculate the ART**T at the specific depth.

The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 2].

The fluence values for 36 and 55 EFPY were used for the development of P-T limit curves contained in this report. Tables 5-1 and 5-2 also contain the 1/4T and 3/4T calculated fluence values and fluence factors, per Regulatory Guide 1.99, Revision 2. The values in these tables will be used to calculate the 36 and 55 EFPY ART values for the Seabrook Unit I reactor vessel materials.

Margin is calculated as M = 2 ,f& + C2

The standard deviation for the initial RTNDT margin term (sT) is 0°F when the initial RTNDT is a measured value, and 170 F when a generic value is available. The standard deviation for the ARTNDT margin term, ca, is 17'F for plates or forgings when surveillance data is not used or is non-credible, and 8.57F (half the value) for plates or forgings when credible surveillance data is used. For welds, aA is equal to 28°F when surveillance capsule data is not used or is non-credible, and is 14'F (half the value) when credible surveillance capsule data is used. The value for cTaneed not exceed 0.5 times the mean value of ARTNDT.

Contained in Tables 5-3 through 5-6 are the 36 and 55 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Seabrook Unit I heatup and cooldown curves.

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-2 Table 5-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Seabrook Unit 1 Reactor Vessel Materials at 36 EFPY Surface 1/4T f 3/4T f Region Fluence, f (x 109 n/cm2FFF (x 1019 n 2 1/4T3/T E>109ne) E > 1.0 MeV) E > 1.0 MeV)

E > 1.0 MeV)

US Plates 0.0484 0.029 0.2146 0.010 0.1114 IS Plates 2.00 1.192 1.0489 0.423 0.7611 LS Plates 2.00 1.192 1.0489 0.423 0.7611 US Long Welds 0.0468 0.028 0.2104 0.010 0.1089 101-122A,B,C US to IS Circ Weld 103-121 0.0484 0.029 0.2146 0.010 0.1114 IS Long Welds 101 - 124A,B,C 1.16 0.691 0.8964 0.245 0.6193 IS to LS Circ Weld 101-171 1.99 1.186 1.0475 0.421 0.7597 LS Long Welds 101 -142A,B,C 1.93 1.150 1.0390 0.408 0.7515 WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-3 Table 5-2 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Seabrook Unit 1 Reactor Vessel Materials at 55 EFPY Surface 1/4T f 3/4T f Fluence, f 1019 2, 1/4T 3/4T Region (x 1019 n/cm2, FF FF E > 1.0 MeV) E > 1.0 MeV)

E>1.MeV)

Inlet Nozzles 0.0102(a) 0.006 0.0773 0.002 0.0349 US Plates 0.0735 0.044 0.2723 0.016 .0.1468 IS Plates 3.05 1.817 1.1639 0.645 0.8772 LS Plates 3.05 1.817 1.1639 0.645 0.8772 Inlet Nozzle to 0.0102 0.006 0.0773 0.002 0.0349 Upper Shell Welds US Long Welds 0.0712 0.042 0.2675 0.015 0.1438 101-122A,B,C US to IS Circ Weld 103-121 0.0735 0.044 0.2723 0.016 0.1468 IS Long Welds 101-124A,BC 1.78 1.061 1.0165 0.377 0.7298 IS to LS Circ Weld 101-171 3.03 1.805 1.1621 0.641 0.8753 LS Long Welds 101-142A,B,C 2.93 1.746 1.1532 0.620 0.8660 Note for Table 5-2:

(a) The limiting Inlet Nozzle to Upper Shell Weld fluence is conservatively applied to the Inlet Nozzles.

October 2011 WCAP- 17441 -NP WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-4 Table 5-3 Adjusted Reference Temperature Evaluation for the Seabrook Unit 1 Reactor Vessel Beltline Materials through 36 EFPY at the 1/4T Location f 2 1/4T RTNDT(U)(a) ARTNDT 71 (a) FA(C) M ART(d)

Rat1/4T Reactor Vessel Material (OF) E>1.0 MeV) FF (OF) (OF) (OF) (OF) (OF) (OF)

US Plate R1807-1 51 0.029 0.2146 30 10.9 0 5.5 10.9 52 US Plate R1807-2 58 0.029 0.2146 30 12.4 0 6.2 12.4 55 US Plate R1807-3 37 0.029 0.2146 10 7.9 0 4.0 7.9 26 IS Plate R1806-1 28.5 1.192 1.0489 40 29.9 0 14.9 29.9 100 IS Plate R1806-2 37 1.192 1.0489 0 38.8 0 17.0 34.0 73 IS Plate R1806-3 47.5 1.192 1.0489 10 49.8 0 17.0 34.0 94 LS Plate R1808-1 37 1.192 1.0489 40 38.8 0 17.0 34.0 113 LS Plate R1808-2 37 1.192 1.0489 10 38.8 0 17.0 34.0 83 LS Plate R1808-3 44 1.192 1.0489 40 46.2 0 17.0 34.0 120 Using Credible Surveillance Data 45.0 1.192 1.0489 *40 47.2 0 8.5 17.0 104 US Long Welds 101-122A.B,C 38.7 0.028 0.2104 -10 8.1 0 4.1 8.1 6 US to IS Circ Weld 103-121 31.3 0.029 0.2146 -56 N 6.7 1 7 (b) 3.4 34.7 -15 IS Long Welds 101-124A,B,C 30.7 0.691 0.8964 -60 27.5 0 13.8 27.5 -5

-- Using Credible Surveillance Data 30.0 0.691 0.8964 -60 26.9 0 13.4 26.9 -6 IS to LS Circ Weld 101-171 30.7 1.186 1.0475 -60I .-.............. ,..... .........32.2 0 16.1... . 32.2 4

-- Using Credible Surveillance Data 30.0 1.186 1.0475 -60 31.4 0 14.0 28.0 -1 LS Long Welds 101-142A,B,C 30.7 1.150 1.0390 -60 31.9 0 15.9 31.9 4

- Using Credible Surveillance Data 30.0 1.150 1.0390 -60 31.2 0 14.0 28.0 -1 Notes for Table 5-3:

(a) Initial RTNDT values are measured values, unless otherwise noted.

(b) Initial RTNDT value is generic; hence a, = 17'F.

(c) Per WCAP-16526-NP, Revision 0 [Ref. 6], the surveillance data of the plate and weld material was deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal GA = 17'F for Position 1.1 and, with credible surveillance data, uA = 8.5°F for Position 2.1; the weld metal GA = 28°F for Position 1.1 and, with credible surveillance data, aA = 14'F for Position 2. 1. However, aA need not exceed 0.5*ARTNDT.

(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART = RTNDT(U) + ARTNDT + Margin.

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-5 Table 5-4 Adjusted Reference Temperature Evaluation for the Seabrook Unit 1 Reactor Vessel Beltline Materials through 36 EFPY at the 3/4T Location 3/4T f 3/4T RTNDT(U)(a) ARTNDT Cai 1 (YA C) M ART(d)

Reactor Vessel Material (CF) (x 10 neCM 2 , FF (F) (OF) (OF) (OF)

M(F) (OF) 0F E>1.0 E MeV)______ (F) () (F) ()

US Plate R1807-1 51 0.010 0.1114 30 5.7 0 2.8 5.7 41 US Plate R1807-2 58 0.010 0.1114 30 6.5 0 3.2 6.5 43 US Plate R1807-3 37 0.010 0.1114 10 4.1 0 2.1 4.1 18 IS Plate R1806-1 28.5 0.423 0.7611 40 21.7 0 10.8 21.7 83 IS Plate R1806-2 37 0.423 0.7611 0 28.2 0 14.1 28.2 56 IS Plate R1806-3 47.5 0.423 0.7611 10 36.2 0 17.0 34.0 80 LS Plate RI1808-1 37 0.423 0.7611 40 28.2 0 14.1 28.2 96 LS Plate R1808-2 37 0.423 0.7611 10 28.2 0 14.1 28.2 66 LS Plate R1808-3 44 0.423 0.7611 40 33.5 0 16.7 33.5 107 Using Credible Surveillance Data 45.0 0.423 0.7611 40 34.2 0 8.5 17.0 91 US Long Welds 101-122A,B,C 38.7 0.010 0.1089 -10 4.2 0 2.1 4.2 US to IS Circ Weld 103-121 31.3 0.010 0.1114 -56(b) 3.5 1 7 (b) 1.7 34.2 -18 IS Long Welds 101-124A,B,C 30.7 0.245 0.6193 -60 19.0 0 9.5 19.0 -22 Using Credible Surveillance Data 30.0 0.245 0.6193 -60 18.6 0 9.3 18.6 -23 IS to LS Circ Weld 101-171 30.7 0.421 0.7597 -60 23.3 0 11.7 23.3 -13 Using Credible Surveillance Data 30.0 0.421 0.7597 -60 22.8 0 11.4 22.8 -14 LS.._ ..

Long.. . .......

Welds .............

. 0!.-.

_4..

10l1-1I42A,B,C .........................

330.7

.97..................

4.......

0.408 _

7._.!....

0.7515...............

%2-6

-60 .............

. ...... 23.1 .......................

................... 0 ....

.... 11.5

.... 23_.........

23.1 -14 Using Credible Surveillance Data 30.0 0.408 0.7515 -60 22.5 0 11.3 22.5 -15 Notes for Table 5-4:

(a) Initial RTNDT values are measured values, unless otherwise noted.

(b) Initial RTNDT value is generic; hence (71= 17°F.

(c) Per WCAP-16526-NP, Revision 0 [Ref. 6], the surveillance data of the plate and weld material was deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal (Ta= 177F for Position 1.1 and, with credible surveillance data., a = 8.5°F for Position 2.1; the weld metal a, = 287 for Position 1.1 and, with credible surveillance data, c7A = 14'F for Position 2.1. However, 0 Aneed not exceed 0.5*ARTNDT.

(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART = RTNDT(u) + ARTNDT + Margin.

WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-6 Table 5-5 Adjusted Reference Temperature Evaluation for the Seabrook Unit 1 Reactor Vessel Beltline Materials through 55 EFPY at the 1/4T Location 101/49T f 2 1/4T RTNDT(U)(a) ARTNDT C"I (a) aA(c) M ART(d)

Reactor Vessel Material (OF) xE> 1.0 MeV) FF (OF) (OF) (OF) (OF) (OF) (OF)

Inlet Nozzles 67 0.006 0.0773 0 5.2 0 2.6 5.2 10 US Plate R1807-1 51 0.044 0.2723 30 13.9 0 6.9 13.9 58 US Plate R1807-2 58 0.044 0.2723 30 15.8 0 7.9 15.8 62 US Plate R1807-3 37 0.044 0.2723 10 10.1 0 5.0 10.1 30 IS Plate R1806-1 28.5 1.817 1.1639 40 33.2 0 16.6 33.2 106 IS Plate R1806-2 37 1.817 1.1639 0 43.1 0 17.0 34.0 77 IS Plate R1806-3 47.5 1.817 1.1639 10 55.3 0 17.0 34.0 99 LS Plate R1808-1 37 1.817 1.1639 40 43.1 0 17.0 34.0 117 LS Plate R1808-2 37 1.817 1.1639 10 43.1 0 17.0 34.0 87 LS Plate R1808-3 44 1.817 1.1639 40 51.2 0 17.0 34.0 125

-- Using Credible Surveillance Data 45.0 1.817 1.1639 40 52.4 0 8.5 17.0 109 Inlet Nozzle to Upper Shell Welds 272 0.006 0.0773 -5 6 "b) 21.0 17 (b) 10.5 40.0 5 US Long Welds 101-122A,B,C 38.7 0.042 0.2675 -10 10.4 0 5.2 10.4 11 US to IS Circ Weld 103-121 31.3 0.044 0.2723 -5 6(b) 8.5 1 7 (b) 4.3 35.1 -12 IS Long Welds 101-124A,B,C 30.7 1.061 1.0165 -60 31.2 0 15.6 31.2 2 Using Credible Surveillance Data 30.0 1.061 1.0165 -60 30.5 0 14.0 28.0 -2 IS to LS Circ Weld 101-171 30.7 1.805 1.1621 -60 35.7 0 17.8 35.7 11 Using Credible Surveillance Data 30.0 1.805 1.1621 -60 34.9 0 14.0 28.0 3 LS Long Welds 101-142A,B,C 30.7 1.746 1.1532

-60 35.4 0 17.7 35.4 11

-- Using Credible Surveillance Data 30.0 1.746 1.1532 -60 34.6 0 14.0 28.0 3 (See next page for notes)

WCAP-1744 1-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-7 Notes for Table 5-5:

(a) Initial RTNOT values are measured values, unless otherwise noted.

(b) Initial RTNDT values are generic; hence a7 = 17'F.

(c) Per WCAP-16526-NP, Revision 0 [Ref. 6], the surveillance data of the plate and weld material was deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal TA= 17'F for Position 1.1 and, with credible surveillance data, caz = 8.5°F for Position 2.1; the weld metal GA ý 28 0 F for Position 1.1 and, with credible surveillance data, GA = 14'F for Position 2.1. However, GA need not exceed 0.5*ARTNDT.

(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART = RTNOT(Uj + ARTNDT + Margin.

WCAP- 17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-8 Table 5-6 Adjusted Reference Temperature Evaluation for the Seabrook Unit I Reactor Vessel Beltline Materials through 55 EFPY at the 3/4T Location RectrVese atril(OF) CF (

E . /4T e) 3/4T FF RTNDT(U)(a)

(OF) ARTNDT (OF) Fa1 (OF) (OF)(C CY "M (OF) ART(d)

(OF)

Reactor Vessel Material Inlet Nozzles 67 0.002 0.0349 0 2.3 0 1.2 2.3 5 US Plate RI 807-1 51 0.016 0.1468 30 7.5 0 3.7 7.5 45 US Plate R1807-2 58 0.016 0.1468 30 8.5 0 4.3 8.5 47 US Plate R1807-3 37 0.016 0.1468 10 5.4 0 2.7 5.4 21 IS Plate R1806-1 28.5 0.645 0.8772 40 25.0 0 12.5 25.0 90 IS Plate R1806-2 37 0.645 0.8772 0 32.5 0 16.2 32.5 65 IS Plate RI 806-3 47.5 0.645 0.8772 10 41.7 0 17.0 34.0 86 LS Plate R1808-1 37 0.645 0.8772 40 32.5 0 16.2 32.5 105 LS Plate R1808-2 37 0.645 0.8772 10 32.5 0 16.2 32.5 75 LS Plate R1808-3 44 0.645 0.8772 40 38.6 0 17.0 34.0 113

- Using Credible Surveillance Data 45.0 0.645 0.8772 40 39.5 0 8.5 17.0 96 Inlet Nozzle to Upper Shell Welds 272 0.002 0.0349 -56".' 9.5 17"b 4.7 35.3 -11 US Long Welds 101-122A,BC 38.7 0.015 0.1438 -10 5.6 0 2.8 5.6 1 US to IS Circ Weld 103-121 31.3 0.016 0.1468 - 5 6 (b) 4.6 1 7 (b) 2.3 34.3 -17 IS Long Welds 10 1-124ABC 30.7 0.377 0.7298 -60 22.4 0 11.2 22.4 -15 Using Credible Surveillance Data 30.0 0.377 0.7298 -60 21.9 0 10.9 21.9 -16 IS to LS Circ Weld 101-171 30.7 0.641 0.8753 -60 26.9 0 13.4 26.9 -6 Using Credible Surveillance Data 30.0 0.641 0.8753 -60 26.3 0 13.1 26.3 -7

............... gWelds

. .W ...

4._ A. ...................................

._..............0...!.:

LS L!

Long.............

... *...C..

10 1-142A,BC... .....

3.

0.......

30.7. .........

... . . . . . . ..0.. ._.6.....

62....

0.......

..... . 0...

...66. ............... -60 0.8660 6O. . . . . . . . . .....

226.-6

.6...... 00 ......

13.3......

66 ...........

26.6 ..........

-77....

-- Using Credible Surveillance Data 30.0 0.620 0.8660 -60 26.0 0 13.0 26.0 -8 (See next page for notes)

WCAP-1744 I-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-9 Notes for Table 5-6:

(a) Initial RTNDT values are measured values, unless otherwise noted.

(b) Initial RTNDT values are generic; hence a, = 17'F.

(c) Per WCAP-16526-NP, Revision 0 [Ref. 6], the surveillance data of the plate and weld material was deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal aA = 17'F for Position 1.1 and, with credible surveillance data, ac = 8.50 F for Position 2.1; the weld metal a* = 28 0 F for Position 1.1 and, with credible surveillance data, aY = 140 F for Position 2.1. However, aA need not exceed 0.5*ARTNDT.

(d) The Regulatory Guide 1.99., Revision 2 methodology was used to calculate ART values. ART = RTNDT(Ul + ARTNDT + Margin.

October2011 WCAP- 17441-NP WCAP- 1744 1-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 5-10 The limiting ART values for Seabrook Unit 1 to be used in the generation of the P-T limit curves are based on Lower Shell Plate R1808-1 (without using surveillance data, Position 1.1). The limiting ART values, using the "Axial Flaw" methodology, for Lower Shell Plate R1808-1 are summarized below in Table 5-7.

Table 5-7 Summary of the Limiting ART Values Used in the Generation of the Seabrook Unit 1 Heatup and Cooldown Curves at 36 and 55 EFPY EFPY 1/4T Limiting ART 3/4T Limiting ART Lower Shell Plate R1808-1 Lower Shell Plate R1808-1 Without Using Surveillance Data, Position 1.1 Without Using Surveillance Data, Position 1.1 36 113 0F 96 0 F Lower Shell Plate R1808-1 Lower Shell Plate RI 808-1 Without Using Surveillance Data, Position 1.1 Without Using Surveillance Data, Position 1.1 55 117°F 105 0 F October 2011 1744 I-NP WCAP- 17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6-1 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-A, Revision 4.

Figures 6-1 and 6-3 present the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 80 and 100°F/hr applicable for 36 and 55 EFPY, without the flange requirements and using the "Axial Flaw" methodology. 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, -80, and -100°F/hr applicable for 36 and 55 EFPY, without the flange requirements and using the "Axial Flaw" methodology.

The heatup and cooldown curves were generated using the 1998 through the 2000 Addenda ASME Code Section X1, Appendix G.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines 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 (heatup curve only). 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 the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G as follows:

1.5 Kim < Kic

where, K1 ,mis the stress intensity factor covered by membrane (pressure.) stress, Ki, = 33.2 + 20.734 e [0.02 (T-RTNT)l, 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 in order to provide additional margin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40'F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 4 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperatures for the inservice hydrostatic leak tests for the Seabrook Unit I reactor vessel at 36 and 55 EFPY are 173TF and 1771F, respectively. The vertical line drawn from these points on the pressure-temperature curve, WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 .6-2 intersecting a curve 40'F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 6-1 through 6-4 define all of the above limits for ensuring prevention of non-ductile failure for the Seabrook Unit I reactor vessel for 36 and 55 EFPY without the flange requirements and without instrumentation uncertainties. The data points used for developing 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- 17441 -NP October 2011 Revision 0

Westin,-house Non-Proprietary Class 3 6-3 Westinghouse Non-Proprietary Class 3 6-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate R1808-1 without using surveillance data, Position 1.1 LIMITING ART VALUES AT 36 EFPY: 1/4T, 113'F (Axial Flaw) 3/4T, 96°F (Axial Flaw)

Figure 6-1 Seabrook Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 and 1001F/hr) Applicable for 36 EFPY (without Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App.

G Methodology (w/ Ki) 2500 OperlimVersion5.3 Run: 15418 Opedim.xlsm Version: 5.3I 50 2250 _______ __ +

Leak Test Limit

2000 . . . I UnaccepatbIle 1750 - Operationi---=

__ Hatup Rate

__o_ 80ODe .F/Hr _ _

_1500 Critical Limit d- 80ODe . FIHr 1250 Heatup Rate _ _ _ _ _ _

01250 10D F/Hr

Critical Limit 100ODen.F/Hr 1000 - Acceptable --

"iOperation 750 500 250 0

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

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower She]l Plate R]808-1 without using surveillance data, Position 1.1 LIMITING ART VALUES AT 36 EFPY: 1/4T, 113'F (Axial Flaw) 3/4T, 96°F (Axial Flaw)

Figure 6-2 Seabrook Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, -80, and -100°F/hr) Applicable for 36 EFPY (without Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1 )

2500 _ , I_ I I Ioperlim Versior:5.3 Run:15418 Oper-im.xlsm Version: 5.3A 2250 2000 1750 1500 Ul)

(L 1250 0.

1000 750 500 250 0

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

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6-5 Table 6-1 Seabrook Unit 1 36 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ Kl,, w/o Flange Notch, and w/o Margins for Instrumentation Errors) 80"F/hr Criticality 100°F/hr Heatup 100'F/hr 80IF/hr Heatup Criticality T P T P T P T P (OF) (psig) (OF) (psig) (IF) (psig) (OF) (psig) 60 0 173 0 60 0 173 0 60 739 173 886 60 726 173 819 65 741 175 900 65 726 175 829 70 741 180 935 70 726 180 856 75 741 185 976 75 726 185 888 80 741 190 1021 80 726 190 923 85 741 195 1071 85 726 195 963 90 742 200 1128 90 726 200 1008 95 746 205 1190 95 726 205 1058 100 753 210 1260 100 728 210 1114 105 764 215 1337 105 734 215 1175 110 778 220 1422 110 742 220 1244 115 795 225 1516 115 753 225 1320 120 816 230 1620 120 767 230 1405 125 840 235 1736 125 785 235 1498 130 868 240 1863 130 805 240 1602 135 900 245 2004 135 829 245 1716 140 935 250 2159 140 856 250 1842 145 976 255 2330 145 888 255 1981 150 1021 150 923 260 2135 155 1071 155 963 265 2304 160 1128 160 1008 165 1190 165 1058 170 1260 170 1114 175 1337 175 1175 180 1422 180 1244

185 1516 185 1320 190 1620 190 1405 195 1736 195 1498 200 1863 -1200 1602 205 2004 205 1716 210 2159 210 1842 215 2330 215 1981 WCAP-17441-NP Octobere 20 Revision 0

Westinghouse Non-Proprietary Class 3 6-6 80°F/hr Heatup 80 0F/hr Criticality 100 0F/hr Heatup 100 F/br Criticality T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 220 2135 1 1__1__1 225 2304 1 1 Leak Test Limit T (OF) P (psig) 156 2000 173 2485 WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6-7 Table 6-2 Seabrook Unit 1 36 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1,, w/o Flange Notch, and w/o Margins for Instrumentation Errors)

Steady State -20°F/hr. -40 0 F/hr. -60 0 F/hr. -80 0 F/hr. -100OF/hr.

T P T P T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 60 0 60 0 60 0 60 0 60 0 60 0 60 739 60 700 60 661 60 622 60 583 60 546 65 753 65 715 65 677 65 640 65 603 65 567 70 768 70 731 70 695 70 660 70 625 70 592 75 785 75 750 75 715 75 682 75 649 75 619 80 804 80 770 80 738 80 706 80 676 80 648 85 824 85 793 85 762 85 733 85 706 85 682 90 847 90 818 90 790 90 763 90 740 90 719 95 872 95 845 95 820 95 797 95 777 95 760 100 900 100 876 100 853 100 834 100 818 100 805 105 931 105 910 105 891 105 875 105 863 105 856 110 965 110 947 110 932 110 920 110 914 110 912 115 1003 115 988 115 977 115 971 115 970 115 970 120 1044 120 1034 120 1028 120 1027 120 1027 120 1027 125 1090 125 1085 125 1084 125 1084 125 1084 125 1084 130 1141 130 1140 130 1140 130 1140 130 1140 130 1140 135 1197 135 1197 135 1197 135 1197 135 1197 135 1197 140 1259 140 1259 140 1259 140 1259 140 1259 140 1259 145 1327 145 1327 145 1327 145 1327 145 1327 145 1327 150 1403 150 1403 150 1403 150 1403 150 1403 150 1403 155 1487 155 1487 155 1487 155 1487 155 1487 155 1487 160 1579 160 1579 160 1579 160 1579 160 1579 160 1579 165 1681 165 1681 165 1681 165 1681 165 1681 165 1681 170 1794 170 1794 170 1794 170 1794 170 1794 170 1794 175 1919 175 1919 175 1919 175 1919 175 1919 175 1919 180 2057 180 2057 180 2057 180 2057 180 2057 180 2057 185 2209 185 2209 185 2209 185 2209 185 2209 185 2209 190 2378 190 2378 190 2378 190 2378 190 2378 190 2378 195 2564 195 2564 195 2564 195 2564 195 2564 195 2564 WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6-8 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate R1808-1 without using surveillance data, Position 1.1 LIMITING ART VALUES AT 55 EFPY: 1/4T, 11 7°F (Axial Flaw) 3/4T, 105'F (Axial Flaw)

Figure 6-3 Seabrook Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 80 and 100 °F/hr) Applicable for 55 EFPY (without Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App.

G Methodology (w/ KI) 2500 lOperlim Version:5.3 Run:2324 Operimxlsm Version: 5.

Leak Test Limit 2250 2000 1750 Unacceptable 1750 __7 Operation O~td* Heatu p Rate 15oo De . F/Hr .

Critical Limit 1250 Heatup Rat8 125 1 ODe .FIHr" V/ Critical Limit 100ODe . FIHr!

=1000 - __

Acceptable 0C.11, 750 500 250 0

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

WCAP- 17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6-9 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate R1808-1 without using surveillance data, Position 1.1 LIMITING ART VALUES AT 55 EFPY: 1/4T, 117'F (Axial Flaw) 3/4T, 105'F (Axial Flaw)

Figure 6-4 Seabrook Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, -80, and -100°F/hr) Applicable for 55 EFPY (without Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ Ki) 2500 -. 1 , ,

Operlim Version 5.3 Run :2324 Opedlim.dsm Version: 5.

2250 -

2000 t i 1750 ,_-_ _ ... .

1500 In an 1250 CL 1000 750 500 250 0

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

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6-10 Table 6-3 Seabrook Unit 1 55 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ Ki,, w/o Flange Notch, and w/o Margins for Instrumentation Errors) 80°F/hr Criticality 1001F/hr Heatup 1001F/hr 801F/hr Heatup Criticality T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 60 0 177 0 60 0 177 0 60 707 177 841 60 691 177 776 65 707 180 860 65 691 180 790 70 707 185 893 70 691 185 815 75 707 190 931 75 691 190 844 80 707 195 972 80 691 195 877 85 707 200 1019 85 691 200 914 90 707 205 1071 90 691 205 955 95 709 210 1129 95 691 210 1001 100 714 215 1193 100 691 215 1053 105 722 220 1264 105 694 220 1110 110 732 225 1342 110 700 225 1173 115 746 230 1429 115 708 230 1243 120 762 235 1525 120 719 235 1321 125 782 240 1631 125 732 240 1407 130 804 245 1749 130 748 245 1502 135 830 250 1878 135 768 250 1607 140 860 255 2021 140 790 255 1723 145 893 260 2179 145 815 260 1851 150 931 265 2353 150 844 265 1993 155 972 155 877 270 2149 160 1019 160 914 275 2321 165 1071 165 955 170 1129 170 1001 175 1193 175 1053 180 1264 180 1110 185 1342 185 1173 190 1429 190 1243 195 1525 195 1321 200 1631 200 1407 205 1749 205 1502 210 1878 210 1607 215 2021 215 1723 WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6-11 80 0F/hr Criticality 100°F/hr Heatup 100OF/hr 801F/hr Heatup Criticality T P T P T P T P (OF) (psig) (OF) (Psig) (OF) (psig) (OF) (psig) 220 2179 220 1851 225 2353 1 225 1993 1 230 2149 235 2321 Leak Test Limit T (OF) P (psig) 160 2000 177 2485 WCAP-17441 -NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 6- 12 Table 6-4 Seabrook Unit 1 55 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1,, w/o Flange Notch, and w/o Margins for Instrumentation Errors)

Steady State -201F/hr. -4 0 °F/hr. -6 0 1F/hr. -80 0 F/hr. -100 0 F/hr.

T P T P T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 60 0 60 0 60 0 60 0 60 0 60 0 60 729 60 689 60 648 60 608 60 569 60 529 65 742 65 702 65 663 65 625 65 587 65 549 70 756 70 718 70 680 70 643 70 607 70 571 75 771 75 735 75 699 75 663 75 629 75 596 80 789 80 754 80 719 80 686 80 654 80 624 85 808 85 774 85 742 85 711 85 682 85 654 90 829 90 797 90 767 90 739 90 712 90 688 95 852 95 823 95 795 95 770 95 746 95 726 100 878 100 851 100 826 100 804 100 784 100 768 105 906 105 882 105 860 105 842 105 826 105 815 110 938 110 917 110 898 110 .883 110 873 110 866 115 972 115 955 115 940 115 930 115 924 115 924 120 1011 120 997 120 987 120 981 120 981 120 981 125 1053 125 1044 125 1039 125 1039 125 1039 125 1039 130 1100 130 1095 130 1095 130 1095 130 1095 130 1095 135 1152 135 1152 135 1152 135 1152 135 1152 135 1152 140 1209 140 1209 140 1209 140 1209 140 1209 140 1209 145 1272 145 1272 145 1272 145 1272 145 1272 145 1272 150 1342 150 1342 150 1342 150 1342 150 1342 150 1342 155 1419 155 1419 155 1419 155 1419 155 1419 155 1419 160 1504 160 1504 160 1504 160 1504 160 1504 160 1504 165 1599 165 1599 165 1599 165 1599 165 1599 165 1599 170 1703 170 1703 170 1703 170 1703 170 1703 170 1703 175 1818 175 1818 175 1818 175 1818 175 1818 175 1818 180 1946 180 1946 180 1946 180 1946 180 1946 180 1946 185 2086 185 2086 185 2086 185 2086 185 2086 185 2086 190 2242 190 2242 190 2242 190 2242 190 2242 190 2242 195 2414 195 2414 195 2414 195 2414 195 2414 195 2414 200 2604 200 2604 200 2604 200 2604 200 2604 200 2604 October 2011 WCAP-17441-NP WCAP-1 7441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 7-1 7 REFERENCES

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

Nuclear Regulatory Commission, May 1988.

2. WCAP-14040-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. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code, Section Xl, Division I, "Fracture Toughness Criteria for Protection Against Failure."

4. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"

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

5. WCAP-17444, Revision 0, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Seabrook Unit 1," A. E. Freed and A. Udyawar, October 2011.

6. WCAP-16526-NP., Revision 0., "Analysis of Capsule V from FPL Energy - Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program," B. N. Burgoset al., March 2006.

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

8. RSICC Computer Code Collection CCC-650, "DOORS 3.2: One-, Two-, and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System," April 1998.

9. RSICC 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.

10. WCAP-15745, Revision 0, "Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, December 2001.

11. "Seabrook Station Updated Final Safety Analysis Report," Revision 14.

12. "Seabrook Station License Renewal Application," May 2010. [ADAMS Accession # ML101590094]

13. CE-NPSD-1039, Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CEOG Task 902, June 1997.

14. CE-NPSD-1119, Revision 1, "Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content," CEOG Task 1054, July 1998.

15. Code of Federal Regulations, 10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60., No. 243, dated December 19, 1995, effective January 18, 1996.

16. ASME Boiler and Pressure Vessel (B&PV) Code, Section 1i1, Division 1, Subsection NB, Section NB-2300, "Fracture Toughness Requirements for Material."

WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 A-]

APPENDIX A THERMAL STRESS INTENSITY FACTORS (Kit)

Tables A-I and A-2 contain the thermal stress intensity factors (K,,) for the maximum heatup and cooldown rates at 36 and 55 EFPY for Seabrook Unit 1. The vessel radii to the I/4T and 3/4T locations are as follows:

1/4T Radius = 88.818 inches

3/4T Radius = 93.133 inches WCAP-1 7441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 A-2 Table A-1 KI, Values for Seabrook Unit 1 at 36 and 55 EFPY 100°F/hr Heatup Curves (w/o Flange Requirements and w/o Margins for Instrument Errors)

Vessel 1/4T Thermal Vessel 3/4T Thermal Water Temperature at Temperature at Temp. 1/4T Location for Stress Intensity Factor 3/4T Location for Stress Intensity Factor (OF) 100 0F/hr Heatup (KSI din.) 100F/hr Heatup (KSI iin.)

(OF) (OF) 60 55.984 -0.995 55.043 0.473 65 58.556 -2.453 55.293 1.438 70 61.618 -3.714 55.960 2.426 75 64.894 -4.912 57.096 3.357 80 68.443 -5.948 58.648 4.192 85 72.104 -6.896 60.581 4.940 90 75.947 -7.719 62.854 5.602 95 79.889 -8.471 65.426 6.196 100 83.963 -9.129 68.265 6.723 105 88.123 -9.729 71.337 7.195 110 92.379 -10.256 74.615 7.615 115 96.708 -10.738 78.073 7.994 120 101.108 -11.164 81.690 8.334 125 105.567 -11.555 85.447 8.642 130 110.078 -11.903 89.328 8.918 135 114.638 -12.223 93.318 9.169 140 119.237 -12.509 97.406 9.396 145 123.876 -12.773 101.578 9.603 150 128.543 -13.011 105.826 9.792 155 133.243 -13.232 110.141 9.964 160 137.964 -13.432 114.515 10.122 165 142.711 -13.620 118.941 10.268 170 147.475 -13.790 123.414 10.403 175 152.259 -13.951 127.927 10.528 180 157.056 -14.099 132.476 10.644 185 161.869 -14.239 137.057 10.752 190 166.692 -14.369 141.666 10.854 195 171.527 -14.493 146.300 10.949 200 176.370 -14.608 150.957 11.039 205 181.223 -14.720 155.633 11.125 210 186.083 -14.824 160.326 11.206 WCAP-17441-NP October 2011 Revision 0

Westinghouse Non-Proprietary Class 3 A-3 Table A-2 Kit Values for Seabrook Unit 1 at 36 and 55 EFPY 100°F/hr Cooldown Curves (w/o Flange Requirements and w/o Margins for Instrument Errors)

Water Vessel Temperature -100°F/hr Cooldown at 1/4T Location for 1/4T Thermal Stress Temp -10 0 'F/hr Cooldown Intensity Factor (OF) (KSI 'in.)

210 237.078 17.166 205 231.993 17.097 200 226.907 17.029 195 221.820 16.959 190 216.734 16.891 185 211.647 16.821 180 206.560 16.752 175 201.473 16.682 170 196.385 16.613 165 191.298 16.543 160 186.210 16.474 155 181.122 16.404 150 176.035 16.335 145 170.947 16.265 140 165.859 16.195 135 160.771 16.126 130 155.684 16.057 125 150.596 15.987 120 145.508 15.918 115 140.421 15.849 110 135.333 15.780 105 130.246 15.711 100 125.159 15.642 95 120.072 15.573 90 114.985 15.505 85 109.898 15.436 80 104.811 15.368 75 99.724 15.299 70 94.637 15.231 65 89.551 15.163 60 84.466 15.095 WCAP-17441-NP October 2011 Revision 0

Attachment 4 Exemption Request

Exemption Request to Allow Elimination of the Reactor Vessel Flange Requirement from the Reactor Coolant System Pressure - Temperature Limits Exemption Request In accordance with 10 CFR 50.12, "Specific exemptions," NextEra is requesting an exemption from the requirements of 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation." The exemption would permit the use of WCAP-17444-P, Rev. 0, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Seabrook Unit 1," in lieu of the methodology required by 10 CFR 50, Appendix G, footnote 2 to Table 1. WCAP-17444 demonstrates that the flange region can tolerate assumed flaws of 0.1 T (thickness) during the heatup, cooldown, and boltup conditions. Additionally, it can be concluded that flaws are unlikely to initiate in the flange region, since there is no known degradation mechanism for the flange region and the fatigue usage in the flange region is less than 0.1.

Furthermore, based on WCAP-1 7444, the alternative flange temperature requirement of 46 0 F is less than the minimum boltup temperature of 60°F for Seabrook Unit 1. Therefore, eliminating the requirement for the reactor vessel/head flange region when determining pressure-temperature (P/T) limits for the reactor vessel is justified. The proposed exemption meets the criteria of 10 CFR 50.12 as discussed below.

Justificationfor Exemption 10 CFR 50.12 states that the Commission may grant an exemption from the requirements contained in 10 CFR 50 provided that the following is met:

1. The requestedexemption is authorized by law.

10 CFR 50.60 (b) allows the use of alternatives to 10 CFR 50, Appendix G when an exemption is granted by the Commission under 10 CFR 50.12.

2. The requestedexemption does not present an undue risk to the public health and safety.

The revised P-T limit curves developed for Seabrook Unit 1 use the methodology described in WCAP-17444. The WCAP methodology uses a higher material fracture toughness, KIc (fracture toughness based on the lower bound of static initiation critical values measured as a function of temperature) instead of Kla (fracture toughness based upon the lower bound of crack arrest critical values measured as a function of temperature), which results in higher allowable pressures. 10 CFR 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange regions. The regulation 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 pre-service hydrostatic test pressure.

Implementing the P/T limit curves that use the KIc material fracture toughness without eliminating the flange requirement of 10 CFR 50, Appendix G, would place a restricted operating window in the temperature range associated with the flange/closure head, i.e., flange RTNDT + 120 0 F. In accordance with WCAP-17444, the KIc toughness has been shown to provide significant margin between the applied stress intensity factor and the fracture toughness of the flange/closure head.

Applying the WCAP-1 7444 methodology to the P/T limits will enhance overall plant safety by expanding the P/T operating window, especially in the region of low temperature operations.

The two primary safety benefits that would be realized are a reduction in the potential challenges to the cold overpressure mitigation system and a reduction in the risk of damaging the reactor coolant pump seals. This will produce a significant improvement in plant safety by reducing the probability of an inadvertent reduction in reactor coolant inventory and in easing the burden on the operators. WCAP-1 7444 concludes that the integrity of the closure head/flange is not a concern for safe unit operation and testing. Therefore, the proposed exemption does not present an undue risk to the public health and safety.

3. The requested exemption is consistent with the common defense and security.

This exemption request does not affect the national defense or security issues.

The common defense and security are not impacted by the approval of this exemption request.

4. Special circumstancesare present which necessitate the request for an exemption to the regulations of 10 CFR 50.60.

In accordance with 10 CFR 50.12(a) (2), the NRC will consider granting an exemption to the regulations if special circumstances are present. This requested exemption meets the special circumstances of 10 CFR 50.12(a) (2)

(ii) that states:

"Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule;"

The underlying purpose of 10 CFR 50.60 and 10 CFR 50, Appendix G is to protect the integrity of the reactor coolant pressure boundary. 10 CFR 50, Appendix G establishes the requirements for the P/T limits for pressure retaining components of the reactor coolant pressure boundary and requirements for the minimum metal temperature of the reactor pressure vessel closure head flange and reactor vessel flange regions. The P/T limits are determined using the methodology of ASME B&PV Code, Section Xl, Appendix G, with additional, more restrictive, flange temperature requirements specified in 10 CFR 50, Appendix G. WCAP-17444

demonstrated that significant margin exists between the applied stress intensity factor and the material fracture toughness when using the KIc toughness, which has been endorsed by the ASME B&PV Code, Section Xl for developing Pressure-Temperature Limit Curves expanding the PIT operating window, enhancing overall plant safety. Another purpose of the requirements of 10 CFR 50, Appendix G is to assure that fracture margins are maintained to protect against service induced cracking due to environmental effects. Since the governing flaw is on the outside surface (the inside surface is in compression) where there are no environmental effects, there is a greater assurance that the fracture margin is maintained.

Therefore, it can be concluded that the integrity of the closure head/flange region is not a concern for Seabrook Unit 1 using the KIc toughness. Additionally, there are no known degradation mechanisms for this region, other than fatigue. The calculated design fatigue usage for this region is less than 0.1, therefore it can be concluded that flaws are unlikely to initiate in this region.

Application of the regulation in this case is not necessary because the use of WCAP-17444 achieves the underlying intent of 10 CFR 50.60 and 10 CFR 50, Appendix G.

ML14216A405

Text

SUMMARY

2.1 INTRODUCTION

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