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WCAP-17035-NP, Revision 2, Watts Bar, Unit 2, Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation
ML100550651
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Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 12/31/2009
From: Rosier B A
Westinghouse
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References
WCAP-17035-NP, Rev 2
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ii WCAP-17035-NP December 2009 Revision 2 RECORD OF REVISION Revision 0: Original Issue Revision 1: This revision was created to correct the initial upper shelf energy values, and to update the end-of-life (EOL) upper shelf energy predictions using the new initial values. These changes are reflected in Table B-1 of Appendix B (Upper Shelf Energy Evaluation). No conclusions changed as a result of using the new initial values to update the EOL upper shelf energy predictions. Revision 2: This revision was created to modify the executive summary. Paragraphs were added to describe the basis for the use of ASME Code Section XI rather than ASME Code Section III in the pressure-temperature limit curve development.

iii WCAP-17035-NP December 2009 Revision 2 TABLE OF CONTENTS LIST OF TABLES.......................................................................................................................................iv LIST OF FIGURES......................................................................................................................................v EXECUTIVE

SUMMARY

..........................................................................................................................vi 1 INTRODUCTION........................................................................................................................1-1 2 FRACTURE TOUGHNESS PROPERTIES.................................................................................2-1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS................3-1 3.1 OVERALL APPROACH.................................................................................................3-1 3.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT............................................................................................................3-1 3.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS...........................................3-5 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE..........................................4-1 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES.......................5-1 6 REFERENCES.............................................................................................................................6-1 APPENDIX A Thermal Stress Intensity Factors (K It)........................................................A-1 APPENDIX B Upper Shelf Energy Evaluation.................................................................B-1 APPENDIX C Pressurized Thermal Shock Evaluation.....................................................C-1 APPENDIX D Emergency Response Guideline Limits Evaluation..................................D-1

iv WCAP-17035-NP December 2009 Revision 2 LIST OF TABLES Table 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT NDT Values for the Watts Bar Unit 2 Reactor Vessel Materials......................................................................2-2 Table 2-2 Summary of the Initial RT NDT Values for the Watts Bar Unit 2 Closure Head and Vessel Flange..............................................................................................................................2-2 Table 2-3 Summary of the Watts Bar Unit 2 Reactor Vessel Beltline Material Chemistry Factors.2-2 Table 4-1 Fluence Values for the Watts Bar Unit 2 Reactor Vessel Beltline Materials....................4-2 Table 4-2 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline Materials through 7 EFPY at the 1/4T Location................................................4-2 Table 4-3 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline Materials through 7 EFPY at the 3/4T Location................................................4-2 Table 4-4 Summary of the Limiting ART Values Used in the Generation of the Watts Bar Unit 2 Heatup/Cooldown Curves................................................................................................4-3 Table 5-1 7 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology (w/K IC, w/Flange Notch and w/o Uncertainties for Instrumentation Errors)

.........................................................................................................................................5-5 Table 5-2 7 EFPY Cooldown Curve Data Points Using 1998 through 2000 Addenda App. G Methodology (w/K IC, w/Flange Notch and w/o Uncertainties for Instrumentation Errors)

.........................................................................................................................................5-6 Table A-1 K It Values for 100F/hr Heatup Curve (w/o Margins for Instrument Errors)..................A-2 Table A-2 K It Values for 100F/hr Cooldown Curve (w/o Margins for Instrument Errors)............A-3 Table B-1 Predicted Position 1.2 Upper Shelf Energy Values at 32 EFPY.....................................B-3 Table C-1 RT PTS Calculations for the Watts Bar Unit 2 Beltline Materials at 32 EFPY..................C-1 Table D-1 Evaluation of Watts Bar Unit 2 ERG Limit Category.....................................................D-1

v WCAP-17035-NP December 2009 Revision 2 LIST OF FIGURES Figure 5-1 Watts Bar Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 7 EFPY (without Margins for Instrumentation Errors) Using 1998 through 2000 Addenda App. G Methodology (w/K Ic)............................................5-3 Figure 5-2 Watts Bar Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 7 EFPY (without Margins for Instrumentation Errors) Using 1998 through 2000 Addenda App. G Methodology (w/K Ic)............................................5-4 Figure B-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper Shelf Energy as a Function of Copper and Fluence.....................................................................................B-2

vi WCAP-17035-NP December 2009 Revision 2 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 Watts Bar Unit 2 reactor vessel. The heatup and cooldown P-T limit curves were generated using the highest adjusted reference temperature (ART) value pertaining to Watts Bar Unit 2. The highest ART value was that of intermediate shell forging 05 at both 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations. The P-T curves made use of the K Ic methodology detailed in the 1998 through 2000 Addenda Edition of the ASME Code,Section XI, Appendix G, and ASME Code Case N-641. The applicable ASME Section III Edition and Addenda for the Watts Bar Unit 2 Reactor Pressure Vessel is the 1971 Edition with Addenda through Winter 1971. This Edition and Addenda did not contain specific information or requirements related to development of the Pressure-Temperature (P-T) limits required by 10 CFR 50, Appendix G. Appendix G of Section III was developed later. Since no guidance existed in the Code of Record for the vessel, the Pressure Temperature Limits Report (PTLR) has been developed in accordance with current methodologies contained in Westinghouse Topical Report WCAP-14040-A, Revision 4, which has been previously accepted by the NRC. The P-T limit curves developed herein utilize the methodology contained in Appendix G of Section XI of the ASME Code. This is the NRC requirement for P-T limit curve development. Particularly, 10 CFR 50, Appendix G (Section IV.A.2.b), requires that P-T limits must be at least as conservative as limits obtained by following the methods of analysis and the margins of safety of Appendix G of Section XI of the ASME Code. Therefore, this NRC requirement is met by the P-T limit curves developed herein. The methods of Appendix G of ASME Code Section XI are described in the NRC-approved Westinghouse methodology (WCAP-14040-A, Revision 4), which was used to develop the P-T limit curves. The NRC Safety Evaluation (SE), contained in the opening pages of WCAP-14040-A, Revision 4, concludes that the contents of WCAP-14040-A, Revision 4, are acceptable for referencing as PTLR methodology (See Section 4.0.b of the SE for the approval statement). WCAP-14040-A, Revision 4 contains guidance on the use of Code Case N-641, which is approved for use without any exemption request per NRC SE Section 4.0.b. In summary, the approach taken herein for P-T limit curve development is in accordance with the applicable NRC requirements. Further, given that ASME Section III, 1971 Edition with Addenda through Winter 1971 does not provide any criteria, this methodology does not conflict with the applicable construction Code requirements. The P-T limit curves were generated for 7 EFPY using heatup rates of 60 and 100F/hr and cooldown rates of 0, 20, 40, 60 and 100F/hr. The curves were developed without margins for instrumentation errors. These curves can be found in Figures 5-1 and 5-2. Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates. Also documented in this report are the upper shelf energy (USE), pressurized thermal shock (PTS), and emergency response guideline (ERG) limit evaluations. These evaluations are included in appendices B, C, and D, respectively.

1-1 WCAP-17035-NP December 2009 Revision 2 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 RT NDT 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 RT NDT, and adding a margin. The unirradiated RT NDT 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 60F. RT NDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT NDT at any time period in the reactor's life, RTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RT NDT (IRTNDT). The extent of the shift in RT NDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" [Reference 1]. Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + RT NDT + margins for uncertainties) at the surface, 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 [Reference 2], "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." Specifically, the K IC methodology of the 1998 through 2000 Addenda Edition of ASME Code,Section XI, Appendix G [Reference 3] was used. The calculated ART values are documented in Tables 4-2 and 4-3 of this report. The design basis fluence projections are based on the values verified by Westinghouse in letter LTR-REA-08-105, Revision 2 [Reference 4]. The purpose of this report is to present the calculations and the development of the Watts Bar Unit 2 heatup and cooldown P-T limit curves for 7 EFPY. This report documents the calculated ART values and the development of the P-T limit curves for normal operation. The P-T curves herein were generated without instrumentation errors. The P-T curves include pressure-temperature limits for the vessel flange region per the requirements of 10 CFR Part 50, Appendix G [Reference 5].

2-1 WCAP-17035-NP December 2009 Revision 2 2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the Watts Bar Unit 2 reactor vessel are presented in Table 2-1. The unirradiated RT NDT values for the closure head and vessel flange are documented in Table 2-2. 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 [Reference 2]. The chemistry factors (CFs) were calculated using Regulatory Guide 1.99 Revision 2, Position 1.1. 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 2-1. Table 2-3 summarizes the Position 1.1 CFs determined for the Watts Bar Unit 2 beltline materials.

2-2 WCAP-17035-NP December 2009 Revision 2 Table 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT NDT Values for the Watts Bar Unit 2 Reactor Vessel Materials Material Description Chemical Composition Reactor Vessel Beltline Region Location Cu wt% Ni wt% Initial RT NDT (a) Intermediate Shell Forging 05 0.05 0.78 14°F Lower Shell Forging 04 0.05 0.81 5°F Intermediate to Lower Shell Circumferential Weld Seam W05 0.05 0.70 -50°F Note: (a) The initial RTNDT values are measured values, taken from WCAP-13830, Revision 1 [Reference 6].

Table 2-2 Summary of the Initial RTNDT Values for the Watts Bar Unit 2 Closure Head and Vessel Flange Material Identification Initial RT NDT (a) Closure Head Flange -40°F Vessel Flange -22°F Note: (a) The initial RTNDT values are measured values, taken from WCAP-13830, Revision 1 [Reference 6].

Table 2-3 Summary of the Watts Bar Unit 2 Reactor Vessel Beltline Material Chemistry Factors Beltline Materials Chemistry Factor Intermediate Shell Forging 05 31°F Lower Shell Forging 04 31°F Intermediate to Lower Shell Circumferential Weld Seam W05 68°F 3-1 WCAP-17035-NP December 2009 Revision 2 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K I, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K Ic , for the metal temperature at that time. K Ic is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Reference 3].

The K Ic curve is given by the following equation:

Ke IcTRT NDT33220734002..*[.()] (1) where, K Ic (ksiin.) = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT This K Ic curve is based on the lower bound of static critical K I 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.

3.2 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* K Im + K It < K Ic (2) where, K Im = stress intensity factor caused by membrane (pressure) stress K It = stress intensity factor caused by the thermal gradients K Ic = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT 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 3-2 WCAP-17035-NP December 2009 Revision 2 For membrane tension, the corresponding K I for the postulated defect is: KMpRt mi Im(/) (3) where, M m for an inside surface flaw is given by:

M m = 1.85 for t < 2, M m = 0.926 t for 2 t3464., M m = 3.21 for t > 3.464 Similarly, M m for an outside surface flaw is given by:

M m = 1.77 for t < 2, M m = 0.893 t for 2 464.3 t , M m = 3.09 for t > 3.464 and p = internal pressure (ksi), Ri = vessel inner radius (in.), and t = vessel wall thickness (in.). For bending stress, the corresponding K I for the postulated defect is:

K Ib = M b

  • Maximum Stress, where M b is two-thirds of M m (4) The maximum K I produced by radial thermal gradient for the postulated inside surface defect of G-2120 is: K It = 0.953x10

-3 x CR x t2.5 (5) where CR is the cooldown rate in F/hr., or for a postulated outside surface defect K It = 0.753x10

-3 x HU x t2.5 (6) where HU is the heatup rate in F/hr. The through-wall temperature difference associated with the maximum thermal K I 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 maximum thermal K I.

3-3 WCAP-17035-NP December 2009 Revision 2 (a) The maximum thermal K I relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2). (b) Alternatively, the K I for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness inside surface defect using the relationship: KCCCCa It(....)*103590632204753038550123 (7) or similarly, K It during heatup for a 1/4-thickness outside surface defect using the relationship:

a C C C C K It*)401.0 481.0 630.0 043.1 (3 2 1 0 (8) where the coefficients C 0 , 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: ()(/)(/)(/)xCCxaCxaCxa012 2 3 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" [Reference 2] Section 2.6 (equations 2.6.2-4 and 2.6.3-1). At any time during the heatup or cooldown transient, K Ic 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 RT NDT, 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, K It, 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 XI 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 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.

3-4 WCAP-17035-NP December 2009 Revision 2 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 T (temperature) across the vessel wall developed during cooldown results in a higher value of K Ic at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K Ic exceeds K It, 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 1/4T 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 K Ic for the inside 1/4T flaw during heatup is lower than the K Ic 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 K Ic 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 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.

3-5 WCAP-17035-NP December 2009 Revision 2 3.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [Reference 5] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RT NDT by at least 120F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3107 psig for Watts Bar Unit 2), which is calculated to be 621 psig. The limiting unirradiated RT NDT of -22F occurs in the vessel flange of the Watts Bar Unit 2 reactor vessel, so the minimum allowable temperature of this region is 98F at pressures greater than 621 psig (without instrument uncertainties). This limit is shown in Figures 5-1 and 5-2 wherever applicable.

4-1 WCAP-17035-NP December 2009 Revision 2 4 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 RT NDT + RT NDT + Margin (10) Initial RT NDT 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 [Reference 7]. If measured values of initial RT NDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class. RT NDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

RT NDT = CF

  • f(0.28 - 0.10 log f) (11) To calculate RT NDT 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 x) = fsurface

  • e (-0.24x) (12) where x inches (vessel beltline thickness is 8.465 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 RT NDT at the specific depth. The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections in LTR-REA-08-105, Revision 2 [Reference 4], and the results are presented in Table 4-1. The evaluation methods used in Reference 4 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" [Reference 2]. Table 4-1 also provides a summary of the vessel fluence projections at the 1/4T and 3/4T locations. Tables 4-2 and 4-3 contain the 1/4T and 3/4T calculated fluences and fluence factors, per Regulatory Guide 1.99, Revision 2, used to calculate the 7 EFPY ART values for all beltline materials in the Watts Bar Unit 2 reactor vessel. Margin is calculated as M = 2 i 22. The standard deviation for the initial RTNDT margin term ( i) is 0F when the initial RT NDT is a measured value and 17F when a generic value is available. The standard deviation for the RTNDT margin term, , is 17F for plates or forgings, and 8.5F for plates or forgings when credible surveillance data is used. For welds, is equal to 28F when surveillance capsule data is not used, and is 14F (half the value) when credible surveillance capsule data is used. need not exceed 0.5 times the mean value of RT NDT. Contained in Tables 4-2 and 4-3 are the Watts Bar Unit 2 7 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the heatup and cooldown curves.

4-2 WCAP-17035-NP December 2009 Revision 2 Table 4-1 Fluence Values for the Watts Bar Unit 2 Reactor Vessel Beltline Materials 7 EFPY Fluence (n/cm 2 , E > 1.0 MeV)

Beltline Materials Inner Wetted Surface 1/4T Location (x=2.116 in.) 3/4T Location (x=6.349 in.) Intermediate Shell Forging 05 6.93E+18 4.17E+18 1.51E+18 Lower Shell Forging 04 6.93E+18 4.17E+18 1.51E+18 Intermediate to Lower Shell Circumferential Weld Seam W05 6.93E+18 4.17E+18 1.51E+18 Table 4-2 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline Materials through 7 EFPY at the 1/4T Location Reactor Vessel Location CF ( F) 1/4T f (n/cm 2 , E > 1.0 MeV) 1/4T FF RT NDT ( F) IRT NDT (a) ( F) i(a) ( F) ( F) M ( F) ART ( F) Intermediate Shell Forging 05 31 4.17E+18 0.757 23.5 14 0 11.7 23.5 61 Lower Shell Forging 04 31 4.17E+18 0.757 23.5 5 0 11.7 23.5 52 Intermediate to Lower Shell Circumferential Weld Seam W05 68 4.17E+18 0.757 51.5 -50 0 25.7 51.5 53 Note: (a) The initial RT NDT values are measured values; therefore, i = 0°F. Table 4-3 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline Materials through 7 EFPY at the 3/4T Location Reactor Vessel Location CF ( F) 3/4T f (n/cm 2 , E > 1.0 MeV) 3/4T FF RT NDT ( F) IRT NDT (a) ( F) i(a) ( F) ( F) M ( F) ART ( F) Intermediate Shell Forging 05 31 1.51E+18 0.504 15.6 14 0 7.8 15.6 45 Lower Shell Forging 04 31 1.51E+18 0.504 15.6 5 0 7.8 15.6 36 Intermediate to Lower Shell Circumferential Weld Seam W05 68 1.51E+18 0.504 34.3 -50 0 17.1 34.3 19 Note: (a) The initial RT NDT values are measured values; therefore, i = 0°F.

4-3 WCAP-17035-NP December 2009 Revision 2 Contained in Table 4-4 is a summary of the limiting ART values used in the generation of the Watts Bar Unit 2 reactor vessel P-T limit curves. The limiting material for both the 1/4T location and the 3/4T location is Intermediate Shell Forging 05.

Table 4-4 Summary of the Limiting ART Values Used in the Generation of the Watts Bar Unit 2 Heatup/Cooldown Curves Limiting ART (F) EFPY 1/4T 3/4T 7 61 45 5-1 WCAP-17035-NP December 2009 Revision 2 5 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 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4. Figure 5-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100F/hr applicable for 7 EFPY with the "Flange-Notch" requirement and using the "Axial-flaw" methodology. This curve was generated using 1998 through 2000 Addenda ASME Code Section XI, Appendix G. Figure 5-2 presents the limiting cooldown curve without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60 and 100°F/hr applicable for 7 EFPY with the "Flange-Notch" requirement. Again, this curve was generated using 1998 through 2000 Addenda ASME Code Section XI, Appendix G. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 and 5-2. 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 Figure 5-1 (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 1998 through 2000 Addenda ASME Code Section XI, Appendix G as follows:

1.5 K Im < K Ic where, K Im is the stress intensity factor covered by membrane (pressure) stress, K Ic = 33.2 + 20.734 e[0.02 (T - RT NDT)], T is the minimum permissible metal temperature, and RT NDT is the metal reference nil-ductility temperature. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production. 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 Watts Bar Unit 2 reactor vessel at 7 EFPY is 122F. The vertical line drawn 5-2 WCAP-17035-NP December 2009 Revision 2 from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel. Figures 5-1 and 5-2 define all of the above limits for ensuring prevention of non-ductile failure for the Watts Bar Unit 2 reactor vessel for 7 EFPY with the "Flange-Notch" requirement, without instrumentation uncertainties

. The data points used for developing the heatup and cooldown pressure-temperature limit curves shown in Figures 5-1 and 5-2 are presented in Tables 5-1 and 5-2.

5-3 WCAP-17035-NP December 2009 Revision 2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Forging 05 LIMITING ART VALUES AT 7 EFPY: 1/4T, 61 F 3/4T, 45 F 0 250 500 750 1000 1250 1500 1750 2000 2250 2500050100150200250300350400450500550Moderator Temperature (Deg. F)Calculated Pressure (PSIG)Operlim Version:5.2 Run:30881 Operlim.xls Version: 5.2UnacceptableOperationAcceptableOperation Criticality Limit based on inservice hydrostatic test temperature (122°F) for the service period up to 7 EFPYHeatup Rate60 Deg. F/HrHeatup Rate100 Deg. F/HrCritical Limit60 Deg. F/HrCritical Limit100 Deg. F/HrLeak Test Limit Figure 5-1 Watts Bar Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 7 EFPY (without Margins for Instrumentation Errors) Using 1998 through 2000 Addenda App. G Methodology (w/K Ic) 5-4 WCAP-17035-NP December 2009 Revision 2 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Forging 05 LIMITING ART VALUES AT 7 EFPY: 1/4T, 61 F 3/4T, 45 F 0 250 500 750 1000 1250 1500 1750 2000 2250 2500050100150200250300350400450500550Moderator Temperature (Deg. F)Calculated Pressure (PSIG)Operlim Version:5.2 Run:30881 Operlim.xls Version: 5.2UnacceptableOperationAcceptableOperationCooldownRates

°F/Hrsteady-state 40-60

-100 Figure 5-2 Watts Bar Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 7 EFPY (without Margins for Instrumentation Errors) Using 1998 through 2000 Addenda App. G Methodology (w/K Ic) 5-5 WCAP-17035-NP December 2009 Revision 2 Table 5-1 7 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology (w/K IC, w/Flange Notch and w/o Uncertainties for Instrumentation Errors) Leak Test Limit 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 105 2000 60 0 122 0 60 0 122 0 105 2000 60 621 122 621 60 621 122 621 122 2485 65 621 122 621 65 621 122 621 122 2485 70 621 122 621 70 621 122 621 75 621 122 621 75 621 122 621 80 621 125 621 80 621 125 621 85 621 130 621 85 621 130 621 90 621 135 621 90 621 135 621 95 621 140 621 95 621 140 621 100 621 140 1256 100 621 140 1128 100 621 145 1314 100 621 145 1160 100 1256 150 1381 100 1128 150 1199 105 1314 155 1458 105 1160 155 1245 110 1381 160 1544 110 1199 160 1298 115 1458 165 1640 115 1245 165 1358 120 1544 170 1748 120 1298 170 1426 125 1640 175 1868 125 1358 175 1503 130 1748 180 2001 130 1426 180 1590 135 1868 185 2149 135 1503 185 1687 140 2001 190 2312 140 1590 190 1795 145 2149 145 1687 195 1915 150 2312 150 1795 200 2048 155 1915 205 2196 160 2048 210 2360 165 2196 170 2360

5-6 WCAP-17035-NP December 2009 Revision 2 Table 5-2 7 EFPY Cooldown Curve Data Points Using 1998 through 2000 Addenda App. G Methodology (w/K IC, w/Flange Notch and w/o Uncertainties for Instrumentation Errors) Steady State 20°F/hr. 40°F/hr. 60°F/hr. 100°F/hr. T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 621 60 621 60 621 60 621 65 621 65 621 65 621 65 621 65 621 70 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 100 1422 100 1422 100 1422 100 1422 100 1422 105 1508 105 1508 105 1508 105 1508 105 1508 110 1603 110 1603 110 1603 110 1603 110 1603 115 1709 115 1709 115 1709 115 1709 115 1709 120 1825 120 1825 120 1825 120 1825 120 1825 125 1954 125 1954 125 1954 125 1954 125 1954 130 2096 130 2096 130 2096 130 2096 130 2096 135 2253 135 2253 135 2253 135 2253 135 2253 140 2427 140 2427 140 2427 140 2427 140 2427 6-1 WCAP-17035-NP December 2009 Revision 2 6 REFERENCES

1. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement 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 XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."
4. Westinghouse Letter LTR-REA-08-105, Revision 2, "Pressure Vessel Design Basis Fluence for Watts Bar Unit 2," M. A. Hunter, dated March 18, 2009.
5. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D. C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
6. WCAP-13830, Revision 1, "Heatup and Cooldown Limit Curves for Normal Operation for Watts Bar Unit 2," J. M. Chicots, et al., February 1995.
7. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Section NB-2300, "Fracture Toughness Requirements for Material."

A-1 WCAP-17035-NP December 2009 Revision 2 APPENDIX A THERMAL STRESS INTENSITY FACTORS (K IT) The following pages contain the thermal stress intensity factors (K It) for the maximum heatup and cooldown rates. The vessel radii to the 1/4T and 3/4T locations are as follows: 1/4T Radius = 88.768" 3/4T Radius = 93.001" A-2 WCAP-17035-NP December 2009 Revision 2 Table A-1 K It Values for 100F/hr Heatup Curve (w/o Margins for Instrument Errors)

Water Temp. ( F) Vessel Temperature @ 1/4T Location for 100F/hr Heatup

( F) 1/4T Thermal StressIntensity Factor (KSI SQ. RT. IN.) Vessel Temperature @ 3/4T Location for 100F/hr Heatup

( F) 3/4T Thermal StressIntensity Factor (KSI SQ. RT. IN.) 60 56.015 -0.994 55.047 0.478 65 58.635 -2.438 55.318 1.443 70 61.728 -3.675 56.029 2.419 75 65.038 -4.846 57.225 3.331 80 68.620 -5.851 58.848 4.142 85 72.314 -6.766 60.858 4.864 90 76.193 -7.556 63.213 5.501 95 80.170 -8.276 65.868 6.069 100 84.280 -8.903 68.790 6.572 105 88.475 -9.472 71.945 7.019 110 92.767 -9.970 75.302 7.416 115 97.129 -10.424 78.839 7.773 120 101.561 -10.823 82.530 8.092 125 106.051 -11.189 86.359 8.379 130 110.592 -11.512 90.308 8.637 135 115.181 -11.809 94.363 8.870 140 119.806 -12.074 98.512 9.080 145 124.470 -12.318 102.742 9.271 150 129.161 -12.537 107.045 9.445 155 133.883 -12.740 111.410 9.603 160 138.625 -12.924 115.832 9.748 165 143.391 -13.096 120.303 9.882 170 148.173 -13.252 124.817 10.005 175 152.974 -13.399 129.369 10.119 180 157.787 -13.534 133.955 10.225 185 162.615 -13.662 138.570 10.324 190 167.452 -13.780 143.211 10.417 195 172.301 -13.894 147.875 10.504 200 177.156 -13.999 152.559 10.587 205 182.021 -14.101 157.261 10.665 210 186.891 -14.197 161.979 10.739 A-3 WCAP-17035-NP December 2009 Revision 2 Table A-2 K It Values for 100F/hr Cooldown Curve (w/o Margins for Instrument Errors)

Water Temp. ( F) Vessel Temperature @ 1/4T Location for 100F/hr Cooldown

( F) 100F/hr Cooldown1/4T Thermal StressIntensity Factor (KSI SQ. RT. IN.) 210 236.000 16.31 205 230.917 16.24 200 225.833 16.18 195 220.750 16.11 190 215.666 16.05 185 210.582 15.98 180 205.497 15.91 175 200.413 15.85 170 195.329 15.78 165 190.244 15.71 160 185.160 15.65 155 180.075 15.58 150 174.990 15.51 145 169.906 15.45 140 164.821 15.38 135 159.737 15.32 130 154.653 15.25 125 149.568 15.18 120 144.484 15.12 115 139.400 15.05 110 134.316 14.99 105 129.232 14.92 100 124.148 14.86 95 119.064 14.79 90 113.981 14.72 85 108.897 14.66 80 103.814 14.59 75 98.731 14.53 70 93.647 14.47 65 88.565 14.40 60 83.483 14.34

B-1 WCAP-17035-NP December 2009 Revision 2 APPENDIX B UPPER SHELF ENERGY EVALUATION Per Regulatory Guide 1.99, Revision 2 [Reference B-1], the Charpy upper shelf energy (USE) is assumed to decrease as a function of fluence and copper content as indicated in Figure 2 of the Guide (Figure B-1 of this report) when surveillance data is not used.

Linear interpolation is permitted.

The 32 EFPY end-of-life (EOL) USE of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the beltline materials, and Figure 2 in Regulatory Guide 1.99, Revision 2. The maximum vessel clad/base metal interface fluence value was used to determine the corresponding 1/4T fluence value at 32 EFPY. The Watts Bar Unit 2 reactor vessel beltline region thickness is 8.465 inches. Per LTR-REA-08-105, Revision 2 [Reference B-2], the maximum vessel clad/base metal interface fluence value is 3.17E+19 n/cm 2 (E > 1.0 MeV). Calculation of the 1/4T vessel surface fluence values at 32 EFPY for the beltline materials is shown as follows: Maximum Vessel Fluence @ 32 EFPY = 3.17E+19 n/cm 2 (E > 1.0 MeV)

1/4T Fluence @ 32 EFPY = (3.17E+19 n/cm

2)
  • e(-0.24 * (8.465 / 4))

= 1.91E+19 n/cm 2 (E > 1.0 MeV)

Table B-1 provides the predicted EOL USE values for 32 EFPY.

B-2 WCAP-17035-NP December 2009 Revision 2 1.010.0100.01.00E+171.00E+181.00E+191.00E+20Neutron Fluence, n/cm 2 (E > 1 MeV)Percentage Drop in USE% Copper Base Metal Weld0.35 0.30 0.30 0.25 0.25 0.20 0.20 0.15 0.15 0.10 0.10 0.05 Upper LimitPeak 32 EFPY 1/4T fluence = 1.91E+19 n/cm 2 (E > 1.0 MeV)

Figure B-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper Shelf Energy as a Function of Copper and Fluence B-3 WCAP-17035-NP December 2009 Revision 2 Table B-1 Predicted Position 1.2 Upper Shelf Energy Values at 32 EFPY Material Weight % Cu 1/4T EOL Fluence (n/cm 2 , E > 1.0 MeV) Unirradiated USE (a) (ft-lb) Projected USE Decrease (d) Projected EOL USE (ft-lb) Intermediate Shell Forging 05 0.05 1.91E+19 (138) 90 (b) 23 (e) 69 Lower Shell Forging 04 0.05 1.91E+19 (162) 105 (b) 23 (e) 81 Intermediate to Lower Shell Circumferential Weld Seam W05 0.05 1.91E+19 127 (c) 23 98 Notes: (a) Information source is CMTR-RV-WBT [Reference B-3]. Reference B-3 reports energy values in units of kgm/cm

2. Per WCAP-9455, Revision 3 [Reference B-4], specimen cross sections are 0.394 in x 0.315 in. The conversion factors used are as follows: 1 kg/cm 2 = 14.223 lb/in 2 1 m = 3.280833 ft (b) According to Reference B-3, the specimens were tested in the strong (tangential) direction. The strong direction initial USE values are listed in parentheses. However, in accordance with the recommendations of NUREG-0800, Revision 1 [Reference B-5], the strong direction values were reduced to 65% in order to approximate the weak (axial) direction values. These values are listed outside the parentheses, and are used in the EOL USE projections. (c) The circumferential weld testing is considered non-directional. Therefore, no percent reduction was performed to determine the initial USE value to be used in the EOL USE projections. (d) Projected USE decreases were calculated in accordance with Regulatory Guide 1.99, Revision 2, Position 1.2. (e) These projected USE decreases were conservatively taken from the base metal 0.10% copper line in Figure 2 of Regulatory Guide 1.99, Revision 2.

USE Conclusion All of the beltline materials in the Watts Bar Unit 2 reactor vessel are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G [Reference B-6]) at 32 EFPY.

B.1 REFERENCES B-1 Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988. B-2 Westinghouse Letter LTR-REA-08-105, Revision 2, "Pressure Vessel Design Basis Fluence for Watts Bar Unit 2," M. A. Hunter, dated March 18, 2009. B-3 CMTR-RV-WBT, Revision 0, "WBT Reactor Vessel Certified Material Test Reports." B-4 WCAP-9455, Revision 3, "Tennessee Valley Authority Watts Bar Unit No. 2 Reactor Vessel Radiation Surveillance Program," B. A. Rosier and B. N. Burgos, September 2009. B-5 NUREG-0800, Revision 1, Section 5.3.2, Branch Technical Position MTEB 5-2, "Fracture Toughness Requirements," U. S. Nuclear Regulatory Commission, Washington, D. C., July 1981.

B-6 Code of Federal Regulations, 10 CFR 50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

C-1 WCAP-17035-NP December 2009 Revision 2 APPENDIX C PRESSURIZED THERMAL SHOCK EVALUATION The PTS Rule, 10 CFR 50.61 [Reference C-1], requires that for each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RT PTS accepted by the NRC for each reactor vessel beltline material at the end-of-life (EOL) fluence of the plant. This assessment must specify the basis for the projected value of RT PTS for each vessel beltline material, including the assumptions regarding core-loading patterns, and must specify the copper and nickel contents and the fluence value used in the calculation. This assessment must be updated whenever there is a significant change in projected values of RT PTS, or upon request for a change in the expiration date for operation of the facility. Changes to RT PTS values are considered significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewed term, if applicable, for the plant. Table C-1 contains the RT PTS calculations for each of the beltline region reactor vessel materials for Watts Bar Unit 2 at 32 EFPY. Table C-1 RT PTS Calculations for the Watts Bar Unit 2 Beltline Materials at 32 EFPY Material CF ( F) 32 EFPY Fluence (n/cm 2 , E > 1.0 MeV)

FF (a) IRT NDT ( F) RT NDT (b) ( F) U (c) (°F) (d) (°F) M (e) ( F) RT PTS (f) ( F) Intermediate Shell Forging 05 31 3.17E+19 1.30 14 40.4 0 17 34 88 Lower Shell Forging 04 31 3.17E+19 1.30 5 40.4 0 17 34 79 Intermediate to Lower Shell Circumferential Weld Seam W05 68 3.17E+19 1.30 -50 88.7 0 28 56 95 Notes: (a) FF = fluence factor = f (0.28 - 0.1 log (f)). (b) RT NDT = RT PTS = CF

  • FF. (c) As indicated in Table 2-1 of this report, the IRTNDT values are measured; hence, according to 10 CFR 50.61, U = 0°F. (d) Per the guidance of 10 CFR 50.61, the base metal = 17°F and the weld metal = 28°F when surveillance data is not utilized. However, need not exceed 0.5*RT NDT. (e) M = Margin = 2 *( U 2 + 2)1/2. (f) RT PTS = IRT NDT + RT PTS + Margin.

PTS Conclusions for RT PTS Values at EOL (32 EFPY) All of the beltline materials in the Watts Bar Unit 2 reactor vessel are below the RT PTS screening criteria values of 270F, for axially oriented welds and plates / forgings, and 300F, for circumferentially oriented welds, (Per 10 CFR 50.61) at 32 EFPY.

C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.

D-1 WCAP-17035-NP December 2009 Revision 2 APPENDIX D EMERGENCY RESPONSE GUIDELINE LIMITS EVALUATION The Emergency Response Guideline (ERG) limits [Reference D-1]

were developed in order to establish guidance for operator action in the event of an emergency situation, such as a PTS event. Generic categories of limits were developed for the guidelines based on the limiting inside surface RT NDT. These generic categories were conservatively generated for the Westinghouse Owners Group (WOG) to be applicable to all Westinghouse plants. The highest end-of-life (EOL) RTNDT for which the generic category ERG limits were developed is 250 F for a longitudinal flaw and 300F for a circumferential flaw. Therefore, if the limiting vessel material has a RT NDT that exceeds 250F for a longitudinal flaw or 300F for a circumferential flaw, plant-specific ERG P-T limits must be developed. The ERG category is determined by the magnitude of the limiting RT NDT value, which is calculated the same way as the RT PTS values were calculated in Appendix C of this report. The material with the highest RT NDT defines the limiting material, which for Watts Bar Unit 2 is the Intermediate to Lower Shell Circumferential Weld Seam W05 (see Table C-1). Table D-1 identifies ERG category limits and the limiting material RT NDT value at 32 EFPY.

Table D-1 Evaluation of Watts Bar Unit 2 ERG Limit Category ERG Pressure-Temperature Limits Applicable RT NDT Value (a) ERG P-T Limit Category RT NDT < 200 F Category I 200F < RT NDT < 250 F Category II 250F < RT NDT < 300 F Category III b Limiting RTNDT Values at 32 EFPY Material RT NDT Value Intermediate to Lower Shell Circumferential Weld Seam W05 95 F Note: (a) Longitudinally oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F. Emergency Response Guideline Limits Conclusion Reviewing the ERG limit categories with the limiting RT NDT value provided in Table D-1 would place Watts Bar Unit 2 in Category I through 32 EFPY.

D.1 REFERENCES D-1 Background Information for Westinghouse Owner's Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Revision 2, April 30, 2005.

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