Text

New England Coalition, Inc. - Contentions 2A and 2B, Volume 1, Prefiled Exhibit NEC-JH_15 (Redacted Version)
	ML082401759
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/28/2008
From:	; New England Coalition
To:	Karlin A, Wendy Reed, Richard Wardwell; Atomic Safety and Licensing Board Panel
	SECY RAS
References
	50-271-LR, ASLBP 06-849-03-LR, Entergy-Intervenor-NEC-JH_15
	Download: ML082401759 (124)
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{{#Wiki_filter:NEC-JH 15 REDACTED VERSION Report No.: SIR-07-130-NPS Revision No.: 0 Project No.: VY-16Q File No.: VY-16Q-401 July 2007 Environmental Fatigue Analysis for the Vermont Yankee Reactor Pressure Vessel Feedwater Nozzles NOTE 7 is document I es vendorpropeta, infor ation Sth in/ rnatio identified r er2vx-P S Project File7 unmb er. n the list o e erenes n-S1 'h refe n res anyed"forn ion in is docun eat where t se refer ices are u 'dare ent so at* inform can be te in a, )rd cc with applica dor proprietary agreements. Prepared for: Entergy Nuclear Operations, Inc. (Contract No. 10150394) Prepared by: Structural Integrity Associates, Inc. Centennial, CO Prepared by: Date: 7/26/2007 M.Q(fn Reviewed by: Date: 7/26/2007 6'j Fý/ .1 Styplis -- Approved by: Date: 7/26/2007 T. J.yiQqrrmann, P.E. V Structural Integrity Associates, Inc. NEC066026

REVISION CONTROL SHEET Document Number: SIR-07-130-NPS

Title:

Environmental Fatigue Analysis for the Vermont Yankee Reactor Pressure Vessel Feedwater Nozzles Client: Entergy Nuclear Vermont Yankee, LLC SI Project Number: VY-16Q Section] Pages JRevision I Date] Comments 1.0 1 1-8 0 07/26/07 Initial Issue 2.0 2-1 4 3.0 3 3-34 4.0 4-1-4-11 5.0 5-1-5-2 6.0 6-1 7.0 7-1 3 v Structural Integrity Associates, Inc. NEC066027

Table of Contents Section Page 1.0 IN T R O D U C T IO N........................................................................................................... 1-1 1.1 G reen's Function M ethodology.................................................................................... 1-2 2.0 FIN ITE ELEM EN T M O DEL.......................................................................................... 2-1 3.0 L O A D D E FIN IT IO N S..................................................................................................... 3-1 3.1 Thermal Loading................................................ 3-1 3.1.1 Heat Transfer Coefficients and Boundar y Fluid Temperatures.......................... 3-1 3.1.2 Green's Function's................................... ........ 3-2 3.1.3 Thermal Transients (for grograin STRESS. EXE)............................................... 3-3 3.2 P ressure L oading........................................................................................................ 3-4 3.3 Piping Loading 3-5 4.0 STRESS AND FATIGUE ANALYSIS RESULTS........................................................ 4-1 5.0 ENVIRONMENTAL FATIGUE ANALYSIS................................................................ 5-1 6.0 C O N C L U SIO N S.............................................................................................................. 6-1 7.0 R E F E R E N C E S................................................................................................................ 7-1 S1R-07-130-NPS, Rev. 0 iii C Structural Integrity Associates, Inc. NEC066028

List of Tables Table Page Table 2-1. Material Properties @ 300'F ......... I.................................................................... 2-2 Table 3-1: Nodal Force Calculation for End Cap Load.............................................................. 3-9 Table 3-2: Maximum Piping Stress Intensity Calculations...................................................... 3-10 Table 3-3: Heat Transfer Coefficients for Region 1 (40% Flow)............................................. 3-11 T able 3-4: B lend R adius T ransients.......................................................................................... 3-12 T able 3-5: Safe E nd T ransient.................................................................................................. 3-13 Table 4-1: Feedwater Nozzle Blend Radius Stress Summary............................. ... 4-3 Table 4-2: Feedwater Nozzle Safe End Stress Summary.......................................................... 4-5 Table 4-3: Fatigue Parameters Used in the Feedwater Nozzle Fatigue Analysis....................... 4-7 Table 4-4: Fatigue Results for Feedwater Nozzle Blend Radius................................................ 4-8 Table 4-5: Fatigue Results for the Feedwater Nozzle Safe End............................................... 4-10 I I I I I I I I I I I I I I I I I I I SIR-07-130-NPS, Rev. 0 iv V Structural Integrity Associates, Inc. NEC066029

List of Figures Fit~ure Page Figure 1-1. Typical Green's Functions for Thermal Transient Stress........................................ 1-7 Figure 1-2. Typical Stress Response Using Green's Functions.................................................. 1-8 Figure 2-1: VY Feedwater Nozzle FEM.............................. .......... 2-3 Figure 2-2: VY Feedwater Nozzle FEM - Safe End/Nozzle Region......................................... 2-4 Figure 3-1: Feedwater Nozzle Internal Pressure Distribution.................................................. 3-14 Figure 3-2: Feedwater Nozzle Pressure Cap Load................................................................... 3-15 Figure 3-3: Feedwater Nozzle Vessel Boundary Condition..................................................... 3-16 Figure 3-4: T herm al R egions.................................................................................................... 3-17 Figure 3-5: Safe End Critical Thermal Stress Location and Linearized Stress Paths............... 3-18 Figure 3-6: Brand Radius Critical Thermal Stress Location and Linearized Stress Paths....... 3-19 Figure 3-7: Safe End Total Stress History for 100% Flow....................................................... 3-20 Figure 3-8: Safe End Membrane Plus Bending Stress History for 100% Flow........................ 3-20 Figure 3-9: Safe End Total Stress History for 40% Flow......................................................... 3-21 Figure 3-10: Safe End Membrane Plus Bending Stress History for 40% Flow........................ 3-21 Figure 3-11: Safe End Total Stress History for 25% Flow....................................................... 3-22 Figure 3-12: Safe End Membrane Plus Bending Stress History for 25% Flow........................ 3-22 Figure 3-13: Blend Radius Total Stress History for 100% Flow.............................................. 3-23 Figure 3-14: Blend Radius Membrane Plus Bending Stress History for 100% Flow.............. 3-23 Figure 3-15: Blend Radius Total Stress History for 40% Flow................................................ 3-24 Figure 3-16: Blend Radius Membrane Plus Bending Stress History for 40% Flow................ 3-24 Figure 3-17: Blend Radius Total Stress History for 25% Flow...... :................... 3-25 Figure 3-18: Blend Radius Membrane Plus Bending Stress History for 25% Flow................ 3-25 F igure 3-19: T ransient 1, B olt-up............................................................................................. 3-26 Figure 3-20: Transient 2, D esign H Y D Test............................................................................. 3-26 F igure 3-2 1: T ransient 3, Startup.............................................................................................. 3-27 Figure 3-22: Transient 4, Turbine Roll and Increased to Rated Power.................................... 3-27 Figure 3-23: Transient 5, Daily Reduction 75% Power............................................................ 3-28 Figure 3-24: Transient 6, Weekly Reduction 50% Power....................................................... 3-28 Figure 3-25: Transient 9, Turbine Trip at 25% Power.............................................................. 3-29 Figure 3-26: Transient 10, Feedw ater Bypass.......................................................................... 3-29 Figure 3-27: Transient 11, Loss of Feedwater Pumps.............................................................. 3-30 Figure 3-28: Transient 12, Turbine Generator Trip.................................................................. 3-30 Figure 3-29: Transient 14, SRV Blow dow n............................................................................. 3-31 Figure 3-30: Transient 19, Reduction to 0% Power................................................................. 3-31 Figure 3-3 1: Transient 20, Hot Standby (Heatup Portion)....................................................... 3-32 Figure 3-32: Transient 20A, Hot Standby (Feedwater Injection Portion)................................ 3-32 Figure 3-33: Transient 21-23, Shutdow n.................................................................................. 3-33 Figure 3-34: Transient 24, H ydrostatic Test............................................................................. 3-33 Figure 3-35: T ransient 25, U nbolt........................................................................................... 3-34 Figure 3-36: External Forces and Moments on the Feedwater Nozzle..................................... 3-34 SIR-07-130-NPS, Rev. 0 v C Structural Integrity Associates, Inc. NEC066030

1.0 INTRODUCTION

In Table 4.3-3 of the Vermont Yankee (VY) License Renewal Application (LRA), the 60-year I cumulative usage factor (CUF) value for the reactor pressure vessel (RPV) feedwater nozzle (FW) is reported as 0.750. Application of an environmentally assisted fatigue (EAF) multiplier, as required for the license renewal period, resulted in an unacceptable EAF CUF value of 2.86. Therefore, further refined analysis was necessitated to show acceptable EAF CUF results for this component. REDACTED 3 The VY FW nozzles were re-evaluated in detail by SI in 2004 for EPU and 60 years of operation. However, that analysis used conservative transient definitions and cyclic projections for 60 years of operation that have since been updated as a part of LRA development. This report documents a refined fatigue evaluation for the VY FW nozzle. The intent of this evaluation is to use refined transient definitions and the revised cyclic transient counts for 60 years for a computation of CUF, including EAF effects, that is more refined than previously performed fatigue analyses. The fatigue-linmiting locations in the FW nozzle and safe end are included in the evaluation, to be consistent with NUREG/CR-6260 [16] needs for EAF evaluation for license renewal. The resulting fatigue results will be used as a replacement to the values previously reported in the VY LRA. I I SIR-07-130-NPS, Rev. 0 1-1 Structural Integrity Associates, Inc. This Page Contains Reference to Vendor Proprietary Information (such information is marked with a "bar" in the right-hand margin) NEC066031

The refined evaluation summarized in this report included development of a detailed finite element model of the FW nozzle, including relevant portions of the safe end, thermal sleeve, and the RPV wall. Thermal and pressure stress histories were developed for relevant transients affecting the FW nozzle, including any effects of EPU, as specified by the VY RPV Design Specification [3], the VY EPU Design Specification [17] and other boiling water reactor (BWR) operating experience. The thermal and pressure stress histories were used to determine total stress and primary plus secondary stress for use in a subsequent fatigue evaluation. Stresses were also included due to loads from the attached piping for application in the stress/fatigue analysis based on the bounding reaction loads obtained from the relevant design documents. The revised fatigue calculation was performed using Section III methodology from the 1998 Edition, 2000 Addenda of the ASME Code [15], and was performed using actual cycles from past plant operation projected out to 60 years of operation. 1.1 Green's Function Methodology In order to provide an overall approach and strategy for evaluating the feedwater nozzle, the Green's Function methodology and associated ASME Code stress and fatigue analyses are described in this section. Revised stress and fatigue analyses are being performed for the feedwater nozzle using ASME Code, Section III methodology. These analyses are being performed to address license renewal requirements to evaluate environmental fatigue for this component in response to Generic Aging Lessons Learned (GALL) Report [22] requirements. The revised analysis is being performed to refine the fatigue usage so that an environmental fatigue factor can be determined for subsequent license renewal efforts. Two sets of rules are available under ASME Code, Section III, Class 1 [15]. Subparagraph NB-3600 of Section III provides simplified rules for analysis of piping components, and NB-3200 allows for more detailed analysis of vessel components. The NB-3600 piping equations combine by absolute sum the stresses due to pressure, moments and through wall thermal gradient effects, regardless of where within the pipe cross-section the maximum value of the components of stress 1-2 V Structural Integrity Associates, Inc. $IR-07-130-NPS, Rev. 0 NEC066032

are located. By considering stress signs, affected surface (inside or outside) and azimuthal position, the stress ranges can be significantly reduced. In addition, NB-3600 assigns stress indices by which the stresses are multiplied to conservatively incorporate the effects of geometric discontinuities. In NB-3200, these are not required, as the stresses are calculated by finite element analysis and any applicable stress concentration factors. This generally results in a net reduction of the stress ranges and consequently, in the fatigue usage. Article 4 [27] methodology was originally used to evaluate the feedwater nozzle. NB-3200 methodology, which is the modern day equivalent to Article 4, is used in this analysis to be consistent with the Section III design bases for this component, as well as to allow a more detailed analysis of this component. In addition, several of the conservatisms originally used in the original feedwater nozzle evaluation (such as grouping of transients) are removed in the current evaluation so as to achieve as accurate a CUF as reasonably achievable. For the feedwater nozzle evaluated as a part of this work, stress histories will be computed by a time integration of the product of a pre-determined Green's Function and the transient data. This Green's Function integration scheme is similar in concept to the well-known Duhamnel theory used in structural dynamics. A detailed derivation of this approach and examples of its application to specific plant locations is contained in Reference [4]. A general outline is provided in this section. The steps involved in the evaluation are as follows: Develop finite element model Develop heat transfer coefficients and boundary conditions for the finite element model Develop Green's Functions Develop thermal transient definitions " Perform stress analysis to determine stresses for all thermal transients

Perform fatigue analysis SJR 30-NPS, Rev. 0 1-3 Structural Integrity Associates, Inc.

NEC066033

A Green's Function is derived by using finite-element methods to determine the transient stress response of the component to a step change in loading (usually a thermal shock). The critical location in the component is identified based on the maximum stress, and the thermal stress response over time is extracted for this location. This response to the input thennal step is the "Green's Function." Figure 1-1 shows a typical set of two Green's Functions, each for a different set of heat transfer coefficients (representing different flow rate conditions). To compute the thermal stress response for an arbitrary transient, the loading parameter (usually local fluid temperature) is deconstructed into a series of step-loadings. By using the Green's Function, the response to each step can be quickly determined. By the principle of superposition, these can be added (algebraically) to determine the response to the original load history. The result is demonstrated in Figure 1-2. The input transient temperature histoiy contains five step-changes of varying size, as shown in the upper plot in Figure 1-2. These five step changes produce the five successive stress responses in the second plot shown in Figure 1-2. By adding all five response curves, the real-time stress response for the input thermal transient is computed. The Green's Function methodology produces identical results compared to running the input transient through the finite element model. The advantage of using Green's Functions is that many individual transients can be run with a significant reduction of effort compared to running all transients through the finite element model. The trade-off in this process is that the Green's Functions are based on constant material properties and heat transfer coefficients. Therefore, these parameters are chosen to bound all transients that constitute the majority of fatigue usage, i.e., the heat transfer coefficients at 300'F bound the cold water injection transient. In addition, the instantaneous value for the coefficient of thermal expansion is used instead of the mean value for the coefficient of thermal expansion. This conservatism is more than offset by the benefit of not having to analyze every transient, which was done in the VY reactor feedwater nozzle evaluation. SIR 30-NPS, Rev. 0 1-4 Structural Integrity Associates, Inc. NEC066034

Once the stress history is obtained for all transients using the Green's Function approach, the remainder of the fatigue analysis is carried out using traditional methodologies in accordance with ASME Code, Section III requirements. Fatigue calculations are performed in accordance with ASME Code, Section III, Subsection NB-3200 methodology. Fatigue analysis is performed for the three limiting locations (two in the safe end and one in the nozzle forging, representing the three materials of the nozzle assembly) using the Green's Functions developed for the three feedwater flow conditions and 60-year projected cycle counts. 3 Three Structural Integrity utility computer programs are used to facilitate the fatigue analysis process: STRESS.EXE, P V.EXE, and FATIGUE.EXE. The first program, STRESS.EXE, calculates a stress history in response to a thermal transient using a Green's Function. The second program, P-V.EXE, reduces the stress history to peaks and valleys, as required by ASME Code fatigue evaluation methods. The third program, FATIGUE.EXE, calculates fatigue from the reduced peak and valley history using ASME Code, Section III range-pair methodology. All three programs are explained in detail and have been independently verified for generic use in the Reference [14] calculation. In order to perform the fatigue analysis, Green's Functions are developed using the finite element model. Then, input files with the necessary data are prepared and the three utility computer programs are run. The first program (STRESS.EXE) requires the following three input files: I Input file "GREEN.DAT": This file contains the Green's Function for the location being evaluated. For each flow condition, two Green's Functions are determined: a membrane plus bending stress intensity Green's Function and a total stress intensity Green's Function. This allows computation of total stress, as well as membrane plus bending stress, which is necessary to compute K, per ASME Code, Section III requirements. 1-5 Structural Integrity Associates, Inc. SIR-07-130-NPS, Rev. 0N NEC066035

Input file "GREEN.CFG": This file is a configuration file containing parameters that define the Green's Function (i.e., number of points, temperature drop analyzed, etc.).

" Input file "TRANSNT.INP": This file contains the input transient history for all thermal transients to be analyzed for the location being evaluated. Pressure and piping stress intensities are also included for each transient case, based on pressure stress results from finite element analysis and attached piping load calculations. The second program (P-V.EXE) simply extracts only the maxima and minima stress (i.e., the peaks and valleys) from the stress histories generated by program STRESS.EXE. The third program (FATIGUE.EXE) performs the ASME Code peak event-pairing required to calculate a fatigue usage value. The input data consists of the output peak and valley history from program P-V.EXE and a configuration input file that provides ASME Code configuration data relevant to the fatigue analysis (i.e., K, parameters, Sin, Young's modulus, etc.). The output is the final fatigue calculation for the location being evaluated. The Green's Function methodology described above uses standard industry stress and fatigue analysis practices, and is the same as the methodology used in typical stress reports. Special approval for the use of this methodology is therefore not required. 1-6 U Structural Integrity Associates, Inc. SIR-07-130-NPS, Rev. 0 NEC066036

0) a-7C) a) a3) 450 400 350 300 250 200 Time (sec) 92825TO Note: A typical set of two Green's Functions is shown, each for a different set of heat transfer coefficients (representing different flow rate conditions). Figure 1-1. Typical Green's Functions for Thermal Transient Stress I I I 1-7 SIR-07-130-NPS, Rev. 0 V Structural Integrity Associates, Inc. NEC066037

a 250. in y 25 20. "*15' 0 4. -10" ir ~ ~

o ft'dwdS m

]mA / /

-{~

SIrm mm i hsw of kddva Skpa Figure 1-2. Typical Stress Response Using Green's Functions 1-8 C Structural Integrity Associates, Inc. SIR-07-130-NPS, Rev. 0 NEC066038

I 2.0 FINITE ELEMENT MODEL A previously generated ANSYS [5] finite element model (FEM) of the VY feedwater nozzle and safe end was used to perform the updated stress and fatigue analyses. The details of the model development are documented in the Reference [6] calculation. A few key points with respect to model development are as follows: " The model is identical to the geometry and mesh of the model previously developed for feedwater nozzle fracture mechanics work perfonned for VY [7]. The boundary condition corresponding to the location of the start of the thermal sleeve in the FEM are consistent with Reference [8]. The materials of the various components of the model are listed below: I

Reactor Pressure Vessel - SA533 Grade B.

I

Reactor Pressure Vessel Cladding - Stainless Steel
Nozzle Forging - ASTM A508 Class II I
Safe End Forging - ASTM A508 Class I
Feedwater Piping - ASTM A106 Grade B I

The FEM model the radius of RPV was increased by a factor of two to account for the fact that the vessel portion of the finite element model is a sphere and the actual geometry is a cylinder. I Material properties were based upon the 1998 ASME Code, Section II, Part D, with 2000 Addenda [9], and are shown in Table 2-1. The properties were evaluated at an average temperature of 300'F. This average temperature is based on a thermal shock of 500'F to 100°F which was applied to the FEM model for Green's Function development. The finite element model is shown in Figures 2-1 and 2-2. SIR-07-130-NPS, Rev. 0 2-1 Structural Integrity Associates, Inc. NEC066039

nal 2-1. m M P roperi m m Table 2-1. Material Properties @~ 3000 F (1 Instantaneous Young's Coefficient of

Density, Conductivity, Specific Heat, Poisson's Material

Modulus, Thermal p

k d Rt [dent. E x 106 Expansion, (lb/in3) k (ft2/hr) (BTU/Ibm-dF) Ratio (psi) a X 10T6 (assumed) (see Note 5) (assumed) (in/in-0 F) SA533 Grade B, A508 Class II 26.7 7.3 0.283 23.4 0.401 0.119 0.3 (see Note 2) SS Clad 27.0 9.8 0.283 9.8 0.160 0.125 0.3 (see Note 3) A508 Class 1 28.1 7.3 0.283 32.3 0.561 0.118 0.3 (see Note 4) A106 Grade B 28.3 7.3 0.283 32.3 0.561 0.118 0.3 (see Note 4) Notes

1. The material properties applied in the analyses are taken from ASME Section II Part D 1998 Edition with 2000 Addenda. This is consistent with information provided in the Design Input Record (page 13 of VY EC No. 1773, SI File No. VY-16Q-209). The use of a later code edition than that used for the original design code is acceptable since later editions typically reflect more accurate material properties than was published in prior Code editions. Material Properties are evaluated at 300'F from the 1998 ASME Code, 2000 Addenda, Section II, Part D [9],

except for density and Poisson's ratio, which are assumed typical values.

2.

Properties of A508 Class II are used (3/4Ni-1/2Mo-1/3Cr-V).

3. Properties of 18Cr-8Ni austenitic stainless steel are used.

4.

Composition = C-Si.

5. Calculated as k/(pd)/1 23.

SIR-07-130-NPS, Rev. 0 2-2 V Structural Integrity Associates, Inc. NEC066040

Figure 2-1: VY Feedwater Nozzle FEM SIR-07-130-NPS, Rev. 0 2-3 V Structural Integrity Associates, Inc. NEC066041

ELEMENTS Feedwater Nozzle Finite Element Model SFP 6 2 C0-16:25 :42 Figure 2-2: VY Feedwater Nozzle FEM - Safe End/Nozzle Region SIR-07-130-NPS, Rev. 0 2-4 C Structural Integrity Associates, Inc. NEC066042

3.0 LOAD DEFINITIONS The pressure and thermal stresses for the feedwater nozzle for the revised fatigue evaluation were developed using the axisymmetric FEM model described in Section 2.0 of this report. The details of the Green's function development and associated stress evaluation are documented in the Reference [10] calculation. 3.1 Thermal Loading U Thermal loads are applied to the feedwater nozzle model. The heat transfer coefficients after I power uprate were determined in Reference [10]. These values were determined for various regions of the finite element model and for 100% (4,590 GPM), 40% (1836 GPM) and 25% (1,148 GPM) [10]. The annulus leakage flow rate is assumed to be 25 GPM for non-EPU conditions and 31 GPM for EPU conditions. The 25 GPM value is calculated by scaling the 23 GPM [Page 6, 13] value up by approximately 9%. The 23 GPM value is scaled up to provide some conservatism and allow for inaccuracies in the determination of leakage flow. The 31 GPM value is calculated by multiplying the 25 GPM value by 1.25 [Page 6, 13]. Based on this, the annulus leakage flow rate is assumed to be 8 GPM for EPU conditions with 25% flow rate and 13 GPM for EPU condition with 40% flow rate. The temperatures used are based upon a thermal shock from 500°F to 100°F. 3.1.1 Heat Transfer Coefficients and Boundaty Fluid Temperatures Referring to Figure 3-4, heat transfer coefficients were applied as follows: " The heat transfer coefficient for the outside surfaces of the FEM (Region 8) was a constant value of 0.2 BTU/hr-ft2-°F (3.858x10-7 BTU/sec-in 2-°F).

Table 3-3 shows a sampling of the heat transfer coefficient calculations for Region 1 for the 40% flow case.

For all Green's Functions, a 500'F to 100°F thermal shock was run to determine the stress response. SIR-07-130-NPS, Rev. 0 3-1 Structural Integrity Associates, Inc. NEC066043

The applied heat transfer coefficients and the initial temperatures for all regions are contained in Reference [10]. 3.1.2 Green's Funclion's Three flow dependent thermal load cases were run on the FEM model with the heat transfer coefficients and the fluid temperature conditions listed above. Two locations were selected for analysis (see Figures 3-5 and 3-6):

1. The critical safe end location was chosen as the node with the highest stress intensity due to thermal loading under high flow conditions. The highest stress intensity due to thermal loading occurred at Node 192 (see Figure 3-5), on the inside diameter of the nozzle safe end, and therefore, this node was selected for analysis. Because the safe end stress response is affected by flow, three flow conditions were analyzed (100%, 40% and 25%).

2. The critical blend radius location was chosen, based upon the highest pressure stress.

Conservatively assuming the cladding has cracked, the critical location is selected as node 657 at base metal of the nozzle, as shown in Figure 3-6. Because theblend radius stress response is affected by flow, three flow conditions were analyzed (100%, 40% and 25%). Two stress intensity time history were developed for each location and each flow case: (I) total stress intensity, and (2) membrane plus bending stress intensity. The stress time histories for the safe end location, where the maximum stress was obtained for each of the flow conditions, are shown in Figures 3-7 through 3-12. The stress time histories for the blend radius location, where the maximum stress was obtained for each of the flow conditions, are shown in Figures 3-13 through 3-18. SIR-07-130-NPS, Rev. 0 3-2 Structural Integrity Associates, Inc. NEC066044

I I I 3.1.3 Thermal Transients (for program STRESS.EXE) The program STRESS.EXE requires the following three input files for analyzing an individual transient: " Green.dat. There are 12 stress history functions obtained from Reference [10]. They represent the membrane plus bending and total stress intensities at the blend radius and safe end locations. Both of the blend radius and the safe end have two stress history functions for each of the following flow conditions; 100%, 40%, and 25% flow. Green.cfg is configured as described in Reference [14].

Transnt.inp. These files are created to represent the transients shown on the thermal cycle diagrams and redefined by power uprate. Note that transients 12, 13, and 15 are nearly identical on the thermal cycle diagram [19] and the results from running transient 12 will be used for all three transients. Transient 16, 17 and 18 will not be considered since there is no temperature change. Tables 3-4 and 3-5 show the thermal history used to represent each transient. Based upon the thermal cycle diagram for the feedwater nozzle [ 19], the transients are split into the following groups based upon flow rate:

o Transients 3, 20, 20A, and 21-23 are run at 25% flow. Although Reference [19] shows 15% flow rate, it is conservative to use 25% flow rate for these transients. Transient 20, Hot Standby, is split up into two parts. The. first portion is "Heatup portion" and the second portion is "Feedwater Injection portion" that are defined from Reference [19]. o Transient 11 is run at 40% flow. Transient 1 1 starts off and ends at 100% flow. o Transients 5, 6, 9, 10, and 19 are run at 100% flow. o Transient 4 is run at 100% flow only to obtain the last stress point. The remainder of the stress points for transient 4 is obtained from the 25% flow stress results. The results are pulled from the two flow case results based upon the flow rates defined in the thermal cycle diagram [19]. o Transients 12, 13, 14 and 15 were run at 100% flow. Heat transfer coefficients were not re-calculated for the 1 minute intervals each of these transients is at 110% flow. The effect of this small flow rate increase for such a relatively short duration should be minor. SIR-07-130-NPS, Rev. 0 3-3 Strmctural Intermritv Assn I I I I I I I I I I I I I I riRtP~ Inc UI NEC066045

o Transients 1, 2, 24, and 25 are set as no thermal stress due to very small temperature changes (70'F to 100'F) at these transients. 3.2 Pressure Loading A uniform pressure of 1,000 psi was applied along the inside surface of the feedwater nozzle and the vessel wall. A pressure load of 1,000 psi was used because it is easily scaled up or down to account for different pressures that occur during transients. In addition, a cap load was applied to the piping at the end of the nozzle. The nodal forces shown in Table 3-1 [10] are defined by the following equation: ý,tenlefl = ;T z(IR )R2 P

R i ) )

SP(OR 2 _ IR12 where: P unit pressure load = 1,000 psi IR inner pipe radius = 4.8345 in OR = outer pipe radius = 5.42 in R = inside radius of element that node is attached to R, outside radius of element that node is attached to Fnocle average of the element forces on either side of the node. Note: The force on the innermost and outermost nodes is calculated as one half oJ the J brce on the element that they are attached to. The calculated nodal forces were applied as positive values so they would exert tension on the end of the model. Figures 3-1, 3-2, and 3-3 show the internal pressure distribution, cap load, and symmetry condition applied to the vessel end of the model, respectively. The pressure stress associated with a 1000 psi internal pressure was determined in Reference [10]. These values are as follows: Pressure stress for the safe end: SIR-07-130-NPS, Rev. 0 3-4 Structural Integrity Associates, Inc. NEC066046

8693 psi membrane plus bending stress intensity. 8891 psi total linearized stress intensity. I Pressure stress for the blend radius: 36653 psi membrane plus bending stress intensity. 37733 psi total linearized stress intensity. I These pressure stress values for each location were linearly scaled with pressure. The actual pressure for column 6 of Tables 4-1 and 4-2 is obtained from Tables 3-4 and 3-5. The scaled 3 pressure stress values are shown in columns 7 and 8 of Tables 4-1 and 4-2. The pressure stress is combined with the thermal and piping loads to calculate the final stress values used for fatigue analysis. The piping load sign is set as the same as the thermal stress sign. I 3.3 Piping Loading I Additionally, the piping stress intensity (stress caused by the attached piping) was determined. These piping forces and moments are determined as shown in Figure 3-36. The following formulas are used to determine the maximum stress intensity in the nozzle at the two locations of interest. From engineering statics, the piping loads at the end of the model can be translated to the first and second cut locations using the following equations: ( M, )I = Mx - F.) Ll For Cut I: (My)L = MY. + t.L (M) 2 =M, - F,L I For Cut II: (My,)2 = + 2 SIR-07-130-NPS, Rev. 0 3-5 Structural Integrity Associates, Inc. NEC066047

The total bending moment and shear loads are obtained using the equations below: = ( ) 2( ),2 For Cut I: FM, = +(F,),2 For Cut II: =(F) 2 +(Fr) 2 The distributed loads for a thin-walled cylinder are obtained using the equations below: N_= 1F+~2R RN N qv = iR--- " 2R To determine the primary stresses, PM, due to internal pressure and piping loads, the following equations are used. For Cut I, using thin-walled equations: SIR-07-130-NPS, Rev. 0 3-6 C Structural Integrity Associates, Inc. NEC066048

2 tx tN (a) RPI !N SI R= MAX + (7, or SMAX =j(P) (P) +(zA)0I where: L1 The length from the end of the nozzle where the piping loads are applied to the location of interest in the safe end. The length from the end of the nozzle where the piping loads are applied to the location of interest in the blend radius. Mxy = The maximum bending moment in the xy plane. Fyx = The maximurn shear force in the xy plane. N, = The normal force per inch of circumference applied to the end of the nozzle in the z direction. qN = The shear force per inch of circumference applied to the nozzle. 3 RN The mid-wall nozzle radius. Because pressure was not considered in this analysis, the equations used for Cut I are valid for Cut II. Furthermore, since the pressure was not considered in this analysis, the equations can be simplified as follows: SIR-07-130-NPS, Rev. 0 3-7 Structural Integrity Associates, Inc. NEC066049

Nz ( )'9 0) (,) - 0 q,=, t~v SIWAX = 2(=r + ) () 51 2 Per Reference [11], the feedwater nozzle piping loads are as follows: F, = 3,000,lbs M.,= 28,000 ft-lb = 336,000 in-lb F, = 15,000 lbs MY = 13,000 ft-lb = 156,000 in-lb F, = 3,200 lbs Mz = 40,000 ft-lb = 480,000 in-lb The loads are applied at the connection of the piping and safe end. Therefore, the Ll is equal to 12.0871 inches and the L2 is equal to 27.572 inches. The calculations for the safe end and blend radius are shown in Table 3-2. The first cut location is the same as the Green's Function cross section per [10] at the safe end, and the second cut is from Node 645 (outside) to Node 501 (inside). The maximum stress intensities due to piping loads are 5707.97 psi at the safe end and 265.47 psi at the blend radius, respectively. These piping stress values are scaled assuming no stress occurs at an ambient temperature of 70'F and the full values are reached at reactor design temperature, 575°F. The scaled piping stress values are shown in columns 9 and 10 of Tables 4-1 and 4-2. Columns 11 and 12 of Tables 4-1 and 4-1 show the summation of all stresses for each thermal peak and valley stress point. SIR-07-130-NPS, Rev. 0 3-8 Structural Integrity Associates, Inc. NEC066050

Table 3-1: Nodal Force Calculation for End Cap Load Node Element Radius A Radius Ro2-Ri2 Felement Fnode Number Number (in) (in) (in (Ib) (lb) 1 5.42 7678.0 1022 0.1171 1.25565 15356.1 2 5.3029 15188.4 1021 0.1171 1.22823 15020.7 3 5.1858 14853.0 1020 0.1171 1.20080 14685.3 4 5.0687 14517.6 1019 0.1171 1.17338 14349.9 5 4.9516 14182.2 1018 0.1171 1.14595 14014.5 6 4.8345 1 7007.3 S1R-07-130-NPS, Rev. 0 3-9 V Structural Integrity Associates, Inc. NEC066051

Table 3-2: Maximum Piping Stress Intensity Calculations Safe End External Piping Loads Parameters Fx 3.00 kips Fy= 15.00 kips Fz 3.20 kips Mx= 336,00 in-kips MV= 156.00 in-kips Mz = 480.00 in-kips OD= 11.86 in ID= 10.409 in RN= 5.57 in L = 12.09 in tN = 0.72 in (Mx)i = 154.69 in-kips (MyIJ = 192.26 in-kips M_ V = 246.77 in-kips F,( = 15.30 kips Nz = 2.63 kips/in qN= -1.59 kips/in Primary Membrane Stress Intensity PMz = 3.63 ksi T = -2.20 ksi Simax = 5.71 ksi Simax = 5707.97 psi Blend Radius External Piping Loads Pa ra meters Fx 300 kips FY= 15.00 kips Fz -- 3,20 kips MX= 336.00 in-kips my= 156.00 in-kips Mz= 480.00 in-kips OD= 22.67 in ID= 10.750 in RN= 8.35 in L = 27.57 in tN = 5.96 in (Mx)2 = -77.58 in-kips (M) 2 = 238.72 in-kips MXV = 251.01 in-kips FY = 15.30 kips Nz= 1.21 kips/in qN -0.51 kips/in Primary Membrane Stress Intensity PMz= 0.20 ksi -U = -0.09 ksi Simax= 0.27 ksi SImax = 265.47 psi Note: The locations for Cut I and Cut II were defined in Reference radius paths, respectively. [10] for safe end and blend Structural Integrity Associates, Inc. S1R-07-130-NPS, Rev. 0 3-10 NEC066052

I I I Table 3-3: Heat Transfer Coefficients for Region 1 (40% Flow) Calculation of Heat Transfer Coefficients for Feedwater Nozzle Pipe Inside Oiameter, D = 6 inches 0.806 ft 0,246 In Flow Path I 100%. ralne (town = *:*;59 / *gpe Flow0 % of rated

4CM, 9 7=.

r*,:.,-.53............... Fluid Velocity, V 8.022 tt/sec = 1,836.0 9Pm 0,793742524 Mlb/hr Charxteisfic Length, = - D 0.806 It _ 0.246 m T_, - T-,_... AT assumed to be 12% of fluid temperature = 8.40 12.00 24.00 36.00 48.00 60.00 72.00 'F =nn. 1 i 4.67 6.67 13.33 20.00 26.67 33.33 40.00 C I u at. FluidTe era..lo/ a 026 It.; a P ur, n s Cooor~tion 70 100 200 300 400 000 800 F Factor 1241 21 11 177 9333 14 AR 204 44 P6000 31no 5qC Water Property k e---.-d * !.y!t........ C. ................ S ci H e t . *o .i. (Vo oelc Rate of Expanfsion) .. ýv o!ý q. .R* e... F E....... .*.

q 9

(Gtavitationat Constant) t * * !... Pr (Prandtt Numrberl 4.1869 10.018 1,8 0.3048 1.4881 0.5997 0.63G0 0.6784 0.6836 0.6611 0.6040 0.5071 8.40 0300 .320? 0.380 0.80 .48 .23 4.185 4.179 4.229 4.313 4.522 4.982 6.322 1:£00.......................

0. 9 8

.................... 01 0.1 ...................... -i O3£2 I..M8............. ............... 1.19 0......... ........ =5i0...... 997.1 994.7 962.7 917.8 858.6 784.9 679.2 62.3 02.1 60.1 57.3 036 49.0 42.4 I.. I., I......

0.......

.... :3..........

s. =...

........

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

........4:............ 1.89E-04 3.24E-04 6.66E-04 1.01E-03 1.40E-03 1.98E-03 3.15E-03 1.05E-04 1.80E-04 3.708-84 5.60E-04 7.80E-04 1.10E-03 1.708-03 ...,.o .* E. ! I 4-0 : E. ~ - 9.806 9.806 9.806 9.806 9.806 9.806 9.806

32. 17 32.17 32.17 32.17 32.17 32.17 32.17 9.68E-04 8.828-04 3.07 04 1..38-04 1.380-04 1.84, 04 8.82E-05

. 09 04.............

0. 4..............

1 2_ 0 ............... 30 4 9.3 0 0 .5............ 05 6.980 4.510 1.910 1.220 0.950 U.859 1.070 8/ / - t' F .!':tf kJ4/g-*C -11y!e kg/m-s [ !. 09/10-s I I I I I I Catculated Parameter Formula 70 t00 200 300 400 S0 600 'F Reynold's Number, Re WVD/u 6.0147E:05 8.7645E805 1.88598E06 2.8491E-06 3.7255E+06 4.5248E006 4.7330E806 Grashof Number', G, gfATL /";) 1.2852E 08 6 6834E÷08 1.2721E÷10 6.59 1BE-10 2.0931E 1 1 5.44 29E +11 1.1372E.12 Roylei/h Number, Ra GrPr 8.9710E803 3.0142E+09 2.4297E010 8.0420E+10 1.9885E81I1 4.6755E+11 1.2108E812 From [24]: Inside Surtace Forced Convection Heat Transfer Coelficient: H-, = 0.023ReP1' kItD 5.132.76 6119.10 8,626.61 10,107.53 108960.57 11,236.63 10,678.39 WlmrC D930 ,77'.086 '1,51.026,- 1700-I803--//-A982' 80 Btt,/hr-t'-Fl 1.7448:03, .2.0798-03. -2.931 E-03 7 2-3/434E-03./3.724E.O3 /.".. 3;817E-03</1*,3 3.628E 03?./: Btu/sec-6n'-°F From [24J: Inside Surface NaturalI Convection Heat Transfer Coefficient: CaseI Enclosed cylinder C 0= -n nc,~ t pd Y H-., C(GrPr)"k8_/ 232.43 330.57 599.85 810.28 998869 ,118.54 1192.73 Wlrn-'C u7.09661

5.

I122L-04 2.039E 04 3.7700 1 359E U,. Onisec-in'.'F I 1 I I I I SR0-30-NPs, Re.o 3-1 Structural Integrity Associates, Inc. NEC066053 I

Table 3-4: Blend Radius Transients T- -, l he T "ioeSlep ,*sc I T1ansiseI C IcsT p IYoSt ]Pressure N -rfs.-, Ti.-n I T p Ti-TeStep P -lss, 00 10 0TLoso JLE I 5 20 N.000 00 L EO 100

0L2 1 0p4C005p i

(0 113 2 Deseost..... -0D Tesl 020 iCycle 0 100 1300 100 0280 1 30 1.. 0860 100* 1000 3000* I o6 ....10..... 50o 70 Cyle lU w000e 23070 0 7070 32 10 1 .1010 50 "0101 HF 100 5 3 '000 5000 "!1 MF 40l HF 0. 300 C;ycles 106 00 164 1010 4 25 210 540 5001 1010 Trled 1l01 100 1010 ...... 85h 1........ 02i o 200 Ig 1010io 300 Coes 362 342 00 1010 LF_25, HF I00 6602 342 5030 1010 S 0ail0 0 3.2 0010 75P.10700 2704 01 W 44 11 mm ;*6o CIO .. Fo i6 ......... 5 00000 Cycles 3800 002 y0 61000 ......HF~i 000 000 002 5000 '1000.... 51eky eul 000 302 600 1010 003.........4"02 900 010... 4.60 Ic12 l0 2121 5 0 1.23 0 " 0 1 W60135 410 ,ý672. 6 .7265 5 20014 1315 5 1135 1 001105 0 1135 0260 " 1015 ..0.. 675... 3500 202i.... 20: 1 Salyb, 0 265 1010 OHcap Porhton) 1 440 1 1010 34 yls 3925 40 3424 16016 LF-25 1325 04 000 103 20A I 0,1 St100by 0 540 1010 0701 Inj55.lion1 Pot0ie) 100 1 1010.. 304 Cycles 100 000 1010 LF25 241 280 66 1010 5451 044 000 10D13 21-230 Shu.own 0 549 1010 3 00l£y~les 620 375 0204 50 LF25 0000 330 60 50 1514 1008-50 301 2644 100 500 0 24 Hy 01Ili1 0 1001 Tes 000 100 00 16 SO0clO ZO 100o 0060 1 563 0402bot 1100 013 50 05.~.0 1001 0 ............. 1002..... 1040 .0 0082o 10 5000 0 19.801100on,100% 0 392 1010 "'0 C-l 265 1000 1011-

0. 05lO~~58 13,p SI 25%

0008, 44141 00100 10803.

265 0020 20 30.. 0 20" 80 1o10 000 ) 1010 "00 101. O 0... 1010..... 000 10 12> 7,,sbeo Cd I 0 0 01 10 t Sep 00 Cycles Fl 44014, 040,40 05 50 10 15 01(10, 0 392 2091 260 o 1037 .1000 [ 043.... Note: 1. The indicated time orpressitre was assumed.

2. 1375 psi isf6r Transient 13 only.

I SIR-07-130-NPS, Rev. 0 3-12 V Structural Integrity Associates, Inc. NEC066054

Table 3-5: Safe End Transient Tras.,,.0 l T,1e Temno : T Se.ep Preossure Tra1set 1 Tie Teelp T0 e1tep:71ressure 1 ra 11,--, T T,:!

TsoTSm P, r*sse

[;* '4, 4 1 o.... ... 0 123 Cysles 10 70 10 0 I7, FQ Hle" 0210,1 200 2570.. 9201 ISO 1010 Olss's5os,- I COslel 03--so -3 Do 500 50 12, Cycles 0 137 100 ...5 00 1...... 0 .. 232100i~ 1000 0005 so VF - H "3-3 F 004"4,4-... 35li's ~ 0 1000 33' olo 0 03 530 103 1010 LF2 1 3 25,0 500 100 O OurC1. e R4.II 0 1.10 1010 54443*e**' 400...1 10 100...... 100i6b... 5 Da 0y 392 100 o.i01

2. 4F 1*

4l 3100 202 500 1C01 6.* el.e: 1 0 302 i 040 73 2700A* 3402 1B 140,0 3.. 3...2. '* 900*....... 1o 2165 224

W275 744.

1100 1 210149 1100 005 500 50 300 1100 559 500 II lOsO .... 0 l 43s.... 4320 4435 3600 0,2 1.003, 07,1 : 132 4040.... -4M2 s,,S 3o. 123 5 4310 F 4425

500 IDIC.

2T)0 1H04 04ar1,41v 4 4 1310 -FW Irl '4t t1 1 31 K00 1340 1o 2*4 0 100 V00 50 H",l orx....... s"o: 00005 50 1 We00 1503 203..14.40.....0. 11 .Q I "5 0 5 425 Ll1 l '0 032 0 O '15411-. 3e 2414n C 4003 30 00 1310 ...... F :i:.....

2. 323:......

00.*551 Th 2 ....3 ; Y 0....... 4'*4 4105 365 2525 00 405 43 5400 232 5905 392 so. ...B1000 500*" 1010 0311,4 049401 TI's. 00 Cecloc 15 5495. 24 00,004311,54, 0~- 422. 15 ........ 0.. 3W2 392 900 1040 130 44. I I I I I I I I I I I I I I I I I Note:

1. These transients are the same as in Table 3-4 with the exception of the 500 second steady state time increment that is used. The transients in Table 3-4 are plotted using a 5000 second steady state increment. The difference is due to the length of the Green's Function for the safe end wvihich is shorter compared to the blend Radius.

2. The indicated time or pressure was assumed.

3. 1375 psi isjbr Transient 13 only.

SIR-07-130-NPS,.Rev. 0 3-13 Structural Integrity Associates, Inc. I NEC066055

Figure 3-1: Feedwater Nozzle Internal Pressure Distribution SIR-07-130-NPS, Rev. 0 3-14 Structural Integrity Associates, Inc. NEC066056

EL EME NT S E LSEP 13 2002 F 12:17:30 FeedwateL-Nozzle Finite Element Model Figure 3-2: Feedwater Nozzle Pressure Cap Load I I I I SIR-07-130-NPS, Rev. 0 3-15 C Structural Integrity Associates, Inc. I NEC066057

ELEMENTS mA. SEP 13 2002 12:20:02 Feedwater Nozzle Finite Element Model Figure 3-3: Feedwater Nozzle Vessel Boundary Condition SIR-07-130-NPS, Rev. 0 3-16 V Structural Integrity Associates, Inc. NEC066058

Region 7 Region 8 F Region 6 Region I Region 5 I I I I I I I I I I I I I I I I I AB 8 0 Notes: Point A: Point B: Point C: Point D: Point E: Point F: End of thermal sleeve = Node 204 = 0.25" from feedwater inlet side of thermal sleeve flat per Reference [8]. Beginning of annulus = Node 252. Beginning of thermal sleeve transition = approximately 4.0" from Point A per Reference [8] = Node 294. End of thermal sleeve transition = approximately 9.5" from Point A per Reference [8] = Node 387. End of inner blend radius (nozzle side) = Node 553. End of inner blend radius (vessel wall side) = Node 779. Figure 3-4: Thernial Regions S1R-07-1.30-NPS, Rev. 0 3-17 V Structural Integrity Associates, Inc. I I NEC066059

~i7~ w1/2g 3 odq" 187 Figure 3-5: Safe End Critical Thermal Stress Location and Linearized Stress Paths SIR-07-130-NPS, Rev. 0 ' 3-18 Structural Integrity Associates, Inc. .NEC066060

I I I I I I I I I I I I Figure 3-6: Brand Radius Critical Thermal Stress Location and Linearized Stress Paths I! SIR-O7-130oNPS, Rev. 0 3-19 Structural I-We9fity Associts, I! I NEC066 0 6 1

Total Stress Intensity 70000 F-s-s G 550000 40000 30000 20000 10000 -\\I-- 0- -4. I. -10000 4-0 4-4- 4-- 100 200 300 400 500 Time (see) Figure 3-7: Safe End Total Stress History for 100% Flow Total Stress Intensity 50000 40000 30000 -10000 0 100 200 300 400 S00 Tire (sec) Figure 3-8: Safe End Membrane Plus Bending Stress History for 100% Flow SIR-07-130-NPS, Rev. 0 3-20 V Structural Integrity Associates, Inc. NEC066062

Total Stress Intensity 200O00 100 200 300 400 500 Time (sec) Figure 3-9: Safe End Total Stress History for 40% Flow Total Stress Intensity 40000 -1 100 200 300 400 500 Time (sec) Figure 3-10: Safe End Membrane Plus Bending Stress History for 40% Flow SIR-07-130-NPS, Rev. 0 3-21 U Structural Integrity Associates, Inc. NEC066063

Total Stress Intensity 50000 40000 30000 20000 10000 0 -10000 100 200 300 400 500 Time (sec) Figure 3-11: Safe End Total Stress History for 25% Flow Total Stress Intensity 40000 30000 20000 F-::27sx g 10000 0 t00 200 300 400 500 Time (sec) Figure 3-12: Safe End Membrane Plus Bending Stress History for 25% Flow SIR-07-130-NPS, Rev. 0 3-22 V Structural Integrity Associates, Inc. NEC066064

Total Stress Intensity 30000 1000 2000 3000 4000 5000 Time (sec) Figure 3-13: Blend Radius Total Stress History for 100% Flow Total Stress Intensity 30000 W* 5000 Time (sec) Figure 3-14: Blend Radius Membrane Plus Bending Stress History for 100% Flow SIR-07-130-NPS, Rev. 0 3-23 V Structural Integrity Associates, Inc. NEC066065

Total Stress Intensity 30000 25000 r 1 I + F F -sz-sx 205000 4F i F 15000 0 1000 2000 3000 4000 5000 Time (sec) Figure 3-15: Blend Radius Total-Stress History for 40% Flow Total Stress Intensity 30000 1000 2000 3000 4000 5000 Time -ser) Figure 3-16: Blend Radius Membrane Plus Bending Stress History for 40% Flow SIR-07-130-NPS, Rev. 0 3-24 V Structural Integrity Associates, Inc. NEC066066

Total Stress Intensity 30000 4

4

.4 .4 L*UU*" 25,300 1 -1 4 4 4 I -*sz-sx; 15000 10000 5000-i_ 4 -i + 4 -t 0 !- i- -4 0 1000 2000 3000 4000 Time (sec) Figure 3-17: Blend Radius Total Stress History for 25% Flow Total Stress intensity 5000 30000 25000 20000 15000 10000 5000 0 1000 2000 3000 4000 5000 Time (sec) Figure 3-18: Blend Radius Membrane Plus Bending Stress History for 25% Flow SIR-07-130-NPS, Rev. 0 3-25 V Structural Integrity Associates, Inc. NEC066067

-I Temp (F) - Pressure (psig(I 80 -1I 70 1 60 50 ._ 40* I-- 30 20 10-4------0 _s D {3-Stress.exe program calculates steady state values at beginning of transients. The time length for this transient can therefore be any value greater than zero. The chosen length of 10 seconds has no significance as there is no temperature change during this transient. -1 0 1 2 3 4 5 Time (seconds) 6 7 8 9 10 Figure 3-19: Transient 1, Bolt-up [- Temp (F) - - Pressure (psig)] 120 100 I\\ E 80 60-Stres..ene programn automatically calculates steady state c..editi.ns at beginning of transients. I I I I 1 I I 1200 1150 1100 1050 1000 950 900 850 800 750 700 650 600 550 500 450 400 350 300 250 20) 150 100 50 0 -50 d: 40 25 0 I 0 1000 2000 3000 4000 5000 Time (seconds) Figure 3-20: Transient 2, Design HYD Test SIR-07-130-NPS, Rev. 0 3-26 Structural Integrity Associates, Inc. NEC066068

-*Temp (°F) -- -- Pressure (psig) 300 E (v I I I I I I I I I I 1 I I 5000 10000 15000 20000 Time (seconds) Figure 3-21: Transient 3, Startup t ~ emp t"F) -- --Pressure psigi] 6000 500 300 200 100 Otress.axe pmrelrm auoxaialycluae steady State candatiass at begnnitng of transients This transient Icgirts at 549"F ard steps ae prtr of end" ss heso r down to 00'F in o.e Secned t pru i aces see od sthat steady atn m isat ischd rechd Tha wanyrathies re"e Kefore She d nextoewhc ilit~r l i1080 10,l0 1000 960 -920 880 040 - 000 -760 720 680 640 2600

560 520 480 440 400 360 320 280 240 200 160 120 BO 40 0

o* 1000 2000 3000 4000 5 Time (seconds) 000 6000 7000 8000 Figure 3-22: Transient 4, Turbine Roll and Increased to Rated Power I I SIR-07-130-NPS, Rev. 0 3-27 V Structural Integrity Associates, Inc. I NEC066069

-I Temp (Ff - - Pressure (psig) 800 - 700 - 600 5000-6 400 E 300 200 - 100 - 1200 1160 1120 1080 1040 1000 960 920 880 840 800

760
720 680 640 toress.exe program calculates steady state values at beginning of transients. The time length for this transient can therefore be any value greater than zero. The chosen length ef 10 seconds has no sinificance as mhere is no temperature change during this transient
560 520 480
440
400
300 320 280 240
200 160 120 80

.40 mg o_ 010-0 1000 0 2000 3000 4000 Time (seconds) 000 6000 7000 8000 Figure 3-23: Transient 5, Daily Reduction 75% Power I--Temp ('F) -: -- Pressure (psq)I 600 - 500 - 400 - 300 200 - 100 - 1080 .1040 1000 9 860 -920 -880 -840 -800 -760 -720 -680 -640 600 -560 -520 -480 .440 -400 -360 --320 .280

240 200

-160 120

80

-40 ._mE o8 0

41.

-0 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 -6500 Time (seconds) Figure 3-24: Transient 6, Weekly Reduction 50% Power SIR-07-130-NPS, Rev. 0 3-28 V Structural Integrity Associates, Inc. NEC066070

I-Temp (*F) -- Pressure (psig) 6 H 1080 1040 1000 960 920 880 840 800 760 720 680 640 600 560 520 U 480 440 o-400 360 320 280 240 200 160 120 80 40 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 Time (seconds) Figure 3-25: Transient 9, Turbine Trip at 25% Power I-Temp ('F) - -Pressure (psig)] S00 450 400 350 300 250 E F-200 150 100 s0 0 1080 1040 1000 960 - 920 880 q-84 L-i -800 760 -720 660 -640 600 -560 -520 -480 -440 400 -360 -320 -280

240
200 160
120

- 80 40 0 1000 2000 3000 4000 5000 6000 Time (seconds) Figure 3-26: Transient 10, Feedwater Bypass

7000, SIR-07-130-NPS, Rev. 0 3-29 V

Structural Integrity Associates, Inc. NEC066071

i-Temp (°F) - - Pressure (psig) 600 550 500 450 400 350 1 300 F 250 200 150 100 50

1200 1000 800 600 400 200 l0 5000 10000 15000 20000 25000 Time (seconds)

Figure 3-27: Transient 11, Loss of Feedwater Pumps Temp ('F) - -Pressure (psig) i-- 100 1020 980 940 900 860 820 780 740 700 9660 6 20 580 540 500 4970 -30 970 1970 2970 3970 Time (seconds) Figure 3-28: Transient 12, Turbine Generator Trip SIR-07-130-NPS, Rev. 0 3-30 V Structural Integrity Associates, Inc. NEC066072

I-Temp (fF) -- Pressure (psig) 450 400 350 E 200 150 100 - 50-o_ 200 150 100 50 0 1000 2000 3000 4000 5000 Ti me (seconds) Figure 3-29: Transient 14, SRV Blowdown I I I I I I I I I I I I I I -*Temp (°F) Pressurre (psig)] 450 400 300 300 250 1080 1040 1000 960 920 880 840 800 760 720 680 640 600 a 5 6 0 520 480 440 0 400 360 320 280 240 200 160 120 80 - 40 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 Time (seconds) Figure 3-30: Transient 19, Reduction to 0% Power SIR-07-130-NPS, Rev. 0 3-31 I C Structural Integrity Associates, Inc. I NEC066073

I I I I U I I I I I I I I I I I (-Temp (F) -- Pressure (psig) 550 500 450 7i 400 E 1100 -1000 500 saThis 7transtent conrtinues at steady tle to18025 7seconds. a 800 700 600 500 ?00 800 900 1000 350 300 100 200 300 400 500 600 Time (seconds) 600 500 Figure 3-31: Transient 20, Hot Standby (Heatup Portion) Temp °F) - Pressure (psig)j This transient confinuns at steady state to 5451,seconds.. 1100 i000 400 300 E 200 Soo 700 600 100 04 0 100 200 300 400 500 600 700 800 Time (seconds) o00 1000 Figure 3-32: Transient 20A, [lot Standby (Feedwater Injection Portion) I I I SIR-07-130-NPS, Rev. 0 3-32 V Structural Integrity Associates, Inc. NEC066074

600-as a straight line for 1150 simplicity. The pressure 1150 actually foltows saturation. 1050 1000 500 950 900 .850 - 600 800 400 750 700 2-00 35 60o .300 050 2!

a.

500 450 400 200 -350 300 250 200 0 0 -50 0 2000 4000 6000 8000 t0000 12000 14000 16000 18000 20000 Time (seconds) Figure 3-33: Transient 21-23, Shutdown To ('F) Pressure (psig) 150 1600 / 100 1500 /\\ 130 / 1400 / k 13oo / 110

1200 1100 90

/ 1000 / 000 -p* in 70 */ 00 0. E 700 50 / 600 / 500 30 400 / / 300 t0 - f200 / 100 -10 0 0 100 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Time (seconds) Figure 3-34: Transient 24, Hydrostatic Test SIR-07-130-NPS, Rev. 0 3-33 Structural Integrity Associates, Inc. NEC066075

-ITemp (F) - - Pressure (psig) 150 130 110 E 90 70 So 30 10 -10 500 400 300 200 a 100 0 6000 0 1000 2000 3000 4000 Time (seconds) Figure 3-35: Transient 25, Unbolt F, RNf 5000 Figure 3-36: External Forces and Moments on the Feedwater Nozzle SIR-07-130-NPS, Rev. 0 3-34 C Structural Integrity Associates, Inc. NEC066076

4.0 STRESS AND FATIGUE ANALYSIS RESULTS Fatigue calculations for the VY FW nozzle were performed in accordance with ASME Code, Section III, Subsection NB-3200 methodology (1998 Edition, 2000 Addenda) [15]. Fatigue analysis was performed in the Reference [23] calculation for the two locations identified in Section 3.1.2 using the Green's Functions developed for these two locations and the 60-year projected cycle counts from Reference [19]. Tables 4-1 and 4-2 show the stresses for each location that were used in the fatigue analysis. Columns 2 tlhrough 5 of Table 4-1 (for the blend radius) and Table 4-2 (for the safe end) show the final thermal peak and valley output. The pressure values for Column 6 in each table were determined from the transient pressures specified in Tables 3-4 and 3-5. The pressure stress I intensities from Section 3.2 were scaled appropriately for each transient case. The scaled piping stress values are shown in Columns 9 and 10 of Tables 4-1 and 4-2. The piping stress intensities I from Section 3.3 were scaled based on the transient case RPV fluid temperature and assuming no stress occurs at an ambient temperature of 70'F. Both of these stress intensities were then added to the thermal stress intensity peak and valley points to calculate the final stress values used for the fatigue analysis. In the case of the piping load stress intensities, the sign of the stress intensity was conservatively set to the same sign as the thermal stress intensity to ensure bounding fatigue usage results. Columns 11 and 12 of Tables 4-1 and 4-2 show the summation of all stresses for each thermal peak and valley stress point. The last column shows the number of cycles associated with each peak or valley based on the cycle counts shown in Tables 3-4 and 3-5. I The program FATIGUE.EXE performs the ASME Code peak event-pairing required to calculate a fatigue usage value. The input data for the configuration input file for FATIGUE.EXE, which is named FATIGUE.CFG, is shown in Table 4-3. I I SIR-07-130-NPS, Rev. 0 4-1 Structural Integrity Associates, Inc. NEC066077

The results of the fatigue analysis are presented in Tables 4-4 and 4-5 for the safe end and blend radius for 60 years, respectively. The blend radius cumulative usage factor (CUF) from system cycling is 0.0636 for 60 years. The safe end CUF is 0.1471 for 60 years. SIR-07-130-NPS, Rev. 0 4-2 C Structural Integrity Associates, Inc. NEC066078

I I Table 4-1: Feedwater Nozzle Blend Radius Stress Summary 1 2 3 4 5 6 7 8 9 10 11 12 13 Total M+B Total M+B Total Total Number Total M+B Pressure Pressure Piping Piping Total M+1B of Transient Time Stress Stress Temperature Pressure Stress Stress Stress Stress Stress Stress Cycles Number is (s (psi) F ipsilg) (i ps p (si) (psi) (60 years) 1 0 0 0 70 0 0 0 0 0 0.00 0.00 123 0 0 0 70 0 0 0 0 0 0.00 0.00 120 2 1680 0 0 100 1100 41506.3 40318.3 15.77042 15.77042 .41522.07 40334.07 120 10880 0 0 100 50 1886.65 1832.65 15.77042 15.77042 1902.42 1848.42 120 0 29166 23676 100 50 1886.65 1832.65 15.77042 15.77042 31068.42 25524.42 300 3 16782.8 -3577 -3138 549 1010 38110.33 37019.53 -251.801 -251.801 34281.53 33629.73 300 21164 -3532 -3138 549 1010 38110.33 37019.53 -251.801 -251.801 34326.53 33629.73 300 0 -3530 -3158 549 1010 38110.33 37019.53 -251.801 -251.801 34328.53 33609.73 300 4 1801.9 29465 22266 244.004 1010 38110.33 37019.53 91.47053 91.47053 67666.80 59377.00 300 8602 7720 6749 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43937.80 300 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 10000 5 2229.8 13598 11941 311.002 1010 38110.33 37019.53 126.6901 126.6901 51835.02 49087.22 10000 8600 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 10000 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 2000 6 2820.3 15742 13892 280.691 1010 38110.33 37019.53 110.7562 110.7562 53963.09 51022.29 2000 10400 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 2000 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 10 9 2524 29006 23417 118.311 1010 38110.33 37019.53 25.39616 25.39616 67141.73 60461.93 10 10400 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 10 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 70 10 1632.4 16828 14701 267.399 1010 38110.33 37019.53 103.7688 103.7688 55042.10 51824.30 70 7070 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 701 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 10 3.5 6620 6632 565 1190 44902.27 43617.07 260.2119 260.2119 51782.48 50509.28 10 4.5 6190 6608 50 1185 44713.61 43433.81 10.51361 10.51361 50914.12 50052.32 10 194.5 31720 21067 109.348 1135 42826.96 41601.16 20.68448 20.68448 74567.64 62688.84 10 2166.3 -4761 -1859 513.483 972 36676.48 35626.72 -233.1304 -233.1304 31682.35 33534.59 10 11 2362.5 31268 22070 102.255 1010 38110.33 37019.53 16.95583 16.95583 69395.29 59106.49 10 6728.3 -4913 -3149 513.448 1010 38110.33 37019.53 -233.112 -233.112 32964.22 33637.42 10 7149.9 32114 21472 83.333 1010 38110.33 37019.53 7.0089 7.0089 70231.34 58498.54 10 18213.3 -3565 -3162 503.978 1010 38110.33 37019.53 -228.1338 -228.1338 34317.20 33629.40 10 19122.6 29156 23083 100.048 1010 38110.33 37019.53 15.79565 15.79565 67282.13 60118.33 10 26814.5 7720 6410 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43598.80 10 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 60 10 7720 6752 392 1135 42826.96 41601.16 169.2692 169.2692 50716.22 48522.42 60 12 30 7720 6752 392 940 35469.02 34453.82 169.2692 169.2692 43358.29 41375.09 60 2033.7 28648 25301 132.007 940 35469.02 34453.82 32.59588 32.59588 64149.62 59787.42 60 9591 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 60 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 1 10 7720 6752 392 1375 51882.88 50397.88 169.2692 169.2692 59772.14 57319.14 1 13 30 7720 6752 392 940 35469.02 34453.82 169.2692 169.2692 43358.29 41375.09 1 2033.7 28648 25301 132.007 1010 38110.33 37019.53 32.59588 32.59588 66790.93 62353.13 1 9591 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 1 14 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 1 5960 28487 25650 100 50 1886.65 1832.65 15.77042 15.77042 30389.42 27498.42 1 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 228 10 7720 6752 392 1135 42826.96 41601.16 169.2692 169.2692 50716.22 48522.42 228 15 30 7720 6752 392 940 35469.02 34453.82 169.2692 169.2692 43358.29 41375.09 228 2033.7 28648 25301 132.007 1010 38110.33 37019.53. 32.59588 32,59588 66790.93 62353.13 028 9591 7720 6752 392 1010 38110.33 37019.53 169.2692 169.2692 45999.60 43940.80 228 19 0 7720 6752 392 1010 38110.33 37019.53 169.2692 169.26921 45999.60 43940-80 300 6800 16752 14971 265 1010 38110.33 37019.53 102.5077 102.5077 54963.84 52093.04 300 20 1 17151 138153 265 1010 38110.33 37019.53 102.5077 102.5077 55363.84 50937.04 300 8925 -3531 -3146 549 1010 38110.33 37019.53 -251.801 -251.801 34327.53 33621.73 300 0 -3530 -3158 549 1010 38110.33 37019.53 -251.801 -251.801 34328.53 33609.73k 300, 20A 183 28102 12153 233 1010 38110.33, 37019.53 85.68595 85.68595 66298.02 49258.22 300 5451 -3530 -3158 549 1010 38110.33 37019.53 -251.801 -251.801 34328.53 33609.73 300 21-23 0 -3530 -3158 5491 1010. 38110,33 37019.53. -251.801 -251M81 34328.53 33609.73 300 20144 29168 23656 100 50 1886.65 1832.65 15.77042 15,77042 31070.42 25504.42 300 0 0 0 100 50 1886.65 1832.65 15.77042 15.77042 1902.42 1848.42 1 24 600 0 0 100 1563 58976.68 57288.64 15.77042 15.77042 58992.45 57304.41 1 1 2400 0 0 100 50 1886.65 1832.65 15.77042 15.77042 1902.42 1848.42 -1 25 0 0 0 100 0 0 0 15.77042 15.77042 15.77 15.77 123 1 1580 0 0 70 0 0 0 0 0 0.00 0.00 123 I I I I I I I I I I I I I I I For notes, see last page of table... SIR-07-130-NPS, Rev. 0 4-3 Structural Integrity Associates, Inc. I NEC066079

Table 4-1: Feedwater Nozzle Blend Radius Stress Summary (continued) NOTES: Column 1: Transient number identification. Colunmn 2: Time during transient where a maxima or minima stress intensity occurs fromn P-V.OUT output file. Column 3: Maxima or minima total stress intensity from P-V.OUT output file. Column 4: Maxima or minima membrane plus bending stress intensity from P-V.OUT output file. Column 5: Temperature per total stress intensity. Column 6: Pressure per Table 3-4. Column 7: Total pressure stress intensity from the quantity (Column 6 x 37733)/1000 [Table3, 101. Column 8: Membrane plus bending pressure stress intensity from the quantity (Column 6 x 36653)/1000 [Table 3, 10]. Column 9: Total external stress orom calculation in Table 3-2, 265.47 psi*(Colmumn 5-70'F)/(5750 F -70"F). Column 10: Same as Column 91 but for M+B stress. Column 11: Sum of total stresses (Columns 3, 7, and 9). Column 12: Sum of membrane plus bending stresses (Columns 4, 8, and 10). Column 13: Number of cycles for the transient (60 years). SIR-07-130-NPS, Rev. 0 4-4 V Structural Integrity Associates, Inc. NEC066080

Table 4-2: Feedwater Nozzle Safe End Stress Summary 1 2 3 4 5 6 7 8 9 10 11 12 13 Total M+B Total M+B Total Total Number Total M+8 Pressure Pressure Piping Piping Total M+B of Transient Time Stress Stress Temperature Pressure Stress Stress Stress Stress Stress Stress Cycles Number (s) (psi) (psi) F psiq) (psi) (psi) (psi) (p (pai) psil (60 years) _1 0 0 0 70 0 0 0 0 0 0.00 0.00 123 0 0 0 70 0 0 0 0 0 0.00 0.00 120 2 1680 0 0 100 1100 9780.1 9562.3 339.0875 339.0875 10119.19 9901.39 120 6960 0 0 100 50 444.55 434.65 339.0875 339.0875 783.64 773.74 120 0 -170 -165 100 50 444.55 434.65 -339.0875 -339.0875 -64.54 -69.44 300 153.2 -235 -212 104.256 50 444.55 434.65 -387.1927 -387.1927 -177.64 -164.54 300 16328.2 2 3 549 1010 8979.91 8779.93 5414.097 5414.097 14396.01 14197.03 300 16664 -1 0 549 1010 8979.91 8779.93 -5414.097 5414.097 3564.81 14194.03 300 0 -3 -2 549 1010 8979.91 8779.93 -5414.097 -5414,097 3562.81 3363.83 300 3.6 44060 30988 100 1010 8979.91 8779.93 339.0875 339.0875 53379.00 40107.02 300 1804.6 -15889 -11224 260.286 1010 8979.91 8779.93 -2150.787 -2150.787 -9059.88 -4594.86 300 4101 214 23 392 1010 8979.91 8779.93 3639.539 3639.539 12640.45 12442.47 300 0 22 23 392 1010 8979.91 8779.93 3639.539 3639.539 12641.45 12442.47 10000 900,1 244 189 310 1010 8979.91 8779.93 2712.7 2712.7 11936.61 11681.63 10000 5 3600 -169 -110 392 1010 8979.91 8779.93 -3639.539 -3639.539 5171.37 5030.39 10000 3684.4 33 35 392 1010 8979.91 8779.93 3639.539 3639.539 12652.45 12454.47 10000 4100 22 23 392 1010 8979.91 8779.93 3639.539 3639.539 12641.45 12442.47 10000 0 22 23 392 1010 8979.91 8779.93 3639.539 3639.539 12641.45 12442.47 2000 1800.1 196 159 280 1010 8979.91 8779.93 2373.612 2373.612 11549.52 11312.54 .2000 6 5400.2 -108 -68 392 1010 8979.91 8779.93 -3639.539 -3639.539 5232.37 5072.39 2000 5496.6 29 31 392 1010 8979.91 8779.93 3639.539 3639.539 12648.45 12450.47 2000 5900 22 23 392 1010 8979.91 8779.93 3639.539 3639.539 12641.45 12442.47 2000 0 22 23 392 1010 8979.91 8779.93 3639.539 3639.539 12641.45 12442.47 10 97.3 180 137 385.135 1010 8979.91 8779.93 3561.945 3561.945 12721.85 12478.87 1t 1884.1 63 65 265 1010 8979.91 8779.93 2204.069 2204.069 11246.98 11049.00 1t 2059.2 1161 859 226.597 1010 8979.91 8779.93 1770.003 1770,003 11910.91 11408.93 1t 9 3420.1 -334 -211 265 1010 8979.91 8779.93 -2204.069 -2204.069 6441.84 6364.86 t1 3490.2 97 98 265 1010 8979.91 8779.93 2204.069 2204.069 11280.98 11082.00 1t 5400.1 -126 -80 392 1010 8979.91 8779.93 -3639.539 -3639.539 5214.37 5060.39 1t 5470.6 31 32 392 1010 8979.91 8779.93 3639.539 3639.539 12650.45 12451.47 t1 5900 22 23 392 1010 8979.91 8779.93 3639.539 3639.539 12641.45 12442.47 10 0 23 22 392 1010 8979.91 8779.93 3639.539 3639.539 12642.45 12441.47 70 77.1 2308 3188 285.461 1010 8979.91 8779.93 2435.338 2435.338 13723.25 14403.27 70 169.4 -12 -13 265 1010 8979.91 8779.93 -2204.069 -2204.069 6763.84 6562.86 70 10 1890 74 72 265 1010 897991 8779.93 2204.069 2204.069 11257.98 11056.00 70 1968.2 -1069 -1511 322.362 1010 8979.91 8779.93 -2852.427 -2852.427 5058.48 4416.50 70 2147.2 91 90 392 1010 8979.91 8779.93 3639.539 3639.539 12710.45 12509.47 70 2570 23 22 392 1010 8979.91 8779.93 3639.539 3639.539 12642.45 12441.47 70 0 -29 -27 392 1010 8979.91 8779.93 -3639.539 -3639.539 5311.37 5113.39 10 2.9 -20317 -13859 565 1147 10197.98 9970.871 -5594.944 -5594.944 -15713.97 -9483.07 10 6.8 42852 29563 565 1172 10420.25 10188.2 5594.944 5594.944 58867.20 45346.14 1t 1567.4 -15216 -10526 565 1135 10091.29 9866.555 -5594.944 -5594.944 -10719.66 -6254.39 1t 2168.4 60377 41773 50 1134 10082.39 9857.862 -226.0583 -226.0583 70233.34 51404.80 10 11 5409.4 -14924 -10329 565 1054 9371.114 9162.422 -5594.944 -5594.944 -11147.83 -6761.52 10 6730.4 60377 41773 50 1133 10073.5 9849.169 -226.0583 -226.0583 70224.44 51396.11 10 724312 -1965 -1434 128.917 675 6001.425 5867,775 -665.9339 -665.9339 3370.49 3767.84 10 18215.4 52636 36417 100 1010 8979.91 8779.93 339.0875 339.0875 61955.00 45536.02 10 20015.5 -24511 -16189 260.183 1010 8979.91 8779.93 -2149.623 -2149.623 -17680.71 -9558.69 1t 22314.5 22 23 392 937 8330.867 8145.341 3639.539 3639.539 11992.41 11807.88 1t 0 23 22 392 1010 8979.91 8779.93 3639.539 3639.539 12642.45 12441.47 60 10 23 22 392 1135 10091.29 9866.555 3639.539 3639.539 13753.82 13528.09 60 30 23 22 392 940 8357.54 8171.42 3639.539 3639.539 12020.08 11832.96 60 90 3174 4383 275 940 8357.54 8171.42 2317.098 2317.098 13848.64 14871.52 60 2793.5 -16189 -24511 260.183 941 8366.431 8180.1131 -2149.623 -2149.623 -9972.19 -18480.51 60 5091 23 22 392 1010 8979.91 8779.93 3639.539 3639.539 12642.45 12441.47 60 0 23 22 392 1010 8979.91 8779.93 3639.539 3639.539 12642.45 12441.47 1 10 23 22 392 1375 12225.13 11952.88 3639.539 3639.539 15887.66 15614.41 1 30 23 22 392 940 8357.54 8171.42 3639.539 3639.539 12020.08 11832.96, 1 90 3174 4383 275 940 8357.54 8171.42 2317.098 2317.098 13848.64 14871.52 2793.5 -16189 -24511 260.183 941 8366.431 8180.113 -2149.623 -2149.623 -9972.19 -18480.51 5091 23 22 392 1010 8979.91 8779.93 3639.539 3639.539 12642.45 12441.47 I I I I I I I I I I I I I I I I I For notes, see last page of table... SIR-07-130-NPS, Rev. 0 4-5 Structural Integrity Associates, Inc. 3 I NEC066081

Table 4-2: Feedwater Nozzle Safe End Stress Summary (continued) 1 2 3 4 5 6 7 8 9 10 11 12 13 Total M+B Total M+B Total Total Number Total M+B Pressure Pressure Piping Piping Total M+B of Transient Time Stress Stress Temperature Pressure Stress Stress Stress Stress Stress Stress Cycles Number (s) (psil (psi) F ipsig) (psi) fpsi) (psi) (psi) fpsi) (psi) (60 years 0 22 23 392 1010 8979.91 8779.93 3639.539 3639-539 12641.45 12442.47 1 60 4383 3174 275 885 7868.535 7693.305 2317.098 2317.098 14568.63 13184.40( 1 14 148 420 300 258.492 803 7139.473 6980.479 2130.509 2130.509 9689.98 9410.99! 1 960 544 424 100 50 444.55 434.65 339.0875 339.0875 1327.64 1197.74 1 1460 137 139 100 50 444.55 434.65 339.0875 339.0875 920.64 912.74 0 23 22 392 1010 8979.91 8779.93 3639.539 3639.539 12642.45 12441.47 228 10 23 22 392 1135 10091.29 9866.555 3639.539 3639.539 13753.82 13528.09 228 30 23 22 392 940 8357.54 8171.42 3639.539 3639.539 12020.08 11832.96 228 90 3174 4383 275 940 8357.54 8171.42 2317.098 2317.098 13848.64 14871.52 228 2793.5 -16189 -24511 260.183 941 8366.431 8180.113 -2149.623 -2149.623 -9972.19 -18480.51 228 5091 23 22 392 1010 8979.91 8779.93 3639.539 3639.539 12642.45 12441.47 228 0 22 23 392 1010 8979.91 8779.93 3639.539 3639.539 12641.45 12442.47 300 19 1800 219 177 265 1010 8979.91 8779.93 2204.069 2204M069 11402.98 11161.00 300 2300 72 74 265 1010 8979.91 8779.93 2204.069 2204.069 11255.98 11058.00 300 0 -109 -105 265 1010 8979.91 8779.93 -2204.069 -2204.069 6666.84 6470.86 300 20 4 -17288 -12189 440.106 1010 8979.91 8779.93 -4183.277 -4183.277 -12491.37 -7592.35 300 4425 -2 -1 549 1010 8979.91 8779.93 -5414.097 -5414.097 3563.81 3364.83 300 0 -3 -2 549 1010 8979.91 8779.93 -5414.097 -5414.097 3562.81 3363.83 300 4 44060 30988 100 1010 8979.91 8779.93 339.0875 339.0875 53379.00 40107.02 3001 20A 241 -7461 -5525 290.247 1010 8979.91 8779.93 -2489.433 -2489.433 -970.52 765.50 300 572 128 132 549 1010 8979.91 8779.93 5414.097 5414.097 14522.01 14326.03 300 951 -3 -2 549 1010 8979.91 8779.93 -5414.097 -5414.097 3562.81 3363.83 300 0 -3 -2 549 1010 8979.91 8779.93 -5414.097 -5414.097 3562.81 3363.83 300 138 62 45 545.167 989 8793.199 8597.377 5370.773 5370.773 14225.97 14013.15 300 21-23 6264 -5 -20 374.97 50 444.55 434.65 -3447.05 -3447.05 -3007.50 -3032.40 300 63901 104 59 366.172 50 444.55 434.65 3347.607 3347.607 3896.16 3841.26 300 15644 -173 -167 100 50 444.55 434.65 -339.0875 -339.0875 -67.54 -71.44 300 0 0 0 100 50 444.55 434.65 339.0875 339.0875 783.64 773.74 1 24 600 0 0 100 1563 13896.63 13587.16 339.0875 339.0875 14235.72 13926.25 1 2400 0 0 100 50 444.55 434.65 339.0875 339.0875 783.64 773.74 1 0 0 0 100 '0 0 0 339.0875 339.0875 339.09 339.09 123 25 1580_ 0 0 70 0 0 0 0 0 0.00 0.00 123 NOTES: Column 1: Transient number identification. Column 2: Time during transient where a maxima or minima stress intensity occurs from P-V.OUT output file. Column 3: Maxima or minima total stress intensity from P-V.OUT output file. Column 4: Maxima or minima membrane plus bending stress intensity from P-V.OUT output file. Column 5: Temperature per total stress intensity. Column 6: Pressure per Table 3-5. Column 7: Total pressure stress intensity from the quantity (Column 6 x 8891)/1000 [Table 3, 101. Column 8: Membrane plus bending pressure stress intensity from the quantity (Column 6 x 8693)/1000 [Table3, 10]. Column 9: Total external stress from calculation in Table 3-2, 5707.97 psi*(Column 5-70°F)/(5750F -70°F). Column 10: Same as Column 9, but for M+B stress. Column 11: Sumn of total stresses (Columns 3, 7, and 9). Column 12: Sum of membrane plus bending stresses (Columns 4, 8, and 10). Colmmn 13: Number of cycles for the transient (60 years). SIR-07-130-NPS, Rev. 0 4-6 U Structural Integrity Associates, Inc. NEC066082

Table 4-3: Fatigue Parameters Used in the Feedwater Nozzle Fatigue Analysis Blend Radius Safe End Parameters in and n for 2.0 & 0.2 (low alloy 3.0 & 0.2 (carbon steel) Computing K, steel) [15] [15] Design Stress Intensity Values, 26700 psi [9] @ 600TF 17800 psi [9] @ 600OF Sm Elastic Modulus from 30.mx106 psi [15] 30.0x106 psi [15] Applicable Fatigue Curve Elastic Modulus Used in Finite 26.7x106 psi [10] 28.1x106 psi [10] Element Model The Geometric Stress CnenGeotrtion Ftoress1.0 1.34 [2, page 35 of S4] Concentration Factor K, I I I I I I I I I I I I I I I I I SIR-07-130-NPS, Rev. 0 4-7 V Structural Integrity Associates, Inc. I I NEC066083

Table 4-4: Fatigue Results for Feedwater Nozzle Blend Radius LOCATION = LOCATION NO. 2 -- BLEND RADIUS FATIGUE CURVE

1 (1 = CARBON/LOW ALLOY, 2 = STAINLESS STEEL) m =2.0 n

.2 Sm = 26700. psi Ecurve = 3.000E+07 psi Eanalysis = 2.670E+07 psi Kt = 1.00 MAX 74568. 70231. 69395. 67667. 67667. 67667. 67282. 67142. 66791. 66791. 66791. 66791. 66298. 66298. 66298. 66298. 66298. 64150. 64150. 59772. 58992. 55364. 55364. 55364. 55364. 55042. 54965. 54965. 54965. 53963. 53963. 53963. 53963. 53963. 53963. 53963. 53963. 53963. 53963. 51835. 51835. MIN 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 16. 1902. 1902. 1902. 1902. 30389. 31068. 31068. 31070. 31070. 31070. 31070. 31682. 32964. 34282. 34282. 34282. 34317. 34327. 34327. 34328. 34329. 34329. 34329. 34329. 41522. 43358. 43358. 43358. 43358. 46000. RANGE 74568. 70231. 69395. 67667. 67667. 67667. 67282. 67142. 66791. 66791. 66775. 64889. 64396. 64396. 64396. 35909. 35230. 33081. 33079. 28702. 27922. 24293. 23681. 22400. 21082. 20761. 20683. 20648. 20638. 19637. 19636. 19635. 19635. 19635. 19635. 12441. 10605. 10605. 10605. 8477. 5835. MEM+BEND 62689. 58499. 59106. 59377. 59377. 59377. 60118. 60462. 62353. 62353. 62337. 60505. 47410. 47410. 47410. 21760. 23734. 34263. 34283. 31815. 31800. 25433. 17402. 17300. 17307. 18195. 18463. 18464. 18463. 17393. 17401. 17413. 17413. 17413. 17413. 10688. 9647. 9647. 9647. 7712. 5149. Ke 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 I.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 I.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 Salt 41892. 39456. 38986. 38015. 3801 5'. 38015. 37799. 37720. 37523. 37523. 3751-4 36454. 36177. 36177. 36177. 20173. 19792. 18585. 18584. 16125. 15687. 13648. 13304. 12584. 11844. 11663. 11620. 11600. 11595. 11032. 11031. 11031. 11031. 11031. 11031. 6989. 5958. 5958. 5958. 4762. 3278. Napplied

1. OOOE+01 1 OOOE+01 1 OOOE+01 9 300E+01

1. 200E+02

8. 700E+01

1. OOOE+01

1. OOOE+00

1. 500E+01 1.230E+02

9. OOOE+01

3. OOOE+01

1. OOOE+00

1. OOOE+00 1. OOOE+00

2. 670E+02

3. 300E+01

2. 700E+01

1. OOOE+00

1. OOOE+00 2.710E+02

1. OOOE+01

9. OOOE+00

7. OOOE+01

2. 210E+02

1. OOOE+01

6. 900E+01

2. 310E+02 3. OOOE+02

3. OOOE+02

3. OOOE+02 1.200E+02

6. OOOE+01

1. OOOE+00 8 800E+01

1. 400E+02

3. OOOE+02 Nallowed 7.488E+03

8. 944E+03

9. 268E+03

9. 988E+03

1. 018E+04

1. 025E+04 1.044E+04 1.044E+04 1.015E+04 1.152E+04 1.182E+04 1.182E+04 1.182E+04

9. 581E+04

1. 038E+05

1. 303E+05

1. 303E+05 2.222E+05

2. 519E+05

4. 757E+05 5 703E+05

9. 414E+05

1. 912E+06 2.231E+06 2.310E+06 2.348E+06

2. 358E+06

3. 757E+06

3. 758E+06

3. 760E+06

3. 760E+06 3.760E+06

1. OOOE+20 1. OOOE+20

1. OOOE+20

1. OOOE+20 U

.0013 .0011 .0011 .0093 .0120 .0087 .0010 .0010 .0001 .0014 .0118 .0078 .0025 .001 0001 .0000 .0026 .0003 .0002 .0000 .0000 .0006 .0000 .0000 .0000 .0000 .0001 .0000 .0000 .0001 .0001 .0001 .0001 .0001 .0001 .0000 .0000 .0000 .0000 .0000 .0000 SIR-07-130-NPS, Rev. 0 4-8 V Structural Integrity Associates, Inc. NEC066084

51835. 51782. 50914. 50716. 50716. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 46000. 5835. 5783. 4915. 4717. 4717. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 5146. 6568. 6112. 4582. 4582. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 I.000 1.000 1.000 1.000 1.000 I.000 1.000 3278. 3249. 2761. 2650. 2650. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0.

9. 560E+03

1. 000E+01

1. 000E+01 6.000E+01

2. 280E+02

1. 320E+02 1.000E+04 2.000E+03 2.000E+03

1. OOOE+01 1.000E+01

7. 000E+01

7. OOOE+01

1. OOOE+01 1.000E+01 6.000E+01

6. OOOE+01 1.000E+00 1.000E+00

1. OOOE+00 2.280E+02

2. 280E+02 1.000E+20 1.000E+20 1.000E+20 1. OOOE+/-20 1.000E+20 1.000E+20

1. OOOE+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20

1. OOOE+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20 1.000E+20

.0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 . 0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0636 TOTAL USAGE FACTOR I I I I I I I I I I I I I I I I I SIR-07-130-NPS, Rev. 0 4-9 Structural Integrity Associates, Inc. I NEC066085

Table 4-5: Fatigue Results for the Feedwater Nozzle Safe End LOCATION = LOCATION NO. 1 SAFE END FATIGUE CURVE = 1 (1 = CARBON/LOW ALLOY, 2 = STAINLESS STEEL) m 3.0 n= .2 Sm = 17800. psi Ecurve = 3.000E+07 psi Eanalysis = 2.810E+07 psi .Kt = 1.34 MAX MIN RANGE MEM+BEND Ke Salt Napplied Nallowed 70233. 70224. 61955. 58867. 53379. 53379. 53379. 53379. 53379. 53379. 53379. 15888. 14569. 14522. 14522. 14396. 14396. 14236. 14226. 14226. 13849. 13849. 13849. 13849. 13754. 13754. 13723. 13723. 12722. 12710. 12652. 12652. 12652. 12652. 12652. 12652. 12652. 12652. 12652. 12652. 12652. 12652. -17681. -15714. -12491. -12491. -12491. -11148. -10720. -9972. -9972. -9972. -9060. -9060. -9060. -9060. -3008. -3008. -971. -971. -971. -178. -178. -178. -178. -68. -68. -68. -68. -65. -65. -65. -65. 0. 0. 0. 339. 784. 784. 784 921. 1328. 3370. 3563. 87914. 85938. 74446. 71359. 65870. 64527. 64099. 63351. 63351. 63351. 62439. 24948. 23629. 23582. 17530. 17404. 15367. 15206. 15196. 14404. 14026. 14026. 14026. 13916. 13821. 13821. 13791. 13788. 12786. 12775. 12717. 12652. 12652. 12652. 12313. 11869. 11869. 11869. 11732. 11325. 9282. 9090. 60963. 60879. 53128. 52938. 47699. 46869. 46361. 58588. 58588. 58588. 44702. 20209. 17779. 18921. 17358. 17229. 13432. 13161. 13248. 14178. 15036. 15036. 15036. 14943. 13600. 13600. 14475. 14473. 12548. 12579. 12524. 12454. 12454. 12454. 12115. 11681. 11681. 11681. 11542. 11257. 8687. 9091. 1.283 1.280 1.000 1.000 1.000 1.000 1.000 1.194 1.194 1.194 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1 000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 I.000 1.000 74422. 72869. 49383. 47700. 43819. 42951. 42631. 53087. 53087. 53087. 41444. 16985. 15840. 16022. 12508. 12417. 10641. 10506. 10516. 10262. 10216. 10216. 10216. 10141. 9846. 9846. 9989. 9987. 9103. 9102. 9061. 9014. 9014. 9014. 8772. 8456. 8456. 8456. 8357. 8088. 6531. 6502.

1. OOOE+01

1. OOOE+01 2.800E+02

1. OOOE+01

6. OOOE+01

1. OOOE+00

2. 280E+02

1. 100E+01

1. OOOE+00

1.

OOOE+00 2.870E+02

1. 300E+01 2.870E+02

1. 300E+01

1. OOOE+00 2.860E+02

1. 400E+01

6. OOOE+01

1. OOOE+00

2. 250E+02

3. OOOE+00

6. OOOE+01

2. 280E+02

9. OOOE+00

6. 100E+01

1. OOOE+01

7. OOOE+01

1. 590E+02 1. 230E+02

1. 200E+02 1. 230E+02 1. 230E+02

1. 200E+02

1. OOOE+00

1. OOOE+01

3. OOOE+02

1. 338E+03

1. 415E+03

4. 568E+03

5. 094E+03

6. 552E+03 6.953E+03

7. 109E+03 3.628E+03 3.628E+03

3. 628E+03 7.731E+03 1.802E+05 2.410E+05

2. 287E+05

9. 944E+05

1. 083E+06 5.165E+06

5. 563E+06 5.531E+06

6. 379E+06

6. 547E+06

6. 837E+06

8. 117E+06 8.117E+06 7.465E+06

7. 474E+06

1. 729E+07 1.730E+07 1.833E+07

1. 959E+07

2. 905E+07 4 952E+07

4. 952E+07

4. 952E+07 5 462E+07

7. 100E+07

1. OOOE+20

1. OOOE+20 U

.0075 .0071 .0022 .0020 .0427 .0014 .0014 .0165 .0003 .0628 .0014 .0000 .0000 .0013 .0000 .0003 .0000 .0000 .0001 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 0000 0000 0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 .0000 SIR-07-130-NPS, Rev. 0 4-10 C Structural Integrity Associates, Inc. NEC066086

I 12652. 3563. 9090. 9091. 1.000 6502. 3.000E+02 1.000E+20 .0000 12652. 3563. 9090. 9091. 1.000 6502. 3.OOOE+02 1.OOOE+20 .0000 12652. 3563. 9090. 9091. 1.000 6502. 3.000E+02 1.OOOE+20 .0000 12652. 3564. 9089. 9090. 1.000 6501. 3.OOOE+02 1.000E+20 .0000 12652. 3565. 9088. -1740. 1.000 4535. 3.000E+02 1.000E+20 .0000 12652. 3896. 8756. 8613. 1.000 6237. 3.OOOE+02 1.OOOE+20 .0000 12652. 5058. 7594. 8038. 1.000 5513. 7.OOOE+01 1.OOOE+20 .0000 12652. 5171. 7481. 7424. 1.000 5341. 7.048E+03 1.OOOE+20 .0000 12650. 5171. 7479. 7421. 1.000 5339. 1.OOOE+01 1.OOOE+20 .0000 I 12648. 5171. 7477. 7420. 1.000 5338. 2.OOOE+03 1.000E+20 .0000 12642. 5171. 7471. 7411. 1.000 5333. 7.OOOE+01 1.000E+20 .0000 12642. 5171. 7471. 7411. 1.000 5333. 7.OOOE+01 1.000E+20 .0000 12642. 5171. 7471. 7411. 1.000 5333. 6.OOOE+01 1.OOOE+20 .0000 I 12642. 5171. 7471. 7411. 1.000 5333. 6.OOOE+01 1.OOOE+20 0000 12642. 5171. 7471. 7411. 1.000 5333. 1.OOOE+00 1.OOOE+20 .0000 12642. 5171. 7471. 7411. 1.000 5333. 1.OOOE+00 1.OOOE+20 .0000 12642. 5171. 7471. 7411. 1.000 5333. 2.280E+02 1.OOOE+20 .0000 I 12642. 5171. 7471. 7411. 1.000 5333. 2.280E+02 1.000E+20 .0000 12641. 5171. 7470. 7412. 1.000 5333. 2.240E+02 1.OOOE+20 .0000 12641. 5214. 7427. 7382. 1.000 5304. 1.OOOE+01 1.000E+20 .0000 12641. 5232. 7409. 7370. 1.000 5293. 2.OOOE+03 1.000E+20 .0000 12641. 5311. 7330. 7329. 1.000 5243. 1.000E+01 1.000E+20 .0000 12641. 6442. 6200. 6078. 1.000 4412. 1.000E+01 1.000E+20 .0000 12641. 6667. 5975. 5972. 1.000 4273. 3.OOOE+02 1.000E+20 .0000 I 12641. 6764. 5878. 5880. 1.000 4205. 7.OOOE+01 1.000E+20 .0000 12641. 9690. 2951. 3031. 1.000 2126. 1.000E+00 1.OOOE+20 .0000 12641. 10119. 2522. 2541. 1.000 1808. 1.200E+02 1.OOOE+20 .0000 12641. 11247. 1394. 1393. 1.000 997. 1.OOOE+01 1.OOOE+20 .0000 i 12641. 11256. 1385. 1384. 1.000 991. 3.OOOE+02 1.000E+20 .0000 12641. 11258. 1383. 1386. 1.000 990. 7.000E+01 1.000E+20 .0000 12641. 11281. 1360. 1360. 1.000 973. 1.OOOE+01 1.000E+20 .0000 12641. 11403. 1238. 1281. 1.000 894. 3.OOOE+02 1.000E+20 .0000 I 12641. 11550. 1092. 1130. 1.000 788. 2.OOOE+03 1.000E+20 .0000 12641. 11911. 731. 1034. 1.000 578. 1.000E+01 1.000E+20 .0000 12641. 11937. 705. 761. 1.000 514. 4.555E+03 1.000E+20 .0000 12641. 11937. 705. 761. 1.000 514. 5.445E+03 1.000E+20 .0000 I 12641. 11992. 649. 635. 1.000 462. 1.000E+01 1.000E+20 .0000 12641. 12020. 621. 610. 1.000 442. 6.000E+01 1.OOOE+20 .0000 12641. 12020. 621. 610. 1.000 442. 1.000E+00 1.OOOE+20 .0000 12641. 12020. 621. 610. 1.000 442. 2.280E502 1.OOOE+20 .0000 I 12641. 12640.

1.

0. 1.000

1. 3.000E+02 1.000E+20

.0000 12641. 12641.

0.

0. 1.000

0. 3.956E+03 1.000E+20

.0000 12641. 12641.

0.

0. 1.000

0. 2.000E+03 1.000E+20

.0000 12641. 12641.

0.

0. 1.000

0. 2.00E+03 1.000E+20

.0000 12641. 12641.

0.

0. 1.000

0. 1.000E+01 1.000E+20

.0000 12641. 12641.

0.

0. 1.000

0. 1.OOOE+01 1.000E+20

.0000 12641. 12641.

0.

0. 1.000

0. 1.000E+00 1.000E+20

.0000 TOTAL USAGE FACTOR = .1471 I SIR-07-130-NPS, Rev. 0 4-11 Structural Integrity Associates, Inc. NEC066087

5.0 ENVIRONMENTAL FATIGUE ANALYSIS In the response to NRC request for additional information (RAI) 4.3-H-02 [19], VYNPS states that they have conservatively assumed that fatigue cracks may be present in the clad. VYNPS manages this cracking by performing periodic inspections that were implemented in response to Generic Letters 80-095 and 81-11, and NUREG-0619. The inspection frequency is based on the calculated fatigue crack growth of a postulated flaw in the nozzle inner blend radius. The VYNPS fatigue crack growth calculation uses methods in compliance with GE BWR Owners Group Topical Report "Alternate BWR Feedwater Nozzle Inspection Requirements", GE-NE-523-A71-0594, Revision 1, August 1999 and the associated NRC Final Safety Evaluation (TAC No. MA6787) dated March 10, 2000. The NRC has reviewed and approved this approach to handling FW nozzle inner blend radius cracking (Letter D.H. Dorman (USNRC) to D.A. Reid (VYNPC),

Subject:

Evaluation of Request for Relief from NUREG-0619 for VYNPS dated 2/6/95, (TAC No. M88803)). The analysis performed for the feedwater nozzle calculated fatigue in the blend radius base metal, not the clad. This is consistent with the VYNPS position stated in the response to RAI 4.3-H-02, and is also consistent with ASME Code methodology since cladding is structurally neglected in fatigue analyses, per ASME Code, Section I1, NB-3 122.3 [15]. Environmental fatigue multipliers were computed for both normal water chemistry (NWC) and hydrogen water chemistry (HWC) conditions in Reference [21 ] for various regions of the VY RPV and attached piping. Based on VY-specific dates for plant startup and HWC implementation, as well as past and future predicted HWC system availability, it was determined that overall HWC availability is 47% over the sixty year operating period for VY. Therefore, for the purposes of the EAF assessment of the FW nozzle, it was assumed that HWC conditions exist for 47% of the time, and NWC conditions exist for 53% of the time over the 60-year operating life of the plant. RPV upper region chemistry was assumed for the FW nozzle blend radius location, since this location experiences reactor conditions for all times. FW line chemistry was assumed for the FW nozzle safe end location, since this location experiences feedwater conditions for all times. SIR-07-130-NPS, Rev. 0 5-1 Structural Integrity Associates, Inc. NEC066088

I For the safe end location, the environmental fatigue factors for pre-HWC and post-HWC are both 1.74 from Table 3 of Reference [21] for the RPV FW line. This results in an EAF adjusted CUF as follows: 60-Year CUF, U60 = 0.1470 (from Table 4-5) U Overall EAF multiplier, Fen = 1.74 60-Year EAF CUF, Uo-env = 0.14709 x 1.74 = 0.2560 The EAF CUF value of 0.2560 for 60 years for the safe end is acceptable (i.e., less than the I allowable value of t.0). The fatigue calculation documented in Section 4.0 for the blend radius location was performed for the nozzle base material since cladding is structurally neglected in modern-day fatigue analyses, per ASME Code, Section III, NB-3122.3 [15]. This is also consistent with Sections 5.7.1 and 5.7.4 of NUREG/CR-6260 [16]. Therefore, the cladding was neglected and EAF I assessment of the nozzle base material was performed for the blend radius location. For the blend radius location, the environmental fatigue factors for pre-HWC and post-HWC are 11.14 and 8.82, respectively, from Table 4 of Reference [21] for the RPV upper region. This results in an EAF adjusted CUF as follows: I 60-Year CUF, U60 = 0.0636 (from Table 4-4) Overall EAF multiplier, Fen (11.14 x 53% + 8.82 x 47%) = 10.05 60-Year EAF CUF, U6 -,env = 0.0636 x 10.05 = 0.6392 I The EAF CUF value of 0.6392 for 60 years for the blend radius is acceptable (i.e., less than the allowable value of 1.0). I SIR-07-130-NPS, Rev. 0 5-2 Structural Integrity Associates, Inc. NEC066089

6.0 CONCLUSION

S This report documents a refined fatigue evaluation for the VY FW nozzle. ý The intent of this evaluation is to use refined transient definitions and the revised cyclic transient counts for 60 years for a computation of CUF, including EAF effects, that is more refined than previously performed fatigue analyses. The fatigue-limiting locations in the FW nozzle and safe end are included in the evaluation, to be consistent with NUREG/CR-6260 [16] needs for EAF evaluation for license renewal. The final fatigue results are considered to be a replacement to the values previously reported in the VY LRA. The fatigue calculations for the VY FW nozzle were performed in accordance with ASME Code, Section III, Subsection NB-3200 methodology (1998 Edition, 2000 Addenda) [15]. The stress evaluation is summarized in Section 3.0, and the fatigue analysis is summarized in Section 4.0. The results in Section 4.0 reveal that the CUF for the limiting safe end location is 0.1470, and the CUF for the limiting blend radius location is 0.0636. Both of these values represent 60 years of plant operation, including all relevant EPU effects. EAF calculations for the VY FW nozzle were also performed, as summarized in Section 5.0. The results in Section 5.0 reveal that the EAFCUF for the limiting safe end location is 0.2560, and the EAF CUF for the limiting blend radius location is 0.6392. Both of these values represent 60 years of plant operation, including all relevant EPU effects. All fatigue allowables, both with and without EAF effects, are met, thus demonstrating acceptability for 60 years of operation. SIR-07-130-NPS, Rev. 0 -1Structural Integrity Associates, Inc. NEC066090

7.0 REFERENCES

I i REDACTED I

2.

CB&I RPV Stress Report, Sections T4 and S4, "Feedwater Nozzle, Vermont Yankee Reactor Vessel, CB&I Contract 9-6201," SI File No. VY-05Q-238. I

3.

GE Design Specification No. 21A 1115, Revision 4, "Vermont Yankee Reactor Pressure Vessel," October 21, 1969, SI File No. VY-05Q-2 10. 3

4.

Kuo, A. Y., Tang, S. S., and Riccardella, P. C., "An On-Line Fatigue Monitoring System for Power Plants, Part I - Direct Calculation of Transient Peak Stress Through Transfer Matrices and Green's Functions," ASME PVP Conference, Chicago, 1986.

5.

ANSYS, Release 8. 1 (w/Service Pack 1), ANSYS, Inc., June 2004.

6.

Structural Integrity Associates Calculation No. VY-10Q-301, Revision 0, "Feedwater Nozzle Finite Element Model and Heat Transfer Coefficients."

7.

Structural Integrity Associates Calculation No. YAEC-13Q-303, Revision 0, "Thermal I Transient Analysis."

8.

VY Drawing No. 5920-9057, Sheet 1, Revision 1, S1 File No. VY-05Q-215. / I

9.

American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section II, Part D, 1998 Edition, 2000 Addenda. I SIR-07-130-NPS, Rev. 0 7-1 Structural Integrity Associates, Inc. NEC066091

10.

Structural Integrity Associates Calculation No. VY-16Q-301, Revision 0, "Feedwater Stress History Development for Nozzle Green's Function."

11.

Vermont Yankee Drawing 5920-00024, Rev. 11, GE Drawing No. 919D294, Revision 11, Sheet No. 7, "Reactor Vessel," SI File No. VY-05Q-241L

12.

CB&I Addenda to RPV Stress Report, "Certification of Addenda to the Stress Report for Vermont Yankee Reactor Vessel," July 9. 1971, SI File No. VY-05Q-238.

13.

VY Calculation Change Notice (CCN), CCN Number 1 for Calculation VYC1005 Revision 2, "This CCN Provides a Basis for the Power Uprate Safety Analysis Report being submitted as part of the power uprate project. The 50.59 assessment will be handled by the EPU design change and NRC SER for this submittal." SI File Number VY-05Q-208.

14.

Structural Integrity Associates Calculation No. SW-SPVF-OIQ-301, Revision 0, "STRESS.EXE, P-V.EXE, and FATIGUE.EXE Software Verification."

15.

American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III Subsection NB, 1998 Edition, 2000 Addenda.

16.

NUREG/CR-6260 (INEL-95/0045), "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," March 1995.

17.

GE Certified Design Specification No. 26A6019, Revision 1, "Reactor Vessel - Extended Power Uprate," August 29, 2003, SI File No. VY-05Q-236.

18.

General Electric Stress Report No. DC22A5583, Revision 0, Section T, "Thermal Analysis FitzPatrick Feedwater Nozzle Modification," SI File No. NYPA-53Q-212. SIR-07-130-NPS, Rev. 0 7-2 Structural Integrity Associates, Inc. NEC066092

19.

Entergy Design Input Record (DIR) Revision 1, EC No. 1773, Revision 0 "Environmental Fatigue Analysis for Vermont Yankee Nuclear Power Station," 7/26/2007, SI File No. VY-16Q-209.

20.

Chicago Bridge & Iron Company Contractor 9-6201, Revision 2, '"Section S4, Stress Analysis Feedwater Nozzle Vermont Yankee Reactor Vessel," SI File No. VY-05Q-238.

21.

Structural Integrity Associates Calculation No. VY-16Q-303, Revision 0, "Environmental Fatigue Evaluation of Reactor Recirculation Inlet Nozzle and Vessel Shell Bottom Head."

22.

NUREG-1801, Revision 1, "Generic Aging Lessons Learned (GALL) Report," U. S. Nuclear Regulatory Commission, September 2005.

23.

Structural Integrity Associates Calculation No. VY-16Q-302, Revision 0, "Fatigue Analysis of Feedwater Nozzle."

24.

J. P. Holman, "Heat Transfer," 4th Edition, McGraw-Hill, 1976.

25.

J. P. Holman, "Heat Transfer," 5th Edition, McGraw-Hill, 1981.

26.

N. P. Cheremisinoff, "Heat Transfer Pocket Handbook," Gulf Publishing Co, 1984.

27.

American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section lII, Subsection A, Article 4, 1965 Edition with Winter 1966 Addenda. I I I I I I I I I I I I I I I I I SIR-07-130-NPS, Rev. 0 7-3 Structural Integrity Associates, Inc. I NEC066093

[REDACTED COPY ý. ý'ý; REDACTED COPY Structural Integrity Associates, Inc. File No.: VY-16Q-303 NEC-JH_06 CALCULATION PACKAGE Project No.: VY-16Q PROJECT NAME: Environmental Fatigue Analysis of VYNPS CONTRACT NO.:

10150394, CLIENT:

PLANT: Enter.gy Nuclear Operations, Inc. Verrnont Yankee CALCULATION TITLE: Environmental Fatigue Evaluation of Reactor Recirculation Inlet Nozzle and Vessel Shell/Bottom Head Document Affected Project Manager Preparer(s) & Revision Pages Revision Description Approval Checker(s) Signature & Date Signatures & Date 01 - 24, Initial issue. Terry J. Herrmann Gary L. Stevens Appendices: 07/05/07 07/05/07 Al -A2, BI - B2 t, In computer files Terry J. Herrmann 07/05/07 onC 09pietar9ma NEC065998 Page 1 of 24 F0306-01 RO

Structural Integrity Associates, Inc. Table of Contents

1.0 INTRODUCTION

/STATEMENT OF PROBLEM/ OBJECTIVE........................................... 3 3 2.0 TECHNICAL APPROACH OR METHODOLOGY................................................................ 3 3.0 ASSUMPTIONS / DESIGN INPUTS...................................................................................... 4 4.0 C A L C U L A T IO N S........................................................................................................................... 6 4.1 R PV Low er H ead...................................................................................................... 7 4.2 R R Inlet N ozzle............................................................................................ .................... 9 5.0 R E SU LT S O F A N A L Y SIS........................................................................................................... 11

6.0 CONCLUSION

S AND DISCUSSION......................................................................................... 11 7.0 RE F E R E N C E S.............................................................................................................................. 12 APPENDIX A VY WATER CHEMISTRY INFORMATION [8].............................................. Al APPENDIX B VY LICENSE DATE [10]................................................................................... B1 List of Tables Table 1: Water Chemistry Calculations........................................................................................ 14 U Table 2: Bounding F,,, Multipliers for Recirculation Line............................................................ 15 Table 3: Bounding Fen Multipliers for Feedwater Line................................................................ 16 Table 4: Bounding Fe,, Multipliers for RPV Upper Region.......................................................... 17 Table 5: Bounding Fon Multipliers for RPV Beltline Region....................................................... 18 Table 6: Bounding F,, Multipliers for RPV Bottom Head Region.............................................. 19 Table 7: EAF Evaluation for RPV Shell/Bottom Head Location................................................. 20 Table 8: EAF Evaluation for Limiting RPV Shell/Shroud Support Location............................... 21 Table 9: EAF Evaluation for RR Inlet Nozzle Forging Location................................................. 22 Table 10: EAF Evaluation for RR Inlet Nozzle Safe End Location............................................. 23 Table 11: Summary of EAF Evaluation Results for VY.............................................................. 24 I File No.: VY-16Q-303 Page 2 of 24 Revision: 0 Cont 's ndor P priet y Infor 10I C 9F0306-OIRO NEC065999

Structural Integrity Associates, Inc.

1.0 INTRODUCTION

/STATEMENT OF PROBLEM/ OBJECTIVE The purpose of this calculation is to perform a plant-specific evaluation of reactor water environmental effects for the reactor recirculation (RR) inlet nozzle and the reactor pressure vessel (RPV) shell/bottom head locations identified within NUREG/CR-6260 [1] for the older vintage General Electric (GE) plant for the Vermont Yankee Nuclear Power Plant (VY). The water chemistry input used in this calculation covers several portions of the RPV, as well as the feedwater and recirculation lines. Although these regions encompass more areas than needed to address the two components of interest in this calculation, environmental fatigue multipliers are developed for all of these regions in this calculation for potential use in other evaluations associated with this project. 2.0 TECHNICAL APPROACH OR METHODOLOGY Per Chapter X, "Time-Limited Aging Analyses Evaluation of Aging Management Programs Under 10 CFR 54.21 (c)(1)(iii)," Section X.M1, "Metal Fatigue of Reactor Coolant Pressure Boundary," of the Generic Aging Lessons Learned (GALL) Report [2], detailed, vintage-specific, fatigue calculations are required for plants applying for license renewal for the locations identified for the appropriate vintage plant in NUREG/CR-6260. In this calculation, detailed environmentally assisted fatigue (EAF) calculations are performed for VY for two of the locations associated with the older vintage GE plant in NUREG/CR-6260. The older-vintage GE plant is the appropriate comparison to VY since the original piping design at VY was in accordance with USAS B31.1 [3], as well as the fact that the older-vintage boiling water reactor (BWR) in NUREG/CR-6260 was a BWR-4 plant, which is the same as VY. Entergy performed an initial assessment of EAF effects for VY in their License Renewal Application (LRA) that was submitted to the NRC in January 2006. Table 4.3-3 of the VY LRA provides the results of those evaluations. All but two of the VY locations evaluated for EAF in the LRA did not yield acceptable results for 60 years of operation. Further refined analyses are currently underway in other calculations associated with this project to address those components. This calculation documents the EAF evaluation for the RR inlet nozzle and RPV shell/bottom head locations, where it is expected that acceptable EAF results can be achieved based on the existing analyses without the need for additional refined evaluations. File No.: VY-16Q-303 Page 3 of 24 Revision: 0 ý tF0306-0IRO NEC066000

V Structural Integrity Associates, Inc. I I 3.0 ASSUMPTIONS / DESIGN INPUTS Per Section X.MI of the GALL Report [2], the EAF evaluation must use the appropriate Fen relationships from NUREG/CR-6583 [4] (for carbon/low alloy steels) and NUREG/CR-5704 [5] (for stainless steels), as appropriate for the material for each location. These expressions are: Fen = exp (0.585 - 0.00124T' - 0.101S*T*O* -*) I For Carbon Steel [4, p. 691: Substituting T' = 25'C in the above expression, as required by NURE.G/CR-6583 to relate room temperature air data to service temperature data in water [6], the following is obtained: Fen exp (0.585 - 0.00124(25°C) - 0.101 S* T* 0* "") - exp (0.554 - 0.101 S* T* 0* 0*) For Lowv Alloy Steel [4, p. 69]: F_, = exp (0.929 - 0.00124T' - 0.101S*T*O* E*) Substituting T= 25'C in the above expression, as required by NUREG/CR-6583 to relate room temperature air data to service tcmperaturc data in water [6], the following is obtained: where [4, pp. 60 and 65]: Fen T* m F~n = exp (0.929 - 0.00124(25°C) - 0.101 S* T* OF:-*) =exp (0.898 - 0.101 S* T* O*c) fatigue life correction factor S for 0 < sulfur content, S < 0.0 15 wt. % 0.015 for S > 0.015 wt. % = 0for T < 150'C (T - 150) for 150 < T < 350'C fluid service temperature (QC) 0 for dissolved oxygen, DO < 0.05 patls per million (ppm) ln(DO/0.04) for 0.05 ppm < DO < 0.5 ppm = ln(l 2.5) for DO >.0.5 pprn O* I I I I U I I I I I I I I = 0 for strain rate, c > I %/sec ln(c*) for 0.001 < C 1 %/sec ln(0.00t) for - < 0.001%/sec File No.: VY-16Q-303 Revision: 0 NECO66001 Page 4 of 24 10306-01 RO

Structural Integrity Associates, Inc. For Types 304 and 316 Stainless Steel [5, p. 31]: Fen =exp (0.935 - T* e* 0*) where [5, pp. 25 and 31]: Fn = fatigue life correction factor T* = 0 for T < 2000C = I forT_ Ž200'C T = fluid service temperature (0C) 6* = 0 for strain rate, E > 0.4%/sec = ln(E/0.4) for 0.0004 _< E: _ 0.4%/sec = ln(0.0004/0.4) for P < 0.0004%/sec 0* = 0.260 for dissolved oxygen, DO < 0.05 parts per million (ppm) = 0.172 for DO >_ 0.05 ppm Bounding Fe, values are determined or, where necessary, computed for each load pair in the detailed fatigue calculation for each component. The environmental fatigue is then determined as Uen,, = (U) (Fe,), where U is the original fatigue usage and Un,,, is the environmentally assisted fatigue (EAF) usage factor. All calculations can be found in Excel spreadsheet "VY-16Q-303 (Env. Fat. Calcs).xls" associated with this calculation. From Reference [7], for the BWR, typical DO levels range from just over 200 ppb for normal water chemistry (NWC) conditions to less than 10 ppb for hydrogen water chemistry (HWC) conditions. Typical HWC system availabilities are greater than 90%. Based on VY-specific water chemistry input for Entergy [8], which is also contained in Appendix A of tlhis calculation, the input shown in Table I is defined for use in this calculation. The water chemistry input covers several portions of the RPV, as well as the feedwater and recirculation lines. Although these regions encompass more areas than needed to address the two components of interest in this calculation, environmental fatigue multipliers are developed for all of these regions in this calculation for potential use in other evaluations associated with this project. Therefore, based on Table 1 and for the purposes of this calculation, the following is assumed: " Over the 60-year operating life of the plant, HWC conditions exist for 47% of the time, and NWC conditions exist for 53% of the time. All operation through 11/1/2003 was assumed as NWC using the dissolved oxygen values from the "Pre-NMCA" column in Appendix A, and all operation after 11/1/2003 was assumed as HWC using the maximum oxygen values from the "Post-NMCA + HWC (OLP)", "Post-NMCA + HWC (EPU)", and "Future Operation" colulmns in Appendix A.

Recirculation line DO is 122 ppb pre-HWC and 48 ppb post-HWC.

Feedwater line DO is 40 ppb for pre-HWC and 40 ppb for post-HWC conditions.

RPV Upper Region DO is 114 ppb pre-HWC and 97 ppb post-HWC.
RPV Beltline DO is 123 ppb pre-HWC and 46 ppb post-HWC.

RPV Bottom Head Region DO is 128 ppb pre-HWC and 69 ppb post-HWC. File No.: VY-16Q-303 Page 5 of 24 Revision: 0 F0306-01R0 NEC066002

Structural Integrity Associates, Inc. Based on the above typical DO levels, bounding Fn multipliers for each of the three applicable materials (carbon, low alloy, and stainless steels) are shown in Tables 2 through 6 for the various RPV and piping regions. The projected number of cycles used in this calculation is based oil the number of cycles actually experienced by the plant in the past and forward-projected with some additional margin for 60 years of operation, as documented in Reference [9]. In addition, the latest governing stress analysis for U each location was utilized, and any relevant effects of Extended Power Uprate (EPU) operation were incorporated as necessaiy. With these assumptions, the cumulative usage factor (CUF) values documented in th-tis calculation are considered applicable for sixty years of operation including all i relevant EAF and EPU effects. U 4.0 CALCULATIONS The analyses for the NUREG/CR-6260 locations identified in Section 2.0 are provided in this i section. As previously noted, the fatigue calculations for 60 years for all locations make use of the 60-year projected cycles for VY from Reference [9], and incorporate EPU effects. 3 Since the Fn methodology documented in References [4] and [5] is relatively "new" technology, it is intended to apply to "modern-day" fatigue analyses, i.e., applied to fatigue analyses that use current ASME Code fatigue curves, etc. Therefore, to be consistent with this approach, the evaluation for the all locations will also utilize modern-day fatigue calculation methodology using the 1998 Edition, 2000 Addenda of the ASME Code [ I I]. This involves applying a Young's Modulus ColTection factor (i.e., Efartguc cur,,c/Eanatysis) to the calculated stresses, applying K. where appropriate, and utilizing the 2000 Addenda fatigue curve. NOTE: It is recognized that some of the references used in this calculation are not the latest I revision;for example, Reference [12] (VYC-378, Revision 0) has been revised. However, the details necessary to petform the evaluations in this calculation are not necessarily contained in the latest revision of all documents. Therefore, wherever necessar', the I appropriate revision of the governing document is referenced in order to obtain all appropriate inputs necessaty to pelform the EAF calculations. So, it should be recognized that, despite using what appear to be outdated revisions of some references, use of these i references is for input data use only. All calculations represent the latest available analyses .tbr all locations. NOTE: Hand calculations may yield results slightly different than the values shown in the tables of this calculation due to round-off based on the significant figures utilized by the spreadsheet used for these calculations. I File No.: VY-16Q-303 Page 6 of 24 Revision: 0 F0306-01 RO NEC066003 3

V Structural Inte/rity Associates, Inc. 4.1 RPV Lower Head The 60-year CUF value (without EAF effects) for the RPV shell/bottom head location was reported in Table 4.3-3 of the VY LRA submittal to be 0.400. The EAF CUF estimated by Entergy for this location was 0.98, based on an overall F,,i of 2.45. Based on this result, further refined analysis would not normally be necessary to show acceptable EAF CUF results for this component. However, the calculation for this location is updated in this section to reflect the updated water chemistry information supplied for this project. The CUF value reported in the VY LRA for the RPV shell/bottom head location is 0.400. This value is the original design basis CUF from the RPV Stress Report, as noted on page B8 of Reference [12]. However, as noted on page A61 of Reference [12], this CUF corresponds to Point 8, which is located on the outside surface of the RPV bottom head at the junction with the support skirt. Therefore, this location is not exposed to the reactor coolant, and EAF effects do not apply. Based on this, evaluation of the limiting location along the inside surface of the RPV bottom head was performed. Based on a review of the primaly plus secondary stresses tabulated for all locations along the bottom head on page A52 of Reference [12], Point 14 was selected for EAF evaluation. Per Section 3.2.1.2 of Reference [ 13], none of the CUF values for the RPV bottom head region were evaluated for the effects of EPU, as the CUF values are below the EPU screening criteria value of 0.5. Therefore, as a part of the evaluation for this location, EPU effects were included. Per References [14] and [19], the RPV shell material is low alloy steel (A-533, Grade B). The new CUF calculation for Point 14 for 40 years, which includes the use of updated methodology and incorporates EPU effects [14], is shown at the top portion of Table 7. The CUF for 40 years (without EAF effects) is 0.0057. The fatigue calculation for 60 years for the RPV shell/bottom head location is also shown in Table 7. The results show a CUE (without EAF effects) of 0.0085 for 60 years. The fatigue calculation for 60 years makes use of the 60-year projected cycles for VY from Reference [9]. The resulting environmental fatigue calculation for the RPV shell/bottom head location is shown in Table 7. Bounding Fn multipliers were applied in the calculations. RPV bottom head water chemistry conditions from Tables 1 and 6 are used for this location. The results show an EAF adjusted CUF of 0.0809 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). The CUF determined for Point 14 is very low. Comparison to other locations of the RPV shell/bottom head region indicates it is not the limiting location from a fatigue perspective. Review of the CUF values in Table 3-1 of Reference [15] reveals that the shroud support (at vessel wall junction) location is potentially more limiting, so EAF evaluation of that location is also performed. Per page $3-99f of Reference [16], the design basis CUF of 0.06 is for Point 9. Page S3-85 of Reference [ 16] reveals that this point is on the RPV shell at the junction of the shroud support plate. Per References [14] and [19], the RPV shell material is low alloy steel (A-533, Grade B). File No.: VY-16Q-303 Page 7 of 24 Revision: 0 ont s ýnd or P>ro r y ýIn fo r ma 0 F0306-01 RO NEC066004

Structural Integrity Associates, Inc. The revised and updated CUF calculation for Point 9 for 40 years, which includes the use of updated methodology and incorporates EPU effects, is shown at the top portion of Table 8. The CUF for 40 years (without EAF effects) is 0.0549. This CUF value is more limiting than the RPV shell/bottom head location evaluated in Table 7, so it is considered to be the governing location for VY with respect to the equivalent NUREG/CR-6260 RPV shell/bottom head location. The fatigue calculation for 60 years for the RPV shell/shroud support location is also shown in Table 8. The results show a CUF (without EAF effects) of 0.0774 for 60 years. The fatigue calculation for 60 years makes use of the 60-year projected cycles for VY from Reference [9]. The resulting environmental fatigue calculation for the RPV shell/shroud support location is shown in Table 8. Bounding F,, multipliers were applied in the calculations. RPV bottom head water chemistry conditions from Table 6 are used for this location. The results show an EAF adjusted CUF of 0.7364 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). I I I I I I I I I I I I I I I I I I I File No.: VY-16Q-303 Revision: 0 Page 8 of 24 F0306-01 RO ~enror~r.ynfori~ NEC066005

Structural Integrity Associates, Inc. 4.2 RR Inlet Nozzle For conservatism due to the different materials involved, two locations are evaluated for the RR inlet nozzle: (1) the limiting location in the nozzle forging, and (2) the limiting location in the safe end. The 60-year CUF value (without EAF effects) for the RR inlet nozzle in the VY LRA submittal is 0.610. However, that analysis used conservative transient definitions and cyclic projections for 60 years of operation that have since been updated. The applicable CUF values are those shown in Table 3-1 of Reference [15] (0.1058 for the safe end, and 0.03 for the nozzle for 40-years), except that these values are pre-EPU. For the RR inlet nozzle forging, the governing CUF calculation is shown on page B28 of Reference [12], where a value of 0.03 was obtained. From pages A269 and A270 of Reference [12], the CUF calculation corresponds to Point 12 in the nozzle forging, which is on the outside surface of the nozzle on the outboard end of the nozzle transition. Although this location is not exposed to the reactor coolant, it will be conservatively evaluated for EAF effects as it is the bounding fatigue location in the nozzle forging. As a part of the evaluation for this location, EPU effects were included. Per page 1-$8-4 of Reference [17], the RR inlet nozzle material is low alloy steel (A-508 Class II). The new CUF calculation for Point 12 for 40 years, which includes the use of updated methodology and incorporates EPU effects [ 14], is shown at the top portion of Table 9. The CUF for 40 years (without EAF effects) is 0.0433. The fatigue calculation for 60 years for the RR inlet nozzle forging location is also shown in Table 9. The results show a CUF (without EAF effects) of 0.0650 for 60 years. The fatigue calculation for 60 years makes use of the 60-year projected cycles for VY from Reference [9]. The resulting environmental fatigue calculation for the RR inlet nozzle forging location is shown in Table 9. Bounding Fe,, multipliers were applied in the calculations. RPV beltline water chernistry conditions from Table 5 are used for this location. The results show an EAF adjusted CUF of 0.5034 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0) For the RR inlet nozzle safe end, the governing CUF calculation is shown on page.B27 of Reference [12], where a value of 0.1058 was obtained. From pages A257 and A259 of Reference [12], the CUF calculation corresponds to Line 6 at the inside surface of the safe end. Page A238 of Reference [12] reveals that this location is location at the nozzle-to-safe end weld. Per Section 3.2.1.2 of Reference [13], the CUF value for the RR inlet nozzle safe end was evaluated for the effects of EPU, since the original CUF calculated in Reference [18] was 0.551 (which was adjusted downward to 0.1,058 by Entergy in Reference [12] based on further refined-evalua-tioin). Therefore, as a part of the evaluation for this location, EPU effects were ificluded. Per page 8 of Reference [18], the RR inlet nozzle safe end material is 316L qtainless steel. File No.: VY-16Q-303 Page 9 of 24 Revision: 0 F0306-0 I RO NEC066006

V Structural Integrity Associates, Inc. The new CUF calculation for the RR inlet nozzle safe end for 40 years, which includes the use of updated methodology and incorporates EPU effects [14], is shown at the top portion of Table 10. The CUF for 40 years (without EAF effects) is 0.00 17. The fatigue calculation for 60 years for the RR inlet nozzle safe end location is also shown in Table 10. The results show a CUF (without EAF effects) of 0.0017 for 60 years. The fatigue calculation for 60 years makes use of the 60-year projected cycles for VY from Reference [9]. The resulting environmental fatigue calculation for the RR inlet nozzle safe end location is shown in Table 10. Bounding F,. multipliers were applied in the calculations. Recirculation line water chemistry conditions from Table 2 are used for this location. The results show an EAF adjusted CUE of 0.0199 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0) I I I I I I I I I I I I I I I I I I I File No.: VY-16Q-303 Revision: 0 Page 10 of 24 NEC066007 F0306-01 RO

Structural Integrity Associates, Inc. 5.0 RESULTS OF ANALYSIS The final environmental fatigue results contained in Sections 4.1 and 4.2 (and associated Tables 7 through 10) for the RPV shell/bottom head and RR inlet nozzle locations are summarized in Table 11.

6.0 CONCLUSION

S AND DISCUSSION In this calculation, EAF calculations were performed in accordance with the GALL Report [2] for the following VY locations: " RR inlet nozzle, consisting of the following bounding locations: o Nozzle forging (low alloy steel) o Safe end (stainless steel) RPV shell/bottom head, consisting of the following bounding locations: o Limiting bottom head shell inside surface location (low alloy steel) o Limiting RPV shell/shroud support location (low alloy steel) The above locations were selected based on the locations identified in NUREG/CR-6260 for the older vintage GE plant and plant-specific fatigue calculations that determined the limiting locations for VY. Calculations for the remaining NUREG/CR-6260 locations will be documented in other analyses performed under this project. The EAF results for the locations identified above are shown in Table 1I. These results indicate that the fatigue usage factors, including environmental effects, are within the allowable value for 60 years of operation for all locations evaluated. The calculations for all locations rnake use of the 60-year projected cycles for VY and incorporate EPU effects. Therefore, no additional evaluation is required for these components, and the GALL requirements are satisfied. File No.: VY-16Q-303 Revision: 0 Page II of 24 . on ta _s-dor ýP r ia ~ry ýIn r ma ti ýn F0306-01 RO NEC066008

Structural Integrity Associates, Inc.

7.0 REFERENCES

I

1. NUREG/CR-6260 (INEL-95/0045), "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," March 1995.

I

2. NUREG-1801, Revision 1, "Generic Aging Lessons Learned (GALL) Report," U. S. Nuclear Regulatory Commission, September 2005.

I

3. USAS B31.1.0 - 1967, USA Standard Code for Pressure Piping, "Power Piping," American Society of Mechanical Engineers, New York.

4. NUREG/CR-6583 (ANL-97/I 8), "Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels," March 1998.

5. NUREG/CR-5704 (ANL-98/31). "Effects of LWR Coolant Environmlents on Fatigue Design Curves of Austenitic Stainless Steels," April 1999.

6. EPRI/BWRViP Memo No. 2005-271, "Potential Error in Existing Fatigue Reactor Water Environmental Effects Analyses," July 1,2005.

REDACTED I ii"

8. "Vermont Yankee Dissolved Oxygen (DO) Levels for Use in EAF Evaluations," page 11 of Entergy Design Input Record (DIR) EC No. 1773, Revision 0, "Environmental Fatigue Analysis I

for Vermont Yankee Nuclear Power Station," 7/3/07, SI File No. VY-16Q-209.

9. "Reactor Thermal Cycles for 60 Years of Operation," Attachment I of Entergy Design Input Record (DIR) EC No. 1773, Revision 0, "Environmental Fatigue Analysis for Vermont Yankee Nuclear Power Station," 7/3/07, S1 File No. VY-16Q-209.

10. VY LRA, page 1-4 (included as Appendix B to this calculalion).

11. American Society of Mechanical Engineers Boiler & Pressure Vessel Code, Section III, Rules for Construction of Nuclear Facility Components, and Section II, Materials, Par-t D, "Properties (Customary)," 1998 Edition including the 2000 Addenda.

1.2. Yankee Atomic Electric Company Calculation No. VYC-378, Revision 0, "Vermont Yankee Reactor Cyclic Limits for Transient Events," 10/16/85, SI File No. VY-05Q-21 1. REDACTED File No.: VY-16Q-303 Page 12 of 24 Revision: 0 3 F0306-OIRO NEC066009

Structural Integrity Associates, Inc.

14. GE Nuclear Energy Certified Design Specification No. 26A6019, Revision 1, "Reactor Vessel -

Extended Power Uprate," June 2, 2003, SI File No. VY-05Q-236.

15. Structural Integrity Associates Report No. SIR-01-130, Rev. 0, "System Review and Recommendations for a Transient and Fatigue Monitoring System at the Vermont Yankee Nuclear Power Station," Februry 2002, S1 File No. W-VY-05Q-401.

16. CB&I RPV Stress Report, Section S3. Revision 4, "Stress Analysis, Shroud Support, Vermont Yankee Reactor Vessel, CB&I Contract 9-620 1," 2-3-70, SI File No. VY-16Q-203.

17. CB&I RPV Stress Report, Section S8, Revision 4, "Stress Analysis, Recirculation Inlet Nozzle, Vermont Yankee Reactor Vessel, CB&I Contract 9-6201," 2-3-70, SI File No. VY-16Q-203.

18. GE Nuclear Energy Certified Stress Report No. 23A4292, Revision 4, "Reactor Vessel -

Recirculation Inlet Safe End Nozzle," March 12, 1986, S1 File No. VY-16Q-203.

19. Entergy Drawing No. 5920-5752, Revision 3 (CB&I Drawing No. RI5, Revision 1), "Vessel &

Attachments Mat'l. Identifications," 1/20/88, S1 File No. VY-16Q-209. File No.: VY-16Q-303 Revision: 0 Page 13 of 24 NECO660O10 F0306-01 RO

V Structural Integrity Associates, Inc. Table 1: Water Chemistry Calculations Date of HWC Implementation: Availability of HWC System Since HWC Implementation: Projected Future HWC System Availability: Recirculation Line DO pre-HWC: post-HWC: Feedwater Line DO pre-HWC: post-HWC: RPV Upper Region DO pre-HWC: post-HWC: RPV Beltline Region DO pre-HWC: post-HWC: RPV Bottom Head Reqion DO 11/01/2003 (see Appendix A) 98.54% (see Appendix A) 98.5% (see Appendix A, assume same as recent experience) 122 48 40 40 114 97 ppb (see Appendix A) ppb (see Appendix A) ppb (see Appendix A) ppb (see Appendix A) ppb (see Appendix A) ppb (see Appendix A) ppb (see Appendix A) ppb (see Appendix A) ppb (see Appendix A) ppb (see Appendix A) I I I I I I I I I I I I I I I pre-HWC: 128 post-HWC: 69 Plant Startup Date: Time at pre-HWC Conditions: Date of Calculations: Time Since HWC Implementation: Projected Future Time for HWC Operation: Overall HWC Availability: 03/22/1972 (see Appendix B) 31.61 years (calculated, includes leap years.) 04/30/2007 3.49 years (calculated, includes leap years.) 24.90 years (calculated, includes leap years.) 47% Note: All operation through 11/1/2003 was assumed as NWC using the dissolved oxygen values from the "Pre-NMCA" column in Appendix A, and all operation after I 1/1/2003 was assumed as HWC using the maximum oxygen values from the "Post-NMCA + HWC (OLP)", "Posi-NMCA + HWC (E1PU)", and "Future Operation" columns in Appendix A. File No.: VY-16Q-303 Revision: 0 Page 14 of 24 -.ontain" en Propr Informati F0306-0I RO NEC066011

V Structural Integrity Associates, Inc. Table 2: Bounding Fun Multipliers for Recirculation Line Low Afloy Steel: Fen = exp(0. 898 - 0. 101S10. c. ) Assume S* = 0.015 (maximum) Assume,,. = In(O.001) = -6.908 (minimum) For a BWR with HWC environment (post-HWC implementation): DO = 48 ppb = 0.048 ppm DO < 0.050 ppm, so 0* = 0 Thus: T (-C) T (-F) Fen 0 32 2.45 50 122 2.45 100 212 2.45 150 302 2.45 200 392 2.45 250 482 2.45 288 550 2.45 For a BWR with NWC environment (pre-HWC implementation): DO = 122 ppb = 0.122 ppm, so 0* = In(0.122/0.04) = 1.115 Thus: T (-C) T (°F) Fen 0 32 2.45 50 122 2.45 100 212 2.45 150 302 2.45 200 392 4,40 250 482 7.89 288 550 12.29 Thus, maximum Fen, 2.45 IT*= (T-150) for T> 150°C] Thus, maximum Fr, = 12.29 Carbon Steel: F, = exp(O.554 - 0.IOtS'T*O*ct) Assume S* = 0.015 (maximum) Assume c. = ln(0.001) = -6.908 (minimum) For a BWR with HWC environment (post-HWC implementation): DO = 48 ppb = 0.048 ppm DO< 0.050 ppm, so 0* = 0 Thus: T ('C) T ('F) Fen 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1.74 200 392 1.74 250 482 1.74 288 550 1.74 For a BWR with NWC environment (pre-HWC implementation). DO = 122 ppb = 0.122 ppm, so O* = ln(0.122/0.04) = 1.115 Thus: T (°C) T (-F) Fn 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1.74 200 392 3.12 250 482 5.59 288 550 8.71 Thus, maximum Fen = 8.71 Thus, maximum Fen,: 1.74 [T*= (T-150) for T t501C] Stainless Stee/. Fen = exp(0.935 - T*,*O*) For a BWR with HWC environment (post-HWC implementation): For a BWR with NWC environment (pre-HWC implementation): DO = 48 ppb = 0.048 ppm < 0.050 ppm, so 0* = 0.260 DO = 122 ppb = 0.122 ppm > 0.05 ppm, so 0* = 0.172 Conservatively use T* = 1 for T, 200'C Conservatively use T* = 1 for T > 200°C Thus: Thus:

0 for c > 0.4%/sec so Fen

2.55 so Fen = 2.55 r.* = ln(r./0.4) for 0.0004 <=. -= 0.4%/sec so Fen ranges from 2.55 so Fen ranges from 2.55 to 15.35 to 8.36 c* = ln(0.0004/0.4) for,: < 0 0004%/sec so Fen = 15.35 so Fn = 8.36 Thus, maximum Fen = 15.35 Thus, maximum Fen, = 8.36 File No.: VY-16Q-303 Revision: 0 Page 15 of 24 Co/ dor op etar r "n F0306-01 RO NEC066012

Structural Integrity Associates, Inc. Table 3: Bounding Fea Multipliers for Feedwater Line Low Alloy Steel: Fen = exp(0.898 -0.101S.T..) For a BWR with HWC environment (post-HWC implementation): DO = 40 ppb = 0.040 ppm < 0.050 ppm so 0* = 0 Thus: T (-C) T (°F) F.n 0 32 2.45 50 122 2.45 100 212 2.45 150 302 2.45 200 392 2.45 250 482 2.45 288 550 2.45 Assume S" = 0.015 (maximum) Assume c: = ln(0.001) = -6.908 (minimum) For a BWR with NWC environment (pre-HWC implementation): DO = 40 ppb = 0.040 ppm < 0.050 ppm so 0* = 0 Thus: T (-C) T (IF) IF_ 0 32 2.45 50 122 2.45 100 212 2.45 150 302 2.45 200 392 2.45 250 482 2.45 288 550 2.45 Thus, maximum F- = 2.45 [P= iT-150) for Ta 150°C) Thus, maximum Fe, = 245 Carbon Steel: F - exp(0.5 54 0.101ST'O't:') Carbon Steel: Fn = exp(0.554 - 0.101S'T*O*1;*) Assume S* = 0.015 (maximum) Assume v* = ln(0.001) = -6.908 (minimum) For a BWR with HWC environment (post-HWC implementation): DO = 40 ppb = 0.040 ppm, 0.050 ppm so 0' = 0 Thus: T (°C) T (-F) FR 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1.74 200 392 1.74 250 482 1.74 288 550 1.74 For a BWR with NWC environment (pre-HWC implementation): DO = 40 ppb = 0.040 ppm < 0.050 ppm so O* = 0 Thus: I I I I I I I I I I I I I I T (-C) T (°F) F00 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1.74 200 392 1.74 250 482 1.74 288 550 1.74 Thus, maximum Fe. 1.74 [IT= iT-150) for Ta> 150°C Thus, maximum F., 1.74 There is no stainless steel in the Class I feedwater line. File No.: VY-16Q-303 Revision: 0 Page 16 of 24 Lopýsdor ýoa i ý fo-s, F0306-01RO NEC066013

Structural Integrity Associates, Inc. Table 4: Bounding Fe. Multipliers for RPV Upper Region Low Alloy Steel: F_=explO.898 - 0.101S*T*0*:1 For a BWR with HWC environment (post-HWC implementation): DO = 97 ppb = 0.097 ppm, so 0' = In(0.097/O.04) = 0.886 Thus: T (-C) T (°F) F_0 0 32 245 50 122 2.45 100 212 2.45 150 302 2.45 200 392 3.90 250 482 6.20 288 550 8.82 Assume S* = 0.015 (maximum) Assume u. = tn(O.01) -6.908 (minimum) For a BWR with NWC environment (pre-HWC implementation): DO = 114 ppb = 0.114 ppm, soO = ln(O.114/0.04) = 1.047 Thus: T (°C) T (°F) F_0 0 32 2.45 50 122 2.45 100 212 2.45 150 302 2.45 200 392 4.25 250 482 7.35 288 550 11.14 Thus, maximum F., 8.82 rT*= (T-150) for T >150'q] Thus, maximum F., = 11.14 Carbon Steel: eup(0.554 - 0.101STOu2 Carbon Steel. F_ exp(O.554 - 0.101 S'T*O'c*) Assume S = 0,015 (maximum) Assume F. In(0.001) = -6.908 (minimum) For a BWR with HWC environment (post-HWC implementation): DO = 97 ppb = 0.097 ppm, so 0' = ln(0.097/0.04) = 0.886 Thus: T (°C) T (-F) F-n 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1.74 200 392 2.77 250 482 4.40 288 550 6.25 For a 13WR with NWC environment (pre-HWC implementation): DO = 114 ppb = 0.114 ppm, so ' = In(0.114/004) = 1047 Thus: T (°C) T (-F) F_ 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1.74 200 392 3.01 250 482 5.21 288 550 7.90 Thus, maximum Fn 6.25 IT*= (T-150) for T 15a0 C] Thus, maximum Fn 7.90 Stainless Steel: F_ = exp(0.935 - T',O*) For a BWR with HWCenvironment (post-HWC implementation): For a BWR with NWC environment (pre-HWC implementation): DO = 97 ppb = 0.097 ppm > 0.050 ppm, so

= 0.172 DO = 114 ppb = 0114 ppm > 0.05 ppm, so O0 = 0.172 Conservatively use T* = 1 for T > 200'C Conservatively use T- = 1 for T > 200°C Thus:

Thus:

0 for r. > 0.4%/sec so Fc.

2.55 so Fun 2.55 SIn(s/0.4) for 0.0004 <= F <= 0.4%/sec so F_0 ranges from 2.55 so F,0 ranges from 2.55 to 8.36 to 8.36

In(0,0004/0.4) for e < 0.0004%/sec so F_

836 so F., = 8.36 Thus, maximum F-0 = 8.36 Thus, maximum F., = 8.36 File No.: VY-16Q-303 Revision: 0 Page 17 of 24 ý IsVrrP~rie~ta~mý F0306-OIRO NEC066014

Structural Integrity Associates, Inc. I I I Table 5: Bounding Fee Multipliers for RPV Beltline Region Low Alloy Steel: F_, = exp(0.898 - 0, 101 S'TO*,*i Ii For a BWR with HWC environment (post-HWC implementation): DO = 46 ppb = 0.046 ppm DO < 0050 ppm, so 0* = 0 Thus: T (°C) T (°F) Fe. 0 32 2.45 50 122 2.45 100 212 2.45 150 302 2.45 200 392 2.45 269.45 517.01 2.45 288 550 2.45 Assume S* = 0.015 (maximum) Assume ro = In(O.O01) - -6.908 (minimum) For a BWR with NWC environment (pre-HWC implementation): DO = 123 ppb = 0.123 ppm, so O* = ln(0.123/0.04) = 1.123 Thus: T (°C) T (°F) F_, 0 32 2.45 50 122 2.45 100 212 245 150 302 2.45 200 392 4.42 269.45 517.01 10.00 288 550 12.43 Thus, maximum F_,, = 12.43 Thus, maximum F0,, 2.45 T= (T 150) for T> 15o0C] Carbon Steel: For a BWR with HWC environment (post-HWC implementation): DO = 46 ppb = 0.046 ppm DO < 0.050 ppm, so 0* - 0 Thus: T (-C) T (oF) F.. 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1.74 200 392 1.74 250 482 1.74 288 550 1.74 F_,, exp(0.554 - 0 101S'T*O*E') Assume S* = 0.015 (maximum) Assume F. = In(O.01) = -6.908 (minimum) For a BWR with NWC environment (pre-HWC implementation): DO = 123 ppb = 0.123 ppm, so O* = ln(0.123!0.04) = 1.123 Thus: T (°C) T (F) F0. 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1.74 200 392 3.13 250 482 5.64 288 550 8.81 I I I I I I I I I I I I I I I Thus, maximum F-: 1.74 )'= (T-1 50) for T,150,q) Thus, maximum F., = 8.81 Stainless Steel: F- = exp(0.935 - T *crO*) For a BWR with HWC environment (posl-HWC implementation): For a BWR with NWC environment (pre-HWC implementation): DO = 46 ppb = 0.046 ppm < 0.050 ppm, so 0* = 0.260 DO = 123 ppb = 0.123 ppm >0.05 ppm, so 0* = 0.172 Conservatively use T* = 1 for T > 200*C Conservatively use T' = 1 for T > 200°C Thus: Thus: r=0 for -. > 0.4%/sec so F_ = 2.55 so F_, = 2.55 C = n(lr'0.4) for 0.0004 <= f- <= 0.4%/sec so F., ranges from 2.55 so Fn ranges from 2.55 to 15.35 to 8.36 n= n(0.0004/0.4) for c. < 0.0004%/sec so F_, = 15.35 so F., = 8.36 Thus, maximum F., = 15.35 Thus, maximum F. = 8.36 File No.: VY-16Q-303 Revision: 0 Page 18 of 24 Uo e~n_ý, ýe r Pr ý' r y ýa t io~n F0306-OIRO NEC066015

V Structural Integrity Associates, Inc. Table 6: Bounding Fen Multipliers for RPV Bottom Head Region Low Alloy Steel: F_, = exp(0.898 -OI 01~S'T-O-t:1 Assume S* = 0.015 (maximum) Assume r:* = ln(0.001) = -6.908 (minimum) For a BWR with HWC environment (post-HWC implementation): DO = 69 ppb = 0.069 ppm, so O0 = In(0.069/0.04) = 0.545 Thus: T (0C) T ('F) F_, 0 32 2.45 50 122 2.45 100 212 2.45 150 302 2.45 200 392 3.27 250 482 4.34 288 550 5.39 For a BWR with NWC environment (pre-HWC implementation): DO = 128 ppb = 0.128 ppm, so O* = ln(0.128/0.04( = 1.163 Thus: T (-C) T (-F) Fe, 0 32 2.45 50 122 2.45 100 212 2.45 150 302 2.45 200 392 4.51 250 482 8.29 288 550 13.17 Thus, maximum F., 5.39 [T'= (T-150) tor Tr 150,C] Thus, maximum F., 13.17 Carbon Steel: F_, = exp(O.554 - 0.101S-T*O*r*) Assume S* = 0.015 (maximum) Assume e' = ln(0.001) = -6.908 (minimum) For a BWR with NWC environment (pre-HWC implementation): DO = 128 ppb = 0.128 ppm. so 0* = ln{O.1280.04) = 1.163 Thus: For a BWR with HWC environment (post-HWC implementation): DO = 69 ppb = 0.069 ppm, so 0' = tn(0.069/0.04) = 0.545 Thus: T (°C) T ('F) F., 0 32 1.74 50 122 1,74 100 212 1.74 150 302 1.74 200 392 2.31 250 482 3.08 288 550 3.82 T (.C) T ('F) F_, 0 32 1.74 50 122 1.74 100 212 1.74 150 302 1 74 200 392 3.20 250 482 5.88 288 550 934 Thus, maximum F-, = 3.82 [T*= (T-150) for T> 150"C] Thus, maximum F-n 9.34 Stainless Steel: F_,= exp(0.935 - TrCO( For a BWR with HWC environment (post-HWC implementation): For a BWR with NWC environment (pre-HWC implementation): DO = 69 ppb = 0.069 ppm > 0,050 ppm, so

= 0.172 DO = 128 ppb = 0.128 ppm a 0.05 ppm, so 0* = 0.172 Conservatively use T' = 1 for T > 2000C Conservatively use T* = 1 for T > 200'C Thus:

Thus: c* = 0 for v, > 0.4%/sec so F,, = 2.55 so F., = 2.55 n* = ln(!0.4) for 0.0004 <= t, <= 0.4%/sec so F-, ranges from 2.55 so F,, ranges from 2.55 to 8.36 to 8.36

ln(0.0004/0.4) for t: < 0.0004%/sec so F,,

8.36 so F_ = 8.36 Thus, maximum F., = 8.36 Thus, maximum F., = 8.36 File No.: VY-16Q-303 - Revision: 0 Page 19 of 24 ~Veo'rropr, r r F0306-OIRO NEC066016

VStructural Integrity Associates, Inc. Table 7: EAF Evaluation for RPV Shell/Bottom Head Location Component: RPV Shell/Bottom Head NUREG/CR-6260 CUF: 0.032 (for reference only) Reterence: NUREG/CR-6260, p. 5-102 Stress Report CUF: 0.0057 (for Point 14, see below) Material: Low Alloy Steel (Material =A-533 Gr. Bper References [14] and[191) Design Basis CUF Calculation for 40 years: Elatigue curve/Eanalysis = Power Uprate = K = n= Sm 1.149 1.0067 1.000 2.0 0.2 26,700 Conservatively used minimum E of 26.1 from Section S2 Appendix of RPV Stress Report. =(549 - 100)1(546-100)per 44.4.b of 25A6019. Rev 1[14] stress concentration factor NB-3228.5 ofASME Code, Section fl/ [11] NB-3228.5 ofASME Code, Section h/ [11] psi (ASME Code. Section fl, Part D [11]) PL+Pe+Q (seeNote 1) Ke(see Note 2) Salt (seeNote3) n (seeNote4) N (see Note 5) U 44,526 1.00 25,762 200 35,300 0.0057 1 Total, U40 = 0.0057 Notes:

1. P, +P,.o is obtained for Point 14 from p. A52 of VYC-378, Rev. 0.

2. K, computed in accordance with NB-3228.5 of ASME Code. Section II1.

3. S,

= 0.5"K K E'K. 'K,

E,,_

Power Uprate "(P *P; +Q).

4. n for 40 years is the number of Heatup-Cooldown cycles, per p. B8 of VYC-378, Rev. 0.

5. N ob rained from Figure 1-9. 1 of Appendix I of ASME Code. Section 1if.

6. n for 60 years is the projected number of Heatup.Cooldovn cycles.

Revised CUF Calculation for 60 Years: PL+P B+Q (see Note 1) K. (see Note 2) Sa., (see Note3) n (see NoteS) N (see Note 4) U 44,526 1.00 25,762 300 35,300 0.0085 Total, U60 = 0.0085 Envronmental CUF Calculation for 60 Years: Maximum FeHWc Multiplier for HWC Conditions = 5.39 (from Table 6) Maximum Fen.NWC Multiplier for NWC Conditions = 13.17 (from Table 6) Uenv.60 = U60 x Fen-NWC X 0.53 + U6 0 X FenHWC X 0.47 = 0.0809 Overall Multiplier = Uenv. 60/U60 = 9.51 I I I I I I I I I I I I I I File No.: VY-16Q-303 Revision: 0 Page 20 of 24 ý n t n s e r r o061 yi n t i o n NEC066017 F0306-01 RO

Structural Integrity Associates, Inc. Table 8: EAF Evaluation for Limiting RPV Shell/Shroud Support Location Component: RPV Shell at Shroud Support NUREG/CR-6260 CUF: 0.032 (for reference only)

Reference:

NUREG!CR-6260, p. 5-102 Stress Report CUF: 0.0549 (for Point 9, see below' Material: Low Alloy Steel (Material =A-533 Gr B per References I(4] and (79]) Design Basis CUF Calculation for 40 years: Hydrotest c, = 26,240 psi (p. S3-97ofRPVStress Report) Hydrotest ns, = -1,250 psi (p. S3-97of RPVStress Report) Stress Concentration Factor, K, = 2.40 Hydrotest Kt 4 = 62,976 Improper Startup I,* = 28,060 Improper Startup a, = -1,025 Improper Startup Skin Stress = 156,099 Improper Startup K, 4, + Skin Stress = 223,443 Warmup c = -5,707 Warmup a= -102 Warmup K,, = -13.696 Elt aligu. _cuv oE cy si = 1.0417 Power Uprate = 1.0067 m= 2.0 n= 0.2 Sm= 26,700 (p. $3-99d of RPV Stress Report) psi (p. S3-97 of RPV Stress Report) psi (p. S3-98 of RPV Stress Report) psi (ip. S3-98 of RPV Stress Report) psi (p. S3-98 of RPV Stress Report) psi (p. $3-98 of RPVStress Report) psi (p. $3-99a of RPV Stress Report) psi (p. $3-99a of RPV Stress Report) psi (p. S3-99a of RPV Stress Report) 30.0 0/2.8 per S3-99f of RPV Stress Report and ASME Code fatigue curve =(549 - tOO) 1(546 - 100) per 4.4. 1.b of 26A6019, Rev. 1 t14] NB-3228.5 of ASME Code. Section /111111 NB-3228. 5 of ASME Code. Section ItfI t11) pSi (ASME Code, Section II, Part D [i]) PL+PF+Q (see Note t) Events Ke (see Note 2) Salt (see Note 3) n (see Note 4) N (see Note 5) U 34,690 Improper Startup - Warmup 1.00 124,825 5 332 0.0151 33,095 Hydrotest - Warmup 1.00 40,804 322 8,095 0.0398 1 Total, U4 0 = 0.0549 Notes:

1. P.+P +Ois computedfor Point 9basedon the[(g - e,),,_

-(. ]s,) ,.,,, ]ctress intensity.

2. K. computed in accordance 4th NS-3228.5 of ASME Code. Section InI.

3. S,: =0.5*

E*K.,

' *Poer Uprate!(K: n, -.[)K,,,,.n -(K rr -

4. n for 40 years is the number of cycles as folloas per p. S3 99e and S3-99f of the RPV Stress Report:

Improper Startup = 5 cycles Hydrotest = 2 cycles Isothermal at 700F and 1.000 psi = 120 cycles (same as number of Startup events) Warmup-Cooldovn = t99 cycles Warmup-Bloadodvtn = I cycle TOTAL = 327 cycles

5. N obtained from Figure 4-9.1 of Appendix I of ASME Code. Section IfI

9. n for 60 years is the projected number of cycles as follovs:

Improper Startup = I cycles Hydrotest = I cycles Isothermal at 70UF and 1,0O0 psi = 300 cycles (same as number of Startup events) Warenup-Cooldoee = 300 cycles Warmup-Blotvdom = I cycle TO TAL = 603 cycles Revised CUF Calculation for 60 Years: SPL+PB+Q (see Note 1) Ke (see Note 2) Sa, (see Note3) n (see Note 6) N (see Note 4) U 34,690 Improper Startup - Warmup 1.00 124,825 1 332 0.0030 33.095 Hydrotest - Warmup 1.00 40,804 602 8,095 0.0744 Total, U6 0 0.0774 Environmental CUF Calculation for 60 Years: Maximum FOnHWC Multiplier for HWC Conditions = 5.39 (from Table 6) Maximum FOn.NWc Multiplier for NWC Conditions = 13.17 (from Table 6) Uenv.&60 1 U6o x FEnijwc x 0.53 + U6 0 x FoeHWC X 0.47 = 0.7364 1 Overall Multiplier = UenV,. 0/U 60 = 9.51 File No.: VY-16Q-303 Revision: 0 Page 21 of 24 F0306-0I RO on nVe o Propr ar Infor sion NEC066018

Structural Integrity Associates, Inc. Table 9: EAF Evaluation for RR Inlet Nozzle Forging Location Component: Recirculation Inlet Nozzle Forging NUREG/CR-6260 CUF: 0.310 (for reference only)

Reference:

NUREG./CR-6260. p. 5-105 Stress Report CUF: 0.0433 (updated for Point 12, see belowl Material: Low Alloy Steel (Material =A-508 Cl II per p. l-S8-4 of CBIN Stress Report Section S8) Design Basis CUF Calculation for 40 years: Efatigue curve/Eanaly sis = 1.1278 Power Uprate = 1.0067 Kt= 1.660 m= 2.0 n= 0.2 S= 26,700 = 30.0 26.6 (per p 1-18-24 of CBON Stress Report Section S8 and ASME Code fatigue curve) =(549 - 100) 1(546 - 100) per 4.4. t.b ot26A6019, Rev. 1(t14 stress concentration factor (p. A270 of VYC-378, Rev. 0 [12]) NB-3228.5 ofASME Code. Section II111 tj NB-3228.5 of ASME Code. Section fli/ tt] psi (ASME Code, Section I1. Part 0[1 t[) PL+PB+Q (see Note t) Skin Stress (see Note 2) K. (see Note 3) Salt (seeNote4) n (see Note S) N (see Note 6) U 43;110 15,145 1.00 49,224 200 4,614 0.0433 1 Total, U4o = 0.0433 Notes: I. P, +P e+Q is obtained for Point 12 from p. A270 ol VYC-378, Rev. 0.

2. Skin Stress is obtained for Point t2 from p. A270 of VYC-378. Rev. 0.

3. K. computed in accordance with NB-3225.5 of ASME Code. Section II.

4. SAI: - 0.5 KE *E,,__

.,_ 'Power Uprate '[ (P..P, 0C) K, + Skin Stress ).

5. n for 40 years is the number of Heatup-Cooldown cycles, per p. B28 of VYC-378, Rev. 0.

6. N obtained from Figure I-9. t of Appendix I of ASME Code. Section II/.

7. n for 60 years is the projected number of Heatup-Cooldomn cycles.

Revised CUF Calculation for 60 Years: I PL+PB+Q (see Nate 1) Skin Stress (see Note 2) Ks (see Note 3) S.11 (see Note 4) n (see Note 7) N (see Note 6) U 43,110 15,145 1.00 49,224 300 4,614 0.0650 Total, Uw 0.0650 Envronmental CUF Calculation for 60 Years: Maximum FenHWC Multiplier for HWC Conditions = 2.45 (from Table 5) Maximum Fen. 5 Wc Multiplier for NWC Conditions 12.43 (from Table 5) Uenv.60 = U60 x Fen.aWC X 0.53 + U60 x Fn_-HWc x5 0.47 0.5034 Overall Multiplier = Ue_-6,iU/11 7.74 i I I I I I i I I I I I I I File No.: VY-16Q-303 Revision: 0 Page 22 of 24 NEC066019 F0306-OIRO

V Structural Integrity Associates, Inc. Table 10: EAF Evaluation for RR Inlet Nozzle Safe End Location Component: Recirculation Inlet Nozzle Safe End NUREG/CR-6260 CUF: 0.310 (for reference only)

Reference:

NUREG/CR-6260. p. 5-105 Stress Report CUF: 0.0017 (updated for Location 6-1, see below) Material: Stainless Steel (316-Lperp. 8of23A4292, Rev. 4) Desian Basis CUF Calculation for 40 years: E I figue ouse/Eanaiysis = 1.1076 =28.3/25,55 (per p. 62 of Reference[t8] and ASME Code fatigue curve) Power Uprate = 1.0067 =(549-100)1(546.- tO0)per 4.4.1.b of26A6019. Rev. 1114] K= 1.280 stress concentration factor (p. B27of VYC-378, Rev. 0[12]) M = 1.7 NB-3228.5 of ASME Code, Section III [t 1] n = 0.3 NB-3228.5 of ASME Code, Section /1 [11] S= 16,600 psi (ASME Code. Section ft. Part D [11]) Ke (see Note 3) Sal1 (see Note 4) n (see Note 5) N (see Note 6) U 1.00 26,385 2,076 1,242,266 0.0017 1 Total, U4. = 0.0017 PL+ PB+O (see Note 1) 47,183 P+Q+F (see Note 2) 36,972 Notes:

1. P, +Po+0 is obtained for Surface I (after weld overlay) tromp. 117of Reference [18).

2. P+Q+F is obtained for Point 6-1 from p. 118 of Reference [18) (BEFORE weld overlay).

3. K, computed in accordance with NB-3228.5 of ASME Code. Section Ill.

4. Sa, = 0.5 -

/EK -E,,_

Power Lprate '[ (P+Q+F) K, ].

5. n for 40 years is the number of cycles as follows per p. B26 of VYC-378, Rev. 0:

Design Hydrotest = 130 Loss of Feedpumps Composite: Startup/Shutdown = 290 SRV Blowdown = 8 Loss of Feedwater Pumps 30 10 events x 3 up/down cycles per event SCRAM= 270 Normal +/- Seismic = I1 10 cycles of upset seismic, plus I Level C seismic event Normal = 739

Sum of all of above events Zeroload

598

Startup/Shutdown, SRVBIowdotw + Scram + LOFP Total number of cycles

2,070

6. N obtained from Figure 1-9.2 of Appendix I of ASME Code. Section Itt.

7 n for 60 years is the projected number of cycles as follows: Design Hydrotest 120 Loss of Feedpumps Composite* StartuplShutdown = 300 SRV Blowdown = 1 Loss of Eeedwater Pumps 30 10 events x 3 up/down cycles per event SCRAM = 289 All remaining scrams Normal /- Seismic = It Assume the same Normal = 751

Sum of all of above events Zeroload

620

Startup/Shutdown + SRV Blovdown + Scram v LOFP Total number of cycles

2, 122 Remised CUF Calculation for 60 Years: PL+ PB+Q (see Note 1) P+Q+F (see Note 2) Ke (see Note 3) Salt (see Note 4) n (see Note.5) N (see Note 7) U 47,183 36,972 1.00 26,385 2,122 1,242,266 0.0017 Total, U6 0 = 0.0017 Environmental CUF Calculation for 60 Years: -- Maximum Fe,.HWc Multiplier for HWC Conditions = Maximum Fen.NWc Multiplier for NWC Conditions = Uenv.6O = U6 o x Fen.NWC X 0.53 + U6 0 x Fen.HWc X 0.47 = Overall Multiplier = Uenv_60/U 6 = 15.35 8.36 0.0199 11.64 (from Table 2) (from Table 2) File No.: VY-16Q-303 Revision: 0 Page 23 of 24 NEC066020 F0306-01 RO A

V Structural Integrity Associates, Inc. Table 11: Summary of EAF Evaluation Results for VY 40-Year 60-Year Overall 60-Year No. Component Material Design CUF (') CUF (2) Environmental Environmental Multiplier CUF (2,3) 1 RPV Shell/Bottom Head Low Alloy Steel 0.0057 0.0085 9.51 0.0809 2 RPV Shell at Shroud Support Low Alloy Steel 0.0549 0.0774 9.51 0.7364 3 Recirculation Inlet Nozzle Safe End Stainless Steel 0.0017 00017 11.64 0.0199 4 Recirculation Inlet Nozzle Forging Low Alloy Steel 0.0433 0.0650 7.74 0.5034 Notes:

1. Updated 40-year CUF calculation based on recent ASME Code methodology and design basis cycles.

2. CUF results using updated ASME Code methodology and actual cycles accumulated to-date and projected to 60 years.

3. An Fe, multiplier was used for each respective component with the following conditions:

+ 47% HWC conditions and 53% NWC conditions i I I I I I I I I I I I I I I I I I I .3 File No.: VY-16Q-303 Revision: 0 Page 24 of 24 Cq ns V Pr i a to F0306-01 RO NEC066021

V Structural Integrity Associates, Inc. APPENDIX A VY WATER CHEMISTRY INFORMATION [8] File No.: VY-16Q-303 Revision: 0 Page A I of A2 on" e or Prrh r o on, NEC066022 F0306-OIRO

V Structural Integrity Associates, Inc. Pre-NMCA Post-NMCA + HWC Post-NMCA + HWC Future Operation 1593 MWth (OLP) 1593 MWth (OLP) 1912 MWth (EPU) Post-NMCA + HWC 1912 MWth (EPU) Location Average Average Average Availability 98.5% Availability 98.5% Availability 99% Implementation Date NMCA Application EPU Implementation = 11/1972 Date = 04/27/2001 Date = 5/2006 HWC Implementation Date = 1 1/01/2003 FW Line 40 ppb 40 ppb 40 ppb 40 ppb Recirc. Line 122 ppb 48 ppb 34 ppb 34 ppb RPV Bottom 128 ppb 69 ppb 55 ppb 55 ppb Head ** RPV Upper 114 ppb 97 ppb 90 ppb 90 ppb Region RPV Beltline 123 ppb 46ppb 31 ppb 31 ppb Region

- RPV Bottom head at "Lower Plenum, Downflow" (i.e. outside core support columns)

File No.: VY-16Q-303 Revision: 0 Page A2 of A2 or o ri et Inf In a n NEC066023 F0306-OIRO

? Structural Integrity Associates, Inc. APPENDIX B VY LICENSE DATE [101 File No.: VY-16Q-303 Revision: 0 Page B I of B2 NEC066024 F0306-01RO

2 ýStructural Integrity Associates, Inc. " ern~cn: Yarkee NuL'caar Power Szaion cicense Renewal Aop~cx'ion Michael A. Baiduzzi Vice Preside.nt - Pilgdrin NLIclear F-c.wer Station Fred R. Dacimo Vice Presidentl - Indian Point oEnerg'y Center Randall K. £dington %/i coe President-Operations Support Ch ristoph er J. S cbwarz Vice Presideni - Cperations Su'ppor, Th-eodore A. SuLliv_'an Vice Presideak - Fitzpatrick Nuclear Power Station Jay K. Thaver Vice President - Vermont Yankee Nuclear Power Station Pilainm iNuclear Power Statiin 600 Rocky Hill Roard Plymouth, Massachusetts 02360 Indian Point Energy Centeir Bleakiey Avenue & Sroadway Buchanaon, New 'York 1' 11 Cooper Nuclear Power 3tation 1200 Prospect Road P.O. BoX H8 Brownsville, Nebrasma 683b21 Entergrv Nuciear Operations. Inc 440 Hamilto't Avenue White Pbains, New York 1001' Fitzpatrick Nuclear Power Sation 268 Lake Road nast Lycoming, Nevw York 13093 Entergy Nuclear ",V"ernmon1t Yankee Cuorporate 0irce P.o. Box 050C) 185 Cld Ferry Road Brattleb..n, 'r, TT 05302-0500 I I I I I I I I I U I I I U I I I 1.1.5 Class and Period of License SoLuuht EN:4 r9'ests rne4e.-al of tne racility operating license **r VYNRS (facility operating license DPR-z.,:,or , eric,d or 20 years. The license s,','as issued under Section i041: of the At.omic Eneroy Act of 1954 js amended. License renew'al would extend the facility operating license fron mid night tia-rth 2 201: to m idnight 2ch*2k2.32. This applic --,'on also applies to renewal of those NRC source materials, special nuclear material, and by-product material licenses that are subsumed or contined with the facility operating j icense. 1.1.6 Alteration Schedule E0 does not propose to construct or alter any production or Utilization facility in connection 'ith this renewal application. .1.Admrninstun'.vE craln FagEw 1-4, File No.: VY-16Q-303 Revision: 0 Page B2 of B2 ~..nao otinsV or F0306-01 RO NEC066025

NEC-JH_18 Report No.: SIR-07-132-NPS Revision No.: 1 Project No.: VY-16Q File No.: VY-16Q-404 December 2007 Summary Report of Plant-Specific Environmental Fatigue Analyses for the Vermont Yankee Nuclear Power Station Prepared for: Entergy Nuclear Operations. Inc. (Contract Order No. 10150394) Prepared by.- Structural Integrity Associates, Inc. Centennial, CO Prepared by: Reviewed by: Approved by: T. J. 4 nann P.E. L. Stevens, P.E. T. J. *r' ann P.E. Date: 12/15/2007 Date: 12/15/2007 Date: 12/15/2007

REVISION CONTROL SHEET Document Number: SIR-07-132-NPS

Title:

Summary Report of Plant-Specific Environmental Fatigue Analyses for the Vermont Yankee Nuclear Power Station Client: Entergy Nuclear Operations, Inc. SI Project Number: VY-16Q Section Pages Revision Date Comments 1.0 1-1 0 7/27/07 Initial issue. 2.0 2-1-2-2 3.0 3 3-18 4.0 4-1 5.0 5-1 2 1.0 1-1 1 12/15/07 Revised based on revision to VY-16Q-2.0 2-1-2-2 309 and VY-16Q-310 associated with 3.0 3-1 18 CAR 07-25 and NCR 07-11. Editorial 4.0 4-1 correction on page 3-5. ,5.0 5-1 2

Table of Contents Section Page

1.0 INTRODUCTION

1-1

2.0 BACKGROUND

2-1 3.0 ENVIRONM ENTAL FATIGUE CALCULATIONS..................................................... 3-1 3.1 Reactor Vessel Shell and Lower Head......................................................................... 3-3 3.2 Reactor Vessel Feedwater Nozzle................................................................................ 3-4 3.3 Reactor Recirculation Piping (Including the Reactor Inlet and Outlet Nozzles)......... 3-5 3.3.1 Reactor Recirculation Piping................................................................................ 3-5 3.3.2 Reactor Recirculation Inlet Nozzle...................................................................... 3-6 3.3.3 Reactor Recirculation Outlet Nozzle.................................................................... 3-7 3.4 Core Spray Line Reactor Vessel Nozzle and Associated Class I Piping..................... 3-7 3.5 RHR Return Line Class 1 Piping................................................................................. 3-8 3.6 Feedwater Line Class I Piping..................................................................................... 3-8 3.7 Summary of Results..................................................................................................... 3-8 4.0 SUM M ARY AND CONCLUSIONS................................................................................ 4-1

5.0 REFERENCES

5-1 SIR-07-132-NPS, Rev. I °°. V Structural Integrity Associates, Inc.

LIST OF TABLES Table Title Page Table 3-1. Environmental Fatigue Evaluation for the Reactor Vessel Shell............................... 3-9 Table 3-2. Environmental Fatigue Evaluation for the Reactor Vessel Shell at................................ S hroud S upport........................................................................................................................... 3-10 Table 3-3. Environmental Fatigue Evaluation for the Reactor Vessel Feedwater Nozzle............... Forging B lend R adius................................................................................................................ 3-11 Table 3-4. Environmental Fatigue Evaluation for the Recirculation/RHR Piping Tee............. 3-12 Table 3-5. Environmental Fatigue Evaluation for the Reactor Recirculation Inlet...................... N ozzle F org ing........................................................................................................................... 3-13 Table 3-6. Environmental Fatigue Evaluation for Reactor Recirculation Inlet Nozzle................... S afe nE n d.............................................................................. 3-14 Table 3-7. Environmental Fatigue Evaluation for Recirculation Outlet Nozzle Forging......... 3-15 Table 3-8. Environmental Fatigue Evaluation for Core Spray Reactor Vessel................... Nozzle Forging Blend Radius, Safe End, and Piping................................................................. 3-16 Table 3-9. Environmental Fatigue Evaluation for the Feedwater Line Class I Piping............. 3-17 Table 3-10. Summary of Environmental Fatigue Calculations for VYNPS............................. 3-18 I I I I I I I I I U I I I I I I I I I SIR-07-132-NPS, Rev. 1 iv V Structural Integrity Associates, Inc.

1.0 INTRODUCTION

This report provides the results of plant-specific environmental fatigue calculations for the Vermont Yankee Nuclear Power Station (VYNPS). These calculations are performed to satisfy Nuclear Regulatory Commission (NRC) requirements for Entergy Nuclear Vermont Yankee's (ENVY's) License Renewal Application for VYNPS, submitted to the NRC in 2006. Generic Safety Issue (GSI) 166 [1], later renumbered as GSI-190 [2], was identified by the NRC staff because of concerns about the effects of reactor water environments on fatigue life during the period of extended operation [3]. GSI-190 was closed in December 1999, based on a memorandum from NRC-RES to NRC-NRR [4]. Timing of issue closure required the first two license renewal applicants - Baltimore Gas & Electric Company for the Calvert Cliffs Nuclear Power Plant and Duke Energy for the Oconee Nuclear Station - to address GSI-190 in their applications prior to issue closure. Each of the applicants developed responses to the NRC staff without the benefit of information from GSI-190 closure. Subsequent license renewal applicants have had the benefit of this information that could be used to guide the resolution of the fatigue design basis and time limited aging analyses (TLAA) issues. This report addresses VYNPS reactor water environmental effects on the fatigue life of selected fatigue-sensitive reactor coolant system (RCS) components, in accordance with the resolution of GSI-190, as required by Chapter X, "Time Limited Aging Analyses Evaluation of Aging Management Programs Under 10CFR54.21(c)(1)(iii), Section X.MI "Metal Fatigue of Reactor Coolant Pressure Boundary", of the Generic Aging Lessons Learned (GALL) Report [5]. Consistent with the requirements of the GALL report, the method chosen for this environmentally-assisted fatigue (EAF) evaluation is based on evaluation of the locations identified in NUREG/CR-6260 [6] and the NRC-accepted EAF relationships generated from laboratory data, as documented in References [7] and [8]. SIR-07-132-NPS, Rev. 1 11-l Structural Integrity Associates, Inc.

2.0 BACKGROUND

As a part of the NRC's Fatigue Action Plan [3], incorporation of environmental fatigue effects 3 originally involved a reduced set of fatigue design curves, such as thoseproposed by Argonne National Laboratory (ANL) in NUREG/CR-5999 [9]. As a part of the effort to close GSI-166 (later GSI-190) for operating nuclear power plants during the current 40-year licensing term, Idaho National Engineering Laboratory (INEL) evaluated fatigue-sensitive component locations at plants designed by all four U. S. nuclear steam supply system (NSSS) vendors. The ANL fatigue curves were used by INEL to recalculate the cumulative usage factors (CUFs) for fatigue-sensitive component locations in early and late vintage Combustion Engineering (CE) pressurized water reactors (PWRs), early and late vintage Westinghouse PWRs, early and late vintage General Electric (GE) boiling water reactors (BWRs), and Babcock & Wilcox Company I (B&W) PWRs. The results of the INEL calculations were published in NUREG/CR-6260 [6]. The INEL calculations took advantage of conservatisms present in governing ASME Code I fatigue calculations, including the numbers of actual plant transients relative to the numbers of design-basis transients, but did not recalculate stress ranges based on actual plant transient profiles. The BWR calculations, especially the early-vintage GE BWR calculations, are directly relevant to VYNPS. The fatigue-sensitive component locations chosen for the older-vintage GE BWR plant were: (1) I the reactor vessel shell and lower head, (2) the reactor vessel feedwater nozzle, (3) the reactor recirculation piping (including the reactor inlet and outlet nozzles), (4) the core spray line reactor vessel nozzle and associated Class 1 piping, (5) the residual heat removal (RHR) return line Class 1 piping, and (6) the feedwater line Class 1 piping. For the recirculation, RHR, and I feedwater piping locations, INEL performed representative design-basis fatigue calculations. This is because no CUF calculations had originally been performed since the piping systems for U the selected BWR plant were initially designed and analyzed in accordance with the criteria of USAS B31.1-1967 [10]. I SIR 132-NPS, Rev. 1 2-1l Structural Integrity Associates, Inc. I

The six RCS component locations described above are evaluated for EAF effects for VYNPS in this report through separate plant-specific analyses of nine VY component locations (with report section numbers indicated): the reactor pressure vessel (RPV) shell and lower head (3.1); the RPV shell at the shroud support junction (3.1); the feedwater nozzle (3.2); the recirculation / residual heat removal Class 1 piping (3.3.1 and 3.5); the recirculation inlet nozzle forging (3.3.2); the recirculation inlet nozzle safe end (3.3.2); the recirculation outlet nozzle forging (3.3.3); the core spray nozzle, safe end, and Class 1 piping (3.4); and the feedwater Class 1 piping (3.6). The calculations reported in NUREG/CR-6260 were based on the interim reduced fatigue design curves given in NUREG/CR-5999 [9]. Such an approach penalizes the component location fatigue analysis unnecessarily, because research has shown that a combination of environmental conditions is required before reactor water environmental effects become pronounced. The strain rate must be sufficiently low and the strain range must be sufficiently high to cause continuing rupture of the passivation layer that protects the exposed surface area. Temperature, dissolved oxygen content, metal sulfur content, and water flow rate are additional variables to be considered. In order to take these parameters into consideration, EPRI and GE jointly developed a method, called the Ft, approach [11], which permits reactor water environmental effects to be applied selectively, as justified by parameter combinations. In 1999, the NRC staff raised a number of issues relative to the use of the EPRI/GE methodology in various industry applications. Those issues, coupled with more recent laboratory fatigue data in simulated LWR reactor water environments generated by ANL for carbon and low-alloy steels and stainless steels, resulted in a revised Fe, methodology, as published in NUREG/CR-6583 [7] for carbon and low alloy steels, and NUREG/CR-5704 [8] for stainless steels. The methodology documented in these reports was used to evaluate environmental effects for VYNPS components, as described in Section 3.0 of this report. SIR-07-132-NPS, Rev. 1 2-2 klý struciurai iegrity tissocidies, mc.

3.0 ENVIRONMENTAL FATIGUE CALCULATIONS Section 2.0 identifies the locations evaluated in NUREG/CR-6260 for the older vintage GE plant, which corresponds to VYNPS. NUREG/CR-6260 provided an assessment of these six selected component locations with respect to environmental fatigue using the older reduced u environmental fatigue curves. Potential reactor water environmental effects are evaluated using the updated Fen methodology on a plant-specific basis in this subsection, in order to address the associated effects on fatigue as required by the GALL Report [5]. For each of the components identified in Section 2.0, environmental fatigue calculations were performed. The details of these calculations are documented in the Reference [12, 17, 18, 21, 22 and 24] calculations. The calculations were carried out using the appropriate methodology I contained in NUREG/CR-6583 for carbon/low alloy steel material, and in NUREG/CR-5704 for stainless steel material. This methodology is as follows: I For Carbon Steel [7]: Fen = exp (0.585 - 0.00124T' - 0.101 S* T* O* *) = exp (0.554 - 0.101 S* T* 0* c*) For Low Alloy Steel [7]: Fen = exp (0.929 - 0.00124T' - 0.101 S* T* 0* ý = exp (0.898 - 0.101 S* T* 0* ý *) Note that the above expressions have been corrected as summarized in Reference [23]. where: Fen fatigue life correction factor T' = 25-C (NUREG/CR-6583, Section 6, Fen relative toair) S* = S for 0 < sulfur content, S < 0.015 wt. % = 0.015 for S > 0.015 wt. % T* = 0 for T < 1500 C (T-150) for 150<*T< 350'C T = fluid service temperature (°C) 0* = 0 for dissolved oxygen, DO < 0.05 parts per million (ppm) ln(DO/0.04) for 0.05 ppm < DO < 0.5 ppm ln(12.5) for DO > 0.5 ppm I SIR 132-NPS, Rev. l 3-1i Structural Integrity Associates, Inc. 3 I

= 0 for strain rate, * > 1%/sec I ln( )for 0.001

1%/sec In(0.001) for ý < 0.001%/sec For Types 304 and 316 Stainless Steel [8]: Fen = exp (0.935 - T* + *O*)

where: Fen = fatigue life correction factor T = fluid service temperature ('C) T* = 0 for T < 2000C = I forT> 200'C E* = 0 for strain rate, r > 0.4%/sec = ln(k/0.4) for 0.0004

k
0.4%/sec

= ln(0.0004/0.4) for i < 0.0004%/sec 0* = 0.260 for dissolved oxygen, DO < 0.05 parts per million (ppm) = 0.172 for DO >_ 0.05 ppm Bounding Fen values are determined or, where necessary, computed for each load pair in a detailed fatigue calculation. The environmental fatigue is then determined as Uenv = (U) (Fen), where U is the original fatigue usage, and Uenv is the EAF usage factor. INFORMATION REDACTED Since implementation of HWC in 2003, VYNPS's availability has exceeded 98.5% and the objective for future HWC system availability is a minimum of 99% [12]. With these considerations, the overall availability for HWC since implementation at VYNPS until the end of the 60-year operating period was estimated at 98.5%. SIR-07-132-NPS, Rev. 1 3-2 Structural Integrity Associates, Inc. This Page Contains Reference to Vendor Proprietary Information (such information is marked with a "bar" in the right-hand margin)

I Some nozzles, (e.g., recirculation outlet nozzle) have three materials: a Ni-Cr-Fe dissimilar metal weld (DMW), a low alloy steel forging, and a stainless steel safe end. To ensure the maximum CUF considering environmental effects was identified, locations in both the safe end and nozzle forging were selected. This selection produces bounding environmental fatigue results for the entire nozzle assembly for the following reasons:

The highest thermal stresses from the finite-element model (FEM) analysis occur in the stainless steel safe end. Stainless steel Fen multipliers at VYNPS are significantly higher than Ni-Cr-Fe multipliers (Fen values are 2.55 or higher for stainless steel [12] vs. a constant value of 1.49 for Ni-Cr-Fe [11]). Therefore, evaluation of the safe end bounds the Ni-Cr-Fe weld material.
The highest pressure stresses from the FEM analysis occur in the low alloy steel nozzle forging. Low alloy steel Fen multipliers at VYNPS are higher than Ni-Cr-Fe multipliers (Fen values are 2.45 or higher for low alloy steel [12] vs. a constant value of 1.49 for Ni-Cr-Fe [11]). Therefore, evaluation of the nozzle forging bounds the Ni-Cr-Fe weld material.

The number of cycles for forty years was adjusted based on the number of cycles actually U experienced by the plant, projected out to 60 years of operation [14]. In addition, VYNPS has implemented extended power uprate (EPU). These effects have been incorporated into the evaluations documented in this report. With the use of this information, the CUF values .documented in this report are applicable for 60 years of operation. The environmental fatigue calculations are shown in Tables 3-1 through 3-9 and summarized in I Table 3-10. Component-specific details are provided in the subsections that follow. 3.1 Reactor Vessel Shell and Lower Head I The environmental fatigue calculations for the reactor vessel shell and lower head location are shown in Table 3-1. The limiting CUF value reported in the VY LRA for the RPV shell/bottom SIR-07-132-NPS, Rev. 1 3-3 Structural Integrity Associates, Inc. U

head location corresponds to a point located on the outside surface of the RPV bottom head at the junction with the support skirt. Therefore, this location is not exposed to the reactor coolant, and EAF effects do not apply. Based on this, evaluation of the limiting location along the inside surface of the RPV bottom head was performed. The calculations shown in Table 3-1 are for the RPV lower head at the area with the highest alternating stress, which represents the limiting RPV bottom head location [12]. Reference [15] is the governing stress report for this low alloy steel location. The design fatigue calculation for the limiting RPV lower head location is reproduced in Table 3-1. The effects of EPU as well as conservative cycle counts for 60 years of plant operation are incorporated in this table. The final results in Table 3-1 show an EAF adjusted CUF of 0.0809 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). The calculations shown in Table 3-2 are for the RPV shell at the RPV shell junction to the shroud support plate, which represents the limiting RPV shell location exposed to the reactor coolant [12]. Reference [16] is the governing stress report for this low alloy steel location. The design fatigue calculation for the limiting RPV shell location is reproduced in Table 3-2, which considers the effects of EPU and conservative cycle counts were used for 60 years of plant operation. The final results in Table 3-2 show an EAF adjusted CUF of 0.7364 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). 3.2 Reactor Vessel Feedwater Nozzle The environmental fatigue calculations for the reactor vessel feedwater nozzle location are summarized in Table 3-3. The calculations summarized in Table 3-3 show both the blend radius, which represents the limiting feedwater nozzle location, and the safe end. Reference [17] contains the governing fatigue calculation for this location. Upper RPV region chemistry was assumed for the feedwater nozzle blend radius location, since this location is exposed to the reactor water chemistry in this region, whereas feedwater line chemistry was assumed for the safe end location. SIR-07-132-NPS, Rev. 1 3-4 .UaMrLM3u1i rnwgrOrly A-SSOuia~e.5, InC.

I The governing fatigue calculation for the limiting feedwater nozzle locations includes the effects of EPU and cycle counts for 60 years of operation obtained from Attachment 1 of Reference [14]. The blend radius cumulative usage factor (CUF) from system cycling is 0.0636 for 60 years. The safe end CUF is 0.1471 for 60 years. Although the carbon steel safe end has a higher CUF prior to considering environmental effects, the environmental multiplier from Table 3-3 results in a higher CUF at the low alloy steel blend radius. For the safe end location, the EAF adjusted CUF is 0.2560 for 60 years. For the blend radius location, EAF adjusted CUF is 0.6392 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). 3 3.3 Reactor Recirculation Piping (Including the Reactor Inlet and Outlet Nozzles) Three locations were identified for the reactor recirculation piping in NUREG/CR-6260: the reactor vessel nozzle (includes both the inlet and outlet nozzles), and the recirculation piping. I The evaluations for each of these components are described in the following subsections. I 3.3.1 Reactor Recirculation Piping I Two locations (both stainless steel) were identified for VY for the reactor recirculation/RJHR piping that correspond to the equivalent locations to those identified in NUREG/CR-6260: the 3 RHR return tee connection to the recirculation piping, and the valve to pipe weld at the RHR isolation valve. Reference [18] contains the governing fatigue calculations for these locations. These analyses determined the limiting location to be at the RHR return tee. The environmental fatigue calculations for the limiting recirculation/RHR piping~location is i summarized in Table 3-4, which includes the effects of EPU and cycle counts for 60 years of plant operation. A review of the shutdown cooling mode of operation since the time of recirculation piping I replacement in 1986 was performed by VYNPS, and the number of cycles per loop was conservatively estimated to be 150 through Year 60 [14]. Based on this, the cycle counts for the SIR-07-132-NPS, Rev. 1 3-5 6 Structural Integrity Associates, Inc. I

Recirculation piping were reduced by a factor of 150/300 (50%) for all transients with the exception of transients that have fewer than 10 transient cycles. To ensure this cycle reduction adequately considered the potential impact on the RHR piping, the full number of transient cycles listed in Attachment I of Reference [14] was initially applied to the PIPESTRESS model and the highest CUF for the RHR piping was lower than the value obtained for the recirculation piping with reduced cycles. Due to replacement of the recirculation piping, HWC conditions exist for 39% of the time, and NWC conditions exist for 61% of the time. This is based on 17.5 years of operation with NWC between March 1986 when the piping was replaced and November 2003 when HWC was implemented, and 46 years of operation from March 1986 to the end of the period of extended operation in March 2032. Using the bounding EAF multipliers (8.36 for HWC and 15.35 for NWC) [12], the overall multiplier is 12.62. Applying this to the 60-Year CUF of 0.0590 results in a total environmentally assisted CUF of 0.7446. 3.3.2 Reactor Recirculation Inlet Nozzle References [15, 19 and 20] are the applicable stress reports for this location. An evaluation was performed for both the inlet nozzle forging (low alloy steel) and the safe end (stainless steel). The environmental fatigue calculations for the recirculation inlet nozzle forging location are shown in Table 3-5. The governing fatigue calculation for the recirculation inlet nozzle location is reproduced in Table 3-5 [12], which includes the effects of EPU and cycle counts for 60 years of plant operation from Attachment I of Reference [14]. The final results show an EAF adjusted CUF of 0.5034 for 60 years, which is acceptable (i.e.,, less than the allowable value of 1.0). The environmental fatigue calculations for the recirculation inlet nozzle safe end location are shown in Table 3-6. The governing fatigue calculation for the recirculation inlet nozzle location is reproduced in Table 3-6 [12], which includes the effects of EPU and cycle counts for 60 years SIR-07-132-NPS, Rev. 1 3-6 O.v lrYIUriI III WYfIly AISSV~dUJSL, 111W

of plant operation from Attachment I of Reference [14]. The final results show an EAF adjusted CUF of 0.0199 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). 3.3.3 Reactor Recirculation Outlet Nozzle 1 The recirculation outlet nozzle was evaluated for environmental fatigue effects. Reference [24] I is the fatigue calculation for this location. An evaluation was performed for both the outlet nozzle safe end (stainless steel) and the nozzle inner comer blend radius (low alloy steel). The n results for the limiting nozzle forging location are reported here. The environmental fatigue calculations for the limiting recirculation outlet nozzle forging blend radius location are shown in Table 3-7 [24], which includes the effects of EPU and cycle counts I for 60 years of plant operation from Attachment I of Reference [14]. The final results in Table 3-7 show an EAF adjusted CUF of 0.0836 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). 3.4 Core Spray Line Reactor Vessel Nozzle and Associated Class 1 Piping Locations that were evaluated in NUREG/CR-6260 included the reactor vessel nozzle blend radius (low alloy steel), the reactor vessel nozzle safe end (Alloy 600) and the core spray piping (stainless steel). Reference [21] is the applicable fatigue calculation for these locations, which shows the nozzle D limiting location to be the blend radius. The design fatigue calculations for the limiting location at the core spray nozzle, safe end, and piping are summarized in Table 3-8 [21], which include the effects of EPU and cycle counts for 60 years of plant operation from Attachment I of Reference [14]. The cumulative fatigue usage, prior to considering environmental effects for the blend radius, is 0.0166. Factoring in the environmental multiplier from Table 3-8 [12], the EAF adjusted CUF is 0.1668 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). SIR-07-132-NPS, Rev. 1 37 Structural Integrity Associates, Inc. I

3.5 RHR Return Line Class 1 Piping The environmental fatigue calculations for the RHR return line Class 1 piping are covered by the calculations in Subsection 3.3.1 above. 3.6 Feedwater Line Class 1 Piping The environmental fatigue calculation for the limiting feedwater Class I piping location (carbon steel) is summarized in Table 3-9. The calculations shown in Table 3-9 are for the limiting feedwater Class 1 piping location. Per Reference [22], the limiting total fatigue usage for the analyzed feedwater/high pressure coolant injection (HPCI) piping system occurs on the riser to the RPV feedwater nozzle N4B. The limiting fatigue usage value for the feedwater Class I piping location is 0.166 1, which includes the effects of EPU and cycle counts for 60 years of plant operation from Attachment 1 of Reference [14]. The final results in Table 3-9 show the EAF adjusted CUF of 0.2890 for 60 years, which is acceptable (i.e., less than the allowable value of 1.0). 3.7 Summary of Results The results of the calculations contained in Tables 3-1 through 3-9 are summarized in Table 3-10. It is noteworthy that the CUF results presented in this section include uniformly applied environmental effects without consideration of threshold criteria that might indicate an absence of conditions that would lead to environmental fatigue effects. Furthermore, conservative values were applied for temperature, strain rate and metal sulfur content in calculating environmental multipliers. Therefore, the environmental adjustments to the CUF results are considered to be conservative. SIR-07-132-NPS, Rev. 1 3-8 Structural Integrity Associates, Inc.

Table 3-1. Environmental Fatigue Evaluation for the Reactor Vessel Shell Component: RPV Shell/Bottom Head NUREG/CR-6260 CUF: 0.032 (or reference oly)

Reference:

NUREG/CR-6260, p. 5-102 Stress Report CUF: 0.0057 (for Point 14, see below) Material: Low Alloy Steel (Material = A-533 Gr. 8) Design Basis CUF Calculation for 40 years: Eots c-E,,yi*, = Power Uprate = K = mI= n= Sn, 1.149 1.0067 1.000 2.0 0.2 26.700 Conseevatively used minimum E of 26.1 from Section S2Appendix ofRPV Stress Report. =(549 - 100)1(546. 100) per 4.4.1.b of 26A6019, Rev. I stress concentration factor NB -3228.5 of ASME Code, Section III NS-3228.5 of ASML Code Section III psi (ASME Code, Section 11, Part D) PL +PB+Q (see Note 1) K. (see Note 2) S81 (see Note 3) n (see Note 4) N (see Note 5) U 44,526 1.00 25,762 200 35,300 0.0057 1 Total, U4o 0.0057 Notes:

t. P, +P8 0 Ois obtained for Point 14 from p. A52 of VYC-378, Rev. 0.

2. Kt, computed in accordance with NB-3228.5 ofASME Code, Section Ill.

3. So,,

0=,0.5

X. 'K
E,*.,/E1..e Poa'erUprate *(P, +Po +O).

4. n for 40 years is the number of Hegtup-Cooldows cycles, per p. 88 of VYC-378, Rev. O.

5. N obtainedfmrom Figure 1-9.1 of Appendix I of ASME Code, Section III,

6. n for 60 years is the projected number of Heatup-Coofdows cycles.

Revised CUF Calculation for 60 Years: P1 +PB-Q (see Note 1) K, (see Note 2) S,, (see Note 3) n (see Note 6) N (see Note 4) U 44,526 1.00 25,762 300 35,300 0.0085 Total, Ue = 0.0085 Environmental CUF Calculation for 60 Years: Maximum F-.-nC Multiplier for HWC Conditions = 5.39 Maximum Fn.NWC Multiplier for NWC Conditions = 13.17 U* nwo, = Uso X F en..wc x 0.53 + Uoo x F en. HWC x 0.47 = 0.0809 Overall Multiplier = U0oý,0 /U0 o = 9.51 I I I i I I I I I I I I I I SIR-07-132-NPS, Rev. I 3-9 4 Structural Integrity Associates, Inc. I I

Table 3-2. Environmental Fatigue Evaluation for the Reactor Vessel Shell at Shroud Support Component: RPV Shell at Shroud Support NUREG/CR-6260 CUF: 0.032 (for reference only)

Reference:

NUREG/CR-6260, p. 5-102 Stress Report CUF: 0.0549 (for Point 9, see below) Material: Low Alloy Steel (Material = A-533 Or B) Design Basis CUP Calculation for 40 years: Hydrotest HI = Hydrotest 4{ = Stress Concentration Factor, K, = Hydrotest K,116 = Improper Startup KI = Improper Startup l, = Improper Startup Skin Stress = Improper Startup KI 11t4 + Skin Stress Warmup it = Warmup HK = Warmup KrI,, = E~at 00e carve/ Eanaivos = Power Uprate = M= n= S, = 26,240 -1,250, 2.40 62,976 28,060 -1,025 156,099 223,443 -5,707 -102 -13,696 1.0417 1.0067 2.0 0.2 26,700 psi (p. S3-47 of RPV Stress Report) psi (p. S3.97 of RPV Staess Report/ (p. S3-e9d of RPV Stress Report) psi (p. S3-7 of RPV Stress Report) psi (p. S3-86 of RPV Stress Report) psi (p. S3-98 of RPV Stress Report) psi (p. S3-8 of RPV Stress Report) psi (p. S3-98 of RPV Stress Report) psi (p. $3-ea of RPtV Stress Report) psi (p. $3-29a of RPV Stress Report) psi po $3-99a of RPV Stress Report) 30,0 / 2 8.8 per S3-90f of RPV Stress Report and ASME Code fatigue curve =(549-100)/(546-tO0) per 4.4. t.b of 26A6019, Rev. I NB-3228.5 of ASME Code, Section H/ NB-32285 of ASME Code, Section III psi (ASME Code, Section If, Part 0) PL+PB+Q (see Note 1) Events K. (se0 Note 2) S,, (see Note 3) n (see Note 4) N (see Note 5) U 34,690 improper Startup - Warmup 1.00 124,825 5 332 0.0151 33,095 Hydrotest - Warmup 1.00 40,804 322 8,095 0.0398 1 Total, Uo,, 0.0549 Notes: I. PL rPs r0 is computed for Point 9 based on the J(tf- - (HH - h).E I stress intensity.

2. K. computae in accordance with NB-3228 5 of ASME Code. Section hif 3 Sa, = 0,5

KE a'

c E.rp,

Power Uprate *I(KtH, - H,)o -

- (K, t, - H,) 0*01.

4., for 40 years is the number of cycles as follows per p, 3-.92 and S3-9gf of the RPV Stress Report:

Improper Startup = 5 cycles Hydrotest = 2 cycles Isothermal at 70°F and 1,000 psi = 120 cycles (same as number of Startup events) Warwup-Coouldow = 199 yres0 Warmup-Blowdow I cycle TOTAL = 327 cycles 5 N obtained from Fgure -9.r of Appandi, t ofASMS Code. Section Ill.

6. n for h0 years is the projected number of cycles as follows Improper Startup =

I cycles Hydrotest t cycles Isothermal at 70'F and 1.000 psi = 300 cycles (same as number of Startup events) Warmup-Cooldovn = 300 cycles Warmup-Biowdown I= cycle TOTAL 603 cycles Revised CUP Calculation for 60 Years: f Pr+PoQ (000 Note t) Ke (ae Note 2) S.,t (see Note 3) n (see Note 6) N (see Note 4) U 34,690 Improper Startup - Warmup 1.00 124,825 1 332 0.0030 33,095 Hydrotest - Warmup 1.00 40,804 602 8,095 0,0744 Total, U60 = 0.0774 Environmental CUF Calculation for 60 Years: Maximum P 0.. Multiptier for HWC Conditions = 5.39 Maximum F0-,wc Multiplier for NWC Conditions = 13.17 U a40

U, 0 x Fen.wc x 0.53 + U60 x F,.Hwc x 0.47

0.7364 Overall Multipliero= Ue,.o0/1.U10 9.51 SIR-07-132-NPS, Rev. I 3-10 Structural Integrity Associates, Inc.

Table 3-3. Environmental Fatigue Evaluation for the Reactor Vessel Feedwater Nozzle Forging Blend Radius Low Alloy Steel: F= exp(0.898 - 0.101S'T'O'E) Assume S" = 0.015 (maximum) Assume [N= ln(O.001) = -6.908 (minimum) For a BWR with HWC environment (post-HWC implementationr: For a BWR with NWC environment (pre-HWC implementation): DO = 97 ppb = 0.097 ppm, so O = In(0.097/0.04) = 0886 DO = 114 ppb = 0.114 ppm. so O= In(0.114/0.04) 1 047 Thus: Thus: T ('C) T (F) F,. T (-C) T (F) F.. 0 32 2.45 0 32 2.45 50 122 2.45 50 122 2.45 100 212 2.45 100 212 2.45 150 302 2.45 150 302 2.45 200 392 3.90 200 392 4.25 250 482 6.20 250 482 7.35 288 550 8.82 288 550 11.14 Thus, maximum F-8.82 (Tr= (-150l for T 150Cl Thus, maximum F.,,= 11.14 Carbon Steel: F_, = exp(0,554 - 0.101S*TO'0) Assume S" = 0.015 (maximum) Assume a = In(0.001) = -6.908 (minimum) II For a 8WR with HWC environment (post-HWC implementation): For a BWR with NWC environment (pre-HWC implementation): DO = 40 ppb = 0.040 ppm 0.050 ppm so 0* = 0 DO = 40 ppb =0.040 ppm < 0.050 ppm so 0* =0 Thus: Thus: T (C) T (°F) F,. T (°C) T )F) F_ 0 32 1.74 0 32 1.74 I 50 122 1.74 50 122 1.74 100 212 1.74 100 212 1.74 150 302

1. 74 150 302 1.74 200 392 1.74 200 392 1.74 250 482 1.74 250 482 1.74 288 550 1.74 288 550 1.74 I

Thus, maximum F, = 1.74 IT'= (T-150) for T 150-C] Thus, maximum F.,= 1.74 Overall 60-Year No. Component Material 6-e Environmental Environmental CUF Multiplier CUF (1,2) 1 Feedwater Nozzle Forging Blend Radius Low Alloy Steel 0.0636 10.05 0.6392 2 Feedwater Nozzle Forging Safe End Carbon Steel 0.1471 1.74 0.2560 Notes:

1. An Fen Multiplier was used for each respective component with the following conditions:

+ 47% HWC conditions and 53% NWC conditions 2: Results using updated ASME Code fatigue calculations and actual cycles accumulated to-date and projected to 60 years. I I SIR-07-132-NPS, Rev. 1 3-1 1 StructuirallIntegrity Associates, Inc. I I

Table 3-4. Environmental Fatigue Evaluation for the Recirculation/RHR Piping Tee Stainless Stee F, = exp(0.935 - T*VsO*) For a BWR with HWC environment (post-HWC implementation): For a BWR with NWC environment (pre-HWC implementation): DO = 46 ppb = 0.046 ppm < 0.050 ppm. so O* = 0.260 D0 z 123 ppb = 0.123 ppm - 0.05 ppm. so O0 = 0.172 Conservatively use T' = 1 for T - 200°C Conservatively use T = 1 for T > 200°C Thus: Thus: -= 0 for E o 0.4%/sec so Fln= 2.55 sO F,, = 2.55 SIn(0/0.4) for 0.0004 v= o <= 0.4%/sec so F_, ranges from 2.55 so F,, ranges from 2.55 - ln(0.0004/0.4) for v 0.0004%/sec so F,, = 15.35 so F,, = 8.36 Thus, maximum F,, = 15.35 Thus, maximum F,, = 8.36 60-Year Overall 60-Year No. Component Material CUF Environmental Environmental Multiplier CUF (1,2) 1 Recirculation IRHR Piping Return Tee Stainless Steel 0.0590 12.62 0.7446 Notes:

1. An Fen multiplier was used for each respective component with the following conditions:

+ 39% HWC conditions and 61% NWC conditions

2. Results using updated ASME Code fatigue calculations and actual cycles accumulated to-date and projected to 60 years.

SIR-07-132-NPS, Rev. I 3-12 C Structural Integrity Associates, Inc.

Table 3-5. Environmental Fatigue Evaluation for the Reactor Recirculation Inlet Nozzle Forging Component: Recirculation Inlet Nozzle Forging NUREG/CR-6260 CUF: 0.310 (for reference only)

Reference:

NUREG/CR-6260, p. 5-105 Stress Report CUF: 0.0433 (updated for Point 12. see below) Material: Low Alloy Steel (Material = A-508 Cl. I1 perp. 1-S8-4 of CBIN Stress Report Section S8) Ilenion Rasis CUF Calculation for 40 vears:

Efatie, E-1y.s 1.1278 Power Uprate =

1.0067 Kn 1.660 m= 2.0 n 0.2 S,= 26,700 = 30.0/26.6 (per p. l-S8-24 of CBIN Stress Report Section S8 and ASME Code fatigue curve) =(549-100)1(546-100) per 4.4.1.b of 26A6019, Rev. 1 stress concentration factor (f. A270 of Vt/C-378, Roe. 0) N13-3228.5 of ASME Code, Section lt NB.3228.5 of ASME Code, Section lit psi (ASME Code, Section It. Part D) PL+PB+Q (See Note t) Skin Stress (see Note 2) K, (see Note 3) Sn, (see Note 4) n (see Note 5) N (see Note 6) U 43,110 15,145 1.00 49,224 200 4,614 0.0433 Total, U40 = 0.0433 Notes: 1. 2. 3. 4. 5. 6. 7. PL +P8 +0 is obtained for Point 12 from p. A270 of VYC-378, Rev, 0. Skin Stress is obtained for Point 12 from p. A270 of IYC-378, Roe. 0. K. computed in accordance with NB-3228.5 of ASME Code, Section ItI. St, = 0 5 'K,

EONa*'

I-/t.,,

Power Uprate ([ (P, +P0 +Q) K, + Skin Stress ].

n for 40 years is the number of Heatup-Cooldown cycles, per p. 828 of VYC-378, Rev. 0, N obtained from Figure 1-9 1 of Appendix I of ASME Code, Section IfI n for 60 yeats is the projected number of Heatup-Cooldown cycles. Revised CUF Calculation for 60 Years: PL+Pe+Q (see Note I) Skin Stress (see Note 2) K, (see Note 3) S,, (see Note 4) n (see Note 5) N (see Note 7) U 43,110 15,145 1.00 49,224 300 4,614 0.0650 Total, U,5 = 0.0650 Environmental CUF Calculation for 60 Years: Maximum Fn-,WC Multiplier for HWC Conditions = 2.45 Maximum Fn,-.wc Multiplier for NWC Conditions = 12.43 Ue,,no = U60 x Fnnwc x 0.53 + Uso x Fn.owc x 0.47 = 0.5034 Overall Multiplier = Uvýo/U.o = 7.74 I I I I I I I I I I I I I I ,I I I SIR-07-132-NPS, Rev. I 3-13 l Structural Integrity Associates, Inc.

Table 3-6. Environmental Fatigue Evaluation for Reactor Recirculation Inlet Nozzle Safe End Component: Recirculation Inlet Nozzle Safe End NUREGICR-6260 CUF: 0.310 (for reference only)

Reference:

NUREG/CR-6260, p. 5-105 Stress.Report CUF: 0.0017 (updated for Location 6-1, see below) Material: Stainless Steel (316L per p. 8 of 23A4292, Rev 4) Design Basis CUF Calculation for 40 years: E~t..._ c/E_ Power Upra Iy,- = 1.1076 te = 1.0067 K,= 1.280 mI 1.7 n= 0.3 = 28& 3 /25.55 (per p. 62 of Refereece [1t] and ASME Code fatigue curve) =(549 - 100) /(546 - 100) per 4.4.1.b of 26A6019, Rev I stress concentration factor (p. 827 of VYC-378, Rev. 0) NB-3228.5 ofASME Code, Section III NB-3228.5 of ASME Code, Section /lI psi (ASME Code, Section 11 Part D) n (see Note 5) N (see Note 6) U 2,076 1,242,266 0.0017 L Total, U4 o = 0.0017 S,,= 16,600 PL+PB+Q (see Note 1) P+Q+F (see Note 2) K, (see Note 3) SSk (see Note 4) 47,183 36,972 1.00 26,385 Notes: I. P, +P8 Q is obtained for Surface I (after weld overlay) from p. 117ofReference [18]. 2, PsQ+F is obtained for Point 6-I from p. 118 of Reference t18] (BEFORE weld overlay).

3. K, computed in accordance with NB-3228.5 of ASME Code, Section Itt.

4, Sa. =. e*Etlo-eEa~lss*wer-Uprate * [(P+Q+F) K, ] 5 n for040 years is the number of cycles as follows per p. B26 of VYC-378. Rev. 0: Design Hydrotest = 130 Loss of Feedpumps Composite: StartupiShutdown = 290 SRV Sowrdown = 8 Loss of Feedwater Pumps 30 110 events x 3 up/down cycles per event SCRAM = 2f0 Normal +/- Seismic = 11 10 cycles of upset seismic, plus I Level C seismic event Normal = 739

Sum of as of above events Zeroload

596

Startup/Shutdown + SRV Blowdown + Scram + LOFP Total number of cycles

2,076

6. N obtained from Figure f-9.2 of Appendix I of ASME Code, Section HI,

7. n for 60 years is file projected number of cycles as follows:

!?!Iigo HydruteV = 72 ................ e.* *. Loss of Feedpumps Composite: Startup/Shutdown = 300 i SRVSlowdown-= I Loss of Feedwater Pumps 30 10 events x 3 up/down cycles per event SCRAM = 200 t'All remaining scrams S. R _ 2.............A l e in n sc s Normal +a-Seismic = 11 Assume the same Normal = 751

Sum of all of above events Zeroload

620

StartuprShutdown + SRV Slowdown 0 Scram + LOFP Total number of cycles

2,122 Revised CUF Calculation for 60 Years: PL+PB+Q (see Note 1) P+Q+F (see Note 2) K. (see Note 3) S%, (see Note 4) n (see Note 5) N (see Note 7) U 47,183 36,972 1.00 26,385 2,122 1,242,266 0.0017 T Total, Us0 = 0.0017 Environmental CUF Calculation for 60 Years: Maximum Fe-uwc Multiplier for HWC Conditions = 15.35 Maximum F-,.NWC Multiplier for NWC Conditions = 8.36 U_,,vo = U6o x Fn,.wC x 0.53 + Uo x F-41HW x 0.47 = 0.0199 Overall Multiplier = UeeO.6oU6o = 11.64 SIR-07-132-NPS, Rev. I 3-14 f Structural Integrity Associates, Inc. A

Table 3-7. Environmental Fatigue Evaluation for Recirculation Outlet Nozzle Forging LOW Alloy Steel: F, = exp(0.898 - 0.101S*T*O*E*) Assume S° = 0,015 (maximum) Assume e. = 1n(0.001) -6.908 (minimum) For a BWR with HWC environment (post-HWC implementation): For a 6WR with NWC environment (pre-HWC implementation): DO = 46 ppb = 0.046 ppm 00 = 123 ppb = 0.123 ppm, so 0* = In(0. 123/0.04) 1.123 DO < 0.050 ppm, so 0* = 0 Thus: Thus: T (-C) T (°F) F_, T (-C) T (°F) F. 0 32 2.45 0 32 2.45 50 122 2.45 50 122 2.45 100 212 2.45 100 212 2.45 150 302 2.45 150 302 2.45 200 392 2.45 200 392 4.42 269.45 517.0t 2.45 269.45 517.01 10.00 288 550 2.45 288 550 12.43 Thus, maximum-F, = 2.45 [T'=(T-150) fof T-10oC] thus, maximum F,, 12.43 Overall 60-Year No. Component Material 60-Year Environmental Environmental CUF Multiplier CUF (1,2) 1 Recirculation Outlet Nozzle Forging Blend Radius Low Alloy Steel 0.0108 7.74 0.0836 Notes:

1. An F., multiplier was used for each respective component with the following conditions:

+ 47% HWC conditions and 53% NWC conditions

2. Results using updated ASME Code fatigue calculations and actual cycles accumulated to-date and projected to 60 years.

I I I I I I I I I I I I 1 I I I I SIR-07-132-NPS, Rev. I 3-15 V Structural Integrity Associates, Inc. I 1

Table 3-8. Environmental Fatigue Evaluation for Core Spray Reactor Vessel Nozzle Forging Blend Radius, Safe End, and Piping Low Alloy Steel:F. = exp(0 89 8 - 0101S*T*O'L) Assume S* = 0,015 (maximum) Assume 2N= In(O.001) -6,908 (minimum) For a BWR with HWC environment (post-HWC implementation): For a BWR with NWC environment (pre-HWC implementation): DO = 97 ppb = 0.097 ppm. soO = ln(0.097/0.04) = 0.886 DO = 114 ppb = 0.114 ppm, so 0* In(0 11410.04) = 1.047 Thus: Thus: T(°-)Q T ("F) F_. T (-C) T (°F) F.. 0 32 2.45 0 32 2.45 50 122 2.45 50 122" 2.45 100 212 2.45 100 212 2.45 150 302 2.45 150 302 245 200 392 3.90 200 392 4.25 250 482 6.20 250 482 7.35 288 550 8.82 288 550 11.14 Thus, maximum Fn = 8.82 (T'= (T-150) f.r T 150-C) Thus, maximum F-, = 11,14 Stainless Steel. F-, = exp(0,935 - T*°*O*) For a BWR with HWC environment (post-HWC implementation): For a 8WR with NWC environment (pre-HWC implementation): DO = 97 ppb =.097 ppm 0.050 ppm, so 0* =0.172 0O = 114 ppb = 0 114 ppm 0.05 ppm, soaO = 0,172 Conservatively use T = 1 for T > 200'C Conservatively use T = 1 for T > 200'C Thus: Thus: ° 0 for o > 0.4%/sec so F_, = 2.55 so F,, = 2.55

in(O/0.4) for 0.0004 <

u 0.4%/sec so F,. ranges from 2 55 so F_, ranges from 2.55

ln(0.0004/0.4) for n0.0004%/sec so F_,

8.36 so F.. = 8.36 Thus, maximum Fn = 836 Thus, maximum F., = 8.36 Overall 60-Year No. Component Material CUF Environmental Environmental Multiplier CUF (1,2) 1 Core Spray Nozzle Forging Blend Radius Low Alloy Steel 0.0166 10.05 0.1668 2 Core Spray Nozzle Safe End Ni-Cr-Fe 0.0398 1.49 0.0593 3 Core Spray Piping Stainless Steel 0.0011 8.36 0.0092 Notes:

1. An Fen Multiplier was used for each respective component with the following conditions:

+ 47% HWC conditions and 53% NWC conditions

2. Results using updated ASME Code fatigue calculations and actual cycles accumulated to-date and projected to 60 years.

SIR-07-132-NPS, Rev. 1 3-Jl6 Structural Integrity Associates, Inc.

Table 3-9. Environmental Fatigue Evaluation for the Feedwater Line Class 1 Piping Carbon Steer: F_, = exp(0.554 - 0.101S*T'O 5 1) AssumeS* = 0.015 (maximum) Assumer. = ln(0.001) = -6.908 (minimum) For a BWR with HWC environment (post-HWC implementation): For a BWR with NWC environment (pre-HWC implementation): DO = 40 ppb = 0.040 ppm < 0.050 ppm so 0* = 0 DO = 40 ppb =0.040 ppm < 0.050 ppm so O* =0 Thus: Thus: T (-C) T (°F) F_, T ('C) T ('F) F,. 0 32 1.74 0 32 1.74 50 122 1.74 o50 122 1.74 100 212 1.74 100 212 1.74 150 302 1.74 150 302 1.74 200 392 1.74 200 392 1.74 250 482 1.74 250 482 1.74 288 550 1.74 288 550 1.74 Thus. maximum F,, = 1.74 IT'= (T-150) fof T 150-C) Thus, maximum F_, = 1.74 Overall 60-Year No. Component Material CUF Environmental Environmental Multiplier CUF (1,2) 1 Feedwater Piping Riser to RPV Nozzle N4B Carbon Steel 0.1661 1.74 0.2890 Notes:

1. An Fenmultiplier was used for each respective component with the following conditions:

+ 47% HWC conditions and 53% NWC conditions

2. Results using updated ASME Code fatigue calculations and actual cycles accumulated to-date and projected to 60 years.

I I I I I I I I I I I I I I SIR-07-132-NPS, Rev. 1 3-17 V Structural Integrity Associates, Inc. U I

Table 3-10. Summary of Environmental Fatigue Calculations for VYNPS Overall 60-Year 40-Year Design 60-Year Eviral Evrneal No. Component Material CUF' 1 CUF "' Environmental Environmental Multiplierl3' CUF 1 RPV Shell/Bottom Head Low Alloy Steel 0.0057 0.0085 9.51 0.0809 2 RPV Shell at Shroud Support Low Alloy Steel 0.0549 0.0774 9.51 0.7364 3 Feedwater Nozzle Blend Radius Low Alloy Steel (4) 0.0636 10.05 0.6392 4 Recirculation/RHR Class 1 Piping (Return Tee) Stainless Steel (4) 0.0590 12.62 0.7446 5 Recirculation Inlet Nozzle Forging Low Alloy Steel 0.0433 0.0650 7.74 0.5034 6 Recirculation Inlet Nozzle Safe End Stainless Steel 0.0017 0.0017 11.64 0.0199 7 Recirculation Outlet Nozzle Forging Low Alloy Steel (4) 0.0108 7.74 0.0836 8 Core Spray Nozzle Forging Blend Radius (5) Low Alloy Steel (4) 0.0166 10.05 0.1668 9 Feedwater Class 1 Piping Carbon Steel (4) 0.1661 1.74 0.2890 Notes:

1. Updated 40-year CUF calculation based on recent ASME Code methodology and design basis cycles.

2. CUF results using updated ASME Code methodology and actual cycles accumulated to-date and projected to 60 years.

3. An F,, multiplier was used for each respective component with the following conditions:

+ 47% HWC conditions and 53% NWC conditions

4. 40 year values were not calculated for these locations

5. Only the highest CUF from Table 3-8 is shown 132-NPS, Rev. 1 3-18 SIR 411 aryUralt ginwry ASSOC~dwS, inc.

4.0

SUMMARY

AND CONCLUSIONS The results of Tables 3-1 through 3-9, as summarized in Table 3-10, demonstrate that the fatigue usage factor, including environmental effects, remainswithin the allowable value of 1.0 for 60 years of operation for the following component locations: J Reactor vessel shell, bottom head and shroud support V Reactor vessel feedwater nozzle v Reactor recirculation piping (including the reactor inlet and outlet nozzles) " Core spray line reactor vessel nozzle and associated Class 1 piping V Feedwater line Class 1 piping Therefore, the environmental fatigue assessment results for all of the NUREG/CR-6260 locations associated with the older vintage BWR plant are acceptable for 60 years of operation for VYNPS. I I I I I I I I I I I I I I I I I SIR-07-132-NPS,,Rev. I 4-1 V Structural Integrity Associates, Inc. I I

\\

5.0 REFERENCES

1.

U. S. Nuclear Regulatory Commission, Generic Safety Issue 166, "Adequacy of Fatigue Life of Metal Components."

2.

U. S. Nuclear Regulatory Commission, Generic Safety Issue 190, "Fatigue Evaluation of Metal Components for 60-Year Plant Life."

3.

SECY-95-245, "Completion of the Fatigue Action Plan," James M. Taylor, Executive Director for Operations, U. S. Nuclear Regulatory Commission, Washington, DC, September 25, 1995.

4.

Memorandum, Ashok C. Thadani, Director, Office of Nuclear Regulatory Research, to William D. Travers, Executive Director for Operations, Closeout of Generic Safety Issue 190, "Fatigue Evaluation of Metal Components for 60 Year Plant Life," U. S. Nuclear Regulatory Commission, Washington, DC, December 26, 1999.

5.

NUREG-1801, Revision 1, "Generic Aging Lessons Learned (GALL) Report," U.S. Nuclear Regulatory Commission, September 2005.

6.

NUREG/CR-6260 (INEL-95/0045), "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," March 1995.

7.

NUREG/CR-6583 (ANL-97/18), "Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels," March 1998.

8.

NUREG/CR-5704 (ANL-98/3 1), "Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels," April 1999.

9.

NUREG/CR-5999 (ANL-93/3), "Interim Fatigue Design Curves for Carbon, Low-Alloy, and Austenitic Stainless Steels in LWR Environments," April 1993.

10.

USAS B31.1 - 1967, USA Standard Code for Pressure Piping, "Power Piping," American Society of Mechanical Engineers, New York.

11.

EPRI Report No. TR-105759, "An Environmental FactorApproach to Account for Reactor Water Effects in Light Water Reactor Pressure Vessel and Piping Fatigue Evaluations," December 1995.

12.

Structural Integrity Associates Calculation No. VY-16Q-303, Revision 0, " Environmental Fatigue Evaluation of Reactor Recirculation Inlet Nozzle and Vessel Shell/Bottom Head." SIR-07-132-NPS, Rev. 1 5-1

' Structural Integrity Associates, Inc.

I I

13.

INFORMATION REDACTED i

14.

Entergy Design Input Record (DIR) Rev. 1, EC No. 1773, Rev. 0, "Environmental Fatigue Analysis for Vermont Yankee Nuclear Power Station," 7/26/07, SI File No. VY-16Q-209.

15.

VY Calculation No. VYC-378, Revision 0, "Vermont Yankee Reactor Cyclic Limits for Transient Events," 10/16/85, SI File No. VY-05Q-21 i.

16.

Chicago Bridge & Iron RPV Stress Report, Section S3, Revision 4, "Stress Analysis, Shroud Support, Vermont Yankee Reactor Vessel, CB&I Contract 9-6201," 2/3/70, SI File No. VY-16Q-203.

17.

Structural Integrity Associates Calculation No. VY-16Q-302, Revision 0, "Fatigue Analysis of Feedwater Nozzle."

18.

Structural Integrity Associates Calculation No. VY-16Q-307, Revision 0, "Recirculation Class I Piping Fatigue and EAF Analysis."

19.

CB&I RPV Stress Report, Section S8, Revision 4, "Stress Analysis, Recirculation Inlet Nozzle, Vermont Yankee Reactor Vessel, CB&I Contract 9-620 1," 2/3/70, SI File No. VY-16Q-203.

20.

GE Nuclear Energy Certified Stress Report No. 23A4292, Revision 0, "Reactor Vessel - Recirculation Inlet Safe End Nozzle," January 21, 1985, SI File No. VY-16Q-203.

21.

Structural Integrity Associates Calculation No. VY-16Q-3 10, Revision 1, "Fatigue I Analysis of Core Spray Nozzle."

22.

Structural Integrity Associates Calculation No. VY-16Q-3 11, Revision 0, "Feedwater Class i 1 Piping Fatigue Analysis."

23.

EPRI/BWRVIP Memo. No. 2005-271, "Potential Error in Existing Fatigue Reactor Water Environmental Effects Analyses," July 1, 2005

24.

Structural Integrity Associates Calculation No. VY-16Q-306, Revision 0, "Fatigue Analysis of Recirculation Outlet Nozzle." I I SIR-07-132-NPS, Rev. 1 5-2 Structural Integrity Associates, Inc. hisa n sRe renc oV or e Inf "n

smared itha "ar"in he igh-had mrgi)U I}}

ML082401759

Text

Title:

1.0 INTRODUCTION

1 (1 = CARBON/LOW ALLOY, 2 = STAINLESS STEEL) m =2.0 n

Subject:

6.0 CONCLUSION

7.0 REFERENCES

1.0 INTRODUCTION

6.0 CONCLUSION

1.0 INTRODUCTION

6.0 CONCLUSION

7.0 REFERENCES

0 for c > 0.4%/sec so Fen

0 for r. > 0.4%/sec so Fc.

In(0,0004/0.4) for e < 0.0004%/sec so F_

ln(0.0004/0.4) for t: < 0.0004%/sec so F,,

Reference:

Reference:

Reference:

Sum of all of above events Zeroload

Startup/Shutdown, SRVBIowdotw + Scram + LOFP Total number of cycles

Sum of all of above events Zeroload

Startup/Shutdown + SRV Blovdown + Scram v LOFP Total number of cycles

Title:

1.0 INTRODUCTION

2.0 BACKGROUND

5.0 REFERENCES

1.0 INTRODUCTION

2.0 BACKGROUND

Reference:

Reference:

U, 0 x Fen.wc x 0.53 + U60 x F,.Hwc x 0.47

Reference:

Reference:

Sum of as of above events Zeroload

Startup/Shutdown + SRV Blowdown + Scram + LOFP Total number of cycles

Sum of all of above events Zeroload

StartuprShutdown + SRV Slowdown 0 Scram + LOFP Total number of cycles

in(O/0.4) for 0.0004 <

ln(0.0004/0.4) for n0.0004%/sec so F_,

SUMMARY

5.0 REFERENCES

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