ML080850074

2008/03/17-New England Coalition, Inc.'S (NEC) Motion to File a Timely New or Amended Contention

ML080850074
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	03/17/2008
From:	Tyler K New England Coalition, Shems, Dunkiel, Kassel, & Saunders, PLLC
To:	Atomic Safety and Licensing Board Panel
SECY/RAS
References
06-849-03-LR, 70-271-LR, RAS M-9
Download: ML080850074 (51)

Text

DOCKETED USNRC UNITED STATES March 17, 2008 (4:28pm)NUCLEAR REGULATORY COMMISSION OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF In the Matter of ))ENTERGY NUCLEAR VERMONT YANKEE, LLC ) Docket No. 50-271-LR and ENTERGY NUCLEAR OPERATIONS, INC. ) ASLB No. 06-849-03-LR

)Vermont Yankee Nuclear Power Station )NEW ENGLAND COALITION, INC.'S (NEC) MOTION TO FILE A TIMELY NEW OR AMENDED CONTENTION The New England Coalition, Inc. (NEC) moves, pursuant to 10 C.F.R. §2.309(f)(2) and the Initial Scheduling Order ¶ 5, for leave to file a timely new or amended contention addressing an additional analysis performed by Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. ("Entergy")

to assess the impact of environmentally assisted metal fatigue (EAF) on the Vermont Yankee plant feedwater nozzle. Entergy produced its reports of this additional analysis (hereinafter referred to as "Entergy's Second CUFen Reanalysis")

to NEC on February 15, 2008.These reports are attached hereto as Exhibits A-C to the Seventh Declaration of Dr.Joram Hopenfeld.

See, Attachment 1.NEC's proposed new or amended contention satisfies requirements of 10 C.F.R. §2.309(f)(2).

Entergy's Second CUFen Reanalysis is materially different from both the EAF analysis reported in Entergy's License Renewal Application, and the Entergy EAF analysis that NEC's Contention 2A addresses.

Entergy employed different methods, and produced different results. See, Attachment 1, Seventh Declaration of Dr. Joram Hopenfeld.

Reports of Entergy's Second CUFen Reanalysis were not available to NEC until Entergy produced them on February 15, 2008. NEC's filing is timely. See, Initial Scheduling Order ¶ 5 (a motion for a new or amended contention shall be deemed timely under 10 C.F.R. § 2.309(f)(2)(iii) if it is filed within thirty (30) days of the date when the new and material information on which it is based first becomes available);

10 C.F.R. §2.306 ("In computing any period of time .... the last day of the period so computed is included unless it is a Saturday, Sunday, or legal holiday. .., in which event the period runs until the end of the next day which is neither a Saturday, Sunday, nor holiday.").

PROPOSED NEW OR AMENDED CONTENTION NEC incorporates by reference its now pending Contentions 2 and 2A (metal fatigue) and its Reply to Entergy's Answers to Contentions 2 and 2A. In addition, NEC contends the following.

(1) Statement of Issue of Law or Fact to be Raised, and Brief Explanation of Basis. 10 C.F.R. §§ 2.309(f)(i), 2.309(f)(ii).

NEC's Contention 2 (metal fatigue) is that data reported in Entergy's License Renewal Application Table 4.3-3 indicate that critical reactor components may fail due to environmentally assisted metal fatigue (EAF) during the period of extended operation, and that Entergy has not proposed an adequate aging management plan addressing this issue as required pursuant to 10 C.F.R. § 54.21.1 NEC's Contention 2A addresses a reanalysis Entergy performed of the impact of EAF on the components listed in License Renewal Application Table 4.3-3 ("Entergy's First CUFen Reanalysis").

Contention 2A is that Entergy's First CUFen Reanalysis is flawed by unjustified assumptions, uncertainties and insufficient conservatism, and fails to prove that the components assessed will not fail due to metal fatigue during the period of Per the Board's Order of November 7, 2007, NEC's Contention 2 is held in abeyance pending resolution of NEC's Contention 2A.2 extended operation.

NEC understands that Entergy has performed its Second CUFen Reanalysis to validate the results of its First CUFen Reanalysis.

NEC now contends, as explained in detail in the attached Seventh Declaration of Dr. Joram Hopenfeld, that Entergy's Second CUFen Reanalysis neither validates the results of Entergy's First CUFen Reanalysis, nor independently demonstrates that CUFens for all components listed in License Renewal Application Table 4.3-3 and all NUREG/CR-6260 locations are less than one. Entergy's Second CUFen Reanalysis addresses only one issue: the uncertainty in calculation of CUF values used in Entergy's First CUFen Reanalysis resulting from use of the Green Function.

It does not address errors in Entergy's First CUFen Reanalysis resulting from the several other factors identified in NEC's Contention 2A and the supporting Sixth Declaration of Dr. Joram Hopenfeld.

See, Attachment 1 ¶¶ 7-10. Further, Entergy's Second CUFen Reanalysis addresses only the feedwater nozzle, and its results are not bounding for other components.

Id. at ¶¶ 11 -12.(2) Scope of the Proceeding and Materiality.

10 C.F.R. §§ 2.309(f)(iii), 2.309(f)(iv).

This contention addresses Entergy'splan, as stated in the License Renewal Application, to monitor and manage the effects of aging on reactor components that are subject to an aging management review, pursuant to 10 C.F.R. § 54.21 (a), and an evaluation of time-limited aging analysis, pursuant to 10 C.F.R. § 54.21(c).

These are issues within the scope of this proceeding, and material to findings the NRC must make in this matter. See, 10 C.F.R. § 54.4; Duke Energy Corp. (McGuire Nuclear Station, Units I and 2; Catawba Nuclear Station, Units 1, 2 and 3), 56 NRC 358, 363-64 (2002).3 (3) Expert or Factual Support. 10 C.F.R. § 2.309(f)(v).

This contention is supported by the attached Declaration of NEC's expert witness, Dr. Joram Hopenfeld.

(4) Genuine Dispute of Material Law or Fact. 10 C.F.R. § 2.309(f)(vi).

The attached Seventh Declaration of Dr. Joram Hopenfeld includes ample information to establish a genuine dispute with the Applicant concerning the validity of Entergy's Second CUFen Reanalysis.

NEC is required to make only "a minimal showing that the material facts are in dispute, thereby demonstrating that an inquiry in depth is appropriate." In Gulf State Utilities Co., 40 NRC 43, 51 (1994).NEC HAS CONSULTED OTHER PARTIES Pursuant to 10 C.F.R. § 2.323(b), NEC has consulted or attempted to consult with all parties to this proceeding concerning this motion. The State of Vermont does not object to the filing of this motion. Entergy considers NEC's motion timely, and will respond to its substance after reviewing NEC's pleading.

The NRC Staff will respond, after reviewing NEC's pleading.

The State of New Hampshire did not take a position.CONCLUSION NEC respectfully requests that the Board grant NEC's Motion to File a Timely New or Amended Contention, and admit NEC's new or amended contention for adjudication in this proceeding.

March 17, 2008 New England Coalition by: -u Andrew Raubvogo Karen Tyler SHEMS DUNKIEL KASSEL & SAUNDERS PLLC For the firm Attorneys for NEC 4 ATTACHMENT 1 UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of ))ENTERGY NUCLEAR VERMONT YANKEE, LLC ) Docket No. 50-271-LR and ENTERGY NUCLEAR OPERATIONS, INC. ) ASLB No. 06-849-03-LR

~)Vermont Yankee Nuclear Power Station )SEVENTH DECLARATION OF DR. JORAM HOPENFELD 1. My name is Dr. Joram Hopenfeld.

The New England Coalition (NEC) has retained me as an expert witness in proceedings concerning the application of Entergy Nuclear Operations, Inc. ("Entergy")

to renew its operating license for the Vermont Yankee Nuclear Power Station ("VYNPS")

for twenty years beyond the current expiration date of March 21, 2012.2. I am a mechanical engineer and hold a doctorate in engineering.

I have 45 years of professional experience in the fields of instrumentation, design, project management, and nuclear safety, including 18 years in the employ of the U.S. Nuclear Regulatory Commission (NRC). My curriculum vitae was previously filed in this proceeding as an attachment to my declaration in support of NEC's Petition to Intervene.

3. I have reviewed the VYNPS License Renewal Application, submitted to the NRC in 2006, and amendments thereto. I have also reviewed the following Structural Integrity Associates, Inc. (SIA) Environmental Fatigue Analysis Calculations and Reports, which I understand that Entergy produced to New England Coalition (NEC) on February 15, 2008.

SIA Calculation or Report No. Title Calculation File No. Design Inputs and Methodology for ASME VY-19Q-301 Code Confirmatory Fatigue Usage Analysis of Reactor Feedwater Nozzle Calculation File No. ASME Code Confirmatory Fatigue VY- 19Q-302 Evaluation of Reactor Feedwater Nozzle Calculation File No. Feedwater Nozzle Environmental Fatigue VY- 19Q-303 Evaluation Copies of these calculations and reports are attached hereto as Exhibits A-C.4. These calculations and reports describe an additional analysis Entergy has performed to assess the impact of environmentally-assisted fatigue (EAF) on the feed water nozzle ("Entergy's Second CUFen Reanalysis").

Entergy represents that this additional analysis confirms the results of Entergy's CUFen Reanalysis that is the subject of NEC's Contention 2A ("Entergy's First CUFen Reanalysis"), which I have addressed in my Declaration in support of NEC's Contention 2A. See, Sixth Declaration of Dr.Joram Hopenfeld (September, 2007).5. Entergy's Second CUFen Reanalysis differs from its, First CUFen Reanalysis in that it does not rely upon the Green Function methodology to obtain stress histories during plan transients.

Instead, it applies the finite element model (FEM) method to calculate CUFs for plant transients.

Entergy claims that this Second CUFen Reanalysis confirms that CUFs for the feedwater, spray and recirculation nozzles that were calculated by the "Green Function" method in its First CUFen Reanalysis are valid, and that CUFens for all NUREG/CR-6260 locations are less than one.6. There are now four different CUFen values for the feedwater nozzle before the Board in this proceeding; summarized in the following Table 1.2 TABLE I- DIFFERENT CUF RESULTS FOR THE FEEDWATER NOZZLE REFERENCE CUF Fen CUFen (Environmental)

License Renewal Application 0.750 3.81 2.86 Table 4.3-3 Entergy First CUFen Reanalysis 0.0636 10.05 0.6392 (subject of NEC's Contention 2A)Entergy Second CUFen Reanalysis 0.0889 3.97 0.3531 Joram Hopenfeld Recalculation, Sixth 0.750 17.00 12.75 Declaration of Dr. Joram Hopenfeld (September, 2007)7. I do not agree that Entergy's Second CUFen Reanalysis confirms the validity of its First CUFen Reanalysis, or independently demonstrates that CUFens for all NUREG/CR-6260 locations are less than one. In my opinion, the results of Entergy s Second CUFen Reanalysis are baseless.8. Entergy's Second CUFen Reanalysis addresses only one issue: the uncertainty in CUF values used in Entergy's First CUFen Reanalysis resulting from use of the Green Function.

Entergy's Second CUFen Reanalysis does not address the errors in calculation of the environmentally corrected usage factor, CUFen, due to the several other factors I have identified in my Sixth Declaration in support of NEC's Contention 2A.9. For instance, no matter how accurate Entergy's FEM-based calculations may be, the accuracy of the final CUFen values also depends on the accuracy of inputs such as the heat transfer coefficients, and the accuracy of the equations that describe the environmental correction factor (Fen). Even though Fen values based on more recent data would have resulted in CUFens greater than 1, Entergy still has not provided justification for using low Fen factors. See, Sixth Declaration of Joram Hopenfeld (September, 2007).3

10. As shown in Table 1, above, the error in the CUFen due to the use of the Green Function is relatively small compared to that resulting from the uncertainties in the Fen values. The CUF as predicted by the Second CUFen Reanalysis methodology is smaller by a factor of 1.81 (0.6391/0.3531) than the CUF which was calculated using the Green Function in Entergy's First CUFen Reanalysis.

On the other hand, Fen is increased by a factor of 4.3 (17/3.97) when more recent data is considered in the calculation of the Fen as described in NEC's Contention 2A. This comparison does not imply that errors in the CUF which were introduced by the Green Function are not important; it only demonstrates that the errors which are introduced by an incorrect Fen are more important.

11. Moreover, I do not agree that results of Entergy's Second CUFen Reanalysis, addressing only the feedwater nozzle, are bounding for the other components addressed by Entergy's First CUFen Reanalysis.

There are considerable differences in geometry and heat transfer characteristics between the feedwater and the spray and recirculation nozzles. These differences could result in different stress distributions which could result in higher CUFs for the spray and recirculation nozzles even if the number of transients for the two nozzles is smaller than the number for the feedwater nozzle. The reports of Entergy's Second CUFen Reanalysis, Exhibits A-C hereto, do not discuss these differences.

Also, in my opinion, the projection of Entergy's results for the feedwater nozzle using the FEM method to other components could only be justified if Entergy could demonstrate an understanding of the reasons for the differences in the CUFs obtained by the simplified "Green Function" analysis and those that were obtained by the FEM method. Entergy's reports of its Second CUFen Reanalysis, Exhibits A-C hereto, 4 do not address this issue. Therefore, Entergy's Second CUFen Reanalysis results, such as they are, should apply only to the feedwater nozzle.12. The differing results stated in the above Table I demonstrate that the Green's Function introduces non-quantified and poorly understood uncertainties in the calculated CUF values. I attended a meeting between Entergy and the NRC on January 8, 2008, at which Entergy's consultants presented and discussed the results of the First CUFen Reanalysis.

My understanding based on the discussion at this meeting is that the NRC asked Entergy to calculate CUFen for the feedwater nozzle without using the Green Function, because Entergy's consultants were not able to justify the generic use of the Green Function for thermal transients.

Until similar calculations are performed for the circulation and the spray nozzles, the associated CUFs will include unquantified errors due to the use of the Green function.13. In conclusion, Entergy's Second CUFen Reanalysis does not address my criticisms of the First CUFen Reanalysis, validate the First CUFen Reanalysis, or independently demonstrate that CUFens for the components listed in License Renewal Application 4.3-3 are less than one. It is my opinion that acceptance of Entergy's results will lead to an unjustified reduction in the scope of fatigue monitoring at the Vermont Yankee plant.5 I declare under penalty of perjury that the foregoing is true and correct.Executed this _Lday of March, 2008 at Rockville, Maryland.

2300 N Street NW Tel 202.663.8142 Pillsbury Washington, DC 20037-1122 Fax 202.663.8007 W inthrop www.pillsburylaw.com Shaw PittmanLl, February 15, 2007 Matias F. Travieso-Diaz Phone: 202.663.8142 matias.travieso-diaz@pillsburylaw.com BY OVERNIGHT MAIL Mary C. Baty, Esq.Office of the General Counsel Mail Stop 0- 15 D21 U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Karen L. Tyler, Esq.Shems Dunkiel Kassel & Saunders PLLC 91 College Street Burlington, VT 05401 In the Matter of Entergy Nuclear Vermont Yankee, LLC, and Entergy Nuclear Operations, Inc.(Vermont Yankee Nuclear Power Station)Docket No. 50-271-LR; ASLBP No. 06-849-03-LR Re: Additional Environmentally Assisted Fatigue Calculations

Dear Ms. Baty and Ms. Tyler:

Enclosed are three additional environmentally assisted fatigue calculations recently performed by Structural Integrity Associates for Entergy relevant to New England Coalition's Contention 2A in the above captioned proceeding.

The enclosed calculations are: VY-19Q-301, "Design Inputs and Methodology for ASME Code Confirmatory Fatigue Analysis of Reactor Feedwater Nozzle;" VY-19Q-302, "ASME Code Confirmatory Fatigue Analysis of Reactor Feedwater Nozzle;" and VY-19Q-303,"Feedwater Nozzle Environmental Fatigue Evaluation." Electronic copies of these calculations are also being forwarded to you today.

Mary C. Baty, Esq. and Karen L. Tyler, Esq.February 15, 2008 Page 2 Copies of these calculations will be included in Entergy's March 2008 Supplemental Disclosure.

Sincerely, Matias F. Travieso-Diaz Counsel for Entergy Enclosures (as noted)Pillsbury Winthrop Shaw Pittman LLP Exhibit A Structural Integrity Associates, inc. Fie No.: VY-19Q-301 CALCULATION PACKAGE Project No.: VY-19Q PROJECT NAME: Provide VY Support for Questions Related to Environmental Fatigue Analyses CONTRACT NO.: 10163217 CLIENT: PLANT: Entergy Nuclear Operations, Inc. Vermont Yankee CALCULATION TITLE: Design Inputs and Methodology for ASME Code Confirmatory Fatigue Usage Analysis of Reactor Feedwater Nozzle Document Affected Project Manager Preparer(s)

&Revision Pages Revision Description Approval Checker(s)

Signature

& Date Signatures

& Date 0 1-18 Original Issue Computer Files J -errmann WF Weitze'\T-,Hr1/VZ/2008 WFW 01/29/2008 A. Chintapalli AC 01/2./2008 Page 1 of 18 F0306-O] RO Structural Integrity Associates, Inc.Table of Contents 1.0 O B JE C T IV E ...............................................................................

...................................................

3 2.0 M E T H O D O L O G Y .......................................................................................................................

3 3.0 ASSUMPTIONS/DESIGN INPUTS .......................................................................................

4 4.0 C A L C U L A T IO N S ......................................................................................................................

13 5.0 RESULTS OF ANALYSIS ..................................................................................................

17 6 .0 R E F E R E N C E S ....................................................

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

18 List of Tables T ab le 1: T ran sien ts ...............................................................................................................................

5 Table 2: Properties of Liquid W ater ...............................................................................................

8 Table 3: Properties of Saturated Steam ..........................................................................................

8 Table 4: Forced Flow and Natural Circulation Heat Transfer Coefficients, Bt/1--ft 2-°F .........8 Table 5: Temperature-Dependent Material Properties

...................................................................

10 Table 6: Condensation Heat Transfer Coefficients, Btu/hr-ft 2-°F ................................................

13 Table 7: M embrane Plus Bending Stresses Due to Piping Loads .................................................

14 List of Figures Figure 1: Nozzle and Vessel W all Thermal and Heat Transfer Boundaries

.....................................

9 Figure 2: Safe End Linearization Path ................................................................................................

11 Figure 3: Nozzle Corner Linearization Path .......................................................................................

11 Figure 4: Coordinate System for Forces and Moments .................................................................

14 Figure 5: Reducer Geometry Parameters

.......................................................................................

15 Figure 6: FW Nozzle Safe End Geometry ......................................................................................

16 FileNo.: VY-19Q-301 Page 2 of 18 Revision:

0 F0306-O1 RO Structural Integrity Associates, Inc.1.0 OBJECTIVE The objective of this calculation package is to establish the design inputs and methodology to be used for an ASME Code,Section III fatigue usage calculation of the reactor pressure vessel (RPV)feedwater (FW) nozzle at Vermont Yankee Nuclear Power Station (VYNPS).2.0 METHODOLOGY A detailed fatigue usage analysis of the FW nozzle will be performed using the methodology of Subarticle NB-3200 of Section III of the ASME Code [1]. The analysis will be used as a confirmnatory analysis for comparison with a previous fatigue usage analysis that was done using simplified methods. Therefore, only the fatigue portion of the ASME Code methodology will be used, and the analysis will be a fatigue assessment only, and not a complete ASME Code analysis.Finite element analysis will be perforlned using a previously-developed axisymmetric finite element model (FEM) of the FW nozzle. Thermnal transient analysis will be performed using the FEM for each defined transient.

Concurrent with the thermal transients are pressure and piping interface loads; for these loads, unit load analyses (finite element analysis for pressure, and manual calculations for piping loads) will be performed.

The stresses from these analyses will be scaled appropriately based on the magnitude of the pressure and piping loads during each thermal transient, and combined with stresses from the thermal transients.

Additional scaling of pressure stresses will be performed to account for nozzle comer contour effects (i.e., the effects of approximating the nozzle-to-RPV intersection of two cylinders with an axisyrnmetric model). Other stress concentration factors (SCFs) will be applied as appropriate.

/All six components of the stress tensor will be used for stress calculations.

The stress components for the non-axisymmetric loads (shear and moment piping loads) can have opposite signs depending upon which side of the nozzle is being examined.

Therefore, when combining stress components from these loads with stress components from thermal transients and other loads, the signs of the stress components will be adjusted to maximize the magnitude of the stress component ranges.The fatigue analysis will be performed at previously-examined locations for direct comparison of results. Stresses will be linearized at these locations.

The linearized primary plus secondary membrane plus bending stress will be used to determine the value of Ke to be used in the simplified elastic-plastic analysis in accordance with ASME Code NB-3200 methodology.

Environmental fatigue multipliers will be applied in accordance with NUREG/CR-6583

[15].FileNo.: VY-19Q-301 Page 3 of 18 Revision:

0 F0306-01RO Structural Integrity Associates, Inc.3.0 ASSUMPTIONS/DESIGN INPUTS 3.1 Assumptions

3.1.1 Power

uprate effects are considered as being applied to the entire period of operation.

The higher pressures, flows, and temperatures at uprate conditions are used in determining and applying heat transfer coefficients

[3, Section 3.2] [2, Section 3.1].3.1.2 The Boltup transient

[2, Tables ] and 2] analysis does not affect the FWnozzle and is therefore excluded from the transients analyzed.3.1.3 Where the flow rates in the thermal cycle diagram are at a value not calculated in Table 2, the next highest flow rate heat transfer coefficient will be used. This results in a higher heat transfer coefficient and is therefore conservative.

3.1.4 The effect of non-uniform geometries is judged to be insigniflcantforflow inside the safe end, because of the smooth transition and small geometry changes as shown in Figure 6. The smaller inner diameter (9.669") at the safe end was used to calculate heat transfer coefficients, resulting in a higher flow velocity and therefore conservative values.3.1.5 The annulus leakage flow rate used is 31 GPMfor EPU conditions

[3, Section 3.2].3.1.6 Density and Poisson's ratio used in the FEM are assumed typical values of p = 0.283 lb/in 3 and 0.3, respectively.

3.1.7 For purposes of linearizing stress at the nozzle corner, the effect of the cladding is conservatively neglected.

3.1.8 Stress

components due to piping loads 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, as was done in the previous analysis [2, Section 3.4].3.2 ASME Code Edition The analysis will be performed in a manner consistent with the fatigue usage rules in NB-3200 of Section III of the ASME Code; the 1998 Edition with Addenda through 2000 [1] will be used, for consistency with the previous analysis [2].3.3 Transients Previously developed thermal and pressure transients

[2, Section 3.1 and Tables 1 and 2] are used for this analysis.

The transients to be evaluated are shown in Table 1. For each transient, the time, nozzle fluid temperature (Tno 0), RPV pressure, percent FW flow rate, and number of cycles are included.

In some cases, flow rates and Tnoz values from the nozzle thermal cycle diagram [10, Attachment 1, p. 3] are used to reduce excess conservatism.

Note that the only difference between the nozzle comer and the safe end transients in the referenced document is the length of the steady state time increment used at the end of the transients.

These steady state periods are not included in Table 1; the analyst will use a value greater than or equal to the largest steady time increment from the referenced document.At the inside surface of the RPV, the Region A temperature from the reactor thermal cycle diagram[10, Attachlnent 1, p. 2] shall be applied. Table 1 also includes these values as TRPv.FileNo.: VY-19Q-301 Page 4 of 18 Revision:

0 F0306-O1RO V Structural Integrity Associates, Inc.Table 1: Transients Time, FW Transient sec T- ... F TRPV, OF P, psig Flow, % Cycles 1. Boltup 0 70 70 0 0% 123 2. Design Hydrotest 0 70 70 0 0% 120 1080 100 100 0 0%1680 100 100 1100 0%5280 100 100 1100 0%5880 100 100 50 0%3. Startup 0 100 100 50 0% 300 16164 549 549 1010 0%4. Turbine Roll and 0 549 549 1010 0% 300 Increased to Rated 1 100 549 1010 25%*Power 1801 100 549 1010 25%*1802 260 549 1010 25%*3602 392 .549 1010 100%5. Daily Reduction 0 392 549 1010 100% 10,000 75% Power 900 310 549 1010 100%*2700 310 549 1010 100%*3600 392 549 1010 100%6. Weekly Reduction 0 392 549 1010 100% 2,000 50% Power 1800 280 549 1010 100%*3600 280 549 1010 100%*5400 392 549 1010 100%9. Turbine Trip at 0 392 549 1010 100% 10 25% Power 1800 265 549 1010 100%1980 265 549 1010 25%*2340 90 549 1010 25%*2520 90 549 1010 25%*3420 265 549 1010 25%*3600 265 549 1010 100%5400 392 549 1010 100%10. FW Heater 0 392 549 1010 100% 70 Bypass 90 265 549 1010 100%1890 265 549 1010 100%2070 392 549 1010 100%File No.: VY-19Q-301 Revision:

0 Page 5 of 18 F0306-01RO Structural integrity Associates, Inc.Transient 11. Loss of FW Pumps Time, sec 0 1 3.5 4.5 13.5 184.5 1564.5 1565.5 2165.5 2166.5 2346.5 5406.5 5407.5 6727.5 6728.5 7148.5 7448.5 11048.5 16411.5 16412.5 18212.5 18213.5 20013.5 20014.5 21814.5 T,..., OF 392 565 565 50 50 50 440 565 565 50 50 440 549 565 50 50 300 400 549 549 549 100 100 260 392 TRPV, "F 549 565 565 565 565 565 565 565 565 565 532 549 549 565 565 502 502 400 549 549 549 549 549 549 549 P, psig 1010 1010 1190 1184.5 1135 1135 1135 1135 1135 1135 885 1055 1055 1135 1135 675 675 232 885 1010 1010 1010 1010 1010 1010 FW Flow, %100%0%0%40%40%40%0%0%0%40%*40%*0%0%0%25%*25%*0%0%0%0%0%25%*25%*25%*100%Cycles 10 12/13/15.

Turbine 0 392 549 1010 100% 289 Generator Trip, 10 392 565/600**

1135/1375**

100%Reactor Overpressure, 15 392 565/600**

1135/1375**

100%Other SCRAMs 30 392 539 940 100%90 275 539 940 25%*990 100 539 940 25%*2790 100 539 940 25%*2791 260 539 940 100%3210 291 549 1010 100%4591 392 549 1010 100%14. SRV Blowdown 0 392 549 1010 100% 1 60 275 531.6 885 100%960 100 365 50 25%*19. Reduction to 0% 0 392 549 1010 100% 300 Power 1800 265 549 1010 25%*20. Hot Standby 0 265 549 1010 25%* 300 (Heatup Portion) 1 440 549 1010 0%3925 549 549 1010 0%20A. Hot Standby 0 549 549 1010 0% 300 (FW Injection Portion) 1 100 549 1010 25%181 100 549 1010 25%241 290 549 1010 0%451 549 549 1010 25%21-23. Shutdown 0 549 549 1010 25%* 300 6264 375 375 50 25%*6864 330 330 50 25%*15144 100 100 .50 0%File No.: VY-19Q-301 Revision:

0 Page 6 of 18 F0306-01RO.

Structural Integrity Associates, Inc.Time, FW Transient sec T OF TRPV, OF P, psig Flow, % Cycles 24. Hydrostatic Test 0 100 100 50 0% 1 600 100 100 1563 0%1200 100 100 1563 0%1800 100 100 50 0%25. Unbolt 0 100 100 0 0% 123 1080 70 70 0 0%* Flow rate is conservatively rounded up to one of the three flow rates considered (25%, 40%, 100%).** The second value applies for one cycle; the first value applies for the rest of the cycles.3.4 Heat Transfer Coefficients, Condensation When steam floods a relatively cold component, the steam condenses on the component surface.Holman [5, p. 413] gives the following equation for average heat transfer coefficient:

h = 0.555 {p(p -pv)gk 3 h'fg/[IID(Tg

-T)]} /' , where p = mass density of liquid, Pv = mass density of vapor, g = acceleration of gravity, k = conductivity of liquid at average temperature, h'fg = hg + 0.68c(Tg -Tw), hfg = heat of condensation at vapor temperature, c = specific heat of liquid at average temperature, Tg = saturated vapor temperature

= Tfmal, T= pipe inner wall temperature

= Tinitial,ý= viscosity of liquid at average temperature D = inner diameter of pipe The portion of the equation inside the brackets, p(p -pv)gk 3 h'fg/[pD(Tg

-T,)], has the following units: (fii)2 (seC 2)(Btu)3 (hr-ft-OF)3 (Btu)(ft-tr)(OF)(ft-hr)obfn it-)(OF)(BtU 4)(36002 sec)(he)12960000 Btu 4 hr 4_ft 8_OF 4 After taking the fourth root, this becomes 60 Btu/hr-ft 2 -F. Steam properties are interpolated at Tg, and water properties are interpolated at Tf, which is taken as the average of Tg and T,. Then, h'fg and heat transfer coefficient h are calculated for each set of steam properties, water properties, Tg and TT.Tables 2 and 3 list selected properties of liquid water [12, Table 1-8] and saturated steam [13], respectively.

File No.: VY-19Q-301 Revision:

0 Page 7 of 18 F0306-01RO VStructural Integrity Associates, Inc.Table 2: Properties of Liquid Water T, OF*300 400 500 600 P, Ibm/ft 3 57.3 53.6 49.0 42.4 C, Btu/lbm-OF 1.03 1.08 1.19 1.51 Ii, Ibm/ft-hr 0.468 0.335 0.252 0.208 v, ft 2/sec 2.27E-06 1.74E-06 1.43E-06 1.37E-06 k, Btu/hr-ft-°F 0.395 0.382 0.349 0.293 Table 3: Properties of Saturated Steam T,, °F v2, ft 3/Ibm hf, Btu/Ibm 545 0.4449 649.6 550 0.4249 641.6 565 0.3703 616.4 3.5 Heat Transfer Coefficients, Forced Flow and Natural Circulation Table 4 summarizes the force flow and natural circulation heat transfer coefficients to be used in the analysis [3, Section 3.2.1 ]. For each flow rate, values are taken at 300'F as in the previous analysis.These values are within 11% of the maximum values for a given flow rate, and are more than 30%greater than the minimum values for a given flow rate [3, Table 4] [4, Tables 4 and 5]. Therefore, the use of heat transfer coefficients at 300'F is bounding for the most severe transients, which occur at a wide range of temperatures.

Figure 1 illustrates the heat transfer coefficient regions [4, Figure 6].Table 4: Forced Flow and Natural Circulation Heat Transfer Coefficients, Btu/hr-ft 2 -F 0% flow, Region 100% flow 40% flow 25% flow water 1 3705 1780 1222 144 2 * * * *3 1489 743 504 109 4 * * * *5 177 89 60 12 6 * * * *7 864 864 864 864 8 0.2 0.2 0.2 0.2* Linearly transition between the values for the adjacent regions.File No.: VY-19Q-301 Revision:

0 Page 8 of 18 F0306-O1RO r Structural Integrity Associates, Inc.Region 7 Region B;-F Region I 0 Region 6 Region 5 A B C 0 EýNotes: 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.Beginning of annulus = Node 252.Beginning of thermal sleeve transition

= approximately 4.0" from Point A = Node 294.End of thermal sleeve transition

= approximately 9.5" from Point A = Node 387.End of inner nozzle corner (nozzle side) = Node 553.End of inner nozzle corner (vessel wall side) = Node 779.Figure 1: Nozzle and Vessel Wall Thermal and Heat Transfer Boundaries

3.6 Finite

Element Model The ANSYS program [6] will be used to perform the finite element analysis.

A previously-developed axisymmetric model will be used [4, file FW.INP], except that temperature-dependent material properties will be used. Table 5 shows the applicable material properties

[14].File No.: VY-19Q-301 Revision:

0 Page 9 of 18 F0306-01RO V Structural Integrity Associates, Inc.Table 5: Temperature-Dependent Material Properties Mean Young's Coefficient of Conductivity, Diffusivity, Specific Heat, Materil Tempera- Modulus, Thermal k c Maeil Description 106 d No. ture, IF E x 10 Expansion, (BTU/hr-ft-°F) 2 (BTU/Ibm-°F)(psi) a X 10.6 (see Note 1)' (see Note 5)(in/in-0 F)I SA533 Grade B, 70 27.8 6.4 23.5 0.458 0.105 A508 Class II 200 27.1 6.7 23.6 0.425 0.114 (see Note 2) 300 26.7 6.9 23.4 0.401 0.119 400 26.1 7.1 23.1 0.378 0.125 500 25.7 7.3 22.7 0.356 0.130 600 25.2 7.4 22.2 0.336 0.135 2 SS Clad 70 28.3 8.5 8.6 0.151 0.116 (see Note 3) 200 27.6 8.9 9.3 0.156 0.122 300 27.0 9.2 9.8 0.160 0.125 400 26.5 9.5 10.4 0.165 0.129 500 25.8 9.7 10.9 0.170 0.131 600 25.3 9.8 11.3 0.174 0.133 3 A508 Class 1 70 29.3 6.4 35.1 0.695 0.103 (see Note 4) 200 28.6 6.7 33.6 0.613 0.112 300 28.1 6.9 32.3 0.561 0.118 400 27.5 7.1 30.9 0.512 0.123 500 27.1 7.3 29.5 0.472 0.128 600 26.5 7.4 28.0 0.433 0.132 4 A 106 Grade B 70 29.3 6.4 35.1 0.695 0.103 (see Note 4) 200 28.6 6.7 33.6 0.613 0.112 300 28.1 6.9 32.3 0.561 0.118 400 27.5 7.1 30.9 0.512 0.123 500 27.1 7.3 29.5 0.472 0.128 600 26.5 7.4 28.0 0.433 0.132 Notes: 1. Convert to BTU/sec-in-0 F for input to ANSYS.2.3.4.5.Properties of A508 Class II are used (3/4Ni-I/2Mo-1/3Cr-V).

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

= C-Si; k and d for plain carbon steel are used [11].Calculated as [k/(pd)]/12 3.Stresses will be extracted and linearized at two locations, both on the inside surface. The critical safe end location is Node 192, which has the highest stress intensity due to thermal loading under high flow conditions

[3, Section 4.0 and Figures 6 and 7]. The corresponding linearization path is from Node 192 to Node 187 (Figure 2 [3, Figure 7]).The critical nozzle corner location is Node 657 at the base metal of the nozzle, chosen based upon the highest pressure stress [3, Section 4.0 and Figures 8 and 9]. The corresponding linearization path is from Node 657 to Node 645 (Figure 3 [3, Figure 9]).File No.: VY-19Q-301 Revision:

0 Page 10 of 18 F0306-01RO Structural Integrity Associates, Inc.ELEMENTS MAT NUN ANSYS 8. IAI MAR 19 2007 13:25:09 Node 187 Feedwater Nozzle Finite Element Model Figure 2: Safe End Linearization Path Figure 3: Nozzle Corner Linearization Path File No.: VY-19Q-301 Revision:

0 Page 11 of 18 F0306-OIRO Structural Integrity Associates, Inc.3.7 Nozzle Corner Effects The axisymnetric model has the effect of modeling the cylindrical RPV as spherical.

To partially counter the resulting reduction in stress in the RPV wall, the radius in the model was increased by a factor of 1.5 [3, p. 8]. This yields a general membrane stress that equals the average of the hoop and axial stress for the cylinder.Stresses from the axisymmetric analysis will need to be increased to account for the three-dimensional (3-D) geometry.

A factor of 1.333 has been established in a previous calculation package that modeled the nozzle [3, p. 9], to achieve an overall pressure multiplication of 2.0. This is consistent with the maximum value used in prior VYNPS analyses [7, Appendix A, p. 4-10].No other SCF is required at the nozzle corner inside surface, since this location has no stress riser.3.8 Piping Interface Loads The previous analysis of the FW nozzle calculated membrane axial and shear stresses due to the piping interface loads by closed form solution, then combined them into stress intensities for the two locations of interest [2, Section 3.4]. All shear stresses were treated as existing in the same plane.In this analysis, the stress components are recalculated in Section 4.3 taking into account through-wall distribution.

Forces and moments are taken from the same reference as before [8, Table 3].3.9 SCFs, Safe End In the previous analysis, an SCF of 1.34 was used for the safe end location for all load conditions

[2]. That value was obtained from the original design basis evaluation for the FW nozzle. For the current analysis, the SCF is updated to reflect modem-day ASME Code fatigue usage analysis methodology for consistency with the rest of the evaluation.

At the safe end inside surface, guidance is taken from the piping analysis rules in Subarticle NB-3600 of Section III of the ASME Code [1]. These rules specify stress indices C 1 , C 2 , and C 3 , which are applied to nominal stress to yield primary plus secondary membrane plus bending stress (P+Q);and K 1 , K 2 , and K 3 , which are applied to nominal stress along with the C factors to yield total stress (P+Q+F). The subscripts indicate the type of loading: I for pressure, 2 for moments, and 3 for thermnal transients.

Stress indices for a reducer are used.Section 4.3 contains calculations of the safe end SCFs. For stresses due to piping loads, the moment stress indices C 2 and K 2 are applied to the nominal stress components at the safe end. For pressure stresses, the ANSYS model is sufficient to account for the effects of gross structural discontinuity such that C 1 is not needed. To account for the effects of local structural discontinuity, K 1 is applied to the linearized P+Q stress to yield P+Q+F. These factors are conservatively applied to all six components of the stress tensor.For thenrmal stresses, C 3 and K 3 are given as 1.0 [1, Table NB-3681 (a)-1]; therefore, no SCF is required.File No.: VY-19Q-301 Page 12 of 18 Revision:

0 F0306-01RO Structural Integrity Associates, Inc.3.10 Environmental Fatigue Multipliers The environmental fatigue multipliers for the safe end and nozzle corner will be calculated in accordance with NUREG/CR-6583 methodology

[15].4.0 CALCULATIONS 4.1 Heat Transfer Coefficients, Condensation Condensation heat transfer coefficients are calculated with the formula shown in Section 3.4 for times during which the nozzle is filled with steam at Region A temperature

[10, Attachnment 1, p. 3].This is done in the sheet labeled "Condensation" in Excel workbook VY-19Q-30.xls in the project computer files. The highest heat transfer coefficient values for the transient temperature range are used. These are provided in Table 6.Table 6: Condensation Heat Transfer Coefficients, Btu/hr-ft 2-°F 0% flow, Region steam 1 598 2 *3 1515 4 *5 874 6 *7 **8 *** Linearly transition between the values for the adjacent regions.** Use values from Table 4, since these are bounding and there is no change in temperature.

4.2 Piping

Interface Loads From general structural mechanics, the membrane plus bending stresses at the inside surface of a thick-walled cylinder are: ca, = axial stress due to axial force = Fz/A u,2 = axial stress due to bending moment = Mxy(ID/2)/I Gz = Gzl + Cyz2-co = shear stress due to torsion = M,(ID/2)/J cr = shear stress due to shearforce

= 2Fxy/A, where F,,, Fy, Fz, Mx, My, and Mz are forces and moments at the pipe-to-safe end weld MxL = moment about x axis translated by length z = -L = Mx -Fy L MyL = moment about y axis translated by length z = -L = My + F, L Mxy = resultant bending moment = (MxL2 + MyL 2)0.5 Fxy = resultant shear force = (Fx 2 + Fy 2)0 5 /ID, OD = inside and outside diameters A = area of cross section = (7r/4)(OD 2 -ID 2)FileNo.: VY-19Q-301 Page 13 of 18 Revision:

0 F0306-01RO Structural Integrity Associates, Inc.I = moment of inertia = (ni/64)(OD 4 -ID 4)J = polar moment of inertia = (7t/32)(OD 4 -ID 4)Figure 4 shows the coordinate system for the forces and moments [8, Figure 1 ]. The shear stresses are expressed in a local coordinate system with r radial (X in ANSYS coordinates), 0 circumferential (Z in ANSYS coordinates), and Z axial (Y in ANSYS coordinates).

Table 7 shows the calculation of stresses; ID, OD, and L are taken firom the previous piping load stress calculations

[2, Section 3.4].Forces and moments are taken from the same reference as before, except that signs are chosen to maximize' stress [8, Table 3].Y tMZ r~FY, Figure 4: Coordinate System for Forces and Moments Table 7: Membrane Plus Bending Stresses Due to Piping Loads Safe Nozzle End Corner F,, kip 3.00 3.00 Fy, kip -15.00 -15.00 F,, kip 3.20 3.20 M., kip-in 336.00 336.00 MY, kip-in 156.00 156.00 M,, kip-in 480.00 480.00 L, in 12.09 27.57 MKL, kip-in 517.31 749.58 MyL, kip-in 192.26 238.72 M'y, kip-in 551.88 786.67 F,, kip-in 15.30 15.30 OD, in 11.86 22.67 ID, in 10.409 10.750 A,in 2 25.28 312.73 I, in 4 393.28 12300.41 J, in 4 786.55 24600.82 7, 1 , ksi 0.127 0.010 cy3,, ksi 7.304 0.344 a,, ksi 7.430 0.354 t~, ksi 3.176 0.105ksi 1.210 0.098 FileNo.: VY-19Q-301 Page 14 of 18 Revision:

0 F0306-01RO Structural Integrity Associates, Inc.4.3 SCFs, Safe End Figure 5 shows the geometry parameters used in calculating stress indices for reducers [1, Figure NB-3683.6-1], and Figure 6 shows the feedwater nozzle safe end geometry [9]. Comparing the two figures gives the following values: Ll = 0", r, = 0.75", D 1 = 12.000", t = (12 -10.515)/2

= 0.7425" L2 = 0", r 2 = 0.75", D 2 = 10.840", t 2 = (10.840 -9.669)/2 = 0.5855" ao= 10°(LI and L 2 are taken as zero because the location of interest is on the radius of curvature.)

Figure 5: Reducer Geometry Parameters FileNo.: VY-19Q-301 Revision:

0 Page 15 of 18 F0306-OIRO Structural Integrity Associates, Inc.-4.4 I Figure 6: FW Nozzle Safe End Geometry FileNo.: VY-19Q-301 Revision:

0 Page 16 of 18 F0306-OIRO Structural Integrity Associates, Inc.Equations for stress indices are taken from the ASME Code [1, NB-3683.6].

For K 1 and K 2 , since the location of interest is not on a weld, the equation for flush welds is used: KI = K 2 = 1.1 -0.1 Lm,/(Dln tn)0'5 , Where Lml/(Din tin)0 5 = the lesser of L1/(Di ti)'5 and L 2/(D 2 t 2)0'5 Since Li = L2 = 0, one finds: Ki = K 2 = 1.1 -0.1 (0) = 1.1 Since rl and r 2 are less than 0.1DI, C 2 is given as: C 2 = 1.0 + 0.0185 a (Dn/tn)0 5 , where Dn/tn = the larger of D 1/tl and D 2/t 2 The bounding D/t value is D 2/t 2 = 10.840/0.5855

= 18.514,\so that: C 2 = 1.0 + 0.0185 (10) (18.514)0.5

= 1.796 C 2 K 2 = 1.796 (1.1) = 1.976 5.0 RESULTS OF ANALYSIS This calculation package specifies the ASME Code edition, finite element model, thermal and pressure transients (Table 1), and heat transfer coefficients (Tables 4 and 6) to be used in a fatigue usage analysis of the FW nozzle at VYNPS. Thermal transient and pressure stress components will be calculated using ANSYS, and piping load stress components are calculated herein using closed fonn solutions (Table 7).Linearized stress components at Nodes 192 (safe end inside surface) and 657 (nozzle comer inside surface) will be used for the fatigue usage analysis.

At the nozzle comer, P+Q and P+Q+F pressure stress components will be increased by a factor of 1.333. For the nozzle corner location, the stresses used in the evaluation shall be for the base metal only; that is, the cladding material should be unselected prior to stress extraction.

At the safe end, linearized P+Q pressure stress components will be multiplied by 1.1 to yield P+Q+F pressure stress components, and nominal stress components due to piping loads are multiplied by 1.796 to yield P+Q stress components and 1.976 to yield P+Q+F stress components.

The fatigue usage analysis will consider all six stress components, and will be performed using the NB-3200 rules of Section III of the ASME Code [1]. Calculated fatigue usage factors will be multiplied by the overall Fen of 1.74 for the safe end .[2, Section 5.0] and values to be developed in a subsequent calculation package, to be assigned file number VY-19Q-303, for the nozzle corner.FileNo.: VY-19Q-301 Page 17 of 18 Revision:

0 F0306-O1RO Structural Integrity Associates, Inc.

6.0 REFERENCES

1. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Subsection NB, 1998 Edition with Addenda through year 2000.2. SI Calculation Package, Fatigue Analysis of Feedwater Nozzle, Revision 0, SI File No. VY- 16Q-302.3. S1 Calculation Package, Feedwater Nozzle Stress Histoiy Development for Greens Functions; Revision 0, SI File No. VY-16Q-301.

4. SI Calculation Package, Feedwater Nozzle Finite Element Model and Heat Transfer Coefficients, Revision 0, SI File No. VY-IOQ-301.

5. Holman, J.P., Heat Transfer, Fifth Edition, McGraw-Hill, 1981.6. ANSYS, Release 8.1 (w/Service Pack 1), ANSYS, Inc., June 2004. (Listed for reference only;this program is not used in this calculation package.)7. Entergy Document VYC-378, Revision 0, Vermont Yankee Reactor Cyclic Limnits for Transient Events, SI File No. VY-05Q-21 1.8. GE Drawing No. 919D294, Revision 11, Sheet 7, Reactor Vessel, Spec. Contrbl, SI File No. VY-05Q-241.9. Ebasco Drawing 5920-234R1, 08/03/67, Safe End Detailfor Nozzles MK. N4A THRUN4D, (CB&I Contract 9-6201, Drawing #M14, Revision 0, 05/02/67), SI File No. VY-05Q-215.

10. Entergy Document EC No. 1773, Revision 0 (Design Input Revision 1), Environmental Fatigue Analysis for Vermont Yankee Nuclear Power Station, SI File No. VY- 16Q-209.11. Letter MLH-08-001 from M.L. Herrera (SI) to N. Lobo (ASME), ASME Code,Section II, Part D, 1998 Edition and Later, Subpart 2, Table TCD, January 10, 2008.12. Cheremisinoff, N., Heat Transfer Pocket Handbook, Gulf Publishing Co., Houston, 1984.13. Keenan, J.H., Keyes, F.G., Hill, P.G., Moore, J.G, Steam Tables, Thermodynamic Properties of Water Including Vapor, Liquid, and Solid Phases (English Units), John Wiley & Sons, 1969.14. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section II, Part D, 1998 Edition with Addenda through year 2000.15. NUREG/CR-6583 (ANL-97/18), Effects ofL WR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels, March 1998.File No.: VY-19Q-301 Page 18 of 18 Revision:

0 F0306-O1RO Exhibit B Structural Integrity Associates, Inc. File No.: VY-19Q-302 CALCULATION PACKAGE Project No.: VY-19Q PROJECT NAME;Provide VY Support for Questions Related to Environmental Fatigue Analyses CONTRACT NO.: 10163217 CLIENT: PLANT: Entergy Nuclear Operations, Inc. Vermont Yankee Nuclear Power Station CALCULATION TITLE: ASME Code Confirmatory Fatigue Evaluation of Reactor Feedwater Nozzle Document Affected Project Manager Preparer(s)

&Revision Pages Revision Description Approval Checker(s), Signature

& Date Signatures

& Date 0 1-12 OriginalIssue Computer Files -w (TI\Herann WF We".'-TJI "01 CJ2008 WFW 01/30/2008

.." L ohse CSL 01/30/2008 Page 1 of 12 F0306-01RO

-j V Structural Integrity Associates, Inc.Table of Contents 1.0 2.0 3.0 4.0 5.0 6.0 7.0 OBJECTIVE

.................................................................................................................................

3 M ETHODOLOGY

...............................................................................................................

3 DESIGN INPUTS .........................................................................................................................

3 CALCULATION S ........................................................................................................................

9 RESULTS OF ANALY SIS .........................................................

...............................................

10 CON CLU SION S AN D DISCU SSION S ....................................................................................

11 REFEREN CES ...........................................................................................................................

12 List of Tables Table 1: Transients as Input to VESLFAT .....................................................................................

7 Table 2: Temperature-Dependent Material Properties, VESLFAT ..................................................

8 Table 3: Carbon/Low Alloy Steel Fatigue Curve ............................................................................

8 Table 4: Stress Components Before SCF, psi ...................................................................................

9 Table 5: Stress Components W ith SCF, psi ......................................................................................

9 Table 6: Fatigue Usage Results for Safe End ................................................................................

10 Table 7: Fatigue Usage Results for Nozzle Corner .......................................................................

II List of Figures Figure 1: ANSYS Finite Element Model ........................................................................................

4 Figure 2: Linearization Paths ...........................................................................................................

5 File No.: VY-19Q-302 Revision:

0 Page 2 of 12 F0306-OIRO Structural Integrity Associates, Inc.1.0 OBJECTIVE The objective of this calculation package is to perform an ASME Code,Section III fatigue usage calculation for the reactor pressure vessel (RPV) feedwater (FW) nozzle at Vermont Yankee Nuclear Power Station (VYNPS).2.0 METHODOLOGY The methodology to be used for this evaluation was established ina previous calculation package[1], and is summarized herein. A previously-developed finite element model (FEM) is analyzed using the ANSYS program [2]. Thermal transient analysis is performed for each defined transient, and the thermal stresses are added to stresses due to pressure and piping loads, which are scaled based on the magnitudes of the pressure and piping loads. Stress concentration factors (SCFs) are applied as appropriate.

All six components of the stress tensor are used for stress calculations.

The fatigue calculation is performed at previously-examined locations, and uses the methodology of Subarticle NB-3200 of Section III of the ASME Code [3]. Environmental fatigue usage analysis will be performed in a separate calculation package.3.0 DESIGN INPUTS 3.1 Finite Element, Analysis A previous calculation package specifies all design input [1]. The FEM input file is taken from the previous analysis of the FW nozzle [4, file FW.INP], and modified to/include temperature-dependent properties

[1, Table 5]. The modified file is named FW-GEOM.INP, and is used as input to the files in which the thermal transient and stress analyses are performed.

Figure 1 shows the FEM [4, Figure 4].For the thermal transient ANSYS analysis, previously defined therimal transients

[1, Table 1] are analyzed, applying heat transfer coefficients

[1, Tables 4 and 6] as appropriate based on flow rate.Bounding reactor temperature is used for Transients 12/13/15 [1, Table 1], called Transient 13 herein. (In VESLFAT, Transient 13 is run separately since it has a higher reactor pressure.)

For ramps during which the flow rate undergoes a ramp change [5, Attachment 1, p. 3], the set of heat transfer coefficients with the largest values is used. This is done because ANSYS always applies changes to the heat transfer coefficients as step changes, even if the temperature undergoes a ramp change.Note that, for three time periods during Transient 11 [1, Table 1], the nozzle is filled with steam at Region A temperature

[5, Attachment 1, p. 3, Note 1], such that heat transfer coefficients for condensation apply [1, Table 6]. Since it takes a finite amount of time for the water to drain and condensation to begin, the condensation heat transfer coefficients are not applied until the load step after the Region A temperature is reached.Stress analysis is perfonned using the temperature distributions calculated in the thermal transient ANSYS analysis as input. At the vessel wall, Y displacement is set to zero, and X displacement is File No.: VY-19Q-302 Page 3 of 12 Revision:

0 F0306-O1RO

" Structural Integrity Associates, Inc.unconstrained, as was previously done [6, Figure 4]. At the FW pipe, Y displacement is coupled to account for the adjacent piping, as was previously done [6, files FWS VY 25.INP, FWS_ VY 4_0.INP, and FWS VY _O0.INP].

Figure 1 shows the locations of these boundary conditions.

ELEMENTS SEP 6 2002 16:23:51 Y=O Feedwater Nozzle Finite Element Model Figure 1: ANSYS Finite Element Model All ANSYS input files, listed below, are saved in the project computer files: FW-GEOMINP:

Geometry and material properties FW-HTBC.INP:

Set heat transfer boundary conditions TRANO2-TINP, TRANO2-S.INP:

TRANO3-T.INP, TRANO3-S.INP:

TRANO4-.T.INP, TRANO4-S.1NP:

TRANO5-TINP, TRANO5-S.INP:

TRANO6-T.INP, TRANO6-S.INP:

TRANO9-T.INP, TRANO9-S.INP:

TRAN1O-T.INP, TRAN1O-S.INP:

TRAN]1-T.INP, TRAN11-S.INP:

TRAN13-T.INP, TRAN13-S.INP:

Transient 2, thermal and stress analysis Transient 3, thermal and stress analysis Transient 4, thermal and stress analysis Transient 5, thermal and stress analysis Transient 6, thermal and stress analysis Transient 9, thernal and stress analysis Transient 10, thermal and stress analysis Transient 11, thermal and stress analysis Transient 12/13/15, thermal and stress analysis File No.: VY-19Q-302 Revision:

0 Page 4 of 12 F0306-O1RO

! Structural Integrity Associates, Inc.TRAN14-T.INP, TRAN14-S.INP:

Transient 14, thermal and stress analysis TRAN19-T.INP, TRAN19-S.INP:

Transient 19, thermal and stress analysis TRAN20-T.INP, TRAN20-S.INP:

Transient 20, thernal and stress analysis TRAN2OAT.INP, TRAN2OAS.INP:

Transient 20A, thermal and stress analysis TRAN21-T.INP, TRAN21-S.INP:

Transient 21, thernal and stress analysis TRAN25-T.INP, TRAN25-S.INP:

Transient 25, thermal and stress analysis 3.2 Stress Calculation Linearized stress components at Nodes 192 (safe end inside surface) and 657 (nozzle comer inside surface) are used for the fatigue usage analysis [1, Section 3.6], as shown in Figure 2 [6, Figures 7 and 9]. For the nozzle comer location, the stresses used in the evaluation are for the base metal only;that is, the cladding material is unselected prior to stress extraction.

The stress components from the thermal stress analyses are combined with stress components due to pressure and piping loads. A unit pressure stress analysis was performed using ANSYS in a previous calculation package [6], and stress component results are taken from that analysis [6, files PSE. OUT and PBLEND. OUT]. Piping load stress components are taken from previous calculations using closed form solutions

[ 1, Table 7].Node 187 Node 657 Figure 2: Linearization Paths SCFs are applied to the pressure and piping load stress components to yield prilnary plus secondary membrane plus bending stress components (P+Q) and the total (primary plus secondary plus peak)stress components (P+Q+F) as specified in the methodology calculation package [1].File No.: VY-19Q-302 Revision:

0 Page 5 of 12 F0306-O1RO

! Structural Integrity Associates, Inc.3.3 Fatigue Usage Analysis, General The VESLFAT program [7] is used to perforn the fatigue usage analysis in accordance with the fatigue usage portion of NB-3200 [3]. VESLFAT performs the analysis required by NB-3222.4(e)

[3] for Service Levels A and B conditions defined by the user. The VESLFAT program computes the primary plus secondary and total stress ranges for all events and performs a correction for elastic-plastic analysis, if appropriate.

The program computes the stress intensity range based on the stress component ranges for all event pairs [3, NB-3216.2].

The program evaluates the stress ranges for primary plus secondary'and primary plus secondary plus peak stress based upon six components of stress (3 direct and 3 shear stresses).

If the primary plus secondary stress intensity range is greater than 3Sin, then the total stress range is increased by the factor K,, as described in NB-3228.5

[3]. The value of S,, is specified as a function of temperature.

The input maximum temperature for both states of a load set pair is used to determine the temperature upon which S,, is determined from the user-defined values.When more than one load set is defined for either of the event pair loadings, the stress differences are determined for all of the potential loadings, saving the maximum for the event pair, based on the pair producing the largest alternating total stress intensity (SaIt), including the effects of Ke. The principal stresses for the stress ranges are deternmined by solving for the roots of the cubic equation: S3 _ (a5x +- O-y -Y + 'S2 +t (Cyx CT Y- GYz -a- +' a, -"Cxy 2 -X "x2 _ TYz 2 )S-(Gx Cy az + 2 txy txzt yz -3z -O'y 2 ,z- _ ayz 2 ) = 0 The stress intensities for the event pairs are reordered in decreasing order of Salt, including a correction for the ratio of modulus of elasticity (E) from the fatigue curve divided by E from the analysis.

This allows a fatigue table to be created to eliminate the number of cycles available for each of the events of an event pair, allowing determination of fatigue usage per NB-3222.4(e)

[3].For each load set pair in the fatigue table, the allowable number of cycles is determined based on Salt.For the VYNPS FW nozzle analysis, transients that consist of both upward and downward temperature and pressure ramps are split so that each successive ramp is treated separately.

Table 1 shows the transients as input to VESLFAT [1, Table 1]. The numnbers of cycles in Table 1 are entered in VESLFAT input files VFA T-1. CYC (safe end) and VFA T-2I. CYC (nozzle corner).File No.: VY-19Q-302 Page 6 of 12 Revision:

0 F0306-01RO Structural Integrity Associates, Inc.Table 1: Transients as Input to VESLFAT VESLFAT Start Time, Temp. Pressure Load Set Transient sec** Change Change Cycles I IBoltup 0 None None *2 2_DesHydrol 0 Upward Upward 120 3 2_DesHydro2 5280 None Downward 120 4 3_Startup 0 Upward Upward 300 5 4 TurbRolll 0 Downward None 300 6 4_TurbRohl2 1801 Upward None 300 7 5_DailyRedl 0 Downward None 10,000 8 5_DailyRed2 2700 Upward None 10,000 9 6_WklyRedl 0 Downward None 2,000 10 6_Wk]yRed2 3600 Upward None 2,000 11 9_TurbTripl 0 Downward None 10 12 9_TurbTrip2 2520 Upward None 10 13 10_FWHBypl 0 Downward None 70 14 10_FWI-HByp2 1890 Upward None 70 15 11_LoFP1 0 Upward Upward 10 16 11 LoFP2 3.5 'Downward Downward 10 17 11_LoFP3 184.5 Upward None 10 18 11 LoFP4 2165.5 Downward Downward 10 19 11_LoFP5 2346.5 Upward Upward 10 20 11 LoFP6 6727.5 Downward Downward 10 21 11_LoFP7 7148.5 Upward Downward 10 22 11_LoFP8 11048.5 Upward Upward 10 23 11 LoFP9 18212.5 Downward None 10 24 11_LoFP1O 20013.5 Upward None 10 25 12_TGTripl 0 None Upward 288 26 12_TGTrip2 15 Downward Downward 288 27 12_TGTrip3 2790 Upward Upward 288 28 13_Overpri 0 None Upward 1 29 13_Overpr2 15 Downward Downward 1 30 13_Overpr3 2790 Upward Upward 1 31 14 SRVBIwdn 0 Downward Downward 1 32 19_RedTo0pct 0 Downward None 300 33 20_HSHeatup 0 Upward None 300 34 20AHSFWInj 1 0 Downward None 300 35 20AHSFWInj2 181 Upward None 300 36 21 Shutdown 0 Downward Downward 300 37 24_HydroTestl 0 None Upward I 38 24_HydroTest2 1200 None Downward 1 39 25 Unbolt 0 Downward None 123* Since this transient does not affect the FW nozzle, it is not considered in the cyclic evaluation.

• • Note that stress peaks may occur after the start of the subsequent ramp.3.4 Material Properties, VESLFAT Material properties are entered in VESLFAT input files VFAT-IJ.FDT (safe end) and VFAT-2I.FDT (nozzle corner). Table 2 lists the temperature-dependent material properties used in the analysis [ 1, Table 5] [8], and Table 3 lists the fatigue curve for the nozzle and safe end materials

[3, Appendix I, Table 1-9.1 and Figure 1-9.1 ]. VESLFAT automatically scales the stresses by the ratio of E on the fatigue curve to E in the analysis, for purposes of determining allowable numbers of cycles, as required by the ASME Code.File No.: VY-"19Q-302 Page 7 of 12 Revision:

0 F0306-01RO Structural Integrity Associates, Inc.Other material properties are input as follows: in = 3.0, n = 0.2, parameters used to calculate factor Ke, safe end [9]m = 2.0, n = 0.2, parameters used to calculate factor Ke, nozzle comer [9]E from fatigue curve = 30,000 ksi [3, Appendix I, Table 1-9.1 and Figure 1-9.1] [9]Table 2: Temperature-Dependent Material Properties, VESLFAT Material A508 Class I (safe end)T, OF 70 200 300 400 500 600 70 200 300 400 500 600 E, psi 29.3(10)6 28.6(10)6 28.1(10)6 27.5(1 0)6 27.1(10)6 26.5(10)6 27.8(10)6 27.1(10)6 26.7(10)6 26.1(10)6 25.7(10)6 25.2(10)6 Sm, ksi 23.3 21.9 21.3 20.6 19.4 17.8 26.7 26.7 26.7 26.7 26.7 26.7 Sv, ksi 36.0 33.0 31.8.30.8 29.3 27.6 50.0 47.0 45.5 44.2 43.2 42.1 A508 Class II (nozzle)Table 3: Carbon/Low Alloy Steel Fatigue Curve Number of Cycles 10 20 50 1.00 200 500 1,000 2,000 5,000 10,000 20,000 50,000 100,000 200,000 500,000 1,000,000 Sý, ksi 580 410 275 205 155 105 83 64 48 38 31 23 20 16.5 13.5 12.5 File No.: VY-19Q-302 Revision:

0 Page 8 of 12 F0306-01 RO Structural integrity Associates, Inc.4.0 CALCULATIONS Table 4 contains the stress components at the locations of interest for the 1,000 psi pressure case [6, files PSE. OUT and PBLEND. OUI] and for the piping loads [ 1, Table 7], corresponding to a reactor temperature of 575°F [1, Section 3.1.8].Table 4: Stress Components Before SCF, psi Loading Type Node S. S, Sz SW,. SNZ SV Unit Membrane 192 -810.7 6116 7853 -450.1 0 0 Pressure, plus Bending 657 -705.2 1198 24020 -3590 0 0 1,000 psi Total 657 -705.2 985.5 27590 -121.1 0 0 Piping Loads Nominal 192 0 7430 0 1210 3176 0 at 575°F 657 0 354 0 98 105 0 SCFs are applied to the pressure and piping load stress components to yield P+Q and P+Q+F stress components as follows [1]: Pressure: Safe end (Node 192): Membrane plus bending from ANSYS equals P+Q Membrane plus bending from ANSYS is multiplied by 1.1 to yield P+Q+F Nozzle comer (Node 657): Membrane plus bending from ANSYS is multiplied by 1.333 to yield P+Q Total stress from ANSYS is multiplied by 1.333 to yield P+Q+F Piping Loads: Safe end (Node 192): Nominal stresses are multiplied by 1.796 to yield P+Q Nominal stresses are multiplied by 1.976 to yield P+Q+F Nozzle comer (Node 657): Nominal stresses are used as is for P+Q and P+Q+F Table 5 shows the stress components with SCFs. The piping load stress components are applied as having negative signs, to yield the largest stress component ranges.Table 5: Stress Components With SCF, psi Membrane plus Bending Total'Load Node IS. S& S. S- S_, SI S. S S. S- S. S.-.Pressure 192 -811 6116 7853 -450 0 0 -892 6728 8638 -495 0 0 657 -940 1597 32019 -4785 0 0 -940 1314 36777 -161 0 0 Piping 192 0 13344 0 2173 5704 0 0 14682 0 2391 6276 0 657 0 354 0 98 105 0 0 354 0 98 105 0 The calculations of VESLFAT stress input are automated in Excel workbooks VFA T-1I.XLS (safe end) and VFAT-2I.XLS (nozzle).

These files are organized with sheets labeled as follows:/

• Overview:

Contains general information.

File No.: VY-19Q-302 Page 9 of 12 Revision:

0 F0306-01RO Structural Integrity Associates, Inc." Other Stresses:

Contains calculation of pressure and piping load as shown in Tables 4 and 5.* Rearranger:

There are 16 Re-arranger sheets, one for each transient as analyzed by ANSYS.In these sheets, thermal stresses are copied from Excel workbook StressResults.xls, which contains the results of the ANSYS stress linearization for each transient, and rearranged to conform to VESLFAT input format (including switching the shear stress components S,, and Sy, as required b6y VESLFAT).

Time-varying scale factors for the piping loads (based on FW nozzle fluid temperature) and pressure are determined, and used to scale the unit load stresses, which are then added to the thermal stresses.

Time-varying pressure is also included in the VESLFAT stress input. The VESLFAT stress input also includes time-varying metal temperature, from the ANSYS output, which is used to determine temperature-dependent properties from the values in Table 2.* VESLFAT: Contains the VESLFAT stress input, obtained from sheets named Rearranger.

Load set numbers are entered on this sheet, as defined in Table 1. These sheets are saved to VESLFAT input files VFAT-II.STR (safe end) and VFAT-2.STR (nozzle corner). To avoid double counting of stress states, the initial time steps of each load set before the first stress peak are not included.The files with extension STR are edited if necessary to remove some intermediate stress points, since VESLFAT has a limit of 3,000 total stress states.5.0 RESULTS OF ANALYSIS Tables 6 and 7 give the detailed fatigue usage results for the safe end and the nozzle comer, respectively, from VESLFAT output files VFAT-]I.FAT (safe end) and VFAT-21.FAT (nozzle comer). All VESLFAT input and output files are saved in the project computer files.Table 6: Fatigue Usage Results for Safe End Load Set A Load Set B n Sil, psi KI S,11, psi N Usage 15 11 LoFP1 18 11 LoFP4 10 61435 1.115 57352 2836.19 0.0035.20 11_LoFP6 27 12_TGTrip3 10 49698 1 40800 8098.01 0.0012 27 12 TGTrip3 34 20AHSFWlnjl 278 42194 1 37182 10769 0.0258 30 13_Overpr3 34 20A_HSFWInjl 1 42194 1 37182 10769 0.0001 33 20_HSHeatup 34 20A HSFWInj1 21 42563 1 35966 12060 0.0017 5 4 TurbRolll 33 20_HSHeatup 279 43986 1 35597 12491 0.0223 64 TurbRoll2 23 11 LoFP9 10 39882 1 32197 17579 0.0006 54 TurbRolll 64 TurbRoll2 21 39842 1 32178 17615 0.0012 16 11 LoFP2 35 20AHSFWInj2 10 40708 1 31762 18413 0.0005 35 20AHSFWInj2 37 24 HydroTestl 1 20956 1 13081 664055 0.0000 17 11_LoFP3 35 20AHSFWInj2 10 20399 1 12667 887275 0.0000 19 11 LoFP5 35 20A HSFWInj2 10 19602 1 12135 infinite 0.0000 TOTAL = 0.0571 FileNo.: VY-19Q-302 Revision:

0 Page 10 of 12 F0306-01RO Structural Integrity Associates, Inc.Table 7: Fatigue Usage Results for Nozzle Corner Load Set A 2 2DesHydrol 2 2DesHydrol 2 2DesHydro 1 2 2DesHydrol 2 2DesHydrol 54 TurbRoIll 3 2_DesHydro2 3 2 DesHydro2 3 2DesHydro2 34 20AHSFWInj 1 34 20AHSFWInj 1 26 12_TGTrip2 21 11 LoFP7 26 12_TGTrip2 22 11 LoFP8 4 3 Startup 4 3 Startup 4 3 Startup 4 3 Startup 32 19_RedTo0pct 13 10_FWHfBypl 13 10_FWHBypl 35 20AHSFWInj2 9 6 WklvRedl Load Set B 16 11 LoFP2 2011 LoFP6 18 11 LoFP4 11 9_TurbTripl 54 TurbRolll 39 25 Unbolt 54 TurbRolll 23 11 LoFP9 34 20AHSFWInj 1 38 24_HydroTest2 36 21_Shutdown 36 21 Shutdown 26 12 TGTrip2 31 14 SRVBlwdn 26 12_TGTrip2 26 12_TGTrip2 29 13_Overpr2 28 13_Overprl 32 19_RedTo0pct 33 20 HSHeatup 33 20_HSHeatup 35 20AHSFWInj2 37 24_HydroTestl 35 20A HSFWInj2 n 10 10 10 10 80 123 97 10 13 1 286 14 10 I 10 253 1 1 45 255 45 25 1 274 Sn, PSi 65109 50344 50150 65712 64296 63308 61437 63138 49069 49097 49111 60379 49395 42902 32212 30212 30212 28966 24083 18765 19637 20388 19850 19341 Table 7: Fatigue Usage Results for Nozzle Corner KýSalt, psi 46047 43990 43205 43011 43008 41430 40391 40101 39657 39622 39616 38556 33091 27518 24687 23513 23513 19423 17665 12883 12679 12624 12359 11952 5655.6477.6832.6924.6925.7738..8343.8524.: 8810.'8833.: 8837.9578 160 288: 402: 467: 467: 1111 1564(7618: 87961 91493 infini infini TOTAL N Usage 78 0.0018 19 0.0015 83 0.0015 58 0.0014 97 0.0116 36 0.0159 98 0.0116 36 0.0012 73 0.0015 38 0.0001 88 0.0324.4 0.0015 15 0.0006 31 0.0000 37 0.0002 28 0.0054'.28 0.0000 18 0.0000 02 0.0003 35 0.0003 15 0.0001 34 0.0000 te 0.0000 te 0.0000= 0.0889

6.0 CONCLUSION

S AND DISCUSSIONS A previously-developed FEM was analyzed using the ANSYS program. Thermal transient analysis was performed for each defined transient, and the thermal stresses were added to stresses due to pressure and piping loads, which were scaled based on the magnitudes of the pressure and piping loads. SCFs were applied as appropriate.

All six components of the stress tensor were used for stress calculations.

The fatigue calculation was performed at previously-examined locations, and used the methodology of Subarticle NB-3200 of Section III of the ASME Code.The 60-year CUF for the safe end location was determined to be 0.0571, and the CUF for the nozzle corner location was determined to be 0.0889. Both values are less than the ASME Code allowable value of 1.0.Enviromnental fatigue usage analysis will be performed in a separate calculation package.File No.: VY-19Q-302 Revision:

0 Page 11 of 12 F0306-01RO Structural Integrity Associates, Inc.

7.0 REFERENCES

1. SI Calculation Package, Design Inputs and Methodologyfor ASME Code Confirmatoly Fatigue Usage Analysis of Reactor Feedwater Nozzle, Revision 0, SI File No. VY- 19Q-301.2. ANSYS, Release 8.1 (w/Service Pack 1), ANSYS, Inc., June 2004.3. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Subsection NB, 1998 Edition with Addenda through year 2000.4. SI Calculation Package, Feedwater Nozzle Finite Element Model and Heat Transfer Coefficients, Revision 0, SI File No. VY-IOQ-301.

5. Entergy Document EC No. 1773, Revision 0 (Design Input Revision 1), Environmental Fatigue Analysis/for Vermnont Yankee Nuclear Power Station, SI File No. VY- 16Q-209.6. SI Calculation Package, Feedwater Nozzle Stress History Development for Green Functions, Revision 0, SI File No. VY-16Q-301.

7. VESLFAT, Version 1.42, 02/06/07, Structural Integrity Associates.

8. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section II, Part D, 1998 Edition with Addenda through year 2000.9. SI Calculation Package, Fatigue Analysis of Feedwater Nozzle, Revision 0, SI File No. VY-16Q-302.File No.: VY-19Q-302 Page 12 of 12 Revision:

0 F0306-OIRO Exhibit C Structural Integrity Associates, Inc. File No.: VY-19Q-303 CALCULATION PACKAGE Project No.: VY-19Q PROJECT NAME: Provide VY Support for Questions Related to Environmental Fatigue Analyses CONTRACT NO.: 10163217 CLIENT: PLANT: Entergy Nuclear Operations, Inc. Vermont Yankee Nuclear Power Station CALCULATION TITLE: Feedwater Nozzle Environmental Fatigue Evaluation Project Manager Preparer(s)

&Document Affected Revision Description Approval Checker(s)

Revision Pages Signature

& Date Signatures

& Date 01 -7 Initial issue. Ten-y J. Herrmann Gary L. Stevens 01/30/2008 01/30/2008 Terry J. Hemnann 01/30/2008 Page 1 of 7 F0306-01RO

-.j V Structural Integrity Associates, Inc.Table of Contents 1.0 INTROD UCTION ...........................................................................................................................

3 2.0 APPROA CH ....................................................................................................................................

3 3.0 M ETH OD OLO GY ..........................................................................................................................

4 4.0 CA LCULATION S ...........................................................................................................................

5 5.0 CON CLU SION S .............................................................................................................................

5 6.0 REFEREN CES ................................................................................................................................

6 List of Tables Table 1: EA F Calculations for the Feedw ater N ozzle Com er ..........................................................

7 File No.: VY-19Q-303 Revision:

0 Page 2 of 7 F0306-OI RO V Structural Integrity Associates, Inc.

1.0 INTRODUCTION

The purpose of this calculation is to perform a plant-specific evaluation of reactor water environmental effects for the reactor pressure vessel (RPV) feedwater.

nozzle identified in NUREG/CR-6260

[1] for the older vintage General Electric (GE) plant for the Vermont Yankee Nuclear Power Station (VY).2.0 APPROACH Per Chapter X, "Time-Limited Aging Analyses Evaluation of Aging Management Programs Under 10 CFR 54.2 l(c)(l)(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 one 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, as they were based on generic analysis results from NUREG/CR-6260 that were not VY-specific.

Plant-specific analyses have been recently completed to address those components for VY. Relevant chemistry input for this calculation is contained in Reference

[5]. This calculation documents the EAF evaluation for the feedwater nozzle locations.

File No.: VY-19Q-303 Revision:

0 Page 3 of 7 F0306-01RO Structural Integrity Associates, Inc.3.0 METHODOLOGY 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), which are the materials under consideration for the feedwater nozzle. Per Figure 2 and Table 2 of Reference

[6], the two locations being evaluated are the feedwater nozzle safe end (carbon steel) and the feedwater nozzle forging corner (low alloy steel). Based on the materials of these locations, the appropriate expressions are: For Carbon Steel [4, p. 69]: Fen =exp (0.585 -0.00124T'

-0.101S*T*O*c*)

(1)Substituting T' 25°C in the above expression, as required by NUREG/CR-6583 to relate room' temperature air data to service temperature data in water [7], the following is obtained: Fen = exp (0.585 -0.00124(25°C)

-0.101 S* T* O*s*) (2)= exp (0.554 -0.101 S* T* O* c*)(3)(4)For Low Alloy Steel [4, p. 69]: Fen = exp (0.929 -0.00124T'

-0.IOS*T*O*s*)

Substituting T' = 25°C in the above expression, as required by NUREG/CR-6583 to relate room temperature air data to service temperature data in water [7], the following is obtained: Fen = exp (0.929 -0.00124(25°C)

-0.101 S* T* O*E*) (5)= exp (0.898 -0.101 S* T* O*'*)(6)where [4, pp. 60 and 65]: Fen -S*T*T -0* =fatigue life correction factor S for 0 < sulfur content, S < 0.015 wt. %0.015 for S > 0.015 wt. %0 for T< 150 0 C (T -150) for 150< T 3501C fluid service temperature (0 C)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 6* = 0 for strain rate, E* > 1%/sec= ln(&*) for0.001 < -* 1%/sec= ln(0.001) for 8* < 0.001%/sec Bounding Fen values were determined in Reference

[5]. The values determined in Table 3 of Reference

[5] will be used for the carbon steel feedwater nozzle safe end location, where feedwater DO levels are low and the(Fn value is a constant value of 1.74 for all temperatures for both hydrogen water chemistry (HWC) and normal water chemistry (NWC) conditions.

For the low alloy steel File No.: VY-19Q-303 Revision:

0 Page 4 of 7 F0306-01 RO Structural Integrity Associates, Inc.nozzle comer location, the applicable Fen values are shown in Table 4 of Reference

[5]. Since there is a significant variation in values with temperature, Fen values will be computed for each load pair in the detailed fatigue calculation for this location.The enviromnental fatigue is determined as Uenv = (U) (Fen), where U is the original fatigue usage and Uev is the environmentally assisted fatigue (EAF) usage factor. All calculations can be found in Excel spreadsheet "VY-19Q-303 (Env. Fat. Calcs).xls" associated with this calculation.

From Table I of Reference

[5], the following water chemistry input applies for the low alloy steel nozzle corner location:* 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.* For the RPV Upper Region, which is applicable to the nozzle corner location, DO is 114 ppb pre-HWC and 97 ppb post-HWC.With these assumptions, the cumulative usage factor (CUF) values documented in this calculation are considered applicable for sixty years of operation including all relevant EAF and EPU effects.4.0 CALCULATIONS From Table 6 of Reference

[6], the CUF for the safe end for 60 years of operation is 0.0571. Thus, the EAF CUF for 60 years is 0.0571 x 1.74 = 0.0994, which is less than the allowable value of 1.0 and is therefore acceptable.

The CUF for the nozzle comer for 60 years of operation is shown in Table 7 of Reference

[6], and has a value of 0.0889. This calculation is reproduced in Table 1, along with EAF calculations on a load pair basis using the Fen expression in Equation (6) above for low alloy steel. The final EAF CUF for 60 years is 0.3531, which is less than the allowable value of 1.0 and is therefore acceptable.

The overall Fen multiplier for this location is 3.97.

5.0 CONCLUSION

S In this calculation, EAF calculations were performed in accordance with the GALL Report [2] for the feedwater nozzle safe end (carbon steel) and nozzle corner (low alloy steel) locations.

These 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 are documented in other calculations.

The EAF results for the locations identified above indicate that the fatigue usage factors, including environmental effects, are within the allowable value for 60 years of operation.

The calculations for both locations make use of the 60-year projected cycles for VY and incorporate EPU effects FileNo.: VY-19Q-303 Page 5 of 7 Revision:

0 F0306-01RO Structural Integrity Associates, Inc.(conservatively assumed to apply for all 60 years of operation).

Therefore, no additional evaluation is required for these components, and the GALL requirements are satisfied.

6.0 REFERENCES

1. NUREG/CR-6260 (INEL-95/0045), "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," March 1995.2. NUREG-1801, Revision 1, "Generic Aging Lessons Learned (GALL) Report," U. S. Nuclear Regulatory Commission, September 2005.3. 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-97/18), "Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels," March 1998.5. SI Calculation Package, Environmental Fatigue Evaluation of Reactor Recirculation Inlet Nozzle and Vessel Shell/Bottomn Head, Revision 0, SI File No. VY-1 6Q-3 03.6. SI Calculation Package, ASME Code Confir7natory Fatigue Evaluation of Reactor Feedwater Nozzle, Revision 0, SI File No. VY-19Q-302.

7. EPRI/BWRVIP Memo No. 2005-271, "Potential Error in Existing Fatigue Reactor Water Envirotnental Effects Analyses," July 1, 2005.File No.: VY-19Q-303 Revision:

0 Page 6 of 7 F0306-O1RO Structural Integrity Associates, Inc.Table 1: EAF Calculations for the Feedwater Nozzle Corner CUF Calculation from Table 7 of Reference 161 EAF lcuations NWC DO NWC DO 97 114 ppb% HWC = 47% 53% = % hwC Index Load #1i Description

1. lt, (cycles) Load #271 escrlption

1. 2! a (cycles) n (cycles'9 Spsii) K, 9. (psi) tUo,,, U 1 2 2 DesHydrol 123 16 1LcFP2 10 10 65.109 1.000 46.047 5.65.78 000177 2 2 , 2 _OsHvdrol 110 20 11 _LoFF 6 10 10 50344 1.000 43.990 6.477.19 0.001.4 3 2 2 DesHydrol 100 18 11 LoFPP 10 10 E0,150 1000 43.20J 6032.83 000146 4 2 ir90 11 10 10 65,712 1.000 43,011 6924.58 0,00144 6 2 2 DesHydrol 80 5 A TurbRol!f 300 80 A4.296 1.000 43.008 6,925.97 0.01166 6 5 4 TurbRclll 220 39 25 Unbolt 123 123 E3.306 1.000 41.430 0.736 36 0 01589 7 3 2 DosHydro2 120 5 4 TfubRolf 97 97 61f437 1.000 40.391 6.343.90 0.01163 8 3 2 DesHydro2 23 23 11 LoFp9 10 10 63.138 1.000 40,101 0.524.36 0 00117 9 3 DeH'ro2 120A HSFWOInjI 300 13 49.069 1.000 39.607 .810.73 0.00148 10 34 302A HSI/Wlnjil 287 '8 24_HydroTeoL2 1 1 49.097 1.000 39.622 00633.38 000011 11 34 20k HSF1,Vnjf1 296 b6 21 Shucdon M00 266 49.111 1.000 39.616 8.837.8 0.03236 12 26 12 TGTrip2 280 36 21 ShutdownI 14 14 60.779 1.000 38.956 9.57940 0.00146 13 21 11 'oF 0? 10 26 2 TGTrip2 1 274 10 49.390 1,000 33,091 16,015.00 0.00002 14 26 1 TGTrip2 264 14 14SRVBhndn I 1 1 42.902 1.000 21.19 20131.00 0.00003 16 22 1 11 LoFPO 10 76 12TGTrip2 263 10 32.212 1.000 24 07 40,237.00 0 00025 16 4 3 Startup 300 26 12 TGTrip2 253 253 30.212 1.000 23.613 46728.00 0.00541 17 4 3 Startup .7 29 13 0.Owrpr2 1 1 30.212 1.000 23,513 46.728.00 0 00002 18 4 1 _Staup 4 28 13overpr1 1 1 28.966 1.000 19.423 111,113.00 0.00001 19 4 3 3Starup -2 19 RedTo 0 pclt 300 45 24.083 1.000 17.655 156,42.00 0.00029 20 32 I0 RedloOp 26620 '.p 3 200 265 18,765 1.000 12.893 761.835.00 0.00070 21 13 10 F'WHByp 70 20_HSHeatup 45 45 19.037 1.000 12.679 879.615.00 0.00005 22 13 10 FWHypI 2H 3 20A 300 25 20.396 1 000 12.024 014,9340 0 00003 23 30 320A HSFWInj2:

275 41 24kHydroTeot1 1 1 19.960 1.000 12.9 inflnite 0.00000 24 9 i 6 'WhlyRedli 2000 35 20A HS"rInI2i 274 274 19.341 1.000 11,952 infinite 000000 4otal. the 9 .68892 Transient fMfoimum Trnoer-turee:

Index Load P1i Description 41: n cycles) Load #2[ Description

1. 2 Index 1 sl jl I I " T2 2 I S, (psi) T ('F)1 2DosHodrol 120 16 11LcFF2 1 2 30 16 14 65,109 36 2 2 2-CenHydrol 110 20 11_LoFP6 2 2 30 20 3 50.3-4 381 3 2 2 DnHydrol 1 100 18 11 LoFP4 3 2 30 18 7 50,150 389 4 2 2 DenHydfol 90 11 9_TuobTTpl 4 2 30 11 6 65.712 351 5 2 2_- 0 sHydrol 80 5 4 TrbRolll 5 2 30 5 12 60,296 360 6 n 4 TuobRolli 220 39 2'5Unbolt 6 5 12 39 23 63.300 360 7 3 2 D.esHydrc2 120 5 4TTusbRollf 7 3 1 0 12 61,437 360 8 3 2_DooHydrc2 23 23 1 1.LcFF9 8 3 1 23 7 63.132 363 9 3 2 DesHydro2 13 34 20A HSFOPWojl 9 3 1 34 21 49,069 380 10 34 20A HSFWInjf1 287 30 24_HydroTesl2 10 34 21 38 1 49,097 368 11 14 20A_HSFlrnj 1 l 266 6 21_Shuldown 11 34 21 36 1569 49.111 389 12 20 1 2TGTcp2 "2a8 3,3 21Shlufdown 12 26 6 36 165 60,379 349 13 21 .11 LcFP7PO 10 26 12"TGToip2 13 21 60 26 7 49,395 424 14 26 12.TGTnp 2 264 31 14_SRVBIhdn 14 26 6 31 41 42,902 349 16 22 11 LcF5 9 10 2a 12 TGTrhp2 16 22 1 26 7 32.212 638 16 4 s-;rtup 300 26 12-7T7Trjp2 16 4 1 26 7 30,212 803 17 4 3 -_SIartup l47 29 130T;erpr2 17 4 1 29 7 30.212 603 18 4 3 Startup 46 28 13 COrpi 18 4 28 I 20.9066 503 19 4 3_Startup J5 32 19_edTo0pci 19 4 1 32 49 24.063 503 20 32 19 RedTotpct 255 33 20 HSHeotup 20 32 45 35 26 10.769 543 13 Q 10 PFWHB6pl 0 33 20_HSHnatup 21 13 23 33 26 19637 040 22 13 10 PO"Hypli 25 35 20AHSFWInj2 22 13 23 26 14 20.309 549 23 15 23A HSFWVInj2 275 37 24_HydroTestl 23 35 14 37 1 19.850 548 24 9 6 0.,klyRedl 4 2000 35 20A HSFWhni2 24 9 16 35 14 19.341 .048 T.w (°F) 'oTjur ('C) HWO F z ttW.- ", I hI 8, '366.0 100.0 3.242 3.410 0.00689 381.0 193.9 3.687 3.971 0 00592 309.0 108.3 3.842 4.169 000688 301.0 177.2 3.159 3308 000468 3500 G 12.2 3.309 3.494 0.03936 360.0 162.2 3.309 3404 0,05416 390.0 1822 3.309 3.494 0 03961 3M3.0 178.3 3.192 3.349 0.00384 38860 197.8 3.823 4.144 0.00589 368.0 197.8 32823 4.144 9.00046 3630 107.8 3.823 4.144 012921 349.0 176.1 3.127 3.298 0 00468 d24.0 2170. 4.001 5.150 0.00309 3490 176 1 3 127 3.2,09 0 00011 53980 281.1 0.277 10 30 0.00233 603.0 201.7 6.912 0.347 0.04104 603.0 261.7 6,912 0.347 000016 503.0 261.7 6.912 P.347 0.00007 503.0 261.7 6.912 6.347 0.00221 543 0 293.9 6.493 10.649 000323 546.0 2660 7 .714 10.078 0.00051 649.0 287.2 9.769 11 045 000027 5480 2067 8.714 10.978 0 00000 548.0 286.7. 8.714 10 978 0.00000 Total. U 8'r -0.35306 Owesall F.n -3.970 Notes: 1. is the maximunn tentperature of the coo paired load states, and represents the metal (nodal) temperature at the Ilcalion being analyzed.

This is determined

'rom the VESLFAT output from Reference

[61.which is included as 7' in theoTrnsient Maximum Temnperatures' table above.2. F., values computed using E0uation (6) with S0 cnnseonatiel; set to a nmaxiu'nm value of 0 015. and the transforned strain rate conservatina ly "et Io a minimum value of In 10 001)0 -8.906 for all load pairs.3. U1", = [U x HI0 Fr. x % HWC6 ] (U x rlVC F,, % NWN0C0 4. T1 and 72 represent the load number for Load d1 and Load 02. respectiomly, and f1 and s2 reoresent tie state number for eanýh of those loads.File No.: VY-19Q-303 Revision:

0 Page 7 of 7 F0306-01 RO UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.(Vermont Yankee Nuclear Power Station))))))Docket No. 50-271-LR ASLBP No. 06-849-03-LR CERTIFICATE OF SERVICE I, Clara Cavitt, hereby certify that copies of NEW ENGLAND COALITION, INC.'S (NEC) MOTION TO FILE A TIMELY NEW OR AMENDED CONTENTION in the above-captioned proceeding were served on the persons listed below, by U.S. Mail, first class, postage prepaid; and, where indicated by an e-mail address below, by electronic mail, on March 17, 2008.Administrative Judge Alex S. Karlin, Esq., Chair Atomic Safety and Licensing Board Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: ask2@nrc.gov Administrative Judge William H. Reed 1819 Edgewood Lane Charlottesville, VA 22902 E-mail: whrcville@embarqmail.com Office of Commission Appellate Adjudication Mail Stop: O-16C 1 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: OCAAmail@nrc.gov Administrative Judge Dr. Richard E. Wardwell Atomic Safety and Licensing Board Panel Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: rew@nrc.gov Office of the Secretary Attn: Rulemaking and Adjudications Staff Mail Stop: O-16C1 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: hearingdocket@nrc.gov Sarah Hofmann, Esq.Director of Public Advocacy Department of Public Service 112 State Street, Drawer 20 Montpelier, VT 05620-2601 E-mail: sarah.hofmann(istate.vt.us Lloyd B. Subin, Esq.Mary C. Baty, Esq.Office of the General Counsel Mail Stop 0- 15 D21 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: lbs3@nrc.gov; mcb1 @nrc.gov Anthony Z. Roisman, Esq.National Legal Scholars Law Firm 84 East'Thetford Road Lyme, NH 03768 E-mail: aroismananationallegalscholars.com Marcia Carpentier, Esq.Atomic Safety and Licensing Board Panel Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail mxc7@nrc.gov David R. Lewis, Esq.Matias F. Travieso-Diaz Pillsbury Winthrop Shaw Pittman LLP 2300 N Street NW Washington, DC 20037-1128 E-mail: david.lewis@pillsburylaw.com matias.travieso-diaz(anillsburvlaw.com Peter C. L. Roth, Esq.Office of the Attorney General 33 Capitol Street Concord, NH 03301 Peter.roth@doi.nh.gov SHEMS DUNKIEL KASSEL & SAUNDERS, PLLC by: 6/A 644 Clara Cavitt, for Andrew Raubvogel, Esq. and Karen Tyler, Esq.91 College Street Burlington, VT 05401 802 860 1003 802 860 1208 (fax)araubvogel@sdkslaw.com ktyler@sdkslaw.com for the firm Attorneys for New England Coalition, Inc.