ML073390247

(PA-LR) E-mail Responses to Open Amp/Amr Audit Questions

ML073390247
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/14/2007
From:	Limberger W Entergy Corp
To:	Rowley J NRC/NRR/ADRO/DLR
References
TAC MD2297
Download: ML073390247 (16)

Text

Jo'na'tha-n-Rowl4ey_ Re es pories to Open~ AMP/AMR Audit Questions Pg From:

"Limberger, Wayne" <wlimber@entergy.com>

To:

<JGR@nrc.gov>

Date:

11/14/2007 4:06:49 PM

Subject:

Responses to Open AMP/AMR Audit Questions

<<Q&A Open Questions.PDF>>

Jonathan, Attached are the last 10 pages of the AMPI/AMR Audit Q&A database report containing the Vermont Yankee updated responses to the six remaining open questions from the EAF portion of the audit (Questions #387 through

1. 392). Also attached are the signed submittal letter and the Attachment cover page. The hard-copy package with the full database report is due to arrive there tomorrow morning.

CC:

"Mannai, David" <dmannai@entergy.com>, "Metell, Mike" <hmetell@entergy.com>,

"Lach, David J" <DLach@entergy.com>, "YOUNG, GARRY G" <GYOUNG4@entergy.com>, "COX, ALAN B" <acox@entergy.com>

I' c:\\temp\\GW}000o3.TMP Paq6 1 !1 Mail Envelope Properties (473B6361.539 :10: 38201)

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Responses to Open AMP/AMR Audit Questions 11/14/2007 4:05:47 PM "Limberger, Wayne" <wlimber(aentergy.com>

wlimber(aentergy.com Recipients nrc.gov TWGWPO03.HQGWDO01 JGR (Jonathan Rowley) entergy.com acox CC (ALAN B COX)

GYOUNG4 CC (GARRY G YOUNG)

DLach CC (David J Lach) hmetell CC (Mike Metell) dmannai CC (David Mannai)

Post Office TWGWPO03.HQGWDO01 Files Size MESSAGE 435 BVY 07-079 Submittal Letter.pdf BVY 07-079 Attachment Page.pdf Q&A Open Questions.PDF 679412 Mime.822 1099928 Route nrc.gov entergy.com Date & Time 11/14/2007 4:05:47 PM 57882 63922 Options Expiration Date:

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Entergy Entergy Nuclear Operations, Inc.

f.

Box 0500 I

5 ld FIV Roa

-I! ý 0rZ/

57 27 November 14, 2007 BVY 07-079 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

1. Aging Management Program/Aging Management Review (AMP/AMR)

Audit Q&A Database, Revision 5, dated March 27, 2007 Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Update of Aging Management Program Audit Q&A Database This letter provides Revision 6 of the AMP/AMR audit question-and-answer (Q&A) database.

In addition to providing a complete current version of the database, this update addresses Staff questions from the Environmentally Assisted Fatigue (EAF) audit portion of the NRC's Vermont Yankee License Renewal Application Aging Management Program review process. Questions # 387 through #392 of this database remain open and are submitted herewith for Staff review.

All other responses were previously reviewed and are considered closed. The enclosed database (Attachment 1) updates and supersedes Reference 1.

This letter contains no new commitments.

Should you have any questions concerning this matter, please contact Mr. David Mannai at (802) 258-5422.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November 14, 2007.

Sincerely, wr,*KA(r" 1;lT, aw VAJ Ted A. Sullivan Site Vice Pr side Vermont Yan'ee Nuclear Power Station Attachment cc listing (next page)

BVY 07-079 Page 2 of 2 cc:

Mr. James Dyer, Director U.S. Nuclear Regulatory Commission Office 05E7 Washington, DC 20555-00001 Mr. Samuel J. Collins Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mr. Jack Strosnider, Director U.S. Nuclear Regulatory Commission Office T8A23 Washington, DC 20555-00001 Mr. Jonathan Rowley, Senior Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike MS-O-1 1 F1 Rockville, MD 20853 Mr. Mike Modes USNRC RI 475 Allendale Road King of Prussia, PA 19406 Mr. James Kim, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0 8 C2A Washington, DC 20555 USNRC Resident Inspector Entergy Nuclear Vermont Yankee, LLC PO Box 157 Vernon, Vermont 05354 Mr. David O'Brien, Commissioner VT Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601 Diane Curran, Esq.

Harmon, Curran, Spielberg & Eisenberg, LLP 1726 M Street, N.W., Suite 600 Washington, DC 20036

BVY 07-079 Attachment I Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

AMP/AMR Audit Q&A Database Revision 6

,Item Reauest

Response

387 The ASME Code defines that stress intensity (SI) from two temperature transients is calculated from the stress components from the two conditions. Please explain how it could be calculated from stress intensities of the two conditions derived from Greens Functions, especially at locations of geometric discontinuity. Also, please justify the validity of combining the thermal transient stress intensities with the stress Intensities from the external loads and pressure loading.

To address Environmentally Assisted Fatigue (EAF) for the NUREG/CR-6260 locations at Vermont Yankee, the stress inputs for the reactor vessel and nozzles were either taken from the design basis stress analyses or new stress analyses were performed. Existing stress analyses were used for the controlling locations on the vessel shell and for the Recirculation Inlet nozzles. New stress analyses were performed for the Feedwater, Reactor Recirculatlon Outlet, and Core Spray nozzles per ASME III, NB-3200. Updated fatigue analyses for the reactor vessel and nozzles were performed per ASME III, Subsection NB-3222.

New fatigue analyses for the Class 1 portions of the Feedwater and Reactor Recirculation/JRHR piping were performed per ASME III, NB-3600.

Finite element models (FEM) using ANSYS were developed for the new fatigue analyses of the Reactor Recirculation Outlet and Core Spray nozzles. The FEM for each nozzle is 2-D axlsymmetric about the centerline of each nozzle. The radius of the vessel in the FEM was multiplied by a factor of two (2) to account for variation in pressure stress for a nozzle oriented normal to the cylindrical vessel shell.

For the Feedwater nozzles, a previously developed, 2-D axlsymmetric ANSYS FEM was used. The vessel radius used in this model was 1.5 times the radius of the vessel. Pressure stresses from this model were factored by (2.0/1.5) = 1.333 to account for variation In pressure stress for a nozzle oriented normal to the cylindrical vessel shell.

For the new fatigue analyses of the Feedwater, Reactor Recirculation Outlet, and Core Spray nozzles, stress Intensities due to internal pressure were calculated directly using the ANSYS FEM model.

The controlling location for thermal stresses at the safe end of each FEM was determined using a 5009F to 1002F temperature step transient at 100% flow conditions. The controlling location in the blend radius of each FEM was taken as the location of maximum stresses due to internal pressure.

Stress Intensities for each thermal transient were determined using Green's function (GF) methodology. The GF at each controlling location was developed from the FEM stress results for the 5002F to 1002F temperature step transient. At each controlling location, absolute values of the component stress differences, (SZ-SX, SY-SX, SZ-SY), were compared to the maximum stress Intensity calculated from ANSYS. For ease of calculation, the stress difference which most closely matched the total stress intensity calculated by ANSYS was used to determine the GF at each location. In most cases the maximum component stress difference with time matched the maximum stress intensity calculated by ANSYS. This shows that shearing stresses are negligible for the thermal transient at that location and the maximum component stress difference is the maximum stress intensity.

Stress components from attached piping loads at the controlling thermal stress locations were calculated separately using standard strength of materials equations. Stress Intensities were calculated from the stress components per ASME Code,Section III, Subsection NB-3215. These stress intensities are referred to as the "hand calculation method" as described below.

To show that the GF approach used to calculate alternating stress intensities for the 11/14/20071:38:00 PM I~age 141 of I 111,4120,07,:1.38:DC PM.

.Pag-e.-4"I": of,15f

t~mRequest

Response

thermal transients obtains results comparable to results from an ASME Code,Section III, Subsection NB-3222 calculation, a comparison with the results a previous fatigue calculation was conducted. This comparison used the identical FEM constructed for the VY Feedwater nozzle.

The VY ASME Code design fatigue calculation (VY-10Q-303) which was performed directly using ANSYS, was compared to the EAF calculation (VY-16Q-302) performed using the GF methodology for the turbine roll transient. This is the most severe design basis transient for VY Feedwater nozzle. To ensure a consistent comparison between the two calculations, the same stress path locations were selected. The Code fatigue calculation alternating stresses (using the limiting Sz-Sx stress difference) were extracted from the ANSYS model at the same paths used in the EAF calculation. To be consistent the Code FEM analysis was re-run with the same heat transfer coefficients and material properties used for the GF calculation.

The comparison showed that the alternating stress Intensities calculated using the FEM with the Code methodology and those calculated with the GF methodology are within 1% at both the safe end and blend radius locations.

Although this comparison was for the feedwater nozzle, the results are considered to be equally applicable to all other nozzle locations based on a BWR Vessel and Internals Project (BWRVIP) study (EPRI Report No. 1003557, "BWRVIP-108: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," Final Report, October 2002.

In BWRVIP-1 08, 3-D models of four different nozzles were developed and analyzed that bounded all nozzle geometries for the BWR fleet. The results of this study showed that for a range of vessel nozzles modeled using the same technique, the ratio of maximum pressure stress intensity at the blend radius to the primary membrane stress intensity at the vessel wall away from the nozzle is nearly a constant with an average ratio that vanes by +/- 3%. This indicates that all different sized BWR vessel nozzles have the same geometric characteristics for calculating peak stresses in the blend radius regions.

Figures 4-30 to 4-33 in BWRVIP-1 08 show the nozzle blend radius stress profiles for pressure and steady state thermal stresses for the four (4) different BWR nozzles.

The figures show a significant variation of pressure stress around the centerline of the nozzle with the peak hoop pressure stresses occurring at the +902 (top) and 0 (bottom) azimuths. This is due to the differences in hoop and axial stresses in a cylindrical vessel. The new FEM models used in the Vermont Yankee EAF evaluations were 2-D axisymmetric about the centerline of each nozzle. The radius of the vessel in the FEM was multiplied by a factor of two (2) to account for variation In pressure stress for a nozzle oriented normal to the cylindrical vessel shell.

Figures 4-30 to 4-33 in BWRVIP-1 08 also show no significant variance in steady state thermal stresses at the nozzle. The figures show the magnitude of axial stress at the 09 & 1809 azimuths is equal to the magnitude of the hoop stress at +909 and -

909 azimuths. This shows that the thermal stress In the blend radius oriented normal to the axis of the nozzle is nearly constant. Thermal transients used in the EAF evaluations are localized to the nozzle safe end, bore, and blend radius regions. Therefore, the use of 2-D axisymmetric modeling vs. the use of a 3-D FEM is adequate to determine thermal transient stresses in both the safe end and blend Aage 142,0/1 -59 I f/144200-71,;38:00 PM

Item Request RcsPonse radius locations.

The adequacy of the hand calculations used to calculate mechanical load stresses is addressed as follows:

Stress intensities from the attached piping loads at the controlling thermal stress locations were calculated from stress components per ASME Section ill, Subsection NB-3215. For the feedwater nozzle of another BWR plant, hand calculations were performed for stresses due to mechanical loads. The hand calculations were performed using the same methodology as used for VY. These were compared to the results from a finite element model which included the mechanical loads applied directly to the model.

The finite element model was an axisymmetric two-dimensional (2-D) finite element model. This model was constructed and meshed in a very similar manner to the VY nozzle FEMs. Non-symmetric loading elements were used and the shear, moment, axial, and torsional loads were applied to the model.

A comparison of the stresses from the hand calculations vs. the FEM Is as follows:

Location: Safe-End - Linearized Membrane + Bending Stress Stress from Hand Calculations (psi): 8863 Stress from FEM (psi): 5852 Difference, Hand Calc vs. FEM: +51.45%

Location: Safe-End - Total Stress Stress from Hand Calculations (psi): 8863 Stress from FEM (psi): 7855 Difference, Hand Calc vs. FEM: +12.83%

Location:

Nozzle Forging - Linearized Membrane + Bending Stress Stress from Hand Calculations (psi): 1042 Stress from FEM (psi): 769 Difference, Hand Calc vs. FEM: +35.50%

Location: Nozzle Forging - Total Stress Stress from Hand Calculations (psi): 1042 Stress from FEM (psi): 554 Difference, Hand Calc vs. FEM: +88.09%

As shown by these results, use of the hand calculations is conservative compared to the FEM results. Stress intensities from attached piping loads are larger in the safe end section of each nozzle and are significantly reduced for the blend radius section due to the larger section thickness provided by the nozzle reinforcement. As shown in Table 3 of calculation VY-16Q-302 for the VY Feedwater nozzles, the maximum stress intensity from attached piping loads Is 5708 psi for the safe end and 265 psi for the blend radius.

A comparison of the maximum stress intensity from the attached piping loads with the total stress Intensity from the significant transients from Tables 4 and 5 of calculation VY-160-302 for the VY Feedwater Nozzle follows:

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Item - Request,

,Response, Location: Table 5: Safe End Attached Piping Maximum Stress Intensity (psi): 5708.

Transient-No. 3: Startup, I = 16,328 sec.

Total Stress Intensity (psi): 14396.

Attached Piping Stress Intensity is 39.6 % of Total Transient No. 4: Turbine Roll, t = 4 sec.

Total Stress Intensity (psi): 53379.

Attached Piping Stress Intensity is 10.7% of Total Transient:

No. 11: Scram - LOFP, t= 2168 sec.

Total Stress Intensity (psi): 70223.

Attached Piping Stress Intensity is 8.1% of Total Transient No. 20A: Hot Standby, t = 4 sec.

Total Stress Intensity (psi): 53379.

Attached Piping Stress Intensity is 10.7% of Total Location: Table 4: Blend Radius Attached Piping Maximum Stress Intensity (psi): 265.

Transient No. 3: Startup t = 16,782 sec Total Stress Intensity (psi): 34282.

Attached Piping Stress Intensity Is 0.8% of Total Transient:

No. 4: Turbine Roll t = 1802 sec.

Total Stress Intensity (psi): 67667.

Attached Piping Stress Intensity is 0.4% of Total Transient:

No. 11: Scram - LOFP t = 195 sec.

Total Stress Intensity (psi): 74567.

Attached Piping Stress Intensity is 0.4% of Total Transient: No. 20A: Hot Standby, t = 183 sec.

Total Stress Intensity (psi): 66298.

Attached Piping Stress Intensity is 0.4% of Total As shown in table above, the contribution to the total stress range from the attached piping loads is more significant for the safe end location and could effectively be Ignored for the blend radius location.

Combining the thermal transient stress intensities directly with the stress intensities calculated from the external loads and pressure loading essentially combines the maximum principal stresses calculated for each load case. This allows for combination of stress results where different methods or models are used to calculate the stresses, and typically produces conservative results compared to combining all stress components and then determining a stress intensity.

The practice of combining stress intensities from thermal transient load cases directly with the stress intensities from mechanical loads vs. combining all stress components and then calculating a combined stress intensity was used by CB&I for

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Item. Request

Response

the original design analyses for Vermont Yankee and by GE In the analyses for the replacement safe ends for the Reactor Recirculation Inlet and Outlet Nozzles.

For the evaluation of EAF for Vermont Yankee, the combination of thermal stress intensities with stress intensities from external loads and pressure was performed as follows:

The stress intensity calculated from external loading is added to the maximum calculated thermal transient stress intensity using the same sign to increase the stress range. This is necessary because the direction of applied loading from external loads is not known. The stress intensity from the external loads for each transient is scaled for the temperature of the transient assuming no stress occurs at 709F and full values are reached at reactor design temperature of 5759F. This maximizes the stress range pairings for the fatigue analysis. The pressure stress intensity value Is added to the stress intensity from the combined thermal and external loadings directly as a positive value, since pressure is always positive due to the known direction of loading.

388 Provide Justification for statement on page 5 of 34 of Calculation No. VY-1 60-302, that "The Greens Function methodology provides identical results compared to running the input transient through the finite element model."

A calculation was performed by the vendor as part of the generic verification for the Green's function approach. The calculation compared the results from an ANSYS FEM analysis of a feedwater nozzle for a turbine roll transient with the results using Green's functions for the same transient. The results showed the stress range differences between the Green's function approach and the ANSYS FEM for the safe end location were between -0.W61 and 3.430% and for the blend radius location were between -1.73% and 1.56%. These differences are considered well within the accuracy range of the analysis.

In addition, a VY specific comparison was made for the Feedwater nozzle as described in the response to question 387 above. The comparison showed that the alternating stress intensities calculated using the ANSYS FEM with the ASME Code methodology and those calculated with the GF methodology for the same transient Inputs are within 1% at both the safe end and blend radius locations.

Further discussion of Green's functions and how they are used in a fatigue monitoring system is available in two papers. The papers are titled "An On-Line Fatigue Monitoring System for Power Plants: Part I - Direct Calculation of Transient Peak Stress Through Transfer Matrices and Green's Functions" and "An On-Line Fatigue Monitoring System for Power Plants: Part II - Development of a Personal Computer Based System for Fatigue Monitoring', ASME Pressure Vessel and Piping Conference, Vol.112, 1986 (Kuo, Tang, and Rlccardella).

The intent of the statement in this and other calculations was to indicate that equivalent stress history results are obtained from each method (Green's function vs. FEM) for a given thermal transient.

1,111412007 1ý38;00 PM Piqe.'.145'ý01W

Item Request

Response

389 For the blend radius for the feedwater nozzle in Calculation No. VY-160-302, Table 4, Page 16: Why are the Total & M+ B stresses for Thermal Transient 3 shown in columns 3 & 4 high at t--0 sec. (zero stress state?) This question also applies to:

Transient 4 at t =1801.9 sec.

Transient 9 at t = 2524 sec.

Transient 21-23 at t= 20144 sec.

This question may also apply to transients 11, 12, and, 14.

The Green's functions are based on constant material properties and heat transfer coefficients. Therefore, parameters were chosen to bound the transients that result in the majority of the fatigue usage. The temperatures in the design transients range from 100F to 5492-.

Material properties and heat transfer coefficients at 3002F were used. These bound the cold water injection events. In addition, the instantaneous value of the coefficient of thermal expansion is used instead of the mean value.

To maximize stresses in the blend radius, the Green's function was based on a fluid temperature shock of 5009 F to 1002F in the nozzle flow path while the vessel wall portion of the model was exposed to a constant fluid temperature of 5002F.

Therefore the reference point stress estimated from this Green's function at an ambient nozzle fluid temperature is non-zero due to the vessel wall being held at 5002F. The resulting stress ranges from the thermal transient analysis using the Green's function methodology are accurate regardless of the reference point used as long as the material properties used are consistent with the transient temperature range. With the above in mind, Table 4 of VY-16Q-302 was set up to yield stress pairings which ensure the calculated stress ranges would be maximized.

These temperature conditions are appropriate for Green's function integration of all feedwater nozzle transients which contribute to fatigue, since they occur with feedwater flow injecting through the nozzle Into a hot vessel. In reality, these conditions are conservative for transients where there is no flow in the nozzle or for transients where the reactor temperature drops below 5000 F. This is due to a large temperature gradient Induced into the nozzle structure due to the temperature difference between the reactor and nozzle flow path portions of the model. This temperature difference leads to the high stress values observed for Transients 3, 14 and 21-23 at ambient temperatures.

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Item Request

Response

390 Explain why there are differences in the calculated CUF values a between Rev. A and Rev. 0 of the Structural Integrity Calculations. Also, why are the CUFs calculated by Structural Integrity different from the CUFs shown in Tables 4.3.1 & 4.3.3 of the Vermont Yankee License Renewal Application?

Calculations VY-160-301 through VY-16Q-311 performed. for VY and issued as Revision A have the Revision Description on each calculation cover sheet labeled as "Initial Draft for Review". These draft calculations were issued for client review and comment. The draft versions of the calculations were never intended to be the Issued version until all inputs were finalized and all external and Internal reviews and comments were incorporated. The Revision A calculations were provided under Entergy's obligation to provide all documents related to Environmentally Assisted Fatigue for NEC Contention 2.

The Revision 0 versions of these calculations were subsequently issued after comments on the draft calculations were resolved and all design inputs finalized and verified. Revision 0 (or later) versions of the calculations are the Calculations of Record. The Revision A drafts are no longer applicable.

The Revision 0 calculations and reports Incorporated reviewer comments which included; expanded descriptions of the methodologies and analyses, additional references, typographical corrections, and component specific technical comments which affected the final CUF values.

Referring to Table 3-10 In the Summary Report, SIR-07-132-NPS (VY-160-404), the most significant changes from Revision A to Revision 0 for 60 year environmental CUFs greater than 0.50 were for the Feedwater nozzle and the Reactor Recirculation / RHR piping.

Specific inputs which affected the CUF for the Feedwater nozzle included Increased pressure stresses and reduced thermal stress inputs for isothermal events. The Revision 0 calculation for the Feedwater nozzle (VY-1 6Q-302) shows the blend radius as the limiting section vs. the safe end as the limiting section in Revision A.

This resulted in an increase In the 60 year CUF from 0.0127 to 0.0636 and an increase in the 60 yr environmental CUF from 0.127 to 0.639 at the Feedwater nozzle blend radius.

Corrections to the transient temperature inputs for the Reactor Recirculation / RHR piping model resulted in the maximum calculated CUFen = 0.7446 at the RHR return tee location. The draft Revision A of the calculation had the RHR suction tee controlling.

The CUFs shown in Tables 4.3.1 & 4.3.3 of the Vermont Yankee License Renewal Application were based on the design basis fatigue evaluations factored to account for the effects of the 1200/6 Extended Power Uprate. For locations with no plant specific CUFs, representative values from NUREGICR-6260 were used.

The CUFs calculated for the Environmentally Assisted Fatigue evaluation are different from the CUFs shown in the VY LRA due to a number of factors specific to each location. These Include:

• Oupdated finite element modeling (FEM) vs. the shell analysis techniques used in the original design analysis,

• Odirect thermal transient analysis using the FEM vs. the separate thermal analyses to determine temperature distributions used in the original design analyses,

-Ouse of updated transient definitions for 60 years of operation. The updated transient deflnitions are shown in Design Input Record (DIR) for EC No. 1773, Rev.

0, "Environmental Fatigue Analysis for Vermont Yankee Nuclear Power Station" 7111412007 ~:38:QD *PIW Page 141 of 1~D 11/114120071.,38,00 P.MW

,Pag 147 ofr 150

Item Recuest

,Response Revision 1, dated 7/26/07.

For the NUREG/CR-6260 locations without existing fatigue analyses, new MY plant-specific ASME III fatigue analyses were performed.

391 On page 1-1 of Report VY-16Q-401 it indicates that refined transient definitions 60 years are used in the computation of the CUF including EAF effects. Please explain the refinements in the transient definitions.

The original design transients for the VY Reactor Vessel are given in Section 5.1.8 and Attachment D to General Electric Purchase Specification No. 21A1 115,

,Reactor Pressure Vessel", Revision 4, 10/21/69 and certified on 10/23169 as contained In the Reactor Pressure Vessel Design Report. This document is the Vermont Yankee Reactor Vessel Design Specification. Additional clarifications and descriptions for the design transients were provided by General Electric in GE Letter W. J. Zarella to OW. Edwards - Yankee Atomic,

Subject:

WV. Y. R.P.V.

Temperature Transient/ Cycling Events", No. G-HB-5-124, dated November 5, 1975.

Earlier versions of the specification made reference to a GE Thermal Cycle Drawing No. 8850941. The final version of the design specification relocated this cycle information to Attachment D of the specification and deleted references to GE drawing No. 8850941.

Comparisons were made between the VY Design Specification transients and the design transients shown on Thermal Cycle Drawings from other GE BWR 4 plants of the same and later vintage. The later plants have more detailed thermal cycle descriptions based on the experience from the earlier GE BWRs.

In general, VY is designed for a smaller spectrum of the most severe transients as compared to the full spectrum of transients used for the later units. As described in General Electric Letter No. G-HB-5-124, the number of cycles for each VY design transient exceeds the number of cycles for the same transient from the typical GE thermal cycles diagram listed in the original VY FSAR. For example; the single severe design transient for the VY Feedwater nozzle of 1500 cycles exceeds the 518 Start-up, Loss of Feedwater Heater, Scram, and Shutdown events listed in the original VY FSAR.

To insure a realistic projection of design thermal transient cycles and events for 60 years of operation, the Thermal Cycle Diagrams used at a number of BWR 4 plants were used as a starting point. The VY Design Specification transients were mapped onto the typical BWR 4 Transient Diagram. Then projections for 60 years were made based on the numbers of cycles in the VY Design Specification, the numbers actually analyzed in the VY Design Certified Stress Report for Vermont Yankee.

Reactor Vessel, Chicago Bridge & Iron, Contract 9-6201, and the number of cycles experienced by VY in approximately 35 years of operation. For all Service Level A &

B events, the 60-year projected cycles for each transient used in the EAF evaluations exceed the actual number of cycles experienced by VY projected to 60 years of operation. The basis for the 60 year transient definitions is documented in Appendix C of calculation VYC-378 Revision 2.

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Request,

Response

392 For the Feedwater Nozzles there are large differences In the CUFs without the Fen The original Design Transient for the VY Feedwater nozzle is given In Attachment 0 factors shown In shown in Table 4.3.1 of the Vermont Yankee License Renewal to GE Specification No. 21A1 115, "Reactor Pressure Vessel", Revision 4, Application and those shown in calculation VY-16Q-302. Section 2.0 of the calculation 10/21/69. It is a single severe design transient intended to envelop all Start-up, on page 4 of 32 states, "...several of the conservatisms originally used in the original Loss of Feedwater Heater, Scram, and Shutdown events. It consists of 1500 cycles feedwater evaluation (such as grouping of transients) are removed.2..

Please explain Of.

what conservatisms were removed.

" Feedwater nozzle at 5464-F steady state and 0% feedwater flow, followed by,

" Step change from 5460F to 1002F with 25% feedwater flow, followed by,

" Feedwater nozzle at 1 00-F steady state and 25% feedwater flow, followed by,

" Step change from 1002F to 260-F, followed by,

" A ramp from 2609F to 3769F at 250W/Hr concurrent with increasing feedwater flow 25% to 100% rated flow.

This transient is equivalent to a Startup and Turbine Roll event combination specified on newer BWR plant Thermal Cycle Diagrams.

As described in GE Letter No. G-HB-5-124, dated November 5, 1975, the 1500 such events considered in the design fatigue evaluation of the Feedwater nozzle exceed the 518 Start-up, Loss of Feedwater Heater, Scram, and Shut-down events listed in the original FSAR.

The CUF for the Feedwater nozzle shown in Table 4.3.1 of the Vermont Yankee License Renewal Application is based on the design basis fatigue evaluations factored to account for the effects of the 120% Extended Power Uprate (EPU).

Changes in temperatures for EPU are from GE Nuclear Energy Certified Design Specification No. 26A6019, "Reactor Vessel - Extended Power Uprate", Rev. 1, 8129/03.

The evaluation of EPU effects on the feedwater nozzle and safe end stress and fatigue analysis is contained in VY Engineering Report, VY-RPT-05-001 00, Rev. 0, "Task T0302 Reactor Vessel Integrity-Stress Evaluation EPU Task Report for ER 1409". Section 3.3.1.1 of GE Report for Task 302, shows the value for the feedwater nozzle safe end EPU CUF for 40 years = 0.75. This is the value shown in Table 4.3.1.

The 0.75 CUF value is based on the original design report. The original design analysis was performed for "loose fit" feedwater spargers where the annular cold gap between the stainless steel thermal sleeve and the nozzle safe end was 0.020 inch. The feedwater spargers and thermal sleeves were replaced in 1976 with new "interference fit" thermal sleeves. The interference fit thermal sleeves significantly reduce leakage flow past the thermal sleeve into the bore region of the nozzles. This reduces the heat transfer from the process fluid to the nozzle base metal, thereby reducing thermal stresses during system thermal transients.

Subsequent to the GE report, a re-analysis of the Feedwater nozzle was performed.

Report No. SIR-04-020 Revision 0, March 2004. "Updated Stress and Fatigue Analysis for the Vermont Yankee Feedwater Nozzles" documents a revised ASME III Stress and Fatigue Analysis for the Feedwater nozzle and safe end. This 1 1:/f 4120t07I" :1;:38, 0PM,'

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Response

analysis included effects of the interference fit thermal sleeve. The analysis was performed for both the original licensed power and system flow rates using the enveloping design transient, "Startup, Loss of Feedwater Heaters, Scram &

Shutdown", from the original Design Specification and for EPU power and flow conditions as modified per the EPU Design Specification. For the nozzle safe end, the 40-year CUF using 1500 cycles of the enveloping transient and including EPU effects = 0.4513 (as compared to the 0.75 value used in the LRA). The primary reason for the decrease In CUF was a result of the improved heat transfer coefficients resulting from the interference fit thermal sleeve.

The calculated fatigue usage for the safe end prior to the installation of the interference fit thermal sleeve using the actual number of startup and shutdown cycles and the allowable number of cycles from the original CB&I design report Is 0.02. As documented in Appendix D to calculation VYC-378, Rev. 2, this has a negligible effect on the revised CUF for the safe end including EPU effects shown above. This is primarily due to conservatismrs in the Updated Feedwater Fatigue Analysis for the number of cycles operating under EPU conditions.

For the Environmentally Assisted Fatigue (EAF) evaluation, (VY-16Q-302), a realistic projection of Design Thermal Transient Cycles and Events for 60 years of operation based on the Feedwater Nozzle Thermal Cycle Diagram from a typical BWR 4 was used. A described in the response to Question 391, the enveloping design transient was mapped to the 'Turbine Roll & Increase to Rated Power" transient Other transients including Loss of Feedwater Heaters and Scram events were taken directly from the typical BWR 4 Feedwater Nozzle Thermal Cycles Diagram using VY specific EPU design pressures and temperatures. The projections for 60 years were based on the number of events in the VY Design Specification, the numbers analyzed in the VY Design Certified Stress Report for VY Reactor Vessel, and the number of cycles experienced by VY in approximately 35 years of operation.

The design transients used in the EAF evaluation for the VY Feedwater nozzle are shown in Attachment I to Design Input Record (DIR) for EC No. 1773, Rev. 0, "Environmental Fatigue Analysis for Vermont Yankee Nuclear Power Station" Revision 1, dated 7/26/07.

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