ML20065G590

Cycle 15 Core Performance Analysis Rept

ML20065G590
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	08/24/1990
From:	Paul Bergeron, Cacciapouti R, Sironen M YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20065G582	List: BVY-90-096, Forwards Rev 0 to Vermont Yankee Nuclear Power Station Cycle 15 Core Operating Limits Rept ML20065G583 ML20065G590
References
YAEC-1749, NUDOCS 9010220150
Download: ML20065G590 (103)

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Vermont Yankee ,

Cycle 15 I Core Performance Analysis Report I  ;

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i August 1990 Major Contributors: V. Chandola -

B. Y. Hubbard-I M. P. LeFrancols D. J. Morin J. Pappas i  ;

I R. C. Potter J. D. Robichaud R. P. Smith i i g K. E. St. John g R. W. Sterner R.-A. Wochlke t l

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Prepared by: k M L4.g_) R 9d M. A. Sironen '(Da(e)

Nuclear Engineering Coordinator Approved by:

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R. J. Cacciapouti, Manager (Date)

Reactor Physics Group Approved by: Yd I P. A. Bergero Manager Transient An ysis Group (Da'te )

Approved by t *,#r f

_ 0s u . l 'l_ $YE3 N LbCAA r p Apprcived by: X, /

B. C. Slyer, Director (Date)

Nuclear Engineering Department 0

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Yankee Atomic Elec.* ric Company I Nuclear Services Division 580 Main Street Bolton, Massachusetts 01740 l WPP40/10

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I DISCIAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company lj

(" Yankee"). The use ot information contained in this document by anyone other '

than Yankee, or the Organization for which this document was prepared under contract, is not authorized and, ylth respect to_any unauthoriltd_uge, neither Yankee nor its officers, directore, agents, or employees assume any I obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained in this document.

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[ AISIMCI This report presents design inform. tion, calculational results, and operating limits pertinent to the operation of Cycle 15 of the Vermont Yankee

( Nuclear Power Station. These include the fuel design and core loading pattern descriptionst calculated reactor power distributions, exposure distributions, shutdown capability, and reactivity datal and the results of safety analyses performed to justify plant operation throughout the cycle.

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I IMLLOLCONIDiTS .

.I fage DISCLAIMER.................................................. iii ABSTRACT.................................................... iv TABLE OF CONTENTS........................................... V LIST OF FIGURES............................................. vii LIST OF TABLES.............................................. ix ACKN0WLEDGEMENTS............................................ xi

1.0 INTRODUCTION

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2.0 RECENT REACTOR OPERATING llIST0RY............................ 2 2.1 Operating Elistory of the Current Cycle................. 2 2.2 Operating flistory of Past Applicable Cycle............. 2 3.0 RELOAD CORE DESIGN DESCRIPTION.............................. 5 3.1 Core Fuel Loading...................................... 5 3.2 Design Reference Core Loading Pattern.................. 5 3.3 Assembly Exposure Distribution......................... 5 4.0 FUEL MECllANICAL AND TilERMAL DESION.......................... 8 4.1 M e c han i c a l De s i s;n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 1 4.2 Thermal Design......................................... 8 4.3 Operating Experience................................... 9 5.0 NUCLEAR DESIGN.............................................. 16 5.1 Core Power D1stributione............................... 16 5.1.1 Italing Power Distribution....................... 16 5.1.2 Rodded Depletion Power Distribution............. 16 5.2 Core Exposure Distributions............................ 17 1 5.3 Cold Shutdown Marg 1n................................... 17 5.4 Standby Liquid Control System Shutdown Capability...... 18 5.5 Maximum Fw for the Spent Fuel Poo1..................... 18 6.0 TilERMA L-l! YDRAUL I C DES I GN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 6.1 S t eady-S ta t e The rma l liyd raul ic s . . . . . . . . . . . . . . . . . . . . . . . . 27 6.2 Reactor Limits Determination........................... 27

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l I l I TABLE OF CONTENTS (Continued)

Pajte 7.0 ACCIDENT ANALYSIS........................................... 29 7.1 Transient Analysis..................................... 29 7.1.1 Methodology..................................... 20 g 7.1.2 Initial Conditions and Assumptions.............. 31 F

W 7.1.3 One-Dimensional Cross Sections and Kinetics Parameters............................. 32 7.1.4 Transients Analyzed............................. 33 7.2 Transient Analysis Results............................. 34 I 7.2.1 7.2.2 Turbine Trip Without Bypass Transient...........

Generator Load Rejection Without Bypass Transient................................

34 35 7.2.3 Loss of Feedwater Heating Transient............. 35 7.3 Overpressurization Analysis Results.................... 36 7.4 Local Rod Withdrawal Error Transient Results........... 37 I 7.5 Misloaded Bundle Error Analysis Results................

7.5.1 Rotated Bundle Error............................

40 40 7.5.2 Mislocated Bundle Error......................... 41 7.6 Control Rcd Drop Accident Results...................... 41 7.7 Refueling Accident Results............................. 43 8.0 LOSS-OF-COOLANT ACCIDENT ANALYSIS........................... 78 9.0 CORE COMPONENT QUALIFICATION PR0 GRAM........................ 79 9.1 Advanced Nuclear Fuels Fuel Assemblies................. 7) 9.2 General Electric Marathon Control Rods................. 79 10.0 STARTUP PR0 GRAM............................................. 81

> REFERENCES.................................................. 82 APPENDIX A CALCULATED OPERATING LIMITS..................... A-1 I

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I I lI I Number LISI_0LEIGURES lille IAge 3.2.1 VY Cycle 15 Design Reference Loading Pattern, Lower Right

! Quadrant 7 I 4.2.1 VY Cycle 15 Core Average Gap Conductance Versus Cycle Exposure 13 4.2.2 VY llot Channel Gap Conductance for CE8X8NB Versus Exposure 14 l 4.2.3 VY llot Channel Cap Conductance for CE8X8EB Versus Exposure 15 l

4 5.1.1 VY Cycle 15 llaling Depletion, E0FPL Bundle Average I, Relative Powers 20 I 5.1.2 VY Cycle 15 !!aling Depletion, EOFPL Core Average Axial Power Distribution 21 5.1.3 VY Cycle 15 Rodded Depletion - ARO at EOFPL, Bundle Average Relative Powers 22 5.1.4 VY Cycle 15 Rodded Depiction - ARO at EOFPL, Core Average Axial Power Distribution 23 5.2.1 VY Cycle 15 !!aling Depletion, E0TPL Bundle Average Exposures 24 5.2.2 VY Cycle 15 Rodded Depletion, E0rPL Bundle Average Exposures 25 l

5.3.1 VY Cycle 15 Cold Shutdown AK in Percent versus Cycle Exposure 26 7.1.1 Flow Chart for the Calculation of ACPR Using the I RETRAN/TCPYA01 Codes 48 I

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L LISI_0E_IIGURES (

(Continued) J

._RumbcI_ 11112 Tage 7.2.1-1 to Turbine Trip Without Bypass. E0FPL15 Transient l

7.2.1-3 Response Versus Time, " Measured" Scram Titne 49-51 7.2.2-1 to Turbine Trip Without Bypass, EOFPL15-1000 MWD /ST 7.2.2-3 Trai. lent Response Versus Time, " Measured" Scram Time 52-54 7.2.3-1 to Turbine Trip Without Bypass, EOFPL15-2000 MWD /ST 1 7.2.3-3 Transient Response Versus Time " Measured" Scram Time 55-57 7.2.4-1 to Generator Load Rejection Without Bypass E0FPL15 7.2.4-3 Transient Response Versus Time, " Measured" Scram Time 58-60 7.2.5-1 to Generator Loni Rejection Without Bypass, E0FPL15-1000 MWD /ST 7.2.5-3 Transient kc' :nse Versus Time, " Measured" Scram Time 61-63 l 7.2.6-1 to Generator Load Rejection Without Dypass, EOFPL15-2000 MWD /ST 7.2.6-3 Transient Response Versus Time, " Measured" Scram Time 64-66 7.2.7-1 to Loss of 100 F Feedwater lleating EOFPL15-1000 MWD /ST 7.2.7-2 (Limiting Case) Transient Response Versus Time 67-68 7.3.1-1 to MSIV Closure, Flux Scram, EOFPL15 Transient Response 7.3.1-3 Versus Time, " Measured" Scram Time 69-71 7.4.1 Reactor Initial Conditions and Transient Summary for the VY Cycle 15 Rod Withdrawal Error Case 1 72 7.4.2 Reactor Initial Conditions and Transient Summary for the VY Cycle 15 Rod Withdrawal Error Case 2 73 7.4.3 VY Cycle 15 RWE Case 1 - Setpoint Intercepts Determined by the A and C Channel 74 7.4.4 VY Cycle 15 RWE Case 1 - Setpoint Intercepts Determined y by the B and D Channel 75 l

7.6.1 First Four Rod Arrays Pulled in the A Sequences 76 7.6.2 First Four Rod Arrays Pulled in the B Sequences 77 I

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LIELQf_1ABl.ES 1

!' Rumber Iitle tage 2.1.1 VY Cycle 14 Operating Highlights 3 2.2.1 VY Cycle 13 Operating flighlights 4 3.1.1 Assumed VY Cycle 15 Fuel Bundle Types and Numbers 6 3.3.1 Design Basis VY Cycle 14 and Cycle 15 Exposures 6 4.1.1 Nominal Fuel Mechanical Design Parameters 10 4.2.1 VY Cycle 15 Cap Conductance Values Used in Transient Analyses 11 4.2.2 Peak Linear llent Generation Rates Corresponding to Incipient Fuel Centerline Melting and 1% Cladding Plastic Strain 12 5.3.1 VY Cycle 15 1;-Effective Values and Shutdown Margin Calculation 19 5.4.1 VY Cycle 15 Standby Liquid Control System Shutdown Capability 19 5.5.1 VY Cycle 15 Maximum Cold Fw of Any Enriched Segment 19 7.1.1 VY Cycle 15 Summary of System Transient Model Initial

g Conditions for Core Wide Transient Analyses 44

g' 7.2.1 VY Cycle 15 Core Wide Transient Analysis Resulta 45 7.3.1 VY Cycle 15 overpressurization Analysis Results 46 7.5.1 VY Cycle 15 Rotated Bundle Analysis Results 46 7.5.2 VY Cycis 15 Mislocated Bundle Analysis Resulta 46 7.6.2 VY Cycle 15 Control Rod Drop Analysis Results 47

! 9.1.1 Nominal ANF-1X Fuel Mechanical Design Param1ters 80 I A.1 Vermont Yankee Nuclear Power Station Cycle 15 MCPR Operating Limits A-2 A.2 MAPLilGR Limits Versus Average Planar Exposure for A-3

I BP8DRB299 A.3 MAPLilGR Limits Versus Average Planar Exposure for A-4 BD324B

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I A.4 MAPLilGR Limits Vercus Average Planar hposure for I A.5 BD326B A-5 MAPLilGk Limits Versus Average Planar h posure for A-6 BP8DWB311-10GZ A.6 MAPLilGR Limite Versus Average Planar h posure for A-7 I BP8tNB311-110Z I

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ACl2iOWLEDGDiENIS The author and major contributors would like to acknowled p the contributions to this work by U. Ansari, C. Hears and the YAEC Word Processing Center. Their assistance in preparing this document is reco6nized and greatly appreciated.

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g 1.0 INTRODUCIION ,

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. This report provides information to support the operation of the Vermont Yankee Nuclear Power Station through the forthcoming Cycle 15. In this report, Cycle 15 will irequently be referred to as the Reload Cycle. The preceding Cycle 14 will frequently be referred to as the Current Cycle. The f refueling between the two will involve the discharge of 128 irradiated fuel bundles and the insertion of 128 new fuel bundles. The resultant core will consist of 128 new fuel bundles and 240 irradiated fuel bundies. The General Electric Company (GE) manufactured all the 1,andles, except four qualification I fuel bundles manufactured by Advanced Nucl.or Fuels. Some of the irradiated fuel was also present in the reactor in Cycle 13. This cycle will frequently be referred to as the Past Cycle.

This report contains descriptions and analyses results pertaining to the mechanical, thermal-hydraulic, physics, and safety aspects of the Reload Cycle. The analyses assumed the reload core contained all GE bundles.

Section 9.0 describes the Reload Cycle Core Component Qualificatior. Program and its impa:t on the analyt.ea.. The MAPLilGR cnd MCPR operating limits I calculated for the Reload Cycle are given in Appendix A.

included in the Coce Operating Limits Report.

These li'mits vill be I

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,g 2.0 RECmLATACIOR_QtERATING II1EIORY

g 2.1 OperatingJlis10ry_01_the_ Current Cyrle The current operating cycle is Cycle 14. To date, the Current Cycle l

has been operated smoothly at, or near, full power with the exception of sequence exchanges, one short repair outage, two scrams, and a coastdown to the end of cycle. The operating history highlights and control rod sequence exchange schedule of the Current Cycle are found in Table 2.1.1.

2.2 OperatingJ11siory_oLInsLApplicable cvcle The irradiated fuel in the Reload Cycle includes some fuel bundles initially inserted in Cycle 13. This Past Cycle operated smoothly at, or near, full power with the exception of sequence exchanges, two short repair outages, and a coastdown to the Snd of cycle. The operating history I highlights of the Past Cycle are found in Table 2.2.1. The Past Cycle is described in detail in Reference 1.

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I IABLLtd.1 ELCYCLLlk_QIISATING II]GIILIGIIIS Beginning of Cycle Date April 8, 1989 End of Cycle Date September 1, 1990*

I Weight of Uranium As-Loaded (Short Tons)

Beginning of cycle Core Average Exposure (WD/ST) 73.94 9195 End of Full Power Core Average Exposure (MWD /ST) 18340*

End of Cycle Core Average Exposure (MWD /ST) 19595*

Number of Fresh Assemblies 136 Number of Irradiated Assemblies 232 I Control Rod Sequence Exchange Schedules I Sequence Date from In June 3, 1989 A2-1 B1-1 July 29, 1989 B1-1 Al-1 September 23, 1989 Al-1 B2-1 November 18, 1989 B2-1 A2-2 January 6, 1990 A2-2 B1-2 March 21, 1990 B1-2 Al-2 May 19, 1990 Al-2 B2-2 July 7, 1990 B2-2 A2-3 i

• Projected Dates and Exposures.

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I IABLE 2.2 J EY_. CYCLE 13. Offl4IUill[lGilliQjiIS I

Beginning of Cycle Date October 2. 1987 End of Cycle Date February 11. 1989 Weight of Uranium As-Loaded (Short Tons)

I Beginning of Cycle Core Average Exposure (MWD /ST) 74.82 8613 End of Full tower Core Average Exposure (MWD /ST) 16901 End of Cycle Core Average Exposure (MWD /ST) 18307 Number of Fresh Assemblies 136 Ntunber of Irradiated Assemblies 232 I Control Rod Sequence Exchange Schedulet I DILLE frQID Sequence 10 December 5, 1987 A2-1 B1-1 January 30, 1988 B1-1 Al-1 March 19, 1988 Al-1 B2-1 April 30, 1988 B2-1 A2-2 I July 2, 1988 August 24, 1988 A2-2 B1-2 B1-2 Al-2 October 29, 1988 Al-2 B2-2 I

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, am 3.0 RELOAILCOELDESLOLDESCMEn0N

.g 3.1 Cotel uelloading The Reload Cycle core will consist of both new and irradiated assemblies. All the assemblies have bypass flow holes drilled in the lower tie plate. Table 3.1.1 characterizes the core by fuel type, batch size, and first cycle loaded. A description of the fuel is found in Reference 2.

3.2 Design Reference CQIc Loading Pattern

.I The Reload Cycle assembly locations are indicated on the map in Figure 3.2.1. For the sake of legibility only the lower right quadrant is shown. The other quadrants are mirror images with bundlet of the same type having nearly identical exposures. The bundles are identified by the reload number in which they were first introduced into the core. If any changes are made to the loading pattern at the time of refueling, they will be evaluated

under 10CFR50.59. The final loading pattern with specific bundle serial numbers will be supplied in the Startup Test Report.

3.3 Assembly _Extosure_ Dis 1ribution The assumed nominal exposure on the fuel bundles in the Reload Cycle design reference loading pattern is given in Figure 3.2.1. To obtain this exposure distribution, the Past Cycle was depleted with the SIMULATE-3 model (3-4) using actual plant operating history.

I For the Current Cycle, plant operating history was used through January 10, 1990. Beyond this date, the exposure was accumulated using a best-estimate rodded depletion analysis to End of Full Power Life (EOFPL) followed by a projected coastdown to End of Cycle (EOC).

I Table 3.3.1 gives the assumed nominal exposure on the Current Cycle and the Beginning of Cycle (B00) core average exposure that results from the shuffle into the Reload Cycle loading pattern. The Reload Cycle E0FPL core average exposure and cycle capability are provided. ,

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I IAELE 3.1d ASSUtiED_VJ fYCLE_ll_I1!EL.3 LWD 11_IYtES AND NLMBERS Fuel Reload Cycle Designation Designatico Loaded Nurober Irradiated BP8DRB299 R12 13 104 BD326B R13A 14 88 BD324B R13B 14 48 New BP8DWB311-10GZ RI4A 15 64 BP8DWB311-110Z R14D 15 64 I

IAllLE_3dil DESIGNJASIS_VY__ CYCLE _LLAND fYCLE 15_ uroSURES I Assurned End of Current Cycle Core Average Exposure 19.59 GWD/ST Assumed Beginning of Reload Cycle Core Average Exposure 10.71 GWD/ST I Italing Calculated End of Full l'ower Life Reload Cycle Core Average Exposure 19.76 GWD/ST Reload Cycle Full Power Exposure Capability I 9.05 GWD/ST I

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! R12 18.708 R140 0.000 R12 17.186 R14B 0.000 R12 19.318 RIAA 0.000 R12 19.924 R14B 0.000 R13B 12.650 R14A 0.000 R12 19.868 22 R14B R13B R14B R13A R14A R13A R140 R13A R14A R13A R12 0.000 13.552 0.000 13.536 0.000 13.314 22.011 20 0.000 13.498 0.000 9.623 R12 R14B R12 R14A R12 RI4A R13B R140 R13B R13A R12 I 17.161 0.000 18.722 0.000 19.659 0.000 13.420 0.000 10.391 13.928 20.621 18 l

R14B R13A R14A R13B R14A R13A R140 R13A R13A R12 16 0.000 9.602 0.000 13.589 0.000 13.422 0.000 11.650 13.566 22.779 R12 R14A R12 R14A R12 R14B R139 R13A R12 14 19.413 0.000 19.651 0.000 20.042 0.000 11.180 12.574 19.893 I R14A 0.000 R13A 13.822 R14A 0.000 R13A 13.516 R14B 0.000 R13A 13.708 R14A 0.000 R13A 13.345 R12 22.881 12 R12 R140 R13B R14B R13B R14A R13A R13B R12 10 19.792 0.000 13.381 J.000 11.204 0.000 13.633 13.071 23.020 R14B R13A R140 R13A R13A R13A R130 R12 08 0.000 13.870 0.000 11.681 12.645 13.259 13.140 20.622 R13B RIAA R13B R13A R12 R12 R12 12.951 0.000 10,434 13.918 20.052 22.804 22.564 06 R14A R13A R13A R12 13,650 BUND'E 3D- FUEL DESIGNATION 04 0.000 13.814 22.741 I R12 BP8DRB299 R12 R12 R12 ........ BUNDLE ID R13A BD326B R13B BD324D I 20.821 21.938 22.894 ....BOC EXPOSURE g34x R14B spgegg333 3ogg BP8DWB311-11GZ

= = 02 23 25 27 29 31 33 35 37 39 41 43 FIGURE 3.2.1 Yl Cy.c.le_15_DRE.ign Ref erence Loading Pattern. Lower PJght Ouadrant WPP40/10

l 4.0 TVELMLCllANICALANDMD11ALDESIGN 4.1 MechanisaLDesign Most of the fuel to be inserted into the Reload Cycle was fabricated by CE. The major mechanical design parameters are given in Table 4.1.1 and Reference 2. Several design changes have been incorporated in the Reload Cycle fuel design. The new fuel bundles differ from the irradiated bundles in I the following ways: 1) the average bundle enrichment has been changed to 3.11 w/o U-23$12) the ferrule spacer is used and 3) the number of water rods has been changed to one large central rod. Detailed descriptions of the fuel rod mechanical design and mechanical design analyses are provided in Reference 2.

These design analyses remain valid with respect to the Reload Cycle operation. Mechanien1 and chemical compatibility of the fuel bundles with the in-service reactor environment is also addressed in Reference 2.

4.2 IhcrmnLDealen The fuel thermal effects calculations were performed using the i FROSSTEY-2 computer code (5). The TROSSTEY-2 code calculates pellet-to-cladding gap conductance and fuel temperatures from a combination of j theoretical and empirical models which include fuel and cladding thermal expansion, fission gas release, pellet swelling, pelle' densification, pellet cracking, and iuel and cladding thermal conductivity.

The thermal effects analysis included the calculation of fuel temperatures and fuel cladding gap conductance under nominal core steady state e-1 peak linear heat generation rate conditions. Figure 4.2.1 provides the core average response of gap conductance. Thuse calculations integrate the respvases of individual fuel batch average operating histories over the core average exposure range of the Reload Cycle. The gap conductance values are weig..ted axially by power distributions and radially by volume. The core-wide I gap conductance values for the RETRAN system simulations, described in lections 7.1 through 7.3, are from this data set at the corresponding exposure statepoints.

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i The gap conductance values input to the hot channel calculations (Section 7.1) were evaluated f or the given f uel bundle type as a function of the ssembly exposure. The calculation assumed a 1.4 chopped cosine axial power shape with the peak power node running at thu maximum average planar linear heat generation rate (MAPulGR) limit defined in Reference 6.

Figures 4.2.2 and 4.2.3 provide the hot channel response of gap conductance.

In Figures 4.2.2 and 4.2.3, " planar exposure" refers to the exposure of the node operating at the MAPLHGR limit. Gap conductance values for the hot I channel analysis were extracted from the figures using the limiting bundle exposure of any minimum critical power ratio (MCPR) limiting bundle within the exposure interval of interent. The $1MU1 ATE rodded depletion (Section 5.1.2) l provides predictions of both limiting MCPR and the associated bundle exposure for the entire cycle.

Table 4.2.1 provides the core average and hot channel gap conductance values used in the transient analyses (Section 7.1). The values for gap conductance are slightly higher than those calculated in previous cycles for the following reasons: 1) the bundles in the core now have the larger diameter pellet, and 2) the new bundle may operate at a higher, relative power level as provided by the MAPUlGR limits.

Fuel rod local linear heat generation rates (UlGR) at fuel centerline

, incipient melt and 17. cladding plastic strain as a function of local axial I segment exposure for the peak gadolinia concentrations used in Vermont Yankee fuel bundles were calculated. These values are displayed in Table 4.2.2.

Initial conditions assumed that fuel rods operated at the local segment power  !

level of the maximum allowable UiGR prior to the power increase.

4.3 OperatinLI.xperience All irradiated f uel bundles scheduled to be reinserted in the Reload Cycle have operated as expected in Past Cycles of Vermont Yankee. Off-gas measurements in the Current Cycle indicate that a number of fuel rod failures may have occurred. Vermont Yankee is planning to identify the failed rods during the outage.

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I IABLLLI.1 I N OMI NALJU EL .ME CilANICALD E S IGILIAPR1E tells l Tuck 1xpcs Irradiated Ney Ivice-lurned Once-turned I Fuel Bundle

• Bundle Types BP8X8R BP8X8EB GE8XbdB I Vendor Designation (Table 3.1.1)

Initial Enrichment, BP8DRB299 2.99 BD324B and BD326B 3.24 and 3.26 BP8"'JB311-10GZ and BP8DWB311-110%

3.11 I w/o U235 Rod Array fuel Rods per 11undle 8X8 62 BX8 60 8X8 60 Outer Tuel Channel Material Zr-4 Zr-4 Zr-2 I Wall Thickness, inches 0.080 0.080 0.080 I

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• Complete bundle, rod, and pellet descriptions are fcund in Reference 2.

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m M M M M M m m e m m m M IABLE 4.2 J V EYLE_J1_9AR_CONDUCIANCE_ VALUES _USED_LN_TRANS I ER_ ANALYSES Irradia_ted_Fpd_Bundits New_Eucl_Eundles Cycle Exposure Core Average Hot Channel Hot ChannelII) Hot Channel Hot Channe1 I2)

Statepoint Gap Conductance Bundle Exposure Gap Couductance Bundle Exposure Gap Conductance (MWDIST) DLTllLHr:F1 - O (MWDIST) IBTUlJr:Ft -O ____QNDIST) IBIUlHr-Ft. -O BOC15 2740 10746(3) 4750 7788 4880 EOFPL15-2000 MWD /ST 4000 17269 4490 857 5010 EOFPL15-1000 MWD /ST 4090 18608 4370 10005(g) 5080 EOFPLIS 4320 19756 4290 10005(3) 5080 HOIES (1) Hot channel gap conductance values are derived for the BD324B fuel type because it is conservative compared to the other fuel types.

(2) Hot channel gap conductance values are derived for the BP8DWB311-10GZ_

(3) Taken as the highest point in the exposure range.

WPP40/10

I IAfLE34242 IIALLIN MRlfAI _GI2iLMIl0!LMIES_CORLS EONDINGl0 INCIEIMI_f1'EL_CMIERLINE MEQUiGJiD_l% CLADD1HC FIASTIC STMIN(I) 4.0 w/o Gd23 0 0.0 w/o Gd20)

I Exposure LMWUDIT]

Melt LLWLLD 1% e LLH1LE)

Melt LLHLLG 1% e LLY11$)

fMel_DTC.llBXBB BP8DkB299 0 24.0 24.0 21.5 24.0 25,000 24.0 24.0 20.5 20.5 50,000 23.5 15.5 19.0 12.0 0.0 w/o Gd230 5.0 w/o Gd 0 2_3-Exposure Melt 1% e Melt 1% e LMWDlMD LLWLLO LLVLLS] LkWLLO LLHIL(,)

fuellypt_GEBXBEB 24.0 24.0 21.0 23.0 I BD326B 0 and 25,000 24.0 24.0 20.0 20.0 BD324B 50,000 24.0 16.0 19.0 12.0 1 4.0 w/o Gd2y 0 0.0w/oGdf)

Exposure Melt 1% c Melt 1% c I LMWDIRT) LLHLLA) LLHLLE) IkWLLO LLWlb facLLTe_GESKBNB BP8DWB311-11GZ 0 24.0 i.0 21.0 23.5 BP8DWB311-10GZ 25,000 24.0 .0 20.0 20.0 50,000 20.5 . 5 16.5 11.0 l

l NGIE (1) Peak linear heat generation rates shown are minimum bounding values to the occurrence of the given condition.

WPP40/10

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EIGURE 4.2.1 yy Ovele 15 Core Average Cap Conductance Versus Ovele Expotur.c WPP40/10

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ricURE 4.L 2 VY Mot ChanacLGap_.Canducianee f or GESK8NB _ Versus AP.2AnIA WPP4J/10 1

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(:1. ,1NWn18) pourpnpuoo dep leuurgo loH I ElqURE 4.2.3 I VY Hot channel cap conduttance _ f or cE8X8EB Versus ExPaauIA WPP40/10

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• L 5.0 NUCLEAR DESIGN I

l 5.1 Cor.c_. foyer Dis t r.ihntions I The Reload Cycle was depleted using SIMUI. ATE-3 (3-4) to give both a l

rodded depletion and an All Rods Out (ARO) llaling depletion. l 5.1.1 lin11ng Poxer Distrib.ution g The llaling depletion serves as the basis for defining core reactivity l characteristics for most transient evaluations. This is primarily because its flat power shape has conservatively weak scram characteristics.

The llaling power distribution is calculated in the ARO condjtion. The llaling iteration converges on a self-consistent power and exposure

, distribution for the burnup step to EOFPL. In principle, this should provide g the overall minimum peaking power shape for the cycle. During the actual B cycle, flatter power distributions might occasionally be achieved by shaping with control rods. Ilowever, such shaping would leave underburned regions in the core which would peak at another point in time. Figures 5.1.1 and 5.1.2 give the llaling radial and axial average power distributions for the Reload Cycle.

5.1.2 Rodded _nepletion P.nwer Distribution The rodded depletion was used to evaluate the mislocated bundle error and the rod withdrawal error because it provides initialising rod patterns and it provides more realistic predictions of initial CPR values. It was also used in the rod drop worth and shutdown margin calculations because it burns the top of the core more realistically than the llaling depletion.

I To generate the rodded depletion, control rod patterns were developed which give critical eigenvalues at each point in the cycle and peaking similar to the llaling calculation. The resulting patterns were frequently more peaked than the llaling, but were below expected operating limits. Ilowever, as stated WPP40/10 l

l

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L above, the underburned regions of the core can exhibit peaking in excess of the llaling peaking when pulling ARO at EOFPL. Figures 5.1.3 and 5.1.4 give b the ARO at EOFPL power distributions for the Reload Cycle rodded depletion.

Note, in Figure 5.1.4, that the average axial power at ARO for the rodded

( depletion is more bottom peaked than the llaling (Figure 5.1.2). The rodded depletion would result in better scram characteristics at EOFPL.

5.2 Cort _Exporturt_ Distributions The Reload Cycle exposures are summarized in Table 3.3.1. The projected BOC radial exposure distribution for the Reload Cycle is given in Figure 3.2.1. The llaling calculation produced the EDTPL radial exposure distribution given in Figure 5.2.1. Since the llaling power shape is constant, it can be held fixed by SIMULATE-3 to give the exposure distributions at various mid-cycle points. B00. E0FPL-2000 MWD /ST, EOFPL-1000 MWD /ST, and EOFPL exposure distributions were uwd to develop reactivity input f or the core wide transient analyses.

The rodded depletion differs from the llaling during the cycle because the rode shape the power differantly. Powever, rod sequences are swapped frequently and the overall exposure distribution at end of cycle is similar to the llaliag. Figure 5.2.2 gives the EOFPL radial exposure distribution for the Reload Cycle rodded depletion.

5.3 Co1LShutdown Margin Technical Specifications [7] state that, for sufficient shutdown margin, the core must be suberitical by at least 0.25% AK + R (defined below) with the strongest worth control rod withdrawn. Using SIMULATE-3, a search was made for the strongest worth control rod at various exposures in the cycle. This is necessary because rod worths change with exposure on adjacent assemblies. Then the cold K,gg with the strongest rod out was calculated at B00 and at the end of each control rod sequence. Subtracting each cold K egg with the strongest rod out from the cold critical K,gg eigenvalue defines the shutdown margin as a function of exposure. Figure 5.3.1 shows the results.

WPP40/10

The cold critical eigenvalue K,gg was defined as the average calculated critical eigenvalue minus a 95% confidence level uncertainty. Then all cold results were normalized to make the critical K,gg eigenvalue equal to 1.000.

{ Because the local reactivity may increase with exposure, the shutdown margin (SDM) may decrease. To account for this and other uncertainties, the value R is calculated. R is defined as Ri plus R2 . Rt is the difference between the cold K,gg with the strongest rod out at BOC and the maximum cold I Kegg with the strongest rod out in the cycle. R2 is a measurement uncertainty in the demonstration of SDM associated with the manufacture of past control blades. It is presently set at .07% AK. The shutdown margin results are summarized in Table 5.3.1.

j 5.4 Sinndbyli quid _ Cont roLSyJLt_cnLShutdom_ Cap abilit y i The shutdown capability of the Standby Liquid Control System (SLCS) is designed to bring the reactor from full power to cold ARO, xenon free shutdown with at least 5% AK margin. Using SIMULATE-3, the ppm of boron was adjusted until the K,gg reached the cold critical K,gg minus .05. Each case I assumed cold, xenon-free conditions, with All Rods Out. The Reload Cycle was searched to find the most reactive point in the cycle. This analysis found that the plant would be suberitical by 5% AK at the worst point in cycle with ices than the 800 ppm of boron required by VY Technical Specifications.

Table 5.4.1 lists the amount of boron concentration and the corresponding shutdown capability of the SLCS.

I 5.5 Maximum rw for the Spent Futi Pool i Section 5.5E of the Technical Specification requires that the 4 for any bundle stored in either the new fuel vault or the spent fuel pool not exceed 1.31 to ensure compliance with the K,gg safety limit of 0.95. The bundles used in the Reload Cycle do not exceed the specifications in Section 5.5E, as shown in Table 5.5.1. These values are obtained from CASMO-3G [8).

WPP40/10

l I&hlE._i d d p VLCICLE 15 L K di VALUES AND SHUTDOWN MARGIN CALCULATION Cold C 4tical K,gg Eigenvalue 1.0000 BOC K,gg - Controlled With

( Strongest Worth Rod Withdrawn .9808 Cycle Minimum Shutdown Margin cucurs at B00 With Strongest Worth Rod Withdrawn 1.92% AK R1 , Maximum Increase in Cold K,gg With Exposure .00% AK IADIE_. bad VY_ CYCLE 15 STANDBY LIOUlD CONTROL SYSTEM BilllIDOWN CAPABILITY ppm of BQIDH ShuldQwn Margin 609 5.0% AK 800 8.55% AK IABLE 5.5 1

( VY CYCLE 15 MAXniUM COLD rw 0F.ANY_EN21CllED SECMENT Aundin_'LYrs MaximunLL BPDDRB299 1.19 BD324B ' 00 BD326B 1.

BP8DWB311-100Z 1.20 BP8DWB311-11GZ 1.20

,[ L WPP40/10

I g n,2 n14B R12 eide R12 R14A n12 R14B R13e Rt A R12 1.124 1.399 1,169 1.399 1.117 1.343 1.046 1.231 0.992 0.895 0.472 22 R148 R13B R148 R13A RIAA R13A R148 R13A R14A R13A R12 1.399 1.248 1.413 1.291 1.3 64 1.199 1.301 1.084 1.081 0.759 0.419 20 R12 R14B R12 R14A R12 RIAA R13B R14B R13B R13A R12 1.168 1.413 1.161 1.399 1.113 1.342 1.136 1.174 0.921 0.663 0.364 18

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I R14B R13A RIAA R130 R14A R13A R14B R13A R13A R12 i

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1. m 1. m 0. m 0.u6 1e 1.398 1.218 1.355 1.167 1.233 0.978 j R12 R14A R12 R14A R12 R14B R13B R13A R12 1.114 1.382 1.111 1.350 1.063 1.263 1.054 0.872 0.567 14 R14A R13A R14A R13A R14B R13A R14A R13A R12 I i.340 1.193 1.339 1.163 1.262 1.054 1.064 0.759 0.43/ 12 R12 R140 R13B R14B R13B R14A R13A R13B R12 1.045 1.298 1.133 1.231 1.052 1.063 0.829 0.626 0.345 10 R14B R13A R148 R13A R13A R13A R13B R12 1.227 1.075 1.170 0.975 0.868 0.758 0.624 0.381 08 R130 R14A F.13B R13A R12 R12 R12 I 0.9 85 1.075 0.916 0.764 0.561 0 434 0.348 06 i R14A 0.887 R13A c.747 R13A 0.660 R12 0.444 BUNDLE ID R12 FUEL DESIGNATION BP8DRB299 04 U

R12 R12 R12 ........ BUNDLE ID R1 B B 0.456 0.413 0.355 ........ EOFPL RELATIVE POWER R14A BP8DWB311-10GZ 02 R14B BP8DWB311-11GZ 23 25 27 29 31 33 35 37 39 41 43 l m m s.1.1 YY Cycle _15 Haling DepleMpn. EOFPL Bundle Average. Relative Pow ns

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, 1.206 1.523 1.249 1.495 1.170 1.387 1.054 1.223 0.969 0.876 0.455 R14B R13B R14B R13A R14A R13A R14B R13A R14A R13A R12 0.732 0.401 20 1.514 1.330 1.515 1.360 1.448 1.214 1.307 1.061 1.059 R12 R14B R12 R14A R12 R14A R13B R14B R13B R13A R12 10 1.239 1.509 1.214 1.467 1.144 1.353 1.121 1.149 0.888 0.634 0.345 R14B R13A R14A R13B R14A R13A R14B R13A R13A Ri2 16 1.481 1.347 1.459 1.242 1.369 1.149 1.208 0.941 0.735 0.422

.- R12 R14A R12 R14A R12 R14B R138 R13A R12 1.156 1.432 1.364 1.059 1.238 1.016 0.831 0.535 I4 1.134 R14A R13A R14A R13A R14B R13A R14A R13A R12 1.368 1.339 1.235 1.014 1.025 0,719 0.409 12 1.195 1.141 R12 R118 R138 R14B R13B R14A R13A R13B R12 I 1.043 1.291 1.110 1.200 1.012 1.023 0.786 0.589 0.320 10 R148 R13A R14B R13A R13A R13A R13B R12 1.209 1.044 1.137 0.932 0.825 0,717 0.587 0.356 08 R13B R14A R138 R13A R12 R12 R12 0.956 1.045 0.878 0.724 0.527 0.406 0.322 06 R14A R13A R13A R12 FUEL DESIGNATION BUNDLE 3 l 0.864 0.717 0.626 0.417 R12 BP8DRB299 R13A BD326B R12 R12 R12 . .. ... BUNDLE ID R13B BD324B

.. . .. EOFPL RELATIVE POWER R14A BP8 "1 02 0.438 0.394 0.334 8 1-23 25 27 29 31 33 35 37 39 - 41 43 FIGURE 5.1.3 VY Cycle 15 Rodded _Hepletion - ARO at EOFPL.

Bundle Average Reintive Powers WPP40/10 r

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" Core Ay m ce Axial Power Distribution WPP40/10 f

R12 R14B R12 R14B R12 RI4A R12 R14B R13B R14A R12 23.985 22-28.525 12.771 27.391 12.771 29.069 12.247 29.053 11.236 21.954 8.161 R14B R13B R14B R13A R14A R13A R148 R13A R14A R13A R12 20 12.773 24.950 12.904 21.428 12.629 24.511 11.879 23.445 9.856 20249 25.669 i R12 R14B R12 R14A R12 R14A R13B R14B R13B R13A R12 10 12.900 28.B57 12.766 29.376 12.239 23.838 10.713 18.840 19.985 23.802 27.365 R14B R13A R14A R13B RIAA R13A R14B R13A R13A R12 20,623 10 12.760 21.401 12.760 24.764 12.323 24.088 11.256 20.593 26.675 R12 R14A R12 R14A R12 R148 R13B R13A R12 14 29.137 12.613 29.357 12.314 29.328 11.530 20.848 20.541 24.844 R14A R13A R14A R13A R14B R13A R14A R13A R12 12 12.227 24.730 12.217 24.218 11.519 23346 9.704 20.279 26.692 R12 R148 R13B R14B R13B R14A R13A R13B R12 i 28,916 11.849 23.776 11.233 20.855 9.697 21.211 18.810 26.029 10 R14B R13A R140 R13A R13A R13A R13B R12 i 11.195 23.699 10.674 20.591 20.584 20.188 18.866 23.947 08 R13B R14A R138 R13A R12 R12 R12 06 21.986 9.801 18.837 20.903 24.946 26.594 25.601 i R14A 8.088 R13A 20.640 R13A 19.680 R12 26.618 DUNDLE -ID FUEL DESIGNATION 04 R12 BP8DRB299 R ^

R12 R12 R12 ...... BUNDLE ID BD32 B R13B

.. ...... EOFPL EXPOSURE R14A BPDDWB311-10GZ 02 24.799 25.542 25,992 R14B BP8DWB311-11GZ 23 25 27 29 31 33 35 37 39 41 43 FIGURE 5 2.d VY Ovele 15 Hn11nc Deoletion. EOFPL Bundle Average Exoosures I

p 22 1B f

I I R12 R14B R12 R14B R12 R14A R12 R14B R13B R14A R12 22 10.807 26.458 11.095 28.512 11.441 29.152 11.208 22.603 8.320 24.357 27.688 R13B R14B R13A R14A R13A R14B R13A R14A R13A R12 R14B 20 11.445 20.800 11.526 24.706 11.697 24.189 10.171 20.941 26.086 11.041 24.238 R12 R14A R12 R14A R138 R14B R13B R13A R12 R12 R14B 18 11.653 29.186 12.052 24.471 11.044 19.738 20.760 24.279 26.680 11.554 28.432 R14A R13A R14B R13A R13A R12 I

R14B R13A R14A R13B 16 11.880 24.814 11.966 24.819 11.619 21.608 21.608 27.282 11.318 21.052 I

I R12 28.775 R14A 11.851 R12 29.375 R14A 12.028 R12 29.710 R14B 11.866 R13B 21.878 R13A 21.617 R12 25.661 14 R14A R13A R14A R13A R14B R13A R14A R13A R12 12 11.727 25.132 12.215 25.010 11.874 24.401 10.389 21.286 27.361 R12 R148 R13B R14B R138 R14A R13A R13B R12 10 29.211 11.870 24.548 11.668 21.914 10.395 22.297 19.733 26.619 R14B R13A R14B R13A R13A R13A R13B R12 08 11.324 24.586 11.150 21.674 21.696 21.213 19.794 24.557 R13B R14A R13B R13A R12 R12 R12 06 22.733 10.247 19.839 21.939 25.794 27.278 26.208

I R14A R13A R13A R12 BUNDLE ID FUEL DESIGNATION 27.260

g 8.317 21.389 20.513 R12-R13A BP8DRB299 BD326B I" '

R12 R12 R12 ........ BUNDLE 10 R13B BD324B

........ EOFPL EXPOSURE R14A BP8DWB311-10GZ 02 25.193 25.988 26.493 R14B BP8DWB311-11GZ

!I 23 25 27 29 31 33 35 37 39 41 43

I FIGURE 5.2 2 a &

WPP40/10 ,

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(%) NIOUW4 NMOO10HS o q q T, o9No "o,oo I FIGURE 5.3.1 VY Ovele 15 Cold Shutdown AK in Percent Versus Cvele Exposure WPP40/10 l l

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6.0 IEBML-11EMVLIC DESIGN 6.1 Steady-State _IhcImal Hydraulics I

l Core steady-state thermal-hydraulic analyses for the Reload Cycle were performed using the FIBWR [9-11) computer code. The FIBWR code incorporates a detailed geometrical representation of the complex flow paths in a BWR core.

and explicitly models the leakage flow to the bypass region and water rod I flow. The FIBWR geometric models for each GE bundle type were benchmarked against vendor-supplied and plant thermal-hydraulic information.

I l Using the fuel bundle geometric models, a power distribution calculated by SIMULATE-3 (3-4) and core inlet enthalpy, the FIBWR code calculates the core pressure drop and total bypass flow for a given total core flow. The core pressure drop and total bypass flow predicted by the FIBWR code were then used in setting the initial conditions for the system transient analysis model.

6.2 Reac. tor _. Limits _Determina11.on The objective for normal operation and anticipated transient events is to maintain nucleate boiling. Avoiding a transition to film boiling protects the fuel cladding integrity. The Fuel Cladding Integrity Safety Limit (FCISL) for Vermont Yankee is a Critical Power Ratio (CPR) of 1.07 [2]. CPR is defined as the ratio of the critical power (bundle power at which some point within the assembly experiences onset of boiling transition) to the operating bundle power. Thermal margin is stated in terms of the minimum value of the i Minimum Critical Power Ratio (MCPR) which corresponds to the most limiting fuel assembly in the core. Both the transient (safety) and normal operating thermal limits, in terms of MCPR, are derived with the GEXL-Plus correlation I

as described in Reference 12.

The Reload Cycle fuel has Linear Heat Generation Rate (LHGR) limits of 13.4 kW/ft for Bundle Type BP8X8R, 14.4 kW/ft for Bundle-Types GE8X8EB and GE8X8NB. The basis for these Maximum LHGR (MLHGR) limits can be found in Reference 2.

WPP40/10

i

' The Reload Cycle fuel has Average Planar Linear llent Generation Rate (APLl!GR) limits shown in Appendix A. The Maximum APLllGR (MAPLilGR) values are the most limiting composite of the fuel mechanical analysis MAPLilGRs and the LOCA analysis MAPLilGRs. The fuel mechanical design analysis, using the methods in Reference 2. demonstrate that all fuel rods in a lattice, operating at the bounding power history, meet the fuel design limits specified in Reference 2. The transients described in Section 7.0 were analyzed to verify that design criteria in the mechanical design analysis methods was not exceeded during the transient. The LOCA analysis is described in Section 8.0.

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L 7.0 ACCIDENT ANALYSIS 7.1 Iranshnt_AnalJnis l Transient simulations are performed to assess the impact of certain transients on the heat transfer characteristics of the fuel. It is the purpose of the analysis to determine the MCPR operating limit, such that the FCISL is not violated for the transients considered.

I 7.1.1 Melhohlogy The analysis requires two types of simulations. A system level simulation is performed to determine the overall plant response. Transient core inlet and exit conditions and normalized power from the system icvel calculation are then used to perform detailed thermal-hydraulic simulations of the fuel, referred to as " hot channel calculations." The hot channel simulations provide the bundle transient ACPR (the initial bundle CPR minus I the MCPR experienced during the transient).

The system level simulations are performed with the model documented in References 13 through 15.

I The hot channel calculations are performed with the RETRAN [16-17] and TCPYA01 (18,11,15] computer codes. The GEXL-Plus correlation (12] is used in TCPYA01 to evaluate critical power ratio. The calculational procedure is outlined below.

The hot channel transient ACPR calculations employ a two-part process, as illustrated by the flow chart in Figure 7.1.1. The first part involves a series of steady-state analyses performed with the FIBWR, RETRAN, and TCPYA01 computer codes. The FIBWR analyses utilize a one-channel model for each fuel type being analyzed, with bypass and water rod flow also modeled. The steady-state FIBWR analyses were performed at several power levels with other WPP40/10

I conditions (i.e., core pressure drop, system pressure, and core inlet I enthalpy) held constant. The FIBWR code result is an active channel flow (AF) and bypass flow (BPF) for each active channel power (AP). j I The FIBWR conditions for channel power, channel flow, and bypass flow i

(

were then used as input to steady-state RETRAN/TCPYA01 hot channel calculations. Other assumptions are consistent with those in the FIBWR analysis. The Initial Critical Power Ratio (ICPR) is the key result for each steady-state RETRAN/TCPYA01 analysis. These results allow for the development of functional relationships, describing AP as a function of ICPR, and AF and I BPF as functions of AP for each fuel type. These relationships are used in the iterative process used during the transient calculations as described i below and shown in Figure 7.1.1.

The second part iterates on the hot channel initial power level. This is necessary because the ACPR for a given transient varies with Initial Critical Power Ratio (ICPR). Ilowever, only the ACPR corresponding to a transient MCPR equal to the FCISL limit (i.e., 1.07 + ACPR = ICPR) is appropriate. The approximate constancy of the ACPR/ICPR ratio is useful in I these iterations. Each iteration requires a RETRAN hot channel run to calculate the transient enthalpies, flows, pressure and saturation properties at each time step. These are required for input to the TCPYA01 code. TCYPA01 is then used to calculate a CPR at each time step during the ttansient, from which a transient ACPR is derived. The hot channel model assumes a chopped l cosine axial power shape with a peak / average ratio of 1.4.

As noted in Section 6.1, analyses for the Reload Cycle included benchmarking the FIBWR model against vendor-supplied thermal-hydraulic information. Therefore, the FIBWR results of AF and BPF for a given AP and core pressure drop are passed directly to RETRAN. As shown in Figure 7.1.1, the current iterative process involves a single loop.

WPP40/10

l L 7.1.2 Inllini Conditions and Assumptions The initial conditions for the system simulations are based on maximum

, turbine capacity of 10$% of rated steam flow. The corresponding reactor conditions are 104.5% core thermal power and 100% core flow. The core axial power distribution for each of the exposure points is based on the 3-dimensional SIMU1 ATE-3 [3-4) predictions associated with the generation of the reactivity data (Section 7.1.3). The core inlet enthalpy is set so that the amount of carryunder from the steam separators and the quality in the liquid region outside the separators is as close to zero as possible. For I fast pressurization transients, this maximizes the initial pressurization rate and predicts a more severe neutron power spike. A summary of the initial operating state used for the system simulations is provided in Table 7.1.1.

Vermont Yankee operators adjust core flow during the cycle for short-tem maneuvering. During this type of operation, core flow may be as low as 87% while at 100% power. To ensure the safety analysis bounds these conditions, transients are reanalyzed at the limiting exposure statepoint (limiting in terms of an increase in ACPR) corresponding to these conditions.

I These analyses are performed at both the " Measured" and the "67B" scram times. The ACPR penalty (defined as the difference in ACPR) generated during this reanalysis is applied to the applicable transient ACPR results.

Assumptions specific to a particular transient are discussed in the section describing the transient. In general, the following assumptions are made for all transients:

1. Scram setpoints are at Technical Specification (7) limits.

1

2. Protective system logic delays are at equipment specification limits.

3. Safety / relief valve and safety valve capacities are based on Technical Specification rated values.

WPP40/10

lI

4. Safety / relief valve and safety valve setpoints are modeled as being at the Technical Specification upper. limit. Valve responses are based on slowest specified response values.

I 5. Control rod drive scram speed is based on the Technical Specification limits. The analysis addresses a dual set of scram speeds, referred to as the " Measured" and the "67B" scram times.

" Measured" refers to the faster scram times given in I Section 3.3.C.1.1 of the Technical Specifications. "67B" refers to the slower scram times given in Section 3.3.C.1.2 of the Technical Specifications.

7.1.3 One-DimensiongL Cross Sections and Kinglics Paramettra The methods used to generate the one-dimensional (1-D) cross sections and kinetics parameters as a function of fuel temperature, moderator density, moderator temperature, and scram are described in detail in Reference 19. The

!I method is outlined below.

A complete set cf 1-D cross sections, 1-D kinetics parameters, the axial power distribution, and the kinetics parameters are generated from base states established for EOFPL, EOFPL-1000 MWD /ST, EOFPL-2000 MWD /ST, and BOC-exposure statopoints. These statepoints are characterized by exposure and I void history distributions, control rod patterns, and core thermal-hydraulic conditions. The latter are consistent with the assumed system transient conditions provided in Table 7.1.1.

l The B00 base state is established by shuffling from the previously defined Current Cycle endpoint into the Reload Cycle loading pattern. A f . criticality search provides an estimate of the BOC critical rod pattern. The E0FPL and intermediate core exposure and void history distributions.are l calculated with a Haling depletion as described in Section 5.2. The EOFPL I

state is unrodded. As such, it is defined sufficiently. However,.the EOFPL-1000 MWD /ST and EOFPL-2000 MWD /ST exposure statw oints require base l' control rod patterns. These are developed to be as " black and white" as WPP40/10 l

1 l

L l

{

I

, possible. That is, beginning with the rodded depletion configuration, all control rods which are more than half inserted are fully inserted, and all control rods which are less than half inserted are fully withdrawn. If the SIMULATE-3 (3-4] calculated parameters are within operating limits, then this configuration becomes the base case. If the limits are exceeded, a minimum number of control rods are adjusted a minimum number of notches until the parameters fall within limits. Using this method, the control rod patterns I and resultant power distributions minimize the scram reactivity and maximize the core average moderator density reactivity coefficient. For the events analyzed, this tends to maximize the transient power response.

l At each exposure statepoint, a SIMULATE-3 initial control state reference case is run. A series of perturbation cases are run with SIMULATE-3 to independently vary the fuel temperature, moderator temperature, and core pressure. All other variables normally associated with the SIMULATE-3 cross sections are held constant at the reference state. To obtain the effect of I the control rod scram, another SIMULATE-3 reference case is run with all-rods-in. The perturbation cases described above are run again from this reference case. For each control state, a data set of kinetics parameters and cross sections is generated as a function of the perturbed variable. There is a table set for each of the 27 neutronic regions, 25 regions to represent the active core and one region each for the bottom and top reflectors.

l Figures 7.2.1 through 7.2.6 show the transient response of scram reactivity in the " Measured" scram time analyses.

7.1.4 Iransients Annivzed Past licensing analysis has shown that the transients which result in the minimum core thermal margins are:

1. Generator load rejection with complete failure of the turbine bypass system.

2. Turbine trip with complete failure of the turbine bypass system.

WPP40/10 I

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3. Loss of feedwater heating.

The "feedwater controller failure" (maximum demand) transient is not a limiting transient for Vermont Yankee, because of the plant's 110% steam flow l bypass system. Past analyses have shown this transient to be considerably less limiting than any of the above for all exposure points. The events reported herein are limiting; no other transients would produce more restrictive MCPR operating limits for the Reload Cycle. Brief descriptions I and the results of the transients analyzed are provided in the following section.

7.2 Iransient Anahsis_Renks The transients selected for consideration were analyzed at exposure points of E0FPL, EOFPL-1000 MWD /ST, and EOFPL-2000 MWD /ST; the loss of feedwater heating transient was also evaluated at BOC conditions. The transient results reported in Table 7.2.1 correspond to the limiting bundle I type in the core. The MCPR limits in Table 7.2.1 are calculated by adding the calculated ACPR to the FCISL.

7.2.1 IurbinedripJilhout Bvynss Trangient (TWOBP) i The transient is initiated by a rapid closure (0.1 second closing time) of the turbine stop valves. It is assumed that the steam bypass valves, which normally open to relieve pressure, remain closed. A reactor protection system signal is generated by the turbine stop valve closure switches. Control rod I drive motion is conservatively assumed to occur 0.27 seconds after the start of turbine stop valve motion. The AWS recirculation pump trip is assumed to occur at a setpoint of 1150 psig dome pressure. A pump trip time delay of 1.0 second is assumed to account for logic delay and M-G set generator field j collapse. In simulating the transient, the bypass piping volume up to the valve chest is lumped into the control volume upst.eam of the turbine stop valves. Predictions of the salleat system parameters at the three exposure l points are shown in Figures 7.2.1 through 7.2.3 for the " Measured" scram time analysis.

WPP40/10

I 7.2.2 Generaint_Loadlej eclionlithonLlypas s_ Iran s i e n t (GLRWOBP)

The transient is initiated by a rapid closure (0.3 seconds closing time) of the turbino control valves. As in the case of the turbine trip transient, the bypass valves are assumed to fall. A reactor protection system signal is generated by the hydraulic fluid pressure switches in the acceleration relay of the turbine control system. Control rod drive motion is conservatively assumed to occur 0.28 seconds after the start of turbine control valve motion. The same modeling regarding the A'NS pump trip and bypass piping is used as in the turbine trip simulation. The influence of the accelerating main turbine generator on the recirculation system is simulated by specifying the main turbine generator electrical frequency as a function of time for the M-G set drive motors. The main turbine generator frequency curve is based on a 100% power plant startup test and is considered representative for the simulation. The system model predictions for the three exposure points are shown in Figures 7.2.4 through 7.2.6 for the " Measured" scram time analysis.

t I

7.2.3 Lags of Feedwater ligating Transient (LOWlO A feedwater heater can be lost in such a way that the steam extraction line to the heater is shut off or the feedwater flow bypasses one of the heaters. In either case, the reactor will receive cooler feedwater, which l

will produce an increase in the core inlet subcooling, resulting in a reactor power increase.

The response of the system due to the loss of 1000F of the feedwater

! heating capability was analyzed. This represents the maximum expected feedwater temperature reduction for a single heater or group of heaters that f can be tripped or bypassed by a single event.

E l

Vermont Yankee has a scram setpoint of 120% of rated power as part of-the Reactor Protection System (RPS) on high neutron flux. In this analysis, no credit was taken for scram on high neutron flux, thereby allowing the WPP40/10 l

t

L reactor power to reach its peak without scram. This approach was selected to F

provide a bounding and conservative analysis for events initiated from any l power level.

I The transient response of the system was evaluated at several exposures l

during the cycle. The transient evaluation at EOFPL-1000 MWD /ST was found to be the limiting case between B00 to EOFPL. The results of the system response to a loss of 1000F feedwater heating capability evaluated at I E0FPL-1000 mwd /ST as predicted by the RETRAN code are presented in Figure 7.2.7.

7.3 QYRIPIts.surization Analysis Resulta I Compliance with ASME vessel code limits is demonstrated by an analysis l

of the Main Steam Isolation Valves (MSIV) closing with failure of the MSIV position switch scram. EOFPL conditions were analyzed. The system model used is the same as that used for the transient analysis (Section 7.1.1). The I initial conditions ar, modeling assumptions discussed in Section 7.1.2 are applicable to t'.is simulation.

The transient is initiated by a simultaneous closure of all four MSIVs. A 3.0 second closing time, which is the Technical Specification i minimum, is assumed. A reactor scram signal is generated on APRM high flux.

Control rod drive motion is conservatively assumed to occur 0.28 seconds after reaching the high flux setpoint. The system response is shown in Figure 7.3.1 for the " Measured" scram time analysis.

i The maximum pressures at the bottom of the reactor vessel calculated for the " Measured" scram time analysis and for the "67B" scram time analysis are given in Table 7.3.1. These results are within the allowable code limit of 10% above vessel design pressure for upset conditions,- or 1375 psig.

WPP40/10

I 7.4 Lcqn1 Rod Withdrawal.. Error _ Transient Results

.I The rod withdrawal error (RWE) is a local core transient caused by an operator erroneously withdrawing a control rod in the continuous withdrawal mode. If the core is operating at its operating limits for MCPR and LHGR at the time of the error, then withdrawal of a control rod could increase both local and core power levels with the potential for overheating the fuel.

I There is a broad spectrum of core conditions and control rod patterns which could be present at the time of such an error. For most normal situations it would be possible to fully withdraw a control rod without exceeding 1% clad plastic strain or violating the FCISL.

To bound the most severe of postulated rod withdrawal error events, a portion of the core MCPR operating limit envelope is specifically defined stich that the cladding limits are not violated. The consequences of the error I depend on the local power increase, the initial MCPd of the neighboring locations and the ability of the Rod Block Monitor (RBM) System to stop the withdrawing rod before MCPR reaches the FCISL.

I The most severe transient postulated begins with the core operating according to normal procedures and within normal operating limits. The operator makes a procedural error and attempts to fully withdraw the maximum worth control rod at maximum withdrawal speed. The core limiting locations are close to the error rod. They experience the spatial power shape transient as well as the overall core power increase.

The core conditions and control rod pattern are conservatively modeled I for the bounding case by specifying the following set of concurrent worst case assumptions:

1. The rod should have high reactivit/ worth. This is provided for by  ;

analysis of the core at several exposure points around the core peak reactivity. The test patterns are developed with xenon-free conditions. The xenon-free condition and the additional control WPP40/10 l l

i l

l

rod inventory needed to maintain criticality exaggerates the worth of the withdrawn control rod when compared to normal operation with nonnal xenon levels.

2. The core is initially at 104.5% power and 100% flow.

I 3. The core power distribution is adjusted with the available control rods to place the locations within the four by four array of I bundles around the error rod as close to the operating limits as possible.

4. Of the many patterns tested, the pattern with the highest ACPR results is selected as the bounding case.

The RBM System's ability to terminate the bounding case is evaluated on the following bases:

I 1. Technical Specifications allow each of the separate RBM channels to remain operable if at least half of the Local Power Range Monitor (LPRM) inputs at every level are operable. For the interior RBM channels tested in this analysis, there are a maximum of-four LPRM inputs per level. One RBM channel averages the inputs from the A and C levels; the other channel averages the inputs from the B and I

l D levels. Considering the inputs for a single-channel, there are eleven failure combinations of none, one and two fallad LPRM strings. The RBM channel responses are evaluated separately at I

l these eleven input failure conditions. Then, for each channel taken separately, the lowest response'as a function of error rod position is chosen for comparison to the RBM setpoint.

l l-

2. The event is analyzed separately in each of the four quadrants of the core due to the differing LPRM string physical locations.

relative to the error rod.

l WPP40/10 l

l

I Technical Specifications require that both RBM channels be operable during normal operation. Thus, the first channel calculated to intercept the RBM setpoint is assumed to stop the rod. To allow for control system delay times, the rod is assumed to move two inches after the intercept and stop at the following notch.

The analysis is performed using SIMULATE-3 (3-4]. Two separate cases are presented from numerous explicit SIMULATE analyses. The reactor conditions and case deceriptions are shown in Figures 7.4.1 and 7.4.2. Case 1 analyzes the bounding event with zero xenon at the most reactive point in the cycle for the worst case abnormal rod pattern configuration. Case 2 is the worst of the 104.5% power conditions modeled with more normal control rod patterns and equilibrium xenon. The transient results, the ACPR and maximum linear heat generation rate (MLHGR) values, are also shown in Figures 7.4.1 and 7.4.2. The ACPR values are evaluated such that the implied MCPR operating limit equals FCISL + ACPR. This is done by conserving the figure of merit

. (ACPR/ICPR) shown by the SIMULATE calculations. The use of this method provides valid ACPR values in the analysis of normal operating states where locations near the assumed error rod are not initially near the MCPR operating limit.

Case 2 is the worst of all the rod withdrawal transients analyzed from 104.5% power, full flow and normal rod pattern conditions. Case 2 is bounded by Case 1 with substantial MCPR margin. The Case 1 RBM channel responses are shown in Figures 7.4.3 and 7.4.4. They also show the. control rod position at the point where the weakest RBM channel response first intercepts the RBM setpoint. For this same bounding case, the operating limit ACPR envelope component versus RBM setpoint is taken from Figure 7.4.1. The same figure shows the resultant LHGR assuming the limiting bundle is placed on the operating limit of 14.4 kW/ft prior to the withdrawal. The calculation includes the 2.2% power spiking penalty. The limiting bundle MLHGR demonstrates margin to the 1% plastic strain limit given the low exposure of the limiting bundle. High exposure bundles which have low 1% plastic strain limits are never limiting.

I ~39 WPP40/10

7.5 Misloaded Bundle Error __ Analysis _ Resgl.ts 7.5.1 Rotated Bundle Err.or I The primary result of a bundle rotation is a large increase in local pin peaking and R-factor as higher enrichment pins are placed adjacent to the surrounding wide water gaps. In addition, there may be a small increase in reactivity, depending on the exposure and void fraction states. The R-factor increase results in a CPR reduction, while the local pin peaking factor increase results in a higher pin LilGR. The objective of the analysis is to

'I ensure that, in the worst possible rotation, the LilGR and CPR safety limits are not violated with the most limiting monitored bundles on their operating

limits.

To analyze the CPR response, rotated bundle R-factors as a function of exposure are developed by adding the largest possible AR-factor resulting from a rotation to the exposure dependent R-factors of the properly oriented bundles (12]. Using these rotated bundle R-f actors, the MCPR values resulting from a bundle rotation are determined using SIMULATE. This is done for each control rod sequence throughout the cycle. The process is repeated with the K-infinity of the limiting bundle modified slightly to account for the increase in reactivity resulting from the rotation. For each sequence, the MCPR for the properly oriented bundles is adjusted by a ratio necessary to place the corresponding rotated CPR on its FCISL. The maximum of these adjusted MCPRs is the rotated bundle operating limit.

To determine the MLl!GR resulting from a rotation, the ratios of the maximum rotated bundle local peaking factor to the maximum properly oriented bundle local peaking are determined for the expected range of exposure and void conditions. The maximum of this ratio is applied to the Ll!GR operating limits of 13.4 kW/ft and 14.4 kW/ft. This maximum rotated bundle LilGR is, in addition, modified to account for the possible reactivity increase resulting from the rotation. It is also increased by the 2.27. power spiking penalty.

I WPP40/10

I The results of the rotated bundle analysis are given in Table 7.5.1.

Comparing Table 7.5.1 to Table 4.2.2, there is sufficient margin to the 1%

plastic strain limit.

7.5.2 Mislocated Bundle Error I Misloading a high reactivity assembly into a region of high neutron I importance results in a location of high relative assembly average power.

Since the assembly is assumed to be properly oriented (not rotated), R-factors used for the misloaded bundle are the standard values for the fuel type.

The analysis uses multiple SIMU1 ATE-3 cases to examine the effects of explicitly mislocating every older interior assembly in a quarter core with a fresh or once-burned assembly. Because of symmetry, the results apply to the whole core. Edge bundles are not examined because they are never limiting, due to neutron leakage.

The effect of the successive mislocations is examined for every control rod sequence throughout the cycle. For each sequence, the MCPR for the properly loaded core is compared to the MCPR of the misloaded core at the misloaded location. The MCPR for the properly loaded core is adjusted by a ratio necessary to place the mislocated assembly on the FCISL. The maximum of l these adjusted MCPRs is the mislocated bundle operating limit.

The results of the mislocated bundle. analysis are given in Table 7.5.2. Comparing Table 7.5.2 to Table 4.2.2, there is sufficient margin to the 1% plastic strain limit.

l 7.6 Contin 1 Rod Drop Accident Results l

The control rod sequences are a series of rod withdrawal and banked.

withdrawal instructions specifically designed to minimize the worths of-(

individual control rods. The sequences are examined so that, in the event of the uncoupling and subsequent free fall of the rod, the incremental rod worth is acceptable. Incremental rod worth refers to the fact that rods beyond WPP40/10 l

I

1 Group 2 are banked out of the core and can only fall the increment from full-in to the rod drive withdrawal position. Acceptable worth is one which produces a maximum fuel enthalpy less than 280 calories / gram.

Some out-of-sequence control rods could accrue potentially high worths. Ilowever, the Rod Worth Minimizer (RWM) will prevent withdrawing an out-of-sequence rod, if accidentally selected. The RWM is functionally tested I before each startup.

The sequence in the RWM will take the plant from All Rods In (AR1) to well above 20% core thermal power. Above 20% power even multiple operator errors will not create a potential rod drop situation above 280 calories per gram [20-22). Below 20% power, however, the sequences must be examined for incrementel rod worth. This is done throughout the cycle using the full core, xenon-free SIMULATE-3 model.

" '"' ^ ""d " S"*" ** **'* **""'"*d "' " "" **Posures E

5 throughout the cycle. For startup, the rods are grouped, as shown in Figures 7.6.1 and 7.6.2, and are pulled in numerical order. All the rods in one group are pulled out before the pulling of the next group begins. The rods in the first two groups are individually pulled from full-in to full-out. Beyond Group 2, the rods are banked out using procedures [23-24) which reduce the rod incremental worths.

The potentially high worths that occur in the pulling of the Group 1 rods are ignored because the reactor is subcritical in Group 1. Therefore, I a rod drops from any configuration in the first group, its excess reactivity if contribution to the Rod Drop Accident (RDA) is zero. Successive reloads of axially zoned fuel have extended this'suberiticality situation to the second group as well.

The second group of rods was examined-using.the analysis procedure

~

described in Reference 25. Relatively few control rod configurations were found to be critical. For conservatism, " critical" was defined as the 2 WPP40/10 I

I

s SIMULATE-3 average cold critical Kerg minus 1% AK (reactivity anomaly criteria). The few potentially criticc1 configurations in Group 2 contributed L less excess reactivity to the RoA than subsequent configurations in Group 3.

[ Most pre-drop cases in Group 3 are critical. Therefore, t'e h entire dropped rod worth contributes toward the RDA excess reactivity insertion. The method used to evaluate Group 3 involved pulling Groups 1 and 2 out and banking Group 3 to varying positions. The types of cases examined includedt-

1. Banked positions 04, 08, 12, and 48 (full-out).

I

2. Group 3 rods pulled out of sequence, creating high flux regions.

l 3. Xenon-free conditions, both cold moderator and " standby" (i.e., 1020 psia).

4. Group 3 rods dropping from 00 (full-in) to the appropriate banked position.

I 5. Stuck rods from previously pulled Group 1 or 2 dropping from 00

to 48.

I The highest worth results, presented in Table 7.6.2, fit under the bounding analysis in References 20 through 22.

l 7.7 Refueline Accident Res.ul_t.E If any assembly is damaged during refueling, then a fraction of the fission product inventory could be released to the environment. The source term for the refueling accident is the maximum gap activity within any bundle. The source term includes contributions from both noble gases and iodines. The calculation of maximum gap activity is based on the MAPLHGRs, the maximum operating fuel centerline temperatures, and maximum bundle burnup.

The fuel rod gap activity for the Reload Cycle is bounded by the values used in Section 14.9 of the FSAR, Reference 26.

WPP40/10 l

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TABLE 7.1.1 VY CYCLE 15

SUMMARY

OF SYSTEM TRANSIENT MODEL INITIAL CONDITIONS FOR TRANSIENT ANALYSES

E Core Thermal Power (MWth) 1664.0 4

Turbine Steam Flow (% NBR) 105 Total Core Flow (10 61bm/hr) 48.0 Core Bypass Flow (106 1bm/hr)* 5.8 Core Inlet Enthalpy (BTU /lbm) 521.6 Steam Dome Pressure (psia) 1034.7 Turbine Inlet Pressure (psia) 986.0 1

Total Recirculation Flow (10 61bm/hr) 23.4-Core Plate Differential Pressure (psi) 19.7 Narrow Range Water Level (in.) 162 Average Fuel Gap Conductance (See Section 4.2)

.I

• Includes water rod flow.

I 1B I

I lI 1

i WPP40/10

i E

TABLE 7.2.1 I VI_. CYCLE _15 TRANSIENT ANALYSIS RESULTS Peak Peak Avg. I Prompt Power Heat Flux Operating (Fraction of (Fraction of I

MCPR Transient ExpDsur.e Initial Value) Initial Value) ACIE __ Limits Turbine Trip EOFPL 3.347 1.244 .18 1.25 I Without Bypass.

" Measured" Scram Time EOFPL-1000 2.453 1.155 .11 1.18 EOFPL-2000 1.712 1.060 .04 1.11 Turbine Trip EOFPL 3.847 1.285 .22 1.29 I Without Bypass.

"67B" EOFPL-1000 2.834 1.206 .15 1.22 E0FPL-2000 2.114 1.121 .08 1.15 i Generator Load E0FPL Rejection Without Bypass, EOFPL-1000

" Measured" 3.240 2.428 1.228 1.143

.16

.10 1.23 1.17 Scram Time E0FPL-2000 1.677 1.040 .02 1.09 Generator Load E0FPL 3.704 1.284 .20 1.27 Rejection Without Bypass, E0FPL-1000 2.990 1.208 .15 1.22 "67B" I. Scram Time E0FPL-2000 2.297 1.117 .07- 1.14 I

f Loss of 100 0F Feedwater Heating E0FPL E0FPL-1000 1.147 1.256 1.148 1.163

.11

.13 1.18 1.20 E0FPL-2000 1.262 1.151 .12 1.19 B00 1.152 1.152 .12 1.19 l

l I

l L i 1

WPP40/10 I

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l TABLE 7.3.1 VY CYCLE 15 OVERPRESSURIZATION ANALYSIS RESULTS Maximum Pressure at Reactor Conditions Vt asci Bottom (esia)

" Measured" Scram Time 1262 "67B" Scram Time 1293 l TABLE 7.5,J VY CYCLE 15 ROTATED BUNDLE ANALYSIS RESULTS I Operatiig MCPR Limit Maximum Attainable LHGR (kW/ft) 1.23 19.95 I

TABLE 7.5 2 VY CYCLE 15 MISLOCATED BUNDLE ANALYSIS RESULTS Maximum Attainable Operating MCPR Limit LHCR (kW/ft) 1.20 19.85

-4 6 -

WPP40/10

I TABLE 7.6 2 I VX_,CICLE 15 CONTROL ROD DR0f_ANALYS.lS_EESULTS-Maximum Incremental Rod Worth 0.67% AK Calculated Cold, Xenon-Free Bounding Analysis Worth for Enthalpy 1.30% AK Less than 280 Calories per Gram (20-22]

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I t______ _ _ _ .-- - - I Estimate New l lCPR3 as:

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STOP FIGURE 7.1.1 Flow Chart for the Calculation of ACPR Usine the RETRAN/TCPYA01 Codgg WPP40/10 _

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-$3-WPP40/10

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~ ~

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i frenerator Load Rejection Without Bypass. EOFPL15 '

Iransient Response Versus Time. " Measured" Scram Time l WPP40/10

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l neaerator toad nejection.Without Bypass. EOFPL15-1,000 MWD /ST Iranti.i.tni._AREP_Qnse Vertus liDLe " Meas.ute.d'.' Scram Time I

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l FIGURE 7.2.5-2 Gangrator Load Rejection Withoutlysgas. EOFPL15-1000 MWD /SI i

IIAnEient.Rttponse Verius Time. "Megsgred" Scram Time l

I l WFP40/10 I

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l flGURE 7.2.5-3 GJtacrttnr_.Lngd Rejection Without Bypass. EOFPL15-1000 MWD /ST Ir_ansient Response Vgraus Time. "Mensured" Scram Time WPP40/10

, , , , o o'

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B3M0d N0hLln3N 'WBON I FIGURE 7.2.6-1 Etnerator Load Rejection Without Bypass. E0FPL15-2000 MWD /ST IIansitnLAtif.QaEC ltrius Time. " Measured" SCIam Time WP140/10 .

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.Gf,ngrator Load Rejection Without Bypag.s. EOFPL15-2000 MWD /ST Transient Response Versus Time. " Measured" scram Time WPP40/10 .

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j. O 3rl*)YA lylilNI 30 N0110YW FIGURE 7.2.7-1 Lgas of 100'F Feedwat_tr_Eggtine. EOFPL15-1000 MWD /ST (Limitine Case)

I Itansient Resconse Versus Time I WPP40/10

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( ngl /nie ) ONn000805137NI 3800 FIGURE 7.2.7-2 Loss of 100'F Feedvater Heatine. EOFPL15-1000 MWD /ST (Limitine Ca d Transient Response Versus Time

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~~

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- - - - - - - - - - - 0 8

s 0

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• 4

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n mg53_8

• yt>poyE -

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Y L

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[ . INITIAL QONDITIONS 43 3g 0 0 35 31- 0 16 16 0 27 - 40 32 32 40 23- 24 0 0 24 19 - 40 32 32 40 1 15 0 16 16 0 11 07 0 0 03 -

i i i 02 06 10 14 to 2.2 26 30 34 38 42 Core Thermal Power = 1664 Wt Core Exit Pressure = 1033 pelo

= 1.384 lI Core Flow Cycle Exposure

= 48 Mlb/hr

= 6600 Wd/st initial MCPR initIof MLHCR = 14.40 kw/f t Zero Xenon RWE Control Rod = 26-23 I

TRANSIENT

SUMMARY

RBM Rod MLHGR.

Setoolnt Pontilen ACP.R (AE/f.i.1 104 10 0.14 15.41 105 12 0.18 15.97 106 12 0.18 15 97 l= 107 14 0.21 16.57 108 18 0.27 17.79 I

El.GUEE_ 2 d.s.1 Re a c t o r In i t i aLCQndili9aa._And._'IIAngie n t Summary for the VY evrle 15 Rod Withdrawal Error Case 1 WPP40/10 s

INITIAL CONDITIONS 43 39 30 35

( 31 14 8 14 27 -

23- 30 8 22 8 30 19 -

15 14 8 14 f 11 07 30 03 i i i E

02 06 10 14 18 22 26 30 34 38 42 Core ThertTol Power = 1664 Wt Core Exit Pressure = 1033 pelo Core Floe =

48 Mlb/hr Initlol MCPR = 1,567 Cycle Exposure - 5400 Wd/s t initial MLHGR = 11,34 kw/fi EquilibrItrn Xenon RWE Control Rod = 30-23 TRANSIENT

SUMMARY

RBM Rod MLHGR.

Setoolnt Positten ACEg (kw/ft) 104 16 0.11 11.35 105 20 0.15 11.38 106 22 0.17 11.30 107 24 0.19 11.44 108 28 0.20 11.57 FIGURE 7.4.2 Reactor initial Condit. ions and Transient Summary for the VY Ovele 15 Rod Withdrawal Error Case 2 WPP40/10

W W W M M M M M M M M M M M M M M M M 7 VY ROD WITHDRAWAL ERROR TEST - CASE S30FLO4 R

5 LICEllSIllG DEM0!1 STRATI 0ft CASE. 4X4 MODE =2.2 RBM RESP 0ftSE TO RVE. A+C _5

~

9_

N' VERSUS o RWE C0f1 TROL ROD POSIT 1011

= ri Y  :

- (  :

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2 e o 2  :

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9

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E' 5  :::

= n n ,

2 w

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-f 3

[f ;_ It0TE: _i

~

E 1. All Intercepts are Determined by the B&D Channel.

2. The Box (D) Shows the Response with

g. tro Instrument Failures. , a "o' co

. s.'oo d.co d.co A.co Am A.co d.co d.oo d.ee w}e RWE C0tiTROL ROD POSIT 10ft

~M M M l O I- T _Ut 1 4

V s

5 VY ROD WITHDRAWAL ERROR TEST - CASE S30FLO4 LICEtJSIf1G DEM0!1 STRATI 0ft CASE. 4X4 MODE =2.2

_5 N 5_

RBM RESPO!3SE TO RWE. B+D ~

i VERSUS 2 RWE C0fiTROL ROD POSIT 10fl

}-

_5 5 Q s:

M e a o cm o -3 G,P 2-c:r 5

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=

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3 dtos RBM Setpoint Intercept is

~

/ 1.

@ 105 Harked with (*). _i n i / E # 2. Rod is Stopped it Notch Following

~

~

/

Two Inches of Free Rod Motion.

[ 3. The Box C) Shows the Response with No Instrument Failures. .

o . . . I It .

so.co -

25.00 40.e0 as.oe too.co c.co s.00 1c.00 as.cc 20.00 25.es RWE C0tJTROL ROD POSITIOtt

1 l

1 I

I I 43 3

39 2 1 1 2 36 4 3 4 3 4 I 31 1 2 2 1 27 - 3 4 3 4 3 23 - 1 2 1 1 2 1 19 - 3 4 3 4 3 15 1 2 2 1 11 4 3 4 3 4 07 2 1 1 2 03 3 l 02 00 10 14 18 1 1 I 22 26 30 34 38 42 I

I I

I FIGURE 7.6.1 Ein t Four Rod Arrava_ht11ed in the A Secuences l WPr40/10 I

l

m I

I

'1 l I

i I 43 3 3  ;

39 2 1 2 35 3 4 4 3 31 2 2

• 1 2 27 , 3 4 3 3 4 3 23 - 1 2 1 2 7 19 - 3 4 3 3 ( 3 15 2 2 2 I

1 1 11 3 4 4 3 07 2 1 2

?,3 3 3 I 0 ". 06 10 14 l

18 l l 22 26 30 34 38 42 -

I I

I I

FIGURE 7.6.2 First Four Rod Arrays Pulled in the B Sequences I WPP40/10

~77-I

L 8.0 LOSS-Of-f 00LANI_ACf1DraiLANALYEIS The results of the complete evaluation of the loss-of-coolant accident for Vermont Yankee, as documented in Reference 27, provide the required support for the operation of the Reload Cycle. The LOCA analysis performed in accordance with 10CFR50, Appendix K, demonstrates that the MAPLilGR values comply with the ECCS limits specified in 100FR50.46. The MAPLilGR limits for all the fuel types in the Reload Cycle, as a function of average planar exposure, are provided in Appendix A. Only the limiting MAPLHGR limits for the zoned fuel are provided in Appendix A. MAPLilGR limits exist for each lattice type and are specified in the process computer.

s 1

1 WPP40/10

I L

9.0 CORE COMPONENT QUALIFICATION PROGRAM 9.1 Advanced Nuclear Fuels Fuel Assemblies Vermont Yankee will be replacing four of the GE BP8DWB311-10G2 bundles with four ANFIX-3.04B-EGZ Qualification Fuel Assemblies (QFAs) (28) to qualify this oundle type for use as a potential reload bundle. The ANF-IX QFAs are I manufactured by Advanct.d Nuclear Fucis Corporation (ANF) and designed to match the GE bundle neutronically and thermal-hydraulically, llowever, they differ from the GE bundles in the following ways: 1) the average bundle enrichment is lower, at 3.04 w/o U-235; 2) the fuel pins are smaller in diameter and their numbers are higher, at 72; and 3) a large square inner water channel is I used rather than a large round water rod. The major mechanical design parameters are given in Table 9.1-1 and in Reference 29.

The bundles will be located at 05-20, 39-20, 05-26, and 39-26. These locations are expected to be nonlimiting with respect to MCPR, MAPLilGR, and MLilGR for the entire cycle. The bundles will be monitored during the cycle to assure that they remain nonlimiting. Reference 30 shows that the use of the ANF bundles does not significantly affect the safety analysis described in Section 7.0. VY specific calculations were also performed to show that the analysis in Section 7.0 bounded the ANF-IX QFAs. Therefore, the ANF-IX QFAs can be monitored as a GE bundle with conservative adjustments to the R-factor tables.

9.2 General Electric Marathon Control Rods Vermont Yankee will also be replacing eight standard control rods with eight Marathon control rods [31]. These longer-life control blades utilize a combination of BA C and hafnium as the neutron-absorbing material. Vermont Yankee has approval to use both materials in its control blades (7]. These control rods have been designed to be direct replacement for any of the current GE control rod assemblies. The control rods will be located in nonlimiting locations with respect to shutdown margin. Reference 31 shows that the use of the Marathon control rods does not significantly affect the safety analysis described in Section 7.0.

1 WPP40/10

IAbird L1 m EQtiLNAL ANF-IX FUEL MECllANICAL DESIGN FuRAMETERS L

a Fuel Bundle

• Bundle Type I Vendor Designation Initial Enrichment, w/o U235 ANF 9X9-IX ANTIX-3.04B-EGZ 3.04 Rod Array 9X9 Fuel Rods per Bundle 72 Outer Fuel Channel Material Zr-2 Wall Thickness, inches 0.080 I

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• Complete bundle, rod, and pellet descriptions found in References 28 and 29.

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WPP40/10 .

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L 10.0 SIARTULIRQGRAM

[ Following refueling and prior to vessel reassembly, fuel assembly position and orientation will be verified and videotaped by underwater f television.

The Vermont Yankee Startup Program will include process computer data checks, shutdown margin demonstration, in-sequence critical measurement, rod I scram tests, power distribution comparisons. TIP reproducibility, and TIP symmetry checks. The content of the Startup Test Report will be similar to that sent to the Office of Inspection and Enforcement in the past (32).

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REFERENCES L 1. D. J. Morin, VmImonLlanken_Cyrin_13_Su-ry Rep _ori, YAEC-1680 June 1989.

2. General Electric Standard Application for Reactor Fuel (GESTARII),

NEDE-24011-P-A-9, GE Company Proprietary, February 1988, as amended.

3. A. S. DiGiovine, J. P. Gorski, and M. A. Tremblay, SIMULATE-3 Validation and Verification, YAEC-1659-A, September 1988.

4.

I R. A. Wochlke, et al., MICBURN-3/CASMO-3/ TABLES-3/ SIMULATE-3 Benchmarking 0.f_.YArmonLlankte Cveles 9 J;hrough_13, YAEC-1683-A, March 19'49.

g 5. VYNPC Letter to USNRC, " Vermont Yankee LOCA Analysis MetFJd FROSSTEY g Fuel Performance Code (FROSSTEY-2)." FVY 87-116, dated **cember 16, 1987.

6. Loss,_of CoolanLAccidenLAnaly111 far _Ytrm0Rt _ Yankee Nuclear Power I Station, NED0-21697, May 1990, as amended.

7. Appendix A to Operating License DPR-28 Technical Specifications and Bases for Vermont Yankee Nuclear Power Station, Docket No. 50-271.

8. A. S. DiGiovine, et al., CASMO-3G Validation, YAEC-1363-A, April 1988.

9. A. A. F. Ansari, Methods for the Analysis of Boiling Water Reactors:

Steady-State Core Flow Distribution Code (FISHE}, YAEC-1234, December 1980.

10. A. A. F. Ansari, R. R. Gay, and B. J. Gitnick, FIBWRt A Steady-State -

Core Flow Distribution Code for Polling Wa1.gr Reactors - Code I g Verification and Oualification Report, EPRI NP-1923, Project 1756-1 Final 3 Report, July 1981.

11. USNRC Letter to J. B. Sinclair, SER, " Acceptance for Referencing in I Licensing Actions for the Vermont Yankee Plant of Reports:- YAEC-1232, YAEC-1238, YAEC-1239P, YAEC-1299P, and YAEC-1234." NVY 82-157, September 15, 1982.

12. General Electric Company, GEXL-Plus Correlation Ap. plication to BWR_2-6 )

Reactors._GE6 through GE9 Fuel NEDE-31598P, GE Company Proprietary, July 1989.

13. A. A. F. Ansari and J. T. Cronin, Methods for the Analysis of Boiling Water Reactors: A Systema Transient Analysis Model (RETRAN), YAEC-1233, April 1981.

14 USNRC Letter to R. L. Smith, SER, " Amendment No. 70 to Facility License No. DPR-28," dated November 27, 1981.

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15. V. Chandola, M. P. LeFrancois, and J. D. Robichaud, Appliation of l Onc-DIEtn110nal_.Ein21.1Cs to BoillDg3&ter Reactor Trangient Analysia

, Malhada, YAEC-1693-A, Revision 1, November _1989.

I

16. EPRI, RETRAN - A Program for One-Dimensional Tranglent Thermal-Hydratnic AnalyJis_QLComplex Fluid Flow SyA1Ama, CCM-5, December 1978.

17.

I USNRC Lettec to T. W. Schnatz, SER, " Acceptance for Referencing of Licensing Topical Reports: EPRI CCM-5 and EPRI NP-1850-CCM,"

Septeuber 4, 1984.

I 18. A. A. F. Ansari, K. J. Burns, and D. K. Beller, Methods for the Analysis nf Boiling Water Rtac1DIai.__IranS. lent Uritical Power RA11Q Analysis (RETRAN-TCEXAQ1l, YAEC-1299P, March 1982.

i

19. J. T. Cronin, Mef, hod for Gf ntration of One-DimensionaLKinetics Data for RETRAN-02, YAEC-1694-A, June 1989.

20. C. J. Paone, et al., Rod Drop Accident Analysis for Large Boilinn Water Reactors, NEDO-10527, March 1972.

I 21. R. C. Stirn, et al. , End_DrQp Accident Analysis for Large Boilinn Waler Rgactors Addendum No. 1. Multiple Enrichment Cores With Axial Gadolinium, NEDO-10527, Supplement 1. July 1972.

22. R. C. Stirn, et al. , Rod Drop Accident _fmalysis for Large Boilinglater Reactor Addendum No. 2 Expaged Cores, NED0-10527, Supplement 2, January 1973.

23. C. J. Paone, Hanked _fasillon Withdrawal _Seguence, NEDO-21231, January 1977.

24. D. Radcliffe and R. E. Bates, " Reduced Notch Worth Procedure," SIL-316, November 1979.

25. M. A. Sironen, Vermont Yankee Cycle 14 Core Performance Analysis Report, YAEC-1706, October 1988.

26. Final Safety Analysis Report Vermont Yankee Nuclear Power Station.

November 1987.

27. Loss-of-Coolant Accident Analysis for Vermont Yankee Nuclear Power Station, NEDO-21697, August 1977, as amended; NEDE-21697, Supplement 1, November 1987; and NEDE-21697, Supplement 2, tiay 1990.

28. M. E. Garrett, K. D. Hartley, and M. H. Smith, Vermont Yankee 9X9-IK Qualifi>ation Fuel Assembiv Desien Report. Mg_chanien1. Thermal-Hydraulic.

and.Jie.utronic Desien, ANF-90-034(P), Revision 0, ANF Company Proprietary, March 1990. 1

29. Egneric Mechanical Designs for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel, ANF-89-014(P), ANF Company Proprietary, May 1989.

WPP40/10

30. M. E. Garrett, K. D. Hartley, and M. H. Smith, EcImonLXankee Qual 1Lication_Enel_Aasemhly_Safc.ty_Analysla_AtPDLt ANF-90-048, May 1990.

L 31. GE Marathca_ Control Rod Anacmkly, NEDE-31758P January 1990.

_ 32. R. W. Capstick Letter to W. T. Russell, Regional Administrator " Cycle 14 Start-Up Test Report." BVY 89-51, dated June 14, 1989.

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I' L APPENDIX A CALOUIATEILO2ERATING LIMITS The MCPR operating limits for the Reload Cycle are calculated by adding I

the calculated ACPR to the FCISL. This is done for each of the analyses in Section 7.0 at each of the exposure statepoints. For an exposure interval between statepoints, the highest MCPR limit at either end is assumed to apply ,

to the whole interval.

Table A.1 provides the highest calculated MCPR limits for the Reload Cycle for each of the exposure intervals for the various scram speeds and for the various rod block lines.

Tables A.2 through A.6 provide the maximum calculated MAPLilGR limits for all the GE assembly types in the Reload Cycle.

I A-1 WPP40/10

ml (Revised)

VER50NI_ YANKEE _HILCLFAR POWER STATION CYCLE 15 M_Cf_R OPERATING LIli1IS Value of "N" in RBM Average Control Rod Cycle MCPR Operating Equation (1) Srmram Time Exp n ute_R_ange _ Limit (2 3) 42% Equal or better BOC to EOFPL-2 GWD/T 1.34 than L.C.O. EOFPL-2 CWDir to EOFPL-1 GWD/T 1.34 3.3 C.1.1 EOFPL-1 GWD/T to EOFPL 1. 34 Equal or better BOC to EOFPL-2 GWD/T 1.34 than L.C.O. EOFPL-2 GWD/T to EOFPL-1 CWD/T 1.34 3.3 C.I.2 E_QFPL-1 GWD/T to E0H L 1.34 41% Equal or better BOC to EOFPL-2 GWD/T 1.28 than L.C.O. EOFPL-2 GWD/T to EOFPL-1 GWD/T 1.28 3.3 C.I.1 EOFPL-1 CWD/T to EOFPL 1.28 Equal or better BOC to EOFPL-2 GWD/T 1.28 than L.C.O. EOFPL-2 GWD/T to EOFPL-1 GWD/T 1.28 3.3 C.1.2 EOFPL-1 CWD/T to EOFPL 1.29 1 40% Equal or better BOC to EOFPL-2 GWD/T 1.25 than L.C.O. EOFPL-2 GWD/T to EOFPL-1 GWD/T 1.25 3.3 C.1.1 EOFPL-1 GWD/T to EOFPL 1. 25 Equal'or better BOC to EOC-2 GWD/T 1.25 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.25 3.3. C.1.2 EOC-1 GWD/T to EOC 1.29 NOTES:

(1) The Rod Block Monitor (RBM) trip setpoints are determined by the equation shown in Table 3.2.5 of the Technical Specifications.

(2) The current analysis for the.MCPR operating limits does not include the 7X7, 8X8, 8X8R, or P8X8R fuel types. On this basis, if any of these fuel types are to be reinserted, they will be evaluated in accordance with 10CFR50.59 to ensure that the above limits are-bounding for these fuel types.

-(3) MCPR Operating Limits are increased by 0.01 for single loop operation. Effective 10/90 A-2 WPP40/10

-j TABLE A.2 I r MAELilGR Versus AvArage Planar _Exposurg L

Plantt V.gImQal.XAnkt.e Fuel Type: REEDRh222 Average Planar MAPLEGR (kW/ft) l Exposure Two-Loop Single-Loop t

(mwd /St) Operation __. Ope ra t ion

• 200.0 10.7 8.8 1,000.0 10.8 8.9 5,000.0 11.4 9.4 10,000.0 12.2 10.1 15,000.0 12.3 10.2 20,000.0 12.2 10.1 25,000.0 11.7 9.7 j 35,000.0 10.6 8.8 41,900.0 9.4 7.8 I

• MAPll!GR for single-loop operation is obtained by multiplying MAPLilGR for I two-loop operation by 0.83.

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I, A-3 I WPP40/10 l

IABLE A.3 t%PLEGR Versus Averare Planar Exposure i

Plant: Yermont Yanken Fuel Type: BD324B-

//terage Planar MAPLHGR (kW/ft)

Exposure Two Loop Single-Loop i (tfddIS11__ Operation _ Operation

• 1 200.0 11.22 9.31 3,000.0 11.83 9.81 8,000.0 12.69 10.53 i 10,000.0- 12.80 10.62 15,000.0 12.74 10.57 20,000.0 12.05 10.00 25,000.0 11.39 9.45 35,000.0 10.12 8.39 45,000.0 8.46 7.02 l 50,000.0 5.99 4.97 I

• MAPLHGR for single-loop operation is obtained~by multiplying MAPLHGR for-I two-loop operation by 0.83.

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'l IABLLAd MAPLHGR Vr,raus_ Average Planar _Fxposure

g l5 Plant

EermaaLInnkee Fuel Type: ED326B 1 l

l l Average Planar MAPLHGR (kW/ft)

Exposure Two-Loop Single-Loop i (mwd /St) DPeIR11QD Operation

• 200.0 11.26 9.34 3,000.0 11.72 9.72

8,000.0 12.76 10.59 10,000.0 12.90 10.70 15,000.0 12.82 - 10,64 20,000.0 12.12 10.05 25,000.0 11.44 9.49 35,000.0 10.15 8.42 45,000.0 8.63 7.16 50,000.0 6.17 5.12 I

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• MAPLHGR for single-loop operation is obtained by multiplying MAPLHGR for l two-loop operation by 0.83.

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I IABLE A.5 MAPLHO]LXcIgus Averagc_._f.lanar_1KpnAure i

Plant: Vermont Yankee Fuel Type BP8DWB311-10GZ Average Planar MAPLHGR (kW/ft)

Exposure Two-Loop Single-Loop Operation.

(mwd /St) Ope re tic]L*

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200.0 11.00 9.13 6,000.0 11.92 9.89 7,000.0 12. 10.05 8,000.0 12.34 10.24 10,000.0 12.83 10.64 12,500.0 13.00 10.79 20,000.0 12.24 10.15

,I 25,000.0 11.55 9.58 fn 45,000.0 8.76 7.27 50,740.0 5.91 4.90 l

• MAPLHGR for single-loop operation is obtained by multiplying MAPLHOR for b two-loop operation by 0.8?.

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L A-6 WPP40/10

IABLE A.6 MAPulGR Versus Averare Planar Exocaurn cl I

Plant: Vermont Yanke.e Fuel Typet BP8DWB311-11GZ Average Planar MAPLHCR (kW/ft)

Exposure Two-Loop Single-Loop i (mwd /St) Operation .Qperation*

4 200.0 11.00 9.13 6,000.0 11.92 9.89 I

7,000.0 12.11 10.05 8,000.0 12.34 10.24 10,000.0 12.83 10.64 12,500.0 12.90 10.70

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15,000.0 12.81 10.63 i

I 35,000.0 10.24 8.49 45,000.0 8.76 7.27 50,740.0 5.91 4.90 f

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• MAPulGR for single-loop operation is obtained by multiplying MAPLHGR for two-loop operation by 0.83.

I A-7 WPP40/10 I