ML20129E551

Cycle Core Performance Analysis Rept

ML20129E551
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	10/11/1996
From:	Paul Bergeron, Cacciapouti R, Sironen M VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20129E542	List: BVY-96-129, Forwards COLR for Vermont Yankee Cycle 19 Which Presents cycle-specific Operating Limits for Operation of Cycle IAW Section 6.7.A.4 ML20129E546 ML20129E551
References
YAEC-1935, NUDOCS 9610280092
Download: ML20129E551 (112)

Text

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l Vermont Yankee J

Cycle 19 l Core Performance Analysis Report l l

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October 1996 Major Contributors: C. Chiu N. Fujita B. Hubbard l D. Kapitz l i M. LeFrancois D. Morin K. Morrissey i J. Neyman L. Schor F. Seiface R. Smith l K.St.khn R. Weader l R. Wochlke i

9610280092 961021 PDR ADOCK 050002 1 P

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Prepared by: I '

An'.t4 - /d /!94 M. . Sfr'onen (Date)

, VY Nuclear Engineering Coordinator Approved by: 'e /* /,///

. Caccia Manager (Date) i actor Phys Group Approved by: # NM P. A. Beigeron, M ager (Date)

Transient Analysis roup Approved by:

R. K. Sundaram, Manager (Date)

LOCA Analysis Group Approved by: L /# I J. apman, tor ('Date)

N lear Enginee g Department Approved by:

. /0[////4 M. Marian (Date)

Nu lear Services Manager

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Yankee Atomic Electric Company Nuclear Services Division 580 Main Street Bolton, Massachusetts 01740

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DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company (" Yankee"). The use of 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, with resoect to any unauthorized pm, neither Yankee nor its officers, directors, agents, or employees assume any 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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ABSTRACT This report presents design infortnation, calculational results, and operating limits pertinent to the operation of Cycle 19 of the Vermont Yankee Nuclear Power Station. These include the fuel design and core loading pattern descriptions; calculated reactor power distributions, exposure distributions, shutdown capability, and reactivity data; and the results of the transient, accident and stability analyses performed to justify plant operation throughout the cycle.

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ACKNOWLEDGMENTS The author and major contributors would like to acknowledge the contributions to this work by the Vermont Yankee Reactor Engineering Depanment for their review of input data and guidance.

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1 i TABLE OF CONTENTS l

r EilEe DISCLAIMER OF RESPONSIBILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii ABSTRACT .... ........................................................iv l ACKNOWLEDGMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi i

l LIST OF TAB LES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ix l LIST OP FIG URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xi l

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... 1 l

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I l 2.0 RECENT REACTOR OPERATINO HISTORY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 i

2.1 Operating History of the Current Cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.2 Operating His' tory of Past Applicable Cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 l

! 3.0 RELOAD CORE DESIGN DESCRHYnON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l l

3.1 Core Fuel Loading .............................................5 l 3.2 Design Reference Core Loading Pattern . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 3.3 Assembly Exposure Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l 4.0 FUEL MECHANICAL AND THERMAL DESIGN . . . . . . . . . . . . . . . . . . . . , . . . . . . . 9 4.1 Mechanical Design .............................................9 4.2 'Ihermal Design ...............................................9 4.3 Operating Experience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 l 2

5.0 NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 j 5.1 Core Power Distributions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.1.1 Haling Power Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 I

5.1.2 Rodded Depletion Power Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.2 Core Exposure Distributions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 5.3 Cold Shutdown Margin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l d 5.4 Maximum K., for the Spent Fuel Pool . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 4

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I TABLE OF CONTENTS (Continued)

Eagt 6.0 THERMAL-HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 6.1 Steady-State Thermal Hydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 6.2 Reactor Limits Determination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 )

7.0 ABNORMAL OPERATIONAL TRANSIENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . 28 i

7.1 Transients Analyzed ...........................................28 7.2 Pressurization Transients Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.2.1 Method ology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 l 7.2.2 Initial Conditions and Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . 30  !

7.2.3 One-Dimensional Cross Sections and Kinetics Parameters . . . . . . . . . . . . 32 7.2.4 Turbine Trip Without Bypass Transient ('1TWOBP) . . . . . . . . . . . . . . . . 33  !

I 7.2.5 Generator Load Rejection Without Bypass Transient (GLRWOBP) . . . . . 33 7.2.6 Pressurization Transient Analysis Results . . . . . . . . . . . . . . . . . . . . . . . 34 l 7.3 Loss of Feedwater Heating Transient Analysis . . . . . . . . . . . . . . . . . . . . . . . . . 34 7.3.1 Loss of a Feedwater Heater (LOFWH) Results . . . . . . . . . . . . . . . . . . . 34 7.3.2 Loss of Stator Cooling (LOSC) Results ........................35 7.4 Overpressurization Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 7.4.1 Compliance with ASME Vessel Code . . . . . . . . . . . . . . . . . . . . . . . . . . 36 7.4.2 Safety Valve Challenges . . . . . . . . . . . . . . . . . . . . . . . . . . . *. . . . . . . . 36 7.5 Local Rod Withdrawal Enror Transient Results . . . . . . . . . . . . . . . . . . . . . . . . . 37 i 7.6 Misloaded Bundle Error Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 7.6.1 Rotated B u ndle Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 7.6.2 Mislocated Bundle Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 7.7 Transient Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 l

8.0 DESIGN BASIS ACCIDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 77

. 8.1 Control Rod Drop Accident Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 77 8.2 Loss-of-Coolant Accident Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 79 8.3 Refueling Accident Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 80 9.0 STABILITY ANALYSIS .............................................86

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TABLE OF CONTENTS (Continued)

EAEG 10.0 STARTUP PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 11.0 CONCLUS ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 APPENDIX A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94

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LIST OF TABLFS Number Title E!gt a 2.1.1 VY CYCG 18 OPERATING HIGHLIGHTS . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2.1 VY CYCG 17 OPERATING HIGHLIGHTS . . . . . . . . . . . . . . . . . . . . . . .... 4 3.1.1 ASSUMED VY CYCG 19 FUEL BUNDG TYPES AND NUMBERS . . . . . . . . 7 3.3.1 DESIGN BASIS VY CYCLE 18 AND CYCLE 19 EXPOSURES . . . . . . . . . . . . 7 4.1.1 NOMINAL FUEL MECHANICAL DESIGN PARAMETERS . . . . . . . . . . . . . . . I1 4.2.1 VY CYCLE 19 CORE AVERAGE GAP CONDUCTANCE VALUES . . . . . . . . 12 4.2.2 VY CYCLE 19 HOT CHANNEL GAP CONDUCTANCE VALUES FOR HALING AXIAL POWER DISTRIBUTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.2.3 VY CYCLE 19 HOT CHANNEL GAP CONDUCTANCE VALUES FOR 1.4 CHOPPED COSINE AXIAL POWER DISTRIBUTION .................14 5.3.1 VY CYCG 19 Km VALUES AND SHUTDOWN MARGIN CALCULATION . 18 5.4.1 VY CYCG 19 MAXIMUM COLD K OF ANY ENRICHED SEGhENT . . . . . 18 7.2.1 VY CYCG 19

SUMMARY

OF SYSTEM TRANSIENT MODEL INITIAL CONDITIONS FOR TRANSIENT ANALYSES . . . . . . . . . . . . . . . . . . . . . . . . 43 7.2.2 VY CYCLE 19 PRESSURIZATION TRANSIENT ANALYSIS RESULTS . . . . . 44 7.3.1 VY CYCG 19 LOSS OF TEEDWATER HEATER TRANSIENT RESULTS . . . 45 7.3.2 VY CYCG 19 LOSS OF STATOR COOLING TRANSIENT RESULTS . . . . . . 45 7.4.1 VY CYCLE 19 AShE COMPLIANCE RESULTS . . . . . . . . . . . . . . . . . . . . . . 46 7.4.2 VY CYCLE 19 SAFETY VALVE SETPOINT CHALLENGE RESULTS . . . . . . 46 7.5.1 VY CYCG 19 ROD WITHDRAWAL ERROR ANALYSIS RESULTS . . . . . . . 47 7.6.1 VY CYCLE 19 ROTATED BUNDLE ANALYSIS RESULTS . . . . . . . . . . . . . . 47 7.6.2 VY CYCLE 19 MISLOCATED BUNDLE ANALYSIS RESULTS . . . . . . . . . . . 47 7.7.1 VY CYCLE 19 LIMITING TRANSIENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 8.1.1 CONTROL ROD DROP ANALYSIS - ROD ARRAY PULL ORDER . . . . . . . . 81

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LIST OF TABLFS (Continued)

Number Iglg Eaga 8.1.2 VY CYCLE 19 CONTROL ROD DROP ANALYSIS RESULTS . . . . . . . . . . . . 81 8.2.1 LOCA ANALYSIS ASSUMITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 82 8.2.2 LOCA ANALYSIS RESULTS, PEAK CLADDING 'IEMPERATURE . . . . . . . . 83 A.1 VERMONT YANKEE NUCLEAR POWER STATION CYCLE 19 MCPR OPERATING LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 A.2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-10GZ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 96 A.3 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-1 1 GZ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97 A.4 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB354-12GZ . . . . . . . . . . . . . . . . . ..........................98 A.5 VERMONT YANKEE NUCLEAR POWE.1 STATION CYCLE 19 STABILITY EXCLUSION AND BUFFER REGIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 99 9

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LIST OF FIGURES, Number lillt EaEt 3.2.1 VY CYCLE 19 DESIGN REFERENCE LOADING PATTERN, LOWER RIGHT QUAL eRANT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 5.1.1 VY CYCLE 19 HALING DEPLETION, EOFPL BUNDLE AVERAGE RELATIVE POWERS . . . . . . . ...........................................19 5.1.2 VY CYCLE 19 HALING DEPLETION, EOFPL CORE AVERAGE AXIAL POWER DISTRIBUTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.1.3 VY CYCLE 19 RODDED DEPLETION - ARO AT EOFPL, BUNDLE AVERAGE RELATIVE POWERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.1 4 VY CYCLE 19 RODDED DEPLETION - ARO AT EOFPL, CORE AVERAGE AXIAL POWER DISTRIBUTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 5.2.1 VY CYCLE 19 HALING DEPLETION, EOFPL BUNDLE AVERAGE EXPOS URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.2.2 VY CYCLE 19 RODDED DEPLETION, EOFPL BUNDLE AVERAGE EXPOS URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 5.3.1 VY CYCLE 19 COLD SHUTDOWN MARGIN, IN %AK, VERSUS CYCLE EXPOS URE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 7.2.1 FLOW CHART FOR THE CALCULATION OF ACPR USING THE RETRAN/TCPYA01 CODES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 7.2.2 TURBINE TRIP WITHOUT BYPASS, EOFPL19 TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIhE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 7.2.3 TURBINE TRIP WITHOUT BYPASS, EOFPL19-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . . . . . . . . . . . 53 7.2.4 TURBINE TRIP WITHOUT BYPASS, EOFPL19-2000 MWD /ST TRANSIEN7' RESPONSE VERSUS TBE, " MEASURED" SCRAM TIME . . . . . . . . . . . . . . 56 7.2.5 GENERATOR LOAD REJECTION WITHOUT BYPASS EOFPL19 TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIhE . . . . . . . . . . . . . . 59 7.2.6 GENERA'IOR LOAD REJECTION WITHOUT BYPASS, EOFPL19-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIhE, " MEASURED" SCRAM TIhE . . . . 62

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LIST OF FIGURES (Continued)

Number Title Ege 7.2.7 GENERATOR LOAD RFRCTION WITHOUT BYPASS, EOFPL19-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . 65 7.3.1 LOSS OF 100*F FEEDWATER HEATER (LIMITING CASE) TRANSIENT RESPONSE VERSUS TL 4E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 7.3.2 LOSS OF STATOR CODLING TRANSIENT RESPONSE VERSUS TIME . . . . 70 7.4.1 M2V CLOSURE, FLU.4 SCRAM, EOFPL19 TRANSIENT RESPONSE VERSUS TIME, 67B' SCRAM '!IME . ...................................72 7.5.1 VY CYCLE 19 NORMAL RWE CASES RESULTS . CPR VERSUS RBM S ETPOINT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75 7.5.2 VY CYCLE 19 ABNORMAL RWE CASES RESULTS- ACPR VERSUS RBM S ETPOINT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 8.1.1 FIRST FOUR ROD ARRAYS PULLED IN THE A SEQUENCES . . . . . . . . . . . 84 8.1.2 FIRST FOUR ROD ARRAYS PUI .IEn IN THE B SEQUENCES . . . . . . . . . . . 85 A.1 VERMONT YANKEE NUCLEAR I'O'AT.F STATION CYCLE 19 STABILITY EXCLUSION AND BUFFER REGIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100

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1.0 INTRODUCTION

This report provides information to suppon the operation of the Vermont Yankee Nuclear Power Station through the fonhcoming Cycle 19. In this repon, Cycle 19 will be referred to as the Reload Cycle. The preceding Cycle 18 will be referred to as the Current Cycle. The Cycle 18/19 refueling will involve the discharge of 120 irradiated fuel bundles and the insenion of 120 new fuel bundles. The resultant core will consist of 120 new fuel bundles and 248 irradiated fuel bundles. The General Electric Company (GE) manufactused all the bundles. Some of the irradiated fuel was also present in the reactor in Cycle 17. This cycle will be irferred to as the Past Cycle.

This report contains descriptions and analyses results penaining to the mechanicsl, thermal-hydraulic, physics, transient analyses, accident analyses and stability analysis of the Reload Cycle. De MAPLHGR and MCPR operating limits and the stability exclusion and buffer regions calculated for the Reload Cycle are given in Appendix A. These limits will be included in the Core Operating Limits Repon.

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2.0 RECENT REACTOR OPERATING HISTORY 2.1 Operatine History of the Current Cycle The current operating cycle is Cycle 18. He Current Cycle operated at, or near, full power with the exception of sequence exchanges, several power reductions, one short repair outage and a coastdown to the end of cycle. De operating history highlights and control rod sequence exchange schedule of the Current Cycle are found in Table 2.1.1.

2.2 Ooeratine History of Past Anolicable Cycle The irradiated fuel in the Reload Cycle includes sorne fuel bundles initially inserted in Cycle

17. His Past Cycle operated at, or near, full power with the exception of sequence exchanges, several short power reductions, four shon repair outages and a coastdown to the end of cycle. De operating history highlights of the Past Cycle are found in Table 2.2.1. The Past Cycle is described in detail in the Cycle 17 Summary Report [1].

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l TABLE 2.1.1 VY CYCLE 18 OPERATING HIGHLIGHTS Beginning of Cycle Date May 2,1995 End of Cycle Date September 7,1996*

Weight of Uranium As-Loaded (Short Tons) 72.12 l

Beginning of Cycle Core Average Exposure ** (mwd /St) 12316*

l End of Full Power Core Average Exposure"(mwd /St) 21930' End of Cycle Core Average Exposure ** (mwd /St) 2265(T Number of Fresh Assemblies 120 Number ofIrradiated Assemblies 248 Control Rod Sequence Exchange Schedule:

Sequence Dam Emm In July 11,1995 A2-1 B2-1 September 11,1995 B2-1 Al-1 November 7,1995 Al-1 B1-1 January 9,1996 ,

B1-1 A2-2 February 27,1996 A2-2 B2-2 April 23,1996 B2-2 Al-2 June 12,1996 Al-2 B1-2

• Projected dates and exposures assumed for licensing. The actual End of Cycle Core Average Exposure was 22,964 mwd /St.

" Exposures based on the Plant Process Computer accounting.

TABLE 2.2.1 VY CYCI.E 17 OPERATING HIGHLIGHTS Beginning of Cycle Date October 24,1993 End of Cycle Date March 17,1995 Weight of Uranium As-Loaded (Short Tons) 72.02 Beginning of Cycle Core Average Exposure * (mwd /St) 11547 End of Full Power Core Average Exposure * (mwd /St) 21603 End of Cycle Core Average Exposure * (mwd /St) 22156 Number of Fresh Assemblies 128 Number ofIrradiated Assemblies 240 Control Rod Sequence Exchange Schedule:

Sequence Date Fmi In January 9,1994 A2-1 B2-1 March 15,1994 B2-1 Al-1 May 17,1994 Al-1 BI-1 July 19,1994 B1-1 A2-2 October 6,1994 A2-2 B2-2 December 2,1994 B2-2 Al-2 January 24,1995 Al-2 B1-2 Exposures based on the Plant Process Computer accounting.

3.0 RELOAD CORE DESIGN DESCRIPTION 3.1 Core Fuel Loading 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 the GE Standard Application for Reactor Fuel [2] and the GE Fuel Bundle Design Reports [3],[4].

3.2 Desien Reference Core Loadine Pattern De Reload Cycle assembly locations are indicated on the map in Figure 3.2.1. The other quadrants are mirror images with bundles of the same type having nearly identical exposures. De bundles are identified by the reload number in which they were first introduced into the core. Table 3.1.1 provides the key, called bundle ID, which identifies what explicit fuel type is found in each bundle location.

If any changes are made to the loading pattem at the time of refueling, they will be evaluated under 10CFR50.59. De final loading pattern with specific fuel bundle serial numbers will be supplied in the Startup Test Report.

3.3 Assembly Exposure Distribution The assumed nominal exposure on the fuel bundles in the Reload Cycle design reference l loading pattern is given in Figure 3.2.1. To obtain this exposure distribution, the Past Cycle was  ;

depleted with the SIMULATE-3 model[5],[6] using actual plant operating history. For the Current Cycle, plant operating history was used through September 18,1995. Beyond this date, the exposure was accumulated using a best-estimate rodded depletion analysis to End of Cycle (EOC).

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Table 3.3.1 gives the assumed nominal exposure on the Current Cycle and the Beginning of Cycle (BOC) core average exposure that results from the shuffle into the Reload Cycle loading pattern.

'Ihe Reload Cycle End of Full Power Life (EOFPL) core average exposure and cycle capability are provided.

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TABLE 3.1.1 ASSUMED VY CYCLE 19 FUEL BUNDLE TYPES AND NUMBERS Fuel Type Reload Cycle Number Desienation Bundle ID Loaded of Bundles Irradiated BP8DWB335-10GZ R16A 17  %

BP8DWB335-1IGZ R16B 17 32 BP8DWB335-10GZ R17A 18 88 BP8DWB335-1IGZ R17B 18 32

, N_m BP8DWB354-12GZ R18 19 120 TABLE 3.3.1 DESIGN B ASIS VY CYCLE 18 AND CYCLE 19 EXPOSURES

• Assumed End of Current Cycle Core Average Exposure with an 22.46+0.6 GWd/St l

Exposure Window of + 600 mwd /St[7]

Assumed Beginning of Reload Cycle Core Average Exposure 12.23 GWd/St Haling Calculated End of Full Power Life Reload Cycle Core Average 22.93 GWd/St Exposure Reload Cycle Full Power Exposure Capability (Haling) 10.70 GWd/St l

l

• Exposures based on the SIMULATE-3 accounting.

R188 R18 R188 R17B R188 R18 R16A R18 R17A R17A R16A 22 21.053 0.000 21.332 11.010 20.790 0.000 23.586 0.000 13.341 12893 23.853 R18 R16A R18 R17A R18 R17A R18 R188 R16 R17A R16A 20 0.000 21217 0.000 11.419 0.000 12.683 0.000 22.000 0.000 13.554 24.502 R188 RIO R178 R18 R16A R16 R17A R18 R17B R17A R16A 18 21.148 0.000 12.063 0A00 21.767 0.000 13.741 0400 12J83 13.586 25.470 R178 R17A R18 R16A R18 R17B RIO R18 R17A R16A 16 11415 11.426 0.000 22452 0.000 12.115 0.000 0.000 13.000 23.818 RI6S R18 R16A R18 R178 R18 RI6A R17A R168 14 20.947 0.000 21.899 0.000 12.124 0.000 22.513 13.013 21.fA5 R16 R17A R18 R17B RIO R17A R18 R17A R16A 12 0.000 13.073 0.000 12.129 0.000 12.325 0400 13.001 23.596 R16A R18 R17A R18 R16A R18 R17A R16A R16A to 23310 0.000 13.705 0.000 22.516 0.000 12.497 23.350 25.225 R18 R168 R18 R18 R17A R17A R16A R16A nn 0.000 22.862 0.000 0.000 13.011 12.806 23.320 24.983 R17A R18 R17B R17A R16A R16A RISA na 13.587 0.000 12256 13.000 21.803 24.566 25481 R17A R17A R17A R16A nd 12.004 13.557 - 13.774 23.791 R16A R16A R16A BUNDLEID ns 23.991 24 60e 254n BOC EXPOSURE (GWO/ST) 23 25 27 29 31 33 35 37 39 41 43 FIGURE 3.2.1 VY CYCLE 19 DESIGN REFERENCE LOADING PATTFRN. LOWER RIGHT OUADRANT

4.0 FUEL MECHANICAL AND THERMAL DESIGN 4.1 Mechanical Desien All of the fuel to be inserted into the Reload Cycle was fabricat' e d by GE. De major mechanical design parameters are given in Table 4.1.1 and References 2 through 4. 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.

Mechanical and chemical compatibility of the fuel bundles with the in-service reactor environment is also addressed in Reference 2.

4.2 Dermal Desien ne fuel thermal effects calculations were performed using the FROSSTEY.2 computer code [8]. The FROSSTEY-2 code calculates pellet-txladding gap conductance and fuel temperatures from a combination of theoretical and empirical models including but not limited to fuel and cladding thermal expansion, fission gas release, pellet swelling, pellet densification, pellet cracking, and fuel and cladding thermal conductivity.

The thermal effects analysis included the calculation of fuel temperatures and pellet-to-cladding gap conductance under core average and. hot channel conditions. The core average calculations integrate the responses of individual fuel batch average operating histories over the core average exposure range of the Reload Cycle. These gap conductance values are weighted axially into 12 axial nodes by power distributions and radially by volume. The core-wide gap conductance values for the RETRAN system simulations, described in Sections 7.! and 7.2, are from this data set at the corresponding exposure statepoints. Table 4.2.1 provides the core average response of gap conductance.

The hot channel gap conductance values, which are input to the hot channel transient calculations (Section 7.1), were evaluated for the limiting fuel bundle type as a function of the assembly exposure for two axial power shapes, a 1.4 chopped cosine and the Reload Cycle's Haling.

The hot channel calculations assumed the following as required by the NRC Safety Evaluation for 9

FROSSTEY-2[9]: 1) appropriate allowances to account for manufacturing uncertainties and 2) the worst axial power shape prior to the transient. The peak power node was placed at the maximum average planar linear heat generation rate (MAPLHGR) limits. Gap conductance values for the hot channel analysis were determined using the limiting bundle exposure. The limiting bundle is dermed as the bundle with the lowest MCPR or the highest power, if different,'within the exposure range of interest. The limiting exposure for the bundle is dermed by the exposure which produces the highest bundle average gap conductance within the interval of interest. The SIMULATE-3 rodded depletion (Section 5.1.2) provided predictions of the limiting bundle exposure for each exposure interval. Table 4.2.2 provides the hot channel gap conductance values for the two axial power shapes. Results are presented for the bounding exposure for the chopped cosine shape and at the four exposure statepoints for the Haling shape.

4.3 Operatine Exoerience All irradiated fuel 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 no fuel rod failure.

TABLE 4.1.1

{

NOMINAL FUEL MECHANICAL DESIGN PARAMETERS Fuel Bundle

• Irradiated Fuel Tvoe New Fuel Tvoes i

Bundle Types GE8X8NB GE8X8NB Vendor Designation GE9B-P8DWB335-10GZ & GE9B-P8DWB354-12GZ GE9B-P8DWB335-11GZ Initial Enrichment, w/o Um 3.35 3.54 Rod Array 8X8 8X8 1

Fuel Rods per Bundle 60 60 I

Outer Fuel Channel Material Zr-2 Zr-2 Wall Thickness, inches 0.080 0.080

• Complete bundle, rod, and pellet descriptions are found in References 2 through 4.

TABLE 4.2.1 VY CYCLE 19 CORE AVERAGE GAP CONDUCTANCE VALUES Gap Conductance (BTU /hr-ft2 ,.p)

Axial E EOFPL-2000 EOFPL-1000 EOFPL N.g[q mwd /St mwd /St i 1125 1890 2035 2095 2 2405 3630 3710 3885 3 2555 3840 3955 4435 4 2580 3860 4015 4610 5 2610 3900 4075 4665 6 2645 3930 4160 4670 7 2605 3895 4075 4620 8 2610 3890 4080 4675 9 2535 3825 3930 4345 10 2450 3710 3805 4035 11 1940 3025 3080 3235 12 685 1085 1145 1245

TABLE 4.2.2 VY CYCLE 19 HOT CHANNEL GAP CONDUCTANCE VALUES

• FOR HALING AXIAL POWER DISTRIBUTION Gap Conductance (BTU /hr-ft' *F)

Axial BOC~* EOFPI 2000 EOFPL-1000 EOFPL*!

Node mwd /St" M W d/St" 7.63 GWd/St"* 11.02 GWd/St"* 11.02 GWd/St"* 12.17 GWd/St*"

1 3970 5280 5280 5080 2 8110 7780 7780 8110 3 9250 10200 10200 9830 4 9280 10200 10200 9830 5 9340 10200 10200 9830 6 9460 10200 10200 9820 7 9350 10200 10200 9820 8 9330 10200 10200 9820 9 9210 10200 10200 9820 10 8470 8160 8160 9030 11 7250 7140 7140 7120 12 2050 2630 2630 3080 The hot channel gap conductance values are derived for the BP8DWB354 fuel type because it is conservative compared to the other fuel types.

" Core Average Exposure.

"* Peak Bundle Exposure.

I TABLE 4.2.3 VY CYCLE 19 HOT CHANNEL GAP CONDUCTANCE VALUES

• FOR 1.4 CHOPPED COSINE AXIAL POWER DISTRIBUTION Gap Conductance (BTU /hrdt2 *F)

Axial HQC" EOFPL-2000 EOFPL-1000 EOFPL" N_ gds mwd /St" mwd /St" 7.15 GWd/St"* 11.03 GWd/St*" 11.43 GWd/St"* 12.80 GWd/St*"

1 1050 1000 1000 990 2 1790 2420 2510 2790 3 4390 6560 6700 6420 4 7880 8070 8000 7740 5 8920 9810 10340 9840 6 9850 10500 10350 9850 7 9490 10500 10350 9850 8 9760 10500 10350 9850 9 8150 10500 8460 9440 10 6640 7320 7420 7380 11 2820 4770 5060 5090 12 1230 1280 1290 1330

• The hot channel gap conductance values are osrived for the BP8DWB354 fuel type because it is conservative compared to the other fuel types.

• • Core Average Exposure.

• • • Peak Bundle Exposure.

I 1

I 5.0 NUCI F AR DESIGN 5.1 Core Power Distributions De Reload Cycle was depleted using SIMULATE-3 to give both a rodded depletion and an All Rods Out(ARO) Haling depletion.

5.1.1 Halino Power Distribution l

l The Haling depletion serves as the basis for defining core reactivity characteristics for most transient evaluations. His is primarily because its flat power shape has conservatively weak scram characteristics. Sensitivity studies have shown that the limiting pressurization transient results are more conservative when calculated using the Haling power distribution as the initial power shape.

The Haling power distribution is calculated in the ARO condition. The Haling iteration converges on a self-consistent power and exposure distribution for the burnup step to EOFPL. In principle, this should provide the overall minimum peaking power shape for the cycle. During the actual cycle, flatter power distributions might occasionally be achieved by shaping with control rods. However, such shaping would leave under bumed regions in the core which would peak at another point in time. Figures 5.1.1 and 5.1.2 give the Haling radial and axial average power distributions for the Reload Cycle.

5.1.2 RndAed Deoletion Power Distribution The rodded depletion was used to evaluate the mistoaded bundle error and the rod withdrawal error because it provides the initial rod patterns and more accurately defines the local characteristics prior to the transient evaluations. It was also used in the rod drop worth and shutdown margin calculations because it depletes the top of the core more realistically than the Haling depletion. 'Ihe rodded depletion also provides the hot channel bundle exposures for the gap conductance calculation.

To generate the rodded depletion, control rod pattems were developed which give critical eigenvalues at several points in the cycle and peaking similar to the Haling calculation. The resulting patterns were frequently more peaked than the Haling, but were below expected operating limits.

However, as stated above, the under burned regions of the core can exhibit peaking in excess of the Haling peaking when pulling ARO at EOFPL. Figures 5.1.3 and 5.1.4 give the ARO radial and axial average power distributions for the Reload Cycle rodded depletion at EOFPL.

5.2 Core Funosure 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 Haling calculation produced the EOFPL radial exposure distribution given in Figure 5.2.1. Since the Haling power shape is constant, it can be held fixed by SIMULATE-3 to give the exposure distributions at various mid-cycle points; that is:

EOFPL-2000 mwd /St and EOFPL-1000 mwd /St. These exposure distributions together with the BOC and EOFPL distributions were used to develop reactivity input for the core wide transient analyses.

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

5.3 Cold Shutdown Margin The shutdown margin (SDM) for the Reload Cycle must exceed the Technical Specifications SDM limit [10]. 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 ( with the strongest rod out was calculated at BOC and at various exposure points during the cycle. Subtracting each cold ( with the strongest rod out from the cold critical (

defines the SDM as a function of exposure. Figure 5.3.1 shows the results.

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

Because the local reactivity may increase with exposure, the SDM may decrease. To account for this and other uncertainties, the value R is calculated. R is defined as Ri plus R 2 . Ri is the difference

between the cold K, with the strongest rod out at BOC and the maximum cold K, with the strongest rod out in the cycle. R is 2 a measurement uncertainty in the demonstration of SDM associated with the manufacture of past control blades. It is presently set at 0.07% AK[11],[12). The shutdown margin results, summarized in Table 5.3.1, are greater than the Technical Specification limit including R.  !

5.4 Maximum K for the Soent Fuel Pool Section 5.5E of the Technical Specifications requires that the K. 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, 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 5.4.1. These values are obtained from CASMO-3G[13].

.. .. .-. _ = . . . . - - - . . - - _ - . . . . -

I TABLE 53.1 i

VY CYCLE 19 Km,r VALUES AND SHUTDOWN MARGIN CALCULATION l '

Cold Critical Km 1.0000 BOC K,- Controlled With Strongest Worth Rod Withdrawn 0.9883 I

Cycle Minimum Shutdown Margin Occurs at 8400 mwd /St With Strongest Worth Rod Withdrawn 1.15% AK l

l Ri , Maximum Increase in Cold ( With Exposure 0.018% AK l I

l l TABLE 5.4.1 VY CYCLE 19 MAXIMUM COLD K_ OF ANY ENRICHED SEGMENT Bundle Tvoe Maximum K_

BP8DWB335-10GZ l.22 BP8DWB335-11GZ 1.22 BP8DWB354-12GZ 1.22 l

l l

R168 R18 R188 R178 R188 R18 R16A R18 R17A R17A R16A 22 1.072 1.373 1.092 1208 1.089 1.367 1.031 1.266 0.972 0.770 0.429 R18 R16A R18 R17A R18 R17A R18 R188 R18 R17A R16A s 1

to 1.379 1.110 1A10 1240 1.401 1204 1.340 0.974 1.102 0.738 0.404 R188 R18 R178 RIO R16A R18 R17A R18 R178 R17A RISA 18 1.005 1.410 1.251 1A26 1.117 1.392 1.150 1.224 0.920 0.868 0.358 R178 R17A R18 R16A R18 R178 R18 R18 R17A R16A 16 1J07 1.240 1.426 1.116 1.407 1.199 1293 1.152 0.807 0.471 R168 R18 R16A R18 R178 R18 R16A R17A R168 14 1.087 1.400 1.117 1.407 1210 1.326 0.956 0.899 0.595 R18 R17A R18 R178 R18 R17A R18 R17A R16A 12 1.367 1202 1.392 1.199 1.326 1.000 1.114 0.767 0.468 R16A R18 R17A R18 R16A R18 R17A R16A R16A 10 1.033 1.340 1.151 1293 0.956 1.114 0444 0.547 0.341 R18 R188 R16 R18 R17A R17A R16A R16A na 1255 0.973 1223 1.151 0.498 0 766 0447 0.382 R17A R18 R178 R17A R16A R104 R16A nn 0.968 1.101 0.919 0.806 0.593 OA it 0.340 R17A R17A R17A R16A n4 0.768 0.737 0.886 0.470 R16A R16A RISA BUNDLE 10 ns 0.426 0A02 0.357 EOFPL RELATIVE POWER 23 25 27 29 31 33 35 37 39 41 43 FIGURE 5.1.1 VY CYCLE 19 HALING DEPLETIO1 EOFPL BUNDLE AVERAGE RELATIVE POWERS 4 alp s 4%r., ,~,,y  %,22 pa_ _a h _a . m,m . 4,_m, _a4 y , , , ,_ __

a a ..-a .am. _a-.sh,,.+ , _ , ,

1 l

25 4 24  !

~

n 21 )

20

)

10 18 i) 17 d l

l f i6 l d 15 i )

1 5 14 0 k) j l g13 m 12 i' i) 11 1 1

l10 g < >

1 q )

l 8 7 1:

6 L l 5 i 4

l 3

/

2

- ' W 1  ; "

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 RELATIVE POWER FIGURE 5.1.2 i

VY CYCLE 19 HALlNG DEPLETION. EOFPL CORE AVERAGE AXIAL l

l B)WER DISTRIBUTION 20-l l

- - . - . ~.- - . . . . - . ~ .- - .- - .,- . ..= - - . ~ _. . . _ . - . . - . - . - , . . . > . . - . _ . .

! I i 1 I

I I

1 I

R188 R10 R188 R178 R188 R18 R16A R18 R17A R17A R16A

)

22 4

1.128 1.400 1.166 1274 1.142 1.411 1.041 1.234 0.931 0.723 0.306 I l

R18 R1SA R18 R17A R18 R17A Ri8 R188 R18 R17A R16A 20 l 1.474 1.177 1.497 1J03 1.486 1.225 1.360 0.966 1.002 0.006 0.373 l

l R188 R18 R178 R18 R16A R18 R17A R18 R178 R17A R16A l 18  !

1.161 1.400 1.314 1.600 1.167 1.414 1.143 1.197 0.882 0.027 0.328  ;

I l R178 R17A R18 RIGA R18 R178 Ris R18 R17A R16A l 16 1.200 1.303 1.498 1.164 1.430 1.191 1273 1.116 0.768 0.439 R188 R16 R16A R18 R178 R18 R16A R17A R188 14 1.145 1.486 1.169 1.441 1.216 1.310 0.930 0.867 0.660

R18 R17A RIS R178 R18 R17A R18 R17A R16A i 12 1.416 1.226 1.419 1.198 1.316 1.066 1.071 0.726 0.426 i R16A R18 R17A R18 R16A R18 R17A R16A R16A

' 10 1.046 1.363 1.147 1280 0.936 1.073 0.799 0.611 0312 R18 . M188 R18 R18 R17A R17A R16A R16A na ,

1234 0.964 1.100 1.119 0.800 0.726 0A12 0.334 R17A R18 R178 R17A R16A R16A R16A 0.828 1.002 0.882 0.769 0.650 0.420 0.312 l f R17A R17A R17A R16A

, n4 l 0.723 0.894 0.626 0.430 i R16A R16A R16A BUNDLEID I

no 0 '95 0.372 0.328 . EOFPL RELATIVE POWER 1

23 26 27 29 31 33 36 37 30 41 43 FIGURE 5.1.3 I r VY CYCLE 19 RODDED DEPLETION - ARO AT EOFPL.

BUNDLE AVERAGE RELATIVE POWERS i

,. n.-._a1 4 s u ..s u. a ta ak.n.ssa s~.,_r aw ,e.s -s-sus... .L_ >-,as--nux, a- m. 6,-- _ n _ . . --. .,.n.., .,,,,,

)

l 1

l l

25 q l 24 l x 4

=

21 )

20 l 19 )

18 < >

17 <

e l16

( 15 5 i I) o g

13 <L m 12 11 l10 <-

b . i 6

5 4

3 2 .

)

1  :

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 RELATIVE POWER FIGURE 5.1.4 VY CYCLE 19 RODDED DEPLETION - ARO AT EOFPL.

CORE AVERAGE AXIAL POWER DISTRBUTION

I Ries R18 R188 R178 R188 R18 R16A R18 R17A R17A R16A 22 32.493 14.634 32.984 23.488 32.406 14.610 34.556 13.321 23.701 20.894 28.422 R18 R16A R18 R17A R18 R17A R18 R188 R18 R17A R16A 20 14.636 33.067 14.000 24.642 14.883 25.723 14.223 32.980 11.894 21.418 98.810 R168 R18 R178 R18 R16A R18 R17A R18 R178 R17A R16A 18 32.827 14.983 26.420 16.130 33.875 14.700 26.006 12.984 22.070 23.890 29.287 R178 R17A R18 R16A R18 R178 R18 R14 R17A R16A 16 23.891 24.649 16.131 34.749 14.929 24.895 13.717 12216 21.886 28.842 R188 R18 R16A R18 R178 R18 RISA R17A R168 14 32.537 14.861 33.816 14.931 26.006 14.088 32.700 22.586 27.991 R18 R17A R18 R178 R18 R17A R14 R17A R16A 12 14.506 25.886 14.767 24.910 14.088 23.940 11.820 21.175 28.478 R16A R18 R17A R18 RI6A R18 R17A R16A R16A 10 34.328 14222 26.972 13.718 32.702 11A19 21.493 29.186 28.866 RIS R188 R18 R18 R17A R17A R16A R16A nn 13J11 33.037 12.977 12.210 22.588 21.182 28.147 28.437 R17A R18 R178 R17A R16A R16A R16A nn 23.908 11.878 22.051 21.864 28.124 29.361 28.901 R17A R17A R17A RISA 04 20.875 21.407 ' 20.838 28.804 R16A R16A R16A BUNDLE 10 no 28.536 28.895 29280 EOFPL EXPOSURE (GWD/ST) 23 25 27 29 31 33 36 37 39 41 43 FIGURE 5.2.1 VY CYCLE 19 HALING DEPLETION. EOFPL BUNDLE AVERAGE EXPOSURES

• I i

1 R168 R18 R188 R178 R188 R18 R16A R16 R17A R17A R16A 22 32.500 13.335 32493 23.370 31.884 13.254 34.487 13.101 24.581 21.889 29.144 l'

R18 R16A R18 R17A R16 R17A R18 R188 M18 R17A R16A 20 I 13283 32.654 13.827 24.181 13.477 26.000 13.528 33.403 11.875 22.291 29.471 i

R188 R16 R178 R18 R18A R18 R17A R10 R178 R17A R16A 32.470 13.587 24.988 13403 33297 14.006 38J88 12.717 22.782 21.614 29.902 a

i R178 R17A MIS R16A R18 R178 R18 R18 R17A RIGA 18 '

23251 24.157 13.863 34.418 13.927 25.328 13.605 12.141 22.459 29.495 R188 R18 R16A R18 R178 R18 R16A R17A R188 14

< 31.891 13.425 33.217 13.885 26.114 13.879 33.332 23.471 28.753 R18 R17A R18 R178 R10 R17A R18 R17A R16A 12 j 13220 25.795 13.924 25.184 13.711 24.886 11.941 22.041 29.100 R16A R16 R17A R18 R16A R16 R17A R16A R16A 10 4

34237 13.492 26.181 13.316 33200 11.800 22.370 29.922 29.442 R18 R188 R18 R18 R17A R17A R16A R16A M \'

13.108 33.600 12.769 12.006 23.407 21.886 28J74 29.410 R17A Rie R178 R17A R16A RISA R18A m ~

j 24A06 11.740 22.820 22.438 28.882 30.011 29.478

. R17A R17A R17A R16A 21 A83 22.302 21.700 29.457 R16A R16A R16A SUNDLEID rs 29262 29.563 29 901 . EOFPL EXPOSURE (GWD/ST) 4 23 25 27 29 31 33 36 37 39 41 43 I I

FIGURE 5.2.2 VY CYQ?..iy RODDED DEPLETION. EOFPL BUNDLE AVERAGE EXPOSURES i

)

l 2.6 i

2.4 -

2.2 -

2.0 -

r i

1.8 c n 1.6 m- g E

1.4 -

N 7 12 ,r x, y e J

6 1.0 -

O Minimum Shutdown Margin 0.8 -

0.6 -

0.4 -

i 0.2 l

0.0 0 1000 2000 3000 4000 5000 0000 7000 0000 9000 10000 11000 CYCLE EXPOSURE (WWO/ST) l FIGURE 5.3.1 i

t I

VY CYCLE 19 COLD SHUTDOWN MARGIN. IN %AK. VERSUS CYCLE EXPOSURE l

l h

i i

6.0 THERMAL-HYDR AULIC DESIGN 6.1 Steady-State Thermal Hydraulics Core steady-state thermal-hydraulic analyses for the Reload Cycle were performed using the FIBWR computer code [14],[15],[16]. De FIBWR code incorporates a detailed geometrical representation of the complex flow paths in a BWR core, and explicitly models the leakage flow te the bypass region and water rod flow. De FIBWR geometric models for each GE bundle type were benchmarked against vendor-supplied and plant thermal-hydraulic information.

Using the fuel bundle geometric models, a power distribution calculated by SIMULATE 3 and core ialet enthalpy, the FIBWR code calculates the core pressure drop and total bypass flow for sev:ral power and flow combinations. De 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.

I 6.2 Reactor Limits Determination Section 3.11 of the Technical Specifications requires that the plant assure the performance of the fuel rods by not exceeding the Minimum Critical Power Ratio (MCPR), the Maximum Linear Heat Generation Rate (MLHGR), and the Maximum Average Planar Linear Heat Generation Rate l (MAPLHGR).

1 The Reload Cycle fuel has the MCPR operating limits shown in Appendix A. The MCPR is a combination of the Fuel Cladding Integrity Safety Limit (FCISL) and the change in a Critical Power Rttio (ACPR) which occurs during an anticipated operational transient. For Vermont Yankee, the FCISL, or commonly called the Safety Limit MCPR, is 1.10 for two loop operation or 1.12 for single loop operation [17],[18] 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.

The objective for normal operation and anticipated transient events is to maintain nucleate boiling.

Avsiding a transition to film boiling protects the fuel cladding integrity. Both the transient and normal MCPR operating limits are derived with the GEXL-Plus correlation [19), with appropriate coefficients representative of the Reload Cycle's fuel types. For core flows other than rated, the MCPR limits must 4

-ee -

r-

be adjusted by a generic factor, K, [19]. The analysis, described in the Section 7.0, determines the Reload Cycle MCPR operating limits.

The Reload Cycle fuel has a Linear Heat Generation Rate (LHGR) limit of 14.4 kW/ft for all bundle types. De basis for this limit can be found in Reference 2.

De Reload Cycle fuel has Average Planar Linear Heat Generation Rate (APLHGR) limits, shown in Appendix A. De Maximum APLHGR (MAPLHGR) values are the most limiting of the fuel rod thermal-mechanical MAPLHGRs [20] and the LOCA analysis MAPLHGRs (Section 8.2).

Da fuel rod thenral-mechanical MAPLHGRs are the result of the GE fuel rod thermal-mechanical design analyses, described in Reference 2. Rese results assume that during steady-state: 1) the maximum LHOR is 14.4 kW/ft,2) the maximum peak pellet exposure is (4).0 GWd/Mt, and 3) maximue operating time is 7.0 years. Dese results also assume that, c'uring an anticipated operational transient, the thermal and mechanical overpower limits [2O are not exceeded De transient analysis, described in Section 7.0, assures that the thernv] and mechanical overpower limits are not exceeded. De LOCA analysis, described in Section 8.0, determines the LOCA analysis MAPLHGRs.

F s

t

l I

7.0 ABNORMAI., OPERATIONAL TREfENT ANALYSIS ,

l 7.1 Irapsients Analyzed Transient simulations are performed to assess the impact of certain transients on the heat transfer characteristics of the fuel. 'Ihe purpose of this analysis is: 1) to determine the MCPR operding limit so that the FCISL is not violated for the transients considered, 2) to assure that the thermal rad mechanical overpower limits are not exceeded during the transient,3) to demonstrate compliance with the ASME vessel code limits and 4) to assure a pressure margin greater than 25 psi below the safety valve actuation settings.

Past licensing analyses have shown that these transients impact the operating MCPR limit:

l

1. Pressurization transients, including the generator load rejection with complete failure of the turbine bypass system and the turbine trip with complete failure of the turbine bypass system;

2. Loss of feedwater heating, including the loss of a feedwater heater aad loss of stator cooling;

3. Continuous Rod Withdrawal During Power Range Operation or commonly called the I local rod withdrawal error; and j

4. Fuel assembly insertion error during refueling or commonly called the misloaded bundle error including the rotated bundle error and the mislocated bundle error.

To demonstrate that the fuel rod thermal and mechanical overpowers are not exceeded, the maximum powers resulting from the pressurization, loss of feedwater heating and rod withdrawal error transients were compared to the criteria. To demonstrate compliance with ASME vessel code limits, the main steam isolation valves (MSIV) closing with failure of the MSIV position switch is also analyzed. To assure that there is a pressure margin greater than 25 psi below the safety valve actuation settings, a generator load rejection with complete failure of the turbine bypass system is

-28

l r

analyzed. To accommodate operation at 100% power with an inoperable safety relief valve (SRV), all l transient analyses assume the lowest setpoint valve fails to open. Brief descriptions and the results of the transients analyzed are provided in the following sections.

i 7.2 Pressurization Transients Analysis 7.2.1 Methodology [

De analysis involves two types of simulations. A system level simulation is performed to determine the overall plant response. Transient crie irkt <,cxi exit conditions and normalized power from the system level calculation are 8.1 en used to Mictm jetailed thermal-hydraulic simulations of the fuel, referred to as ' hot channel calculations." The hit charu,el simulations provide the bundle ,

transient t CPR (the initial bundle CPR minus the MCPR experienced during the transient).

De system level simulations are performed with the one dimensional (1-D) kinetics RETRAN ,

model [22],[23],[24]. De hot channel calculations are performed with the RETRAN [25],[26] and .

TCPYA0l[27],[16],[23] computer codes. De OEXIePlus correlation [19], contained in TCPYA01,  !

r evaluates the transient critical power ratio. ,

The hot channel transient ACPR calculations employ a two-part process, as illustrated by the flow chart in Figure 7.2.1. De 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 odel for each fuel type being analyzed, with bypass and water rod flow also modeled. De steady-state FIBWR analyses were performed at several power levels with other conditions (i.e., core pressure ,

drop, system pressure, and core inlet enthalpy) held constant. De FIBWR code results provide a steady-state CPR, active channel flow (AF) and bypass flow (BPF) for each active channel power (AP). ]

ne FIBWR conditions for channel power, channel flow, and bypass flow were then used as input to steady-state RETRAN/ICPYA01 hot channel calculations. Other assumptions are consistent

]

with those in the FIBWR analysis. De Initial Critical Power Ratio (ICPR) is the result of the

]

l steady-state RETRAN/I'CPYA01 analysis. Dese results allow for the development of functional i

f

i relationships, describing AP as a function of ICPR, and AF and BPF as functions of AP for each fuel type. Dese relationships are used in the iterative process for determining the transient CPR, as shown in Figure 7.2.1.

De second part of the hot channel calculations determines the transient CPR performance.

Because the ACPR for a given transient varies with Initial Critical Power Ratio (ICPR), the hot channel analysis is an iterative process. He objective of the hot channel iteration for each transient is to determine the hot channel initial conditions which result in reaching the FCISL. Each iteration requires a RETRAN hot channel run to calculate the transient enthalpies, flows, pressere and saturation properties at each time step. Dese are required for input to the TCPYA01 code. 'ICPYA01 is then used to calculate a CPR at each time step during the transient, from which a transient ACPR is derived.

In response to the NRC Safety Evaluation for FROSSTEY-2 [9], the hot channel methodology has considered the assumption of both fixed and time-varying power shapes. He fixed power shape assumes a 1.4 chopped cosine axial distribution which remains constant throughout the transient. De initial power shape for the time-varying power shape methodology is the Haling axial distribution used in the core wide analysis. The time-varying hot channel power distribution is assumed to be the same as that in the core wide analysis to account for the effects of transient power feedbacks and the scram.

The transient MCPR limits are defined as the more conservative results from the fixed and varying shape analyses. ,

7.2.2 Initial Conditions and Assumptions ne initial conditions for the Reload Cycle are based on a reactor power level of 1664 MW.

which includes a 4.5% margin on the current licensed reactor power level of 1593 MW.. This margin conservatively bounds the expected 2% calorimetne uncertainty. The reactor core flow is assumed to be 100% of rated. De core axial power distribution for each of the exposure points is based on the 3-dimensional SIMULATE-3 predictions associated with the generation of the reactivity data (Section 7.2.3). The core inlet enthalpy is set so that the amount of carry under from the steam separators and the quality in the liquid region outside the separators is as close to zero as possible. For fast pressurization tranrSots, this maximizes the initial pressurization rate and results in a more severe l

d 3 neutron power spike. ' A summary of the initial operating state used for the system simulations is provided in Table 7.2.1.

During the cycle, Vermont Yankee can adjust the core flow to account for reactivity changes rather than using the control rods. 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 also analyzed at the limiting exposure statepoint at 1664 MW. power and 87% flow. Limiting exposure is  !

defined as the exposure which had the highest ACPR.  ;

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

'I. Scram setpoints are at Technical Specification [10] limits.

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.

4. Safety / relief valve (SRV) and safety valve (SV) setpoints are modeled as being at the l Technical Specification setpoint plus 3%. Valve responses are based on slowest specified response values. One inoperable SRV is assumed in the analysis.

5. Control rod drive scram speed is based on the Technical Specification limits, except for the notch position 46 which is assumed to have a scram speed of 0.5 seconds. 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 Section 3.3.C.I.1 of the Technical Specifications. "67B" refers to the slower scram times given in Technical Specifications 3.3.C.I.2.

7.2.3 One-Dimensional Cross Sections and Kinetics Parameters The one-dimensional (1-D) cross sections and kinetics parameters are generated as functions of fuel temperature, moderator density, and scram. The method [28] is outlined below.

A complete set of 1-D cross sections, kinetics parameters, and axial power distributions are g nerated from base states using the Haling depletion established for EOFPL, EOFPL-1000 mwd /St, EOFPL 2000 mwd /St, and BOC exposure statepoints. Dese statepoints are characterized by exposure and void history distributions, contml rod pattems, and core thermal-hydraulic conditions. The latter are consistent with the assumed system transient conditions provided in Table 7.2.1.

De BOC base state is established by shuffling from the previously defined Current Cycle endpoint into the Reload Cycle loading pattem. A criticality search provides an estimate of the BOC critical rod pattem. De EOFPL and intermediate core exposure and void history distributiorss are calculated with a Haling depletion as described in Section 5.2. He EOFPL state is unrodded He EOFPL 1000 mwd /St and EOFPL-2000 mwd /St exposure statepoints require base control rod patterns. Dese are developed to be as " black and white" as possible to minimize the scram reactivity, maximize the core average moderator density reactivity coefficient and, therefore, maximize the transient power response. 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 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.

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 the control rod scram, another SIMULA'IE-3 reference case is run with all-rods-in. De 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. Dere is a table set l

for each of the 27 neutronic regions,25 regions to represent the active core and one region each for the bottom and top reflectors, 7.2.4 Turbine Trio Without Bvoass Transient (ITWOBP)  !

The transient is initiated by a rapid closure (0.1 second closing time) of the turbine stop l

valves. It is assumed that the steam bypass valves, which normally open to relieve pressure, remain closed. A reacter protection system signal is generated by the turbine stop valve closure switches.

Control rod drive motion is conservatively assumed to occur 0.27 seconds after the start of turbine '

l stop valve motion. He ATWS 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 ,

1 set generator field collapse. In sirrdating the transient, the bypass piping volume up to the valve chest is lumped into the control volume upstream of the turbine stop valves. Predictions of the salient system parameters at the three exposure points are shown in Figures 7.2.2 through 7.2.4 for the +

" Measured" scram time analysis.

7.2.5 Generator Lead Reiection Without Bvoass Transient (GLRWOBP) i The transient is initiated by a rapid closure (0.3 seconds closing time) of the turbine control valves. As in the case of the turbine trip transient, the bypass valves are assumed to fail. 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 i seconds after the start of turbine control valve motion. De same modeling regarding the ATWS pump trip and bypass piping is used as in the turbine trip simulation. He influence of the accelerating main turbine generator on the recirculation systein is simulated by specifying the main j turbine generator electrical frequency as a function of time for the M-G set drive motors. De 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 7.2.5 through 7.2.7 for the " Measured" scram time analysis. )

I

l 7.2.6 Pressurization Transient Analysis Results The transients selected for consideration were analyzed at exposure points of EOFPL, EOFPL-1000 mwd /St, and EOFPL-2000 mwd /St. De MCPR limits, in Table 7.2.2, are calculated by adding the calculated ACPR to the FCISL for the limiting bundle type in the core. The worst ACPR for the pressurization transients account for an exposure window of 0.0 to +600 mwd /St on Current Cycle and its impact on the Reload Cycle [7]. De fuel rod thermal and mechanical overpowers are not exceeded for these transients.

7.3 Loss of Feedwater Heatine Transient Analysis 7.3.1 Loss of a Feedwater Heater (LOFWH) Results A feedwater heater can be lost in such a way that the steam extraction line to the heater is shut cff or the feedwater flow bypasses one of the heaters. In either case, the reactor will receive cooler feedwater, which will produce an increase in the core inlet suocooling, resulting in a reactor power -

increase.

De response of the system due to the loss of 1007 of the feedwater heatmg capability was analyzed. His repitsents the maximum expected feedwater temperature reduction for a single heater or group of heaters that can be tripped or bypassed by a single event. The system model used is the same as that used for the pressurization transient analysis (Section 7.2.1). The initial conditions and modeling assumptions discussed in Section 7.2.2 are applicable to this simulation.

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 reactor power to reach its peak without scram. His approach was selected to provide a bounding and conservative analysis for events initiated from any power level.

The transient response of the system was evaluated at several exposures during the cycle, EOFPL, EOFPL 1000 mwd /St, EOFPL-2000 mwd /St, and BOC. Tx transient results, corresponding to the limiting bundle type in the core, are listed in Ti.ble 7.3.1. The MCPR limits in

i Table 7.3.1 are calculated by adding the calculated ACPR to the FCISL. De transient evaluation at EOFPL was found to be the limiting case. The results of the system response to a loss of 1007 feedwater heating capability evaluated at EOFPL, as predicted by the RETRAN code, are presented in Figure 7.3.1. The fuel rod thermal and mechanical overpowers are not exceeded for this transient.

l 7.3.2 Loss of Stator Cooline (LOSC) Results In response to a loss of stator cooling, a turbine runback is initiated to reduce generator output to less than 29% of rated output. His runback is accomplished by bypassing main steam from the turbine directly to the main condenser. Since heating steam to the feedwater heaters is supplied from the turbine stages, the amount of steam available for feedwater heating is significantly reduced. The

)

reduction of heating steam to the feedwater heaters results in a severe subcooling event.

For the analysis, the loss of stator cooling event is initiated at, or near, rated thermal power j I

(maximum 104.5%). It is assumed that an instantaneous loss of extraction steam occurs to the Nos.

14 feedwater heaters of both feedwrier trains. His is a conservative assumption, since there would not be a total loss of steam to the fudwater heaters, and the reduction in heating steam would occur over the several minutes required for the turbine runback. Also, no credit is taken for the heat l I

capacity of structural materials in the process piping or feedwater heaters. His results in a stepwise decrease in feedwater inlet temperature as the feedwater travels through the feedwater piping to the reactor vessel.

De decrease in feedwater temperature results in a subsequent reduction in core inlet temperature. Due to the negative void coefficient, core inermal power increases. De transient is terminated by APRM high flux trip at 120% of rated core thermal power.

The transient response of the system was evaluated at several exposures during the cycle, EOFPL, EOFPL-1000 mwd /St, EOFPL-2000 mwd /St and BOC. The transient results, corresponding l l

to the limiting bundle type in the core, are listed in Table 7.3.2. The MCPR limits in Table 7.3.2 are calculated by adding the calculated ACPR to the FCISL. De transient evaluation at several er.posures, including EOFPL, were found to be limiting. He results of the system response to a loss of stator l cooling evaluated at EOFPL, as predicted by the RETRAN code, are presented in Figure 7.3.2 as an

exunple. To assure that the thermal overpower limits are not exceeded, the MAPLHGR limits were modified.

7.4 Ovemressurization Analysis Results 7.4.1 Comoliance with ASME Vessel Code Compliance with ASME vessel code limits is demonstrated by an analysis of the Main Steam Isol tion 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 7.2.1). De initial conditions and modeling assumptions discussed in Section 7.2.2 are applicable to this simulation.

The transient is initiated by a simultaneous closure of all MSIVs. A 3.0 second closing time, which is the minimum time in Technical Specification Table 4.7.2, is assumed. A reactor scram signal is generated on APRM high flux. Control rod drive motion is conservatively assumed to initiate 0.28 seconds after reaching the high flux setpoint. One safety relief valve (SRV) is assumed to be -

inoperable. De system response is shown in Figure 7.4.1 for the "67B" scram time analysis.

The maximum pressures at the bottom of the reactor vessel calculated for the "67B" scram time analysis are given in Table 7.4.1. Dese results are within the ASME code overpressure design limit which is 110% of the vessel design pressure. Vermont Yankee's design pressure is 1250 psig so the maximum pressure limit is 1375 psig.

7.4.2 Safetv Valve Challences An overpressure analysis was performed to assure that sufficient pressure margin exists below the safety valve setpoint with one inoperable SRV during the limiting AOT. For this purpose, the limiting AOT for the Reload Cycle is the GLRWOBP at EOFPL conditions. The system model used is the same as that used for the transient analysis (Section 7.2.1). De transient scenario is the same as described in Section 7.2.5. De analysis is typically performed with best estimate assumptions.

However, this analysis was performed with the following conservative assumptions:

1 1

l 1

1) the same power and flow conditions described in Section 7.2.2,

2) an inoperable SRV,

3) all of the SRVs opening setpoint was assumed to have drifted up to 1110 psig (1% above 1100),

4) the safety valves opening setpoint was assumed to have drifted down to 1227.6 psig (1%

below 1240 psig), and

5) the " Measured" scram insertion times.

De acceptance criteria for this event is a pressure margin of greater than 25 psi to the safety valve setpoint. The results are greater than 25 psi as shown in Table 7.4.2. In conclusion, operation at full power with an inoperable SRV will not cause Safety Valve challenges during an AOT.

7.5 Local Rod Withdrawal Error Transient Results ne 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 MCPR operating limit at the time of the error, the continuous vdthdrawal of a control rod could increase both local and core average power levels with the potential for overheating the fuel. He consequences of the error depend on the local and overall core power increases, the initial MCPRs found in locations close to the error rod, and the ability of the Rod Block Monitor (RBM) System to stop the withdrawal of the rod before MCPR reaches the FCISL.

De purpose of this analysis is to develop a MCPR operating limit for the RWE, defined at each RBM setpoint, so that the FCISL is not violated. A broad spectrum of core conditions and rod patterns can exist at the time of such an error. This analysis evaluates the RWE, as initiated from all the " normal" rod patterns and power conditions, as defined in the rodded depletion (Section 5.1.2). In addition, a significant number of " abnormal" control rod partems and initial conditions are developed and analyzed. Dese abnormal cases serve the purpose of exaggerating the worth and ACPR impact of the error rod to be withdrawn, especially at the higher RBM setpoints.

De analysis for a given statepoint is performed using SIMULA'IE-3 in a quasi-static mode.

% the xenon distribution is held fixed at the initial state, but void and power distributions are

.- - - - - - .- - =._- . .- .. -. . . - -

ellowed to equilibrate at each new control rod withdrawal position during the transient. A power search increases the core average power in the model with each withdrawal step by searching on the critical eigenvalue of the initial state.

An evaluation of the instrument response of the Rod Block Mohitor (RBM) System to the rod withdrawal determines at what position the error rod is blocked. De RBM System's ability to terminate the RWE is evaluated on the following bases:

1. Technical Specifications [10] allow each of the separate RBM channels to remain operable if half, or more of the Local Power Range Monitor (LPRM) inputs on each level are operable. One RBM channel averages the inputs from the A and C levels; i the other channel averages the inputs from the B and D levels. For the interior locations tested in this analysis, there are four LPRM inputs per level. Thus, there are eleven failure combinations of none, one and two completely failed LPRM strings.

For each RBM channel, the instrument response as a function of error rod position is chosen to be the minimum response for these eleven LPRM string failure modes. ,

2. De event is analyzed separately in each of the four quadrants of the core, because of the different physical locations of the LPRM strings relative to the error rod.

3. Technical Specifications require that both RBM channels be operable during normal operation. Hus, the first channel whose response is calculated to intercept the RBM setpoint is assumed to stop the rod. To allow for control system delay times, the rod is assumed to continue moving for a minimum of two inches after the signal intercepts the RBM setpoint; the rod stops at the next available notch position.

Once the notch at which the rod stops moving is determined, the ACPR versus RBM setpoint can be calculated. De ACPR is calculated such that the implied MCPR operating limit equals the

]

FCISL + ACPR. This is done by conserving the figure of merit (ACPR/ICPR) found in the l

SIMULA*E-3 calculations. I i

As already mentioned, two types of RWE cases are evaluated - normal and abnormal.

- - - .-. - - . - - _ - .. - - - . _ ~ - - - . -

he normal cases are initiated from the expected rod patterns found in the rodded depletion. The normal cases have the following bases:

1. The core model is initially at full power, full flow and equilibrium xenon.

2. Every control rod sequence in the rodded depletion is examined. Within each l sequence a representative pattern is chosen. All deeply inserted rods (i.e., inserted above the core mid-plane) are withdrawn in separate cases.

I i

i Numerous normal cases are analyzed, as shown in Figure 7.5.1. An envelope is drawn that encompasses the most limiting ACPR results of all the normal cases. For conservatism, at least 0.02 ACPR margin is added by the envelope to assure that exposure uncertainties in the Current Cycle and the Reload Cycle are accounted for [7]. Table 7.5.1 provides the ACPR versus RBM setpoint i envelope for the normal cases, including the 0.02 ACPR uncertainty.

Figure 7.5.? provides the envelope of the worst of the abnormal case results. He abnormal cases are run to determine the sensitivity of the analysis to power level uncertainty and rod pattern variation. The abnormal cases are developed from the worst of the normal cases by specifying the follow'mg set of concurrent worst case assumptions:

1. The abnormal initial rod pattems are developed with xenon-free conditions. He xenon-free condition and the additional control rod inventory needed to maintain criticality exaggerates the worth of the withdrawn control rod when compared to normal operation with normal xenon levels.

2. The core is modeled at IN.5% power and 100% flow.

3. The core power distribution is adjusted with the available control rods (other than the error rod) to force assemblies within the four by four array of bundles around the error rod to be as close to the operating limits as possible.

-39 r

I l

I 1

l l

I Of the many abnormal pattern cases tested, an envelope of the most limiting ACPR results is created I 1

l by rounding-up to the closest 0.01 ACPR. As shown in Table 7.5.1, the ACPR derived from the l abnormal cases envelope is important in establishing the limits at the higher RBM setpoints. He final operating MCPR limit shown in Table 7.5.1 is based on the maximum ACPR of both the normal cases envelope (worst + 0.02 ACPR), and the abnormal cases envelope. De ' fuel rod mechanical overpowers are not exceeded for this transient.

7.6 Misloaded Bundle Error Analysis Results 7.6.1 Rotated Bundle Error i

ne primary result of a bundle rotation is a large increase in local pin peaking and the associated R-factor as higher enrichment pins are placed adjacent to the surrounding wide water gaps.

l 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. The objcetive of the analysis is to ensure that, in the worst possible rotation, the FCISL is not violated with the most limiting bundles on their operating limits.

To analyze the CPR response, rotated bundle R-factors as a function of exposure are l

developed by adding the largest possible AR-factor resulting from a rotation to the exposure dependent i R-factors of the properly oriented bundles. Using these rotated bundle R-factors, the MCPR values .

1 resulting from a bundle rotation are determined using SIMULATE-3. This is done for each control rod sequence throughout the cycle. The process is rcpeated with the K.,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 bundle's CPR on its FCISL. The adjusted MCPRs at each exposure is the rot-ted bundle operating limit for the rotated bundle error. He operating MCPR limit resulting from a rot: tion at any exposure is presented in Table 7.6.1.

l l

I i

7.6.2 Mislocated Bundle Error i l

l i

Mislocating a high reactivity assembly into a region of high neutron 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 mislocated bundle are the standard values for the given l i

fuel type.

i ne analysis uses multiple SIMULATE-3 cases to examine the effects of explicitly mislocating every older interior assembly in a quarter core with a fresh or once-bumed 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.

[

ne 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 mistoaded core at the misloaded location. De MCPR for the properly loaded core is adjusted by a ratio necessary to place the mislocated assembly on the FCISL. De maximum of these adjusted MCPRs is the mislocated bundle operating limit. De results of the mislocated bundle analysis are given in Table 7.6.2.

l 7.7 Transient Analysis Results The results of this transient analysis has: 1) determined the MCPR operating limit so that the FCISL is not violated for the transients considered, 2) assured that the thermal and rnechanical overpower limits are not exceeded during the transient,3) demonstrated compliance with the ASME vessel code limits and 4) assured there is a pressure margin greater than 25 psi below the safety valve actuation settings.

The MCPR operating limits for the Reload Cycle are calculated by adding the calculated ACPR to the FCISL at each of the exposure statepoints for each transient. Table 7.7.1 lists the limiting transient for each statepoint. For an exposure interval between statepoints, the highest MCPR l limit at either end is assumed to apply to the whole interval. De highest calculated MCPR limits for the Reload Cycle for e3ch of the exposure intervals for the various scram speeds and for the various

! 1 I

l

\ - . - , -- - -- ..

rod block lines are provided in Appendix A. These MCPR operating limits are valid for operation of the Reload Cycle at full power up to 10701 mwd /St and for operation during coastdown beyond EOFPL.

9

i l

l l

TABLE 7.2.1  !

i yY CYCLE 19

SUMMARY

OF SYSTEM TRANSIENT MODEL INITIAL CONDITIONS FOR TRANSIENT ANALYSES Core Thermal Power (MW.) 1664.0 l l

Turbine Steam Flow (10'Ib,/hr) 6 75 l

6 Total Core Flow (101b,/hr) 48.0 1

Core Bypass Flow (10'Ib /hr)* 6.28 Core Inlet Enthalpy (BTU /lb ) 523.2 i Steam Dome Pressure (psia) 1034.7 Turbine Inlet Pressure (psia) 985.7 6

Total Recirculation Drive Flow (101b,/hr) 23.7 l

Core Plate Differential Pressure (psi) 20.4 l l

Narrow Range Water Level (in.) 162.0 )

Average Fuel Gap Conductance (See Section 4.2)

• Includes water rod flow.

l

l l

TABLE 7.2.2 VY CYCLE 19 PRESSURIZATION TRANSIENT ANALYSIS RESULTS Peak Prompt Power Peak Average Heat Exposure (Fraction of Flux (Fraction of Transient Transient Statenoint initial Valuel Initial Value ACPR* MCPR Limits Turbine Trip Without EOFPL 2.88354 1.20382 0.24 1.34 Bypass, " Measured" EOFPL 1000 23 8439 1.14673 0.18 1.28 Scram Time EOFPL 2000 1.80549 1.06437 0.10 1.20 Turbine Trip Without EOFPL 3.21942 1.25097 0.31 1.41 Bypass, "67B" Scram EOFPL-1000 2.84459 1.20546 0.22 1.32 Time EOFPL-2000 2.23193 1.12560 0.14 1.24 Generator Load EOFPL 2.74082 1.17762 0.23 1.33 Rejection Without EOFPL-1000 2.15236 1.10768 0.16 1.26 l

Bypass," Measured" 1 EOFPL 2000 1.43162 1.01683 0.08 1.18 l Scram Time l

Generator Load EOFPL 3.28186 1.23935 0.30 1.40 Rejection Without EOFPL 1000 2.83940 1.18845 0.22 1.32 Bypass, "67B" Scram EOFPL-2000 2.07830 1.10680 0.13 1.23 Time

• The ACPR accounts for an exposure window of 0 to +600 mwd /St on the Current Cycle and its impact on the Reload Cycle [7].

l

TABLE 7.3.1 VY CYCLE 19 LOSS OF FEEDWATER HEATER TRANSIENT RESULTS 5

Peak Prompt Power Peak Average Heat Exposure (Fraction of Flux (Fraction of Transient Transient Statepoint Initial Value) Initial Value) M MCPR Limits

[

less of 100*F EOFPL 1.18034 1.18159 0.15 1.25 Feedwater Heating EOFPL-1000 1.19809 1.14579 0.12 1.22 EOFPL-2000 1.20639 1.14546 0.12 1.22 ,

BOC 1.14536 1.14565 0.12 1.22  ;

TABLE 712 VY CYCLE 19 LOSS OF STATOR COOLING TRANSIENT RESULTS Peak Prompt Power Peak Average Heat l

' Exposure (Fraction of Flux (Fraction of Transient l Transient Statenoint initial Value) Initial Value) M MCPR Limits l

Loss of Stator Cooling EOFPL 1.20089 1.18898 0.16 1.26 EOFPL 1000 1.20100 1.18894 0.16 1.26 EOFPL-2000 1.21592 1.11061 0.09 1.19 BOC 1.20187 1.18756 0.16 1.26 I

l TABLE 7.4.1 l

VY CYCLE 19 ASME COMPLIANCE RESULTS 1

l l

Maximum Pressure at Reactor Conditions Vessel Bottom (osin)

"67B" Scram Time 1319 TABLE 7.4.2 VY CYCLE 19 SAFETY VALVE SETPOINT CHAI IFNGE RESULTS 4 l

i Safety Valve Setooint (-1%)

Peak Steam Line Pressure Margin kiiE) kiial f2Eil 1190.8 1227.6 36.8

.m .._ . __ _ _ __.-_ - _ - _ _. _ __ _ _ _ . _ _ _ . _ _ _ _ . . . . _ . .

TABLE 7.5.1 VY CYCLE 19 ROD WITHDRAWAL ERROR ANALYSIS RESULTS Rod Block Monitor ACPR Envelone from ACPR Envelooe from Transient MCPR Setooint Normal Cases Abnormal Cases Limits (Worst + 0.02) (Worst Rounded Uo) 104 0.11 0.10 1.21 105 0.14 0.12 1.24 106 0.16 0.13 1.26 107 0.18 0.17 1.28 ,

108 0.19 0.22 1.32  ;

TABLE 7.6.1 VY CYCLE 19 ROTA*ED BUNDLE ANALYSIS RESULTS Transient MCPR Limit ,

f

'1.29 TABLE 7.6.2 VY CYCLE 19 MISLOCATED BUNDLE ANALYSIS RESULTS_

Transient MCPR Limit. _

i 1.17 .

t' l

TABLE 7.7.1 VY CYCLE 191.IMITING TRANSIENTS Rod Block Scram Exoosure Limitine Transient Monitor Setooint lialf (GWd/St) Transient MCPR Limit '

108 Measured 0.0 to 9.7 Rod Withdrawal Error 1.32 I 9.7 to 10.7 Turbine Trip 1.34 l

1 108 "67B" 0.0 to 9.7 Rod Withdrawal Error 1.32 l 9.7 to 10.7 Tustine Trip 1,41 1 107 Measured 0.0 to 9.7 Rotated Bundle 1.29 9.7 to 10.7 Turbine Trip 1.34 107 "67B" 0.0 to 8.7 Rotated Bundle 1.29 8.7 to 9.7 Turbine Trip 1.32 9.7 to 10.7 Turbine Trip 1.41 106 Measured 0.0 to 9.7 Rotated Bundle 1.29 9.7 to 10.7 Turbine Trip 1.34 106 *67B" 0.0 to 8.7 Rotated Bundle 1.29 )

8.7 to 9.7 Turbine Trip 1.32  !

9.7 to 10.7 Turbine Trip 1.41 Part I e e l

Part II l

l FIBWR and l RETRAN/ .

Estimate Initial TCPYA01 Power for St g tate Transient Analysis l l l l l l

' Functional

' Relationships ' RETRAN/TCPYA01 l AP = f(ICPR) l Transient Hot i

AF = f(AP) .

- Channel Analysis

' BPF = f(AP) l l l Estimate New ICPRI as:

'1 ~

ACPRo ICPRo No sient R=FCISL7 Yes a

STOP FIGURE 7.2.1 FLOW CHART FOR THE CALCULATION OF ACPR USING THE RETRANRCPYA01 CODES l

k TTWOBP,EOFPi.,MST TTWOBP,EOFPL,MST

....i.... ....i.... ....i.... ....i....i. ..i....i. .i. ..

4 93 3- -

125 - -

I w

> h ,, ..

a ,

< e a ,

• i .

b z e '.

- - 1.0 * -

'.' . . . ~ . .

I2- . '-' . .

0 i '.,

E .

1 J -

0.75 - \.

t

- 00Riiiturnow l- NORM. POWER l "" AVE. HEAT RUX 0 .- . .. ..

.i.. .i. ...... 0.5 .... . .. ..

.i. ,,

0.0 0.5 1.0 1.5 2.0 25 10 0.0 0.5 1.0 1.5 2.0 2.5 3.0 TNE(SEQ TNE(SEQ FIGURE 7.2.2 TURBINE TRIP WITHOUT BYPASS. EOFPL19 TRANSIENT RESPONSE VERSUS TIME ' MEASURED" SCRAM TIME

~ =_ _ . . _ - , . ~ . _ . _ _ . . _ _ ___~ __ m_. _. . . _ _ , .

TTWOBP,EOFPL,MST T1WOBP,EOFPL,MST

..l. ..t... 1. .t. .t. ..

g .f. .t. .l.. .I, .l. g .

T.

1500 -  !

- !m 2 4.

.::  :: :11  :  !: '

u:  :  ::

1000 -

i: i: :! y..!  : : 9 m -

. :n i

c.1200 -

O  :. .  :  :. 1.::

- }

j!

E 500 - ! ! j . i: ii -

o 5 i : i m  :

W m  :

a W o

: i G.

W T i i li: ,

g  : :

y  :: .. ,

c: :  ::

o -s00 - c;i.:  ; -

2 a ... .

1100 -

- u.

li

I

1. ! L^ -

1000 - t J isa - __, -

--* FEEDWATER l- STEAM DOWE PRES.l - VES$aSTW.0UT ,

1x0 ........ ... ,. ... 2000-- .,. . .. .,. ,

0 1 2 3 4 5 6 0 1 2 3 4 5 6 TIME (SEC) TIME (SEC) i l

l FIGURE 7.2.2 (Continued)

TURBINE TRIP WITHOUT BYPASS. EOFPL19 TRANSTENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME

-- . . .. . - . . = -..- - - _ -

p i  !

l I

1 J

r TTWOBP,EOFPL,MST 2h 12 i  ! a i

12 4 -

1.4 i -

1.2 !

.**.g 14 4 / i l.

I  %  :

0.8 d l \j >

l f 0.6 i ')

$ 0.4 i -

b 02 d -

6 0.04 .

> gy 4

...,~.

P ,'.' ~

1 04.4i ,

's g 42 i i

s

)>

4.8 i is a

i i

14 i -

i i

~

1.2 ! -

s 1.4 i I

- Tots.- i 1 A .j "" 00PPLE g

. . mm i i

1.8 i ... m -

s

-2.0 ,,- ,, ..

0.0 0.5 14 1.5 2.0 2.5 3.0 TNE(SEC) l l

l FIGURE 7.2.2 l

(Continued)

I TURBINE TRIP WITHOUT BYPASS. EOFPL19 TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TTME

-52

6 f

i i

TTWOBP,EOFPL 1,MST TTWOBP,EOFPL 1,MST

.i. .s.... .. .i. .i.

g3 ... .. . .. .i.. .i. .i.. .

4 i

i i

i 3- -

125 - - '

W r D  :

J

< f. >

> e

'. c k j ,I l

F i 1

i e '.

2- - -

1 .0 ., l ., - t b .

I Z i i j '**..., l E \., j

. i 1- -

0.75 - ',

(

.,' , i l

t

- CORE KET FLOW [

l-IORM POWER l *"* AVE. HEAT FLUX _.

0 .. .............. ........ , 0.5 .... .

0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 l

TNE(SEC) TNE(SEC)  !

i FIGURE 7.2.3  !

TURBINE TRIP WITHOUT BYPASS. EOFPL19-1000 MWDGT TRANSIENT RESPONSE VERSUS TIME.

• MEASURED" SCRAM TP.fE

l l

TTWOBP,EOFPL 1,MST TTWOBP,EOFPL 1,MST ,

3, ...i. . . .. .i. ..... .

• y****=....,......,,,*...*..

e f.

i.

i -

, 1500 - g  :

i lii  !!

:. . :. .r i.

,a -

!fi

! \ !?:.  : : : 4 l:\ -

(  :.  : : J :

in :i . .i.!  !! i g  :

i:,i' il:

9;. 1200 - i ,

w W 500 -  ! -

g . e 1 o i :..i t m e -

: 1

• i.:*i *i m d W

g e 0 -

c. W . >  ;

F  :  : -

g ': .

(-  :::  !

-500 - 5.i  !.!

2 Es }

1100 - - 14. . i; {

1.:

m 3. u  !

1000 - i -

l 1

I

~

1500 -

- DRVALVE(NEG)

.... FEED M TER l- STEAW DOME PRES.l ,

-- VESSEL STW.0UT 1000 . . ,, . . .,,. ..

,,. 2000 .. .. .. ..,. . ..

0 1 2 3 4 5 6 0 1 2 3 4 5 6 TNE(SEQ TNE(SEQ FIGURE 7.2.3 (Continued) i TURBINE TRIP WITHOUT BYPASS. EOFPL19-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME 1

I l

54-

1 1

1 l

t l

I TTWOBP,EOFPL-1,MST 24 - ' :. ' '

18 4  ! '

1.64  ! -

1.44  !

s  !

1.2 i / \  ! -

i 14 4 i i - 1

.?

0.84 / H  ;

60.6i

~

c: l 0.4 i F ,

02 i 0.0 ! >

4, .

' ..~.. .... .......... ~..**.

P o c.4 ! '. -

5 E 0.6 i i, i '

4.8 ! i

\ '

1.0 i i -

\ '

1.2 i i -

\

1.44 gg 1.6 i "" DOPPLER

--- WODERATOR '

1.8 ! m ,,

-2.0 ...... ...' ... .. . . .. i 0.0 0.5 1.0 1.5 2.0 2.5 3.0 TNE(SEC) 1 I

i l

l 1

FIGURE 7.2.3 (Continued)

TURBINE TRIP WITHOUT BYPASS. EOFPL19-1000 MWD /ST l TRANSIENT RESPONSE VERSUS TIME. MEASURED" SCRAM TIME

- . _ . . - . _ =

t TTWOBP, EOFPL-2, MST /0BP, EOFPL-2, MST

. ..i. ......i.... . .. . ' '' '

4 1.5 3- -

125 - -

W E D '

Y E >

.J k (

g g i , .,

Z 1 '. -

80 lz 2- ' .:/

g ,

i c '.

E b '

g ( **..,

E ,

i j -

0.75 - '..,

e l- = mal - amunux 0 .. ... ... .,. .... .,. . e.5 ...,. ..,.. .,. ... ......

0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 TNE(SEC) TNE(SEC)

FIGURE 7.2.4 TURBINE TRIP WITHOUT BYPASS. EOFPL19-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME TTWOBP, EOFFL-2, MST TTWOBP, EOFPL-2, MST

.i....i....e. ,i., .e.. ' l' ''

gg .. '

2000

-,-*****....g....,,...., r.

1500 -  : ., 3

r.  ::  ::

. :3  ::

• i i !! ^

1000 - El : i ! *!: !! -

5

! ! ! k.

m ^  :-

A 1200 -

O i 1. :: 1 m W 500 -  !  :

E . i ii 'i 3  : : :

m e  : : .

I m d .

w g e o

- i. :.i E $  : i c un  !!

.t g  :'..

2 O -500 - .i. :j i 1100 -

jj i a

e 3 L_

1000 - t 1

~ ~

- SAVALVE(NEG)

--- FEEDwtTER l- STEAM DOME PRES.l -- VESSELSTW.OW 1000 ... .,, .i.

.,. 2000 .,,. . .. .,, ,,, , ,,..

0 1 2 3 4 5 6 0 1 2 3 4 5 6 TNE(SEC) TNE(SEC)

FIGURE 7.2.4 (Continued) l TURBINE TRIP WITHOUT BYPASS. EOFPL19-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME ,

l i

l l

l i l l 1

l l

TTWOBP,EOFPL 2,MST 2.0 1.8 i -

I

/ \

1.6 i

+

1.4 .

l *. ) .,,

1.2 i l

1 1.0 i - 1 0.8 ! '-

l n

1 m 0.6 -i -

c  !

0.4 i F 0.2 ! **..<

l -

p 0.0 5 42; \

P *$. l 00.4i i F l 6 ' )

g 4.6 i ii F 42i i i 1.04 i s 1.2 i ii F

1.42 -

_ 7c73t 1.6 i " DOPPLER '-

- MODERATOR '

1.8 i .. m -

2.0 ........,'. -

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TWE(SEC)

FIGURE 7.2.4 (Continued)  ;

i TURBINE TRIP WITHOUT BYPASS. EOFPL19-2000 MWD /ST l TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME I l

f W

l l

Gl.RWOBP, EOFPL,MST GLRWOBP,EOFPL,MST

....i. ..i....i. .i....i. . ....i. .i.. .i. ..i. .s.. .

4 93 L

'! l P

3- -

1.25 -  ;

m E

h.

3 a o

< s a ,' -

8 E .

2 .

i 2- -

1 .0 - ., l .

h s.*

W Z

2. * '

s

< '. 1 E ',,

1- -

0.75 - -

y

- CORE NH ROW l- NOW.P0FR l -*

• AVE. lEAT RUX i 0 ... ... ... ..

... 0.5 ...... .... . ,,. ...

0.0 03 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 13 2.0 23 3.0 l i

TNE(SEC) TNE(SEC) i 1

i l

i I

k I

FIGURE 7.2.5 GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL19  !

TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME r i

a l

1 GLRWOBP,EOFPL,MST GLRWOBP, EOFPL,MST gg .. .i. .i. .i. .i.

2000

-r...............

4 i  :.

1500 - i- :i

:. f, i F. 19 :i

V.

\

1000 -

i ii. .i : .! i. .i ):~

i ii: i

! !/ i 1200 - - -

0 ii

.. ./ :l!

i. .

W W 500 -  !!:  !!; V -

e  : : :

3  : : :

I e  : : :

e d i !!

W  : : : :

m o .

ii i

]

e  !-  :.

i  ?

-600 - i; -

lE g  :-

y g 1100-

" L-3o00 ,

~1* ~ ~

- ssivatvE PEG)

.... so miin l- STEAW DOME PRES.l --- VESSELSTM.0LH 1000 ,,,,. ,,. .,, ,,. ,,. 2000 ,,, ,,,,, .,,, ,,, ,,.

0 1 2 3 4 5 6 0 1 2 3 4 5 6 TIME (SEC) TIME (SEC) e

• 4 FIGURE 7.2.5 (Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL19 TRANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME

( -

GLRWOBP,EOFPL,MST 2.0 '.'

1A :

f ~

1.6 i  ! -

1.4 i  !

12 ;  ! -

.~.

1.0 :  !.i1  ! -

0.8 :  !

\. /

60.6:  !

C i 50.4i ,

02 : ,*..

- 4 O.04 -

4y .; s,... ................*..* -

b 0.4 :

g l-g 0.6i s.

I 4A4 '. F i

\

1b i -

12 i '.i 1.4 i E

- -- TOTE \-s 1.6 -! "" DOPPLER i  :.

-- MODERATOR  ! -

1.8 ! .m t -

i

-2.0 . .

.... ..  ? . . , , '. ..

0.0 0.5 1.0 1.5 2.0 2.5 3.0 ThE(SEC) 1 FIGURE 7.2.5 (Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL19 TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME 1

GLRWOBP,EOFPL-1, MST GLRWOBP,E0FPL 1,MST 4

....i.... .... . .. .. . .... 1.5 (

1 1

i 3- -

1.25 - -

m E D l Y

.3

@ < , 'sl' E

8 ,

f 2- -

f 1.0 ., l ',* -

8 o .,- , 1 z z '. l 3

.Q ,...

g  :

< \ l E \. ,

l 1

0.75 - ',, -

l

- CORE N.ET R0W I

+ l AVF.HEATRUX l- NORW. POWER l 0 .. ... ...

... ,,. 0.5 .

.... ... ... ,,. ... l 0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 l TIME (SEC)  !

TIME (SEC)

I 1

i FIGURE 7.2.6 GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL19-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME 1

4 i

i i

i l

l 1

1 GLRWOBP, EOFPL-1, MST GLRWOBP, EOFPL-1, MST

.i. .i. ..i. .i.... ' ~

gg 2000 f............... y

'i 1500 - i ,

.  : f.

r.  :. '.

!! i i /-

i ! I.'. i i i I.:

1000 - i i' .i 'i ji: ~ -

l 5 .i  ! i/ i i m  : !

a 1200 - - ^

O i i.: ./  !. .!

W W Soo .

,i :j -

a ,

3 I m m m

• W  :

E

• 0 .

A W H

W <  :/

E il 3 . I -

O -500 -  !

2 - g -

1100 - ii e L 1000- -

• ~ ~

~ SEVRVE(NEG)

-* FEEDWATER l- STEAW DOWE PRES.l - VESSEL STW.0VT 1000 ....,. . ...,,.. ,,.. ,,.. -2000 . .. ,,.. ,,,. ., ... . .

0 1 2 3 4 5 6 0 1 2 3 4 5 6 TNE(SEC) TNE(SEC) l l

l i

i FIGURE 7.2.6 (Continued) i GENER ATOR LOAD REJECTION WITHOUT BYPASS. EOFPL19-1000 MWD /ST TR ANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME

l GLRWOBP,EOFPL 1,MST 2.0 ,

1.8 i -

1.6 i -

i 1.4 i ,5 12 4 l \ *s l. -

1.0 i l -

0.8 i f -

6 0.6- l l

c  :

l 0.4 i l F 0.2 i f '

-5 O.0 (2-0.2 . .',

b 0.4 i ','- -

0.6 i ',

i 424 i F i

1.0 i i F i

12 i i '

F 1.4 i - ToiAL ~ l 1.6 i , p

""..DOPPLER NTOR 'i '

1.8 i . .. m \

i 2.0 .

,4 ... ... ... ..

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC)

FIGURE 7.2.6 (Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL19-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME 64-

b I

I 4

i GLRWOBP,EOFPL 2,MST 'GLRWOBP, EOFPL 2, MST  ;

..n....i.. .i. . ....i. ..i. .i. ..i. .. .. -

4,u ..i.... . 33 3- -

1.25 - -

W >

g 3 =

I E .

E -

1 , -  !

2

$- 80

'.../  !

2 2 .,  ;

i '. l E

E e

u.

I 1

0.75 - ',

i

- CDRE NET ROW .

l- NORK POWER l **= AVE.HEATRUX 0 ........,....,. .. .. ,,. . 0.5 .... . .......,...........

0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 TNE(SEC) THE(SEC) l FIGURE 7.2.7 GENERATOR LOAD RFJECTION WITHOUT BYPASS. EOFPL19-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME

• MEASURED" SCRAM TIME GLRWOBP,EOFPL 2,MST GLRWOBP,EOFPL 2,MST

,,,ifi, .l. .f, ,f,,iit,,,, ,,I,,,,f,, ,t,,,,f,,,,1, gg g ,

-r***==......... ,,,

• a..*

5

. 1500 -  !! -

35 l S. I

.!!. i. .i . .i:

1000 -

I i ' l E. !! -

E m  :

i l 1 l '.

,i :: -

E 1200 -

g W

l i

~

jf il w 500 . -  !!

• E -

3 .  :

m m e s w 1 g m o .

1

c. W .

F .

W <  !!

E  : !!

3:  : ! y 0 -500 - ;s) 2 .

1100 - -

i d i 1000 -

~

'I# ~

- SR VALVE (NEG)

-

• FEEDWATER l- STEAM DOMEPRES l --- VESSEL STM. 0bi 1000 ,

,,.... . . ...,,.. .. -2000 .

0 1 2 3 4 5 6 0 1 2 3 4 5 6 TIME (SEC) TIME (SEC)

FIGURE 7.2.7 (Continued)

GENERATOR LOAD REIECT10N WITHOUT BYPASS. EOFPL19-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME i

GLRWOBP,EOFPL-2,MST ,

2.0 ''' / '

i '

184 i 1.6 -i

/"*' F -

l 1.4 i e 12 2  ! F 1.0 -!

l 0.8 i  ! F .

to 0.62  !

E  :

0.4 ! l F

02 i l -

0.0 --.................=**..*

l '

> 42 - i F b4 .4 i i l

$ t g 4.6 i * -

i 4.8 i i s F ,

i 1.0 i s '

1.2 i \s

~~

1.4 i - - TcTE 1.6 i *-*- DOPPLER l

M TOR '

1.82, -

. .. m 2.0 ........... ,,...... ....

l 0.0 0.5 1.0 1.5 - 2.0 2.5 3.0 l

l TIME (SEC) 1 FIGURE 7.2.7

, (Continued)

GENERATOR LOAD RFJECTION WITHOUT BYPASS. EOFPL19-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME 1

l l

l 1

l 1

LOFWH, EOFPL LOFWH, EOFPL i

i 1.25 O.5 O.4 - -

)

03 - -

l l """""""*~~~--

t w

3 .

/ 0.2 -

t

/ ** ****"'~ -

J

• m / 1

<  ! g i 1

>

• s

.J

! 0.1 - /

E .-  ;-

Z

u. 1.o .

0.0

. - - . - . . . . . . . . . . . . . . . _ l O >

2 -

0 $ -0.1 -

8 6 '.'.

E .

b 0.2 - -

N.....

l 4.3- -

1

- TOTAL l - ou P0wm .... DO R ER -

.o.4

.--- CORE KET R0W -- MODERATOR

." AVE.HEATFLUX -- - SCRAM 0.75 ..

... .. .... ... -0.5 .... . ........... ...

0 25 50 75 100 125 150 0 25 50 75 100 125 150 TIME (SEC) TIME (SEC) l l

1 FIGURE 7.3.1 LOSS OF 100*F FEEDWATER HEATER (LIMITING CASE)

TRANSIENT RESPONSE VERSUS TIME 68

. _- - . _ = _ . _ .- - .

l l

1 i

j i l

t l LOFWH,EOFPL LOFWH, EOFPL

...i. .i. .i.. .i. .i.... g .... . .. . ..e.. .i. .. .

jg l

g. . .

30

^

n%- - -

g' c 70 - -

S 6 e w g 60 -

g. .

a) 2000 - -

g g <

0: ,.*

u) <

g 40 ,

.3 I

w g. .

8 1750 - -

o 20 -

10 -

- FEED WATER R0W l- COREIN SUBC00UNGl **** VESSEL STR OWLET 0 . . .. . .

.i. . ......... 1500 ............ ............ . ,

0 25 50 75 100 125 150 0 25 50 75 100 125 150 THE(SEC) TNE(SEC) i l

l FIGURE 73.1 (Continued)

LOSS OF 100*F FEEDWATER HEATER RIMITING CASE)

TRANSIENT RESPONSE VERSUS TIME I

LOSC, EOFPL, MST LOSC, EOFPL, MST 93 ....e....t....t. .t.. . . '

03 I:s

'8 e

0.4 - ll -

10 i e

,e

/

/ 0 ,1

./ .e 0.3 -  ;

, /**'

W o

/! 0.2 - / l -

4 ll m

/ l

/. i

> t E j /  ! $ 0.1 - l

s /

l t- .- 8 .**.,,

2 0.0 -W--- -- u-- - - - --

i

~

1 * - * - !

80- l e

l i

g  % i i

! U 0.1- l -

e g *.., ,

g e .,  !

[ 0.2 - ',

. l

\

l 0.3 -

- TOTAL 1

- NORK POWER , 0.4 - * * *

• DOPPLER -

.... m g g gow .

NTOR AVE. m i nux --- m 0.75 ......... .... . ... .... . 0.5 .............. , ..... .....

0 25 50 75 100 125 150 0 25 50 75 100 125 150 TIME (SEC) TIME (SEC) l e

i l

FIGURE 7.3.2 LOSS OF STATOR COOLING TRANSIENT RESPONSE VERSUS TIME l  !

LOSC,EOFPL,MST LOSC,EOFPL,MST

.t. ..i., .t. .1,. .t. g ...t. .t. ..t.. .I....I..

gg . ..

g. .

80 -

2250 -

{- -

70 -

6

• 8 W g60-h d O

o 50 -

@ 2000 -

m F 3 4

0) E 3

Y 40 - o - -* ' ........ ----l fw i e l

g. .

l 8 1750 - l o ,

m. ,

I e

10 - - I

- FEED WATER R0W l l- CORE IN SUB0000NGl - - VESSEL STW.0UTLET l 0 . ., ......... ... 1500 ......... ... , , ' . ....

0 25 50 75 100 125 150 0 25 50 75 100 125 150 TIME (SEC) TNE(SEC) l l

l FIGURE 7.3.2 l (Continued)

LOSS OF STATOR COOLING TRANSIENT RESPONSE VERSUS TIME 1

l l

....i. .i.. .. .i. ..i. . .i. ..i.. .i....i....i. .

4 33 l

I 3- -

1.25 -

g g s ,/, ,*** .,*..

a l

@ (

F /

2 ---- ~~'.**

2- - ~

1 .0 w b 2 2 i O E b l 0 <

l 2 e

! L 1

0.75 -

l l

- CORE RETROW l- NORM. POWER l **** AVE HEAT RUX 0 . .,, ,,,. ,,.. ... ... 0.5 .

0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC) TIME (SEC) l l

1 l

i FIGURE 7.4.1 MSIV CLOSURE. FLUX SCRAM. EOFPL19 TRANSIENT RESPONSE VERSUS TIME '67B" SCRAM TIME n

l l

f f

1330 2000

\

1500 - i ...*.N.

i:

1 1000 - { is -

7 i f.!iM v.. .~...

m  : &-

E 1230 - o  ;

W W 500 - i -

E  : -

3 w

g m 0

Al$EIl.!

4 g 3

(

e

~ ~

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- STEM DOME PRES.l - VESSEL STM.0UT i

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,,,,,,,,,..,,,.,,,, ,i. 2000 ,,..,. ,,,,,,,,,.,,,. ,.,, ,

0 1 2 3 4 5 6 0 1 2 3 4 5 6 TNE(SEC) TNE(SEC)

FIGURE 7.4.1 ,

j (Continued) l MSIV CLOSURE. FLUX SCRAM. EOFPL19 TRANSIENT RESPONSE VERSUS TIME. "67B" SCRAM TTME I

73-l t

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0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC) i l

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FIGURE 7.4.1 (Continued)

MSIV CLOSURE. FLUX SCRAM. EOFPL19 TRANSIENT RESPONSE VERSUS TIME. "67B" SCRAM TIME i

1 l

CYCLE 19 NORMAL CASES RWE RESULTS

- NORMAL PATTERN CASES a ENVELOPE (Worst + 0.02) 0.22 l

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/ A 0.,2 y)y z x -

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0.06 ., y 0.04 104 105 106 107 108 ROD BLOCK MONITOR SETPOINT FIGURE 7.5.1 VY CYCLE 19 NORMAL RWE CASES RESULTS

- ACPR VERSUS RBM SETPOINT l

l l

l l

)

CYCLE 19 ABNORMAL CASES RWE RESULTS

- ABNORMAL PATTERN CASES m ENVELOPE (Rounded up)

I O.22 -

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f 0.06 f/ f 0.04 104 105 106 107 108 ROD BLOCK MONITOR SETPOINT FIGURE 7.5.2 VY CYCLE 19 ABNORMAL RWE CASES RESULTS

- 6CPR VERSUS RBM SETPOINT 8.0 DESIGN BASIS ACCIDENT ANALYSIS 8.1 Control Rod Droo Accident Results ne control rod sequences are a series of rod withdrawal and banked withdrawal instructions specifically designed to minimize the worths of individual control rods. De 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 refear. to the fact that rods beyond Group 2 are leaked 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 per gram.

Some out-of-sequence control rods could accrue potentially high worths. However, the Rod Worth Minimizer (RWM) will prevent withdrawing an out-of-sequence rod, if accidentally selected.

De RWM is functionally tested before each startup.

De sequence in the RWM will take the plant from All Rods In (ARI) 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 [29],[30],[31]. Below 20% power, however, the sequences must be examined for incremental rod worth. His is done throughout the cycle using the full core, xenon-free SIMULATE-3 model.

Both the A and B sequences were examined at various exposures throughout the cycle. For staitup, the rods are grouped, as shown in Figures 8.1.1 and 8.1.2, and are pulled in numerical order.

All the rods in one group are pulled out before the pulling of the next group begins. De 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 which reduce the rod incremental worths [32),[33].

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

i i

l l

l De second group of rods was examined using the following analysis method [34). Both the A and B sequences were examined. It was found that the highest worth rod was the first rod in the  :

second group. Any of the first four rod arrays, shown in Figures 8.1.1 and 8.1.2, may be designated  !

as the first group pulled. However, a specific second group must follow as Table 8.1.1 illustrates. For added conurvatism, each of the high worth rods in the second group were checked; i.e., one at a time, j they were assigned to be the first rod pulled. His assures that in any sequence the actual worths will always be less than those calculated here. l Only that portion of the control rod worth above the SIMULATE-3 cold critical eigenvalue j contributes to the rod drop accident. For conservatism, " critical" was defined as the SIMULATE-3 1

cverage cold critical K,,r minus 1% AK (reactivity anomaly criteria). De results of the Group 2

]

calculations, as presented in Table 8.1.2, fit under the bounding analysis of References 29 through 31. i l

l Beyond Group 2, the rods are banked out of the core. His generally limits the incremental-  !

I worth of a single rod drop; however, virtually all of the pre-drop cases in Group 3 are critical.

nerefore, the 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. He types of cases examined included:

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

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

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

l

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

5. Stuck rods from previously pulled Group 1 or 2 dropping from 00 to 48.

1 The highest worth results from the Group 3 analysis fit under the Group 2 results, presented in Table 8.1.2.

l 4

8.2 Loss-of-Coolant Accident Analysis "Ihe LOCA analysis, performed in accordance with 10CFR50 Appendix K and the Safety Evaluation Reports [35),[36), demonstrates that Vermont Yankee, operating within the assumed conditions, complies with the LOCA limits specified in 10CFR50.46.

The LOCA analysis for the Reload Cycle is a combination of cycle specific analysis and a base analysis for Cycle 17[37). Both analyses use the NRC-approved codes, FROSSTEY-2[9] and RELAP5YA[38). The base analysis provided the break spectmm and the single failure conditions.

The Reload Cycle analysis provided the verification that the base analysis was valid for the Reload Cycle changes.

Several changes have been made to the model during Cycle 18 since the Cycle 18 Core Performance Analysis Report [39): 1) an error in the fuel behavior model in RELAP5YA was corrected,2) the recirculation bypass valves were added as being open,3) the LPCI flow was reduced by 500 gpm to account for the RHR minimum flow valve being open at all times except during shutdown cooling, and 4) a change in the trip logic for ECCS pressure permissive and ADS actuation to represent the containment under harsh environmental conditions. The model changes for Cycle 19 were: 1) the core spray flow was reduced by 300 gpm to allow the core spray minimum flow bypass valves to be open at all times and 2) minor cycle-specific changes to the physics data, stored energy and the central average region. All other assumed initial conditions and assumptions are the same for both analyses. Table 8.2.1 lists some of the key input assumptions but Reference 37 provides a more detailed listing cf the input assumptions.

The Cycle 17 base analysis was performed for a combination of break size, break location, and single failure conditions. De break sizes range from 0.05 ft2 to 7.28 ft2 . Five break locations were analyzed: main steam line, core spray line, feedwater line, recirculation loop suction and recirculation loop discharge. Five possible single failures were evaluated: low pressure coolant injection valve, high pressure coolant injection, DC power supply, diesel generator and one automatic depressurization system valve. De impact of the Od O 2 on 3 initial volume average temperature and material properties was included Table 8.2.2 lists the individual PCT results for each of the changes identified above. The Reload Cycle analysis was performed for the limiting break size and two single failure conditions.

He Reload Cycle analysis results show that the limiting break is 0.6 ft2 in the recirculation loop at the pump discharge with one DC power supply as the single failure and loss of offsite power coincident with the break opening. The maximum PCT results was 1801.7 F. He Peload Cycle analysis also showed that the break spectrum performed for the base analysis remains valid for the Reload Cycle.

The analysis shows compliance with the other 10CFR 50.46 limits: the calculated peak clad temperatures are well below the 2200*F limit, total cladding oxidation at the peak location is less than 17%; hydrogen generated in the core is less than 1%; and the core retains a coolable geometry with no clad rupture.

During the cycle, Vermont Yankee can adjust the core flow to account for reactivity changes rather than using the control rods. 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, the LOCA analysis was analyzed at 1698 MW. power and 87% flow. He results showed that the 100% flow case bounded the low flow case.

He analysis showed that the MAPLHGR limits are not limited by a LOCA. Herefore, the MAPLHGR limits are set based on the thermal-mechanical analysis of the bundle. They are provided in Appendix A for all the fuel types in the Reload Cycle, as a function of average planar er.posure.

The analysis also verified that the single loop MAPLHGR multiplier,0.83, is valid for the Reload Cycle.

83 Refueline Accident Results If any assembly is damaged during refueling, then a fraction of the fission product inventory could be released to the envimnment. He 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 aedvity is based on the MAPLHGR operating limits, tie maximum operating fuel centerline temperatures, and maximuni bundle bumup. The fuel rod gap activity, intemal pressure and centerline temperature for the Reload Cycle are bounded by the values used in Secdon 14.9 of the FSAR(40).

y- - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

TABLE 8.1.1 CONTROL ROD DROP ANALYSIS - ROD ARRAY PULL ORDER The order in whi-h rod arrays are pulled is specific once the choice of the first group is made.

First Group Second Group Successive Group Pulled Is: Pulled Must Be: Is Banked Qui Array 1 Array 2 Arrays 3 or 4 Array 2 Array 1 Arrays 3 or 4 Array 3 Array 4 Arrays 1 or 2 Array 4 Array 3 Arrays 1 or 2 TABLS 8.1.2 VY CYCLE 19 CONTROL ROD DROP ANALYSIS RESULTS Maximum Incremertal Rod Worth Calculated 0.94% AK Cold, Xenon-Free Bounding Analysis Wonh for Enthalpy Less than 1.30% AK 280 Calories per Gram [29),[30],[31]

TABLE 8.2.1 LOCA ANALYSIS ASSUMirTIONS Core Thermal Power (MW.) 1698.3 6

Total Core Flow (101b/hr) 48.0 Reactor Vessel Pressure (psia) 1067.0 6

Recirculation Loop Flow (101b/hr) - Each Loop 12.3 6

Feedwater Flow (101bdhr) 6.93 Feedwater Temperature (*F) 377.0 Water Level Above Top of Enriched Fuel (in.) 130.0 Containment Drywell Pressure (psia) 16.5 Containment Wetwell Pressure (psia) 14.7 Con:ainment Wetwell Liquid Temperature ('F) 165.0 Maximum Bundle Power (MW.) 7.3 Maximum Averace Planar Linean Heat Generation Rate (kW/ft) 13.6*

Maximum Linear Heat Generation Rate (kW/ft) 14.4"

• Plus Calorimetric and TIP Reading Uncertainties (8.9%): 13.6 x 1.089 = 14.8 kW/ft

" Plus Calorimetric and TIP Reading Uncertainties (9.2%): 14.4 x 1.092 = 15.72 kW/ft TABLE 8.2.2 LOCA ANALYSIS RESULTS. PEAK CLADDING TEMPERATURE Descriptions of Model Chance PCT (*F)

Correct conservative error in the fuel behavior model in 1778.1 RELAP5YA*

Accounting for the harsh environmental conditions 1780.5 l Opening the Recirculation Bypass Valves 1764.5 LPCI flow reduction with the opening of the RHR 1793.9 Minimum Flow Valve Changing the fuel for Cycle 19 and LPCS flow reduction 1801.7 with the opening of the Core Spray Minimum Flow Bypass Valve and the above three changes b

W

• Already reported to the NRC in accordance with 10CFR50.46(a)(3)(ii). -

49 43 3 e 39 2 1 1 2 M 4 3 4 3 4 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 II 4 3 4 3 4 07 2 1 1 2 03 3 02 06 10 14 18 22 26 30 34 38 42 FIGURE 8.1.1 FIRST FOUR ROD ARRAYS PULLED IN THE A SEOUENCES 84

43 3 3 39 2 1 2 35 3 4 4 3 31 2 1 2 1 2 l

l 27 3 4 3 3 4 3 23 - 1 2 1 2 1 19 3 4 3 3 4 3 15 2 1 2 1 2 11 3 4 4 3 07 2 1 2 03 3 3 02 06 10 14 18 22 26 30 34 38 42 FIGURE 8.1.2 FIRST FOUR ROD ARRAYS PULLED IN THE B SEOUENCES

9.0 STABILITY ANALYSIS The reload core design is analyzed each fuel cycle to assure that the potential occurrence of a thermal hydraulic oscillation or instability can be prevented through avoidance of operation in conditions susceptible to an instability. An assessment of the APRM flow biased scram setpoints is also made to assure that for the expected mode of oscillation in Vermont Yankee, the core wide mode, the setpoints are adequate to prevent exceeding the FCISL if an oscillation is not prevented and remains undetected by an operator. The identification of operating conditions susceptible to instabilities is conservatively predicted for the operating cycle through application of the BWROG stability exclusion region methodology [41]

and with the LAPUR code [42]. The BWROG application involves probing high powered, low flow conditions for a decay ratio of 0.8, the upper limit for acceptable operation of the reactor at off nominal conditions. Several operating states from natural circulation to approximately 60% flow and power are calculated to form a boundary at a 0.8 decay ratio. Operation above this boundary or in the " exclusion region" is prohibited for normal operation. For Cycle 19, LAPUR was applied to evaluate the relative stability characteristics of the reload core to confirm that the exclusion boundary calculated with the BWROG methodology remains valid.

Results of the analysis for the Cycle 19 exclusion region are shown in Figure A.1, with the corresponding data points composing the boundary contained in Table A.S. Also, contained in Figure A.1 is an additional boundary referred to as the buffer region. This region is an area ofoperation where, under unexpected operating circumstances, it may be possible to experience an instability. Operation within the buffer region requires the use of the plant process computer stability monitor. The stability monitor provides the decay ratio for the actual operating conditions. If while operating in the buffer region the monitor calculates a decay ratio of 0.8, the operator is directed to exit the buffer region.

Should the stability monitor fail to function, the buffer region is considered part of the exclusion region.

Justification of the APRM flow biased scram for detecting and suppressing a core wide mode instability is carried out using the BWROG detect and suppress reload methodology [41],[43]. The purpose of the reload methodology is to confirm that the combination of APRM trip setpoints and operating MCPR limits provides a high confidence that the FCISL will not be violated for the anticipated core wide oscillations. It consists of checking the operating MCPR limits, the fuel type and that the hot bundle oscillation characteristics have not changed. For Cycle 19, the MCPR operating limits have been

The increase provides a more conservative initial operating state. The fuel type, GE9B, has not changed since the last analysis ibr Cycle 15 [44]. The hot bundle oscillation characteristics have also not changed since Cycle 15, nor have the. flow biased scram serpoints, the two primary factors which would influence the hot bundle oscillation magnitude prior to power suppression by the APRM scram. Thus, it is concluded that the APRM flow biased scram setpoints remain adequate for oscillation suppression prior to exceeding the MCPR.

O s

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

'Ihe Vermont Yankee Startup Program will include process computer data checks, shutdown magin demonstration, in-sequence critical measurement, rod 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 ofInspection and Enforcement in the past[45).

11.0 CONCLUSION

His report presented the design information, calculational results, and operating limits pertinent to the operation of the Reload Cycle. The core is designed to consist of120 new GE-9B fuel bundles and 248 irradiated GE-9B fuel bundles. He shutdown margin for the Reload Cycle is greater than the Technical Specification limit. The bundles used in the Reload Cycle do not exceed the Technical Specification limit of 1.31 K. for storage in the spent fuel pool or the new fuel storage facility. He transient analysis has: 1) determined the MCPR operating limits so that the FCISL is not violated for the transients considered, 2) assured that the thermal and mechanical overpower limits are not exceeded during the transient,3) demonstrated compliance with the ASME vessel code limits and 4) assured a pressure margin greater than 25 psi below the safety valve actuation settings. The control rod drop worth is less than the bounding analysis which demonstrates a maximum fuel enthalpy less than the Technical Specification limit of 280 calories per gram. He LOCA analysis demonstrates compliance

, with the acceptance criteria specified in 10CFR50.46. The fuel rod gap activity, intemal pressure and centerline temperature are bounded by the values used in Section 14.9 of the FSAR which demonstrates the limits of 10CFR100 are not exceeded for a refueling accident. He stability analysis has defined an exclusion and buffer regions that assure the 10CFR50 Appendix A, General Design Criteria 10 and 12 are met for the Reload Cycle.

REFERENCES

1. R. C. Paulson, Vermont Yankee Cycle 17 Summary Report. YAEC-1919 (October 1995).

2. General Electric Company, General Electric Standard Application for Reactor Fuel (GESTARII), NEDE-240ll-P-A-13, GE Company Propriet q, as amended (August 1996).

3. General Electric Company, General Electric Fuel Bun / . sesigns, NEDE-31152P, Revision 5, GE Company Proprietary, as amended (June 1996).

4. GE Letter, P. J. Savoia to R. T. Yee, " Transmittal of Documentation for Vermont Yankee Reload 18 Fuel Bundle," PJS95133 (August 29,1995).

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

6. R. A. Wochlke, et al.; MICBURN-3/CASMO-3/ TABLES-3/ SIMULATE-3 Benchmarkinc of Vermont Yankee Cycles 9 throuch 13: YAEC-1683-A (March 1989).

7. B. Y. Hubbard, et al.; End-of-Full-Power-Life Sensitivity Study for the Revised SWR Licensine Methodolocy; YAEC-1822 (October 1991).

8. K. E. StJohn, S. P. Schultz and R. P. Smith; Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects: YAEC-1912P-A (January 1995).

9. USNRC Letter to L. A. Tremblay, SER, "Vermor.t Yankee Nuclear Power Station, Safety Evaluation of FROSSTEY-2 Computer Code (TAC No. M68216)," NVY 92-178 (September 24, 1992).

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

11. VYNPC Letter to USNRC, " Inverted Control Rod Poison Tubes at Vermont Yankee," WVY 75-51 (May 16,1975).

12. USNRC I.etter to G. C. Andognini, " Change to Bases," (June 6,1975).

13. A. S. DiGiovine, et al.; CASMO-3 Validation; YAEC-1363-A (April 1988).

14. A. A. F. Ansari, Methods for the Analysis of Boiline Water Reactors: Steady-State Core Flow Distribution Code (FIBWRL YAEC-1234 (December 1980).

l

15. A. A. F. Ansai, R. R. Gay, and B. J. Gitnick; FIBWR: A Steady-State Core Flow Distribution Code for Boiline Water Reactors - Code Verification and Oualification Report; EPRI NP-1923; Project 1754-1 Final Repon (July 1981).

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

17. GE Letter, J. L. Tuttle to R. T. Yee, "SLMCPR Calculation for Vermont Yankee Reload 18/ Cycle 19," JL'196035 (August 5,1996).

18. USNRC Letter to D. A. Reid, " Issuance of Amendment for Vermont Yankee Nuclear Power Station (TAC No. M96304)," NVY96-154 (October 4,1996).

19. General Electric Company, GEXL-Plus Correlation Aeolication to BWR 2-6 Reactors GE6 throuch GE9 Fuel. NEDE-31598P, GE Company Proprietary (July 1989).

20. GE Letter, J. L. Tuttle to R. T. Yee, " Revised Thermal-Mechanical MAPLHGR Limits for Vermont Yankee Reload 18/ Cycle 19," JL'I96044 (August 14, 1996).

21. GE Letter, D. T. Weiss to R. T. Yee, " Fuel . Rod Thermal-Mechanical Performance Limits,"

DTW92260 (November 19, 1992).

22. A. A. F. Ansari and J. T. Cronin, Methods for the Analysis of Boiline Water Reactors: A.

Systems Transient Analysis Model (RETR AN). YAEC-1233, (April 1981).

23. USNRC Letter to R. L. Smith, SER, " Amendment No. 70 to Facility License No. DPR-28,"

(November 27,1981).

24. V. Chandola, M. P.12Francois, and J. D. Robichaud; Apolication of One-Dimensional Kinetics to Boiline Water Reactor Transient Analysis Methods; YAEC-1693-A, Revision 1 (November 1989).

25. Electric Power Research Institute, RETRAN - A Procram for One-Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems. CCM-5 (December 1978).

26. USNRC Letter to T. W. Schnatz, SER, " Acceptance for Referencing of Licensing Topical Repons: EPRI CCM-5 and EPRI NP-1850-CCM," (September 4,1984).

27. A. A. F. Ansari, K. J. Bums, and D. K. Beller; Methods for the Analysis of Boiline Water Reactors: Transient Critical Power Ratio Analysis (RETRAN-TCPYA01): YAEC-1299P (March 1982).

28. J. T. Cronin, Method for Generation of One-Dimensional Kinetics Data for RETRAN-02.

YAEC-1694-A (June 1989).

29. General Electric Company; C. J. Paone, et al.; Rod Dron Accident Analysis for Laree Boiline Water Reactors: NEDO-10527 (March 1972).

30. General Electric Company; R. C. Stirn, et al.; Rod Dron Accident Analysis for Larce Boiline Water Reactors Addendum No.1. Multiple Enrichment Cores with Axial Gadolinium: NEDO-10527, Supplement 1 (July 1972).

31. General Electric Company; R. C. Stirn, et al.; Rod Dron Accident Analysis for I2rce Boiline Water Reactor Addendum No. 2 Exoosed Cores: NEDO-10527, Supplement 2 (January 1973).

32. General Electric Company, C. J. Paine, Banked Position Withdrawal Seauence. NEDO-21231 (January 1977).

33. General Electric Company, D. Radcliffe and R. E. Bates, Reduced Notch Worth Procedure. SIL-316 (November 1979).

34. M. A. Sironen, Vennont Yankee Cycle 14 Core Performance Analysis Reoort. YAEC-1706 (October 1988).

35. USNRC Letter to L. A. Tremblay, SER, " Safety Evaluation for Vermont Yankee Nuclear Power Station RELAPSYA LOCA Analysis Methodology (TAC No. M74595)," NVY 92-192 (October 21,1992).

36. USNRC Letter to R. W. Capstick, SER, " Approval of Use of Thermal-Hydraulic Code RELAPSYA (TAC No. 60193)," NVY 87-136 (August 25,1987).

37. L. Schor, et al.; Vermont Yankee Loss-of-Coolant A, -ident Analvsis: YAEC-1772 (June 1993).

38. Report, RELAPSYA. A Computer Procram for Licht-Water Reactor System Thermal-Hydrau_li;q Analysis. YAEC-1300P-A, Revision 0, October 1982; Revision 1 (July 1993).

39. M. A. Sironen, Vermont Yankee Cycle 18 Core Performance Analysis Report. YAEC-1908, Rev.1 (November 1995).

40. Vermont Yankee Nuclear Power Station Final Safety Analysis Report. Rev. 13,1995.

41. Report, General Electric Nuclear Energy, BWR Owners' Group Lone-Term Solutions Licensine Methodolocv, NEDO-31960-A and Supplement 1 (November 1995).

l l

42. N. Fujita, M. P. leFrancois and R. J. Weader, Method for Power / Flow Exclusion Recion Calculation Usine the LAPUR5 Computer Code. YAEC-1926 (September 1995).

43. Report, General Electric Nuclear Energy BWR Owners' Group Reactor Stability Detect and Suppress Solutions Licensine Basis Methodolocy and Reload Anolications. NEDO-32465 (May 1995).

44. Report, General Electric Nuclear Energy, Annlication of the "Recional Exclusion with Flow Biased Scram APRM Neutron Flux Scram" Stability Solution (Ontion 1-D) to the Vermont Yankee Nuclear Power Plant. GENE-637-018-0793 (July 1993).

45. VYNPC Letter to USNRC, " Cycle 18 Startup Test Report," BVY 95-81 (July 25,1995).

l l

l 1

APPENDIX A CALCULATED OPERATING LIMITS The MCPR operating limits for the Reload Cycle are calculated by adding 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 en 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. Rese MCPR operating limits are valid for operation of the Reload Cycle at full power up to 10701 mwd /St and for operation during coastdown beyond EOFPL.

Tables A.2 through A.4 provide the most limiting calculated MAPLHGR limits for all the fuel types in the Reload Cycle. These values bound the lattice-specific MAPLHGR limits for all the enriched lattice zones in each fuel type. He MAPLHGR limits were revised for the LOSC transient results.

Table A.5 provides the actual power and flow values used in the generation of the stability exclusion and buffer regions. Figure A.I provides the stability exclusion and buffer region curves.

TAllLE A.1 VERMONT YANKEE NUCLEAR POWER STATION CYCLE 19 MCPR OPERATING LIMITS Value of"N" in Averace Control Rod MCPR.

RBM Eauntion' Scram Time Cycle Exoosure Rance Ooeratine Limit 25 Equal to or 0.0 to 9701 mwd /St 1.32 9701 to 10701 mwd /St 1.34 L 0 3 .C.I.1 Equal to or 0.0 to 9701 mwd /St 1.32 9701 to 10701 mwd /St 1.41 L O 3 .C.I.2 Equal to or 0.0 to 9701 mwd /St 1.29 9701 to 10701 mwd /St 1.34 O 3 .C.1.1 Equal to or 0.0 to 8701 mwd /St 1.29 better than 8701 to 9701 mwd /St 1.32 L.C.O. 3.3.C. I .2 9701 to 10701 mwd /St 1.41 Equal to or 0.0 to 9701 mwd /St 1.29 9701 to 10701 mwd /St 1.34 O 3 .C.l.1 Equal to or 0.0 to 8701 mwd /St 1.29 better than 8701 to 9701 mwd /St 1.32 L.C.O. 3.3.C.I .2 9701 to 10701 mwd /St 1.41 (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 should be increased by 0.02 for the single loop operation.

TABLE A.2 MAPLHOR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-10GZ Plant: Vermont Yankee Fuel Type: BP8DWB335-10GZ Averace Planar Exoosure M APLHGR Limits (kW/ft)

(mwd /St) Two-Loon Operation Sincle-Loon Operation

• 0.00 11.29 9.37 200.00 11.34 9.41 1,000.00 11.48 9.53 2,000.00 11.69 9.70 3,000.00 11.92 9.89 4,000.00 12.17 10.10 5,000.00 12.43 10.32 6,000.00 12.68 10.52 7,000.00 12.87 10.68 8,000.00 13.06 10.84 '

9,000,00 13.20 10.%

10,000.00 12.79 10.62 12,500.00 12.65 10.50 15,000.00 12.47 10.35 20,000.00 11.76 9.76 25,000.00 11.09 9.20 35,000.00 '9.88 8.20 45,000.00 8.38 6.96 50,590.00 5.65 4.69 o

MAPLHGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLHGR limits by 0.83.

TABLE A.3 MAPLHGR VERSUS AVEPAGE PLANAR EXPOSURE FOR BP8DWB335-11GZ Plant: Vermont Yankee Fuel Type: BP8DWB335-11GZ Averace Planar Exposure MAPLHGR Limits OcW/ft)

Two-Loco Ooeration Sincle-Looo Operation

• afWd/St) 0.00 11.28 9.36 200.00 11.33 9.40 1,000.00 11.43 9.49 2,000.00 11.60 9.63 3,000.00 11.80 9.80 4,000.00 12.04 9.99 5,000.00 12.30 10.21 6,000.00 12.53 10.40 7,000.00 12.73 10.57 8,000.00 12.94 10.74 9,000.00 13.12 10.89 10,000.00 12.79 10.62 12,500.00 12.65 10.50 15,000.00 12.47 10.35 20,000.00 11.76 9.76 25,000.00 11.09 9.20 35,000.00 9.88 8.20 45,000.00 8.38 6.96 50,590.00 5.65 4.69

• MAPLHGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLHGR limits by 0.83.

TABLE A.4 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB354-12GZ Plant: Vermont Yankee Fuel Type: BP8DWB354-12GZ Averace Planar Exposure MAPLHOR Limits 0:W/ft)

(mwd /St) Two-Loon Operation Sincle-Looo Operation

• 0.00 10.96 9.10 200.00 '11.04 9.16 1,000.00 11.18 9.28 2,000.00 11.40 9.46 3,000.00 11.63 9.65 4,000.00 11.81 9.80 5,000.00 12.01 9.97 6,000.00 12.14 10.08 7,000.00 12.26 10.18 8,000.00 12.37 10.27 9,000.00 12.46 10.34 10,000.00 12.52 10.39 12,500.00 12.40 10.29 15,000.00 12.10 10.04 20,000.00 11.40 9.46 25,000.00 10.72 8.90 35,000.00 9.44 7.84 45,000.00 7.24 6.01 48,200.00 5.67 4.71 MAPLHGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLHGR limits by 0.83.

TABLE A.5 VERMONT YANKEE NUCLEAR POWER STATION CYCLE 19 STABILITY EXCLUSION AND BUFFER REGIONS Exclusion Region Buffer Region Flow Power Flow Power 27.000 34.2602 27.8420 29.4008 28.000 34.5125 29.5823 29.8049 29.000 34.9275 31.2232 30.4723 30.000 35.5053 32.7519 31.3477 31.000 36.2458 34.1750 32.3963 >

32.000 37.1492 35.5083 33.5970 33.000 38.2154 36.7696 34.9390 34.000 39.4443 37.9747 36.4180 35.000 40.8360 39.1369 38.0338 36.000 42.3905 40.2661 39.7879 37.480 44.9900 41.8927 42.6430 38.000 45.9879 42.4546 43.7209 39.000 48.0308 43.5238 45.9047 40.000 50.2364 44.5812 48.2365 41.000 52.6049 45.6292 50.7181 41.970 55.0578 46.6385 53.2701 43.000 57.8301 47.7040 56.1376 44.000 60.6869 48.7334 59.0781 45.000 63.7065 49.7587 62.1739 ,

46.000 66.8889 50.7806 65.4260 47.000 70.2340 51.7998 68.8350 48.000 73.7420 52.8166 72.4017 ,

49.700 80.0791 54.5407 78.8284 50.000 81.2461 54.8445 80.0102 51.000 85.2425 55.8561 84.0529 ,

51.660 88.0000 56.8665 88.2552 57.7330 92.0000 57.8758 92.6173

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-100-