ML20096E291

VT Yankee Cycle 18 Core Performance Analysis Rept

ML20096E291
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/29/1995
From:	Paul Bergeron, Cacciapouti R, Sironen M VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20096E279	List: ML20096E277 ML20096E291
References
YAEC-1908, YAEC-1908-R01, YAEC-1908-R1, NUDOCS 9601220107
Download: ML20096E291 (107)

Text

. . . - - - - ... . . . . . . . -- . . . . . .. .

lI e

a:

I b

$ i Vermont Yankee I Cycle 18 Core Perfonnance Analysis Report

!I

!I

'I i

' November 1995 Major Cbntributors: C. Oilu

N. Fujita B. Hubbard

.l

'E D. Kapitz M. LeFrancois D. Morin {

g K. Morrissey l

J. Neyman l i}3 L. Schor

g F.Seiface l jE R. Smith 4

K. StJohn 4

R. Wochlke l JI i

i

.' **n 1

9601220107 960116 l l PDR ADOCK 05000271

P PDR

h. .

I Prepared by: N AOW I A / /d M. A. Sk6nen ' (Date)

VY Nuclear Engineering Coordinator Approved by: . 42 //M '//W

. Cacciapouti[ anager '(Date) g ctor Physics roup g Approved by: -7 0" 9

anager (Date)

P. A. Bergeron,[is Transient Analys Group Appmved by: W" b b!29!@C '

R. K. Sundaram, Man /ger (Date)

LOCA Analysis Group Approved by: d[46 '

d /' # f f~

Nu Department I

i .

I

! l I

g Yankee Atomic Electric Company Nuclear Services Division B 580 Main Street g Bolton, Massachusetts 01740 R4(ND I

I

DISCLAIMER OF RESPONSIBILITY

'Ihis 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 respect to any unauthorized g, 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 cf the material contained in this document.

1 l

i a4M3 -Ill.

i

I ABSTRACT 4

This report presents design infonnation, calculational results, and operating i t .ts pertinent to the operation of Cycle 18 of the Vermont Yankee Nuclear Power Station. 'Ihese include the fuel design and core loading pattem descriptions; calculated reactor power distributions, exposure g distributions, shutdown capability, and reactivity data; and the results of safety analyses performed to 3 justify plant operation thmughout the cycle.

! g

'Ihis report was revised to incorporate the loss of stator cooling transient analysis description T t.nd results. The revised MAPLHOR limits are included.

I l

I-

I

I, I

E I

I I

t I,

l u 1

ACKNOWLEDGEMENTS The author and major contdbutors would like to acknowledge the contributions to this woi by the Vennont Yankee Reactor & Computer Engineering Department for their review ofinput data and guidance.

i 5

j r

l t

IWIMS -V-i

e TABLE OF CONTENTS fiUL DISCLAIMER OF RESPONSIBILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111 AB STRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv ACKNOWLEDGEMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . VI LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vili LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 RECENT REACTOR OPERATING HISTORY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1 Operating History of the Current Cycle ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.2 Operating History of Past Applicable Cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.0 RELOAD CORE DESIGN DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.1 Core Fuel Loading . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.2 Design Reference Cort Loading Pattern . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.3 Assembly Exposure Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

. 4.0 FUEL MECHANICAL AND THERMAL DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.1 Mechanica1 Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.2 'Ihe nn al Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 -

4.3 Operating Experience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 .

! 5.0 NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

5.1 Cort Power Distributions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.1.1 Haling Power Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.1.2 Rodded Depletion Power Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . 15 l 5.?. Core Exposure Distributions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 5.3 Cold Shutdown Margin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 5.4 Maximum K , for the Spent Fuel Pool . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 6.0 THERMAL-HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 mess -vi-I I

1 i

i i

l TABLE OF CONTENTS (Continued)

East 6.1 Steady-State 7hermal Hydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26-6.2 Reactor Limits Determination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 7.0 ABNORMAL OPERATIONAL 'IRANSIENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . 28 7.1 Transients Analyzed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 l 7.2 Pressurization Transients Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.2.1 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 1 7.2.2 Initial Conditions and Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 l 7.2.3 One. Dimensional Cross Sections and Kinetics Parameters . . . . . . . . . . . 31 '

l 7.2.4 Turbine Trip Without Bypass Transient (7TWOBP) . . . . . . . . . . . . . . . . 33 7.2.5 Generator Load Rejection Without Bypass Transient (GLRWOBP) . . . . . 33 'i 7.2.6 Pressurization Transient Analysis Results . . . . . . . . . . . . . . . . . . . . . . . 33 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 7.5 local Rod Withdrawal Error Transient Results . . . . . . . . . . . . . . . . . . . . . . . . . 36 7.6 Misloaded Bundle Enor Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39  !

7.6.1 Rotated Bundle Enor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 7.6.2 Mislocated Bundle Error . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 7.7 Transient Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 8.0 DESIGN BASIS ACCIDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 8.1 Control Rod Drop Accident Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 8.2 Loss.of-Coolant Accident Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 78 8.3 Refueling Accident Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 79 1

9.0 STARTUP PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 85

10.0 CONCLUSION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86  !

APPENDIX A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 1

i i mass -YII-8 6

t

---____m_____

I LIST OF TABLES Number Title ,Page a

2.1.1 VY CYCLE 17 OPERATING HIGHLIGHTS . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2.1 VY CYCLE 16 OPERATING HIGHLIGHTS .......................... 4 3.1.1 ASSUMED VY CYCLE 18 FUEL BUNDLE TYPES AND NUMBEPS . . . . . . . . 7 l

3.3.1 DESIGN BASIS VY CYCLE 17 AND CYCLE 18 EXPOSURES . . . . . . . . . . . . 7 l 4.1.1 NOMINAL FUEL MECHANICAL DESIGN PARAMETERS . . . . . . . . . . . . . . 11 4.2.1 VY CYCLE 18 CORE AVERAGE GAP CONDUCTANCE VALUES . . . . . . . . 12 5 4.2.2 VY CYCLE 18 HOT CHANNEL GAP CONDUCTANCE VALUES FOR HALING E AXIAL POWER DISTRIBUTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 5 4.2.3 VY CYCLE 18 HOT CHANNEL GAP CONDUCTANCE VALUES FOR 1.4 CHOPPED COSINE AXIAL POWER DISTRIBUTION . . . . . . . . . . . . . . . . . . 14 VY CYCLE 18 Kp,pp VALUES AND SHUTDOWN MARGIN CALCULATION 18 5.3.1 5.4.1 VY CYCLE 18 MAXIMUM COLD K., OF ANY ENRICHED SEGMENT . . . . 18 7.2.1 VY CYCLE 18

SUMMARY

OF SYSTEM TRANSIENT MODEL INITIAL CONDITIONS FOR TRANSIENT ANALYSES . . . . . . . . . . . . . . . . . . . . . . . . 42 7.2.2 VY CYCLE 18 PRESSURIZATION TRANSIENT ANALYSIS RESULTS . . . . . 43 7.3.1 VY CYCLE 18 LOSS OF FEEDWATER HEATER TRANSIENT RESULTS . . . 44 7.3.2 VY CYCLE 18 LOSS OF STATOR COOLING TRANSIENT RESULTS . . . . . . 44 7.4.1 VY CYCLE 18 OVERPRESSURIZATION ANALYSIS RESULTS . . . . . . . . . . 45 7.5.1 VY CYCLE 18 ROD WITHDRAWAL ERROR ANALYSIS RESULTS . . . . . . . 45 7.6.1 VY CYCLE 18 ROTA7ED BUNDLE ANALYSIS RESULTS . . . . . . . . . . . . . . 46 7.6.2 VY CYCLE 18 MISLOCATED BUNDLE ANALYSIS RESULTS . . . . . . . . . . . 46 7.7.1 VY CYCLE 18 LIMITING TRANSIENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 8.1.1 CONTROL ROD DROP ANALYSIS - ROD ARRAY PULL ORDER , . . . . . . . 80 8.1.2 VY CYCLE 18 CONTROL ROD DROP ANALYSIS RESULTS . . . . . . . . . . . . 80 mess -vili-I Il

f LIST OF TABLES (Continued)

Number 3 Eggg, 8.2.1 LOCA ANALYSIS ASSUMF1' IONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 A.1 ' VERMONT YANKEE NUCLEAR POWER STATION CYCLE 18 MCPR OPERATING LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 l

A.2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB311-10GZ92

A.3 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BF8DWB311-11GZ93 A.4 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-10GZ94 A.5 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-11GZ95 1

4 9

1 I

i i

l t

I i

e 1

i m .ix.

i LIST OF FIGURES

Number .Iblg Egge 3.2.1 VY CYCLE 18 DESIGN REFERENCE LOADING PATTERN, LOWER RIGHT Q UADRANT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 5.1.1 VY CYCLE 18 HALING DEPLETION, EOFPL BUNDLE AVERAGE RELATIVE POWERS..............'..................................... 19 I,

5.1.2 VY CYCLE 18 HALING DEPLLTION EOFPL CORE AVERAGE AXIAL POWER 4 DISTRIBUTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

5.1.3 VY CYCLE 18 RODDED DEPLETION - ARO AT EOFPL, BUNDLE AVERAGE RELATIVE POWERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.1.4 VY CYCLE 18 RODDED DEPLETION - ARO AT EOFPL, CORE AVERAGE 4

AXI AL POWER DISTRIBUTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 5.2.1 VY CYCLE 18 HALING DEPLETION, EOFPL BUNDLE AVERAGE EXPOSURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 i 5.2.2 VY CYCLE 18 RODDED DEPLETION, EOFPL BUNDLE AVERAGE EXPO SURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 5.3.1 VY CYCLE 18 COLD SHUTDOWN MARGIN,IN %AK, VERSUS CYCLE EXPO S URE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 7.2.1 FLOW CHART FOR THE CALCULATION OF ACPR USING THE

, RETRAN/TCPYA01 CODES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 7.2.2 TURBINE TRIP WITHOUT BYPASS, EOFPL18 TRANSIENT RESPONSE VERSUS l TIME, " MEASURED" SCRAM TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 3 7.2.3 TURBINE TRIP WITHOUT BYPASS, EOFPL18-1000 MWD /ST TRANSIENT E RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . . . . . . . . . . . 52 5 7.2.4 TURBINE TRIP WITHOUT BYPASS, EOFPL18 2000 MWD /ST TRANSIENT g RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . . . . . . . . . . . 55 g 7.2.5 GENERATOR LOAD REJECTION WITHOUT BYPASS, EOFPL18 TRANSIENT g RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME . . . . . . . . . . . . . . 58 5 7.2.6 GENERATOR LOAD REJECTION WITHOUT BYPASS, EOFPL18-1000 MWD /ST I

TRANSIENT RESPONSE VERSUS TIME, MEASURED" SCRAM TIME . . . . 61 I

I I

l l

l l

LIST OF FIGURES 1 l (Condnued) i  ;

Nmnber Tult East 7.2.7 GENERATOR LOAD REJECTION WITHOUT BYPASS, EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME , . . . 64 7.3.1 LOSS OF 100*F FEEDWATER HEATER (LIMITINO CASE) TRANSIENT RESPONSE VERSUS TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 7.3.2 LOSS OF STATOR COOLINO TRANSIENT RESPONSE VERSUS TIME . . . . 69 7.4.1 MSIV CLOSURE, PLUX SCRAM, EOFPL18 TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 7.5.1 VY CYCLE 18 RWE CASE 1 - SETPOINT INTERCEPTS DETERMINED BY THE A AND C CHANNELS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 7.5.2 VY CYCLE 18 RWE CASE 1 - SETPOINT INTERCEPTS DETERMINED BY THE B AND D CHANNELS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75 8.1.1 FIRST FOUR ROD ARRAYS PULLED IN THE A SEQUENCES . . . . . . . . . . 82 8.1.2 FIRST FOUR ROD ARRAYS PULLED IN THE B SEQUENCES . . . . . . . . . . . 83 8.2.1 LOCA ANALYSIS RESULTS, PEAK CLADDING TEMPERATURE VERSUS B REAK SIZE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84 1 1

R4tMS -XI-

l

1.0 INTRODUCTION

l This report provides information to support the operation of the Vermont Yankee Nuclear g Power Station through the forthcoming Cycle 18. In this report, Cycle 18 will be referred to as the N Reload Cycle. 'Ihe preceding Cycle 17 will be referred to as the Current Cycle. The Cycle 17/18 g

refueling will involve the discharge of 120 Irradiated fuel bundles and the insertion of 120 new fuel g bundles. 'Ihe resultant core will consist of 120 new fuel bundles and 248 irradiated fuel bundles. The General Electric Company (GE) manufactured all the bundles. Some of the irradiated fuel was also present in the reactor in Cycle 16. This cycle will be refened to as the Past Cycle.

This report contains descriptions and analyses results pertaining to the merhanien1; thermal-hydraulic, physics, and safety aspects of the Reload Cycle. The MAPLHGR and MCPR operating limits calculated for the Reload Cycle are given in Appendix A. 'Ihese limits will be included in the Core Operating Limits Report.

This report was revised to incorporate the loss of stator cooling transient analysis description and results. 'Ihe MAPLHGR operating limits in Appendix A were revised as a result of this transient.

I I

I i

E l

I a- .,.

E I

I

^

l 2.0 RECENT REACTOR OPERATING HISTORY i

i 2.1 Operating History of the Current Cycle l

'Ihe current operating cycle is Cycle 17. To date, the Cunent Cycle has been operating at, or l

near, full power with the exception of sequence exchanges, several power reductions, and four shon repair outages. 'Ihe operating history highlights and control rod sequence exchange schedule of the Cunent Cycle are found in Table 2.1.1. )

1 l

2.2 Operating History of Past Apolicable Cycle

'Ihe irradiated fuel in the Reload Cycle includes some fuel bundles initially insened in Cycle j

16. This Past Cycle operated at, or near, full power with the exception of sequence exchanges, several I short power reductions, one short repair outage and a coastdown to the end of cycle. The operating history highlights of the Past Cycle are found in Table 2.2.1. The Past Cycle is described in detail in the Cycle 16 Summary Repon[1].

1 e

maus l 1

I TABLE 2.1.1 VY CYCLE 17 OPERATING HIGHLIGITTS October 24,1993 Beginning of Cycle Date g End of Cycle Date March 18,1995* 5 Weight of Uranium As-Imaded (Short Tons) 72.02 l Beginning of Cycle Core Average Exposure ** (mwd /St) 11547 End of Full Power Core Average Exposure ** (mwd /St) 21997*

End of Cycle Core Average Exposure ** (mwd /St) 21997*

Number of Fmsh Assemblies 128 Number of Irradiated Assemblies 240 Control Rod Sequence Exchange Schedule:

Sequence Date From Toq 4 January 9,1994 A2-1 B2-1

$ March 15,1994 B2-1 Al 1 May 17,1994 Al-1 B1-1 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

I I

Projected dates and exrnsures.

Exposures based on the Plant Process Computer accounting.

E men I

I

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

a TABLE 2.2.1 VY CYN 16 OPERATING HIGHLIGHTS l

1 Beginning of Cycle Date April 19,1992 r End of Cycle Date August 28,1993 Weight of Uranium As-Imaded (Short Tons) 72.06 Beginning of Cycle Core Average Exposure * (mwd /St) 11417 End of Full Power Core Average Exposure * (mwd /St) 21103 End of Cycle Core Average Exposure * (mwd /St) 21878 Number of Fresh Assemblies 128 Number ofIrradiated Assemblies 240 Contml Rod Sequence Exchange Schedule:

Sequence M Fmm Tg June 14,1992 A2-1 B2-1 August 9,1992 B2-1 Al-1 i October 15,1992 Al-1 B1-1 ,

December 7,1992 B1-1 A2-2  ;

February 9,1993 A2-2 B2-2 April 6,1993 B2-2 Al-2 June 6,1993 Al-2 B1-2 O Exposures based on the Plant Process Computer accounting.

men .-.

E1 i

3.0 RELOAD CORE DESIGN DESCRWTION l i

Com Fuel Loading 3.1 i

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 fuct is found in the GE Standard Application for Reactor Fuel [2] and the GE Fuel Bundle Design Reports [3][4]. i 3.2 Design Refemnce Core Loading Pattem

'Ih Reload Cycle assembly locations are indicated on the map in Figure 3.2.1. For the sake of legibility only the lower right quadrant is shown. The other quadrants are mirror images with bundles of the same type having nearly identical exposures. The bundles are identified by the reload number in which they were first introduced into the core. 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. The final loading pattem with specific fuel bundle serial numbers will be supplied in the Startup Test Report.

3.3 Assemb1v Exoosure Distribution E

The assumed nominal exposure on the fuel bundles in the Reload Cycle design reference loading pattem is given in Figure 3.2.1. To obtain this exposure distribution, the Past Cycle was 3 depleted with the SIMULATE-3 model[5],[6] using actual plant operating history. For the Current Cycle, plant openting history was used through April 22,1994. Beyond this date, the exposure was l accumulated using a best-estimate rodded depletion analysis to End of Cycle (EOC).

I I

g men I' I.

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

d 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 pattem.

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

men -6

I TABLE 3.1.1 E g

ASSUMED VY CYr1 F 18 FUEL BUNDLE TYPES AND NUMBERS Fuel Type Reload Cycle Number

' Designation Bundle ID IAaded of Bundles Irradiated BP8DWB311-10GZ RISA 16 40 BPSDWB311-110Z RISB 16 80 BP8DWB335-10GZ R16A 17 96 BP8DWB335-11GZ R16B 17 32 New BP8DWB335-10GZ R17A 18 88 BP8DWB335-110Z R17B 18 32

! I s TABLE 3.3.1 DESIGN BASIS VY CYCLE 17 AND CYCLE 18 EXPOSURES

• 4 Assumed End of Current Cycle Core Average Exposure wit an 21.92i.6 GWd/St Exposum Window ofi 600 mwd /St[7]

j Assumed Beginning of Reload Cycle Core Average Exposure 12.13 GWd/St I

Haling Calculated End of Full Power Life Reload Cycle Core Average 22.16 GWd/St g Exposum 3 Reload Cycle Full Power Exposure Capability (Haling) 10.035 GWd/St

~

I Exposures based on the SIMULATE-3 accounting.

mess ,

I I

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

l i

i 4

R185 R17A R100 R14A R16A R17A R16A R175 R16A R14A RISO 30 064 0 000 21 A00 11.381 21.794 OA00 21433 0 000 33.110 13 4:2 24AaB R17A R14A R17A R14A R17A AleA R17A R175 RieA Ries A16A 30 SADO 11 A48 OA00 13.738 0.000 12.475 OA00 OA00 12.730 13A02 34A00 i R198 R17A Ries R17A Rigg R17A Rim R175 R14A Ales R169 18 2125 SA00 21A01 SADO 21 A81 SA00 N.387 4A00 13A51 13A03 38 048 R14A R18A R17A RieA R17A AleA R17A R17A RieA R185 18 11A87 11.738 OA00 11 ABO 0.000 12.701 OA00 SADO 13.198 34.104 J R168 R17A R188 R17A Ries R17A R1ES R14A Ries 3

21A01 OA00 21 408 0 000 21 A03 OA00 33A31 13178 13 842

R17A R14A R17A RieA - R17A AleA R175 R145 R16A 12 0000 12A10 Om 13A10 0 410 1928 OA00 13 m 34.1H A16A R17A Ries R17A RIGA R175 R14A RISB R1ES 2

10

. 21.330 SADO 22.448 OA00 St.the 0.000 13434 13A79 N.130 R175 R175 R178 R17A RIGA AleB R184 R10B na CADO OA00 0A00 CA00 13.107 13 873 13.737 N 000 R16A R14A R14A R14A R14A R16A ASES na 33J04 12 000 13.886 13A13 13.006 23.387 NA00 R14A Alet A145 R185 64 13 448 13.077 13 Jet 34.136 R16A R165 Atas . BUIOLE C ns 34 442 24 ses as.s7e _ soc EXPoeURE phDST)

SS N 27 38 31 M 36 37 N 41 43 FIGURE 3.2.1 VY CYCLE 18 DESIGN REFERENCE LOADING PATTERN. LOWER RIGHT OUADRANT nea -8

4.0 FUEL MECHANICAL AND THERMAL DESIGN 4.1 Mechanical Design All of the fuel to be inserted into the Reload Cycle was fabricated by GE. The major mechanical design parameters art given in Table 4.1.1 and Reference 2. Detailed descriptions of the w fuel rod mechanical design and mechanical design analyses are provided in Reference 2. Dese 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 Thennal Design The fuel thermal effects calculations were performed using the FROSSTEY-2 computer code [8],[9],[10]. De FROSSTEY-2 code calculates pellet-to-cladding gap conductance and fuel temperatures from a combination of theoretical and empirical models including but not limited to fuel and cladding thennal 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. Dese gap conductance values are weighted axially into 12 axial nodes by power distributions and radially by volume. De core-wide gap conductance values for the RETRAN system simulations, described in Sections 7.1 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.

De 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.

De hot channel calculations assumed the following as required by the NRC Safety Evaluation for ams I I

FROSSTEY-2[ll): 1) appropriate allowances to account for manufacturing uncertainties and 2) the

! worst axial power shape prior to the transient. 'Ihe peak power node was placed at the maximum average planar linear heat generation rate (MAPLHOR) limits. Gap conductance values for the hot channel analysis were determined using the limiting bundle exposure. 'Ihe limiting bundle is defined as the bundle with the lowest MCPR or the highest power, if different, within the exposure range of interest. 'Ihe limiting exposure for the bundle is defined by the exposure which produces the highest 4

bundle average gap conductance within the interval of interest. 'Ihe SIMULA*IE-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.

i 4.3 Operating Experience 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 i fuel rod failure.

i i I l

J A

o i

1 men .

TABLE 4.1.1 NOMINAL FUEL MECHANICAL DESIGN PAR AMETERS l

i l

Fuel Bundle

• Irradiated Fuel Tvoe New & Irradiated Fuel Tvoes i Bundle Types GE8X8NB GE8X8NB

! Vendor Designation BP8DWB311-10GZ & BP8DWB33510GZ &

I BP8DWB311-110Z BP8DWB335-11GZ Initial Endchment,w/o U235 3.11 3.35 Rod Array 8X8 8X8 Fuel Rods per Bundle 60 60 Outer Fuel 01annel I

Material Zr-2 Zr-2 Wammss. - o.080 0.080 g I

I I

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

aan I I

TABLE 4.2.1 VY CYOP 18 CORE AVERAGE GAP CONDUCTANCE VALUES .

2 Gap Conductance (BTU /hr-ft ,.p)

Axial 1Q, Q EOFPL-2000 EOFPL-1000 EOFPL 1 ligds, mwd /St mwd /St 1 1190 1830 1960 2115 2 2345 3600 3715 3840 l 3 2445 3810 3875 4140 ,

\

4 2455 3820 3895 4185 5 2495 3860 3955 4325 6 2610 3945 4150 4560 7 2600 3940 4140 4555 8 2625 3960 4180 4565 9 2525 3880 4005 4455 1

10 2420 3760 3850 4080 11 1880 2860 2990 3125 12 675 1015 1120 1225 L

nes . l

TABLE 4.2.2 l VY CYCLE 18 HOT CHANNEL GAP CONDUCTANCE VALUES

• Wr l

FOR HALING AXIAL POWER DISTRIBUTION 2

Gap Conductance (BTU /hr ft ,.p) ,j g

Axial },QQ" EOFPL-2000 EOFPL-1000 EOFPL" E2d.9. mwd /St mwd /St 10.95 GWd/St 10.016 GWd/St 10.95 GWd/St 11.472 GWd/St 1 3360 3040 3360 3590 2 7850 7520 7850 8330 3 9630 9530 9630 9500 4 9630 9780 9630 9500 5 9630 9880 9630 9500 6 9650 9890 9650 9500 7 9650 9890 9650 9500 8 9650 9890 9650 9500 9 9650 9890 9650 9500 10 9450 8150 9450 9500 11 6580 6130 6580 6930 12 1530 1460 1530 1570 I

I

'Ihe hot channel gap conductance values are derived for the BP8DWB335 fuel type because it is conservative compared to the other fuct types.

• • Core Average Exposure.

• • • Peak Bundle Exposure.

m. I-I'

l 1

TABLE 4.2.3 I

VY CYCLE 18 HOT CHANNEL GAP CONDUCTANCE VALUES

• FOR 1.4 CHOPPED COSINE AXTAL POWER DISTRIBUTION l i

l i

2 Gap Conductance (BTU /hr ft ,.p)  !

Axial E.QQ" EOFPL-2000 EOFPL-1000 EOFPL" u.gr uwd/st" Mwa/st"  ;

12.15 GWd/St"* 10.008 GWd/St"* 11.472 GWd/St " 12.15 GWd/St"*

1 750 790 760 750 i l

2 1510 1410 1480 1510 l l

3 5040 3690 4680 5040 4 7920 8000 7500 7920 ,

4 l

5 9730 10450 9970 7930 6 9780 10450 10020 9780  ;

7 9810 10450 10020 9810 l l

I 10020 9810 8 9810 10450

. 9 9810 '8570 9150 9810

' 10 7230 6210 7070 7230

, 11 2640 2260 2520 2640 12 950 960 950 950

'Ihe hot channel gap conductance values are derived for the BP8DWB335 fuel type because it is conservative compared to the other fuel types.

I

• • Core Average Exposure.  !

. 1 C" Peak Bundle Exposure, men  !

l

I 5.0 NUCLEAR DESIGN 5.1 Core Power Distributions The Reload Cycle was depleted using SIMULATE-3 to give both a rodded depletion and an All Rods Out (ARO) Haling depletion. W 1

5.1.1 Haling Power Distribution

The Haling depletion serves as the basis for defining core reactivity characteristics for most transient evaluations. This 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.

1 'Ihe Haling power distribution is calculated in the ARO condition. The Haling iteration I

converges on a self consistent power and exposure distribution for the burnup step to EOFPL. In principic, this should provide the overall minimum peaking power shape for the cycle. Dudng the actual cycle, flatter power distributions might occasionally be achieved by shaping with contml rods.

However, such shaping would leave underbumed regions in t!e core which would peak at anotler 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 Rodded Depletion Power Distribution The rodded depletion was used to evaluate the mistoaded bundle error and the rod withdrawal enor because it provides the initial rod pattems and more accurately defm' es the local characteristics g

prior to the transient evaluations. It was also used in the rod drop worth and shutdown margin 5 calculations because it depletes the top of the core more realistically than the Haling depletion. The rodded depletion also provides the hot channel bundle exposures for the gap conductance calculation.

l 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 non I Il l

l .

j patterns were frequently more peaked than the Haling, but were below expected operating limits.

l However, as stated above, the underbumed regions of the core can exhibit peaking in excess of the Haling peaking when pulling ARO at EOPPL. 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 Exposure Distributions The Reload Cycle exposures are summarized in Table 3.3.1. 'Ihe 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. BOC, EOFPL-2000 mwd /St EOFPL-1000 mwd /St, and EOFPL exposure distributions were

! used to develop reactivity input for the core wide transient analyses.

1 i

'Ihe rodded depletion differs from the Haling during the cycle because the rods shape the l power differently. However, rod sequences are swapped frequently and the overall exposure I

distribution at end of cycle is similar to the Haling. Figure 5.2.2 gives the EOFPL radial exposure j distribution for the Reload Cycle rodded depletion.

l 5.3 Cold Shutdown Marrin

Technical Specifications [12] state that, for sufficient shutdown margin (SDM), the core must f be subcritical by at least 0.25% AK + R (defined below) with the strongest worth contml rod

withdrawn. Using SIMULATE-3, a search was made for the stmngest worth contml rod at various l exposures in the cycle. 'Ihis is necessary because rod worths change with exposure on adjacent i assemblies. Then the cold K,g with the stmngest rod out was calculated at BOC and at the end of j each control rod sequence. Subtracting each cold K,g with the strongest rod out from the cold critical K,g defines the SDM as a function of exposure. Figure 5.3.1 shows the results.

The cold critical K,g was defined as the average calculated critical K,g minus a 95%

, confidence level uncertainty. 'Ihen all cold results were normalized to make the critical K,g equal to i 1.000.

me,n f

I Because the local reactivity may increase with expositre, the SDM may decrease. To account for this and other uncertainties, the value R is calculated. R is dermed as3 R plus 2R . R3 is the difference between the cold K,g with the stmngest rod out at BOC and the maximum cold Ke n with the strongest rod out in the cycle. R 2is a measurement uncertainty in the demonstration of SDM l l

associated with the manufacture of past control blades. It is presently set at 0.07% AK[13],[14]. 'Ihe shutdown margin results, summarized in Table 5.3.1, show that the shutdown margin for the Reload Cycle is greater than the Technical Specification limit of 0.32% AK.

5.4 Maximum K_ for the Soent Fuel Pool Section 5.5E of the Technical Specifications requires that the K,, for any burxile stored in I

either the new fuel vault or the spent fuel pool not exceed 1.31 to ensure compliance with the K,g 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. 'Ihese values are obtained from CASMO-3G[15].

I I

! I I

~

I

I.

I I

man I I

I

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

TABLE 5.3.1 VY CYCLE 18 K_ opp VALUES AND SHUTDOWN MARGIN CALCULATION Cold Critical K,g 1.0000 BOC K,g - Controlled With Strongest Wonh Rod Wiihdrawn 0.9872

. Cycle Minimum Shutdown Margin Occurs at BOC With  !

Strongest Worth Rod Withdrawn 1.28% AK l

R 3, Maximum Increase in Cold K,g With Exposure 0.00% AK TABLE 5.4.1 VY CYCLE 18 MAXIMUM COLD K_ OF ANY ENRICHED SEGMENT Bundle Tvoe Maximum K_

BP8DWB311-10GZ 1.20 BP8DWB311-11GZ 1.20 BP8DWB335-10GZ 1.22 j BP8DWB335-11GZ 1.22 I

men .

i i

I I

R158 R17A R158 R14A R15A R17A R15A R178 R15A R14A R158 I

1870 1203 1 A01 1218 1 A06 1384 1D42 1243 0434 0.738 0.306 R17A R16A R17A R14A R17A R14A R17A R178 RIGA R188 R16A

' IJe3 1206 1 A10 1J51 1.302 1230 1340 1240 0364 0.732 0378 R158 R17A R188 R17A R158 R17A R188 R178 R14A R148 R168 1DB3 1A10 1.121 1 A30 1.108 1.375 1D3B 1220 0.913 0.083 0.342 Ri4A R14A R17A R14A R17A R16A R17A R17A RIM R158 18 12'8 1251 1 A30 1271 1J94 1.194 1280 1.162 CA27 0 403 R158 R17A R158 R17A R158 R17A R158 R10A R188 1 064 1J01 1.107 1303 1.070 1204 0336 Cats OA04 R17A R14A R17A R14A R17A R14A R178 R188 R15A 1203 1 210 1274 1.192 1.304 1D06 1.116 0.791 0 470 R15A R17A R158 R17A R15A R178 R14A R188 R158 1D44 1230 1 A07 1 279 0.938 1.116 OA83 0.868 0.366 R178 R178 R178 R17A R14A R188 R1M R158 na 1242 1244 1226 1.161 0 918 0.792 0 867 0J64 RISA R14A RIGA RisA RIGA R15A R158 na 0 830 0 983 0 211 OA27 0 000 0 471 0.384 3

R16A R168 R148 R158 0.738 0 730 0.002 0.487 R15A R168 R188 ~ 8UNDLE D OJe8 0.375 0342 _ EOFPL RELATNE POWER l l

1 23 36 27 as 31 33 36 37 39 41 43 i

FIGURE 5.1.1 VY CYCLE 18 HALING DEPLETION.

EOFPL BUNDLE AVERAGE RELATIVE POWERS a*m I I

l -

M N  %

i 24 1  %

as  %

n 21 N

is i 10 i l 17 i 18 l l

16 11 N ,

is il I e tt I 11 '

s 10 i g < >

7 t '

i e ii 5 i 4

I 1

3

, 2 ._

1 F 4 0 1 0.0 0.2 0.4 0.8 0.8 14 1.2 1.4 1.8 RELAT1VE PCWER l

FIGURE 5.1.2 4

VY CYCLE 18 HALING DEPLETION. EOFPL CORE AVERAGE AXIAL POWER DISTRIBLTTION man 1

I I

R158 R17A R188 MleA Rt&A R17A R16A R178 R15A R14A Rtl8 I

1.1M 1 A08 1.142 1200 1.087 1M7 1844 1221 OJ13 Om1 0370 R17A RISA R17A RieA R174 R16A R17A R178 R14A R198 R15A 30 1J08 1346 1 A06 12M 1 Ass 1234 1.334 1223 0327 0 408 0.366 R188 R17A R188 R17A RIES R17A RISS R178 R14A R188 R168 1.143 1 A00 1.173 1 A70 1.127 1.300 1A30 1.1M OA00 0A00 0319 R14A R14A R17A R14A R17A RISA R17A R17A R14A R168 18 1274 1J04 1A79 1298 1A13 1.1M 1200 1.110 0.792 OA38 R168 R17A R188 R17A R168 R17A R158 R14A R198 1.101 1 441 1.137 1 A17 1.075 1230 0 917 0 866 OA60 R17A R14A R17A R14A R17A R14A R178 R188 RIAA 1200 1232 1.300 1.1N 1J00 1 A84 1488 0.750 0 444 Rt&A R17A R158 R17A RIAA R178 R14A R188 R158 10

,_ ,_ 1. .1 01 ,_ O_ O_ .3 R178 R178 R178 R17A R14A R188 RieA R188 nn 12a6 1_ 1_ 1.130 0.eN 0.7ei O sas 0_

RISA R16A R14A R14A RIM RIEA R158 na 0 00s Om1 Die O_ O Mi 044e 024:

R14A R148 R188 RIS8 n4 0 802 0As7 0 830 0 442 Ri&A R168 R188 --. 8U80LE D ns 0374 0.363 0J21 EOFPL RELATIVE POWER i

23 26 27 30 31 33 36 37 39 41 43 I

I FIGURE 5.1.3

, VY CYCLE 18 RODDED DEPLETION - ARO AT EOFPL.

BUNDLE AVERAGE RELATIVE POWERS mess I I

= ., - - - . . - -

l l

FIGURE S.1 A W CYCLE 18 ROODED DEPLETlON - ARO AT EOFPL CORE AVERAGE AX1AL POWER DISTRIBUTION as , i 24 x

• • M 21 J

20 l 4

18 i l

18 < >

17 i g,8 i t'-

,,. t 1

13 '>

1. 12 11 l10 8

8 i 7 4 '

1 5

! 4 1

0.0 0.2 0.4 0.8 0.8 1.0 1.2 1.4 1.8 RELATWE POWER l

FIGURE 5.1.4 .

VY CYCLE 18 RODDED DEPLETION - ARO AT EOFPL.

CORE AVERAGE AXIAL POWER DISTRIBUTION mess

,E _ - 4.- . -- --m .m

. I I

R168 R17A R188 R1eA R164 R17A R15A R178 RtEA R14A Rt68 31.700 13318 32Jg7 33J40 32.418 13 023 31313 12A14 31477 3 .700 20 170

]

i R17A R14A Rt7A R14A R17A R14A Rt7A R178 RieA Rie8 R15A 18882 23 M7 teses M.37 13.307 M.000 13.375 12.479 30264 30220 24.309 q

M188 M17A R188 R17A RtS8 R17A R158 Rt78 R10A R188 R158 M.161 14.001 32.843 14.148 32.806 13.733 N.738 12274 22J71 18At5 20.000 R14A AleA R17A Rt4A R17A R14A R17A R17A Al8A R168

' 13222 M.830 12.781 11 A01 a1Att 38.814

23a22 24230 14.148 23363 R168 R17A R158 R17A R188 R17A R188 RisA Rie8 32.403 13 001 32.833 13 917 32 326 13.024 32360 32.344 30.006 R17A R16A R17A R14A R17A R14A R178 R188 RISA 13.418 34 702 13.727 34.723 13.3 4 M.006 11.138 21.840 38414 R16A RI7A RISS Rt7A RiaA R178 R18A R188 R158 31 744 13.373 32.757 12.779 32208 11.1 2 22.048 30.153 28.757 R178 R178 Rt?8 R17A R14A R108 R14A R158 na 12 404 12.471 12E71 11A00 22.330 21.757 30.3W 38 938 R15A R14A R14A R14A RieA R15A R188 na 31AS4 22 226 23.067 21 A72 30.848 28.057 20 938 R14A R188 R148 R188 64 30.701 30275 19.800 38.790 R18A R158 R188 . SUNDLE D ns a asie7 2s sat se see EOFPL EXPOSURE AWD/ST) m3 26 27 se Si 33 as 37 3e 41 43 m

I I

,mRE m g

VY CYCLE 18 HALING DEPLLTON. EOFPL BUNDLE AVERAGE EXPOSURES R253 ,

I

I

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

f i

b s

4 RtGS R17A RtOS R144 R14A RITA R15A R175 R164 R14A R198 f =

1 30JW6 11 ABO 31.713 38.419 MABB 18 843 91 AGE 18A73 Eles 21418 MA01 R17A R10A R17A RteA R17A AleA R17A R15 R144 R188 R18A  ;

i 30 a St.1N ESAM ta m MM 13AB7 WM1 18Ag5 12A50 M.148 21.100 38 083  ;

j l R198 R17A Rim R17A Rtes R17A R15 R175 RieA R145 R1M

is t 1 91A07 18Jg8 33.385 13AM EMB 19.314 MA08 ISAM 35.710 30.756 30A18 t

! AleA R144 R17A R14A R17A R184 R17A R17A RIGA R188 d

is

! 3841 33.800 18A00 NAAR 19 373 35.008 18AM 11A00 Ette NJB4 l

1

< R100 R17A Ries R17# Rim R17A R185 R14A RieB  !

l 14 3BJes itsas $3A71 1sA14 Seas st eos mass asAM 31 Ass i

R17A AleA R17A RteA R17A R14A R175 Rim RIGA 13404 34Ago 1325 alist 1854 34440 10230 Stade 3A75 1

) RIGA R17A R188 R17A RIGA R175 RieA RieB R188  !

1 10 1 31 A18 18A57 NA17 itJe N.742 10.087 NA04 NAIS Mas ,

l ~

R170 R175 R175 R17A R14A Ale R144 RIES if itA16 18.141 ISA00 11370 38333 Re5 MASB NAS1 a

i RisA AleA RteA R14A R14A RtEA R150 na M 146 SEAN 33.740 M 175 21.00f 3B MS 38 494 R14A Ritt h185 Ate alsn 21.tts 30 Ass Esme

]

R1.A R1 8 Rt. __ -u o i no asaos So.ist Seer? __ eOFPL EXPCOURE POST)

! N N 37 N St 38 36 SF as 41 es i

4 l

l 4

I

! MOURE 5.2.2 l i

l  ;

. VY CYCLE 18 RODDED DEPLETION. EOFPL BUNDLE AVERAGE EXPOSURES G

i i 34:43 -24 l

t

a a I

I

... g

.4 2.2 t.0

/

/ \ g 4.. -

[  % ~ ) g E

1.4 -

t

,, o Minimum shadoen usqsn

- Tem speamonion umn 1.0 -

0..

3 I

0..

0.4 -

0.2 0.0 0 1000 2000 3000 4000 5000 0000 7000 0000 9000 10000 11000 CYCLE EXPO.URE (uWD/ST) m O U m e s.3.,

g VY CYCLE 18 COLD SHUTDOWN MARGIN. IN %AK. VERSUS CYCLE EXPOSURE man 25 B

I

4

6.0 THERMAL-HYDRAULIC DESIGN i

j 6.1 Steady-State 1hennal Hydraulics i

i

! Core steady-state thermal-hydraulic analyses for the Reload Cycle were performed using the

FIBWR[16],[17],[18] computer code. The FIBWR code inmrporates a detallad geometrical

representation of the complex flow paths in a BWR core, and explicitly models the leakage flow to the l bypass region and water rod flow.1he FIBWR geometric models for each GE bundle type were I benchmarked against vendor-supplied and plant thermal-hydraulic information. i

. l l l i

i Using the fuel bundle geometric models, a power distribution calculated by SIMULATE-3 and 4 1

4 core inlet enthalpy, the FIBWR code calculates the core pressure dmp and total bypass flow for several power and flow combinations.1he core pressure drop and total bypass flow predicted by the j FIBWR code were then used in setting the initial conditions for the system transient analysis model i

6.2 Reactor Limits Determination 4 ,

4 i  !

i Section 3.11 of the Technical Specifications requires that tne plant assure the performance of j the fuel rods by not exceeding the Minimum Critical Power Ratio (MTR), the Maximum Linear Heat Generation Rate (MLHGR), and the Maximum Average Planar Linear Heat Generadon Rate (MAPLHGR).  ;

e .

i i

l 1he Reload Cycle fuel has MCPR operating limits, shown in Appendix A. The MCPR is a combination of the Fuel Cladding.Integdty Safety Limit (FCISL) and the change in a Critical Power I Ratio (ACPR) which occurs during an anticipated operational transient. For Vennont Yankee FCISL j is 1.07 [2]. OR 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 i

objective for nonnal operation and anticipated transient events is to maintain nucleate bolling.

Avoiding a transition to film boiling protects the fuel cladding integrity. Both the transient and normal MCPR operating limits are derived with the GEXL-Plus conelation[19), with appmpriate coefficients

, representative of the Reload Cycle's fuel types. For core flows other than rated, the MCPR limits 1

! man ,

i 1

l must be adjusted by a generic factor, Kf19]. "De analysis, described in tre Section 7.0, determines the Reload Cycle MCPR operating limits.  ;

1

'Ihe Reload Cycle fuel has a Linear Heat Generation Rate (LHGR) limit of 14.4 kW/ft for all Ii; 1

tamdle types. 'Ihe basis for this limit can be found in Reference 2.  ;

'lhe Reload Cycle fuel has Average Planar Linear Heat Generation Rate (APLHGR) limits. ,

shown in Appendix A. The Maximum APLHGR'(MAPLHGR) values are the most limiting of the

- fuel rod thermal-mechanical MAPLHGRs[20] and the LOCA analysis MAPLHORs (Section 8.2). 'Ihe fuel rod thermal-mechanical MAPLHGRs are the result of the GE fuel rod thermal-mechanical design l analyses, described in Reference 2. 'Ihese results assume that during steady-state: 1) the maximum j LIIGR is 14.4 kw/ft,2) the maximum peak pellet exposure is 60.0 GWd/Mt. and 3) maximum operating time is 7.0 yean. These results also assume that, during an anticipated operational transient, the thermal and mechanical overpower limits [21] are not exceeded. Tim transient analysis, described in Section 7.0, assures that the thermal and mechanical overpower limits are not exceeded. The

{

LOCA analysis, described in Section 8.0. detennines the LOCA analysis MAPLHGRs.  !

I 1

! I I

I I

I

- 2,.

I

r 7.0 ABNORMAL OPERATIONAL TRANSIENT ANALYSIS 7.1 Transients Analyzed Transient simulations are performed to assess the impact of certain transients on the heat tansfer characteristics of the fuel. 'Ihe purpose of this analysis is: 1) to detennine the MCPR t

operating limit so that the PCISL is not violated for the transients considered 2) to assure that the j thermal and mechanical overpower limits are not ew=W during the transient, and 3) to demonstrate  !

f.

j compliance with the ASME vessel code limits. ,

i J Past licensing analyses have shown that these transients result in the maximum MCPR:

j i

1. Pressurization transients, including the generator load rejec, tion with complete failure of the turbine bypass system and the turbine trip with complete failure of the turbine l bypass system; l 2. Loss of feedwater heating; l,. 3. Local rod withdrawal error; and

, 4. Misloaded bundle enor, including the rotated bundle error and the mislocated bundle enor. .

I To demonstrate that the fuel rod thermal and mechanical overpowers are not exceeded, the l l maximum powers resulting from the pressurization, loss of feedwater heating and rod withdrawal enor i i

! transients were compared to the criteria. To demonstrate compliance with ASME vessel code limits, j

! the main steam isolation valves (MSIV) closing with failure of the MSIV position switch is also analyzed. Brief descriptions and the results of the transients analyzed are provided in the following l l sections, j l

I amm f l

4

1 I

i 7.2 Pressurization Transients Analysis 7.2.1 Methodology g 5i The analysis involves two types of simulations. A system level simulation is performed to 1

determine the overall plant response. Transient core inlet and exit conditions and nonnalized power j from the system level calculation are then used to perform detailed thermal-hydraulic simulations of the fuel, referred to as hot channel calculations." The hot channel simulations provide the bundle transient ACPR (the initial bundle CPR minus the MCPR experienced during the transient).

The system level simulations are performed with the one dimensional (1-D) kinetics RETRAN model[22),[23],[24]. The hot channel calculations are perfonned with the RETRAN[25],[26] and TCPYA0l[27],[18],[23] computer codes. The GEXL-Plus correlation [19], contained in TCPYA01, evaluates the transient cdtical power ratio.

The hot channel transient ACPR calculations employ a two-part process, as illustrated by the flow chart in Figure 7.2.1. The first part involves a series of steady-state analyses perfonned with the RBWR, RETRAN, and TCPYA01 computer codes. The HBWR analyses utilize a one-channel model for each fuel type being analyzed, with bypass and water rod flow also modeled. The steady-state FIBWR analyses were performed at several power levels with other conditions (i.e., core pressure drop, system pressure, and core inlet enthalpy) held constant. The HBWR code results provide a steady-state CPR, active channel flow (AF) and bypass flow (BPF) for each active channel power l

I (AP).

The FIBWR conditions for channel power, channel flow, and bypass flow were then used as input to steady-state RETRAN/TCPYA01 hot channel calculations. Other assumptions are consistent with those in the MBWR analysis. The Initial Critical Power Ratio (ICPR) is the result of the steady-state RETRAN/TCPYA01 analysis. These results allow for the development of functional relationships, describing AP as a function of ICPR, and AF and BPF as functions of AP for each fuel type. These relationships are used in the iterative process for determ* ming the transient CPR, as shown in Figure 7.2.1.

l l

man I I'

k The second part of the hot channel calculations detennines the transient CPR perfonnance.

Cecause the AOR for a given transient vades with Initial Critical Power Ratio (ICPR), the hot channel analysis is an iteradve process. 'Ihe objective of the hot channel iteradon for each transient is

{ to determine the hot channel initial conditions which result in reaching the FCISL Each iteration 1 requires a REIRAN hot channel run to calculate the transient enthalples, flows, pressure and saturation properties at each time step. 'Ihese are required for input to the TCPYA01 code. TCPYA01 is then used to calculate a CPR at each time step during the transient, from which a transient A&R is derived.

i

] In response to Reference 11, NRC Safety Evaluation for FROSSFTEY-2, the hot channel l

methodology has considered the assumption of both fixed and time-varying power shapes. The fixed

. power shape assumes a 1.4 chopped cosine axial distribution which remains constant throughout the l transient. 'Ihe initial power shape for the time-varying power shape methodology is the Haling axial distribution used in the core wide analysis. 'Ihe time-varying hot channel power distribution is i 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 dermed as the more conservative results from the fixed and varying shape analyses.

i *

7.2.2 Initial Conditions and Assumotions 4

'Ihe initial conditions for the Reload Cycle are based on a reactor power level of 1664 MWe which includes a 4.5% margin on the current licensed reactor powerlevel of 1593 MW m . 'Ihis margin conservatively bounds the expected 2% calorimetric uncertainty. 'Ihe reactor core flow is assumed to

, be 100% of rated. 'Ihe 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). 'Ihe core inlet enthalpy is set so that the amount of carr>under from the steam separators and

the quality in the liquid region outside the separators is as close to zero as possible. For fast j pressurization transients, this maximizes the initial pressudzation rate and results in a more severe neutron power spike. A summary of the initial operating state used for the system simulations is provided in Table 7.2.1.

aan I'

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 MWg 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:

1. Scram setpoints are at Technical Specification [12] 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 and safety valve setpoints are modeled as being at the Technical I

Specification upper limit. Valve responses are based on slowest specified response values.

5. Control rod drive scram speed is based on the Technical Specification limits. The analysis addresses a dual set of scram speeds, referred to as the " Measured" and the "67B" scram times. " Measured" refers to the faster scram times given in Section 3.3.C.1.1 of the Technical Specifications. "67B" refers to the slower scram times given in Section 3.3.C.I.2 of the Technical Specifications.

7.2.3 One-Dimensional Cross Sections and Kinetics Parameters

'Ihe one-dimensional (1-D) cross sections and kinetics parameters are generated as functions of fuel tempenture, moderator density, and scram. The method [28) is outlined below.

I

._ 31 I

I

I A complete set of 1 D cross sections, kinetics parameters, and axial power distdbutions are pnerated imm base states using the Haling depletion established for EOFPL, EOFPL-1000 mwd /St, EOFPL 2000 mwd /St, and BOC exposure stalepoints. '!hese statepoints art, characterized by exposure and void history distributions, control rod pattems, and core thermal-hydraude conditions. The latter i are consistent with the assumed system transient conditions pmvided in Table 7.2.1.

1 i i

'Ihe BOC base state is established by shuffling from the previously defined Curmnt Cycle

{

j endpoint into the Reload Cycle loading pattem. A criticality search provides an estimate of the BOC critical rod pattem. 'Ihe EOFPL and intermediate mre exposure and void history distributions are j calculated with a Haling depletion as described in Section 5.2. 'Ihe EOFPL state is unrodded. 'Ihe i EOFPL-1000 mwd /St and EOFPL-2000 mwd /St exposure statepoints require base control rod

! patterns. 'Ihese 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 contml 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 withir. 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 l 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 l SIMULATE-3 cross sections are held constant at the reference state. To obtain the effect of the contml rod scram, another SIMULATE-3 reference case is run with all-rods-in. 'Ihe perturbation cases 4

described above are run again from this reference case. For each control state, a data set of kinetics

parameters and cross sections is generated as a function of the perturbed variable. There is a table set l for each of the 27 neutronic regions,25 regions to represent the active core and one region each for q the bottom and top reflectors.

4 4

4 men I

7.2.4 Turbine Trio Without Bypass Tmnsient OTWOBP)

The transient is initiated by a rapid closure (0.1 second closing time) of the turbine stop valves. It is assumed that the steam bypass valves, which normally open to relieve pressure, remain closed. A reactor protection system signal is generated by the turbine stop valve closure switches.

Control rod drive motion is conservatively assumed to occur 0.27 seconds after the start of turbine stop valve motion. 'Ihe ATWS recirculation pump trip is assumed to occur at a setpoint of 1150 psig dome pr:ssure. A pump trip time delay of 1.0 second is assumed to account for logic delay and M-G set generator field collapse in simulating 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 Imad Relection Without Byoass Transient (GLRWOBP)

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 hydraulle fluid pressure switches in the acceleration relay of the turbine contml system. Control rod drive motion is conservatively assumed to occur 0.28 seconds after the start of turbine control valve motion. The same modeling regarding the ATWS pump trip and bypass piping is used as in the turbine trip simulation. The influence of the accelerating main turbine generator on the recirculation system is simulated by specifying the main turbine generator electrical frequency as a function of time for the M-G set drive motors. The main turbine genemtor 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.

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. 'Ihe tansient results, reported in Table 7.2.2, correspond to the limiting bundle type in the core. The MCPR limits, in Table 7.2.2, are calculated by men I

adding the calculated ACPR to the PCISL 1he worst ACPR for the pressurization transients include an adjustment to allow for the exposure window of *600 mwd /St on Cunent Cycle and the exposure j uncertainty on the Reload Cycle [7).

7.3 Loss of Feedwater Heating Transient Analysis f, l

7.3.1 pss of a Feedwater Heater (LOFW10 Results I

A feedwater heater can be lost in such a way that the steam extraction line to the heater is shut 1

off or the feedwater Dow bypasses one of the heatess. In either case, the reactor will receive cooler

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

The response of the system due to the loss of 100*F of the feedwater heating capability was analyzed. This represents the maximum expected feedwater temperature reduction for a single heater or group of heaters that can be tripped or bypassed N y tingle event. The system model used is the same as that used for the pressurization transient analyr {Section 7.2.1). The initial conditions and ,

modeling assumptions discussed in Section 7.2.2 are applicable to this simulation.

4 Vennont Yankee has a scram setpoint of 120% of rated power as part of the Reactor  ;

f l 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. This approach was 1 l

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-1000 mwd /St, EOFPL-2000 mwd /St, and BOC.1he transient results, corresponding to the j limiting bundle type in the core, are listed in Table 7.3.1.1he MCPR limits in Table 7.3.1 are calculated by adding the calculated ACPR to the FCISL. The transient evaluation at j

{. EOFPL-1000 mwd /st was found to be the limitirig case between BOC to EOFPL-1000 mwd /St. The results of the system response to a loss of 100*F feedwater heating capability evaluated at EOFPL-1000 mwd /St as predicted by the RETRAN code are presented in Figure 7.3.1.

! men l l

1

I 7.3.2 Loss of Stator Cooling (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. 'Ihis 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 (maximum 104.5%). It is assumed that an instantaneous loss of extraction steam occurs to the Nos.

14 feedwater heamrs of both feedwater trains. This is a conservative assumption, since there would g

not be a total loss of steam to the feedwater heaters, and the reduction in heating steam would occur 5 over the several minutes required for the turbine runback. Also, no credit is taken for the heat capacity 9f structural materials in the process piping or feedwater heaters. This results in a stepwise decrease in feedwater inlet temperature as the feedwater travels through the feedwater piping to the i reactor vessel.

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

'Ihe transient response of the system was evaluated at several exposures during the cycle, EOFPL-1000 mwd /St, EOFPL-2000 mwd /St, and BOC. The transient results, contsponding to the limiting bundle type in the core, are listed hi Table 7.3.2. The MCPR limits in Table 7.3.2 are calculated by adding the calculated ACPR to the FCISL. The transient evaluation at BOC was found to be limiting case between BOC and EOFPL-1000 mwd /St. The results of the system response to a loss of stator cooling evaluated at BOC as predicted by the RETRAN code are presented in Figure 7.3.2. To assure that the thermal overpower limits are not exceeded, the MAPLHGR limits in Appendix A were modified according to Reference 41.

I aam I

i I 1

7.4 Ovemressurization Analysis Results '

l Compliance with ASME vessel code limits is demonstrated by an analysis of the Main Steam Isolation Valves (MSIV) closing with failure of the MSIV position switch scram. EOFPL conditions were analyzed. The system model used is the same as that used for the transient analysis (Section j 7.2.1). The initial conditions and modeling assumptions discussed in Section 7.2.2 are applicable to this simulation. l l

1he 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 I l

is generated on APRM high flux. Control rod drive motion is conservatively assumed to initiate 0.28 seconds aRer reaching the high flux setpoint. The system response is shown in Figure 7.4.1 for the

" Measured' A ram time analysis.

1he maximum pressures at the bottom of the reactor vessel calm 1=*ad for the " Measured" scram time analysis and for the "67B" scram time analysis are given in Table 7.4.1. These results are within the ASME code overpressure design limit which is 110% of the vessel design piessure.

Vermont Yankee's design pressure is 1250 psig so the maximum pressure limit is 1375 psig.

1 7.5 Imcal Rod Withdrawal Error Transient Results I The rod withdrawal ermr (RWE) is a local core transient caused by an operator erroneously withdrawing a control rod in the continuous withdrawal mode. If the core is operating at its operating limits for MCPR at the time of the error, then withdrawal of a contml md could increase both local j and core power levels with the potential for overheating the fuet 1here is a broad spectrum of core conditions and control md patterns which could be present at the time of such an error. For most normal situations it would be possible to fully withdraw a control rod without violating the FCISL.

The MCPR operating limit for the RWE is defined at each Rod Block Monitor (RBM) System setpoint so that the FCISL is not violated. The consequences of the error depend on the local power men l

I increase, the initial MCPR of the nelgi ooring locations and the ability of the RBM to stop the 1 withdrawing rod before MTR reaches the FCISL. l The most severe transient postulated begins with the core operating according to normal procedures and within normal operating limits. The operator makes a procedural error and attempts to fully withdraw the maximum worth contml md at maximum withdrawal speed. The core limiting locations are close to the error rod. 'Ihey experience the spatial power shape transient as well as the overall core power increase. 5 The core conditions and contml rod pattem are conservatively modeled for the licensing bounding case by specifying the following set of concurrent worst case assumptions:

1he rod should have high reactivity worth. The worst rod is identified by running the I

1.

full RWE analysis for the control rods as found in the normal contml rod pattems of the rodded depletion. Every control rod sequence is che@ad. From this examination, the control rods that result in the highest worth and highest ACPR are identified.

Licensing test case rod pattems are then developed to further exaggerate the worth and ACPR impact of the rod to be withdrawn.

The test pattems are developed with xenon-free conditions. The xenon-free condition and the additional control rod inventory needed to maintain criticality exaggerates tie I

worth of the withdrawn control rod when compared to nonnal operation with normal xenon levels.

l

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

3. The core power distribution is adjusted with the available contml rods to place the locations within the four by four array of bundles amund the error rod as close to the operating Ilmits as possible.

4. Of the many pattems tested, the pattem with the most limiting ACPR results is selected as the bounding case.

m I I

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

I  !

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

f

1. Technical Specifications allow each of the separate RBM channels to remain operable if at least half of the Imal Power Range Monitor (LPRM) inputs on each level are  !

operable. For the interior locations tested in this analysis, there are a maximurn of (

I l four LPRM inputs per level. One RBM channel averages the inputs from the A and C levels; the other channel averages the inputs from the B and D levels. Considering the i l ' inputs for a single h; there are eleven failure combinations of none, one and two failed I.PRM strings. The RBM channel responses are evaluated sepamtely at these  ;

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

2.' 'Ihe event is analyzed separately in each of the four quadrants of the core due to the ,

differing LPRM string physical locations relative to the enor rod.

l 4

Technical Specifications require that both RBM channels be operable during nonnal operation.

, Thus, the first channel calculated to intercept the RBM setpoint is assumed to stop the rod. To allow l l for control system delay times, the rod is assumed to move two inches after the intercept and stop at f

! the following notch.

i The analysis is perfonned using SIMULATE-3. 'Ihe two separate cases presented here are i

! selected from numerous explicit SIMULATE-3 analyses. Case 1 analyzes the bounding event with  ;

1 zero xenon, initiated from 104.5% power and 100% flow. This case also assumes the worst case abnormal rod pattem configuration which results in the initial MCPR being as low as possible. Case 2 i is the worst of all the rod withdrawal transients analyzed from 100% power,100% flow, equilibrium j xenon, and normal rod patterns used in the rodded depletion. The worst transient ACPR results for

} both cases are shown in Table 7.5.1. The ACPR values are evaluated such that the implied MCPR

operating limit equals PCISL + ACPR. 'Ihis is done by conserving the figure of merit (ACPR/ICPR) shown by the SIMULATE-3 calculations. The transient ACPR results for Case I will be used to set

the operating MCPR limits. Case 2 results are bounded by the Case I results by at least 0.02 ACPR margin to assure that the exposure uncertainties on the Current Cycle and the Reload Cycle are i

non (

i

I accounted for. This method also provides valid operating MCPR values that bound expected operating conditions.

The Case 1 (bounding event) RBM channel responses are shown in Figures 7.5.1 and 7.5.2.

I 1 hey also show the contml rod position at the point where the RBM channel response first intercepts the RBM setpoint.

7.6 Misloaded Bundle Ermr Analysis Results t

7.6.1 Rotated Bundle Error u The 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 sunnunding wide water gaps.

In addition, there may be a small increase in reactivity, depending on the exposure and void fraction states. 'lhe R-factor increase results in a CPR reduction. The objective of the analysis is to ensure i

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 I

developed by adding the largest possible AR-factor resulting from a rotation to the exposure dependent R-factors of the properly oriented bundles. Using these rotated bundle R-factors, the MCPR values resulting from a bundle rotation are determined using SIMULATE-3. This is done for each control rod sequence throughout the cycle. The process is repeated with the K-infinity of the limiting bundle modified slightly to account for the increase in reactivity resulting from the rotation. For each sequence, the MCPR for the properly oriented bundles is adjusted by a ratio necessary to place the corresponding rotated bundle's CPR on its FCISL. The adjusted MCPRs at each exposure is the rotated bundle operating limit for the rotated bundle enor.

Because the BP8DWB335 fuel designs exhibit a significant increase in R factor with rotation early in exposure, the impact upon the rotated bundle ACPR is high at BOC. This effect soon drops off with exposure. Therefore, the operating MCPR limit resulting from a rotation is presented in Table 7.6.1 versus cycle exposure, men i

7.6.2 Mislocated Bundle Enor Mislocating a high reactivity assembly into a region of high neutron imponance results in a location of high relative assembly average power. Since the assembly is assumed to be propedy oriented (not rotated), R-factors used for the mislocated bundle are the standard values for the given  ;

fuel type.

'lhe analysis uses multiple SIMULA'IE-3 cases to examine the effects of explicitly mislocating every older interior assembly in a quarter core with a fiesh or once-burned assembly. Because of symmetry, the results apply to the whole core. Edge bundles are not namined because they are never lim 8. ting, due to neutron leakage.

'Ihe effect of the successive mislocations is examined for every control rod sequence thmughout 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. 'Ihe MCPR for the propedy loaded core is adjusted by a ratio me='y to place the mislocated assembly on the FCISL. 'Ihe maximum of these adjusted MCPRs is the mislocated bundle operating limit. The results of the mislocated bundle ar$alysis are given in Table 7.6.2.

7.7 Transient Analysis Results

'Ihe 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 mechanical overpower limits are not exceeded during the transient, and 3) demonstrated compliance with the ASME vessel code limits.

'Ihe 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 limit at either end is assumed to apply to the whole interval. 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 are provided in Appendix A. These MCPR operating limits are valid for operation of mem .

i I  !

i i the Reload Cycle at full power up to 10644 mwd /St and for operation during coastdown beyond i

EOFPL. l 4

i d

i t

}

e 4

il k

i i

1 I

l 1

s i

u a

i n

i 4

4 I

.I d

l men l l

1 I

l TABLE 7.2.1 VY CYCLE 18

SUMMARY

OF SYSTEM TRANSIENT MODEL INITIAL CONDITIONS FOR TRANSIENT ANALYSES Core Thermal Power (MWg ) 1664.0 i

hrbine Steam Flow (105bA) 6.75 Total Core Flow (10$bA) 48.0 r

8 Core Bypass Flow (101bA)* 6.28 Core Inlet Enthalpy (BTU /lb,) 523.2 Steam Dome Pressure (psia) 1034.7 Turbine Inlet Pressure (psia) 985.7 8

Total Recirculation Drive Flow (101bA) 23.7 Core Plate Differential Pressure (psi) 20.4 Narrow Range Water Level (in.) 162.0 Average Puel Gap Conductance (See Section 4.2)

Includes water rod flow.

amas TABLE 7.2.2 i VY CYCLE 18 PRESSURIZATION TRANSIENT ANALYSIS RESULTS Peak Prompt Power Peak Average Heat Exposure (Fraction of Flux (Fraction of Transient Transient Statenoint Initial Value) Initial Value) ACPR* MCPR Limits Turbine Trip Without EOFPL 2.80378 1.19541 0.25 132 Bypass, " Measured" EOFPL-1000 2.24381 1.14296 0.20 1.27

! Scram Time EOFPL-2000 1.25154 1.00000 0.04 1.11 Turbine Trip Without EOFPL 3.12172 1.23971 0.27 134 Bypass "67B" Scram EOFPL-1000 2.63805 1.19536 0.23 130 Time EOFPL-2000 1.64655 1.04145 0.08 1.15 Generator Imad EOFPL 2.81519 1.17663 0.23 130 Rejection Without EOFPL 1000 2 34636 1.12670 0.19 1.26 Bypass, " Measured" Scram Time EOFPL-2000 1.13637 1.00000 0.02 1.09 Generator Imad EOFPL 3 3 0268 1.23750 0.27 134 Rejection Without EOFPL 1000 2.92611 1.19323 0.23 130 Bypass, "67B" Scram Time EOFPL-2000 1.60336 1.01760 0.05 1.12

'Ihe worst ACPR for'ITWOBP and GLRWOBP includes a 0.01 ACPR adjustment to allow for the exposure window of 1600 mwd /St on Current Cycle and the exposure uncertainty on the Reload Cycle, mass

.. . _ ~ . _ . __ _ _ _ . _ . . . . . _ . . _ m.

TABLE 7.3.1 VY CYCLE 18 LOSS OF FEEDWATER HEATER TRANSIENT RESULTS 4

.i I

Peak Prompt Power Peak Average Heat i Exposure (Fraction of Flux (Fraction of Transient Transient Statenoint Initial Value) Initial Value) M MCPR Linuts i

l IAss of 100*F EOPPL-1000 1.24680 ~1.15373 0.12 1.19 Mwm Heating BOPPL-2000 1.14354 1.14440 0.11 1.18  ;

2 BOC 1.17041 1.14530 0.11 1.18 i

I i

4 TABLE 7.3.2 j VY CYCLE 18 LOSS OF STATOR COOLING TRANSIENT RESULTS Peak Prornpt Power Peak Average Heat

! Exposure (Fraction of Flux (Fraction of Transient I

Transient Statcooint Initial Value) Initial Value) M MCPR Limits i

1 Ioss of Stator Cooling EOFPL-1000 1.23165 1.12990 0.11 1.18 EOFPL-2000 1.19910 1.18716 0.13 1.20 BOC 1.19806 1.18498 0.13 1.20 aan 1 1

f l

I

! TABLE 7.4.1 4

~

VY CYCLE 18 OVERPRESSURIZATION ANALYSIS RESULTS Maximum Pmssum at Reactor j Conditions Vessel Bottom (osin) 4

" Measured" Scram Time 1251 "67B" Scram Time 1278 i

l TABLE 7.5.1

=

VY CYCLE 18 ROD WITHDRAWAL ERROR ANALYSIS RESULTS i

Rod Block Monitor Transient MCPR I

Setoolnt Bounding Case ACPR Worst Normal ACPR Limits 104 0.15 0.13 1.22 105 0.16 0.14 1.23 106 0.16 0.14 1.23 i

107 0.20 0.18 1.27 108 0.26 0.18 1.33 I

I I

mem I

16),LE 7.6.1 ,

VY CYCLE 18 ROTATED BUNDLE ANALYSIS RESULTS Exoosure (GWd/St) Transient MCPR Limit -

0.0 1.39 4.0 1.35 5.5 1.29 6.5 1.25 10.0 1.25 TABLE 7.6.2 VY CYCLE 18 MISLOCATED BUNr>LE ANALYSIS RESULTS Transient MCPR Limit 1.15 I

- 1 l

I TABLE 7.7.1 VY CYCLE 18 LIMITTNG TRANSIENTS Rod Block Monitor Scram Time BxDosure (GWd/SO Limiting Transient Transient MCPR g Setpoint Limit E 108 Measured 0.0 Rotated Eundle 139 4D Rotated Bundle 135 v 5.5* Rod Withdrawal Error 133 108 "67B" 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 5.5 Rod Withdrawal Error 133 g 9.035* Turbine Trip 134 5 107 Measured 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 g 5.5 Rotated Bundle 1.29 m 6.5 Rod Withdrawal Error 1.27 9.035* Turbine Trip 132 107 "67B" 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 5.5 Rotated Bundle 1.29 6.5 Rod Withdrawal Error 1.27 8.035 Turbine Trip 130 g 9.035* Turbine Trip 134 5 106 Measured 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 5.5 Rotated Bundle 1.29 6.5 Rotated Bundle 1.25 8.035 Turbine Trip 1.27 9.035* Turbine ' nip 132 106 "67B" 0.0 Rotated Bundle 139 g 4.0 Rotated Bundle 135 g 5.5 Rotated Bundle 1.29 6.5 Rotated Bundle 1.25 8.035 Turbine Trip 130 9.035* Turbine Trip 134

• Transient MCPR is limiting from this exposure to EOFPL.

a_ a,.

I

Part I e s l e Part 11 l R and l e RAN/ e l CPVA01 e Es tima,t,e, p, I,n,i,tt el 5 % y te l Transient Analysis l

s l l l l

Functional Relettenshiss e RETRAN/TCPYA01 e AP = f(ICPR) l Transtant Not e AF = f . Channel Assalyste l BPF = (AP) f(AP) l

. . . . . . , ..... 2 Estiente New ICPRg es:

ACP ,

2 2CPn, N,mm, ,.

l5T0P l l

l i

FIGURE 7.2.1 l 1

FLOW CHART FOR THE CALCUI.ATION OF ACPR USING THE RETRAN/TCPYA01 CODES 1

l i

man ,

i l

1 I

E I

TTWOBP, EOFPL, MST TTWOBP, EOFPL, MST 10F 5 20F5 g a 1.5 g 5-1.25 -

oc " $ l' W

e W

a E

/.\ -

l

[1.0 ',

lw 3- o

, i z.

g ..

g .,

@2- '.,

0.75 - ',

j 1 ,

l CORE INLET FLOW l NORM. 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 TIME (SEC) TIME (SEC)

I FIGURE 7.2.2 TURBINE TRIP WITHOUT BYPASS. EOFPL18 TRANSTENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME man I I

I a

TTWOBP, EOFPL, MST TTWOBP, EOFPL, MST 30F5 4 0F 5 j 1500 2000 p---.... ,,, ..

1500 - s

• s L.* ii >

1400 - A.

i i

1000 Y!

I a.  ;

3 G rd  ; \

W 1300 -

500 -

V j 'v y

m 0 1 .

2 1200 -

500 -

Ill ::

. 1000 -

l 1100 - _

' ' ~

S/R VALVE (NEG) 1 ---- FEED WATER I

l STEAM DOME PRES.l -~~~ VESSEL STM.OUT l 1000 . .

.. . . . -2000 .. .

. . . . l 0 1 2 3 4 5 6 0 1 2 3 4 5 6  !

TIME (SEC) TIME (SEC) ,

l l

3 FIGURE 7.2.2 (Continued)

'l TURRINE TRIP WITHOUT BYPASS. EOFPL18

'IRANSIENT RESPONSE VERSUS TIhE. " MEASURED" SCRAM TIME i l

m -l M

I I

! I i

t i

J TTWOBP, EOFPL, MST g 5OF5 2.0 j j

1.8 i 1.6 i

[ g g

1.4 i f

1.2 -! -

j 1.0 -i j

\

i 4

0.8 i \ /

~'

6 0.6 i 0.4 i

] f 0.2 -! ,'

g 0.0 -

.., , ***,, M 0.2 i \., --- --------

-0.4 i *~.

l g -0.6 i s, l 0.8-i '\

1.0 ! g 1.2 ! i

\

1.4 i TOTAL \.

1.6 ! ---- DOPPLER ')

........ MODERATOR i 1.8 i .... 8 CRAM i

2.0 ., .

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC) en E

E FIGURE 7.2.2 (Continued)

TURBINE TRIP WITHOUT BYPASS. EOFPL18 TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME

- -51 g

I i

C w

T1WOBP, EOFPL-1, MST TTWOBP, EOFPL-1, MST  !

10F 5 20F5 e 1.0 5-l 1.25 - t 4-Y ,l '.

d '

h

,.0 - .

y  : .

l 2- .

O.75 - .

\'.

1 i

$ . CORE INLET FLOW l t

l NORM POWER l - = = - AVE. HEAT FLUX  !

I o .... .

, . 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 l

TNE (SEC) TWE (SEC) l l

i FIGURE 7.2.3 TURBINE TRIP WITHOUT BYPASS. EOFPL18-1000 MWD /ST

TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME

! man -

3 e

I l

l I

I TTWOBP, EOFPL-1, MST TTWOBP, EOFPL-1, MST ,

30F5 4 0F 5 2000 1500 1500 - g 81 4 l i $ 11 .

1400 - i } ti l ,

l 1000 - l l l k l!!I!

E I il ~i

{ -  :  ! $f 21

• W b l

@ 1300 - lm 500 - l!!

l j l

!l!!

8 i{!

$ d w

E M 0

! l-- li:

1200 - 3:

-500 -

l!

l

~

1000 -

1100 -

' " ~

8/R VALVE (NEG)

-== = FEED WATER l l- STEAM DOME PRES.l - -

• VESSEL STM. DUT 1000 2000 l

0 1

2 3 4 5 6 0 1

2 3 4

5 6 5 l

l TIME (SEC) TIME (SEC)

I FIGURE 7.2.3 (Continued) 1 TURBINE TRIP WITHOUT BYPASS. EOFPL181000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRA'M TIhE

_ 33 L_______________-___--_____-_-____--_-_-_________________-_-____________-_______________

TTWOBP, EOFPL-1, MST i 5OF5 2.0 .

1.s i

, 1.6 !

1.4 !

1.2i 1.0 .

['i \ .s f 0.8 !

^ 0.6 i '

0.4 !

f 0.2-!

~

0.0

' - . . , ' ----------- ~~,,.*

-0.2 i -

g 0.4 i 'i% I i

a -0.6 i  %,,'

-0.8-!

\

1.0 ! \

1.2 i

\

1.4 i TOTA 'g I 1.6-! ---- DOPPLER 1

... . MODERATOR '

1.8 ! ....- seRAu 'i 2.0 ........ ,, \ , .. . .

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC)

FIGURE 7.2.3 (Continued)

TURBINE TRIP WITHOUT BYPASS. EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME

- - 54-i m._____.__________ _ _ _ _ _ _ . _ _ _ _ _ _ __ _ _ _ _ _-_ _ _ ,

I I

I i

T1WOBP, EOFPL-2, MST TTWOBP, EOFPL-2, MST 10F 5 ,

20F5 6 1.5 .

5-I 1.25 -

4- $

4 k

8 E g 3- f1.0 o

w ,' ',.

z g * <

l

? .

G '.'.

@2- g '., ,

0.75 - .

j 1 i CORE INLET FLOW l NORM. POWER l  %

---

AVE. HEAT FLUX 0 ........ .... . ...... 0.5 .

O.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)

I FIGURE 7.2.4 I)

TURBINE TRIP WITHOUT BYPASS. EOFPL18-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME i

I

_ .ss.

r i

1 i

a

' TTWOBP, EOFPL-2, MST TTWOBP, EOFPL-2, MST SOF5 4 0F 5 -

! 1500 2000 s, ,s.

I i -

1500 - lI l .

. .A i.:

i' ,A il 1400 - l

'i  !! Ii li (t

!. 'Il4 l i i a 1000

}l\,

e . .  !\l i i

s' 1 r' I. k!!\

500 -  : f' ti T 1300 -  : *  !! e

.l i a .

I si M 0 8!

"' )

1200 -  !!:

500 -  : I

$ i 1 i

L_- _

. -1000 -

1100- -

4

' ~

SS VALVE (NEG) l

---

FEED WATER l STEAM DOW, PRES.l - *-* VESSEL STM.OUT 1000 . .,.. .,....,... ,....,,, . 2000 ....,.. ,,... ,... ,... ,

0 1 2 3 4 5 8 0 1 2 3 4 5 8 TIME (SEC) TIME (SEC)

I FIGURE 7.2.4 I j-(CDntinued)

TURBINE TRIP WITHOUT BYPASS. EOFPL18-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME nem i

I: l Il !

I; 1

TTWOBP, EOFPL-2, MST l l

4 5OF5 2.0 l 1.8 i ^/

J

1.6 !

1.4 i [

t l

1.2 i /

1.0 i 0.8 i f

j f

i ~ 0.6 i f i . 04d f ,,

• 0.2 8 ..

~

0.0 ** - ... ********"

-0.2 4

0.4 i g -0.6 i

gl g

l {

-0.8 i \

1.0 i t,

. 1.2 i \

1.4 i TOTAL l 1.6 i * **

• DOPPLER

... . E DERATOR

-1.8 i ..... scRAu  :

1

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

0.0 0.5 1.0 1.5 2.0 2.5 3.0 I, ,

TIME (SEC)

I FIGURE 7.2.4 l (Continued)

TURBINE TRIP WITHOUT BYPASS. EOFPL18-2000 MWD /ST Il '

TRANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME ness \

I I

4 4

GLRWOBP, EOFPL, MST GLRWOBP, EOFPL, MST ,

I 10F 5 2 OF 5 s 1.5 1 p 5-

~

1.25 -

I.

W

1. ~

> ,l ' .

! (E l g '.*

3- 1 0- , , .

8 '.' ,l '.

g - --.,

2-

\. .

l I

1 O.75 - '.

1 I .

- CORE INLET FLOW l NORM. POWER l ---- AVE. HEAT FLUX i 0.5 0 ................ .......... ....... ........ ...........

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 i TIME (SEC) TIME (SEC) i FIGURE 7,2,5 GENER ATOR LOAD REJECTION WITHOLTT BYPASS. EOFPL18

'IRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME amm I i

I l I

I:

l GLRWOBP, EOFPL, MST . GLRWOBP, EOFPL, MST 3 OF 5 4 0F 5 1500 2000 1500 -

, ,i iif.

t.

- 5 4* -  !

!3 A, l

\

\\

I j 11

\ \ \ '\i t

'/

l\

i l

h ^

! p/ t $t w 500 -  ! !' I I!' 'v'

@ 1300 - h j W

. m N

g 8 t l

E m 0 l:l A. i j W i :

1 !

1200 - 3: lf o 500 - i .:

g e r

~

1000 -

l 1100 - *

• ~ ~

SH VALVE (NEG)

---

FEED WATER l STEAM DOME PRES.l -- VESSEL STM. OUT 1000 .

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

,,,. ,,., , . ,, g 0 1 2 3 4 5 6 0 1 2 3 4 5 6 g TIME (SEC) TIME (SEC)

I FIGURE 7.2.5 (Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18 TR ANSiENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME mem I i

t i

i e

4 E

GLRWOBP, EOFPL MST SW5 i

2.0

/

1.8 i 1.8 i 1.4 ! i

,b

[g\, If 1.2 i 1.0 i 0.8 !

[ \ .

~ 0.8 -!

0.4 !

i I

[

0.2 i i

~

0.0 -

,,, y , .. .............

-0.4 -!

i

%,s

' n. 0.8 i \g

-0.8 i 'g 4

1.0 i \,

t 1.2 !

\

1.4 i Toyn \g

. , .8 ;' . . . . oo, i MODERATok k f 1.8 i ..- - SCRAM

\-

t 2.0 ... -

ii t ' '

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC) 1 l

FIGURE 7.2.5 j (Continued) 1 l

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18 j TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME j l

l men - . - _

\

L I

! I I

I GLRWOBP, EOFPL-1, MST GLRWOBP, EOFPL-1, MST 10F 5 20F5 g 6 1.5 j 3 5-I 1.25 -

w

~

2 W l\

E l <

3- 1.0 ., j

+

! E 2 './ \- ., ,

i E .

@2- \.,

0.75 - ..

CORE INLET FLOW l NOA% POWER l - = = - AVE. HEAT FLUX 0- -v .- , .., .

i. ..... . 0.5 ..... ..... . ..... ..... g 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 3.03 TIME (SEC) TIME (SEC)

I FIGURE 7.2.6 GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME '

am I I

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

A i

1 l

l l

l t

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

S OF 5 4 0F 5 1500 2000 g.................,***....

} 4, ,

1500 -

l 1400 -

g

+

l

!n! j\: ::

^

i=0- I  ! :I j \  ::\

<W

( 500 -

lllj,II t

if 1 I

I

• v!

\

g 1300 - l: l: 1:t*

l

'd d

" o l lj  !

I!

if i w '

: i l38 1200 -

t o -500 - i' )8 d  ;[

to -

i i . 1000 -

1100 -

, . 1500 -

aR VALVE (NEG) ,

. .. . FEED WATER

~~-a VE88EL STM. OUT l l STEAM DOME PRES.l q 1000 .

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

l.

0 1 2 3 4 5 6 0 1 2 3 4 5 6

, TIME (SEC) TIME (SEC) i FIGURE 7.2.6 (Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME i

< i m .(Q,.

i

I 4

I i I i

I l l

GLRWOBP, EOFPL-1, MST 5OF5 l

2.0 j 1.8 i 1.6 i j 1.4 -i ,.., /

1.2-i

. \'v/

j i

1.0 i O.8 i /

~ 0.6 i 1

0.4 ! ,

f 0.2 i [ ',,'

~

0.0

-0.2 i

\

• ----------- ~~'

i -0.4 i 'g s.

a m 0.6 i '..,

j 0.8 ! N 1.0 -i 'g 1.2 i g 3

1.4 i 7o7g 'g

! 1.6 i ---* DOPPLER  !,

j uODERATOR i

-1.8 . ....- SCRAM i

.....,l,. I

-2.0 ....... ,

4 0.0 0.5 1.0 1.5 2.0 2.5 3.0 l TIME (SEC)

I ,

FIGURE 7.2.6 (Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME I

sen I

t 4

I' i

i 4

6 I

GLRWOBP, EOFPL-2, MST GLRWOBP, EOFPL-2, MST 10F 5 2 OF 5 8 1.5 t .'

~

] 5-1.25 -

'- 8 l

W l k

. E  ;

3-Ly $. 1.0 '.. ~.

g .

}' 's,  ;

i 2-A l

0.75 - .\.'.

l .

4 1

j . .,

1 - CORE INLET FLOW l NORM 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 TIME (SEC) TIME (SEC)

. ELQiLRE 7.2,7 GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME

M 64-s i

l l

l1 I'

I GLRWOBP, EOFPL-2, MST GLRWOBP, EOFPL-2, MST SOF5 40F5 1500 2000 g.................,,.....

t ,,

}  ::

1500 - g 11 1400 -

i j%1 li f4 f.

[

1000 - l

'-- '\j~ l '1 sr l'lij'!1/\

E w

@ 1300 -

E 500 - \ ' \ ,l 'ksf \lb i l ? ,c tl

• l

• i i i i j i 00 - 3

] $ -500 - l[

8 1000 -

\

. 1100 -

0-8/R VALVE (NEG)

.... FEED WATER l STEAM DOME PRES.l - *= VESSEL STM.0UT

. 1000 ....... , .

........ . .. -2000 .

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

I FIGURE 7.2.7 (Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18-1000 MWD /ST TR ANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME mrom 65

GLRWOBP, EOFPL 2, MST 5OF5 2.0 1.8 i f

1.6 ! /

1.4 i 1.2 !

1.0 !

0.8 i

~ 0.6 i I I 0.4 i 0.2 i

~..**..

0.0 j

'-~~ -------- *****,,,...

-0.2 i

0.4 !

t' g 0.6 !

0. 8 \,

1.0 i \t 1.2 ! i

s.

,! 1.4-i TOTE

! 1.6 i ---- DOPPLER

... MODEMTOR l 1.8i ..... SCMM 2.0 ., .... '... ,.l., i. .....

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC) l l FIGURE 7.2.7 (Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL181000 MWD /ST b

TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME i

[

man _ . - _ - .

l I

I 4

I I'

LOFWH, EOFPL 1 LOFWH, EOFPL 1 4

10F 4 20F4' g 1.5 1.0 W 0.8 4 0.64 ,

0.4 -

1.25 - 0.2 - .

us

~ 0.0

• • • ,,... -~ ~ """*~~ % -. g

y ,,,..-

0.2 : - .......................------ 5

-0.4 4 g, .*[.......... - - ...... - - ........ -

8 0.62

-0.8 :

E 1.0 :

1.2 O.75 -

j 1.4 -

4 1,6 TOTAL M

---

DOPPLER j

NORM. POWER

• *-- CORE INLET FLOW 1,3 .: ~~~* MODERATOR E

j a~a

• AVE. HEAT FLUX - - - 8 CRAM j 0.5 ....... .

, 2.0 ...., , .

........ ...... E 0 25 50 75 100 125 150 0 25 50 75 100 125 150 5 TIME (SEC) TIME (SEC)

I FIGURE 7.3.1 I

LOSS OF 100*F FEEDWATER HEATER (LIMITING CASE)

TRANSIENT RESPONSE VERSUS TIME I

mem -67 I  ;

r

?

l l

LOFWH, EOFPL-1 LOFWH, EOFPL-1 3OF4 4OF4 100 2500 90 -

i 80 -

~ '

2250 -

70 - -

8 e

. E e

I50 m

n-

. g 40 -

• 2w0-g ' 'p ,.. ----t w ....... ...,,,,,,,

g  : e t

w 30 1750 -

20 -

10 -

FEED WATER FLOW l CORE IN SUBC00UNG l - -- VESSEL STM. OUTLET 0 .. ......, ... .... , .. 1500 .

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

FIGURE 7.3.1 (Continued)

LOSS OF 100*F FEEDWATER HEATER a.IMITING CASE)

TRANSIENT RESPONSE VERSUS TIME

-m 1 J

I I

l i

I l

l LOSC,000 LOSC,BOC 742 742 10F4 2OF4

! 1A 1.0  ;

0.8 d f

b DA d $

0.4 .  ;

1.26 -

0.2 k l /

/ 42.

44-0

= ...........** ...,

' l gy , ........w *******'

8 48-p 484 l

• 1.0 i

,,,. ,o

• 1.4 i l 1.3 f - TOTAL

- seasua powtR * *** DOPPLEA

• • • • COME INLET PLOW 1,3 J'

- nCDERATOM

= AVE. HEAT PLUM **=== SCAAM

, 0.s ... ... ... ... ... 2.0 ..

0 to to 75 100 125 150 0 25 80 75 100 126 150 TWE (SEC) TNE (SEC) 4 I l Ii I

FIGURE 7.3.2 LOSS OF STATOR COOLINO TRANSIENT RESPONSE VERSUS TIME I

ame I l

i l

l l

LOSC,BOC 14SC,BOC 742 742 SOF4 40F4 50 -

2250 -

70 -

e 4

g l

I50 g- q n

g ........ - ........-

1750 - l 30 - l l

~

l - FEED WAM EDW l- oona a suecoouna j *

• VES$EL STM OUTUET l {

l 0 ... ... ... ... ... 1N2 ... ... ... .;. ...

0 25 50 75 100 125 150 0 05 50 75 100 125 150 TNE (SEC) TNE (SEC)

FIGURE 7.3.2 (Continued)

LOSS OF STATOR COOLING TRANSIENT RESPONSE VERSUS TTME m_

I I

I I

i MSIVC, EOFPL, MST MSIVC, EOFPL, MST 10F 5 2OF5 6- 1.5 I

~

5-4 g.

1.25 - W w

i E **

80 3 1 w .

i, 2 -

5 j

i

@2-I

0.75 - '.,

g CORE INLET FLOW .

= = = = AVE. HEAT FLUX

• l NORM. POWER l .

] 0 ....... ........ , .

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

0 1 2 3 4 5 6 0 1 2 3 4 5 6

]

TIME (SEC) TIME (SEC)

I FIGURE 7.4.1 I

MSIV CL,OSURE. FLUX SCRAM. EOFPL18 TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TINE I

men .

MSIVC, EOFPL, MST MSIVC, EOFPL, MST 3 OF 5 4 0F 5 1500 2000 1500 -

i l

\ ~

1400 - -

{ A

,000 -

\ \\lyyvv. . ...

Q t tW g 500 -

\i l

l @ 1300 - h i l 0

^

g

• 0

\l))s'

?

1200 -

500 -

I 1000 -

1100 -

' ~

! &R VALVE (NEG)

---

FEED WATER

---a VESSEL STM. DUT l STEAM DOME PRES.l 1000 .... .... ..

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

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

FIGURE 7.4.1 (Continued)

MSIV CLOSURE, FLUX SCR AM. EOFPL18 TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME men _ _ _ _ _ _ _ _ _ _ _ _

I' I

s 1

6 MSIVC, EOFPL, MST E i 5OF5

, 2.0 ,

1.8 i f

1.6 : f 1.4-!

f f

. 1.2 i / g 1.0 i / W 0.8 4 ,,./

^

[

4 lg 0.6 0.4 4 On g 0.0 0.2 4 l

-0.4 -! '.' ..f.......

{

g -0.6 i

-0.8-!

(

t, 1.0 i i.2 4

{N 1

1.4 -i

TOTAL \ l

,' 1.6 i = = = = DOPPLER j g MODERATOR t I 1.8 .: "...~.*"- SCRAM j*

j 2.0 ,

0 1 2 3 4 5 6 I

TIME (SEC)

FIGURE 7.4.1 I

(Continued)

MSIV CLOSURE, FLUX SCRAM, EOFPL18 TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME m ,

i I

l FINAL LICENSING CASE fl42,4x4 MODE-2,2 RESPONSE OF RBM CHANNEL A (A+C)

ERROR ROD 34 23

,,, i e i i i 1.3 -  ! U

• h""I M*** l -

1.25 - - .

1.2 - -

l4 x 1.15 - -

x-1.1 - -

NOTE:

1. An intercepts are determined

• 1.06 - -

1.0 , , , , ,

0 8 16 24 32 40 48 ERROR ROD POSITION FIGURE 7.5.1 VY CYCLE 18 RWE CASE 1 - SETPOINT INTERCEPTS DETERMINED BY THE A AND C CHANNELS nano Ii I

I FINAL LICENSING CASE fl42,4x4 MODE-2,2 RESPONSE OF RBM CHANNEL B (B+D)

ERROR ROD 34-23 i i e i e

33_

g,3 l 0 No instrument Failuree l 1.26 -

1.2 -

'~

~

1.16 -

r

/

1.1 - M

1. NOTES:

' I 1. RSM Segmint internoptle marted f .

• *th ist- g 8 107 -

1.06 - 2. Rod is sagged at rumn cosawing

g , '". hohee of Fr Rod uomen,

,, 1os l

,,o i 1 0 3 g'6 24 32 4o 4g

• ERROR ROO PostTiON i

I I

FIGURE 7.5.2 VY CYCLE 18 RWE CASE 1 - SETPOINT INTERCElrFS DETERMINED BY THE B AND D CHANNELS I

I

i j

8.0 DESIGN BASIS ACCIDENT ANALYSIS 8.1 Contml Rod Droo Accident Results l

The contml md sequences are a series of rod withdrawal and banked withdrawal instmetions specifically designed to minimize the worths of individual control rods. The sequences are exammed so that, in the event of the uncoupling and subsequent free fall of the rod, the incremental rod worth is acceptable. Incremental md worth refers to the fact that rods beyond Group 2 are banked out of the core and can only fall the increment from full in to the rod drive withdrawal position. Acceptable worth is one which produces a maximum fuel enthalpy less than 280 calories 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.

l The RWM is functionally tested before each startup.

The 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 enors will not create a potential md drop situation above 280 calories per gram [29],[30],[31]. Below 20% power, however, the sequences must be examined for incremental rod worth. This 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 startup, the rods are grouped, as shown in Figures 8.1.1 and 8.1.2, and are pulled in numerical onier.

All the rods in one group are pulled out before the pulling of the next group begins. The rods in the first two gmups are individually pulled from full-in to full-out. Beyond Group 2, the rods are banked out using procedures [32),[33] which reduce the rod incremental worths.

The potentially high worths that occur in pulling the rods in Group 1 are ignored because the reactor is subcritical in Group 1. 'Iherefore, if a rod drops fmm 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.

aan I

The second gmup of mds 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.l'.1 illustrates. For added conservatism, each of the high worth rods in the second gmup were checked; i.e., one at a time, they were assigned to be the first rod pulled. 'Ihis assures that in any sequence the actual worths will always be less than those calculated here.

Only that portion of the control rod worth above the SIMULATE-3 cold critical eigenvalue contributes to the rod drop accident. For conservatism, " critical" was defined as the SIMULATE-3 average cold critical K,g minus 1% AK (reactivity anomaly criteria). 'Ihe results of the Group 2 calculations, as presented in Table 8.1.2, fit under the bounding analysis of References 29 through 31.

Beyorad Group 2, the rods are banked out of the core. This generally limits the incremental worth of a single rod drop; however, virtually all of the pre-drop cases in Group 3 are critical Therefore, the entire dropped rod worth contributes toward the RDA excess reactivity insertion. 'Ihe method used to evaluate Group 3 involved pulling Groups 1 and 2 out and banking Group 3 to varying positions. 'Ihe 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).

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

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

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

nau I I

I

l 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 4

conditions, complies with the LOCA limits specified in 10CFR50. 46.

'Ihe 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[11] and RELAP5YA[38]. 'Ihe base analysis provided the break spectrum and the single failure conditions.

The Reload Cycle analysis provided the verification that the base analysis was valid for the Reload Cycle given changes in the reactivity and the UNIX system configuration. 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 of the input assumptions.

The base analysis was performed for a combination of break size, break location, and single 2

l failure conditions. The break sizes range from 0.05 ft 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. P impact of the Gd O23 on initial volume average temperature and material properties was included. 'Ihe PCT results for the limiting break 0.6 ft2 with loss of DC power was 1778.I'F.

'Ihe Reload Cycle analysis was perfonned for the limiting break size and two single failure conditions. The PCT results for the limiting break 0.6 ft2 with loss of DC power was 1788.9'F which j is a 10.8'F increase in PCT compared to the base analysis results. For the same size break with LPCI injection valve failure, the PCT for the Reload Cycle was 1770.4'F, an increase of 26.2*F compared to the base analysis. '!he Reload Cycle analysis also showed that the break spectrum performed for the

{

base analysis remains valid for the Reload Cycle.

'Ihe combined analysis results, in terms of peak cladding temperature (PCT), are shown in Figure 8.2.1. "Ihe break spectrun PCT results for Reload Cycle were obtained by increasing the base analysis results by the maximum change in PCT from the Reload Cycle analysis,26.2'F. These men _ .______ _ __ __- ____ - _______ ___ - _ ___________ _ _-

I!

2 results show that the limiting break is 0.6 ft 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.

Overall, the calculated peak clad temperatures are well below the 2200'F limit of 10CFR 50.46. 'Ihe  ;

analysis also shows compliance with the other 10CFR 50.46 limits: 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.

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, the LOCA analysis was W analyzed at 1698 MWm power and 87% flow. 'Ihe results showed that the 100% flow case bounded the low flow case.

'Ihe analysis showed that the MAPLHGR limits are not limited by a LOCA. Therefore, the MAPLHOR limits are set based on the thermal-mechanical analysis of the bundle fmm Reference 18.

They are provided in Appendix A for all the fuel types in the Reload Cycle, as a function of average planar exposure. 'Ihe analysis also verified that the single loop MAPLHGR multiplier,0.83,is valid for the Reload Cycle.

8.3 Refueling Accident Results If any assembly is damaged during refueling, then a fraction of the fission product inventory could be released to the environment. The source term for the refueling accident is the maximum gap I

activity within any bundle. 'Ihe source term includes contributions from both noble gases and lodines.

'Ihe calculation of maximum gap activity is based on the MAPLHGRs, tim maximum operating fuel centerline temperatures, and maximum bundle bumup.

l The fuel rod gap activity,intemal pressure and centerline temperature for the Reload Cycle are bounded by the values used in Section 14.9 of the FSAR(39).

\

l l non I I

TABLE 8.1.1 CONTROL ROD DROP ANALYSIS - ROD ARRAY PULL ORDER The order in which 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 Out Amy1 Array 2 Arrays 3 or 4 Amy2 Array 1 Arrays 3 or 4 Amy3 Amy4 Amys 1 or 2 Amy4 Amy3 Arrays 1 or 2 f

TABLE 8.1.2 VY CYCLE 18 CONTROL ROD DROP ANALYSIS RESULTS Maximum Incremental Rod Worth Calculated 0.80% AK Cold, Xenon-Free Bounding Analysis Worth for Enthalpy Less than 1.20% AK 280 Calories per Gram [29],[30],[31]

men I

TABLE 8.2.1 LOCA ANALYSIS ASSUMPTION 1 l

1698.3 Core 'Ihermal Power (MWm) i 48.0 l Total Core Flow (10$bA)

Reactor Vessel Pressure (psla) 1067.0 Recirculation loop Flow (l(fibA) - Each IAop 12.3 6.93 Feedwater Flow (10$bA)

Feedwater Temperature ('F) 377.0 l Water Level Above Top of Enriched Fuel (in.) 130.0 l

Containment Drywell Pressure (psla) 16.5 I

i 14.7 Containment Wetwell Pressure (psia)

Containment Wetwell Liquid Temperature (*F) 165.0 l

Maximum Bundle Power (MWm) 7.3 l Maximum Average Planar Linear Heat Generation Rate (kW/ft) 13.6*

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

• I I

Plus Calorimetric and TIP Reading Uncertainties (8.9%)

Plus Calorimetric and TIP Reading Uncertainties (9.2%)

mes3 I I

n- . - _ . - . . .. - - _. . - .

l 19 3 M 2 1 1 2 M '

4 3 4 3 4 31 1 2 2 1 27 3 4 3 4 3 +

l 23-- 1 2 1 1 2 1 19-.- 3 4 3 4 3 in 1 2 2 1 11 4 3 4 3 4 07 2 1 1 2 i

@ 3 02 06 10 '4 18 22 26 30 34 38 42 I

f t

FIGURE 8.1.1  !

l FIRST FOUR ROD ARRAYS PULLED IN THE A SEOUENCES i

~

_ .s2 ,

J h

. . _. .. . . . =.

I I

I~

49 3 3 39 2 1 2 SE 3 4 4 3 31 2 1 2 1 2 27 3 4 3 3 4 3 23 - 1 2 1 2 1 19 - 3 4 3 3 4 3 in 2 1 2 1 2 11 4 4 3 3

07 2 1 2

@ 3 3 02 06 10 14 18 22 26 30 34 38 42 l

I m O U R E 8. m I

HRST FOUR ROD ARRAYS PULLED IN THE B SEOUENCES 1

l l

I

!. men l I l

I

1 2400 , .

Fkelrc.Ploe LowPressure : Discharee  :

Break, Location CSCS Credited

• N Appen&c K Umt -

2200_ . .

• m.6 ,e awC j t0 j 2000 -_::5. g--m----j- .

,3- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

u_ -

l l x DI.6.,e aLPCS . 0.8 a1800 as - --------------------I------------------;.--------------------

O -

l l-v A DI. der ,e ILPCS

• EPCI l 10

: a 32)1600 - ------------------ -------- ----- -----

m u

3.--------

3.-

as . .

a . .

E 18 -------------------- '

as . ------------------i.-------'------------

F- l l 1200 -- - - - - - - - - - - - - - - - - - - ' - - - - - - - - - - - - - - - - - - '--------------------

1000 - ------------------- =

800 . . . . . . ..; . . . ..... . . . . . . . .

U.01 0.1 1 10 Break Size (ft2) ,

11GURE 8.2.1 LOCA ANALYSIS RESULTS. PEAK CLADDING TEMPERATURE VERSUS BREAK SIZE mess  !

9.0 STARTUP PROGRAM l

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

The Vennont k'ankee Startup Program will include process computer data checks, shutdown margin demonstration, in-sequence critical measurement, rod scram tests, power distribution .

comparisons, TIP reproducibility, and TIP symmetry checks. 'De content of the Startup Test Report  ;

will be similar to that sent to the Office of Inspection and Enforcement in the pastI40].

I I

I I

I i

I I

men I I

I

4 1

10.0 CONCLUSION

i i This report presented the design information, calculational results, and operatmg limits pertinent to the operation of the Reload Cycle. The core is designed to consist of 120 new GE-9B fuel bundles and 248 irradsated GE-9B fuel bundles. The shutdown margin for the Reload Cycle is 3 ]

greater than the Technical Specification limit of 0.32% AK.1he 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. The 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, and 3) demonstrated compliance with the ASME vessel code limits. The contml 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 i gram. The LOCA analysis demonstrates compliance with the acceptance criteria specified in

)

10CFR50.46. The fuel md gap activity, intemal pressure and centedine temperature are bounded by r

4 the values used in Section 14.9 of the PSAR which demonstrates the limits of 10CFR100 are not

! exceeded for a refueling Eddant-4 i

a t

f 1 I 4 ,

l l

4 9

1

amm {

i i I

I REFERENCES

1. K. J. Morrissey, Vennont Yankee Cycle 16 Summary Report YAEC-1878 (April 1994).

2. General Electric Company, General Electric Standard Apolication for Reactor Fuel (GESTARID. NEDE-24011-P-A 10, GE Company Proprietary February 1991, as amended.

3. Ixtrer, D. T. Weiss to J. M. Buchheit, "GE9B Bundle Nuclear Design Reports for Reload 14,"

DTW89168. October 6,1989.

4. Letter, D. T. Weiss to R. T. Yee, " Fuel Bundle Nuclear Design Reports for Vermont Yankee Reload 16," IyrW92194, September 3,1992.

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 Benchmarking of Vermont Yankee Cycles 9 through 13; YAEC-1683-A (March 1989).

~

7. B. Y. Hubbard, et al.; End-of-Full-Power-Life Sensitivity Study for the Revised BWR Licensing Methodolory: YAEC-1822 (October 1991).

8. VYNPC Ixtter to USNRC, " Vermont Yankee LOCA Analysis Method: FROSSTEY Fuel

. Performance Code (FROSSTEY-2)," FVY 87-116 (December 16,1987).

9. VYNPC Letter to USNRC, " Response:. to Request for Additional Infonnation - FROSSTEY-2 Fuel Performance Code," BVY 91-024 (March 6,1991).

10. VYNPC 12tter to USNRC, "FROSSTEY-2 Fuel Perfonnance Code - Vermont Yankee Response to Remaining Concems," BVY 92-54 (May 15,1992).

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

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

13. VYNPC Letter to USNRC, " Inverted Control Rod Poison Tubes at Vermont Yankee," WVY

, 75-51 (May 16,1975).

14. USNRC Letter to G. C. Andognini, " Change to Bases," (June 6,1975).

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

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

aan I I

EEEEEENEF4 (Condnued)

17. A. A. F. Ansad, R. R. Gay, and B. J. Gitnick; FIBWR: A Steady-State Core Flow Distributic n Code for Boillne Water Reactors - Code Verification and Oualification Report: EPRI NP-1923; Project 1754-1 Final Report, July 1981.

18. USNRC letter to J. B. Sinclair, SER, " Acceptance for Referencing in Licensing Actions for the Vennont Yankee Plant of Reports: YAEC-1232,YAEC-1238, YAEC-1299P, and YAEC- ,

1234," NVY 82-157 (September 15,1982).

l

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

20. Letter, D. T. Weiss to R. T. Yee, " Mechanical MAPLHGRs for Vermont Yankee Reload 16," l DTW93136, June 3,1993.  !

21. Letter, D. T. Weiss to R. T. Yee, " Fuel Rod 'Ihermal-Mechanical Performance Limits,"

DTW92260, November 19,1992.  ;

22. A. A. F. Ansad and J. T. Cronin, Methods for the Analysis of Boiling Water Reactors: A Systems Transient Analysis Model GETRAN). YAEC-1233, (April 1981).

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

(November 27,1981).

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

25. Electric Power Research Institute, RETRAN - A Program for One-Dimensional Transient Thennal-Hydraulle Analysis of Complex Fluid Flow Systems. CCM-5, December 1978.

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

27. A. A. F. Ansad, K. J. Bums, and D. K. Beller, Methods for the Analysis of Boiling 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 Large Boiling Water Reactors: NEDO-10527; March 1972.

mem .

l l

REFERENCES (Cbntinued)

30. General Electric Company; R. C. Stim, et al.; Rod Droo Accident Analysis for Large Boiling

! Water Reactors Addendum No.1. Multiole Enrichment Cores with Axial Gadolinium: NEDO-10527, Supplement 1; July 1972.

31. General Electric Company; R. C. Stim, et al.; Rod Droo Accident Analysis for Large Bolling Water Reactor Addendum No. 2 Exposed Cores: NEDO-10527, Supplement 2; January 1973.

32. General Electric Company, C. J. Paine Banked Position Withdrawal Secuence. NEDO-21231, q

January 1977.

33. General Electric Company, D. Radcliffe and R. E. Bates, Reduced Notch Worth Procedure.

. SIL-316, November 1979.

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

i

35. USNRC Letter to L. A. Tremblay, SER, " Safety Evaluation for Vennont Yankee Nuclear Power Station RELAP5YA LOCA Analysis Methodology (TAC No. M74595)," NVY 92-192 (October 21,1992).

36. USNRC 12tter to R. W. Capstick, SER, " Approval of Use of Thennal-Hydraulic Code RELAPSYA (TAC No. 60193)," NVY 87-136 (August 25,1987).

! 37. L Schor, et al.; Vermont Yankee less-of-Coolant Accident Analysis; YAEC-1772 (June 1993).

38. Report. RELAPSYA. A Computer Program for Light-Water Reactor System Thennal. Hydraulic Analysis. YAEC-1300P-A, Revision 0, October 1982; Revision 1, July 1993.

39. Vermont Yankee Nuclear Power Station Final Safety Analysis Report, December 1991.

40. Letter from L. A. Tremblay to USNRC, " Cycle 17 Startup Test Report," BVY 94-03 (January 13, 1994)..

41. Letter, P. J. Savola, GE, to R. T. Yee, " Transmittal of Modified Thermal Mechanical MAPLHGR Limits for Vennont Yankee Cycle 18 Loss of Stator Cooling Event," PJS 95106, dated July 25,1995.

I I

man I I

APPENDIX A .

CALCULATED OPERATING LIMTTS

'Ihe MCPR operating limits for the Reload Cycle are calculated by adding the calculated ACPR to the FCISL. 'Ihis is done for each of the analyses in Section 7.0 at each of the exposure statepoints. For an exoosure interval between statepoints, the highest MOR 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.1hese MCPR operating limits are valid for operation of the Reload Cycle at full power up to 10644 mwd /St and for operation during coastdown beyond EOFPL.

Tables A.2 through A.5 provide the most limiting calculated MAPLHGR limits for all the fuel types in the Reload Cycle. 'lhese values bound the lattice-specific MAPLHGR limits for all the' enriched lattice zones in each fuel type.1he MAPLHGR limits were revised for the LOSC transient results[41).

l k

l I

4 I

1 men .n l

I TABLE A.1 VERMONT YANKEE NUCLEAR POWER STATION CYCLE 18 MCPR OPERATING LIMITS Value of "N" in Average Control Rod Cycle Exposure Ranne MCPR Operating RBM Eaustion! Scram Time Limit 2.3 Equal to or 0.0 to 4000 mwd /St 139 42% better than 4000 to 5500 mwd /St 135 g L.C.O. 33.C.1.1 5500 to 10644 mwd /St 133 m Equal to or un so ouuu mwet IJv 4000 to 5500 mwd /St 135 better than E L.C.O. 33.C.1.2 5500 to 9035 mwd /St 133 5 9035 to 10644 mwd /St 134 Equal to or un to 4uuu mwet Iav g i 41% better than 4000 to 5500 mwd /St 135 g L.C.O. 33.C.1.1 5500 to 6500 mwd /St 1.29 6500 to 9035 mwd /St 1.27 9035 to 10644 mwd /St 132 Equal to or un to 4uuu mwet 139 better than 4000 to 5500 MW4/St 135 g L.C.O. 33.C.1.2 5500 to 6500 MW4/St 1.29 3 6500 to 8035 MW4/St 1.27 8035 to 9035 mwd /St 130 9035 e 10644 mwd /St 134 Equal to or unsosuuu mw e: 1av s40% better than 4000 to 5500 MW4/St 135 L.C.O. 33.C.1.1 5500 to 6500 mwd /St 1.29 6500 to 8035 mwd /St 1.25 8035 to 9035 mwd /St 1.27 9035 to 10644 mwd /St 132

' Equal to or um so 4uuu m w e t 4av better than 4000 to 5500 mwd /St 135 L.C.O. 33.C.1.2 5500 to 6500 mwd /St 1.29 6500 to 8035 mwd /St 1.25 8035 to 9035 MW4/St 130 9035 to 10635 mwd /St 134 (1) 'Ihe 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 g fuel types. On this basis, if any of these fuel types are to be reinserted, they will be evaluated in 5 accordance with 10CFR50.59 to ensure that the above limits are bounding for these fuel types.

(3) MCPR operating limits should be increased by 0.01 for the single loop operation.

aen I I

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

i TABLE A.2 .

MAPLHOR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB311-10GZ 4

) Plant: Vennont Yankee Fuel Type: BP8DWB311-10GZ 1

1 l Av.trandanatE2nosa MAPLHGR Limits &W/fu >

. (mwd /St) Two-Imo Operation Single-Looo Operation

• 0.00 10.93 9.07 200.00 11.00 9.13 j 1,000.00 11.13 9.24 2,000.00 11.32 9.40 3,000.00 11.52 9.56 4,000.00 11.64 9.66 5,000.00 11.77 9.77 i 6,000.00 11.92 9.89 .

i ,

7,000.00 12.11 10.05 8,000.00 12.34 10.24 f

9,000.00 12.59 10.45 10,000.00 12.83 10.65 L

12,500.00 13.00 10.79 15,000.00 12.81 10.63 20,000.00 12.24 10.16 25,000.00 11.55 9.59 4

35,000.00 10.2d 8.50 45,000.00 8.76 7.27 50,735.00 5.91 4.91 l MAPLHGR limits for single-loop operation are obtained by multiplying the two-loop operation  :

MAPLHGR limits by 0.83.

aan i l

l I, 1

TABLE A.3 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB311-110Z t

)

Plant: Vennont Yankee Fuel Type: BP8DWB311-11GZ Average Planar Exposure MAPLHGR Limits (kW/ft) 04Wd/St) Two-Looo Oceration Single-Imoo Operation' 0.00 10.93 9.07

200.00 11.00 9.13 1,000.00 11.13 9.24 I 2,000.00 11.32 9.40 3,000.00 11.52 9.56 4,000.00 11.64 9.66 5,000.00 11.77 9.77 6,000.00 11.92 9.89 7,000.00 12.11 10.05 8,000.00 12.34 10.24 9,000.00 12.59 10.45 10,000.00 12.83 10.65 12,500.00 13.00 10.79 i 15,000.00 12.81 10.63 20,000.00 12.24 10.16 l

25,000.00 11.55 9.59 35,000.00 10.24 8.50 45,000.00 8.76 7.27 50,735.00 5.91 4.91 MAPLHGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLHGR limits by 0.83.

aan .

I

TABLE A.4 MAPLHOR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-10GZ Plant: Vennont Yankee Fuel Type: BP8DWB335-10GZ Averane Planar Exposure MAPLHGR Limits (kW/fD (mwd /St) Two-Looo Operation Single-Loop 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 13.24 10.99

. 10,000.00 12.99 10.78

! 12,500.00 12.84 10.66 15,000.00 12.65 10.50 20,000.00 11.93 9.90 25,000.00 11.26 9.35 35,000.00 9.88 8.20 45,000.00 8.38 6.%

50,593.00 5.65 4.69 MAPLHGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLHGR limits by 0.83.

i

. 94

i l

TABLE A.5 ,

MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-11GZ Plant: Vermont Yankee Fuel Type: BP8DWB335-11GZ Average Planar Exoosum MAPLHGR Limits (kW/ft)

(mwd /SO Two-Imo Operation Single-Imoo Operation

• 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 9.79 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 l 9,000.00 13.13 10.90 i 10,000.00 12.99 10.78

! 12,500.00 12.84 10.66 15,000.00 12.65 10.50 20,000.00 11.93 9.90 25,000.00 11.26 9.35 35,000.00 9.88 8.20 i

45,000.00 8.38 6.96 50,593.00 5.65 4.69 MAPLHGR limits for single-loop operation am obtained by multiplying the two-loop operation MAPLHGR limits by 0.83.

aan I