ML20010F459

Cycle 9 Core Performance Analysis.

ML20010F459
Person / Time
Site:	Vermont Yankee
Issue date:	08/31/1981
From:	Ansari A, Burns K, Candon J VERMONT YANKEE NUCLEAR POWER CORP.
To:
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ML20010F453	List: ML20010F452 ML20010F457 ML20010F459
References
YAEC-1275, NUDOCS 8109100237
Download: ML20010F459 (105)

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I e Vermont Yankee Cycle 9 I

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Core Performance Analysis August 1981

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I Major Contributors:

A. A. F. Ansari M. J. Hebert K. E. St. John I K. J. Burns J. D. Candon J. M. Holzer D. M. Kapitz M. A. Sironen D. M. VerPlanck J. T. Cronin E. E. Pilat R. A. k'oehlke R. Habert I

Approved by: 8 .5'/ /

R. J.Caccia[uti, Manager / (Da'te)

Reactor Physics Group

. Approved by:

S. P'. Sch ul t z , TanagerN

))31f$l(Date)

BWR Transient Analysis Group Approved by: ctaJ2._ M , 5/ / P/

A. Husain, Manager [/ / (Da t'e )

LOCA Group Approvec by: _

B. C. Sli r, Manager (Date)

Nuclear Engineering l

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DISCIAIMER OF RESPONSIBILITY I This document was prepared by Yankee Atomic Electric Company for its own use and on behalf of Vermont Yankee Nuclear Power Corporation. This document is believed to be completely true and accurate to the best of our knowledge and informa tion. It is authorized for use specifically by Yankee I Atomic Electric Company, Vermont Yankee Nuclear Power Corporation and/or the appropriate subdivisions within the Nuclear Regulatory Commission only.

With regard to any unauthorized use whatsoever , Yankee Atomic Electric Company, Vermont Yankee Nuclear Power Corpora tion and their o fficers ,

directors, agents and employees assume no liability nor nake any warranty or representa tion with respect to the contents of this document or to its accuracy or completeness.

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I ABSTRACT

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This report presents design informa tion and calculational analysis results pertinent to the operation of Cycle 9 of the Vernont Yankee Nuclear Power Sta tion. These include the fuel design and core loading pattern descriptions; calculated reactor power distributions , power peaking, shutdown capability and reactivity functions; and the results of the several safety analyses per formed to jus ti fy plant operation throughout Cycle 9.

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I I TABLE OF CONTENTS Page ii D I S C lA I M E R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

I ABSTRACT..................................................

TAB lE OF C0 N TEN T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

iii iv vi LI S T O F F IGU RES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

I viii LIST OF TABLES............................................ ix A CK N 0WlE GEME NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 1.0 INTRODUCT0N...............................................

2 2.0 RECENT REACTOR OPERATING HIST 0RY..........................

2 2.1 Opera ting His tory o f the Curr ent Cycle . . . . . . . . . . . . . . .

2 2.2 Operating History of Recent Applicable Cycles .......

4 3.0 RELOAD CORE D ES IGN D ES CRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4 3.1 C or e Fue l Loa d i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4 3.2 Design Re f erence Core Loa ding Pa t tern . . . . . . . . . . . . . . . 4 3.3 As s embl y Exposur e an d Cycle 8 His tory . . . . . . . . . . . . . . .

8 4.0 FUEL MECHANICAL AND THERMAL DESIGN . . . . . . . . . . . . . . . . . . . . . . .

8 4.1 Mechanical Design ...................................

8 4.2 Thermsl Design ......................................

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

15 5.0 NU C LE AR D ES IG N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

15 5.1 Core Power Distributions ............................ 15 5.1.1 Haling Power Distribution ....................

15 5 .1. 2 Rodded De pletion Power Dis tr ibution . . . . . . . . . .

16 5.2 Core Exposure Distributions ........................ 17 5.3 Col d Core Rea ctivity and Shutdown Margin . . . . . . . . . . . .

17 5.4 Standby Liquid Control System Shutdown Capability ...

...... 27 6.0 THERMAL-HYDRAU LIC D ES IGN . . . . . . . . . . . . . . . . . . . . . . . . . .

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I I TABLE OF CONTENTS (Con tinued )

Page 7.0 A C C ID EN T AN A LYS I S , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.1 Cor e Wide Tr an s ie n t Anal ys is . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.1.1 Me th o d o l o gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 I

7.1.2 Initial Conditions and Assumptions . . . . . . . . . . . . 31 7.1.3 Reactivity Functions ......................... 32 7.1.4 T r an s ie n t s An a l yz ed . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 7.2 Core-Wide Transient Analysis Results ................ 35 7.2.1 Turbine Trip Without Bypass Transient ........ 35 7.2.2 Generator Load Rejection Without I 7.2.3 Bypass Transient..............................

Loss of Fecdwater Hea ting Transient 7.3 Over pr essur iza tion Anal ysis Re sul ts . . . . . . . . . . . . . . . . .

36 37 38 7.4 Local Rod Withdrawal Error Transient Results ........ 38 I 7.5 Misloaded Bundle Error Analysis 7.5.1 Rota ted Bundle Errcr Analysis Results ........

Mislocated Bundle Error Analysis Results .....

42 42 44 7.5.2 I 7.6 7.7 Control Red Drop Accident Results ...................

Stability Analysis Results ..........................

45 46 8.0 STARTUP PROGRAM .......................................... 89 9.0 LOS S-OF-COOLANT ACCIDENT ANALYS IS . . . . . . . . . . . . . . . . . . . . . . . . 90 APPENDIX A PLANT TECHNICAL SPECIFICATION CHANGES ........ 91 REFERENCES ............................................... 94 I

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l LIST OF FIGURES Number Title Page j

3.2.1 Design Reference Loading Pattern, Northeast Quadrant 7 4.2.1 Core Averaged Gap Conductance versus Cycle Burnup 13 4.2.2 Core Averaged Volume Average Temperature versus Cycle Burnup 14

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5.1.1 VY Cycle 9 Haling Depletion EOC Bundle Average Relative Powers 20 5.1.2 Cycle 9 Core Average Axial Power Distribution Taken from the Haling Calculation to EOFPL9 21 I

l 5.1.3 VY Cycle 9 Rodded Depletion - ARO at EOFPL9, Bundle Average Rela tive Powers 22 l 5.1.4 Cycle 9 Core Average Axial Power Distribution, Rodded Depletion - ARO at EOFPL9 23 l

5.2.1 VY Cycle 9 Haling Depletion, EOC Bundle Average Exposures 24 5.2.2 VY Cycle 9 Rodded Depletion, EOC Bundle Average Exposures 25 VY Shutdown Margin for Cycle 9, Cold Percent Shutdown I 5.3.1 Del ta K Ver sus Cycle Average Exposure 26 7.1.1 Flow Chart for the Calcula tion of ACPR Using I 7.1.2 RETRAN/MAYU4 Codes Inserted Rod Worth and Rod Position versus Scram Time 53 at EOC, " Measured" Scram Time 54 7.1.3 Inserted Rod Worth and Rod Position versus Scram Time at EOC-1000 MWD /ST, " Measured" Scram Time 55 7.1.4 Inserted Rod Worth and Rod Position versus Scram Time at EOC-2000 MWD /ST, " Measured" Scram Time 56 7.1.5 Inserted Rod Worth and Rod Position versus Scram Time at EOC, "67B" Scram Time - 57 7.1.6 Inserted Rod Worth and Rod Position versus Scram Time at EOC-1000 MWD /ST, "67B" Scram Time 58 7.1.7 Inserted Rod Worth and Rod Position versus Scram Time at EOC-2000 MWD /ST, "67B" Scram Time 59 I -vi-I I i

I LIST OF FIGURES I Number Title Page 7.2.1 Turbine Trip Without Bypass, EOC Trans ien t Response versus Time, " Measured" Scram Time 60 I 7.2.2 Turbine Trip Without Bypass, EOC-1000 MWD /ST Transient Response versus Time , " Measured" Scram Time 63 7.2.3 Turbine Trip Without Bypass, EOC-2000 MWD /ST Transient Response versus Time , " Measured" Scram Time 66 7.2.4 Generator Load Rejection Without Bypass, EOC Transient Response versus Tiin , " Measured" Scram Time 69

-I 7.2.5 Generator Load Rejection Without Bypass, EOC-1000 MWD /ST Transient Response versus Time , " Measured" Scram Time 72 7.2.6 Generator Load R2jection Without Bypass, EOC-2000 MWD /ST Transient Response versus Time , " Measured" Scram Time 75 7.2.7 Loss of Feedwa ter Hea ting, EOC-2000 MWD /ST Transient Response versus Time 78 7.3.1 MSIV Closure, Flux Scram 80 Transient Response versus Time , " Measured" Scram Time 83 7.4.1 Reactor Initial Conditions for Rod Withdrawal Error Case 1 84 7.4.2 Reactor Initial Conditions for Rod Withdrawal Error Case 2 7.4.3 Case 1 - Setpoint Intercepts Determined by the A+C 85 RBM Channel 7.4.4 Case 1 - Setpoint Intercepts Determined by the B+D 86 RBM Channel First Four Rod Arrays Pulled in the A Sequences 87 7.6.1 First Four Rod Arrays Pulled in the B Sequences 87 7.6.2 I 7.7.1 Reactor Core Decay Ra tio ver sus Power 88 I

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I LIST OF TABLES i

Number Title Page 3.1.1 Cycle 9 Fuel Bundle Types and Numbers 6 3.3.1 Desita Basis Cycle 8 and Cycle 9 E.xposures 6 4.1.1 Noninal Fuel Mechanical Design Parameters 11

'4. 2.1 Pe.ak Linear Heat Generation Rates Corresponding I to Incipient Fuel Centerline Melting and 1% Clad Plas tic Strain 12 5.3.1 K-E f fe c tive Values , Shu tdown Margin Calcula tion 19 5.4.1 Standby Liquid Control Sys tem Shutdown Capability 19 7.1.1 Summary of Sys tem Transient Model Initial Condition for Core Wide Transient Analyses 47 7.1.2 Transient Analysis Reactivity Coe fficients 48 7.2.1 Core Wide Transient Analysis Results 49

7. s .1 Rod Withdrawal Error Transient Summary (With Limitieg Instrument Failure) 50 7.5.1 Rota ted Bundle Analysis Results 51 7.6.1 Control Rod Drop Analysis - Rod Array Pull Order 52 7.6.2 Control Rod Drop Analysis Results 52 A.1 Cycle 9 MCPR Operating Limits 92 A.2 MAPLHGR Opera ting Limits Versus Average Planar Exposure 93 I

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i ltl ACKNOWLEDGEMENTS il jg 4

The authors and principal contributors would like to express th eir jg gra titude in the acknowledgement of contributions to this work by B. G. Baharynejad , K. G. Gavin , Q. A. Haque , J. Pappas , and R. K. Sch mi d t .

. Their assistance in performing the required eniculations, documenting the resul ts , and preparing figures and displays for th is document is recognized i

! and gr ea tl y a ppre c ia ted . Special apprecia tion is expressed to S. M. Henchey I

and D. L. Nich ols for preparation of the draft and final manuscript.

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1.0 INTRODUCTION

I This report provides jus ti fication to support the operation of the Vermont Yankee Nuclear Power Sta tion through the forthcoming fuel reload cycle (Cycle 9). The refueling preceding Cycla 9 will involve the discharge of 120 irradia ted fuel bundles and the insertion of 120 new fuel bundles. The resultant reload cycle will consist of 120 new fuel bundles of the pressurized retrofit 8X8 design (P8DPB289), 176 irradia ted fuel bundles of the pressurized retrofit 8X8 design, 60 irradiated bundles of the retrofit 8X8 Jesign (8DPB289), and 12 irradiated bundles of the earlier 8X8 design (8D274). All fuel bundles for Cycle 9 operation have been fabricated by General Electric (GE). The introduc tion of the new fuel is necesary in order to maintain sufficient reactivity for continued operation at full rated power.

I This report contains descriptions and analyses results pertaining to the mechanical, thermal-hydraulic, physics , and sa fety analyses aspects of the Cycle 9 r eload .

I The cycle de pendent parameters and operating limits to be incorporated into the Vermont Yankee Plant Technical Specifica tions for Cycle 9 are given in Appendix A.

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I I 2.0 RECENT RFACTOR OPERATING HISTORY 2.1 Opera ting His tory of the Current Cycle I The current operating cycle is Cycle 8. The reactor was started up for th i's cyc le on De c emb er 24, 1980 and is projected to be shut down for re fueling on October 17, 1981. During this period, control rod sequence exchanges were performed on the following schedule:

I SEQUENCE from to Mar ch 14 , 1981 Al B2 May 11, 1981 B2 A2 Au gus t 1, 1981 A2 B1 I

During the week following the control rod sequence exchange in Mar ch , th e reactor was operated at reduced power to allow special testing including a recirculation pump trip tes t and reactor stability tes ting [1] . To th is time with th ese prov is ions , the reactor has been operated smoothly and uninterrupted at full power with the exception of a scram on May 11, 1981.

I 2.2 Opera ting His ory of Recent Applicable Cycles I Fuel to be re-irradia ted in Cycle 9 includes fuel bundles which were initially inserted into the reactor in Cycle 6, Cycle 7, and Cycle 8. Cycle 7 reactor operation proceeded at full power with normal maintenance and I

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I operational maneuvers with the exception of a 3 day outage in February 1980 to implement plan t modifications required by the NRC (IMI fix). A total of six rea ctor scrams occurred during the cycle opera tions.

Cycle 6 operation was interrupted at about mid-cycle for an outage of approxima tely three weeks to identify and replace failed and suspect fuel rods in several assemblies (Mar ch 17 to Apr il 3, 197 9) . The reactor operated at reduced power levels (v90% ra ted power) for two weeks prior to the outage.

Operation prior to this outage has been designated Cycle 6. Subsequent to the return to power on April 3, coolant activity levels were at normally low levels and full power operation proceeded through the end-o f-cycle. This I period of opera tion has been designated Cycle 6A.

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I 3.0 RELOAD CORE DESIGN DESCRIPTION 3.1 Core Fuel Loading Reload 8 (Cycle 9) will discharge 120 spent fuel assemblies out of a core total of 363. Thus the Cycle 9 core will consist of 120 new assemblies and 248 irradia ted assemblies. All assemblies have bypass flow holes in the lower Table 3.1.1 characterizes the core by fuel type, batch size, and tie pla te.

first cycle loaded. A description of each fuel type is found in Reference 2.

I 3.2 Design Reference Core Loading Pa ttern The location of the assemblies in Cycle 9 is indicated by the map in Figure 3.2.1. For the sake of legibility only the northeast quadranc is I shown. The other quadrants are mirror images with bundles of the same type having nearly identical exposure histories. The new bundles have been identified as R8XX. Similarly, irradiated bundles are designated by the reload number in which they were first introduced into the core. When the reloading is completed , specific bundle ID's will appear in these locations.

Ch anges to the loading pa ttern at the time of refueling will be checked and verified accept:ble under 10CFR50.59. The final loading pattern will be supplied with the Startup Test Report.

I 3.3 Ajsembly Exposure and Cycle 8 History The exposure on the fuel bundles in the design reference loading pattern l

is given in Figure 3.2.1. To obtain this exposure distribution, previous cycles up to Cycle 8 were depleted in the SIMULATE model using actual plant

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, oper ating h is tory [ 3,4 ] . For Cycle 8, plant operating history was used through 5/18/81, tha t is, a core average exposure of 13.641 CWD/ST. Beyond 5/18/81 the exposure was accumulated using a bes t-es timate rodded depletion anal ys is to the end of Cycle 8.

- Table 3.3.1 gives the assumed minimum burnup on Cycle 8 and the BOC9 exposure that results from the shu f fle. In this table, as in th e r es t o f th i s report, the terms "End of Cycle (EOC)" and "End of Full Power Life (EOFPL)",

as applied to Cycle 9, are used interchangeably.

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I TABE 3.1.1 CYCE 9 FL'EL BUND 1E TYPES AND NUMBERS i

1 Fuel Cycle Span of Possible Desi gna tion Loa ded Number Bundle ID's l IRRADIATED 8DB274H 6 12 LJ7057-LJ7133 8DPB289 6 60 LJB710-LJB769 LJG930-LJG999 I

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P8DPB289 7 70 P8DPB289 7 24 LJH001-LJH024 PSDPB289 7 2 LJL746 and LJL747 P8DPB289 8 78 LJP191-L.P268 P8DPB289 8 2 LJU719 and LJU720 NEW P8DPB289 9 120 Not yet specified I

TABLE 3. 3.1 DESIGN BASIS CYCLE 8 AND CYCLE 9 EXPOSURES lI j

Assumed previotis cycle core average exposure End of Cycle 8 16.67 GWD/ST I

Assumed reload cycle core average exposure Beginning of Cycle 9 9.02 CWD/ST Haling calculated core average exposure at End of Cycle 9 (Excluding Coas tdown) 17.06 GWD/ST lI l

Cycle 9 Capability to EOFPL9 8.04 GWD/ST i

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VERMONT YANKEE I CYCLE 9 BOC SUNOLE AVERROE EXP03URES I

RSX2 R7XX R5X2 RSXX R5X2 R7XX R5X2 R7XX R6XX R8XX R7XX 19.57 5 33 20.01 0.00 20.21 5 72 20.57 8 74 14.23 0.00 7.40 R7XX RSX1 R8XX R6XX RSXX RSXX R8XX R8XX R8XX R7XX RSX2 5.45 15 58 0.00 12.62 0.00 13 20 0.00 13.41 0 00 7 39 22.21 I RSX2 R8XX R6XX RSXX R6XX RSXX R6XX R8XX R6XX RSXX 0 00 R5XI 21 87 20 12 0 00 12.68 0 00 12 52 0 00 13 03 0 00 14 35 R8XX RSXX R8XX R6XX R8XX RSXX R8XX R7XX R8XX R5X2 0 00 12 70 0 00 12 84 0.00 12 93 0.00 8 68 0 00 20.61 R5X2 R8XX RSXX RSXX R6XX R7XX RSXX R8XX R7XX 20.27 0 00 12 60 0 00 12.36 6.27 13.82 0.00 7.64 R7XX R6XX R8XX R8XX R77X R6XX R8XX R7XX R5X2 5.78 13 42 0 00 13 08 8.39 12 38 0.00 8 73 20 26 R5X2 R8XX RSXX R8XX MSXX R8XX RSXX R7XX RSX2 20.87 0.00 13.32 0.00 13.98 0.00 14.55 7 48 22.85 R7XX R6XX R8XX R7XX R8XX R7XX R7XX R6XX I 6 52 RSXX 13.69 R8XX 0 00 RSXX 8.98 RSXX 0 00 R7XX 6 92 R5X2 7 64 RSX2 14.36 14.58 0.00 14.47 0.00 8.10 20.54 21.79 R8XX R7XX R8XX R5X2 RSX1 - 808274H. RELOAD 5 0 00 7 83 0 00 20.89 I .

R5X2 - 80P8289. RELORD 5 R7XX R5X2 R5X1 FUEL TYPE 10 RSXX - P80P8289. RELOAD 8 I 7 51 22.58 21.97 EXPOSURE (OWD/8T)

R7XX - P80P8289. RELOAD 7 R8XX - P80P8289. RELORD 8 FIGURE 3.2.1 DESIGN REFERENCE CORE LOADING PATTERN, NORTHEAST QUAD 9 ANT g .,-

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I 4.0 FUEL MECHANICAL AND THERMAL. DESIGN i 4.1 Mechanical Design I One. hun dred and twenty (120) fresh fuel bundles fabricated by General Eiectric Co. will be inserted into the Vermont Yankee reactor for Cycle 9 o per a tion . The mechanical design parameters are identical to the General Electric fabricated bundles which were first inser ted and irradiated during Cycles 7 and 8. Table 4.1.1 identifies the design parameters for all irradiated and fresh fuel bundle types in Cycle 9.

I Further descriptions of the fuel rod mechanical design and mechanical design analyses are provided in Reference 2. These design analyses remain valid with respect to Cycle 9 ructor operation. Mechanical ud chemical compa tib ility of the fuel assemblies with the in-service reactor environment is also addressed in Re ference 2.

I 4.2 Thermal Design I The fuel thermal eff ects calculations were per formed using the FROSSTEY compu ter code [ 5-7 ] . The FROSSTEY code calcula tes pellet-to-clad gap conductance and fuel temperatures from a combination of theoretical and empirical models which include fuel and cladding thermal expans ion, fission gas release , pellet swelling, pellet densification, pellet cracking, and fuel and cladding thermal conductivity.

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I The thermal effects analysis included the calculation of fuel temperatures and fuel cladding gap conductance under nominal core s teady sta te and peak linear heat generation rate conditions. Figures 4. 2.1 and 4. 2. 2 provide the core-average response of gap conductance and fuel temperature, r es pect ivel y. These calculations integrate the responses of individual fuel ba.tch average opera ting his tories over the core average exposure range of Cycle 9. The gap conductance values are weighted axially by power dis tribu tions and radially by volume. The core-wide gap conductance values for the RETRAN system sit:ulations deseribed in Seetion 7.1 are from this data set at the particular exposure paint of interest. The fuel temperature values presented in Figure 4.2.2 are weighted axially and radially by volume.

I The gap conductance value input to the RETRAN Hot-Channel /MAYU4 calcula tions was evaluated for the P8X8R fresh fuel bundle type at the peak assembly power to the cycle exposure point of peak bundle reactivity. Gap con du ct an ce calculated at this point was bounded by a value of 1000 2 With consideration for the hot channel transient response BTU /hr-ft - F.

to bundle power level and gap conductance values calculated for all other fuel I 2 types in Cycle 9, a gap conductance value of 1000 BTU /hr-f t - F was u tilized for all RETRAN Hot-Channel /MAYU4 calcula tions at all exposure point .

and for all fuel bundle types.

lI Fuel rod local linear hea t generation rates at fuel centerline incipient mel t were calcula ted as a function of local axial segment exposure for the gadolinia concentrations in Vermont Yankee fuel bundles and are displayed in I

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'I Table 4.2.1. Initial conditions assumed that fuel rods opcrate to the local segment power level of the maximum allowable linear heat generation rate (13.4 kw/ f r ) pr ior to a power increase.

4.3 Opera ting Experience All fuel bundles scheduled to be reloaded in Cycle 9 have operated as expected in previous cycles of Vermont Yankee. Since the Cycle 6/6A outage, offgas measurements are at normally low levels indicating that no fuel failures are present.

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TABLE 4.1.1 I NOMINAL FUEL MECHANICAL DESIGN PARAMETERS FUEL TYPE I

8X8 8X8R P8X8R Fuel Pellets I Fuel Material (sintered Pellets)

Initin1 Enr ichment, UO2 2.74 UO2 2.89 UO2 2.89 I w/o U-235 Pel1et Density,

% theoretical 95.0 95.0 0.410 95.0 0.410 Pellet Diameter , inch es 0.416 Fuel Rod Active Length , inches 144.0 150.0 150.0 9.5 I Plenum Length , inches Fuel Rod Pitch, inch es Diametral Gap (cold),

16.0 0.640 0.009 9.5 0.640 0.009 0.640 0.009 I inch es Fill Gas Fill Gas Pressure, psig Helium Hel ium Helium I Cladding Ma te r ial Outside Diameter , inch es Zr-2 0.493 Zr-2 0.483 Zr-2 0.483 0.034 0.032 0.032 I Th icknes s , inch es Ins ide Diame ter ,

inch es 0.425 0.419 0.419 I Fuel Channel Ma te r ial Zr-4 Zr-4 Zr-4 Inside Dimension, inch es 5.278 5.278 5.278 Wall Thickness , inch es 0.080 0.080 0.080 Fuel Assembly 8x8 8x8 8x8 I Fuel Rod Array Fuel Rods per Assembly Spacer Grid Material 63 Zr-4 62 2r-4 62 Zr-4 I ** GE Proprietary I

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TABLE 4. 2.1 PEAK LINEAR HEAT GENERATION RATES CORRESPONDING TO INCIPIENT FUEL CENTERLINE MELTING AND 1% CLAD PLASTIC STRAIN (1) 0.0 w/o Gd23 0 3.0 w/o Gd23 0

C Exposure Melt p Melt p (M'JD /MT) (kw/ft) (kw/ f t ) (kw/ft) (kw/ f t )

I FUEL TYPE 8XS 1000 21.5 21.5 21.5 21.5 25000 21.5 21.5 21.0 21.5 50000 21.0 17.5 19.5 17.5 FUEL TYPE 8X8R 21.5 21.5 I 1000 25000 21.5 21.5 21.5 21.5 21.0 21.5 i 50000 21.5 19.0 20.0 18.5 FUEL T/PE P8X8R I 1000 21.5 21.5 21.5 21.5 25000 21.5 21.5 21.5 21.5 50000 21.5 18.5 20.5 18.5

I Note (1): Peak linear heat genera tion ra tes shown are minimum bounding values to the o, currence of the given condition.

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I 5.0 NUCLEAR DESIGN 5.1 Core Power Distributiens The cycle was depleted using SIMULATE [3] to give both a rodded depletion

'and an All-Rods-Out ( ARO) Haling deple tion.

• he Haling depletion serves i as the basis for de fining core reactivity characte ristics for mos t transient and accident evalua tions. This is due primarily to its flat power shape which has weak scram characteristics. Because of the more realistic prediction of initial CPR values, the rodded deple tions were used to evalua te the misloaded bundle error and rod withdrawal error.

I 5.1.1 Haling Power Dis tribu tion The Haling power distribution is calculated in the All-Rods-Out c on di tion. The Haling itera tion converges on a self-consis tent power and exposure shape for the exposure step to end of cycle. In principle , this should provide the overall minimum peaking peser shape for the cycle. This does not mean that during the actual cycle fla tter power dis tributions cannot be achieved by shaping with control rods. However, such shaping would leave i

underburned regions in the core which would peak at another point in time .

i I Figures 5.1.1 and 5.1.2 give the Haling radial and axial average pcwer di s tr ibu tions .

i 1

5.1.2 Rodded Depletion Power Dis tribution To genera te the rodded de pletion, control rod pa tterns were devel3 ped which gave critical eigenvalues at each point in the cycle and gave peaking similar to the Haling calcula tion. 'Ine resul ting pa tterns were generally more I

peaked than the Haling but were not in excess of the bundle and local power peaks seen in past cycles. Hceever, as s ta ted above , the underburned regions of the core can exhibit peaking in excess of the Haling when pulling ARO at End of Cyc'- Figures 5.1.3 ar.d 5.1.4 give the ARO, end o f cycle pow,:r distributions for the rodded depletion, Note in Figure 5.1.4 that the average axial power is more bottor peaked, thus the better scram characteristics.

5.2 Core Exposure Distributions I Cycle 9 was calculated to be capable of a cycle exposure of 8037 MWD /ST at EOFPL (no coastdown). Table 3.3.1 summarizes the resultant core average exposures.

As stated , the Haling power shape produces a sel f-consistent change in exposure. The BOC radial exposure distribution 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 for smaller exposure steps to give ?.he exposure distributions at various mid-cycle points. BOC, EOC-2000 MWD /ST, EOC-1000 MWD /ST, and EOC conditions were used to develop reactivity input for core wide trans ien t

! analysis.

l l

' Wl The rodded depletion may differ considerably from the Haling during the cycle due to the shaping of the power by the rod . However, rod sequences are swa pped frequently and the overall exposure distribution at end of cycle is similar to the Haling. Figure 5.2.2 gives the EOFPL radial exposure distribution for the rodded depletion.

I I .

~ -

5.3 Cold Core Reactivity and Shutdown Margin I The cold K, ff with all rods withdrawn (ARO) and the cold K,fg with all rods insertel ( ARI) at BOC were calculated using the SIMULATE code and are shown in Table 5.3.1. K, f g with ARO minus the cold critical K,ff [3] is the amount of excess core reactivity. K fg with ARI minus the cold critical K is the worth of all the control rods.

eff Technical Specifications [8] state that the core must be suberitical by at leas t 0.25% +R (de fined below) with the stronges t worth control rod withdrawn for sufficient shu tdown margin. Again using SDIULATE, a search was made for the strongest worth control rod at various exposures in th e cyc le .

Th is is necessary because rod worths change with expesure. Then the cold K

eff with the s tronges t rod out was calculated every 1000 MWD /ST through the cycle. Subtracting each cold K with Lhe s tronges t rod out from the cold eff critical K derines the shutdown margin as a function of exposure. Figure e ff

5. 3.1 .s th e r e s ul t . Because the core reactivity increases with exposure, th e cold K 's increase and therefore, the shutdown margin decreases. To ff a cc oun t f or th ia , the value R is calculated as the difference between the cold I K ff with the s tronges t rod ou t at BOC and the maximum cold K eff with the s tronges t rod out in the cycle. The R for Cycle 9 is given in Table 5.3.1.

5.4 Standby i,igaid Control Sys tem Shutdown capability i

l l

l The shutdown capability of the standby liquid control system (SLCS) is l

l designed to bring the reactor fr om full power to cold , ARO, xenon free shu tdown wi th a t least 5% margin. Using the boron concentra tion search option lI in SIMUIATE , the ppm of boron was adjus ted mtil the Keff reached the cold critical K gg minus .05 minus a 95% confidence level uncertainty. This case assumed cold , xenon-free conditions. with All-Rods-Out at the mos t reactive time in th e cycle . The criticai. cy search found that the plant would be 57 suberitical st th e wor s t point in time with 702 ppm of boron injected. VY Technical Specifications require that 800 ppm of boron be available for in je c tion. Table 5.4.1 lists the amount o f boron concentra tion and the shutdown margin.

I I

I I

I I

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I TAB 1E 5.3.1 I K, f f VALUES: SHUTDOWN MARGIN CALCULATION i

l 1.10P4 j BOC Keff - Uncontrolled

.9605 BOC eK << - Contrclied Cold Critical Keff - Search Eigenvalue .9954

.9810 BOC eK gg - Controlled Str ongest Worth Rod Withdrawn BOC Minir:um Shutdown Margin 1.44% A K R, Maxir:um increase in Cold Keff 0.66% A K With Exposure 0.78% A K Cycle Minir:um Shutdown Margin at 5037 MWD /ST I

I TABLE 5.4.1 STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY I ppm of Boron Shutdown Margin I 702 .050 800 .068 I

I I

I _ _ _ _ _ _ - - . _ _ _ _ _ _ -

CYCLE N ETION EOC BUNDLE RVERROE RELRTIVE POWERS R5X2 R7XX R5X2 R8XX REX 2 R7XX R5X2 R7XX R8XX R8XX R7XX 0 989 1 183 1 029 1 308 1 038 1 182 0 972 1 085 0.938 0 938 0 808 R7XX R5X1 RSXX R8XX R8XX R8XX R8XX R8XX RSXX R7XX RSX2 1 181 1 051 1 311 1.171 1.326 1 131 1.251 1 034 1.082 0 841 0.459 RSX2 RSXX RSXX RSXX R8XX R8XX R8XX R8XX RSXX R8XX RSX1 1 028 1 310 1 181 1 385 1 184 1 319 1 107 1 189 0 901 0.270 0.391 R8XX R8XX R8XX RSXX R8XX RSXX R8XX R7XX R8XX RSX2 1 304 1.789 1.384 1 197 1.348 1 137 1.248 1 080 0.949 0 553 RSX2 R8XX RBXX R8XX R8XX R7XX R8XX A8XX R7XX 1 033 1 323 1 182 i.344 1 149 1 174 1 028 1.081 0 771 R7XX RSXX RSXX R8XX R7XX RSXX RSXX R7XX RSX2 1.178 1 125 1 318 1 133 1 171 1.047 1 104 0.888 0.529 R5X2 RSXX RSXX R8XX RSXX R8XX RSXX R7XX RSX2 0.988 1.248 1 100 1 244 1.023 1 102 0.844 0 724 0.404 R7XX R8XX RSXX R7XX R8XX R7XX R7XX R8XX NORTH 1 078 1 025 1 184 1 072 1 057 0.882 0 723 0.519 RSXX R8XX R8XX RBXX R7XX R5X2 R5X2 0.928 1 077 0.897 0.944 0.782 0.526 0.407 I R8XX R7XX E8XX R5X2 R5X1 - 808274H. RELDAD 5 0.930 0.835 0.798 0 548 RSX2 - 8DP8289. RELOAD 5 R7XX REX 2 RSX1 FUEL TYPE 10 R8XX - P8DP8289. RELOAD 8 0.801 0.453 0.388 RELATIVE POWERC I R7XX - P8DP8289. RELOAD 7 RSXX - P8DP8289. RELOAD 8 FIGURE 5.1.1 VY CYCLE 9 HALING DEPLETION EOC BUNDLE AVERAGE RELATIVE POWERS I

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I I VERMONT YANKEE CYCLE 9 8 RRbE L E W RS

I

'I REX 2 R7XX REX 2 A8XX R5X2 R7XX R5X2 R7XX R8XX R8XX R7XX 1 045 1 255 1 099 1 389 1 078 1 207 0.975 1 070 0 919 0 918 0.588 R7XX R5X1 R8XX R8XX RSXX A8XX R8XX R8XX R8XX R7XX RSX2 1 280 1 133 1 410 1.235 1 384 1 158 1 285 1.024 1 087 0.824 0.448 I R5X2 1 100 R8XX 1 405 R8XX 1 251 R8XX 1 429 R8XX 1 219 R8XX 1 345 R8XX 1 112 RSXX 1 179 R8XX 0 888 R8XX 0 785 R5X1 0 379 R8XX R8XX R8XX RSXX R8XX R8XX R8XX R7XX R8XX RSX2 1 374 1 218 1 414 1 222 1 38? 1 138 1 242 1 082 0.933 0 538 R5X2 R8XX RSXX R8XX R8XX R7XX R8XX R8XX R7XX 1 058 1 352 1.195 1 348 1 139 1.153 1.007 1.039 0.748 R7XX RSXX R8kX RSXX R7XX R8XX R8XX R7XX R5X2 1 173 1 120 1 305 1 113 1 139 1 017 1.073 0 859 0.509 R5X2 R8XX R8XX R8XX RSXX R8XX RSXX R7XX R5X2 l

0.948 1 222 1.075 1 210 0.990 1.085 0.814 0 895 0.383 R7XX R8XX R8XX R7XX R8XX R7XX R7XX R8XX NORTH 1 040 0 993 1.147 1 034 1.018 0 848 0.891 0 488 R8XX R8XX RSXX RSXX R7XX R5X2 R5X2 0.894 1 039 0 888 0 910 0 724 0 498 0.383 4 R8XX R7XX R8XX R5X2 RSX1 - BDB274H REL0A0 5 0 895 0 803 0 786 0.520 R5X2 - 8DPB289. RELDRD 5 R7XX RSX2 RSX1 FUEL TYPE 10 RSXX - P90PB289. RELOAD 8 0.571 0 431 0.388 RELATIVE POWERS R8XX - P80P8289. RELOAO 8 FIGURE 5.1.3 VY CYCLE 9 RODDED DEPLETION - ARO AT EOFPL9, BUNDLE AVERAGE RELATIVE F0WERS I

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I I VERMONT YANKEE CYCLE 9 HALING DEPLETION E0C BUNDLE AVERROE EXPOSURES I

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RSX2 R7XX R5X2 R8XX R5X2 R7XX R5X2 R7XX RSXX ROXX R7XX 27 54 14.85 28 39 10.58 28.54 15 20 28 38 15 48 21 78 7 52 12 28 R7XX R5X1 R8XX R8XX RSXX R8XX RSXX R8XX R8XX R7XX R5X2 14.96 24 10 10.81 22.04 10.82 22.24 10.02 21 71 8.70 14.15 25.90 REX 2 R8XX R8XX R8XX R8XX R8XX R6XX R8XX R6XX R8XX R5X1 28.49 10.81 22 15 10.91 21 95 10.E2 21 88 9.ti4 21 80 6 43 24 99 R8XX R3XX RSXX R8XX R8XX R8XX RSXX R7XX R8XX RSX2 10.55 22 10 10.30 22 17 10.70 22 00 10.00 15 38 7 84 25.06 R5X2 R8XX R6XX R8XX RSXX R7XX R6XX .RSXX R7XX 28.58 10.80 22 02 10 89 21 51 16.08 22.07 8.55 13.87 R7XX REXX R8XX R8XX R7XX RSXX R8XX R7XX R5X2 15.23 22 41 10.50 22 11 15 78 20 78 8 92 13 89 24 53 l

R5X2 R8XX R8XX R8XX R8XX R8XX R8XX R7XX R5X2 28.43 9 98 22 12 9.97 22.19 S.90 21 38 13.35 28.12 R7XX R6XX R8XX R7XX R8XX R7XX R7XX R6XX 15.58 21 92 9.50 15.59 8 52 14.05 13.49 18 58 R6XX R8XX RSXX RSXX R7XX RSX2 R5X2 l

22.04 8.88 21.88 7.80 14.25 24.78 25.08 lI R8XX R7XX R8XX RSX2 REX 1 - BDB274H. RELORD 5 7 47 14.34 8.39 25 31 RSX2 - 80PB289. RELOAD 5

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R7XX RSX2 R5XI FUEL TYPE ID R8XX - PSDPS289. RELOAD 8 12.33 28.22 15 07 EXPOSURE (OND/ST)

R7XX - P80P8289. RELOAD 7 R8XX - P80P8289. RELORD 8 l'IGURE 5.2.1 VY CYCLE 9 HALING DEPLETION, EOC BUNDLE AVERAGE Exf'OSURES 1g -u-

I I

VERMONT YRNMEE CYCLE 9 RODDED DEPLET!ON EOC BUNDLE RVERRDE EXPOSURES I

R7XX REX 2 R8XX R5X2 I R7XX RSX2 R7XX R8XX R8XX R7XX REX 2 27 04 14.20 27 80 9 44 28 32 15 25 28 80 15 98 22.13 7 30 12 38 R8XX R8XX R8XX R8XX R6XX R8XX R8XX R7XX RSX2 R7XX REX 1 14.03 23 28 9 11 21 38 9.84 22.09 9.b3 22 08 8 82 14.31 25.99 I RSX2 R8XX RSXX 21 34 R8XX 9 82 R8XX 21 81 R8XX 9 90 R8XX 21 88 R8XX 9 43 REXX 21 80 RSXX 8 15 RSX1 24.99 27 88 9 10 I RBXX R8XX R8XX R8XX R8XX R8XX R8XX R7XX R8XX 7 49 R5X2 25 08 9 48 21.88 10 08 22 09 10 35 22 28 9 85 15 71 I R5XZ R8XX RSXX R8XX R8XX R7XX R8XX R8XX R7XX 28.51 10.05 22 04 10.51 21 98 18.24 22 51 8.82 14.18 R7XX R8XX R8XX R8XX R7XX RSXX R8XX R7XX R5X2 10.82 22 78 10.48 22 89 18 52 21 50 9 14 14 42 24.84 RSX2 R8XX R8XX RSXX R8XX R8XX R8XX R7XX R5X2 f

28.94 10.08 22 59 10.21 22 84 9.20 21 98 13 93 28.44 R7XX R8XX R8XX R7XX R8XX R7XX R7XX R8XX t r NORTH 18.28 22 48 9 84 18 18 8 75 14 85 14.09 19 05 R8XX R8XX R8XX R8XX R7XX R5X2 R5X2 22.50 8.70 22 00 7.80 14.84 25.14 25.43

.I RSXX R7XX R8XX R5X2 R5X1 - 808274H. RELOAD 5 7 38 14 61 8 22 25 39 , p R7XX R5X2 RSX1 FUEL TYPE ID 98XX - P8DP8289, RELOAD 8 1 12.51 28.37 25 13 EXPOSURE (OHD/ST1 R7XX - P8DP9289. P~ LORD 7 R8XX - P8DP8289. REL0A0 8 FIGURE 5.2.2 VY CYCLE 9 RODDED DEPLETION, EOC BUNDLE AVERAGE EXPOSURES l

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M M M M M M M M M M M M M M VERMONT YANKEE SHUT 00WN MARGIN FOR CYCLE 9 COLD PERCENT SHUTOUWN DELTA K VS. CYCLE AVERAGE EXPOSURE 2.000 - - - - - - - - - -

1 900.~ .

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1 700" 1.600" 1 500' '

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.400 300' TFCH SPFC LIMI T TFCH SP TC LIMIT

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' 7030 Cl 1030 2030 3030 4030 6030 8030 9030 ' 10C'00 CYCLE EXPOSURE (MH0/ST1 FIGURE 5.1.I VY SIIUTD_QW_LTARGTN EOR CYCI.E 9 COTD PERCENT SistTTDOWN DET,TA K VERSifS CYCT.E AVERACE EXPOSURE

I 6.0 THERMAL-HYDRAULIC DESIGN I The thercul hydraulic evaluation of the reload cycle was performed using the methods described in the following section.

I 6.1 Steady-State Thermal Hydraulics

~

Core s teady state thermal-hydraulic analyses were performed using .he FIIMR [9,10] compu ter code. The FIBWR code incorporates a detailed geometrical representation of the complex flow paths in a BWR core, and explicitly models the leakage fl ow to the bypass region. FIEWR calcula tes the core pressure drop and total bypass flow for a given total core flow.

The power dis tribu tion, inlet enthalpy, and geometry are presumed known and are supplied to FIBWR. Power dis tribution is derived by the 3-D neutronic I s imula tor SIMULATE [3]. Core pressure drop and total leakage flow predicted by the FIBWR code were used in setting the initial conditions for the sys tem's transient analys is model. Further details are provided in Re ference 9.

1 1

= 6.2 Reactor Limits Determina tion The objective for normal operation and transient events is to maintain nuclea te boiling and thus avoid a tr e: s i t ion to film boiling, th ereby l protecting the fuel cladding integrity. Based on Reference 11, the fuel cladding in tegrity safety limit for Vermont Yankee is a lowest allowable i minimum critical power ratio (IAMCPR) of 1.07 for 8X8R/P8X8P. reload l

l 1 '

l I

cycles. Opera ting limits are specif *..d to maintain adequate margin to th e onset of the boiling transition. The figure of merit utilized for plant opera tion is the critical power ra tio (CPR). This is defined as the ra tio of the critical power (bundle power at which some point within the assembly experiences onset of boiling transition) to the operating bundle power.

Thermal cargin is stated in terms of the minimum value of the critical power ra tio MCPR, which corresponds to the most limiting fuel assembly in the core. Both the transient (sa fety) and normal operating thermal limits in terms of MCPR are derived based on the GEXL correla tion as described in Re fe r en ce 11.

Vercunt Yankee Technical Speci fication limits the operation of 8X8, 8X8R, and P8X8R fuel types to a maximum linear heat genera tion ra te (MLHGR) of 13.4 KW/ft. The basis for a MLHGR o f 13.4 KW/f t can be found in Re ference 2.

1 l

l l

I l

l l

l

• I 1 -

7.0 ArCIDENT ANALYSIS .

I 7.1 Core Wide Transient Analysis I Core wide transient simulations are performed to assess the impact of the part cular transient on th e h ea t transfer characteristics of the fuel. The figure of merit used is the critical power ratio [11] .

More specifically , it is the purpose of the analysis to determine th e minimum critical power ratio for each fuel type such that the sa fety limit critical power ratio is not exceeded for the transients considered.

I 7.1.1 Meth odol o gy I The analysis requires two types of simulations. A system level simula tion is performed to determine the overall plant response. Tr ans ien t core inlet and exit pressures aad normalized power from the system level calculation are used to perform detailed thermal hydraulic simulations (r e ferr M to as " hot channel calculations") of each fuel type. The results of each of these latter simulations is the bundle tr ans ien t ACPR (the in itial bundle CPR minus the minimum CPR experienced during the transient).

I 1e sxst m 1 ve1 s1mu1ations ar P rform d w1th the moea1 eocum.ntee 1n Re ference 12.

I I

The hot channel calculations are performed with the RETRAN [13] and I MAYU4-YAEC [14] computer codes. The CEXL correlation [11] is used in MAYU4-YAEC to evaluate critical power ratio. The calculational procedure is outlined below.

The hot ch ann el transient ACPR calculations are performed via a series of

" inner" and "ou ter" i tera tions , as illus trated by the flow chart in Figure 7.1.1. The outer loop represents iterations on the hot channel initial power lev el . These iterations are necessary because the ACPR for a given transient varies with Initital Critical Power Ratio (ICPR) , yet only the ACPR corresponding to a transient MCPR equal to the safety limit (i .e . , 1.07 + ACPR

= ICPR) is appropria te. The approxicate cons tancy of the ACPR/ICPR ratio is use ful in these iterations. Each outer iteration requires a RETRAN hot channel run to establish the transient boundary conditions of inlet flow, inlet enthalpy, channel pressure and power level required for irrut to the MAYU4-YMC code . MAYU4-YAEC is then used to calculate a CPR at ea ch time-s tep during the tr ans ient , from which a transient ACPR is derived. The hot ch ann el is modeled using a chopped cosine axial power shape with a peak / average ratio of 1.4.

The inner loop represents iterations on the hot channel inlet flow.

I There itera tions are necessary because the RETRAN hot channel model c..lculates 1

an exit loss coe fficent when given the inital power level, flow, and pressure drop as input. The pressure drop is assumed equal to the core average pressure drop, and the flow is varied for a g,"ven power level until the exit loss coe f ficient is correct. FIIMR [9] is utilized to estimate the correct inlet flow for a particular power level and pressure drop.

7.1.2 Initial Conditions and Assumptions 1'

The initial conditions for the system simulations are based on 105% rated steam flow (maximum turbine capacity) and 100% core flow. The core axial power distribution for each of the exposure points is based on llating-mode 3-D SIMULATE predictions associated with the generation of the reactivity data (Section 7.1.3). The core ' inlet enthalpy is set so that th e amoun t o f c a rryun der from the steam separators and the quality in the liquid region ou t s ide the separators is as close to zero as possible. For fast pr es sur iza t ion tr ans ien t s , this maximizes the initial pressuriza tion ra te r.nd predicts a more severe neutron power spike. A summary of the initial opera ting s ta te used f er the sys tem simula tions is provided in Table 7.1.1.

Assumptions specific to a particular transient are discussed in the sec tion describing the trans ient. In general, the following assumptions are ma de for all transients.

1. Scram setpoints are at Technical Specification limits.

I

2. Protective system logic delays are at equipment specification limits, j 3. Sa fety/Relie f Valve and Sa fety Valve capacities are based on Technical Specifica tion ra ted values.

P t

4. Sa fety/Relie f Valve and Sa fety Valve setpoints are nxadeled as being 1% above the Technical Specifica tion upper limit. Valve responses are based on slowes t specified response values.

I 5. Control rod drive scram speed is based on the proposed Technical Speci fication limits. The analysis addresses a dual set of scram speeds as given in the proposed Technical Specifications. These are referred to as the " measured" and "67B" scram time sets.

I 7.1.3 Reactivity Functions I The methods used to generate the fuel temperature , moderator dens ity , and scram reactivity functions are described in Reference 15 and are specifica ly .

as outlined in Figures 2.1 through 2.J of that document. A complete set of reactivity functions , the axial power dis tribu tion, and kinetics parameters are generated from "de fined" base s tates es tablished for EOC, EOC-1000 MWD /ST, EOC-2000 MWD /ST, and BOC cycle exposure conditions. These s ta tes are characterized by exposure and void history distributions , control rod pattern, and core thermal-hydraulic conditions.

l The Cycle 9 BOC hot , full power core s tate is es tablished from the pre-de fined Cycle 8 endpoint , the Cycle 9 reload pa ttern, and an es tima te of the BOC control rod pattern. The EOC and intermedia te core exposure and void his tory distributions were calcula ted via a Haling depletion as described in Section 5.2. The EOC state is unrodded and, as such , is de fined su f fic ien tly . However, EOC-1000 MWD /ST and EOC-2000 MWD /ST exposure points require a control rod pattern and such was developed using the following al gor i thm. Beginning with the rodded depletion control configuration, all control rods which are mare than hal f-inserted are fully inserted and all

.I control rods which are less th an half-inserted are f ully withdrawn. If th e SIMULATE-calculated parameters are less than or close to limits, then this configura tion becomes the base case. If the limits are greatly exceeded in this SIMULATE calculation, a minimum number of control rods are adjus ted a min imum numb er o f no tch es un t il the parame ters are less than or close to I. limits. Using this mthodology , the control rod patterns and resul tan t power dis tribu tions are es tablished so as to minimize the scram reactivity function and to maximize the core averaga moderator density reactivity coe fficient.

For the turbine trip without bypass, generator load rejection without bypass, and loss of feedwater hea ting transient analyses , these configurations ma x imize the power response.

I In generating the fuel reactivity function data for RETRAN, 12 unique v olume-spe ci f ic cable sets are produced which are analagous to that shown in Figure 3.7 of Re ference 15. h ooderator and relative moderator density iI functions also are 12 unique volume-specific tables, analagous to Figures 3.10 and 3.11 in Re ference 15. A moderator density set is generated specifically for each tr ans ient type. The density reactivity functions for the subcooling transient are generated by quasi-statically varying the inlet subcooling only. The modera tor enthalpy source dis tribution is in equilibrium with the cal cula ted nu clear power. The density reactivity functions of the pr e s sur iza tion trans ients are genera ted by quasi-sta tically varying the core pressure. A series of the cal culations are per formed for various inlet modera tor t empera tur e s . T1. 2 modera tor enthalpy source dis tribution is tha t o f th e bas e s ta te cas e .

I In order to qualitatively compare the core reactivity characteristics between di fferent base configura tions, core average reactivity coe fficients I are calculated and provided in Table 7.1.2. Calculated kinetics parameters for RETRAN are also provided in the Table.

The reactivities versus scram insertion is calculated at cons tant, pre-trans ient modera tor conditions. These calculated data are fit and evaluated to yield the highly derailed scram reactivity curves. These are then combined with the rod position versus time curves to es tab..sh the final RETRAN scram reactivity function. Figures 7.1.2 through 7.1.4 display the inserted rod worths and rod positions as functions of scram time f or th e

" measured" scr am time analysis. Fi',ure 7.1.5 through 7.1.7 display similar curves for the "67B" scram time analys is.

I 7.1.4 Transients Anal yzed I Pas t licensing experience has sh own th a t the core wide tr ans ien ts mos t likely to resul t in the minimum core thermal margins are:

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

2. Turbine trip with complete failure of the turbine bypass sys tem.

I 3. Loss o f feedwa ter h ea ting.

I I

I The "feedwater controller failure (maximum demand)" transient is not a severe transient for Vermont Yankee, because of the plant's 110% stern flow b ypas s s ys te m. Pas t analyses have shown this transient to be considerably less severe than any of the above for all exposure points. Brief descriptions and the results of the core wide transients analyzed are provided in the

'f ollowing sec tion.

I 7.2 Core Wide Transient Analysis Resul ts I The transients selected for consideration were analyzed at exposure points of end of cycle (EOC), EOC-1000 MWD /ST, and EOC-2000 MWD /ST; the loss of feedwa ter hea ting was also evalua ted at beginning of cycle conditions. A summary of the results of the analyses is provided in Table 7.2.1. The turbine trip without bypass transient is the most severe tr ans ient for the EOC condition; and the loss of feedwater heating transient is the mos t severe for the EOC-1000 MWD /ST and EOC-2000 MWD /ST conditions.

I 7.2.1 Turbine Trip Without Bypass Transient (ITWOB )

The tr ans ient is initiated by a rapid closure (0.1 sec. closing time) of the turbine s top valves . It is assumed that the s team bypass valves , which normally open to relieve pressure , remain closed. A reactor protection sys tem signal is generated by the turbine stop valve closure switches. Control rod drive motion is conservatively assumed to occur 0.27 seconds a f ter the start o f turbine stop valve motion. The ATWS recirculation pump trip is assumed to occur at a setpoint of 1150 psig dome pressure. A pump trip time delay of 1.0 second is assumed to accoun t for logic delay and M-G set generator field I

I

I 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 v alv e s . Predictions of the sys tem parameters of main interest are shown in Figures 7.2.1 through 7.2.3 for the three exposure points for the " measured" scram time analys is.

t E 7.2.2 Generator Load Rejection Without Bypass Transient (GLRWOB)

I The tr ans ient is initiated by a rapid closure (0.3 seconds closing time) o f the turbine control valves. As in the case of the turbine trip transient ,

the bypass valves are assumed to f ail . A reactor protection sys tem signal is I generated by the hydraulic fluid pressure switches in the acceleration relay of the turbine control sys tem. Control rod drive motion is conserva tively assumed to occur 0.28 seconds after the start of turbine control valve motion. The same modeling regarding the A1WS pump trip and bypass piping are us ed in the simula tion as in the turbine trip simulation. The influence of I the accelera ting main turbine-genera tor on the recircula tion sys tem is simulated by specifying the main turbine generator electrical frequency as a

,I f un c t ion o f time f or the M-G set dr ive mo tor s . The main turbine genera tor frequency curve is based on a 100% power plant startup test and is considered repr es enta tive for the s imula tion. The sys tem model predictions for the three

.l exposure points are shown in Figures 7.2.4 through 7.2.6 for the " measured" E

scram time an al y s is .

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I 7.2.3 Loss of Feedwater Heating Transient (LOFWH)

I A f eedwa ter h ea ter c an be los t in such a way that the steam extraction line to the heater is shut off or the feedwater flow bypasses one of the h ea ter s . In either c ase , the reactor will receive cooler feedwater, which will produce an increase in the core inlet subcooling, resulting in a reactor pbwer increase.

The response of the system due to the loss of 700 F of the feedwater hea ting capability was analyzed. This represents the current licensing assumption for the maximum expected single heater or group of heaters that can be tripped or bypassed by a single event. The reactor is assumed to be on manual recirculation flow control when the heater is lost.

I The transient response of the sys tem was evaluated at several exposures during Cycle 9. Transient evaluation at EOC-2000 MWD /ST was found to be the limiting case between BOC to EOC.

Vermont Yankee has a scram set point (120% of NBR) as part of the reactor protection sys tem (RPS) on high neutron flux. In this analysis, no credit was taken for scram on high neutron flux, thereby allowing the reactor power to r ea ch its peak without scram. This aFproach was selected to previde a bounding and conservative analysis.

I -

The results of the system response to a loss of 100 F feedwater heating capability evaluated at EOC-2000 MWD /ST as predicted by the RETRAN cr de are presenteo in Figure 7.2.7.

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I 7.3 Overpressuriza tion Analysis Resul ts I

Compliance with ASME vessel code limits is demons trated by an analysis of the main s team isola tion valves (MSIV) closing with failure of the MSIV position switch scram. End of cycle conditions were analyzed. The system model used is the same as tha t used for the core wide transient analys is (Section 7.1.1). The initial conditions and modeling assumptions discussed in Section 7.1.2 are applicable to this s imula tion. The maximum pressure at th e bottom of the reactor vessel is calculated to be 1280 psig for the " measured" scram time analysis and 1310 psig for the "6T scram time analysis. This r esul t is within the allowable code limit of 1375 psig,10% above vessel design pressure for upset conditions.

I Th e tr an s ie n t is initia ted by a simul taneous closure of all four MSlV's.

A 3.0 second closing time , which is th e Te ch S pe c . min imum , is assumed. A reactor scram signal is generated on APRM high flux. Control rod drive motion is conserva tively assumed to occur 0.28 seconds a f ter reaching the high flux setpoint. The system response is shown in Figure 7.3.1 for the " measured" scram time analys is.

7.4 Rod Withdrawal Error Transient Re sul ts The rod withdrawal error is a local core tr ans ient caused by an opera tor erroneously withdrawing a control rod in the continuous withdrawal mode. If the core is operr. ting at its opera ting limits for MCPR and LHGR at the time of the err or , then withdrawal of a control rod could increase both local and core power levels with the potential f uc overhea ting the fuel.

I I

There is a broad spectrum of core conditions .:nd control rod patterns which could be present at the time of such an error. For many situations it would be possible to fully withdraw a control r 'd without violating either fuel cladding integrity safety limit.

I To bound the mos t severe of pos tulated rod withdrawal error events , a portion af the core MCPR opera ting limit envelope is specifically defined t. c5 th a t the cladding limits are not violated. The consequences of the error depend on the local power increase , the initial MCPR of the neighboring I locations and the ability of the Rod Block Monito: Sys tem t'o stop the withdrawing rod before MCPR reaches 1.07.

The mos t severe pos tulated transient begins with the core operating according to normal procedures and within normal operating limits. The operator makes a procedural err or and attempts to fully withdraw the maximum worth control rod at maximum withdrawal speed. The core limiting loca tions

.tre close to the error rod and therefore experience the spatial power shape tr ans ien t as well as the overall core power increase.

I The core conditions and control rod pattern for the bounding case are specified using the f ollowing set of concurrent worst case assumptions:

I

1. The rod should have high reactivity worth. This is provided for by analysis of the core at the exposure correspondi to ma x it..um c on tr ol inventory with the xenon-free condition superimposed. The xenon-free condition and the additional control rod inventory needed to maintain

, I I

I criticality exaggerates the worth of control rods substantially when compared to normal opera tion with norm.1 xenon levels. A fully inserted high worth rod is selected as the err a rod. i l

2. 'Ihe core is initially at rated power and flow.

I

3. The core power distribution is adjus ted with the available control rods

.I to place the locations within the four by four array of bundles around the error rod as nearly on the operating limits as practical .

The Rod Block Monitor Sys tem's ability to termina te the bounding case is i

evalua ted on the following bases:

1. Technical Specifications allow each of the separate RBM channels to remain operable if at least half of the LPRM inputs at every level are operable. For the interior RBM channels tes ted in this analysis , there are a maximum of four LPRM inputs per level. One RBM channel averages

'. the inputs from the A and C levels; the other channel averages th e inputs fr om the B and D levels. Considering the inputs for a single channel, there are eleven failure combinations of none, one and two failed LPRM strings. The RBM channel responses are evaluated separa tely at th ese eleven input failure conditions. Then, for each channel taken separa tely, the lowest response as a function of error rod position is chosen for comparison to the RBM setpoint.

I

2. The event is analyzed separa tely in each of the four quadrants 'f th e cc _a to the differing LPRM string physical locations relative to th e error rod.

I Technical Specifications require that both RBM channels be operable during normal opera tion. Th us , the first channel calculated to intercept the RBM setpoint is as sumed to s to p th e rod . To allow for control system delay times, the rod is assumed to move two inches af ter the intercept and s top at the following notch.

I The analysis is performed using the three dimensional steady state Necessary proper ties of tha t SIMULATE core model demons tra ted in Re ference 3.

model for use in this analysis are:

1. Accurate bundle power calculation as shown by the PDQ and gamma scan J3 comparisons.

!I 2. Accurate LPRM signal calcula tion as shown by the detailed TIP trace comparisons.

3. Accurate control rod worths and core power coe fficient as shown by the cons is tent core eigenvalues.

The Two separate cases are presented from explicit SIMUIATE analyses.

reactor conditions and case descriptions are shown in Figures 7.4.1 and 7.4.2. Case 1 analyzes the bounding event with the concurrent abnormal xenon condition and rod pattern configura tion provided to increase the wor th o f th e error rod. The initial conditions for Case 2 approximate the expected full power conditions at the mos t reac tive point in the cycle; the control rod The ACPR density is at its maxinum at the normal equilibrium xenon condition.

The ACPR values are and MLHGR values for both cases are shown in Table 7.4.1.

~I I

evalua ted such that the implied operating limit MCPR equals 1.07 + ACPR, conserving the Figure of Merit (ACPR/ Initial CPR) shown by the SIMUI. ATE cal cula tions . The use of this method provides valid ACPR values in the analysis of normal opera ting sta tes where locations near the assumed error rod are not initially near the MCPR operating limit. Case 2 is the worst of three

. rod withdrawal transients analyzed from the same initial full power, full flow and rod pattern conditions. Cases using the rod at coordinates (22,27) and at (22,35) were found to be less severe than Case 2 and all were bounded by Case 1 with subs tan tial MCPR margin.

I As an example for Case 1, Figures 7.4.3 and 7.4.4 show the determinations of the end of transient control rod position at the point where the weakes t RBM channel response first intercepts the RBM setpoint.

The opera ting limit ACPR envelope component versus Rod Block Monitor setpoint is taken from the Table 7.4.1 valu<_s for the bounding Case 1. The same table demons tra tes margin to the 1% plas tic strain limit including the 2.2% power spiking penalty.

l 7.5 Mi sl oa ded Bundle Error Analysis I 7.5.1 Rotated Bundle Error I

I The primary result of an assembly rotation is a large increase in local pin peaking and R-factor as higher enrichment pins are placed adjacent to the surrounding wide water gaps. In addition, there may be a small increase in reactivity , depending on the exposure and void fraction s ta tes. The R-factor 1

. I l

w -

I increase results in a CPR reduction, while the local pin peaking factor increase results in a higher pin linear heat generation rate. The objective of the analysis is to insure that in the worst possible rotation, the sa fety limit lin ear h ea t genera tion ra te and CPR are not viola ted with the mos t limiting monitored bundles on their operating limits.

To analyze the CPR response, rota ted bundle R-factors as a function of exposure are developed by adding the largest possible AR-factor increase resul ting from a rota tion to the exposure dependent R-factors of the properly oriented bundles [11] . Using these rotated bundle R-factors, the minimum CPR values resul ting from a bundle rota tion are de termined using the SIMULATE code, f or ea ch c on tr ol r od s e que n ce thr ou gh ou t the cycle . These minimum CPR values are in addition modified slightly to account for the change in reactivity resulting from the rotation. For each sequence , the MCPR for the properly oriented assemblies is ratioed by a factor necessary to p! ace the minimum rotated CPR on its 1.07 sa fety limit . The maximum of these adjus ted MCPR's is the rota ted bund)e opera ting 1imit.

To determine the maximum linear heat generation rate (MLHGR) resul ting from a rota tion, the ratios of the maximum rotated local peaking factor to the maxir:um unrotated local peaking are determined for the expected range of exposure and void conditions. The maximum of this ratio is applied to the operating limit LHGR. This maximum rotated bundle LHGR is in addition modified to account for (1) the possible reactivity increase resul ting from the rotation and (2) a 2.2% power spiking penalty.

s,. r. sui ts o f the rota t.e bunei. ana1ysis ar. as g1v.n 1n Tas1. ,.s.1.

g I

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I 7.5.2 Misloca ted Bundle Misloa. ling a high reactivity assembly into a region of high neutron importance results in a loca tion of high relative assembly average power.

Since the assembly is assumed to be properly oriented (not rotated), R-factors

'used for the misloaded bundle are the standard values for the misloaded fuel type.

I The analysis for Cycle 9 consists of an iterative procedure which successively elimina tes potential misloading locations from any MCPR safety limit violations . Th e fir s t step is to use SIMULATE to determine the larges t possible ACPR which could result at any location as the result of misloading a high reactivity assembly into the location. This maximum ACPR is then applied to all core CPR's for various cycle exposure sta tes. Even with this maximum ACPR applied , some locations will never exceed the MCPR sa fety limit o f 1.07.

I Th e ne xt iteration consists of applying the same procedure to the locations which did not result in CPR's above tne sa fety limit when the j maximum ACPR from the firs t itera tion was applied. Since these locations are o f higher rea ctivity than those clin.inated in the first iteration, t?ey will r es ul t in a lower ACPR from a misloading. Using this new maximum ACPR, some of the remaining loc tions will be eliminated from potential CPR sa fety limit v iola t ions . This procedure is con tinued un til all loca tions are shown to be above the MCPR sa fety limit due to a misloading, or until a limiting location l is iden ti f ied .

I l

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I Using the above procedure, it has been demons trated that for Cycle 9 all possible misloca tions resul ted in calcula ted MCPR's above the 1.07 a fe ty limit, assuming an initial operating CPR limit of 1.24, 7.6 Control Rod Drop Accident Re sul t s

~

I The control rod sequences are a series of rod withdrawal and banked with dr awal instructicas specifically designed to minimize the worths of individual control rods. The sequences are examined so that, in the event of I

the uncoupling and subsequent free f all of the rod , the incremental rod worth is acce ptable. Incremental worth refers to the fact that rods beyond Group 2 are banked out of the core and can only f all the increment from all-in to th e rod orive withdrawal positicn. Acceptable worth is one which produces a ma ximum fuel en th al py l e ss than 280 calories / gram. This provides considerable margin to the sudden fuel pin rupture threshold of 425 calories / gram [16] .

I Some out-of-sequence control rods could accrue potential high worths.

However, the Rod Worth Minimizer (RWM) will prevent withdrawing an out-of-sequence rod if accidentally selected. The RWM is func tionally tes ted before ea ch s tartup.

I The sequence entered into the RWM will take the plant from All-Rods-In (ARI) to well above 20% CTP. Above 20% power even mul tiple opera tor errors will not crea te a potential rod drop situation above 280 calories per 1

gr am [ 17 ) . Below 20% CTP, however , the sequences must be examined for incremental rod worth. Th is is done using the full core, xenon free SIMULATE j

model at the projected most reactive point in the cycle. Th is assures tha t l the maximum amount of reactivity is held in the rods.

lg3

I l

Both the A and B sequences were examined. It was found that the highes t worth occurred in the first rod pull of the second group. Any o f the first four rod arrays shown in Figures 7.6.1 and 7.6.2 may be designated as th e L first stoop pulled. But, then a specific second aroup =ust follow as Table 7.6.1 illus trate s . For added conservatism, th e h : gh es t wor th rod in th e r

'second group was deliberately assigned to be the first rod pulled. This assures that in i r sequence followed at the plant, the worths will always be less than those ca.' cula ted here.

1 Beyond Group 2, banking pcocedures [18] apply which severely limit th e l rod incremental worths. Th is is why the xenon free , hot s tandby worth is much less than the cold xenon free worth. The results of the calculations are presented in Table 7.6.2.

7.7 Stability Analysis Resul ts The analysis of reactor stability has been performed by General Electric as described in Section 5.4 of Reference 2. The 105% rod line was analyzed and the resultant decay ratio as a function of reactor power level is provided in Figure 7.7.1.

I The reactor core stability decay ratio, x2 /*0, is calculated to be 0.83. The channel hydrodynamic performance decay ratios , 2x /*0, are calculated to be 0.38 and 0.30 for the EX8 channel and the 8X8R/P8X8R channel, r es pect ivel y.

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TABLE 7.1.1

SUMMARY

OF SYSTEM TRANSIENT MODEL INITIAL CONDIT0NS FOR CORE WIDE TRANSIENT ANALYSES Core Thermal Power (MWth) 1664.0 Turbine Steam Flow (% NBR) 105 Total Core Flow (1061bm/hr ) 48.0 Core Bypass Flow (10 61bm/hr ) 5.3 Core Inlet Enthalpy (BTU /lbm) 520.5 Steam Dome Pressure (psia) 1034.7 Turbine Inlet Pressure (ps ia) 991.6 Total Recircula tion Flow (1061bm/hr ) 23.5 Core Pla te Differential Pressure (psi) 18.6 Average Fuel Gap Conductance sBTU/hr-ft2 _p)

EOC 900.0 EOC-1000 MW9/ST 890.0 EOC-2000 MWD /ST 885.0 Narrow Range Wa ter Level (in. ) 35 I

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m M M M M M M M M M M M M M M M M M m Table 7.1.2 TRANSIENT ANALYSIS REACTIVITY COEFFICIENTS VERMONT YANKEE CYCLE 9 Cycle Exposure Poitit (MWD /ST)

Calcula ted Parameter EOC (EOC-1000) (EOC-2000) BOC Axial Shape Index(l) -0.0490 -0.2057 -0.2498 -0.1220 Modera tor Dens ity Coe fficient 21.0 24.1 25.3 20.4 (Subcooling), //Au(2)

Pressure = 1050 ps ia Subcooling = 30 BTU /lbm l Moderator Density Coe fficient 22.9 24.7 25.9 -

4 I (Pressuriza tion), //Au s- Pressure = 1050 psia Inlet Enthal py = 520 BTU / lbm Fuel Temperature Coe fficient -0.272 -0.267 -0.275 -0.249 at 1130 F, // F Effective Delayed 0.005383 0.005471 0.005533 0.005999 Neutron Fraction Prompt Neutron Generation 42.74 43.13 42.57 39.69 Time , microseconds P -P Notes: (1) Axial 5 ape Index ( ASI) = p p (2) Au = change in density, Ibm /f t3

I I TABLE 7.2.1 CORE WIDE TRANSIENT ANALYSIS RESULTS I

Peak Peak Avg.

I ~ Tr ans ient Exposure Prcxnpt Power (Fraction of Initial Value) hat Flux (Fraction of Initial Value) 8X8 ACPR 8X8R/P8X8R I Turbine Trip EOC 3.50 1.23 0.20 0.20 With out Bypass, I 'heasured" Scram Time EOC-1000 EOC-2000 2.86 2.20 1.18 1.10 0.15 0.01 0.16 0.07 I Turbine Trip EOC 3.97 1.28 0.25 0.24 With ou t Bypass, I "67B" Scram Time EOC-1000 3.38 1.24 0.21 0.20 EOC-2000 2. 9 1.21 0.13 0.13 Generator Load EOC 3.35 1.22 0.20 0.19 I Rejection With ou t Bypas s ,

" Measured" EOC-1000 2.78 1.17 0.16 0.15 Scram Time EOC-2000 2.08 1.08 0.06 0.06 I Generator Load EOC 3.93 1.27 0.26 0.25 I Rejection With ou t Bypass, "67B" EOC-1000 3.40 1.23 0.23 0.23 Scram Time EOC-2000 2.78 1.14 0.15 0.13 i Loss of 100 F EOC 1.21 -

0.15 0.15 l Feedwater EOC-1000 1.22 -

0.17 0.16 l Heating EOC-2000 1.23 1.22 0.17 0.17 BOC 1.21 -

0.15 0.15

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I TABLE 7.4.1 t I ROD WITHDRAWAL ERROR TRANSIENT

SUMMARY

(WITH LIMITING INSTRUMENT FAILURE)

I Case 1 Conditions in Figure 7.4.1 I Setpoint RBM Rod Position ACPR 8X8/ 8X8R/P8X8R MLHCR (kw/ft) 8X8/8X8R/P8X8R y 104 08 .08 13.8 g 105 10 .11 14.5 106 12 .14 15.2 107 14 .18 15.8 108 18 ,22 16.9 I

Case 2 Con di t ions in Figure 7.4.2 RBM Rod ACPR MLHGR (kw/ft)*

Setraint Position 8X8/8X8R/P8X8R 8X8/8X8R/P8X8R 104 16 .06 11.8

.08 12.1 I 105 18 106 18 .08 12.1 l 107 24 .12 12.6 108 26 .14 12.6

• Not initially on limits I

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TABLE 7.5.1 ROTATED BUNDIE ANALYSIS RESULTS I

I Initial MCPR Resul ting MCPR Re sul ting LHGR (kw/ft) 1.24 1.07 17.47 I

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TABLE 7.6.1 CONTROL ROD DROP ANALYSIS - ROD ARRAY PULL ORDER I The order in which rod arrays are pulled is specific once the choice of

- first gr oup is ma de .

First Group Second Group Third Group Pulled is: Pulled Must Be: Is Banked Out Array 1 Array 2 Array 3 or 4 Array 2 Array 1 Array 3 or 4 I Array 3 Array 4 Array 3 Array 1 or 2 Array 1 or 2 Array 4 I

TABLE 7.6.2 CONTROI ROD DROP ANALYSIS RESULTS Maximum Incremental Rod Worth:

I Cold Xenon Free .86% AK Hot Standby, Xenon Fr ee .28% AK I Bounding Analysis Worth for Enthalpy less than 280 Calories per Gram 1.30% AK (Re ferences 17 and 19)

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choose ICPR l'

Estimate Power 4 I

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I 1f 1 Estimate Flow with 71BWR 1 r I RETRAN Flov Initialization Run I ..

Exit Loss Coefficient Revise Correct? Flow I 3r Yes RETRAN/MAYU4-YAEC Ect Channel Run I

Ra.

I f.CFR Converged?

N y g, g g Yes STOP lI FIGURE 7.1.1 FLOW CHART FOR TliE CALCULATION OF ACPR USING RETRAN/MAYU4 CODES l

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FIGURE 7.1.2 INSERTED ROD WORTH AND ROD POSITION VERSUS SCRAM TIME AT EOC 1

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INSERTED ROD WORTH AND ROD POSITION VERSUS SCRAM TIME AT EOC-1000 MRD/ST 1 _ - _ _ . _ -

" MEASURED" SCRAM TIME I 8 E

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FIGURE 7.1.4 INSERTED ROD WORTH AND POSITION VEPSUS SCRAM TIME AT EOC-2000 MWD /ST I - - -

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I "67B" Scram Time i

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Figure 7.1.5 Inserted Rod Worth and Rod Position Versus Scram Time at EOC

"67B" Scram Time

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' Inserted Rod Worth and Rod Position Versus Scram Time At EOC-1000 MWD /ST I .

l "67B" Scram Time 1

= 8 1

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!I Figure 7.1.7 Inserted Rod Worth and Rod Position Versus Scram Time At EOC-2000 WD/ST

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FIGURE 7.2.7-2 LOSS OF 100 F FEEDVATER llEATING, EOC9-2000 MWD /ST (LD1ITING CASE)

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Numbers Indicate Number of Positions Withdrawn Out of 48, Blank is a Withdrawn Rod.

I Reactor Conditions:

Core Thermal Power = 1664 Mwt Core Flow = 48 M1b/hr I Cycle Exposure = 4900 MWD /T Xenon Free Initial MCPR = 1.237 I Initial LHGR = 13.8 kw/ft Case Description I o Operator attempts full withdrawal of the fully inserted rod at coordina tes ( 26, 27) .

o Bounding Case .

FIGURE 7.4.1 REACTOR INITIAL CONDITIONS FOR RWE CASE 1 I

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CONTROL ROD PATTERN I

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24 14 24 3 42 l

2 6 10 14 18 22 26 30 34 38 42 lW Numbers Indicate Number of Positions Withdrawn Out of 48. Blank is a Withdrawn Rod.

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,l Reactor Conditions:

Wir Core Thermal Power = 1664 Mwt Core Flow = 48 M1b/hr I Cycle Exposure Equilibrium Xenon Initial MCPR

= 4900 MWD /T

= 1.406 Initial LHGR = 11.9 kw/ft I Case Description o Operator attempts full withdrawal of the partially inserted rod I o at (30,27).

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.I FIGURE 7.4.2 REACTOR INITIAL CONDITIONS FOR RWE CASE 2 I

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I 19 3 4 3 4 3 15 1 2 2 1 11 4 3 4 3 4 07 2 1 1 2 03 3 02 06 10 14 18 22 26 30 34 38 42 FIGURE 7.6.1 FIRST FOUR R0D ARRAYS PULLED IN THE A SEQUENCES I

g 43 3 3 I 39 2 1 2 35 3 4 4 3 31 2 1 2 1 2 2; 3 4 3 3 4 3 23 1 2 1 2 1 g

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02 06 10 14 18 22 26 30 34 38 42 FIGURE 7.6.2 FIRST FOUR ROD ARRAYS PULLED IN THE B SEQUENCES I -

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ULTIMATE PERFORMANCE LIMIT

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I FIGURE 7.7.1 REACTOR CORE DTCAY RATIO VERSUS POWER I

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I I 8.0 STARTUP PROGRAM Following refueling and prior to vessel rearsembly, fuel assembly poeition and orienta tion will be verifie:J by underwater television and video ta ped .

~

I The Vermont Yankee 5tartup Program will int'ude process computer data checks , shu tdown margin demons tra tion, in-sequence critical measurement, rod scram tes ts, power dis tribution comparisons , TIP reproducibility, and TIP s ymme try checks. The content of the Startup Test Report [20) will be similar to that sent to the Office of Inspection and Enforcement subsequent to th e start of Cycle 8.

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I 9.0 LOSS-OF-COOLANT ACCIDENT ANALYSIS The resul ts of the complete evaluation of the loss-of-coolant accident for Vertnont Yankee as documented in Reference 21 pr ovide required support for the operation of Vermont Yankee Cycle 9. The MAPLHGR limits as a function of average planar exposure are provided for each fuel type in Table A.2 of Appendix A.

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1 APPENDIX A iI i

PLANT 'IECHNICAL SPECIFICATION CHANGES j The proposed operating, limits changes to the Plant Technical I

! Specifications for Vermont Yankee Cycle 9 operation are documented in the Appendix.

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M M M M -m M M M M M M M M M M M M TABLE A.1 VERMONT YANKEE NUCLEAR POWER STATION CYCLE 9 MCPR OPERATION LIMITS MCPR Operating Limit for Value of "N" in RBM Average Control Rod Cycle Fuel Type (2)

Equa tion (l ) Scram Time Exposure Range 8X8 8X8R P8X8R 42% Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1,29 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.29 1.29 1.29 3.3 C.l.1 EOC-1 GWD/T to EOC 1.29 1.29 1.29 Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1.29 th an L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.30 1.30 1.30 3.3 C.I.2 EOC-1 GWD/T to EOC 1.33 1.32 1.32 41% Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.25 1.25 1.25 3.3 C.l.1 EOC-1 GWD/T to EOC 1.27 1.27 1.27 Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 th an L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.30 1.30 1.30

' 3.3 C.l.2 EOC-1 GWD/T to EDC 1.33 1.32 1.32 o $40% Equal or better BOC to EOC-2 GWD/T 1.24 1.24 1.24 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.24 1.23 1.23 3.3 C.I.1 EOC-1 GWD/T to EOC 1.27 1.27 1.27 Equal or better BOC to EOC-2 GWD/T 1.24 1.24 1.24 th an L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.30 1.30 1.30 3.3 C.I.2 EOC-1 GWD/T to E0C 1.33 1.32 1.32 75% Special Testing at Natural Circulation (Note 3, 4) 1.30 1.31 1.31 (1) The Rod Block Monitor (RBM) trip secpoints are determined by the equation shown in Table 3.2.5 of the Technical Speci f ica t ions.

(2) The current analysee for MCPR Operating Limits do not include 7X7 fuel. On this basis further evaluation of MCPR opi. mti ng limits is required before 7X7 fuel can be used in Reactor Power Operation.

(3) For the duration of pump trip and stability testing.

(4) Kg factors are not applied during the pump trip and stability testing.

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, I I TABLE A.2

~

I VERMONT YANKEE MAPLHGR OPERATING LIMITS VERSUS AVERAGE PLANAR EXPOSURE

• I Average Planar MAPLHGR by Fuel Type (kw/ f t )

I Exposure (mwd /t) 8D219 8D274 (High Gd) 8DRB289 P8DRB289 200 11 .4 11.1 11.2 I 1,000 5,000 11.5 11.9 11.1 11.6 11.2 11.8 10,000 12.1 12.1 12.0 15,000 12.3 12.2 12.1 20,000 12.1 12.1 11.8 25,000 11.3 11.6 11.3 30,000 10.2 10.6 11.1 35,000 9.6 10.0 10.4 40,000 - 9.4 9.8 I

• From Re ference 21 I I I

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I REFERENCES

1. B. Bu teau , " Stability and Recirculation Pump Trip Test, Special Test Procedure No. 81-01", Vermont 'lankee Nuclear Power Corporation, February 1981.

2. General Electric Boiling Water Reactor Generic Reload Fuel Application, NEDE-24011-P , GE Company Proprietary, July 1979.

3. D. M. VerPlanck , Methods for the Analysis of Boiling Water Reactors S teady Sta te Core Physics , YASC-1238, March 1981.

4. E. E. Pil a t , Me th od s for the Analysis of Boiling Water Reactors Lattice Physics , YAEC-1232, December 1980.

5. S. P. Schul tz and K. E. S t . John , Me th od s for the Analysis of Oxide Fuel Rod Steady-Sta te Thermal Effects (FROSSTEY) Code /Model Description Manual , YAE C-1249P, Apr il 1981.

6. S. P. Schul tz and K. E. St . John , Me th od s for the Analysis of Oxide Fuel Rod Steady-Sta te Thermal Effects (FROSSIEY) Code Qualifica tion and Applica tion , YAEC-1265P, June 1981.

7. D. C. Albright, H20DA: An Improved Wa ter Proper ties Package, YAEC-1237, March 1981.

8. Appendix A to Operating License DPR-28 Technical Specifications and Bases f or Vermont Yankee Nuclear Power Sta tion, Docket No. 50-271.

9. A. A. F. Ansar i, Meth od s for the Analysis of Boiling Water Reactors:

Steady-Sta te Core Flow Dis tribu tion Code (FIEWR), YAEC-1234, December 1980.

10. A. A. F. Ansari, R. R. Gay, and B. J. Gitnick , FIBWR: A Steady-State Core Flow Dis tribu tion Code for Boiling Water Reactor s - Code Verification and Qualification Report, EPRI NP-1923, Project 1754-1 Final I 11.

Re por t , J uly 1981.

General Electric Compacy, GEXL Correlation Application to BWR 2-6 Rea ctor s , NEDE-25422, GE 6.cipev Proprietary, June 1981.

12. A. A. F. Ansari and J. T. Cronin, Methods for the Analysis of Boiling I Wa ter Rea c tor s : A Sys tems Tr ans ient Analysis Model (RETRAN), YAEC-1233, Apr il 1981.

13. EPRI, RETRAN - A Program for One-Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, CCM-5, December 1978.

14. K. J. Burns , Meth od s for the Analysis of Boiling Water Reactors:

I Transient Thermal Margin Analysis Code (MAYU04-YAEC), YAEC-1235, December 1980.

- 9. .

15. J. M. Holzer , Mathods for the Analysis of Boiling Water Reactors Transient Core Physics, YAEC-1239P, August 1981.

16. Vermont Yankee Nuclear Power Station Final Safety Analysis Report, Section 3.6.6

17. C. J. Paone , et.al . , Rod Drop Accident Anal ysis for Large Boiling Water Reactors , NEDO-10527, March 1972.

I 18. C. J. Paone, Banked Position Withdrawal Sequence, NED0-21231, J anuary , 1977.

R. C. Stirn, et.al., Rod Drop Accident Analysis for Large Boiling Water I ~ 19.

Reactor Addendum No. 2 Exposed Cores, NEDO-10527, Supplement 2, January 1973.

20. Letter , FVY 81-52, dated March 31, 1981, R. L. Smith to Boyce d. Grier of USNRC Region 1, " Cycle VIII Startup Test Re por t" .

I 21. Loss-of-Coolant Accident Analysis for Vernont Yankee Nuclear Power Sta tion, NEDO-21697, Augus t 1977, as amended.

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