Text

Cycle 11 Core Performance Analysis
	ML20084S466
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Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/12/1984
From:	Paul Bergeron, Cacciapouti R, Stephen Schultz; YANKEE ATOMIC ELECTRIC CO.
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VERMONT YANKEE CYCLE 11 CORE PERFORMANCE ANALYSIS April 1984 Major Contributors:

D. C. Albright D. M. Kapitz B. G. Baharynejad J. Pappas K. J. Burns K. E. St. John J. T. Cronin M. A. Sironen D. P. Heinrichs R. A. Woehlke Approved by:

R.p.Cacc4e/outi, Manager ( Ddt e )'

Reactor Phttics Group Approved by: M. Co h k/2!M m y

P.A.Bergeron[anager (Date) r Transient Analfsis Group Approved by: '

I /82 /8 4-S. P' Schultz, Marfager % (Date)

Nuclear Evaluations and Support Group Approved by: f I A. Husain, Mankger '(Date)

LOCA Group Approved by: \ .

B. C."Slifer, Director ()

mz/n (Date)

Nuclear Engineering DepaMment 8405250088 840517 DR ADOCK 05000

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

With regard to any unauthorized use whatsoever, Yankee Atomic Electric Company, Vermont Yankee Nuclear Power Corporation and their officers, directors, agents and employees assume no liability nor make any warranty or representation with respect to the contento of this document or to its accuracy or completeness.

f ABSTRACT l This report presents design information and calculational results pertinent to the operation of Cycle 11 of the Vermont Yankee Nuclear power Station. These include the fuel design and core loading pattern descriptions; calculated reactor power distributions, exposure distributions, shutdown capability and reactivity functions; and the results of safety analyses performed to justify plant operation throughout the cycle.

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TABLE OF CONTENTS Eard!

DISCLAIMER.................................................. 11 '

ABSTRACT.................................................... iii j I

i TABLE OF CONTENTS........................................... iv LIST OF FIGURES............................................. Vi LIST OF TABLES.............................................. Vill ACKN0WLEDGEMENTS............................................ ix

1.0 INTRODUCTION

................................................ 1 2.0 RECENT REACTOR OPERATING HIST 0RY............................ 2 2.1 Operating History of the Current Cycle................. 2 2.2 Operating History of Past Applicable Cycles............ 2 3.0 RELOAD CORE DESIGN DESCRIPTION.............................. 6

. 3.1 Core Fuel Loading...................................... 6 3.2 Design Reference Core Loading Pattern.................. 6 3.3 Assembly Exposure Distribution......................... 6 4.0 FUEL MECHANICAL AND THERMAL DESIGN.......................... 9 4.1 Mechanical Design...................................... 9 4.2 Thermal Design......................................... 9 4.3 Operating Experience................................... 10 5.0 NUCLEAR DESIGN.............................................. 15 5.1 Core Power Distributions............................... 15 5.1.1 Haling Power Distribution....................... 15 5.1.2 Rodded Depletion Power Distribution............. 15 5.2 Core Exposure Distributions............................ 16 5.3 Cold Core Reactivity and Shutdown Margin............... 16 I 5.4 Standby Liquid Control System Shutdown Capability...... 17 6.0 THERMAL-HYDRAULIC DESIGN.................................... 26 6.1 Steady-State Thermal Hydraulics........................ 26 6.2 Reactor Limits Determination........................... 26

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TABLE OF CONTENTS (Continued)

P.are 7.0 ACCIDENT ANALYSIS........................................... 28 7.1 Core Wide Transient Analysis........................... 28 7.1.1 Methodology..................................... 28 7.1.2 Initial Conditions and Assumptions.............. 29 7.1.3 Reactivity Functions............................ 30 7.1.4 Transients Analyzed............................. 32 7.2 Core Wide Transient Analysis Results................... 32 7.2.1 Turbine Trip Without Bypass Transient........... 32 7.2.2 Generator Load Rejection Without Bypass Transient................................ 33 7.2.3 Loss of Feedwater Heating Transient............. 33 7.3 Overpressurization Analysis Results . . . . . . . . . . . . . . . . . . . . 34 7.4 Local Rod Withdrawal Error Transient Results........... 34 7.5 Mistoaded Bundle Error Analysis Results................ 37 7.5.1 Rotated Bundle Error............................ 37 7.5.2 Mislocated Bundle Error......................... 38 7.6 Control Rod Drop Accident Results...................... 39 7.7 Stability Analysis Results............................. 40 8.0 STARTUP PR0 GRAM............................................. 82 9.0 LOSS-OF-COOLANT ACCIDENT ANALYSIS........................... 83 APPENDIX A CALCULATED CYCLE DEPENDENT LIMITS............... 84 REFERENCES.................................................. 87

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LIST OF FIGURES Number Title Page 3.2.1 VY Cycle 11 Design Reference Loading Pattern, Lower Right 8

. Quadrant 4.2.1 VY Cycle 11 Core Average Cap Conductance versus Cycle 13 Exposure 4.2.2 VY Hot Channel Cap Conductance for P8X8R versus Exposure 14 5.1.1 VY Cycle 11 Haling Depletion, EOFPL Bundle Average 19 Relative Powers 5.1.2 VY Cycle 11 Core Average Axial Power Distribution Taken 20 from the Haling Calculation to EOFPL 5.1.3 VY Cycle 11 Rodded Depletion - ARO at EOFPL 21 Bundle Average Relative Powers 5.1.4 VY Cycle 11 Core Average Axial Power Distribution, 22 Rodded Depletion - ARO at EOFPL 5.2.1 VY Cycle 11 Haling Depletion, EOFPL Bundle Average Exposures 23 5.2.2 VY Cycle 11 Rodded Depletion EOFPL Bundle Average Exposures 24 5.3.1 VY Cycle 11 Cold Shutdown Delta K in Percent versus Cycle 25 Exposure 7.1.1 Flow Chart for the Calculation of ACPR Using the 46 RETRAN/TCPYA01 Codes 7.1.2 Inserted Rod Worth and Rod Position versus Time From 47 Initial Rod Movement at EOFPL11. " Measured" Scram Time 7.1.3 Inserted Rod Worth and Rod Position versus Time From 48 Initial Rod Movement at EOFPL11-1000 MWD /ST, " Measured" Scram Time 7.1.4 Inserted Rod Worth and Rod Position versus Time From 49 Initial Rod Movement at EOFPL11-2000 MWD /ST, " Measured" Scram Time -

7.1.5 Inserted Rod Worth and Rod Position versus Time From 50 Initial Rod Movement at EOFPL11. "67B" Scram Time 7.1.6 Inserted Rod Worth and Rod Position versus Time From 51 Initial Rod Movement at EOFPL11-1000 MWD /ST, "67B" Scram Time 7.1.7 Inserted Rod Worth and Rod Position versus Time From 52 Initial Rod Movement at EOFPL11-2000 MWD /ST, "67B" Scram Time

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LTST OF FfGURES [

Number Title Page E55

_ my 7.2.1 Turbine Trip Without Bypass, EOFPL11 Transient 53 Response versus Time, " Measured" Scram Time Turbine Trip Without Bypass, EOFPL11-1000 NWD/ST 56 - "

7.2.2 --

Transient Response versus Time, " Measured" Scram Time 59

=

7.2.3 Turbine Trip Without Bypass, EOFPL11-2000 MWD /ST Transient Response versus Time, " Measured" Scram Time 62 i 7.2.4 Cenerator Load Rejection Without Bypass, EOFPL11 _.

Transient Response versus Time, " Measured" Scram Time ri 7.2.5 Cenerator Load Rejection Without Bypass, EOFPL11-1000 MWD /ST 65 c--

Transient Response versus Time, " Measured" Scram Time ,

--k 7.2.6 Generator Load Rejection Without Bypass, EOFPL11-2000 Y1D/ST 68 i Transient Response versus Time, " Measured" Scram Time _ __

7.2.7 Loss of 100 F0 Feedwater Heating, EOFPL11-2000 MWD /ST 71 '

(Limiting Case) Transient Response versus Time

._ {

7.3.1 MSIV Closure, Flux Scram, EOFPL11 Transient Response 73 x; versus Time, " Measured" Scram Time (_,

7.4.1 Reactor Initial Conditions for the VY Cycle 11 76 Rod Withdrawal Error Case 1 _

7.4.2 Reactor Initial Conditions for the VY Cycle 11 77 Rod Withdrawal Error Case 2 7 1

7.4.3 VY Cycle 11 RWE Case 1 - Setpoint Intercepts Determined 78 --

by the A+C Channel s_

7.4.4 VY Cycle 11 RWE Case 1 - Setpoint Intercepts Determined 79 by the B+D Channel _

7.6.1 First Four Rod Arrays Pulled in the A Sequences 80 ---

7.6.2 First Four Rod Arrays Pulled in the B Sequences 80 ,

7.7.1 VY Cycle 11 Reactor Core Decay Ratio versus Power 81

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LIST OF TABLES Number Title Page 4

3.1.1 VY Cycle 10 Operating Highlights 3 2.2.1 VY Cycle 9 Operating Highlights 4 3.2.2 VY Cycle 8 Operating Highlights 5 3.1.1 VY Cycle 11 Fuel Bundle Types and Nunbers 7 3.3.1 Design Basis VY Cycle 10 and Cycle 11 Exposures 7 4.1.1 Nominal Fuel Mechanical Design Parameters 11 4.2.1 Cap Conductance Values Used in VY Cycle 11 Transient Analyses 12 5.3.1 VY Cycle 11 K-Effective Values and Shutdown Margin 18 Calculation 5.4.1 VY Cycle 11 Standby Liquid Control System Shutdown Capability 18 7.1.1 VY Cycle 11 Summary of System Transient Model Initial 41 Conditions for Core Wide Transient Analyses 7.1.2 VY Cycle 11 Transient Analysis Reactivity Coefficients 42 at Selected Conditions 7.2.1 VY Cycle 11 Core Wide Transient Analysis Results 43 7.3.1 VY Cycle 11 Overpressurization Analysis Results 44 7.4.1 VY Cycle 11 Rod Withdrawal Error Transient Summary 44 With Limiting Instrument Failure 7.5.1 VY Cycle 11 Rotated Bundle Analysis Results 44 7.6.1 Control Rod Drop Analysis - Rod Array Pull Order 45 7.6.2 VY Cycle 11 Control Rod Drop Analysis Results 45 7.7.1 VY Cycle 11 Stability Analysis Results 45 A.1 Vermont Yankee Nuclear Power Station Limiting Cycle 11 85 MCPR Results A.2 Vermont Yankee Nuclear Power Station Technical 86 Specification MCPR Operating Limits

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ACKNOWLEDGEMENTS The authors and principal contributors would like to acknowledge the centributions to this work by W. J. Watetuan, S. B. Bowman, and the YAEC Word Processing Center. Their assistance in preparing figures and text for this

'd:cument.is recognized and greatly appreciated.

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1.0 INTRODUCTION

This report provides information to support the operation of the Vermont Yankee Nuclear Power Station through the forthcoming Cycle 11. In this report, Cycle 11 will frequently be referred to as the Reload Cycle. The preceding Cycle 10 will frequently be referred to as the current Cycle. The refueling between the two will involve the discharge of 104 irradiated fuel bundles and the insertion of 104 new fuel bundles. The resultant core will consist of 104 new fuel bundles and 264 irradiated fuel bundles. Some of the irradiated fuel was present in the reactor in Cycles 8 and 9, as well as the Current Cycle. These cycles will frequently be referred to as Past Cycles.

This report contains descriptions and analyses results pertaining to the mechanical, thermal-hydraulic, physics, and safety aspects of the Reload Cycle (Cycle 11).

The cycle-dependent operating limits calculated for the Reload Cycle cre bounded by the Vermont Yankee Plant Technical Specifications. Both are given in Appendix A.

2.0 RECENT REACTOR OPERATING HISTORY 2.1 Operating History of the Current Cycle The currently operating cycle is Cycle 10. During the current Cycle, the reactor has operated smoothly at, or near, full power with the exception of normal maintenance, sequence exchanges, and a few scrams. The operating history highlights and control rod sequence exchange schedule of the current cycle are found on Table 2.1.1.

2.2 Operating History of Past Applicable Cycles The irradiated fuel in the Reload Cycle includes some fuel bundles initially inserted in Cycles 8 and 9. These Past Cycles operated smoothly at, or near, full power with the exception of normal maintenance, sequence exchanges, a few scrams, and coastdown to the end of cycle. The highlights of the Past Cycles are found in Tables 2.2.1 and 2.2.2. The Past Cycles are described in detail in References 1 and 2.

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TABLE 2.1.1 VY CYCLE 10 OPERATING HIGHLIGHTS Beginning of Cycle Date June 17, 1983 End of Cycle Date June 15, 1984*

Weight of Uranium As-Loaded (Short Tons) 74.13 Beginning of Cycle Core Average Exposure (NWD/ST) 10463.

End of Full Power Core Average Exposure (NWD/ST) 17188.*

End of Cycle Core Average Exposure (NWD/ST) 17963.*

Bumber of Fresh Assemblies 108 Number of Irradiated Assemblies 260 Control Rod Sequence Exchange Schedule:

Sequence Date From Io August 13, 1983 Al-1 B2-1 October 1, 1983 B2-1 A2-1 November 5, 1983 A2-1 B1-1 December 17, 1983 B1-1 Al-2 January 24, 1984 Al-2 B2-2 March 3, 1984 B2-2 A2-2 April 14, 1984* A2-2 B1-2 3

Projected Dates and Exposures.

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TABLE 2.2.1 VY CYCLE 9 OPERATING HIGHLIGHTS Beginning of Cycle Date December 1, 1981 End of Cycle Date March 5, 1983 Weight of Uranium As-Loaded (Short Tons) 74.15 Beginning of Cycle Core Average Exposure (MWD /ST) 9192.

End of Full Power Core Average Exposure (MWD /ST) 16595.

End of Cycle Core Average Exposure (MWD /ST) 18137.

Capacity Factor While Operating (%) 90.9 Number of Fresh Assemblies 120 Number of Irradiated Assemblies 248 Control Rod Sequence Exchange Schedule:

Sequence Date f_rpom r To January 28, 1982 Al-1 B2-1 March 13, 1982 B2-1 A2-1 April 24, 1982 A2-1 B1-1 June 10, 1982 B1-1 Al-2 July 24, 1982 Al-2 B2-2 September 11, 1982 B2-2 A2-2 October 30, 1982 A2-2 B1-2

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TABLE 2.2.2 VY CYCLE 8 OPERATING HIGHLIGHTS Beginning of Cycle Date December 23, 1980 End of Cycle Date October 16, 1981 Weight of Uranium As-Loaded (Short Tons) 74.34 Beginning of Cycle Core Average Exposure (MWD /ST) 10888.

End of Full Power Core Average Exposure (MWD /ST) 16211.

End of Cycle Core Average Exposure (MWD /ST) 16794.

Capacity Factor While Operating (%) 94.1 Number of Fresh Assemblies 80 Number of Irradiated Assemblies from Cycle 7 276 Number of Assemblies Reinserted from the Spent Fuel Pool 12 Control Rod Sequence Exchange Schedule:

Sequence Date From To March 14, 1981 Al B2 May 12, 1981 B2 A2 August 1, 1981 A2 B1

d 3.0 RELOAD CORE DESIGN DESCRIPTION ._

5 3.1 Core Fuel Loading The Reload Cycle core will consist of both new and irradiated '

essemblies. All the assemblies have bypass flow holes drilled in the lower tie plate. Table 3.1.1 characterizes the core by fuel type, batch size, and first cycle loaded. A description of the fuel is found in Reference 3.

s 3.2 Design Reference Core Loading Pattern The Reload Cycle assembly locations are indicated on the map in Figure f 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. If any changes are made to the loading pattern at the time of refueling, they will be checked and verified acceptable under 10CFR50.59. The final loading pattern with specific ,

bundle serial numbers will be supplied with the Startup Test Report.

3.3 Assembly Exposure Distribution The assumed nominal exposure on the fuel bundles in the Reload Cycle design reference loading pattern is given in Figure 3.2.1. To obtain this _

exposure distribution, Past Cycles were depleted with the SIMULATE model [4,5) --

using actual plant operating history. For the Current Cycle, plant operating _

history was used through August 16, 1993. Beyond this date, the exposure was accumulated using a best-estimate rodded depletion analysis to End of Full -

Power Life (EOFPL) followed by a projected coastdown to End of Cycle (EOC). -'

Table 3.3.1 gives the assumed nominal exposure on the Current Cycle and .

the Beginning of Cycle (BOC) core average exposure that results from the shuffle into the Reload Cycle loading pattern. The Reload Cycle EOFPL core everage exposure and cycle capability are provided.

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TABLE 3.1.1 VY CYCLE 11 FUEL BUNDLE TYPES AND NUMBERS l Fuel Reload Cycle Possible - _ .

Desirnation Desir. nation Loaded Number Bundle ID's -

IRRADIATED P8DPB289 R7 8 36 LJPKXX, LJUKKX P8DPB289 R8 9 120 LJTKXX, LJZKKX _ .

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P8DPB289 R9 10 108 LY4KXX NEW P8DPB289 RIO 11 104 LY6KXX, LY7KXX 6

NOTE: KXX stands for the last three digits of the bundle serial number.

TABLE 3.3.1 ~

DESIGN BASIS VY CYCLE 10 AND CYCLE 11 EXPOSURES Assumed Current Cycle Core Average Exposure End of Cycle 10 17.96 GWD/ST Assumed Reload Cycle Core Average Exposure Beginning of Cycle 11 10.62 GWD/ST Haling Calculated Core Average Exposure at End of Full Power Life, Cycle 11 17.48 GWD/ST Cycle 11 Exposure Capability 6.96 CWD/ST 4

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VERMONT YANKEE CYCLE 11 BOC BUNOLC AVERADE EXPOSURES R8 R9 R8 R8 RIO R8 R9 R10 R8 R10 RB

--22 19 64 8.75 19 50 15 93 0.00 18 56 8.63 0.00 16 27 0.00 16.42 R9 R8 R10 R9 R8 R9 R10 R9 R10 R9 R7

-- 2 0 8 68 16.88 0 00 8 54 18 80 9 51 0 00 9 38 0 00 9 39 22.25 R8 R10 R8 RIO R9 R8 R9 RIO R8 R9 R7

--18 19.55 0 00 16 97 0 00 9 13 20 19 9 60 0 00 16 87 9.78 22 32 R8 R9 R10 R8 R10 R9 R10 R9 R9 R7 16 15.95 8.44 0.00 16.83 0.00 9.22 0 00 7.86 9.75 22 16 R10 R8 R9 R10 R8 R10 R8 R10 R8 34 0 00 18 73 9 72 0 00 17 25 0.00 17 24 0 00 18 40 R8 R9 R8 R9 R10 R9 R10 R9 R8 12 18 56 9 81 20 05 9 02 0 00 7 02 0 00 9.63 17 57 R9 RIO R9 R10 RB R10 RB R8 R7 10 8 54 0 00 9 34 0 00 17 38 0 00 19.76 16 20 23 15 RIO R9 R10 R9 R10 R9 R8 R7 g

0.00 9 55 0 00 7 78 0 00 9 51 16 35 21.21 R8 RIO R8 R9 R8 R8 R7 6

16 20 0.00 17.09 9 50 18 26 17 58 23 37 l

R10 R9 R9 R7

--- 4 0.00 9 25 9 62 22 24 R7 - P8DPB289, RELOAD 7 R8 - P8DPB289, RELOAD 8 R9 - P8DPB289 RELOAD 9 R8 R7 R7 8UNOLE 10 R10 - P8DPB289. RELOAD 10 --- 2 16.36 22 29 22 18 EXPOSURE (DWO/ST1 I i I i i 1 23 25 27 29 31 33 35 37 39 41 43 FIGURE 3.2.1 VY CYCLE 11 DESIGN REFERENCE LOADING PATTERN, lok'ER RIGHT OUADRANT

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4.0 yUEL MECHANICAL AND THERMAL DESIGN 4.1 Mechanical Design All fuel to be inserted into the Reload Cycle was fabricated by the General Electric Company (GE). The major mechanical design parameters are given in Table 4.1.1. Detailed descriptions of the fuel rod mechanical design and mechanical design analyses are provided in Reference 3. These design analyses renmin valid with respect to the Reload Cycle operation. Mechanical and chemical compatibill'y of the fuel assemblies with the in-service reactor environment is also adaressed in Reference 3.

4.2 Thermal Design The fuel thermal effects calculations were performed using the FROSSTEY computer code [6-8). The FROSSTEY code calculates pellet-to-clad gap conductance and fuel temperatures from a combination of theoretical and empirical models which include fuel and cladding thermal expansion, fission gas release, pellet swelling, pellet densification, pellet cracking, and fuel and cladding thermal conductivity.

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

The gap conductance values input to the hot channel calculations (Section 7.1) were evaluated for the given fuel bundle type as a function of the assembly exposure. The calculation assumed a 1.4 chopped cosine axial power shape with the peak power node running at the MApLHGR limit defined in Reference 9. Figure 4.2.2 provides the hot channel response of gap

conductance. In Figure 4.2.2, " planar exposure" refers to the exposure of the node running at the MAPLHGR limit. Gap conductance values for the hot channel analysis were extracted from Figure 4.2.2 using the maximum bundle exposure of any MCPR limiting bundle within the exposure interval of interest. The SINULATE rodded depletion (Section 5.1.2) provides predictions of both limiting MCPR and the associated bundle exposure for the entire cycle.

Table 4.2.1 provides the core average and hot channel gap conductance values used in the transient analyses (Section 7.1).

Fuel rod local linear heat generation rates at fuel centerline incipient melt and 1% clad plastic strain as a function of local axial segment cxposure for the gadolinia concentrations used in Vermont Yankee fuel were previously reported in Reference 10.

4.3 Operatint 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 are at normally low levels indicating that no fuel failures are present.

TABLE 4.1.1 NOMINAL FUEL MECHANICAL DESIGN PARAMETERS FUEL TYPE P8K8R Vendor Designations (Table 3.1.1)

New P8DPB289 Irradiated P8DPB289 Fuel Pellets Fuel Material (sintered Pellets) UO2 Initial Enrichment, w/o U-235 2.89 Pellet Density, % theoretical 95.0 Pellet Diameter, inches 0.410 Fuel Rod Active Length, inches 150.0 Plenum Length, inchei 9.5 Fuel Rod Pitch, inches 0.640 Diametral Gap (cold), inches 0.009 Fill Gas Helium Fill Gas Pressure, psig [See Ref. 3)

Cladding Material Zr-2 Outside Diameter, inches 0.483 Thickness, inches 0.032 Inside Diameter, inches 0.419 Fuel Channel Material Zr-4 Inside Dimension, inches 5.278 Wall Thickness, inches 0.080 Fuel Assembly Fuel Rod Array 8x8 Fuel Rods per Assembly 62 Spacer Grid Material Zr-4

TABLE 4.2.1

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GAP CONDUCTANCE VALUES USED IN VY CYCLE 11 TRANSIENT ANALYSES Hot Channel 4 Cycle Exposure Core Average Hot Channel Statepoint Cap Conductance Bundle Exposure Gap Conductance

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(MWD /ST) (BTU /Hr-Ft - F) (MWD /ST) (BTU /Hr-Ft - F)

BOC11 755 11240(1) 1490 EOFPL11-2000 MWD /ST 980 6320 1040 6 EOPPL11-1000 MWD /ST 995 7460 1120 1010 8720 1250 EOFPL11 l(

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(1) Between BOC11 and EOFPL11-2000 MWD /ST, the highest exposure limiting hot '

channel bundle is once-burned. '

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VY CYCLE 11 FUEL. PERFORMANCE CORE AVERAGE GRP CONDUCTRNCE 1050

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4 5.0 NUCLEAR DESIGN

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! 5.1 Core Power Distributions The Reload Cycle was depleted using SINULATE [4] to give both a rodded depletion and an All Rods Out (ARO) Haling depletion. The Haling depletion

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cerves as the basis for defining core reactivity characteristics for most ttansient evaluaticns. This is primarily because its flat power shape has conservatively weak scram characteristics. The rodded depletion was used to

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, cvaluate the misloaded bundle error and the rod withdrawal error. This is because of the more realistic predictions it makes of initial CPR values. It ges also used in the rod drop worth and shutdown margin calculations because it burns the top of the core more realistically than the Haling.

5.1.1 Haling Power Distribution I

The Haling power distribution is calculated in the All Rods Out (ARO) condition. The Haling iteration converges on a self-consistent power and sxposure shape for the burnup step to EOFPL. In principle, this should provide the overall minimum peaking power shape for the cycle. During the cetual cycle, flatter power distributions might occasionally be achieved by chaping with control rods. However, such shaping would leavs underburned f regions in the core which would peak at another point in time. Figures 5.1.1 cnd 5.1.2 give the Haling radial and axial average power distributions for the Reload Cycl'e.

5.1.2 Rodded Depletion Power Distribution To generate the rodded depletion, control rod patterns were developed which gave critical eigenvalues at each point in the cycle and gave peaking similar to the Haling calculation. The resulting patterns were frequently more peaked than the Haling, but were below expected operating limits.

However; es stated above, the underburned regions of the core can exhibit peaking'in excess of the Haling peaking when pulling ARO at EOFPL. Figures 5.1.3 and 5.1.4 give the ARO at EOFPL power distributions for the Reload Cycle rodded depletion. Note in Figure 5.1.4 that the average axial power at ARO i

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

' 5.2 Core Exposure Distributions l l

l l The Reload Cycle exposures are summarized in Table 3.3.1. Coastdown is l

nst included. The projected BOC radial exposure distribution for the Reload

( Cycle is given in Figure 3.2.1. The Haling calculation produced the EOFPL rrdial exposure distribution given in Figure 5.2.1. Since the Haling power chape is constant, it can be held fixed by SINULATE to give the exposure distributions at various mid-cycle points. BOC, EOFPL-2000 MWD /ST, EOFPL-1000 NWD/ST, and EOFPL exposure distributions were used to develop reactivity input for the core wide transient analyses.

The rodded depletion differs from the Haling during the cycle due to power shaping by the rods. However, rod sequences are swapped frequently and the overall exposure distribution at end of cycle is similar to the Haling.

Figure 5.2.2 gives the EOFPL radial exposure distribution for the Reload Cycle radded depletion.

5.3 Cold Core Reactivity and Shutdown Martin The cold K,gg with ARO and the cold K,gg with All Rods Inserted (ARI) at BOC were calculated using the SINULATE code [4,5) and are shown in Table 5.3.1. K,gg with ARO minus the cold critical K gg is the amount of cxcess core reactivity. K,gg with ARI minus the K gg with ARO is the worth of all the control rods.

The cold critical eigenvalue K,gg was defined as the average calculated critical eigenvalue minus a 95% confidence level uncertainty. Then oil cold results were normalized to make the criticci K,gg eigenvalue equal j 1

to 1.000.

Technical Specifications _[9] state that, for' sufficient shutdown margin, the core must be suberitical' by at least 0.2f% AK + R (defined below) with the strongest worth control rod withdrawn. Again, using SINULATE, a

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ccarch was made for the strongest worth control rod at various exposures in the cycle. This is necessary because rod worths change with exposure. Then the cold K,gg with the strongest rod out was calculated at BOC and at the cnd of each control rod sequence. Subtracting each cold K,gg with the otrongest rod out from the cold critical K,gg eigenvalue defines the chutdown margin as a function of exposure. Figure 5.3.1 shows the result.

B:cause the local reactivity may increase with exposure, the shutdown margin (SDN) may deerosse. To account for this, and other uncertainties, the value R l 10 calculated. R is defined as R plus R . R is the difference l between the cold K,gg with the strongest rod out at BOC and the maximum cold K with the strongest rod out in the cycle. R is a measurement gg uncertainty in the demonstration of SDN. It is presently set at .07% AK. The I shutdown margin results are summarized in Table 5.3.1. ,

t 5.4 Standby Liquid control System Shutdown capability The shutdown capability of the Standby Liquid Control System (SLCS) is dssigned to bring the reactor from full power to cold, ARO, xenon free shutdown with at least 5% AK margin. Using the boron concentration search cption in SINULATE [4], the ppm of boron was adjusted until the K,gg reached the cold critical K, g minus .05. This case assumed cold, xenon free conditions, with All Rods Out at the most reactive time in the cycle. The criticality search found that the plant would be suberitical by 5% AK at the worst point in time with less than the 800 ppm of boron required by VY Technical Specifications (9). Table 5.4.1 lists the amount of boron concentration and the corresponding shutdown cepability of the SLCS.

d a

s

TABLE 5.3.1 VY CYCLE 11 VALUES AND SHUTDOWN MARGIN CALCULATION l

BOC K,gg - Uncontrolled 1.1172 BOC K,gg - Controlled .9700 Cold Critical K gg e Eigenvalue 1.0000 BOC eK gg - Controlled With .9881 l Strongest Worth Rod Withdrawn Cycle Minimum Shutdown Margin Occurs at 1.19% AK BOC With Strongest Worth Rod Withdrawn R1 , Maximum Increase in Cold K,gg .00% AK-With Exposure TABLE 5.4.1 VY CYCLE 11 STANDBY LIOUID CONTROL SYSTEM SHUTDOWN CAPABILITY ppm of Boron Shutdown Martin 680 .050 AK 800 .073 AK t

'I i

VERMONT YANKEE CYCLE 11 HRLING DEPLETION E0FPL BUNDLE AVERADE RELATIVE POWERS R8 R9 R8 R8 R10 R8 R9 R10 R8 R10 R8

- 22 0.98 1 13 1.04 1 10 1 32 1.08 1 22 1 29 0 96 0.04 0.52 R9 R8 RIO R9 R8 R9 R10 R9 R10 R9 R7

- 20 1 13 1 07 1.32 1 22 1 09 1 20 1 32 1 17 1 13 0.83 0.45 R8 Rio R8 R10 R9 R8 R9 R10 R8 R9 R7

- 18 1 04 1 32 1 13 1 36 1 23 1.07 1 19 1 23 0.88 0.71 0.37 R8 R9 RIO R8 Rio R9 R10 R9 R9 R7 16 1 10 1 22 1 36 1 15 1 36 1 21 1 28 1 07 0 84 0 30 R10 R8 R9 R10 R8 R10 RB R10 R8 g4 1 32 1 09 1.22 1 36 1 12 1 30 1 01 1 05 0 64 R8 R9 R8 R9 R10 R9 R10 R9 R8 12 1 08 1.19 1 07 1 22 1.30 1 16 1 12 0.84 0 52 R9 RIO R9 R10 R8 R10 R8 R8 R7 10 1 22 1.32 1 19 1.28 1 01 1 12 0.78 0 61 0 37 R10 R9 R10 R9 R10 R9 RB R7 8

1.29 1.16 1 23 1.07 1 05 0 85 0.61 0.41 R8 R10 R8 R9 R8 R8 R7 6

0 97 1.13 0.87 0.84 0.65 0 52 0.36 R10 R9 R9 R7 R7 - P8DPB289, RELOAD 7 _4 0.95 0 83 0.71 0.50 RS - P8DPB289, RELOAD 8 R9 - P8DPB289, RELOAD 9 R8 R7 R7 8UNDLE 10 R10 - P8DPB289, RELOAD 10 -2 0 52 0 45 0.37 RELRTIVE POWERS l l l l l l 23 26 27 29 31 33 35 37 39 41 43 FIGURE 5.1.1 VY CYCLE 11 HALING DEPLETION, EOFPL BUNDLE AVERAGE RELATIVE POWERS

VY CYCLE 11 CORE AVERAGE RXIRL POWER DISTRIBUTION TRKEN FROM THE HALING CRLCULRTION TO EOFPL 1.5 .

1.4 1.3 1.2 I N e- i x 4 $ / \\

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= 0.7 ,/

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E 0.4 0.3 0.2 0.1 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 PLRNT AND SIMULATE RXIAL NODE (24-TOP)

FIGURE 5.1.2 VY CYCI E 11 CORE AVERAGE AXIAL POWER DISTRIBUTTON TAKEN FROM Tile IIAT.ING CAT.Clif ATION TO EOFPL

~

VERMONT YANKEE CYCLE 11 RODDED DEPLETION EOFPL BUNDLE AVERAGE RELRTIVE POWERS l

R8 RS R8 R8 RIO R8 R9 RIO R8 RIO R8

--22 1.01 1.17 1.07 1 12 1.35 1.09 1 22 1.29 0.96 0 94 0 62 ,

R9 R8 R10 R9 R8 R9 R10 R9 R10 R9 R7

--20 1 16 1 10 1 36 1 24 1.10 1 20 1 33 1 16 1 12 0.82 0.44 R8 R10 R8 R10 R9 R8 R9 R10 R8 R9 R7 __gg 1.06 1 35 1.15 1.38 1.24 1 07 1 18 1.22 0.87 0 71 0.37 R8 R9 R10 R8 R10 R9 R10 R9 R9 R7 16 1 11 1 24 1 39 1.16 1.36 1 21 1 27 1 06 0.83 0.50 R10 R8 R9 R10 R8 RIO R8 RIO R8 g4 1 34 1.10 1 23 1 36 1.12 1 30 1 00 1 04 0 64 R8 R9 R8 R9 Rio R9 R10 R9 R8 12 1 08 1.19 1 06 1.21 1.29 1.14 1 10 0.83 0.61 A9 RIO R9 R10 R8 RIO R8 R8 R7 10 1.21 1.32 1 18 1 27 1 00 1 10 0 77 0.60 0.36 R10 R9 R10 R9 R10 R9 R8 R7 g

1.28 1.15 1.22 1.06 1.04 0.83 0.60 0.40 R8 RIO R8 R9 R8 R8 R7 6

0.96 1 12 0.87 0.83 0.64 0.61 0.36 R10 R9 R9 R7

--- 4 0.94 0.82 0.70 0.49 R7 - P8DFB289, RELOAD 7 R8 - P8DPB289, RELOAD 8 R9 - P8DPB289, RELOAD 9 R8 R7 R7 8UNOLE 10 R10 - P8DPB289, RELOAD 10 --- 2 0 52 0.44 0 37 RELATIVE POWERS I I I I I I 23 25 27 29 31 33 3G 37 39 41 43 FIGURE 5.1.3 VY CYCLE 11 RODDED DEPLETION-ARO AT EOFPL BUNDLE AVERAGE RELATIVE POWERS

VY CYCLE 11 CORE AVERAGE AXIAL POWER DISTRIBUTION RODDED DEPLETION -- ALL RODS OUT RT EOFPL 1.5 1.4 1.3 1.2 j# "'

- m 1.1 %

2 1'0 I g

E 0.9 / N 3 i b 0.8 Y E 0.7 /

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. PLANT AND SIMULRTE RXIRL NODE (24-TOP)

,FTCURE 5.1.4 VY CYCI.E 11 CORE AVERAGE AXTAL POWER DISTRIntrrTON, RODDED DEPLETION-ARO AT EOFPI,

VERMONT YANKEE CYCLE 11 HALING DEPLETION E0FPL BUNDLE AVERAGE EXPOSURES R8 R9 R8 R8 R10 R8 R9 R10 R8 R10 R8

- 22 28 39 16.52 26.82 23.45 9 06 25.98 18.97 8.84 22.89 6.48 20 01 I

R9 R8 R10 R9 R8 R9 R10 R9 R10 R9 R7

- 20 18.46 24.24 9.03 16.92 26 28 17.71 9.08 17.37 7.72 15.05 25.31 R8 Rio R8 R10 R9 R8 R9 R10 R8 R9 R7 _gg 26 67 9.03 24.72 9.35 17 58 27.52 17 73 8.41 22.89 14.66 24.87 R8 R9 R10 R8 R10 R9 R10 R9 R9 R7 16 23.47 16.83 9 34 24 53 9.33 17 55 8.74 15.21 15.48 25.60 R10 R8 R9 R10 R8 Rio R8 R10 R8 34 9 05 26 20 18.09 9.33 24 94 8 93 24 18 7 19 22 81 l R8 R9 R8 R9 R10 R9 R10 R9 R8 12 25 97 17.9S 27 39 17 37 8.94 14.96 7 66 15.42 21 13 R9 R10 R9 R10 R8 R10 R8 R8 R7 10 16.88 9.07 17.50 8.75 24.32 7.65 25 13 20.39 25.65 R10 R9 R10 R9 R10 R9 R8 R7 g 8.84 17 51 8.41 15 14 7 19 15 30 20.52 24.04 R8 R10 R8 R9 R8 R8 R7 6

22.82 7.71 23.08 15 26 22.69 21 14 25.86 R10 R9 R9 R7

-4 8.48 14.92 14.51 25.87 R7 - P8DPB289, RELOAD 7 R8 - P8DPB289, RELOAD 8 R8 R7 R7 SUNOLE 10 R10 - P8DPB289, RELOAD 10 -- 2 19 96 25 34 24.74 EXPOSURE (DN0/ST) l I i i l I l 23 25 27 29 31 33 35 37 39 41 43 FIGURE 5.2.1 VY CYCLE 11 HALING DEPLETION, EOFPL BUNDLE AVERAGE EXPOSURES s

VERMONT YANKEE CYCLE 11 RODDED DEPLETION E0FPL BUNDLE AVERAGE EXPOSURES R8 R9 R8 R8 RIO R8 R9 RIO RB R10 R8

- 22 26 16 16.21 26.29 23.27 8.44 26 14 17 17 8.64 22.98 6 09 19.99 :

R9 RB R10 R9 R8 R9 R10 R9 R10 R9 R7

- 20 16 35 23 95 8 18 16.85 26.37 17.93 8.83 17.57 7.49 15 15 25.38 R8 R10 R8 R10 R9 R8 R9 RIO R8 R9 R7 _gg 26.55 8.21 24.39 8.86 17 72 27.73 17.90 8.26 23 10 14.90 25.01 R8 R9 R10 RB R10 R9 R10 R9 R9 R7 16 23.41 16.68 8.71 24.54 9.09 17.82 8 59 15.46 15 79 25.84 R10 R8 R9 R10 RB R10 R8 R10 R8 g 8 56 26 22 18 11 9 06 25 07 8 76 24.33 7 02 23.00 RB R9 RB R9 R10 R9 R10 R9 RB 12 26 20 18 23 27.63 17 65 8 77 15 17 7 48 15.61 21 30 R9 R10 R9 RIO RB R10 R8 R8 R7 10 17 17 8.92 17.73 8.62 24.47 7 48 25 30 20.61 25 82 R10 R9 R10 R9 RIO R9 R8 R7 g 8.70 17.76 8.31 15.42 7.04 15 50 20 74 24 23 R8 R10 R8 R9 RB R8 R7 6

22 96 7.53 23.12 15 80 22.89 21 32 26 03 RIO R9 R9 R7

-4 6 14 15 05 14.78 25 93 R7 - P8DPB289, RELOAD 7 RS - P8DPB2G9, RELOAD S R9 - P8DPB289. RELOAD 9 R8 R7 R7 BUNOLE 10 R10 - P8DPB289, RELOAD 10 -2 19.97 25.44 24.90 EXPOSURE (DWD/ST) l I i l l I 23 25 27 29 31 33- 35 37 39 41 43 FIGURE 5.2.2 VY CYCLE 11 RODDE3 DEPLETION, EOFPL BUNDLE AVERAGE EXPOSURES

VERHONT YANKEE CYCLE 11 COLO SHUTOOWN PERCENT DELTA K VS. CYCLE AVERAGE EXPOSURE 2.00

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l 0 1030 2030 3030 4030 5030 6030 7030 8030 CYCLE EXPOSURE j (MWO/ST) i FIC11RE 5.3.1 i

VY CYCLE 11 COLD SilliTDOWN DET TA K IN PERCENT VERSLIS CYCLE EXPOS 11RE l

6.0 THERMAL-HYDRAULIC DESIGN The thermal-hydraulic evaluation of the Reload Cycle was performed using the methods described in the following section.

6.1 Steady-State Thermal Hydraulics Core steady-state thermal-hydraulic analyses were' performed using the FIBWR [11,12] computer code. The FIBWR code incorporates a detailed geometrical representation of the complex flow paths in a BWR core, and cxplicitly models the leakage flow to the bypass region. FIBWR calculates the core pressure drop and total bypass flow for a given total core flow. The power distribution, inlet enthalpy, and geometry are presumed known and are supplied to FIBWR. The power distribution is derived by the 3-D neutronic cimulator SINULATE [4]. Core pressure drop and total leakage flow predicted by the FIBWR code were used in setting the initial conditions for the system's transient analysis model.

6.2 Reactor Limits Determination The objective for normal operation and anticipated transient events is to maintain nucleate boiling. Avoiding a transition to film boiling protects the fuel cladding integrity. Based on Reference 13, the fuel cladding integrity safety limit for Vermont Yankee is a Lowest Allowable Minimum Critical Power Ratio (LAMCPR) of 1.07. Operating limits are specified to maintain adequate margin to onset of the boiling transition. The figure of merit utilized for plant operation is the Critical Power Ratio (CPR). This is defined as the ratio of the critical power (bundle power at which some point within the assembly experiences onset of boiling transition) to the operating bundle power. Thermal margin is stated in terms of the minimum value of the critical power ratio, MCPR, which corresponds to the most limiting fuel cssembly in the. core. Both the transient (safety) and normal operating thermal limits in terms of MCPR are derived based on the GEIL correlation as described in Reference 13.

s

Vermont Yankee Technical Specifications [9] limit the operation of the Reload Cycle fuel to a Maximum Linear Heat Generation Rate (MLHGR) of 13.4 KW/ft. The basis for a MLHGR of 13.4 KW/ft can be found in Reference 3.

i es

7.0 ACCIDENT ANALYSIS 1.1 Cord Wide Transient Analysis core wide transient simulations are performed to assess the impact of csetain transients on the heat transfer characteristics of the fuel. The figure of merit used is the Critical Power Ratio (CPR). It is the purpose of the analysis to determine the minimum critical power ratio such that the c0fety limit is not violated for the transients considered.

7.1.1 Methodology The analysis requires two types of simulations. A system level simulation is performed to determine the overall plant response. Transient core inlet and exit conditions and normalized power from the system level calculation are 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 minimum CPR experienced during the transient).

The system level simulations are performed with the model documented in Rsference 14.

The hot channel calculations are performed with the RETRAN (15] and TCPYA01 [16] computer codes. The GEXL correlation [13] is used in TCPYA01 to svaluate critical power ratio. The calculational procedure is outlined below.

4 l

The hot channel transient ACPR calculations employ a series of " inner" cnd " outer" iterations, as illustrated by the flow chart in Figure 7.1.1. The cuter loop iterates on the hot channel initial power level. This is necessary because the ACPR for a given transient varies with Initial Critical Power .

R tio (ICPR). However, only the ACPR corresponding to a transient MCPR equal to the safety limit (i.e. ,1.07 + ' ACPR = ICPR) is appropriate.- The cpproximate constancy of the ACFR/ICPR ratio is useful in these iterations.

Etch outer iteration requires a RETRAN hot channel run to calculate the transient enthalples, flows, pressure and saturation properties at each time step. These are' required for_ input to the TCPYA01 code. TCYPA01 is then used

-28 =

to calculate a CPR at each time step during the transient, from which a transient ACPR is derived. The hot channel model assumes a chopped cosine cxial power shape with a peak / average ratio of 1.4.

The inner loop iterates on the hot channel inlet flow. These iterations are necessary, because the RETRAN hot channel model calculates the cntrance loss coefficient when given the initial power level, flow, and I pressure drop as input. The pressure drop is assumed equal to the core everage pressure drop, and the ficw is varied for a given power level until the calculated entrance loss coefficient is correct. FIBWR [11. 12] is utilized to estimate the correct inlet flow for a particular power level and l

! pressure drop.

l 7.1.2 Initial Conditions and Assumptions 4 l

The initial conditions for the system simulations are based on maximum l turbine capacity of 105% of rated steam flow. The corresponding reactor l

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

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

I l

1. Scram setpoints are at Technical Specification [9] limits.

f

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

i i

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 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 as given in the Technical Specifications. These are referred to as the " Measured" and the (slower) "67B" scram times.

7.1.3 Resetivity Functions The methods used to generate the fuel temperature, moderator density, end scram reactivity functions are described in detail in Reference 17. The method is outlined below.

A complete set of reactivity functions, the axial power distribution, and the kinetics parameters are generated from base states established for EOFPL, EOFPL-1000 MWD /ST, EOFPL-2000 MWD /ST, and BOC exposure statepoints.

These statepoints are characterized by exposure and void history distributions, control rod patterns, and core thermal-hydraulic conditions.

The latter are consistent with the assumed system transient conditions provided in Table 7.1.1.

The BOC base state is established by shuffling from the previously defined Current Cycle endpoint into the Reload Cycle loading pattern. A criticality search provides an estimate of the BOC critical rod pattern. The EOFPL and intermediate core exposure and void history distributions are calculated with a Haling depletion as described in Section 5.2. The EOFPL state is unrodded. As such, it is defined sufficiently. However, EOFPL-1000 MWD /ST and EOFPL-2000 MWD /ST exposure statepoints require base control rod patterns. These are developed to be as " black and white" as possible. That is, beginning with the rodded depletion configuration,~all control rods which cre more than half inserted are fully inserted, and all control rods which are less than half inserted are fully withdrawn. If the SINULATE calculated t

. _ _ . . . _ .)

A parameters are within operating limits, then this configuration becomes the base case. If the limits are exceeded, a minimum number,of control rods are cdjusted a minimum number of notches until the parameters fall within limits.

Using this method, the control rod patterns and resultant power distributions cre established which minimize the scram reactivity function and maximize the core average moderator density reactivity coefficient. For the transients cnalyzed, this tends to maximize the power response.

At each exposure statepoint, reactivity function table sets are l produced for the 12 core-volumes of the Vermont Yankee RETRAN model. The fuel temperature (Doppler) data set is generated by fixing the power distribution while varying the fuel temperature associated with that power. A moderator density table set is generated specifically for each transient type. The moderator density reactivity functions for the subcooling transient are generated by quasi-statically varying the inlet subcooling only. The moderator enthalpy source distribution is in equilibrium with the calculated nuclear power. The moderator density reactivity functions for the 3

pressurization transients are generated by quasi-statically varying the core I pressure. A series of calculations are performed for various inlet moderator

(

temperatures. The moderator enthalpy source distribution is that of the base state case.

In order to qualitatively compare the core reactivity characteristics between different base configurations, core average reactivity coefficients at selected conditions are provided in Table 7.1.2. Calculated point kinetics parameters for RETRAN are also provided.

The reactivities versus scram insertion are calculated at constant, pre-transient moderator conditions. These are fitted to yield highly detailed scram reactivity curves. The curves are combined with the appropriate rod position versus time data to generate the final RETRAN scram reactivity functions. Figures 7.1.2 through 7.1.4 display the inserted rod worths and rod positions as functions of scram time for the " Measured" scram time analysis. Figures' 7.1.5 through 7.1.7 display similar curves for the "67B" ceram-time analysis.

7.1.4 Transients Analyzed l

Past licensing experience has shown that the core wide transients which r:sult in the minimum core thermal margins are:

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

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

3. Loss of feedwater heating.

The "feedwater controller failure" (maximum demand) transient is not a ecvere transient for Vermont Yankee, because of the plant's 110% steam flow bypass system. Past analyses have shown this transient to be considerably 1sss severe than any of the above for all exposure points. Brief descriptions end the results of the core wide transients analyzed are provided in the following section.

7.2 Core Wide Transient Analysis Results The transients selected for consideration were analyzed at exposure points of EOFPL, EOFPL-1000 MWD /ST, and EOFPL-2000 MWD /ST; the loss of feedwater heating transient was also evaluated at BOC conditions. A summary of the results of the analyses is provided in Table 7.2.1.

7.2.1 Turbine Trip Without Bypass Transient (TTWOBP)

The transient is initiated by a rapid closure (0.1 second closing time) cf the turbine stop valves. It is assumed that the steam bypass valves, which l

normally open to relieve pressure, remain closed. A reactor protection system f

l 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 cf turbine stop valve motion. The ATWS recirculation pump trip is assumed to l

cccur at a setpoint of 1150 psig dome pressure. A pump trip time delay of 1.0 second is assumed to account for logic delay and M-G set generator field I

collapse. In simulating the transient, the bypass piping volume up to the s

valve chest is lumped into the control volume upstream of the turbine stop valver. Predictions of the salient system parameters at the three exposure points are shown in Figures 7.2.1 through 7.2.3 for the " Measured" scram time analysis.

7.2.2 Generator Load Rejection Without Bypass 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 hydraulic fluid pressure switches in the ceceleration relay of the turbine control system. Control rod drive motion is conservatively assumed to cecur 0.28 seconds af ter 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 cceelerating main turbine generator on the recirculation system is simulated by specifying the main turbine generator electrical frequency as a function of time for the M-G set drive motors. The main turbine generator frequency curve is based on a 100% power plant startup test and is considered representative for the simulation. The system model predictions for the three exposure points are shown in Figures 7.2.4 through 7.2.6 for the " Measured" scram time enalysis.

7.2.3 Loss of Feedwater Heating Transient (LOFWH1 A feedwater heater can be lost in such a way that the steam extraction line to the heater is shut off or the feedwater flow bypasses one of the heaters. In either case, the reactor will receive cooler feedwater, which 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 current licensing casumption-for the maximum expected single heater or.gecup of heaters that can be tripped or bypassed by a single event.

~

Vermont Y nkse hrs s scram satpoint of 120% of retsd powar es pset of the Reactor Protection System (RPS) on high neutron flux. In this analysis, no credit was taken for scram on high neutron flux, thereby allowing the reactor power to reach its peak without scram. This approach was selected to provide a bounding and conservative analysis.

The transient response of the system was evaluated at several exposures during the cyclo. The transient evaluation at EOFPL-2000 MWD /ST was found to be the limiting case between BOC to EOFPL. The results of the system response o

to a loss'of 100 F feedwater heating capability evaluated at EOFPL-2000 MWD /ST as predicted by the RETRAN code are presented in Figure 7.2.7.

7.3 Overpressurization Analysis Results 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 core wide transient analysis (Section 7.1.1).

The initial conditions and modeling assumptions discussed in Section 7.1.2 are cpplicable to this simulation.

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

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

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

7.4 Local Rod Withdrawal Error Transient Results The rod withdrawal error is a local core transient caused by an cperator erroneously withdrawing a control rod in the continuous withdrawal

mode. If the core is operating at its operating limits for MCPR and LHGR at the time of the error, then withdrawal of a control rod could increase both local and core power levels with the potential for overheating the fuel.

There is a broad spectrum of core conditions and control rod patterns which could be present at the time of such an error. For most normal cituations it would be possible to fully withdraw a control rod without cxceeding 1% clad plastic strain or violating the CPR based fuel cladding integrity safety limit.

To bound the most severe of postulated rod withdrawal error events, a portion of the core MCPR operating limit envelope is specifically defined such that the cladding limits are not violated. The consequences of the error depend on the local power increase, the initial MCPR of the neighboring locations and the ability of the Rod Block Monitor System to stop the withdrawing rod before MCPR reaches 1.07.

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

The core conditions and control rod pattern for the bounding case are epecified using the following set of concurrent worst case assumptions:

1. The rod should have high reactivity worth. This is provided for by analysis of the core at peak reactivity exposure for each test pattern with xenon free conditions superimposed. The xenon free conditions and the additional control rod inventory needed to maintain criticality exaggerates the worth of control rods substantially when compared to normal operation with normal xenon levels. A fully inserted high worth rod is selected as the error rod.

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

l I

3. The core power distribution is adjusted with the available control rods to place the locatione within the four by four array of bundles around the error rod as nearly on the operating limits as practical.

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

1. Technical Specifications [9] 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 tested in this analysis, there are a maximum of four LPRM inputs per level.

One RBM channel averages the inputs from the A and C levels; the other channel averages the inputs from the B and D levels.

Considering the inputs for a single channel, there are eleven failure combinations of none, one and two failed LPRM stringc. The RBM channel responses are evaluated separately at these eleven input failure conditions. Then, for each channel taken separately, the lowest response as a function of error rod position is chosen for comparison to the RBM setpoint.

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

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

The analysis is performed using the three dimensional steady state SIMULATE core model (4]. Necessary properties of that model for use in this analysis are:

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

2. Accurate LPRM signal calculation as shown by the detailed TIP trace comparisons.

3. Accurate control rod worths and core power coefficient as shown by the consistent core eigenvalues.

Two separate cases are presented from numerous explicit SIMULATE analyses. The reactor conditions and case descriptions are shown in Figures 7.4.1 and 7.4.2. Case 1 analyzes the bounding event with the abnormal xenon condition at the most reactive point in the cycle for the given rod pattern configuration. The initial conditions for Case 2 approximate the 104.5% power conditions with an expected control rod pattern and equilibrium xenon. The ACPR and MLHGR values for both cases are shown in Table 7.4.1. The ACPR values are evaluated such that the implied operating limit MCPR equals 1.07 +

ACPR. This is done by conserving the figure of merit (ACPR/ Initial CPR) shown by the SIMULATE calculations. The use of this method provides valid ACPR values in the analysis of normal operating states where locations near the assumed error rod are not initially near the MCPR operating limit. Case 2 is the worst of all the rod withdrawal transients analyzed from 104.5% pcwcr, full flow and normal rod pattern conditions. Care 2 is bounded by case 1 with substantial MCPR margin.

For Case 1; Figures 7.4.3 and 7.4.4 show the end of transient control rod position. This is determined from the point where the weakest RBM channel response first intercepts the RBM setpoint. For this same bounding case, the i operating limit ACPR envelope component versus Rod Block Monitor setpoint is taken from the Table 7.4.1. The same table demonstrates margin to the 1%

plastic strain limit. The MLHGR values include the 2.2% power spiking penalty.

7.5 Misloaded Bundle Error Analysis Results 7.5.1 Rotated Bundle Error 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 states. The R-factor increase results in a CPR reduction, while the local pin peaking factor ircrease results in a higher pin linear heat generation rate. The objective of the analysis is to insure that in the worst possible rotation, the safety limit linear heat generation rate and CPR are not violated with the most limiting monitored bundles on their operating limits.

To analyze the CPR response, rotated bundle R-factors as a function of cxposure are developed by adding the largest possible AR-factor resulting from a rotation to the exposure dependent R-factors of the properly oriented bundles [13]. Using these rotated bundle R-factors, the MCPR values resulting from a bundle rotation are determined using SIMULATE. This is done for each control rod sequence throughout the cycle. These MCPR values are, in cddition, modified slightly to account for the change in reactivity resulting l from the rotation. For each sequence, the MCPR for the properly oriented essemblies is adjusted by a ratio necessary to place the corresponding rotated CPR on its 1.07 safety limit. The maximum of these adjusted MCPR's is the rotated bundle operating limit.

To determine the Maximum Linear Heat Generation Rate (MLHGR) resulting from a rotation, the ratios of the maximum rotated bundle local peaking factor to the maximum properly oriented bundle local peaking are determined for the expected range of exposure and void conditions. The maximum of this ratio is applied to the operating limit LHGR of 13.4 kw/ft. This maximum rotated bundle LHGR is in addition modified to account for the possible reactivity increase resulting from the rotation. It is also increased by the 2.2% power spiking penalty.

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

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

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

1

I The analysis consists of an iterative procedure which successivaly oliminates potential misloading locations from any MCPR safety limit violations. The first step is to use SIMULATE to determine the largest 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 tubtracted from all the other bundle CPR's in the core. This is done at the various cycle exposures. Even with this maximum ACPR applied, some locations will never exceed the MCPR safety limit of 1.07. These locations are oliminated from further investigation.

The next iteration consists of applying the same procedure to the locations which appeared to violate the safety limit when the maximum ACPR from the first iteration was applied. Since these locations are of higher reactivity than those eliminated in the first iteration, they will result in a smaller ACPR when misloaded. Using this smaller ACPR, some of the remaining locations will be eliminated from potential CPR safety limit violations. This

! procedure is continued until all locations are shown to be above the MCPR safety limit due to a misloading, or until a limiting location is identified.

Using the above procedure, it has been demonstrated that for the Reload Cycle all possible mislocations result in calculated MCPR's above the 1.07 safety limit, assuming an initial operating CPR limit of 1.22. This makes the mislocated bundle analysis less limiting than the rotated bundle analysis.

l 7.6 Control Rod Drop Accident Resultg The control rod sequences are a series of rod withdrawal and banked withdrawal instructions specifically designed to minimize the worths of individual control rods. The sequences are examined so that, in the event of the uncoupling and subsequent free fall of the rod, the incremental rod worth is acceptable. Incremental worth refers to the fact that rods beyond Group 2 cre banked out of the core and can only fall the increment from all in to the rod drive withdrawal position. Acceptable worth is one which produces a maximum fuel enthalpy less than 280 calories / gram.

Some out-of-sequence control rods could accrue potentially high

' worths. However, the Rod Worth Minimizer (RWM) will prevent withdrawing an l

I

out-of-sequence rod if accidentally selected. The RWM is functionally tested before each startup.

The sequence entered into the RWM will take the plant from All Pods In (ARI) to well above 20% core thermal power. Above 20% power even multiple operator errors will not create a potential rod drop situation abcVe 280 7, calories per gram (18, 19]. Below 20% power, however, the sequences must be examined for incremental rod worth. This is done using the full core, xenon free SINULfTE model at the projected most reactive point in the cycle. This casures that the maximum amount of reactivity is held in the rods.

Both the A and B sequences were examined. It was found that the highest worth occurred in the first red pull of the second group. Any of the first four rod arrays shown in Figures 7.6.1 and 7.6.2 may be designated ask the first group pulled. But, then a specific second group must follow as ~

Table 7.6.1 illustrates. For added conservatism, the highest worth rod in the second group was deliberately assigned to be the first rod pulled. This assures that in any sequence followed at the plant, the worths will always be less than those calculated here. The results of the calculations are presented in Table 7.6.2.

s Beyond Group 2, procedures (20] apply which severely reduce the rod incremental worths. This makes the xenon free, hot staniby worths much less ,

than the cold, xenon free worths as demonstrated in Ref'e'rance 10.

i s

7.7 Stability Analysis Results ,

The analysis of reactor stability has been performed by General '

Electric as described in Section S.2.4 of Reference 3. The 105% power rod I line was analyzed and the resultant decay ratio as a function of reactor power ;

level is provided in Figure 7.7.1. 4 t

s The reactor core stability decay ratio is calculated at natural- ,, <

circulation conditions and a power level corresponding to the 105% power rod line. The calculated core stability decay ratio and the channel hydrodynamic 't.. 'l' performance decay ratio is given in Table 7.7.1.

s I

^%,,.

TABLE 7.1.1 l l

VY CYCLE 11

SUMMARY

OF SYSTEM TRANSIENT MODEL INITIAL CONDITIONS FOR CORE WIDE TRANSIENT ANALYSES

(

Core Thermal Power (NWth) 1664.0 Turbine Steam Flow (% NBR) b 105 .

Total Core Flow (10 61bm/hr) 48.0 Core Bypass Flow (1061bm/hr) 5.3

Core Inlet Enthalpy (BTU /lbm) 520.9 Steam Dome Pressure (psia) 1034.7 Turbine Inlet Pressure (psia) 986.0 Total Recirculation Flow (10 1bm/he) 6 23.4 Core Plate Differential Pressure (psi) 18.5 Narrow Range Water Level (in.) 35 I

Average Fuel Cap Conductance (See Section 4.2) c, 5

i M .

, e -r - - ,-

TABLE 7.1.2 VY CYCLE 11 TRANSIENT ANALYSIS REACTIVITY COEFFICIENTS AT SELECTED CONDITIONS Cycle Exposure Point (MWD /ST)

Calculated Parameter EOFPL EOFPL-1000 EOFPL-2000 BOC Axial Shape Index(l) -0.0805 -0.1854 -0.1938 -0.0760 Moderator Density Coefficient 20.75 20.64 26.04 20.13 (Subcooling), d/au(2)

Pressure = 1050 psia Subcooling = 30 BTU /lbm Moderator Density Coefficient 23.12 22.10 28.69 (3)

(Pressurization), g/au Pressure = 1050 psia Inlet Enthalpy = 520 BTU /lbm

),. Fuel Temperature Coefficient -0.278 -0.277 -0.276 -0.259 7 at 1130 0F, d/ UF Effective Delayed 0.005400 0.005494 0.005547 0.005958 Neutron Fraction Prompt Neutron Generation 42'.64 42.52 41.90 39.65 Time in Microseconds

- ,~ -

" w p -p NotesE;(1)-AxialShapeIndex(AkI)=P+P B

w (2) au = change in density, in percent

'(3) Pressurization transients are not calculated at BOC y

g.

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TABLE 7.2.1 l

ho' VY CYCLE 11 CORE WIDE TRANSIENT ANALYSIS RESULTS Peak Peak Avg.

Prompt Power Heat Flux (Fraction of (Fraction of ACPR Transient Exposure Initial Value) Initial Value) P8X8R Turbine Trip EOFPL 2.936 1.186 .19 Without Bypass,

'; Measured" EOFPL-1000 2.097 1.108 .13

( Scram Time EOFPL-2000 1.007 1.000 .00 j

Turbine Trip EOFPL 3.466 1.236 .25 Without Bypass, "67B" EOFPL-1000 2.574 1.189 .20 Scram Time l EOFPL-2000 1.433 1.007 .01 G:nerator Load EOFPL 2.802 1.168 .19 Rajection Without Bypass, EOFPL-1000 2.008 1.094 .11

" Measured" Scram Time EOFPL-2000 1.000 1.000 .00 Generator Load EOFPL 3.405 1.227 .26 Raj ection Without Bypass, EOFPL-1000 2.590 1.193 .20 "67B" Scram Tine- EOFPL-2000 1.310 1.000 .00 Loss of 100 0F EOFPL 1.203 1.196 '.16 Feedwater

~ ' Heating ' EOFPL-1000 1.216 1.208 .17 EOFPL-2000 1.221 1.213 .18 BOC 1.199 1.191 .15 t

.4

b. fA

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TABLE 7.3.1 e, VY CYCLE 11 OVERPRESSURIZATION ANALYSIS RESULTS l

Maximum Pressure at Reactor ,

Conditions Vessel Bottom (psix) l

" Measured" Scram Time 1281 "67B" Scram Time 1306 l

TABLE 7.4.1 VY CYCLE 11 ROD WITHDRAWAL ERROR TRANSIENT

SUMMARY

f WITH LIMITING INSTRUMENT FAILURE i

Case 1 Conditions in Figure 7.4.1 RBM Rod ACPR MLHGR (kw/ft)

Setpoint Position P8X8R P8X8R 104 10 .12 15.4 105 10 .12 15.4 106 12 .16 16.4 107 12 .16 16.4 108 18 .22 17.8 Case 2 Conditions in Figure 7.4.2 RBM Rod ACPR MLHGR (kw/ft)*

Setpoint Position P8X8R P8X8R 104 30 10 8 13.1 105 38 .11 14.1 106 40 .11 14.4 107 42 .12 14.6 108 44 .12 14.7 Q Not initially on limits t

TABLE 7.5.1-VY CYCLE 11 ROTATED BUNDLE ANALYSIS RESULTS Resulting Initial MCPR Resultinz MCPR LHGR (kw/ft) 1.24 1.07 17.46 s

TABLE 7.6.1 CONTROL ROD DROP ANALYSIS - h0D ARRAY PULL ORDER l

The order in which rod arrays are pulled is specific once the choice of first group is made.

- First Group Second Group Successive Group Pulled is: Pulled Must Be: Is Banked Out I

j Array 1 Array 2 Array 3 or 4 I l

l Array 2 Array 1 Array 3 cr 4 l

l Array 3 Array 4 Array 1 or 2 Array 4 Array 3 Array 1 or 2 l

TABLE 7.6.2 VY CYCLE 11 CONTROL ROD DROP ANALYSIS RESULTS Maximum Incremental Rod Worth .83% AK Calculated Cold, Xenon Free ,

Bounding Analysis Worth for Enthalpy 1.30% AK Less than 280 Calories per Gram (References 18, 19 and 20)

TABLE 7.7.1 VY CYCLE 11 STABILITY ANALYSIS RESULTS Decay Ratio Reactor Total Co?9 Stability .83 Channel Hydrodynamic Performance .30 U

. - e-r , w

Choose ICPR i f Estimate Power 4 I

If 7

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initialization Run '

1 is Loss No Revise Coefficient :

Correct? Flow Yes II RETRAN/TCPYA01 Hot Channel Run 1

Has ACPR No Converged? New ICPR Yes STOP FIGURE 7.1.1 FLOW GIART FOR Tile CALCULATION OF ACPR USING Tile RETRAN/TCPYA01 CODES v

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FIGURE 7.1.6 INSERTED ROD WORTH AND ROD POSITION VERSUS TIME FROM INITIAL ROD MOVEMENT AT EOFPL11-1000 IMD/ST, "67B" SCPJOf TIMF VY CYCLE 11 - 678 SCRAM, EOFPL-2000 l o.

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FIGURE 7.1.7 INSERTED ROD WORTH AND ROD POSITION VERSUS TIME FROM INITIAL ROD MOVEMENT AT EOFPL11-2000 MWD /ST, "67B" SCRAM TIME 9

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CONTROL ROD PATTERN 43 8 10 14 10

" 20 20 18 18 10 10 87 18 00 00 18

' 88 14 14 a 18 00 00 18 15 10 10 21 20 18 18 20 o?-- 10 14 10 03 02 04 10 14 la 22 26 30 34 34 42 Reactor Conditions:

Core Thermal Power = 1664 Mwt Core Flow = 48 M1b/hr Cycle Exposure = 2565 MWD /ST Xenon Free Initial MCPR' = 1.357 Initial LHGR = 13.4 kw/ft Case Description o operator attempts full withdrawal of.the fully inserted rod at coordinates (26, 27),

o Bounding Case.

FIGURE 7.4.1 REACTOR INTTIM CONDITIONS FOR THE VY CYCLE 11 ROD WITHDRAWAL ERROR CASE 1 ;

l 1

J.

CONTROL ROD PATTERN 43 8 40 40 35 36 16 16 36 27 n 28 06 06 28 19 u 36 16 16 36 11 c7 40 40 o3 C2 04 10 to 18 22 26 3C 34 30 42 Reactor Conditions:

Core Thermal Power = 1664 Mwt Core Flow - 48 M1b/hr Cycle Exposure = 4275 MWD /ST l Equilibrium Xenon l

Initial MCPR = 1.386 Initial LHGR = 12.8 kw/ft Case Description l o Operator attempts full withdrawal of the partially inserted i rod at coordinates (26, 31).

o Normal Xenon condition and control rod pattern.

FIGURE 7.4.2 REACTOR INITIAL CONDITIONS FOR THE VY CYCLE 11 ROD WITHDRAWAL ERROR CASE 2

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3M NOTES:

107

1. RBM Setpoint Intercept is Marked with (e).

g 105

2. Rod is Stopped at Notch Following Z

".- 104 Two Inches of Free Rod Motlon. ~~

, 3. The Box (D) Shows the Response with No Instrument Failures.

O O 9, 10 12 la 9 T.00 5'.00 10.00 l'5.00 2'O.00 2'5.00 3'O.00 3'5.00 4'0 00 4h.00 50700 RWE CONTROL ROD POSITION FIGURE 7.4.4 g__CYCJ.E_11 R_WE CASE 1- _SETPOINT_INTERCEP_TS _D.ETERt11N_ED_ BY_._Ti!E_ B+D CifANNEl.

43 3 39 2 1 1 2 35 4 3 4 3 4 31 1 2 2 1 27 3 4 3 4 3 23 1 2 1 1 2 1 19 3 4 3 4 3 15 1 2 2 1 11 4 3 4 3 4 07 2 1 1 2 03 3 02 06 10 14 18 22 26 30 34 38 42 FIGUEE 7.6.1 FIRST FOUR ROD ARRAYS PULLED IN THE A SEQUENCES 43 3 3 39 2 1 2 35 3 4 4 3 31 2 1 2 1 2 i

27 3 4 3 3 4 3 23 1 2 1 2 1 19 3 4 3 3 4 3 15 2 1 2 1 2 11 3 4 4 3 07 2 1 2 03 3 3 02 06 10 14 18 22 26 30 34 38 42 l

FIGURE 7.6.2 -FIRST FOUR ROD ARRAYS PULLED IN THE B SEQUENCES l

)

_._____m_ _ _ _ __ _ _ _ _ . _ _L. ___ _ _ _ _ _ _ . , , , ,_, _ _ _ , _ _ _ _ _ , , _

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I I A iC5 E l:ATL FER. E::I AL[ICI:CULATI'.iN RCD L[lNE C ULT. PE F O:.M A:.C E LIMIT 1.00 ( ,.

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T-O 0 20 40 60 80 100 120 PERCENT POWER FIGURE 7.7.1 VY CYCLE 11 CORE DECAY RATIO VERSUS POWER 1

I 4

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

The Vermont Yankee 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. The content of the Startup Test Report will be similar to that sent to the Office of Inspection and Enforcement in the past (22].

l I

9.0 LOSS-OF-COOLANT ACCIDENT ANALYSIS The results of the complete evaluation of the loss-of-coolant accident for Vermont Yankee as documented in Reference 23 provide required support for the operation of the Reload Cycle. No new fuel types have been introduced in this reload, therefore, the MAPLHGR limits as a function of average planar cxposure remain the same as in the Current Cycle [9].

APPENDIX A CALCULATED CYCLE DEPENDENT LIMITS The MCPR limits appropriate for the Reload Cycle are calculated by cdding the calculated ACPR to the safety limit LAMCPR of 1.07. This is done fcr each of the analyses in Section 7 at each of the exposure statepoints.

Far an exposure interval between statepoints, the highest MCPR limit at either cnd 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.

With regard to MAPLHGR, no new fuel types have been introduced. The MAPLHGR limits given in Reference 9 apply to the Reload Cycle. The MCPR limits in Reference 9 are also bounding for the Reload Cycle. These are found in Reference 9 as Table 3.11-2 and are reproduced here as Table A.2. On Table A.2, as in the Technical Specifications, End of Cycle (EOC) is understood to ,

mean End of Full Power Life (EOFPL).

TA!LE A.1 VERMONT YANKEE NUCLEAR POWER STATION LIMITING CYCLE 11 MCPR RESULTS Value of "N" in RBM Average Control Rod Cycle MCPR for Equation (1) Scram Time Exposure Range P8X8R Fuel 42% BOC to EOFPL-2 GWD/T 1.29

" MEASURED" EOFPL-2 GWD/T to EOFPL-1 CWD/T 1.29 EOFPL-1 GWD/T to EOFPL 1.29 BOC to EOFPL-2 GWD/T 1.29 "57B" EOFPL-2 GWD/T to EOFPL-1 GWD/T 1.29 EOFPL-1 GWD/T to EOFPL 1.33 41% BOC to EOFPL-2 GWD/T 1.25

" MEASURED" EOFPL-2 GWD/T to EOFPL-1 GWD/T 1.25 EOFPL-1 GWD/T to EOPPL 1.26 BOC to EOFPL-2 GWD/T 1.25 5 "67B" EOFPL-2 GWD/T to EOFPL-1 CWD/T 1.27 y EOFPL-1 GWD/T to EOFPL 1.33

$40% BOC to EOFPL-2 GWD/T 1.25

" MEASURED" EOFPL-2 GWD/T to EOFPL-1 GWD/ 1.25 EOFPL-1 GWD/T to EOFPL 1.26 BOC to EOFPL-2 GWD/T 1.25 "67B" EOFPL-2 GWD/T to EOFPL-1 GWD/T 1.27 EOFPL-1 GWD/T to EOFPL 1.33 NOTES:

(1) The Rod Block Monitor (RBM) trip setpoints are determined by the equation shown in Table 3.2.5 of the Technical Specifications [ Reference 9].

TABLE A.2 VERMONT YANKEE NUCLEAR POWER STATION TECHNICAL SPECIFICATION MCPR OPERATING LIMITS MCPR Operating Limit for Value of "N" in RBM Average Control Rod Cycle Fuel Type (2)

Equation (1) 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.1.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 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 1.33 1.31 1.31 3.3 C.1.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 41% Equal or better BOC to EOC-2 CWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.26 1.25 1.25 3.3 C.1.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 4 than L.C.O. EOC-2 GWD/T to EOC-1 CWD/T 1.33 1.31 1.31 f 3.3 C.I.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 f 40% Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 CWD/T to EOC-1 GWD/T 1.26 1.25 1.25 3.3 C.1.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOO to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1. 3 '.

3.3. C.I.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 75% Special Testing at Natural Circulation (Notes 3, 4) 1.30 1.31 1.31 NOTES:

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

(2) The curtent analyses for MCPR Operating Limits do not include 7X7 fuel. On this basis, further evaluation of MCPR optrating 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.

/

REtERENCES l

1. M.' A. Sironen, R. A. Woehlke, and I . A. McGahan, Vermont Yankee Cycle 8

. Summary Report, YAEC-1305, August 1982.

- 2. M. A. Sironen and P. A. McGahan, Vetmont Yankee Cycle 9 Summary Report, YAEC-1367, June 1983.

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

! NEDE-24011-P-A-5, GE Company Proprietary, August 1982.

4. D. M. VerPlanck, Methods for the Analysis of Boilinz Water Reactors .

Steady State Core Physics, YAEC-1238, March 1981. '

5. E. E. Pilat, Methods for the Analysis of Boilina Water Reactors Lattice Physics YAEC-1232, December 1980.

6. S. P. Schultz and K. E. St. John, Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY) Code /Model Description Manual, YAEC-1249P, April 1981.

7. S. P. Schultz and K. E. St. John, Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY) Code Oualification and Application, YAEC-1265P, June 1981.

l 8. D. C. Albright, H20DA: An Improved Water Properties Packaze, YAEC-1237, March 1981.

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

10. A. A. F. Ansari, et al., Vermont Yankee Cycle 9 Core Performance Analysis, YAEC-1275, August 1981.

11. A. A. F. Ansari, Methods for the Analysis of Boilina Water Peactors:

Steady-State Core Flow Distribution Code (FIBWR),' YAEC-1234, December 1980.

12. A. A. F. Ansari, R. R. Gay, and B. J. Gitnick, FIBWR: A Steady-State Core Flow Distribution Code for Boilinx Water Reactors - Code Verification and-Oualification Report, EPRI NP-1923, Project 1754-1 Final

-Report,-July 1981.

13. General Electric Company, GEIL Correlation Application to BWR 2 Reactors, NEDE-25422, GE Company Proprietary, June 1981.

l

14. A. A. F. Ansari and J. T. Cronin, Methods for the Analysis of Boilina '

Water Reactors: A Systems Transient Analysis Model (RETRAN). YAEC-1233, l April 1981.

15. RPRI, RETRAN - A Promram for One-Dimensional Transient Thermal-Hydraulic

, &gglysis of-Complex Fluid Flow Systegg. CCM-5, December 1978.

[.

(

., f -- . . - _ . .-, .4 , ,

16. A. A. F. Anscri, K. J. Burns, cnd D. K. Bellce, Mathwds fer th, An+1ysim of Boiling Water Reactors: Transient Critical Power Ratio Analysis (RETRAN-TCPYA01), YAEC-1299P, March 1982.

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

18. C. J. Paone, et al., Red Drop Accident Analysis for Large Boilinz Water Reactors. NEDO-10527, March 1972.

19. R. C. Stirn, et al., Rod Drop Accident Analysis for Large Boilinz Water Reactors Addendum No. 1. Multiple Enrichment Cores With Axial Gadolinium, NEDO-10527, Supplement 1, July 1972.

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

21. R. C. Stirn, et al., Rod Drop Accident Analysis for Large Boiling Water Reactor Addendum No. 2 Exposed Cores, NEDO-10527, Supplement 2 January 1973.

22. Letter, FVY 83-100, dated September 15, 1983, W. P. Murphy to T. E. Murley, Regional Administrator, " Cycle X Startup Test Report".

23. Loss-of-Coolant Accident Analysis for Vermont Yankee Nuclear Power l Station, NEDO-21697, August 1977, as amended.

1

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ML20084S466

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