ML20071A817
| ML20071A817 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 02/09/1983 |
| From: | Paul Bergeron, Cacciapouti R, Stephen Schultz VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20071A814 | List: |
| References | |
| YAEC-1342, NUDOCS 8302250167 | |
| Download: ML20071A817 (95) | |
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VERMONT YANKEE 3
s CYCLE 10 CORE PERFORMANCE ANALYSIS _
u-January 1983 '?
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Major Contributors: __
B. G. Baharynejad D. M. Kapitz --
D. K. Beller J. M. Kendall -
K. E. St. John ; -
K. J. Burns J. T. Cronin M. A. Sironen G. E. Jarka D. M. VerPlanck R. A. Woehlke ,
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Approved by: f 83 b RgJ.Caccigtouti, f
Manager '(D# ate)
Reactor Physics Group Approved by: M ~
P. A. Berge n, Manager (Date)
TransientAn[alysisGroup Approved by: 1/T)03 S.4. Schultz,1(anageM (Date) ..
Nuclear Evaluations and Support Group i Approved by: {_ h_ 1l9 TS t A. Huiain, Manager f(Date)
LOCA Group _'
Approved by: 'J ,. 2 1 1 B. CT SlifAr, 3anager ,
[ ' (Date)
Nuclear Engineering ' ;
8302250167 830222 PDR ADOCK 05000271 P PDR
DISCLAIMER OF RESPONSIBILITY 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 information. It is authorized for use specifically by Yankee 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 Corporation and their officers, directors, agents and employees assume no liability nor make any warranty or representation with respect to the contents of this document or to its accuracy or completeness.
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ABSTRACT This report presents design information and calculational results t
partinent to the operation of Cycle 10 of the Vermont Yankee Nuclear Power Station. These include the fuel design and core loading pattern descrip': ions; calculated reactor power distributions, power peaking, shutdown capability and reactivity functions; and the results of safety analyses perfctmed to justify plant operation throughe ' Cycle 10.
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TABLE OF CONIENTS Page DISCLAIMER OF RESPONS IBILLTY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 ABSTRACT.................................................... iii T ABL E O F CONTE NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv LIST OF FIGURES............................................. vi LIST OF TABLES.............................................. viii ACKN0WLEDGEMENTS............................................ ix
1.0 INTRODUCTION
................................................ 1 2
2.0 RECENT REACTOR OPERATING HIST 0RY............................
2.1 Operating History of the Current Cycle................. 2 2.2 Operating History of Recent Applicable Cycles.......... 2 3.0 RELOAD CORE DESIGN DESCRIPTION.............................. 4 3.1 Core Fuel Loading...................................... 4 3.2 Design Reference Core Loading Pattern.................. 4 3.3 Assembly Exposure and Cycle 9 History . . . . . . . . . . . . . . . . . . 4 ,
4.0 FUEL MECHANICAL AND THERMAL DESIGN. . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.1 Mechanical Design...................................... 7 4.2 Thermal Design................................ ........ 7 .
4.3 Operating Experience................................... 8 "4 -
W 5.0 N UCLE AR D ES I GN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 h.
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5.1 Core Power Distributions............................... 13 'ir.- G^u
%.e 5.1.1 Haling Power Distribution....................... 13 e. "vs 9 5.1.2 Rodded Depletion Power Distribution............. 13 +W g . . .
5.2 Core Exposure Distributions............................ 14 $C -
5.3 Cold Core Reactivity and Shutdown Marg in. . . . . . . . . . . . . . . 14 5.4 Standby Liquid Control System Shutdown Capability...... 15 6.0 THERMAL-HYDRAULIC DESIGN.................................... 24 6.1 Steady-S ta te Thermal Hy araulic s. . . . . . . . . . . . . . . . . . . . . . . . 24 6.2 Reactor Limits Determination........................... 24
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6 TABLE OF CONTENTS V (Continued)
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E 7.0 ACCIDENT ANALYSIS........................................... 26 7.1 Core Wide Transient Analysis........................... 26 --
I 7.1.1 Methodology..................................... 26 7.1.2 Initial Conditions and Assumptions.............. 27 7.1.3 Reactivity Functions............................ 28 7.1.4 Transients Analyzed............................. 30 7.2 Core-Wide Transient Analysis Results................... 30
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E 7.2.1 Turbine Trip Without Bypass Transient........... 30 7.2.2 Generator Load Rejection Without E Bypass Transient................................ 31 7.2.3 Loss of Feedwater Heating Transient............. 31 P 7.3 Ove rpressurization Analysis Results. . . . . . . . . . . . . . . . . . . . 32 7.4 Local Rod Withdrawal Error Transient Results........... 32
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7.5 Misloaded Bundle Error Analysis Results................ 35
[ 7.5.1 Rotated Bundle Error............................ 35 7 7.5.2 Mislocated Bundle Error......................... 36 7.6 Control Rod Drop Accident Results...................... 37 7.7 Stability Analysis Results............................. 38
[ 8.0 STARTUP PR0 GRAM............................................. 80
= 9.0 LOSS-OF-COOLANT ACCIDEffI ANALYSIS........................... 81 APPENDIX A CALCULATED CYCLE DEPENDENT LIMITS............... 82 REFERENCES.................................................. 85
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L LIST OF FIGURES 3 Number Title M 2
3.2.1 VY Cycle D Design Reference Leading Pattern, Lower Right 6 Quadrant VY Cycle 10 Core Average Gap Conductance versus Cycle 11 4.2.1 -
Burnup 4.2.2 Vermont Yankee Hot Channel Gap Conductance for P8X8R versus 12 Exposure 5.1.1 VY Cycle 10 Haling Depletion EOC Bundle Average 17 _
i Relative Powers .
5.1.2 VY Cycle 10 Core Average Axial Power Distribution Taken 18 from the Haling Calculation to EOFPL T L
5.1.3 VY Cycle 10 Rodded Depletion - ARO at EOFPL 19 Bundle Average Relative Powers }
5.1.4 VY Cycle 10 Core Average Axial Power Distribution, 20 Rodded Depletion - ARO at EOFPL _
5.2.1 VY Cycle 10 Haling Depletion, EOC Bundle Average Exposures 21 5.2.2 VY Cycle 10 Rodded Depletion, EOC Bundle Average Exposures 22 5.3.1 VY Cycle 10 Cold Shutdown Delta K in Percent versus Cycle 23 Exposure 7.1.1 Flow Chart for the Calculation of a CPR Using the 44 RETRAN/TCPYA01 Codes 7.1.2 Inserted Rod Worth and Rod Position versus Time From 45 Initial Rod Movement at EOC10, " Measured" Scram Time 7.1.3 Inserted Rod Worth and Rod Position versus Time From 46 Initial Rod Movement at EOC10-1000 MWD /ST, " Measured" Scram Time <
7.1.4 Inserted Rod Worth and Rod Position versus Time From 47 Initial Rod Movement at EOC10-2000 MWD /ST, " Measured"
, Scram Time 7.1.5 Inserted Rod Worth and Rod Position versus Time From 48 Initial Rod Movement at EOC10, "67B" Scram Time 7.1.6 Inserted Rod Worth and Rod Position versus Time From 49 Initial Rod Movement at EOC10-1000 MWD /ST, "67B" Scram Time 7.1.7 Inserted Rod Worth and Rod Position versus Time From 50 Initial Rod Movement at EOC10-2000 MWD /ST, "67B" Scram Time
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LIST OF FIGURES (Continued) Page Number Title Turbine Trip Without Bypass, EOC10 Transient 51 7.2.1 Response versus Time, " Measured" Scram Time Turbine Trip Without Bypass, EOC10-1000 MWD /ST 54 7.2.2 Transient Response versus Time, " Measured" Scram Time Turbine Trip Without Bypass, EOC10-2000 MWD /ST 57 7.2.3 Transient Response versus Time, " Measured" Scram Time Generator Load Rejection Without Bypass, EOC10 60 7.2.4 Transient Response verous Time, " Measured" Scram Time Generator Load Rejection Without Bypass, EOC10-1000 MWD /ST 63 7.2.5 Transient Response versus Time, " Measured" Scram Time Generator Load Rejection Without Bypass, EOC10-2000 MWD /ST 66 7.2.6 Transient Response versus Time, " Measured" Scram Time Loss of 100 F0 Feedwater Heating, EOC10-2000 MWD /ST 69 7.2.7 (Limiting Case) Transient Response versus Time 7.3.1 MSIV Closure, Flux Scram, EOC10 Transient Response versus 71 Time, " Measured" Scram Time 7.4.1 Reactor Initial Conditions for the VY Cycle 10 74 Rod Withdrawal Error Case 1 7.4.2 Reactor Initial Conditions for the VY Cycle 10 75 Rod Withdrawal Error Case 2 VY Cycle 10 RWE Case 1 - Setpoint Intercepts Determined 76 7.4.3 by the A+C Channel VY Cycle 10 RWE Case 1 - Setpoint Intercepts Determined 77 7.4.4 by the B+D Channel 7.6.1 First Four Rod Arrays Pulled in the A Sequences 78 First Four Rod Arrays Pulled in the B Sequences 78 7.6.2 Reactor Core Decay Ratio versus Power 79 7.7.1
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LIST OF TABLES Page Number Title VY Cycle 10 Fuel Bundle Types and Numbers 5 3.1.1 Design Basis VY Cycle 9 and Cycle 10 Exposures 5 3.3.1 4.1.1 Nominal Fuel Mechanical Design Parameters 9 4.2.1 Gap Conductance Values used in VY Cycle 10 Transient Analyses 10 5.3.1 VY Cycle 10 K-Effective Values and Shutdown Margin 16 Calculation 5.4.1 VY Cycle 10 Standby Liquid Control System Shutdown Capability 16 7.1.1 VY Cycle 10 Summary of System Transient Model Initial 39 Conditions for Core Wide Transient Analyses 7.1.2 VY Cycle 10 Transient Analysis Reactivity Coefficients 40 VY Cycle 10 Core Wide Transient Analysis Results 41 7.2.1 VY Cycle 10 Rod Withdrawal Error Transient Summary 42 7.4.1 (With Limiting Instrument Failure)
Rotated Bundle Analysis Results 42 7.5.1 Control Rod Drop Analysis - Rod Array Pull Order 43 7.6.1 VY Cycle 10 Control Rod Drop Analysis Results 43 7.6.2 Vermont Yankee Nuclear Power Station Calculated Cycle 10 MCPR 83 A.1 Limits A.2 The MCPR Operating Limits for Cycle 9 are Bounding for 84 Cycle 10.
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ACKNOWLEDGEMENTS The authors and principal contributors would like to acknowledge the contributions to this work by P. A. McGahan, K. E. Mitchell, J. Pappas, and tha YAEC Word Processing Center. Their assistance in preparing figures and text for this document is recognized and greatly appreciated.
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1.0 INTRODUCTION
This report provides information to support the operation of the Varmont Yankee Nuclear Power Station through the forthcoming fuel reload cycle (called Cycle 10). The refueling preceding Cycle 10 (called Reload 9) will isvolve the discharge of 108 irradiated fuel bundles and the insertion of 108 tow fuel bundles. The resultant core will consist of 108 new fuel bundles and 260 irradiated fuel bundles of the pressurized retrofit 8X8 design (P8DPB289). All fuel bundles for Cycle 10 operation have been fabricated by General Electric (CE).
This report contains descriptions and analyses results pertaining to tha mechanical, thermal-hydraulic, physics, and safety aspects of Cycle 10.
The cycle dependent operating limits as calculated for Cycle 10 are given in Appendix A.
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2.0 RECENT REACTOR OPERATING HISTORY 2.1 Operating History of the Current Cycle The current operating cycle is Cycle 9. The reactor was started up for this cycle on December 1,1981 and is projected to be shut down for refueling os March 5, 1983. During this period, control rod sequence exchanges were parformed on the following schedule:
SEQUENCE from tjt 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 The reactor has been operated smoothly and at full power with the exception of normal maintenance and a few scrams. The control rod sequence cxchanges in January, April and June were perf ormed following scrams. The rest of the exchanges occurred at minimum flow, reduced power. The reactor otorted coastdown on December 16, 1982 with four rods at position 30. The7e care pulled out on January 14, 1983. The remainder of the cycle will be in ths All-Rods-Out (ARO) condition.
2.2 Operating History of Recent Applicable Cycles -.
Fuel to be re-irradiated in Cycle 10 includes fuel bundles which were initially inserted into the reactor in Cycle s 7, 8, and 9.
Cycle 7 reactor operation proceeded at full power with normal maintenance and operational maneuvers with the exception of a three day outage in February 1980 to implement plant modifications required by the NRC (TMI fix). A total of four control rod sequences were used during the cycle.
Cycle 8 operation [1] also proceeded at full power with normal ocintenance and operational saneuvers. Four control rod sequences were used in Cycle 8. Two sequence exchanges were performed at minimum flow, reduced power. Following the exchange in March 1981, the reactor was operated at l
reduced power for five days to allow for special testing; including, r: circulation pump trip testing and reactor stability testing.
3.0 RELOAD CORE DESIGN DESCRIPTION 3.1 Core Fuel Loading Reload 9 (Cycle 10) will discharge 108 spent fuel assemblies out of a core total of 368. Thus, the Cycle 10 core will consist of 108 new assemblies cad 260 irradiated assemblies. All assemblies have bypass flow holes in the 1swer tie plate. Table 3.1.1 characterizes the core by fuel type, batch size, and first cycle loaded. A description of the fuel is found in Reference 2.
3.2 Design Reference Core Loading Pattern The Cycle 10 assembly locations are indicated by the map in Figure 3.2.1. For the sake of legibility only the lower right quadrant is shown.
Th2 other quadrants are mirror images with bundles of the same type having noctly identical exposures. The new bundles (inserted during Reload 9) have bssn identified as R9. Similarly, irradiated bundles are designated by the raload number in which they were first introduced into the core. If any ch:nges are made to the loading pattern at the time of refueling, they will be chrcked 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 and Cycle 9 History The assumed nominal exposure on the fuel bundler in the design rsference loading pattern is given in Figure 3.2.1. To obtain this exposure diotribution, previous cycles up to Cycle 9 were dep.leted with the SIMULATE model [3,4] using actual plant operating history. For Cycle 9, plant cparating history was used through 8/19/82; that is, a core average exposure of 14.339 GWD/ST. Beyond 8/19/82 the exposure was accumulated using a bact-estimate rodded depletion analysis to EOFPL9. This was followed by a projected coastdown to EOC9 on 3/5/83.
Table 3.3.1 gives the assumed nominal burnup on Cycle 9 and the BOC10 cuposure that results from the shuffle. In this table, as in the rest of this rcport, the terms "End of Cycle (EOC)" and "End of Full Power Life (EOFPL)",
cc applied to Cycle 10, are used interchangeably.
TABLE 3.1.1 VY CYCLE 10 PUEL BUNDLE TYPES AND NUMBERS Puel Cycle Possible Designation Loaded Number Bundle ID's IRRADIATED P8DPB289 7 60 LJGXXX, LJHXXX, LJLXXX F8DFB289 8 80 LJPXXX, LJUXXX P8DPB289 9 120 LJTXXX, LJZXXX NEW P8DPB289 10 108 LY4XXX NOTE: XXX stands for the last three digits of the bundle serial number.
TABLE 3.3.1 DESIGN BASIS VY CYCLE 9 AND CYCLE 10 EXPOSURES Assumed Previous Cycle Core Average Exposure End of Cycle 9 18.19 GWD/ST Assumed Reload Cycle Core Average Exposure Beginning of Cycle 10 10.48 CWD/ST Haling Calculated Core Average Exposure at End of Cycle 10 17.70 GWD/ST Cycle 10 Capability ,
7.22 GWD/ST VERMONT YANKEE CYCLE 10 00C SUWOLE AVERROC EXPG8URES 35 37 39 41 43 23 25 27 29 31 33 PLANT coOaD R7 R8 R8 R8 R7 RS R8 Rt R$ R8 R$
22 15 34 0 00 22 88 8 47 14.83 0 00 11 32 0 00 0.71 23 08 18 75 Rt R9 R7 RS R8 R8 R8 R7 R8 R8 R8
~O 8 98 0 00 18 74 0 00 13 47 0 00 10 83 10 18 22.78 15 40 0 00 R7 R9 R8 R9 R8 R8 R9 R8 R7 R8 R9 18 17 87 10 83 0 00 15 58 0.00 11 07 0 00 11 80 23.28 0.00 8 97 R7 R9 R8 R9 R8 R8 R8 R9 R8 R9 16 0 00 10 82 0 00 17.08 0.00 9.87 0.00 11 18 22 80 22 71 R9 R7 R8 R7 R9 .R8 R7 RR R7 0 00 17 19 8 28 15 43 0 00 11 85 17 38 8 44 18.89 R9 R7 R8 R7 R8 R$
R7 R9 R7 12 15 54 0 00 15.50 8.25 15 03 10.70 22 89 14.87 0 00 R8 R9 R7 R9 R8 R8 R9 R7 R9 0 00 9.85 0.00 15.13 0.00 10.97 23.15 0 00 13.77 R8 R9 R8 R8 R8 R7 R8 R9 g
0 00 11 18 0 00 11 82 10 82 11 18 17.48 11 20 06 22 78 23 20 O.00 10 84 0.00 11 22 17 21 R8 R8 R$ R$
04 9 87 10 18 11 40 22 78 R6 - P8DPB289, RELOAD 6 R7 - P8DPB289, RELOAD 7 R8 R8 R$ SUNOLE 10 R8 - P8DPB289, RELOAD 8 02 R9'- PSDPB289, RELOAD 9 22 98 23 27 23.13 EXPO 8URE 10W0/ST)
FIGURE 3.2.1_
VY CYCLE 10 DESIGN REFERENCE LOADING PATTERN, LOWER RIGHT QUADRANT
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4.0 FUEL M!CHANICAL AND THERMAL DESIGN 4.1 Mechanical Design i
One hundred and eight (108) fresh fuel bundles fabricated by the General Electric Co. will be inserted into the Vermont Yankee reactor for Cycle 10 operation. The mechanical design parameters are identical to the General Electric fabricated bundles which were inserted and irradiated during Cycles 7, 8 and 9. Table 4.1.1 identifies the major design parameters.
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 10 reactor operation. Mechanical and chemical compatibility of the fuel assemblies with the in-service reactor environment is also addressed in Reference 2.
4.2 Thermal Design The fuel thermal effecte calculations were performed using the FROSSTEY computer codo [ 5-7] . The FROSETEY code calculatec pellet-to-clad gap i 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.
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The thermal effects analysis included the calculation of fuel j l
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 c,alculations integrate the responses of individual fuel batch average operating histories over the core average exposure range of Cycle 10. The gap conductance values are weighted axially by power distributions and radially by volume. The cure-wide gap conductance values for the RETRAN systen simulations, described in Sections
?.1 and 7.2, are from this data set at the particular exposure statepoints.
The gap conductance values input to the hot channel (RETRAN/TCPYA01) 1 calculations were evaluated for the P818R fuel bundle type as a function of the assembly exposure. The calculation assumed a 1.4 chopped cosine axial
p:wer shape with the peak power node running at the MAPLHGR limit defined in Reference 8 for the P8X8R fuel type. Figure 4.2.2 provides the hot channel r:cponse of gap conductance. In Figure 4.2.2, " planar exposure" refers to the cuposure 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 eithin 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 (rection 7.1).
Fuel rod local linear heat generation rates at fuel centerline incipient seit and 1% clad plastic strain as a function of local axial segment exposure for the gadolinia concentrations used in Vermont Yankee P8X8R fuel ccre previously reported in Reference 1.
4.3 Operating Experience All fuel bundles scheduled to be reloaded in Cycle 10 have operated as cxpected in previous cycles of Vermont Yankee. Off-gas measurements are at normally low levels indicating that no fuel failures are present.
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TABLE 4.1.1 NOMINAL FUEL MECHANICAL DESIGN PARAMETERS FUEL TYPE P8X8R 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, inches 9.5 Fuel Rod Pitch, inches 0.640 Diametral Gap (cold), inches 0.009 Fill Gas Helium Fill Gas Pressure, psig (See Ref. 2)
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.C80 Fuel Assembly Fuel Rod Array 8x8 Fuel Rods per Assembly 62 Spacer Grid Material - Zr-4 TABLE 4.2.1 CAP CONDUCTANCE VALUES USED IN VY CYCLE 10 TRANSIENT ANALYSES Cycle Exposure Core Average Hot Channel Hot Channel Statepoint Gap Condugtance Bundle Exposure Gap Condugtance #
(BTU /Hr-Ft - F) (MWD /ST) (BTU /Hr-Ft -
F)
(MWD /ST) 750 9680(1) 1370 BOC10 E0C10-2000 MWD /ST 960 6580 1050 975 7670 1140 EOC10-1000 MWD /ST 990 8650 1240 EOC10 NOTE (1) Between BOC and EOC-2000 MWD /ST, the highest exposure limiting hot channel bandle is once-burned.
VERMONT YANKEE - CORE RVERAGE GRP CONDUCTRNCE P8X8R FUEL -- GRP CONDUCTRNCE VS EXPOSURE sen.no num. roen se
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! CYCLE EXPOSURE (GWD/ST)
FICURE 4.2.1
' VY CYCLE 10 CORE AVERACE CAP CONDUCTANCE VERSUS CYCI.E BURNUP
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, VERMONT YRNKEE HOT CHRNNEL GRP CONDUCTRNCE E
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5 BE N f k0 1 2 3 4 5 8 7 8 9 10 11 12 13 14 15 18 17 18 19 20 21 22 23 24 25 28 27 28 PLfMIR EXPOSURE (GWD/ST) 5 7 5 5 [0 11 [2 b [1 b [8 l'7 18 19 b d i 2 h 4 $
BUNDLE EXPOSURE (GWD/ST)
FIGURE 4.2.2 VERMONT YANKEE HOT C11ANNEL CAP CONDUCTANCE FOR P8X8R VERSUS EXPOSURE
5.0 NUCLEAR DESIGN 5.1 Core Power Distributions l
The cycle was depleted using SIMULATE (3] to give both a rodded d2pletion and an All-Rods-Out (ARO) Haling depletion. The Haling depletion carves as the basis for defining core reactivity characteristics for most trcasient and accident evaluations. This is primarily because its flat power chape has conservatively weak scram chara:teristics. The rodded depletion was ussd to evaluate the misloaded bundle error and the rod withdrawal error.
This is because of the more realistic predictions it makes of initial CPR vclues. It was used in the calculation of rod drop worth because it burns the tcp of the core more realistically than the Haling.
5.1.1 Haling Power Distribution The Haling power distribution is calculated in the All-Rods-Out ccndition. The Haling iteration converges on a self-consistent power and exposure shape for the exposure step to end of cycle. In principle, this chould provide the overall minimum peaking power shape for the cycle. During the actual cycle, flatter power distributions cight occasionally be achieved by shaping with control rods. However, such shaping would leave underburned ragions 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.
l 5.1.2 Rodded Depletion Power Distribution To generate the rodded depletion, control rod, patterns were developed chich gave critical eigenvalues at each point in the cycle and gave peaking cinilar to the Haling calculation. The resulting patterns were frequently ocre peaked than the Haling, but were not in excess of expected operating lisits. However, as stated above, the underburned regions of the core can exhibit peaking in excess of the Haling peaking when pulling ARO at end of cycle. Figures 5.1.3 and 5.1.4 give the ARO end of cycle power distributions fcr the rodded depletion. Note in Figure 5.1.4 that the average axial power et ARO for the redded depletion is more bottom peaked than the Haling (Figure 5.1.2). This would result in better scram characteristics.
i 5.2 Core Exposure Distributions l
Cycle 10 was calculated to be capable of a cycle exposure of 7222 MWD /ST at EOFPL (no coastdown) Table 3.3.1 summarizes the resultant core j cverage exposures. The projected 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 to give the exposure distributions at various l cid-cycle points. BOC, EOC-2000 MWD /ST, EOC-1000 MWD /ST, and EOC conditions were used to develop reactivity input for the core wide transient analyses.
The rodded depletion may differ from the Haling during the cycle due to th2 shaping of the power by the rods. However, rod sequences are swapped fraquently and the overall exposure distribution at end of cycle is similar to ths Haling. Figure 5.2.2 gives the EOFPL radial exposure distribution for the redded depletion.
5.3 Cold Core Reactivity and Shutdown Margin The cold K,ff with all rods withdrawn (ARO) and the cold K,ff with j all rods inserted (ARI) at BOC were calculated using the SIMULATE code [3,4]
cnd are shown in T.M. J.3.1. K,ff with ARO minus the cold critical K,ff io the amount of excess core reactivity. K,ff with ARI sinus the K,ff with ARO is the worth of all the control rods.
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The cold critical eigenvalue K,ff was defined as the average calculated critical eigenvalue minus a 95% confidence level uncertainty. Then I cil cold results were normalized to make the critical K,f f equal to 1.000.
Technical Specifications [8] state that, for sufficient shutdown ccrgin, the core must be auberitical by at least 0.25% +R (defined below) with tha strongest worth control rod withdrawn. Again, using SIMULATE, a search esa 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,ff with the strongest rod out was. calculated at approximately 900.
MWD /ST intervals through the cycle. Subtracting each cold K df with the str:ngest rod out from the cold critical K,ff defines the shutdown margin as
o function of exposure. Figure 5.3.1 is the result. Because the local rsactivity may increase with exposure, the shutdown margin may decrease. To cccount for this, the value R is calculated as the difference between the cold K gg with the strongest rod out at BOC and the maximum cold K,gg with the strongest rod out in the cycle. The R for Cycle 10 is given in Table 5.3.1.
5.4 Standby Liquid Control System Shutdown Capability The shutdown capability of the standby liquid control system (SLCS) is d2 signed to bring the reactor from full power to cold, ARO, xenon free shutdown with at least 5% margin. Using the boron concentration search option in SINULATE [3), the ppa of boron was adjusted until the K,ff reached the This case assumed cold, xenon-free cold critical K,ff minus .05.
ecoditions, with All-Rods-Out at the most reactive time in the cycle. The criticality search found that the plant would be 5% suberitical at the worst point in time with 670 ppm of boron injected. VY Technical Specifications [8]
rsquire a minimum of 800 ppm of boron be available for injection. Table 5.4.1 lists the amount of boron concentration and the corresponding shutdown margin ecpability of the SLCS.
TABLE 5.3.1 VY CYCLE 10 K,gg VALUES AND SHUTDOWN MARGIN CALCULATION BOC K,fg - Uncontrolled 1.1142 BOC K,ff - Controlled .9676 Cold Critical K,ff Eigenvalue 1.0000 BOC K,ff - Controlled With .9854 Strongest Worth Rod Withdrawn BOC Minimum Shutdown Margin With 1.46% a K Strongest Worth Rod Withdrawn R, Maximum increase in Cold K,ff .28% AK With Exposure Cycle Minimum Shutdown Margin 1.18% a K at 6000 MWD /ST With Strongest Worth Rod Withdrawn TABLE 5.4.1 VY CYCLE 10 STANDBY LIQUID CONTROL SYSTEM SHUIDOWN CAPABILITY ppa of Boron Shutdown Margin 670 .050A K 800 .0,76, A K VERMON! YANNEE CYCLE 10 HALING DEPLETION ECO BUNOLE AVERADE RELATIVE POWERS PLANT 23 25 27 29 31 33 35 37 39 41 43
- OORD R8 R7 R9 R8 R8 R7 R8 R8 R9 R8 R8 1 002 1 120 1 358 1 042 1 211 1 143 1 348 1 153 1 189 0.837 0.444 R7 R9 R8 R9 R7 R9 R7 R9 R8 R6 R8 1 119 1 381 1 272 1 387 1 134 1 354 1.157 1 270 1 021 0.792 0.415 R9 R8 R7 R8 R9 R7 RS R8 R9 R8 R8 g
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VERMONT YANNEE CYCLE 10 HALING DEPLETION EOC SUNOLE AVERROE EXPOSURES l PLANT 23 25 27 29 31 33 35 37 39 41 43 COORD R8 R7 R9 R8 R8 R7 R8 R8 R9 R8 R6 22 8 44 26 28 28 99 23 43 9 79 30 18 17.20 23 09 9 72 19.00 15.76 R7 R9 R8 R9 R7 R9 R7 R9 R8 R8 R6 23 49 9.81 16 15 9.86 24.94 9 75 21.85 9 17 18.01 15.89 25.75 R9 R8 R7 R8 R9 R7 R8 R8 R8 R8 R8 18 25 88 9.79 16.18 25 81 19.53 9.79 23.78 9.46 19 03 7 67 16.86 R6 R9 R8 R9 R7 R9 R8 R9 R8 RB 16 30.24 9 85 19 52 9.81 25.09 9 41 17.88 8.27 17 13 26 42 R8 A7 R9 R7 R8 R7 R9 R8 R7 17 18 24.89 9.79 25 20 16.88 22 39 8.42 18 18 21 86 R7 R9 R7 R9 R7 R8 R7 R8 R6 22 95 9.75 23.74 9 41 23 06 15 84 21 53 16 35 26 15 R9 R7 R9 R8 R9 R7 R9 R8 R6 9.72 22 11 9.48 17.88 8 42 21.82 8 71 15.70 25.73
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VEFMONT YRNMEE CYCLE 10 R000E0 DEPLETION EOC SUNOLE RVERADE EXPOSURES PLANT 23 25 27 29 31 33 35 37 39 41 43 COORD R8 A7 RS R8 R8 R7 RS R8 R$ R8 R$
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6.0 THERMAL-HYDRAULIC DESIGN The thermal-hydraulic evaluation of the reload cyc1r was performed scing the methods described in the following section.
6.1 Steady-State Thermal Hydraulics i
Core steady-state thermal-hydraulic analyses were performed using ,
ths FIBWR [9,10] computer code. The FIBWR code incorporates a detailed gsometrical 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.
Th3 power distribution, inlet enthalpy, and geometry are presumed known and ora supplied to FIBWR. The power distribution is derived by the 3-D asutronic simulator SIMULATE [3]. Core pressure drop and total leakage ficw predicted by the FIBWR code were used in setting the initial ecnditions for th* system's transient analysis model.
6.2 Reactor Limits Determination i
The objective for normal operation and anticipated transient events in to meintain nucleate boiling and thus avoid a transition,to film bailing, thereby protecting the fuel cladding integrity. Based on Reference 11, the fuel cladding integrity safety limit for Vermont Yankee is a lowest allowabic minimum critical power ratio (LAMCPR) of 1.07 for P8K8R reload fuel. Operating limits are specified to maintain adequate cargin to onset of the boiling transition. The figure of merit utilized for plant operation is the critical power ratio (CPR). This is defined as th2 ratio of the critical power (bundle power at which some point within tha sesembly experiences onset of boiling transition) to the operating bundle power. Thermal margin is stated in terms of the minimum value of tha critical power ratio NCPR, which corresponds to the most limiting fuel ,
coceably in the core. Both the transient (safety) and normal operating th2rmal limits in terms of MCPR are derived based on the GEXL correlation os described in Reference 11. .
Vermont Yankee Technical Specifications [8] limit the operation of P8X8R fuel to a maximum linear heat generation rate (EHCR) of 13.4 KW/f t.
The basis for a EHGR of 13.4 KW/f t can be found in Reference 2.
j
7.0 ACCIDENT ANAL *lSIS 7.1 Core Wide Transient Analysis Core wide transient simulations are performed to assess the impact of th3 particular transient on the heat transfer characteristics of the fuel.
!h3 figure of serit used is the critical power ratio (CPR). It is the purpose of the analysis to determine the minimum critical power ratio such that the esfety limit critical power ratio (LAMCPR) is not violated for the transients considered.
7.1.1 Methodology The analysis requires two types of simulations. A system level siculation is performed to determine the overall plant response. Transient core inlet and exit conditions and normalized power from the system level esiculation are used to perfora detailed thermal-hydraulic simulations of the fu21 (referred to as " hot channel calculations"). The result of each of these latter simulations is the bundle transient a CPR (the initial bundle CPR minus th2 minimum CPR experienced during the transient).
The system level simulations are performed with the model documented in Reference 12.
The hot channel calculations are performed with the RETRAN [13] and TCPYA01 [14] computer codes. The CEXL correlation [11] is used in TCPYA01 to evaluate critical power ratio. The calculational procedure is outlined below.
The hot channel transient A CPR calculations are performed via a series of " inner" and " outer" iterations, as illustrated by the flow chart in Figure 7.1.1. The outer loop represents iterations on the hot channel initial p wer level. These iterations are necessary, because the o CPR for a given transient varies with Initial Critical Power Ratio (ICPR), yet only theA CPR ccrresponding to a transient MCPR equal to the safety limit (i.e.,1.07 +
A CI1t = ICPR) is appropriate. The approximate constancy of thea CPR/1CPR ratio is useful in these iterations. Each outer iteration requires a RETRAN h2t channel run to calculate the transient enthalpies, flows, pressure and
caturation properties at each time step. These are required for input to the TCPYA01 code. TCYPA01 is then used to calculate a CPR at each time step daring the transient, from which a transient ACPR is derived. The hot channel to modeled using a chopped cosine axial power shape with a peak / average ratio ef 1.4.
The inner loep represents iterations on the hot channel inlet flow.
Th:se iterations are necessary, because the RETRAN hot channel model calculates the entrance loss coefficient when given the initial 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 given power level until the calculated entrance loss coefficient is correct. FIBWR [9] is utilized to estimate the correct inlet flow for a particular power level and prassure drop.
7.1.2 Initial Conditions and Assumptions The initial conditions for the system simulations are based on 105%
reted steam flow (maximum turbine capacity) and 100% core flow. The core cxial power distribution for each of the exposure points is based on Haling-mode 3-D SINULATE predictions associated with the generation of the racetivity data (Section 7.1.3). The core inlet enthalpy is set so that the crount of carryunder from the steam separators and the quality in the liquid ragion outside 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 opsrating state used for the system simulations is provided in Table 7.1.1.
Assumptions specific to a particular transient are discussed in the 02ction describing the transient. In general, the following assumptions are t orde for all transients:
- 1. Scraa setpoints are at Technical Specification limits.
- 2. Protective system logic delays are at equipment specification limits.
i l
l 1
- 3. Safety / relief valve and safety valve capacities are based on i Technical Specification rated values. l 1
- 4. Safety / relief valve and safety valve setpoints are modeled as being 1% above the Technical Specification upper limit. Valve responses are based on slowest specified recponse 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 r
referred to as the " measured" and "67B" scram time sets.
7.1.3 Reactivity Functions The methods used to generate the fuel temperature, moderator density, cad scram reactivity functions are described in Reference 15 and are outlined in Figures 2.1 through 2.3 of that document. A complete set of reactivity functions, the axial power distribution, and the kinetics parameters are gsnerated from base states established for EOC, EOC-1000 MWD /ST, E00-2000 MWD /ST, and BOC exposure conditions. These states are characterized by cxposure and void history distributions, control rod pattern, and core thsreal-hydraulic conditions. The latter core conditions are consistent with tha assumed system transient conditions provided in Table 7.1.1.
The BOC base state is established from the previously defined Cycle 9 andpoint, the Cycle 10 reload pattern, and an estimate of the BOC10 critical red pattern. The EOC and intermediate core exposure and void history distributions were calculated via a Haling depletion,as described in Section 5.2. The EOC state is unrodded and, as such, is defined bufficiently.
Hswever, EOC-1000 MWD /ST and EOC-2000 MWD /ST exposure points require base control rod patterns. These are developed to be as " black and white" as pseeible. That is, beginning with the rodded depletion configuration, all ecstrol rods which are more than half inserted are fully inserted, and all ccatrol rods which are less than half inserted are fully withdrawn. If the SINULATE-calculated parameters are within operating limits, then this configuration becomes the base case. If the limits are exceeded, a minimum nu:ber of control rods are adjusted a minimum number of notches until the parameters fall within limits. Using this method, the control rod patterns cod resultant power distributions are established so as to minimize the scram rscetivity function and to maximize the core average moderator denalty rssetivity coefficient. For the transients analyzed, this tends to maximize f th2 power response.
I In generating the fuel react ivity function data for RETRAN, twelve unique volume-specific table sets ara produced which are analogous to those ohown in Figure 3.7 of Reference 15. The moderator and relative moderator d2nsity functions also are twelve unique volume-specific tables, analogous to Figures 3.10 and 3.11 in Reference 15. A mcderator density set is generated cpicifically for each transient type. The density reactivity functions for tha subco' ling transient are generated by quasi-statically varying the inlet subcooling only. The s. '.arator enthalpy source distribution is in equilibrium with the calculated nuclear power. The density reactivity functions of the prassurization transients are generated by quasi-statically varying the core
-pressure. A series of the calculations are performed for various inlet cederator temperatures. The moderator enthalpy source distribution is that of ths base state case.
in order to qualitatively compare the cr,re reactivity characteristics between different base configurations, core average reactivity coefficients era calculated and provided in Table 7.1.2. Calculated point kinetics parameters for RETRAN are also provided in the table.
The reactivities versus scram insertion are calculated at constant, pre-transient moderator conditions. These calculated data are fit and evaluated to yield highly detailed scram reactivity curves. These are then combined with the appropriate rod position versus time curves to establish the final RETRAN scras reactivity functions. Figures 7.1.2 through 7.1.4 display tha inserted rod worths and rod positions as functions of scram time for the "otasured" scram time analysis. Figures 7.1.5 through 7.1.i display similar curves for the "67B" scram time analysis.
7.1.4 Transients Analyzed Past licensing experience has shown that the core wide transients which
! racult in the minimum core thermal margins are:
1
- 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 covere transient for Vermont Yankee, because of the plant's 110% steam flow bypass system. Past analyses have shown this transient to be considerably 1 sos severe than any of the above for all exposure points. Brief descriptions and tne results of the core wide transients analyzed are provided in the following section.
I 7.2 Core Wide Transient Analysis Results 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 feedwater heating was also evaluated at beginning of cycle (BOC) conditions. A summary of the results of the analyses is provided in Table 7.2.1.
l 7.2.1 Turbine Trip Without Bypass Transient (TTWOB) l t
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 acrually open to relieve pressure, remain closed. A reactor protection system oignal is generated by the turbine stop valve closure switches. Control rod drive motion is conservatively assumed to occur 0.27 seconds after the start of turbine stop valve motion. The ATWS recirculation pump trip is assumed to secur at a setpoint of 1150 psig done pressure. A pump trip time delay of 1.0 second is assumed to account for logic delay and M-G set generator field
collapse. In simulating the transient, the bypass piping volume up to the valve chest is lumped into the control volume upstream of the turbine stop velves. As an example, predictions of the salient system parameters are shown in Figures 7.2.1 through 7.2.3 for the three exposure points for the
" measured" scram time analysis.
7.2.2 Generator Load Rejection Without Bypass Transient (GLRWOB)
The transient is initiated by a rapid closure (0.3 seconds closing tire) 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 cignal is generated by the hydraulic fluid pressure switches in the acceleration relay of the turbine control system. Control rod drive motion is censervatively assumed to occur 0.28 seconds after the start of turbine control valve motion. The same modeling regarding the ATWS pump trip and bypass piping is used as in the turbine trip simulation. The influence of the eccelerating main turbine generator on the recirculation system is simulated by specifying the main turbine generator electrical frequency as a function of tine for the M-G set drive motors. The main turbine generator frequency curve in based on a 100% power plant startup test and is considered representative fcr the simulation. As an example, the system model predictions for the three exposure points are shown in Figures 7.2.4 through 7.2.6 fcr the " measured" ceram time analysis.
7.2.3 Loss of Feedwater Heating Transient (LOFWH)
A feedwater heater can be lost in such a way that the steam extraction line to the heater is shut of f or the feedwater flow bypasses one of the hanters. In either asse, the reactor will receive cooler feedwater, which will produce an increase in the core inlet subcooling, resulting in a reactor pswer increase.
The response of the system due to the loss of 100*F of the feedwater h cting capability was analyzed. This represents the current licensing occumption for the maximum expected single heater or group of heaters that can b2 tripped or bypassed by a single event.
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I Vermont Yankee has a screa setpoint of 120% ef rated power as part of th2 Reactor Protection System (RPS) on high neutron flux. In this analysis, to credit was taken for scram on high neutron flux, thereby allowing the esector power to reach its peak without scran. This approach was selected to prcvide a bounding and conservative analysis.
The transient response of the system was evaluated at several exposures during the cycle. Transient evaluation at EOC-2000 MWD /ST was found to be the liciting case between BOC to EOC. The results of the system response to a less of 100*F feedwater heating capability evaluated at EOC-2000 MWD /ST as pradicted by the RETRAN code are presented in Figure 7.2.7.
7.3 Overpressurization Analysis Results Compliance with ASNE vessel code limits is demonstrated by an analysis of the main steam isolation valves (MSIV) closing with failure of the MSIV position switch scras. End of cycle conditions were analyzed. The system cedel 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 applicable to this simulation. The maximum pressure at the bettom of the reactor vessel is calculated to be 1266 psig for the " measured" ceran time analysis and 1291 psig for the ~67B" scram time analysis. These racults are within the allowable code limit of 10% above vessel design pressure for upset conditions, or 1375 psig.
The transient is initiated by a simultaneous closure of all four MSIV's. A 3.0 second closing time, which is the Technical Specification cinimum, is assumed. A reactor scraa signal is gepe, rated on APRM high flux.
Control rod drive motion is conservatively assumed to occur 0.28 seconds after racching the high flux setpoint. The system response is shown in Figure 7.3.1 for the " measured" scram time analysis.
7.4 Local Rod withdrawal Error Transient Results The rod withdrawal error is a loca1 core transient caused by an cp rator erroneously withdrawing a control rod in the continuous withdrawal code. If the core is operating at its operating limits for MCPR and LHGR at
I l
l the time of the error, then withdrawal of a control rod could increase both l Iccal 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 many rituations it would be possible to fully withdraw a control rod without exceeding 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 p rtion of the core MCPR operating limit envelope is specifically defined such that the cladding limits are not violated. The consequences of the error dapend on the local power increase, the initial MCPR of the neighboring 1ccations and the ability of the Rod Block Monitor System to stop the withdrawing rod before MCPR reaches 1.07.
The most severe postulated transient begins with the core operating eccording to normal procedures and within normal operating limits. The aparator makes a procedural error and attempts to fully withdraw the maximum worth control rod at maximum withdrawal speed. The core limiting locations sie close to the error rod and therefore experience the spatial power shape trcnsient as well as the overall core power increase.
l The core conditions and control rod pattern for the bounding case are cpscified 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 the exposure corresponding to maximum control inventory with the xenon-free con,d,ition superimposed. The xenon-free condition 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 105% power and rated flow.
- 3. The core power distribution is adjusted with the available control rods to place the locations within the four by four array of bundles around the error rod as nearly on the operating limits as practical. j The Rod Block Monitor System's ability to terminate the bounding case
! to evaluated 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 tested in this analysis, there are a maximum of four LPRM inputs per level. One RBM channel avereges 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 i combinations of none, one and two failed LPRM strings. 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 rv*
Technical Specifications require en t 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 j tha following notch.
The analysis is performed using the three dimensional steady state
- 21NULATE core model demonstrated in Reference 3. Necessary properties of that todal for use in this analysis are
i
- 1. Accurate bundle power calculation as shown by the PDQ and gamma scan comparisons.
- comparisons.
- 3. Accurate control rod worths and core power coefficient as shown by the consistent core eigenvalues.
l Two separate cases are presented from numerous explicit SIMULATE cualyses. The reactor conditions and case descriptions are shown in Figures 7.4.1 and 7.4.2. Case 1 maalyzes the bounding event with the concurrent abnormal menon condition and rod pattern configuration necessary to increase the worth of the error rod. The initial conditions for Case 2 approximate the
" normal" 105% power conditions at the most reactive point in the cycle; the ccatrol rod density is at its maximum at the normal equilibrium xenon condition. The A CPR 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 + A CPR, conserving the figure of merit (ACPR/ Initial CPR) chown by the SINULATE calculations. The use of this method provides valid ACPR values in the analysis of neraal operating states where locations near the assumed error rod are not initially near the MCPR operating limit. Case 2 is the worst of thirteen rod withdrawal transients analyzed from normal initial 105% power, full flow and rod pattern conditions at various exposure I points throughout the cycle. Case 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 rarponse first intercepts the RBM setpoint. For this same bounding case, the opsrating limit ACPR envelope component versus Rod. Block Monitor setpoint is teken from the Table '.4.1. The same table demonstrates margin to the 1%
plcstic 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 i
_ ~ . . . _ _ _ _ , _ _ . _ . _ . _ . _ , . _ _ . _ . _ _ _ _ _ _ _ . . . . . -_ _ __ _ _ __
r currounding vide water gaps. In addition, there may be a small increase in r activity, depending on the exposure and void fraction states. The R-factor iteresse results in a CPR reduction, while che local pin peaking factor increase results in a higher pin linear heat generation rate. The objective cf 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 liatting 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 increase raculting from a rotation to the exposure dependent R-factors of the properly criented bundles [11). Using these rotated bundle R-factors, the miniaua CPR values resulting from a bundle rotation are determined using SIMULATE. This to done for each control rod sequence throughout the cycle. These miniaua CPR values are, in addition, modified slightly to account for the change in racctivity resulting from the rotation. For each sequence, the MCPR for-the properly oriented assemblies is adjusted by a r&tio necessary to place the corresponding rotated CPR on its 1.07 safety 3imit. The maximum of these i edjusted MCPR's is the rotated bundle operating limit.
To determine the acximum linear heat generation rate (MLHGR) resulting frca a rotation, the ratios of the maximum rotated local peaking factor to the ccximum unrotated local peaking are determined for the expected range of exposure and void conditions. The maximum of this ratio is applied to the cparating limit LHGR of 13.4 kw/ft. This maximum rotated bundle LHGR is in codition modified to account for the possible reactivity increase resulting frca the rotation. It is also increased by the 2.2% power spiking penalty.
The results of the rotated bundle analysis are as given in Table 7.5.1.
f 7.5.2 Mislocated Bundle Error i
I Misloading a high reactivity assembly into a region of high neutron l
1:psrtance results in a location of high relative assembly average power.
i Since the assembly is assumed to be properly oriented (not rotated), R-factors utod for the misloaded bundle are the standard values for the fuel type.
The analysis for Cycle 10 consists of an iterative procedure which l ouccessively eliminates potential misloading locations from any MCPR safety limit violations. The first step is to use SIMULATE to determine the largest l j possible ACPR which could result, at any location, as the result of misloading ,
I a high reactivity assembly into the location. This maximum ACPR is then l
! cubtracted 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 cill never exceed the MCPR safety limit of 1.07. These locations are l
oliminated from further investigation.
The next iteration consists of applying the same procedure to the 1ccations which appeared to violate the safety limit when the maximum ACPR frem the first iteration was applied. Since these locations are of higher racctivity than those eliminated in the first iteration, they will result in a caaller ACPR when misloaded. Using this smaller ACPR, some of the remaining Jocations will be eliminated from potential CPR safety limit violations. This procedure is continued until all locations are shown to be above the MCPR cafety limit due to a misloading, or until a limiting location is identified.
Using the above procedure, it has been demonstrated that for Cycle 10 all possible mislocations resulted in calculated MCPR's above the 1.07 safety licit, assuming an initial operating CPR limit of 1.22. This makes the ciclocated bundle analysis less limiting than the rotated bundle analysis.
7.6 Control Rod Drop Accident Results 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 worch refers to the fr u: that rods beyond Group 2 cro banked out of the core and can only fall the increment from all in to the red drive withdrawal position. Acceptable worth is one which produces a maximum fuel enthalpy less than 280 calories / gram.
4
1 Some out-of-sequence control rods could accrue potentially high t:srths. However, the Rod Worth Minimizer (RWM) will prevent withdrawing an cut-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-Rods-In r
(ARI) to well above 20% core thermal power. Above 20% power even multiple i cparator errors will not create a potential rod drop situation above 280 colories per gram. [16, 17] Below 20% power, however, the sequences sust be cxcained for incremental rod worth. This is done using the full core, xenon i frse SIMULATE model at the projected most reactive point in the cycle. This escures that the maximum amount of reactivity is held in the rods.
Both the A and 8 sequences were examined. It was found that the highest worth occurred in the first rod 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 as tha 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 accond 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.
i Beyond Group 2, procedures [18] apply which severely reduce the rod incremental worths. This makes the xenon free, hot standby worths much less then the cold xenon free worths. [1]
7.7 Stability Analysis Results J f The analysis of reactor stability has been performed by General Electric as described in Section S.2.4 of Reference 2. The 105% power rod line was analyzed and the resultant decay ratio as a function of reactor power laval is provided in Figure 7.7.1.
The reactor core stability decay ratio at natural circulation c:nditions and a power level corresponding to the 105% power rod line, is calculated to be 0.87. The channel hydrodynamic performance decay ratio coacciated with this condition is 0.30.
TABLE 7.1.1 VY CYCLE 10
SUMMARY
OF SYSTEM TRANSIENT MODEL INITIAL CONDITIONS FOR CORE WIDE TRANSIENT ANALYSES Core Thermal Fower (MWth) 1664.0 Turbine Steam Flow (% NBR) 105 Total Core Flow (10 61bm/hr) 48.0 Core Bypass Flow (1061bm/hr) 5.3 Core Inlet Enthalpy (BTU /lba) 520.9 i
Steam Dome Pressure (psia) 1034.7 Turbine Inlet Pressure (psia) 986.0 Total Recirculation Flow (10 61bm/hr) 23.4 Core Plate Differential Pressure (psi) 18.5 Narrow Range Water Level (in.) 35 Average Fuel Gap Conductance (See Section 4.2) d
TABLE 7.1.2 VY CYCLE 10 TRANSIENT ANALYSIS REACTIVITY COEFFICIENTS Calculated Parameter Cycle Exposure Point (MWD /ST)
EOC (EOC-1000) (EOC-2000) POC Axial Shape Index(l) -0.0824 -0.1802 -0.2137 -0.2463 Moderator Density Coefficient 20.8 22.0 (Subcooling), 4/Au(2) 24.4 19.7 Pressure = 1050 psia i Subcooling = 30 BTU /lbe Moderator Density Coefficient 23.3 24.3 (Pressurization), 4/Au 26.3 (3)
Pressure = 1050 psia
, Inlet Enthalpy - 520 BTU /lba l $'
Fuel Temperature Coefficient -0.282 -0.283 at 11300F, l/0F -0.283 -0.261 Effective Delayed 0.005390 0.005473 0.005557 0.006028
+
Neutron Fraction Prompt Neutron Generation 42.67 42.49 41.81 40.07 Time in Microseconds 1
P -P Notes: T (1) Axial Shape Index (ASI) = p ,,B j (2) Au = change in density, in percent I
(3) Pressurization transients are not calculated at BOC
l TABLE 7.2.1 VY CYCLE 10 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 EOC 2.931 1.183 .19 Without Bypass,
" Measured" EOC-1000 2.189 1.102 .10 Scram Time EOC-2000 1.208 1.000 .00 Turbine Trip EOC 3.433 1.234 .25 Without Bypass, "67B" EOC-1000 2.694 1.168 .17 Scram Time EOC-2000 1.634 1.031 .03 Generator Load EOC 2.793 1.164 .18 Rejection Without Bypass, EOC-1000 2.076 1.078 .09
" Measured' Scram Time EOC-2000 1.086 1.000 .00 l
Generator Load EOC 3.381 1.225 .26 Rejection Without Bypass, EOC-1000 2.658 1.155 .18 "67B" Scram Time EOC-2000 1.522 1.012 .02 Luss of 1000F EOC 1.198 1.190 .15 Fasdwater H2cting EOC-1000 1.210 4 203 .16 EOC-2000 1.215 1.207 .17 BOC 1.201 1.193 .15 TABLE 7.4.1 VY CYCLE 10 ROD WITHDRAWAL ERROR TRANSIENT
SUMMARY
(WITH LIMITING INSTRUMENT FAILURE) 1 Case 1 Conditions in Figure 7.4.1 RBM Rod ACPR MLHGR (kw/ft)
Setpoint Position P8X8R P8X8R 104 10 .09 13.7 105 12 .12 13.9 106 14 .17 15.8 107 14 .17 15.8 108 16 .21 17.7 Case 2 Conditions in Figure 7.4.2 RBM Rod LCPR MLHGR (kw/ft)*
Setpoint Position P8X8R P8X8R 104 22 .06 12.7 105 24 .07 12.7 106 26 .09 12.7 107 30 .11 12.7 108 34 .15 13.4 0 Not initially on limits I
l
- TABLE 7.5.1 ROTATED BUNDLE ANALYSIS RESULTS Resulting Initial MCPR Resulting MCPR LHGR (kw/ft) 1.24 1.07 17.47
TABLE 7.6.1 CONTROL ROD DROP ANALYSIS - ROD ARRAY PULL ORDER 1
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 Array 1 Array 2 Array 3 or 4 Array 2 Array 1 Array 3 or 4 Array 3 Array 4 Array 1 or 2 Array 4 Array 3 Array 1 or 2 TABLE 7.6.2 VY CYCLE 10 CONTROL ROD DROP ANALYSIS RESULTS i
Maximum Incremental Rod Worth .84% AK
, Calculated Cold Xenon Free l
Bounding Analysis Worth for Enthalpy 1.30% AK Less than 280 Calories per Gram (References 16, 17 and 19) l l
l l
Choose ICPR l
I f Estimate Power F !
I f Estimate Flow With FIBWR I f RETRAN Flow ,
Initialization Run I f is No Revise
=
Coefficient 2 Correct Flow Yes if RETRAN/TCPYA61 Hot Channel Run 1
Has ACPR No Converged? New ICPR Yes STOP FIGURE 7.1.1 FLOW CHART FOR THE CALCULATION OF ACPR USING THE RETRAN/TCPYA01 CODES
53 = f :)
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FIGURE 7.1.2 INSERTED ROD WORTH AND ROD POSITION VERSUS TIME FROM INITIAL ROD MOVEMENT AT EOC10, " MEASURED" SCRAM TIME i
S
, E= -
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FIGURE 7.1.3 INSERTED ROD WORTH AFD ROD POSITION VERSUS TIME FROM INITIAL ROD MOVEMENT AT EOC10-1000 MWD /ST, " MEASURED" SCRAM TIME
8 iR o
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FIGURE 7.1.4 INSERTED ROD WORTH AND ROD POSITION VERSUS TIME FROM INITIAL ROD MOVEMENT AT EOC10-2000 MWD /ST, " MEASURED" SCRAM TIME
~ - . _ . . _ . _ - _ _ _ .
VY CYCLE 10 - 678 SCRAM, E00 S
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FIGURE 7.1.5 l
INSERTED ROD WORTH AND ROD POSITION VERSUS TIME FROM INITIAL ROD MOVDIENT AT EOC10, "67B" SCRAM TIME i
i VY CYCLE 10 - 678 SCRAM, E00-1 1 1
- R S
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FIGURE 7.1.6 INSERTED ROD WORTH AND ROD POSITION VERSUS TIME FROM INITIAL ROD MOVEMENT AT EOC10-1000 MVD/ST, "67B" SCRAM TLMF
VY CYCLE 10 - 678 SCRRM, E00-2
- E l
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FIGURE 7.1.7 INSERTED ROD WORTH AND ROD POSITION VERSUS TIME FROM INITIAL ROD MOVEMENT AT EOC10-2000 MWD /ST, "67B" SCRAM TIME
4 l
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N EOC E FIGURE 7.2.1-1 N EOC W TURBINE TRIP WITIIOUT BYPASS EOC10 1
1 TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME I
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FIGURE 7.2.1-2 TTEBP EOC MST TURBINE TRIP WITHOUT BYPASS, EOC10 TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME i
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TURBI_NE TRIP WITHOUT BYPASS, EOC10-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME t
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_ TURBINE TRIP WITHOUT BYPASS, EOC10-2000 MWD /ST I,
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e Operator attempts full withdrawal of the fully inserted rod at coordinates (22, 35).
e Bounding Case.
FIGURE 7.4.1 REACTOR INITIAL CONDITIONS FOR THE VY CYCLE 10 RWE CASE 1
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CONTROL ROD PATTERN 43 30
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Reactor Conditions:
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e Normal Xenon condition and control rod pattern.
FIGURE 7.4.2 REACTOR INITIAL CONDITIONS FOR THE VY CYCLE 10 RWE CASE 2
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- : A 0
0 20 40 60 80 100 120 PERCENT POWER FIGURE 7.7.1 REACTOR CORE DECAY RATIO VERSUS POWER
.s 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 -
ceram tests, power distribution comparisons TIP reproducibility, and TIP oyametry checks. The content of the Startup Test Report will be similar to that sent to the Of fice of Inspection and Enforcement subsequent to the start of Cycle 9 [20).
i Q
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 21 provide required support for the operation of Vermont Yankee Cycle 10. No new fuel types have been introduced in this reload, therefore, the MAPLHGR limits as a function of overage planar exposure remain the same as in the previous cycle. [1,8)
O
APPENDlX A CALCULATED CYCLE DEPENDENT LIMITS The MCPR limits appropriate for Cycle 10 are calculated by adding the calculated ACPR to the safety limit LAMCPR of 1.07. This is done for each of the analyses in Section 7 at each of the exposure statepoints. For an exposure interval between statepoints, the highest MCPR limit at either end is assumed to apply to the whole interval.
Table A.1 provides the highest calculated MCPR limits for Cycle 10 for ecch of the exposure intervals for the various scram speeds and for the
'crious rod block lines.
With regard to MAPLHGR, no new fuel types have been introduced. The MAPLHGR limits given f n Reference 8 f or the P8X8R fuel type apply to Cycle 10.
The MCPR limits presently employed in Cycle 9 are also bounding for Cycle 10.
These are given in Reference 8 and are reproduced here as Table A 2.
TABLE A.1 VERMONT YANKEE NUCLEAR POWER STATION CALCULATED CYCLE 10 MCPR LIMITS Average Control Rod Cycle MCPR Lisift for Value of "N" in RBM P8K8R Fue!
Equation (1) Scram Time Exposure Range 42% BOC to EOC-2 CWD/T 3.28
" MEASURED" EOC-2 CWD/T to EOC-1 CWD/T 3.28 l
EOC-1 CWD/T to EOC 1.28 BOC to EOC-2 CWD/T 1.28 "67B" EOC-2 CWD/T to EOC-1 CWD/T 3.28 EOC-1 CWD/T to EOC 1,33 41% BOC to EOC-2 CWD/T 1.24
" MEASURED" EOC-2 CWD/T to EOC-1 CWD/T 1.24 EOC-1 CWD/T to EOC 1.26 BOC to EOC-2 CWD/T 1.24 "67B" EOC-2 CWD/T to EOC-1 CWD/T 1.25 l
EOC-1 CWD/T to EOC 1.33 e < 40% BOC to EOC-2 CWD/T 1.24 O'
~
" MEASURED" EOC-2 CWD/T to EOC-1 CWD/T 1.24 EOC-1 CWD/T to EOC 1.26 EOC to EOC-2 CWD/T 1.24 "67B" EOC-2 CWD/T to EOC-1 CWD/T 3.25 EOC-1 CWD/T to EOC 1.33 tg 9 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 8].
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TABLE A.2 THE MCPR OPERATINC LIMITS FOR CYCLE 9 ARE BOUNDINC FOR CYCLE 10. THESE ARE FOUND IN REFERENCE 8 AS TABLE 3.11-2 MCPR Operating Limit for Value of "N" in RM Average Control Rod Cycle Fuel Type (2)
Equation (1) Scram Time Exposure Range 838 838R P838R 42% Equal or better BOC to EOC-2 WD/T 1.29 1.29 1.29 than L.C.O. EOC-2 CWD/T to EOC-1 CWD/T 1.29 1.29 1.29 3.3 C.1.1 EOC-1 WD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 CWD/T 1.29 1.29 1.29 than L.C.O. EOC-2 CWD/T to EOC-1 CWD/T 1.33 1.31 1.31 3.3 C.1.2 EOC-1 CWD/T to EOC 1.36 1.35 1.35 41% Equal or better BOC to EOC-2 WD/T 1.25 1.25 1.25 than L.C.O. EOC-2 WD/T to EOC-1 CWD/T .1.26 1.25 1.25 i 3.3 C.1.1 EOC-1 CWD/T to EOC 1.30 1.30 1.30 E Equal or better BOC to EOC-2 CWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 CWD/T to EOC-1 WD/T 1.33 1.31 1.31 3.3 C.1.2 EOC-1 WD/T to EOC 1.36 1.35 1.35
< 40% Equal or better BOC to EOC-2 CWD/T 1.25 1.25 1.25 than L.C.O. EOC-2 WD/T to EOC-1 CWD/T 1.26 1.25 1.25 3.3 C.1.1 EOC-1 CWD/T to EOC 1.30 1.30 1.30 Equal er better 800 to EOC-2 GWD/T 1 25 1.25 1.25 than L'.C.O. EOC-2 WD/T to EOC-1 CWD/T 1.33 1.31 1.31 3.3 C.I.2 EOC-1 WD/T to EOC 1.36 1.35 1.35 75% Special Testing at Natural Circulation (Note 3, 4) 1.30 1.31 1.31 (1) The Rod Block Monitor (RM) trip setpoints are determined by the equation shown in Table 3.2.5 of the Technical Specifications.
(2) The current analyses for MCPR Operating Limits do not include 727 fuel. On this basis further evaluation of MCPR operating 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 f actors are not applied during the pump trip and stability testing.
REFERENCES I
l 1. A. A. F. Ansari, et al., Vermont lankee Cycle 9 Core Performance Analysis, YAEC-1275, August 1981.
- 2. General Electric Standard Application for Reactor Fuel (CESTARII),
NEDE-240ll-P-A-5, GE Company Proprietary, August 1982.
l 3. D. M. VerPlanck, Methods for the Analysis of Boiling Water Reactors l Steady State Core Physics, YAEC-1238, March 1981.
- 4. E. E. Pilat, Methods for the Analysis of Boiling Water Reactors Lattice Physics, YAEC-1232, December 1980.
- 5. S. P. Schultz and K. E. St. John, Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Ef fects (FROSSTEY) Code /Model Description Manual, YAEC-1249P, April 1981.
- 6. S. P. Schultz and K. E. St. John, Methods for the Analysis of Oxide Fuel Rod Steady-State Thermal Effects (FROSSTEY) Code Qualification and Application, YAEC-1265P, June 1981.
- 7. D. C. Albright, H20DA: An Improved Water Properties Package, YAEC-1237, March 1981.
- 8. Appendix A tu Operating License DPR-28 Technical Specifications and Bases for Vermont Yankee Nuclear Power Station, Docket No. 50-271.
- 9. A. A. F. Ansari, Methods for the Analysis of Boiling Water Reactors:
Steady-State Core Flow Distribution Code (FIBWR), YAEC-1234, December 1980.
10 A. A. F. Ansari, R. R. Gay, and B. J. Citnick, FIBWR: A Steady-State Core Flow Distribution Code for Boiling Water Reactors - Code Verification and Qualification Report, EPRI NP-1923, Project 1754-1 Final Repo'rt, July 1981.
- 11. General Electric Company, CEKL Correlation Application to BWR 2-6 Reactors, NEDE-25422, GE Company Proprietery, June 1981.
'12. A. A. F. Ansari and J. T. Cronin, Methods for the Analysis of Boiling Water Reactors: A Systems Transient Analysis Model (RETRAN), YAEC-1233, April 1981.
- 13. EPRI, RETRAN - A Program for One-Dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, CCM-5, December 1978.
14 A. A. F. Ansari, K. J. Burns, and D. K. Beller, Methods for the Analysis i of Boiling Water Reactors: Transient Critical Power Ratio Analysis (RETRAN-TCPYA01), YAEC-1299P, March 1982.
! 15. J. M. Holzer, Methods for the Analysis of Boiling Water Reactors Transient Core Physics, YAEC-1239P, August 1981.
l l
i
- 16. C. J. Paone, et al., Rod Drop Accident Analysis for Large Boiling Water Reactors, NEDO-10527, March 1972.
17 R. C. Stirn, et al. , Rod Drop Accident Analysis for Large Boiling Water Reactors Addendum No. 1, Multiple Enrichment Cores With Axial Gadolinium, NEDO-10527, Supplement 1, July 1972.
18 D. Radclif fe and R. E. Bates, " Reduced Notch Worth Procedure", SIL-316, November 1979.
- 19. 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.
20 Letter, FVY 82-21, dated February 25, 1982, E. W. Jackson to R. C.
Haynes, Director of USNRC Region I, " Cycle IX Startup Test Report".
- 21. Loss-of-Coolant Accident Analysis for Vermont Yankee Nuclear Power Station, NEDC-21697, August 1977, as amended.
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