ML20153H632

Yankee Nuclear Power Station,Core 20,Performance Analysis

ML20153H632
Person / Time
Site:	Yankee Rowe
Issue date:	08/31/1988
From:	Carpenito F, Crofton S, Distefano J YANKEE ATOMIC ELECTRIC CO.
To:
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ML20153H623	List: ML20153H619 ML20153H632
References
YAEC-1652, NUDOCS 8809090274
Download: ML20153H632 (84)

Text

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I L '**- Sc1 ear' Power Station Core 20 Performance Analysis 4

t August 1988 Major Contributors: F. L. Carpenito .

S. P. Crofton J. DiStefano  !

N. Fujita R. C. Harvey G. E. Jarka S. M. Mihaiu K. J. Morrissey R. C. Paulson P. B. Perez D. A. Rice K. E. St. John

> M. E. Towell i

i 3825R/4.329 8809090274 080019 PDR ADOCK 05000029 P PNU ,

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Per, pared By: 1 A.21$,,ub Wa//7% ~

T. ^. Schmid [. Cycle'2'O Reload Coordinator (Date)

Approved By: #4W 7/27 kb P'.A.Bergeron,ganager (Date)

Transient Analysis Group Approved By: <"ldd/ 4 i,' Manager 7df R . Caccia / (Da(e )

ctor Phys s Group Approved By: . f. _ .

D 88 S. P'. Schultz, hanagel (Date)

LOCA Group

, Approved By: k/ mum

• 7/E2/E B. C. Slifer, Directg (Date)

Nuclear Engineering Department i

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i Yankee Atomic Electric Company Nuclear Services Division

'571 Worcester Road Framingham, Massachusetts 01701 l

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DISCLAIMER OF RESPONSIBILITY l

~ This document was prepared by Yankee Atomic Electric Company and is completely true and accurate to'the best of our knowledge, information and balief. It is authorized for use specifically by Yankee Atomic Electric Company, and the appropriate subdivisions within the Nuclear Regulatory Commission only.

Nith r.6ard to say unauthorized use whatsoever, Yankee Atomic Electric l- Company, and its officers, directors, agents and employees assume no liability nor make any warranty or representation with respect to the conter.ts of this document or to its accuracy or completeness.

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ABSTRACT This report describes the mechanical, thermal-hydraulic, physics, and scfety analysos necessary for the support of the Core 20 reload cycir for the Yrnkee Nuclear Power Station.

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,t q' TABLE OF CONTENTS i S:ction Page DISCLAIMER OF RESPONSIBILITY..................................... iii ABSTRACT......................................................... iv LIST OF FIGURES.................................................. vii LIST OF TABLES................................................... vili ACKN0WLEDGEMENTS................................................. ix

1.0 INTRODUCTION

..................................................... 1 2.0 OPERATING HISTORY OF CURRENT CYCLE............................... 2 3.0 G ENE RAL DES C R I PT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.0 FUEL MECHANICAL AND THERMAL DESIGN............................... 8 4.1 Mechanical Design.......................................... 8 4.2 Thermal Design............................................. 8 4.3 Operating Experience....................................... 9 5.0 NUCLEAR DESIGN................................................... 16

! 5.1 Physics Characteristics.................................... 16 5.2 Reactor Physics Analytical Computer Codes.................. 17 i 5.3 Changes in Analytical Methods.............................. 18 6.0 TH ERMAL-H YD RAU L IC D E S I GN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 ,

1 7.0 ACCIDENT ANALYSIS................................................ 41 7.1 Introduction............................................... 41 7.1.1 Initial Operating conditions....................... 41 7.1.2 Reactor Trip Setpoints and Instrumentation Delays............................................. 42 7.1.3 Reactivity Coefficients............................ 42 7.2 Control Rod Wi.hdrawal Incident............................ 42 r

J 7.3 Boron Dilution Incident.................................... 43 l 7.3.1 Introduction....................................... 43 7.3.2 Analysis and Results............................... 43

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TAbtE OF CONTENTS (continued)

Stetion Page

+ 7.3.2.1 Boron Dilution During Modes 1 and 2...... 43 7.3.2.2 Boron Dilution During Mode 3............. 44 7.3.2.3 Boron Dilution During Modes 4 and 5...... 45 7.3.2.4 Boron Dilution During Mode 6..........~... 45 7.3.2.5 Failure to Borate Prior to Cooldown...... 47 s

7.3.3 Conclusions........................................ 48 7.4 Control Rod Drop Incident............................... .. 48 7.5 Isolated Loop Startup Incident............................. 49 "7.6 Loss of' Load Incident...................................... 49 7.7 Lorn of Feedwater Flow Itcident............................ 49 7.8 Loss of Coolunt Flow Incident.............................. 50 t 7.9 Control Rod Ejcction Accident.............................. 51 o 7.10 Steam Line Break Accident.................................. 51 7.11 Stean Cenerator Tube Rupture Incident...................... 52 7.12- Fue l lland l in g I nc id en t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 7.13 Other Accidents and Transients............................. 52

[j, '7.14 Transient Analysis Summary................................. 52 e

8.0 STAR!?P Pa0 CRAM.................................................. 66 i

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9.4 LOSS'0F COOLANT ACCIDENT......................................... 69 f i'

9.1 Introduction............................................... 69 .

9.2 Smal? Brogk L0CA........................................... 69 9.3 La r g a B r e c k L0C A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69 94 Break Spectrum Analysis.................................... 70 '

9.3 Burnup Senaltivity Study................................... 70 9.6 Sunaary of Results......................................... 71 9.7 Radiological Consequences of Design Basis LOCA and  ;

Post-LOCA Hydrogen Contro1................................. 71

10.0 REFERENCES

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LIST OF FIGURES Number Title Page

.2-1 Yankee Core 19 Comparison of Measured and Calculated Relative Assembly Average Power Distribution 3 3-1 Yankee Core 19 Loading Pattern 5 3-2 Yankee Core 20 BOC Assembly Average Burnup 6 3-3 Position of Core 19 Assemblies in Core 20 7 j 4-1 Yankes Core 20 Core Locations of Modified Assemblies 12 f 4-2 Yankee Core 20 Lattice Locations of Solid Rods 13 4-3 Yankee Core 20 B0C Fuel Centerline Temperature 14 4-6 Yankee Core 20 EOFPL Fuel Centerline Temperature 15  !

l 5-1 Yankee Core 20 Relative Radial Power Distribution 500 mwd /Mtu All-Rods-Out 23

, 5-2 Yankee Core 20 Relative Radial Power Distribution 8,000 mwd /Mtu All-Rods-Out 24 i 5-3 Yankee Core 20 Relative Radial Power Distribution i 13,000 mwd /Mtu All-Rods-Out 25  !

, 5-4 Yankee Core 20 Relative Radial Power Distribution 500 mwd /Mtu Group C Inserted 26 l j 5-5 Yankee Core 20 Relative Axial Power Versus Group C .

Insertion 27  ;

i 5-6 Core 20 Control Rod Group Identification 28 l i

i $-7 Yankee Core 20 Rod Insertion Limit Versus Power Level 29  ;

5-8 Yankee core 20 Xenon Redistribution Factor 30 t 5-9 Yankee Core 20 Reduced Load Multiplier 31 l 6-1 Reactor Core Safety Limit - All Loops in Operation 40 1

j 9-1 Core 20 Allowable Peak Rod LHGR Versus Cycle Burnup 73

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LIST OF TABLES Number Title Pm 4-1 Core 20 Nominal Mechanical Design Parameters 10 5-1 Sununary of Core 19 and Core 20 Nuclear Characteristics 19 5-2 Core 19 and Core 20 Shutdown Requirements, %Ap 21 5-3 Comparison of Core 19 and Core 20 Control Rod Worths 22 6-1 Thermal-Hydraulic Data Sheet for Yankee Core 20 During 4-Loop Operation 34 Thermal-Hydraulic Data Sheet for Comparison of Core 19 6-2 and Core 20 Design Characteristics During 4-Loop Operation 36 .

I i 6-3 Sununary of Hot Spot and Hot Channel Factors for Yankee Core 20 Versus Design 38

, 6-4 Nominal Hot Channel DNBR, F6H and Fg as Functions of '

Group C Position for Core 20 39 i

7-1 Four-Loop Initial Operating Conditions 53 7-2 Reactor Trip Setpoints and Instrumentation Delays 54 )

l 7-3 Modes 1 and 2 Boron Dilution 55

7-4 Mode 3 Boron Dilution 56 l

7-5 Modes 4 and 5 Boron Dilution 57 7-6 Mode 6 Boron Dilution - All Loops Isolated, Minimum Water 58 Level

+ 7-7 Mode 6 Boron Dilution - Water Volume Including Shield Tank $9 ,

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7-8 Control Rod Drop Incident Parameters 60 7-9 Loss of Coolant Flow Parameters 61 7-10 HZP Rod Ejection Accident Parameters 62 i

7-11 HTP Rod Ejection Accident Parameters 63 7-12 Yankee Core 20 Safety Analysis Summary of Results 64 l 8-1 Yankee Core 20 Startup Test Acceptance Criteria 68 9-1 Core 20 Burnup Sensitivity Study Results 72 T

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ACKNOWLEDGEMENTS The authors would like to express their gratitude to the Word Processing Center for their fine typing of this report.

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k 1.0 INTRODUCTI_0N This report describes the basis for operation of the Yankee Nuclear Power Station (YNPS) through the forthcoming Core 20 (Reload Cycle). This report is also being submitted to the Nuclear Regulatory Commission for information only. This reload egele will contain forty (40) fresh fuel t assemblies and thirty-six (36) recycled assemblies from Core 19 (Current Cycle) fabricated by Combustion E'ngineering (C-E). The introduction of the nsw fuel is necessary in order to maintain sufficient reactivlty for continued operation at full rated power.

The following sections of this report describe the mechanical, thermal-hydraulic, physics, and safety analysis aspects of the Reload Cycle. i The approach to licensing the reload core presented in this report is ,

similar to that used since Core 11. l i

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2.0 OPERATING HISTORY OF CURRENT CYCLE .o The current operating cycle is Core 19. Core 19 started producing power on July 3, 1987, and is scheduled to shut down on November 12, 1988. .

During the period of operation to date, the core has operated normally in an assentially all-rods-out condition. Both plant measured data and Reactor t Physics calculations have confirmed that the gross power distribution changes only slightly as a function of time during core life. The middle of life  ;

power distribution shown in Figure 2-1 is, therefore, representative of the [

cntire cycle. No abnormalities which would adversely affect the core power distribution or reactivity have been detected thus far during the Current Cycle operation. ,  !

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FIGURE 2-1 YANKEE CORE 19 COMPARISON OF MEASURED AND CALCULATED RELATIVE ASSEMBLY AVERAGE POWER DISTRIBUTION 0.628 0.804 0.820 0.639 0.635 0.8 12 0.825 0.641

-1. 0 --1.0 -0.6 -0.2 1.07 3 1.025 1.M7 0.778 0.760 1.115 1.080 1.028 1.M6 0.775 0.765 1.127

-0.6 -0.2 0.1 0.4

-0.6 -1.0 1.UO 1.071 1.U9 0.775 0.7 64 1.159 1.043 1.17 1 1.168 1.068 1.75 0.772 0.760 1.166 1.043 1.168 0.2 0.3 0.4 0.4

-0.6 -0.6 0.0 0.3 1.089 1.042 1.M0 0.641 1.068 1.080 1.18 8 1.152 0.635 1.132 1.038 1.136 0.639 1.078 1.181 1.M6 1.085 0.638 1.138 1.072 0.4 0.5 0.4 0.4 0.4

-0. 6 -0.6 -0.4 0.3 0.6 1.219 1.113 1.073 0.828 1.159 1,218 1.111 1.119 0.817 1.078 0.8 27 1.111 1.210 1.105 1.070 1.085 1.158 1.211 1.104 0.822 0.7 0.7 0.2 0.6 0.7 0.6 0.7

-0. 6 -0.6 0.1 1.184 1.166 1.059 0.816 1.155 1.117 1.1M 0.810 1.068 1.14 4 1.0 57 0.8M 1.110 1.107 1.U6 1.160 0.815 1.074 1.135 1.M 6 0.7 0.5 0.2 0.2 0.8 0.8 0.7 0.6

-0.6 -0.6 1.219 1.079 1.071 1.14 2 9660 '

0.610 1.12 8 1.036 1.082 1.2 14 1.213 1.075 1.069 1.M0 0.r59 1.032 1.07 8 1.208 0.630 1.12 4 0.2 0.2 0(

0.4 0.4 0.5 0.5 0.3

-3.2 0.4 1.109 1.040 1.U0 0.796 0 530 1.M t 1.062 1.167 1.107 1.041 1.00 0.794 0.755 1.16 3 1.062 1162 0.4 0.2 0.0 0.0 0.2

-3.2 -1.4 0.0 _

1.090 1.078 1.130 0.789 WEASUREMD4T (6735 MWD /WN 0.737 1.12 0 1.079 1.132 0.790 CALCOLATION (6735 M #D/MTU) 0.759 1.136 1.086

-1.4 04 -0.1 -0.1 -0.1 PERCENT DIFFTAU4CE

-2.9 0.6M 0.820 0.825 0.653 0.6M 0.8U 0.826 0.654

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l 3.0 GENERAL DESCRIPTION I t )

Figure 3-1 is a schematic of the YNPS core showing the Current Cycle loading pattern. In this scheme, the inner region consists of 40 recycled assemblies and the outer region consists of 36 fresh assemblies.- The Reload j Cycle core will utilize 40 fresh assemblies and 36 recycled assemblies. The-fresh assemblies fabricate / by C-E have an initial enrichment of 3.9 w/o U-235 for 36 of the assemblies and 3.7 w/o U-235 for four of the assemblies. The racycled assemblies also fabricated by C-E all had an initial enrichment of 3.8 w/o U-235. The Reload Cycle core average exposure for Basinning-Of-Cycle (BOC) is 6608 mwd /Mtu compared to 7315 mwd /Mtu for the Current Cycle. The Reload Cycle full power lifetime is estimated to be i 14,100 mwd /Mtu compared to 13,000 mwd /Mtu for the Current Cycle. [

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Figure 3-2 shows the Reload Cycle loading pattern giving the location ,

of both fresh and recycled fuel, as well as the recycled fuel assembly exposure. Figure 3-3 shows the position that each recycled assembly occupied

. r in the Current Cycle.

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FIGURE 3-1 YANKEE CORE 19 LOADING PATTERN XXX X XX XX XX XX .

XX

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XX X X X X XX XX XX XX XX XX

-FRESH FUEL

-RECYCLED FUEL 3825R/4.329

FIGURE 3-2 YANKEE CORE 20 BOC ASSEMBLY AVERAGE BURNUP

0. O. O. O.

12420. O. O. O.

0. O.

O. O. 12883. 12537. 18281. U936. O. O.

10256. 12781. O. O.

O. O. U916. 10046. 10 318. 12986.

+ O.

12910, 17825. T16#. 10528. 12351. 1219 8.

0. O. 18460.

12737. 18370. O. O.

0. 12619. 12394. 10071. 17756. 17697.

12729, 10431. U842. O. O.

O. O. 12906, 10237. 10310.

12H2. 12795. O. O.

O. O. 17916, 18408.

O. 12459. O. O.

0. O.

O. O. O. O.

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FIGURE 3-3 POSmON OF CORE 19 ASSEMBUES IN CORE 20 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 11 3 59 60 57 20 21 22 23 24 25 26 27 28 19 9 28 58 75 76 48 32 33 34 35 36 37 38 29 30 31 65 38 71 50 73 5 10 40 42 43 44 45 46 47 48 39 41 07 72 4 27 6 39 12 52 53 54 55 56 57 58 49 50 51 29 1 2 19 49 66 59 60 61 62 63 64 65 66 20 17 18 74 67 66 69 70 71 72 66 73 74 75 76 ASSEMBD' NUWBER CORE 19 LOCATION 3825R/4.329

4 4.0 FUEL MECHANICAL AND THERMAL DESIGN 4.1 Mechanical Design l

Forty fresh assemblies manufactured by C-E will be inserted into the rsector for the Reload Cycle. This is the third batch of C-E fuel (Batch C)-

to be used at the YNPS. The mechanical design parameters are described in '

dstall in Reference 4.1. Table 4-1 lists the design parameters for the recycled (Batch B) and fresh (Batch C) fuel assemblies.

Like the first two batches of C-E fuel, Batch C continues to use solid Zircaloy rods and special guide bars in some preselected peripheral assembly locations. The special guide bars with a fixed spacing device are used to provide extra rigidity for the solid Zircaloy rods. The locations of the essemblies with these design features are chown in Figure 4-1. Figure 4-2 shows the assembly lattice locations for the salid Zircaloy rods and special guide bars.

Further descriptions of the recycled fuel rod mechanical design and analyses are provided in Reference 4.2. Mechanical and chemical compatibility of the fuel assemblies with the in-service reactor environment is also addressed in References 4.1 and 4.2.

4.2 Thermal. Design The fuel thermal offects calculations were performed using t'.ne GAPEXX digital computer code (Reference 4.3). The Reload Cycle calculation mathodology is unchanged from previous reload analyses.

The GAPEXX code calculates pellet-to-clad gap conductance f rom a combination of theoretical and empirical models which predict fuel and cladding thermal expansion, fission gas release, pellet swelling, pellet d:nsification, pellet cracking, and fuel and cladding thermal conductivity.

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The thermal effects analysis encompassed a study of fuel rod tsmperature response as a function of the detailed cycle burnup and power.

Tha fuel rod types and power histories examined in detail include:

High-Power FreJh Pin: fuel rod with the maximum average power for the fresh (reload) fuel, High-Power Exposed Pin: fuel rod with the maximum average power for the exposed (recycle) fuel.

Figure 4-3 illustrates the effect of Linear Heat Generation Rate (LHCR) on calculated temperatures at BOC. Figure 4-4 illustrates the effect of LHGR on calculated temperatures for End-0f-Full-Power-Life (EOFPL) conditions.

Th2se resultant temperatures are similar to those which have been reported in previous reload analyses (References 4.4 and 4.5).

These results demonstrate that the BOC conditions yield the maximum prsdicted fuel temperatures. This is due to a prediction of the maximum diametral gap at the time when the predicted fuel thermal conductivity is the lowest. The calculated internal fuel rod pressures are less than operating coolant system pressure throughout Reload Cycle operation.

4.3 Operating _ Experience The batch average burnup of the C-E fuel to be discharged during this refueling outage will be approximately 31.600 mwd /Mtu with a peak assembly burnup of approximately 35,600 mwd /Mtu. The batch of Exxon fuel discharged from the reactor during the last refueling outage had a batch average burnup of 30.985 mwd /Mtu with a peak assembly burnup of 34,394 mwd /Mtu. The first two batches of C-E nuclear fuel have performed as expected.

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- TABLE 4-1 CdRE20NOMINALMECHANICALDESIGNPARAMETERS Recycled Fuel Fresh Fuel ,

(36 assemblies) (40 assemblies)

Fual Pellets Fuel Material (sintered UO2 UO2 pellets)

Initial Enrichment. 3.8 3.9/3.7*

w/o U-235 .

Pellet Density, 94.75 94.75

% theoretical-Pellet Diameter, inches 0.3105 0.3105 Fu21 Rod y.

Active Length, inches 91.0 91.0 Overall Rod Length, 95.40 95.40 inches Upper Plenum Length, 1.54 1.54 inches Fuel Rod Pitch, inches -

0.472 0.472 '

Diametral Cap (cold). 0.0065 0.0065 inches Fill Gas Helium Helium Fill Gas Pressure, psig' 250.0 250.0 Cicdding Material 2r-4 2r-4 Outside Diameter, inches 0.365 0.365 Thickness, inches 0.024 0.024 Inside T1ameter. 0.317 0.317 inches Guide Bars Material 2r-4 2r-4 Number per Assembly 8 8 Length, inches 96.52 96.52 0 36 assemblies have an initial enrichment of 3.9 w/o U-235 and 4 assemblies are at 3.7 w/o U-235.

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TABLE 4-1 CORE 20 NOMINAL FMCHANICAL DESIGN PARAMETERS (Continued)

Recycled Fuel Fresh Fuel (36 assemblies) (40 assemblies)

Fuel Assembly Number of Assemblies 36 40 Fuel Rod Array 16x16 16x16 Fuel Rods per Assembly Type A 231 231 Type B 230 230 Fuel Rod Axial Clearance, 1.217 1.217 inches Outside Dimensions Assembly Cross 7.614x7.156 7.614x7.156 Section, inches Overall Length, 111.75'/ 111.787 inches Spacer Grids Material Zr-4 Zr-4 Number per Assembly 6 6 Weight of Contained Uranium Type A. Kg U 231.3 231.3 Type B. Kg U 230.3 230.3 i

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RGURE 4-1 YANKEE CORE 20 CORE LOCATIONS OF MODlRED ASSEMBUES l

l 1 2 3 4 A-1 M

5 6 7 6 9 to A-1 A-1 M M 14 15 16 17 18 11 12 13 A-1 0-1 M (12) 23 24 25 26 27 28 19 20 21 22 A-1 5 -

33 34 35 36 37 38 29 30 31 32 B-2 B-1 ,A-1 (6)

(12) M 43 44 45 46 47 48 39 40 41 42 A-1 B-1 (7) (12) 53 54 55 56 57 58 49 50 51 52 A-1 A-1 0-1 A-2 B-1 M M (12) M (12) 62 63 64 65 66 59 60 61 A-2 B-1 (7)

(t2) 67 63 69 70 71 72 W-1 A-2 (t2) (T) 73 74 75 76 ASSEMBLY NUMBER B-1 A-2 B-3 A-2 ASSEMBLY TYPE (12) M (6) M f 0F REPLACED RODS 3825R/4.329

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FIGURE 4-2 YANKEE CORE 20 LATTICE LOCABONS OF SOUD RODS THII R$508.7 TYPE A 1 TPr*II R$5Cre.T TTPC A 2 x x '.: x x x

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5.0 N1l CLEAR DESIGN l

5.1 Physics Characteristics e Table 5-1 prasents a summary of pertinent physics data for the Reload Cycle. The data is similar to that of the Current Cycle with only minor ,

variations due to the different core loading pattern. The Reload Cycle radial power distributions at hot-full-power, all-rods-out condition, are presented for the cycle average exposures of 500; 8.000; and 15.000 FNd/Mtu. ..

respectively. In Figures 5-1 through 5-3. the maximum unrodded radial pin '

power peak. Fxy is 1.590 at 500 FNd/Mtu. 1.529 at 8.000 FNd/Mtu, and 1.437 at 15.000 FNd/Mtu. The radial power distribution at 500 mwd /Mtu with control rod Group C fully inserted is given in Figure 5-4 In this case, the peak F is 1.746. Figure 5-3 shows the relative axial power distr?bution as control rod Group C is inserted for the assembly containing the I,tak F I (Location 53 (E-7)).

Table 5-1 presents various calculated reactivity coetficients for both j .e

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BOC and E0FPL reactor condi; ions. The moderator temperature coefficient at c E0C for the Reload Cycle is less negative than the Current Cycle value primarily due to the larger value of critical boron concentration. The change in total control rod worth is due mostly to the change in the power distribution and critical boron concentration. The effective delayed neutron fractions for the Reload Cycle are similar to those for the Current Cycle. ,

All of the physics values used in the accident analysis are chosen in a conservative manner for each analysis.

The Reload Cycle like the past several cycles, uses control rod ,

Group C as its controlling rod group. Table 5-2 presents a sun: mary of calculated control rod group worths for the Reload Cycle and comparable Currect Cycle data. The reactivity allowances are lisud for the Reload Cycle # ,

and the Current Cycle with the resulting excess shutdown margin also I tabulated. The control rod group configuration remains unchanged from g previous cores and is presented in Figure 5-6. Ths calculated reactivity 3825R/4.329 ,,

worth for each control group at hot-full-power conditions is given in ..

Table 5-3 fcr both BOC and EOFPL reactor conditions. The power dependent control rod insertion limit curve for the Reload Cycle 's unchanged from that used during the Current Cycle operation and is giver, in F,gure 5-7  :

Xenon redistribution ef fects are accounted for by ti e use of two y Y factors: the Xenon Redistribution Factor (XRF) and the Red need Load _

Multiplier (RLM). The XFf accounts for changes in axial peaking due tn contro rod Group C motion in the full power operating band t 80 to 90 inches withdrawn). This multiplicative factor is defined as the ratis of the maximum value of F2in the analytically derived top-peaked xenon-inducec axial shape to the value of F of the nominal operating axial chape. The min. mum value 7

This is consistent with the methodology u'ed to I of this factor is unity.

derive the LHGR limits which were generated based on the worst top-pssked axial shapes. The top-peaked axial shapes bound both nominal and bottom-peaked shapes in terms of LHCR limits. The RLM designates a powei level at which the plant must remain at or below f or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if control red Group C is inserted below 80 inches withdrawn. This time period allows for dampening of the xenon tr..asient aad, therefore, the power redistribution.

• The factors to be used is a function of the Reload Cycle average burnup are presented in Figures 5-8 and 5-9. These factors are unchanged from the previous cycle values as they bound those values calculated for the Reload Cycle.

5.2 Reactor Physics Analytical Computer Codes YNPS core depletion calculations since Core 13 have been made using a PDQ/ HARMONY (References 5.1, 5.2) model with each fuel rod in an assembly 'S represented explic;tly. Thus, when the Current Cycle assemblies were reshuffled into the Reload Cycle, the recycled fuel had the burnup of each fuel rod represented explicitly. Few group cross sections, for use in PDQ.

were obtained with the LEOPARD (Reference 5.3) program. The full core PDQ model of the YNPS core, which was introduced in Core 18, was again used for ,

the Reload Cycle.

The S1ML' LATE (Ref erence 5.4) pr1 gram has been used for the calculation of core average reactivity parameters such as moderator temperature coefficients, fuel teeperature coeff.cients, boron worth and critical baron .

3825R/4.329 .

concentrations. Axial power distributions and xenon-induced redistribution effects are also calculated with the SIMULATE program. Comparisons between plant measurements and the calculated SIMULATE reactivity parameters and axial power distributions since Core 14 have been in excellent agreement.

For analysis of flux measurements. YNPS uses the INCORE (Reference 5.5) program. INCORE requires theoretical data (PDQ) in the form of radial asserrbly and f uel rod powers, plus the f ast and thermal f! ttxes in the instrumentation thimbles. The edits in the PDQ/ HARMONY are arranged to give these data as a function of lifetime for each assembly location in the core.

5.3 Changes in Analytical Methoda There have not been any analytical methodology changes initiated for the Reload Cycle analysis.

3S25R/4.329

TABLE 5-1

SUMMARY

OF CORE 19 AND CORE 20 NUCLFAR CHARACTERISTICS Core 19 Core 20 Total Control Rod Worth, %Sp Hot Full Power BOC 11.13 11.61 Hot Full Power EOFPL 11.66 12.21 ,

Dissolved 3oron Dissolved Boron Content for Criticality BOC, All-Rods-Out, ppm 680F, No Xe, Peak Sm 1908 2107 Hot Zero Power, No Xe, Peak Sm 2091 1851 J Hot Full Power. Eq. Xe and Sm 1490 1660 Dissolved Boron Content for Refueling, ppm 1901 20S1 BOC, All Rods In, K = .93 ,

Inverse Boron Worth, ppm /%ap ,

680F, BOC No Xe, Peak Sm 103 106 BOC, Hot Zero Power, No Xe, Peak Sm 133 137 e BOC, Hot Full Power. Eq. Xe 134 137

• Reactivity Coefficients (All-Rods-Out)

Moderator Temperature Coefficients, ap/0F Hot Full Power, BOC -1.15x10-4 -0.80x10-4 '

Hot Full Power. EOFPL, 0 ppm -3.13x10-4 -3.08x10-4

• Fuel Temperature Coefficients, ap/0F Hot Full Power, BOC -1.46x10-5 -1.56x10-5 Hot Full Power. EOFPL, 3 ppm -1.47x10-5 -1.75x10-5 Modarator */oid Coefficients, ap/1 void Hot Full Power, BOC -0. 88x10-3 -0.61x10-3 Hot Full Power, EOFPL, 0 ppm -2,48x10-3 -2.35x10-3 m

<a 3825R/4.329

TABLE 5 1 SLSNARY OF CORE 19 AND CORE 20 NUCLEAR CHARACTERISTICS (Continued)

Core 19 Core 20 Moderator Pressure Coefficients, op/ psi Hot Full Power, BOC 1.11x10-6 J.86x10-6 Hot Full Power, EOFPL, 0 ppm 3.27x10-6 3.30x10-6 Effective Delayed Neutron Fraction BOC, HZP .006383 .006543 Effective Delayed Neutron Traction. EOFPL, HZP .005437 .005411 Prompt Neutron Lifetime, psec, BOC, HZP 20.43 19.90 Prompt Neutron Lifetime, psec. EOFPL, HZP 23.32 23.15 20-3825R/4.329

TABLE 5-2 CORE 19 AND CORE 20 SHUTDOWN REQUIREMENTS. 13p (HTP)

Core 19 Core 20 BOC EOFPL f00 EOFPL ,

Rod Worths

1. Total Control Rod 'Jorth 11.13 11.66 11.61 12.21

2. Worth of Stuck Rod 3.03 3.21 3.31 3.44

3. Total Worth less Stuck Rod (1-2) 8.10 8.45 8.30 8.77

4. Total Worth less 7.5% Uncertainty 7.49 7.82 7.68 8.11 ,

(No. 3 x 0.925)

5. Total Allowances

• 6.60 7.17 6.48 7.23

6. Excess Shutdown Margin (4-5) 0.89 0.65 1.20 .SS Allowances

• L 5.1 Fuel Temperature Variation .73 .76 .80 .90 5.2 Moderator Temperature Variation .27 .72 .11 .67 5.3 Moderator Voids .03 .07 .02 .07 5.4 operational Maneuvering Band .07 .12 .05 .09 5.5 Shutdown Margin 5.50 5.50 5,50 5.50 Total Allowances 6.60 7.17 6.48 7.23 3825R/4.329

TABLE 5-3 COMPARISON OF CORE 19 AND CORE 20 CONTROL ROD WORTHS (HFP)

Core 19 Core 20 BOC EOFPL BOC EOFPL Mo -. M e U p. Up Gr:up C 1.78 1.88 1.77 1.86 Grcup A 1.37 1.46 1.23 1.49 Gr:up B 2.31 2.35 2.40 2.49 Group D 5.67 5.97 6.21 6.37 Total 11.13 11.66 11.61 12.21 Stuck Rod 3.03 3.21 3.31 3.44 Total Less Stuck Rod 8.10 8.45 8.30 8.77 3825R/4.329

FIGURE 5-1 YANKEE CORE 20 RELATIVE RADIAL POWER DISTRIBUTION 500 MWD /MTU, ALL-RODS-OUT 0.594 0.788 0.835 0.6M 0.743 1.161 1.062 1.267 1.U6 0.739 l

i 1.088 1.045 1.050 1.208 0.749 0.735 1.192 1.167 1.166 1.U2 1.101 1.U1 1.167 1.17 5 0.598 0.611 1.160 1.048 a .

1.0 21 1.030 1.121 1.12 3 1.095 0.810 0.816 1.263 1.035 1.10 8 1.166 1.028 1.021 1.114 1.046 1.283 0.824 0.782 1.036 1.142 i

1.141 1.112 , t 'e.1 1.114 1.0 57 1.U2 0.619 0.592 1.16 3 1.165 1.048 1.046 1.M6 1.Us 1.199 0.741 0.74 3 1.19 8 i

l 0.737 1.167 1.268 1.090 1.16 3 0.747 E 0.615 0.810 0.791 0.596 3825R/4.329 i

i i

RGURE 5-2 YANKEE CORE 20 RELATIVE RADIAL POWER DISTRIBUTION -

8000 MWD /MTU, ALL-RODS-0UT 4

a k

0.627 0.609 0.641 0.640 ,

I 1.051 1.218 1.136 0.754 .t 0.755 1.134 '

a ,

1.062 1.048 1.165 0.760 0.748 1.15 6 1.14 6 1.092

. I L

1.17 5 1.137 1.137 0.630 1.i73 1.192 1.13 3

' O.636 1.12 5 1 052 t 1.136 1.119 1.073 0.923 .

1.054 1.14 0 1.076 1.085 0.822 1.219 l- r i l

1.184 1.082 1.07 5 1.1H 1.057 1.224 0.025 (

0.802 1.025 1.14 4 l l

i 1.19 1052 1.12 5 0.6 37 [

1.146 1.142 0.621 1.131 1.14 0 [ 1.16 3 I i - ._

1 s , 1.14 4 1.154 0.748 0.747 1.155 1.046 1.059 j

I t 1.2 14 1.07's 1.12 5 0.753 0.741 1.12 3  ;

t 0.631 0 .8 11 0.803 0.625 ,

J 1

. J I

t f 387.5R/4.329  :

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y 1 t

j 1

RGURE 5-3 YANKEE CORE 20

RELATIVE RADIAL POWER DISTRIBUTION 15,000 MWD /MTU, ALL-RODS-0UT [

1 f

f 0.671 1 0.643 0.067 0.687 ,

I 4.760 1.12 0 1.046 1.191 1.124 0.792 l t

P 0.775 1.133 1.12 5 1.078 1.062 1.047 1.147 0.796 L

l 0.672 1.107 1.047 1.15 3 1.17 2 1.124 1.15 8 1.118 1.125 0.661 r

!. 0.630 1.187 1.051 1.12 5 1.075 1.085 1.116 1.105 1.069 0.656

i

}

0.*324 1.015 1.12 6 1.162 1.080 1.075 1.135 1.057 1.195 0.852 f

. 0.655 1.112 1.116 1.12 5 1.12 9 1.14 2 1.099 1.049 1.110 0.676 [

f 0.763 1.12 8 1 040 1.055 1.125 1.122 1.13 3 0.775 I

i d

f l 0.759 1.10 3 1.1 A1 1.062 1.109 0.779 1

i -

t 0.660 0.827 0.826 0.666  !

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l l

FIGURE 5-4 YANKEE CORE 20 RELATIVE RADIAL POWER DISTRIBUTION 500 MWD /MTU, GROUP C INSERTED <

r 0 601 0.724 0.758 0.609 l

0 431 L193 0 829 0 968 1t76 0 412 ,

i 1

,.450 0 92$ 0 461 1.10 3 1347 0 429 O $ 17 134$

r 1.283 1199 1288 1235 1196 O c00 0 613 1178 1110 1293 i

l 1212 1189 1195 1 213 0.947 0 d49 0 '40 0 756 0 912 0 462 1196 11!4 1210 0 866 0 986 0 749 O 131 0 #18 0 922 1 279 t 1 I 1251 1266 1214 1244 t222 tii4 1177 0 tt) c 607 L209

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0 83?

i 0 424 t194 0 993 0.458 t195 0.430 3 i

0.753 0.736 0.604 [

0 620 i

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RGURE 5-6 YANKEE CORE 20 CONTROL ROD GROUP IDENTIRCATION l

3 4 1 _2 D

5 6 7 8 9 ]O 4

~

D C t 16 17 18 12 13 '14 15 11 ,

l .

l B B D  :

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' 22 '3 24 25 19 20 21 '

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, 61 62 63

! 59 60 'J i C D 72 68 69 70 71 ,

67 a ,

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D 74 75 76 73 ASSDGLY POSITION f l

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3825R/4.329 I

FIGURE S-7 YANKEE CORE 20 ROD INSERTION LIMIT VERSUS POWER LEVEL 200 -

90-j 60 -

E a

{ 70 -

5 i5 .

a e0 -

d 5

3 50 -

d u .

O 40 -

, 2

'f 10 f

i . . i i 0- i . i 20 30 40 50 80 70 60 SO 0 10 GROUP C POSITION (INCHES WITHDRRWN) 3825R/4.329

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3825R/4.329

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6.0 THERMAL-HYDRAULIC DESIGN The thermal-hydraulic evaluation of the Reload Cycle has been performed utilising the same methodology as all of the reload analyses since Core 11.

The thermal-hydraulic analysis of the Reload Cycle was performed by cdjusting code input to reflect the Reload Cycle power distributions and thormal-hydraulic characteristics. Table 6-1 contains the pertinent nominal cnd design thermal-hydraulic parameters for the Reload Cycle and Table 6-2 contains a comparison of the design information for the Reload and Current Cycles. Table 6-3 summarises both the predicted hot spot and hot channel fcetors for the Reload Cycle and the design hot spot and hot channel factors.

Tcble 6-4 indicates the behavior of the hot channel DNBR, F and Fq at AH full power conditions versus Rod Group C position for the Reload Cycle.

Prodicted hot channel factors are based on power distributions at the limiting exposure for power peaking when Rod Group C is inserted up to 25%, even though rod restrictions do not permit operation at full power for this rod group position.

The information provided in Tables 6-1 through 6-4 shows that the Reload Cycle has significant margin to DNB, coolant quality and fuel conterline melt limits. The design DNBR for the Reload Cycle is slightly lower than the previous analysis DNER, 3.02 versus 3.04 at full power 4-loop opsration. This difference is due primarily to the insertion of 206 solid Zircaloy rods and special guide bars in the Reload Core versus only 128 solid rods in the previous analysis. The resulting increase in core average linear host generation rate leads to a slightly reduced va'ue of DNER.

Safety limit curves for the Reload Cycle are provided in Figure 6-1.

Th3se curves were developed in the Core 15 analysis to conservatively bound future reload cores. These curves continue to be bounding for the Reload Cycle operation.

The effect of rod bow has been considered for the Reload Cycle op2 rat ion. As required in Reference 6.1, a 34% DNER credit is needed to offset the very conservatively applied full-elosure rod bow penalty. Generic credit of 13.2% DNER margin, based on conservative thermal-hydraulic 3825R/4.329

cnalytical methods, was accepted in Reference 6.1. The most limiting cnticipated transient is the 2-out-of-4 pump loss of flow. Based on design c nditions, this event results in a minimum DNBR of 2.01. Thus, sufficient margin to a DNBR of 1.3 exists for this limiting event, which can be applied to the remaining 20.8% margin requirement for rod bow.

A rod bowing evaluation was performed which demonstrates that the full-closure bowing penalty remains conservative. Reference 6.2 indicates that rod bow resulting in less than 50% channel closure has no adverse impact en the predleted DNBR. The maximum predicted channel closure for the fresh full in the Reload Cycle remains less than 50%. For the recycled fuel, the maximum closure is approximately 63%. requiring a penalty of 8.8% DNBR margin. This value conservatively assumes that the bowing penalty is a linear function of channel closure, as indicated in Reference 6.1 with a 0% penalty et 50% closure and a 341 penalty at full closure. Therefore, application of ,

th2 34% full closure penalty to Reload Cycle operation is very conservative.

a b

l 4

i t

3S25R/4.329

TABLE 6-1 THERMAL-HYDRAULIC DATA SHEET FOR YANKEE CORE 20 DURING 4-L90P OPERATION General Cha rac teris tics Nominal Design T:tal Core Power, 600 618 MWt Frcetion of Heat. .973 .973 Generated in Fuel Main Coolant 2,000 1,925 Pressure, psig Main Coolant Inlet 520 524 Temperature. OF R::ctor Vessel 563 568 Outlet Temperature.

CF Average Core Outlet 567 572 Temperature. OF ,

Av: rage Core 58.5 60.3 ,

Enthalpy Rise, Btu /lb Tstal Coolant Flow 38.3x106 38.3x106 Rate, Ib/hr H:st Trsnsfer Flow 35.0x106 35.0x106 Rate, Ib/hr Av: rage Mass 2.296x106 2.296x106 Velocity, ib/hr-ft 2 Av; rage Coolant 13.5 13.6 Velocity in Core, ft/see C:re Pressure Drop, 13.2 13.2 psi R::ctor Vessel 31.3 31.4 Pressure Drop, psi Av rage Rod Heat 158,845 163.610 T1ux, Btu /hr-ft2 Average Film 5.624 5.645 Coefficient.

Btu /hr-ft2 _oy Avetage Film 28.2 29.0 Temperature Difference. OF Average Linear 4.45 4.58 Rod Power, kW/ft Av: rage Specific 34.6 35.7 Power, kW/kgU Power Density. 90.1 92.8 kW/ liter Hydraulic Diameter, in 0.412 0.412 Assembly Heat 165.1 165.1 Transfer Area, ft2 3825R/4.529

1 TABLE 6-1 (Continued)

Hot Channel and Hot Spot Parameters Nominal Design Maximum Heat, Flux. 390,123 451,564 Btu /hr-ft2 Maximum Linear Rod Power, 10.92 12.6 -

kW/ft Maximum Clad Surface 632 637 Temperature, OF Maximum Centerline Pellet 2,916 3,290 Temperature. OF Hat Channel Outlet 588 606 Temperature. OF Minimum W-3 DNBR 4.16 3.02 3S25R/4.329

l TABLE {-2 THERMAL-HYDRAULIC DATA SHEET FOR COMPARISON OF CORE 19 AND CORE 20 DESIGN CHARACTERISTICS DURING 4-LOOP OPERATION General Characteristics Core 20 Core 19 ,

Total Core Power. 618 618 MWt Frcetion of Heat 0.973 0.973 Generated in Fuel Main Coolant 1.925 1.925 Pressure, psig Main Coolant Inlet 524 524 Temperature. OF R:cetor vessel 568 568 Outlet Temperature, cF Av2 rage Core Outlet 572 572 Temperature. OF Avsrage Core 60.3 60.2 Enthalpy Rise.

Btu /lb Total Coolant Flow 38.3x106 38.3x106 Rate lb/hr H2st Transfer Flow 35.0x106 35,oxio6 Rate. Ib/hr Avorage Mass 2.296x106 2.293x106 Velocity 1b/hr-ft2 Avsrage Coolant 13.6 13.6 Velocity in Core, ft/see core Pressure Drop, 13.2 13.1 psi R30ctor Vessel 31.4 31.3 Fressure Drop. psi Av3 rage Rod Heat 163.610 162.813 Flux. Btu /hr-ft2 Assembly Heat 165.1 165.8 Transfer Area, ft2 Average Film 5.645 5.692 Coefficient.

Stu/hr-ft 2.oF Average Film 29.0 28.6 Toeperature Dif ference. OF Averase Linear 4.58 4.56 Rod Power. kW/f t Avorage Specific 35.7 35.5 Power, kW/kgU Pswer Density. 92.8 92.8 kW/ liter Nominal Channel Hydraulic 0.412 0.412 Diameter, in 3825R/4.329

TABLE 6-2 (Io'ntinued)

Not Channel and Hot Spot Parameters go).e 20 Core 19 Maximum Heat Flux, 451.564 449.364  :

Btu /hr-ft2 ,

Maxinum Linear Rod Power, 12.6 12.6  !

kW/ft Maximum Clad Surface 637 637 Temperature. OF  !

Maximum Centerline Pellet 3.290 3.281 Temperature. OF '

l H t Channel Outlet 606 606 Temperature. OF Minimum 14-3 DNBR 3.02 3.04 I

P l

i i

i 3S25R/4.329

TABLE 6-3

.SLHMARY OF HOT SPOT AND HOT CHANNEL FACTORS FOR YANKEE CORE 20 VERSUS DESIGN Nominal Design Heat T1ux Factors Nuclear Heat Flux Factor 2.27 2.59 Fuel Densification Factor 1.03 1.03 Engineering Heat Flux Factor 1.04 1.04 Total Heat Flux Factor 2.46 2./6 Inthalpy Rise factors Statistical F.thalpy Rise Factor 1.08 1.08 Lower Plenum Tsetor 1.05 1.05 Nuclear Enthalpy Rise Factor 1.59 1.80 l

1 3S25R/4.329

TAbiE o-4 NOMINAL HOT CHANNEL DNBR, FoH, AND Fq AS FUNCTIONS OF GROUP C POSITION FOR CORE 20 i

Group C Fag F Posit.lon *(inches inserted) _ Q DNBR 0 1.517 2.27 4.39 7.5 1.512 2.29 4.34 15.0 1.505 2.37 4.29 22.1 1.499 2.46 4.16 CPower Dependent Insertion Limit. Restrict Group C Insertion to 10 inches at Full Power.

3825R/4.329

FIGURE 6-1 RE ACTOR CORE S ATETY LIMIT - ALL LOOPS IN OPERATION 660 .-

~ =.= " '

640 w .

. 620 :. ',' .

1 w 1 y ,

0 KAIN CCCUJ!

h c-

[

600 b' .

~__

Si s! L*.

P F.15 5 JF.E H q x_ _-

-~_'

u. ._

2600 psia 6 '- .

a '( _. ._

'E 580 - s

-+1 v

o .

2400 psia c ___-~

--n -n 3 . T -.' ' ._

' 'w- s._

E  %

. . 2200 rsts

", 560 '.__, . _ _ _

N.-,~

~

o

=-.h_ ,.

~--' 7 -

~-

^ -

'- 2000 ;s!a o N u ._ . _ .__ _ w -

e 540 -

x. ,

~_ .___'x E l - - -

. _ _ , '1 . 1800 psia

_? __

520 -

___ ~_.__. _

+;

- - - - w--

^ 1600 psia

__._~-

_ _ ____.=g --s - - 3 -

s- ,%

4 500

, 70 80 90 100 110 120 130 Indicated Reactor Power, Percent 3825R/4.329

' 7. 0 nCCIDENT ANALYSIS 7.1 Introduction The safety analysis of the Reload Cycle is presented in this section.

Ecch transient in the following subsections is compared with the most recent.

reference analysis. The reference analysis is presented in the Final Safety-Analysis Report (FSAR, Reference 7.1), which is updated to include the most ,

recent transient reviews. Where appropriate, these investigations serve as tha reference analysis for the Reload Cycle review, and are indicated as such. Each event is evaluated considering the most limiting time in core life. This evaluation includes conditions during coastdown beyond the EOFPL.

7.1.1 Initial Operatina Conditions Table 7-1 provides the initial operating cond'.tions that apply to most f i

of the transients analyzed. Any deviations from the values indicated are

' noted in the discussion of the specific transient.

T Only minor differences in the basic plant parameters exist between the Reload Cycle and the reference analysis (Reference 7.1). These minor ,

differences are tho following:

1. The core average LHGR is slightly higher for the Deload Cycle than i for the reference analysis. This difference resulted from the addition of the 206 solid Zircaloy rods and special guide bars to

! selected assenblies.

1

2. The minimum DNB ratio at design conditions for the Reload Cycle is ,

r marginally lower than for the reference analysis. The impact of this minor reduction in design DNBR will be addressed in the review f of wach appropriate transient. l L

[

i 3825R/4.329  !

t l

p

7.1.2 Reactor Trip Setpoints and Instepmentation Delays Table 7-2 presents t.ie reactor trip setpoints and instrumentation dalays applied in the transient analysis.

7.1.3 Reactivity Coefficients The moderator and fuel temperature coefficients, where they are inportant to the analysis,'will be discussed on an event-by-event basis.

7.2 Control Rod Withdrawal Incident For the reference analysis (FSAR), a bounding snalysis was performed with the following assumed conditions:

1. 7esign peaking factors were-used even though lower peaking factors exist during the incidant, as shown in Table 6-3;

2. Core powar was assumed to be at the overpower trip setpoint;

3. Cociant pressure was assumed to be at the lower end of the operating band; 4 Coolant temperatures were assumed to be at the values consistent I with steady-state operation at the overpower trip setpoint.

To the Reload Cycle, these conservatively assumed conditions remain bounding. However, the steady-state design DNBR is marginally lowe- far the l Roload Cycle than was calculated for the reference analysis. The reference enclysis shows that the minimum DNBR for this event remains significantly

cbove 1.3. Accounting for the slightly lower design DNBR for the Reload i Cycle, the consequences of this event are still within fuel design limits.

i 3825R/4.329

n, m . . - - _.

i l

7.3 Boron Dilution Incident i ~ 7. 3.1' Introduction ,.

Due to th'e nature of key reactivity parameters which changa slightly from cycle to cycle, a complete evaluation of the boron dilution event la parformed for each reload cycle to assure acceptance criteria are met fnr each oparational mode. The most recent review of the time to loss of shutdown ,

margin was performed for the Current Cycle which was submitted to the NRC in Rsference 7.2. This analysis reviewed each mode of operation allowed by Technical Specifications. An additional analysis presented in the FSAR j (Raference.7.1) investigated the minimum DNBR which could occur for a boron j dilution event in Mode 1.

l For the Reload Cycle, each mode of operation has been analyzed using ,

4 t eppropriate initial and critical boron concentrations. The initial boron i

' concentration in each mode is that needed to provide the minimum shutdown .

margin required by Technical Specifications. In all cases, the dilution is ,

assuuwd to proceed at the maximum capacity of the charging pumps (100 gpm).

The results are described as follows for each operational mode. These results l

are for the limiting BOC conditions. As the fuel exposure increases during j the cycle, the time to a loss of shutdown margin increases.

1 7.3.2 Analysis and Results

! t a

7.3.2.1 Boron Dilution During Modes 1 and 2 In Mode 1 or 2, there must be at least 15 minutes available for the l f l oparator to terminate the dilution prior to a complete loss of shutdown j i margin. The significant parameters for the Reload Cycle during Modes 1 and 2 l l st the most limiting condition are shown in Table 7-3. As shown, the Reload l Cycle remains well above the minimum allowable time criterion of 15 minutes, i

In addition to the tirce to loss of shutdown margin analysis, the l

maximum reactivity insertion rate was investigated for Modes 1 and 2 to ensure l l that the plant response for a boron dilution is bounded by the control rod  ;

l withdrawal event. The maximum reactivity insertion rate for a Mode 1 or 2 ,

I l 3825R/4.329  !

'I

boron dilution was last analyzed in support of the FSAR. The analysis used a higher initial bdron concentration than that assumed for the time to loss of shutdown margin analysis, since higher concentrations result in higher rasetivity insertion rates. Even for the high borou concentration assumed (2527 ppm), the FSAR analysis demonstrated that th: maximum reactivity insertion rate for a boron dilution is bounded by the control rod withdrawal evsnt. Since the maximum initial boron concentration for the Reload Cycle is isss than that assumed in the FSAR analysis, the plant response for a Mode 1 or 2 boron dilution is bounded by the results of the control rod withdrawal ovcat.

For the Mode 1 or 2 boron dilution event, the Low Pressure Surge Tcnk (LPST) level, temperature and pressure indication, as well as the high clarms on these three parameters, would provide information to aid the oparator in diagnosing the condition. Should the reactor be in the automatic control mode during a dilution ac power, the control rod group currently selected would insert to offset any temperature increase resulting from the core power / steam flow mismatch. Audible indication of this control rod motion would be available from the Containment Sound Monitoring System. Even if the control rods insert at the maxinum rate of 6 inches / minute, it would take approximately 15 minutes for the control rod group to be fully incerted.

During this time, the operator would be alerted by the high average temperature and high neutron flux alarms. Because of the available alarms and s indications, there is ample time to allow the operator to diagnose the situation, terminate the d11ution and restore adequate shutdown margin.

7.3.2.2 Boron Dilution During Mode 3 In Mode 3, there must be at least 15 minutes available for the operator to terminate the dilution prior to a complate loss of shutdown margin. The enslysis for the Reload Cycle boron dilution during Mode 3 used the wame cpproach as the FSAR and the previous reload analyscs. The significant pcrameters for the limiting case are presented in Table 7-4 As shown, the reload cycle in the limiting ca.e remains well above the minimura allowable time criterion of 15 minutes for this mode.

3825R/4.329

Close surveillance is required for a feed and bleed operation performed at the maximum capacity of the charging pumps (100 gpm). The possibility of insdvertently feeding unborated water at this magnitude is extremely remote.

H, wever, assuming this unlikely situation did occur, t.he operator would have a minimum of four alarms to alert him to an inadvectent dilution. These alarms .

include the.high neutron flux alarm and the LPST high level, temperature and-

~ pressure alarm'.

s Additional alarms would also be expected from the bleed line radiation monitor and possibly the main coolant pump low cooling flow alarm.

The latter alarm could be expected due to increased component cooling flow to cool the LPST in responso to its high temperature, resulting in a decrease in the cooling flow to the main coolant pumps. This event can be terminated quickly and easily from the Control Room by isolating the charging line, shutting off the charging pumps, or isolating the source of demineralized water.

7.3.2.3 Boron Dilution During Modes 6 and 5 In Modes 4 and 5, there must be at least 15 minutes available for the operator to terminate the d!Pation prior to a complete loss of shutdown margin. The most limiting boron dilution for these two modes of operation is from Mode 5 conditions as demonstrated in the analysis of previous cores. The significant parameters for the limiting case f rom these conditions are shown in Table 7-5. The Reload Cycle results remain well above the minimum allowable time criterion of 15 minutes.

Table 7-5 shows that ample time exists for the operator to acknowledge the high neutron flux alarm. In addition, high level, temperaturc and pressure alarms on the LPST would key the operator to the event. The operator would then take corrective action to terminate the dilution from the Control Room by isolating the charging line, shutting off the charging pumps, or isolating the source af demineralized water.

7.3.2.4 Boron Dilution During Mode 6 (Refueling)

In Mode 6, there must be at least 30 minutes available for the operator to terminate the dilution prior to a complete loss of shutdown margin. The current reference analysis for the boron dilution incident from Mode 6 3825R/4.329

conditions was performed in suppert of the FSAR (Reference 7.1). In both the reference are t.., Reload Cycle analyses, two cases were examined for Mode 6 conditions. The first case assumes the minimum main coolant volume when the rsactor vessel is drained to just below the head flange to allow head removal or installation. The second case represents the majority of time spent in l Mode 6 when the shield tanka c'vity is filled with borated water.

For the first case examined, the analysis assumed the minimum volume of l

water to be diluted, four isolated reactor coolant loops with the upper head drained in preparation for head removal or installation. The analysis also "

assumes the largest possible dilution rate of 100 gpm unborated water. For this case, a shutdown nargin of 7.,%dp is required in order to provide at leest 30 minutes for the operator to terminate the dilution. The significant paremeters for the minimum water volume case of the Mode 6 boron dilution incident are given in Table 7-6. The Reload Cycle remains above the minimum allowable time criterion of 30 minutes.

It is important to note that this combination of conditions could only occur when the reactor vessel head is about to be removed or installed. The 4 conditions during the bulk of time in Mode 6 include 32 feet of water above the top of the irradiated fuel assemblies and the tagging out of service of equipment that would make possible inadvertent reactivity increases, as 4

required by Technical Specifications. A more realistic way, therefore, to examine the Mode 6 boron dilution event was used as an additional case in the l

- FSAR analysis, and is repeated for the Roload Cycle. In this additional case, t

the .etive dilution volume assumes that all 4 loops are isolated, and the shleid tank cavity is filled to 32 feet above the top of the fuel assemblies l

with one-half of the shield tank volume contributing to the active dilution j volume. For this case, a 514p shutdown margin is adequate to provide the  !

required 30 minutes for the operator to terminate the dilution. The 7 significant parameters for this case are shown in Table 7-7. For this case,  !

substantial time is available for operator action.

I In Mode 6, the operator is provided with indication of a possible boron dilution via the audible count rate signal, which would increase, and the high nsutron flux alarm. The operator can then take correctivo action from the l 3825R/4.329 n ,- - - - - - - . , _ . , - - , _ , _ , , _ _ _ _ _ . _ , _ _ . .

,. s; Centrol Room by isolating the charging line, shutting off the charging pumps or isolating the source of' demineralized water.

-7;3.2.5 , Failure to Borate Prior to Cooldewn Because of the large negative temperature coefficient of reactivity at thz end of cycle, any decrease in main coolant system temperature increases the core raactivity state. Consequently, during the process of cooldown of tha main coolant system, it is necessary to ensure adequate shutdown margin by snsuring that adequate boron concentration and/or control rod worth is '

e oveilable.

The failure to ensure adequate shutdown margin prior to cooldown was re-evaluated for che Reload Cycle with the following basic assumptions:

a) The moderator defect versus temperature curve is used in assessing the reactivity addition, since the moderator temperature coefficient is a function of temperature.

b) The reactor is initially 1% suberitical at an average temperature of 520 F.

c) The shutdown margin at 520 F .s 5.514p, the minimum required by the Technical Specifications, for Modes 1 and 2.

d) The main coolant system temperature is reduced at the rate of 50 F/hr, the maximum cooling rate permitted by the Technical Specifications for a normal cooldown.

In order to make the reactor' critical from these initial conditions, th2 average coolant temperature must be reduced to approximately 495 F.

This temperature reduction requires approximately 27 minutes to accomplish.

This is ample time for the operator to diagnose the condition and take n:cessary corrective action. For a complete loss of shutdown margin, a cooldown to approximately 335 F would be required. which would take more th:n 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

t 3825R/4.329 i

1

, . . _ - - . . , . , - - _ - - - - - - - , . . , , , - - - - - - ,,rr- , - -,.wan,-- -- - , , - , -. . - - , , , _ , . - , , , _ , - , , ,

i 7.3.3 Conclusions l

The probability of erroneous dilution is considered very small because of the equipment, controls, and administrative procedures provided for boron dilution activities. However, in tne unlikely event that an unintentional dilution of boron in the main coolant system occurs, numerous alarms and indications are available to alert the operator of the candition. These alarms include high level, temperature and pressure on the Low Pressure Surge Tank, and the high neutron flux alarta, as well as the audible count rate signal in place during Mode 6 conditions. If the reactor is initially critical at the time dilution begins, the automatic safety features of the reactor protection system would ensure acceptable plant performance. For baron dilutions initiated during any operational mode, adequate time exists for the operator to determine the cause of the dilution and take corrective l

action before a complete loss of shutdown margin occurs.

7.4 Control Rod Drop Incident For the reference analysis (Reference 7.!), a bounding analysis was performed assuming steady-state operation at a conservatively determined f

! limiting end point for the transient. The following assumptions were made in l

l the analysis:

1. The highest calculated radial peaking factors with uncertainties i

i for any dropped rod were used;

2. The design core power of 618 MWt was assumed as the power level;

3. Main coolant pressure was assumed to be at the low pressure trip setpoint to minimize the DNB ratio;

4. Core inlet coolant temperature was assumed to be at the design value consistent with the design core power.

Table 7-8 shows a higher maximum LHGR for the Reload Cycle compared to the reference analysis due to an increase in the post-drop peaking. The resulting increase in the maximum LHGR required that the Control Rod Drop 3825R/4.329

event be analyzed f the Reload Cycle. The Reload Cycle evaluation results provided in Table 7-8 show that fuel performance is well within the acceptable limits for this event.

7.5 Isolated Loop Startup Incident Three-loop operation in Mode 1 or 2 is not presently allowed by the Technical Specifications. This transient is no longer applicable to plant operation and, therefore, was not considered for the Reload Cycle.

7.6 Loss of Load Incident For the reference analysis (FSAR), a bounding analysis was performed using the design Technical Specification value for the Moderator Temperature Coefficient (MTC). The FSAR analysis assumes an upper limit on the pressurizer safety valve setpoint tolerance of +3%. The main coolant system high pressure trip was also credited in the analysis. In general, this incident is not sensitive to minor changes in core parameters. In addition to the reference analysis, numerous parametric analyses were performed in support of the Core 14 reload submittal (Reference 7.3). These sensitivity studies on MTC and Doppler coefficient demonstrated the minor impact of core physics parameters on the loss of load transient. The Reload Cycle physics parameters are bounded by those assumed in the reference analysis. Thus, the Reload Cycle plant response to a loss of load is within system design limits, and therefore acceptable.

7.7 Loss of Feedwater Flow Incident The most recent review of the loss of feedwater flow incident was performed in support of the FSAR (Reference 7.1). Each of the assumptions made in this analysis bounds the Reload Cycle system characteristics. For this event, the most significant requirement is maintenance of a steam generator heat sink. The analysis provided in Reference 7.1 concluded that plant performance for this event was acceptable. It was concluded that the combination of the reactor protection system and emergency feedwater system assured the integrity of the core and primary and secondary system pressure boundaries by 1) reactor trip on low steam generator water level, and 2) 3825R/4.329

auxiliary feedwater flow sufficient to assure adequate steam generator liquid ,

inventory for primary system cooldown, decay heat removal, and main coolant pump heat removal for the entire course of the event.

Since the Reload Cycle operating conditions are bounded by this analysis, it is concluded that the Reload Cycle response to a loss of feedwater is acceptable.

7.8 Loss of Coolant Flow Incident The reference analysis for the loss of coolant flow transient was performed in support of the Current Cycle (Reference 7.2). The event is sensitive to core parameters, Reactor Protection System setpoints, and steady-state thermal margin.

The analysis assumes a least negative MTC and most negative Fuel /

Temperature Coefficient (FTC). Table 7-9 shnws that the FTC including uncertainty is more negative for the Reload Cycle than the reference analysis, requiring the event to be reanalyzed for the Reload Cycle. The total amount of fuel pin failures predicted for the 4-of-4 pump loss of flow event for the a-reload core remains below 10.0%. Thus, radiological evaluation for rod ejection events, which assumes 10% fuel failures, bounds the results of the ,'

complete loss of flow event.

It is important to note that the 4-of-4 pump loss of flow transient is .

considered to oe a very unlikely event due to the main coolant pump power source diversification. Two main coolant pumps are powered by the generator while the two other pumps are powered by the two separate off-site ac lines.

Even with a complete loss of off-site power, two main coolant pumps would produce near full flow while the generator is coasting down, and thus reduce -

the .everity of this transient.

Anticipated 1o.1 of flow events include the 1-of-4 and 2-of-4 pump losses. The 1-of-4 pump loss of flow event does not result in a reactor trip. The 2-of-4 pump loss of flow transient results in a reactor trip on low flow. Both of these cases produce similar DNB results. For the Reload Cycle.

,P 3825R/4.329

k' pt' the 2-of-4 loss of flow is the limiting case with a minimum DNBR of 2.01. The results of both cases are well within the acceptable liruits for this event.

) -

7.9 Control Rod Ejection Accident The most recent control rod ejection accident analysis was performed in support of the FSAR (Reference 7.1). A comparison of pertinent parameters, including calculational uncertainties, affecting the event for the reference analysis and the reload core is provided in Tables 7-10 and 7-11. ' e s

Both the zero power and full power rod ejection events required reanalysis for the Reload Cycle. The zero power rod ejection required reanalysis since the MTC and post-ejection peaking are not bounded by the reference values. For the full power case, the post-ejection peaking is not bounded by the reference value. The results of both analyses show the average enthalpy of the hottest fuel pin is less than 200 cal /gm, and the centerline enthalpy is below 250 cal /gm. Therefore, no fuel cladding damage is predicted to occur for either event. [,

7.10 Steam Line Break Accident The reference analyses for the main steam line break transient were performed in the Current Cycle reload analysis (Reference 7.2) and the FSAR (Reference 7.1). The moderator and Doppler reactivity feedback parameters and f:

inverse borun worths for the Reload Cycle were found to be more limiting for the steam line break than the values generated for the reference analysis. L The scram rod worths for the Reload Cycle have improved over the values irem ,,

the reference analysis. A reanalysis of the main steam line break was performed for the Reload Cycle to show the net effect of the change in reactivity. The reanalysis was performed using a constant loop average temperature based on the calculated average temperature at the current ..

limiting full power conditions. This results in a higher average temperature at low power levels than is currently allowed by plant Technical Specifications, thereby making this analysis conservative for the current -

operating conditions. The reanalysis bounds current operation and supports potential operation with an average loop temperature limit instead of the \ \

current cold leg temperature limit. The results for the limiting case show a 3825R/4.329

s minimum vr.beriticality of -0.223%4p. Therefore, there is no return to power following the stear,liae break. This precludes CNB and fuel cladding damage. ,

7.11 Steam Cenerator Tube Rupture Incident C

The reference analysis for this event, Core 11 de.nonstrated that the results are not sensitive to the core design. Thus, the results of the analysis presented in Reference 7.4 will also apply to the Reload Cycle. .

,- v 7.12 Fuel Handling Incident j ,

The source terto used for the fuel handling incident is the gap activity within the highest tated fuel assembly to be moved during the reteeling -[

outage. The cource term includes contributions from both long and short ,'"

half-life fission gas products. The guidance used for calculating maximum gap .

activity is Regulatory Guide 1.25 (Reference 7.5), and is based on the peak linear power density, the maximum operating fuel centerline temperatures, and .'

assembly burnup.

The fuel rod gap activity for the Reload Cycle is bounded by the values gg used in the FSAR analysis (Reference 7.1).

7.13 Other Accidents and Transients Previous analyses (Reference 7.6) have demonstrated that the Waste Gas Incident and Reactor Containment Pressure Analysis are not sensitive to core design changes and, therefore, the results presented in Reference 7.7 still apply to the Reload Cycle.

l l 7.14 Transient Analysis Summary 1

The results of the transient analysis review of the Reload Cycle are shown in Table 7 52. Provided are the criteria for each incident, and the results for the reference analysis as well as the Reload Cycle analysis. .

These results provide assurance of continued safe operation of the YNPS.

3825R/4.329 4

TABLE 7-1 FOUR-LOOP INITIAL OPERATING CONDITIONS Reference Analysis Parametnr Reload Core FSAR Reactor Power, MWt 600 + 18 600 + 18 Core Inlet Temperature OF 520 + 4 520 + 4 Main Coolant Pressure, psia 2015 1 75* 2015 1 75*

Minimum Reactor Coolant Flow. 35.0 35.0 10$ lb/hr Axial Heat Flux Profile Cosine Cosine Total Hea'. Flux Factor 2.76 2.76 Nuclear Enthalpy Rise 1.80 1.80 Factor l

l l

l t

o Includes 50 psi instrument uncertainty and 25 psi operating deadband.

3825R/4.329

TABLE 7-2 REACTOR TRIP SETPOINTS AND INSTRUMENTATION DELAYS Trip Functions Setpoint Delay Time (sec)

High Startup Rate 5.2 decades / min. 0.3 High Neutron Flux 112% 0.4 Pump Current Deviatior.s high/ low current 0.6 on two pumps High Pressurizer Water Level 209 inches 0.65 Low Main Coolant Pressure 1735 psig 0.6 Low Steam Generator Water Level -13.0 inches 2.0 on two steam generators Main Steam Line Isolation Trip 200 psig 0.128 High Main Coolant Pressure 2350 psig 2.0 l

3825R/4.329

TABLE 7-3 MODES 1 AND 2 BORON DILUTION

• Parameter Reload Cycle FSAR Minimum Coolant Volume, ft 3 ** 2400 2400 Limiting Initial Boron 2038 1858 Concentration, ppm Minimum Required Shutdown 5.5 5.5 tbrgin (14p)

Time to Loso of Shutdown 51.7 58.0 Margin, Minutes

• These operating modes are defined as follows:

Mode 1 2 Description Power Operation Startup Reactivity, K gg e ).99 >.99 Power Level, % Rated 12 <2 Coolant Temperature,0F 1330 1330

• • Assumes one main coolant loop isolated.

3825R/4.329 l _ - _ _ - _ _ _ -

,T/BLE 7-4 MODE 3 BORON DILUTION

• Parameter Reload Cycle FSAR Minimwn Coolant Volume, f t 3.* 1700 1700 Limiting Initial Boron 1981 1770 Concentration, ppm Minimum Require 3 Shutdown 5.0 5.0 Margin (14p)

Time to Loss of Shutdown 33.8 36.3 Margin, Minutes l

l

• Mode 3 is defined as follows:

Description Hot Standby Reactivity, K err <.99 Power Level, % Rated 0 Coolant Temperature,0F 1330

• • Asetsmes three main coolant loops isolated.

3825R/4.329 I . - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

TABLE 7-5 MODES 4 AND 5 BORON DILUTION

• Parameter Reload Cycle FSAR n

Minimum Coolant Volume, ft 3 ** 1276 1276 Limiting Initial Boron 2092 1932 Concentration, ppm Minimum Required Shutdown 5.0 5.0 Margin (%dp)

Time to Loss of Shutdown 23.9 26.2 Margin, Minutes o These operating modes are defined as follows:

Mode 4 5 Description Hot Shutdown Cold Shutdown Reactivity, K egg 1 96 (.96 Power Level, % Rated 0 0 Coolant Temperature,0F 200 < T 1 330 1200 0* tasumes four main coolant loops isolated, with provisions for upper reactor l vessel head draining.

3825R/4.329

TABLE 7-6 MODE 6 BORON DILUTION

• ALL LOOPS ISOLATED, ,,

MINIMUM WATER LEVEL Parameter Reload Cycle FSAR Minimum Coolant Volume, ft3 1276 1276 Limiting Initial Boron 2352 2137 Concentration, ppm Minimum Required Shutdown ** 7.5 7.5 .

Margin (%dp)

Time to Loss of Shutdown 33.6 34.5 Margin, Minutes d

9 s

F l

l

• Mode 6 la defined as follows:

Description Refueling ,

Reactivity, K egt 1 95 Power Level, % Rated 0 Coolant Temperature,0F 1140

• • Required by Technicel Specification 3.9.1.

J 3825R/4.329

?

, TABLE 7-7 MODE 6 BORON DILUTION

• WATER VOLUME INCLUDING SHIELD TANK Pa rane te r Reload Cycle FSAR Minimum Coolant Volume, ft 3** 8650 8650 Limiting Initial Boron 2092 1932 ~

Concentration, ppm Assumed Shutdown Margin (14p) 5.0 5.0 Time to Loss of Shutdown 162.1 177.0 Margin, Minutes

(

4

/

• Mode 6 is defined as follows:

Description Refueling Reactivity, K gg e 1 95 Power Level, % Rated 0 l

Coolant Temperature,oF 1140 l

A* Assumes 4 loops isolated and shield tank cavity filled to 32 feet above fuel assemblies, with 1/2 of the shield tank volume contributing to the active dilution volume.

(

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b TABLE 7-8 CONTROL R0D DROP INCIDENT PARAMETERS t)

Parameter Reload Cycle Reference Analysis Core Power, MWt 618 618 Core Inlet Temperature, OF 524 524 Main Coolant Pressure, psia 1750 1750 Maximum Linear Heat Rate 14.6 14.4 (post-drop, with uncertainties),

kW/ft Maximum Fuel Centerline 3699 3704 Temperature, OF Minimum W-3 DNB Ratio 2.04 2.17 1

e 3825R/4.329

~

TABLE 7-9 LOSS OF C001 ANT FLOW PARAMETERS Parameter Reload Cycle Reference Analysis a Moderator Temperature Coefficient

(.1.0-hop / 0F ) -0.30 0.0 Fuel Temperature ^

Coefficient '

(10-5apfor) -2.19 -1.84 Scram Rod Worth Less Group C Integral -

Worth (U p) 7.56 6.0 I

3825R/4.329

TABLE 7-10 HZP ROD EJECTION ACCIDENT PARAMETERS Parameter Reload Cycle Reference Analysis BOC EOFPL Moderator Temperature Coefficient (10-4dp/0F) +0.32* -2.28 +0.27*

Doppler Coefficient (10-5ap/oF) -1.39 -1.55 -1.10 Ejected Rod Worth (lap) .82 .84 0.93 Delayed Neutron Fraction (D) .00o543 .005411 .005490 p/S 1.250 1.558 1.694 Fg Following Rod Ejection 4.42 4.19 4.22 l

• This includes an uncertainty of 0.5 x 10-4dp/0F. Technical Specifications require MTC to be negative under Hot Zero Power conditions.

3825R/4.329

TABLE 7-11 HFP ROD EJECTION ACCIDENT PARAMETERS Reload Cycle Reference Analysis Parameter BOC EOFPL Moderator Temperature Coefficient

,(10-44p/0F) -0.30 -2.58 0.0*

Doppler Coefficient (10-5ap/oF) -1.17 -1.31 -0.766 Ejected Rod Worth

(%dp) .21 .21 .50 Delayed Neutron Fraction (Q) .006543 .005411 .005743 p/S .315 .380 .871 Fq Following Rod Ejection 3.22 2.36 3.17 l

l l

• Technical Specifications require MTC to be more negative than

-0.2 x 10'44p/0F at Hot Full Power conditions.

3825R/4.329

es TABLE 7-12

?"

z~ -

• YANKEE CORE 20 w

SAFETY ANALYSIS o"

SUMMARY

OF RESULTS Criteria Reference Analysis Core 20 Incident Section MDNBR > 1.30 MDNBR > 2.00 MDNBR greater than 2.00 Control Rod 7.2 RCS pressure RCS pressure RCS pressure Withdrawal less than 2300 psia

< 2750 psia < 2300 psia Suberitical: Suberitical:

Boron Dilution 7.3 Suberitical:

Sufficient time for Creater than 15 minutes - Greater than 15 min. -

operator action Modes 3, 4 and 5 Modes 3, 4 and 5 Greater than 30 min. - Greater than 30 min. -

Mode 6 Mode 6 Critical: Critical: Critical:

Reactivity addition Bounded by control rod Bounded by control rod S withdrawal i' rate withdrawal MDNBR = 2.17 MDNBR greater than 2.00 Control Rod Drop 7.4 MDNER > 1.30 No fuel centerline Fuel centerline Fuel centerline melt temperature = 37040F temperature = 36990F 7.5 MDNER > 1.30 MDNBR greater than 2.97 Power operation with loop Isolated Loop out of service prohibited Startup No fuel centerline Fuel centerline i melt temperature = 34850F by Technical l I

Specifications Maximum RCS pressure = Not analyzed for Core 20 Loss of Load 7.6 RCS pressure less than 2750 psia 2663 psia (see text) l 1

l 1

l l

l l

l l

o$

21 TABLE 7-12 U YANKEE CORE 20

• SAFETY ANALYSIS

SUMMARY

OF RESULTS (Continued)

Reference Analysis Core 20 Incident Section Criteria Suff'.ciert time for Emergency feedwater Emergency feedwater Loss of Feedwater 7.7 required 15 minutes required 15 minutes initiation of emergency feedwater following event following event 7.8 10CFR100 Less than 10.0% fuel Less than 10.0% fuel Loss of Coolant failure Flow failure No clad damage No clad damage Control Rod 7.9 10CFR100 4, Ejection Steam Line 7.10 Maintain feel rod No return to critical No return to critical Rupture integrity 7.11 10CFR100 Radiological doses well Radiological doses well Steam Generator within 10CFR100 Tube Rupture within 10CFR100 7.7.2 10CFR100 Radiological doses well Radiological doses well Fuel Handling within 10CFR100 within 10CFR100 MS

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

The Startup Program for the Reload Cycle will include the following tests:

1. Control rod operability tests will be performed by moving each rod group in turn from 0" to 90" to 0" and verifying control rod movement by the rod position indicators.

2. Control rod drop time measurements will be conducted by withdrawing one rod group at a time to 90 inches, dropping it and measuring individual rod drop times with a recording oscillograph.

3. Just critical boron concentration is determined by placing the reactor just critical, allowing for system equilibrium and takir,g a series of main coolant boron samples. This will be done as close as possit a to the conditions of all rods out and Group C inserted.

4. Control rod group worths of Group C Group A, and Group B will be determined. This is done by establishing a boron change, balancing the reactivity change with a control rod position change and measu.i.1g the reactivity worth of the rod steps with the reactivity computer.

l l

5. Isothermal temperature coeffic! .t meas..ements are performed by changing main coolant temperat -c r. sutir.g the reactivity change with the reactivity computer. Measurements are taken At the equilibrium boron concentrations which correspond to both unrodded and rodded core conditions.

6. Power and xenon defects are inferred uoing a reactivity balance before and after power ascension.

3825R/4.329

7. Power distributions will be measured as soon as the reactor is at steady-state power (50% 1 power level 1 75%). This is done with the Incore Instrumentation System. A power distribution map will at.so be taken at a low power leve'. to check for gross quadrant tilt.

8. A startup test report on the above will be submitt6d to the NRC 90 days after startup The acceptance criterir. for the prediction of key core parameters is d3 fined ad Table 8 6. The permissible deviation from predicted values are salected to insure the adequacy of the safety analysis. In these tests, the nominal measured value is compared to the nominal calculated value.

Ce rections are made for any differences between the measurement and cciculational conditions.

If the criteri.a in Table 8-1 are not met, the deviations are evaluated relative to the assumptivns in '.he satety analyr I for the given core 4 ,,crameters. The Ptant Operations Review Committee reviews the evaluation prior to power operation.

1

~ ~

- t./ 329

t t

TABLE 8-1  :

YANKEE CORE 20 STARTUP TEST ACCEPTANCE CRITERIA l 4

i Measurement Conditions Criteria to Control Rod Drop Time Operating temperature Drop times no greater than 2.5 seconds l r

2. Critical Boron Hot zero power, near Measurement within Concentration all rods out 210% of predicted  ;

value  ;

3. Control Rod Group Not zero power Groups Worth of each group  !

Worths C. A and B within ! 7.5% of the I predicted value j 4 Con *.rol Rod Group Hot zero power. Groups If the criteria in l Warths C. A and B Measurement (3) is  ;

not met, the total i

' sorth of all Groups i measured must be ,

i within 275% of the i tredicted value t

! 5. Isothermal Temperature Hot zero power, near Measurement within ,

Coefficient all rods out 105x10-'Ap/0F l

of predicted value i

4

6. Radial Power Above 50% power with all Thc measured reaction s Distribution rod groups greater than rates within 15% of

. 80 inches withdrawn the predicted value {

! in the high power a assemblies  :

4 l

a r i

t f i i  ;

i I I I i

i r

i i  !

l l

i i

ja, 3825R/4.329 l l

4  !

i  !

j.  !

9.0 LOSS OF COOLANT ACCIDENT 9.1 Introduction For the Reload Cycle, the fuel is similar in design to the fuel used since Core 18. Therefore, the only differences in reactor performance to a Loss of Coolant Accident (19CA) are related to minor changes in physics parameters. As such, much of the previous analysis results .an be used to support the Reload Cycle operation.

9.2 Small Break LOCA 1

The assumptions made in performing the Core 13 small break analysis (Raference 9.1) conservatively encompass any minor changes in physics parameters that occur from cycle to cycle. The core 13 small break LOCA cnslysis was performed with the limiting fuel stored energy which was calculated to occur for the fresh fuel at EOC conditions. A cosine axial power distribution was used for this analysis. Break sizes ranging fron 10 inches down to 2-1/4 inches in diameter were analyzed with PLHCR of 12.85 kW/ft. The 4-inch bteak with a Peak Cladding Tecperature (PCT) of 1793 F was found to be the limiting small break, i

For Core 18 (Reference 9.2), analysis was performed to determine the icpact of top-skewed core power distributions and a 5 F increase in the ellowable core inlet temperature. The ar"lysis was performed for the limiting brcak size, and several axial power shapes were anai* zed. The results were aompared to the corresponding large break results which demonstrated that smell breaks are less limiting than latet breaks.

Since the axial power shapes for the Reload Cycle are very similar to those of Core 18, the Core 18 analysis results demonstrate that for the Reload Cycle, the small break spectrum will continue to be nonlimiting.

9.3 g Break LOCA 1

The larg' break LOCA analysis consists of a break spectrum analysis and i o burnup sensitivity study.

i 3825R/4.329

The break spectrum analysis is used to determine the limiting break size and type. The analysis is performed utilizing BOC conditions which rcsult in the maximu:n fuel stored energy. Since cosine $tnd top-skewed axial power shapes are possible over the core lifetime, both shapes are analyzed in ths break spectrum analysis.

The burnup sensitivity study ir performed to establish the LHGR limits cs a function of c/cle exposure. The wordt case axial shape possible at each burnup is used in calculating the limit.

9.4 Break Spectrum Analysis The Reload Cycle fuel is nearly identical in design and in hydraulic parformance to previous cycles. The pnysics parameters used at BOC condi, ions for Core 16 and Core 18 were compared to the Reload Cycle values. An cycluation was made which showed the Reload Cycle values either were boundeo by the previous values or had an insignificant effect on analysis results.

Th0refore, the break spectrums performed for Core 16 (Reference 9.3) and Core 18 (Reference 9.4) remain valid for this cycle. The Core 16 analysis provides the limiting break characteristics for the cosine power shapes while tha Core 18 analysis provides those for the top-skewed shapes.

9.5 Burnup Sensitivity Study The burnup sensitivity study addressed both fresh fuel and recycled fusi performance in the Reload Cycle. For burnup points at less then 4 CWd/Mtu, a cosine axial power shape is assumed since the "worst case" xenon j shapes closely approximste a cosine. Beyond 4 GWd/Mtu, the shapes become top-skewed and theref ore, the actual "worst c.ase" xenon shapes are ubed.

Tne results of the sensitivity study are given in Table 9-1. Where the input parameters of the previous cycles t.3 nded the Reinsd Cycle parameters, no analysis was performed.

The analysis for the Reload Cycle shows the LOCA limi u of the previous cycles (Core 18 and Core 19) to be valid for the Reload Cycle. Therefore.

Roload Cycle operation using previous LOCA limits is acceptable. These limits 3825R/4.329

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cro given in Figure 9-1 where all LHGR values were multiplied by 0.973 to cecount for heat deposition in the fuel rod.

9.6 Summary of Results Based on the analysis presented in Sections 9.0 through 9.5, it is concluded that the Current Cycle LHGR LOCA limits are conservative with I respect to the Reload Cycle. Therefore, no change to the LOCA limits are required for operation of the Reload Cycle. Operation within the limits specified in Figure 9-1 yields LOCA results within the specification of 10CFR50.46.

9.7 Padiological Consequences _ of a Deslan Basis LOCA and Post-LOCA Hydrogen Control The radiological consequences from the design basis LOCA and Post-LOCA Hydrogen Control must conform to the guideline values specified within 10CTR100. The calculation of the source term for these events assumes a total fission product inventory for an equilibrium core. Thus, the calculation of ths source term is dependent on reactor power during operation. The reactor

, power level for the Reload Cycle will not be greater than the reactor power level assumed In the FSAR analysis for this event (Reference 9.5). Therefore, tha results of the analysis presented in the FSAk also apply to the Reload Cycle.

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TABLE 9-1 CORE 20 BURNUP SENSITIVITY STUDY RE$ULTS CAB Fuel PLHCR Limit Core 20 Reference mwd /Mtu Type kW/ft PCT ( F) Cycle PCT (UF) 0.0 Fresh 10.20

• 2099=*

0.25 Fresh 11.25 2189 2145 1.00 Fresh 11.80

• 2150**

4.00 Fresh 11.00 1933 1971 10.00 Fresh 9.60

• 1973 14.00 Fresh 9.40

• 2146 0.0 Recycled 11.45

• 2164**

4.0 Recycled 10.50

• 1951 10.0 Recycled 9.30

• 2177 14.0 Recycled 9.10

• 2093 17.5 Recycled 8.00

• 1565 i

i t

O Bounded by previous cycle results, point not reanalyzed.

00 Calculation performed at PLHCR greater than the PLHGR limit.

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10.0 REFERENCES

4.1 C-E Report, "Yankee Nuclear Fuel Design Report Update for Batch C."

June 1988.

4.2 C-E Report, "Yankee Nuclear Fuel Design Report Update for Batch B,"

February 1987.

4.3 K. P. Galbraith, "GAPEXX: A Computer Code for Predicting Pellet-to-Cladding Heat Transfer Coefficients", XN-73-25, August 13, 1973.

4,4 YAEC-1496, "Yankee Nuclear Power Station Core XVIII Performance Analysis," August 1985.

4.5 YAEC-15P3, "Yankee Nuclear Power Station Core XIX Performance Analysis,"

January 1987.

5.1 W. R. Cadwell, "PDQ-7 Reference Manual", WAPD-TM-678. January 1967.

5.2 R. J. Breen, O. J. Marlowe and C. J. Pfeifer. "HARMONY: System for Nuclear Reactor Depletion Computation", WAPD-TM-478, January 1965.

5.3 R. F. Barry "LEOPARD - A Spectrum Dependent Nonspatial Depletion Program", WCAP-2795, March 1965.

5.4 D. M. VerPlanck, "SIMULATE A Nodal Core Analysis Program for Light Water Reactors," July 1982.

5.5 W. D. Leggett and L. D. Eisenhart. "The INCORE Code", WCAP-7149 December 1967.

6.1 USNRC Letter. D. Crutchfield to J. A. Kay, dated July 22, 1981.

6.2 ASME Paper, S. S. Markowski, et al., "Effect of Rod Bowing on CHF in PWR Fuel Assemblies," AICHE-ASME Heat Transf er Conf erence Salt Lake City, Utah August 15-17, 1977.

7.1 Final Safety Analysis Report, Yankee Nuclear Power Station, July 1987.

7.2 YAEC-1583, ' Yankee Nuclear Power Station, Core XIX Performance Analysis,"

January 1987.

7.3 Letter, WYR 78-99, dated November 21, 1978 D. E. Vandenburgh to USNRC, "Additional Information - Core XIV Refueling."

7,4 Proposed Change No. 115. "Core XI Refueling," submitted DOL /AEC on March 29, 1974.

7.5 Regulatory Guide 1.25, "Assumptior.a Used for Evaluating the Potential Radiological Consequenc.s of a Fuel Handling Accident in the Fuel Handling and Storage Facility for BWRs and PWRs," dated March 23, 1972.

3825R/4.329

7.6 Change No. 97 to License DPR-3 (Docket No. 50-29). -[

'7.7 Amendment No. 9. Letter f rom K. R. Go11er. DOL /AEC to YAEC, Attention: G. C. Andognini, July 30, 1974.

9.1 Proposed Change No. 145 Supplement No. 7, WYR 77/90, "Additional Yankee Rowe Core XIII Small Break Analysis," September 21, 1977.

. 9.2 Letter, FYR 85-131, dated November 19, 1985, G. Papanic, Jr., YAEC to .

l USNRC," Core XVIII LOCA Analysis - Additional Information."

.9.3 YAEC-1325, "Yankee Nuclear Power Station Core XVI Performance Analysis,"

September 1982. P 9.4 YAEC-1496 "Yankee Nuclear Power Station Core XVIII Performance Analysis," August 1985. L 9.5 Final Safety Analysis Report, Yankee Nuclear Power Stativn July 1987. t I

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