ML20055G701

Yankee Nuclear Power Station Core 21 Performance Analysis

ML20055G701
Person / Time
Site:	Yankee Rowe
Issue date:	07/31/1990
From:	Paul Bergeron, Cacciapouti R, Sundarem R YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20055G697	List: ML20055G696 ML20055G701
References
YAEC-1731, NUDOCS 9007240084
Download: ML20055G701 (85)

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I Iankee Nuclear. Power Station Core 21 Performance Analysis I

l July 1990 Major Contributors:

.'J. DiStefano N. Fujita I

R.~C. Harvey K. J. Morrissey l

M. E. Napolitano I

R. C. Paulson P. B. Perez D. A. Rice I

J. D. Robichaud R. P. Smith K. E. St. John M. A. Volk I

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i l I WPP40/4 I

1

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I 7 /I fd Approved By:

I P. A. Berg'eron, nager (Ifate)

Tran lent Analys s Group Approved By:

1 dN bt 7

'7 /96 1}fJ. Caccia uti, Manager

' (Da t'e )

Reactor Phys es Group "0**

7/'7[90 Approved By:

R. K. Sundaram, Manager (Date)

LOCA Group b 0 'I ' 6l**/N 7//4)io Approved By:

B. C. Slifer, Director (Date)-

Nuclear Engineering Department I

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I Yankee Atomic Electric-Company I

Nuclear Services Division 580 Main Street Bolton, Massachusetts 01740 I

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, WPP40/4

I DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company and is completely true and accurate to the best of our knowledge, information and I

belief.

It is authorized for use specifically by Yankee Atomic Electric Company, and the appropriate subdivisions within the Nuclear Regulatory Commission only.

With regard to any unauthorized use whatsoever, Yankee Atomic Electric Company, and its officers, directors, agents and employees assume no liability I

nor make any warranty or representation with respect to the contents of this document or to its accuracy or completeness.

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I WPP40/4

I ABSTRACT I

This report describes the mechanical, thermal-hydraulic, physics, and safety analyses necessary for the support of the Core 21 reload cycle for the Yankee Nuclear Power Station.

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t TABLE OF CONTENTS I

East Section iii DISCLAIMER OF RESPONSIBILITY.....................................

Av ABSTRACT.........................................................

vii LIST OF FIGURES..................................................

vili LIST OF TABLES...................................................

ix ACKN0WLEDGEMENTS.................................................

I 1

1.0 INTRODUCTION

2 2.0 OPERATING HISTORY OF CURRENT CYCLE...............................

I 4

3.0 GENERAL DESCRIPTION..............................................

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

8 8

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

9 4.2 Thermal Design.............................................

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

5.0 NUCLEAR DESIGN...................................................

17 5.1 Physics Characteristics....................................

17 5.2 Reactor Physics Analytical Computer Codes..................

18 5.3 Changes in Analytical Me thods..............................

19 6.0 THERMAL-HYDRAULIC DESIGN.........................................

33-7.0 ACCIDENT ANALYSIS................................................

40 7.1 Introduction...............................................

40 7.1.1 Initial Operating Conditions.......................

40 7.1.2 Reactor Trip Setpoints and Instrumentation 40 Delays.............................................

I 7.1.3 Reactivity Coefficients............................

41 t

l 7.2 Control Rod Withdrawal Incident............................

41-i 7.3' Boron Dilution Incident....................................

41 7.3.1 Introduction.......................................

41 I

42 7.3.2 Analysis and Results...............................

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I TABLE OF CONTENTE

. I (continued)

Section Page 7.3.2.1 Time to Loss of Shutdown Margin..........

42 7.3.2.2 Maximum Reactivity Insertion Rate........

44 I

7.3.2.3 Failure to Borate Prior to Cooldown......

45 7.3.3 Conclusions........................................

46 7.4 Control Rod Drop Incident'..................................

46 7.5 Isolated Loop Startup Incident.............................

47 I

7.6 Lo s s o f Lo a d In c i d en t......................................

47 7.7 Loss of Feedwater Flow Incident............................

48 7.8 Loss of Coolant Flow Incident..............................

48 7.9 Control Rod' Ejection Accident..............................

50 I

7.10 Steam Line Break Accident..................................-

50 7.11 Steam Generator Tube Rupture Incident......................

51 7.12 Transient Analysis Summary.................................

51 8.0 STARTUP PR0 GRAM..................................................

66 9.0 LOSS OF COOLANT ACCIDENT.........................................

69 I-9.1 Introduction...............................................

69 9.2 Small Break L0CA...........................................

69 I

9.3 Large Break L0CA...........................................

69 9.4 Break Spectrum Analysis....................................

70 9.5 Burnup Sensitivity Study...................................

70 9.6 Summary of Results.........................................

71 9.7 Radiological Consequences of Design Basis LOCA and Post-LOCA Hydrogen Contro1.................................

71

10.0 REFERENCES

74 I

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I LIST OF FIGURES Humher Iltl.c East 2-1 Comparison of Measured and Calculated Relativ6 Relative Assembly Average Power Distribution 3-3-1 Current Cycle Loading Pattern 5

3-2 B00 Assembly Average Burnup 6

3-3 Position of Recycled Assemblies in Reload Cycle 7

4-1 Core Locations of Modified Assemblies 13 4-2 Lattice Locations of Solid Rods 14 4-3 B00 ruel Centerline Temperature 15 4-4 EOC Fuel Centerline Temperature 16 5-1 Relative Radial Power Distribution 500 mwd /Mtu All-Rods-Out 24 5-2 Relative Radial Power Distribution 8,000 mwd /Mtu l

All-Rods-Out 25 5-3 Relative Radial Power Distribution 14,500 mwd /Mtu All-Rods-Out 26 5-4 Relative Radial Power Distribution 500 mwd /Mtu Group C Inserted 27 5-5 Relative Axial Power Versus Group C Insertion 28 5-6 Control Rod Group Identification 29 5-7 Rod Insertion Limit Versus Power Level 30 5-8 Xenon Redistribution Factor 31 5-9 Reduced Load Multiplier 32 6-1 Reactor Core Safety Limit - All-Loops in Operation 39 I

9-1 Allowable Peak Rod LHGR Versus Cycle Burnup 73 I

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-vil-WPP40/4 I

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LIST OF TABLES I

Number Iilla Ease 4-1 Nominal Mechanical Design Parameters 11 5-1 Summary of Curreat and Reload Cycle Nuclear Characteristics 20 5-2 Current and Reload Cycle Shuidem Requirements, %Ap 22 5-3 Comparison of Current and Reload Cycle Control Rod Worths 23 6-1 Thermal-Hydraulic Data Sheet During 4-Loop Operation 35 6-2 Thermal-Hydraulic Data Sheet for Comparison of Design I-Characteristics During 4-Loop Operation 36 6-3 Summary of Hot Spot and Hot Channel Factors Versus Design 37 6-4 Nominal Hot Channel DNBR, FAH, and Fg as Functions of Group C Position 38 7-1 Four-Loop Initial Operating Conditions 52 7-2 Reactor Trip Setpoints and Instrumentation Delays 53 7-3 Modes 1 and 2 Boron Dilution 54 7-4 Mode 3 Boron Dilution 55 1

7-5 Modes 4 and 5 Boron Dilution 56 7-6 Mode 6 Boron Dilution - All Loops Isolated Minimum Water 57 Level 7-7 Mode 6 Boron Dilution - Water Volume Including Shield Tank 58 7-8 Boron Dilution During Reduced Level Operation

• 59 7-9 Control Rod Drop Incident Parameters 60 7-10 Loss of Coolant Flow Parameters 61 7-11 HZP Rod Ejection Accident Parameters 62 7-12 HFP Rod Ejection Accident Parameters 63 7-13 Safety Analysis Sumary of Results 64 8-1 Startup Test Acceptance Criteria-68 ll g 9-1 Bt.rnup Sensitivity Study Results 72 l g WPP40/4 I

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

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_ix.

WFP40/4 I

1.0 IRIRODUCTION This report describes the basis for operation of the Yankee Nuclear Powar Station (YNPS) through the forthcoming Core 21 (Reload Cycle).

This report is also being submitted to the Nuclear Regulatory Commission for information only.

This reload cycle will contain f orty (40) f resh f uel assemblies and thirty-six (36) recycled assemblies from Core 20 (current Cycle) fabricated by Combustion Engineering (C-E).

The introduction of the new fuel is necessary in order to maintain sufficient reactivity for continued operation at full rated power.

l The following sections of this report describe the mechanical, thermal-hydraulic, physics, and safety analysis aspects of the Reload Cycle.

The approach to licensing the Reload Cycle presented in this report is 1

similar to that used since Core 11.

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' I 2.0 OPERATING HISIQRY OF CURRENT CYCLE I

1 The current operating cycle is Core 20.

Core 20 started producing power on January 15, 1989, and shut down on June 23, 1990.

During this period of operation, the core has operated normally in an essentially all-rods-out condition.

Both plant measured data and Reactor Physics calculations have l

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 entire cycle. No abnormalities which would adversely affect the core power distribution or reactivity were detected during the Current Cycle operation.

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I FIGURE 21 l

YANKEE CORE 20 COMPARISON OF MEASURED AND CALCULATED RELATIVE ASSEMBLY AVERAGE POWER DISTRIBUTION I

AT 6877 MWD /MTU l

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0.754 1.136 1.041 1.197 1.116 0.734 0.751 1.139 1.054 1.224 1.140 0.748 1

0.398 0.264 1.249 2.256 2.151 1.907 0.748 1.166 1.156 1.092 1.050 1.026 1.156 0.7. -

3*

0.744 1.161 1.151 1.093 1.060 1.048 1.169 0.755 0.535 0.429 0.433 0.092 0.952 2,144 1.125 0.667 0.638 1.135 1.061 1.185 1.202 1.140 1.181 1.144 1.144 0.624 4-0.631 1.130 1.052 1.175 1.192 1.130 1.176 1.141 1.142 0.623 1.097 0.441 0.848 0.844 0.832 0.877 0.423 0.262 0.175 0.160

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0.831 1.239 1.065 1.152 1.081 1.087 1.142 1.125 1.080 0.822

E 5-0.822 1.225 1.053 1.138 1.070 1.078 1.136 1.120 1.077 0.821 3

1.083 1.130 1.127 1.215 1.018 0.828 0.525 0.444 0.278 0.122 0.809 1.040 1.160 1.200 1.085 1.076 1.149 1.063 1.239 0.825 I6 0.799 1.028 1.146 1.185 1.076 1.069 1.141 1.055 1.230 0.823 1.236 1.154 1.207 1.250 0.829 0.651 0.696 0.753 0.726 0.242 0.624 1.149 1.158 1.161 1.145 1.167 1.123 1.062 1.142 0.639 I

.7 0.616 1.136 1.145 1.148 1.140 1.163 1.118 1.052 1.130 0.632 1.282 1.131 1.123 1.120 0.437 0.343 0.445 0.942 1.051 1.095 0.728 1.135 1.027 1.049 1.140 1.144 1.162 0.752 0.745 1.161 1.047 1.057 1.144 1.148 1.159 0.743 8

2.335 2.291

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-0.350 0.258 1.197 0.722 1.107 1.205 1.072 1.133 0.746 0.739 1.129 1.221 1.075 1.131 0.750 2.355

-1.987 1.328

-0.280' O.177

-0.536 l

0.615 0.809 0.810 0.627 MEASUREMENT 0.627 0.810 0.802 0.619 CALCULATION I

10-1.951

-0.124 0.988 1.276 PERCENT DIFFERENCE I

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lI 3.0 GENERAL DESCRifl103 g

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l Tigure 3-1 is a schematic of the YNPS core showing the Current Cycle j

loading pattern.

In this scheme, the inner region consists of 36 recycled assemblies and the outer region consists of 40 fresh assemblies. The Reload Cycle core will also utilize 40 fresh assemblies and 36 recycled assemblies.

The iresh assemblies fabricated by C-E have an initial enrichment of 3.9 w/o U-235 for 36 of the assemblies and 3.5 w/o U-235 for four of the assemblies.

2 The recycled assemblies also fabricated by C-E had 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 I

assemblies.

The Reload Cycle core average exposure for Beginning-Of-Cycle (B00) is 7,201 mwd /Mtu compared to 6,608 mwd /Mtu for the Current Cycle.

The Reload Cycle full power lifetime is estimated to be 14,000 mwd /Mtu compared to 14.100 mwd /Mtu for the Current Cycle.

I Figure 3-2 shows the recycled fuel assembly exposures for the Reload Cycle. Figure 3-3 shows the position that each recycled assembly occupied in the Current Cycle and the Reload Cycle loading pattern.

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FIGURE 31 i

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I 4.0 DEL MECHANICAL AND THEMAL DESIGN 4.1 Mechanical Desien I

Forty fresh assemblies manufactured by C-E will be inserted into the reactor for the Reload Cycle. This is the fourth batch of C-E fuel (Batch D) to be used at the YNPS. The mechanical design parameters are described in detail in Reference 4.1.

Table 4-1 lists the design parameters for the I

recycled (Batch C)*and fresh (Batch D) fuel assemblies.

4 i

Batch D is the first YNPS batch of fuel to incorporate a debris-resistant fuel design. This design is similar to the previous C-E Batch C fuel bundle assembly design with a few significant differences.

These differences include changes to both the fuel rod and the fuel assembly cage.

J l

The debris-resistant fuel rod design features includet 1.

Longer end cap.

2.

Longer overall length.

3.

Reduced volume plenum spring.

4.

Removal of the pellet spacer at the bottom of the fuel column.

5.

Spacer disk placement at the bottom of the fue; column instead of I

the top.

6.

Maintenance of active fuel length location.

The major debris-resistant fuel assembly design feature change is the l

lowering of the bottom Inconel grid. This repositioning, along with the l

longer fuel rod end cap allows the lower Inconel grid to act as a debris filter and capture the debris against a solid bar of Zircaloy, where debris fretting will not cause a breach in the fuel rod clad.

I WPP40/4 I

I Like the first three batches of C-E fuel, Batch D 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 assemblies with these design features are shown in Figure 4-1.

Figure 4-2 shows the assembly lattice locations for the solid Zircaloy rods and special guide bars.

I 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, 4.2, and 4.3.

I 4.2 Thermal Design I

The fuel thermal effects calculations were performed using the GAPEXX digital computer code (Reference 4.4).

The Reload Cycle calculation I

methodology is unchanged from previous reload analyses.

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

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

I The fuel rod types and power histories examined in detail include:

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I High-Power Fresh Pint fuel rod with the maximum average power for the fresh (reload) fuel.

High-Power Exposed Pint fuci rod with the maximum average power for the exposed (recycle) fuel.

I Figure 4-3 illustrates the effect of Linear Heat Generation Rate (LHGR) on calculated temperatures for Beginning-of-Cycle (B00) conditions. Figure I

4-4 illustrates the effect of LHGR on calculated temperatures for End-of-Cycle (E00) conditions. These resultant temperatures are similar to those which have been reported in previous reload analyses (References 4.5 and 4.6).

These results demonstrate that the BOC conditions yield the maximum predicted 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.

I 4.3 Qp n ating lxpirience The batch average burnup of the C-E fuel to be discharged during this refueling outage will be approximately 32,200 mwd /Mtu with a peak assembly burnup of apptoximately 35,700 mwd /Mtu. The batch of C-E fuel discharged from the reactor during the last refueling outage had a batch average burnup of 31,318 mwd /Mtu with a peak assembly burnup of 35.359 mwd /Mtu. The first three batches of C-E nuclear fuel have performed as expected.

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WPP40/4 I

TABLE 4-1 CORE 21 NOMINAL MECHANICAL DESIGN PKRAMETERS

]

Recycled Fuel Fresh Fuel

(

(36 assemblies)

(40 assemblies)

Fuel Pellets 002 Fuel Material (sintered pellets. UO2 Initial Enrichment, w/o U-235 3.9 (32 assemblies) 3.9 (36 assemblies) 3.7 ( 4 assemblies) 3.5 ( 4 assemblies)

Pellet Density, % theoretical 94.75 95.25 Pellet Diameter, inches 0.3105 0.3105 g

fuel _ Rad i

Active Length, inches 91.0 91.0 Overall Rod Length, inches 95.40 95.617 1

Upper Plenum Length, inches 1.54 1.942 Fuel Rod Pitch, inches 0.472 0.472 Diametral Gap (cold), inches 0.0065 0.0065 Fill Gas Helium Helium Fill Gas Pressure, psig 250.0 300.0 Cladding Material Zr-4 Zr-4 i

Outside Diameter, inches 0.365 0.365 Thickness, inches 0.024 0.024 Inside Diameter, inches 0.317 0.317 I

Guide Bars Material Zr-4 Zr-4 Number per Assembly 8

8 Length, inches 96.52 96.52 WPP40/4

I TABLE 4-1 CQEE 21 NOMINAL MECHANICAL DESIGN PARAMETERS i

Recycled Tuel Fresh fuel (36 maammh11egj (40 mainablieA.)

Fuel Annembly N aber of Assemblies 36 40 Fuel Rod Array 16x16 16x16 Fuel Rods per Assembly i

I Type A 231 231 Type B 230 230 Fuel Rod Axial Clearance, inches 1.217 1.000 Outside Dirrensions Assembly Cross Section, inches 7.116/7.578 7.116/7.578 I

x 7.116/7.578 x 7.116/7.578 Overall Length, inches 111.787 111.787 8

Spacer Grids Material Zr-4 Zr-4 1

Number per Assembly 6

6 Weight of Contained Uranium Type A Kg U 231.3 232.5 Type B. Kg U 230.3 231.6 I

I I

I I

lI E l WPP40/4 I

lE l

FIGURE 41

'g YANKEE CORE 21 i5 CORE LOCATIONS OF MODIFIED ASSEMBLIES l

'I iI i

1 2

3 4

ia A1

E (7) 5 6

7 8

9 10 ll A1 A1 j

(7)

(7)

I-11 12 13 14 15 16 17 18 A1 B1 (7)

(12) 19 20 21 22 23 24 25 26 27 28 A1 A2 B3 A2 (7)

(7)

(6)

(7)

'I 29 30 31 32 33 34 35 36 37 38 B2 B1 A1

'g (6)

(12)

(7) 39 40 41 42 43 44 45 46 47 48 1 m A1 Bi A2 A1 5

(7)

(12)

(7)

(7) 49 50 51 52 53 54 55 56 57

'58 l

B1 A1 A1 B1 A2 (12)

(7)

(7)

(12)

(7)

I 59 60 61 62 63 64 65 66 B1 A2 (12)

(7) l 67 68 69 70 71 72 B1 A2 A2 (12)

(7)

(7) 73 74 75 76 ASSEMBLY NUMBER B1 A2 B3 A2 ASSEMBLY TYPE

!g (12)

(7)

(6)

(7)

1. OF REPLACED RODS I

.m.

I

I FIGURE 4 2 I

YANKEE CORE 21 LATTICE LOCATIONS OF SOLID RODS

,A%ttfl A&&l4.9 TYP4 A 1 vAhn$$ AMS4.vTYPl A 3 i..

ui n ',

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lT. INSTRUMENTATION TUBE

08. QUIDE BAR No. NEW GUIDE BAR 1R. INERT ROD I

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I 5.0 NUci. EAR DESIGN I

5.1 Physles Characteristics Table 5-1 presents 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 conditions are presented i

for the cycle average exposures of $001 8,000; and 14,500 mwd /Mtu, respectively.

In Figures 5-1 through 5-3, the maximum unrodded radial pin power peak Txy. is 1.570 at 500 mwd /Mtu, 1.541 at 8,000 mwd /Mtu, and 1.446 at 14,500 mwd /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 T is xy 1.784.

Figure 5-5 shows the relr.tive axial power distribution as control rod Croup C is inserted for the assembly containing the peak F37 (Location 42 4

(D-6)).

I Table 5-1 presents various calculated reactivity coefficients for both BOC and EOFPL reactor conditions. The moderator temperature coefficient at I

BOC for the Reload Cycle is slightly more negative than the Current Cycle value primarily due to the lower 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 Reloed 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 I

Group C as its controlling rod group. Table 5-2 presents a summary of calculated control rod group worths for the Reload Cycle and comparable Current Cycle data. The reactivity allowances are listed for the Reload Cycle and the Current Cycle with the resulting excess shutdown margin also tabulated. The control rod group configuration remains unchanged from previous cores and is presented in Figure 5-6.

The calculated reactivity worth for each control group at hot-full-power conditions is given in Table 5-3 for both BOC and EOFPL reactor conditions.

The power dependent l

,(g g WPP40/4 I

I

1 E

control rod insertion limit curve for the Reload Cycle is unchanged f rom that j

used during the Current Cycle operation and is given in Figure 5-7.

Xenon redistribution effects are accounted for by the use of two factors:

the Xenon Redistribution Factor (XRF) and the Reduced Load Multiplier (Riti). The XRT accounts for changes in axial peaking due to I

control rod Group C motion in the full power operating band (80 to 90 inches withdrawn).

This multiplicative factor is defined as the ratio of the maximum value of Fg in the analytically derived top-peaked xenon-induced axial shape to the value of Fg of the nominal operating axial shape. The minimum value of this factor is unity.

This is consistent with the methodology used to derive the LHGR limits which were generated based on the worst top-peaked axial shapes. The top-peaked axial shapes bound both nominal and bottom-peaked shapes in terms of LHGR limits.

The Riti designates a power level at which the plant must remain at or below for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if control rod Group C is inserted I

below 80 inches withdrawn.

This time period allows for dampening of the xenon transient and, therefore, the power redistribution.

The factors to be used as 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 Codga YNPS core depletion calculations since Core 13 have been made using a I

PDQ/ HARMONY (References 5.1, 5.2) model with each fuel rod in an assembly represented explicitly. 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 SIMULATE (Reference 5.4) program has been used for the calculation I

of core average reactivity parameters such as moderator temperature coefficients, fuel temperature coefficients, boron worth and critical boron concentrations. Axial power distributions and xenon-induced redistribution,

l WPr40/4 I

I

I ef fects are also calculated with the SIMUIATE program.

Comparisons between plant measurements and the calculated SIMUIATE 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 assembly and fuel rod powers, plus the fast and thermal fluxes 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 Methods There have not been any analytical methodology changes initiated for the Reload Cycle analysis.

I I

E I

I i l I

E I

WPP40/4 I

I TABLE 5-1 SM1ARLDLCUPADiT AND AELOAD_CICLE_EUCLEAR CHAPACITAISTICS Current Reload Core 20 Core 21 Total Control Rod Worth, %Ap Hot Full Power BOC 11.61 11.31 Hot Full Power EOFPL 12.21 12.11 Dissolved Boron Dissolved Boron Content for Criticality BOC, All-Rods-Out, ppm I

68'F, No Xe, Peak Sm 2,107 2,045 l

Hot Zero Power, Nc Xe, Peak Sm 2,091 2,014 Hot Full Power, Eq. Xe and Sm 1,660 1.576 Dissolved Boron Content for Refueling, ppm 2,081 2,046 BOC, All Rods In, K =.93 Inverse Boron Worth, ppm /%Ap 66'F, B00 No Xe, Peak Sm 106 105 l I BOC, Hot Zero Power, No Xe, Peak Sm 137 136 BOC, Hot Full Power, Eq. Xe 137 135 1

l Reactivity Coefficients (All-Rods-Out)

Moderator Temperature Coefficients Ap/'F Hot Full Power, BOC

-0.80x10-4

-0.91x10-'

Hot Full Power, E0FPL, 0 ppm

-3.08x10-4

-3.11x10-4 Fuel Temperature Coefficients. Ap/'T Hot Full Power, BOC

-1.56x10-5 1,49x10-5 Hot Full Power EOFPL, 0 ppm

-1.75x10-5

-1.66x10-5 l

Moderator Void Coefficients, Ap/1 volo Eot Full Power, BOC

-0.61x10-3

-0.69x10-3 I

Hot Full Power, EOFPL, 0 ppm

-2.35x10-3

-2.37x10-3 l

l Moderator Pressure Coefficients, Ap/ psi Hot Full Power, BOC 0.86x10-6 0.97x10-6 Hot Full Power, EOFPL, O ppm 3.30x10-6 3,33xto-6

I WPP40/4 I

.~.

. + _ _, -,

<c-,

~ _.

I TABLE $-1

$1% MARY 0F CURRENT AND RELQAD CYCLE NUCLEAR CEARACTERISTICS us (Continued)

I Current Reload i

Core 20 Core 21 Ef f ective Delayed Neutron Fraction BOC, HZP

.006543

.006518 i

Effective Delayed Neutron Traction, E0TPL, RZP

.005411

.005404 Prompt Neutron Lifetime, psec, BOC, HZP 19.90 20.01 Prompt Neutron Lifetime, psec, E0FPL, HIP 23.15 23.27 l

!I j

I i i lI I

l I

I I

I WPP40/4 lI I

I f

TABLE $-2 CURRENT AND RELOAD CYCLE SHUTDOWN REOUIRD1ENTS. W

(!!FP)

Current Reload I

Core 20 Core 21 Rod Worth:

1.

Total Control Rod Worth 11.61 12.21 11.31 12.11 2.

Worth of Stuck Rod 3.31 3.44 3.04 3.27 3.

Total Worth less Stuck Rod (1-2) 8.30 8.77 8.27 8.84 4.

Total Worth Less 7.5% Uncertainty 7.68 8.11 7.65 8.18 (No. 3 x 0.925) l 5.

Total Allowances

• 6.48 7.23 6.49 7.20 6.

Excess Shutdown Margin (4-5) 1.20

.88 1.16 0.98 I

Allowances

• 5.1 Fuel Temperature Variation

.80

.90

.76

.85 I

5.2 Moderator Temperature Variation

.11

.67

.14

.67 5.3 Moderator Voids

.02

.07

.02

.07 5.4 operational Maneuvering Band

.05

.09

.07

.11 5.5 Shutdown Margin

.bjiD idQ LliD idD Total Allowances 6.48 7.23 6.49 7.20 I

I

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TABLE 5-3 I

COMPARISON OF CURRENT AND RELOAD CYCLE CONTROL ROD WORTHS

(!!TP)

Current Reload Core 20 Core 21 i

BOC EOFPL BOC EOFPL W

MP W

MD 1.77 1.86 1.80 1.87 Group C 1.23 1.49 1.28 1.53 Group A 2.40 2.49 2.34 2.49 Group B Group D

.6.21 6.37 5.89

.6.22 Total 11.61 12.21 11.31 12.11 Stuck Rod

,32._LL4 3.04

.3.27

)

Total Less Stuck Rod 8.30 8.77 8.27 8.84 iI I

8 I

1E I

I

-23 l3 m m 0,4

i FIGURE 51

I E YANKEE CORE 21 m

RELATIVE RADIAL POWER DISTRIBUTION 500 MWD /MTU, ALL RODS OUT l

.A.

8

.C-D-

E-F-

G-H-

J-K-

l g

• 1*

0.609 0.802 0.837 0.610

'2' O.739 1.137 1.144 1.242 1.159 0.724 l

! l

+3-l 0.742 1.160 1.078 1.151 1.009 1.040 1.165 0.722 i

i I 4-0.624 1.151 1.037 1.137 1.187 1.114 1.139 1.074 1.139 0.596 I

0.844 1.235 1.008 1.115 1.054 1.049 1.196 1.149 1.143 0.818 I

5-l J

6-0.816 1.146 1.149 1.180 1.049 1.045 1.117 1.009 1.247 0.827 3

.7 0.e23 1.,31 1.087 4.,43

,.124 1.189 1.ui 4.040 1.158 0.e30 I8 0.758 1.159 1.040 1.009 1.145 1.077 1.161 0.745-lI 0.764 1.151 1.241 1.145 1.131 0.740

.g.

'I 0.644 0.835 0.831 0.608

' g

.10-E I 1

FIGURE 5 2 1'

YANKEE CORE 21 RELATIVE RADIAL POWER DISTRIBUTION i

8000 MWD /MTU, ALL RODS OUT 1

A-B-

C-D-

E-F-

G-H.

J.

K-

~1*

0.639 0.811 0.833 0.632 l

2-0.756 1.115 1.116 1.189 1.118 0.742

. 3-0.756 1.132 1.074 1.153 1.027 1.041 1.137 0.742 1

l 4-0.644 1.111 1.039 1.148 1.202 1.143 1.153 1.068 1.111 0.625 i

5-0.839 1.185 1.025 1.139 1.098 1.096 1.213 1.149 1.109 0.822 I

6-0.825 1.119 1.149 1.190 1.090 1.092 1.147 1.025 1.190 0.822 l

7 0.649 1.106 1.081 1.152 1.148 1.202 1.155 1.040 1.113 0.646

-8 0.769 1.127 1.040 1.024 1.143 1.071 1.131 0.756 I

0.770 1.106 1.187 1.114 1.105 0.755

.g.

E 0.658 0.829 0.836 0.636 10-I I I

I FIGURE 5 5

!3 YANKEE CORE 21

!E RELATIVE RADIAL POWER DISTRIBUTION 14500 MWD /MTU, ALL RODS OUT I

i I

^-

g

• l-0.678 0.836 0.854 0.674

t 2-0.784 1.105 1.104 1.163 1.106 0.778 i

l 3-0.782 1.116 1.066 1.140 1.029 1.042 1.126 0.779 i

4-0.676 1.095 1.035 1.134 1.181 1.134 1.142 1.062 1.104 0.670 iI 0.850 1.154 1.024 1.126 1.091 1.093 1.195 1.137 1.097 0.847

, i

.5-

!l 6-0.841 1.102 1.132 1.167 1.084 1.088 1.139 1.027 1.163 0.842 7

0.680 1.092 1.069 1.136 1.134 1.181 1.142 1.039 1.098 0.681 I

8 0.788 1.107 1.034 1.022 1.127 1.062 1.117 0.783 I

0.787 1.086 1.155 1.097 1.093 0.783

.g.

0.684 0.840 0.852 0.673 10-I I

-u.

I

?

FIGURE 5 4

)

YANKEE CORE 21

!g RELATIVE RADIAL POWER DISTRIBUTION 500 MWD /MTU, GROUP C INSERTED i

i

.A.

.g.

.C.

D-E-

F-G-

H-J-

K-I

• 1*

0.615 0.735 0.760 0.606

!; I

.2-0.827 1.161 0.891 0.951 1.159 0.798 3*

0.828 1.307 1.151 0.979 0.838 1.093 1.297 0.798 i

4-0.629 1.170 1.099 1.262 1.302 1.216 1.253 1.132 1.148 0.595 I

.5-0.782 0.963 0.841 1.223 1.234 1.225 1.302 0.968 0.882 0.744 i l 6-0.763 0.904 0.983 1.298 1.228 1.220 1,220 0.839 0.963 0.753

.I 0.636 1.169 1.165 1.272 1.234 1.298 1.258 1.099 1.166 0.628 I

.7 l

8 0.855 1.314 1.110 0.845 0.970 1.138 1.299 0.825 0.857 1.179 0.975 0.894 1.153 0.823

.g.

I 0.653 0.777 0.770 0.614 10-I I

I 2,.

I

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09 m

TTT NNN T

EEE e

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\\\\'

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O

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1 E

2G H

5ES 0 C 5

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OS

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IE m

KR 0 H I

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4 E g

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A I

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/

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?

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/

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M 0

~

e

~

4 m

0 5

0 5

0 2

1 0

0

-gy3a gm m

m

.Y m

_ _ _ ~ _. _. - - -

FIGURE 5 6 YANKEE CORE 21 l-CONTROL ROD GROUP IDENTIFICATION I

I ASSEMBLY NUMBER 1

2 3

4 i

6 6

7 8

9 10 l

_I F

D C

11 12 13 14 15 16 17 18 4

B B

D 19 20 21 22 23 24 25 26 27 28 l

D B

A B

29 30 31 32 33

.4 36 36 37 38 C

A-A C

lg 39 40 41 42 43 44 45 40 47 48 B

A B

D 49

$0 61 52 63 54 65 56 67 58 l

D B

B 59 60 61 62 63 64 66 B6 C

D 67 68 69 70 71 72 I

D-I 73 74 76 76

.mm I

i FIGURE 5 7 YANKEE CORE 21 ROD INSERTION LIMIT VERSUS POWER LEVEL g

I l

400 I

90 -

_I I

80 -

et I

70 -

E I

g i

F 60 -

w l

6 g 50 -

s N

I 8

8 40 -

l!

8 e

I 30 -

Y 10 -

g g,

g i ig

'l l

i' O

10 20 30 40 50 60

' 70 80 90 Q

GROUP C POSITION (INCHES WITHDRAWN)

I.

g g

g 8

1 g

6 1

g g

4 1

R 2

)

1 U

O T

T M

C A

D F

W c

2N G(

1 e

8E O 0 P 1

U I

5 RT N

OU R

E CB U

R B

I UER E

e GET G

K S

, '8 A

I I

FND R

E A. E V

a

' R A

N EL O

C N

Y E

,6 C

e X

e

,4

,2 W

W 0

~~

G 4

1 5

0 59 0

1 1

W 1

0 Oo2 5 5sNeOxZo6x

,e' M

l M

-M M

-M M

a m.m_

g

- g g

g g

g i

l

'i l

l I

FIGURE 5-9 L

YANKEE CORE 21 REDUCED LOAD MULTIPLIER i

1.0 p

0.95 -

1A3 5'

s J

I 3

~

g I

W 2

w O 0.9 -

l-O s

O til I,

3o I

SAD i

4 E

f l

0.85 -

i, 1

1

\\

~

t i

I 1

i l

i OS j

0 2

4 6

8 10 12 14-16 18 I

CYCLE AVERAGE BURNUP (GWDMTU) 4 ei 1

a 4

'r

- --w

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

The thermal-hydraulic analysis of the Reload Cycle was performed by adjusting code input to reflect the Reload Cycle power distributions and thermal-hydraulic characteristics. Table 6-1 contains the pertinent nominal I

and 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 summarizes both the predicted hot spot and hot channel factors for the Reload Cycle and the design hot spot and hot channel factors.

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

Predicted hot channel factors are based on power distributions at the limiting exposure I

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 centerline melt limits. The design DNBR for the Reload Cycle is slightly lower than the previous analysis DNBR, 3.01 versus 3.02 at full power 4-loop operation. This difference is due primarily to the insertion of 247 solid Zircaloy rods and special guide bars in the Reload Cycle versus 206 in the I

previous analysis. The resulting increase in core average linear heat generation rate leads to a slightly reduced value of DNBR.

The design of the new C-E fuel in the Reload Cycle has been modified slightly from the previous cycle. The new fuel design includes a bottom Inconel spacer grid which has been lowered and longer end caps on the bottom of the fuel rods.

This slightly increases the hydraulic resistance of the new fuel. The increased hydraulic resistance is reflected in all of the thermal hydraulic analyses for the Relord Cycle. The increased resistance was found I

to have a negligible impact on JNB.

I WPP40/4

I 1

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

These 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 operation. As required in Reference 6.1, a 34% DNBR credit is needed to offset the very conservatively applied full-closure rod bow penalty.

Generic I

credit of 13.2% DNBR margin, based on conservative thermal-hydraulic analytical methods, was accepted in Reference 6.1.

The most limiting j

anticipated transient is the 2-out-of-4 pump loss of flow.

Based on design conditions, 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 I

that rod bow resulting in less than 50% channel closure has no adverse impact on the predicted DNBR.

The maximum predicted channel closure for the fresh fuel in the Reload Cycle is 50.1%, requiring a penalty of 0.1% DNBR margin.

For the recycled fuel, the maximum closure is 63.2%, requiring a penalty of 9.0% DNBR margin.

These values conservatively assume that the bowing penalty is a linear function of channel closure, as indicated in Reference 6.1, with a 0% penalty at 50% closure and a 34% penalty at full closure. Therefore, application of the 34% full closure penalty to' Reload Cycle operation is very conservative.

I I

I I

WPP40/4 I

TABLE 6-1 IllERMAL-HYDRAULIC DATA SHEET FOR YANKEE CORE 21 DURING 4-LOOP OPERATION I

General Characteristics Nominal Design l

I Total Core Power, MWt 600 618 1

Fraction of Heat, Generated in Fuel

.973

.973 Main Coolant Pressure, psig 2,000 1,925 Main Coolant Inlet Temperature,

's 520 524 Reactor Vessel Outlet Temperature, 'F 563 568 Average Core Outlet Temperature.'F_

567 572 Average Core Enthalpy Rise, Btu /lb 58.5 60.3

)

6 6

38.3x10 j

Total Coolant Flow Rate, Ib/hr 38.3x10 6

6 Heat Transfer Flow Rate, Ib/hr 35.0x10 35.0x10 1

I 6

2 6

2.296x10 Average M6ss Velocity, lb/hr-ft 2.296x10 Average Coolant Velocity in Core, ft/see 13.5 13.6 Core Pressure Drop, psi 13.3 13.3 l

Reactor Vessel Pressure Drop, psi 31.4 31.5 2

Average Rod Heat Flux, Btu /hr-ft 159,207 163,983 2

Average Film Coefficient, Btu /hr-ft

.F 5.662 5,683 Average Film Temperature Difference, 'F 28.1 28.9 Average Linear Rod Pover, kW/ft 4.46 4.59 Average Specific Power, kW/kgU 34.7 35.7 I

Power Density, kW/ liter 90.1 92.8 Hydraulic Diameter, in 0.412 0.412 2

Assembly Heat Transfer Area, ft 164.7 164.7 2

daximum Heat Flux, 3tu/hr-ft 380,663 452,593 M aimum Linear Rod Power, kW/ft 10.66 12.67 Maximum Clad Surface Temperature, 'F 629 637 Maximum Centerline Pellet Temperature, 'F 2,833 3,265 Hot Channel' Outlet Temperature, 'F 589 606' Minimum W-3 DNBR 4.32 3.01 l

i I WPP40/4 I

]

TABLE 6-2 l

THERMAL-HYDRAULIC DATA SHEET FOR COMPARISON OF CORE 20 AND CORE 21 DESIGN CHARACTERISTICS DURING 4-LOOP OPERATION I

General ChaIActeristics Core 21 Core 20 Total Core Power, MWt 618 618 Fraction of Heat Generated in Fuel-0.973 0.973 Main Coolant Pressure, psig 1,925 1,925 I

Main Coolant Inlet Temperature. 'F 524 524 Reactor Vessel Outlet Temperature, 'F 568 568 Average Core Outlet Temperature. 'F 572 572 Average Core Enthalpy Rise, Bt g '1b 60.3 60.3 6

6 Total Coolant Flow Rate, Ib/hr 38.3x10 38.3x10 Heat Transfer Flow Rate, Ib/hr 35.0x106 35.0x106

)

Average Mass Velocity, Ib/hr-ft 2.296x106 2.296x10 2

6 Average Coolant Velocity in Core, ft/sec 13.6 13.6 Core Pressure Drop, psi 13.3 13.2 I

Reactor Vessel Pressure Drop, psi 31.5 31.4 2

Average Rod Heat Flux, Btu /hr-ft 163,983 163,610 2

Assembly Heat Transfer Area, ft 164.7 165.1 2

Average Film Coefficient, Btu /hr-ft

'F 5,683 5,645 Average Film Temperature Difference, 'F 28.9 29.0 Average Linear Rod Power, kW/ft 4.59 4.58 Average Specific Power, kW/kgU-35.7 35.7 Power Density, kW/ liter 92.8 92.8 Nominal Channel Hydraulic Diameter, in 0.412 0.412 I

Maximum Heat Flux, Btu /hr-ft2 452,593 451,564 Maximum Linear Rod Power, kW/ft 12.67 12.6 Maximum Clad Surface Temperature, 'F 637 637 Maximum Centerline Pellet Temperature, 'F 3,265 3,290 Hot Channel Outlet. Temperature, 'F 606 606 Minimum W-3 DNBR 3.01 3.02 I

~

WPP40/4 N

I TABLE 6-3 I

SUMMARY

OF HOT SPOT AND HOT CHANNEL FACTDAS FOR YANKEE CORE 21 VERSUS DESIGN I

Nominal Design RCAt Flux Factora I

Nuclear Heat Flux Factor 2.21 2.59 Fuel Densification Factor 1.03 1.03 Engineering Heat Flux Factor 1.04 1.04 Total Heat Flux Factor 2.39 2.76 Enthalpv Rise Factors Statistical Enthalpy Rise Factor 1.08 1.08 Lower Plenum Factor 1.05 1.05 Nuclear Enthalpy Rise Factor 1.55 1.80 I

I l

' I

' I I

g I

.I I,

WPP40/4 I

i i>.

TABLE 6-4 NOMINAL HOT CRANNEL DNBR, FAH, MD FO AS FUNCTIONS OF GROUP C POSITION FOR CORE 21 Group C

-i Eosition *(inches inserted)

Fjlgg F_g pg l

0:

1.494 2.20 4.55 7.5 1.488 2.22 4.50 15.0 1.479 2.29 4.44 22.5 1.468 2.39 4.32 I

I I

.I

-g I

I I

I

• Power Dependent Insertion Limits Restrict Group C Insertion to 10 inches at Full Power.

l WPP40/4 I

I l

I FIGURE 6-1 REACTOR CORE SAFETY LIMIT-I ALL LOOPS IN OPERATION I

i 660 l

640

- I I

~_

~

~

620 w.

i I

x_

y b

MAIN COOLANT I

c.

600 SYSTEM PRESSURE g

g I

y

-~___

3 580 2600 psia O

I 2400 psia g

y 5

x 2200 psia I

560 3

c t

h 2000 psia e

540 g

I-1800 psia 520 I

I 1600 psia 500 70 80 90 100 110 120 130 I

Indicated Recctor Power, Percent I

I

-se-1

7.0 ACCIDENT ANA1.YSIS I

7.1 Introduction I

r The safety analysis of the Reload Cycle is presented in this section.

Each transient in the following subsect' ions is compared with the most recent I

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 the referenes analysis for the Reload Cycle review, and are indicated as i

such.

Each event is evaluated considering the most limiting time in core life. This evaluation includes conditions during constdown beyond the 50FPL.

7.1.1 Initini operatine Conditions Table 7-1 provides the initial operating conditions that apply to most of the transients analyzed. Any deviations from the values indicated are nowd in the discussion of the specific transient.

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

These minor differences are the following:

I 4

1.

The core average LHGR is slightly higher for the Reload Cycle than for the reference analysis. This difference resulted from the addition of 247 solid Zircaloy rods and special guide bars to selected assemblies.

2.

The minimum DNB ratio at design conditions for the Reload Cycle is marginally lower than for the reference analysis. The impact of.

this minor reduction in design DNBR will be addressed in the review I

of each appropriate transient.

7.1.2 Reactor Trip Setpoints and Instrumentation Delava Table 7-2 presents the reactor trip setpoints and instrumentation delays applied in the transient analysis.

I WPP40/4

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

7.2 Control kod Withdrawal Incident For the reference analysis (FSAR), a bounding analysis was performed I

with the following essumed initial conditions:

1.

Design peaking factors were used even though lower peaking factors exist during the incident; 2.

Core power was assumed to be at the overpower trip setpoint; I

3.

Coolant pressure was assumed to be at the lower end of the operating band; I

4.

Coolant temperatures were assumed to be at the values consistent with steady-state operation at the overpower trip setpoint.

For the Reload Cycle, these conservatively assumed conditions remain bounding. However, the steady-state design DNBR is marginally lower for the Reload Cycle than was calculated for the reference analysis. The reference analysis shows that the minimum DNBR for this event remains significantly above 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.

7.3 Boron Dilution Incident 7.3.1 Introduction Due to the nature of key reactivity parameters which change slightly from cycle to cycle, a complete evaluation of the boron dilution event is performed for each Reload Cycle to assure acceptance criteria are met for each I

operational 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 WPP40/4 I

i

Reference 7.2.

This analysis reviewed each mode of operation allowed by Technical Specifications. An additional analysis presented in the FSAR (Reference 7.1) investigated the minimum DNBR which could occur for a boron dilution event in Mode 1.

For the Reload Cycle, each mode of operation has been analyzed using appropriate initial and critical boron concentrations.

The initial boron concentration in each mode is that needed to provide the minimum shutdown margin required by Technical Specifications.

In all cases, the dilution is assumed to proceed at the maximum capacity of the charging pumps (100 spm).

The results are described below for each operational mode.

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

7.3.2 Analysis and Results 7.3.2.1 Time to_ Loss of Shutdown Margin In Modes 1 through 5 there must be at least 15 minutes available for the operator to terminate the dilution prior to a complete loss of shutdown margin.

In Mode 6, at least 30 minutes must be available.

The boron dilution analysis for the Reload Cycle used the same approach as the FSAR and previous reload analyses. The analysis used the minimum shutdown margin and main coolant volume allowed by Technical Specifications for each mode.

The significant parameters for the limiting case in each operational mode are presented in Tables 7-3 through 7-7.

As shown, the Reload Cycle remains above the minimum allowable time criteria in all cases.

In both the reference and Reload Cycle analyses, two cases were examined for Mode 6 conditions. The first case assumes that all four loops are isolated with the reactor vessel head drained in preparation for head removal or installation.

The second case represents the majority of the time spent in Mode 6.

For the first case, presented in Table 7-6, a shutdown margin of 7.5% op is required to provide at least 30 minutes for the operator to terminate the dilution. WPP40/4

I The first case is applicable only when the reactor vessel head is about 4

to be removed or installed.

The 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 required by Technical Specifications. A more realistic way, therefore, to examine the Mode 6 boron dilution event was used I

as an additional case in the FSAR analysis, and is repeated for the Reload Cycle.

In this additional case, the active dilution volume assumes that all 4 I

i loops are isolated, and the shield tank cavity is filled to 32 feet above the top of the fuel assemblies with one-half of the shield tank volume contributing to the active dilution volume.

For this case, a 5%Ap shutdown margin is adequate to provide the required 30 minutes for the operator to terminate the dilution. The significant parameters for this case are shown in Table 7-7.

I In addition to the analysis presented above, boron dilution during reduced level operation in Modes 5 and 6 was evaluated for the Reload Cycle.

The significant parameters are provided in Table 7-86 Reduced level operation is the condition that exists whenever the reactor vessel water level is lower than two feet below the reactor vessel head flange with irradiated fuel in the vessel.

The volume used in the analysis assumes that all four main coolant loops are isolated with the main coolant water level at the centerline of the loop piping. This corresponds to the elevation of the shutdown cooling suction connection to the Loop 4 hot leg. During reduced level operation in I

Mode 6, all control rods must be fully inserted to ensure that at least 30 minutes is available for the operator to terminate the dilution. All other boron dilution cases, including reduced level operation in Mode 5, satisfy the minimum time criteria assuming all rods in with the most reactive rod stuck in the fully withdrawn position.

Since close surveillance is required for a feed and bleed operation performed at the maximum capacity of the charging pumps (100 gpm), the possibility of inadvertently feeding unborated water at this magnitude is I

extremely remote. However, assuming this unlikely situation did occur, the operator would have several alarms to alert him to an inadvertent dilution. WPP40/4 I

i In Modes 1 through 5, the high level, temperature, and pressure alarms on the Low Pressure Surge Tank (LPST), would alert the operator.

In Mode 6, the operator is provided with indication of a possible boron dilution via the audible count rate signal, which would increase.

Also, during power operation with the Rod Control System in the automatic mode the selected control rod group would insert to offset the temperature increase resulting from the core power / steam flow mismatch.

In I

this situation, the operator would have additional indication of the dilution from the high average temperature alarm and audible indication of the control rod motion from the Containment Sound Monitoring System.

For any boron dilution, the event can be terminated quickly and easily from the Control Room by isolating the charging line, tripping the charging pumps, or isolating the source of demineralized water.

7.3.2.2 Maximum Reactivity Insertion Rate I

In addition to the time to loss of shutdown margin analysis, the maximum reactivity insertion rate was investigated for the FSAR and has been reviewed for the Reload Cycle. The analysis used a higher initial boron concentration than for the time to loss of shutdown margin analysis since higher concentrations result in higher reactivity insertion rates.

For the Reload Cycle, the analysis assumed the minimum water volume, the minimum inverse boron worth, and the maximum charging flow rate for any plant operational mode, including reduced level operation. This maximizes the reactivity insertion rate.

I For the Reload Cycle, the boron concentration must be in excess of 5,000 ppm to result in the maximum reactivity insertion rate (1.5 x 10~4 op/sec) allowed by Technical Specification 3.1.1.3.

Since the maximum concentration expected during the cycle is less than 5,000 ppm and the control rod withdrawal analysis is based on a reactivity insertion rate in excess of the current Technical Specification limit, the plant response to a Mode 1 or 2 boron dilution is bounded by the control rod withdrawal event for the Reload Cycle.

I WPt ' 0/4 I

I 7.3.2.3 Enihtte to_Borattfrior to_ Con 1sinEn Because of the large negative temperature coefficient of reactivity at the end,of cycle, any decrease in main coolant system temperature increases the core reactivity state. Consequently, during a main coolant system cooldown adequate shutdown margin must be maintained by ensuring that adequate boron concentration and/or control rod worth is available.

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

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

I b) The reactor is initially at an average temperature of 536'F.

I c) The shutdown margin at 536'F is 5.5Mp, the minimum required by the Technical Specifications for Modes 1 and 2.

d) The main coolant system temperature is reduced at the rate of 100'F/hr, the maximum rate permitted by Technical Specifications.

For a complete loss of shutdown margin, a cooldown to approximately 376*F would be required, which would take more than one hour. This is ample time for the operator to diagnose the condition and take necessary corrective I

action.

I I

I I

WPP40/4 I

I 7.3.3 Conclusions The probability of erroneous dilution is considered very small because of the equipment, controls, and administrative procedures provided for boron dilution activities.

However, in the unlikely event that an unintentional dilution of boron in the main coolant system occurs, nwerous alarms and indications are available to alert the operator of the condition.

These alarms include high level, temperature and pressure on the Low Pressure Surge I

Tank, 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. Therefore, for boron dilutions initiated during any operational mode, adequate time exists for the operator to determine the cause of the dilution and take corrective action before a complete loss of shutdown margin occurs.

7.4 Control Rod Drop Incident The reference analysis for the Control Rod Drop event was performed in support of the Current Cycle (Reference 7.2).

The reference analysis assumed steady-state operation at a conservatively determined limiting end point for the transient. The following assumptions were made in the analysis 1.

The highest calculated radial peaking factors with uncertainties for any dropped rod were used; I

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; I

4.

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

I WPP40/4 I

l l

I Table 7-9 summarizes the significant parameters for this event.

The information provided in Table 7-9 demonstrates that the Reload Cycle is conservatively assumed to be at essentially the same bounding operating conditions as the reference analysis, but with lower power peaking. While the Reload Cycle has a slightly lower design DNBR than the reference analysis at full power steady-state conditions, the minimum DNBR for this event remains significantly above 1.3.

Accounting for the marginally lower design DNBR for the Reload Cycle, the fuel performance remains well within the acceptable B

limits for this event.

7.5 h olated Laop 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 I

For the reference analysis, performed in support of the FSAR (Reference 7.1), a bounding analysis was performed using the design Technical Specification velue for the Moderator Temperature Coefficient (MTC).

The reference 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.

The reference analysis demonstrated that the peak primary and secondary system pressures are less than 110% of design pressures.

It also demonstrated that both DNBR and fuel centerline temperatures improve during the transient relative to the initial steady-state values.

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 I

WPP40/4 I

I demonstrated the minor impact of core physics parameters on the loss of load transient.

Since the DNBR improves during this event, the marginally lower design DNBR for the Reload Cycle does not adversely affect the reference analysis results. Also, 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 Less 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 requir9 ment 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 I

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) auxiliary feedwater flow sufficient to assure adequate steam generator liquid inventory for primary system cooldown, de::ay heat removal, and main coolant pump heat removal for the entire ccurse of the event.

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

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.

I WPP40/4 I

.l

I The analysis assumes a least negative MTC and most negative Fuel Temperature Coefficient (FTC). Table 7-10 shows that both of these parameters are more favorable for the Reload Cycle than for the reference analysis.

While the total scram worth is marginally lower for the Reload Cycle than for the reference analysis, the scram worth versus rod insertion is more favorable for the Reload Cycle.

That is, more negative reactivity is inserted earlier I

in the scram for the Reload Cycle.

Thus, the power would decrease slightly faster following the scram, improving the fuel performance.

The fuel pin failures for the 4-of-4 pump loss of flow in the Reload Cycle were evaluated based on the plant response from the reference analysis.

The total amount of fuel pin failures predicted is well below 10.0%.

Accounting for the slightly lower design DNBR for the Reload Cycle, the amount of fuel pin failures remains below 10%.

Thus, the radiological evaluation for rod ejection events, which assumes 10% fuel failures, bounds the results of I

the complete loss of flow event.

It is important to note that the 4-of-4 pump loss of flow transient is considered to be 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 severity of this transient.

Anticipated loss 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, the 2-of-4 loss of flow is the limiting case with a minimum DNBR of 2.01.

Accounting for the marginal reduction in design DNBR for the Rs. load Cycle, the results of both cases remain well within the acceptable limits for this event.

I I WPP40/4 1 I

I 7.9 Control Rod Eiection Accident I

The most recent control rod ejection acci@nt 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 Cycle is provided in Tables 7-11 and 7-12.

Both the zero power and full power rod ejection events required I

reanalysis for the Reload Oycle.

The zero power rod ejection required reanalysis since the p/B f raction 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. For rod ejections initiated from less than 15 MWe, the reactor was assumed to trip when core power increased to 36.1% (35% power plus 3% uncertainty). The results of both analyses show the average enthalpy of the hottest fuel pin is less than 200 cal /gm, and the I

centerline enthalpy is below 250 cal /gm.

Therefore, no fuel cladding damage or fuel centerline melting 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 support of the Current Cycle (Reference 7.2) and Proposed Change 226 (Reference 7.5).

The proposed change was approved via the issuance of Amendment 127 to the facility operating license (Reference 7.6).

The I

amendment replaced the cold leg temperature limit with a limit on average MCS temperature. This results in higher allowable MCS temperatures during part power operation.

The moderator reactivity feedback for the Reload Cycle was found to be more limiting for the steam line break than the values generated for the reference analysis. The scram rod worth, Doppler reactivity feedback, and inverse boron worth for the Reload Cycle have improved over the values from the reference analysis. A ret.nalysis of the main steam line break was I

performed for the Reload Cycle to show the net effect of the change in reactivity. The reanalysis included full power, zero power, and Mode 3 cases. WPP40/4 I

I as well as the most limiting part power conditions.

The results for the' limiting case show a minimum suberiticality of -0.233%Ap.

Therefore, there is

'no return to power following the steam line break.

This precludes DNB and fuel cladding damage.

7.11 Steam Generator Tube Rupture Incident The reference analysis for this event, Core-11, demonstrated that the I

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.

7.12 IIAD11gaLAnaly31s S"mmary I

The results of the transient analysis review of the Reload Cycle are shown in Table 7-13.

Provided are the criteria for each incident, and the results for the reference analysis as well as the Reload Cycle analysis.

These results providt 3ssurance of continued safe. operation of the YNPS.

I I

I

I I

I I

E WPP40/4 1 E

l I-TABLE 7-1

~

FOUR-LOOP INITIAL OPERATING CONDITIONS Reference Analysis Parameter Reload Core FSAR Reactor Power, MWt 600 + 18 600 + 18 l

Core Inlet Temperature. 'F 520 + 4 520 + 4 Main Coolant Pressure, psia 2,015 1 75*

2,015 A 75*

Minimum Reactor Coolant Flow.'

35.0 35.0 106 lb/br Axial lleat Flux Profile Cosine Cosini I

Total lieat Flux Factor 2.76 2.76 Nuclear Enthalpy Rise-1.80 1.80 I

Factor I

II I

I I

l.

l 1

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

E WPP40/4 E

4

l I

TABLE 7-2 REACTOR TRIP SETPOINTS AND INSTRUMENTATION DELAYS Irip Functions Setpoint Delav Time (see)

High Startup Rate 5.2 decades / min.*

0.3 High Neutron Flux 112%

0.4 I

Pump Current Deviations high/ low current 0.6 on two pumps High Pressurizer Water Level 209 inches

• 0.65 I

Low Main Coolant Pressure 1,735 psig 0.6 I.

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 2,350 psig 2.0 High Neutron Flux (low setpoint) 36.1%

0.4 I

I I

• Not specifically credited in analysis.

I

. WPP40/4 I

IABLE_1-3 MODES 1 AND 2 BORON DILUTION

• EAIARIC11r Raioad Ovels ISAB 3

Minimum Coolant Volume, f t **

2,400 2.400 Limiting Initial Boron 1,979 1,858 Concentration, ppm Minimwn Required Shutdown 5.5 5.5 Margin (%Ap) 58.0-Time to Loss of Shutdown 52.6

~

Margin, Minutes I

I I

I I

• These operating modes are defined as follows:

Mode 1

2 Deccription Power Operation Startup Reactivity K,

1 99 2 99 Power Level, %ggRated

>2 12 Coolant Temperature,*F 2330 1330

• • Assumes one main coolant loop isolated.

( WPP40/4 l

1

n

?

TABLE 7-4

B MODE 3 BORON DILUTION *

=

FArameter Reload Cycle ESAE 3

Minimum Coolant Volume, ft **

1,700 1,700 Limiting Initial Boron 1,923 1,770 Concentration, ppm Minimum Required Shutdown 5.0 5.5 Margin (%Ap)

Time to Loss of Shutdown 34.4 36.3 Margin, Minutes I

=

I a

f i

I I

g I

• Mode 3 is defined as follows:

Description Hot Standby

- I Reactivity, K,gg

<.99 Power Level, % Rated 0

Coolant Temperature,'F 1330

=

• • Assumes three main coolant loops isolated.

I WPP40/4

I TABLE 7-5 MODES 4 AND $ BORON DILUTION

• I ERIAmtitI Reload Ovele ISAR 3

Minimum Coolant Volume, it **

1,276 1,276 Limiting Initial Boron 2,039 1,932 Concentration, ppm Minimum Required Shu'.down 5.0 5.0 Margin (1Ap)

Time to Loss of Shutdown 24.3 26.2 Margin, Minutes I

I I

I I

• These operating modes are defined as follows:

Mode 4

5 q

Description llot Shutdown Cold Shutdown Reactivity, K,gg

<.96

<.96 Power Leve), 7. Rated 0

0 1

Coolant Temperature,*F 200 < T < 330 1200 I

i

• • Assumes four main coolant loops isolated, with provisions for upper reactor vessel head draining.

WPP40/4 I

['

I TABLE 7-6 MODE 6 BORON DILVIJON*

ALL_ LOOPS = ISOLATED.

MINIMUM WATER LEVIL FAramt11I Rgload Cycl 3

[$&R 3

1,276 1.276 Minimum Coole.nt Volume, ft I

i Limiting Initial Boron 2,298 2,137 Concentration, ppm Minimum Required Shutdown **

7.5 7.5 Margin (%$p)

Time to Loss of Shutdown 34.2 34.5 Margin, Minutes I

1

• Mode 6 is defined as follows:

1%scription Refueling Reactivity, K gr

.i.95 PowerLevel,$ Rated 0

Coolant Temperature,'F

.(140

• • Required by Technical Specification 3.9.1.

WPP40/4

I IABLE 7-7 I

MODE 6 BORON DILUTION

• WATER VOLtME INCLUDING EHIFI.D TANK CAVITY j

E& tame _ttr Reload Cvele TEAR 3

l Minituum Coolant Volume, f t **

8,630 8,650 i

Limiting Initial Boron 2,039 1,932 Concentration, ppm Assumed Shutdown Margin (%Ap) 5.0 5.0 1

Time to Loss of Shutdown 165.0 177.0 Margin, Minutes 4

1 i

I I

1 I

i I

• Mode 6 is defined as follows:

Description Refueling I

Reactivity, K,gt 1 95 Power Level, % Rated 0

Coolant Temperature,'F 1140

• • Assumeo 4 loops isolated and shield tank cavity filled to 32 feet above fuel

.:ssemblies, with 1/2 of the shield tank volume contributing to the active

{

diluu on volume. WPP40/4

I I

TABLE 7-8 EDRON_D1LUIIOR_DURING REDUCED LEVEL OPE 14TIDN*

Paramti.tr Mode 5 Mode 6 3

Mininum Coolant Volume Ft **

1,012 1,012 Limiting Initial Boron 2,039 2,013+

Concentration, ppm Minimum Required Shutdown 5.0 7.5 Margin (% Ap)

Time to Loss of Shutdow-19.3 31.3 Margin, Minuteo 1

I I

I Reduced level operation is the condition that exists whenever the reactor vessel water level is lower than two feet below the reactor vessel head flange, with irradiated fuel in the vessel.

Assumes reactor vessel water level is at the centerline of the loop riping.

+

Assumes all control rods fully inserted.

I 1 WPP40/4 I

I IABLE 7-9 CONTROL ROD DROP INCIDENT PAMMETERE I

f&IAmtitI Reload Ovele Reference Analysis Core Power, MWt 618 618 Core Inlet Temperature

'F 524 524 Main Coolant Pressure, psia 1,750 1,750 Maximum Linear Heat Rate 14.20 14.62 I

(post-drop, with uncertainties),

kW/ft Maximum Tuel Centerline 3,581 3,699 I

Temperature, 'F Minimum W-3 DNB Ratio

>2.04 2.04 I

I I

I I

I I WPP40/4 I

- -. ~. -

IABLE 7-10 I

LOSS OF. CQQ1 ANT FLOW 'ARAMETER$

j l

Parameter Eticad Ovele Rtierence Analvsla 4

i Moderator Temperature J

l Coef{op/'F) 1cient (10'

-0.41 0.0

~ l Fuel Temperature 3

Coef{Ap/'F) icient m

(10~

-2.08

-2.19 Scram Rod Worth With Group C at PDIL (1Ap) 7.60 7.61 I

I I

I I

I I

I I

I WPP40/4 I

TABLE 7-11 i

H2P ROD Ej.ECTION ACCIDENT PARAMETERE I

Earameter Reloat Cycle Rafarance Analvals l

10f EDTPL Moderator Temperature Coef{Ap/'F) icient (10~

+0.23*

-2.16

+0.27*

Coefficient DopplegAp/'F)

(10~

-1.16

-1.30

-1.10 Ejected Rod Worth (top)

.83

.86 0.93 I

Delayed Neutron Fraction (D)

.005866

.004874

.005490 p/3 1.415 1.756 1.694 F Folloung Rod qEjection 4.46 4.25 4.22 I

'I I

I I

llI l'

• This includes an uncertainty of 0.5 x 10-44p/'F.

Technical Specifications require MTC to be negative under Hot Zero Power conditions.

l I '

WPP40/4 I

I TABLE 7-12

}((f,_RQD EJECTION ACCIDENT PARAMETERS j

faramgity Reload Ovele Reference Analysis

=

om

, I Moderator Temperature Coef{Ap/'F) icient (10'

-0.41

-2.61 0.0*

Dopple Coefficient (10' Ap/'F)

-1.08

-1.15

-0.766 i

Ejected Rod Worth

(%Ap)

.25

.36

.50 i

Dolayed Neutron Fraction (B)

.005866

.004874

.005743 p/B

.418

.728

.871 Fg Following Rod Ejection 3.42 3.11 3.17 I

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TechnicalgpecificationsrequireMTCtobemorenegativethan

-0.2 x 10 Ap/'F at Hot Full Power conditions.

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W IABLE 7-13 TANKEE_CQREll SAFEIT_ ANALYSIS s

SIM3ARLRF_RESIILTS 4

l Inridenl Section Criter_ia Refirenst_ Analysis Cort.11 Control Rod 7.2 MDNBR > 1.30 MDNBR > 2.00 MDNBR greater than 2.00 Withdrawal RCS pressure RCS pressure RCS pressure

< 2750 psia

< 2300 psia less than 2300 psia Boron Dilution

7.3 Subcritical

Subcritical:

Suberitical:

Sufficient time for Greater 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 rate withdrawal withdrawal Control Rod Drop 7.4 MDNBR > 1.30 MDNBR >2.00 MDNBR greater than 2.00 i

No fuel centerline Fuel centarline Fuel centerline melt temperature <3,700*F temperature <3,700'F Isolated Loop 7.5 MDNBR > 1.30 MDNBR greater than 2.97 Power operation with loop Startup No fuel centerline Fuel centerline out of service prohibited

)

seit temperature = 3,485'F by Technical Specifications Loss of Load 7.6

' M

!ssure Maximum RCS pressure =

Maximum RCS pressure =

less than 2,750 psia 2,663 psia 2,663 psia i

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M M

M M

M M

M M

M M

M M

M M

M TABLE _Z-13 (Continued)

IANIEE__ CORE _n l

SAEEU_ANALISIS l

SIRf3ARI_DLRES_UJJS t

j Incident Section Crlieria Refitence_ analysis Co_te_H i

Loss of Feedwater 7.7 Sufficient time for Emergency feedwater Emergency feedwater initiation of required 15 minutes required 15 minutes emergency feedwater following event following event Loss of Coolant 7.8 10CFR100 Less than 10.0% fuel Iess than 10.0% fuel Flow failure failure 4

Control Rod 7.9 10CFR100 No clad damage No clad damage Ejection Steam Line 7.10 Maintain fuel rod No return to critical Minimum suberit. of Rupture integrity

-0.233% p Steam Generator 7.11 10CFR100 Reference doses well Radiological doses well l

4 Tube Rupture within 10CHt100 within 10CFR100 Fuel Handling 10CFR100 Radiological doses well Reference analysis l

within 10CHt100 unchanged by reload i

Waste Gas 10CFR100 Radiological doses well Reference analysis within 10 Cat 100 unchanged by reload i

. Containment Pressure Peak pressure less Peak pressure <34.5 psig Reference analysts i

than 110% of design unchanged by reload (i.e., less than 34.5 psig) i i WPP40/4

I 8.0 STARTUP PROGBAM 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:

i 1.

Control rod operability tests will be performed by moving each rod group in turn from 0" to 9" 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 I

reactor just critical, allowing for system equilibrium and taking a

(

series of main coolant boron samples. This will be done as close I

as possible 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 n termined. This is done by establishing a boron change, balancing l

the reactivity change with a control rod position change and measuring the reactivity worth of the rod steps with the reactivity a

computer.

g l

5.

Isothermal temperature coefficient measurements are performed by changing main coolant temperature ar.d measuring the reactivity change with the reactivity computer. Measurements are taken at the equilibrium boron concentrations which correspond to both unrodded and rodded core conditions.

I 6.

Power and xenon defects are inferred using a reactivity balance before and after power accension.

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I 7.

Power distributions will be measured as soon as the reactor is at I

steady-state power (50% 1 power level I 75%).

This is done with the Incore Instrumentation System. A power distribution map will also be taken at a low power level to check for gross quadrant tilt.

8.

A startup test report on the above will be submitted to the NRC 90 days after startup.

I The acceptance criteria for the prediction of key core parameters is defined in Table 8-1; The permissible deviation from predicted values are

]

selected to insure the adequacy of the safety analysis.

In these tests, the nominal measured value is compared to the nominal calculated value.

Corrections are made for any differences between the measurement and calculational craditions.

I If the criteria in Table 8-1 are not met, the deviations are evaluated

]

relative to the assumptions in the safety analysis for the given core parameters. The Plant Operations Review Committee reviews tae evaluation

)

prior to power operation, j

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IABLE 8-1 YANTIE CORE 21 l

STARTUP TEST ACCEPTANCE CRITERIA lisasmsmant Conditions criteria I

1.

Control Rod Drop Time Operating temperature Drop times no greater than 2.5 seconds 2.

Critical Boron Hot zero power, near Measurement within I

Concentration all rods out A10% of predicted value 3.

Control Rod Group Hot zero power, Groups Worth of each group Worths C, A and B within 17.5% of the predicted value

I 4.

Control Rod Group Hot zero power, Groups If the criteria in Worths C. A and B Measurement (3) is not met, the total I

worth of all Groups measured must be within 27.5% of the

'I predicted value 5.

Isothermal Temperature Hot zero power, near Measuremgntwithin Coefficient all rode out 10.5x10~ Ap/*F of predicted value 6.

Radial Power Above 50% power with all The measured reaction I

Distribution rod groups greater than rates within 35% of 80 inches withdrawn the predicted value in the high power assemblies I

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I 9.0 LOSS _DLC00LANLACCIDENT 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 (LOCA) are related to minor changes in physics parameters and fuel design. As such, much of the previous analysis results I

can be used to support the Reload Cycle operation.

9.2 Rmm11 Break.LOCn The small break LOCA analysis (Reference 9.1) was performed using Core 20 physics parameters. The assumptions made in performing the analysis conservatively encompass any minor chaages in physics parameters that occur from cycle to cycle.

The small break analysis was performed with the limiting fuel stored energy which was calculated to occur for the fresh fuel at BOC I

conditions. A bounding axial power distribution was used for this analysis.

The power shape and hot rod peaking was selected to envelope the peak hot rod locations determined from the large break analysis.

A spectrum of cold leg breaks ranging from 10 inches down to 2-1/4 inches in diameter were analyzed with a PLilGR of 13.36 kW/fL. The Peak Cladding Temperature (ccT) for the small break LOCA cases was determined to be significantly lower than the large break results.

The limiting small break LOCA case was found to be an 8-1/2 inch break with a PCT of 1601'F.

I Therefore, for the reload cycle, the small break LOCA spectrum will continue to be nonlimiting.

9.3 Large Break LOCA I

The large break LOCA analysis consists of a break spectrum analysis and a burnup sensitivity study.

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I The break spectrum analysis is used to determine the limiting break I

size and type. The analysis is performed utilizing B00 conditions which result in the maximum fuel stored energy.

Since cosine and top-skewed axial power shapes are possible over the core lifetime, both shapes are analyzed in the break spectrum analysis.

, I The burnup sensitivity study is performed to establish the IJiGR limits as a function of cycle exposure. The worst case axial shape possible at each j

burnup is used in calculating the limit.

I 9.4 Dreak_SpectruntAnalysis The Reload Cycle fuel is similar in design and in hydraulic performance to previous cycles.

The physics parameters used at BOC conditions for Core 16 and Core 18 were compared to the Reload Cycle values. An evaluation was made which showed the Reloed Cycle values either were bounded by the previous values or had an insignificant effect on analysis results. Therefore, the I

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 g

g break characteristics for the cosine power shapes while the Core 18 analysis provides those for the top-skewed shapes.

9.5 Burnup_Sentitivity study I

The burnup sensitivity study addressed both f resh fuel and recycled fuel performance in the Reload Cycle.

For burnup points at less than I

4 GWd/Mtu, a cosine axial power shape is assumed since the " worst case" xenon shapes closely approximate a cosine.

Beyond 4 GWd/Mtu, the shapes become top-skewed and therefore, the actual " worst case" xenon shapes are used.

The results of the sensitivity study are given in Table 9-1.

Where the input parameters of the previous cycles bounded the Reload Cycle parameters, no analysis was performed.

I,,,,,,,

I

The analysis for the recycled fuel shows the LOCA limits of the previous cycles (Core 18 through Core 20) to be valid for the Reload Cycle.

For the fresh iuel, the allowable LOCA limits for the reload cycle are about O.4 kW/ft lower than the current cycle.

This change is the result of a shift in the location of the highest temperature rod to a region with a smaller radiation heat sink.

These limits are given in Figure 9-1 where all LHCR values were multiplied by 0.973 to account for heat deposition in the fuel rod.

9.6 Swxparv of Ruulla Based on the analysis presented in Sections 9.0 through 9.5.

Operation within the limits specified in Figure 9-1 yields LOCA results within the specification of 10CFR50.46.

9.7 Radiological _Constr.utacas of a Des _ign Basis LOCA and Posi-l&CA_1tdrosen ContInl The radiological consequences from the design basis LOCA and Post-LOCA Hydrogen Control must conform to the guideline values specified within 10CFR100.

The calculation of the source term for these events assumes a total fission product inventory for an equilibrium core. Thus, the calculation of the 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, the results of the analysis presented in the FSAR also apply to the Reload Cycle.

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I IABLE 9-1 CORE 21 BURNUP SENSITIVITY STUDY RESULTS I

CAB Fuel PLHGR Limit Reference Reference CWd/Mtu ly.gg kW/ft evele Cvele PCT ("F) 0.0 Fresh 9.60*

21 2132 0.25 Fresh 10.65*

21 2160 1.00 Fresh 11.40 21 2124 4.00 Fresh 10.60*

20 1933 10.00 Fresh 9.20*

18 1973 14.00 Fresh 9.10 21 2197 17.50 Fresh 8.00 21 1636 0.0 Recycled 11.50 19 2164 4.0 Recycled 10.50 19 1951 10.0 Recycled 9.30 21 2181 14.0 Recycled 9.10 18 2093 17.5 Recycled 8.00 18 1565 I

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• PCT's calculated at higher PLEGh.

Limit PHLGR reduced to account for shift in location of high temperature rod based on results of calculations for fresh fuel at CAB's of 1.00 GWd/Mtu and 14.00 GWd/Mtu.

~~

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i i,g FIGURE 91 W

CORE 21 ALLOWABLE PEAK ROD LHGR j

VS. CYCLE BURNUP 12

I r

]

11 -

o FRr.5H FUEL i

D EXPOSED FUEL g 10 -

o g

ga

.g9-1 W

l 8-W G

E G

l T

3 Y

Y Y

v v g y

s y T u gT g3 T

T T

T T T 3 T T U V W W

T T T T E Y Y T 1 g' W g'T T

Y Y g3 O

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18

,I CYCLE BURNUP (GWD/MTU)

, I

i 2

10.0 EEFERf.NCES 4.1 "Rowe Batch D Reload Fuel Design and Development Report " CE-NPSD-582-P, January 1990.

I C-E Report, "First C-E Supplied Rowe Fuel Batch Design and Development 4.2 Report, May 1985.

l 4.3 C-E Report, Yankee Rowe Nuclear Fuel Desiga Report Update for Batch C."

June 1988.

j 4.4 K. P. Calbraith, "GAPEXX: A Computer Code for Predicting

)

Pellet-to-Cladding Heat Transfer Coefficients", XN-73-25. August 13, 1973.

i 4.5 YAEC-1583, " Yankee Nuclear Power Station Core XIX Performance Analysis,"

I January 1987.

4.6 YAEC-1652, " Yankee Nuclear Power Station Core 20 Performance Analysis,"

I August 1988.

5.1 W. R. Cadwell, "PDQ-7 Reference Manual". WAPD-2-678, January 1967.

5.2 R. J. Breen, O. J. Marlowe and C. J. Pfeifer. " HARMONY:

System for Nuclear Reactor Depletion Computation", WAPD-E -478. January 1965.

1 5.3 R. F. Barry, " LEOPARD - A Spectrum Dependent Nonspatial Depletion Program", WCAP-2795, March 1965.

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

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

December 1967.

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

6.2 ASME Paper. E. S. Markowski, et al., "Effect of Rod Boving on CHF in PWR Fuel Assemblics " AICHE-ASME Heat Transfer Conference Salt Lake City, Utah, August 15-17, 1977.

7.1 Final Safety Analysis Report. Yankee Nuclear Power Station.

7.2 YAEC-1652, " Yankee Nuclear Power Station, Core 20 Performance Analysis "

August 1988.

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 Proposed Change No. 226, BYR 89-120. " Request for Change to Technical Specification 3.2.4," July 24, 1989. WPP40/4

I 7.6 Amendment No. 127, Letter from P. M. Sears, USNRC, to YAEC, NYR 89-326,

" Issuance of Amendment No. 127 to Facility Operating License No. DPR-3,"

October 27, 1989.

9.1 YAEC-1732, " Yankee Plant Small Break LOCA Analysis," July 1990.

9.2 Letter, FYR 85-131, dated November 19, 1985 G. Papanic, Jr., YAEC to USNRC." Core XVIII LOCA Analysis - Additional Information."

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

September 1982.

9.4 YAEC-1496. " Yankee Nuclear Power Station Core XVIII Performance 1

Analysis," August 1985.

9.5 Final Safety Analysis Report, Yankee Nuclear Power Station.

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