ML20127B083

Cycle 9 Core Performance Analysis

ML20127B083
Person / Time
Site:	Maine Yankee
Issue date:	04/30/1985
From:	Paul Bergeron, Cacciapouti R, Husain A Maine Yankee
To:
Shared Package
ML20127B075	List: ML20127B072 ML20127B077 ML20127B083
References
YAEC-1479, NUDOCS 8506210353
Download: ML20127B083 (189)

Text

.

1 6

MAINE YANKEE CYCLE 9 CORE PERFORMANCE ANALYSIS April 1985 Major Contributors:

Nuclear Engineering Department Reactor Physics Group Transient Analysis Group G. M. Solan M. W. Scott D. G. Adli .P. J. Guimond J. P. Gorski G. E. Jarka B. Y. Hubbard M. P. LeFrancois K. S. Spinney T. D. Radcliff S. VanVolkinburg K. R. Rousseau LOCA Analysis Group Nuclear Evaluations and Support Group l

J. Ghaus K. E. St. John R. C. Harvey G. E. Jarka Yankee Atomic Electric Company l Nuclear Services Division 1671 Worcester Road Framingham, Massachusetts 01701 8506210353 850614 PDR P

ADOCK 05000309 PDR

3 APPROVALS I

Approved By: .

P. A. Bergerog Manager (Date)

Transient Analysis Group Approved By: dd Pf R.' J(/ Cacciapou,t[, Manager '/(Date$

Reactor PhysicFGroup Approved By: k('Date')

23f SI A. Husain, Madager LOCA Analysis Group Approved By: b bS S. f. Schultz, Mdiager U (Date)

Nuclear Evaluations and Support Group Approved By:

B.C.Slifer, Direct {r ' (Date)

Nuclear Engineering Uepartment

~

DISCLAIMER This document was prepared by Yankee Atomic Electric Company (" Yankee")

pursuant to a contract between Yankee and Maine Yankee Atomic Power Company

("Naine Yankee"). The use of information contained in this document by anyone cther than Maine Yankee, or for any purpose other than for which it is intended, is not authorized, and with respect to any unauthorized use, neither Yankee nor its officers, directors, agents, or employees assume any cbligation, responsibility, or liability, or makes any warranty or representation concerning the contents of this document or its accuracy or completeness.

-lii-

ABSTRACT This report presents design and analysis results pertinent to the cperation of Cycle 9 of the Maine Yankee Atomic Power Station. These include core fuel loading, fuel description, reactor power distributions, control rod worths, reactivity coefficients, the results of the safety analyses performed to justify plant operation, the startup test program and the Reactor Protective System (RPS) setpoints assumed in the safety analysis. The cnalysis results, in conjunction with the startup test results, RPS setpoints cnd Technical Specifications, serve as the basis for ensuring safe operation cf Maine Yankee during Cycle 9.

4 l -iv-

ACKNOWLEDGEMENTS The authors would like to express their gratitude to W. J. Szymezak for his assistance.

-V-

TABLE OF CONTENTS

.P_agg APPROVALS........................................................ 11 DISCLAIMER....................................................... iii ABSTRACT......................................................... iv ACKN0WLEDGEMENTS................................................. v TABLE OF CONTENTS................................................ Vi LIST OF TABLES................................................... ix LIST OF FIGURES.................................................. xiii

1.0 INTRODUCTION

..................................................... 1 2.0 OPERATING HIST 0RY................................................ 3 2.1 Cycles 1 and 1A............................................ 3 2.2 Cycle 2.................................................... 3 2.3 Cycles 3 and 4............................................. 4 2.4 Cycles 5 and 6............................................. 4 2.5 Cycles 7 and 8............................................. 4 3.0 RELOAD CORE DESIGN............................................... 7 3.1 General Description........................................ 1 3.1.1 Core Fuel Loading.................................. 7 3.1.2 Core Burnable Poison Loading. . . . . . . . . . . . . . . . . . . . . . . 7 3.1.3 Core Loading Pattern............................... 8 3.1.4 As s emb ly E xp o su re H i s t o ry . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1.5 CEA Group Configuration............................ 9 3.2 Fuel System Design......................................... 10 3.2.1 Fuel Mechanical Design............................. 10 3.2.2 Thermal Design..................................... 12 3.2.3 Thermal-Hydraulic Design........................... 13 4.0 PHYSICS ANALYSIS................................................. 36 4.1 Fuel Management............................................ 36 4.2 Core Physics Characteristics............................... 36 4.3 Power Distributions........................................ 36 4.4 CEA Group Reactivity Worths............'.................... 37 4.5 Doppler Reactivity Coefficients and Defects................ 37 4.6 Moderator Reactivity Coefficients and Defects.............. 38 4.7 Soluble Boron and Burnable Poison Reactivity Effects....... 39 4.8 Kinetics Parameters........................................ 40 4.9 Safety-Related Characteristics............................. 40 4.9.1 CEA Group Insertion Limits......................... 40 4.9.2 CEA Ejection Results............................... 41

- vi-

TABLE OF CONTENTS (continued)

Este 4.9.3 CEA Drop Results................................... 42 4.9.3.1 Design Analysis Results.................. 42 4.9.3.2 Post-CEA Drop Restrictions............... 42 4.9.4 Available Scram Reactivity......................... 43 4.9.5 Shutdown Margin Requirements....................... 45 4.9.6 Augmentation Factors............................... 46 4.10 Pressure Vessel F1uence.................................... 47 4.11 Methodology and Methodology Revisions...................... 48 4.11.1 Summary of Physics Methodology Documentation....... 48 4.11.2 CEA Ejection Results from Partial insertions. . . . . . . 48 4.11.3 Moderator Density Reactivity Defect Analysis for L0CA............................................... 49 5.0 SAFETY ANALYSIS.................................................. 83 5.1 Genera 1.................................................... 83 5.1.1 Initial Operating Conditions....................... 83 5.1.2 Core Power Distributions........................... 84 5.1.3 Reactivity Coefficients............................ 85 5.1.4 Shutdown CEA Characteristics....................... 86 5.1.5 Reactor Protective System Setpoints and Time Delays.................................... 88 5.2 Summary.................................................... 88 5.3 Anticipated Operational Occurrences for Which the RPS Assures No Violation of SAFDLs............................. 89 5.3.1 Control Element Assembly Bank Withdrawal........... 90 5.3.2 Boron DL1ution..................................... 91 5.3.2.1 Dilution During Refueling................ 92 5.3.2.2 Dilution During Cold, Transthermal, and Hot Shutdown With the RCS Fliled..... 92 5.3.2.3 Dilution During Cold, Transthermal, and Hot Shutdown With Drained RCS Conditions............................... 93 5.3.2.4 Dilution During Hot standby, Startup, and Power Operation...................... 94 5.3.2.5 Failure to Borate Prior to Cooldown...... 94

-vil-

TABLE OF CONTENTS (continued)

P.112 5.3.3 Excess Load Incident............................... 95 5.3.4 Loss of Load Incident.............................. 96 5.3.5 Loss of Feedwater Incident......................... 96 5.4 Anticipated Operational Occurrences Which are Dependent on Initial Overpower Margin for Protection Against Violation of SAFDLs........................................ 97 5.4.1 Loss-of-Coolant F1ow............................... 97 5.4.2 Full Length CEA Drop............................... 98 5.5 Postulated Accidents....................................... 100 5.5.1 Steam Line Rupture................................. 100 5.5.2 Steam Generator Tube Rupture....................... 103 5.5.3 Seized Rotor Accident.............................. 104 5.5.4 CEA Ejection....................................... 104 5.5.5 Loss of Coolant.................................... 105 5.5.5.1 Introduction and Summary................. 105 5.5.5.2 Large Break LOCA Analysis................ 106 5.5.5.2.1 Break Spectrum Analysis. . . . . . 107 5.5.5.2.2 Burnup sensitivity Studies... 101 5.5.5.2.3 Cosine Axial Power Distribution Study........... 108 5.5.5.3 Small Break LOCA Analyses................ 108 5.6 Plant Hardware Modifications............................... 108 6.0 STARTUP TEST PR0 CRAM............................................. 163 6.1 Low Power Physics Tests.................................... 163 6.2 Power Eccalation Tests..................................... 164 6.3 Acceptance Criteria........................................ 164 7.0 CONCLUS10NS...................................................... 168

8.0 REFERENCES

....................................................... 169

-viii-

LIST OF TABLES Number Title Page 2.1 Operating Mistory Summary 5 2.2 Fuel Assembly Types by Cycle 6 3.1 Cycle 9 Assembly Description 16 3.2 Cycle 9 Core Loading 17 3.3 Mechanical Design Features of Cycle 9 Fuel 18 3.4 Centerline and UO2 Melt Temperature Comparison 19 3.5 Cycle 9 Ratio of Maximum Radial Relative Pin Powers -

Maximum in Type E Fuel to Maximum in Core 20 3.6 Cycle 9 Ratio of Maximum Radial Relative Pin Powers -

Maximum in Type L Fuel to Maximum in Core 21 3.7 Cycle 9 Ratio of Maximum Radial Relative Pin Powces -

Maximum in Type M Fuel to Maximum in Core 22 3.8 Cycle 9 Thermal-Hydraulic Parameters at Full Power 23 4.1 Cycles 3, 8, and 9 Nuclear Characteristics 51 4.2 Cycles 3, 8, and 9 CEA Group Worths at HFP 52 4.3 Cycles 3, 8 and 9 Core Average Doppler Defect 53 4.4 Cycles 3, 8 and 9 Core Average Doppler coefficient 54 4.5 Cycles 3, 8 and 9 Moderator Temperature Coefficients 55 4.6 Cycles 3, 8 and 9 ARI Moderator Defect with Worst Stuck CEA 56 4.7 Cycles 3 and 9 Kinetics Parameters 51 4.8 Cycles 8 and 9 CEA Ejection Results from Full Insertions 58 4.9 Cycle 9 CEA Ejection Results from Partial Insertions 59

-ix-

LIST OF TABLES (continued)

Number Title Egge 4.10 Cycles 8 and 9 CEA Drop Results at BOC 60 4.11 Cycles 8 and 9 CEA Drop Results at EOC 61 4.12 Cycles 8 ani 9 Dropped CEA with Power Level Restriction -

Most Limiting Peaking Cases 62 4.13 Cycle 9 Available Scram Reactivity 63 4.14 Cycle 9 Required Scram Reactivity 64 4.15 Cycles 8 and 9 Augmentation Factors 65 4.16 Cycles 8 and 9 Core Radial Pin Power Census for Augmentation Factor Calculation 66 4.17 Cycles 6 through 9 Relative Pressure Vessel Fluence Comparison 67 Physics Methodology Documentation 4.18 68 5.1 Maine Yankee safety Parameters 109 5.2 Cycle 9 - Incidents considered 113 5.3 Cycle 9 Safety Analysis - Summary of Results 114 5.4 Required Initial RCS Boron Concentrations to Allow Fifteen Minutes Margin to Criticality for Dilutions from Shutdown Conditions with the RCS Filled 117 5.5 Required Initial hCS Boron Concentrations to Allow Thirty Minutes Margin to Criticality for Dilutions from Shutdown Conditions with the PCS Drained 119 5.6 Summary of Boron Dilution Incident Results for Cycle 9 121 5.7 Nominal Rod Worths to Prevent a Return to Power During a Steam Line Rupture Accident 122 5.8 Cycle 9 CEA Ejection Accident Results 123 5.9 Cycle 9 Burnup Sensitivity Study Results 124 5.10 Cycle 9 Cosine Axial Power shapo Study Results 125

-x-

LIST OF TABLES (continued)

Number I}tle g 5.11 Comparison of Themal Margin for Limiting Cycle 9 Power Distributions to FSAR Design Power Distribution 126 5.12 Reactor Protective System Trips Assumed in the Cycle 9 Safety Analysis 127 5.13 Available Shutdown Margin Assumed in Cycle 9 Safety Analysis 128 6.1 Cycle 9 Startup Test Acceptance Criteria 166 i

e

.xt.

LIST OF FIGURES Number Title Page 3.1 Cycle 9 Burnable Poison Shim Assembly Locations 25 3.2 Cycle 9 Assembly Loading Pattern 26 3.3 Cycle 9 Calculated Assembly Exposures at BOC 27 3.4 Cycle 8 Burnup Distribution by Assembly near 6000 MWD /NT 28 3.5 Cycle 9 CEA Group Locations 29 3.6 Cycle 9 Centerline Temperature Vs. LHGR at BOC for Type N Fuel 30 3.7 Cycle 9 Centerline Temperature Vs. LHCR at BOC for Type M Fuel 31 3.8 Cycle 9 Centerline Temperature Vs. LHCR at BOC for Type L Fuel 32 3.9 Cycle 9 Centerline Temperature Vs. LHCR at EOC for Type N Fuel 33 3.10 Cycle 9 Centerline Temperature Vs. LHGR at EOC for Type M Fuel 34 3.11 Cycle 9 Centerline Temperature Vs. LHGR at EOC for Type L Fuel 35 4.1 Cycle 9 Assembly Relative Power Densities BOC (500 MWD /MT). HFP, ARO 69 4.2 Cycle 9 Assembly Relative Power Densities MOC (6,000 MWD /MT) HFP, ARO 70 4.3 Cycle 9 Assembly Relative Power Densities EOC (14,000 MWD /MT) HFP, ARO 71 4.4 Cycle 9 Assembly Relative Power Densities BOC (500 MWD /MT) HFP, CEA Bank 5 Inserted 72 4.5 Cycle 9 Assembly Relative Power Densities MOC (6,000 MWD /MT). HFP, CEA Bank 5 Inserted 73 4.6 Cycle 9 Assembly Relative Power Densities ROC (14,000 MWD /MT) HFP, CEA bank 5 Inserted 74 4.7 Cycle 9 Allowable Unrodded Radial Peak Vs. Cycle Average Burnup 75

-xii-

LIST OF FIGURES (continued)

Number Title Egge

4.8 Cycle 9 Modcrator Temperature Coefficient Limits vs.

Power Level 76 l

4.9 Cycle 9 Power Dependent Insertion Limit (PDIL) for CEAs 77 4.10 Cycle 9 Reference Power Level vs. Nominal Cold Les Temperature 78 4.11 Cycles 8 and 9 Maximum Radial Peaking Vs.

Dropped CIA Worth from Specified Power Levels 79 4.12 Cycle 9 Shutdown Margin Equation and Required Scram Reactivity 80 4.13 Cycle 9 Required Shutdown Margin Vs. RCS Boron Concentration 81 4.14 Cycles 5 and 9 Moderator Density Defect Curves for LOCA Analysic 82 5.1 Cycic 9 Allowablo 3 Loop Steady-State Coolant Conditions 129 5.2 Design Power Distributions 130 5.3 Normalized Roactivity Worth vs. Position Assumed in BOC CEA Ejection Analysin 131 5.4 Normalized Reactivity Worth vs. Position Assumed in EOC CEA Ejection Analysin 132 5.5 TM/LP Trip Sctroint (Yg Versus Ag) 133 5.6 TM/LP Trip Sotpoint Part 2 134 5.7 Symmetric Offset Trip Function 135  !

5.8 RETRAN Primary System Model 136 5.9 RETRAN Secondary System Model '

137 5.10 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Pressurizer Pressure 138 5.11 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Core Temperaturn 139 l

-Mill-I

LIST OF FIGURES (continued)

Number I,Lt.le P_gge 5.12 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Faulted Loop Cold Les TemFerature 140 5.13 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Core Flow 141 5.14 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Pressuriser Level 142 5.15 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Upper Head Level 143 5.16 cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Faulted Loop SG Break Flow 144 5.17 Cycle 9 MSLB HZP MFWRV Pailure With Electric FW Pumps -

SG Pressure 145 5.18 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

SG Inventory 146 5.19 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Faulted Loop SC Heat Input 141 5.20 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Faulted Loop FW Flow 148 5.21 Cycle 9 MSLB HZP MFWRV Failure With Electric FW Pumps -

Auxiliary FW Flow 149 5.22 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Relative Power 150 5.23 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Pressurizer Pressurn 151 5.24 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Core Temperature 152 5.25 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Faulted Loop Cold Let Temperatura 153 5.26 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Core Flow 154

-xiv-

LIST OF FIGURES (continued)

Number 11,(19. EALe 5.27 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Pressuriser Level 155 5.20 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Upper Mead Level 156 5.29 Cycle 9 MSLB HFP MFWEV Failure With Electric FW Pumps -

Faulted Loop SG Break Flow 157 5.30 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

SG Pressure 158 5.31 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

SG Inventory 159 5.32 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Faulted Loop SC Heat Input 160 5.33 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Faulted Loop FW Flow 161 5.34 Cycle 9 MSLB HFP MFWRV Failure With Electric FW Pumps -

Auxiliary FW Flow 162

-xv- ,

1.0 INTRODUCTION

This report provides justification for the operation of Maine Yankee during the next fuel cycle Cycle 9. The Cycle 9 refueling will involve the discharge of 73 assemblies and the insertion of 72 new fuel assemblies and one burned Type E assembly from Core 2. The new fuel assemblies (designated Batch W) are being provided by Combustion Engineering (CE) and are similar in design to the CE Datch H fuel provided for Cycle 4. The Types L and M fuel, remaining from Cycles 7 and 8, were provided by Exxon Nuclear Corporation (ENC).

The CE fuel designs are similar but not identical to the IWC design.

Small differences exist in both the mechanical and hydraulic characteristics.

The differences in the mechanical design and the hydraulic characteristics are discussed in Section 3.2.1 and Section 3.2.3 of (33).

I The proposed operating conditions for Cycle 9 are a rated core thermal ,

power of 2630 MWt, at a steady state operating pressure of 2225 pois to 2275 psia, at a maximum indicated core inlet temperature of 552"F. In addition, operation is allowed over a pressure range from 2075 psia to 2225 psia by imposing a limit on the maximum core inlet temperature at the lower pressures t; preserve the DNB margin. This assures that DN8 performance is the same for all possible limiting temperature and pressure combinations. These conditions are consistent with the " Stretch power" condstlons proposed in (1) and l l currently allowed by the Maine Yankec operating license, with the exception of the increase in the maximum indicated ccre inlet temperature from 550 F to 552 F.

This report contains sections dealing with the fue mechanical, thermal-hydraulic, physics and safety analysis aspects of the operation of Cycle 9. A description of the Startup Test program is also included. Except as noted, the methods used in these analyses are in accordance with those described in (8-16), (35), and (50). These methods have been approved by the NRC for use on Maine Yankee in (11) and (18). Methods used in safety-related analyses for the fuel mechanical design evaluations are based on the Combustion Engineering and Rxxon Nuclear generic models which have received prior approval by the NRC. New methods were used in the steam line rupture 1

l cccident and CEA Ejection Analysis, as described in (57) and (55),

respectively.

The significant features of Cycle 9 aret

1. Continued implementation of lower-leakage core designs as initiated in Cycle 7 (Sections 3.1 and 4.1).

2. Replacement of part-strength with full-strength CEAs in the non-scrasenable locations (subgroup 55) in CEA Bank 5 (Section 3.1.5).

i

3. An increase in the maximum allowable indicated core inlet temperature from 550 F to 552 F.

I Details of each change are provided in the sections indicated.

i i

r I

.i ,

i

!' f f

I i

l A

e2

2.0 OPERATING HISTORY The operating history of Maine Yankee has consisted of nine cycles d:signated as 1, 1A, and 2 through 8. The significant operating conditions and durations of the cycles are defined in Table 2.1. The fuel assembly typec 1:4ded by cycle are given in Table 2.2.

2.1 Cycles 1 and if The initial Maine Yankee core consisted of unpressurized, low density fuel designated as core 1 design fuel assemblies (Types A. R, and C). Cycle 1 l

operation was restricted and terminated due to leaking fuel assemblies.

4 Cycle 1A consisted of operation after the leaking fuel assemblies from I

the initial core were replaced with fresh fuel designated as Replacement Fuel (Type RF) assemblies. The mechanical design of the Type RF assemblies was cssentially the same as core 1 design fuel. The significant difference in the d: sign was the pressurization of the fuel rod with helium sufficient to prevent creep collapse of the fuel rod cladding and improve gap heat transfer. The replacement fuel assemblies performed successfully during cycle 1A.

2.2 Cycle 2 Cycle 2 consisted entirely of fresh assemblies designated as Core 2 d: sign fuel (Types D, g, and F). Mechanical design changes were made in comparison to the Coro 1 design fuel. These comprised pre pressurization, ,

higher fuel density, and smaller diameter pellets. A detailed discussion of

! the design changes was provided in (30). The Core 2 design fuel performed successfully. Subsequent to Cycle 2 operation, burnable poison shim f ailures were discovered in the Type a sesemblies. Corrective action consisted of replacement of all Type a shims with water-filled strcaloy rods prior to r0 insertion in subsequent cycles.

i I

3- r

2.3 Cycles 3 and 4 Cycle 3 consisted of fresh fuel assemblies of the Core 2 design (Types l

G and H) and Replacement Fuel assemblies reinserted from Cycle IA. The l performance integrity of the Cycle 3 fuel had been demonstrated through l

irradiation in Cycles 2 and 1A respectively. All fuel performed successfully during Cycle 3.

Cycle 4 consisted of all fuel assemblier of the Core 2 design, slight i

design changes to the fresh Type I fuel were made and discussed in section 3.2.1 of (7). New fuel and once-burned fuel assemblies from Cycle 2 were inserted and the replacement fuel discharged. A small number of leaking fuel i

Ossemblies were discovered near end-of-cycle.

2.4 Cycles 5 and 6 Cycles 5 and 6 consisted of fuel assemblies of the Core 2 CE design and fresh assemblies designed by RNC (Types J and X) . A detailed discussion of the EN:: design assemblies was provided in (33). Five Core 2 design 1 caking Casomblies returned to the core in cycle 5 were repaired by replacement of fuel rods with fresh, low enrichment Core 2 design fuel (34 rods) or water-filled zircaloy rods (10 rods). The fuel perfomed successfully during Cycles 5 and 6.

l 2.5 Cycles 7 % g Cycles 7 and 8 consisted almost entirely of ENC-designed fuel. One Type E assembly of the Core 2 Cs design with Cycle 2 exposure was inserted in the core center location. The fresh gNC batches (Types L and M) represented an increase in enrichment to 3.30 w/o 17 235. The Cycle 7 design was the first low-leakage, low-fluence core design. Minor fuel design changes for the Types l

L and M fuel were discussed in (50). All fuel has perfomed successfully i during Cycles 7 and 8.

l l

4

TABLE 2.1 l MAINE YANKEE OPERATING HISTORY SUIGRARY Date of Core Power Level Cycle Power Licensed Operated turnup GXgle Escalation (Itit) M) (WD/MT) 1 11/3/72 2440 50-80(1) 10367 1A 10/12/74 2440 80(1) 4500 2 6/29/75 2440 100 17395 3 6/17/77 2630(2) 93 11075 4 8/23/78 2630 97(3) 10496 5 3/11/80 2630 97 10796 6 7/20/81 2630 97 11580 7 12/12/02 2630 100 12466 l ,

8 6/20/84 2630 100 12000(4) l l

t (1) Power decrease and primary system pressure decrease to 1800-2000 psia due to leaking fuel (2) Licensed power increase from 2440 Wt/2100 puis operation to 2630 i MWt/2250 psia operation (3) Power restriction due to secondary plant limitations (turbine)

(4) Estimated l -s-i i

TABLE 2.2 i

l MAINE YANKEE FUEL ASSEMBLY TYPES BY CYCLE Assembly 9 - % r of Fuel Assemblies by Cycle Fuel Enrichment Mechanical 1333 (w/o U-235) Deslan Tree J M J J ,,1 _}, _i .], _1 A 2.01 CE-Core 1 69 51 - - - -

B 2.40 CE-Core 1 80 24 - - -

C 2.95 CE-Core 1 68 64 - - -

RF 2.33 CE-RF - 2 - -

RF 1.93 CE-RF - 70 - 65 - - - -

D 1.95 CE-Core 2 - - 69 - - - -

E 2.52 CE-Core 2 - - 80 12 61 1 1 1 1 F 2.90 CE-Core 2 - - 68 68 12 - - -

l C 2.73 CE-Core 2 - - - 32 32 32 - -

H 3.03 CE-Core 2 - - - 40 40 40 - - -

3.03 CE-Core 2 - - - - 72 72 72 - -

1

- 12 72 12 .

J 3.00 ENC - - - -

- - 12 12 12 K 3.00 ENC - - - -

- - - 72 12 L 3.30 ENC - - - -

- - - - - 12 M 3.30 ENC - - -

l 6-l r

3.0 RELOAD CORE DESIGN 3.1 General Description l

3.1.1 Core Puel Loading The core of Maine Yankoo Cycle 9 consists of 217 fuel assemblies of the j type and quantity detailed in Table 3.1. The single Type I assembly is of the Core 2 mechanical design and has irradiation exposure from Cycle 2. Assembly Type L was introduced in Cycle 7 and has irradiation exposure from Cycles 7 i and 8. Assembly Type M was introduced in Cycle 8. Assembly Type N is fresh

fuel to be introduced in Cycic 9. Assembly Types L and M are ENC design fuel. Assembly Type N is fabricated by CE and is designated Core 2 design fuel. The Type N fuel initial enrichment and shim loading are equal to thosc ef the Types L and M fuel. The total number of fuel rods by assembly type for Cycle 9 is also given in Tabic 3.1. The core loading by fuel type is given in Tabic 3.2. .

, I l

3.1.2 Core Burnable Poison Loading j I burnable poison shim rods are located in selected assemblies in Cycle 9. The total number of shim rods and locations by assembly type is detailed in Table 3.1. The shim locations in the assemblies are illustrated -

i ('

in Figure 3.1. As described in Ecction 2.2, the single Type E assembly shim rod locations contain water-filled strcaloy rods with end plugs to restrict the flow in these rods.

All burnable poison shims are composed of 8 C in A10 . The gNC

] 4 23 '

design shim irradiation integrity has been demonstrated in the Types J, K, L, and M assemblies during Cycles 5 through 8. The CE design shim irradiation a

i integrity has been demonstrated in the Types 0 and I assemblies during Cycles ,

3 through 6.

I L

f 1

I i

r 1~ j i

3.1.3 Core Loading Pattern The fuel assembly locations designated for Maine Yankee Cycle 9 are given for the first quadrant in Figure 3.2. They are given relative to the previous locations of the Type E assembly in Cycle 2 and the Types L and M Casomblies in Cycle 8. The appropriate rotation index relative to the previous assembly position in the core is also given for each assembly. The loading and rotations of the other quadrants are such that mirror symmetry cxists with respect to the quadrant boundary lines.

The Cycle 9 loading pattern incorporates a low-leakage design, achieved by placement of fresh fuel assemblies in selected core interior locations and burned fuel assemblies on the core edge. The Cycle 9 loading pattern is a further improvement upon the Cycle 7 and 8 low-Icakage loading patterns. The benefits of such a core design arci

! 1) Reduced irradiation exposure to the reactor pressure vessel, thus reducing the rate of it.adiation embrittlement:

2) Extended cycle full-pob..e lifetime due to reduced neutron leakagen l

3) preferred fuel rod power and exposure histories from fuel performance and mocianical integrity considerations

4) Improved stability to axial xenon oscillations nose end-of-cycles and S) A less severe moderator defect with cooldown at end-of-cycle, providing greater shutdown margin for cooldown transients.

3.1.4 AtnEhly Inosure Illstory The calculated exposure history of the cycle 8 fuel assemblies at beginning-of-Cycle (BOC) is given in Fiture 3.3. The exposures are based on an expected cycle length of 12,000 mwd /Mt for Cycle 8 and the achieved cycle length of 12,466 mwd /Mt for Cycle 1. Table 3.2 gives DOC average exposures by 8

fuel type. The Cycle 9 BOC average exposure for the core is approximately 12,900 MWD /MT.

The exposure history of the assemblies utilized in the analysis is demonstrated to be accurate by comparison with incore detector plant data.

Figure 3.4 is a comparison of predicted and actual burnup assembly data in the ciddle of Cycle 8. The excellent agreement demonstrates a high confidence in the prediction of the core depletion behavior.

3.1.5 CEA Group Configuration The Control Element Assembly (CEA) group configuration for the lead regulating bank, CEA Bank 5 had been modified for Cycle 7. This configuration was described in detail in (50) and was unchanged for Cycle 8.

F r Cycle 9, the configuration is unchanged, but the four part-strength CEAs in Subgroup SB have been replaced with full-strength CEAs. Figure 3.5 shows the CEA group locations in the quarter core. The Bank 5 configuration thus consists of

1) Nine full-strength CEAs, designated Cubgroup 5A, which are scrammable CEAs and contribute to the available serrm reactivity.

2) Four full-strength CEAs, designated Cubgroup SD, adted to Dank 5 for local power distribution control. These four CEA locations are non-scrammable and do not centribute to the available scram reactivity.

As in Cycles 7 and 8, subgroups SA and SD are independently moveable and not directly connected as a single CEA bank. As such, their movements are administrativo1y controlled for positioning as a single CEA bank. To cecommodate this movement, the physics input to the Heactor protective System (RPs) setroint analysis has included power distribution cases sufficient to j:stify differences in insertion between those two regulating subgroups subject to the CEA group insertion limits, as discussed in section 4.9.1.

9

1 I

3.2 Fuel System Desian i 3.2.1 Fuel Mechanical Deslan i

Fuel assembly types and quantitics for this Maine Yankee core are given in Table 3.1. The mechanical design of the fuel assemblies present in this  ;

I cycle is documented in references ((7), (15), (20), (30), and (61)). The fuel )

1 j tesemblies have been designed to maintain mechanical, material, chemical, and i therinal-hydraulic compatibility with all other fuel and structures in the .

1

]

reactor core. Table 3.3 lists the mechanical design features and vendors of l cil fuel types.  ;

The detailed fuel assembly description and mechanical design criteria l i and the design considerations for the recycled reload fuel have been described j

in (33) and (50). The fresh reload fuel is provided by Combustion gngineering l

(Cs). Compared to previous CE Batch H, the cold beginning-of-life shoulder j gap has been increased by reducing the overall length of the fuel rod and the height of the lower end fitting. These changes will allow for additional fuel l l

rod growth clearance. The overall length of the poison rods is increased due to the reduction in lower end fitting height. l The mechanical decir,n analyses performed for the Cs tatch N fuel ,

include the following: .

fuel Claddine Collarjtg a.

i The fuel is designed to preclude collapse of the fuel rod cladding (

l as a result of the long-term ovality increase due to creep under l enternal pressure on an unsupported length of fuel rod cladding if  !

l

< axial gaps form in the fuel pellet column. Irradiation creep of l the cladding is a function of temperature, stress, and fast neutron flux that accentuates the ovality with time, A point of j instability may be reached when the ove11ty reaches a critical value and the cladding flattens completely into the gap. The '

l results of the design analysis indicate that the collapse resistance of the fuel rods is sufficient to preclude collapse during the projected design lifetime of the fuel. (

i 10-j f I I I I

_J

i

.a

b. Irradiation-Induced Dimensional Channes Fuel assembly length change results from two distinct mechanisms in I the Zircaloy guide tubest irradiation induced growth and l compressive creep.

The overall elongation of a Zircaloy clad fuel rod is due to a combination of the stress-free irradiation growth of the Zircaloy I cladding, mechanical interaction between the U0 fuel pellets and 2

the Zircaloy cladding and creepdown of the cladding under external coolant pressure. sach of these factors is related to the time of ,

l operation through accumulated fluence.

Design analyses ches the following (1) the beginning-of-life cold shoulder sap is adequate to assure sufficient clearance between the [

l top of the fuel rods and the bottom of the upper-end fitting flow plate while accommodating fuel rod and guide tube growth without [

l interference at end-of-life. (2) the fuel bundle has adequate clearance at end-of-Life to preclude interference of the bundle between the fuel alignment plate and the core support plate, (3) j and the fuel bundle has adequate clearance at end-of-life to l preclude having the hold down springs reach their solid height. l l

l <

Because the poison rods are shorter than the fuel rods, and are not affected by pellet clad interaction, they are predicted to have greater shoulder gaps than the fuel rods at end-of-life. I 1

l

c. Cladding Strain and Fatinue Analysis Cladding tensile strain occurs when the combination of cladding  ;

creepdown and pellet swelling have closed the diametral gap between ,.

the pellets and cladding, pomanent (unrecoverable) strain of the l l

l cladding takes place if the stress produced in the cladding by a pellet diameter increase exceeds the yield strength of the f cladding, or if the stress remains in the cladding long enough for I creep to occur.

i 1

! I 1

11-l

Fatigue damage occurs due to repeated application of cyclic stress levels above a certain value, known as the endurance limit.

Materials testing is used to establish both the endurance limit and the critical number of cycles.

The design analyses show that the unrecoverable circumferential strain is less than 1% and the cumulative fatigue damage does not exceed 0.8 for all rods,

d. Maximum Puel Rod Internal Pressure The fuel rod internal gas pressure will increase with burnup as fission gases are released from the fuel matrix and as internal void space is filled by fuel swelling and clad creepdown. In addition, the hot gas pressure is dependent on the operating power level due to relative thermal expansion of the fuel rod components.

The results of the design analyses indicate:

1. The primary stress in the cladding will not exceed the design stress limit.

I 2. The computed fuel rod internal hot gas pressure remains below possible coolant pressure values throughout its life-time (i.e.,

gas pressure is less than 2050 psia). The maximum pressure is insensitive to the operating coolant system pressure, so that the relative margin to exceeding coolant system pressure increases as the coolant system pressure increases.

3.2.2 Thermal Design I

The thermal effects analysis encompassed a study of fuel rod response cs a function of the detailed cycle burnup and power. The fuel rod types and pcwer histories examined in detail include the maximum power rod of each fuel type. The calculation methodology is the same as that employed in the previous cycle.

Fuel thermal calculations were performed for all fuel types using the CAPEX (21) computer program. The GAPEX code calculates pellet-to-clad gap ccnductance from a combination of theoretical and empirical models which predict fuel and cladding thermal expansion, fission gas release, pellet twelling, pellet densification, pellet cracking, and fuel and cladding thermal ccnductivity.

Figures 3.6 through 3.8 demonstrate the effect of Linear Heat G:neration Rate (LHGR) on fuel temperature at Beginning-of-Cycle (BOC) ccnditions. Figures 3.9 through 3.11 demonstrate the effect of LHGR on fuel temperature at End-of-Cycle (EOC) conditions. Table 3.4 lists UO melting 2

temperature and centerline temperature for the rods of interest at selected points in life and selected power levels. The calculated internal fuel rod pressures are less than Operating Coolant System pressure throughout cycle operation [(48), (51), and (61)).

The result of the fuel performance calculations indicates that the thernal performance is similar to that previously reported in (53), (33),

(15), and (7). The Specified Acceptable Fuel Design Limit (SAFDL) for fuel egnterline melt for each fuel type is indicated in Table 3.4. Tables 3.5, 3.6, and 3.7 provide a comparison of the maximum radial relative pin power for the recycled fuel types to the core maximum radial pin power during this cycle. The fresh fuel SAFDL is bounding for all fuel types since:

1. it contains the core-wide maximum power pin throughout the cycle, and

2. the SAFDL for any previously exposed fuel batch is greater than or equal to the ratio of the peak power of that batch divided by the peak power of the fresh fuel batch multiplied by the SAFDL for the fresh fuel batch.

3.2.3 C..ermal-Hydraulic Design Steady-state and transient DNBR analysis of the Cycle 9 core have been ptrformed using the COBRA-IIIC computer program (22), in the manner described in (8) and (9), and as described below. The models reflect the intended Cycle

9 coolant conditions and powar distributiens, thz assambly flow distribution due to differences in hydraulic characteristics and inlet flow maldistribution, and the specific geometry of the Maine Yankee fuel assemblies.

A new eighth core COBRA-IIIC model was used to determine hot assembly cnthalpy rise flow factors. This model explicitly represents each fuel casombly in the one-eighth (1/8) core in the specific location it will reside fcr Cycle 9 operation, and accounts for the differences in hydraulic characteristics between the CE and the ENC fuel assemblies. In Cycle 8 (53),

cnly the low flow region of the core was explicitly represented due to a limitation in the dimensional array size in the COBRA-III-C code. However, fer this calculation, the dimensional array size was expanded to allow cxplicit representation of the entire one-eighth core. Test cases were run with the revised COBRA-III-C code. The results were compared to results from the original code and were found to be consistent. The inlet flow maldistribution imposed on this model is based on the results of flow measurements taken in scale model flow tests of the Maine Yankee reactor vsssel reported in (36) and the FSAR (2). The assumed hot assembly enthalpy rise flow factor was 1.0 for the fresh CE fuel assemblies at all power distributions. A 0.964 enthalpy rise flow factor is applied to all ENC sesemblies due to higher spacer loss coefficients relative to the CE fuel.

These factors are applied to the inlet mass velocity in the hot channel model in predicting DNB performance.

The potential effects of fuel rod bow on thermal-hydraulic performance has also been evaluated for Cycle 9 operation. Using the channel closure correlation in (54), the maximum channel gap closure due to fuel rod bowing for the CE fuel assembly with the highest burnup during Cycle 9, the single E-16 assembly, was calculated to be 21.9%. Tests performed at Columbia (24) indicate that a degradation in DNB performance is not experienced until channel closures exceed 50%. Therefore, no penalty is required for fuel rod b:w considerations.

Allowances for rod pitch, bow and clad diameter variations for the ENC fuel are accommodated as follows. Allowances for manufacturing tolerances on red pitch and clad diameter, if considered in the most adverse situation, wruld result in a maximum channel closure in the vicinity of 10%. Using the

-p- - - , - y ,-,-,,-,-----,-w e- , , , - ,7- ,,-.,.e-, . , - , - - we , vs--.-me_--- ,-----e---.----.e---,-w,-----,,---ee.r---- e----

methodology of (34), the maximum channel gap closure due to fuel rod bowing fcr the ENC fuel assembly with the highest burnup during Cycle 9 is less than 30%. Therefore, a flow factor of 1.0 is justified for the ENC fuel in Cycle 9 cince the channel closure resulting from rod pitch, bow and clad diameter c:nsiderations for any ENC fuel during Cycle 9 will be less than 50%.

Table 3.8 contains a list of the pertinent thermal-hydraulic design p;rameters used for both safety analysis and for generating Reactor Protection System (RPS) setpoint information. The list also includes the corresponding thermal-hydraulic parameters from (3), and (53), Cycle 8, for comparison.

J TABLE 3.1 MAINE YANKEE CYCLE 9 ASSEMBLY DESCRIPTION Number Initial of Fuel Initial Number of mg B-10 Number of Total Total Assembly Exposure Rods per w/o U-235 Shim Locations per inch Assemblies Shim Fuel Designation History Assembly Fuel per Assembly in Shims in Core Locations Rods E-16 Cycle 2 160 2.52 16 29.0* 1 16* 160 L-0 Cycle 7,8 176 3.30 0 - 16 0 2,816 L-4 Cycle 7,8 172 3.30 4 23.8 12 48 2,064 L-8 Cycle 7,8 168 3.30 8 23.8 40 320 6,720 L-12 Cycle 7,8 164 3.30 12 23.8 4 48 656 M-0 Cycle 8 176 3.30 0 - 8 0 1,408 M-4 Cycle 8 172 3.30 4 23.8 28 112 4,816 M-8 Cycle 8 168 3.30 8 23.8 36 288 6,048 N-0 Fresh 176 3.30 0 - 4 0 704 N-4 Fresh 172 3.30 4 23.8 24 96 4,128 N-8 Fresh 168 3.30 8 23.8 44 352 7,392 C:ra Totals 217 1,280 36,912 8 E-16 shims replaced with water- filled rods for Cycle 9 I

I TABLE 3.2 MAINE YANKEE CYCLE 9 CORE LOADING Uranium per Uranium Exposure Assembly Number of Assembly Total at BOC*

TYoe Assemblies (KGU) (KGU) (NWD/MT)

E-16 1 353.7 354 20,184 L-0 16 381.1 6,098 21,192

~L-4 12 372.5 4,470 27,061 L-8 40 363.8 14,552 26,040 L-12 4 355.1 1,420 28,546 M-0 8 381.1 3,049 11,498 M-4 28 372.5 10,430 12,273 M-8 36 363.8 13,097 14,913 N-0 4 388.7 1,555 0 N-4 24 379.9 9,118 0 N-8 -44 371.0 16,324 0 80,467 12,865 0 Based on End-of-Cycle 8 at 12,000 MWD /MT 9

TABLE 3.3 Mechanical Design Features of Cycle 9 Fuel Type E Types L and M Type N Fuel Vendor C-E ENC C-E Fuel Assembly Overall length 156.718* 156.718 156.718 Spacer grid size (maximum square) 8.115 8.115 8.115 Number of zircaloy grids 8 0 8 Number of inconel grids 1 0 1 Number of bimetallic grids 0 9 0 Fuel rod growth clearance 1.021 1.300 min. 1.600 Fuel Rod Active fuel length 136.7 136.7 136.7 Plenum length 8.575 8.8 8.375 Clad OD 0.440 0.440 0.440 Clad ID 0.384 0.378 0.384 Clad wall thickness 0.028 0.031 0.028 Pallet OD 0.3765 0.370 0.3765 Pallet length 0.450 (62) 0.450 Dish depth 0.023 0.008 0.021 Clad material Zr-4 2r-4 Zr-4 Initial pellet density 95.0% 94.0% 94.75%

Initial pressure (15) (62) (61)

Poison Rods Overall rod length 146.513 146.500 146.629 Clad OD 0.440 0.440 0.440 Clad ID 0.388 0.378 0.384 Clad wall thickness 0.026 0.031 0.028 Pallet OD 0.376 0.353 0.362 Clad material 2r-4 Zr-4 Zr-4 l

i l *All length dimensions are in inches l C-E - Combustion Engineering ENC - Exxon Nuclear Corporation l '

l

TABLE 3.4 Centerline and U02 Melt Temperature Comparison Melt Centerline Fuel Ter.perature LHGR Temperature ruel Type Vendor ( F) (kW/ft) ( F)

BOC E C-E 4904 19 4440 20 4654 21* 4858 L ENC 4852 19 4421 20 4635 21* 4840 M ENC 4941 19 4332 20 4547 21 4753 21.9* 4932 N C-E 5049 19 4418 20 4572 21 4718 22 4858 23* 5022 EOC E C-E 4827 19 4477 20* 4691 21 4894 L ENC 4760 19 4555 19.9* 4748 20 4767 21 4968 M ENC 4843 19 4439 20 4653 20.9* 4839 21 4856 N C-E 4932 19 4253 20 4467 21 4672 22 4871 22.3* 4928 OL:!GR kW/f t SAFDL Limit E!sC Exxon Nuclear Corporation C-M Combustion Engineering

<! -19 Y

e

TABLE 3.5 MAINE YANKEE CYCLE 9 RATIO OF MAXIMUM RADIAL RELATIVE PIN POWERS MAXIMUM IN TYPE E FUEL TO MAXIMUM IN CORE Rodded Condition HFP, Equilibrium Conditions of ARO Ratio of Maximum Radial Relative Pin Powers Regulating BOC BOC MOC EOC Banks Inserted 50 mwd /Mt 500 mwd /Mt 6K mwd /Mt 14K mwd /Mt tRO 0.667 0.706 0.711 0.710 B:nk 5 0.500 0.512 0.527 0.493 Banks 5 + 4 0.243 0.255 0.250 0.220 I

B nks 5 + 4 + 3 0.267 0.284 0.271 0.227 l

B nks 5 + 4 + 3 + 2 0.206 0.224 0.207 0.159 f

B:nks 5 + 4 + 3 + 2 + 1 0.342 0.368 0.336 0.266 i

l TABLE 3.6 MAINE YANKEE CYCLE 9 RATIO OF MAXIMUM RADIAL RELATIVE PIN POWERS MAXIMUM IN TYPE L FUEL TO MAXIMUM IN CORE Rodded Condition HFP, Equilibrium conditions of ARO Ratio of Maximum Radial Relative Pin Powers Regulating BOC BOC MOC EOC Banks Inserted 50 Ed/Mt 500 Wd/Mt 6K Wd/Mt 14K Wd/Mt ARO 0.801 0.837 0.802 0.747 Bank 5 0.866 0.880 0.853 0.746 Jtnks 5 + 4 0.732 0.732 0.735 0.722 Jcnks 5 + 4 + 3 0.741 0.744 0.744 0.716 Jtnks 5 + 4 + 3 + 2 0.706 0.715 0.715 0.678 Jcnks 5 + 4 + 3 + 2 + 1 0.764 0.768 0.758 0.726 t

i l

l l

l l

i l

TABLE 3.7 MAINE YANKEE CYCLE 9 RATIO OF MAXIMUM RADIAL RELATIVE PIN POWERS MAXIMUM IN TYPE M FUEL To MAXIMUM IN CORE Rodded Condition HFP, Equilibrium Conditions of ARO Ratio of Maxinum Radial Relative Pin Powers R:gulating BOC BOC MOC EOC Banks Inserted 50 MWD /MT 500 mwd /Mt 6K mwd /Mt 14K mwd /Mt 20 0.920 0.916 0.899 0.856 lank 5 0.904 0.902 0.873 0.843

nks 5 + 4 0.930 0.922 0.890 0.853 lanks 5 + 4 + 3 0.917 0.911 0.889 0.847 lanks 5 + 4 + 3 + 2 0.938 0.940 0.918 0.860 lanks 5 + 4 + 3 + 2 + 1 0.955 0.958 0.933 0.860 f

l l

l l

I I

1 i

i l

l

T~blo 3.8 Maine Yankee Cycle 9 Thermal-Hydraulic Parameters at Full Power General characteristles Units cycle 1 Cycle 8 Cycle 9 Total Heat Output MWT 2630 2630 2630 106 Btu /hr 8976 8976 8976 Fraction of Heat Generated in Fuel Rod 0.975 0.975 0.975 Pressure Nominal psig 2235 2235 2235 Minimum in Steady-State psig 2185 2060 2060 Maximum in Steady-State psig 2285 2260 2260 Design inlet Temperature (steady-state) 0F 554 546-554 548-556 Total Reactor Coolant Flow (design) 106 lb/hr 134.6 136.0-134.6 135.8-134.2 E Coolant Flow Through Core (design) 106 lb/hr 130.7 132.1-130.7 131.9-130.3 Y Hydraulic Diameter (nominal channel) fL 0.044 0.044 0.044 Average Mass Velocity 106 lb/hr-ft2 2.444 2.47-2.444 2.46-2.436 Pressure Drop Across Core (design flow) psi 9.7 10.35 10.18 Total Pressure Drop Across vessel (Based on nominal dimensions and design flow) psi 32.4 33.1 32.9 Core Average Heat Flux

• Btu /hr-ft2 178,742 180,851 182,184 Total Heat Transfer Area

• ft2 48,978 48,392 48,038 Film Coefficient at Average

?

Btu /hr-ft OF 5,640 5,636 5,636 ConditLons Maximum Clad Surface Temperature OF 656 657 657 Average Film Temperature Difference OF 31.7 32.1 32.3 Average Linear Heat Rate of Rod

• kW/ft 6.03 6.10 6.15 Average Core Enthalpy Rise Btu /lb 68.7 68.7 68.7

Table 3 8 (Continued)

Haine Yankee Cycle 9 Thermal-Hydraulic Parameters at Full Power Cycle 3 Cycle 8 Cycle 9 I

Calculational Factors ,CE ENC M EBC g Engineering Heat Flux Factor 1.03 1.03 1.03 1.03 1.03 Engineering Factor on Hot Channel

! Heat Input 1.03 1.03 1.03 1.03 1.03 Flow Factors Inlet Plenum Bon-Uniform Distribution 1.05 1.00 1.00 1.00 1.00 Rod Pitch Bowing and Clad Diameter 1.065 1.00 1.065 1.00 1.065

• Allows for axial shrinhage due to fuel densification.

i t

1 l

1 l

1 I

i i

e

FIGURE 3.1 Maine Yankee Cycle 9 Burnable Poison Shim Assembly Locations O O O O I

O O O O X R O O O O X X O O O. O 16 Water Rod Arsembly (E-16) 4 Shim Assembly (L-4, M-4, S -4) i X X X X __ __

X X X X X X X

X X X X

~

X 8 Shim Assembly (L-8, M-8, N-8) 12 Shim Assembly (L-12) h Water-filled rods i B4 C in A123 0 shim rods FIGURI 3.2 MAINE YANKEE CYCLE 9 ASSEMBLY LOADING PATTERN FUEL TYPE E E-16 62 --Cycle 9 location 36 -Cycle 2 Location 0 --Rotation Index*

L,M L-8 3 --Cycle 9 Location L-8 1 L-0 2 43 --Cycle 8 Location 10 2 0 --Rotation Index** 2 0 N N-4 5 --Cycle 9 Location L-8 3 M-4 4 N-4 5 N-4 6 N-8 7 43 5 0 2 L-8 8 N-4 9 N-8 10 M-0 11lM-8 12 M-8 13 42 4 22 17 0 0 0 0 L-8 14 N-0 15 M-4 16 L-0 17 N-8 18 L-4 19 N-8 20 36 14 35 26 0 1 1 1 L-8 21 N-4 22 M-4 23 M-8 24 N-8 25 L-8 *6 M-4 27 M-8 28 51 8 53 1 6 33 0 3 2 1 1 2 M-4 29 N-8 30 L-0 31 N-8 32 L-4 33 N-8 34 L-8 35 M-4 36 37 50 28 3 15 2 3 0 1 2 N-4 37 M-0 35 N-8 39 L-8 40 N-8 41 M-8 42 M-8 43 L-12 44 29 38 13 19 24 0 3 2 3 0 L-8 45 30 2 M-8 47 L-4 48 M-4 49 L-8 50 M-8 51 M-8 52 N-8 53 N-4 46 9 40 46 21 48 17 L-0 54 0 3 3 3 1 2 54 0 N-8 55 M-8 56 N-8 57 M-8 58 M-4 59 L-12 60 N-8 61 E-16 62 31 33 15 24 36 0 2 2 0 0 0 clockwise multiple of 900 relative to cycle 2 location in quadrant co clockwise multiple of 90 relative to cycle 8 location in quadrant

Figuro 3.3 Maine Yankee Cycle 9 Calculated Assembly Exposures at BOC CYCLE AVERAGE EXPOStftE = 0.0 CORE AERACE EIPOSLRE = 1286e.7 MAI ASSEMBLY EXP05UIE = 29744.0 t LOC 1 CORE POSIT! Cpi / ASSEMBLY IUtBER . I A14 1 1 A12 21 FtEL TYPE . . . . . . . . . . . . . . . . . . . . . I L-8 I L-0 1 CEA BAft: TYPE ................. !

l l AST9LY IUMP (!WD/MT) ...... I 29744.0 1 17622.0 1 1 B17 3 i B16 4 I B15 5 i B13 6 I Bil 71 i L-8 i M-4 i N-4 I N-4 ! N-8 !

1 1*Ce 1 Ie1e ! I I I I I  !  !

I 27933.0 1 12938.0 1 0.0 1 0.0 1 0.0 1 1 C18 8 I C17 9 I C16 10 I C15 11 1 Cl3 12 I Cll 13 I I L-8 i N-4 ! N-8 I M-0 I M-8 i M-8 I I I eAe ! IeCe ! I e5e I

!  !  !  !  !  ! I

! 28794.0 1 0.0 1 0.0 I 11498.0 1 13604.0 1 15506.0 I I D19 14 1 018 15 1 017 16 1 016 17 1 015 18 1 D13 19 I Dil 20 I I L-8 1 N-0 I M-4 I L-0 1 N-8 I L-4 I N-8 1 1 i e5*1 1eAs ! I e3e !  !

I I I I I 1  !  !

I 28794.0 1 0.0 1 9121.0 1 24761.0 1 0.0 1 26703.0 1 0.0 I i'E20211E19221E18231E1724!E1625IE1526IE1327IEll 28 I I L-8 I N-4 I M-4 1 M-8 1 N-8 1 L-8 i M-4 1 M-8 I I I eA* I I I I e2* I I I I I I I I 1 1 1 1 1 27933.0 1 0.0 1 9121.0 1 14526.0 1 0.0 1 27734.0 1 13807.0 1 14907.0 t i F20 29 I F19 301 F18 311 F17 321 F16 33 I F15 341 F13 351 Fil 36 I I M-4 I N-8 1 L-0 1 N-8 i L-4 I N-8 I L-8 1 M-4 I I eCe ! I eA* 1 Ie5* I I I I I I I I I I  !  ! I I 12938.0 1 0.0 1 24761.0 1 0.0 1 27778.0 1 0.0 1 15994.0 1 14177.0 1 1020 371019 381018 391017 401016 411015 421013 431011 441 1 N-4 1 M-0 I N-8  ! L-8 I N-8 I M-8 1 M-B I L-12 1 1 1eCe ! I e2e i I eBe I I e4e i I H21 45 1 1 I I I I I I  !

I L-8 I 0.0 1 11498.0 1 0.0 1 27734.0 1 0.0 1 15570,0 1 15498.0 1 28546.0 1 1

1 1J2046IJ19471J1848IJ1749IJ1650IJ1551!J13521Jtt53l

! 29744.0 I N-4 I M-8 1 L-4 I M-4 1 L-8 i M-8 I M-8 1 N-8 1 Iele ! I e3e ! I- 1 IeBe i 1 1 K21 54 I I I I I I 1 I L-0 1 0.0I13604.0126703.0113807.0115994.0115498.0115506.0l 0.0 1 l

I I L20 55 i L19 56 I L18 57 I Ll7 58 I Ll6 59 I Ll5 60 I L13 61 1 Lil 62 1 117622.01 N-8 I M-8 I N-8 I M-8 1 M-4 I L-12 i M-9 I E-16 1 1 Ie5e i 1 1 1e4e I Ie5e !

I I I i  !  ! 1 1 I I 0.0 1 15506.0 1 0.0I14907.0I14177.0128546.01 0.0120184y BATCH M7CH 9ATCH AYEMOE IUSER ID EIPOM 1 E-16 20184.0 2 L-0 21191.5 3 L-4 27061.3 4 L-8 26039.8

! 5 L-12 25546.0 I 6 M-0 11498.0 7 M-4 12272.7 8 M-9 19911.2 9 50 0.0 10 N-4 0.0 11 N-8 0.0 Figure 3.4 Maine Yankee Cycle 8 Burnup Distribution by Assembly INCA versus Predicted Near 6,000 MWD /MT Cycle Exposure Assembly Type and IWCA Location . . . . . . . . K-8 8 L-0 21 INCA at 99% Power. 5992 mwd /Mt . ....... .30375 13673 Predicted at 100% Power, 6000 mwd /Mt . .. . , 30004 13945 Percent Difference .............. -1.2 2.0 L-8 15 M-0 31 M-4 11 M-4 25 K-8 4 12507

~

5752 6591 6770 32336 12178 5773 6462 6831 31917

-2.6 0.4 -2.0 0.9 -1.3 M-4 16 M-8 33 L-8 13 L-8 28 K-0 7 M-8 20 4452 6659 22260 20877 28447 7677 4350 6559 22904 20667 27418 ss47

-2.3 -1.5 2.9 -1.0 -3.6 -0.4 M-4 34 K-0 14 M-8 30 K-0 10 M-8 24 K-4 3 7020 27625 7525 29669 7589 32608 6919 27503 7615 30051 7696 32755

-1.4 -0.4 1.2 1.3 1.4 0.5 L-12 32 K-0 12 L-4 27 K-0 6 L-4 19 21924 31862 20341 28972 21611 21810 31681 20229 28652 21352

-0.5 -0.6 -0.6 -1.1 -1.2 M-8 29 K-0 9 L-0 23 L-8 2 7277 29474 18214 22039 7413 29747 18364 22435 1.9 0.9 0.8 1.8 Fuel Exposure Exposure L-8 26 L-8 5 K-8 18 Irpe IUCA Predicted Difference (%) 23040 21841 31800 22556 21680 31862 E-16 25894 25723 -171 (-0.7%) -2.1 -0.7 0.2 K-0 29341 29175 -166 (-0.6%)

E-4 32608 32765 147 ( 0.5%) K-8 22 M-8 1 K-8 31292 31059 -233 (-0.7%) 31573 7143 L-0 15944 16154 210 ( 1.3%) 31510 7206 L-4 20764 20603 -161 (-0.8%) -0.2 0.9 L-8 20005 19985 - 20 (-0.1%)

L-12 21924 21810 -114 (-0.5%) E-16 17

M-0 5752 5773 21 ( 0.4%) 25894 M-4 6092 6029 - 63 (-1.0%) 25723 M-8 7294 7334 40 ( 0.5%) -0.7 C

re 18640 18708 68 ( 0.4%)

Cycle 5992 6000 8 ( 0.1%)

Figure 3.5 Maine Yankee Cycle 9 CEA Group Locations Regulating Shutdown Dual CEA Groups CEA Groups

~ "

5 (5A and 5B) C L-8 1 L-0 2 4 B 3 A 2

1 L-8 3 M-4 4 N-4 5 N-4 6 N-8 7 C 1 L-8 8 N-4 9 N-8 10 M-0 11 M-8 12 M-8 13 A C 5A L-6 14 N-0 15 M-4 16 L-0 17 N-8 18 L-4 19 N-8 20 SA A 3 L-8 21 N-4 22 M-4 23 M-8 24 N-S 25 L-8 26 M-4 27 M-8 28 A 2 M-4 29 N-8 30 L-0 31 N-8 32 L-4 33 N-8 34 L-8 35 M-4 36 C A 5B N-4 37 M-0 38 N-S 39 L-8 40 N-8 41 M-8 42 M-8 43 L-12 44 L-8 45 C 2 B 4 N-4 46 M-8 47 L-4 46 M-4 49 L-8 50 M-8 51 M-8 52 N-8 53 B

L-0 54 1 3 N-8 55 M-8 56 N-8 57 M-8 58 M-4 59 L-12 60 N-8 61 E-16 62 5A 4 5A

• non-scrammable CEA locations (Subgroup 5B) with full-strength CEAs replacing part-strength CEAs for Cycle 9

T* y*

4 2

N E 0 2

P Y

T )

L T F

E /

U W F K

[

9 R E G T E H L L

- 6R 1, R C S Y V -

N C O I

E E T E R R K U R N T E

_ A A R N Y E E G

E P M 2 N E 1 T I ' R A T E M E H N R I A L E R N E I T L N

E C 8 C

O B

4 0 0 0

0 0

5 e m = 0 1

C $_ Ma$@Es N5O

.Y

MAINE YANKEE CYCLE 9 BOC CENTERLINE TEMPERATURE VS LHGR; FUEL TYPE M s000 ts_

CD

[d 4000

\

M a -

M 3000 w 5

- / ~

~

N 3

z /

a i

4 8 12 16 20 24 LINEAR HERT GENERATION RATE (KW/FT)

NAINE YANKEE CYCLE 9 BOC CENTERLINE TEMPERATURE VS LHGR; FUEL TYPE L SIX0

/

c 8 axo B /

=

il 2 sE

' 5 E 3000 e 5

/ -

3 h"

8

/

1000 4 8 12 .16 20 24 LINEAR HERT GENERATION RATE [KW/FT1

"! r 4 2

N E 0 P 2 Y

T )

L T E F U /

F W K

(

9 R G E E H T L L 6A 1

C - R Y S C V N O

E E I E R T K U -

R N T R A A E Y R N E E E P G N M 2 I E 1 T A T R M E E H N

I R L - A R E E N T -

I N L E

C C

/ 8 O

E 4

o )

X 0 0 o 0 0 o G 0 0 s 4 3 1

~tu e[d ie aib =30i s s'

MAINE YANKEE CYCLE 9 E00 CENTERLINE TEMPERATURE VS LHGR; FUEL TYPE N 5000

/

c 8 <xo 8 /

e ,

a -

sE 5

E 3000 .-

5 s

/

M J

h" 8

/

sooo . . . . . . . . . . . .

i 8 12 16 20 24 LINEAR HEAT GENERATION RATE [KW/PT)

2! e 4 2

G L E 0 P 2 Y

T )

L T E F U /

F W K

(

9 R E G T E H / 6A L L 1 R C S Y V N C .

O I

E E T E R A K U R N T A E A R N Y E E G

E N

P M y 2 E 1 T I

T R A E M E -

H N R I

L A R E E N I

T N

L E

C - 8 0

0 E

- 4 0 O 0 o 0 0

0 5 " m = 0 1

E h dei't5s I al 3$5o s'

4.0 PHYSICS ANALYSIS 4.1 Fuel Management Maine Yankee Cycle 9 consists of irradiated and fresh fuel assemblies Cs described in Section 3.1.1. The core layout is given in Figure 3.2.

Cycle 9 is expected to attain a cycle average full power lifetime of 13,400 MWD /NT. A low-leakage loading pattern is employed, as described in Section 3.1.3.

4.2 pore Physics Characteristics The primary physics characteristics of the reference physics cycles (Cycle 3, Cycle 8 and Cycle 9) are given in Table 4.1. The Cycles 8 and 9 characteristics differ from those of Cycle 3 based on the following significant changes:

1) Increased fuel enrichment;

2) Low-leakage fuel management;

3) Increased use of burnable poison shims; and

4) Increased reactivity worth resulting from the reconfiguration of CEA Bank 5.

The impact of the these changes on the physics characteristics are discussed in the following sections.

4.3 Power Distributions Assembly relative power densities for Cycle 9 at Hot Full Power (HFP),

equilibrium xenon conditions are presented for unrodded and rodded (CgA Bank 5 iCserted) configurations at Beginning, Middle, and gnd-of-Cycle (BOC, NOC, ROC). Figure 3.5 shows the locations of the CgA groups.

The unrodded power distributions at BOC (500 KWD/MT), MOC (6000 MWD /MT) cnd EOC (14,000 MWD /MT) are presented in Figures 4.1 through 4.3. The rodded (CEA Bank 5 inserted) power distributions at BOC, MOC and EOC are presented in Figures 4.4 through 4.6. The unrodded radial peaking is comparable and the r:dded radial peaking is 2-4% increased relative to the Cycle 8 radial peaking.

The allowable unrodded radial peaking (with uncertainties) versus cxposure for Cycle 9 is included in the plant Technical Specifications for the purpose of comparison to measured values. This assures peaking will not Cxceed the values used in safety analysis. The values are shown in Figure 4.7 and show that the maximum radial peaking occurs early in the cycle.

The core power distributions are slightly asymmetric due to non-octant symmetric burnup gradients across the octant and quadrant boundary Cssemblies. The quadrant analycic presented overpredicts the slight asymmetry the full core will exhibit, providing a conservative analysis of the peaking cffects.

4.4 CEA Croup Reactivity Worths The CEA group configurationc were shown in Figure 3.5. The CEA group worths at l{FP are presented in Tabic 4.2 for Cycles 3, 8, and 9. The CEA Bank 5 modifications for Cycle 7 provided the large increase in CEA group worth compared to previous cycles. The low-leakage fuel management in these cycles C1so contributes to a general increasc in CEA worths due to power distribution weighting effects. In general, the regulating CEA group worths for Cycle 9 are comparable to those of Cycle 8, but the shutdown and total CEA group worths are less. This is primarily due to the placement of more fresh fuel in interior unrodded locations in Cycle 9 to improve the core leakage characteristics. CEA group reactivity worths are verified by the startup test program and the associated acceptance criteria.

4.5 poppler Resetivity Coefficients and Defects The fuel temperature, or Doppler, components of reactivity are l

presented in Tables 4.3 and 4.4 for nominal conditions in cycles 3, 8, and 9.

The total core average Doppler defect from 4000 F is given in Table 4.3 and l

the core average Doppler coefficient in Table 4.4. The valu:s in Cycles 3, 8, cnd 9 are similar. Uncertainties of 25% are conservatively applied to the [

coefficient and defect values prior to transient analysis applications, except  ;

cs noted below.

Separate core average Doppler defect calculations appropriate for application to the ejected CgA cases have been perforined for Cycles 7 and 8 and were discussed in detail in (50). The calculations utilise the three-dimension 1 pre-ejected configuration weighting to obtain a more representative Doppler defect. For Cycle 9 the CgA ejection methodology utilises the core average unrodded Doppler defects from Table 4.3 and a Doppler weighting factor technique, as described in detail in (55). An [

uncertainty of 15% is applied prior to transient analysis application of the Doppler defects for CgA ejection, which is unchanged from Cycle 8.  !

4.6 Moderator Reactivity Coefficients and Defects The Moderator Temperature Coefficients (MTCs) at nominal operating Hot Full Power (MFP) and Hot Zero Power (HZP), critical boron conditions are presented in Table 4.5 for Cycles 3, 8, and 9. Relative to Cycle 8, the Cycic 9 MTCs at DOC are more positive, due primarily to the increased BOC critical boron concentration resultin5 from lower leakage and more excess reactivity in the core. The end of Cycles 8 and 9 MTC values are similar. An uncertainty cf i 0.5 x 10 delta rho / F is conservatively applied to calculated MTC values for use in transient analysis. The startup test program demonstrates the validity of such values. f The moderator density defect curve used in the LOCA analysis infers specific NTC values in the operating range which must not be exceeded. New ,

NTC Technical specification limits are provided for Cycle 9 based on a new reference LOCA analysis moderator density defect curve.

The cycle 5 moderator density defect curve was the reference curve for l the Cycle 5 through 8 LOCA analyses and determined the Cycle 8 MTC Technical i specification limits of +0.39, +0.27, and +0.12 x 10 delta rho / F at 0, l 50, and 100% rated power, respectively. The new reference moderator density f defect curve generated for the Cycle 9 LOCA analysis utilises the results of l

l l i

l l

l l d: tailed two-dimensional PDQ r: dict analysis at solcetcd mod:rctor d:nsity conditions. This methodology change is discussed in detail in Section 4.10.

The new MTC Technical Specification limits for Cycle 9, presented in Figure 4.8, are consistent with the new moderator density defect curve and the MTC cssumptions used in safety analysis. The startup test program will demonstrate that these MTC limits are not exceeded.

The moderator defect appropriate to the scranned (ARI) less worst stuck CEA configuration is given in Table 4.6 for Cycles 3, 8, and 9. This defect curve yields a conservative moderator reactivity increase versus temperature ce density, while accounting for the effects of loss in total CEA worth and the worst stuck CEA. 9 tarting in Cycle 6 this calculation has been performed Ct BOC, high solubl. sron and EOC, no soluble boron conditions. The EOC moderator defect for Cycle 9 is less severe than the Cycle 8 moderator defect due primarily to the improved, lower-leakage core. Since the soluble boron concentration is the most significant factor in determining the moderator d:fect, a boron-concentration dependent minimum required shutdown margin was incorporated in the plant Technical Specifications for Cycle 7, as discussed in Section 4.9.5. An uncertainty of 15% is applied to the moderator defect values in cooldown transients from HZP. An uncertainty of 25% is applied in croldown transients from HFP, where moderator redistribution effects are an additional reactivity component. These uncertainties are unchanged for Cycle 9.

4.7 Soluble Boron and Burnable Poison Reactivity Effects The soluble boron and burnable poison shim reactivity effects are shown in Table 4.1 for Cycles 3, 8, and 9. The critical boron concentrations for Cycle 9 at BOC are greater than those of Cycle 8, due primarily to a lower c re leakage and greater excess reactivity in the core. There are more total burnable poison pins in Cycle 9 although there are a similar number of fresh poison pins and a similar burnable poison pin reactivity worth at BOC for Cycles 8 and 9. The inverse boron worths for Cycles 8 and 9 reflect the different soluble boron levels and core characteristics.

l l

L

4.8 Kinetics Parameters The total delayed neutron fractions and prompt neutron generation time fcr Cycles 3, 8, and 9 are presented in Table 4.1. The values are comparable and the differences reflect the effects of core average exposure and power weighting. Table 4.7 details the delayed neutron fractions and lifetimes by dalayed neutron group for Cycles 3 and 9 at HFP, All-Rods-Out.(ARO) etnditions. Kinetics parameters for HFP and HZP conditions, both unrodded and r:dded, are calculated for appropriate application in transient analysis cases cnd a 10% uncertainty is applied in a conservative manner.

4.9 Safety-Related Characteristics 4.9.1 CEA Group Insertion Limits The CEA group insertion limits are given in the Technical Specifications and Figure 4.9. The Power-Dependent Insertion Limit (PDIL) for CEAs provides for sufficient available scram reactivity at all power levels cnd times-in-cycle-life. It also specifies the allowable CEA configurations for analysis of dropped, ejected, and withdrawn CEAs. The CEA group insertion limits for Cycle 9 are more restrictive than for Cycle 8, resulting from increased radial peaking under rodded and post-CEA drop conditions.

The allowable CEA insertion is determined by the maximum of either:

1) the actual operating power level, or

2) the reference power level, given in Figure 4.10, which is the power level normally associated with the actual operating cold les temperature.

This definition is required to assure that sufficient available scram P

reactivity is maintained when operation deviates from the normal cold les temperatures. For Cycle 9, the normal cold leg temperature assumed at HFP has been increased 2 F from 550 F to 552 F. This increase is reflected in Figure 4.10 and is conservatively incorporated in all safety analysis.

4.9.2 CEA Ejection Results The calculated worths and planar radial maximum 1-pin powers resulting from the worst ejected CEAs for BOC and EOC are shown in Table 4.8 for Cycles 8 cnd 9. MFP and HZP conditions are considered for these comparable CEA insertion cases. No credit is taken for feedback effects in these sciculations. These calculations assume full CEA insertion of CEA Bank 5 at HFP and up to CEA Banks 5 + 4 + 3 at HZP, which are conservative relative to th7 allowable insertions at these power levels, given by the insertion limits cf Section 4.9.1. The Cycle 9 values are generally decreased relative to Cycle 8 due primarily to the decreased reactivity of fresh assemblies in the vicinity of the limiting ejected CEA locations (INCA locations 20 and 34, as id:ntified in Figure 3.4). The Cycle 9 HFP ejections are the only case with more severe results than Cycle 8.

4 A change in methodology for the analysis of the CEA ejection is d: scribed for Cycle 9 in (55). As part of this change, CEA ejection results cre calculated from the actual partial CEA insertions permitted by the CEA group insertion limits, rather than assuming conservative, full insertion CEA b nk configurations. This results in less severe higher power CEA ejections, since CEA insertion is normally very restricted at high powers. The analysis method for calculation of these results is detailed in Section 4.11.2.

The results of the CEA ejections from partial insertions are presented in Table 4.9. Specific CEA insertions along the Power-Dependent Insertion Limit (PDIL) are chosen and the worst CEA ejection location determined. A p:wer level which is either determined by or conservatively high relative to f the PDIL is used in the analysis, as indicated. The maximum ejected worth and l maximum 1-pin radial peak are determined. The maximum axial peak in the core f is determined and conservatively assumed to be coincident at the maximum r: dial peak location in the analysis. The raximum axial peak increases with d: creasing power level since more poritive symmetric offset and, hence, more ttp-peaked shapes are allowed by the symmetric offset trip limits.

Comparison cf the full and partial insertion CEA ejection results are summarized in Tables 4.8 and 4.9. Due to res;riction of CEA insertion at I

higher power levels, results comparable to the full insertion, MFP values are

n:t reached until near the 60% power level range. The results at the lower power levels are comparable to the full insertion HZP valves. The results in T;ble 4.9 are utilized in the CEA ejection analysis for Cycle 9.

4.9.3 CEA Drop Results 4.9.3.1 Desian Analysis Results The calculated worths of the most limiting dropped CEAs for Cycles 8 cnd 9, with the resulting maximum 1-pin radial powers, are given in Tables 4.10 and 4.11 for BOC and EOC. Since Cycle 4, this analysis has utilized a local pinwise Doppler feedback methodology which was verified by a special ct-power CEA drop test performed during the Cycle 4 startup physics tests (27).

The calculations are performed for all CEA drops at 20% increments in power level. CEA drops from ARO, Bank 5 Banks 5 + 4 and Banks 5 + 4 + 3 in cre those considered for Cycle 9 based on conservatively assumed CEA insertion limits with power level. CEA drops from ARO and Bank 5 are the most important due to the higher power levels permitted in these CEA configurations.

The CEA drop results in Tables 4.10 and 4.11 are compared for Doppler feedback conditions of 80% of rated thermal power. Detailed separate cnvelopes of maximum percent increase in radial peaking versus reactivity worth of the dropped CEA are calculated for various power levels and presented in Figure 4.11. The resulting peaking increase is slightly higher than cycle 8, also presented in the figure for the 100% power case.

In the design analysis for dropped CEAs, the two-dimensional radial peaking increases in Figure 4.11 are combined with the most limiting radial cnd axial peaking allowed by the symmetric offset limits to obtain total peaking for the given power level. This peaking, increased by 10% for uncertainties, is accommodated in the RPS setpoint generation.

4.9.3.2 Post-CKA Drop Restrictions Analyses for Cycle 7 were performed and presented in (50) to determine the required rate of power level reduction which the design snalysis method in S:ction 4.9.3.1 would bound. Three-dimensional nodal calculations were used to determine the required rate of power reduction. The results indicated that the following actions are required to maintain the core within the limits of the design analysis following a dropped CEA:

1) Decrease thermal power by at least 10% of rated power within one-half hour;

2) Decrease thermal power by at least 20% of rated power within one hour;

3) Maintain thermal power at or below this reduced power level; and

4) Limit CEA insertion to the insertion level corresponding to the pre-drop thermal power.

The power reductions described above assure that proper limits are maintained fcr operation up to four hours post-drop. The plant Technical Specifications r:flect these restrictions. Similar calculations are performed for Cycle 9 to quantify the peaking increases under these power level restrictions for proper incorporation in the RpS setpoint generation.

The worst of the core peripheral (CEA Type A dual) or core central (CEA Type B dual) CEA drop peaking increases are presented in Table 4.12 for Cycles 8 and 9. These peaking increases are accommodated in the same manner as the d: sign analysis instantaneous peaking increases in Section 4.9.3.1.

4.9.4 Available Scram Reactivity The available scram reactivity from both HFp and HZp conditions at BOC cnd EOC is tabulated in Table 4.13. Allowances for the worst stuck CEA and the power dependent insertion limit for CEAs are included. The CEA programming allowance corresponds to the loss in available scram reactivity due to movement of all CEAs a maximum of 3 inches (4 steps) into the active etre.

The available scram reactivity with uncertainties at EOC is less for Cycle 9 by 0.70% delta rho at HFP and 0.52% delta rho at HZP conditions r:1stive to Cycle 8. This is due primarily to the decreased total CEA worth, which is discussed in Section 4.4.

The required scram reactivity at the HZP condition is determined from the requirements of the steam line rupture analysis in Section 5.5.1 and the cther safety analyses in Section 5. The required scram reactivity at HZP must be sufficient to prevent a return-to-criticality following the most limiting steam line rupture event from HZP. It also must be greater than assumed in Cther safety analyses from HZP. The available scram reactivity at HZP, from Tcble 4.13, nust be greater than the required scram reactivity at HZP.

In addition, the required scram reactivity at HZP, when added to the cdditional scram reactivity provided by the CEA insertion limits versus power from Figure 4.9, must be suf ficient to prevent a return-to-criticality fOllowing a steam line rupture event from any power level. It nust also be greater than the value assumed in other safety analyses from at-power ccnditions.

The steam line rupture analyses are performed from both HFP and HZP c:nditions, as discussed in Section 5.5.1. They explicitly account for the moderator defect as a function of moderator density, and Doppler defect as a function of fuel temperature, with the uncertainties stated in Sections 4.5 cnd 4.6. Other safety analyses are also performed from both HZP and HFP c;nditions. The CEA insertion limits versus power are designed to provide increased available scram reactivity proportional to the increased power level. This assures that intermediate power level conditions are covered by cnalysis of the HZP and HFP cases.

The steam line rupture analysis provides the minimum required worth in CEAs for cooldown events f rom HFP and HZP conditions to maintain suberiticality. In addition, other safety analyses have implicitly assumed Cinimum required worth in CEAs, as stated in Section 5.1.4. The minimum required worths in CEAs are compared, in Table 4.14, to the available scram reactivity from Table 4.13. The table demonstrates that, in each condition i

i and time in cycle life, the available scram re ctivity is ges0 tor thin ths r: quired scram reactivity.

l Available scram reactivities with uncertainties are compared in the l

' ttble to the values assumed in the analyses. A 10% uncertainty component is included in the determination of the minimum required worth in CEAs for the steam line rupture analysis, as part of the statistical combination of uncertainties described in (57). Compliance with the startup test criteria on CEA worths demonstrates the available scram reactivity in Table 4.13. As l

such, it also demonstrates CEA worth in excess of the required scram r: activities.

l l The minimum required worth in CEAs for the steam line rupture analysis is calculated at typical beginning and end-of-cycle conditions, corresponding to Cycle 9 RCS soluble boron conditions of 1015 and 0 ppm, respectively. The

baron concentration determines the magnitude of the moderator temperature d

fect and has the most direct impact on the minimum required worth in CEAs.

The result is that the minimum required shutdown margin, as discussed in the n:xt section, can be expressed as a function of RCS soluble boron ccncentration in the Technical Specifications, r

4.9.5 Shutdown Martin Requirements shutdown margin is defined as the sum of:

l l

1) the reactivity by which the reactor is suberitical in its present condition, and l

2) the reactivity associated with the withdrawn trippable CEAs less i

i the reactivity associated with the highest worth withdrawn trippable CEA.

l For a critical reactor, the shutdown margin must be maintained by

sufficient available scram reactivity. The required and available scram reactivity comparison in Tabic 4.14 is the result of calculations which demonstrate adequate shutdown margin by bounding all the critical operating ccnditions for Cycle 9. Adequate shutdown margin exists, provided the CEA l

_ _ , . . _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ . . _ . . _ - . . _ . . ._-___-.-.._m._ _ . - _ . - - . . _ _ .

insertion limits and assumptions inherent in them are fulfilled. These cssumptions are:

1) the available scram reactivity calculations,

2) the operability of all trippable CEAs, and

3) the CEA drop time to 90% of full insertion in less than 2.7 seconds.

The shutdown margin requirement is expressed in the Technical Specifications, as shown in Figure 4.13. The equation representation in the figure allows for calculation of the minimum required shutdown margin for any RCS boron concentration, power level and core inlet temperature. The Cycle 9 cquations are unchanged from Cycle 8.

This shutdown margin representation is demonstrated, in Figure 4.12, to b:und the required scram reactivities of Tabic 4.14 from both HFP and HZP ccnditions. Based on the discussion in Section 4.9.4, meeting the startup test criteria on CEA worths demonstrates the calculated available scram r: activity with uncertainties and thus demonstrates compliance with the r: quired shutdown margin.

The minimum required shutdown margin is given for selected power icvels in Figure 4.13 and the Technical Specifications to provide a well-defined r;quirement as a function of key plant parameters. This specification permits the development of procedures which preserve the minimum required shutdown margin. Under normal operating conditions, the CEA insertion limits provide such assurance. In the event of an inoperable or slow CEA, such procedures would apply.

4.9.6 Auamentation Factors The set of augmentation factors applied to Cycle 9 has been determined for noncollapsed fuel clad using the CE calculational model described in (28). This is the same methodology applied to CE fuel in Cycles 1 through 4.

The ENC methodology has been used in Cycles 5 through 8. Augmentation factors have been conservatively calculated for the cycle by statistically combining restrictive single gap peaking factors and the worst (i.e., flattest) radial pin power census during the cycle from Table 4.16. The pin census at this time in cycle life is compared to the most limiting Cycle 8 pin census the table. The Cycle 9 pin census shows a generally large number of pins in the higher power intervals.

The augmentation factors are calculated for a 1% increase in fuel density during irradiation. This is the densification characteristic of modern CE manufactured fuel which is considered applicable to Type N fuel.

Rosintering test data for the Types L and M fuel assembly pellet lots also justify use of augmentation factors appropriate to a 1.0% fuel density change. The augmentation factors for Cycles 8 and 9 are given in Table 4.15.

The Cycle 8 augmentation factors assumed a change in fuel density of 1.5%

based on the Type K fuel, which has been discharged for Cycle 9. Considering the compensating differences in density change and pin census data, the augmentation factors are comparable. l The augmentation f actors are incorporated as a power spike penalty in all calculations of core power to incipient fuel centerline melt, as part of the RPS setpoint analyses.

4.10 Pressure Vessel Fluence A program for reduction in pressure vessel fluence has been in place for Maine Yankee since Cycle 7 to limit the potential for Pressurized Thermal Shock (PTS) conc e rns . The Cycles 7, 8, and 9 core designs have been a progression of lower leakage loading patterns with particular emphasis on reduced fluence in the area of the critical longitudinal weld, which is positioned at 10 degrees from a perpendicular line to the core shroud fla s.

The core shroud flats are the core boundary lines defined by assembly numbers 1 and 2 (or 45 and 54)'in Figure 3.2. The 0 to 10 degree reston is the high fluence area.

The program for fluence reduction has been detailed in (59) and (60),

with target fluence reductions for Cycles 7 and 8 and subsequent cycles relative to the Cycle 6 fluence level as a reference. The Cycle 6 out-in fuel management provided relative fluences in the 0-10 degree region which were similar to the fluence history accumulated from Cycles 1, 1A, and 2 through 5.

The fluence reductions, expressed as flux reduction factors relative to the Cycles 1 through 6 fluence history, are shown in Table 4.17. The target fluence reductions in (59) for Cycles 7, 8, and 9 are compared to the actual c:re design fluence reductions obtained by a view-factor weighting technique cf the average quarter-assembly powers calculated for the cycles. The result is that the cumulative fluence reduction factor target to end-of-Cycle 9 has been achieved for both the 0 and 10 degree azimuthal angles. At the critical longitudinal weld at 10 degrees, the target cumulative fluence reduction is 14% relative to the case in which no fluence reduction measures were instituted. This is achieved by a Cycle 9 flux reduction factor in excess of 50%. Similar flux reduction factors are expected for future cycles to meet the targets set forth in (59).

4.11 Methodology and Methodology Revisions 4.11.1 Summary of Physics Methodology Documentation A summary of the reference report and supplemental documentation for the application of physics methodology to Maine Yankee since Cycle 3 ic given in Table 4.18. The reference physics methodology report is YAEC-1115 (14).

There are two changes to the reactor physics methodology included for Cycle

9. The methodology changes represent an extension of existing methods and no new calculational methods are proposed.

4.11.2 CEA E3ection Results from Partial Insertions The results of the CEA ejections from partial insertions allowed by the PDIL are discussed in detail in Section 4.9.2. A complete discussion of the CEA ejection physics methodology is contained in (50). As part of the CEA ojection methodology change for Cycle 9, described in (55), the SINULATE nodal model is used to represent the three-dimensional, full core CEA ejection from partial CEA insertion conditions. This is the same SINULATE nodal model described in (14), (7), and (27) in application to Maine Yankee. As with the two-dimensional PDQ analysis, the CEA ejection calculation is a static i

f I

calculation in which there are no changes in feedback effects due to the 4 ojection itself. The local moderator and fuel temperatures from the I pre-ejection configuration are fixed prior to the CEA ejection. This results in conservative CEA ejection consequences.

The CIA ejections from full insertions in Table 4.8 analyzed with PDQ cre also analysed with SIMULATE. SIMULATE is then used to determine the changes in CEA ejection worth and peaking for partial insertion relative to I the full insertion condition. In this way, the accuracy of the more detailed l

! PDQ analysis is preserved. SIMULATE provides an axial distribution for the j core average and maximum axial peak locations. Pre-ejection top-peaked core xenon conditions and inlet moderator temperatures are chosen within allowable operating conditions to maximize the ejected worth and peaking results.

The uncertainties assigned to the static ejected CEA worth, the maximum

]

pin power peaking, the core average Doppler defect curve and the delayed f

) neutron parameters are unchanged from the previous analysis values specified l j in (50). As discussed in (55), the revised CEA ejection methodology for ,

2 Cycle 9 utilizes nominal Doppler defects and a Doppler weighting factor

technique instead of the three-dimensional pre-ejected configuration weighting I

used in Cycles 7 and 8.

a P

j 4.11.3 Moderator Density Reactivity Defect Analysis for LOCA I

i i

A new reference moderator density reactivity defect curve is generated

, for the Cycle 9 LOCA analysis. As disctssed in Section 4.6, the new curve utilizes the results of detailed, pinwise, two-dimensional PDQ radial analysis at selected moderator density conditions. previous cycle moderator density

]

reactivity curves were calculated using FOC, a code with a one-dimensional j cylindrical core representation, as discussed in (14). The result of the f methodology change is a vast improvement in the spatial detail of the problem and a more precise treatment of the radial core leakage effects which j influence the shape of the moderator density reactivity defect curve.

F l The Cycles 5 and 9 reference moderator density reactivity defect curves i cre compared in Figure 4.14. The curves are calculated at beginning of cycle.

l high soluble boron concentration conditions to provide the smallest negative

-(

-._---.,-__,_.-.---,.-_-___,-.---.-,,,,m_.- _ _ _ , . _ . . _ , , _ _ _ _ - - , , _ __

rosctivity defect due to decreasing moderator density. Although the curves represent different cycles and a different calculational technique, the Cycle 9 curve shows comparable negative reactivity down to approximately 15 lbm/ft . The Cycle 9 curve supports more positive MTC limits in the operating range due to its greater curvature around 40-50 lbm/ft . The i Cycle 9 curve is consistent with the new NTC Technical Specification limits in Figure 4.8. As in previous analyses, a 10% uncertainty is conservatively Cpplied to the moderator density reactivity defect curve prior to use in LOCA analysis.

l t

l l

TABLE 4.1 HAINE YANKEE CYCLES 3. 8 and 9 KUCLEAR CHARACTERISTICS Cycle 3 Cycle 8 Cycle 9 gpre Characteristics Exposure (MWD /NT)

Core Average at BOC 7,000 12,700 12,900 Cycle Length at Full Power 10,200 11,000 13,400 R: activity Coefficients - ARO 4 Moderator Temperature Coefficient (10 A0/0F)

HFP,BOC -0.34* -0.77 -0.41 HFP.EOC -1.98* -2.27 -2.31 Fuel Temperature Coefficient (10 AC/0F)

HZP,80C -1.70 -1.64 -1.62 HFP,BOC -1.30 -1.27 -1.25 HZP.EOC -1.80 -1.77 -1.76 HFP EOC -1.37 -1.37 -1.36 Kinetics Parameters - ARO Total Delayed Neutron Fraction (Entr)

HFF,BOC 0.00611 0.00607 0.00618 HFP.EOC 0.00517 0.00518 0.00511 Prompt Neutron Generation Time (10-6 see,)

HFP BOC 29.3 26.4 25.4 HFP,EOC 32.3 30.4 30.2 Control Characteristics control Elements Assemblico Number Full /Part Length 77/8 81/0** 81/0**

Total CEA Scrammable Worth (%1r )

HFP BOC 9.18 9.22 8.73 HFP.EOC 9.56 10.39 9.55 Burnable Poison Rods 756 992 1264 Number B4C-A1 023 Total Worth at HFP BOC (%it ) 1.4 2.2 2.5 Critical Soluble Boron at BOC.ARO (ppm)

HZP.NoXe Pksm 1,075 1.121 1,355 HFP NoXe.PkSm 995 1.016 1,261 HFP, Equilibrium Xe 762 781 1,014 Inverse Boron Worths (ppm /%'t )

HZP,BOC 84 96 101 HFP BOC 89 102 107 74 81 82 HZP.EOC 79 86 88 HFP.EOC 6 Conditions of 2440 MWt/2100 psia operation C* Four full-length CEAc are non-scrammable in Cycles 8 and 9 TABLE 4.2 NAINE YANKEE CYCLES 3. 8 and 9 CEA GROUP WORTHS AT HFP Worths (%AO)

Cycle 3 Cycle 8 Cycle 9 BOC ROC BOC 20C BOC T.0_q Shutdown CEA Croups Banks C + B + A 5.86 6.02 5.53 6.75 5.16 6.13 Ranulatina CEA Groups Bank 5* 0.55 0.64 1.29 1.54 1.47 1.56 Banks 5 + 4 0.90 0.97- 1.67 1.88 1.84 1.88 Banks 5 + 4 + 3 1.74 1.90 2.66 2.91 2.81 2.83 Banks 5 + 4 + 3 + 2 2.49 2.73 3.34 3.55 3.56 3.55 Banks 5 + 4 + 3 + 2 + 1 3.32 3.54 4.29 4.51 4.51 4.69 A11 CEA Groups Banks 5 + 4 + 3 + 2 + 1 +

C+B+A 9.18 9.56 9.82 11.26 9.67 10.82

• Bank 5 was redesigned in Cycle 7 to provide additional reactivity worth

- a__________ - _ - .

TABLE 4.3 MAIME YAMKEE CYCLES 3. 8. 9 i CORE AVERAGE DOPPLER DEFECT Doppler Defect (x 10-4 4p )

l l Puel Cycle 3* Cycle 8 cvele 9 l Resonance l Temperature o

F M E E E E E ,

4000 0 0 0 0 0 0 3750 19.4 20.9 - - - -

3500 39.4 42.5 41.5 44.8 40.8 44.6 3250 59.9 64.8 - - - -

3000 81.2 87.8 85.4 92.3 83.8 91.9 2750 103.1 111.6 - - - -

2500 125.9 136.2 132.4 143.1 129.9 142.4 i

2250 149.7 161.9 - - - -

2000 174.5 188.6 183.4 198.0 179.8 197.1 1750 200.5 216.7 - - - -

1500 228.1 246.5 239.4 258.6 234.8 257.2 1232 - - 272.3 294.0 267.1 292.4 1000 288.7 311.9 302.8 326.9 296.9 325.1 >

800 - - 331.0 357.2 324.5 355.4 532 351 5* 387.4* 372.5 401.8 365.4 399.8 t

300 394.3 426.0 412.9 445.6 405.1 443.0 f

200 - - 432.3 466.2 423.9 463.6 100 - - 453.1 488.6 444.2 485.9 0 - - 475.6 512.9 466.4 510.2

• at 5250F l

l L-__-___________________________________________________________

TABLE 4.4 MAINE YANKEE CYCLES 3. 8. 9 CORE AVERAGE DOPPLER COEFFICIENT Doppler coefficient (x 10-460/ r)

Fuel Cycle 3* cvele 8 Cycle 9 '

Resonance Temperature F M g g g g g 100 - - 0.2168 0.2333 0.2126 0.2330 200 - - 0.2007 0.2140 0.1960 0.2145 300 - - 0.1874 0.2010 0.1835 0.1996 400 - - 0.1762 0.1902 0.1729 0.1881 532 0.159* 0.172* 0.1644 0.1767 0.1618 0.1764 800 0.144** 0.156** 0.1465 0.1579 0.1437 0.1565

(

1000 0.131 0.141 0.1359 0.1468 0.1335 0.1458 l

l 1232 0.121*** 0.121*** 0.1269 0.1372 0.1251 0.1364 1500 0.114 0.123 0.1186 0.1271 0.1164 0.1272 l 2000 0.102 0.110 0.1070 0.1155 0.1049 0.1148 2500 0.093 0.101 0.0900 0.1051 0.0960 0.1052 l ,

3000 0.086 0.094 0.0909 0.0902 0.0891 0.0917 -

3500 0.081 0.088 0.0854 0.0923 0.0838 0.0919 l

i l

l I

f

• st 5250r  !

• • at 7500r C** at 12500 r [

l t

i TABLE 4.5 MAINE YANKEE CYCLES 3. 8 MD 9 BODEEATOE_ TEMPERATURE IQEFFICIENTI Conditions: HFP and HZP, ARO, Critical Boron Concentrations

- NTC (1C 440/0F)

Cycle 3* Cycle 8 cvele 9 Case conditions g g g g g g HFP, EqXe, Eqsm -0.47 -2,24 -0.77 -2.27 -0.41 -2.31 HZP Note, Pksm +0.24 - +0.06 - +0.41 -

HZP, NoXe, EqSm - -1.40 - -1.21 - -1.20 i

i 4

I i

4

• Cycle 3 HZP values at $250F, cycles 8 and 9 HZP values at 5320r a I 4

, I l

l TABLE 4.6 NAIME Yamver CYCLES 3. 8. 9 ARI NODERATOR DEFECT WITH WORST STUCK CEA l

Moderator Defect (x 10-4 60 )**

l Nederator Cycle 3* Cycle 8 Cycle 9 l

Temperature F M $9,g g g g g

- O ppm 790 ppm 0 ppm 1015 ppm 0 ppm 576.4 - -130.0 -67.2 -131.3 -47.3 -125.6 532 - (525)* - 0.0 0.0 0.0 0.0 0.0 l

500 - 43.5 24.7 90.0 23.0 72.0 450 - 111.0 39.0 160.0 42.0 133.0 400 - 167.3 51.0 218.0 45.0 183.0 350 - 217.0 57.0 260.0 40.0 224.0 i

! 300 - 261.6 58.9 299.0 33.3 254.6 1

250 - 298.5 66.0 328.0 25.0 288.0 .

l-l 200 - - 77.0 352.0 19.0 310.0 !

l 150 - - 82.0 375.0 14.0 325.0 l

100 - - 94.0 390.0 9.0 333.0 ,

68 - - $6.4 397.7 5.8 335.3 l

l t

• Cycle 3 values referenced to 0 at 5250F

• • Moderator defect at a constant 2250 psia for the specified temperatures e

TABLE 4.7 BAINE YANKEE CYCLES 3 and 9 k!NETICS PARAMETERS Conditions: HFP, AR0, Critical boron Delayed Cvele 3 Cycle 4 Time in Neutron Effective Lifetige Effective Lifetige Cvele Life _ Croup Fraction. (See ) Fraction (See )

DOC 1 0.00018 0.0126 0.00018 0.0126 2 0.00128 0.0305 0.00129 0.0305 3 0.00116 0.1163 0.00118 0.1168 4 0.00237 0.3116 0.00240 0.3131 5 0.00083 1.1652 0.00083 1.1730 6 0.00029 3.0253 0.00029 3.0230 TOTAL 0.00611 0.00616 40C 1 0.00014 0.0126 0.00014 0.0127 2 0.00111 0.0304 n.00110 0.0304 3 0.00097 0.1193 0.00097 0.1200 4 0.00197 0.3185 0.00195 0.3206 5 0.00077 1.1833 0.00071 1.1959 6 0.00025 2.9831 0.00024 2.9856 TOTAL 0.00517 0.00511 TABLE 4.8 MAINE YANKEE CYCLES 8 AND 9 CEA EJECTION RESULTS FROM FULL INSERTIONS Cycle 8 Cycle 9 Max N '= I-Pin Radial Peth 19E 59C E 195 HFP Bank 5 In 3.47 4.09 3.95 4.17 Ejected 5 (INCA Location 20)

HZP Banks 5 + 4 In -- -- 5.80 5.82 Ejected 5 (INCA Location 20)

HZP Banks 5 + 4 + 3 In 7.24 8.70 6.85 5.52 Ejected 5 (INCA Location 34)

Maximum E3ected Worth (%3r)

HFP Bank 5 In 0.296 0.409 0.340 0.385 Ejected 5 (INCA Location 20)

HZP Banks 5 + 4 In -- -- 0.499 0.528 Ejected 5 (INCA Location 20)

HZP Banks 5 + 4 + 3 In 0.618 0.860 0.507 0.437 Ejected 5 (INCA Location 34)

I TABLE 4.9 I

MAINE YANKEE CYCLE 9 CEA EJECTION RESULTS FROM PARTIAL INSERTIONS Worst Maximum Maximum Maximum CEA Insertion Ejected Ejected 1-Pin Axial -

Power by Oroup (%) CEA at Worth Radial Peak in Level INCA (Mp) Peak Core 31,,_, 5 4 3 Loeation 994 EE E9G A E99. 50.C0q i 100 10 0 0 20 0.008 0.014 1.69 1.69 1,59 1.55

80 25 0 0 20 0.069 0.103 2.31 2.55 2.19 2.05 60 65 5 0 20 0.352 0.409 4.45 4.84 2.05 2.14 40 100 40 0 20 0.462 0.525 5.40 5.81 1.95 2.07 l

20 100 80 20 20 0.446 0.456 5.75 5.85 2.15 2.03 0 100 100 60 34 0.476 0.445 6.57 5.71 2.00 2.34 I

t I

l l

l I

I L

i

. _ . . . . ~ . . _ _ _ _ _ _ . . . _ . . _ _ _ . _ . _ _ _ . . - _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ . . _ - - . _ . - _ , . _ _ . . _ _ . .

i i

TABLE 4.10 NAINE YANKEE CYCLES 8 AND 9 I

CIA DROP RESULTS AT DOC CEA Group Dropped Dropped CEA Naximum 1-Pin Positions CEA Worth (%S0) Radial Power

• l Refort_Dres Tree Cvele 8 Cvele 9 Cvele 8 Cvele 9 ARO A 0.114 0.107 1.72 1.73 ARO B 0.146 0.169 1.71 1.77 ARO C 0.105 0.095 1.71 1.71 ARO 1 0.070 0.066 1.63 1.64 tank 5 In A 0.106 0.101 1.82 1.83 Bank 5 In B 0.162 0.196 1.85 1.91 Bank 5 In C 0.110 0.100 1.83 1.83 Bank 5 In 1 0.069 0.062 1.73 1.74 0 Pre-Drop Maximum 1-Pin radial powers:

&RQ Bank 5 In Cycle 8 1.496 1.590 Cycle 9 1.511 1.620 Post-Drop Maximum 1-Pin radial power represent 80% of 2630 MWt feedback conditions.

l l

i l

! TABLE 4.11 NAINE YANKEE CYCLES 8 AND 9 l CEA DROP RESULTS AT IOC j l

CEA Group Dropped Dropped CEA Nazimum 1-Pin Positions CEA Worth (12.0) hadial Power

• Tyne cycle 8 cycle 9 CVCle 8 cycle 9 l Before Drom ARO A 0.125 0.118 1.71 1.67 ARO B 0.154 0.174 1.68 1.66 l

ARO C 0.107 0.104 1.68 1.64 Amo 1 0.072 0.075 1.61 1.58 l

l Bank 5 In A 0.117 0.116 1.78 1.82 l

tank 5 In B 0.161 0.177 1.77 1.86 Sank 5 In C 0.114 0.114 1.78 1.82

(

! Dank 5 In 1 0.071 0.073 1.69 1.72 l

l l

l l

l

• Pre-Drop Maximum 1-Pin radial powerst

&RQ Sank 5 In l Cycle 8 1.477 1.550 l Cycle 9 1.459 1.589 1

Post-Drop Maximum 1-Pin radial power represent 80% of 2630 MWt feedback l I

l conditions.

i

-ci-j

1 i TABLE 4.12 ,

l MAINE YANKEE CYCLES 8 AND 9 j DROPPED CEA WITH POWER LEVEL RESTRICTION MOST LIMITING PEAKING CASES 1

CEA Drop Time Maximum Power Percent Increase in Maximum j From Power Post-Drop Level Permitted 1-Pin Peaking Level (1) (Hrs) (%) Cycle 8 Cycle 9 ,_,

100 0.5 100 13.05 14.50 l 1.0 90 16.31 18.91 2.0 to 22.42 25.83 3.0 to 24.64 28.96 4.0 80 25.07 30.19 90 0.5 90 - 14.52 1.0 80 - 18.00 2.0 70 - 24.00 3.0 70 - 28.46 4.0 70 - 31.66 to 0.5 80 13.39 14.75 1.0 70 19.43 18.96 2.0 60 29.76 26.39 3.0 60 36.15 31.23 4.0 60 41.36 35.68 70 0 . *> 70 - 15.42 1.V 60 - 20.26 2.0 50 - 29.84 3.0 50 - 36.17 4.0 50 - 42.49 60 0.5 60 14.13 16.25 1.0 50 18.79 21.52 2.0 40 23.55 32.00 3.0 40 26.02 38.74 4.0 40 28.29 45.40

l }&BLE 4.13 l

l MalME YAMERE CYCLE 9 AVAILABLE SCRAM REACTIVITY l

j Worths (U a) 30c Roc l M E lEl M Scremenable CEA Worth

• 8.73 8.30 9.55 9.07 Stuck CEA Worth 1.49 1.42 1.73 1.65 POIL CEA Wortha* 0.00 2.15 0.17 2.48 CEA Progransning Allowance 0.04 0.04 0.11 0.22 Available scram CIA Worth - Nominal 7.12 4.69 7.54 4.72

- With Uncertainties *** 6.41 4.22 6.79 4.25

• AR1 CEA worth less non scraminable CEA worth (four subgroup 55 CEAs)

• • PDIL CEA insertion limit for NFP is 10% of Group 5 inserted PDIL CIA insertion limit for HZP is 100% of Groups 5 and 4 and 53% of Group 3 inserted

• • • Uncertainty factor of 0.9 l

l 1

TABLE 4.14 NAIME YANKEE CYCLE 9 REQUIRED SCRAM REACTIVITY Worth (%0 0) for l Time in Cycle Life and i RC8 Soluble Boron Concentration DOC ROC j 1015 som Q_333 E E E E Available Scram 6.41 4.22 6.79 4.25 i Beactivity with  ;

Uncertainties ,

(Table 4.13)

Cinimum Required Worth in CEAs Assumed -

- Steam Line Rupture 2.52 1.16 6.02 3.62 l l

twent*  !

l (section 5.5.1)

- Safety Analyses 5.70 3.20 5.70 3.20 I (Section 5)  ;

l Required Scram 5.70 3.20 6.02 3.62 l l

ReactLvity** ,

I i i Excess from Required 0.71 1.02 0.77 0.63 l to AvaL1able scram l Reactivity l

t i

o An uncertainty f actor of 0.9 is applied to the nominal minimum required worth 1 l

in CEAs for the steam line rupture event f rom Table 5.7 for comparison to i the available scram reactivity with uncertainties. This uncertalnty component is statistically combined with the other uncertainty components to derive the nominal minimum required worth in CEAs, as discussed in (51).  ;

ca Maximum of either the minimum required worth in CEAs assumed for the steam line rupture event or other safety analyses in Section 5 l i I

l

l TABLE 4.15 R&lME YANEEE CYCLES 8 AND 9 AUGMENTA110E FACTORS L

Core Core Channe in hel Density Due to Densification Height Height 1.5% 1.0% ,

(U (inches) Cvele 8 cvele 9  !

2 98.5 134.7 1.049 1.040 7 86.0 110.6 1.044 1.044 77.9 106.5 1.039 1.041 66.2 90.5 1.034 1.036 54.4 14.4 1.028 1.031 45.6 62.3 1.024 1.027 33.8 46.2 1.019 1.021  !

22.1 30.2 1.013 1.015 ,

13.2 18.1 1.009 1.010 1.S 2.0 1.002 1.001 l

i 1

l I

i 1

TABLE 4,16 MIME YAPKEE CYCLES 8 AND 9 CORE RADIAL PIM POWER CENSUS FOR AUGHENTATION FACTOR CALCULATION

" " r of PLns in Core Radial Pin Power Cycle O Cycle 9 Interval at 2000 WD/NT* pt 8.000 WD/NT*

1.4501 - 1.5000 192 68 1.4001 - 1.4500 416 664 1.3501 - 1.4000 604 964 1 1.3001 - 1.3500 1160 1628 1.2501 - 1.3000 1760 2576 f

1.2001 - 1.2500 2524 2020 1.1501 - 1.2000 3624 3008 1.1001 - 1.1500 3912 4428 1.0501 - 1.1000 400s 4400 1.0001 - 1.0500 2284 3448 l

0 - 1.0000 liLQi M Total 37184 36912

• F1sttest pin census during the cycle l

I l

l l

L>

TAaLE 4,17 MAINE YANKEE CYCLES 6-9 RELATIVE PRESSURE VESSEL FLUENCE C0HPARISONS Flux Reduction Factors at Asimuthat Angle from Perpendicular to Core Shroud Flats Total Effective Full-Power ------ 00 -------- ------ 100 --------

G2fdag Years _to_ ROC

• Tarist Desianed Iggggi Deslaned 1-6 6.51 1.00 1.00 1.00 1.00 7 7.50 1.02 1.05 1.20 1.21 8 8.55 1.35 1.42 1.51 1.43 9 9.72 1.35 1.55 1.51 1.56 Future Cycles -- 1.35 -- 1.51 --

Cycle 1-9 Average 9.72 1.08 1.12 1.14 1.14 0 Based on 2.630 MWt full power operatLon e

Table 4.18 Maine Yankee Physics Methodology Documentation Supporting Application Descriptien of Methodology Documentation Reference in Cycle R; actor Physics Methods - YAEC-1115 14 3 Reference Report R: actor Protective System Setpoint YAEC-1110 8 3 Analysis - Reference Report Extension of Fine Mesh Diffusion PC No. 64, 7 4 Theory and Nodal Physics Methods to Section 4.8 R activity Parameter Calculations WMY 78-102, 27 cnd a Change in the Nodal Neutronic Attachment B Crupling Model Introduction of Local Pointwise PC No. 64, 7 4 Doppler Feedback Effects in Section 4.8 Two-Dimensional Pinwise Diffusion WMY 78-102, 27 Theory Calculations for Dropped CEAs Attachment C cnd Special CEA Drop Test at 50%

P:wer for Method Verification Uncertainty Applied to Moderator YAEC-1259, 37 6 Reactivity Defect from Hot Zero Power Section 4.7 Reduced from 25 to 15%

Doppler Defects for CEA Ejections YAEC-1324, 50 7 Calculated with Explicit Pre-Ejected Section 4.10 Local Power Weighting and Uncertainty Reduced from 25 to 15%

l Figure 4.1 Maine Yankee Cycle 9 Assembly Relative Power Densities BOC (500 MWD /MT)

HFP,ARO C@E 9.rMARY OF MIIM FLEL ASSEMBLY POWER DESCRIPTION MI.VAll.E ASSEMBLY ASSEMBLY AVG. 1.28025 20 MI. FLEL R0D 1.48541 39 MI. CHAf0EL 1.41267 25 CORE POSITION / ASSEMBLY MBER . ! A14 1 I A12 2I FUEL TYPE ..................... I L-8 ! L-0 1 CEA BAf0C TYPE . . . . . . . . . . . . . . . . . I I I ASSEMBLY AVERAGE P0lER ....... 1 .32611 1 .48044 1 MIIM FLEL ROD P0lER ....... 1 .61989 I .76808 I MIIM CHAf0EL P0lER . . . . . . . . . I .60014 I .76138 I I B17 3 I B16 4 I B15 5 I B13 6 i B11 7I I L-8 1 W41 S4I S41 W81 I I eC* I I *1e ! I I .31783 I .63371 I .97188 1 1.13796 1 1.13492 I I .64590 1 1. % 921 1 1.42673 1 1.46412 I 1.44203 I I .62276 1 1.05230 1 1.36056 I 1.39541 1 1.37064 1 1 C18 8 I C17 9 I C16 10 I C15 11 1 C13 12 I C11 13 1 1 L-8 i N-4 I N-8 I M-0 I M-8 I M-8 I I I eAs ! I eC* I I e5e !

I .40453 I .95648 1 1.16287 1 1.20670 I 1.19799 1 1.17527 I I .76496 I 1.37204 1 1.45511 1 1.35253 1 1.32847 1 1.26518 I I .74508 1 1.30771 1 1.38201 1 1.31130 1 1.28530 1 1.20120 1

! D19 14 I D18 15 I D17 16 1 D16 17 I D15 18 I D13 19 I D11 20 I I L-8 I N-0 I M-4 I L-0 1 N-8 I L-4 I K-8 I I 1 55* ! I *As ! I e3e ! I I .40673 1 1.09163 1 1.19607 I 1.00731 I 1.27464 1 .98130 1 1.28025 I I .76617 I 1.47539 I 1.35929 1 1.09645 I 1.48528 1 1.00075 I 1.45509 I I .74628 1 1.39866 1 1.30676 I 1.08265 1 1.41140 1 1.07153 1 1.38364 1 1 E20 21 1 E19 22 I E18 23 I E17 24 I E16 25 I E15 26 I E13 27 I E11 28 I I L-8 I N-4 1 M-4 1 M-8 i N-8 I L-8 I M-4 I M-8 I I I *As I I I I e25 I I I I .31817 I .95770 1 1.19698 1 1.20755 I 1.27846 I .96977 I 1.13345 1 1.17022 1 1 .64632 1 1.37331 1 1.36056 1 1.30720 1 1.48433 1 1.02633 1 1.27021 ! 1.26485 I I .62319 1 1.30895 I 1.30797 1 1.24211 1 1.41267 ! 1.00477 1 1.22012 1 1.20016 I I F20 291 F19 30 I F18 31 I F17 32 I F16 33 i F15 34 I F13 35 I Fil 36 I I M-4 I N-8 I L-0 I N-8 I L-4 I N-8 I L-8 I M-4 1 1 *Ce 1 I eAs I I e5e 1 I I I I .63399 I 1.16345 I 1.00777 I 1.27752 I .97974 1 1.27999 I 1.15074 1 1.12538 I I 1.06950 1 1.45553 1 1.09681 1-1.48376 1 1.03533 1 1.48000 1 1.24384 1 1.23497 I I 1.05262 1 1.38239 I 1.08298 1 1.41226 1 1.02403 1 1.40456 I 1.19252 1 1.17579 1 1 020 37 I G19 38 I G18 39 I G17 40 I G16 41 I G15 42 1 013 43 I G11 44 I I N-4 I M-0 I N-8 I L-8 I N-8 I M-8 I M-8 I L-12 I I I eCa 1 I e2e I IeBe ! I e4e !

I H21 45 I .97201 1 1.20681 1 1.27442 I .96871 1 1.27850 1 1.17131 1 1.11770 I .93200 1

< I L-8 1 1.42684 1 1.35261 1 1.48541 I 1.02566 I 1.47963 1 1.27208 I 1.22547 I .97655 I I I 1.36067 1 1.31139 I 1.41152 1 1.00422 I 1.40381 1 1.20710 1 1.16542 I .95479 I I .32612 I I .61990 I J20 46 I J19 47 I J18 48 1 J17 49 I J16 50 I J15 51 1 J13 52 1 J11 53 I I .60015 I N-4 1 M-8 I L-4 I M-4 I L-8 I M-8 i M-8 I N-8 I I ele ! I e3e ! I I I eBe 1 I I K21 54 1 1.13797 I 1.19794 I .98111 1 1.13282 I 1.15020 1 1.11739 1 1.13437 1 1.25894 I I L-0 1 1.46409 1 1.32848 1 1.00068 I 1.26954 1 1.24342 1 1.22564 1 1.22939 1 1.45878 I I I 1.39538 1 1.28533 1 1.07145 I 1.21950 1 1.19226 I 1.16505 1 1.16773 I 1.40830 1 I .40041 I l .76004 I L20 55 I L19 56 I L18 57 I L17 58 I L16 59 I L15 60 I L13 61 1 L11 62 I

[

I .76134 I N-8 I M-8 I N-8 I M-8 i M-4 I L-12 I N-8 I E-16 I I I e5e I I I I e4e ! I 558 1 1 1.13489 1 1.17517 I 1.27994 1 1.16972 1 1.12484 1 .93133 1 1.25003 I .97497 I i 1.44198 I 1.26504 I 1.45485 I 1.26434 1 1.23438 I .97561 1 1.45838 1 1.04940 1 I 1.37058 1 1.20108 1 1.38342 1 1.19966 1 1.17523 I .95387 1 1.40799 1 1.01799 I 69-7, - - - . - . - - - - , . , . --y . -

Figure 4.2 Maine Yankee Cycle 9 Assembly Relative Power Densities MOC (6,000 MWD /MT)

HFP,ARO CORE St#9tARY OF MAllful FEL ASSEMBLY POER DESCRIPTION MAI. VALE ASSEMBLY ASSENLY AVG. 1.32001 25 MAI. FEL R0D 1.47515 25 MAI. CHA10EL 1.42332 25 CORE POSITION /ASSDELY IU SER . I A14 1 I A12 2I FE L TYPE ..................... I L-8 1 L-0 1 CEA BAIGC TYPE . . . . . . . . . . . . . . . . . I I I ASSDELY AVERACE POER ....... 1 .35298 1 .51659 I MAlllUI FlEL R00 POER ........ I .64451 I .79920 I Mall!UI CHAf0EL POER ......... I .62881 I .79523 1 I B17 3 I B16 4 I B15 5 I B13 6 I B11 7I I L-8 I M-4 I N-4 I N-4 I N-8 1

' I I eCe 1 I *1e! I I .34722 I .66556 I .98788 1 1.15452 1 1.17856 I I .67432 I 1.05118 I 1.36788 1 1.42352 1 1.41962 I I .65540 1 1.04083 1 1.31812 1 1.37321 1 1.36904 I I C18 8 1 C17 9 I C16 10 1 C15 11 1 C13 12 I C11 13 I I L-8 I N-4 1 N-8 I HI M-8 I M-8 1 1 I eA* ! I eC*I I e5* I I .41942 1 .98639 1 1.19874 ! 1.16031 1 1.17048 I 1.15492 1 I .75181 1 1.34331 1 1.42600 I 1.26745 1 1.28998 I 1.22743 I I .73721 1 1.29154 1 1.37185 I 1.23404 1 1.23063 1 1.18198 I I D19 14 1 D18 15 I D17 16 1 D16 17 I D15 18 1 D13 19 I D11 20 I I L-8 I N-0 I M-4 I L-0 1 N-8 I L-4 I N-8 I I I e5* I I eAe ! I e3e 1 I I .42147 1 1.05801 I 1.17169 1 1.00377 I 1.29538 I .96222 1 1.28259 I I .75325 I 1.38595 1 1.32514 1 1.07460 1 1.45000 1 1.04480 1 1.41137 I 1 .73865 1 1.33184 1 1.27554 I 1.06232 1 1.39519 1 1.03662 1 1.36159 I I E20 211 E19 22 I E18 231 E17 241 E16 25 I E15 26 I E13 27 I EttM-8 28 II I L-8 1 N-4 1 M-4 I M-8 I N-8 1 L-8 I M-4 1 I I *As ! I I I e2e ! I I I .34750 I .98734 1 1.17242 1 1.19465 1 1.32001 I .96628 I 1.09044 1 1.11485 I I .67459 1 1.34416 1 1.32614 I 1.29479 1 1.47515 1 1.00917 1 1.19902 1 1.15897 I I .65569 1 1.29238 I 1.27635 I 1.24342 1 1.42332 1 .98851 ! 1.14826 I 1.15042 I I F20 29 I F19 30 I F18 31 ! F17 32 I F16 33 I FIS 34 I F13 35 I F11 36 I I M-4 I N-8 I L-0 1 N-8 i L-4 I N-8 1 L-8 I M-4 1 I eCe 1 I *As  ! I e5e I I I I I .66575 I 1.19915 1 1.00411 1 1.31927 I .98985 1 1.30218 I 1.09969 I 1.06792 1

!  ! 1.05139 1 1.42706 I 1.07485 1-1.47474 1 1.03811 1 1.44796 1 1.18348 1 1.15360 I i 1.04103 1 1.37210 1 1.06258 1 1.42303 I 1.02587 1 1.39555 I 1.13460 1 1.10659 I I G20 37 I G19 38 I G18 39 I G17 40 I G16 41 I G15 42 I G13 43 I Gil 44 I I N-4 1 HI N-8 I .L-8 I N-8 I M-8 I M-8 I L-12 I I I eCe ! I e2e ! I eBe I Ie4e I I H21 45 I .98797 I 1.16039 1 1.29525 I .96552 1 1.30110 I 1.14550 1 1.08977 1 .91267 I I L-8 I 1.36795 1 1.26751 1 1.45009 1 1.00867 1 1.44727 I 1.23848 1 1.17714 I .95585 1 1 1 1.31820 1 1.23410 I 1.39528 1 .98006 1 1.39488 1 1.18955 I 1.12441 I .93904 I I 35299 I I . 64453 I J20 46 I J19 47 I J18 48 I J17 49 I J16 50 I J15 51 I J13 52 I J11HI 53 I I 62883 I N-4 I HI L-4 1 M-4 I L-8 i HI HI I ele I I e3e ! I I IeBe ! I I K21 54 I 1.15455 I 1.17050 I .96217 1 1.01013 1 1.09946 1 1.08954 I 1.12005 1 1.28898 I i I L-0 1 1.42355 1 1.29003 1 1.04483 1 1;19874 1 1.18336 1 1.17747 1 1.21071 I 1.44622 I I I 1.37323 1 1.23068 I 1.03664 1 1.14000 1 1.13460 1 1.12479 I 1.16111 1 1.41128 I l 1 51658 I I 79918 I L20 55 I L19 56 I L18 57 I L17 50 I L16 59 I L15 60 I L13 61 I L11E-16

. 62 1I I 79521 I

. N-8 I M-8 I N-8 I HI M-4 I L-12 i HI I

I I e5e! I I Ie4e ! I e5e!

l

! I 1.17857 1 1.15492 1 1.28250 I 1.11468 1 1.06767 I .91208 1 1.28799 I .99068 I I 1.41963 1 1.22743 1 1.41133 I 1.19879 1 1.15337 I .95493 1 1.44578 1 1.04809 1 1 1.36906 1 1.18198 1 1.36155 1 1.15025 I 1.10637 I .93812 I 1.41090 1 1.02559 I l

. - _m - _ , _ .. _ , , . , , _ . _ _ , , - - - _. -

Figure 4.3 '

Maine Yankee Cycle 9 -

Assembly Relative Power Densities EOC (14,000 MWD /MT)

HFP,ARO CORE SIM9RY OF MIIM FEL ASSO9LY POER DESCRIPTION MI. VALE ASSDELY ASSO R.Y AVG. 1.34205 25 MI. FlfL R00 1.45352 25 MI. CHAP 0EL 1.38507 25 CGIE POSIT!WUASSDELY MBER . I A14 1 I A12 2I FE L TYPE ..................... I L-8 I L-0 1 CEA BApK TYPE . . . . . . . . . . . . . . . . . I I I ASSOR.Y AYDIAGE PGER ........ I .40392 I .58226 I MIIM FEL R00 PGER ........ I .70259 I .86924 1 MIIM CHAPIEL POER . . . . . . . . . I .68895 I .86227 I I B17 3 I B16 4 I B15 5 I B13 6 I B11 71 1 L-8 I M-4 I N-4 I N-4 I N-8 I I IeCe! I ele ! I I .39961 I .72323 1 1.03216 I 1.18866 1 1.24676 I I .73629 1 1.06586 I 1.31112 I 1.39065 I 1.42283 I I .71919 I 1.05990 1 1.27661 1 1.32840 1 1.34773 I

! C18 8 I C17 9 I C16 10 1 C15 11 I C13 12 I C11 13 I I L-8 I N-4 I N-8 i M-0 1 M-8 1 M-8 I I I eAs 1 I *Ce ! I e5e 1 I .44977 I 1.03621 1 1.26123 I 1.13244 1 1.12315 1 1.11519 I I .75811 I 1.30313 I 1.42900 1 1.19434 I 1.20023 1 1.16064 I I .74625 1 1.26659 1 1.35754 1 1.17710 1 1.17027 I 1.13883 I I D19 14 1 D18 15 1 D17 16 I D16 17 I D15 18 I D13 19 I D11 20 1 I L-8 I N-0 I M-4 I L-0 I N-8 I L-4 I N-8 1 1 I e5e ! I eAs ! I e3e I I I .45160 1 1.01339 1 1.11717 I 1.00663 1 1.32117 I .95014 I 1.27964 !

! .75954 I 1.25377 I 1.24806 1 1.06500 1 1.44814 1 1.01902 I 1.39044 I I .74770 1 1.22332 1 1.22093 I 1.06164 I 1.37414 I 1.01689 I 1.31920 I I E20 21 I E19 22 1 E18 23 I E17 24 I E16 25 I E15 26 I E13 27 I E11 28 I I L-8 1 N-4 I M-4 I M-B I N-8 I L-8 I M-4 1 M-8 I I I eAs I I I I e2*I I I I .39978 I 1.03677 I 1.11758 I 1.13657 1 1.34205 I .96837 I 1.03806 I 1.04500 I I .73636 1 1.30355 1 1.24845 I 1.20855 1 1.45852 1 1.00316 I 1.09925 1 1.10517 I I .71928 1 1.26689 I 1.22127 I 1.18280 1 1.38507 I .99665 1 1.07802 1 1.07863 1 1 F20 29 I F19 30 I F18 311 F17 32 I F16 33 I F15 34 I F13 35 I Fil 36 I I M-4 I N-8 I L-0 1 N-8 I L-4 I HI L-8 I M-4 I I eCe I I eAe ! I e5e ! I I I I .72328 I 1.26137 1 1.00600 1 1.34162 I .99712 I 1.30599 I 1.03153 I .99484 I I 1.06588 1 1.42906 I 1.06510 1-1.45829 1 1.02992 1 1.42068 I 1.09006 I 1.04092 I I 1.05993 1 1.35759 1 1.06175 1 1.38 % 9 I 1.02 4 9 I 1.34934 I 1.07024 I 1.01641 I I G20 37 I G19 38 I G18 39 I G17 40 I G16 411 G15 42 I G13 43 I Gil 44 I I N-4 1 HI HI L-8 I N-8 I HI M-8 I L-12 I I I eCe ! I e2e ! I eBe I I e4* I I H21 45 I 1.03216 I 1.13245 I 1.32115 I .96798 1 1.30545 I 1.08159 I 1.02315 I .88235 I I L-8 I 1.31112 I 1.194331 1.44815 I 1.002951 1.42030 I 1.15088 I 1.07381 I .92132 I I I 1.27660 1 1.17709 I 1.37415 I .99647 I 1.34869 I 1.12477 I 1.05017 I .91596 I I .40391 1 I .70258 I J20 % 1 J19 47 I J18 48 I J17 49 I J16 50 I J15 51 I J13 52 I J11 53 I I .68894 I N-4 I M-9 I L-4 I M-4 I L-4 I HI HI HI I ele! Ie3e !  !  ! IeB*I I I K21 54 1 1.19865 I 1.12318 I .95021 1 1.0 1 1.03157 I 1.02308 1 1.06605 I 1.29228 I I L-0 1 1.39865 1 1.20025 I 1.01909 I 1. I 1.09018 1 1.07413 1 1.13474 1 1.41444 I I I 1.32841 1 1.17027 1 1.01697 1 1.07808 1 1.07037 1 1.05048 I 1.10739 I 1.36165 I I .58223 I I .86920 I L20 551 L19 56 I L18 57 I L17 58 I L16 59 I L15 60 I L13 611 L11 62 I I .86223 I N-8 I M-8 I HI HI M-4 I L-12 I HI E-16 I I I e$eI  !  ! I e4e ! I e5* 1 I 1.2 4 76 I 1.11523 1 1.27973 1 1.04508 I .99487 I .88203 I 1.29167 1 .99867 I I 1.42284 1 1.16066 1 1.39053 I 1.10524 1 1.04099 I .92074 1 1.41407 1 1.03502 I i 1.34775 1 1.13886 1 1.31929 1 1.07871 1 1.01648 I .91536 1 1.36143 I 1.01996 1

Figure 4.4 Maine Yankee Cycle 9 Assembly Relative Power Densities BOC (500 MWD /MT)

HFP, CEA Bank 5 Inserted CORE

SUMMARY

OF MAII?Of FLEL ASSEMBLY PGER CESCRIPTICN MAI. VALE ASSEMBLY ASSE55LY AVG. 1.35340 39 FAI. FLEL RCD 1.64463 41 11AI. CHANNEL 1.56413 41 CCRE FG3ITION/ ASSEMBLY NUMBER . I A14 1 I A12 2I FLEL TYPE ..................... I L-3 I L-0 I CEA EANK T M ................. I I I ASSEMBLY AWFACE PMER ........ I .3'729 . I .47633 I PAXIMLN FUEL RCD POER ........ ! .624E5 I 74401 I PAIIMLM CHANNEL TGER . . . .. . . . . I .60738 I .73233 I I B17 3 I B16 4 I B15 5 I B13 6 I B11 7I I L-0 I M-4 I N-4 I N-4 I N-S I I I eCe I I e1s I I I .32027 I .6S344 I 1.03350 I 1.10967 I 1.00024 I I .6 4 95 I 1.16096 I 1.49334 I 1.47236 I 1.21061 I I .6t095 I 1.14204 I 1.43233 I 1.39904 I 1.16tSS I I CIS S I C17 9 I C16 10 I C15 11 I Cl3 12 I Cll 13 I I L-3 I N-4 I N-S I M-0 I M-8 I H-S I I I eAs I I eC I I e5* I 1 .25436 I .87981 1 1.21313 1 1.23633 1 1.12699 I .74096 I I .53993 I 1.31300 1 1.54946 I 1.41980 1 1.34503 I .90717 I

, I .52299 I 1.25407 I 1.47486 I 1. 2 027 I 1.31040 I .05340 I D19 14 I D13 15 I D17 16 I D16 17 I D15 18 I D13 19 I D11 20 I I L-S I N-0 I M-4 I L-0 I N-S I L-4 I N-S I I I e5* I I eAs ! I a3a I I I .28649 I .56007 I 1.00473 I 1.00631 I 1.35006 I 1.01818 I 1.24766 I I .54101 I .90394 I 1.27454 I 1.15229 I 1.53633 I 1.13004 I 1.55132 I I .52397 I .05643 I 1.23076 I 1.14179 I 1.50987 I 1.12541 I 1.432S5 I I E20 21 I E19 22 I E!S 23 I E17 24 I E16 25 I EIS 26 I E13M-4 27I I E!! M-3 23 II I L-S I N-4 I M-4 I M-3 I N-8 I L-8 I I I eAe I I I I e2* I I I I .32065 I .SS!05 I 1.00572 1 1.04666 I 1.15546 I 1.00317 I 1.23198 I 1.34'24 ,

I I .66650 I 1.31472 1 1.27599 I 1.15000 I 1.40736 I 1.10216 I 1.47111 1 1.47613 I I .64150 1 1.25540 I 1.23216 I 1.10555 I 1.33845 I 1.08724 I 1.42725 I 1.39545 I I F20 29 I F19 30 I F13 31 I F17 32 I F16 33 I FIS 34 I F13 35 I FilM-4 36 II I M-4 I N-S I L-0 I N-S I L-4 I N-8 I L-8 I I ec# 1 I *Ae I I e5* I I I I I .68391 1 1.21904 I 1.00707 I 1.15515 I .64513 1 1.29996 I 1.345S3 1 1.35730 I I 1.16162 I 1.55037 I 1.15303 I 1.40792 I .86943 1 1.64638 I 1.44680 I 1.43515 I I 1.14272 I 1.47571 I 1.14250 I 1.33860 I .80440 I 1.56391 I 1.3S443 --

I I.41283 I

!UO37I019 38 I 018 39 I G17 40 I G16 411015 421013 M-8 43 II 011L-12 44 II I N-4 I M-0 I N-8 I L-8 I N-8 I M-8 I

-I I eCa I I e2e ! I eBe ! I e4* I I H21 45 I 1.03899 I 1.28690 1 1.35340 1 1.00771 1 1.29918 I 1.30225 I 1.32074 I 1.11251 I I L-3 J 1.49943 I 1.42038 I 1.587531 1.10203 I 1.64663 I 1.43258 I 1.43389 I 1.17458 I I I 1.43345 I 1.33085 I 1.51053 I 1.08720 I 1.56413 1 1.35717 I 1.35788 -

I 1.14497 -

I I .33942 I-----

I .62509 I J20 44 I J19 47 I J18 48 I J17 49 I J16 50 I J15M-8 51 II J13M-3 52 II J11N-8 53 II I .60761 I N-4 I M-8 I L-4 I M-4 I L-3 I

-I e1e I I e3e I I I I aBe ! I I K21 54 1 1.11009 I 1.12736 I 1.01S39 -I 1.28183-I 1.34569 I 1.32055 I 1.28783 I 1.33883 I I L-0 I 1.473441 1.345551 1.13041 1 1.47095 I 1.44864 I 1.43408 I 1.42012 I 1.60545 I I I 1.39958 I 1.31090 1 1.12577 I 1.42709 I 1.38422 I 1.35804 I 1.34567 --

I I.53132 I I .47648 I -

I .74427 I L20 55 I L19 56 I L18 57 I L17 58 I L16 59 I L15L-12 60 II L13N-8 61 II L11E-16 62 II I .73256 I N-8 I M-S I N-8 I M-8 I M-4 I

---

- ! I e5a ! I I I e4e I I a5e I I 1.00059 I .74117 1 1.24782 I 1.34318 I 1.35715 I 1.!!!88 I 1.33782 I .70295 I I 1.21303 I 90739 I 1.55140 I 1.47902 I 1.48495 I 1.17422 I 1.60377 I .54420 I I 1.16728 I .85369 I 1.48292- I- 1.39836-1 1.41266 I 1.14463 I 1.52981

- - - I .76542- I

Figure 4.5 Maine Yankee Cycle 9 Assembly Relative Power Densities MOC (6,000 MWD /MT)

HFP, CEA Bank 5 Inserted CORE SiffMRY OF MAIIM FLEL ASSEMBLY FGER LESCRIPi!0N MI.VALLE A3SEMBLY ASSEMBLY AVG. 1.33920 39 MI. FLEL ROD 1.56843 39 MAI. CHANEL 1.51115 39 CORE POSITION / ASSEMBLY M.tGER . I A14 1 I A12 2I FUEL TYPE ..................... I L-S I L-0 I CEA SA E TYPE ................. I I I ASSEMR.Y AVERAGE PGdER ........ I .37416 I .52223 I MIIM FLEL ROD POER ........ I .66138 I .78130 I MIIMlM CHAEL F0ldER . . . . . . . . . I .64793 I .77940 I I B17 3 I B16 4IB 5 I B13 6 I Bil 7I I L-8 I M-4 I )5N-4 I N-4 I N-8 I I I eCe I I e!e I I I .36011 I .73340 1 1.07411 I 1.14320 I 1.05555 I I .71349 I 1.16340 I 1.45S53 1 1.43553 I 1.21197 I I .69245 I 1.15144 I 1.40873 I 1.30076 I 1.17097 I I CIS 8 I C17 9 I C16 10 I C15 !! I C13 12 I Cl! M-8 13 II ~

I L-8 I N-4 I N-8 I M-0 I M-8 I I I *Ae I I eCe ! I e5e I I .30907 I .93589 I 1.28177 I 1.25349 I 1.10SO9 I .72197 I I .55473 I 1.32266 1 1.54392 1 1.352S9 I 1.30976 I .87547 I I .54117 I 1.27482 I 1.43754 I 1.31706 I 1.26839 I .82182 I I D19 14 I D18 15 I D17 16 I D16 17 I D15 18 I D13 19 I Dl! N-8 20 II I L-8 I N-0 I M-4 I L-0 I N-8 I L-4 I I I e5e I I eAe ! I e3* I I I .30963 I .56103 I 1.0!!63 I 1.01S10 I 1.3SS75 I .99653 I 1.24041 I I .55436 I .91647 I 1.26084 I 1.14626 I 1.56776 I I.09750 I 1.47471 I I .54179 I .85673 I 1.23048 I 1.13535 I 1.51051 I 1.09383 I 1.42737 I

-I E20 21 I E19 22 I EIS 23 I E17 24 I E16 25 I E15 26 I E13 27 I Ell M-S M-4 I 28 !I I L-8 I N-4 I M-4 I M-8 I N-8 I L-8 I I I eAe I I I I e2e ! I I I .36044 I .93693 I 1.01247 I 1.05254 I 1.19669 I .99698 I !.21850 I 1.26241 I I .71392 I 1.32371 I 1.26191 1 1.15588 I 1.40955 I I.07777 I !.36903 I 1.36206 I I .69288 I 1.27582 I 1.23153 I 1.10338 I 1.35885 I 1.06235 I 1.31775 I 1.30773 I b20 29 I F19 30 I FIS 31 I F17 32 I F16 33 I FI5 34 I F13 35 I Fl! M-4 06 II I M-4 I N-8 I L-0 I N-3 I L-4 I N-8 I L-3 I I eCe ! I eAe I _ I e5e ! I I I I .73379 I 1.282'4 I 1.01875 I 1.19656 I .M628 I 1.29373 I 1.259751 1.262S4 I I 1.16396 1 1.54470 I 1.14689 I 1.40979 I .83382 I 1.55363 I 1.33842 I 1.36242 I I 1.15199 I 1.48828 I 1.13597 I 1.35914 I .77296 I 1.49638 I 1.28696 I 1.30612 I I 020 37 I G19 38 I G18 39 I G17 40 I 016 41 I G15 42 I G13M-8 43 II G11 44 II L-12 I N-4 I M-0 I N-8 I L-8 I N-8 I M-8 I I I eCe I I e2* I I eBe ! I e4e I I H21 45 I 1.07457 I 1.25402 I !.33920 I .99683 I 1.29335 I 1.24110 I 1.25399 I 1.06010 I I L-8 I 1.45912 I 1.35348 I 1.56843 1 1.07792 I 1.55421 1 1.34687 I 1.34108 1 1.09960 I I I 1.40935 I 1.31759 I 1.51115 I 1.06255 I 1.49741 1 1.29250 1 1.28667 I 1.07936 I I .37430 I I .66163 I J20 46 I J19 47 I J18 48 I J17 49 I J16 50 I J15M-8 51 II J13M-8 52 II J11N-8 53 II I .64818 I N-4 I M-8 I L-4 I M'-4 I L-8 I

-I e!e ! I e3e1 I I IeBe I I I X21 54 I 1.14365 I 1.10853 I .99690 1 1.21872' I 1.25997 I 1.25390 I 1.23150 I 1.32292 I I L-0 I 1.43609 ! 1.31028 I 1.09797 I 1.36931 I 1.33860 I 1.34164 I !.34009 I 1.52663 I I I 1.38131 I 1.26940 I 1.09429 I 1.31802 I 1.28716 I 1.28706 I 1.28369 I 1.47446 I .

I .52240 I - -

I .78156 I L20 55 I L19 56 I L18 57 I L17 59 I L16 59 I L15 60 I L13N-8 61 II L11E-16 62 II I .77966 I N-8 I M-3 I N-8 I M-8 I M-4 I L-12 I I I e5e I I I I e4e ! I e5e !

I 1.05594 I .72224 I 1.24001 I 1.26273 1 1.26296 I 1.05959 I 1.32185 I .67868 I I 1.21243 I .87580 I 1.47516 I 1.36239 I 1.36265 I 1.09947 I 1.52503 I .C2609 I I 1.17141 I .32213..1 1.42830 I 1.30005 I 1.30634 I !.07924 I 1.47298 I .74817 I

Figure 4.6 Maine Yankee Cycle 9 Assembly Relative Power Densities EOC (14,000 MWD /MT)

HFP, CEA Bank 5 Inserted CORE SLM1ARY Cf MAIIMUM FLEL ASSEFAY FCER DE3CRIPTION t1AI.VALUE ASSEMBLY AS E EtY AVG. 1.42373 39 FAX. FLEL F0D 1.58873 59 VAI. CH E EL 1.*0485 39 CME Ft1 SIT!W/ASSEFRY PUGER . I A!4 1 I A12 2I FLEL TYPE . . . . . . . . . . . . . . . . . . . . . I L-3 I L-0 I

&A BA* TYPE . . . . . . . . . . . . . . . . . I I I ASSEMBLY AVERAGE POER ........ ! 44132 I .60803 I MAX 1 rut FLEL RCD #0ER ........ I .74232 I .87562 I FAXIMUM CWNEL F0WER . . . . . . . . . I .73146 I .87229 I I B17 3 I B16 4 I B15 5 I B13 6 I B11 N-4 I 7I N-S I I L-8 I M-4 I N-4 I I I sC* I I e!* I I I .42889 I .81995 I 1.15251 I 1.20SS4 I 1.14905 !

! .S0392 I 1.20921 1 1.44933 I 1.42415 I 1.27613 I I .78431 I 1.23335 I 1.39726 I 1.36034 I J.20547 I iCIS 3 I C17 9 I C16 10 I C15 11 I C13 12 I C11 M-S 13 II I L-8 I N-4 I N-8 I M-0 I M-S I I I eAe ! I eCe ! I e5e I I .3445S I 1.01305 1 1.35418 I 1.24309 I 1.07667 I .68861 I I .58623 I 1.32618 I 1.5S443 I 1.33843 I 1.24114 I .86248 I I .57223 I 1.30075 I 1.50326 I 1.31403 I 1.23095 I .80045 I I 019 14 I DIS 15 I D17 16 I D16 17 I D15 18 I D13 L-4 19 II D11 N-S20 II I L-S I N-0 I M-4 I L-0 I N-8 I I I e5* I I eAs 1 I e3e  ! I I .3?.609 I .54490 I .98S11 I 1.03332 I 1.42321 I .98353 I 1.22750 I I .58659 I .S7310 I 1.22370 I 1.15779 I 1.58317 I 1.08749 I 1.41453 I I .57342 I .81188 I 1.20587 I 1.15663 I 1.50434 I 1.0S196 I 1.34007 I I E20 hl E19 22 I EIS 23 I E17 24 I E16N-S 25 II EISL-8 26 II E13M-4 27I I Ell M-8 29 II I L-S I N-4 I M-4 I M-S I I I I e2* I I I 1 I eAe !

I .42912 I 1.01S73 I .93364 I !.0!!52 I 1.21324 I .99741 ! 1.14114 I 1.16211 I I .S0413 1 1.32672 I 1.22425 I 1.08221 I 1.39868 I 1.05815 I 1.22206 I !.20790 I I .7S453 I 1.30128 I 1.20944 I 1.05370 I !.34549 I 1.0539S I 1.19401 I 1.18714 I I F20 29 I F19 30 I FIS 3 I F17 32 I F16 33 I F15 34 I F13 N-8 I L-8 35 I I F11M-4 I36 I I M-4 I N-8 I L-0 I N-8 I L-4 I I eAe I I e5* I I I I I eCe !

I .82019 I 1.3S465 I 1.03SSO I 1.21335 I .61248 I 1.25374 1 1.14665 I 1.14362 I I 1.20954 I 1.58498 I 1.15825 I 1.39910 I .79335 I 1.43908 I 1.18454 I 1.19S45 I I 1.20367 I 1.50374 I 1.15708 I 1.34592 I .73173 I 1.36921 I 1.16505 I 1.16994 I I G20 37 I G19 38 I G18 39 I G17 40 I G16 41 I G15M-8 42 II G13M-8 43 II G11L-12 44 II I N-4 I M-0 I N-8 I L-8 I N-8 I

-I I eCe I I e2* I I eBe ! I e4e I I H21 45 I 1.15234 I 1.24648 I 1.42S73 I .98758 I 1.25382 I 1.12586 I 1.13134 I .9S381 I I L-8 I 1.45025 I 1.33S86 I 1.58873 J 1.05854 I 1.43968 I 1.18155 I 1.17919 I 1.00240 I I I 1.39767 I 1.31440 I 1.50485 I 1.05437 I 1.36986 I 1.15734 1 1.15290 I .99311 I I .44144 I - - -

! .74303 I J20 46 I J19 47 I J18 48 I J17 49 I J16 L-8 50 II J15 M-8 51 II J13 M-8 52 II J11 N-8 53 II I .73166 I N-4 I M-8 I L-4 I M-4 I I

- I e!* I I e3*J .I I I eBe I I K21 54 I 1.20920 I 1.07706 I .98:198 I 1.14131.I 1.14710 1 1.13142 I 1.!!456 I 1.25271 I I L-0 I 1.42457 I 1.24155 I 1.08797 I 1.22260 I 1.18508 I 1.17969 I 1.17327 I 1.40818 I I I 1.36126 I 1.23134 I 1.08244 I 1.19454 I 1.16556 I 1.15331 I 1.14905 I 1.34649 I I .60818 I I .875S5 I L20 55 I L19 56 I L18 57 I L17 58 I L16M-4 59 II L15 60 II L13N-8 L-12 61 II L11E-16 62 II I .S7251 I N-8 I M-8 I N-8 I M-8 I I I e5* I I I I e4e I I e5e  !

I 1.14939 I .68887 I 1.22805 I 1.16265 I 1.14401 I .98362 I 1.25208 I .62609 I i 1.27650 I .86276 I 1.41519 I 1.20848 ! 1.19895 I 1.00253 I 1.40731 I .78393 I I 1.20582 I .80071 I 1.34870 I 1.18769 I 1.17043 I .99325 1 1.34569 I .70917 I

NOTE: 1. THIS CURVE INCLUDES A CALCULATIONAL UNCERTAINTY FACTOR OF 1.10 AND A TILT FACTOR OF 1.03 T

2. MEASURED FgSHOULD INCLUDE A MEASUREMENT UNCERTAINTY FACTOR OF 1.08 BEFORE COMPARISON TO THIS CURVE 1.77 I I ll 1 l lIN ILllB!!

l l 1.76- -

I i

l ip I l l l l l '

!pl l Ilgj! g K

COORDINATES (KMWD/WT FR )

1.75- F IIh II I  !! II II II I (o.co 1.7o0) (o.os ,1.7o0)

(o.so .1.7o0) (1.oo ,1.7o2) l l l ll

'74 L ll j:

l l

j llll jjg j (2.00 .1.704) (4.00 .1.89s) gy >

i l jj tj j (s.oo ,1.ess) (s.oo.1.eso) i sx

s (12.oo ,1.as4 )  !

Ill ti I I Ii i U llII I I (10.00.1.e93) g;j.

w $73- gl! !pq q l l ll l gig yl l l (14.00.1.e70 ) (15.00.1.670) l thlll niilllil ' ilHL ili l in li l l !!llIll lll111111lill1lilllill I ifInl kk !f l i ll h

@ llill ll !!I i l ii I !Il lli 1 0 !Il j  !

!! l lljl Mil

$ [Ilf Hf l' l il ll ih Ih i ll d il I !i ,

1! l Uljl l il llal l l il l li h I i

l10 iii! n !I l I l I il  ! l IUUIM h!liin

$ ' LNillllli! ll  !!IINI!!' [

ll Olllllll lHiin s "JHDd%E I 1I i H[ i Il II I

!lih[iliiN!Ill & ll HhlW Mil 5 l.

np I  ! l ll Il E I lh 2 II H lll W I!!@M l1 1 i l # ioi Nini Minh:lil:ir li}!  !  !

l@lh ll ll ll l ll !, ! llll0 @ IM! II I , , f I l } l!h !@ ll!l!!%l l lll e 1.68 '

U l Mill JI;!:

ilh 'l I! l lHlll ll!

I!N

'!!!M ,:h Ul lh I Mu I i  ; I n W!$

hi i l l Hill  !! l l I l l l H I d!!! Ni![

I ill l IlillH :llill!?

IIh !;I Wh!I IW li I l.

lf lp I ll  !: i l ! giugHit S q!N I I! h t mli ll [ lH I l l I li l 1.65 9 12 13 14 15 0 1 2 3 4 5 a 7 s 10 11 CYCLE AVERAGE EXPOSURE (KMWD/MT)

Figure MAINE YANKEE Allowoble Unrodded Rodial Peak

• '7 Cvele 9 Cycle Average Bumup ,

17. Above Nominal Case

0.6 UNACCEPTABT OPERAVON ~::

0.5  : .

X g __

g g

g (i COORDINATES b 0.4

                                                                                      ,                                         ( 0,0.50)         --

Oo w -

                                                                                        \ A

( 30,0.50) 3h 0: \ (100,0.12) -- P_

           <             e@                                                                        '

s b 2

                  $ 0.3                                                                                        '

(,.

           # e-d                                                                                                            'g a: o                                                                                                                4 ST                                                                                                                     N
           @E 0.2                                                                                                                     X
                                                                                                                                        \

z i 7 ( I \ ACCEPTABLE OPERAVON '% i 1 ) 0.1  ; l , 0.0 0 10 20 30 40 50 60 70 80 90 100 POWER EVEL (X OF RATED POWER) l l l WAINE YANKEE Moderator Temperature Coefficient Umits ! Figure Versus Cycle 9 Power Level 48

I I v o 4 -p

• 1g. ._.... ..... _

l g _l

                                    !!!iiiiiiiil 5'piiiiiii"
                                                                    ~  ~

E. _ D 5 2 -! 4 e o

                                                                                                                                                                            -5 g!

Is _g -E g

                                                                                                                                                                          .                      n E
                                                                                                                                                                    ~* g -!                      E g                                                                                                                                                                  o o                        g 5                                                                                                                      .
                                                                                                                                                                    -E 5                         v a.

u. o

                                                                                                                                    --                              -I      -E                   o o

v x . 8

                                                                                                                                                                  ~   o     _*                   g g          _    ___,

g -s j

                                                      '       '                                                                                                   g         _E                   *
   ,5 g

d

            ;::#~+ q
                                                                                                                                                                            -g                   a 2        r---+                               - ooooooooooo                                                         Izo                              I jE 8- :;::: *I ,sesssssssss s                                                                 si 5                                                 ooooooooooo                                                               g                                          -   o                  <
            +--.,                   .

sessssssses _* g e 8 t-:tf __14 gl2n oooooooo+++ x ' Io E o o

                           +4                        ssessssssoa                                                      o                                             -e 8
                   ,- ++                ;$mo e         -_

_g -E o tt! 05+ EEEEEEENSM g _2 y mn gnggegeoooo _4 s , - - __- Sf ,- - o

                                                                                                                                                                            .o

,l _ __+. g _g a g g

               -    ----             +

( - E ooooooSoooo r i 3 DM omosee nn-o - - - - o a v- - , a ,,_ I II IU!I II I "7YI II I P o o o o o o o o o o o e o o a e 5 m o t M N (W3 Mod 031YW 30 %) 01 y 3Hn0L4 MOW.:1 M W3 Mod 30N383338 WO WR10Y 30 MnMIXYM MAINE YANKEE Power Dependent insertion Umlt Figure i Cycle 9 (PDIL) 4.9 t for CEA's (

100  ? 90 ~ . 5 m 80- --

       %                                                                           i

• COORDINATES  !

E 70 ' g (500, 0) g (532, 0)  ! g 60 (552,100) - 5 50-l h@ R .l 8 30 5 - g i e 20 . 5 l f 10 . O d 500 510 520 530 540 550 560 l ! NOMINAL COLD LEG TEMPERATURE (INDICATED) 0F 1 MAINE YANKEE Referonos Power Level Figure Versus Cycle 9 Nominal Cold Leg Temperature 4.10

1 1 1 FIGURE 4.11 l MY CYCLES 8, 9 MAXIMUM PEAKING VS. DROPPED CER WORTH FROM SPECIFIED POWER LEVELS 48;""''''""'""""'""'"'"'""""'"''"'"'"''"'"'"'""',. 48 i  : 44 i  :

: P0k'ER 42 POINTS OF' INTERSECTION 7 LEVEL (E)

(.0728 39.96)  :

0 40 (.09 30.89)

(.1083 22.86)  :

Wg 38 - (.1212 19.77) o - (.1281 17.48)  : O 38 i (.1338 18.22)

                                     .- (.1254 14.88) z
                   $ 32i:                                                                                                                                                                      F

20

o. 30 -: -

x  : 2 o Mi - b l 1 l L Mi F r . o 24- - 1 r  :  : 40 E e 22i . F E 20 i F 80

z. 18 -

                   .-                                                                                                                                                                           - 80

: 100 w

(n 18 ------------ C  :  :

                                                                                                                        -/

W 14 - l z  : S m

                         - 12i        .

F x 10 i I 8-' F 8 --- CYCLE 8 RT 100% P0lER 7 CYCLE 9  :. si . 2i - F O"......................................,.........,.........,.........,.........' O.00 0.02 0.04 0.08 0.08 0.10 0.12 ' 0.14 0.18 DROPPED CER WORTH (PERCENT DELTA RHO)

FIGURE 4.12 MAINE YANKEE CYCLE 9 SHUTDOWN MARGIN EQUATION AND REQUIRED SCRAM REACTMTY 7.0 6.5 - 6.0 i. i s, .HFP, P=100 ( O 5.5 - F- A ,' ,, 5.0 -

                                                                      '*,'s, 4.5-                                                                ', , HFP g

{ 4.0- ' ,, 6 m 35 ,' , '.=-c,- - ' ,\ HZP,P=0 O g 3.0-( l-- o ,, b c.

                                                                                    ,HZP                                                              ,e 2.5 -

2,0 - ,

                                                                                                         ,N LEGEND 1.5 -                                                                               ,               ,' ,,-

SHUTDOWN MARGIN EQUATION ,'e 1.0 - REQUIRED SCRAM REACTIVITY (TABLE 4.14) 0.5 - e -- - - - - - -e STEAM UNE RUPRJRE EVENT G.------E] OTHER SAFETY ANALYSES 0.0 .- . . . . . . .- , . , . 0 100 200 300 400 500 600 700 800 900 1000 110 0 1200 1300 i RCS SOLUBLE BORON CONCENTRATION - PPM i l

7.0 i k i I iiI  ! I i 1 i I h.

• i 5 . I I i t i I i 3

                                                                                                                                                                                                       !           i . !                I          I                     I*

i  ! 1I  !

• I . I i .

I i ' .! 3 I ' 3 1 i r i i e i f f I Y . _ i i i 1 i i r i i i l i  ! !

            - wir                                                                                                                                                              i 6.5                                                                                                                                , ,                    ,                        ,                ,                                    .
                                                                                    , .,                      , , .                                                                                                                                                      i

_ i . x . I , I i

                                                                                                                               !               i
            !I e                     N             !       .          I     1       * ! i                     ! !                     I i              i                !                        !
                                                                                                                                                                                                                                             '      ,                           t i!                             N              i           ! I            I            '_          f i
                                                                                                                                               !              i                1                        !          !

Y' . 3 mA i i

                                                                             '       r i !         t i 3           e i            ,            i         I
             'N-                                   g

l ;ji P =100

                  ;                  s 3C           ,       i h j;'i 1i6                     T i

i 1  !!l O e '%.+! I r N i iI ! I ' i

                                                                                                                                                                                                                    !         I I

i%_

                          .s
                                             ,                     1,m                            i            ,

i i P = 80  ! I i

             !            !I         'C              t       _ ! _7 <3-                       _

i i 11 I I i Ni i I I ! '

                                                                                                                                                                 ! ! !                        I                      I I!                       6 i e i                 . i C . i iNi                              IihmNi i       !
                                                             ' ! . t i X i          ! i i

i  ; i r i I i p, g i T. i i t f 3 , 1

                                     ' C                                                                i                                              i         i
                    !                                 1 1

3<3~ 1 N  ! I I i  ! i n '% mi i l I '

                                                                                                                                  !              i!              ,                                                                                      i i
           %wi . .                                                                                                                               r               i i                                                            i A                                                                       6 4*5        i NI tA             it i t i i i .A                      !,ii i .1 1       5
                                                                                                                                        ! r              i        i t i i i i i1 i

pg g iii 1 1I I .1 3  ! t i !1 %i i I "i <!- I i i !Ni. i i it i i i i _  ! 3 I C! . t + .A . i.

                .%'                                      '%                                                                                                                        ,    . ,                                . 1 4.0 -        1 , s i ,                                         s.                    i,                           , , .                       , i                  .

i i , i P = 20 ii l , 1 i w. i x ; i i i . i i i i i i i z I I 3 i  ! ' h g 3.5 I i

                                                          %m x, m i !

F i

                                                                                                                       . , ,                                                 i
                                                                                                                                                                                                  .                         .             P=0             .       i
                                                                                                                                                                                                                                                                             ,1 i
                                                                                                                                                                             ,                                              i                             !        ,

g

                                                                                                           .           i ,
                                   ,                                      i       x1
                                                                          !               I     i                      . i            l        ll         Eii6 I                     I i !l6 9!1 l1 !l!le i I                      -

i 3.0 Z  ::: h :a SDW = 4.21 - 0.00310 C + 0.0255 P  ;; when C is less than 326 PPM a 2.5 - Z -g g

• 7 SDM = 3.20 + 0.0253 P when C ie greater than or equal to 326 PPM 4 y

4 where 2 1 m SDM is b required shutdown morg percent reactivity) 2.0 - E  ::: O I: Cis RCS boron concentration On P is the maximum of either the octu power level E 14"

              +-                                                                            or the reference power level correagonding to the                                                                                                                                :r D              7                                                                                                                                                                                                                                                              ;;

temperature from Figure 4.10 8  ::: nominal cold ed power) ;g z - On Peroent of o T 0.5 , I 0.0 0 100 200 300 400 500 600 700 800 900 1000 110 0 1200 1300 ACTUAL RCS BORON CONCENTRATION - PPM Required Shutdown Margin Figure MAINE YANKEE Versus Cvcle 9 4.13 RCS Boron Concentration

                                                                                                                                                                -- +            -.                   .           _ - -                                    _.                                                                 _                                     _ _ _ _ _ _ _

l Figure 4.14 Maine Yankee Cycle 5 and 9 l J Moderator Density Defect Curves for LOCA Analysis 1 0.1 0.0 ,, , , . 7 _ %

                                                                                       %(
               -0.1 6-*
               -0.2                                                                                                      g        i<

W L M L\

                -0.3                                                                                                                    '1L\

O M Y < 1 CD 8 ' gg s -0.4 \) s t i .EJ

\ "c'      N l
                                                                                                                                                         \\ ,
            + -0.5-
                                                                                                                                                                 \

t

   #                                                                                                                                                           t I

Id

>e
                 -0.6                                                                                                                                            I
                                                                                                                                                                 \_
                                                                                                                                                                     \

I N -0.7 - - CYCLE 5

                                                                  - CYCE 9
                 -0.8

{

                                                                                                                                                                      \     t
                  -0.9 l                                                                                                                                                                            I
                     .0                                                                   E.0                                        10.0                               0.0 j                        00.0   50.0                  40.0           E.0 3                                                                            i
         -                                                         MODERATOR DENSITY (LBM/FT )

5.0 SAFETY ANALYSIS 5.1 General A review of the safety analysis for operation of Maine Yankee during Cycle 9 is presented in this section. The parameters which influence the results of the safety analysis are listed in Table 5.1. Values are provided fcr the Reference Safety Analysis, for Cycle 8 and Cycle 9. The Reference S:fety Analysis for Maine Yankee consists of the Cycle 3 stretch power cnalysis (3), or any specific safety analysis completely redone since the power uprate. Table 5.3 lists the Reference Safety Analysis for each event. The safety parameters may be divided as follows: 1) initial operating etnditions, 2) core power distributions, 3) reactivity coefficients, 4) shutdown CEA characteristics, and 5) Reactor Protection System setpoints and time delays. A discussion of the differences between Cycle 8. Cycle 9, and the Reference Safety Analysis values for the parameters listed above is ccntained in Subsections 5.1.1 through 5.1.5. L l Values reported herein for MDNBR were determined using the YAEC-1 CHF correlation (40). Application of this correlation with a DNBR limit of 1.20 was approved for Maine Yankee in (52). l 5.1.1 Initial Operatinr Conditions l ! The initial conditions assumed in the safety evaluations considered in this section are listed in Table 5.1. These conditions are conservative with ! respect to intended Cycle 9 operation in that uncertainties are included to cccount for measurement errors associated with plant instrumentation. The l uncertainties include: l l a) A two percent allowance for calorimetric error in core thermal power. b) A four degree allowance for measurement error on reactor coolant temperature. I 1 c) A twenty-five psi allowance for measurement error on main coolant pressure. 1 The maximum allowable core inlet temperature has been increased from 550 F to 552 F for Cycle 9. Allowable core inlet temperature and pressure ccnditions during power operation are specified in Technical Specifications. These are based on preservinF/ DNB overpower margin for all possible combinations of temperature and pressure. The preservation of DNB overpower margin is accomplished by reducing the allowable core inlet temperature when cperating at lower pressures. This assures that the minimum DNBR reported for ccch of the incidents consider <3 remains conservative for operation at the lower system pressures. The safety analysis for Cycle 9 assumed that 180 tubes are plugged in cach steam generator for the loss of load event, and 250 tubes plugged per steam generator for all other events. The steam generator normal water level cssumed for the Cycle 9 safety analysis was 66% narrow range. 5.1.2 Core Power Distributions The power distribution in the core, and in particular, the peak heat flux and enthalpy rise, are of major importance in determining core thermal margin. The procedure used in the safety analysis was to set the initial conditions (inlet temperature, power, pressure, CEA insertion, and axial power distribution) and, through analysis, assure that sufficient initial overpower margin is available to prevent the violation of acceptable criteria for each incident analyzed. This procedure is continued for Cycle 9. If the available overpower margin is not sufficient for the set of initial conditions, new power distributions are selected, by either modifying the symmetric offset limiting condition for operation (S/0 LCO) or by modifying the allowable CEA insertion limit versus power until it is demonstrated that sufficient margin exists. As a starting point the safety analysis assumes the FSAR design power W 9 s wn n Figure 5.2. In distribution (F, = 1.68 and F delta H = . ' most cases considered, acceptable performance was demonstrated with the use of 4 {

the design power distribution. As indicated in Table 5.11, this power distribution, evaluated at the full power heat flux, results in a lower DNBR than any of the Cye,le 9 predicted power distributions within the S/O LCO band. cvaluated at their respective maximum power level limit as defined in the PDIL for Cycle 9 (Figure 4.9). In addition, values are presented in Table 5.11 for Pd-Po, the percent-rated thermal power margin between Pd, the power level at which the MDNBR for o given power distribution would equal the SAFDL on DNB, and Po, the initial maximum power level allowed by the CEA insertion limit for that rod configuration. Because of variation in the subchannel location in which MDNBR is predicted at nominal conditions versus limiting conditions, this is a more precise indicator of relative DNB margin between power distributions than initial steady-state MDNBR. For Cycle 9, the thermal power margin (Pd-Po) for the FSAR design power distribution at full power conditions is lower than the thermal margin for the limiting Cycle 9 power distributions within the LCO band at both full and lower power levels. Hence, thermal margins calculated using the FSAR design power distribution are conservative for Cycle 9. Power peaking associated with the CEA drop and the CEA ejection events for Cycle 9 are compared with reference values in Table 5.1. The effect of differences between Cycle 9, cycle 8, and the Reference Safety Analysis for the CEA drop and CEA ejection are discussed in Sections 5.4.2 and 5.5.4. 5.1.3 Reactivity Coefficients The transient response of the reactor system is dependent on reactivity feedback effects, in particular, the moderator and the fuel temperature reactivity coefficients. Nominal values for each of the above feedback coefficients are given in Sections 4.5 and 4.6. The Doppler coefficients for Cycle 9 are essentially identical with those of Cycle 8. Variations in the cbove parameters will influence each transient in a different manner. Therefore, the effect of the difference in reactivity coefficients is discussed on an event-by-event basis. For Cycle 9, the allowable positive values for MTC in the power range see detailed in Figure 4.8. The analyses, limited by a positive MTC, performed in (6) and Cycle 9, with the exception of the CEA ejection analysis, were conservatively performed at HFP conditions with NTC equal to

 +0.5 x 10~   delta rho / F and bound the values in Figure 4.8. The CEA ejection analysis for Cycle 9 (Section 5.5.4) assunes the most positive of Gither the predicted NTC (Table 4.5) with +0.5 x 10~ delta rho / F uncertainty added or the value specified in Figure 4.8. Events limited by a negative NTC are discussed in their respective sections.

The effective neutron lifetime, delayed neutron fractions, and decay constants are functions of fuel burnup and the fuel loading pattern. The Cycle 9 kinetics parameters are compared to the corresponding reference cycle values in Section 4.8. Small differences that are experienced from cycle to cycle have an insignificant impact on the response of the plant for all transients except the CEA ejection. In the CEA ejection accident, the ratio of the ejected rod worth to effective delayed neutron fraction is extremely sensitive in determining the course of the power response. An evaluation of this event for Cycle 9 is provided in Section 5.5.4. 5.1.4 Shutdown CEA Characteristics The negative reactivity insertion following a reactor trip is a function of the acceleration of the CEA and the variation of CEA worth as a function of position. The safety analysis considers this function in three separate parts: 1) the CEA position versus time, 2) the normalized reactivity worth versus rod position, and 3) the total negative reactivity inserted following a scram. The CEA position versus time assumed in the Reference Safety Analysis was provided as Figure 4.2 in (3). This curve reflects a conservative rod insertion time of 3.0 seconds. This curve is based on results from plant measurements and is not expected to change from cycle to cycle. Furthermore, CEA drop times are measured at each refueling as part of the startup test program to verify this assumption. l l The normalized reactivity worth versus rod position assumed in the Reference Safety Analysis was provided as Figure 4.3 in (3). This curve is sensitive to axial power distribution and is based on the minimum reactivity insertion for a variety of axisl power distributions. The normalized r activity worth versus rod position was calculated for limiting Cycle 9 axial power distributions and was compared to the curve assumed in the Reference Safety Analysis. The normalized reactivity worth versus rod position assumed in the Reference Safety Analysis is conservative for Cycle 9 for events limiting at HFP. The normalized control rod negative reactivity insertion versus time curve presented in Figure 4.4 of (3), which was obtained from a cynthesis of the aforementioned functions, is likewise conservative in j cpplication to Cycle 9 for HFP events. The normalized reactivity worth versus position curve from (3) was modified for the HZP condition for Cycle 6 (37). A more conservative function l was derived from Cycle 6 power distributions at HZP. This was compared with the normalized reactivity worth versus rod position curves determined for limiting cycle 9 axial power distributions at zero power conditions. The Cycle 6 curve bounds all but the most bottom-peaked EOC power distribution for Cycle 9. Events sensitive to changes in scram worth versus position are the loss of flow and seized rotor events, which are not limiting at zero power, and the CEA ejection incident, which is discussed below. The normalized reactivity worths versus position used in the Cycle 9 4 CEA Ejection Transient (Section 5.5.4) are shown in Figures 5.3 and 5.4. These were derived from the limiting bottom peaked axial power distributions for Cycle 9. The HFP curves bound the range of symmetric offset allowed by the LCO band at intermediate power levels. The HZP curves correspond to the m st negative offset case seen in the RPS setpoint axial oscillation study for ' Cycle 9, and assume no restriction on allowed symmetric offset. Values assumed in the Reference Safety Analysis and for Cycle 9 for the total negative reactivity inserted following a scram are given in Tables 5.1 cnd 5.13. Comparison of the scram worths assumed in the Reference Safety 1 Analyses and the values assumed in the Cycle 9 safety analysis indicate that the Cycle 9 values bound the Reference Safety Analysis values. The values of ceram reactivity specified in Table 4.14 bound those assumed in the safety cnalysis supporting operation of Cycle 9. 1 5.1.5 Reactor Protective System Setpoints and Time Delays 1 The reactor is protected by the Reactor Protective System (RPS) and Engineered Safeguards Features (ESF). In the event of an abnormal transient, the Reactor Protective System is set to trip the reactor and prevent unacceptable core damage. The elapsed time between the time when the setpoint candition exists at the sensor and the time when the trip breakers are open, is defined as the trip delay time. The values of the trip setpoints and instrumentation delay times used in the Reference Safety Analysis are provided '# in Table 4.7 of (3). The setpoints assumed for Cycle 9 are given in Table 5.12 and Figures 5.5, 5.6, and 5.7. The only difference between these values and those assumed in the Reference Safety Analysis is the low pressurizer pressure (floor of the thermal margin) trip setpoint which was increased from 1750 psia to 1850 psia in Cycle 4. Since this was a conservative change, the values for all these setpoints for Cycle 9 are either the same as or bound those used in the Reference Safety Analysis. As indicated in (3) the Reference Safety Analysis assumes no credit for the high rate of change of power trip function. This remains unchanged for Cycle 9. Credit is taken for the functioning of the Variable Overpower (V0PT). Thermal Margin / Low Pressure (TM/LP) and Symmetric Offset Trip Systems (SOTS) in several areas. First, the V0PT is credited in limiting the initial power distributions considered in setting the Symmetric Offset Trip System cetpoints as a function of power level. Second, the V0PT is also credited in limiting the power increase and power distribution changes possible during CEA Bank Withdrawal, Excess Load, and CEA Drop transients, as discussed in Sections 5.3.1, 5.3.3 and 5.4.2. Credit is also taken for the functioning of the V0PT in the analysis of the CEA Ejection transient Section 5.5.4. The TM/LP and Symmetric Offset Trips are cycle dependent. They are derived from the predicted core behavior as described in (8). The Cycle 9 cetpoints for the TM/LP and Symmetric Offset Trips for 3-loop operation are presented in Figures 5.5, 5.6, and 5.7. Both the Symmetric Offset Trip setpoints and the TM/LP setpoints have been modified for Cycle 9. The low pressure floor of the TM/LP trip continues to be assumed to trip the reactor in the analysis of the SGTR accident, Section 5.5.2. 5.2 Summary Each transient and accident considered in (3) and (6) has been reviewed cnd/or re-evaluated for Cycle 9. The incidents considered are categorized as follows:

1) Anticipated Operational Occurrences (A00) for which the Reactor Protection System (RPS) assures that no violation of Specified Acceptable Fuel Design Limits (SAFDL) will occur.

2) Anticipated Operational Occurrences (A00) for Which sufficient initial steady-state overpower margin must be unintained in order to assure acceptable results.

. 3) Postulated Accidents. The incidents considered are listed in Table 5.2. In most cases the parameters considered in (3) and (6) and for Cycle 6 in (37) and (38) bound the Cycle 9 values. For those transients where the i parameters for Cycle 9 are outside the bounds considered in previous Safety Analyses, a new or revised analysis has been performed. These are:

1) Boron Dilution

2) CEA Ejection

3) CEA Withdrawal

4) CEA Drop

5) Steam Line Rupture

6) Large Break LOCA Other transients that require a partial reanalysis or review included:

1) Seized RCP Rotor

2) Excess Lond

3) Loss of Load

4) Loss of Feedwater

5) Loss of Coolant Flow

6) Steam Generator Tube Rupture

7) Small Break LOCA A summary of results for Cycle 9 is presented in Table 5.3.

5.3 Anticipated Operational Occurrences for which the RPS Assures No Violation of SAFDLs The incidents in this category were analyzed in the Reference Safety Analyses for the 2630 MWt tiprate and Positive MTC submittals for Maine Yankee, (3) and (6). Selected cases were reanalyzed in (7) to account for char.ges in the Cycle 4 core physics characteristics. These analyses showed that the incidents in this category do not violate the SAFDLs; the primary coolant system pressure limit; or the 10CFR20 site boundary dose limits. The changes considered in the present analysis do not significantly affect the NSSS r2sponse during these transients. This assures that the conclusions relative to primary system pressure and site boundary dose remain valid. Protection against violation of the SAFDLs continues to be assured by the RPS. Setpoints are generated for the TM/LP and Symmetric Offset Trips which include the changes in power distributions associated with cycle 9. Ssetions 5.3.1 through 5.3.5 review the Anticipated Operational Occurrences for which the RPS assures no violation of the SAFDLs. 5.3.1 Control Element Assembly Bank Withdrawal The Reference Safety Analysis for this event demonstrates that the most cevere CEA withdrawal transient occurs for a combination of reactivity tddition rate and time in core life that results in the slowest reactor power rise to a level just below the Variable overpower Trip. This combination of parameters maximizes the core thermal heat flux and core inlet temperature and results in the mininum DNBR. i The Reference Safety Analysis considered parametric analyses at full power (2630 MWt) for Moderator Temperature Coefficient (MTC) and Reactivity Addition Rate. The ranges analyzed were +0.5 x 10' delta rho / F to

    -3.0 x 10~ delta rho / F and 0 to 0.7 x 10- delta rho /sec. As indicated in Table 5.1 the Cycle 9 predicted value of MTC, with uncertainty, ic -2.81 x 10~ delta rho / F. Reference 3 showed the MDNBR to occur at
                    ~

En MTC of -2.9 x 10 delta rho / F for this event, with less negative MIC r:sulting in higher MDNBR. Table 5.1 also shows a higher maximum rate of reactivity addition for Cycle 9. Reference (3) showed that high rates of rcactivity addition result in a faster rise of core power to the Variable Overpower Trip Setpoint and values of MDNBR less limiting than for slower transients. The CEA bank 5 division into two separately moveable subgroups for Cycle 7 created a new class of CEA bank withdrawal events, withdrawal of a CEA bank subgroup. Complete withdrawals of each of the CEA bank 5 subgroups were cnalyzed from initial conditions corresponding to the limiting power distributions within the S/O LCO alarm band for each power level. The withdrawal of CEA bank 5B presents a particularly severs transient. Bank 5B is symmetrically located around the center in the inner portions of the core. Withdrawal of this group causes an increase in peaking towards the center of the core as the flux shifts inward from the outer portions of the core. This chift also causes a reduction in the contribution to the excore monitors. Thuc,'the actual overall core power can increase beyond the 10% setting of the variable overpower trip that the excores will see. The MDNBR for a CEA bank withdrawal event for Cycle 9 occurs for the withdrawal of the CEA bank 5B subgroup from an initial power level of 47% rated power, assuming the CEAs to be initially positioned at the corresponding insertion limit. The MDNBR for this event is 1.42. The peak RCS pressure for o CEA bank withdrawal is provided in Table 5.3 and is less than the ASME design overpressure limit of 2750 psia. 5.3.2 Boron Dilution The Boron Dilution Incident was addressed in (3), (29), (43) and the FSAR. Inadvertent dilution of the Reactor Coolant System was considered under o variety of plant conditions which could result in either an inadvertent power generation or loss of shutdown margin if sufficient time were not cvailable for the operator to take corrective action. Small changes in boron concentrations resulting from the Cycle 9 reload have an insignificant impact on the conclusions reached in the Reference Analysis. An evaluation of this incident was performed for Cycle 9 for events pastulated during refueling, shutdown, startup, hot standby and power cperation conditions. Table 5.6 presents a summary of the results of this r: view for Cycle 9. 5.3.2.1 Dilution DurinK Refueling Assumptions made in the Cycle 9 evaluation for dilutions during c: fueling are consistent with those rade in (3) and (29). The limiting dilution in (3) was based on the maximum capacity of the CVCS via the normal makeup and letdown flow paths (200 spm each). The limiting dilution event in (29) was based on the maximum flow of the Primary Water Makeup System (250 spm). Both analyses assumed letdown flows equal to the dilution flowrates and minimum reactor vessel water volumes of 2599 ft (volume below lower lip of reactor vessel nozzles). Hence, the Primary Makeup Water System dilution is the limiting dilution under refueling conditions. Based on the Cycle 9 core loading, the critical boron concentration under cold conditions (68 F) during refueling is 1086 ppm. The minimum initial reactor vessel boron concentration which will prevent an inadvertent criticality within 30 minutes is 1644 ppm (Case No. 3 Dilution, Reference 29). Therefore, it is concluded that if the reactor vessel boron concentration is maintained at or greater than 1644 ppm during Cycle 9 refueling, it would require a continuous dilution at the maximua possible rate for 30 minutes to cchieve an inadvertent criticality. This is ample time for the operator to ccknowledge the audible count rate signal and take corrective action to cut off the source of the dilution. 5.3.2.2 Dilution During Cold. Transthermal. and Hot Shutdown with RCS Filled Dilutions during cold, transthermal, and hot shutdown were addressed in (43). The assumptions in (43) remain unchanged for Cycle 9. The limiting dilution is via the CVCS (200 spm), and the RCS is assumed to be filled (no

                                           -92

l i credit taken for pressurizer volume). The highest worth CEA is assumed to be l stuck out of the .: ore, the loop stop valves open and either RHR or RCP on. R: quired minimum Reactor Coolant System initial boron concentrations to allow 15 minutes margin to criticality are listed in Table 5.4, along with the boron , cencontration required to meet the Technical Specification 5% delta K/K suberiticality requirement for shutdown conditions. The boron concentrations r: quired by the Technical Specification 5% delta K/K suberiticality 4 r:quirement conservatively bound those required to meet the 15-minute i' rsquirement for margin to criticality during boron dilution events. 5.3.2.3 Dilution Durint Cold. Transthermal. and Hot Shutdown with Drained RCS Conditions f 1 4 Dilutions during shutdown conditions with the RCS partially drained were addressed in (29) and (43). In order to conservatively bound any partially drained configuration with one or more reactor coolant loop isolated, the assumption is made that only the portion of the reactor vessel , below the lower lip of the nozzle is filled. With the exception of the CEA of highest worth, which is assumed to be stuck out of the core, and 1% delta K/K cf Bank A, which is procedurally withdrawn during cooldowns from hot standby to approximately 350 F, all CEAs are assumed to be inserted in the core. [ The limiting dilution in this situation is case No. 3 of (29), i The required initial Reactor Coolant System boron concentrations to allow 30 minutes margin to criticality during drained RCS conditions are given , in Table 5.5. Thirty minutes margin is used to bound mid-cycle " refueling" i situations where the reactor vessel head may be removed to perform maintenance l cperations. Table 5.5 also shows the boron concentrations required to meet I the 5% delta K/K Technical Specification suberiticality requirement for I shutdown conditions. Administrative procedures ensure that the higher of the two values in Table 5.5 are used during drained RCS conditions, thus a minimum cf 30 minutes margin to criticality will be provided for the limiting boron dilution event from drained conditions. i I l 4 l _ . _ _ . . . ._. _. - . _ . , _ _ _ _ . _ _ - - , _ . _ _ _ _ _ _ . . ~ . . , _ _ . . _ _ _ . , _ . _ _ . _ . _ _ _ ~ _ . _ . . - . . _

5.3.2.4 Dilution Durinz Hot Standby. Startue. and Power Oooration The assumptions made for boron dilution events during hot standby, startup, and power operation in (3) remain the same (except for inverse boron worth) for Cycle 9. However, the hot standby critical boron concentration with uncertainty is higher, 1557 ppm versus 1275 ppm. The results for Cycle 9 using Figures 4.3-4 and 4.3-5 of (3) are sumanarized below: Maximum Reactivity Insertion Rate Dilution at Hot Standby 10.40 x 10" delta rho /see

                                                ~

Dilution at Po' der 9.28 x 10 delta rho /sec The consequences of events with such small reactivity addition rates cre bounded by the results reported in Section 5.3.1 for the CEA Withdrawal Incident. Based on the maximum reactivity addition rate it would take cyproximately 56 minutes of continuous dilution at the maximum charging rate to completely absorb a 3.2% delta K/K shutdown margin. Because of the cvailable' alarms and indications, there is ample time and information to allow the operator to take corrective action. 5.3.2.5 Failure to Borate Prior to Cooldown Because of the large negative moderator temperature coefficient at EOL, cny decrease in primary coolant temperature adds reactivity to the reactor core. Consequently, during the process of cooling down the Primary System for refueling or repairs, it is necessary to borate in order to compensate for this reactivity addition. The failure to add boron during cooldown was evaluated on the basis of the following assumptions: (a) The moderator temperature coefficient is the most negative value expected with all rods in the core, including uncertainties. (b) The reactor is initially 3.2% suberitical at an average temperature of 550 F (a more conservative condition than the nominal 532 F). i (c) The primary system temperature is reduced at the rate of 100 F/hr, the maximum cooling rate permitted. i In order to make the reactor critical from these initial conditions, , the average coolant temperature must be reduced from 550 F to about 445 F. This temperature reduction requires approximately 63 minutes to l cecomplish. This is ample time for the operat'or to diagnose the condition and f take the necessary corrective action. 5.3.3 Excess Load Incident l An Excess Load In. 1ent is an event where a power-energy removal nisaatch is established leading to a decrease in the reactor coolant average i temperature and pressure. Hence, when the moderator temperature coefficient I cf reactivity is negative, unintentional increases in reactor power may l cccur. Thus, the Excess Load Incident as reported in (3) is most limiting at l EOC where the moderator temperature coefficient is most negative. i i The Cycle 9 MTC with uncertainty is bounded by the Reference Safety Analysis value of 3.17 x 10 delta rho / F which is more negative than the 7

value predicted for cycle 9 including uncertainty.

! Analyses of the Excess Load transient do not credit the action of the I turbine load limiter in mitigating the potential overpower consequences of 1 inadvertent openings of the turbine admission valves at power to less than 10% I tbove the initial power. A general approach to the excess load transient is j used which credits RPS functions only. At various power levels, the limiting positive and negative syneetric offset initial power distributions are j gbtained from the S/0 LCO band in Figure 5.7. From any initial power level f the excess load transient is assumed to start at the S/0 LCO band and { terminate at the symmetric offset and/or variable overpower trip limit. The f MDNBR for the most limiting Excess Load event is 1.63 and corresponds to an

cvent initiated from the positive edge of symmetric offset band at I typroximately the one-hundred percent power level which results in a power increase to the variable overpower trip setpoint.

5.3.4 Loss of Load Incident A Loss of Load transient occurs when the turbine trips While the plant is at power. For Cycle 9 two changes are being implemented which affect this transient. To increase the turbine header pressure, the RCS cold leg temperature will be increased 2 F. In addition, the number of steam scnerator tubes that could be plugged and still remain below a peak system pressure of 2750 psia was determined, thereby allowing margin for future tube plugging. Peak RCS pressure is noted in Table 5.3. Assuming that the pressurizer ssfety' valves open at 2675 psia, the system pressure remains below 2750 psia throughout the transient. The pressure response is not sensitive to the initial RCS temperature; however, plugging steam generator tubes degrades the h at removal from the RCS and forces the pressure up. No more than 180 tubes par steam generator can be plugged before the RCS pressure exceeds 2750 psia. The MDNBR for this event is affected by the increase in RCS temperature. The Loss of Load transient MDNBR is greater than 1.69. 5.3.5 Loss of Feedwater Incident A Loss of Feedwater transient occurs when the feedwater supply to the steam generators is discontinued while the plant is at power. For Cycle 9 two changes are being implemented Which affect this transient. To increase the turbine header pressure. the RCS cold leg temperature will be increased 2 F. In addition, 250 tubes per steam generator are assumed to be plugged, thereby allowing margin for future tube plugging. Peak RCS pressure for the Loss of Feedwater transient is bounded by the Loss of Load transient pressure of less than 2750 psia.

Both of the Cycle 9 specific changes tend to increase the RCS temperature, which decreases the transient MDNBR. A COBRAIIIC analysis with FSAR peaking was performed to determine the MDNBR for Cycle 9. The core inlet temperature was increased to account for the direct RCS temperature increase cnd for the increase in primary to secondary delta T caused by the degradation in steam generator heat transfer. The predicted MDNBR of 1.69, noted in Table 5.3, is well above the YAEC-1 correlation limit of 1.20. For a loss of feed transient from full power with the single failure of cne auxiliary feedwater pump, the steam generator level reaches a mininum of 36.4% of the tube bundle height 15.5 minutes after the low level trip occurs. This level provides adequate heat sink throughout the transient. 5.4 Anticipated Operational Occurrences Which are Dependent on Initial Overpower Martin for Protection Against Violation of SAFDLs The incidents in this category rely on the provision of adequate initial overpower margin to assure that they do not result in violation of the Specified Acceptable Fuel Design Limits (SAFDL). These incidents are reviewed here, with the parameters listed in Table 5.1, in order to demonstrate that the incidents of this category do not violate the SAFDLs, primary system pressure limits, or site boundary dose limits (10CFR20) under Cycle 9 conditions. 5.4.1 Loss-of-coolant Flow Results of the Loss-of-Coolant Flow analysis are sensitive to initial overpower DNB margin, rate of flow degradation, low reactor coolant flow reactor trip setpoint, available scram reactivity, and moderator temperature coefficient. The limiting overpower DNB margin within the LCO envelope for Cycle 9 is greater than that assumed in the Reference Safety Analysis of this event inherent in the use of the FSAR design power distribution, as discussed in Section 5.1.2. The assumptions pertaining to MTC, low reactor coolant flow trip setpoint, and rate of coolant flow degradation remain the same as in the Reference Safety Analysis for this event. The available scram reactivity assumed for Cycle 9 (Table 5.1) bounds that assumed for the Reference Safety Analysis. Thus, the minimum DNBR for the three pump loss of flow from 100% power using the FSAR design power distribution is greater than 1.51. 5.4.2 Full Length CEA Drop The drop of a full length CEA results in a distortion of the core power distribution and could lead to the violation of SAFDL. As discussed in S3ction 6.4.1, the LCO symmetric offset band is designed to restrict p;rmissible initial operating conditions such that the SAFDL for DNB and fuel j centerline melt are not exceeded for this incident. i The Reference Safety Analysis of this incident identified the limiting transient as one initiated from near full power. To cover all potentially limiting conditions the CEA drop for Cycle 9 was evaluated from power levels r:nging from 0% to 100% of 2630 MWt. Power distributions used in the evaluation of DNBR and proximity to fuel centerline melt were selected at each power level from the limiting cases within the S/0 LCO band. The initial percent increase in peaking as a function of dropped CEA worth for Cycle 9 is given in Figure 4.11. The value for the maximum increase in peaking for any dropped CEA from Figure 4.11 was conservatively (Section 4.9.3.1) applied at each power level considered. The CEA drop analysis also considers the increased peaking which r0sults from xenon redistribution during the period of time operation with a dropped CEA is allowed by the Technical Specifications, see Section 4.9.3.2. The percent increase in peaking from Figure 4.11 was conservatively augmented by the increase in peaking due to xenon redistribution at subsequent points in time, assuming operation consistent with the power level reductions required by the Technical Specifications. The margins to the SAFDLs were then determined for the limiting power distributions within the symmetric offset LCO band allowed for the existing power level at any point in time assuming the CEAs to be inserted no deeper than allowed by the insertion limit essociated with the pre-drop power level. The worst case full length CEA drop, with respect to DNB reported in (3), was the minimum worth CEA that results in the maximum increase in peaking. Thus, for conservatism the plant response assumed in the Cycle 9 ovaluation was based on a worth of 0.10% delta rho.

                                                                                                                          --p%,.- u----.--,

The results of the DNB evaluation for Cycle 9 indicate that the limiting DNBR during a full length CEA drop occurs 4.0 hours after a drop from 100%, after the power has been reduced by 20%. A KDNBR of 1.29 occurs at the p;sitive edge of the offset band at this time. The worst case full length CEA drop with respect to fuel centerline melt is one initiated from power distributions at the edge of the S/O LCO band et each power level. The maximum allowable steady-state linear heat scneration rate is limited to assure that the maximum post-drop linear heat g neration rate does not violate the limiting centerline melt SAFDL. These limits are reflected in deriving the LCO band on symmetric offset. The safety analysis of the CEA drop event assumes that control of the turbine admission valves is performed manually. Following the drop, core power initially decreases due to the negative reactivity associated with the inserted CEA. The power level is assumed to return to its pre-drop level under the influence of the moderator reactivity feedback induced by the lowering of the core average temperature by continued heat removal to match the constant steam demand of the turbine throttle setting. During operation with turbine admission valve control in Impulse IN mode,-(INPIN), the admission valves automatically react to changes in steam flow or pressure to maintain impulse pressure at the inlet to the first stage of the high pressure turbine constant. Thus, if the inlet steam flow and pressure were to drop, as they do immediately after a CEA drop, the throttle valves would open in an attempt to restore the impulse pressure to its set value. While operating in the IMPIN mode the potential exists, due to either a single failure of the IMPIN control logic or overshoot of the controller during the return of core power and SG pressure for the transient steam demand, to exceed the pre-drop power level. This could cause core power to return to values higher than the pre-drop level. The potential power overshoot is limited by the variable overpower trip ! setpoint to a maximum of 10 percent above the initial power level. Since the allowable range of plant operation with respect to core power distributions (i.e., symmetric offset LCO band) is determined by the locus of points at which the DNB or LHGR SAFDLs would not exceed their limits following a CEA drop. Any decrease in margin due to power levels returning above the pre-drop level affects this LCO. Consequently, a suitably conservative S/0 LCO operating band for the IMPIN operating mode which protects both the DNB and LHGR SAFDLs has been d veloped by lowering the normal S/0 LCO by an amount equal to the maximum patential increase in post-drop power level (determined by the V0PT setpoint) fcr the IMPIN mode. Both the nornal and IMPIN S/0 LCO bands are shown in Figure 5.7. 5.5 Postulated Accidents The incidents in this category were previously analyzed in (3), (6), (7), (37), (38), (50), and (53). For the conditions in those reports it was demonstrated that each of these incidents met the appropriate accident criteria. Each of these incidents have been reviewed below and results of new cnalyses reported when Cycle 9 conditions warranted reanalysis of the accident. 5.5.1 Steam Line Rupture Accident The last complete analysis of the Steam Line Rupture (SLR) accident was performed for Cycle 6. The analysis was reviewed and approved by the NRC . The system analysis code RETRAN-01 MOD 3 was used for Cycle 6 to predict the consequences of a double-ended guillotine break in the main steam line coincident with a single failure in the safety grade systems. The goal cf the RETRAN analysis is to determine if the core returns to criticality citer the initial reactor scram. This is very conservative with respect to the actual criteria, which require that fuel damage be of sufficiently limited cxtent that the core will remain in place and intact with no loss of core cooling capability and the calculated offsite doses not exceed the guidelines values of 10CFR Part 100.

                                                                         -100-

_ _ . . . - _ . . ._ _ _ . _ , _ _ _ _ , _ ~ _ - _ , - . - _.

New analytical methodology has been adopted for the Cycle 9 SLR cecident. Central to the Cycle 9 analysis is the use of RETRAN-02 MOD 2. This version of RETRAN incorporates improvements to the non-equilibrium volume model and steam generator heat transfer options which allow better prediction cf the SLR transient response. More detailed nodalization of the reactor v ssel upper head and cold les regions of the RETRAN model was implemented to improve natural circulation flow predictions, while more detailed nodalization cf the boron transport model allowed more accurate prediction of the boron concentrations. Adequate margin to suberiticality is demonstrated it the cvailable scram reactivity and boron worth is larger than the reactivity due to moderator and Doppler defects at all times during the accident. Uncertainties in each of these reactivity components was statistically combined by the root-mean-square method and subtracted from the available margin. Figures 5.8 and 5.9 illustrate the Cycle 9 RETRAN-02 model. A more complete discussion of this model is given in YAEC-1447, ' which was transmitted to the NRC via. Since the SLR transient is an overcooling event, the assumptions which d fine the limiting case must force the worst possible RCS cooldown while limiting the availability of boron injection into the RCS. Many single failure scenarios were considered in defining the limiting SLR case. Careful study of the main steam isolation system showed that no single failure would cllow more than one steam generator to blow down through a break anywhere in the steam line. Postulating the break at the steam generator outlet nozzle icwers friction resistance and maximizes the steam flow, so this location was used in all cases. Single failures of safety grade components in the fcedwater, safety injection, main steam isolation, and auxiliary feed systems were considered. Both the loss of offsite power and offsite power available cituations were considered. Non-safety grade equipment was assumed to fail unless its operation would aggravate the transient. Five single failure cases were chosen to represent the worst possible combinations of RCS cooldown and low boron concentrations. These cases are: a) Main feedwater regulating valve fails open at Hot Full Power (HFP) conditions. This failure allows the feedwater system to provide

                                              -101-                                                     -

the maximum possible amount of feedwater flow to the faulted steam generator, thereby maximizing the RCS cooldown from HFP. b) Main feedwater regulating valve fails open at Hot Zero Power (HZP) conditions. c) Main feedwater bypass valve fails open at HZP startup conditions with a coincident loss cf off-site power. d) High Pressure Safety Injection (HPSI) pump fails at HZP conditions with a coincident loss of off-site power. The loss of one HPSI pump causes the worst case reduction in the RCS boron concentration during a SLB transient. This allows a longer RCS cooldown before the injected boron reactivity overrides the moderator and Doppler defects. e) Excess Flow Check Valve (EFCV) fails open at HFP conditions with a coincident loss of off-site power. In this case, an EFCV which controls steam flow from one of the intact steam generators fails open, providing a path for high pressure steam to reach the turbine driven main feedwater pump. This pump continues to deliver feedwater to the faulted steam generator until the Feedwater Isolation Signal is received, even though off-site power has been lost. Several conservative key assumptions were made in this analysis. These cre: a) The trip setpoints and valve ramp times were chosen to maximize the cooldown and minimize the boron injection in all cases studied. b) Pure steam blow-down through the ruptured steam line. This maximizes the energy removed from the system. c) The reactivity balance was based on fluid conditions in the faulted loop cold leg. No credit was taken for coolant mixing in the downcomer, inlet plenum, or core.

                                       -102-

d) The safety injection pipes were assumed to be deborated to the reverse flow check valves upstream of the HPSI pumps. e) The SLR transient was initiated from the maximum allowable RCS temperature. This increases the moderator defect, since the moderator density change per degree is larger at higher temperatures. The RETRAN analysis of the five cases described shows that a feedwater regulating valve failure coincident with a SLR from HZP (Case b) gives the least margin to recriticality. This limiting case for Cycle 9 differs from the Cycle 6 limiting case because of changes in the steam generator heat transfer options, as discussed in (57). Figures 5.10 through 5.21 illustrate the response of several important system parameters during the limiting transient. A regulating valve failure from full power is the most limiting HFP case and is only slightly less limiting than the zero power case. Figures 5.22 through 5.34 show the response of this transient. Table 5.7 gives the nominal scram reactivity necessary to avoid recriticality for full and HZP cases at BOC and EOC, along with the nominal cvailable scram reactivities for Cycle 9. The minimum margin is 0.70% delta rho. Since the nuclear uncertainties in the available scram reactivities have been statistically combined in the SLR analysis, the available scram reactivities listed in Table 5.7 are nominal values as measured at the plant. 5.5.2 Steam Generator Tube Rupture The analysis of the SGTR event performed in the Reference Safety Analysis was reviewed for its applicability to Cycle 9. The primary system response is mainly a function of the initial system pressure and the time of reactor trip. The nominal operating pressure remains unchanged from the value cssumed in the Reference Safety Analysis. In the Reference Safety Analysis the reactor trip occurred at the thermal margin trip setpoint. The results of the Reference Safety Analysis adequately represent the primary system response to SGTR during Cycle 9.

                                        -103-

5.5.3 Seized Rotor Accident The consequences of the seized rotor accident are sensitive to the initial overpower DNB margin, core power distribution, radial pin peak census, ccsumed rate of flow degradation, reactor coolant low flow trip setpoint, MTC, cnd the primary to secondary leakage rate. Most of these factors remain unchanged in Cycle 9. The important differences for Cycle 9 are a reduction in initial cverpower DNB margin, due to higher RCS temperatures and an increased number of shim pins, and changes in the radial pin peak census. The MTC is more p:sitive for Cycle 9, but is bounded by the assumed value of +0.5 x 10~ d21ta rho / F. The fraction of fuel failure predicted during a seized rotor event was cnalyzed using the FSAR power distribution and pin census for Cycle 9. The rssults for Cycle 9 indicate that less than 7.5% of the fuel pins experience DNB as compared to 4.0% for Cycle 8. These results were calculated assuming a censervative time for low flow reactor trip based on the rate of flow decrease from a single pump coastdown rather than an impeller seizure. Radiological release analyses show that the release limits of 10CFR100 will not be exceeded even if all pins experiencing DNB fail; therefore, the predicted consequences of a seized rotor event during Cycle 9 are acceptable. 5.5.4 CEA Eiection The consequences of a CEA ejection accident are most sensitive to ojected CEA worth, effective delayed neutron fraction Geff), and post-ejection peaking. Specifically, the severity of the transient increases for higher ejected CEA worths, smaller delayed neutron fractions, and increased post-ejected peaking. A comparison between the values assumed in the Reference Safety Analysis, and those predicted for Cycle 9, including uncertainty, is presented in Table 5.1. In each case, the values for Cycle 9 cxceed one or more of the parameters assumed in the Reference Safety

                                        -104-

Analyses. Thus, a reanalysis of each case has been performed for Cycle 9 using the modified methodology described in (55). As discussed in Section 5.1.3, the values for MTC assumed in the CEA ojection analysis correspond to the more positive of either the value given in Table 4.5, with +0.5 x 10~ delta rho / F uncertainty added, or those shown in Figure 4.8. A +15% uncertainty continues to be applied to the ejected CEA worth. CEA ejections were specifically analyzed at intermediate power levels ranging from 0 to 100%. Similarly, the CEA scram worth versus position curves used (Figures 5.4 and 5.5) correspond to the negative edge (i.e. bottom skewed power distribution) of the symmetric offset trip band for Cycle 9 at low power levels, thereby minimizing the inserted worth versus time after scram. A summary of the results for the HFP and HZP cases is presented in Table 5.8. All cases investigated resulted in a radially averaged fuel cnthalphy of less than 280 cal /gm at any axial location in any fuel pin. A bounding radiological release calculation shows the resulting off-site doses to be within 10CFR100. 5.5.5 Loss of Coolant 5.5.5.1 Introduction and Summary The LOCA analyses performed for Cycle 5 through Cycle 8 presented in (33, 37, 50, 53) serve as the reference LOCA analyses for Cycle 9. Additional analyses were also performed to account for minor differences between Cycle 9 and reference cycle system configuration, physics, and thermal hydraulic parameters. Based on the results of these analyses it is concluded that Appendix K criteria are met for Cycle 9 fuel types operating upto the following limits: Fresh Fuel 13.5 kW/ft for X greater than 0.50 and CAB L less than or equal to 792 MWD /MTU 14.0 kW/ft for I greater than 0.50 and CAB L greater than 792 MWD /MTU

                                           -105-

16.0 kW/ft for I less than or equal to 0.50 L Exposed Fuel 14.0 kW/ft for X greater than 0.50 L 16.0 kW/ft for I less than or equal to 0.50 L where X is fraction of core height and CAB is cycle average burnup. L Fresh fuel is comprised only of CE Type N fuel. Exposed fuel is comprised of ENC Types L and M fuel and CE Type E fuel. 5.5.5.2 Larne Break LOCA Analysis Large break LOCA calculations were performed for Cycle 5 through Cycle 9 using YAEC's WREM based Generic PWR ECCS Evaluation Model (16). These calculations consisted of break spectrum, burnup, and axial power distribution cnalyses for Cycle 5 (33) and selected sensitivity studies for Cycle 6 through Cycle 8 (37, 50, 53). The system configuration, physics and thermal-hydraulic parameters for Cycle 9 are slightly different from those of previously analyzed cycles. These differences may be classified into two categories; those affecting the system response to LOCA, thereby af fecting the break spectrum analysis and, 'those specific to a fuel type, thereby affecting the burnup study. Parameters af fecting break spectrum analysis: o An increase in cold leg temperature of 2 F over that used in previous analysis. o 250 plugged tubes per steam generator as compared to 28 assumed for previous analyses, o A moderator density defect curve that is not bounded by Cycle 5 curve at densities lower than 26 lb/ft .

                                          -106-

o The core loading in Cycle 9 is modified as shown in Table 3.1. Parameters affecting burnup study: o Increased fission gas plenum volume in CE fresh fuel, o The burnup on three cycle exposed fuel (ENC-L) at EOC 9 exceeds previously analyzed burnups. 5.5.5.2.1 Break Spectrum Analysis The break spectrum analysis performed for Cycle 5 showed 1.0 DECLS to b2 the most limiting, i.e., yields the highest PCT (1978 F). The 1.0 DECLS ccse was re-analyzed to quantify the effect of system changes on PCT. The re-analysis yielded a PCT of 1977 F. This is within 20 F of the previous cnalysis result, thereby proving the applicability of the Cycle 5 break cpectrum analysis to Cycle 9. 5.5.5.2.2 Burnup Sensitivity Studies The burnup sensitivity analyses performed for Cycles 5, 6, 7, and 8 were reviewed for their applicability to Cycle 9. The result of the review is given in Table 5.9. Three additional cases were analyzed for Cycle 9. These cre NBOC9*, W25D9, and LEOC9. The PCT for all these cases were predicted to be within 2200 F, and are given in Table 5.9. The Batch L fuel was analyzed at EOC9 conditions, for top skew and cosine axial power shapes. For the top skew shape, the hot assembly was calculated to rupture just prior to the start of refill phase (Hot Channel Analysis). The flow blockage caused by the rupture was conservatively treated in the Flood Calculation. In the " flood model", the core is assumed to consist of all hot essemblies (16). Therefore, the total core flow area was reduced by the eKey example - NBOC9 is Batch N fuel at beginning of Cycle 9 conditions.

                                        -107-

blockage fraction calculated by the hot channel model. The core loss c:efficients were also modified to account for the blockage. The resulting flooding rates were used in the T00DEE-2 hot rod model to predict the PCT. As reported in Table 5.10, the PCT was calculated to be 2046 F. 5.5.5.2.3 Cosine Axial Power Distribution Study The cosine axial power distribution study performed in the referenced Cnalyses was similarly re-examined for its applicability to Cycle 9. The results of the review are given in Table 5.10. The previously performed Cnalyses for IBOC5 and IBOC6 cases were used to represent the MBOC9 and LBOC9 cases. The NBOC9 and LEOC9 burnup points were analyzed, the results from these analyses are included in Table 5.9. 5.5.5.3 Small Break LOCA Analyses The small break LOCA analysis performed by Combustion Engineering (32) for Cycle 4 considered a spectrum of cold les breaks varying in size from 0.1 to 0.5 ft . Results showed that the limiting break size is the 0.5 ft break with a peak clad temperature of 1348 F, well below the acceptance criteria of 10CFR50.46. A demonstration analysis of the limiting break performed for Cycle 5 (33) utilizing Yankee methodology yielded a peak clad temperature of 1230 F, well below the 10CFR50.46 acceptance criteria and Maine Yankee large break results. Thus, small break LOCAs for Maine Yankee were shown to be not limiting. j The small break results are primarily decay heat driven. The results cf previous analysis are applicable to Cycle 9 because (1) these results are insensitive to fuel type, (2) the slight differences in Cycle 9 and reference cycle system configuration is expected to have minor effect on PCT and (3) the PCT predicted by previous analysis is well below the 10CFR50.46 criteria. Hence, the minor system changes will not make the small break a limiting scenario. 5.6 Plant Modifications There are no plant modifications that need be considered for the Cycle 9 safety analysis.

                                        -108-

Tablo 5.1 Maine Yankee Safety Parameters Reference Cycle 3 Cycle 8 Cycle 9 Including Including Including Uncertainties Uncertainties Uncertainties Parameter Units

 - Planar Radial Peaking Factor                   1.68(2)(3)             1,77(2)             1.81 Bank 5 Inserted to 100% PDIL                   1.72(7)
 - Axial Peak for Shape Resulting in IWWBR at 100% RTP                 1.42(2)                1.48(2)             1.47
 - Augmentation Factors                           1.0 to 1.067(1)        1.0 to 1.049(1) 1.0 to 1.048(1)

- Moderator Temperature 10-4 delta rho /0F 0 to -2.74 +.5 to -2.77 +.5 to -2.81 5 Coefficient

?

 - Ejected CEA Worth BOC Zero Power              %  delta  rho         .396                    .711            .583 BOC Pull Power              %  delta  rho         .210                    .340            .391 EOC Zero Power              %  delta  rho         .544                    .989            .607 EOC Pull Power              %  delta  rho         .230                    .470            .443
 - Ejected CEA 3D Peak BOC Zero Power                                  13.32                 13.38            15.50 90C Pull Power                                   5.53                   6.41             7.20 EOC Zero Power                                  14.08                  16.08           14.40 EOC Full Power                                   5.59                    7.56            7.35
 - Dropped CEA Integral Worth % delta rho            O to .30             0 to .20         0 to .20 i

e

l Table 5.1 (Continued) Reference Cycle 3 Cycle 8 Cycle 9 l Including Including Including l Units Uncertainties Uncertainties Uncertainties Parameter i Figure 4.4-1 of Figure 4.11 of Figure 4.11

    - Dropped CRA Integral Reference 53 Radial Peak                                                                          Reference 3
    - Power Level (including 2% uncertainty)                     ledt                                                2683                                               2683            2683
    - Maximan Reactor Coolant                                                                                                                                          548 - 556 Inlet Temperature                  OF                                                  554                       546 - 554
    - Reactor Coolant System Pressure                            psia                                           2200 - 2300          2050 - 2300                                   2050 - 2300 l
    - Reactor Coolant System l 2               Flow mate                           10' lbs/hr                                    134.6(5)          134.6(5)-136.0(6)                               134.2(5)-135.8(6)

I E' Synnetric Figure 6.3-1 Figure 6.3 of Figure 5.7 ! - Axial Fouer Distribution Limit Offset of Reference 3 Reference 53 Figure 4.9 of Figure 4.9 of Figure 4.9

    - Power E_,- " zt Insertion Limit                                                                  Reference 20      Reference 53
    - Initial Steady-State (4)                                                                                       1.946(5)                                        1,395(5) l                  Minimamm D W Ratio                 YAEC-1                                        1.977             2.092(6)                                        1.911(6) 1
     - Maximan Possible Rate of Reactivity Addition (9) delta rho /sec                                       0.7x10-4          1.22x10-4                                       1.40x10-4 l

Nominal

     - Steam Generator                               psia                                          871               877 - 610                                       877 - 610 l                  Pressure (1001) i l

l l \ _ - _ - . _ _ _ . . . _ _ .

Table 5.1 (Continued) Reference Cycle 3 Cycle 8 Cycle 9' Including Including Including Parameter Units Uncertainties Uncertainties Uncertainties

                     - nominst Steam Generator Level (Berrow Bange)       %                                                       66.                              66.-48.         66.

Steen Geneestor (SC) hates Plugged /SC - - ISO

                     - Available Scram Beectivity Assumed in Safety Analysis % delta rho IFF. 30C                                                                                   4.0                         5.73         5.70 e

N2F. BOC 2.0 3.20 3.20 NFF.EOC 5.7 6.74 6.02 6.5(8) NZP.BDC 2.9 4.21 3.62 s

Table 5.1 (Continued) potts.

1) Applies only in fuel centerline melt calculations.

2) With limiting cycle dependent power distribution as limited by the associated cycle's symmetric offset pre-trip alarm. Power level refers to conditions allowed by PDIL for that cycle.

3) Values shown in Reference 50 did not include uncertainty.

4) FSARdesi5npowerdistribution(F\eltaH=1.49.Fg = 1.68).

Includes 2% calorimetric power uncertainty and 3% allowance for maximum tilt allowed by Technical Specification 3.10.

5) Based on Reactor Coolant System pressure of 2200 psia and reactor coolant inlet temperature of 5560F for Cycle 9 or 5540F for earlier cycles.

6) Based on Reactor Coolant System pressure of 2050 psia and reactor coolant inlet temperature of 5480 F for Cycle 9 or 546 0F for earlier cycles.

7) Banks 5 and 4 inserted to PDIL at 100% per Cycle 3 PDIL.

8) EOC, MFP steam line break assumed 6.5% delta rho.

9) For CEA bank withdrawal transient.

l i l 1 -112-

.c . _ ~ _ - _ _. . . Table 5.2 Maine Yankee Cycle 9 - Incidents Considered A. Anticipated Operational Occurrences for which the RPS assures no violation of SAFDLs:

1. Control Element Assembly Bank and Subgroup Withdrawal

2. Boron Dilution

3. Excess Load

4. Loss of Load

5. Loss of Feedwater B. Anticipated Operational Occurrences which are dependent on Initial Overpower Mars i n for protection against violation of SAFDLs:

1. Loss of Coolant Flow

2. Full Len5th CEA Drop C. Postulated Accidents:

1. CEA Ejection

2. Steam Line Rupture

3. Steam Generator Tube Rupture

4. Seized Rotor

5. Loss of Coolant

                                                         -113-

T-blo 5.3 Maine Yankee Cycle 9 Safety Analysis Summary of Results Reference Safety Analysis Cycle 8 Cycle 9 Incident Section Criteria Cycle 3 MDNBR = 1.51* MDNBR = 1.64 MDNBR = 1.42 CEA With- 5.3.1 MDNBR = 1.20 RCS pressure RCS pressure RCS pressure drawal RCS pressure 2570 psia 2570 psia 2750 psia 2570 psia Not exceeded Not exceeded Not exceeded LHCR SAFDL Cycle 3 Suberitical: Suberitical: Suberitical: Boron 5.3.2 Suberitical: Refueling-65 min. Refueling-30 min. Refueling-30 min. Dilution Sufficient time Startup-3.2 hours Startup-20 minutes Startup-15 minutes for operator action Critical: Bounded Critical: Bounded Critical: Bounded .'. by CEA withdrawal by CEA withdrawal E Critical: MDNBR by CEA withdrawal 8 1.20 Cycle 3 MDNBR = 1.20 MDNBR = 1.36* MDNBR = 1.54 MDNBR = 1.29 CEA Drop 5.4.2 LHGR SAFDL Not exceeded Not exceeded Not exceeded. Cycle 3 MDNBR = 1.2 MDNBR = 1.50* MDNBR = 1.57 MDNBR = 1.51 Lo23 of 5.4.1 Coolcnt Flow Cycle 5 Seized Pump 5.5.3 A sufficiently 8.4% of rods with 4% of rods with 7.5% of rods with low fraction of MDhBR less than 1.3* MDNBR less than 1.2 MDNBR less than 1.2 C:ter rods with MDNBR less than 1.2 Cycle 4 MDNBR = 1.2 MDNBR = 1.7* MDNBR = 1,67 MDNBR = 1.63 Exc= s Load 5.3.3

T-bl9 5.3 (Continued) Section Criteria Reference Safety Analysis Cycle 8 Cycle 9 Incident Cycle 3 Loss of Load 5.3.4 MDNBR = 1.2 MDNBR = 1.85* MDNBR = 1.81 MDNBR = 1.69 RCS pressure RCS pressure RCS pressure RCS pressure

                           <2750 psia        2689 psia                   2689 psia           <2750 psia cycle 3 Loss of         5.3.5   RCS pressure      Peak RCS pressure           Peak RCS pressure   Peak RCS pressure Feedwater               <2750 psia         2600 psia                  2600 psia           <2750 psia MDNBR = 1.20                                  MDNBR = 1.78        MDNBR = 1.69 Cycle 6 Steam Line      5.5.1   Maintain fuel rod Fuel rod integrity          Fuel rod integrity  Fuel rod integrity Rupture                 integrity          is maintained since        is maintained since is maintained since reactor does not           reactor does not    reactor does not return critical            return critical     return critical C.                                                      Cycle 3
'                          10CFR100           Radiological dose          Radiological dose   Radiological dose Steam Gener-    5.5.2 ctor Tube                                  within 10CFR100            within 10CFR100     within 10CFR100 Rupture Cycle 8 CEA Ejection    5.5.4   10CFR100           Radiological dose          Radiological dose   Radiological dose within 10CFR100            within 10CFR100     within 10CFR100 Cycle 5 LOCA            5.5.5   10CFR100           Radiological dose          Radiological dose   Radiological dose within 10CFR100            within 10CFR100     within 10CFR100 PCT =21890F clad           PCT =21490F clad    PCT =21490 F clad 10CFR50.46 oxidation = 15.59%;        oxidation = 5.52%;  oxidation = 5.52%;

Less than 1% Less than 1% Less than 1% hydrogen generation hydrogen generation hydrogen generation Cycle 3 Steam Line - 10CFR100 Radiological dose Reference analysis Reference analysis Rupture Outside within 10CFR100 unchanged by Cycle unchanged by Cycle Containment 8 reload 9 reload

q T-blo 5.3 (Continued) Criteria Reference Safety Analysis Cycle 8 Cycle 9 Incident Section 4 Cycle 3 Feedwater - 10CFR100 Bounded by steam line Reference analysis Reference analysis Line Rupture rupture unchanged by Cycle unchanged by Cycle Outside 8 reload 9 reload Containment Cycle 3 Peak pressure less Reference analysis Reference analysis Containment - Peak pressure Pressure less than 55 psig than 55 psig unchanged by Cycle unchanged by Cycle containment 8 reload 9 reload design pressure 1 1 Cycle 3 Fuel Handling - 10CER100 Radiological dose Reference analysis Reference analysis j i Incident within 10CFR100 unchanged by Cycle unchanged by Cycle 1C 8 reload 9 reload }i i Cycle 3 I Waste Cas - 10CFR100 Radiological dose Reference analysis Reference analysis f Sy: Item within 10CFR100 unchanged by Cycle unchanged by Cycle Failure 8 reload 9 reload Spent Fuel - 10CFR100 NA NA NA j Ccsk Drop i Cycle 3 l Radiological dose Reference analysis Reference analysis l Radioactive - 10CFR100 1.iquid Weste within 10CFR100 unchanged by Cycle unchanged by Cycle j i System Leak 8 reload 9 reload 4 i ) n W-3 DNBR shown for Reference Safety Analysis result, YAEC-1 for Cycles 8 and 9. i i

Table 5.4 Cycle 9 Required Initial RCS Boron Concentrations to Allow 15 Minutes Margin to Criticality for Dilutions from Shutdown Conditions with the RCS Filled Required Initial Concentration (opm) Boron Dilution 5% Delta K/K*

ARI, BOC 5320F 806 1234 3000F 945 1276 680F 974 1265

ARI, EOC 5320F 0 208 3000F 83 384 680F 150 409 ARI Less 1 Stuck CEA, BOC 5320F 968 1401 3000F 1066 1403 680F 1101 1395 ARI Less 1 Stuck CEA, EOC 5320F 0 372 3000F 210 522 680F 275 543 ARI w/ Withdrawn Bank,**

BOC 5320F 1114 1525 3000F 1238 1541 680F 1256 1512

                                          -117-

k Table 5.4 (Cont'd) Required Initial Concentration (com) Boron Dilution 5% Delta K/K* cARI w/ Withdrawn Bank,** EOC 5320F 128 498 3000F 366 629 680F 411 629 ARO, BOC 5320F 1684 2164 3000F 1582 1965 680F 1516 1855 ARO, EOC 5320F 579 1026 3000F 616 963 680F 589 891 i

   # Nargin of subcriticality required by Technical Specifications for shutdown conditions.
   ** The critical boron concentrations for 2 stuck CEAs bound those for a withdrawn bank.                 Therefore, the boron concentrations provided for 2 stuck CEAs were used in the calculations of required concentration for boron dilution events. The 2 stuck calculations also bound intermediate combinations of 2 or more CEAs withdrawn during CEA rod drop testing, i
                                                                                    -    118-v .- - - - --. = -     . . - - .     -   ..-,,--,,-.,,-_,,..,,,7...,_,.-.-,,,,-_s                    --_--,_y_,      . - - . , , , , _ _ , . _ , - _ , - - . . _ , , _ _ . , _ . , , , . , . _ , - .

Table 5.5 Cycle 9 Required Initial RCS Boron Concentrations to Allow 30 Minutes Margin to Criticality for Dilutions from Shutdown Conditions with the RCS Drained

• Required Initial Concentration (com)

Boron Dilution 5% Delta K/K* ARI, BOC 5320F 1235 1234 3000F 1330 1276 680F 1325 1265 ARI, EOC 5320F 0 208 3000F 117 384 680F 204 409 ARI, Less 1 Stuck CEA, BOC 5320F 1482 1401 3000F 1501 1403 680F 1497 1395 ARI, Less 1 Stuck CEA. EOC 5320F 0 372 3000F 297 522 680F 374 543 ARI w/ Withdrawn Bank,*** BOC 5320F 1705 1525 l 3000F 1743 1541 680F 1709 1512

                                                          -119-

Table 5.5 (Cont'd) Required Initial Concentration (com) Boron Dilution 5% Delta K/K* ARI, w/ Withdrawn Bank,*** EOC 5320F 195 498 3000F 516 629 680F 559 629

ARO, BOC 5320F 2578 2164 3000F 2228 1965 680F 2062 1855

ARO, EOC 5320F 887 1026 3000F 868 963 680F 800 891 o Level = lower lip of RV nozzles, c* Margin of suberiticality required by Technical Specifications for shutdown conditions.

C** The critical boron concentrations for 2 stuck CEAs bound those for a withdrawal bank. Therefore, the boron concentrations provided for 2 stuck CEAs were used in the calculations of required concentrations for boron dilution events. The 2 stuck calculations also bound intermediate combinations of 2 or more CEAs withdrawn during CEA rod drop testin5

                                       -120-

Table 5.6 Summary of Boron Dilution Incident Results for Cycle 9 (A) (B) (C) (D) Minimum Technical Minimum Time Operating Specification Shutdown to Absorb (B) Acceptance Mode Margin Requirement Minutes Criteria (5) Refueling 5% delta K/K 30 30 C21d Shutdown Filled RCS 5% delta K/K l', 15 Drained RCS 5% delta K/K(1) 30(1) 30(1) Hot Shutdown Filled RCS 5% delta K/K 15 15 Drained RCS 5% delta K/ kcl) 30(1) 30(1) Startup 3.2% delta K/K(6) 15(2) 15 Hot Standby 3.2% delta K/K(6) 56(3) 15 Power Operation 3.2% delta K/K(6) 56(3) 15 Failure to Borate Prior to Cooldown 3.2% delta K/K(6) 63(4) 15 (1) 30 minutes margin is used to provide sufficient margin for drained conditions where the head is removed. These are classed as " refueling" conditions in the Technical Specification. Margin quoted is for initial boron concentrations administratively required for these conditions. (2) Margin quoted assumes initial boron concentration at refueling value for Cycle 9, 1644 ppm. (3) Time to absorb minimum specified 3.2% delta K/K shutdown margin. (4) Cooldown rate assumed to be 1000F/hr. (5) Time span between event initiation and criticality. (6) Minimum value for shutdown margin specified in Technical Specification.

                                           -121-

Isble 5.7 Nominal Scram Reactivity Worths Required to Prevent a Return to Power During a Steam Line Rupture Accident Nominal Scram Reactivity Gaat Required Available 2.8% 7.12% HFP BOC NFP EOC 6.69% 7.54% HZP BOC 1.29% 4.69% HZP EOC 4.02% 4.72% i

                                                                                                                                                   -122-t

r d l Isble 5.8 Cycle 9 CEA Ejection Accident Results Full power ! E E Fraction of Rods that Suffer Clad 7.3 0.5 l Damage (Radial Average Enthalpy Above 200 cal /sm), % I Fractior. of Fuel Volume Exceeding 1.5 0.1 ! Incipient Neiting Criteria (Enthalpy greater than 250 cal /sm), % l 3ero power Fraction of Rods that Suffer Clad 0.0 0.0 Damage (Radial Average Enthalpy Above 200 cal /ga), % Fraction of Fuel Volume Exceedin5 0.0 0.0 l Incipient Melting Criteria (Enthalpy treater than 250 cal /sm), % i I i

                                          -123-

Table 5.9 Maine Yankee Cycle 9 Burnup Sensitivity Study R_esults Case

• LHGR Hot Pin Peak Cladding Reference Description (kW/ft) Burnup (MWD /MTU) Temperature, OF Analysis 4 NBOC9 13.5 0.0 2045 1.0 DECLS9 N25D9 14.0 1229. 2098 --

NEOC9 14.0 30473. 2135 IEOC5 MBOC9 14.0 17452. 2149 JEOCS MEOC9 14.0 33573. 2109 JEOC6 LBOC9 14.0 33573. 2109 JEOC6 LEOC9 14.0 50062. 2046 --

• Key example - MBOC9 is Batch M fuel at beginning of Cycle 9 conditions.

                                                                 -124-

Table 5.10 Maine Yankee Cycle 9 Cosine Axial Power Shape Study Results Case

• LHGR Hot Pin Peak Cladding Cycle 9 Description (kW/ft) Burnup (MWD /MTU) Temperature, OF Equivalent NBOC9 16.0 0.0 1954 _-

MBOC9 16.0 17452. 1854 IBOCS LBOC9 16.0 30473. 1768 IBOC6 LEOC9 16.0 50062. 1738 --

• Key example - MBOC9 is Batch M fuel at beginning of Cycle 9 conditions.

                                       -125-

Table 5.11 Comparison of Thermal Margin for Limiting Cycle 9 Power Distributions to FSAR Design Power Distribution Power Power YAEC-1 YAEC-1 Level Distribution MDNBR (Pd-Po) 100 FSAR(3) 1.895 30 100 (2)(3) 2.018 32 93 (2)(3) 1.997 31 73 (2)(3) 2.150 35 64 (2)(3) 2.915 48 47 (2)(3) 3.675 39 100 FSAR(7)(4) 1.946 33 100 FSAR(7)(5) 1.962 34 100 FSAR(7)(6) 1.919 31 100 FSAR(7)(8) 1.911 31 (1) Includes allowances for 2% calorimetric power uncertainty. 3% tilt, and

 .       10% physics radial peaking uncertainty on non-FSAR cases.

(2) Limiting cycle 9 power distribution within symmetric offset pre-trip alarm band plus uncertainty for indicated power level for Cycle 9. j (3) At 2200 psia, 556 degrees F. (4) Cycle 8. (5) Cycle 3. (6) Cycle 5. (7) At 2200 psia, 554 degrees F. (8) Cycle 6

                                            -126-

Table 5.12 Reactor Protective System Trips Assumed in the Cycle 9 Safety Analysis Delay Setpoint Uncertainty Time (Sec) High youtron Flux 106.5% i5.5% 0.4 Low Reactor Coolant Flow 93% 12% 0.65 High Pressurizer Pressure 2400 psia i22 psi 0.9 Low Steam Generator Pressure 500 psia 122 psi 0.9 Low Steam Generator Water Level 35% NR 110 in 0.9 Low Pressurizer Pressure

• 1850 psia 22 psi 0.9 Safety Injection Signal 1600 psia 122 psi **

Thermal Margin Trip Figure 5.5 (8) 0.9 and 5.6 Symmetric Offset Trip Figure 5.7 .04 asiu 0.9 Variable Overpower Trip Q + 10%*** 15.5% 0.4

• Low limit of thermal margin trip.

** See specific accident for time delay assumed for safety injection delivery.
***Q = Initial indicated power level in percent thermal or nuclear power.
                                         -127-

Table 5.13 Required Scram Reactivity Assumed in Cycle 9 Safety Analysis Required Scram Reactivity (% delta rho) Event BOC. HZP EOC. HZP BOC. HFP EOC. HFP CEA Withdrawal - 4.00 4.00 B2ron Dilution 3.20 3.20 - Excess Load 3.20 3.20 - 5.70 L:ss of Load - 4.00 - Loss of Feedwater - 4.00 - Loss of Coolant Flow - 5.50 - CEA Drop - CEA Ejection 3.20 3.20 5.70 5.70 Steam Line Rupture

• 1.16 3.62 2.52 6.02 Steam Generator Tube Rupture 2.00 2.40 4.00 5.70 Saized Rotor - 4.00 -

Less of Coolant - Maximum Assumed in Any Event 3.20 3.62 5.70 6.02

• An uncertainty factor of 0.9 is applied to the nominal required scram reactivities assumed for the Steam Line Rupture event from Table 5.7 for comparison to the available scram reactivities with uncertainties assumed for the other events. This uncertainty component is statistically combined in the Steam Line Rupture analysis with the other uncertainty components to derive the nominal required scram reactivities for that event as discussed in (57).

                                               -128-

2300  !

                                                                                                                                                 , t _                 .          ,            i .                          .i.
                       'i
                       .i                UNACCEPTABLE OPERATION                                                                                  ii.'                                          !. 'i
                                                                                                                                                                                                                    !                     I
                                                                                         .              I I            t             i i         l il                             l            Ii                  i               t i I I i i                       -            i i i i i                                       '4             i i i i                   i i i                            i             i i                   ,i l i
                       ! } l i,                      ,                                   i t)                 ;                i           . . .                                .                                        i                            !          !             . ' i' l i i l                .             !               i                    i i
                                                                                                                                          -            i i i                      i                                         i!

ii ! ;  ; i i i i i ! i E i i i i  ! i i *  ! . i  ! i i

                              . '                                           .                    i i i                  i            I i               t i .

A ' ' '

                                                            ! i Ve 2250                                i                                                        i                    i                   , i ,
                                                                                                                                                                                   ',                   .                   i              i r                                              i
                                                                                          .                                                      f f '                              i                   .                   .

M . e i + . . i i l o i i , , , , i _ i , , i , ,

         .2             '

ii .

                                                                                                            !           !! I i                   i               .

J i i i 6 .i

                                                                                                                                                                                                                            !              i i

i . t ' 4 i

         .E                    !     ' !               !

ACCEPTABLE OPERATION . i ; i i i . . v . . I i t i .  ; i

                                                                                                                             i      I
                                                                                                                                                   ' '                   l            [          l        i                  i
           @                                                                                                                                                             i i                             i L.

a 2200 , , , , , , m y) i i i  ! ! t i i i , I l

                                                                                                                                                                                                 .        i           i i            ,
                                                        !     ,t                             ,      i         t           i i 6 .                       . 4                           i 0                           a
                                                                                                                                                         .                             .         i        ,           ,

i L. .

0. '

BREAK POINTS i , i i

                                                                      ;                                 (2275 , 500)                                    ;          ,
                                                                                                                                                                                      ;                   l           , ,

E . o i (2275 , 552)  ! i a ' ' m (2225 , 552) ' g 220  ; l' (2075 , 544) l i  !

           ,                      l l

(2075 , 500) , l , 4 c . . i i C i i

                                          '                                                                                                 1              ' -               !                                                  <

o o i . o i ii i i i ,

                                                                                                                                                                     . .                                    i                   i
                                                                                                                                        ,                   i ,

6. o 210 0 , , , , , ,

           "                                                                       ,           .                                        , ,                 ,                 i           .
                                                                                                                                                                                                                         .       4 i

O . ,

                                                                                                          ,,               . ,               .                       . .                  i g                             ,
                                                          '                                                      i
                                                                                                                                             '      .        , i i                                           i            ,                   .

o .  : , i

                                                                                               '. i
                                                                                                                           .       4 1

i, (r .i . ,, . i i i

                                                                                                                                                                      ! i i

i "i t i i 3

                          .         i ii .i!                                     i           ! t > i                             . ! I             _

i 6 i . i i4 i e i . l 6 i i I , j UNACCEPTABLE OPERATION l l  ! l l 2050 490 5b0 5k0 520 530 540 550 560 Nominal Cold Leg Temperature (indicated), deg F MAINE YANKEE Allowable 3 Loop Steady State Figure Coolont Conditions 5.1

                                                                                                        -129-

N kh

                    ~~                          .J r       ,
                                                                                                        =

l ~ I I I I. I I

              ~

l l I - m l n l I I \ - ~ l o I I  ; I - m I I I i

                                                                                                     . 8 y I                                                                        O  @

I s C f

                .e                                                                                         e
              -5     l                                                                                ' n e 8
                .e                                          x          ,                                   O 5

5 I I W39e J ~ J C

                                                                                                          .9
                                                                                                          ~
              ~

C  ;

                                                                                                      - 9 o

8 0 a ( -  % o

              .

• N o

_ - 9 O C o O N w ' , E o 52o O N c' C 53.E..M3N O z<co AO7hpog pr O. Figure Design Axial Power Distribution 5.2

                                              -130-                                                           Rev. 6/71

l FIGURE 5.3 MRINE YRNKEE CYCLE 9 NORMALI2ED RERCTIVITY WORTH VS PERCENT CER INSERTION CER EJECTION: BOC, SCRAM RT HFP RND HZP 1.0 0.9 2

0. 8 -

% 0.7 I 8 i g 0.6 - { 5 i: - 00.5- - M B c 0.4: - id E - @ 0. 3 - HFP

0. 2 -

H2P _ 0.1_ g egu 'l' 'l' 'l' 'l' 'l' 'l'

                       =#'   'l' O    10   20   30    40       50     60    70   80       90     100 PERCENT CER INSERTION
                                     -131-

FIGGE 5.4 MAINE YANKEE CYCLE 9 NORMALIZED REACTIVITY WORTH VS PERCENT.CER INSERTION CER EJECTION: EOC, SCRAM RT HFP AND HZP 1, g , 0.9 2 4 0.8 2 % 0.7 - 5r  : 0.6 W !3 - a- 0.5 - e N 0. 4 - d E - -

@  0.3 -

HFP 0.2 - HZP 0.1 : r e wgv vga wgu vgw wgw g ugi igi 0 10 20 30 40 50 60 70 80 90 100 PERCENT CER INSERTION

                                      -132-

WHERE: Q=A3

• QR3 TRIP = 1959.2 10053.0 AND P + 17.9T -

WR C T = COLD LEG TEMPERATURE, F C 1.40 __ _ _ .__. 1.35 L'

i l

1.30 - .' >

                                                                                                                                                                 /           i          :i
                                                                                                                                                             /               1          ll l                                           Ag(+)=.74021(S.04)+1.02635                                                   -/'

l;; ' 1.25 - 1 o . 1

                                                                                                                     ;   ;___q______j
                                                                                                                                                                             ;i I
                    ,                                                 i                  i                                                        /!                               1 Fi I          i                   I Y 1

< 1.20 - .  ! l l

                                                                                                                          .                I
                                                                                                                                            /i i       j____             _    _    _   _._      a_   ._   .i. _  , __

_y

                                                                                                                                           ':                                            i M: Aj(-)= .35535(S.O.)+.98735                                                                                                /

1.15 *3 kI i1 E

                             \'!                                                                                                    f-l i                 %::                    ---

1 [M-l I i I 2 [^ E i I 1.10 _, _ _ _ . _ _ ____j i

                                                                                                                       .J
           +v                                      $                 _        __-

I 1 I I I I 'L I ;F1 1.05

                                                                     \..

j' .i X,j  ! i 1

                                                                     ----Q----j
                                                                     ~~--

_4 r_ j ,j

                      +.         -_,__           _ .__.__._               _    _._  _                      . _    __        ___ _._ ____ _.                   . _ _  . _.
         -0.5 -0. 4                         -0.3                -0.2               -0.1                   0.0               0.1          0.2           0.3              0. 4                      0.5 Excore Symmetric Offset MAINE YANKEE                                    Thermal Margin / Low Pressure Trlp Setpoint                                                                                                                  Figure 5-                                                                                                            Port 1                                                                                          5.5 (A verm Excore SpmeMe OffseQ 1
                                                                                                        -133-

l l WHERE: QDNB " A1

• 1 TRIP - 10053.0 AND P VAR = 1959.2 QDNS+ 17.9TC T = COLD LEG TEMPERATURE, F C

J 1.2 .a

/

                                                       '--"--- --~~

1.0 ~ ;r 3r-

                                                                           --V
                                                                     /

E 0.6 l O "

.- , f.

0.6 ( 120 ,t2000 ) 2-o, , _ -- -- --,'

                                      -  -/
                                             --                                      ( 100 ,10000 )          __
                       ~
                                    /                                                ( 0.61 .0.8315 ) EE     --
                         - -- 9 0.2 ggg                                                            ( 0.44 .0.6273 )

( O.10 ,0.1557 ) ==

                                                                                                             ]

I

                                                                        ~~
                          ~~~    ~ -'

0.0 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Roted Thermal Power Thermal Margin / Low Pressure Figure MAINE YANKEE 5.6 , Trip Setpoint Port 2 (QR1 versus Fraction of Roted Thermal Power)

                                                                  -134-

120 y f

                                                                                .                                                                      i           .

i , i i i f i , h I I  :

                                                                                                                                                                                                                  !                                    t
                                                                           ' f 110
                                                                                                                               ,A                        ,
                                                                                                                                                                                                       ,., ,                                    i l         . ii:                                        /,             ,

i ', TRIP LIMIT

                                                                                        ,4                            .,

(-0.5000 ,20.0 )

                                                                                  '.# r'/y                       .
                                                                                                                         .,6 "              ti-x s

(-0.5000 ,70.0 ) 100  ; ': i

                                                                                                                                                                                     ;                                                           ,               (-0.2400 ,100.0 )
                                                                                                          .                                                       i
                                                                                                 /                            ,        ..     >

( 0.0000 ,110.0 ) A"N g T ( O.2000 ,100.0 )

                                                                           " '                                                                                                                                                                    l 0                                                                                               g

( 0.5000 ,70.0 )

                                                           'l i                d                                                  . .                .               T 90                                                        ./                  .                   -                                                  -                    ,                ,

( 0.5000 .20.0 ) f ;f;  :  ;- '

                                                                                                                                                                     ,                   ,7
                                          ;                  /                 !!                f--                            l
                  ,!                      !               /                  f:.                " !!
                                                                                                   ;I!

i '.* 6

                                                                                                                                                                                                 ~A m

i i ii I F F ' t t  !! E  ; _  :/ LCO PRE-TRIP w ,, , , 80 , . 1 .. . ALARM l 7 l r-c , l R l L' ' i .. o y r y

                                                                    =
                                                                               'yi
                                                                                                                                                                                                                                             ;;                   (-0.4000 ,20.0 )

a '

                                                                                                                                                                                                                                             'lii-                (-0.4000 ,70.0 )

11

                                                                            /1                      . . -
                                                                                                                                          ,     l                                                 i m

1 i _J . . ,, :ii 1 . U  : (-0.1400 ,100.0 )

      <    70                        :

(- 0.010 0 ,10 5.0 ) 3

/; ,

                                                                                                                                                                                                               ',-                                                ( 0.1000 ,100.0 )
      @                                                                          !;j                                                i i

( 0.4000 ,70.0 ) F l f: I 4 ( 0.4000 ,20.0 ) a 80 . W  !. ' ' '

                                                                                                                            !;                    l-                                                                           l                     ;

g .

                                                                                                                      ,                           ii i!

LCO PRE-TRIP g ALARM 50 IMPIN MODE o ' - (-C.4000 ,60.0 ) g (-0.1400 ,90.0 ) I (-0.0100 ,95.0 )

             #                                                                                                                                                                                                                                                     ( 0.1000 ,90.0 )

( 0.4000 ,60.0 ) 30 1 20

               .-C.8-0.5 -0.4 -0.3 -0.2 0.1 0.0 0.1 0.2 0.3 0.4 0.5 0.6 Symmetric Offset Y, = Ae(U-L/U+L)+8 Symmetrie Offeet Trlp Function                                                                                                                                                                   Figure MAIPE YAPEEE                                                                                                                                                                                                                                                              5.7 Three Pump Operation
                                                                                                                                      -135-

FIGURE 5.8 RETRAN PRIMARY SYSTEM MODEL ( x s-._ e - -- >_ .

                                                    ~

0 e s

                 @, @t_@t                     _                    3
                                                                                                          -f 3
                              @      =e
                  =       ,          e         .                                                e ua                                       :         :          @-1                       >

j u si l.l ,e e . 4

                  .. . .s.

vv" vv W T i g ?2U 1, t@ 31 ,.

                                 >-b 00 aN*=
                                                                 "i G      n-        :
      't> {2L                                           j a               2 m e                                             r
                                                                                    ...K.

__j --l y a -- gg g

• n e) ' '

g, - e N

                                                    ' ' 91 n                @l~                         *
                                        @                   . l@                                           ._-

c , _ 2 g __F d /

                                            <                                                                   /
                                                      ---lc g.l*1= =
                                                               ,       yni*

H e g . a T f g_- _= , e -- - s a = e 3

 )
                     @.26 s@- 1                                 \

i- as , - t

                     @*-- @"-- @'                                                       @                  y ec              r             ,
                 /
                                              -136-

FIGURE 5.9 RETRAN SECONDARY SYSTEM MODEL o 11f E5o

                                 $3                      r                  (g                              _ ,3 4,O                         c
                                                                                                -        ~r, z f
                                       ~                                                   >                            N m               o                   e                     > *                           \

s  ; -

                                                                                %          v    u W        /\
                                     $                     n                  f-'s m,        15
                                                                                             ,x          -

as

             -                              ,                              1          s e     v                    1 w               i na?D                                    1 "h5    5 si E                                            !                                        -
                               .g.'@
                                   } S-                                                                              1 s.i,                       --
                                                              .                  @                   O--         @ liy 0                                             N Y

M b

 $        7            'i Y       &                                                 n
              ~

f) Yh g g r-- e-Kq

                                                                                  @.,- @ eN/

E

                                        @                                                          n 3- .
          *             ,n, =

1 < *# gt gM - III , ,;

                                             ,;                                                   I; e
                            -      ,          ,,                 =

3 2 ' " " #

        /

{ .-

                     ;                y i       's s        --

ln  ? ~

                                                     \
      ;; E         h            >

l

                                                                                     ~

El i, f , j %, e _ _

                                                        ,                             s 8

t 1 gi t s i j .

                                                                                        #                \

N_____--- g

                                                              ~~

gy_, e t@ - c @ g  ! 2 1 - t x.

                                                                                                                       )

a> s

                                                                         -137-

dIDE3 S'IO o i i i g i g

• o m -

7 E - 1T-m O N m o DC I

                                                         \

g-m m 0: C 3 "r

  .-.                                                                     t-E                                                                   o Z
  <                                                                       m 0::

E hC m Z 3 OI o2

   !t-

r. i-N 2

IIC

   ~r                                                              -   o hc        -                                                         m 3

O.t m

    ~r o                                                              -o
                                                                       +

O s I I I I I 0002 0051 0001 005 @ 0052

                )RISP( ERUSSERP REZIRUSSERP SPMUP RETAWDEEF CIRTCELE HTIW ERULIAF EVLAV GNITALUGER RETAWDEEF NIAM *
                             -IC8-

2o . 0 2 _ _ ~ _ 3 0 i

                                      !     8 2

0 4 S i 2 P W F ) E C / E W 6S i

                                       !    0 2(

E R E U . M L I I T R o F i I sT i N V E R I W S F N M 0A i 2R P 1 T Z H B L S i

                                        !    0 M                                             8 9

E L C I 0 Y i 4 C _ _ ~ _ b g= g= o O- gN OE m- m m o

FIGURE 5.12 o i I I I N o g w w a_ 3 b ^ b.J O N W 3

                                            -     8m
                                                  ~-

w M W D z _J - .-. W C W -

                                             -    8g z

> W M # 3 W L_ z r oC NM 1 n m I LD J W - g r - m LLJ i J U - o y -

• O I i i 1 009 03S Ott 09E 093 Od2 (3) dW31 037 0703 13

                         -140-

__q FIGURE 5.13 I O M I I I o g O w - Z n_ Z

                                                              ~

LL LLJ C) N LL.I x yk _ gm u-L'l I z $ Lu E D

• H C o e _

                                                          =H z

> L1J z >-* 3 CO LL- Z E oC N CE' O. "H N I CD CO - o co r CD LLI _] O _ o O I l l l 1 00006 00004 00009 0000E 00001 0000@ (S/97) Mold 3803

                         -141-

CYCLE 9 MSLB HZP MFWRV FAILURE W/EFWPS i i i i I i i o N H - i 13 m _J W N a i W _ m b' U v. e g W N e-. t M - w. en W E Q. - I I i i i I I 160 200 240 280 d20 c0 40 80 120 TRANSIENT TIME (SEC) i

FIGURE 5.15 e I I i 1

 -                                                                   o
                                                               -     e a

w - g n_ 1 - u_ o W w N

                                                           .      . 8m u-m                                                                      w E                                                                       r D                                                                      -
   -s                                                                      s-
   ~

e C - I e, p Lt. - z w z m 1 z u_ oc r o-iw e 1 r N r m

    .s                                                             mee m        _

r cn w J _e u

• u I i i t

9

• 3 E OT 9

                 ,    ( .L .-l )  13A31 063H hl3ddn
                                      -143-

o CYCLE 9 MSLB HZP MFWRV FRILURE W/EFWPS o I I I I I I I m (D o l N8 (D 4D i J 1 w en 3 I 3 5

      % Oc         -

S l T LL.J8to tn M 5 C to Eo - (D o - 1 0 (D J i ko - o - i o b - 1 I r 1 I I i i 120 160 200 240 280 320 c0 40 80

TRANSIENT TIME (SEC) 1 i

l i i

ZeNwh 0 2

                          -        -        _           3 0

C  ? 8 i 2 0 S I 4 4 2 P W F P ) O E O C

 /                               L                           E W                               T C                  i 0S 0

I A 2( E T N R I E U M L I I T A 0 F I i 61 T N V E . R P I W O O S F L N M ) T, 0A I F - i 12R P T. T fi Z l H F B L S I i 0 8 M 9 E L C i 0 Y i 4 C 0

                 $#         e e-     E )u    gN            c o$-                                _

e C.WQ" wmD mw O &D e A08

                            ,sEsc' wL=

0 2 _ _ _ - 3 0

                                                 .-     I      8 i

O 2 0 1 I 4 S i 2 P W

                                                                     )

F E C / P E W O O 0S 0 i L 1 I 2( E D R E E T M U L L U I A I F T A 0 F i 1 I 6 T 1 N V P E R O I W O S L F T N M P i C A T N __ I 0A 2R 1 T I Z H B L S I 0 i 8 M 9 E L C 1 0 Y i 4 C 0 88, 88M 88Z o8o3 o 8. c e - rJ- 3Eo&Zu>z~ o(o

                          -            t
                                  ,ET

           )il!                  I m$s w.G 0

2

                                   -             -             ~          _                        3 0

I 8 i 2 0 I 4 S i 2 P W ) F E C

          /                                                                                                  E W                                                                                I 0S 0

i 2( E R E U - M I L - I T A 0 F i I 6 1 T N V E R I W S F N M 0A i I 2R 1 P T Z H B L 0 S i f 8 M 9 E L C i 0 i 4 Y C _ 0

• c

- o*

• e ]

N

                          ,gINDFm nWO*

c ~ sDQZ H1WI bM ( .u . b' llllll l Ill llIll l il! ll ,' '

FIGURE 5.20 C N l l l I M l l i l O g O e - Z a. 3

                                                                    ^

Lt. LLJ O N W 3 _ 8m N-LAJ M LLJ p r

            ._f                                                     ~
            ~                                                       6--

C O L -

                                                           ~

a Z

            >                                                        LLJ M                                                        ~

3 U') LL Z E oC

                                                            - Ng L
                     ~
                                                                 "W N

I CD

             ._J v3                                              - a g        -

m CD LLJ l J O - o p - O I I I I 000t 003E 00t3 0091 008 9 (S/97) M0ld Md ld

                                     -148-

Il U!m *

• 0 2

                   ~                -                -          -         3 0

I i 82 0 S I

                 .-                                                    6 4 2

. P W F ) E C

      /                                                                          E W

E I 6 0 0S 2( R E U M L I I T A 0 P 6 61 T F I O O N L V E R T I C W A T S F N N M I 0A I 6 12R P T Z H B L S I 60 8 M P O 9 O L D E E L T L C U 60 4 Y I A F C

              // (s 0

Om oN  :- . o- @ c nDN ( o o J_ " 2oJL 2L X01 t 2>$ l llII ' ll l1,l1 ll  ! 11

l l UgoM: w 0 2 _ - ~ 3 0 I I 8" 2 0 I 4 S I 2 P W F ) E C

       /                                                                         E N                                                            I 0S 0

I 2(

   '   E R                                                                         E U                                                                         M L                                                                         I I

T R 0 F I I 6 T 1 N V E R I W S F N M 0A i I 2R P 1 T F H B L S i I 0 M 8 9 E L ' C l 0 Y i 4 C i 0 b- ,m , O N c

                       ^ N w eJ2Oa.

L L

                                            - C J.J e
                                                  .L L
                                      ,- o8 1                      II    l                  il l   1. l !'        l        l 1

lljl-M@N * . U 0 2 _ _ 3 0 l 8 i 2 0 l 4 S i 2 P W _ ) F C E

               /                                                                                                       E W                                                                                       l 0S 0

i 2( E R E U M I L T T A 0 l 6 T F i 1 N V E R I W S F N M 0A i l 2R 1 P T F H B L S t 0 i 8 M 9 E L C e t 0 4 Y C 0 8g oo N oo3 E og c

                                - CCwD(t" WMDmnWxG    (
                                                                        -         nn MwN~EDicWMO L*,

7 llll ll lI' f1l .l1lll'\,1lIli l l l l

U O E u*e> o 0 2 _ - - _ - 3 0 I I 8 2 0 I 4 S I 2 P W F ) E C

   /                                                        E W                                              I 0

0 S I 2( E R E U M L I I T R 0 F I I 6 T 1 N V E R I W S F N M 0A I I 2R 1 P T F H B L S I I 0 M 8 9 E L C I I 0 Y 4 C O. gm  ?* 0m gu c% 3 mi mrW>~ W 0U i 'i

Ug wU 0 2

           -       -          ~            _               3 0

I 8 i 2 0 I 4 S P i

                                                ~

2 W F ) E C / E W I 0S 0 i 2( E R E U M L I I T A I 0 6 r i 1 T N V E R I W S F N M 0A i I 2R 1 P T F H B L S i I 0 8 M 9 E L C 1 0 4 Y i C

             -        -         ~            -

gN h oE ge o o Om n_ ~Ww I dw a o !_. to L

g (nh. 0 2 _ - _ _ 3 0 I i 8 2 a 0 4 S I i 2 P W F ) E C

 /                                                             E W                                                          0S I

i 0 2( E R E U M L I I T A 0 F I i 6 1 T N V E R I W S F N M 0A I i 2R P 1 T F H B L S I i 0 M 8 9 E L C i 0 Y i 4 C O o 8 c) m 8gr 8gm o8e 8g- 88T mN - 2Oa orOU to i; 1

mEE[ U 0 2

                     -          ~         _   3 0

i 8 2 0 S , 4 2 P W F ) E C

   /                                             E W                                          0S i                                    l 0 E                                          2(

R E U M L I I T R 0 F i I 6 T 1 V N R E I W S F N M 0A P i I 2R 1 T F H B L S i I 0 M 8 9 t L C I 0 Y i 4 C

               )

o 0 g S c c d ) ji gmlEQ .

             - 5 m>mJ     etN h Y'

l CYCLE 9 MSLB HFP MFWRV FRILURE W/EFWPS I I I I I I I I e-O 4

                            ^                                                                                      _

H LL. D m

                            -I i

W m

                            >                                                                                           E c

1 i W - G Je $ T o v.

E ls W

I - M* W Q. Q. 3 4 N ' l 1 1 1 1 - - c0 40 80 120 160 200 240 280 ~320 TRANSIENT TIME (SEC) 4 i

mseM Yw* 0 2

                     -            -        -           -         3 0

i 8 i 2 0 I 4 S i 2 P . W  ;

                                                                       )

F ., E C

 /                                                                     E W                                                          I 0S 0

i 2( E R E U M L I I T R I 0 6 T F i 1 N V E R I W S F N M 0A i I 2R 1 P T F H B L S i I 0 8 M 9 E L C I 0 Y i 4 C 0 g8 o: cgw

                                  )        og*         0gN           c m

m I._ w 1OJL %CIEo L L ein J_.1 1$

mhw $ 0 2

                           -      -        r               3 0

T 8 1 C 2 0 C T 4 S 3 2 P W ) F E C / E W T 02( 0S l P O E O R L E U T M C I L A P T I T O N 0 R I O L . I 6 T F l 1 D E N V T E L I R U W A F S F N M 0A 2 I 2R 1 P l T F H B L 0 S I I 8 M 9 E L C I 0 I 4 Y C 0 ay 8us gN c o N- $- e c<m0* cs W DWmW bD C 1wY

FIGURE 5.31 o I I I I k O o <i - g o o <\ - ~ en - N CL 2 u_ _ W L >.. . g

• m _ 8 E

_ <i 8 u-w ' z o W 3 w r J

 ~

w A C b_ _ <l - e g Z

 >                                                                 W z                                                                 ~

2 8; y S oC g <i - we e < ~

u. M r

cm

 .1 2        -                          i                       -

g en w _.J M - ) - g o I

                                       /

l i I i 00000t 0000ZE 0000t2 000091 00008 3

             .        (W97) A801N3ANI 09
                               -159-

FIGURE 5.32 I O m I I I o g i e _ g a.- 3 ^ LL. h' W  ! o s o Z W

                                                              =h _

% U D E _J ~ C _ S g W _ Z > W 3 M Lt. Z E oC we e u_ I CD

• _J (D -

                                                          ~

E en W J

                                                           -   o f

o 1 I i 1 8 9 t 3 9 01 (8H/019 630*1) 10dNI 183H OS ld

                            -160-

FIGURE 5.33 S I I I I m o g . o_ L o _  ?

                                                       ~

E N 3 - o Z W W C U) e - g_ o W

    >                                                      r M                                                      -

1 & LL o r _

                                                       =~

Z t W a_ - LL U) I Z o CC

       =

Nx D W J l i (D r N r _ g N O o, I i l 1 000t 003E 00tZ 0091 008 9

                     -(S/97) M0ld MJ 13
                               -161-

m$E YY ' 0 2 _ _ ~ _ 3 0 i 6 82 0 S i 6 42 P W F ) E C

 /                                                                  E W                                                              0S i        ^                                            6 0 2(

E R E U M L I I T R 0 F i 6 6 1 T N V P E R 0 0, I W I S F T N M C A 0A I T N 6 12R P I T F H B L P O S I O 6 08 M. L D 9 E T L U E A F L C 60 4 Y I C 0 O CN o"- g c e C J. "

                     -(O N D       EIk 2    2       XoT 19

6.0 STARTUP TEST PROGRAM The startup test program includes low power physics and power cccalation tests for the purposes of:

1) Verifying that the core is correctly loaded and there are no anomalies present which could cause problems later in the cycle;

2) Verifying that the calculational model used will correctly predict core behavior during the cycle.

The low power physics tests are conducted at a power level less than 2.0% of r;ted full power with a primary system temperature and pressure of cpproximately 532 F and 2250 psia, respectively. 6.1 Low Power Physics Tests The following reactor parameters are measured at the low power ccnditions:

1) Critical boron concentration is determined at unrodded and, if required, selected rodded positions.

2) The integral worth at the hot zero power condition of control rod groups 1, 2, 3, 4 and 5 in the non-overlap condition. The total of the worths of these groups must be within i10% of the predicted value. If this condition is not met, then Banks A, B and C will be measured and the sum of the worths of all the groups must be within 110% of predicted.

3) The most limiting near full power ejected CEA worth is measured at the pre-ejection conditions by the boron dilution technique.

4) The isothermal temperature coefficient is measured by trending moderator temperature and reactivity changes. The measurement is performed at unrodded and, if required, a rodded condition.

                                           -163-

5) Control rod drop times are measured by monitoring reed switch voltage for position indication versus time. All scrammable full length CEA drop times are measured.

6.2 Power Escalation Tests y Power escalation tests assure the performance of various primary and secondary plant systems. Plant parameters are stabilized and test data taken ct approximately 48% and approximately 100% of rated power. The following plant parameters are evaluated at the above power levels, er as indicated:

1) Core radial power distributions at essentially unrodded conditions at the above power levels are determined using the fixed incore detector system.

2) Isothermal temperature coefficients, if required, are derived at 48% power by partially closing the steam turbine governor valves which increase reactor coolant system temperature. The result is a change in moderator temperature and power level from which the coefficient is inferred.

6.3 Acceptance Criteria Acceptance criteria for the prediction of key core parameters are defined in Table 6.1. The permissible deviations 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, the latter corrected for any difference between the measurement and calculational conditions. If the criteria in Table 6.1 are met, verification is obtained that the core characteristics conform to those assumed in the safety analysis. In particular, compliance with the shutdown margin Technical Specification is demonstrated by the CEA worth and drop time measurements, provided all trippable CEAs remain operable. The acceptance criteria for Cycle 9 are un: hanged from Cycle 8.

                                         -164-

If the initial criteria in Table 6.1 are not met, additional measurements, as prescribed by the table, are performed. In addition, any d:viations are evaluated relative to the assumptions in the safety analysis fcr the given core parameters. Such deviations and their evaluations are reported to the Staff. A startup test summary report will be available en-site within 90 days of the completion of the startup tests.

                                        -165-

Table 6.1 Maine Yankee Cycle 9 Startup Test Acceptance Criteria Measurement Conditions Criteria

1. Measurement within 11%

Critical Boron Hot zero power, near Concentration all-rods-out delta rho of predicted value

2. CEA Bank Worths - Hot zero power, CEA Total worth within 110%

Regulating Banks 1+2+3+4+5 in the of the predicted value non-overlap condition

3. CEA Bank Worths - Hot zero power, CEA This measurement is not Shutdown Banks A+B+C+1+2+3+4+5 tequired if the criteria in in the non-overlap Measurement (2) is met. If condition the criteria in Measurement (2) is not met, the total worth of all CEA banks must be within 110% of the predicted value

4. Ejected CEA Worth Hot zero power, pre- Ejected CEA worth no ejection CEA banks more than 15% greater inserted for measure- than the predicted value ment of the most limiting near full power ejected CEA

5. Isothernal Hot zero power, near ITC measurement within Temperature all-rods-out 10.5 x 10-4 delta rho /0F Coefficient at of predicted value and the HZP MTC is within the acceptable region specified in Figure 4.8

6. Isothermal At or slightly below 50% This measurement is not Temperature power, near all-rods-out required if both criteria L

Coefficient at in Measurement (5) are met. 50% Power If either criteria in Measurement (5) are not net, the MTC must be in the acceptable region specified in Figure 4.8

7. Control Rod Drop Operating temperature Drop times to 90% insertion Times no greater than 2.70 seconds

                                          -166-

Table 6.1 (Cont'd) Maine Yankee Cycle 9 Startup Test Acceptance Criteria Nessurement Conditions Criteria

8. Radial Power At or siishtly below Each assembly average Distribution 50% power, near all- power within 110% of rods-out predicted value

9. Tilt Monitoring 5-48% rated power, near Tilt trends to less than for Symmetry all-rods-out, tilt is 3.0% for greater than 50%

Verification monitored at 5% power power operation, as intervals indicated by the relative changes in excore detector readings or incore detectors

                                  -167-

7.0 CONCLUSION

The results of analyses presented herein have demonstrated that design criteria as specified in the FSAR and the NRC ECCS Acceptance criteria (31) will be met for operation of Maine Yankee during cycle 9. Table 5.3 rummarises the results of each incident analyzed; including the Reference Cycle result and the appropriate design limit. This table illustrates that Specified Acceptable Fuel Design Limits (SAFDL) on DNB and fuel centerline molt, the primary coolant system ASME code pressure limit, and the 10CFR100 site boundary dose limits are not violated for any of the incidents considered. 1 1

                                           -168-

8.0 REFERENCES

1. Maine Yankee Letter to USNRC, WMY 77-75, dated August 1, 1977.

2. Maine Yankee Atomic Power Station Final Safety Analysis Report (FSAR).

3. P. A. Bergeron, P. J. Guimond, J. DiStefano, " Justification for 2630 MWt Operation of the Maine Yankee Atomic Power Station", YAEC-1132, dated July 1977.

4. USNRC Letter to Yankee Atomic dated January 17, 1978. USNRC Letter to Yankee Atomic dated April 11, 1978.

5. ACRS Letter to J. M. Hendrie, Chaiman USNRC from S. Lawroski, Chairman ACRS, dated June 7, 1978.

6. P. J. Guimond, P. A. Bergeron, " Justification for Operation of the Maine Yankee Atomic Power Station with a Positive Moderator Temperature Corfficient", YAEC-1148, dated April 1978.

7. Maine Yankee Letter to USNRC, WNY 78-62, dated June 26, 1978, " Maine tankee Proposed Change No. 64".

8. P. A. Bergeron, D. J. Denver, " Maine Yankee Reactor Protection System Setpoint Methodology", YAEC-1110, dated September 1976.

l 9. R. N. Gupta, " Maine Yankee Core Themal-Hydraulic Model Using COBRA IIIC", YAEC-1102, dated June 1976. l

10. R. W. Gupta, " Maine Yankee Core Analysis Model Using CHIC-XIN", YAEC 1103, dated September 1976.

11. T. R. Hencey, " GEMINI-II - A Modified Version of the CEMINI Computer Program", YAEC-1068, dated April 1974.

12. P. A. Bergeron, " Maine Yankee Plant Analysis Model Using GEMINI-II",

YAEC-1101, dated June 1976.

                                           -169-

13. W. J. Szymezak, " Maine Yankee Plant Accident Analysis Model Using FLASH-4", YAEC-1104, dated November 1976,

14. D. J. Denver, E. E. Pilat, R. J. Cacciapouti, " Application of Yankee's Reactor Physics Methods to Maine Yankee", YAEC-1115, dated October 1976.

15. P. A. Bergeron, " Maine Yankee Fuel Thermal Performance Evaluation Model",

YAEC-1099P, dated February 1976 (Proprietary).

16. YAEC-1160, " Application of YANKEE WREM-BASED Generic PWR ECCS Evaluation Model to Maine Yankee", July 1978.

17. USNRC Letter, R. W. Reid to R. H. Groce, dated May 27, 1977.

18. USNRC Letter to Yankee Atomic dated January 17, 1979.

19. Letter, R. H. Groce (YAEC) to B. H. Grier (NRC), " Transmittal of MY Startup Test Report, Cycle 5", WMY 80-93, (June 4, 1980).

I 20. Maine Yankee Letter to USNRC, WNY 76-129, dated November 18, 1976, " Maine l Yankee Proposed Change No. 52".

21. KN-73-25 GAPEK: A Computer Program for Predicting Pellet-to-Cladding Heat Transfer coefficients, August 13, 1973.

i 22. D. S. Rowe, " COBRA IIIC: A Digital Computer Program for Steady-State and Transient Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements", BNWL-1695 (March 1973).

23. USNRC Letter to Yankee Atomic, dated January 7,1977.

24. E. S. Markowski, L. Lee, R. Biderman, J. E. Casterlin, "Effect of Rod Bowing on CHF in PWR Fuel Assemblies", ASME paper 77-HT-91.

25. CEN-93(M)-P, " Maine Yankee Reactor Operation with Modified CEA Guide Tubes", June 21, 1978.

                                               -170-

26. Maine Yankee Letter to USNRC, dated October 12, 1979.

27. Keine Yankee Letter to USNRC, WMY 78-102, dated November 15, 1978, " Maine Yankee Startup Test Report".

28. CENPD-139-P-A, "CE Fuel Evaluation Model Topical Report", July 1, 1974.

29. Maine Yankee Letter to USNRC, WNY 78-2, dated January 5, 1978.

30. Maine Yankee Letter to USNRC, WNY 75-28, dated March 27, 1975, " Maine l

Yankee Core 2 Performance Analysis".

31. " Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled Nuclear Power Reactors" Federal R* sister, Vol. 39, No. 3-Friday, January 4, 1974.

l

32. Mainc Yankee Letter to USNRC, WMY 77-87, dated September 22, 1977.

i

33. Maine Yankee Letter to USNRC, WMY 79-143, dated December 5, 1979; Attachment A, YAEC-1202, " Maine Yankee Cycle 5 Core Performance Analysis", P. Bergeron, et al.

34. KN-75-32 (NP) Supplement 2 " Computation Procedures for Evaluating Fuel Rod Bowing", July 1979.

35. R. E. Helfrich, " Thermal-Hydraulic Analysis of PWR Fuel-Element Transients Using the CHIC-KIN Code" YAEC-1241 March 1981.

36. Combustion Engineering Report. TR-DT-34. "The Hydraulic Performance of the Maine Yankee Reactor Model", June 1971.

37. YAEC-1259, " Maine Yankee Cycle 6 Core Performance Analysis" Attachment to NYAPC Letter to USNRC, FMY-81-65, Proposed Change No. 84, dated April 28, 1981.

                                           -171-

38. Cycle 6 MSLB Analysis, Attachment to NYAPC Letter to USNRC, FMY 81-162, dated October 29, 1981.

39. C. L. Wheeler, et al., " COBRA-IV-I: An Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores", BMWL-1962 (March 1976).

40. J. Handschuh, "DNBR Limit Methodology and Application to the Maine Yankee Plant," YAEC-1296P, January 1982 Attachment to YAEC Letter to USNRC, FYR-82-41, MM-32-78, dated April 8, 1982.

41. R. VanHouten, " Fuel Rod Failure as a Consequence of Departure from Nucleate Boiling or Dryout", NUREG-0562, Office of Nuclear Regulatory Research, USNRC, Washington, D. C. 20555 (June 1979).

42. MYAPC Letter to USNRC, MN-82-165, " Maine Yankee Reportable Occurrence 82-23/01T-0", dated August 19, 1982.

43. MYAPC Letter to USNRC, MN-82-53, " Boron Dilution During Hot and Cold Shutdown (Mode 5 Operation)," dated March 18, 1982.

44. CENPD-190, "CE Method for Control Element Assembly Ejection Analysis",

Combustion Engineering, January 1976.

45. S. Bian, " Application of Reactivity Weigiting to Rod Ejection Accident Analysis in a Pressurized Water Reactor" Nucl. Tech., Vol. 41, 401-407 (1978).

46. MYAPC Letter to USNRC, MN-82-157, " Maine Yankee Reportable Occurrence 82-21/01T-0", dated August 11, 1982.

47. USNRC letter to NYAPC, dated December 11, 1982.

       " Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 60 to License No. DPR-36, Maine Yankee Atomic Power Company, Maine Yankee Atomic Power Station, Docket No. 50-309", attached to USNRC Letter to MYAPC, dated December 11, 1981.
                                         -172-

48. XM-NF-81-85 " Mechanical Design Report Supplement fcr Exxin Nuclece Main 3 Yankee IN-1 through IN-4 Extended Burnup Program", November 1981.

l ! 49. J. B. Yasinsky, "On the Use of Point Kinetics for the Analysis of Rod-Ejection Accidents", Nuclear Science and Engineering: 39, 241-256 (1970).

50. YAEC-1324, " Maine Yankee Cycle 7 Core Performance Analysis",

September 1982.

51. ENC Letter to YAEC, " Reverification of Fuel Mechanical Design Analysis Results (IN-NF-81-85) using NRC Approved Methods", TJH-82-423, November 15, 1982.

52. USNRC Letter to MYAPC, dated March 9, 1983, NNY 83-62, " Topical Report YAEC-1296P, "DNBR limit Methodology and Application to the Maine Yankee Plant".

53. YAEC-1396, " Maine Yankee Cycle 6 Core Performance Analysis", January 1984.

54. CEND-414. "The Evaluation and Demonstration of Methods for Improved Fuel Utilization", October 1983.

55. YAEC-1464, " Modified Method for CEA Ejection Analysis of Maine Yankee Plant", Dece'.:ber 1984.

56. FMY-81-162 Decket No. 59-309. December 11, 1981.

57. YAEC-1447, " Application of RETRAN-02 Mod 2 to the Analysis of the MSLB Accident at MYAPC," T. D. Radcliff, M. P. LeFrancois September 1984.

l 58. MM-84-204, Docket No. 50-309, "RETRAN-02 Main Steam Line Break Model Application for Maine Yankee", Cover Letter for YAEC-1447 Transmittal. October 26, 1984.

                                                                       -173-
                                                                                                                         )

59. MYAPC L*;tter to USNRC, MN-83-76, "R ceter V 0331 Pr ssuriz:d Th rnal Shock (PTS)", April 22, 1983, with Enclosures A, B, and C.

60. MYAPC Letter to USNRC, MM-84-88, " Reactor Vessel Pressurized Thermal Shock (PTS)", June 1, 1984, with Enclosures A, B, C, and D.

61. Maine Yankee Batch N Reload Fuel Design Report, April 1985.

62. IN-NF-79-52, " Maine Yankee Reload Fuel Design Report / Mechanical Thermal-Hydraulic and Neutronic Analyses", 1979.

                                       -174-}}