Text

Cycle 10 Core Performance Analysis
	ML20207P536
Person / Time
Site:	Maine Yankee
Issue date:	12/31/1986
From:	Paul Bergeron, Cacciapouti R, Stephen Schultz; Maine Yankee
To:	;
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ML20207P516	List: ML20207P515; ML20207P530; ML20207P536;
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	YAEC-1573, NUDOCS 8701160226
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MAINE YANKEE CYCLE 10 CORE PERFORMANCE ANALYSIS December 1986 Major Contributors:

Nuclear Engineering Department Reactor Physics Group Transient Analysis Group G. M. Solan M. W. Scott D. G. Adli V. M. Esquillo E. B. Bartlett D. C. Fan B. Y. Hubbard A. S. Fatemi K. B. Spinney M. P. LeFrancois

=

S. VanVolkinburg T. D. Radcliff I

K. R. Rousseau LOCA Analysis Group K. E. St. John i

G. E. Jarka J. Ghaus I

Yankee Atomic Electric Company Nuclear Services Division

-I 1671 Worcester Road Framingham, Massachusetts 01701 I

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j APPROVALS I

Approved By:

/

/M P. A. Bergeron p nager (Date)

Transient Analysis Group 4

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Approved By:

I WJ.Cacci ti, Manager

'(Dafe)

Reactor Ph es Group Approved By:

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2 /3c/89 S. P.' Schultz, Man'ager (Date)

LOCA Analysis Group I

Approved By:

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/ 2.3/ $47 B.C.slifer/ Director

/(Da(e)

NuclearEngfneeringDepartme[nt I

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I DISCLAIMER OF RESPONSIBILITY I

This document was prepared by Yankee Atomic Electric Company

(" Yankee"). The use of information contained in this document by anyone other I

than Yankee, or the Organizatio-for which this document was prepared under contract, is not authorized and, with respect to any unauthorized use, neither Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or I

representation as to the accuracy or completeness of the material contained in this document.

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ABSTRACT 1

This report presents design and analysis results pertinent to the operation of Cycle 10 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 I

to justify plant operation, the startup test program and the Reactor Protective System (RPS) setpoints assumed in the safety analysis. The I

analysis results, in conjunction with the startup test results, RPS setpoints and Technical Specifications,. serve as the basis for ensuring safe operation of Maine Yankee during Cycle 10.

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TABLE OF CONTENTS l

Page I

APPR0VALS........................................................

11 DISCLAIMER.......................................................

iii ABSTRACT.........................................................

iv TABLE OF C0NTENTS................................................

v LIST OF TABLES...................................................

viii LIST OF FIGURES..................................................

xi

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 I

2.4 Cycles 5 and 6.............................................

4 2.5 Cycles 7 and 8.............................................

4 2.6 Cycle 9....................................................

5 3.0 RELOAD CORE DESIGN...............................................

8 3.1 General Description........................................

8 3.1.1 Core Fuel Loading..................................

8 8

3.1.2 Core Burnable Poison Loading.......................

8 3.1.3 Core Loading Pattern...............................

9 3.1.4 Assembly Exposure History..........................

9 l

3.1.5 CEA Group Configuration............................

10 i

l 3.2 Fuel System Design.........................................

11 l

l l l 3.2.1 Fuel Mechanical Design.............................

11

! E 3.2.2 Fuel Thermal Analysis..............................

14 l

3.2.3 Thermal-Hydraulic Design...........................

15 t

4.0 PHYSICS ANALYSIS.................................................

33 4.1 Fuel Management............................................

33 I

4.2 Core Physics Characteristics...............................

33 l

l 4.3 Power Distributions........................................

33 l

4.4 CEA Group Reactivity Worths................................

34 l

4.5 Doppler Reactivity Coefficients and Defects................

34 i

4.6 Moderator Reactivity Coefficients and Defects..............

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4.7 Soluble Boron and Burnable Poison Reactivity Effects.......

36 4.8 Kinetics Parameters........................................

36 4.9 Safety-Related Characteristics.............................

37 j

I 4.9.1 CEA Croup Insertion Lim 1ts.........................

37 4.9.2 CEA Ejection Results...............................

37

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TABLE OF CONTENTS (continued)

Page 4.9.3 CEA Drop Results...................................

38 4.9.3.1 Design Analysis Results..................

38 4.9.3.2 Post-CEA Drop Restrictions...............

38 4.9.4 Available Scram Reactivity.........................

39 8

4.9.5 Shutdown Margin Requirements.......................

41 4.9.6 Augmentation Factors...............................

43 4.10 Pressure Vessel F1uence....................................

43 4.11 Methodology and Methodology Revisions......................

44 4.11.1 Summary of Physics Methodology Documentation.......

44 4.11.2 Removal of Augmentation Factors....................

44 5.0 SAFETY ANALYSIS..................................................

74 5.1 Genera1....................................................

74 5.1.1 Initial Operating Conditions.......................

74 5.1.2 Core Power Distributions...........................

75 8

5.1.3 Reactivity Coefficients............................

76 5.1.4 Shutdown CEA Characteristics.......................

77 5.1.5 Reactor Protective System Setpoints and Time Delays....................................

79 5.2 Summary....................................................

80 I

5.3 Anticipated Operational Occurrences for Which the RPS Assures No Violation of SAFDLs.............................

81 5.3.1 Control Element Assembly Bank Withdrawal...........

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5.3.2 Boron Dilution.....................................

82 5.3.2.1 Dilution During Refueling................

83 5

5.3.2.2 Dilution During Cold, Transthermal, and Hot Shutdown With the RCS Filled.....

83 5.3.2.3 Dilution During Cold, Transthermal, I

and Hot Shutdown With Drained RCS Conditions...............................

84 5.3.2.4 Dilution During Hot Standby, Startup, I

and Power Operation......................

84 5.3.2.5 Failure to Borate Prior to Cooldown......

85 1

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TABLE OF CONTENTS (continued) i Page 5.3.3 Excess Load Incident...............................

86 5.3.4 Los s of Load Inciden t..............................

86 5.3.5 Loss of Feedwater Incident.........................

87 5.4 Anticipated Operational Occurrences Which are Dependent on Initial Overpower Margin for Protection Against Violation of SAFDLs........................................

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5.4.1 Loss-of-Coolant F10w...............................

87 3e..J 2

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5.4.2 Full Length CEA Drop...............................

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- 2 5.5 Postulated Accidents.......................................

90

$.f, 5.5.1 Steam Line Rupture.................................

90 k

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\\'J 5.5.2 Steam Generator Tube Rupture.......................

91 N..S 5.5.3 Seized Rotor Accident..............................

92 p.AT..

A 5.5.4 CEA Ejection.......................................

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5.5.5 Loss of Coolant....................................

93

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6.0 STARTUP TEST PR0 GRAM.............................................

119 1.*c."

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- i 6.1 Low Power Physics Tests....................................

119 av<j 6.2 Power Escalation Tests.....................................

120

.'..2 i p +,.i 2 6.3 Acceptance Criteria........................................

120

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7.0 CONCLUSION

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123

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8.0 REFERENCES

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I LIST OF TABLES 1

Number Title Page 2.1 Operating History Summary 6

2.2 Fuel Assembly Types by Cycle 7

3.1 Cycle 10 Assembly Description 17 3.2 Cycle 10 Core Loading 18 3.3 Mechanical Design Features of Cycle 10 Fuel 19 I

3.4 Maine Yankee Cycle 10 Centerline and UO2 Melt Temperature Comparison 20 Cycle 10 Ratio of Maximum Radial Relative Pin Powers -

I" 3.5 Maximum in Type E Fuel to Maximum in Core 21 3.6 Cycle 10 Ratio of Maximum Radial Relative Pin Powers -

Maximum in Type M Fuel to Maximum in Core 22 3.7 Cycle 10 Ratio of Maximum Radial Relative Pin Powers -

Maximum in Type N Fuel to Maximum in Core 23 3.8 Cycle 10 Thermal-Hydraulic Parameters at Full Power 24 4.1 Cycles 3, 9, and 10 Nuclear Characteristics 45 4.2 Cycles 3, 9, and 10 CEA Group Worths at HFP 46 I,

4.3 Cycles 3, 9 and 10 Core Average Doppler Defect 47 Cycles 3, 9 and 10 Core Average Doppler Coefficient 48 I

4.4 4.5 Cycles 3, 9 and 10 Moderator Temperature Coefficients 49 l

l l 4.6 Cycles 3, 9 and 10 ARI Moderator Defect with Worst

'W Stuck CEA 50 4.7 Cycles 3 and 10 Kinetics Parameters 51 4.8 Cycles 9 and 10 CEA Ejection Results from Full Insertions 52 E

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I LIST OF TABLES (continued)

Number Title Page 8

4.9 Cycles 9 and 10 CEA Drop Results at BOC 53 1

4.10 Cycles 9 and 10 CEA Drop Results at E0C 54 4.11 Cycles 9 and 10 Dropped CEA with Power Level Restriction -

Most Limiting Peaking Cases 55 4.12 Cycle 10 Available Scram Reactivity 56 4.13 Cycle 10 Required Scram Reactivity 57 4.14 Cycles 6 through 10 Relative Pressure Vessel Fluence Comparison 58 4.15 Physics Methodology Documentation 59 5.1 Maine Yankee Safety Parameters 94 5.2 Cycle 10 - Incidents Considered 98 5.3 Cycle 10 Safety Analysis - Summary of Results 99 I

5.4 Required Initial RCS Boron Concentrations to Allow Fifteen Minutes Margin to Criticality for Dilutions from Shutdown Conditions with the RCS Filled 102 5.5 Required Initial RCS Boron Concentrations to Allow Thirty Minutes Margin to Criticality for Dilutions from Shutdown Conditions with the RCS Drained 104 5.6 Summary of Boron Dilution Incident Results for Cycle 10 106 8

5.7 Nominal Rod Worths to Prevent a Return to Power During a Steam Line Rupture Accident 107 5.8 Cycle 10 CF.A Ejection Ac it's :: 1-uits 108 8

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LIST OF TABLES (continued)

(

i Number Title Page e

5.9 Comparison of Thermal ifargin for Limiting Cycle 10 Power L

Distribations to FSAR Design Power Distribution 109 j

5.10 Reactor Protective System Trips Assumed in the Cycle 10 l

Safety Analysis 110 L

5.11 Available Shutdown Margin Assumed in Cycle 10 Safety Analysis 111 I

L 6.1 Cycle 10 Startup Test Acceptance Criteria 122 L

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LIST OF FIGURES I

Number Title Pm 3.1 Cycle 10 Burnable Poison Shim Assembly Locations 26 3.2 Cycle 10 Assembly Loading Pattern 27 3.3 Cycle 10 Calculated Assembly Exposures at BOC 28 3.4 Cycle 9 Burnup Distribution by Assembly near 6000 MWD /MT 29 I

3.5 Cycle 10 CEA Croup Locations 30 3.6 Maine Yankee Cycle 10 BOC Centerline Temperature Vs. LHGR 31 3.7 Maine Yankee Cycle 10 E0C Centerline Temperature Vs. LHGR 32 4.1 Cycle 10 Assembly Relative Power Densities BOC (500 MWD /MT), HFP, ARO 61 I

4.2 Cycle 10 Assembly Relative Power Densities MOC (6,000 MWD /MT), HFP, ARO 62 5

4.3 Cycle 10 Assembly Relative Power Densities EOC (13,000 MWD /MT), HFP, ARO 63 4.4 Cycle 10 Assembly Relative Power Densities BOC 8

(500 MWD /MT), HFP, CEA Bank 5 Inserted 64 4.5 Cycle 10 Assembly Relative Power Densities MOC (6,000 MWD /MT), HFP, CEA Bank 5 Inserted 65 4.6 Cycle 10 Assembly Relative Power Densities EOC (13,000 MWD /MT), HFP, CEA Bank 5 Inserted 66 4.7 Cycle 10 Allowable Unrodded Radial Peak Vs. Cycle Average Burnup 67 4.8 Cycle 10 Moderator Temperature Coefficient Limits Vs.

Power Level 68 4.9 Cycle 10 Power Dependent Insertion Limit (PDIL) for CEAs 69 4.10 Cycle 10 Reference Power Level Vs. Nominal Cold Leg 8

Temperature 70 4.11 Cycles 9 and 10 Maximum Radial Peaking Vs. Dropped CEA Worth from Specified Power Levels 71

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LIST OF FIGURES 4

(continued) l i

Number Title h

i 4.12 Cycle 10 Shutdown Margin Equation and Required Scram Reactivity 72 4.13 Cycle 10 Required Shutdown Margin Vs. RCS Boron Concentration 73 5.1 Cycle 10 Allowable 3 Loop Steady-State Coolant conditions 112 5.2 Design Power Distributions 113 5.3 Normalized Reactivity Worth Vs. Position Assumed in BOC CEA Ejection Analysis 114 5.4 Normalized Reactivity Worth Vs. Position Assumed in EOC CEA Ejection Analysis 115 5.5 TM/LP Trip Setpoint (Y1 Versus A )

116 1

5.6 TM/LP Trip Setpoint Part 2 117 5.7 Symmetric Offset Trip Function 118 I

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1.0 INTRODUCTION

I als report provides justification for the operation of Maine Yankee during the s fuel cycle, Cycle 10.

The Cycle 10 refueling will involve the discharge of 73 assemblies and the insertion of 72 new fuel assemblies and one burned Type E assembly previously irradiated in Core 2.

The new fuel f

assemblies (designated Batch P) are being fabricated by Combustion Engineering (CE) and are similar in design to the CE Batch N fuel provided for Cycle 9.

The Types L and M fuel, remaining from Cycles 7 and 8, were fabricated by Exxon Nuclear Corporation (ENC).

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

Small differences exist in both the mechanical and hydraulic characteristics.

The differences in the mechanical design and the hydraulic characteristics are I

discussed in Section 3.2.1 and Section 3.2.3 of (1).

The proposed operating conditions for Cycle 10 are a rated core thermal power of 2630 MWt, at a steady-state operating pressure of 2225 psia 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 to preserve the DNB margin. This assures that DNB performance is the same for all possible limiting temperature and pressure combinations. These conditions I

are consistent with the " Stretch Power" conditions proposed in (2). An increase in the allowable maximum indicated core inlet temperature from 550 F to 552 F was made in Cycle 9 (3).

This report contains sections dealing with the fuel mechanical, thermal-hydraulic, physics and safety analysis aspects of the operation of Cycle 10.

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 (4-14).

These methods have been approved by the NRC for use on i

Maine Yankee in (15-18). Methods used in safety-related analyses for the fuel mechanical design evaluations are based on the CombucLion Engineering and Exxon Nuclear generic models which have received prior approval by the NRC. I

The significant features of Cycle 10 are:

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

~ 2) An increase in fresh fuel enrichment from 3.30 to 3.50 w/o U-235 (Section 3.1).

Details of each change are provided in the sections indicated.

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2.0 OPERATING HISTORY The operating history of Maine Yankee has consisted of ten cycles designated as 1, lA, and 2 through 9.

The significant operating conditions and durations of the cycles are defined in Table 2.1.

The fuel assembly types l

loaded by cycle are given in Table 2.2.

2.1 Cycles 1 and 1A y

The initial Maine Yankee core consisted of unpressurized, low density fuel designated as Core 1 design fuel assemblies (Types A, B, and C).

Cycle 1 operation was restricted and teiminated due to leaking fuel assemblies.

Cycle 1A consisted of operation after the leaking fuel assemblies from the initial core were replaced with fresh fuel designated as Replacement Fuel (Type RF) assemblies. The mechanical design of the Type RF assemblies was essentially the same as Core 1 design fuel. The significant difference in the design was the pressurization of the fuel rod with helium sufficient to L

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 PA Cycle 2 consisted entirely of fresh assemblies designated as Core 2 design fuel (Types D, E, and F).

Mechanical design changes were made in comparison to the Core 1 design fuel. These comprised prepressurization, higher fuel density, and smaller diameter pellets. A detailed discussion of the design changes was provided in (19). The Core 2 design fuel performed successfully. Subsequent to Cycle 2 operation, burnable poison shim failures were discovered in the Type E assemblies. Corrective action consisted of replacement of all Type E shims with water-filled zircaloy rods prior to reinsertion in subsequent cycles.

I 2.3 Cycles 3 and 4 Cycle 3 consisted of fresh fuel assemblies of the Core 2 design (Types G and H) and Replacement Fuel assemblies reinserted from Cycle 1A.

The performance integrity of the Cycle 3 fuel had been demonstrated through I

irradiation in Cycles 2 and 1A, respectively.

All fuel performed successfully during Cycle 3.

Cycle 4 consisted of all fuel assemblies of the Core 2 design. Slight design changes to the fresh Type I fuel were made and discussed in Section 3.2.1 of (20). New fuel and once-burned fuel assemblies from Cycle 2 were inserted and the replacement fuel discharged. A small number of leaking fuel assemblies 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 I

fresh assemblies designed by ENC (Types J and K).

A detailed discussion of the ENC design assemblies was provided in (1). Five Core 2 design leaking assemblies returned to the core in Cycle 5 were repaired by replacement of fuel rods with fresh, low enrictanent Core 2 design fuel (34 rods) or water-filled zircaloy rods (10 rods). The fuel performed successfully during Cycles 5 and 6.

2.5 Cycles 7 and 8 I

Cycles 7 and 8 consisted almost entirely of ENC-designed fuel. One Type E assembly of the Core 2 CE design with Cycle 2 exposure was inserted in the core center location. The fresh ENC batches (Types L and M) represented en increase in enrichment to 3.30 w/o U-235.

The Cycle 7 design was the first low-leakage, low-fluence core design. Minor fuel design changes for the Types L and M fuel were discussed in (12). All fuel has performed cuccessfully during Cycles 7 and 8.

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5 2.6 Cycle 9 1

Cycle 9 consisted of fresh fuel assemblies designed by CE (Type N) and reinserted assemblies designed by ENC (Types L and M) at 3.30 w/o U-235 enrichment. Cne Type E assembly with Cycle 2 exposure was inserted in the core center location. The Cycle 9 design was a continuation of the low i

leakage, low fluence core designs. The fuel design of the Type N fuel was I

discussed in (21).

A small number of leaking fuel rods were discovered during Cycle 9 operation.

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TABLE 2.1 MAINE YANKEE OPERATING HISTORY

SUMMARY

Date of Core Power Level Cycle Power Licensed Operated Burnup

)

Cycle Escalation (MWt)

(%)

(MWD /MT) 1 11/3/72 2440 50-60(1) 10367 1A 10/12/74 2440 80(1) 4500 L

2 6/29/75 2440 100 17395

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3 6/17/77 2630(2) 93 11075 4

8/23/78 2630 97(3) 10496 r'

5 3/17/80 2630 97 10796 f

6 7/20/81 2630 97 11580 7

12/12/82 2630 100 12466 8

6/20/84 2630 100 12458 9

10/25/85 2630 100 14200(4) s F

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(1) Power decrease and primary system pressure decrease to 1800-2000 psia due to leaking fuel (3) Licensed po.ter increase from 2440 MWt/2100 psia operation to 2630 MWt/2250 psia operation (3) Power restriction due to secondary plant limitations (turbine)

(4) Estimated m

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TABLE 2 2 MAINE YANEEE FUEL ASSEMBLY TYFES BY CYCLE Assembly Number of Fuel Assemblies by Cycle Fuel Enrichment Mechanical Type (w/o U-235) Design Type 1

1A 2

3 4

5 6

7 8

9 A

2.01 CE-Core 1 69 57 B

2.40 CE-Core 1 80 24 C

2.95 CE-Core 1 68 64 l

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RF 2.33 CE-RF 70 65 RF 1.93 CE-RF 69 D

1.95 CE-Core 2 1

80 12 61 1

1 1

E 2.52 CE-Core 2 68 68 12 F

2.90 CE-Core 2 C

2.73 CE-Core 2 32 32 32 40 40 40 H

3.03 CE-Core 2 I

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

3.00 ENC K

3.00 ENC 72 72 72 L

3.30 ENC 72 72 72 72 72 M

3.30 ENC 72 N

3.30 CE-Core 2

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3.0 RELOAD CORE DESIGN 3.1 General Description I

3.1.1 Core Fuel Loading E

The core of Maine Yankee Cycle 10 consists of 217 fuel assemblies of the type and quantity detailed in Table 3.1.

The single Type E assembly is of I

the Core 2 mechanical design and was previously irradiated in Cycle 2.

Assembly Type L was introduced in Cycle 7 and was irradiated in Cycles 7, 8, and 9.

Assembly Type M was introduced in Cycle 8.

Assembly Type N was introduced in Cycle 9.

Assembly Type P is fresh fuel to be introduced in Cycle 10.

Assembly Types L and M are ENC design fuel. Assembly Types N.and P are fabricated by CE and are designated Core 2 design fuel. The Type N fuel initial enrichment of 3.30 w/o U-235 and shim loading of 23.8 milligrams of B-10 per inch of active shim rod are equal to those of the Types L and M fuel. The Type P fuel enrichment is 3.50 w/o U-235 and the shim loading is I

31.4 milligrams of B-10 per inch of active shim rod. The total number of fuel rods by assembly type for Cycle 10 is also given in Table 3.1.

The core loading by fuel type is given in Table 3.2.

3.1.2 Core Burnable Poison Loading Burnable poison shim rods are located in selected assemblies in Cycle 10.

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 Section 2.2, the single Type E assembly shim rod locations contain water-filled zircaloy rods with end plugs to restrict the flow in these rods.

All burnable poison shims are composed of B C in A1 0. The ENC 4

23 design shim irradiation integrity has been demonstrated in the Types J, K, L, and M assemblies during Cycles 5 through 9.

The CE design shim irradiation integrity has been demonstrated in the Types G and I assemblies during Cycles 3 through 6 and the Type N assemblies during Cycle 9.

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3.1.3 Core Loading Pattern The fuel assembly locations designated for Maine Yankee Cycle 10 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, M, and N assemblies in Cycle 9.

The appropriate rotation index relative to the previous assembly position in the core is also given for each assembly. The L

loading and rotations of the other quadrants are such that mirror symmetry exists with respect to the quadrant boundary lines.

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The Cycle 10 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 10 loading pattern is similar to the Cycle 7, 8, and 9 low-leakage loading patterns. The benefits of such a core design are:

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1) Reduced irradiation exposure to the reactor pressure vessel, thus

)

reducing the rate of irradiation embrittlement;

2) Extended cycle full-power lifetime due to reduced neutron leakage;

3) Preferred fuel rod power and exposure histories from fuel performance and mecnsnical integrity considerations (i.e., higher relative powers at lower burnups);

4) Improved stability to axial xenon oscillations near end-of-cycle; and

5) A less severe moderator defect with cooldown at end-of-cycle, providing greater shutdown margin for cooldown transients.

3.1.4 Assembly Exposure History The calculated exposure history of the Cycle 10 fuel assemblies at Beginning-of-Cycle (B0C) is given in Figure 3.3.

The exposures are based on an expected cycle length of 14,200 mwd /Mt for Cycle 9 and the achieved cycle _ _ _... _. _...

I length of 12.458 mwd /Mt for Cycle 8.

Table 3.2 gives BOC average exposures by fuel type. The Cycle 10 BOC average exposure for the core is approximately 15,000 MWD /MT.

I The exposure history of the assemblies utilized in the analysis is demonstrated to be accurate by comparison with incore detector data. Figure 3.4 is a comparison of predicted and actual burnup assembly data at I

approximately the middle of Cycle 9.

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 Cycle 10 is unchanged from Cycle 9.

Figure 3.5 shows the CEA group locations in the quarter core. The Bank 5 configuration consists of:

1) Nine full-strength CEAs, designated Subgroup 5A, which are I

scrammable CEAs and contribute to the available scram reactivity.

2) Four full-strength CEAs, designated Subgroup 5B, added to Bank 5 for local power distribution control. These four CEA locations are nonscrammable and do not contribute to the available scram reactivity.

I As in Cycles 7, 8, and 9, Subgroups SA and SB are independently g

moveable and not directly connected as a single CEA bank. As such, their 5

movements are administrative 1y controlled for positioning as a single CEA bank. To accommodate this movement, the physics input to the Reactor Protective System (RPS) setpoint analysis has included power distribution cases sufficient to justify differences in insertion between these two l

regulating subgroups subject to the CEA group insertion limits, as discussed in Section 4.9.1.

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I 3.2 Fuel System Design I

3.2.1 Fuel Mechanical Design I

Fuel assembly types and quantities for this core are given in Table 3.1.

The fuel assemblies, described in (21-24), have been designed to I

maintain mechanical, material, chemical, and thermal-hydraulic compatibility with all other fuel and structures in the reactor core. Table 3.3 lists the mechanical design features and vendors of all fuel batches.

The detailed fuel assembly descriptions and mechanical design criteria for the recycled reload fuel have been described in (21-26). The fresh reload fuel is provided by CE.

The fresh reload fuel, Batch P, being inserted in the Maine Yankee core I

is similar to the previously supplied reload fuel with one exception, a modification to the poison pin design. Compared to the previous CE Batch N, the overall length of the poison rods is decreased to allow for additional rod growth clearance. Also, additional free volume was provided to reduce the calculated internal poison pin pressure. The free volume was provided by replacing the two solid internal spacers with two hollow spacers.

I The Exxon supplied fuel Batches L and M will schieve exposures higher than previously encountered at Maine Yankee and have been analyzed to I

demonstrate compliance with the appropriate design criteria at these higher exposures. These analyses are documented in (26) and employ the methods described in (27) which have been reviewed by the NRC staff in (28).

The results of the extended burnup analyses for the Exxon fuel batches are summarized as follows:

I

1) Fuel Cladding Collapse I

The fuel cladding collapse is precluded during the design lifetime of the fuel. This is accomplished by limits on cladding initial I I

ovality and the use of a plenum spring which maintains the fuel pellet column in compression while the fuel pellet column is

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undergoing densification.

2) Irradiation Induced Dimensional Changes The fuel assembly length change results from two distinct mechanisms in the Zircaloy guide tubes:

a) Irradiation induced growth, and a

b) Compressive creep.

The analysis of fuel bundle growth demonstrates that there is sufficient clearance at the end of the assembly design lifetime to preclude interference of the bundle between the fuel alignment plate and the core support plate.

The fuel rod length change results from the irradiation growth of Zircaloy. The analysis of the differential growth of the fuel rods and the assembly guide tubes demonstrates that there is sufficient space (shoulder gap) at Beginning-of-Life (BOL) between the upper and lower tie plates to accommodate rod growth without interference at end of the design lifetime.

3) Cladding Strain and Fatigue Analysis The cladding experiences tenslie strain when the combination fuel pellet expansion and cladding creepdown have closed the gap between the fuel pellets and the cladding. Permanent strain results when the cladding yield strength is exceeded, or if the stress is

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maintained for a sufficient time.

i The analysis of cladding strain demonstrates that the maximum steady-state strain is well below the 1.0% design limit. The transient strain is less than the 1% circumferential strain limit. -

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L The cladding can experience stress levels during ramps which can be significts The fuel design incorporates allowance for these

(

stresses by limiting the calculated stress intensities to a value significantly below those calculated to cause fuel failure during H

experimental ramp tests with irradiated fuel.

L The analysis of the transient stresses demonstrated that the calculated ramp stresses remain below the design limit.

Fatigue damage is the result of the repeated application of cyclic stresses above the endurance limit. The endurance limit and the critical number of cycles are established by material testing.

The analysis of the cumulative usage factor is well below the design limit.

4) Maximum Fuel Rod Internal Pressure The internal pressure of the fuel rod is predominantly a function of the initial rod prepressurization level, internal rod free

(.

volume and temperature distribution, and the release of fission product gases during irradiation. The release of fission product gases is evaluated as a function of the rod operating history.

The analyses of the fuct rod internal pressure demonstrate:

a) The calculated stress intensities during the steady-state operation meet the ASME boiler and pressure vessel requirements.

b) The calculated rod internal pressure is below the applicable criteria for analyzed operating histories which are chosen to bound expected operation. t

l I

5) Fuel Rod Corrosion I

The Zircaloy cladding experiences waterside corrosion and hydrogen absorption. The combination of these effects reduces the wall thickness of the cladding and its strength. Appropriate design limits are specified to prevent the loss of cladding integrity from I

these effects.

The analysis of these effects demonstrates that the design limits are met.

I 3.2.2 Fuel Thermal Analysis I

The thermal effects analysis encompassed a study of fuel rod response cs a function of the detailed cycle exposure and power. The fuel rod types I

cnd power histories examined in detail are the maximum power rod of each fuel batch. The Batch L fuel was not analyzed explicitly because its exposure and power histories were bounded by Batch M.

The calculation methodology is the came as that employed in the previous cycle (3).

I Fuel thermal calculations were performed using the GAPEX (29) computer program. The CAPEX code calculates pellet-to-clad gap conductance from a I

combination of theoretical and empirical models which predict fuel and I

cladding thermal expansion, fission gas release, pellet swelling, pellet densification, pellet cracking, and fuel and cladding thermal conductivity.

Figure 3.6 demonstrates the effect of Linear Heat Generation Rate (LHGR) on fuel temperature at Beginning-of-Cycle (BOC) conditions. Figure 3.7 demonstrates the effect of LHGR on fuel temperature at End-of-Cycle (E0C) conditions. Table 3.4 lists UO melting temperature and centerline 2

temperature for the rods of interest at selected points in life and power levels.

I The result of the fuel performance calculations indicates that the thermal performance is similar to that reported for the previous cycle (3). I I

E The Specified Acceptable Fuel Design Limit (SAFDL) for fuel centerline melt for each fuel batch 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 batches to the core maximum radial pin power during this cycle. The fresh fuel SAFDL is bounding for all fuel batches since:

I a) The fresh fuel contains the core-wide maximum power pin throughout the cycle, and I

b) 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 Thermal-Hydraulic Design Steady-state and transient DNBR analysis of the Cycle 10 core have been I

performed using the COBRA-IIIC computer program (30), in the manner described in (4) and (5), and as described below. The models reflect the intended Cycle 10 coolant conditions and power distributions, the assembly flow distribution due to differences in hydraulic characteristics and inlet flow maldistribution, and the specific geometry of the Maine Yankee fuel assemblies.

An eighth core COBRA-IIIC model was used to determine hot assembly cnthalpy rise flow factors. This model explicitly represents each fuel I

cssembly in the one-eighth (1/8) core in the specific location it will reside for Cycle 10 operation, and accounts for the differences in hydraulic characteristics between the CE and the ENC fuel assemblies. 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 vessel reported in (31) and the FSAR (32). The assumed hot assembly enthalpy rise flow factor was 1.0 for the fresh CE fuel assemblies at all power distributions. A 0.950 enthalpy rise flow factor is applied to all ENC casemblies 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 I

in predicting DNB performance.

l l

E The potential effects of fuel rod bow on thermal-hydraulic performance has also been evaluated for Cycle 10 operation. Using the channel closure correlation in (33), the maximum channel gap closure due to fuel rod bowing for the CE fuel assembly with the highest burnup during cycle 10 a Type N assembly, was calculated to be 22.0%.

Tests performed at Columbia (34) indicate that a degradation in DNB performance is not experienced until channel closures exceed 50%. Therefore, no penalty is required for fuel rod bow considerations.

Allowances for rod pitch, bow and clad diameter variations for the ENC fuel are accommodated as follows. Allowances for manufacturing tolerances on rod pitch and clad diameter, if considered in the most adverse situation, would result in a maximum channel closure in the vicinity of 10%. Using the methodology of (35), the maximum channel gap closure due to fuel rod bowing for the ENC fuel assembly with the highest burnup during Cycle 10 is less than 32%. Therefore, a flow factor of 1.0 is justified for the ENC fuel in Cycle 10 since the channel closure resulting from rod pitch, bow and clad I

diameter considerations for any ENC fuel during Cycle 10 will be less than 50%.

Table 3.8 contains a list of the pertinent thermal-hydraulic design parameters used for both safety analysis and for generating Reactor Protection System (RPS) setpoint information. The list also includes the corresponding thermal-hydraulic parameters for Cycle 3 (36) and Cycle 9 (3) for comparison.

I I

I I I

M M

M TABLE 3.1 MAINE YANKEE CYCLE 10 i

ASSEMBLY DESCRIPTION Number Initial 4

of Fuel Initial Number of og B-10 Number of Total Total l

Assembly Exposure Rods per w/o U-235 Shim Locations per inch Assemblies Shim Fuel i

Designation History Assembly Fuel per Assembly in Shims in Core Locations Rods i

E-16 Cycle 2 160 2.52 16 29.0*

1 16*

160 L-0 Cycle 7,8,9 176 3.30 0

8 0

1,408 M-0 Cycle 8,9 176 3.30 0

23.8 8

0 1,408 1

M-4 Cycle 8,9 172 3.30 4

23.8 28 112 4.816 I

M-8 Cycle 8,9 168 3.30 8

23.8 28 224 4,704 N-0 Cycle 9 176 3.30 0

4

- 0 704 N-4 Cycle 9 172 3.30 4

23.8 24 96 4,128 O

N-8 Cycle 9 168 3.30 8

23.8 44 352 7,392 I

l P-0 Fresh 176 3.50 0

28 0

4,928 i

P-4 Fresh 172 3.50 4

31.4 20 80 3,440 i

P-8 Fresh 168 3.50 8

31.4 24 192 4,032 Core Totals 217 1,072 37,120 1

E-16 shims replaced with water-filled rods for Cycle 10.

l 1

i t

4

I TABLE 3.2 MAINE YANKEE CYCLE 10 CORE LOADING Uranium per Uranium Exposure Assembly Number of Assembly Total at BOC*,

Type Assemblies (KGU)

(KGU)

(mwd /Mt)

E-16 1

353.7 354 20,184 L-0 8

381.1 3.049 25,362 M-0 8

381.1 3,049 28,578 M-4 28 372.5 10.430 26,772 M-8 28 363.8 10,186 31,458 M-0 4

388.7 1,555 14,916 I

N-4 24 379.9 9,118 14,811 N-8 44 371.0 16,324 17,893 P-0 28 388.7 10,884 0

I P-4 20 379.9 7,598 0

P-8 24 371.0 8,904 0

81,450 14,998

Based on End-of-Cycle 9 at 14,200 MWD /MT I

I

I I

I I I

I TABLE 3.3 Mechanical Design Features of Cycle 10 Fuel I

Type E Types L and M Type N Type P Fuel Vendor C-E ENC C-E C-E Fuel Assembly Overall length 156.718*

156.718 156.718 156.718 Spacer grid size (maximum square) 8.115 8.115 8.115 8.115 Number of zircaloy grids 8

0 8

8 I

Number of inconel grids 1

0 1

1 Number of bimetallic grids 0

9 0

0 Fuel rod growth clearance 1.021 1.300 min.

1.600 1.600 Fuel Rod Active-fuel length 136.7 136.7 136.7 136.7 I

Plenum length 8.575 8.8 8.375 8.375 Clad OD 0.440 0.440 0.440 0.440 Clad ID 0.384 0.378 0.384 0.384 I

Clad wall thickness 0.028 0.031 0.028 0.028 Pellet OD 0.3765 0.370 0.3765 0.3765 Pellet length 0.450 (62) 0.450 0.450 I

Dish depth 0.023 0.008 0.021 0.021 Clad naterial Zr-4 Zr-4 Zr-4 Zr-4 Initial pellet density 95.0%

94.0%

94.75%

Initial pressure (10)

(37)

(21)

(24)

Poison Rods I

Clad OD Overall rod length 146.513 146.500 146.629 146.322 0.440 0.440 0.440 0.440 Clad ID 0.388 0.378 0.384 0.384 Clad wall thickness 0.026 0.031 0.028 0.028 Pellet OD 0.376 0.353 0.362 0.362 Clad nmterial Zr-4 Zr-4 Zr-4 Zr-4 I

All length dimensions are in inches I

C-E - Combustion Engineering ENC - Exxon Nuclear Corporation I I I

m

l I

TABLE 3.4 Maine Yankee Cycle 10 Centerline and UO9 Melt Temperature Comparison Melt Centerline Fuel Temperature LHGR Temperature Fuel Type Vendor

( F)

(kW/ft)

( F)

BOC E C-E 4910 19 4415 20 4630 21 4834 21.3*

4893 I

M.

ENC 4833 19 4435 20 4649 20.9*

4833 N

C-E 4924 19 4271 20 4485 21 4690 I

22 4885 22.2*

4923 P

C-E 5049 19 4420 I

20 4574 21 4720 22 4858 23.1*

5040 EOC E C-E 4826 19 4487 E

20 4700 W

20.6*

4824 M

ENC 4748 19 4573 I

19.8*

4743 N

C-E 4824 19 4410 I

20 4623 20.9*

4806 I

P C-E 4931 19 4264 20 4478 21 4683 22 4879 I

22.2*

4917 I

ALHCR kU/ft SAFDL Limit ENC Exxon Nuclear Corporation I

C-E Combustion Engineering I I

M M

I i

j i

TABLE 3.5 l

}

MAINE YANKEE CYCLE 10 I

RATIO OF MAXIMUM RADIAL RELATIVE PIN POWERS MAXIMUM IN TYPE E FUEL TO MAXIMUM IN CORE l

1 Rodded Condition HFP, Equilibrium conditions of ARO i

Ratio of Maximum Radial Relative Pin Powers j

Regulating BOC BOC MOC EOC Banks Inserted 50 ledd/Mt 500 mwd /Mt 6K ledd/Mt 13K mwd /Mt ARO 0.667 0.706 0.7?4 0.694 i

l Bank 5 0.446 0.474 0.487 0.487 Banks 5 + 4 0.205 0.215 0.237 0.240 0

7 Banks 5 + 4 + 3 0.212 0.226 0.252 0.249 Banks 5 + 4 + 3 + 2 0.139 0.149 0.164 0.169 I

Banks 5 + 4 + 3 + 2 + 1 0.225 0.239 0.263 0.269 l

1 i

l l

M M

j i

j TABLE 3.6 1

I I

MAINE YANKEE CYCLE 10 l

RATIO OF MAXIMUM RADIAL RELATIVE PIN POWERS MAXIMUM IN TYPE M FUEL TO MAXIMUM IN CORE Rodded Condition HFP, Equilibrium Conditions of ARO k

Ratio of Maximum Radial Relative Pin Powers

]

Regulating BOC BOC MOC EOC Banks Inserted 50 mwd /Mt 500 mwd /Mt 6K mwd /Mt 13K ledd/Mt t'

j ARO 0.705 0.725 0.718 0.710 I

Bank 5 0.657 0.700 0.705 0.688 lI b

7 Banks 5 + 4 0.573 0.588 0.638 0.672 I

Banks 5 + 4 + 3 0.586 0.604 0.650 0.655 I

Banks 5 + 4 + 3 + 2 0.496 0.501 0.529 0.579 1

Banks 5 + 4 + 3 + 2 + 1 0.564 0.572 0.596 0.626 l

l 1

l l

1 l

1 t

M M

=

M M

i i

TABLE 3.7 1

MAINE YANEEE CYCLE 10 RATIO OF MAXIMUM RADIAL RELATIVE PIN POWERS MAXIMUM IN TYPE N 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 /ut 500 mwd /Mt 6K mwd /Mt 13E mwd /Mt 1

ARO 0.872 0.876 0.843 0.806 i

Bank 5 0.880 0.886 0.892 0.834 5

1 Banks 5 + 4 0.876 0.882 0.903 0.862 i

4 Banks 5 + 4 + 3 0.869 0.876 0.885 0.825 Banks 5 + 4 + 3 + 2 0.856 0.862 0.868 0.878 Banks 5 + 4 + 3 + 2 + 1 0.798 0.802 0.808 0.808 i

I i

i i

l i

l

}

3

n n

__ O D

R fl Ul C

TABLE 3.8 Maine Yankee Cycle 10 Thermal-Hydraulic Parameters at Full Power General Characteristics Units Cycle 3 Cycle 9 Cycle 10 Total Heat Output MWT 2630 2630 2630 106 Btu /hr 8976 8976 8976 Fraction of Heat Generated in 0.975 0.975 0.975 l

Fuel Rod 4

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 554 548-556 548-556 Total Reactor Coolant Flow (design) 106 lb/hr 134.6 135.8-134.2 135.8-134.2 Coolant Flow Through Core (design) 106 lb/hr 130.7 131.9-130.3 131.9-130.15 Hydraulic Diameter (nominal channel) ft 0.044 0.044 0.044 2

2.444 2.46-2.436 2.46-2.433 Z

Average Mass Velocity 106 lb/hr-ft Pressure Drop Across Core (design flow) psi 9.7 10.18 9.99 Total Pressure Drop Across Vessel (Based on nominal dimensions and design flow) psi 32.4 32.9 32.6 Core Average Heat Flux

Btu /hr-ft2 178,742 182,184 180,211 Total Heat Transfer Area

ft2 48,978 48,038 48,309 Film Coefficient at Average Conditions Btu /hr-ft20F 5,640 5,636 5,698 Maximum Clad Surface Temperature OF 656 657 657 Average Film Temperature Difference OF 31.7 32.3 31.5 Average Linear Heat Rate of Rod

kW/ft 6.03 6.15 6.11 Average Core Enthalpy Rise Btu /lb 68.7 68.7 68.9

W M

M M

M TABLE 3.8 (Continued)

Maine Yankee Cycle 10 Thermal-Hydraulic Parameters at Full Power Cycle 3 Cycle 9 Cycle 10 Calculational Factors CE ENC CE ENC CE 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 Non-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 shrinkage due to fuel densification.

I I

FIGURE 3.1 Maine Yankee Cycle 10 Burnable Poison Shim Assembly Locations I

I g~

O O

l O

O O

O 1

g O

O O

O l

O O

O.

O 16 Water Rod Assembly (E-16)

O Shim Assembly (L'-0,M-0,N-0,P-0)

I I

x x

' X I

X X

X I

I 4 Shim Assembly (M-4,N-4,P-4) 8 Shim Assembly (M-8,N-8,P-8) b Water filled rods g

a

,,c i. m,0, sm m r.es I

FIGURE 3.2 MAINE YANKEE CYCLE 10 ASSEMBLY LOADING PATTERN Ra31 NPe E-16 62 -- Cy c l e~ 10 Location E

36 --Cycle 2 Location g

0 --Rotation Index*

M-8 3 -- Cycle 10 Location M-S 1

L-0 2

L,M.N 12 - Cycle 9 Location 24 2

1 3 --Rotation Index**

O 2

P-0 5 --Cycle 10 Location M-8 3

P-0 4 P-0 5 P-4 6 M-4 7

12 27 3

0 1

M-0 8 P-0 9 N-8 10 N-4 11 N-8 12 P-8 13 38 34 37 10 2

0 0

1 M-0 14 P-0 15 N-8 16 P-8 17 M-8 18 P-4 19 M-8 20 11 18 43 13 2

1 1

0 M-8 21 P-0 22 N-8 23 N-8 24 M-4 25 P-8 26 M-4 27 N-8 28 h'

47 39 7

4 23 53 a

1 3

0 0

3 2

3 <,;..;. s IU -, (i '

I P-0 29 N-8 30 P-8 31 H-4 32 N-8 33 N-8 34 N-4 35 N-0 36 i?

41 29 20 25 9

15 i6'. ' 'Y. (

[1. t S '.

0 0

1 3

2 2

.p.M U I

n.-

u..

' ge..f9,h

,j;j'[';'

P-0 37 N-4 38 M-8 39 P-8 40 N-8 41 P-8 42 N-4 43 M-4 44 5

51 32 46 36 M-8 45 0

3 1

2 0

yks,4 4 Oh);; f};'

r _'

^

52 O

P-4 46 N-8 47 P-4 48 9-4 49 9-4 50 N-4 51 H-4 52 p-4 53 1

,j.y.f l;'i 30 16 22 6

49 yllUg-fY L-0 54 54

.1.

^

l 2 M-4 55 P-8 56 M-8 57 9-8 58 4-0 59 M-4 60 p-4 61 E-16 62

..;;.y.L," *

),, f."..].

27 56 6

15 59 36 3

0 2

2 0

0

" !u hf;g%

3.:v:g y_

f1, k',fj.h j

Clockwise multiple of 90 relative to Cycle 2 location in quadrant.

N.jh. -

m

- Clockwise multiple of 90 relative to Cycle 9 location in quadrant.

j. -;

.b

.f. Y...... u ~<.-

I Wfy&:-

3;.-

I Figure 3.3 Maine Yankee Cycle 10 Calculated Assembly Exposures at BOC I

CYCLE AYERAGE EXPOSURE =

0.0 CORE AYERA6E EXPOSURE = 14998.0 I

MAI ASSEM8LY EIPDSURE = 32462.0 8 LOC 20 i

CORE POSITION / ASSEMBLY NUMBES. I A14 1 I A12 2I FUEL TYPE................... I M-8 I

L-0 I

CEA SANE TYPE................. I I

I I

I ASSEMBLY 80tNUP (NM0/MT)...... I 31853.0 I 25362.0 I

! 817 3 I 816 4 I 815 5 I 813 6 I til 7I i

I M-8 I

P-0 I

P-4 I

M-4 I

I I aC8 I I eIe I I

I I

j I C18 8I C16 I C15 I C13 I

I I

M-0 1

P-0 I

N-8 I

N-4 I

N-8 I

P-8 I

I I aAs I I eC8 I I a58 I I

I I

I 28578.0 I 0.0 1 18207.0 I 14049.0 1 17039.0 1 0.0 I I 019 14 I 018 15 I 017 16 I 016 17 I 015 18 I 013 19 I 011 20 I I

P-0 I

N-8 I

P-8 I

M-8 I

P-4 I

M-8 I

I I e5: I I eAs !

I e3s !

I I

I I 28578.0 I 0.0 1 18284.0 1 0.0 I 31257.0 I 0.0 I 32462.0 I I E20 21 I EIS 22 I E13 23 I E17 24 I E16 25 I E15 26 I E13 27 I F11 28 I I

I M-8 I

P-0 I

N-8 I

M-4 I

P-8 I

M-4 I

N-8 I

I I eA I

I I

I e2e I I

I I

I 30763.0 1 0.0 I 18284 0 I 16734.0 1 22975.0 1 0 0 I 26189.0 I 17898.0 I I F20 29 I Fil 30 I F18 31 I F17 32 I F16 33 I F15 34 I F13 35 I Ft1 36 I I

P-0 I

N-8 I

P-8 I

M-4 I

N-8 I

N-4 I

N-0 I

I eCe I I aAa I I

58 I I

I I

I 0.0 I 18207.0 1 0.0 I 22975.0 1 17979 0 1 18578.0 1 13997.0 1 14916.0 I I 620 37 I G19 38 I 818 39 I G17 40 I G16 41 I 815 42 I G13 43 I G11 44 I I

P-0 I

N-4 1

M-8 I

P-8 I

N-8 I

P-8 I

N-4 I

M-4 I

-I I eCa !

I 2e I I e8: I I a4e 1 I

I H21 45 I I

I I

I l

I M-B I

0.0 1 14049.0 I 31257.0 I 0.0 I 18578 0 I 0.0 1 16387.0 2 29675.0 I l

l I

I J20 46 I J19 47 I J18 48 I J17 49 I J16 50 I J15 51 I J13 52 I J11 53 I l

I 31853.0 I P-4 I

N-8 I

P-4 I

M-4 I

N-4 I

M-4 I

P-4 I

I

--~~-I a s !

I a3s !

I s8s !

I I E21 54 I I

I I

L-0 1

0.0 1 17039.0 1 0.0 I 26185.0 I 13997 0 I 16387.0 1 29700.0 1 0.0 I I

........................ ~... _

I I L20 55 I L19 56 I L18 57 I L17 58 I L16 59 I L15 60 I L13 61 I L11 62 I I 25362.0 I M-4 I

P-8 I

M-B I

N-8 I

N-O I

M-4 I

P-4 I

E-16 I

-....--I I e5: I I

I I s4: I I e5* I I

I I

i I 29700.0 I 0.0 I 32462.0 1 17898.0 1 14916.0 1 29675 0 I 0.0 1 20184 0 I t

SATCH SATCH SATCH AVERAGE NUPBER ID EXPCSURE I

1 E-16 20184.0 2

L-0 25362.0 3

M-0 28578.0 4

M-4 26771.9 5

M-8 31459.3 6

N-0 14916.0 7

N-4 14811.0 l

8 h.8 17893 4 9 P-0 00 10 P-4 00 11 P-8 CC

i:

FIGURE 35 r

i NAINE YRNKEE CYCLE 9 BURNUP DISTRIBUTION BY RSSEMBLY INCR VS PREDICIED

=

2 6001 MWD /MT CYCLE EXPOSURE i

y '. -

d "a

RSSEMBLY TYPE AND INCA LOCATION L-8 8

L-0 21 INCR ASSEMBLY EXPOSURE (MWO/MT).

* = *

31645 20882

=**

32228 20877 PREDIETED ASSEMBLY EXPOSURE (ISID/MT) a PERCENT DIFFERENCE.********=**=

1.8 0.9 g.

L-8 15 M-4 31 N-4 11 N-4 25 N-8 4

30688 17753 5944 6854 6909 4<

30322 17290 5867 6875 8909 A

-1.2

-2.6

-1.3 0.3 0.0 L-8 16 N-4 33 N-8 13 M-0 28 M-8 7

M-8 20 na MAXIMUM EXPOSURE 31770 5816 7100 19101 21641 23073 31648 5810 7048 19093 21298 23111 4;

1 OCTANT LOCATION 18 1.4

-0.1 1.7 0.0

-1.6 0.2 MCASURED 34608" N-0 34 M-4 14 L-0 30 N-8 10 L-4 24 N-8 3

PREDICTED 34454w 6447 17032 31166 7534 33005 7438 1

j

% DIFFERENCE

-0.4 6496 16698 31260 7625 32968 7585 0.8

-2.0 0.3 1.2

-0.1 20 r;

M-8 32 N-8 12 L-8 27 M-4 6

M-8 19 22308 7506 34218 20971 21994 22331 7702 33988 20945 22265

~;

0.1 1.8

-0.7

-0.1 1.2 FUCL EXPOSURE (MWD /MT)

PERCENT Q26 y

E Mmm DIrrt e 04 23 40 21

=

E-16 25864 25937 0.3 34098 7593 23012 21220 s

L-0 25924 26069 0.6

-1.0

-1.4

-0.5 4"

L-4 33479 33345

-0.4 L-8 30332 30240

-0.3 M-8 26 M-8 5

L-1218 e

L-12 34608 34454

-0.4 23349 22520 34608m G

M-0 19101 19093 0.0 23031 22594 34454u d

M-4 18976 18727

-1,3

-1.4 0.3

-0.4

=

M-8 22401 22363

-0.2 g

N-0 8447 8496 0.8 M-8 22 N-8 1

N-4 6205 8184

-0.3 22564 7139

-3 N-8 7390 7443 0.7 22746 7445 g

0.8 4.3 CORE 19191 19146

-0.2 ABSOLUTE AVERAGE 0.98 E-1617 h

)

STANDAR0 DEVIATION 1.28 25864 3

25937 g

0.3 N OI

- INCR PERCENT DIrrERENCE X 100 p m

E.

I

~

Figure 3.5 Maine Yankee Cycle 10 CEA Group Locotions Regulating Shutdown Dual CEA Groups CEA Groups I

1 2

5 (SA and SB)

C 4

B 3

A 2

1 M-8 3

P-0 4 P-0 5 P-4 6

M-4 7

C 1

M-0 8 P-0 9 N-8 10 N-4 11' N-8 12 P-8 13 A

C 5A M-0 14 P-0 15 N-8 16 P-8 17 M-8 18 P-4 19 M-8 20 5A A

3 M-8 21 P-0 22 N-B 23 N-8 24 M-4 25 P-8 26 M-4 27 N-8 28

'A 2

P-0 29 N-8 30 P-8 31 M-4 32 N-8 33 N-8 34 N-4 35 N-0 36 C

A 5B*

P-0 37 N-4 38 M-8 39 P-8 40 N-B 41 P-8 42 N-4 43 M-4 44 M-8 45 C

2 B

4 P-4 46 N-8 47 P-4 48 M-4 49 9-4 50 N-4 51 M-4 52 p-4 53 L-0 54 1

3 B

M-4 55 P-8 56 M-8 57 9-8 58 s-0 59 M-4 60 p-4 61 E-16 62 SA 4

5A l

non-scramable CEA locations (Subgroup SB) 1 I I

L

[

FIGURE 3.6 I

o_ z r w Z

l NCCCC Siiii 004+

8 T

C xe s

I 3

L 3

g in k

M O

C e a

SE d e z

g o

e p

o a-a C

E E

y M

M y

e z

Z

(

b Z

"M W 2 3

e C

I r z E

M C

C U

U H

8 J

t

\\

+

~

[

~

(J 930) 3801983dW31 3NI1831N30._

I 4.0 PHYSICS ANALYSIS 4.1 Fuel Management Maine Yankee Cycle 10 consists of irradiated and fresh fuel assemblies cs described in Section 3.1.1.

The core layout is given in Figure 3.2.

Cycle 10 is expected to attain a cycle average full power lifetime of 11,900 mwd /Mt. A low-leakage loading pattern is employed, as described in Section 3.1.3.

4.2 Core Physics Characteristics The primary nuclear characteristics of the reference physics cycles (Cycle 3. Cycle 9 and Cycle 10) are given in Table 4.1.

The Cycles 9 and 10 characteristics differ from those of Cycle 3 based on the following significant changes:

I

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 l

CEA Bank 5.

l l

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

4.3 Power Distributions I

Assembly relative power densities for Cycle 10 at Hot Full Power (HFP),

cquilibrium xenon conditions are presented for unrodded and rodded (CEA Bank 5 inserted) configurations at Beginning, Middle, and End-of-Cycle (BOC, MOC, E00). Figure 3.5 shows the locations of the CEA groups.

l 8

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m uma e cup Ime m

a u m igue m a e

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SW MAINE YANKEE CYCLE 10 EOC CENTERLINE TEMPERATURE VS LHGR 5000 E

O 4000 g

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$ 20

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

E The unrodded power distributions at BOC (500 mwd /Mt) MOC (6000 mwd /Mt) and EOC (13,000 mwd /Mt) are presented in Figures 4.1 through 4.3.

The rodded (CEA Bank 5 inserted) power distributions at BOC, MOC and E0C are presented in Figures 4.4 through 4.6.

The unrodded and rodded radial peaking is approximately 2% above the maximum Cycle 9 radial peaking.

The allowable unrodded radial peaking (with uncertainties) versus exposure for Cycle 10 is included in the plant Technical Specifications for i

the purpose of comparison to measured values. This assures peaking will not exceed 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 assemblies. The quadrant analysis presented overpredicts the slight asymmetry the full core will exhibit, providing a conservative analysis of the peaking effects.

4.4 CEA Group Reactivity Worths The CEA group configurations were shown in Figure 3.5.

The CEA group worths at HFP are presented in Table 4.2 for Cycles 3, 9, and 10.

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 also contributes to a general increase in CEA worths due to power distribution weighting effects.

In general, the regulating CEA group worths I

for Cycle 10 are increased relative to those of Cycle 9.

This is primarily duetotheplacementofmorefreshfuelinrodde4.locationsinCycle10. This I

assembly placement overcomes the tendency for reduced CEA worths resulting from increased core enrichment. CEA group reactivity worths are verified by the startup test program and the associated acceptance criteria.

4.5 Doppler Reactivity Coefficients and Defects The fuel temperature, or Doppler, components of reactivity are I

presented in Tables 4.3 and 4.4 for nominal conditions in Cycles 3, 9, and I

=10.

The total core average Doppler defect from 4000 F is given in Table 4.3 cnd the core average Doppler coefficient in Table 4.4.

The values in Cycles 3, 9, and 10 are similar. Uncertainties of 25% are conservatively applied to th) coefficient and defect values prior to transient analysis applications, cxespt as noted below.

The CEA ejection methodology utilizes the core average unrodded Doppler d:;fsets from Table 4.3 and a Doppler weighting factor technique, as described in detail in'(13). An uncertainty of 15% is applied prior to transient cnalysis application of the Doppler defects for CEA ejection.

h 4.6 Moderator Reactivity Coefficients and Defects The Moderator Temperature Coefficients (MTCs) at nominal operating Hot Full Power (HFP) and Hot Zero Power (HZP), critical boron conditions are prssented in Table 4.5 for Cycles 3, 9, and 10.

Relative to Cycle 9, the Cycle 10 MTCs at BOC are more negative, due primarily to the decreased BOC critical boron concentration resulting from less excess reactivity in the core. The end of Cycle 10 MTCs are more negative than Cycle 9's due primarily

-0 to the. increased core average exposure. An uncertainty of + 0.5 x 10 dalta rho / F is conservatively applied to calculated MTC values for use in tr nsient analysis. The startup test program demonstrates the validity of cuch values.

L The moderator density defect curve used in the LOCA analysis infers specific MTC values in the operating range which must not be exceeded. The MTC Technical Specification limits for Cycle 10 are unchanged from Cycle 9 and era consistent with the LOCA moderator density defect curve and MTC cccumptions used in the safety analysis. These limits are provided in Figure 4.8.

The startup test program demonstrates that these if=its will not b2 exceeded.

The moderator defect appropriate to the scrammed (ARI) less worst stuck CEA configuration is given in Table 4.6 for Cycles 3, 9, and 10.

This defect curve yields a conservative moderator reactivity increase versus temperature I

or density, while accounting for the effects of loss in total CEA worth and the worst stuck CEA. Starting in Cycle 6, this calculation has been performed at BOC, high soluble boron and EOC, no soluble boron conditions, since the I

moderator defect is strongly dependent upon soluble boron concentration.

The EOC moderator defect for Cycle 10 is slightly less than the Cycle 9 moderator defect. The increased placement of fresh assemblies in CEA locations in i

Cycle 10 has overcame the tendency of the moderator defects to increase due to the increased fresh fuel enrichment. 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 cooldown transients from HFP, where I

moderator redistribution ef fects are an additional reactivity ccmponent.

These uncertainties are unchanged for Cycle 10.

I 4.7 Soluble Baron and Burnable Poison Reactivity Effects The soluble boron and burnable poison shim reactivity effects are shown in Table 4.1 for Cycles 3, 9, and 10.

The critical boron concentrations for Cycle 10 at BOC are less than those of Cycle 9, due primarily to a less excess reactivity in the core. There are fewer burnable poison pins in Cycle 10, and a smaller burnable poison pin reactivity worth at BOC. The inverse boron worths for Cycles 9 and 10 reflect the different soluble boron levels and core I

characteristics.

4.8 Kinetics Parameters The total delayed neutron fractions and prompt neutron generation time for Cycles 3, 9, and 10 are presented in Table 4.1.

The values are comparable and the differences reflect the effects of core average exposure nnd power I

weighting.

Table 4.7 details the delayed neutron fractions and lifetimes by delayed neutron group for Cycles 3 and 10 at HFP, All-Rods-Out (ARO)

Kinetics parameters for HFP and HZP conditions, both unrodded and I

conditions.

rodded, are calculated for appropriate application in transient analysis cases and a 10% uncertainty is applied in a conservative manner. I I

B 4.9 Safety-Related Characteristics 4.9.1 CEA Group Insertion Limits The CEA group insertion limits are given in the Technical i

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 CEA configurations from which the consequences of dropped, ejected, or withdrawn CEAs are acceptable. The CEA group insertion limits for Cycle 10 are similar to those of Cycle 9.

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

Ij,

1) the actual operating power level, or

2) the reference power level, given in Figure 4.10, which is the power f

level normally associated with the actual operating cold leg tempera ture.

l This definition is required to assure that sufficient available scram

(

rcactivity is maintained when operation deviates from the normal cold leg t;mperatures.

l l

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 E00 are shown in Table 4.8 for Cycles 9 and 10.

HFP and HZP conditions are considered for these comparable CEA insertion cases. No credit is taken for feedback effects in these calculations. 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 the allows.ble insertions at these power levels, given by the insertion limits I

cf Section 4.9.1.

The Cycle 10 values are generally equivalent or increased relative to Cycle 9, due primarily to the increased worth of the CEA groups, capecially with greater CEA group insertion.,

I 4.9.3 CEA Drop Results 4.9.3.1 Design Analysis Results The calculated worths of the most limiting dropped CEAs for Cycles 9 cnd 10, with the resulting maximum 1-pin radial powers, are given in Tables 4.9 and 4.10 for BOC and EOC. Since Cycle 4 this analysis has utilized a local pinwise Doppler feedback methodology which was verified by a special I

ct-power CEA drop test performed during the Cycle 4 startup physics tests (38).

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 ere those considered for Cycle 10 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 4.9 and 4.10 are compared for Doppler I

feedback conditions of 80% of rated thermal power.

Detailed separate envelopes 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 less than Cycle 9's, 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 end axial peaking allowed by the symmetric offset limits to obtain total I-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-CEA Drop Restrictions I

Analyses for Cycle 7 were performed and presented in (12) to determine the required rate of power level reduction which the design analysis method in Section 4.9.3.1 would bound. Three-dimensional nodal calculations were used I

to determine the required rate of power reduction.

The results indicated that I

I 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; I

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 maximum allowable insertion level corresponding to the predrop thermal power.

Tt.;

ower reductions described above assure that proper limits are I

maintainec.or operation up to four hours post-drop. The plant Technical Specifications reflect these restrictions. Similar calculations are performed i

for Cycle 10 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 4.11 for Cycles 9 and 10.

These peaking increases are accommodated in the same manner as the design 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 4.12.

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 core.

I a

I The available scram reactivity with uncertainties at EOC is greater for Cycle 10 by 0.37% delta rho at HFP and HZP conditions relative to Cycle 9.

This is due primarily to the increased 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 other safety analyses in Section 5.

The required scram reactivity at HZP must I

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 I

other safety analyses from HZP. The available scram reactivity at HZP, from Table 4.12, must be greater than the required scram reactivity at HZP.

In addition, the required scram reactivity at HZP, when added to the additional scram reactivity provided by the CEA insertion limits versus power from Figure 4.9, must be sufficient to prevent a return-to-criticality following a steam line rupture event from any power level. It must also be I

greater than the value assumed in other safety analyses from at-power conditions.

I The steam line rupture analyses are performed from both HFP and HZP conditions, 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 and 4.6.

Other safety analyses are also performed from both HZP and HFP conditions. The CEA insertion limits versus power are designed to provide I

increased available scram reactivity proportional to the increased power level. This assures that intermediate power level conditions are covered by 8

analysis of the HZP and HFP cases.

The steam tine rupture analysis provides the minimum required worth in CEAs for cooldown events from HFP and HZP conditions to maintain suberiticality.

In addition, other safety analyses have implicitly assumed minimum required worth in CEAs, as stated in Section 5.1.4.

The minimum required worths in CEAs are compared, in Table 4.13, to the available scram I 8

I reactivity from Table 4.12.

The table demonstrates that, in each condition and time in cycle life, the available scram reactivity is greater than the required scram reactivity.

I Available scram reactivities with uncertainties are compared in the

{

table 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 I

uncertainties described in (14). Compliance with the startup test criteria on CEA worths demonstrates the available scram reactivity in Table 4.12.

As such, it also demonstrates CEA worth in excess of the required scram reactivities.

I 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 10 RCS soluble boron conditions of 1000 and 0 ppm, respectively. The I

boron concentration determines the magnitude of the moderator temperature defect 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 next section, can be expressed as a function of RCS soluble boron concentration in the Technical Specifications.

I 4.9.5 Shutdown Margin Requirements I

Shutdown norgin is defined as the sum of:

0

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

2) the reactivity associated with the withdrawn trippable CEA3 less the reactivity associated with the highest worth withdrawn trippable CEA.

I For a critical reactor, the shutdown margin must be maintained by sufficient available scram reactivity. The required and available scram I I

I reactivity comparison in Table 4.13 is the result of calculations which demonstrate adequate shutdown margin by bounding all the critical operating conditions for Cycle 10. Adequate shutdown margin exists, provided the CEA insertion limits and assumptions inherent in them are fulfilled. These assumptions are:

I

1) the available scram reactivity calculations, I

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 10 equations are unchanged from Cycles 8 or 9.

This shutdown margin representation is demonstrated, in Figure 4.12, to I

bound the required scram reactivities of Table 4.13 from both HFP and HZP conditions. Based on the discussion in Section 4.9.4, meeting the startup test criteria on CEA worths demonstrates the calculated available scram reactivity with uncertainties and thus demonstrates compliance with the required shutdown margin.

The minimum required shutdown margin is given for selected power levels in Figure 4.13 and the Technical Specifications to provide a well-defined requirement as a function of key plant parameters. This specification permits 8

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.

I I

--4 2-I

l 4.9.6 Augmentation Factors l

l The augmentation factors have been eliminated as a power spike penalty in all calculations of core power to incipient fuel centerline melt. The removal of augmentation factors is discussed in Section 4.11.2.

4.10 Pressure Vessel Fluence A program for reduction in pressure vessel fluence has been in place l

for Maine Yankee since Cycle 7 to address Pressurized Thermal Shock (PTS)

The Cycles 7 through 10 core designs have been a progression of concerns.

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 flats. 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 O to 10 degree region is the high fluence area.

The program for fluence reduction has been detailed in (39) and (40),

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, lA, and 2 through 5.

I The fluence reductions, expressed as flux reduction factors relative to the Cycles 1 through 6 fluence history, are shown in Table 4.14.

The inverse of the flux reduction factor is the fraction by which the flux is reduced I

relative to the fluence history Cycles 1 through 6. The target fluence reductions in (39) for Cycles 7 through 10 are compared to the actual core design fluence reductions obtained by a view-factor weighting technique of the average quarter-assembly powers calculated for the cycles. The result is that the cumulative fluence reduction factor target to end-of-Cycle 10 has been achieved for both the 0 and 10 degree azimuthal angles. At the critical I

longitudinal weld at 10 degrees, the target cumulative fluence reduction factor is 1.14 relative to the case in which no fluence reduction measures were instituted. This is achieved by a Cycle 10 flux reduction factor of _

i B

1.73 relative to average flux in Cycles 1 through 6.

Similar flux reduction factors are expected for future cycles to meet the targets set forth in (39).

4.11 Methodology and Methodology Revisions l

4.11.1 Summary of Physics Methodology Documentation A summary of the reference report and supplemental documentation for l

the application of physics methodology to Maine Yankee since Cycle 3 is given i

in Table 4.15.

The reference physics methodology report is YAEC-1115 (9).

There is one change to the reactor physics methodology included for Cycle 10.

The methodology change is a removal of augmentation factors as a power spike penalty. No new calculational methods are proposed.

1 4.11.2 Removal of Augmentation Factors l

The use of axially dependant flux augmentation factors is no longer appropriate. These factors account for an effect that is no longer present in modern design PWR fuel rods used in the Maine Yankee Atomic Power Station, as I

described in (41). Therefore, these factors have been removed from the presently approved operating limit determination calculations of kW/ft limit margin to fuel cetterline melt limit. Removal of these factors has been approved by the NRC in (42).

E I

I I

I I l

TLB'sE 4.1 MIt:E YANKEE CYCLES 3. 9 and 10 NUCLEAR CHARACTERISTICS Core Characteristics Exposure (WWD/NT)

I Core Average at BOC 7,000 12,900 15,000 Cycle Length at Full Power 10,200 13,400 11,900 Reactivity Coefficients - ARO I

Moderator Temperature Coefficient (10-4 delta rho /0F)

-0.34*

-0.41

-0.63 HFP,BOC

-1.98*

-2.31

-2.46 i

HFP.EOC Fuel Temperature Coefficient (10-5 delta rho / F) 0 I

HZP.BOC

-1.70

-1.62

-1.63 HFP,BOC

-1.30

-1.25

-1.26 HZP.EOC

-1.80

-1.76

-1.78 HFP.EOC

-1.37

-1.36

-1.38 Kinetics Parameters - ARO Total Delayed Neutron Fraction (B,gg) 8 HFP.BOC 0.00611 0.00618 0.00618 HFP.EOC 0.00517 0.00511 0.00516 Prompt Neutron Generation Time (10-6,,e)

HFP.BOC 29.3 25.4 25.2 HFP.EOC 32.3 30.2 29.8 Control Characteristics Control Elements Assemblies Number Full /Part Length 77/8 81/0**

81/0**

8 Total CEA Scrammable Worth (% delta rho)

HFP,BOC 9.18 8.73 9. 5'6 HFP.EOC 9.56 9.55 10.55 Burnable Poison Rods 756 1264 1072 Number B C-A1 023 4

Total Worth at HFP BOC (% delta rho) 1.4 2.5 1.7 I

1 Critical Solublo Boron at BOC ARO (ppm)

HZP.NoIe,PkSm 1,075 1.355 1,345 HFP NoIe.PkSm 995 1,261 1,243 i

I HFP. Equilibrium Xe 782 1,014 993 Inverse Boron Worths (ppm /% delta rho) 5 HZP BOC 84 101 102

]

HFP.BOC 89 107 107 1

HZP.EOC 74 82 83 HFP.EOC 79 88 89 Conditions of 2440 MWL/2100 psia operation

- Four full-length CEAs are non-scrammable in Cycles 9 and 10,

5 TABLE 4.2 MAINE YANKEE CYCLES 3. 9 AND 10 CEA CROUP WORTHS AT HFP Worths (% delta rho)

I Cycle 3 Cycle 9 Cycle 10 BOC EOC BOC EOC BOC EOC Shutdown CEA Croups Banks C + B + A 5.86 6.02 5.16 6.13 6.06 6.59 Regulating CEA Croups Bank 5*

0.55 0.64 1.47 1.56 1.46 1.59 Banks 5 + 4 0.90 0.97 1.84 1.88 1.79 1.97 Banks 5 + 4 + 3 1,74 1.90 2.81 2.83 2.81 3.12 Banks 5 + 4 + 3 + 2 2.49 2.73 3.56 3.55 3.52 4.01 Banks 5 + 4 + 3 + 2 + 1 3.32 3.54 4.51 4.69 4.44 5.02 All CEA Croups Banks 5 + 4 + 3 + 2 + 1 +

C+B+A 9.18 9.56 9.67 10.82 10.50 11.(1 I

I I

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

' I

<L TABLE 4.3 MAINE YANKEE CYCLES 3. 9. 10 GORE AVERACE DOPPLER DEFECT a

Doppler Defect. (x 10-4 delta cho)

Fuel Cycle 3*

Cycle 9 Cycle 10 Resonance i

Temperature F

39C EOC RQC_

E goc EOC 4000 0

0 0

0 1

3750 19.4 20.9 3500 39.4 42.5 40.8 44.6 41.1 45.3 3250 59.9 64.8 3000 81.2 87.8 83.8 91.9 84.8 93.2 2750 103.1 111.6 2500 125.9 136.2 129.9 142.4 131.4 144.5 2250 149.7 161.9 2000 174.5 188.6 179.8 197.1 181.9 200.0 1750 200.5 216.7 1500 228.1 246.5 234.8 257.2 237.5 261.1 267.1 292.4 270.1 296.8 1232 1000 288.7 311.9 296.9 325.1 300.4 329.9 324.5 355.4 328.4 360.5 800 532 358.5*

387.4*

365.4 399.8 369.4 405.4 300 394.3 426.0 405.1 443.0 409.5 449.1 200 423.9 463.6 428.6 470.0 444.2 485.9 449.1 492.4 100 466.4 510.2 471.5 516.8 0

at 5250F --

i

=

2 TABLE 4.4 s

5 MAINE YANKEE CYCLES 3. 9,10 CORE AVERAGE DOPPLER COEFFICIENT ii Doppler coefficient (x 10-4 delta rho per OF) n Fuel Cycle 3*

Cycle 9 Cycle 10 e

Resonance Temperature b

F E

g g

EOC 0.2126 0.2330 0.2142 0.2340 100 0.1960 0.2145 0.1980 0.2164 200 0.1835 0.1996 0.1850 0.2020 300 0.1729 0.1881 0.1744 0.1902 400 532 0.159*

0.172*

0.1618 0.1764 0.1630 0.1777

.d 800 0.144**

0.156**

0.1437 0.1565 0.1449 0.1587

?

1000 0.131 0.141 0.1335 0.1458 0.1352 0.1480 1232 0.121***

0.121***

0.1251 0.1364 0.1263 0.1377 m, y[ g g ryRy f ;

'p,g]g)?

1500 0.114 0.123 0.1164 0.1272 0.1176 0.1290 Z

2000 0.102 0.110 0.1049 0.1148 0.1061 0.1166 2500 0.093 0.101 0.0960 0.1052 0.0972 0.1068 Sy ', f 3000 0.086 0.094 0.0891 0.0977 0.0903 0.0992

['(' ' '

3:0 4.,,[

3 3500 0.081 0.088 0.0838 0.0919 0.0847 0.0932

,_ e. ',4, h4 f.U

, J.70 $. :,'

{g%$

M %. :l-w,. :,..

f,4x,[. f

?

.?t%p7 w.'

ijk J

=

at 5250F

- at 7500F

- - at 12500F 5

AM

- di E

I TABLE 4.5 MAINE YANYEE CYCLES 3. 9 IJfD 10 MODERATOR TEMPERATURE COEFFICIENTS

)

l I

Conditions: HFP and HZP, ARO, Critical Boron Concentrations MTC (10-4 delta rho /0F)

Cycle 3*

Cycle 9 Cycle 10 Case Conditions E

E BOC EOC E

EOC HFP, EqIe, Eqsm

-0.47

-2.24

-0.41

-2.31

-0.63

-2.46 HZP, NoIe, PkSm

+0.24

+0.41

+0.23 HZP, NoIe, Eqsm

-1.40

-1.20

-1.28 I

I I

I

!I I

Cycle 3 HZP values at 5250F, Cycles 9 and 10 HZP values at 5320F I 1

I TABLE 4.6 MAINE YANKEE CYCLES 3. 9. 10 ARI MODERATOR DEFECT WITH WORST STUCK CEA I

Moderator Defect (x 10-4 delta rho)**

I Moderator Cycle 3*

Cycle 9 Cycle 10 Temperature

( F)

BOC EOC joC EOC BOC EOC I

O ppm 1015 ppm 0 ppm 1000 ppm 0 ppm 576.4

-138.0

-47.3

-125.6

-54.0

-130.0 532 - (525)*

0.0 0.0 0.0 0.0 0.0 500 43.5 23.0 72.0 23.0 60.0 450 111.0 42.0 133.0 35.0 128.0 400 167.3 45.0 183.0 35.0 180.0 217.0 40.0 224.0 35.0 213.0 350 261.6 33.3 254.6 35.0 238.0 300 298.5 25.0 288.0 35.0 260.0 250 19.0 310.0 35.0 280.0 200 150 14.0 325.0 35.0 295.0 9.0 333.0 35.0 306.0 100 5.8 335.3 35.0 312.0 68 l

I I

Cycle 3 values referenced to 0 at 5250F

- Moderator defect. at a constant 2250 psia for the specified temperatures I

~

I TABLE 4_.7 MAINE YANKEE CYCLES 3 and 10 KINETICS PARAMETERS Conditions: HFP, ARD, Critical Boron Delayed Cycle 3 Cycle 10 Effective Lifetige Lifetip)

Time in Neutron Effective (See Fraction (See )

Cycle Life Group Fraction BOC 1

0.00018 0.0126 0.00018 0.0126 2

0.00128 0.0305 0.00130 0.0305 3

0.00116 0.1163 0.00118 0.1172 4

0.00237 0.3116 0.00240 0.3142 3

5 0.00083 1.1652 0.00083 1.1776 6

0.00029 3.0253 0.00029 3.0212 TOT /.L 0.00611 0.00618 I

EOC 1

0.00014 0.0126 0.00014 0.0127 2

0.00111 0.0304 0.00111 0.0304 3

0.00097 0.1193 0.00098 0.1200 4

0.00197 0.3185 0.00197 0.3209 5

0.00072 1.1833 0.00072 1.19G5

~ " ' '

~

I TOTAL 0.00517 0.00516 I

I I

I

-s1-g S

I

-W W

W W

W W T M

M M

M R

TABLE 4.8 MAINE YANKEE CYCLES 9 AND 10 CEA EJECTION RESULTS FROM FULL INSERTIONS Cycle 9 Cycle 10 Maximum I-Pin Radial Peak BOC EOC BOC EOC HFP Bank 5 In 3.95 4.17 3.81 4.12 Ejected 5 (INCA Location 20)

HZP Banks 5 + 4 In 5.80 5.82 5.63 5.78 Ejected 5 (INCA Location 20)

HZP Banks 5 + 4 + 3 In 6.85 5.52 7.88 6.45 Ejected 5 (INCA Location 34) b Y

Maximum Elected Worth (% delta rho) 0.340 0.385 0.307 0.402 HFP Bank 5 In Ejected 5 (INCA Location 20)

HZP Banks 5 + 4 In 0.499 0.528 0.432 0.573 Ejected 5 (INCA Location 20)

HZP Banks 5 + 4 + 3 In 0.507 0.437 0.613 0.535 Ejected 5 (INCA Location 34)

I TABLE 4.9

]ULINE YANKEE CYCLES 9 AND 10 CEA DROP RESULTS AT BOC CEA Croup Dropped Dropped CEA Worth Maxinaam 1-Pin I

Positions CEA

(% delta rho)

Radial Power

Before Drop Type Cycle 9 Cycle 10 Cycle 9 Cycle 10 ARO A

0.107 0.113 1.73 1.77 ARO B

0.169 0.167 1.77 1.79 l

ARO C

0.095 0.100 1.71 1.74 ARO 1

0.066 0.062 1.64 1.65 Bank 5 in A

0.101 0.106 1.83 1.92 Bank 5 in B

O.196 0.178 1.91 1.99 Bank 5 in C

0.100 0.103 1.83 1.91 Bank 5 in 1

0.062 0.058 1.74 1.80 I

I I

Pre-Drop Maximum 1-Pin radial powers:

ARO Bank 5 In Cycle 9 1.511 1.620 Cycle 10 1.530 1.678 Post-Drop Maximum 1-Pin radial power at 80% of 2630 MWt power level conditions. I E

q w-

--ww-n-

TABLE 4.10 MAINE YANKEE CYCLES 9 AND 10 CEA DROP RESULTS AT EOC CEA Croup Dropped Dropped CEA Worth Maximum 1-Pin Positions CEA

(% delta rho)

Radial Power

Before Drop Type Cycle 9 Cycle 10 Cycle 9 Cycle 10-

~

ARO A

0.118 0.118 1.67 1.74 ARO B

0.174 0.180 1.66 1.75 lm ARO C

0.104 0.100 1.64 1.71 m

ARO 1

0.075 0.067 1.58 1.64 L_,

B:nk 5 In A

0.116 0.114 1.82 1.87 Bank 5 In B

0.177 0.195 1.86 1.83 Eank 5 In C

0.114 0.104 1.82 1.85 Bank 5 In 1

0.073 0.061 1.72 1.77 I

I I

Pre-Drop Maximum 1-Pin radial powers:

ARO Bank 5 In Cycle 9 1.459 1.589 I

Cycle 10 1.522 1.675 Post-Drop Maximum 1-Pin radial power at 80% of 2630 MWt power level conditions.

_S.

I

I TABLE 4.11 MAINE YANKEE CYCLES 9 AND 10 DROPPED CEA WITH POWER LEVEL RESTRICTION MOST LIMITING PEAKING CASES CEA Drop Time Maximum Power Percent Increase in Maximum I

From Power Post-Drop Level Perinitted 1-Pin Peaking Level (%)

(Hrs)

(%)

Cycle 9 Cycle 10 100 0.5 100 14.58 13.64 I

1.0 90 18.91 17.08 2.0 80 25.83 22.46 3.0 80 28.96 24.50 I

4.0 80 30.19 26.07 i

90 0.5 90 14.52 14.39 I

1.0 80 18.00 18.12 2.0 70 24.80 24.02 3.0 70 28.46 27.05 4.0 70 31.66 28.98 80 0.5 80 14.75 13.92 1.0 70 18.96 16.70 l

2.0 60 2 39 20.75 3.0 60 31.23 21.52 4.0 60 35.68 21.81 70 0.5 70 15.42 15.74 1.0 60 20.26 19.75 2.0 50 29.84 26.31 5

3.0 50 36.17 29.71 4.0 50 42.49 33.12 I

60 0.5 60 16.25 17.12 1.0 50 21.52 22.59 2.0 40 32.00 30.61 3.0 40 38.74 34.99 I

4.0 40 45.48 39.17 I

I I

~

I I 1

I LABLE 4.12 MAINE YANKEE CYCLE 10 AVAILABLE SCRAM REACTIVITY Worths (% delta rho)

BOC EOC HFP HZP HFP HZP Scrammable CEA Worth

9.56 9.13 10.55 10.06 Stuck CEA Worth 1.88 1.84 2.32 2.28 PDIL CEA Worth **

0.10 1.85 0.16 2.45

)

CEA Programming Allowance 0.06 0.05 0.11 0.20 Available Scraa CEA Worth

- Nominal 7.52 5.39 7.96 5.13

- With Uncertainties ***

6.77 4.85 7.16 4.62 I

I I

I ARI CEA worth less non-scrammable CEA worth (four subgroup SB CEAs)

- PDIL CEA insertion limit for HFP is 10% of Group 5 inserted PDIL CEA insertion limit for HZP is 100% of Group 5, 90% of Group 4, and 30% of Group 3 inserted

- - Uncertainty factor of 0.9 I I

.. f=..

I TABLE 4.13 I

MAINE YANKEE CYCLE 10 REQUIRED SCRAM REACTIVITY Worth (% delta rho) for-I Time in Cycle Life and RCS Soluble Boron Concentration I

BOC EOC 1000 ppm O_ ppm EFP BZP H,E HZP Available Scram 6.77 4.85 7.16 4.62 Reactivity with I

Uncertainties.

(Table 4.12)

Minimum Required I

Worth in CEAs Assumed -

- Steam Line Rupture 2.77 1.31 5.88 3.25 I

Event *

(Section 5.5.1)

- Safety Analyses 5.70 3.20 5.70 3.20 I

(Section 5)

Required Scram 5.70 3.20 5.88 3.25 Reactivity **

Excess from Required 1.07 1.65 1.28 1.37 I

to Available Scram Reactivity I

An uncertainty factor of 0.9 is applied to the nominal minimum required I

worth ir. CEAs for the steam line rupture event from Table 5.7 for comparison to the available scram reactivity with uncertainties. This uncertainty component is statistically combined with the other uncertainty I

components to derive the nominal minimum required worth in CEAs, as dircussed in (14).

- !Lx mum of either the r,!nirru:a re ;uired worth in CF1.:.,rtt::d for tl.c stcam 8

I line ruptt.re event, or other safety ane. lyses in Section 5 I

I TABLE 4.14 MAINE YANKEE CYCLES 6-10 RELATIVE PRESSURE VESSEL FLUENCE COMPARISONS I

Flux Reduction Factors ** at Azimuthal g

i g

Angle from Perpendicular to Core Shroud Flats Total Effective Full-Power Years (EFPY)

00 =-

100 Cycles to EOC*

Target Designed Target Designed I

1-6 6.51 1.00 1.00 1.00 1.00 7

7.56 1.02 1.05 1.28 1.21 8

8.60 1.35 1.42 1.51 1.43 9

9.79 1.35 1.55 1.51 1.56 10 10.93 1.35 2.03 1.51 1.73 Future Cycles 1.35 1.51 Cycle 1-10 Average *** 10.93 1.09 1.14 1.14 1.15 I

Based on 2,630 MWt full power operation.

- Inverse of fractional flux relative to Cycles 1-6.

I Inverse of EFPY - weighted fractional fluxes.

I I

I I I

TABLE 4.15 Maine Yankee Physics Methodology Documentation

-l Supporting Application

{-.

Description of Methodology

-Documentation Reference in Cycle Re ctor Physics Methods -'

YAEC-1115 9

3

{

Reference Report

,R= ctor Protective System Setpoint YAEC-1110 4

3 Analysis - Reference Report

[

Extcasion of Fine Mesh Diffusion PC No. 64, 20 4

(-

Thsory and Noda1' Physics Methods to Section 4.8 Ructivity Parameter Calculations WMY 78-102, 38

{

cnd a Change in the Nodal Neutronic Attachment B C:upling Model

[

Introduction of. Local Pointwise PC No. 64, 20 4

Doppler Feedback Effects in Section 4.8 Two-Dimensional Pinwise Diffusion WMY 78-102, 38 Thtory Calculations for Dropped CEAs Attachment C h

cnd Special CEA Drop Test at 50%

P wer for Method Verification Uncertainty Applied to Moderator YAEC-1259, 43 6

Racetivity Defect from Hot Zero Power Section 4.7 Rsduced from 25 to 15%

g

[.

~

TABLE 4.15 (Continued)

Maine Yankee Physics Methodology Documentation

~

Supporting Application L

Description of Methodology Documentation Reference in Cycle c

Doppler Defects for CEA Ejections YAEC-1324, 12 7

Calculated with Explicit Pre-Ejected Section 4.10 Local Power Weighting and Uncertainty

(

R duced from 25 to 15%

CEA Ejections Calculated From YAEC-1479, 3

9 P;rtial CEA Insertions Section 4.11 Moderator Density Defect for YAEC-1479, 3

9 LOCA Analysis Calculated Using Section 4.11 Fine Mesh Diffusion Theory

[

r E

Figure 4.1 l

Maine Yankee Cycle 10 l

Assembly Relative Power Densities BGC (500 MWD /MT), HFP, ARO CORE

SUMMARY

OF MAXIMUM Ft;EL ASSEMBLY POWER l

DESCt!PTION MAX.VALUE ASSEMBLY ASSEMBLY AUG.

l.33401 42 MAX. FUEL 800 1.52952 5

MAX. CHANNEL 1.48322 61 I

CORE POSIT 10N/ASSEM8LY NUMBER I A14 1 I A12 2I FUEL TYPE................... I M-9 I L-0 I CEA SANK TYPE................ I I

I ASSEMBLY AVERAGE POWEt........ I.30781 I 34375 I I

MAXIMUM FUEL 800 POWER...... I.56335 I 60817 I MAXIMUM CHANNEL POWER

...I 54547 I 58385 I I $17 3 I 816 4 I $15 5 I 813 6 I $11 7I I

M-8 I P-0 I P-0 I P-4 I M-4 I I

I eC8 I I e e I I

I 38369 I 92910 1 1.14723 I 1.07902 I 70693 I I 73097 I 1 42029 1 1.52952 I 1.47655 I 86729 I I 71166 1 1.34894 1 1.45531 I 1 40281 I 56222 I I

I C18 9 I C17 9 I C16 10 I C15 11 I Cl3 12 I C11 13 I I

M-0 I P-0 I N-8 I N-4 I N-8 I P-8 I I

I aAa I I eC8 I I e5 I

I.44823 I 1 13420 1 1.10625 I 1.18671 1 1.16213 I 1.26298 I I

I 88775 I 1.51345 I 1.22245 I 1.34027 I 1.26799 I b48329 I I 35999 I 1.43701 I 1.16269 I 1.30186 I 1 19%5 I l.40719 I I D19 14 I 018 15 I D17 16 I D16 17 1 D15 18 I D13 19 I D11 20 I I

M-0 I P-0 I N-8 I P-8 I M-8 I P-4 I M-8 I I

I I

5e !

I eAe !

I 3a 1 I

I.44765 I 1.15980 I 1.13289 I 1.31386 I 94035 I 1.33275 I 91079 I I 88628 I 1.50799 1 1.23320 I 1.52259 I 98103 1 1 52622 I 9527C I I 35061 1 1.43039 I 1.17033 I 1.44265 I 96440 I 1.45212 I 94411 I I

I E20 21 I Ett 22 I E18 23 I E17 24 I E16 25 I E15 26 I E13 27 I Ett 28 I I

M-8 I P-0 I N-8 I N-8 I M-4 I P-8 I M-4 I N-8 I I

I 8Aa I I

I I a2s I I

I I 38294 I 1.13165 I 1.12903 1 1.13053 1 1.02049 I 1.28511 I 96414 1 1.04382 I I

I.72955 I 1.50930 I 1.22865 I 1.24865 I 1.11040 I l.47004 I 1.03484 I 1.13311 I I.71026 I 1.43300 1 1.16630 I 1.19764 I 1.08814 I 1.39574 1 1.02711 I 1.07600 I I F20 21 I F19 30 I F18 31 I F17 32 I F16 33 I F15 34 I F13 35 I Fil 36 I I

P-0 I N-8 I P-8 I M-4 I N-8 I N-8 I N-4 I N-0 I I 8Ce 3 I eAe 3 I a58 I I

I I

I 92749 I 1 10395 I 1.30920 I 1.01764 I 1.10650 I 1.13867 I 1.14342 1 1 08279 I I 1.41783 1 1.21945 1 1.51752 1 1.10487 I 1.20618 I 1.23306 I 1.27886 I 1.18882 I I 1.34658 1 1.15966 I 1.43771 I 1.08184 I 1.14539 1 1.17259 I 1.21631 I 1.15318 I I

I 820 37 I 819 38 I 818 39 1 817 40 I 816 41 ! 815 42 I 813 43 I C11 44 I I

P-0 I N-4 I M-8 I P-8 I N-8 I P-8 I N-4 I M-4 I

I I aCa I I a28 I I a8e I I e4a I I H21 45 1 1.14553 I 1.18472 I 93872 I 1.28389 I 1.13904 I 1.33401 1 1.10161 I 90069 I I

M-8 1 1.52721 1 1.33797 I 97977 I t.46919 1 1.23367 1 1.52594 1 1.21540 I 94847 I I

I 1.45311 1 1.29959 I 96308 I l.39506 I l.17313 I 1.44577 I 1.15177 I 93229 I I 30773 I--------

I 56293 I J20 46 I J19 47 I J18 48 I J17 49 I J16 50 I J15 511 J13 52 I J11 53 I I

I 54416 I P-4 I N-8 I P-4 I M-4 I N-4 I N-4 I M-4 I P-4 1

I eie !

I a3e I I

I I e8s !

I I E21 54 1 1.07822 I l.16130 1 1.33143 I 96397 1 1.14424 1 1.104C6 I 92131 1 1.30638 I I

L-0 1 1.47438 I 1.26638 1 1.52457 I 1.03502 I 1.28040 1 1.21717 I 19640 I 1.52422 I I

I 1.40076 I 1.19809 I 1.45131 I 1.02734 I 1.21818 I 1.16140 I 98853 1 1.48137 I I

I 34320I------------------------------~~----------------------------------------

I.f0766 I L20 55 I L19 56 I L18 57 I L17 58 I L16 59 ! Ll5 60 I L13 61 ! L11 62 I I 58317 I M-4 I P-8 I M-8 I N-8 I N-0 1 M-4 I P-4 1 E-16 I

! 85 1

I I

I e4 I

I e5e !

I

!.70591 1 1.26143 I 91008 I 1.04406 1 1.08419 I 90414 I l.31071 I 99994 I I 86611 1 1.48159 I 95182 1 1.13465 1 1.19076 I 95296 1 1.52632 1 1 07945 I I 86104 ! 1.40635 I 94312 I 1.07653 1 1.15493 I 93677 I 1.48322 1 1.04864 I I

Figure 4. 2 Maine Yankee cycle 10 I'

Assembly Relative Power Densities MOC (6000 MWD /MT), HFPe ARO C08E

SUMMARY

OF MAXIMUM FUEL ASSEM8LY PDuEt I

DESCRIPI!ON MAX.YALUE ASSEMBLY ASSEMBLY AVS.

1.37200 42 MAX. FUEL ROD l.51637 42 MAX. CHANNEL 1.47960 61 I

CORE POSITION /ASSEMILY NUMBER. I A14 1 I A12 2I FUEL TYPE........

..........I M-B I L-0 I CE A BANI TYPE................ I I

I ASSEMBLY AVERAGE POWEt....... I 33114 I 30505 I MAXIMUM FUEL 900 POWER........ I 58190 I 63779 I MAXIMUM CHANNEL POWER........ I 56748 I 62195 1 1 It? 3 I Il6 4 I IIS 5 I 813 6 1 311 71 1

M-8 I P-0 I P-0 I P-4 I M-4 I I

I eC8 I I a1e !

I I.38623 I 89167 I 1.89324 I 1.10438 I 75161 I I.70294 I 1.31692 I 1.43229 I 1.42091 I 89432 I I

I 68842 I 1.26837 I 1 37985 1 1.37087 I 89082 I I C18 8 I C17 9 I C16 10 I C15 11 I C13 12 I C11 13 I I

M-0 I P-0 I N8 I N-4 I N-I I P-8 I I

I 8Aa !

I age !

I a58 I I.44397 I 1.07847 I 1.05651 I 1 14745 I 1.15435 I 1.31795 I I 83188 I 1.40734 I 1.16386 I 1.27705 I 1.25189 I 1 48871 I I 81165 I 1.35379 I 1.11742 I 1.22305 I l.20087 1 1.43179 I I

I D19 14 I D18 15 I D17 16 I D16 17 1 D15 18 I 013 19 I D11 20 I I

M-0 I P-0 I N-8 I P-8 I 5-8 I P-4 I M-B I I

I a58 I I aA I a3e !

I I.44373 1 1.09494 I 1.08109 I 1.32272 I 95229 I 1.25144 I 93565 I I 83117 1 1.39312 1 1.17073 I 1.45882 I 99132 1 1.50779 I.97347 I I

I 81100 1 1.33892 1 1.12675 I l.40464 I 97381 I 1.45255 I 96202 I I E20 21 I EIS 22 1 E18 23 I E17 24 I E16 25 I E15 26 I E13 27 I [11 28 I I

M-8 I P-0 I N-8 I N-B I M-4 I P-8 I M-4 I N-8 I I

I eA: I I

I I e2a I I

I

! 38609 1 1.07755 1 1.07843 I l.10747 1 1.02259 I 1.33198 I.98096 I 1.06041 I l

I 70270 I 1.40558 I 1.16593 1 1.19967 I 1 08886 I 1.47077 I 1.04214 1 1.12913 I I

I 68818 I 1.35117 I 1 12239 I 1.15486 1 1.06472 1 1.41516 3 1.03745 I 1.08721 I l

.. ~.. -

l I F20 29 I FIS 30 I F18 31 1 F17 32 I F16 33 I FIS 34 I F13 35 I Fil 36 I I

P-0 1 N-8 I P-8 I M-4 I N-8 I N-8 I N-4 I N-0 I I ece I I eAs I I e5e I I

I I

I 89168 I 1.05806 I 1.31992 I 1.02051 I 1.11384 1 1.14443 I l.14697 I 1.07903 I i

I 1 31700 I 1 16309 I 1.45640 1 1 08667 I 1.1 % 09 I 1.21735 I 1.27755 I 1.16564 I I 1.26839 I 1.11675 I 1.40266 1 1.06238 I 1.15012 I 1.17594 I 1.21779 I l.13847 I I G20 37 I 819 38 I Gl8 39 I G17 40 I G16 411815 42 I G13 43 I G11 44 I I

P-0 I N-4 I M-8 I P-9 I N-B I P-8 I N-4 I M-4 I

--I I 8C8 I I e2: I I ege !

I e4: I I H21 45 I 1.09355 1 1 14740 I 95184 1 1.33160 1 1.14488 I 1.37200 1 1.10760 I.91749 I I

M-8 1 1.43264 I 1 27706 I 99117 1 1.47060 1 1.21787 I 1.51637 I 1.19927 I.95897 I I

I 1.38017 1 1 22303 I 97367 1 1.41521 1 1.17645 I 1.4 % 84 1 1.15769 I 94765 I I 3 316 8 I -------------------

I 58130 I J20 4i I Jll 47 I J18 48 I J17 49 I J16 50 I J15 51 I J13 52 I J11 53 I I 56709 I P-4 I N-8 I P-4 I M-4 I N-4 I N-4 I M-4 I P-4 I e a I I a3: 1 I

I I ega I I

I E21 54 1 1.10461 1 1.15448 I 1 35163 I 98130 1 1.14758 1 1.10939 I 94146 ! 1.32932 I I

I L-0 I 1.42125 1 1.25208 I 1.50812 1 1.04260 I 1.27843 I 1.20046 I 1.00331 1 1.50902 I I

I 1.37118 1 1.20104 1 1.45285 I 1.03794 1 1.21870 1 1.15881 I 99487 I 1.47827 I I 3 8 510 1 --------------------------- ----- -- -

! 63816 I L20 55 I L19 56 I L18 57 I L17 58 I L16 59 I L15 60 I L13 61 I L11 62 I I 62203 I M-4 I P-8 I M-8 I N-8 I N-0 1 M-4 I P-4 I E-16 I I

..~.-...I I a58 I I

I I 84 I

I 8 5: I I.75175 I 1 31823 I 93591 1 1.06109 1 1.00006 I 92003 1 1 33260 1 1 02009 I I 89451 I 1 48907 I 97371 I 1 13005 I 1 16688 1 % 223 1 1 51050 1 1 062 % I I 89101 ! 1 43212 1 %224 I l CE8C21 1.13%2 I 95c% I 1479601 1.05992 I l

=...

\\,

Figura 4. 3 Maine Yankee Cycle 10 Asseinbly Relative Power Densities EOC (13000 MWD /MT), HFP, ARO CORE SUMMAlf 0F max! MUM FUEL ASSEMBLY POWER DESCRIPTION MX.VALUE ASSEMILY I

ASSEM8LY AYG.

1.39701 56 MAX. FUEL 800 1.52188 56 MAX. CHAuNEL 1.44674 56 CORE POSITION / ASSEMBLY NUMBER I A14 1 I A12 2I FUEL TYPE................. I M-B I L-0 I CEA SANI TYPE I

I I

ASSEMBLY AYERAGE POWER...... ! 36938 I.44689 I

-I MAXIMUM FUEL B00 POWER....... I 62683 I 69898 I MAXIMUM CHANNEL POWEt........ ! 61476 I 68522 I I B17 3 I 816 4 I 815 5 I 813 6 I B11 7I I

I M-8 I P-0 I P0 I P-4 I M-4 I I

I 8Ce !

I ela I I

I 39879 I 86330 I 1.05088 I 1.15159 I 8170' I I 69578 I 1.21550 I 1.33404 I 1.37373 I 94475 I I 68298 I 1.18651 I 1.30072 I 1.33934 I 94371 I I

I CIS 8 I C17 9 I C16 10 I C15 11 I C13 12 I C11 13 I I

M-0 I P-0 I N-8 I N-4 I N-B I P-8 I I

I aAs !

I aC8 I I a5s !

I, I.45079 I 1.032M I 1.02222 I 1.10909 I 1.14692 I 1.39747 I I 80350 1 1.30434 I l.11351 I 1.18489 I 1.22591 I 1.52155 I I 78619 I 1.27130 1 1.08743 I 1.15079 I 1.19456 I 1.44640 I I D19 14 I D18 15 I 017 16 1 016 17 I 015 18 I 013 19 I D11 20 I I

I M-0 I P-0 I N-8 I P-8 I M-8 I P-4 I M-8 I I

I e58 I I 8As !

I e3a !

I I 45072 1 1.03834 1 1.03918 I 1.35864 I 97396 I 1.36962 I.96278 I I 803201 1.27629 I l.1166111.4761B I l.00716 I 1.51654 I 1.00170 I I.70592 1 1.24316 I 1.09108 I 1.41120 1 1.00361 I 1.43026 I 99736 I I E20 21 I Ett 22 I E18 23 I E17 24 I E16 25 I E15 26 I E13 27 I Ett 28 I I

M-B I P-0 I N-B !

N-8 I M-4 I P-8 I M-4 I N-8 I I

I eA I

I I

I a2e !

I l

I 29884 I 1.03215 1 1.03755 I 1.07647 I 1.02686 I 1.38876 I 9684B I 1.04904 I l

I 69588 I 1.30378 I 1.11363 I 1.13700 1 1.07988 1 1.49291 I 1.03430 I 1.08354 I I 68308 1 1.27063 I 1.08794 I l.11690 I 1.07490 1 1.42198 1 1.03265 1 1.06418 I I F20 29 I FIS 30 I F1B 31 I F17 32 I F16 33 I F15 34 I F13 35 I F11 36 I I

I P-0 I N-8 I P-8 I M-4 I N-8 I N-8 I N-4 I N-0 I I aCa I I aAs I I e5: I I

I I

I 86361 1 1.02231 I 1.3 % 90 1 1.02548 I 1.09746 I 1.12656 1 1.112 % ! 1.04578 I I 1.215 % ! 1.11342 1 1.47428 I 1.07859 I 1.14657 I 1.16797 I 1.19419 I 1.10472 1 I

! 1.18692 I 1.08728 I 1.41024 1 1.07376 I 1 12325 I 1 14688 1 1.15998 I 1.08902 I I G20 37 I GIS 38 I GtB 39 I G17 40 I B16 41 I 015 42 I G13 43 I Gil 44 1 1

P-0 I N-4 I M-8 I P-B I N-8 I P-B I N-4 I M-4 I

...--..I I eCe I I e28 I I 8Ie !

! 848 I

' I I M21 45 I 1.05137 1 1.10934 I 97384 I 1.38861 I 1.12685 I 1.39384 1 1.08108 I 90979 1

(

I M-8 I 1.33455 1 1.18526 I 1.00715 I 1 49247 1 1.16808 I 1.51090 1 1.14063 I 93947 I I

I 1.30120 I 1 15110 1 1.00350 1 1.42151 I l.14707 I l.43848 I 1.17055 I 93 % 4 I I 3 7 015 I ---------- ~~-----------------------

I I 62517 I J20 46 I Jtt 47 I J18 48 I J17 49 I J16 50 I J15 51 1 J13 52 I J11 53 I I 61290 I P-4 I N-8 I P-4 I M-4 I N-4 I N-4 I M-4 I P-4 I I a1a !

I e3e !

I I

I ege !

I I E21 54 I 1.15195 1 1.14714 I 1.36985 I 98867 I 1.11267 I 1.08208 I 93870 1 1.31759 I I

L-0 1 1.37415 1 1.22614 I 1.51684 I 1.03456 I 1.19428 I 1.14115 I 98837 I 1.45510 I I

I I 1.33172 I 1.19479 I 1 43854 1 1.03290 I 1.16019 1 1.12109 I 98598 1 1.41173 I I

.44702I----------------*-----=--------------------------------------------~~----------

r l

1 69913 I L20 55 I L19 56 I L18 57 I L17 58 I L16 59 I L15 60 I L13 61 I L11 62 I i

I 68536 I M-4 I P-B I M-9 I N-8 I N-0 I M-4 I P-4 1

E-16 I l

I s5a I I

I I e4e !

I s58 !

j

! 81726 I t.39781 I %298 I t 05019 I 104606 I 91128 ! 1.3t971 I 1.01612 I

(

I 94499 1 1 52188 I 1 00192 I 1 08390 I 1 10510 I 94140 1 1 45695 1 1 0! % 2 I l

I 143% I t 446741 19758 I 1 M451 I 109937 I 93765 I 1 41249 I 1.04195 I l

t 1 l

I f

Figure 4.4 Maine Yar.kee Cycle 10 Assembly Relative Power Densities I

BOC (500 MWD /MT), 5FP, CEA Bank 5 Inserted 1

P0tt suMMetY OF MAXIMUM FUEL ASSEMBLY P8 vet DESCt!PTION MAX.MLUE 48SEMl' Y

,3EMBLY AUG.

1.44121 42 4AX FUEL 800 1 67786 42 MAX. CNANNEL 1.59 % 8 5

C00E POSITION / ASSEMBLY NUMBER I A14 1 1 A12 2I

..... I M-B I L-0 1 FUEL TYPE.......

I I

CEA teNE TYPE a5SEMBLY AYEt4GE P8WER.

I 33287 I 35034 1 max] MUM FUEL 80D POWER

! 6M49 I 63050 I MAXIMUM CHANNEL POWER

! 58650 I 60372 I I

I I17 3 I 816 4 I B15 5 I 813 6 I I11 7I I

M-I 1 P-0 I P-0 I P-4 1 M-4 I I

I ece g 9 e1: I I

I.40075 I t.035461 1.267781 1.084581.62651 1 I

I.77709 1 1 58859 I 1.67509 1 1.54726 I.72597 1 I.75568 1 1.50720 I l.59568 1 1.47534 I.72131 I I CtB B I C17 9 I tt6 10 I C15 11 I C13 12 I C11 13 1 I

M-0 I f-0 I N-8 I N4 I N-B I P-8 !

1 I

I se e !

I ega !

I e5e !

I 33063 1 1.07635 1 1.18326 1 1 28683 1 1.10737 I 00134 I I 65217 1 1.40009 1 1.72656 I 1.48548 I 1.27054 1 1.07420 1 1 63072 I 1.42029 1 1.26513 1 1.44165 1 1 21808 1 1 013!? :

1 1 Ott 14 1 018 15 I 017 16 1 016 17 1 015 18 1 013 19 1 011 20 I I

M-0 1 P-0 1 N-I I P-8 1 M-8 I P-4 1 M-E I I

I 858 I I

A8 1

1 838 I I

I 33003 I 60786 I.96750 I 1.32775 1 1.00159 I 1.37610 1 87715 I I

I 6506B I 91807 1 1.15644 1 1.57205 1 1.05130 1 1.59415 1 94575 1 I 62932 I 87321 1 1 12513 I t.493021 1.03 11 1 1.52059 ! 93877 I I E20 21 I Ett 221 E18 23 I E17 24 I E16 25 I E15 261 E13 27 I Ett 2B :

I M-9 I P-0 I N-8 1 N-6 I M-4 I P-8 I M-4 I N-B I I

I I 8A8 1 1

I I e2e !

I I

I 3999: 1 1 07362 I 96372 I 98555 I 923d01 1.322:31 1 M924 : 1.17593 :

I 77547 1 1 48362 1 1 15185 1 1 09279 1 1 07277 1 1 % 303 I 1.17280 1 1 30962 1 I 75409 1 1.41629 1 1.12100 1 1.06950 1 1.06517 1 1 40498 I 1.15970 1 1.23595 1 1F20 29 1 F19 301 F18 31 1 F17 32 I Ftb 33 I F15 34 I F13 35 1 F11 36 I I

P-0 I N-B I P-B I M-4 I N-O !

N-6 I W4 !

k-0 I I 8Ce I I eAe 1 1 e58 I I

I I

I 1 03360 1 1.19074 1 1 32301 1 92100 1.71365 I 1.12859 1 1.30424 1 1 27353 I I

I 1.58585 1 1.32417 1 1 56807 1 1 06967 I 93771 1 1.30914 : 1 48541 I t 40072 1 1504701 1262961 14t921 1 1.M197 I 87553 1 1 25291 1 1.41905 1 1.3!772 I I 620 37 1 619 30 1 G1B 39 ! $17 40 I 616 41 1 St! 42 1 613 43 I G11 44 !

1 P-0 I W-4 I M-8 1 PI I N-8 1 P8 !

W-4 I M-4 I

I eCe I I e2e !

I e3e I I e4e !

I M21 45 1 1.265E! ! 1.284e7 1 99989 1 1 32090 1 1.12905 1 1 44121 I 1 25965 1 1 03766 !

M-B I 1.672361 1482901 1.050031 15628C I i 31008 I I 67786 I t-39370 I t 076tt :

I I 1.593041 1.43911 1 1.029761 148464 I t.253751 1.592571 1.327251 105899 I I 332E' I- - - ~ - - - - ~ - ~ - - - = - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - - - - - - - = - - - - - - ~ - - - - - - - - - - - -

I -60773 ! J20 46 1 J19 47 1 J18 40 I J17 49 I J16 50 I J15 511 J13 52 I Jtt 53 1 1 56606 I P-4 I N.8 I

P-4 I M-4 I N-4 I W-4 I M-4 I

P-4 I e1e !

I e3e 1 1

1 I eas g I

I R21 54 1 1 08295 1 1 totes I 1374891 1 M9231 1305361 1262!91 999t2 I 1.31962 1 1

L-0 I 1.54495 1 1.26857 I 1 59350 1 1 17362 1 1.48742 1 1 39708 1 1 05321 1 1.56274 I 1 1.473141 1 21615 I 1519131 1.t5952 I 1.4?i421 1.330261 1.03671 1 1 9 s2 :

349771.........................................................................41 I

I I 62956 I L20 55 I L19 % ! Lif 57 I L17 $8 I L16 59 I L15 60 1 L13 61 I L11 62 1 1 60300 I M-4 I P-8 I M9 I NB 1 W-0 I M4 I P-4 I E-16 I I e5e I I

I I e4e 1 I e5e I 1 62558 1 00040 1 87664 1 1 17645 I 1 27538 1 1 04179 1 1 32466 1 6?9'1 1 I 72494 1 1 07313 I 94538 I 1 3:072 1 1 40320 1 1 08336 I 1 57163 I 79 st :

I 72026 I 1 01255 I 939431123662 I t 35995 I 1 M2 911500011 74a43 I

I

~

Figure 4.5 Maine Yankee Cycle 10 Assembly Relative Power Densities MOC (6000 MWD /MT)e HFPe CEA Bank 5 Inserted I

CORE

SUMMARY

OF MI! MUM FUEL ASSEMILY POWER DES:tIPTION MX.VALUE ASSEMBLY ASSEMILY AVG.

1.48718 42 MI. FUEL B00 1.67043 42 M X CMNNEL l.60772 42 C0tE POS!!!DN/ASSEMILY NUM81R I A14 1 1 A12 2I FUEL TYPE.

..I R-t I L-0 I CEA BANK TYPE I

I I

ASSEMBLY AVERAGE POWEt I.35403 I 30764 I MXIMUM FUEL B00 POWER..

! 61510 I 65201 I MIIMUM CMNNEL POWER I 60064 I 63210 1 I B17 3 I B16 4 I II: 5 I B13 6 I B11 71 I

M-8 I P-0 I P-0 1 P-4 I M-4 I I

I eCe !

I e1e !

I.40442 I 99459 I 1.19911 1 1.09206 I 65550 I I

~

! 75199 1 1.47010 1 1.53736 I 1 45761 I 73503 I I.73569 I 1 41393 1 1 48387 1 1.39686 I.73354 1 1 CIB 8 I C17 9 I C16 10 1 C15 11 I C13 12 I C11 13 I I

R-0 I P-0 I N-B I N-4 I N-B I P-8 I I

I I eae !

I eCa g 3 s5: I I 33234 1 1.0343B I 1.13694 1 1.23825 1 1.08185 I 51358 I I 62392 1 1.39868 1 1.25614 1 1 40324 I 1 21781 I 1.06432 I I 606561 1.352781 1.206041 135151 1 1.1691B ! 1.01559 I I

I 019 14 I DIB 15 I 017 16 I D16 17 1 015 18 I D13 19 I Ott 20 I I

M-0 I P-0 I N-B I P-8 I M-B I P-4 I p-f I I

I 85e !

I eae I I e3e !

I I 33209 I 58165 I 93736 I 1.34753 I 1.01491 I I 38740 I 09249 I I

I 62325 I 90619 I t.10704 I I 53642 I 1.06185 I 1.58277 I 96503 I I 60591 I 04960 1 1 07954 I 1.40025 1 1.04126 I 1 5252s !

96'13 I I E20 21 I Ett 22 I EIB 23 I E17 241 E16 25 I E15 26 I E13 27 I Ett 28 I

, I I

n-B I P-0 I N-I I N-B I n-4 I P-3 I M-4 I

N-B I I

I sAe I I

I I e28 I I

I

! 40625 I 1 03340 I 934?6 I 97754 I 93124 1 1.37517 I l 06914 1 1.19326 I l

! 751701 1.397031 1 10431 1 1077751 1074721 1.572t1 1 1 17677 I 1 30455 I I

I 73540 I 1.35123 I 1.07723 1 1.05247 1 1.06522 I 1 51418 1 1.16745 1 1.24931 I l

I F20 29 I F19 301 Fil 31 1 F17 32 I F16 33 I F15 34 I F13 35 I F11 36 I l

1 P-0 I N-8 I P-I I M-4 I N-B I N-8 I N-4 I

le-O I l

1 *Ca I I eAe !

I e5s !

I I

1 l

1 99454 I 1.13633 I 1.34463 I 92918 1 71137 I 1.13667 I 1.31145 1 1.27179 I l

I 1 47009 1 1.25557 1 1 53496 1 1.07287 I 94262 1 1.29731 1 1 48785 1 1 37700 I l

I 1 41306 I 1.20544 I l.47083 1 1.06341 I 87769 1 1 25469 1 1.41619 1 1 34414 1 1

I 620 37 1 Ett 38 I G18 39 I 617 40 1 616 41 1 615 42 I G13 43 I Gil 44 I I

P-0 I N-4 I h-B I P-8 I N-8 I P-8 I N-4 I M-4 I I

I eCa 1 1 828 I 1 8Ie !

I a4: I I N21 45 I 1.19933 1 1 23809 I 1.01440 I l.37472 I 1 13710 1 1.48718 I 1.26995 1 1.05932 I I

M-I I 1.53758 I ?.40315 1 1 06163 1 1 57231 1 1.29806 I 1 67043 I l 38113 I 1.09'9 1 1

I 1.48397 I 1.35140 1 1 04099 I 1 51438 1 1.25541 1 1 60772 1 1.32019 1 1.67649 I I

I 35459 I--------------------------------------

I 61557 I J20 46 I J11 47 I Jll 40 I J17 49 I J16 50 I J15 51 I J13 52 I Jll 53 I I 60041 1 P4 I N-O I P-4 I M-4 I N-4 I R-4 I M-4 I P-4 I

...I

te I I

3e I I e3: I I

I K21 54 1 1 09213 1 1 08103 3 1 38749 1 1 08950 I 1.31215 1 1.27203 I l.02195 1 1 340?3 !

I I

L-0 I 1 45779 I 1 21774 1 1 58200 I 1.17753 1 1.48999 1 1 39410 I 1 06845 1 1.55796 !

I I l 397031 1 16907 I t.525341 1 168201 141725 I 133018 I 1.056091 150372 2 1 39762 I

--- =-- --------- ---------------- ------=--*- ---

! 65229 I L20 55 I L19 56 I L18 57 I L17 $8 I Lt6 59 I L15 60 I tt3 6t I L11 62 !

I 63232 I n-4 I P-B I n-B I N-8 I N-0 1 M-4 I P-4 E-16 I I e$s I I

I e4e !

I e5e I 1 65551 1 81364 I 99279 I 1 19400 1 1 27300 1 1 06229 I 1 3439i ! 67920 !

I 73506 I 1 09442 I 96531 ! 1 30560 1 1 37849 1 1 09572 1 1 56423 ! 01305 !

l 1 73357 I 1.01570 I 96142 1 1.24972 I t.34598 I t 07961 1 1 50982 1 75049 !

M g '

t

Figure 4.6

- I Maine Yankee Cycle 10 Assembly Relative Power Denrities EOC (13000 MWD /MT), HFP, CEA Bank 5 Inserted l

CORE

SUMMARY

OF MXIMUM FUEL ASSEMBLY PONER DESCt!PTION MI UALUE ASSEM8LY ASSEMILY AVG.

l.51903 42 I

M1. FLTL 900 1.67541 42 MX. CHANNEL 1 59303 42

)

Coat P051710N/ASSEnBLY NUMBER I A14 1 I A12 2I I

FUEL TYPE I

R-8 I

.L-0 I

CEA BAWk TYPE I

I I

ASSEMILY AVERAGE P0dEt I 39021 I.44356 I MIIMu9 FUEL 900 POWER I 65MS I 69796 I MIIMU9 CHANNEL POWEt 1 63771 I 69279 I I

I B17 3 I B16 4 I B15 5 I I13 6 I It1 7I I

M-9 I P-0 I P-0 I P-4 1 M-4 1

1 I

8C* I I eie I I

I I 42538 I 96064 I t 14541 1 1.11731 I 69:6e !

I 75356 I 1.36209 1 l 41335 I 1 37315 I 75691 I I.73906 I 1 32833 1 1 38163 1 1 334:2I 75460 !

I CIB B 1 C17 9 I Cf6 10 I C15 11 1 C13 12 I C11 13 :

I 1

M-0 I P-0 I N-S I N4 I N-B 1 PB 1 I

I

A8 I

I 8C8 I I e58 !

! 34342 1 1 00657 1 1.10843 1 1.193 9 1 1.05016 ! 81589 1 I 61557 I t 31992 1 1.20676 1 1.30190 1 1.15507 1 1 06051 :

I I 59787 I 1.29286 I 1.17639 1 1.27287 1 1.13410 I 90657 I I D19 14 I D18 15 1 017 16 I Ot6 17 I D15 18 I Ot3 11 I D11 20 I M-0 I P-0 I h-B !

P-8 I M-B I P-4 I M-8 I i

I I e58 I I ea8 I I 838 I 1

. l I 34332 I 554E1 I 91973 I 1 40485 1 1 04187 I 1.39373 I 0010! I 3

I 61529 I 88)i? ! 1.0'247 I 1.55267 I l.08180 1 1.58101 I 97513 I l

I 59760 I 81627 1 1 06697 I 1.49108 I 1 07378 ! 1 50612 I 97352 :

I E20 21 1 Ett 22 I ElB 23 I E17 24 I E16 25 I E15 24 1 E13 27 I E1' 28 !

I M-8 I P-0 I N-B 1 N-B I n4 I P-B I M-4 I N-E I I

I eae I I e28 !

! 42541 1 1.00630 I 91810 I 96829 I 94737 1 1.44433 1 1.09654 ! 1 177B? !

! 75365 I 1 3t042 1 1 07091 1 1 04914 1 1.0942t I 1.59976 1 1 15134 I 1.2fC60 I I

I 73913 1 1.29239 1 1.06550 1 1.04371 1 1.09337 ! 1 52770 1 1.14861 1 1.22100 I 1 F20 29 I F19 30 I FlB 31 1 F17 32 1 F16 33 I Ft 34 I F13 35 I F11 36 1 P-0 I N-B I P8 I N4 I N-8 I N-8 1 N-4 I N-0 I i

i I eC8 I I eae !

I e5e !

I I

I

! 96931 1 1 10850 1 1.40308 I 94596 I -69172 I 1 12344 1 1 27697 I 1 236!7 I l

I 1 36248 1 1.20663 1 1.55167 1 1 09307 I 9329e I 1 24457 1 1 39673 1 1.31051 I I 1 32869 I 1.17624 1 1.48053 1 1 0923? I........................................................................29182 I

66110 1 I 22601 1 1 35417 I l I 620 37 I 619 301 GIB 39 I G17 40 1 616 41 I G15 42 I st? 43 I G'1 4e :

I 1

P-0 I N-4 1 M-B 1 P-8 1 N-9 I P-B 1 N-4 1 M-4 1

I eCe I I e2e !

I ega !

I a4e !

! N21 45 1 1-14580 1 1 19373 1 1.04166 1 1 44402 I 1 12363 ! 1 51933 1 1 24479 1 1.05403 I j

1 M-8 I 141370 I 1.30217 I l 09169 I 1599611 124487 I 1675411 13tME ! 1072791

' E I

I 1 30196 1 1.27313 1 1.07358 1 1.52760 1 1 22627 1 1.19302 1 1.28369 1 1 06i?9 I E

i--------------------------- ------------------------- --------- -----------------

! 650?2 I J20 46 I Jt9 47 I J1B 48 I Jt? 49 I Jt6 50 I Jt5 51 I J13 52 1 J11 53 1 1 63715 I P-4 I N-8 I P-4 I p-4 I N-4 I N-4 I M-4 I P-4 :

I 8 1*

I I e3e I I

I I ege !

I

[

I #21 54 I 1 11749 I 1 05021 I 1 39377 1 1.096(2 1 1 27699 I 1 24591 1 1 02039 1 1 324'7 :

I L-0 1 1 37342 1 1.15512 1 1 58100 I 1.15156 I 1 39674 I 1 31156 I 1 H 562 1 1 51593 I I

I 1 33479 1 1.13413 1 1 S M 09 I 1 14890 1 1 35429 1 ' 26426 I I 06292 I 1 45395 1 44361 1-----------------------------------

I 69217 I L20 55 I Lt1 56 I LtB $71 Ll? $$ I L16 59 I Lt! 60 I L13 61 1 L 11 62 I I

I 68285 I n-4 I P-8 1 n-I I N-g I N-0 I M4 1 P-4 1 e$6I E1 I e5e !

I I

I e4e I I

i l

I 69570 I B15?5 I 90107 I 1 170'5 I t 23689 I 1 c5573 ! ' 32727 I 65653 I l

1 75618 I 1 H 057 1 97521 I 1 25095 I 1 3:086 I 1 07:0' I 1 520 4 I 01!2i.

I 1.75467 I 98664 ! 97360 I l 22132 I I 21243 1 1 H t4* 1 1 45744 I 74005 I

(

. NOTE: 1. THIS CURVE INCLUDES 10% CALCULAT!ONAL UNCERTAINTY T

P

2. F, p X 1.03 p

g

3. MEASURED F SHOULD BE AUGMENTED BY MEASUREMENT R

UNCERTAINTY (8%) BEFORE COMPARISON TO THIS CURVE.

i.s2

[

3 I

i i

l l

I

(

1.81 3

g i

3, COORDINATES (KMWD/MT,F )

I R

(0.00,1.761)

(0.50,1.761) i

(

1.80

,l l

e i

lll l

l l

(1.00,1.761)

(2.00,1.761)

(4.00,1.752)

(6.00,1.745)

I I !',l (6.00,1.7341 (10.00,1.735) i, i l i

I!

( Y l

l I

i i

i "q

i t 12100,1.745)

(13.00,1.7521 l

l l

l i h, l' i

g l

(14.00,1.752) i h

I M 1 i ! l l M M

% H I IM

(

g 3,73 3

i h

l Il i

I lI l

l I

o i

i I!

[

d

.I lil i

l 1.77 -

2 i

l l;

I l

I

(

i 2

[

.I

! ! l kl I l lbj l

! i y

I !

b l l

! I f

!i il lilli. ! 1! li24llh!

i ! ilk; 3i ili ! ilki iii

!!i! !l :liil hiM 'N U ltihili llhliilll i0 I c260 iiii I"k lll hh in: % iillW!

!if

! illi!! !!ili M U l li ll ilili. I i! I il MililN li

! U i. !!!illi' d!ill!

I[i !! ':!ll!k 'Iull h! II l!!l! !!'l ! l ll!! N i!! l

!!III! [

!h!!

d ii l

!!! !!! !! ! 4Inilli!hih llil luh l !!

!Ih I'inic!I! #h i'l l I; i I i b ! iil il I ill I if l 10h iNIH li h!!

j S !@it it!I Hil I :h l011l il ll! ! i iH I

'l ji! !!ll

! !) l!

l' il:;!l l' i

! t'i! l,, i!i!j! !i !.l

' llI

.; I '

' ly

' ll !; i!ll l'l!li.

i!'

i i

i 1.72 O

1 2

3 4

5 6

7 8

9 10 11 12 13 14

{

CYCLE AVERAGE EXPOSURE (KMWD/MT)

MAINE YANKEE Allowable Unrodded Radial Peak Versus Figure Cycle Average Burnup 4,7 Cycle 10 (1.6% Above Nominal)

I

-...n l

0.6 UNACCEPTA81.E OPERATION 0.5 4\\

'(

lo

~

i COORDINATES 0.4 -

'g

( 0,0.50) i

's

( 30,0.50)

((

-s X,

(100,0.12) h

\\

i h "< 0.3

\\

N g

t-T b

\\

o

(

0.2 -

2 i

N_____

\\

ACCEPTABLE OPERAT10N

'4 1 ?

l 0.1 --

?

I l

0.0 0

10 20 30 40 50 60 70 80 90 100 i

POWER LEVEL (X OF RATED POWER) 4 WAINE YANKEE Woderator Temperature Coefficient Umits re Versus Cycle 10 Puer W 4.8 -

a. y v s. ; a

.y..

.. -4 r

r p

MAXIMUM OF ACTUAL OR REFERENCE POWER LEVEL ( X ) VS, CEA WITHDRAWAL (STEPS) m ny y 100 r

m:

r ++

a@m

_Q. ll y

- d..

+

_ p n

+

_ f pn

g: :-y

+.

q j 1.1 f

1-7 z

y

.. qE

~

*211 11 s

/

~

4

~

~~

~

~~

~

~~:

~

Eh -:---

t:

~

~~

~

3 :

o

' ' ~

~

p it:

LEVEL BY GROUPS (STEPS) 4

~

J

~

~~~

".}i

~'

90- 'iii

~

POWER CEA WITHDRAWAL xito r

~:~

~

"T t

~

c:--4:

ut c

1 1:~~ ":

CI :

~

3 7.R:H

" t:::

Q 80- m$

(X) 5 4

3 2

1

~

_ r.s:

c

a W

_..f'

...* h iii 3l.

~

~~

]

E

!9 100 162 180 180 180 180

~

90 153 180 180 180 180

~

.7p

e. 'l d.!1!

~

6 i::.

~

~~

+

~~

nti 3E 16 80 141180 180 180 180 R

@ ^ 70- if 70 129 180 180 180 180

_2

~

~ j

- Wii

~

E ihl

_..d w tt E

60 116 180 180 180 180

~

V

~

jj! " 60- fii 50 90 180 180 180 180

_E

~:-

~

E :~ NE-Ei ? :

~

2111D1 o

~

0 1

ti f'

~

- l}nM 77:

40 63 171180 180 180

- :f

- E

--c

- c W

e 4

Q -2 ip 20 0 62 170 180 180 f_

~

~~

~ p. *:":

_" d$.: -:11 o

d

$1

@l 30 36 144 180 180 180

~

~

i

~

c_

_.j

~

npm

" w 50- 7:2 10 0 48 156 180 180 c_,

c

. g c

e 4 'y o,y jai 0

0 18 126 180 180 f

..f

~~

E__

f; l.J

i E 11 hie!EEElli) : ::liitil H 3

_ E

~

c J

'~

~

~~

~

~ :- S i

~

o $ ?.

g

+

f:

nfEtj.t1,!

$ C, =r px r+r-

- 2-t

_ 3 7-

- c- +

"g f

1 f'

F I i

r ::-

c 4

-:-g....g}E 1

_ y}

y.:33,34[{

t

~

n-3

$g

- g.

~

--Ef-__

y g

a 30

.g

-,mt;;. g 4

.g~

. q o

3

. ij jf.!.

g f-f

_ y-3_,

g p_

2_

+{4 g

g.

t

- 3 s

n

_ _:-c z

f J

~

g

~

- J.l

~

ACCEPTABLE i

10 Y1-l~~ 7 1. ~ * '

UNACCEPTABLE..f g

~

OPERATION

-- jf.

fE:-:-:

OPERATION

_ _[

3

~

} {titt$$f r

c

[

~

[_

.hd

~2

~

'y

.:.:g

. t'itttttt:{jj. t :.

0 GROUP 1 GROUP 3 OROUP 5 (SA AND 50) 0 50 100 150 180 0 50 100 15 0 180 0 50 100 15 0 180 3

GROUP 2 OROUP 4 E

i i

e e

i e

i s

e 0

50 100 ISO 180 0 50 10 0 150 180 CEA WITHDRAWAL BY GROUP (STEPS)

I 1

l l

1 100

=

~'

90-80-m COORDINATES E

70-g (soo, c) l l

g (632. c) 60-g (2 2,100) 50-j l

i 30-l 20-i

?

10-j

?

i

_E 0

1 500 510 520 530 540 550 560 NOMINAL COLD LEC TEMPERATURE (INDICATED)'F I

WAINC YANKEE Reference Power Level n re Versus 4.10 cycle 10 Nominci Cold Log Temperature

{

Figuro 4.11 MY CYCLES 9,1.0

[

MAXIMUM PEAKING VS. DROPPED CER WORTH FROM SPECIFIED POWER LEVELS

('

...............................................................i......................

g POER I

I i LIM 1-qq.:

. (%)

(

42 i POINTS OF INTERSEDTI(M E

0

(.0898 40.81) 40 -

(.0918 31.18)

{

(.1098 23.39) g 38 -

(.1208 20.09) go

(.1221 17.17) l O 38

(.1306 15.00)

(.1336 18.22)

[

2 34 y

Z 32 :

20 a.

6 30 !

i 2 28 -

o a-28i r

[

{

]

40 p 22i E

{

c E 20 i p

80 2

18 -

80 M 18.:

- 100 C

W 14 :

E O 12i F

POIL PERMITTED

~

N l

POWER LIMIT 8i

(%)

f

[

Bi CYCLE 9 ffT 100% POWER q

CYCLE 10 2i F

giy gV gW V W WW W W WW W WW W W W W W

WW W W W iW y W y W g y y y y y y y y y y

(

W WW W W WW W W W W W W W W W W

W WW W W W i 3 y 0.00 0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16 DROPPED CER WORTH - PERCENT DELTR RHO l -

L rL E

[

FIGURE 4.12 MAINE YANKEE CYCE10 SHUTDOWN MARGIN EQUATION

~

AND REQUIRED SCRAM REACTIVITY 7.0 6.5 -

l 6.0; HFP, P=100 i

gp % ___.--.

c

\\.,,-

5.5 -

5.0 -

4.5 -'

's,

,**HFP I

%g' 4.0 l

35i

'\\

HZP,P=0 l h,

~.--

3.0

's, HZP 2.5 -

2,0 -

g LEGEND 1.5 SHUTDOWN MARGIN EQUATION 1.0 -

REQUIRED SCRAM REACTMTY (TABLE 4.13) 0.5 -

e--------e STEAM UNE RUPTURE EVENT G-------E]

OTHER SAFETY ANALYSES I

0.0 '

0 100 200 300 400 500 600 700 800 900 1000 110 0 1200 1300 RCS SOLUBLE BORON CONCENTRATION - PPM i

E I

I l

l l

7.0 s '

s 5-

s. ' '

P =100 8.0 i'

i s

1 t

[

5.5 -

i l

P = 80

' ', s

,c s,

i m

O i

I P = 60 5.0 --

X l

'.i P = 40 l

s g_

P L,'

hr

~

\\

1 4-0 -- %

t Z

s 3.5 -

's-P=0 I

3.0 -

l-SDW = 4.21 - 0.00310 C + 0.0253 P I'15

n when C1s !ses than 326 PPW y

-g SDW = 3.20 + 0.0253 P L_

when C1s greater than or equal to 326 PPW I:

ih where

Z 2.0 - 7:

SDW 1e the required shutdown mor percent reactivity)

n C is RCS boron concentration 0 I:

P ls the maximum of either the octu power. level

Z or the reference power level corresponding to the

+-

I3 13' Z:

nominolcoldlog temperature from Figure 4.10 i:

g On por ont of rated power) i 1.0 0.5 E

300 400 500 000 700 000 900 1000 110 0 1200 1300 O.0 0

10 0 200 ACTUAL RCS BORON CONCENTRA110N - PPM Figure MAINE YANKEE Required Shutdown Norgin Venue 4.13 cvele 10 RCS Boron Concentration g I

5.0 SAFETY ANALYSIS

(

5.1 General A review of the safety analysis for operation of Maine Yankee during Cycle 10 is presented in this section. The parameters which influence the results of the safety analysis are listed in Table 5.1.

Values are provided

[

for the Reference Safety Analysis, for Cycle 9 and Cycle 10. The Reference Safety Analyais for Maine Yankee consists of the Cycle 3 stretch power analysis (36), 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 conditions, 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 9. Cycle 10, and the Reference Safety Ana?ysis values for the parameters listed above is contained in Subsections 5.1.1 throuc,h 5.1.5.

Values reported herein for MDNBR were determined using the YAEC-1 CHF

(

correlation (44). Application of this correlation with a DNBR limit of 1.20 was approved for Maine Yankee in (45).

5.1.1 Initial Operating conditions 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 10 operation in that uncertainties are included to account for measurement errors associated with plant instrumentation. The uncertainties include:

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 c) A twenty-five psi allowance for measurement error on main coolant pressure.

d) An uncertainty factor of 0.9 applied to the nominal available scram CEA worth.

I Allowable core inlet temperature and pressure conditions during power operation are specified in Technical Specifications. These are based on preserving 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 operating at lower pressures. This assures that the minimum DNBR reported for each of the incidents considered remains conservative for operation at the lower system pressures.

I The hot and cold leg RTD response times, which affect the AT power input to the RPS functions, were considered in the Cycle 10 analysis.

The safety analysis for Cycle 10 evaluated 250 plugged tubes in each steam generator for all events.

5.1.2 Core Power Distributions I

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 l

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.

I This procedure is continued for Cycle 10.

If the available overpower I

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.

I

-75 lI 1

\\

I As a starting point the safety analysis assumes the FSAR design power distribution (F, = 1.68 and F elta H =

shown in N gure 5.2.

In most cases considered, acceptable performance was demonstrated with the use of 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 Cycle 10 predicted power distributions within the S/0 LCO I

band, evaluated at their respective maximum power level limit as defined in the PDIL for Cycle 10 (Figure 4.9).

I 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 a 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 10, the thermal power margin (Pd-Po) for the 100% power PDIL case (Figure 4.9) is lower than the thermal margin for the FSAR design power distribution at full power conditions. Hence, thermal margins calculated using the 100% power PDIL power distribution are conservative for Cycle 10.

I Power peaking associated with the CEA drop and the CEA ejection eventr for Cycle 10 are compared with reference values in Table 5.1.

The effect of I

differences between Cycle 10 Cycle 9, and the Reference Safety Analysis for the CEA drop and CEA ejection are discussed in Sections 5.4.2 and 5.5.4.

I 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 10 are essentially identical with those of Cycle 9.

Variations in the above parameters will influence each transient in a dif ferent mimner.

I Therefore, the effect of the difference in reactivity coefficients is discussed on an event-by-event basis.

I For Cycle 10, the allowable positive values for MTC in the power range s.re detailed in Figure 4.8.

The analyses, limited by a positive MTC, performed in (46) and Cycle 10, with the exception of the CEA ejection analysis, were conservatively performed at 'dFP conditions with MTC equal to

+0.5 x 10 ' delta rho / F and bound the values in Figure 4.8.

The CEA I

ejection analysis for Cycle 10 (Section 5.5.4) assumes the most positive of either the predicted MTC (.able 4.5) with +0.5 x 10 delta rho / F I

uncertainty added or the value specified in Figure 4.8.

Events limited by a negative MTC 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 10 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 I

sensitive in determining the course of the power response. An evaluation of this event for Cycle 10 is provided in Section 5.5.4 5.1.4 Shutdown CEA Characteristics I

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 I

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 (36). 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.

I I I

I The normalized reactivity worth versus rod position assumed in the Reference Safety Analysis was provided as Figure 4.3 in (36). This curve is sensitive to axial power distribution and is based on the minimum reactivity insertion for a variety of axial power distributions. The normalized I

reactivity worth versus rod position was calculated for limiting Cycle 10 axial power distributions and was compared to the curve assumed in the Reference Safety Analysis. The normalized reactivity worth versus rod pcsition assumed in the Reference Safety Analysis is conservative for Cycle 10 for events limiting at HFP. The normalized control rod negative reactivity insertion versus time curve presented in Figure 4.4 of (36), which was obtained from a synthesis of the aforementioned functions, is likewise conservative in application to Cycle 10 for HFP events.

I The normalized reactivity worth versus position curve from (36) was modified for the HZP condition for Cycle 6 (43). A more conservative function 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 10 axial power distributions at zero power conditions. The Cycle 6 curve bounds all but the most bottom-peaked EOC power distribution for Cycle 10.

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 10 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 10.

The HFP curves bound the range of symmetric offset allowed by the LC0 band at intermediate power levels. The HZP curves correspond to the most negative offset case seen in the RPS setpoint axial oscillation study for Cycle 10, and assume no restriction on allowed symmetric offset.

I Values assumed in the Reference Safety Analysis and for Cycle 10 for the total negative reactivity inserted following a scram are given in Tables t

5.1 and 5.13.

Comparison of the scram worths assumed in the Reference Safety Analyses and the values assumed in the Cycle 10 safety analysis indicate that

{

the Cycle 10 values bound the Reference Safety Analysis values. The values of l !

i I

I scram reactivity specified in Table 4.13 bound those assumed in the safety analysis supporting operation of Cycle 10.

5.1.5 Reactor Protective System Setpoints and Time Delays The reactor is protected by the Reactor Protective System (RPS) and Engineered Safeguards Features (ESF).

In the event of an abnormal translent, the Reactor Protective System is set to trip the reactor and prevent I

unacceptable core damage. The elapsed time between the time when the setpoint condition 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 (36). The setpoints assumed for Cycle 10 are given in Table 5.12 and Figures 5.5, 5.6, and 5.7.

I The only difference between these values and those assumed in the Reference Safety Analysis is the low pressurizer pressure (floor of the I

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 10 are either the same as or bound those used in the Reference Safety Analysis.

I As indicated in (36) the Reference Safety Analysis assumes no credit for the high rate of change of power trip function. This remains unchanged for Cycle 10.

Credit is taken for the functioning of the Variable Overpower (V0PT), Thermal Margin / Low Pressure (TM/LP) and Symmetric Offset Trip Systems I

(SOTS) in several areas. First, the V0PT is credited in limiting the initihl power distributions considered in setting the Symmetric Offset Trip System setpoints 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 I

derived from the predicted core behavior as described in (4). The Cycle 10 setpoints for the TM/LP and Symmetric Offset Trips for 3-loop operation are E ; I

I presented in Figures 5.5, 5.6, and 5.7.

The Symmetric Offset Trip setpoints are consistent with the Technical Specifications and the TM/LP setpoints remain the same as 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.

I 5.2 Sumary Each transient and accident considered in (36) and (46) has been reviewed and/or re-evaluated for Cycle 10.

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 maintained in order to assure acceptable results.

3) Postulated Accidents.

I The incidents considered are listed in Table 5.2.

I For those transients where the parameters for Cycle 10 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) Large Break LOCA I

6) Loss of Load I I I

i s Other transients that require a partial reanalysis or review included:

i g

1) Seized RCP Rotor

2) Excess Load

3) Loss of Feedwater

4) Loss of Coolant Flow h B

5) Steam Line Rupture f

6) Steam Generator Tube Rupture

7) Small Break LOCA t-

[

A summary of results for Cycle 10 is presented in Table 5.3.

E r

5 5.3 Anticipated Operational Occurrences for which the RPS Assures No I'

Violation of SAFDLs e

The incidents in this category were analyzed in the Reference Safety r

I Analyses for the 2630 MWt Uprate and Positive MTC submittals for Maine Yankee, (36) and (46). Selected cases were reanalyzed in (20) to account for changes

[

in the Cycle 4 core physics characteristics. These analyses showed that the f

incidents in this category do not violate the SAFDLs; the primary coolant f

system pressure limit; or the 10CFR20 site boundary dose limits. The changes f

considered in the present analysis do not significantly affect the NSSS

[

response during these transients. This assures that the conclusions relative

[

g to primary system pressure and site boundary dose remain valid, b

3 v

[~.

Protection against. violation of the SAFDLs continues to be assured by I

the RPS. Setpoints are generated for the TM/LP and Symmetric Offset Trips I

which include the changes in power distributions associated with Cycle 10.

Sections 5.3.1 through 5.3.5 review the Anticipated Operational Occurrences for which the RPS assures no violation of the SAFDLs.

g 5.3.1 Control Element Assembly Bank Withdrawal t

I The Reference Safety Analysis for this event demonstrates that the most f ?.*

34

~

severe CEA withdrawal transient occurs for a combination of reactivity h.;%,

[

- I y.

addition rate and time in core life that results in the slowest reactor power

'WN tO rise to a level just below the Variable Overpower Trip. This combination of g

K P 1.[.[

=

r g;%

l 2?

?

gg

I parameters maximizes the core thermal heat flux and core inlet temperature and results in the minimum DNBR.

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 10 predicted value of MTC, with uncertainty, is -2.96 x 10-delta rho / F.

Reference (36) showed the MDNBR to occur at an MTC of -2.9 x 10- delta rho / F for this event, with less. negative MTC resulting in higher MDNBR. Table 5.1 also shows a higher maximum rate of reactivity addition for Cycle 10.

Reference (36) showed that high rates of reactivity addition result in a faster rise of core power to the Variable 5..

Overpower Trip Setpoint and values of MDNBR less limiting than for slower transients.

)i The MDNBR for a CEA bank withdrawal event for Cycle 10 occurs from an initial power level of 96.4% rated power, assuming the CEAs to be initially positioned at the corresponding insertion limit. The MDNBR for this event is

>1.20.

The peak RCS pressure for a CEA bank withdrawal is listed in Table 5.3 as less than the ASME design overpressure limit of 2750 psia.

I 5.3.2 Boron Dilution The Boron Dilution Incident was addressed in (36), (47), (48) and the FSAR.

Inadvertent dilution of the Reactor Coolant System was considered under a variety of plant conditions which could result in either an inadvertent power generation or loss of shutdown margin if sufficient time were not available for the operator to take corrective action.

MI Small changes in boron concentrations resulting from the Cycle 10 reload have an insignificant impact on the conclusions reached. An evaluation I

of this incident was performed for Cycle 10 for events postulated during refueling, shutdown, startup, hot standby and power operation conditions.

Table 5.6 presents a summary of the results of this review for Cycle 10.

I e g

(

I

I 5.3.2.1 Dilution During Refueling I

Assumptions made in the Cycle 10 evaluation for dilutions during refueling are consistent with those made in (36) and (47).

The limiting dilution in (36) was based on the maximum capacity of the CVCS via the normal makeup and letdown flow paths (200 gpm each). The limiting dilution event in (47) was based on the maximum flow of the Primary Water Makeup System (250 gpm). Both analyses assumed letdown flows equal to 3

the dilution flow rates 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 10 core loading, the critical boron concentration under cold conditions (68 F) during refueling is 1109 ppm. The minimum I

initial reactor vessel boron concentration which will prevent an inadvert,.nt criticality within 30 minutes is 1651 ppm (Case No. 3 Dilution, Reference (47)).

Therefore, it is concluded that if the reactor vessel boron concentration is maintained at or greater than 1651 ppm during Cycle 10 refueling, it would require a continuous dilution at the maximum possible rate for 30 minutes to achieve an inadvertent criticality. This is ample time for the operator to acknowledge the audible count rate signal and take corrective action to cut I

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 (48). The assumptions in (48) remain unchanged for Cycle 10.

The limiting I

dilution is via the CVCS (200 gpm), and the RCS is assumed to be filled (no credit taken for pressurizer volume). The highest worth CEA is assumed to be stuck out of the core, the loop stop valves open and either RHR or RCP on.

I Required minimum Reactor Coolant System initial boron concentrations to allow 15 minutes margin to criticality are listed in Table 5.4, along with the boron concentration reqaired to meet the Technical Specification 5% delta K/K suberiticality requirement for shutdown conditions. The boron concentrations I

I required by the Technical Specification 5% delta K/K suberiticality requirement conservatively bound those required to meet the 15-minute requirement for margin to criticality during boron dilution events.

I 5.3.2.3 Dilution During Cold Transthermal, and Hot Shutdown with Drained RCS Condition Dilutions during shutdown conditions with the RCS partially drained I

were addressed in (47) and (48).

In order to conservatively bound any partially drained configuration with one or more reactor coolant loop j

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 of 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 (47).

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" situations where the reactor vessel head may be removed to perform maintenance operations. Tabic 5.5 also shows the boron concentrations required to meet the 5% delta K/K Technical Specification suberiticality requirement for 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

' I of 30 minutes margin to criticality will be provided for the limiting boron dilution event from drained conditions.

5.3.2.4 Dilution During Hot Standby, Startup, and Power Operation The assumptions made for boron dilution events during hot standby, startup, and power operation in (36) remain the same (except for inverse boron worth) for Cycle 10.

However, the hot standby critical boron concentration f

with uncertainty is higher, 1571 ppm versus 1275 ppm.

The results for Cycle i

10 using Figures 4.3-4 and 4.3-5 of (36) are summarized below:

l

, lI l

I Maximum Reactivity Insertion Rate

-6 Dilution at Hot Standby 10.38 x 10 delta rho /sec

-6 Dilution at Power 8.81 x 10 delta rho /sec I

The consequences of events with such small reactivity addition rates are bounded by the results reported in Section 5.3.1 for the CEA Withdrawal Incident. Based on -Jae maximum reactivity addition rate it would take I

approximately 59 minutes of continuous dilution at the maximum charging rate to completely absorb a 3.2% delta K/K shutdown margin.

Because of the available 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 I

Because of the large negative moderator temperature coefficient at EOL, any 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.

I (b) The reactor is initially 3.2% suberitical at an average temperature of 550 F (a more conservative condition than the nominal 532 F).

(c) The primary system temperature is reduced at the rate of 100 F/hr, the maximum cooling rate permitted.

In order to make the reactor critical from these initial conditions, the average coolant temperat6re must be reduced from 550 F to about 430 F.

This temperature reduction requires approximately 72 minutes to I

I accomplish. This is ample time for the operator to diagnose the condition and take the necessary corrective action.

5.3.3 Excess Load Incident An Excess Load Incident is an event where a power-energy removal I

mismatch is established leading to a decrease in the reactor coolant average temperature and prer.sure. Hence, when the moderator temperature coefficient of reactivity is negative, unintentional increases in reactor power may Thus, the Excess Load Incident as reported in (36) is most limiting at occur.

EOC where the moderator temperature coefficient is most negative.

I The Cycle 10 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 value predicted for Cycle 10 including uncertainty.

Analyses of the Excess Load transient do not credit the action of the turbine load limiter in mitigating the potential overpower consequences of inadvertent openings of the turbine admission valves at power to less than 10%

above the initial power. A general approach to the excess load transient is used which credits RPS functions only. At various power levels, the limiting positive and negative symmetric offset initial power distributions are obtained from the S/0 LCO band in Figure 5.7.

From any initial power level the excess load transient is assumed to start at the S/0 LCO band and I

terminate at the symmetric offset and/or variable overpower trip limit. The MDNBR for the most limiting Excess Load event is >1.20 and corresponds to an event initiated from the positive edge of symmetric offset band at approximately 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 l

is at power.

l I l

I For Cycle 10, one change is being implemented which affects this transient. The number of steam generator tubes that could be plugged and still remain below a peak system pressure of 2750 psia was increased to 250 tubes per steam generator. The system pressure remains below 2750 psia throughout the transient with this change.

I The Loss of Load transient MDNBR is >1.20.

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.

I Feak RCS pressure for the Loss of Feedwater transient is bounded by the Loss of Load transient pressure of less than 2750 psia.

A COBRAIIIC analysis with peaking consistent with the 100% power PDIL I

was performed to determine the MDNBR for Cycle 10.

The predicted MDNBR 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 one auxiliary feedwater pump, the steam generator level reaches a minimum of 36.7% of the tube bundle height 19.3 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 Margin 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 10 conditions.

, I

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 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 10 (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 100% power PDIL power distribution is >l.20.

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 Section 5.1.2, the LCO symmetric offset band is designed to restrict permissible initial operating conditions such that the SAFDL for DNB and fuel centerline melt are not exceeded for this incident.

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 10 was evaluated from power levels ranging 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 10 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.

l 1

The CEA drop analysis also considers the increased peaking which g

results 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.

I -

I 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 l

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 tho existing power level at any point in time assuming the CEAs to be inserted no deeper than allowed by the insertion limit associated with the predrop power level.

The worst case full length CEA drop, with respect to DNB reported in (36), was the minimum worth CEA that results in the maximum increase in peaking. Thus, for conservatism the plant response assumed in the Cycle 10 evaluation was based on a worth of 0.10% delta rho.

The results of the DNB evaluation for Cycle 10 indicate that the I

limiting DNBR during a full length CEA drop cccurs 4.0 hours0 days 0 hours 0 weeks 0 months after a drop from 100%, after the power has been reduced by 20%. A MDNBR of >1.20 occurs at the positive edge of the S/0 LCO 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/0 LC0 band at each power level. The maximum allowable steady-state linear heat generation rate is limited to assure that the maximum post-drop linear heat generation rate does not violate the limiting centerline melt SAFDL. These I

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 predrop level under I

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.

I During operation with turbine admission valve control in Impulse IN mode, (IMPIN), the admission valves automatically react to changes in steam flow or pressure to maintain impulse pressure at the inlet to the first stage I

I 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 opea 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 predrop power level. This could cause core power to return to values higher than the predrop level.

The potential power overshoot is limited by the variable overpower trip setpoint to a maximum of 10 percent above the initial power level.

The allowable range of plant operat!on 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 predrop 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 developed by lowering the normal S/0 LCO by an amount equal to the maximum potential increase in post-drop power level (determined by the V0PT setpoint) for the IMPIN mode.

Both the normal and IMPIN S/0 LCO bands are shown in Figure 5.7.

l 5.5 Postulated Accidents The incidents in this category were previously analyzed in (3), (12),

(20), (36), (43), (46), (49), and (50).

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 analyses reported when Cycle 10 conditions warranted reanalysis of the accident.

l l I

5.5.1 Steam Line Rupture Accident The system analysis code, RETRAN-02 MOD 2, was used in the most recent complete Steam Line Rupture (SLR) analysis (3) to predict the consequences of a double-ended guillotine break in the main steam line coincident with a I

single failure.

The worst single failure was determined to be a feedwater regulating valve failure.

The goal of the this analysis is to determine if the core returns to criticality after the initial reactor scram. Adequate margin to suberiticality is demonstrated if the available scram reactivity and boron worth is larger than the reactivity due to moderator and Doppler defects at all times during the accident. This is conservative with respect to the actual criteria, which require that fuel damage be of sufficiently limited extent that the core will remain intact with no loss of core cooling capability and the calculated off-site doses not exceed the guidelines values of 10CFR P'rt 100.

a A system analysis was not required for Cycle 10 because none of the thermal-hydraulic characteristics have changed, making the thermal-hydraulic response predicted in the reference analysis (3) still valid.

Table 5.7 gives the nominal scram reactivity necessary to avoid recriticality for HFP and HZP cases at BOC and E00, along with the nominal available scram reactivities for Cycle 10.

The most limiting case for Cycle 10 was HFP at EOC. The minimum margin for the HFP at E0C case is 1.43%

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.

In all cases, the required scram reactivities j

calculated for Cycle 10 are within the required shutdown margin Technical Specification.

I 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 10.

The primary system I

I 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 assumed 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 I

to SGTR during Cycle 10.

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, assumed rate of flow degradation, reactor coolant low flow trip setpoint, MTC, and the primary to secondary leakage rate. Most of these factors remain unchanged in Cycle 10.

I The important differences for Cycle 10 ace a reduction in initial overpower DNB margin due to changes in the radial pin peak census and a more I

limiting 100% power PDIL power distribution. The MTC for Cycle 10 is bounded by the assumed value of +0.5 x 10- delta rho / F.

The fraction of fuel failure predicted during a seized rotor event was analyzed using the power distribution consistent with 100% power PDIL and pin census for Cycle 10.

The results for Cycle 10 are included in Table 5.3.

They indicate an increase in the fraction of fuel failure predicted for Cycle 10 as compared to Cycle 9.

These results were calculated assuming a conservative 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 10 are acceptable.

5.5.4 CEA Ejection I

The consequences of a CEA ejection accident are most sensitive to ejected CEA worth, effective delayed neutron fraction (beta effective), and post-ejection peaking. Specifically, the severity of the transient increases I I

I 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 10, including In each case, t' e values for Cycle 10 uncertainty, is presented in Table 5.1.

h exceed one or more of the parameters assumed in the Reference Safety Analyses. Thus, a reanalysis of each case has been performed for Cycle 10 I

using the modified methodology described in (13) and approved in (17).

As discussed in Section 5.1.3, the values for MTC assumed in the CEA ejection 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. The CEA scram worth versus position curves used (Figures 5.3 and 5.4) correspond to the negative edge (i.e. bottom skewed power distribution) of the symmetric offset trip band for Cycle 10 at low power levels, thereby I

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 enthalpy 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 I

Proposed changes to Maine Yankee LOCA methods are currently under review by the NRC. Maine Yankee plans to submit revised LOCA Technical Specifications to the NRC f 11owing approval of the proposed changes.

In the interim, Maine Yankee will continue to utilize administrative LOCA limits developed using NRC-approved methods.

I I

I I I

M M

M Tcble 5.1 Maine Yankee Safety Parameters Cycle 3 Cycle 9 Cycle 10 Including Including Including Uncertainties Uncertainties Uncertainties Parameter Units

- Planar Radial Peaking Factor 1.68(2)(3) 1.81 1.85 Bank 5 Inserted to 100% PDIL 1.72(7)

- Axial Peak for Shap, Resulting in MDNBR at 100% RTP 1.42(2) 1.47 1.42

- Augmentation Factors 1.0 to 1.067(1) 1.0 to 1.048(1)

None(10)

- Moderator Temperature Coefficient 10-4 delta rho /0F 0 to -2.74

+.5 to -2.81

+.5 to -2.96 1

i' - Ejected CEA Worth BOC Zero Power

% delta rho

.396

.583

.718 BOC Full Power

% delta rho

.210

.391

.371 EOC Zero Power

% delta rho

.544

.607

.627 EOC Full Power

% delta rho

.230

.443

.465

- Ejected CEA 3D Peak BOC Zero Power 13.32 15.50 16.34 BOC Full Power 5.53 7.20 6.26 EOC 2ero Power 14.08 14.40 15.18 EOC Full Power 5.59 7.35 6.44

- Dropped CEA Integral Worth % delta rho O to.30 0 to.20 0 to.20 i

e e

me m

M M

M W

M M

M E

Trblo 5.1 (Continued)

I Cycle 3 Cycle 9 Cycle 10 l

Including Including Including i

Parameter Units Uncertainties Uncertainties Uncertainties l

l

- Dropped CEA Integral Figure 4.4-1 of Figure 4.11 of Figure 4.11 Radial Peak Reference 3 Reference 64 l

- Power Level (including 2% uncertainty)

MWt 2683 2683 2683

- Maximum Reactor Coolant Inlet Temperature CF 554 548 - 556 548 - 556

- Reactor Coolant System l

Pressure psia 2200 - 2300 2050 - 2300 2050 - 2300

- Reactor Coolant System Flow Rate 106 lbs/hr 134.6(5) 134.2(5)-135.8(6) 134.2(5)-135.8(6)

- Axial Power Distribution Symmetric Figure 6.3-1 Figure 5.7 of Figure 5.7 Limit Offset of Reference 3 Reference 64

- Power Dependent Figure 4.9 of Figure 4.9 of Figure 4.9 Insertion Limit Reference 20 Reference 64

- Initial Steady-State (4) 1.895(5) 1.897(5)

Minimum DNB Ratio YAEC-1 1.977 1.911(6) 1.915(6)

- Maximum Possible Rate of Reactivity Addition (9) delta rho /sec 0.7x10-4 1.40x10-4 1.29x10-4 Nominal

- Steam Generator psia 877 877 - 610 877 - 610 Pressure (100%)

M M

m m

M M

m W

m mm m

M M

Table 5.1 (Continued) i Cycle 3 Cycle 9 Cycle 10 Including.

Including Including Parameter Units Uncertainties Uncertainties Uncertainties Steam Generator (SG)

Tubes Plugged /SG 180 250

- Minimum Required Worth in CEAs Assumed in Safety Analysis % delta rho i

i HFP, BOC 4.0 5.70 5.70 HZP, BOC 2.0 3.20 3.20 HFP EOC 5.7 6.02 5.85 6.5(8) i i

l HZP ECO 2.9 3.62 3.22 l

I l

l

I Table 5.1 I

(Continued)

Notes 1)

Applies only in fuel centerline melt calculations.

2)

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

3)

Values shown in Reference 12 did not include uncertainty.

4)

FSARdesignpowerdistribution(F[eltaH=1.49,Fz = 1.68).

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

5)

I Based on Reactor Coolant System pressure of 2200 psia, and temperature of 5560F.

6)

Based on Feactor Coolant System pressure of 2050 psia, and temperature of 5480F.

7)

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

8)

EOC, HFP steam line break assumed 6.5% delta rho.

9)

For CEA bank withdrawal transient.

10) Augmentation factors have been removed for Cycle 10, see Section 4.11.2.

I I

I t I I I l I

Table 5.2 Maine Yankee Cycle 10 - Incidents Considered I

A.

Anticipated Operational Occurrences for which the RPS assures no violation of SAFDLs:

1.

Control Element Assembly Bank and Subgroup Withdrawal I

2.

Boron Dilution 3.

Excess load 4.

Loss of Load 5.

Loss of Feedwater B.

Anticipated Operational Occurrences which are dependent on Initial Overpower Margin for protection against violation of SAFDLs:

1.

Loss of Coolant Flow 2.

Full Length CEA Drop C.

Postulated Accidents:

1.

CEA Ejection 2.

Steam Line Rupture 3.

Steam Generator Tube Rupture I

4.

Seized Rotor 5.

Loss of Coolant I

I I

I I I I

M M

M m

m m

M M

Table 5.3 Maine Yankee Cycle 10 Safety Analysis Summary of Results Incident Section Criteria Reference Safety Analysis Cycle 9 Cycle 10 I

Cycle 3 CEA Withdrawal 5.3.1 MDNBR = 1.20 MDNBR = 1.51*

MDNBR = 1.42 MDNBR = 1.47 RCS pressure RCS pressure RCS pressure RCS pressure 2750 psia 2570 psia 2570 psia 2570 psia LHGR SAFDL Not exceeded Not exceeded Not exceeded Cycle 3 Boron

5.3.2 Suberitical

Suberitical:

Dilution Sufficient time Refueling-65 min.

Refueling-30 min.

for operator Startup-3.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Startup-15 minutes Startup-20 minutes action Critical: Bounded Critical: Bounded Critical: Bounded Critical: MDNBR by CEA withdrawal by CEA withdrawal by CEA withdrawal 1.20 Cycle 3 i

$ CEA Drop 5.4.2 MDNBR = 1.20 MDNBR = 1.36*

MDNBR = 1,29 MDNBR = 1.42 LHGR SAFDL Not exceeded Not exceeded Not exceeded.

Cycle 3 Loss of 5.4.1 MDNBR = 1.2 MDNBR = 1.50*

MDNBR = 1.51 MDNBR = 1.38 Coolant Flow Cycle 5 Seized Pump 5.5.3 10CFR100 8.4% of rods with 7.5% of rods with 10.3% of rods with Roto?

MDNBR less than 1.3*

MDNBR less than 1.2 MDNBR less than 1.2 Radiological dose Radiological dose Radiological dose within 10CFR100 within 10CFR100 within 10CFR100 Cycle 4 Excess Load 5.3.3 MDNBR = 1.2 MDNBR = 1.7*

MDNBR = 1.63 MDNBR = 1.42

W W

W m

M M

m W

W W

W Tcbic 5.3 (Continu-d) l l

Incident Section Criteria Reference Safety Analysis Cycle 9 Cycle 10 i

Cycle 3 i

Loss of Load 5.3.4 MDNBR = 1.2 MDNBR = 1.85*

MDNBR = 1.69 MDNBR = 1.92 RCS pressure RCS pressure RCS pressure RCS pressure

<2750 psia 2689 psia

<2750 psia (2750 psia l

l Cycle 3 l

Loss of 5.3.5 RCS pressure Peak RCS pressure Peak RCS pressure Peak RCS pressure Feedwater

<2750 psia 2600 psia

<2750 psia

<2750 psia MDNBR = 1.20 MDNBR = 1.69 MDNBR = 1.61 Cycle 9 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 Cycle 3 Steam Gener-5.5.2 10CFR100 Radiological dose Radiological dose Radiological dose y ator Tube within 10CFR100 within 10CFR100 within 10CFR100 g Rupture Cycle 9 CEA Ejection 5.5.4 10CFR100 Radiological dose Radiological dose Radiological dose within 10CFR100 within 10CFR100 within 10CFR100 Cycle 3 Steam Line 10CFR100 Radiological dose Reference analysis Reference analysis Rupture Outside within 10CFR100 unchanged by Cycle unchanged by Cycle Containment 9 reload 10 reload

m m

M M

M m

m m

M M

Table 5.3 (Continued)

Incident Section Criteria Reference Safety Analysis Cycle 9 Cycle 10 Cycle 3 Feedwater 10CFR100 Bounded by steam line Reference analysis Reference analysis Line Rupture rupture unchanged by Cycle unchanged by Cycle Outside 9 reload 10 reload Containment Cycle 3 Containment Peak pressure Peak pressure less Reference analysis Reference analysis Pressure less than 55 psig than 55 psig unchanged by Cycle unchanged by Cycle containment 9 reload 10 reload design pressure Cycle 3 Fuel Handling 10CFR100 Radiological dose Reference analysis Reference analysis Incident within 10CFR100 unchanged by Cycle unchanged by Cycle 9 reload 10 reload i

Cycle 3

$ Waste Gas 10CFR100 Radiological dose Reference analysis Reference analysis y System within 10CFR100 unchanged by Cycle unchanged by Cycle Failure 9 reload 10 reload Spent Fuel 10CFR100 NA NA NA Cask Drop Cycle 3 Radioactive 10CFR100 Radiological dose Reference analysis Reference analysis Liquid Waste within 10CFR100 unchanged by Cycle unchanged by Cycle System Leak 9 reload 10 reload W-3 DNBR shown for Reference Safety Analysis result. YAEC-1 for Cycles 9 and 10.

I Table 5.4 Cycle 10 Required Initial RCS Boron Concentrations to Allow 15 Minutes Margin to Criticality for Dilutions from Shutdown Conditions I

with the RCS Filled I

Required Initial Concentration (ppm)

Boron Dilution 5% Delta K/K*

I

ARI, B00 5320F 744 1140 3000F 897 1190 680F 918 1169

ARI, I

E00 5320F 36 115 3000F 40 309 I

680F 135 339 ARI Less 1 Stuck CEA, BOC 5320F 952 1352 3000F 1066 1366 680F 1129 1383 I

ARI Less 1 Stuck CEA, EOC 5320F 36 333 3000F 248 485 680F 344 516 ARI w/ Withdrawn Bank,**

BOC 5320F 1177 1487 3000F 1313 1513 1

680F 1321 1478

-102-I I

I Table 5.4 (Cont'd)

Required Initial Concentration (ppm)

Boron Dilution 5% Delta K/K*

ARI w/ Withdrawn Bank,**

EOC 5320F 138 431 3000F 475 584 680F 548 596 I

ARO, BOC 5320F 1748 2161 3000F 1677 1962 680F 1605 1844

ARO, I

EOC 5320F 743 1000 3000F 799 941 680F 770 873 I

I I

I g

Margin of suberiticality required by Technical Specifications for shutdown l g conditions.

l

- The critical boron concentrations for 2 stuck CEAs bound those for a withdrawn bank. Therefore, the boron concentrations provided for 2 stuck l

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.

-103-1 I

I Table 5.5 Cycle 10 Required Initial RCS Boron Concentrations to Allow 30 Minutes Margin to criticality for Dilutions from Shutdown Conditions with the RCS Drained

Required Initial Concentration (ppm)

Boron Dilution 5% Delta K/K*,

I

ARI, I

BOC 5320F 1117 1140 3000F 1246 1190 680F 1234 1169 I

ARI, 5320F 0

115 3000F 55 309 680F 142 339 ARI, Less 1 Stuck CEA, BOC 5320F 1435 1352 3000F 1484 1366 I

680F 1521 1383 i

ARI, Less 1 Stuck CEA, E0C 5320F 0

333 I

3000F 271 485 680F 365 516 I

ARI w/ Withdrawn Bank,***

5320F 1783 1487 3000F 1832 1513 680F 1783 1478 I

-104-

Tsble 5.5 (Cont'd)

I Required Initial Concentration (ppm)

Boron Dilution 5% Delta K/K**-

ARI, w/ Withdrawn Bank,***

EOC 5320F 158 431 3000F 522 584 680F 583 596

ARO, BOC 5320F 2605 2161 3000F 2263 1962 680F 2088 1844 EOC 5320F 868 1000 3000F 864 941 680F 805 873 I

o Level = lower lip of RV nozzles.

I

- Margin of suberiticality required by Technical Specifications for shutdown conditions.

- - The critical boron concentrations for 2 stuck CEAs bound those for a I

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 testing.

-105-I

I Table 5.6 Summary of Boron Dilution Incident Results for Cycle 10 (A)

(B)

(C)

(D)

Minimum Technical Minimum Time Operating Specification Shutdown to Absorb (B)

Acceptance I

Mode Margin Requirement Minutes Criteria (5)

Refueling 5% delta K/K 30 30 Cold Shutdown Filled RCS 5% delta K/K 15 15 Drained RCS 5% delta K/K(1) 30(1) 30(1)

Hot Shutdown Filled RCS 5% delta K/K 15 15 I

Drained RCS 5% delta K/K(1) 30(1) 30(1)

Startup 3.2% delta K/K(6) 20(2) 15 Hot Standby 3.2% delta K/K(6) 59(3) 15 Power Operation 3.2% delta K/K(6) 59(3) 15 Failure to Borate Prior to Cooldown 3.2% delta K/K(6) 72(4) 15 l

(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 administrative 1y required for these conditions.

(2) Margin quoted assumes initial boron concentration at refueling value for Cycle 10, 1651 ppm.

(3) Time to absorb minimum specified 3.2% delta K/K shutdown margin.

l (4) Cooldown rate assumed to be 1000F/hr.

i (5) Time span between event initiation and criticality.

(6) Minimum value for shutdown margin specified in Technical Specification.

-106-

I

Table 5.7 Nominal Scram Reactivity Worths Required to Prevent a-Return to Power During a Steam Line Rupture Accident Nominal Scram Reactivity Case Required Available HFP BOC 3.08%

7.52%

HFP EOC 6.53%

7.96%

I HZP BOC 1.46%

5.39%

HZP E0C 3.61%

5.13%

I I

-107 -

Table 5.8 Cycle 10 CEA Ejection Accident Results I

Full Power BOC Eg Fraction of Rods that Suffer Clad 3.6 0.0 Damage (Radial Average Enthalpy Above 200 cal /gm). %

Fraction of Fuel Volume Exceedfug 0.0 0.0 Incipient Melting Criteria (Enthalpy greater than 250 cal /gm), %

2ero Power Fraction of Rods that Suffer Clad 0.0 0.0 Damage (Radial Average Enthalpy Above 200 cal /gm), %

Fraction of Fuel Volume Exceeding 0.0 0.0 Incipient Melting Criteria (Enthalpy greater than 250 cal /gm), %

I

! I I

-108-I

Table 5.9 Comparison of Thermal Margin for Limiting Cycle 10 Power Distributions to FSAR Design Power Distribution I

Power Power YAEC-1(

YAEC-1( }

Level Distribution MDNBR (Pd-Po) 100 FSAR(3) 1.897 28 100 (2)(3) 1.875 24 94 (2)(3) 1.891 25 I

t

{

83 (2)(3) 2.047 31 75 (2)(3) 2.123 36 57 (2)(3) 3.093 45 100 FSAR(7)(4) 1.946 33 l

100 FSAR(7)(5) 1.962 34 l

100 FSAR(7)(6) 1.919 31 100 FSAR(7)(8) 1.911 31 4

s' (1) Includes allowances for 2% calorimetric power uncertainty, 3% tilt, and 10% physics radial peaking uncertainty on non--FSAR ca'ses.

(2) Limiting Cycle 10 power distribution within sytnmetric offset pretrip alarm band plus uncertainty for indicated power level for Cycle 10.

(3) At 2200 psia, 556 degrees F.

(4) Cycle 9.

(5) Cycle 3.

(6) Cycle 5.

(7) At 2200 psia, 554 degrees F.

(8) Cycle 6

-109-I

l Table 5.1^

Reactor Protective System Trips Assumed in the Cycle 10 Safety Analysis I

Delay Setpoint Uncertainty Time (Sec)

I High Neutron Flux 106.5%

5.5%

0.4 Low Reactor Coolant Flow 93%

2%

0.65 High Pressurizer Pressure 2400 psia 22 psi 0.9 Low Steam Generator Pressure 500 psia 122 psi 0.9 I

Low Steam Generator Water Level 35% NR 10 in 0.9 Low Pressurizer Pressure

1850 psia 22 psi 0.9 Safety Injection Signal 1600 psia 22 psi I

Thermal Margin Trip Figure 5.5 (4) 0.9 and 5.6 Synn.etric Offset Trip Figure 5.7

.04 asiu 0.9 Variable Overpower Trip Q + 10%***

5.5%

0.4 I

I I

Low limit of thermal margin trip.

I

- See specific accident for time delay assumed for safety injection delivery.

- - Q = Initial indicated power level in percent thermal or nuclear power.

-110-I l

e Table 5.11 Required Scram Reactivity Assumed in Cycle 10 Safety Analysis I

Required Scram Reactivity (% delta rho)

Event BOC, HZP EOC, HZP BOC, HFP EOC, HFP CEA Withdrawal 4.00 4.00

I Boron Dilution 3.20 3.20 Excess Load 3.20 3.20 5.70 Loss of Load 4.00 Loss of Feedwater 4.00 I

Loss of Coolant Flow 5.50 CEA Drop I

CEA Ejection 3.20 3.20 5.70 5.70 Steam Line Rupture

1.29 3.22 2.75 5.85 I

Steam Generator Tube Rupture 2.00 2.40 4.00 5.70 Seized Rotor 4.00 Loss of Coolant Maximum Assumed in Any Event 3.20 3.22 5.70 5.85 1

An uncertainty factor of 0.9 is applied to the nominal required scram reactivities ass'.ae' 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 (14).

I I

I

-111-

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2300 I

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MAINE YANKEE Aliowable 3 Loop Steady State Figure Coolant Conditions 5.1 I

-112-

+ -,,, - -.., - -.,

,w-a -

sus num aus mum uma muu num unas muu num num mum uma em uma uma uma sam uma b

ea BM o.<

l Normalized Axial Power "m

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=

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l ncuat 5.3 MRINE YANKEE OYCLE 10 NORMRLIZED RERCTIVITY WORIH VS PERCENT CER IN5ERTION g

CER EJECTION: BOC, SCRAM AT HFP RND HZP g

g,g I

0.9 -

I 0.8 -

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0.0 - e-,..

0 10 20 30 40 50 60 70 80 90 100 g

PCRCCNT CER INSERTION

-114-I

FIGURE 5.4 g

MRINE YANKEE CYCLE 10 NORMALIZED RERCTIVITY WORTH VS PERCENT CER INSERTION CER EJECTION: E00, SCRAM RT HFP RND HZP 1.0 0.9 I

0.8 -

I 0.7 -

- I E

o2 U.

0.6 -

I Eg 0.5-I m

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0.4 2 3

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0.2 -

i HZP 0.1 -

l 0.0

..c 0

10 20 30 40 50 60 70 80 90 1T I

PCRCENT CER INSERTION l

-its-

- %. s. %

TRIP 10053.0 AND Pg = 1959.2

+ U.9T C

C = COLD LEG TEMPERATURE,'F T

1.40 l

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=

l 1.30 -

d' i

i Ag(+)=.74011(S.0d+1.02635

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: :: ::: ::: :: a-- :::::::::: ::::: _ __ _-._+__

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Excore Symmetric Offset I

WAINE YANKEE Thermal Margin / Low Pressure Trip Setpoint Figure 5.5 5-Port 1 (A versus Excore Symmetric Offset) 3

LI 1I I

WHERE: QDNB = A e QR1 1

TRIP

- 10053.0 AND P

= 1959.2 0DB+ 17.9TC Wt l

T = COLD LEG TEMPERATURE,oF C

I I

I

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l V

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==

I 0.d 0.2 0.4 0.8 0.8 10 1.2 Fraction of Roted Thermal Power I

MAINE YANKEE Thermot Margin / Low Pressure Figure 5.6 Trip Setpoint Pert 2.

(QR1 versus Froetion of I

Roted Thermal Power) 117-E

l i

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-0.6-0,5 -0,4 -0,3 -0,2 -0,1 0.0 0.1 0,2 0,0 0,4 0,5 0,8 Symmetric Offset Y, = A*(U-L/U+L)+B l

AINE YANKEE Symmetrie Offset Trip Funetion Figure l

l Three Pump Operation 5.7

-118-I

I 6.0 STARTUP TEST PROGRAM The startup test program includes low power physics and power escalation 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; I

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 rated full power with a primary system temperature and pressure of approximately 532 F and 2250 psia, respectively.

I 6.1 Low Power Physics Tests I

The following reactor parameters are measured at the low power conditions:

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 CEA groups 1, 2, 3, 4 and 5 in the nonoverlap condition. The total of the worths of these groups must be within !10% of the predicted value.

If I

this condition is not met, then Banks B and C will be measured and the sum of the worths of all measured groups must be within 10% of predicted. Bank A is reserved to maintain shutdown margin requirements and has been eliminated from measurement for Cycle 10.

I

3) 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.

I

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I

4) CEA drop times are measured by monitoring reed switch valtage for position indication versus time. All scrammable full length CEA drop times are measured.

The most limiting near full power ejected CEA worth has been measured J

at the pre-ejection conditions by the boron dilution technique in all previous cycles and has always met the acceptance criteria. Since the ejected CEA worths have been well predicted and the CEA group measurements provide an I

indication of maximum ejected CEA worth, the ejected CEA vorth measurement has been eliminated for Cycle 10.

6.2 Power Escalation Tests Power escalation tests assure the performance of various primary and secondary plant systems. Plant parameters are stabilized and test data taken at approximately 48% and approximately 100% of rated power.

I The following plant parameters are evaluated at the above power levels, or as indicated:

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

I

2) Isothermal temperature coefficients, if required, are derived at 48% power by partially closing the steam turbine governor valves I

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 I

Acceptance criteria for the prediction of key core pararaeters 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 I

nominal measured value is compared to the nominal calculated value, the latter

-120-

I 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 I

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 values for Cycle 10 are unchanged from Cycle 9.

If the initial criteria in Table 6.1 are not met, additional measurements, as prescribed by the table, are performed.

In addition, any deviations are evaluated relative to the assumptions in the safety analysis for the given core parameters. Such deviations and their evaluations are I

reported to the Staff. A startup test summary report will be available on-site within 90 days of the completion of the startup tests.

I I

t!I I

I

-121-I

,I 1

Table 6.1 Maine Yankee Cycle 10 Startup Test Acceptance Criteria I

Measurement Conditions Criteria I

1.

Critical Boron Hot zero power, near Measurement within 1%

Concentration all-rods-out delta rho of predicted value 2.

CEA Bank Worths -

Hot zero power, CEA Total worth within 10%

Regulating Banks 1+2+3+4+5 in the of the predicted value nonoverlap condition I

3.

CEA Bank Worths -

Hot zero power, CEA This measurement is not Shutdown Banks B+C+1+2+3+4+5 required if the criteria in in the nonoverlap Measurement (2) is met.

If condition the criteria in Measurement (2) is not met, the total worth of all CEA banks I

measured must be within 110% of the predicted value I

4.

Isothermal Hot zero power, near ITC measurement within Temperature all-rods-out 0.5 x 10-4 delta rho /0F Coefficient at of predicted value and the I

HZP MTC is within the acceptable region specified in Figure 4.8 5.

Isothermal At or slightly below 50%

This measurement is not Temperature power, near all-rods-out required if both criteria Coefficient at in Measurement (4) are met.

50% Power If either criteria in Measurement (4) are not met, the MTC must be in the l 3 acceptable region specified j g in Figure 4.8 i

6.

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

Radial Power At or slightly below Each assembly average I

Distribution 50% power, near all-power within !10% of rods-out predicted value 8.

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%

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

-122-I 1

I

7.0 CONCLUSION

I The results of analyses presented herein have demonstrated that design criteria as specified in the FSAR will be met for operation of Maine Yankee during Cycle 10.

Table 5.3 sunurarizes the results of each incident analyzed; including the Reference Cycle result and the appropriate design limit. This I

table illustrates that Specified Acceptable Fuel Design Limits (SAFDL) on DNB and fuel centerline melt, the primary coolant system ASME code pressure limit, and the 10CFR100 site boundary dose limits are not violated for any of the incidents considered.

I I

I

-123-I

I

8.0 REFERENCES

1.

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.

I 2.

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

I 3.

YAEC-1479, " Maine Yankee Cycle 9 Core Performance Analysis," April 1985.

4.

P. A. Bergeron, D. J. Denver, " Maine Yankee Reactor Protection System Setpoint Methodology", YAEC-1110, dated September 19 76.

I 5.

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

6.

R. N. Gupta, " Maine Yankee Core Analysis Model Using CHIC-KIN", YAEC I

1103, dated September 1976.

7.

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

I 8.

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

YAEC-1101, dated June 1976.

i l

9.

D. J. Denver, E. E. Pilat, R. J. Cacciapouti. " Application of Yankee's I

Reactor Physics Methods to Maine Yankee", YAEC-1115, dated October 1976.

l l

l 10.

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

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

!I 11.

YAEC-1160, " Application of YANKEE WREM-BASED Generic PWR ECCS Evaluation l

Model to Maine Yankee", July 1978.

12.

YAEC-1324 " Maine Yankee Cycle 7 Core Performance Analysis",

September 1982.

-124-

I 13.

YAEC-1464, " Modified Method for CEA Ejection Analysis of Maine Yankee Plant", December 1984.

14.

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.

I 15.

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

I 16.

USNRC Letter to Yankee Atomic dated January 17, 1979.

17.

USNRC Memorandum to G. C. Lainas from L. S. Rubenstein, dated June 20, 1985, " Safety Evaluation Report of YAEC-1464 Maine Yankee Modified Method for CEA Ejection Analysis."

18. NMY 85-166, " Safety Evaluation of the Maine Yankee Atomic Power Corporation (MYAPC) Report YAEC-1447, Applications of RETRAN02/ MOD 02 and BIRP to the Analysis of the MSLB Accident at MYAPC," E. J. Butcher, I

October 2, 1985.

19.

Maine Yankee Letter to USNRC, WMY 75-28, dated March 27, 1975, " Maine Yankee Core 2 Performance Analysis".

I 20.

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

21. Maine Yankee Batch N Reload Fuel Design Report, April 1985.

22.

Letter, W. P. Johnson (MYAPC) to USNRC, WMY 75-28, dated June 26, 1975,

" Maine Yankee Proposed Change No. 27 (Cycle 2 Core Performance Analysis Report)."

!I 23.

C. A. Brown, " Generic Mechanical Design Report for Exxon Nuclear Maine Yankee 14 x 14 Reload Fuel Assemblies," XN-NF-81-39P, June 1981.

24.

Maine Yankee Batch P Reload Fuel Design Report.

l

-125-I

l I 25.

XN-NF-81-85, " Mechanical Design Report Supplement for Exxon Nuclear Maine Yankee XN-1 through XN-4 Extended Burnup Program", November 1981.

26.

XN-NF-86-94(P), " Mechanical Design Report Supplement for Exxon Nuclear Maine Yankee XN-3 and XN-4 Extended Burnup," September 1986.

I 27.

M. J. Ades, " Qualification of Exxon Nuclear Fuel for Extended Burnup,"

XN-NF-82-06(P), Revision 1 June 1982.

28.

" Safety Evaluation of the Exxon Nuclear Company Topical Report, XN-NF-86-06(P), " Qualification of Exxon Nuclear Fuel for Extended Burnup," July 1986.

I 29.

XN-73-25 GAPEX: A Computer Program for Predicting Pellet-to-Cladding Heat Transfer Coefficients August 13, 1973.

1 30.

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).

31.

Combustion Engineering Report, TR-DT-34, "The Hydraulic Performance of the Maine Yankee Reactor Model", June 1971.

32.

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

33. CEND-414. "The Evaluation and Demonstration of Methods for Improved Fuel Utilization", October 1983.

34.

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.

I

35. XN-75-32 (NP) Supplement 2

" Computation Procedures for Evaluating Fuel Rod Bowing", July 1979.

36.

P. A. Bergeron, P. J. Guimond, J. DiStefano, " Justification for 2630 MWt I

Operation of the Maine Yankee Atomic Power Station", YAEC-Il32, dated July 1977.

-126-

I 37.

XN-NF-79-52, " Maine Yankee Reload Fuel Design Report / Mechanical Thermal-Hydraulic cnd Neutronic Analyses", 1979.

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

I

39. MYAPC Letter to USNRC, MN-83-76, " Reactor Vessel Pressurized Thermal Shock (PTS)", April 22, 1983, with Enclosures A, B, and C.

40.

MYAPC Letter to USNRC, MN-84-88, " Reactor Vessel Pressurized Thermal Shock (PTS)", June 1, 1984, with Enclosures A, B, C, and D.

41.

MYAPC Letter to USNRC, MN-86-69, "Angmentation Factor Removal,"

May 20, 1986.

I 42.

USNRC Letter to MYAPC, dated June 20, 1986, " Augmentation Factor Removal," with Attached Safety Evaluation Report.

43.

YAEC-1259, " Maine Yankee Cycle 6 Core Performance Analysis", Attachment to MYAPC Letter to USNRC, FMY-81-65, Proposed Change No. 84, dated April 28, 1981.

44.

J. Handschuh, "DNBR Limit Methodology and Application to the Maine Yankee I

Plant," YAEC-1296P, January 1982, Attachment to YAEC Letter to USNRC, FYR-82-41, MN-82-78, dated April 8, 1982.

45.

USNRC Letter to MYAPC, dated March 9,1983, NMY 83-62, " Topical Report YAEC-1296P, "DNBR limit Methodology and Application to the Maine Yankee Plant".

l I

l 46.

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

II l

47. Maine Yankee Letter to USNRC, WMY 78-2, dated January 5, 1978.

l

-127-i

I

48. MYAPC Letter to USNRC, MN-82-53, " Boron Dilution During Hot and Cold Shutdown (Mode 5 Operation)," dated March 18, 1982.

49.

Cycle 6 MSLB Analysis, Attachment to MYAPC Letter to USNRC, FMY 81-162, dated October 29, 1981.

50. YAEC-1396, " Maine Yankee Cycle 8 Core Performance Analysis", January 1984.

51.

Maine Yankee Letter to USNRC, WMY 77-87, dated September 22, 1977.

52.

" Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled Nuclear Power Reactors", Federal Register, Vol. 39, No. 3-Friday, January 4, 1974.

I

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- - - -

ML20207P536

Contents

Text

1.0 INTRODUCTION

7.0 CONCLUSION

8.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

5.3.2 Suberitical

7.0 CONCLUSION

8.0 REFERENCES

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ML20207P536

Text

1.0 INTRODUCTION

7.0 CONCLUSION

8.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

5.3.2 Suberitical

7.0 CONCLUSION

8.0 REFERENCES

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