ML20150E906
| ML20150E906 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 07/31/1988 |
| From: | Paul Bergeron, Cacciapouti R, Michael Scott Maine Yankee |
| To: | |
| Shared Package | |
| ML20150E900 | List: |
| References | |
| YAEC-1648, NUDOCS 8807180045 | |
| Download: ML20150E906 (142) | |
Text
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0 MAINE YANKEE CYCLE 11 l
CORE PERFORMANCE ANALYSIS July 1988 Major Contributors:
Reactor Phy_ sics Group Transient Analysis Group K. B. Spinney K. R. Rousseau G. M. Solan V. M. Esquillo D. G. Adli A. S. Fatemi M. R. Durrenberger T. D. Radcliff S. VanVolkinburg M. A. Volk LOCA Analysis Group J. M. Ghaus Q. A. Haque G. E. Jarka K. E. St. John P. A. Theriault Yankee Atomic Electric Company Nuclear Services Division 1671 Wercester Road Framingham, Massachusetts 01701 6341R/23.193 8807180045 880630 PDR ADOCK 05000309 P PDC
l APPROVALS i
Prepared By: M _/ /[ k M. W. S'cott, Nuclear Engineering Coordinator /(Dale)
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Nuclear Engineering Department Approved By: -
M P. A. Bergeron, Mar ger (Date)
TransientAnalysis[ Group Approved By: *
, }*L IP[N!II
- f. J. Racciasouti, Manager F F (Date)
Reactor Physics Group
/.pproved By: 1 . b lI4 !EE S. P.'1chultz, Mandger % (Date)
LOCA Analysis Group Approved By: .), '
h b B.'C.Slifer, Direct [G (Date)
Nuclear Engineering Mpartment l
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1 l
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I 6341R/23.193 I
DISCLAIMER OF RESPONSIBILITY This document was prepare'd by Yankee. Atomic Electric Company
("Yankee"). The use of information contained in this document by anyone other than Yankee, or the Organization for which this document was prepared under.
- contract, is not authorized and, with respect to any unauthorized use, neither Yankee nor'eits officers, directors, agents, or. employees assume any obligation, responsibility, or liability'or make any warranty or representation as to the accuracy'or completeness of the material contained in.this document.
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ABSTRACT This report presents design and analysis results pertinent to-the operation of Cycle 11 of the Maine Yankee Atomic Power Station. These include core fuel loading, fuel description, reactor power distributions control' rod worths, reactivity coefficients, the results of the safety analyses performed to justify plant operation, the startup test program and the Reactor Protective System (RPS) setpoints assumed in the safety analysis. The 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 11.
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l' TABLE OF CONTENTS l
l Page APPR0VALS........................................................ 11 D I S C LA I ME R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii ABSTRACT......................................................... iv-TABLE OF C0NTENTS................................................ v LIST OF TABLES................................................... vili LIST OF FIGUEES.................................................. 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 2.4 Cycles 5 and 6............................................. 4 2.5 Cycles 7 and 8............................................. 4 2.6 Cyc'.e 9.................................................... 5 2.7 Cycle 10................................................... 5 3.0 RELOAD CORE DESIGN............................................... 8
.1 General Description........................................ 8 3.1.1 Core Fuel Loading.................................. 8 3.1.2 Core Burnable Poison Loading....................... 8 3.1.3 Core Loading Pattern............................... 9 3.1.4 Assembly Exposure History.......................... 9 3.1.5 CEA Group Configuration............................ 10 3.2 Fuel System Design......................................... 11 3.2.1 Fuel Mechanical Design............................. 11 3.2.2 Fuel Thermal Analysis.............................. 11 3.2.3 The rma l-Hyd ra ul i c De s i gn . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4.0 PHYSICS ANALYSIS................................................. 31 4.1 Fuel Management............................................ 31 4.2 Core Phyaics Characteristics............................... 31 4.3 Power Distributions........................................ 31 4.4 CEA Group Reactivity Worths................................ 32 4.5 Doppler Reactivity Coefficients and Defects................ 32 4.6 Moderator Reactivity Coefficients and Defects.............. 33 4.7 Soluble Boron and Burnable Poison Reactivity Effects....... 34 4.8 Kinetics Parameters........................................ 34 4.9 Safety-Related Characteristics............................. 35 4.9.1 CEA Group Insertion Limits......................... 35 4.9.2 CEA Ejection Results............................... 35
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TABLE OF CONTENTS (continued) l Page L 4.9.3 CEA Drop Results................................... 36 4.9.3.11 Design Analysis Results.................. 36 4.9.3.2 Post-CFA Drop Restrictions............... 37 4.9.4 Available Scram Reactivity......................... 37 4.9.5 Shutdown Margin Requirements....................... 39 4.10 Pressure Vessel Fluence.................................... 41 4.11 Methodology and Methodology Revisions...................... 41 5.0 SAFETY ANALYSIS.................................................. 71 5.1 General.................................................... 71 5.1.1 Initial Operating Conditions....................... 71 5.1.2 Core Power Distributions........................... 72 5.1.3 Reactivity Coefficients............................ 73 5.1.4 Shutdown CEA Characteristics....................... 74 5.1.5 Reactor Protective System Setpoints and Time Delays.................................... 75 5.2 Summary.................................................... 76 5.3 Anticipated Operational Occurrences for Which the RPS Assures No Violation of SAFDLs............................. 77 5.3.1 Control Element Assembly Bank Withdrawal........... 78 5.3.2 Boron Dilution..................................... 79 5.3.2.1 Dilution During Refueling................ 79 5.3.2.2 Dilution During Cold, Transthermal, and Hot Shutdown With the RCS Filled..... 80 5.3.2.3 Dilution During Cold, Transthermal, and Hot Shutdown With Drained RCS Conditions............................... 80 5.3.2.4 Dilution During Hot Standby, Startup, and Power 0poration...................... 81 5.3.2.5 Failure to Borate Prior to Cooldown...... 82 5.3.3 Excess Load Incident............................... 82 5.3.4 Loss of Load Incident.............................. 83 5.3.5 Loss of Feedwater Incident......................... 84
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l TABLE OF CONTENTS (continued)
Page 5.4 Anticipated Operational Occurrences.Which are Dependent on Initial Overpower Margin for Protection Against Violation of SAFDLs........................................ 84 5.4.1 Loss-of-Coolant F1ow............................... 85 5.4.2 Full Length CEA Drop............................... 85 5.5 Postulated Accidents....................................... 87 5.5.1 Steam Line Rupture................................. 88 5.5.2 Steam Generator Tube Rupture....................... 89 5.5.3 Seized Rotor Accident.............................. 89
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5.5.4 CEA Ejection....................................... 90 5.5.5 Loss of Coolant.................................... 90 5.5.5.1 Introduction and Summary................. 90 5.5.5.2 Large Break LOCA Analys is . . . . . . . . . . . . . . . . 91 5.5.5.2.1 Break Spectrum Analysis....... 92 5.5.5.2.2 LOCA Limit Calculations....... 92 5.5.5.2.3 Sensitivity of LOCA Limits to Radial Peaking Factor......,.. 92 5.5.5.3 Small Break LOCA Analyses................ 93 6.0- STARTUP TEST PR0 GRAM............................................. 122 6.1 Low Power Physics Tests.................................... 122 6.2 Power Escalation Tests..................................... 123 6.3 Acceptance Criteria........................................ 123
7.0 CONCLUSION
S...................................................... 126
8.0 REFERENCES
....................................................... 127 l
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l -LIST OF TABLES L
Number Title Page 2.1 Operating History Summary 6
'2.2 Fuel Assembly Types by Cycle 7 3.1 Cycle 11 Assembly Description 15
~ 3. 2 Cycle 11 Core Loading 16 3.3 Mechanical Design Features of Cycle 11 Fuel 17 3.4 Maine Yankee Cycle 11 Centerline and UO2 Me?t Temperature Comparison 18 3.5 Cycle 11 Ratio of Maximum Radial Relative Pin Powers -
Maximum in Types L and M Fuel to Maximum in Core 19 3.6 Cycle 11 Ratio of Maximum Radial Relative Pin Powers -
Ma:timum in Type N Fuel to Maximum in Core 20 3.7 Cycle 11-Ratio of Maximum Radial Relative Pin Powers -
Maximurc. in Type P Fuel to Maximum in Core 21 3.8 Cycle 11 Thermal-Hydraulic Parameters at Full Power 22 4.1- Cycles 3, 10, and 11 Nuclear Characteristics 43 4.2 Cycles 3, 10, and 11 CEA Group Worths at HFP 44 4.3 Cycles 3,10 and 11 Core Average Doppler Defect 45 4.4 cycles 3,10 and 11 Core Average Doppler Coef ficient 46 4.5 Cycles 3, 10 and 11 Moderator Temperature Coefficients 47 4.6 Cycles 3, 10 and 11 ARI Moderator Defect with Worst Stuck CEA 48 4.7 Cycles 3 and 11 Kinetics Parameters 49 4.8 Cycles 10 and 11 CEA Ejection Results from Full Insertions 50 4.9 Cycles 10 and 11 CEA Drop Results at BOC 51 4.10 Cycles 10 and 11 CEA Drop Results at E00 52 4.11 Cycles 10 and 11 Dropped CEA with Power Level Restriction -
Most Limiting Peaking Cases 53
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-LIST OF TABLES (continued)
Number Title Page l.
4.12 Cycle 11 Available Scram Reactivity 54' 4.13 Cycle 11: Required-Scram Reactivity 55 4.14 Cycles'6 through 11 Relative Pressure Vessel Fluence Comparison '56 4.15 Physics Methodology Documentation 57
- 5 .1' Maine Yankee Safety Parameters 94 5.2 Cycle 11 - Incidents Considered 98 5.3. Cycle 11 Safety Analysis - Summary of Results 99 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 11 106 5.7 Nominal Rod Worths to Prevent a Return to Power During a Steam Line Rupture Accident 107 5.8 Cycle 11 CEA Ejection Accident Results 108 5.9 Comparison of Thermal Margin for Limiting Cycle 11 Power Distributions to FSAR Design Power Distribution 109 5.10 Reactor Protective System Trips Assumed in the Cycle 11 Safety Analysis 110 5.11 Available Shutdown Margin Assumed in Cycle 11 Safety Analysis 111 5.12 Break Spectrum Analysis Results 112 5.13 Maine Yanxee Cycle 11 Large Break LOCA Analysis Results 113 6.1 Cycle 11 Startup Test Acceptance Criteria 125 i
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l I LIST OF FIGURES
.1 Number. Title Page 3.1 Cycle 11 Burnable Poison Shim Assembly Locations 24 3.2 Cycle 11 Assembly Loading Pattern- 25 3'.3 - Cycle 11 Calculated Assembly Exposures at'BOC 26 3.4 Cycle 10 Burnup Distribution by Assembly near 6,000 mwd /Mt 27 3.5' Cycle 11 CEA Group Locations 28 3.6 Cycle 11 BOC Centerline Temperature Versus LHGR 29 3.7 Cycle 11 EOC Centerline Temperature Versus LHGR 30 6.1 Cycle 11-Assembly Relative Power Densities BOC (500 MWD /MT), HFP, ARO 59 4.2 Cycle 11 Assembly Relative Power Densities MOC (6,000 MWD /MT), HFP, ARO 60 4.3 Cycle 11 Assembly Relative Power Densities E0C (14,000 MWD /MT), HFP, ARO 61 4.4 Cycle 11 Assembly Relative Power Densities BOC (500 MWD /MT), HFP, CEA Bank 5 Inserted 62 4.5 Cycle 11 Assembly Relative Power Densities MOC (6,000 MWD /MT), HFP, CEA Bank 5 Inserted 63 4.6 Cycle 11 Assembly Relative Power Densities EOC (14,000 MWD /MT), HFP, CEA Bank 5 Inserted 64 6.7 Cycle 11 Allowable Unrodded Radial Peak Versus Cycle Average Burnup 65 4.8 Cycle 11 Moderator Temperature Coefficient Upper Limits Versus Power Level 66 4.9 Cycle 11 Power Dependent Insertion Limit (PDIL) for CEAs 67 4.10 Cycles 10 and 11 Maximum Peaking Versus Dropped CEA Worth from Specified Power Levels 68
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shj LIST OF FIGURES (continued) l L Number - Title Page 4.11 Cycle 11 Shutdown Margin Equation:and. Required Scram
' Reactivity 69 4.12. Cycle 11 Required Shutdown Margin Versus RCS Boron Concentration 70 5.1 Cycle 11 Allowable 3 Loop' Steady-State Coolant Conditions 114
-5.2 Design Power Distributions 115 5.3 Normalized Reactivity Worth Versus Position Assumed in BOC CEA Ejection Analysis 116 5.4 Normalized Reactivity Worth Versus Position Assumed in EOC CEA Ejection' Analysis 117 5.5 TM/LP Trip.Setpoint (YI Versus A1 ) 118 5.6 TM/LP Trip Setpoint Part 2 119 5.7 Symmetric Offset Trip Function 120
.5.8 Linear Heat Generation Rate (LHGR) Limits Versus Core Height- 121
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1.0 INTRODUCTION
This report provides justification for the-operation of Maine Yankee during the next fuel cycle, Cycle 11. The Cycle 11 refueling will involve the
- 41. charge of 73 assemblies and the insertion of 72 new fuel assemblies and one burned Type M assembly from Cycle 9. The new fuel assemblies (designated Batch Q) are being fabricated by Combustion Engineering (CE) and are similar in design to the CE Batch N and Batch P fuel provided for Cycles 9 and 10, respectively. The Types L and M fuel remaining f rom Cycles 7, 8, and 9 were provided 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 discussed in Section'3.2.1 and Section 3.2.3 of (1).
The proposed operating conditions for Cycle 11 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 are consistent with the "Stretch Power" conditions proposed in (2) and also bound expected coastdown conditions. The allowable maximum indicated core inlet temperature of 552 F remains unchanged from Cycle 10. However, the inlet temperature programming, which defined the allowable maximum indicated inlet temperature as a function of core power, is eliminated. This provides a flexible core inlet *emperature range from 500 F to 552 F# which is accounted for in the Cycle 11 safety analyses.
6341R/23.193
,, This report contains sections dealing with the fuel mechanical, thermal-hydraulic, physics and safety analysis aspects of the operation of Cycle 11. A_ description of the Startup Test Program is also included. Except as noted, the methods used in these analyses are in accordance with'those I_ described in (4-14). These methods have been approved by the NRC for use on Maine Yankee in (15-18). Methods used in safety-related analyses for the fuel mechanical design evaluations are based on the Combustion Engineeric rnd Exxon Nuclear generic models which'have received prior approval by the NRC.
The significant features of Cycle 11 are:
- 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 3.50 to 3.70 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 eleven cycles designated as 1, 1A, and 2 through 10. The significant operating conditions 1
and durations of the cycles are defined in Table 2.1. The fue1~ assembly types loaded by cycle are given in Table 2.2.
2.1 Cycles 1 and 1A 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 terminated 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 Replacament 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 prevent creep collapse of the fuel rod cladding and improve gap heat transfer. The replacement fuel assemblies performed successfully during Cycle 1A.
2.2 Cycle 2 Cycle 2 consisted entirely of fresh assemblies designated as Core 2 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.
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2.3 Cycles 3 and 4 Cycle 3 consisted of fresh. fuel assemblics 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 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.
9 2.4 Cycles 5 and 6 Cycles 5 and 6 consisted of fuel assemblies of the Core 2 CE design and fresh assemblies designed by 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 enrichment 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 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 an increase in enrichment to 3.30 w/o U-235. The Cycle 7 design was the first low-leakage, low-fluence core deaign. Minor fuel design changes for the Types L and M fuel were discussed in (12). All fuel performed successfully during Cycles 7 and 8.
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2.6 Cycle 9
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l Cycle 9 consisted nf fresh fuel assemblies designed by CE (Type N) and
!- reinserted cesemblies designed by ENC (Types L and M) at 3.30.w/o U-235 enrichment. One Type E assembly with Cycle 2 exposure was inserted in the core center location.. The Cycle 9 design was a continuation of the low leakage, low fluence core decigns. The fuel design of the Type N fuel was discussed in (21). A small number of leaking fuel rods were discovered during Cycle 9 nperation.
2.7 Cycle 10 Cycle 10 consisted o' fresh (Type P) and reinserted (Type N) fuel assemblies designed by CE, and reinserted assemblies designed by ENC (Type L and M). The fresh Type P fuel had a 3.50 w/o U-235 enrichment. One Type E assembly with Cycle 2 exposure was inserted in the core center location. The Cycle 10 design was a continuation of the low leakage, low fluence core designs. The fuel design of the Type P fuel was discussed in (24). A small fuel leak is suspected from the iodine concentrations observed during Cycle 10 operation and is not considered significant.
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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 50-80(1) 10367 1A 10/12/74 2440 80(1) 4500 2 '6/29/75 2440 100 17395 3 6/17/77 2630(2) 93 11075 4 8/23/78 2630 97(3) 10496 5 3/17/80 2630 97(3) 10796 6 7/20/81 2630 97(3) 11580 7 12/12/62 2630 100 12466 8 6/20/84 2630 100 12458 9 10/25/85 2630 100 14362 10 6/18/87 2630 100 13000(4)
(1) Power decrease and primary system pressure decrease to 1800-2000 psia due to leaking fuel.
(2) Licensed power increase from 2440 MWt/2100 psia operation to 2630 MWt/2250 psia operation.
1 (3) Power restriction due to secondary plant limitations (turbine).
(4) Estimated.
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TABLE 2.2 MAINE YANKEE
. FUEL ASSEMBLY TYPES BY CYCLE-Assembly Fuel Enrichment Mechanical Number of Fuel Assemblies by Cycle L Type (w/o U '35)
. Design Type 1 1A 2 3 4 5- 6 7 8 9 10 A 2.01 CE-Core 1 -69 57 - - - - - - - - -
B 2.40 CE-Core 1 80 24 - - - - - - - - -
C 2.95 CE-Core 1 68 64 - - - - - - - - -
RF 2.33 CE-RF -
2 - - - - - - - - -
RF 1.93 CE-RF -
70 -
65 - - - - - - -
D 1. 95 CE-Core 2 - -
69 - - - - - - - -
E 2.52 CE-Core 2 - -
80 12 61 1 1 1 1 1 1 F 2.90 CE-Core 2 - -
68 68 12 - - - - - -
G 2.73 CE-Cot e 2 - - - -
32 32 32 - - - - -
H 3.03 CE-Core 2 - - -
40 40 40 - - - - -
I 3.03 CE-Core 2 - - - -
72 72 72 - - - -
J 3.00 ENC - - - - -
72 72 72 - - -
K 3.00 ENC - - - - - -
72 72 72 - -
L 3.30 ENC - - - - - - -
72 72 72 8 M 3.30 ENC - - - - - - - -
72 72 64 N 3.30 CE-Core 2 - - - - - - - - -
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l 3.0 RELOAD CORE DESIGN 3.1 General Description 3.1.1 Core Fuel Loading The core of Maine Yankee Cycle 11 consists of 217 fuel assemblies of the type and quantity detailed in Table 3.1. Assembly Type L is of the ENC mechanical design and has irradiation exposure from Cycles 7, 8, 9, and 10.
Assembly Type M is also ENC design fuel and was irradiated in Cycles 8 and 9.-
Assembly Types N and P were introduced in Cycles 9 and 10, respectively, and assembly Type Q is fresh fuel to be introduced in Cycle 11. Assembly Types N.
P, and Q are fabricated by CE and are designated Core 2 design fuel. The Type N fuel initial enrichment and shim loading are equal to those of the Types L and M fuel. The Type P initial fuel enrichment was increased to 3.50 w/o U-235 and the initial shim loading is 31.4 milligrams of B-10 per inch of active shim length. The Type Q fuel enrichment has been increased to 3.70 w/o U-235 and the shim loading is the same as the Type P fuel. The total number of fuel rods by assembly type for Cycle 11 is given in Table 3.1, and 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 11. 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 in Figure 3.1.
5 All burnable poison shims are composed of B4 C in A123 0 . The ENC design shim irradiation integrity has been demonstrated in the Types J, K, L, and M assemblies during Cycles 5 through 10. The CE design shim irradiation integrity has i.een demonstrated in the Types N and P assemblies during Cycles 9 and 10.
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3.1.3 Core Loading Pattern The fuel assembly locations designated for Maine Yankee Cycle 11 are
! - given for the first quadrant in Figure 3.2. They are given relative to the previous locations of the Type M assembly in Cycle 9 and the Types L N, and P assemblies in Cycle 10. The appropriate rotation index relative to the previous assembly position in the core is also given for each assembly. The loading ar.d rotations of the other quadrants are such that mirror symmetry exists with respect to the quadrant boundary lines.
The Cycle 11 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 cor' edge. The Cycle 11 loading pattern is similar to the Cycles 7 through 10 low-leakage loading patterns.
The benefits of such a core design are:
- 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; 4
- 3) Preferred fuel rod power and exposure histories for fuel performance and mechanical 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 11 fuel assemblies at Beginning-of-Cycle (BOC) is given in Figure 3.3. The exposures are based on an expected cycle length of 13,000 mwd /Mt for Cycle 10 and the achieved cycle
' length of 14,362 mwd /Mt for Cycle 9. Table 3.2 gives BOC average exposures by
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fuel' type. The Cycle 11 BOC average exposure for the core is approximately 15,500 mwd /Mt.
l l .The exposure history of the assemblies utilized in the analysis is l demonstrated to be accurate by comparison with incore detector data.
Figure 3.4 is a comparison of predicted and actual burnup assembly data at approximately the middle of Cycle 10. 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 11 is unchanged from Cycle 10. 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 SA, which are 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.
As in Cycles 7 through 10, Subgroups SA and 5B are independently moveable and not directly connected as a single CEA bank. As such, their 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 regulating subgroups subject to the CEA group insertion limits, as discussed i
in Section 4.9.1.
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3.2 Fuel Sy: tem Design 3.2.1 Fuel Mechanical Design The fuel assemblies, described in (21, 23, 24, 53), have been designed to 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 in Cycle 11.
The detailed fuel assembly descriptions and mechanical design criteria for the recycled reload fuel have been described in (21, 23, 24, 25, 26). The fresh reload fuel is supplied by CE and is similar to the previous cycle reload fuel.
The ENC supplied fuel Batches L and M will achieve exposures higher than the original design analysis and have been reanalyzed to demonstrate compliance with the appropriate design criteria at thesc 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).
3.2.2 Fuel Thermal Analysis The thermal effects analysis encompassed a rtudy of fuel rod response as a function of the detailed cycle exposure and power. The fuel rod types and 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
(
l same as that employed in the previous cycle (54).
I Fuel thermal calculations were performed using the CAPEX (29) computer program. The CAPEX code calculates pellet-to-clad ge- conductance from a I combication of theoretical and empirical models which predict fuel and l
l cladding thermal expansion, fission gas release, pellet swelling, pellet densification, pellet cracking, and fuel and cladding thermal conductivity.
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. 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 -ef tect of LHGR on fuel temperature at End-of-Cycle (E0C) conditions. Table 3.4-lists UO meltinS 2 temperature and centerline temperature for the rods of interest at selected points in life and power levels.
The result of the fuel performance calculations indicates that the thermal performance is similar to that reported for the previous cycle (54).
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:
a) The fresh fuel contains the core-wide maximum power pin throughout the cycle, and 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 tre fresh fuel batch.
The margins to SAFDLs demonstrated by this use of the ratios in Tables 3.5, 3.6, and 3.7 are maintained under transient conditions, such as CEA drop or withdrawal.
3.2.3 Thermal-Hydraulic Design I
Steady-state and transient DNBR analysis of the Cycle 11 core havc been f 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 11 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 fuel assemblies.
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A COBRA-IIIC model was used to determine hot assembly enthalpy rise flow factors. This model explicitly represents each fuel assembly in the one-eighth (1/8) core in the specific location it will reside for Cycle 11 operation, and accounts for the differences in hydraulic characteristics l
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 hot assembly enthalpy rise flow factor was calculated to be 0.972 for the CE fuel assemblies with bottom peaked power cistributions and 0.990 for the CE fuel assemblies with top peaked power distributions. A 0.950 enthalpy rise flow factor is applied to all ENC assemblies due to higher spacer loss coefficients relative to the CE fuel. These factors are applied to the inlet mass velocity in the hot channel model in predicting DNB performance.
The potential effects of fuel rod bow on thermal-hydraulic perfcrmance has also been evaluated for Cycle 11 operation. Using the channel closure correlation in (33), the maximum channel gap closure due to fuel rod bowing for the CE fuel as embly with the highest burnup during Cycle 11, a Type N assembly, was calculated to be 24.8%. 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 11 is less than 1
33%. Therefore, no penalty is required for fuel rod bow considerations since l the channel closure resulting from rod pitch, bow and clad diameter l
l considerations for any ENC fuel during Cycle 11 will be less than 50%.
I 1
63419/23.193
Table 3.8 contains a list of the pertinent thermal-hydraulic design parameters used for both safety analysis and for generating Reactor Protection l System (RPS) setpoint information. The list also includes the corresponding thermal-hydraulic parameters for Cycle 3 (36) and Cycle 10 (54) for comparison.
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l l 6341R/23.193
l TABLE 3.1 Maine Yankee Cycle 11 Assembly Description Cycles Number Init*al j j of of Fuel Initial Number of mg B-10 Number of Total Total j
! Assembly Exposure Rods per w/o U-235 Shim Locations per inch Assemblies Shim Fuel j
.Desianation History Asiembly Fuel per Assembly _ in Shims in Core Locations Rods l l
L-0 7,8,9,10 176 3.30 0 - 8 0 1,40S M-8 8,9 168 3.30 8 23.8 1 8 16d N-0 9,10 176 3.30 0 - 4 0 704 l N-4 9,10 172 3.30 4 23.8 .24 96 4,128 l N-6. 9,10 168 3.30 8 23.8 36 288 6,048 l P-0 10 176 3.50 0 - 28 0 4,925 P-4 10 172 3.50 4 31.4 20 80 3.440 P-8 10 168 3.50 8 31.4 24 192 4,032 Q-0 Fresh 176 3.70 0 -
28 0 4,928 Fresh 172 3.70 4 31.4 36 144 6,192 Q-4 Fresh 168 3.70 8 31.4 8 64 1,344 Q-8 Core Totals 217 872 37,320 l
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1 6341R/23.193 i
1ABLE 3.2 Maine Yankee Cycle 11 Core Loading Uranium per Uranium Exposure Assembly Number of Assembly Total at BOC*
Type Assemblies (kg U) (kg U) (mwd /Mt) l L-0 8 381.1 3,049 30,554 M-8 1 363.8 364 32,414 N-0 4 388.7 1,555 28,914 N-4 24 379.9 9,117 29,594 N-8 36 371.0 13.356 32,375 P-0 28 388.7 10,884 13,465 P-4 20 379.9 7,597 16,151 P-8 24 371.0 8.905 17,221 Q-0 28 388.7 10,884 0 Q-4 36 379.9 13,675 0 Q-8 8 371.0 2,968 0 82,354 15,479**
- Based on End-of-Cycle 10 at 13,000 mwd /Mt
- Core average exposure is based on a uranium-mass weighting of the batch exposures l
1 63L1R/23.193 I
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l TABLE 3.3 Mechanical Design Features of Cycle 11 Fuel
' Types L and M Type N Types P and Q Fuel Vendor ENC CE CE Fuel Assembly overall length 156.718* 156.718 156.718 Spacer grid' size (maximum square) 8.115 8.115 8.115 Number of zircalcy grids 0 8 8 Number of inconel grids 0 1 1 Number of bimetallic grids 9 0 0 Fuel rod growth clearance 1.300 min. 1.600 1.600 Fuel Rod Active fuel length 136.7 136.7 136.7 Plenum length 8.8 8.375 8.375
-Clad OD 0.440 0.440 0.440 Clad ID 0.378 0.384 J.384 Clad wall thickness 0.031 0.028 0.028 Pellet OD 0.370 0.3765 0.3765 Pellet length (37) 0.43u 0.450 Dish depth 0.008 0.021 0.021 Clad material Zr-4 Zr-4 Zr-4 Initial pellet density 94.0% 94.75% 94.75%
Initial pressure (37) (21) (24,53)
Poison Rods Overall rod length 146.500 146.629 146.322 Clad OD 0.440 0.440 0.440 Clad ID 0.378 0.384 0.384 Clad wall thickness 0.031 0.028 0.028 Pellet OD 0.353 0.362 0.362 Clad material Zr Zr-4 Zr-4
- All length dimensions are in inches CE - Combustion Engineering
( ENC - Exxon Nuclear Corporation l
6341R/23.193
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TABLE 3.4 l-Maine Yankee Cycle 11 Centerline and UO2 Melt Temperature Comparison Melt Centerline Fuel Temperature LHGR Temperature Fuel Type Vendor (*F) (kW/ft) (#T)
BOC M -ENC 4835 19 4455 20 4668 20.8* 4832
^
N CE 4827 19 4360 20 4573 j 21 4777 21.2* 4816 P CE 4931 19 4237 20 4452 21 4657-22 4853 22.3* 4910 Q CE# 5049 19 4420 20 4574 21 4720 22 4859 23 5009 23.2* 5043 EOC M ENC 4756 19 4500 20 4713
.'0.2* 4754 N CE 4741 19 4472 20 4683 20.2* 4725 P CE 4823 19 4368 20 4581 21 4784 21.1* 4804 Q CE 4922 19 4245 20 4459 21 4664 22 4860 22.2* 4898
^
TABLE 3.5 Maine Yankee Cycle 11 Ratio cf Maximum Radial Relative Pin Powers Maximum in Types L and M Fuel to Maximum in Core Rodded Condition HFP, Equilibrium Conditions of ARO Ratio of Maximum Radial Relative Fin Powers
, Regulating BOC BOC MOC EOC Banks Inser:ed 50 mwd /Mt- 500 mwd /Mt 6K mwd /Mt 14K mwd /Mt ARO 0.573 0.614 0.621 0.620 B:nk 5 0.433 0.470 0.478 0.472 Banks 5 + 4 0.173 0.185 0.192' O.196
~
B:nks 5 + 4.+ 3 0.168 0.182 0.192 0.189 Banks 5 + 4 + 3 + 2 0.104 0.114 0.120- 0.117 Binks 5 + 4 + 3 + 2 + 1 0.182 0.197 0.200 0.196 6341R/23.193
TABLE 3.6 Maine Yankee Cycle'll Ratio of Maximum Radial Relative Pin Powers Maximun. in Type N Fuel to Maximum in Core Rodded Condition.
HFP, Equilibrium Conditions of ARO Ratio of Maximu- Radial Relative Pin Powers Regulating 300 BOC MOC EOC Banks Inserted 50 mwd /Mt 500 MWdLMt 6K mwd /Mt 14K mwd /Mt ARO 0.646 0.668 0.679 0.682 B:nk 5 :0.585 0.633 0.665 0.657 Banks 5 + 4 0.530 0.545 0.591 0.646 Banks 5 + 4 + 3 0.500 0.513 0.558 0.601 Banks 5 + 4 + 3 + 2 0.493 0.498 0.524 0.571 Banks 5 + 4 + 3 + 2 + 1 0.535 0.533 0.545 0.582 l
6341R/23.193 I
TABLE 13.7
- y,..
' Maine Yankee Cycle 11 Ratio of Maximum Radial Relative Pin Powers Maximum in Type P> Fuel to Maximum in Core Rodded Condition HFP, Equilibrium Conditions of ARO Ratio of Maximum Radial Relative Pin Powers Regulating BOC BOC MOC EOC Banks Inserted 50 mwd /Mt 500 mwd /Mt 6K mwd /Mt. 14K mwd /Mt ARO 0.839 0.889 0.856. 0.820 B:nk 5 0.860 0.934 0.917 0.852 Btnks 5 + 4 0.844 0.850 0.847 0.846 Bruks 5 + 4 + 3 0.860 0.865 0.870 0.851 Banks 5 + 4 + 3 + 2 0.847 0.852 0.853 O.861 Binks 5 + 4 + 3 + 2 + 1 0.823 0.818 0.802 0.810 Q
3 6341R/23.193
- y TABLE 3.8 Maine Yankee Cycle 11 Thermal-Hydraulic Parameters at Full Power Ceneral Characteristics Units Cycle 3 Cycle 10 Cycle 11 :l
.1 1
Total Heat Output MWt 2630 2630 2630 106 Btu /hr 8976 8976 8976 .;
Fraction of Heat Generated in Fuel Rod 0.975 0.975 0.975 Pressure Nominal psig 2235 2235 2235 o l 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 130.7 131.9-130.15 131.9-130.15 Hydraulic Diameter (nominal channel) ft 0.044 0.044 0.044
. Average Mass Velocity 106 lb/hr-ft2 2.444 2.46-2.433 2.46-2.433
! Pressure Drop Across Core i (design flow) psi 9.7 9.99 9.82 Total Pressure Drop Across Vessel (Based on nominal dimensions and design flow) psi 32.4 32.6 32.4 Core Average Heat Flux
- Btu /hr-ft2 178,742 180,211 179,245 Total Heat Transfer Area
- ft2 48,978 48,309 48,826 Film Coefficient at Average Conditions Btu /hr-ft20F 5.640 5,698 5,700 Maximum Clad Surface Temperature OF 656 657 657 l
) Average Film Temperature Difference OF 31.7 31.5 31.5 i Average Linear Heat Rate of Rod
- kW/ft 6.03 6.11 6.05 Average Core Enthalpy Rise Btu /lb 68.7 68.9 68.9 i
e i
- 6341R/23.193 c ,- n,
i TABLE 3.8-(Continued)
-i Maine Yankee Cycle 11 13ermal-Hydraulic Parameters at Full' Power i
Cycle 3 Cycle 10 Cycle 11 i
Calculational Factors CE ENC CE . ENC CE i
Engineering Heat Flux Factor 1.03 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 Flow Factors:
Inlet Plenum Non-Uniform Distribution 1.05 1.05 1.05 1.05 1.05
! Rod Pitch, Bowing and Clad Diameter 1.065 1.00 1.065 1.00 1.065 1
i 4
i I
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- Allows for axial shrinkage due to fuel densification.
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i 1 6341R/23.193 i
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__ _ ~ __ . . . . _ _ -. - - . _ _
j FIGURE 3.1 MAINE YAEKEE CYCLE 11 BURNABLE POICON SEIM ASSEMBLY LOCAT10NS A A B B l
C C l
l D D r E X X F F G C H H I I J J X X K K L L M M N N 1 2 3 4 5 6 7 8 9 10 11 12 13 M 1 2 3 4 56 7 8 91011121314 O SHIM ASSEMBLY (L-0 N-0, P-0, Q-0) 4 SHIM ASSEMBLY (N-4, P-4, Q-4)
A B
e X X D
E r X X 0
H i X X J
K L
X X M
N i 1 2 d 4 56 7 8 910111213 %
8 SHIM ASST.;iBLY (M-8, N-8, P-8, Q-8)
FIGURE 3.2 MAINE YANKEE CYCLE 11 ASSEMBLY LOADING PATTERN RJEL IYP.I M M-8 62 Cycle 11 Location 42 Cycle 9 Location 0 Rotation index*
L L-0 2 Cycle 11 Locotion N-8 1 L-0 2 10 2 N 2 Cycle 10 Location 2 0 P O Rotation index' Q Q-0 4 Cycle 11 Location N-8 3 Q-0 4 Q-0 5 Q-4 6 Q-4 7 12 I I N-8 8 Q-0 9 P-0 10 P-4 11 P-4 12 N-4 13 28 9 19 46 43 0 1 0 1 0 N-8 14 Q-0 15 P-8 16 Q-8 T7 N-8 18 Q-4 19 N-4 20 '
26 34 51 33 0 2 2 1 N-8 21 0 -0 22 P-8 23 Pr-0 24 N-4 25 Q-4 26 P-8 27 N-0 28 l 40 15 11 17 36 47 3 2 3 1 0 3
0 -0 29 P-0 30 0 - 8 31 N-4 32 Q-4 33 N-4 34 P-0 35 P-4 36 22 38 35 4 53 ;
3 1 1 1 0i Q-0 37 P-4 38 N-8 39 Q-4 40 N-4 41 P-8 42 P-0 43 P-8 44 l 48 41 50 13 5 42 '
0 3 0 0 2 N-8 45 0 30 i 2 Q-4 46 P-4 47 Q-4 48 P-8 49 P-0 50 P-0 51 N-B 52 Q-4 53 6 31 29 37 24 l L-0 54 3 3 3 0 3 l 54 M-8 62 0 Q-4 55 N-4 56 N-4 57 N-0 58 P-4 59 P-8 60 Q-4 61 43 59 61 42 42 51 2 0 0 2 0 0
- Clockwise rnuttiple of 90 degrees relative to previous cycle of insertion indicated.
i FIGURE 3.3 MRINE YANKEE CYCLE 11 CALCULATED RSSEMBLY EXPOSURES BOC (O MWD /MT)
ASSEMBLY TYPE AND CORE POSITION. . . N-8 1 L-0 2 855EMBLY AVERAGE EXPO 5URE. . . .
- 32220 30554 N-8 3 0-0 4 0-0 5 0-4 6 0-4 7 32226 0 0 0 O li-8 8 0-0 9 P-0 10 P-4 11 P-4 12 N-4 13 32108 0 14060 17479 14349 30784 N-8 14 0-0 15 P 9 16 0-8 17 N-8 18 0-4 19 N-4 20 0 171's 0 33496 0 30784 32108 0-0 22 P-8 23 P-0 24 N-4 25 0-4 26 P-8 27 N-0 28 N-B 21 17174 14279 29140 0 17152 28914 32226 0 P-0 30 0-8 31 N-4 32 0-4 33 N-4 34 P-0 35 P-4 36 0-0 29 0 29140 0 28858 11637 17102 0 14060 P-4 38 N-8 39 0-4 40 N-4 41 P-8 42 P-0 43 P-8 44 0-0 37 E 96 0 28858 17075 14289 17596 0 17479 N-8 45 32220 P-4 47 0-4 48 P-8 49 P-0 50 P-0 51 Na 52 0-4 53 0-4 46 0 14349 0 17152 11637 14289 31a/I O L-0 54 30554 N-4 57 N-0 58 P-4 59 P-8 60 0-4 61 M-8 62 0-4 55 N-4 56 28914 17102 17596 0 32414 0 30784 30784 l
l l
FIGURE 3.4 MAINE YANKEE CYCLE 10 BURNUP DISTRIBUTION BY ASSEMBLY INCR VS PREDICTED 5949 MWD /MT CYCLE EXPOSURE M5SEMBLY TYPE RNO INCR LOCRTION * * * * * *
- M-8 8 L-0 21 INCR R5SEMBLY EXPOSURE (MWO/MT) * ***** ***** 33768 27688 PREDICTED RS5EMBLY EXPOSURE (MWO/MT) 33989 27645 PERCENT DIFPERENCE =************ 0.7 -0.2 M-8 15 P-0 31 P-0 11 P-4 25 l M-4 4 33653 5410 6672 6370 34158 33315 5480 6760 6501 34281
-1.0 1.3 1.3 2.1 0.4 M-0 16 P-0 33 N-8 13 N-4 28 N-8 7 P-8 20 w MAXIMUM EXPOSURE 31476 6606 25053 21273 24200 7575 31505 6649 24934 21224 24176 7633 0.1 0.7 -0.5 -0.2 -0.1 0.8 OCTANT LOCATION 3 NERSURED 38007w P-0 34' R-8 14 P-8 30 M-B 10 P-4 24 M-8 3 PREDICTED 38214w 6776 25077 7833 37113 7769 38007w 6779 25139 7807 37147 7942 38214w
% DIPTERENCE 0.5 0.2 -0.3 0.1 2.2 0.5 0.0 N-8 32 H-4 12 P-8 27 M-4 6 N-B 19 23563 29794 7715 32518 23G67 23608 29247 7672 32211 24298 0.2 -1.8 -0.6 -0.9 2.7 h h hg$yf h Dir E fCE N-8 29 24583 N-8 25234 9 N-4 23 20828 N-0 2 21098 E-16 26555 26093 -1.7 24731 25513 20888 21419 L-0 27688 2764C -0.2 0.6 1.1 0.3 1.5 M-0 31476 31505 0.1 M-4 32766 32556 -0.6 P-8 26 N-4 5 M-4 18 i M-8 35296 35302 0.0 7940 22801 35140 N-0 21098 21419 1.5 7887 23060 35264 i i N-4 21634 21724 0.4 -o,7 g,g o,4 !
N-8 24531 24742 0.5 P-0 6307 6365 0.9 M-4 22 P-4 1 P-4 7239 7322 1.2 35441 7917 P-8 7769 7746 -0.3 36430 7726 0.0 -2.4 l CORE 21068 21091 0.1 0.81 E-1617 RBSOLUTE RVERAGE 1.04 26555 l STRNDARD DEVIRTION 26093
-1.7 PREDIC - INCA X 100 '
PERCENT DIFFERENCE i
FIGURE 3.5 MAINE YANKEE CYCLE 11 CEA GROUP LOCATIONS REGULATING SHUT 00HN 00RL N-8 1 L-0 2 CEA GROUPS CEA GROUPS 5 (5A AND 5B) C i B 0-0 0-0 5 0-4 6 0-4 7
! N-8 3 4 C 1 N-8 8 0-0 9 P-0 10 P-4 !! P-4 12 N 4 23 A C 58 N-8 14 0-0 15 P-8 16 0-8 17 N-8 18 0-4 19 N-4 20 5A A 3 N-8 21 0-0 22 P-8 23 P-0 24 N-4 25 0-4 26 P-8 27 N-0 28 A 2 0-0 29 P-0 30 0-8 31 N-4 32 0-4 33 N-4 34 P-0 35 P-4 36 C A 5Bx 0-0 37 P-4 38 N-8 39 0-4 40 N-4 41 P-8 42 P-0 43 P-8 44 0 2 B 4 N-8 45 l
0-4 46 P-4 47 0-4 48 P-8 49 P-0 50 P-0 51 N-8 52 0-4 53 l
1 3 B L-0 54 0-4 55 N-4 56 N-4 57 N-0 58 P-4 59 P-8 60 0-4 61 M-8 62 4 58 5A M NON-SCRR.W13LE CER LOCATIONS (SUBGROUP 58)
MRINE YANKEE CYCLE 11 800 CENTERLINE TEMPERATURE VS LHGR
- 5000 c
O 4000 e .
(f w .
5 .
s' E 3000
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- w 5
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/ [ LEGEND o - TYPE O :/
l o - TYPE P 1 l A - TYPE N i
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+ - TYPE M 1000 . , . .
20 24 8 12 16 4
LINEAR HERT GENERATION RATE [KW/FT)
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FIGURE 3,7 5
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4.0~ PHYSICS ANALYSIS l
4.1 Fuel Management l
Maine Yankee Cycle 11 consists of irradiated and fresh fuel assemblies as described in Section 3.1.1. The core layout is given in Figure 3.2.
Cycle 11 is expected to attain a cycle average full power lifetime of 12,750 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 10, and Cycle 11) are given in Table 4.1. Nominal Hot Full Power (HFP) anJ Hot Zero Power (HZP) conditions are based on ncminal core inlet temperatures of 552 and 532 F, respectively. The Cycles 10 and 11 characteristics differ from those of Cycle 3 based on the following significant changes:
- 1) Increased fuel enrichment;
- 2) Low-leakage fuel management;
- 3) Increased use of burnable poison shims; and
- 4) Increased reactivity worth resulting from the reconfiguration of CEA Bank 5-The impact of these changes on the physics characteristics are discussed in the following sections.
4.3 Power Distributions Assembly relative power densities for Cycle 11 at HFP, equilibrium xenon conditions are presented for unrodded and rodded (CEA Bank 5 inserted) configurations at Beginning, Middle, and End-of-Cycle (BOC, MOC, EOC).
Figure 3.5 shows the locations of the CEA groups.
' 6341R/23.193 1
1
The unrodded power distributions at BOC (500 mwd /Mt), MOC (6000 mwd /Mt) and EOC (14,000 mwd /Mt) are presented in Figures 4.1 through 4.3. The rodded (CEA Bank 5 inserted) power distributions at B00 MOC and E0C are presented in Figures 4.4 through 4.6. The unrodded and rodded radial peakings are a maximum of approximately 1% above the Cycle 10 radial peakings. -
The allowable unrodded radial peaking (with uncertainties) versus exposure for Cycle 11 is included in the plant Technical Specifications for the purpos_e of comparison to ceasured 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 syurnetric 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 Croup 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 3,10, and 11. The CEA Bank 5 modifications for Cycle 7 and subsequent cycles 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 for Cycle 11 are decreased relative to those of Cycle 10. This is primarily due to the movement of four fresh fuel assembly locations from rodded (Locations 13 and 56) to unrodded (Locations 7 and 55),
I as shown in Figure 3.5. Reduced CEA worths also result from increased core enrichment and higher core average exposure. CEA group reactivity worths are verified by the startup test program and the associated acceptance criteria.
l
' 4.5 Doppler Reactivity Coefficients and Defects l
The fuel temperature, or Doppler, components of reactivity are presented in Tables 4.3 and 4.4 for nominal conditions in Cycles 3, 10, and
- 11. The total core average Doppler defect from 4,000 F is given in 6341R/23.193
[
l Table 4.3 and the core average Doppler coefficient in Table 4.4. The values l in Cycles 3, 10, and 11.are similar. Uncertainties of 25% are conservatively applied to the coefficient and defect values prior to transient analysis applications, except as noted below.
The CEA ejection methodology utilizes the core average unrodded Doppler defects f rom Table 4.3 and a Doppler weighting factor technique, as described in detail in (13). An uncertainty of 15% is applied prior to transient analysis application of the Doppler defects for CEA ejection.
4.6 Moderator Reactivity Coefficients and Defects The Moderator Temperature Coefficient (MTC) at nominal operating HFP and HZP critical boron conditions are presented in Table 4.5 for Cycles 3, 10, and 11. Relative to Cycle 10, the Cycle 11 MTCs at BOC are slightly more positive, primarily due to higher critical boron concentration resulting from more excess reactivity in the core. The end of Cycle 11 MTCs are more negative than Cycle 10's due to the higher average enrichment and the
, increased core average exposure. An uncertainty of 1 0.5 x 10 delta ehe!"F s eens& vatively applied to calculated MTC values, or, if more
, limiting, the MTC technical Specification limits are used for transient analysis. The startup test program demonstrates the validity of such values.
The Moderator Density Defect (MDD) 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 11 are consistent with the LOCA MDD curve and MTC assumptions used in the safety analysis. These limits are provided in Figure 4.8. The startup test program demonstrates that these limits will not be exceeded.
The moderator defect appropriate to the scranned (ARI) less worst stuck CEA configuration is given in Table 4.6 for Cycles 3, 10, and 11. This defect curve yields a conservative moderator reactivity increase versus temperature 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 I
at BOC, high soluble boron and E00, no soluble boron conditions, since the 6341R/23.193 i
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moderator defect is strongly dependent upon soluble boron concentration. The moderator defects for Cycle 11 are increased relative to Cycle 10 due to the placement of fresh fuel nearer the core periphery and the increased core average enrichment and exposure. 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 moderator redistribution effects are an additional reactivity component.
These uncertainties are unchanged for Cycle 11.
4.7 Soluble Boron and Burnable Poison Reactivity Effects The soluble boron and burnable poison shim reactitity effects are shown in Table 4.1 for Cycles 3, 10, and 11. The critical boron concentrations for Cycle 11 at BOC are higher than those of Cycle 10, due primarily to more excess reactivity in the core. There are fewer burnable poison pins in Cycle 11, and a smaller burnable poison pin reactivity worth at BOC. The inverse boron worths for Cycles 10 and 11 reflect the different soluble boron levels and core characteristics.
4.8 Kinetics Parameters The total delayed neutron fractions and prompt neutron generation time for Cycles 3, 10, and 11 are presented in Table 4.1. The values are comparable and the differences reflect the effects of core average exposure and power weighting. Table 4.7 details the delayed neutron fractions and lifetimes by delayed neutron group for Cycles 3 and 11 at HFP, All-Rods-Out (ARO) conditions. Kinetics parameters for HFP and HZP conditions, both unrodded and rodded, are calculated for appropriate application in transient analysis cases and a 10% uncertainty is applied in a conservative manner.
6341R/23.193
4.9 Safety-Related Characteristics ')
4.9.1 CEA Group Insertion Limits The CEA group insertion limits are given in the Technical
' Specifications and Figure 4.9. The Power-Dependent Insertion Limit (PDIL) for-r CEAs provides for sufficient available scram reactivity at all power levels and times-in-cycle-life. It also specifies the allowable CEA configurations for analysis of dropped, ejected, and withdrawn CEAs. The CEA group insertion limits have been changed'for Cycle 11. These limits allow for:
o A 15% insertion of Bank 5 at HFP conditions, an increase from 10%
insertion in Cycle 10. ,
o 100% of Bank 5 inserted and 40% of Bank 4 inserted at HZP condicions. This limit is more restrictive than the Cycle 10 limit which allowed all of Banks 5, 90% of Bank 4, and 30% of Bank 3 inserted at HZP conditions.
The Cycle 11 PDIL allows more operational flexibility for symmetric ,
offset control at HFP, while accommodating the removal of core inlet reference i temperature requirements and reducing the effects of CEA ejections at HZP.
r 1
The allowable CEA insertion is determined from the actual operating power level. For Cycle 11, the Technical Specification limiting the reference power level as a function of nominal cold leg temperature is removed.
Therefore, the calculation of required scram reactivity includes an evaluation of the impact of operating between the Technical Specification minimum and ;
maximum nominal cold leg temperatures of 500 F and 552 F, respectively.
i 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 E0C are shown in Table 4.8 for Cycles 10 and 11. HFP and HZP conditions are considered for these comparable l
CEA insertion cases. No credit is taken for feedback effects in these calculations. These calculations assume full CEA insertion of CEA Bank 5 at 6341R/23.193 1
1 l
HFP and CEA Banks 5 + 4 at HZP, which are conservative relative to the ellowable insertions at these power levels, given by the insertion limits of Section 4.9.1. The Cycle 11 values are generally less than Cycle 10, due primarily to the movement of the four fresh fuel assembly locations, as discussed in Section 4.4.
'4.9.3 CEA Drop Results 4,9.3.1 Design Analysis Results The calculated worths of the most limiting droppeu CEAs for Cycles 10 end 11, 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 at-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, and Banks 5 + 4 are those considered for Cycle 11 based on conservatively assumed CEA insertion limits with power level. CEA drops frcm 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 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 4.10. The resulting peaking increase is generally slightly less than Cycle 10, 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 4.10 are combined with the most limiting radial and axial pea' zing allowed by the symnetric offset limits to obtain total peaking for the given power level. This peaking, increased by 10% for uncertainties, is accommodated in the safety analysis.
6341R/23.193
4.9.3.2 Post-CEA Drop Restrictions 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 to determine the required rate of power reduction. The results indicated that the following actions are required to maintain the core within the limits of the design analysis following a dropped CEA:
- 1) Decrease thermal power by at least 10% of rated power within one-half hour;
- 2) Decrease thermal power by at least 20% of rated power within one hour;
- 3) Maintain thermal power at or below this reduced power level; and
- 4) Limit CEA insertion to the maximum allowable insertion level corresponding to the predrop thermal power.
The power reductions described above assure that proper limits are maintained for operation up to four hours post-drop. The plant Technical Specifications reflect these restrictions. Similar calculations were performed for Cycle 11 to quantify the peaking increases under these power level restrictions for proper incorporation in the safety analysis.
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 10 and 11. 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 and EOC is tabulated in Table 4.12. Allowances for the '.orst stuck CEA and the power dependent insertion limit for CEAs are included. The CEA programming allowance corresponds to the loss in available scram reactivity 6341R/23.193
due to movement of all CEAs a maximum of 3 inches (4 steps) into the active core.
The available scram reactivity with uncertainties at EOC is more for Cycle 11 by 0.27% delta rho at HFP and 1.11% delta rho at HZP conditions relative to Cycle 10. This is generally due to the decreased stuck CEA worths, which more than offset r. lightly lower scrammablo CEA worths. At HZP, the PDIL restrictions for Cycle 11 eliminate Bank 3 inst r clon, thus reducing the PDIL CEA worth penalty and significantly increasing the available sceam reactivity.
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 requited scram reactivity at HZP must be sufficient to prevent a return-to-criticality 1011owing the most limiting steam line rupture event from HZP. It also must be greater than assumed in other safety analyses from HZP. The available scram reactivity at HZP, from Table 4.13, must be greater than the required scram reactivity at HZP.
In addition, the required scram reactivity at HZP, when afded to the cdditional 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 greater than the value assumed in other safety analyses f rom at-power conditions.
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 cnd 4.6. The analysis explicitly includes an evaluation of the impact nf operating between the Technical Specification minimum and maximum cold leg temperutures of 500 and $52 r, respectively. Other safety analyses are also performed f rom both HZP and HFP ct iditions. The CEA insertion limits versus power are designed to provide increased available scram reactivity proportional to the increased power level.
6341R/23.193
The steam line rupture analysis provides the minimum required worth in CEAs for cooldown events from HFP and HZP conditions to maintain subcriticality. 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 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 for nominal HFP and HZP conditions. This is also true for the range of allowable core inlet temperatures from 500 to 552 F for both HFP and HZP conditions. l l
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 dercribed in (14). Compliance with the startup test criteria on CEA worths demonstrates the available scram reactivity in Table 4.13. As such, it also demonstrates CEA worth in excess of the required scram reactivities.
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 11 RCS soluble boron conditions of 1150 and 0 ppm, respectively. The 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.
4.9.5 Shutdown Margin Requirements Shutdown margin is defined as the sum of:
- 1) the reactivity by which the reactor is suberitical in its present condition, and l
l l 6341R/23.193 1
- 2) the reactivity associated with the withdrawn trippable CEAs less the reactivity associated with the highest worth withdrawn trippable CEA.
For a critical reactor, the shutdown margin must be maintained by sufficient available scram reactivity. The required and available scram 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 11. Adequate shutdown margin exists, provided the CEA insertion limits and assumptions inherent in them are fulfilled. These assumptions are:
- 1) the available scram reactivity calculations,-
- 2) the operability of all trippable CEAs, and
- 3) the CEA drop time to 90% of full insertion in less than 2.7 seconds.
The shutdown margin requirement is expressed in the Technical Specifications, as shown in Figure 4.12. The equation representation in the figure allows for calculation of the minimum required shutdown margin for any RCS beron concentration and power level. The shutdown margin requirements at lower boron concentrations are increased for Cycle 11 by 0.55% delta rho.
This increase is the result of the increased moderator defect with cooldown discussed in Section 4.6.
This shutdown margin representation is demonstrated, in Figure 4.11, to 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 4.12 and the Technical Specifications to provide a well-defined requirement as a function of key plant parameters. This Specification permits the development of procedures which preserve the minimum required shutdown 6341R/23.193
margin. Ur. der normal operating conditions, the CEA insertion limits provide such assurance. In the event of an inoperable or slow CEA, such procedures would apply.
4.10' Pressure Vessel Fluence A prog.am for reduction in pressure vessel fluence has been in place for Maine Yankee since Cycle 7 to address Pressurized Thermal Shock (PTS) concerns. The Cycles 7 through 11 core designs have been a progression to lower leakage loading patterns with particular emphasis on the high fluence-2 area from 0 to 10 degrees from a perpendicular line to the core shroud flats.
The core shroud flats are the core boundary lines defined by assembly locations 1 and 2 (or 45 and 54) in Figure 3.2.
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 f rom Cycles 1,1 A, and 2 through 5. Given these target fluence reductions, the materials assessment
- submitted to the NRC in (58) concludes that the circumferential weld seam between the middle and lower shells will not reach the PTS screening criteria until well beycao the expiration of the current plant license.
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 relative to the fluence history of Cycles 1 thrcugh C. The target fluence reductions in (39) for Cycles 7 through 11 are compared to the actual core design fluence reductions obtained by a view-f etor weighting technique of the average quarter-assembly powers calculated for the cycles. The result is that the cumulative fluence reduction factor target te end-of-Cycle 11 has been achieved for both the 0 and 10 degree azimuthal at gles. At the critical circumferential weld at 0 degrees, the target cumulative fluence reduction factor is 1.11 relative to the case in which no fluence reduction measures were instituted. This is achieved by a Cycle 11 flux reduction factor of 1.93 6341R/23.193
relativetoaverage.fluxinCycles1througI16. Similar flux reduction
. factor.s are expected for future cycles to meet the targets set forth in (39).
4.11 Methodology and Methodology Revisions A summary of the reference report and supplemental documentation for the application of physics methodology to Maine Yankee since Cycle 3 is given in Table 4.15. The reference physics methodology report is YAEC-1115 (9).
g There are no new changes to the reactor physics methodology for Cycle 11.
6341R/23.193
-= . -- . - , _ - _ _ , _ - - . _ . _ _ - . - . . _ . . _ , _ _ _ . _ - _ - _ _ . _ - . - - _ - - _ , _ _ , . . _ . _ , _ _
4 '
TABLE 4.1 Maine Yanlee Cycles 3, 10 and 11 Nuclear Characteristics Cycle 3 Cycle 10 Cycle il Core Characteristics s Exposure (mwd /Mt)
Core Average at BOC 7,000 15,150 15,500 Cycle Length at Full Power '0,200 11,200 12,750 Reactivity Coefficients - ARO Moderator Temperature Coefficient f (10-4 delta rho /JF) ;
HFP BOC -0.34* -0.63 -0.57 l HFP E0C -1 .988' -2.46 -2.67 l Fuel Temperature Coefficient l (10-5 delta rho /0F) 1 HZP,B0C
. 10 -1.63 -1.65 l
> HFP,BOC -1.30 -1.26 -1.28 l HZP,E0C -1.80 -1.78 -1.80 l HFP,E00 -1.37 -1.38 -1.40 Kinetics Par'ameters - ARO Total Delayed Neutron Fraction (Be rg)
HFP,BOC 0.00611 0.00618 0.00625 HFP,E00 0.00517 0.00516 0.00517 Prompt Neutron Generation Time (10-6 sec)
HFP BOC 29.3 25.2 24.1 HFP,EOC 32.3 29.8 28.7 y Control Characteristics Control Elements Assemblies Number Full /Part Length 77/* 81/0** 81/0**
Total CEA Scrammable Worth (% delta rho)
HFP,BOC 9.56 ,9.21 HFP,EOC 5 10.55 10.17 Burnable Poison Rods Number B40-A1 023 756 1072 872 Total Worth at HFP,BOC (% delta rho) 1.4 1.7 1.4 Critical Soluble Boron at BOC,ARO (ppm)
HZP,NoXe,PkSm 1,075 1,345 1,516 HFP,NoXe,PkSm 995 1,243 1,410 HFP, Equilibrium Xe 782 993 1,149 Inverse Boron Worths (ppm /% delta rho)
HZP,B9C 84 102 108 HFP,BOC 89 107 114 HZP,EOC 74 83 87 HFP,EOC 79 89 92 1
- Conditions of 2440 MWt/2100 psia operation
- Four f ull-length CEAs are nonscrammable in Cycles 10 and 11 ruvuns&Jun . _ _ __ _ _ _
_ . _ _ _ _ - _ - . _ _ - - _ . . . . ._ m .
TABLE 4.2 Maine Yankee Cycles 3, 10. and 11 CEA Group Worths at HFP Worths (% delta rho)
Cycle 3 Cycle 10 Cycle 11 BOC EOC BOC EOC BOC E00
~ Shutdown CEA Groups Banks C + B + A 5.86 6.02 6.06 6.59 5.76 6.43 Regulating CEA Groups Bank 5* 0.55 0.64 1.46 1.59 1.31 1.47 Bar,ks 5 + 4 0.90 0.97 1.79 1.97 1.67 1.85
. Banks 5 + 4 + 3 1.74 1.90 2.81 3.12 2.67 - 2.98 Banks 5 + 4 + 3 + 2 2.49 2.73 3.52 4.01 3.30 3.71 Banks 5 + 4 + 3 + 2 + 1 3.32 3.54 4.44 5.02 4.46 4.97 All CEA Groups Banks 5 + 4 + 3 + 2 + 1 +
C+B+A 9.18 9.56 10.50 11.61 10.22 11.40
- Bank 5 was redesigned in Cycle 7 to provide additional raectivity worth.
6341R/23.193
\
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, - . . , _ . _ - _ . . - _ . . - _ . _ , _ _ _ . . __ . ._, . _ ..l
[r
"- TABLE 4.3 Maine Yankee Cycles 3, 10. and 11 Core Average Doppler Defect Doppler Defect (x 10-4 delta rho)
Fuel Cycle 3* Cycle 10 Cycle 11 Resonance Temperature F BOC EOC B00. EOC BOC EOC-4000 0 0 0 0 0 0 3750 19.4 20.9 - - - -
3500 39.4 42.5 41.1 45.3 41.6 45.9 3250 59.9 64.8 - - - -
3000 81.2 87.8 84.8 93.2 85.8 '94.5 2750 103.1 111.6 - - - -
2500 125.9 136.2 131.4 144.5 133.0 146.5 2250 149.7 161.9 - - - -
2000 174.5 188.6 181.9 200.0 184.0 202.7 1750 200.5 216.7 - - - -
1500 228.1 246.5 237.5 261.1 240.3- 264.6 1232 - - 270.1 296.8 273.3 300.8 1000 28F. 7 311.9 300.4 329.9 303.9 334.3 800 - -
328.4 360.5 33_.2 -365.4 532 358.5* 387.4* 369.4 405.4 373.7 411.0 300 394.3 426.0 409.5 449.1 414.4 455.4 200 - - 428.6 470.0 433.7 476.7 100 - - 449.1 492.4 454.5 499.4 0 - - 471.5 516.8 477.0 524.2
- at 5250F
- j. 6341R/23.193
L e TABLE 4.4
' Maine Yankee Cycles 3, 10 and 11 Core Average Doppler Coefficient Doppler Coefficient (x 10-4 delta rho per OF)
. Fuel- Cycle 3* Cycle 10 Cycle-11 Resonance Temperature F BOC E0C .BOC E0C BOC EOC 100 - -
0.2142 0.2340 0.2168 0.2378 200 - - 0.1980 0.2164 0.2006 0.2198 300 - - 0.1850 0.2020 0.1876 0.2055 400 - -
0.1744 0.1902 0.1768 0.1933 532 0.159* 0.172* 0.1630 0.1777 0.1646 6.1803 ,
800 0.144** 0.156** 0.1449 0.1587 0.1465 0.1610 1000 0.131 0.141 0.1352 0.1480 0.1369 0.1501 1232 0.121*** 0.121*** 0.1263 0.1377 0.1277 0.1398 1500 0.114 0.123 0.1176 0.1290 0.1191 0.1307 2000 0.102 0.110 0.1061 0.1166 0.1073 0.1181 2500 0.093 0.101 0.0972 0.1068 0.0982 0.1082.
3000 0.086 0.094 0.0903 0.0992 0.0913 0.1006 3500 0.081 0.088 0.0847 0.0932 0.0856 0.0945 !
- At 5250F
- At 7500F
- At 12500F 6341R/23.193
75J l
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TABLE 4.5 Maine Yankee Cycles 3. 10. and 11 I Moderator Temperature Coefficients l
-Conditions: HFP and HZP, ARO, Critical Boron Concentrations
MTC (10-4 delta rho /0F)----==== =-
Cycle 3* Cycle 10 Cycle 11 Case Conditions BOC EOC BOC EOC BOC EOC HFP, EqXe, EqSm -0.47 -2.24 -0.63 -2.46 -0.57 -
-2.67 HZP, NoXe, PkSm +0.24 -
+0.23 -
+0.24 -
HZP, NoXe, Eqsm -
-1.40 -
-1.28 -
-1.33
- Cycle 3 HZP values at 5250F, Cycles 10 and 11 HZP values at 5320F.
i l-l l 6341R/23.193
~
E~ l t.
TABLE 4.6 Maine Yankee Cycles 3. 10, and 11 l ARI Moderator Defect With Warst Stuck CEA Moderator Defect (x-10-4 delta rho)**
Moderator Cycle 3* Cycle TO Cycle 11 Temperature
( F) BOC' EOC BOC EOC BOC EOC O ppm -1000 ppm 0 ppm 1150 ppm 0 ppm 576.4 -
-138.0 -54.0 -130.0 -55.0 -145.0 532 -
0.0* 0.0 0.0 '0.0 0.0 500 -
43.5 23.0 60.0 18.0 75.0 450 - 111.0 35.0 128.0 33.0 160.0 400 -
167.3 35.0 180.0 40.0 220.0 350 -
217.0 35.0 213.0 44.0 270.0 300 -
261.6- 35.0 238.0 47.0 311.0 250 -
298.5 35.0 260.0 48.0 340.0 200 - -
35.0 280.0 49.0 368.0 150 - -
35.0 '295.0 50.0- 385.0 100 - -
35.0 306.0 50.0 396.0 68' - -
35.0 312.0 51.0 401.0 l'
l i
f l.
- Cycle 3 values referenced to 0 at 5250F L
O* Moderator defect at a constant 2250 psia for the specified temperatures I ,
6341R/23.193
TABLE 4.7 Malne Yankee Cycles 3 and 11
-Kinetics Parameters Conditions: 'HFP, ARO, Critical' Boron Delayed Cycle 3 Cycle 11 Time in Neutron Effective Lifetige . Effective Lifetige Cycle Life Group Fraction (See ) ' Fraction (See )
800 1 0.00018 0.0126 0.00018 0.0126 2 0.00128 0.0305 0.00131 0.0305 3 0.00116 0.1163 0.00119 0.1172 4 0.00237 0.3116 0.00243 0.3142 5 0.00083- 1.1652 0.00084 '1.1787 6 0.00029 3.0253 0.00030 '3.0247 TOTAL 0.00611 0.00625 E0C 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 0.1202 4 0.00197 0.3185 0.00197 0.3211 5 0.00072 1.1833 0.00072 1.1999 6 0.00025 2.9831 0.00024 2.9904 TOTAL 0.00517 0.00517 l
6341R/23.193
. TABLE 4.8 Maine Yankee Cycles-10 AND 11 CEA Ejection Results from Full Insertions Cycle 10 Cycle 11
-Maximum 1-Pin Radial Peak BOC 'EOC BOC EOC HFP Bank 5 In 3.81 4.12 3.50 3.83 Ejected 5 (INCA Location 20)
HZP Banks 5 + 4 In 5.63 5.78 5.51 15.72 Ejected 5 (INCA Location 20)
Maximum Ejected Worth (% delta rho)
HFP Bank 5 In 0.307 0.402 0.216 0.301 Ejected 5 (INCA Location 20)
HZP Banks 5 A 4 In 0.432 0.573 0.316 0.434 Ejected 5 (INCA Location 20) l' l
t i ~50-6341R/23.193
3 TABLE 4.9 Maine Yankee Cycles 10 and 11 CEA Drop Results at BOC CEA Group Dropped Dropped CEA Worth Maximum 1-Pin Positions CEA (1 delta rho) Radial Power
- Before Drop Type- Cycle 10- Cycle 11 . Cycle 10 Cycle 11 ARO A 0.113 0.120 1.77 1.77 ARO B 0.167 0.153 1.79 1.75 ARO C 0.100 0.106 1.74 1.75 ARO 1 0.062 0.072 1.65 '1.68 Bank 5 l[n A 0.106 0.110 1.92 1.93 Bank 5 In B 0.178 0.160 1.99 1. 94 Bank 5 In C 0.103 0.113 1.91 1.95 Bank 5 In 1 0.058 0.075 1.80 1.85
- Pre-Drop Maximum 1-Pin radial powers:
Cycle 10 Cycle 11 ARO 1.530 1.547 Bank 5 In 1.678 1.693 Post-Drop Maximum 1-Pin radial power at 80% of 2630 MWt power level conditions.
6341R/23.193
i l
I TABLE 4.10 g Maine Yankee Cycles 10 and 11 CEA Drop Results at EOC CEA~ Group Dropped Dropped CEA Worth Maximum 1-Pin Positions CEA .01 delta rho) Radial Power *-
TB efore Drop Type Cycle 10 Cycle 11 Cycle'10 Cycle 11
.ARO- A 0.118 'O.127 1.74 1.74 ARO B 0.180 0.164 1.75 1.69 ARO ' C 0.100 0.107 1.71 1.70
' ARO 1 0.067 0.078 1.64 1.64 Bank 5 In A 0.114 0.121 1.87 1.84 Bank 5 In B 0.195 0.171 1.63 1,81 Bank 5 In -C 0.104 0.117 1.85 1.84 Bank 5 In 1 0.061 0.081 1.77 1.76 i
i
- Pre-Drop Maximum 1-Pin radial powers:
Cycle 10 Cycle 11 ARO 1.522 1.513 Bank 5 In 1.675 1.613 l
l Post-Drop Maximum 1-Pin radial power at 80% of 2630 MWt power level conditions.
6341R/23.193
TABLE 4.11 Maine Yankee Cycles 10 and 11 Dropped CEA With Power Level Restriction Jiost Limiting Peaking Cases CEA Drop Time Maximum Power Percent Increase in Maxim ^im
-From Power Post-Drop . Level Permitted 1-Pin Peaking Level-(%) (Hrs) (%) Cycle 10 Cycle 11 100 0.5 100 13.64 11.72-1.0 90 17.08 14.49 2.0 80 22.46 18.42 3.0 80 24.50 20.00 4.0 80 26.07 21.88 90 0.5 90 14.39 12.47 1.0 80 18.12 15.50 2.0 70 24.02 20.04 3.0 70 27.05 22.10 4.0 70 28.98 23.58 80 0.5 80 13.82 13.35 1.0- 70 16.70 16.49 2.0 60 20.75 21.40 3.0 60 21.52 23.80 4.0 60 21.81 25.90 70 0.5 70 15.74 14.11 1.0 60 19.75 17.55 2.0 50 26.31 22.56 3.0 50 29.71 25.25 4.0 50 33.12 27.95 60 0.5 60 17.12 14.69 1.0 50 22.59 18.40 2.0 40 30.61 23.75 3.0 40 34.99 26.32 l
4.0 40 39.17 28.67 6341R/23.193
e TABLE 4.12 Maine Yankee Cycle'll Available Scram Reactivity Worths (% delta rho) l BOC EOC l HFP HZP HFP- HZP
'Scrammable CEA. Worth
- 9.21 8.77 10.17 9.67 l
l Stuck CEA Worth 1.32 1.26 1.58 1.51 l
PDIL CEA Worth ** 0.13 1.42 0.24 1.64
'CEA Programming Allowance 0.05 0.04 0.09 0.15 Available Scram CEA Worth
- Nominal 7.71 6.05 8.26 6.37.
- With Uncertainties *** 6.94 5.45 7.43 5.73 s
- Uncertainty factor of 0.9 6341R/23.193 i
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TABLE 4.13 Maine Yankee Cycle 11 Required Scram Reactivity Worth (% delta rho) for Time in Cycle Life and RCS Soluble Boron Concentration BOC- E00 1150 ppm 0 ppm
-HFP HZP HFP H2P Available Scram 6.94 5.45 7.43 5.73 Reactivity with Uncertainties (Tables 4.12)'
Minimum Required Worth in CEAs Assumed -
- Steam Line Rupture 2.96 1.68 6.94 4.75 Event *
(Section 5.5.1)
- Safety Analyses 5.70 3.20 6.93 4.50 (Section 5)
Required Scram 5.70 3.20 6.94 4.75 Reactivity **
Excess from Required 1.24 2.25 0.49 0.98 to Available Scram Reactivity
- An uncertainty factor of 0.9 is applied to the nominal minimum required worth in CEAs for the steam line rupture event from Table 5.7 for comparison to the available scrat reactivity with uncertainties. This uncertainty component is statistically combined with the other uncertainty components to derive the nominal minimum required worth in CEAs, as discussed in (14).
- Maximum of either the minimum required worth in CEAs assumed for the steam line rupture event or other safety analyses in Section 5 6341R/23.193
TABLE 4.14 Maine Yankee Cycles 6-11 Relative Pressure Vessel Fluence Comparison Flux Reduction' Factors ** at Azimuthal Angle from Perpendicular to Core Shroud Flats-Total Effective Full-Power
. Years (EFPY) ------ 00 --
100 --------
Cycles to EOC* Target Designed Target Designed 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 9.80 1.35 1.55 1.51 1.56 10 10.90*** 1.35 2.03 1,51 1,73 11 12.10*** 1.35 1.93 1.51 1.67 Future Cycles 1.35 1.51 Cycle 1-11 Average **** 1.11 1.19 1.17 1.18
- Based on 2,630 MWt full power operation
- Inverse of fractional flux relative to Cycles 1 through 6
- Estimated cycle lengths
- Inverse of EFPY-weighted fractional fluxes l
i l 6341R/23.193 l-
TABLE 4.15 Maine Yankee Physics Methodology Documentation Supporting /.pplication Description of Methodology Documentation Reference in Cycle Reactor Physics Methods - YAEC-1115 9 3 Reference Report Reactor Protective System Setpoint YAEC-1110' 4 3 Analysis - Reference Report Extension of Fine Mesh Diffusion PC No. 64, 20 4 Theory and Nodal Physics Methods to Section 4.8 Reactivity Parameter Calculations WMY 78-102, 38 and a Change in the Nodal Neutronic Attachment B Coupling 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 Theory Calculations for Dropped CEAs Attachment C and Special CEA Drop Test at 50%
Power for Method Verification Uncertainty Applied to Moderator YAEC-1259, 43 6 Reactivity Defect from Hot Zero Power Section 4.7 Reduced from 25 to 15%
6341R/23.193
e.
TABLE 4.15 (Continued)
-Maine Yankee Physics Methodology Documentation Supporting < Application
~ Description of Methodology Documentation Reference in Cycle Doppler Defects for. CEA Ejections YAEC-1324, 12 7 Calculated with Explicit Pre-Ejected Section 4.10 Local Power' Weighting and Uncertainty Reduced from 25 to 15%
CEA Ejections Calculated From YAEC-1479, 3 9 Partial CEA Insertions Section 4.11 Moderator Density Defect for YAEC-1479, 3 '9
'LOCA Analysis Calculated Using Section 4.11 Fine Mesh Diffusion Theory Augmentation Factors Eliminated YAEC-1573, 54 10 as a Power Spike Penalty Section 4.11 i
i 1
6341R/23.193 4 l
L
- FIGURE 4.1 MRINE YANKEE CYCLE 11 RSSEMBLY' RELATIVE POWER DENSITIES HFP, RRO, 800 (S00 MHD/MT) l ASSEMBLY TYPE AND CORE POSITION
- N-8 i L-0 2 ASSEMBLY AVERAGE RELATIVE POWER
- 0.317 0.352 MAXIMUM FUEL ROD RELATIVE POWER
- 0.586 0.590 MAXIMUM CHANNEL RELATIVE POWER.
- 0.568 0.571 N-8 3 0-0 4 0-0 5 0-4 6 0-4 7 0.369 0.938 1.150 1.153 1.107 0.745 1.440 1.542 1.513 1.394 5UMMARY OF MAXIMUM POWER 5 PDWER VALUE LOCATION N-8 8 0-0 9 P-0 10 P-4 11 P- i 12 N-4 13 ASSEMBLY 1.359 40 0.400 1.108 1.135 1.127 1.135 0.845 FUEL ROD 1.547 40 0.732 1.503 1.303 1.253 1.306 0.886 CHANNEL 1.469 40 0.718 1.423 1.257 1.194 1.263 0.879 N-8 14 0-0 15 P-8 16 0-8 17 N-8 18 0-4 19 N-4 20 0.396 1.101 1.109 1.299 0.905 1.304 0.830 0.727 1.458 1.231 1.520 0.942 1.525 0.870 O.713 1.380 1.169 1.437 0.938 1.448 0.873 N-8 21 0-0 22 P-8 23 P-0 24 N-4 25 0-4 26 P-8 27 N-0 28 0.368 1.107 1.108 1.125 0.947 1.358 1.132 0.856 0.744 1.502 1.230 1.260 1.033 1.546 1.266 0.937 0.721 1.422 1.168 1.217 1.027 1.469 1.202 0.926 0-0 29 P-0 30 0-8 31 N-4 32 0-4 33 N-4 34 P-0 35 P-4 36 0.938 1.135 1.299 0.947 1.338 0.944 1.205 1.129 1.440 1.303 1.519 1.033 1.516 1.005 1.373 1.255 1.366 1.257 1.437 1.027 1.441 0.995 1.333 1.195
~
0-0 37 P-4 38 N-8 39 0-4 40 N-4 41 P-8 42 P-0 43 P-8 44 1.158 1.128 0.905 1.359x 0.944 1.089 1.147 1.166 1.543 1.253 0.943 1.547x 1.005 1.179 1.308 1.255 N-8 45 1.463 1.195 0.938 1.469x 0.994 1.122 1.269 1.191 0.318 0.58 0-4 46 P-4 47 0-4 48 P-8 49 P-0 50 P-0 51 N-8 52 0-4 53 '
O.568 1.154 1.135 1.305 1.133 1.206 1.150 0.926 1.346 1.513 1.306 1.526 1.267 1.375 1.311 0.973 1.522 L-0 54 1.434 1.263 1.449 1.204 1.335 1.270 0.960 1.447 0.352 0.590 0-4 55 N-4 56 N-4 57 N-0 58 P-4 59 P-8 60 0-4 61 M-8 62 0.571 1.107 0.845 0.831 0.857 1.130 1.170 1.350 0.925 1.394 0.886 0.880 0.937 1.257 1.261 1.530 0.950 ,
1.324 0.879 0.873 0.927 1.196 1.197 1.455 0.938 1
l-I FIGURE 4.2 1 MRINE YANKEE CYCLE 11 l ,
i RSSEMBLY RELATIVE POWER DENSITIES HPP, RRO, MOC (6K MWD /MT)
RSSEMBLY TYPE RND CORE POSITION
. . N-B 1 L-0 2 RSSEMBLY RVERAGE RELATIVE POWER
- 0.341 0.390 NRXIMUM FUEL R00 RELATIVE POWER
- 0.608 0.625 NAXIMUM CHANNEL RELATIVE POWER. . . . 0.595 0.610
" N-8 3 0-0 4 0-0 5' 0-4 6 0-4 7 0.378 0.906 1.110 1.168 1.149 0.725 1.341 1.454 1.457 1.395 0.708 1.287 1.397 1.@0 1.339 ,
SUMMARY
OF MAXIMUM POWERS POWER VBLT LOCATION N-8 8 D-D 9 P-0 10 P-4 11 P-4 12 N-4 13 40 0.418 1.084 1.092 1.107 1.144 0.881 RSSEMBLY 1.367 1.211 1.296 0.917 FUEL R00 1.521 40 0.735 1.434 1.220 0.723 1.377 1.191 1.161 1.233 0.910 CHANNEL 1.462 40 N-8 14 0-0 15 P-8 16 0-8 17 N-8 18 0-4 19 N-4 20 1.091 1.109 1.324 0.925 1.327 0.866 0.414 1.506 0.910 0.731 1.412 1.222 1.480 0.960 1.35s 1.165 1.419 0.952 1.446 0.902 0.719 P-8 23 P-0 24 N-4 25 0-4 26 P-8 27 N-0 28 N-8 21 0-0 22 1.367 1.142 0.881 0.378 1.083 1.109 1.122 0.962 1.222 1.237 1.031 1.520 1.265 0.943 0.725 1.434 1.204 0.937 0.707 1.376 1.165 1.205 1.025 1.462 0-8 31 N-4 32 0-4 33 N-4 34 P-0 35 P-4 36 0-0 29 P-0 30 1.167 1.10.7 0.906 1.092 1.324 0.962 1.344 0.950 1.480 1.032 1.487 1.003 1.302 1.212 1.341 1.220 1.267 1.161 1.287 1.191 1.419 1.025 1.430 0.992 N-8 39 0-4 40 N-4 41 P-8 42 P O 43 P-8 44 0-0 37 P-4 38 1.090 1.112 1.141 1.110 1.107 0.925 1.367x 0.950 0.960 1.521x 1.003 1,167 1.237 1.218 1.454 1.211 1.171 N-8 45 1.161 0.952 1.462w 0.992 1.117 1.217 1.397 0.341 0.609 P-4 47 0-4 48 P-8 49 P-0 50 P-0 51 N-8 52 0-4 53 ,
0.595 0-4 46 1.167 1.113 0.920 1.326 1.168 1.144 1.327 1.143 1.267 1.302 1.237 0.959 1.461
'. 1.457 1.296 1.506 1.406 64 1.233 1.447 1.205 1.267 1.217 0.946 1.400
- 0. 90 M-8 62 0.625 0-4 55 N-4 56 N-4 57 N-0 58 F-4 59 P-8 60 0-4 61 0.610 0.881 1.107 1.144 1.329 0.923 1.149 0.881 0.867 0.945 0.917 0.911 0.943 1.211 1.222 1.466 1.394 1.175 1.410 0.935 1.339 0.910 0.902 0.937 1.161 l
FIGURE 4.3 MRINE YANKEE CYCLE 11 l RSSEMBLY RELATIVE POWER DENSITIES HFP, RRO, EOC I14K MWD /MT) l l
RSSEMBLY TYPE RNO CORE POSITION - N-8 1 L-0 2 l ASSEMBLY RVERAGE RELATIVE POWER 0.388 0.460 M8XIMUM RJEL R00 RELATIVE POWER -
0.675 0.707 M8XIMUM CH8NNEL RELRTIVE POWER. . . . 0.661 0.694 j
l N-8 3 0-0 4 0-0 5 0-4 6 0-4 7l 0.402 0.887 1.075 1.216 1.229 1 1 0.731 1.242 1.356 1.443 1.427 i 0.72 1.210 1.322 1.368 1.MB l
SUMMARY
OF MAXIMUM POWERS EDE8 VALUE LOCATION N-8 8 0-0 9 P-0 10 P-4 11 P-4 12 N-4 13 RSSEMBLY 1.378 17 0.443 1.057 1.065 1.084 1.140 0.925 FUEL RDO 1.513 26 0.745 1.341 1.158 1.147 1.241 0.956 CH8NNEL 1.434 26 0.734 1.308 1.141 1.122 1.205 0.954 i N-8 14 0-0 15 P-8 16 0-8 17 N-B 18 0-4 19 N-4 20 '
O.440 1.060 1.086 1.378x 0.954 1.342 0.901 i 0.743 1.314 1.176 1.488 0.978 1.494 0.939 i 0.731 1.279 1.146 1.416 0.974 1.413 0.936 !
N-8 21 0-0 22 P-8 23 P-0 24 N-4 25 0-4 26 P-8 27 N-0 28 0.402 1.056 1.086 1.109 0.987 1.371 1.114 0.893 ,
0.730 1.341 1.176 1.200 1.031 1.513x 1.201 0.927 '
O.715 1.307 1.146 1.184 1.031 1.434x 1.169 0.925 ,
0-0 29 P-0 30 0-8 31 N-4 32 0-4 33 N-4 34 P-0 35 P-4 36 f 1.065 1.378 0.987 1.349 0.950 1.106 1.050 ;
l 0.887 1.107 1.242 1.159 1.487 1.031 1.487 1.000 1.203 :
1.210 1.142 1.416 1.031 1.413 0.995 1.184 1.081 i
0-0 37 P-4 38 N-8 39 0-4 40 N-4 41 P-8 42 P-0 43 P-8 44 1.075 1.084 0.954 1.371 0.949 1.053 1.056 1.077 1.356 1.147 0.978 1.513 0.999 1.106 1.147 1.126 N-8 45 1.322 1.122 0.974 1.434 0.994 1.078 1.139 1.105 i O.388 0.675 0-4 46 P-4 47 0-4 48 P-8 49 P-0 50 P-0 51 N-8 52 0-4 53 0.661 1.140 1.342 1.114 1.106 1.057 0.910 1.301 1.216 1.426 1.443 1.241 1.494 1.202 1.202 1.146 0.935 64 1.368 1.205 1.413 1.170 1.184 1.138 0.931 1.350 O. 60 0.707 0-4 55 N-4 56 N-4 57 N-0 58 P-4 59 P-8 60 0-4 61 M-8 62 0.694 0.925 0.902 0.890 1.049 1.078 1.303 0.926 1.229 0.938 1.426 0.956 0.940 0.927 1.106 1.128 1.427 0.954 0.936 0.925 1.080 1.106 1.351 0.936 1.358 1
FIGURE 4.4 MAINE i'RNKEE CYCLE 11 ASSEMBLY RELRTIVE POWER DENSITIES I HFP, CER BANK 5 INSERTED, 800 (500 MWD /MT)
ASSEMBLY TYPE RND CORE POSITION . . . N-8 1 L-0 2 ASSEMBLY AVERAGE RELATIVE POWER . . . 0.349 0.371 MAXIMUM FUEL R00 RELATIVE POWER -
0.640 0.621 MAXIMUM CHANNEL RELATIVE P0HER. . . - 0.621 0.602
, N-8 3 0-0 4 0-0 5 0-4 6 0-4 7 0.383 1.042 1.290 1.200 1.057 0.787 1.604 1.692 1.623 1.291 0.761 1.518 1.606 1.542 1.238
SUMMARY
OF HAX1 MUM POWERS POWER YfLUE LOCRTION N-8 8 0-0 9 P-0 10 P-4 11 P-4 12 N-4 13 ASSEMBLY 1.376 61 0.293 1.045 1.206 1.228 1.131 0.583 FUEL R30 1.693 37 0.539 1.466 1.437 1.378 1.381 0.750 CHANNEL 1.607 37 0.526 1.399 1.386 1.315 1.332 0.697 N-8 14 0-0 15 P-8 16 0-8 17 N-8 18 0-4 19 N-4 20 0.290 0.576 0.931 1.286 0.957 1.374 0.830 0.534 0.879 1.137 1.545 1.005 1.644 0.894 0.521 0.836 1.107 1.467 1.000 1.562 0.892 N-8 21 0-0 22 P-8 23 P-0 24 N-4 25 0-4 26 P-8 27 N-0 28 0.383 1.045 0.931 0.946 0.815 1.362 1.243 0.962 0.787 1.466 1.137 1.047 0.894 1.644 1.370 1.071 0.761 1.399 1.107 1.019 0.886 1.564 1.294 1.060 0-0 29 P-0 30 0-8 31 N-4 32 0-4 33 N-4 34 P-0 35 P-4 36 1.042 1.206 1.287 0.815 0.814 0.892 1.349 1.306 1.605 1.438 1.546 0.894 1.027 0.990 1.579 1.457 1.519 1.387 1.468 0.886 0.971 0.980 1.535 1.386 0-0 37 P-4 38 N-8 39 0-4 40 N-4 41 P-8 42 P-0 43 P-8 44 1.291 1.229 0.958 1.363 0.892 1.139 1.290 1.329 1.693w 1.379 1.006 1.645 0.989 1.286 1.500 1.446 N-8 45 1.607x 1.316 1.001 1.566 0.980 1.218 1.452 1.373 0.350 0.640 0-4 46 P-4 47 0-4 48 P-8 49 P-0 50 P-0 51 N-8 52 0-4 53 0.622 1.201 1.132 1.375 1.245 1.350 1.294 0.999 1.371 1.624 1.382 1.646 1.372 1.582 1.503 1.063 1.664 L-0 54 1.543 1.333 1.563 1.296 1. 2 1.455 1.055 1.587 0.371 0.622 0-4 55 N-4 56 N-4 57 N-0 58 P-4 59 P-8 60 0-4 61 M-8 62 0.603 1.058 0.583 0.831 0.963 1.308 1.334 1.376x 0.652 1.292 0.751 0.895 1.072 1.459 1.451 1.673 0.796 i
1.239 0.697 0.894 1.061 1.388 1.377 1.596 0.746 l
l
FIGURE 4.5 MRINE YRNKEE CYCLE 11 RSSEMBLY RELATIVE POWER DENSITIES HFP, CER BANK 5 INSERTED, M00 (6K MWD /MT)
ASSEMBLY TYPE AND CORE POSITION - N-B 1 L-0 2 ASSEMBLY AVERRGE RELATIVE POWER -
0.375 0.411 MAXIMUM FUEL R00 RELATIVE POWER -
0.661 0.655 MAXIMUM CH8NNEL RELATIVE POWER -
0.645 0.640 N-8 3 0-0 4 0-0 5 0-4 6 0-4 7 0.395 1.008 1.237 1.211 1.092 0.769 1.497 1.593 1.548 1.283
SUMMARY
OF MRXIMUM POWERS
- W *
- POWER VALUE LOCATION N-8 8 0-0 9 P-0 10 P-4 11 P-4 12 N-4 13 ASSEMBLY 1.399 48 0.307 1.025 1.164 1.207 1.134 0.594 FUEL ROD 1.629 48 0.544 1.403 1.349 1.323 1.343 0.758 CH8NNEL 1.567 46 0.533 1.353 1.316 1.271 1.294 0.704 N-8 14 0-0 15 P-8 16 0-8 17 N-8 18 0-4 19 N-4 20 .
0.304 0.572 0.938 1.319 0.981 1.398 0.864 l 0.540 0.895 1.134 1.524 1.028 1.628 0.931 0.529 0.840 1.091 1.462 1.018 1.566 0.927 N-8 21 0-0 22 P-8 23 P-0 24 N-4 25 0-4 26 P-8 27 N-0 28 0.395 1.025 0.938 0.950 0.833 1.376 1.259 0.993 0.769 1.403 1.134 1.041 0.919 1.622 1.374 1.082 l 0.749 1.354 1.091 1.021 0.911 1.563 1.315 1.074 0-0 29 P-0 30 0-8 31 N-4 32 0-4 33 N-4 34 P-0 35 P-4 36 l 1.009 1.165 1.319 0.833 0.810 0.898 1.309 1.285 l 1.498 1.350 1.524 0.920 1.037 0.992 1.492 1.411 1.435 1.317 1.462 0.911 0.972 0.987 1.464 1.351 0-0 37 P-4 38 N-8 39 0-4 40 N-4 41 P-8 42 P-0 43 P-8 44 1.237 1.207 0.982 1.376 0.898 1.139 1.252 1.303 1.594 1.324 1.029 1.623 0.992 1.269 1.421 1.407 N-8 45 1.533 1.271 1.019 1.564 0.987 1.210 1.386 1.349 0.375
- 0,.661 0-4 46 P-4 47 0-4 48 P-8 49 P-0 50 P-0 51 N-8 52 0-4 53 c 646 1.212 1.134 1.'599x 1.260 1.310 1.254 0.994 1.347 l 1.051 1.597 i 1.549 1.344 1.629x 1.376 1.493 1.422 L-0 54 1.480 1.295 1.567x 1.316 1.465 1.388 1.045 1.542 0.411 l 0.656 0-4 55 N-4 56 N-4 57 N-0 58 P-4 59 P-8 60 0-4 61 H-8 62 0.640 1.092 0.594 0.865 0.993 1.285 1.306 1.351 0.636 l 1.603 0.779 1.284 0.758 0.932 1.083 1.411 1.409 1.236 0.705 0.928 1.074 1.352 1.351 1.548 0.728 f
l - - -. , _ . _ - - - -
l FIGURE 4.6 MAINE YANKEE CYCLE 11 l RSSEMBLY RELATIVE POWER DENSITIES HFP, CER BANK 5 INSERTED, EOC (14K MWD /MT)
I RSSEMBLY TYPE RNO CORE POSITION N-8 1 L-0 2 RSSEMBLY RVERAGE RELATIVE POWER 0.430 0.489 MAXIMUM FUEL R00 RELRTIVE P0HER 0.730 0.739 MAXIMUM CHRNNEL RELRTIVE POWER.
0.717 0.724 N-8 3 0-0 4 0-0 5 0-4 6 0-4 7 0.427 1.003 1.209 1.267 1.172 ;
0.787 1.406 1.496 1.512 1.325
- 0. m 1. W 1.460 1.M3 1.262 SUMMRRY OF MAXIMUM POWERS POWER VRLUE LOCRTION N-8 8 0-0 9 P-0 10 P-4 11 P-4 12 N-4 13 RSSEMBLY 1.414 48 0.331 1.016 1.151 1.192 1.126 0.604 FUEL R00 1.613 48 0.561 1.333 1.282 1.258 1.286 0.789 '
CHRNNEL 1.526 48 0.549 1.306 1.267 1.234 1.259 0.728 N-8 14 0-0 15 P-8 16 0-8 17 N-8 18 0-4'19 N-4 20 0.328 0.555 0.934 1.390 1.018 1.413 0.892 0.558 0.881 1.105 1.553 1.050 1.612 0.958 0.546 0.817 1.089 1.478 1.045 1.526 0.957 N-8 21 0-0 22 P-8 23 P-0 24 N-4 25 0-4 26 P-8 27 N-0 28 '
O.427 1.015 0.934 0.950 0.857 1.378 1.225 1.000 0.788 1.333 1.105 1.028 0.959 1.592 1.304 1.060 l 0.770 1.306 1.090 1.021 0.955 1.512 1.275 1.057 l 0-0 29 P-0 30 0-8 31 N-4 32 0-4 33 N-4 34 P-0 35 P-4 36 '
1.003 1.152 1.391 0.857 0.779 0.885 1.234 1.214 1.407 1.283 1.554 0.959 1.008 0.982 1.374 1.283 i 1.368 1.268 1.478 0.956 0.936 0.980 1.355 1.254 0-0 37 P-4 38 N-8 39 0-4 40 N-4 41 P-8 42 P-0 43 P-8 44 1.210 1.192 1.018 1.378 0.884 1.085 1.178 1.219 1.497 1.258 1.050 1.592 0.982 1.173 1.294 1.278 N-8 45 1.460 1.235 1.045 1.5?3 0.980 1.144 1.277 1.254 l 0.431 l 0.730 0-4 46 P-4 47 0-4 48 P-8 49 P-0 50 P-0 51 N-8 52 0-4 53 '
O.717 1.268 1.126 1.414x 1.226 1.233 1.179 0.968 1.295 1.512 1.286 1.613x 1.305 1.373 1.294 1.012 1.501 L-0 54 1.444 1.259 1.526x 1.276 1.355 1.277 1.011 1.434 l 0.489 0.739 0-4 55 N-4 56 N-4 57 N-0 58 P-4 59 P-8 60 0-4 61 M-8 62 i 0.724 1.173 0.604 0.893 1.000 1.213 1.220 1.297 0.606 1.326 0.789 0.959 1.059 1.282 1.279 1.504 0.761 1.263 0.728 0.958 1.056 1.253 1.254 1.436 0.704 L
NOTE: 1. THIS CURVE INCLUDES 10% CALCULAT10NAL UNCERTAINTY
- 2. F =F X 1.03 R
- 3. MEASURED F SHOULD BE AUGMENTED BY MEASUREMENT R
UNCERTAINTY (8%) BEFORE COMF%RISON TO THIS CURVE.
' 85 p '
i n H I i l i 1.84 ,
i l T 1 as t C ORDINATES (KMWD/MT,Fp )
u o I ' (0.00 ,1.770) to.50 ,1.770 )
1'82 L 1.00 ,1.769 ) (2.00 ,1.767) l l
- j utp (4.00 ,1.755) (6.00 ,1.741) j[ ,
}
1.81 t8.00 ,1.722) I10.00,1.721) i b, t14.00,1.731)
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g 0 1 2 3 4 5 6 7 8 9 10 11 12 1'3 M 15 l l
CYCLE AVERAGE EXPOSURE (KMWD/ aft) l l l
MAINE YANKEE Allowable Unrodded Radial Peak Versus Figure Cycle 11 Cycle Average Burnup 4.7
( 1.0% Above Nominal) 0.6 ,1 i, ,, , , ,
mi ,1 4 i,N . I I !
I
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% a \
l A i
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COORDINATES :
t
' . POWER MIC :
g ,- ,,
0 0.60 O h '
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3 ACCEPTABLE OPERATION i .1 I I ! I Ii i l11 0.1 - ,
l 4
l
- I 0.0 0 10 20 30 40 50 60 70 80 90 10 0 POWER LEVEL (%, OF RATED POWER)
Moderator Temperature Coefficient Upper Umits Figure MAINE YANKEE 4.8 Cycle 11 Versus Power Leve!
l l
1 i
r h MAXIMUM POWER LE/EL (% OF RATED POWER) VS. CEA WITH0RAWAL (STEPS)
M7
" ra ' - -
100
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l POWER CEA WITHDRAWAL I l li:b- :i
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- 12
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l h 10 0 90 153 180 180 180 180 140 180 180 180 180
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E j :. m _Ofi;j 4
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- 2 60 101180 180 180 180 j . . ..
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att o 50 88 180 180 180 180 .
F g 60-E:t 40 75 180 180 180 180 :}:L JL pp. :Lp :7 '
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t 20 49 157 180 180 180 } }; . ; i g 50- =it! d: :
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OPERATION f
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OROUP 3 OROUP S (5A ANO 58)
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E B E 5 5 5 100 15 0 ISO 0 50 100 15 0 180 I
50 15 0 180 0 50 0 10 0 OROUP 4 3 h OROUP 2 s a e s i i e i gE i i 15 0 160 0 50 10 0 15 0 160
= 0 50 10 0 CEA WITHDRAWAL BY GROUP (STEPS)
l I
FIGURE 4.10 MAINE YANKEE CYCLES 10 RND 11 MAXIMUM PEAKING VS. DROPPED CER WORTH FROM SPECIFIED POWER LEVELS g ... ... . .i.
1
- l 1
46 : POWER
- : LIVD
(%)
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CYCLE 11 L'NVELOPES 1 l
POINTS Or INTERSECTION ~
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-(.0983'23.59)
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.........,............................i....................................'
O.00 0.02 0.04 0.06 0.08 0.10 0.12 0.14 0.16
' DROPPED CER WORTH - PERCENT DELTA RHO FIGURE 4.11 MAINE YANKEE CYCLE 11 SHUTDOWN MARGIN EQUATION AND REQUIRED SCRAM REACTMTY S.o
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). .
e s- .,.,
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. . . *o 1.5-LEGEND SHUTDOWN MARGIN EQUATION REQUIRED SCRAM REACTMTY (TABLE 4.13) o 5-o.................O STEAM UNE RUPTURE EVENT D OTHER SAFETY ANALYSES
~
'o ioo 26o 36o' 46o s60 e6o 76o e6e e6o to'o'o 11bo' troo' ts'oo' $4bo ts' oc te'oo 17b~o isoo RCS S01 UBLE BORON CONCENTRATION (PPM)
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=
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.___ .A_.w ;
2.0 SDM = 4.76 - 0.00260C + 0.0253P .
i when C is less than 600 PPM ;d b '
l 1.5 SOM = 3.20 + 0.0253P s_ +
l w
l 4___ when C is greater than or equal to 600 PPM ::
s_+.
- . ; {.
1.0 i Where M#
1 3
- j SDM is the required shutdown rnorgin in percent reactivity E ,
I' C is the RCS boron concentration in PPM i
0.5 g..____ ~~_*
- + g
, P is the power levelin percent of rated power :
I 11.
IIi!
I' 0.0 , , , , , , , , , , , , , , ,
0 10 0 200 300 400 500 600 700 800 9001000110012001300140015001600 f7001800 ACTUAL RCS BORON CONCENTRATION (PPM) l 1
MAINE YANKEE Required Shutdown Margin Figure Cycle 11 Versus 4.12 l
I l
_ 1
5.0 SAFETY ANALYSIS 5.1 General A review of the safety analysis for operation of Maine Yankee during ,
(
Cycle 11 is presented in this section. The parcmeters which influence the i results of the safety analysis are listed in Table 5.1. Values are provided for the Reference Safety Analysis, for Cycle 10 and Cycle 11. The Reference Safety Analysis 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 fer 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 10, Cycle 11, and
.he Reference Safety Analysis values for the parameters listed above is contained in Sections 5.1.1 through 5.1.5.
Values reported herein for MDNBR were determined using the YAEC-1 CHF correlation (44). Application of this correlation with a DN3R limit of 1.20 was approved for Maine Yankee in (45).
5.1.1 Initial Operating Cnnditions 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 11 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.
6348R/4.317
, . ~. - -. . - .
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.
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.
The hot and cold leg RTD response times, which affect the AT power input to'the RPS functions, were considered in the Cycle 11 analysis.
The safety analysis for Cycle 11 evaluated 250 plugged tubes in each steam generator for all events and accounts for various operating conditions, including coastdown out to a maximum burnup of 14,000 mwd /MT.
5.1.2 Core Power Distributions The power distribution in the core, and in particular, the peak heat flux and enthalpy rise, are of major importance in determining core thermal margin. The procedure used in the safety analysis was to set the initial j conditions (inlet temperature, power, pressure, CEA insertion, and axial power distribution) and, through analysis, assure that sufficient initial overpower margin is available to present the violation of the acceptance criteria for l
j each incident analyzed.
This procedure is continued for Cycle 11. If the available overpower margin is not sufficient for the set of initial conditions, new power 6348R/4.317 i
l L. }
d'istributions are selected, by either modifying the symmetric offset' limiting..
condition for operation (S/0 LCO) or by modifying the allowable CEA insertion limit versus power until it is demonstrated that sufficient margin exists.
As a starting point the safety analysis assumes the FSAR design power distribution (F, = 1.68 and F elta H = . shown k N gure L 2. As indicated in Table 5.9, the Cycle 11 predicted power distribution within the S/0 LCO band at the 100% power PDIL, defined in Figure 4.9, results in the lowest DNBR for Cycle 11.
In addition, values are presented in Table 5.9 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 precice indicator of relative DNB margin between power distributions than initial steady-state MDNBR. For Cycle 11, 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 11.
Power peaking associated with the CEA drop and the CEA ejection events for Cycle 11 are compared with reference alues in Table 5.1. The effect of differences between Cycle 11, Cycle 10, and the Reference Safety Analysis for l
f the CEA drop and CEA ejection are discussed in Sections 5.4.2 and 5.5.4.
5.1.3 Reactivity _ Coefficients The transient response of the reactor system is dependent on reactivity feedback effecta, 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 11 are essentially identical with those of Cycle 10. Variations in the 6348R/4.317 i
above parameters will influence each transient in a different manner.
Therefore, the effect of the difference in reactivity coefficients is discussed on an event-by-event basis.
For Cycle 11, the allowable positive values for MTC in the power range are detailed in Figure 4.8. The analyses, limited by a positive MTC, performed in (46) and Cycle 11, with the exception of the CEA ejection analysis, were conservatively performed at HFP conditions with MTC equal to
-0
+0.5 x 10 delta rho / F and bound the values in Figure 4.8. The CEA ejection analysis for Cycle 11 (Section 5.5.4) assumes the most positive of
~0 either the predicted MTC (Table 4.5) with +0.5 x 10 delta rho / F uncertainty added or the value specified in Figure 4.8. Events limited by a negative MTC are discussed in their respective sections.
The effective neutror lifetime, delayed neutron fractions, and decay constants are functions of fiel burnup and the fuel loading pattern. The Cycle 11 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, the ratio of the ejected rodworth to the effective delayed neutron fraction is sensitive in determining the course of the power response. An evaluation of this event for Cycle 11 is provided in Section 5.5.4.
5.1.4 Shutdown CEA Characteristics The negative reactivity insertion following a reactor trip is a function of the acceleration of the CEA and the variation of CEA worth as a function of position. The safety analysis considers this function in three separate parts: 1) the CEA position versus time, 2) the normalized reactivity worth versus rod position, and 3) the total negative reactivity inserted following a scram.
The CEA position versus time assumed in the Reference Safety Analysis was provided as Figure 4.2 in (36). This curve reflects a conservative rod I 6348R/4.317
l insertion time of 3.0' seconds. .This curve is based on results from plant
- measurements and is not expected to . change f rom cycle to cycle. Furthermore.
-CEA drop times are measured at each refueling as part of the startup test program to verify this assumption.
The limiting normalized reactivity worth _versus rod position was calculated for various operating conditions in Cycle 11. For each event, the Cycle 11 reactivity worth versus rod position was shown'to be bounded by the.
Reference Safety Analysis. For events which have been completely reanalyzed for Cycle 11, normalized reactivity worth versus rod position curves which are conservative relative to the Cycle 11 curves were used. The normalized reactivity worths versus rod position calculated for Cycle 11 are shown in Figures 5.3 and 5.4.
Values assumed in the Reference Safety Analysis and for Cycle 11 for the total negative reactivity inserted following a scram are given in Tables
~
5.1 and 5.11. Comparison of the scram worths assumed in the Reference Safety Analyses and the values assumed in the Cycle 11 safety analysis indicate that, with the exception of events which have been specifically reanalyzed, the Cycle 11 values are bounded by the Reference Safety Analysis values. The values of scram reactivity specified in Table 4.13 bound those assumed in the safety analysis supporting operation of Cycle 11.
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 transient, the Reactor Protective System is set to trip the reactor and prevent unacceptable core damage. The elapsed time between the time when the setpoint 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 11 are given in Table 5.10 and Figures 5.5, 5.6, and 5.7. The values for all these setpoints for Cycle 11 are either the same as or bound those used in the Reference Safety Analysis.
6348R/4.317
. . ~. _ . . _ _ _
a 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 11. Credit is taken for the functioning of the Variable Overpower !
(V0PT) Thermal Margin / Low Pressure (TM/LP) and Symmetric Of f set Trip Systems (SOTS) in several areas. First, the V0PT is credited in limiting the initial '
power distributions considered in setting the_ Symmetric Offset Trip System 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 derived from the predicted core behavior as described in (4). The Cycle 11 setpoints for the TM/LP and Symmetric Offset Trips for 3-loop operation are presented in Figures 5.5, 5.6, and 5.7. The required Symmetric Offset and the TM/LP Trip setpoints are bounded by the Technical Specifications and remain the~same as for Cycle 10. The low pressure floor of the TM/LP trip continues to be assumed to trip the reactor in the analysis of the SGTR accident, Section 5.5.2.
5.2 Summary Each transient and accident considered in (36) and (46) has been reviewed and/or re-evaluated for Cycle 11. The incidents considered are l 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.
6348R/4.317 l
- 3) -Postulated-Accidents. *
'The. incidents considered are listed in Table 5.2.
Those incidents that required a new or revised analysis included:
- 1) Boron Dilution
- 2) Excess Load Other incidents that require a partial reanalysis or review included:
'l) . Seized RCP Rotor
- 2) CEA Withdrawal
- 3) Loss of Feedwater
- 4) Loss of Coolant Flow
- 5) Steam'Line Rupture
- 6) Steam Generator Tube Rupture
- 7) LOCA
- 8) CEA Drop
- 9) Loss of Load
- 10) CEA Ejection A summary of results for Cycle 11 is presented in Table 5.3.
5.3 Anticipated Operational Occurrences for which the RPS Assures No Violation of SAFDLs The incidents in this category were analyzed in the Reference Safety Analyses for the 2630 MWt 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 incidents in this category do not violate the SAFDLs; the primary coolant system pressure limit; or the 10CFR100 site boundary dose limits. The changes considered in the present analysis do not significantly affect the NSSS response during these transients. This assures that the conclusions relative to primary system pressure and site boundary dose remain valid.
6348R/4.317
Protection against violation of the SAFDLs continues to be assured by the RPS. Setpoints are generated for the TM/LP and Symetric Offset-Trips which include the changes in power distributions associated with Cycle 11.
Sections 5.3.1 through 5.3.5 review the Anticipated Operational Occurrences for which the RPS assures no violation of the SAFDLs.
5.3.1 Control Element Assembly Bank Withdrawal The Reference Safety Analysis for this event demonstrates that the most severe CEA withdrawal transient occurs for a combination of reactivity addition rate and time in core life that results in the slowest reactor power rise to a level just below the Variable Overpower Trip. This combination of parameters maximizes the core thermal heat flux and core inlet temperature and results in the minimum DNBR.
The Reference Safety Analysis considered parametric analyses at full power (2630 MWt) for Moderator Temperature Coefficient (MTC) and Reactivity
-0 Addition Rate. The ranges evaluated 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 11 predicted value of MTC, with uncertainty, is -3.17 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 more negative MTC resulting in higher MDNBR. Table 5.1 also shows a higher maximum rate of reactivity addition for Cycle 11. Reference (36) showed that high rates of reactivity addition result in a faster rise of core power to the Variable Overpower Trip Setpoint and values of MDNBR less limiting than for slower transients.
The MDNBR for a CEA bank withdrawal event for Cycle 11 occurs from an initial power level of 1007. 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.
6348R/4.317
- 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.
Small changes in boron concentrations resulting from the Cycle 11 reload have an insignificant impact on the conclusions reached. An evaluation of this incident was performed for Cycle 11 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 11.
i' 5.3.2.1 Dilution During Refueling Assumptions made in the Cycle 11 evaluation for dilutions during refueling are consistent with those made in (36) and (47). i 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 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 11 core loading, the critical boron concentration under cold conditions (68 F) during refueling is 1346 ppm with the two most reactive rods withdrawn or 905 ppm for ARI. The minimum initial reactor vessel boron concentration which will prevent an inadvertent criticality within 30 minutes is 1979 ppm with the two most reactive rods withdrawn or 1331 ppm for ARI (Case No. 3 Dilution, Reference (47)).
634SR/4.317 i
1
Therefore, it is cencluded that if the reactor vessel boron concentration-is maintained at or greater than 1979 ppm with the two most reactive rods withdrawn or 1331 ppm for ARI during Cycle 11 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 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 11. The limiting 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.
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 required to meet the Technical Spec fication 5% delta K/K i
suberiticality requirement for shutdown ennditions. The boron concentrations 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.
5.3.2.3 Dilution During Cold, Transthermal, and Hot Shutdown with Drained RCS
' Conditions Dilutions during shutdown conditions with the RCS partially drained l were addressed in (47) and (48). In order to conservatively bound any partially drained configuration with one or more reactor coolant loop l 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 a maximum 1%
delta K/K of bank withdrawal, all CEAs are assumed to be inserted in the l
l core. The limiting dilution in this situation is Case No. 3 of (47).
6348R/4.317 l
l l
E -
t 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. Table 5.5 also shows the boron concentrations required to meet the 5% delta K/K Technica1' 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 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 11. However, the hot standby critical boron concentration with uncertainty is higher, 1758 ppm versus 1571 ppm. The results for Cycle 11 using Figures 4.3-4 and 4.3-5 of (36) are summarized below:
Maximum Reactivity Insertion Rate Dilution at Hot Standby 11.12 x 10-6 delta rho /sec
-6 Dilution at Power 9.46 x 10 delta rho /sec 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 the maximum reactivity addition rate it would take
- approximately 53 minutes of continuous dilution at the maximum charging rate l
l to completely ab. orb 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.
6348R/4.317
5.3.2.5 Failure to Borate Prior to Cooldown Because of the large negative moderator temperature coefficient at EOC, 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.
(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 temperature must be reduced from 550 F to about 448 F. This temperature reduction requires approximately 61 minutes to 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 mismatch is established leading to a decrease in the reactor coolant average temperature and pressure. Hence, when the moderator temperature coefficient of reactivity is negative, unintentional increases in reactor power may occur. Thus, the Excess Load Incident as reported in (36) was analyzed over a wide range of power levels and negative MTCs to determine the minimum margin to the LHCR and DNBR SAFDLs.
6348R/4.317
The' Cycle 11 MTC with uncertainty is bounded by the Reference Safety Analysis value of -3.17 x 10~ delta rho / F.
A general approach to the excess load transient is used which credits RPS functions only. ' The most severe consequences f rom an excess load event result from an inadvertent opening of the Steam Dump and Bypass System near full power or an inadvertent opening of the turbine admission valve without credit for the turbine load limits at. lower power levels. 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 terminate when the delta-T power or excore power signals reach the symmetric offset and/or variable overpower trip limit. Since the excore power signal is decalibrated by the transient-induced cooldown in the downcomer and the delta-T power signal lags slightly behind the power excursion, the peak power achieved durink this transient can overshoot the RPS trip setpoints. The maximum power varies as a function of MTC, so a wide range of negative MTCs are considered. For Cycle 11, the peak power and closest approach to the LHGR SAFDL occurs at MTC values expected mid-cycle when the transient is initiated near full power and at the end of cycle for lower powers. The minimum DNBR occurs at the end of cycle at all powers because the peak power coincides with the minimum DNBR benefit from the plant cooldown. 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 near full power which results in a power increase to the variable overpower trip setpoint. The closest approach to fuel centerline melt corresponds to an event initiated from the negative edge of the symmetric offset band near full power resulting in a power increase to the variable overpower trip setpoint.
5.3.4 Loss of Load Incident A Loss of Load transient occurs when the turbine trips while the plant is at power.
6348R/4.317 l
l-
b For Cycle 11, no changes have been implemented which affect this transient. The number.of steam generator tubes that could be plugged and still remain below a peak system pressure of 2750 psia is 250 tubes per steam generator.
The Loss of Load transient MDNBR is :1.20.
5.3.5 Loss of Feedwater Incident A Loss of Feedwater transient occurs when the main feedwater supply to the steam generators is discontinued while the plant is at power.
Peak 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 was performed to determine the MDNBR for Cycle 11. 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 (10CFR100) under Cycle 11 conditions.
l I
6348R/4.317 i
s
'5.4.1 Loss-of4 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 11 (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 >1.20.
5.4.2 Full Length CEA Drop The drop of a full length CEA results in a distortion of the core power i
. 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 I
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 11 was evaluated from power levels ranging f rota 0% to 100% of 2630 MWt.
i Power distributions used in the evaluation of DNBR and proximity to ,
- j. 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 11 is given in Figure 4.10. The value for the maximum increase in peaking for any dropped CEA from Figure 4.10 was conservatively l
(Section 4.9.3.1) applied at each power level considered.
l l
l l
6348R/4.317 i
-- -n, , , , - . , , - , , . . , - - - - - . . - . - , , - ,.-e.,,,-.,,. - - . , - , _ _ . - , , , _ , - _ _ _ , _ , , , , _ _ . _ , , . . , - , , _ , _ - , , , ,
The CEA drop analysis also considers the increased peaking which 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.
The percent increase in peaking from Figure 4.10 was conservatively augmented by the increase in peaking due to xenon redistribution at subsequent points in-time, assuming operation consistent with the power level reductions required by the. Technical Specifications. The margins to the SAFDLs were then determined for the limiting power distributions within the symmetric offset LCO band allowed for the existing power level at any point in time assuming the CEAs to be inserted no deeper than allowed by the insertion limit 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 tae maximum increase in peaking. Thus, for conservatism the plant response assumed in the Cycle 11 '
evaluation was based on a worth of 0.10% delta rho.
The results of the DNB evaluation for Cycle 11 indicate that the limiting DNBR during a full length CEA drop occurs 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after a drop from 100%. 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 1
melt is one initiated from power distributions at the edge of the S/0 LCO 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 L 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 the influence of the moderatoc reactivity feedback induced by the lowering of the core average tempetature by continued heat removal to match the constant steam demand of the t-*' ,
throttle rsetting.
-h6- l 6348R/4.317 i __
, . .=
During operation with turbine admission valve control in Impulse IN sede (IMPIN), the admission valves automatically react ' to changes in steam flow or pressure to maintain impulse pressure at the inlet to the first stage of the high pressure turbine ~ constant. Thus, if the inlet steam flow and pressure were to drop, as they do immediately after a CEA drop, the throttle valves would open in an attempt to restore the impulse pressure to its set ,
value. While operating in the IMPIN mode the potential exists to exceed the predrop power level, 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. This could cause core power to return to values higher than the predrop level.
l 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 operation with respect to core power distributions (i.e., symmetric offset LCO band) is determined by the locus of points at which the DNB or LHCR SAFDLs would not exceed their limits following a CEA drop. Any decrease in margin due to power levels returning above the predrop level af fects this LCO.
Consequently, a suitably conservative S/0 LC0 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 LC0 bands are shown in Figure 5.7.
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 l
6348R/4.317
cnalyses reported when Cycle 11 conditions warranted reanalysis of the tecident. a 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 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 Part 100.
A system analysis was not required for Cycle 11 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 EOC, along with the nominal available scram reactivities for Cycle 11. The most limiting case for
! Cycle 11 was HFP at E00. The minimum margin for the HFP at EOC case is 0.55%
l 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 determined in Table 4.12. The required scram reactivities calculated for Cycle 11 are accounted for in the shutdown margin Technical Specification.
l l l l 6348R/4.317 ,
t
-.n .n- wer ,,,.,-.-.e ..nn- --p. , .
-+-r.,,.e-y.e,-r -
r-- ., m,, . . , , - - , - - . - ,n-.,wm-e e-,- , - a,--,-~ --, ,, ,., + e v --~'
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 11. The primary system response is mainly a function of the initial system pressure and the time of reactorLtrip. 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
[ to SGTR during Cycle 11.
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 power 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 11.
The important differences for Cycle 11 are a reduction in initial overpower DNB margin due to changes in the radial pin power census and a more limiting 100% power PDIL power distribution. The MTC at full power for Cycle 11 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 11. The results for Cycle 11 are included in Table 5.3.
l They indicate a slight increase in the fraction of fuel failure predicted for d
Cycle 11 as compared to Cycle 10. 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 11 are acceptable. .
i
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6348R/4.317 t
t
. , , .,--,,--.-.----y,, -- - - , - ~ - _ ,.,,% , .,-y-- p. .__. .,,_,,-.-,,r, . - - -
5.5.4 C16 +.=y 'on
'tuences of a CEA ejection' accident are most sensitive to eject..d. > .arth, effective delayed neutron fraction (beta effective), and
- post-ejected peaking. Specifically, the. severity of the transient increases for higher ejected CEA worths, smaller delayed neutron fractions, and increased post-ejected peaking. A comparison between the values assumed in the Cycle 10 Reference Safety Analysis, and those predicted for Cycle 11, including uncertainty, is presented in Table 5.1. In each case, the values for Cycle 11 are bounded by all of the parameters from Cycle 10. Thus, a reanalysis of each case is not necessary for Cycle 11.
A summary of the bounding results from Cycle 10 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 5.5.5.1 Introduction and Summarv Large break Loss-of-Coolant Accident (LOCA) calculations for Maine Yankee were performed for Cycle 5 through Cycle 9 using the Yankee Atomic Electric Company (YAEC) WREM-based generic PWR ECCS evaluation model (1). The Cycle 5 calculations consisted of a complete break spectrum analysis and the calculation of cycle-specific limits. Cycle 6 through Cycle 9 evaluations I
- demonstrated that the Cycle 5 break spectrum analysis was applicable and the l
- . results of this analysis were then used in calculating the LOCA limits for each cycle.
For Cycle 10 and subsequent analyses, the YAEC LOCA methodology was modified to include a more complete spectrum of possible power shapes than that previously analyzed. This model improvement required break spectrum t
! analyses for four different power shapes. The results of the break spectrum i
6348R/4.317 L_
r.
\ cnalyses were then used in calculating the LOCA limit curve for Cycle 10.
/ Detailed discussions of the break spectrum analysis and the LOCA limit calculations for Cycle 10 are provided in (55) and (56), respectively.
h The Cycle 10 LOCA analysis serves as a reference LOCA analysis for Cycle 11. A quantit:tive review of the reference analysis, presented ir *,he followirg sections, indicates that the Cycle 10 break spectrum analysis is opplicable~to Cycle 11.
For the. limiting breaks, determined in Cycle 10 break spectrum analysis, LOCA calculations were performed with Cycle 11 specific input.
Based on the results of-these analyses, it is concluded that Appendix K criteria are met-for the Cycle 11 fuel types operating within the linear heat generation rate limits specified in Figure 5.8. Moreover, it should be noted that the Cycle 11 LOCA limit curve is identical to that for Cycle 10; no effort was made to further optimize this curve.
5.5.5.2 Large Break LOCA Analysis To determine the LOCA limit at an axial location in the core, a limiting axial power shape with its peak at the specified location is used in the LOCA analysis. If the results of the analysis meet the LOCA design criteria, the Peak Linear Heat Generation Rate (PLHGR) is the derived limit at
! the specified location. The process is repeated for several axial locations so that a curve through the established PLHGRs forms the axially dependent LOCA limit curve.
For Cycle 11, as in Cycle 10, the LOCA limit curve was generated by calculating the limits at four axial locations: 52%, 65%, 73%, and 85% of the core height. To calculate the limits at the four specified locations, the limiting break for each limiting power shape needs to be identified. A brief discussion of the break spectrum analysis follows.
6348R/4.317
5.5.5.2.1 -Break Spectrum Analysis-The break spectrum analysis performed for Cycle 10 was found to be
~
cpplicable to Cycle 11. As in Cycle 10, the Cycle 11 core is predominantly fuel supplied by CE so that the hydraulic characteristics of the core have not changed significantly.
Of the reactor physics parameters used in the LOCA analysis, the Moderator Density Defect (MDD) has the greatest impact on the LOCA results.
The variation of MDD between Cycles 10 and 11 is less than 2%, which is well within the data uncertainty range. The remaining physics parameters in the Cycle 10 reference analysis are representative of Cycle 11. Therefore, the Cycle 10 break spectrum analysis results are applicable to Cycle 11. These results are presented in Table 5.12.
5.5.5.2.2 LOCA Limit Caltulations The Peak Linear Heat Generation Rate (PLHGR) values used in the break spectrum analysis are repres*nt.tive values. The PLHGR limits are optimized in the LOCA limit calculations. For each of the limiting breaks identified in the break spectrum analysis, a LOCA calculation was performed with input data specifically for Cycle 11. The results of the analysis for each axial power shape are provided in Table 5.13. The calculated cladding temperature, cladding oxidation values, and hydrogen generation results demonstrate compliance with Appendix K criteria. The PLHGR values are plotted as a function of core height in Figure 5.8. The LOCA limit is assumed to be constant between 0% and 52% core height, linear between the analyzed points, and is determined assuming a conservative fall-off rate between 85% and 100%
core height. These assumptions are discussed in detail in Reference (57).
5.5.5.2.3 Sensitivity of LOCA Limits to Radial Peaking Factor Calculations were performed to determine the sensitivity of LOCA limits to radial peaking factor. This sensitivity study showed that the LOCA limits are inversely proportional to the radial peaking factor and the relationship is approximately linear. A calculation with 3% higher radial peaking factor i
6348R/4. 317
^
\
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, - - - , - - - - , _ , . - _ , , _ . ,,,,7_, . , _ . _ . . , , , , , , _ _ . , , , , _ , , , ,, , . , , , , , _ , , , _ _ , _ , , . , , _ . .
l l
cnd 3% lower LOCA liraits (PLHGRs) resulted in PCTs below 2,200 F and comparable to those given in Table 5.13.
5.5.5.3 Small Break LOCA Analyses The small break LOCA analysis performed by Combustion Engineering for Cycle 4 considered a spectrum of cold leg breaks varying in size from 0.1 to 0.5 ft2 (22). Results showed that the limiting break size is the 0.5 ft 2 break with a peak clad temperature of 1,348 F, well below the acceptance criteria of 10CFR50.46. A demonstration analysis of the limiting break performed for Cycle 5 (1) utilizing YAEC methodology yielded a peak clad temperature of 1,230 F, well below the 10CFR50.46 acceptance criteria and Maine Yankee large break results. In that analysis, a 68% peak top skew design shape and a PLHGR of 16 kW/f t were used. The analysis predicted a short period of core uncovery and resultant cladding heatup. Thus, small break LOCAs for Maine Yankee were shown to be nonlimiting.
The results of previous analyses are applicable to Cycle 11 because they are determined primarily by the decay heat values which are insensitive to fuel type. Additionally, slight differences in Cycle 11 and Cycle 5 system configuration would not significantly affect on the PCT which was predicted to be well below the 10CFR50.46 criteria. Hence, the minor system changes will not make the small break LOCA a limiting scenario.
l l
6348R/4.317
TABLE 5.1 Maine Yankee Safety Parameters Assumptions Cycle 3 Cycle 10 Cycle 11 Includir.g Including Including Uncertainties Uncertainties Uncertainties Parameter Units
- Planar Radial Peaking Factor 1.68(2)(3) 1.85 1.79 Bank 5 Inserted to 100% PDIL 1.72(7)
- Axial Peak for Shape Resulting in MDNBR at 100% LI? 1.42(2) 1.42 1.50
- Augmentation factors 1.0 to 1.067(1) None(10) None(10)
- Moderator Temperature Coefficient 10-4 delta rho /0F 0 to -2.74 +.5 to -2.96 +.80(11) to -3.17
- Ejected CEA Worth BOC 2ero Power % delta rho .396 .718 .368 BOC Full Pcuer % delta rho .210 .371 .259 EOC Zero Power % delta rho .544 .627 .502 EOC Full Power % delta rho .230 .465 .350
- Ejected CEA 3D Peak P.0C Zero Power 13.32 16.34 10.68 BOC Full Power 5.53 6.26 6.03 EOC Zero Power 14.08 15.18 13.00.
EOC Full Power 5.59 6.44 6.11
- Droyned CEA Integral Worth % delta rho O to .30 0 to .20 0 to .20 6348R/4.317 G --___s w
TABLE 501 (Continued) ,
Cycle 3- Cycls 10 Cycle'll
- . Including. Inclu11ng Including 2
Parameter Units Uncertainties Uncertainties- Uncertainties Dropped CEA Integral . Figure 4.4-1 of. Figure 4.11 of Figure 4.10 j Radial Peak Reference 3- Reference 54 i
, - Power Level (including 2% uncertainty) MWt 2683 2683 2683 4-I
- Maximum Reactor Coolant Inlet Temperature OF 554 548 - 556 548 .556 i
j - Reactor Coolant System Pressure psia 2200 - 2300 2050 - 2300. 2050 - 2300.
I
- Reactor Coolant System Flow Rate 106 3 Sr 134.6(5) 134.2(5)-135.8(6) 134.2(5)-135.8(6)
- Axial Power Distribution S, .aric Figure 6.3-1 Figure 5.7 of Figure 5.7.
Limit Offset- of Reference 3 Reference 54
- Power Dependent Figure 4.9 of . Figure 4.9 of Figure 4.9 j Insertion Limit Reference 20. Reference 54 ~_,. ,
- Initial Steady-Sta!.e(4) 1.897(5) 1.894(5)
Minimum DNB Ratio YAEC-1 1.977 1.915(6) 1,907(6)
- Maximum Possible Rate of Reactivity Addition (9) delta rho /sec 0.7x10-4 1.29x10-4 1.36x10 !
Nominal
- Steam Generator , psia 877 877 - 610 877 - 610(12) j Pressure (100%)
I o q
- 5 6348R/4.317 i
t
1 , r TABLE 5.1 (Continued) ..
~
Cycle 3 Cycle 10 Cycle 11 Including Including Includin',
Parameter Units Uncertainties Uncertainties Uncertsiaties
- Nominal Steam Generator Level (Narrow Range) % 66 66 - 48 66-48 Steam Generator (SG)
Tubes Plugged /SG --
180 250
- Minimum Required Worth in CEAs .
Ass;.,med in Safety Analysis % delta rho 5.70 HFP, BOC 4,, 9 5.70 il2P, BOC 2.0 3.20 3.20 HFP,EOC ~
5.7 5.85 6.94 6.5(8)
Il2P,EOC 2.9 3.22 4.75 6348R/4.317
l TABLE 5.1 (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 sht in Reference 12 did not include. uncertainty.
. 4). FSAR design power distribution (F kelta H = 1.49 Fz = 1.68).
Includes,2% calorimetric power uncertainty and 3% allowance for maximum tilt allowed by Technical Specification 3.10.
- 5) Based on reactor Coolant System pressure of 2200 psia, and temperature of 5560F.
-6) Based on Reactor 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) E00, HFP steam line break assumed 6.5% delta rho.
9). For CEA bank withdrawal transient.
- 10) Augmentation factors were removed in Cycle 10 (54).
- 12) The lower pressure is evaluated to account for coastdown conditions.
l-l 6348R/4.317
TABLE 5.2 Maine Yankee
-; Cycle 11 - Incidents Considered
- A. Anticipated Operational Occurrences for which the RPS assures no violation of SAFDLs:
- 1. Contro1' Element Assembly Bank and Subgroup Withdrawal
- 2. Boron Dilution
- 3. -Excess Load 4.- Loss'of Load
- 5. Loss of Feedwater B. Anticipated Operational Occurrences which are dependent on Initial Overpower Margin for protection against violation of SAFDLs:
- 1. Loss of Coolant Flow -
- 2. Tull Length CEA Drop C. Postulated Accidents:
- 1. CEA Ejection
- 2. Steam Line Rupture
- 3. Steam Generator tube Rupture
- 4. Seized Rotor
- 5. Loss of Coolant i
l i
l l -
6348R/4.317 f
I-_ .]
TABLE 5.3
~
Maine Yankee Cycle 11 Safety Analysis Summary of Results Reference Safety Analysis Cycle 10 Cycle 11 Incident Section Criteria Cyc?e 3 MDNBR = 1.51* MDNBR = 1.47 MDNBR = 1.37 CEA Withdrawal 5.3.1 MDNBR = 1.20 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 Subcritical: Subcritical: Subcritical:
Boron 5.3.2 Subcriticci: Refueling-30 mia.
Sufficient time Refueling-65 min. Refueling-30 min.
Dilution Startup-32 minutes Startup-15 minutes for operator Startup-3.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action Critical: Bounded Critical: Bounded Critical: Bounded l by CEA withdrawal by CEA withdrawal by CEA withdrawal Critical: MDNBR 1.20 Cycle 3 MDNBR = 1.36* MDNBR = 1.42 MDNBR = 1.37 CEA Drop 5.4.2 MDNBR = 1.20 Not exceeded Not exceeded Not exceeded.
LHCR SAFDL Cycle 3 MDNBR = 1.38 MDNBR r- 1.32 Less of 5.4.1 MDNBR = 1.2 MDNBR = 1.50*
Coolant Flow Cycle 5 10CFR100 8.4% of rods with 10.3% of rods with 10.8% of rods with Seized Pump 5.5.3 I
MDNBR less than 1.3* MDNBR less than 1.2 MDNBR less than 1.2 -
I Rotor Radiological dose Radiological dose Radiological dose-within 10CFR100 within 10CFR100 within 10CFR100 l
l Cycle 4 MDNBR = 1.7* MDNBR = 1.42 MDNBR = 1.33 Excess Load 5.3.3 MDNBR.= 1.2 i
6348R/4.317 l
TABLE 5.3 (Continurd)
Incident Section Criteria Reference Safety Analysis Cycle-10 Cycle 11 Cycle 3 Loss of Load 5.3.4 MDNBR = 1.2 MDNBR = 1.85* MDNBR = 1.92 MDNBR = 1.79 RCS pressure RCS pressure RCS pressure RCS pressure
<2750 psia 2639 psia <2750 psia <2750 psia Cycle 3 Loss of 5.3.5 RCS pressure Peak RCS pressure Peak RCS pressure Peak RCS pressure Feedwater <2750 psia 2600 psia <2750 psia . <2750 psia.
MDNBR = 1.20 MDNBR = 1.61 MDNBR = 1.54 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 ator Tube within 10CFR100 within 10CFR100 within 10CFR100 Rupture Cycle 10 CEA Ejection 5.5.4 10CFR100 Radiological dose Radiological dose Radiological dose within 10CFR100 within 10CFR100 within 10CFR100 LOCA 5.5.5 10CFR100 Cycle 10 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 unchanged by Containment Cycle 10 reload Cycle 11 reload
-100-634RR/4.317
TABLE So3 (Continued)
Incident Section Criteria Reference Safety Analysis Cycle 10 Cycle 11 Cycle 3 Feedwater -
10CFR100 Bounded by steam line. Reference analysis Reference analysis
, Line Rupture rupture unchanged by unchanged by Outside Cycle.10 reload Cycle 11 reload.
Containment Cycle 3 Containnent -
Peak pressure Peak-pressure less Reference analysis Reference analysis
- Pressure less than 55 psig _than 55 psig unchanged by unchanged by.
- containment Cycle 10 reload Cycle 11 reload j design pressure
- Cycle 3 Fuel Handling -
10CFR100 Radiological dose Reference analysis. Reference analysis Incident within 10CFR100 unchanged by unchanged by
) Cycle 10 reload Cycle 11 reload
$ Cycle 3 Waste cas -
10CFR100 Radiological dose Reference analysis Reference analysis System within 10CFR100 unchanged by. unchanged by
,t Failure Cycle 10 reload Cycle 11 reload i
j Spent Fuel -
10CFR100 NA NA NA j Cask Drop j
i Cycle 3 i Radioactive -
10CFR100 Radiological dose Reference analysis Reference analysis.
i Liquid Waste within 10CFR100 unchanged by unchanged by i System Leak Cycle 10 reload Cycle 11 reload l
- W-3 DNBR shown for Reference Safety Analysis result, YAEC-1 for Cycles 10 and .11.
-101-6348R/4.317 4
TABLE 5.4 Cycle 11 Required Initial RCS Boron Concentrations to Allow 15 Minutes. Margin to Criticality for Dilutions from Shutdown Conditions With the RCS Filled Required Initial Concentration (ppm)
' Boron Dilution 5% Delta K/K*
ARI, BOC 5320F 907 1047 3000F 1049 1140 680F 1062 1137 ARI, E00-5320F 0 0 3000F 91 213 i
680F 169 269 ASRI..Less 1 Stuck CEA, B0C 5320F 1175 1303
'3000F 1296 1377 680F 1338 1400 ASRI Less 1 Stuck CEA, E00 5320F 13 186 3000F 301 418 680F 366 462
.ASRI w/ Withdrawn Bank,**
BOC 5320F 1453 1368 3000F 1555 1442 680F 1542 1427
-102-6348R/4.317
L.
TABLE 5.4 (Cont'd)
Required Initial Concentration (ppm)
Boron Dilution 5% Delta K/K*
j; ASRI w/; Withdrawn Bank,**
'EOC 5320F. 238 267 3000F 484 480
-680F 528 514 ARO, BOC 5320F 1901 2082 3000F 1789 1922 680F 1707 -1821 AR0, E00 5320F 635 849 3000F 690 848, 680F 667 802' i
O Margin of suberiticality required by Technical Specifications for shutdown conditions.
C* The critical boron concentrations for 2 stuck CEAs bound those for a withdrawn bank. Therefore, the boron concentrations provided for 2 stuck CEAs were used in the calculations of required concentration for boron dilution events. The 2 stuck calculations also bound intermediate
-combinations of 2 or more CEAs withdrawn during CEA rod drop testing. ;
-103- l 6348R/4.317
TABLE 5_5 Cycle 11 Required Initial RCS Boron Concentrations to Allow 30 Minutes Margin to Criticality for Dilutions from Shutdown Conditions With the RCS Drained
- l Required Initial Concentration (ppm)
Boron Dilution 5% Delta K/K*
ARI, i BOC 5320F 1389 1047 3000F 1476 1140 680F 1444 1137 ARI, E0C 5320F 0 0 3000F 128 213 680F 229 269 ARI, Less 1 Stuck CEA, BOC 5320F 1800 1303 3000F 1825 1377 680F 1821 1400 ARI, Less 1 Stuck CEA, E0C 5320F 20 186 3000F 423 418 680F 499 462 ARI w/ Withdrawn Bank,***
BOC 5320F 2225 1368 3000F 2190 1442 680F 2097 1427
-104-6348R/4.317
TABLE 5.5 (Cont'd)
Required Initial Concentration (ppm)
Boron Dilution 5% Delta K/K**
ARI, w/ Withdrawn Bank,***
EOC 5320F 364 267 3000F 682 480 680F 718 514 AR0, BOC 5320F 2911 2082 3000F 2519 1922 680F 2322 1821 ARO, EOC 5320F 972 849 3000F 971 848 680F 907 802
- Level = lower lip of RV nozzles.
}
- Margin of suberiticality required by Technical Specifications for shutdown conditions.
combinations of 2 or more CEAs withdrawn during CEA rod drop testing.
-105-6348R/4.317 a.
TABLE 5.6 Summary of Boron Dilution Incident Results
'- for Cycle 11
( (A) (B) (C) (D) l Minimum Technical Minimum Time Operating Specification Shutdown to Absorb (B) Acceptance 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 Drained RCS 5% delta K/K(1) 30(1) 30(1)
Startup W/2 Rods Withdrawn 3.2% delta K/K(6) 32(2) 15 ARI 3.2% delta K/K(6) 161(2) 15 Hot Standby 3.2% delta K/K(6) 53(3) 15 Power Operation 3.2% delta K/K(6) 53(3) 15 Failure to Borate Prior to Cooldown 3.2% delta K/K(6) 61(4) 15 (1) 30 minutes margin is used to provide sufficient margin for drained conditions where the head is removed. These are classed as "refueling" conditions in the Technical Specification. Margin quoted is for initial boron concentrations administrative 1y required for these conditions.
(2) Margin quoted assumes initial boron concentration at refueling value for Cycle 11 of 1979 ppm for 2 rods withdrawn and 1331 ppm for ARI.
(3) Time to absorb minimum specified 3.2% delta K/K shutdown margin.
(4) Cooldown rate assumed to be 1000F/hr.
(5) Time span between event initiation and criticality.
(6) Minimum value for shutdown margin specified in Technical Specification.
-106-6348R/4.317
TABLE 5.7 Nominal Scram Reactivity Worths Required to Prevent a Return to Power During a Steam Line Rupture Accident i Nominal Scram Reactivity (% Delta Rho)
Case Required Available HFP BOC 3.29 7.71 HFP E0C 7.71 8.26 HZP BOC 1.87 6.05 HZP EOC 5.28 6.37
-107-6348R/4.317
_ TABLE 5.8 Cycle 11*
I CEA Ejection Accident Results Full Power BOC EOC Fraction of Rods that Suffer Clad 3.6 0.0 Damage (Radial Average Enthalpy Above 200 cal /gm), %
Fraction of Fuel Volume Exceeding 0.0 0.0 Incipient Melting Criteria s (Enthalpy greater than 250 cal /gm), %
Zero 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), % .
1P s
- These results are derived from the bounding Cycle 10 CEA ejection analysis.
-108-6348R/4.317
TABLE 5.9 Comparison of Thermal Margin for Limiting Cycle 11 Power Distributions to FSAR Design Power Distribution Power Power YAEC-1( } YAEC-1(
Level Distribution MDNBR (Pd-Po) 100 FSAR(3) 1.894 28 100 ~(2)(3) 1.856 25 93 (2)(3) 1.957 29 79 (2)(3) 2.296 39 66 (2)(3) 2.655 40 53 (2)(3) 3.122 46 100 FSAR(7)(4) 1.897 28 100 FSAR(7)(5) 1.962 34 100 FSAR(7)(6) 1,919 31 100 FSAR(7)(8) 1.911 31 (1) Includes allowances for 2% calorimetric power uncertainty, 3% tilt, and 10% physics radial peaking uncertainty on non-FSAR cases.
(2) Limiting Cycle 11 power distribution within symmetric offset pretrip alarm band plus uncertainty for indicated power level for Cycle 11.
(3) At 2200 psia, 556 degrees F.
(4) Cycle 10.
(5) Cycle 3.
(6) . Cycle 5.
(7) At 2200 psia, 554 degrees F.
(8) Cycle 6
-109-6348R/4.317
TABLE 5.10 Reactor Protective System Trips Assumed in the Cycle 11 Safety Analysis Setpoint Uncertainty Time (Sec)
High Neutron Flux 106.5% ! 5.5% 0.4 Low Reactor Coolant Flow 93% 12% 0.65 High Prassurizer Pressure 2400 psia !22 psi 0.9 Low Steam Generator Pressure 500 psia 122 psi 0.9 Low Steam Generator Water Level 35% NR 210 in 0.9 Low Pressurizer Pressure
- 1850 psia 122 psi 0.9 1600 psia !22 psi **
Safety Injection Signal Thermal Margin Trip (TM/LP) Figure 5.5 (4) 0.9 and 5.6 Symmetric Offset Trip (SOTS) Figure 5.7 .04 asiu 0.9 Variable Overpower Trip (V0PT) Q + 10%*** ! .5%
5 0.4
- Low limit of thermal margin trip.
- See specific accident for time delay assumed for safety injection delivery.
- Q = Initial indicated power level in percent thermal or nuclear power.
-110-6348R/4.317
TABLE 5.11 Scram Reactivity Assumed in l Cycle 11 Safety Analysis Scram Reactivity (% delta rho) l
(
Event B0C, HZP EOC H2P BOC, HFP EOC, HFP CEA Withdrawal - -
4.00 4.00 Boron Dilution 3.20 3.20 - -
Excess Load 3.20 4.50 5.70 6.93 Loss of Load - -
4.00 -
Loss of Feedwater - - 4.00 -
Loss of Coolant Flow - -
5.50 -
CEA Drop - - - -
CEA Ejection 3.20 3.20 5.70 5.70 Steam Line Rupture
- 1.68 4.75 2.96 6.94 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 4.75 5.70 6.94
- An uncertainty factor of 0.9 is applied to the nominal required scram reactivities assumed for the Steam Line Rupture event from Table 5.7 for comparison to the available scram reactivities with uncertainties assumed-for the other events. This uncertainty component is statistically combined in the Steam Line Rupture analysis with the other uncertainty components to derive the nominal required scram reactivities for that event as discussed in (14).
-111-6348R/4.317
TABLE 5.12 Break Spectrum Analysis Results PLHGR Axial Power Shape (kW/ft) Limiting Break 52% 16.0 0.8G 65% 13.5 0.8G -
73% 13.0 1.0S 85% 11.5 1.0S gA OL t
e-g
-112-6348R/4.317
TABLE 5.13 Maine Yankee Cycle 11 Large Break LOCA Analysis Results Cladding Axial LOCA Peak Cladding Power Limiting Limit Temperature Oxidation Shape Break (kW/ft) ( F) (% Thickness) 52% 0.8G 16.0 1982 1.6 65% 0.8G 15.0 2107 2.5 73% 1.0S 14.4 2127 3.2 85% 1.0S 13.0 2009 2.3 Notes (1) All analyser are performed at beginning-of-life conditiore.
(2) Less than one percent hydrogen generation is predicted in all ,
analyses.
-113-6348R/4.317 h
l
\ .
[
l 2300 . .. , , , ,, , , ,
l III UNACCEPTABLE OPERATION l l i , il i, i i I ! i i i 2
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UNACCEPTABLE OPERATION l 2050 490 500 SM 520 530 540 650 560 Nominal Cold Leg Temperature (indicated), deg F MAINE YANKEE ARoweble 3 Loop Stoody State Figure Coolont Conditions 5.1
-114-
\I 8 7 2
, 6 4 1 1 1 1 0
- 1
, - - . 9 U
8 U
7 0
t 6 h 0 g i
e H
n e r
i o o
,t 5 u
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z iou g 0 lPb a 0 3 0 3 i
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mlat r ris oxi .
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n E n ! *'8' $ ,_w o N e. m ! 9.". & Cgm . .
~
w$. @D-
. - 'Y ,.
f jl\' 1
FIGURE 5.3 MAI'NE YflNKEE CYCLE 11 .
NORMALIZED REACTIVITY WORTH VS PERCENT CEA INSERTION BOC SCRAM AT RFP RND HZP
, 3,g , .............. ......... .... . .... ...........................................................
i :
0.9 i )
0.8 i.
7 e
0.7 h h
o . .
2 : : -
s : :
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6 -4 - -
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e :
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\ .
0.1 p 0.0 ............. .. ........, . . ..... ,,... . ....... ... ,,,,,...... .. ,,,,,,,, . ,,.. ,, .
0 10 20 30 40 50 60 70 80 90 100 -
PERCENT CER INSERTION
-116-
FIGURE 5.4 MAINE YANKEE CYCLE 11 ;
NORMALIZED REACTIVITY WORTH VS PERCENT DER INSERTION EOC SCRAM RT HFP AND HZP
.... ... ......... .........i....................................... .. ...... ......... .........
1,g i :
0.92 'r 1
0.8 i -
I s 0.74 :
e : :
o .
A 5 5 b
U H
0.6 k -
s : .
o : :
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y : :
e o : :
u -
i N 0.45 s
a : .
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r e :
@ 0.3-' r 0.2 ; HFP HZP i i
\ :
0.1-~
0.0 .........,. . ............ ........,,,,.......,...........,......... .. ....,,,..... ........
s 0 10 20 30 40 50 60 70 80 90 100 PERCENT CER INSERTION
-117-
I ,
l I
ww.RE: g-g.g T - 10053.0 NO P , RIP = 2014.3'Qg 17.9TC TC " "D
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-118-
i mat: o, - 4, . on, TRIP = - 10053.0 AND Pg 2014.3 Og+ 17.9T C T
C = CM LEC WNM l
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- Froction of Roted Thermal Power Figure Thermot Margin / Low Pressure 5.6 MAINE YANKEE Trip Setpoint Port 2 (ort versus Froction of Roted Thermal Power)
-119-
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.x 40 ,
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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 0.6 Symmetric Offset Y, = A*(U-L/U+L)+B Symmetri: Offset Trip Function Figure MAINE. YANKE.E Three Purnp Operation 5.7 p.
-120-
t 20
{O 1918 -
UNACCEPTABLE OPERATION Q- 16 :' m 15 '=-
_=
gu N g 13 6CCEPTABLE OPERABON _
E 12-
~
N 11 \ ,
d 10- .
I)
$9 e COORDINATES
~~~~~
( 0,16.0) "~~~
7 ---
( 52,16.0) . . ___
6 ( 6 5,15.0) s 5 = ( 7 3,14.4) 7_
64 I
( 85,13.0) g 3 ( 9 3, 11.8)
(10 0, 9.2) 62-Z 1
3 0
0 10 20 30 40 50 60 70 80 90 10 0 CORE HEIGHT (%)
MAINE YANKEE Unear Heat Generation Rate (LHGR) Umits Cycle 11 Versus 5.8 Core Height
-121-
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 l l
anomalies present which could cause frelems later in the cycle;.
- 2) Verifying that the calculational model used will correctly predict core behavior during the cycle.
The low power physics tests are conducted at a power level less than 2.0% of rated full power with a primary system temperature and pressure of j spproximately 532 F and 2250 psia, respectively.
4 .
6.1 Low Power Physics Tests 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 non-overlap condition. The total of the worths of these groups mrwt be within 110% of the predicted value.
If 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 110% of predicted.
- 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.
- 4) CEA drop times are measured by monitoring reed switch voltage for position ind! cation versus time. All scrammable full length CEA drop times are measured.
-122-6348R/4.317
, _ - = 3 6.2 Power Escalation Tests-Power escalation tests assure the performance of vsrious primary and secondary plant systems. Plant parameters are stabilized and test data taken at approximately 48% and approximately 100% of rated power.
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.
- 2) Isothermal temperature coefficients, if required, are derived at 48% power by partially closing the steam turbine governor valves which '.ncrease reactor coolant system temperature. The resnit is a chant,e in moderator temperature and power level from which the coe.ficient is inferred.
6.3 Acceptr.nce Criteria Acceptance criteria for the prediction of key core parameters are defined in Table 6.1. The acceptance criteria for Cycle 11 are unchanged from Cycle 10. The permissible deviations from predicted values ara selected to insure the adequacy of the safety analycis. In these tests, the nominal measured value is compared to the nominal calculated value, the latter corrected for any difference between the measurement and calculational conditions.
If the criteria in Table 6.1 are met, verification is obtained that the core characteristics conform to those assumed in the safety analysis. In particular, compliance with the shutdown margin Technical Specification is l' demonstrated by the CEA worth and drop time measurements, provided all l
l trippable CEAs remain operable.
If the initial criteria in Table 6.1 are not met, additional l measurements, as prescribed by the table, are performed. In addition, any
- -123-l 6348R/4.317 l-
.._,, .. .,, ___m _
deviations are evaluated relative to the assumptions in the safety analysis for the given core parameters. Such deviations and their evaluations are reported to the Staff. A startup test summary report will be available on-site within 90 days of the completion of the startup tests.
-124-65'8R/4.317
O TABLE 6.1 Maine Yankee Cycle 11 Startup Test Acceptance Criteria Measurement . Conditions Criteria
-1. Critical Boron Hot zero power, near Mer.surement within 1%
Concentration all-reds-out delta rho of predicted value
Regulating Banks 1+2+3+4+5 in the of the predicted value nonoverlap condition
- 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 is, in the nonoverlap Measurement (2) is met.- If condition the criteria in Measurement (2) is not met, the total worth of t~1 CEA banks measured mast be within 110% of the predicted value
- 4. Isothermal Hot zero power, near ITC measurement within Temperature all-rods-out 105 x 10-4 delta rho /0F Coefficient at of predicted value and the HZP MTC is within the i 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
acceptable region specified in Figure 4.8 I 6. Control Rod Drep Operating temperature Drop times to 90% insertion Times no greater than 2.70 seconds
- 7. Radial Power At or slightly below Each assembly r.erage 50% power, near all- power within 140% of l Distribution rods-out predicted value
- 8. Tilt Monitoring 5-48% ratad power, near Tilt trends to less than for Symmetry all-rods-out, tilt is 3.0% for greater than 50%
Verification monitored at 5% power power operation, ac
- l. intervals indicated by the relative changes in excore detector l
readings or incore detectors
-125-6348R/4.317
- w. , -
17.0 CONCLUSION
f 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 11. Table 5.3 summarizes the results of each incident analyzed;
-including the Reference Cycle. result and the~ appropriate design limit'. This table illustrates that Specified Acceptable' Fuel Design Limits _(2AFDL) on DNB and fuel' center 11re 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.
l
-126-63483/4.317
t
' 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.
- 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 1976.
15 . 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 ifsing CHIC-KIN", YAEC 1103, dated September 1976.
- 7. T. R. Hencey, "GEMINI-II - A Modified Version of the GEMINI Computer Program", YAEC-1068, dated April 1974.
- 8. P. A. Bergeron, "Maine Yankee Plant Analysis Model Using GEMINI-II",
YAEC-1101, dated June 1976.
- 9. D. J. Denver, E. E. Pilat, R. J. Cacciapouti, "Application of Yankee's Reactor Physics Methods to Maine Yankee", YAEC-1115, dated October 1976.
l 10. P. A. Bergeron, "Maine Yankee Fuel Thermal Performance Evaluation Model",
! YAEC-1099P, dated February 1976 (Proprietary).
- 11. YAEC-1160, "Application of YANKEE WREM-BASED Generic PWR ECCS Evaluation Model to Maine Yankee", July 1978.
I i 12. YAEC-1324, "Maine Yankee Cycle 7 Core Performance Analysis",
September 1982.
-127-6348R/4.317
. 0-5h
- 13. -; YAEC-1464, "Modified Method for CEA Ejection Analysis of Phine 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.
- 15. USNRC Letter, R. W. Reid to R. H. Groce, dated May 27, 1977.-
.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, October 2, 1985.
- 19. Maine Yankee Letter to USNRC, WMY 75-28, dated March 27, 1975, Proposed Change No. 27. "Maine Yankee Core 2 Performance Analysis".
- 20. Maine Yankee Letter to USNRC, WMY 78-62, dated June 26, 1978, "Maine Yankee Proposed Change No. 64".
- 21. Letter,_E. L. Trapp (CE) to R. T. Yee (MYAPC), M-CE-R-011. "Maine Yankee Batch N' Reload Fuel Design Report," dated April 3, 1985.
- 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.
i
, 24. Letter, R. S. Freeman (CE) to R. T. Yee (MYAPC), M-CE-R-118, "Maine Yankee Nuclear Fuel Design Report - Update for Batch P," dated October 28, 1986.
128-6348R/4.317 .
c - .- . . 7.s , ,-,-.,,- - --, , r- ,m., -, y - , ,.-r- , . , - . .- . - .- . . -
7
- 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.
- 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.
- 29. XN-73-25 GAPEX: A Computer Program for Predicting Pellet-to-Cladding Heat Transfer Coefficients, August 13, 1973.
- 30. D. S. Rowe, "COBRA IIIC: A Digital Computer Program for Steady-State and Transient Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Flements", BNWL-1695 (March 1973).
- 31. Combustion Engineering keport, TR-DT-34 "The Hydraulic Performance of the Maine Yankee Reactor Model", June 1971.
- 33. CEND-414. "The Evaluation and Demonstration of Methods for Improved Fuel Utilization", October 1983.
l 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.
- 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 Operation of the Maine Yankee Atomic Power Station", YAEC-1132, dated July 1977.
-129-6348R/4.317
l l
l
" l
- 37. XN-NF-79-52, "Maine Yankee Reload Fuel Design Report / Mechanical Thermal-Hydraulic and Neutronic Analyses", 1979.
- 38. Maine Yankee Letter to USNRC, WMY 78-102, dated November 15, 1978, "Maine Yankee Startup Test Report".
- 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. >
May 20, 1986,
- 42. USNRC Letter to MYAPC, dated June 20, 1986, "Augmentation Factor Removal " with Attached eOfety 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, "DNER Limit Methodology and Application to the Maine Yankee 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 NYAPC, dated March 9, 1983, NMY 83-62, "Topical Report YAEC-1296P, "DNBR limit Methodology and Application to the Maine Yankee Plant".
- 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.
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- 48. ~ NYAPC Letter to USNRC, MN-82-53, "Boron Dilution During Hot and Cold Shutdown (Mode 5~0peration)," dated March 18, 1982.
- 50. YAEC-1396, "Maine Yankee Cycle 8 Core . Performance Analysis", January 1984.
- 52. "Acceptance Criteria for Emergency Cote Cooling Systems for Light-Water Cooled Nuclear Power Reactors", Federal Register, Vol. 39 No. 3-Friday, January 4,1974.
- 53. "Maine Yankee Nuclear Reload Fuel Design Report - Update for Batch Q."
- 54. YAEC-1573, "Maine Yankee Cycle 10 Core Performance Analysis "
December 1986.
- 55. Letter from G. D. Whittier (MYAPC) to A. C. Thadani (USNRC), "Maine Yankee LOCA Analysis," MN-87-15, dated February 24, 1987.
- 56. Letter from John B. Randazza (MYAPC) to Director Nuclear Reactor Regulation, USNRC, MN-87-18, dated February 24, 1987.
- 57. Letter from G. D. Whittier (MYAPC) to A. C. Thadani (USNRC), MN-86-141, dated November 10, 1986.
- 58. - Letter from G. D. Whither (MYAPC) to A. C. Thadani (USNRC), "Response to Requirement of 10CFR50.61 (Pressurized Thermal Shock Rule)," MN-86-14, dated January 21, 1986.
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