Text

License Amendment Request 187, Conforming Technical Specification Changes for Use of Westinghouse Vantage+ Fuel. Attachment 5, Chapter 14, Which Includes Revised Pages for Updated Safety Analysis Report
	ML022200575
Person / Time
Site:	Kewaunee
Issue date:	07/26/2002
From:	Warner M; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NRC-02-067
	Download: ML022200575 (194)
	v • d • e

ATTACHMENT 5 Letter from M. E. Warner (NMC)

To Document Control Desk (NRC)

Dated July 26, 2002 License Amendment Request 187 Revised Pages for Updated Safety Analysis Report (USAR)

TABLE OF CONTENTS PA gB SECTION TITLE

!CAGE 14.*,

SAFETY ANALYSIS...................................................................................................

14.0-1 SAFETY ANALYSIS OVERVIEW..............................................................................

14.0-1 SAFETY ANALYSIS ASSUMPTIONS.................................

14.0-5 SAFETY ANALYSIS AND CORE RELOAD METHODOLOGY.............................. 14.0-8 REFERENCES - SECTION 14.0............................................................................................

14.0-9 14.1 CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS..................... 14.1-1 14.1.1 UNCONTROLLED RCCA WITHDRAWAL FROM A SUB-CRITICAL COND ITION...................................................................................................

14.1-1 14.1.2 UNCONTROLLED RCCA WITHDRAWAL AT POWER...........................

14.1-4 14.1.3 RCCA MISALIGNMENT...............................................................................

14.1-7 14.1.4 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION...... 14.1-8 14.1.5 STARTUP OF AN INACTIVE REACTOR COOLANT LOOP.................. 14.1-12 14.1.6 EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM M ALFUNCTIONS........................................................................................

14.1-13 14.1.7 EXCESSIVE LOAD INCREASE INCIDENT.............................................

14.1-17 14.1.8 LOSS OF REACTOR COOLANT FLOW...................................................

14.1-19 14.1.9 LOSS OF EXTERNAL ELECTRICAL LOAD............................................

14.1-24 14.1.10 LOSS OF NORMAL FEEDWATER............................................................

14.1-27 14.1.11 ANTICIPATED TRANSIENTS WITHOUT SCRAM.................................

14.1-30 14.1.12 LOSS OF AC POWER TO THE PLANT AUXILIARIES........................... 14.1-31 REFERENCES - SECTION 14.1..........................................................................................

14.1-33 14.2 STANDBY SAFETY FEATURES ANALYSIS.........................................................

14.2-1 14.2.1 FUEL HANDLING ACCIDENTS..................................................................

14.2-1 Rev. 16 14-1 12/01/2000 L -- ----

TABLE OF CONTENTS pc a

7-0 BE UPDATE()

SECTION TITLE

&Y mc-PAGE

-14.2.2 ACCIDENTAL RELEASE-RECYCLE OF WASTE LIQUID...................... 14.2-6 14.2.3 ACCIDENTAL RELEASE-WASTE GAS.....................................................

14.2-8 14.2.4 STEAM GENERATOR TUBE RUPTURE..................................................

14.2-10 14.2.5 STEAM LINE BREAK.................................................................................

14.2-15 14.2.6 RUPTURE OF A CONTROL ROD DRIVE MECHANISM HOUSING (RCCA EJECTION)...................................................................................................

14.2-23 14.2.7 TURBINE MISSILE DAMAGE TO SPENT FUEL POOL (DELETED)... 14.2-29 REFERENCES - SECTION 14.2..........................................................................................

14.2-33 14.3 REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT A C CID EN T)..................................................................................................................

14.3-1 14.3.1 LOSS OF REACTOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES WHICH ACTUATES EMERGENCY CORE COOLING SYSTEM.......................................................................................

14.3-1 14.3.2 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS-OF-COOLANT ACCIDENT).............................................................

14.3-6 14.3.3 CORE AND INTERNALS INTEGRITY ANALYSIS.................................

14.3-18 14.3.4 CONTAINMENT INTEGRITY EVALUATION.........................................

14.3-26 14.3.5 OFF-SITE DOSE CONSEQUENCES..........................................................

14.3-36 14.3.6 DELETED.....................................................................................................

14.3 46 14.3.7 DELETED.................................................................................................

14.346 14.3.8 CHARCOAL FILTER IGNITION HAZARD DUE TO IODINE ABSORPTION 14.3-4 6 14.3.9 GENERATION AND DISPOSITION OF HYDROGEN..............

14.348 14.3.10 STEAM GENERATOR TUBE PLUGGING................................................

14.3-60 Rev. 16 14-2 12/01/2000

TABLE OF CONTENTS

'PAC3-

'_MBEeWi,6 TO 5E UPDTb.4~t SECTION TITLE B

PAGE 14.3.11 STEAM GENERATOR TUBE LEEVNG.................................................

14.3-61 14.3.12 STEAM GENERATOR TUBE FATIGUE ANALYSIS..............................

14.3-64 14.3.13 VOLTAGE BASED REPAIR CRITERIA FOR STEAM GENERATOR TUBES 14.3-64 14.3.14 F* AND ELEVATED F* ALTERNATE REPAIR CRITERIA FOR STEAM GENERATOR TUBES..................................................................................

14.3-64 14.3.15 STEAM GENERATOR TUBE REMOVAL................................................

14.3-65 REFERENCES - SECTION 14.3..........................................................................................

14.3-66 Rev. 16 14-3 12/01/2000

TABLE 14.0-1 14.1.8-1 14.1.10-1 14.2.5-1 14.2.5-2 14.2.6-1 14.2-4 Small Break LOCA Time Sequence of Events Small Break LOCA Fuel Cladding Results Initial Conditions for the Kewaunee Nuclear Plant WCOBRA/TRAC Large Break LOCA Analysis Assumptions Used in the Appendix K Calculation Assumptions Used in the Superbounded Calculation Kewaunee Large Break LOCA Containment Data Kewaunee Large Break LOCA Fan Cooler Performance Data (Per Fan Cooler)

Kewaunee Large Break LOCA Structural Heat Sink Data Kewaunee Large Break LOCA Mass and Energy Releases Kewaunee Large Break LOCA Sequence of Events Kewaunee Large Break LOCA Results DELETED Components Nomenclature Maximum Deflections Under Blowdown (Inches)

Rev. 16 12/01/2000 L14-1 CHAPTER 14 LIST OF TABLES TITLE(N keS instrumentation Drift and Calorimetric Errors Nuclear Overpower Trip Channel DELETED 14-0-3r Ued 7

Nommo. Valwe~ss 4-.

of O

Peinn SOOLer4er-lr, sseny-,

DELETED Cjerfiorn Aee, 'tn+

DELETED I

1 14.3-1 14.3-la 14.3.2-1 14.3.2-2 14.3.2-3 14.3.2-4 14.3.2-5 14.3.2-6 14.3.2-7 14.3.2-8 14.3.2-9 14.3.2-10 14.3-3 14.3-3a 7-

CHAPTER 14 LIST OF TABLES___

TABLE TITLE L

14.3-3b Summary of Maximum Stress Intensities (psi) 14.3-4 Structural Heat Sinks 14.3-5 Energy Sources 14.3-6 Integrated Energy Balance 14.3-7 Internal Energy Balance 14.3-8 Major Assumptions for Design Basis LOCA Analysis 14.3-8b Activity in the Core for Design Basis LOCA at the Kewaunee Nuclear Power Plant for Advanced Nuclear Fuels 14.3-9 Doses for Design Basis Loss-of-Coolant Accident 14.3-10 Exposed Zinc and Aluminum Bearing Surfaces within Kewaunee Containment Vessel 14.3-11 Hydrogen Gas Production in Acidic and Basic Solutions 14.3-12 Sources and Assumptions for Hydrogen Calculations 14.3-13 Venting Requirements and Venting Doses without Pressure Increase 14.3-14 Effects of Pressure Increase on Venting Doses 14.3-15 Steam Generator Tube Removal Rev. 16 L14-2 12/01/2000

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.0.1 Scram Reactivity Insertion Rate Negative Reactivity vs Time 14.1._-1 ncontrolled RCCA Withdrawal from a Sub-Critical Condition Reactor Power vs. Time 14.1.1-2 Uncontrolled RCCA Withdrawal from a Sub-Critical Condition Heat Flux vs. Time 14.1.1-3 Uncontrolled RCCA Withdrawal from a Sub-Critical Condition 14.1.1-4 Uncontrolled RCCA Withdrawal from a Sub-Critical Condition vs. ~Time H bS~t

_5

%keI Averaqe ~pr~-~C 14.1.1-5 nmcontroled RCCA Withdrawal from a Sub-Critical Condition 1.-

emperature vs. Time 14.1.2-1 Uncontrolled RCCA Withdrawal I Fast Rate 100% Power Reactor Power vs. Time 14.1.2-2 Uncontrolled RCCA Withdrawal I Fast Rate 100% Power Pressurizer Pressure vs. Time 14.1.2-3 Uncontrolled RCCA Withdrawal I Fast Rate 100% Power Tave vs. Time 14.1.2-4 Uncontrolled RCCA Withdrawal l Fast Rate 100% Power Minimum DNBR vs. Time 14.1.2-5 Uncontrolled RCCA Withdrawal I Slow Rate 100% Power Reactor Power vs. Time 14.1.2-6 Uncontrolled RCCA Withdrawal I Slow Rate 100% Power Pressurizer Pressure vs. Time 14.1.2-7 Uncontrolled RCCA Withdrawal I Slow Rate 100% Power "Tave vs. Time 14.1.2-8 Uncontrolled RCCA Withdrawal I Slow Rate 100% Power Minimum DNBR vs. Time Rev. 16 L14-3 12/01/2000

CILAJTERM1 LIST OF FIGURES FIGURE TITLE 14.1.2-9 Uncontrolled RCCA Withdrawal 9

ýoe 14.1.2-10 Uncontrolled RCCA Withdrawal I(--stR-a)60% Power 14.1.2-11 Uncontrolled RCCA Withdrawal I0Power 14.1.2-12 TT

ý-!

ifdmm-l--

14.1.2-13 UnotoldRC-C-Wihraa Slow Rate 600% Powei 14.1.2-14 U13Cmointrmolked RC-GA JWithdAw-lwRt 60%" w 14.1.2-15

/UcnoleRCAWtdaa SowRate 6A-PowA,0 14.1.2-16 Uncontrle RCC WihdA--l I -Slo

-it-0 oe 14.1.3-1 DELETEBD Doppect RcCA i:ýeprerdA'.Triise(

VS Nt'hc le-or Po~wer vs.

e 14.1.3-2 14.1.3-3 14.1.4-1

-DELETED-1brop~iec( RCCA

-p.et1r{V r-mylsienl P-es1po"SL Cort e-x Hr-fI=k v:S-T~?mC.

-DEIET9ED-broppcc{ P.CCA Rpresenfnl tv-'

7T,'1V15'en+ R-rSpon~e Prt es urZcer~ Pressurie-e vs. T.,nie 14.1.4-2 14.1.4-3 14.1.4-4 D)rvppec( RCCA Rpres1Cd~

-Ti-cmitenf Re.-oS~e Ves~eIl Avcrzie-T'L c

VS. TTme v

Rev. 16 12/01/2000 L14-4

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.1.5-1 Startup of Inactive RX Coolant Loop Tinlet vs. Time 14.1.5-2 Startup of Inactive RX Coolant Loop Tave vs. Time 14.1.5-3 Startup of Inactive RX Coolant Loop Reactor Power vs. Time 14.1.5-4 Startup of Inactive RX Coolant Loop Pressurizer Pressure vs. Time 14.1.5-5 Startup of Inactive RX Coolant Loop Heat Flux vs. Time 14.1.6-1 Excessive Heat Removal I Feedwater System Malfunction Iv anual Control Reactor Power vs. Time 14.1.6-2 Excessive Heat Removal l Feedwater System Malfunction I] M anual Control Pressurizer Pressure vs. Time 14.1.6-3 Excessive Heat Removal I Feedwater System Malfunction I9

ýmanual Control Tavs.s. Time Core Anr 5e Ter rrohttre 14.1.6-4 Excessive Heat Removal l Feedwater System Malfunction flýManuao Control T

vs. ime 14.1.6-5 Excessive Heat Removal Feedwater System Malfunction IgManual Control Minimum DNBR vs. Time 14.1.6-6 Excessive Heat Removal I Feedwater System Malfunction Il* Auto Control Reactor Power vs. Time 14.1.6-7 Excessive Heat Removal IFeedwater Sysiemn Malfunction IqýAuto Control Pressurizer Pressure vs. Time 14.1.6-8 Excessive Heat Removal I Feedwater System Malfunction I1@3Auto Control Rev. 16 L14-5 12/01/2000

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.1.6-9 Excessive Heat Removal j Feedwater System Malfunction I@JAuto Control 14.1.6-10 Excessive Heat Removal I Feedwater System Malfunction Ij;3Zuto Control Minimum DNBR vs Time 14.1.7-1 Excessive Load Increase I BOC Manual Control Reactor Power vs. Time 14.1.7-2 Excessive Load Increase I BOC Manual Control Pressurizer Pressure vs. Time 14.1.7-3 Excessive Load Increase I BOC Manual Control Delta T Core vs. Time 14.1.7-4 Excessive Load Increase I BOC Manual Control Minimum DNBR vs. Time 14.1.7-5 Excessive Load Increase EOC Manual Control Reactor Power vs. Time 14.1.7-6 Excessive Load Increase I EOC Manual Control b

cPesue seTm Pressurizer Pressure vs. Time

"ý 14.1.7-7 Excessive Load Increase I EOC Manual Control Delta T Core vs. Time 14.1.7-8 Excessive Load Increase I EOC Manual Control Minimum DNBR vs. Time 14.1.7-9 Excessive Load Increase I BOC Auto Control Reactor Power vs. Time 14.1.7-10 Excessive Load Increase I BOC Auto Control Pressurizer Pressure vs. Time 14.1.7-11 Excessive Load Increase 1 BOC Auto Control Delta T Core vs. Time 14.1.7-12 Excessive Load Increase I BOC Auto Control Tave vs. Time Rev. 16 L14-6 12/01/2000

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.1.7-13 Excessive Load Increase j BOC Auto Control Minimum DNBR vs. Time 14.1.7-14 Excessive Load Increase I EOC Auto Control Reactor Power vs. Time 14.1.7-15 Excessive Load Increase I EOC Auto Control Pressurizer Pressure vs. Time 14.1.7-16 Excessive Load Increase I EOC Auto Control Delta T Core vs. Time 14.1.7-17 Excessive Load Increase EOC Auto Control Tave vs. Time 14.1.7-18 Excessive Load Increase I EOC Auto Control Minimum DNBR vs Time Rev. 16 12/01/2000 L14-7

& oo &aytyts 4-o+ýve s e 3v r-e-s eLrf-2,etu t re-4

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.1.8-91 C-er-lows,- Time f

Tine 14.1.8-11

-n A A vs. Time Nut!eo.r Power 14.1.8-12 I

'M

jrqs, e FI~k 14.1.8-13 ROW

,,r,,*

,.,a*

m c:.. vs. Tim e 0

.S P Pressur ~-ze Fres-zdre

-Tsrt 14.1-oc.-I 14.1.9-1 Loss of External Electrical Loa.

Power vs. Time 14.1.9-2 Loss of External Electrical LoadB 11111ls Time 0

14.1.9-3 Loss of External Electrical Load PressurizereWat vs. Time 14.1.9-Loss of External Electrical Loadfi-*

14.1.9-6 LossofEtraElcralLa__

Pressurizer Wat e

vs. Time 14.1.9-5 Loss of External Electrical LoaA (vs.hTime 14.1.9-6 Loss of External Electrical Loa

.RoG 6

W s. Time 14.1.9-7 Loss of External Electrical Loa 5--z,,

VeS~e( Av*r'j*avc Cbre. 3f*Id--r rfr 14.1.9-9 Loss of External Electrical Load Pressurizer(

v Time 14-Rev. 16 F

V

USAR Insert 14.TOC-1 14.1.8-14 Complete Loss of Flow - Frequency Decay in Two Pumps (CLOF-UF)

RCS Loop Temperature vs. Time 14.1.8-15 Complete Loss of Flow - Frequency Decay in Two Pumps (CLOF-UF)

Hot Channel Heat Flux vs. Time 14.1.8-16 Complete Loss of Flow - Frequency Decay in Two Pumps (CLOF-UF)

DNBR vs. Time 14.1.8-17 Locked Rotor / Shaft Break-RCS Pressure / PCT Case Total Core Inlet Flow vs. Time 14.1.8-18 Locked Rotor / Shaft Break - RCS Pressure / PCT Case RCS Loop Flow vs. Time 14.1.8-19 Locked Rotor I Shaft Break - RCS Pressure / PCT Case Nuclear Power vs. Time 14.1.8-20 Locked Rotor / Shaft Break - RCS Pressure / PCT Case Core Average Heat Flux vs. Time 14.1.8-21 Locked Rotor / Shaft Break - RCS Pressure / PCT Case Pressurizer Pressure vs. Time 14.1.8-22 Locked Rotor / Shaft Break - RCS Pressure / PCT Case Vessel Lower Plenum Pressure vs. Time 14.1.8-23 Locked Rotor / Shaft Break - RCS Pressure / PCT Case RCS Loop Temperature vs. Time 14.1.8-24 Locked Rotor / Shaft Break - RCS Pressure / PCT Case Hot Channel Heat Flux vs. Time 14.1.8-25 Locked Rotor / Shaft Break - RCS Pressure / PCT Case Hot Spot Cladding Inner Temperature vs. Time

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.1.9-10 Loss of External Electrical Loa eEA toCno Time 14.1.9-11LoaLoss of External Electrical Loa*tr°t*-*

14.1.9-12 Loss of External Electrical Loa 14.1.9-13 Loss of External Electrical Loa*

14.1.9-14 Loss of External E 14.1.9-15 Loss of External Electrical Loa 14.1.9-16 Loss of External Electrical Loa 14.1.9-17

-, -^

,^

^1

i**,"

IX rf;..ý

>.^

4-14.1.9-18 14.1.9-19 14.1.9-20

, :^,.:*

^

i C

t M rý KST,*

D,"ý

",T TIm 14.1.10-1 Loss of Normal Feedwater 14.1.10-2 Loss of Normal Feedwater Ivs.

Time

'ess.

Avc~rcLe ae Cm ucw-Tnl e

14.1.10-3 Loss of Normal Feedwater v

s. Time Rev. 16 L1.4-9 12/01/2000

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.1.10-4 Loss of Normal Feedwater (r

Z;W e

v

s. Time 14.1.10-5 Loss of Normal Feedwater r vs. Time Act& 05/R Z-hscrf 14 S 14.2.4-1 r

or 114.2.4-1 Break Flow and Safety Ijection Fow (Iwo 1ups) vs Reactor 00olant ressure Main Steam Line-Break Variation of Reactivity with Power at Constant Core Average Temperature Main Steam Line Break Safety Injection Flow Rate vs. Reactor Coolant Pressure Main Steam Line Brealk Tave vs. Time Main Steam Line Brea4 Pressurizer Pressure vs.

Main Steam Line Breal(

Heat Flux vs. Time 14.2.5-1 14.2.5-2 14.2.5-3 14.2.5-4 14.2.5-5 14.2.5-6 14.2.5-7 14.2.5-8 14.2.5-9 Main Steam Line Tave vs. Time 14.2.5-10 Main Steam Line Break*eý Pressurizer Pressure vs. Time 14.2.5-11 Main Steam Line Break.<

Heat Flux vs. Time Z

CP Rev. 16 12/01/2000 L14-10 Main Steam Line Breal41psti 4I e"Z:) vs. Timi Main Steam Line BreakS] Ugs Reactivity vs. Time J

USAR Insert 14.TOC-2 14.1.10-6 Loss of Normal Feedwater Steam Generator Mass vs. Time 14.1.12-1 Loss of AC Power to the Plant Auxiliaries Nuclear Power vs. Time 14.1.12-2 Loss of AC Power to the Plant Auxiliaries Vessel Average Temperature vs. Time 14.1.12-3 Loss of AC Power to the Plant Auxiliaries Pressurizer Pressure vs. Time 14.1.12-4 Loss of AC Power to the Plant Auxiliaries Pressurizer Water Volume vs. Time 14.1.12-5 Loss of AC Power to the Plant Auxiliaries Steam Generator Pressure vs. Time 14.1.12-6 Loss of AC Power to the Plant Auxiliaries Steam Generator Mass vs. Time

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.2.5-12 Main Steam Line Break(N F.lowYs. Time 14.2.5-13 Main Steam Line Bre 0

Reactivity vs. Time 14.2.5-14 ainSt L

rc S

u c

fc' 1

r it

" u I.,.

'I o

1 rs r,l P.

t RCCA Ejection I BOC Full Power Reactor Power vs. Time 14.2.6-fl RCCA Ejection I BOC Full Power STemperaturel vs. Time 14.2.6*

G RCCA Ejection BOC Zero Power Reactor Power vs. Time I

-e ffý AA-ý-C-l A-J I J-RW%-

ýýA vy ý1 T-. -

T-.

-2..

Owel 2-,V,"CLV1 11,11egrateu A Rev. 16 12/01/2000 L14-11 14.2.5-16 14.2.5-16 14.2.5-17 14.2.5-18 14.2.5-19 Main Steam Line Break Containment Pressure Response Pressure vs. Time 14.2.5-20 Main Steam Line Break Containment Temperature Response Temperature vs. Time 14.2.6-1 ooI-r rA SCOPE

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.2.6@

14.2.6-&

14.2.6f 14.2.6-6 14.2.6-12 14.2-27 14.2-28 RCCA Ejection BOC Zero Power RCCA Ejection I EOC Full Power Reactor Power vs. Time "R1-)/' CA "TP ^.;

6 i "E"OG" Full.1 Power...

RCCA Ection I EOC Full Power Temperaturel vs. Time RCCA EjectionEOC Zero Power Reactor Power vs. Time DELETED DELETED Pumped Safety Injection Flow (Spill to Containment)

Pumped Safety Injection Flow (Spill to RCS)

RCS Depressurization Transient (3 in.)

Block Diagram for Satan-V Refill Calculation Core Mixture Height (3 in.)

Clad Temperatures Transient (3 in.)

Steam Flow (3 in.)

Core Heat Transfer Coefficient (3 in)

Hot Spot Fluid Temperature (3 in.)

Rev. 16 12/01/2000 L14-12 14.3-1a 1I4.3-1b 14.3-1c 14.3-2 14.3-2a 14.3-2b 14.3-3a 14.3-3b 14.3-4a (E]OPE 1I

CHA-PTER 14 LIST OF FIGURES FIGURE TITLE 14.3-4b Core Power 14.3-5a RCS Depressurization Transient (2 in.)

14.3-5b RCS Depressurization Transient (4 in.)

14.3-6a Core Mixture Height (2 in.)

14.3-6b Core Mixture Height (4 in.)

14.3-7b Clad Temperature Transient (4 in.)

14.3.2-1 Large Break LOCA Sequence of Events 14.3.2-2 Large Break LOCA Containment Pressure Curve 14.3.2-3 Large Break LOCA Appendix K Power Shape 14.3.2-4 Appendix K Calculation Peak Cladding Temperature 14.3.2-5 Appendix K Calculation Core Pressure 14.3.2-6 Appendix K Calculation Vessel Inventory 14.3.2-7 Appendix K Calculation Loop Side Break Flow 14.3.2-8 Appendix K Calculation Vessel Side Break Flow 14.3.2-9 Appendix K Calculation Accumulator Flow 14.3.2-10 Appendix K Calculation High Head Safety Injection Flow 14.3.2-11 Appendix K Calculation Low Head Safety Injection Flow 14.3.2-12 Appendix K Calculation Liquid Flow at Top of Core Channel 10 (OR/SC) 14.3.2-13 Appendix K Calculation Vapor Flow at Top of Core Channel 10 (OH/SC) 14.3.2-14 Appendix K Calculation Liquid Flow at Top of Core Channel 11 (GT) 14.3.2-15 Appendix K Calculation Vapor Flow at Top of Core Channel II (GT)

Rev. 16 L14-13 12/01/2000

FIGURE 14.3.2-16 14.3.2-17 14.3.2-18 14.3.2-19 14.3.2-20 14.3.2-21 14.3.2-22 14.3.2-23 14.3.2-24 14.3.2-25 14.3.2-26 14.3.2-27 14.3.2-28 14.3.2-29 14.3.2-30 14.3.2-31 14.3.2-32 14.3.2-33 14.3.2-34 14.3.2-35 14.3.2-36 CHAPTER 14 LIST OF FIGURES Wo-T -rA SroP TITLE Appendix K Calculation Liquid Flow at Top of Core Channel 12 (HA)

Appendix K Calculation Vapor Flow at Top of Core Channel 12 (HA)

Appendix K Calculation Liquid Flow at Top of Core Channel 13 (LP)

Appendix K Calculation Vapor Flow at Top of Core Channel 13 (LP)

Appendix K Calculation Downcomer Liquid Level Appendix K Calculation Lower Plenum Liquid Level Appendix K Calculation Core Liquid Level Appendix K Calculation Void Fraction in Lower Plenum Appendix K Calculation Void Fraction at Bottom of Core Superbounded Calculation Peak Cladding Temperature Superbounded Calculation Core Pressure Superbounded Calculation Vessel Inventory Superbounded Calculation Loop Side Break Flow Superbounded Calculation Vessel Side Break Flow Superbounded Calculation Accumulator Flow Superbounded Calculation High Head Safety Injection Flow Superbounded Calculation Low Head Safety Injection Flow Superbounded Calculation Liquid Flow at Top of Core Channel 10 (OH/SC)

Superbounded Calculation Vapor Flow at Top of Core Channel 10 (OH/SC)

Superbounded Calculation Liquid Flow at Top of Core Channel 11 (GT)

Superbounded Calculation Vapor Flow at Top of Core Channel 11 (GT)

Rev. 16 L14-14 12/01/2000

CHAPTER 14 LIST OF FIGURES FIGURE TITLE 14.3.2-37 Superbounded Calculation Liquid Flow at Top of Core Channel 12 (HA) 14.3.2-38 Superbounded Calculation Vapor Flow at Top Core Channel 12 (HA) 14.3.2-39 Superbounded Calculation Liquid Flow at Top of Core Channel 13 (LP) 14.3.2-40 Superbounded Calculation Vapor Flow at Top of Core Channel 13 (LP) 14.3.2-41 Superbounded Calculation Downcomer Liquid Level 14.3.2-42 Superbounded Calculation Lower Plenum Liquid Level 14.3.2-43 Superbounded Calculation Core Liquid Level 14.3.2-44 Superbounded Calculation Void Fraction in Lower Plenum 14.3.2-45 Superbounded Calculation Void Fraction at Bottom of Core 14.3.2-46 Peak Clad Temperature Comparison Superbounded Calculations without Uncertainies 14.3.2-47 GT Assembly Clad Temperature Comparison Superbounded Calculations without Uncertainties 14.3.2-48 OH/SC Assembly Clad Temperature Comparison Superbounded Calculations without Uncertainties 14.3.2-49 through DELETED 14.3.2-66 14.3-21 Loop Reactor Mathematical Model for Vertical Response 14.3-21a Reactor Internals Allowable Stress Criteria 14.3-21b Notes for Figure 14.3-21a (Sheet 1 of 2 & Sheet 2 of 2) 14.3-22 Fan Cooler Heat Removal Rate vs Containment Pressure 14.3-23 Containment Pressure Transient Rev. 16 L14-15 12/01/2000

CHArTER 14 LIST OF FIGURES FIGURE TITLE 14.3-24 Containment Capability Study Containment Pressure vs Steam Air Internal Energy Volume = 1.32 x 106 ft3 14.3-25 Structural Heat Transfer Coefficient 14.3-26 Containment capabiiity Study containment Pressure vs Steam Air Internal Energy Volume= 1.32 x 106 fW 14.3-27 Containment Capability Study All Available Energies 14.3-28 Containment Capability Study Zr-H20 Reaction (32.3%)

14.3-29 Containment Pressure Transients 14.3-30 Containment Capability Study Rate of Energy Addition 14.3-31 Containment Capability - Case 1 14.3-32 Containment Capability - Case 2 14.3-33 Individual Contributions to Heat Removal 14.3-34 Containment Capability Study Heat Source 14.3-35 Shield Building Ventilation System 14.3-36 Shield Building Ventilation Performance Test Curve and Shield Building Performance Curve 14.3-37 Annulus Pressure with Steel Shell Expansion Design Basis Transient 14.3-37a Flow Out of the SBVs with Steel Shell Expansion Design Basis Transient 14.3-38 Sensitivity of Base Case to Filter Efficiency 14.3-39 Sensitivity of Base Case to Recirculation*Flow 14.3-40 Sensitivity of Base Case to Participation Fraction 14.3-41 Sensitivity of Base Case to Containment Leak Rate 14.3-42 Containment Sump Water Temperature Transient Rev. 16 L14-16 12/01/2000

CHAPTER 14 LIST OF FIGURES FIGURE 14.3-43 14.3-44 14.3-45 14.3-46 14.3-46a TITLE Hydrogen Control by Purging and Venting through Shield Building Vent System without Pressure Increase Effects of Pressure Increase on Venting Requirements Containment Leakage and Venting Requirement as Affected by Pressure Increase Provisions for Mixing, Sampling and Venting of Containment Gases Reactor Building Ventilation System Post LOCA H2 Control - Flow Diagram Rev. 16 L14-17 12/01/2000 Cýfý

14.0 SAFETY ANALYSIS SAFETY ANALYSIS OVERVIEW Q(33js In this section the safety aspects of the plant are evaluated to demonstrate that the plant can be

-operated safely and that radiological consequences from postulated accidents do not exceed the guidelines of 10 CFR 100.

The American Nuclear Society (ANS), Reference 1, has classified plant conditions into four categories in accordance with the anticipated frequency of occurrence and potential radiological consequences to the public. The four categories are as follows:

Condition I:

Condition II:

+ Condition III:

Condition IV:

Normal Operation and Operational Transients Incidents of Moderate Frequency Infrequent Incidents Limiting Faults A description of each category including design requirements, acceptance criteria, and the applicable design basis transient events is provided below:

Condition I: Normal Operation and Operational Transients Definition Condition I occurrences are operations that are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant.

Design Requirements Condition I occurrences shall be accommodated with margin between any plant parameter and the value of that parameter which would require either automatic or manual protective action.

Events Normal Operation (Base Load and Load Follow)

Acceptance Criteria

+ No Clad Damage/Fuel Melting

Reactor Coolant System Pressure < Design Limits

Main Steam System Pressure < Design Limits

+ Containment Pressure and Temperature < Design Limits Rev. 16 12/01/2000 14.0-1

Condition 1I: Incidents of Moderate Frequency Definition Condition II occurrences include incidents, any one of which may occur during a calendar year for a particular plant.

Design Requirements Condition fl incidents shall be accommodated with, at most, a shutdown of the reactor with the plant capable of returning to operation after corrective action. Any release of radioactive materials in effluents to unrestricted areas shall be in conformance with Paragraph 20.1 of 10 CFR Part 20, "Standards for Protection Against Radiation".

By itself, a Condition HI incident cannot generate a more serious incident of the Condition III or IV type without other incidents occurring independently. A single Condition II incident shall not cause consequential loss of function of any barrier to the escape of radioactive products. (No fuel rod failure or RCS overpressurization).

Transient Events

Uncontrolled RCCA Withdrawal From Sub-critical

Uncontrolled RCCA Withdrawal at Power

+ RCCA Misalignment (Dropped/Static)

+ Chemical and Volume Control System Malfunction 4 Startup of Inactive Reactor Coolant Loop

+ Feedwater System Malfunction

+ Excessive Load Increase

+ Partial Loss of Reactor Coolant Flow

+ Loss of External Load

+ Loss of Normal Feedwater

+ Loss of AC Power to Plant Auxiliaries Acceptance Criteria

+ Reactor Coolant System Pressure < 110% of Design (2750 psia)

+ MDNBR > MDNBR Limit

Fuel Centerline Temp < 4700TF

Dose Consequences < 10CFR20

Main Steam System Pressure < 110% of Design (1210 psia)

+ Containment Pressure and Temperature < Design Limits Rev. 16 14.0-2 12/01/2000

Condition III: Infrequent Incidents Definition Condition IH occurrences include incidents, any one of which may occur during the lifetime of a particular plant.

Design Requirements Condition III incidents shall not cause more than a small fraction of the fuel elements in the reactor to be damaged, although sufficient fuel element damage might occur to preclude resumption of operation for a considerable outage time.

The release of radioactive material due to Condition HI incidents may exceed guidelines of 10 CFR Part 20, "Standards for Protection Against Radiation", but shall not be sufficient to interrupt or restrict public use of those areas beyond the exclusion radius.

A Condition IIM incident shall not, by itself, generate a Condition IV fault or result in a consequential loss of function of the Reactor Coolant System or reactor containment barriers.

Transient Events

+ SmallLOCA

+ Small Steam Line Break

+ Complete Loss of Reactor Coolant Flow

+ Single RCCA Withdrawal at Power

+ Fuel Assembly Misloading

+ Volume Control Tank Rupture Acceptance Criteria Most incidents use Condition H criteria, which are more limiting than the Condition Inl criteria. If these are not satisfied, the following criteri l_*ied:

+ MDNBR < MDNBR Limit - Small Fraction of Fuel Rods (< 5%)

+ Dose Consequences < 10% of 10CFR100

+ RCS Pressure < 2900 psia

+ Containment Pressure and Temperature < Design Limits Condition IV: Limiting Faults Definition Condition IV occurrences are faults that are not expected to occur but are postulated because their consequences would include the potential for the release of significant amounts of Rev. 16 14.0-3 12/01/2000

radioactive material. Condition TV faults are the most drastic, which must be designed against, and thus represent the limiting design cases.

Design Requirements Condition IV faults shall not cause a release of radioactive material that results in an undue risk to public health and safety exceeding the guidelines of 10 CFR 100, "Reactor Site Criteria". A single Condition IV fault shall not cause a consequential loss of required functions of systems needed to cope with the fault including those of the Reactor Coolant System and the Reactor Containment System.

Events

+ Large LOCA

+ Steam Generator Tube Rupture

+ Main Steam Line Break (MSLB)

+ Locked Rotor

RCCA Ejection

Fuel Handling Acceptance Criteria

+ Dose Consequences < IOCFRIOO

RCS Pressure < 2900 psia (emergency)

S< 4000 psia (faulted)

Containment Pressure and Temperature < Design Limits The following events have event specific limits that are more limiting than the Condition IV criteria:

Main Steam Line Break MDNBR > MDNBR Limit (MSLB)

Locked Rotor Peak Clad Temperature < 2700'F Percentage of Fuel Rods Experiencing DNB <

R5A Eection A

a eFuelEnthal

<200 call The basic principle applied in relating design requirements to each of the conditions is that the most frequent occurrences must yield little or no radiological risk to the public and those extreme situations having the potential for the greatest risk to the public shall be those least likely to occur. Where applicable, Reactor Protection System and Engineered Safeguards functioning is assumed to the extent allowed by considerations such as the single failure criterion in fulfilling this principle.

Rev. 16 14.0-4 12/01/2000

In the evaluation of the radiological consequences associated with initiation of a spectrum of accident conditions numerous assumptions must be postulated. In many instances, these assumptions are a product of extremely conservative judgments. This is due to the fact that.

many physical phenomena, in particular fission product transport under accident conditions, are not understood to the extent that accurate predictions can be made. Therefore, the set of assumptions postulated would predominantly determine the accident classification.

This section is divided into three subsections, dealing with various behavior categories:

Core and Coolant Boundary Protection Analysis, Section 14.1 The abnormalities presented in Section 14.1 have no off-site radiation consequences.

Standby Safety Features Analysis, Section 14.2 The accidents presented in Section 14.2 are more severe than those discussed in 14.1 and may cause release of radioactive material to the environment.

+ Rupture of a Reactor Coolant Pipe, Section 14.3 The accident presented in Section 14.3, the rupture of a reactor coolant pipe, is the worst-case accident analyzed and is the primary basis for the design of engineered safety features. It is shown that the consequences of even this accident are within the guidelines of 10 CFR100.

SAFETY ANALYSIS ASSUMPTIONS Parameters and assumptions that are common to the safety analyses are described below to avoid repetition in subsequent sections.

Operating Parameters FV~.4Ote+or ccoldon 177Z C 77T2.)

ower t

+/-2% of t for calorimetric error (eeperature F

deadband and measurement error essurep.

e 220pi 50.1psi for steady-state fluctuations and 56 measurement error The in-itial. activ~e core fl-ov rate is9 consr.atvl ee o

ount fbr-increased eere yasfo d uh to thible plug removal and inereased steam generater-Iube pluggig. Unloss othewe sttdin thoi Method of-Analysis Scto b

HiXu he. R65 miad C-z flow rvates~

R~EPLACE woVTH LJSAR rn-set-t 14.0- Z Rev. 16 14.0-5 12/01/2000

The movable in-core instrumentation system is employed to veridny that actual hot channel factors are, in fact, no higher than the limiting values of the Technical Specifications. These limits on hot channel factors are designed to conservatively bound the assumptions used in the accident aa sestor Heation.System

-~5I AcLUSReSer 14.o

-43 Rhemactor Pr

-otetionSstmemtt ytmi mlydt eif htata o hne atr A reactor trip signal acts to open the two series trip breakers feeding power to the control rod drive mechanisms. The loss of power to the mechanism coils causes the mechanism to release the control rods, which then fall by gravity into the core. There are various instrumentation delays associated with each tripping function including delays in signal actuation, in opening the trip breakers and in the release of the rods by the mechanisms. The total delay to trip is defined as the time delay from the time that trip conditions are reached to the time the rods are free and begin to fall. The time delay and setpoint assumed for each tripping function used in the analysis are as follows:

Reactor Trip Function Setpoint Power Range Negative Rate NIM*

Power Range Low Setpoint 35%

Power Range High Setpoint 118%

Overpower Delta T-%

VOCIO Overtemperature Delta T RCS Low Flow B6.5 0 Of loop fib' High Pressurizer Level 100 feey-si Low Pressurizer Pressure 185 High Pressurizer Pressure pr

_- z42 Low-Low Steam Generator Level 0.0% of level sp RXCP Undervoltage N/M*

RXCP Underfrequency NIM*

Turbine Trip N/M*

N/M* - not explicitly modeled in safety analysis Time Delay (sec)

NIA 6.0 1.0 1.5 NIA N/A N/A Poý qOw.-

ýRtv-nr3Posý,-i~ic ?P-ac Nr0

/

Rev. 16 12/01/2000 14.0-6

The difference between the limiting trip setpoint assumed for the analysis and the actual trip setpoint represents a conservative allowance for instrumentation channel and setpoint errors. Results of surveillance tests demonstrate that actual instrument errors are equal to or less than the assumed values.

The instrumentation drift and calorimetric errors used in eseablishineghemax um pTable 14.0.1.

Trip is defined for analytical purposes as the insertion of all full-length rod control cluster assemblies (RCCAs) except the most reactive RCCA, which is assumed to remain in the fully withdrawn position. This is to provide shutdown margin capability against the remote possibility of a stuck RCCA condition existing at a time when shutdown is required.

The negative reactivity insertion following a reactor trip is a function of the acceleration of the control rods and the variation in rod worth as a function of rod position. Control rod positions after trip have been determined experimentally as a function of time using an actual prototype assembly under simulated flow conditions. The resulting rod positions were combined with rod worths to define the negative reactivity insertion as a function of time, as shown in Figure 14.0.1.

In summary, reactor protection is designed to prevent cladding damage in all transients and abnormalities. The most probable modes of failure in each protection channel result in a signal calling for the protective trip. Coincidence of two-out-of-three (or two-out-of-four) signals is required where single channel malfunction could cause spurious trips while at power. A single component or channel failure in the protection system itself coincident with one stuck RCCA is always permissible as a contingent failure and does not cause violation of the protection criteri e eactor rotection systems are designed in accordance with Reference 2.

Unless oteri tated in the section describing aspecific accident, the fol g

ta Geeao safety valve s with 15% blowdown and rated saeyvl aaiiswere assumed:

Valve Nominal Safety Valve Setting Aalsis Pressure Selpoint (psig) b-si-).(Pressure at SIG) 2 11 3

1105 1167 4

1120 1183 f

1127 1193 Calorimetric Error Instrumentation Accuracy The calorimetric error is the error assumed in the determination of core thermal power as obtained from secondary plant measurements. The total ion chamber current (sum of the top and bottom sections) is calibrated (set equal) to this measured power on a periodic basis. The Rev. 16 14.0-7 12/01/2000

secondary power is obtained from measurement of feedwater flow, feedwater inlet temperature to the steam generator and steam pressure. High accuracy plant instrumentation is provided for these measurements with accuracy tolerances more restrictive than that which would be required to only control the feedwater flow. Each feedwater flow venturi is laboratory calibrated and certified. The expected accuracies are tabulated below with their effect on the overall power measurement.

Variable Accuracy

,Equivalent Percent of Rated Power Feedwater temperature

+ 2-F (%%)

Feedwater pressure (Small

+/-5%

0.3% = Total effect correction on enthalpy)

Steam pressure (Small

+/-2%

correction on enthalpy)

Feedwater flow

+ 1.25%

1.25%

1.55% =Total error Note that the errors have been added directly; statistical combination of errors indicate better accuracy. Corrections for moisture carry-over in the steam (0.25% design basis) can be made which would yield a lower measured power level. This effect can be conservatively neglected!

Qfl

__e Ie KNPP feedwater bypass line (FBL) is a full flow normal feedwater bypass loop designed to accurately measure total feedwater flow at KNPP. The FBL contains a flow section, which includes a flow straightener and a laboratory calibrated flow nozzle. The flow section is accurate to 0.25%.

t The total uncertainty of this feedwater measurement is a function of the uncertainty of the FBL calibration, MM the venturi repeatabili e uncertainty of total feedwater flow, as it contributes to the uncertainty of overall reactor power, is significantly less than the required 1.25%.

S ANAYSI ORE RELOAD METHODOLOGY rdated March 27, 1987, WPS submitted for NRC review a topical reporte "Reload Sa e uation Methods for Application to Kewaunee". Additiona ation was submitted to the NR ruary 12 and March 7, 1988. The r Includes methods for analyzing plant accidens, transi d setpoints ex

.g the loss-of-coolant accident (LOCA) and the fuel mishandling accident.

C Safety Evaluation Report provided in Reference 3 reviewed the descripti performan DYNODE-P (Version 5.4), the RETRAN-02, the the TOODEE-2 codes employ analyses. In addition, the analyses, r es and the results of specific calculations and reloa tions; were exam i e NRC found that the topical report was acceptable for referencing in icensing submittals.

.. r. r*.,,,vAl OV1 iehodaoy Is escrielb-

ý'

&Rrvc-.

10.

Rev. 16 140-8 12/01/2000

REFERENCES - SECTION 14.0

1.

ANSI Ni18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants"

2.

IEEE 279, "Standard for Nuclear Plant Protection Systemns", August 1968

3.

NRC Safety Evaluation Report, JG Guitter (NRC) to DC Hintz (NRC), Letter No. K-S 8-67 dated A p ril 1 1, 1 9 8 8 R y

.) " e i e t T e m l D s 3 r c e h r - f W A - 1 9 -

L.IoZ F ej R-4.

W C AP- -7 9O 8-A ) Deceme 19 69.

6.

Hut iel, S.s.,j d. "REMtrt-oz Mo~elVn3 vincl Qual~fic~tiof 4ý Wds wJOSL W 987 P -A (prfro'i cia.y ), Apr;I 1999.

7-B~rnetfTWTf et zd.~ "/LpOFAtJ Code tkSc-rip+"oo) WCAP.-7co7-P-A (Propr~e-4-ry)

,'And WCAP-7907-A

("npofie-v)

April 1984+.

8.

'Risher, lR.I.

-Tr-. o~nd betry, R.,F, "-WTa4YLE A

"",f..'ýl~OfI C~rofl YKjYe+,'cs Cor-npvft, Code-," WcAPF-7q79q P-A (Proprift-h~rY) eknot wCAP-80-i--A (ZJv-Pori4.oy)

Jo ko y 19 7 5 9-5-~'--

t.+ 61.) " v'/,TRP-01 r-Modelen 5 wmd

-FIiY 4

or~

F~- re:S1SiV*ZeCL cifi~vo RCvCf-Vr-14n-LacA Theral-Hyc(P-a~~e 5aretfy AiiU/ySIS,.

WscAP-1456*.- P-A (Proprie4.ry)., or-tober 1999.

'WCAP-77'7TZ-P-A(prc

),

H'5 Rev. 16 14.0-9 12/01/2000

USAR Insert 14.0.-i For most accidents that are DNB limited, nominal values of initial conditions are assumed. The allowances on power, temperature, and pressure are determined on a statistical basis and are included in the limit DNBR, as described in WCAP-11397 (Reference 4). This procedure is known as the "Revised Thermal Design Procedure," and is discussed more fully in Section 3.2.

For accidents in which the Revised Thermal Design Procedure is not employed, the initial conditions are obtained by applying the maximum steady-state errors to the rated values. The following rated values and conservative steady-state errors were assumed in the analyses:

USAR Insert 14.0.-2 Tables 14.0-2 and 14.0-3 summarize initial conditions and computer codes used in non-LOCA accident analyses, and identify which DNB limited transients were analyzed using the Revised Thermal Design Procedure (RTDP).

USAR Insert 14.0.-3 Power Distribution The transient response of the reactor system is dependent on the initial power distribution. The nuclear design of the reactor core minimizes adverse power distribution through the placement of control rods and operating instructions. The radial peaking factor (F17) and the total peaking factor (F0) characterize the power distribution. The peaking factor limits are provided in the Technical Specifications.

For transients that may be DNB limited, the radial peaking factor is of importance. The radial peaking factor increases with decreasing power level due to rod insertion. This increase in F&H is included in the core limits illustrated in Figure 14.0.2. All transients that may be DNB limited are assumed to begin with an FH consistent with the initial power level defined in the Technical Specifications. The axial power shape used in the DNB calculations is discussed in Section 3.2.

Also, the radial and axial power distributions are input to the VIPRE code as described in Section 3.2.

For transients that may be overpower limited, the total peaking factor (Fa) is of importance.

These transients are assumed to begin with plant conditions, including power distributions, that are consistent with reactor operation as defined in the Technical Specifications.

For overpower transients that are slow with respect to the fuel rod thermal time constant (for example, the Chemical and Volume Control System malfunction that results in a decrease in the boron concentration of the reactor coolant system, lasting many minutes), the fuel rod thermal evaluations are performed as discussed in Section 3.2. For overpower transients that are fast with respect to the fuel rod thermal time constant (for example, the uncontrolled rod cluster control assembly (RCCA) bank withdrawal from subcritical and the RCCA ejection incidents that result in a large power rise over a few seconds), a detailed fuel heat transfer calculation is performed. Although the fuel rod thermal time constant is a function of system conditions, fuel burnup, and rod power, a typical value at beginning-of-life for high power rods is approximately five seconds.

USAR Insert 14.0.-4 Reactivity Coefficients The transient response of the reactor system is dependent on reactivity feedback effects, in particular the moderator temperature coefficient and the Doppler power coefficient. These reactivity coefficients and their values are discussed in detail in Section 3.2.

In the analysis of certain events, conservatism requires the use of large reactivity coefficient values, whereas in the analysis of other events, conservatism requires the use of small reactivity coefficient values. Some analyses, such as loss of coolant from cracks or ruptures in the Reactor Coolant System, do not depend on reactivity feedback effects. The justification for use of conservatively large versus small reactivity coefficient values is treated on an event-by event basis. In some cases, conservative combinations of parameters are used to bound the effects of core life, although these combinations may represent unrealistic situations.

USAR Insert 14.0.-5 Reference is made above to Overpower and Overtemperature Delta T (AT) variable reactor trip setpoints illustrated in Figure 14.0.2. This Figure presents the allowable reactor coolant loop average temperature and AT for the design flow and power distribution, as described in Section 3.2, as a function of primary coolant pressure. The boundaries of operation defined by the overpower AT trip and the overtemperature AT trip are represented as "Protection Lines" on this diagram. The protection lines are drawn to include all adverse instrumentation and setpoint errors so that under nominal conditions a trip would occur well within the area bounded by these lines. The utility of this diagram is in the fact that the limit imposed by any given DNBR can be represented as a line. The DNB lines represent the locus of conditions for which the DNBR equals the limit value (1.34 for the thimble cell and 1.34 for the typical cell). All points below and to the left of a DNB line for a given pressure have a DNBR greater than the limit value. The diagram shows that DNB is prevented for all cases if the applicable DNBR line at any point does not traverse the area enclosed with the maximum protection lines. The area of permissible operation (power, pressure, and temperature) is bounded by the following combination of reactor trips: high neutron flux (fixed setpoint), high pressurizer pressure (fixed setpoint), low pressurizer pressure (fixed setpoint), overpower AT (variable setpoint) and overtemperature AT (variable setpoint). The DNBR limit value, which was used as the DNBR limit for all accidents analyzed with the Revised Thermal Design Procedure (see Table 14.0-2), is conservative compared to the actual design DNBR value required to meet the DNB design basis as discussed in Section 3.2.

USAR Insert 14.0.-6 Computer Codes Utilized Summaries of some of the principal computer codes used in transient analyses are given below.

Other codes, such as those used in the analysis of reactor coolant system pipe ruptures (Section 14.3), are summarized in the respective accident analyses sections. Table 14.0-2 provides a list of codes used for each transient analysis.

FACTRAN (Ref. 5)

FACTRAN calculates the transient temperature distribution in a cross-section of a metal clad U0 2 fuel rod and the transient heat flux at the surface of the cladding, using as input the nuclear power and the time-dependent coolant parameters of pressure, flow, temperature and density.

The code uses a fuel model that simultaneously contains the following features:

a.

A sufficiently large number of radial space increments to handle fast transients such as a rod ejection accident;

b.

Material properties which are functions of temperature and a sophisticated fuel-to-cladding gap heat transfer calculation; and

c.

The necessary calculations to handle post-DNB transients: film boiling heat transfer correlations, Zircaloy-water reaction, and partial melting of the fuel.

RETRAN (Ref. 6)

RETRAN is used for studies of transient response of a pressurized water reactor (PWR) system to specified perturbations in process parameters. This code simulates a multi-loop system by a lumped parameter model containing the reactor vessel, hot and cold leg piping, reactor coolant pumps, steam generators (tube and shell sides), steam lines, and the pressurizer.

The pressurizer heaters, spray, relief valves, and safety valves may also be modeled. RETRAN includes a point neutron kinetics model and reactivity effects of the moderator, fuel, boron, and control rods. The secondary side of the steam generator uses a detailed nodalization for the thermal transients. The reactor protection system (RPS) simulated in the code includes reactor trips on high neutron flux, overtemperature and overpower AT (OTAT/OPAT), low reactor coolant system (RCS) flow, high and low pressurizer pressure, high pressurizer level, and Io-lo steam generator water level.

Control systems are also simulated including rod control and pressurizer pressure control.

Parts of the safety injection system (SIS), including the accumulators, may also be modeled. RETRAN approximates the transient value of departure from nucleate boiling ratio (DNBR) based on input from the core thermal safety limits.

LOFTRAN (Ref. 7)

Transient response studies of a pressurized water reactor (PWR) to specified perturbations in process parameters use the LOFTRAN computer code.

This code simulates a multi-loop system by a model containing the reactor vessel, hot and cold leg piping, steam generators (tube and shell sides), the pressurizer and the pressurizer heaters, spray, relief valves, and safety valves. LOFTRAN also includes a point neutron kinetics model and reactivity effects of the moderator, fuel, boron, and rods. The secondary side of the steam generator uses a homogeneous, saturated mixture for the thermal transients. The code simulates the reactor protection system (RPS) which includes reactor trips on high neutron flux, OTAT, OPAT, high and low pressurizer pressure, low reactor coolant system (RCS) flow, Io-lo steam generator water level, and high pressurizer level.

Control systems are also simulated including rod control, steam dump, and pressurizer pressure control.

The safety injection system (SIS),

including the accumulators, is also modeled. LOFTRAN also approximates the transient value

of departure from nucleate boiling ratio (DNBR) based on the input from the core thermal safety limits.

TWINKLE (Ref. 8)

TWINKLE is a multi-dimensional spatial neutron kinetics code. The code uses an implicit finite difference method to solve the two-group transient neutron diffusion equations in one, two, and three dimensions. The code uses six delayed neutron groups and contains a detailed multi region fuel-cladding-coolant heat transfer model for calculating pointwise Doppler and moderator feedback effects. The code handles up to 8,000 spatial points and performs its own steady-state initialization.

Aside from basic cross-section data and thermal-hydraulic parameters, the code accepts as input basic driving functions such as inlet temperature, pressure, flow, boron concentration, control rod motion, and others. The code provides various output, e.g., channelwise power, axial offset, enthalpy, volumetric surge, pointwise power and fuel temperatures.

It also predicts the kinetic behavior of a reactor for transients that cause a major perturbation in the spatial neutron flux distribution.

VIPRE (Ref. 9)

The VIPRE computer program performs thermal-hydraulic calculations. This code calculates coolant density, mass velocity, enthalpy, void fractions, static pressure and departure from nucleate boiling ratio (DNBR) distributions along flow channels within a reactor core. Additional discussion of the VIPRE code is provided in Section 3.2.

USAR Insert 14.0.-7 The core reload methodology is described in Reference 10.

TABLE 14.0-1 INSTRUMENTATION DRIFT AND CALORIMETRIC ERRORS NUCLEAR OVERPOWER TRIP CHANNEL Set Point and Error Estimated Instrument Allowances:

Errors:

(% of rated power)

Nominal set point 109 Calorimetric error 2

1.55 Axial power distribution effects on total ion chamber current 5

3 Instrumentation channel drift and set point reproducibility 2

1.0 Maximum overpower trip point assuming all individual errors are simultaneously in the most adverse direction 118 Rev. 16 12/01/2000 Page 1 of 1

Table 14.0-2 I-ab 1-4o.

Aide-i' Summary of Initial Conditions and Computer Codes Used for Non-LOCA Accident Analyses Computer Revised Initial Core Vessel Coolant Vessel Avg.

Pressurizer Transient/Event Codes DNB Thermal Design Power Flow Coolant Pressure Used Correlation Procedure

(% 1772 MWt)

(gpm)

Temp. (°F)

(psla)

Uncontrolled RCCA TWINKLE Withdrawal from a Subcritical FACTRAN W-3V1)

No 0

79,922 547.0 2160 Condition VIPRE WRB-1t2 Uncontrolled RCCA 100 (DNB) 573.0 (100%)

Withdrawal at Power RETRAN WRB-1 Yes (DNB) 60 (DNB) 186,000 (DNB) 562.6 (60%)

2250 (DNB)

No (Pressure) 10 (DNB) 178,000 (Pressure) 549.6 (10%)

2200 (Pressure)(3) 8 (Pressure) 555.6 (8%)

RCCA Misalignment LOFTRAN (Dropped Rod)

VIPRE WRB-1 Yes 100 186,000 573.0 2250 Chemical and Volume Control 579.0 (Power) 2250 (Power)

System Malfunction N/A N/A N/A N/A N/A 554.3 (Startup) 2250 (Startup) 140.0 (Refueling) 14.7 (Refueling)

Startup of an Inactive Reactor Event precluded by the Technical Specifications Coolant Loop Reduction in Feedwater Event bounded by the Excessive Load Increase Incident Temperature Increase in Feedwater Flow RETRAN WRB-1 (HFP)

Yes (HFP) 100 (HFP) 186,000 (HFP) 573.0 (HFP)

VIPRE W-3 (HZP)

No (HZP) 0 (HZP) 178,000 (HZP) 547.0 (HZP) 2250 Excessive Load Increase N/A WRB-1 Yes 100 186,000 573.0 2250 Loss of Reactor Coolant Flow RETRAN VIPRE WRB-1 Yes 100 186,000 573.0 2250 Locked Rotor RETRAN VIPRE WRB-I Yes (DNB) 100 (DNB) 186,000 (DNB) 573.0 (DNB) 2250 (DNB)

FACTRAN No (Hot Spot) 102 (Hot Spot) 178,000 (Hot Spot) 579.0 (Hot Spot)'

2300 (Hot Spot)03)

Loss of External Electrical Yes (DNB) 100 (DNB) 186,000 (DNB) 573.0 (DNB) 2250 (DNB)

Load RETRAN WRB-1 No (Pressure) 102 (Pressure) 178,000 (Pressure) 579.0 (Pressure) 2200 (Pressure( 3)

Loss of Normal Feedwater RETRAN N/A No 102 178,000 579.0 2300(3)

Anticipated Transients Without NMC Scram Scope Loss of AC Power to the Plant Auxiliaries RETRAN N/A No 102 178,000 579.0 2300(31 Steam Generator Tube Not TA Rupture Scope Steam Line Break RETRAN VIPRE W-3 No 0

178,000 547.0 2250 RCCA Ejection TWINKLE 102 (HFP) 178,000 (HFP) 579.0 (HFP) 2200(')

_FACTRAN N/A No 0 (HZP) 79,922 (HZP) 547.0 (HZP) t"*Below the first mixing vane grid. ")Above the first mixing vane grid.

""An additional 0.1 psi uncertainty has been evaluated.

Table 14.0-3 Nominal Values of Pertinent Parameters for Non-LOCA Accident Analyses Maximum T-avg Maximum T-avg Minimum T-avg Minimum T-avg Parameter with RTDP non-RTDP with RTDP non-RTDP Thermal Output of NSSS (MWt) 1780 1780 1780 1780 Maximum Core Power (MWI) 1772 1772 1772 1772 Vessel Average Coolant Temperature (°F)(1) 573.0 573.016.0 556.3 556.3+/-6.0 Pressurizer Pressure (psia) 2250.0 2250.0+/-50.1 2250.0 2250.0+/-50.1 Reactor Coolant Loop Flow (GPM) 93,000 89,000 93,000 89,000 Steam Generator Tube Plugging 0 to 10%

0 to 10%

Steam Generator Outlet Pressure 771 (0% SGTP) 771 (0% SGTP) 656 (0% SGTP) 656 (0% SGTP)

(psia) 747 (10% SGTP) 747 (10% SGTP) 634 (10% SGTP) 634 (10% SGTP)

Assumed Feedwater Temperature at Steam Generator Inlet (OF) 437.1 437.1 437.1 437.1 Average Core Heat Flux (Btu/hr-ft2) 206,585 206,585 206,585 206,585 i"The accident analyses support a full power T-avg range from 556.3°F to 573.0°F.

SCRAM REACTIVITY INSERTION RATE NEGATIVE REACTIVITY vs TIME REPLACE WVTH WIVCW

, uR" 14..1 1.0 1.5 Time From Rod Release (Sec)

FIGURE 14.0.1 NOV 0 1 1997 2.0 1.8 1.6 1.4 1.2 1.0 0.8

(

0 Z o.6 0.4 0.2 0.0 0.0 2.0

Scram Reactivity Insertion Rate Negative Reactivity vs Time 35 I

I I

3 I

I I

I 2.54 -----

0 f-I

_0 4-1 I

I I

I 0 5 0

0 0.5 1

1.5 2

Time From Rod Release (seconds)

Figure 14.0.1 2.5 J

z I

B i

i i

Illustration of Overtemperature and Overpower AT Protection 90 OPAT 80 Protection 70 Normal 50 Operating Condition 40 SG Safety Valve Protection 30-20 OTAT Protection Core Limits Figure 14.0.2

14.1 CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS The following anticipated events are abnormal operational transients resulting from component failure or operator error. They are anti.cpated to occur sometime in the design life of the plant.

In these events the reactor control and protection system and engineered safeguards are relied upon to protect the core and reactor coolant system boundary from damage.

4 Uncontrolled RCCA Withdrawal from a Sub-critical Condition (Section 14.1.1)

Uncontrolled RCCA Withdrawal at Power (Section 14.1.2)

+ RCCA Misalignment (Section 14.1.3)

+ Chemical and Volume Control System Malfunction (Section 14.1.4)

+ Startup of an Inactive Reactor Coolant Loop (Section 14.1.5)

+ Excessive Heat Removal Due to Feedwater System Malfunctions (Section 14.1.6)

+ Excessive Load Increase Incident (Section 14.1.7)

+ Loss of Reactor Coolant Flow (Section 14.1.8)

+ Loss of External Electrical Load (Section 14.1.9)

+ Loss of Normal Feedwater (Section 14.1.10)

Loss of all AC Power to the Plant Auxiliaries (Section 14.1.12) 14.1.1 UNCONTROLLED RCCA WITHDRAWAL FROM A SUBCRITICAL CONDITION Accident Description A RCCA withdrawal incident is defined as an uncontrolled addition of reactivity to the reactor core by withdrawal of RCCAs resulting in a power excursion. While the probability of this type of a transient is extremely low, such a transient could be caused by a malfunction of the reactor control or control rod drive systems. This could occur with the reactor either sub-critical or at power. The "at power" case is discussed in Section 14.1.2.

Reactivity is added at a prescribed and controlled rate in bringing the reactor from a shutdown condition to a low power level during startup by RCCA withdrawal. Although the initial startup procedure used the method of boron dilution, the normal startup is with RCCA withdrawal. RCCA motion can cause much faster changes in reactivity than can be made by changing boron concentration.

The control rod drive mechanisms are wired into pre-selected bank configurations, which are not altered. The RCCAs are therefore physically prevented from withdrawing in other than their respective banks. Power supplied to the rod banks is controlled such that no more than two banks can be withdrawn at any time. The rod drive mechanism is of the magnetic latch type and the coil actuation is sequenced to provide variable speed rod travel.

The nuclear power response to a continuous reactivity insertion is characterized by a very fast rise terminated by the reactivity feedback effect of the negative fuel temperature coefficient. This self-limitation of the initial power burst results from a fast negative fuel Rev. 16 14.1-1 12/01/2000

coefficient. This self-limitation of the initial power burst results from a fast negative fuel temperature feedback (Doppler effect) and is of prime importance during a startup accident, since it limits the power to a tolerable level prior to external control action. After the initial power burst, the nuclear power is momentarily reduced. If the accident is not terminated by a reactor trip, the nuclear power increases again, but at a much slower rate.

Should a continuous RCCA withdrawal be initiated, the transient will be terminated by the following automatic P1oQtection

,r control stem actio

a. Source Range Hi eutron Flux Reactor Trip - This trip is actuated when either of two independent sou e range channels indicates a flux level above a pre-selected, manually adjustable valu. This trip function may be manually bypassed when either intermediate range flux channel indicates a flux level above

. It is automatically reinstated when both intermediate range channels indicate a ux level below Q* eeie~d*..

b. Intermediate Range High Neutron Flux Rod Stop - This rod stop is actuated when either of two independent intermediate range channels indicates a flux level above a pre-selected, manually adjustable value. This rod stop may be manually bypassed when two out of the four power range channels indicate a power level above approximately 10% power. It is automatically reinstated when three of the four power range channels are below this value.

c. Although the actuation logic, bypass and automatic reinstatement conditions are the same for the Intermediate Range High Neutron Flux Rod Stop and Intermediate Range High Neutron Flux Reactor Trip, the rod stop is generated at

35% full power unless manually bypassed above permissive 10 (10% full power). The reactor trip will be actuated at *40%

full power unless it has been manually bypassed above permissive 10.

d. Power Range High Neutron Flux Reactor Trip (low setting) - Trip is actuated when two out of the four power range channels indicate a power level above approximately 25%. This trip function may be manually bypassed when two of the four power range channels indicate a power level above approximately 10% power and is automatically reinstated when three of the four channels indicate a power level below this value.

e. Power Range High Neutron Flux Rod Stop - This rod stop is actuated when one-out-of four power range channels indicates a power level above a preset setpoint. This function is always active.

f Power Range High Neutron Flux Reactor Trip (high setting) - Trip is actuated when two out-of-four power range channels indicate a power level above a preset setpoint. This trip function is always active.

Termination of the startup accident by the above protection channels prevents core damage. In addition, the reactor trip from high pressurizer pressure serves as a backup to terminate the accident before an overpressure condition could occur.

Rev. 16 14.1-2 12/01/2000

Method of Analysis RP1JACEWiTII USR nser' Ana transient is erformed by digital computation incorporating the neutron kinetics (including six delayed neutron groups ore thermal and hydraulic equations. In k....-"/-*laddition to the nuclear flux response, the average fuel, clad and water heatpc S*lu response were computed.

In order to give conservative results for a st

-u accid win assu tions are ma e concerning the initial reactor conditions-(,J~sel o-".

it.)

V(a]O u e -*"4 3

a. Since the magnitude of the nuclear power pe reac e during e initial part of t transient is for any given rate of reactivity insertion strongly dependent on the Doppler reactivity coefficient, a conservatively low 5

st~aap-aeeidenA. The less negative Deppier coefficent reducer th*Doppler feedback effectc bezebicreas4 the nuclear flux peak.

b. The contribution of the moderator reactivity coefficient is negligib uring the initial part of the transient because the heat transfer time constant betwe e fuel and the moderator is much longer than the nuclear flux response time co ant. However, after the initial nuclear flux peak, the succeeding rate of power rease is affected by the moderator reactivity coefficient. A conservative value of-ihas been used in the analysis rince the pgiti;'c &'aluc w'-2iyield the maximum peak core heat flux.

(~~~N-6j)

(v~Owa tomf.erg..tue a 4 5jV7 0fr

c. The reactor is assu te-i--d-o be a-ot zero powe. This assumption is more conservative than that of a lower initial system temperature. The higher initial system temperature yields larger fuel to water heat transfer, larger fuel thermal capacity, and less negative (smaller absolute magnitude) Doppler coefficient. The high nuclear flux peak combined with a high fuel thermal capacity an large thermal conductivity yields a larger peak heat flux. The initial ultiplication-Q* a assumed to be 1.0 since this results in the maximum nuclear flux peak. *e-v'j§

d. The most adverse combination of instrument and setpoint errors, as well as delays for trip signal actuation and rod release, are taken into account. A 10% increase has been assumed for the power range flux trip setpoint (low setting) raising it from the nominal value of 25%

to 35%. Reference to Figure 14.1.1-1, however, shows that the rise in nuclear flux is so rapid that the effect of errors in the trip setpoint on the actual time at which the rods are released is negligible.

e. A maximum reactivity insertion rate is assumedj(&29-44Wsee ich is greater than that for the simultaneous withdrawal at maximum speed of the combination of the two RCCA banks having the greatest combined worth.

f. Initial power level of 1.0E-;Wmultiplied by the nominal full power level is assumed to maximize the heat flux peak.

14.

n1s--rt Rev. 16 14..12.14.1-3 12/01/2000

USAR Insert 14.1.1-1 The analysis of the uncontrolled RCCA bank withdrawal from subcritical accident is performed in three stages: first, an average core nuclear power transient calculation; then, an average core heat transfer calculation; and finally, the DNBR calculation. The average nuclear power transient with respect to time calculation is performed using a spatial neutron kinetics code, TWINKLE, which includes the various total core feedback effects, i.e., Doppler and moderator reactivity. The FACTRAN code is then used to calculate the thermal heat flux transient, based on the nuclear power transient calculated by TWINKLE. FACTRAN also calculates the fuel and cladding temperatures. The average heat flux is next used in VIPRE for transient DNBR calculations.

USAR Insert 14.1.1-2

g.

The most limiting axial and radial power shapes, associated with having the two highest combined worth sequential banks in their highest worth position, are assumed for DNB analysis.

h.

One reactor coolant pump is assumed to be in operation. This lowest initial flow minimizes the resulting DNBR.

i.

A core flow reduction of 1.1 percent, which addresses the potential reactor coolant flow asymmetry associated with a maximum loop-to-loop steam generator tube plugging imbalance of 10 percent, has been applied.

USAR Insert 14.1.1-3 With the reactor tripped, the plant returns to a stable condition. The plant may subsequently be cooled down further by following normal plant shutdown procedures.

Conclusions In the event of a RCCA withdrawal accident from the subcritical condition, the core and the reactor coolant system are not adversely affected. The minimum departure from nucleate boiling ratio remains above the limit value and thus, no fuel or clad damage is predicted.

R esults

-I. e ;

R'V'x 91 Figures 14. 1. 1-1 through 14.1.1-5 ýbow the transient behavior of key, parameters for aivi

(.acc

,-lc,

r ta t of 8.2E I A k

ýsec. " accident i term inated by a reactor trip at 35% pow er.

The nuclear power overshoots nominal full power, but only for a very short time period. Hence, the energy release and the fuel temperature increases are small. The heat flux response, of interest for DNB considerations, is shown in Figure 14.1.1-2. The beneficial effect of the inherent thermal lag of the fuel is evidenced by a peak heat flux that is less than the nominal full power heat flux. There is a large margin to DNB during the transient since the rod surface heat flux remains below the full power design value, and there is a high degree of sub- -.

-ti cooling at all times in the core. Figres 14.1.1-3, 14.1.1-4, and 14.1.1-5 show the response of '~

t.i "h e co re

-av g c f l, c l

-aA an d c lad d in g tem p eratu r T h e av erag e e tem p eratu re" increases to a value that is lower than the nominal full power value.--*

"The following table shows the comparison of the important calculated safety parameters to their respective acceptance criteria (Calculated Value/Acceptance Criterion 0NBR RCS Pressure MSS Pressure Uncon rod withdrawal from sub-critical 3.218/1.14 2358/2750 1156/1210 Conctini Considering the conservative ass ons used ceident analysis, it is concluded that in the unlikely event of a control r raw.

dent the core and reactor coolant systems are not adversely a he peak heat flux reached rema n thenominal full power value._N is well above its Imiting value. The peak average clad tempera s than]

its nominal full power value, and thus there is no possibility of fuel or clad damage.

1 UNCONTROLLED RCCA WITHDRAWAL AT POWER 5-t-k-01 Acciden cription hr.\\...

An uncontrolled RCCA w h awal at powe u ts in an increase in core heat flux. Since the led heat extraction from the steam ge r remains constant until the steam generator pressure A3reaches the relief, or saf alve setpo' there is a net increase in reactor coolant L W temperature. Unle riated by manual or au tic action, the power mismatch and resultant co t temperature rise would eventually result mn

. Therefore, to prevent the pos ty of damage to the cladding, the Reactor Protection Systemn i ned to terminate any such transient before the DNBR falls below its limit.

Rev. 16 12/01/2000 14.1-4

\\

oooz/1oUz[

g[ -^Aaa(([1[anl[

[s] aw~l O0 9

Z2 8g ?17 O

91.

L 8

17 0

0 0 1

~oo og

\\3 S.

....... i

. a-S................ :

7

-- 0 0O '1-

-o 009*?L

.0 O OS T'*-'

"I

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

I...........

Io I...........

C0"i

[ I I i

J 00"9 OW!j -SA JaMOd JO13eaU U ipuoo leoi lpo-qns V WoJJ IeMeJpqf VOOU PallOJIUO3.ol o o Pa,,-ou

Uncontrolled RCCA Withdrawal From A Sub-Critical Condition Reactor Power vs T1ime 1.65

3.

3..

3 3

3 I

3 3

3

3...

I 3o 1 8*..

303Ii 3-33

""3 3

3 3

I 3

3 3

I3 3

o II 3

3 I

I 3

3 3

3 I

3'

3 3

"Time [s]

Figure 14.1.1-1

oooz/fo/zi 91 *

[tl

-'1;17f3 1

[s] aw~lI Ot1 9,

-C 83 t'z Oz 9[.

3 8

V 0L0 0603 O

ij SA XflJ B PV J

0 I ----------...........

0T

.......... ------r..

0 6 *"

0" 00 L

k2 3

lv

Uncontrolled RCCA Withdrawal From A Sub-Critical Condition Heat Flux vs Time

.4 S

E O

O 3

3 3

o I3 I

2 I

I I

3 10 15 02 3

I me I[I Figur 14..-

oooz/1o/zl 91Aa C-1*7F 3t'f

~i1l!

O 96 3

8g tV3 O

9[

L 8 9 0

0*00L S..........

........... e -------

4 0..

S....... i........

0 0 0

-; il O

' O 009LOO S..........

.............=.......,....

0 "0 0 6

].....................................................................

0o o 0 6 i L

~ ~

~------ ------

0 *0.....

...................6o o i

4 *° "0"0£0 L L I

I ewPOL

-s^ WJnJeiadwM

_L lon_.U

Olllpuoo laielplo-qnS V uwoJ-.I iemejqiVOOE PallOJiUO:aUfn t't' { '0'

Uncontrolled RCCA Withdrawal From A Sub-Critical Condition Hot-Spot Fuel Centerline Temperature vs Time 3000)

II I

3 I

I I

I Ii 3

I I

I 3

I I

I 21500--------4 I

I 3

3 I

"I I

I 3

3 I

I 200---------5 3Time 3[3 Figure 14.1.1-3

9OOZ/1;o/Z r Ov 96 Z6 83 vl [So0

/ 9L 3

8 V

0 II I

,I-u

-~~~~~ ~ - --

-I..

0"6--

CD

-----------i C :..

" £ "CD 6.

--- 0 "L9 9 0,999 Uo!1puQo ieo!lpo-qns V WOJI iemeJPLi!M WP_*.Io l

°uO3U

_ I.1.

tl q,_* *-A

Uncontrolled RCCA Withdrawal From A Sub-Critical Condition Hot-Spot Fuel Average Temperature vs. Time 2500 3

3 3

3 I

I 3

3 3

3 I

3 3

3 I

I 3

I I

3 3

I I

I 3

I I

3 I

I 3

I I

I 3

I I

I 3

3 1

I 3

3 3

3 I

3 3

2000-------------+-----4---------

3 3

3 I

3 3

3 I

I 3

3 I

3 3

I I

I 3

3 I

I 3

3 I

3 3

I 3

3 3

L.

I 3

3 I

3 3

3 I

I 3

3 3

0

3 3

L I

3 3

3 I

I I

moo-------------------------------------3 L

I I

I 3

3D I

3 3

I 3

3 3

3 I

E a

1-3 I

I I

3 I

I I

3 I

I I

3 3

I I

3 3

3 I

3 3

I 3

3 3

I 3

3 3

I 3

3 3

I 3

3 3

I 3

3 3

1000-------------+-----4-----4---

I 3

3 3

I 3

3 3

3 I

3 3

I I

3 3

I I

I 3

3 3

I 3

I I

I 3

I I

3 3

I I

500-3 3

3 I

3 3

I 3

3 I

I I

0 5

10 15 20 25 Time [si Figure 14.1.1-4

/017[s]

aw~lI OV 98 z

8z 1'z Oz 9L.

U1I 8

1 0

0 '93 9I.

CD 1CD S....

1*..........

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

CD "0 8 £ OW l

-SA wnieiadweL pe l adS IOH uoil!puoo leoillj3-qnS V WOJd le

Uncontrolled RCCA Withdrawal From A Sub-Critical Condition Hot-Spot Cladding Temperature vs. Time 750 3

3 I3 3

3 3

3' 3

3 3 3 I

3 3

I 3

3 I

3 3

I 3

I I

I 3

I I

3,II, "3

365 I..

3

~I 3

700 I

3 3

I 3

I I--

3 I

550 I

I I

I 3

I,1 I

3 3

55-----------------4--

ITi me 3[s Figure 14.1.1-5

Results jsecerf'-)

Figures 14.1.1-1 through 14.1.1-5 show the transient behavior.of key parameters for a reactiit, insr-rtion rate of &.--4

-MAsec.

The accident is terminated by a reactcr tri,

t '2%

power.

The nuclear power overshoots nominal full power, but only for a very short time period. Hence, the energy release and the fuel temperature increases are small. The beat flux response, of interest for DNB considerations, is shown in Figure 14.1.1-2. The beneficial effect of the inherent thermal lag of the fuel is evidenced by a peak heat flux that is less than the nominal full power heat flux. There is a large margin to DNB during the transient since the rod surface heat flux remains below the full power design value, and there is a high degree of sub-cooling at all times in the core. Figures 14.1.1-3, 14.1.1-4, and 14.1.1-5 show the response of the core average fuel, coolant and cladding temperature. The average fuel temperature increases to a value that is lower thian the nominal full power value. The average coolant temperature increases to a value that is also less than the full power nominal value.

The following table shows the comparison of the important calculated safety parameters to their respective acceptance criteria (Calculated Value/Acceptance Criterion):

MDNBR RCS Pressure MSS Pressure buS. a)

(usi~a Uncontrolled rod withdrawal from sub-critical 3.218/1.14 M-2750 Ml1210 Conclusions Considering the conservative assumptions used in the accident analysis, it is concluded that in the unlikely event of a control rod withdrawal accident the core and reactor coolant systems are not adversely affected. The peak heat flux reached remains less than the nominal full power value. DNBR is well above its limiting value. The peak average clad temperature is less than its nominal full power value, and thus there is no possibility of fuel or clad damage.

14.1.2 UNCONTROLLED RCCA WITHDRAWAL AT POWER Accident Description An uncontrolled RCCA withdrawal at power results in an increase in core heat flux. Since the heat extraction from the steam generator remains constant until the steam generator pressure reaches the relief or safety valve setpoint, there is a net increase in reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise would eventually result in DN3. Therefore, to prevent the possibility of damage to the cladding, the Reactor Protection System is designed to terminate any such transient before the DNBR falls below its limit.

Rev. 16 14.1-4 12/01/2000

The automatic features of the Reactor Protection System which prevent core damage in an RCCA withdrawal incident at power include the following:

1. Nuclear power range instrumentation actuates a reactor trip if two-out-of-four channels exceed an overpower setpoint.

2. Reactor trip is actuated if any two-out-of-four AT channels exceed an overtemperature AT setpoint. This setpoint is automatically varied with axial power distribution, temperature and pressure to protect against DNB.

3. Reactor trip is actuated if any two-out-of-four AT channels exceed an overpower AT setpoint. This setpoint is automatically varied with axial power distribution and temperature to ensure that the allowable fuel power rating is not exceeded.

4. A high-pressure reactor trip, actuated from any two-out-of-three pressure channels, is set at a fixed point. This set pressure is less than the set pressure for the pressurizer safety valves.

5. A high pressurizer water level reactor trip, actuated from any two-out-of-three level channels, is set at a fixed point. This affords additional protection for RCCA withdrawal incidents.

6. In addition to the above listed reactor trips, there are the following control rod assembly withdrawal blocks:

+ High nuclear power (one-out-of-four)

+ High overpower AT (two-out-of-four)

+ High overtemperature AT (two-out-of-four)

Method of Analysis The purpose of this analysis is to demonstrate the manner in which the above protection systems function for various reactivity insertion rates from different initial conditions. Reactivity insertion rates and initial conditions govern which protective function a

s performed using several digital computer codes.'The reactor protection f ris are "mcorpo dto the transient analysis digital simulation of the Nucle a

Supply System. The system se to the transient is then used as a trans ttring function for the fuel thermal hydraulic ana and DNBR assessment.

In order to obtain conservatively low ollowing assumptions are made:

1. Initial conditions as aximumi power and reactor cool eratures and minimum pressure; i.e.

ower is assumed 2% high, the average temperature i ed 4-F high, and ressure is assumed [M psi low. This gives the minimum initial mar D

Rev. 16 14.1-5 12/01/2000

Insert A This transient is analyzed using the RETRAN code. This code simulates the neutron kinetics, RCS, presssurizer relief and safety valves, pressurizer spray, SG, and SG safety valves. The code computes pertinent plant variables including temperatures, pressures and power level. The core limits, as illustrated on Figure 14.0.2, are used to develop input to RETRAN to determine the minimum DNBR during the transient.

In order to obtain a conservative value for the minimum DNBR, the following analysis assumptions are made:

I.

This accident is analyzed with the Revised Thermal Design Procedure (Section 3.2).

Therefore, initial reactor power, pressure, and RCS temperatures are assumed to be at their nominal values. Uncertainties in initial conditions are included in the limit DNBR.

2.

Reactivity Coefficients - Two cases are analyzed.

a.

Minimum Reactivity Feedback - A zero moderator temperature coefficient of reactivity (0 pcm/°F) is assumed at full power. For power levels less than or equal to 60% power, a positive moderator temperature coefficient of reactivity

(+5 pcm/ 0F) is conservatively assumed, corresponding to the beginning of core life. A conservatively small (in absolute magnitude) Doppler power coefficient is used in the analysis.

b.

Maximum Reactivity Feedback - A conservatively large positive moderator density coefficient and a large (in absolute magnitude) negative Doppler power coefficient are assumed.

3.

The reactor trip on high neutron flux is actuated at a conservative value of 118% of nominal full power. The overtemperature AT trip includes all adverse instrumentation and setpoint errors. The delays for trip actuation are assumed to be the maximum values.

No credit was taken for the other expected trip functions.

4.

The rod cluster control assembly trip insertion characteristic is based on the assumption that the highest worth assembly is stuck in its fully withdrawn position.

5.

A range of reactivity insertion rates is examined. The maximum positive reactivity insertion rate is greater than that which would be obtained from the simultaneous withdrawal of the two control rod banks having the maximum combined differential rod worth at a conservative speed (45 inches/minute, which corresponds to 72 steps/minute).

6.

Power levels of 10%, 60% and 100% of full power are considered.

7.

The impact of a full power RCS vessel Tvg window was considered for the uncontrolled RCCA bank withdrawal at power analysis. A conservative calculation modeling the high end of the RCS vessel T.vg window was explicitly analyzed.

The effect of rod cluster control assembly movement on-the axial core power distribution is accounted for by causing a decrease in the overtvmperature AT trip setpoint proportional to a decrease in margin to DNB.

Results Figures 14.1.2-1 through 14.1.2-4 show the response of nuclear ower, pressure, average coolant temperature, and DNBR to a rapid RCCA withdrawal ksec) incident starting from full power. This reactivity insertion rate is greater than that for the two highest worth banks, both assumed in their highest incremental worth region, withdrawn at their maximum speed. Reactor Trip on high nuclear power occurs less than 2.0 seconds from the start of the accident. Since this is rapid with respect to the thermal time constants small changes in T.vg and pressure result. A large margin to the MDNBR limit is maintained.

The response of nuclear power, pressure, average coolant temperature, and DNBR for a slow RCCA withdrawal (Jf.0E-5 Ak/sec) from full power is shown in Figures 14.1.2-5 through 14.1.2-8. Reactor Trip occurs on overtemperature AT. The rise in temperature and pressure is larger than for the rapid RCCA withdrawal. The minimum DNBR reached during the transient is greater than the "DNBR limit.

nuclear power, RCS pressure, coolant average temperature, and DNBR responses for RCC jthdrawal from 60% power are shown in Figures 14.1.2-9 through 14.1.2-Yor a rapid withdr Irate (8.2E-4 Ak/sec) and in Figures 14.1.2-13 through 14.1.2-6for a slow withdrawal rate (1.t

-5 Ak/sec). The results demonstrate that the ovg temperature AT and high nuclear flux trip fin--

ns adequately protect the fuel. The urn DNBR reached is above the MDNBR limit.

The following table shows the compariso -o th portant calculated safety parameters to their respective acceptance criteria (Calcul edAcceptance Criterion):

Uncontrolled rod withdrawal MMDNBR RC essure MSS Pressure Fast Rate Full Pow*pr--

I 1.14 Mt750 I

121 Slow Rate lPdwer 1.362/1.14 E02750 Fast termediate Power HM1.14 MM/2750 1

Rate Intermediate Power

/1.14 2350/2750 1182/110 Conclusions In the unlikely event of an RCCA withdrawal incident during power operation, the core and Reactor Coolant System are not adversely affected since the minimum value of the DNBR reached is greater than the DNBR limit for all RCCA reactivity rates. Protection is provided by the high nuclear flux,`.

and overtemperature AT trip functions.

Rev. 16 14 1-_6 12/01/2000 tor fficient of reactivity is assumed correspondin' t

ngo core life. A conservativeyWo1 a

ýolute oppler reactivity coefficient is ;e

eref, v

cients resu u.r of negative feedback repa

ý,

eref~ore, higher peak powers and temperatures.

,It1 V

Insert B Figures 14.1.2-9 through 14.1.2-11 show the minimum departure from nucleate boiling ratio as a function of the reactivity insertion rate for the three initial power levels (100%, 60%, and 10%)

and minimum and maximum reactivity feedback. It can be seen that the high neutron flux and overtemperature AT trip channels provide protection over the whole range of reactivity insertion rates. The minimum DNBR is never less than the limit value.

In the referenced figures, the shape of the curves of minimum departure from nuclear boiling ratio versus reactivity insertion rate is due both to reactor core and coolant system transient response and to protection system action in initiating a reactor trip.

Referring to Figure 14.1.2-11 for example, it is noted that:

1. For high reactivity insertion rates (i.e, between -100 pcrn/second and -30 pcm/second) when modeling minimum reactivity feedback, reactor trip is initiated by the high neutron flux trip. The neutron flux level in the core rises rapidly for these insertion rates, while core heat flux and coolant system temperature lag behind due to the thermal capacity of the fuel and coolant system fluid. Thus, the reactor is tripped prior to significant increase in heat flux or water temperature with resultant high minimum departure from nucleate boiling ratios during the transient. Within this range, as the reactivity insertion rate decreases, core heat flux and coolant temperatures can remain more nearly in equilibrium with the neutron flux; minimum DNBR during the transient thus decreases with decreasing insertion rate.

2. With further decrease in reactivity insertion rate, the overtemperature AT and high neutron flux trips become equally effective in terminating the transient (e.g., at an approximately 30 pcm/second reactivity insertion rate).

The overtemperature AT reactor trip circuit initiates a reactor trip when measured coolant trip AT exceeds a setpoint based on measured reactor coolant system average temperature and pressure.

It is important in this context to note, however, that the average temperature contribution to the circuit is lead-lag compensated in order to decrease the effect of the thermal capacity of the reactor coolant system in response to power increases.

For reactivity insertion rates between -30 pcm/second and -8 pen/second, the effectiveness of the overtemperature AT trip increases (in terms of increased minimum departure from nucleate boiling ratio) due to the fact that, with lower insertion rates, the power increase rate is slower, the rate of rise of average coolant temperature is slower, and the system lags and delays become less significant.

3. For reactivity insertion rates of -8 pcmlsecond and lower, the rise in reactor coolant temperature is sufficiently high so that there is more steam relief through the steam generator safety valves prior to trip. This steam relief acts as an additional heat sink on

the reactor coolant system and sharply decreases the rate of iise of reactor coolant system average temperature. This causes the overtemperature AT trip setpoint to be reached later with resulting lower minimum departure from nucleate boiling ratios.

oooz/o/zit 91

-Aa) 5.' -

V"I-b ~*.

[s] aW!l 8L 9L Vt Z L

0 t 8

9 V,

0 0

I 0

S.......

i

. ° 09LC0 0001

9 9"0

.~..... :.....

q

....... ~...

=. u

.. 09'L LC'L OW!J -SA JaMOd JO3leaN m d%0O L-oeu ise-i - ietaeJPqi!M voOu PliiojiUOoUR

-u 0

(D

=3

-4' 0

3 0o

0 0

E (D

I 0

0 0

3 Z' t V, I L-Z-L--vL a-inbij

[S] awl I

DWIL SA J;9mOd -IOIOD@d

-J@mOd %00 1 GIDd JSDJ -

II)MDJpql!M Vood p@jjo-jjuooufj

2400.

2350.0 2300.0 2250.0 2200.0 2150.0 2100.0 2050.0 2000.0 1950.0 1900.0 1850.0 180.0 Figure 14.1.2-2 Rev.

16 12/0112000 C2 CD U)

EL Uncontrolled RCCA Withdrawal - Fast Rate 100% Pow r Pressurizer Pressure vs. Time t

t.

0..-...................

0 2

4 6

8 10 12 14 16 18 Time [s]

Uncontrolled RCCA Withdrawal -

Fast Rate 100% Power Pressurizer Pressure vs Time 2400 S2250

-+

I I

3 3

I 3

3 3

I 3

215

- -l I

I

,'I I

i 3

I 2300 -- -.

+-------------------------

21000 SI I

SI I

II I

3 I

II I

3-I I

0I I

I 20 0,

I I

- I I

I Time [s]

Figure 14.1.2-2

oooz/1o/zl 91 -^al c-v'rtr mnf!d.

[s] aU 8[

9L VL t IL 01-8 9

Vz0/

I I

./

........ {..

...........4 ------------ --.

~~~~~~~~--

=----

o'SZ 0"£179 0"999 0"999 0§SLS 0.s99 CD B

CD ru.

CD CD

-rn 0"S69 0"909 O"£L9 0"9[9 OWJUL -SA SAe2jI od

%OO L eieu ise8l - IeMJpqLi!M VOOB PaIOJIUOUOUN N

'~rrfl-o 5Jj. I/P )1 24j tI Cf 8

9 17 Oi fJI

\\

uI-L

Uncontrolled RCCA Withdrawal -

Fast Rate 100%o Power Tave vs Time 590 58

-+

1 3i,3 3I 3

Ip

-I 580 I

3I I

3.

3-3 33 57 3

58-

+-----

I I

3 I

565

+

560 Time [s]

Figure 14.1.2-3

oooz/Io/ZT 91 -AZU x*

[s] eWIJl

8.

9L tL 1L

01.

8 9

v 3

0 S---------- -.

S---------

aWJ

-SA "s

fINO WflwnwuiJu oGM~

d %00. PH SlEs - leMeJpql!V OH pailoJiuooun 0L'O 00" l 3=

3 0 z wl xJ 0S'l GL't SLgZ oo0 V-Zr-pj;mnýttj

Uncontrolled RCCA Withdrawal -

Fast Rate 100% Power Minimum DNBR vs Time Time [s]

Figure 14.1.2-4 5

4.2 W

0 E

E3 E

3.4 2.6 1.8

Uncontrolled RCCA Withdrawal - Slow Rate 100% Power Reactor Power Vs. Time 1.375\\

/

1.250

',,u.............

1.

125 ----------

1.000.........

0 7 5 0

"...."* """I"I Il 0.625 0 5 0 0"

.......... I u,.o".'*

0.

5 nnnn 6

12 18 24 30 36 42 48 54 60 Time [s]

Figure 14.1.2-5 Rev.

16 12/01/2000 F

0 0

C'

'4 o,

0 N

Uncontrolled RCCA Withdrawal -

Slow Rate 100% Power Reactor Power vs Time 1.4 I

1.2-

+----------------

I o

I I

'4

.43 -

3+

L J--.

3 I

I I I

I I

SI I

I I

I i

I 30 Time [s]

Figure 14.1.2-5

Uncontrolled RCCA Withdrawal - Slow Rate 100% Power Pressurizer Pressure vs. Time 2400.0

"/ l 2 3 0 0...0 Ziiiiiiil S

2 50.0 2300.0................... i.....

iii i_..

2250.0 2150.0

~~~~~~~~~~~~~~~~-

...i".

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

2100.0 0

2 0 5 0.0........

19 5 0.0 --------

1900.0 - ----.....

....t......

18

0.

0) 6 12 18 24 30 Time [s]

36 42 48 54 Figure 14.1.2-6 Rev.

16 12/01/2000

0)

L..

C,,

Uy)

(D2 I,. ci_

Uncontrolled RCCA Withdrawal -

Slow Rate 100% Power Pressurizer Pressure vs Time 2400 C./

2250 -

V1)

CL CL 4 --

I I2150 I

II 2100

-I SIII 2050 I

2000 Time [s]

Figure 14.1.2-6

oooz/lo/zt 91 -_aa L-Z" FI'*T ain$i!g

[s] GW!Il 09 V9 8t7 gf7 9C 0S lz 8L L

U*

L-9 0

u.ssg 9

I------..

S S

0 9 9 S...................................................................................

- ------- O 0

9 0 SWIj

-SAGOA'ej

=

+/-v..

1ý ý -n A.....

, i i i

i o.

l.........

.....i.......".........'.......i ".......i.......

n. /--,,

-I CD 3

-a CD (0

"1

Uncontrolled RCCA Withdrawal -

Slow Rate 100% Power Tave vs Time 590 I J 3

I-3 E"

575 -

-+

-1 565 ------------------------------------------------------

560 I

3 I

3 I,3 Time [s]

Figure 14.1.2-7

oooz/[o/zt 9t

-Aaa

[s] owl j

)9 IV9 8V7 gzv 96 06 17;

81.

?L 9

0

-2

/n fnvA i.......

Z96' HEISNGV]

Wnlwul.j S.........

.~~~ ~ ~...........

t.

V I':[ = 11 WI.N" ' d/IH I

-/

S.........

I I

ii i

W~i_

-SA HEING wflw!u!iA mOd %00 L. 011j mOIS -

emeJPqI!M VO U a8JjoJo3uf 1

c

-rUJLL -t"s EItN Akuntu u`Pv V

%

J 93,0 09"0 OT'O OL0 001L 09 i-K 3 0 a

3)

I (flC'"

L 8-Z'['p[ a.infi!,A

Uncontrolled RCCA Withdrawal Slow Rate 1 00% Power Minimum DNBR vs Time 5

3 3

.t E

3 IP

-II3 3

33

-II333 4.2 -

1 Time [s]

Figure 14.1.2-8

Unco-ntrolled RCCA Withdrawal - Fast Rate 60% Power Reactor Power vs. Time 1 25.......................

1.250 0.8 7 5 -.-

E

" 0.6 2 5.........

05 y

0.0 0

-- 0.2 5 0.............................-

-- 7 --

0.1 2 5 0.3 5 i -.---...................

0.000 2

4 6

8 10 12 14 16 18 0

Time [s]

Figure 14.1.2-9 Rev. 16 12/01/2000

Uncontrolled RCCA Withdrawal -

Full Power Minimum DNBR vs Reactivity Insertion Rate Minimum Reactivity Feedback Maximum Reactivity Feedback 1.75 1.7 4---

HIGH NEOTRON FLUX TRIP 1.6 -

1.55 1.5 -

1.45 OTDT TRIP I

I I

. I

-1 10 I,

0 10 Reactivity I

I I I II

HIGH NEUTRON FLUX TRIP SI------------------------

OTDT TRIP I I I I I II II I

I I I I 3

10 2

10 10 Insertion Rate [pcm/second]

Figure 14.1.2-9 1.65 -

/

I I

I E

E 3

I

oooz/[o/zl 9t "A*I 0

81.

91.

K1

[s] ewu!i EL 01 8

9 E

0 (VAAo I

..........."........... }............ ----------- --....

S...

~----------------

...............t...........................-...t.....

ii Ii I I Iii ii O*Oso L O'O061 O'OOtZ o0,0oo CD O'OOZ 0"00)

\\

0"006Z n'nni7

UW!l -SA ojnssaJd JaZ!JnSsOJd JaMOd %09 EleBl ISE-I - leaMeJpql!m VOOl P8IIOJiUOUfl SC\\Cmo

~

Ai

.YI 7V'1 tSI

-Y 0 [-z'fI'P[ aan1U

Uncontrolled RCCA Withdrawal -

60% Power Minimum DNBR vs Reactivity Insertion Rate Minimum Reactivity Feedback

-Maximum Reactivity Feedback I

I

+-----------------------------+--------------

I I

'I OTDT TRIP 1.7 2

1.9 1.8-II

'I 1.6 ------------

OTDT TRIP I

I III III.

-1 10 r III ii!

0 10 10 Reactivity Insertion Rate I

III!

II.

I I

I I I 1 3

10 2

10

[pcm/second]

Figure 14.1.2-10

'I OTDT TRIP I-12::

z E

E3 E

1.5 1.4 I

I I

I I I

'/

I l I

S~Uncontrolled RCCA Withdrawal - Fast Rate 60%Y Power 625.0 Tare vs. Time

/

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

655.0 ----------

S5 5 0 CL.0 5 6.0...............

i-.........

I..

5 5 5.....

0 -.

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

5 4.-...---

I ---

525.0 S2 4

6 8

10 12 14 16 18 Time [s]

Figure 14.1.2-11 Rev.

16 12/01/2000

Uncontrolled RCCA Withdrawal -

10% Power Minimum DNBR vs Reactivity Insertion Rate Minimum Reactivity Feedback S....-

Maximum Reaciivity Feedback I

I I

I ii I

1 I

I I

ii I

I I

+----------------------------+---------------------------------------------------------

I I

I II I

I I

If I

I I

ii.

--------------+------------------------+-----------------------

- -- - --+--- -

OTDT TRIP 1.7 ----------

1.6 -

1.54 HIGH NEUTRON FLUX TRIP ii liii I

I I

0 1

10 10 Reactivity Insertion Rate liii I II

1 2 10

[pcm/second]

I I

I I I I 3

10 Figure 14.1.2-11 2.2 2.1 2-1.9 1.8 z

E

=3 E"

1.4 I

I i I I

-1 10 T

i T

I I

I I I I I I I I T

V r I W T

=

/k I

! Jl I,

l

OOoz/10/zl 91 -&.aa 038\\9 9L t4

[s] awl I zL-0L 8

9 V

3 0

.~~~~~~~~

~ ~

}..............

~~~~~~~~.

/.....

"IPI

=CI 1duR euWIl -SA UI8NCI uwnwiuil~l

/J/mOd %09 e0eH lsl-! - lemuipqlm voouI PliiOJiUooun 09'0 00" L 3

C::

3 0 z m

10 00"71 00",6 zt-z" [pt[a-in'ýt.

Uncontrolled RCCA Withdrawal - Slow Rate 60% Power Reactor Power vs. Time 1.375 1.250 1.125 1.000 0.875 0.750 0.625 0.500 0.375 0.250 0.125 n nnn 25 50 75 100 125 150 175 200 Time [s]

225 2 Figure 14.1.2-13 Rev. 16 12/01/2000 E

0 0

0 a_

/..-........

°.....°°.°°

.o..°°*

~~

°...........

---.. --- I......

[s] awlUI SL L 091 93 L 00L GL 09 I-.fIfI I

U UUU 0,0981L 0*0061 0'096 L 0,0003 0,0903 0,00 2~

z-- -

i

...........l...

JgM OWIJ "SA oJnssaJcl jazunssaJd 0,4d %09 aieu~ mols - loeimeJlm VOOu~ PallOJiUOOUN O'OOZZ 0.09cz 0"008E ooozifolzi 91 "A*I 0\\9

-u CD Co CD 7o f J) 933 003 P I-Z"

'P n*.m*4

Uncontrolled RCCA Withdrawal - Slow Rate 60% Powe!

Tave vs. Time 625.0 6 15.0 r----- -----

595.0.......................

"- 585.0.................................... -

575."

545.0....................

535.

"............I 545.0 ------

5 3.----

i I..

525.0 0

25 50 75 100 125 150 175 200 225 0

Time [s]

Figure 14.1.2-15 Rev. 16 12/01/2000

Uncontrolled RCCA Withdrawal - Slow Rate 60% Power Minimum DNBR vs. Time 2.70 I......

1.0*PD i

i~~i i

I..

.o......

2.5........

C.c "2::

0 E

0.50

\\

0.25.........-.-...............

1.00 -----

Figure 14.1.2-16 Rv16o 0./0/00

14.1.3 RCCA MISALIGNMENT Accident Description RCCA misalignment accidents include:

a. dropped full-length RCCAs;

b. dropped full-length RCCA banks; and

c. statically misaligned full-length RCCAs.

Each RCCA has a rod position.indicator channel which displays position of the assembly. The displays of assembly positions are grouped for the operator's convenience. Fully inserted assemblies are further indicated by rod bottom lights. Bank demand position is also indicated. The full-length assemblies are always moved in pre-selected banks and the banks are always moved in the same pre-selected sequence.

Dropped assemblies or banks are detected by:

a. sudden drop in the core power level

b. asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples

c. rod bottom light(s)

d. rod deviation alarm (if the plant computer is in operation).

Misaligned assemblies are detected by:

a. asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples

b. rod deviation alarm (if the plant computer is in operation).

The resolution of the rod position indicator channel is +/- 5% of span or 7.2 inches (span equals 12 feet). Deviation of any assembly from its bank by twice this distance, 10% of span, or 14.4 inches, will not cause power distributions worse than the design limits.

If one or more rod position indicator channels is not operable, the operator will be fully aware of the I of the channel, and special surveillance of core power tilt indications, using established procedures and relying on ex-core nuclear detectors and/or movable in-core detectors, will be used to verify power distribution symmetry.

Rev. 16 14.1-7 12/01/2000

Method of Analysis REPý- 5 R

S T

ýr t powthermnal hydraulic analysis. The peak fuel rod F. in the analysis is i r*eased to adequately bo-aukthe core power distribution anticipated during the s y state RCCA misalignment and drop core conditions. The reactor is as not to trip.

it is necessary to show on a reloa cle specific basi at the worst'dropped or misaligned RCCA does not result in a peak fuel rod p that is greater than the F. assumed in the safety analysis.

Rod drop in automatic con is not analyzed since restrictio n control rods in automatic control are imposed en reactor power is > 90% and control ro e inserted to < 215 steps. This re

ction ensures that the rod drop in automatic control acci e

bounded by the sta '

CA misaligniment described above. Removal of the rod drop in automa" ontrol Setermined to be acceptable by the NRC in Reference 1.

Results PAEWT SRTne in al hws the compari son--of the important calculae-sft Stheir respectivec rtei Clulated Valu/cepac C t

MDN CS Pressure MSS Pressure Control Rod Drop a Misa "

ý- 2.02 1.142/1.14 2250/2750 750 Conclusions Dropped or misaligned RCCAs are not deemed to be a hazard to the safe operation of the plant because these events are clearly indicated to the operator, and the analyzed cases of the worst misaligned and dropped rod do not result in a DNBR less than the #DNBR limit.

14.1.4 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION Accident Description Reactivity can be added to the core with the Chemical and Volume Control System by feeding reactor makeup water into the Reactor Coolant System via the Reactor Makeup Control System. Boron dilution is a manual operation. A boric acid blend system is provided to permit the operator to match the concentration of reactor coolant makeup water to that existing in the coolant at the time.

45u1-8 Rev. 16 14.1-8 12/01/2000

USAR Insert 14.1.3-1 For the dropped RCCA(s) event, the transient response is calculated using the LOFTRAN code.

The code simulates the neutron kinetics, reactor coolant system, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generators, and steam generator safety valves.

The code predicts pertinent plant variables including temperatures, pressures and power level.

Dropped RCCA(s) statepoints are calculated and nuclear models are used to obtain a hot channel factor consistent with the primary system conditions and reactor power. Using the primary conditions from the transient analysis and the hot channel factor from the nuclear analysis, the VIPRE code is used to calculate the minimum DNBR to demonstrate that the DNB design basis is satisfied. The transient response, nuclear peaking factor analysis, and DNB design basis confirmation are performed in accordance with the methodology described in Reference 16.

For the RCCA misalignment event, steady-state power distributions are analyzed using the appropriate nuclear physics computer codes. The peaking factors are then used as input to the VIPRE code to calculate the DNB ratio (DNBR). The following cases are examined in the analysis assuming the reactor is initially at full power: the worst rod withdrawn with bank D inserted at the insertion limit, the worst rod dropped with bank D inserted at the insertion limit, and the worst rod dropped with all other rods out. It is assumed that the incident occurs at the time in the cycle at which the maximum all-rods-out FA& occurs. This assures a conservative FAN for the misaligned RCCA configuration.

USAR Insert 14.1.3-2 One or More Dropped RCCAs Single or multiple dropped RCCAs within the same group result in a negative reactivity insertion.

The core is not adversely affected during this period since power is decreasing rapidly. Either reactivity feedback or control bank withdrawal will re-establish power.

Following a dropped rod event in manual rod control mode, the plant will establish a new equilibrium condition. Without control system interaction, a new equilibrium is achieved at a reduced power level and reduced primary temperature. Thus, the automatic rod control mode of operation is the limiting case.

For a dropped RCCA event in the automatic rod control mode, the rod control system detects the drop in power and initiates control bank withdrawal. Power overshoot may occur due to this action by the automatic rod controller, after which the control system will insert the control bank to restore nominal power. Figures 14.1.3-1 through 14.1.3-4 show a typical transient response to a dropped RCCA event with the reactor in automatic rod control. In all cases, the minimum DNBR remains above the limit value.

Dropped RCCA Bank A dropped RCCA bank results in a negative reactivity insertion greater than 500 pcm. The core is not adversely affected during the insertion period because power is decreasing rapidly. The transient will proceed similar to that described previously for the one or more dropped RCCAs scenario, but the return to power will be less due to the greater negative reactivity worth of an entire RCCA bank. The power transient for a dropped RCCA bank is symmetric.

Statically Misaligned RCCA The most severe RCCA misalignment situations with respect to DNB at significant power levels are associated with cases in which one RCCA is fully inserted with either all rods out or bank D at the insertion limit, or where bank D is inserted to the insertion limit and one RCCA is fully

withdrawn. Multiple independent alarms, including a bank insertion limit alarm, alert the operator well before the transient approaches the postulated conditions.

The insertion limits in the Technical Specifications may vary from time-to-time depending on several limiting criteria. The full power insertion limits on control bank D must be chosen to be above that position which meets the minimum DNBR and peaking factors. The full power insertion limit is usually dictated by other criteria. Detailed results will vary from cycle-to-cycle depending on fuel arrangements.

For each case, DNB does not occur for the RCCA misalignment incident, and thus there is no reduction in the ability of the primary coolant to remove heat from the fuel rod. The peak fuel temperature corresponds to a linear heat generation rate based on the radial peaking factor penalty associated with the misaligned RCCA and the design axial power distribution. The resulting linear heat generation rate is well below that which would cause fuel melting.

Dropped RCCA Representative Transient Response - Nuclear Power vs. Time 0

50 100 Time (s) 150 200 Figure 14.1.3-1 1 9 1.1) 1.

.9 -

0 3:

CL.

=3

.74-

.6-4

.5 I

I I

P I

T i

.8-I

Dropped RCCA Representative Transient Response - Core Heat Flux vs. Time 1.2 S1.1

.9 L._

(15 0

.7 C-)

.6 I II I

I II III I

I Time (s)

Figure 14.1.3-2

Dropped RCCA Dropped RCCA Representative Transient Response - Pressurizer 2400 -r 0

50 100 Time (s)

Pressure vs. Time 150 Figure 14.1.3-3 1

2300 a 2200 L.

2100 N4 "E 2000 U)

U)

L 0-. 1900 1800 I

I I

I 200

ENew Fiý\\xtia be A-dW~e Dropped RCCA Representative Transient Response - Vessel Average Temperature vs. Time 600 c'580

- 560

()

E S540

__ 520 500 100 lime (s)

Figure 14.1.3-4

SMethod of Analysis peo ersafety analysis for the RCCA misalignment and RCCA drop accidents involves aý er fuel thermal hydraulic analysis. The peak fuel rod F. in the analysis is incre ed to adeq ately bound the core power distribution anticipated during the steady st RCCA misali uent Md RCCA drop core conditions. The reactor is assumed not to

p.

It is necess to show on a reload cycle specific basis that the worst dr ped or misaligned RCCA does n result in a peak fuel rod power (F.) that is greater th e FA assumed in the safety analysis.

Rod drop in automa i control is not analyzed since restrict ns on control rods in automatic control are imposed en reactor power is > 90% an control rods are inserted to < 215 steps. This restriction es that the rod drop in a matic control accident is bounded by the static RCCA rmsalignm t described above, oval of the rod drop in automatic control was determined to be accepta e by the NRC i eference 1.

d 4.

1.3 Results 77 The following table shows the copan n of the important calculate safety parameters to their respective acceptance crit a (Calcu ted Value/Acceptance Criterion):

MDNBR RCS Pressure MSS Pressure

/i(s Control Rod Drop and Misalignment F

.02 1.142/1.14 22 2750 750/1210 Conclusions Dropped n

RCCAs are not deemed to be a hazard to the s operation of the plant because ese events are clearly indicated to the operator, and the analyz cases of the worst rmi~salit ed and dropped rod do not result in a DNBR less than the MI)NBRt F

all cases of dropped banks, the reactor is tripped by the power range negative neu n flux rate trip and consequently dropped banks do not cause core damage.

14.1.4 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION Accident Description Reactivity can be added to the core with the Chemical and Volume Control System by feeding reactor makeup water into the Reactor Coolant System via the Reactor Makeup Control System. Boron dilution is a manual operation. A boric acid blend system is provided to permit the operator to match the concentration of reactor coolant makeup water to that existing in the coolant at the time.

Rev. 16 14.1-8 12/01/2000

The Chemical and Volume Control System is designed to limit, even under various postulated failure modes, the potential rate of dilution to a value that after indication through alarms and instrumentation, provides the operator sufficient time to correct the situation in a safe and orderly manner.

The source of reactor makeup water for the Reactor Coolant System is the reactor makeup water storage tanks.

Inadvertent dilution can be readily terminated by isolating this source. The operation of the reactor makeup water pumps, which take suction from these tanks, provides the only supply of makeup water to the Reactor Coolant System. In order for makeup to be added to the Reactor Coolant System the charging pumps must be running in addition to the reactor makeup water pumps.

There are three positive displacement variable speed drive charging pumps, manually or automatically controlled. When in automatic, each is provided with a high and low speed alarm. However, only one of them is automatically controlled at any one time, as dictated by procedure.

The rate of addition of unborated makeup water to the reactor coolant system is limited by the capacity of the charging pumps and by the capacity of the control valve between the two makeup water pumps and the three charging pumps. The maximum dilution flow (80 gpm) occurs with two charging pumps operating and three letdown orifices in-service. -For-the-During normal operation, two charging pumps are operated; one in manual and one in automatic control. The speed of the pump selected for automatic control is controlled by the pressurizer level error signal. During load changes the pressurizer level set point varies automatically with T.v, such that the charging pump speed) remains relatively constant.

%r-j-r,

r1-1eyi.

S t

,.i 1

,m dCtQ-r O,, 410t0 The boric acid from the boric acid tank is blended with the reactor makeup water in the blender and the composition is determined by the preset flow rates of boric acid and reactor makeup water on the Reactor Makeup Controller. Two separate operations are required. First, the operator must switch from the automatic makeup mode to the dilute mode. Second, the control switch must be actuated. Omitting either step would prevent dilution. This makes the probability of inadvertent dilution very small.

Information on the status of the reactor coolant makeup is continuously displayed. Lights are provided on the control board to indicate the operating condition of pumps in the Chemical and Volume Control System. Alarms are actuated to warn the operator if boric acid or demineralized water flow rates deviate from pre-set values as a result of system malfunction.

To cover all phases of plant operation, boron dilution during refueling, startup, and power operation are considered in this analysis.

Rev. 16 14.1-9 12/01/2000

Method of Analysis and Results Dilution During Refueling During refueling the following conditions exist:

a. One residual heat removal pump is running to ensure continuous mixing in the reactor

vessel,

b. The valve in the seal water header to the reactor coolant pumps is closed,

c. The valves on the suction side of the charging pumps are adjusted for addition of concentrated boric acid solution,

d. The boron concentration of the refueling water is a minimum of Pm, corresponding to a shutdown of at least 5% Ak/k with all control rods in; periodic sampling ensures that this concentration is maintained,

e. The source range detectors outside the reactor vessel are active and provide an audible count rate.

The operator has prompt and definite indication of any boron dilution from the audible count rate instrumentation. High-count rate is alarmed in the reactor containment and the main control room. The count rate increase is proportional to the inverse core multiplication factor. Assuming the reactor is 5% shutdown at the required refueling boron concentration of 2.,5 ppm, the time to reach critical conditions is> 30 minutes. This is ample time for the operator to recognize the audible high-count rate signal and isolate the reactor makeup water source by closing valves and stopping the reactor makeup water pumps.

Dilution During Startup During startup the following are assumed for a boron dilution event:

+ Core monitoring of neutron flux is provided by the excore detectors.

+ Reactor coolant is mixed by operation of the reactor coolant pumps.

FO

.f30ieecharging pumps are running, delivering a maximum dilution flow rate ofJ8(J gpm.

+ The boron endpoint with all rods inserted is 4300pm.

+ Initial reactor boron concentration isimppm.

0(C0 An evaluation of the reactor shows that the minimum time required to reduce the reactor coolant boron concentration to a concentration at which the reactor could go critical with all RCCAs in is > 15 minutes. This provides adequate time for the operator to respond to the high-count rate signal and terminate dilution flow.

Rev. 16 14.1-10 12/01/2000

Dilution at Power 46" -

70 The reactiviy addition te gM a boron dilution flow of.,.lOgppm at full ower uall charging pumps runninf This is a c t

high Sreactivity insertion rate for the assumed at power boron concentration of 1-0-ppm.

With the reactor in automatic control, at full power, the power and temperature increase from the boron dilution results in the insertion of the controlling RCCA bank and a decrease in shutdown margin. A continuation of the dilution and RCCA insertion would cause the rods to reach the lower limit of the maneuvering band. Before reaching this point, however, two alarms would be actuated to warn the operator of the potential accident condition. These two alarms, the low RCCA insertion limit alarm and the low-low RCCA insertion limit alarm, alert the operator to initiate normal boration.

I With no boration, the required shutdown margin is maintained for at least 10 minutes d i g a

ntinuous boron dilution. Therefore, ample time is available following the al for the opera to determine the cause, isolate the reactor water makeup sou

,and initiate reboration.

J.

If rod control is in m-uarpal, and the operator takes no action, th - ower rises to the high neutron flux trip setpoint and the ctor trips. Figures 14.1.4-1 ough 14.1.4-5 show the response of nuclear power, pressure, caolant average temr e, heat flux, and DNBR to a boron dilution event in manual control.

boron tion in this case is essentially identical to a rod withdrawal accident. The reactiviy ertion rate due to the boron dilution is within the range of reactivity insertioný rates nside in Section 14.1.2 - Uncontrolled RCCA Withdrawal at Power. Assumnin

% shutdown

,there is ample time available for the operator to terminate the dil *on before the reactor can to criticality following the trip.

The following table ows the comparison of the important calc ted safiety parameters to their respective ceptance criteria (Calculated Value/Acceptance C

'on):

RCS Pressure MS essure MDNBR (sia')

bi emical & Volume Control 13711 R25 System Malfunction 1.347/1.14 2750 E/1210 Conclusions Because of the procedures involved in the dilution process, an erroneous dilution is considered unlikely. Nevertheless, if an unintentional dilution of boron in the reactor coolant does occur, numerous alarms and indications are available to alert the operator to the condition. The maximum reactivity addition rate due to the dilution is slow enough to allow the operator adequate time to determine the cause of the dilution and take corrective action before required shutdown margin is lost. The dilution event at power is showi.n to have adequ,-M nargin to thpe

.MD9,PNB imit-.

Rev. 16 14.1-11 12/01/2000

Insert I for FSAR Section 14.1.4:

The low alarm is set sufficiently above the low-low alarm to allow normal boration without the need for emergency procedures. If dilution continues after reaching the low-low alarm, it takes approximately 37.58 minutes before the total shutdown margin is lost due to dilution. Adequate time is therefore available following the alarms for the operator to determine the cause, isolate the reactor makeup water source, and initiate reboration.

With the reactor in manual control, if no operator action is taken, the power and temperature rise causes the reactor to reach the OTAT trip setpoint. The boron dilution accident in this case is essentially identical to a rod cluster control assembly withdrawal accident at power. Prior to the OTAT trip, an OTAT alarm and turbine runback would be actuated. There is time available

(- 34.76 minutes) after a reactor trip for the operator to determine the cause of dilution, isolate the reactor makeup water source, and initiate reboration before the reactor can return to criticality.

oooz/fo/zi

-- ip[a[*,

808

[s] auJ{1 9[SI"8W!I O0 06 09 OL 09 09 01V 0C 3 ;

0 L 0 I00"0 I

00C

.. 0

'0 OSL *O 00£0 o...........................................

9" c

~~~~~- ---

...... T "0

OW!I "SA JaMOd 13ooa~B lojluoo lenuleVq - JaMOd IB uotinli}C - uOllounilevil SOAO qA-IA-i3KL J

3 3

13 N

oooz/lo/zt

,I,

/

gT Z-P

[

's] aan!,.[

001t 06 09 OL 09 0

OV 06 0

0L 0

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

2..................

i..........

I.........

9 X -- --------

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

s..

... ° ° z 000LD

-0 I............

1------...... oo r...........

o tz OWIj -SA ainss8Jd iazuflss@Jd loiluoo lenuelVl - JOMOd le uo!inl!a] - uoQlounlllVl SOAO

oooz/To/T£-zl1 91

-AaHj-)Jp

-i~t

[s] OWIJL 00 0 06 03 OL 09 09 OV 06 O0

01.

0 S.

ii -...........

i..........

i........o o I

I....................

1 i..........

0...

t ti..........

- CD Q~~9 CD 7 -"....."........

t..........

9 1CD

0..

8.

lunUelNq - OMOd 18 uoilnl!Cl - uollounilel~l SOAO q -

-ýt-- E LIE CVCS Malfunction - Dilution at Power - Manual Control Heat Flux vs. Time 20 30 40 50 60 70 80 90 Time [s]

Figure 14.1.4-4 Rev.

16 12/01/2000 1.12 1.00 0.88 0.75 0.62 03 E

0 r-0 X

r4 i:3 4

o o,

0.50 0.38 0.25 0.12 0.00 10

9 1

&aa

-'*l lI.

S[s]

owl1I 0

0608 OL09 O

Ot 06 OZ 0 L 0

0.

.... u..., ---.......

.......... o o

"ii..........

............ O L 'O

'DtCf L. IjSNCGV4 Wflwiuiv4

e

~ *e

-O

~ H

÷........ *..........

9 'I O '::

.0.....0 o

~owlJ~l "SA) i3NCO unwiu!iVl jOiluoo lenue*q - Jamod Ie uo!lnl!Ci - uopounlleNh SOAO

14.1.5 STARTUP OF AN INACTIVE REACTOR COOLANT LOOP Accident Description Operation of the plant with an inactive loop causes reversed flow through the inactive loop because there are no isolation valves or check valves in the reactor coolant loops.

If the reactor is operated at power in this condition there is reverse flow through the inactive loop due to the pressure difference across the reactor vessel. The cold leg temperature in the inactive loop is identical to the cold leg temperatures of the active loop and to the reactor core inlet temperature. If the reactor is operated at power, there is a temperature drop across the steam generator in the inactive loop, and with the reverse flow, the hot leg temperature of the inactive loop is lower than the reactor core inlet temperature.

The protection system prohibits continuous operation of the plant above approximately 10%

with one inactive loop. The starting of the idle reactor coolant pump results in the injection of cold water into the core and this causes a rapid reactivity and power increase. However, for power on the order of 10%, the hot leg temperature of the inactive loop is close to the core inlet thus limitýin the severity of the resulting transient.

"-sumptit-iis anU Method of Analysis The following assumptions are made:

a. Following the start of the idle pump, the inactive loop flow accelerates linearly to its nominal full flow value over a period of 10 seconds.

b. A conservative negative moderator coefficient of-4.0E1-4 Ak/°F is assumed.

c. A conservative low Doppler temperature coefficient of-I.01E-5 Ak/°F is assumed.

d. The reactor is assumed to be initially at 12% of 1650 MWt with the secondary side of both steam generators at the same pressure and with reverse reactor coolant flow through the idle loop steam generator. The 12% includes 2% allowance for calibration and instrument errors. The high initial power assumed is conservative since it gives the greatest temperature difference between the core inlet temperature and the inactive loop hot leg temperature.

e. The initial Reactor Coolant System average temperature in the active loops is 4*F above the programmed value for 12% power. This is a conservatively high value for the initial average temperature including instrument errors and results in the minimum margin to core DNB limits.

f. The initial Reactor Coolant System pressure is M psi below nominal.

This is a conservatively low value for the initial pressure including instrument errors and results in the minimum margin to core DNB limits.

Rev. 16 14.1-12 12/01/2000

USAR Insert 14.1.5-1 The Kewaunee Nuclear Power Plant Technical Specifications require that both reactor coolant pumps (RCPs) be operating when the reactor is in the OPERATING mode. One pump operation is not permitted except for tests. In the event that one RCP trips with the power being less than 10% of full power, the Technical Specifications require that the core power be reduced to a level below the maximum power determined for zero power testing. If an RCP trips above 10% power, an automatic reactor trip will be initiated. The maximum, initial core power level for the startup of an inactive reactor coolant loop is limited to less than 2%. Under these conditions, there can be no significant reactivity insertion because the reactor coolant system is initially at a nearly uniform temperature. Based on this, an analysis of this event was determined not to be necessary. The discussion presented below corresponds to an analysis previously performed assuming an initial power level of 12% of 1650 MWt and is retained for historical purposes.

USAR Insert 14.1.5-2 The Kewaunee Nuclear Power Plant Technical Specifications require that both reactor coolant pumps (RCPs) be operating when the reactor is in the OPERATING mode. Based on the.

Technical Specification requirements, an analysis of this event is not required to show that the DNB design basis is satisfied.

A detailed digital simulation of the plant, including heat transfer to the steam generators of the active and inactive loop, and reactor coolant flow transit times, was used to study the transient following pump startup in the inactive loop.

Results The results following the startup of an idle loop with the assumptions listed above are shown in Figures 14.1.5-1 through 14.1.5-5.

The heat flux response, of interest for DNB considerations, indicates that the peak heat flux reaches a value that is less than the nominal full power value. This low heat flux combined with a high degree of sub-cooling in the core at all times results in no adverse effects to the core by the transient. No reactor trip occurs.

It is expected that the actual transient effects would be less severe than those shown because of alleviating factors, which have not been taken into account. For example, the actual starting time of the Reactor Coolant Pump is likely to be about 20 seconds rather than the 10 seconds assumed in the analysis. This means that the change in core temperature would occur more gradually than shown in the figures. Furthermore, the water entering the core is assumed to exhibit the temperature of the water in the inactive loop, providing the analysis with a high degree of conservatism.

The average temperature of the reactor coolant water increases because of the positive reactivity insertion and power increase brought about by the entry into the core of the cold water in the inactive loop. This leads to an increase in pressurizer pressure. 'The maximum pressure reached is well below the acceptance criteria of 2750 psia.

The following table shows the comparison of the important calculated safety parameters to their respective acceptance criteria (Calculated Valve/Acceptance Criterion):

MDNBR RCS Pressure MSS Pressure Startup of Inactive Loop 5.878/E 2313/2750 1153/1210 Conclusions The results show that for startup of an inactive loop, the power and the temperature excursions are not severe. There is a considerable margin to the limiting MDNBR. Therefore, no undue restriction needs to be placed on the plant when starting a reactor coolant pump at power levels to 12% ower.

14.1.6 EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTIONS Accident Description Reductions in feedwater temperature or additions of excessive feedwater are means of increasing core power above full power. Such transients are attenuated by the thermal capacity Rev. 16 14.1-13 12/01/2000

of the secondary system and of the Reactor Coolant System. The Reactor Protection System trip functions prevent any power increase that could lead to a DNBR less than the MDNBR limit.

An extreme example of excessive heat removal from the Reactor Coolant System is the transient'associated with the accidental opening of the feedwater bypass valve, which diverts flow around the low-pressure feedwater heaters. The function of this valve is to maintain net positive suction head on the main feedwater pump in the event that the heater drain pump flow is lost, e.g., following a large load decrease.

In the event of an accidental opening, there is a sudden reduction in feedwater inlet temperature to the steam generators. This increased sub-cooling would create a greater load demand on the Reactor Coolant System due to the increased heat transfer in the steam generator.

Another example of excessive heat removal from the Reactor Coolant System is a common mode failure in the feedwater control system, which leads to the accidental opening of the feedwater regulating valves (FW-7A and FW-7B) to both steam generators (see Reference 2).

FW-7A and FW-7B could fail open due to a high output signal to the feedwater control system from any one of the following components:

+ PT-485 First Stage Turbine Pressure Transmitter

+ PM-485A MI Converter

+ LM-463F Steam Generator Level Auto Programmer Mode Controller

+ LM-4Q3H Steam Generator Level Program Median Selector

+ LM-463D Current Source for Steam Generator Level Minimum Setpoint

+ LM-463C Lead/Lag Circuit This results in the valves stepping open 20% from their current position followed by a 20%

step open every 5 minutes after that until full open.

Accidental opening of the feedwater regulating valves results in an increase of feedwater flow to both steam generators, causing excessive heat removal from the reactor coolant system. The resultant decrease in the average temperature of the core causes an increase in core power due to moderator and control system feedback.

Continuous addition of cold feedwater after a reactor trip is prevented since the reduction of Reactor Coolant System temperature, pressure, and pressurizer level leads to the actuation of safety injection on low pressurizer pressure. The safety injection signal trips the main feedwater pumps, closes the feedwater pump discharge valves, and closes the main feedwater control valves.

Rev. 16 14.1-14 12/01/2000

USAR Insert 14.1.6-1 With the plant at no-load conditions, the addition of cold feedwater may cause a decrease in RCS temperature and thus a reactivity insertion due to the effects of the negative moderator temperature coefficient. However, the rate of energy change is reduced as load and feedwater flow decrease, so that the transient is less severe than the full power case.

The net effect on the RCS due to a reduction in feedwater temperature is similar to the effect of increasing secondary steam flow, i.e.,-the reactor will reach a new equilibrium condition at a power level corresponding to the new steam generator AT.

The protection available to mitigate the consequences of a decrease in feedwater temperature is the same as that for an excessive load increase.

USAR Insert 14.1.6-2

j c

TMc' -o The reduction in feedwater temperature is determined by computing conditions at the feedwater pump inlet following the opening of the heater bypass valve. These feedwater conditions are then used to recalculate a heat balance through the high pressure heaters. This heat balance gives the new feedwater conditions at the steam generator inlet.

The following assumptions are made:

A. Initial power level of 1780 MWt.

B. Low pressure heater bypass valve opens, resulting in condensate flow splitting between the bypass line and the low pressure heaters; the flow through each path is proportional to the pressure drops.

An evaluation method was applied that demonstrates the decreased enthalpy caused by the feedwater temperature reduction is bounded by an equivalent enthalpy reduction that results from an excessive load increase incident (Section 14.1.7).

USAR Insert 14.1.6-3 The opening of a low pressure heater bypass valve causes a reduction in feedwater temperature which increases the thermal load on the primary system. The reduction in feedwater temperature is less than 33°F resulting in an increase in heat load on the primary system of less than 10% of full power. The reduction in feedwater temperature due to a 10%

step load increase is greater than 33 0F. The increased thermal load, due to the opening of the low pressure heater bypass valve, thus results in a transient very similar, but of reduced magnitude, to the 10% step load increase incident described in Section 14.1.7. Therefore, the transient results are not presented.

USAR Insert 14.1.6-4 With respect to the feedwater temperature reduction transient (accidental opening of the feedwater bypass valve), it was determined to be less severe than the excessive load increase incident (see Section 14.1.7). Based on results presented in Section 14.1.7, the applicable acceptance criteria for the feedwater temperature reduction transient have been met.

Accidental Opening of the Feedwater Bypass Valve Method of Analysis caa1bilty Th se.cn1 d case is show ate tora ns nt automatic ith oervtivelnt lo a

negative moderator reactivity coefficient r

b n

y.

Initial pressurizer pressure, re coolant average tea prn es and reactor power are consistent with steady full power operation, allowing for -a eion and instrument errors. This re in the minimum margin to core DNB at the start of sient. The

}ansalys incrfaeiormed usin as dtaied digia l simultioo thruhte platea incl erator.-h~s Results RETPLA~CE WiTt-+ LJSA9 ;1WStaRr 141.

es 14.1.6-1 through 14.1.6-5 show the transient without automatic reactor control with a o moderator reactivity coefficient representing beginning of cycle cond o ns. As expected, t verage reactor coolant temperature and pressurizer pres show rapid decreases as the sec ab eat extraction remains greater than the core er generation. The core power level increas lowly and eventually comes to e

" r ium. at a value slightly above the nominal full ower

e. There is an incre margn to DNB because of the accompanying reduction in coolant a e temper

e. The reactor does not trip. There is a small increase in core AT as the heat trans creases through the steam generator.

h ic vity insertio rate at no-load from andexcessive feedwatea flowl icrease also analyze he following assumptions:

1. A step increase in feedwater flow steam gen omi zero to the nominal full-load flow.

2. The mostne ati e ivty moderator coefficient at end-of-li e.

.v 3,.---6"~stant feedwater temperature of 70°F Rev. 16 14.1-15 12/01/2000

Accidental Opening of Feedwater Regulating Valves Method of Anal sis TnFhe fo owing assuwmp ions are made for the analysis of for a feedwater malfinction event invlvn te cc ntal co en*

o the dwater regulating valves:

--PL~c--E wi-_ ush

_4 jj~nsev-t Mi4.i6 Thep ntisonram-I ower allowin2 for instrument and calibration uncertaipt-ire feedwater control system in automatic mode. The safety anal ogivejte highest feedwater flow rate.

The feedwater in headers B is at a consistent with no=

=ant conditions.

4) Feed er flow increases 50% in both loops from 3.6 to 5.4 MELBM/HR.-Tjis o-nservative because pump runout flow is 5.0 MILBM/HR.

The analysis is performed using a detailed digital simulation of the plant including core kinetics, Reactor Coolant System and the Main Steam and Feedwater Systems.

Rev. 16 12/01/2000 14.1-16

Results Cp 1LpcE IVqT$*

LJSoe nsTrt 1. I. -7 R-Reaetor-power-............

slightl abe',,e-th-............pcc v

a............. -c r~tc 4

LUUIdUV fj. wlIULL+/--. i.-

2 db h ccsiv dwatcrflown Pobth aem IX=atw Av I-resultj-MInIrntImu DNBR deereases slightly but is well above the limi~ting Tm~iirnt=u DNIMR Conclusions 4L.

L Feedwater system malfunction transients involving a reduction in feedwater temperature or an

=~

increase in feedwater flow rate have been analyzed. The analyses show an increase in reactor e

power from the reactor temperature reduction due to the excessive heat removal in the steam g generators. -,,

mo.mt limitint is 4-h k 4-on

-1.

1 -eanifins:Analyses demonstrate "5

that considerable margin to the safety anaysis acceptance criteria, (MDNBR, primary and secondary pressure), exists throughout the transient. Therefore, there is no radioactive release or public hazard in the event of a feedwater malfunction event.

14.1.7 EXCESSIVE LOAD INCREASE INCIDENT fe ore'vVOL Accident Description An excessive load increase incident is defined as a rapid increase in steam generator steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. The Reactor Control System is designed to accommodate a 10% step load increase or a 5% per minute ramp load increase (without a reactor trip) in the range of 15 to 95% of full power. Any loading rate in excess of these values may cause a reactor trip actuated by the Reactor Protection System. If the load increase exceeds the capability of the Reactor Control System; the transient is terminated in sufficient time to prevent the DNBR from being reduced below the MDNBR limit. An excessive load increase incident could result from either an administrative violation such as excessive loading by the operator or an equipment malfunction in the steam dump control or turbine speed control.

For excessive loading by the operator or by system demand, the turbine load limiter keeps maximum turbine load from exceeding 100% rated load.

During power operation, steam dump to the condenser is controlled by reactor coolant condition signals; i.e., high reactor coolant temperature indicates a need for steam dump. A single controller malfunction does not cause steam dump; an interlock is provided which blocks the opening of the valves unless a large turbine load decrease or a turbine trip has occurred.

Load increases caused by a hypothetical steam-line break are analyzed in Section 14.2.5.

Rev. 16 14.1-17 12/01/2000

USAR Insert 14.1.6-5 The system is analyzed to demonstrate acceptable consequences in the event of an excessive feedwater addition, due to a control system malfunction or operator error which allows all feedwater control valves to open fully. The following cases have been analyzed:

1 a.

Accidental full opening of all feedwater control valves with the reactor at full power assuming manual rod control and a conservatively large negative moderator temperature coefficient of reactivity.

lb.

Accidental full opening of all feedwater control valves with the reactor at full power assuming automatic rod control and a conservatively large negative moderator temperature coefficient of reactivity.

2.

Accidental opening of a feedwater control valve with the reactor at no load (hot zero power) conditions and assuming a conservatively large negative moderator temperature coefficient of reactivity with minimum available shutdown margin.

This accident is analyzed using the Revised Thermal Design Procedure (RTDP) methodology.

USAR Insert 14.1.6-6

1)

Initial reactor power, pressure, and RCS temperatures are assumed to be at their conservative nominal values. Uncertainties in initial conditions are included in the MDNBR limit.

2)

Feedwater control valves are assumed to malfunction resulting in a step increase from 100% to 150% of nominal feedwater flow delivering flow to both steam generators.

3)

For the feedwater control valve accident at full power conditions (cases 1 a and I b), a feedwater temperature of 437.1°F is assumed, consistent with nominal plant conditions.

4)

For the feedwater control valve accident at no load conditions (case 2), a feedwater temperature of 198.0°F is assumed, consistent with no load plant conditions.

5)

No credit is taken for the heat capacity of the RCS and steam generator thick metal in attenuating the resulting plant cooldown.

6)

The feedwater flow resulting from a fully open control valve is terminated by the steam generator high-high water level signal that closes the associated feedwater main control and feedwater control-bypass valves, indirectly closes all feedwater pump discharge valves, and trips the main feedwater pumps and turbine generator.

Normal reactor control system and engineered safety systems are not required to function. The reactor protection system may function to trip the reactor due to power-range high neutron flux, overpower or turbine trip on high-high steam generator water level conditions.

USAR Insert 14.1.6-7 The feedwater flow malfunction at hot zero power conditions result in an increased nominal heat flux and reduced RCS pressure due to the reactor cooldown, which is caused by the excessive feedwater flow to both steam generators. The results of the DNB analysis yielded a minimum DNBR above the safety analysis limit, however it was found that this case is bounded by the excessive feedwater flow cases analyzed at full power initial conditions.

The most limiting case is the excessive feedwater flow from a full power initial condition with automatic rod control. This case gives the largest reactivity feedback and results in the greatest power increase. A turbine trip, which results in a reactor trip, is actuated when the steam generator water level in either steam generator reaches the high-high water level setpoint. Assuming the reactor to be in manual rod control results in a slightly less severe transient. The rod control system is not required to function for this event; however assuming that the rod control system is operable, yields a slightly more limiting transient.

For each excessive feedwater flow case analyzed, continuous addition of cold feedwater is prevented by automatic closure of the associated feedwater control valves, closure of all feedwater bypass valves, a trip of the feedwater pumps, and a turbine trip on high-high steam generator water level. In addition, the feedwater discharge isolation valves will automatically close upon receipt of the feedwater pump trip signal.

Transient results, Figures 14.1.6-1 through 14.1.6-10, show the reactor power, pressurizer pressure, core average temperature, vessel inlet and outlet temperature and minimum DNB conditions throughout the transient for the full power cases (while in manual and automatic rod control). Though the reactor power increases slightly above the nominal full power value during the transient, the DNBR does not drop below the safety analysis limit value.

RELpAC c -

v I

i4, ld-( (L RLe 4L-J06 PAG* )*

xcessive Heat Removal - Feedwater System Malfunction - BOC Manual Control Reactor Power vs. Time 050 1 0 4 0............

1.045 1.040....---

.~.....

CO 1.035.

E 0 ~~~~.

o 1.030 C:

ciii

.o 1.0 20 0"

15

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

1. 0 10

_L "0

30 60 90 120 150 180 210 240 270 300 Time [s]

Figure 14.1.6-1 Rev.

16 12/01/2000

Excessive Heat Removal -

Feedwater System Malfunction Manual Control Reactor Power vs. Time 1.2

-E E

0 t'--

t---------

.2

- I B

B_

I I_

I B

0.

In 0

Time [s]

Figure 14.1.6-1

oooz.fo/ ziZ-9"'f o-,-*!

91 -. allZ9TVT3I2' S[s]

aW~l_

\\Z3 OLt7 0LZ 09L 091 OL 06 09 OC 0

I

"------I-----

O'OL L[

O'SL 2 0098L 0"S8L2 0"061.

"-u "CI)

C Cn cin 0"96L*

0*0LFI OWIl -SA eJnSSeJd jgzijnsseJci Ioiluoo lenuelhN 008 - uoflounlleBA waIsAS JaleMPaaj - i1AOWo8 12a"H aA!SSa3x3

Excessive Heat Removal -

Feedwater System Malfunction Manual Control Pressurizer Pressure vs. Time 2300 2200 - -- - - - - - - -- - - - - - -----

-- -- -- -I-- - - --

,, 2100 - - - - - - - - - - - - - - - - - -

I Q)

V')

I I

I 1900 I

I 1800 Time [s]

Figure 14.1.6-2

T'ý'IPtAC-C WArt-rOF(L

/-I -lI b &-3 (c, Ij 1 J;,Lc,&AIRACCE Excessive Heat Removal - Feedwater System Malfunction - BOC Manual Control Tave vs. Time 570.5 570.0 569.E 569.*

573.5 573.0 0

30 60 90 120 150 180 210 240 270 00 Time [s]

Figure 14.1.6-3 Rev. 16 12/01/2000 572.5 572.0 571.5 571.0

0) a)

CL.

E (D

Excessive Heat Removal -

Feedwater System Malfunction -

Manual Control Core Average Temperature vs. Time 580 570 --

560 -+

E*

- 540 -+

5 II 53I I

+I-I 520 Time [s]

Figure 14.1.6-3

~oooU/to/ztl

,,1

,au 17-9" ['1V oa-nýIzl

- ,"[s]

aw~l 00\\C6 L3 OVZ O

0O 0801 00 00.9

-~~~~~

............------I------- ----------- ---------

g9

---I CD 0CD0

'L9 u~

j 008-----

walAS J8BM 8 ---------------- o 8

0 "8 9 7

OI I'AGO

-U1U joiluoo

°eu~

01onie4wlA O'mPO BO@j I@

AS8x 1

( -1 0...

..... :3 v-

- I.

.. -:ý..

f,"

- )

ý o6

Excessive Heat Removal -

Feedwater System Malfunction -

Manual Control Vessel Inlet and Outlet Temperature vs. Time Vessel out leI Vessel inlet 620 0

0

+

CL E 560 --- - - - - - - - - - - - - - - - - - -

S 540 520 I

II 540-"-

I I

I1 I

I I

20 I

I I

1 Time [s]

Figure 14.1.6-4

Excessive Heat Removal - Feedwater System Malfunction - BOC Manual Control Minimum DNBR vs. Time 2.00 2.7 5 2.50 2.25 2.00 1.75 1.50 1.25 1.00 0.75 0.50 0.25 0.0

-T MDNBR " t= 1.4

--I.....

........... I --------

"Mihimunh MDN

=1.681.

i.

.um S.;...........

p 0

30 60 90 120 150 180 210 240 270 300 Time [s]

Figure 14.1.6-5 Rev. 16 12/01/2000 Cc 1=3 E

E-

Excessive Heat Removal -

Feedwater System Malfunction -

Manual Control Minimum DNBR vs. Time 5

4 I

M 3.5 - -- - - - - -- - - - - -- - - - - - -- - - - - -- - -

E a

E 2-I 2.5 -----------

+

2-I 1.5 V

Time [s]

Figure 14.1.6-5

Excessive Heat Removal - Feedwater System Malfunction - EOC Auto Control Reactor Power vs. Time 1.50 1.045 1.040 1.035 1.030 1.025 1.020 1.015 S......................

0 20 40 60 80 100 120 140 160 Time [s]

Figure 14.1.6-6

[2 180 zooU Rev.

16

/01/2000 C-0 0

0 L._

(I 0

1.010

A

Excessive Heat Removal ' Feedwater System Malfunction Auto Control Reactor Power vs. Time 1.2 E

I aI r,.)B I

I II I

I.

I2 I

I I

I B

I oI B

I I

I Time [s]

Figure 14.1.6-6

9 L -9 *1 i17 3.n !d,

[s] OewIl 00 08L 09t 01t4 0 L 001L 08 09 OV 03 0

a......

-N

/----

OW!J SA JnSSiJd Jaz~urzýGSSJd IoJluoo olnv 003 - Uo013unllhp" WuSeISIS ja1eMpa@Oj - IeAOWOU leaH 0"00S3 0"£0g8gx

(-?

+/- C) ryrrQ) j L.-

!t7I 0"9 2 O'0Lt 2 O'SLt 2 0'08[3 0'06LI 7O CD C=

CD 70 Gn.

3 -?j L-'*"7 I'

Excessive Heat 2300 2200 Cf)

3 Q)

I(I)

(I)

L.

[3-2100 2000 1900 1800 Removal -

Feedwater System Malfunction -

Auto Control Pressurizer Pressure vs. Time 180 Time [s]

Figure 14.1.6-7

Excessive Heat Removal - Feedwater System Malfunction - EOC Auto Control Tave vs. Time

7.................................

572.0 I -------

0 571.5 -------

5 1 -.................

572.

v -

1..

E 5 7 0.0 o-Figure 14.1.6-8

r.

12/01/2000

Excessive Heat Removal -

Feedwater System Malfunction -

Auto Control Core Average Temperature vs. Time 580 570 -- - - - - - - - - - - - - - - - -

560 -- - - - - --

Lj S

550 -*

=3 (1)

E 540 - - - - - - - - - - - - - -

530 - -

--f-520 Time [s]

Figure 14.1.6-8

I'-,?CPLACC wn, F -0L L-0 L A)

IJC,

P A & )-

Excessive Heat Removal - Feedwater System Malfunction - EOC Auto Control Delta T Core vs. Time 69.5 69.0 68.5 68.0 67.5 67.0 66.5 66.0 65.5 65.

70.0 LL t0) 0 I

CL E

U)

I--

.......... i.................*......*.....

./

.t.............

0 20 40 60 80 100 120 140 160 18 200 Time [s]

Figure 14.1.6-9

,2Rev.o16

14. 1. ý-* 1 6,

Excessive Heat Removal -

Feedwater System Malfunction -

Auto Control Vessel Inlet and Outlet Temperature vs. Time Vessel out let Vessel inlet 620 500

+--------------------

I a

Ia L-54 0 a

aI Sa I

I aIa 54I a-rI-520 180 Time [s]

Figure 14.1.6-9

"cessive Heat Removal - Feedwater System Malfunction - EOC Auto Control Minimum DNBR vs. Time 3.0 2.75 2.50 ----------..

1.............

2.00 1.75 E

1.2 5 ----------------

HTPMDNBRLi t 1.14

1........................................... -

0.75 Mihimuni MDN 1.647 0

20 40 60 80 100 120 140 160 180 200 Time [s]

Figure 14.1.6-10 Rev. 16 14.16-1012/01/2000

Excessive Heat Removal -

Feedwater System Malfunction -

Auto Control Minimum DNBR vs. Time 5

4.5-4 E

.*3 2I-- -

- I - -

2 2I+

+-

1.5 180 Time [s]

Figure 14.1.6-10

Method of Analysis Four cases are analyzed to demonstrate the plant behavior for a 20% step increase from rated load. The first two cases are for a manually controlled reactor at beginning of cycle (BOC, ccm = zero AMF) and end of cycle (EOC, ctm = -4.OE-4 Ak/F) conditions (a, is the moderator reactivity co-efficient). Beginning of cycle represents a condition when the plant has the smallest moderator temperature coefficient of reactivity and, therefore, the least inherent transient capability. Two cases are analyzed for an automatic control situation at BOC and EOC conditions with control rods initially inserted to the power dependent insertion limits. A conservative limit on the turbine valve opening was assumed corresponding to 1.2 times nominal steam flow at nominal steam pressure. Initial pressurizer pressure, reactor coolant average temperature and power are assumed at extreme values consistent with steady state, full-power operation, allowing for calibration and instrument errors. This results in the minimum margin to core DNB at the start of the transient. The analyses are performed using a detailed digital simulation of the plant including core kinetics, Reactor Coolant System, and the Steam and Feedwater Systems.

Results Figures 14.1.7-1 through 14.1.7-8 illustrate the transient with the reactor in the manual control mode. As expected, for the BOC case with a very slight power increase, the core average temperature shows a large decrease. For the EOC case, there is a much larger increase in reactor power due to the moderator feedback. Both of the manual control cases demonstrate adequate #:DNBR margin.

Figures 14.1.7-9 through 14.1.7-18 illustrate the transient assuming the reactor is in automatic control. In automatic control the reactor power transient is greater than for the corresponding case in manual control. The automatic control cases still show adequate margin to the

DNBR limit.

The following table shows the comparison of the important calculated safety parameters to their respective acceptance criteria (Calculated Value/Acceptance Criterion):

Excessive Load Increase MDNBR RCS Pressure MSS Pressure (usia')

(sial BOC Manual Control 1.681/E 2200/2750 751/1210 BOC Auto Control 1.430/Z 2200/2750 751/1210 EOC Manual Control 1.478/E 220012750 751/1210 EOC Auto Control 1.438/M 2200/2750 751/1210 Conclusions

(

5/

[

The four cases analyzed show a considerable margin to the limiting MDNBR. It is concluded that reactor integrity is maintained throughout lifetime for the excessive load increase incident.

Rev. 16 14.1-18 12/01/2000

USAR Insert 14.1.7-1 Based on historical precedence, this event does not lead to a serious challenge to the acceptance criteria and a reactor trip is not typically generated. As such, it has been determined that a detailed reanalysis of this event is not necessary to support a core power rating of 1772 MWt. A simplified statepoint evaluation, assuming a 10% step load increase, was performed and the results confirmed that core DNB limits are not challenged following this event. The discussion presented below corresponds to the analysis previously performed for this event and is retained for historical purposes.

USAR Insert 14.1.7-2 Furthermore, the results of a simplified statepoint evaluation performed for a 10% step load increase with a nominal core power of 1772 MWt confirm that the core thermal limit lines are not challenged, and that the minimum DNBR during this transient will remain above the safety analysis limit value.

14.1.8 LOSS OF REACTOR COOLANT FLOW Accident Description A loss-of-coolant flow incident can result from a mechanic;,l or electrical failure in one or more reactor coolant pumps (RXCPs), or from a fault in the power supply to these pumps. If the reactor is at power at the time of the incident, the immediate effect of loss-of-coolant flow is a rapid increase in coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor is not tripped promptly. The following trip circuits provide the necessary protection against a loss-of-coolant flow incident:

+ Low voltage on pump power supply bus

+ Pump circuit breaker opening (low frequency on pump power supply bus opens pump circuit breaker)

+ Low reactor coolant flow These trip circuits and their redundancy are further described in Section 7.2, Reactor Control and Protection System.

Simultaneous loss of electrical power to all RXCPs at full power is the most severe credible loss-of-coolant flow condition. For this condition, reactor trip togetherwith flow sustained by the inertia of the coolant and rotating pump parts will be sufficient to prevent fuel failure, Reactor Coolant System overpressure, and prevent the DNBR from going below its limit.

P'LR(e tJ rOT4 Two types of flow coastdown accidents were alyzed, loss of two RXCPs at nominal l~i*-T" frequency and loss of two RXCPs at low fr uency. These two types of flow coastdown

-,.1analyses are described separately under Loss f Reactor Coolant Flow-Nominal Frequency and under Loss of Reactor Coolant Flow Lo requency.

Loss of Reactor Coolant Flow-Nominal Frequency Method of Analysis JJO)/o i

772 The following nominal frequency loss of coolant flow ase is analzed: Loss of two pumps from a Reactor Coolant System heat output of of if MWt with two loops operating. This case represents the worst credible coolant flow loss.

The normal power supplies for the pumps are the two buses connected to the generator, each of which supplies power to one of the two pumps. When a generator trip occurs, the pumps are automatically transferred to a bus supplied from external power lines. Therefore, the simultaneous loss of power to both reactor coolant pumps is a highly unlikely event.

Following any turbine trip, when there are no electrical faults requiring tripping the generator from the grid, the generator remains connected to the grid for at least thirty seconds. Since both pumps are not on the same bus, a single bus fault does not result in the loss of both pumps.

Rev. 16 14.1-19 12/01/2000

j?PLv

tJt1lA fwe,ZT iq,1 18 -

F Rev. 16 12/01/2000 14.1-20 A full plant simulation is used in the analysis to compute he core average and hot spot heat flux transient responses, including flow coastdown, t erature, reactivity, and control rod insertion effects.

These data are then used in a detailed thermal hydra ic computation to compute the margin to DNB. This computation solves the continuity, mentum, and energy equations of fluid flow together with the DNB correlation.

The ollowing assumptions are made in the calculations:

a. The initial operating conditions, which are ed to be most adverse with respect to the margin to DNB, are maximum steady-stat power level, minimum steady-state pressure, and maximum steady-state inlet tempera

e.

b. The largest negative initial value of th oppler coefficient (-2.32E-5 Ak/0F) and a zero moderator coefficient (0.0 Ak/°F) are sumed since these result in the maximum heat flux during the initial part of the transien, when the minimum DNB ratio is reached.

c. A reactor trip is actuated by low fl w. The time from the initiation of low-flow signal to initiation of RCCA motion is 0.6 onds. The trip signal is assumed to be initiated at 87%

of full-loop flow, allowing at le t 3% for flow instrumentation errors.

Upon reactor trip, it is also sumed that the most reactive RCCA is stuck in its fully withdrawn position, hence r uting in a minimum insertion of negative reactivity.

d. The overall heat transfer etween the fuel and the water varies considerably during the transient mostly as a res It of the change of fuel gap conductance. A conservatively evaluated overall heat tr sfer coefficient is used in the analysis.

Results r

Reactor coolant flow coastdown curve is shown in Fi ure 14.1.8-1. Reactor coolant flow is calculated based on a momentum balance in the R ctor Coolant System combined with a pump momentum balance.

The following table shows the comparison of e important calculated safety parameters to their respective acceptance criteria (Calculat Value/Acceptance Criterion):

Loss of Flow MIDNBR RCS Pressure MSS Pressure (psia')

(psial 2/2 Pump Trip I

/1.14 M

2750 W1210 Figures 14.1.8-2 and 14.1.8-3 show he nuclear power and the average heat flux response for the two-pump loss of flow. Figu 14.1.8-4 shows the MI)NBR as a function of time.

r '71EPLA(E UJ I

,ULhderlnu)c'f cj Loss of Coolant Flow -

Fr.1 e4 cy Method of Analysis The underfiequency event is analyzed using systems analysis that calculates the loop and core flow, nuclear power, and primary sys em pressure and temperature.transients. The M[DNBR is calculated by performing a de iled fuel thermal hydraulic simulation using as transient forcing functions the core heat

, core flow, core inlet temperature, and Reactor Coolant System pressure from the syst s analysis.

f2_'-Cf*-

/a.

The initial operating conditions, w ch are assumed to be most adverse with respect to the margin to DNB, are maximum sti dy-state power level, minimum steady-state pressure, and maximum steady-state aver e temperature.

b. A conservatively large absolu value of the Doppler only power co-efficient and a zero moderator coefficient (0.0 TF) are assumed since these result in the maximum hot channel heat flux during the itial part of the transient, when the MDNBR is reached.

c. A constant frequency decay rate of 5 Hz/sec is assumed. Reference 3 determined that this is the maximum credible frequency decay rate that could occur on a typical electrical grid. Analysis of the Wisconsin-Upper Michigan transmission system indicates that the worst-case frequency decay rate is [3E] approximately 2

Hz/sec (see Reference 4). Therefore, 5 Hzlsec is a very conservative decay rate. In addition, the assumption of a constant rate is conservative, since Reference 3 also shows that the expected grid frequency decay rate actually decreases during the transient.

Prior to the opening of the RXCP breaker, the RXCP speed is assumed to be directly proportional to the power supply frequency. As discussed in Reference 5, this is a conservative assumption, since the speed coasidown will lag the frequency coastdown due to the effects of pump inertia and induction motor slip. During steady state operation the pump motor speed is below the synchronous speed because of induction motor slip. After the frequency decay starts, the deceleration of the pump-motor-flywheel combination provides a positive driving torque to the pump so that the required electrical torque decreases. The reduction in electrical torque reduces the induction motor slip, thus resulting in a higher speed than that assumed in the analysis. The degree of conservatism varies directly with the assumed decay rate because the inertia torque increases directly with the decay rate. At 5 Hzlsec the expected speed is approximately 1.2% higher than the analysis value.

Reactor Coolant System flow is calculated based on a momentumbalance in the Reactor

?*PL/AC-_

Coolant System combined with a pump momentum balance.

W'*'Y-* *td. No credit is taken for the RXCP trip on underfrequency.

r45 P_ P-T

t. I,

- 5

e. Upon reactor trip, it is assumed that the most reactive RCCA is stuck at its fully withdrawn position, resulting in a minimum insertion of negative reactivity.

Rev. 16 14.1-21 12/01/2000

In Reference 7, the NRC approved ue eWS loss of flow underfrequency ti methodology.

Results

-j-'

1.34517

"4*T J4.15.(,

The following table shows the comparison of th important calculated safety parameters to their respective acceptance criteria (Calculated talueAcceptance Criterion):

Loss of Flow MDNBR CS Pressure MSS Pressure (psia)

(pia)

Underfrequency Trip O

1.14 M /2750 011210 Figures 14.1.8-5 through 14.1.8-8 shows he nuclear power, average channel heat flux, core flow, and MDNBR transient responses r the underfrequency event.

MDNBR is always above the MD limit. Therefore, fuel rod integrity and safe plant shutdown are ensured by an underfre uency trip setting of 54.5 Hz.

Conclusions Since DNB does not occur, there is no cladding damage and no release of fission products into the reactor coolant. Therefore, once the fault is corrected the plant can be returned to service in the normal manner. The absence of fuel failures would, of course, be verified by analysis of reactor coolant samples. In the loss of reactor coolant flow accidents, it has been shown that there is adequate reactor coolant flow to maintain a MDNBR greater than the MDNBR limit.

Locked Rotor Accident Accident Description A transient analysis is performed for the hypothetical instantaneous seizure of a reactor coolant pump rotor. Flow through.the Reactor Coolant System is rapidly reduced, leading to a reactor trip on a low-flow signal. Following the trip, beat stored in the fuel rods continues to pass into the core coolant causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generator is reduced, first because the reduced flow results in a decreased tube side film coefficient and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the coolant in the reactor core, combined with the reduced heat transfer in the steam generator causes an insurge into the pressurizer and a pressure increase throughout the Reactor Coolant System. The insurge into the pressurizer compresses the steam volume, actuates the automatic sjray system, opens the power-operated relief valves, and opens the pressurizer safety valves, in that sequence. The two power-operated relief valves are designed for reliable operation and would be expected to function properly during the accident. However, for conservatism, their pressure-reducing effect is not included in the analysis.

Rev. 16 14.1-22 12/01/2000

Method of Analysis At the beginning of the postulated locked rotor ccident, i.e., at the time the shaft in one of the RXC-Ps seizes, the plant is assumed to beiin operation under the steady-state operating conditions that are most adverse with resp to MDNBR margin. The plant is assumed to be operating at maximum steady-state po r, minimum steady-state pressure, and maximum steady-state core inlet temperature.

After pump seizure, nuclear power is rapidly reduced because of void shutdown and the RCCA insertion upon reactor trip.

No credit is taken for the pressure reducing effect of the pressurizer relief valves, pressurizer spray, steam dump, or controlled feedwater flow after plant trip. Although these operations are expected to occur and would result in a lower peak pressure, an additional degree of conservatism is provided by ignoring their effect.

(ic-ACI The pressurizer safety valves start operating/

2500 psia and relieve steam at their rated 0 TJ7 t capacities. Additional sensitivity analyses w/e performed at pressurizer safety valve settings of +6% and -4% of the nominal setpoint t account for the effects of steam accumulation and IYNS."

selpoint drift. The critical safety p meters were shown to be acceptable under these 1-, 1,1- *,

assumptions.

Calculations of the extent of DNB in the core during the accident are performed using the heat flux, the coolant flow decay and the coolant pressure and temperature as transient forcing functions.

In order to estimate the severity of the accident in the core as far as the integrity of the fuel rods is concerned the thermal behavior of the fuel located at the hot spot after DNB was investigated. Results obtained from an analysis of this "hot spot" condition represent the upper limit with respect to clad temperature, clad melting and zirconium-steam reaction.

C PL-A(q fOrIT/

Results The coolant flow through the core is rapidly reduc to < 50% of its initial value (see Figure 14.1.8-9).

The reactor coolant pressure vs. time for a locked rotor accident is shown in Figure 14.1.8-11. The minimum DNBR fora elrod having an initial F. value of is shown in Figure 14.1.8-12. The M F

rod reaches a MDITNBR of slightly above the MDNBR limit The MDNBR for the 1.70

. fuel rod is less than the MDNBR limit, and the "fuel rod is assumed to fail. Up to 40% o the fuel rods in the core can go below the MDNBR limit with acceptable radiological cons quences (Reference 8). Fuel rod power census curves are generated for each reload to asses the percentage of fuel rods that are expected to go below the MDNBR limit of this accident Rev. 16 1

A I ')1 12/01/2000 F e, c W IT*-H j R ~r --

(

l q,l.3* -*

1 "-I'. * -,t.,J

Figure 14.1.8-13 shows the clad temperature transient t the hot spot. Since in the worst case examined, the clad temperature does not exceed I 00F, it is not necessary io consider the possibility of a zirconium-steam reaction. The z conium-steam reaction is only significant above this temperature./

The following table shows the comparison f the important calculated safety parameters to their respective acceptance criteria (Caac ted Valve/Acceptance Criterion):

% Fuel Rods Max Clad RCS Pressure MS Pressure

< DNB Limit /

Temp.

(°sia)

(psia')

Locked Rotor 2700 2750 104411210

Percentage of Fuel Rods w 2it Conclusions Since the peak pressure reached during the transient is < 110% of design, the integrity of the Reactor Coolant System is not endangered. The pressure can be considered as an upper limit because of the following conservative assumptions used in the study:

1. Credit is not taken for the negative moderator coefficient.

2. It is assumed that the pressurizer relief valves were inoperative.

3. The steam dump is assumed to be inoperative.

The peak clad temperature calculated for the hot spot, can also be considered an upper limit because of the following:

1. The hot spot is assumed to be in DNB at the start of the accident.

2. A high gap coefficient isused during the transient.

3. The nuclear heat released in the fuel at the hot spot is based on a zero moderator coefficient.

114.1.9 LOSS OFEXTERNALELECTICAL LOAD-T O

0U 5c~O141-3,9 Chaofrt Accident Description l]

C The loss of external electrical load may result from an abnormal increase in network frequency, opening of the main breaker from the generator, which causes a rapid large Nuclear Steam Supply System load reduction by the action of the turbine control, orby a trip of the turbine generator.

The plant is designed to accept a full-load rejection without actuating a reactor trip. The automatic steam dump system with 85% steam dump capacity (401/6 to the condenser and 45%

Rev. 16 l1A 1_'7A 12/01/2000

.11 --at JL dd o

Inserts for USAR Section 14.1.8 Insert 14.1.8-1 Two types of loss of flow accidents were analyzed: complete loss of flow due to the loss of two RXCPs and complete loss of flow due to a frequency decay (underfrequency). These two types of flow coastdown analyses are described separately under Loss of Reactor Coolant Flow - Nominal Frequency and under Loss of Reactor Coolant Flow - Underfrequency.

Insert 14.1.8-2 This transient is analyzed with two computer codes. First, the RETRAN computer code is used to calculate the loop and core flow during the transient, the time of reactor trip based on the calculated flows, the nuclear power transient, and the primary system pressure and temperature transients. The VIPIRE computer code is then used to calculate the heat flux and DNBR transients based on the nuclear power and RCS temperature (enthalpy), pressure, and flow from RETRAIN. The DNBR transients presented represent the minimum of the typical or thimble cell for the fuel.

This event is analyzed with the Revised Thermal Design Procedure (RTDP).

The following assumptions are made in the calculations:

a.

Consistent with the RTDP methodology, the initial operating conditions are assumed to be at their nominal values, including the steady-state power level, RCS pressure, and RCS vessel average temperature. Minimum Measured Flow (MMF) is also assumed.

b.

The largest negative value of the Doppler Power Coefficient and a zero moderator coefficient are assumed since these maximize the heat flux during the initial part of the transient, when the minimum DNBR is reached.

c.

A reactor trip is actuated by low flow. The time from the initiation of the low flow signal to initiation of RCCA motion is 0.75 seconds. The trip signal is assumed to be initiated at 86.5% of full loop flow, allowing 3.5% for flow instrumentation errors.

d Upon reactor trip, it is assumed that the most reactive RCCA is stuck in its fully withdrawn position, hence resulting in a minimum insertion of negative reactivity.

e.

No credit is taken for the reactor trip on reactor coolant pump motor breaker open due to low voltage, or the reactor trip directly on undervoltage.

Insert 14.1.8-3 The reactor coolant flow coastdown is presented in Figures 14.1.8-1 and 14.1.8-2. Reactor coolant flow is calculated based on a momentum balance in the Reactor Coolant System combined with a pump momentum balance. The nuclear power and core average heat flux transients are presented in Figures 14.1.8-3 and 14.1.8-4, and the pressurizer pressure and RCS loop temperature transients areshown in Figures 14.1.8-5 and 14.1.8-6. Finally, the hot channel heat flux and DNBR transients are presented in Figures 14.1.8-7 and 14.1.8-8.

The acceptance criteria for this event, the minimum DNBR limit and the maximum RCS pressure limit of 2750 psia, are met.

Insert 14.1.8-4 As with the complete loss of flow case, the underfrequency transient is analyzed with the RETRAN and VIPRE computer codes, as well as the RTDP methodology.

The following assumptions are made in the calculations:

a.

Consistent with the RTDP methodology, the initial operating conditions are assumed to be at their nominal values, including the steady-state power level, RCS pressure, and RCS vessel average temperature.

b.

The largest negative value of the Doppler Power Coefficient and a zero moderator coefficient are assumed since these maximize the heat flux during the initial part of the transient, when the minimum DNBR is reached.

Insert 14.1.8-5

d.

A reactor trip is actuated by low flow. The time from the initiation of the low flow signal to initiation of RCCA motion is 0.75 seconds. The trip signal is assumed to be initiated at 86.5% of full loop flow, allowing 3.5% for flow instrumentation errors. No credit is taken for the RXCP trip on underfrequency.

Insert 14.1.8-6 The reactor coolant flow coastdown is presented in Figures 14.1.8-9 and 14.1.8-10. Reactor coolant flow is calculated based on a momentum balance in the Reactor Coolant System combined with a pump momentum balance. The nuclear power and core average heat flux transients are presented in Figures 14.1.8-11 and 14.1.8-12, and the pressurizer pressure and RCS loop temperature transients are shown in Figures 14.1.8-13 and 14.1.8-14. Finally, the hot channel heat flux and DNBR transients are presented in Figures 14.1.8-15 and 14.1.8-16, respectively.

The acceptance criteria for this event, the minimum DNBR limit and the maximum RCS pressure limit of 2750 psia, are met.

Insert 14.1.8-7 The Locked Rotor transient is analyzed with three computer codes. First, the RETRAN computer code is used to calculate the loop and core flow during the transient, the time of reactor trip based on the calculated flows, the nuclear power transient, and the primary system pressure and temperature transients. The FACTRAN computer code is then used to calculate the thermal behavior of the fuel located at the core hot spot based on the nuclear power and RCS temperature (enthalpy), pressure, and flow from RETRAN. The FACTRAN computer code includes a film boiling heat transfer coefficient. Finally, the VIPRE code is used to calculate the "Rods-in-DNB" using the nuclear power and RCS flow from RETRAN.

At the beginning of the postulated RCP Locked Rotor accident, the plant is assumed to be in operation under the most adverse steady state operating conditions, i.e., a maximum steady state thermal power, maximum steady state pressure, and maximum steady state coolant average temperature. The analysis is performed to bound operation with a maximum uniform steam generator tube plugging level of 10%. However, a core flow reduction of 1.1 percent, which addresses the potential reactor coolant flow asymmetry associated with a maximum loop-to-loop steam generator tube plugging imbalance of 10 percent, was applied.

A conservatively large absolute value of the Doppler-only Power Coefficient is used, along with the most-positive moderator temperature coefficient limit for full power operation (0 pcm/IF). These assumptions maximize core power during the initial part of the transient when the peak RCS pressures and hot spot results are reached.

A conservatively low trip reactivity value (3.5% Ap) is used to minimize the effect of rod insertion following reactor trip and maximize the heat flux statepoint used in the DNBR evaluation for this event. This value is based on the assumption that the highest worth RCCA is stuck in its fully withdrawn position. A conservative trip reactivity worth versus rod position was modeled in addition to a conservative rod drop time (1.8 seconds to dashpot).

A loss of offsite power is assumed with the unaffected RCP losing power instantaneously at the time ofreactor trip.

Insert 14.1.8-8 The pressurizer safety valves start operating at 2500 psia and relieve steam at their rated capacities. A safety valve set pressure tolerance of+1% and a set pressure shift of+1% are modeled. Also, a sensitivity analysis was performed assuming a pressurizer safety valve tolerance of +6% and a set pressure shift of +1%. The critical safety parameters were shown to be acceptable under these assumptions.

Insert 14.1.8-9 Figures 14.1.8-17 through 14.1.8-25 illustrate the transient response for the Locked Rotor event. The results shown are for the peak RCS pressure/PCT case. The coolant flow through the core is rapidly reduced to less than fifty percent of its initial value (Figure 14.1.8-17). As shown in Figure 14.1.8-22, the peak RCS pressure is less than the acceptance criterion of 2750 psia. Also, Figure 14 1.8-25 shows that the peak cladding temperature is considerably less than the limit of 27000F. The zirconium-water reaction at the hot spot meets the criterion of less than 16% zirconium-water reaction. This transient trips on a low primary reactor coolant flow trip setpoint which is assumed to be 86.5% of the initial flow.

Calculations performed with the VIPRE code demonstrate that the maximum percentage of rod-in-DNB for this event is less than 50%. This calculation is based upon the RTDP methodology and utilizes a generic rod census curve.

Loss of Reactor Coolant Flow - Two Pump Trip Core Flow vs. Time 75.0 67.5 60.0 52.5 45.0 37.5 30.0 22.5 15.0 7.5 2

3 4

5 6

Time [s]

0 1

Figure 14.1.8-1 Rev.

16 12/01/2000 Ii

............~

~~~.....

~~~~~-

iiiiii iii..............

L..Q

.0 U_

J 7

8 9

1 u A

Complete Loss of Flow - Two Pumps Coasting Down (CLOF)

Total Core Inlet Flow vs. Time 400

"-280 L j 0

1I0!

I f

f I

I I

0 2

0 -. -.

-12 Time [seconds]

Figure 14.1.8-1

Loss of Reactor Coolant Flow - Two Pump Trip Reactor Power vs. Time 1.100 1.0 0.900 0.800 1

2 3

4 5

6 Time [s]

7 8

9 10 Figure 14.1.8-2 Rev.

16 12101/2000 T........

2

............ ;.................... i........... -: ---------- ---------

i 0

0 L..

0, 0

0E.

0.700 0.600 0.500 0.400 0.300 0.200 0.000 0

-8TlL ean6j 96

[spuo0es] ewUJLl Z'L R".;

0 0

- t*z" K

ZL" 0

0 -o 0

"1ol i-3 CD

2.

0

-qi 0

-I-I 0

z I

a a a a a a

a a I

I I

I a

I I

I a

I I

I a

a a

I a

I I

I a

a a

I I

I a

I I

I a

I

(

I I

a a

I I

I

-----+-----+-----+-----4.

-+-----+-----4.-----+

-+-----+-----+-----4.----

a I

a a

a I

a a

-

a a

a I

I I

a a

I I

S4.------------------+------------------+------------------+

a a

u a

a a

I I

a a

a I

a a

a I

I a

a a

a aWu!-

SA MOld doo-j SOU (d10l) umoo BuUlseoo sdwnd OM. - MOld JO sso-1 aejidwoo ZI 96" Vt I

I

!z

Loss of Reactor Coolant Flow - Two Pump Trip Heat Flux vs. Time 1.10 0

0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.01 0

1 2

3 4

5 Time [s]

6 7

8 9

10 Figure 14.1.8-3 Rev.

16 12/01/2000

('S E

0 0

X (3.

"a:

ce) 0

-j L.,

0 a

0) 0 C.,

E 0

u-c.J

0)

I..-

[NOd].J9MOd.jDeI~loN

0)

I.

0)

U-E 0)

Loss of Reactor Coolant Flow - Two Pump Trip Minimum DNBR vs. Time 3.00 T

a.00 2.50 2.25 2.00 1.75 -

z 0

.5 C

1.50................................

1.2 5 ----------------

.HTP MDNB 1-----.. 1.05.......... "...............................

0.75.............

MihimurfI MDNBR 1.291 0.50 o.............-......

......i-......-

0.0 "0

1 2

3 4

5 6

7 Time [s]

Figure 14.1.8-4 Rev.

16 12/01/2000

c-J a

-I I- -

01

.S E

o I

CL E.

LL 6 )

0 u

.. +.

+..............I" 0

0 0

II II 0

I

~II II (flr-I I

0 o~c

--tII 0

[

n HO

[O I InjJ9HGBJA JC

oooz/fo/zi 91 -"aI S-8w11r 3-n.1J 01[

6 8

L

[S] ow£ Ll 9

£ 913 I

0 I

.- 4IIII e

o...........

ewij "SA MO._- aWo3 diijJ AouanbaJpapufl'-

MOl-I lUejOQQ Jo1OBBH Jo sso-I

-7,19j~

0-)q f-JO3f-2 0*0 O'sL

"-r 0

5:

31 0.0£ S'9L O'st; fY09 0"09 O'SL

Complete Loss of Flow-Two Pumps Coasting Down (CLOF)

Pressurizer Pressure vs. Time 2400 2340 -

2.4 1

1 4.8 7.2 Time [seconds]

9.6 Figure 14.1.8-5 I

I I

3 I

I 3

3 3

I 3

3 3

I 3

I I

I 3

I I

3 3

I I

3 I

I I

3 3

I 3

I I

3 3

3 I

------------------+-------------------------------------+------------------4-----------------

3 r

3 I

I 3

3 I

3 3

3 I

I 3

3 I

I 3

3 I

I 3

I I

I 3

3 I

3 3

3 I

I 3

I I

-+------------------4-------------------------------------4-----------------

I I

3 3

I I

I 3

3 I

I 3

3 I

3 3

3 I

I I

3 I

I 3

I I

3 I

I I

3 3

1 I

3 3

I I

3 3

I I

-+------------------+------------------+------------------------------------

3 I

I 3

3 I

I 3

3 I

3 3

I I

3 3

I I

3 I

I 3

3 I

I I

3 I

I I

3 I

I I

3 3

I 3

I

-+-------------------------------------4----------------

3 3

3 I

3 3

3 I

I I

3 3

I 1

I 3

3 3

3 I

I 3

3 I

3 3

I I

3 3

3 I

3 3

I I

I 3

3 I

I I

I 3

3 I

I I

3 J J 3

3 I

3 3

3 CL 0~

2280 CD IL C..

t CD 0

22

£1-2160 2100 12 l

i T

I T

T i

l l

=

I I

Loss of Reactor Coolant Flow - Underfreqefency Trip Reactor Power vs. Time

\\

1.100 1.0 0.900 0.800 0.700 0.600 0.500 0.400 0.300 0.200 0.1 0.00o

°*

~

~ ~ ~ ~ ~~~~°...................

S........... :-.......... ---

_L 0

1 2

3 4

5 6

7 8

9 10 Time [s]

Figure 14.1.8-6 Rev.

16 12/01/2000 E

0 0

0 a"

O.

1

~1_

I I

o I

I u,,

I 1

!III I

-_ I,...

I

/

I, I

SI I

I (U

i I

I I

0 0

o-------------------------- ----------------------1---

I E

C I

I I

i i '

I..0 I"

+

"I I.

0 I

0 0- 0 n,

.I I

LL 7

' 3, CU I

I 1

i i

l l

i II I

I 1

I I

0 I

0 In I.

C 000 0

[.A -69o] gJnID.JedwBJ. dool SO*I N

oooz/folzt 91

-Aaa L-8"i'3f ;ann!,_j

[s] aw.l.

8 L

9 9

V c

z L

0 I

v I. )00 o L 6

............ - l....

.... i li'............

S.....

a)WI I"s^ xnl~l luaH diJ.L AouenbalJapufl - MOU 1UeIOOO Jo10eeU 10 sso-1 L-(3('1'hl 9~

r4"ifJ Hj-11r a9)4ji~J:

0"'0 Wo 09'0 090 0L0o CD

£13

-n

-- Z

=3 0

£3

=3 08"0 00 OV'1

L.-8VVL1 81an6!=l (spuo3as) ~~

6 9

0 IIiI I

a S

I I

a a

a

-'I I

a I

a II a° I

a a

-r a

a a

a aI I

a a

a a-

£

-+-------o4].-------sdu+-l-o-.-.--o--

osoim l

Luoo

.a"^xn-aeHieuq oO

Loss of Reactor Coolant Flow - Underfrequency Trip Minimum DNBR vs. Time "3.00

.5 2.50 2.25 2.00 1.75 1.50 1.25 1.0(

0.7.

0.5' 0.2 4 ----------

.i 5..

0 0

1 2

3 4

5 6

7 8

9 Time [s]

10 Figure 14.1.8-8 Rev.

16 12/01/2000 W

E E-

Complete Loss of Flow - Two Pumps Coasting Down (CLOF)

DNBR vs. Time Transient DNBR DNBR Limit 4 -

3 Z2 2.5 2

1.5 I1 I

I I

I I -----------------------4-----------------------4----------------------

I I

K I

I

/

3 I

I

~I I

I I

~II

~I I

I I!

I I

0 I

I 3

I I

lime (seconds)

Figure 14.1.8-8 12 L

A 1

i 4*

t

! 9

oooz1tolz [

91 I:"aS 6-8"-'LP aan2!j

[s] oW!l 9

V 6

z 0

U p

L

--......... L.....

°.

S.

i / :....

i" i....

0"09 0"L9 OWjl "SA MOI- 0100 joloHj pavjo-I - Moi-luueJooo JO1OEolH 10 SSO-I Al SI

-nl 0

3 5:"

)'0 Y'St 0"06 SLC O**t;

oI 2

I 2_

I I

I 2

m ~~~~-----

+--

0 2

E 0

C 0

I I2 II 2

2 I

2 2

0 U

1-0 U.~

I 2

2 0

0 oO2

0) 0 0

C 0

[9Ss/24i] MCIA 19IUI B.J03 IDJQL C.)

oooz/1to/Z 91 -&aa Of-"IT' oIn_,*

[s] aW!.L 01L 6

8 L

9 9

t' 3

L-0 I.......

II 10 0

O 0g'0

. i C

0 0L"0 C.........oL L

O~3_.

owIUL -SA xnH, IUaH JoloH Pa3Oo1 - MoiI luelooo Jo4o8au 10 sso1 9T 0--U 417',,-/

eYt J1p v

2J

L 0

-i L)

0.

0

a. 10

.2>*

C.-

LI) 0>

C.)

0 c"

0 U-.C 14 0

CI) 0 0

._-I 0

E 0

u to a,

c'!,

I I

i

£ i

I I

I l

l I

I I

i Ii/

I iI

/

i II

/i iI

/I II

/II II I

I SII I

I I

I SI I

II I

I I

I 7

I I

i t

I I I

1 I

0

-o E

4 -

CD

[NOd] *olI oij.gwnlOA dooi Soý I0 a,

c-

C 0

C) 0 C,,

I

.Ir C4.

ýz CD

000Z1/ 0/iT 91

-Aa3

[s] GW!j 8

L 9

v 6

Z L

0 aWIj -SA aJnssaJd j3z!InssaJdl JoIoQ paP)OI - MO!-l lUelOOo JOlOEaB*B o sso']

fT -- i'llh I,,

YOScLg

)SL S..

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

.."°..........

S....

r......

S............

(D Cl)

Cl)

(D "7)

Q3 oo"0Cg O'O0cz 0"0933 0"0SL2 o-oocg -

09N ýIjjfO -Alldý)

y

ý 2J I [-j'['t[

Ia 3n*.!A IlL 6

Complete Loss of Flow - Frequency Decay in Two Pumps (CLOF-UF)

Nuclear Power vs. Time 1.2 3

I I

I 3

I I

3 3

I33 3

I3I SI III I

III I

I I

oI7

,I..

I...

o 3

3_

S

_3 I+

I I

~I3 I

I

,I t

Figure 14.1.8-11

ooozlTolzI 91 ATo/[

[s] owlU I L

6-6 8"8 L'L 9-9 9-9 17't, CT C

'Z VL 0 00"0

\\ ~ ~ ~ ------

..................,.............i...........-.

.. -------I..

..............- ', o

. o 9 "o

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

i........

.......... o

2 V V-t9Q~

=

nW WINd-L 03

_ --L ------

00 "

GWI :SA HlSNG uwflu!UW4 JO1OU. po...o MOld

.U..OO..

JO1 jo SSO7

Complete Loss of Flow - Frequency Decay in Two Pumps (CLOF-UF)

Core Average Heat Flux vs. Time 1.2

.9G -

2.4 4.8 7.2 Time [seconds]

9.6 12 Figure 14.1.8-12 z

0 1,

X c,)

4

-1 a,

(D (D

0 0,

.72

.48 -

.24 0

0 I

3 3

3 I

I..

i I,,

I..

I iI v

I l

f

=

l T

r r

Loss of Reactor Coolant Flow - Locked Rotor Hot Spot Clad Temperature vs. Time 1600.0

\\500.b*

1400.0 1300.0 1200.0 1100.0 1000.0 900.0 800.0 700.0 0

1 2

3 4

5 6

7 8

9 10 Time [s]

Rev.

16 1210112000 Figure 14.1.8-13 1]_

r)

=:3 a) 0E

0)

H--

° °

°--

1-S............

i.....

Complete Loss of Flow - Frequency Decay in Two Pumps (CLOF-UF)

Pressurizer Pressure vs. Time 2400 2340--------

L 2280 --------------

+-

4--------------+---------

U/)

N m

I I!

a) i,,. 2160 -

2100 21 0

t i

=

i I

I I

I f

02.4 4.8 7.2 9.6 12 Time [seconds]

Figure 14.1.8-13

Complete Loss of Flow - Frequency Decay in Two Pumps (CLOF-UF)

RCS Loop Temperature vs. Time Loop 1 Hot Leg Loop 1 Cold Leg Ii CD 1..

.4" 0

E CD

0.

E 3D I_

0 0

-J C,)

0-650 620 590 560 -

530 500 2.4 1

1 4.8 7.2 Time [seconds]

9.6 12 Figure 14.1.8-14 3

3 3

3 1

3 3

I 3

3 3

I 3

3 3

3 I

3 3

3 I

I 3

3 3

3 I

I 3

I I

I 3

I I

3 3

3 I

I

------------------+------------------4------------------+------------------+

I 3

3 I

I 3

I I

I 3

I I

I 3

S4------------------+------------------4------------------+----------------

3 I

3 3

3 I

I 3

3 I

3 3

I I

3 I

I 3

I I

I 3

3 I

I I

3 3

I I

3 3

I I

I 3

I I

-+------------------+------------------+------------------+----------------

I I

I 3

3 I

3 3

I I

I 3

I I

3 3

I I

3 I

I I

3 I

I I

3 I-3 3

I I

I S4------------------+------------------+------------------+----------------

I I

3 I

I I

3 I

I I

3 3

I I

3 I

I I

3 I

I I

3 3

I I

3 I

I I

i

sL I-g"!,'17 !,

nJll!Ij (spuo3Gs) uWLL gL6 9

£0 9I I1 E-r B

IVI B

(_-I11-i0-10) sdwund omj1 u! Aegroa ANounbaj_ -I -mo11 jo sso-1 aoJldtuoC)

GW!.L -SA xnl~t IeGH lauueqo IOH

Complete Loss of Flow - Frequency Decay in Two Pumps (CLOF-UF)

DNBR vs. Time Transient DNBR DNBR Limit 0

3 6

Time (seconds)

Figure 14.1.8-16 J. 1 I 32 25 -

Of 2-a I

aI I

II I

a 1I I

I I

I II a

II I

I I

I a

I I

I a

I I

II I

I a

I I

I a

I I

n I

m I

i 2

2 2

I 2

I

/

i I

I/

0)I I

/

Si 2

I I

II 2

II I

2 SiI 2

I i

/

0)I I

II

12.

L.i I

i i l

i 1-I 2

t-.I SI 2

0 i

CLL (IOI 2

2 I

-~0-----

+----------------------t---------t-----

+o 0

0) te LC)

[oeS/£U] MOA.IUI.lJO:)

ID4Oj

I I

+----------------------------------------------+--------------------

I I

S+----------------------4.

I I

4.----------------------+-------------------------------------------------------------------

I I

-4.-

+

-4.-

4.

I I

[NO.I] MOlJ 3.9!.wnlOA dool pe4,lnD..

a.)

th C.)

a au E 0

LL CL 0

0 0

-j, 0

C 0

0 0

-~E co 0

C)

LL.

,,co r-:

Locked Rotor I Shaft Break - RCS Pressure / PCT Case Nuclear Power vs. Time Figure 14.1.8-19 1.2

.96 0

a, L.

0 C)

(D

.72

.48

.24 0

0oz-q8" I,'* ! jn6!.l

[spuooes] ew!1L 9"6 "L

g' V"Z 0

I I

Io 9a aa I!

I II I

I a a

+-----+-----+------

a 0

1

+-----4

4.

+

I 1 0

L'J I

TI a

IJL -SA xnl:a1IeOH oaeJAV WOO oseo l~d I oinssaJd SON -

eoJI3 :Ueqs I JO010H P85130-1

Locked Rotor I Shaft Break - RCS Pressure I PCT Case Pressurizer Pressure vs. Time 2700 3

33i 26 0 -*

3..

I I

SI 3

3!

I 3

I II 3

3 w

3

3.

III I

II I

I

)

III L2 I

I C,,

I CI I

/i I

i oII I

3I 3

3 3

I 3

3 3

I I

S 2400-

+-+

220 i

Ih I

I I

Figure 14.1.8-21

Locked Rotor I Shaft Break - RCS Pressure I PCT Case Vessel Lower Plenum Pressure vs. Timne 2800 2700

-3 3

3I

_3I I3 1

3I 3

I2600 I

I I

II_

E 2700-----------+--

-+--+-

23 0,

I I

Figure 14.1.8-22

Locked Rotor / Shaft Break - RCS Pressure I PCT Case RCS Loop Temperature vs. Time Loop 1 Hot Leg Loop 1 Cold Leg 2.4 1

1 4.8 7.2 Time [seconds]

9.6 12 Figure 14.1.8-23 650 620 a,

(D

'*J L.

0 a-a, I-

0.

0 0

_J U) 13::

590 560 530 500 '

I 3

3 3

I 3

3 3

I 3

3 3

I 3

3 I

I 3

3 3

I I

3 3

I I

I 3

I I

I 3

3 S+-----------4-----4----

4----------

3 3

I 3

3 I

3 3

I 3

I I

3 I

3 3

3 1

3 3

I I

3 I

I I

3 I

I I

I 3

-----4-----+-----+-----4----

3 3

3 I

I 3

I I

3 3

I I

I 3

I

_-------t I

I I

3 I

I I

3 3

I I

I 3

4-----4-----4-----4----

I I

I 3

I I

3 I

I I

I 3

I I

3 3

I I

3 3

I 3

I I

3 I

I I

3 3

I I

3 3

I I

I 3

I I

I

____ ii I

I

Hot Channel Heat Flux vs. Time Locked Rotor ! Shaft Break* - RCS Pressure ! PCT Case 3-2 3

3 "3

3 X

3 3

3..

I 3

I II I

II 3

I

-2 Time (seconds)

Figure 14.1.8-24

Locked Rotor I Shaft Break - RCS Pressure I PCT Case Hot Spot Cladding Inner Temperature vs. Time 2000 I

I I

I..

"I..

16I..

-'I L.

.21400 E

c 1200---------------

r-1000-I I

OO 600 Time (seconds)

Figure 14.1.8-25

igure 14.1.8-13 shows the clad temperature transient at the hot spot. Since in the worst case ex ed, the clad temperature does not exceed 1800°F, it is not necessary to consider t possi ity of a zirconium-steam reaction. The zirconium-steam reaction is only signi ant above t temperature.

The following le shows the comparison of the important calculated safety eters to their respective ac tance criteria (Calculated Valve/Acceptance Criterio o Uel Rods Max Clad RCS Pressure S Pressure Limit Temp. (0F)

(psial bfp.

Locked Rotor

/2700 EM275 1044/1210

- S

Percentage of Fuel Rods with L-LI oŽ*

Conclusions Since the peak pressure reached during the tr tis < 110% of design, the integrity of the Reactor Coolant System is not endangered.

e p sure can be considered as an upper limit because of the following conservative ptions u in the study:

1. Credit is not taken for the negaf e moderator coefficie

2. It is assumed that the pre rizer relief valves were inoperativ

3. The steam dump is sumed to be inoperative.

The peak clad te erature calculated for the hot spot, can also be considered upper limit because of the olowing:

1. The t spot is assumed to be in DNB at the start of the accident.

2.

high gap coefficient is used during the transient.

3. The nuclear heat released in the fuel at the hot spot is based on a zero moderator coefficient.

14.1.9 LOSS OF EXTERNAL ELECTRICAL LOAD ccident Description e

4e,-w1" ctIo

_ 5eA i -x The loss of ex ectrical load may result from an abnormal increas~e in netwo

uency, opening of the main bre a-m.

the generator, which causes a e Nuclear Steam generator.

The plant is designed to acc

-load rejection without a a reactor trip. The automatic steam dum em with 85% steam dump capacity (40% to the ser and 45%

Rev. 16 14.1-24 12/01/2000

11 o L3 to the atmosphere) is able to accommodate this load rejection by reducing the transient posed upon the Reactor Coolant System. The reactor power is reduced to the n eq flibrium power level at a rate consistent with the capability of the Rod Co ol Syst

. The pressurizer relief valves may be actuated, but the pressurizer safety valv and the st generator safety valves do not lift for a step loss of load with steam p to auxili oad.

In the event e steam dump valves fail to open following a large load loss, the s generator safety valves y lift and the reactor may be tripped by the high pressuriz pressure signal or the high pr ep er level signal. The steam generator shell side pressure d reactor coolant temperatures woul increase rapidly. The pressurizer safety valves and s am generator safety valves are, however, 'zed to protect the Reactor Coolant System and ear generator against overpressure for all lo losses without taking credit for the steam p system.

The most likely source of a omplete loss of load on the Nuclear team Supply System is a trip of the turbine generator.

this case, there is a direct re tor trip signal (unless below approximately 10% power) d *ved from either the turbine to-stop oil pressure or a closure of the turbine stop valves. Rea or coolant temperatur and pressure do not significantly increase if the steam dump an pressurizer press e control system are functioning properly. However, in this analysis, 2e behavior of e plant is evaluated for a complete loss of load from 102% of full power witho t a direct r ctor trip primarily to show the adequacy of the pressure relieving devices and als to sho that no core damage occurs. The Reactor Coolant System and Steam System pressur r eying capacities are designed to insure safety of the plant without requiring the automatic d control, pressurizer pressure control and/or steam dump control systems.

Method of Analysis The total loss of load transient are analyzed by em loying a detailed digital computer program. The program describ the neutron kinetics, R ctor Coolant System, pressurizer, pressurizer relief and safety Ives, pressurizer spray, ste generator, and steam generator safety valves.

The objectives of this alysis are to demonstrate margins to c protection limits and to demonstrate the ad acy of the plant pressure relieving devices.

a. The initial r ctor power and Reactor Coolant System temperaturesre assumed at their maximum alues consistent with steady-state full power operation, in uding allowances for calib tion and instrument errors. The initial Reactor Coolant Sy em pressure is assums at the mininmum value consistent with steady-state full power op

on, including allo ances for calibration and instrument errors. This results in the maxiurn power di *rence for the load loss, and the minimum margin to core protection its at the

.ftiatin of the total loss-of-load accident.

The total loss of load is analyzed for both BOC and EOC conditions. At BOC, a ero moderator coefficient (0.0 Ak/0F) is used; and at EOC, a moderator coefficient value f Rev. 16 14.1-25 12/01/2000 14.1.

-4.OE-4 Ak/OF is used. A conservatively large absolute value of the Doppler coeffici i used for all cases with a negative moderator coefficient. For the cases in which e

m erator coefficient is zero, a conservatively small absolute value of the Do ler coe *ent is used.

c. Two cas for both the beginning and end-of-life are analyzed as follows:

The react is assumed to be in normal automatic control with the co ol rods in the minimum emental worth region.

9 The reactor i ed to be in manual control with no control d insertion until a reactor trip ocC

d. No credit is taken for of the steam dump valves or power-perated steam generator relief valves. The steam erator pressure rises to the safe.t alve set point where steam release through safety valv limits secondary steam pres e at the set point.

e. Two cases for both the beginni and end-of-life are lyzed as follows:

Full credit is taken for the effect pressurizer ray and power-operated relief valves in reducing or limiting coolant pre ure.

"* No credit is taken for the effect of pr sun spray and power-operated relief valves in reducing or limiting coolant pressur A nominal pressurizer safety valve etpo t of 2500 psia is assumed. Sensitivity analyses were performed at press zer safet valve settings of +6% and -4% of the nominal setpoint -to account fothe effects steam accumulation and setpoint drift. The critical safety p eters were sho to be acceptable under these assumptions.

Results Figures 14.1.9-1 through 1.1.9-5 show the transient responses r a total loss of load at beginning of cycle with omoderator temperature coefficient ass g full credit for the pressurzer spray,.pre er power-operated relief valves, and aomatic control rod insertion. No cred taken for the steam dump system.

Figures 14.1.9-6 ough 14.1.9-10 show the responses for the total loss of lo tend of cycle with the most gative moderator temperature coefficient (-4.OE--4 Ak/OF).

rest of the plant operat conditions are the same as the case above.

The loss-f-load accident is also analyzed assuming manual RCCA control. In addi n, no credit i taken for the pressurizer spray, pressurizer power-operated relief valves, or s am dum system. Figures 14.1.9-11 through 14.1.9-15 show the manual control beginning f cy transient with zero moderator coefficient. Figures 14.1.9-16 through 14.1.9-20 show th nual control transient results at end of cycle.

Rev. 16 i A 1-2i 12/01/2000

to following table shows the comparison of the important calculated s espective acceptance criteria (Calculated Value/Acceptance Criteri Loss of d

MDNBR RCS Pressure SS Pre, (p s ia)

(psi n

BOCManualContr 1.681/1.14 2501/.0 1182/1210 BOC Auto Control

.681/1.14 750 iM1210 eeAýO"I EOC Manual Control

1.

.14 10/2750 I

/1210 EOC Auto Control 1.681/

2377/2750 1198/1210 Conclusions The safety an is indicates that a total loss of load without ect or immediate reactor trip presents o hazard to the integrity of the Reactor Coolan ystem or the Steam Pressure relieving devices incorporated in the two systems are eate to limit the m-,,faximnum pressures to within safety analysis limits. The integrity of the co

s maintained by the Reactor Protection System. The MDNBR does not fall below -its initiali is above the MDNBR limit.

14.1.10 LOSS OF NORMAL FEEDWATER (Se.dion 14-J.I.O chnnw je Accident Description SL&3eS+CL

,cuhr, A loss of normal feedwater (from a pipe break, pump failure, valve malfunctio, or loss of off-site power) results in a reduction in capability of the secondary system t ove the heat g

erated in the reactor core. If the reactor is not tripped during this acci t, Reactor Coolant Syst damage could possibly occur from a sudden loss of heat sink.

an alternative supply of feedwer is not supplied to the plant, residual heat following ctor trip heats the coolant to the point ere water relief from the pressurizer occurs. S ficant loss of water from the Reactor Coolan stem could conceivably lead to core age.

o L /TT-The following provide e necessary protection nst a loss of normal feedwater:

1. Reactor trip on Low-Low ter level ither steam generator.

2. Reactor trip on steam flow-f a

flow mismatch in coincidence with low water level in either steam generator.

3. Two motor driven a iliary feedwater pumps hich are started automatically on:

a) Low-Lo evel in either steam generator, or b) nig of both feedwater pump circuit breakers, or c

Safety Injection signal, or 14.1-27 12/01/2000

ML022200575

Text

4

Navigation menu

Search