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{{#Wiki_filter:Enclosure I to ET 13-0023WCAP-17658-NP, "Wolf Creek Generating Station Transition of Methods for Core Designand Safety Analyses
-Licensing Report" (Non-Proprietary)
(454 pages)
Westinghouse Non-Proprietary Class 3WCAP-17658-NP August 20'Revision 0Wolf Creek Generating StationTransition of Methods forCore Design and SafetyAnalyses
-Licensing ReportWestinghouse 13 WESTINGHOUSE NON-PROPRIETARY CLASS 3WCAP-17658-NP Revision 0Wolf Creek Generating StationTransition of Methods for Core Design andSafety Analyses
-Licensing ReportAugust 2013Prepared:
Denise Solomon*,
EngineerEngineering, Equipment and Major ProjectsPlant Licensing Approved:
Dewey Olinski*,
ManagerEngineering, Equipment and Major ProjectsPlant Licensing
*Electronically approved records are authenticated in the electronic document management system.Westinghouse Electric Company LLC1000 Westinghouse DriveCranberry
: Township, PA 16066, USA© 2013 Westinghouse Electric Company LLCAll Rights Reserved WESTINGHOUSE NON-PROPRIETARY CLASS 3ivTABLE OF CONTENTSL IS T O F T A B L E S ......................................................................................................................................
v iiL IS T O F F IG U R E S ......................................................................................................................................
xLIST OF ACRONYM S ..............................................................................................................................
xixPROFESSIONAL ENGINEERING STAMPS .........................................................................................
xxiiINTRODUCTION
........................................................................................................................
1-11.1 NUCLEAR STEAM SUPPLY SYSTEM PARAM ETERS ............................................
1-21.1.1 In tro du ctio n .....................................................................................................
1-21.1.2 Input Parameters, Assumptions, and Acceptance Criteria
...............................
1-21.1.3 Description of Analyses and Evaluation
..........................................................
1-31.1.4 C on clu sion s .....................................................................................................
1-42 ACCIDENT AND TRANSIENT ANALYSIS
..............................................................................
2-12.1 NON-LOCA ANALYSES INTRODUCTION
................................................................
2-12.1.1 Program Features
.............................................................................................
2-12.1.2 Non-LOCA Transient Events Considered
.......................................................
2-22.1.3 Analysis M ethodology
.....................................................................................
2-52.1.4 Computer Codes Used .....................................................................................
2-82.1.5 Initial Conditions
...........................................................................................
2-102.1.6 Fuel Design Description
................................................................................
2-122.1.7 Power Distribution Peaking Factors .........................................................
2-132.1.8 Reactivity Feedback
......................................................................................
2-132.1.9 Pressure Relief M odeling ..........................................................................
2-132.1.10 RTS and ESFAS Functions
............................................................................
2-152.1.11 Reactor Trip Characteristics
..........................................................................
2-162.1.12 Operator Actions Credited
.............................................................................
2-162.1.13 Results Summary ...........................................................................................
2-172 .1.14 R eferences
.....................................................................................................
2-172.2 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM .....................
2-372.2.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (USAR Section 15.1.1) .............................................................
2-372.2.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow (USAR Section 15.1.2) .........................................................................
2-452.2.3 Excessive Increase in Secondary Steam Flow (USAR Section 15.1.3) .........
2-542.2.4 Inadvertent Opening of a Steam Generator Atmospheric Relief or SafetyValve (USAR Section 15.1.4) ........................................................................
2-682.2.5 Steam System Piping Failure (USAR Section 15.1.5) ..................................
2-81WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3VTABLE OF CONTENTS (cont.)2.3 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM ..................
2-1122.3.1 Loss of External Electrical Load, Turbine Trip, Inadvertent Closure ofMain Steam Isolation Valves, Loss of Condenser Vacuum and OtherEvents Resulting in Turbine Trip (USAR Sections 15.2.2, 15.2.3,15 .2 .4 , and 15 .2 .5) .......................................................................................
2-1122.3.2 Loss of Non-Emergency AC Power to the Station Auxiliaries (USARS ection 15 .2 .6) .............................................................................................
2-1282.3.3 Loss of Normal Feedwater Flow (USAR Section 15.2.7) ...........................
2-1442.3.4 Feedwater System Pipe Break (USAR Section 15.2.8) ...............................
2-1612.4 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE ...........................
2-1792.4.1 Partial and Complete Loss of Forced Reactor Coolant Flow(U SA R Sections 15.3.1 and 15.3.2) .............................................................
2-1792.4.2 Reactor Coolant Pump Shaft Seizure (Locked Rotor) and Shaft Break(U SA R Sections 15.3.3 and 15.3.4) .............................................................
2-2052.5 REACTIVITY AND POWER DISTRIBUTION ANOMALIES
...............................
2-2242.5.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from aSubcritical or Low Power Startup Condition (USAR Section 15.4.1) ........
2-2242.5.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal atPow er (U SA R Section 15.4.2) .....................................................................
2-2342.5.3 Control Rod Misoperation (USAR Section 15.4.3) .....................................
2-2552.5.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Tem perature (U SA R Section 15.4.4) ...........................................................
2-2632.5.5 Chemical and Volume Control System Malfunction Resulting in aDecrease in Boron Concentration in the Reactor Coolant(U SA R Section 15.4.6) ................................................................................
2-2632.5.6 Spectrum of Rod Cluster Control Assembly Ejection Accidents (U SA R Section 15.4.8) ................................................................................
2-2722.6 INCREASE IN REACTOR COOLANT INVENTORY
.............................................
2-2852.6.1 Inadvertent Operation of the Emergency Core Cooling System DuringPower Operation (USAR Section 15.5.1) ....................................................
2-2852.6.2 Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory (USAR Chapter 15.5.2) ....................................
2-2942.7 DECREASE IN REACTOR COOLANT INVENTORY
............................................
2-3072.7.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve(U SA R Section 15.6.1) ................................................................................
2-3072.7.2 Steam Generator Tube Rupture Margin to Overfill(U SA R Section 15.6.3) ................................................................................
2-3 132.7.3 Steam Generator Tube Rupture -Input to Dose (USAR Section 15.6.3) ... 2-3342.7.4 Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary(U SA R Section 15.6.5) ................................................................................
2-355WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3viTABLE OF CONTENTS (cont.)2.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM (USAR SECTION 15.8) ....... 2-3692.8.1 Technical E valuation
...................................................................................
2-3692 .8.2 C onclusion
...................................................................................................
2-37 12 .8.3 R eferences
...................................................................................................
2-37 12.9 RA D IO LO G IC A L D O SES ..........................................................................................
2-3812.10 INSTRUMENT UNCERTAINTIES
...........................................................................
2-3812.10.1 Reactor Trip System/Engineered Safety Feature Actuation SetpointU ncertainties
................................................................................................
2-3 8 12.10.2 Initial C ondition U ncertainties
....................................................................
2-3832.11 CONTROL SYSTEMS ANALYSIS
...........................................................................
2-3842.11.1 NSSS Pressure Control Component Sizing(USAR Chapters 5.4, 7.7, & 10.4.4) ............................................................
2-3842.11.2 Operational Analysis and Margin to Trip (USAR Chapter 7.7) ..................
2-3902.12 THERMAL AND HYDRAULIC DESIGN ................................................................
2-3952 .12 .1 Introduction
.................................................................................................
2-3952.12.2 Input Parameters and Acceptance Criteria
...................................................
2-3952.12.3 Description of Analyses and Evaluations
....................................................
2-3982 .12 .4 R esu lts .........................................................................................................
2-4 0 12 .12 .5 C onclusion
...................................................................................................
2-4022 .12 .6 R eferences
...................................................................................................
2-402APPENDIX A SAFETY EVALUATION REPORT COMPLIANCE
...............................................
A-1WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3viiLIST OF TABLESTable 1.1-1Table 1.1-2Table 2.1-1Table 2.1-2Table 2.1-3Table 2.1-4Table 2.1-5Table 2.1-6Table 2.2.1-1Table 2.2.1-2Table 2.2.2-1Table 2.2.2-2Table 2.2.2-3Table 2.2.3-1Table 2.2.3-2Table 2.2.4-1Table 2.2.4-2Table 2.2.5.1-1 Table 2.2.5.1-2 Table 2.2.5.2-1 Table 2.2.5.2-2 Table 2.3.1-1Table 2.3.1-2Table 2.3.2-1Table 2.3.3-1NSSS Design Parameters for WCGS TM Program ......................................................
1-5NSSS Design Parameters for WCGS TM Program ......................................................
1-6Non-LOCA Transient Events Analyzed or Evaluated
................................................
2-19Summary of Initial Conditions and Computer Codes Used .......................................
2-20Core Kinetics Parameters and Reactivity Feedback Coefficients
...............................
2-23Summary of RTS and ESFAS Functions Actuated
.....................................................
2-24Parameters Related to OTAT and OPAT RT Setpoints
...............................................
2-27N on-LO C A Results Sum m ary ....................................................................................
2-28Time Sequence of Events -Decrease In Tfted (Manual Rod Control)
........................
2-41Decrease in Tfeed Minimum DNBR and Peak Core Average Thermal PowerR e su lts ........................................................................................................................
2 -4 1Increase in FW Flow Cases A nalyzed ........................................................................
2-49Time Sequence of Events -Increase in FW Flow (HFP, Multi-Loop, ManualR od C o ntro l) ...............................................................................................................
2-4 9HFP FWM Flow Increase Minimum DNBR and Peak Core Average ThermalP ow er R esu lts .............................................................................................................
2-4 9Excessive Load Increase Incident Summary of Results .............................................
2-58Time Sequence of Events for the Excessive Load Increase Incident
.........................
2-59Time Sequence of Events -Accidental Depressurization of the MSS at HZPC o n d itio n s ...................................................................................................................
2 -7 3Limiting Results -Accidental Depressurization of the MSS at HZP Conditions
...... 2-73Time Sequence of Events -Steam System Piping Failure at HZP Conditions
.........
2-87Limiting Results -Steam System Piping Failure at HZP Conditions
........................
2-88Time Sequence of Events -Steam System Piping Failure at HFP Conditions
........
2-107Limiting Results -Steam System Piping Failure at HFP Conditions
.......................
2-107Time Sequence of Events -Loss of External Electrical Load and/orT u rb in e T rip ..............................................................................................................
2 -118Limiting Results -Loss of External Electrical Load and/or Turbine Trip ...............
2-118Time Sequence of Events for Limiting LOAC Case ................................................
2-133Time Sequence of Events for Limiting LONE Case .................................................
2-150WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3viiiWESTINGHOUSE NON-PROPRIETARY CLASS 3 viiiLIST OF TABLES (cont.)Table 2.3.4-1Table 2.3.4-2Table 2.4.1-1Table 2.4.1-2Table 2.4.2-1Table 2.4.2-2Table 2.5.1-1Table 2.5.2-1Table 2.5.2-2Table 2.5.3-1Table 2.5.3-2Table 2.5.5-1Table 2.5.6-1Table 2.5.6-2Table 2.6.1-1Table 2.6.2-1Table 2.7.1-1Table 2.7.1-2Table 2.7.2-1Table 2.7.2-2Table 2.7.2-3Table 2.7.2-4Table 2.7.2-5Time Sequence of Events for Limiting Feed Line Break Case With OffsitePow er A vailable
........................................................................................................
2-168Time Sequence of Events for Limiting Feed Line Break Case Without OffsitePow er A vailable
........................................................................................................
2-168Time Sequence of Events -Loss of Forced Reactor Coolant Flow .........................
2-183Results -Loss of Forced Reactor Coolant Flow ......................................................
2-183Time Sequence of Events -RCP Locked Rotor/Shaft Break ...................................
2-210Limiting Results -RCP Locked Rotor/Shaft Break .................................................
2-211Time Sequence of Events -Uncontrolled RCCA Bank Withdrawal from aSubcritical C ondition
................................................................................................
2-229Time Sequence of Events -Uncontrolled RCCA Bank Withdrawal at Power .........
2-241Uncontrolled RCCA Bank Withdrawal at Power -Limiting Results .......................
2-242Non-LOCA Analysis Limits and Analysis Results for the Dropped Rod Event ....... 2-262Summary of Initial Conditions and Computer Codes Used for the DroppedR od E v en t .................................................................................................................
2 -2 62CVCS Malfunction Boron Dilution Event Results -Event Alarm to Loss ofShutdow n M argin .....................................................................................................
2-27 1Selected Input and Results of the Limiting RCCA Ejection Analyses
.....................
2-279Time Sequence of Events -RCCA Ejection
............................................................
2-280Time Sequence of Events -Inadvertent ECCS ........................................................
2-290Time Sequence of Events -CVCS Malfunction
......................................................
2-298Time Sequence of Events -Accidental Depressurization of the RCS ......................
2-310Results -Accidental Depressurization of the RCS ...................................................
2-310AFW Flows for Design Basis SGTR Analyses MDAFW Failure, All AFWPum ps O perating
......................................................................................................
2-323AFW Flows for Design Basis SGTR Analyses MDAFW Failure, TDAFWPump Stopped, MDAFW Pumps Operating
.............................................................
2-323AFW Flows for Design Basis SGTR Analyses MDAFW Failure, RupturedSG Isolated, TDAFW Pump Stopped, MDAFW Pumps Operating DuringC o o ld o w n .................................................................................................................
2 -32 3SI Flows for Design Basis SGTR Analyses
..............................................................
2-324Operator Action Times for Design Basis SGTR Margin to Overfill Analyses
.........
2-325WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3ixWESTINGHOUSE NON-PROPRIETARY CLASS 3 lxTable 2.7.2-6Table 2.7.3-1Table 2.7.3-2Table 2.7.3-3Table 2.7.3-4Table 2.7.4-1Table 2.7.4-2Table 2.7.4-3Table 2.8.1-1Table 2.8.1-2Table 2.12-1Table 2.12-2Table 2.12-3LIST OF TABLES (cont.)Sequence of Events for Limiting Margin to Overfill Analyses
................................
2-325Operator Action Times for Design Basis SGTR T/H Analyses
................................
2-342Sequence of Events for Limiting Input to Radiological Consequences A n aly ses ...................................................................................................................
2 -34 2Break Flow and Flashed Break Flow ........................................................................
2-343Intact and Ruptured SG Steam Flow to Atmosphere
................................................
2-343Subcriticality Analysis Input Parameters
..................................................................
2-362LTC A nalysis Input Param eters ................................................................................
2-362Boric Acid Solution Solubility Limit Data ...............................................................
2-363LOL ATWS Time Sequence of Events .....................................................................
2-372LONF ATWS Time Sequence of Events ..................................................................
2-372T/H Design Parameters Comparison
........................................................................
2-403Limiting Parameter Direction for DNB ....................................................................
2-405RTDP DNBR Margin Surmnary
...............................................................................
2-406WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3XWESTINGHOUSE NON-PROPRIETARY CLASS 3 xLIST OF FIGURESFigure 2.1 -1Figure 2.1-2Figure 2.1-3Figure 2.1-4Figure 2.1-5Figure 2.1-6Figure 2.2.1-1Figure 2.2.1-2Figure 2.2.1-3Figure 2.2.2-1Figure 2.2.2-2Figure 2.2.2-3Figure 2.2.2-4Figure 2.2.3-1Figure 2.2.3-2Figure 2.2.3-3Figure 2.2.3-4Figure 2.2.4-1Figure 2.2.4-2Figure 2.2.4-3Integrated D P C ........................................................................................................
2 -3 1R eactor C ore Safety L im its .....................................................................................
2-32Illustration of OTAT and OPAT Protection
.............................................................
2-33Fractional Rod Insertion versus Time from Release ...............................................
2-34Normalized RCCA Reactivity Worth versus Fractional Rod Insertion
...................
2-35Normalized RCCA Reactivity Worth versus Time from Release ............................
2-36Decrease in T'eed at Full Power -Nuclear Power and Core Heat Fluxversu s T im e .............................................................................................................
2-4 2Decrease in Tfred at Full Power -Vessel Delta-T and Core Average Moderator Tem perature versus T im e ........................................................................................
2-43Decrease in Tfd at Full Power -Pressurizer Pressure and DNBRv ersu s T im e .............................................................................................................
2 -4 4Increase in FW Flow at Full Power -Multi-Loop Manual Rod Control -Nuclear Power and Core Heat Flux Versus Time ....................................................
2-50Increase in FW Flow at Full Power -Multi-Loop Manual Rod Control -Core Average Moderator Temperature and Pressurizer Pressure Versus Time ........
2-51Increase in FW Flow at Full Power -Multi-Loop Manual Rod Control -SG Mass Inventory and Pressure Versus Time ........................................................
2-52Increase in FW Flow at Full Power -Multi-Loop Manual Rod Control -D N B R V ersus T im e .................................................................................................
2-5310% Step Increase in Heat Removal by Secondary System -MinimumReactivity
: Feedback, Manual Reactor Control .......................................................
2-6010% Step Increase in Heat Removal by Secondary System -MinimumReactivity
: Feedback, Automatic Reactor Control ...................................................
2-6210% Step Increase in Heat Removal by Secondary System -MaximumReactivity
: Feedback, Manual Reactor Control .......................................................
2-6410% Step Increase in Heat Removal by Secondary System -MaximumReactivity
: Feedback, Automatic Reactor Control ...................................................
2-66Accidental Depressurization of the MSS at HZPNuclear Power and Core Heat Flux versus Time ....................................................
2-74Accidental Depressurization of the MSS at HZPReactor Vessel Inlet Temperature and Core Average Temperature versus Time ..... 2-75Accidental Depressurization of the MSS at HZPPressurizer Pressure and Pressurizer Water Volume versus Time ...........................
2-76WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3XiLIST OF FIGURES (cont.)Figure 2.2.4-4Figure 2.2.4-5Figure 2.2.4-6Figure 2.2.4-7Figure 2.2.5.1-1 Figure 2.2.5.1-2 Figure 2.2.5.1-3 Figure 2.2.5.1-4 Figure 2.2.5.1-5.
Figure 2.2.5.1-6 Figure 2.2.5.1-7 Figure 2.2.5.1-8 Figure 2.2.5.1-9 Figure 2.2.5.1-10 Figure 2.2.5.1-11 Figure 2.2.5.1-12 Figure 2.2.5.1-13 Accidental Depressurization of the MSS at HZPCore Boron Concentration and Reactivity versus Time ..........................................
2-77Accidental Depressurization of the MSS at HZPSteam Pressure and Steam (Break) Flow versus Time ............................................
2-78Accidental Depressurization of the MSS at HZPFW Flow and SG M ass versus Tim e .......................................................................
2-79Accidental Depressurization of the MSS at HZPC ore Flow versus T im e ............................................................................................
2-80Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)
Nuclear Power and Core Heat Flux versus Time ...................................
2-89Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)
Reactor Vessel Inlet Temperature and Core Average Temperature versu s T im e .............................................................................................................
2-90Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)
Pressurizer Pressure and Pressurizer Water Volume versus Time ..........
2-91Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)
Core Boron Concentration and Reactivity versus Time .........................
2-92Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)
Steam Pressure and Steam (Break) Flow versus Time ..........................
2-93Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)
FW Flow and SG Mass versus Time ......................................................
2-94Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)
Core Flow versus Tim e ..........................................................................
2-95Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)
Nuclear Power and Core Heat Flux versus Time ...................................
2-96Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)
Reactor Vessel Inlet Temperature and Core Average Temperature v ersu s T im e .............................................................................................................
2 -9 7Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)
Pressurizer Pressure and Pressurizer Water Volume versus Time ..........
2-98Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)
Core Boron Concentration and Reactivity versus Time .........................
2-99Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)
Steam Pressure and Steam (Break) Flow versus Time ........................
2-100Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)
FW Flow and SG M ass versus Time ....................................................
2-101WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xiiWESTINGHOUSE NON-PROPRIETARY CLASS 3 xiiLIST OF FIGURES (cont.)Figure 2.2.5.1-14 Figure 2.2.5.2-1 Figure 2.2.5.2-2 Figure 2.2.5.2-3 Figure 2.2.5.2-4 Figure 2.3.1 -1Figure 2.3.1-2Figure 2.3.1-3Figure 2.3.1-4Figure 2.3.1-5Figure 2.3.1-6Figure 2.3.1-7Figure 2.3.1-8Figure 2.3.1-9Figure 2.3.2-1Figure 2.3.2-2Figure 2.3.2-3Figure 2.3.2-4Figure 2.3.2-5Figure 2.3.2-6Figure 2.3.2-7Figure 2.3.2-8Figure 2.3.2-9Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)
Core Flow versus Tim e ........................................................................
2-102Steam System Piping Failure at HFP (1.04 ft2 Break) Nuclear Power andC ore H eat Flux versus T im e ..................................................................................
2-108Steam System Piping Failure at HFP (1.04 ft2 Break) Pressurizer Pressure andPressurizer W ater Volum e versus Tim e ................................................................
2-109Steam System Piping Failure at HFP (1.04 ft2 Break) Reactor Vessel InletTemperature and Loop Average Temperature versus Time ...................................
2-110Steam System Piping Failure at HFP (1.04 ft2 Break) Steam Pressure andB reak Flow versus T im e ........................................................................................
2-111LOL/TT, Minimum DNBR Case Nuclear Power and SG Pressure versus Time.. 2-119LOL/TT, Minimum DNBR Case Pressurizer Pressure and Pressurizer W ater Volum e versus Tim e ....................................................................................
2-120LOL/TT, Minimum DNBR Case RCS Temperatures and DNBR versus Time .... 2-121LOL/TT, Peak MSS Pressure Case Nuclear Power and SG Pressurev ersu s T im e ...........................................................................................................
2 -12 2LOL/TT, Peak MSS Pressure Case Pressurizer Pressure and Pressurizer W ater Volum e versus Tim e ....................................................................................
2-123LOL/TT, Peak MSS Pressure Case RCS Temperatures versus Time ....................
2-124LOL/TT, Peak RCS Pressure Case Nuclear Power and SG Pressureversu s T im e ...........................................................................................................
2-12 5LOL/TT, Peak RCS Pressure Case RCS Pressures and Pressurizer WaterV olum e versus T im e ..............................................................................................
2-126LOL/TT, Peak RCS Pressure Case RCS Temperatures versus Time ....................
2-127LOAC -Nuclear Power versus Tim e ....................................................................
2-134LOAC -Core Average Heat Flux versus Time .....................................................
2-135LOAC -Reactor Coolant Loop Flow versus Time ...............................................
2-136LOAC -HL and CL Temperatures versus Time ...................................................
2-137LOAC -Actual Pressurizer Pressure versus Time ................................................
2-138LOAC -Pressurizer Water Volume versus Time ..................................................
2-139LOAC -SG Pressure versus Tim e ........................................................................
2-140LOAC -Indicated SG Level versus Tim e .............................................................
2-141LO A C -SG M ass versus Tim e .............................................................................
2-142WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xiiiWESTINGHOUSE NON-PROPRIETARY CLASS 3 XIIILIST OF FIGURES (cont.)Figure 2.3.2-10Figure 2.3.3-1Figure 2.3.3-2Figure 2.3.3-3Figure 2.3.3-4Figure 2.3.3-5Figure 2.3.3-6Figure 2.3.3-7Figure 2.3.3-8Figure 2.3.3-9Figure 2.3.3-10Figure 2.3.4-1Figure 2.3.4-2Figure 2.3.4-3Figure 2.3.4-4Figure 2.3.4-5Figure 2.3.4-6Figure 2.3.4-7Figure 2.3.4-8Figure 2.3.4-9Figure 2.3.4-10Figure 2.4.1-1Figure 2.4.1-2LOAC -Loop AFW Flow versus Time ................................................................
2-143LON F -N uclear Pow er versus Tim e ....................................................................
2-151LONF -Core Average Heat Flux versus Time ......................................................
2-152LONF -Reactor Coolant Loop Flow versus Time ...............................................
2-153LONF -HL and CL Temperatures versus Time ....................................................
2-154LONF -Actual Pressurizer Pressure versus Time ................................................
2-155LONF -Pressurizer Water Volume versus Time ...................................................
2-156LO N F -SG Pressure versus Tim e ........................................................................
2-157LON F -Indicated SG Level versus Tim e .............................................................
2-158LO N F -SG M ass versus Tim e ..............................................................................
2-159LONF -Loop AFW Flow versus Time .................................................................
2-160Feed Line Break with Offsite Power Available Nuclear Power, CoreHeat Flux and Total Core Reactivity versus Time .................................................
2-169Feed Line Break with Offsite Power Available Pressurizer Pressure andPressurizer W ater Volum e versus Tim e .................................................................
2-170Feed Line Break with Offsite Power Available Reactor Coolant Flow andFW Line Break Flow versus Tim e .........................................................................
2-171Feed Line Break with Offsite Power Available Faulted Loop and IntactLoop Reactor Coolant Temperatures versus Time ................................................
2-172Feed Line Break with Offsite Power Available SG Shell Pressureversu s T im e ...........................................................................................................
2-173Feed Line Break without Offsite Power Nuclear Power, Core Heat Fluxand Total Core Reactivity versus Tim e ..................................................................
2-174Feed Line Break without Offsite Power Pressurizer Pressure andPressurizer W ater Volume versus Tim e .................................................................
2-175Feed Line Break without Offsite Power Reactor Coolant Flow andFW Line Break Flow versus Tim e .........................................................................
2-176Feed Line Break without Offsite Power Faulted Loop and Intact LoopReactor Coolant Temperatures versus Time ..........................................................
2-177Feed Line Break without Offsite Power SG Shell Pressure versus Time ..............
2-178PLOF -Core Volumetric Flow Rate versus Time .................................................
2-184PLOF -Loop Volumetric Flow Rates versus Time ...............................................
2-185WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xivLIST OF FIGURES (cont.)Figure 2.4.1-3Figure 2.4.1-4Figure 2.4.1-5Figure 2.4.1-6Figure 2.4.1-7Figure 2.4.1-8Figure 2.4.1-9Figure 2.4.1-10Figure 2.4.1-11Figure 2.4.1-12Figure 2.4.1-13Figure 2.4.1-14Figure 2.4.1-15Figure 2.4.1-16Figure 2.4.1-17Figure 2.4.1-18Figure 2.4.1-19Figure 2.4.1-20Figure 2.4.1-21Figure 2.4.2-1Figure 2.4.2-2Figure 2.4.2-3Figure 2.4.2-4Figure 2.4.2-5Figure 2.4.2-6PLO F -N uclear Power versus Tim e .....................................................................
2-186PLOF -Pressurizer Pressure versus Tim e ............................................................
2-187PLOF -Core Average Heat Flux versus Time ......................................................
2-188PLOF -Hot Channel Heat Flux versus Time ........................................................
2-189PLOF -Minimum DNBR versus Time .................................................................
2-190CLOF -Core Volumetric Flow Rate versus Time .............................................
2-191CLOF -Loop Volumetric Flow Rates versus Time ...........................................
2-192CLOF -Nuclear Power versus Time ...............................................................
2-193CLOF -Pressurizer Pressure versus Tim e ............................................................
2-194CLOF -Core Average Heat Flux versus Time ......................................................
2-195CLOF -Hot Channel Heat Flux versus Time ..................................................
2-196CLOF -Minimum DNBR versus Time .............................................................
2-197CLOF-UF -Core Volumetric Flow Rate versus Time ..........................................
2-198CLOF-UF -Loop Volumetric Flow Rates versus Time ........................................
2-199CLOF-UF -Nuclear Power versus Time ..............................................................
2-200CLOF-UF -Pressurizer Pressure versus Time ......................................................
2-201CLOF-UF -Core Average Heat Flux versus Time ...............................................
2-202CLOF-UF -Hot Channel Heat Flux versus Time .................................................
2-203CLOF-UF -Minimum DNBR versus Time ..........................................................
2-204RCP Locked Rotor/Shaft Break Overpressurization!PCT Case -CoreVolum etric Flow Rates versus Tim e ......................................................................
2-212RCP Locked Rotor/Shaft Break Overpressurization/PCT Case -FaultedLoop Volumetric Flow Rates versus Time ............................................................
2-213RCP Locked Rotor/Shaft Break Overpressurization!PCT Case -MaximumR C S Pressure versus Tim e ....................................................................................
2-214RCP Locked Rotor/Shaft Break Overpressurization!PCT Case -NuclearPow er versus T im e ................................................................................................
2-2 15RCP Locked Rotor/Shaft Break Overpressurization!PCT Case -Core HeatF lux versus T im e ...................................................................................................
2-2 16RCP Locked Rotor/Shaft Break Overpressurization/PCT Case -PCTv ersu s T im e ...........................................................................................................
2 -2 17WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3XVLIST OF FIGURES (cont.)Figure 2.4.2-7Figure 2.4.2-8Figure 2.4.2-9Figure 2.4.2-10Figure 2.4.2-11Figure 2.4.2-12Figure 2.5.1-1Figure 2.5.1-2Figure 2.5.1-3Figure 2.5.1-4Figure 2.5.2-1Figure 2.5.2-2Figure 2.5.2-3Figure 2.5.2-4Figure 2.5.2-5Figure 2.5.2-6Figure 2.5.2-7Figure 2.5.2-8Figure 2.5.2-9RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Core Volumetric FlowR ate versus T im e ...................................................................................................
2-2 18RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Loop Volumetric FlowR ates versus T im e ..................................................................................................
2-2 19RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Pressurizer Pressure versus T im e .............................................................................................
2-220RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Nuclear Powerv ersu s T im e ...........................................................................................................
2 -2 2 1RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Core Average HeatF lux versus T im e ...................................................................................................
2-222RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Hot Channel HeatF lux versus T im e ...................................................................................................
2-223Rod Withdrawal from Subcritical
-Nuclear Power Transient
..............................
2-230Rod Withdrawal from Subcritical
-Core Average Heat Flux Transient
...............
2-231Rod Withdrawal from Subcritical
-Fuel Average Temperature Transient
............
2-232Rod Withdrawal from Subcritical
-Cladding Surface Temperature Transient
..... 2-233Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -110 pcm/sec Nuclear Power and Core Heat Flux Versus Time ...............
2-243Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -110 pcm/sec Pressurizer Pressure and Water Volume Versus Time ........
2-244Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -110 pcm/sec Vessel Average Temperature and DNB R Versus Time ....... 2-245Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -1 pcm/sec Nuclear Power and Core Heat Flux Versus Time ...................
2-246Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -1 pcm/sec Pressurizer Pressure and Water Volume Versus Time ............
2-247Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -1 pcm/sec Vessel Average Temperature and DNBR Versus Time ...........
2-248Bank Withdrawal at Power -100 Percent Power Minimum DNBR VersusR eactivity Insertion R ate .......................................................................................
2-249Bank Withdrawal at Power -60 Percent Power Minimum DNBR VersusR eactivity Insertion R ate .......................................................................................
2-250Bank Withdrawal at Power -10 Percent Power Minimum DNBR VersusR eactivity Insertion R ate .......................................................................................
2-251WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xviLIST OF FIGURES (cont.)Figure 2.5.2-10Figure 2.5.2-11Figure 2.5.2-12Figure 2.5.6-1Figure 2.5.6-2Figure 2.5.6-3Figure 2.5.6-4Figure 2.6.1-1Figure 2.6.1-2Figure 2.6.1-3Figure 2.6.2-1Figure 2.6.2-2Figure 2.6.2-3Figure 2.6.2-4Figure 2.6.2-5Figure 2.6.2-6Figure 2.6.2-7Figure 2.6.2-8Figure 2.7.1-1Figure 2.7.1-2Figure 2.7.1-3Bank Withdrawal at Power -Minimum Reactivity Feedback LimitingOverpressure Case Nuclear Power and Core Heat Flux Versus Time ...................
2-252Bank Withdrawal at Power -Minimum Reactivity Feedback LimitingOverpressure Case Pressurizer Pressure and Water Volume Versus Time .............
2-253Bank Withdrawal at Power -Minimum Reactivity Feedback LimitingOverpressure Case Vessel Average Temperature and Peak RCS Pressure
............
2-254R od Ejection
-B O L/H Z P .....................................................................................
2-281R od Ejection
-B O L/H FP ......................................................................................
2-282R od Ejection
-EO L/H Z P ......................................................................................
2-283R od Ejection
-EO L/H FP ......................................................................................
2-284Inadvertent ECCS -Nuclear Power and Tavg versus Time ....................................
2-291Inadvertent ECCS -Pressurizer Pressure and Water Volume versus Time ...........
2-292Inadvertent ECCS -Total Steam Flow and Total Flow Injected to the RCSversu s T im e ...........................................................................................................
2-293CVCS Malfunction, Maximum Reactivity
: Feedback, With Pressurizer SprayN uclear Power and Tavg versus Tim e .....................................................................
2-299CVCS Malfunction, Maximum Reactivity
: Feedback, With Pressurizer Spray Pressurizer Pressure and Water Volume versus Time ..................................
2-300CVCS Malfunction, Maximum Reactivity
: Feedback, Without Pressurizer Spray Nuclear Power and Tavg versus Tim e ...........................................................
2-301CVCS Malfunction, Maximum Reactivity
: Feedback, Without Pressurizer Spray Pressurizer Pressure and Water Volume versus Time ..................................
2-302CVCS Malfunction, Minimum Reactivity
: Feedback, With Pressurizer SprayN uclear Power and Tavg versus Tim e .....................................................................
2-303CVCS Malfunction, Minimum Reactivity
: Feedback, With Pressurizer SprayPressurizer Pressure and Water Volume versus Time ............................................
2-304CVCS Malfunction, Minimum Reactivity
: Feedback, Without Pressurizer Spray Nuclear Power and Tavg versus Time ...........................................................
2-305CVCS Malfunction, Minimum Reactivity
: Feedback, Without Pressurizer Spray Pressurizer Pressure and Water Volume versus Time ..................................
2-306RCS Depressurization
-Nuclear Power versus Time ...........................................
2-311RCS Depressurization
-Pressurizer Pressure versus Time ...................................
2-311RCS Depressurization
-Indicated Loop Average Temperature versus Time ........
2-312WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xviiWESTINGHOUSE NON-PROPRIETARY CLASS 3 XVIILIST OF FIGURES (cont.)Figure 2.7.1-4Figure 2.7.2-1Figure 2.7.2-2Figure 2.7.2-3Figure 2.7.2-4Figure 2.7.2-5Figure 2.7.2-6Figure 2.7.2-7Figure 2.7.2-8Figure 2.7.3-1Figure 2.7.3-2Figure 2.7.3-3Figure 2.7.3-4Figure 2.7.3-5Figure 2.7.3-6Figure 2.7.3-7Figure 2.7.3-8Figure 2.7.3-9Figure 2.7.3-10Figure 2.7.3-11Figure 2.7.4-1Figure 2.7.4-2Figure 2.7.4-3Figure 2.7.4-4Figure 2.7.4-5RCS Depressurization
-DNBR versus Time ........................................................
2-312Pressurizer Level -M argin to Overfill Analysis
...................................................
2-326Pressurizer Pressure
-Margin to Overfill Analysis
...............................................
2-327Secondary Pressure
-Margin to Overfill Analysis
...............................................
2-328Primary to Secondary Break Flow -Margin to Overfill Analysis
........................
2-329SG Water Volumes -Margin to Overfill Analysis
................................................
2-330SG Steam Releases
-Margin to Overfill Analysis
................................................
2-331Ruptured Loop RCS Temperature
-Margin to Overfill Analysis
.........................
2-332Intact Loops RCS Temperature
-Margin to Overfill Analysis
.............................
2-333Pressurizer Level -Input to Radiological Consequences Analysis
.......................
2-344Pressurizer Pressure
-Input to Radiological Consequences Analysis
..................
2-345Secondary Pressure
-Input to Radiological Consequences Analysis
...................
2-346Primary to Secondary Break Flow -Input to Radiological Consequences A n a ly sis .................................................................................................................
2 -34 7SG Steam Releases
-Input to Radiological Consequences Analysis
...................
2-348Ruptured Loop HL and CL Temperatures
-Input to Radiological C onsequences A nalysis .........................................................................................
2-349Intact Loop HL and CL Temperatures
-Input to Radiological Consequences A n a ly sis .................................................................................................................
2 -3 5 0Break Flow Flashing Fraction
-Input to Radiological Consequences A n a ly sis .................................................................................................................
2 -3 5 1Integrated Flashed Break Flow -Input to Radiological Consequences A n a ly sis .................................................................................................................
2 -3 5 2Ruptured SG Fluid Mass -Input to Radiological Consequences Analysis
...........
2-353Ruptured SG Water Volume -Input to Radiological Consequences Analysis
...... 2-354Post-LOCA Subcriticality Boron Limit Curve ......................................................
2-364B oric A cid Solubility L im it ...................................................................................
2-365LBLOCA Boric Acid Concentration Analysis
-Vessel Boric AcidConcentration, Boil-off, and Flushing Flow versus Time .....................................
2-366SBLOCA Boric Acid Concentration Analysis
-Vessel Boric AcidConcentration, Boil-off, and Flushing Flow versus Time .....................................
2-367Core Dilution at 12 Hours for SBLOCA Pressure Hangup ...................................
2-368WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xviiiFigure 2.8.1-1Figure 2.8.1-2Figure 2.8.1-3Figure 2.8.1-4Figure 2.8.1-5Figure 2.8.1-6Figure 2.8.1-7Figure 2.8.1-8Figure 2.8.1-9Figure 2.8.1-10Figure 2.8.1-11Figure 2.8.1-12Figure 2.8.1-13Figure 2.8.1-14Figure 2.8.1-15Figure 2.8.1-16LIST OF FIGURES (cont.)Nuclear Power versus Time for LOL ATWS .........................................................
2-373Core Heat Flux versus Time for LOL ATWS ........................................................
2-373RCS Pressure versus Time for LOL ATWS ...........................................................
2-374Pressurizer Water Volume versus Time for LOL ATWS .......................................
2-374Vessel Inlet Temperature versus Time for LOL ATWS .........................................
2-375RCS Flow versus Time for LOL ATWS ................................................................
2-375SG Pressure versus Time for LOL ATWS .............................................................
2-376SG M ass versus Tim e for LOL ATW S ..................................................................
2-376Nuclear Power versus Time for LONF ATWS ......................................................
2-377Core Heat Flux versus Time for LONF ATWS .....................................................
2-377RCS Pressure versus Time for LONF ATWS ........................................................
2-378Pressurizer Water Volume versus Time for LONF ATWS ....................................
2-378Vessel Inlet Temperature versus Time for LONF ATWS ......................................
2-379RCS Flow versus Time for LONF ATWS .............................................................
2-379SG Pressure versus Time for LONF ATWS ..........................................................
2-380SG Mass versus Time for LONF ATWS ...............................................................
2-380WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xixWESTINGHOUSE NON-PROPRIETARY CLASS 3 xixLIST OF ACRONYMSACAECAFWAMSACANSANSIAOOsAORARVASME B&PVASTATWSalternating currentAtomic Energy Commission auxiliary feedwater ATWS mitigation system actuation circuitry American Nuclear SocietyAmerican National Standards Institute anticipated operational occurrences analysis of recordatmospheric relief valveAmerican Society of Mechanical Engineers Boiler & Pressure VesselAlternative Source Termanticipated transient without scramBAPCIpeffBOCBOLCCPCDSACFRCHFCLCLOFCOLRCRDMCSTCVCSDCDHRDNBDNBRDPCDTCECCSEOCEOLEOPESFASFCVFNAHFOIFONFQboric acid precipitation controleffective delayed neutron fractionbeginning of cyclebeginning of lifecentrifugal charging pumpcore design and safety analysesCode Federal Regulations critical heat fluxcold legcomplete loss of flowCore Operating Limits Reportcontrol rod drive mechanism condensate storage tankchemical and volume control systemdirect currentdecay heat removaldeparture from nucleate boilingdeparture from nucleate boiling ratioDoppler power coefficient Doppler temperature coefficient emergency core cooling systemend of cycleend of lifeemergency operating procedure engineered safety features actuation systemflow control valveradial peaking factorfraction of initialfraction of nominaltotal peaking factorWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xxLIST OF ACRONYMS (cont.)FW feedwater FWI feedwater isolation FWM feedwater malfunction GDC general design criterion HFP hot full powerHL hot legHLSO hot leg switchover HZP hot zero powerIFM intermediate flow mixing vanesITC isothermal temperature coefficienteffective multiplication factorLBLOCA large-break LOCALCO limiting condition for operation LOAC loss of non-emergency ACLOCA loss-of-coolant accidentLOL loss of loadLOL/TT loss of load/turbine tripLONF loss of normal feedwater LOOP loss of offsite powerLTC long-term coolingMDAFW motor-driven AFWMMF minimum measured flowMSIV main steam isolation valveMSS main steam systemMSSV main steam safety valveMTC moderator temperature coefficient MUR measurement uncertainty recapture NRS narrow-range spanNSSS nuclear steam supply systemOTAT overtemperature ATOPAT overpower ATpcm percent millirhoPCT peak cladding temperature PLOF partial loss of flowPORV power operated relief valveppm part per millionPSV pressurizer safety valvePWR pressurized water reactorRCCA rod cluster control assemblyRCP reactor coolant pumpRCPB reactor coolant pressure boundaryRCS reactor coolant systemWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xxiLIST OF ACRONYMS (cont.)RFA-2 Robust Fuel Assembly-2 RG Regulatory GuideRHR residual heat removalRHRS residual heat removal systemRMWS reactor makeup water systemRPS reactor protection systemRPV reactor pressure vesselRSE Reload Safety Evaluation RT reactor tripRTD resistance temperature detectors RTDP Revised Thermal Design Procedure RTP rated thermal powerRTS reactor trip systemRWAP rod withdrawal at powerRWST refueling water storage tankSAFDLs specified acceptable fuel design limitsSAL safety analysis limitSBLOCA small-break LOCASER Safety Evaluation ReportSG steam generator SGTP steam generator tube pluggingSGTR steam generator tube ruptureSI safety injection SLB steam line breakSLI steam line isolation STDP Standard Thermal Design Procedure Tavg (reactor) vessel average temperature TCD thermal conductivity degradation T/H thermal-hydraulic TDAFW turbine-driven AFWTDF thermal design flowTreed feedwater temperature TM Transition of MethodsTPI thimble plugs installed TPR thimble plugs removedTS Technical Specification TT turbine trip (Section 2.1 Tables only)UF underfrequency USAR Updated Safety Analysis ReportUSNRC United States Nuclear Regulatory Commission UV undervoltage VCT volume control tankWCAP Westinghouse Commercial Atomic Power (topical report)WCGS Wolf Creek Generating StationWCNOC Wolf Creek Nuclear Operating Corporation WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3xxiiPROFESSIONAL ENGINEERING STAMPSThis section contains the State of Kansas Professional Engineer certifications for each of the sectionspertaining to technical services scope supporting the Wolf Creek Generating Station plant design ordesign configuration.
Each Professional Engineer has designated applicable scope sections for whichthey provided Practice of Engineering oversight and for which their certification applies.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 31, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:
Section 1, Section 1 .1Certified By: Melanie Rose Fici, P.E.License Number: 21873State: KS Expiration Date: 4/30/2014 WCAP-1 7658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:
Section 2.1, Section 2.2.1, Section 2.2.2, Section 2.2.3, Section 2.3.2,Section 2.3.3, Section 2.3.4, Section 2.8, Appendix A Table A. l-1 Item No. 1 through 7, Appendix ATables A.2-1, A.3-1, and A.4-1Certified By: James A. StewartLicense Number: 21814State: KS Expiration Date: 04/30/2015 WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief theresults herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:
Section 2.2.4, Section 2.2.5, Section 2.3.1, Section 2.6.1, Section 2.6.2Certified By: William D. Higby, P.E.License Number: 22319 State: KS Expiration Date: 04/30/2014
,,,,WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:
Section 2.5.1, Section 2.5.3, Section 2.5.4, Section 2.5.5, Section 2.5.6Certified By: Chris J. McHughLicense Number: 21957State: KS Expiration Date: 04/30/15WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:
Section 2.4.1, Section 2.4.2, Section 2.5.2, Section 2.7.1Certified By: Andrew R Detar, P.E.License Number: 22109State: KS Expiration Date: April 30, 2014WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:
Section 2.7.2, Section 2.7.3, Section 2.9, Appendix A Table A. 1-1 Item No. 8Certified By: John C Reck, P.E.License Number: 21960State: KS Expiration Date: 2015-04-30 WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 31, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Section being Certified:
Section 2.7.4Certified By: David J. Fink, P.E.License Number: 18714 State: KSExpiration Date: 04-30-2014 WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Section being Certified:
Section 2.10Certified By: Ryan Paul Rossman, P.E.License Number: 18724State: KS Expiration Date: April 30, 2015WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional
: Engineer, certify that to the best of myknowledge and belief the results herein do not jeopardize the protection of life, health, property, and welfare of the public.Section being Certified:
Section 2.11Certified By: Shamsul M. AbedinLicense Number: 21983State: KS Expiration Date: April 30, 2014WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional
: Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:
Section 2.12, Appendix A Tables A.5-1 and A.6-1Certified By: Kevin Regis McAtee, P.E.License Number: 18582State: KS Expiration Date: 2015-04-30 WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 31-11 INTRODUCTION General OverviewThis Transition of Methods (TM) for Core Design and Safety Analyses (CDSA) licensing report isprovided by Wolf Creek Nuclear Operating Corporation (WCNOC) in support of the TM licenseamendment application for the Wolf Creek Generating Station (WCGS). WCNOC plans to transition fromits current methodology for performing core design, non-loss-of-coolant-accident (non-LOCA) andLOCA safety analyses (Post-LOCA Subcriticality and Cooling) to the Westinghouse methodologies forperforming these analyses.
Westinghouse currently holds the analysis of record (AOR) for both the small-break (SB) andlarge-break (LB) LOCA; therefore SBLOCA and LBLOCA are not included in the transition effort.For safety analyses that were reanalyzed, they were conservatively reanalyzed at the higher nominalpower level associated with a Measurement Uncertainty Recapture (MUR) power uprate. The reanalysis effort did not assume any other plant or analysis input changes that may be required to support an actualMUR power uprate. Also, the core design effort did not assume any other plant or analysis input changesthat may be required to support an actual MUR power uprate.Note that even though some analyses were performed at an uprated power (representative of an MUR),the MUR conditions (i.e., NSSS power) would be bounding for plant operation at current rated thermalpower (RTP).It is not the intent of this licensing amendment application to request approval of an MIUR power uprate.This document addresses the transition to the approved Westinghouse methodologies only.This report summarizes the analyses that were perforned to confirm that applicable acceptance criteriaare met. Sections 2.0 through 2.12 of this TM CDSA licensing report provide the results of the accidentanalyses and core design efforts.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 31-21.1 NUCLEAR STEAM SUPPLY SYSTEM PARAMETERS
====1.1.1 Introduction====
The nuclear steam supply system (NSSS) design parameters are the fundamental parameters used as inputin all of the NSSS accident analyses.
The current WCGS NSSS design parameters are summarized inTable 5,1- 1 of the WCGS Updated Safety Analysis Report (USAR). The NSSS design parameters providethe primary and secondary side system conditions (thermal power, temperatures, pressures, and flows)that serve as the basis for all of the NSSS analyses and evaluations.
As a result of the TM Program, theWCGS NSSS design parameters have been revised, as shown in Tables 1.1-1 and 1.1-2. Tables 1.1-1and 1.1-2 provide information for the eight cases associated with the TM Program at current power andMUR Uprate conditions, respectively.
These parameters have been incorporated, as appropriate, into theapplicable CDSA, performed in support of the TM Program.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe NSSS design parameters provide the reactor coolant system (RCS) and secondary system conditions (thermal power, temperatures, pressures, and flows) that are used as the basis for the NSSS designtransients,
: systems, structures, components,
: accident, and fuel analyses and evaluations.
For the TM Program at the current licensed power level, the established major input parameters andassumptions used to calculate the four cases of NSSS design parameters are summarized as follows:1. The parameters are based on Westinghouse Model F steam generators (SGs).2. The NSSS power level of 3579 MWt (3565 MWt reactor core power + 14 MWt net heat input)was assumed.3. A nominal feedwater temperature (Tfeed) range of 400.0'F to 446.0°F was selected.
: 4. Two design core bypass flows were used: 8.4 percent, which accounts for fuel with thimble plugsremoved (TPR) and intermediate flow mixing vanes (IFMs); and 6.4 percent, which accounts forfuel with thimble plugs installed (TPI) and IFMs.5. A thermal design flow (TDF) of 90,300 gpm/loop was assumed based on a TDF of90,324 gpm/loop rounded to the nearest hundred gpm/loop.
: 6. A full-power normal operating vessel average temperature (Tavg) range of 570.7°F to 588.4°F wasassumed.
This provides the basis for the WCGS to operate within this window. Any exceptions tothese values will be addressed in the affected sections.
: 7. Steam generator tube plugging (SGTP) levels of 0 and 10 percent were assumed.8. A maximum SG moisture carryover of 0.25 percent was utilized.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 31-3For the TM Program at the MUR Uprate power level, the established major input parameters andassumptions used to calculate the four cases of NSSS design parameters are summarized as follows:1. The parameters are based on Westinghouse Model F SGs.2. An uprated NSSS power level of 3651 MWt (3637 MWt reactor core power + 14 MWt net heatinput) was assumed for MUR Uprate conditions.
: 3. A nominal Tfeed range of 400.0°F to 448.6°F was selected.
: 4. Two design core bypass flows were used: 8.4 percent, which accounts for fuel with TPR andIFMs; and 6.4 percent, which accounts for fuel with TPI and IFMs.5. A TDF of 90,300 gpm/loop was assumed based on a TDF of 90,324 gpm/loop rounded to thenearest hundred gpm/loop.
: 6. A full-power normal operating Tavg range of 570.7°F to 588.4°F was assumed.
This provides thebasis for the WCGS to operate within this window. Any exceptions to these values will beaddressed in the affected sections.
: 7. SGTP levels of 0 and 10 percent were assumed.8. A maximum SG moisture carryover of 0.25 percent was utilized.
Acceptance CriteriaThe acceptance criteria for determining the NSSS design parameters were that the results of the accidentanalyses and evaluations continue to comply with all WCGS applicable industry and regulatory requirements, and that they provide WCGS with adequate flexibility and margin during plant operation.
1.1.3 Description of Analyses and Evaluation Table 1.1-1 provides the NSSS design parameter cases that were generated and serve as the WCGS basisfor the analyses considering the current licensed power level. These cases are as follows:* Cases 1 and 2 of Table 1.1-1 represent parameters based on a Tavg of 570.7°F.
Case 2, which isbased on an average 10 percent SGTP, yields the minimum secondary side SG pressure andtemperature.
Note that all primary side temperatures are identical for these two cases.* Cases 3 and 4 of Table 1.1-1 represent parameters based on the Tavg of 588.4°F.
Case 3, which isbased on 0 percent SGTP, yields the higher secondary side SG pressure performance conditions.
Note that all primary side temperatures are identical for these two cases. As provided viafootnote 4 of Table 1.1-1, for instances where an absolute upper limit SG outlet pressure isconservative for any analyses, these data are based on the Case 3 parameters with 0 percent SGTPand also assume a SG fouling factor of 0 hr-ft2-OF/BTU.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 31-4Table 1. 1-2 provides the NSSS design parameter cases that were generated and serve as the WCGS basisfor the analyses considering MUR Uprate conditions.
These cases are as follows:Cases 1 and 2 of Table 1.1-2 represent parameters based on a Ta,,g of 570.7°F.
Case 2, which isbased on an average 10 percent SGTP, yields the minimum secondary side SG pressure andtemperature.
Note that all primary side temperatures are identical for these two cases.Cases 3 and 4 of Table 1.1-2 represent parameters based on the Tavg of 588.4°F.
Case 3, which isbased on 0 percent SGTP, yields the highest secondary side SG pressure performance conditions.
Note that all primary side temperatures are identical for these two cases. As provided viafootnote 4 of Table 1.1-2, for instances where an absolute upper limit SG outlet pressure isconservative for any analyses, these data are based on the Case 3 parameters with 0 percent SGTPand also assume a SG fouling factor of 0 hr-ft2-°F/BTU.1.1.4 Conclusions The resulting NSSS design parameters (Tables 1.1-1 and 1. 1-2) were used by Westinghouse as the basisfor CDSA efforts.
Westinghouse performed the analyses and evaluations based on the parameter sets thatwere most limiting, so that the analyses would support operation over the entire range of conditions specified.
In cases where the analyses performed do not bound the entire range of conditions specified (such as a restricted Tavg operating range), the applicable report section identifies the range of conditions analyzed for the TM Program.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 31-5Table 1.1-1 NSSS Design Parameters for WCGS TM ProgramCurrent Power -Safety Analysis OnlyThermal Design Parameters Case I Case 2 Case 3 Case 4NSSS Power, MWt 3579 3579 3579 3579106 Btu/hr 12,212 12,212 12,212 12,212Reactor Power, MWt 3565 3565 3565 3565106 Btu/hr 12,164 12,164 12,164 12,164Thermal Design Flow, gpm/loop 90,300 90,300 90,300 90,300Reactor 106 lb/hr 138.2 138.2 134.7 134.7Reactor Coolant Pressure, psia 2250 2250 2250 2250Core Bypass, % 8.4(1,2) 8.4( 2) 8.4 1,3) 8.4(1,3Reactor Coolant Temperature, OFCore Outlet 609.8(2) 609.8(2) 626.2(3) 626.213)Vessel Outlet 604.3 604.3 621.0 621.0Core Average 575.3(2) 575.3(2) 593.2(3) 593.2(')Vessel Average 570.7 570.7 588.4 588.4Vessel/Core Inlet 537.1 537.1 555.8 555.8Steam Generator Outlet 536.8 536.8 555.5 555.5Steam Generator Steam Outlet Temperature.,
F 520.8 518.3 539.9(4) 537.5Steam Outlet Pressure, psia 818 801 962(4) 943Steam Outlet Flow, 106 lb/hr total 14.86/15.83 14.86/15.82 14.95/15.93(4) 14.94/15.91 Feed Temperature, OF 400.0/446.0 400.0/446.0 400.0/446.0 400.0/446.0 Steam Outlet Moisture,
% max. 0.25 0.25 0.25 0.25Tube Plugging Level, % 0 10 0 10Zero-Load Temperature, OF 557 557 557 557Hydraulic Design Parameters Mechanical Design Flow, gpm/loop 104,200Minimum Measured Flow, gpm total 371,000Notes:I. Core bypass flow accounts for TPR and IFMs.2. If thimble plugs are installed, the core bypass flow is 6.4%, core outlet temperature is 608.4'F, and core averagetemperature is 574.5'F.3. If thimble plugs are installed, the core bypass flow is 6.4%, core outlet temperature is 624.9'F, and core averagetemperature is 592.5'F.4. Where appropriate for NSSS analyses, a greater steam outlet pressure of 984 psia, steam outlet temperature of 542.6°Fand total steam outlet flow of 15.94 x 106 lb/hr may be assumed.
This envelops the possibility that the plant could operatewith more efficient SG performance.
WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 31-6Table 1.1-2 NSSS Design Parameters for WCGS TM ProgramMUR Uprate Power -Safety Analysis OnlyThermal Design Parameters Case 1 Case 2 Case 3 Case 4NSSS Power, MWt 3651 3651 3651 3651106 Btu/hr 12,458 12,458 12,458 12,458Reactor Power, MWt 3637 3637 3637 3637106 Btu/hr 12,410 12,410 12,410 12,410Thermal Design Flow, gpm/loop 90,300 90,300 90,300 90,300Reactor 106 lb/hr 138.3 138.3 134.9 134.9Reactor Coolant Pressure, psia 2250 2250 2250 2250Core Bypass, % 8.4(1,2) 8.4'1,2) 8.4(13) 8.4(1.3)Reactor Coolant Temperature, IFCore Outlet 610.5 2) 610.512) 626.9"'1 626.9(3)Vessel Outlet 604.9 604.9 621.6 621.6Core Average 575.4)2 575.42) 593.33) 593.3")Vessel Average 570.7 570.7 588.4 588.4Vessel/Core Inlet 536.5 536.5 555.2 555.2Steam Generator Outlet 536.2 536.2 554.9 554.9Steam Generator Steam Outlet Temperature, IF 519.7 517.2 538.914) 536.4Steam Outlet Pressure, psia 810 793 954(4) 934Steam Outlet Flow, 106 lb/hr total 15.16/16.21 15.15/16.20 15.24/16.304.)
15.23/16.29 Feed Temperature, IF 400.0/448.6 400.0/448.6 400.0/448.6 400.0/448.6 Steam Outlet Moisture,
% max. 0.25 0.25 0.25 0.25Tube Plugging Level, % 0 10 0 10Zero Load Temperature, IF 557 557 557 557Hydraulic Design Parameters Mechanical Design Flow, gpm/Ioop 104,200Minimum Measured Flow, gpm total 371,000Notes:I. Core bypass flow accounts for TPR and IFMs.2. If thimble plugs are installed, the core bypass flow is 6.4%, core outlet temperature is 609. I°F, and core averagetemperature is 574.6°F.3. If thimble plugs are installed, the core bypass flow is 6.4%, core outlet temperature is 625.6°F, and core averagetemperature is 592.6°F.4. Where appropriate for NSSS analyses, a greater steam outlet pressure of 976 psia, steam outlet temperature of 541.6°Fand total steam outlet flow of 16.32 x 106 lb/hr may be assumed.
This envelops the possibility that the plant could operatewith more efficient SG performance.
WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-I2 ACCIDENT AND TRANSIENT ANALYSIS2.1 NON-LOCA ANALYSES INTRODUCTION Chapter 15, "Accident Analysis,"
of the WCGS USAR (Reference
: 1) identifies the non-LOCA transient events that have been analyzed as part of the current WCGS licensing basis. In support of the TMProgram for the WCGS, most of the non-LOCA licensing basis events have been reanalyzed usingWestinghouse Electric Company safety analysis methods previously approved by the United StatesNuclear Regulatory Commission (USNRC).
The non-LOCA events summarized in this introduction section are those discussed in greater detail in Sections 2.2 through 2.7. 1, as well as the Anticipated Transient without Scram (ATWS) event discussed in Section 2.8. Other non-LOCA events, i.e., steamgenerator tube rupture (SGTR), are discussed elsewhere in this report.2.1.1 Program FeaturesKey features of the TM Program that were considered in the non-LOCA transient analyses are as follows.0 A NSSS power level of 3651 MWt, which includes all applicable uncertainties and a nominalreactor coolant pump (RCP) net heat input of 14 MWt (or 20 MWt for events where higher RCPheat is conservative)
* Westinghouse 17x 17 Robust Fuel Assembly (RFA-2) fuel design with IFMs and thimble plugseither removed or installed (see Note below)0 A nominal, full-power Tavg window of 570.7°F to 588.4°F0 A RCS TDF of 361,200 gpm (90,300 gpm/loop),
and a minimum measured flow (MMF) of376,000 gpm (94,000 gpm/loop)
As indicated in Table 2.1-2, a bounding MMF value of 371,000 gpm (92,750 gpm/loop) wasapplied in all but one analysis where MMF is used as the RCS flow.* Westinghouse Model F SGs, with a maximum SGTP level of 10 percent0 A nominal, full-power main Treed window of 400'F to 448.6°F0 A nominal operating pressurizer pressure of 2250 psia0 A design core bypass flow of 8.4 percent and a statistical core bypass flow of 6.61 percent,conservatively corresponding to having the core TPR (see Note below)Whereas the statistical core bypass flow is used in some departure from nucleate boilingratio (DNBR) analyses, the design core bypass flow is used for all other non-LOCA analyses; see Section 2.1.5, "Initial Conditions,"
for additional details.Note: Except for the limiting DNBR analysis of the uncontrolled rod cluster control assembly (RCCA)bank withdrawal at power event, all analyses covered the bounding scenario of having the coreTPR. As a result of the exception, the plant may be required to operate with the core TPI.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-22.1.2 Non-LOCA Transient Events Considered The non-LOCA transient events considered in support of the TM Program for the WCGS are identified inthe plant condition classification discussion presented below. As noted at the beginning of Section 2. 1, thenon-LOCA events discussed in this section are a subset of the non-LOCA licensing basis events.Plant Condition Classification The American Nuclear Society (ANS) Standard ANS-51.1-1973 (ANSI-N 18.2) (Reference
: 2) providesclassification of plant conditions that are divided into four categories based on the anticipated frequency of occurrence and the potential radiological consequences to the public. The four categories, orconditions, are:* Condition I -Normal Operation and Operational Transients
* Condition II -Faults of Moderate Frequency
* Condition III -Infrequent Faults* Condition IV -Limiting FaultsThe basic principle applied in relating design requirements to each of the conditions is that the mostprobable occurrences should yield the least 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. Whereapplicable, and to the extent allowed, the reactor trip system (RTS) and/or engineered safeguards featuresare applied in fulfilling this principle.
Each condition is described in more detail as follows.Condition I -Normal Operation and Operational Transients Condition I occurrences are those that are expected frequently or regularly during power operation, refueling, maintenance, or maneuvering of the plant. Condition I occurrences are accommodated withmargin between any plant parameter and the value of the parameter that would require either automatic ormanual protective action. As Condition I events occur frequently, they must be considered from the pointof view of their effect on the consequences of fault conditions (Conditions II, III, and IV). In this regard,analysis of each fault condition described is generally based on a conservative set of initial conditions corresponding to adverse conditions that can occur during Condition I operation.
A typical list ofCondition I events is given below.Steady state and shutdown operations
-Power operation
-Startup-Hot standby-Hot shutdown-Cold shutdown-Refueling WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3Operation with permissible deviations Various deviations from nonnal operation but specifically allowed by the Technical Specifications (TS) that may occur during continued operation are considered in conjunction withother operational modes. These include:-Operation with components or systems out of service (such as an inoperable RCCA)-Leakage from fuel with limited clad defects-Excessive radioactivity in the reactor coolant" Fission products* Corrosion products" Tritium-Operation with SG leaks-Testing* Operational transients
-Plant heatup and cooldown-Step load changes (up to +/- 10 percent)-Ramp load changes (up to 5 percent per minute)-Load rejection up to and including design full-load rejection transient Condition II -Faults of Moderate Frequency Condition II faults (or events) occur with moderate frequency during the life of the plant, any one ofwhich may occur during a calendar year. These events, at worst, result in a reactor trip (RT) with the plantbeing capable of returning to operation after corrective action. A Condition II event, by itself, does notpropagate to a more serious event of the Condition III or Condition IV type without the occurrence ofother independent incidents.
In addition, Condition II events should not cause the loss of any barrier to theescape of radioactive products.
The following list identifies the Condition II non-LOCA events considered herein in support of the TM Program for the WCGS.* Feedwater (FW) system malfunctions that result in a decrease in Treed(USAR Section 15.1.1)* FW system malfunctions that result in an increase in FW flow(USAR Section 15.1.2)* Excessive increase in secondary steam flow (USAR Section 15.1.3)WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-4WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-4* Inadvertent opening of a SG atmospheric relief or safety valve(USAR Section 15.1.4)0 Loss of external electrical load (USAR Section 15.2.2)0 Turbine trip (USAR Section 15.2.3)* Inadvertent closure of main steam isolation valves (MSIVs) (USAR Section 15.2.4)* Loss of condenser vacuum and other events resulting in turbine trip (USAR Section 15.2.5)* Loss of non-emergency AC power to the station auxiliaries (USAR Section 15.2.6)* Loss of normal FW flow (USAR Section 15.2.7)* Partial loss of forced reactor coolant flow (USAR Section 15.3.1)* Uncontrolled RCCA bank withdrawal from a subcritical or low power startup condition (USAR Section 15.4. 1)* Uncontrolled RCCA bank withdrawal at power (USAR Section 15.4.2)* RCCA misoperation (dropped RCCA, dropped RCCA bank, and statically misaligned RCCA)(USAR Section 15.4.3)* Startup of an inactive RCP at an incorrect temperature (USAR Section 15.4.4)* Chemical and volume control system (CVCS) malfunction that results in a decrease in the boronconcentration in the reactor coolant (boron dilution)
(USAR Section 15.4.6)* Inadvertent operation of the emergency core cooling system (ECCS) during power operation (USAR Section 15.5.1)* CVCS malfunction that increases reactor coolant inventory (USAR Section 15.5.2)* Inadvertent opening of a pressurizer safety or relief valve (USAR Section 15.6.1)Condition III -Infrequent FaultsCondition III events occur very infrequently during the life of the plant, any one of which may occurduring the plant's lifetime.
Condition III events can be accommodated with the failure of only a smallfraction of the fuel rods, although sufficient fuel damage might occur to preclude resumption of operation for a considerable outage time. The release of radioactivity due to a Condition III event will not besufficient to interrupt or restrict public use of those areas beyond the exclusion area boundary.
ACondition III event does not, by itself, generate a Condition IV event or result in a consequential loss ofWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-5WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-5function of the RCS or containment barriers.
The following list identifies the Condition III non-LOCAevents considered herein in support of the TM Program for the WCGS.* Steam system piping failure (minor) (USAR Section 15.1.5)* Complete loss of forced reactor coolant flow (USAR Section 15.3.2)* RCCA misoperation (withdrawal of a single RCCA) (USAR Section 15.4.3)Condition IV -Limiting FaultsCondition IV events are not expected to occur, but are postulated because their consequences have thepotential for the release of significant amounts of radioactive material.
Condition IV events are the mostdrastic occurrences that must be designed
: against, and represent the limiting design cases. Condition IVevents should not cause a fission product release to the environment resulting in an undue risk to publichealth and safety in excess of the guideline values in 10 CFR 100 (Code of Federal Regulations).
A singleCondition IV event shall not cause a consequential loss of required functions of the systems needed tocope with the event, including those of the ECCS and the reactor containment system. The following listidentifies the Condition IV non-LOCA events considered herein in support of the TM Program for theWCGS.* Steam system piping failure (major) (USAR Section 15.1.5)* FW system pipe break (USAR Section 15.2.8)* RCP shaft seizure (locked rotor) (USAR Section 15.3.3)* RCP shaft break (USAR Section 15.3.4)0 Spectrum of RCCA ejection accidents (USAR Section 15.4.8)Summary of Non-LOCA Events Considered Table 2.1-1 presents a list of all the non-LOCA transient events that were considered in support of the TMProgram for the WCGS to which this introductory discussion applies.
Also included in Table 2.1-1 arecross references to the applicable USAR sections, cross references to the event-specific sections withinthis report, and assertions as to which events were analyzed versus evaluated.
2.1.3 Analysis Methodology The transient-specific analysis methodologies that were applied in analyzing the non-LOCA transient events have been reviewed and approved by the USNRC via transient-specific topical reports,e.g., WCAPs, and/or through the review and approval of various licensing amendment request submittals for the WCGS or other plants. The following non-LOCA transients analyzed for the WCGS have anapproved transient-specific topical report, and each topical report is identified and discussed below.* Steam system piping failure (steam line break (SLB)) (USAR Sections 15.1.4 and 15.1.5)* Dropped RCCA/dropped RCCA bank (dropped rod) (USAR Section 15.4.3)* RCCA ejection (USAR Section 15.4.8)WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-6WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-6Steam Line Break Analysis Methodology The SLB licensing topical report, WCAP-9226-P-A Revision 1 (Reference 3), was approved by theUSNRC via a safety evaluation report (SER) from A. C. Thadani (NRC) to W. J. Johnson (Westinghouse),
dated January 31, 1989. The SLB SER identifies two conditions of acceptance, which are summarized below, along with justification for application to the WCGS.1. "Only those codes which have been accepted by the USNRC should be used for licensing application."
Justification As identified in Table 2.1-2, the computer codes used in the analysis of the SLB event areRETRAN, ANC, and VIPRE. These computer codes are discussed in Section 2.1.4, "Computer Codes Used," and it is confirmed that these codes have been accepted by the USNRC. Therefore, this condition of acceptance is satisfied for the WCGS.2. "For the pressure between 500 and 1000 psia, the 95/95 DNBR limit for the W-3 correlation is 1.45."Justification As discussed in Section 2.12, "Thermal and Hydraulic Design,"
the W-3 DNB correlation hasbeen replaced with the WLOP DNB (departure from nucleate boiling) correlation, which has adifferent 95/95 DNBR limit. Table 2.1-6 presents the DNBR safety analysis limit (SAL) appliedin the SLB analysis for which the WLOP DNB correlation was used. No further justification isrequired for the WCGS.Dropped Rod Analysis Methodology The dropped rod licensing topical report, WCAP-1 1394-P-A (Reference 4), was approved by the USNRCvia an SER from A. C. Thadani (NRC) to R. A. Newton (Westinghouse Owners Group), datedOctober 23, 1989. The dropped rod SER identifies one condition of acceptance, which is summarized below along with justification for application to the WCGS.1. "The Westinghouse
: analysis, results, and comparisons are reactor and cycle specific.
No credit istaken for any direct RT due to dropped RCCA(s).
Also, the analysis assumes no automatic powerreduction features are actuated by the dropped RCCA(s).
A further review by the staff (for eachcycle) is not necessary, given the utility assertion that the analysis described by Westinghouse hasbeen performed and the required comparisons have been made with favorable results."
Justification For the reference cycle assumed in the WCGS TM Program, the methodology described inWCAP-1 1394-P-A was applied and the required comparisons have been made with acceptable results (DNBR remains greater than the limit). Future fuel cycles will be assessed as part of theReload Safety Evaluation (RSE) process described in Reference 7.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-7WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-7RCCA Ejection Analysis Methodology The RCCA ejection licensing topical report, WCAP-7588 Revision 1-A (Reference 5), was approved bythe Atomic Energy Commission (AEC) via an SER from D. B. Vassallo (AEC) to R. Salvatori (Westinghouse),
dated August 28, 1973. The RCCA ejection SER identifies two conditions of acceptance, which are summarized below, along with justification for application to the WCGS.1. "The staff position, as well as that of the reactor vendors over the last several years, has been tolimit the average fuel pellet enthalpy at the hot spot following a rod ejection accident to280 cal/gm. This was based primarily on the results of the SPERT tests, which showed that, ingeneral, fuel failure consequences for U02 have been insignificant below 300 cal/gm for bothirradiated and unirradiated fuel rods as far as rapid fragmentation and dispersal of fuel andcladding into the coolant are concerned.
In this report, Westinghouse has decreased their limitingfuel failure criterion from 280 cal/gm (somewhat less than the threshold of significant conversion of the fuel thermal energy to mechanical energy) to 225 cal/gm for unirradiated rods and200 cal/gm for irradiated rods. Since this is a conservative revision on the side of safety, the staffconcludes that it is an acceptable fuel failure criterion."
Justification The maximum fuel pellet enthalpy at the hot spot calculated for each WCGS-specific RCCAejection case was less than 200 cal/gm (see Table 2.1-6). These results satisfy thecurrently-accepted fuel failure criterion.
: 2. "Westinghouse proposes a clad temperature limitation of 2700'F as the temperature above whichclad embrittlement may be expected.
Although this is several hundred degrees above themaximum clad temperature limitation imposed in the AEC ECCS Interim Acceptance
: Criteria, this is felt to be adequate in view of the relatively short time at temperature and the highlylocalized effect of a reactivity transient."
Justification As discussed in Westinghouse letter NS-NRC-89-3466 to the NRC (Reference 6), the 2700'Fcladding temperature limit was historically applied by Westinghouse to demonstrate that the coreremains in a coolable geometry during an RCCA ejection transient.
This limit was never used todemonstrate compliance with fuel failure limits and is no longer used to demonstrate corecoolability.
The RCCA ejection acceptance criteria applied by Westinghouse to demonstrate long-term core coolability and compliance with applicable offsite dose requirements are identified in Section 2.5.6, "Spectrum of Rod Cluster Control Assembly Ejection Accidents."
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-8WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-82.1.4 Computer Codes UsedSummnary descriptions of the principal computer codes used in the non-LOCA transient analyses areprovided below. Table 2.1-2 lists the computer codes used in each of the non-LOCA analyses.
FACTRANFACTRAN calculates the transient temperature distribution in a cross-section of a metal-clad UO2 fuelrod and the transient heat flux at the surface of the cladding.
The inputs are the nuclear power and thetime-dependent coolant parameters of pressure, flow, temperature, and density.
This code uses a fuelmodel with the following features:
* A sufficiently large number of radial space increments to handle fast transients such as an RCCAejection accident* Material properties that are functions of temperature
* A sophisticated fuel-to-cladding gap heat transfer calculation
* Calculations to address post-DNB conditions (film boiling heat transfer correlations, zircaloy-water
: reaction, and partial melting of the fuel)The FACTRAN licensing topical report, WCAP-7908-A (Reference 8), was approved by the USNRC viaan SER from C. E. Rossi (NRC) to E. P. Rahe (Westinghouse),
dated September 30, 1986. TheFACTRAN SER identifies seven conditions of acceptance, which are summarized in Appendix A.2,"FACTRAN for Non-LOCA Thermal Transients,"
along with justifications for application to the WCGS.RETRANRETRAN is used for studies of a pressurized water reactor (PWR) system transient response 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, RCPs, SGs (tube and shell sides), mainsteam lines, and pressurizer.
The pressurizer
: heaters, spray, relief valves, and safety valves can also bemodeled.
RETRAN includes a point neutron kinetics model and reactivity effects of the moderator, fuel,boron, and control rods. The secondary side of the SG uses a detailed nodalization for the thermaltransients.
The reactor protection system (RPS) simulated in the code includes RTs on high neutron flux,high neutron flux rate, overtemperature AT (OTAT), overpower AT (OPAT), low reactor coolant flow,high pressurizer
: pressure, low pressurizer
: pressure, high pressurizer level, safety injection (SI) actuation, and low-low SG water level. Control systems are also simulated including rod control and pressurizer pressure control.
Parts of the SI system, including the accumulators, are also modeled.
Also, aconservative approximation of the transient DNBR, based on the core thermal limits, is calculated byRETRAN.WCAP-17658-NP August 2013Licensing Report Revision 0
WESUNGHOUSE NON-PROPRIETARY CLASS 32-9WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-9The RETRAN licensing topical report, WCAP-14882-P-A (Reference 9), was approved by the USNRCvia an SER from F. Akstulewicz (NRC) to H. Sepp (Westinghouse),
dated February 11, 1999. TheRETRAN SER identifies three conditions of acceptance, which are summarized in Appendix A.3,"RETRAN for Non-LOCA Safety Analysis,"
along with justifications for application to the WCGS.Note that the RETRAN nodalization modeling used in the WCGS analyses is consistent with theWestinghouse plant nodalization model described in WCAP-14882-P-A, except for the nodalization of theRCS hot legs (HLs). Since the approval of WCAP-14882-P-A, the HL modeling was enhanced tominimize code instabilities attributed to pressurizer insurge and outsurge.
This HL model enhancement, which has been applied in other RETRAN analyses performed by Westinghouse, consisted of dividingeach HL control volume into three equal control volumes.
Although it was needed only for the HLconnected to the pressurizer, all HLs were divided in the same manner.LOFTRANThe LOFTRAN computer code is used to study the transient response of a PWR to specified perturbations in process parameters.
This code simulates a multi-loop system by a model containing the reactor vessel,hot- and cold-leg piping, SGs (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 SG uses a homogeneous, saturated mixture for the thermal transients.
The code simulates the RPS, which includes RTs on highneutron flux, OTAT and OPAT, high and low pressurizer
: pressure, low RCS flow, low-low SG waterlevel, and high pressurizer level. Control systems are also simulated, including rod control, steam dump,and pressurizer pressure control.
The SI system, including the accumulators, is also modeled.
Also, aconservative approximation of the transient DNBR, based on the core thermal limits, is calculated byLOFTRAN.The LOFTRAN licensing topical report, WCAP-7907-P-A (Reference 10), was approved by the USNRCvia an SER from C. 0. Thomas (NRC) to E. P. Rahe (Westinghouse),
dated July 29, 1983. TheLOFTRAN SER identifies one condition of acceptance, which is summarized in Appendix A.4,"LOFTRAN for Non-LOCA Safety Analysis,"
along with justification for application to the WCGS.TWINKLETWINKLE is a multi-dimensional spatial neutron kinetics code. This code uses an implicitfinite-difference method to solve the two-group transient neutron diffusion equations in one, two, andthree 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 feedbackeffects.
The code handles up to 8000 spatial points and performs steady-state initialization.
Besides basiccross-section data and thermal-hydraulic (T/H) parameters, the code accepts as input basic drivingfunctions such as inlet temperature,
: pressure, flow, boron concentration, and control rod motion. The codeprovides various outputs, such as 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 amajor perturbation in the spatial neutron flux distribution.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-10WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-10The TWINKLE licensing topical report, WCAP-7979-P-A (Reference 11), was approved by the AEC viaan SER from D. B. Vassallo (AEC) to R. Salvatori (Westinghouse),
dated July 29, 1974. The TWINKLESER does not identify any conditions, restrictions, or limitations that need to be addressed for application to the WCGS.ANCANC is an advanced nodal code capable of two-dimensional (2-D) and three-dimensional (3-D)neutronics calculations.
ANC is the reference model for certain safety analysis calculations, powerdistributions, peaking factors, critical boron concentrations, control rod worths, and reactivity coefficients.
In addition, 3-D ANC validates one-dimensional (l-D) and 2-D results and providesinformation about radial (x-y) peaking factors as a function of axial position.
It can calculate discrete pinpowers from nodal information as well.The ANC licensing topical report, WCAP- 10965-P-A (Reference 12), was approved by the USNRC viaan SER from C. Berlinger (NRC) to E. P. Rahe (Westinghouse),
dated June 23, 1986. The ANC SER doesnot identify any conditions, restrictions, or limitations that need to be addressed for application to theWCGS.VIPREThe VIPRE computer program performs T/H calculations.
This code calculates coolant density, massvelocity,
: enthalpy, void fractions, static pressure, and DNBR distributions along flow channels within areactor core.The VIPRE licensing topical report, WCAP-14565-P-A (Reference 13), was approved by the USNRC viaan SER from T. H. Essig (NRC) to H. Sepp (Westinghouse),
dated January 19, 1999. The VIPRE SERidentifies four conditions of acceptance, which are summarized in Appendix A.5, "VIPRE for Non-LOCAThermal/Hydraulics,"
along with justifications for application to the WCGS.2.1.5 Initial Conditions The initial conditions applied in non-LOCA transient analyses are dependent on the analysis methodology employed for each transient.
For the purpose of this discussion, the non-LOCA analyses are categorized as either DNB or non-DNB.
DNB analyses include the transient cases analyzed for DNB concerns, andnon-DNB analyses include the transient cases analyzed for concerns other than DNB, e.g., RCSoverpressure.
For most DNB analyses, the Revised Thermal Design Procedure (RTDP) methodology ofReference 14 was employed.
With this methodology, nominal values are applied as the initial RCSconditions of power (see Note below), temperature,
: pressure, and flow, and the corresponding uncertainty allowances (identified later in this section) are accounted for statistically in defining the design limitDNBR. In RTDP DNB analyses, the nominal RCS flow is the MMF value and the core bypass flow is thestatistical value (see Section 2.1.1, Program Features, for the MMF and bypass flow values).Note: The reactor power applied in all analyses is consistent with the NSSS power of 3651 MWt, whichincludes a bounding uncertainty of up to 2 percent.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-11As discussed in Section 2.12, "Thermal and Hydraulic Design,"
uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, computer codes, and DNBcorrelation predictions were combined statistically to obtain the overall DNB uncertainty factor, whichwas used to define the design limit DNBR. In other words, the design limit DNBR is a DNBR value thatis greater than the WRB-2 DNB correlation limit by an amount that accounts for the RTDP uncertainties.
To provide DNBR margin to offset various penalties such as those due to rod bow and instrument bias,and to provide flexibility in design and operation of the plant, the design limit DNBR was conservatively increased to a value designated as the safety analysis limit DNBR, to which transient-specific DNBRvalues were compared.
For DNB analyses where RTDP is not employed, which are those DNB analyses that are initiated fromzero power conditions, the initial conditions were defined by applying
: maximum, steady-state uncertainties to the nominal values in the most conservative direction, as appropriate; this is known asStandard Thermal Design Procedure (STDP) methodology, or non-RTDP.
In non-RTDP DNB analyses, the initial RCS flow is the TDF value and the core bypass flow is the design value (see Section 2.1.1,"Program Features,"
for the TDF and bypass flow values).
As discussed in Section 2.12, "Thermal andHydraulic Design,"
the DNBR limits for non-RTDP DNB analyses correspond to the appropriate DNBcorrelation limit increased by sufficient margin to offset any applicable DNBR penalties.
For each DNBanalysis, Table 2.1-6 identifies whether RTDP or non-RTDP (STDP) was applied, the DNB correlation, and the DNBR limit.For non-DNB analyses, the initial conditions were defined by applying
: maximum, steady-state uncertainties to the nominal values in the most conservative direction.
In these analyses, the initial RCSflow is the TDF value and the core bypass flow is the design value (see Section 2.1.1, "ProgramFeatures,"
for the TDF and bypass flow values).Steady-State Initial Condition Uncertainties The following bulleted items identify the maximum steady-state initial condition uncertainties for corepower, RCS flow, Tayg. and pressurizer pressure that had to be accounted for in the non-LOCA safetyanalyses.
More limiting (bounding) uncertainties than those presented below may have been applied insome analyses.
Table 2.1-2 summarizes the initial conditions applied in each analysis.
* 0 percent core power allowance for calorimetric measurement uncertainty
* As indicated above, all applicable uncertainties are accounted for in the applied initial core powervalue.* +/-2.7 percent RCS flow allowance for steady-state fluctuations and measurement uncertainties
* +6.5/-4.0°F Tavg allowance for deadband and system measurement uncertainties and bias* +50/-35 psi pressurizer pressure allowance for steady-state fluctuations and measurement uncertainties WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-12WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-12Pressurizer Level Initial Condition The nominal pressurizer water level program used at the WCGS varies linearly from 27 percent of span atthe no-load Tag of 557°F to 57 percent of span at a Tavg of 586.5°F.
For Tavg values above 586.5°F andbelow 557°F, the program level is constant at the respective levels of 57 percent of span and 27 percent ofspan. For analysis
: purposes, the upper end of the pressurizer water level program was conservatively extrapolated out to 59 percent of span at the maximum full-power Tavg value of 588.4°F.
An uncertainty ofat least 5 percent of span was applied when conservative.
Steam Generator Initial Conditions The steam flow rate and steam pressure initial conditions are dependent on the initial conditions of power,Tavg, RCS flow (TDF or MMF), Tfed, SG water level, and SGTP level. The analyses considered a fullpower Tfed range of 400.0°F to 448.6°F, a constant SG water level program of 50 percent narrow rangespan (NRS), and a SGTP level range of 0 percent to 10 percent.
An uncertainty was applied to the initialSG levels when it was conservative to do so; the level uncertainties considered were + 10 percent NRSand -12 percent NRS, which correspond to initial levels of 60 percent NRS and 38 percent NRS,respectively.
Residual Decay HeatThe fission product contribution to decay heat applied in the non-LOCA analyses is consistent with theAmerican National Standards Institute (ANSI)/(ANS standard ANSI/ANS-5.1-1979 for decay heat powerin light water reactors (Reference 15), including two standard deviations of uncertainty.
2.1.6 Fuel Design Description The fuel currently in use at the WCGS and considered in the safety analyses described herein is theWestinghouse 17x 17 RFA-2 fuel design with IFMs and thimble plugs either removed or installed (seeNote below). The RFA-2 fuel rods contain enriched uranium dioxide (UO2) fuel pellets and have ZIRLOHigh Performance Fuel Cladding Material")
with an outer diameter of 0.374 inch. ZIRLO material is alsoused as the material for the mid-grids, guide thimble tubes, and instrumentation tubes. More detailedinformation on the RFA-2 fuel design is provided in Chapter 4.0 of the WCGS USAR (Reference 1). Withrespect to the non-LOCA transient
: analyses, the effects of fuel design mechanical features were accounted for in fuel-related input parameters such as fuel and cladding dimensions, cladding
: material, fueltemperatures, and core bypass flow.Note: Except for the limiting DNBR analysis of the uncontrolled RCCA bank withdrawal at powerevent, all analyses covered the bounding scenario of having the core TPR. As a result of theexception, the plant may be required to operate with the core TPI.( ZIRLO is a registered trademark of Westinghouse Electric Company LLC, its Affiliates and/or its Subsidiaries inthe United States of America and may be registered in other countries throughout the world. All rights reserved.
Unauthorized use is strictly prohibited.
Other names may be trademarks of their respective owners.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESUNGHOUSE NON-PROPRIETARY CLASS 32-13WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-13Regarding the issue of fuel thermal conductivity degradation (TCD) with Westinghouse codes andmethods, Westinghouse provided a discussion of the TCD impact in Reference 16 and justified thecontinued safe operation of the plants analyzed with Westinghouse codes and methods.
The Westinghouse codes and methods applied in the non-LOCA analyses discussed herein are consistent with thoseevaluated for TCD in Reference 16, and therefore the conclusions presented in Reference 16 areapplicable to the WCGS.2.1.7 Power Distribution Peaking FactorsRelative to the fuel, the power distribution is characterized by nuclear enthalpy rise hot channel factors(radial peaking factor, FNAH) of 1.59 for RTDP DNB analyses and 1.65 for non-RTDP DNB analyses, anda full-power heat flux hot channel factor (total peaking factor, FQ) of 2.50. FNAH is important for transients that are analyzed for DNB concerns.
The DNB transients as well as the DNB methodology applied(RTDP or non-RTDP) in the DNB analyses are identified in Table 2.1-6. As FNAH increases withdecreasing power level, due to rod insertion, all transients analyzed for DNB concerns are assumed tobegin with an FNAH consistent with the FNAH defined in the Core Operating Limits Report (COLR) for theassumed nominal power level. The FQ, for which the limits are specified in the COLR, is important fortransients that are analyzed for overpower
: concerns, for example RCCA ejection.
2.1.8 Reactivity FeedbackThe transient response of the reactor core is dependent on reactivity feedback
: effects, in particular themoderator temperature coefficient (MTC), Doppler temperature coefficient (DTC), and the Dopplerpower coefficient (DPC). Depending upon event-specific characteristics, conservatism dictates the use ofeither maximum or minimum reactivity coefficient values. Justification for the use of the reactivity coefficient values was treated on an event-specific basis. Table 2.1-3 presents the core kinetics parameters and reactivity feedback coefficients applied in the non-LOCA analyses.
The maximum and minimumintegrated DPCs applied in the safety analyses are provided in Figure 2.1-1. Note that a different DPC(not shown in Figure 2.1-1) was applied in the zero power SLB core response and zero power feedwater malfunction (FWM) analyses; this DPC is based on an RCCA being stuck out of the core.2.1.9 Pressure Relief ModelingRCS Pressure ReliefPlant components that provide RCS pressure relief in the non-LOCA analyses include the pressurizer sprays, pressurizer power-operated relief valves (PORVs),
and the pressurizer safety valves (PSVs). Themodeling of these components in each non-LOCA safety analysis is dependent on the type of transient being analyzed and the applicable analysis methodology.
Note that the sprays and PORVs are not safetygrade components, and thus were modeled only if doing so lead to more limiting
: results, i.e., no creditwas taken for the operation of these components if such operation were to mitigate transient results.
Ingeneral, maximum RCS pressure relief is modeled when a minimum RCS pressure is conservative, e.g., for transients that are analyzed for DNB concerns, and minimum RCS pressure relief is modeledWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-14WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-14when a maximum RCS pressure is conservative.
Modeling details for the sprays, PORVs, and PSVs areprovided as follows, but note that more conservative modeling may have been applied in some analyses.
Sprays -The pressurizer sprays were modeled via a control valve that was set to initially open ona proportional-integral-derivative (PID) pressure signal of +25 psid from the nominal reference pressure of 2250 psia, and ramping to full-open when the PID pressure signal reaches +75 psid.This spray control logic is consistent with the as-designed logic.PORVs -Each of the two pressurizer PORVs were modeled based on a relief capacity of210,000 Ibm/hr at a pressure of 2350 psia. One PORV was modeled to actuate on an indicated pressure signal of 2350 psia and the other PORV was modeled to actuate on a PID pressure signalof 100 psid from the nominal reference pressure of 2250 psia.PSVs -Each of the three PSVs was modeled based on a relief capacity of 420,000 lbm/hr at apressure of 2575 psia. Depending on the direction of conservatism for a given analysis, thenominal opening setpoint of 2460 psig was either increased by 2.9 percent, which accounts for a+2.0 percent setpoint tolerance and a +0.9 percent set pressure shift associated with thewater-filled PSV loop seals (see WCAP-12910, Reference 17), or decreased by 2.0 percent,which accounts for a -2.0 percent setpoint tolerance.
Also, when conservative, a PSV openingdelay of 1.153 seconds was modeled to account for the purging of the water in the PSV loopseals.The pressurizer
: heaters, which include proportional heaters and backup heaters, are related to the RCSpressure relief components in that they are included as part of the pressurizer pressure control system. Thepressurizer heaters were modeled as-designed if doing so causes transient results to be more limiting.
Theproportional heaters were modeled with a maximum capacity of 416 kW and the backup heaters weremodeled with a maximum capacity of 1384 kW. The heat output of the proportional heaters varies linearlyas a function of the PID pressure signal. The proportional heaters are on at 50 percent capacity when the PIDpressure signal is 0 psid, 100 percent capacity when the PID pressure signal is -15 psid, and 0 percentcapacity when the PID pressure signal is +15 psid. The backup heaters turn on at full capacity when the PIDpressure signal is -25 psid or if the pressurizer level deviates from the program level by +5 percent of span.Main Steam System (MSS) Pressure ReliefPlant components that provide MSS pressure relief in the non-LOCA analyses include the control gradeatmospheric relief valves (ARVs) and the safety grade main steam safety valves (MSSVs).
No credit istaken for the automatic actuation of the ARVs. Rather, operator action to open an ARV is credited in theanalysis described in Section 2.6.1, "Inadvertent Operation of the Emergency Core Cooling SystemDuring Power Operation."
General modeling details for the MSSVs are provided as follows, but note thatmore conservative modeling may have been applied in some analyses.
Five MSSVs per loop were modeled with opening setpoints based on nominal lift settings of1185, 1197, 1210, 1222, and 1234 psig. Each MSSV was modeled with a +3.0 percent setpointtolerance and a 5 psi ramp from closed to full-open, which accounts for accumulation.
Becausenone of the non-LOCA transients is limiting with minimum MSSV setpoints, a negative setpointtolerance was not explicitly modeled.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-15WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-152.1.10 RTS and ESFAS Functions Table 2.1-4 summarizes the RTS and engineered safety features actuation system (ESFAS) functions actuated in the non-LOCA transient analyses.
The setpoints applied in the safety analyses and theassociated time delays of each function are also presented in Table 2.1-4. Additional information relatedto the OTAT and OPAT RT setpoints is provided as follows.OTAT and OPAT Reactor Trip Setpoints Using the methodology described in WCAP-8745-P-A (Reference 18), the current OTAT and OPAT RTsetpoints were evaluated for the TM Program.
The evaluation process first involved using conservative core thermal limits, developed based on the RTDP DNB methodology (as described in Section 2.12,"Thermal and Hydraulic Design"),
to determine, under steady-state conditions, whether the OTAT andOPAT RT setpoints provide sufficient protection for the core thermal limits. Based on this initialevaluation, it was determined that one coefficient of the OTAT RT setpoint
: equation, the pressure termcoefficient, had to be increased from 0.00067 1/psi to 0.00095/psi to ensure that the core thermal limits arefully protected.
The applied core thermal limits are presented in Figure 2.1-2. The OTAT and OPAT RTsetpoints are illustrated in Figure 2.1-3 and presented in Table 2.1-5.The boundaries of operation defined by the OTAT and OPAT trips are represented as "protection lines" inFigure 2.1-3. The protection lines were drawn to include all adverse instrumentation and setpoint errors sothat under nominal conditions, a trip would occur well within the area bounded by these lines. Theseprotection lines are based upon the OTAT and OPAT RT setpoints applied in the safety analyses, whichare the TS nominal values with allowances for instrumentation errors and acceptable drift betweeninstrument calibrations.
The diagram of Figure 2.1-3 is useful in the fact that the limit imposed by anygiven DNBR can be represented as a line (Tavg versus AT). The DNB lines represent the locus ofconditions for which the DNBR equals the limit value. All points below and to the left of a DNB line for agiven pressure have a DNBR greater than the SAL DNBR value. The area of permissible operation (power, temperature, and pressure) is bounded by the combination of the high neutron flux (fixedsetpoint)
RT, the high and low pressurizer pressure RTs (fixed setpoints),
the OTAT (variable setpoint) and OPAT (variable setpoint)
RTs, and the opening of the MSSVs, which limits the maximum RCSaverage temperature.
The final determination of the adequacy of the OTAT and OPAT RT setpoints is demonstrated by showingthat the design bases for DNB and fuel melting are met in the analyses of those events that credit thesefunctions for accident mitigation.
Table 2.1-4 identifies the event analyses that credit the OTAT andOPAT RT functions.
In these analyses, the dynamic compensation tenrms of the OTAT and OPAT setpointequations, which compensate for inherent instrumentation delays and piping lags between the reactor coreand the loop temperature
: sensors, were modeled.
As the analysis results presented in Table 2.1-6 showthat all applicable limits are met for the analyses that credit OTAT and OPAT, the OTAT and OPAT RTsetpoints (see Table 2.1-5) are confirmed to be adequate.
Note that the OTAT penalty function that is usedto compensate for expected variations in the axial power shape, f(AI), although not explicitly credited inthe analyses, was separately confirmed to be acceptable based on the method described inWCAP-8745-P-A (Reference 18).WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-16WESTIINGHOUSE NON-PROPRIETARY CLASS 3 2-16Related to the OTAT and OPAT setpoint functions, the functional temperature ranges of the TcoId, Thor, andTavg resistance temperature detector instrumentation were reviewed to ensure that they cover the expectedtemperature ranges. It was determined that the current TCoId and Tavg ranges were acceptable, but the Thotrange required an adjustment, as indicated below.* Tcojd -510-F -630'F (same as current range)* Thot -540'F -660'F (requires a revision from the current range of 530°F-650°F)
* Tavg -530°F -630°F (same as current range)Finally, note that a temperature difference of up to 3°F between the nominal (reference) temperature usedfor the OTAT and OPAT RT setpoints and the loop-specific, indicated, full-power Tavg values has beencovered for the analyses that rely on these RT functions for protection.
2.1.11 Reactor Trip Characteristics The negative reactivity insertion following a RT is a function of the acceleration of the RCCAs and thevariation in rod worth as a function of rod position.
With respect to the non-LOCA transient
: analyses, thecritical parameter is the time from the start of RCCA insertion to when the RCCAs reach the dashpotregion, which is located at an insertion point corresponding to approximately 86 percent of the totalRCCA travel distance.
For the non-LOCA
: analyses, the RCCA insertion time from filly withdrawn todashpot entry was modeled as 2.7 seconds.
The applied negative reactivity insertion following RT isbased on having the most reactive RCCA stuck in the fully withdrawn position.
Three figures relating to RCCA drop time and reactivity worth are presented in this report. The RCCAposition (fraction of full insertion) versus the time from release is presented in Figure 2.1-4. Thenormalized reactivity worth applied in the safety analyses is shown in Figure 2.1-5 as a function of rodinsertion fraction and in Figure 2.1-6 as a function of time. A total negative trip reactivity worth of4.0 percent Ak was modeled in the non-LOCA
: analyses, unless noted otherwise.
In the analyses of zeropower transients that have a potential for a return-to-power (FW system malfunction, steam system pipingfailure, inadvertent opening of a SG atmospheric relief or safety valve), a minimum shutdown margin of1.3 percent Ak was conservatively modeled.2.1.12 Operator Actions CreditedTo help demonstrate compliance with applicable acceptance
: criteria, operator actions were credited in theanalysis of the inadvertent operation of the ECCS during power operation event; see Section 2.6.1,"Inadvertent Operation of the Emergency Core Cooling System During Power Operation,"
for details ofthe credited operator actions.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-17In addition, there are two events that were analyzed to demonstrate that there is sufficient time available for operators to recognize the event is in progress and to take necessary actions to terminate the eventprior to reaching plant conditions that fail to comply with applicable acceptance criteria.
Althoughoperator actions are not modeled in the analyses of these events, actions by the plant operators areultimately required to ensure plant safety is maintained.
These two events are as follows.Boron dilution (Section 2.5.5, "Chemical and Volume Control System Malfunction Resulting in aDecrease in Boron Concentration in the Reactor Coolant")
CVCS malfunction that increases reactor coolant inventory (Section 2.6.2, "Chemical andVolume Control System Malfunction that Increases Reactor Coolant Inventory")
2.1.13 Results SummaryTable 2.1-6 summarizes the results obtained for each of the non-LOCA transient analyses.
The resultsdemonstrate that all applicable safety analysis acceptance criteria are satisfied for the WCGS. Althoughthe analyses and evaluations were performed with the intent to make them cycle-independent, the RSEprocess described in Reference 7 will be applied for future fuel reloads to verify that reload-related safetyanalysis inputs remain bounding.
2.1.14 References
: 1. "Wolf Creek Updated Safety Analysis Report,"
Revision 26, March 2013.2. ANS-51.1-1973 (ANSI-N 18.2), "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
August 1973.3. WCAP-9226-P-A, Revision 1, "Reactor Core Response to Excessive Secondary Steam Releases,"
February 1998.4. WCAP- 11394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.5. WCAP-7588, Revision I-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods,"
January 1975.6. NS-NRC-89-3466, Letter from W. J. Johnson (Westinghouse) to R. C. Jones (NRC), "Use of2700'F PCT Acceptance Limit in Non-LOCAAccidents,"
October 23, 1989.7. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"
July 1985.8. WCAP-7908-A, "FACTRAN
-A FORTRAN IV Code for Thermal Transients in a UO2Fuel Rod," December 1989.9. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-18WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1810. WCAP-7907-P-A, "LOFTRAN Code Description,"
April 1984.11. WCAP-7979-P-A, "TWINKLE
-A Multi-Dimensional Neutron Kinetics Computer Code,"January 1975.12. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.13. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.14. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"
April 1989.15. ANSI/ANS-5.1-1979, "American National Standard for Decay Heat Power in Light WaterReactors,"
August 29, 1979.16. LTR-NRC-12-18, Letter from J. A. Gresham (Westinghouse) to USNRC Document Control Desk,"Westinghouse Response to December 16, 2011 NRC Letter Regarding Nuclear Fuel ThermalConductivity Degradation (TAC No. ME5186) (Proprietary),"
February 17, 2012.17. WCAP-12910, Revision 1-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.18. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,"
September 1986.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTfNGHOUSE NON-PROPRIETARY CLASS 32-19Table 2.1-1 Non-LOCA Transient Events Analyzed or Evaluated Report USAR Analyzed orTransient Event Section Section Evaluated?
Feedwater system malfunctions that result in a decrease in 2.2.1 15.1.1 Analyzedfeedwater temperature Feedwater system malfunctions that result in an increase in 2.2.2 15.1.2 Analyzedfeedwater flowExcessive increase in secondary steam flow 2.2.3 15.1.3 AnalyzedInadvertent opening of a steam generator atmospheric relief or 2.2.4 15.1.4 Analyzedsafety valveSteam system piping failure (SLB) at zero power 2.2.5.1 15.1.5 AnalyzedSteam system piping failure (SLB) at full power 2.2.5.2 15.1.6 AnalyzedLoss of external electrical load, turbine trip, inadvertent closure of 2.3.1 15.2.2 Analyzedmain steam isolation valves, and loss of condenser vacuum 15.2.315.2.415.2.5Loss of non-emergency AC power to the station auxiliaries 2.3.2 15.2.6 AnalyzedLoss of normal feedwater flow 2.3.3 15.2.7 AnalyzedFeedwater system pipe break 2.3.4 15.2.8 AnalyzedPartial loss of forced reactor coolant flow 2.4.1 15.3.1 AnalyzedComplete loss of forced reactor coolant flow 2.4.1 15.3.2 AnalyzedRCP shaft seizure (locked rotor) and RCP shaft break 2.4.2 15.3.3 Analyzed15.3.4Uncontrolled RCCA bank withdrawal from a subcritical or low 2.5.1 15.4.1 Analyzedpower startup condition Uncontrolled RCCA bank withdrawal at power 2.5.2 15.4.2 AnalyzedRCCA misoperation (dropped RCCA, dropped RCCA bank, 2.5.3 15.4.3 Analyzedstatically misaligned RCCA, single RCCA withdrawal)
Startup of an inactive RCP at an incorrect temperature 2.5.4 15.4.4 Evaluated CVCS malfunction that results in a decrease in the boron 2.5.5 15.4.6 Analyzedconcentration in the reactor coolant (boron dilution)
Spectrum of RCCA ejection accidents 2.5.6 15.4.8 AnalyzedInadvertent operation of the ECCS during power operation 2.6.1 15.5.1 AnalyzedCVCS malfunction that increases reactor coolant inventory 2.6.2 15.5.2 AnalyzedInadvertent opening of a pressurizer safety or relief valve 2.7.1 15.6.1 AnalyzedATWS 2.8 15.8 AnalyzedWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-20WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-20Table 2.1-2 Summary of Initial Conditions and Computer Codes UsedInitial Reactor Vessel Reactor Vessel Average RCSComputer Power Coolant t1  Coolant Temperature(2)
Pressure(
3)Event Case Distinction Code(s) Used (%) (gpm) (OF) (psia)Feedwater system malfunctions Bounding RETRAN 100 371,000 588.4 2250that result in a decrease infeedwater temperature Feedwater system malfunctions Zero power RETRAN 0 361,200 557.0 2250that result in an increase in ANCfeedwater flow VIPREFull power RETRAN 100 371,000 588.4 2250Excessive increase in secondary Bounding RETRAN 100 371,000 588.4 2250steam flowInadvertent opening of a steam Bounding RETRAN 0 361,200 557.0 2250generator atmospheric relief or ANCsafety valve VIPRESteam system piping failure Zero power RETRAN 0 361,200 557.0 2250(SLB) ANC(core response only) VIPREFull power RETRAN 100 371,000 588.4 2250ANCVIPRELoss of external electrical load, Minimum DNBR RETRAN 100 371,000 588.4 2250turbine trip, inadvertent closure of Peak RCS Pressure RETRAN 100 361,200 581.9 2215main steam isolation valves, andloss of condenser vacuum Peak MSS Pressure RETRAN 100 361,200 594.9 2200Loss of non-emergency AC Bounding RETRAN 100 361,200 564.2 2300power to the station auxiliaries Loss of normal feedwater flow Bounding RETRAN 100 361,200 564.2 2300Feedwater system pipe break Bounding RETRAN 100 361,200 594.9 2200Partial loss of forced reactor Bounding RETRAN 100 371,000 588.4 2250coolant flow VIPREComplete loss of forced reactor Bounding RETRAN 100 371,000 588.4 2250coolant flow VIPREWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-21WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-21Table 2.1-2 Summary of Initial Conditions and Computer Codes Used (cont.)Initial Reactor Vessel Reactor Vessel Average RCSComputer Power Coolant Flow"') Coolant Temperature(
: 2) Pressure(
3)Event Case Distinction Code(s) Used (%) (gpm) (OF) (psia)RCP shaft seizure (locked rotor) DNB RETRAN 100 371,000 588.4 2250and RCP shaft break VIPREPeak RCS pressure/
RETRAN 100 361,200 594.9 2300PCT VIPREUncontrolled RCCA bank Bounding TWINKLE 0 160,662 557.0 2200withdrawal from a subcritical or FACTRANlow power startup condition VIPREUncontrolled RCCA bank Minimum DNBR RETRAN 100 371,000 588.4 2250withdrawal at power VIPRE 60 and 575.8376,000(4) 10 560.1______
Peak RCS Pressure LOFTRAN Various'5' 361,200 Various(5) Various(5'Dropped RCCA(s) and Dropped All LOFTRAN 100 371,000 588.4 2250RCCA bank ANCVIPREStatically misaligned RCCA All ANC 100 371,000 588.4 2250VIPRESingle RCCA withdrawal Manual Rod Control ANC 100 371,000 588.4 2250VIPREStartup of an inactive RCP at an No analysis was performed; incorrect temperature see Section 2.5.4, "Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature."
CVCS malfunction that results in Mode 1 N/A 100 N/A 594.9 2250a decrease in the boron Mode 2 5 565.1concentration in the reactorcoolant (boron dilution)
Mode 3 0 350.0, 557.0Mode 4 0 200.0Mode 5 0 68.0Mode 6 No analysis was performed; a Mode 6 boron dilution is precluded by administrative controls.
Spectrum of RCCA ejection Full power TWINKLE 100 361,200 594.9 2200accidents Zero power FACTRAN 0 160,662 557.0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-22WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-22Table 2.1-2 Summary of Initial Conditions and Computer Codes Used (cont.)Initial Reactor Vessel Reactor Vessel Average RCSComputer Power Coolant Flow01  Coolant Temperature(
: 2) Pressuret3)Event Case Distinction Code(s) Used (%) (gpm) (OF) (psia)Inadvertent operation of the Bounding RETRAN 100 361,200 566.7 2215ECCS during power operation CVCS malfunction that increases Bounding RETRAN 100 361,200 564.2 2200reactor coolant inventory Inadvertent opening of a Bounding RETRAN 100 371,000 588.4 2250pressurizer safety or relief valveATWS Bounding LOFTRAN 100 361,200 588.4 2250Notes:1. 361,200 gpm-TDF371,000 gpm Bounding MMF376,000 gpm MMF160,662 gpm Reactor vessel flow provided by two RCPs = 0.4448xTDF.
: 2. 594.9°F = High nominal full power Ta,,g (588.40F) + 6.50F588.4°F = High nominal full power Ta,,g581.90F = High nominal full power Ta,.g (588.40F) -6.5°F575.8°F = 60% power Tavg (linearly interpolated between Tno-load and the high nominal full power Tag of 588.40F)570.7°F = Low nominal full power Tang566.70F = Low nominal full power Ta,,g (570.70F) -4.0°F565. IF = 5% power Ta.g (linearly interpolated between To0_1-ad and the high nominal full power Ta,,g of 588.40F) + 6.50F564_20F = Low nominal full power Tang (570.70F) -6.50F560. 1F = 10% power T.vg (linearly interpolated between Tno-load and the high nominal full power Ta,,g of 588.40F)557.00F = 0% power Tavg = Tno-load
= Mode 3 maximum Ta,,g350.0°F = Mode 3 minimum Tavg200.0°F = Mode 4 minimum Ta,,g68.0°F ý Mode 5 minimum Tavg3. 2300 psia = Nominal + 50 psi2250 psia = Nominal2215 psia = Nominal -35 psi2200 psia = Nominal -50 psi4. As indicated in Section 2.12, "Thermal and Hydraulic Design,"
for the most limiting case in the uncontrolled RCCA withdrawal at power DNBR analysis, credit was takenfor the higher MMF of 376,000 gpm to help demonstrate that the DNB design basis was met with adequate margin.5. For the uncontrolled RCCA withdrawal at power peak RCS pressure
: analysis, a spectrum of initial power levels ranging from 8 to 100% was analyzed.
The corresponding initial Tavgs were based on the high nominal full power Ta,,g of 588.4°F (linear between 588.4°F at 100% power and 557°F at 0% power) plus uncertainty (6.50F). Caseswere analyzed with initial pressurizer pressures of 2200 psia (nominal minus uncertainty) and 2300 psia (nominal plus uncertainty).
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-23WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-23Table 2.1-3 Core Kinetics Parameters and Reactivity Feedback Coefficients BOC EOCParameter (Minimum Feedback)
(Maximum Feedback)
Moderator temperature coefficient, pcm/&deg;F 6.0 (< 70% RTP)") N/A0.0 (> 70% RTP)Moderator density coefficient, Ak/(g/cc)
N/A 0.47Doppler temperature coefficient, pcm/&deg;F -1.0 -3.5Doppler-only power coefficient, pcm/percent power -10.13 + 0.0342Q -19.33 + 0.0662Q(Q = power in %)Delayed neutron fraction 0.0075 (maximum) 0.0044 (minimum)
Doppler power defect, pcm* RCCA ejection 1007 925* RCCA withdrawal from subcritical 1007 N/ANote:1. RTP -- rated thermal powerWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32 -24WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-24Table 2.1-4 Summary of RTS and ESFAS Functions ActuatedDelayEvent Case Distinction RTS or ESFAS Signal(s)
Actuated Analysis Setpoint (seconds)
Feedwater system malfunctions Bounding OPAT RT See Table 2.1-5 7.0(1that result in a decrease infeedwater temperature Low pressurizer pressure SI with feedwater 1715.0 psia 17.0 (FWI)isolation (FWI) on SI(2)Feedwater system malfunctions Zero power Hi-hi SG water level turbine trip (TT) and FWI 100% NRS 2.5 (TT)that result in an increase in 17.0 (FWI)feedwater flow Full powerExcessive increase in secondary Bounding None N/A N/Asteam flowInadvertent opening of a steam Bounding Low pressurizer pressure SI with FWI on SI 1715.0 psia 27.0 (SI)generator atmospheric relief or 17.0 (FWI)safety valveSteam system piping failure (SLB) Zero power Low steam line pressure SI and steam line isolation 375.0 psia 27.0 (SI)(core response only) (SLI) with FWI on SI (lead/lag
= 50/5 sec) 17.0 (SLI)17.0 (FWI)Full power OPAT RT See Table 2.1-5 7.0(1)Loss of external electrical load, Minimum DNBR OTAT RT See Table 2.1-5 7.01)turbine trip, inadvertent closure ofmain steam isolation valves, and Peak RCS Pressure High pressurizer pressure RT 2425.0 psia 1.0loss of condenser vacuum Peak MSS Pressure OTAT RT See Table 2.1-5 7.0"'Loss of non-emergency AC power Bounding Low-low SG water level RT and AFW system 0% NRS 2.0 (RT)to the station auxiliaries actuation 60.0 (AFW)Loss of normal feedwater flow BoundingFeedwater system pipe break Bounding Low-low SG water level RT and AFW system 0% NRS 2.0 (RT)actuation 60.0 (AFW)Partial loss of forced reactor Bounding Low reactor coolant loop flow RT 86.3% MMF 1.0coolant flowWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRJETARY CLASS 32-25WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-25Table 2.1-4 Summary of RTS and ESFAS Functions Actuated (cont.)DelayEvent Case Distinction RTS or ESFAS Signal(s)
Actuated Analysis Setpoint (seconds)
Complete loss of forced reactor Bounding RCP undervoltage (UV) RT (3) 1.5coolant flowRCP shaft seizure (locked rotor) DNB Low reactor coolant loop flow RT 86.3% MMF 1.0and RCP shaft breakPeak RCS pressure!
peak claddingtemperature (PCT)Uncontrolled RCCA bank Bounding 35% RTP 0.5withdrawal from a subcritical or Power range neutron flux (low setting)
RTlow power startup condition Uncontrolled RCCA bank Minimum DNBR Power-range high neutron flux RT (high setting) 116.5% RTP 0.5withdrawal at power OTAT RT See Table 2.1-5 6.25(l)Peak RCS Pressure Power-range high neutron flux RT (high setting) 116.5% RTP 0.5OTAT RT See Table 2.1-5 7.0"'Power range neutron flux rate (high positive rate) 6.9% RTP with a 1.0RT 2.0-second time constantHigh pressurizer pressure RT 2425.0 2.0Dropped RCCA(s) and Dropped See Note 4RCCA bankStatically misaligned RCCA All None N/A N/ASingle RCCA withdrawal Manual Rod Control None N/A N/AStartup of an inactive RCP at an No analysis was performed; incorrect temperature see Section 2.5.4, "Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature."
CVCS malfunction that results in a Mode 1-manual OTAT RT See Table 2.1-5 7.0(1)decrease in the borontdcesintebrnMode 1 -auto None N/A N/Aconcentration in the reactorcoolant (boron dilution)
Mode 2 None N/A N/AWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-26WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-26Table 2.1-4 Summary of RTS and ESFAS Functions Actuated (cont.)DelayEvent Case Distinction RTS or ESFAS Signal(s)
Actuated Analysis Setpoint (seconds)
Mode 3 None N/A N/AMode 4 None N/A N/AMode 5 None N/A N/AMode 6 N/A N/A N/ASpectrum of RCCA ejection Full power Power range neutron flux (high setting)
RT 118% RTP 0.5accidents Zero power Power range neutron flux (low setting)
RT 35% RTP 0.5Inadvertent operation of the ECCS Bounding See Note 5 N/A N/Aduring power operation CVCS malfunction that increases Bounding None N/A N/Areactor coolant inventory Inadvertent opening of a Bounding OTAT RT See Table 2.1-5 7.0")pressurizer safety or relief valveATWS All See Note 6 N/A N/ANotes:I. The OTAT and OPAT RT response times were modeled with a time constant (first order lag) of 4.0 seconds to account for the response of the resistance temperature detectors (RTDs), the RTD bypass piping fluid transport time, and the RTD bypass piping heatup thermal lag, and a pure delay of at least 2.25 seconds to account forprotection system electronics delays, RT breaker opening, and RCCA gripper release.
A pure delay of 3.0 seconds was conservatively modeled in some analyses.
: 2. No SI flow was modeled because the transient is terminated by FWI before SI flow would be initiated.
: 3. The RCP UV RT (initiation of rod motion) was assumed to occur 1.5 seconds following the loss of bus voltage.4. Multiple cases were analyzed to cover bounding values for MTC, dropped RCCA(s) worth, and D-bank worth. The limiting cases do not result in actuation of any RTS orESFAS functions.
: However, the low pressurizer pressure RT, with an analysis setpoint of 1875 psia, was actuated for some cases that are non-limiting with respect toDNBR, e.g., dropped RCCA bank cases.5. A RT is conservatively modeled coincident with event initiation; see Section 2.6.1, "Inadvertent Operation of the Emergency Core Cooling System During PowerOperation,"
for more information.
No ESFAS functions are actuated for event mitigation.
: 6. The ATWS mitigation system actuation circuitry (AMSAC) is credited in the ATWS analysis; see Section 2.8, "Anticipated Transients Without Scram," for moreinformation.
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-27WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-27Table 2.1-5 Parameters Related to OTAT and OPAT RT Setpoints OTAT K, (safety analysis value) 1.205OTAT K2  0.0137/0FOTAT K3  0.00095/psi OTAT f(AI) deadband
-23% Al to +5% AlOTAT f(AI) negative gain -2.27 %/% AlOTAT f(AI) positive gain +1.84 %/% AlT' (OTAT) and T" (OPAT) Note IP' (OTAT) 2250 psiaOPAT K4 (safety analysis value) 1.169OPAT K5  -for decreasing Tasg 0.0/0F-for increasing Tavg 0.02/0FOPAT K6  -for Tavg > T" 0.00 128/0F-for Ta,g < T" 0.0/&deg;FAllowable full-power Tavg range 570.70F to 588.40FPressurizer pressure range of applicability for OTAT and OPAT 1924.7 psia to 2459.7 psia 2)Notes:I. The analyzed initial Tasvg is used as the reference (T' and T") in the OTAT and OPAT setpoint equations.
: 2. Values correspond to bounding SAL for the low and high pressurizer pressure RT setpoints.
WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-28Table 2.1-6 Non-LOCA Results SummarySafety Analysis Safety AnalysisEvent Case Distinction Parameter Description Result LimitFeedwater system malfunctions that Bounding Minimum DNBR (RTDP, WRB-2 correlation) 1.77 1.521"result in a decrease in feedwater temperature Maximum core average heat flux, FOI 1.192 1.21(I)Feedwater system malfunctions that Zero power Minimum DNBR (non-RTDP, WLOP correlation)
See Note 2result in an increase in feedwater flow Maximum linear heat generation, kW/flFull power Minimum DNBR (RTDP, WRB-2 correlation) 2.04 1.52(5)Maximum core average heat flux, FOI 1.098 1.21 ")Excessive increase in secondary steam Bounding Minimum DNBR (RTDP, WRB-2 correlation) 1.97 1.52 5)flow Maximum core average heat flux, FOI 1.11 1.21 "'Inadvertent opening of a SG Bounding Minimum DNBR (non-RTDP, WLOP correlation) 5.10 1.18atmospheric relief or safety valveMaximum linear heat generation, kW/ft 6.924 22.4"1Steam system piping failure (SLB) Zero power Minimum DNBR (non-RTDP, WLOP correlation) 1.80 1.18(core response only) Maximum linear heat generation, kW/ft 15.829 22.4Full power Minimum DNBR (RTDP, WRB-2 correlation) 2.026 1.52"5)Maximum linear heat generation, kW/ft 21.8 22.4"'Loss of external electrical load, turbine Minimum DNBR Minimum DNBR (RTDP, WRB-2 correlation) 1.72 1.52(5)trip, inadvertent closure of main steam Peak RCS Pressure Maximum RCS pressure, psia 2746.8 2750.0isolation valves, and loss of condenser vacuum Peak MSS Pressure Maximum MSS pressure, psia 1297.4 1318.5Loss of non-emergency AC power to Bounding Maximum pressurizer mixture volume, ft3 1623.2 1800.0the station auxiliaries Loss of normal feedwater flow Bounding Maximum pressurizer mixture volume, ft3 1384.1 1800.0Feedwater system pipe break Bounding Minimum margin to hot leg saturation,
'F 40.5 >0.0Partial loss of forced reactor coolant Bounding Minimum DNBR (RTDP, WRB-2 correlation) 1.82 1.52...flowComplete loss of forced reactor coolant Bounding Minimum DNBR (RTDP, WRB-2 correlation) 1.69 1.52(5)flowWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-29WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-29Table 2.1-6 Non-LOCA Results Summary (cont.)Safety Analysis Safety AnalysisEvent Case Distinction Parameter Description Result LimitRCP shaft seizure (locked rotor) and DNB Maximum number of rods-in-DNB,
% 0.7 5.0RCP shaft break Peak RCS Maximum RCS pressure, psia 2675.1 2750.0pressure/PCT Maximum cladding temperature,
'F 1786.6 2700.0Maximum zirconium-water
: reaction,
% of zirconium 0.29 16.0reacted by weightUncontrolled RCCA bank withdrawal Bounding Minimum DNBR below first mixing vane grid 1.83 1.13from a subcritical or low power startup (non-RTDP, ABB-NV correlation) condition Minimum DNBR above first mixing vane grid 1.66 1.17(non-RTDP, WRB-2 correlation)
Maximum fuel centerline temperature,
'F 2342 4800.014, Uncontrolled RCCA bank withdrawal Minimum DNBR Minimum DNBR (RTDP, WRB-2 correlation)
See Note 6 1.52("at power Maximum core average heat flux, fraction of 1.183 1.21...analyzed full powerPeak RCS Pressure Maximum RCS pressure, psia 2707.4 2750.0Dropped RCCA(s) and Dropped RCCA All Minimum DNBR (RTDP, WRB-2 correlation)
>1.52 1.5215,bank Maximum linear heat generation, kW/ft <22.4 22.4(3"Maximum uniform cladding strain, % <1.0 1.0Statically misaligned RCCA All Minimum DNBR (RTDP, WRB-2 correlation)
>1.52 1.52 15Single RCCA withdrawal Manual Rod Maximum number of rods-in-DNB,
% <5.0 5.0ControlStartup of an inactive RCP at an No analysis was performed; incorrect temperature see Section 2.5.4, "Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature."
WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-30WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-30Table 2.1-6 Non-LOCA Results Summary (cont.)Safety Analysis Safety AnalysisEvent Case Distinction Parameter Description Result LimitCVCS malfunction that results in a Mode 1-manual Minimum time from alarm to loss of shutdown 50.3 15decrease in the boron concentration in Mode 1-auto margin, minutes 112.9the reactor coolant (boron dilution)
Mode 2 56.0Mode 3 15.6Mode 4 15.8Mode 5 15.7Mode 6 No analysis was performed.
N/A N/ASpectrum of RCCA ejection accidents Full power Maximum fuel pellet average enthalpy, cal/gm 176.4 200.0Maximum fuel melt at the hot spot, % 4.62 10.0Zero power Maximum fuel pellet average enthalpy, cal/gm 145.2 200.0Maximum fuel melt at the hot spot, % 0.0 10.0Inadvertent operation of the ECCS Bounding Maximum pressurizer mixture volume, ft3 1786.5 1800.0during power operation CVCS malfunction that increases Bounding Minimum time from alarm to filling the pressurizer 8.5 8.0reactor coolant inventory water-solid, minutesInadvertent opening of a pressurizer Bounding Minimum DNBR (RTDP, WRB-2 correlation) 2.00 1.52 51safety or relief valveATWS Bounding Maximum RCS pressure, psia 3129.0 3215.0Notes:1. The 1.21 fraction of initial power (or analyzed full power for part-power conditions) limit was confirmed to be less than that which would correspond to melting conditions at the fuel centerline.
: 2. The results for this case were determined to be bounded by the results of the zero power steam system piping failure case.3. Corresponds to a conservative fuel melting temperature of 4700'F associated with a conservative EOC U02 peak bumup at the hot spot of -65,000 MWD/MTU.4. 4800'F is the fuel melting temperature corresponding to an EOC UO2 peak bumup at the hot spot of -48,276 MWD/MTU.5. This SAL DNBR is conservatively used to demonstrate that the DNB design basis is satisfied for analyses performed using RTDP methods.
Sufficient margin is maintained between the SAL DNBR and the design limit DNBR to offset the effects of rod bow, lower plenum flow anomaly, and plant instrumentation biases, as well as to provideflexibility in the design and operation of the plant. See Section 2.12, "'Thermal and Hydraulic Design,"
for additional information.
: 6. As discussed in Section 2.5.2, "Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power," a detailed DNBR analysis of the most limiting case wasperformed using the VIPRE computer code. This was necessary because the minimum DNBR calculated with the RETRAN computer code was less than the SAL DNBR.Per Section 2.12, "Thermal and Hydraulic Design,"
the detailed DNBR analysis confirmed that the DNB design basis is met and sufficient DNBR margin was retained toallow for flexibility in the design and operation of the plant.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-31WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-31=---Minimum Feedback
-Maximum Feedback0.000-0.002-0.004.-0.006" -0.008CD0C.)o -0.0100a.o-0.01210-.WCD -0.014-0.016-0.018-0.020020 40 60 80Power (%)100120Figure 2.1-1 Integrated DPCWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-32WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-32680660h_ 640-0L)600-5800 0.2 0.4 0.6 0.8Fraction of Rated Thermal Power1.2Figure 2.1-2 Reactor Core Safety LimitsWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-33WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-33wL&#xfd;0560 580 600 620 640Tcvg (7F)660Figure 2.1-3 Illustration of OTAT and OPAT Protection WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-341.000.8000U)C0.60L-000.,21..,- 0.400:=00.200.000.00.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0Time From Release (seconds)
Figure 2.1-4 Fractional Rod Insertion versus Time from ReleaseWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-351.001rT0.80 -----------
----------
I--------------------
------ --00 .6 --- ------------------------------------------------I )CDI0.40 ~ ~ --- ----------------- ---------------------------ERo InetoIFato o ulIsrinFigure~~~~
2.- omlzdRC eciiyWrt essFatoa o netoIC--76 N Auus 2013Liesn Reor Reiio WESTINGHOUSE NON-PROPRIETARY CLASS 32-361.000.800S0.60UN0.40E0z0.200.000.00.5 1.0 1.5 2.0 2.5 3.0 3.5Time From Release (seconds) 4.0Figure 2.1-6 Normalized RCCA Reactivity Worth versus Time from ReleaseWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-372.2 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM2.2.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (USAR Section 15.1.1)2.2.1.1 Technical Evaluation 2.2.1.1.1 Introduction The opening of a low-pressure FW heater bypass valve, the tripping of the FW heater drain pumps, orisolating all high-pressure extraction steam will cause a reduction in Tfeed that increases the thermal loadon the primary system. For this event, there is a sudden decrease in Tfeed into the SGs.At power, the increased subcooling caused by the decreased Treed creates a greater load demand on theRCS. With the plant at no-load conditions, the addition of cold FW may cause a decrease in RCStemperature, and thus a reactivity insertion because of the negative MTC of reactivity.
: However, becausethe rate of energy change decreases as the load and FW flow decrease, the no-load transient is less severethan the full-power case.Depending on the magnitude of the temperature decrease and the operation of the automatic rod controlsystem, the net effect on the RCS can be similar to the effect of increasing secondary steam flow; that is,the reactor will reach a new equilibrium condition at a power level corresponding to the new temperature difference across the secondary-side of the SG. For large Tfeed reductions, the OPAT RT function willprevent a power increase that could lead to a DNBR that is lower than the SAL value.2.2.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe decrease in Tfeed event is analyzed to confirm that the minimum DNBR and fuel centerline temperature design bases are met. Therefore, the analysis uses the following key modeling inputs andassumptions:
The RTDP (Reference
: 1) was used. The initial RCS pressure and RCS temperature were assumedto be at the nominal values consistent with steady-state full-power operation.
The reactor coolantMMF was also modeled.
Uncertainties for these initial conditions were accounted for in theDNBR SAL as described in Reference 1.The analyses were performed at an initial NSSS power of 3651 MWt, which includes a nominalreactor coolant pump (RCP) net heat input of 14 MWt and all applicable uncertainties.
The analyses model the WCGS SGs (Westinghouse Model F). An initial water levelcorresponding to the nominal level minus uncertainties was modeled in all four SGs.All Tfeed reduction cases modeled a symmetric decrease in Treed to all four SGs. At the start of thetransient, the FW enthalpy was reduced to bound a temperature reduction of 200'F (step change)and the FW mass flow remained constant throughout the event. The 200'F temperature reduction WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-38conservatively bounds the spurious opening of the low pressure FW heater bypass valve and theresulting bypass of all flow through the low pressure FW heaters.Pressurizer sprays and PORVs were modeled to reduce RCS pressure, resulting in a conservative evaluation of the margin to the DNBR SAL.All Tteed reduction cases were initiated from hot full-power.
Cases modeling manual andautomatic rod control were analyzed.
In addition, sensitivities were analyzed to ensure thatconservative vessel mixing characteristics were used.* The OPAT RT function was credited for this event.Based on its frequency of occurrence, the decrease in Treed event is considered to be a Condition II eventas defined by the ANS document "Nuclear Safety Criteria for the Design of Stationary Pressurized WaterReactor Plants,"
ANSI N 18.2-1973.
As such, the applicable acceptance criteria for this incident are:Pressures in the RCS and the MSS should be maintained below 110 percent of the respective design pressures.
Fuel cladding integrity is maintained by ensuring that the minimum DNBR remains greater thanthe 95/95 DNBR SAL of 1.52. Additionally, fuel melting is precluded by ensuring that themaximum transient core average thermal power does not exceed a value that would result inexceeding the kW/ft value corresponding to fuel centerline melting at the core hot spot. For theWCGS, it has been confirmed that power levels up to 121 percent of the analyzed power levelmeet this criterion.
An incident of moderate frequency should not generate a more serious plant condition withoutother faults occurring independently.
Demonstrating that the pressurizer does not becomewater-solid ensures a more serious plant condition is not generated.
Because this event results in acooldown of the RCS, the reactor coolant experiences a reduction in volume, and therefore pressurizer filling is not a concern.The primary acceptance criteria used in this analysis is that the minimum DNBR remains greater than theSAL and that the maximum transient core average thermal power does not exceed the value that couldpotentially result in fuel melt at the core hot spot. The event does not challenge the primary-orsecondary-side pressure limits because the increased heat removal results in an RCS cooldown.
The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC). Brief discussions ofthe specific GDCs that are related to the FWM event acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the FWM event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-39GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the reactor coolant pressureboundary (RCPB) are not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the FWM event, this is shown to be met bydemonstrating that the peak RCS pressure is less than 110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires that one of thereactivity control systems consists of control rods capable of reliably controlling reactivity changes with appropriate margin for malfunctions like stuck rods so that specified acceptable fueldesign limits are not exceeded under conditions of normal operation, including anticipated operational occurrences.
For the FWM event, this is shown to be met by demonstrating that thefuel cladding integrity is maintained.
2.2.1.1.3 Description of Analyses and Evaluations The excessive heat removal due to a Tfeed reduction transient was analyzed with the RETRAN computercode (Reference 2). This code simulates a multi-loop RCS, core neutron kinetics, the pressurizer, pressurizer relief and safety valves, pressurizer spray and heaters, SGs, and MSSVs. The code computespertinent plant variables including temperatures, pressures, and power level. RETRAN (Reference
: 2) isused to conservatively predict DNBR.The Tfeed reduction analysis accounts for the spurious opening of the low pressure FW heater bypass valvewhich results in a maximum Tfeed reduction of 200'F to all SGs. Cases modeled both automatic andmanual rod control.
All cases assume a conservatively large moderator density coefficient characteristic of end-of-life (EOL) conditions.
2.2.1.1.4 ResultsComparison of results for both analyzed cases confirms that the Tfeed reduction case modeling manual rodcontrol and design vessel mixing coefficients is the most limiting case. This case produces the largestreactivity
: feedback, and therefore results in the greatest power increase.
For both analyzed Tfed reduction transient cases, the reactor trips on the OPAT function which thencauses a consequential turbine trip. Minimum DNBR and peak core average thermal power are reachedshortly after the RT. Following RT, the event is terminated as a consequence of a SI trip due to lowpressurizer pressure.
This SI trip causes automatic FW isolation ending the event.Table 2.2.1 -1 shows the time sequence of events for the limiting Tfed decrease transient.
Table 2.2.1-2provides minimum DNBR and peak core average thermal power results of both analyzed cases.Figures 2.2.1 -1 through 2.2.1-3 show the transient responses of various system parameters for the limitingTfeed decrease transient.
2.2.1.2 Conclusions For the excessive Tfted decrease event, the results show that the minimum DNBR remains above theapplicable SAL and that the core average thermal power does not exceed a value that results in exceeding WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-40the kW/ft limit corresponding to fuel centerline melting at the core hot spot. Therefore, no fuel damage ispredicted and all applicable acceptance criteria are satisfied for the WCGS. Based on this, it is concluded that the plant will continue to meet the requirements of GDCs 10, 15., and 26.2.2.1.3 References
: 1. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"
April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
May 1999.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-41Table 2.2.1-1 Time Sequence of Events -Decrease In Treed (Manual Rod Control)Event Time (seconds)
Low Pressure FW Heater Bypass Valves Open 00.01OPAT Setpoint Reached in Two Loops 35.2RT (Rod Motion Starts) on OPAT 38.2Minimum DNBR Reached 38.5Low Pressurizer Pressure SI Setpoint Reached 83.1FW Isolation Initiated 100.0Table 2.2.1-2 Decrease in Tfeed Minimum DNBR and Peak Core Average Thermal Power ResultsTime ofMinimum Peak Core Average Time of Peak CoreMinimum DNBR Thermal Power(2) Average Thermal PowerTfeed Decrease Case DNBR(1  (seconds)
(FOI) (seconds)
Automatic Rod Control 1.86 47.5 1.173 47.5Manual Rod Control 1.77 38.5 1.192 38.5Notes:1. The SAL for DNBR is 1.52.2. The SAL for peak core average thermal power (fraction of initial) is 1.21 of the analyzed power level.WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-42WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-4200U0C_.)0 20 40 60 90 100 120 140 160Time (seconds) 0 20 40 60 80 100 120 140 160Time (seconds)
Figure 2.2.1-1 Decrease in Tfeed at Full Power -Nuclear Power and Core Heat Flux versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-43WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-43LaI-V4)0C,)C,)4)U-ci~0ci)0~EI-I-00ci.)-v0ci)1.~0C-)0 20 40 60 80 100 120 140 160Time (seconds)
Knan.1J7V580-570-560-550-540-530-.... ... .... .. ....... ... .. ... .... ...JIM0 20 40 60 80 100Time (seconds) 120 140 160Figure 2.2.1-2Decrease in Treed at Full Power -Vessel Delta-T and Core Average Moderator Temperature versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-44CLa,0,0 20 40 60 80 100Time (seconds12 14 100 20 40 60 8O 100 120 140 160rime (seconds)
Figure 2.2.1-3 Decrease in Trd at Full Power -Pressurizer Pressure and DNBR versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-452.2.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow(USAR Section 15.1.2)2.2.2.1 Technical Evaluation 2.2.2.1.1 Introduction The addition of excessive FW will cause an increase in heat removal from the RCS. An example ofexcessive FW flow would be a full opening of a main FW flow control valve (FCV) due to a FW controlsystem malfunction or an operator error. At power, this excess flow causes a greater load demand on theRCS due to increased subcooling in the SG. With the plant at no-load conditions, the addition of excessFW may cause a decrease in RCS temperature, and thus a reactivity insertion due to the effects of thenegative MTC of reactivity.
2.2.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe FW flow increase event is analyzed to confirm that the minimum DNBR and fuel centerline temperature design bases are met. Therefore, the analysis uses the following key modeling inputs andassumptions:
The RTDP (Reference
: 1) was used for the cases initiated from full power. The initial RCSpressure and RCS temperature were assumed to be at the nominal values consistent withsteady-state full-power operation.
The reactor coolant MMF was also modeled.
Uncertainties forthese initial conditions were accounted for in the DNBR SAL as described in Reference 1.The analyses were performed at an initial NSSS power of 3651 MWt, which includes a nominalRCP net heat input of 14 MWt and all applicable uncertainties.
* The analyses model the WCGS SGs (Westinghouse Model F).For the single-loop FW flow increase event at full-power, one FW control valve was assumed tomalfunction, resulting in a step increase to 200 percent of the nominal full-power FW flow toone SG.For the multiple-loop FW flow increase event at full-power, two FW control valves were assumedto malfunction, resulting in a step increase to 200 percent of the nominal full-power FW flow totwo SGs.The increase in FW flow rate results in a decrease in the Tfeed (enthalpy) due to the reducedefficiency of the FW heaters.
For full power, a 25 Btu/lbm decrease in the FW enthalpy wasconservatively assumed to occur coincident with the FW flow increase.
For the single-loop FW flow increase event at no-load conditions, one FW control valve wasassumed to malfunction, resulting in a step increase to 250 percent of the full-power nominal flowto one SG.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-46For the multiple-loop FW flow increase event at no-load conditions, two FW control valves wereassumed to malfunction, resulting in a step increase to 250 percent of the full-power nominal flowto two SGs.For the cases initiated at zero power, initial reactor power, RCS pressure, and RCS temperature were assumed to be at levels corresponding to no-load conditions.
TDF was also modeled.
Inaddition, the reactor was assumed to be at the minimum shutdown margin condition of-0.013 Ak/k.For the full-power cases, an initial water level corresponding to the nominal level minusuncertainties was modeled in all four SGs, whereas an initial water level corresponding to thenominal level was modeled for the zero-power cases.Pressurizer sprays and PORVs were modeled to reduce RCS pressure, resulting in a conservative evaluation of the margin to the DNBR SAL.The full-power cases were analyzed with manual and automatic rod control.For cases at zero-power conditions, the initial Tfeed was assumed to be 35'F.The heat capacities of the RCS and SG thick metal were not considered, thereby maximizing thepotential temperature reduction of the reactor coolant.Based on its frequency of occurrence, the increase in FW flow event is considered to be a Condition IIevent as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized WaterReactor Plants,"
ANSI N18.2-1973.
As such, the applicable acceptance criteria for this incident are:Pressure in the RCS and MSS should be maintained below 110 percent of the design pressures.
Fuel cladding integrity is maintained by ensuring that the minimum DNBR remains greater thanthe 95/95 DNBR SAL of 1.52 in the limiting fuel rods and that the centerline temperature of thefuel rods with the peak linear heat rate (kW/ft) does not exceed the UO2 melting temperature.
Fuel melting is precluded by ensuring that the maximum transient core average thermal powerdoes not exceed a value that would result in exceeding the kW/ft value corresponding to fuelcenterline melting at the core hot spot. For the WCGS, it has been confirmed that power levels upto 121 percent of the initial value meet this criterion.
An incident of moderate frequency should not generate a more serious plant condition without otherfaults occurring independently.
Demonstrating that the pressurizer does not become water-solid ensures a more serious plant condition is not generated.
Because this event results in a cooldown ofthe RCS, the reactor coolant volume decreases, and therefore pressurizer filling is not a concern.The primary acceptance criterion used in this analysis is that the minimum DNBR remains greater thanthe SAL, thus ensuring fuel cladding integrity is maintained.
The fuel cladding integrity is also assured byensuring that the maximum transient core average thermal power does not exceed the value that wouldresult in exceeding the kW/ft value corresponding to fuel centerline melting at the core hot spot. TheWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32 -47event does not challenge the primary-or secondary-side pressure limits because the increased heatremoval results in an RCS cooldown.
The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the FWM event acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the FWM event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.
GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.
For theFWM event, this is shown to be met by demonstrating that the peak RCS pressure is less than110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires that one of thereactivity control systems consist of control rods capable of reliably controlling reactivity changeswith appropriate margin for malfunctions like stuck rods so that specified acceptable fuel designlimits are not exceeded under conditions of normal operation, including anticipated operational occurrences.
For the FWM event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.
Note that a RT is not modeled in the FWM analysis performed for theWCGS.2.2.2.1.3 Description of Analyses and Evaluations The excessive heat removal due to a FW flow increase transient was analyzed with the RETRANcomputer code (Reference 2). This code simulates a multi-loop RCS, core neutron kinetics, thepressurizer, pressurizer relief and safety valves, pressurizer spray and heaters, SGs, and MSSVs. The codecomputes pertinent plant variables including temperatures, pressures, and power level.The VIPRE computer code (Reference
: 3) is used to verify that the DNBR remains above the DNBR SALfor hot zero power (HZP) cases. For hot full power (HFP) cases, RETRAN (Reference
: 2) is used toconservatively predict DNBR.The excessive FW flow event assumes an accidental opening of one or more FW control valves with thereactor at full- and zero-power conditions, and with automatic and manual rod control, where applicable.
Both the automatic and manual rod control cases assume a conservatively large moderator densitycoefficient characteristic of EOL conditions.
Table 2.2.2-1 summarizes the analyzed cases.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-482.2.2.1.4 ResultsFor the cases initiated at HFP conditions, a comparison of the multiple-loop (failure of two FW controlvalves) and single-loop (failure of one FW control valve) cases demonstrates that the two-loop failurecase with manual rod control is more limiting.
The two-loop FW flow increase case with manual rodcontrol produces the largest reactivity
: feedback, and therefore results in the greatest power increase.
The cases initiated at HZP conditions are less limiting than the HZP SLB analysis described inSection 2.2.5. Therefore, the results of this case are not presented.
Continuous addition of excessive FW is prevented by the SG high-high level trip, which initiates FWI andtrips the turbine and main FW pumps. Subsequent to FWI initiated by a SG high-high level trip, thereactor continues to operate until the low-low SG level setpoint is reached.
: However, the RT on low-lowSG level is not modeled in the analysis because it occurs after the time of interest for the event.Table 2.2.2-2 shows the time sequence of events for the limiting multi-loop, full-power FW flow increasetransient with manual rod control; Table 2.2.2-3 provides minimum DNBR and peak core average thermalpower results of all cases. Figures 2.2.2-1 through 2.2.2-4 show the transient responses of various systemparameters for the limiting multi-loop FW flow increase initiated from full-power conditions with manualrod control.2.2.2.2 Conclusion For the excessive increase in FW flow event, the results show that the DNBRs encountered are above theapplicable SAL value and that the core average thermal power does not exceed a value that results inexceeding the kW/ft limit corresponding to fuel centerline melting at the core hot spot. Therefore, no fueldamage is predicted and all applicable acceptance criteria are satisfied for the WCGS. Based on this, it isconcluded that the plant will continue to meet the requirements of GDCs 10, 15, and 26.2.2.2.3 References
: 1. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"
April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
May 1999.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-49Table 2.2.2-1 Increase in FW Flow Cases AnalyzedCase Power Level Failure Affected Loop(s) Rod Control1 HFP 1 FCV Loop 1 Manual2 HFP 1 FCV Loop 1 Automatic 3 HFP 2 FCVs Loops 1 and 2 Manual4 H-FP 2 FCVs Loops 1 and 2 Automatic 5 HZP 1 FCV Loop 1 Manual6 HZP 2 FCVs Loops 1 and 2 ManualTable 2.2.2-2 Time Sequence of Events -Increase in FW Flow (HFP, Multi-Loop, ManualRod Control)Event Time (seconds)
Two FW Control Valves Fail Full-Open (Event Initiation) 0.01SG Level Reaches High-High Setpoint of 100% NRS 36.9Turbine Trip Initiated (from High-High SG Level Trip) 39.3Minimum DNBR Occurs 41.5FWI Initiated (from High-High SG Level Trip) 53.8Table 2.2.2-3 HFP FWM Flow Increase Minimum DNBR and Peak Core Average Thermal PowerResultsTime of Peak CoreTime of Peak Core Average Average ThermalHFP FW Flow Increase Minimum Minimum DNBR Thermal Power(2) PowerCase DNBR(1) (seconds)
(FOI) (seconds)
Single Loop Auto Control 2.15 26.0 1.046 39.5Single Loop Manual Control 2.10 28.0 1.073 45.5Multiple Loop Auto Control 2.11 39.5 1.065 44.5Multiple Loop Manual 2.04 41.5 1.098 45.0ControlNotes:1. The SAL for DNBR is 1.52.2. The SAL for peak core average thermal power (FOI) is 1.21.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-50WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-504 ,~,.i.'016.110A,...... ......................................
..............0A"U.S020 40 60 80 100Time (seconds) 120 140160OSi.I.40C4-0C0C-)04-'C1=04)4)0C-)1-...........
.........................
n-j.%f.&0 20 40 60 80 100Time (seconds) 120 140 160Figure 2.2.2-1Increase in FW Flow at Full Power -Multi-Loop Manual Rod ControlNuclear Power and Core Heat Flux Versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-51WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-512600'E594-590............................................... ...................................................................................................................................................... ....... .................imOo 20 40 6O 80 100Time (seconds) 120 140160af)cocaco60 80 100Time (seconds)
Figure 2.2.2-2Increase in FW Flow at Full Power -Multi-Loop Manual Rod ControlCore Average Moderator Temperature and Pressurizer Pressure Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-52WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-52L oo ps 1 Loops 3-4S140DOO......
t12DDDD041000D-01002o 1 0 20.. .. ...100-~1000.. ......9150-60 80 00Time (seconds)
Figure 2.2.2-3Increase in FW Flow at Full Power -Multi-Loop Manual Rod ControlSG Mass Inventory and Pressure Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-53WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-530 20 40 60 80 100 120 140 160Time (seconds)
Figure 2.2.2-4Increase in FW Flow at Full Power -Multi-Loop Manual Rod ControlDNBR Versus TimeWCAP- 17658-NP August 2013WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-542.2.3 Excessive Increase in Secondary Steam Flow (USAR Section 15.1.3)2.2.3.1 Technical Evaluation 2.2.3.1.1 Introduction An excessive load increase incident is defined as a rapid increase in steam flow that causes a powermismatch between the reactor core power and the SG load demand. The RCS is designed to accommodate a 10 percent step-load increase or a 5percent-per-minute ramp-load increase in the range of 15 to100 percent of full power. Any loading rate in excess of these values may cause a RT actuated by the RTsystem. If the load increase exceeds the capability of the RCS, the transient would be terminated insufficient time to prevent the DNB design basis from being violated.
This incident could result from eitheran administrative violation such as excessive loading by the operator or an equipment malfunction in thesteam bypass control system, or turbine speed control.During power operation, steam dump to the condenser is controlled by reactor coolant condition signals,such as a high reactor coolant temperature, which indicates a need for steam dump. A single controller malfunction will not cause steam dump valves to open; an interlock is provided that blocks the opening ofthe valves unless a large turbine load decrease or a turbine trip has occurred.
For all cases, the plantrapidly reaches a stabilized condition at a higher power level. Normal plant operating procedures wouldbe followed to reduce power. The excessive load increase incident is an overpower transient for which thefuel temperatures will rise. RT may not occur for some cases, and the plant will reach a new equilibrium condition at a higher power level corresponding to the increase in steam flow. Protection against anexcessive load increase
: incident, if necessary, is provided by the following RT signals:* OPAT* OTAT* Power range high neutron flux2.2.3.1.2 Input Parameters, Assumptions, and Acceptance CriteriaAn evaluation was performed to show that the DNB design basis is satisfied for the excessive loadincrease incident.
Key aspects of the evaluation are provided below.The RTDP (Reference
: 1) was applied.
Initial reactor power, RCS pressure, and RCS temperature wereassumed to be at their nominal values, consistent with steady-state full-power operation.
MMF was alsoassumed.
Uncertainties in initial conditions were accounted for in the safety analysis DNBR limit value,as described in Reference 1.The evaluation was performed for a step-load increase of 10 percent steam flow from 100 percent of corepower.The higher end of the Ta,,g range (570.77F to 588.4&deg;F) is applied because it minimizes the initial margin tothe DNBR limit.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-55The higher end of the full-power T'eed range (400.0&deg;F to 448.6&deg;F) is applied, although the event is not verysensitive to Tfeed.Zero percent SGTP is modeled because this maximizes the primary-to-secondary heat transfer area, whichis conservative for maximizing the cooldown of the RCS.The pressurizer heaters are not modeled because the pressurizer heaters would actuate to try and raise thepressurizer
: pressure, which is not conservative with respect to minimizing DNBR.The pressurizer sprays and PORVs are modeled to limit any RCS pressure increase.
A lower RCSpressure is conservative for DNBR calculations.
Although the OTAT, OP AT, and power range high neutron flux RTs are available to mitigate the event,the analysis conservatively does not credit these trips.No credit is taken for the heat capacity of the RCS and SG metal mass in attenuating the resulting plantcooldown.
This event is analyzed with automatic and manual rod control.Because the event is not sensitive to the initial pressurizer and SG levels, the pressurizer level and SGlevel are modeled to be at the nominal values consistent with steady-state full-power operation.
The event is analyzed for both the beginning-of-life
((BOL) minimum reactivity feedback) and EOL(maximum reactivity feedback) conditions.
Based on its frequency of occurrence, the excessive load increase incident is considered to be aCondition II event as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
The following items summarize the acceptance criteria associated with this event:* The critical heat flux (CHF) should not be exceeded.
This is met by demonstrating that theminimum DNBR does not go below the SAL value at any time during the transient.
* Pressures in the RCS and MSS should be maintained below 110 percent of the respective designpressures.
* The peak linear heat generation rate (expressed in kW/ft) should not exceed a value that wouldcause fuel centerline melt. This criterion is satisfied by demonstrating that the core average heatflux remains below the limit of 121 percent of the applied nominal core thermal power during theevent.* An incident of moderate frequency should not generate a more serious plant condition withoutother faults occurring independently.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-562.2.3.1.3 Description of Analyses and Evaluations The excessive load increase event is analyzed using the RETRAN computer code described inWCAP-14882-P-A (Reference 2). The RETRAN code model simulates the RCS, neutron kinetics, pressurizer, pressurizer relief and safety valves, pressurizer
: heaters, pressurizer spray, SG, FWsystem, andMSSVs. The code computes pertinent plant variables including SG mass, pressurizer water volume,reactor coolant average temperature, RCS pressure, and SG pressure.
For the minimum reactivity feedback cases, the core has the least-negative MTC of reactivity and theleast-negative Doppler-only power coefficient curve, and, therefore, the least-inherent transient responsecapability.
For the maximum reactivity feedback cases, the core has the most-negative MTC of reactivity and the most-negative Doppler-only power coefficient curve. This results in the largest amount ofreactivity feedback due to changes in coolant temperature.
Normal reactor control systems and engineered safety systems are not required to function.
2.2.3.1.4 ResultsThe analysis results for the 10 percent load increase event from full-power conditions show that in allcases analyzed the minimum DNBR remains above the SAL value and the peak linear heat generation does not exceed the limit value, thus demonstrating that the fuel cladding integrity and fuel centerline melt acceptance criteria are met. The peak pressurizer water volume remains below the total volume ofthe pressurizer, demonstrating that this event does not generate a more serious plant condition.
Following the initial load increase, the plant reaches a stabilized (steady-state) condition.
The increase in the MSSflow rate results in a cooldown of the RCS and a decrease in the MSS pressure.
The RCS and MSSpressure limits are not challenged during the event. The analysis inputs are intended to minimize theresultant minimum DNBR and not to maximize RCS and MSS pressures.
The case that models minimum reactivity feedback conditions with automatic rod control is the mostlimiting case with respect to minimum DNBR. The key results are summarized in Table 2.2.3-1.
The timesequence of events for each case is provided in Table 2.2.3-2.
The transient responses for the four casesare shown in Figures 2.2.3-1 through 2.2.3-4.2.2.3.2 Conclusion The excessive load increase analysis demonstrates that for this event at the WCGS, the DNBR does notdecrease below the SAL value at any time during the transient for all cases. Also, the peak core averagepower (heat flux) remains below the limit of 121 percent of the applied nominal core thermal power; thus,no fuel or cladding damage is predicted.
The event does not challenge the primary and secondary sidepressure limits because the increased heat removal cools the RCS and depressurizes the MSS. The peakpressurizer water volume remains below the total volume of the pressurizer, demonstrating that this eventdoes not generate a more serious plant condition.
All applicable acceptance criteria are met for theWCGS.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-572.2.3.3 References
: 1. WCAP-11397-P-A (Proprietary) and WCAP- 11397-A (Non-Proprietary),
"Revised ThermalDesign Procedure,"
April 1989.2. WCAP- 14882-P-A (Proprietary) and WCAP- 15234-A (Non-Proprietary),
"RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA SafetyAnalyses,"
April 1999 and May 1999, respectively.
WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-58Table 2.2.3-1 Excessive Load Increase Incident Summary of ResultsCore Heat FluxCase Minimum DNBR (FOI)Limits 1.52 1.21Minimum reactivity
: feedback, manual rod control 2.29 1.02Minimum reactivity
: feedback, automatic rod control 1.97 1.11Maximum reactivity
: feedback, manual rod control 2.03 1.10Maximum reactivity
: feedback, automatic rod control 2.00 1.10WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-59WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-59Table 2.2.3-2 Time Sequence of Events for the Excessive Load Increase IncidentCase Event Time of Event (seconds)
Minimum reactivity
: feedback, manual rod 10% step load increase 0.0controlMinimum DNBR reached 6.7Steady-state conditions reached -300"(approximate)
Peak heat flux reached 382.5Minimum reactivity
: feedback, automatic 10% step load increase 0.0rod controlPeak heat flux reached 265.6Steady-state conditions reached -300"(approximate)
Minimum DNBR reached 377.6Maximum reactivity
: feedback, manual rod 10% step load increase 0.0controlSteady-state conditions reached -350("(approximate)
Peak heat flux reached 362.8Minimum DNBR reached 395.7Maximum reactivity
: feedback, automatic 10% step load increase 0.0rod controlPeak heat flux reached 368.5Minimum DNBR reached 398.2Steady-state conditions reached -400")(approximate)
Note:I. Time of equilibrium (steady-state conditions reached) was selected based on when the nuclear power leveled out aftertransient initiation.
WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-60a,=3EnCn10501000950S900,9o0i 850" 800"750'0 100 200 300Time (s)400Figure 2.2.3-110% Step Increase in Heat Removal by Secondary SystemMinimum Reactivity
: Feedback, Manual Reactor ControlWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-61WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-61100 200 300Time (s)400595i-590-585", 580"--575".570"in 565-.. ........................ ...........
.. ..................................
.....................................
..............
.....................................................................................................................................I..0100200Time (s)300400Figure 2.2.3-110% Step Increase in Heat Removal by Secondary SystemMinimum Reactivity
: Feedback, Manual Reactor Control (cont.)WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-62a)00~0U).- 2260"0_2250-I 2240"2230-2220080-1070"1060'E-Z 1050"" " 1040"1030--10200 l2O0 100 200 300Time (s)400Figure 2.2.3-210% Step Increase in Heat Removal by Secondary SystemMinimum Reactivity
: Feedback, Automatic Reactor ControlWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-630 100 200 300Time (s)400rnrEC3Uf)590-585-580-575-570-565-...........................................................
............................................................................
..........................................................F .............................-JVu01002;0Time (s)300400Figure 2.2.3-210% Step Increase in Heat Removal by Secondary SystemMinimum Reactivity
: Feedback, Automatic Reactor Control (cont.)WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-64C:)(-0nCd,f-)200Time (s)Figure 2.2.3-3 10% Step Increase in Heat Removal by Secondary SystemMaximum Reactivity
: Feedback, Manual Reactor ControlWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-65z:0 100 200 300Time (s)400rnr590-585-580-E575-a:)> 570-g' 565-............................................
.............
.........................................................
.. ..............................................................................JDoU0100200Ti me (s)300400Figure 2.2.3-310% Step Increase in Heat Removal by Secondary SystemMaximum Reactivity
: Feedback, Manual Reactor Control (cont.)WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-660a)00~0a)=3.225022452240"&_ 2235-2230106010501040"011030"U 1020"(n9a)0'i200Time (s)Figure 2.2.3-410% Step Increase in Heat Removal by Secondary SystemMaximum Reactivity
: Feedback, Automatic Reactor ControlWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-67WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-67Z0 100 200 300Time (s)400595: 590-585-_ 580-EF---575-> 570-o565-.................
...............
.................
.................
.................
...........................
............
.......................................................................................................0100200Time (s)300400Figure 2.2.3-4 10% Step Increase in Heat Removal by Secondary SystemMaximum Reactivity
: Feedback, Automatic Reactor Control (cont.)WCAP-17658-NP Licensing ReportALIgust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-682.2.4 Inadvertent Opening of a Steam Generator Atmospheric Relief or Safety Valve(USAR Section 15.1.4)2.2.4.1 Technical Evaluation 2.2.4.1.1 Introduction The most severe core conditions resulting from an accidental depressurization of the MSS are associated with an inadvertent opening of a SG atmospheric relief or safety valve. More specifically, an accidental depressurization of the MSS is a transient that is analyzed to bound the opening of a single turbine bypassvalve or SG ARV because the inadvertent opening of one of these control valves is most likely to occur.Conversely, the inadvertent opening of a SG safety valve is not nearly as likely to occur because thedesign of a spring-loaded safety valve is passive in nature. However, because the relief capacity of aSG safety valve is larger than those associated with either of the other two types of valves, the failure of aSG safety valve is also conservatively considered in the analysis of an accidental depressurization of theMSS. The analyses that consider a major rupture of a main steam pipe are presented in Section 2.2.5.The steam release, as a consequence of an accidental depressurization of the MSS, results in an initialincrease in steam flow, followed by a decrease in steam flow during the rest of the accident as theSG pressure decreases.
The increased energy removal from the RCS causes a decrease in the reactorcoolant temperature and pressure.
In the presence of a negative MTC, the cooldown results in a positivereactivity insertion.
The primary design features that provide protection for accidental depressurizations of the MSS are:Actuation of the SI system on any of the following:
-Two-out-of-four low pressurizer pressure signals-Two-out-of-three low steam line pressure signals in any one loopActuation of a RT from the overpower signals (neutron flux and AT) or upon the receipt of an SIsignal.Redundant isolation of the main FW lines to prevent sustained main FW flow, which would causeadditional cooldown.
In addition to the primary means of protection, where an SI signal closes themain FW isolation valves, an SI signal will also rapidly close all main FW control valves andcontrol bypass valves, trip the main FW pumps, and close the FW pump discharge valves.Closure of the fast-acting MSIVs on the following:
-Two-out-of-three low steam line pressure signals in any one loop(above Permissive P-1 1)-Two-out-of-three high negative steam pressure rate signals in any one loop(below Permissive P-I 1)WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-692.2.4.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe following summarizes the major input parameters and/or assumptions used in the analysis of anaccidental depressurization of the MSS event at HZP conditions:
HZP conditions were modeled with four loops in service and with offsite power available.
A steam flow rate of approximately 240 Ibm/sec at 1075 psia (corresponding to 268 lbm/sec at1200 psia) was analyzed.
This flow rate corresponds to the maximum capacity of any singleturbine bypass, atmospheric relief, or safety valve.Minimum SGTP (0 percent) was modeled to conservatively maximize primary-to-secondary heattransfer.
* All control rods were modeled to be inserted except the most reactive RCCA, which was assumedto be stuck out of the core.* A minimum, EOL shutdown margin corresponding to 1.30% Ak/k was modeled at eventinitiation.
* The SI system was modeled with a conservatively low flow capability, corresponding to only onehigh-head SI (centrifugal charging) pump injecting through the cold legs (CLs).* The flow from the SI system that is delivered to the RCS was modeled with a temperature andboron concentration consistent with the minimum values for the refueling water storage tank, asrequired by the TS.* The low pressurizer pressure signal was credited for SI system actuation.
In addition, the SIsignal that results from the low pressurizer pressure signal was credited for isolation of the mainFW lines.* The accumulators were modeled to be available;
: however, CL pressures never decreased to thepoint where flow from the accumulators was injected.
* The AFW system was modeled with a conservatively high flow capability, corresponding to allAFW pumps operating at maximum capacity and the maximum flow possible being delivered tothe faulted SG.An accidental depressurization of the MSS is classified as a Condition II event, an incident of moderatefrequency, as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
The pressure limits for the primary and secondary systems are not challenged for this accident because thepressures in these systems each decrease from their initial values during the transient.
The only criterion that has the potential to be challenged during this event is that associated with fuel damage. The analysisdemonstrates that this criterion is met by showing that the DNB design basis is met. That is, this analysisWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-70shows that the minimum DNBR does not go below the limit value at any time during the transient.
Inaddition, it has been historical practice to assume that fuel failure will occur if centerline melting takesplace. Therefore, the analysis also demonstrates that the peak linear heat generation rate (expressed in kW/ft) does not exceed the value that would cause fuel centerline melt.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the accidental depressurization of the MSS event acceptance criteria areprovided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the accidental depressurization of the MSS event, this is shown to be met bydemonstrating that the fuel damage criterion is satisfied.
GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.
For theaccidental depressurization of the MSS event, this is shown to be met by demonstrating that thepeak RCS pressure is less than 110 percent of the design pressure.
GDC 20 (Protection System Functions) requires that the protection system be designed to initiateautomatically the operation of appropriate systems including the reactivity control systems, sospecified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences, and to sense accident conditions and initiate the operation of systems andcomponents important to safety. For the accidental depressurization of the MSS event, this isshown to be met by demonstrating that the fuel damage criterion is satisfied.
GDC 26 (Reactivity Control System Redundancy and Capability) requires that one of thereactivity control systems uses control rods capable of reliably controlling reactivity changes withappropriate margin for malfunctions like stuck rods so that specified acceptable fuel design limitsare not exceeded under conditions of normal operation, including anticipated operational occurrences.
For the accidental depressurization of the MSS event, this is shown to be met bydemonstrating that the fuel damage criterion is satisfied.
2.2.4.1.3 Description of Analyses and Evaluations A detailed analysis was performed using the RETRAN computer code (Reference
: 1) to determine theplant transient conditions following an accidental depressurization of the MSS. The RETRAN modelsimulates the core neutron kinetics, RCS, pressurizer, SGs, SI system and AFW system. To properlymodel the system response to this event and prevent any non-physical behavior from being predicted when the pressurizer
: refills, the pressurizer and surgeline were modeled as a single volume, as comparedto the nodalization documented in Reference
: 1. The code computes pertinent plant variables, including the core heat flux and reactor coolant temperature and pressure.
A detailed core analysis was thenperformed using the ANC code (Reference
: 2) to confirm the validity of the RETRAN-predicted reactivity WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-71feedback model. The core models developed in ANC were also used to calculate the power peakingfactors that were used as input to the detailed T/H digital computer code, VIPRE (Reference 3), whichwas used with a DNB correlation applicable to the low pressure condition (Reference
: 4) to determine ifthe DNB design basis was met. In addition, the core models developed in ANC were used to calculate thepeak linear heat generation rate.2.2.4.1.4 ResultsThe calculated sequence of events for an accidental depressurization of the MSS at HZP initial conditions with offsite power available is shown in Table 2.2.4-1, and the limiting results are presented inTable 2.2.4-2.Figures 2.2.4-1 through 2.2.4-7 show the transient results for an accidental depressurization of the MSS.Because offsite power was assumed to be available in this analysis, there is full reactor coolant flow.If the core were to be critical at or near HZP conditions when the accidental depressurization occurs, theinitiation of SI via a low pressurizer pressure signal would trip the reactor.
In addition, sustained main FWflow is prevented by the isolation of the main FW lines on the SI signal that results from the lowpressurizer pressure signal.As shown in Figure 2.2.4-4, the core attains criticality with the RCCAs inserted (i.e., with the plant shutdown assuming one stuck RCCA) before the transient is effectively terminated by boron injected from theSI system.The results of the analysis of an accidental depressurization of the MSS event demonstrate that the DNBdesign basis is met. The calculated minimum DNBR is well above the limit value. In addition, the peaklinear heat generation rate (expressed in kW/ft) does not exceed the value that would cause fuel centerline melt. The pressure limits for the primary and secondary systems are not challenged for this accidentbecause the pressures in these systems each decrease from their initial values during the transient.
Therefore, this event does not adversely affect the core or the RCS, and all applicable acceptance criteriaare met.2.2.4.2 Conclusion The analysis of the accidental depressurization of the MSS described above has been reviewed.
It isconcluded that the analysis has adequately accounted for operation of the plant at the analyzed powerlevel and was performed using acceptable analytical models. It is further concluded that the analysis hasdemonstrated that the reactor protection and safety systems will continue to ensure that the applicable safety analysis design limits and the RCPB pressure limits will not be exceeded as a result of this event.Based on this, the conclusion is that the plant will continue to meet the requirements of GDCs 10, 15, 20,and 26.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-722.2.4.3 References
: 1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.2. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.4. WCAP-14565-P-A Addendum 2-P-A, "Addendum 2 to WCAP-14565-P-A Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low PressureApplications,"
April 2008.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-73Table 2.2.4-1 Time Sequence of Events -Accidental Depressurization of the MSS at HZP Conditions Case Event Time (sec)Inadvertent Opening of a Accidental Depressurization of the MSS Occurs 0.0Single Turbine Bypass,Atmospheric Relief, or Safety Pressurizer Empties 209.5Valve Low Pressurizer Pressure Setpoint Reached 216.3SI Signal Generated (on low pressurizer pressure) 218.3FW Isolation Complete 233.3SI Flow Initiated 243.3Core Re-criticality Occurs 270.5Borated Water from SI System Reaches the Core 578.8Peak Core Heat Flux Reached 581.5Core Becomes Subcritical 590.5Table 2.2.4-2 Limiting Results -Accidental Depressurization of the MSS at HZP Conditions Case Parameter Analysis Value LimitInadvertent Opening of a Single Minimum DNBR 5.10 1.18Turbine Bypass, Atmospheric Relief, or Safety Valve Peak Linear Heat Generation 6.924 22.4(kW/ft)WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-74E00<C)U-0.-50 100 200 300 400 500 600 700Time (sec)E000C.)U-2,i,0,-t-0 100 200 300 400Time (sec)500 600 700Figure 2.2.4-1Accidental Depressurization of the MSS at HZPNuclear Power and Core Heat Flux versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-75Faulted LoopIntact LoopsC-0,E(D0)0)E0)c-0)C,)a-)0 100 200 300 400Time (sec)500 600 700Qnn .U-0)v- 550-E 500-I--0).......................................'fU0 100 200 300Time1400(sec)500 600700Figure 2.2.4-2Accidental Depressurization of the MSS at HZPReactor Vessel Inlet Temperature and Core Average Temperature versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-76WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-76Cl)(r)CL):30nCl)Cl)E-5a,0 IO0 200 300 400Time (sec)500 600 7000 100 200 300 400Time (sec)500 600 700Figure 2.2.4-3Accidental Depressurization of the MSS at HZPPressurizer Pressure and Pressurizer Water Volume versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-77F-M_0000Of0 100 200 300 400Time (sec)500 600 7000 100 200 300 400Time (sec)500 600 700Figure 2.2.4-4Accidental Depressurization of the MSS at HZPCore Boron Concentration and Reactivity versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-78WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-78Foul t ed LoopI... Intact Loopsca-)Cl)UPECD,00 100 200 300 400 500 600 700Time (sec)0 100 200 300 400Time (sec)500 600 700Figure 2.2.4-5Accidental Depressurization of the MSS at HZPSteam Pressure and Steam (Break) Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-79Foul ted LoopIntact Loops')1A-.0200150Eo 100-50ELtE .... ... 1.............
.............
.............
I~~ ~ ............U1 I00 200 300Time400(sec)500 600700Foul ted LoopIntact LoopsLJUUUUu~00E-4--U-)200000-150000-.............................
......................................
1 M nn ..I ..I .IIUUUUU0 100 200 300 400Time (sec)500 600700Figure 2.2.4-6Accidental Depressurization of the MSS at HZPFW Flow and SG Mass versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-80WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-801.4U--.0U-01.2-10.80.6-0.4-0.2-..........I I I lU0 100 200 300Time400(sec)500 600700Figure 2.2.4-7Accidental Depressurization of the MSS at HZPCore Flow versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCkP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-812.2.5 Steam System Piping Failure (USAR Section 15.1.5)2.2.5.1 Steam System Piping Failure at Hot Zero Power Conditions 2.2.5.1.1 Technical Evaluation 2.2.5.1.1.1 Introduction The steam release following a main steam pipe rupture would result in an initial increase in steam flowthat decreases during the accident as the steam pressure decreases.
The increased energy removal from theRCS causes a decrease in the reactor coolant temperature and pressure.
In the presence of a negativeMTC, the cooldown results in a positive reactivity insertion and subsequent reduction in core shutdownmargin. If the most-reactive RCCA is assumed stuck in its fully withdrawn position after RT, there is anincreased possibility that the core will become critical and return to power. A return to power following asteam pipe rupture is a concern primarily because of the high power peaking factors that would exist withthe most-reactive RCCA assumed to be stuck in its fully withdrawn position.
The major rupture of a main steam pipe is the most limiting cooldown transient.
It is analyzed at HZPconditions with no decay heat (decay heat would retard the cooldown, thus reducing the potential returnto power). A detailed discussion of this transient with the most limiting break size (i.e., a double-ended rupture) is presented below.The primary design features that provide protection for steam pipe ruptures are:* Actuation of the SI system on any of the following:
-Two-out-of-four low pressurizer pressure signals-Two-out-of-three low steam line pressure signals in any one loop-Two-out-of-three high-1 containment pressure signals* Actuation of a RT from the overpower signals (neutron flux and AT) or upon the receipt of an SIsignal.Redundant isolation of the main FW lines to prevent sustained main FW flow, which would causeadditional cooldown.
In addition to the primary means of protection, where an SI signal closes themain FW isolation valves, an SI signal will also rapidly close all main FW control valves andcontrol bypass valves, trip the main FW pumps, and close the FW pump discharge valves.Closure of the fast-acting MSIVs on the following:
-Two-out-of-three high-2 containment pressure signalsWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-82Two-out-of-three low steam line pressure signals in any one loop(above Permissive P- 1i)Two-out-of-three high negative steam pressure rate signals in any one loop(below Permissive P-11)For any break, in any location, no more than one SG would experience an uncontrolled
: blowdown, even ifone of the MSIVs were to fail to close. For breaks downstream of the MSIVs, closure of all MSIVs wouldcompletely terminate the blowdown from all of the SGs. Thus, even with the worst possible breaklocation (i.e., upstream of an MSIV), only one SG can blow down, minimizing the potential steam releaseand resultant RCS cooldown and depressurization.
The remaining SGs would still be available fordissipation of decay heat after the initial transient is over.Following blowdown of the faulted SG, the plant can be brought to a stabilized, hot standby condition through control of AFW flow and SI flow, as prescribed by plant operating procedures.
The operating procedures call for operator action to limit RCS pressure and pressurizer level by terminating SI flow, andto control SG level and reactor coolant temperature using the AFW system.2.2.5.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe following summarizes the major input parameters and/or assumptions used in the analysis of a majorrupture of a main steam pipe event at HZP conditions:
* HZP conditions were modeled with four loops in service, both with and without offsite poweravailable.
0 A 1.388 ft2 break size was analyzed.
This break size corresponds to the maximum effective throatarea of the integral flow restrictor that is built into the steam outlet nozzle of each SG.0 Minimum SGTP (0 percent) was modeled to conservatively maximize primary-to-secondary heattransfer.
* All control rods were modeled to be inserted except the most reactive RCCA, which was assumedto be stuck out of the core.0 A minimum, end-of-life shutdown margin corresponding to 1.30% Ak/k was modeled at eventinitiation.
* The SI system was modeled with a conservatively low flow capability, corresponding to only onehigh-head SI (centrifugal charging) pump injecting through the CLs.* The flow from the SI system that is delivered to the RCS was modeled with a temperature andboron concentration consistent with the minimum values for the refueling water storagetank (RWST), as required by the TS.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-83The low steam line pressure signal was credited for SI system actuation and closure of theMSIVs. In addition, the SI signal that results from the low steam line pressure signal was creditedfor isolation of the main FW lines.The accumulators were modeled to be available;
: however, the flow that is injected from theaccumulators was conservatively modeled to have a boron concentration of 0.0 ppm.The AFW system was modeled with a conservatively high flow capability, corresponding to allAFW pumps operating at maximum capacity and the maximum flow possible being delivered tothe faulted SG.A major break in a steam system pipe is classified as a Condition IV event, a limiting fault, as defined bythe ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"ANSI N 18.2-1973.
Minor secondary system pipe breaks are classified as ANS Condition III events,infrequent incidents.
: However, the major rupture of a main steam pipe event was conservatively analyzedto meet the more restrictive acceptance criteria associated with a Condition II event.The pressure limits for the primary and secondary systems are not challenged for this accident because thepressures in these systems each decrease from their initial values during the transient.
The only criterion that has the potential to be challenged during this event is that associated with fuel damage. The analysisdemonstrates that this criterion is met by showing that the DNB design basis is met. That is, this analysisshows that the minimum DNBR does not go below the limit value at any time during the transient.
Inaddition, it has been historical practice to assume that fuel failure will occur if centerline melting takesplace. Therefore, the analysis also demonstrates that the peak linear heat generation rate (expressed in kW/ft) does not exceed the value that would cause fuel centerline melt.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the major rupture of a main steam pipe event acceptance criteria areprovided as follows.GDC 27 (Combined Reactivity Control Systems Capability) requires that the reactivity controlsystems be designed to have a combined capability, in conjunction with poison addition by theECCS, of reliably controlling reactivity changes so that, under postulated accident conditions andwith appropriate margin for stuck rods, the capability to cool the core is maintained.
For thesteam system piping failure at HZP conditions event, this is shown to be met by demonstrating that the fuel damage criterion is satisfied, which ultimately ensures that the ability to insertcontrol rods is maintained.
GDC 28 (Reactivity Limits) requires that the reactivity control system be designed withappropriate limits on the potential amount and rate of reactivity increase so that the effects ofpostulated reactivity accidents can neither result in damage to the RCPB greater than limited localyielding, nor sufficiently disturb the core, its support structures, or other reactor pressurevessel (RPV) internals to impair significantly the capability to cool the core. For the steam systempiping failure at HZP conditions event, this is shown to be met by demonstrating that the peakWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-84RCS pressure is less than 110 percent of the design pressure, which ultimately ensures that theRCPB pressure limits are not exceeded.
GDC 35 (Emergency Core Cooling) requires that the RCS and associated auxiliaries be designedwith a safety system able to provide abundant emergency core cooling.
For the steam systempiping failure at HZP conditions event, this is shown to be met by demonstrating that the fueldamage criterion is met, which ultimately shows that the ECCS provides abundant core cooling,even with the most-limiting single failure considered.
2.2.5.1.1.3 Description of Analyses and Evaluations A detailed analysis was performed using the RETRAN computer code (Reference
: 1) to determine theplant transient conditions following a major rupture of a main steam pipe with and without offsite poweravailable.
The RETRAN model simulates the core neutron kinetics, RCS, pressurizer, SGs, SI system, andAFW system. To properly model the system response to this event and prevent any non-physical behaviorfrom being predicted when the pressurizer
: refills, the pressurizer and surgeline were modeled as a singlevolume, as compared to the nodalization documented in Reference
: 1. The code computes pertinent plantvariables, including the core heat flux and reactor coolant temperature and pressure.
A detailed coreanalysis was then performed for the case that assumes offsite power is available using the ANC code(Reference
: 2) to confirm the validity of the RETRAN-predicted reactivity feedback model. The coremodels developed in ANC were also used to calculate the power peaking factors that were used as inputto the detailed T/H digital computer code, VIPRE (Reference 3), which was used with a DNB correlation applicable to the low pressure condition (Reference
: 4) to determine if the DNB design basis was met. Inaddition, core models developed in ANC were used to calculate the peak linear heat generation rate.The detailed core and DNB analyses for the case that assumes a loss of offsite power (LOOP) are notperformed as the transient resulting at low RCS flow conditions has been judged to be less limiting thanthat resulting when full RCS flow is maintained.
This is based on the fact that, as RCS forced flowdecreases, heat transfer across the SG tubes also decreases.
This decrease in heat transfer significantly reduces the rate and magnitude of the RCS cooldown and, consequently, the final return to power level isalso lower. The drop in RCS pressure is not as significant for this case as well. Furthermore, the loss offorced reactor coolant flow allows for more uniform flow and temperature distributions at the lowerreactor plenum such that, as the coolant travels up into the core inlet and through the active core region,the radial temperature asymmetry is not as significant as in the case where forced flow is maintained.
Thisreduced asymmetric temperature distribution results in less significant power peaking factors.The reduced power peaking in the reactor core, combined with a lower return to power and higher RCSpressure, more than offsets the penalty associated with reduced flow at the location of minimum DNBR.As such, the minimum DNBR for such a case would be higher than that calculated for the case whereoffsite power remains available.
This conclusion has also been previously validated with the linkedneutronics and T/H code systems.
Therefore, consistent with the previous discussion, only the DNBR ofthe most limiting case (with offsite power available) is presented herein.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-852.2.5.1.1.4 ResultsThe calculated sequence of events for the complete severance of a main steam pipe at HZP initialconditions, both with and without offsite power available, is shown in Table 2.2.5.1-1.
The results for themost limiting case, the case with offsite power available, are presented in Table 2.2.5.1-2.
Figures 2.2.5.1-1 through 2.2.5.1-7 show the transient results for the case with offsite power available.
Because offsite power was assumed to be available in this case, there is full reactor coolant flow.Figures 2.2.5.1-8 through 2.2.5.1-14 show the transient results for the case without offsite poweravailable.
Because offsite power was assumed to be lost in this case, the RCPs coast down and there is adecrease in reactor coolant flow. As can be seen from Figures 2.2.5.1-1 and 2.2.5.1-8, the return to poweris considerably less significant for the case without offsite power, even when using the conservative RETRAN reactivity model calibrated for conditions where forced RCS flow was maintained.
Similarly, Figures 2.2.5.1-3 and 2.2.5.1-10 show that RCS pressure is higher for the case with reduced RCS flow.If the core were to be critical at or near HZP conditions when the rupture occurs, the initiation of SI via alow steam line pressure signal would trip the reactor.
Steam release from more than one SG is prevented by the automatic closure of the MSIVs in the steam lines on a low steam line pressure signal. In addition, sustained main FW flow is prevented by the isolation of the main FW lines on the SI signal that resultsfrom the low steam line pressure signal.As shown in Figures 2.2.5.1-4 and 2.2.5.1-11 for the cases with and without offsite power available, respectively, the core attains criticality with the RCCAs inserted (i.e., with the plant shut down assumingone stuck RCCA) before the transient is effectively terminated by boron injected from the SI system.The results of the analysis of a major rupture of a main steam pipe event demonstrate that the DNB designbasis is met. The calculated minimum DNBR is well above the limit value for the limiting case thatassumes offsite power is available.
In addition, the peak linear heat generation rate (expressed in kW/ft)does not exceed the value that would cause fuel centerline melt. The pressure limits for the primary andsecondary systems are not challenged for this accident because the pressures in these systems eachdecrease from their initial values during the transient.
Therefore, this event does not adversely affect thecore or the RCS, and all applicable acceptance criteria are met.2.2.5.1.2 Conclusion The analysis of the steam system piping failure at HZP conditions described above has been reviewed.
Itis concluded that the analysis has adequately accounted for operation of the plant at the analyzed powerlevel and was performed using acceptable analytical models. It is further concluded that the analysis hasdemonstrated that the reactor protection and safety systems will continue to ensure that the ability toinsert control rods is maintained, the RCPB pressure limits will not be exceeded, and abundant corecooling will be provided.
Based on this, the conclusion is that the plant will continue to meet therequirements of GDCs 27, 28 and 35.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-862.2.5.1.3 References
: 1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.2. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.3. WCAP-14565-P-A, "VIPRE-0l Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.4. WCAP- 14565-P-A Addendum 2-P-A, "Addendum 2 to WCAP- 14565-P-A Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low PressureApplications,"
April 2008.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-87Table 2.2.5.1-1 Time Sequence of Events -Steam System Piping Failure at HZP Conditions Case Event Time (sec)Double-Ended Rupture SLB Occurs 0.0(1.388 ft2) with Offsite PowerAvailable Low Steam Line Pressure Setpoint Reached in the FaultedLoop (lead-lagged) 0.6SI Signal Generated (on low steam line pressure) 2.6Pressurizer Empties 15.8Steam Line and FW Isolation Complete 17.6Core Re-criticality Occurs 19.3SI Flow Initiated 27.6Peak Core Heat Flux Reached 320.0Borated Water from SI System Reaches the Core 320.8Core Becomes Subcritical 327.5Accumulators Begin to Inject (Unborated Water) 343.5Double-Ended Rupture SLB Occurs 0.0(1.388 ft2) without.
OffsitePower Available Low Steam Line Pressure Setpoint Reached in the Faulted 0.6Loop (lead-lagged)
SI Signal Generated (on low steam line pressure) 2.6RCPs Begin to Coast Down 3.0Steam Line and FW Isolation Complete 17.6Pressurizer Empties 18.0Core Re-criticality Occurs 26.0SI Flow Initiated 39.6Borated Water from SI System Reaches the Core 366.5Peak Core Heat Flux Reached 374.5Core Becomes Subcritical 384.3WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-88WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-88Table 2.2.5.1-2 Limiting Results -Steam System Piping Failure at HZP Conditions Case Parameter Analysis Value LimitDouble-Ended Rupture Minimum DNBR 1.80 1.18(1.388 ft2) with Offsite PowerAvailable Peak Linear Heat Generation (kW/ft) 15.829 22.4WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-89WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-89&#xa2;-E000(-)LL_0n00~&#xa2;-E00oU-&#xa2;C-)a 100 200 300Time (sec)4000 100 200 300Time (sec)400Figure 2.2.5.1-1 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)
Nuclear Power and Core Heat Flux versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-90Faulted LoopIntact LoopsQ)M.)Ea)c-Q)Q)0-0>a,0'C.)U_)0:O)0 100 200 300Time (sec)4000 100 200 300Time (sec)400Figure 2.2.5.1-2 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)
Reactor Vessel Inlet Temperature and Core Average Temperature versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-910na,)C/)0nEa,'P,CI-0 100 200 300Time (sec)4000 100 200 300 400Time (sec)Figure 2.2.5.1-3 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)
Pressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-920~0ciCDW-c)&#xfd;0 100 200 300Time (sec)4000 100 200 300Time (sec)400Figure 2.2.5.1-4 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)
Core Boron Concentration and Reactivity versus TimeWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-93Four ted LoopI ntact LoopsCI)0.)Cj)cn0 1O0 200 300Time (sec)400Foul ted LoopIntact Loops (Tot a )E0ECD0.)0 100 200 300Time (sec)400Figure 2.2.5.1-5 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)
Steam Pressure and Steam (Break) Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-94WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-94Fau I ted LoopIntact LoopsI DUUCO,E-001200800400..............
......................................................................................................................-7 -'r -i- -T ---00100200Time (sec)300400Faur ted LoopIntact Loopsnnnnn .A4U.-UCf,Ct)0EV.)150000-1000001------ --------- ----------------- ---...........................................................
.............................................................................................................50000"U0100200Time (sec)300400Figure 2.2.5.1-6 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)
FW Flow and SG Mass versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-951.41.2-..........
..............................
C30C_)1-0.8-0.6-0.4-0.2-7............
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V0100200Time (sec)300400Figure 2.2.5.1-7 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)
Core Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-96E00D0r)F-00~C-)0 100 200 300Time (sec)4000 1O0 200 300Time (sec)400Figure 2.2.5.1-8 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)
Nuclear Power and Core Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-97Faul ted LoopIntact Loops----- 6007" 550--4 500 -............
..E 450-cI---400 ... ....._....co 350 ..............
2 300-0OnajI,C-Q_0U.200Time (sec)Figure 2.2.5.1-9 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)
Reactor Vessel Inlet Temperature and Core Average Temperature versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-98CL0na,a.)EUf)0 100 200 300Time (sec)400Figure 2.2.5.1-10 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)
Pressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32 -990~00F-)0 100 200 300Time (sec)4000 100 200 300Time (sec)400Figure 2.2.5.1-11 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)
Core Boron Concentration and Reactivity versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-100WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-100Faul ted LoopIntact LoopsV)U)U-30 100 200 300Time (sec)400Faul ted LoopIntact Loops (Total)C-)E0En0 100 200 300Time (sec)400Figure 2.2.5.1-12 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)
Steam Pressure and Steam (Break) Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-101Fau ted LoopIntact Loops1 rflf II UUUEa-)1200-800"r- -12 i --, -T- -I-- r400-U0100200Time (sec)300400Faul ted LoopIntact Loops'ntAAAA.LUUUUUC/)CI)E0n150000------------
-r --0I100200Time (sec)300400Figure 2.2.5.1-13 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)
FW Flow and SG Mass versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-102WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-102C.-)U-"0U-0:C--0 100 200 300Time (sec)400Figure 2.2.5.1-14 Steam System Piping Failure at HZP(1.388 fte Break without Offsite Power Available)
Core Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1032.2.5.2 Steam System Piping Failure at Hot Full Power Conditions 2.2.5.2.1 Technical Evaluation 2.2.5.2.1.1 Introduction A rupture in the MSS piping from an at-power condition creates an increased steam load, which extractsan increased amount of heat from the RCS via the SGs. This results in decreased RCS temperature andpressure.
In the presence of a strong negative MTC, typical of EOC conditions, the colder core inletcoolant temperature causes the core power to increase from its initial level due to the positive reactivity insertion.
The power approaches a level equal to the total steam flow. Depending on the break size, a RTmay occur due to overpower conditions or as a result of a SLB protection function actuation.
The steam system piping failure accident analysis described in Section 2.2.5.1 is performed assuming aHZP initial condition with the control rods inserted in the core, except for the most reactive rod in thefully withdrawn position.
Such a condition could occur the following ways:* When the reactor is at hot shutdown at the minimum required shutdown margin* After the plant has been tripped automatically by the reactor protection system* Manually by the operator.
For an at-power SLB, the analysis of Section 2.2.5.1 represents the limiting condition with respect to coreprotection for the time period following RT. The purpose of this section is to describe the analysis of asteam system piping failure occurring from at-power initial conditions, which demonstrates that coreprotection is maintained prior to and immediately following RT.2.2.5.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe following assumptions are made in the analysis of a main steam line rupture accident at full power:The initial reactor power, pressurizer
: pressure, and RCS Tag are assumed to be at the nominal fullpower values. The full power condition is more limiting than part power. The reactor coolant flowrate is the MMF value. The initial loop flows were assumed to be symmetric.
Uncertainties forthe initial conditions of pressurizer
: pressure, RCS Tavg, and reactor coolant flow are statistically accounted for in the DNBR limit calculated using the RTDP methodology (Reference 4). InitialNSSS power was conservatively modeled to be at 3651 MWt, which includes all applicable uncertainties.
The full power RCS Tvg range is from 570.70F to 588.40F. Because the full power steam linerupture event is primarily a DNB event, assuming a maximum RCS average temperature islimiting.
Therefore, an initial RCS average temperature of 588.4&deg;F was assumed.The main FW analytical temperature range is from 400'F to 448.6'F.
A higher Tfeed is morelimiting for this event. Thus, a Tfeed of 448.6&deg;F was assumed.WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-104A spectrum of break sizes was analyzed.
Typically, small breaks do not result in a RT; in this casecore power stabilizes at an increased level corresponding to the increased steam flow.Intermediate size breaks may result in a RT on OPAT as a result of the increasing core power.Larger break sizes result in a RT soon after the break from the SI signal actuated by low steamline pressure, which includes lead/lag dynamic compensation.
The limiting break size is thelargest break that does not trip on a low steam pressure SI signal.* To maximize the primary-to-secondary heat transfer rate, 0 percent SGTP is assumed.Maximum moderator reactivity feedback and minimum Doppler power feedback are assumed tomaximize the power increase following the break.The protection system features that mitigate the effects of a SLB are described in Section 2.2.5.1.This analysis only considers the initial phase of the transient from at-power conditions.
Protection in this phase of the transient is provided by RT, if necessary.
Section 2.2.5.1 presents the analysisof the bounding transient following RT, where other protection system features are actuated tomitigate the effects of the SLB.In general, the results would be less severe as a result of normal control system operation.
Therefore, the mitigation effects of control systems have been ignored in the analysis.
However,the main FW control system is assumed to operate in that FW flow is assumed to equal the steamflow prior to RT.Depending on the size of the break, a rupture in a main steam line is classified as either a Condition III(infrequent fault) or Condition IV (limiting fault) event, as defined by the ANS's "Nuclear Safety Criteriafor the Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
: However, for ease ofinterpreting the results, the more restrictive criteria associated with Condition II events are applied.
Theapplicable acceptance criteria that may be challenged are that fuel damage due to DNB or fuel centerline melting should be precluded.
Fuel cladding integrity and the prevention of fuel failure is demonstrated byshowing that the calculated minimum DNBR is greater than the applicable limit value. The centerline temperature of the fuel rods with the peak linear heat rate (kW/ft) must not exceed the U02 meltingtemperature.
The pressure limits for the primary and secondary systems are not challenged for thisaccident.
The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the major rupture of a main steam pipe event acceptance criteria areprovided below.GDC 27 (Combined Reactivity Control Systems Capability) requires that the reactivity controlsystems be designed to have a combined capability, in conjunction with poison addition by theECCS, of reliably controlling reactivity changes so that, under postulated accident conditions andwith appropriate margin for stuck rods, the capability to cool the core is maintained.
For thesteam system piping failure at HFP conditions event, this is shown to be met by demonstrating that the fuel damage criterion is satisfied, which ultimately ensures that the ability to insertcontrol rods is maintained.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-105GDC 28 (Reactivity Limits) requires that the reactivity control system be designed withappropriate limits on the potential amount and rate of reactivity increase so that the effects ofpostulated reactivity accidents can neither result in damage to the RCPB greater than limited localyielding, nor sufficiently disturb the core, its support structures, or other RPV internals to impairsignificantly the capability to cool the core. For the steam system piping failure at HFP conditions event, this is shown to be met by demonstrating that the peak RCS pressure is less thanI 10 percent of the design pressure, which ultimately ensures that the RCPB pressure limits arenot exceeded.
2.2.5.2.1.3 Description of Analyses and Evaluations A detailed analysis was performed using the RETRAN computer code to determine the plant transient conditions following a main steam line rupture at full power. Details of the RETRAN model aredocumented in Reference
: 1. The code computes pertinent variables, including the core power and reactorcoolant temperature and pressure.
Statepoints from RETRAN, consisting of core heat flux, RCS loop inlettemperatures,
: pressure, and core flow, are used as input to the DNB analysis and the calculation of thepeak linear heat rate (kW/ft).
A detailed core analysis was performed using the ANC code (Reference 2)to confirm the validity of the RETRAN-predicted reactivity feedback model. The core models developed in ANC were also used to calculate the power peaking factors for input to the DNB analysis and thecalculation of the peak kW/ft. The detailed T/H digital computer code VIPRE (Reference
: 3) was used tocalculate the DNBR for the limiting time in the transient.
The DNBR calculations were performed usingthe WRB-2 DNB correlation and RTDP methodology (Reference 4).2.2.5.2.1.4 ResultsThe calculated sequence of events for the most limiting break size (1.04 fte) for a main steam line ruptureat full power event is shown in Table 2.2.5.2-1.
This is the largest break that does not trip on a low steamline pressure SI signal. The results for this case are presented in Table 2.2.5.2-2.
Figures 2.2.5.2-1 through 2.2.5.2-4 show the transient response for selected parameters.
The results of the analysis of a major rupture of a main steam pipe event at full power demonstrate thatthe DNB design basis is met. The calculated minimum DNBR is above the limit value. In addition, thepeak linear heat generation rate (expressed in kW/ft) does not exceed the value that would cause fuelcenterline melt. The pressure limits for the primary and secondary systems are not challenged for thisaccident because the pressures in these systems each decrease from their initial values during thetransient.
The steam pressure does not increase significantly following turbine trip. Therefore, this eventdoes not adversely affect the core or the RCS, and all applicable acceptance criteria are met.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1062.2.5.2.2 Conclusion The analysis of the steam system piping failure at full power conditions described above has beenreviewed.
It is concluded that the analysis has adequately accounted for operation of the plant at theanalyzed power level and was performed using acceptable analytical models. It is further concluded thatthe analysis has demonstrated that the reactor protection and safety systems will continue to ensure thatthe ability to insert control rods is maintained, the RCPB pressure limits will not be exceeded, andabundant core cooling will be provided.
Based on this, the conclusion is that the plant will continue tomeet the requirements of GDCs 27 and 28.Although a discussion of the steam system piping failure at full power analysis is not included in thecurrent USAR, Section 15.1.6 will be revised to reflect the analysis described herein.2.2.5.2.3 References
: 1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.2. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.4. WCAP-11397-P-A, "Revised Thermal Design Procedure,"
April 1989.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-107WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-107Table 2.2.5.2-1 Time Sequence of Events -Steam System Piping Failure at HFP Conditions Case Event Time (sec)Limiting Break Size (1.04 ft-) Steam Line Ruptures 0.0OPAT RT Setpoint Reached 17.7(in two loops)Rods Begin to Drop 20.7Peak Core Heat Flux Occurs 21.5Minimum DNBR Occurs 21.5Table 2.2.5.2-2 Limiting Results -Steam System Piping Failure at HFP Conditions Case Parameter Analysis Value LimitLimiting Break Size (1.04 ft2) Minimum DNBR 2.026 1.52Peak Linear Heat Generation (kW/fi) 21.8 22.4WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-108WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1081.4CD0C-)C-)1.21.00.80.60.4-0.2I I I I I I I I I I I I I &#xfd; I I I I I I I I I I &#xfd; I I &#xfd; I I I I I0.0051015Time20(sec)2530351.4Q(9P0(C90(9-EIF,(9l0(91.2-1.0-0.8-0.4-0.2-0.010U15 20Time (sec)3530Figure 2.2.5.2-1 Steam System Piping Failure at HFP (1.04 ft2 Break)Nuclear Power and Core Heat Flux versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1092300.C 2200-Caa1)f 2100-a:)2000-c)_ 1900-I I I I I I I I I I I I I I III I I I I I I I I I I I I I I I11001000-900-800-700-600-P0515 20Time (Sec)253035I0a)a)500-I II I ; I I I I I I I 105IU15 20Time (see)Figure 2.2.5.2-2 Steam System Piping Failure at HFP (1.04 fe Break)Pressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-1 7658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-110ai tedn c- co op)60C)C)C)C)(1<C-)I I I I I I I I I I I I I I(11G 15 2? 5 31,c, L/Figure 2.2.5.2-3 Steam System Piping Failure at HFP (1.04 ft2 Break)Reactor Vessel Inlet Temperature and Loop Average Temperature versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-111F a "I~ " c 10(n 7 C C c CP1000-~--~ ~crP ~(f~)(U.I I I I I &#xfd; &#xfd; I &#xfd; I I I0 515 2 0 7n C1&#xfd;1>0(I.Figure 2.2.5.2-4 Steam System Piping Failure at HFP (1.04 ft2 Break)Steam Pressure and Break Flow versus TimeWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1122.3 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM2.3.1 Loss of External Electrical Load, Turbine Trip, Inadvertent Closure of Main SteamIsolation Valves, Loss of Condenser Vacuum and Other Events Resulting in TurbineTrip (USAR Sections 15.2.2, 15.2.3, 15.2.4, and 15.2.5)2.3.1.1 Technical Evaluation 2.3.1.1.1 Introduction A major load loss on the plant can result from either a loss of external electrical load or from a turbinetrip. A loss of external electrical load can result from an abnormal variation in network frequency or otheradverse network operating conditions.
In either case, offsite power is available for the continued operation of plant components such as the RCPs.The plant is designed to accept a 50 percent loss of electrical load while operating at full power, or acomplete loss of load (LOL) while operating below the P-9 setpoint without actuating a RT with all NSSScontrol systems in automatic.
A 50 percent loss of electrical load is handled by the following:
Steam dump system, which accommodates 40 percent of the nominal full-power load,Rod control system, which accommodates the remaining 10 percent of the load rejection bydriving rods in to reduce coolant average temperature, Pressurizer, which absorbs the change in coolant volume due to the heat addition resulting fromthe load rejection.
Should a 100 percent LOL occur from full power, the reactor protection system automatically actuates aRT. Based on this, a complete LOL from 100 percent power represents the most severe challenge to thesystem and, as such, it is the case explicitly analyzed and described in this section.The most likely source of a complete LOL on the NSSS is a trip of the turbine generator.
In this case, ifthe reactor is operating above the P-9 setpoint, there is a direct RT signal from either the turbine low fluidoil pressure or the turbine stop valve closure.
Reactor temperature and pressure do not increasesignificantly if the steam dump system and pressurizer pressure control system are functioning properly.
: However, the RCS and MSS pressure-relieving capacities are designed to ensure the safety of the plantwithout requiring the use of automatic rod control, pressurizer pressure
: control, or steam dump controlsystems.
In this analysis, the behavior of the plant is evaluated for a 100 percent loss of steam load fromfull power without direct RT in order to demonstrate the adequacy of the pressure-relieving devices andcore protection margins.In the event the steam dump valves fail to open following a large LOL, the MSSVs can lift and the reactorcan be tripped by the high pressurizer pressure signal, the OTAT signal, or the OPAT signal. The SGshell-side pressure and reactor coolant temperatures increase rapidly.
The PSVs and MSSVs are sized toprotect the RCS and SGs against overpressurization for all load losses without assuming the operation ofWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-113the steam dump system, pressurizer sprays, pressurizer PORVs, automatic rod control, or the direct RT onturbine trip.2.3.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThree cases were analyzed for a loss of load / turbine trip (LOL/TT) event from full-power conditions.
* Maximum SGTP with automatic pressurizer pressure control (minimum DNBR case)* Minimum SGTP with automatic pressurizer pressure control (peak MSS pressure case)* Maximum SGTP without automatic pressurizer pressure control (peak RCS pressure case)The minimum DNBR case was analyzed using the RTDP (Reference 1). The initial NSSS power wasconservatively modeled to be at 3651 MWt, which includes all applicable uncertainties.
RCS temperature and pressurizer pressure were assumed to be at their nominal values consistent with steady-state, full-power operation.
MMF was modeled.
Uncertainties in initial conditions were included in the safetyanalysis DNBR limit, as described in Reference 1.The peak RCS and MSS pressure cases were analyzed with uncertainties on RCS temperature andpressurizer pressure applied in the direction required to obtain the most conservative initial plantconditions for the transient.
Both cases modeled TDF.The LOL/TT transient was conservatively analyzed with minimum reactivity feedback (beginning of corelife). All cases assumed the least-negative DPC and a 0 pcm/&deg;F MTC, which bounds part-power conditions with a positive MTC. Minimum reactivity feedback conditions are conservative becausereactor power is maintained until the time of RT, which exacerbates the calculated minimum DNBR andpeak RCS and MSS pressures.
Manual rod control was modeled for all cases. If the reactor had been in automatic rod control, the controlrod banks would have been driven into the core prior to RT, thereby reducing the severity of the transient.
The LOL/TT event was analyzed both with and without automatic pressurizer pressure control.
Thepressurizer PORVs and sprays were assumed to be operable for the minimum DNBR case to minimize theincrease in RCS pressure, which is conservative for the calculation of the minimum DNBR. Thepressurizer PORVs and sprays were also assumed to be operable for the peak MSS pressure case tominimize the increase in RCS pressure.
This delays or completely prevents a RT from occurring on a highpressurizer pressure signal, which results in a conservative calculation of the peak MSS pressure.
Thepeak RCS pressure case was analyzed without automatic pressurizer pressure control to conservatively maximize the RCS pressure increase.
In all cases, the MSSVs and PSVs were assumed to be operable.
A total PSV setpoint tolerance of +2 percent was accounted for in the analysis.
For the minimum DNBRcase and the peak MSS pressure case, the negative tolerance was applied to conservatively reduce thesetpoint.
For the peak RCS pressure case, the positive tolerance was applied to conservatively increase thesetpoint.
In addition, the peak RCS pressure case includes a 0.9 percent setpoint shift and a 1.153-second purge time delay to account for the existence of PSV water-filled loop seals, as described in Reference 2.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-114Main FW flow to the SGs was assumed to be lost at the time of turbine trip. The AFW system would beavailable for long-term heat removal.
: However, operation of the AFW system is not credited in thetimeframe considered for this analysis.
The following RT functions are assumed to be operable:
* High pressurizer pressure* OTAT* OPATThe MSSVs were modeled with opening setpoints that account for a maximum setpoint tolerance of3 percent and all appropriate line losses. Valve accumulation was modeled via a 5-psi ramp of the valveflow area from closed to full-open.
The limiting single failure is the failure of one train of the reactor protection system. The remaining (operable) train trips the reactor.
As described in USAR Section 3.1.1, the MSSVs and PSVs (that is, codesafety valves) are considered to be qualified components exempt from active failure and are assumed toopen on demand. Control systems are assumed to function only if their operation results in more severetransient conditions.
Thus, a failure of a control system is not applicable as a limiting single failure.
FWisolation (redundant valves),
AFW (multiple pumps) and SI (multiple pumps) are susceptible to a singlefailure.
: However, none of these systems provides any mitigation for a LOL/TT event. Thus, these systemsare not applicable as a limiting single failure.
Furthermore, the protection system is designed to besingle-failure-proof.
Maximum SGTP (10 percent) is assumed in the minimum DNBR case and peak RCS pressure casebecause it maximizes the RCS temperature increase following event initiation.
: However, the peak MSSpressure case is analyzed with zero SGTP because this conservatively maximizes theprimary-to-secondary side heat transfer; this assumption is slightly more limiting with respect to thesecondary-side pressure transient.
Based on its frequency of occurrence, the LOL/TT accident is considered a Condition II event, an incidentof moderate frequency, as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
The specific criteria for this accident are asfollows:Pressures in the RCS and MSS are maintained below 110 percent of their respective design values(for the WCGS, this represents an RCS pressure limit of 2750 psia and MSS pressure limit of1318.5 psia).Fuel cladding integrity is maintained by demonstrating that the minimum DNBR remains abovethe 95/95 DNBR limit for PWRs (for the WCGS, the applicable safety analysis DNBR limitis 1.52).WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-115An incident of moderate frequency does not generate a more serious plant condition without otherfaults occurring independently.
This criterion is conservatively satisfied by verifying that thepressurizer does not fill.An incident of moderate frequency, in combination with any single active component failure orsingle operator error, is considered an event for which an estimate of the number of potential fuelfailures is provided for radiological dose calculations.
For such accidents, fuel failure is assumedfor all rods for which the DNBR decreases below those values cited above for cladding integrity unless it can be shown that, based on an acceptable fuel damage model, fewer failures occur.There shall be no loss of function of any fission product barrier other than the fuel cladding.
Thiscriterion is satisfied by verifying that the minimum DNBR remains above the 95/95 DNBR limit,which is discussed above.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the LOL/TT acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the LOL/TT event, this is shown to be met by demonstrating that the fuelcladding integrity is maintained.
GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the reactor coolant pressureboundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the LOL/TT event, this is shown to be met by demonstrating that thepeak RCS pressure is less than 110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires that one of thereactivity control systems consist of control rods capable of reliably controlling reactivity changeswith appropriate margin for malfunctions like stuck rods so that specified acceptable fuel designlimits are not exceeded under conditions of normal operation, including anticipated operational occurrences.
For the LOL/TT event, which results in a RT, this is shown to be met bydemonstrating that the fuel cladding integrity is maintained.
2.3.1.1.3 Description of Analyses and Evaluations A detailed analysis using the RETRAN computer code (Reference
: 3) was performed to determine theplant transient conditions following a total LOL due to turbine trip without credit for a direct RT. TheRETRAN model simulates the core neutron kinetics, RCS, pressurizer, pressurizer PORVs and sprays,PSVs, SGs, MSSVs, and the AFW system. The code computes pertinent plant variables, including RCSpressures and temperatures, and SG pressure.
The Westinghouse RETRAN model has been approved by the NRC for the analysis of the LOL/TTtransient (Reference 3).WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1162.3.1.1.4 ResultsThe calculated sequence of events for each of the cases is listed in Table 2.3.1-1, and the limiting resultsfor each of the cases are presented in Table 2.3.1-2.2.3.1.1.4.1 Minimum DNBR CaseThe minimum DNBR case was analyzed at the high nominal Tavg (i.e., 588.4&deg;F),
nominal pressurizer pressure (i.e., 2250 psia), MMF, 10 percent SGTP, and the high main Tfeed (448.6&deg;F) with automatic pressurizer pressure control operable.
Plots of the transient response to a LOL/TT event for the minimum DNBR case are shown inFigures 2.3.1-1 through 2.3.1-3.
The reactor was tripped on the OTAT RT function.
The nuclear powerremained essentially constant at full power prior to the RT. The pressurizer sprays, PORVs, and safetyvalves actuated to minimize the RCS pressure transient, which is conservative for the calculation of theminimum DNBR. Although the DNBR decreased below its initial value, it remained well above the SALthroughout the entire transient.
The peak pressurizer water volume remained below the total volume ofthe pressurizer, demonstrating that this event does not generate a more serious plant condition.
TheMSSVs actuated to maintain the MSS pressure below 110 percent of the design value.2.3.1.1.4.2 Peak MSS Pressure CaseThe peak MSS pressure case was analyzed at the high nominal Tavg plus uncertainties (i.e., 588.4'F +6.5&deg;F), nominal pressurizer pressure minus uncertainties (i.e., 2250 psia -50 psi), TDF, 0 percent SGTPand the high main Tfeed (448.60F) with automatic pressurizer pressure control operable.
Plots of the transient response to a LOL/TT event for the peak MSS pressure case are shown inFigures 2.3.1-4 through 2.3.1-6.
The reactor was tripped on the OTAT RT function.
The nuclear powerremained essentially constant at full power prior to the RT. The pressurizer sprays, PORVs, and safetyvalves actuated to minimize the RCS pressure transient, which is conservative because it prevented a RTfrom occurring on high pressurizer pressure and exacerbated the peak MSS pressure.
The MSSVsactuated to maintain the MSS pressure below 110 percent of the design value. The peak pressurizer watervolume remained below the total volume of the pressurizer, demonstrating that this event does notgenerate a more serious plant condition.
2.3.1.1.4.3 Peak RCS Pressure CaseThe most limiting peak RCS pressure case was that analyzed at the high nominal Tavg minus uncertainties (i.e., 588.4&deg;F -6.5&deg;F), nominal pressurizer pressure minus uncertainties (i.e., 2250 psi -35 psi), TDF,10 percent SGTP and the high main Tfeed (448.6&deg;F) with automatic pressurizer pressure controlinoperable.
Plots of the transient response to a LOL/TT event for the limiting peak RCS pressure case are shown inFigures 2.3.1-7 through 2.3.1-9.
The reactor was tripped on the high pressurizer pressure RT function.
Thenuclear power remained essentially constant at full power prior to the RT. The PSVs actuated to maintainthe RCS pressure below 110 percent of the design value. The MSSVs also actuated to maintain the MSSWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-117pressure below 110 percent of the design value. The peak pressurizer water volume remained below thetotal volume of the pressurizer, demonstrating that this event does not generate a more serious plantcondition.
2.3.1.2 Conclusions From a review of the updated analyses for the LOL/TT event, it is concluded that these analyses haveadequately accounted for operation of the plant at the analyzed power level and that they were performed using acceptable analytical models. The calculated results demonstrate that the reactor protection andsafety systems will continue to ensure that the safety analysis DNBR limit is met and the RCS and MSSpressure boundary limits will not be exceeded as a result of the LOL/TT event. Furthermore, this eventwill not generate a more serious plant condition.
Based on this, the WCGS will continue to meet therequirements of GDCs 10, 15, and 26.2.3.1.3 References
: 1. WCAP- I 1397-P-A, "Revised Thermal Design Procedure,"
April 1989.2. WCAP-129 10, Rev. I-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.3. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-118Table 2.3.1-1 Time Sequence of Events -Loss of External Electrical Load and/or Turbine TripCase Event Time (seconds)
Minimum DNBR Case Loss of Electrical Load/Turbine Trip Occurs 0.0Pressurizer PORVs Open 2.8PSVs Open 9.7MSSVs Open 9.7OTAT RT Setpoint Reached 17.0Minimum DNBR Occurs 19.8Rods Begin to Drop 20.0Peak MSS Pressure Case Loss of Electrical Load/Turbine Trip Occurs 0.0Pressurizer PORVs Open 1.7MSSVs Open 4.7PSVs Open 9.6OTAT RT Setpoint Reached 15.8Rods Begin to Drop 18.8Peak Secondary Side Pressure Occurs 22.0Peak RCS Pressure Case Loss of Electrical Load/Turbine Trip Occurs 0.0High Pressurizer Pressure RT Setpoint Reached 6.6Rods Begin to Drop 7.6PSVs Open 8.2Peak RCS Pressure Occurs 9.7MSSVs Open 12.1Table 2.3.1-2 Limiting Results -Loss of External Electrical Load and/or Turbine TripCase Parameter Analysis Value LimitMinimum DNBR Case Minimum DNBR 1.72 1.52Peak MSS Pressure Case Peak MSS Pressure (psia) 1297.9 1318.5Peak RCS Pressure Case Peak RCS Pressure (psia) 2746.8 2750.0WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-119I.0C0(.)L..a,300a,(-)=3CnCncnI.-0 20 40 60 80Time (sec)100Figure 2.3.1-1LOLUTT, Minimum DNBR CaseNuclear Power and SG Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-120WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-120CnIU,0-r4I-Cn(Ur4)0ncn0 20 40 60 80Time (sec)1000 20 40 60 80 100Time (sec)Figure 2.3.1-2LOL/TT, Minimum DNBR CasePressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3Hot Leg Temperature Vessel Average Temperature Cold Leg Temperature 700U-LJ....650 .. ..............
.. .. .. .. ... .. ..... .... .. ... ....-6 0 0 ............... ...................... .E550.......................
C/,2-1210 20 40 60 80Time (sec)1000 20 40 61Time (seconds80100Figure 2.3.1-3LOL/TT, Minimum DNBR CaseRCS Temperatures and DNBR versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-122WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-12200!<D)CUcn(UL..0 20 40 60 80 100Time (sec)a 20 40 60 80Time (sec)100Figure 2.3.1-4LOL/TT, Peak MSS Pressure CaseNuclear Power and SG Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-123WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-123~211.0-0 20 40 60 80 100Time (sec)0 20 40 60 80Time (sec)100Figure 2.3.1-5LOL/TT, Peak MSS Pressure CasePressurizer Pressure and Pressurizer Water Volume versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-124WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-124Hot Leg Temperature Vessel Average Temperature Cold Leg Temperature LL_Cfl=3CLE0 20 40 60 80Time (sec)100Figure 2.3.1-6LOL/TT, Peak MSS Pressure CaseRCS Temperatures versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-125WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-125004--0L.0~3ra_0~C,,C,,E0 20 40 60 80Time (seconds) 100Figure 2.3.1-7LOL/TT, Peak RCS Pressure CaseNuclear Power and SG Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-126WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-126Pressurizer PressureMaximum RCS PressureC,,C,,C,,V)C-,)=3E=3C,,Cfl0 20 40 60 80Time (sec)LOL/TT, Peak RCS Pressure CaseRCS Pressures and Pressurizer Water Volume versus Time100Figure 2.3.1-8WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-127WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-127Hot Leg Temperature Vessel Average Temperature Cold Leg Temperature rn. r600E 550-------- ------ -- --- ---d0-II II Ii I II II I I-.1/u020406080100Time (sec)Figure 2.3.1-9LOLITT, Peak RCS Pressure CaseRCS Temperatures versus TimeWCAP-17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1282.3.2 Loss of Non-Emergency AC Power to the Station Auxiliaries (USAR Section 15.2.6)2.3.2.1 Technical Evaluation 2.3.2.1.1 Introduction A complete loss of non-emergency alternating current (AC) power (LOAC) may result in a loss of powerto the plant auxiliaries, which include the RCPs, main FW pumps, condensate pumps, etc. The loss ofpower may be caused by a complete loss of the offsite grid accompanied by a turbine generator trip at thestation, or by a loss of the onsite AC distribution system.The LOAC event is analyzed as a LONF with a loss of power to the RCPs as a result of the RT becausethis is a more severe event relative to long-term consequences than the LOAC event. In the LOAC event,the RCPs lose power at the beginning of the event and the reactor trips soon thereafter on low reactorcoolant loop flow. The short-term consequences are bounded by those of the complete loss of reactorcoolant flow event described in Section 2.4.1, "Partial and Complete Loss of Forced Reactor CoolantFlow." The immediate consequence following a loss of FW is a reduction in the SG water level, which, ifleft unmitigated, will ultimately result in a RT and AFW system actuation on the low-low SG water levelsignal. Following RT, the rate of heat generation in the RCS (core residual (decay) heat) may exceed theheat removal capability of the secondary system. If this occurs, the RCS heats up, and the resulting thermal expansion of the reactor coolant causes an insurge to the pressurizer and an increase in thepressurizer water level. This trend generally continues until the RCS heat generation rate decreases belowthe secondary-side heat removal capability, at which time a cooldown of the RCS commences.
The LONFevent without a LOOP is addressed in Section 2.3.3, "Loss of Normal Feedwater Flow."The expected events following an LOAC with turbine and reactor trips are described in the sequencelisted as follows.Plant vital instruments are supplied by emergency direct current (DC) power sources.The SG ARVs are automatically opened to the atmosphere as the MSS pressure increases following the trip. The condenser is assumed to be unavailable for steam dump. If the steam flowrate through the ARVs is not sufficient or if the ARVs are not available, the MSSVs may lift todissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in thereactor.The SG ARVs (or safety valves, if the ARVs are not available) are used to dissipate the residualdecay heat and to maintain the plant at the Mode 3 (hot standby) condition as the no-loadtemperature is approached.
The diesel generators start on a loss of voltage to the plant engineered safety features busses andbegin to supply plant vital loads.The AFW system is automatically actuated.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-129The plant safety features that are available to mitigate the consequences of an LOAC event are as follows.* A RT can be initiated by one of the following.
-Two-out-of-four low-low water level signals in any one SG-Two-out-of-four high pressurizer pressure signals-Two-out-of-three high pressurizer level signals-Two-out-of-four OTAT signalsThe MSSVs open and provide secondary-side pressure protection and a heat sink source thathelps limit the RCS heatup.* The PSVs may open to provide primary-side pressure protection.
Backup FW for the SGs is provided by the AFW system, which is composed of two motor-driven AFW (MDAFW) pumps and one turbine-driven AFW (TDAFW) pump.-The two MDAFW pumps are started on any of the following:
* Two-out-of-four low-low water level signals in any one SG* Trip of both main FW pumps* SI signal* LOOP* Manual pump start" Manual AFW system actuation
-The TDAFW pump is started on any of the following:
" Two-out-of-four low-low water level signals in each of two SGs* LOOP* Manual pump start* Manual AFW system actuation The MDAFW pumps are supplied power by the diesel generators, and the TDAFW pump utilizessteam from the secondary system. The pump turbine exhausts the secondary steam to theatmosphere.
: Normally, the AFW pumps take suction from the condensate storage tank (CST), butif the CST is unavailable, the essential service water system is used as the water source for theAFW pumps.After power to the RCPs is lost, coolant flow necessary for core cooling and the removal of core decayheat is maintained by natural circulation in the RCS loops. Following the RCP coastdown, the naturalcirculation capability of the RCS will remove decay heat from the core, aided by the AFW flow in thesecondary system. Demonstrating acceptable analysis results for this event proves that the resultant natural circulation flow in the RCS and the AFW flow are sufficient for removing the decay heat from thecore.WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1302.3.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The major inputs and assumptions applied in the analysis of the LOAC event are identical to thoseapplied in the analysis of the LONF event described in Section 2.3.3, "Loss of Normal Feedwater Flow,"with the following exceptions:
The initial RCP heat is the nominal value of 14 MWt. Nominal RCP heat is conservative for theLOAC event because the initial core power is slightly higher compared to that associated withmaximum RCP heat, and this translates into slightly higher core decay heat, which is the primaryheat source of concern for this event; after coastdown, the RCPs cease to add heat to the primarycoolant, and so it is conservative to maximize the core decay heat.The loss of power to the RCPs is assumed to be the result of an electrical disturbance on theoffsite power grid caused by the RT. The RCPs were assumed to lose power and begin coastingdown 2 seconds after the start of rod motion. This time delay is considered to be reasonable, but itis not a critical parameter in the analysis because it is short relative to the overall transient time.Acceptance CriteriaBased on the expected frequency of occurrence, the LOAC event is considered to be a Condition II eventas defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water ReactorPlants,"
ANSI N 18.2-1973.
The following items summarize the acceptance criteria associated with theanalysis of this event:Pressures in the RCS and MSS must remain less than 110 percent of the respective designpressures.
With respect to peak RCS and MSS pressures, the LOAC event is bounded by the LOL/TT eventdescribed in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip, Inadvertent Closureof a Main Steam Isolation Valve, and Loss of Condenser Vacuum,"
in which assumptions aremade to conservatively calculate the RCS and MSS pressure transients.
For the LOAC event,turbine trip occurs after RT, whereas for LOL/TT, the turbine trip is the initiating incident.
Therefore, the power mismatch between the primary and secondary sides and the resultant temperature and pressure transients of the RCS and MSS are always more severe for LOL/TTthan for LOAC. Based on this, no explicit calculation of maximum RCS or MSS pressure isperformed for this event.Fuel cladding integrity must be maintained by ensuring that the minimum DNBR remains abovethe 95/95 DNBR limit.With respect to the DNBR, the LOAC event is bounded by the complete loss of reactor coolantflow event described in Section 2.4.1, "Partial and Complete Loss of Forced Reactor CoolantFlow." Whereas the LOAC event has RCP coastdown (reactor coolant flow reduction) occurring after rod motion, the complete loss of reactor coolant flow event begins with a coastdown of allWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-131the RCPs, and RT occurs after the core coolant flow has already degraded.
As the limiting ratio ofthe core power to core flow is greater for the complete loss of reactor coolant flow event, it ismore limiting with respect to the DNBR. Based on this, no explicit calculation of minimumDNBR is performed for this event.An incident of moderate frequency must not generate a more serious plant condition withoutother faults occurring independently.
This criterion is conservatively demonstrated to be met if the pressurizer does not becomewater-solid.
The concern with filling the pressurizer water-solid is that it could lead to the failingopen of one or more PSVs, which would provide an unisolable path for the loss of reactorcoolant, and a LOCA is a more serious plant condition.
Satisfying this criterion demonstrates thepreclusion of a more serious plant condition, ensures that the RCS and MSS pressure criteria andminimum DNBR criterion are satisfied for the long-term portion of the event, and confirms theAFW system is adequate for long-term heat removal.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the LOAC acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the LOAC event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.
GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.
For theLOAC event, this is shown to be met by demonstrating that the peak RCS pressure is less than110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.
For the LOAC event, whichresults in a RT, this is shown to be met by demonstrating that the fuel cladding integrity ismaintained with a trip reactivity that accounts for the most reactive rod stuck out of the core.2.3.2.1.3 Description of Analyses and Evaluations A detailed analysis using the RETRAN computer code (Reference
: 1) was performed to determine theplant transient conditions for the LOAC event. A RETRAN input model specific to the WCGS wasdeveloped to simulate the core neutron kinetics, RCS, pressurizer, pressurizer
: heaters, pressurizer sprays,SGs, MSSVs, and the AFW system. Several LOAC cases were modeled for various combinations ofinitial conditions and pressurizer PORV availability, and the RETRAN code computed the time-dependent WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-132trends of pertinent variables, including the pressurizer
: pressure, pressurizer water volume, SG mass, andreactor coolant temperatures.
2.3.2.1.4 ResultsThe most limiting LOAC case was with an initial Ta&#xfd;,g of 564.2&deg;F (low end of the full-power Tavgwindow (570.7&deg;F) minus uncertainties),
an initial pressurizer pressure of 2300 psia (nominal (2250 psia)plus uncertainties),
an initial main Tfeed of 400'F (low full-power value), maximum (10 percent)
SGTP,and the pressurizer PORVs not available.
The calculated sequence of events for the limiting LOAC case is presented in Table 2.3.2-1, and transient plots of the significant plant parameters are provided in Figures 2.3.2-1 through 2.3.2-10.
Following theloss of FW from full power, the SG water level decreases to the low-low setpoint at 37.7 seconds, whichactuates a RT and the AFW system. The lack of FW causes the RCS temperature to increase.
Rod motionand turbine trip are initiated at 39.7 seconds and the RCPs begin coasting down at 41.7 seconds.
Althougha temporary cooldown of the RCS occurs as a result of the RT, the RCS heats up rapidly in response to thecontinued lack of FW and also the turbine trip. The MSSVs open at 73.2 seconds to help dissipate thestored and generated heat, and at 97.7 seconds, one minute after being actuated, the AFW system beginsto deliver 220 gpm of AFW flow to each SG. The pressurizer water volume reaches a maximum value of1623.2 ft3 at 2953.5 seconds after event initiation.
As the maximum pressurizer water volume value is lessthan the total pressurizer volume of 1800 ft3, it is confirmed that the pressurizer does not reach awater-solid condition.
2.3.2.2 Conclusions Based on the above information, it is concluded that the LOAC event will not progress into a moreserious plant condition.
Thus, all applicable event acceptance criteria are satisfied, and the AFW systemwith natural circulation reactor coolant flow are confirmed to be adequate for long-term heat removalfollowing an LOAC event. Therefore, it has been demonstrated that the reactor protection and safetysystems ensure that the acceptable fuel design limits are met, and the RCS and MSS pressure limits willnot be exceeded as a result of an LOAC event. Based on this, the plant continues to meet the requirements of GDCs 10, 15, and 26.2.3.2.3 References
: 1. WCAP- 14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-133Table 2.3.2-1 Time Sequence of Events for Limiting LOAC CaseEvent Time (seconds)
Main FW Flow Stops 0.0Low-Low SG Water Level RT Setpoint Reached 37.7Rods Begin to Drop and Turbine Trip Initiated 39.7RCPs Begin Coasting Down 41.7On Each Loop, the MSSV with the Lowest Setting Opens 73.2Flow from Two MDAFW Pumps Initiated 97.7SG Inventory Reduction Reverses 221.5Maximum Pressurizer Water Volume Occurs 2953.5WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-134WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1341 )~1*ci)00~C)(-3ID0.80.60.40.2-I I I I , -I I I I I I I I11 I I ( I 1 [II I- -I---- ~ I -i I I I010110210ime (seconds) 310410Figure 2.3.2-1 LOAC -Nuclear Power versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-135WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1351.2-1QC)CTC)C)0.6-0.4--- 0.2 -0I I I I I I IT~~TI~--
I I2010110210ime (seconds) 310410Figure 2.3.2-2 LOAC -Core Average Heat Flux versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-136WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1361 -LbI _ I I i I I III I I I I IFigure 2.3.2-3 LOAC -Reactor Coolant Loop Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-137WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-137620~17D600tCn58-S560-_C)0540-520/01010z10mime (seconds) 310410Figure 2.3.2-4 LOAC -HL and CL Temperatures versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-138WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1382600-IC)u~)C)enenC)CLC)inU;enC)CL-C)C)2500-2400-2J00-2200-2100-kI I III I I~. I I Ill2000I I I I I 1 i I I I I t I I t I I II i I t201010210Time (seconds) 310410Figure 2.3.2-5 LOAC -Actual Pressurizer Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-139WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1391~f~fl *1I UUU01600-1400-1200-1000-I I f I I I I I I I I I I I I I .I I I I I800I I I I I I I 1 t t I t IT i I I2010110210Time (seconds) 310410Figure 2.3.2-6 LOAC -Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-140WESTiNGHOUSE NON-PROPRIETARY CLASS 3 2-1401300C)cn&#xa3;012001100-1000900-800I I700'2010110210Time (seconds)
'310410Figure 2.3.2-7 LOAC -SG Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-141WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-14 1C/)0C)DEC)7D.0T10080-60-40-20-~1I I I I I I I IU010I10210Time (seconds) 310410Figure 2.3.2-8 LOAC -Indicated SG Level versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-142WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1422000000C:)150000-100000-50000 -0I i I I I I III I I I I I I II I I I I I I) I I I I I .I I I I I I I I I I I I I I I I201010210Firme (seconds) 10410Figure 2.3.2-9 LOAC -SG Mass versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-14325(^-IJc7- 2000(IDC)-C)C)C)x0.00-O150-100-50-0--50I i I I I I i I 1 i f I , I I I I I I t i I I I2010110210Time (seconds) 310410Figure 2.3.2-10 LOAC -Loop AFW Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1442.3.3 Loss of Normal Feedwater Flow (USAR Section 15.2.7)2.3.3.1 Technical Evaluation 2.3.3.1.1 Introduction A LONF flow (from pump failures, valve malfunctions, or a complete LOAC power) results in areduction in the capability of the secondary system to remove the heat generated in the reactor core. If analternative supply of FW is not provided, core residual (decay) heat following RT would heat the primarysystem water to the point where water relief from the pressurizer could occur, resulting in a substantial loss of water from the RCS.The expected events following an LONF (caused by either pump failures or valve malfunctions) withturbine and reactor trips are described in the sequence listed as follows:The SG ARVs are automatically opened to the atmosphere as the MSS pressure increases following the trip. The condenser is assumed to be unavailable for steam dump. If the steam flowrate through the ARVs is not sufficient or if the ARVs are not available, the MSSVs may lift todissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in thereactor.The SG ARVs (or safety valves, if the ARVs are not available) are used to dissipate the residualdecay heat and to maintain the plant at the Mode 3 (hot standby) condition as the no-loadtemperature is approached.
* The AFW system is actuated automatically.
The plant safety features that are available to mitigate the consequences of an LONF event are as follows.* A RT can be initiated by one of the following.
-Two-out-of-four low-low water level signals in any one SG-Two-out-of-four high pressurizer pressure signals-Two-out-of-three high pressurizer level signals-Two-out-of-four OTAT signalsThe MSSVs open and provide secondary-side pressure protection and a heat sink source thathelps limit the RCS heatup.* The PSVs may open to provide primary-side pressure protection.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-145Backup FW for the SGs is provided by the AFW system, which is composed of two MDAFWpumps and one TDAFW pump.The two MDAFW pumps are started on any of the following:
-Two-out-of-four low-low water level signals in any one SG-Trip of both main FW pumps-SI signal-LOOP-Manual pump start-Manual AFW system actuation The TDAFW pump is started on any of the following:
-Two-out-of-four low-low water level signals in each of two SGs-LOOP-Manual pump start-Manual AFW system actuation The MDAFW pumps are supplied power by offsite power sources, and the TDAFW pumputilizes steam from the secondary system. The pump turbine exhausts the secondary steam to theatmosphere.
: Normally, the AFW pumps take suction from the CST, but if the CST is unavailable, the essential service water system is used as the water source for the AFW pumps.The analysis of the LONF event demonstrates that the AFW system is capable of removing the stored andresidual heat, and consequently ensures the core will remain covered with water, and the RCS and MSSwill not overpressurize.
With this, the plant is shown to be able to return to a safe condition following aLONF event.2.3.3.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The following inputs and assumptions were applied in the analysis of the LONF event:An initial NSSS power of 3651 MWt, which includes all applicable uncertainties The initial RCP heat is the maximum value of 20 MWt. Maximum RCP heat is conservative forthe LONF event because the RCPs operate continuously throughout the transient.
The constantheat generated by the RCPs, in combination with the core decay heat, are the primary-side heatsources that provide the challenge to the long-term cooling (LTC) capability of the plant.Two initial full-power main Tfeed:-400.0&deg;F (low)-448.60F (high)WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-146Four initial full-power Tavg values that cover the full range of the full-power Tayg window(570.7&deg;F to 588.4&deg;F) including uncertainty
(+/-6.5&deg;F):
-594.90F-581.90F-577.20F-564.20F(high Tavg plus uncertainty)
(high Tvg minus uncertainty)
(low Tavg plus uncertainty)
(low Ta,,g minus uncertainty)
Two initial pressurizer pressure values that cover the +/-50 psi uncertainty associated with thenominal operating value of 2250 psia:-2300 psia-2200 psia(nominal plus uncertainty)
(nominal minus uncertainty)
Two initial pressurizer water level values, which are dependent on the full-power Ta,,g value, thatcover the +7 percent span uncertainty associated with the nominal values of 59 percent span forhigh Tavg cases and 41 percent span for low Tavg cases:66 percent span48 percent span(high nominal plus uncertainty (high Tavg cases))(low nominal plus uncertainty (low Tavg cases))SGTP levels of 0 and 10 percentA minimum low-low SG water level setpoint of 0 percent NRS for RT and AFW system actuation A maximum delay for RT (rod motion) of 2 secondsA maximum delay for AFW flow initiation of 60 secondsA minimum total AFW flow of 880 gpm split evenly between the four loopsThis flow corresponds to having both MDAFW pumps available for event mitigation.
As it is theworst single active failure for this analysis, the TDAFW pump was assumed to fail.A maximum AFW enthalpy of 96 Btu/lbm, which corresponds to a temperature of 125&deg;F.The pressurizer proportional and backup heaters were modeled to maximize the heatup andthermal expansion of the water within the pressurizer.
In addition, the pressurizer sprays wereassumed to be operable, and cases were analyzed with and without the pressurizer PORVsavailable.
Secondary system steam relief is achieved through the self-actuated MSSVs. Note that steamrelief would normally be provided by the SG ARVs or condenser dump valves, but these wereconservatively assumed to be unavailable.
WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-147The MSSVs were modeled with opening setpoints that account for a maximum setpoint tolerance of 3 percent and appropriate line losses. Valve accumulation was modeled via a 5 psi ramp of thevalve open area from closed to full-open.
The reactivity feedback parameters were chosen to maximize the heatup of the RCS. Thisincluded modeling a least-negative MTC, a least-negative DTC, and a most-negative Doppler-only power coefficient.
Note that the applied MTC value, 0 pcm/0F, is the least-negative limit value for full power conditions; the application of a zero MTC at full power conditions isbounding compared to the application of a positive MTC at part power conditions.
Core residual/decay heat generation was based on the 1979 version of ANS 5.1 (Reference 1).ANSI/ANS-5.1-1979 is a conservative representation of the decay energy release rates.Long-term operation at the initial power level preceding the trip was assumed.Acceptance CriteriaBased on the expected frequency of occurrence, the LONF event is considered to be a Condition II eventas defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water ReactorPlants,"
ANSI N 18.2-1973.
The following items summarize the acceptance criteria associated with theanalysis of this event:Pressures in the RCS and MSS must remain less than 110 percent of the respective designpressures.
With respect to peak RCS and MSS pressures, the LONF event is bounded by the LOL/TT eventdescribed in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip, Inadvertent Closureof a Main Steam Isolation Valve, and Loss of Condenser Vacuum,"
in which assumptions aremade to conservatively calculate the RCS and MSS pressure transients.
For the LONF event,turbine trip occurs after RT, whereas for LOL/TT, is the initiating incident.
Therefore, the powermismatch between the primary and secondary sides and the resultant temperature and pressuretransients of the RCS and MSS are always more severe for LOL/TT than for LONF. Based onthis, no explicit calculation of maximum RCS or MSS pressure is performed for this event.Fuel cladding integrity must be maintained by ensuring that the minimum DNBR remains abovethe 95/95 DNBR limit.With respect to the DNBR, the LONF event is bounded by the LOL/TT event described inSection 2.3.1, "Loss of External Electrical Load, Turbine Trip, Inadvertent Closure of a MainSteam Isolation Valve, and Loss of Condenser Vacuum."
Each of these two events represents areduction in the heat removal capability of the secondary system. For the LONF event, the RCStemperature increases gradually as the SGs boil down to the low-low water level trip setpoint, atwhich time RT occurs, followed by turbine trip. For the LOL/TT event, the turbine trip is theinitiating event, and the loss of heat sink is much more severe. As such, the initial RCS heatupwill be much more severe for the LOL/TT event than for the LONF event, and the LOL/TT eventwill always be more severe with respect to the minimum DNBR criterion.
Based on this, noexplicit calculation of minimum DNBR is performed for this event.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-148An incident of moderate frequency must not generate a more serious plant condition withoutother faults occurring independently.
This criterion is conservatively demonstrated to be met if the pressurizer does not becomewater-solid.
The concern with filling the pressurizer water-solid is that it could lead to the failingopen of one or more PSVs, which would provide an unisolable path for the loss of reactorcoolant, and a loss of coolant accident is a more serious plant condition.
Satisfying this criterion demonstrates the preclusion of a more serious plant condition, ensures that the RCS and MSSpressure criteria and minimum DNBR criterion are satisfied for the long-term portion of theevent, and confirms the AFW system is adequate for long-term heat removal.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the LONF acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the LONF event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.
GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.
For theLONF event, this is shown to be met by demonstrating that the peak RCS pressure is less than110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.
For the LONF event, whichresults in a RT, this is shown to be met by demonstrating that the fuel cladding integrity ismaintained with a trip reactivity that accounts for the most reactive rod stuck out of the core.2.3.3.1.3 Description of Analyses and Evaluations A detailed analysis using the RETRAN computer code (Reference
: 2) was performed to determine theplant transient conditions for the LONF event. A RETRAN input model specific to the WCGS wasdeveloped to simulate the core neutron kinetics, RCS, pressurizer, pressurizer
: heaters, pressurizer sprays,SGs, MSSVs, and the AFW system. Several LONF cases were modeled for various combinations ofinitial conditions and pressurizer PORV availability, and the RETRAN code computed the time-dependent trends of pertinent variables, including the pressurizer
: pressure, pressurizer water volume, SG mass, andreactor coolant temperatures.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1492.3.3.1.4 ResultsThe most limiting LONF case was with an initial Ta,,g of 564.2&deg;F (low end of the full-power Tag window(570.7&deg;F) minus uncertainties),
an initial pressurizer pressure of 2300 psia (nominal (2250 psia) plusuncertainties),
an initial main Tfeed temperature of 400'F (low full-power value), and minimum (0 percent)SGTP. Although the pressurizer PORVs were modeled as being available in this limiting case, thepressurizer sprays were sufficient in controlling the pressurizer pressure below the setpoint of the PORVs.The calculated sequence of events for the limiting LONF case is presented in Table 2.3.3-1, and transient plots of the significant plant parameters are provided in Figures 2.3.3-1 through 2.3.3-10.
Following theloss of FW from full power, the SG water level decreases to the low-low setpoint at 37.8 seconds, whichactuates a RT and the AFW system. The lack of FW causes the RCS temperature to increase.
Rod motionand turbine trip are initiated at 39.8 seconds and the RCPs continue running.
Although a temporary cooldown of the RCS occurs as a result of the RT, the RCS heats up rapidly in response to the continued lack of FW and also the turbine trip. The MSSVs open at 68.0 seconds to help dissipate the stored andgenerated heat, and at 97.8 seconds, one minute after being actuated, the AFW system begins to deliver220 gpm of AFW flow to each SG. The RCS heatup turns around shortly after the MSSVs open, and it isfurther controlled by the cooling effect of the AFW flow. The pressurizer water volume reaches amaximum value of 1384.1 ft3 at 1372.0 seconds after event initiation.
As the maximum pressurizer watervolume value is less than the total pressurizer volume of 1800 ft3, it is confirmed that the pressurizer doesnot reach a water-solid condition.
2.3.3.2 Conclusions Based on the above information, it is concluded that the LONF event will not progress into a more seriousplant condition.
Thus, all applicable event acceptance criteria are satisfied, and the AFW system isconfirmed to be adequate for long-term heat removal following an LONF event. Therefore, it has beendemonstrated that the reactor protection and safety systems ensure that the acceptable fuel design limitsare met, and the RCS and MSS pressure limits will not be exceeded as a result of an LONF event. Basedon this, the plant continues to meet the requirements of GDCs 10, 15 and 26.2.3.3.3 References
: 1. ANSI/ANS-5.1
-1979, "American National Standard for Decay Heat Power in Light WaterReactors,"
August 1979.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-150Table 2.3.3-1 Time Sequence of Events for Limiting LONF CaseEvent Time (seconds)
Main FW Flow Stops 0.0Low-Low SG Water Level RT Setpoint Reached 37.8Rods Begin to Drop and Turbine Trip Initiated 39.8On Each Loop, the MSSV with the Lowest Setting Opens 68.0Flow from Two MDAFW Pumps Initiated 97.8SG Inventory Reduction Reverses 431.0Maximum Pressurizer Water Volume Occurs 1372.0WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-151WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-15 11^-IC)0U)I.L1-0.80.6-0.4-0.2Ii ! i I II I i tU010110210Time (seconds) 310410Figure 2.3.3-1 LONF -Nuclear Power versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-152WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1521.2-vC)C)C-)(9.1-0.80.60.4-0.2-I I I I I IIII I I I I I I I I11 I i I I I ---- I- I ---- t-f, I II I0010110210Time (seconds) 310410Figure 2.3.3-2 LONF -Core Average Heat Flux versus TimeWCAP-1 7658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-153I.()wC-)C)C)-JCDC)C)C---)(~ .)C--)CD(1)Liii0.6-III I II II I I I I ISre Le-ont cFigure 2.3.3-3 LONF -Reactor Coolant Loop Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-154WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1546?n~ ,KL/H-QThC)-C)600-GD580-EGD5600(-3540-0I I I I I I I I010110210Time (seconds) 310410Figure 2.3.3-4 LONF -HL and CL Temperatures versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-155WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-15526002500-~1C)IDcnC)C)IDUDC)ID(92400-2300-2200-2100--- --~-- --I liii2000I I I I I I I I [ 1 1I I I T I I I I I I I I2010110210Time (seconds) 310410Figure 2.3.3-5 LONF -Actual Pressurizer Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1561800C)C)C)C)0C)C)C)C)C)C)C)a-160014001200-1000-I I I I I I I II I , 1 1 , ,I I I I I I I I I I I I I I I I;~flf\ -~I I [I I I III I I I I I II2LUUu010110210Time (seconds) 310410Figure 2.3.3-6 LONF -Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-157WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1571300~1.EDCfuCn7-)U-)12001100-1000900-800I I I I I I I II III I II I I II I I I I I I i I1 1 I I I I I700010110210Time (seconds) 1510410Figure 2.3.3-7 LONF -SG Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-158WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-158(f)C)C)0C)C)C)C)C)C)cn~C)C)C)C-)-C)C)10080-60-40-20-fI X~ 1111 I 1111 I I III0t I I I I I II I I I I I I I II I I I II I I I I I I I I I I I201010210Time (seconds) 310410Figure 2.3.3-8 LONF -Indicated SG Level versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-159WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-159200000~1*EC-)0c-a-a-F150000-100000-50000-00101 210 10Time (seconds)
Figure 2.3.3-9 LONF -SG Mass versus Time310410WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-160WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-160250cm 200-0C)0~0-1150-100-50-0--50-4 .I I I I I , I I t I ...010110210Time (seconds) 310410Figure 2.3.3-10 LONF -Loop AFW Flow versus TimeWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1612.3.4 Feedwater System Pipe Break (USAR Section 15.2.8)2.3.4.1 Technical Evaluation 2.3.4.1.1 Introduction A major feed line break is defined as a break in a FW line large enough to prevent the addition ofsufficient FW to maintain shell-side fluid inventory in the SGs. If the break is postulated in a feed linebetween the check valve and the SG, fluid from the SG will be discharged through the break.Furthermore, depending upon the arrangement of the AFW piping, a break in this location could precludethe subsequent addition of AFW to the affected SG. A break upstream of the feed line check valve wouldaffect the RCS only as a LONF accident.
This event is addressed by the LONF analysis presented inSection 2.3.3, "Loss of Normal Feedwater Flow."Depending upon the size of the break and the plant operating conditions at the time of the rupture, thebreak could cause either an RCS cooldown or heatup. The potential RCS cooldown resulting from asecondary pipe rupture is evaluated in the SLB analysis presented in Section 2.2.5.1, "Steam SystemPiping Failure at HZP." Therefore, only the RCS heatup effects are evaluated for a feed line break.A feed line break reduces the ability of the secondary system to remove heat generated by the core fromthe RCS for the following reasons:Reduction in FW flow to the SGs. The degradation in FW flow can cause the reactor coolanttemperature to increase prior to RT.Fluid inventory of the faulted SG may be discharged through the break, and therefore, would notbe available for decay heat removal following RT.The AFW system is provided to ensure that adequate FW is available to provide decay heat removal.Thus, the primary function of the feed line break analysis is to verify that the capacity of the AFW systemis adequate.
The AFW system is intended to provide an adequate supply of FW to ensure that:No substantial overpressurization of the RCS and MSS occurs , andSufficient liquid is maintained in the RCS so that the core remains in place and geometrically intact with no loss of core cooling capability.
The most limiting single failure in this event is the loss of one AFW train that results in the loss of oneAFW pump, thus reducing the heat removal capability of the AFW system. The AFW flow rate modeledin the analysis bounds the consequences from the loss of either a MDAFW pump or a TDAFW pump.The feed line break event is analyzed at full power conditions that bound all other power levels and allother Modes, because decay heat and stored energy are most limiting following a trip from a full powercondition.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-162The severity of the feed line rupture transient depends on a number of parameters, including break size,initial reactor power, and the functioning of various control and safety systems.
Sensitivity studiespresented in Reference 1 illustrate many of the limiting assumptions for the feed line rupture.In the analysis, the main FW control system is assumed to fail due to an adverse environment.
As a result,the water levels in all the SGs decrease equally until the low-low SG water level RT setpoint is reached.After RT, a double-ended rupture of the largest FW line is modeled.
This combination of eventsconservatively bounds the most limiting feed line break scenario that can occur. Analyses have beenperformed at full power, with and without LOOP, with credit taken for the pressurizer PORVs, but no SIactuation modeled.
For the case without offsite power available, the power is assumed to be lost at thetime of RT. This is more conservative than the case in which power is lost at the initiation of the event.The plant safety features that are available to mitigate the consequences of a feed line break event are asfollows.A RT can be initiated by one of the following.
-Two-out-of-four low-low water level signals in any one SG-Two-out-of-four high pressurizer pressure signals-Two-out-of-three high pressurizer level signals-Two-out-of-four OTAT signals-SI signal (from one of the following
[1] two-out-of-three low steamline pressure signals inany one loop or [2] two-out-of-three high containment pressure signals)-Two-out-of-four low pressurizer pressure signalsThe MSSVs open and provide secondary-side pressure protection and a heat sink source thathelps limit the RCS heatup.The PSVs may open to provide primary-side pressure protection.
Backup FW for the SGs is provided by the AFW system, which is composed of two MDAFWpumps and one TDAFW pump.The two MDAFW pumps are started on any of the following:
-Two-out-of-four low-low water level signals in any one SG-Trip of both main FW pumps-SI signal-LOOP-Manual pump start-Manual AFW system actuation WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-163The TDAFW pump is started on any of the following:
-Two-out-of-four low-low water level signals in each of two SGs-LOOP-Manual pump start-Manual AFW system actuation The MDAFW pumps are supplied power by offsite power sources, and the TDAFW pumputilizes steam from the secondary system. The pump turbine exhausts the secondary steam to theatmosphere.
: Normally, the AFW pumps take suction from the CST, but if the CST is unavailable, the essential service water system is used as the water source for the AFW pumps.The analysis of the feed line break event demonstrates that the AFW system is capable of removing thestored and residual heat, and consequently ensures the core will remain covered with water. With this, theplant is capable of returning to a safe condition following a feed line break event.2.3.4.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The following inputs and assumptions were applied in the analysis of the feed line break event:* An initial NSSS power of 3651 MWt, which includes all applicable uncertainties.
The initial RCP heat is the maximum value of 20 MWt for the case with offsite power available throughout the event. With offsite power available, maximum RCP heat is conservative for thefeed line break event because the RCPs operate continuously throughout the transient.
For thecase with a LOOP, the nominal value of 14 MWt for the initial RCP heat is used because it ismore conservative to model a slightly initial higher core power that increases the subsequent post-trip decay heat. Any post-trip heat generated by the RCPs, in combination with the coredecay heat, are the primary-side heat sources that provide the challenge to the long-term coolingcapability of the plant.A maximum initial full-power main Tfeed of 448.6&deg;F.A maximum initial full-power Tag value of 594.90F, which represents the nominal high T,,vg plusthe uncertainty of 6.5&deg;F.Main FW is assumed to be lost to all SGs at event initiation due to the feed line break. Thereverse blowdown of the faulted SG is conservatively delayed and begins when the SG inventory reaches 0 percent NRS. The combination of conditions modeled is defined to produce the mostsevere feed line break transient with the control and protection interaction considered.
The worst possible break area is modeled to maximize the blowdown discharge rate following thetime of RT, which maximizes the resultant heatup of the reactor coolant.
Choked flow is modeledat the break.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-164Operation of the pressurizer PORVs is modeled to minimize RCS pressure, which produces lower(more limiting) saturation temperatures within the RCS. To ensure the conservatism of thisassumption, SI flow, which would increase with reduced RCS pressure, is set to zero. Pressurizer spray and heaters are both assumed to be inoperable.
A minimum initial pressurizer pressure value of 2200 psia, which represents the nominal pressureof 2250 psia minus the uncertainty of 50 psi. The use of a low initial RCS pressure is consistent with modeling the operation of the pressurizer PORVs.A maximum initial pressurizer water level of 66 percent span, which conservatively bounds thehigh Tavg nominal value plus the uncertainty of 7 percent span. The initial water level in all SGs isset at 60 percent span, which is the nominal value (50 percent span) plus a level uncertainty of10 percent span.SGTP level of 10 percent (a maximum value).A minimum low-low SG water level setpoint of 0 percent NRS for RT and AFW systemactuation.
A maximum delay for RT (rod motion) of 2 seconds.A maximum delay for AFW flow initiation of 60 seconds.
The analysis conservatively accountsfor the purging of the hotter main FW in the FW piping, which delays delivery of the relatively cold auxiliary FW flow to the SGs.Failure of one protection train is taken as the worst single failure and results in one AFW pumpbeing inoperable.
The total AFW flow modeled is 594.4 gpm delivered to the three intact SGs.This is a conservative minimum value for the AFW flow following a feed line break and boundseither a single failure of the TDAFW pump or one MDAFW pump. The distribution of AFWflow is 222.7 gpm to each of two intact SGs with the third intact SG receiving 149 gpm. No flowis modeled as being delivered to the SG in the faulted loop.A maximum AFW enthalpy of 96 Btu/lbm, which corresponds to a conservatively hightemperature of 1257F.Secondary system steam relief is achieved through the self-actuated MSSVs. Note that steamrelief would normally be provided by the SG ARVs or condenser dump valves, but these wereconservatively assumed to be unavailable.
The MSSVs were modeled with opening setpoints that account for a maximum setpoint tolerance of 3 percent and appropriate line losses. Valve accumulation was modeled via a 5 psi ramp of thevalve open area from closed to full-open.
Both minimum and maximum reactivity feedback conditions have been considered.
Consistent with the limiting conditions determined for each scenario, the case analyzed with offsite poweravailable throughout the transient models maximum reactivity feedback while the case withoutWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-165offsite power (i.e., with a LOOP) models minimum reactivity feedback.
Note that the minimumreactivity feedback input includes a least negative moderator MTC value of 0 pcm/0F.Core residual/decay heat generation was based on the 1979 version of ANS 5.1 (Reference 2).ANSI/ANS-5.1-1979 is a conservative representation of the decay energy release rates.Long-term operation at the initial power level preceding the trip was assumed.No credit is taken for heat energy deposition in the RCS metal during the RCS heatup phase ofthe transient.
* No credit is taken for charging or letdown.No credit is taken for the following potential protection logic signals to mitigate the consequences of the accident:
-High pressurizer pressure-OTAT-High pressurizer level-High containment pressureAcceptance CriteriaBased on the expected frequency of occurrence, the feed line break event is considered to be aCondition IV event as defined by "Nuclear Safety Criteria for the Design of Stationary Pressurized WaterReactor Plants,"
ANSI N 18.2-1973.
The following items summarize the acceptance criteria associated with the analysis of this event:* Pressures in the RCS and MSS must remain less than 110 percent of the respective designpressures.
With respect to peak RCS and MSS pressures, the feed line break event is bounded by theLOL/TT event described in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip,Inadvertent Closure of a Main Steam Isolation Valve, and Loss of Condenser Vacuum,"
in whichassumptions are made to conservatively calculate the RCS and MSS pressure transients.
For thefeed line break event, turbine trip occurs after RT, whereas for the LOL/TT event, the turbine tripis the initiating fault. Therefore, the primary to secondary power mismatch and resultant RCS andMSS heatup and pressurization transients are always more severe for the LOL/TT event. Basedon this, no explicit calculation of maximum RCS or MSS pressure is performed for this event.* Any fuel damage calculated to occur must be sufficiently limited to the extent that the core willremain in place and intact with no loss of core cooling capability.
With respect to fuel damage due to "dryout,"
where the water level in the vessel drops below thetop of the core, Westinghouse has established an internal criterion that no bulk boiling occurs inthe primary coolant system prior to event turn around. Turn around occurs when the heat removalcapability of the SGs, being fed AFW, exceeds NSSS heat generation.
This conservatively ensuresWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-166that the core remains covered with water and thereby will remain in place and geometrically intact with no loss of core cooling capability.
This criterion is very limiting and is adopted forconvenience in interpreting the results of this study. Compliance can be determined by checkingthe temperature plots (HL saturation, HL, and CL) for all the loops and verifying that neither theHL nor CL temperatures exceed the saturation temperature prior to turn around. It should benoted that precluding bulk boiling in the RCS is the limiting criterion that is considered in thefeed line break analysis.
With respect to possible fuel damage due to DNB, the feed line breakevent would be bounded by either the SLB analysis presented in Section 2.2.5.1 or the LOL/TTanalysis in Section 2.3.1.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the feed line break acceptance criteria are provided as follows.GDC 27 (Combined Reactivity Control Systems Capability) requires that the reactivity controlsystems be designed to have a combined capability, in conjunction with poison addition by theECCS, of reliably controlling reactivity changes so that, under postulated accident conditions andwith appropriate margin for stuck rods, the capability to cool the core is maintained.
For the feedline break event, this is shown to be met by demonstrating that the applicable fuel damagecriterion is satisfied.
It should be noted that in the feed line break analysis cases presented, poisonaddition via SI actuation is conservatively not credited but is available and would be actuated.
GDC 28 (Reactivity Limits) requires that the reactivity control system be designed withappropriate limits on the potential amount and rate of reactivity increase so that the effects ofpostulated reactivity accidents can neither result in damage to the RCPB greater than limited localyielding, nor sufficiently disturb the core, its support structures, or other RPV internals to impairsignificantly the capability to cool the core. For the feed line break event, this is shown to be metby demonstrating that the peak RCS pressure is less than 110 percent of the design pressure, which ultimately ensures that the RCPB pressure limits are not exceeded, and by confirming thatthe fuel damage requirements are met.GDC 35 (Emergency Core Cooling) requires that the RCS and associated auxiliaries be designedwith a safety system able to provide abundant emergency core cooling.
For the feed line breakevent, this is shown to be met by demonstrating that the fuel damage criterion is met, whichconfirms that the AFW system provides abundant cooling for the RCS, even with themost-limiting single failure considered.
2.3.4.1.3 Description of Analyses and Evaluations A detailed analysis using the RETRAN computer code (Reference
: 3) was performed to determine theplant transient conditions for the feed line break event. A RETRAN input model specific to WCGS wasdeveloped to simulate the core neutron kinetics, RCS, pressurizer, pressurizer sprays, SGs, MSSVs, andthe AFW system. Feed line break cases were modeled to address minimum and maximum reactivity feedback conditions with and without a LOOP occurring.
The RETRAN code computed thetime-dependent trends of pertinent variables, including the pressurizer
: pressure, pressurizer water volume,SG mass, and reactor coolant temperatures.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1672.3.4.1.4 ResultsCalculated plant parameters following a major feed line break are shown in Figures 2.3.4-1through 2.3.4-10.
Results for the limiting case with offsite power available are presented inFigures 2.3.4-1 through 2.3.4-5.
Results for the limiting case where offsite power is lost are presented inFigures 2.3.4-6 through 2.3.4-10.
The calculated sequence of events for each of the reported cases isprovided in Tables 2.3.4-1 and 2.3.4-2.The system response following the feed line break is similar for both cases analyzed.
In both cases, the primaryand secondary pressures increase prior to RT. After RT occurs on low-low SG water level, the pressuredecreases
: sharply, due to the cooldown caused by the break, until SLI occurs. The pressure in the faulted SGcontinues to decrease, whereas the pressure in the intact SGs and the primary side begins to increase until thesafety valve settings are reached.
Results presented in Figures 2.3.4-2 and 2.3.4-5 (with offsite poweravailable) and Figures 2.3.4-7 and 2.3.4-10 (without offsite power) show the predicted pressures in the RCSand MSS. As was previously discussed in Section 2.3.4.1.3, the feed line break analysis is bounded byLOL/TT with respect to meeting maximum pressure limits for the RCS and MSS. This is especially truebecause the reported cases for the feed line break analysis model operation of the pressurizer PORVs to reducetransient RCS pressure which is conservative with respect to bulk boiling concerns.
The primary temperatures are stable or increase slightly prior to RT and decrease sharply duringcooldown after RT. Once the heat-up begins, the primary temperature increases until the heat removalcapability of the intact SGs, with the inventory maintained by the AFW System, equals the decay heatgenerated in the core plus pump heat ("turn around" time). The peak primary temperature remains belowthe saturation temperature although the margin to boiling is decreased.
At the predicted time of turnaround, the minimum margin to HL saturation for the case with offsite power is 65.3&deg;F, and 40.57F for thecase without offsite power. Thus, there is no bulk boiling in the RCS.2.3.4.2 Conclusions Based on the above information, it is concluded that for the postulated feed line break event the modeledAFW system performance is adequate for long-term heat removal.
The results confirm that for the feedline break event, the AFW system can adequately remove decay heat to preclude uncovering of thereactor core. Based on this, the plant continues to meet the requirements of GDCs 27, 28 and 35.2.3.4.3 References
: 1. WCAP-9230, "Report on the Consequences of a Postulated Main Feedline Rupture,"
January 1978.2. ANSI/ANS-5.1
-1979, "American National Standard for Decay Heat Power in Light WaterReactors,"
August 1979.3. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-168Table 2.3.4-1 Time Sequence of Events for Limiting Feed Line Break Case With Offsite PowerAvailable Event/Parameter Time (sec)Feed Line Break Occurs Causing Loss of FW to all SGs due to Harsh Environment 0.0Pressurizer PORV Opens (First Occurrence) 13.9Low-Low SG Water Level RT Setpoint Reached in all SGs 36.2Rods Begin to Drop and Faulted SG Begins Discharging Fluid Directly Out the Break 38.2Pressurizer PORV Closes (First Occurrence) 41.7AFW Flow is Delivered to Intact SGs 96.2Low Steam Line Pressure SI Setpoint Reached in Ruptured SG 132.7Main Steam Line Isolation Valves Closed 149.7SG Safety Valve Setpoint Reached in Intact SGs (First Occurrence) 579.0Core Decay Heat plus RCP Heat Decreased to AFW Heat Removal Capacity
-1700.0Table 2.3.4-2 Time Sequence of Events for Limiting Feed Line Break Case Without Offsite PowerAvailable Event/Parameter Time (sec)Feed Line Break Occurs Causing Loss of FW to all SGs due to Harsh Environment 0.0Pressurizer PORV Opens (First Occurrence) 13.9Low-Low SG Water Level RT Setpoint Reached in all SGs 36.3Rods Begin to Drop and Faulted SG Begins Discharging Fluid Directly out the Break 38.3Power Lost to RCPs 40.3Pressurizer PORV Closes (First Occurrence) 42.1AFW Flow is Delivered to Intact SGs 96.3Low Steam Line Pressure SI Setpoint Reached in Ruptured SG 105.6Main Steam Line Isolation Valves Closed 122.6SG Safety Valve Setpoint Reached (first occurrence) 553.4Core Decay Heat Decreased to AFW Heat Removal Capacity
-1100WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-169WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-169Li-00~C-)1.21'0.8-0.6-0.4-0.2-....11 ..AI I .0 1 2 3 410 10 10 10 10Time (seconds) 41*0.8-0.6-0.4-0.2-..............I I I I [ I I I I [ I l l[A.'.U01010210Time (seconds) 3104102, 0.1E-010av -0.1E-01-0.2E-01g -0.3E-01-0.4E-01--0.5E-01.....................-U.DL-U I010110210Time (seconds) 31010Figure 2.3.4-1Feed Line Break with Offsite Power Available Nuclear Power, Core Heat Flux and Total Core Reactivity versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-170WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1702,co,2~1U)NJCn0-0 1 2 310 10 10 10Time (seconds) 41001 2 310 10 10 10Time (seconds) 410Figure 2.3.4-2Feed Line Break with Offsite Power Available Pressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-171WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-17 11.2a,0cnC,)Cl)C-,1.1-1-0.90.8-0.7-..............................................................................
...............
0.t" I I i i I I i l i I I I I I I .I I l I i .I i I I I2O010110210Time (seconds) 310410Fraction of Ini tial Total Plant Feedwoter Moss Flownr .U-co2co,n iU.-1-I I II* i.I liiiI I I I I II-1010I10210Time (seconds) 310410Figure 2.3.4-3Feed Line Break with Offsite Power Available Reactor Coolant Flow and FW Line Break Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-172WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-172Faul ted Loop Hot Leg Temp.Fau lted Loop Cold Leg Temp.-Faulted Loop Hot Leg T-sat700-it 650 ...... ...--"----*--;-.
.....S 6 0 0 ................... -.E 550 ..................t-- -0 1 2 3 410 10 10 10 10Time (seconds)
Intact Loop Hot Leg Temp.Intact Loop Cold Leg Temp.Intact Loop Hot Leg T-sat1-1CDCLE0 1 2 310 10 10 10Time (seconds) 410Figure 2.3.4-4Feed Line Break with Offsite Power Available Faulted Loop and Intact Loop Reactor Coolant Temperatures versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-173Loop 1 F Foul ted)Loop 2 (Intoct)Loop 3 (Intact)Loop 4 I ntact)0ci,0~0~V-c(I)00VVC-,E0VU)0 1 2 310 10 10 10Time (seconds) 410Figure 2.3.4-5Feed Line Break with Offsite Power Available SG Shell Pressure versus TimeWCAP- 17658-NP August 2013WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-174WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-174C)L)CI,UC-,1.210.80.6-0.4-0.2-............0 IiII I I III0 1 2 3 410 10 10 10 10Time (seconds) 41-0.8-0.6-0.4-0.2-0I I I I LI III I I I 1 1 1 -- l , t01 2 3 410 10 10 10 10Time (seconds)
Ir*["nit.)n.V-0.1E-01
--0.2E-01
--0.3E-01
--0.4E-01
--0.5E-01
--0.6E-01.....................010110210Time (seconds) 310I410Figure 2.3.4-6Feed Line Break without Offsite PowerNuclear Power, Core Heat Flux and Total Core Reactivity versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-175WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-175C')CnC')a,0 1 2 3 410 10 10 10 10Time (seconds) 0 1 2 310 10 10 10Time (seconds) 410Figure 2.3.4-7Feed Line Break without Offsite PowerPressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1761-a,-M-0 1 2 310 10 10 10Time (seconds) 410Fraction ofI i tial Total PIant Feedwater Mass FlowCL)C',co,0WL2U.,J'-.4 UU-0.5--1-...............
...............
-2010I &#xfd; I .I .I I I I I I I I I Ii I I i I I t I I [ i r I i I I i i I I I i I I i210210Time (seconds) 310410Figure 2.3.4-8Feed Line Break without Offsite PowerReactor Coolant Flow and FW Line Break Flow versus TimeWCAP- 17658-NP August 2013WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-177WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-177Faulted Loop Hot Leg Temp.Faulted Loop Cold Leg Temp.-Faul ted Loop Hot Leg T-sotCLE0 1 2 310 10 10 10Time (seconds) 410Intact LoopI Intact LoopIntact Loop700650"~600 ..........o 55 0 .. ......F--Hot Leg Temp.Cold Leg Temp.Hot Leg T-sat0 1 2 310 10 10 10Time (seconds) 410Figure 2.3.4-9Feed Line Break without Offsite PowerFaulted Loop and Intact Loop Reactor Coolant Temperatures versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-178WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-178LoopLoopL oo pLoopE1400-1200.1000T13.-800.600.= 400.E 200 --- ...200CD65) ~1234IFo Ited)In tactIntact)Intact)01 2 310 10 10 10Time (seconds) 410Figure 2.3.4-10Feed Line Break without Offsite PowerSG Shell Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-1792.4 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE2.4.1 Partial and Complete Loss of Forced Reactor Coolant Flow (USAR Sections 15.3.1and 15.3.2)2.4.1.1 Technical Evaluation 2.4.1.1.1 Introduction A loss of forced reactor coolant flow accident (USAR Sections 15.3.1 and 15.3.2) can result from thefollowing:
* Mechanical or electrical failure in one or more RCPs* Interruption in the power supplying one or more of the RCPs* Reduction in RCP motor supply frequency If the reactor is at power at the time of the event, the immediate effect from the loss of forced coolantflow is a rapid increase in the coolant temperature.
This increase in coolant temperature could result inDNB, with subsequent fuel damage, if the reactor is not promptly tripped.The following signals provide protection against a loss of forced reactor coolant flow incident:
* Low reactor coolant loop flow RT* UV on RCP power supply busses RT* Underfrequency (UF) on RCP power supply busses RTThe RT on low reactor coolant loop flow provides primary protection against partial loss-of-flow (PLOF)conditions.
This function is generated by two-out-of-three low-flow signals in any reactor coolant loop.Above Permissive P-8, low flow in any loop will actuate a RT. Between approximately 10 percent power(Permissive P-7) and the power level corresponding to Permissive P-8, low flow in two loops will actuatea RT. RT on low flow is blocked below Permissive P-7 because there is insufficient heat production to beconcerned about DNB.The RT on RCP UV is provided to protect against conditions that can cause a loss of voltage to all RCPs,that is, LOOP. An UV RT serves as an anticipatory backup to the low reactor coolant loop flow trip. TheUV trip function is blocked below approximately 10 percent power (Permissive P-7).The RCP UF RT is provided to trip the reactor for an UF condition resulting from frequency disturbances on the power grid. The RCP UF RT function is blocked below Permissive P-7. This trip function alsoserves as an anticipatory backup to the low reactor coolant loop flow trip.2.4.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThis accident was analyzed using the RTDP methodology (Reference 1). Initial NSSS power wasconservatively modeled to be at 3651 MWt, which includes all applicable uncertainties.
The RCSpressure and vessel average temperature were assumed to be at their nominal values. MMF was alsoWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-180assumed.
Uncertainties in initial conditions were accounted for in the DNBR limit value as described inthe RTDP.A conservatively large absolute value of the Doppler-only power coefficient was used. The analysis alsoassumed a conservative MTC of 0 pcm/&deg;F at HFP conditions.
This resulted in the maximum core powerand hot spot heat flux during the initial part of the transient when the minimum DNBR is reached.Engineered safety systems (such as SI) are not required to function.
No single active failure in any systemor component required for mitigation will adversely affect the consequences of this event.A partial loss of forced reactor coolant flow incident is classified as a Condition II event as defined by theANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSIN 18.2-1973.
A complete loss of forced reactor coolant flow incident is classified by the ANS as aCondition III event. However, for conservatism, the incident was analyzed to Condition II criteria.
Theimmediate effect from a complete loss of forced reactor coolant flow is a rapid increase in the reactorcoolant temperature and subsequent increase in RCS pressure.
The following two items identify theacceptance criteria associated with the analysis of the loss of flow events:The CHF is not to be exceeded.
This is met by demonstrating that the minimum DNBR does notdecrease below the SAL value at any time during the transient.
Pressures in the RCS and MSS are maintained below 110 percent of their respective designpressures.
With respect to peak RCS and MSS pressures, the loss of forced reactor coolant flow event is boundedby the LOL/TT event described in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip,Inadvertent Closure of a Main Steam Isolation Valve, and Loss of Condenser Vacuum,"
in whichassumptions are made to conservatively calculate the RCS and MSS pressure transients.
For the loss offorced reactor coolant flow event, turbine trip occurs after RT, whereas for loss of load, the turbine tripis the initiating incident.
Therefore, the power mismatch between the primary and secondary sides andthe resultant temperature and pressure transients of the RCS and MSS are always more severe forLOL/TT than for the loss of forced reactor coolant flow. Based on this, no explicit calculation ofmaximum RCS or MSS pressure is performed for this event.The above acceptance criteria are based on meeting the relevant regulatory requirements of 10 CFR 50,Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of the specific GDCsthat are related to the loss of forced reactor coolant flow acceptance criteria are provided as follows.The specific acceptance criteria for this event are as follows:GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the loss of forced reactor coolant flow event, this is shown to be met bydemonstrating that the DNBR remains above the 95/95 DNBR limit at all times during thetransient.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-181GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.
For theloss of forced reactor coolant flow event, this is shown to be met by demonstrating that the peakRCS pressure is less than 110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.
For the loss of forced reactorcoolant flow event, which results in a RT, this is shown to be met by demonstrating that theDNBR remains above the 95/95 DNBR limit at all times during the transient with a trip reactivity that accounts for the most reactive rod stuck out of the core.2.4.1.1.3 Description of Analyses and Evaluations The following loss of forced reactor coolant flow cases were analyzed:
* Loss of power to two RCPs (PLOF)* Loss of power to all RCPs (complete loss of flow (CLOF))* 5 Hz/second frequency decay of the RCPs power supply (CLOF-underfrequency (CLOF-UF))
The transients were analyzed with two computer codes. First, the RETRAN computer code (Reference 2)was used to calculate the following:
* Loop and core flows during the transient
* Time of RT based on the calculated flows* Nuclear power transient a Primary system pressure and temperature transients The VIPRE computer code (Reference
: 3) was then used to calculate the heat flux and DNBR based on thenuclear power and RCS temperature (enthalpy),
: pressure, and flow from the output of the RETRANtransient run. The DNBR transients presented represent the minimum of the typical or thimble cell for thefuel.An evaluation of the P-8 permissive setpoint was performed and it was determined that the currentplant-specific value continued to provide adequate protection.
No change to the existing setpoint wasdeemed necessary.
Additionally, the effects of loop-to-loop flow asymmetry due to 10 percent SGTP imbalance have beenconsidered in the analysis.
2.4.1.1.4 ResultsThe PLOF case resulted in a low reactor coolant loop flow RT signal and the CLOF case resulted in anUV RCP RT signal. The CLOF-UF case resulted in an UF RCP RT signal. The VIPRE (Reference 3)WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-182analysis for these scenarios confirmed that the minimum DNBR acceptance criterion was met. Fuelcladding damage criteria were not challenged in any of the loss of forced reactor coolant flow casesbecause the DNB criterion was met.The analyses of the loss of flow events also demonstrated that the peak RCS and MSS pressures werewell below their respective limits.The most limiting of these cases in terms of the minimum calculated DNBR was the CLOF case. Thetransient results for each case are presented in Figures 2.4.1 -1 through 2.4.1-21.
The sequence of eventsfor each case is presented in Table 2.4. 1-1. Numerical results for the analyses are shown in Table 2.4.1-2.The analysis demonstrates that, for the aforementioned loss of flow cases, the DNBR did not decreasebelow the SAL value at any time during the transients.
Therefore, no fuel or cladding damage is predicted.
Also, the peak RCS and MSS pressures remained below their respective limits at all times. All applicable acceptance criteria were therefore met.The protection features identified in Section 2.4. 1. 1. 1 provide mitigation for the loss of forced reactorcoolant flow transients such that the above criteria are satisfied.
Furthermore, the results and conclusions of the loss of flow analysis will be confirmed on a cycle-specific basis as part of the normal RSE process.2.4.1.2 Conclusion The analyses of the decrease in forced reactor coolant flow event have been reviewed.
It is concluded thatthe analyses have adequately accounted for plant operations at a nominal NSSS power of up to3651 MWt, and were performed using acceptable analytical models. The review further concludes that theevaluation has demonstrated that the reactor protection and safety systems will continue to ensure that thespecified acceptable fuel design limits and the RCS and MSS pressure limits will not be exceeded as aresult of this event. Based on this, it is concluded that the plant will continue to meet the requirements ofGDCs 10, 15, and 26.2.4.1.3 References I. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"
April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRJETARY CLASS 32-183Table 2.4.1-1 Time Sequence of Events -Loss of Forced Reactor Coolant FlowCase Event Time (seconds)
Loss of Power to Two RCPs (PLOF) Flow Coastdown Begins 0.0Reactor Coolant Low-Flow Trip Setpoint Reached 1.5Rods Begin to Drop 2.5Minimum DNBR Occurs 4.3Loss of Power to All RCPs (CLOF) Flow Coastdown Begins 0.0Rods Begin to Drop(D 1.5Minimum DNBR Occurs 3.55 Hz/sec Frequency Decay of the Frequency Decay Begins 0.0RCPs Power Supply (CLOF-UF)
Underfrequency RT Setpoint Reached 0.6Rods Begin to Drop 1.2Minimum DNBR Occurs 3.3Note:I. LUV RT (rods begin to drop) is assumed to occur 1.5 seconds following the loss of bus voltage.Table 2.4.1-2 Results -Loss of Forced Reactor Coolant FlowMinimum DNBR Limit ValueLoss of Power to Two RCPs (PLOF) 1.82 1.52Loss of Power to All RCPs (CLOF) 1.69 1.525 Hz/sec Frequency Decay of the RCPs Power Supply (CLOF-UF) 1.73 1.52WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-184WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-184I0I.f-.....................................
0 .8 -..............................................i-,E<C.)0.61 ..........................................................
0.41 ...........
...........
............
0.2-1...............................................n-i0246810Time (sec)Figure 2.4.1-1 PLOF -Core Volumetric Flow Rate versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-185Loop 1Loop 2-Loop 3Loop 4E-50 2 4 6 8lime (sec)10Figure 2.4.1-2 PLOF -Loop Volumetric Flow Rates versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-186w0~0 2 4 6 8lime (sec)10Figure 2.4.1-3 PLOF -Nuclear Power versus TimeWCAP-l 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-187WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-187CnC',C/3P,0'l0 2 46810Time (sec)Figure 2.4.1-4 PLOF -Pressurizer Pressure versus TimeWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportALIgust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-188CD)6Time (sec)10Figure 2.4.1-5 PLOF -Core Average Heat Flux versus TimeWCAP- 17658-NP August 2013WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-189WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-189wa,czC-,0 2 4 6 8Time (sec)10Figure 2.4.1-6 PLOF -Hot Channel Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-190WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-190EE0 2 46810Time (sec)Figure 2.4.1-7 PLOF -Minimum DNBR versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-191WESTJNGHOUSE NON-PROPRIETARY CLASS 3 2-1911.2U-I-.Ea-,C-,0 2 4 6 8Time (see)10Figure 2.4.1-8 CLOF -Core Volumetric Flow Rate versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-192WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-192Loop 1Loop 2Loop 3Loop 40C,0E_..J2 46810Time (sec)Figure 2.4.1-9 CLOF -Loop Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-193WESTINGHOUSE NON-PROPRiETARY CLASS 3 2-193wC-)Time (sec)Figure 2.4.1-10 CLOF -Nuclear Power versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-194WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1942-C/)C/)M~0 2 4 6 8Time (sec)10Figure 2.4.1-11 CLOF -Pressurizer Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-195a.,)C-D0 2 4 6 8 10Time (sec)Figure 2.4.1-12 CLOF- Core Average Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-196WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1961.2WQ-):izciC-,ci0 2 4 6 8Time (sec)10Figure 2.4.1-13 CLOF -Hot Channel Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-197WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-197EE02 46810Time (sec)Figure 2.4.1-14 CLOF -Minimum DNBR versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-198Lj~a.)C)C-)a.)E-5a.)(-)0 2 4 6 8lime (sec)10Figure 2.4.1-15 CLOF-UF -Core Volumetric Flow Rate versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-199Loop 1Loop 2Loop 3Loop 4CD,czl0 2 46810Time (sec)Figure 2.4.1-16 CLOF-UF -Loop Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-200WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-200_L__CD0 2 4 6 8Time (sec)10Figure 2.4.1-17 CLOF-UF -Nuclear Power versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-201WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-201Cf)C/)r14C/)M)0 2 4 6 8 10Time (sec)Figure 2.4.1-18 CLOF-UF-Pressurizer Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-202WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-202LLa,a,a,a,C-)0 2 4 6 8Time (sec)10Figure 2.4.1-19 CLOF-UF -Core Average Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-203wa,C-)0 2 4 6 8Time (see)10Figure 2.4.1-20 CLOF-UF -Hot Channel Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-204Q__0 2 4 6 8Time (see)10Figure 2.4.1-21 CLOF-UF-Minimum DNBR versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2052.4.2 Reactor Coolant Pump Shaft Seizure (Locked Rotor) and Shaft Break(USAR Sections 15.3.3 and 15.3.4)2.4.2.1 Technical Evaluation 2.4.2.1.1 Introduction The event postulated is an instantaneous seizure of a RCP rotor or the sudden break of the RCP shaft.Flow through the affected reactor coolant loop is rapidly reduced, leading to initiation of a RT on a lowreactor coolant flow signal.Following initiation of the RT, heat stored in the fuel rods continues to be transferred to the coolant,causing the coolant to expand. At the same time, heat transfer to the shell side of the SGs is reduced, firstbecause the reduced flow results in a decreased tube-side film heat transfer coefficient, and secondbecause the temperature differential between the reactor coolant in the tubes and the shell-side fluid isdecreased.
The rapid expansion of the coolant in the reactor core causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steamvolume, actuates the automatic spray system, opens the PORVs, and opens the PSVs, in that sequence.
The PORVs are designed for reliable operation and are expected to function properly during the event.However, for conservatism, their pressure-reducing effect, as well as the pressure-reducing effect of thepressurizer spray, was not included in the analysis.
The consequences of a locked rotor (that is, an instantaneous seizure of a pump shaft) are very similar tothose of a pump shaft break. The initial rate of the reduction in coolant flow is slightly greater for thelocked rotor event. However, with a broken shaft, the impeller could conceivably be free to spin in thereverse direction.
The effect of reverse spinning is a reduced core flow when compared to the locked rotorscenario.
The analysis considers only one scenario; it represents the most limiting (conservative) combination of conditions for the locked rotor and pump shaft break events.2.4.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions There were three locked rotor cases analyzed, with each being applicable to the WCGS: one for peak RCSpressure and PCT concerns with LOOP, one for peak RCS pressure and PCT concerns without LOOP, anda third to determine the percentage of rods-in-DNB.
For the cases performed to evaluate peak RCSpressure and PCT concerns, one locked rotor and shaft break was simulated with all reactor coolant loopsin operation; one of these cases accounted for a LOOP and the other considered a continued supply ofoffsite power. Inputs for these cases were designed to maximize the RCS pressure and claddingtemperature transients; the STDP was applied for these cases. Initial core power, reactor coolanttemperature, and pressurizer pressure were modeled to be at their maximum values consistent withfull-power conditions, including allowances for calibration and instrument errors. The initial reactorcoolant flow was the TDF. These inputs resulted in a conservative calculation of the coolant insurge intothe pressurizer, which in turn resulted in a maximum calculated peak RCS pressure.
The case thatconsidered a LOOP conservatively modeled the intact RCPs as being tripped coincident with RT.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-206The third case was run to confirm that the percentage of rods that experience DNB, also referred to aspercentage of rods-in-DNB, with LOOP is less than that considered in the radiological analysis.
As in thepeak RCS pressure/PCT case, one locked rotor and shaft break was simulated with all reactor coolantloops in operation.
Initial NSSS power was conservatively modeled to be at 3651 MWt, which includesall applicable uncertainties.
The pressurizer pressure and Tavg were modeled to be at their nominal values.The initial reactor coolant flow was the MMF. Uncertainties in initial pressure and temperature conditions were accounted for in the DNBR SAL value as described in the RTDP (Reference 1).A least negative MTC and a conservatively large (absolute value) Doppler-only power coefficient weremodeled in the analysis.
The negative reactivity from control rod insertion/scram was based on4.0 percent Ak/k trip reactivity from HFP conditions.
Engineered safety systems (such as SI) are not required to function.
No single active failure in any systemor component required for mitigation will adversely affect the consequences of this event.Acceptance CriteriaThe RCP locked rotor/shaft break accident is classified as a Condition IV event as defined by the ANS's"Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSIN 18.2-1973.
An RCP locked rotor/shaft break results in a rapid reduction in forced reactor coolant loopflow that increases the reactor coolant temperature and subsequently causes the fuel cladding temperature and RCS pressure to increase.
The following items summarize the acceptance criteria for the analysis ofthis event:The potential for fuel cladding damage due to the combination of high cladding temperature and theexothermic zirconium oxidation process (zirconium-water reaction) provides a failure mechanism thatcould affect the core cooling capability.
With respect to fuel cladding temperature, the maximum claddingtemperature at the core hot spot must remain below 2700'F, and the zirconium-water reaction at the corehot spot must be less than 16 percent by weight. Satisfying these criteria conservatively ensures that thecore will remain in place and geometrically intact with no loss of core cooling capability.
Pressures in the RCS and MSS are to be maintained below 110 percent of the respective design pressures.
With respect to the MSS pressure transient, this event is bounded by the LOL/TT event discussed inSection 2.3.1, "Loss of External Electrical Load, Turbine Trip, Inadvertent Closure of a Main SteamIsolation Valve, and Loss of Condenser Vacuum."
This is because the turbine trip occurs later, coincident with RT, in this event compared to the LOL/TT event where the turbine trip is the initiating fault. Thegreater mismatch between the primary-side power and secondary-side power makes the MSS pressuretransient more severe for the LOL/TT event. Thus, the peak MSS pressure is not reported for the lockedrotor analysis.
Because it is a Condition IV event, the locked rotor/shaft break transient is allowed to result in a minimalrelease of radioactive material such that the calculated doses at the site boundary are within acceptable limits (see Section 4.3.5 of Enclosure VI of this LAR). For dose considerations, fuel failure isconservatively assumed for all fuel rods that are shown to experience DNB. In the dose analysis (seeSection 4.3.5 of Enclosure VI of this LAR), 5 percent of the fuel rods were assumed to have failed andWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-207released radioactive material as a result of a locked rotor/shaft break event. Therefore, the total percentage of rods-in-DNB must be less than the 5 percent value used in the dose analysis.
The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the locked rotor/shaft break acceptance criteria are provided as follows.* GDC 27 (Combined Reactivity Control Systems Capability) requires that the reactivity controlsystems be designed to have a combined capability, in conjunction with poison addition by theECCS, of reliably controlling reactivity changes under postulated accident conditions, withappropriate margin for stuck rods, to assure the capability to cool the core is maintained.
In thelocked rotor event analysis, the applied trip reactivity accounts for the highest worth rod beingstuck fully out of the core, and the capability to cool the core is demonstrated by showing that thelimits for fuel cladding temperature, zirconium-water
: reaction, and RCS pressure are met.GDC 28 (Reactivity Limits) requires that the reactivity control systems be designed to assure thatthe effects of postulated reactivity accidents can neither result in damage to the RCPB greaterthan limited local yielding, nor disturb the core, its support structures, or other reactor vesselinternals so as to significantly impair the capability to cool the core. For the locked rotor event,this is shown to be met by demonstrating that the limits for fuel cladding temperature, zirconium-water
: reaction, and RCS pressure are met.GDC 31 (Fracture Prevention of Reactor Coolant Pressure Boundary) requires that the RCPB bedesigned with sufficient margin to assure that, under specified conditions, it will behave in anon-brittle manner and the probability of a rapidly propagating fracture is minimized.
For thelocked rotor event, this is shown to be met by demonstrating that the RCS pressure limit is met.2.4.2.1.3 Description of Analyses and Evaluations The locked rotor transient was analyzed with two primary computer codes. First, the RETRAN computercode (Reference
: 2) was used to calculate the loop and core flows during the transient, the time of RTbased on the calculated flows, the nuclear power transient, and the primary system pressure andtemperature transients.
The VIPRE code (Reference
: 3) was then used to calculate the PCT using thenuclear power and RCS temperature (enthalpy),
: pressure, and flow from RETRAN.For the peak RCS pressure evaluation, the initial pressure was conservatively estimated to be 50 psi abovethe nominal pressure of 2250 psia, which accounts for initial condition uncertainties in the pressurizer pressure measurement and control channels.
This was done to obtain the highest possible rise in thecoolant pressure during the transient.
The pressure response reported in Table 2.4.2-2 corresponds to thelocation in the RCS that has the maximum pressure, that is, in the lower plenum of the reactor vessel.No credit was taken for the pressure-reducing effect of the pressurizer PORVs, pressurizer spray, or steamdump. Although these systems are expected to function and would result in a lower peak pressure, anadditional degree of conservatism was provided by not including their effect. The PSV model includeda +2 percent valve opening tolerance above the nominal setpoint of 2460 psig plus a 1 percent setWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-208pressure shift and a 1. 153-second purge time delay to account for the existence of PSV water-filled loopseals, as described in Reference 4.Post-DNB heat transfer is limited to film boiling, and the film boiling coefficient was calculated in theVIPRE code (Reference
: 3) using the Bishop-Sandberg-Tong heat transfer correlation.
The fluid properties were evaluated at film temperature.
The code calculated the film coefficient at every time step based uponthe actual heat transfer conditions at the time. The nuclear power, system pressure, bulk density, and RCSflow rate as a function of time were based on the RETRAN results.The magnitude and time dependence of the heat transfer coefficient between the fuel and cladding (gapcoefficient) has a pronounced influence on the thermal results.
The larger the value of the gap coefficient, the more heat is transferred between the pellet and cladding.
Based on investigations on the effect of thegap coefficient upon the maximum cladding temperature during the transient, the gap coefficient wasassumed to increase from a steady-state value consistent with the initial fuel temperature to approximately 10,000 Btuihr-ft 2-OF at the initiation of the transient.
Therefore, the large amount of energy stored in thefuel because of the small initial gap coefficient was released to the cladding at the initiation of thetransient.
The zirconium-water reaction can become significant above 1800'F (cladding temperature).
TheBaker-Just parabolic rate equation was used to define the rate of zirconium-water reaction.
The effect ofthe zirconium-water reaction was included in the calculation of the PCT temperature transient.
2.4.2.1.4 ResultsWith respect to the peak RCS pressure, PCT and zirconium-water
: reaction, the analysis demonstrated thatall applicable acceptance criteria were met for the WCGS. The calculated sequence of events is presented in Table 2.4.2-1 for the locked rotor/shaft break event. The results of the calculations (peak pressure, PCTand zirconium-water reaction) for the limiting case (with a LOOP) are summarized in Table 2.4.2-2.
Thetransient results for the peak pressure/PCT cases (with a LOOP and without a LOOP) are provided inFigures 2.4.2-1 through 2.4.2-6, and the transient results for the rods-in-DNB case are provided inFigures 2.4.2-7 through 2.4.2-12.
The locked rotor/shaft break analysis performed for the WCGS demonstrated that the PCT calculated forthe hot spot remained considerably less than 2700'F, and the amount of zirconium-water reaction wassmall. Under such conditions, the core would remain in place and intact with no loss of core coolingcapability.
The analysis also confirmed that the peak RCS pressure reached during the transient was less than theacceptance limit, and thereby, the integrity of the primary coolant system was demonstrated.
The totalnumber of rods-in-DNB was less than 5 percent.
The low reactor coolant flow RT function providedmitigation for the locked rotor/shaft break transient such that the above criteria were satisfied.
Furthermore, the results and conclusions of this analysis will be confirmed on a cycle-specific basis aspart of the normal RSE process.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2092.4.2.2 Conclusion The analyses of the sudden decrease in core coolant flow due to a locked rotor/shaft break event havebeen examined.
It is concluded that the analyses have adequately accounted for plant operation at theanalyzed power level and were performed using acceptable analytical models. It is further concluded thatthe evaluation has demonstrated that the reactor protection and safety systems will continue to ensure thatthe ability to insert control rods is maintained, the RCS pressure limit will not be exceeded, the RCPBwill behave in a nonbrittle manner, the probability of propagating fracture of the RCPB is minimized, andadequate core cooling will be provided.
Based on this, it is concluded that the plant will continue to meetthe requirements of GDCs 27, 28, and 31.2.4.2.3 References
: 1. WCAP-1 1397-P-A, "Revised Thermal Design Procedure,"
April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.4. WCAP-129 10, Rev. I-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-210WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 10Table 2.4.2-1 Time Sequence of Events -RCP Locked Rotor/Shaft BreakCase Event Time (seconds)
Locked Rotor/Shaft Break -Rotor on One Pump Locks or the Shaft 0.0Overpressurization/PCT with LOOP BreaksReactor Coolant Low-Flow RT Setpoint 0.04ReachedRods Begin to Drop 1.04Undamaged RCPs Lose Power and Begin to 1.04Coast DownM')Maximum Cladding Temperature Occurs 3.90Maximum RCS Pressure Occurs 4.75Locked Rotor/Shaft Break -Rotor on One Pump Locks or the Shaft 0.0Overpressurization/PCT without BreaksLOOPReactor Coolant Low-Flow RT Setpoint 0.04ReachedRods Begin to Drop 1.04Maximum Cladding Temperature Occurs 3.50Maximum RCS Pressure Occurs 4.08Locked Rotor/Shaft Break -Rotor on One Pump Locks or the Shaft 0.0Rods-in-DNB BreaksReactor Coolant Low-Flow RT Setpoint 0.04ReachedRods Begin to Drop 1.04Undamaged RCPs Lose Power and Begin to 1.05Coast Downt'1Minimum DNBR Occurs 3.20Note:1. The undamaged RCPs were modeled to trip coincident with rod motion, but slight differences, which are considered tobe negligible, occur because of the RETRAN code trip logic.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-211Table 2.4.2-2 Limiting Results -RCP Locked Rotor/Shaft BreakCriterion Analysis Value LimitPCT at Core Hot Spot (&deg;F) 1786.6 2700Maximum Zirconium-Water Reaction at Core Hot Spot (%) 0.29 16.0Maximum RCS Pressure (psia) 2675.1 2750Maximum Number of Rods-in-DNB
(%) 0.7 5WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-212WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 12with Of fsi te Powerwi thou t Of f s Ite Power1CD0000)C-)0.8-0.4-0.20510Time (seconds) 1520Figure 2.4.2-1RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseCore Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-213WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 13wI th Of fs te Powerw ithout Of f s i t e Power1.200CDiC)Lci10.8-0.4-0.2-0--0.4--0.6I I I fI I III I I I I I I I I I5010Time (seconds) 1520Figure 2.4.2-2RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseFaulted Loop Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-214WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 14w ith Of f s ite Powerw ithout Of f s ite Power27F~A -LUu2600ot0)EI0)2500 t2400 t2300-I I I I I I I I I22000510Time (seconds) 1520Figure 2.4.2-3RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseMaximum RCS Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-215WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-215wi th Of f s i te Powerwithout Of fsite Powe-1.2cici0ci0cicici0m0cicici0.60.4-T0.20I I I I II I I I I I I --I I i I I I I0510Time (seconds) 1520Figure 2.4.2-4RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseNuclear Power versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-216WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 16with Of fsi te Powerwi thout Of f si te Powee1.20)C)0.6-t0.4-0.2-I I I I II If I I I I I I I I I I IUL/0510Time (seconds) 1520Figure 2.4.2-5RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseCore Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-217WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-217with Of fsi te Powerwi thout Of s te Powe-'1f~f'd~
-,zUUU1800-1600t01)0.)E0)-0)0)1400-1200-000-t800-6000510T;me (seconds) 1520Figure 2.4.2-6RCP Locked Rotor/Shaft Break Overpressurization/PCT CasePCT versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-218WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 181.2C_CD00i0(DQ0--I -0.80.6tVI I0.4-0./010Time (seconds) 15)0Figure 2.4.2-7RCP Locked Rotor/Shaft Break Rods-in-DNB CaseCore Volumetric Flow Rate versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-219WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 19Fau I ted LoopIntac t Loops1.2CDC300Qi1-0.80.60.4-0.20-0.2040510Time (seconds) 1520Figure 2.4.2-8RCP Locked Rotor/Shaft Break Rods-in-DNB CaseLoop Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-220WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-22025002450-2400-(n)Cf)C)C/)23502300-2250-2200-2150I I I I I I I I I I I I I I051520Time (seconds)
Figure 2.4.2-9RCP Locked Rotor/Shaft Break Rods-in-DNB CasePressurizer Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-221WESTiNGHOUSE NON-PROPRIETARY CLASS 3 2-22 11.2C)ci0ci0(9ciG)00C)C)C)0.80.60.2-00510Time (seconds) 1520Figure 2.4.2-10RCP Locked Rotor/Shaft Break Rods-in-DNB CaseNuclear Power versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-222WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2221-) -VILZ0C_00.6-0.4-0.2-I II I 1 1UI010Time (seconds) 1520Figure 2.4.2-11RCP Locked Rotor/Shaft Break Rods-in-DNB CaseCore Average Heat Flux versus TimeWCAP- 17658-NP August 2013WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-223I') -I./_cici0ci0C)ciciwciciILcicicici-ci(-I0IL0.4-0.2 tI I I If I I I I I I I I I I11 -~U0510Time (seconds) 1520Figure 2.4.2-12RCP Locked Rotor/Shaft Break Rods-in-DNB CaseHot Channel Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2242.5 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 2.5.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical orLow Power Startup Condition (USAR Section 15.4.1)2.5.1.1 Technical Evaluation The specific acceptance criteria applied for this event are as follows:The DNBR should remain above the applicable 95/95 DNBR limits at all times during thetransient.
Demonstrating that the DNBR limits are met satisfies the requirements of GDC 10.Per GDC 20, the protection system should be designed to automatically initiate the operation ofappropriate
: systems, including the reactivity control systems, to ensure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and to senseaccident conditions and initiate the operation of safety-related systems and components.
For thisevent, protection is provided via the high neutron flux (low setting).
GDC 25 requires that the protection system is designed to ensure that specified acceptable fueldesign limits are not exceeded for any single malfunction of the reactivity control systems, suchas accidental withdrawal (not ejection or dropout) of control rods. Demonstrating that the fueldesign limits (that is, DNBR) are met satisfies the requirements of GDC 25.The following discussion demonstrates that all applicable acceptance criteria are met for this event for theWCGS.2.5.1.1.1 Introduction An uncontrolled RCCA withdrawal incident is defined as an uncontrolled addition of reactivity to thereactor core by withdrawal of RCCAs, resulting in a power excursion.
Although the probability of atransient of this type is extremely low, such a transient could be caused by a malfunction of the reactorcontrol rod drive system. This could occur with the reactor either subcritical or at power. The "at power"occurrence is discussed in Section 2.5.2. The uncontrolled RCCA withdrawal from a subcritical condition is classified as a Condition II event, a fault of moderate frequency, as defined by the ANS's "NuclearSafety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
During startup, when bringing the reactor from a shutdown condition to a low-power level, reactivity isadded at a prescribed and controlled rate by RCCA withdrawal or by reducing the core boronconcentration.
RCCA motion can cause much faster changes in reactivity than can result from changingboron concentration.
The rods are physically prevented from withdrawing in other than their respective banks. Power suppliedto the rod banks is controlled such that no more than two banks can be withdrawn at any time. The controlrod drive mechanism (CRDM) is of the magnetic latch type and the coil actuation is sequenced to providevariable speed rod travel. The maximum reactivity insertion rate is analyzed in the detailed plant analysisWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-225assuming the simultaneous withdrawal of the combination of the two rod banks with the maximumcombined worth at maximum speed.The neutron flux response to a continuous reactivity insertion is characterized by a very fast flux increaseterminated by the reactivity feedback effect of the negative Doppler coefficient.
This self-limitation of theinitial power increase results from a fast negative fuel temperature feedback (Doppler effect) and is ofprime importance during a startup transient because it limits the power to an acceptable level prior toprotection system action. After the initial power increase, the nuclear power is momentarily reduced andthen, if the incident is not terminated by a RT, the nuclear power increases again, but at a much slowerrate.Should a continuous RCCA withdrawal be initiated, the transient will be terminated by one of thefollowing automatic protective functions:
Source range neutron flux RT -actuated when either of two independent source range channelsindicates a flux level above a preselected, manually adjustable setpoint.
This trip function may bemanually bypassed only after an intermediate range neutron flux channel indicates a flux levelabove the source range cutoff power level. It is automatically reinstated when both intermediate channels indicate a flux level below the source range cutoff power level.Intermediate range neutron flux RT -actuated when either of two independent intermediate rangechannels indicates a flux level above a preselected, manually adjustable setpoint.
This tripfunction may be manually bypassed when two of the four power range channels are readingabove approximately 10 percent of full power and is automatically reinstated when three of thefour channels indicate a power level below this value.Power range neutron flux RT (low setting)
-actuated when two of the four power range channelsindicate a power level above approximately 25 percent of full power. This trip function may bemanually bypassed when two of the four power range channels indicate a power level aboveapproximately 10 percent of full power. This trip function is automatically reinstated when threeof the four channels indicate a power level below 10 percent power.Power range neutron flux RT (high setting)
-actuated when two out of the four power rangechannels indicate a power level above approximately 109 percent of full power. This trip functionis active in Modes 1 and 2, when the low setting is bypassed.
High nuclear flux rate RT -actuated when the positive rate of change of neutron flux on two outof four nuclear power range channels indicates a rate above the preset nominal setpoint ofapproximately 4.0 percent in 2 seconds.
This trip function is always active in Modes 1 and 2, andit is not explicitly modeled in the analysis of this event.In addition, control rod stops on high intermediate range flux level (one out of two) and high power rangeflux level (one out of four) serve to discontinue rod withdrawal and prevent the need to actuate theintermediate range flux level trip and the power range flux level trip, respectively.
This analysis creditsthe power range neutron flux trip (low setting) to initiate the RT.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2262.5.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe accident analysis uses the STDP methodology because the conditions resulting from the transient areoutside the range of applicability of the RTDP methodology.
To obtain conservative results for theanalysis of the uncontrolled RCCA bank withdrawal from subcritical event, the following inputparameters and initial conditions are modeled:The magnitude of the nuclear power peak reached during the initial part of the transient, for anygiven reactivity insertion rate, is strongly dependent on the Doppler-only power defect. Therefore, a conservatively low absolute value is used (1007 pcm) to maximize the nuclear power transient.
A most-positive MTC (+6 pcm/nF) is used because this yields the maximum rate of powerincrease.
The contribution of the moderator temperature coefficient is negligible during the initialpart of the transient because the heat transfer time constant between the fuel and moderator ismuch longer than the nuclear flux response time constant.
: However, after the initial neutron fluxpeak, the succeeding rate of power increase is affected by the moderator reactivity coefficient.
The analysis assumes the reactor to be at HZP conditions with a nominal no-load temperature of557&deg;F. This assumption is more conservative than that of a lower initial system temperature (thatis, shutdown conditions).
The higher initial system temperature yields a larger fuel-to-moderator heat transfer coefficient, a larger specific heat of the moderator and fuel, and a less-negative (smaller absolute magnitude)
Doppler defect. The less-negative Doppler defect reduces theDoppler feedback effect, thereby increasing the neutron flux peak. The high neutron flux peakcombined with a high fuel specific heat and larger heat transfer coefficient yields a larger peakheat flux.The analysis assumes the initial effective multiplication factor (Kff) to be 1.0 because itmaximizes the peak neutron flux and results in the most severe nuclear power transient.
RT is assumed on power range high neutron flux (low setting).
A conservative combination ofinstrumentation error, setpoint error, delay for trip signal actuation, and delay for control rodassembly release is modeled.
The analysis assumes a 10 percent uncertainty in the power rangeflux trip setpoint (low setting),
increasing it from the nominal value of 25 percent of full power to35 percent of full power. A delay time of 0.5 seconds is assumed for trip signal actuation andcontrol rod assembly release.
No credit is taken for the source range or intermediate rangeprotection.
During the transient, the increase in nuclear power is so rapid that the effect of errorsin the trip setpoint on the actual time at which the rods release is negligible.
In addition, the totalRT reactivity is based on the assumption that the highest worth RCCA is stuck in its fullywithdrawn position.
The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the two sequential control banks having the greatest combined worth at themaximum rod withdrawal speed. The assumed reactivity insertion rate is 75 pcm/sec, which isbased on a rod worth of 100 pcm/inch and a maximum rod speed of 72 steps per minute.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-227The DNB analysis assumes the most limiting axial and radial power shapes possible during thefuel cycle associated with having the two highest combined worth banks in their highest worthposition.
The analysis assumes the initial power level to be below the power level expected for anyshutdown condition (10-9 fraction of nominal power). The combination of highest reactivity insertion rate and low initial power produces the highest peak heat flux.The analysis assumes two of the four RCPs to be in operation.
This is conservative with respect tothe DNB transient.
This accident analysis uses the STDP methodology.
The use of the STDP stipulates that the RCSflow rates will be based on a fraction of the thermal design flow for two pumps operating.
Because the event is analyzed from HZP, the steady-state non-RTDP uncertainties are notconsidered in defining the initial conditions.
The uncontrolled RCCA bank withdrawal from subcritical event is considered an ANS Condition II event,a fault of moderate frequency, and is analyzed to show that the core and RCS are not adversely affectedby the event. This is demonstrated by showing that the DNB design basis is not violated and subsequently that there is little likelihood of core damage. It must also be shown that the peak hot spot fuel centerline temperature remains within the acceptable limit (48000F), although for this event, the heatup is relatively non-limiting.
2.5.1.1.3 Description of Analyses and Evaluations The analysis of the uncontrolled RCCA bank withdrawal from subcritical conditions is performed in threestages. First, a spatial neutron kinetics computer code, TWINKLE (Reference 1), is used to calculate thecore average nuclear power transient, including the various core feedback effects; that is, Doppler andmoderator reactivity.
Next, the FACTRAN computer code (Reference
: 2) uses the average nuclear powercalculated by TWINKLE and performs a fuel rod transient heat transfer calculation to determine the coreaverage heat flux and hot spot fuel temperature transients.
: Finally, the core average heat flux calculated by FACTRAN is used in the VIPRE computer code (Reference
: 3) for transient DNBR calculations.
2.5.1.1.4 ResultsThe analysis shows that all applicable acceptance criteria are met for the WCGS. The minimum DNBRnever decreases below the applicable limit values and the peak fuel centerline temperature is 2342'F. Thepeak temperatures are well below the minimum temperature at which fuel melting would be expected(4800&deg;F).
Figure 2.5. 1-1 shows the nuclear power transient, Figure 2.5.1-2 shows the core average heat fluxtransient, and Figures 2.5.1-3 and 2.5.1-4 show the fuel average and cladding surface temperature transients at the hot spot.The time sequence of events for both cases is presented in Table 2.5.1-1.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-228In the event of an RCCA withdrawal event from subcritical conditions, the core and the RCS are notadversely affected because the combination of thermal power and coolant temperature results in aminimum DNBR greater than the SAL value. Furthermore, because the maximum fuel temperatures predicted to occur during this event are much less than those required for fuel melting to occur, no fueldamage is predicted as a result of this transient.
Cladding damage is also precluded.
2.5.1.2 Conclusions Based on a review of the analysis of the uncontrolled RCCA withdrawal from a subcritical or low-power startup condition, it is concluded that the analysis adequately accounted for plant operation at the statedpower level and were performed using acceptable analytical models. It is further concluded that theanalysis demonstrates that the reactor protection and safety systems will continue to ensure that thespecified acceptable fuel design limits are not exceeded.
Based on this, it is concluded that the plant willcontinue to meet the requirements of GDCs 10, 20, and 25.2.5.1.3 References
: 1. WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non-Proprietary),
"TWINKLEMulti-dimensional Neutron Kinetics Computer Code," January 1975.A2. WCAP-7908-A, "FACTRAN
-A FORTRAN IV Code for Thermal Transients in a UO2 FuelRod," December 1989.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-229Table 2.5.1-1 Time Sequence of Events -Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition Event Time (seconds)
Initiation of Uncontrolled Rod Withdrawal 0.0Power Range High Neutron Flux Low Setpoint is Reached 10.43Peak Nuclear Power Occurs 10.57Rod Motion Begins 10.93Peak Heat Flux Occurs (0.3642) 12.73Minimum DNBR Occurs (1.66) 12.73Peak Average Cladding Temperature Occurs (683&deg;F) 13.06Peak Average Fuel Temperature Occurs (1934&deg;F) 13.26Peak Fuel Centerline Temperature Occurs (2342&deg;F) 13.71WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17659-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-230WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-230I.E00-0'4--00DO3-0.8"0.4-0.2'0 5 10 15 20 25Time (seconds) 30Figure 2.5.1-1 Rod Withdrawal from Subcritical
-Nuclear Power Transient WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-231WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-23 10F-0r0 5 10 15 20 25Time (seconds)
Figure 2.5.1-2 Rod Withdrawal from Subcritical
-Core Average Heat Flux Transient WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-232WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-232CLC,)EQ)F-a,C/_-0a-"U-0 5 10 15 20 25 30Time (seconds)
Figure 2.5.1-3 Rod Withdrawal from Subcritical
-Fuel Average Temperature Transient WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-233WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-233U-C,,a)a)L..a)a)I.-DL..a)0~Ea)a)C)U)00~U)00 5 10 15 20 25Time (seconds) 30Figure 2.5.1-4 Rod Withdrawal from Subcritical
-Cladding Surface Temperature Transient WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2342.5.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (USARSection 15.4.2)2.5.2.1 Technical Evaluation 2.5.2.1.1 Introduction An uncontrolled RCCA bank withdrawal at power that causes an increase in core heat flux can result fromfaulty operator action or a malfunction in the rod control system. Immediately following the initiation ofthe transient, the SG heat removal rate lags behind the core power generation rate until the SG pressurereaches the setpoint of the SG relief or safety valves. This imbalance between heat removal and heatgeneration rate causes the reactor coolant temperature to increase.
Unless terminated, the power mismatchand resultant coolant temperature increase could eventually result in a violation of the DNBR SAL, fuelcenterline melt, and/or RCS overpressurization.
Therefore, to avoid core damage, the reactor protection system is designed to automatically terminate any such transient before the DNBR falls below the limitvalue, or the fuel rod linear heat generation rate (kW/ft) limit is exceeded.
The reactor protection systemand PSVs are designed to preclude exceeding the RCS pressure boundary safety limit.The automatic features of the reactor protection system that prevent core damage and preclude RCSoverpressurization during an RCCA bank withdrawal incident at power include the following:
Power range neutron flux instrumentation actuates a RT if two-out-of-four channels exceed anoverpower setpoint.
RT actuates if any two-out-of-four channels exceed the power range neutron flux high positiverate setpoint.
RT actuates if any two-out-of-four OTAT channels exceed the corresponding setpoint.
Thissetpoint is automatically varied with axial power distribution, coolant average temperature, andpressure to help protect the DNB design basis.RT actuates if any two-out-of-four OPAT channels exceed the corresponding setpoint.
Thissetpoint is capable of being automatically varied with axial power imbalance to help ensure thatthe allowable heat generation rate (kW/ft) is not exceeded.
A high pressurizer pressure RT actuates if any two-out-of-four pressure channels exceed thecorresponding
: setpoint, which is set at a fixed point. This pressure setpoint is less than the setpressure for the PSVs.MSSVs can open for this event and provide additional steam flow.A high pressurizer water level RT actuates if any two-out-of-three channels exceed the tripsetpoint, which is set at a fixed value, when the reactor power is above approximately 10 percent(Permissive P-7).WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-235In addition to the above listed automatic
: features, there are the following RCCA withdrawal blocks:* Power range neutron flux (one-out-of-four power range)* OPAT (two-out-of-four)
* OTAT (two-out-of-four) 2.5.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaA number of cases were analyzed assuming a range of reactivity insertion rates for both minimum andmaximum reactivity feedback conditions at various power levels for DNB and RCS overpressure considerations.
The cases presented below are representative for this event.The following assumptions were made for the analysis of the uncontrolled RCCA bank withdrawal atpower transient in order to obtain conservative results with respect to core damage:This transient was analyzed with the RTDP (Reference 1). Initial RCS pressure and temperature were assumed to be at their nominal values. An initial NSSS power of 3651 MWt, which includesall applicable uncertainties, was modeled.
MMF was also modeled.
Uncertainties in initialconditions, with the exception of power, were included in the DNBR SAL as described in theRTDP.* For reactivity coefficients, two sets were analyzed.
-Minimum reactivity feedback:
A least negative or positive value of the MTC of reactivity is assumed corresponding to the beginning of core life. A conservatively small (in absolutemagnitude) value of the Doppler coefficient is assumed.-Maximum reactivity feedback:
A conservatively large positive moderator densitycoefficient and a large (in absolute magnitude) negative Doppler coefficient are assumed.* The RT on power range neutron flux (high setpoint) was assumed to be actuated at the SAL of116.5 percent of the analyzed full power level.* The OTAT and OPAT trips included all adverse instrumentation and setpoint errors, and thedelays for the trip signal actuation were assumed at their maximum values.0 The RCCA trip insertion characteristic was based on the assumption that the highest-worth RCCAwas stuck in its fully-withdrawn position.
* A range of reactivity insertion rates was examined.
The maximum positive reactivity insertion rate was greater than that which would be obtained from the simultaneous withdrawal of the twocontrol rod banks having the maximum combined worth at a conservative speed(48.125 inches/minute, which corresponds to 77 steps/minute).
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-236To be conservative with respect to DNB, the pressurizer sprays and relief valves were assumedoperational because they limit the reactor coolant pressure increase.
Power levels of 10, 60, and 100 percent of the assumed NSSS power were considered.
For the RCS overpressure
: analysis, the preceding assumptions still apply with the following differences inorder to obtain conservative results with respect to RCS overpressurization:
This case is analyzed with the STDP. Initial RCS pressure and temperature were assumed to bewithin their respective allowable operating ranges with uncertainties applied in the conservative directions.
As was done in the DNB case, an NSSS power of 3651 MWt, which includes allapplicable uncertainties, was modeled.
TDF was also modeled.Minimum reactivity feedback conditions were modeled.The pressurizer sprays were not modeled because operation of the pressurizer spray valves wouldminimize the pressure increase during the transient.
Ranges of initial power levels (from 8 to 100 percent of the analyzed power level) and reactivity insertion rates (1 to 110 pcm/sec) were analyzed.
The RT on power range neutron flux high positive rate trip was assumed to be actuated at aconservative rate setpoint of 6.9 percent of the analyzed power level, with a conservative ratetime constant and delay time.The PSVs were modeled with a positive set pressure tolerance (2 percent) applied toconservatively increase the opening pressure.
In addition, this case includes a 1 percent setpointshift and a 1.153-second purge time delay to account for the existence of PSV water-filled loopseals, as described in Reference
: 2. The pressurizer PORVs were not modeled because theiroperation would minimize the pressure increase during the transient.
The MSSVs were modeled with bounding opening pressures in order to prolong the mismatchbetween core heat generation and secondary heat removal capability.
Based on its frequency of occurrence, the uncontrolled RCCA bank withdrawal at power transient isconsidered to be a Condition II event as defined by the ANS's "Nuclear Safety Criteria for the Design ofStationary Pressurized water Reactor Plants,"
ANSI N 18.2-1973.
The following items summarize theacceptance criteria associated with the analysis of this event:Pressures in the RCS and MSS must remain less than 110 percent of the respective designpressures.
With respect to peak MSS pressures, the RCCA withdrawal at power event is bounded by theLOL/TT event described in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip,Inadvertent Closure of Main Steam Isolation Valves, Loss of Condenser Vacuum and OtherEvents Resulting in Turbine Trip," in which assumptions are made to conservatively calculate theWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-237MSS pressure transients.
For the LOL/TT event, the turbine trip is the initiating incident thatmaximizes the resultant power mismatch between the primary and secondary sides, and theresultant temperature and pressure transients of the MSS are always more severe for LOL/TTevents than for RCCA withdrawal at power events. Based on this, no explicit calculation ofmaximum MSS pressure is performed for this event.Fuel cladding integrity must be maintained by ensuring that the DNBR remains above the95/95 DNBR limit. In addition, it has been historical practice to assume that fuel failure willoccur if centerline melting takes place. Therefore, the analysis evaluates whether the peak linearheat generation rate exceeds the value that would cause fuel centerline melt. For the WCGS, thisis met by ensuring that the peak core average heat flux does not exceed 121 percent of theanalyzed full power level.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the uncontrolled RCCA bank withdrawal at power acceptance criteriaare provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the RCCA bank withdrawal at power event, this is shown to be met bydemonstrating that the fuel damage criterion is satisfied.
GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.
For theRCCA bank withdrawal at power event, this is shown to be met by demonstrating that the peakRCS pressure is less than 110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.
For the RCCA bank withdrawal at power event, which results in a RT, this is shown to be met by demonstrating that the fueldamage criterion is satisfied.
The protection features presented in subsection 2.5.2.1.1 provide mitigation of the uncontrolled RCCAbank withdrawal at power transient such that the above criteria are satisfied.
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-2382.5.2.1.3 Description of Analyses and Evaluations The purpose of this analysis was to demonstrate the manner in which the protection functions described above actuate for various combinations of reactivity insertion rates and initial conditions.
Insertion rateand initial conditions determined which trip function actuated first.The uncontrolled RCCA bank withdrawal at power event was analyzed with the RETRAN computer code(Reference
: 3) to demonstrate the manner in which the previously described protection functions provideadequate protection from core damage. The RETRAN model simulates the core neutron kinetics, RCS,.pressurizer, pressurizer relief and safety valves, pressurizer pressure control systems, SGs, and MSSVs.The code computes pertinent plant variables, including temperatures, pressures, power level, and coreboron concentration.
For the most limiting case analyzed, a detailed DNBR evaluation using the detailedT/H digital computer code, VIPRE (Reference 4), was used to determine if the DNB design basis wasmet.An analysis to confirm that the RCS pressure safety limit is protected was performed using theLOFTRAN code (Reference 5). Similar to the RETRAN model, the LOFTRAN model simulates the coreneutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer pressure controlsystems, SGs, and MSSVs.2.5.2.1.4 ResultsFigures 2.5.2-1 through 2.5.2-3 show the transient response for a rapid uncontrolled RCCA bankwithdrawal incident (110 pcm/sec) starting from 100 percent power with minimum reactivity feedback.
The neutron flux level in the core rises rapidly while the core heat flux and coolant system temperature lag behind due to the thermal capacity of the fuel and coolant system fluid. RT on power range neutronflux (high setpoint) occurs shortly after the start of the transient prior to significant increases in the heatflux and water temperature, and the resultant DNBR remains well above the SAL value throughout thetransient.
The transient response for a slow uncontrolled RCCA bank withdrawal (I pcm/sec) from 100 percentpower with minimum feedback is shown in Figures 2.5.2-4 through 2.5.2-6.
With a lower insertion rate,the power increase rate is slower, the rate of increase of the average coolant temperature is slower, and thesystem lags and delays become less significant.
ART on OTAT occurs after a longer period of time thanfor a rapid RCCA bank withdrawal.
Again, the DNBR remains greater than the SAL value throughout thetransient.
Figure 2.5.2-7 shows the minimum DNBR as a function of reactivity insertion rate from 100 percentpower for both minimum and maximum reactivity feedback conditions.
The high neutron flux and OTATRT functions provide DNB protection over the analyzed range of reactivity insertion rates because theminimum DNBR is never less than the SAL value.Figures 2.5.2-8 and 2.5.2-9 show the minimum DNBR as a function of reactivity insertion rate for RCCAbank withdrawal incidents starting at 60 and 10 percent power, respectively.
The results are similar to the100-percent power case. However, as the initial power level is decreased, the range over which the OTATtrip is effective is increased.
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-239Note that the conservative minimum DNBR approximation calculated by RETRAN for a number of the60 and 10 percent power cases did not meet the SAL DNBR value. This limit was conservatively definedto demonstrate that the DNB design basis is satisfied for analyses performed using RTDP methods.Sufficient margin is maintained between the SAL DNBR and the design limit DNBR to offset the effectsof rod bow, lower plenum flow anomaly, and plant instrumentation biases, as well as to provide flexibility in the design and operation of the plant. See Section 2.12, "Thermal and Hydraulic Design,"
for additional information.
To demonstrate that the DNB design basis was satisfied, the VIPRE code was used toperform a detailed DNBR calculation of the most limiting part-power case (reactivity insertion rate of13 pcm/sec from 10 percent power with minimum feedback).
The results confirmed that the DNB designbasis continues to be met and that sufficient DNBR margin is retained to allow for flexibility in the designand operation of the plant. To increase the amount of DNBR margin retained, the DNBR calculations performed for this event credited the following changes:A higher MMF of 376,000 gpmThimble plugs were assumed to remain installed to reduce the core bypass flow (all othernon-LOCA analyses covered the bounding scenario of having the core TPR)Finally, Figures 2.5.2-10 through 2.5.2-12 show the transient responses for the limiting RCS pressure caseindicating that the applicable limit is met.The calculated sequences of events for four cases are shown in Table 2.5.2-1; the four cases include:0 a rapid RCCA bank withdrawal (110 pcm/sec) from 100 percent power with minimum feedback,
* a slow RCCA bank withdrawal (1 pcm/sec) from 100 percent power with minimum feedback,
* the limiting DNB case (withdrawal rate of 13 pcm/sec from 10 percent power with minimumfeedback),
* the limiting overpressure case (withdrawal rate of 21 pcm/sec from 74 percent power withminimum feedback).
With the reactor tripped, the plant eventually returns to a stable condition.
The plant could subsequently be cooled down further by following normal plant shutdown procedures.
The limiting results of theuncontrolled RCCA bank withdrawal at power analyses are shown in Table 2.5.2-2.For the DNB cases, the power range neutron flux and OTAT RT functions provided adequate protection over the entire range of possible reactivity insertion rates. The results show that the DNB design basis ismet and the peak linear heat generation rate is less than the limit.For the RCS overpressure cases, the power range neutron flux, OTAT, power range neutron flux highpositive rate, and high pressurizer pressure RT functions, in conjunction with the PSVs and MSSVs,provide adequate protection over the entire range of possible reactivity insertion rates. The results showedthat the peak RCS pressure remains below 110 percent of the design pressure.
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-240Therefore, the results of the analysis show that an uncontrolled RCCA bank withdrawal at power does notadversely affect the core, the RCS, or the MSS.2.5.2.2 Conclusions Based on the above information, it is concluded that the analysis has adequately accounted for operation of the plant at the analyzed power level and was performed using acceptable analytical models. Thisanalysis has also demonstrated that the reactor protection and safety systems will continue to ensure thatthe specified acceptable fuel design limits and RCS pressure safety limit are not exceeded.
Based on this,it can be concluded that the plant will continue to meet the requirements of GDCs 10, 15, and 26.2.5.2.3 References
: 1. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"
April 1989.2. WCAP-12910, Rev. I-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.3. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.4. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.5. WCAP-7907-P-A, "LOFTRAN Code Description,"
April 1984.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-241Table 2.5.2-1 Time Sequence of Events -Uncontrolled RCCA Bank Withdrawal at PowerCase Event Time (sec)DNB Case Initiation of Uncontrolled RCCA Bank Withdrawal 0.0100 Percent Power, Power Range Neutron Flux -High Setpoint 1.25Minimum Feedback, Rapid ReachedRCCA Bank Withdrawal (110 pcm/sec)
Rods Begin to Drop 1.75Minimum DNBR Occurs 3.05DNB Case Initiation of Uncontrolled RCCA Bank Withdrawal 0.0100 Percent Power, OTAT Setpoint Reached 99.3Minimum Feedback, Slow RCCA Bank Rods Begin to Drop 101.6Withdrawal (1 pcm/sec)
Minimum DNBR Occurs 102.0Limiting DNB Case Initiation of Uncontrolled RCCA Bank Withdrawal 0.010 Percent Power, Minimum Power Range Neutron Flux -High Setpoint 55.18Feedback, Slow RCCA Bank ReachedWithdrawal (13 pcm/sec)Rods Begin to Drop 55.68Minimum DNBR Occurs 56.42Limiting Overpressure Case Initiation of Uncontrolled RCCA Bank Withdrawal 0.074 Percent Power, Minimum Power Range Neutron Flux High Positive Rate 12.54Feedback, Rapid RCCA Setpoint ReachedBank Withdrawal (21 pcm/sec)
Rods Begin to Drop 13.54Maximum RCS Pressure Occurs 15.50WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-242WESTIINGHOU SE NON-PROPRIETARY CLASS 3 2-242Table 2.5.2-2 Uncontrolled RCCA Bank Withdrawal at Power -Limiting ResultsLimiting value Analysis Limit CaseMinimum DNBR See notetl) See note1.) 10% power, minimum feedback, 13 pcm/sec reactivity insertion ratePeak Core Heat Flux (fraction of 1.183 1.21 10% power, minimum feedback, analyzed full power) 100 pcm/secPeak Primary System Pressure 2707.4 2750.0 74% power, minimum feedback(psia) 21 pcm/secNote:1. A detailed DNBR evaluation was performed using the VIPRE code confirming that the DNB design basis was satisfied for the limiting case.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-243000C-,0U-w00~0a,C-,00C-,0Lj~U-0a,U,C-,0 2 3 4 5 6Time (s)0 1 2 3 4 5 6 7Time (s)Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -110 pcm/secNuclear Power and Core Heat Flux Versus TimeFigure 2.5.2-1WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-244U)V)L.Dt-4co(n4Time (s)(DE-50Q)L4LU)U)a)0nI ,UL/1150-1100-1050-1000-950-..............................................................................................
....................I ...............................................
oIv'-t0 1 2 3 4Time (s)5 6 7Figure 2.5.2-2Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -110 pcm/secPressurizer Pressure and Water Volume Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-245WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-245U.)En0 1 2 3 4 5 6 7Time (s)0 2 3 4Time (s)5 6 7Figure 2.5.2-3Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power- 110 pcm/secVessel Average Temperature and DNBR Versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2460c-0-0C-,.0.--(V0C~_)cvCa-,C..)0 20 40 60Time (s)80 100 1200 20 40 60Time (s)80 100 120Figure 2.5.2-4Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -1 pcm/secNuclear Power and Core Heat Flux Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-247U3Enco1PcnV)0 20 40 60Time (s)80 100 1200 20 40 60 80 100 120Time (s)Figure 2.5.2-5Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -1 pcm/secPressurizer Pressure and Water Volume Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-248WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-248E-0Cnco0 20 40 60Time (s)80 100 1200 20 40 60Time (s)80 100 120Figure 2.5.2-6Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -1 pcm/secVessel Average Temperature and DNBR Versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-249WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-249Maximum reactivity
---- ---- Minimum r e a c t i v i t y1.951.90.S1.85--E .._ 1.80- /E..1 .. ..........feedbackfeedback0 20 40 60 80 100Reactivity Insertion Rate (pcm/sec) 120Figure 2.5.2-7Bank Withdrawal at Power -100 Percent PowerMinimum DNBR Versus Reactivity Insertion RateNote that the minimum DNBR values presented were calculated using the RETRAN code and arerepresentative of the trends for minimum DNBR and not indicative of final calculated DNBR values. Thedetailed T/H code VIPRE was used to confirm that the limiting case (10 percent power, minimumreactivity
: feedback, 13 pcm/sec reactivity insertion rate) satisfied the DNB safety design basis.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-250Maximum reactivi ty----- --M in im um r e a c t i v it y3.002.80.2.60m 2.40E 2.20-E2.00 ...feedbackfeedback0 20 40 60 80 100Reactivity Insertion Rate (pcm/sec) 120Figure 2.5.2-8Bank Withdrawal at Power -60 Percent PowerMinimum DNBR Versus Reactivity Insertion RateNote that the minimum DNBR values presented were calculated using the RETRAN code and arerepresentative of the trends for minimum DNBR and not indicative of final calculated DNBR values. Thedetailed T/H code VIPRE was used to confirm that. the limiting case (10 percent power, minimumreactivity
: feedback, 13 pcm/sec reactivity insertion rate) satisfied the DNB design basis.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-251WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-25 1Maximum reactivity
----------
M inimum r e a c t i v it y4.50-4.00-3 .5 0 "3.00E ..............feedbackfeedback0 20 40 60 80 100Reactivity Insertion Rate (pcm/sec) 120Figure 2.5.2-9Bank Withdrawal at Power -10 Percent PowerMinimum DNBR Versus Reactivity Insertion RateNote that the minimum DNBR values presented were calculated using the RETRAN code and arerepresentative of the trends for minimum DNBR and not indicative of final calculated DNBR values. Thedetailed T/H code VIPRE was used to confirm that the limiting case (10 percent power, minimumreactivity
: feedback, 13 pcm/sec reactivity insertion rate) satisfied the DNB design basis.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-252.)0U-00C-)U-a.)C.U'-v (C--t0 5 10 15Time (s)200 5 10 15 20Time (s)Figure 2.5.2-10Bank Withdrawal at Power -Minimum Reactivity FeedbackLimiting Overpressure CaseNuclear Power and Core Heat Flux Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-253WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-253r'4Cl)coQ)Cl)C,,ni0 5 10 15 20Time (s)0 5 10 15 20Time (s)Figure 2.5.2-11Bank Withdrawal at Power -Minimum Reactivity FeedbackLimiting Overpressure CasePressurizer Pressure and Water Volume Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-254WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-254EZ30.0~IL~5 10 15Time (s)200 10 15Time (s)20Figure 2.5.2-12Bank Withdrawal at Power -Minimum Reactivity FeedbackLimiting Overpressure CaseVessel Average Temperature and Peak RCS PressureWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2552.5.3 Control Rod Misoperation (USAR Section 15.4.3)2.5.3.1 Technical Evaluation 2.5.3.1.1 Introduction The RCCA misalignment events include the following:
* One or more dropped RCCAs from the same group0 A dropped RCCA bank* A statically misaligned RCCA* Withdrawal of a single RCCAEach RCCA has a position indicator channel that displays the position of the assembly.
The displays ofassembly positions are grouped for the operator's convenience.
Fully inserted assemblies are furtherindicated by a rod at bottom signal, which actuates an alarm and a control room annunciator.
Groupdemand position is also indicated.
Full-length RCCAs are moved in preselected banks, and the banks are moved in the same preselected sequence.
Each control bank of RCCAs is divided into two groups. The rods comprising a group operatein parallel through multiplexing thyristors.
The two groups in a bank move sequentially such that the firstgroup is always within one step of the second group in the bank. A definite schedule of actuation (ordeactuation of the stationary
: gripper, movable gripper, and lift coils of a mechanism) is required towithdraw the RCCA attached to the mechanism.
Because the stationary
: gripper, movable gripper, and liftcoils associated with the four RCCAs of a rod group are driven in parallel, any single failure that wouldcause rod withdrawal would affect a minimum of one group. Mechanical failures are in the direction ofinsertion, or immobility.
A dropped RCCA or RCCA bank is detected by one or more of the following:
Sudden drop in the core power level as seen by the nuclear instrumentation systemAsymmetric power distribution as seen on out-of-core neutron detectors or core exitthermocouples Rod-at-bottom signalRod deviation alarmRod position indication Dropping of a full-length RCCA is assumed to be initiated by a single electrical or mechanical failure thatcauses any number and combination of rods from the same group of a given control bank to drop to thebottom of the core. The resulting negative reactivity insertion causes nuclear power to rapidly decrease.
An increase in the hot channel factor can occur due to the skewed power distribution representative of aWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-256dropped rod configuration.
For this event, it must be shown that the DNB design basis is met for thecombination of power, hot channel factor, and other system conditions that exist following a dropped rod.Misaligned assemblies are detected by:Asymmetric power distribution as seen on out-of-core neutron detectors or core exitthermocouples Rod deviation alarm* Rod position indicators For the WCGS, rod position is displayed in 6-step increments with an accuracy of +/-4 steps. Deviation ofany RCCA from its group by twice this distance (12 steps) will not cause power distributions worse thanthe design limits. The deviation alarm alerts the operator to rod deviation with respect to the groupposition in excess of 12 steps. If the rod deviation alarm is not functional, the operator is required to takeaction per the Technical Requirements Manual (or equivalent document).
If one or more rod position indicators should be out of service, detailed operating instructions shall befollowed to assure the alignment of the non-indicated RCCAs. The operator is also required to take actionper the TS.In the extremely unlikely event of simultaneous electrical failures that could result in single RCCAwithdrawal, rod deviation and rod control urgent failure would both be displayed on the plant annunciator, and the rod position indicators would indicate the relative positions of the assemblies in the bank. Theurgent failure alarm also inhibits automatic rod motion in the group in which it occurs. Withdrawal of asingle RCCA by operator action, whether deliberate or by a combination of errors, would result inactivation of the same alarm and the same visual indications.
Withdrawal of a single RCCA results in bothpositive reactivity insertion tending to increase core power, and an increase in local power density in thecore area associated with the RCCA. Automatic protection for this event is provided by the OTAT RT, butbecause the local power density increases, it is not possible in all cases to provide assurance that the coresafety limits will not be violated.
2.5.3.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe dropped RCCA, dropped RCCA bank, and statically misaligned RCCA events are classified asCondition II events (faults of moderate frequency) as defined by the ANS's "Nuclear Safety Criteria forthe Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
The single RCCAwithdrawal incident is classified as an ANS Condition III event, as discussed below.The acceptance criteria applied to this event are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the dropped RCCA, dropped RCCA bank, and statically misaligned RCCA events acceptance criteria are provided as follows.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-257GDC 10, insofar as it requires that the reactor core be designed with appropriate margin to assurethat specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition ofnormal operation, including the effects of anticipated operational occurrences (AOOs)GDC 20, insofar as it requires that the protection system be designed to initiate the reactivity control systems automatically to assure that acceptable fuel design limits are not exceeded as aresult of AOOs and to initiate automatically operation of systems and components important tosafety under accident conditions GDC 25, insofar as it requires that the protection system be designed to assure that SAFDLs arenot exceeded for any single malfunction of the reactivity control systems.The following discussion demonstrates that all applicable acceptance criteria are met for this event at theWCGS.No single electrical or mechanical failure in the rod control system could cause the accidental withdrawal of a single RCCA from the inserted bank at full-power operation.
The operator could deliberately withdraw a single RCCA in the control bank since this feature is necessary in order to retrieve anassembly should one be accidentally dropped.
The event analyzed could only occur from multiple wiringfailures or multiple deliberate operator actions and subsequent and repeated operator disregard of eventindication.
The probability of such a combination of conditions is so low that it would be acceptable forthe consequences to include slight fuel damage. Thus, consistent with the philosophy and format ofANSI N 18.2, the event is classified as a Condition Ill event. By definition "Condition II occurrences include incidents, any one of which may occur during the lifetime of a particular plant," and "shall notcause more than a small fraction of fuel elements in the reactor to be damaged."
See Tables 2.5.3-1 and 2.5.3-2 for detailed acceptance criteria and initial conditions used in the droppedRCCA/dropped RCCA bank analysis.
For the statically misaligned RCCA and single RCCA withdrawal events, see the analysis descriptions and results in Sections 2.5.3.1.3 and 2.5.3.1.4 for details of the inputsand acceptance criteria.
2.5.3.1.3 Description of Analyses and Evaluations One or More Dropped RCCAs from the Same GroupThe LOFTRAN computer code (Reference I) calculates transient system responses for the evaluation of adropped RCCA event. The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief andsafety valves, pressurizer spray, SG, and MSSVs. The code computes pertinent plant variables including temperatures, pressures, and power levels.Transient RCS statepoints (temperature,
: pressure, and power) are calculated by LOFTRAN.
Nuclearmodels are used to obtain a hot-channel factor consistent with the primary-system conditions and reactorpower. By incorporating the primary conditions from the transient analysis and the hot-channel factorfrom the nuclear analysis, it is shown that the DNB design basis is met using dropped rod limit linesdeveloped with the VIPRE code (Reference 2). The transient response
: analysis, nuclear peaking factorWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-258analysis, and performance of the DNB design basis confirmation are performed in accordance with theapproved methodology described in Reference 3.Dropped RCCA BankA dropped RCCA bank results in a symmetric power change in the core. Assumptions made in themethodology (Reference
: 3) for the dropped RCCA(s) analysis provide a bounding analysis for thedropped RCCA bank.Statically Misaligned RCCASteady-state power distributions are analyzed using the appropriate nuclear physics computer codes. Thepeaking factors are then compared to peaking factor limits developed using the VIPRE code, which arebased on meeting the DNBR design criterion.
The following cases are examined in the analysis assumingthe reactor is at full power: the worst rod withdrawn with bank D inserted at the insertion limit, the worstrod 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 with maximum predicted peaking factors.This assures a conservative FAH for the misaligned RCCA configuration.
Single RCCA Withdrawal Power distributions within the core are calculated.
The peaking factors are then used by VIPRE tocalculate the DNBR for the event. The case of the worst rod withdrawn from bank D inserted at theinsertion limit, with the reactor initially at full power, was analyzed.
This incident is assumed to occur atBOL because this condition results in a minimum MTC. This assumption maximizes the power increaseand minimizes the tendency of increased moderator temperature to flatten the power distribution.
2.5.3.1.4 Control Rod Misalignment ResultsOne or More Dropped RCCAsSingle or multiple dropped RCCAs within the same group result in a negative reactivity insertion.
Thecore is not adversely affected during this period because power is decreasing rapidly.
Either reactivity feedback or control bank withdrawal will re-establish power.For a dropped RCCA event in the automatic rod control mode, the rod control system detects the drop inpower and initiates control bank withdrawal.
Power overshoot may occur due to rod control movement, after which the control system will insert the control bank to restore nominal power. In all cases, theminimum DNBR remains above the limit value.Following a dropped rod event in manual rod control, the plant will establish a new equilibrium condition.
The equilibrium process without control system interaction is monotonic, thus removing powerovershoot as a concern, and establishing the automatic rod control mode of operation as the limiting case.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-259Dropped RCCA BankA dropped RCCA bank results in a large negative reactivity insertion.
Due to the relatively large worth ofthe dropped bank, and if the turbine load is constant, a RT may occur on low pressurizer pressure due tothe mismatch between the reactor power and the turbine power. The core is not adversely affected duringthis period because power is decreasing rapidly.
In the event a RT does not occur, the initial powerreduction from a dropped RCCA bank is large and the power return due to reactivity feedback and controlbank withdrawal is far less than seen from one or more dropped RCCAs from the same group. In eitherinstance, the minimum DNBR remains above the limit value.Following plant stabilization, the operator may manually retrieve the RCCA(s) by following applicable plant procedures.
Statically Misaligned RCCAThe most severe misalignment situations with respect to DNBR at significant power levels arise fromcases in which one RCCA is fully inserted, or where bank D is fully inserted with one RCCA fullywithdrawn.
Multiple independent alarms, including a bank insertion limit alarm, alert the operator wellbefore the postulated conditions are approached.
The bank can be inserted to its insertion limit with anyone assembly fully withdrawn without the DNBR decreasing below the limit value.The insertion limits in the TS may vary depending on a number of limiting criteria.
It is preferable, therefore, to analyze the misaligned RCCA case at full power for a control bank insertion position that isas deep as allowed by the DNBR and power peaking factor limits. The full power insertion limits oncontrol bank D must then be chosen to be above that position and will usually be dictated by othercriteria.
Detailed results will vary from cycle to cycle depending on fuel arrangements.
For this RCCA misalignment, with bank D inserted to its full-power insertion limit and one RCCA fullywithdrawn, the DNBR does not decrease below the limit value. This case is analyzed assuming the initialreactor power and RCS pressure and temperature are at their nominal values including uncertainties, butwith the increased radial peaking factor associated with the misaligned RCCA.DNB calculations have not been performed specifically for RCCAs missing from other banks. However,power shape calculations have been done as required for the RCCA ejection analysis.
Inspection of thepower shapes shows that the DNB and peak kW/ft situation is less severe than the bank D case discussed above, assuming insertion limits on the other banks are equivalent to a bank D full-in insertion limit.For RCCA misalignments with one RCCA fully inserted, the DNBR does not decrease below the limitvalue. This case is analyzed assuming the initial reactor power and RCS pressure and temperature are attheir nominal values including uncertainties, but with the increased radial peaking factor associated withthe misaligned RCCA.DNB does not occur for the RCCA misalignment
: incident, and thus the ability of the primary coolant toremove heat from the fuel rod is not reduced.
The peak fuel temperature corresponds to a linear heatgeneration rate based on the radial peaking factor penalty associated with the misaligned RCCA and theWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-260design axial power distribution.
The resulting linear heat generation is well below that which would causefuel melting.Following the identification of an RCCA group misalignment condition, the operator is required to takeaction per the plant TS and applicable plant procedures.
Single RCCA Withdrawal For the single rod withdrawal event, two cases have been considered as follows:1. If the reactor is in the manual control mode, continuous withdrawal of a single RCCA results inboth an increase in core power and coolant temperature, and an increase in the local hot channelfactor in the area of the withdrawing RCCA. In terms of the overall system response, this case issimilar to those presented for the uncontrolled RCCA bank withdrawal at power event. However,the increased local power peaking in the area of the withdrawn RCCA results in lower minimumDNBRs than for the withdrawn bank cases. Depending on initial bank insertion and location ofthe withdrawn RCCA, automatic RT may not occur sufficiently fast enough to prevent theminimum DNBR from decreasing below the limit value. Evaluation of this case at the power andcoolant conditions at which the OTAT trip would be expected to trip the plant shows that anupper limit for the number of fuel rods with a DNBR less than the limit value is 5 percent of thetotal rods in the core.2. If the reactor is in the automatic control mode, the multiple failures that result in the withdrawal of a single RCCA will result in the immobility of the other RCCAs in the controlling bank. Thetransient will then proceed in the same manner as Case (1) described above.For such cases as above, a RT will ultimately ensue, although not sufficiently fast enough in all cases toprevent a minimum DNBR in the core of less than the limit value. Following RT, normal shutdownprocedures are followed.
No single failure of the RT system will negate the protection functions requiredfor the single RCCA withdrawal
: accident, or adversely affect the consequences of the accident.
2.5.3.1.5 ResultsThe evaluation of the dropped rod event using the methodology in Reference 3, encompassing all possibledropped RCCA or RCCA bank worths delineated in Reference 3, concluded that the minimum DNBRremains above the SAL value for the WCGS. For all cases of any single RCCA fully inserted, or bank Dinserted to the rod insertion limit and any single RCCA in that bank fully withdrawn (staticmisalignment),
the minimum DNBR remains above the limit value for the WCGS. Therefore, the DNBdesign criterion is met and the RCCA misalignments do not result in core damage. For the case of theaccidental withdrawal of a single RCCA, with the reactor in the automatic or manual control mode andinitially operating at full power with bank D at the insertion limit, an upper bound of the number of fuelrods experiencing DNB is 5 percent of the total number of fuel rods in the core for the WCGS.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2612.5.3.2 Conclusion The analyses of control rod misalignment events have been reviewed and it has been concluded that thesewere performed using acceptable analytical models. It was further concluded that the analyses havedemonstrated that the reactor protection and safety systems will continue to ensure that the SAFDL's willnot be exceeded during normal or anticipated operational transients.
Based on this, it is concluded that theplant will continue to meet the requirements of GDCs 10, 20, and 25.2.5.3.3 References
: 1. WCAP-7907-P-A, "LOFTRAN Code Description,"
April 1984.2. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.3. WCAP-11394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-262Table 2.5.3-1 Non-LOCA Analysis Limits and Analysis Results for the Dropped Rod EventAnalysis ResultResult Parameter Analysis Limit Limiting CaseMinimum DNBR (RTDP, WRB-2) 1.52 > 1.52Peak Linear Heat Generation (kW/ft) 22.4(" < 22.4Peak Uniform Cladding Strain (%) 1.0 < 1.0Note:1. Corresponds to a conservative UO2 fuel melting temperature of 4700'F.Table 2.5.3-2 Summary of Initial Conditions and Computer Codes Used for the Dropped Rod EventVessel Vessel Average RCSComputer DNB Initial Power Coolant Flow Coolant Temp PressureCodes Used Correlation RTDP (%) (gpm) (OF) (psia)LOFTRAN WRB-2 Yes 100 371,000 588.4 2250.0ANC (3637 MWt -VIPRE core power)WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2632.5.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (USARSection 15.4.4)2.5.4.1 Technical Evaluation As described in USAR Section 15.4.4, the Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature event has historically been analyzed for the WCGS. If the plant is operating with one pumpout of service, there is reverse flow through the inactive loop due to the pressure difference across thereactor vessel. The CL temperature in an inactive loop is identical to the CL temperature of the activeloops (the reactor core inlet temperature).
If the reactor is operated at power, and assuming the secondary side of the SG in the inactive loop is not isolated, there is a temperature drop across the SG in the inactiveloop and, with the reverse flow, the HL temperature of the inactive loop is lower than the reactor coreinlet temperature.
Starting of an idle RCP without bringing the inactive loop HL temperature close to thecore inlet temperature would result in the injection of cold water into the core, which would cause areactivity insertion and subsequent power increase.
Because the WCGS TS (LCO 3.4.4) require all four RCS loops to be in operation while at power or instartup conditions (Modes 1 and 2), the Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature event is administratively precluded for the WCGS. Therefore, no explicit analysis of thisevent is required, and no further evaluation is necessary.
2.5.4.2 Conclusions Based on the above information, it is concluded that the Startup of an Inactive Reactor Coolant Pump atan Incorrect Temperature event is administratively precluded by the WCGS TS and no analysis of theevent is required.
2.5.5 Chemical and Volume Control System Malfunction Resulting in a Decrease in BoronConcentration in the Reactor Coolant (USAR Section 15.4.6)2.5.5.1 Technical Evaluation The specific acceptance criterion applied for the CVCS malfunction (also referred to as boron dilution) events is that adequate operator action time is available prior to a complete loss of shutdown margin. Forboron dilution events in Modes 1 through 5, there must be at least 15 minutes from operator notification (that is, first alarm) until shutdown margin is lost. For the WCGS, a boron dilution event cannot occurduring Mode 6 (Refueling) due to administrative controls that isolate the RCS from the potential sourcesof unborated water. Additionally, for conditions when no RCP is in operation, all dilution sources areisolated or under administrative control.
Hence, a boron dilution event cannot occur during Mode 5 (ColdShutdown) or Mode 4 (Hot Shutdown) once operation on the residual heat removal system (RHRS)begins. This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS (Reference 1). With shutdown margin maintained, there is no return to critical andno violation of the 95/95 DNBR limit (GDC 10), as well as no violation of the primary and secondary pressures limits (GDC 15). Furthermore, because a return to critical is precluded and fuel design limits arenot exceeded, the requirements of GDC 26 are satisfied.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-264For Modes 1 through 5, the boron dilution analysis is performed to ensure that adequate time is available from alarm to total loss of shutdown margin for the operator to identify and terminate the dilution.
Thediscussion below demonstrates that all applicable acceptance criteria are met for this event at the WCGSin operating Modes 1 through 5.2.5.5.1.1 Introduction Reactivity can be added to the core by feeding primary-grade water into the RCS via the reactor makeupportion of the CVCS. Boron dilution is a manual operation under strict administrative controls with procedures calling for a limit on the rate and duration of dilution.
A boric acid blend system is provided to allow theoperator to match the boron concentration of the reactor coolant makeup water during normal charging to theRCS boron concentration.
As discussed below, the CVCS 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.2.5.5.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe opening of the primary water makeup control valves provides makeup to the CVCS and subsequently to the RCS, which can dilute the reactor coolant.
Inadvertent dilution from this source can be readilyterminated by closing the control valve. In order for makeup water to be added to the RCS at pressure, atleast one charging pump must be running in addition to a primary makeup water pump.The limiting dilution flow path is identified as the lowest resistance flow path for an unintentional dilution.
The boron dilution analysis excludes deliberate dilution operations from considerations.
Duringintentional boron dilution operations, the plant operators are keenly aware of and continually monitor thedilution process in progress for any sign of deviation or malfunction, such that the possibility of anundetected malfunction is considered remote. This is a standard assumption in the boron dilution analysismethodology.
Thus, the limiting boron dilution flow path does not include either the normal dilute or thealternative dilute flow paths (these paths are used only for deliberate dilution operations).
The limitingboron dilution flow path is the makeup flow path of the reactor makeup water system (RMWS) used innormal boration/blend operations.
The most common causes of an inadvertent boron dilution are the opening of the primary water makeupcontrol valve and failure of the blend system, either by controller or mechanical failure.
The CVCS andthe RMWS are designed to limit, even under various postulated failure modes, the potential rate ofdilution to values that will allow sufficient time for operator response to terminate the dilution.
Aninadvertent dilution from the RMWS may be terminated by closing the primary water makeup controlvalve. All expected sources of dilution may be terminated by closing isolation valves in the CVCS. Thelost shutdown margin may be regained by the opening of isolation valves to the RWST, thus allowing theaddition of borated water to the RCS.The rate at which unborated water can be added to the RCS is limited by the design of the CVCS andRMWS. The maximum (limiting) boron dilution flow rate is 245 gpm for Modes 1 and 2 with rod controlin manual mode, and 120 gpm in Mode 1 with rod control in automatic mode. For Modes 3 through 5, themaximum boron dilution flow rate is 157.5 gpm.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-265Information on the status of the reactor coolant makeup is continually available to the operator.
Lights areprovided on the control board to indicate the operating condition of the pumps in the CVCS. Alarms areactuated to warn the operator when boric acid or makeup water flow rates deviate from preset values as aresult of system malfunction.
A CVCS malfunction is classified as an ANS Condition II event, a fault of moderate frequency as definedby the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"ANSI N 18.2-1973.
Criteria established for Condition II events are as follows:The CHF must not be exceeded.
This is met by demonstrating that the minimum DNBR does notdecrease below the limit value at any time during the transient.
Pressure in the RCS and MSS must be maintained below 110 percent of the respective designpressures.
Fuel temperature and fuel cladding strain limits must not be exceeded.
The peak linear heatgeneration rate should not exceed a value that would cause fuel centerline melt.This event is analyzed to show that there is sufficient time for mitigation of an inadvertent boron dilutionprior to a complete loss of shutdown margin. A complete loss of plant shutdown margin results in a returnof the core to a critical condition causing an increase in the RCS temperature and heat flux. This couldviolate the SAL DNBR value and challenge the fuel and fuel cladding integrity.
A complete loss of plantshutdown margin could also result in a return of the core to a critical condition causing an increase inRCS pressure.
This could challenge the pressure design limits for the RCS and/or MSS.If the shutdown margin is shown not to be lost, the condition of the plant at any point in the transient iswithin the bounds of those calculated for other Condition II transients.
By showing that the above criteriaare met for those Condition II events, it can be concluded that they are also met for the boron dilutionevent. Operator action is relied upon to preclude a complete loss of plant shutdown margin.2.5.5.1.3 Description of Analyses and Evaluations Dilution During Mode 6 -An analysis is not performed for an uncontrolled boron dilution accidentduring refueling.
In this mode, the event is prevented by administrative controls that isolate the RCS fromthe potential source of unborated water.Dilution During Mode 5 Drained -The RCS water level can be dropped to the mid-plane of the HL formaintenance work that requires the SGs to be drained.
When the water level is drained down to themid-plane of the HL from a filled and vented condition in cold shutdown, an uncontrolled boron dilutionaccident is prevented by administrative controls that isolate the RCS from the potential source ofunborated water. Consequently, an analysis is not performed in this configuration.
Dilution During Mode 5 Filled -Typically, the plant is maintained in the Cold Shutdown mode whenRCS ambient temperatures are required.
Occasionally, reduced RCS inventory may be necessary.
Mode 5can also be a transition mode to either Refueling (Mode 6) or Hot Shutdown (Mode 4). Through the cycle,the plant may enter Mode 5 if reduced temperatures are required in containment or as the result of a TSWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-266required action. The plant is maintained in Mode 5 at the beginning of each cycle for startup testing ofcertain systems.
During this mode of operation, the control banks are fully inserted.
The following conditions are assumed for an uncontrolled boron dilution during cold shutdown.
The assumed dilution flow (157.5 gpm) is the maximum flow from the RMWS assuming multiplesimultaneous failures of control valves.The active RCS water volume for the WCGS is 8639.0 ft3.This active volume assumes at leastone RCP is in operation, with the volume of the pressurizer and surge line excluded to assure thatconservative estimates are made. Additionally, because no consideration is given to mixing in thereactor vessel upper head region, the volumes for the upper head and the downcomer from the topof the CLs to the bottom of the upper head spray nozzles are also excluded.
When no RCP is in operation, all dilution sources are isolated or under administrative control.This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS.The volume control tank (VCT) high water level alarm alerts the operators that a boron dilutionmay be in progress.
This is consistent with inadvertent boron dilution event analysis methodology approved by the USNRC for the WCGS.The initial boron concentration is assumed to be 1925 ppm (parts per million) at an RCStemperature of 68&deg;F, with all rods inserted (minus the most reactive RCCA), no xenon andshutdown margin of 1.3 percent Ak/k.The critical boron concentration is assumed to be 1800 ppm at an RCS temperature of 68&deg;F, withall rods inserted (minus the most reactive RCCA), and no xenon. The 125 ppm change from theinitial boron concentration noted above is a conservative minimum value.Dilution During Mode 4 -In Mode 4, the plant is being taken from a short-term mode of operation, ColdShutdown (Mode 5), to a long-term mode of operation, Hot Standby (Mode 3). Typically, the plant ismaintained in the Hot Shutdown mode to achieve plant heatup before entering Mode 3. The plant ismaintained in Mode 4 at the beginning of each cycle for startup testing of certain systems.
Throughout thecycle, the plant will enter Mode 4 if reduced temperatures are required in containment or as a result of aTS required action. During this mode of operation, the control banks are fully inserted.
In Mode 4, theprimary coolant forced flow that provides mixing can be provided by either the RHRS or a RCP,depending on system pressure.
The following conditions are assumed for an uncontrolled boron dilutionduring Hot Shutdown:
The assumed dilution flow (157.5 gpm) is the maximum flow from the RMWS assumingmultiple, simultaneous failures of control valves.The active RCS water volume for the WCGS is 8639.0 ft3.This active volume assumes at leastone RCP is in operation, with the volume of the pressurizer and surge line excluded to assure thatconservative estimates are made. Additionally, because no consideration is given to mixing in theWCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-267reactor vessel upper head region, the volumes for the upper head and the downcomer from the topof the CLs to the bottom of the upper head spray nozzles are also excluded.
When no RCP is in operation, all dilution sources are isolated or under administrative control.This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS (Reference 1).The VCT high water level alarm alerts the operators that a boron dilution may be in progress.
This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS.The initial boron concentration is assumed to be 1930 ppm at an RCS temperature of 200'F,with all rods inserted (minus the most reactive RCCA), no xenon and shutdown margin of1.3 percent Ak/k.The critical boron concentration is assumed to be 1800 ppm at an RCS temperature of 200'F,with all rods inserted (minus the most reactive RCCA), and no xenon. The 130 ppm change fromthe initial boron concentration noted above is a conservative minimum value.Dilution During Mode 3 -During this mode, rod control is in manual and the rods can be eitherwithdrawn or inserted.
In Mode 3, all RCPs may not be in operation.
In an effort to balance the heat lossthrough the RCS and the heat removal of the SGs, one or more of the pumps may be off to decrease heatinput into the system. In the approach to Mode 2, the operator must manually withdraw the control rodsand may initiate a limited dilution according to shutdown margin requirements, but not simultaneously.
Ifthe shutdown or control banks are withdrawn, the dilution scenario is similar to the Mode 2 analysiswhere the failure to block the source range trip results in a RT and immediate shutdown of the reactor.The dilution scenario is more limiting if the control rods are not withdrawn and the reactor is shut downby boron to the TS minimum requirement for Mode 3. The following conditions are assumed for anuncontrolled boron dilution during hot standby:The assumed dilution flow (157.5 gpm) is the maximum flow from the RMWS assumingmultiple, simultaneous failures of control valves.The active RCS water volume for the WCGS is 8639.0 ft3.This active volume assumes at leastone RCP is in operation, with the volume of the pressurizer and surge line excluded to assure thatconservative estimates are made. Additionally, because no consideration is given to mixing in thereactor vessel upper head region, the volumes for the upper head and the downcomer from the topof the CLs to the bottom of the upper head spray nozzles are also excluded.
The VCT high water level alarm alerts the operators that a boron dilution may be in progress.
This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS.The initial boron concentration is assumed to be 1645 ppm and 1940 ppm at RCS temperatures of557&deg;F and 350'F, respectively, with all rods inserted (minus the most reactive RCCA), no xenonand shutdown margin of 1.3 percent Ak/k.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-268The critical boron concentration is assumed to be 1500 ppm and 1800 ppm at RCS temperatures of 557&deg;F and 350'F, respectively, with all rods inserted (minus the most reactive RCCA), and noxenon. The changes between the associated initial and critical boron concentrations noted aboveare conservative minimum values.Dilution During Mode 2 -In this mode, the plant is being taken from one long-term mode ofoperation (Mode 3) to another (Mode 1). The plant is maintained in the Startup mode only for the purposeof startup testing at the beginning of each cycle. All normal actions required to change power level, eitherup or down, require operator initiation.
Assumed conditions at startup require the reactor to have available at least 1.3 percent Ak/k shutdown margin. The following conditions are assumed for an uncontrolled boron dilution during startup:* The assumed dilution flow (245 gpm) is the maximum flow from the RMWS assuming
: multiple, simultaneous failures of control valves.* Conservative estimates of the minimum active RCS water volume are made by excluding thepressurizer and surge line. For the WCGS, the active RCS water volume is 9810 ft3.* The RT on source range neutron flux level alerts the operators that a boron dilution may be inprogress.
* The initial boron concentration is assumed to be 1935 ppm, which is a conservative maximumvalue for the critical concentration at the condition of HZP, with the rods at the insertion limits,and no xenon.* The critical boron concentration following RT is assumed to be 1500 ppm, corresponding to HZP,all rods inserted (minus the most reactive RCCA), no xenon conditions.
The 435 ppm changefrom the initial condition noted above is a conservative minimum value.Mode 2 is a transitory operational mode in which the operator intentionally dilutes and withdraws controlrods to take the plant critical.
During this mode, the plant is in manual control with the operator requiredto maintain a high awareness of the plant status. For a normal approach to criticality, the operator mustmanually initiate a limited dilution and withdraw the control rods, a process that takes several hours. Priorto approaching criticality, the TS require that the predicted position of the rods is within the rod insertion limits. This ensures that the reactor did not go critical with the control rods below the insertion limits.Once critical, the power escalation must be sufficiently slow to allow the operator to manually block thesource range RT (nominally at 105 cps) after reaching permissive P-6. Too fast of a power escalation (dueto an unknown dilution) would result in reaching P-6 unexpectedly, leaving insufficient time to manuallyblock the source range RT. Failure to perform this manual action results in a RT and immediate shutdownof the reactor.However, in the event of an unplanned approach to criticality or dilution during power escalation while inMode 2, the plant status is such that minimal impact will result. The plant will slowly escalate in power toa RT on the power range neutron flux low setpoint.
After RT, more than 15 minutes is available foroperator action prior to return to criticality.
Mode 2 results are summarized in Table 2.5.5-1.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-269Dilution During Mode 1 -In this mode, the plant can be operated in either automatic or manual rodcontrol.
With the reactor in manual control and no operator action taken to terminate the transient, thepower and temperature increase will cause the reactor to reach the power range high neutron flux tripsetpoint or the OTAT trip setpoint, resulting in a RT. In this case, the boron dilution transient up to thetime of trip is essentially equivalent to an uncontrolled RCCA bank withdrawal at power. Following RT,there is at least 15 minutes prior to criticality.
This is sufficient time for the operator to determine thecause of dilution and isolate the reactor makeup water source before the available shutdown margin islost.With the reactor in automatic rod control, the power and temperature increase from the boron dilutionresults in insertion of the control rods and a decrease in the available shutdown margin. As the dilutionand rod insertion
: continue, the rod insertion limit alarms (low and low-low settings) and axial fluxdifference alarm alert the operator at least 15 minutes prior to criticality that a dilution is in progress andthat the TS requirement for shutdown margin may be challenged.
This is sufficient time to determine thecause of dilution and isolate the reactor makeup water source before the available shutdown margin islost.The effective reactivity addition rate is primarily a function of the dilution rate, boron concentration, andboron worth. The following conditions are assumed for an uncontrolled boron dilution during full power:* The assumed dilution flow (245 gpm with rod control in manual mode, and 120 gpm for rodcontrol in automatic mode) is the maximum flow from the RMWS assuming
: multiple, simultaneous failures of control valves.0 Conservative estimates of the minimum active RCS water volume are made by excluding thepressurizer and surge line. For the WCGS, the active RCS water volume is 9810 ft3.* The RT on power range neutron flux high or OTAT alerts the operators that a boron dilution maybe in progress.
a The initial boron concentration is assumed to be 1954 ppm, which is a conservative maximumvalue for the initial concentration at the condition of HFP, with the rods at the insertion limits, andno xenon.* The critical boron concentration following RT is assumed to be 1500 ppm, corresponding to theHZP, all rods inserted (minus the most reactive RCCA), and no xenon condition.
The 454 ppmchange from the initial condition noted above is a conservative minimum value.* A 1.3 percent Ak/k minimum shutdown margin is assumed in the analysis.
* Bounding boron worths of -15 pcm/ppm and -5 pcm/ppm are conservatively considered.
Thelarger absolute value maximizes the reactivity insertion rate, whereas the smaller absolute valueminimizes the reactivity insertion rate thereby delaying the time to reach the RT setpoint.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2702.5.5.1.4 ResultsThe boron dilution analysis concluded that all applicable acceptance criteria are met for the WCGS. Thismeans that operator action to terminate the dilution flow within 15 minutes from operator notification (first alarm) in Modes 1, 2, 3, 4 and 5, precludes a complete loss of shutdown margin. The results of theboron dilution analysis are provided in Table 2.5.5-1.No analysis is presented for the Mode 5 drained condition or Mode 6 operation because dilution isprecluded by administrative controls.
If an unintentional dilution of boron in the RCS does occur, numerous alarms and indications areavailable to alert the operator to the condition.
The maximum reactivity addition due to the dilution isslow enough to allow the operator sufficient time to determine the cause of the addition and takecorrective action before shutdown margin is lost. The acceptance criteria as specified in Section 2.5.5.1.2 are met.2.5.5.2 Conclusion The analyses of the decrease in boron concentration in the reactor coolant due to a CVCS malfunction have been reviewed.
It is concluded that the analyses have adequately accounted for plant operation at thecurrent and proposed uprated power levels and were performed using acceptable analytical models. Also,when there is a decrease in boron concentration event, the analyses demonstrate that the reactor protection and safety systems will continue to ensure that the specified acceptable fuel design limits and theRCS and MSS pressure limits will not be exceeded.
Based on this, it is concluded that the WCGS willcontinue to meet the requirements of GDCs 10, 15, and 26.2.5.5.3 References I. Letter from James C. Stone (USNRC) to Neil S. Cams (WCNOC),
"Wolf Creek Generating Station -Amendment No. 96 to Facility Operating License No. NPF-42 (TAC No. M94112),"
March 1, 1996.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESUNGHOUSE NON-PROPRIETARY CLASS 32-271Table 2.5.5-1 CVCS Malfunction Boron Dilution Event Results -Event Alarm to Loss of ShutdownMarginAvailable OperatorAction Time LimitOperating Mode (minutes)
(minutes)
Mode 1 -Manual Rod Control 50.3 15Mode 1 -Automatic Rod Control 112.9 15Mode 2 56.0 15Mode 3 -557&deg;F 15.8 15Mode 3 -350'F 15.6 15Mode 4 -200'F 15.8 15Mode 5 -68&deg;F -Filled 15.7 15Mode 5 -68&deg;F -DrainedMode 6Note:1. No analysis is presented for the Mode 5 drained condition or Mode 6 because boron dilution is precluded byadministrative controls.
WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2722.5.6 Spectrum of Rod Cluster Control Assembly Ejection Accidents (USAR Section 15.4.8)2.5.6.1 Technical Evaluation The criterion applied to ensure the core remains in a coolable geometry following a rod ejection incidentis that the average fuel pellet enthalpy at the hot spot must remain less than 200 cal/gm (360 Btu/lbm).
The use of the initial conditions presented in Table 2.5.6-1 resulted in conservative calculations of the fuelpellet enthalpy.
The results of the licensing basis analyses demonstrated that the fuel pellet enthalpy doesnot exceed 360 Btu/lbm for any of the rod ejection cases analyzed.
Overpressurization of the RCS during a rod ejection event is generically addressed in WCAP-7588, Revision 1-A (Reference 1).Another applicable acceptance criterion is that fuel melting must be limited to less than the innermost 10 percent of the fuel pellet at the hot spot, even if the average fuel pellet enthalpy at the hot spot is lessthan the limit of 360 Btu/lbm.
Conservative fuel melt temperatures of 4900'F and 4800'F were assumedfor the hot spot for the BOL and EOL cases, respectively.
These fuel melting temperatures correspond to aspecific burnup limit at the hot spot. The peak UO2 burnup at the hot spot is based on the assembly withthe maximum post-ejection FQ, which is typically a fresh fuel assembly.
Therefore, the fuel meltingtemperatures represent bounding values for the assumed UO2 burnup at the hot spot. The maximumburnup at the hot spot at BOL and EOL is confirmed to be below these values as part of the reloadprocess.
This assumption does not affect the maximum licensed fuel burnup limit. The results of thelicensing basis rod ejection analyses demonstrated that the amount of fuel melting was limited to less than10 percent of the fuel pellet at the hot spot for each of the rod ejection cases.2.5.6.1.1 Introduction This accident is defined as a mechanical failure of a CRDM pressure housing resulting in the ejection ofthe RCCA and drive shaft. The consequence of this mechanical failure is a rapid, positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.The resultant core thermal power excursion is limited by the Doppler reactivity effect of the increased fuel temperature and terminated by RT actuated by high nuclear power signals.A failure of a CRDM housing sufficient to allow a control rod to be rapidly ejected from the core is notconsidered credible for the following reasons:* Each full-length CRDM housing is completely assembled and shop tested at 4100 psig.* The mechanism housings are individually hydrotested after they are attached to the head adaptersin the reactor vessel head and checked during the hydrotest of the completed RCS.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-273Stress levels in the mechanism are not affected by anticipated system transients at power or by thethermal movement of the coolant loops. Moments induced by the design earthquake can beaccepted within the allowable primary working stress ranges specified in the American Society ofMechanical Engineers Boiler and Pressure Vessel Code (ASME B&PV), Section III, for Class Icomponents.
The latch mechanism housing and rod travel housing are each a single length of forged type-304stainless steel. This material exhibits excellent notch toughness at all temperatures that will beencountered.
A significant amount of margin of strength in the elastic range, together with the large energy absorption capability in the plastic range, gives additional assurance that the gross failure of the housing will notoccur. The joints between the latch mechanism housing and rod housing are threaded joints reinforced bycanopy-type rod welds.In general, the reactor is operated with the RCCAs inserted only far enough to permit load follow.Reactivity changes caused by the core depletion are compensated by boron changes.
Furthermore, thelocation and grouping of control rod banks are selected during the nuclear design to lessen the severity ofan RCCA ejection accident.
Therefore, if an RCCA is ejected from its normal position during full-power operation, only a minor reactivity excursion, at worst, could be expected to occur. The position of all ofthe RCCAs is continuously indicated in the control room. An alarm will occur if a bank of RCCAsapproaches its insertion limit or if one control rod assembly deviates from its bank. There are low andlow-low level insertion alarm circuits for each bank. The control rod position monitoring and alarmsystems are described in Reference 1.2.5.6.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput parameters for the analysis were conservatively selected on the basis of values calculated for thistype of core. The most important parameters are discussed below. Table 2.5.6-1 presents the parameters used in this analysis.
Ejected Rod Worths and Hot Channel FactorsThe values for the ejected rod worths and hot channel factors were calculated using either 3-D staticmethods or a synthesis of I-D and 2-D calculations.
Standard nuclear design codes were used in theanalysis.
No credit was taken for the flux-flattening effects of reactivity feedback.
The calculation wasperformed for the maximum allowed bank insertion at a given power level, as determined by the rodinsertion limits. The analysis assumed adverse xenon distributions to provide worst-case results.Appropriate margins were added to the ejected rod worth and hot channel factors to account for anycalculational uncertainties.
Delayed Neutron Fraction, NeffCalculations of the effective delayed neutron fraction (I3eff) typically yield values of approximately 0.75 percent at BOL and 0.40 percent at EOL. The ejected rod accident is sensitive to I3eff if the ejectedWCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-274rod worth is equal to or greater than I3erf, as in the zero-power transients.
In order to allow for future fuelcycle flexibility, conservative estimates of 3elf of 0.49 percent at beginning of cycle and 0.44 percent atend of cycle were used in the analysis.
Reactivity Weighting FactorThe largest temperature rises, and therefore the largest reactivity feedbacks, occur in channels where thepower is higher than average.
Since the weight of a region is dependent on flux, these regions have highweights.
This means that the reactivity feedback is larger than that indicated by a simple channel analysis.
Physics calculations have been carried out for temperature changes with a flat temperature distribution, and with a large number of axial and radial temperature distributions.
Reactivity changes were compared and effective weighting factors determined.
These weighting factorstake the form of multipliers which, when applied to single-channel feedbacks, correct them to effective whole-core feedbacks for the appropriate flux shape. In this analysis, a I -D (axial) spatial kinetics methodwas employed.
Therefore, axial weighting is not necessary if the initial condition is made to match theejected rod configuration.
In addition, no weighting was applied to the moderator feedback.
Aconservative radial weighting factor was applied to the transient fuel temperature to obtain an effective fuel temperature as a function of time accounting for the missing spatial dimension.
These weighting factors have also been shown to be conservative compared to 3-D analysis.
Moderator and Doppler Coefficient The MTC and the DTC are combined and input as an isothermal temperature coefficient (ITC). The ITCsthat were modeled are +7.695 pcm/&deg;F at zero-power nominal Ta,,g and +5.247 pcm/&deg;F at full-power Tavgfor the BOL cases. These are very conservative values that easily bound the BOL moderator temperature limit of +5 pcm/&deg;F. For the EOL cases, the applicable zero-power ITC was -16.817 pcm/&deg;F and the full-power MTC was -22.920 pcm/&deg;F.The Doppler reactivity defect as a function of power level was adjusted in the nuclear code to aconservative design value using a Doppler weighting factor of 1.0. The Doppler weighting factor wasincreased under accident conditions, as discussed above.Heat Transfer DataThe FACTRAN code (Reference 2), which contains standard curves of thermal conductivity versus fueltemperature, is used to determine the hot spot transient.
During the transient, the peak centerline fueltemperature is nearly independent of the gap conductance.
The cladding temperature is, however, stronglydependent on the gap conductance and is highest for high gap conductance.
For conservatism, a lowinitial gap heat transfer coefficient was used at the beginning of the transient to maximize the initial fueltemperature and a high gap heat transfer coefficient value of 10,000 Btu/hr-ft 2 was used for the remainder of the transient to maximize the cladding temperature.
This high gap heat transfer coefficient corresponds to a negligible gap resistance, and a further increase would have essentially no effect on the rate of heattransfer.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-275Coolant Mass Flow RatesWhen the core is operating at full power, all four RCPs are always operational.
For zero-power conditions, the system was conservatively assumed to be operating with two pumps. The principal effectof operating at reduced flow is to reduce the film boiling heat transfer coefficient.
This resulted in higherPCTs, but did not affect the peak centerline fuel temperature.
Reduced flow also lowers the CHF.However, since DNB was always assumed at the hot spot, and since the heat flux rose very rapidly duringthe transient, this produced only second-order changes in the cladding and centerline fuel temperatures.
Trip Reactivity Insertion The trip reactivity insertion was assumed to be 4.0 percent Ak from HFP conditions and 2.0 percent Akfrom HZP conditions, including the effect of one stuck RCCA. These values were also reduced by theejected rod reactivity.
The shutdown reactivity was simulated by dropping a rod of the required worth intothe core. The start of rod motion occurred 0.5 seconds after reaching the power range high neutron fluxtrip setpoint.
It was assumed that insertion to dashpot did not occur until 2.7 seconds after the rods beganto fall. The time delay to full insertion combined with the 0.5 second trip delay conservatively delayedinsertion of shutdown reactivity into the core.Due to the extremely low probability of an RCCA ejection
: accident, this event is classified as aCondition IV event as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
As such, some fuel damage is considered anacceptable consequence.
The real physical limits of this accident are that the rod ejection event and any consequential damage toeither the core or the RCS must not prevent long-term core cooling.
More specific and restrictive criteriaare applied to ensure that there is no fuel dispersal in the coolant and that gross lattice distortion or severeshock waves do not occur. Based on experimental data, Reference 1 concludes that the acceptance criteriato be applied for an RCCA ejection are:Average fuel pellet enthalpy at the hot spot must remain below 200 cal/gm for irradiated fuel.This bounds non-irradiated fuel, which has a slightly higher enthalpy limit.Peak reactor coolant pressure must be less than that which could cause RCS stresses to exceed thefaulted-condition stress limits (Note: the peak pressure aspects of the rod ejection transient areaddressed generically in Reference 1).Fuel melting is limited to less than the innermost 10 percent of the pellet volume at the hot spoteven if the average fuel pellet enthalpy at the hot spot is below the 200 cal/gm fuel enthalpy limit.2.5.6.1.3 Description of Analyses and Evaluations This section describes the models used in the analysis of the rod ejection accident.
Only the initial fewseconds of the power transient are discussed since the long-term considerations are the same as those for asmall LOCA.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-276The calculation of the RCCA ejection transient was performed in two stages: first an average core channelcalculation, and then a hot spot calculation.
The average core calculation used spatial neutron-kinetics methods to determine the average power generation with time, including the various total core feedbackeffects; that is, Doppler reactivity and moderator reactivity.
Enthalpy and temperature transients at the hotspot were then determined by multiplying the average core energy generation by the hot channel factorand performing a fuel rod transient heat transfer calculation.
The power distribution calculated withoutfeedback was conservatively assumed to continue throughout the transient.
A detailed discussion of themethod of analysis can be found in Reference 1.Average CoreThe spatial-kinetics computer code TWINKLE (Reference
: 3) was used for the average core transient analysis.
This code solves the two-group neutron diffusion theory kinetic equation in one, two, or threespatial dimensions (rectangular coordinates) for six delayed neutron groups and up to 2000 spatial points.The computer code includes a detailed, multi-region, transient fuel-clad-coolant heat transfer model forcalculation of pointwise Doppler and moderator feedback effects.
This analysis used the code as al-D axial kinetics code since it allows a more realistic representation of the spatial effects of axialmoderator feedback and RCCA movement.
: However, since the radial dimension was missing, it was stillnecessary to employ very conservative methods (described below) of calculating the ejected rod worthand hot channel factor.Hot Spot AnalysisIn the hot spot analysis, the initial heat flux is equal to the nominal heat flux times the design hot channelfactor. During the transient, the heat flux hot channel factor is linearly increased to the transient value in0.1 second, the time for full ejection of the rod. Therefore, the assumption is made that the hot spot beforeand after ejection are coincident.
This is very conservative since the peak after ejection will occur in oradjacent to the assembly with the ejected rod, and prior to ejection the power in this region willnecessarily be depressed.
The average core energy addition, calculated as described above, was multiplied by the appropriate hotchannel factors.
The hot spot analysis used the detailed fuel and cladding transient heat transfer computercode FACTRAN (Reference 2). This computer code calculates the transient temperature distribution in across section of a metal-clad U02 fuel rod, and the heat flux at the surface of the rod, using the nuclearpower versus time and local coolant conditions as input. The zirconium-water reaction is explicitly represented, and all material properties are represented as functions of temperature.
A parabolic radialpower distribution was assumed within the fuel rod.FACTRAN uses the Dittus-Boelter or Jens-Lottes correlation to determine the film heat transfer beforeDNB, and the Bishop-Sandberg-Tong correlation to determine the film boiling coefficient after DNB. TheBishop-Sandberg-Tong correlation was conservatively used assuming zero bulk fluid quality.
The DNBheat flux was not calculated.
: Instead, the code was forced into DNB by specifying a conservative DNBheat flux. The gap heat transfer coefficient could be calculated by the code. However, it was adjusted toforce the full-power, steady-state temperature distribution to agree with fuel heat transfer design codes.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-277Reactor Protection The protection for this accident, as explicitly modeled in the analysis, is provided by the power rangeneutron flux trip (high and low settings).
The power range high neutron flux positive rate tripcomplements the high flux trip function (high and low settings) to ensure that the criteria are met for rodejection from partial power.2.5.6.1.4 ResultsThe results of the analyses performed for the rod ejection event, which cover BOL and EOL conditions atHFP and HZP for the WCGS,. are discussed below.Beginning of Cycle, Zero PowerThe worst ejected rod worth and hot channel factor were conservatively calculated to be 0.78 percent Akand 13.0, respectively.
The peak hot spot average fuel pellet enthalpy reached 254.7 Btu/lbm(141.5 cal/gm).
The peak fuel centerline temperature never reached the BOL melt temperature of 4900'F.Therefore, no fuel melting is predicted.
Beginning of Cycle, Full PowerControl bank D was assumed to be inserted to its insertion limit. The worst ejected rod worth and hotchannel factor were conservatively calculated to be 0.23 percent Ak and 6.6, respectively.
The peak hotspot average fuel pellet enthalpy reached 317.6 Btu/lbm (176.4 cal/gm).
The peak fuel centerline temperature reached the BOL melt temperature of 4900'F. However, fuel melting remained well belowthe limiting criterion of 10 percent of total pellet volume at the hot spot.End of Cycle, Zero PowerThe worst ejected rod worth and hot channel factor were conservatively calculated to be 0.86 percent Akand 21.0, respectively.
The peak hot spot average fuel pellet enthalpy reached 261.4 Btu/lbm(145.2 cal/gm).
The peak fuel centerline temperature never reached the EOL melt temperature of 4800'F.Therefore, no fuel melting is predicted.
End of Cycle, Full PowerControl bank D was assumed to be inserted to its insertion limit. The ejected rod worth and hot channelfactors were conservatively calculated to be 0.25 percent Ak and 7.1, respectively.
The peak hot spotaverage fuel pellet enthalpy reached 305.4 Btu/lbm (169.7 cal/gm).
The peak fuel centerline temperature reached the EOL melting temperature of 4800'F. However, fuel melting remained well below the limitingcriterion of 10 percent of total pellet volume at the hot spot.A summary of the parameters used in the rod ejection
: analyses, and the analyses
: results, are presented inTable 2.5.6-1.
The sequence of events for all four cases is presented in Table 2.5.6-2.
Figure 2.5.6-1shows the results for the BOL/HZP case and Figure 2.5.6-2 shows the BOL/HFP plot results.
TheEOL/HZP and EOL/HFP results are presented in Figures 2.5.6-3 and 2.5.6-4, respectively.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-278A detailed calculation of the pressure surge for an ejected rod worth of 1 dollar at BOL HFP indicates thatthe peak pressure did not exceed that which would cause the RPV stress to exceed the faulted condition stress limits (Reference 1). Since the severity of the present analysis did not exceed the worst-case
: analysis, the accident for this plant will not result in an excessive pressure rise or further adverse effectson the RCS.2.5.6.2 Conclusion Despite the conservative assumptions, the analyses indicate that the described fuel and cladding limitswere not exceeded.
It is concluded that there is no danger of sudden fuel dispersal into the coolant.
Sincethe peak pressure did not exceed that which would cause stresses to exceed the faulted condition stresslimits, it is concluded that there is no danger of further consequential damage to the RCS. Genericanalyses demonstrated that the fission product release as a result of fuel rods entering DNB was limited toless than 10 percent of the fuel rods in the core.The results and conclusions of the analyses performed for the rupture of a CRDM housing RCCA ejectionsupport operation up to the analyzed reactor core power of 3637 MWt.Based on the review of the analyses of the rod ejection
: accident, it was concluded that the analyses haveadequately accounted for plant operation at the stated power level and were performed using acceptable analytical models. It is further concluded that appropriate reactor protection and safety systems willprevent postulated reactivity accidents that could result in damage to the RCPB greater than limited localyielding, or cause sufficient damage that would significantly impair the capability to cool the core.2.5.6.3 References
: 1. WCAP-7588, Revision 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Special Kinetics Methods,"
January 1975.2. WCAP-7908-A, "FACTRAN
-A FORTRAN IV Code for Thermal Transients in a UO2 FuelRod," December 1989.3. WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non-Proprietary),
"TWINKLE
-A Multi-Dimensional Neutron Kinetics Computer Code," January 1975.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-279Table 2.5.6-1 Selected Input and Results of the Limiting RCCA Ejection AnalysesInput BOC BOC EOC EOCInitial Reactor Core Power Level (MWt) 3637 0 3637 0Ejected Rod Worth (%Ak) 0.23 0.78 0.25 0.86Delayed Neutron Fraction
(%) 0.49 0.49 0.44 0.44Doppler Reactivity Weighting 1.433 2.309 1.499 3.078Trip Reactivity
(%Ak) 4.0 2.0 4.0 2.0FQ Before Rod Ejection (fraction) 2.50 -- 2.50 --FQ After Rod Ejection (fraction) 6.6 13.0 7.1 21.0Number of Operational Pumps 4 2 4 2Results BOC BOC EOC EOCMaximum Fuel Pellet Average Temperature (0F) 4041 3357 3911 3432Maximum Fuel Centerline Temperature
(&deg;F) 4965 3867 4864 3869Maximum Cladding Average Temperature
(&deg;F) 2270 2498 2191 2626Maximum Fuel Stored Energy (cal/gm) 176.4 141.5 169.7 145.2Maximum Fuel Melt at the Hot Spot (%) 4.62 0.00 3.96 0.00WCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-280Table 2.5.6-2 Time Sequence of Events -RCCA EjectionTime (seconds)
Event BOL HFP EOL HFPInitiation of Rod Ejection 0.0 0.0Power Range Neutron Flux Setpoint Reached 0.05 0.04Peak Nuclear Power Occurs 0.13 0.14Rods Begin to Fall 0.55 0.54Peak Fuel Average Temperature Occurs 2.22 2.33PCT Occurs 2.28 2.37Peak Heat Flux Occurs 2.30 2.38Peak Fuel Centerline Temperature Occurs 4.00 4.09BOL HZP EOL HZPInitiation of Rod Ejection 0.0 0.0Power Range Neutron Flux Setpoint Reached 0.24 0.19Peak Nuclear Power Occurs 0.28 0.22Rods Begin to Fall 0.74 0.69PCT Occurs 2.18 1.56Peak Heat Flux Occurs 2.18 1.57Peak Fuel Average Temperature Occurs 2.38 1.82Peak Fuel Centerline Temperature Occurs 3.09 2.87WCAP-l 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-281WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-28 1-,00U--(3-)0,0v0_201'15-10-5 .................I. ..........U024Time6(Seconds)
I810FuelFuelCladMelt5000----4 00 .....=3'/2000 .E1000 ......i,Center I i neAve rageOutering = 4900 DegreesF0 2 4 6 8Time (Seconds) 10Figure 2.5.6-1 Rod Ejection
-BOL/HZPWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-282E000 2 4 6 8Time (Seconds) 10FuelFuelCladMeltCenter I i neAve rageOutering = 4900 DegreesF~r'J'Jv.Or)(DCIa,LEa,C-U-_4000-3000-2000-7 -------..... /' ..... ....... L...................
.................-.....
S.......
.... .............
..........
.. ... ... .I nnA.,024 6Time (Seconds) 810Figure 2.5.6-2 Rod Ejection
-BOL/HFPWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-283WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-28300U..-0O320"15-10-5-A --I I -I I I I I I I I IU024T 6Time (Seconds) 810Fuel Center IineFuel AverageClad OuterMel t ing = 4800 Degrees FU-C,)L.C,)L..=30L..~I)0~EH-U-0 2 4 6 8 10Time (Seconds)
Figure 2.5.6-3 Rod Ejection
-EOL/HZPWCAP- 17658-NP August 2013WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-284WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-284EC0C>w_0)C)0 2 4 6 8Time (Seconds) 10Fuel--- -Fuel-~CladMelt5000:4000".(3)r/CDC,,.--- 3000"/2000-Dc-- -Center I i neAverageOutering = 4800 DegreesF0 2 4 6 8Time (Seconds) 10Figure 2.5.6-4 Rod Ejection
-EOL/HFPWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2852.6 INCREASE IN REACTOR COOLANT INVENTORY 2.6.1 Inadvertent Operation of the Emergency Core Cooling System During PowerOperation (USAR Section 15.5.1)2.6.1.1 Technical Evaluation 2.6.1.1.1 Introduction An inadvertent actuation of the ECCS at power event results in an increase in RCS inventory, leading tothe potential filling of the pressurizer.
Operator error or a spurious electrical actuating signal could causethe event.Following the actuation signal, the SI system is actuated, which results in borated water being pumpedinto the CL of each RCS loop. Normally, an SI actuation signal results in an immediate and automatic RT,which in turn generates a turbine trip. However, even without an immediate RT, the reactor willexperience a negative reactivity excursion as a result of the borated water being injected.
This negativereactivity results in a decrease in reactor power.In manual rod control, the primary-to-secondary system power mismatch causes a decrease in coolanttemperature and a contraction of the reactor coolant.
Assuming an immediate RT signal is not received, the RCS responds with a decrease in pressurizer pressure and water level, and the turbine load willdecrease because of reduced steam pressure once the turbine throttle valves are fully open. The decreasein RCS pressure results in an increase in SI flow because of the SI pump performance characteristics.
In automatic rod control, RCCA withdrawal may compensate for the above effects as the control systemresponds to maintain programmed Tavg. Once the rods have been fully withdrawn, the event continues asdescribed for operation in manual rod control.The Inadvertent ECCS actuation at power event is performed to demonstrate that sufficient time isavailable for the appropriate operator actions to be taken to preclude a pressurizer water-solid condition.
2.6.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The following inputs and assumptions were applied in the analysis of the Inadvertent ECCS event:The initial NSSS power is 3651 MWt, which includes all applicable uncertainties.
A full power T,,,g range of 570.7&deg;F to 588.4&deg;F was considered in the analysis.
The limiting initialTavg is 566.7'F, which corresponds to the low nominal full power Tavg minus uncertainties (including bias). The lower initial temperature corresponds to a higher reactor coolant mass,which leads to a more severe pressurizer water volume transient.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-286The initial Tfred is 448.6'F, which corresponds to the high end of the full power Tfeed range(400.0&deg;F to 448.6&deg;F).
The initial pressurizer water level is 46 percent span, which is the nominal pressurizer water levelof 41 percent span at the low full power Tavg of 570.7&deg;F plus 5 percent span uncertainty.
The initial pressurizer pressure is 2215 psia, which is the nominal value of 2250 psia minus 35 psiuncertainty.
A lower initial RCS pressure is conservative because it allows higher SI flows to beinjected into the RCS.The pressurizer proportional heaters and pressurizer sprays were modeled to function as-designed because their operation generates a more limiting condition with respect to filling the pressurizer.
Because an SI signal causes the pressurizer backup heaters to be shed from their electric powersupply, and they are not loaded onto another power supply automatically or manually untilletdown flow is re-established, the backup heaters were not modeled.* A maximum SGTP level of 10 percent was modeled.The total flow initially injected to the RCS corresponds to maximum flow from two centrifugal charging pumps (CCPs) and one normal charging pump; this total flow is reflective of SI flow tothe CLs plus RCP seal injection flow.An immediate RT on the SI actuation signal and a turbine trip derived from the RT were modeledbecause these limit the primary-to-secondary heat transfer rate, thus minimizing the magnitude ofthe initial reactor coolant shrinkage.
Within 6 minutes from event initiation, the plant operators are assumed to initiate actions tocontrol RCS (CL) temperature to 557&deg;F. This is conservatively modeled in the analysis byopening one of the four SG ARVs at 6 minutes after event initiation to control Too1d to atemperature of 557&deg;F.At 8 minutes after event initiation, operator action to terminate SI flow to the CLs was credited.
After the SI flow is terminated, the only source of flow injection to the RCS is maximum RCPseal injection flow from one CCP.At 29.5 minutes after event initiation, operator action to re-establish letdown flow was credited.
This was conservatively modeled by terminating the remaining flow injection to the RCS; noRCS inventory reduction was modeled in the analysis.
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-287Acceptance CriteriaBased on the expected frequency of occurrence, the Inadvertent ECCS event is considered to be aCondition II event as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
The following items summarize the acceptance criteria associated with the analysis of this event:Fuel cladding integrity is maintained by ensuring that the minimum DNBR remains above the95/95 DNBR SAL.Based on historical precedence, the Inadvertent ECCS event does not lead to a serious challenge of the DNB design basis. The conditions do not approach the core thermal DNB limits, as thecore power, RCS pressure and RCS temperatures remain relatively unchanged.
Therefore, theDNBR typically increases and does not approach the DNBR SAL following event initiation.
Assuch, no explicit analysis of the event was performed to calculate a minimum DNBR value.* Pressures in the RCS and MSS are maintained below 110 percent of the design pressures.
With respect to the overpressure evaluation, the Inadvertent ECCS event is bounded by theLOL/TT event, discussed in Section 2.3.1, in which assumptions are made to conservatively maximize the RCS and MSS pressure transients.
For the Inadvertent ECCS event, turbine tripoccurs following RT, whereas for the LOL/TT event, the turbine trip is the initiating fault.Therefore, the primary-to-secondary power mismatch and resultant RCS and MSS heatup andpressurization transients are always more severe for the LOL/TT event. For this reason, it is notnecessary to calculate the maximum RCS or MSS pressures for the Inadvertent ECCS event.An incident of moderate frequency does not generate a more serious plant condition without otherfaults occurring independently.
The major concern from an Inadvertent ECCS event is that associated with pressurizer filling.The pressurizer water volume increases for this event as a result of the flow injected to the RCS.This event is analyzed to demonstrate that sufficient time is available for the appropriate operatoractions to be taken to preclude a pressurizer water-solid condition.
The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the Inadvertent ECCS acceptance criteria are provided as follows.* GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the Inadvertent ECCS event, this is shown to be met by demonstrating that thefuel cladding integrity is maintained.
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-288GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.
For theInadvertent ECCS event, this is shown to be met by demonstrating that the peak RCS pressure isless than 110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions suchas stuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.
For the Inadvertent ECCS event,which results in a RT, this is shown to be met by demonstrating that the fuel cladding integrity ismaintained with a trip reactivity that accounts for the most reactive rod stuck out of the core.2.6.1.1.3 Description of Analyses and Evaluations The Inadvertent ECCS event was analyzed using the RETRAN computer code (Reference 1). TheRETRAN model simulates the RCS, neutron kinetics, pressurizer, pressurizer relief and safety valves,pressurizer
: heaters, pressurizer spray, SI system, SGs, FW system, and MSSVs. The code computespertinent plant variables including nuclear power, reactor coolant average temperature, RCS pressure, pressurizer water volume, and SG pressure.
2.6.1.1.4 ResultsThe calculated sequence of events for the limiting Inadvertent ECCS case is presented in Table 2.6.1-1and transient plots of the significant plant parameters are provided in Figures 2.6.1-1 through 2.6.1-3.
RToccurs at the event initiation followed by a rapid cooldown of the RCS. The initial coolant contraction results in a short-term reduction in pressurizer pressure and water level. The combination of the RCSheatup, due to residual RCS heat generation, and ECCS injected flow causes the pressure and leveltransients to rapidly turn around. The RCS heatup continues until a source of cooling is established, firstvia the automatic opening of the lowest-setting MSSV of each loop, and then via the single ARV that wasmodeled at 6 minutes after event initiation to simulate the plant operators taking action to control theRCS (CL) temperature to 557&deg;F. The pressurizer water level increases rapidly until SI flow to the CLs isterminated at 8 minutes as a result of crediting operator action, after which the pressurizer level continues to increase at a much slower rate until letdown flow is assumed to be re-established (via operator action)at 29.5 minutes.
The results of the analysis show that the pressurizer does not reach a water-solid condition provided that the plant operators initiate the required operator actions within the assumed timelimits.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2892.6.1.2 Conclusion Based on the above information, it is concluded that the Inadvertent ECCS event will not progress into amore serious plant condition.
Thus, all applicable event acceptance criteria are satisfied.
It has beendemonstrated that the reactor protection and safety systems ensure that the specified acceptable fueldesign limits are met and the RCPB pressure limits will not be exceeded as a result of the Inadvertent ECCS event. Based on this, the plant will continue to meet the requirements of GDCs 10, 15, and 26.2.6.1.3 References
: 1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-290Table 2.6.1-1 Time Sequence of Events -Inadvertent ECCSTimeEvent Seconds MinutesInadvertent ECCS/SI Signal Actuation 0.0 0.0RT due to SI SignalTurbine Trip from RTLetdown Isolation On Each Loop, the MSSV with the Lowest Setting Opens 236.4 3.9One ARV Begins to Open (1st Operator Action) 360.0 6.0All MSSVs Closed 373.5 6.2SI Flow to CLs Terminated (2nd Operator Action.)
480.0 8.0Letdown Flow Re-established (3rd Operator Action) 1770.0 29.5Maximum Pressurizer Water Volume (1786.5 ft3) Reached 2450.0 40.8WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-29113:0C-)0.80.6-0.4-0.2-0590mo57CD5805D-570CD<2 560U)CD550I [ [- " I -t- I .. ...-I. .1120002400 3600Time (seconds) 48006000I I II I IIII I I I I I I012001 12400 3600Time (seconds) 448006000Figure 2.6.1-1 Inadvertent ECCS -Nuclear Power and Tavg versus TimeWCAP-1 7658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-29224002300-o" 2200-cnG-)2100-Cn2000I I II t II I II I I1 QI (1012002400 3600Time (seconds) 48006E0CIO0-18001600140012001000/000000800600J012002400 3600lime (seconds) 448006Figure 2.6.1-2 Inadvertent ECCS -Pressurizer Pressure and Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-293WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-293I3:0CDU-)00.80,6-0.4-0.2-012002400 3600Time (seconds) 48006000500400300-200100-I-~~~ I II0012002400Ti roe3600(seconds) 48006000Figure 2.6.1-3Inadvertent ECCS -Total Steam Flow and Total Flow Injected to the RCSversus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-2942.6.2 Chemical and Volume Control System Malfunction that Increases Reactor CoolantInventory (USAR Section 15.5.2)2.6.2.1 Technical Evaluation 2.6.2.1.1 Introduction Increases in reactor coolant inventory caused by a malfunction of the CVCS may be postulated to resultfrom operator error or a false electrical signal. The transients examined in this section are characterized by increasing pressurizer level, increasing pressurizer
: pressure, and maintaining a constant boronconcentration.
The transients analyzed in this section are done to demonstrate that there is adequate timefor the operator to take corrective action to prevent filling the pressurizer.
An increase in reactor coolantinventory, which results from the addition of cold, unborated water to the RCS, is analyzed inSection 2.5.5, "Chemical and Volume Control System Malfunction That Results in a Decrease in BoronConcentration in the Reactor Coolant (USAR Section 15.4.6)."
The most limiting case occurs if the charging system is in automatic control and the pressurizer levelchannel being used for charging control fails in a low direction.
This causes the maximum charging flowto be delivered to the RCS and letdown flow to be isolated.
The worst single failure for this event is asecond pressurizer level channel failing in an as-is condition or a low condition.
This defeats the RT ontwo-out-of-three high pressurizer level channels.
To prevent filling the pressurizer the operator must berelied upon to terminate charging flow.2.6.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The following inputs and assumptions were applied in the analysis of the CVCS Malfunction event:The initial NSSS power is 3651 MWt, which includes all applicable uncertainties.
A full power Tavg range of 570.7'F to 588.4'F was considered in the analysis.
The limiting initialTavg is 564.2'F, which corresponds to the low nominal full power Tavg minus uncertainties (including bias). The lower initial temperature corresponds to a higher reactor coolant mass,which leads to a more severe pressurizer water volume transient.
The initial Tfeed is 448.6&deg;F, which corresponds to the high end of the full power Tfeed range(400.0&deg;F to 448.6'F).
The initial pressurizer water level is 46 percent level span, which is the nominal pressurizer waterlevel of 41 percent span at the low full power Tavg of 570.7&deg;F plus 5 percent span uncertainty.
The initial pressurizer pressure is 2200 psia, which is the nominal value of 2250 psia minus 50 psiuncertainty.
A lower initial RCS pressure is conservative because it allows higher charging flowsto be injected into the RCS.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-295* The pressurizer heaters are modeled to function because their operation generates a more limitingcondition with respect to filling the pressurizer.
* Cases were analyzed both with and without automatic pressurizer spray modeled.* A maximum SGTP level of 10 percent was modeled.a The flow injected to the RCS corresponds to maximum flow from one CCP.* No RT at event initiation.
* Cases were analyzed with both maximum and minimum reactivity feedback conditions.
0 Cases were analyzed both with and without automatic rod control.Acceptance CriteriaBased on the frequency of occurrence, the CVCS Malfunction event is considered to be a Condition IIevent as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized WaterReactor Plants,"
ANSI N 18.2-1973.
The following items summarize the acceptance criteria associated with the analysis of this event:Fuel cladding integrity is maintained by ensuring that the minimum DNBR remains above the95/95 DNBR SAL.Based on historical precedence, the CVCS Malfunction event does not lead to a serious challenge of the DNB design basis. The conditions do not approach the core thermal DNB limits, as thecore power, RCS pressure, and RCS temperatures remain relatively unchanged.
Therefore, theDNBR typically increases and does not approach the DNBR SAL following event initiation.
Assuch, no explicit analysis of the event was performed to calculate a minimum DNBR value.Pressures in the RCS and MSS are maintained below 110 percent of the design pressures.
With respect to the overpressure evaluation, the CVCS Malfunction event is bounded by theLOL/TT event, discussed in Section 2.3.1, in which assumptions are made to conservatively maximize the RCS and MSS pressure transients.
For this event, a turbine trip would occurfollowing a RT, whereas for the LOL/TT event, the turbine trip is the initiating fault. Therefore, the primary-to-secondary power mismatch and resultant RCS and MSS heatup and pressurization transients are always more severe for the LOL/TT event. For this reason, it is not necessary tocalculate the maximum RCS or MSS pressures for the CVCS Malfunction event.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-296An incident of moderate frequency does not generate a more serious plant condition without otherfaults occurring independently.
The major concern from a CVCS Malfunction event is that associated with pressurizer filling.The pressurizer water volume increases for this event as a result of the flow injected to the RCS.This event is analyzed to demonstrate that sufficient time is available for the appropriate operatoractions to be taken to preclude a pressurizer water-solid condition.
The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
Brief discussions of thespecific GDCs that are related to the CVCS Malfunction acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated
: coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.
For the CVCV Malfunction event, this is shown to be met by demonstrating that thefuel cladding integrity is maintained.
GDC 15 (RCS Design) requires that the RCS and associated auxiliary,
: control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.
For theCVCS Malfunction event, this is shown to be met by demonstrating that the peak RCS pressure isless than 110 percent of the design pressure.
GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.
For the CVCS Malfunction event, this is shown to be met by demonstrating that the fuel cladding integrity is maintained.
2.6.2.1.3 Description of Analyses and Evaluations The CVCS Malfunction event was analyzed using the RETRAN computer code (Reference 1). TheRETRAN model simulates the RCS, neutron kinetics, pressurizer, pressurizer relief and safety valves,pressurizer
: heaters, pressurizer spray, SI system, SGs, FW system, and MSSVs. The code computespertinent plant variables including nuclear power, reactor coolant average temperature, RCS pressure, pressurizer water volume, and SG pressure.
2.6.2.1.4 ResultsIn all cases analyzed, the core power and RCS temperatures remain relatively constant.
Cases both withand without automatic rod control were examined.
Because there was little or no change in core powerand RCS average temperatures, the results showed that automatic control has no effect on the cases thatmodel maximum reactivity feedback conditions and a relatively negligible effect on the cases that modelminimum reactivity feedback conditions.
Figures 2.6.2-1 through 2.6.2-8 show the transient responses forWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-297the cases with automatic rod control modeled.
The calculated sequences of events for these cases arepresented in Table 2.6.2-1.The cases that model maximum reactivity feedback show that the pressurizer level increases at arelatively constant rate; whereas, the cases that model minimum reactivity feedback show that thepressurizer level increases at a somewhat varying rate. This is because the reactivity feedback in themaximum feedback cases is of a magnitude that it is able to maintain Tavg within the temperature deadband for the automatic rod control system, and no rod movement is predicted in these cases.However, rod movement is predicted in the minimum feedback cases, resulting in slight variations in Tavgand ultimately in the pressurizer level increase in these cases.The pressurizer level rate of increase is slightly faster in the cases where the pressurizer spray is modeledoperable, as compared to the cases in which the pressurizer sprays are modeled inoperable, because sprayactuation tends to keep the RCS pressure lower for several minutes, which allows the charging pumps todeliver more flow to the RCS. However, pressurizer pressure does eventually increase enough to open therelief valves in the cases with the pressurizer spray modeled operable.
The limiting case, shown in Figures 2.6.2-5 and 2.6.2-6, models minimum reactivity feedback conditions and the pressurizer sprays operable.
In this case, the pressurizer high level alarm is reached inapproximately 8.8 minutes and the pressurizer reaches a water-solid condition at approximately 17.3 minutes.
This allows the operators 8.5 minutes from the time the pressurizer high level alarm isreached to terminate normal charging flow before pressurizer filling occurs. Thus, with respect to thecriterion of precluding the generation of a more serious plant condition, there is sufficient time for theoperators (more than 8 minutes) to respond to the event and terminate the reactor coolant inventory addition.
2.6.2.2 Conclusion Based on the above information, it is concluded that the CVCS Malfunction event will not progress into amore serious plant condition.
Thus, all applicable event acceptance criteria are satisfied.
It has beendemonstrated that the reactor protection and safety systems ensure that the specified acceptable fueldesign limits are met and the RCPB pressure limits will not be exceeded as a result of the CVCSMalfunction event. Based on this, the plant will continue to meet the requirements of GDCs 10, 15,and 26.2.6.2.3 References
: 1. WCAP-14882
-P-A "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-298Table 2.6.2-1 Time Sequence of Events -CVCS Malfunction TimeCase Event Seconds MinutesMaximum Reactivity Two Pressurizer Level Channels Fail Low 0.0 0.0Feedback, with pressurizer Maximum Charging Flow from One CCP Begins 0.0 0.0sprayLetdown is Isolated 0.0 0.0Low-Low Pressurizer Level Alarm 0.0 0.0High Pressurizer Level Alarm 514.6 8.6Pressurizer Fills 1049.5 17.5Pressurizer Relief Valve Setpoint Reached 1054.3 17.6End of Transient 1800.0 30.0Maximum Reactivity Two Pressurizer Level Channels Fail Low 0.0 0.0Feedback, without Maximum Charging Flow from One CCP Begins 0.0 0.0pressurizer sprayLetdown is Isolated 0.0 0.0Low-Low Pressurizer Level Alarm 0.0 0.0Pressurizer Relief Valve Setpoint Reached 15.4 0.3High Pressurizer Level Alarm 609.1 10.2Pressurizer Fills 1365.7 22.8End of Transient 1800.0 30.0Minimum Reactivity Two Pressurizer Level Channels Fail Low 0.0 0.0Feedback, with pressurizer Maximum Charging Flow from One CCP Begins 0.0 0.0sprayLetdown is Isolated 0.0 0.0Low-Low Pressurizer Level Alarm 0.0 0.0High Pressurizer Level Alarm 529.3 8.8Pressurizer Fills 1036.4 17.3Pressurizer Relief Valve Setpoint Reached 1239.8 20.7End of Transient 1800.0 30.0Minimum Reactivity Two Pressurizer Level Channels Fail Low 0.0 0.0Feedback, without Maximum Charging Flow from One CCP Begins 0.0 0.0pressurizer sprayLetdown is Isolated 0.0 0.0Low-Low Pressurizer Level Alarm 0.0 0.0Pressurizer Relief Valve Setpoint Reached 24.5 0.4High Pressurizer Level Alarm 597.6 10.0Pressurizer Fills 1345.7 22.4End of Transient 1800.0 30.0WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-299WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-299c.)0C-_r)1.21-0.80.6-0.4-0.2-...........
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I I I I......................................................................I I I Ino4U05001000Time (s)15002000C ^I.-a)_ECi.cvUuu590-580-570"560"550-C A AJI'U0500I1000Time (s)15002000Figure 2.6.2-1CVCS Malfunction, Maximum Reactivity
: Feedback, With Pressurizer SprayNuclear Power and Tavg versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-300WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-300zltUU.- 2300"2200U)cn 2100",- 2000-NJ1900U 1S1800-...............
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..........I170005500I1000Time (s)115002000E0U)L.I0 500 1000 1500Time (s)2000Figure 2.6.2-2CVCS Malfunction, Maximum Reactivity
: Feedback, With Pressurizer SprayPressurizer Pressure and Water Volume versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3011.24-I0~75~C-3I0.8-0.6-0.4-0.2-...........
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I I I I.................................................U05;01000Time (s)15002000C-EC,,DUU590-580-570-560-550-r AA-....................................................0500I1000Time (s)115002000Figure 2.6.2-3CVCS Malfunction, Maximum Reactivity
: Feedback, Without Pressurizer SprayNuclear Power and Tavg versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-302n~A nCnC,,0~CtDEnC-VtUU2300-2200-2100-2000-1900-1800-1700.......................................................................
...................... .................................................. ............................................... ....................' ............I05500I1000Time (s)1150020000 500 1000 1500Time (s)2000Figure 2.6.2-4CVCS Malfunction, Maximum Reactivity
: Feedback, Without Pressurizer SprayPressurizer Pressure and Water Volume versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-303WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-303.4-a,00~C-)1.21-0.8-0.6-0.4-0.2-.....................................................
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A -, I I I i .I IU0500I1000Time (s)15002000CAA-4Q_EUVU590-580-570-560-550-540.........................................................
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0500I1000Time (s)15002000Figure 2.6.2-5CVCS Malfunction, Minimum Reactivity
: Feedback, With Pressurizer SprayNuclear Power and Tavg versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-304WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 040n(nC/)(nC/)C/)(DCLL&#xb6;4UU2300-..............
nnw% _& ................/ZUU2100-2000-1900-1800-1700..............
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05500I1000Time (s)1150020000 500 1000 1500Time (s)2000Figure 2.6.2-6CVCS Malfunction, Minimum Reactivity
: Feedback, With Pressurizer SprayPressurizer Pressure and Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-305WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3054-&#xfd;00~0-1.21 -0.80.6"0.4-0.2-...........................................................
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L&#xfd;Afl.r''u05001000Time (s)15002000Figure 2.6.2-7CVCS Malfunction, Minimum Reactivity
: Feedback, Without Pressurizer SprayNuclear Power and T.vg versus TimeWCAP-1 7658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-306WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-306C,)0na,C/)L)E,0~0-24002 3 0 0 ...................................... ...............23002200 ...2 0 ............................................................2100-2 0 0 0 ..................................................2000 ........
......18 0 0 .................................................1900,1800-05001000Time (s)150020000 500 1000 1500Time (s)2000Figure 2.6.2-8CVCS Malfunction, Minimum Reactivity
: Feedback, Without Pressurizer SprayPressurizer Pressure and Water Volume versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3072.7 DECREASE IN REACTOR COOLANT INVENTORY 2.7.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve (USAR Section 15.6.1)2.7.1.1 Technical Evaluation 2.7.1.1.1 Introduction An accidental depressurization of the RCS could occur as a result of an inadvertent opening of apressurizer relief or spray valve. To conservatively bound this scenario, the Westinghouse methodology models the failure of a PSV, because a PSV is sized to relieve approximately twice the steam flow of apressurizer PORV and thus results in a much more rapid depressurization upon opening.
Thedepressurization resulting from an open PSV is also much more rapid than would occur from theaccidental actuation of pressurizer spray. Therefore, the failure of a PSV yields the most severe coreconditions resulting from an accidental depressurization of the RCS. It should be noted that a stuck-open PSV is not an event of moderate frequency (i.e., Condition II event) such as a control system failurewould be. A stuck-open PSV is considered to be a SBLOCA (i.e., Condition III event) during which theRCS cannot be isolated, whereas the failure of a PORV can be overridden by the closure of the PORVblock valve. The results of this analysis are shown to comply with the more restrictive Condition IIacceptance criterion of ensuring that the DNB design basis is met.Initially, the event results in a rapidly decreasing RCS pressure, which could reach HL saturation conditions without reactor protection system intervention.
If saturated conditions are reached, the rate ofdepressurization is slowed considerably.
: However, the pressure continues to decrease throughout theevent. The power remains essentially constant throughout the initial stages of the transient.
The reactor may be tripped by the following RTS signals:* OTAT* Pressurizer low pressure2.7.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaTo produce conservative results in calculating the DNBR during the transient, the following assumptions were made:The accident was analyzed using the RTDP (Reference 1). Pressurizer pressure and RCStemperature were assumed to be at their nominal values, consistent with steady-state full-power operation.
Reactor coolant minimum measured flow was modeled.
Uncertainties in initialconditions were included in the DNBR SAL as described in Reference
: 1. The event isconservatively analyzed at an initial NSSS power level of 3651 MWt, which includes nominalRCP net heat input; no additional uncertainty on core power is modeled.A zero moderator coefficient of reactivity was assumed.
This is conservative for BOL operation in order to provide a conservatively low amount of negative reactivity feedback due to changes inmoderator temperature.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-308A small (absolute value) Doppler coefficient of reactivity is assumed, such that the resultant amount of negative feedback is conservatively low in order to maximize any power increase dueto moderator feedback.
The spatial effect of voids resulting from local or subcooled boiling was not considered in theanalysis with respect to reactivity feedback or core power shape. In fact, it should be noted thatthe power peaking factors were kept constant at their design values, while the void formation andresulting core feedback effects would result in considerable flattening of the power distribution.
Although this would increase the calculated DNBR, no credit was taken for this effect.The analysis performed assumes that the rod control system is in automatic.
: However, no rodmotion occurs during the transient because the conditions do not change enough to demand anyrod motion from the rod control system. Therefore, the transient results are identical with orwithout automatic rod control.Based on its frequency of occurrence, the accidental depressurization of the RCS accident is considered tobe a Condition II event as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"
ANSI N 18.2-1973.
The following items summarize the acceptance criteria associated with this event:The critical heat flux should not be exceeded.
This criterion was met by demonstrating that theminimum DNBR does not go below the limit value at any time during the transient.
Pressure in the RCS and MSS should be maintained below 110 percent of the design pressures.
Note that because this event is a depressurization event, these limits are not challenged.
Bothprimary and secondary pressures decrease for the entire duration of the event.As discussed above, the accidental depressurization of the RCS event has historically beenanalyzed to show that the minimum DNBR limit is not exceeded.
: However, during the licensing pre-application meetings between Westinghouse.,
WCNOC, and the USNRC, the USNRCrequested that the potential for pressurizer filling during the event, due to the actuation of the SIsystem, be considered.
Consistent with that request, an additional sensitivity was performed toshow that the pressurizer would not overfill such that the transient would not transition to a moreserious plant condition.
2.7.1.1.3 Description of Analyses and Evaluations The purpose of this analysis was to demonstrate that the RTS functions and mitigates the consequences ofthe RCS depressurization event. This analysis is concerned with the transient from initiation through justpast the time of RT. With respect to long-term post-accident
: recovery, it is assumed that operators followapproved plant procedures to bring the plant to a safe post-accident condition.
The accident was analyzed by using the detailed digital computer code RETRAN (Reference 2). Thiscode simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, SG, and SG safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3092.7.1.1.4 ResultsThe system response to an inadvertent opening of a PSV is shown in Figures 2.7.1-1 through 2.7.1-4.Figure 2.7.1-1 illustrates thenuclear power transient.
Nuclear power remains essentially unchanged untilthe RT occurs on OTAT. The pressurizer pressure transient is illustrated in Figure 2.7.1-2.
Pressuredecreases continuously throughout the transient.
: However, pressure decreases more rapidly after core heatgeneration is reduced via the RT. Figure 2.7.1-3 shows the loop average temperature transient.
The loopaverage temperature decreases slowly until the RT occurs. The DNBR decreases initially, but increases rapidly following the RT as demonstrated in Figure 2.7.1-4.
The DNBR remains above the SALthroughout the transient.
The calculated sequence of events is shown in Table 2.7.1-1.
The calculated minimum DNBR value isprovided in Table 2.7.1-2.The results of the analysis show that the OTAT RTS function provides adequate protection against theRCS depressurization event because the minimum DNBR remains above the SAL throughout thetransient.
Therefore, no cladding damage or release of fission products to the RCS is predicted for thisevent.With regards to overfill, the WCGS has a pressurizer PORV interlock that is set to 2185 psig. When thepressurizer pressure reaches the PORV interlock
: setpoint, the PORV block valves are closed. A detailedRETRAN analysis was performed to demonstrate that the PORV block valves close prior to thepressurizer pressure reaching the Pressurizer Pressure
-Low, S1 setpoint.
This will prevent the Si systemfrom actuating; without the addition of SI, pressurizer filling does not occur. The RETRAN analysisconservatively modeled signal processing time, valve stroke time, and instrument uncertainties to increasethe likelihood of SI actuation.
2.7.1.2 Conclusion The RCS depressurization analysis demonstrates that for this event at WCGS, the DNBR does notdecrease below the SAL value at any time. The event does not challenge the primary and secondary sidepressure limits because this is a depressurization event. Thus, all applicable acceptance criteria for thisevent are met for WCGS operating at a nominal NSSS power of up to 3651 MWt.2.7.1.3 References
: 1. WCAP-11397-P-A, "Revised Thermal Design Procedure,"
April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-310Table 2.7.1-1 Time Sequence of Events -Accidental Depressurization of the RCSEvent Time (seconds)
PSV opens fully 0.0OTAT RT setpoint reached 21.4Rods begin to drop 24.4Minimum DNBR 25.0Table 2.7.1-2 Results -Accidental Depressurization of the RCSMinimum Calculated DNBR DNBR SAL2.001 1.52WCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-31100CD0 10 20 30 40Time (seconds)
Figure 2.7.1-1 RCS Depressurization
-Nuclear Power versus Time50C)C-U-C)ELC)K-_C)K-_EL_0 10 20 30 40Time (seconds)
Figure 2.7.1-2 RCS Depressurization
-Pressurizer Pressure versus Time50WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-312WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 12590c 585D 580EW 575U')-5700o 5650560555I I I I I I I I I I I I I I I I I I I I01020Time30(seconds) 4050Figure 2.7.1-3 RCS Depressurization
-Indicated Loop Average Temperature versus Time4.5.4.03.5-= 3.0-r --0 10 2O 30 40 50Time (seconds)
Figure 2.7.1-4 RCS Depressurization
-DNBR versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3132.7.2 Steam Generator Tube Rupture Margin to Overfill (USAR Section 15.6.3)2.7.2.1 Technical Evaluation 2.7.2.1.1 Introduction The major hazard associated with a SGTR event is the radiological consequences resulting from thetransfer of radioactive primary coolant to the secondary side of the ruptured SG and subsequent release ofradioactivity to the atmosphere.
One major concern for an SGTR is the possibility of ruptured SGoverfills because this could potentially result in a significant increase in the radiological consequences.
Therefore, an analysis was performed to demonstrate that the ruptured SG does not overfill and releasewater from the main steam relief valves, assuming the limiting single failure relative to overfill.
Theanalysis confirmed that water releases through the SG safety valves did not occur.The SGTR margin to overfill transient analysis was performed using the RETRAN computer program(Reference
: 1) following the methodology developed in WCAP-10698-P-A and its Supplement I(References 2 and 3). Modifications were made to address NSAL-07-11 (Reference 4), which identified apotential non-conservative assumption.
This regards the direction of conservatism for decay heat in theReference 2 methodology for demonstrating margin to overfill.
The plant response to the SGTR was modeled using conservative assumptions of break size and location, condenser availability, and initial secondary water mass. The analyses include the simulation of theoperator actions for recovery from an SGTR based on the WCGS Emergency Operating Procedures (EOPs), which are based on the Westinghouse Owners Group Emergency ResponseGuidelines.
The SGTR margin to overfill analysis was performed for the time period from the SGTR until the primaryand secondary pressures equalized (break flow termination).
In the ruptured SG secondary side, the watervolume was calculated as a function of time to demonstrate that overfill did not occur.The SGTR margin to overfill analysis supports operation at a core power up to 3637 MWt. The analysissupports a full power RCS Tavg operating range from 575.00 to 588.4&deg;F, and a main Tfeed range from 4000to 448.6'F, with up to 10 percent of the SG tubes plugged.2.7.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe margin to overfill analysis modeled the plant operating at the lower end of the Tavg range. A loweroperating temperature results in a higher mass flow rate through the broken tube and less steam releasedfrom the ruptured SG. The analysis assumed that the plant was operating with the T'eed at the low end ofthe temperature range. This results in a higher mass of water in the SG at the start of the event, whichlimits the amount of break flow and AFW that can accumulate in the ruptured SG without forcing waterinto the steam lines. The maximum SGTP was modeled because it reduces heat transfer to the rupturedSG, which reduces the mass released by steaming, which in turn reduces margin to overfill.
The reducedheat transfer also prolongs the cooldown period, leading to delayed break flow termination.
Sensitivity runs were made to confirm the conservative nature of these plant operating assumptions.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3142.7.2.1.2.1 Design Basis AccidentThe accident modeled was a double-ended break of one SG tube located at the top of the tube sheet on theoutlet (CL) side of the SG. The location of the break on the cold side of the SG results in higher primaryto secondary break flow than a break on the hot side of the SG. It was also assumed that a LOOP occurs atthe time of RT, and the highest worth control rod assembly was assumed to be stuck in its fully withdrawn position at RT.2.7.2.1.2.2 Single Failure Considerations An evaluation was performed to determine the limiting single failure with respect to margin to SG overfillfor an SGTR. To identify the limiting single failure, sensitivity runs were performed considering thefollowing failures:
Failure of an Intact SG ARV or Failure of Multiple SG ARVsThis scenario considered the failure of an ARV to open on one of the intact SGs when theoperator performed the RCS cooldown.
Because offsite power was assumed to be lost at RT forthe SGTR analyses, the SG ARVs were relied upon to cool the RCS. Failure of an ARV on anintact SG to open on demand reduced the steam release capability provided by the ARVs becauseonly two intact SG ARVs are available for the cooldown.
This increased the time required for thecooldown, resulting in increased break flow.A single failure that results in the failure of multiple SG ARVs does not exist in the WCGSdesign. Each SG ARV can be actuated by an independent safety-related compressed gas supply. Afailure of any one of the four compressed gas supplies would only affect the associated SG ARV,and would not affect the other three SG ARVs.* Failure of the MDAFW Control ValveThis scenario considered the failure of the MDAFW control valve to isolate MDAFW flow to theruptured SG when the TDAFW flow is isolated.
This required additional operator action tomanually isolate the MDAFW flow, resulting in increased AFW flow to the ruptured SG. Thus,the mass in the ruptured SG increased in relation to the intact SGs prior to RT. Although thisadditional mass would be expected to provide early identification and isolation of AFW flow tothe ruptured SG following RT, no reduction in the operator action time for AFW isolation wascredited.
The initial secondary SG water mass was not increased to account for the impact ofturbine runback.
This modeling is consistent with the AFW flow control valve failure presented inReference
: 2. The cooldown was performed using all three of the ARVs on the intact SGs.The MDAFW control valve failure was determined to be the limiting single failure.
The penalty from thedelay to terminate AFW flow to the ruptured SG that resulted from the AFW control valve failure resultedin the largest secondary side inventory.
The effects of adding more inventory to the ruptured SG throughlonger AFW flow duration offset the effects of the other intact SG ARV failure, which prolonged cooldown and break flow termination.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3152.7.2.1.2.3 Conservative Assumptions Plant responses until break flow termination were calculated using the RETRAN computer code. Theconservative conditions and assumptions used in Reference 2 were also used in the analysis to determine margin to SG overfill with the exception of the following differences:
Turbine RunbackThe mass increase due to turbine runback is the mass corresponding to power at the end ofturbine runback minus the mass at 100 percent power. A power reduction of 10%/min for turbinerunback is assumed, where the turbine runback duration is the smaller value of RT time or3 minutes.
The reduction in power that would result from the turbine runback during that periodis used to develop a secondary mass penalty associated with the runback.SG Secondary MassA higher initial secondary water mass in the ruptured SG was determined by Reference 2 to beconservative for overfill.
The increase in mass that would result from a turbine runback to a lowerpower (discussed in the prior item) and the consideration of mass uncertainties are added to theinitial secondary water mass.Intact SG Target Pressures The intact SG target pressures are calculated based on the target temperature and the intact loopAT at the time cooldown is terminated instead of at the start of cooldown.
This exception isjustified because it is consistent with the WCGS EOPs.AFW Isolation Based Solely on SG LevelThe analysis modeled AFW flow isolation based on ruptured SG level with no consideration of atime component.
The SG level for AFW isolation from the WCGS EOPs is 6 percent NRS. (Notethat the analysis used a conservative SG level of 15 percent NRS with the stipulation that theaction not be taken before 2 minutes after RT.) This exception is justified because it is a morerealistic modeling of the operator response to an SGTR accident.
Decay Heat and NSAL-07-11 NSAL-07-11 (Reference
: 4) identifies a potential non-conservative assumption regarding thedirection of conservatism for decay heat in the Reference 2 methodology for evaluating margin tooverfill.
For the margin to overfill
: analysis, higher decay heat yields a benefit by increasing steamreleases from the ruptured SG, but results in a penalty from a longer cooldown and aconservatively delayed break flow termination.
Conversely, lower decay heat yields a penalty byreducing steam releases from the ruptured SG, but results in a benefit from a shorter cooldownand earlier break flow termination.
Similar impacts were identified for the AFW and SI flowenthalpies.
The relative importance of these competing effects is plant-specific, and plant-specific analyses are required to determine the conservative assumption.
Plant-specific sensitivities WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-316performed for WCGS showed the following to be conservative with respect to margin to overfillfor the limiting cases:-1979-2a ANS decay heat was conservative compared to the 1971+20%
ANS decay heatmodel specified in Reference
: 2. For this analysis, the 1979 ANS decay heat model minus2a uncertainty was used.-Minimum AFW enthalpy was conservative compared to the maximum AFW enthalpyspecified by Reference
: 2. For this analysis, the minimum AFW enthalpy of 18.1 Btu/lbmwas modeled.-Minimum SI enthalpy was conservative compared to the maximum SI enthalpy modeled inReference
: 2. For this analysis, the minimum SI enthalpy of 11.53 Btu!lbm was modeled.2.7.2.1.2.4 Plant InputThe following significant WCGS input was used in the analysis:
: 1. SG ARVIt was assumed that a LOOP occurs at RT for the SGTR analyses, and thus the SG ARVs open tolimit the secondary pressure.
The ARV pressure setpoint is 1139.7 psia (1125 psig). The ARVcapacity modeled in the analysis is 594,642 lbm/hr/valve at a reference pressure of 1107 psia(1092.3 psig).2. Pressurizer PORV CapacityIt was assumed that a LOOP occurs at RT for the SGTR analyses, and thus the pressurizer PORVwas relied upon to depressurize the RCS. The capacity of 210,000 Ibm/hr at 2350 psia was usedin the analysis.
: 3. AFW System Operation and Associated Single Failure Considerations The WCGS AFW system consists of two MDAFW pumps and one TDAFW pump. EachMDAFW pump normally feeds two SGs and the TDAFW pump feeds all four SGs. There is acontrol valve in the flow path from the MDAFW pump to each SG and a control valve in the flowpath from the TDAFW pump to each SG. The control valves in the MDAFW pump andTDAFW pump flow paths are used to control the inventory in the SGs, and are closed to isolateAFW flow to the ruptured SG in accordance with WCGS Emergency Mitigation Guideline E-3for SGTR recovery.
The control valves for the MDAFW pumps are controlled to throttle the flowas required to maintain the level in the associated SG between 29 percent and 50 percent NRS,and can also be manually controlled by the operator to adjust flow to the SGs. Also, the automatic level control function of the MDAFW pump control valves is not credited to reduce the liquidinventory in the ruptured SG. The control valves for the TDAFW pump have no automatic controlfeatures.
They are manually throttled to adjust flow to maintain the desired level in the SGs.WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-317The AFW flow rates for WCGS are dependent on the number of MDAFW and TDAFW pumpsthat are operating as well as the SG pressure.
Flow to an SG from one or more pumps may bethrottled or isolated depending on the time in the transient progression, the level in the SG, andthe single failure being considered.
All AFW pumps are assumed to be operating following RTand LOOP. The operators will isolate flow from the TDAFW pump and then isolate flow from theMDAFW pump to the ruptured SG. The single failures are discussed in detail later in this report.The associated AFW flows are outlined below.a. MDAFW Failure AFW FlowsThe MDAFW failure is the failure of the MDAFW control valve to throttle or isolate ondemand by the control system or by manual operator action. This failure would causeadditional AFW flow to be delivered to the ruptured SG until the associated MDAFWpump is stopped.
This terminates all AFW flow to the ruptured SG and one intact SG, andleaves the MDAFW pumps providing flow to two intact SGs. It is assumed that the AFWflow to the third intact SG is restored by the start of the cooldown.
The AFW flow rates to the ruptured and intact SGs for the MDAFW failure are shown inTables 2.7.2-1 through 2.7.2-3.b. Intact SG ARV Failure AFW FlowsThe intact SG ARV failure is the failure of an ARV on one intact SG to open for cooldown.
This failure has no impact on the AFW flows and the AFW system. A constant AFW flowof 320 gpm is used throughout the event.4. SI FlowsThe maximum SI flow was assumed to be initiated at the low pressurizer pressure setpoint of2004.7 psia. The flow rates are presented in Table 2.7.2-4.2.7.2.1.2.5 Operator Action TimesIn the event of an SGTR, the operator is required to take actions to stabilize the plant and terminate theprimary to secondary break flow. The operator actions for SGTR recovery are provided in the WCGSEOPs, and major actions were explicitly modeled in these analyses.
The operator actions modeled includeisolation of the ruptured SG, cooldown of the RCS, depressurization of the RCS to restore inventory, andtermination of SI to stop primary to secondary break flow. These operator actions are described below.1. Identify the Ruptured SGHigh secondary side activity, as indicated by the main steamline radiation monitor (or othersecondary monitors) or high SG sample activity typically will provide the first indication of anSGTR event. The ruptured SG can be identified by a mismatch between steam and FW flows,high activity in an SG water sample, or a high radiation indication on the corresponding mainsteamline radiation monitor.
For an SGTR that results in a RT at high power as assumed in theseWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-318analyses, the SG water level, as indicated on the narrow range, will decrease significantly for allof the SGs. The AFW flow will begin to refill the SGs, distributing approximately equal flow toeach of the SGs. Because primary to secondary break flow adds additional inventory to theruptured SG, the water level will increase more rapidly in that SG. This response, as displayed bythe SG water level instrumentation, provides confirmation of an SGTR event and also identifies the ruptured SG.2. Isolate the Ruptured SGOnce the ruptured SG has been identified, recovery actions begin by isolating AFW flow to theruptured SG and closing the MSIV on the ruptured SG steamline.
In addition to minimizing radiological
: releases, this also reduces the possibility of filling the ruptured SG by minimizing theaccumulation of AFW. The operator can also establish a pressure differential between theruptured and intact SGs, which is a necessary step toward terminating primary to secondary breakflow. In the WCGS EOPs for SGTR, the operator is directed to verify that the level in theruptured SG is greater than a specified level on the NRS prior to isolating AFW. The requiredlevel is 6 percent NRS. (Note that the analysis used a conservative SG level of 15 percent NRSwith the stipulation that the action not be taken before 2 minutes after RT.) For the single failureinvolving the failure of MDAFW control valves, an additional 30 seconds is added after thenarrow range level is reached until MDAFW isolation.
For the single failure involving the failureof the SG ARV to open, the TDAFW and MDAFW are isolated simultaneously once the narrowrange level is reached.
All SG MSIVs were assumed to be closed at 8 minutes from RT for bothsingle failure scenarios.
: 3. Cooldown the RCS using the Intact SGsAfter isolation of the ruptured SG MSIV, dumping steam from only the intact SGs cools the RCSas rapidly as possible to less than the saturation temperature corresponding to the rupturedSG pressure.
This ensures adequate subcooling in the RCS after depressurization to the rupturedSG pressure in subsequent actions.
If offsite power is available, the normal steam dump system tothe condenser can be used to perform this cooldown.
If offsite power is lost, the RCS would becooled using the ARVs on the intact SGs. The availability of relief valves for cooldown isdependent on the single failure assumption being modeled (See Section 2.7.2.1.2.5).
The analysis assumed 23 minutes elapse from the time of RT until the cooldown was initiated viathe ARVs. The cooldown is terminated when the required core exit temperature for cooldowntermination (without adverse environment) corresponding to the ruptured SG pressure is reached.The temperature is identified in the WCGS EOPs for SGTR. The ARVs on the intact SGs werethen used as necessary to maintain that temperature.
: 4. Depressurize the RCS to Restore Inventory When the cooldown is completed, SI flow will tend to increase RCS pressure until break flowmatches SI flow. Consequently, SI flow must be terminated to stop primary to secondary breakflow. However, adequate inventory must first be assured.
This includes both sufficient RCSsubcooling and pressurizer inventory to maintain a reliable pressurizer level indication after SIWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-319flow is stopped.
Because break flow from the primary side will continue after SI flow is stoppeduntil RCS and ruptured SG pressures
: equalize, an excess amount of inventory is needed to ensurethat the pressurizer level remains on span. The excess amount required depends on the RCSpressure, and reduces to zero when the RCS pressure equals the pressure in the ruptured SG.The analyses assumed that 5 minutes elapsed from the time the cooldown was terminated untilthe depressurization was initiated.
The RCS depressurization is performed using normalpressurizer spray if the RCPs are running.
Because offsite power was assumed to be lost at thetime of RT, the RCPs were not running and thus normal pressurizer spray was not available.
Therefore, the depressurization was modeled using a pressurizer PORV.The RCS depressurization is continued until any of the following three conditions in the WCGSEOPs for SGTR (using setpoints without adverse environment) are satisfied:
RCS pressure is lessthan the ruptured SG pressure and pressurizer level is greater than 6 percent, pressurizer level isgreater than 75 percent, or RCS subcooling is less than required to address the subcooling uncertainty.
: 5. Terminate SI to Stop Primary to Secondary Break FlowThe previous actions will have established adequate RCS subcooling, a secondary side heat sink,and sufficient RCS inventory to ensure that the SI flow is no longer needed. When these actionshave been completed, the SI flow must be stopped to terminate primary to secondary break flow.The analyses assumed that 4 minutes elapsed from the time the depressurization was terminated until SI could be stopped.
SI can be stopped provided the following conditions in the WCGSEOPs for SGTR (using setpoints without adverse environment) are satisfied:
RCS pressure isstable or rising, pressurizer level is greater than 6 percent, RCS subcooling is greater thanrequired to address the subcooling uncertainty, and a secondary heat sink is confirmed.
After SI termination, the analyses do not model specific actions leading to break flowtermination, consistent with the Reference 2 method. The primary to secondary break flowcontinues after the SI flow is stopped until the RCS and ruptured SG pressures equalize.
The total time required to complete the recovery operations consists of both operator action time andsystem, or plant, response time. For instance, the time for each of the major recovery operations (i.e., RCScooldown) is primarily due to the time required for the system response, whereas the operator action timeis reflected by the time required for the operator to perform the intermediate action steps.The operator action times to isolate AFW flow to the ruptured SG, to isolate the MSIV on the rupturedSG, to initiate RCS cooldown, to initiate RCS depressurization, and to terminate SI were developed forthe design basis analyses.
WCGS has determined the corresponding operator action times to performthese operations.
The operator actions and the corresponding operator action times used for the analysesare summarized in Table 2.7.2-5.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3202.7.2.1.2.6 Acceptance CriteriaThe analyses were performed to demonstrate that the secondary side of the ruptured SG did notcompletely fill with water. The available secondary side volume of a single SG is 5852 ft3.Margin tooverfill is demonstrated, provided the transient-calculated SG secondary side water volume is less thanthis value. No credit is taken for the volume of the nozzle or any steam piping.2.7.2.1.3 Description of Analyses and Evaluations The RETRAN analysis for the limiting margin to overfill case is described below. The limiting case withrespect to margin to SG overfill considered operation at the minimum operating temperature (575.0&deg;F),
with the minimum main FW temperature (400.07F),
the maximum SGTP level (10 percent),
low AFWenthalpy, low SI enthalpy, low decay heat (1979-2a),
and the failure of the MDAFW control valve toclose. The sequences of events for these transients are presented in Table 2.7.2-6.Following the tube rupture, water flowed from the primary into the secondary side of the ruptured SGbecause the primary pressure is greater than the SG pressure.
In response to this loss of coolant,pressurizer level decreased (Figure 2.7.2-1).
The RCS pressure (represented by the pressurizer pressure) also decreased (Figure 2.7.2-2) as the steam bubble in the pressurizer expanded.
As the RCS pressuredecreased due to the continued primary to secondary break flow, an automatic RT occurred on an OTATtrip signal.After RT, core power rapidly decreased to decay heat levels. The turbine stop valves closed and steamflow to the turbine was terminated.
The steam dump system is designed to actuate following RT to limitthe increase in secondary
: pressure, but the steam dump valves remained closed due to the loss ofcondenser vacuum resulting from the assumed LOOP at the time of RT. Thus, the energy transfer from theprimary system caused the secondary side pressure to increase rapidly after RT (Figure 2.7.2-3),
until theSG ARVs lifted to dissipate the energy. As a result of the assumed LOOP, main FW flow was assumed tobe terminated and AFW flow was assumed to be automatically initiated following RT.The RCS pressure and pressurizer level continued to decrease after RT as energy transfer to the secondary system shrank the primary coolant and the tube rupture break flow continued to deplete primaryinventory.
The decrease in RCS inventory resulted in a low pressurizer pressure SI signal. The SI flowincreased the RCS inventory and the RCS pressure trended toward the equilibrium value, where the SIflow rate would equal the break flow rate.TDAFW flow to the ruptured SG was isolated at 368 seconds, MDAFW flow to the ruptured SG wasisolated at 398 seconds, and the ruptured SG MSIV was closed at 625 seconds.
The ruptured SG level waswell above the level required for identification and isolation by these times as a conservatively high SGlevel was assumed in the analysis.
Cooldown of the RCS was initiated 23 minutes after RT. It was therefore assumed that the ARVs on threeintact SGs were opened for the RCS cooldown at 1525 seconds (Figure 2.7.2-6).
The cooldown wascontinued until the cooldown termination temperature obtained from WCGS EOPs was reached.
Whenthis condition was satisfied, the operator closed the ARVs to terminate the cooldown.
This cooldownensured that there would be adequate subcooling in the RCS after the subsequent depressurization of theWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-321RCS to the ruptured SG pressure.
The reduction in the intact SG pressure required to accomplish thecooldown is shown in Figure 2.7.2-3.
The pressurizer level and RCS pressure also decreased during thiscooldown process due to shrinkage of the RCS (Figures 2.7.2-1 and 2.7.2-2).
The ARVs on the intact SGs that were used for the cooldown also automatically opened as necessary tomaintain the prescribed RCS temperature to ensure that subcooling was maintained.
When the ARVs wereopened, the increased energy transfer from the RCS to the secondary system also aided in thedepressurization of the RCS to the ruptured SG pressure after the SI flow was terminated.
After termination of the cooldown, a 5-minute operator action time was imposed prior to the RCSdepressurization.
In these analyses, the RCS depressurization was terminated when the RCS pressure wasreduced to less than the ruptured SG pressure and the pressurizer level was above the required value,because there was adequate subcooling margin and the high pressurizer level setpoint was not reached.The RCS depressurization is shown in Figure 2.7.2-2.
The depressurization reduced the break flow(Figure 2.7.2-4) and increased SI flow to refill the pressurizer (Figure 2.7.2-1).
After termination of the depressurization, a 4-minute operator action time was imposed prior to SItermination.
The SI flow was terminated at this time because the requirements for SI termination weresatisfied.
(RCS subcooling was greater than the required allowance for subcooling uncertainty, minimumAFW flow was available or at least one intact SG level was in the narrow range, the RCS pressure wasstable or increasing, and the pressurizer level was greater than the required value.) After SI termination, the RCS pressure began to decrease (Figure 2.7.2-2).
Break flow was terminated at 4132 seconds.2.7.2.1.4 ResultsThe primary to secondary break flow rate throughout the recovery operations is presented inFigure 2.7.2-4.
The water volume in the ruptured SG is presented as a function of time in Figure 2.7.2-5.The ruptured loop RCS temperature is presented in Figure 2.7.2-7.
The intact loops RCS temperature ispresented in Figure 2.7.2-8.
The peak ruptured SG water volume is 5789 ft3 resulting in 63 ft3 of marginto overfill.
Therefore, it is concluded that overfill of the ruptured SG will not occur for a design basisSGTR for WCGS.2.7.2.2 Conclusion It is concluded that overfill of the ruptured SG causing water to pass through the main steam relief valveswill not occur for a design basis SGTR for WCGS.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3222.7.2.3 References
: 1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.2. WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,"
August 1987.3. Supplement I to WCAP-10698-P-A, "Evaluation of Offsite Radiation Doses for a SteamGenerator Tube Rupture Accident,"
March 1986.4. NSAL-07-1 1, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology,"
November 2007.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-323Table 2.7.2-1 AFW Flows for Design Basis SGTR Analyses MDAFW Failure, All AFW PumpsOperating SG Pressure AFW Flow to Ruptured SG AFW Flow to Intact SGs(psig) (gpm) (gpm/SG)1125 684.76 320.001000 787.93 332.67900 862.25 367.13800 931.00 398.64700 995.51 428.37Table 2.7.2-2 AFW Flows for Design Basis SGTR Analyses MDAFW Failure, TDAFW Pump Stopped,MDAFW Pumps Operating SG Pressure AFW Flow to Ruptured SG AFW Flow to Intact SGs(psig) (gpm) (gpm/SG)1125 440.50 320.001000 500.80 332.67900 544.60 367.13800 585.50 398.64700 623.90 428.37Table 2.7.2-3 AFW Flows for Design Basis SGTR Analyses MDAFW Failure, Ruptured SG Isolated, TDAFW Pump Stopped, MDAFW Pumps Operating During CooldownSG Pressure AFW Flow to Ruptured SG(psig) (gpm)1125 320.001000 332.67900 367.13800 398.64700 428.37WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-324Table 2.7.2-4 SI Flows for Design Basis SGTR AnalysesPressure (psig) Total Injection Flow Rate (gpm)900 12821000 12241100 11651200 11001300 10291400 9511500 8581600 7361700 5381800 5141900 4892000 4632100 4362200 4062235 3952300 3742400 3403000 340WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-325Table 2.7.2-5 Operator Action Times for Design Basis SGTR Margin to Overfill AnalysesAction TimeOperator action time to isolate TDAFW flow to 15% NRSruptured SG")Operator action time to isolate MDAFW flow to AFW control valve single failure -30 seconds fromruptured SG"' TDAFW isolation SG ARV single failure -coincident with TDAFWisolation Operator action time to isolate MSIV on ruptured SG 8 minutes from RTOperator action time to initiate cooldown 23 minutes from RTCooldown Calculated by RETRANOperator action time to initiate depressurization 5 minutes from end of cooldownDepressurization Calculated by RETRANOperator action time to terminate SI following 4 minutes firom end of depressurization depressurization Pressure equalization Calculated by RETRANNote:I. Isolation is assumed to occur no earlier than 2 minutes after RT.Table 2.7.2-6 Sequence of Events for Limiting Margin to Overfill AnalysesEvent Time (seconds)
SGTR 100RT (OTAT) and LOOP 145AFW Initiated 145SI Actuated 304TDAFW Flow to Ruptured SG Isolated 368MDAFW Flow to Ruptured SG Isolated 398Ruptured SG MSIV Closed 625RCS Cooldown Initiated 1525RCS Cooldown Terminated 2149RCS Depressurization Initiated 2449RCS Depressurization Terminated 2481SI Terminated 2721Break Flow Terminated 4132WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3261IU-C)_0 1000 2000 300o 4000Time (s)5000Figure 2.7.2-1 Pressurizer Level -Margin to Overfill AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-327WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-327C/')C)0 1000 2000 3000 4000Time (s)5000Figure 2.7.2-2 Pressurizer Pressure
-Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-328WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-328Intact SGsR...Rup tured SG2500-2000-o 1500-c_C !)0-1000-0 1000 2000 3000 4000Time (s)5000Figure 2.7.2-3 Secondary Pressure
-Margin to Overfill AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-329WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-329504030200ElCD10-I0 1000 2000 3000 4000Time (s)5000Figure 2.7.2-4 Primary to Secondary Break Flow -Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-330WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-330Av i clab I eRuptured SI60005000-4000-ILD3000-2000-1000-0010002000Time7000(S)40005000Figure 2.7.2-5 SG Water Volumes -Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-331WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-33 1Ruptured SGTotal IntactSGsQ90)Ct)co0 1000 2 o00 3000 4000Time (s)5000Figure 2.7.2-6 SG Steam Releases
-Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTLNGHOUSE NON-PROPRIETARY CLASS 32-332WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 32Hot Leg InCo d Leg II eetn Ie t700600500-40)cB400-:73000)a-G~)K-~A\ ,~'N300200-100-01000200003000040005000Time (s)Figure 2.7.2-7 Ruptured Loop RCS Temperature
-Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-333Hot Leg InIetCo d Leg Inlet700-600500WS400-C)500-Cr,C)0 1000 2000 3000 4000Time (s)5000Figure 2.7.2-8 Intact Loops RCS Temperature
-Margin to Overfill AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3342.7.3 Steam Generator Tube Rupture -Input to Dose (USAR Section 15.6.3)2.7.3.1 Technical Evaluation 2.7.3.1.1 Introduction The major hazard associated with an SGTR event is the radiological consequences resulting from thetransfer of radioactive primary coolant to the secondary side of the ruptured SG and subsequent release ofradioactivity to the atmosphere.
A T/H SGTR analysis was performed to determine the mass releases foruse in calculating the radiological consequences, assuming the limiting single failure relative toradiological consequences without ruptured SG overfill.
Section 2.7.2 confirmed that ruptured SG overfilldid not occur.The SGTR T/H transient analysis was performed using the RETRAN computer program (Reference 1)following the methodology developed in Reference 2 and its Supplement 1 (Reference 3). The plantresponse to the event was modeled using conservative assumptions of break size and location, condenser availability, and initial secondary water mass. The analyses include the simulation of the operator actionsfor recovery from an SGTR based on the WCGS EOPs, which are based on the Westinghouse OwnersGroup Emergency Response Guidelines.
A detailed SGTR T/H analysis was performed for the time period from the SGTR until the primary andsecondary pressures equalized (break flow termination).
In the T/H analysis, the primary to secondary break flow and the steam releases to the atmosphere from the ruptured and intact SGs were calculated foruse in determining the activity released to the atmosphere.
The mass releases were calculated with theRETRAN computer code from the initiation of the event until break flow termination.
For the time periodfrom break flow termination until all releases are terminated, steam releases from the intact and rupturedSGs were determined from a mass and energy balance.The SGTR T/H analysis supports operation at a core power up to 3637 MWt. The analysis supports a fullpower RCS Tavg operating range from 570.7&deg;F to 588.4&deg;F, and a main Tfeed range from 400'F to 448.6&deg;F,with up to 10 percent of the SG tubes plugged.2.7.3.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe T/H analyses, which determined the mass releases for the radiological consequences
: analyses, modeled the plant operating at the higher end of the Ta,,g range. A higher operating temperature results inincreased steaming from the ruptured SG and a higher fraction of the break flow flashing to steam insidethe ruptured SG. The analyses also assumed that the plant was operating at the high end of the Tfeed range.This results in increased steaming from the ruptured SG. A SG tube plugging level of 10 percent wasmodeled in the analyses because this results in a higher fraction of the break flow flashing to steam insidethe ruptured SG.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3352.7.3.1.2.1 Design Basis AccidentThe design basis accident modeled was a double-ended break of one SG tube located at the top of the tubesheet on the outlet (CL) side of the SG. The location of the break on the cold side of the SG results inhigher primary to secondary break flow than a break on the hot side of the SG, as determined byReference
: 2. However, the break flow flashing fraction was conservatively calculated, assuming that allof the break flow comes from the HL side of the SG. The combination of these conservative assumptions results in a very conservative calculation of the radiological consequences.
It was also assumed thatLOOP occurred at the time of RT, and the highest worth control rod assembly was assumed to be stuck inits fully withdrawn position at RT. Due to the assumed LOOP, the condenser was not available for steamreleases once the reactor was tripped.
Consequently, after RT, steam was released to the atmosphere through the SG ARVs.2.7.3.1.2.2 Single Failure Consideration Based on Reference 3, the most limiting single failure with respect to radiological consequences (assuming no overfill) is a failed-open ARV on the ruptured SG. Failure of this ARV causes anuncontrolled depressurization of the SG, which increases primary to secondary break flow and the steamrelease to the atmosphere.
The lower secondary pressure also results in a higher break flow flashingfraction.
Pressure in the ruptured SG will remain below that in the primary system until the failed ARVcan be isolated, and recovery actions completed.
2.7.3.1.2.3 Conservative Assumptions This section includes a discussion of the methods and assumptions used to analyze the SGTR event and tocalculate the mass released, the sequence of events during the recovery operations, and the calculated results.Most of the assumptions used for the margin to overfill analyses are also conservative for the radiological consequences analyses.
The major differences in the assumptions that were used for the RETRANanalyses for radiological consequences compared to those used in the margin to overfill analyses arediscussed below.1. SG Secondary MassA low secondary mass is conservative for the dose analyses because it promotes steam releasefrom the ruptured SG. A low secondary mass also results in a lower ruptured SG pressure whenthe ruptured SG ARV is failed open. This was considered in the confirmation that the pressure didnot decrease below 275 psig as noted in the operator action time discussion below.2. Decay Heat and NSAL-07-11 As noted in NSAL 11 (Reference 4), SGTR T/H analyses for input to the radiological consequences analyses have no competing effects with respect to decay heat. Higher decay resultsin increased steam releases from the ruptured SG and a longer cooldown, leading to a later breakflow termination.
These effects are conservative for the SGTR radiological consequences WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-336calculation, and thus, lower decay heat was not considered.
Similarly, the maximum AFW and SIenthalpies were used. The following changes were made to the related assumptions used in themargin to overfill analyses:
-The 1971+20%
ANS decay heat model specified by Reference 2 was used for theseanalyses.
-Maximum AFW enthalpy is conservative consistent with Reference
: 2. For this analysis, themaximum AFW enthalpy of 96.0 Btu/lbm was modeled.-Maximum SI enthalpy is conservative consistent with Reference
: 2. For this analysis, themaximum SI enthalpy of 73.91 Btu/lbm was modeled.3. Flashing FractionWhen calculating the fraction of break flow that flashes to steam, 100 percent of the break flowwas assumed to come from the HL side of the break. Because the tube rupture flow actuallyconsists of flow from the HL and CL sides of the SG, the temperature of the combined flow willbe less than the HL temperature and the flashing fraction would be correspondingly lower. Thus,this assumption is conservative.
2.7.3.1.2.4 Plant InputThe significant WCGS input is the same as modeled in the margin to overfill analyses except for the AFWflow. It was assumed that the minimum AFW flow (285 gpm/SG) was delivered to the SGs following RTand LOOP with a maximum delay (60 seconds).
The maximum purge volume (138.4 ft3) was modeled todelay delivery of cold AFW to the SGs and maximize steam release.
Flow to the ruptured SG continued atthis rate until it was isolated by the operators.
Flow to the intact SGs was throttled to maintain the levelbelow 50 percent NRS.2.7.3.1.2.5 Operator Action TimesThe major operator actions required for the recovery from an SGTR are discussed in Section 2.7.2.1.2.5, and the operator action times used for the analyses are presented in Table 2.7.3-1.
With the exception ofthe time to isolate AFW flow to the ruptured SG, the operator action times assumed for the margin tooverfill analyses were also used for the radiological consequences analyses.
Earlier AFW isolation resultsin higher releases, so it was assumed that AFW flow to the ruptured SG was isolated when level in the SGreached the WCGS required level (but not before 8 minutes because earlier isolation is considered unrealistic).
Assuming a minimum of 8 minutes from event initiation until AFW isolation used in theinput to dose analyses is not a critical operator action time, and does not impose a requirement on theoperators.
This time constraint was included to avoid unrealistic AFW isolation times.For the radiological consequences
: analyses, the ARV on the ruptured SG was assumed to fail open at thetime the ruptured SG is isolated.
Before proceeding with the recovery operations, the failed-open ARV onthe ruptured SG was assumed to be isolated by locally closing the associated block valve. An operator canlocally close the block valve for the ARV on the ruptured SG within 30 minutes after the failure.
Thus, itWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-337was assumed that the ruptured SG ARV was isolated at 30 minutes after the valve is assumed to fail open.After the ruptured SG ARV was isolated, an additional delay time of 10 minutes (Table 2.7.3-1) wasassumed before initiation of the RCS cooldown.
The cooldown was performed using the ARVs on allthree of the intact SGs. The cooldown target temperature was selected based on the ruptured SG pressure.
As specified in the WCGS EOPs for SGTR, this pressure must be above 275 psig.2.7.3.1.2.6 Mass Release Calculations The mass releases were determined for use in evaluating the offsite and control room radiological consequences of the SGTR using the methodology of Reference
: 3. The steam releases from the rupturedand intact SGs, and primary to secondary break flow into the ruptured SG and the associated flashingfraction, were determined for the period from accident initiation until break flow termination and frombreak flow termination to 12 hours after the accident.
In the RETRAN analyses, the SGTR recovery actions in the WCGS EOPs were simulated until thetermination of primary to secondary break flow. After the primary to secondary break flow is terminated, the operators will continue the SGTR recovery actions.
The plant is then cooled and depressurized to coldshutdown conditions.
In accordance with the methodology in Reference 3, it was assumed that theoperators perform the post-SGTR cooldown using steam dump to the atmosphere.
This method results ina conservative evaluation of the long-term releases for use in the radiological consequences analysescompared to the other cooldown methods in the WCGS EOPs. This procedure for depressurizing theruptured SG was assumed even though the RETRAN analyses performed to calculate releases up untilbreak flow termination assumed ruptured SG ARV isolation.
The high level actions for the post-SGTR cooldown method using steam dump are discussed below.I. Prepare for Cooldown to Cold ShutdownThe initial steps to prepare for cooldown to cold shutdown will be continued if they have notalready been completed.
A few additional steps are also performed prior to initiating cooldown.
These include isolating the CL SI accumulators to prevent unnecessary injection, energizing pressurizer heaters as necessary to saturate the pressurizer water and to provide for better pressurecontrol, and ensuring shutdown margin in the event of a potential boron dilution due to in-leakage from the ruptured SG.2. Cooldown RCS to RHRS Temperature The RCS is cooled by releasing steam from the intact SGs similar to a normal cooldown.
Becauseall immediate safety concerns have been resolved, the cooldown rate should be maintained lessthan the maximum allowable rate of 100lF/hr.
The preferred means for cooling the RCS is viasteam dump to the condenser, because this minimizes the radiological releases and conserves FWsupply. The ARVs on the intact SGs can also be used if steam dump to the condenser isunavailable.
Because a LOOP is assumed, it is assumed that the cooldown is performed usingsteam dump to the atmosphere via the ARVs on the intact SGs. When the RHRS operating temperature is reached, the cooldown is stopped until RCS pressure can also be decreased.
Thisensures that pressure/temperature limits will not be exceeded.
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-3383. Depressurize RCS to RHRS PressureWhen the cooldown to RHRS temperature is completed, the pressure in the ruptured SG isdecreased by releasing steam from the ruptured SG. It was assumed that the ruptured SG isdepressurized by releasing steam via the ARV. As the ruptured SG pressure is reduced, the RCSpressure is maintained equal to the pressure in the ruptured SG in order to prevent excessive in-leakage of secondary side water or additional primary to secondary break flow. Although normalpressurizer spray is the preferred means of RCS pressure
: control, auxiliary spray or a pressurizer PORV can be used to control RCS pressure if pressurizer spray is not available.
: 4. Cooldown to Cold ShutdownWhen RCS temperature and pressure have been reduced to the RHRS in-service values, RHRScooling is initiated to complete the cooldown to cold shutdown.
When cold shutdown conditions are achieved, the pressurizer can be cooled to terminate the event.2.7.3.1.2.7 Acceptance CriteriaThe analyses were performed to calculate the mass transfer data for input to the radiological consequences analyses.
As such, no acceptance criteria are defined.
The results of the analyses were usedas input to the radiological consequences analyses.
2.7.3.1.3 Description of Analyses and Evaluations The RETRAN results for the limiting input to dose analysis are described below. The limiting case withrespect to the input to dose considered operation at the maximum operating temperature (588.40F), withthe maximum main Tfeed (448.60F), the maximum SGTP level (10 percent),
and the failure of the ARV onthe ruptured SG in the full open position when the operator closes the MSIV. The sequences of events forthese transients are presented in Table 2.7.3-2.Following the tube rupture, water flowed from the primary into the secondary side of the ruptured SGbecause the primary pressure was greater than the SG pressure.
In response to this loss of coolant,pressurizer level decreased (Figure 2.7.3-1).
The RCS pressure (represented by the pressurizer pressure) also decreased (Figure 2.7.3-2) as the steam bubble in the pressurizer expanded.
As the RCS pressuredecreased due to the continued primary to secondary break flow, automatic RT occurred on an OTAT tripsignal.After RT, core power rapidly decreased to decay heat levels. The turbine stop valves closed and steamflow to the turbine was terminated.
The steam dump system is designed to actuate following RT to limitthe increase in secondary
: pressure, but the steam dump valves remained closed due to the loss ofcondenser vacuum resulting from the assumed LOOP at the time of RT. Thus, the energy transfer from theprimary system caused the secondary side pressure to increase rapidly after RT (Figure 2.7.3-3) until theSG ARVs lift to dissipate the energy (Figure 2.7.3-5).
As a result of the assumed LOOP, main FW flowwas assumed to be terminated and AFW flow was assumed to be automatically initiated following RT.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-339The RCS pressure and pressurizer level continued to decrease after RT as energy transfer to the secondary system shrunk the RCS and the tube rupture break flow continued to deplete primary inventory.
Thedecrease in RCS inventory resulted in a low pressurizer pressure SI signal. The SI flow increased the RCSinventory and the RCS pressure trended toward the equilibrium value where the SI flow rate would equalthe break flow rate.AFW flow to the ruptured SG was isolated when the ruptured SG level reached 6 percent NRS, and theruptured SG MSIV was closed at 8 minutes after RT.The ruptured SG ARV was assumed to fail open when the MSIV was closed at the time the ruptured SGlevel reached 6 percent NRS. The failure caused the ruptured SG to depressurize
: rapidly, which resultedin an increase in primary to secondary break flow. The depressurization of the ruptured SG increased thebreak flow and energy transfer from primary to secondary, which resulted in RCS pressure andtemperature decreasing more rapidly than in the margin to overfill analyses.
The ruptured SGdepressurization caused a cooldown in the intact SGs loops. The operators identified that the ruptured SGARV had failed open and closed the associated block valve 30 minutes after the failure.
Once the rupturedSG ARV block valve was closed, the ruptured SG pressure began to increase (Figure 2.7.3-3).
Theruptured SG pressure was confirmed to be above 275 psig at all times in the transient.
This was alsoconfirmed for transients run with less limiting mass transfer
: results, but greater ruptured SG pressurereductions.
The lowest ruptured SG pressure for all cases analyzed was greater than 360 psia.After the block valve for the ruptured SG ARV was closed, a 10-minute operator action time was imposedprior to initiating the cooldown.
The ARVs on all three of the intact SGs were opened at approximately 60 minutes for the RCS cooldown (Figure 2.7.3-5).
The depressurization of the ruptured SG due to thefailed-open ARV affected the RCS cooldown target temperature.
The target temperature was determined based upon the pressure in the ruptured SG at the time the cooldown was initiated.
The cooldown wascontinued until the cooldown termination temperature obtained from WCGS EOPs was reached.
Whenthis condition was satisfied, the operators closed the ARVs to terminate the cooldown.
The cooldownensured that there would be adequate subcooling in the RCS after the subsequent depressurization of theRCS to the ruptured SG pressure.
The reduction in the intact SG pressure required to accomplish thecooldown is shown in Figure 2.7.3-3.
The pressurizer level and RCS pressure also decreased during thiscooldown process due to shrinkage of the RCS (Figure 2.7.3-1).
The ARVs on the intact SGs also automatically opened to maintain the prescribed RCS temperature toensure that subcooling was maintained.
When the ARVs were opened, the increased energy transfer fromthe RCS to the secondary system also aided in the depressurization of the RCS to the ruptured SGpressure after the SI flow was terminated.
After termination of the cooldown, a 5-minute operator action time was imposed prior to the RCSdepressurization.
In these analyses, the RCS depressurization was terminated when the RCS pressure wasreduced to less than the ruptured SG pressure and the pressurizer level was above the required value,because there was adequate subcooling margin and the high pressurizer level setpoint was not reached.The RCS depressurization is shown in Figure 2.7.3-2.
The depressurization reduced the break flow(Figure 2.7.3-4) and increased SI flow to refill the pressurizer (Figure 2.7.3-1).
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-340After termination of the depressurization, a 4-minute operator action time was imposed prior to SItermination.
The SI flow was terminated at this time because the requirements for SI termination weresatisfied.
(RCS subcooling was greater than the required allowance for subcooling uncertainty, minimumAFW flow was available or at least one intact SG level was in the narrow range, the RCS pressure wasstable or increasing, and the pressurizer level was greater than the required value.) After SI termination, the RCS pressure began to decrease (Figure 2.7.3-2).
2.7.3.1.3.1 Calculation of Mass ReleasesThe operator actions for the SGTR recovery up to the termination of primary to secondary break flowwere simulated in the RETRAN analyses.
Thus, the steam releases from the ruptured and intact SGs alongwith the break flow into the ruptured SG were determined from the RETRAN results for the period fromthe initiation of the accident until the break flow was terminated.
Following the termination of break flow, it was assumed that the RCS and intact SG conditions weremaintained stable until the cooldown to cold shutdown was initiated.
The ARVs for the intact SGs werethen assumed to be used to start to cool down the RCS to the RHRS operating temperature of 350'F, atthe maximum allowable cooldown rate of 100&deg;F/hr.
The RCS cooldown was assumed to continue until the RHRS operating temperature of 350'F wasreached.
Depressurization of the ruptured SG was then assumed to be performed immediately following the completion of the RCS cooldown.
The ruptured SG was assumed to be depressurized to the RHRSoperating pressure (using a bounding value of 375 psia) via steam release from the ruptured SG ARV. Thismaximizes the steam release from the ruptured SG to the atmosphere, which is conservative for theevaluation of the radiological consequences.
The RCS pressure was also assumed to be reducedconcurrently as the ruptured SG is depressurized.
It was assumed that the RCS cooldown anddepressurization to RHRS operating conditions were completed within 12 hours after the accident.
Thesteam releases from break flow termination to 12 hours were determined for the intact SGs from a massand energy balance using the RCS and intact SG conditions at break flow termination and at the RHRSin-service conditions.
The steam released from the ruptured SG from break flow termination to 12 hourswas determined based on a mass and energy balance for the ruptured SG using the conditions at the timeof break flow termination and saturated conditions at the RHRS operating pressure.
After 12 hours, it was assumed that further plant cooldown to cold shutdown as well as LTC was providedby the RHRS. Therefore, the steam releases to the atmosphere were terminated at 12 hours.2.7.3.1.4 Results2.7.3.1.4.1 Results of RETRAN AnalysesThe primary to secondary break flow rate throughout the recovery operations is presented inFigure 2.7.3-4.
The break flow flashing fraction was calculated using the ruptured HL loop temperature (Figure 2.7.3-6).
The intact HL loop temperature is presented in Figure 2.7.3-7.
The flashing fraction ispresented in Figure 2.7.3-8.
The integrated flashed break flow is presented in Figure 2.7.3-9.
The rupturedSG ARV steam release is presented in Figure 2.7.3-5.
The ruptured SG fluid mass is shown inFigure 2.7.3-10 and ruptured SG water volume is shown in Figure 2.7.3-11.
WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3412.7.3.1.4.2 Mass Release ResultsThe mass release calculations were performed using the methodology discussed in Section 2.7.3.1.3.1.
For the time period from initiation of the accident until break flow termination, the releases weredetermined from the RETRAN results for the time prior to RT and following RT. Because the condenser was in service until RT, any radioactivity released to the atmosphere prior to RT would be through thecondenser vacuum exhaust.
After RT, the releases to the atmosphere were assumed to be via the SGARVs.The transfer and release data are presented in Tables 2.7.3-3 and 2.7.3-4.2.7.3.2 Conclusion The analyses performed to calculate the mass transfer data for input to the radiological consequences analyses were completed and the data were tabulated for the limiting cases. The results of the analyseswere used as input to the radiological consequences analyses.
2.7.3.3 References
: 1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
April 1999.2. WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,"
August 1987.3. Supplement I to WCAP-10698-P-A, "Evaluation of Offsite Radiation Doses for a SteamGenerator Tube Rupture Accident,"
March 1986.4. NSAL-07-1 1, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology,"
November 2007.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-342WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-342Table 2.7.3-1 Operator Action Times for Design Basis SGTR T/H AnalysesAction TimeOperator action time to isolate TDAFW flow to ruptured SG(" 6% NRSOperator action time to isolate MDAFW flow to ruptured SGt1) Coincident with TDAFW isolation Operator action time to isolate MSIV on ruptured SG 8 minutes from RTOperator action time to identify and isolate the failed-open ARV 30 minutes from AFW flow isolation Operator action time to initiate cooldown 10 minutes from closure of the failed-open SG ARVCooldown Calculated by RETRANOperator action time to initiate depressurization 5 minutes from end of cooldownDepressurization Calculated by RETRANOperator action time to terminate SI following depressurization 4 minutes from end of depressurization Pressure equalization Calculated by RETRANNote:I. Isolation is assumed to occur no earlier than 2 minutes after RT.Table 2.7.3-2 Sequence of Events for Limiting Input to Radiological Consequences AnalysesEvent Time (seconds)
SGTR 100RT (OTAT) and LOOP 152AFW Actuated 212SI Actuated 425Ruptured SG MSIV Closed 632AFW Flow to Ruptured SG Isolated 1202Ruptured SG ARV Fails Open 1202Ruptured SG ARV Block Valve Closed 3002RCS Cooldown Initiated 3602Break Flow Flashing Terminated 3946RCS Cooldown Terminated 4955RCS Depressurization Initiated 5255RCS Depressurization Terminated 5377SI Terminated 5616Break Flow Terminated 7627Time RHRS Takes Over Cooling 43,300WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-343Table 2.7.3-3 Break Flow and Flashed Break FlowTotal Break Flow during Total Flashed Break FlowStart of Period (sec) End of Period (sec) Period (Ibm) during Period (Ibm)0 152 2424 399152 1202 41,847 26381202 3002 92,563 12,2113002 3602 32,388 23963602 3946 17,119 5513946 5255 55,9305255 7627 37,4247627 43,300Table 2.7.3-4 Intact and Ruptured SG Steam Flow to Atmosphere Total Intact SGs Steam Total Ruptured SG SteamFlow to Atmosphere Flow to Atmosphere Start of Period (sec) End of Period (sec) during Period (Ibm) during Period (Ibm)0 152 511,500 171,000152 1202 63,525 24,9721202 3002 0 136,2283002 3602 0 03602 3946 85,734 03946 5255 118,909 05255 7627 89,233 07627 43,300 1,496,300 2300Note:Pre-trip steam releases are through the condenser.
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WESTINGHOUSE NON-PROPRIETARY CLASS 32-3441CiC)GC)CJ0 2000 4000 6000Time (s)8000Figure 2.7.3-1 Pressurizer Level -Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-345WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3452400220020001800-V~)16001400.1200'1000 I0 2000 4000 6000 8000Time (s)Figure 2.7.3-2 Pressurizer Pressure
-Input to Radiological Consequences AnalysisWCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-346Intac t SGs.u Rup t ured SG1) [&#xfd; ()njVUU2000-1500-cn1000I "\\I -I I I500-0-02000 4000 6000Time (s)8000Figure 2.7.3-3 Secondary Pressure
-Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-347WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3476050-40--IC-)LEU~)co~C)20-10-0-I I II I II I II I-10I I I I I I I I I I I02000 4000 6000Time (s)8000Figure 2.7.3-4 Primary to Secondary Break Flow -Input to Radiological Consequences AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-348WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-348Ruptured SG....Total Intact SGs1600-14001200-(.9n 1000-E800-0C-nU6o 600-0 2000 4000 6000Time (s)8000Figure 2.7.3-5SG Steam Releases
-Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-349H Hot Leg InletCold Leg Inlet700600-500-0L)0z)co0)H-400-300-200100-'N.........................................................
~-'-'.7I, / AI I II I II I IU02200014000Time (s)060008000Figure 2.7.3-6Ruptured Loop HL and CL Temperatures
-Input to Radiological Consequences AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-350WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-350Hot Leq InIetCo d Leg Inlet-7AF .I L060;U50orLF-400-.-0 ---.' ---- -_-\ , \0-0-0SI I302010U020004000Time (s)60008000Figure 2.7.3-7Intact Loop HL and CL Temperatures
-Input to Radiological Consequences AnalysisWCAP-I 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-351WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-35 10.20.1500(90.10.5E-010 2000 4000 6000Time (s)8000Figure 2.7.3-8 Break Flow Flashing Fraction
-Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3522000015000-10000-K//rCoCO0I5000-/I II I II I II I I002000 4000 6000Time (s)8000Figure 2.7.3-9 Integrated Flashed Break Flow -Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-353EC-)C-n0 2000 4000 6000Time (s)8000Figure 2.7.3-10 Ruptured SG Fluid Mass -Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-354WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-354AvoR u p eRuptured S(Or.,MnUUUU5000-4000-3~000-(GDO2000-1000-I I II I I020004000Time (s)60008000Figure 2.7.3-11 Ruptured SG Water Volume -Input to Radiological Consequences AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3552.7.4 Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breakswithin the Reactor Coolant Pressure Boundary (USAR Section 15.6.5)2.7.4.1 Post-LOCA Subcriticality 2.7.4.1.1 Technical Evaluation 2.7.4.1.1.1 Introduction Post-LOCA subcriticality sump boron calculations were performed in support of the TM CDSA Program.The methodology used to demonstrate compliance with the requirements of 10 CFR 50.46(b) isdocumented in WCAP-8339 (Reference 1). Reference 1 states that the core will remain subcritical post-LOCA by borated water from the various injected ECCS water sources.
Post-LOCA sump boroncalculations demonstrate the core will remain subcritical upon entering, and during, the sumprecirculation phase of ECCS injection.
Containment sump boron concentration calculations are used todevelop a core reactivity limit that is confirmed as part of the Westinghouse RSE Methodology (Reference 2).2.7.4.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe major input parameters and assumptions used in the sump boron calculations are given inTable 2.7.4-1.
The sump boron concentration model is based on the following assumptions:
The calculation of the sump mixed mean boron concentration assumes minimum mass and minimumboron concentrations for significant boron sources and maximum mass and minimum boron concentration for significant dilution sources.Boron is mixed uniformly in the sump. The post-LOCA sump inventory is made up of constituents thatare equally likely to return to the containment sump; that is, selective holdup in containment is neglected.
The sump mixed mean boron concentration is calculated as a function of the pre-trip RCS conditions.
There are no specific acceptance criteria when calculating the post-LOCA sump boron concentration.
Theresulting sump boron concentration, which is calculated as a function of the pre-LOCA RCS boronconcentration, is reviewed for each cycle-specific core design to confirm that adequate boron exists tomaintain subcriticality in the long-term post-LOCA.
2.7.4.1.1.3 Description of Analyses and Evaluations A post-LOCA subcriticality boron limit curve was developed using Westinghouse methodology.
Providedthat the cycle-specific maximum critical boron concentration remains below the post-LOCA sump boronconcentration limit curve (for all rods out, no Xenon, 68&deg;F-212&deg;F),
the core will remain subcritical post-LOCA and the only heat generation will be that due to the remaining long-lived radioactivity.
Thiscriterion will be evaluated on a cycle-specific basis in accordance with the RSE Methodology (Reference 2).WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3562.7.4.1.1.4 ResultsA post-LOCA subcriticality boron limit curve was developed for the TM CDSA Program.
The post-LOCA subcriticality boron limit curve is shown in Figure 2.7.4-1.2.7.4.1.2 Conclusion A post-LOCA subcriticality sump boron curve was generated.
The methodology used to generate thecurve aids in demonstrating compliance with 10 CFR 50.46(b).
The post-LOCA subcriticality sump boroncurve will be tracked on a cycle-specific basis using the Westinghouse RSE Methodology and will aid indemonstrating continued compliance with 10 CFR 50.46(b).
2.7.4.1.3 References
: 1. WCAP-8339, "Westinghouse Emergency Core Cooling System Evaluation Model -Summary,"
June 1974.2. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"
July 1985.2.7.4.2 Post-LOCA Long-term Cooling2.7.4.2.1 Technical Evaluation 2.7.4.2.1.1 Introduction A post-LOCA LTC analysis was performed for the methodology transition program.
There are twoaspects to an LTC analysis:
6Boric acid precipitation control (BAPC)Long-term decay heat removal (DHR)This analysis satisfies the requirements of 10 CFR 50.46(b),
Item (5). The 10 CFR 50 GDC contribute tosupporting the conclusions that the following requirements are met:(5) Long-term cooling.
After any calculated successful initial operation of the ECCS, thecalculated core temperature shall be maintained at an acceptably low value and decay heatshall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.The injection and sump recirculation ECCS modes are described in USAR Section 6.3: Emergency CoreCooling System.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3572.7.4.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaLong-Term CoolingThe major inputs to the boric acid precipitation calculation include core power assumptions, assumptions for boron concentrations, and water volume/mass assumptions for significant contributors to thecontainment sump. The input parameters used in the WCGS methodology transition boric acidprecipitation calculations are given in Table 2.7.4-2.
The accumulator maximum boron concentration utilized in the safety analysis is 2500 ppm and the RWST maximum boron concentration utilized in thesafety analysis is 2500 ppm.The boric acid precipitation model is based on the following assumptions:
meets USNRC guidance aspresented in Reference 1, and is consistent with the interim methodology reported in Reference 2.Additional detailed input assumptions are given as follows:The boric acid concentration in the core region was computed over time by considering the effectof core voiding on liquid mixing volume.0 The boric acid concentration limit is the experimentally determined boric acid solubility limit asreported in Reference 3 and summarized in Table 2.7.4-3 and Figure 2.7.4-2.
For large breaks,containment back pressure is not credited and the RCS is assumed atmospheric.
The boric acidsolubility limit credits an increased boiling point of 21 8&deg;F (boiling point of saturated boric acidsolution under atmospheric conditions).
For break sizes where the RCS pressure might remainelevated (or instances where RCS depressurization is not complete),
the boric acid solubility limitunder atmospheric conditions is assumed.* The liquid mixing volume used in the calculation includes 50 percent of the lower plenum asjustified in Reference 2 and Reference 4.* For SBLOCA scenarios, this analysis does not assume a specific start time forcooldown/depressurization emergency procedures.
In reality, it is anticipated that operators willbegin cooldown/depressurization within 1 hour of the initiation of the event.& The effect of containment sump pH additives on increasing the boric acid solubility limit is notcredited.
0 The boric acid concentration of the makeup containment sump coolant during recirculation is acalculated mixed mean boron concentration.
The calculation of the sump mixed mean boronconcentration assumes maximum mass and maximum boron concentrations for significant boronsources, and minimum mass and maximum boron concentrations for significant dilution sources.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-358USNRC requirements pertaining to the decay heat generation rate for both boric acidaccumulation and decay heat removal (1971 ANS Standard for an infinite operating time with20 percent uncertainty) was considered when performing the boric acid precipitation calculations.
The assumed core power includes a multiplier to address uncertainty as identified by Section L.Aof 10 CFR 50, Appendix K.ECCS recirculation flows are evaluated by comparing the limiting single-failure minimum SIpump flows to the flows necessary to dilute the core and replace core boil-off, thus keeping thecore quenched and amenable to cooling.The acceptance criteria for the LTC analysis are demonstrated by the capability to keep the core cool aftera LOCA and by calculating a time to initiate the BAPC plan with methods, plant design assumptions, andoperating parameters that are consistent with the interim methodology reported in Reference 2.2.7.4.2.1.3 Description of Analyses and Evaluations The LTC phase of the accident begins at the transfer to CL recirculation.
Prior to sump recirculation, corecooling is addressed for the full spectrum of break sizes by the LBLOCA and SBLOCA analyses areas.This satisfies the 10 CFR 50.46 acceptance criteria pertaining to PCT, maximum cladding oxidation, andmaximum hydrogen generation.
DHR checks are performed for the transfer to CL recirculation at transient times based upon full ECCSinjection for the injection phase. Full ECCS injection for the injection phase consists of two residual heatremoval (RHR) pumps (for low head injection),
two SI pumps (for intermediate head injection),
twocentrifugal charging pumps (CCP) (for high head injection),
and two containment spray pumps.Maximizing the flow in the RWST drain down calculation conservatively bounds entry to CLrecirculation.
The earliest entry to CL recirculation for the LBLOCA scenario was conservatively assumed to be 13 minutes.
The earliest entry to CL recirculation for the SBLOCA scenario wasconservatively assumed to be 25 minutes.The adequacy of the CL recirculation flow is checked at the earliest entry to CL recirculation.
Minimumflows are generated assuming the failure of a diesel generator.
One RHR pump takes suction from thecontainment sump, while one IHSI pump and one CCP pump take suction from the RHR pump discharge.
The RHR pump, SI pump, and CCP inject to all four CLs. The spilling line is assumed to be atatmospheric pressure to conservatively minimize the ECCS available for core cooling.The latest acceptable time to enter HL recirculation is determined by the calculated incipient boric acidprecipitation time. Consistent with regulator expectations, the earliest acceptable time to enter HLrecirculation is determined by subtracting 1 hour from the latest acceptable HL recirculation time. Thelatest acceptable HL recirculation time is determined from the calculated incipient boric acid precipitation time. An HL recirculation window provides the operators 1 hour of margin to establish HL recirculation.
The adequacy of the earliest acceptable time to prevent core uncovery and to effectively remove the decayheat generated in the core is then confirmed.
The earliest acceptable time to transfer to HL recirculation isalso confirmed against the entrainment threshold.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-359The transfer to the HL recirculation procedure instructs the operators to transfer the discharge of the SIpump from three CLs to two HLs. The SI pump is stopped to perform this transfer.
The limiting flows forperforming HL recirculation flow checks with respect to DHR occur when the HL recirculation transferprocess is complete.
This is due to the RHR pump providing boosting to the SI pump.The limiting scenario for post-LOCA LTC for HL recirculation consists of one RHR pump taking suctionfrom the containment sump and discharging to all four CLs. The CCP continues to take suction from theRHR pump discharge and inject to all four CLs. The SI pump takes suction from the RHR pumpdischarge and injects to two of the four HLs. The adequacy to effectively remove decay heat at the earlyentry to HL recirculation was checked for both a CL and HL break. The adequacy of the HL flow to haltand reverse the concentration of boric acid in the core was checked at the late entry to HL recirculation for a CL break. An HL break is not a concern with respect to the concentration of boric acid in the core.This is due to CL recirculation flow always being present and providing a forward flushing path.For small breaks, emergency procedures instruct operators to take action to depressurize and cool downthe RCS. Although this depressurization and cooldown process typically begins within 1 hour after theevent, the LTC analysis makes no specific assumptions regarding time to depressurize.
: However, the post-LOCA LTC safety analyses do assume the rate of cooldown is limited by the operating procedures to100lF/hr.
Depressurization to 120 psia (the threshold for boric acid precipitation concerns) may occurbefore or after the prescribed HL switchover (HLSO) time.2.7.4.2.1.4 ResultsAn incipient boric acid concentration time was calculated.
This value is conservatively rounded down tothe nearest half hour to determine the latest acceptable time to complete the transfer to HL recirculation.
The latest acceptable time to complete the transfer to HL recirculation was determined to be 7.5 hoursfrom the initiation of the accident.
Figures 2.7.4-3 and 2.7.4-4 show the buildup of boric acid in the corealong with the impacts of the HL dilution flow at the latest entry to HL recirculation.
It is shown that theHL recirculation flow with one CL spilling is adequate to halt and reverse the concentration of boric acidin the core.Consistent with regulator expectations relative to the interim methodology, 1 hour of margin wasprovided to the operators to complete the steps necessary to transfer from CL recirculation to HLrecirculation.
The adequacy of the ECCS to effectively remove decay heat at an early entry to HLrecirculation of 6.5 hours from the initiation of the accident was checked.
HL recirculation flow wasshown to be adequate to effectively remove decay heat for a CL break, whereas CL recirculation flow wasshown to be adequate to effectively remove decay heat for an HL break. The earliest HL recirculation time of 6.5 hours is well after the entrainment threshold.
Calculations were performed for a condition where HL dilution flow is not established until 12 hoursfrom the initiation of the accident.
This demonstrates the effectiveness of HL dilution flow for thescenario where the RCS remains at an elevated pressure for an extended period. Figure 2.7.4-5 shows theboric acid concentration in the core with the RCS at 120 psia for 12 hours assuming no SG heat removal,no dilution flow, and no benefit of reduced steaming due to SI subcooling.
At 12 hours, the boric acidconcentration is still below the boric acid solubility limit at the saturation temperature of concentrated boric acid associated with 120 psia.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-360Figure 2.7.4-5 shows HL flow at 12 hours with the RCS at saturation conditions.
The RCS is then cooled(with corresponding depressurization) at the maximum cooldown rate of 1 00&deg;F/hr.
It is shown that thecore boric acid concentration is still maintained below the incipient boric acid precipitation limit at thesaturation temperature of concentrated boric acid at the associated pressure.
A higher core region pressure has two significant effects on the calculation of the incipient precipitation time. A higher pressure would decrease core voiding and increase the available mixing volume. With nocredit for subcooling, a higher pressure would increase the core boil-off due to the heat of vaporization decreasing with increasing pressure and thus increase the rate of concentration of boric acid in the core.Loop seal refilling would be significant to the calculations only if the loop seal closure was sustained.
: However, neither LOCA ECCS evaluation models nor observations during the Rig-of-Safety Assessment tests (Reference
: 5) predict sustained loop seal closure, but instead predict cyclic loop seal refilling andclearing.
Cyclic loop seal refilling/clearing would promote mixing in the vessel by forcing liquid from thecore region to the lower plenum and downcomer.
Effective mixing resulting from this type of oscillatory behavior was observed in the modified VEERA test facility (Reference 6).In summary, the WCGS post-LOCA BAPC calculations used a conservative methodology to establish a6.5 to 7.5 hour timeframe to realign the ECCS to provide flushing flow to the HLs. Flushing flow to theHLs provides effective core dilution to halt and reverse the concentration of boric acid in the core prior toreaching the boric acid precipitation limit. This realignment addresses the requirements of10 CFR 50.46(b),
Item (5) LTC. ECCS flows during sump recirculation were shown to be sufficient toremove decay heat after a LOCA.The post-LOCA LTC analyses for the methodology transition are applicable with the following modification to the Emergency Procedures (EMGs):The modification of the transfer to an HL recirculation time of 10 hours to an early initiation time of6.5 hours and a latest completion time of 7.5 hours2.7.4.2.2 Conclusion A post-LOCA LTC analysis was completed.
The capability to keep the core cool in the long-term postLOCA was shown and compliance with 10 CFR 50.46(b),
Item (5) was demonstrated.
A BAPC plan wasestablished to keep the core cool post-LOCA and demonstrate compliance with 10 CFR 50.46(b),
Item (5).WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3612.7.4.2.3 References
: 1. U.S. NRC letter dated November 23, 2005, D.S. Collins to J.A. Gresham, "Clarification of NRCLetter Dated August 1, 2005, Suspension of NRC Approval for Use of Westinghouse TopicalReport CENPD-254-P,
'Post-LOCA Long-Term Cooling Model,' Due to Discovery ofNon-Conservative Modeling Assumptions During Calculations Audit (TAC NO. MB 1365),"(U.S. NRC ADAMS Accession Number ML052930272).
: 2. Letter dated October 3, 2006, "Slides for the Summary of August 23, 2006 Meeting with thePressurized Water Reactor Owners Group (PWROG) to Discuss the Status of Program toEstablish Consistent Criteria for Post Loss-of-Coolant (LOCA) Calculations,"
(U.S. NRCADAMS Accession Number ML062720565).
: 3. WCAP-1570, "Literature Values for Selected Chemical/Physical Properties of Aqueous BoricAcid Solutions,"
May 1960.4. Supplement W3FI-2005-0007 dated February 5, 2005, "Supplement to Amendment RequestNPF-38-249, Extended Power Uprate, Waterford Steam Electric
: Station, Unit 3," (U.S. NRCADAMS Accession Number ML050400463).
: 5. NSD-NRC-97-5092, "Core Uncovery Due to Loop Seal Re-Plugging During Post-LOCA Recovery,"
March 1997.6. Tuunanen, J.; Tuomisto, H.; Raussi, P., "Experimental and Analytical Studies of Boric AcidConcentrations in a VVER-440 Reactor During the Long-Term Cooling Period ofLoss-of-Coolant Accidents,"
Nuclear Engineering and Design, Vol. 148, July 1994, pgs. 217-231.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-362Table 2.7.4-1 Subcriticality Analysis Input Parameters Parameter Current ValueRWST Boron Concentration, Minimum (ppm) 2400RWST Delivered Volume, Minimum (gallons) 236,993RWST Temperature, Maximum ('F) 120Accumulator Boron Concentration, Minimum (ppm) 2300Accumulator Liquid Volume, Minimum (gallons) 6122Number of Accumulators 4Accumulator Tank Temperature, Maximum ('F) 120Table 2.7.4-2 LTC Analysis Input Parameters Parameter Current ValueAnalyzed Core Power (MWt) 3565Analyzed Core Power Uncertainty
(%) 2.0Decay Heat Standard 1971 ANS, Infinite Operation, plus 20%(10 CFR 50 Appendix K)H3B03 Solubility Limit (wt %) See Table 2.7.4-3RWST Boron Concentration, Maximum (ppm) 2500RWST Delivered Volume, Maximum (gallons) 419,000RWST Temperature, Minimum ('F) 37Accumulator Boron Concentration, Maximum (ppm) 2500Accumulator Liquid Volume, Maximum (gallons) 26,376Accumulator Tank Temperature, Minimum ('F) 50WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-363Table 2.7.4-3 Boric Acid Solution Solubility Limit DataSolubility Solubility Temperature g H3BO3/100 g of Temperature g H3BO3/100 g ofOC (OF) Solution H20 &deg;C (OF) Solution H20P = I Atmosphere 75 (167)0(32) 2.70 80(176)5(41) 3.14 85(185)10(50) 3.51 90(194)15 (59) 4.17 95(203)20(68) 4.65 100(212)25 (77) 5.43 103.3 (217.9)30 (86) 6.34 P = PSAT35 (95) 7.19 107.8 (226.0)40(104) 8.17 117.1 (242.8)45 (113) 9.32 126.7 (260.1)50 (122) 10.23 136.3 (277.3)55(131) 11.54 143.3 (289.9)60(140) 12.97 151.5 (304.7)65 (149) 14.42 159.4 (318.9)70 (158) 15.57 171 (339.8) = Congruent Melting of H3B03WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3642,2502,1500OP0doan0.E0LeC2,0501,9501,8501,7502,0001,6500 250 500 750 1,000 1,250 1,500 1,750RCS Boron Concentration, Peak Xe (ppm)Figure 2.7.4-1 Post-LOCA Subcriticality Boron Limit CurveWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-365WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-365706050:~4030Saturated Boric Acid SolutionBoiling Point, 218 *FP=PATM P=PSAT..20100050100150200250300350Temperature
(*F)Figure 2.7.4-2 Boric Acid Solubility LimitWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-366Boric Acid Concentration (wt .%)NO HL DILUTION FLOW475 GPM OF HOT LEG DILUTION FLOWBORIC ACID SOLUBILITY LIMIT 29.27Mass Flow Rate (Ibm/sec)
CORE BOILOFFHL SI FLOWWT%000M-0nU)Cr?(D0 2 4 6 8 10Time (hr)Figure 2.7.4-3LBLOCA Boric Acid Concentration Analysis
-Vessel Boric Acid Concentration, Boil-off, and Flushing Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-367WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-367Boric AMass FI50-40--30-.0C-)C- _0o201-cocid Concentration (wi .t- )NO HL DILUTION FLOW475 GPM OF HOT LEG DILUATION FLOWBORIC ACID SOLUBILITY LIMIT 29.27 WT%ow Rate (Ibm/sec)
CORE BOILOFFHL SI FLOW0CD-nCD0 2 4 6Time (hr)10Figure 2.7.4-4SBLOCA Boric Acid Concentration Analysis
-Vessel Boric Acid Concentration, Boil-off, and Flushing Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-368Boric Acid Concentration (wt.I %)NO HL DILUTION FLOW450 GPM OF HL DILUTION FLOWBORIC ACID SOL LIMIT W/ 1OOF/HR COOLDOWNTemperature (F )TEMP W/ 1OOF/HR COOLDOWN10080--r350--- --- --- --+\iI\,iI""S IVSI0C-)00300250200-15060-40-----------------------------------------------------------__4CD-0C'DCD20-5'-I-10050I I I9-n5)111213Time (hr)1415Figure 2.7.4-5 Core Dilution at 12 Hours for SBLOCA Pressure HangupWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3692.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM (USAR SECTION 15.8)2.8.1 Technical Evaluation 2.8.1.1 Introduction The Final ATWS Rule, 10 CFR 50.62(c)(1)
(Reference 1), requires the incorporation of a diverse (fromthe RT system) actuation of the AFW system and turbine trip for Westinghouse-designed plants. Theinstallation of the USNRC-approved AMSAC satisfies this Final ATWS Rule. However, it must also bedemonstrated that the deterministic ATWS analyses that form the basis for this rule and the AMSACdesign remain valid for the plant. This is typically done by confirming that the analyses documented in NS-TMA-2182 (Reference
: 2) remain valid or by performing new deterministic analyses for theproposed plant state.For the WCGS, the LOL and LONF ATWS events were analyzed to ensure that the analytical basis for theFinal ATWS Rule continues to be met. The LOL and LONF ATWS events are the two most limiting RCSoverpressure transients reported in NS-TMA-2182.
The objective is to show that the ATWS pressure limitof 3200 psig is met for at least 95 percent of the cycle, and therefore the analytical basis for the FinalATWS Rule continues to be met.2.8.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe LOL and LONF ATWS analyses for the WCGS used a plant-specific ATWS model consistent withthe methodology described in NS-TMA-2182.
The following analysis assumptions were used:The nominal and initial conditions were set consistent with the design parameters for an NSSSpower of 3651 MWt.* Westinghouse Model F SG characteristics were used.Consistent with the analysis basis for the Final ATWS Rule (NS-TMA-2182):
-TDF is assumed, no uncertainties are applied to the initial power, RCS average temperature or RCS pressure.
-0 percent SGTP is assumed.
0 percent SGTP is more limiting (that is, results in a higherpeak RCS pressure) for ATWS events.-Control rod insertion was not assumed.-100 percent pressurizer PORV capacity was assumed.WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-370Turbine trip and AFW actuation are modeled to occur at plant-specific times after eventinitiation, using the WCGS AMSAC setpoint and delays.A 25 second AMSAC response time was assumed.
This delay time is added to the time atwhich the SG water mass reaches a mass equivalent to the water level at the AMSAC lowSG water level setpoint of 12 percent of span. The AFW initiation time was determined byadding an additional 60-second delay to account for the time to get the AFW pumps up tospeed, sensor delays, and logic delays. The turbine trip initiation time was determined byadding an additional 5-second delay.A WCGS best-estimate AFW flow of 1200 gpm was assumed.A WCGS-specific MTC of-8 pcm/&deg;F was modeled to bound 95 percent of the cycle. This value isconsistent with that assumed in generic ATWS analyses (Reference 2). The ATWS MTC limit isconfirmed each cycle as part of the reload process.To remain consistent with the basis of the Final ATWS Rule and the supporting analyses documented inNS-TMA-2182,.
the peak RCS pressure reached in the WCGS ATWS evaluations should not exceed theASME B&PV Code, Service Level C stress limit criterion of 3200 psig. This value corresponds to themaximum allowable pressure for the weakest component in the RPV (the nozzle safe end).2.8.1.3 Description of Analyses and Evaluations The LONF and LOL ATWS events were analyzed based on a conservative NSSS power of 3651 MWt.The LOFTRAN computer code (Reference
: 3) was used to perform the WCGS ATWS analyses, consistent with the analysis basis for the Final ATWS Rule.2.8.1.4 ResultsTo remain consistent with the basis of the Final ATWS Rule (10 CFR 50.62), the peak RCS pressurecalculated in both the LOL and the LONF ATWS analyses shall be less than 3200 psig (or 3215 psia). Thecalculated peak RCS pressure obtained for the LOL and LONF ATWS analyses is 2897.9 psia and3129.0 psia, respectively.
The time sequence of events is documented in Table 2.8.1-1 for the LOL ATWSand in Table 2.8.1-2 for the LONF ATWS. Key transient parameters are shown in Figures 2.8. 1-1through 2.8.1-8 for the LOL ATWS and in Figures 2.8.1-9 through 2.8.1-16 for the LONF ATWS. Basedon these results, it has been demonstrated that the analytical basis for the Final ATWS Rule continues tobe met for operation of the WCGS at an NSSS power level as high as 3651 MWt.WCAP-17658-NP August 2013Licensing Report Revision 0
WESUNGHOUSE NON-PROMETARY CLASS 32-3712.8.2 Conclusion The information related to ATWS has been reviewed and it was concluded that it has adequately accounted for the WCGS plant-specific effects on ATWS. The evaluation has demonstrated that theAMSAC continues to meet the requirements of 10 CFR 50.62. The evaluation has shown that the plant isnot required by 10 CFR 50.62 to have a diverse scram system. Additionally, the evaluation has shown thatthe ATWS pressure limit of 3200 psig will be met for at least 95 percent of the cycle. The MTC assumedin this analysis will continue to be checked on a cycle-specific basis. Therefore, the WCGS is acceptable with respect to ATWS.2.8.3 References
: 1. 10 CFR 50.62 and Supplementary Information
: Package, "Requirements for Reduction of Riskfrom ATWS Events for Light Water-Cooled Nuclear Power Plants."2. NS-TMA-2182, "ATWS Submittal,"
December 1979.3. WCAP-7907-P-A.,
"LOFTRAN Code Description,"
April 1984.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-372Table 2.8.1-1 LOL ATWS Time Sequence of EventsEvent Time (seconds)
Turbine trip occurs 1.0Loss of FW flow initiated 4.0Peak RCS pressure reached [2897.9 psia] 104.7AFW initiated 131.0Table 2.8.1-2 LONF ATWS Time Sequence of EventsEvent Time (seconds)
Loss of FW flow initiated 4.0Turbine trip occurs 61.0Peak RCS pressure reached [3129.0 psia] 90.2AFW initiated 116.0WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-373WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3731.20.80.6<D 0.40.201.2o 1x< 0.8t0.60.40C-) n.90 50 100 150 200 250Time (s)Figure 2.8.1-1 Nuclear Power versus Time for LOL ATWS3000 50 100 150 200 250Time (s)Figure 2.8.1-2 Core Heat Flux versus Time for LOL ATWS300WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-374WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-374CL.Co0 50 100 150 200 250Time (s)Figure 2.8.1-3 RCS Pressure versus Time for LOL ATWS3001-1Cl)M00 50 100 150 200 250Time (s)Figure 2.8.14 Pressurizer Water Volume versus Time for LOL ATWS300WCAP-1 7658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-375WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-375LL0~F-Q)0 50 100 150 200 250Time (s)Figure 2.8.1-5 Vessel Inlet Temperature versus Time for LOL ATWS3000C-)V.)0 50 100 150 200 250Time (s)300Figure 2.8.1-6 RCS Flow versus Time for LOL ATWSWCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-376WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-376C)C/)CDV)C-DC/")0 50 100 150 200 250Time (s)Figure 2.8.1-7 SG Pressure versus Time for LOL ATWS3000 50 100 150 200 250Time (s)Figure 2.8.1-8 SG Mass versus Time for LOL ATWS300WCAP- 17658-NP August 2013WCAP- 17658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-377WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-377i0::00~':1)0 50 100 150 200 250Time (s)Figure 2.8.1-9 Nuclear Power versus Time for LONF ATWS3000 50 100 150 200 250Time (s)Figure 2.8.1-10 Core Heat Flux versus Time for LONF ATWS300WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-378WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-378a,-C/)C/)C--)0 50 100 150 200 250 300Time (s)Figure 2.8.1-11 RCS Pressure versus Time for LONF ATWS2,EC')0 50 100 150 200 250Time (s)Figure 2.8.1-12 Pressurizer Water Volume versus Time for LONF ATWS300WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-379EC',a)0 50 100 150 200 250 300Time (s)Figure 2.8.1-13 Vessel Inlet Temperature versus Time for LONF ATWS0c-?O-0050 100 150 200Time (s)Figure 2.8.1-14 RCS Flow versus Time for LONF ATWS250300WCAP- 17658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-380WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-380CLCnCoa,C-oCo0 50 100 150 200 250Time (s)Figure 2.8.1-15 SG Pressure versus Time for LONF ATWS3000 50 100 150 200 250Time (s)Figure 2.8.1-16 SG Mass versus Time for LONF ATWS300WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3812.9 RADIOLOGICAL DOSESThe WCGS radiological consequences analyses have been performed to follow Regulatory Guide (RG) 1.183 (Reference
: 1) which provides guidance on the application of Alternative SourceTerm (AST) methodology, as allowed by 10 CFR 50.67. The AST methodology is being used to calculate the offsite, control room, and Technical Support Center radiological consequences for WCGS to supportthe CDSA TM Program.
The following accidents are analyzed:
0 MSLB* LOAC* Locked rotor0 Rod ejection* Letdown line break* SGTR* LOCA0 Waste gas decay tank failure* Liquid waste tank failure0 Fuel handling accident (FHA)Detailed discussion of the input parameters, assumptions, event descriptions, acceptance
: criteria, analysisresults, and conclusions for each accident is presented in Section 4.3 of Enclosure VI of this LAR.References
: 1. Regulatory Guide 1. 183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,"
July 2000.2.10 INSTRUMENT UNCERTAINTIES 2.10.1 Reactor Trip System/Engineered Safety Feature Actuation Setpoint Uncertainties 2.10.1.1 Introduction and Background The TS Reactor Trip System (RTS)/Engineered Safety Feature Actuation System (ESFAS) and Loss ofPower Diesel Generator Start setpoints have been reviewed, and TS changes have been identified consistent with the Westinghouse setpoint methodology defined in WCAP- 17746-P "Westinghouse Setpoint Methodology as Applied to the Wolf Creek Generating Station,"
Revision 0 (Reference 1). Forthe WCGS TSs, the Allowable Values have been removed and the Nominal Trip Setpoints, which are theWestinghouse defined Limiting Safety System Setting required by 10 CFR 50.36, have been added.Modified footnotes in TSTF-493 Revision 4 Option A have been added to the TS.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3822.10.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe RTS/ESFAS and Loss of Power Diesel Generator Start uncertainty calculations for WCGS wereperformed based on WCGS specific instrumentation, plant calibration procedures, and verified designinputs. The acceptance criterion for the RTS/ESFAS and Loss of Power Diesel Generator Start setpoints isthat the calculated margin, defined as Total Allowance
-Channel Statistical Allowance (defined inReference 1), is >_ 0 % of span.2.10.1.3 Description of Analyses and Evaluations The setpoint analysis uses the square root sum of the squares technique to combine the uncertainty components of an instrument channel in an appropriate combination of those components, or groups ofcomponents, that are statistically independent.
Those uncertainties that are statistically dependent arearithmetically summed to produce statistically independent groups that are then statistically combined.
The methodology used for WCGS is defined in Reference
: 1. Uncertainty calculations were performed forthe RTS/ESFAS and the Loss of Power Diesel Generator Start functions based on the design inputparameters and the Nominal Trip Setpoints noted in the revised TS Tables 3.3.1-1 and 3.3.2-1 andLCO 3.3.5 and confirmed that the acceptance criterion was satisfied for each protection function.
Thecalculations for the RTS/ESFAS and Loss of Power Diesel Generator Start functions are provided inWCAP-17602-P "Westinghouse Setpoint Calculations for the Wolf Creek Generating Station Control,Protection and Indication Systems,"
Revision 0 (Reference 3).2.10.1.4 ResultsThe results of the setpoint calculations for each RTS/ESFAS and Loss of Power Diesel Generator Startfunction are provided in Reference
: 3. The Westinghouse setpoint calculations are based on using theNominal Trip Setpoint as the Limiting Safety System Setting.
For WCGS TS Tables 3.3.1-1 and 3.3.2-1and LCO 3.3.5, the Allowable Values are to be removed and the Nominal Trip Setpoints are to be added.This is consistent with the Westinghouse definition of the Limiting Safety System Setting required in 10CFR 50.36 and the Westinghouse setpoint methodology.
The footnotes defined in TSTF-493 Revision 4Option A are to be added to the TS. The Allowable Values for WCGS are to be replaced with As Left andAs Found operability criteria consistent with TSTF-493 Revision 4, Option A requirements.
The As Leftand As Found operability criteria for the RTS/ESFAS and Loss of Power Diesel Generator Startinstrumentation have been defined in Reference 3 and are consistent with the Westinghouse setpointmethodology defined in Reference 1.2.10.1.5 Conclusions All RTS/ESFAS functions meet the acceptance criterion.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTfNGHOUSE NON-PROPRIETARY CLASS 32-3832.10.1.6 References
: 1. WCAP- 1 7746-P, Revision 0, "Westinghouse Setpoint Methodology as Applied to the Wolf CreekGenerating Station."
: 2. TSTF-493, Revision 4 "Clarify Application of Setpoint Methodology for LSSS Functions,"
April 2010.3. WCAP- 1 7602-P, Revision 0, "Westinghouse Setpoint Calculations for the Wolf Creek Generating Station Control, Protection and Indication Systems."
2.10.2 Initial Condition Uncertainties 2.10.2.1 Introduction and Background Initial condition uncertainties are conservative steady-state instrumentation measurement uncertainties that are applied to nominal parameter values to obtain conservative initial conditions for use in the safety(accident) analyses.
The following initial condition parameter uncertainties were calculated for WCGS foruse in the accident analyses to assess the safety analyses acceptability.
* Daily Power Measurement
* RCS Flow Measurement
* Pressurizer Pressure Control* Tavg Control* SG Level Control* Pressurizer Level ControlThe results of the uncertainty calculations are presented in WCAP- 17602-P "Westinghouse SetpointCalculations for the Wolf Creek Generating Station Control, Protection and Indication Systems,"
Revision0 (Reference 1). The final uncertainty calculations confirm that the initial condition values utilized in thesafety analysis are bounding.
2.10.2.2 Input Parameters, Assumptions, and Acceptance CriteriaThe initial condition uncertainty calculations for WCGS were performed based on WCGS specificinstrumentation, plant calibration procedures, and verified design inputs. The acceptance criteria for theinitial condition uncertainties are that the calculated uncertainty must be less than or equal to theuncertainty values used in the safety analyses.
2.10.2.3 Description of Analyses and Evaluations The initial condition uncertainty analysis uses the square root sum of the squares technique to combinethe uncertainty components of an instrument channel in an appropriate combination of those components, or groups of components, that are statistically independent.
Those uncertainties that are statistically dependent are arithmetically summed to produce statistically independent groups that are then statistically WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-384combined.
The Westinghouse setpoint methodology is defined in WCAP- 17746-P "Westinghouse Setpoint Methodology as Applied to the Wolf Creek Generating Station,"
Revision 0 (Reference 2).2.10.2.4 ResultsThe results of this analysis are summarized in Reference 1.2.10.2.5 Conclusions As demonstrated by the calculations provided in Reference 1, the acceptance criteria have been satisfied.
2.10.2.6 References
: 1. WCAP-17602-P, Revision 0, "Westinghouse Setpoint Calculations for the Wolf Creek Generating Station Control, Protection and Indication Systems."
: 2. WCAP- 1 7746-P, Revision 0, "Westinghouse Setpoint Methodology as Applied to the Wolf CreekGenerating Station."
2.11 CONTROL SYSTEMS ANALYSIS2.11.1 NSSS Pressure Control Component Sizing (USAR Sections 5.4, 7.7, & 10.4.4)2.11.1.1 Technical Evaluation 2.11.1.1.1 Introduction The following pressure control components were evaluated for the TM Program.
The purpose of thisevaluation is to ensure that the NSSS pressure control system component installed capacities are adequateand meet the plant design basis sizing requirements.
* Pressurizer PORVs* Pressurizer spray valves* Pressurizer heaters* Steam dump valvesThe analyses were performed to envelop the window of operating conditions, which include thefull-power Ta,,g window, the full-power Tfeed window, and 0 to 10 percent average SGTP levels. Theanalyses utilized the potential MUR power level and conditions; the analysis results therefore, bound thecurrent power level conditions.
The pressure control components are described in the USAR, Section 5.4 ("Component and Subsystem Design"),
Section 7.7 ("Control Systems not Required for Safety"),
and Section 10.4.4 ("Turbine BypassSystem").
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3852.11.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaPressurizer PORVsThe pressurizer PORV sizing analysis was performed at the MUR operating conditions shown inSection 1.1. The analysis at MUR power conditions bounds current power conditions.
The pressurizer PORV sizing analysis was performed to confirm that the installed PORV capacity is adequate to meet theapplicable sizing criteria for the TM Program.
The analysis included the following key input parameters and assumptions:
* The design basis transient is modeled as a 50 percent step load rejection from full power.* The analysis is performed at a full power Tavg of 588.4&deg;F (high Tavg), 0 percent SGTP, and a Tfeedof 448.60F, which bounds all other normal operating conditions.
0 The analysis is best estimate and best estimate conditions are assumed except a 4&deg;F Tavguncertainty and 0.6 percent power uncertainty were applied for conservatism.
* The secondary side water mass was reduced by 10 percent for conservatism.
0 The initial pressurizer pressure is at the nominal pressure of 2250 psia.* The initial pressurizer water level is at the nominal setpoint.
* There are a total of two pressurizer PORVs, each with a rated capacity of 2 10,000 lbm/hr at2335 psig.0 The NSSS control systems (rod, pressurizer
: pressure, pressurizer level, SG level, and steam dumpcontrol systems) are assumed to be operational and functioning as designed.
* The pressurizer PORV sizing analysis is completed using best estimate BOL fuel reactivity data.BOL parameters have lower differential rod worths and least negative MTC; thus, using BOLparameters yields conservative results and bounds the entire fuel cycle.Acceptance CriteriaThe Westinghouse sizing basis for the pressurizer PORVs is to prevent the pressurizer pressure fromreaching the high pressurizer pressure RT setpoint during the design basis large load rejection transient.
This design basis large load rejection is defined as a 50 percent step load reduction from full power. Thesizing criterion is conservatively met if the maximum pressurizer insurge during the transient is less thanthe total capacity of the PORVs. The sizing basis for the PORVs is documented inUSAR Sections 5.4.13.1 and 7.7.1.5 and is consistent with the Westinghouse sizing basis.WCAP-17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-386Pressurizer Spray ValvesThe pressurizer spray valves sizing analysis was performed at the MUR power conditions shown inSection 1.1. The analysis at MUR power conditions bounds current power conditions.
The pressurizer spray valves sizing analysis was performed to confirm that the installed spray valves capacity is adequateto meet the applicable sizing criteria for the TM Program.
The analysis included the key input parameters and assumptions listed below:0 The design basis transient is modeled as a 10 percent step load decrease from full power.* The analysis is perforned at a full power Tavg of 588.4&deg;F (high Tavg), 0 percent SGTP, and a T'eedof 448.6&deg;F, which bounds all other normal operating conditions.
* The analysis is best estimate and best estimate conditions are assumed except a 4&deg;F Taguncertainty and 0.6 percent power uncertainty were applied for conservatism.
* The secondary side water mass was reduced by 10 percent for conservatism.
* The initial pressurizer pressure is at the nominal pressure of 2250 psia.0 The initial pressurizer water level is at the nominal setpoint applicable to the full power Tavgoperating conditions.
* There are two pressurizer spray valves with a combined total capacity of 896 gpm.* The NSSS control systems (rod, pressurizer level, pressurizer
: pressure, and SG level) areassumed to be operational and functioning as designed.
The steam dump system is not actuatedfor load changes less than 10 percent; therefore, steam dump is not modeled for this analysis.
* The pressurizer spray valve sizing analysis is completed using best estimate BOL fuel reactivity data. BOL parameters have lower differential rod worths and least negative MTC; thus, usingBOL parameters yields conservative results and bounds the entire cycle.Acceptance CriteriaThe Westinghouse sizing basis for the pressurizer spray valves is to prevent the pressurizer pressure fromreaching the pressurizer PORV actuation setpoint of 2335 psig (2350 psia) for the design basis 10 percentstep load decrease transient.
The sizing basis for the spray valves is documented in USARSection 5.4.10.3.4 and is consistent with the Westinghouse sizing basis.Pressurizer HeatersThe evaluation included the following key input parameters and assumptions:
* The total backup heater design capacity is 1384 kW and the total proportional heater designcapacity is 416 kW. This provides a total heater design capacity of 1800 kW.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-387The currently installed backup and proportional heater capacities are 1315 kW and 347 kW,respectively.
The total pressurizer internal volume is 1800.0 ft3.Acceptance CriteriaThe pressurizer heater capacity should be sufficient to 1) provide heat during cold plant startup,2) regulate pressure to avoid reaching applicable RT and ESFAS setpoints during Condition I transients, and 3) counteract the steady-state heat loss that occurs within the pressurizer to maintain steady stateoperating pressure.
The design basis pressurizer heater capacity to meet these requirements is 1 kW ofheater capacity per cubic foot of pressurizer volume. Additionally, limiting condition for operation (LCO)3.4.9.b requires two groups of backup heaters to be operable with the capacity of each group > 150 kW toensure the normal operating pressure can be maintained after accounting for heat losses.Steam Dump ValvesThe steam dump valves sizing analysis was performed at the MUR power conditions shown inSection 1.1. The analysis at MUR power conditions bounds current power conditions.
The steam dumpvalves sizing analysis was performed to confirm that the installed steam dump system capacity isadequate to meet the applicable sizing criteria for the TM Program.Two design basis transients were analyzed for the steam dump capacity sizing: the 50 percent loadrejection from full power transient and the plant trip from full power transient.
The 50 percent loadrejection transient was modeled as a 50 percent step rejection in turbine load from full power and theplant trip was modeled as a turbine trip followed by a RT from full power. The 50 percent load rejection transient was analyzed as part of the margin to trip analysis and the details of the input parameters andassumptions for this transient are defined in Section 2.11.2. The analysis of the plant trip transient included the following key input parameters and assumptions:
0 A two second delay is conservatively assumed for RT on turbine trip.a The analysis is performed at a full power Tavg of 588.4&deg;F (high Tavg), 0 percent SGTP, and a Tfeedof 448.6&deg;F, which bounds all other normal operating conditions.
* The analysis is best estimate and best estimate conditions are assumed except a 4&deg;F Tavguncertainty and 0.6 percent power uncertainty were applied for conservatism.
* The secondary side water mass was reduced by 10 percent for conservatism.
0 The SG PORVs are not modeled in the analysis.
* The NSSS control systems (rod, pressurizer level, pressurizer
: pressure, and SG level) areassumed to be operational and functioning as designed.
The steam dump system is operating inthe RT controller mode.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-388Acceptance CriteriaThe Westinghouse sizing basis of the steam dump system is to enable the plant to survive a 50 percentload rejection without generating a RT or challenging the MSSV, and by being able to survive a turbinetrip with a RT without challenging the MSSV actuation setpoint of 1185 psig (1200 psia). The sizing basisfor the steam dump system is documented in USAR Section 10.4.4.1 "Power Generation Design BasisThree" and is consistent with the Westinghouse sizing basis.2.11.1.1.3 Description of Analyses and Evaluations Transients for the pressure control component sizing analysis were analyzed using the LOFTRANcomputer code (Reference 1). This computer code is a system-level program code that models the overallNSSS, including detailed modeling for control and protection systems.
A LOFTRAN computer modelwas developed for the WCGS. The key input parameters and assumptions for each transient are given inSection 2.11.1.1.2.
No computer analyses were performed specifically for the pressurizer heater sizing evaluation; however,the operational and margin to trip analyses in Section 2.11.2 were performed using the current heatercapacity and show that an adequate transient response is maintained with the reduced pressurizer heatercapacity.
Additionally, the impact of the reduced heater capacity is evaluated in Section 2.11.1.1.4.
2.11.1.1.4 ResultsPressurizer PORVsThe results of the analysis show a maximum pressurizer pressure of 2351 psia, which is less than the highpressurizer pressure RT setpoint of 2400 psia. The results of the analysis also show that the maximumpressurizer insurge flow rate is 152,900 lbm/hr compared to the installed PORV capacity of420,000 lbrm/hr.The calculated peak pressurizer pressure on a design basis 50 percent load rejection was less than the highpressure RT setpoint.
Therefore, the PORVs have sufficient relief capacity to avoid a RT on highpressurizer pressure for the design basis load rejection.
Similarly, it was shown that the required PORVcapacity (i.e., the pressurizer insurge) during the transient was less than the total installed capacity.
ThePORVs are therefore adequately sized for the TM program.Pressurizer Spray ValvesThe results of the 10 percent step load decrease from full power show a maximum pressurizer pressure of2330 psia, which is less than the pressurizer PORV actuation setpoint of 2350 psia. Because the peakpressurizer pressure was less than the PORV actuation setpoint of 2350 psia, the total installed sprayvalves capacity of 896 gpm is adequate to avoid actuation of the pressurizer PORV during a 10 percentstep load decrease from full power transient.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-389Pressurizer HeatersThe total design heater capacity of 1800 kW meets the design criteria of 1 kW/ ft3 of pressurizer volume;however, as discussed in Section 2.11.1.1.2, the current heater capacities are 1315 kW for the backupheaters and 347 kW for the proportional
: heaters, for a total heater capacity of 1662 kW. Based on thecurrent capacity of the heaters, the criterion of one kilowatt per one cubic foot is not met. Therefore, thesizing evaluation was performed for a total heater capacity of 1662 kW to confirm that the plant responseto the design basis transients would remain acceptable with the reduced heater capacity.
Design basis transients resulting in pressurizer insurges/outsurges such as loadings/unloadings, loadrejections, and RTs show pressurizer pressure changes that are too rapid for the pressurizer heaters tosignificantly influence.
In addition, past analyses have demonstrated only small differences in themaximum/minimum pressurizer pressure when it is assumed that a fraction of the pressurizer heaters areout of service.
Analyses have demonstrated that a reduced heater capacity results in increased times forplant heatup. A reduction in pressurizer heater capacity of this magnitude is acceptable for transient mitigation based on the results of the operational transient analysis in Section 2.11.2. The currentproportional heater capacity of 347 kW is greater than the 300 kW specified in LCO 3.4.9.b; therefore, theproportional heaters are capable of maintaining the normal operating pressure as designed.
It is concluded that a total heater capacity of 1662 kW is acceptable and this conclusion is furtherdemonstrated by the results of the operational transient analyses documented in Section 2.11.2.Steam Dump ValvesThe results of the turbine trip followed by a RT from full power analysis show a maximum SG pressure of1158 psia, which provides 42 psi of margin to the first MSSV lift setpoint of 1200 psia. The results of the50 percent load rejection analysis discussed in more detail in Section 2.11.2 show that acceptable marginis maintained to all applicable RT setpoints, and the first MSSV lift setpoint is not exceeded.
Because the SG pressure was less than the lowest MSSV actuation setpoint of 1200 psia for both designbasis transients and no RT setpoints are reached during the 50 percent load rejection transient, the totalinstalled capacity steam dump system is adequate for the TM Program.2.11.1.2 Conclusions Pressure control component sizing analyses for the pressurizer PORV, pressurizer spray valves,pressurizer
: heaters, and steam dump valves were performed using Westinghouse methodology as part ofthe WCGS TM Program.
The results of these analyses showed that the installed capacities of thesecomponents at WCGS are adequate at the current and MUR power levels.2.11.1.3 References
: 1. WCAP-7907-P-A, "LOFTRAN Code Description,"
April 1984.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3902.11.2 Operational Analysis and Margin to Trip (USAR Section 7.7)2.11.2.1 Technical Evaluation 2.11.2.1.1 Introduction The following transients were analyzed or evaluated for the TM Program.
These transients are listed inUSAR Section 7.7.2 ("Analysis"),
as design basis operational transients that the plant control systemsdescribed in USAR Section 7.7.1 should regulate without actuation of plant safety systems.* 5 percent per minute unit loading and unloading (Condition I)* 10 percent step load decrease (Condition I)* 10 percent step load increase (Condition I)0 Large load rejection (Condition I)Additionally, the turbine trip without a RT transient from the P-9 permissive setpoint was analyzed todemonstrate that adequate margin is maintained to the pressurizer PORVs actuation setpoint(NUREG-0737, item II.K.3.10).
The transients were analyzed to envelop the window of operating conditions that include the full-power Tavg window, the full-power Tfeed window, and 0 to 10 percent average SGTP levels. The analyses wereperformed at the potential MUR power uprate conditions and the results of the analyses bound the currentpower level conditions.
2.11.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe analyses included the following key input parameters and assumptions:
The analyses were performed at the MUR operating conditions shown in Section 1.1. Theanalyses were based on the MUR NSSS power level of 3651 MWt and bound the current NSSSpower level of 3579 MWt. Additionally, the analyses cover a full-power Tag range from 570.71Fto 588.4'F, full-power Tfeed range from 400.00F to 448.60F, and average SGTP levels between0 percent and 10 percent.The plant operational analysis is a best-estimate analysis; therefore, the plant conditions are at thenominal values and instrument uncertainties are not applied.
: However, a 0.6 percent powerallowance was applied for conservatism.
All NSSS control systems (reactor, pressurizer
: pressure, pressurizer level, SG level, andsteam dump) are assumed to be operational and functioning as designed.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTrNGHOUSE NON-PROPRIETARY CLASS 32-391Best-estimate reactor kinetics data (such as MTC, Doppler power defect, and control rod worth)were used as input to the analysis.
BOL core conditions were used, which are bounding for theentire cycle.The analyses were performed using the currently installed control system settings and the RPSsettings as discussed in the WCAPs referenced in Section 2.10.Acceptance CriteriaThe acceptance criteria for the NSSS control systems are based on GDC 13, which requires thatinstrumentation and control systems be provided to monitor variables and systems over their anticipated ranges during normal operation as well as anticipated operational occurrences, and maintain thesevariables and systems within prescribed operating ranges.There should be adequate operating margin to the relevant RTS and ESFAS setpoints during andfollowing the Condition I (normal operating) transients.
All control system responses should be smoothand stable without diverging oscillations.
In addition to the limiting RTS and ESFAS setpoints, the10 percent step load decrease transient should not challenge the pressurizer PORVs or MSSV liftsetpoints.
The turbine trip without a RT from the P-9 setpoint analysis is performed to demonstrate that adequatemargin is maintained to the pressurizer PORVs actuation setpoint of 2350 psia. The results of this analysisare used to demonstrate that the requirements of NUREG-0737, item II.K.3.10 (Reference
: 1) are satisfied.
Although it is not a design requirement of the turbine trip without a RT from the P-9 permissive transient, it is desirable to ensure that the MSSVs do not lift during this transient.
This is demonstrated bymaintaining adequate margin to the first MSSV lift setting of 1185 psig (1200 psia).2.11.2.1.3 Description of Analyses and Evaluations The transients were analyzed using the LOFTRAN computer code (Reference 2). This computer code is asystem-level program code that models the overall NSSS, including detailed modeling for control andprotection systems.
A LOFTRAN computer model was developed for the WCGS. The key inputparameters and assumptions for the analyses are given in subsection 2.11.2.1.2.
5 Percent per Minute Unit Loading and Unloading The 5 percent per minute loading and unloading transients are not limiting transients and are enveloped by the 10 percent step load increase and decrease transients, respectively.
Therefore, no specific analyseswere performed for the 5 percent per minute loading and unloading transients.
10 Percent Step Load DecreaseThis transient was analyzed as a step decrease in turbine load from 100 to 90 percent power which boundslower power levels. The primary control systems that mitigate this transient are the reactor control systemand pressurizer pressure control system. The steam dump system is blocked on load decreases less than10 percent.
The 10 percent step load decrease transient should not result in challenges to the pressurizer WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-392PORVs actuation setpoint and the maximum steam pressure should not challenge the first MSSV liftsetpoint.
10 Percent Step Load IncreaseThis transient was analyzed as a step increase in turbine load from 90 to 100 percent power, which boundslower power levels. The primary control systems that mitigate this transient are the reactor control systemand pressurizer pressure control system. The 10 percent step load increase transient should not result in anautomatic actuation of an RTS or ESFAS function.
The limiting RTS and ESFAS functions are the highneutron flux, low pressurizer
: pressure, OTAT, OPAT, and low steam line pressure RTS and ESFASsetpoints.
Large Load Rejection The large load rejection is the most severe operational transient and was analyzed as a step decrease inturbine load from 100 to 50 percent power, which bounds lower power levels. The primary controlsystems that mitigate this transient are the reactor control system, pressurizer pressure control system, andsteam dump control system. The steam dump control system maintains the RCS temperature within thecontrol range until a new equilibrium condition is reached.
The RTS functions that are most limiting forthis transient are the OTAT, OPAT, low pressurizer
: pressure, and high pressurizer pressure setpoints.
Turbine Trip without Reactor Trip from P-9 SetpointThis transient was analyzed as a step change in steam flow from the P-9 setpoint of 50 to 0 percent power.A turbine trip from 50 percent power bounds all lower power levels. The analysis is performed todemonstrate that the pressurizer PORVs do not lift following a turbine trip without a RT transient from theP-9 permissive setpoint to address NUREG-0737, Item II.K.3.10 (Reference 1). The analysis is best-estimate and credits the reactor control system, pressurizer pressure control system, and steam dumpcontrol systems.2.11.2.1.4 ResultsThe results of the analyses show that the current control system setpoints and RPS settings, as discussed in the WCAPs referenced in Section 2.10, are acceptable for the TM Program and enable the plant tosatisfy the requirements of the design basis operational transients.
Ten Percent Step Load DecreaseFollowing the 10 percent step load decrease, the secondary side steam pressure and temperature initially
: increase, resulting in a Tavg and pressurizer pressure increase.
The control system automatically inserts thecontrol rods to restore Tavg to the programmed value. Pressurizer spray restores the pressurizer pressure tothe nominal value.Based on the results, a 10 percent step load decrease transient can be accommodated without challenging the pressurizer PORVs setpoint of 2350 psia and the first MSSV setpoint of 1200 psia. The maximumpressurizer pressure was 2326 psia and the maximum steam line pressure was 1060 psia. This resulted inWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-393minimum margins of 24 psi to the pressurizer PORVs and 140 psi to the nominal MSSV actuation setpoints.
The results indicated that no RTS or ESFAS setpoints were challenged and the control systemresponses were smooth and not oscillatory.
Therefore, the 10 percent step load decrease transient can besuccessfully accommodated for the TM Program.Ten Percent Step Load IncreaseFollowing the 10 percent step load increase, the secondary side steam pressure and temperature initially decrease and the Tavg and pressure also initially decrease.
The control system automatically withdraws thecontrol rods to restore Tavg to the programmed value. Pressurizer heaters restore the pressurizer pressureto the nominal value.The 50 seconds/5 seconds lead/lag compensated steam pressure reached a minimum of 668.2 psia, whichis above the low steam line pressure SI actuation setpoint of 630 psia. Therefore,
.the criterion of notchallenging the low steam line pressure SI on a design basis 10 percent step load increase transient is met.The minimum pressurizer pressure was 2211 psia, which is greater than the low pressure RTS setpoint of1955 psia. The power range neutron flux reached a maximum value of 104.4 percent, which is less thanthe RTS setpoint of 109 percent.
Acceptable margins of 9.68 and 5.0 percent were maintained to theOTAT and OPAT RTS setpoints, respectively.
The results indicated that no RTS or ESFAS setpoints werechallenged and the control system response was stable and not oscillatory.
Therefore, the 10 percent stepload increase transient can be successfully accommodated for the TM Program.Large Load Rejection Following the large load rejection, the secondary side steam pressure and temperature initially
: increase, resulting in a Tavg and pressurizer pressure increase.
The steam dump valves open to mitigate the RCStemperature increase and the reactor control system automatically inserts the control rods to decreasereactor power and restore Tavg to the programmed value. The steam dump valves modulate closed as theplant is brought to a new equilibrium condition.
Pressurizer spray and relief valves prevent the pressurefrom reaching the high pressurizer pressure RTS setpoint.
Based on the results, a 50 percent step load rejection can be sustained over the range of operating conditions.
Minimum margins of 6.02 and 8.88 percent of nominal AT were maintained to the OTAT andOPAT RT setpoints, respectively, which is acceptable.
The pressurizer PORVs open and limit the pressureto less than the high pressurizer pressure RTS setpoint of 2400 psia. Following the opening of thepressurizer PORVs, the minimum pressurizer pressure of 2139 psia maintains adequate margin to the lowpressurizer pressure RTS setpoint of 1955 psia.The results indicated that no RTS or ESFAS setpoints were challenged and the control system responsewas stable and not oscillatory.
Therefore, the large load rejection transient can be successfully accommodated for the TM Program.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-394Turbine Trip without Reactor Trip from the P-9 Permissive Following the turbine trip, the secondary side steam pressure and temperature initially
: increase, resulting in a Tavg and pressurizer pressure increase.
The steam dump valves open to mitigate the RCS temperature increase and the reactor control system automatically inserts the control rods to decrease reactor powerand Tavg. Pressurizer spray prevents the pressure from reaching the pressurizer PORVs actuation setpoint.
This transient is modeled as a step change in steam flow from the P-9 setpoint power level to 0 percentpower. The analysis was performed from the current P-9 permissive setpoint of 50 percent RTP andcovered the range of operating conditions.
The current P-9 setpoint was found to be acceptable at conditions corresponding to the high full powerTavg of 588.4&deg;F;
: however, the results indicated that the pressurizer PORVs actuation setpoint waschallenged for the case corresponding to low full- power Ta,,g (i.e., 570.7&deg;F) conditions.
Further analysesshowed that acceptable results are obtained for the current P-9 setpoint if the full-power Tavg is restricted to no lower than 575&deg;F.Therefore, the current P-9 setpoint of 50 percent RTP is acceptable for plant operation over a restricted full-power Tavg window of 575&deg;F to 588.4&deg;F.
The limiting full-power Tag for the current P-9 setpoint is5750F and the pressurizer pressure increases to a maximum pressure of 2339 psia, which results inacceptable margin of approximately 11 psi to the PORVs actuation setpoint.
The case corresponding tohigh Tavg (588.4&deg;F) conditions resulted in a maximum pressurizer pressure of 2302 psia, which is wellbelow the pressurizer PORVs actuation setpoint.
The maximum SG pressure for all cases analyzed is1071 psia, which is well below the first MSSV lift setting of 1200 psia. The analyses indicate the controlsystem response was smooth during the transient with no oscillatory response noted. Therefore, theturbine trip without RT transient from the P-9 permissive setpoint of 50 percent RTP can be successfully accommodated for the TM Program over a restricted full-power Tavg window of 575&deg;F to 588.4&deg;F.WCNOC will administratively limit full-power Tavg to greater than or equal to 575&deg;F.2.11.2.1.5 Conclusions Plant operational margin to trip analyses were performed using Westinghouse methodology as part of theWCGS TM Program.
The results of these analyses conclude that the plant response is acceptable andsufficient margin exists to the relevant RTS and ESFAS setpoints during the design basis operational transients as described in USAR 7.7.2 at the WCGS for the current and MUR power levels. The results ofthe analyses show that the current control system setpoints and RPS settings, as discussed in the WCAPsreferenced in Section 2.10, are acceptable for the TM Program and enable the plant to satisfy therequirements of the design basis operational transients.
The current P-9 permissive setpoint of 50 percent RTP is acceptable for a restricted full-power Tavgwindow between 575&deg;F and 588.4&deg;F.
The results show that a turbine trip from the P-9 setpoint will notchallenge the pressurizer PORVs actuation setpoint with all NSSS control systems operable in theautomatic mode; therefore, the requirements of NUREG-0737, item II.K.3.10 are satisfied.
The WCGSwill administratively limit full-power Tavg to greater than or equal to 575&deg;F.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3952.11.2.1.6 References
: 1. NUREG-0737, "Clarification of TMI Action Plan Requirements,"
Item II.K.3.10, ProposedAnticipatory Trip Modification, October 1980.2. WCAP-7907, "LOFTRAN Code Description,"
April 1984.2.12 THERMAL AND HYDRAULIC DESIGN2.12.1 Introduction This section describes the T/H analysis performed to support operation of the WCGS with a corecontaining 17x17 RFA-2 fuel at the nominal conditions described in Table 2.12-1.2.12.2 Input Parameters and Acceptance CriteriaFor the purposes of the WCGS methodology transition, bounding fuel-related safety and designparameters have been chosen. These bounding parameters have been used in the analyses discussed in thissection.
Table 2.12-1 summarizes the current T/H design parameters used in the DNB analyses.
Thelimiting direction for these parameters for DNB is shown in Table 2.12-2. The core inlet temperature usedin the DNB analyses is based on the upper bound of the RCS temperature range for conditions corresponding to a conservatively higher core power.The current licensing basis for T/H design for the WCGS includes the prevention of DNB on the limitingfuel rod with a 95-percent probability at a 95-percent confidence level (95/95) and criteria to ensure fuelcladding integrity.
The licensing basis is documented in USAR Section 4.4, Thermal and Hydraulic Design. The DNB analysis for the methodology transition is based on this licensing basis whileincorporating a conservatively higher core power. The analysis addresses DNB performance and theeffects of fuel rod bow, bypass flow and lower plenum flow anomalies.
2.12.2.1 Design Basis and Methodology The T/H DNB analysis of the fuel at the WCGS is based on the RTDP (Reference
: 1) and the WRB-2DNB correlation (Reference
: 2) using the Westinghouse version of the VIPRE-01 subchannel analysiscode (Reference 3). The STDP is used when RTDP is not applicable.
For analyses that are outside of therange of applicability of the WRB-2 correlation, a W-3 alternative DNB correlation (ABB-NV or WLOP)is used (Reference 4). The RTDP methodology is applicable to accidents that initiate from normaloperating conditions whereas the STDP methodology is typically applied to events that are initiated fromshutdown conditions.
The WRB-2 correlation is used for analysis of fuel regions above the first mixingvane grid whereas the ABB-NV correlation is used for analysis of fuel regions below the first mixingvane grid. The WLOP correlation is used when the coolant conditions are outside the range ofapplicability of the WRB-2 and ABB-NV correlations.
Specific methodologies and correlations forspecific events are identified in other portions of Section 2.0 of this report. The analyses demonstrate thatthe 95/95 DNB design basis is met for the core in operation at the maximum analyzed core power inTable 2.12-1.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESUNGHOUSE NON-PROPRIETARY CLASS 32-3962.12.2.1.1 Subchannel Analysis CodeThe Westinghouse version of the VIPRE-01 code (VIPRE, Reference
: 3) is used for DNBR calculations.
The use of VIPRE for the methodology transition is in full compliance with the conditions specified in theUSNRC SER in WCAP-14565-P-A (Reference
: 3) for THINC and FACTRAN replacement.
See AppendixA of this WCAP for code applicability.
2.12.2.1.2 DNB Methodology With the RTDP methodology, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, computer codes, and DNB correlation predictions are statistically considered to obtain the overall DNB uncertainty factors.
Proprietary DNBR sensitivity
: factors, which are used todevelop DNB uncertainty
: factors, are calculated over ranges of conditions that bound the events for whichRTDP methodology is applied.
Based on the DNB uncertainty
: factors, RTDP design limit DNBR valuesare determined such that there is a 95-percent probability with a 95-percent confidence level that DNBwill not occur on the most limiting fuel rod during normal operation, operational transients, or transient conditions arising from faults of moderate frequency.
The uncertainties included in the overall DNB uncertainty factor are:* nuclear enthalpy rise hot channel factor, (FNAH)0 enthalpy rise engineering hot channel factor, (FEAH)* uncertainties in the DNB correlations and the computer codes* vessel coolant flow* effective core flow fraction (1 -bypass flow fraction)
* core thermal power* coolant temperature
* system pressureInstrumentation uncertainties in core thermal power, RCS flow, pressure, and inlet temperature weretaken into account for the methodology transition.
Only the random portion of each plant operating parameter uncertainty is included in the statistical combination for RTDP. Any adverse instrumentation bias is treated either as a DNBR penalty or a direct analysis input.In addition to the above considerations for uncertainties, DNBR margin is retained by performing thesafety analyses to DNBR limits higher than the design limit DNBR values. Sufficient DNBR margin isconservatively maintained in the safety analysis DNBR limits to offset penalties for rod bow, lowerplenum flow anomalies, and plant instrumentation biases and to provide flexibility in design andoperation of the plant. Table 2.12-3 provides a summary of the DNBR margin and penalties applicable atnominal conditions.
The STDP is used for those analyses where RTDP is not applicable.
The DNBR limit for STDP is theappropriate DNB correlation limit increased by sufficient margin to offset the applicable DNBR penalties.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3972.12.2.1.3 DNB Correlations and LimitsThe WRB-2 DNB correlation is based entirely on rod bundle data and takes credit for the significant improvements in DNB performance due to the mixing vane grid effects.
USNRC acceptance of a95/95 WRB-2 correlation DNBR limit of 1.17 for 17x17 RFA-2 fuel is documented in References 2, 3,and 5.For the methodology transition, the WRB-2 RTDP design limit DNBR is 1.24 for the 17x 17 RFA-2 fuel atthe WCGS. The RTDP DNB analyses are performed to a higher DNBR limit referred to as the SALDNBR that includes additional margin. For cases in which WRB-2 is not applicable, the W-3 alternative correlations (ABB-NV or WLOP) are used as approved in Reference 4.For events in which STDP is used, the 95/95 correlation DNBR limits are 1. 17 for WRB-2, 1.13 forABB-NV, and 1.18 for WLOP.The reactor core is designed to meet the following limiting T/H criteria:
There is at least a 95-percent probability, at a 95-percent confidence level, that DNB will notoccur during any anticipated normal operating condition, operational transients, or any condition of moderate frequency.
Fuel melting will not occur during any anticipated normal operating condition, operational transients, or any conditions of moderate frequency.
Thermo-hydrodynamic instabilities will not occur during any anticipated normal operating condition, operational transients, or any conditions of moderate frequency.
DNBR is defined as the ratio of the heat flux causing DNB at a particular location in the core, aspredicted by a DNB correlation, to the actual heat flux at the same location.
Analytical assurance thatDNB will not occur is provided by showing the calculated DNBR to be higher than the 95/95 limit DNBRfor all conditions of normal operation, operational transients, and transient conditions of moderatefrequency.
The Design Limit DNBR is calculated by using the USNRC-approved RTDP methodology (Reference 1). Meeting the Design Limit assures compliance with the aforementioned DNB criteria.
A SAL DNBR, which is higher than the Design Limit DNBR, is conservatively used in safety analyses toprovide DNBR margin to offset the effect of rod bow, lower plenum flow anomalies, and plantinstrumentation biases and to provide flexibility in the design and operation of the plant.The RCS lower plenum anomaly is applicable to the WCGS. The probable cause of the flow anomaly isan unsteady, vortex flow disturbance in the reactor vessel lower plenum. The vortex restricts flow into thecore in the perturbed region and causes coolant temperature increases in the affected fuel assemblies.
Thehigher coolant temperatures then depress the local neutron fluxes due to reactivity feedback.
The flowdisturbance in the reactor vessel lower plenum also increases the overall hydraulic resistance of thereactor, and thus decreases the flow rate to all loops. The effect of the flow anomaly is a DNBR penalty,which is offset by available margin.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3982.12.3 Description of Analyses and Evaluations For the WCGS methodology transition, a DNB re-analysis was required to define new core limits, axialoffset limits, and Condition II and IV accident acceptability.
The core limits, axial offset limits, anddropped rod limit lines are generated based on the SAL DNBR and a design FN H limit of 1.65. This FNAHlimit incorporates all applicable uncertainties, including a measurement uncertainty (Reference 6), and isadjusted for the power level using the following equation:
FAN, = 1.65 * [1 + 0.3(1 -P)]where P is the fraction of full power.Various DNB analyses that were performed in support of the WCGS methodology transition are described below; all analyses were performed at a conservatively high core power. The descriptions belowsupplement the write-up already provided in previous sections.
2.12.3.1 Core Thermal LimitsThe core thermal limits are required for the generation of the OTAT and OPAT trip setpoints.
The corethermal limits define the loci of points of thermal power, primary system pressure, and coolant inlettemperature that satisfy the following criteria:
* The minimum DNBR is not less than the SAL DNBR.* The hot channel exit quality is not greater than the upper limit of the quality range of the DNBcorrelation (adjusted for the analysis-specific quality uncertainty).
* Vessel Th., < Tst to ensure that the difference between Thor and Tcold remains proportional to power.For the transition to Westinghouse methods and to support operation at a conservatively higher power,new core thermal limits were generated for the 17x 17 RFA-2 fuel at the WCGS. The DNB-limited portionof the core thermal limits was generated with the VIPRE code using the WRB-2 DNB correlation and theRTDP methodology.
2.12.3.2 Axial Offset LimitsThe axial offset limits are used to reduce the core DNB limit lines to account for the effect of adverseaxial power distributions that are more limiting for DNB than the axial power shape used to generate thecore thermal limits. For the transition to Westinghouse methods and to support operation at aconservatively higher power, new axial offset limits were generated for the 17x 17 RFA-2 fuel at theWCGS. The axial offset limits were generated with the VIPRE code using the RTDP methodology.
Forthe DNB analysis of axial power distributions that were limiting in the fuel region above the first mixingvane grid, the WRB-2 DNB correlation was used. For the DNB analysis of axial power distributions thatwere limiting in the fuel region below the first mixing vane grid, the ABB-NV DNB correlation was used.The axial offset limits were used to define the f(AI) reset function in the OTAT reactor trip function suchthat the DNB design criterion is met for accidents terminated by the OTAT reactor trip function.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-3992.12.3.3 Loss of FlowThe DNB analysis of the loss-of-flow accident was performed using RTDP for three different cases,including partial loss of flow, complete loss of flow, and UF. Each case was checked to ensure that thelimiting scenario was identified.
The effect of fuel temperatures is included in the analysis of this event.The complete loss of flow case results in the lowest minimum DNBR. The minimum DNBRs calculated for each of the three cases are greater than the SAL, thereby demonstrating compliance with the DNBdesign criterion for this event.DNB analysis was also performed to confirm that the DNB criterion was met for low flow conditions supporting the P-8 setpoint.
2.12.3.4 Locked RotorThe locked rotor accident is classified as a Condition IV event. To calculate the radiation release as aconsequence of the accident, calculations are performed using RTDP to quantify the inventory of rods thatwould experience DNB. Any rods in DNB are conservatively presumed to fail. For the WCGS, theanalysis indicates that there would be less than 1.0 percent rods in DNB due to the locked rotor accident.
The radiological consequences analysis conservatively assumes 5 percent of the fuel rods have failed andshows that the site dose limits are met.The Locked Rotor PCT analysis is performed using STDP and the VIPRE code. The acceptance criterion for this analysis is to demonstrate that the PCT is less than 2700'F. The PCT analysis for the WCGSsatisfied this acceptance criterion, thus, confirming that the fuel melt limit for ZIRLO (2700'F) highperformance fuel cladding materials is met.2.12.3.5 RCCA Drop/Misoperation This section supplements the methodology discussion in Section 2.5.3, "Control Rod Misoperation."
The USNRC-approved Westinghouse analysis methods in Reference 7 were used for analyzing the RCCAdrop event. The dropped rod limit lines were generated to define the loci of points that would result in theRTDP SAL DNBR for a wide range of core conditions (inlet temperature, power, and pressure).
Per themethodology described in Reference 8, these lines are used to verify that the DNB design basis is meteach cycle.The maximum allowable FNAH limit for RCCA misalignment was calculated using RTDP methodology.
This is the value of FNAH at normal operating conditions that results in a minimum DNBR equal to theRTDP SAL DNBR.The limits provided for the RCCA drop and RCCA misalignment events were used to confirm that theDNB design basis was met for the Wolf Creek methodology transition.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-4002.12.3.6 Steam Line Break AccidentThe event descriptions for the HZP and HFP events have been provided in Sections 2.2.5.1 and 2.2.5.2respectively.
SLB cases were analyzed at both HZP and HFP conditions.
For each of these cases, the appropriate methodology was applied.For the HFP cases, the RTDP methodology and the WRB-2 correlation are used. The DNB analysisshowed that the minimum DNBR values were above the SAL, thereby demonstrating that the DNBRdesign basis was met.For the HZP cases, STDP and the WLOP DNB correlation were applied.
The DNBR limit is thecorrelation limit increased by a small amount to account for any DNB penalties applicable at theseconditions.
The analysis showed that the minimum DNBR was greater than the DNBR limit, therebydemonstrating that the DNBR design basis was met.2.12,3.7 Feedwater Malfunction The HZP FWM event is analyzed using the same method that is used for the HZP SLB analysis.
For theWCGS methodology transition, Nuclear Design analyses indicated that the HZP FWM event wasbounded by the HZP SLB event. The DNBR design basis was met for the HZP SLB event, therebyindicating that the DNBR design basis was met for the HZP FWM event at the WCGS.2.12.3.8 Uncontrolled Rod Cluster Control Assembly Withdrawal from Subcritical Because the event is initiated from HZP conditions, the analysis for the uncontrolled RCCA withdrawal from subcritical accident is based on the STDP methodology.
Results and additional information arecontained in Section 2.5.1.This transient results in a power excursion and a bottom-skewed power shape due to the withdrawal of arod bank. A conservative accident-specific power shape was applied.
Two DNBR calculations arerequired for this accident.
The ABB-NV correlation is applied for fuel assembly spans below the firstmixing vane grid. The WRB-2 correlation is applied for spans above the first mixing vane grid. For theSTDP application, the DNBR limits applied are the correlation limits for ABB-NV and WRB-2, increased by any applicable DNBR penalties.
The results of the calculations showed that the calculated DNBRvalues remain above the respective DNBR limits, thereby demonstrating that the DNB design basis ismet.2.12.3.9 Rod Withdrawal at PowerA detailed DNB analysis of the rod withdrawal at power (RWAP) event was performed using the RTDPmethodology.
Statepoints for the limiting case (a low reactivity insertion rate case initiated from 10percent power, as described in Section 2.5.2, "Uncontrolled Rod Cluster Control Assembly BankWithdrawal at Power") were analyzed using the VIPRE code. The DNB design basis was met withmargin. Furthermore, in order to provide WCGS with adequate future flexibility in the design andWCAP- 17658-NP August 2013Licensing Report Revision 0
WESTfNGHOUSE NON-PROPRIETARY CLASS 32-401loperation of the plant relative to DNBR margin, the following credits were applied to the DNB analysisfor this particular event:* a reduced thimble bypass flow (since WCGS is currently operating with TPI), and* an increased MMF of 376,000 gpm, since WCNOC decided to raise the MMF value by5000 gpm.2.12.3.10 Bypass FlowTwo different bypass flow rates are used in the T/H design analysis.
The thermal design bypass flow is theconservatively high core bypass flow used in conjunction with the TDF in power capability analyses thatuse standard (non-statistical) methods.
The best estimate bypass flow is the core bypass flow that wouldbe expected using nominal values for dimensions and operating parameters that affect bypass flowwithout applying uncertainty factors.
The best estimate bypass flow is used in conjunction with the vesselMMF for power capability analyses using the RTDP design procedures.
As discussed inSection 2.12.2.1.2, for RTDP, the bypass flow uncertainty is included in the statistical combination for theRTDP design limit DNBR.2.12.3.11 Effects of Fuel Rod Bow on DNBRRod bow can occur between mid-grids, reducing the spacing between adjacent fuel rods and reducing themargin to DNB. Rod bow must be accounted for in the DNB safety analysis of Condition I andCondition II events. Westinghouse has conducted tests to determine the impact of rod bow on DNBperformance.
The testing and subsequent analyses are documented in References 9 and 10.The rod bow penalties are included in the DNBR margin summary shown in Table 2.12-3. In the spanscontaining IFM grids in the 17x17 RFA-2 fuel, no rod bow penalty is necessary due to the short spacingbetween grids. The maximum rod bow penalty accounted for in the safety analysis is a function ofassembly average bumup (References 9 and 10). Credit may also be taken for the effect of F AHburndown due to the decrease in fissionable isotopes and the buildup of fission products (Reference 11).2.12.3.12 License Renewal Impact Evaluation A review of T/H design for impact on plant license renewal evaluations was not necessary becausecontinued applicability of the safety analysis for the Westinghouse 17x 17 RFA-2 fuiel assemblies isre-evaluated during the RE process for each reload cycle. The reload design methodology includesevaluation of the reload core key safety parameters that comprise input to the safety evaluation for eachreload cycle.2.12.4 ResultsAnalyses described in the previous sections show that the DNB design basis is met for the WCGSmethodology transition.
The DNBR limits and margin summary are listed in Table 2.12-3. Cycle specificevaluation is to be performed in accordance with Reference 8.WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-4022.12.5 Conclusion WCNOC has reviewed the analyses related to the effects of the proposed methodology transition on theT/H design of the core and the RCS. WCNOC concludes that the analyses demonstrated that the design(1) has been accomplished using acceptable analytical
: methods, (2) is equivalent to proven designs,(3) provides acceptable margins of safety from conditions that would lead to fuel damage during normalreactor operation and AOOs, and (4) is not susceptible to thenno-hydrodynamic instability.
WCNOCfurther concludes that the analyses have adequately accounted for the effects of the proposedmethodology transition on the hydraulic loads on the core and RCS components.
Based on this, WCNOCconcludes that the T/H design will continue to meet the requirements of GDCs 6 and 7 following implementation of the proposed methodology transition.
Therefore, WCNOC finds the proposedmethodology transition acceptable with respect to T/H design.2.12.6 References
: 1. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"
April 1989.2. WCAP-10444-P-A, "Reference Core Report -VANTAGE 5 Fuel Assembly,"
September 1985.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
October 1999.4. WCAP-14565-P-A Addendum 2-P-A, "Extended Application of ABB-NV Correlation andModified ABB-NV Correlation WLOP for PWR Low Pressure Applications,"
April 2008.5. LTR-NRC-02-55, "Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design,Revision 1," November 2002.6. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties,"
June 1988.7. WCAP-1 1394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.8. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"
July 1985.9. WCAP-8691-RI "Fuel Rod Bow Evaluation,"
July 1979.10. Letter from Rahe, E. P., Jr. (Westinghouse) to Miller, J. R. (USNRC),
"Partial Response toRequest Number 1 for Additional Information on WCAP-8691, Revision 1," NS-EPR-2515, October 9, 1981; and Letter from Rahe, E. P., Jr. (Westinghouse) to Miller, J. R. (USNRC),"Remaining Response to Request Number 1 for Additional Information on WCAP-8691, Revision 1," NS-EPR-2572, March 16, 1982.11. Letter from Berlinger, C. (USNRC) to Rahe, E. P., Jr. (Westinghouse),
"Request for Reduction inFuel Assembly Burnup Limit for Calculation of Maximum Rod Bow Penalty,"
NS-NRC-85-3901 NRC Response, June 18, 1986.WCAP-1 7658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-403Table 2.12-1 T/H Design Parameters Comparison Methods Transition T/H Design Parameters Current Design Value Analysis ValueReactor Core Heat Output, MWt 3565 3637Reactor Core Heat Output, 106 BTU/hr 12,164 12,410Heat Generated in Fuel, % 97.4 97.4Core Exit Pressure,
: Nominal, psia 2270 2270Pressurizer Pressure.,
: Nominal, psia 2250 2250Radial Power Distribution, FNAH I1 1.65[ 1+0.3(1-P)]
1.65[ 1+0.3(1 -P)]HFP Nominal Coolant Conditions (uncertainties and biases not included)
Vessel TDF Rate (including bypass)106 Ibm/hr 134.7 134.9gpm 361,200 361,200Core Flow Rate (excluding bypass)12)106 lbm/hr 123.4 123.6gpm 330,859 330,859Core Flow Area, ft2  51.08 51.08Core Inlet Mass Velocity, 106 Ibm/hr-ft 2  2.416 2.419Nominal Vessel/Core Inlet Temperature, IF 555.8 555.2Vessel Average Temperature, IF 588.4 588.4Core Average Temperature.,
F 593.2 593.4Vessel Outlet Temperature, IF 621.0 621.7Core Outlet Temperature, IF 626.2 627.0Average Temperature Rise in Vessel, IF 65.2 66.5Average Temperature Rise in Core, OF 70.4 71.8Heat TransferActive Heat Transfer Surface Area, ft2  59,742 59,742Average Heat Flux, BTU/hr-ft 2  198,315 202,326Average Linear Power, kW/ft 5.691 5.806Peak Linear Power for Normal Operation, 3" kW/ft 14.23 14.52WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-404Table 2.12-1 T/H Design Parameters Comparison (cont.)Methods Transition T/H Design Parameters Current Design Value Analysis ValuePeak Linear Power for Prevention of Centerline Melt, 22.4 22.4kW/ftPressure Drop Across Core, psi(4) 28.7 28.7Notes:Thermal Power1. P=Rated Thermal Power2. A design bypass flow of 8.4 percent was used.3. Based on maximum FQ of 2.5.4. The core pressure drop calculations are based on the same best estimate flows for 3565 MWt and 3637 MWt and full coresof 17x 17 RFA-2 fuel.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-405Table 2.12-2 Limiting Parameter Direction for DNBParameter Limiting Direction for DNBF NAH, nuclear enthalpy rise hot-channel factor maximumHeat generated in fuel (%) maximumReactor core heat output (MWt) maximumHeat flux (BTU/hr-ft
: 2) maximumVessel/core inlet temperature
('F) maximumCore pressure (psia) minimumPressurizer pressure (psia) minimumTDF for non-RTDP analyses (gpm) minimumMMF for RTDP analyses (gpm) minimumBypass flow maximumWCAP-l 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 32-406Table 2.12-3 RTDP DNBR Margin SummaryCurrent Operation with 17x17 RFA-2 fuel(3565 MWt)t)DNB Correlation WRB-2DNBR Correlation Limit 1.17DNBR Design Limit(2) 1.24Total DNBR Penalties (due to rod bow, instrumentation biases, 13.6%lower plenum flow anomaly and RWAP)Total DNBR Margin(3)
> 13.6%Notes:1. DNBR analyses for the WCGS methodology transition were performed at a bounding power level of 3637 MWt.2. Design limit DNBR calculations are based on the measurement uncertainties and the sensitivity to changes in theparameters.
: 3. DNBR margin is the margin that exists between the SAL and the Design Limit DNBRs.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-1APPENDIX ASAFETY EVALUATION REPORT COMPLIANCE A.1 SAFETY EVALUATION REPORT COMPLIANCE INTRODUCTION This Appendix is a summary of USNRC-approved codes used in the LR. This appendix addresses compliance with the limitations, restrictions, and conditions specified in the approving safety evaluation of the applicable codes.Table A. 1-1 presents an overview of the SER by codes. For each SER, the applicable report subsections and Appendix A subsections are listed.Table A.1-1 Safety Evaluation Report Compliance SummaryLimitation, Licensing Topical Report Restriction, Report Appendix ANo. Subject (Reference)
Code(s) Condition Section Section1. Non-LOCA WCAP-7908-A FACTRAN Yes 2.5.1 A.2Thermal (Reference A. 1-1) 2.5.6Transients
: 2. Non-LOCA WCAP-14882-P-A RETRAN Yes 2.2.1 A.3Safety Analysis (Reference A. 1-2) 2.2.22.2.32.2.42.2.5.12.2.5.22.3.12.3.22.3.32.3.42.4.12.4.22.5.22.6.12.6.22.7.13. Non-LOCA WCAP-7907-P-A LOFTRAN Yes 2.5.2 A.4Safety Analysis (Reference A. 1-3) 2.5.32.84. Non-LOCA WCAP-1 1397-P-A RTDP Yes 2.12 A.6Thermal / Reference A. 1-5Hydraulics
: 5. Neutron Kinetics WCAP-7979-P-A TWINKLE None for 2.5.1 Not(Reference A. 1-4) Non-LOCA
====2.5.6 Applicable====
Transient AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-2Table A.1-1 Safety Evaluation Report Compliance Summary(cont.)Limitation, Licensing Topical Report Restriction, Report Appendix ANo. Subject (Reference)
Code(s) Condition Section Section6. Multi- WCAP-10965-P-A ANC None for 2.2.2 Notdimensional (Reference A. 1-6) Non-LOCA
====2.2.4 Applicable====
Neutronics Transient 2.2.5Analysis 2.5.37. Non-LOCA WCAP- 14565-P-A VIPRE Yes 2.2.2 A.5Thermal / (Reference A. 1-7) 2.2.4Hydraulics 2.2.52.4.12.4.22.5.12.5.22.5.32.128. Steam Generator WCAP- 10698-P-A RETRAN None for 2.7.2 NotTube Rupture (Reference A.1-8) Steam 2.7.3 Applicable Generator WCAP-14882-P-A Tube A.3(Reference A. 1-2) RuptureReferences A. 1-1 WCAP-7908-A, "FACTRAN
-A FORTRAN IV Code for Thermal Transients in a UO2Fuel Rod," H. G. Hargrove, December 1989.A. 1-2 WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"
D. S. Huegel, et al., April 1999.A.1-3 WCAP-7907-P-A, "LOFTRAN Code Description,"
T. W. T. Burnett, et al., April 1984.A. 1-4 WCAP-7979-P-A, "TWINKLE
-A Multi-Dimensional Neutron Kinetics Computer Code,"D. H. Risher, Jr. and R. F. Barry, January 1975.A. 1-5 WCAP-I 1397-P-A, "Revised Thermal Design Procedure,"
A. J. Friedland and S. Ray, April 1989.A. 1-6 WCAP- 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," Y. S. Liu,et al., September 1986.A.1-7 WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
Y. X. Sung, et al., October 1999.A. 1-8 WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,"
R. N. Lewis, et al., August 1987.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-3WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3A.2 FACTRAN FOR NON-LOCA THERMAL TRANSIENTS Table A.2-1 FACTRAN for Non-LOCA Thermal Transients Limitations, Restrictions, and Conditions
: 1. "The fuel volume-averaged temperature or surface temperature can be chosen at a desired value whichincludes conservatisms reviewed and approved by the USNRC."Justification The FACTRAN code was used in the analyses of the following two transients for the WCGS:Uncontrolled RCCA Withdrawal from a Subcritical Condition (USAR Section 15.4.1) and RCCA Ejection(USAR Section 15.4.8).
Initial fuel temperatures used as FACTRAN input in the RCCA Ejection analysiswere calculated using the USNRC-approved PAD 4.0 computer code as described in WCAP-15063-P-A Revision 1 (Reference A.2-1). As indicated in WCAP-15063-P-A Revision 1., the USNRC has approvedthe method of determining uncertainties for PAD 4.0 fuel temperatures.
: 2. "Table 2 presents the guidelines used to select initial temperatures."
Justification In summary, Table 2 of the SER specifies that the initial fuel temperatures assumed in the FACTRANanalyses of the following transients should be "High" and include uncertainties:
Loss of Flow, LockedRotor, and Rod Ejection.
As discussed above, fuel temperatures were used as input to the FACTRAN codein the RCCA Ejection analysis for the WCGS. The assumed fuel temperatures, which were calculated using the PAD 4.0 computer code (Reference A.2-1), include uncertainties and are conservatively high.FACTRAN was not used in the Loss of Flow and Locked Rotor analyses.
: 3. "The gap heat transfer coefficient may be held at the initial constant value or can be varied as afunction of time as specified in the input."Justification The gap heat transfer coefficients applied in the FACTRAN analyses are consistent with SER Table 2. Forthe RCCA Withdrawal from a Subcritical Condition transient, the gap heat transfer coefficient is kept at aconservative constant value throughout the transient; a high constant value is assumed to maximize thepeak heat flux (for DNB concerns) and a low constant value is assumed to maximize fuel temperatures.
For the RCCA Ejection transient, the initial gap heat transfer coefficient is based on the predicted initialfuel surface temperature, and is ramped rapidly to a very high value at the beginning of the transient tosimulate clad collapse onto the fuel pellet.4. "...the Bishop-Sandberg-Tong correlation is sufficiently conservative and can be used in theFA CTRAN code. It should be cautioned that since these correlations are applicable for local conditions only, it is necessary to use input to the FA CTRAN code which reflects the local conditions.
If the inputvalues reflecting average conditions are used, there must be sufficient conservatism in the input valuesto make the overall method conservative."
Justification Local conditions related to temperature, heat flux, peaking factors and channel information were input toFACTRAN for each of the two transients analyzed for the WCGS: RCCA Withdrawal from a Subcritical Condition (USAR Section 15.4.1) and RCCA Ejection (USAR Section 15.4.8).
Therefore, additional justification is not required.
WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-4Table A.2-1 FACTRAN for Non-LOCA Thermal Transients (cont.)Limitations, Restrictions, and Conditions
: 5. "The fuel rod is divided into a number of concentric rings. The maximum number of rings used torepresent the fuel is 10. Based on our audit calculations we require that the minimum of 6 should beused in the analyses."
Justification At least 6 concentric rings were assumed in FACTRAN for each of the two transients analyzed for theWCGS: RCCA Withdrawal from a Subcritical Condition (USAR Section 15.4.1) and RCCA Ejection(USAR Section 15.4.8).
Therefore, additional justification is not required.
: 6. "Although time-independent mechanical behavior (e.g., thermal expansion, elastic deformation) of thecladding are considered in FACTRAN, time-dependent mechanical behavior (e.g., plastic deformation) is not considered in the code....for those events in which the FACTRAN code is applied (see Table 1),significant time-dependent deformation of the cladding is not expected to occur due to the shortduration of these events or low cladding temperatures involved (where DNBR Limits apply), or the gapheat transfer coefficient is adjusted to a high value to simulate clad collapse onto the fuel pellet."Justification The two transients that were analyzed with FACTRAN for the WCGS (RCCA Withdrawal from aSubcritical Condition (USAR Section 15.4.1) and RCCA Ejection (USAR Section 15.4.8))
are included inthe list of transients provided in Table 1 of the SER; each of these transients is of short duration.
For theRCCA Withdrawal from a Subcritical Condition transient, relatively low cladding temperatures areinvolved, and the gap heat transfer coefficient is kept constant throughout the transient.
For the RCCAEjection transient, a high gap heat transfer coefficient is applied to simulate clad collapse onto the fuelpellet. The gap heat transfer coefficients applied in the FACTRAN analyses are consistent with SERTable 2.7. "The one group diffusion theory model in tire FACTRAN code slightly overestimates at beginning oflife (BOL) and underestimates at end of life (EOL) the magnitude offlux depression in the fuel whencompared to the LASER code predictions for the same fuel enrichment.
The LASER code uses transport theory. There is a difference of about 3 percent in the flux depression calculated using these two codes.When JT(centerline)
-T(Surface)l is on the order of 3000F, which can occur at the hot spot, thedifference between the two codes will give an error of 100lF. When the fuel surface temperature isfixved, this will result in a 100VF lower prediction of the centerline temperature in FA CTRAN. We haveindicated this apparent nonconservatism to Westinghouse.
In the letter NS-TMA-2026, datedJanuary 12, 1979, Westinghouse proposed to incorporate the LASER-calculated power distribution shapes in FACTRAN to eliminate this non-conservatism.
We find the use of the LASER-calculated power distribution in the FACTRAN code acceptable."
Justification The condition of concern (T(centerline)
-T(surface) on the order of 3,000&deg;F) is expected for transients that reach, or come close to, the fuel melt temperature.
As this applies only to the RCCA ejection transient, the LASER-calculated power distributions were used in the FACTRAN analysis of the RCCA ejectiontransient for the WCGS.Reference A.2-1 WCAP- 15063-P-A (Proprietary) and WCAP- 15064-NP-A (Non-Proprietary),
Revision 1 (withErrata) "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0),"July 2000.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRJETARY CLASS 3A-5A.3 RETRAN FOR NON-LOCA SAFETY ANALYSISTable A.3-1 RETRAN for Non-LOCA Safety AnalysisLimitations, Restrictions, and Conditions I. "The transients and accidents that Westinghouse proposes to analyze with RETRAN are listed in thisSER (Table 1) and the NRC staff review of RE TRAN usage by Westinghouse was limited to this set. Useof the code for other analytical purposes will require additional justification."
Justification The transients listed in Table I of the SER are:* Feedwater system malfunctions
" Excessive increase in steam flow" Inadvertent opening of a steam generator relief or safety valve* Steam line break* Loss of external load/turbine trip* Loss of offsite power* Loss of normal feedwater flow* Feedwater line rupture* Loss of forced reactor coolant flow* Locked reactor coolant pump rotor/sheared shaft* Control rod cluster withdrawal at power* Dropped control rod cluster/dropped control bank* Inadvertent increase in coolant inventory
* Inadvertent opening of a pressurizer relief or safety valve* Steam generator tube ruptureThe transients explicitly analyzed for the WCGS using RETRAN are:* Feedwater system malfunctions (USAR Sections
: 15. 1.1 and 15.1.2),* Excessive increase in secondary steam flow (USAR Section 15.1.3),* Inadvertent opening of a steam generator atmospheric relief or safety valve(USAR Section 15.1.4),* Steam system piping failure (steam line break) (USAR Section 15.1.5),* Loss of external electrical load/turbine trip (USAR Sections 15.2.2, 15.2.3, 15.2.4, and 15.2.5),* Loss of non-emergency alternating current (AC) power to the station auxiliaries (loss of offsitepower) (USAR Section 15.2.6)," Loss of normal feedwater flow (USAR Section 15.2.7)," Feedwater system pipe break (feedwater line rupture)
(USAR Section 15.2.8)," Loss of forced reactor coolant flow (USAR Sections 15.3.1 and 15.3.2),* Locked reactor coolant pump rotor/shaft break (USAR Sections 15.3.3 and 15.3.4),* Uncontrolled RCCA bank withdrawal at power (USAR Section 15.4.2),* Inadvertent operation of the ECCS (increase in coolant inventory)
(USAR Section 15.5.1),* CVCS malfunction that increases reactor coolant inventory (USAR Section 15.5.2),* Inadvertent opening of a pressurizer safety or relief valve (USAR Section 15.6. 1),* Steam generator tube rupture (USAR Section 15.6.3).As each transient analyzed for the WCGS using RETRAN matches one of the transients listed in Table 1of the SER, additional justification is not required.
WCAP- 17658-NP August 2013Licensing Report Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-6Table A.3-1 RETRAN for Non-LOCA Safety Analysis(cont.)Limitations, Restrictions, and Conditions
: 2. "WCAP-14882 describes modeling of Westinghouse designed 4-, 3, and 2-loop plants of the type thatare currently operating.
Use of the code to analyze other designs, including the Westinghouse AP600,will require additional justification."
Justification The WCGS consists of one 4-loop Westinghouse-designed unit that was "currently operating" at the timethe SER was written (February 11, 1999). Therefore, additional justification is not required.
: 3. "Conservative safety analyses using RETRAN are dependent on the selection of conservative input.Acceptable methodology for developing plant-specific input is discussed in WCAP-14882 and inReference 14 [WCAP-92 72-P-Al.
Licensing applications using RETRAN should include the source ofand justification for the input data used in the analysis."
Justification The input data used in the RETRAN analyses performed by Westinghouse came from both WCNOC andWestinghouse sources.
Assurance that the RETRAN input data is conservative for the WCGS is providedvia Westinghouse's use of transient-specific analysis guidance documents.
Each analysis guidancedocument provides a description of the subject transient, a discussion of the plant protection systems thatare expected to function, a list of the applicable event acceptance
: criteria, a list of the analysis inputassumptions, e.g., directions of conservatism for initial condition values, a detailed description of thetransient model development method, and a discussion of the expected transient analysis results.
Based onthe analysis guidance documents, conservative plant-specific input values were requested and collected from the responsible WCNOC and Westinghouse sources.
Consistent with the Westinghouse ReloadEvaluation Methodology described in WCAP-9272 (Reference A.3-1), the safety analysis input valuesused in the WCGS analyses were selected to conservatively bound the values expected in subsequent operating cycles.Reference A.3-1 WCAP-9272-P-A (Proprietary) and WCAP-9273-NP-A (Non-Proprietary),
"Westinghouse Reload Safety Evaluation Methodology,"
July 1985.WCAP-17658-NP Licensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-7A.4 LOFTRAN FOR NON-LOCA SAFETY ANALYSISTable A.4-1 LOFTRAN for Non-LOCA Safety AnalysisLimitations, Restrictions, and Conditions
: 1. "LOFTRAN is used to simulate plant response to many of the postulated events reported in Chapter 15of PSARs and FSARs, to simulate anticipated transients without scram, for equipment sizing studies,and to define mass/energy releases for containment pressure analysis.
The Chapter 15 events analyzedwith LOFTRAN are:* Feedwater System Malfunction
* Excessive Increase in Steam Flow" Inadvertent Opening of a Steam Generator Relief or Safety Valve* Steamline Break* Loss of External Load* Loss of Offsite Power* Loss of Normal Feedwater
* Feedwater Line Rupture* Loss of Forced Reactor Coolant Flow* Locked Pump Rotor* Rod Withdrawal at Power* Rod Drop* Startup of an Inactive Pump* Inadvertent ECCS Actuation
* Inadvertent Opening of a Pressurizer Relief or Safety ValveThis review is limited to the use of LOFTRANfor the licensee safety analyses of the Chapter 15 eventslisted above, and for a steam generator tube rupture..."
Justification For the WCGS, the LOFTRAN code was used in the analysis of the uncontrolled RCCA bank withdrawal at power transient (USAR Section 15.4.2),
in the analysis of the dropped rod transient (USAR Section 15.4.3),
and in the analysis of the anticipated transients without scram(USAR Section 15.8). As each of these transients match one of the transients listed in the SER, additional justification is not required.
WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-8WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-8A.5 VIPRE FOR NON-LOCA THERMAL/HYDRAULICS Table A.5-1 VIPRIE for Non-LOCA Thermal/Hydraulics Limitations, Restrictions, and Conditions 1."Selection of the appropriate CHF correlation, DNBR limit, engineered hot channel factors forenthalpy rise and other fuel-dependent parameters for a specific plant application should be justified with each submittal."
Justification The WRB-2 correlation with a 95/95 correlation limit of 1.17 was used in the DNB analyses for the WCGS17x 17 RFA-2 fuel. The use of the WRB-2 DNB correlation was approved in WCAP- 10444-P-A (Reference A.5-2). Applicability of the WRB-2 to 17x17 RFA-2 fuel was established through the FuelCriterion Evaluation Process (FCEP) in LTR-NRC-02-55 (Reference A.5-3). For conditions where WRB-2is not applicable, analyses were performed using approved secondary CHF correlations (such as ABB-NVand WLOP) in compliance with the SER conditions licensed for use in the VIPRE code.(WCAP-14565-P-A and its Addendum 2-P-A, Reference A.5-4).The use of the plant specific hot channel factors and other fuel dependent parameters in the DNB analysisfor the WCGS 17x 17 RFA-2 fuel were justified using the same methodologies as for previously approvedsafety evaluations of other Westinghouse four-loop plants using the same fuel design.2. "Reactor core boundary conditions determined using other computer codes are generally input intoVIPRE for reactor transient analyses.
These inputs include core inlet coolant flow and enthalpy, coreaverage power, power shape and nuclear peaking factors.
These inputs should be justified asconservative for each use of VIPRE."Justification The core boundary conditions for the VIPRE calculations for the 17x 17 RFA-2 fuel are all generated fromUSNRC-approved codes and analysis methodologies.
Conservative reactor core boundary conditions werejustified for use as input to VIPRE. Continued applicability of the input assumptions is verified on acycle-by-cycle basis using the Westinghouse reload methodology described in WCAP-9272-P-A (Reference A.5-1).3. "The NRC Staffs generic SER for VIPRE set requirements for use of new CHF correlations withVIPRE. Westinghouse has met these requirements for using WRB-1, WRB-2 and WRB-2Mcorrelations.
The DNBR limit for WRB-1 and WRB-2 is 1.17. The WRB-2M correlation has a DNBRlimit of 1.14. Use of other CHF correlations rot currently included in VIPRE will require additional justification."
Justification As discussed in response to Condition 1, the WRB-2 correlation with a limit of 1.17 was used as theprimary correlation in the DNB analyses of 17x 17 RFA-2 fuel for WCGS. For conditions where theWRB-2 is not applicable, analyses were performed using approved secondary CHF correlations licensedfor the VIPRE code in Reference A.5-4.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-9WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-9Table A.5-1 VIPRE for Non-LOCA Thermal/Hydraulics (cont.)Limitations, Restrictions, and Conditions
: 4. "Westinghouse proposes to use the VIPRE code to evaluate fuel performance folow#ving postulated design-basis accidents, including beyond-CHF heat transfer conditions.
These evaluations arenecessary to evaluate tihe extent of core damage and to ensure that the core maintains a coolablegeometry in the evaluation of certain accident scenarios.
The NRC Staffs generic review of VIPRE didnot extend to post CHF calculations.
VIPRE does not model the time-dependent physical changes thatmay occur within the fuel rods at elevated temperatures.
Westinghouse proposes to use conservative input in order to account for these effects.
The NRC Staff requires that appropriate justification besubmitted with each usage of VIPRE in the post-CHF region to ensure that conservative results areobtained."
Justification For application to Wolf Creek safety analysis, the use of VIPRE in the post-critical heat flux region islimited to the PCT calculation for the locked rotor transient.
The calculation demonstrated that the PCT inthe reactor core is well below the allowable limit to prevent clad embrittlement.
VIPRE modeling of thefuel rod is consistent with the model described in WCAP-14565-P-A (Reference A.5-4) and included thefollowing conservative assumptions:
* DNB was assumed to occur at the beginning of the transient,
* Film boiling was calculated using the Bishop-Sandberg-Tong correlation,
* The Baker-Just correlation accounted for heat generation in fuel cladding due to zirconium-water reaction.
Conservative results were further ensured with the following input:* Fuel rod input based on the maximum fuel temperature at the given power,* The hot spot power factor was equal to or greater than the design linear heat rate,* Uncertainties were applied to the initial operating conditions in the limiting direction.
References A.5-1 WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"
S. L. Davidson (Ed.),July 1985.A.5-2 WCAP-10444-P-A, "Reference Core Report -VANTAGE 5 Fuel Assembly,"
S. L. Davidson(Editor),
September 1985.A.5-3 LTR-NRC-02-55, "Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design,Revision 1," November 2002.A.5-4 WCAP- 14565-P-A Addendum 2-P-A, "Extended Application of ABB-NV Correlation andModified ABB-NV Correlation WLOP for PWR Low Pressure Applications,"
A. Leidich,et. al., April 2008.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-10A.6 REVISED THERMAL DESIGN PROCEDURE FOR NON-LOCA THERMALHYDRAULICS Table A.6-1 Revised Thermal Design Procedure for Non-LOCA Thermal Hydraulics Limitations, Restrictions, and Conditions
: 1. "Sensitivity factors for a particular plant and their ranges of applicability should be included in theSafety Analysis Report or reload submittal.
Justification Sensitivity factors were evaluated using the WRB-2 and ABB-NV correlations and the VIPRE code forparameter values applicable to the 17x 17 RFA-2 fuel at conditions corresponding to a conservatively higher nominal core power of 3637 MWt. These sensitivity factors were used to determine the RTDPdesign limit DNBR values which are to be included in the WCGS USAR.2. "Anj' changes in DNB correlation, THINC-IV correlations, orparameter values listed in Table 3-1 ofWCAP-11397 outside of previously demonstrated acceptable ranges require re-evaluation of thesensitivity factors and of the use of Equation (2-3) of the topical report."Justification Because the VIPRE code was used to replace the THINC-IV code, sensitivity factors were evaluated forusing the VIPRE code. VIPRE has been demonstrated to be equivalent to the THINC-IV code inWCAP-14565-P-A (Reference A.6-1). See the response to condition 3 for a discussion of the use ofEquation (2-3) of the topical report. Evaluations using both WRB-2 and ABB-NV correlations were donein compliance with the methodology described in WCAP- 11397-P-A (Reference A.6-2).3. "If the sensitivity factors are changed as a result of correlation changes or changes in the application oruse of the THINC code, then the use of an uncertainty allowance for application of Equation (2-3) mustbe re-evaluated and the linearity assumption made to obtain Equation (2-17) of the topical report mustbe validated.
Justification Equation (2-3) of WCAP-I 1397-P-A (Reference A.6-2) and the linearity approximation made to obtainEquation (2-17) were confirmed to be valid for the WCGS using the combination of the VIPRE code andthe WRB-2 and ABB-NV correlations, at conditions corresponding to a conservatively higher nominalcore power of 3637 MWt.4. "Variances and distributions for input parameters must be justified on a plant-by-plant basis untilgeneric approval is obtained."
Justification The plant specific variances and distributions were justified for use at conditions corresponding to aconservatively higher nominal core power of 3637 MWt and are presented in Section 2.12 of the LR.5. "Nominal initial condition assumptions apply only to DNBR analyses using RTDP. Other analyses, such as overpressure calculations, require the appropriate conservative initial condition assumptions."
Justification Nominal conditions were only applied to the DNBR analyses which used RTDP.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0
WESTINGHOUSE NON-PROPRIETARY CLASS 3A-11Table A.6-1 Revised Thermal Design Procedure for Non-LOCA Thermal Hydraulics (cont.)Limitations, Restrictions, and Conditions
: 6. "Nominal conditions chosen for use in analyses should bound all permitted methods of plant operation.
Justification Bounding nominal conditions corresponding to a conservatively higher nominal core power of 3637 MWtwere used in the DNBR analyses using RTDP, consistent with the current methods of plant operation atWCGS.7. "The code uncertainties specified in Table 3-1 (of WCAP-1139 7-P) (+/- 4 percent for THINC-IV and+/- 1 percent for transients) must be included in the DNBR analyses using R TDP."Justification The code uncertainties specified in Table 3-1 of WCAP-1 1397-P-A (Reference A.6-2) remainedunchanged and were included in the DNBR analyses using RTDP. The THFNC-IV uncertainty was appliedto VIPRE, based on the equivalence of the VIPRE model approved in WCAP-14565-P-A to THINC-IV.
References A.6-1 WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"
Y. X. Sung, et al., October 1999.A.6-2 WCAP-1 1397-P-A, "Revised Thermal Design Procedure,"
Friedland, A. J. and Ray, S.,April 1999.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0}}

Revision as of 08:35, 4 July 2018

WCAP-17658-NP, Rev. 0, Wolf Creek Generating Station Transition of Methods for Core Design and Safety Analyses - Licensing Report
ML13247A077
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 08/31/2013
From: Solomon D
Westinghouse
To:
Office of Nuclear Reactor Regulation
References
ET 13-0023 WCAP-17658-NP, Rev 0
Download: ML13247A077 (455)
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Text

Enclosure I to ET 13-0023WCAP-17658-NP, "Wolf Creek Generating Station Transition of Methods for Core Designand Safety Analyses

-Licensing Report" (Non-Proprietary)

(454 pages)

Westinghouse Non-Proprietary Class 3WCAP-17658-NP August 20'Revision 0Wolf Creek Generating StationTransition of Methods forCore Design and SafetyAnalyses

-Licensing ReportWestinghouse 13 WESTINGHOUSE NON-PROPRIETARY CLASS 3WCAP-17658-NP Revision 0Wolf Creek Generating StationTransition of Methods for Core Design andSafety Analyses

-Licensing ReportAugust 2013Prepared:

Denise Solomon*,

EngineerEngineering, Equipment and Major ProjectsPlant Licensing Approved:

Dewey Olinski*,

ManagerEngineering, Equipment and Major ProjectsPlant Licensing

  • Electronically approved records are authenticated in the electronic document management system.Westinghouse Electric Company LLC1000 Westinghouse DriveCranberry
Township, PA 16066, USA© 2013 Westinghouse Electric Company LLCAll Rights Reserved WESTINGHOUSE NON-PROPRIETARY CLASS 3ivTABLE OF CONTENTSL IS T O F T A B L E S ......................................................................................................................................

v iiL IS T O F F IG U R E S ......................................................................................................................................

xLIST OF ACRONYM S ..............................................................................................................................

xixPROFESSIONAL ENGINEERING STAMPS .........................................................................................

xxiiINTRODUCTION

........................................................................................................................

1-11.1 NUCLEAR STEAM SUPPLY SYSTEM PARAM ETERS ............................................

1-21.1.1 In tro du ctio n .....................................................................................................

1-21.1.2 Input Parameters, Assumptions, and Acceptance Criteria

...............................

1-21.1.3 Description of Analyses and Evaluation

..........................................................

1-31.1.4 C on clu sion s .....................................................................................................

1-42 ACCIDENT AND TRANSIENT ANALYSIS

..............................................................................

2-12.1 NON-LOCA ANALYSES INTRODUCTION

................................................................

2-12.1.1 Program Features

.............................................................................................

2-12.1.2 Non-LOCA Transient Events Considered

.......................................................

2-22.1.3 Analysis M ethodology

.....................................................................................

2-52.1.4 Computer Codes Used .....................................................................................

2-82.1.5 Initial Conditions

...........................................................................................

2-102.1.6 Fuel Design Description

................................................................................

2-122.1.7 Power Distribution Peaking Factors .........................................................

2-132.1.8 Reactivity Feedback

......................................................................................

2-132.1.9 Pressure Relief M odeling ..........................................................................

2-132.1.10 RTS and ESFAS Functions

............................................................................

2-152.1.11 Reactor Trip Characteristics

..........................................................................

2-162.1.12 Operator Actions Credited

.............................................................................

2-162.1.13 Results Summary ...........................................................................................

2-172 .1.14 R eferences

.....................................................................................................

2-172.2 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM .....................

2-372.2.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (USAR Section 15.1.1) .............................................................

2-372.2.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow (USAR Section 15.1.2) .........................................................................

2-452.2.3 Excessive Increase in Secondary Steam Flow (USAR Section 15.1.3) .........

2-542.2.4 Inadvertent Opening of a Steam Generator Atmospheric Relief or SafetyValve (USAR Section 15.1.4) ........................................................................

2-682.2.5 Steam System Piping Failure (USAR Section 15.1.5) ..................................

2-81WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3VTABLE OF CONTENTS (cont.)2.3 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM ..................

2-1122.3.1 Loss of External Electrical Load, Turbine Trip, Inadvertent Closure ofMain Steam Isolation Valves, Loss of Condenser Vacuum and OtherEvents Resulting in Turbine Trip (USAR Sections 15.2.2, 15.2.3,15 .2 .4 , and 15 .2 .5) .......................................................................................

2-1122.3.2 Loss of Non-Emergency AC Power to the Station Auxiliaries (USARS ection 15 .2 .6) .............................................................................................

2-1282.3.3 Loss of Normal Feedwater Flow (USAR Section 15.2.7) ...........................

2-1442.3.4 Feedwater System Pipe Break (USAR Section 15.2.8) ...............................

2-1612.4 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE ...........................

2-1792.4.1 Partial and Complete Loss of Forced Reactor Coolant Flow(U SA R Sections 15.3.1 and 15.3.2) .............................................................

2-1792.4.2 Reactor Coolant Pump Shaft Seizure (Locked Rotor) and Shaft Break(U SA R Sections 15.3.3 and 15.3.4) .............................................................

2-2052.5 REACTIVITY AND POWER DISTRIBUTION ANOMALIES

...............................

2-2242.5.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from aSubcritical or Low Power Startup Condition (USAR Section 15.4.1) ........

2-2242.5.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal atPow er (U SA R Section 15.4.2) .....................................................................

2-2342.5.3 Control Rod Misoperation (USAR Section 15.4.3) .....................................

2-2552.5.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Tem perature (U SA R Section 15.4.4) ...........................................................

2-2632.5.5 Chemical and Volume Control System Malfunction Resulting in aDecrease in Boron Concentration in the Reactor Coolant(U SA R Section 15.4.6) ................................................................................

2-2632.5.6 Spectrum of Rod Cluster Control Assembly Ejection Accidents (U SA R Section 15.4.8) ................................................................................

2-2722.6 INCREASE IN REACTOR COOLANT INVENTORY

.............................................

2-2852.6.1 Inadvertent Operation of the Emergency Core Cooling System DuringPower Operation (USAR Section 15.5.1) ....................................................

2-2852.6.2 Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory (USAR Chapter 15.5.2) ....................................

2-2942.7 DECREASE IN REACTOR COOLANT INVENTORY

............................................

2-3072.7.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve(U SA R Section 15.6.1) ................................................................................

2-3072.7.2 Steam Generator Tube Rupture Margin to Overfill(U SA R Section 15.6.3) ................................................................................

2-3 132.7.3 Steam Generator Tube Rupture -Input to Dose (USAR Section 15.6.3) ... 2-3342.7.4 Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary(U SA R Section 15.6.5) ................................................................................

2-355WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3viTABLE OF CONTENTS (cont.)2.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM (USAR SECTION 15.8) ....... 2-3692.8.1 Technical E valuation

...................................................................................

2-3692 .8.2 C onclusion

...................................................................................................

2-37 12 .8.3 R eferences

...................................................................................................

2-37 12.9 RA D IO LO G IC A L D O SES ..........................................................................................

2-3812.10 INSTRUMENT UNCERTAINTIES

...........................................................................

2-3812.10.1 Reactor Trip System/Engineered Safety Feature Actuation SetpointU ncertainties

................................................................................................

2-3 8 12.10.2 Initial C ondition U ncertainties

....................................................................

2-3832.11 CONTROL SYSTEMS ANALYSIS

...........................................................................

2-3842.11.1 NSSS Pressure Control Component Sizing(USAR Chapters 5.4, 7.7, & 10.4.4) ............................................................

2-3842.11.2 Operational Analysis and Margin to Trip (USAR Chapter 7.7) ..................

2-3902.12 THERMAL AND HYDRAULIC DESIGN ................................................................

2-3952 .12 .1 Introduction

.................................................................................................

2-3952.12.2 Input Parameters and Acceptance Criteria

...................................................

2-3952.12.3 Description of Analyses and Evaluations

....................................................

2-3982 .12 .4 R esu lts .........................................................................................................

2-4 0 12 .12 .5 C onclusion

...................................................................................................

2-4022 .12 .6 R eferences

...................................................................................................

2-402APPENDIX A SAFETY EVALUATION REPORT COMPLIANCE

...............................................

A-1WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3viiLIST OF TABLESTable 1.1-1Table 1.1-2Table 2.1-1Table 2.1-2Table 2.1-3Table 2.1-4Table 2.1-5Table 2.1-6Table 2.2.1-1Table 2.2.1-2Table 2.2.2-1Table 2.2.2-2Table 2.2.2-3Table 2.2.3-1Table 2.2.3-2Table 2.2.4-1Table 2.2.4-2Table 2.2.5.1-1 Table 2.2.5.1-2 Table 2.2.5.2-1 Table 2.2.5.2-2 Table 2.3.1-1Table 2.3.1-2Table 2.3.2-1Table 2.3.3-1NSSS Design Parameters for WCGS TM Program ......................................................

1-5NSSS Design Parameters for WCGS TM Program ......................................................

1-6Non-LOCA Transient Events Analyzed or Evaluated

................................................

2-19Summary of Initial Conditions and Computer Codes Used .......................................

2-20Core Kinetics Parameters and Reactivity Feedback Coefficients

...............................

2-23Summary of RTS and ESFAS Functions Actuated

.....................................................

2-24Parameters Related to OTAT and OPAT RT Setpoints

...............................................

2-27N on-LO C A Results Sum m ary ....................................................................................

2-28Time Sequence of Events -Decrease In Tfted (Manual Rod Control)

........................

2-41Decrease in Tfeed Minimum DNBR and Peak Core Average Thermal PowerR e su lts ........................................................................................................................

2 -4 1Increase in FW Flow Cases A nalyzed ........................................................................

2-49Time Sequence of Events -Increase in FW Flow (HFP, Multi-Loop, ManualR od C o ntro l) ...............................................................................................................

2-4 9HFP FWM Flow Increase Minimum DNBR and Peak Core Average ThermalP ow er R esu lts .............................................................................................................

2-4 9Excessive Load Increase Incident Summary of Results .............................................

2-58Time Sequence of Events for the Excessive Load Increase Incident

.........................

2-59Time Sequence of Events -Accidental Depressurization of the MSS at HZPC o n d itio n s ...................................................................................................................

2 -7 3Limiting Results -Accidental Depressurization of the MSS at HZP Conditions

...... 2-73Time Sequence of Events -Steam System Piping Failure at HZP Conditions

.........

2-87Limiting Results -Steam System Piping Failure at HZP Conditions

........................

2-88Time Sequence of Events -Steam System Piping Failure at HFP Conditions

........

2-107Limiting Results -Steam System Piping Failure at HFP Conditions

.......................

2-107Time Sequence of Events -Loss of External Electrical Load and/orT u rb in e T rip ..............................................................................................................

2 -118Limiting Results -Loss of External Electrical Load and/or Turbine Trip ...............

2-118Time Sequence of Events for Limiting LOAC Case ................................................

2-133Time Sequence of Events for Limiting LONE Case .................................................

2-150WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3viiiWESTINGHOUSE NON-PROPRIETARY CLASS 3 viiiLIST OF TABLES (cont.)Table 2.3.4-1Table 2.3.4-2Table 2.4.1-1Table 2.4.1-2Table 2.4.2-1Table 2.4.2-2Table 2.5.1-1Table 2.5.2-1Table 2.5.2-2Table 2.5.3-1Table 2.5.3-2Table 2.5.5-1Table 2.5.6-1Table 2.5.6-2Table 2.6.1-1Table 2.6.2-1Table 2.7.1-1Table 2.7.1-2Table 2.7.2-1Table 2.7.2-2Table 2.7.2-3Table 2.7.2-4Table 2.7.2-5Time Sequence of Events for Limiting Feed Line Break Case With OffsitePow er A vailable

........................................................................................................

2-168Time Sequence of Events for Limiting Feed Line Break Case Without OffsitePow er A vailable

........................................................................................................

2-168Time Sequence of Events -Loss of Forced Reactor Coolant Flow .........................

2-183Results -Loss of Forced Reactor Coolant Flow ......................................................

2-183Time Sequence of Events -RCP Locked Rotor/Shaft Break ...................................

2-210Limiting Results -RCP Locked Rotor/Shaft Break .................................................

2-211Time Sequence of Events -Uncontrolled RCCA Bank Withdrawal from aSubcritical C ondition

................................................................................................

2-229Time Sequence of Events -Uncontrolled RCCA Bank Withdrawal at Power .........

2-241Uncontrolled RCCA Bank Withdrawal at Power -Limiting Results .......................

2-242Non-LOCA Analysis Limits and Analysis Results for the Dropped Rod Event ....... 2-262Summary of Initial Conditions and Computer Codes Used for the DroppedR od E v en t .................................................................................................................

2 -2 62CVCS Malfunction Boron Dilution Event Results -Event Alarm to Loss ofShutdow n M argin .....................................................................................................

2-27 1Selected Input and Results of the Limiting RCCA Ejection Analyses

.....................

2-279Time Sequence of Events -RCCA Ejection

............................................................

2-280Time Sequence of Events -Inadvertent ECCS ........................................................

2-290Time Sequence of Events -CVCS Malfunction

......................................................

2-298Time Sequence of Events -Accidental Depressurization of the RCS ......................

2-310Results -Accidental Depressurization of the RCS ...................................................

2-310AFW Flows for Design Basis SGTR Analyses MDAFW Failure, All AFWPum ps O perating

......................................................................................................

2-323AFW Flows for Design Basis SGTR Analyses MDAFW Failure, TDAFWPump Stopped, MDAFW Pumps Operating

.............................................................

2-323AFW Flows for Design Basis SGTR Analyses MDAFW Failure, RupturedSG Isolated, TDAFW Pump Stopped, MDAFW Pumps Operating DuringC o o ld o w n .................................................................................................................

2 -32 3SI Flows for Design Basis SGTR Analyses

..............................................................

2-324Operator Action Times for Design Basis SGTR Margin to Overfill Analyses

.........

2-325WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3ixWESTINGHOUSE NON-PROPRIETARY CLASS 3 lxTable 2.7.2-6Table 2.7.3-1Table 2.7.3-2Table 2.7.3-3Table 2.7.3-4Table 2.7.4-1Table 2.7.4-2Table 2.7.4-3Table 2.8.1-1Table 2.8.1-2Table 2.12-1Table 2.12-2Table 2.12-3LIST OF TABLES (cont.)Sequence of Events for Limiting Margin to Overfill Analyses

................................

2-325Operator Action Times for Design Basis SGTR T/H Analyses

................................

2-342Sequence of Events for Limiting Input to Radiological Consequences A n aly ses ...................................................................................................................

2 -34 2Break Flow and Flashed Break Flow ........................................................................

2-343Intact and Ruptured SG Steam Flow to Atmosphere

................................................

2-343Subcriticality Analysis Input Parameters

..................................................................

2-362LTC A nalysis Input Param eters ................................................................................

2-362Boric Acid Solution Solubility Limit Data ...............................................................

2-363LOL ATWS Time Sequence of Events .....................................................................

2-372LONF ATWS Time Sequence of Events ..................................................................

2-372T/H Design Parameters Comparison

........................................................................

2-403Limiting Parameter Direction for DNB ....................................................................

2-405RTDP DNBR Margin Surmnary

...............................................................................

2-406WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3XWESTINGHOUSE NON-PROPRIETARY CLASS 3 xLIST OF FIGURESFigure 2.1 -1Figure 2.1-2Figure 2.1-3Figure 2.1-4Figure 2.1-5Figure 2.1-6Figure 2.2.1-1Figure 2.2.1-2Figure 2.2.1-3Figure 2.2.2-1Figure 2.2.2-2Figure 2.2.2-3Figure 2.2.2-4Figure 2.2.3-1Figure 2.2.3-2Figure 2.2.3-3Figure 2.2.3-4Figure 2.2.4-1Figure 2.2.4-2Figure 2.2.4-3Integrated D P C ........................................................................................................

2 -3 1R eactor C ore Safety L im its .....................................................................................

2-32Illustration of OTAT and OPAT Protection

.............................................................

2-33Fractional Rod Insertion versus Time from Release ...............................................

2-34Normalized RCCA Reactivity Worth versus Fractional Rod Insertion

...................

2-35Normalized RCCA Reactivity Worth versus Time from Release ............................

2-36Decrease in T'eed at Full Power -Nuclear Power and Core Heat Fluxversu s T im e .............................................................................................................

2-4 2Decrease in Tfred at Full Power -Vessel Delta-T and Core Average Moderator Tem perature versus T im e ........................................................................................

2-43Decrease in Tfd at Full Power -Pressurizer Pressure and DNBRv ersu s T im e .............................................................................................................

2 -4 4Increase in FW Flow at Full Power -Multi-Loop Manual Rod Control -Nuclear Power and Core Heat Flux Versus Time ....................................................

2-50Increase in FW Flow at Full Power -Multi-Loop Manual Rod Control -Core Average Moderator Temperature and Pressurizer Pressure Versus Time ........

2-51Increase in FW Flow at Full Power -Multi-Loop Manual Rod Control -SG Mass Inventory and Pressure Versus Time ........................................................

2-52Increase in FW Flow at Full Power -Multi-Loop Manual Rod Control -D N B R V ersus T im e .................................................................................................

2-5310% Step Increase in Heat Removal by Secondary System -MinimumReactivity

Feedback, Manual Reactor Control .......................................................

2-6010% Step Increase in Heat Removal by Secondary System -MinimumReactivity

Feedback, Automatic Reactor Control ...................................................

2-6210% Step Increase in Heat Removal by Secondary System -MaximumReactivity

Feedback, Manual Reactor Control .......................................................

2-6410% Step Increase in Heat Removal by Secondary System -MaximumReactivity

Feedback, Automatic Reactor Control ...................................................

2-66Accidental Depressurization of the MSS at HZPNuclear Power and Core Heat Flux versus Time ....................................................

2-74Accidental Depressurization of the MSS at HZPReactor Vessel Inlet Temperature and Core Average Temperature versus Time ..... 2-75Accidental Depressurization of the MSS at HZPPressurizer Pressure and Pressurizer Water Volume versus Time ...........................

2-76WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3XiLIST OF FIGURES (cont.)Figure 2.2.4-4Figure 2.2.4-5Figure 2.2.4-6Figure 2.2.4-7Figure 2.2.5.1-1 Figure 2.2.5.1-2 Figure 2.2.5.1-3 Figure 2.2.5.1-4 Figure 2.2.5.1-5.

Figure 2.2.5.1-6 Figure 2.2.5.1-7 Figure 2.2.5.1-8 Figure 2.2.5.1-9 Figure 2.2.5.1-10 Figure 2.2.5.1-11 Figure 2.2.5.1-12 Figure 2.2.5.1-13 Accidental Depressurization of the MSS at HZPCore Boron Concentration and Reactivity versus Time ..........................................

2-77Accidental Depressurization of the MSS at HZPSteam Pressure and Steam (Break) Flow versus Time ............................................

2-78Accidental Depressurization of the MSS at HZPFW Flow and SG M ass versus Tim e .......................................................................

2-79Accidental Depressurization of the MSS at HZPC ore Flow versus T im e ............................................................................................

2-80Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)

Nuclear Power and Core Heat Flux versus Time ...................................

2-89Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)

Reactor Vessel Inlet Temperature and Core Average Temperature versu s T im e .............................................................................................................

2-90Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)

Pressurizer Pressure and Pressurizer Water Volume versus Time ..........

2-91Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)

Core Boron Concentration and Reactivity versus Time .........................

2-92Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)

Steam Pressure and Steam (Break) Flow versus Time ..........................

2-93Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)

FW Flow and SG Mass versus Time ......................................................

2-94Steam System Piping Failure at HZP (1.388 ft2 Break with Offsite PowerAvailable)

Core Flow versus Tim e ..........................................................................

2-95Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)

Nuclear Power and Core Heat Flux versus Time ...................................

2-96Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)

Reactor Vessel Inlet Temperature and Core Average Temperature v ersu s T im e .............................................................................................................

2 -9 7Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)

Pressurizer Pressure and Pressurizer Water Volume versus Time ..........

2-98Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)

Core Boron Concentration and Reactivity versus Time .........................

2-99Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)

Steam Pressure and Steam (Break) Flow versus Time ........................

2-100Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)

FW Flow and SG M ass versus Time ....................................................

2-101WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xiiWESTINGHOUSE NON-PROPRIETARY CLASS 3 xiiLIST OF FIGURES (cont.)Figure 2.2.5.1-14 Figure 2.2.5.2-1 Figure 2.2.5.2-2 Figure 2.2.5.2-3 Figure 2.2.5.2-4 Figure 2.3.1 -1Figure 2.3.1-2Figure 2.3.1-3Figure 2.3.1-4Figure 2.3.1-5Figure 2.3.1-6Figure 2.3.1-7Figure 2.3.1-8Figure 2.3.1-9Figure 2.3.2-1Figure 2.3.2-2Figure 2.3.2-3Figure 2.3.2-4Figure 2.3.2-5Figure 2.3.2-6Figure 2.3.2-7Figure 2.3.2-8Figure 2.3.2-9Steam System Piping Failure at HZP (1.388 ft2 Break without Offsite PowerAvailable)

Core Flow versus Tim e ........................................................................

2-102Steam System Piping Failure at HFP (1.04 ft2 Break) Nuclear Power andC ore H eat Flux versus T im e ..................................................................................

2-108Steam System Piping Failure at HFP (1.04 ft2 Break) Pressurizer Pressure andPressurizer W ater Volum e versus Tim e ................................................................

2-109Steam System Piping Failure at HFP (1.04 ft2 Break) Reactor Vessel InletTemperature and Loop Average Temperature versus Time ...................................

2-110Steam System Piping Failure at HFP (1.04 ft2 Break) Steam Pressure andB reak Flow versus T im e ........................................................................................

2-111LOL/TT, Minimum DNBR Case Nuclear Power and SG Pressure versus Time.. 2-119LOL/TT, Minimum DNBR Case Pressurizer Pressure and Pressurizer W ater Volum e versus Tim e ....................................................................................

2-120LOL/TT, Minimum DNBR Case RCS Temperatures and DNBR versus Time .... 2-121LOL/TT, Peak MSS Pressure Case Nuclear Power and SG Pressurev ersu s T im e ...........................................................................................................

2 -12 2LOL/TT, Peak MSS Pressure Case Pressurizer Pressure and Pressurizer W ater Volum e versus Tim e ....................................................................................

2-123LOL/TT, Peak MSS Pressure Case RCS Temperatures versus Time ....................

2-124LOL/TT, Peak RCS Pressure Case Nuclear Power and SG Pressureversu s T im e ...........................................................................................................

2-12 5LOL/TT, Peak RCS Pressure Case RCS Pressures and Pressurizer WaterV olum e versus T im e ..............................................................................................

2-126LOL/TT, Peak RCS Pressure Case RCS Temperatures versus Time ....................

2-127LOAC -Nuclear Power versus Tim e ....................................................................

2-134LOAC -Core Average Heat Flux versus Time .....................................................

2-135LOAC -Reactor Coolant Loop Flow versus Time ...............................................

2-136LOAC -HL and CL Temperatures versus Time ...................................................

2-137LOAC -Actual Pressurizer Pressure versus Time ................................................

2-138LOAC -Pressurizer Water Volume versus Time ..................................................

2-139LOAC -SG Pressure versus Tim e ........................................................................

2-140LOAC -Indicated SG Level versus Tim e .............................................................

2-141LO A C -SG M ass versus Tim e .............................................................................

2-142WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xiiiWESTINGHOUSE NON-PROPRIETARY CLASS 3 XIIILIST OF FIGURES (cont.)Figure 2.3.2-10Figure 2.3.3-1Figure 2.3.3-2Figure 2.3.3-3Figure 2.3.3-4Figure 2.3.3-5Figure 2.3.3-6Figure 2.3.3-7Figure 2.3.3-8Figure 2.3.3-9Figure 2.3.3-10Figure 2.3.4-1Figure 2.3.4-2Figure 2.3.4-3Figure 2.3.4-4Figure 2.3.4-5Figure 2.3.4-6Figure 2.3.4-7Figure 2.3.4-8Figure 2.3.4-9Figure 2.3.4-10Figure 2.4.1-1Figure 2.4.1-2LOAC -Loop AFW Flow versus Time ................................................................

2-143LON F -N uclear Pow er versus Tim e ....................................................................

2-151LONF -Core Average Heat Flux versus Time ......................................................

2-152LONF -Reactor Coolant Loop Flow versus Time ...............................................

2-153LONF -HL and CL Temperatures versus Time ....................................................

2-154LONF -Actual Pressurizer Pressure versus Time ................................................

2-155LONF -Pressurizer Water Volume versus Time ...................................................

2-156LO N F -SG Pressure versus Tim e ........................................................................

2-157LON F -Indicated SG Level versus Tim e .............................................................

2-158LO N F -SG M ass versus Tim e ..............................................................................

2-159LONF -Loop AFW Flow versus Time .................................................................

2-160Feed Line Break with Offsite Power Available Nuclear Power, CoreHeat Flux and Total Core Reactivity versus Time .................................................

2-169Feed Line Break with Offsite Power Available Pressurizer Pressure andPressurizer W ater Volum e versus Tim e .................................................................

2-170Feed Line Break with Offsite Power Available Reactor Coolant Flow andFW Line Break Flow versus Tim e .........................................................................

2-171Feed Line Break with Offsite Power Available Faulted Loop and IntactLoop Reactor Coolant Temperatures versus Time ................................................

2-172Feed Line Break with Offsite Power Available SG Shell Pressureversu s T im e ...........................................................................................................

2-173Feed Line Break without Offsite Power Nuclear Power, Core Heat Fluxand Total Core Reactivity versus Tim e ..................................................................

2-174Feed Line Break without Offsite Power Pressurizer Pressure andPressurizer W ater Volume versus Tim e .................................................................

2-175Feed Line Break without Offsite Power Reactor Coolant Flow andFW Line Break Flow versus Tim e .........................................................................

2-176Feed Line Break without Offsite Power Faulted Loop and Intact LoopReactor Coolant Temperatures versus Time ..........................................................

2-177Feed Line Break without Offsite Power SG Shell Pressure versus Time ..............

2-178PLOF -Core Volumetric Flow Rate versus Time .................................................

2-184PLOF -Loop Volumetric Flow Rates versus Time ...............................................

2-185WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xivLIST OF FIGURES (cont.)Figure 2.4.1-3Figure 2.4.1-4Figure 2.4.1-5Figure 2.4.1-6Figure 2.4.1-7Figure 2.4.1-8Figure 2.4.1-9Figure 2.4.1-10Figure 2.4.1-11Figure 2.4.1-12Figure 2.4.1-13Figure 2.4.1-14Figure 2.4.1-15Figure 2.4.1-16Figure 2.4.1-17Figure 2.4.1-18Figure 2.4.1-19Figure 2.4.1-20Figure 2.4.1-21Figure 2.4.2-1Figure 2.4.2-2Figure 2.4.2-3Figure 2.4.2-4Figure 2.4.2-5Figure 2.4.2-6PLO F -N uclear Power versus Tim e .....................................................................

2-186PLOF -Pressurizer Pressure versus Tim e ............................................................

2-187PLOF -Core Average Heat Flux versus Time ......................................................

2-188PLOF -Hot Channel Heat Flux versus Time ........................................................

2-189PLOF -Minimum DNBR versus Time .................................................................

2-190CLOF -Core Volumetric Flow Rate versus Time .............................................

2-191CLOF -Loop Volumetric Flow Rates versus Time ...........................................

2-192CLOF -Nuclear Power versus Time ...............................................................

2-193CLOF -Pressurizer Pressure versus Tim e ............................................................

2-194CLOF -Core Average Heat Flux versus Time ......................................................

2-195CLOF -Hot Channel Heat Flux versus Time ..................................................

2-196CLOF -Minimum DNBR versus Time .............................................................

2-197CLOF-UF -Core Volumetric Flow Rate versus Time ..........................................

2-198CLOF-UF -Loop Volumetric Flow Rates versus Time ........................................

2-199CLOF-UF -Nuclear Power versus Time ..............................................................

2-200CLOF-UF -Pressurizer Pressure versus Time ......................................................

2-201CLOF-UF -Core Average Heat Flux versus Time ...............................................

2-202CLOF-UF -Hot Channel Heat Flux versus Time .................................................

2-203CLOF-UF -Minimum DNBR versus Time ..........................................................

2-204RCP Locked Rotor/Shaft Break Overpressurization!PCT Case -CoreVolum etric Flow Rates versus Tim e ......................................................................

2-212RCP Locked Rotor/Shaft Break Overpressurization/PCT Case -FaultedLoop Volumetric Flow Rates versus Time ............................................................

2-213RCP Locked Rotor/Shaft Break Overpressurization!PCT Case -MaximumR C S Pressure versus Tim e ....................................................................................

2-214RCP Locked Rotor/Shaft Break Overpressurization!PCT Case -NuclearPow er versus T im e ................................................................................................

2-2 15RCP Locked Rotor/Shaft Break Overpressurization!PCT Case -Core HeatF lux versus T im e ...................................................................................................

2-2 16RCP Locked Rotor/Shaft Break Overpressurization/PCT Case -PCTv ersu s T im e ...........................................................................................................

2 -2 17WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3XVLIST OF FIGURES (cont.)Figure 2.4.2-7Figure 2.4.2-8Figure 2.4.2-9Figure 2.4.2-10Figure 2.4.2-11Figure 2.4.2-12Figure 2.5.1-1Figure 2.5.1-2Figure 2.5.1-3Figure 2.5.1-4Figure 2.5.2-1Figure 2.5.2-2Figure 2.5.2-3Figure 2.5.2-4Figure 2.5.2-5Figure 2.5.2-6Figure 2.5.2-7Figure 2.5.2-8Figure 2.5.2-9RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Core Volumetric FlowR ate versus T im e ...................................................................................................

2-2 18RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Loop Volumetric FlowR ates versus T im e ..................................................................................................

2-2 19RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Pressurizer Pressure versus T im e .............................................................................................

2-220RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Nuclear Powerv ersu s T im e ...........................................................................................................

2 -2 2 1RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Core Average HeatF lux versus T im e ...................................................................................................

2-222RCP Locked Rotor/Shaft Break Rods-in-DNB Case -Hot Channel HeatF lux versus T im e ...................................................................................................

2-223Rod Withdrawal from Subcritical

-Nuclear Power Transient

..............................

2-230Rod Withdrawal from Subcritical

-Core Average Heat Flux Transient

...............

2-231Rod Withdrawal from Subcritical

-Fuel Average Temperature Transient

............

2-232Rod Withdrawal from Subcritical

-Cladding Surface Temperature Transient

..... 2-233Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -110 pcm/sec Nuclear Power and Core Heat Flux Versus Time ...............

2-243Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -110 pcm/sec Pressurizer Pressure and Water Volume Versus Time ........

2-244Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -110 pcm/sec Vessel Average Temperature and DNB R Versus Time ....... 2-245Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -1 pcm/sec Nuclear Power and Core Heat Flux Versus Time ...................

2-246Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -1 pcm/sec Pressurizer Pressure and Water Volume Versus Time ............

2-247Bank Withdrawal at Power -Minimum Reactivity Feedback 100 PercentPower -1 pcm/sec Vessel Average Temperature and DNBR Versus Time ...........

2-248Bank Withdrawal at Power -100 Percent Power Minimum DNBR VersusR eactivity Insertion R ate .......................................................................................

2-249Bank Withdrawal at Power -60 Percent Power Minimum DNBR VersusR eactivity Insertion R ate .......................................................................................

2-250Bank Withdrawal at Power -10 Percent Power Minimum DNBR VersusR eactivity Insertion R ate .......................................................................................

2-251WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xviLIST OF FIGURES (cont.)Figure 2.5.2-10Figure 2.5.2-11Figure 2.5.2-12Figure 2.5.6-1Figure 2.5.6-2Figure 2.5.6-3Figure 2.5.6-4Figure 2.6.1-1Figure 2.6.1-2Figure 2.6.1-3Figure 2.6.2-1Figure 2.6.2-2Figure 2.6.2-3Figure 2.6.2-4Figure 2.6.2-5Figure 2.6.2-6Figure 2.6.2-7Figure 2.6.2-8Figure 2.7.1-1Figure 2.7.1-2Figure 2.7.1-3Bank Withdrawal at Power -Minimum Reactivity Feedback LimitingOverpressure Case Nuclear Power and Core Heat Flux Versus Time ...................

2-252Bank Withdrawal at Power -Minimum Reactivity Feedback LimitingOverpressure Case Pressurizer Pressure and Water Volume Versus Time .............

2-253Bank Withdrawal at Power -Minimum Reactivity Feedback LimitingOverpressure Case Vessel Average Temperature and Peak RCS Pressure

............

2-254R od Ejection

-B O L/H Z P .....................................................................................

2-281R od Ejection

-B O L/H FP ......................................................................................

2-282R od Ejection

-EO L/H Z P ......................................................................................

2-283R od Ejection

-EO L/H FP ......................................................................................

2-284Inadvertent ECCS -Nuclear Power and Tavg versus Time ....................................

2-291Inadvertent ECCS -Pressurizer Pressure and Water Volume versus Time ...........

2-292Inadvertent ECCS -Total Steam Flow and Total Flow Injected to the RCSversu s T im e ...........................................................................................................

2-293CVCS Malfunction, Maximum Reactivity

Feedback, With Pressurizer SprayN uclear Power and Tavg versus Tim e .....................................................................

2-299CVCS Malfunction, Maximum Reactivity

Feedback, With Pressurizer Spray Pressurizer Pressure and Water Volume versus Time ..................................

2-300CVCS Malfunction, Maximum Reactivity

Feedback, Without Pressurizer Spray Nuclear Power and Tavg versus Tim e ...........................................................

2-301CVCS Malfunction, Maximum Reactivity

Feedback, Without Pressurizer Spray Pressurizer Pressure and Water Volume versus Time ..................................

2-302CVCS Malfunction, Minimum Reactivity

Feedback, With Pressurizer SprayN uclear Power and Tavg versus Tim e .....................................................................

2-303CVCS Malfunction, Minimum Reactivity

Feedback, With Pressurizer SprayPressurizer Pressure and Water Volume versus Time ............................................

2-304CVCS Malfunction, Minimum Reactivity

Feedback, Without Pressurizer Spray Nuclear Power and Tavg versus Time ...........................................................

2-305CVCS Malfunction, Minimum Reactivity

Feedback, Without Pressurizer Spray Pressurizer Pressure and Water Volume versus Time ..................................

2-306RCS Depressurization

-Nuclear Power versus Time ...........................................

2-311RCS Depressurization

-Pressurizer Pressure versus Time ...................................

2-311RCS Depressurization

-Indicated Loop Average Temperature versus Time ........

2-312WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xviiWESTINGHOUSE NON-PROPRIETARY CLASS 3 XVIILIST OF FIGURES (cont.)Figure 2.7.1-4Figure 2.7.2-1Figure 2.7.2-2Figure 2.7.2-3Figure 2.7.2-4Figure 2.7.2-5Figure 2.7.2-6Figure 2.7.2-7Figure 2.7.2-8Figure 2.7.3-1Figure 2.7.3-2Figure 2.7.3-3Figure 2.7.3-4Figure 2.7.3-5Figure 2.7.3-6Figure 2.7.3-7Figure 2.7.3-8Figure 2.7.3-9Figure 2.7.3-10Figure 2.7.3-11Figure 2.7.4-1Figure 2.7.4-2Figure 2.7.4-3Figure 2.7.4-4Figure 2.7.4-5RCS Depressurization

-DNBR versus Time ........................................................

2-312Pressurizer Level -M argin to Overfill Analysis

...................................................

2-326Pressurizer Pressure

-Margin to Overfill Analysis

...............................................

2-327Secondary Pressure

-Margin to Overfill Analysis

...............................................

2-328Primary to Secondary Break Flow -Margin to Overfill Analysis

........................

2-329SG Water Volumes -Margin to Overfill Analysis

................................................

2-330SG Steam Releases

-Margin to Overfill Analysis

................................................

2-331Ruptured Loop RCS Temperature

-Margin to Overfill Analysis

.........................

2-332Intact Loops RCS Temperature

-Margin to Overfill Analysis

.............................

2-333Pressurizer Level -Input to Radiological Consequences Analysis

.......................

2-344Pressurizer Pressure

-Input to Radiological Consequences Analysis

..................

2-345Secondary Pressure

-Input to Radiological Consequences Analysis

...................

2-346Primary to Secondary Break Flow -Input to Radiological Consequences A n a ly sis .................................................................................................................

2 -34 7SG Steam Releases

-Input to Radiological Consequences Analysis

...................

2-348Ruptured Loop HL and CL Temperatures

-Input to Radiological C onsequences A nalysis .........................................................................................

2-349Intact Loop HL and CL Temperatures

-Input to Radiological Consequences A n a ly sis .................................................................................................................

2 -3 5 0Break Flow Flashing Fraction

-Input to Radiological Consequences A n a ly sis .................................................................................................................

2 -3 5 1Integrated Flashed Break Flow -Input to Radiological Consequences A n a ly sis .................................................................................................................

2 -3 5 2Ruptured SG Fluid Mass -Input to Radiological Consequences Analysis

...........

2-353Ruptured SG Water Volume -Input to Radiological Consequences Analysis

...... 2-354Post-LOCA Subcriticality Boron Limit Curve ......................................................

2-364B oric A cid Solubility L im it ...................................................................................

2-365LBLOCA Boric Acid Concentration Analysis

-Vessel Boric AcidConcentration, Boil-off, and Flushing Flow versus Time .....................................

2-366SBLOCA Boric Acid Concentration Analysis

-Vessel Boric AcidConcentration, Boil-off, and Flushing Flow versus Time .....................................

2-367Core Dilution at 12 Hours for SBLOCA Pressure Hangup ...................................

2-368WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xviiiFigure 2.8.1-1Figure 2.8.1-2Figure 2.8.1-3Figure 2.8.1-4Figure 2.8.1-5Figure 2.8.1-6Figure 2.8.1-7Figure 2.8.1-8Figure 2.8.1-9Figure 2.8.1-10Figure 2.8.1-11Figure 2.8.1-12Figure 2.8.1-13Figure 2.8.1-14Figure 2.8.1-15Figure 2.8.1-16LIST OF FIGURES (cont.)Nuclear Power versus Time for LOL ATWS .........................................................

2-373Core Heat Flux versus Time for LOL ATWS ........................................................

2-373RCS Pressure versus Time for LOL ATWS ...........................................................

2-374Pressurizer Water Volume versus Time for LOL ATWS .......................................

2-374Vessel Inlet Temperature versus Time for LOL ATWS .........................................

2-375RCS Flow versus Time for LOL ATWS ................................................................

2-375SG Pressure versus Time for LOL ATWS .............................................................

2-376SG M ass versus Tim e for LOL ATW S ..................................................................

2-376Nuclear Power versus Time for LONF ATWS ......................................................

2-377Core Heat Flux versus Time for LONF ATWS .....................................................

2-377RCS Pressure versus Time for LONF ATWS ........................................................

2-378Pressurizer Water Volume versus Time for LONF ATWS ....................................

2-378Vessel Inlet Temperature versus Time for LONF ATWS ......................................

2-379RCS Flow versus Time for LONF ATWS .............................................................

2-379SG Pressure versus Time for LONF ATWS ..........................................................

2-380SG Mass versus Time for LONF ATWS ...............................................................

2-380WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xixWESTINGHOUSE NON-PROPRIETARY CLASS 3 xixLIST OF ACRONYMSACAECAFWAMSACANSANSIAOOsAORARVASME B&PVASTATWSalternating currentAtomic Energy Commission auxiliary feedwater ATWS mitigation system actuation circuitry American Nuclear SocietyAmerican National Standards Institute anticipated operational occurrences analysis of recordatmospheric relief valveAmerican Society of Mechanical Engineers Boiler & Pressure VesselAlternative Source Termanticipated transient without scramBAPCIpeffBOCBOLCCPCDSACFRCHFCLCLOFCOLRCRDMCSTCVCSDCDHRDNBDNBRDPCDTCECCSEOCEOLEOPESFASFCVFNAHFOIFONFQboric acid precipitation controleffective delayed neutron fractionbeginning of cyclebeginning of lifecentrifugal charging pumpcore design and safety analysesCode Federal Regulations critical heat fluxcold legcomplete loss of flowCore Operating Limits Reportcontrol rod drive mechanism condensate storage tankchemical and volume control systemdirect currentdecay heat removaldeparture from nucleate boilingdeparture from nucleate boiling ratioDoppler power coefficient Doppler temperature coefficient emergency core cooling systemend of cycleend of lifeemergency operating procedure engineered safety features actuation systemflow control valveradial peaking factorfraction of initialfraction of nominaltotal peaking factorWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xxLIST OF ACRONYMS (cont.)FW feedwater FWI feedwater isolation FWM feedwater malfunction GDC general design criterion HFP hot full powerHL hot legHLSO hot leg switchover HZP hot zero powerIFM intermediate flow mixing vanesITC isothermal temperature coefficienteffective multiplication factorLBLOCA large-break LOCALCO limiting condition for operation LOAC loss of non-emergency ACLOCA loss-of-coolant accidentLOL loss of loadLOL/TT loss of load/turbine tripLONF loss of normal feedwater LOOP loss of offsite powerLTC long-term coolingMDAFW motor-driven AFWMMF minimum measured flowMSIV main steam isolation valveMSS main steam systemMSSV main steam safety valveMTC moderator temperature coefficient MUR measurement uncertainty recapture NRS narrow-range spanNSSS nuclear steam supply systemOTAT overtemperature ATOPAT overpower ATpcm percent millirhoPCT peak cladding temperature PLOF partial loss of flowPORV power operated relief valveppm part per millionPSV pressurizer safety valvePWR pressurized water reactorRCCA rod cluster control assemblyRCP reactor coolant pumpRCPB reactor coolant pressure boundaryRCS reactor coolant systemWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xxiLIST OF ACRONYMS (cont.)RFA-2 Robust Fuel Assembly-2 RG Regulatory GuideRHR residual heat removalRHRS residual heat removal systemRMWS reactor makeup water systemRPS reactor protection systemRPV reactor pressure vesselRSE Reload Safety Evaluation RT reactor tripRTD resistance temperature detectors RTDP Revised Thermal Design Procedure RTP rated thermal powerRTS reactor trip systemRWAP rod withdrawal at powerRWST refueling water storage tankSAFDLs specified acceptable fuel design limitsSAL safety analysis limitSBLOCA small-break LOCASER Safety Evaluation ReportSG steam generator SGTP steam generator tube pluggingSGTR steam generator tube ruptureSI safety injection SLB steam line breakSLI steam line isolation STDP Standard Thermal Design Procedure Tavg (reactor) vessel average temperature TCD thermal conductivity degradation T/H thermal-hydraulic TDAFW turbine-driven AFWTDF thermal design flowTreed feedwater temperature TM Transition of MethodsTPI thimble plugs installed TPR thimble plugs removedTS Technical Specification TT turbine trip (Section 2.1 Tables only)UF underfrequency USAR Updated Safety Analysis ReportUSNRC United States Nuclear Regulatory Commission UV undervoltage VCT volume control tankWCAP Westinghouse Commercial Atomic Power (topical report)WCGS Wolf Creek Generating StationWCNOC Wolf Creek Nuclear Operating Corporation WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3xxiiPROFESSIONAL ENGINEERING STAMPSThis section contains the State of Kansas Professional Engineer certifications for each of the sectionspertaining to technical services scope supporting the Wolf Creek Generating Station plant design ordesign configuration.

Each Professional Engineer has designated applicable scope sections for whichthey provided Practice of Engineering oversight and for which their certification applies.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 31, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:

Section 1, Section 1 .1Certified By: Melanie Rose Fici, P.E.License Number: 21873State: KS Expiration Date: 4/30/2014 WCAP-1 7658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:

Section 2.1, Section 2.2.1, Section 2.2.2, Section 2.2.3, Section 2.3.2,Section 2.3.3, Section 2.3.4, Section 2.8, Appendix A Table A. l-1 Item No. 1 through 7, Appendix ATables A.2-1, A.3-1, and A.4-1Certified By: James A. StewartLicense Number: 21814State: KS Expiration Date: 04/30/2015 WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief theresults herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:

Section 2.2.4, Section 2.2.5, Section 2.3.1, Section 2.6.1, Section 2.6.2Certified By: William D. Higby, P.E.License Number: 22319 State: KS Expiration Date: 04/30/2014

,,,,WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:

Section 2.5.1, Section 2.5.3, Section 2.5.4, Section 2.5.5, Section 2.5.6Certified By: Chris J. McHughLicense Number: 21957State: KS Expiration Date: 04/30/15WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:

Section 2.4.1, Section 2.4.2, Section 2.5.2, Section 2.7.1Certified By: Andrew R Detar, P.E.License Number: 22109State: KS Expiration Date: April 30, 2014WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:

Section 2.7.2, Section 2.7.3, Section 2.9, Appendix A Table A. 1-1 Item No. 8Certified By: John C Reck, P.E.License Number: 21960State: KS Expiration Date: 2015-04-30 WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 31, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Section being Certified:

Section 2.7.4Certified By: David J. Fink, P.E.License Number: 18714 State: KSExpiration Date: 04-30-2014 WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Section being Certified:

Section 2.10Certified By: Ryan Paul Rossman, P.E.License Number: 18724State: KS Expiration Date: April 30, 2015WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional

Engineer, certify that to the best of myknowledge and belief the results herein do not jeopardize the protection of life, health, property, and welfare of the public.Section being Certified:

Section 2.11Certified By: Shamsul M. AbedinLicense Number: 21983State: KS Expiration Date: April 30, 2014WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3I, the undersigned, being a registered Professional

Engineer, certify that to the best of my knowledge andbelief the results herein do not jeopardize the protection of life, health, property, and welfare of thepublic.Sections being Certified:

Section 2.12, Appendix A Tables A.5-1 and A.6-1Certified By: Kevin Regis McAtee, P.E.License Number: 18582State: KS Expiration Date: 2015-04-30 WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 31-11 INTRODUCTION General OverviewThis Transition of Methods (TM) for Core Design and Safety Analyses (CDSA) licensing report isprovided by Wolf Creek Nuclear Operating Corporation (WCNOC) in support of the TM licenseamendment application for the Wolf Creek Generating Station (WCGS). WCNOC plans to transition fromits current methodology for performing core design, non-loss-of-coolant-accident (non-LOCA) andLOCA safety analyses (Post-LOCA Subcriticality and Cooling) to the Westinghouse methodologies forperforming these analyses.

Westinghouse currently holds the analysis of record (AOR) for both the small-break (SB) andlarge-break (LB) LOCA; therefore SBLOCA and LBLOCA are not included in the transition effort.For safety analyses that were reanalyzed, they were conservatively reanalyzed at the higher nominalpower level associated with a Measurement Uncertainty Recapture (MUR) power uprate. The reanalysis effort did not assume any other plant or analysis input changes that may be required to support an actualMUR power uprate. Also, the core design effort did not assume any other plant or analysis input changesthat may be required to support an actual MUR power uprate.Note that even though some analyses were performed at an uprated power (representative of an MUR),the MUR conditions (i.e., NSSS power) would be bounding for plant operation at current rated thermalpower (RTP).It is not the intent of this licensing amendment application to request approval of an MIUR power uprate.This document addresses the transition to the approved Westinghouse methodologies only.This report summarizes the analyses that were perforned to confirm that applicable acceptance criteriaare met. Sections 2.0 through 2.12 of this TM CDSA licensing report provide the results of the accidentanalyses and core design efforts.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 31-21.1 NUCLEAR STEAM SUPPLY SYSTEM PARAMETERS

1.1.1 Introduction

The nuclear steam supply system (NSSS) design parameters are the fundamental parameters used as inputin all of the NSSS accident analyses.

The current WCGS NSSS design parameters are summarized inTable 5,1- 1 of the WCGS Updated Safety Analysis Report (USAR). The NSSS design parameters providethe primary and secondary side system conditions (thermal power, temperatures, pressures, and flows)that serve as the basis for all of the NSSS analyses and evaluations.

As a result of the TM Program, theWCGS NSSS design parameters have been revised, as shown in Tables 1.1-1 and 1.1-2. Tables 1.1-1and 1.1-2 provide information for the eight cases associated with the TM Program at current power andMUR Uprate conditions, respectively.

These parameters have been incorporated, as appropriate, into theapplicable CDSA, performed in support of the TM Program.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe NSSS design parameters provide the reactor coolant system (RCS) and secondary system conditions (thermal power, temperatures, pressures, and flows) that are used as the basis for the NSSS designtransients,

systems, structures, components,
accident, and fuel analyses and evaluations.

For the TM Program at the current licensed power level, the established major input parameters andassumptions used to calculate the four cases of NSSS design parameters are summarized as follows:1. The parameters are based on Westinghouse Model F steam generators (SGs).2. The NSSS power level of 3579 MWt (3565 MWt reactor core power + 14 MWt net heat input)was assumed.3. A nominal feedwater temperature (Tfeed) range of 400.0'F to 446.0°F was selected.

4. Two design core bypass flows were used: 8.4 percent, which accounts for fuel with thimble plugsremoved (TPR) and intermediate flow mixing vanes (IFMs); and 6.4 percent, which accounts forfuel with thimble plugs installed (TPI) and IFMs.5. A thermal design flow (TDF) of 90,300 gpm/loop was assumed based on a TDF of90,324 gpm/loop rounded to the nearest hundred gpm/loop.
6. A full-power normal operating vessel average temperature (Tavg) range of 570.7°F to 588.4°F wasassumed.

This provides the basis for the WCGS to operate within this window. Any exceptions tothese values will be addressed in the affected sections.

7. Steam generator tube plugging (SGTP) levels of 0 and 10 percent were assumed.8. A maximum SG moisture carryover of 0.25 percent was utilized.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 31-3For the TM Program at the MUR Uprate power level, the established major input parameters andassumptions used to calculate the four cases of NSSS design parameters are summarized as follows:1. The parameters are based on Westinghouse Model F SGs.2. An uprated NSSS power level of 3651 MWt (3637 MWt reactor core power + 14 MWt net heatinput) was assumed for MUR Uprate conditions.

3. A nominal Tfeed range of 400.0°F to 448.6°F was selected.
4. Two design core bypass flows were used: 8.4 percent, which accounts for fuel with TPR andIFMs; and 6.4 percent, which accounts for fuel with TPI and IFMs.5. A TDF of 90,300 gpm/loop was assumed based on a TDF of 90,324 gpm/loop rounded to thenearest hundred gpm/loop.
6. A full-power normal operating Tavg range of 570.7°F to 588.4°F was assumed.

This provides thebasis for the WCGS to operate within this window. Any exceptions to these values will beaddressed in the affected sections.

7. SGTP levels of 0 and 10 percent were assumed.8. A maximum SG moisture carryover of 0.25 percent was utilized.

Acceptance CriteriaThe acceptance criteria for determining the NSSS design parameters were that the results of the accidentanalyses and evaluations continue to comply with all WCGS applicable industry and regulatory requirements, and that they provide WCGS with adequate flexibility and margin during plant operation.

1.1.3 Description of Analyses and Evaluation Table 1.1-1 provides the NSSS design parameter cases that were generated and serve as the WCGS basisfor the analyses considering the current licensed power level. These cases are as follows:* Cases 1 and 2 of Table 1.1-1 represent parameters based on a Tavg of 570.7°F.

Case 2, which isbased on an average 10 percent SGTP, yields the minimum secondary side SG pressure andtemperature.

Note that all primary side temperatures are identical for these two cases.* Cases 3 and 4 of Table 1.1-1 represent parameters based on the Tavg of 588.4°F.

Case 3, which isbased on 0 percent SGTP, yields the higher secondary side SG pressure performance conditions.

Note that all primary side temperatures are identical for these two cases. As provided viafootnote 4 of Table 1.1-1, for instances where an absolute upper limit SG outlet pressure isconservative for any analyses, these data are based on the Case 3 parameters with 0 percent SGTPand also assume a SG fouling factor of 0 hr-ft2-OF/BTU.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 31-4Table 1. 1-2 provides the NSSS design parameter cases that were generated and serve as the WCGS basisfor the analyses considering MUR Uprate conditions.

These cases are as follows:Cases 1 and 2 of Table 1.1-2 represent parameters based on a Ta,,g of 570.7°F.

Case 2, which isbased on an average 10 percent SGTP, yields the minimum secondary side SG pressure andtemperature.

Note that all primary side temperatures are identical for these two cases.Cases 3 and 4 of Table 1.1-2 represent parameters based on the Tavg of 588.4°F.

Case 3, which isbased on 0 percent SGTP, yields the highest secondary side SG pressure performance conditions.

Note that all primary side temperatures are identical for these two cases. As provided viafootnote 4 of Table 1.1-2, for instances where an absolute upper limit SG outlet pressure isconservative for any analyses, these data are based on the Case 3 parameters with 0 percent SGTPand also assume a SG fouling factor of 0 hr-ft2-°F/BTU.1.1.4 Conclusions The resulting NSSS design parameters (Tables 1.1-1 and 1. 1-2) were used by Westinghouse as the basisfor CDSA efforts.

Westinghouse performed the analyses and evaluations based on the parameter sets thatwere most limiting, so that the analyses would support operation over the entire range of conditions specified.

In cases where the analyses performed do not bound the entire range of conditions specified (such as a restricted Tavg operating range), the applicable report section identifies the range of conditions analyzed for the TM Program.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 31-5Table 1.1-1 NSSS Design Parameters for WCGS TM ProgramCurrent Power -Safety Analysis OnlyThermal Design Parameters Case I Case 2 Case 3 Case 4NSSS Power, MWt 3579 3579 3579 3579106 Btu/hr 12,212 12,212 12,212 12,212Reactor Power, MWt 3565 3565 3565 3565106 Btu/hr 12,164 12,164 12,164 12,164Thermal Design Flow, gpm/loop 90,300 90,300 90,300 90,300Reactor 106 lb/hr 138.2 138.2 134.7 134.7Reactor Coolant Pressure, psia 2250 2250 2250 2250Core Bypass, % 8.4(1,2) 8.4( 2) 8.4 1,3) 8.4(1,3Reactor Coolant Temperature, OFCore Outlet 609.8(2) 609.8(2) 626.2(3) 626.213)Vessel Outlet 604.3 604.3 621.0 621.0Core Average 575.3(2) 575.3(2) 593.2(3) 593.2(')Vessel Average 570.7 570.7 588.4 588.4Vessel/Core Inlet 537.1 537.1 555.8 555.8Steam Generator Outlet 536.8 536.8 555.5 555.5Steam Generator Steam Outlet Temperature.,

F 520.8 518.3 539.9(4) 537.5Steam Outlet Pressure, psia 818 801 962(4) 943Steam Outlet Flow, 106 lb/hr total 14.86/15.83 14.86/15.82 14.95/15.93(4) 14.94/15.91 Feed Temperature, OF 400.0/446.0 400.0/446.0 400.0/446.0 400.0/446.0 Steam Outlet Moisture,

% max. 0.25 0.25 0.25 0.25Tube Plugging Level, % 0 10 0 10Zero-Load Temperature, OF 557 557 557 557Hydraulic Design Parameters Mechanical Design Flow, gpm/loop 104,200Minimum Measured Flow, gpm total 371,000Notes:I. Core bypass flow accounts for TPR and IFMs.2. If thimble plugs are installed, the core bypass flow is 6.4%, core outlet temperature is 608.4'F, and core averagetemperature is 574.5'F.3. If thimble plugs are installed, the core bypass flow is 6.4%, core outlet temperature is 624.9'F, and core averagetemperature is 592.5'F.4. Where appropriate for NSSS analyses, a greater steam outlet pressure of 984 psia, steam outlet temperature of 542.6°Fand total steam outlet flow of 15.94 x 106 lb/hr may be assumed.

This envelops the possibility that the plant could operatewith more efficient SG performance.

WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 31-6Table 1.1-2 NSSS Design Parameters for WCGS TM ProgramMUR Uprate Power -Safety Analysis OnlyThermal Design Parameters Case 1 Case 2 Case 3 Case 4NSSS Power, MWt 3651 3651 3651 3651106 Btu/hr 12,458 12,458 12,458 12,458Reactor Power, MWt 3637 3637 3637 3637106 Btu/hr 12,410 12,410 12,410 12,410Thermal Design Flow, gpm/loop 90,300 90,300 90,300 90,300Reactor 106 lb/hr 138.3 138.3 134.9 134.9Reactor Coolant Pressure, psia 2250 2250 2250 2250Core Bypass, % 8.4(1,2) 8.4'1,2) 8.4(13) 8.4(1.3)Reactor Coolant Temperature, IFCore Outlet 610.5 2) 610.512) 626.9"'1 626.9(3)Vessel Outlet 604.9 604.9 621.6 621.6Core Average 575.4)2 575.42) 593.33) 593.3")Vessel Average 570.7 570.7 588.4 588.4Vessel/Core Inlet 536.5 536.5 555.2 555.2Steam Generator Outlet 536.2 536.2 554.9 554.9Steam Generator Steam Outlet Temperature, IF 519.7 517.2 538.914) 536.4Steam Outlet Pressure, psia 810 793 954(4) 934Steam Outlet Flow, 106 lb/hr total 15.16/16.21 15.15/16.20 15.24/16.304.)

15.23/16.29 Feed Temperature, IF 400.0/448.6 400.0/448.6 400.0/448.6 400.0/448.6 Steam Outlet Moisture,

% max. 0.25 0.25 0.25 0.25Tube Plugging Level, % 0 10 0 10Zero Load Temperature, IF 557 557 557 557Hydraulic Design Parameters Mechanical Design Flow, gpm/Ioop 104,200Minimum Measured Flow, gpm total 371,000Notes:I. Core bypass flow accounts for TPR and IFMs.2. If thimble plugs are installed, the core bypass flow is 6.4%, core outlet temperature is 609. I°F, and core averagetemperature is 574.6°F.3. If thimble plugs are installed, the core bypass flow is 6.4%, core outlet temperature is 625.6°F, and core averagetemperature is 592.6°F.4. Where appropriate for NSSS analyses, a greater steam outlet pressure of 976 psia, steam outlet temperature of 541.6°Fand total steam outlet flow of 16.32 x 106 lb/hr may be assumed.

This envelops the possibility that the plant could operatewith more efficient SG performance.

WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-I2 ACCIDENT AND TRANSIENT ANALYSIS2.1 NON-LOCA ANALYSES INTRODUCTION Chapter 15, "Accident Analysis,"

of the WCGS USAR (Reference

1) identifies the non-LOCA transient events that have been analyzed as part of the current WCGS licensing basis. In support of the TMProgram for the WCGS, most of the non-LOCA licensing basis events have been reanalyzed usingWestinghouse Electric Company safety analysis methods previously approved by the United StatesNuclear Regulatory Commission (USNRC).

The non-LOCA events summarized in this introduction section are those discussed in greater detail in Sections 2.2 through 2.7. 1, as well as the Anticipated Transient without Scram (ATWS) event discussed in Section 2.8. Other non-LOCA events, i.e., steamgenerator tube rupture (SGTR), are discussed elsewhere in this report.2.1.1 Program FeaturesKey features of the TM Program that were considered in the non-LOCA transient analyses are as follows.0 A NSSS power level of 3651 MWt, which includes all applicable uncertainties and a nominalreactor coolant pump (RCP) net heat input of 14 MWt (or 20 MWt for events where higher RCPheat is conservative)

  • Westinghouse 17x 17 Robust Fuel Assembly (RFA-2) fuel design with IFMs and thimble plugseither removed or installed (see Note below)0 A nominal, full-power Tavg window of 570.7°F to 588.4°F0 A RCS TDF of 361,200 gpm (90,300 gpm/loop),

and a minimum measured flow (MMF) of376,000 gpm (94,000 gpm/loop)

As indicated in Table 2.1-2, a bounding MMF value of 371,000 gpm (92,750 gpm/loop) wasapplied in all but one analysis where MMF is used as the RCS flow.* Westinghouse Model F SGs, with a maximum SGTP level of 10 percent0 A nominal, full-power main Treed window of 400'F to 448.6°F0 A nominal operating pressurizer pressure of 2250 psia0 A design core bypass flow of 8.4 percent and a statistical core bypass flow of 6.61 percent,conservatively corresponding to having the core TPR (see Note below)Whereas the statistical core bypass flow is used in some departure from nucleate boilingratio (DNBR) analyses, the design core bypass flow is used for all other non-LOCA analyses; see Section 2.1.5, "Initial Conditions,"

for additional details.Note: Except for the limiting DNBR analysis of the uncontrolled rod cluster control assembly (RCCA)bank withdrawal at power event, all analyses covered the bounding scenario of having the coreTPR. As a result of the exception, the plant may be required to operate with the core TPI.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-22.1.2 Non-LOCA Transient Events Considered The non-LOCA transient events considered in support of the TM Program for the WCGS are identified inthe plant condition classification discussion presented below. As noted at the beginning of Section 2. 1, thenon-LOCA events discussed in this section are a subset of the non-LOCA licensing basis events.Plant Condition Classification The American Nuclear Society (ANS) Standard ANS-51.1-1973 (ANSI-N 18.2) (Reference

2) providesclassification of plant conditions that are divided into four categories based on the anticipated frequency of occurrence and the potential radiological consequences to the public. The four categories, orconditions, are:* Condition I -Normal Operation and Operational Transients
  • Condition II -Faults of Moderate Frequency
  • Condition III -Infrequent Faults* Condition IV -Limiting FaultsThe basic principle applied in relating design requirements to each of the conditions is that the mostprobable occurrences should yield the least 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. Whereapplicable, and to the extent allowed, the reactor trip system (RTS) and/or engineered safeguards featuresare applied in fulfilling this principle.

Each condition is described in more detail as follows.Condition I -Normal Operation and Operational Transients Condition I occurrences are those that are expected frequently or regularly during power operation, refueling, maintenance, or maneuvering of the plant. Condition I occurrences are accommodated withmargin between any plant parameter and the value of the parameter that would require either automatic ormanual protective action. As Condition I events occur frequently, they must be considered from the pointof view of their effect on the consequences of fault conditions (Conditions II, III, and IV). In this regard,analysis of each fault condition described is generally based on a conservative set of initial conditions corresponding to adverse conditions that can occur during Condition I operation.

A typical list ofCondition I events is given below.Steady state and shutdown operations

-Power operation

-Startup-Hot standby-Hot shutdown-Cold shutdown-Refueling WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3Operation with permissible deviations Various deviations from nonnal operation but specifically allowed by the Technical Specifications (TS) that may occur during continued operation are considered in conjunction withother operational modes. These include:-Operation with components or systems out of service (such as an inoperable RCCA)-Leakage from fuel with limited clad defects-Excessive radioactivity in the reactor coolant" Fission products* Corrosion products" Tritium-Operation with SG leaks-Testing* Operational transients

-Plant heatup and cooldown-Step load changes (up to +/- 10 percent)-Ramp load changes (up to 5 percent per minute)-Load rejection up to and including design full-load rejection transient Condition II -Faults of Moderate Frequency Condition II faults (or events) occur with moderate frequency during the life of the plant, any one ofwhich may occur during a calendar year. These events, at worst, result in a reactor trip (RT) with the plantbeing capable of returning to operation after corrective action. A Condition II event, by itself, does notpropagate to a more serious event of the Condition III or Condition IV type without the occurrence ofother independent incidents.

In addition, Condition II events should not cause the loss of any barrier to theescape of radioactive products.

The following list identifies the Condition II non-LOCA events considered herein in support of the TM Program for the WCGS.* Feedwater (FW) system malfunctions that result in a decrease in Treed(USAR Section 15.1.1)* FW system malfunctions that result in an increase in FW flow(USAR Section 15.1.2)* Excessive increase in secondary steam flow (USAR Section 15.1.3)WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-4WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-4* Inadvertent opening of a SG atmospheric relief or safety valve(USAR Section 15.1.4)0 Loss of external electrical load (USAR Section 15.2.2)0 Turbine trip (USAR Section 15.2.3)* Inadvertent closure of main steam isolation valves (MSIVs) (USAR Section 15.2.4)* Loss of condenser vacuum and other events resulting in turbine trip (USAR Section 15.2.5)* Loss of non-emergency AC power to the station auxiliaries (USAR Section 15.2.6)* Loss of normal FW flow (USAR Section 15.2.7)* Partial loss of forced reactor coolant flow (USAR Section 15.3.1)* Uncontrolled RCCA bank withdrawal from a subcritical or low power startup condition (USAR Section 15.4. 1)* Uncontrolled RCCA bank withdrawal at power (USAR Section 15.4.2)* RCCA misoperation (dropped RCCA, dropped RCCA bank, and statically misaligned RCCA)(USAR Section 15.4.3)* Startup of an inactive RCP at an incorrect temperature (USAR Section 15.4.4)* Chemical and volume control system (CVCS) malfunction that results in a decrease in the boronconcentration in the reactor coolant (boron dilution)

(USAR Section 15.4.6)* Inadvertent operation of the emergency core cooling system (ECCS) during power operation (USAR Section 15.5.1)* CVCS malfunction that increases reactor coolant inventory (USAR Section 15.5.2)* Inadvertent opening of a pressurizer safety or relief valve (USAR Section 15.6.1)Condition III -Infrequent FaultsCondition III events occur very infrequently during the life of the plant, any one of which may occurduring the plant's lifetime.

Condition III events can be accommodated with the failure of only a smallfraction of the fuel rods, although sufficient fuel damage might occur to preclude resumption of operation for a considerable outage time. The release of radioactivity due to a Condition III event will not besufficient to interrupt or restrict public use of those areas beyond the exclusion area boundary.

ACondition III event does not, by itself, generate a Condition IV event or result in a consequential loss ofWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-5WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-5function of the RCS or containment barriers.

The following list identifies the Condition III non-LOCAevents considered herein in support of the TM Program for the WCGS.* Steam system piping failure (minor) (USAR Section 15.1.5)* Complete loss of forced reactor coolant flow (USAR Section 15.3.2)* RCCA misoperation (withdrawal of a single RCCA) (USAR Section 15.4.3)Condition IV -Limiting FaultsCondition IV events are not expected to occur, but are postulated because their consequences have thepotential for the release of significant amounts of radioactive material.

Condition IV events are the mostdrastic occurrences that must be designed

against, and represent the limiting design cases. Condition IVevents should not cause a fission product release to the environment resulting in an undue risk to publichealth and safety in excess of the guideline values in 10 CFR 100 (Code of Federal Regulations).

A singleCondition IV event shall not cause a consequential loss of required functions of the systems needed tocope with the event, including those of the ECCS and the reactor containment system. The following listidentifies the Condition IV non-LOCA events considered herein in support of the TM Program for theWCGS.* Steam system piping failure (major) (USAR Section 15.1.5)* FW system pipe break (USAR Section 15.2.8)* RCP shaft seizure (locked rotor) (USAR Section 15.3.3)* RCP shaft break (USAR Section 15.3.4)0 Spectrum of RCCA ejection accidents (USAR Section 15.4.8)Summary of Non-LOCA Events Considered Table 2.1-1 presents a list of all the non-LOCA transient events that were considered in support of the TMProgram for the WCGS to which this introductory discussion applies.

Also included in Table 2.1-1 arecross references to the applicable USAR sections, cross references to the event-specific sections withinthis report, and assertions as to which events were analyzed versus evaluated.

2.1.3 Analysis Methodology The transient-specific analysis methodologies that were applied in analyzing the non-LOCA transient events have been reviewed and approved by the USNRC via transient-specific topical reports,e.g., WCAPs, and/or through the review and approval of various licensing amendment request submittals for the WCGS or other plants. The following non-LOCA transients analyzed for the WCGS have anapproved transient-specific topical report, and each topical report is identified and discussed below.* Steam system piping failure (steam line break (SLB)) (USAR Sections 15.1.4 and 15.1.5)* Dropped RCCA/dropped RCCA bank (dropped rod) (USAR Section 15.4.3)* RCCA ejection (USAR Section 15.4.8)WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-6WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-6Steam Line Break Analysis Methodology The SLB licensing topical report, WCAP-9226-P-A Revision 1 (Reference 3), was approved by theUSNRC via a safety evaluation report (SER) from A. C. Thadani (NRC) to W. J. Johnson (Westinghouse),

dated January 31, 1989. The SLB SER identifies two conditions of acceptance, which are summarized below, along with justification for application to the WCGS.1. "Only those codes which have been accepted by the USNRC should be used for licensing application."

Justification As identified in Table 2.1-2, the computer codes used in the analysis of the SLB event areRETRAN, ANC, and VIPRE. These computer codes are discussed in Section 2.1.4, "Computer Codes Used," and it is confirmed that these codes have been accepted by the USNRC. Therefore, this condition of acceptance is satisfied for the WCGS.2. "For the pressure between 500 and 1000 psia, the 95/95 DNBR limit for the W-3 correlation is 1.45."Justification As discussed in Section 2.12, "Thermal and Hydraulic Design,"

the W-3 DNB correlation hasbeen replaced with the WLOP DNB (departure from nucleate boiling) correlation, which has adifferent 95/95 DNBR limit. Table 2.1-6 presents the DNBR safety analysis limit (SAL) appliedin the SLB analysis for which the WLOP DNB correlation was used. No further justification isrequired for the WCGS.Dropped Rod Analysis Methodology The dropped rod licensing topical report, WCAP-1 1394-P-A (Reference 4), was approved by the USNRCvia an SER from A. C. Thadani (NRC) to R. A. Newton (Westinghouse Owners Group), datedOctober 23, 1989. The dropped rod SER identifies one condition of acceptance, which is summarized below along with justification for application to the WCGS.1. "The Westinghouse

analysis, results, and comparisons are reactor and cycle specific.

No credit istaken for any direct RT due to dropped RCCA(s).

Also, the analysis assumes no automatic powerreduction features are actuated by the dropped RCCA(s).

A further review by the staff (for eachcycle) is not necessary, given the utility assertion that the analysis described by Westinghouse hasbeen performed and the required comparisons have been made with favorable results."

Justification For the reference cycle assumed in the WCGS TM Program, the methodology described inWCAP-1 1394-P-A was applied and the required comparisons have been made with acceptable results (DNBR remains greater than the limit). Future fuel cycles will be assessed as part of theReload Safety Evaluation (RSE) process described in Reference 7.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-7WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-7RCCA Ejection Analysis Methodology The RCCA ejection licensing topical report, WCAP-7588 Revision 1-A (Reference 5), was approved bythe Atomic Energy Commission (AEC) via an SER from D. B. Vassallo (AEC) to R. Salvatori (Westinghouse),

dated August 28, 1973. The RCCA ejection SER identifies two conditions of acceptance, which are summarized below, along with justification for application to the WCGS.1. "The staff position, as well as that of the reactor vendors over the last several years, has been tolimit the average fuel pellet enthalpy at the hot spot following a rod ejection accident to280 cal/gm. This was based primarily on the results of the SPERT tests, which showed that, ingeneral, fuel failure consequences for U02 have been insignificant below 300 cal/gm for bothirradiated and unirradiated fuel rods as far as rapid fragmentation and dispersal of fuel andcladding into the coolant are concerned.

In this report, Westinghouse has decreased their limitingfuel failure criterion from 280 cal/gm (somewhat less than the threshold of significant conversion of the fuel thermal energy to mechanical energy) to 225 cal/gm for unirradiated rods and200 cal/gm for irradiated rods. Since this is a conservative revision on the side of safety, the staffconcludes that it is an acceptable fuel failure criterion."

Justification The maximum fuel pellet enthalpy at the hot spot calculated for each WCGS-specific RCCAejection case was less than 200 cal/gm (see Table 2.1-6). These results satisfy thecurrently-accepted fuel failure criterion.

2. "Westinghouse proposes a clad temperature limitation of 2700'F as the temperature above whichclad embrittlement may be expected.

Although this is several hundred degrees above themaximum clad temperature limitation imposed in the AEC ECCS Interim Acceptance

Criteria, this is felt to be adequate in view of the relatively short time at temperature and the highlylocalized effect of a reactivity transient."

Justification As discussed in Westinghouse letter NS-NRC-89-3466 to the NRC (Reference 6), the 2700'Fcladding temperature limit was historically applied by Westinghouse to demonstrate that the coreremains in a coolable geometry during an RCCA ejection transient.

This limit was never used todemonstrate compliance with fuel failure limits and is no longer used to demonstrate corecoolability.

The RCCA ejection acceptance criteria applied by Westinghouse to demonstrate long-term core coolability and compliance with applicable offsite dose requirements are identified in Section 2.5.6, "Spectrum of Rod Cluster Control Assembly Ejection Accidents."

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-8WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-82.1.4 Computer Codes UsedSummnary descriptions of the principal computer codes used in the non-LOCA transient analyses areprovided below. Table 2.1-2 lists the computer codes used in each of the non-LOCA analyses.

FACTRANFACTRAN calculates the transient temperature distribution in a cross-section of a metal-clad UO2 fuelrod and the transient heat flux at the surface of the cladding.

The inputs are the nuclear power and thetime-dependent coolant parameters of pressure, flow, temperature, and density.

This code uses a fuelmodel with the following features:

  • A sufficiently large number of radial space increments to handle fast transients such as an RCCAejection accident* Material properties that are functions of temperature
  • A sophisticated fuel-to-cladding gap heat transfer calculation
  • Calculations to address post-DNB conditions (film boiling heat transfer correlations, zircaloy-water
reaction, and partial melting of the fuel)The FACTRAN licensing topical report, WCAP-7908-A (Reference 8), was approved by the USNRC viaan SER from C. E. Rossi (NRC) to E. P. Rahe (Westinghouse),

dated September 30, 1986. TheFACTRAN SER identifies seven conditions of acceptance, which are summarized in Appendix A.2,"FACTRAN for Non-LOCA Thermal Transients,"

along with justifications for application to the WCGS.RETRANRETRAN is used for studies of a pressurized water reactor (PWR) system transient response 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, RCPs, SGs (tube and shell sides), mainsteam lines, and pressurizer.

The pressurizer

heaters, spray, relief valves, and safety valves can also bemodeled.

RETRAN includes a point neutron kinetics model and reactivity effects of the moderator, fuel,boron, and control rods. The secondary side of the SG uses a detailed nodalization for the thermaltransients.

The reactor protection system (RPS) simulated in the code includes RTs on high neutron flux,high neutron flux rate, overtemperature AT (OTAT), overpower AT (OPAT), low reactor coolant flow,high pressurizer

pressure, low pressurizer
pressure, high pressurizer level, safety injection (SI) actuation, and low-low SG water level. Control systems are also simulated including rod control and pressurizer pressure control.

Parts of the SI system, including the accumulators, are also modeled.

Also, aconservative approximation of the transient DNBR, based on the core thermal limits, is calculated byRETRAN.WCAP-17658-NP August 2013Licensing Report Revision 0

WESUNGHOUSE NON-PROPRIETARY CLASS 32-9WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-9The RETRAN licensing topical report, WCAP-14882-P-A (Reference 9), was approved by the USNRCvia an SER from F. Akstulewicz (NRC) to H. Sepp (Westinghouse),

dated February 11, 1999. TheRETRAN SER identifies three conditions of acceptance, which are summarized in Appendix A.3,"RETRAN for Non-LOCA Safety Analysis,"

along with justifications for application to the WCGS.Note that the RETRAN nodalization modeling used in the WCGS analyses is consistent with theWestinghouse plant nodalization model described in WCAP-14882-P-A, except for the nodalization of theRCS hot legs (HLs). Since the approval of WCAP-14882-P-A, the HL modeling was enhanced tominimize code instabilities attributed to pressurizer insurge and outsurge.

This HL model enhancement, which has been applied in other RETRAN analyses performed by Westinghouse, consisted of dividingeach HL control volume into three equal control volumes.

Although it was needed only for the HLconnected to the pressurizer, all HLs were divided in the same manner.LOFTRANThe LOFTRAN computer code is used to study the transient response of a PWR to specified perturbations in process parameters.

This code simulates a multi-loop system by a model containing the reactor vessel,hot- and cold-leg piping, SGs (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 SG uses a homogeneous, saturated mixture for the thermal transients.

The code simulates the RPS, which includes RTs on highneutron flux, OTAT and OPAT, high and low pressurizer

pressure, low RCS flow, low-low SG waterlevel, and high pressurizer level. Control systems are also simulated, including rod control, steam dump,and pressurizer pressure control.

The SI system, including the accumulators, is also modeled.

Also, aconservative approximation of the transient DNBR, based on the core thermal limits, is calculated byLOFTRAN.The LOFTRAN licensing topical report, WCAP-7907-P-A (Reference 10), was approved by the USNRCvia an SER from C. 0. Thomas (NRC) to E. P. Rahe (Westinghouse),

dated July 29, 1983. TheLOFTRAN SER identifies one condition of acceptance, which is summarized in Appendix A.4,"LOFTRAN for Non-LOCA Safety Analysis,"

along with justification for application to the WCGS.TWINKLETWINKLE is a multi-dimensional spatial neutron kinetics code. This code uses an implicitfinite-difference method to solve the two-group transient neutron diffusion equations in one, two, andthree 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 feedbackeffects.

The code handles up to 8000 spatial points and performs steady-state initialization.

Besides basiccross-section data and thermal-hydraulic (T/H) parameters, the code accepts as input basic drivingfunctions such as inlet temperature,

pressure, flow, boron concentration, and control rod motion. The codeprovides various outputs, such as 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 amajor perturbation in the spatial neutron flux distribution.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-10WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-10The TWINKLE licensing topical report, WCAP-7979-P-A (Reference 11), was approved by the AEC viaan SER from D. B. Vassallo (AEC) to R. Salvatori (Westinghouse),

dated July 29, 1974. The TWINKLESER does not identify any conditions, restrictions, or limitations that need to be addressed for application to the WCGS.ANCANC is an advanced nodal code capable of two-dimensional (2-D) and three-dimensional (3-D)neutronics calculations.

ANC is the reference model for certain safety analysis calculations, powerdistributions, peaking factors, critical boron concentrations, control rod worths, and reactivity coefficients.

In addition, 3-D ANC validates one-dimensional (l-D) and 2-D results and providesinformation about radial (x-y) peaking factors as a function of axial position.

It can calculate discrete pinpowers from nodal information as well.The ANC licensing topical report, WCAP- 10965-P-A (Reference 12), was approved by the USNRC viaan SER from C. Berlinger (NRC) to E. P. Rahe (Westinghouse),

dated June 23, 1986. The ANC SER doesnot identify any conditions, restrictions, or limitations that need to be addressed for application to theWCGS.VIPREThe VIPRE computer program performs T/H calculations.

This code calculates coolant density, massvelocity,

enthalpy, void fractions, static pressure, and DNBR distributions along flow channels within areactor core.The VIPRE licensing topical report, WCAP-14565-P-A (Reference 13), was approved by the USNRC viaan SER from T. H. Essig (NRC) to H. Sepp (Westinghouse),

dated January 19, 1999. The VIPRE SERidentifies four conditions of acceptance, which are summarized in Appendix A.5, "VIPRE for Non-LOCAThermal/Hydraulics,"

along with justifications for application to the WCGS.2.1.5 Initial Conditions The initial conditions applied in non-LOCA transient analyses are dependent on the analysis methodology employed for each transient.

For the purpose of this discussion, the non-LOCA analyses are categorized as either DNB or non-DNB.

DNB analyses include the transient cases analyzed for DNB concerns, andnon-DNB analyses include the transient cases analyzed for concerns other than DNB, e.g., RCSoverpressure.

For most DNB analyses, the Revised Thermal Design Procedure (RTDP) methodology ofReference 14 was employed.

With this methodology, nominal values are applied as the initial RCSconditions of power (see Note below), temperature,

pressure, and flow, and the corresponding uncertainty allowances (identified later in this section) are accounted for statistically in defining the design limitDNBR. In RTDP DNB analyses, the nominal RCS flow is the MMF value and the core bypass flow is thestatistical value (see Section 2.1.1, Program Features, for the MMF and bypass flow values).Note: The reactor power applied in all analyses is consistent with the NSSS power of 3651 MWt, whichincludes a bounding uncertainty of up to 2 percent.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-11As discussed in Section 2.12, "Thermal and Hydraulic Design,"

uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, computer codes, and DNBcorrelation predictions were combined statistically to obtain the overall DNB uncertainty factor, whichwas used to define the design limit DNBR. In other words, the design limit DNBR is a DNBR value thatis greater than the WRB-2 DNB correlation limit by an amount that accounts for the RTDP uncertainties.

To provide DNBR margin to offset various penalties such as those due to rod bow and instrument bias,and to provide flexibility in design and operation of the plant, the design limit DNBR was conservatively increased to a value designated as the safety analysis limit DNBR, to which transient-specific DNBRvalues were compared.

For DNB analyses where RTDP is not employed, which are those DNB analyses that are initiated fromzero power conditions, the initial conditions were defined by applying

maximum, steady-state uncertainties to the nominal values in the most conservative direction, as appropriate; this is known asStandard Thermal Design Procedure (STDP) methodology, or non-RTDP.

In non-RTDP DNB analyses, the initial RCS flow is the TDF value and the core bypass flow is the design value (see Section 2.1.1,"Program Features,"

for the TDF and bypass flow values).

As discussed in Section 2.12, "Thermal andHydraulic Design,"

the DNBR limits for non-RTDP DNB analyses correspond to the appropriate DNBcorrelation limit increased by sufficient margin to offset any applicable DNBR penalties.

For each DNBanalysis, Table 2.1-6 identifies whether RTDP or non-RTDP (STDP) was applied, the DNB correlation, and the DNBR limit.For non-DNB analyses, the initial conditions were defined by applying

maximum, steady-state uncertainties to the nominal values in the most conservative direction.

In these analyses, the initial RCSflow is the TDF value and the core bypass flow is the design value (see Section 2.1.1, "ProgramFeatures,"

for the TDF and bypass flow values).Steady-State Initial Condition Uncertainties The following bulleted items identify the maximum steady-state initial condition uncertainties for corepower, RCS flow, Tayg. and pressurizer pressure that had to be accounted for in the non-LOCA safetyanalyses.

More limiting (bounding) uncertainties than those presented below may have been applied insome analyses.

Table 2.1-2 summarizes the initial conditions applied in each analysis.

  • 0 percent core power allowance for calorimetric measurement uncertainty
  • As indicated above, all applicable uncertainties are accounted for in the applied initial core powervalue.* +/-2.7 percent RCS flow allowance for steady-state fluctuations and measurement uncertainties
  • +6.5/-4.0°F Tavg allowance for deadband and system measurement uncertainties and bias* +50/-35 psi pressurizer pressure allowance for steady-state fluctuations and measurement uncertainties WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-12WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-12Pressurizer Level Initial Condition The nominal pressurizer water level program used at the WCGS varies linearly from 27 percent of span atthe no-load Tag of 557°F to 57 percent of span at a Tavg of 586.5°F.

For Tavg values above 586.5°F andbelow 557°F, the program level is constant at the respective levels of 57 percent of span and 27 percent ofspan. For analysis

purposes, the upper end of the pressurizer water level program was conservatively extrapolated out to 59 percent of span at the maximum full-power Tavg value of 588.4°F.

An uncertainty ofat least 5 percent of span was applied when conservative.

Steam Generator Initial Conditions The steam flow rate and steam pressure initial conditions are dependent on the initial conditions of power,Tavg, RCS flow (TDF or MMF), Tfed, SG water level, and SGTP level. The analyses considered a fullpower Tfed range of 400.0°F to 448.6°F, a constant SG water level program of 50 percent narrow rangespan (NRS), and a SGTP level range of 0 percent to 10 percent.

An uncertainty was applied to the initialSG levels when it was conservative to do so; the level uncertainties considered were + 10 percent NRSand -12 percent NRS, which correspond to initial levels of 60 percent NRS and 38 percent NRS,respectively.

Residual Decay HeatThe fission product contribution to decay heat applied in the non-LOCA analyses is consistent with theAmerican National Standards Institute (ANSI)/(ANS standard ANSI/ANS-5.1-1979 for decay heat powerin light water reactors (Reference 15), including two standard deviations of uncertainty.

2.1.6 Fuel Design Description The fuel currently in use at the WCGS and considered in the safety analyses described herein is theWestinghouse 17x 17 RFA-2 fuel design with IFMs and thimble plugs either removed or installed (seeNote below). The RFA-2 fuel rods contain enriched uranium dioxide (UO2) fuel pellets and have ZIRLOHigh Performance Fuel Cladding Material")

with an outer diameter of 0.374 inch. ZIRLO material is alsoused as the material for the mid-grids, guide thimble tubes, and instrumentation tubes. More detailedinformation on the RFA-2 fuel design is provided in Chapter 4.0 of the WCGS USAR (Reference 1). Withrespect to the non-LOCA transient

analyses, the effects of fuel design mechanical features were accounted for in fuel-related input parameters such as fuel and cladding dimensions, cladding
material, fueltemperatures, and core bypass flow.Note: Except for the limiting DNBR analysis of the uncontrolled RCCA bank withdrawal at powerevent, all analyses covered the bounding scenario of having the core TPR. As a result of theexception, the plant may be required to operate with the core TPI.( ZIRLO is a registered trademark of Westinghouse Electric Company LLC, its Affiliates and/or its Subsidiaries inthe United States of America and may be registered in other countries throughout the world. All rights reserved.

Unauthorized use is strictly prohibited.

Other names may be trademarks of their respective owners.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESUNGHOUSE NON-PROPRIETARY CLASS 32-13WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-13Regarding the issue of fuel thermal conductivity degradation (TCD) with Westinghouse codes andmethods, Westinghouse provided a discussion of the TCD impact in Reference 16 and justified thecontinued safe operation of the plants analyzed with Westinghouse codes and methods.

The Westinghouse codes and methods applied in the non-LOCA analyses discussed herein are consistent with thoseevaluated for TCD in Reference 16, and therefore the conclusions presented in Reference 16 areapplicable to the WCGS.2.1.7 Power Distribution Peaking FactorsRelative to the fuel, the power distribution is characterized by nuclear enthalpy rise hot channel factors(radial peaking factor, FNAH) of 1.59 for RTDP DNB analyses and 1.65 for non-RTDP DNB analyses, anda full-power heat flux hot channel factor (total peaking factor, FQ) of 2.50. FNAH is important for transients that are analyzed for DNB concerns.

The DNB transients as well as the DNB methodology applied(RTDP or non-RTDP) in the DNB analyses are identified in Table 2.1-6. As FNAH increases withdecreasing power level, due to rod insertion, all transients analyzed for DNB concerns are assumed tobegin with an FNAH consistent with the FNAH defined in the Core Operating Limits Report (COLR) for theassumed nominal power level. The FQ, for which the limits are specified in the COLR, is important fortransients that are analyzed for overpower

concerns, for example RCCA ejection.

2.1.8 Reactivity FeedbackThe transient response of the reactor core is dependent on reactivity feedback

effects, in particular themoderator temperature coefficient (MTC), Doppler temperature coefficient (DTC), and the Dopplerpower coefficient (DPC). Depending upon event-specific characteristics, conservatism dictates the use ofeither maximum or minimum reactivity coefficient values. Justification for the use of the reactivity coefficient values was treated on an event-specific basis. Table 2.1-3 presents the core kinetics parameters and reactivity feedback coefficients applied in the non-LOCA analyses.

The maximum and minimumintegrated DPCs applied in the safety analyses are provided in Figure 2.1-1. Note that a different DPC(not shown in Figure 2.1-1) was applied in the zero power SLB core response and zero power feedwater malfunction (FWM) analyses; this DPC is based on an RCCA being stuck out of the core.2.1.9 Pressure Relief ModelingRCS Pressure ReliefPlant components that provide RCS pressure relief in the non-LOCA analyses include the pressurizer sprays, pressurizer power-operated relief valves (PORVs),

and the pressurizer safety valves (PSVs). Themodeling of these components in each non-LOCA safety analysis is dependent on the type of transient being analyzed and the applicable analysis methodology.

Note that the sprays and PORVs are not safetygrade components, and thus were modeled only if doing so lead to more limiting

results, i.e., no creditwas taken for the operation of these components if such operation were to mitigate transient results.

Ingeneral, maximum RCS pressure relief is modeled when a minimum RCS pressure is conservative, e.g., for transients that are analyzed for DNB concerns, and minimum RCS pressure relief is modeledWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-14WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-14when a maximum RCS pressure is conservative.

Modeling details for the sprays, PORVs, and PSVs areprovided as follows, but note that more conservative modeling may have been applied in some analyses.

Sprays -The pressurizer sprays were modeled via a control valve that was set to initially open ona proportional-integral-derivative (PID) pressure signal of +25 psid from the nominal reference pressure of 2250 psia, and ramping to full-open when the PID pressure signal reaches +75 psid.This spray control logic is consistent with the as-designed logic.PORVs -Each of the two pressurizer PORVs were modeled based on a relief capacity of210,000 Ibm/hr at a pressure of 2350 psia. One PORV was modeled to actuate on an indicated pressure signal of 2350 psia and the other PORV was modeled to actuate on a PID pressure signalof 100 psid from the nominal reference pressure of 2250 psia.PSVs -Each of the three PSVs was modeled based on a relief capacity of 420,000 lbm/hr at apressure of 2575 psia. Depending on the direction of conservatism for a given analysis, thenominal opening setpoint of 2460 psig was either increased by 2.9 percent, which accounts for a+2.0 percent setpoint tolerance and a +0.9 percent set pressure shift associated with thewater-filled PSV loop seals (see WCAP-12910, Reference 17), or decreased by 2.0 percent,which accounts for a -2.0 percent setpoint tolerance.

Also, when conservative, a PSV openingdelay of 1.153 seconds was modeled to account for the purging of the water in the PSV loopseals.The pressurizer

heaters, which include proportional heaters and backup heaters, are related to the RCSpressure relief components in that they are included as part of the pressurizer pressure control system. Thepressurizer heaters were modeled as-designed if doing so causes transient results to be more limiting.

Theproportional heaters were modeled with a maximum capacity of 416 kW and the backup heaters weremodeled with a maximum capacity of 1384 kW. The heat output of the proportional heaters varies linearlyas a function of the PID pressure signal. The proportional heaters are on at 50 percent capacity when the PIDpressure signal is 0 psid, 100 percent capacity when the PID pressure signal is -15 psid, and 0 percentcapacity when the PID pressure signal is +15 psid. The backup heaters turn on at full capacity when the PIDpressure signal is -25 psid or if the pressurizer level deviates from the program level by +5 percent of span.Main Steam System (MSS) Pressure ReliefPlant components that provide MSS pressure relief in the non-LOCA analyses include the control gradeatmospheric relief valves (ARVs) and the safety grade main steam safety valves (MSSVs).

No credit istaken for the automatic actuation of the ARVs. Rather, operator action to open an ARV is credited in theanalysis described in Section 2.6.1, "Inadvertent Operation of the Emergency Core Cooling SystemDuring Power Operation."

General modeling details for the MSSVs are provided as follows, but note thatmore conservative modeling may have been applied in some analyses.

Five MSSVs per loop were modeled with opening setpoints based on nominal lift settings of1185, 1197, 1210, 1222, and 1234 psig. Each MSSV was modeled with a +3.0 percent setpointtolerance and a 5 psi ramp from closed to full-open, which accounts for accumulation.

Becausenone of the non-LOCA transients is limiting with minimum MSSV setpoints, a negative setpointtolerance was not explicitly modeled.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-15WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-152.1.10 RTS and ESFAS Functions Table 2.1-4 summarizes the RTS and engineered safety features actuation system (ESFAS) functions actuated in the non-LOCA transient analyses.

The setpoints applied in the safety analyses and theassociated time delays of each function are also presented in Table 2.1-4. Additional information relatedto the OTAT and OPAT RT setpoints is provided as follows.OTAT and OPAT Reactor Trip Setpoints Using the methodology described in WCAP-8745-P-A (Reference 18), the current OTAT and OPAT RTsetpoints were evaluated for the TM Program.

The evaluation process first involved using conservative core thermal limits, developed based on the RTDP DNB methodology (as described in Section 2.12,"Thermal and Hydraulic Design"),

to determine, under steady-state conditions, whether the OTAT andOPAT RT setpoints provide sufficient protection for the core thermal limits. Based on this initialevaluation, it was determined that one coefficient of the OTAT RT setpoint

equation, the pressure termcoefficient, had to be increased from 0.00067 1/psi to 0.00095/psi to ensure that the core thermal limits arefully protected.

The applied core thermal limits are presented in Figure 2.1-2. The OTAT and OPAT RTsetpoints are illustrated in Figure 2.1-3 and presented in Table 2.1-5.The boundaries of operation defined by the OTAT and OPAT trips are represented as "protection lines" inFigure 2.1-3. The protection lines were drawn to include all adverse instrumentation and setpoint errors sothat under nominal conditions, a trip would occur well within the area bounded by these lines. Theseprotection lines are based upon the OTAT and OPAT RT setpoints applied in the safety analyses, whichare the TS nominal values with allowances for instrumentation errors and acceptable drift betweeninstrument calibrations.

The diagram of Figure 2.1-3 is useful in the fact that the limit imposed by anygiven DNBR can be represented as a line (Tavg versus AT). The DNB lines represent the locus ofconditions for which the DNBR equals the limit value. All points below and to the left of a DNB line for agiven pressure have a DNBR greater than the SAL DNBR value. The area of permissible operation (power, temperature, and pressure) is bounded by the combination of the high neutron flux (fixedsetpoint)

RT, the high and low pressurizer pressure RTs (fixed setpoints),

the OTAT (variable setpoint) and OPAT (variable setpoint)

RTs, and the opening of the MSSVs, which limits the maximum RCSaverage temperature.

The final determination of the adequacy of the OTAT and OPAT RT setpoints is demonstrated by showingthat the design bases for DNB and fuel melting are met in the analyses of those events that credit thesefunctions for accident mitigation.

Table 2.1-4 identifies the event analyses that credit the OTAT andOPAT RT functions.

In these analyses, the dynamic compensation tenrms of the OTAT and OPAT setpointequations, which compensate for inherent instrumentation delays and piping lags between the reactor coreand the loop temperature

sensors, were modeled.

As the analysis results presented in Table 2.1-6 showthat all applicable limits are met for the analyses that credit OTAT and OPAT, the OTAT and OPAT RTsetpoints (see Table 2.1-5) are confirmed to be adequate.

Note that the OTAT penalty function that is usedto compensate for expected variations in the axial power shape, f(AI), although not explicitly credited inthe analyses, was separately confirmed to be acceptable based on the method described inWCAP-8745-P-A (Reference 18).WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-16WESTIINGHOUSE NON-PROPRIETARY CLASS 3 2-16Related to the OTAT and OPAT setpoint functions, the functional temperature ranges of the TcoId, Thor, andTavg resistance temperature detector instrumentation were reviewed to ensure that they cover the expectedtemperature ranges. It was determined that the current TCoId and Tavg ranges were acceptable, but the Thotrange required an adjustment, as indicated below.* Tcojd -510-F -630'F (same as current range)* Thot -540'F -660'F (requires a revision from the current range of 530°F-650°F)

  • Tavg -530°F -630°F (same as current range)Finally, note that a temperature difference of up to 3°F between the nominal (reference) temperature usedfor the OTAT and OPAT RT setpoints and the loop-specific, indicated, full-power Tavg values has beencovered for the analyses that rely on these RT functions for protection.

2.1.11 Reactor Trip Characteristics The negative reactivity insertion following a RT is a function of the acceleration of the RCCAs and thevariation in rod worth as a function of rod position.

With respect to the non-LOCA transient

analyses, thecritical parameter is the time from the start of RCCA insertion to when the RCCAs reach the dashpotregion, which is located at an insertion point corresponding to approximately 86 percent of the totalRCCA travel distance.

For the non-LOCA

analyses, the RCCA insertion time from filly withdrawn todashpot entry was modeled as 2.7 seconds.

The applied negative reactivity insertion following RT isbased on having the most reactive RCCA stuck in the fully withdrawn position.

Three figures relating to RCCA drop time and reactivity worth are presented in this report. The RCCAposition (fraction of full insertion) versus the time from release is presented in Figure 2.1-4. Thenormalized reactivity worth applied in the safety analyses is shown in Figure 2.1-5 as a function of rodinsertion fraction and in Figure 2.1-6 as a function of time. A total negative trip reactivity worth of4.0 percent Ak was modeled in the non-LOCA

analyses, unless noted otherwise.

In the analyses of zeropower transients that have a potential for a return-to-power (FW system malfunction, steam system pipingfailure, inadvertent opening of a SG atmospheric relief or safety valve), a minimum shutdown margin of1.3 percent Ak was conservatively modeled.2.1.12 Operator Actions CreditedTo help demonstrate compliance with applicable acceptance

criteria, operator actions were credited in theanalysis of the inadvertent operation of the ECCS during power operation event; see Section 2.6.1,"Inadvertent Operation of the Emergency Core Cooling System During Power Operation,"

for details ofthe credited operator actions.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-17In addition, there are two events that were analyzed to demonstrate that there is sufficient time available for operators to recognize the event is in progress and to take necessary actions to terminate the eventprior to reaching plant conditions that fail to comply with applicable acceptance criteria.

Althoughoperator actions are not modeled in the analyses of these events, actions by the plant operators areultimately required to ensure plant safety is maintained.

These two events are as follows.Boron dilution (Section 2.5.5, "Chemical and Volume Control System Malfunction Resulting in aDecrease in Boron Concentration in the Reactor Coolant")

CVCS malfunction that increases reactor coolant inventory (Section 2.6.2, "Chemical andVolume Control System Malfunction that Increases Reactor Coolant Inventory")

2.1.13 Results SummaryTable 2.1-6 summarizes the results obtained for each of the non-LOCA transient analyses.

The resultsdemonstrate that all applicable safety analysis acceptance criteria are satisfied for the WCGS. Althoughthe analyses and evaluations were performed with the intent to make them cycle-independent, the RSEprocess described in Reference 7 will be applied for future fuel reloads to verify that reload-related safetyanalysis inputs remain bounding.

2.1.14 References

1. "Wolf Creek Updated Safety Analysis Report,"

Revision 26, March 2013.2. ANS-51.1-1973 (ANSI-N 18.2), "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

August 1973.3. WCAP-9226-P-A, Revision 1, "Reactor Core Response to Excessive Secondary Steam Releases,"

February 1998.4. WCAP- 11394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.5. WCAP-7588, Revision I-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods,"

January 1975.6. NS-NRC-89-3466, Letter from W. J. Johnson (Westinghouse) to R. C. Jones (NRC), "Use of2700'F PCT Acceptance Limit in Non-LOCAAccidents,"

October 23, 1989.7. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

July 1985.8. WCAP-7908-A, "FACTRAN

-A FORTRAN IV Code for Thermal Transients in a UO2Fuel Rod," December 1989.9. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-18WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1810. WCAP-7907-P-A, "LOFTRAN Code Description,"

April 1984.11. WCAP-7979-P-A, "TWINKLE

-A Multi-Dimensional Neutron Kinetics Computer Code,"January 1975.12. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.13. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.14. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.15. ANSI/ANS-5.1-1979, "American National Standard for Decay Heat Power in Light WaterReactors,"

August 29, 1979.16. LTR-NRC-12-18, Letter from J. A. Gresham (Westinghouse) to USNRC Document Control Desk,"Westinghouse Response to December 16, 2011 NRC Letter Regarding Nuclear Fuel ThermalConductivity Degradation (TAC No. ME5186) (Proprietary),"

February 17, 2012.17. WCAP-12910, Revision 1-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.18. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,"

September 1986.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTfNGHOUSE NON-PROPRIETARY CLASS 32-19Table 2.1-1 Non-LOCA Transient Events Analyzed or Evaluated Report USAR Analyzed orTransient Event Section Section Evaluated?

Feedwater system malfunctions that result in a decrease in 2.2.1 15.1.1 Analyzedfeedwater temperature Feedwater system malfunctions that result in an increase in 2.2.2 15.1.2 Analyzedfeedwater flowExcessive increase in secondary steam flow 2.2.3 15.1.3 AnalyzedInadvertent opening of a steam generator atmospheric relief or 2.2.4 15.1.4 Analyzedsafety valveSteam system piping failure (SLB) at zero power 2.2.5.1 15.1.5 AnalyzedSteam system piping failure (SLB) at full power 2.2.5.2 15.1.6 AnalyzedLoss of external electrical load, turbine trip, inadvertent closure of 2.3.1 15.2.2 Analyzedmain steam isolation valves, and loss of condenser vacuum 15.2.315.2.415.2.5Loss of non-emergency AC power to the station auxiliaries 2.3.2 15.2.6 AnalyzedLoss of normal feedwater flow 2.3.3 15.2.7 AnalyzedFeedwater system pipe break 2.3.4 15.2.8 AnalyzedPartial loss of forced reactor coolant flow 2.4.1 15.3.1 AnalyzedComplete loss of forced reactor coolant flow 2.4.1 15.3.2 AnalyzedRCP shaft seizure (locked rotor) and RCP shaft break 2.4.2 15.3.3 Analyzed15.3.4Uncontrolled RCCA bank withdrawal from a subcritical or low 2.5.1 15.4.1 Analyzedpower startup condition Uncontrolled RCCA bank withdrawal at power 2.5.2 15.4.2 AnalyzedRCCA misoperation (dropped RCCA, dropped RCCA bank, 2.5.3 15.4.3 Analyzedstatically misaligned RCCA, single RCCA withdrawal)

Startup of an inactive RCP at an incorrect temperature 2.5.4 15.4.4 Evaluated CVCS malfunction that results in a decrease in the boron 2.5.5 15.4.6 Analyzedconcentration in the reactor coolant (boron dilution)

Spectrum of RCCA ejection accidents 2.5.6 15.4.8 AnalyzedInadvertent operation of the ECCS during power operation 2.6.1 15.5.1 AnalyzedCVCS malfunction that increases reactor coolant inventory 2.6.2 15.5.2 AnalyzedInadvertent opening of a pressurizer safety or relief valve 2.7.1 15.6.1 AnalyzedATWS 2.8 15.8 AnalyzedWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-20WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-20Table 2.1-2 Summary of Initial Conditions and Computer Codes UsedInitial Reactor Vessel Reactor Vessel Average RCSComputer Power Coolant t1 Coolant Temperature(2)

Pressure(

3)Event Case Distinction Code(s) Used (%) (gpm) (OF) (psia)Feedwater system malfunctions Bounding RETRAN 100 371,000 588.4 2250that result in a decrease infeedwater temperature Feedwater system malfunctions Zero power RETRAN 0 361,200 557.0 2250that result in an increase in ANCfeedwater flow VIPREFull power RETRAN 100 371,000 588.4 2250Excessive increase in secondary Bounding RETRAN 100 371,000 588.4 2250steam flowInadvertent opening of a steam Bounding RETRAN 0 361,200 557.0 2250generator atmospheric relief or ANCsafety valve VIPRESteam system piping failure Zero power RETRAN 0 361,200 557.0 2250(SLB) ANC(core response only) VIPREFull power RETRAN 100 371,000 588.4 2250ANCVIPRELoss of external electrical load, Minimum DNBR RETRAN 100 371,000 588.4 2250turbine trip, inadvertent closure of Peak RCS Pressure RETRAN 100 361,200 581.9 2215main steam isolation valves, andloss of condenser vacuum Peak MSS Pressure RETRAN 100 361,200 594.9 2200Loss of non-emergency AC Bounding RETRAN 100 361,200 564.2 2300power to the station auxiliaries Loss of normal feedwater flow Bounding RETRAN 100 361,200 564.2 2300Feedwater system pipe break Bounding RETRAN 100 361,200 594.9 2200Partial loss of forced reactor Bounding RETRAN 100 371,000 588.4 2250coolant flow VIPREComplete loss of forced reactor Bounding RETRAN 100 371,000 588.4 2250coolant flow VIPREWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-21WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-21Table 2.1-2 Summary of Initial Conditions and Computer Codes Used (cont.)Initial Reactor Vessel Reactor Vessel Average RCSComputer Power Coolant Flow"') Coolant Temperature(

2) Pressure(

3)Event Case Distinction Code(s) Used (%) (gpm) (OF) (psia)RCP shaft seizure (locked rotor) DNB RETRAN 100 371,000 588.4 2250and RCP shaft break VIPREPeak RCS pressure/

RETRAN 100 361,200 594.9 2300PCT VIPREUncontrolled RCCA bank Bounding TWINKLE 0 160,662 557.0 2200withdrawal from a subcritical or FACTRANlow power startup condition VIPREUncontrolled RCCA bank Minimum DNBR RETRAN 100 371,000 588.4 2250withdrawal at power VIPRE 60 and 575.8376,000(4) 10 560.1______

Peak RCS Pressure LOFTRAN Various'5' 361,200 Various(5) Various(5'Dropped RCCA(s) and Dropped All LOFTRAN 100 371,000 588.4 2250RCCA bank ANCVIPREStatically misaligned RCCA All ANC 100 371,000 588.4 2250VIPRESingle RCCA withdrawal Manual Rod Control ANC 100 371,000 588.4 2250VIPREStartup of an inactive RCP at an No analysis was performed; incorrect temperature see Section 2.5.4, "Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature."

CVCS malfunction that results in Mode 1 N/A 100 N/A 594.9 2250a decrease in the boron Mode 2 5 565.1concentration in the reactorcoolant (boron dilution)

Mode 3 0 350.0, 557.0Mode 4 0 200.0Mode 5 0 68.0Mode 6 No analysis was performed; a Mode 6 boron dilution is precluded by administrative controls.

Spectrum of RCCA ejection Full power TWINKLE 100 361,200 594.9 2200accidents Zero power FACTRAN 0 160,662 557.0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-22WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-22Table 2.1-2 Summary of Initial Conditions and Computer Codes Used (cont.)Initial Reactor Vessel Reactor Vessel Average RCSComputer Power Coolant Flow01 Coolant Temperature(

2) Pressuret3)Event Case Distinction Code(s) Used (%) (gpm) (OF) (psia)Inadvertent operation of the Bounding RETRAN 100 361,200 566.7 2215ECCS during power operation CVCS malfunction that increases Bounding RETRAN 100 361,200 564.2 2200reactor coolant inventory Inadvertent opening of a Bounding RETRAN 100 371,000 588.4 2250pressurizer safety or relief valveATWS Bounding LOFTRAN 100 361,200 588.4 2250Notes:1. 361,200 gpm-TDF371,000 gpm Bounding MMF376,000 gpm MMF160,662 gpm Reactor vessel flow provided by two RCPs = 0.4448xTDF.
2. 594.9°F = High nominal full power Ta,,g (588.40F) + 6.50F588.4°F = High nominal full power Ta,,g581.90F = High nominal full power Ta,.g (588.40F) -6.5°F575.8°F = 60% power Tavg (linearly interpolated between Tno-load and the high nominal full power Tag of 588.40F)570.7°F = Low nominal full power Tang566.70F = Low nominal full power Ta,,g (570.70F) -4.0°F565. IF = 5% power Ta.g (linearly interpolated between To0_1-ad and the high nominal full power Ta,,g of 588.40F) + 6.50F564_20F = Low nominal full power Tang (570.70F) -6.50F560. 1F = 10% power T.vg (linearly interpolated between Tno-load and the high nominal full power Ta,,g of 588.40F)557.00F = 0% power Tavg = Tno-load

= Mode 3 maximum Ta,,g350.0°F = Mode 3 minimum Tavg200.0°F = Mode 4 minimum Ta,,g68.0°F ý Mode 5 minimum Tavg3. 2300 psia = Nominal + 50 psi2250 psia = Nominal2215 psia = Nominal -35 psi2200 psia = Nominal -50 psi4. As indicated in Section 2.12, "Thermal and Hydraulic Design,"

for the most limiting case in the uncontrolled RCCA withdrawal at power DNBR analysis, credit was takenfor the higher MMF of 376,000 gpm to help demonstrate that the DNB design basis was met with adequate margin.5. For the uncontrolled RCCA withdrawal at power peak RCS pressure

analysis, a spectrum of initial power levels ranging from 8 to 100% was analyzed.

The corresponding initial Tavgs were based on the high nominal full power Ta,,g of 588.4°F (linear between 588.4°F at 100% power and 557°F at 0% power) plus uncertainty (6.50F). Caseswere analyzed with initial pressurizer pressures of 2200 psia (nominal minus uncertainty) and 2300 psia (nominal plus uncertainty).

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-23WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-23Table 2.1-3 Core Kinetics Parameters and Reactivity Feedback Coefficients BOC EOCParameter (Minimum Feedback)

(Maximum Feedback)

Moderator temperature coefficient, pcm/°F 6.0 (< 70% RTP)") N/A0.0 (> 70% RTP)Moderator density coefficient, Ak/(g/cc)

N/A 0.47Doppler temperature coefficient, pcm/°F -1.0 -3.5Doppler-only power coefficient, pcm/percent power -10.13 + 0.0342Q -19.33 + 0.0662Q(Q = power in %)Delayed neutron fraction 0.0075 (maximum) 0.0044 (minimum)

Doppler power defect, pcm* RCCA ejection 1007 925* RCCA withdrawal from subcritical 1007 N/ANote:1. RTP -- rated thermal powerWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32 -24WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-24Table 2.1-4 Summary of RTS and ESFAS Functions ActuatedDelayEvent Case Distinction RTS or ESFAS Signal(s)

Actuated Analysis Setpoint (seconds)

Feedwater system malfunctions Bounding OPAT RT See Table 2.1-5 7.0(1that result in a decrease infeedwater temperature Low pressurizer pressure SI with feedwater 1715.0 psia 17.0 (FWI)isolation (FWI) on SI(2)Feedwater system malfunctions Zero power Hi-hi SG water level turbine trip (TT) and FWI 100% NRS 2.5 (TT)that result in an increase in 17.0 (FWI)feedwater flow Full powerExcessive increase in secondary Bounding None N/A N/Asteam flowInadvertent opening of a steam Bounding Low pressurizer pressure SI with FWI on SI 1715.0 psia 27.0 (SI)generator atmospheric relief or 17.0 (FWI)safety valveSteam system piping failure (SLB) Zero power Low steam line pressure SI and steam line isolation 375.0 psia 27.0 (SI)(core response only) (SLI) with FWI on SI (lead/lag

= 50/5 sec) 17.0 (SLI)17.0 (FWI)Full power OPAT RT See Table 2.1-5 7.0(1)Loss of external electrical load, Minimum DNBR OTAT RT See Table 2.1-5 7.01)turbine trip, inadvertent closure ofmain steam isolation valves, and Peak RCS Pressure High pressurizer pressure RT 2425.0 psia 1.0loss of condenser vacuum Peak MSS Pressure OTAT RT See Table 2.1-5 7.0"'Loss of non-emergency AC power Bounding Low-low SG water level RT and AFW system 0% NRS 2.0 (RT)to the station auxiliaries actuation 60.0 (AFW)Loss of normal feedwater flow BoundingFeedwater system pipe break Bounding Low-low SG water level RT and AFW system 0% NRS 2.0 (RT)actuation 60.0 (AFW)Partial loss of forced reactor Bounding Low reactor coolant loop flow RT 86.3% MMF 1.0coolant flowWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRJETARY CLASS 32-25WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-25Table 2.1-4 Summary of RTS and ESFAS Functions Actuated (cont.)DelayEvent Case Distinction RTS or ESFAS Signal(s)

Actuated Analysis Setpoint (seconds)

Complete loss of forced reactor Bounding RCP undervoltage (UV) RT (3) 1.5coolant flowRCP shaft seizure (locked rotor) DNB Low reactor coolant loop flow RT 86.3% MMF 1.0and RCP shaft breakPeak RCS pressure!

peak claddingtemperature (PCT)Uncontrolled RCCA bank Bounding 35% RTP 0.5withdrawal from a subcritical or Power range neutron flux (low setting)

RTlow power startup condition Uncontrolled RCCA bank Minimum DNBR Power-range high neutron flux RT (high setting) 116.5% RTP 0.5withdrawal at power OTAT RT See Table 2.1-5 6.25(l)Peak RCS Pressure Power-range high neutron flux RT (high setting) 116.5% RTP 0.5OTAT RT See Table 2.1-5 7.0"'Power range neutron flux rate (high positive rate) 6.9% RTP with a 1.0RT 2.0-second time constantHigh pressurizer pressure RT 2425.0 2.0Dropped RCCA(s) and Dropped See Note 4RCCA bankStatically misaligned RCCA All None N/A N/ASingle RCCA withdrawal Manual Rod Control None N/A N/AStartup of an inactive RCP at an No analysis was performed; incorrect temperature see Section 2.5.4, "Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature."

CVCS malfunction that results in a Mode 1-manual OTAT RT See Table 2.1-5 7.0(1)decrease in the borontdcesintebrnMode 1 -auto None N/A N/Aconcentration in the reactorcoolant (boron dilution)

Mode 2 None N/A N/AWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-26WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-26Table 2.1-4 Summary of RTS and ESFAS Functions Actuated (cont.)DelayEvent Case Distinction RTS or ESFAS Signal(s)

Actuated Analysis Setpoint (seconds)

Mode 3 None N/A N/AMode 4 None N/A N/AMode 5 None N/A N/AMode 6 N/A N/A N/ASpectrum of RCCA ejection Full power Power range neutron flux (high setting)

RT 118% RTP 0.5accidents Zero power Power range neutron flux (low setting)

RT 35% RTP 0.5Inadvertent operation of the ECCS Bounding See Note 5 N/A N/Aduring power operation CVCS malfunction that increases Bounding None N/A N/Areactor coolant inventory Inadvertent opening of a Bounding OTAT RT See Table 2.1-5 7.0")pressurizer safety or relief valveATWS All See Note 6 N/A N/ANotes:I. The OTAT and OPAT RT response times were modeled with a time constant (first order lag) of 4.0 seconds to account for the response of the resistance temperature detectors (RTDs), the RTD bypass piping fluid transport time, and the RTD bypass piping heatup thermal lag, and a pure delay of at least 2.25 seconds to account forprotection system electronics delays, RT breaker opening, and RCCA gripper release.

A pure delay of 3.0 seconds was conservatively modeled in some analyses.

2. No SI flow was modeled because the transient is terminated by FWI before SI flow would be initiated.
3. The RCP UV RT (initiation of rod motion) was assumed to occur 1.5 seconds following the loss of bus voltage.4. Multiple cases were analyzed to cover bounding values for MTC, dropped RCCA(s) worth, and D-bank worth. The limiting cases do not result in actuation of any RTS orESFAS functions.
However, the low pressurizer pressure RT, with an analysis setpoint of 1875 psia, was actuated for some cases that are non-limiting with respect toDNBR, e.g., dropped RCCA bank cases.5. A RT is conservatively modeled coincident with event initiation; see Section 2.6.1, "Inadvertent Operation of the Emergency Core Cooling System During PowerOperation,"

for more information.

No ESFAS functions are actuated for event mitigation.

6. The ATWS mitigation system actuation circuitry (AMSAC) is credited in the ATWS analysis; see Section 2.8, "Anticipated Transients Without Scram," for moreinformation.

WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-27WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-27Table 2.1-5 Parameters Related to OTAT and OPAT RT Setpoints OTAT K, (safety analysis value) 1.205OTAT K2 0.0137/0FOTAT K3 0.00095/psi OTAT f(AI) deadband

-23% Al to +5% AlOTAT f(AI) negative gain -2.27 %/% AlOTAT f(AI) positive gain +1.84 %/% AlT' (OTAT) and T" (OPAT) Note IP' (OTAT) 2250 psiaOPAT K4 (safety analysis value) 1.169OPAT K5 -for decreasing Tasg 0.0/0F-for increasing Tavg 0.02/0FOPAT K6 -for Tavg > T" 0.00 128/0F-for Ta,g < T" 0.0/°FAllowable full-power Tavg range 570.70F to 588.40FPressurizer pressure range of applicability for OTAT and OPAT 1924.7 psia to 2459.7 psia 2)Notes:I. The analyzed initial Tasvg is used as the reference (T' and T") in the OTAT and OPAT setpoint equations.

2. Values correspond to bounding SAL for the low and high pressurizer pressure RT setpoints.

WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-28Table 2.1-6 Non-LOCA Results SummarySafety Analysis Safety AnalysisEvent Case Distinction Parameter Description Result LimitFeedwater system malfunctions that Bounding Minimum DNBR (RTDP, WRB-2 correlation) 1.77 1.521"result in a decrease in feedwater temperature Maximum core average heat flux, FOI 1.192 1.21(I)Feedwater system malfunctions that Zero power Minimum DNBR (non-RTDP, WLOP correlation)

See Note 2result in an increase in feedwater flow Maximum linear heat generation, kW/flFull power Minimum DNBR (RTDP, WRB-2 correlation) 2.04 1.52(5)Maximum core average heat flux, FOI 1.098 1.21 ")Excessive increase in secondary steam Bounding Minimum DNBR (RTDP, WRB-2 correlation) 1.97 1.52 5)flow Maximum core average heat flux, FOI 1.11 1.21 "'Inadvertent opening of a SG Bounding Minimum DNBR (non-RTDP, WLOP correlation) 5.10 1.18atmospheric relief or safety valveMaximum linear heat generation, kW/ft 6.924 22.4"1Steam system piping failure (SLB) Zero power Minimum DNBR (non-RTDP, WLOP correlation) 1.80 1.18(core response only) Maximum linear heat generation, kW/ft 15.829 22.4Full power Minimum DNBR (RTDP, WRB-2 correlation) 2.026 1.52"5)Maximum linear heat generation, kW/ft 21.8 22.4"'Loss of external electrical load, turbine Minimum DNBR Minimum DNBR (RTDP, WRB-2 correlation) 1.72 1.52(5)trip, inadvertent closure of main steam Peak RCS Pressure Maximum RCS pressure, psia 2746.8 2750.0isolation valves, and loss of condenser vacuum Peak MSS Pressure Maximum MSS pressure, psia 1297.4 1318.5Loss of non-emergency AC power to Bounding Maximum pressurizer mixture volume, ft3 1623.2 1800.0the station auxiliaries Loss of normal feedwater flow Bounding Maximum pressurizer mixture volume, ft3 1384.1 1800.0Feedwater system pipe break Bounding Minimum margin to hot leg saturation,

'F 40.5 >0.0Partial loss of forced reactor coolant Bounding Minimum DNBR (RTDP, WRB-2 correlation) 1.82 1.52...flowComplete loss of forced reactor coolant Bounding Minimum DNBR (RTDP, WRB-2 correlation) 1.69 1.52(5)flowWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-29WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-29Table 2.1-6 Non-LOCA Results Summary (cont.)Safety Analysis Safety AnalysisEvent Case Distinction Parameter Description Result LimitRCP shaft seizure (locked rotor) and DNB Maximum number of rods-in-DNB,

% 0.7 5.0RCP shaft break Peak RCS Maximum RCS pressure, psia 2675.1 2750.0pressure/PCT Maximum cladding temperature,

'F 1786.6 2700.0Maximum zirconium-water

reaction,

% of zirconium 0.29 16.0reacted by weightUncontrolled RCCA bank withdrawal Bounding Minimum DNBR below first mixing vane grid 1.83 1.13from a subcritical or low power startup (non-RTDP, ABB-NV correlation) condition Minimum DNBR above first mixing vane grid 1.66 1.17(non-RTDP, WRB-2 correlation)

Maximum fuel centerline temperature,

'F 2342 4800.014, Uncontrolled RCCA bank withdrawal Minimum DNBR Minimum DNBR (RTDP, WRB-2 correlation)

See Note 6 1.52("at power Maximum core average heat flux, fraction of 1.183 1.21...analyzed full powerPeak RCS Pressure Maximum RCS pressure, psia 2707.4 2750.0Dropped RCCA(s) and Dropped RCCA All Minimum DNBR (RTDP, WRB-2 correlation)

>1.52 1.5215,bank Maximum linear heat generation, kW/ft <22.4 22.4(3"Maximum uniform cladding strain, % <1.0 1.0Statically misaligned RCCA All Minimum DNBR (RTDP, WRB-2 correlation)

>1.52 1.52 15Single RCCA withdrawal Manual Rod Maximum number of rods-in-DNB,

% <5.0 5.0ControlStartup of an inactive RCP at an No analysis was performed; incorrect temperature see Section 2.5.4, "Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature."

WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-30WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-30Table 2.1-6 Non-LOCA Results Summary (cont.)Safety Analysis Safety AnalysisEvent Case Distinction Parameter Description Result LimitCVCS malfunction that results in a Mode 1-manual Minimum time from alarm to loss of shutdown 50.3 15decrease in the boron concentration in Mode 1-auto margin, minutes 112.9the reactor coolant (boron dilution)

Mode 2 56.0Mode 3 15.6Mode 4 15.8Mode 5 15.7Mode 6 No analysis was performed.

N/A N/ASpectrum of RCCA ejection accidents Full power Maximum fuel pellet average enthalpy, cal/gm 176.4 200.0Maximum fuel melt at the hot spot, % 4.62 10.0Zero power Maximum fuel pellet average enthalpy, cal/gm 145.2 200.0Maximum fuel melt at the hot spot, % 0.0 10.0Inadvertent operation of the ECCS Bounding Maximum pressurizer mixture volume, ft3 1786.5 1800.0during power operation CVCS malfunction that increases Bounding Minimum time from alarm to filling the pressurizer 8.5 8.0reactor coolant inventory water-solid, minutesInadvertent opening of a pressurizer Bounding Minimum DNBR (RTDP, WRB-2 correlation) 2.00 1.52 51safety or relief valveATWS Bounding Maximum RCS pressure, psia 3129.0 3215.0Notes:1. The 1.21 fraction of initial power (or analyzed full power for part-power conditions) limit was confirmed to be less than that which would correspond to melting conditions at the fuel centerline.

2. The results for this case were determined to be bounded by the results of the zero power steam system piping failure case.3. Corresponds to a conservative fuel melting temperature of 4700'F associated with a conservative EOC U02 peak bumup at the hot spot of -65,000 MWD/MTU.4. 4800'F is the fuel melting temperature corresponding to an EOC UO2 peak bumup at the hot spot of -48,276 MWD/MTU.5. This SAL DNBR is conservatively used to demonstrate that the DNB design basis is satisfied for analyses performed using RTDP methods.

Sufficient margin is maintained between the SAL DNBR and the design limit DNBR to offset the effects of rod bow, lower plenum flow anomaly, and plant instrumentation biases, as well as to provideflexibility in the design and operation of the plant. See Section 2.12, "'Thermal and Hydraulic Design,"

for additional information.

6. As discussed in Section 2.5.2, "Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power," a detailed DNBR analysis of the most limiting case wasperformed using the VIPRE computer code. This was necessary because the minimum DNBR calculated with the RETRAN computer code was less than the SAL DNBR.Per Section 2.12, "Thermal and Hydraulic Design,"

the detailed DNBR analysis confirmed that the DNB design basis is met and sufficient DNBR margin was retained toallow for flexibility in the design and operation of the plant.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-31WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-31=---Minimum Feedback

-Maximum Feedback0.000-0.002-0.004.-0.006" -0.008CD0C.)o -0.0100a.o-0.01210-.WCD -0.014-0.016-0.018-0.020020 40 60 80Power (%)100120Figure 2.1-1 Integrated DPCWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-32WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-32680660h_ 640-0L)600-5800 0.2 0.4 0.6 0.8Fraction of Rated Thermal Power1.2Figure 2.1-2 Reactor Core Safety LimitsWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-33WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-33wLý0560 580 600 620 640Tcvg (7F)660Figure 2.1-3 Illustration of OTAT and OPAT Protection WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-341.000.8000U)C0.60L-000.,21..,- 0.400:=00.200.000.00.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0Time From Release (seconds)

Figure 2.1-4 Fractional Rod Insertion versus Time from ReleaseWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-351.001rT0.80 -----------

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

--00 .6 --- ------------------------------------------------I )CDI0.40 ~ ~ --- ----------------- ---------------------------ERo InetoIFato o ulIsrinFigure~~~~

2.- omlzdRC eciiyWrt essFatoa o netoIC--76 N Auus 2013Liesn Reor Reiio WESTINGHOUSE NON-PROPRIETARY CLASS32-361.000.800S0.60UN0.40E0z0.200.000.00.5 1.0 1.5 2.0 2.5 3.0 3.5Time From Release (seconds) 4.0Figure 2.1-6 Normalized RCCA Reactivity Worth versus Time from ReleaseWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-372.2 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM2.2.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (USAR Section 15.1.1)2.2.1.1 Technical Evaluation 2.2.1.1.1 Introduction The opening of a low-pressure FW heater bypass valve, the tripping of the FW heater drain pumps, orisolating all high-pressure extraction steam will cause a reduction in Tfeed that increases the thermal loadon the primary system. For this event, there is a sudden decrease in Tfeed into the SGs.At power, the increased subcooling caused by the decreased Treed creates a greater load demand on theRCS. With the plant at no-load conditions, the addition of cold FW may cause a decrease in RCStemperature, and thus a reactivity insertion because of the negative MTC of reactivity.

However, becausethe rate of energy change decreases as the load and FW flow decrease, the no-load transient is less severethan the full-power case.Depending on the magnitude of the temperature decrease and the operation of the automatic rod controlsystem, the net effect on the RCS can be similar to the effect of increasing secondary steam flow; that is,the reactor will reach a new equilibrium condition at a power level corresponding to the new temperature difference across the secondary-side of the SG. For large Tfeed reductions, the OPAT RT function willprevent a power increase that could lead to a DNBR that is lower than the SAL value.2.2.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe decrease in Tfeed event is analyzed to confirm that the minimum DNBR and fuel centerline temperature design bases are met. Therefore, the analysis uses the following key modeling inputs andassumptions:

The RTDP (Reference

1) was used. The initial RCS pressure and RCS temperature were assumedto be at the nominal values consistent with steady-state full-power operation.

The reactor coolantMMF was also modeled.

Uncertainties for these initial conditions were accounted for in theDNBR SAL as described in Reference 1.The analyses were performed at an initial NSSS power of 3651 MWt, which includes a nominalreactor coolant pump (RCP) net heat input of 14 MWt and all applicable uncertainties.

The analyses model the WCGS SGs (Westinghouse Model F). An initial water levelcorresponding to the nominal level minus uncertainties was modeled in all four SGs.All Tfeed reduction cases modeled a symmetric decrease in Treed to all four SGs. At the start of thetransient, the FW enthalpy was reduced to bound a temperature reduction of 200'F (step change)and the FW mass flow remained constant throughout the event. The 200'F temperature reduction WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-38conservatively bounds the spurious opening of the low pressure FW heater bypass valve and theresulting bypass of all flow through the low pressure FW heaters.Pressurizer sprays and PORVs were modeled to reduce RCS pressure, resulting in a conservative evaluation of the margin to the DNBR SAL.All Tteed reduction cases were initiated from hot full-power.

Cases modeling manual andautomatic rod control were analyzed.

In addition, sensitivities were analyzed to ensure thatconservative vessel mixing characteristics were used.* The OPAT RT function was credited for this event.Based on its frequency of occurrence, the decrease in Treed event is considered to be a Condition II eventas defined by the ANS document "Nuclear Safety Criteria for the Design of Stationary Pressurized WaterReactor Plants,"

ANSI N 18.2-1973.

As such, the applicable acceptance criteria for this incident are:Pressures in the RCS and the MSS should be maintained below 110 percent of the respective design pressures.

Fuel cladding integrity is maintained by ensuring that the minimum DNBR remains greater thanthe 95/95 DNBR SAL of 1.52. Additionally, fuel melting is precluded by ensuring that themaximum transient core average thermal power does not exceed a value that would result inexceeding the kW/ft value corresponding to fuel centerline melting at the core hot spot. For theWCGS, it has been confirmed that power levels up to 121 percent of the analyzed power levelmeet this criterion.

An incident of moderate frequency should not generate a more serious plant condition withoutother faults occurring independently.

Demonstrating that the pressurizer does not becomewater-solid ensures a more serious plant condition is not generated.

Because this event results in acooldown of the RCS, the reactor coolant experiences a reduction in volume, and therefore pressurizer filling is not a concern.The primary acceptance criteria used in this analysis is that the minimum DNBR remains greater than theSAL and that the maximum transient core average thermal power does not exceed the value that couldpotentially result in fuel melt at the core hot spot. The event does not challenge the primary-orsecondary-side pressure limits because the increased heat removal results in an RCS cooldown.

The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC). Brief discussions ofthe specific GDCs that are related to the FWM event acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the FWM event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-39GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the reactor coolant pressureboundary (RCPB) are not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the FWM event, this is shown to be met bydemonstrating that the peak RCS pressure is less than 110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires that one of thereactivity control systems consists of control rods capable of reliably controlling reactivity changes with appropriate margin for malfunctions like stuck rods so that specified acceptable fueldesign limits are not exceeded under conditions of normal operation, including anticipated operational occurrences.

For the FWM event, this is shown to be met by demonstrating that thefuel cladding integrity is maintained.

2.2.1.1.3 Description of Analyses and Evaluations The excessive heat removal due to a Tfeed reduction transient was analyzed with the RETRAN computercode (Reference 2). This code simulates a multi-loop RCS, core neutron kinetics, the pressurizer, pressurizer relief and safety valves, pressurizer spray and heaters, SGs, and MSSVs. The code computespertinent plant variables including temperatures, pressures, and power level. RETRAN (Reference

2) isused to conservatively predict DNBR.The Tfeed reduction analysis accounts for the spurious opening of the low pressure FW heater bypass valvewhich results in a maximum Tfeed reduction of 200'F to all SGs. Cases modeled both automatic andmanual rod control.

All cases assume a conservatively large moderator density coefficient characteristic of end-of-life (EOL) conditions.

2.2.1.1.4 ResultsComparison of results for both analyzed cases confirms that the Tfeed reduction case modeling manual rodcontrol and design vessel mixing coefficients is the most limiting case. This case produces the largestreactivity

feedback, and therefore results in the greatest power increase.

For both analyzed Tfed reduction transient cases, the reactor trips on the OPAT function which thencauses a consequential turbine trip. Minimum DNBR and peak core average thermal power are reachedshortly after the RT. Following RT, the event is terminated as a consequence of a SI trip due to lowpressurizer pressure.

This SI trip causes automatic FW isolation ending the event.Table 2.2.1 -1 shows the time sequence of events for the limiting Tfed decrease transient.

Table 2.2.1-2provides minimum DNBR and peak core average thermal power results of both analyzed cases.Figures 2.2.1 -1 through 2.2.1-3 show the transient responses of various system parameters for the limitingTfeed decrease transient.

2.2.1.2 Conclusions For the excessive Tfted decrease event, the results show that the minimum DNBR remains above theapplicable SAL and that the core average thermal power does not exceed a value that results in exceeding WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-40the kW/ft limit corresponding to fuel centerline melting at the core hot spot. Therefore, no fuel damage ispredicted and all applicable acceptance criteria are satisfied for the WCGS. Based on this, it is concluded that the plant will continue to meet the requirements of GDCs 10, 15., and 26.2.2.1.3 References

1. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

May 1999.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-41Table 2.2.1-1 Time Sequence of Events -Decrease In Treed (Manual Rod Control)Event Time (seconds)

Low Pressure FW Heater Bypass Valves Open 00.01OPAT Setpoint Reached in Two Loops 35.2RT (Rod Motion Starts) on OPAT 38.2Minimum DNBR Reached 38.5Low Pressurizer Pressure SI Setpoint Reached 83.1FW Isolation Initiated 100.0Table 2.2.1-2 Decrease in Tfeed Minimum DNBR and Peak Core Average Thermal Power ResultsTime ofMinimum Peak Core Average Time of Peak CoreMinimum DNBR Thermal Power(2) Average Thermal PowerTfeed Decrease Case DNBR(1 (seconds)

(FOI) (seconds)

Automatic Rod Control 1.86 47.5 1.173 47.5Manual Rod Control 1.77 38.5 1.192 38.5Notes:1. The SAL for DNBR is 1.52.2. The SAL for peak core average thermal power (fraction of initial) is 1.21 of the analyzed power level.WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-42WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-4200U0C_.)0 20 40 60 90 100 120 140 160Time (seconds) 0 20 40 60 80 100 120 140 160Time (seconds)

Figure 2.2.1-1 Decrease in Tfeed at Full Power -Nuclear Power and Core Heat Flux versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-43WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-43LaI-V4)0C,)C,)4)U-ci~0ci)0~EI-I-00ci.)-v0ci)1.~0C-)0 20 40 60 80 100 120 140 160Time (seconds)

Knan.1J7V580-570-560-550-540-530-.... ... .... .. ....... ... .. ... .... ...JIM0 20 40 60 80 100Time (seconds) 120 140 160Figure 2.2.1-2Decrease in Treed at Full Power -Vessel Delta-T and Core Average Moderator Temperature versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-44CLa,0,0 20 40 60 80 100Time (seconds12 14 100 20 40 60 8O 100 120 140 160rime (seconds)

Figure 2.2.1-3 Decrease in Trd at Full Power -Pressurizer Pressure and DNBR versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS32-452.2.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow(USAR Section 15.1.2)2.2.2.1 Technical Evaluation 2.2.2.1.1 Introduction The addition of excessive FW will cause an increase in heat removal from the RCS. An example ofexcessive FW flow would be a full opening of a main FW flow control valve (FCV) due to a FW controlsystem malfunction or an operator error. At power, this excess flow causes a greater load demand on theRCS due to increased subcooling in the SG. With the plant at no-load conditions, the addition of excessFW may cause a decrease in RCS temperature, and thus a reactivity insertion due to the effects of thenegative MTC of reactivity.

2.2.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe FW flow increase event is analyzed to confirm that the minimum DNBR and fuel centerline temperature design bases are met. Therefore, the analysis uses the following key modeling inputs andassumptions:

The RTDP (Reference

1) was used for the cases initiated from full power. The initial RCSpressure and RCS temperature were assumed to be at the nominal values consistent withsteady-state full-power operation.

The reactor coolant MMF was also modeled.

Uncertainties forthese initial conditions were accounted for in the DNBR SAL as described in Reference 1.The analyses were performed at an initial NSSS power of 3651 MWt, which includes a nominalRCP net heat input of 14 MWt and all applicable uncertainties.

  • The analyses model the WCGS SGs (Westinghouse Model F).For the single-loop FW flow increase event at full-power, one FW control valve was assumed tomalfunction, resulting in a step increase to 200 percent of the nominal full-power FW flow toone SG.For the multiple-loop FW flow increase event at full-power, two FW control valves were assumedto malfunction, resulting in a step increase to 200 percent of the nominal full-power FW flow totwo SGs.The increase in FW flow rate results in a decrease in the Tfeed (enthalpy) due to the reducedefficiency of the FW heaters.

For full power, a 25 Btu/lbm decrease in the FW enthalpy wasconservatively assumed to occur coincident with the FW flow increase.

For the single-loop FW flow increase event at no-load conditions, one FW control valve wasassumed to malfunction, resulting in a step increase to 250 percent of the full-power nominal flowto one SG.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-46For the multiple-loop FW flow increase event at no-load conditions, two FW control valves wereassumed to malfunction, resulting in a step increase to 250 percent of the full-power nominal flowto two SGs.For the cases initiated at zero power, initial reactor power, RCS pressure, and RCS temperature were assumed to be at levels corresponding to no-load conditions.

TDF was also modeled.

Inaddition, the reactor was assumed to be at the minimum shutdown margin condition of-0.013 Ak/k.For the full-power cases, an initial water level corresponding to the nominal level minusuncertainties was modeled in all four SGs, whereas an initial water level corresponding to thenominal level was modeled for the zero-power cases.Pressurizer sprays and PORVs were modeled to reduce RCS pressure, resulting in a conservative evaluation of the margin to the DNBR SAL.The full-power cases were analyzed with manual and automatic rod control.For cases at zero-power conditions, the initial Tfeed was assumed to be 35'F.The heat capacities of the RCS and SG thick metal were not considered, thereby maximizing thepotential temperature reduction of the reactor coolant.Based on its frequency of occurrence, the increase in FW flow event is considered to be a Condition IIevent as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized WaterReactor Plants,"

ANSI N18.2-1973.

As such, the applicable acceptance criteria for this incident are:Pressure in the RCS and MSS should be maintained below 110 percent of the design pressures.

Fuel cladding integrity is maintained by ensuring that the minimum DNBR remains greater thanthe 95/95 DNBR SAL of 1.52 in the limiting fuel rods and that the centerline temperature of thefuel rods with the peak linear heat rate (kW/ft) does not exceed the UO2 melting temperature.

Fuel melting is precluded by ensuring that the maximum transient core average thermal powerdoes not exceed a value that would result in exceeding the kW/ft value corresponding to fuelcenterline melting at the core hot spot. For the WCGS, it has been confirmed that power levels upto 121 percent of the initial value meet this criterion.

An incident of moderate frequency should not generate a more serious plant condition without otherfaults occurring independently.

Demonstrating that the pressurizer does not become water-solid ensures a more serious plant condition is not generated.

Because this event results in a cooldown ofthe RCS, the reactor coolant volume decreases, and therefore pressurizer filling is not a concern.The primary acceptance criterion used in this analysis is that the minimum DNBR remains greater thanthe SAL, thus ensuring fuel cladding integrity is maintained.

The fuel cladding integrity is also assured byensuring that the maximum transient core average thermal power does not exceed the value that wouldresult in exceeding the kW/ft value corresponding to fuel centerline melting at the core hot spot. TheWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32 -47event does not challenge the primary-or secondary-side pressure limits because the increased heatremoval results in an RCS cooldown.

The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the FWM event acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the FWM event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.

GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.

For theFWM event, this is shown to be met by demonstrating that the peak RCS pressure is less than110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires that one of thereactivity control systems consist of control rods capable of reliably controlling reactivity changeswith appropriate margin for malfunctions like stuck rods so that specified acceptable fuel designlimits are not exceeded under conditions of normal operation, including anticipated operational occurrences.

For the FWM event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.

Note that a RT is not modeled in the FWM analysis performed for theWCGS.2.2.2.1.3 Description of Analyses and Evaluations The excessive heat removal due to a FW flow increase transient was analyzed with the RETRANcomputer code (Reference 2). This code simulates a multi-loop RCS, core neutron kinetics, thepressurizer, pressurizer relief and safety valves, pressurizer spray and heaters, SGs, and MSSVs. The codecomputes pertinent plant variables including temperatures, pressures, and power level.The VIPRE computer code (Reference

3) is used to verify that the DNBR remains above the DNBR SALfor hot zero power (HZP) cases. For hot full power (HFP) cases, RETRAN (Reference
2) is used toconservatively predict DNBR.The excessive FW flow event assumes an accidental opening of one or more FW control valves with thereactor at full- and zero-power conditions, and with automatic and manual rod control, where applicable.

Both the automatic and manual rod control cases assume a conservatively large moderator densitycoefficient characteristic of EOL conditions.

Table 2.2.2-1 summarizes the analyzed cases.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-482.2.2.1.4 ResultsFor the cases initiated at HFP conditions, a comparison of the multiple-loop (failure of two FW controlvalves) and single-loop (failure of one FW control valve) cases demonstrates that the two-loop failurecase with manual rod control is more limiting.

The two-loop FW flow increase case with manual rodcontrol produces the largest reactivity

feedback, and therefore results in the greatest power increase.

The cases initiated at HZP conditions are less limiting than the HZP SLB analysis described inSection 2.2.5. Therefore, the results of this case are not presented.

Continuous addition of excessive FW is prevented by the SG high-high level trip, which initiates FWI andtrips the turbine and main FW pumps. Subsequent to FWI initiated by a SG high-high level trip, thereactor continues to operate until the low-low SG level setpoint is reached.

However, the RT on low-lowSG level is not modeled in the analysis because it occurs after the time of interest for the event.Table 2.2.2-2 shows the time sequence of events for the limiting multi-loop, full-power FW flow increasetransient with manual rod control; Table 2.2.2-3 provides minimum DNBR and peak core average thermalpower results of all cases. Figures 2.2.2-1 through 2.2.2-4 show the transient responses of various systemparameters for the limiting multi-loop FW flow increase initiated from full-power conditions with manualrod control.2.2.2.2 Conclusion For the excessive increase in FW flow event, the results show that the DNBRs encountered are above theapplicable SAL value and that the core average thermal power does not exceed a value that results inexceeding the kW/ft limit corresponding to fuel centerline melting at the core hot spot. Therefore, no fueldamage is predicted and all applicable acceptance criteria are satisfied for the WCGS. Based on this, it isconcluded that the plant will continue to meet the requirements of GDCs 10, 15, and 26.2.2.2.3 References
1. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

May 1999.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-49Table 2.2.2-1 Increase in FW Flow Cases AnalyzedCase Power Level Failure Affected Loop(s) Rod Control1 HFP 1 FCV Loop 1 Manual2 HFP 1 FCV Loop 1 Automatic 3 HFP 2 FCVs Loops 1 and 2 Manual4 H-FP 2 FCVs Loops 1 and 2 Automatic 5 HZP 1 FCV Loop 1 Manual6 HZP 2 FCVs Loops 1 and 2 ManualTable 2.2.2-2 Time Sequence of Events -Increase in FW Flow (HFP, Multi-Loop, ManualRod Control)Event Time (seconds)

Two FW Control Valves Fail Full-Open (Event Initiation) 0.01SG Level Reaches High-High Setpoint of 100% NRS 36.9Turbine Trip Initiated (from High-High SG Level Trip) 39.3Minimum DNBR Occurs 41.5FWI Initiated (from High-High SG Level Trip) 53.8Table 2.2.2-3 HFP FWM Flow Increase Minimum DNBR and Peak Core Average Thermal PowerResultsTime of Peak CoreTime of Peak Core Average Average ThermalHFP FW Flow Increase Minimum Minimum DNBR Thermal Power(2) PowerCase DNBR(1) (seconds)

(FOI) (seconds)

Single Loop Auto Control 2.15 26.0 1.046 39.5Single Loop Manual Control 2.10 28.0 1.073 45.5Multiple Loop Auto Control 2.11 39.5 1.065 44.5Multiple Loop Manual 2.04 41.5 1.098 45.0ControlNotes:1. The SAL for DNBR is 1.52.2. The SAL for peak core average thermal power (FOI) is 1.21.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-50WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-504 ,~,.i.'016.110A,...... ......................................

..............0A"U.S020 40 60 80 100Time (seconds) 120 140160OSi.I.40C4-0C0C-)04-'C1=04)4)0C-)1-...........

.........................

n-j.%f.&0 20 40 60 80 100Time (seconds) 120 140 160Figure 2.2.2-1Increase in FW Flow at Full Power -Multi-Loop Manual Rod ControlNuclear Power and Core Heat Flux Versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-51WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-512600'E594-590............................................... ...................................................................................................................................................... ....... .................imOo 20 40 6O 80 100Time (seconds) 120 140160af)cocaco60 80 100Time (seconds)

Figure 2.2.2-2Increase in FW Flow at Full Power -Multi-Loop Manual Rod ControlCore Average Moderator Temperature and Pressurizer Pressure Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-52WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-52L oo ps 1 Loops 3-4S140DOO......

t12DDDD041000D-01002o 1 0 20.. .. ...100-~1000.. ......9150-60 80 00Time (seconds)

Figure 2.2.2-3Increase in FW Flow at Full Power -Multi-Loop Manual Rod ControlSG Mass Inventory and Pressure Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-53WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-530 20 40 60 80 100 120 140 160Time (seconds)

Figure 2.2.2-4Increase in FW Flow at Full Power -Multi-Loop Manual Rod ControlDNBR Versus TimeWCAP- 17658-NP August 2013WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-542.2.3 Excessive Increase in Secondary Steam Flow (USAR Section 15.1.3)2.2.3.1 Technical Evaluation 2.2.3.1.1 Introduction An excessive load increase incident is defined as a rapid increase in steam flow that causes a powermismatch between the reactor core power and the SG load demand. The RCS is designed to accommodate a 10 percent step-load increase or a 5percent-per-minute ramp-load increase in the range of 15 to100 percent of full power. Any loading rate in excess of these values may cause a RT actuated by the RTsystem. If the load increase exceeds the capability of the RCS, the transient would be terminated insufficient time to prevent the DNB design basis from being violated.

This incident could result from eitheran administrative violation such as excessive loading by the operator or an equipment malfunction in thesteam bypass control system, or turbine speed control.During power operation, steam dump to the condenser is controlled by reactor coolant condition signals,such as a high reactor coolant temperature, which indicates a need for steam dump. A single controller malfunction will not cause steam dump valves to open; an interlock is provided that blocks the opening ofthe valves unless a large turbine load decrease or a turbine trip has occurred.

For all cases, the plantrapidly reaches a stabilized condition at a higher power level. Normal plant operating procedures wouldbe followed to reduce power. The excessive load increase incident is an overpower transient for which thefuel temperatures will rise. RT may not occur for some cases, and the plant will reach a new equilibrium condition at a higher power level corresponding to the increase in steam flow. Protection against anexcessive load increase

incident, if necessary, is provided by the following RT signals:* OPAT* OTAT* Power range high neutron flux2.2.3.1.2 Input Parameters, Assumptions, and Acceptance CriteriaAn evaluation was performed to show that the DNB design basis is satisfied for the excessive loadincrease incident.

Key aspects of the evaluation are provided below.The RTDP (Reference

1) was applied.

Initial reactor power, RCS pressure, and RCS temperature wereassumed to be at their nominal values, consistent with steady-state full-power operation.

MMF was alsoassumed.

Uncertainties in initial conditions were accounted for in the safety analysis DNBR limit value,as described in Reference 1.The evaluation was performed for a step-load increase of 10 percent steam flow from 100 percent of corepower.The higher end of the Ta,,g range (570.77F to 588.4°F) is applied because it minimizes the initial margin tothe DNBR limit.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-55The higher end of the full-power T'eed range (400.0°F to 448.6°F) is applied, although the event is not verysensitive to Tfeed.Zero percent SGTP is modeled because this maximizes the primary-to-secondary heat transfer area, whichis conservative for maximizing the cooldown of the RCS.The pressurizer heaters are not modeled because the pressurizer heaters would actuate to try and raise thepressurizer

pressure, which is not conservative with respect to minimizing DNBR.The pressurizer sprays and PORVs are modeled to limit any RCS pressure increase.

A lower RCSpressure is conservative for DNBR calculations.

Although the OTAT, OP AT, and power range high neutron flux RTs are available to mitigate the event,the analysis conservatively does not credit these trips.No credit is taken for the heat capacity of the RCS and SG metal mass in attenuating the resulting plantcooldown.

This event is analyzed with automatic and manual rod control.Because the event is not sensitive to the initial pressurizer and SG levels, the pressurizer level and SGlevel are modeled to be at the nominal values consistent with steady-state full-power operation.

The event is analyzed for both the beginning-of-life

((BOL) minimum reactivity feedback) and EOL(maximum reactivity feedback) conditions.

Based on its frequency of occurrence, the excessive load increase incident is considered to be aCondition II event as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

The following items summarize the acceptance criteria associated with this event:* The critical heat flux (CHF) should not be exceeded.

This is met by demonstrating that theminimum DNBR does not go below the SAL value at any time during the transient.

  • Pressures in the RCS and MSS should be maintained below 110 percent of the respective designpressures.
  • The peak linear heat generation rate (expressed in kW/ft) should not exceed a value that wouldcause fuel centerline melt. This criterion is satisfied by demonstrating that the core average heatflux remains below the limit of 121 percent of the applied nominal core thermal power during theevent.* An incident of moderate frequency should not generate a more serious plant condition withoutother faults occurring independently.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-562.2.3.1.3 Description of Analyses and Evaluations The excessive load increase event is analyzed using the RETRAN computer code described inWCAP-14882-P-A (Reference 2). The RETRAN code model simulates the RCS, neutron kinetics, pressurizer, pressurizer relief and safety valves, pressurizer

heaters, pressurizer spray, SG, FWsystem, andMSSVs. The code computes pertinent plant variables including SG mass, pressurizer water volume,reactor coolant average temperature, RCS pressure, and SG pressure.

For the minimum reactivity feedback cases, the core has the least-negative MTC of reactivity and theleast-negative Doppler-only power coefficient curve, and, therefore, the least-inherent transient responsecapability.

For the maximum reactivity feedback cases, the core has the most-negative MTC of reactivity and the most-negative Doppler-only power coefficient curve. This results in the largest amount ofreactivity feedback due to changes in coolant temperature.

Normal reactor control systems and engineered safety systems are not required to function.

2.2.3.1.4 ResultsThe analysis results for the 10 percent load increase event from full-power conditions show that in allcases analyzed the minimum DNBR remains above the SAL value and the peak linear heat generation does not exceed the limit value, thus demonstrating that the fuel cladding integrity and fuel centerline melt acceptance criteria are met. The peak pressurizer water volume remains below the total volume ofthe pressurizer, demonstrating that this event does not generate a more serious plant condition.

Following the initial load increase, the plant reaches a stabilized (steady-state) condition.

The increase in the MSSflow rate results in a cooldown of the RCS and a decrease in the MSS pressure.

The RCS and MSSpressure limits are not challenged during the event. The analysis inputs are intended to minimize theresultant minimum DNBR and not to maximize RCS and MSS pressures.

The case that models minimum reactivity feedback conditions with automatic rod control is the mostlimiting case with respect to minimum DNBR. The key results are summarized in Table 2.2.3-1.

The timesequence of events for each case is provided in Table 2.2.3-2.

The transient responses for the four casesare shown in Figures 2.2.3-1 through 2.2.3-4.2.2.3.2 Conclusion The excessive load increase analysis demonstrates that for this event at the WCGS, the DNBR does notdecrease below the SAL value at any time during the transient for all cases. Also, the peak core averagepower (heat flux) remains below the limit of 121 percent of the applied nominal core thermal power; thus,no fuel or cladding damage is predicted.

The event does not challenge the primary and secondary sidepressure limits because the increased heat removal cools the RCS and depressurizes the MSS. The peakpressurizer water volume remains below the total volume of the pressurizer, demonstrating that this eventdoes not generate a more serious plant condition.

All applicable acceptance criteria are met for theWCGS.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-572.2.3.3 References

1. WCAP-11397-P-A (Proprietary) and WCAP- 11397-A (Non-Proprietary),

"Revised ThermalDesign Procedure,"

April 1989.2. WCAP- 14882-P-A (Proprietary) and WCAP- 15234-A (Non-Proprietary),

"RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA SafetyAnalyses,"

April 1999 and May 1999, respectively.

WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-58Table 2.2.3-1 Excessive Load Increase Incident Summary of ResultsCore Heat FluxCase Minimum DNBR (FOI)Limits 1.52 1.21Minimum reactivity

feedback, manual rod control 2.29 1.02Minimum reactivity
feedback, automatic rod control 1.97 1.11Maximum reactivity
feedback, manual rod control 2.03 1.10Maximum reactivity
feedback, automatic rod control 2.00 1.10WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-59WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-59Table 2.2.3-2 Time Sequence of Events for the Excessive Load Increase IncidentCase Event Time of Event (seconds)

Minimum reactivity

feedback, manual rod 10% step load increase 0.0controlMinimum DNBR reached 6.7Steady-state conditions reached -300"(approximate)

Peak heat flux reached 382.5Minimum reactivity

feedback, automatic 10% step load increase 0.0rod controlPeak heat flux reached 265.6Steady-state conditions reached -300"(approximate)

Minimum DNBR reached 377.6Maximum reactivity

feedback, manual rod 10% step load increase 0.0controlSteady-state conditions reached -350("(approximate)

Peak heat flux reached 362.8Minimum DNBR reached 395.7Maximum reactivity

feedback, automatic 10% step load increase 0.0rod controlPeak heat flux reached 368.5Minimum DNBR reached 398.2Steady-state conditions reached -400")(approximate)

Note:I. Time of equilibrium (steady-state conditions reached) was selected based on when the nuclear power leveled out aftertransient initiation.

WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-60a,=3EnCn10501000950S900,9o0i 850" 800"750'0 100 200 300Time (s)400Figure 2.2.3-110% Step Increase in Heat Removal by Secondary SystemMinimum Reactivity

Feedback, Manual Reactor ControlWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-61WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-61100 200 300Time (s)400595i-590-585", 580"--575".570"in 565-.. ........................ ...........

.. ..................................

.....................................

..............

.....................................................................................................................................I..0100200Time (s)300400Figure 2.2.3-110% Step Increase in Heat Removal by Secondary SystemMinimum Reactivity

Feedback, Manual Reactor Control (cont.)WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-62a)00~0U).- 2260"0_2250-I 2240"2230-2220080-1070"1060'E-Z 1050"" " 1040"1030--10200 l2O0 100 200 300Time (s)400Figure 2.2.3-210% Step Increase in Heat Removal by Secondary SystemMinimum Reactivity

Feedback, Automatic Reactor ControlWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-630 100 200 300Time (s)400rnrEC3Uf)590-585-580-575-570-565-...........................................................

............................................................................

..........................................................F .............................-JVu01002;0Time (s)300400Figure 2.2.3-210% Step Increase in Heat Removal by Secondary SystemMinimum Reactivity

Feedback, Automatic Reactor Control (cont.)WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-64C:)(-0nCd,f-)200Time (s)Figure 2.2.3-3 10% Step Increase in Heat Removal by Secondary SystemMaximum Reactivity

Feedback, Manual Reactor ControlWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-65z:0 100 200 300Time (s)400rnr590-585-580-E575-a:)> 570-g' 565-............................................

.............

.........................................................

.. ..............................................................................JDoU0100200Ti me (s)300400Figure 2.2.3-310% Step Increase in Heat Removal by Secondary SystemMaximum Reactivity

Feedback, Manual Reactor Control (cont.)WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-660a)00~0a)=3.225022452240"&_ 2235-2230106010501040"011030"U 1020"(n9a)0'i200Time (s)Figure 2.2.3-410% Step Increase in Heat Removal by Secondary SystemMaximum Reactivity

Feedback, Automatic Reactor ControlWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-67WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-67Z0 100 200 300Time (s)400595: 590-585-_ 580-EF---575-> 570-o565-.................

...............

.................

.................

.................

...........................

............

.......................................................................................................0100200Time (s)300400Figure 2.2.3-4 10% Step Increase in Heat Removal by Secondary SystemMaximum Reactivity

Feedback, Automatic Reactor Control (cont.)WCAP-17658-NP Licensing ReportALIgust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-682.2.4 Inadvertent Opening of a Steam Generator Atmospheric Relief or Safety Valve(USAR Section 15.1.4)2.2.4.1 Technical Evaluation 2.2.4.1.1 Introduction The most severe core conditions resulting from an accidental depressurization of the MSS are associated with an inadvertent opening of a SG atmospheric relief or safety valve. More specifically, an accidental depressurization of the MSS is a transient that is analyzed to bound the opening of a single turbine bypassvalve or SG ARV because the inadvertent opening of one of these control valves is most likely to occur.Conversely, the inadvertent opening of a SG safety valve is not nearly as likely to occur because thedesign of a spring-loaded safety valve is passive in nature. However, because the relief capacity of aSG safety valve is larger than those associated with either of the other two types of valves, the failure of aSG safety valve is also conservatively considered in the analysis of an accidental depressurization of theMSS. The analyses that consider a major rupture of a main steam pipe are presented in Section 2.2.5.The steam release, as a consequence of an accidental depressurization of the MSS, results in an initialincrease in steam flow, followed by a decrease in steam flow during the rest of the accident as theSG pressure decreases.

The increased energy removal from the RCS causes a decrease in the reactorcoolant temperature and pressure.

In the presence of a negative MTC, the cooldown results in a positivereactivity insertion.

The primary design features that provide protection for accidental depressurizations of the MSS are:Actuation of the SI system on any of the following:

-Two-out-of-four low pressurizer pressure signals-Two-out-of-three low steam line pressure signals in any one loopActuation of a RT from the overpower signals (neutron flux and AT) or upon the receipt of an SIsignal.Redundant isolation of the main FW lines to prevent sustained main FW flow, which would causeadditional cooldown.

In addition to the primary means of protection, where an SI signal closes themain FW isolation valves, an SI signal will also rapidly close all main FW control valves andcontrol bypass valves, trip the main FW pumps, and close the FW pump discharge valves.Closure of the fast-acting MSIVs on the following:

-Two-out-of-three low steam line pressure signals in any one loop(above Permissive P-1 1)-Two-out-of-three high negative steam pressure rate signals in any one loop(below Permissive P-I 1)WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-692.2.4.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe following summarizes the major input parameters and/or assumptions used in the analysis of anaccidental depressurization of the MSS event at HZP conditions:

HZP conditions were modeled with four loops in service and with offsite power available.

A steam flow rate of approximately 240 Ibm/sec at 1075 psia (corresponding to 268 lbm/sec at1200 psia) was analyzed.

This flow rate corresponds to the maximum capacity of any singleturbine bypass, atmospheric relief, or safety valve.Minimum SGTP (0 percent) was modeled to conservatively maximize primary-to-secondary heattransfer.

  • All control rods were modeled to be inserted except the most reactive RCCA, which was assumedto be stuck out of the core.* A minimum, EOL shutdown margin corresponding to 1.30% Ak/k was modeled at eventinitiation.
  • The SI system was modeled with a conservatively low flow capability, corresponding to only onehigh-head SI (centrifugal charging) pump injecting through the cold legs (CLs).* The flow from the SI system that is delivered to the RCS was modeled with a temperature andboron concentration consistent with the minimum values for the refueling water storage tank, asrequired by the TS.* The low pressurizer pressure signal was credited for SI system actuation.

In addition, the SIsignal that results from the low pressurizer pressure signal was credited for isolation of the mainFW lines.* The accumulators were modeled to be available;

however, CL pressures never decreased to thepoint where flow from the accumulators was injected.
  • The AFW system was modeled with a conservatively high flow capability, corresponding to allAFW pumps operating at maximum capacity and the maximum flow possible being delivered tothe faulted SG.An accidental depressurization of the MSS is classified as a Condition II event, an incident of moderatefrequency, as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

The pressure limits for the primary and secondary systems are not challenged for this accident because thepressures in these systems each decrease from their initial values during the transient.

The only criterion that has the potential to be challenged during this event is that associated with fuel damage. The analysisdemonstrates that this criterion is met by showing that the DNB design basis is met. That is, this analysisWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-70shows that the minimum DNBR does not go below the limit value at any time during the transient.

Inaddition, it has been historical practice to assume that fuel failure will occur if centerline melting takesplace. Therefore, the analysis also demonstrates that the peak linear heat generation rate (expressed in kW/ft) does not exceed the value that would cause fuel centerline melt.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the accidental depressurization of the MSS event acceptance criteria areprovided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the accidental depressurization of the MSS event, this is shown to be met bydemonstrating that the fuel damage criterion is satisfied.

GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.

For theaccidental depressurization of the MSS event, this is shown to be met by demonstrating that thepeak RCS pressure is less than 110 percent of the design pressure.

GDC 20 (Protection System Functions) requires that the protection system be designed to initiateautomatically the operation of appropriate systems including the reactivity control systems, sospecified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences, and to sense accident conditions and initiate the operation of systems andcomponents important to safety. For the accidental depressurization of the MSS event, this isshown to be met by demonstrating that the fuel damage criterion is satisfied.

GDC 26 (Reactivity Control System Redundancy and Capability) requires that one of thereactivity control systems uses control rods capable of reliably controlling reactivity changes withappropriate margin for malfunctions like stuck rods so that specified acceptable fuel design limitsare not exceeded under conditions of normal operation, including anticipated operational occurrences.

For the accidental depressurization of the MSS event, this is shown to be met bydemonstrating that the fuel damage criterion is satisfied.

2.2.4.1.3 Description of Analyses and Evaluations A detailed analysis was performed using the RETRAN computer code (Reference

1) to determine theplant transient conditions following an accidental depressurization of the MSS. The RETRAN modelsimulates the core neutron kinetics, RCS, pressurizer, SGs, SI system and AFW system. To properlymodel the system response to this event and prevent any non-physical behavior from being predicted when the pressurizer
refills, the pressurizer and surgeline were modeled as a single volume, as comparedto the nodalization documented in Reference
1. The code computes pertinent plant variables, including the core heat flux and reactor coolant temperature and pressure.

A detailed core analysis was thenperformed using the ANC code (Reference

2) to confirm the validity of the RETRAN-predicted reactivity WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-71feedback model. The core models developed in ANC were also used to calculate the power peakingfactors that were used as input to the detailed T/H digital computer code, VIPRE (Reference 3), whichwas used with a DNB correlation applicable to the low pressure condition (Reference

4) to determine ifthe DNB design basis was met. In addition, the core models developed in ANC were used to calculate thepeak linear heat generation rate.2.2.4.1.4 ResultsThe calculated sequence of events for an accidental depressurization of the MSS at HZP initial conditions with offsite power available is shown in Table 2.2.4-1, and the limiting results are presented inTable 2.2.4-2.Figures 2.2.4-1 through 2.2.4-7 show the transient results for an accidental depressurization of the MSS.Because offsite power was assumed to be available in this analysis, there is full reactor coolant flow.If the core were to be critical at or near HZP conditions when the accidental depressurization occurs, theinitiation of SI via a low pressurizer pressure signal would trip the reactor.

In addition, sustained main FWflow is prevented by the isolation of the main FW lines on the SI signal that results from the lowpressurizer pressure signal.As shown in Figure 2.2.4-4, the core attains criticality with the RCCAs inserted (i.e., with the plant shutdown assuming one stuck RCCA) before the transient is effectively terminated by boron injected from theSI system.The results of the analysis of an accidental depressurization of the MSS event demonstrate that the DNBdesign basis is met. The calculated minimum DNBR is well above the limit value. In addition, the peaklinear heat generation rate (expressed in kW/ft) does not exceed the value that would cause fuel centerline melt. The pressure limits for the primary and secondary systems are not challenged for this accidentbecause the pressures in these systems each decrease from their initial values during the transient.

Therefore, this event does not adversely affect the core or the RCS, and all applicable acceptance criteriaare met.2.2.4.2 Conclusion The analysis of the accidental depressurization of the MSS described above has been reviewed.

It isconcluded that the analysis has adequately accounted for operation of the plant at the analyzed powerlevel and was performed using acceptable analytical models. It is further concluded that the analysis hasdemonstrated that the reactor protection and safety systems will continue to ensure that the applicable safety analysis design limits and the RCPB pressure limits will not be exceeded as a result of this event.Based on this, the conclusion is that the plant will continue to meet the requirements of GDCs 10, 15, 20,and 26.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-722.2.4.3 References

1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.2. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.4. WCAP-14565-P-A Addendum 2-P-A, "Addendum 2 to WCAP-14565-P-A Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low PressureApplications,"

April 2008.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-73Table 2.2.4-1 Time Sequence of Events -Accidental Depressurization of the MSS at HZP Conditions Case Event Time (sec)Inadvertent Opening of a Accidental Depressurization of the MSS Occurs 0.0Single Turbine Bypass,Atmospheric Relief, or Safety Pressurizer Empties 209.5Valve Low Pressurizer Pressure Setpoint Reached 216.3SI Signal Generated (on low pressurizer pressure) 218.3FW Isolation Complete 233.3SI Flow Initiated 243.3Core Re-criticality Occurs 270.5Borated Water from SI System Reaches the Core 578.8Peak Core Heat Flux Reached 581.5Core Becomes Subcritical 590.5Table 2.2.4-2 Limiting Results -Accidental Depressurization of the MSS at HZP Conditions Case Parameter Analysis Value LimitInadvertent Opening of a Single Minimum DNBR 5.10 1.18Turbine Bypass, Atmospheric Relief, or Safety Valve Peak Linear Heat Generation 6.924 22.4(kW/ft)WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-74E00<C)U-0.-50 100 200 300 400 500 600 700Time (sec)E000C.)U-2,i,0,-t-0 100 200 300 400Time (sec)500 600 700Figure 2.2.4-1Accidental Depressurization of the MSS at HZPNuclear Power and Core Heat Flux versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-75Faulted LoopIntact LoopsC-0,E(D0)0)E0)c-0)C,)a-)0 100 200 300 400Time (sec)500 600 700Qnn .U-0)v- 550-E 500-I--0).......................................'fU0 100 200 300Time1400(sec)500 600700Figure 2.2.4-2Accidental Depressurization of the MSS at HZPReactor Vessel Inlet Temperature and Core Average Temperature versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-76WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-76Cl)(r)CL):30nCl)Cl)E-5a,0 IO0 200 300 400Time (sec)500 600 7000 100 200 300 400Time (sec)500 600 700Figure 2.2.4-3Accidental Depressurization of the MSS at HZPPressurizer Pressure and Pressurizer Water Volume versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-77F-M_0000Of0 100 200 300 400Time (sec)500 600 7000 100 200 300 400Time (sec)500 600 700Figure 2.2.4-4Accidental Depressurization of the MSS at HZPCore Boron Concentration and Reactivity versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-78WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-78Foul t ed LoopI... Intact Loopsca-)Cl)UPECD,00 100 200 300 400 500 600 700Time (sec)0 100 200 300 400Time (sec)500 600 700Figure 2.2.4-5Accidental Depressurization of the MSS at HZPSteam Pressure and Steam (Break) Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-79Foul ted LoopIntact Loops')1A-.0200150Eo 100-50ELtE .... ... 1.............

.............

.............

I~~ ~ ............U1 I00 200 300Time400(sec)500 600700Foul ted LoopIntact LoopsLJUUUUu~00E-4--U-)200000-150000-.............................

......................................

1 M nn ..I ..I .IIUUUUU0 100 200 300 400Time (sec)500 600700Figure 2.2.4-6Accidental Depressurization of the MSS at HZPFW Flow and SG Mass versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-80WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-801.4U--.0U-01.2-10.80.6-0.4-0.2-..........I I I lU0 100 200 300Time400(sec)500 600700Figure 2.2.4-7Accidental Depressurization of the MSS at HZPCore Flow versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCkP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-812.2.5 Steam System Piping Failure (USAR Section 15.1.5)2.2.5.1 Steam System Piping Failure at Hot Zero Power Conditions 2.2.5.1.1 Technical Evaluation 2.2.5.1.1.1 Introduction The steam release following a main steam pipe rupture would result in an initial increase in steam flowthat decreases during the accident as the steam pressure decreases.

The increased energy removal from theRCS causes a decrease in the reactor coolant temperature and pressure.

In the presence of a negativeMTC, the cooldown results in a positive reactivity insertion and subsequent reduction in core shutdownmargin. If the most-reactive RCCA is assumed stuck in its fully withdrawn position after RT, there is anincreased possibility that the core will become critical and return to power. A return to power following asteam pipe rupture is a concern primarily because of the high power peaking factors that would exist withthe most-reactive RCCA assumed to be stuck in its fully withdrawn position.

The major rupture of a main steam pipe is the most limiting cooldown transient.

It is analyzed at HZPconditions with no decay heat (decay heat would retard the cooldown, thus reducing the potential returnto power). A detailed discussion of this transient with the most limiting break size (i.e., a double-ended rupture) is presented below.The primary design features that provide protection for steam pipe ruptures are:* Actuation of the SI system on any of the following:

-Two-out-of-four low pressurizer pressure signals-Two-out-of-three low steam line pressure signals in any one loop-Two-out-of-three high-1 containment pressure signals* Actuation of a RT from the overpower signals (neutron flux and AT) or upon the receipt of an SIsignal.Redundant isolation of the main FW lines to prevent sustained main FW flow, which would causeadditional cooldown.

In addition to the primary means of protection, where an SI signal closes themain FW isolation valves, an SI signal will also rapidly close all main FW control valves andcontrol bypass valves, trip the main FW pumps, and close the FW pump discharge valves.Closure of the fast-acting MSIVs on the following:

-Two-out-of-three high-2 containment pressure signalsWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-82Two-out-of-three low steam line pressure signals in any one loop(above Permissive P- 1i)Two-out-of-three high negative steam pressure rate signals in any one loop(below Permissive P-11)For any break, in any location, no more than one SG would experience an uncontrolled

blowdown, even ifone of the MSIVs were to fail to close. For breaks downstream of the MSIVs, closure of all MSIVs wouldcompletely terminate the blowdown from all of the SGs. Thus, even with the worst possible breaklocation (i.e., upstream of an MSIV), only one SG can blow down, minimizing the potential steam releaseand resultant RCS cooldown and depressurization.

The remaining SGs would still be available fordissipation of decay heat after the initial transient is over.Following blowdown of the faulted SG, the plant can be brought to a stabilized, hot standby condition through control of AFW flow and SI flow, as prescribed by plant operating procedures.

The operating procedures call for operator action to limit RCS pressure and pressurizer level by terminating SI flow, andto control SG level and reactor coolant temperature using the AFW system.2.2.5.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe following summarizes the major input parameters and/or assumptions used in the analysis of a majorrupture of a main steam pipe event at HZP conditions:

  • HZP conditions were modeled with four loops in service, both with and without offsite poweravailable.

0 A 1.388 ft2 break size was analyzed.

This break size corresponds to the maximum effective throatarea of the integral flow restrictor that is built into the steam outlet nozzle of each SG.0 Minimum SGTP (0 percent) was modeled to conservatively maximize primary-to-secondary heattransfer.

  • All control rods were modeled to be inserted except the most reactive RCCA, which was assumedto be stuck out of the core.0 A minimum, end-of-life shutdown margin corresponding to 1.30% Ak/k was modeled at eventinitiation.
  • The SI system was modeled with a conservatively low flow capability, corresponding to only onehigh-head SI (centrifugal charging) pump injecting through the CLs.* The flow from the SI system that is delivered to the RCS was modeled with a temperature andboron concentration consistent with the minimum values for the refueling water storagetank (RWST), as required by the TS.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-83The low steam line pressure signal was credited for SI system actuation and closure of theMSIVs. In addition, the SI signal that results from the low steam line pressure signal was creditedfor isolation of the main FW lines.The accumulators were modeled to be available;

however, the flow that is injected from theaccumulators was conservatively modeled to have a boron concentration of 0.0 ppm.The AFW system was modeled with a conservatively high flow capability, corresponding to allAFW pumps operating at maximum capacity and the maximum flow possible being delivered tothe faulted SG.A major break in a steam system pipe is classified as a Condition IV event, a limiting fault, as defined bythe ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"ANSI N 18.2-1973.

Minor secondary system pipe breaks are classified as ANS Condition III events,infrequent incidents.

However, the major rupture of a main steam pipe event was conservatively analyzedto meet the more restrictive acceptance criteria associated with a Condition II event.The pressure limits for the primary and secondary systems are not challenged for this accident because thepressures in these systems each decrease from their initial values during the transient.

The only criterion that has the potential to be challenged during this event is that associated with fuel damage. The analysisdemonstrates that this criterion is met by showing that the DNB design basis is met. That is, this analysisshows that the minimum DNBR does not go below the limit value at any time during the transient.

Inaddition, it has been historical practice to assume that fuel failure will occur if centerline melting takesplace. Therefore, the analysis also demonstrates that the peak linear heat generation rate (expressed in kW/ft) does not exceed the value that would cause fuel centerline melt.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the major rupture of a main steam pipe event acceptance criteria areprovided as follows.GDC 27 (Combined Reactivity Control Systems Capability) requires that the reactivity controlsystems be designed to have a combined capability, in conjunction with poison addition by theECCS, of reliably controlling reactivity changes so that, under postulated accident conditions andwith appropriate margin for stuck rods, the capability to cool the core is maintained.

For thesteam system piping failure at HZP conditions event, this is shown to be met by demonstrating that the fuel damage criterion is satisfied, which ultimately ensures that the ability to insertcontrol rods is maintained.

GDC 28 (Reactivity Limits) requires that the reactivity control system be designed withappropriate limits on the potential amount and rate of reactivity increase so that the effects ofpostulated reactivity accidents can neither result in damage to the RCPB greater than limited localyielding, nor sufficiently disturb the core, its support structures, or other reactor pressurevessel (RPV) internals to impair significantly the capability to cool the core. For the steam systempiping failure at HZP conditions event, this is shown to be met by demonstrating that the peakWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-84RCS pressure is less than 110 percent of the design pressure, which ultimately ensures that theRCPB pressure limits are not exceeded.

GDC 35 (Emergency Core Cooling) requires that the RCS and associated auxiliaries be designedwith a safety system able to provide abundant emergency core cooling.

For the steam systempiping failure at HZP conditions event, this is shown to be met by demonstrating that the fueldamage criterion is met, which ultimately shows that the ECCS provides abundant core cooling,even with the most-limiting single failure considered.

2.2.5.1.1.3 Description of Analyses and Evaluations A detailed analysis was performed using the RETRAN computer code (Reference

1) to determine theplant transient conditions following a major rupture of a main steam pipe with and without offsite poweravailable.

The RETRAN model simulates the core neutron kinetics, RCS, pressurizer, SGs, SI system, andAFW system. To properly model the system response to this event and prevent any non-physical behaviorfrom being predicted when the pressurizer

refills, the pressurizer and surgeline were modeled as a singlevolume, as compared to the nodalization documented in Reference
1. The code computes pertinent plantvariables, including the core heat flux and reactor coolant temperature and pressure.

A detailed coreanalysis was then performed for the case that assumes offsite power is available using the ANC code(Reference

2) to confirm the validity of the RETRAN-predicted reactivity feedback model. The coremodels developed in ANC were also used to calculate the power peaking factors that were used as inputto the detailed T/H digital computer code, VIPRE (Reference 3), which was used with a DNB correlation applicable to the low pressure condition (Reference
4) to determine if the DNB design basis was met. Inaddition, core models developed in ANC were used to calculate the peak linear heat generation rate.The detailed core and DNB analyses for the case that assumes a loss of offsite power (LOOP) are notperformed as the transient resulting at low RCS flow conditions has been judged to be less limiting thanthat resulting when full RCS flow is maintained.

This is based on the fact that, as RCS forced flowdecreases, heat transfer across the SG tubes also decreases.

This decrease in heat transfer significantly reduces the rate and magnitude of the RCS cooldown and, consequently, the final return to power level isalso lower. The drop in RCS pressure is not as significant for this case as well. Furthermore, the loss offorced reactor coolant flow allows for more uniform flow and temperature distributions at the lowerreactor plenum such that, as the coolant travels up into the core inlet and through the active core region,the radial temperature asymmetry is not as significant as in the case where forced flow is maintained.

Thisreduced asymmetric temperature distribution results in less significant power peaking factors.The reduced power peaking in the reactor core, combined with a lower return to power and higher RCSpressure, more than offsets the penalty associated with reduced flow at the location of minimum DNBR.As such, the minimum DNBR for such a case would be higher than that calculated for the case whereoffsite power remains available.

This conclusion has also been previously validated with the linkedneutronics and T/H code systems.

Therefore, consistent with the previous discussion, only the DNBR ofthe most limiting case (with offsite power available) is presented herein.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-852.2.5.1.1.4 ResultsThe calculated sequence of events for the complete severance of a main steam pipe at HZP initialconditions, both with and without offsite power available, is shown in Table 2.2.5.1-1.

The results for themost limiting case, the case with offsite power available, are presented in Table 2.2.5.1-2.

Figures 2.2.5.1-1 through 2.2.5.1-7 show the transient results for the case with offsite power available.

Because offsite power was assumed to be available in this case, there is full reactor coolant flow.Figures 2.2.5.1-8 through 2.2.5.1-14 show the transient results for the case without offsite poweravailable.

Because offsite power was assumed to be lost in this case, the RCPs coast down and there is adecrease in reactor coolant flow. As can be seen from Figures 2.2.5.1-1 and 2.2.5.1-8, the return to poweris considerably less significant for the case without offsite power, even when using the conservative RETRAN reactivity model calibrated for conditions where forced RCS flow was maintained.

Similarly, Figures 2.2.5.1-3 and 2.2.5.1-10 show that RCS pressure is higher for the case with reduced RCS flow.If the core were to be critical at or near HZP conditions when the rupture occurs, the initiation of SI via alow steam line pressure signal would trip the reactor.

Steam release from more than one SG is prevented by the automatic closure of the MSIVs in the steam lines on a low steam line pressure signal. In addition, sustained main FW flow is prevented by the isolation of the main FW lines on the SI signal that resultsfrom the low steam line pressure signal.As shown in Figures 2.2.5.1-4 and 2.2.5.1-11 for the cases with and without offsite power available, respectively, the core attains criticality with the RCCAs inserted (i.e., with the plant shut down assumingone stuck RCCA) before the transient is effectively terminated by boron injected from the SI system.The results of the analysis of a major rupture of a main steam pipe event demonstrate that the DNB designbasis is met. The calculated minimum DNBR is well above the limit value for the limiting case thatassumes offsite power is available.

In addition, the peak linear heat generation rate (expressed in kW/ft)does not exceed the value that would cause fuel centerline melt. The pressure limits for the primary andsecondary systems are not challenged for this accident because the pressures in these systems eachdecrease from their initial values during the transient.

Therefore, this event does not adversely affect thecore or the RCS, and all applicable acceptance criteria are met.2.2.5.1.2 Conclusion The analysis of the steam system piping failure at HZP conditions described above has been reviewed.

Itis concluded that the analysis has adequately accounted for operation of the plant at the analyzed powerlevel and was performed using acceptable analytical models. It is further concluded that the analysis hasdemonstrated that the reactor protection and safety systems will continue to ensure that the ability toinsert control rods is maintained, the RCPB pressure limits will not be exceeded, and abundant corecooling will be provided.

Based on this, the conclusion is that the plant will continue to meet therequirements of GDCs 27, 28 and 35.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-862.2.5.1.3 References

1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.2. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.3. WCAP-14565-P-A, "VIPRE-0l Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.4. WCAP- 14565-P-A Addendum 2-P-A, "Addendum 2 to WCAP- 14565-P-A Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low PressureApplications,"

April 2008.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-87Table 2.2.5.1-1 Time Sequence of Events -Steam System Piping Failure at HZP Conditions Case Event Time (sec)Double-Ended Rupture SLB Occurs 0.0(1.388 ft2) with Offsite PowerAvailable Low Steam Line Pressure Setpoint Reached in the FaultedLoop (lead-lagged) 0.6SI Signal Generated (on low steam line pressure) 2.6Pressurizer Empties 15.8Steam Line and FW Isolation Complete 17.6Core Re-criticality Occurs 19.3SI Flow Initiated 27.6Peak Core Heat Flux Reached 320.0Borated Water from SI System Reaches the Core 320.8Core Becomes Subcritical 327.5Accumulators Begin to Inject (Unborated Water) 343.5Double-Ended Rupture SLB Occurs 0.0(1.388 ft2) without.

OffsitePower Available Low Steam Line Pressure Setpoint Reached in the Faulted 0.6Loop (lead-lagged)

SI Signal Generated (on low steam line pressure) 2.6RCPs Begin to Coast Down 3.0Steam Line and FW Isolation Complete 17.6Pressurizer Empties 18.0Core Re-criticality Occurs 26.0SI Flow Initiated 39.6Borated Water from SI System Reaches the Core 366.5Peak Core Heat Flux Reached 374.5Core Becomes Subcritical 384.3WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-88WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-88Table 2.2.5.1-2 Limiting Results -Steam System Piping Failure at HZP Conditions Case Parameter Analysis Value LimitDouble-Ended Rupture Minimum DNBR 1.80 1.18(1.388 ft2) with Offsite PowerAvailable Peak Linear Heat Generation (kW/ft) 15.829 22.4WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-89WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-89¢-E000(-)LL_0n00~¢-E00oU-¢C-)a 100 200 300Time (sec)4000 100 200 300Time (sec)400Figure 2.2.5.1-1 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)

Nuclear Power and Core Heat Flux versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-90Faulted LoopIntact LoopsQ)M.)Ea)c-Q)Q)0-0>a,0'C.)U_)0:O)0 100 200 300Time (sec)4000 100 200 300Time (sec)400Figure 2.2.5.1-2 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)

Reactor Vessel Inlet Temperature and Core Average Temperature versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-910na,)C/)0nEa,'P,CI-0 100 200 300Time (sec)4000 100 200 300 400Time (sec)Figure 2.2.5.1-3 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)

Pressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-920~0ciCDW-c)ý0 100 200 300Time (sec)4000 100 200 300Time (sec)400Figure 2.2.5.1-4 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)

Core Boron Concentration and Reactivity versus TimeWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-93Four ted LoopI ntact LoopsCI)0.)Cj)cn0 1O0 200 300Time (sec)400Foul ted LoopIntact Loops (Tot a )E0ECD0.)0 100 200 300Time (sec)400Figure 2.2.5.1-5 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)

Steam Pressure and Steam (Break) Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-94WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-94Fau I ted LoopIntact LoopsI DUUCO,E-001200800400..............

......................................................................................................................-7 -'r -i- -T ---00100200Time (sec)300400Faur ted LoopIntact Loopsnnnnn .A4U.-UCf,Ct)0EV.)150000-1000001------ --------- ----------------- ---...........................................................

.............................................................................................................50000"U0100200Time (sec)300400Figure 2.2.5.1-6 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)

FW Flow and SG Mass versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-951.41.2-..........

..............................

C30C_)1-0.8-0.6-0.4-0.2-7............

............

............

............

.................

.................

.................

.................

V0100200Time (sec)300400Figure 2.2.5.1-7 Steam System Piping Failure at HZP(1.388 ft2 Break with Offsite Power Available)

Core Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-96E00D0r)F-00~C-)0 100 200 300Time (sec)4000 1O0 200 300Time (sec)400Figure 2.2.5.1-8 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)

Nuclear Power and Core Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-97Faul ted LoopIntact Loops----- 6007" 550--4 500 -............

..E 450-cI---400 ... ....._....co 350 ..............

2 300-0OnajI,C-Q_0U.200Time (sec)Figure 2.2.5.1-9 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)

Reactor Vessel Inlet Temperature and Core Average Temperature versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-98CL0na,a.)EUf)0 100 200 300Time (sec)400Figure 2.2.5.1-10 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)

Pressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32 -990~00F-)0 100 200 300Time (sec)4000 100 200 300Time (sec)400Figure 2.2.5.1-11 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)

Core Boron Concentration and Reactivity versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-100WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-100Faul ted LoopIntact LoopsV)U)U-30 100 200 300Time (sec)400Faul ted LoopIntact Loops (Total)C-)E0En0 100 200 300Time (sec)400Figure 2.2.5.1-12 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)

Steam Pressure and Steam (Break) Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-101Fau ted LoopIntact Loops1 rflf II UUUEa-)1200-800"r- -12 i --, -T- -I-- r400-U0100200Time (sec)300400Faul ted LoopIntact Loops'ntAAAA.LUUUUUC/)CI)E0n150000------------

-r --0I100200Time (sec)300400Figure 2.2.5.1-13 Steam System Piping Failure at HZP(1.388 ft2 Break without Offsite Power Available)

FW Flow and SG Mass versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-102WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-102C.-)U-"0U-0:C--0 100 200 300Time (sec)400Figure 2.2.5.1-14 Steam System Piping Failure at HZP(1.388 fte Break without Offsite Power Available)

Core Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1032.2.5.2 Steam System Piping Failure at Hot Full Power Conditions 2.2.5.2.1 Technical Evaluation 2.2.5.2.1.1 Introduction A rupture in the MSS piping from an at-power condition creates an increased steam load, which extractsan increased amount of heat from the RCS via the SGs. This results in decreased RCS temperature andpressure.

In the presence of a strong negative MTC, typical of EOC conditions, the colder core inletcoolant temperature causes the core power to increase from its initial level due to the positive reactivity insertion.

The power approaches a level equal to the total steam flow. Depending on the break size, a RTmay occur due to overpower conditions or as a result of a SLB protection function actuation.

The steam system piping failure accident analysis described in Section 2.2.5.1 is performed assuming aHZP initial condition with the control rods inserted in the core, except for the most reactive rod in thefully withdrawn position.

Such a condition could occur the following ways:* When the reactor is at hot shutdown at the minimum required shutdown margin* After the plant has been tripped automatically by the reactor protection system* Manually by the operator.

For an at-power SLB, the analysis of Section 2.2.5.1 represents the limiting condition with respect to coreprotection for the time period following RT. The purpose of this section is to describe the analysis of asteam system piping failure occurring from at-power initial conditions, which demonstrates that coreprotection is maintained prior to and immediately following RT.2.2.5.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe following assumptions are made in the analysis of a main steam line rupture accident at full power:The initial reactor power, pressurizer

pressure, and RCS Tag are assumed to be at the nominal fullpower values. The full power condition is more limiting than part power. The reactor coolant flowrate is the MMF value. The initial loop flows were assumed to be symmetric.

Uncertainties forthe initial conditions of pressurizer

pressure, RCS Tavg, and reactor coolant flow are statistically accounted for in the DNBR limit calculated using the RTDP methodology (Reference 4). InitialNSSS power was conservatively modeled to be at 3651 MWt, which includes all applicable uncertainties.

The full power RCS Tvg range is from 570.70F to 588.40F. Because the full power steam linerupture event is primarily a DNB event, assuming a maximum RCS average temperature islimiting.

Therefore, an initial RCS average temperature of 588.4°F was assumed.The main FW analytical temperature range is from 400'F to 448.6'F.

A higher Tfeed is morelimiting for this event. Thus, a Tfeed of 448.6°F was assumed.WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-104A spectrum of break sizes was analyzed.

Typically, small breaks do not result in a RT; in this casecore power stabilizes at an increased level corresponding to the increased steam flow.Intermediate size breaks may result in a RT on OPAT as a result of the increasing core power.Larger break sizes result in a RT soon after the break from the SI signal actuated by low steamline pressure, which includes lead/lag dynamic compensation.

The limiting break size is thelargest break that does not trip on a low steam pressure SI signal.* To maximize the primary-to-secondary heat transfer rate, 0 percent SGTP is assumed.Maximum moderator reactivity feedback and minimum Doppler power feedback are assumed tomaximize the power increase following the break.The protection system features that mitigate the effects of a SLB are described in Section 2.2.5.1.This analysis only considers the initial phase of the transient from at-power conditions.

Protection in this phase of the transient is provided by RT, if necessary.

Section 2.2.5.1 presents the analysisof the bounding transient following RT, where other protection system features are actuated tomitigate the effects of the SLB.In general, the results would be less severe as a result of normal control system operation.

Therefore, the mitigation effects of control systems have been ignored in the analysis.

However,the main FW control system is assumed to operate in that FW flow is assumed to equal the steamflow prior to RT.Depending on the size of the break, a rupture in a main steam line is classified as either a Condition III(infrequent fault) or Condition IV (limiting fault) event, as defined by the ANS's "Nuclear Safety Criteriafor the Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

However, for ease ofinterpreting the results, the more restrictive criteria associated with Condition II events are applied.

Theapplicable acceptance criteria that may be challenged are that fuel damage due to DNB or fuel centerline melting should be precluded.

Fuel cladding integrity and the prevention of fuel failure is demonstrated byshowing that the calculated minimum DNBR is greater than the applicable limit value. The centerline temperature of the fuel rods with the peak linear heat rate (kW/ft) must not exceed the U02 meltingtemperature.

The pressure limits for the primary and secondary systems are not challenged for thisaccident.

The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the major rupture of a main steam pipe event acceptance criteria areprovided below.GDC 27 (Combined Reactivity Control Systems Capability) requires that the reactivity controlsystems be designed to have a combined capability, in conjunction with poison addition by theECCS, of reliably controlling reactivity changes so that, under postulated accident conditions andwith appropriate margin for stuck rods, the capability to cool the core is maintained.

For thesteam system piping failure at HFP conditions event, this is shown to be met by demonstrating that the fuel damage criterion is satisfied, which ultimately ensures that the ability to insertcontrol rods is maintained.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-105GDC 28 (Reactivity Limits) requires that the reactivity control system be designed withappropriate limits on the potential amount and rate of reactivity increase so that the effects ofpostulated reactivity accidents can neither result in damage to the RCPB greater than limited localyielding, nor sufficiently disturb the core, its support structures, or other RPV internals to impairsignificantly the capability to cool the core. For the steam system piping failure at HFP conditions event, this is shown to be met by demonstrating that the peak RCS pressure is less thanI 10 percent of the design pressure, which ultimately ensures that the RCPB pressure limits arenot exceeded.

2.2.5.2.1.3 Description of Analyses and Evaluations A detailed analysis was performed using the RETRAN computer code to determine the plant transient conditions following a main steam line rupture at full power. Details of the RETRAN model aredocumented in Reference

1. The code computes pertinent variables, including the core power and reactorcoolant temperature and pressure.

Statepoints from RETRAN, consisting of core heat flux, RCS loop inlettemperatures,

pressure, and core flow, are used as input to the DNB analysis and the calculation of thepeak linear heat rate (kW/ft).

A detailed core analysis was performed using the ANC code (Reference 2)to confirm the validity of the RETRAN-predicted reactivity feedback model. The core models developed in ANC were also used to calculate the power peaking factors for input to the DNB analysis and thecalculation of the peak kW/ft. The detailed T/H digital computer code VIPRE (Reference

3) was used tocalculate the DNBR for the limiting time in the transient.

The DNBR calculations were performed usingthe WRB-2 DNB correlation and RTDP methodology (Reference 4).2.2.5.2.1.4 ResultsThe calculated sequence of events for the most limiting break size (1.04 fte) for a main steam line ruptureat full power event is shown in Table 2.2.5.2-1.

This is the largest break that does not trip on a low steamline pressure SI signal. The results for this case are presented in Table 2.2.5.2-2.

Figures 2.2.5.2-1 through 2.2.5.2-4 show the transient response for selected parameters.

The results of the analysis of a major rupture of a main steam pipe event at full power demonstrate thatthe DNB design basis is met. The calculated minimum DNBR is above the limit value. In addition, thepeak linear heat generation rate (expressed in kW/ft) does not exceed the value that would cause fuelcenterline melt. The pressure limits for the primary and secondary systems are not challenged for thisaccident because the pressures in these systems each decrease from their initial values during thetransient.

The steam pressure does not increase significantly following turbine trip. Therefore, this eventdoes not adversely affect the core or the RCS, and all applicable acceptance criteria are met.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1062.2.5.2.2 Conclusion The analysis of the steam system piping failure at full power conditions described above has beenreviewed.

It is concluded that the analysis has adequately accounted for operation of the plant at theanalyzed power level and was performed using acceptable analytical models. It is further concluded thatthe analysis has demonstrated that the reactor protection and safety systems will continue to ensure thatthe ability to insert control rods is maintained, the RCPB pressure limits will not be exceeded, andabundant core cooling will be provided.

Based on this, the conclusion is that the plant will continue tomeet the requirements of GDCs 27 and 28.Although a discussion of the steam system piping failure at full power analysis is not included in thecurrent USAR, Section 15.1.6 will be revised to reflect the analysis described herein.2.2.5.2.3 References

1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.2. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.4. WCAP-11397-P-A, "Revised Thermal Design Procedure,"

April 1989.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-107WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-107Table 2.2.5.2-1 Time Sequence of Events -Steam System Piping Failure at HFP Conditions Case Event Time (sec)Limiting Break Size (1.04 ft-) Steam Line Ruptures 0.0OPAT RT Setpoint Reached 17.7(in two loops)Rods Begin to Drop 20.7Peak Core Heat Flux Occurs 21.5Minimum DNBR Occurs 21.5Table 2.2.5.2-2 Limiting Results -Steam System Piping Failure at HFP Conditions Case Parameter Analysis Value LimitLimiting Break Size (1.04 ft2) Minimum DNBR 2.026 1.52Peak Linear Heat Generation (kW/fi) 21.8 22.4WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-108WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1081.4CD0C-)C-)1.21.00.80.60.4-0.2I I I I I I I I I I I I I ý I I I I I I I I I I ý I I ý I I I I I0.0051015Time20(sec)2530351.4Q(9P0(C90(9-EIF,(9l0(91.2-1.0-0.8-0.4-0.2-0.010U15 20Time (sec)3530Figure 2.2.5.2-1 Steam System Piping Failure at HFP (1.04 ft2 Break)Nuclear Power and Core Heat Flux versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1092300.C 2200-Caa1)f 2100-a:)2000-c)_ 1900-I I I I I I I I I I I I I I III I I I I I I I I I I I I I I I11001000-900-800-700-600-P0515 20Time (Sec)253035I0a)a)500-I II I ; I I I I I I I 105IU15 20Time (see)Figure 2.2.5.2-2 Steam System Piping Failure at HFP (1.04 fe Break)Pressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-1 7658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-110ai tedn c- co op)60C)C)C)C)(1<C-)I I I I I I I I I I I I I I(11G 15 2? 5 31,c, L/Figure 2.2.5.2-3 Steam System Piping Failure at HFP (1.04 ft2 Break)Reactor Vessel Inlet Temperature and Loop Average Temperature versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-111F a "I~ " c 10(n 7 C C c CP1000-~--~ ~crP ~(f~)(U.I I I I I ý ý I ý I I I0 515 2 0 7n C1ý1>0(I.Figure 2.2.5.2-4 Steam System Piping Failure at HFP (1.04 ft2 Break)Steam Pressure and Break Flow versus TimeWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1122.3 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM2.3.1 Loss of External Electrical Load, Turbine Trip, Inadvertent Closure of Main SteamIsolation Valves, Loss of Condenser Vacuum and Other Events Resulting in TurbineTrip (USAR Sections 15.2.2, 15.2.3, 15.2.4, and 15.2.5)2.3.1.1 Technical Evaluation 2.3.1.1.1 Introduction A major load loss on the plant can result from either a loss of external electrical load or from a turbinetrip. A loss of external electrical load can result from an abnormal variation in network frequency or otheradverse network operating conditions.

In either case, offsite power is available for the continued operation of plant components such as the RCPs.The plant is designed to accept a 50 percent loss of electrical load while operating at full power, or acomplete loss of load (LOL) while operating below the P-9 setpoint without actuating a RT with all NSSScontrol systems in automatic.

A 50 percent loss of electrical load is handled by the following:

Steam dump system, which accommodates 40 percent of the nominal full-power load,Rod control system, which accommodates the remaining 10 percent of the load rejection bydriving rods in to reduce coolant average temperature, Pressurizer, which absorbs the change in coolant volume due to the heat addition resulting fromthe load rejection.

Should a 100 percent LOL occur from full power, the reactor protection system automatically actuates aRT. Based on this, a complete LOL from 100 percent power represents the most severe challenge to thesystem and, as such, it is the case explicitly analyzed and described in this section.The most likely source of a complete LOL on the NSSS is a trip of the turbine generator.

In this case, ifthe reactor is operating above the P-9 setpoint, there is a direct RT signal from either the turbine low fluidoil pressure or the turbine stop valve closure.

Reactor temperature and pressure do not increasesignificantly if the steam dump system and pressurizer pressure control system are functioning properly.

However, the RCS and MSS pressure-relieving capacities are designed to ensure the safety of the plantwithout requiring the use of automatic rod control, pressurizer pressure
control, or steam dump controlsystems.

In this analysis, the behavior of the plant is evaluated for a 100 percent loss of steam load fromfull power without direct RT in order to demonstrate the adequacy of the pressure-relieving devices andcore protection margins.In the event the steam dump valves fail to open following a large LOL, the MSSVs can lift and the reactorcan be tripped by the high pressurizer pressure signal, the OTAT signal, or the OPAT signal. The SGshell-side pressure and reactor coolant temperatures increase rapidly.

The PSVs and MSSVs are sized toprotect the RCS and SGs against overpressurization for all load losses without assuming the operation ofWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-113the steam dump system, pressurizer sprays, pressurizer PORVs, automatic rod control, or the direct RT onturbine trip.2.3.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThree cases were analyzed for a loss of load / turbine trip (LOL/TT) event from full-power conditions.

  • Maximum SGTP with automatic pressurizer pressure control (minimum DNBR case)* Minimum SGTP with automatic pressurizer pressure control (peak MSS pressure case)* Maximum SGTP without automatic pressurizer pressure control (peak RCS pressure case)The minimum DNBR case was analyzed using the RTDP (Reference 1). The initial NSSS power wasconservatively modeled to be at 3651 MWt, which includes all applicable uncertainties.

RCS temperature and pressurizer pressure were assumed to be at their nominal values consistent with steady-state, full-power operation.

MMF was modeled.

Uncertainties in initial conditions were included in the safetyanalysis DNBR limit, as described in Reference 1.The peak RCS and MSS pressure cases were analyzed with uncertainties on RCS temperature andpressurizer pressure applied in the direction required to obtain the most conservative initial plantconditions for the transient.

Both cases modeled TDF.The LOL/TT transient was conservatively analyzed with minimum reactivity feedback (beginning of corelife). All cases assumed the least-negative DPC and a 0 pcm/°F MTC, which bounds part-power conditions with a positive MTC. Minimum reactivity feedback conditions are conservative becausereactor power is maintained until the time of RT, which exacerbates the calculated minimum DNBR andpeak RCS and MSS pressures.

Manual rod control was modeled for all cases. If the reactor had been in automatic rod control, the controlrod banks would have been driven into the core prior to RT, thereby reducing the severity of the transient.

The LOL/TT event was analyzed both with and without automatic pressurizer pressure control.

Thepressurizer PORVs and sprays were assumed to be operable for the minimum DNBR case to minimize theincrease in RCS pressure, which is conservative for the calculation of the minimum DNBR. Thepressurizer PORVs and sprays were also assumed to be operable for the peak MSS pressure case tominimize the increase in RCS pressure.

This delays or completely prevents a RT from occurring on a highpressurizer pressure signal, which results in a conservative calculation of the peak MSS pressure.

Thepeak RCS pressure case was analyzed without automatic pressurizer pressure control to conservatively maximize the RCS pressure increase.

In all cases, the MSSVs and PSVs were assumed to be operable.

A total PSV setpoint tolerance of +2 percent was accounted for in the analysis.

For the minimum DNBRcase and the peak MSS pressure case, the negative tolerance was applied to conservatively reduce thesetpoint.

For the peak RCS pressure case, the positive tolerance was applied to conservatively increase thesetpoint.

In addition, the peak RCS pressure case includes a 0.9 percent setpoint shift and a 1.153-second purge time delay to account for the existence of PSV water-filled loop seals, as described in Reference 2.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-114Main FW flow to the SGs was assumed to be lost at the time of turbine trip. The AFW system would beavailable for long-term heat removal.

However, operation of the AFW system is not credited in thetimeframe considered for this analysis.

The following RT functions are assumed to be operable:

  • High pressurizer pressure* OTAT* OPATThe MSSVs were modeled with opening setpoints that account for a maximum setpoint tolerance of3 percent and all appropriate line losses. Valve accumulation was modeled via a 5-psi ramp of the valveflow area from closed to full-open.

The limiting single failure is the failure of one train of the reactor protection system. The remaining (operable) train trips the reactor.

As described in USAR Section 3.1.1, the MSSVs and PSVs (that is, codesafety valves) are considered to be qualified components exempt from active failure and are assumed toopen on demand. Control systems are assumed to function only if their operation results in more severetransient conditions.

Thus, a failure of a control system is not applicable as a limiting single failure.

FWisolation (redundant valves),

AFW (multiple pumps) and SI (multiple pumps) are susceptible to a singlefailure.

However, none of these systems provides any mitigation for a LOL/TT event. Thus, these systemsare not applicable as a limiting single failure.

Furthermore, the protection system is designed to besingle-failure-proof.

Maximum SGTP (10 percent) is assumed in the minimum DNBR case and peak RCS pressure casebecause it maximizes the RCS temperature increase following event initiation.

However, the peak MSSpressure case is analyzed with zero SGTP because this conservatively maximizes theprimary-to-secondary side heat transfer; this assumption is slightly more limiting with respect to thesecondary-side pressure transient.

Based on its frequency of occurrence, the LOL/TT accident is considered a Condition II event, an incidentof moderate frequency, as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

The specific criteria for this accident are asfollows:Pressures in the RCS and MSS are maintained below 110 percent of their respective design values(for the WCGS, this represents an RCS pressure limit of 2750 psia and MSS pressure limit of1318.5 psia).Fuel cladding integrity is maintained by demonstrating that the minimum DNBR remains abovethe 95/95 DNBR limit for PWRs (for the WCGS, the applicable safety analysis DNBR limitis 1.52).WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-115An incident of moderate frequency does not generate a more serious plant condition without otherfaults occurring independently.

This criterion is conservatively satisfied by verifying that thepressurizer does not fill.An incident of moderate frequency, in combination with any single active component failure orsingle operator error, is considered an event for which an estimate of the number of potential fuelfailures is provided for radiological dose calculations.

For such accidents, fuel failure is assumedfor all rods for which the DNBR decreases below those values cited above for cladding integrity unless it can be shown that, based on an acceptable fuel damage model, fewer failures occur.There shall be no loss of function of any fission product barrier other than the fuel cladding.

Thiscriterion is satisfied by verifying that the minimum DNBR remains above the 95/95 DNBR limit,which is discussed above.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the LOL/TT acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the LOL/TT event, this is shown to be met by demonstrating that the fuelcladding integrity is maintained.

GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the reactor coolant pressureboundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the LOL/TT event, this is shown to be met by demonstrating that thepeak RCS pressure is less than 110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires that one of thereactivity control systems consist of control rods capable of reliably controlling reactivity changeswith appropriate margin for malfunctions like stuck rods so that specified acceptable fuel designlimits are not exceeded under conditions of normal operation, including anticipated operational occurrences.

For the LOL/TT event, which results in a RT, this is shown to be met bydemonstrating that the fuel cladding integrity is maintained.

2.3.1.1.3 Description of Analyses and Evaluations A detailed analysis using the RETRAN computer code (Reference

3) was performed to determine theplant transient conditions following a total LOL due to turbine trip without credit for a direct RT. TheRETRAN model simulates the core neutron kinetics, RCS, pressurizer, pressurizer PORVs and sprays,PSVs, SGs, MSSVs, and the AFW system. The code computes pertinent plant variables, including RCSpressures and temperatures, and SG pressure.

The Westinghouse RETRAN model has been approved by the NRC for the analysis of the LOL/TTtransient (Reference 3).WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1162.3.1.1.4 ResultsThe calculated sequence of events for each of the cases is listed in Table 2.3.1-1, and the limiting resultsfor each of the cases are presented in Table 2.3.1-2.2.3.1.1.4.1 Minimum DNBR CaseThe minimum DNBR case was analyzed at the high nominal Tavg (i.e., 588.4°F),

nominal pressurizer pressure (i.e., 2250 psia), MMF, 10 percent SGTP, and the high main Tfeed (448.6°F) with automatic pressurizer pressure control operable.

Plots of the transient response to a LOL/TT event for the minimum DNBR case are shown inFigures 2.3.1-1 through 2.3.1-3.

The reactor was tripped on the OTAT RT function.

The nuclear powerremained essentially constant at full power prior to the RT. The pressurizer sprays, PORVs, and safetyvalves actuated to minimize the RCS pressure transient, which is conservative for the calculation of theminimum DNBR. Although the DNBR decreased below its initial value, it remained well above the SALthroughout the entire transient.

The peak pressurizer water volume remained below the total volume ofthe pressurizer, demonstrating that this event does not generate a more serious plant condition.

TheMSSVs actuated to maintain the MSS pressure below 110 percent of the design value.2.3.1.1.4.2 Peak MSS Pressure CaseThe peak MSS pressure case was analyzed at the high nominal Tavg plus uncertainties (i.e., 588.4'F +6.5°F), nominal pressurizer pressure minus uncertainties (i.e., 2250 psia -50 psi), TDF, 0 percent SGTPand the high main Tfeed (448.60F) with automatic pressurizer pressure control operable.

Plots of the transient response to a LOL/TT event for the peak MSS pressure case are shown inFigures 2.3.1-4 through 2.3.1-6.

The reactor was tripped on the OTAT RT function.

The nuclear powerremained essentially constant at full power prior to the RT. The pressurizer sprays, PORVs, and safetyvalves actuated to minimize the RCS pressure transient, which is conservative because it prevented a RTfrom occurring on high pressurizer pressure and exacerbated the peak MSS pressure.

The MSSVsactuated to maintain the MSS pressure below 110 percent of the design value. The peak pressurizer watervolume remained below the total volume of the pressurizer, demonstrating that this event does notgenerate a more serious plant condition.

2.3.1.1.4.3 Peak RCS Pressure CaseThe most limiting peak RCS pressure case was that analyzed at the high nominal Tavg minus uncertainties (i.e., 588.4°F -6.5°F), nominal pressurizer pressure minus uncertainties (i.e., 2250 psi -35 psi), TDF,10 percent SGTP and the high main Tfeed (448.6°F) with automatic pressurizer pressure controlinoperable.

Plots of the transient response to a LOL/TT event for the limiting peak RCS pressure case are shown inFigures 2.3.1-7 through 2.3.1-9.

The reactor was tripped on the high pressurizer pressure RT function.

Thenuclear power remained essentially constant at full power prior to the RT. The PSVs actuated to maintainthe RCS pressure below 110 percent of the design value. The MSSVs also actuated to maintain the MSSWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-117pressure below 110 percent of the design value. The peak pressurizer water volume remained below thetotal volume of the pressurizer, demonstrating that this event does not generate a more serious plantcondition.

2.3.1.2 Conclusions From a review of the updated analyses for the LOL/TT event, it is concluded that these analyses haveadequately accounted for operation of the plant at the analyzed power level and that they were performed using acceptable analytical models. The calculated results demonstrate that the reactor protection andsafety systems will continue to ensure that the safety analysis DNBR limit is met and the RCS and MSSpressure boundary limits will not be exceeded as a result of the LOL/TT event. Furthermore, this eventwill not generate a more serious plant condition.

Based on this, the WCGS will continue to meet therequirements of GDCs 10, 15, and 26.2.3.1.3 References

1. WCAP- I 1397-P-A, "Revised Thermal Design Procedure,"

April 1989.2. WCAP-129 10, Rev. I-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.3. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-118Table 2.3.1-1 Time Sequence of Events -Loss of External Electrical Load and/or Turbine TripCase Event Time (seconds)

Minimum DNBR Case Loss of Electrical Load/Turbine Trip Occurs 0.0Pressurizer PORVs Open 2.8PSVs Open 9.7MSSVs Open 9.7OTAT RT Setpoint Reached 17.0Minimum DNBR Occurs 19.8Rods Begin to Drop 20.0Peak MSS Pressure Case Loss of Electrical Load/Turbine Trip Occurs 0.0Pressurizer PORVs Open 1.7MSSVs Open 4.7PSVs Open 9.6OTAT RT Setpoint Reached 15.8Rods Begin to Drop 18.8Peak Secondary Side Pressure Occurs 22.0Peak RCS Pressure Case Loss of Electrical Load/Turbine Trip Occurs 0.0High Pressurizer Pressure RT Setpoint Reached 6.6Rods Begin to Drop 7.6PSVs Open 8.2Peak RCS Pressure Occurs 9.7MSSVs Open 12.1Table 2.3.1-2 Limiting Results -Loss of External Electrical Load and/or Turbine TripCase Parameter Analysis Value LimitMinimum DNBR Case Minimum DNBR 1.72 1.52Peak MSS Pressure Case Peak MSS Pressure (psia) 1297.9 1318.5Peak RCS Pressure Case Peak RCS Pressure (psia) 2746.8 2750.0WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-119I.0C0(.)L..a,300a,(-)=3CnCncnI.-0 20 40 60 80Time (sec)100Figure 2.3.1-1LOLUTT, Minimum DNBR CaseNuclear Power and SG Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-120WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-120CnIU,0-r4I-Cn(Ur4)0ncn0 20 40 60 80Time (sec)1000 20 40 60 80 100Time (sec)Figure 2.3.1-2LOL/TT, Minimum DNBR CasePressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3Hot Leg Temperature Vessel Average Temperature Cold Leg Temperature 700U-LJ....650 .. ..............

.. .. .. .. ... .. ..... .... .. ... ....-6 0 0 ............... ...................... .E550.......................

C/,2-1210 20 40 60 80Time (sec)1000 20 40 61Time (seconds80100Figure 2.3.1-3LOL/TT, Minimum DNBR CaseRCS Temperatures and DNBR versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-122WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-12200!<D)CUcn(UL..0 20 40 60 80 100Time (sec)a 20 40 60 80Time (sec)100Figure 2.3.1-4LOL/TT, Peak MSS Pressure CaseNuclear Power and SG Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-123WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-123~211.0-0 20 40 60 80 100Time (sec)0 20 40 60 80Time (sec)100Figure 2.3.1-5LOL/TT, Peak MSS Pressure CasePressurizer Pressure and Pressurizer Water Volume versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-124WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-124Hot Leg Temperature Vessel Average Temperature Cold Leg Temperature LL_Cfl=3CLE0 20 40 60 80Time (sec)100Figure 2.3.1-6LOL/TT, Peak MSS Pressure CaseRCS Temperatures versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-125WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-125004--0L.0~3ra_0~C,,C,,E0 20 40 60 80Time (seconds) 100Figure 2.3.1-7LOL/TT, Peak RCS Pressure CaseNuclear Power and SG Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-126WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-126Pressurizer PressureMaximum RCS PressureC,,C,,C,,V)C-,)=3E=3C,,Cfl0 20 40 60 80Time (sec)LOL/TT, Peak RCS Pressure CaseRCS Pressures and Pressurizer Water Volume versus Time100Figure 2.3.1-8WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-127WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-127Hot Leg Temperature Vessel Average Temperature Cold Leg Temperature rn. r600E 550-------- ------ -- --- ---d0-II II Ii I II II I I-.1/u020406080100Time (sec)Figure 2.3.1-9LOLITT, Peak RCS Pressure CaseRCS Temperatures versus TimeWCAP-17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1282.3.2 Loss of Non-Emergency AC Power to the Station Auxiliaries (USAR Section 15.2.6)2.3.2.1 Technical Evaluation 2.3.2.1.1 Introduction A complete loss of non-emergency alternating current (AC) power (LOAC) may result in a loss of powerto the plant auxiliaries, which include the RCPs, main FW pumps, condensate pumps, etc. The loss ofpower may be caused by a complete loss of the offsite grid accompanied by a turbine generator trip at thestation, or by a loss of the onsite AC distribution system.The LOAC event is analyzed as a LONF with a loss of power to the RCPs as a result of the RT becausethis is a more severe event relative to long-term consequences than the LOAC event. In the LOAC event,the RCPs lose power at the beginning of the event and the reactor trips soon thereafter on low reactorcoolant loop flow. The short-term consequences are bounded by those of the complete loss of reactorcoolant flow event described in Section 2.4.1, "Partial and Complete Loss of Forced Reactor CoolantFlow." The immediate consequence following a loss of FW is a reduction in the SG water level, which, ifleft unmitigated, will ultimately result in a RT and AFW system actuation on the low-low SG water levelsignal. Following RT, the rate of heat generation in the RCS (core residual (decay) heat) may exceed theheat removal capability of the secondary system. If this occurs, the RCS heats up, and the resulting thermal expansion of the reactor coolant causes an insurge to the pressurizer and an increase in thepressurizer water level. This trend generally continues until the RCS heat generation rate decreases belowthe secondary-side heat removal capability, at which time a cooldown of the RCS commences.

The LONFevent without a LOOP is addressed in Section 2.3.3, "Loss of Normal Feedwater Flow."The expected events following an LOAC with turbine and reactor trips are described in the sequencelisted as follows.Plant vital instruments are supplied by emergency direct current (DC) power sources.The SG ARVs are automatically opened to the atmosphere as the MSS pressure increases following the trip. The condenser is assumed to be unavailable for steam dump. If the steam flowrate through the ARVs is not sufficient or if the ARVs are not available, the MSSVs may lift todissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in thereactor.The SG ARVs (or safety valves, if the ARVs are not available) are used to dissipate the residualdecay heat and to maintain the plant at the Mode 3 (hot standby) condition as the no-loadtemperature is approached.

The diesel generators start on a loss of voltage to the plant engineered safety features busses andbegin to supply plant vital loads.The AFW system is automatically actuated.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-129The plant safety features that are available to mitigate the consequences of an LOAC event are as follows.* A RT can be initiated by one of the following.

-Two-out-of-four low-low water level signals in any one SG-Two-out-of-four high pressurizer pressure signals-Two-out-of-three high pressurizer level signals-Two-out-of-four OTAT signalsThe MSSVs open and provide secondary-side pressure protection and a heat sink source thathelps limit the RCS heatup.* The PSVs may open to provide primary-side pressure protection.

Backup FW for the SGs is provided by the AFW system, which is composed of two motor-driven AFW (MDAFW) pumps and one turbine-driven AFW (TDAFW) pump.-The two MDAFW pumps are started on any of the following:

  • Two-out-of-four low-low water level signals in any one SG* Trip of both main FW pumps* SI signal* LOOP* Manual pump start" Manual AFW system actuation

-The TDAFW pump is started on any of the following:

" Two-out-of-four low-low water level signals in each of two SGs* LOOP* Manual pump start* Manual AFW system actuation The MDAFW pumps are supplied power by the diesel generators, and the TDAFW pump utilizessteam from the secondary system. The pump turbine exhausts the secondary steam to theatmosphere.

Normally, the AFW pumps take suction from the condensate storage tank (CST), butif the CST is unavailable, the essential service water system is used as the water source for theAFW pumps.After power to the RCPs is lost, coolant flow necessary for core cooling and the removal of core decayheat is maintained by natural circulation in the RCS loops. Following the RCP coastdown, the naturalcirculation capability of the RCS will remove decay heat from the core, aided by the AFW flow in thesecondary system. Demonstrating acceptable analysis results for this event proves that the resultant natural circulation flow in the RCS and the AFW flow are sufficient for removing the decay heat from thecore.WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1302.3.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The major inputs and assumptions applied in the analysis of the LOAC event are identical to thoseapplied in the analysis of the LONF event described in Section 2.3.3, "Loss of Normal Feedwater Flow,"with the following exceptions:

The initial RCP heat is the nominal value of 14 MWt. Nominal RCP heat is conservative for theLOAC event because the initial core power is slightly higher compared to that associated withmaximum RCP heat, and this translates into slightly higher core decay heat, which is the primaryheat source of concern for this event; after coastdown, the RCPs cease to add heat to the primarycoolant, and so it is conservative to maximize the core decay heat.The loss of power to the RCPs is assumed to be the result of an electrical disturbance on theoffsite power grid caused by the RT. The RCPs were assumed to lose power and begin coastingdown 2 seconds after the start of rod motion. This time delay is considered to be reasonable, but itis not a critical parameter in the analysis because it is short relative to the overall transient time.Acceptance CriteriaBased on the expected frequency of occurrence, the LOAC event is considered to be a Condition II eventas defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water ReactorPlants,"

ANSI N 18.2-1973.

The following items summarize the acceptance criteria associated with theanalysis of this event:Pressures in the RCS and MSS must remain less than 110 percent of the respective designpressures.

With respect to peak RCS and MSS pressures, the LOAC event is bounded by the LOL/TT eventdescribed in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip, Inadvertent Closureof a Main Steam Isolation Valve, and Loss of Condenser Vacuum,"

in which assumptions aremade to conservatively calculate the RCS and MSS pressure transients.

For the LOAC event,turbine trip occurs after RT, whereas for LOL/TT, the turbine trip is the initiating incident.

Therefore, the power mismatch between the primary and secondary sides and the resultant temperature and pressure transients of the RCS and MSS are always more severe for LOL/TTthan for LOAC. Based on this, no explicit calculation of maximum RCS or MSS pressure isperformed for this event.Fuel cladding integrity must be maintained by ensuring that the minimum DNBR remains abovethe 95/95 DNBR limit.With respect to the DNBR, the LOAC event is bounded by the complete loss of reactor coolantflow event described in Section 2.4.1, "Partial and Complete Loss of Forced Reactor CoolantFlow." Whereas the LOAC event has RCP coastdown (reactor coolant flow reduction) occurring after rod motion, the complete loss of reactor coolant flow event begins with a coastdown of allWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-131the RCPs, and RT occurs after the core coolant flow has already degraded.

As the limiting ratio ofthe core power to core flow is greater for the complete loss of reactor coolant flow event, it ismore limiting with respect to the DNBR. Based on this, no explicit calculation of minimumDNBR is performed for this event.An incident of moderate frequency must not generate a more serious plant condition withoutother faults occurring independently.

This criterion is conservatively demonstrated to be met if the pressurizer does not becomewater-solid.

The concern with filling the pressurizer water-solid is that it could lead to the failingopen of one or more PSVs, which would provide an unisolable path for the loss of reactorcoolant, and a LOCA is a more serious plant condition.

Satisfying this criterion demonstrates thepreclusion of a more serious plant condition, ensures that the RCS and MSS pressure criteria andminimum DNBR criterion are satisfied for the long-term portion of the event, and confirms theAFW system is adequate for long-term heat removal.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the LOAC acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the LOAC event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.

GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.

For theLOAC event, this is shown to be met by demonstrating that the peak RCS pressure is less than110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.

For the LOAC event, whichresults in a RT, this is shown to be met by demonstrating that the fuel cladding integrity ismaintained with a trip reactivity that accounts for the most reactive rod stuck out of the core.2.3.2.1.3 Description of Analyses and Evaluations A detailed analysis using the RETRAN computer code (Reference

1) was performed to determine theplant transient conditions for the LOAC event. A RETRAN input model specific to the WCGS wasdeveloped to simulate the core neutron kinetics, RCS, pressurizer, pressurizer
heaters, pressurizer sprays,SGs, MSSVs, and the AFW system. Several LOAC cases were modeled for various combinations ofinitial conditions and pressurizer PORV availability, and the RETRAN code computed the time-dependent WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-132trends of pertinent variables, including the pressurizer

pressure, pressurizer water volume, SG mass, andreactor coolant temperatures.

2.3.2.1.4 ResultsThe most limiting LOAC case was with an initial Taý,g of 564.2°F (low end of the full-power Tavgwindow (570.7°F) minus uncertainties),

an initial pressurizer pressure of 2300 psia (nominal (2250 psia)plus uncertainties),

an initial main Tfeed of 400'F (low full-power value), maximum (10 percent)

SGTP,and the pressurizer PORVs not available.

The calculated sequence of events for the limiting LOAC case is presented in Table 2.3.2-1, and transient plots of the significant plant parameters are provided in Figures 2.3.2-1 through 2.3.2-10.

Following theloss of FW from full power, the SG water level decreases to the low-low setpoint at 37.7 seconds, whichactuates a RT and the AFW system. The lack of FW causes the RCS temperature to increase.

Rod motionand turbine trip are initiated at 39.7 seconds and the RCPs begin coasting down at 41.7 seconds.

Althougha temporary cooldown of the RCS occurs as a result of the RT, the RCS heats up rapidly in response to thecontinued lack of FW and also the turbine trip. The MSSVs open at 73.2 seconds to help dissipate thestored and generated heat, and at 97.7 seconds, one minute after being actuated, the AFW system beginsto deliver 220 gpm of AFW flow to each SG. The pressurizer water volume reaches a maximum value of1623.2 ft3 at 2953.5 seconds after event initiation.

As the maximum pressurizer water volume value is lessthan the total pressurizer volume of 1800 ft3, it is confirmed that the pressurizer does not reach awater-solid condition.

2.3.2.2 Conclusions Based on the above information, it is concluded that the LOAC event will not progress into a moreserious plant condition.

Thus, all applicable event acceptance criteria are satisfied, and the AFW systemwith natural circulation reactor coolant flow are confirmed to be adequate for long-term heat removalfollowing an LOAC event. Therefore, it has been demonstrated that the reactor protection and safetysystems ensure that the acceptable fuel design limits are met, and the RCS and MSS pressure limits willnot be exceeded as a result of an LOAC event. Based on this, the plant continues to meet the requirements of GDCs 10, 15, and 26.2.3.2.3 References

1. WCAP- 14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-133Table 2.3.2-1 Time Sequence of Events for Limiting LOAC CaseEvent Time (seconds)

Main FW Flow Stops 0.0Low-Low SG Water Level RT Setpoint Reached 37.7Rods Begin to Drop and Turbine Trip Initiated 39.7RCPs Begin Coasting Down 41.7On Each Loop, the MSSV with the Lowest Setting Opens 73.2Flow from Two MDAFW Pumps Initiated 97.7SG Inventory Reduction Reverses 221.5Maximum Pressurizer Water Volume Occurs 2953.5WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-134WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1341 )~1*ci)00~C)(-3ID0.80.60.40.2-I I I I , -I I I I I I I I11 I I ( I 1 [II I- -I---- ~ I -i I I I010110210ime (seconds) 310410Figure 2.3.2-1 LOAC -Nuclear Power versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-135WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1351.2-1QC)CTC)C)0.6-0.4--- 0.2 -0I I I I I I IT~~TI~--

I I2010110210ime (seconds) 310410Figure 2.3.2-2 LOAC -Core Average Heat Flux versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-136WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1361 -LbI _ I I i I I III I I I I IFigure 2.3.2-3 LOAC -Reactor Coolant Loop Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-137WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-137620~17D600tCn58-S560-_C)0540-520/01010z10mime (seconds) 310410Figure 2.3.2-4 LOAC -HL and CL Temperatures versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-138WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1382600-IC)u~)C)enenC)CLC)inU;enC)CL-C)C)2500-2400-2J00-2200-2100-kI I III I I~. I I Ill2000I I I I I 1 i I I I I t I I t I I II i I t201010210Time (seconds) 310410Figure 2.3.2-5 LOAC -Actual Pressurizer Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-139WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1391~f~fl *1I UUU01600-1400-1200-1000-I I f I I I I I I I I I I I I I .I I I I I800I I I I I I I 1 t t I t IT i I I2010110210Time (seconds) 310410Figure 2.3.2-6 LOAC -Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-140WESTiNGHOUSE NON-PROPRIETARY CLASS 3 2-1401300C)cn£012001100-1000900-800I I700'2010110210Time (seconds)

'310410Figure 2.3.2-7 LOAC -SG Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-141WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-14 1C/)0C)DEC)7D.0T10080-60-40-20-~1I I I I I I I IU010I10210Time (seconds) 310410Figure 2.3.2-8 LOAC -Indicated SG Level versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-142WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1422000000C:)150000-100000-50000 -0I i I I I I III I I I I I I II I I I I I I) I I I I I .I I I I I I I I I I I I I I I I201010210Firme (seconds) 10410Figure 2.3.2-9 LOAC -SG Mass versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-14325(^-IJc7- 2000(IDC)-C)C)C)x0.00-O150-100-50-0--50I i I I I I i I 1 i f I , I I I I I I t i I I I2010110210Time (seconds) 310410Figure 2.3.2-10 LOAC -Loop AFW Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1442.3.3 Loss of Normal Feedwater Flow (USAR Section 15.2.7)2.3.3.1 Technical Evaluation 2.3.3.1.1 Introduction A LONF flow (from pump failures, valve malfunctions, or a complete LOAC power) results in areduction in the capability of the secondary system to remove the heat generated in the reactor core. If analternative supply of FW is not provided, core residual (decay) heat following RT would heat the primarysystem water to the point where water relief from the pressurizer could occur, resulting in a substantial loss of water from the RCS.The expected events following an LONF (caused by either pump failures or valve malfunctions) withturbine and reactor trips are described in the sequence listed as follows:The SG ARVs are automatically opened to the atmosphere as the MSS pressure increases following the trip. The condenser is assumed to be unavailable for steam dump. If the steam flowrate through the ARVs is not sufficient or if the ARVs are not available, the MSSVs may lift todissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in thereactor.The SG ARVs (or safety valves, if the ARVs are not available) are used to dissipate the residualdecay heat and to maintain the plant at the Mode 3 (hot standby) condition as the no-loadtemperature is approached.

  • The AFW system is actuated automatically.

The plant safety features that are available to mitigate the consequences of an LONF event are as follows.* A RT can be initiated by one of the following.

-Two-out-of-four low-low water level signals in any one SG-Two-out-of-four high pressurizer pressure signals-Two-out-of-three high pressurizer level signals-Two-out-of-four OTAT signalsThe MSSVs open and provide secondary-side pressure protection and a heat sink source thathelps limit the RCS heatup.* The PSVs may open to provide primary-side pressure protection.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-145Backup FW for the SGs is provided by the AFW system, which is composed of two MDAFWpumps and one TDAFW pump.The two MDAFW pumps are started on any of the following:

-Two-out-of-four low-low water level signals in any one SG-Trip of both main FW pumps-SI signal-LOOP-Manual pump start-Manual AFW system actuation The TDAFW pump is started on any of the following:

-Two-out-of-four low-low water level signals in each of two SGs-LOOP-Manual pump start-Manual AFW system actuation The MDAFW pumps are supplied power by offsite power sources, and the TDAFW pumputilizes steam from the secondary system. The pump turbine exhausts the secondary steam to theatmosphere.

Normally, the AFW pumps take suction from the CST, but if the CST is unavailable, the essential service water system is used as the water source for the AFW pumps.The analysis of the LONF event demonstrates that the AFW system is capable of removing the stored andresidual heat, and consequently ensures the core will remain covered with water, and the RCS and MSSwill not overpressurize.

With this, the plant is shown to be able to return to a safe condition following aLONF event.2.3.3.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The following inputs and assumptions were applied in the analysis of the LONF event:An initial NSSS power of 3651 MWt, which includes all applicable uncertainties The initial RCP heat is the maximum value of 20 MWt. Maximum RCP heat is conservative forthe LONF event because the RCPs operate continuously throughout the transient.

The constantheat generated by the RCPs, in combination with the core decay heat, are the primary-side heatsources that provide the challenge to the long-term cooling (LTC) capability of the plant.Two initial full-power main Tfeed:-400.0°F (low)-448.60F (high)WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-146Four initial full-power Tavg values that cover the full range of the full-power Tayg window(570.7°F to 588.4°F) including uncertainty

(+/-6.5°F):

-594.90F-581.90F-577.20F-564.20F(high Tavg plus uncertainty)

(high Tvg minus uncertainty)

(low Tavg plus uncertainty)

(low Ta,,g minus uncertainty)

Two initial pressurizer pressure values that cover the +/-50 psi uncertainty associated with thenominal operating value of 2250 psia:-2300 psia-2200 psia(nominal plus uncertainty)

(nominal minus uncertainty)

Two initial pressurizer water level values, which are dependent on the full-power Ta,,g value, thatcover the +7 percent span uncertainty associated with the nominal values of 59 percent span forhigh Tavg cases and 41 percent span for low Tavg cases:66 percent span48 percent span(high nominal plus uncertainty (high Tavg cases))(low nominal plus uncertainty (low Tavg cases))SGTP levels of 0 and 10 percentA minimum low-low SG water level setpoint of 0 percent NRS for RT and AFW system actuation A maximum delay for RT (rod motion) of 2 secondsA maximum delay for AFW flow initiation of 60 secondsA minimum total AFW flow of 880 gpm split evenly between the four loopsThis flow corresponds to having both MDAFW pumps available for event mitigation.

As it is theworst single active failure for this analysis, the TDAFW pump was assumed to fail.A maximum AFW enthalpy of 96 Btu/lbm, which corresponds to a temperature of 125°F.The pressurizer proportional and backup heaters were modeled to maximize the heatup andthermal expansion of the water within the pressurizer.

In addition, the pressurizer sprays wereassumed to be operable, and cases were analyzed with and without the pressurizer PORVsavailable.

Secondary system steam relief is achieved through the self-actuated MSSVs. Note that steamrelief would normally be provided by the SG ARVs or condenser dump valves, but these wereconservatively assumed to be unavailable.

WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-147The MSSVs were modeled with opening setpoints that account for a maximum setpoint tolerance of 3 percent and appropriate line losses. Valve accumulation was modeled via a 5 psi ramp of thevalve open area from closed to full-open.

The reactivity feedback parameters were chosen to maximize the heatup of the RCS. Thisincluded modeling a least-negative MTC, a least-negative DTC, and a most-negative Doppler-only power coefficient.

Note that the applied MTC value, 0 pcm/0F, is the least-negative limit value for full power conditions; the application of a zero MTC at full power conditions isbounding compared to the application of a positive MTC at part power conditions.

Core residual/decay heat generation was based on the 1979 version of ANS 5.1 (Reference 1).ANSI/ANS-5.1-1979 is a conservative representation of the decay energy release rates.Long-term operation at the initial power level preceding the trip was assumed.Acceptance CriteriaBased on the expected frequency of occurrence, the LONF event is considered to be a Condition II eventas defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water ReactorPlants,"

ANSI N 18.2-1973.

The following items summarize the acceptance criteria associated with theanalysis of this event:Pressures in the RCS and MSS must remain less than 110 percent of the respective designpressures.

With respect to peak RCS and MSS pressures, the LONF event is bounded by the LOL/TT eventdescribed in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip, Inadvertent Closureof a Main Steam Isolation Valve, and Loss of Condenser Vacuum,"

in which assumptions aremade to conservatively calculate the RCS and MSS pressure transients.

For the LONF event,turbine trip occurs after RT, whereas for LOL/TT, is the initiating incident.

Therefore, the powermismatch between the primary and secondary sides and the resultant temperature and pressuretransients of the RCS and MSS are always more severe for LOL/TT than for LONF. Based onthis, no explicit calculation of maximum RCS or MSS pressure is performed for this event.Fuel cladding integrity must be maintained by ensuring that the minimum DNBR remains abovethe 95/95 DNBR limit.With respect to the DNBR, the LONF event is bounded by the LOL/TT event described inSection 2.3.1, "Loss of External Electrical Load, Turbine Trip, Inadvertent Closure of a MainSteam Isolation Valve, and Loss of Condenser Vacuum."

Each of these two events represents areduction in the heat removal capability of the secondary system. For the LONF event, the RCStemperature increases gradually as the SGs boil down to the low-low water level trip setpoint, atwhich time RT occurs, followed by turbine trip. For the LOL/TT event, the turbine trip is theinitiating event, and the loss of heat sink is much more severe. As such, the initial RCS heatupwill be much more severe for the LOL/TT event than for the LONF event, and the LOL/TT eventwill always be more severe with respect to the minimum DNBR criterion.

Based on this, noexplicit calculation of minimum DNBR is performed for this event.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-148An incident of moderate frequency must not generate a more serious plant condition withoutother faults occurring independently.

This criterion is conservatively demonstrated to be met if the pressurizer does not becomewater-solid.

The concern with filling the pressurizer water-solid is that it could lead to the failingopen of one or more PSVs, which would provide an unisolable path for the loss of reactorcoolant, and a loss of coolant accident is a more serious plant condition.

Satisfying this criterion demonstrates the preclusion of a more serious plant condition, ensures that the RCS and MSSpressure criteria and minimum DNBR criterion are satisfied for the long-term portion of theevent, and confirms the AFW system is adequate for long-term heat removal.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the LONF acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the LONF event, this is shown to be met by demonstrating that the fuel claddingintegrity is maintained.

GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.

For theLONF event, this is shown to be met by demonstrating that the peak RCS pressure is less than110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.

For the LONF event, whichresults in a RT, this is shown to be met by demonstrating that the fuel cladding integrity ismaintained with a trip reactivity that accounts for the most reactive rod stuck out of the core.2.3.3.1.3 Description of Analyses and Evaluations A detailed analysis using the RETRAN computer code (Reference

2) was performed to determine theplant transient conditions for the LONF event. A RETRAN input model specific to the WCGS wasdeveloped to simulate the core neutron kinetics, RCS, pressurizer, pressurizer
heaters, pressurizer sprays,SGs, MSSVs, and the AFW system. Several LONF cases were modeled for various combinations ofinitial conditions and pressurizer PORV availability, and the RETRAN code computed the time-dependent trends of pertinent variables, including the pressurizer
pressure, pressurizer water volume, SG mass, andreactor coolant temperatures.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1492.3.3.1.4 ResultsThe most limiting LONF case was with an initial Ta,,g of 564.2°F (low end of the full-power Tag window(570.7°F) minus uncertainties),

an initial pressurizer pressure of 2300 psia (nominal (2250 psia) plusuncertainties),

an initial main Tfeed temperature of 400'F (low full-power value), and minimum (0 percent)SGTP. Although the pressurizer PORVs were modeled as being available in this limiting case, thepressurizer sprays were sufficient in controlling the pressurizer pressure below the setpoint of the PORVs.The calculated sequence of events for the limiting LONF case is presented in Table 2.3.3-1, and transient plots of the significant plant parameters are provided in Figures 2.3.3-1 through 2.3.3-10.

Following theloss of FW from full power, the SG water level decreases to the low-low setpoint at 37.8 seconds, whichactuates a RT and the AFW system. The lack of FW causes the RCS temperature to increase.

Rod motionand turbine trip are initiated at 39.8 seconds and the RCPs continue running.

Although a temporary cooldown of the RCS occurs as a result of the RT, the RCS heats up rapidly in response to the continued lack of FW and also the turbine trip. The MSSVs open at 68.0 seconds to help dissipate the stored andgenerated heat, and at 97.8 seconds, one minute after being actuated, the AFW system begins to deliver220 gpm of AFW flow to each SG. The RCS heatup turns around shortly after the MSSVs open, and it isfurther controlled by the cooling effect of the AFW flow. The pressurizer water volume reaches amaximum value of 1384.1 ft3 at 1372.0 seconds after event initiation.

As the maximum pressurizer watervolume value is less than the total pressurizer volume of 1800 ft3, it is confirmed that the pressurizer doesnot reach a water-solid condition.

2.3.3.2 Conclusions Based on the above information, it is concluded that the LONF event will not progress into a more seriousplant condition.

Thus, all applicable event acceptance criteria are satisfied, and the AFW system isconfirmed to be adequate for long-term heat removal following an LONF event. Therefore, it has beendemonstrated that the reactor protection and safety systems ensure that the acceptable fuel design limitsare met, and the RCS and MSS pressure limits will not be exceeded as a result of an LONF event. Basedon this, the plant continues to meet the requirements of GDCs 10, 15 and 26.2.3.3.3 References

1. ANSI/ANS-5.1

-1979, "American National Standard for Decay Heat Power in Light WaterReactors,"

August 1979.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-150Table 2.3.3-1 Time Sequence of Events for Limiting LONF CaseEvent Time (seconds)

Main FW Flow Stops 0.0Low-Low SG Water Level RT Setpoint Reached 37.8Rods Begin to Drop and Turbine Trip Initiated 39.8On Each Loop, the MSSV with the Lowest Setting Opens 68.0Flow from Two MDAFW Pumps Initiated 97.8SG Inventory Reduction Reverses 431.0Maximum Pressurizer Water Volume Occurs 1372.0WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-151WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-15 11^-IC)0U)I.L1-0.80.6-0.4-0.2Ii ! i I II I i tU010110210Time (seconds) 310410Figure 2.3.3-1 LONF -Nuclear Power versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-152WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1521.2-vC)C)C-)(9.1-0.80.60.4-0.2-I I I I I IIII I I I I I I I I11 I i I I I ---- I- I ---- t-f, I II I0010110210Time (seconds) 310410Figure 2.3.3-2 LONF -Core Average Heat Flux versus TimeWCAP-1 7658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-153I.()wC-)C)C)-JCDC)C)C---)(~ .)C--)CD(1)Liii0.6-III I II II I I I I ISre Le-ont cFigure 2.3.3-3 LONF -Reactor Coolant Loop Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-154WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1546?n~ ,KL/H-QThC)-C)600-GD580-EGD5600(-3540-0I I I I I I I I010110210Time (seconds) 310410Figure 2.3.3-4 LONF -HL and CL Temperatures versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-155WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-15526002500-~1C)IDcnC)C)IDUDC)ID(92400-2300-2200-2100--- --~-- --I liii2000I I I I I I I I [ 1 1I I I T I I I I I I I I2010110210Time (seconds) 310410Figure 2.3.3-5 LONF -Actual Pressurizer Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1561800C)C)C)C)0C)C)C)C)C)C)C)a-160014001200-1000-I I I I I I I II I , 1 1 , ,I I I I I I I I I I I I I I I I;~flf\ -~I I [I I I III I I I I I II2LUUu010110210Time (seconds) 310410Figure 2.3.3-6 LONF -Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-157WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1571300~1.EDCfuCn7-)U-)12001100-1000900-800I I I I I I I II III I II I I II I I I I I I i I1 1 I I I I I700010110210Time (seconds) 1510410Figure 2.3.3-7 LONF -SG Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-158WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-158(f)C)C)0C)C)C)C)C)C)cn~C)C)C)C-)-C)C)10080-60-40-20-fI X~ 1111 I 1111 I I III0t I I I I I II I I I I I I I II I I I II I I I I I I I I I I I201010210Time (seconds) 310410Figure 2.3.3-8 LONF -Indicated SG Level versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-159WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-159200000~1*EC-)0c-a-a-F150000-100000-50000-00101 210 10Time (seconds)

Figure 2.3.3-9 LONF -SG Mass versus Time310410WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-160WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-160250cm 200-0C)0~0-1150-100-50-0--50-4 .I I I I I , I I t I ...010110210Time (seconds) 310410Figure 2.3.3-10 LONF -Loop AFW Flow versus TimeWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1612.3.4 Feedwater System Pipe Break (USAR Section 15.2.8)2.3.4.1 Technical Evaluation 2.3.4.1.1 Introduction A major feed line break is defined as a break in a FW line large enough to prevent the addition ofsufficient FW to maintain shell-side fluid inventory in the SGs. If the break is postulated in a feed linebetween the check valve and the SG, fluid from the SG will be discharged through the break.Furthermore, depending upon the arrangement of the AFW piping, a break in this location could precludethe subsequent addition of AFW to the affected SG. A break upstream of the feed line check valve wouldaffect the RCS only as a LONF accident.

This event is addressed by the LONF analysis presented inSection 2.3.3, "Loss of Normal Feedwater Flow."Depending upon the size of the break and the plant operating conditions at the time of the rupture, thebreak could cause either an RCS cooldown or heatup. The potential RCS cooldown resulting from asecondary pipe rupture is evaluated in the SLB analysis presented in Section 2.2.5.1, "Steam SystemPiping Failure at HZP." Therefore, only the RCS heatup effects are evaluated for a feed line break.A feed line break reduces the ability of the secondary system to remove heat generated by the core fromthe RCS for the following reasons:Reduction in FW flow to the SGs. The degradation in FW flow can cause the reactor coolanttemperature to increase prior to RT.Fluid inventory of the faulted SG may be discharged through the break, and therefore, would notbe available for decay heat removal following RT.The AFW system is provided to ensure that adequate FW is available to provide decay heat removal.Thus, the primary function of the feed line break analysis is to verify that the capacity of the AFW systemis adequate.

The AFW system is intended to provide an adequate supply of FW to ensure that:No substantial overpressurization of the RCS and MSS occurs , andSufficient liquid is maintained in the RCS so that the core remains in place and geometrically intact with no loss of core cooling capability.

The most limiting single failure in this event is the loss of one AFW train that results in the loss of oneAFW pump, thus reducing the heat removal capability of the AFW system. The AFW flow rate modeledin the analysis bounds the consequences from the loss of either a MDAFW pump or a TDAFW pump.The feed line break event is analyzed at full power conditions that bound all other power levels and allother Modes, because decay heat and stored energy are most limiting following a trip from a full powercondition.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-162The severity of the feed line rupture transient depends on a number of parameters, including break size,initial reactor power, and the functioning of various control and safety systems.

Sensitivity studiespresented in Reference 1 illustrate many of the limiting assumptions for the feed line rupture.In the analysis, the main FW control system is assumed to fail due to an adverse environment.

As a result,the water levels in all the SGs decrease equally until the low-low SG water level RT setpoint is reached.After RT, a double-ended rupture of the largest FW line is modeled.

This combination of eventsconservatively bounds the most limiting feed line break scenario that can occur. Analyses have beenperformed at full power, with and without LOOP, with credit taken for the pressurizer PORVs, but no SIactuation modeled.

For the case without offsite power available, the power is assumed to be lost at thetime of RT. This is more conservative than the case in which power is lost at the initiation of the event.The plant safety features that are available to mitigate the consequences of a feed line break event are asfollows.A RT can be initiated by one of the following.

-Two-out-of-four low-low water level signals in any one SG-Two-out-of-four high pressurizer pressure signals-Two-out-of-three high pressurizer level signals-Two-out-of-four OTAT signals-SI signal (from one of the following

[1] two-out-of-three low steamline pressure signals inany one loop or [2] two-out-of-three high containment pressure signals)-Two-out-of-four low pressurizer pressure signalsThe MSSVs open and provide secondary-side pressure protection and a heat sink source thathelps limit the RCS heatup.The PSVs may open to provide primary-side pressure protection.

Backup FW for the SGs is provided by the AFW system, which is composed of two MDAFWpumps and one TDAFW pump.The two MDAFW pumps are started on any of the following:

-Two-out-of-four low-low water level signals in any one SG-Trip of both main FW pumps-SI signal-LOOP-Manual pump start-Manual AFW system actuation WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-163The TDAFW pump is started on any of the following:

-Two-out-of-four low-low water level signals in each of two SGs-LOOP-Manual pump start-Manual AFW system actuation The MDAFW pumps are supplied power by offsite power sources, and the TDAFW pumputilizes steam from the secondary system. The pump turbine exhausts the secondary steam to theatmosphere.

Normally, the AFW pumps take suction from the CST, but if the CST is unavailable, the essential service water system is used as the water source for the AFW pumps.The analysis of the feed line break event demonstrates that the AFW system is capable of removing thestored and residual heat, and consequently ensures the core will remain covered with water. With this, theplant is capable of returning to a safe condition following a feed line break event.2.3.4.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The following inputs and assumptions were applied in the analysis of the feed line break event:* An initial NSSS power of 3651 MWt, which includes all applicable uncertainties.

The initial RCP heat is the maximum value of 20 MWt for the case with offsite power available throughout the event. With offsite power available, maximum RCP heat is conservative for thefeed line break event because the RCPs operate continuously throughout the transient.

For thecase with a LOOP, the nominal value of 14 MWt for the initial RCP heat is used because it ismore conservative to model a slightly initial higher core power that increases the subsequent post-trip decay heat. Any post-trip heat generated by the RCPs, in combination with the coredecay heat, are the primary-side heat sources that provide the challenge to the long-term coolingcapability of the plant.A maximum initial full-power main Tfeed of 448.6°F.A maximum initial full-power Tag value of 594.90F, which represents the nominal high T,,vg plusthe uncertainty of 6.5°F.Main FW is assumed to be lost to all SGs at event initiation due to the feed line break. Thereverse blowdown of the faulted SG is conservatively delayed and begins when the SG inventory reaches 0 percent NRS. The combination of conditions modeled is defined to produce the mostsevere feed line break transient with the control and protection interaction considered.

The worst possible break area is modeled to maximize the blowdown discharge rate following thetime of RT, which maximizes the resultant heatup of the reactor coolant.

Choked flow is modeledat the break.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-164Operation of the pressurizer PORVs is modeled to minimize RCS pressure, which produces lower(more limiting) saturation temperatures within the RCS. To ensure the conservatism of thisassumption, SI flow, which would increase with reduced RCS pressure, is set to zero. Pressurizer spray and heaters are both assumed to be inoperable.

A minimum initial pressurizer pressure value of 2200 psia, which represents the nominal pressureof 2250 psia minus the uncertainty of 50 psi. The use of a low initial RCS pressure is consistent with modeling the operation of the pressurizer PORVs.A maximum initial pressurizer water level of 66 percent span, which conservatively bounds thehigh Tavg nominal value plus the uncertainty of 7 percent span. The initial water level in all SGs isset at 60 percent span, which is the nominal value (50 percent span) plus a level uncertainty of10 percent span.SGTP level of 10 percent (a maximum value).A minimum low-low SG water level setpoint of 0 percent NRS for RT and AFW systemactuation.

A maximum delay for RT (rod motion) of 2 seconds.A maximum delay for AFW flow initiation of 60 seconds.

The analysis conservatively accountsfor the purging of the hotter main FW in the FW piping, which delays delivery of the relatively cold auxiliary FW flow to the SGs.Failure of one protection train is taken as the worst single failure and results in one AFW pumpbeing inoperable.

The total AFW flow modeled is 594.4 gpm delivered to the three intact SGs.This is a conservative minimum value for the AFW flow following a feed line break and boundseither a single failure of the TDAFW pump or one MDAFW pump. The distribution of AFWflow is 222.7 gpm to each of two intact SGs with the third intact SG receiving 149 gpm. No flowis modeled as being delivered to the SG in the faulted loop.A maximum AFW enthalpy of 96 Btu/lbm, which corresponds to a conservatively hightemperature of 1257F.Secondary system steam relief is achieved through the self-actuated MSSVs. Note that steamrelief would normally be provided by the SG ARVs or condenser dump valves, but these wereconservatively assumed to be unavailable.

The MSSVs were modeled with opening setpoints that account for a maximum setpoint tolerance of 3 percent and appropriate line losses. Valve accumulation was modeled via a 5 psi ramp of thevalve open area from closed to full-open.

Both minimum and maximum reactivity feedback conditions have been considered.

Consistent with the limiting conditions determined for each scenario, the case analyzed with offsite poweravailable throughout the transient models maximum reactivity feedback while the case withoutWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-165offsite power (i.e., with a LOOP) models minimum reactivity feedback.

Note that the minimumreactivity feedback input includes a least negative moderator MTC value of 0 pcm/0F.Core residual/decay heat generation was based on the 1979 version of ANS 5.1 (Reference 2).ANSI/ANS-5.1-1979 is a conservative representation of the decay energy release rates.Long-term operation at the initial power level preceding the trip was assumed.No credit is taken for heat energy deposition in the RCS metal during the RCS heatup phase ofthe transient.

  • No credit is taken for charging or letdown.No credit is taken for the following potential protection logic signals to mitigate the consequences of the accident:

-High pressurizer pressure-OTAT-High pressurizer level-High containment pressureAcceptance CriteriaBased on the expected frequency of occurrence, the feed line break event is considered to be aCondition IV event as defined by "Nuclear Safety Criteria for the Design of Stationary Pressurized WaterReactor Plants,"

ANSI N 18.2-1973.

The following items summarize the acceptance criteria associated with the analysis of this event:* Pressures in the RCS and MSS must remain less than 110 percent of the respective designpressures.

With respect to peak RCS and MSS pressures, the feed line break event is bounded by theLOL/TT event described in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip,Inadvertent Closure of a Main Steam Isolation Valve, and Loss of Condenser Vacuum,"

in whichassumptions are made to conservatively calculate the RCS and MSS pressure transients.

For thefeed line break event, turbine trip occurs after RT, whereas for the LOL/TT event, the turbine tripis the initiating fault. Therefore, the primary to secondary power mismatch and resultant RCS andMSS heatup and pressurization transients are always more severe for the LOL/TT event. Basedon this, no explicit calculation of maximum RCS or MSS pressure is performed for this event.* Any fuel damage calculated to occur must be sufficiently limited to the extent that the core willremain in place and intact with no loss of core cooling capability.

With respect to fuel damage due to "dryout,"

where the water level in the vessel drops below thetop of the core, Westinghouse has established an internal criterion that no bulk boiling occurs inthe primary coolant system prior to event turn around. Turn around occurs when the heat removalcapability of the SGs, being fed AFW, exceeds NSSS heat generation.

This conservatively ensuresWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-166that the core remains covered with water and thereby will remain in place and geometrically intact with no loss of core cooling capability.

This criterion is very limiting and is adopted forconvenience in interpreting the results of this study. Compliance can be determined by checkingthe temperature plots (HL saturation, HL, and CL) for all the loops and verifying that neither theHL nor CL temperatures exceed the saturation temperature prior to turn around. It should benoted that precluding bulk boiling in the RCS is the limiting criterion that is considered in thefeed line break analysis.

With respect to possible fuel damage due to DNB, the feed line breakevent would be bounded by either the SLB analysis presented in Section 2.2.5.1 or the LOL/TTanalysis in Section 2.3.1.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the feed line break acceptance criteria are provided as follows.GDC 27 (Combined Reactivity Control Systems Capability) requires that the reactivity controlsystems be designed to have a combined capability, in conjunction with poison addition by theECCS, of reliably controlling reactivity changes so that, under postulated accident conditions andwith appropriate margin for stuck rods, the capability to cool the core is maintained.

For the feedline break event, this is shown to be met by demonstrating that the applicable fuel damagecriterion is satisfied.

It should be noted that in the feed line break analysis cases presented, poisonaddition via SI actuation is conservatively not credited but is available and would be actuated.

GDC 28 (Reactivity Limits) requires that the reactivity control system be designed withappropriate limits on the potential amount and rate of reactivity increase so that the effects ofpostulated reactivity accidents can neither result in damage to the RCPB greater than limited localyielding, nor sufficiently disturb the core, its support structures, or other RPV internals to impairsignificantly the capability to cool the core. For the feed line break event, this is shown to be metby demonstrating that the peak RCS pressure is less than 110 percent of the design pressure, which ultimately ensures that the RCPB pressure limits are not exceeded, and by confirming thatthe fuel damage requirements are met.GDC 35 (Emergency Core Cooling) requires that the RCS and associated auxiliaries be designedwith a safety system able to provide abundant emergency core cooling.

For the feed line breakevent, this is shown to be met by demonstrating that the fuel damage criterion is met, whichconfirms that the AFW system provides abundant cooling for the RCS, even with themost-limiting single failure considered.

2.3.4.1.3 Description of Analyses and Evaluations A detailed analysis using the RETRAN computer code (Reference

3) was performed to determine theplant transient conditions for the feed line break event. A RETRAN input model specific to WCGS wasdeveloped to simulate the core neutron kinetics, RCS, pressurizer, pressurizer sprays, SGs, MSSVs, andthe AFW system. Feed line break cases were modeled to address minimum and maximum reactivity feedback conditions with and without a LOOP occurring.

The RETRAN code computed thetime-dependent trends of pertinent variables, including the pressurizer

pressure, pressurizer water volume,SG mass, and reactor coolant temperatures.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1672.3.4.1.4 ResultsCalculated plant parameters following a major feed line break are shown in Figures 2.3.4-1through 2.3.4-10.

Results for the limiting case with offsite power available are presented inFigures 2.3.4-1 through 2.3.4-5.

Results for the limiting case where offsite power is lost are presented inFigures 2.3.4-6 through 2.3.4-10.

The calculated sequence of events for each of the reported cases isprovided in Tables 2.3.4-1 and 2.3.4-2.The system response following the feed line break is similar for both cases analyzed.

In both cases, the primaryand secondary pressures increase prior to RT. After RT occurs on low-low SG water level, the pressuredecreases

sharply, due to the cooldown caused by the break, until SLI occurs. The pressure in the faulted SGcontinues to decrease, whereas the pressure in the intact SGs and the primary side begins to increase until thesafety valve settings are reached.

Results presented in Figures 2.3.4-2 and 2.3.4-5 (with offsite poweravailable) and Figures 2.3.4-7 and 2.3.4-10 (without offsite power) show the predicted pressures in the RCSand MSS. As was previously discussed in Section 2.3.4.1.3, the feed line break analysis is bounded byLOL/TT with respect to meeting maximum pressure limits for the RCS and MSS. This is especially truebecause the reported cases for the feed line break analysis model operation of the pressurizer PORVs to reducetransient RCS pressure which is conservative with respect to bulk boiling concerns.

The primary temperatures are stable or increase slightly prior to RT and decrease sharply duringcooldown after RT. Once the heat-up begins, the primary temperature increases until the heat removalcapability of the intact SGs, with the inventory maintained by the AFW System, equals the decay heatgenerated in the core plus pump heat ("turn around" time). The peak primary temperature remains belowthe saturation temperature although the margin to boiling is decreased.

At the predicted time of turnaround, the minimum margin to HL saturation for the case with offsite power is 65.3°F, and 40.57F for thecase without offsite power. Thus, there is no bulk boiling in the RCS.2.3.4.2 Conclusions Based on the above information, it is concluded that for the postulated feed line break event the modeledAFW system performance is adequate for long-term heat removal.

The results confirm that for the feedline break event, the AFW system can adequately remove decay heat to preclude uncovering of thereactor core. Based on this, the plant continues to meet the requirements of GDCs 27, 28 and 35.2.3.4.3 References

1. WCAP-9230, "Report on the Consequences of a Postulated Main Feedline Rupture,"

January 1978.2. ANSI/ANS-5.1

-1979, "American National Standard for Decay Heat Power in Light WaterReactors,"

August 1979.3. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-168Table 2.3.4-1 Time Sequence of Events for Limiting Feed Line Break Case With Offsite PowerAvailable Event/Parameter Time (sec)Feed Line Break Occurs Causing Loss of FW to all SGs due to Harsh Environment 0.0Pressurizer PORV Opens (First Occurrence) 13.9Low-Low SG Water Level RT Setpoint Reached in all SGs 36.2Rods Begin to Drop and Faulted SG Begins Discharging Fluid Directly Out the Break 38.2Pressurizer PORV Closes (First Occurrence) 41.7AFW Flow is Delivered to Intact SGs 96.2Low Steam Line Pressure SI Setpoint Reached in Ruptured SG 132.7Main Steam Line Isolation Valves Closed 149.7SG Safety Valve Setpoint Reached in Intact SGs (First Occurrence) 579.0Core Decay Heat plus RCP Heat Decreased to AFW Heat Removal Capacity

-1700.0Table 2.3.4-2 Time Sequence of Events for Limiting Feed Line Break Case Without Offsite PowerAvailable Event/Parameter Time (sec)Feed Line Break Occurs Causing Loss of FW to all SGs due to Harsh Environment 0.0Pressurizer PORV Opens (First Occurrence) 13.9Low-Low SG Water Level RT Setpoint Reached in all SGs 36.3Rods Begin to Drop and Faulted SG Begins Discharging Fluid Directly out the Break 38.3Power Lost to RCPs 40.3Pressurizer PORV Closes (First Occurrence) 42.1AFW Flow is Delivered to Intact SGs 96.3Low Steam Line Pressure SI Setpoint Reached in Ruptured SG 105.6Main Steam Line Isolation Valves Closed 122.6SG Safety Valve Setpoint Reached (first occurrence) 553.4Core Decay Heat Decreased to AFW Heat Removal Capacity

-1100WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-169WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-169Li-00~C-)1.21'0.8-0.6-0.4-0.2-....11 ..AI I .0 1 2 3 410 10 10 10 10Time (seconds) 41*0.8-0.6-0.4-0.2-..............I I I I [ I I I I [ I l l[A.'.U01010210Time (seconds) 3104102, 0.1E-010av -0.1E-01-0.2E-01g -0.3E-01-0.4E-01--0.5E-01.....................-U.DL-U I010110210Time (seconds) 31010Figure 2.3.4-1Feed Line Break with Offsite Power Available Nuclear Power, Core Heat Flux and Total Core Reactivity versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-170WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1702,co,2~1U)NJCn0-0 1 2 310 10 10 10Time (seconds) 41001 2 310 10 10 10Time (seconds) 410Figure 2.3.4-2Feed Line Break with Offsite Power Available Pressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-171WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-17 11.2a,0cnC,)Cl)C-,1.1-1-0.90.8-0.7-..............................................................................

...............

0.t" I I i i I I i l i I I I I I I .I I l I i .I i I I I2O010110210Time (seconds) 310410Fraction of Ini tial Total Plant Feedwoter Moss Flownr .U-co2co,n iU.-1-I I II* i.I liiiI I I I I II-1010I10210Time (seconds) 310410Figure 2.3.4-3Feed Line Break with Offsite Power Available Reactor Coolant Flow and FW Line Break Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-172WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-172Faul ted Loop Hot Leg Temp.Fau lted Loop Cold Leg Temp.-Faulted Loop Hot Leg T-sat700-it 650 ...... ...--"----*--;-.

.....S 6 0 0 ................... -.E 550 ..................t-- -0 1 2 3 410 10 10 10 10Time (seconds)

Intact Loop Hot Leg Temp.Intact Loop Cold Leg Temp.Intact Loop Hot Leg T-sat1-1CDCLE0 1 2 310 10 10 10Time (seconds) 410Figure 2.3.4-4Feed Line Break with Offsite Power Available Faulted Loop and Intact Loop Reactor Coolant Temperatures versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-173Loop 1 F Foul ted)Loop 2 (Intoct)Loop 3 (Intact)Loop 4 I ntact)0ci,0~0~V-c(I)00VVC-,E0VU)0 1 2 310 10 10 10Time (seconds) 410Figure 2.3.4-5Feed Line Break with Offsite Power Available SG Shell Pressure versus TimeWCAP- 17658-NP August 2013WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-174WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-174C)L)CI,UC-,1.210.80.6-0.4-0.2-............0 IiII I I III0 1 2 3 410 10 10 10 10Time (seconds) 41-0.8-0.6-0.4-0.2-0I I I I LI III I I I 1 1 1 -- l , t01 2 3 410 10 10 10 10Time (seconds)

Ir*["nit.)n.V-0.1E-01

--0.2E-01

--0.3E-01

--0.4E-01

--0.5E-01

--0.6E-01.....................010110210Time (seconds) 310I410Figure 2.3.4-6Feed Line Break without Offsite PowerNuclear Power, Core Heat Flux and Total Core Reactivity versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-175WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-175C')CnC')a,0 1 2 3 410 10 10 10 10Time (seconds) 0 1 2 310 10 10 10Time (seconds) 410Figure 2.3.4-7Feed Line Break without Offsite PowerPressurizer Pressure and Pressurizer Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1761-a,-M-0 1 2 310 10 10 10Time (seconds) 410Fraction ofI i tial Total PIant Feedwater Mass FlowCL)C',co,0WL2U.,J'-.4 UU-0.5--1-...............

...............

-2010I ý I .I .I I I I I I I I I Ii I I i I I t I I [ i r I i I I i i I I I i I I i210210Time (seconds) 310410Figure 2.3.4-8Feed Line Break without Offsite PowerReactor Coolant Flow and FW Line Break Flow versus TimeWCAP- 17658-NP August 2013WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-177WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-177Faulted Loop Hot Leg Temp.Faulted Loop Cold Leg Temp.-Faul ted Loop Hot Leg T-sotCLE0 1 2 310 10 10 10Time (seconds) 410Intact LoopI Intact LoopIntact Loop700650"~600 ..........o 55 0 .. ......F--Hot Leg Temp.Cold Leg Temp.Hot Leg T-sat0 1 2 310 10 10 10Time (seconds) 410Figure 2.3.4-9Feed Line Break without Offsite PowerFaulted Loop and Intact Loop Reactor Coolant Temperatures versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-178WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-178LoopLoopL oo pLoopE1400-1200.1000T13.-800.600.= 400.E 200 --- ...200CD65) ~1234IFo Ited)In tactIntact)Intact)01 2 310 10 10 10Time (seconds) 410Figure 2.3.4-10Feed Line Break without Offsite PowerSG Shell Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-1792.4 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE2.4.1 Partial and Complete Loss of Forced Reactor Coolant Flow (USAR Sections 15.3.1and 15.3.2)2.4.1.1 Technical Evaluation 2.4.1.1.1 Introduction A loss of forced reactor coolant flow accident (USAR Sections 15.3.1 and 15.3.2) can result from thefollowing:

  • Mechanical or electrical failure in one or more RCPs* Interruption in the power supplying one or more of the RCPs* Reduction in RCP motor supply frequency If the reactor is at power at the time of the event, the immediate effect from the loss of forced coolantflow is a rapid increase in the coolant temperature.

This increase in coolant temperature could result inDNB, with subsequent fuel damage, if the reactor is not promptly tripped.The following signals provide protection against a loss of forced reactor coolant flow incident:

  • Low reactor coolant loop flow RT* UV on RCP power supply busses RT* Underfrequency (UF) on RCP power supply busses RTThe RT on low reactor coolant loop flow provides primary protection against partial loss-of-flow (PLOF)conditions.

This function is generated by two-out-of-three low-flow signals in any reactor coolant loop.Above Permissive P-8, low flow in any loop will actuate a RT. Between approximately 10 percent power(Permissive P-7) and the power level corresponding to Permissive P-8, low flow in two loops will actuatea RT. RT on low flow is blocked below Permissive P-7 because there is insufficient heat production to beconcerned about DNB.The RT on RCP UV is provided to protect against conditions that can cause a loss of voltage to all RCPs,that is, LOOP. An UV RT serves as an anticipatory backup to the low reactor coolant loop flow trip. TheUV trip function is blocked below approximately 10 percent power (Permissive P-7).The RCP UF RT is provided to trip the reactor for an UF condition resulting from frequency disturbances on the power grid. The RCP UF RT function is blocked below Permissive P-7. This trip function alsoserves as an anticipatory backup to the low reactor coolant loop flow trip.2.4.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThis accident was analyzed using the RTDP methodology (Reference 1). Initial NSSS power wasconservatively modeled to be at 3651 MWt, which includes all applicable uncertainties.

The RCSpressure and vessel average temperature were assumed to be at their nominal values. MMF was alsoWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-180assumed.

Uncertainties in initial conditions were accounted for in the DNBR limit value as described inthe RTDP.A conservatively large absolute value of the Doppler-only power coefficient was used. The analysis alsoassumed a conservative MTC of 0 pcm/°F at HFP conditions.

This resulted in the maximum core powerand hot spot heat flux during the initial part of the transient when the minimum DNBR is reached.Engineered safety systems (such as SI) are not required to function.

No single active failure in any systemor component required for mitigation will adversely affect the consequences of this event.A partial loss of forced reactor coolant flow incident is classified as a Condition II event as defined by theANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSIN 18.2-1973.

A complete loss of forced reactor coolant flow incident is classified by the ANS as aCondition III event. However, for conservatism, the incident was analyzed to Condition II criteria.

Theimmediate effect from a complete loss of forced reactor coolant flow is a rapid increase in the reactorcoolant temperature and subsequent increase in RCS pressure.

The following two items identify theacceptance criteria associated with the analysis of the loss of flow events:The CHF is not to be exceeded.

This is met by demonstrating that the minimum DNBR does notdecrease below the SAL value at any time during the transient.

Pressures in the RCS and MSS are maintained below 110 percent of their respective designpressures.

With respect to peak RCS and MSS pressures, the loss of forced reactor coolant flow event is boundedby the LOL/TT event described in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip,Inadvertent Closure of a Main Steam Isolation Valve, and Loss of Condenser Vacuum,"

in whichassumptions are made to conservatively calculate the RCS and MSS pressure transients.

For the loss offorced reactor coolant flow event, turbine trip occurs after RT, whereas for loss of load, the turbine tripis the initiating incident.

Therefore, the power mismatch between the primary and secondary sides andthe resultant temperature and pressure transients of the RCS and MSS are always more severe forLOL/TT than for the loss of forced reactor coolant flow. Based on this, no explicit calculation ofmaximum RCS or MSS pressure is performed for this event.The above acceptance criteria are based on meeting the relevant regulatory requirements of 10 CFR 50,Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of the specific GDCsthat are related to the loss of forced reactor coolant flow acceptance criteria are provided as follows.The specific acceptance criteria for this event are as follows:GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the loss of forced reactor coolant flow event, this is shown to be met bydemonstrating that the DNBR remains above the 95/95 DNBR limit at all times during thetransient.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-181GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.

For theloss of forced reactor coolant flow event, this is shown to be met by demonstrating that the peakRCS pressure is less than 110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.

For the loss of forced reactorcoolant flow event, which results in a RT, this is shown to be met by demonstrating that theDNBR remains above the 95/95 DNBR limit at all times during the transient with a trip reactivity that accounts for the most reactive rod stuck out of the core.2.4.1.1.3 Description of Analyses and Evaluations The following loss of forced reactor coolant flow cases were analyzed:

  • Loss of power to two RCPs (PLOF)* Loss of power to all RCPs (complete loss of flow (CLOF))* 5 Hz/second frequency decay of the RCPs power supply (CLOF-underfrequency (CLOF-UF))

The transients were analyzed with two computer codes. First, the RETRAN computer code (Reference 2)was used to calculate the following:

  • Time of RT based on the calculated flows* Nuclear power transient a Primary system pressure and temperature transients The VIPRE computer code (Reference
3) was then used to calculate the heat flux and DNBR based on thenuclear power and RCS temperature (enthalpy),
pressure, and flow from the output of the RETRANtransient run. The DNBR transients presented represent the minimum of the typical or thimble cell for thefuel.An evaluation of the P-8 permissive setpoint was performed and it was determined that the currentplant-specific value continued to provide adequate protection.

No change to the existing setpoint wasdeemed necessary.

Additionally, the effects of loop-to-loop flow asymmetry due to 10 percent SGTP imbalance have beenconsidered in the analysis.

2.4.1.1.4 ResultsThe PLOF case resulted in a low reactor coolant loop flow RT signal and the CLOF case resulted in anUV RCP RT signal. The CLOF-UF case resulted in an UF RCP RT signal. The VIPRE (Reference 3)WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-182analysis for these scenarios confirmed that the minimum DNBR acceptance criterion was met. Fuelcladding damage criteria were not challenged in any of the loss of forced reactor coolant flow casesbecause the DNB criterion was met.The analyses of the loss of flow events also demonstrated that the peak RCS and MSS pressures werewell below their respective limits.The most limiting of these cases in terms of the minimum calculated DNBR was the CLOF case. Thetransient results for each case are presented in Figures 2.4.1 -1 through 2.4.1-21.

The sequence of eventsfor each case is presented in Table 2.4. 1-1. Numerical results for the analyses are shown in Table 2.4.1-2.The analysis demonstrates that, for the aforementioned loss of flow cases, the DNBR did not decreasebelow the SAL value at any time during the transients.

Therefore, no fuel or cladding damage is predicted.

Also, the peak RCS and MSS pressures remained below their respective limits at all times. All applicable acceptance criteria were therefore met.The protection features identified in Section 2.4. 1. 1. 1 provide mitigation for the loss of forced reactorcoolant flow transients such that the above criteria are satisfied.

Furthermore, the results and conclusions of the loss of flow analysis will be confirmed on a cycle-specific basis as part of the normal RSE process.2.4.1.2 Conclusion The analyses of the decrease in forced reactor coolant flow event have been reviewed.

It is concluded thatthe analyses have adequately accounted for plant operations at a nominal NSSS power of up to3651 MWt, and were performed using acceptable analytical models. The review further concludes that theevaluation has demonstrated that the reactor protection and safety systems will continue to ensure that thespecified acceptable fuel design limits and the RCS and MSS pressure limits will not be exceeded as aresult of this event. Based on this, it is concluded that the plant will continue to meet the requirements ofGDCs 10, 15, and 26.2.4.1.3 References I. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRJETARY CLASS 32-183Table 2.4.1-1 Time Sequence of Events -Loss of Forced Reactor Coolant FlowCase Event Time (seconds)

Loss of Power to Two RCPs (PLOF) Flow Coastdown Begins 0.0Reactor Coolant Low-Flow Trip Setpoint Reached 1.5Rods Begin to Drop 2.5Minimum DNBR Occurs 4.3Loss of Power to All RCPs (CLOF) Flow Coastdown Begins 0.0Rods Begin to Drop(D 1.5Minimum DNBR Occurs 3.55 Hz/sec Frequency Decay of the Frequency Decay Begins 0.0RCPs Power Supply (CLOF-UF)

Underfrequency RT Setpoint Reached 0.6Rods Begin to Drop 1.2Minimum DNBR Occurs 3.3Note:I. LUV RT (rods begin to drop) is assumed to occur 1.5 seconds following the loss of bus voltage.Table 2.4.1-2 Results -Loss of Forced Reactor Coolant FlowMinimum DNBR Limit ValueLoss of Power to Two RCPs (PLOF) 1.82 1.52Loss of Power to All RCPs (CLOF) 1.69 1.525 Hz/sec Frequency Decay of the RCPs Power Supply (CLOF-UF) 1.73 1.52WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-184WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-184I0I.f-.....................................

0 .8 -..............................................i-,E<C.)0.61 ..........................................................

0.41 ...........

...........

............

0.2-1...............................................n-i0246810Time (sec)Figure 2.4.1-1 PLOF -Core Volumetric Flow Rate versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-185Loop 1Loop 2-Loop 3Loop 4E-50 2 4 6 8lime (sec)10Figure 2.4.1-2 PLOF -Loop Volumetric Flow Rates versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-186w0~0 2 4 6 8lime (sec)10Figure 2.4.1-3 PLOF -Nuclear Power versus TimeWCAP-l 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-187WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-187CnC',C/3P,0'l0 2 46810Time (sec)Figure 2.4.1-4 PLOF -Pressurizer Pressure versus TimeWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportALIgust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-188CD)6Time (sec)10Figure 2.4.1-5 PLOF -Core Average Heat Flux versus TimeWCAP- 17658-NP August 2013WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-189WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-189wa,czC-,0 2 4 6 8Time (sec)10Figure 2.4.1-6 PLOF -Hot Channel Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-190WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-190EE0 2 46810Time (sec)Figure 2.4.1-7 PLOF -Minimum DNBR versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-191WESTJNGHOUSE NON-PROPRIETARY CLASS 3 2-1911.2U-I-.Ea-,C-,0 2 4 6 8Time (see)10Figure 2.4.1-8 CLOF -Core Volumetric Flow Rate versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-192WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-192Loop 1Loop 2Loop 3Loop 40C,0E_..J2 46810Time (sec)Figure 2.4.1-9 CLOF -Loop Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-193WESTINGHOUSE NON-PROPRiETARY CLASS 3 2-193wC-)Time (sec)Figure 2.4.1-10 CLOF -Nuclear Power versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-194WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1942-C/)C/)M~0 2 4 6 8Time (sec)10Figure 2.4.1-11 CLOF -Pressurizer Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-195a.,)C-D0 2 4 6 8 10Time (sec)Figure 2.4.1-12 CLOF- Core Average Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-196WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1961.2WQ-):izciC-,ci0 2 4 6 8Time (sec)10Figure 2.4.1-13 CLOF -Hot Channel Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-197WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-197EE02 46810Time (sec)Figure 2.4.1-14 CLOF -Minimum DNBR versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-198Lj~a.)C)C-)a.)E-5a.)(-)0 2 4 6 8lime (sec)10Figure 2.4.1-15 CLOF-UF -Core Volumetric Flow Rate versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-199Loop 1Loop 2Loop 3Loop 4CD,czl0 2 46810Time (sec)Figure 2.4.1-16 CLOF-UF -Loop Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-200WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-200_L__CD0 2 4 6 8Time (sec)10Figure 2.4.1-17 CLOF-UF -Nuclear Power versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-201WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-201Cf)C/)r14C/)M)0 2 4 6 8 10Time (sec)Figure 2.4.1-18 CLOF-UF-Pressurizer Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-202WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-202LLa,a,a,a,C-)0 2 4 6 8Time (sec)10Figure 2.4.1-19 CLOF-UF -Core Average Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-203wa,C-)0 2 4 6 8Time (see)10Figure 2.4.1-20 CLOF-UF -Hot Channel Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-204Q__0 2 4 6 8Time (see)10Figure 2.4.1-21 CLOF-UF-Minimum DNBR versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2052.4.2 Reactor Coolant Pump Shaft Seizure (Locked Rotor) and Shaft Break(USAR Sections 15.3.3 and 15.3.4)2.4.2.1 Technical Evaluation 2.4.2.1.1 Introduction The event postulated is an instantaneous seizure of a RCP rotor or the sudden break of the RCP shaft.Flow through the affected reactor coolant loop is rapidly reduced, leading to initiation of a RT on a lowreactor coolant flow signal.Following initiation of the RT, heat stored in the fuel rods continues to be transferred to the coolant,causing the coolant to expand. At the same time, heat transfer to the shell side of the SGs is reduced, firstbecause the reduced flow results in a decreased tube-side film heat transfer coefficient, and secondbecause the temperature differential between the reactor coolant in the tubes and the shell-side fluid isdecreased.

The rapid expansion of the coolant in the reactor core causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steamvolume, actuates the automatic spray system, opens the PORVs, and opens the PSVs, in that sequence.

The PORVs are designed for reliable operation and are expected to function properly during the event.However, for conservatism, their pressure-reducing effect, as well as the pressure-reducing effect of thepressurizer spray, was not included in the analysis.

The consequences of a locked rotor (that is, an instantaneous seizure of a pump shaft) are very similar tothose of a pump shaft break. The initial rate of the reduction in coolant flow is slightly greater for thelocked rotor event. However, with a broken shaft, the impeller could conceivably be free to spin in thereverse direction.

The effect of reverse spinning is a reduced core flow when compared to the locked rotorscenario.

The analysis considers only one scenario; it represents the most limiting (conservative) combination of conditions for the locked rotor and pump shaft break events.2.4.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions There were three locked rotor cases analyzed, with each being applicable to the WCGS: one for peak RCSpressure and PCT concerns with LOOP, one for peak RCS pressure and PCT concerns without LOOP, anda third to determine the percentage of rods-in-DNB.

For the cases performed to evaluate peak RCSpressure and PCT concerns, one locked rotor and shaft break was simulated with all reactor coolant loopsin operation; one of these cases accounted for a LOOP and the other considered a continued supply ofoffsite power. Inputs for these cases were designed to maximize the RCS pressure and claddingtemperature transients; the STDP was applied for these cases. Initial core power, reactor coolanttemperature, and pressurizer pressure were modeled to be at their maximum values consistent withfull-power conditions, including allowances for calibration and instrument errors. The initial reactorcoolant flow was the TDF. These inputs resulted in a conservative calculation of the coolant insurge intothe pressurizer, which in turn resulted in a maximum calculated peak RCS pressure.

The case thatconsidered a LOOP conservatively modeled the intact RCPs as being tripped coincident with RT.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-206The third case was run to confirm that the percentage of rods that experience DNB, also referred to aspercentage of rods-in-DNB, with LOOP is less than that considered in the radiological analysis.

As in thepeak RCS pressure/PCT case, one locked rotor and shaft break was simulated with all reactor coolantloops in operation.

Initial NSSS power was conservatively modeled to be at 3651 MWt, which includesall applicable uncertainties.

The pressurizer pressure and Tavg were modeled to be at their nominal values.The initial reactor coolant flow was the MMF. Uncertainties in initial pressure and temperature conditions were accounted for in the DNBR SAL value as described in the RTDP (Reference 1).A least negative MTC and a conservatively large (absolute value) Doppler-only power coefficient weremodeled in the analysis.

The negative reactivity from control rod insertion/scram was based on4.0 percent Ak/k trip reactivity from HFP conditions.

Engineered safety systems (such as SI) are not required to function.

No single active failure in any systemor component required for mitigation will adversely affect the consequences of this event.Acceptance CriteriaThe RCP locked rotor/shaft break accident is classified as a Condition IV event as defined by the ANS's"Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSIN 18.2-1973.

An RCP locked rotor/shaft break results in a rapid reduction in forced reactor coolant loopflow that increases the reactor coolant temperature and subsequently causes the fuel cladding temperature and RCS pressure to increase.

The following items summarize the acceptance criteria for the analysis ofthis event:The potential for fuel cladding damage due to the combination of high cladding temperature and theexothermic zirconium oxidation process (zirconium-water reaction) provides a failure mechanism thatcould affect the core cooling capability.

With respect to fuel cladding temperature, the maximum claddingtemperature at the core hot spot must remain below 2700'F, and the zirconium-water reaction at the corehot spot must be less than 16 percent by weight. Satisfying these criteria conservatively ensures that thecore will remain in place and geometrically intact with no loss of core cooling capability.

Pressures in the RCS and MSS are to be maintained below 110 percent of the respective design pressures.

With respect to the MSS pressure transient, this event is bounded by the LOL/TT event discussed inSection 2.3.1, "Loss of External Electrical Load, Turbine Trip, Inadvertent Closure of a Main SteamIsolation Valve, and Loss of Condenser Vacuum."

This is because the turbine trip occurs later, coincident with RT, in this event compared to the LOL/TT event where the turbine trip is the initiating fault. Thegreater mismatch between the primary-side power and secondary-side power makes the MSS pressuretransient more severe for the LOL/TT event. Thus, the peak MSS pressure is not reported for the lockedrotor analysis.

Because it is a Condition IV event, the locked rotor/shaft break transient is allowed to result in a minimalrelease of radioactive material such that the calculated doses at the site boundary are within acceptable limits (see Section 4.3.5 of Enclosure VI of this LAR). For dose considerations, fuel failure isconservatively assumed for all fuel rods that are shown to experience DNB. In the dose analysis (seeSection 4.3.5 of Enclosure VI of this LAR), 5 percent of the fuel rods were assumed to have failed andWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-207released radioactive material as a result of a locked rotor/shaft break event. Therefore, the total percentage of rods-in-DNB must be less than the 5 percent value used in the dose analysis.

The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the locked rotor/shaft break acceptance criteria are provided as follows.* GDC 27 (Combined Reactivity Control Systems Capability) requires that the reactivity controlsystems be designed to have a combined capability, in conjunction with poison addition by theECCS, of reliably controlling reactivity changes under postulated accident conditions, withappropriate margin for stuck rods, to assure the capability to cool the core is maintained.

In thelocked rotor event analysis, the applied trip reactivity accounts for the highest worth rod beingstuck fully out of the core, and the capability to cool the core is demonstrated by showing that thelimits for fuel cladding temperature, zirconium-water

reaction, and RCS pressure are met.GDC 28 (Reactivity Limits) requires that the reactivity control systems be designed to assure thatthe effects of postulated reactivity accidents can neither result in damage to the RCPB greaterthan limited local yielding, nor disturb the core, its support structures, or other reactor vesselinternals so as to significantly impair the capability to cool the core. For the locked rotor event,this is shown to be met by demonstrating that the limits for fuel cladding temperature, zirconium-water
reaction, and RCS pressure are met.GDC 31 (Fracture Prevention of Reactor Coolant Pressure Boundary) requires that the RCPB bedesigned with sufficient margin to assure that, under specified conditions, it will behave in anon-brittle manner and the probability of a rapidly propagating fracture is minimized.

For thelocked rotor event, this is shown to be met by demonstrating that the RCS pressure limit is met.2.4.2.1.3 Description of Analyses and Evaluations The locked rotor transient was analyzed with two primary computer codes. First, the RETRAN computercode (Reference

2) was used to calculate the loop and core flows during the transient, the time of RTbased on the calculated flows, the nuclear power transient, and the primary system pressure andtemperature transients.

The VIPRE code (Reference

3) was then used to calculate the PCT using thenuclear power and RCS temperature (enthalpy),
pressure, and flow from RETRAN.For the peak RCS pressure evaluation, the initial pressure was conservatively estimated to be 50 psi abovethe nominal pressure of 2250 psia, which accounts for initial condition uncertainties in the pressurizer pressure measurement and control channels.

This was done to obtain the highest possible rise in thecoolant pressure during the transient.

The pressure response reported in Table 2.4.2-2 corresponds to thelocation in the RCS that has the maximum pressure, that is, in the lower plenum of the reactor vessel.No credit was taken for the pressure-reducing effect of the pressurizer PORVs, pressurizer spray, or steamdump. Although these systems are expected to function and would result in a lower peak pressure, anadditional degree of conservatism was provided by not including their effect. The PSV model includeda +2 percent valve opening tolerance above the nominal setpoint of 2460 psig plus a 1 percent setWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-208pressure shift and a 1. 153-second purge time delay to account for the existence of PSV water-filled loopseals, as described in Reference 4.Post-DNB heat transfer is limited to film boiling, and the film boiling coefficient was calculated in theVIPRE code (Reference

3) using the Bishop-Sandberg-Tong heat transfer correlation.

The fluid properties were evaluated at film temperature.

The code calculated the film coefficient at every time step based uponthe actual heat transfer conditions at the time. The nuclear power, system pressure, bulk density, and RCSflow rate as a function of time were based on the RETRAN results.The magnitude and time dependence of the heat transfer coefficient between the fuel and cladding (gapcoefficient) has a pronounced influence on the thermal results.

The larger the value of the gap coefficient, the more heat is transferred between the pellet and cladding.

Based on investigations on the effect of thegap coefficient upon the maximum cladding temperature during the transient, the gap coefficient wasassumed to increase from a steady-state value consistent with the initial fuel temperature to approximately 10,000 Btuihr-ft 2-OF at the initiation of the transient.

Therefore, the large amount of energy stored in thefuel because of the small initial gap coefficient was released to the cladding at the initiation of thetransient.

The zirconium-water reaction can become significant above 1800'F (cladding temperature).

TheBaker-Just parabolic rate equation was used to define the rate of zirconium-water reaction.

The effect ofthe zirconium-water reaction was included in the calculation of the PCT temperature transient.

2.4.2.1.4 ResultsWith respect to the peak RCS pressure, PCT and zirconium-water

reaction, the analysis demonstrated thatall applicable acceptance criteria were met for the WCGS. The calculated sequence of events is presented in Table 2.4.2-1 for the locked rotor/shaft break event. The results of the calculations (peak pressure, PCTand zirconium-water reaction) for the limiting case (with a LOOP) are summarized in Table 2.4.2-2.

Thetransient results for the peak pressure/PCT cases (with a LOOP and without a LOOP) are provided inFigures 2.4.2-1 through 2.4.2-6, and the transient results for the rods-in-DNB case are provided inFigures 2.4.2-7 through 2.4.2-12.

The locked rotor/shaft break analysis performed for the WCGS demonstrated that the PCT calculated forthe hot spot remained considerably less than 2700'F, and the amount of zirconium-water reaction wassmall. Under such conditions, the core would remain in place and intact with no loss of core coolingcapability.

The analysis also confirmed that the peak RCS pressure reached during the transient was less than theacceptance limit, and thereby, the integrity of the primary coolant system was demonstrated.

The totalnumber of rods-in-DNB was less than 5 percent.

The low reactor coolant flow RT function providedmitigation for the locked rotor/shaft break transient such that the above criteria were satisfied.

Furthermore, the results and conclusions of this analysis will be confirmed on a cycle-specific basis aspart of the normal RSE process.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2092.4.2.2 Conclusion The analyses of the sudden decrease in core coolant flow due to a locked rotor/shaft break event havebeen examined.

It is concluded that the analyses have adequately accounted for plant operation at theanalyzed power level and were performed using acceptable analytical models. It is further concluded thatthe evaluation has demonstrated that the reactor protection and safety systems will continue to ensure thatthe ability to insert control rods is maintained, the RCS pressure limit will not be exceeded, the RCPBwill behave in a nonbrittle manner, the probability of propagating fracture of the RCPB is minimized, andadequate core cooling will be provided.

Based on this, it is concluded that the plant will continue to meetthe requirements of GDCs 27, 28, and 31.2.4.2.3 References

1. WCAP-1 1397-P-A, "Revised Thermal Design Procedure,"

April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.4. WCAP-129 10, Rev. I-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-210WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 10Table 2.4.2-1 Time Sequence of Events -RCP Locked Rotor/Shaft BreakCase Event Time (seconds)

Locked Rotor/Shaft Break -Rotor on One Pump Locks or the Shaft 0.0Overpressurization/PCT with LOOP BreaksReactor Coolant Low-Flow RT Setpoint 0.04ReachedRods Begin to Drop 1.04Undamaged RCPs Lose Power and Begin to 1.04Coast DownM')Maximum Cladding Temperature Occurs 3.90Maximum RCS Pressure Occurs 4.75Locked Rotor/Shaft Break -Rotor on One Pump Locks or the Shaft 0.0Overpressurization/PCT without BreaksLOOPReactor Coolant Low-Flow RT Setpoint 0.04ReachedRods Begin to Drop 1.04Maximum Cladding Temperature Occurs 3.50Maximum RCS Pressure Occurs 4.08Locked Rotor/Shaft Break -Rotor on One Pump Locks or the Shaft 0.0Rods-in-DNB BreaksReactor Coolant Low-Flow RT Setpoint 0.04ReachedRods Begin to Drop 1.04Undamaged RCPs Lose Power and Begin to 1.05Coast Downt'1Minimum DNBR Occurs 3.20Note:1. The undamaged RCPs were modeled to trip coincident with rod motion, but slight differences, which are considered tobe negligible, occur because of the RETRAN code trip logic.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-211Table 2.4.2-2 Limiting Results -RCP Locked Rotor/Shaft BreakCriterion Analysis Value LimitPCT at Core Hot Spot (°F) 1786.6 2700Maximum Zirconium-Water Reaction at Core Hot Spot (%) 0.29 16.0Maximum RCS Pressure (psia) 2675.1 2750Maximum Number of Rods-in-DNB

(%) 0.7 5WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-212WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 12with Of fsi te Powerwi thou t Of f s Ite Power1CD0000)C-)0.8-0.4-0.20510Time (seconds) 1520Figure 2.4.2-1RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseCore Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-213WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 13wI th Of fs te Powerw ithout Of f s i t e Power1.200CDiC)Lci10.8-0.4-0.2-0--0.4--0.6I I I fI I III I I I I I I I I I5010Time (seconds) 1520Figure 2.4.2-2RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseFaulted Loop Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-214WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 14w ith Of f s ite Powerw ithout Of f s ite Power27F~A -LUu2600ot0)EI0)2500 t2400 t2300-I I I I I I I I I22000510Time (seconds) 1520Figure 2.4.2-3RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseMaximum RCS Pressure versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-215WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-215wi th Of f s i te Powerwithout Of fsite Powe-1.2cici0ci0cicici0m0cicici0.60.4-T0.20I I I I II I I I I I I --I I i I I I I0510Time (seconds) 1520Figure 2.4.2-4RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseNuclear Power versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-216WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 16with Of fsi te Powerwi thout Of f si te Powee1.20)C)0.6-t0.4-0.2-I I I I II If I I I I I I I I I I IUL/0510Time (seconds) 1520Figure 2.4.2-5RCP Locked Rotor/Shaft Break Overpressurization/PCT CaseCore Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-217WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-217with Of fsi te Powerwi thout Of s te Powe-'1f~f'd~

-,zUUU1800-1600t01)0.)E0)-0)0)1400-1200-000-t800-6000510T;me (seconds) 1520Figure 2.4.2-6RCP Locked Rotor/Shaft Break Overpressurization/PCT CasePCT versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-218WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 181.2C_CD00i0(DQ0--I -0.80.6tVI I0.4-0./010Time (seconds) 15)0Figure 2.4.2-7RCP Locked Rotor/Shaft Break Rods-in-DNB CaseCore Volumetric Flow Rate versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-219WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 19Fau I ted LoopIntac t Loops1.2CDC300Qi1-0.80.60.4-0.20-0.2040510Time (seconds) 1520Figure 2.4.2-8RCP Locked Rotor/Shaft Break Rods-in-DNB CaseLoop Volumetric Flow Rates versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-220WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-22025002450-2400-(n)Cf)C)C/)23502300-2250-2200-2150I I I I I I I I I I I I I I051520Time (seconds)

Figure 2.4.2-9RCP Locked Rotor/Shaft Break Rods-in-DNB CasePressurizer Pressure versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-221WESTiNGHOUSE NON-PROPRIETARY CLASS 3 2-22 11.2C)ci0ci0(9ciG)00C)C)C)0.80.60.2-00510Time (seconds) 1520Figure 2.4.2-10RCP Locked Rotor/Shaft Break Rods-in-DNB CaseNuclear Power versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-222WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2221-) -VILZ0C_00.6-0.4-0.2-I II I 1 1UI010Time (seconds) 1520Figure 2.4.2-11RCP Locked Rotor/Shaft Break Rods-in-DNB CaseCore Average Heat Flux versus TimeWCAP- 17658-NP August 2013WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-223I') -I./_cici0ci0C)ciciwciciILcicicici-ci(-I0IL0.4-0.2 tI I I If I I I I I I I I I I11 -~U0510Time (seconds) 1520Figure 2.4.2-12RCP Locked Rotor/Shaft Break Rods-in-DNB CaseHot Channel Heat Flux versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2242.5 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 2.5.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical orLow Power Startup Condition (USAR Section 15.4.1)2.5.1.1 Technical Evaluation The specific acceptance criteria applied for this event are as follows:The DNBR should remain above the applicable 95/95 DNBR limits at all times during thetransient.

Demonstrating that the DNBR limits are met satisfies the requirements of GDC 10.Per GDC 20, the protection system should be designed to automatically initiate the operation ofappropriate

systems, including the reactivity control systems, to ensure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and to senseaccident conditions and initiate the operation of safety-related systems and components.

For thisevent, protection is provided via the high neutron flux (low setting).

GDC 25 requires that the protection system is designed to ensure that specified acceptable fueldesign limits are not exceeded for any single malfunction of the reactivity control systems, suchas accidental withdrawal (not ejection or dropout) of control rods. Demonstrating that the fueldesign limits (that is, DNBR) are met satisfies the requirements of GDC 25.The following discussion demonstrates that all applicable acceptance criteria are met for this event for theWCGS.2.5.1.1.1 Introduction An uncontrolled RCCA withdrawal incident is defined as an uncontrolled addition of reactivity to thereactor core by withdrawal of RCCAs, resulting in a power excursion.

Although the probability of atransient of this type is extremely low, such a transient could be caused by a malfunction of the reactorcontrol rod drive system. This could occur with the reactor either subcritical or at power. The "at power"occurrence is discussed in Section 2.5.2. The uncontrolled RCCA withdrawal from a subcritical condition is classified as a Condition II event, a fault of moderate frequency, as defined by the ANS's "NuclearSafety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

During startup, when bringing the reactor from a shutdown condition to a low-power level, reactivity isadded at a prescribed and controlled rate by RCCA withdrawal or by reducing the core boronconcentration.

RCCA motion can cause much faster changes in reactivity than can result from changingboron concentration.

The rods are physically prevented from withdrawing in other than their respective banks. Power suppliedto the rod banks is controlled such that no more than two banks can be withdrawn at any time. The controlrod drive mechanism (CRDM) is of the magnetic latch type and the coil actuation is sequenced to providevariable speed rod travel. The maximum reactivity insertion rate is analyzed in the detailed plant analysisWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-225assuming the simultaneous withdrawal of the combination of the two rod banks with the maximumcombined worth at maximum speed.The neutron flux response to a continuous reactivity insertion is characterized by a very fast flux increaseterminated by the reactivity feedback effect of the negative Doppler coefficient.

This self-limitation of theinitial power increase results from a fast negative fuel temperature feedback (Doppler effect) and is ofprime importance during a startup transient because it limits the power to an acceptable level prior toprotection system action. After the initial power increase, the nuclear power is momentarily reduced andthen, if the incident is not terminated by a RT, the nuclear power increases again, but at a much slowerrate.Should a continuous RCCA withdrawal be initiated, the transient will be terminated by one of thefollowing automatic protective functions:

Source range neutron flux RT -actuated when either of two independent source range channelsindicates a flux level above a preselected, manually adjustable setpoint.

This trip function may bemanually bypassed only after an intermediate range neutron flux channel indicates a flux levelabove the source range cutoff power level. It is automatically reinstated when both intermediate channels indicate a flux level below the source range cutoff power level.Intermediate range neutron flux RT -actuated when either of two independent intermediate rangechannels indicates a flux level above a preselected, manually adjustable setpoint.

This tripfunction may be manually bypassed when two of the four power range channels are readingabove approximately 10 percent of full power and is automatically reinstated when three of thefour channels indicate a power level below this value.Power range neutron flux RT (low setting)

-actuated when two of the four power range channelsindicate a power level above approximately 25 percent of full power. This trip function may bemanually bypassed when two of the four power range channels indicate a power level aboveapproximately 10 percent of full power. This trip function is automatically reinstated when threeof the four channels indicate a power level below 10 percent power.Power range neutron flux RT (high setting)

-actuated when two out of the four power rangechannels indicate a power level above approximately 109 percent of full power. This trip functionis active in Modes 1 and 2, when the low setting is bypassed.

High nuclear flux rate RT -actuated when the positive rate of change of neutron flux on two outof four nuclear power range channels indicates a rate above the preset nominal setpoint ofapproximately 4.0 percent in 2 seconds.

This trip function is always active in Modes 1 and 2, andit is not explicitly modeled in the analysis of this event.In addition, control rod stops on high intermediate range flux level (one out of two) and high power rangeflux level (one out of four) serve to discontinue rod withdrawal and prevent the need to actuate theintermediate range flux level trip and the power range flux level trip, respectively.

This analysis creditsthe power range neutron flux trip (low setting) to initiate the RT.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2262.5.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe accident analysis uses the STDP methodology because the conditions resulting from the transient areoutside the range of applicability of the RTDP methodology.

To obtain conservative results for theanalysis of the uncontrolled RCCA bank withdrawal from subcritical event, the following inputparameters and initial conditions are modeled:The magnitude of the nuclear power peak reached during the initial part of the transient, for anygiven reactivity insertion rate, is strongly dependent on the Doppler-only power defect. Therefore, a conservatively low absolute value is used (1007 pcm) to maximize the nuclear power transient.

A most-positive MTC (+6 pcm/nF) is used because this yields the maximum rate of powerincrease.

The contribution of the moderator temperature coefficient is negligible during the initialpart of the transient because the heat transfer time constant between the fuel and moderator ismuch longer than the nuclear flux response time constant.

However, after the initial neutron fluxpeak, the succeeding rate of power increase is affected by the moderator reactivity coefficient.

The analysis assumes the reactor to be at HZP conditions with a nominal no-load temperature of557°F. This assumption is more conservative than that of a lower initial system temperature (thatis, shutdown conditions).

The higher initial system temperature yields a larger fuel-to-moderator heat transfer coefficient, a larger specific heat of the moderator and fuel, and a less-negative (smaller absolute magnitude)

Doppler defect. The less-negative Doppler defect reduces theDoppler feedback effect, thereby increasing the neutron flux peak. The high neutron flux peakcombined with a high fuel specific heat and larger heat transfer coefficient yields a larger peakheat flux.The analysis assumes the initial effective multiplication factor (Kff) to be 1.0 because itmaximizes the peak neutron flux and results in the most severe nuclear power transient.

RT is assumed on power range high neutron flux (low setting).

A conservative combination ofinstrumentation error, setpoint error, delay for trip signal actuation, and delay for control rodassembly release is modeled.

The analysis assumes a 10 percent uncertainty in the power rangeflux trip setpoint (low setting),

increasing it from the nominal value of 25 percent of full power to35 percent of full power. A delay time of 0.5 seconds is assumed for trip signal actuation andcontrol rod assembly release.

No credit is taken for the source range or intermediate rangeprotection.

During the transient, the increase in nuclear power is so rapid that the effect of errorsin the trip setpoint on the actual time at which the rods release is negligible.

In addition, the totalRT reactivity is based on the assumption that the highest worth RCCA is stuck in its fullywithdrawn position.

The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the two sequential control banks having the greatest combined worth at themaximum rod withdrawal speed. The assumed reactivity insertion rate is 75 pcm/sec, which isbased on a rod worth of 100 pcm/inch and a maximum rod speed of 72 steps per minute.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-227The DNB analysis assumes the most limiting axial and radial power shapes possible during thefuel cycle associated with having the two highest combined worth banks in their highest worthposition.

The analysis assumes the initial power level to be below the power level expected for anyshutdown condition (10-9 fraction of nominal power). The combination of highest reactivity insertion rate and low initial power produces the highest peak heat flux.The analysis assumes two of the four RCPs to be in operation.

This is conservative with respect tothe DNB transient.

This accident analysis uses the STDP methodology.

The use of the STDP stipulates that the RCSflow rates will be based on a fraction of the thermal design flow for two pumps operating.

Because the event is analyzed from HZP, the steady-state non-RTDP uncertainties are notconsidered in defining the initial conditions.

The uncontrolled RCCA bank withdrawal from subcritical event is considered an ANS Condition II event,a fault of moderate frequency, and is analyzed to show that the core and RCS are not adversely affectedby the event. This is demonstrated by showing that the DNB design basis is not violated and subsequently that there is little likelihood of core damage. It must also be shown that the peak hot spot fuel centerline temperature remains within the acceptable limit (48000F), although for this event, the heatup is relatively non-limiting.

2.5.1.1.3 Description of Analyses and Evaluations The analysis of the uncontrolled RCCA bank withdrawal from subcritical conditions is performed in threestages. First, a spatial neutron kinetics computer code, TWINKLE (Reference 1), is used to calculate thecore average nuclear power transient, including the various core feedback effects; that is, Doppler andmoderator reactivity.

Next, the FACTRAN computer code (Reference

2) uses the average nuclear powercalculated by TWINKLE and performs a fuel rod transient heat transfer calculation to determine the coreaverage heat flux and hot spot fuel temperature transients.
Finally, the core average heat flux calculated by FACTRAN is used in the VIPRE computer code (Reference
3) for transient DNBR calculations.

2.5.1.1.4 ResultsThe analysis shows that all applicable acceptance criteria are met for the WCGS. The minimum DNBRnever decreases below the applicable limit values and the peak fuel centerline temperature is 2342'F. Thepeak temperatures are well below the minimum temperature at which fuel melting would be expected(4800°F).

Figure 2.5. 1-1 shows the nuclear power transient, Figure 2.5.1-2 shows the core average heat fluxtransient, and Figures 2.5.1-3 and 2.5.1-4 show the fuel average and cladding surface temperature transients at the hot spot.The time sequence of events for both cases is presented in Table 2.5.1-1.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-228In the event of an RCCA withdrawal event from subcritical conditions, the core and the RCS are notadversely affected because the combination of thermal power and coolant temperature results in aminimum DNBR greater than the SAL value. Furthermore, because the maximum fuel temperatures predicted to occur during this event are much less than those required for fuel melting to occur, no fueldamage is predicted as a result of this transient.

Cladding damage is also precluded.

2.5.1.2 Conclusions Based on a review of the analysis of the uncontrolled RCCA withdrawal from a subcritical or low-power startup condition, it is concluded that the analysis adequately accounted for plant operation at the statedpower level and were performed using acceptable analytical models. It is further concluded that theanalysis demonstrates that the reactor protection and safety systems will continue to ensure that thespecified acceptable fuel design limits are not exceeded.

Based on this, it is concluded that the plant willcontinue to meet the requirements of GDCs 10, 20, and 25.2.5.1.3 References

1. WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non-Proprietary),

"TWINKLEMulti-dimensional Neutron Kinetics Computer Code," January 1975.A2. WCAP-7908-A, "FACTRAN

-A FORTRAN IV Code for Thermal Transients in a UO2 FuelRod," December 1989.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-229Table 2.5.1-1 Time Sequence of Events -Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition Event Time (seconds)

Initiation of Uncontrolled Rod Withdrawal 0.0Power Range High Neutron Flux Low Setpoint is Reached 10.43Peak Nuclear Power Occurs 10.57Rod Motion Begins 10.93Peak Heat Flux Occurs (0.3642) 12.73Minimum DNBR Occurs (1.66) 12.73Peak Average Cladding Temperature Occurs (683°F) 13.06Peak Average Fuel Temperature Occurs (1934°F) 13.26Peak Fuel Centerline Temperature Occurs (2342°F) 13.71WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17659-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-230WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-230I.E00-0'4--00DO3-0.8"0.4-0.2'0 5 10 15 20 25Time (seconds) 30Figure 2.5.1-1 Rod Withdrawal from Subcritical

-Nuclear Power Transient WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-231WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-23 10F-0r0 5 10 15 20 25Time (seconds)

Figure 2.5.1-2 Rod Withdrawal from Subcritical

-Core Average Heat Flux Transient WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-232WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-232CLC,)EQ)F-a,C/_-0a-"U-0 5 10 15 20 25 30Time (seconds)

Figure 2.5.1-3 Rod Withdrawal from Subcritical

-Fuel Average Temperature Transient WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-233WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-233U-C,,a)a)L..a)a)I.-DL..a)0~Ea)a)C)U)00~U)00 5 10 15 20 25Time (seconds) 30Figure 2.5.1-4 Rod Withdrawal from Subcritical

-Cladding Surface Temperature Transient WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2342.5.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (USARSection 15.4.2)2.5.2.1 Technical Evaluation 2.5.2.1.1 Introduction An uncontrolled RCCA bank withdrawal at power that causes an increase in core heat flux can result fromfaulty operator action or a malfunction in the rod control system. Immediately following the initiation ofthe transient, the SG heat removal rate lags behind the core power generation rate until the SG pressurereaches the setpoint of the SG relief or safety valves. This imbalance between heat removal and heatgeneration rate causes the reactor coolant temperature to increase.

Unless terminated, the power mismatchand resultant coolant temperature increase could eventually result in a violation of the DNBR SAL, fuelcenterline melt, and/or RCS overpressurization.

Therefore, to avoid core damage, the reactor protection system is designed to automatically terminate any such transient before the DNBR falls below the limitvalue, or the fuel rod linear heat generation rate (kW/ft) limit is exceeded.

The reactor protection systemand PSVs are designed to preclude exceeding the RCS pressure boundary safety limit.The automatic features of the reactor protection system that prevent core damage and preclude RCSoverpressurization during an RCCA bank withdrawal incident at power include the following:

Power range neutron flux instrumentation actuates a RT if two-out-of-four channels exceed anoverpower setpoint.

RT actuates if any two-out-of-four channels exceed the power range neutron flux high positiverate setpoint.

RT actuates if any two-out-of-four OTAT channels exceed the corresponding setpoint.

Thissetpoint is automatically varied with axial power distribution, coolant average temperature, andpressure to help protect the DNB design basis.RT actuates if any two-out-of-four OPAT channels exceed the corresponding setpoint.

Thissetpoint is capable of being automatically varied with axial power imbalance to help ensure thatthe allowable heat generation rate (kW/ft) is not exceeded.

A high pressurizer pressure RT actuates if any two-out-of-four pressure channels exceed thecorresponding

setpoint, which is set at a fixed point. This pressure setpoint is less than the setpressure for the PSVs.MSSVs can open for this event and provide additional steam flow.A high pressurizer water level RT actuates if any two-out-of-three channels exceed the tripsetpoint, which is set at a fixed value, when the reactor power is above approximately 10 percent(Permissive P-7).WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-235In addition to the above listed automatic

features, there are the following RCCA withdrawal blocks:* Power range neutron flux (one-out-of-four power range)* OPAT (two-out-of-four)
  • OTAT (two-out-of-four) 2.5.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaA number of cases were analyzed assuming a range of reactivity insertion rates for both minimum andmaximum reactivity feedback conditions at various power levels for DNB and RCS overpressure considerations.

The cases presented below are representative for this event.The following assumptions were made for the analysis of the uncontrolled RCCA bank withdrawal atpower transient in order to obtain conservative results with respect to core damage:This transient was analyzed with the RTDP (Reference 1). Initial RCS pressure and temperature were assumed to be at their nominal values. An initial NSSS power of 3651 MWt, which includesall applicable uncertainties, was modeled.

MMF was also modeled.

Uncertainties in initialconditions, with the exception of power, were included in the DNBR SAL as described in theRTDP.* For reactivity coefficients, two sets were analyzed.

-Minimum reactivity feedback:

A least negative or positive value of the MTC of reactivity is assumed corresponding to the beginning of core life. A conservatively small (in absolutemagnitude) value of the Doppler coefficient is assumed.-Maximum reactivity feedback:

A conservatively large positive moderator densitycoefficient and a large (in absolute magnitude) negative Doppler coefficient are assumed.* The RT on power range neutron flux (high setpoint) was assumed to be actuated at the SAL of116.5 percent of the analyzed full power level.* The OTAT and OPAT trips included all adverse instrumentation and setpoint errors, and thedelays for the trip signal actuation were assumed at their maximum values.0 The RCCA trip insertion characteristic was based on the assumption that the highest-worth RCCAwas stuck in its fully-withdrawn position.

  • A range of reactivity insertion rates was examined.

The maximum positive reactivity insertion rate was greater than that which would be obtained from the simultaneous withdrawal of the twocontrol rod banks having the maximum combined worth at a conservative speed(48.125 inches/minute, which corresponds to 77 steps/minute).

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-236To be conservative with respect to DNB, the pressurizer sprays and relief valves were assumedoperational because they limit the reactor coolant pressure increase.

Power levels of 10, 60, and 100 percent of the assumed NSSS power were considered.

For the RCS overpressure

analysis, the preceding assumptions still apply with the following differences inorder to obtain conservative results with respect to RCS overpressurization:

This case is analyzed with the STDP. Initial RCS pressure and temperature were assumed to bewithin their respective allowable operating ranges with uncertainties applied in the conservative directions.

As was done in the DNB case, an NSSS power of 3651 MWt, which includes allapplicable uncertainties, was modeled.

TDF was also modeled.Minimum reactivity feedback conditions were modeled.The pressurizer sprays were not modeled because operation of the pressurizer spray valves wouldminimize the pressure increase during the transient.

Ranges of initial power levels (from 8 to 100 percent of the analyzed power level) and reactivity insertion rates (1 to 110 pcm/sec) were analyzed.

The RT on power range neutron flux high positive rate trip was assumed to be actuated at aconservative rate setpoint of 6.9 percent of the analyzed power level, with a conservative ratetime constant and delay time.The PSVs were modeled with a positive set pressure tolerance (2 percent) applied toconservatively increase the opening pressure.

In addition, this case includes a 1 percent setpointshift and a 1.153-second purge time delay to account for the existence of PSV water-filled loopseals, as described in Reference

2. The pressurizer PORVs were not modeled because theiroperation would minimize the pressure increase during the transient.

The MSSVs were modeled with bounding opening pressures in order to prolong the mismatchbetween core heat generation and secondary heat removal capability.

Based on its frequency of occurrence, the uncontrolled RCCA bank withdrawal at power transient isconsidered to be a Condition II event as defined by the ANS's "Nuclear Safety Criteria for the Design ofStationary Pressurized water Reactor Plants,"

ANSI N 18.2-1973.

The following items summarize theacceptance criteria associated with the analysis of this event:Pressures in the RCS and MSS must remain less than 110 percent of the respective designpressures.

With respect to peak MSS pressures, the RCCA withdrawal at power event is bounded by theLOL/TT event described in Section 2.3.1, "Loss of External Electrical Load, Turbine Trip,Inadvertent Closure of Main Steam Isolation Valves, Loss of Condenser Vacuum and OtherEvents Resulting in Turbine Trip," in which assumptions are made to conservatively calculate theWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-237MSS pressure transients.

For the LOL/TT event, the turbine trip is the initiating incident thatmaximizes the resultant power mismatch between the primary and secondary sides, and theresultant temperature and pressure transients of the MSS are always more severe for LOL/TTevents than for RCCA withdrawal at power events. Based on this, no explicit calculation ofmaximum MSS pressure is performed for this event.Fuel cladding integrity must be maintained by ensuring that the DNBR remains above the95/95 DNBR limit. In addition, it has been historical practice to assume that fuel failure willoccur if centerline melting takes place. Therefore, the analysis evaluates whether the peak linearheat generation rate exceeds the value that would cause fuel centerline melt. For the WCGS, thisis met by ensuring that the peak core average heat flux does not exceed 121 percent of theanalyzed full power level.The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the uncontrolled RCCA bank withdrawal at power acceptance criteriaare provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the RCCA bank withdrawal at power event, this is shown to be met bydemonstrating that the fuel damage criterion is satisfied.

GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.

For theRCCA bank withdrawal at power event, this is shown to be met by demonstrating that the peakRCS pressure is less than 110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.

For the RCCA bank withdrawal at power event, which results in a RT, this is shown to be met by demonstrating that the fueldamage criterion is satisfied.

The protection features presented in subsection 2.5.2.1.1 provide mitigation of the uncontrolled RCCAbank withdrawal at power transient such that the above criteria are satisfied.

WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2382.5.2.1.3 Description of Analyses and Evaluations The purpose of this analysis was to demonstrate the manner in which the protection functions described above actuate for various combinations of reactivity insertion rates and initial conditions.

Insertion rateand initial conditions determined which trip function actuated first.The uncontrolled RCCA bank withdrawal at power event was analyzed with the RETRAN computer code(Reference

3) to demonstrate the manner in which the previously described protection functions provideadequate protection from core damage. The RETRAN model simulates the core neutron kinetics, RCS,.pressurizer, pressurizer relief and safety valves, pressurizer pressure control systems, SGs, and MSSVs.The code computes pertinent plant variables, including temperatures, pressures, power level, and coreboron concentration.

For the most limiting case analyzed, a detailed DNBR evaluation using the detailedT/H digital computer code, VIPRE (Reference 4), was used to determine if the DNB design basis wasmet.An analysis to confirm that the RCS pressure safety limit is protected was performed using theLOFTRAN code (Reference 5). Similar to the RETRAN model, the LOFTRAN model simulates the coreneutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer pressure controlsystems, SGs, and MSSVs.2.5.2.1.4 ResultsFigures 2.5.2-1 through 2.5.2-3 show the transient response for a rapid uncontrolled RCCA bankwithdrawal incident (110 pcm/sec) starting from 100 percent power with minimum reactivity feedback.

The neutron flux level in the core rises rapidly while the core heat flux and coolant system temperature lag behind due to the thermal capacity of the fuel and coolant system fluid. RT on power range neutronflux (high setpoint) occurs shortly after the start of the transient prior to significant increases in the heatflux and water temperature, and the resultant DNBR remains well above the SAL value throughout thetransient.

The transient response for a slow uncontrolled RCCA bank withdrawal (I pcm/sec) from 100 percentpower with minimum feedback is shown in Figures 2.5.2-4 through 2.5.2-6.

With a lower insertion rate,the power increase rate is slower, the rate of increase of the average coolant temperature is slower, and thesystem lags and delays become less significant.

ART on OTAT occurs after a longer period of time thanfor a rapid RCCA bank withdrawal.

Again, the DNBR remains greater than the SAL value throughout thetransient.

Figure 2.5.2-7 shows the minimum DNBR as a function of reactivity insertion rate from 100 percentpower for both minimum and maximum reactivity feedback conditions.

The high neutron flux and OTATRT functions provide DNB protection over the analyzed range of reactivity insertion rates because theminimum DNBR is never less than the SAL value.Figures 2.5.2-8 and 2.5.2-9 show the minimum DNBR as a function of reactivity insertion rate for RCCAbank withdrawal incidents starting at 60 and 10 percent power, respectively.

The results are similar to the100-percent power case. However, as the initial power level is decreased, the range over which the OTATtrip is effective is increased.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-239Note that the conservative minimum DNBR approximation calculated by RETRAN for a number of the60 and 10 percent power cases did not meet the SAL DNBR value. This limit was conservatively definedto demonstrate that the DNB design basis is satisfied for analyses performed using RTDP methods.Sufficient margin is maintained between the SAL DNBR and the design limit DNBR to offset the effectsof rod bow, lower plenum flow anomaly, and plant instrumentation biases, as well as to provide flexibility in the design and operation of the plant. See Section 2.12, "Thermal and Hydraulic Design,"

for additional information.

To demonstrate that the DNB design basis was satisfied, the VIPRE code was used toperform a detailed DNBR calculation of the most limiting part-power case (reactivity insertion rate of13 pcm/sec from 10 percent power with minimum feedback).

The results confirmed that the DNB designbasis continues to be met and that sufficient DNBR margin is retained to allow for flexibility in the designand operation of the plant. To increase the amount of DNBR margin retained, the DNBR calculations performed for this event credited the following changes:A higher MMF of 376,000 gpmThimble plugs were assumed to remain installed to reduce the core bypass flow (all othernon-LOCA analyses covered the bounding scenario of having the core TPR)Finally, Figures 2.5.2-10 through 2.5.2-12 show the transient responses for the limiting RCS pressure caseindicating that the applicable limit is met.The calculated sequences of events for four cases are shown in Table 2.5.2-1; the four cases include:0 a rapid RCCA bank withdrawal (110 pcm/sec) from 100 percent power with minimum feedback,

  • a slow RCCA bank withdrawal (1 pcm/sec) from 100 percent power with minimum feedback,
  • the limiting DNB case (withdrawal rate of 13 pcm/sec from 10 percent power with minimumfeedback),
  • the limiting overpressure case (withdrawal rate of 21 pcm/sec from 74 percent power withminimum feedback).

With the reactor tripped, the plant eventually returns to a stable condition.

The plant could subsequently be cooled down further by following normal plant shutdown procedures.

The limiting results of theuncontrolled RCCA bank withdrawal at power analyses are shown in Table 2.5.2-2.For the DNB cases, the power range neutron flux and OTAT RT functions provided adequate protection over the entire range of possible reactivity insertion rates. The results show that the DNB design basis ismet and the peak linear heat generation rate is less than the limit.For the RCS overpressure cases, the power range neutron flux, OTAT, power range neutron flux highpositive rate, and high pressurizer pressure RT functions, in conjunction with the PSVs and MSSVs,provide adequate protection over the entire range of possible reactivity insertion rates. The results showedthat the peak RCS pressure remains below 110 percent of the design pressure.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-240Therefore, the results of the analysis show that an uncontrolled RCCA bank withdrawal at power does notadversely affect the core, the RCS, or the MSS.2.5.2.2 Conclusions Based on the above information, it is concluded that the analysis has adequately accounted for operation of the plant at the analyzed power level and was performed using acceptable analytical models. Thisanalysis has also demonstrated that the reactor protection and safety systems will continue to ensure thatthe specified acceptable fuel design limits and RCS pressure safety limit are not exceeded.

Based on this,it can be concluded that the plant will continue to meet the requirements of GDCs 10, 15, and 26.2.5.2.3 References

1. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.2. WCAP-12910, Rev. I-A, "Pressurizer Safety Valve Set Pressure Shift," May 1993.3. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.4. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.5. WCAP-7907-P-A, "LOFTRAN Code Description,"

April 1984.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-241Table 2.5.2-1 Time Sequence of Events -Uncontrolled RCCA Bank Withdrawal at PowerCase Event Time (sec)DNB Case Initiation of Uncontrolled RCCA Bank Withdrawal 0.0100 Percent Power, Power Range Neutron Flux -High Setpoint 1.25Minimum Feedback, Rapid ReachedRCCA Bank Withdrawal (110 pcm/sec)

Rods Begin to Drop 1.75Minimum DNBR Occurs 3.05DNB Case Initiation of Uncontrolled RCCA Bank Withdrawal 0.0100 Percent Power, OTAT Setpoint Reached 99.3Minimum Feedback, Slow RCCA Bank Rods Begin to Drop 101.6Withdrawal (1 pcm/sec)

Minimum DNBR Occurs 102.0Limiting DNB Case Initiation of Uncontrolled RCCA Bank Withdrawal 0.010 Percent Power, Minimum Power Range Neutron Flux -High Setpoint 55.18Feedback, Slow RCCA Bank ReachedWithdrawal (13 pcm/sec)Rods Begin to Drop 55.68Minimum DNBR Occurs 56.42Limiting Overpressure Case Initiation of Uncontrolled RCCA Bank Withdrawal 0.074 Percent Power, Minimum Power Range Neutron Flux High Positive Rate 12.54Feedback, Rapid RCCA Setpoint ReachedBank Withdrawal (21 pcm/sec)

Rods Begin to Drop 13.54Maximum RCS Pressure Occurs 15.50WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-242WESTIINGHOU SE NON-PROPRIETARY CLASS 3 2-242Table 2.5.2-2 Uncontrolled RCCA Bank Withdrawal at Power -Limiting ResultsLimiting value Analysis Limit CaseMinimum DNBR See notetl) See note1.) 10% power, minimum feedback, 13 pcm/sec reactivity insertion ratePeak Core Heat Flux (fraction of 1.183 1.21 10% power, minimum feedback, analyzed full power) 100 pcm/secPeak Primary System Pressure 2707.4 2750.0 74% power, minimum feedback(psia) 21 pcm/secNote:1. A detailed DNBR evaluation was performed using the VIPRE code confirming that the DNB design basis was satisfied for the limiting case.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-243000C-,0U-w00~0a,C-,00C-,0Lj~U-0a,U,C-,0 2 3 4 5 6Time (s)0 1 2 3 4 5 6 7Time (s)Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -110 pcm/secNuclear Power and Core Heat Flux Versus TimeFigure 2.5.2-1WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-244U)V)L.Dt-4co(n4Time (s)(DE-50Q)L4LU)U)a)0nI ,UL/1150-1100-1050-1000-950-..............................................................................................

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

oIv'-t0 1 2 3 4Time (s)5 6 7Figure 2.5.2-2Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -110 pcm/secPressurizer Pressure and Water Volume Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-245WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-245U.)En0 1 2 3 4 5 6 7Time (s)0 2 3 4Time (s)5 6 7Figure 2.5.2-3Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power- 110 pcm/secVessel Average Temperature and DNBR Versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2460c-0-0C-,.0.--(V0C~_)cvCa-,C..)0 20 40 60Time (s)80 100 1200 20 40 60Time (s)80 100 120Figure 2.5.2-4Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -1 pcm/secNuclear Power and Core Heat Flux Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-247U3Enco1PcnV)0 20 40 60Time (s)80 100 1200 20 40 60 80 100 120Time (s)Figure 2.5.2-5Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -1 pcm/secPressurizer Pressure and Water Volume Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-248WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-248E-0Cnco0 20 40 60Time (s)80 100 1200 20 40 60Time (s)80 100 120Figure 2.5.2-6Bank Withdrawal at Power -Minimum Reactivity Feedback100 Percent Power -1 pcm/secVessel Average Temperature and DNBR Versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-249WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-249Maximum reactivity

---- Minimum r e a c t i v i t y1.951.90.S1.85--E .._ 1.80- /E..1 .. ..........feedbackfeedback0 20 40 60 80 100Reactivity Insertion Rate (pcm/sec) 120Figure 2.5.2-7Bank Withdrawal at Power -100 Percent PowerMinimum DNBR Versus Reactivity Insertion RateNote that the minimum DNBR values presented were calculated using the RETRAN code and arerepresentative of the trends for minimum DNBR and not indicative of final calculated DNBR values. Thedetailed T/H code VIPRE was used to confirm that the limiting case (10 percent power, minimumreactivity

feedback, 13 pcm/sec reactivity insertion rate) satisfied the DNB safety design basis.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-250Maximum reactivi ty----- --M in im um r e a c t i v it y3.002.80.2.60m 2.40E 2.20-E2.00 ...feedbackfeedback0 20 40 60 80 100Reactivity Insertion Rate (pcm/sec) 120Figure 2.5.2-8Bank Withdrawal at Power -60 Percent PowerMinimum DNBR Versus Reactivity Insertion RateNote that the minimum DNBR values presented were calculated using the RETRAN code and arerepresentative of the trends for minimum DNBR and not indicative of final calculated DNBR values. Thedetailed T/H code VIPRE was used to confirm that. the limiting case (10 percent power, minimumreactivity

feedback, 13 pcm/sec reactivity insertion rate) satisfied the DNB design basis.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-251WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-25 1Maximum reactivity

M inimum r e a c t i v it y4.50-4.00-3 .5 0 "3.00E ..............feedbackfeedback0 20 40 60 80 100Reactivity Insertion Rate (pcm/sec) 120Figure 2.5.2-9Bank Withdrawal at Power -10 Percent PowerMinimum DNBR Versus Reactivity Insertion RateNote that the minimum DNBR values presented were calculated using the RETRAN code and arerepresentative of the trends for minimum DNBR and not indicative of final calculated DNBR values. Thedetailed T/H code VIPRE was used to confirm that the limiting case (10 percent power, minimumreactivity

feedback, 13 pcm/sec reactivity insertion rate) satisfied the DNB design basis.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-252.)0U-00C-)U-a.)C.U'-v (C--t0 5 10 15Time (s)200 5 10 15 20Time (s)Figure 2.5.2-10Bank Withdrawal at Power -Minimum Reactivity FeedbackLimiting Overpressure CaseNuclear Power and Core Heat Flux Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-253WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-253r'4Cl)coQ)Cl)C,,ni0 5 10 15 20Time (s)0 5 10 15 20Time (s)Figure 2.5.2-11Bank Withdrawal at Power -Minimum Reactivity FeedbackLimiting Overpressure CasePressurizer Pressure and Water Volume Versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-254WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-254EZ30.0~IL~5 10 15Time (s)200 10 15Time (s)20Figure 2.5.2-12Bank Withdrawal at Power -Minimum Reactivity FeedbackLimiting Overpressure CaseVessel Average Temperature and Peak RCS PressureWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2552.5.3 Control Rod Misoperation (USAR Section 15.4.3)2.5.3.1 Technical Evaluation 2.5.3.1.1 Introduction The RCCA misalignment events include the following:

  • One or more dropped RCCAs from the same group0 A dropped RCCA bank* A statically misaligned RCCA* Withdrawal of a single RCCAEach RCCA has a position indicator channel that displays the position of the assembly.

The displays ofassembly positions are grouped for the operator's convenience.

Fully inserted assemblies are furtherindicated by a rod at bottom signal, which actuates an alarm and a control room annunciator.

Groupdemand position is also indicated.

Full-length RCCAs are moved in preselected banks, and the banks are moved in the same preselected sequence.

Each control bank of RCCAs is divided into two groups. The rods comprising a group operatein parallel through multiplexing thyristors.

The two groups in a bank move sequentially such that the firstgroup is always within one step of the second group in the bank. A definite schedule of actuation (ordeactuation of the stationary

gripper, movable gripper, and lift coils of a mechanism) is required towithdraw the RCCA attached to the mechanism.

Because the stationary

gripper, movable gripper, and liftcoils associated with the four RCCAs of a rod group are driven in parallel, any single failure that wouldcause rod withdrawal would affect a minimum of one group. Mechanical failures are in the direction ofinsertion, or immobility.

A dropped RCCA or RCCA bank is detected by one or more of the following:

Sudden drop in the core power level as seen by the nuclear instrumentation systemAsymmetric power distribution as seen on out-of-core neutron detectors or core exitthermocouples Rod-at-bottom signalRod deviation alarmRod position indication Dropping of a full-length RCCA is assumed to be initiated by a single electrical or mechanical failure thatcauses any number and combination of rods from the same group of a given control bank to drop to thebottom of the core. The resulting negative reactivity insertion causes nuclear power to rapidly decrease.

An increase in the hot channel factor can occur due to the skewed power distribution representative of aWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-256dropped rod configuration.

For this event, it must be shown that the DNB design basis is met for thecombination of power, hot channel factor, and other system conditions that exist following a dropped rod.Misaligned assemblies are detected by:Asymmetric power distribution as seen on out-of-core neutron detectors or core exitthermocouples Rod deviation alarm* Rod position indicators For the WCGS, rod position is displayed in 6-step increments with an accuracy of +/-4 steps. Deviation ofany RCCA from its group by twice this distance (12 steps) will not cause power distributions worse thanthe design limits. The deviation alarm alerts the operator to rod deviation with respect to the groupposition in excess of 12 steps. If the rod deviation alarm is not functional, the operator is required to takeaction per the Technical Requirements Manual (or equivalent document).

If one or more rod position indicators should be out of service, detailed operating instructions shall befollowed to assure the alignment of the non-indicated RCCAs. The operator is also required to take actionper the TS.In the extremely unlikely event of simultaneous electrical failures that could result in single RCCAwithdrawal, rod deviation and rod control urgent failure would both be displayed on the plant annunciator, and the rod position indicators would indicate the relative positions of the assemblies in the bank. Theurgent failure alarm also inhibits automatic rod motion in the group in which it occurs. Withdrawal of asingle RCCA by operator action, whether deliberate or by a combination of errors, would result inactivation of the same alarm and the same visual indications.

Withdrawal of a single RCCA results in bothpositive reactivity insertion tending to increase core power, and an increase in local power density in thecore area associated with the RCCA. Automatic protection for this event is provided by the OTAT RT, butbecause the local power density increases, it is not possible in all cases to provide assurance that the coresafety limits will not be violated.

2.5.3.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe dropped RCCA, dropped RCCA bank, and statically misaligned RCCA events are classified asCondition II events (faults of moderate frequency) as defined by the ANS's "Nuclear Safety Criteria forthe Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

The single RCCAwithdrawal incident is classified as an ANS Condition III event, as discussed below.The acceptance criteria applied to this event are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the dropped RCCA, dropped RCCA bank, and statically misaligned RCCA events acceptance criteria are provided as follows.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-257GDC 10, insofar as it requires that the reactor core be designed with appropriate margin to assurethat specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition ofnormal operation, including the effects of anticipated operational occurrences (AOOs)GDC 20, insofar as it requires that the protection system be designed to initiate the reactivity control systems automatically to assure that acceptable fuel design limits are not exceeded as aresult of AOOs and to initiate automatically operation of systems and components important tosafety under accident conditions GDC 25, insofar as it requires that the protection system be designed to assure that SAFDLs arenot exceeded for any single malfunction of the reactivity control systems.The following discussion demonstrates that all applicable acceptance criteria are met for this event at theWCGS.No single electrical or mechanical failure in the rod control system could cause the accidental withdrawal of a single RCCA from the inserted bank at full-power operation.

The operator could deliberately withdraw a single RCCA in the control bank since this feature is necessary in order to retrieve anassembly should one be accidentally dropped.

The event analyzed could only occur from multiple wiringfailures or multiple deliberate operator actions and subsequent and repeated operator disregard of eventindication.

The probability of such a combination of conditions is so low that it would be acceptable forthe consequences to include slight fuel damage. Thus, consistent with the philosophy and format ofANSI N 18.2, the event is classified as a Condition Ill event. By definition "Condition II occurrences include incidents, any one of which may occur during the lifetime of a particular plant," and "shall notcause more than a small fraction of fuel elements in the reactor to be damaged."

See Tables 2.5.3-1 and 2.5.3-2 for detailed acceptance criteria and initial conditions used in the droppedRCCA/dropped RCCA bank analysis.

For the statically misaligned RCCA and single RCCA withdrawal events, see the analysis descriptions and results in Sections 2.5.3.1.3 and 2.5.3.1.4 for details of the inputsand acceptance criteria.

2.5.3.1.3 Description of Analyses and Evaluations One or More Dropped RCCAs from the Same GroupThe LOFTRAN computer code (Reference I) calculates transient system responses for the evaluation of adropped RCCA event. The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief andsafety valves, pressurizer spray, SG, and MSSVs. The code computes pertinent plant variables including temperatures, pressures, and power levels.Transient RCS statepoints (temperature,

pressure, and power) are calculated by LOFTRAN.

Nuclearmodels are used to obtain a hot-channel factor consistent with the primary-system conditions and reactorpower. By incorporating the primary conditions from the transient analysis and the hot-channel factorfrom the nuclear analysis, it is shown that the DNB design basis is met using dropped rod limit linesdeveloped with the VIPRE code (Reference 2). The transient response

analysis, nuclear peaking factorWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-258analysis, and performance of the DNB design basis confirmation are performed in accordance with theapproved methodology described in Reference 3.Dropped RCCA BankA dropped RCCA bank results in a symmetric power change in the core. Assumptions made in themethodology (Reference

3) for the dropped RCCA(s) analysis provide a bounding analysis for thedropped RCCA bank.Statically Misaligned RCCASteady-state power distributions are analyzed using the appropriate nuclear physics computer codes. Thepeaking factors are then compared to peaking factor limits developed using the VIPRE code, which arebased on meeting the DNBR design criterion.

The following cases are examined in the analysis assumingthe reactor is at full power: the worst rod withdrawn with bank D inserted at the insertion limit, the worstrod 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 with maximum predicted peaking factors.This assures a conservative FAH for the misaligned RCCA configuration.

Single RCCA Withdrawal Power distributions within the core are calculated.

The peaking factors are then used by VIPRE tocalculate the DNBR for the event. The case of the worst rod withdrawn from bank D inserted at theinsertion limit, with the reactor initially at full power, was analyzed.

This incident is assumed to occur atBOL because this condition results in a minimum MTC. This assumption maximizes the power increaseand minimizes the tendency of increased moderator temperature to flatten the power distribution.

2.5.3.1.4 Control Rod Misalignment ResultsOne or More Dropped RCCAsSingle or multiple dropped RCCAs within the same group result in a negative reactivity insertion.

Thecore is not adversely affected during this period because power is decreasing rapidly.

Either reactivity feedback or control bank withdrawal will re-establish power.For a dropped RCCA event in the automatic rod control mode, the rod control system detects the drop inpower and initiates control bank withdrawal.

Power overshoot may occur due to rod control movement, after which the control system will insert the control bank to restore nominal power. In all cases, theminimum DNBR remains above the limit value.Following a dropped rod event in manual rod control, the plant will establish a new equilibrium condition.

The equilibrium process without control system interaction is monotonic, thus removing powerovershoot as a concern, and establishing the automatic rod control mode of operation as the limiting case.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-259Dropped RCCA BankA dropped RCCA bank results in a large negative reactivity insertion.

Due to the relatively large worth ofthe dropped bank, and if the turbine load is constant, a RT may occur on low pressurizer pressure due tothe mismatch between the reactor power and the turbine power. The core is not adversely affected duringthis period because power is decreasing rapidly.

In the event a RT does not occur, the initial powerreduction from a dropped RCCA bank is large and the power return due to reactivity feedback and controlbank withdrawal is far less than seen from one or more dropped RCCAs from the same group. In eitherinstance, the minimum DNBR remains above the limit value.Following plant stabilization, the operator may manually retrieve the RCCA(s) by following applicable plant procedures.

Statically Misaligned RCCAThe most severe misalignment situations with respect to DNBR at significant power levels arise fromcases in which one RCCA is fully inserted, or where bank D is fully inserted with one RCCA fullywithdrawn.

Multiple independent alarms, including a bank insertion limit alarm, alert the operator wellbefore the postulated conditions are approached.

The bank can be inserted to its insertion limit with anyone assembly fully withdrawn without the DNBR decreasing below the limit value.The insertion limits in the TS may vary depending on a number of limiting criteria.

It is preferable, therefore, to analyze the misaligned RCCA case at full power for a control bank insertion position that isas deep as allowed by the DNBR and power peaking factor limits. The full power insertion limits oncontrol bank D must then be chosen to be above that position and will usually be dictated by othercriteria.

Detailed results will vary from cycle to cycle depending on fuel arrangements.

For this RCCA misalignment, with bank D inserted to its full-power insertion limit and one RCCA fullywithdrawn, the DNBR does not decrease below the limit value. This case is analyzed assuming the initialreactor power and RCS pressure and temperature are at their nominal values including uncertainties, butwith the increased radial peaking factor associated with the misaligned RCCA.DNB calculations have not been performed specifically for RCCAs missing from other banks. However,power shape calculations have been done as required for the RCCA ejection analysis.

Inspection of thepower shapes shows that the DNB and peak kW/ft situation is less severe than the bank D case discussed above, assuming insertion limits on the other banks are equivalent to a bank D full-in insertion limit.For RCCA misalignments with one RCCA fully inserted, the DNBR does not decrease below the limitvalue. This case is analyzed assuming the initial reactor power and RCS pressure and temperature are attheir nominal values including uncertainties, but with the increased radial peaking factor associated withthe misaligned RCCA.DNB does not occur for the RCCA misalignment

incident, and thus the ability of the primary coolant toremove heat from the fuel rod is not reduced.

The peak fuel temperature corresponds to a linear heatgeneration rate based on the radial peaking factor penalty associated with the misaligned RCCA and theWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-260design axial power distribution.

The resulting linear heat generation is well below that which would causefuel melting.Following the identification of an RCCA group misalignment condition, the operator is required to takeaction per the plant TS and applicable plant procedures.

Single RCCA Withdrawal For the single rod withdrawal event, two cases have been considered as follows:1. If the reactor is in the manual control mode, continuous withdrawal of a single RCCA results inboth an increase in core power and coolant temperature, and an increase in the local hot channelfactor in the area of the withdrawing RCCA. In terms of the overall system response, this case issimilar to those presented for the uncontrolled RCCA bank withdrawal at power event. However,the increased local power peaking in the area of the withdrawn RCCA results in lower minimumDNBRs than for the withdrawn bank cases. Depending on initial bank insertion and location ofthe withdrawn RCCA, automatic RT may not occur sufficiently fast enough to prevent theminimum DNBR from decreasing below the limit value. Evaluation of this case at the power andcoolant conditions at which the OTAT trip would be expected to trip the plant shows that anupper limit for the number of fuel rods with a DNBR less than the limit value is 5 percent of thetotal rods in the core.2. If the reactor is in the automatic control mode, the multiple failures that result in the withdrawal of a single RCCA will result in the immobility of the other RCCAs in the controlling bank. Thetransient will then proceed in the same manner as Case (1) described above.For such cases as above, a RT will ultimately ensue, although not sufficiently fast enough in all cases toprevent a minimum DNBR in the core of less than the limit value. Following RT, normal shutdownprocedures are followed.

No single failure of the RT system will negate the protection functions requiredfor the single RCCA withdrawal

accident, or adversely affect the consequences of the accident.

2.5.3.1.5 ResultsThe evaluation of the dropped rod event using the methodology in Reference 3, encompassing all possibledropped RCCA or RCCA bank worths delineated in Reference 3, concluded that the minimum DNBRremains above the SAL value for the WCGS. For all cases of any single RCCA fully inserted, or bank Dinserted to the rod insertion limit and any single RCCA in that bank fully withdrawn (staticmisalignment),

the minimum DNBR remains above the limit value for the WCGS. Therefore, the DNBdesign criterion is met and the RCCA misalignments do not result in core damage. For the case of theaccidental withdrawal of a single RCCA, with the reactor in the automatic or manual control mode andinitially operating at full power with bank D at the insertion limit, an upper bound of the number of fuelrods experiencing DNB is 5 percent of the total number of fuel rods in the core for the WCGS.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2612.5.3.2 Conclusion The analyses of control rod misalignment events have been reviewed and it has been concluded that thesewere performed using acceptable analytical models. It was further concluded that the analyses havedemonstrated that the reactor protection and safety systems will continue to ensure that the SAFDL's willnot be exceeded during normal or anticipated operational transients.

Based on this, it is concluded that theplant will continue to meet the requirements of GDCs 10, 20, and 25.2.5.3.3 References

1. WCAP-7907-P-A, "LOFTRAN Code Description,"

April 1984.2. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.3. WCAP-11394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-262Table 2.5.3-1 Non-LOCA Analysis Limits and Analysis Results for the Dropped Rod EventAnalysis ResultResult Parameter Analysis Limit Limiting CaseMinimum DNBR (RTDP, WRB-2) 1.52 > 1.52Peak Linear Heat Generation (kW/ft) 22.4(" < 22.4Peak Uniform Cladding Strain (%) 1.0 < 1.0Note:1. Corresponds to a conservative UO2 fuel melting temperature of 4700'F.Table 2.5.3-2 Summary of Initial Conditions and Computer Codes Used for the Dropped Rod EventVessel Vessel Average RCSComputer DNB Initial Power Coolant Flow Coolant Temp PressureCodes Used Correlation RTDP (%) (gpm) (OF) (psia)LOFTRAN WRB-2 Yes 100 371,000 588.4 2250.0ANC (3637 MWt -VIPRE core power)WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2632.5.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (USARSection 15.4.4)2.5.4.1 Technical Evaluation As described in USAR Section 15.4.4, the Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature event has historically been analyzed for the WCGS. If the plant is operating with one pumpout of service, there is reverse flow through the inactive loop due to the pressure difference across thereactor vessel. The CL temperature in an inactive loop is identical to the CL temperature of the activeloops (the reactor core inlet temperature).

If the reactor is operated at power, and assuming the secondary side of the SG in the inactive loop is not isolated, there is a temperature drop across the SG in the inactiveloop and, with the reverse flow, the HL temperature of the inactive loop is lower than the reactor coreinlet temperature.

Starting of an idle RCP without bringing the inactive loop HL temperature close to thecore inlet temperature would result in the injection of cold water into the core, which would cause areactivity insertion and subsequent power increase.

Because the WCGS TS (LCO 3.4.4) require all four RCS loops to be in operation while at power or instartup conditions (Modes 1 and 2), the Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature event is administratively precluded for the WCGS. Therefore, no explicit analysis of thisevent is required, and no further evaluation is necessary.

2.5.4.2 Conclusions Based on the above information, it is concluded that the Startup of an Inactive Reactor Coolant Pump atan Incorrect Temperature event is administratively precluded by the WCGS TS and no analysis of theevent is required.

2.5.5 Chemical and Volume Control System Malfunction Resulting in a Decrease in BoronConcentration in the Reactor Coolant (USAR Section 15.4.6)2.5.5.1 Technical Evaluation The specific acceptance criterion applied for the CVCS malfunction (also referred to as boron dilution) events is that adequate operator action time is available prior to a complete loss of shutdown margin. Forboron dilution events in Modes 1 through 5, there must be at least 15 minutes from operator notification (that is, first alarm) until shutdown margin is lost. For the WCGS, a boron dilution event cannot occurduring Mode 6 (Refueling) due to administrative controls that isolate the RCS from the potential sourcesof unborated water. Additionally, for conditions when no RCP is in operation, all dilution sources areisolated or under administrative control.

Hence, a boron dilution event cannot occur during Mode 5 (ColdShutdown) or Mode 4 (Hot Shutdown) once operation on the residual heat removal system (RHRS)begins. This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS (Reference 1). With shutdown margin maintained, there is no return to critical andno violation of the 95/95 DNBR limit (GDC 10), as well as no violation of the primary and secondary pressures limits (GDC 15). Furthermore, because a return to critical is precluded and fuel design limits arenot exceeded, the requirements of GDC 26 are satisfied.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-264For Modes 1 through 5, the boron dilution analysis is performed to ensure that adequate time is available from alarm to total loss of shutdown margin for the operator to identify and terminate the dilution.

Thediscussion below demonstrates that all applicable acceptance criteria are met for this event at the WCGSin operating Modes 1 through 5.2.5.5.1.1 Introduction Reactivity can be added to the core by feeding primary-grade water into the RCS via the reactor makeupportion of the CVCS. Boron dilution is a manual operation under strict administrative controls with procedures calling for a limit on the rate and duration of dilution.

A boric acid blend system is provided to allow theoperator to match the boron concentration of the reactor coolant makeup water during normal charging to theRCS boron concentration.

As discussed below, the CVCS 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.2.5.5.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe opening of the primary water makeup control valves provides makeup to the CVCS and subsequently to the RCS, which can dilute the reactor coolant.

Inadvertent dilution from this source can be readilyterminated by closing the control valve. In order for makeup water to be added to the RCS at pressure, atleast one charging pump must be running in addition to a primary makeup water pump.The limiting dilution flow path is identified as the lowest resistance flow path for an unintentional dilution.

The boron dilution analysis excludes deliberate dilution operations from considerations.

Duringintentional boron dilution operations, the plant operators are keenly aware of and continually monitor thedilution process in progress for any sign of deviation or malfunction, such that the possibility of anundetected malfunction is considered remote. This is a standard assumption in the boron dilution analysismethodology.

Thus, the limiting boron dilution flow path does not include either the normal dilute or thealternative dilute flow paths (these paths are used only for deliberate dilution operations).

The limitingboron dilution flow path is the makeup flow path of the reactor makeup water system (RMWS) used innormal boration/blend operations.

The most common causes of an inadvertent boron dilution are the opening of the primary water makeupcontrol valve and failure of the blend system, either by controller or mechanical failure.

The CVCS andthe RMWS are designed to limit, even under various postulated failure modes, the potential rate ofdilution to values that will allow sufficient time for operator response to terminate the dilution.

Aninadvertent dilution from the RMWS may be terminated by closing the primary water makeup controlvalve. All expected sources of dilution may be terminated by closing isolation valves in the CVCS. Thelost shutdown margin may be regained by the opening of isolation valves to the RWST, thus allowing theaddition of borated water to the RCS.The rate at which unborated water can be added to the RCS is limited by the design of the CVCS andRMWS. The maximum (limiting) boron dilution flow rate is 245 gpm for Modes 1 and 2 with rod controlin manual mode, and 120 gpm in Mode 1 with rod control in automatic mode. For Modes 3 through 5, themaximum boron dilution flow rate is 157.5 gpm.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-265Information on the status of the reactor coolant makeup is continually available to the operator.

Lights areprovided on the control board to indicate the operating condition of the pumps in the CVCS. Alarms areactuated to warn the operator when boric acid or makeup water flow rates deviate from preset values as aresult of system malfunction.

A CVCS malfunction is classified as an ANS Condition II event, a fault of moderate frequency as definedby the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"ANSI N 18.2-1973.

Criteria established for Condition II events are as follows:The CHF must not be exceeded.

This is met by demonstrating that the minimum DNBR does notdecrease below the limit value at any time during the transient.

Pressure in the RCS and MSS must be maintained below 110 percent of the respective designpressures.

Fuel temperature and fuel cladding strain limits must not be exceeded.

The peak linear heatgeneration rate should not exceed a value that would cause fuel centerline melt.This event is analyzed to show that there is sufficient time for mitigation of an inadvertent boron dilutionprior to a complete loss of shutdown margin. A complete loss of plant shutdown margin results in a returnof the core to a critical condition causing an increase in the RCS temperature and heat flux. This couldviolate the SAL DNBR value and challenge the fuel and fuel cladding integrity.

A complete loss of plantshutdown margin could also result in a return of the core to a critical condition causing an increase inRCS pressure.

This could challenge the pressure design limits for the RCS and/or MSS.If the shutdown margin is shown not to be lost, the condition of the plant at any point in the transient iswithin the bounds of those calculated for other Condition II transients.

By showing that the above criteriaare met for those Condition II events, it can be concluded that they are also met for the boron dilutionevent. Operator action is relied upon to preclude a complete loss of plant shutdown margin.2.5.5.1.3 Description of Analyses and Evaluations Dilution During Mode 6 -An analysis is not performed for an uncontrolled boron dilution accidentduring refueling.

In this mode, the event is prevented by administrative controls that isolate the RCS fromthe potential source of unborated water.Dilution During Mode 5 Drained -The RCS water level can be dropped to the mid-plane of the HL formaintenance work that requires the SGs to be drained.

When the water level is drained down to themid-plane of the HL from a filled and vented condition in cold shutdown, an uncontrolled boron dilutionaccident is prevented by administrative controls that isolate the RCS from the potential source ofunborated water. Consequently, an analysis is not performed in this configuration.

Dilution During Mode 5 Filled -Typically, the plant is maintained in the Cold Shutdown mode whenRCS ambient temperatures are required.

Occasionally, reduced RCS inventory may be necessary.

Mode 5can also be a transition mode to either Refueling (Mode 6) or Hot Shutdown (Mode 4). Through the cycle,the plant may enter Mode 5 if reduced temperatures are required in containment or as the result of a TSWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-266required action. The plant is maintained in Mode 5 at the beginning of each cycle for startup testing ofcertain systems.

During this mode of operation, the control banks are fully inserted.

The following conditions are assumed for an uncontrolled boron dilution during cold shutdown.

The assumed dilution flow (157.5 gpm) is the maximum flow from the RMWS assuming multiplesimultaneous failures of control valves.The active RCS water volume for the WCGS is 8639.0 ft3.This active volume assumes at leastone RCP is in operation, with the volume of the pressurizer and surge line excluded to assure thatconservative estimates are made. Additionally, because no consideration is given to mixing in thereactor vessel upper head region, the volumes for the upper head and the downcomer from the topof the CLs to the bottom of the upper head spray nozzles are also excluded.

When no RCP is in operation, all dilution sources are isolated or under administrative control.This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS.The volume control tank (VCT) high water level alarm alerts the operators that a boron dilutionmay be in progress.

This is consistent with inadvertent boron dilution event analysis methodology approved by the USNRC for the WCGS.The initial boron concentration is assumed to be 1925 ppm (parts per million) at an RCStemperature of 68°F, with all rods inserted (minus the most reactive RCCA), no xenon andshutdown margin of 1.3 percent Ak/k.The critical boron concentration is assumed to be 1800 ppm at an RCS temperature of 68°F, withall rods inserted (minus the most reactive RCCA), and no xenon. The 125 ppm change from theinitial boron concentration noted above is a conservative minimum value.Dilution During Mode 4 -In Mode 4, the plant is being taken from a short-term mode of operation, ColdShutdown (Mode 5), to a long-term mode of operation, Hot Standby (Mode 3). Typically, the plant ismaintained in the Hot Shutdown mode to achieve plant heatup before entering Mode 3. The plant ismaintained in Mode 4 at the beginning of each cycle for startup testing of certain systems.

Throughout thecycle, the plant will enter Mode 4 if reduced temperatures are required in containment or as a result of aTS required action. During this mode of operation, the control banks are fully inserted.

In Mode 4, theprimary coolant forced flow that provides mixing can be provided by either the RHRS or a RCP,depending on system pressure.

The following conditions are assumed for an uncontrolled boron dilutionduring Hot Shutdown:

The assumed dilution flow (157.5 gpm) is the maximum flow from the RMWS assumingmultiple, simultaneous failures of control valves.The active RCS water volume for the WCGS is 8639.0 ft3.This active volume assumes at leastone RCP is in operation, with the volume of the pressurizer and surge line excluded to assure thatconservative estimates are made. Additionally, because no consideration is given to mixing in theWCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-267reactor vessel upper head region, the volumes for the upper head and the downcomer from the topof the CLs to the bottom of the upper head spray nozzles are also excluded.

When no RCP is in operation, all dilution sources are isolated or under administrative control.This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS (Reference 1).The VCT high water level alarm alerts the operators that a boron dilution may be in progress.

This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS.The initial boron concentration is assumed to be 1930 ppm at an RCS temperature of 200'F,with all rods inserted (minus the most reactive RCCA), no xenon and shutdown margin of1.3 percent Ak/k.The critical boron concentration is assumed to be 1800 ppm at an RCS temperature of 200'F,with all rods inserted (minus the most reactive RCCA), and no xenon. The 130 ppm change fromthe initial boron concentration noted above is a conservative minimum value.Dilution During Mode 3 -During this mode, rod control is in manual and the rods can be eitherwithdrawn or inserted.

In Mode 3, all RCPs may not be in operation.

In an effort to balance the heat lossthrough the RCS and the heat removal of the SGs, one or more of the pumps may be off to decrease heatinput into the system. In the approach to Mode 2, the operator must manually withdraw the control rodsand may initiate a limited dilution according to shutdown margin requirements, but not simultaneously.

Ifthe shutdown or control banks are withdrawn, the dilution scenario is similar to the Mode 2 analysiswhere the failure to block the source range trip results in a RT and immediate shutdown of the reactor.The dilution scenario is more limiting if the control rods are not withdrawn and the reactor is shut downby boron to the TS minimum requirement for Mode 3. The following conditions are assumed for anuncontrolled boron dilution during hot standby:The assumed dilution flow (157.5 gpm) is the maximum flow from the RMWS assumingmultiple, simultaneous failures of control valves.The active RCS water volume for the WCGS is 8639.0 ft3.This active volume assumes at leastone RCP is in operation, with the volume of the pressurizer and surge line excluded to assure thatconservative estimates are made. Additionally, because no consideration is given to mixing in thereactor vessel upper head region, the volumes for the upper head and the downcomer from the topof the CLs to the bottom of the upper head spray nozzles are also excluded.

The VCT high water level alarm alerts the operators that a boron dilution may be in progress.

This is consistent with inadvertent boron dilution event analysis methodology approved by theUSNRC for the WCGS.The initial boron concentration is assumed to be 1645 ppm and 1940 ppm at RCS temperatures of557°F and 350'F, respectively, with all rods inserted (minus the most reactive RCCA), no xenonand shutdown margin of 1.3 percent Ak/k.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-268The critical boron concentration is assumed to be 1500 ppm and 1800 ppm at RCS temperatures of 557°F and 350'F, respectively, with all rods inserted (minus the most reactive RCCA), and noxenon. The changes between the associated initial and critical boron concentrations noted aboveare conservative minimum values.Dilution During Mode 2 -In this mode, the plant is being taken from one long-term mode ofoperation (Mode 3) to another (Mode 1). The plant is maintained in the Startup mode only for the purposeof startup testing at the beginning of each cycle. All normal actions required to change power level, eitherup or down, require operator initiation.

Assumed conditions at startup require the reactor to have available at least 1.3 percent Ak/k shutdown margin. The following conditions are assumed for an uncontrolled boron dilution during startup:* The assumed dilution flow (245 gpm) is the maximum flow from the RMWS assuming

multiple, simultaneous failures of control valves.* Conservative estimates of the minimum active RCS water volume are made by excluding thepressurizer and surge line. For the WCGS, the active RCS water volume is 9810 ft3.* The RT on source range neutron flux level alerts the operators that a boron dilution may be inprogress.
  • The initial boron concentration is assumed to be 1935 ppm, which is a conservative maximumvalue for the critical concentration at the condition of HZP, with the rods at the insertion limits,and no xenon.* The critical boron concentration following RT is assumed to be 1500 ppm, corresponding to HZP,all rods inserted (minus the most reactive RCCA), no xenon conditions.

The 435 ppm changefrom the initial condition noted above is a conservative minimum value.Mode 2 is a transitory operational mode in which the operator intentionally dilutes and withdraws controlrods to take the plant critical.

During this mode, the plant is in manual control with the operator requiredto maintain a high awareness of the plant status. For a normal approach to criticality, the operator mustmanually initiate a limited dilution and withdraw the control rods, a process that takes several hours. Priorto approaching criticality, the TS require that the predicted position of the rods is within the rod insertion limits. This ensures that the reactor did not go critical with the control rods below the insertion limits.Once critical, the power escalation must be sufficiently slow to allow the operator to manually block thesource range RT (nominally at 105 cps) after reaching permissive P-6. Too fast of a power escalation (dueto an unknown dilution) would result in reaching P-6 unexpectedly, leaving insufficient time to manuallyblock the source range RT. Failure to perform this manual action results in a RT and immediate shutdownof the reactor.However, in the event of an unplanned approach to criticality or dilution during power escalation while inMode 2, the plant status is such that minimal impact will result. The plant will slowly escalate in power toa RT on the power range neutron flux low setpoint.

After RT, more than 15 minutes is available foroperator action prior to return to criticality.

Mode 2 results are summarized in Table 2.5.5-1.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-269Dilution During Mode 1 -In this mode, the plant can be operated in either automatic or manual rodcontrol.

With the reactor in manual control and no operator action taken to terminate the transient, thepower and temperature increase will cause the reactor to reach the power range high neutron flux tripsetpoint or the OTAT trip setpoint, resulting in a RT. In this case, the boron dilution transient up to thetime of trip is essentially equivalent to an uncontrolled RCCA bank withdrawal at power. Following RT,there is at least 15 minutes prior to criticality.

This is sufficient time for the operator to determine thecause of dilution and isolate the reactor makeup water source before the available shutdown margin islost.With the reactor in automatic rod control, the power and temperature increase from the boron dilutionresults in insertion of the control rods and a decrease in the available shutdown margin. As the dilutionand rod insertion

continue, the rod insertion limit alarms (low and low-low settings) and axial fluxdifference alarm alert the operator at least 15 minutes prior to criticality that a dilution is in progress andthat the TS requirement for shutdown margin may be challenged.

This is sufficient time to determine thecause of dilution and isolate the reactor makeup water source before the available shutdown margin islost.The effective reactivity addition rate is primarily a function of the dilution rate, boron concentration, andboron worth. The following conditions are assumed for an uncontrolled boron dilution during full power:* The assumed dilution flow (245 gpm with rod control in manual mode, and 120 gpm for rodcontrol in automatic mode) is the maximum flow from the RMWS assuming

multiple, simultaneous failures of control valves.0 Conservative estimates of the minimum active RCS water volume are made by excluding thepressurizer and surge line. For the WCGS, the active RCS water volume is 9810 ft3.* The RT on power range neutron flux high or OTAT alerts the operators that a boron dilution maybe in progress.

a The initial boron concentration is assumed to be 1954 ppm, which is a conservative maximumvalue for the initial concentration at the condition of HFP, with the rods at the insertion limits, andno xenon.* The critical boron concentration following RT is assumed to be 1500 ppm, corresponding to theHZP, all rods inserted (minus the most reactive RCCA), and no xenon condition.

The 454 ppmchange from the initial condition noted above is a conservative minimum value.* A 1.3 percent Ak/k minimum shutdown margin is assumed in the analysis.

  • Bounding boron worths of -15 pcm/ppm and -5 pcm/ppm are conservatively considered.

Thelarger absolute value maximizes the reactivity insertion rate, whereas the smaller absolute valueminimizes the reactivity insertion rate thereby delaying the time to reach the RT setpoint.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2702.5.5.1.4 ResultsThe boron dilution analysis concluded that all applicable acceptance criteria are met for the WCGS. Thismeans that operator action to terminate the dilution flow within 15 minutes from operator notification (first alarm) in Modes 1, 2, 3, 4 and 5, precludes a complete loss of shutdown margin. The results of theboron dilution analysis are provided in Table 2.5.5-1.No analysis is presented for the Mode 5 drained condition or Mode 6 operation because dilution isprecluded by administrative controls.

If an unintentional dilution of boron in the RCS does occur, numerous alarms and indications areavailable to alert the operator to the condition.

The maximum reactivity addition due to the dilution isslow enough to allow the operator sufficient time to determine the cause of the addition and takecorrective action before shutdown margin is lost. The acceptance criteria as specified in Section 2.5.5.1.2 are met.2.5.5.2 Conclusion The analyses of the decrease in boron concentration in the reactor coolant due to a CVCS malfunction have been reviewed.

It is concluded that the analyses have adequately accounted for plant operation at thecurrent and proposed uprated power levels and were performed using acceptable analytical models. Also,when there is a decrease in boron concentration event, the analyses demonstrate that the reactor protection and safety systems will continue to ensure that the specified acceptable fuel design limits and theRCS and MSS pressure limits will not be exceeded.

Based on this, it is concluded that the WCGS willcontinue to meet the requirements of GDCs 10, 15, and 26.2.5.5.3 References I. Letter from James C. Stone (USNRC) to Neil S. Cams (WCNOC),

"Wolf Creek Generating Station -Amendment No. 96 to Facility Operating License No. NPF-42 (TAC No. M94112),"

March 1, 1996.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESUNGHOUSE NON-PROPRIETARY CLASS 32-271Table 2.5.5-1 CVCS Malfunction Boron Dilution Event Results -Event Alarm to Loss of ShutdownMarginAvailable OperatorAction Time LimitOperating Mode (minutes)

(minutes)

Mode 1 -Manual Rod Control 50.3 15Mode 1 -Automatic Rod Control 112.9 15Mode 2 56.0 15Mode 3 -557°F 15.8 15Mode 3 -350'F 15.6 15Mode 4 -200'F 15.8 15Mode 5 -68°F -Filled 15.7 15Mode 5 -68°F -DrainedMode 6Note:1. No analysis is presented for the Mode 5 drained condition or Mode 6 because boron dilution is precluded byadministrative controls.

WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2722.5.6 Spectrum of Rod Cluster Control Assembly Ejection Accidents (USAR Section 15.4.8)2.5.6.1 Technical Evaluation The criterion applied to ensure the core remains in a coolable geometry following a rod ejection incidentis that the average fuel pellet enthalpy at the hot spot must remain less than 200 cal/gm (360 Btu/lbm).

The use of the initial conditions presented in Table 2.5.6-1 resulted in conservative calculations of the fuelpellet enthalpy.

The results of the licensing basis analyses demonstrated that the fuel pellet enthalpy doesnot exceed 360 Btu/lbm for any of the rod ejection cases analyzed.

Overpressurization of the RCS during a rod ejection event is generically addressed in WCAP-7588, Revision 1-A (Reference 1).Another applicable acceptance criterion is that fuel melting must be limited to less than the innermost 10 percent of the fuel pellet at the hot spot, even if the average fuel pellet enthalpy at the hot spot is lessthan the limit of 360 Btu/lbm.

Conservative fuel melt temperatures of 4900'F and 4800'F were assumedfor the hot spot for the BOL and EOL cases, respectively.

These fuel melting temperatures correspond to aspecific burnup limit at the hot spot. The peak UO2 burnup at the hot spot is based on the assembly withthe maximum post-ejection FQ, which is typically a fresh fuel assembly.

Therefore, the fuel meltingtemperatures represent bounding values for the assumed UO2 burnup at the hot spot. The maximumburnup at the hot spot at BOL and EOL is confirmed to be below these values as part of the reloadprocess.

This assumption does not affect the maximum licensed fuel burnup limit. The results of thelicensing basis rod ejection analyses demonstrated that the amount of fuel melting was limited to less than10 percent of the fuel pellet at the hot spot for each of the rod ejection cases.2.5.6.1.1 Introduction This accident is defined as a mechanical failure of a CRDM pressure housing resulting in the ejection ofthe RCCA and drive shaft. The consequence of this mechanical failure is a rapid, positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.The resultant core thermal power excursion is limited by the Doppler reactivity effect of the increased fuel temperature and terminated by RT actuated by high nuclear power signals.A failure of a CRDM housing sufficient to allow a control rod to be rapidly ejected from the core is notconsidered credible for the following reasons:* Each full-length CRDM housing is completely assembled and shop tested at 4100 psig.* The mechanism housings are individually hydrotested after they are attached to the head adaptersin the reactor vessel head and checked during the hydrotest of the completed RCS.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS32-273Stress levels in the mechanism are not affected by anticipated system transients at power or by thethermal movement of the coolant loops. Moments induced by the design earthquake can beaccepted within the allowable primary working stress ranges specified in the American Society ofMechanical Engineers Boiler and Pressure Vessel Code (ASME B&PV),Section III, for Class Icomponents.

The latch mechanism housing and rod travel housing are each a single length of forged type-304stainless steel. This material exhibits excellent notch toughness at all temperatures that will beencountered.

A significant amount of margin of strength in the elastic range, together with the large energy absorption capability in the plastic range, gives additional assurance that the gross failure of the housing will notoccur. The joints between the latch mechanism housing and rod housing are threaded joints reinforced bycanopy-type rod welds.In general, the reactor is operated with the RCCAs inserted only far enough to permit load follow.Reactivity changes caused by the core depletion are compensated by boron changes.

Furthermore, thelocation and grouping of control rod banks are selected during the nuclear design to lessen the severity ofan RCCA ejection accident.

Therefore, if an RCCA is ejected from its normal position during full-power operation, only a minor reactivity excursion, at worst, could be expected to occur. The position of all ofthe RCCAs is continuously indicated in the control room. An alarm will occur if a bank of RCCAsapproaches its insertion limit or if one control rod assembly deviates from its bank. There are low andlow-low level insertion alarm circuits for each bank. The control rod position monitoring and alarmsystems are described in Reference 1.2.5.6.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput parameters for the analysis were conservatively selected on the basis of values calculated for thistype of core. The most important parameters are discussed below. Table 2.5.6-1 presents the parameters used in this analysis.

Ejected Rod Worths and Hot Channel FactorsThe values for the ejected rod worths and hot channel factors were calculated using either 3-D staticmethods or a synthesis of I-D and 2-D calculations.

Standard nuclear design codes were used in theanalysis.

No credit was taken for the flux-flattening effects of reactivity feedback.

The calculation wasperformed for the maximum allowed bank insertion at a given power level, as determined by the rodinsertion limits. The analysis assumed adverse xenon distributions to provide worst-case results.Appropriate margins were added to the ejected rod worth and hot channel factors to account for anycalculational uncertainties.

Delayed Neutron Fraction, NeffCalculations of the effective delayed neutron fraction (I3eff) typically yield values of approximately 0.75 percent at BOL and 0.40 percent at EOL. The ejected rod accident is sensitive to I3eff if the ejectedWCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-274rod worth is equal to or greater than I3erf, as in the zero-power transients.

In order to allow for future fuelcycle flexibility, conservative estimates of 3elf of 0.49 percent at beginning of cycle and 0.44 percent atend of cycle were used in the analysis.

Reactivity Weighting FactorThe largest temperature rises, and therefore the largest reactivity feedbacks, occur in channels where thepower is higher than average.

Since the weight of a region is dependent on flux, these regions have highweights.

This means that the reactivity feedback is larger than that indicated by a simple channel analysis.

Physics calculations have been carried out for temperature changes with a flat temperature distribution, and with a large number of axial and radial temperature distributions.

Reactivity changes were compared and effective weighting factors determined.

These weighting factorstake the form of multipliers which, when applied to single-channel feedbacks, correct them to effective whole-core feedbacks for the appropriate flux shape. In this analysis, a I -D (axial) spatial kinetics methodwas employed.

Therefore, axial weighting is not necessary if the initial condition is made to match theejected rod configuration.

In addition, no weighting was applied to the moderator feedback.

Aconservative radial weighting factor was applied to the transient fuel temperature to obtain an effective fuel temperature as a function of time accounting for the missing spatial dimension.

These weighting factors have also been shown to be conservative compared to 3-D analysis.

Moderator and Doppler Coefficient The MTC and the DTC are combined and input as an isothermal temperature coefficient (ITC). The ITCsthat were modeled are +7.695 pcm/°F at zero-power nominal Ta,,g and +5.247 pcm/°F at full-power Tavgfor the BOL cases. These are very conservative values that easily bound the BOL moderator temperature limit of +5 pcm/°F. For the EOL cases, the applicable zero-power ITC was -16.817 pcm/°F and the full-power MTC was -22.920 pcm/°F.The Doppler reactivity defect as a function of power level was adjusted in the nuclear code to aconservative design value using a Doppler weighting factor of 1.0. The Doppler weighting factor wasincreased under accident conditions, as discussed above.Heat Transfer DataThe FACTRAN code (Reference 2), which contains standard curves of thermal conductivity versus fueltemperature, is used to determine the hot spot transient.

During the transient, the peak centerline fueltemperature is nearly independent of the gap conductance.

The cladding temperature is, however, stronglydependent on the gap conductance and is highest for high gap conductance.

For conservatism, a lowinitial gap heat transfer coefficient was used at the beginning of the transient to maximize the initial fueltemperature and a high gap heat transfer coefficient value of 10,000 Btu/hr-ft 2 was used for the remainder of the transient to maximize the cladding temperature.

This high gap heat transfer coefficient corresponds to a negligible gap resistance, and a further increase would have essentially no effect on the rate of heattransfer.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-275Coolant Mass Flow RatesWhen the core is operating at full power, all four RCPs are always operational.

For zero-power conditions, the system was conservatively assumed to be operating with two pumps. The principal effectof operating at reduced flow is to reduce the film boiling heat transfer coefficient.

This resulted in higherPCTs, but did not affect the peak centerline fuel temperature.

Reduced flow also lowers the CHF.However, since DNB was always assumed at the hot spot, and since the heat flux rose very rapidly duringthe transient, this produced only second-order changes in the cladding and centerline fuel temperatures.

Trip Reactivity Insertion The trip reactivity insertion was assumed to be 4.0 percent Ak from HFP conditions and 2.0 percent Akfrom HZP conditions, including the effect of one stuck RCCA. These values were also reduced by theejected rod reactivity.

The shutdown reactivity was simulated by dropping a rod of the required worth intothe core. The start of rod motion occurred 0.5 seconds after reaching the power range high neutron fluxtrip setpoint.

It was assumed that insertion to dashpot did not occur until 2.7 seconds after the rods beganto fall. The time delay to full insertion combined with the 0.5 second trip delay conservatively delayedinsertion of shutdown reactivity into the core.Due to the extremely low probability of an RCCA ejection

accident, this event is classified as aCondition IV event as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

As such, some fuel damage is considered anacceptable consequence.

The real physical limits of this accident are that the rod ejection event and any consequential damage toeither the core or the RCS must not prevent long-term core cooling.

More specific and restrictive criteriaare applied to ensure that there is no fuel dispersal in the coolant and that gross lattice distortion or severeshock waves do not occur. Based on experimental data, Reference 1 concludes that the acceptance criteriato be applied for an RCCA ejection are:Average fuel pellet enthalpy at the hot spot must remain below 200 cal/gm for irradiated fuel.This bounds non-irradiated fuel, which has a slightly higher enthalpy limit.Peak reactor coolant pressure must be less than that which could cause RCS stresses to exceed thefaulted-condition stress limits (Note: the peak pressure aspects of the rod ejection transient areaddressed generically in Reference 1).Fuel melting is limited to less than the innermost 10 percent of the pellet volume at the hot spoteven if the average fuel pellet enthalpy at the hot spot is below the 200 cal/gm fuel enthalpy limit.2.5.6.1.3 Description of Analyses and Evaluations This section describes the models used in the analysis of the rod ejection accident.

Only the initial fewseconds of the power transient are discussed since the long-term considerations are the same as those for asmall LOCA.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-276The calculation of the RCCA ejection transient was performed in two stages: first an average core channelcalculation, and then a hot spot calculation.

The average core calculation used spatial neutron-kinetics methods to determine the average power generation with time, including the various total core feedbackeffects; that is, Doppler reactivity and moderator reactivity.

Enthalpy and temperature transients at the hotspot were then determined by multiplying the average core energy generation by the hot channel factorand performing a fuel rod transient heat transfer calculation.

The power distribution calculated withoutfeedback was conservatively assumed to continue throughout the transient.

A detailed discussion of themethod of analysis can be found in Reference 1.Average CoreThe spatial-kinetics computer code TWINKLE (Reference

3) was used for the average core transient analysis.

This code solves the two-group neutron diffusion theory kinetic equation in one, two, or threespatial dimensions (rectangular coordinates) for six delayed neutron groups and up to 2000 spatial points.The computer code includes a detailed, multi-region, transient fuel-clad-coolant heat transfer model forcalculation of pointwise Doppler and moderator feedback effects.

This analysis used the code as al-D axial kinetics code since it allows a more realistic representation of the spatial effects of axialmoderator feedback and RCCA movement.

However, since the radial dimension was missing, it was stillnecessary to employ very conservative methods (described below) of calculating the ejected rod worthand hot channel factor.Hot Spot AnalysisIn the hot spot analysis, the initial heat flux is equal to the nominal heat flux times the design hot channelfactor. During the transient, the heat flux hot channel factor is linearly increased to the transient value in0.1 second, the time for full ejection of the rod. Therefore, the assumption is made that the hot spot beforeand after ejection are coincident.

This is very conservative since the peak after ejection will occur in oradjacent to the assembly with the ejected rod, and prior to ejection the power in this region willnecessarily be depressed.

The average core energy addition, calculated as described above, was multiplied by the appropriate hotchannel factors.

The hot spot analysis used the detailed fuel and cladding transient heat transfer computercode FACTRAN (Reference 2). This computer code calculates the transient temperature distribution in across section of a metal-clad U02 fuel rod, and the heat flux at the surface of the rod, using the nuclearpower versus time and local coolant conditions as input. The zirconium-water reaction is explicitly represented, and all material properties are represented as functions of temperature.

A parabolic radialpower distribution was assumed within the fuel rod.FACTRAN uses the Dittus-Boelter or Jens-Lottes correlation to determine the film heat transfer beforeDNB, and the Bishop-Sandberg-Tong correlation to determine the film boiling coefficient after DNB. TheBishop-Sandberg-Tong correlation was conservatively used assuming zero bulk fluid quality.

The DNBheat flux was not calculated.

Instead, the code was forced into DNB by specifying a conservative DNBheat flux. The gap heat transfer coefficient could be calculated by the code. However, it was adjusted toforce the full-power, steady-state temperature distribution to agree with fuel heat transfer design codes.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-277Reactor Protection The protection for this accident, as explicitly modeled in the analysis, is provided by the power rangeneutron flux trip (high and low settings).

The power range high neutron flux positive rate tripcomplements the high flux trip function (high and low settings) to ensure that the criteria are met for rodejection from partial power.2.5.6.1.4 ResultsThe results of the analyses performed for the rod ejection event, which cover BOL and EOL conditions atHFP and HZP for the WCGS,. are discussed below.Beginning of Cycle, Zero PowerThe worst ejected rod worth and hot channel factor were conservatively calculated to be 0.78 percent Akand 13.0, respectively.

The peak hot spot average fuel pellet enthalpy reached 254.7 Btu/lbm(141.5 cal/gm).

The peak fuel centerline temperature never reached the BOL melt temperature of 4900'F.Therefore, no fuel melting is predicted.

Beginning of Cycle, Full PowerControl bank D was assumed to be inserted to its insertion limit. The worst ejected rod worth and hotchannel factor were conservatively calculated to be 0.23 percent Ak and 6.6, respectively.

The peak hotspot average fuel pellet enthalpy reached 317.6 Btu/lbm (176.4 cal/gm).

The peak fuel centerline temperature reached the BOL melt temperature of 4900'F. However, fuel melting remained well belowthe limiting criterion of 10 percent of total pellet volume at the hot spot.End of Cycle, Zero PowerThe worst ejected rod worth and hot channel factor were conservatively calculated to be 0.86 percent Akand 21.0, respectively.

The peak hot spot average fuel pellet enthalpy reached 261.4 Btu/lbm(145.2 cal/gm).

The peak fuel centerline temperature never reached the EOL melt temperature of 4800'F.Therefore, no fuel melting is predicted.

End of Cycle, Full PowerControl bank D was assumed to be inserted to its insertion limit. The ejected rod worth and hot channelfactors were conservatively calculated to be 0.25 percent Ak and 7.1, respectively.

The peak hot spotaverage fuel pellet enthalpy reached 305.4 Btu/lbm (169.7 cal/gm).

The peak fuel centerline temperature reached the EOL melting temperature of 4800'F. However, fuel melting remained well below the limitingcriterion of 10 percent of total pellet volume at the hot spot.A summary of the parameters used in the rod ejection

analyses, and the analyses
results, are presented inTable 2.5.6-1.

The sequence of events for all four cases is presented in Table 2.5.6-2.

Figure 2.5.6-1shows the results for the BOL/HZP case and Figure 2.5.6-2 shows the BOL/HFP plot results.

TheEOL/HZP and EOL/HFP results are presented in Figures 2.5.6-3 and 2.5.6-4, respectively.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-278A detailed calculation of the pressure surge for an ejected rod worth of 1 dollar at BOL HFP indicates thatthe peak pressure did not exceed that which would cause the RPV stress to exceed the faulted condition stress limits (Reference 1). Since the severity of the present analysis did not exceed the worst-case

analysis, the accident for this plant will not result in an excessive pressure rise or further adverse effectson the RCS.2.5.6.2 Conclusion Despite the conservative assumptions, the analyses indicate that the described fuel and cladding limitswere not exceeded.

It is concluded that there is no danger of sudden fuel dispersal into the coolant.

Sincethe peak pressure did not exceed that which would cause stresses to exceed the faulted condition stresslimits, it is concluded that there is no danger of further consequential damage to the RCS. Genericanalyses demonstrated that the fission product release as a result of fuel rods entering DNB was limited toless than 10 percent of the fuel rods in the core.The results and conclusions of the analyses performed for the rupture of a CRDM housing RCCA ejectionsupport operation up to the analyzed reactor core power of 3637 MWt.Based on the review of the analyses of the rod ejection

accident, it was concluded that the analyses haveadequately accounted for plant operation at the stated power level and were performed using acceptable analytical models. It is further concluded that appropriate reactor protection and safety systems willprevent postulated reactivity accidents that could result in damage to the RCPB greater than limited localyielding, or cause sufficient damage that would significantly impair the capability to cool the core.2.5.6.3 References
1. WCAP-7588, Revision 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Special Kinetics Methods,"

January 1975.2. WCAP-7908-A, "FACTRAN

-A FORTRAN IV Code for Thermal Transients in a UO2 FuelRod," December 1989.3. WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non-Proprietary),

"TWINKLE

-A Multi-Dimensional Neutron Kinetics Computer Code," January 1975.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-279Table 2.5.6-1 Selected Input and Results of the Limiting RCCA Ejection AnalysesInput BOC BOC EOC EOCInitial Reactor Core Power Level (MWt) 3637 0 3637 0Ejected Rod Worth (%Ak) 0.23 0.78 0.25 0.86Delayed Neutron Fraction

(%) 0.49 0.49 0.44 0.44Doppler Reactivity Weighting 1.433 2.309 1.499 3.078Trip Reactivity

(%Ak) 4.0 2.0 4.0 2.0FQ Before Rod Ejection (fraction) 2.50 -- 2.50 --FQ After Rod Ejection (fraction) 6.6 13.0 7.1 21.0Number of Operational Pumps 4 2 4 2Results BOC BOC EOC EOCMaximum Fuel Pellet Average Temperature (0F) 4041 3357 3911 3432Maximum Fuel Centerline Temperature

(°F) 4965 3867 4864 3869Maximum Cladding Average Temperature

(°F) 2270 2498 2191 2626Maximum Fuel Stored Energy (cal/gm) 176.4 141.5 169.7 145.2Maximum Fuel Melt at the Hot Spot (%) 4.62 0.00 3.96 0.00WCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-280Table 2.5.6-2 Time Sequence of Events -RCCA EjectionTime (seconds)

Event BOL HFP EOL HFPInitiation of Rod Ejection 0.0 0.0Power Range Neutron Flux Setpoint Reached 0.05 0.04Peak Nuclear Power Occurs 0.13 0.14Rods Begin to Fall 0.55 0.54Peak Fuel Average Temperature Occurs 2.22 2.33PCT Occurs 2.28 2.37Peak Heat Flux Occurs 2.30 2.38Peak Fuel Centerline Temperature Occurs 4.00 4.09BOL HZP EOL HZPInitiation of Rod Ejection 0.0 0.0Power Range Neutron Flux Setpoint Reached 0.24 0.19Peak Nuclear Power Occurs 0.28 0.22Rods Begin to Fall 0.74 0.69PCT Occurs 2.18 1.56Peak Heat Flux Occurs 2.18 1.57Peak Fuel Average Temperature Occurs 2.38 1.82Peak Fuel Centerline Temperature Occurs 3.09 2.87WCAP-l 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-281WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-28 1-,00U--(3-)0,0v0_201'15-10-5 .................I. ..........U024Time6(Seconds)

I810FuelFuelCladMelt5000----4 00 .....=3'/2000 .E1000 ......i,Center I i neAve rageOutering = 4900 DegreesF0 2 4 6 8Time (Seconds) 10Figure 2.5.6-1 Rod Ejection

-BOL/HZPWCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-282E000 2 4 6 8Time (Seconds) 10FuelFuelCladMeltCenter I i neAve rageOutering = 4900 DegreesF~r'J'Jv.Or)(DCIa,LEa,C-U-_4000-3000-2000-7 -------..... /' ..... ....... L...................

.................-.....

S.......

.... .............

..........

.. ... ... .I nnA.,024 6Time (Seconds) 810Figure 2.5.6-2 Rod Ejection

-BOL/HFPWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-283WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-28300U..-0O320"15-10-5-A --I I -I I I I I I I I IU024T 6Time (Seconds) 810Fuel Center IineFuel AverageClad OuterMel t ing = 4800 Degrees FU-C,)L.C,)L..=30L..~I)0~EH-U-0 2 4 6 8 10Time (Seconds)

Figure 2.5.6-3 Rod Ejection

-EOL/HZPWCAP- 17658-NP August 2013WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-284WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-284EC0C>w_0)C)0 2 4 6 8Time (Seconds) 10Fuel--- -Fuel-~CladMelt5000:4000".(3)r/CDC,,.--- 3000"/2000-Dc-- -Center I i neAverageOutering = 4800 DegreesF0 2 4 6 8Time (Seconds) 10Figure 2.5.6-4 Rod Ejection

-EOL/HFPWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2852.6 INCREASE IN REACTOR COOLANT INVENTORY 2.6.1 Inadvertent Operation of the Emergency Core Cooling System During PowerOperation (USAR Section 15.5.1)2.6.1.1 Technical Evaluation 2.6.1.1.1 Introduction An inadvertent actuation of the ECCS at power event results in an increase in RCS inventory, leading tothe potential filling of the pressurizer.

Operator error or a spurious electrical actuating signal could causethe event.Following the actuation signal, the SI system is actuated, which results in borated water being pumpedinto the CL of each RCS loop. Normally, an SI actuation signal results in an immediate and automatic RT,which in turn generates a turbine trip. However, even without an immediate RT, the reactor willexperience a negative reactivity excursion as a result of the borated water being injected.

This negativereactivity results in a decrease in reactor power.In manual rod control, the primary-to-secondary system power mismatch causes a decrease in coolanttemperature and a contraction of the reactor coolant.

Assuming an immediate RT signal is not received, the RCS responds with a decrease in pressurizer pressure and water level, and the turbine load willdecrease because of reduced steam pressure once the turbine throttle valves are fully open. The decreasein RCS pressure results in an increase in SI flow because of the SI pump performance characteristics.

In automatic rod control, RCCA withdrawal may compensate for the above effects as the control systemresponds to maintain programmed Tavg. Once the rods have been fully withdrawn, the event continues asdescribed for operation in manual rod control.The Inadvertent ECCS actuation at power event is performed to demonstrate that sufficient time isavailable for the appropriate operator actions to be taken to preclude a pressurizer water-solid condition.

2.6.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The following inputs and assumptions were applied in the analysis of the Inadvertent ECCS event:The initial NSSS power is 3651 MWt, which includes all applicable uncertainties.

A full power T,,,g range of 570.7°F to 588.4°F was considered in the analysis.

The limiting initialTavg is 566.7'F, which corresponds to the low nominal full power Tavg minus uncertainties (including bias). The lower initial temperature corresponds to a higher reactor coolant mass,which leads to a more severe pressurizer water volume transient.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-286The initial Tfred is 448.6'F, which corresponds to the high end of the full power Tfeed range(400.0°F to 448.6°F).

The initial pressurizer water level is 46 percent span, which is the nominal pressurizer water levelof 41 percent span at the low full power Tavg of 570.7°F plus 5 percent span uncertainty.

The initial pressurizer pressure is 2215 psia, which is the nominal value of 2250 psia minus 35 psiuncertainty.

A lower initial RCS pressure is conservative because it allows higher SI flows to beinjected into the RCS.The pressurizer proportional heaters and pressurizer sprays were modeled to function as-designed because their operation generates a more limiting condition with respect to filling the pressurizer.

Because an SI signal causes the pressurizer backup heaters to be shed from their electric powersupply, and they are not loaded onto another power supply automatically or manually untilletdown flow is re-established, the backup heaters were not modeled.* A maximum SGTP level of 10 percent was modeled.The total flow initially injected to the RCS corresponds to maximum flow from two centrifugal charging pumps (CCPs) and one normal charging pump; this total flow is reflective of SI flow tothe CLs plus RCP seal injection flow.An immediate RT on the SI actuation signal and a turbine trip derived from the RT were modeledbecause these limit the primary-to-secondary heat transfer rate, thus minimizing the magnitude ofthe initial reactor coolant shrinkage.

Within 6 minutes from event initiation, the plant operators are assumed to initiate actions tocontrol RCS (CL) temperature to 557°F. This is conservatively modeled in the analysis byopening one of the four SG ARVs at 6 minutes after event initiation to control Too1d to atemperature of 557°F.At 8 minutes after event initiation, operator action to terminate SI flow to the CLs was credited.

After the SI flow is terminated, the only source of flow injection to the RCS is maximum RCPseal injection flow from one CCP.At 29.5 minutes after event initiation, operator action to re-establish letdown flow was credited.

This was conservatively modeled by terminating the remaining flow injection to the RCS; noRCS inventory reduction was modeled in the analysis.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-287Acceptance CriteriaBased on the expected frequency of occurrence, the Inadvertent ECCS event is considered to be aCondition II event as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

The following items summarize the acceptance criteria associated with the analysis of this event:Fuel cladding integrity is maintained by ensuring that the minimum DNBR remains above the95/95 DNBR SAL.Based on historical precedence, the Inadvertent ECCS event does not lead to a serious challenge of the DNB design basis. The conditions do not approach the core thermal DNB limits, as thecore power, RCS pressure and RCS temperatures remain relatively unchanged.

Therefore, theDNBR typically increases and does not approach the DNBR SAL following event initiation.

Assuch, no explicit analysis of the event was performed to calculate a minimum DNBR value.* Pressures in the RCS and MSS are maintained below 110 percent of the design pressures.

With respect to the overpressure evaluation, the Inadvertent ECCS event is bounded by theLOL/TT event, discussed in Section 2.3.1, in which assumptions are made to conservatively maximize the RCS and MSS pressure transients.

For the Inadvertent ECCS event, turbine tripoccurs following RT, whereas for the LOL/TT event, the turbine trip is the initiating fault.Therefore, the primary-to-secondary power mismatch and resultant RCS and MSS heatup andpressurization transients are always more severe for the LOL/TT event. For this reason, it is notnecessary to calculate the maximum RCS or MSS pressures for the Inadvertent ECCS event.An incident of moderate frequency does not generate a more serious plant condition without otherfaults occurring independently.

The major concern from an Inadvertent ECCS event is that associated with pressurizer filling.The pressurizer water volume increases for this event as a result of the flow injected to the RCS.This event is analyzed to demonstrate that sufficient time is available for the appropriate operatoractions to be taken to preclude a pressurizer water-solid condition.

The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the Inadvertent ECCS acceptance criteria are provided as follows.* GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the Inadvertent ECCS event, this is shown to be met by demonstrating that thefuel cladding integrity is maintained.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-288GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.

For theInadvertent ECCS event, this is shown to be met by demonstrating that the peak RCS pressure isless than 110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions suchas stuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.

For the Inadvertent ECCS event,which results in a RT, this is shown to be met by demonstrating that the fuel cladding integrity ismaintained with a trip reactivity that accounts for the most reactive rod stuck out of the core.2.6.1.1.3 Description of Analyses and Evaluations The Inadvertent ECCS event was analyzed using the RETRAN computer code (Reference 1). TheRETRAN model simulates the RCS, neutron kinetics, pressurizer, pressurizer relief and safety valves,pressurizer

heaters, pressurizer spray, SI system, SGs, FW system, and MSSVs. The code computespertinent plant variables including nuclear power, reactor coolant average temperature, RCS pressure, pressurizer water volume, and SG pressure.

2.6.1.1.4 ResultsThe calculated sequence of events for the limiting Inadvertent ECCS case is presented in Table 2.6.1-1and transient plots of the significant plant parameters are provided in Figures 2.6.1-1 through 2.6.1-3.

RToccurs at the event initiation followed by a rapid cooldown of the RCS. The initial coolant contraction results in a short-term reduction in pressurizer pressure and water level. The combination of the RCSheatup, due to residual RCS heat generation, and ECCS injected flow causes the pressure and leveltransients to rapidly turn around. The RCS heatup continues until a source of cooling is established, firstvia the automatic opening of the lowest-setting MSSV of each loop, and then via the single ARV that wasmodeled at 6 minutes after event initiation to simulate the plant operators taking action to control theRCS (CL) temperature to 557°F. The pressurizer water level increases rapidly until SI flow to the CLs isterminated at 8 minutes as a result of crediting operator action, after which the pressurizer level continues to increase at a much slower rate until letdown flow is assumed to be re-established (via operator action)at 29.5 minutes.

The results of the analysis show that the pressurizer does not reach a water-solid condition provided that the plant operators initiate the required operator actions within the assumed timelimits.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2892.6.1.2 Conclusion Based on the above information, it is concluded that the Inadvertent ECCS event will not progress into amore serious plant condition.

Thus, all applicable event acceptance criteria are satisfied.

It has beendemonstrated that the reactor protection and safety systems ensure that the specified acceptable fueldesign limits are met and the RCPB pressure limits will not be exceeded as a result of the Inadvertent ECCS event. Based on this, the plant will continue to meet the requirements of GDCs 10, 15, and 26.2.6.1.3 References

1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-290Table 2.6.1-1 Time Sequence of Events -Inadvertent ECCSTimeEvent Seconds MinutesInadvertent ECCS/SI Signal Actuation 0.0 0.0RT due to SI SignalTurbine Trip from RTLetdown Isolation On Each Loop, the MSSV with the Lowest Setting Opens 236.4 3.9One ARV Begins to Open (1st Operator Action) 360.0 6.0All MSSVs Closed 373.5 6.2SI Flow to CLs Terminated (2nd Operator Action.)

480.0 8.0Letdown Flow Re-established (3rd Operator Action) 1770.0 29.5Maximum Pressurizer Water Volume (1786.5 ft3) Reached 2450.0 40.8WCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-29113:0C-)0.80.6-0.4-0.2-0590mo57CD5805D-570CD<2 560U)CD550I [ [- " I -t- I .. ...-I. .1120002400 3600Time (seconds) 48006000I I II I IIII I I I I I I012001 12400 3600Time (seconds) 448006000Figure 2.6.1-1 Inadvertent ECCS -Nuclear Power and Tavg versus TimeWCAP-1 7658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-29224002300-o" 2200-cnG-)2100-Cn2000I I II t II I II I I1 QI (1012002400 3600Time (seconds) 48006E0CIO0-18001600140012001000/000000800600J012002400 3600lime (seconds) 448006Figure 2.6.1-2 Inadvertent ECCS -Pressurizer Pressure and Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-293WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-293I3:0CDU-)00.80,6-0.4-0.2-012002400 3600Time (seconds) 48006000500400300-200100-I-~~~ I II0012002400Ti roe3600(seconds) 48006000Figure 2.6.1-3Inadvertent ECCS -Total Steam Flow and Total Flow Injected to the RCSversus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-2942.6.2 Chemical and Volume Control System Malfunction that Increases Reactor CoolantInventory (USAR Section 15.5.2)2.6.2.1 Technical Evaluation 2.6.2.1.1 Introduction Increases in reactor coolant inventory caused by a malfunction of the CVCS may be postulated to resultfrom operator error or a false electrical signal. The transients examined in this section are characterized by increasing pressurizer level, increasing pressurizer

pressure, and maintaining a constant boronconcentration.

The transients analyzed in this section are done to demonstrate that there is adequate timefor the operator to take corrective action to prevent filling the pressurizer.

An increase in reactor coolantinventory, which results from the addition of cold, unborated water to the RCS, is analyzed inSection 2.5.5, "Chemical and Volume Control System Malfunction That Results in a Decrease in BoronConcentration in the Reactor Coolant (USAR Section 15.4.6)."

The most limiting case occurs if the charging system is in automatic control and the pressurizer levelchannel being used for charging control fails in a low direction.

This causes the maximum charging flowto be delivered to the RCS and letdown flow to be isolated.

The worst single failure for this event is asecond pressurizer level channel failing in an as-is condition or a low condition.

This defeats the RT ontwo-out-of-three high pressurizer level channels.

To prevent filling the pressurizer the operator must berelied upon to terminate charging flow.2.6.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaInput Parameters and Assumptions The following inputs and assumptions were applied in the analysis of the CVCS Malfunction event:The initial NSSS power is 3651 MWt, which includes all applicable uncertainties.

A full power Tavg range of 570.7'F to 588.4'F was considered in the analysis.

The limiting initialTavg is 564.2'F, which corresponds to the low nominal full power Tavg minus uncertainties (including bias). The lower initial temperature corresponds to a higher reactor coolant mass,which leads to a more severe pressurizer water volume transient.

The initial Tfeed is 448.6°F, which corresponds to the high end of the full power Tfeed range(400.0°F to 448.6'F).

The initial pressurizer water level is 46 percent level span, which is the nominal pressurizer waterlevel of 41 percent span at the low full power Tavg of 570.7°F plus 5 percent span uncertainty.

The initial pressurizer pressure is 2200 psia, which is the nominal value of 2250 psia minus 50 psiuncertainty.

A lower initial RCS pressure is conservative because it allows higher charging flowsto be injected into the RCS.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-295* The pressurizer heaters are modeled to function because their operation generates a more limitingcondition with respect to filling the pressurizer.

  • Cases were analyzed both with and without automatic pressurizer spray modeled.* A maximum SGTP level of 10 percent was modeled.a The flow injected to the RCS corresponds to maximum flow from one CCP.* No RT at event initiation.
  • Cases were analyzed with both maximum and minimum reactivity feedback conditions.

0 Cases were analyzed both with and without automatic rod control.Acceptance CriteriaBased on the frequency of occurrence, the CVCS Malfunction event is considered to be a Condition IIevent as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized WaterReactor Plants,"

ANSI N 18.2-1973.

The following items summarize the acceptance criteria associated with the analysis of this event:Fuel cladding integrity is maintained by ensuring that the minimum DNBR remains above the95/95 DNBR SAL.Based on historical precedence, the CVCS Malfunction event does not lead to a serious challenge of the DNB design basis. The conditions do not approach the core thermal DNB limits, as thecore power, RCS pressure, and RCS temperatures remain relatively unchanged.

Therefore, theDNBR typically increases and does not approach the DNBR SAL following event initiation.

Assuch, no explicit analysis of the event was performed to calculate a minimum DNBR value.Pressures in the RCS and MSS are maintained below 110 percent of the design pressures.

With respect to the overpressure evaluation, the CVCS Malfunction event is bounded by theLOL/TT event, discussed in Section 2.3.1, in which assumptions are made to conservatively maximize the RCS and MSS pressure transients.

For this event, a turbine trip would occurfollowing a RT, whereas for the LOL/TT event, the turbine trip is the initiating fault. Therefore, the primary-to-secondary power mismatch and resultant RCS and MSS heatup and pressurization transients are always more severe for the LOL/TT event. For this reason, it is not necessary tocalculate the maximum RCS or MSS pressures for the CVCS Malfunction event.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-296An incident of moderate frequency does not generate a more serious plant condition without otherfaults occurring independently.

The major concern from a CVCS Malfunction event is that associated with pressurizer filling.The pressurizer water volume increases for this event as a result of the flow injected to the RCS.This event is analyzed to demonstrate that sufficient time is available for the appropriate operatoractions to be taken to preclude a pressurizer water-solid condition.

The acceptance criteria identified above are based on meeting the relevant regulatory requirements of10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

Brief discussions of thespecific GDCs that are related to the CVCS Malfunction acceptance criteria are provided as follows.GDC 10 (Reactor Design) requires that the reactor core and associated

coolant, control, andprotection systems be designed with appropriate margin so specified acceptable fuel design limitsare not exceeded during any condition of normal operation, including anticipated operational occurrences.

For the CVCV Malfunction event, this is shown to be met by demonstrating that thefuel cladding integrity is maintained.

GDC 15 (RCS Design) requires that the RCS and associated auxiliary,

control, and protection systems be designed with sufficient margin so design conditions of the RCPB are not exceededduring any condition of normal operation, including anticipated operational occurrences.

For theCVCS Malfunction event, this is shown to be met by demonstrating that the peak RCS pressure isless than 110 percent of the design pressure.

GDC 26 (Reactivity Control System Redundancy and Capability) requires the use of control rodscapable of reliably controlling reactivity changes with appropriate margin for malfunctions likestuck rods so that specified acceptable fuel design limits are not exceeded under conditions ofnormal operation, including anticipated operational occurrences.

For the CVCS Malfunction event, this is shown to be met by demonstrating that the fuel cladding integrity is maintained.

2.6.2.1.3 Description of Analyses and Evaluations The CVCS Malfunction event was analyzed using the RETRAN computer code (Reference 1). TheRETRAN model simulates the RCS, neutron kinetics, pressurizer, pressurizer relief and safety valves,pressurizer

heaters, pressurizer spray, SI system, SGs, FW system, and MSSVs. The code computespertinent plant variables including nuclear power, reactor coolant average temperature, RCS pressure, pressurizer water volume, and SG pressure.

2.6.2.1.4 ResultsIn all cases analyzed, the core power and RCS temperatures remain relatively constant.

Cases both withand without automatic rod control were examined.

Because there was little or no change in core powerand RCS average temperatures, the results showed that automatic control has no effect on the cases thatmodel maximum reactivity feedback conditions and a relatively negligible effect on the cases that modelminimum reactivity feedback conditions.

Figures 2.6.2-1 through 2.6.2-8 show the transient responses forWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-297the cases with automatic rod control modeled.

The calculated sequences of events for these cases arepresented in Table 2.6.2-1.The cases that model maximum reactivity feedback show that the pressurizer level increases at arelatively constant rate; whereas, the cases that model minimum reactivity feedback show that thepressurizer level increases at a somewhat varying rate. This is because the reactivity feedback in themaximum feedback cases is of a magnitude that it is able to maintain Tavg within the temperature deadband for the automatic rod control system, and no rod movement is predicted in these cases.However, rod movement is predicted in the minimum feedback cases, resulting in slight variations in Tavgand ultimately in the pressurizer level increase in these cases.The pressurizer level rate of increase is slightly faster in the cases where the pressurizer spray is modeledoperable, as compared to the cases in which the pressurizer sprays are modeled inoperable, because sprayactuation tends to keep the RCS pressure lower for several minutes, which allows the charging pumps todeliver more flow to the RCS. However, pressurizer pressure does eventually increase enough to open therelief valves in the cases with the pressurizer spray modeled operable.

The limiting case, shown in Figures 2.6.2-5 and 2.6.2-6, models minimum reactivity feedback conditions and the pressurizer sprays operable.

In this case, the pressurizer high level alarm is reached inapproximately 8.8 minutes and the pressurizer reaches a water-solid condition at approximately 17.3 minutes.

This allows the operators 8.5 minutes from the time the pressurizer high level alarm isreached to terminate normal charging flow before pressurizer filling occurs. Thus, with respect to thecriterion of precluding the generation of a more serious plant condition, there is sufficient time for theoperators (more than 8 minutes) to respond to the event and terminate the reactor coolant inventory addition.

2.6.2.2 Conclusion Based on the above information, it is concluded that the CVCS Malfunction event will not progress into amore serious plant condition.

Thus, all applicable event acceptance criteria are satisfied.

It has beendemonstrated that the reactor protection and safety systems ensure that the specified acceptable fueldesign limits are met and the RCPB pressure limits will not be exceeded as a result of the CVCSMalfunction event. Based on this, the plant will continue to meet the requirements of GDCs 10, 15,and 26.2.6.2.3 References

1. WCAP-14882

-P-A "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-298Table 2.6.2-1 Time Sequence of Events -CVCS Malfunction TimeCase Event Seconds MinutesMaximum Reactivity Two Pressurizer Level Channels Fail Low 0.0 0.0Feedback, with pressurizer Maximum Charging Flow from One CCP Begins 0.0 0.0sprayLetdown is Isolated 0.0 0.0Low-Low Pressurizer Level Alarm 0.0 0.0High Pressurizer Level Alarm 514.6 8.6Pressurizer Fills 1049.5 17.5Pressurizer Relief Valve Setpoint Reached 1054.3 17.6End of Transient 1800.0 30.0Maximum Reactivity Two Pressurizer Level Channels Fail Low 0.0 0.0Feedback, without Maximum Charging Flow from One CCP Begins 0.0 0.0pressurizer sprayLetdown is Isolated 0.0 0.0Low-Low Pressurizer Level Alarm 0.0 0.0Pressurizer Relief Valve Setpoint Reached 15.4 0.3High Pressurizer Level Alarm 609.1 10.2Pressurizer Fills 1365.7 22.8End of Transient 1800.0 30.0Minimum Reactivity Two Pressurizer Level Channels Fail Low 0.0 0.0Feedback, with pressurizer Maximum Charging Flow from One CCP Begins 0.0 0.0sprayLetdown is Isolated 0.0 0.0Low-Low Pressurizer Level Alarm 0.0 0.0High Pressurizer Level Alarm 529.3 8.8Pressurizer Fills 1036.4 17.3Pressurizer Relief Valve Setpoint Reached 1239.8 20.7End of Transient 1800.0 30.0Minimum Reactivity Two Pressurizer Level Channels Fail Low 0.0 0.0Feedback, without Maximum Charging Flow from One CCP Begins 0.0 0.0pressurizer sprayLetdown is Isolated 0.0 0.0Low-Low Pressurizer Level Alarm 0.0 0.0Pressurizer Relief Valve Setpoint Reached 24.5 0.4High Pressurizer Level Alarm 597.6 10.0Pressurizer Fills 1345.7 22.4End of Transient 1800.0 30.0WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-299WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-299c.)0C-_r)1.21-0.80.6-0.4-0.2-...........

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I I I I......................................................................I I I Ino4U05001000Time (s)15002000C ^I.-a)_ECi.cvUuu590-580-570"560"550-C A AJI'U0500I1000Time (s)15002000Figure 2.6.2-1CVCS Malfunction, Maximum Reactivity

Feedback, With Pressurizer SprayNuclear Power and Tavg versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-300WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-300zltUU.- 2300"2200U)cn 2100",- 2000-NJ1900U 1S1800-...............

..........................

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..........I170005500I1000Time (s)115002000E0U)L.I0 500 1000 1500Time (s)2000Figure 2.6.2-2CVCS Malfunction, Maximum Reactivity

Feedback, With Pressurizer SprayPressurizer Pressure and Water Volume versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3011.24-I0~75~C-3I0.8-0.6-0.4-0.2-...........

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I I I I.................................................U05;01000Time (s)15002000C-EC,,DUU590-580-570-560-550-r AA-....................................................0500I1000Time (s)115002000Figure 2.6.2-3CVCS Malfunction, Maximum Reactivity

Feedback, Without Pressurizer SprayNuclear Power and Tavg versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-302n~A nCnC,,0~CtDEnC-VtUU2300-2200-2100-2000-1900-1800-1700.......................................................................

...................... .................................................. ............................................... ....................' ............I05500I1000Time (s)1150020000 500 1000 1500Time (s)2000Figure 2.6.2-4CVCS Malfunction, Maximum Reactivity

Feedback, Without Pressurizer SprayPressurizer Pressure and Water Volume versus TimeWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-303WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-303.4-a,00~C-)1.21-0.8-0.6-0.4-0.2-.....................................................

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A -, I I I i .I IU0500I1000Time (s)15002000CAA-4Q_EUVU590-580-570-560-550-540.........................................................

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0500I1000Time (s)15002000Figure 2.6.2-5CVCS Malfunction, Minimum Reactivity

Feedback, With Pressurizer SprayNuclear Power and Tavg versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-304WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 040n(nC/)(nC/)C/)(DCLL¶4UU2300-..............

nnw% _& ................/ZUU2100-2000-1900-1800-1700..............

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05500I1000Time (s)1150020000 500 1000 1500Time (s)2000Figure 2.6.2-6CVCS Malfunction, Minimum Reactivity

Feedback, With Pressurizer SprayPressurizer Pressure and Water Volume versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-305WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3054-ý00~0-1.21 -0.80.6"0.4-0.2-...........................................................

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  • I r IU05001000Time (s)15002000(D,CL)-Cna,600590-580-570"560"550-..........................................................

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LýAfl.ru05001000Time (s)15002000Figure 2.6.2-7CVCS Malfunction, Minimum Reactivity

Feedback, Without Pressurizer SprayNuclear Power and T.vg versus TimeWCAP-1 7658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-306WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-306C,)0na,C/)L)E,0~0-24002 3 0 0 ...................................... ...............23002200 ...2 0 ............................................................2100-2 0 0 0 ..................................................2000 ........

......18 0 0 .................................................1900,1800-05001000Time (s)150020000 500 1000 1500Time (s)2000Figure 2.6.2-8CVCS Malfunction, Minimum Reactivity

Feedback, Without Pressurizer SprayPressurizer Pressure and Water Volume versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3072.7 DECREASE IN REACTOR COOLANT INVENTORY 2.7.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve (USAR Section 15.6.1)2.7.1.1 Technical Evaluation 2.7.1.1.1 Introduction An accidental depressurization of the RCS could occur as a result of an inadvertent opening of apressurizer relief or spray valve. To conservatively bound this scenario, the Westinghouse methodology models the failure of a PSV, because a PSV is sized to relieve approximately twice the steam flow of apressurizer PORV and thus results in a much more rapid depressurization upon opening.

Thedepressurization resulting from an open PSV is also much more rapid than would occur from theaccidental actuation of pressurizer spray. Therefore, the failure of a PSV yields the most severe coreconditions resulting from an accidental depressurization of the RCS. It should be noted that a stuck-open PSV is not an event of moderate frequency (i.e., Condition II event) such as a control system failurewould be. A stuck-open PSV is considered to be a SBLOCA (i.e., Condition III event) during which theRCS cannot be isolated, whereas the failure of a PORV can be overridden by the closure of the PORVblock valve. The results of this analysis are shown to comply with the more restrictive Condition IIacceptance criterion of ensuring that the DNB design basis is met.Initially, the event results in a rapidly decreasing RCS pressure, which could reach HL saturation conditions without reactor protection system intervention.

If saturated conditions are reached, the rate ofdepressurization is slowed considerably.

However, the pressure continues to decrease throughout theevent. The power remains essentially constant throughout the initial stages of the transient.

The reactor may be tripped by the following RTS signals:* OTAT* Pressurizer low pressure2.7.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaTo produce conservative results in calculating the DNBR during the transient, the following assumptions were made:The accident was analyzed using the RTDP (Reference 1). Pressurizer pressure and RCStemperature were assumed to be at their nominal values, consistent with steady-state full-power operation.

Reactor coolant minimum measured flow was modeled.

Uncertainties in initialconditions were included in the DNBR SAL as described in Reference

1. The event isconservatively analyzed at an initial NSSS power level of 3651 MWt, which includes nominalRCP net heat input; no additional uncertainty on core power is modeled.A zero moderator coefficient of reactivity was assumed.

This is conservative for BOL operation in order to provide a conservatively low amount of negative reactivity feedback due to changes inmoderator temperature.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-308A small (absolute value) Doppler coefficient of reactivity is assumed, such that the resultant amount of negative feedback is conservatively low in order to maximize any power increase dueto moderator feedback.

The spatial effect of voids resulting from local or subcooled boiling was not considered in theanalysis with respect to reactivity feedback or core power shape. In fact, it should be noted thatthe power peaking factors were kept constant at their design values, while the void formation andresulting core feedback effects would result in considerable flattening of the power distribution.

Although this would increase the calculated DNBR, no credit was taken for this effect.The analysis performed assumes that the rod control system is in automatic.

However, no rodmotion occurs during the transient because the conditions do not change enough to demand anyrod motion from the rod control system. Therefore, the transient results are identical with orwithout automatic rod control.Based on its frequency of occurrence, the accidental depressurization of the RCS accident is considered tobe a Condition II event as defined by the ANS's "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,"

ANSI N 18.2-1973.

The following items summarize the acceptance criteria associated with this event:The critical heat flux should not be exceeded.

This criterion was met by demonstrating that theminimum DNBR does not go below the limit value at any time during the transient.

Pressure in the RCS and MSS should be maintained below 110 percent of the design pressures.

Note that because this event is a depressurization event, these limits are not challenged.

Bothprimary and secondary pressures decrease for the entire duration of the event.As discussed above, the accidental depressurization of the RCS event has historically beenanalyzed to show that the minimum DNBR limit is not exceeded.

However, during the licensing pre-application meetings between Westinghouse.,

WCNOC, and the USNRC, the USNRCrequested that the potential for pressurizer filling during the event, due to the actuation of the SIsystem, be considered.

Consistent with that request, an additional sensitivity was performed toshow that the pressurizer would not overfill such that the transient would not transition to a moreserious plant condition.

2.7.1.1.3 Description of Analyses and Evaluations The purpose of this analysis was to demonstrate that the RTS functions and mitigates the consequences ofthe RCS depressurization event. This analysis is concerned with the transient from initiation through justpast the time of RT. With respect to long-term post-accident

recovery, it is assumed that operators followapproved plant procedures to bring the plant to a safe post-accident condition.

The accident was analyzed by using the detailed digital computer code RETRAN (Reference 2). Thiscode simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, SG, and SG safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3092.7.1.1.4 ResultsThe system response to an inadvertent opening of a PSV is shown in Figures 2.7.1-1 through 2.7.1-4.Figure 2.7.1-1 illustrates thenuclear power transient.

Nuclear power remains essentially unchanged untilthe RT occurs on OTAT. The pressurizer pressure transient is illustrated in Figure 2.7.1-2.

Pressuredecreases continuously throughout the transient.

However, pressure decreases more rapidly after core heatgeneration is reduced via the RT. Figure 2.7.1-3 shows the loop average temperature transient.

The loopaverage temperature decreases slowly until the RT occurs. The DNBR decreases initially, but increases rapidly following the RT as demonstrated in Figure 2.7.1-4.

The DNBR remains above the SALthroughout the transient.

The calculated sequence of events is shown in Table 2.7.1-1.

The calculated minimum DNBR value isprovided in Table 2.7.1-2.The results of the analysis show that the OTAT RTS function provides adequate protection against theRCS depressurization event because the minimum DNBR remains above the SAL throughout thetransient.

Therefore, no cladding damage or release of fission products to the RCS is predicted for thisevent.With regards to overfill, the WCGS has a pressurizer PORV interlock that is set to 2185 psig. When thepressurizer pressure reaches the PORV interlock

setpoint, the PORV block valves are closed. A detailedRETRAN analysis was performed to demonstrate that the PORV block valves close prior to thepressurizer pressure reaching the Pressurizer Pressure

-Low, S1 setpoint.

This will prevent the Si systemfrom actuating; without the addition of SI, pressurizer filling does not occur. The RETRAN analysisconservatively modeled signal processing time, valve stroke time, and instrument uncertainties to increasethe likelihood of SI actuation.

2.7.1.2 Conclusion The RCS depressurization analysis demonstrates that for this event at WCGS, the DNBR does notdecrease below the SAL value at any time. The event does not challenge the primary and secondary sidepressure limits because this is a depressurization event. Thus, all applicable acceptance criteria for thisevent are met for WCGS operating at a nominal NSSS power of up to 3651 MWt.2.7.1.3 References

1. WCAP-11397-P-A, "Revised Thermal Design Procedure,"

April 1989.2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-310Table 2.7.1-1 Time Sequence of Events -Accidental Depressurization of the RCSEvent Time (seconds)

PSV opens fully 0.0OTAT RT setpoint reached 21.4Rods begin to drop 24.4Minimum DNBR 25.0Table 2.7.1-2 Results -Accidental Depressurization of the RCSMinimum Calculated DNBR DNBR SAL2.001 1.52WCAP-1 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-31100CD0 10 20 30 40Time (seconds)

Figure 2.7.1-1 RCS Depressurization

-Nuclear Power versus Time50C)C-U-C)ELC)K-_C)K-_EL_0 10 20 30 40Time (seconds)

Figure 2.7.1-2 RCS Depressurization

-Pressurizer Pressure versus Time50WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-312WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 12590c 585D 580EW 575U')-5700o 5650560555I I I I I I I I I I I I I I I I I I I I01020Time30(seconds) 4050Figure 2.7.1-3 RCS Depressurization

-Indicated Loop Average Temperature versus Time4.5.4.03.5-= 3.0-r --0 10 2O 30 40 50Time (seconds)

Figure 2.7.1-4 RCS Depressurization

-DNBR versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3132.7.2 Steam Generator Tube Rupture Margin to Overfill (USAR Section 15.6.3)2.7.2.1 Technical Evaluation 2.7.2.1.1 Introduction The major hazard associated with a SGTR event is the radiological consequences resulting from thetransfer of radioactive primary coolant to the secondary side of the ruptured SG and subsequent release ofradioactivity to the atmosphere.

One major concern for an SGTR is the possibility of ruptured SGoverfills because this could potentially result in a significant increase in the radiological consequences.

Therefore, an analysis was performed to demonstrate that the ruptured SG does not overfill and releasewater from the main steam relief valves, assuming the limiting single failure relative to overfill.

Theanalysis confirmed that water releases through the SG safety valves did not occur.The SGTR margin to overfill transient analysis was performed using the RETRAN computer program(Reference

1) following the methodology developed in WCAP-10698-P-A and its Supplement I(References 2 and 3). Modifications were made to address NSAL-07-11 (Reference 4), which identified apotential non-conservative assumption.

This regards the direction of conservatism for decay heat in theReference 2 methodology for demonstrating margin to overfill.

The plant response to the SGTR was modeled using conservative assumptions of break size and location, condenser availability, and initial secondary water mass. The analyses include the simulation of theoperator actions for recovery from an SGTR based on the WCGS Emergency Operating Procedures (EOPs), which are based on the Westinghouse Owners Group Emergency ResponseGuidelines.

The SGTR margin to overfill analysis was performed for the time period from the SGTR until the primaryand secondary pressures equalized (break flow termination).

In the ruptured SG secondary side, the watervolume was calculated as a function of time to demonstrate that overfill did not occur.The SGTR margin to overfill analysis supports operation at a core power up to 3637 MWt. The analysissupports a full power RCS Tavg operating range from 575.00 to 588.4°F, and a main Tfeed range from 4000to 448.6'F, with up to 10 percent of the SG tubes plugged.2.7.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe margin to overfill analysis modeled the plant operating at the lower end of the Tavg range. A loweroperating temperature results in a higher mass flow rate through the broken tube and less steam releasedfrom the ruptured SG. The analysis assumed that the plant was operating with the T'eed at the low end ofthe temperature range. This results in a higher mass of water in the SG at the start of the event, whichlimits the amount of break flow and AFW that can accumulate in the ruptured SG without forcing waterinto the steam lines. The maximum SGTP was modeled because it reduces heat transfer to the rupturedSG, which reduces the mass released by steaming, which in turn reduces margin to overfill.

The reducedheat transfer also prolongs the cooldown period, leading to delayed break flow termination.

Sensitivity runs were made to confirm the conservative nature of these plant operating assumptions.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3142.7.2.1.2.1 Design Basis AccidentThe accident modeled was a double-ended break of one SG tube located at the top of the tube sheet on theoutlet (CL) side of the SG. The location of the break on the cold side of the SG results in higher primaryto secondary break flow than a break on the hot side of the SG. It was also assumed that a LOOP occurs atthe time of RT, and the highest worth control rod assembly was assumed to be stuck in its fully withdrawn position at RT.2.7.2.1.2.2 Single Failure Considerations An evaluation was performed to determine the limiting single failure with respect to margin to SG overfillfor an SGTR. To identify the limiting single failure, sensitivity runs were performed considering thefollowing failures:

Failure of an Intact SG ARV or Failure of Multiple SG ARVsThis scenario considered the failure of an ARV to open on one of the intact SGs when theoperator performed the RCS cooldown.

Because offsite power was assumed to be lost at RT forthe SGTR analyses, the SG ARVs were relied upon to cool the RCS. Failure of an ARV on anintact SG to open on demand reduced the steam release capability provided by the ARVs becauseonly two intact SG ARVs are available for the cooldown.

This increased the time required for thecooldown, resulting in increased break flow.A single failure that results in the failure of multiple SG ARVs does not exist in the WCGSdesign. Each SG ARV can be actuated by an independent safety-related compressed gas supply. Afailure of any one of the four compressed gas supplies would only affect the associated SG ARV,and would not affect the other three SG ARVs.* Failure of the MDAFW Control ValveThis scenario considered the failure of the MDAFW control valve to isolate MDAFW flow to theruptured SG when the TDAFW flow is isolated.

This required additional operator action tomanually isolate the MDAFW flow, resulting in increased AFW flow to the ruptured SG. Thus,the mass in the ruptured SG increased in relation to the intact SGs prior to RT. Although thisadditional mass would be expected to provide early identification and isolation of AFW flow tothe ruptured SG following RT, no reduction in the operator action time for AFW isolation wascredited.

The initial secondary SG water mass was not increased to account for the impact ofturbine runback.

This modeling is consistent with the AFW flow control valve failure presented inReference

2. The cooldown was performed using all three of the ARVs on the intact SGs.The MDAFW control valve failure was determined to be the limiting single failure.

The penalty from thedelay to terminate AFW flow to the ruptured SG that resulted from the AFW control valve failure resultedin the largest secondary side inventory.

The effects of adding more inventory to the ruptured SG throughlonger AFW flow duration offset the effects of the other intact SG ARV failure, which prolonged cooldown and break flow termination.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3152.7.2.1.2.3 Conservative Assumptions Plant responses until break flow termination were calculated using the RETRAN computer code. Theconservative conditions and assumptions used in Reference 2 were also used in the analysis to determine margin to SG overfill with the exception of the following differences:

Turbine RunbackThe mass increase due to turbine runback is the mass corresponding to power at the end ofturbine runback minus the mass at 100 percent power. A power reduction of 10%/min for turbinerunback is assumed, where the turbine runback duration is the smaller value of RT time or3 minutes.

The reduction in power that would result from the turbine runback during that periodis used to develop a secondary mass penalty associated with the runback.SG Secondary MassA higher initial secondary water mass in the ruptured SG was determined by Reference 2 to beconservative for overfill.

The increase in mass that would result from a turbine runback to a lowerpower (discussed in the prior item) and the consideration of mass uncertainties are added to theinitial secondary water mass.Intact SG Target Pressures The intact SG target pressures are calculated based on the target temperature and the intact loopAT at the time cooldown is terminated instead of at the start of cooldown.

This exception isjustified because it is consistent with the WCGS EOPs.AFW Isolation Based Solely on SG LevelThe analysis modeled AFW flow isolation based on ruptured SG level with no consideration of atime component.

The SG level for AFW isolation from the WCGS EOPs is 6 percent NRS. (Notethat the analysis used a conservative SG level of 15 percent NRS with the stipulation that theaction not be taken before 2 minutes after RT.) This exception is justified because it is a morerealistic modeling of the operator response to an SGTR accident.

Decay Heat and NSAL-07-11 NSAL-07-11 (Reference

4) identifies a potential non-conservative assumption regarding thedirection of conservatism for decay heat in the Reference 2 methodology for evaluating margin tooverfill.

For the margin to overfill

analysis, higher decay heat yields a benefit by increasing steamreleases from the ruptured SG, but results in a penalty from a longer cooldown and aconservatively delayed break flow termination.

Conversely, lower decay heat yields a penalty byreducing steam releases from the ruptured SG, but results in a benefit from a shorter cooldownand earlier break flow termination.

Similar impacts were identified for the AFW and SI flowenthalpies.

The relative importance of these competing effects is plant-specific, and plant-specific analyses are required to determine the conservative assumption.

Plant-specific sensitivities WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-316performed for WCGS showed the following to be conservative with respect to margin to overfillfor the limiting cases:-1979-2a ANS decay heat was conservative compared to the 1971+20%

ANS decay heatmodel specified in Reference

2. For this analysis, the 1979 ANS decay heat model minus2a uncertainty was used.-Minimum AFW enthalpy was conservative compared to the maximum AFW enthalpyspecified by Reference
2. For this analysis, the minimum AFW enthalpy of 18.1 Btu/lbmwas modeled.-Minimum SI enthalpy was conservative compared to the maximum SI enthalpy modeled inReference
2. For this analysis, the minimum SI enthalpy of 11.53 Btu!lbm was modeled.2.7.2.1.2.4 Plant InputThe following significant WCGS input was used in the analysis:
1. SG ARVIt was assumed that a LOOP occurs at RT for the SGTR analyses, and thus the SG ARVs open tolimit the secondary pressure.

The ARV pressure setpoint is 1139.7 psia (1125 psig). The ARVcapacity modeled in the analysis is 594,642 lbm/hr/valve at a reference pressure of 1107 psia(1092.3 psig).2. Pressurizer PORV CapacityIt was assumed that a LOOP occurs at RT for the SGTR analyses, and thus the pressurizer PORVwas relied upon to depressurize the RCS. The capacity of 210,000 Ibm/hr at 2350 psia was usedin the analysis.

3. AFW System Operation and Associated Single Failure Considerations The WCGS AFW system consists of two MDAFW pumps and one TDAFW pump. EachMDAFW pump normally feeds two SGs and the TDAFW pump feeds all four SGs. There is acontrol valve in the flow path from the MDAFW pump to each SG and a control valve in the flowpath from the TDAFW pump to each SG. The control valves in the MDAFW pump andTDAFW pump flow paths are used to control the inventory in the SGs, and are closed to isolateAFW flow to the ruptured SG in accordance with WCGS Emergency Mitigation Guideline E-3for SGTR recovery.

The control valves for the MDAFW pumps are controlled to throttle the flowas required to maintain the level in the associated SG between 29 percent and 50 percent NRS,and can also be manually controlled by the operator to adjust flow to the SGs. Also, the automatic level control function of the MDAFW pump control valves is not credited to reduce the liquidinventory in the ruptured SG. The control valves for the TDAFW pump have no automatic controlfeatures.

They are manually throttled to adjust flow to maintain the desired level in the SGs.WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-317The AFW flow rates for WCGS are dependent on the number of MDAFW and TDAFW pumpsthat are operating as well as the SG pressure.

Flow to an SG from one or more pumps may bethrottled or isolated depending on the time in the transient progression, the level in the SG, andthe single failure being considered.

All AFW pumps are assumed to be operating following RTand LOOP. The operators will isolate flow from the TDAFW pump and then isolate flow from theMDAFW pump to the ruptured SG. The single failures are discussed in detail later in this report.The associated AFW flows are outlined below.a. MDAFW Failure AFW FlowsThe MDAFW failure is the failure of the MDAFW control valve to throttle or isolate ondemand by the control system or by manual operator action. This failure would causeadditional AFW flow to be delivered to the ruptured SG until the associated MDAFWpump is stopped.

This terminates all AFW flow to the ruptured SG and one intact SG, andleaves the MDAFW pumps providing flow to two intact SGs. It is assumed that the AFWflow to the third intact SG is restored by the start of the cooldown.

The AFW flow rates to the ruptured and intact SGs for the MDAFW failure are shown inTables 2.7.2-1 through 2.7.2-3.b. Intact SG ARV Failure AFW FlowsThe intact SG ARV failure is the failure of an ARV on one intact SG to open for cooldown.

This failure has no impact on the AFW flows and the AFW system. A constant AFW flowof 320 gpm is used throughout the event.4. SI FlowsThe maximum SI flow was assumed to be initiated at the low pressurizer pressure setpoint of2004.7 psia. The flow rates are presented in Table 2.7.2-4.2.7.2.1.2.5 Operator Action TimesIn the event of an SGTR, the operator is required to take actions to stabilize the plant and terminate theprimary to secondary break flow. The operator actions for SGTR recovery are provided in the WCGSEOPs, and major actions were explicitly modeled in these analyses.

The operator actions modeled includeisolation of the ruptured SG, cooldown of the RCS, depressurization of the RCS to restore inventory, andtermination of SI to stop primary to secondary break flow. These operator actions are described below.1. Identify the Ruptured SGHigh secondary side activity, as indicated by the main steamline radiation monitor (or othersecondary monitors) or high SG sample activity typically will provide the first indication of anSGTR event. The ruptured SG can be identified by a mismatch between steam and FW flows,high activity in an SG water sample, or a high radiation indication on the corresponding mainsteamline radiation monitor.

For an SGTR that results in a RT at high power as assumed in theseWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-318analyses, the SG water level, as indicated on the narrow range, will decrease significantly for allof the SGs. The AFW flow will begin to refill the SGs, distributing approximately equal flow toeach of the SGs. Because primary to secondary break flow adds additional inventory to theruptured SG, the water level will increase more rapidly in that SG. This response, as displayed bythe SG water level instrumentation, provides confirmation of an SGTR event and also identifies the ruptured SG.2. Isolate the Ruptured SGOnce the ruptured SG has been identified, recovery actions begin by isolating AFW flow to theruptured SG and closing the MSIV on the ruptured SG steamline.

In addition to minimizing radiological

releases, this also reduces the possibility of filling the ruptured SG by minimizing theaccumulation of AFW. The operator can also establish a pressure differential between theruptured and intact SGs, which is a necessary step toward terminating primary to secondary breakflow. In the WCGS EOPs for SGTR, the operator is directed to verify that the level in theruptured SG is greater than a specified level on the NRS prior to isolating AFW. The requiredlevel is 6 percent NRS. (Note that the analysis used a conservative SG level of 15 percent NRSwith the stipulation that the action not be taken before 2 minutes after RT.) For the single failureinvolving the failure of MDAFW control valves, an additional 30 seconds is added after thenarrow range level is reached until MDAFW isolation.

For the single failure involving the failureof the SG ARV to open, the TDAFW and MDAFW are isolated simultaneously once the narrowrange level is reached.

All SG MSIVs were assumed to be closed at 8 minutes from RT for bothsingle failure scenarios.

3. Cooldown the RCS using the Intact SGsAfter isolation of the ruptured SG MSIV, dumping steam from only the intact SGs cools the RCSas rapidly as possible to less than the saturation temperature corresponding to the rupturedSG pressure.

This ensures adequate subcooling in the RCS after depressurization to the rupturedSG pressure in subsequent actions.

If offsite power is available, the normal steam dump system tothe condenser can be used to perform this cooldown.

If offsite power is lost, the RCS would becooled using the ARVs on the intact SGs. The availability of relief valves for cooldown isdependent on the single failure assumption being modeled (See Section 2.7.2.1.2.5).

The analysis assumed 23 minutes elapse from the time of RT until the cooldown was initiated viathe ARVs. The cooldown is terminated when the required core exit temperature for cooldowntermination (without adverse environment) corresponding to the ruptured SG pressure is reached.The temperature is identified in the WCGS EOPs for SGTR. The ARVs on the intact SGs werethen used as necessary to maintain that temperature.

4. Depressurize the RCS to Restore Inventory When the cooldown is completed, SI flow will tend to increase RCS pressure until break flowmatches SI flow. Consequently, SI flow must be terminated to stop primary to secondary breakflow. However, adequate inventory must first be assured.

This includes both sufficient RCSsubcooling and pressurizer inventory to maintain a reliable pressurizer level indication after SIWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-319flow is stopped.

Because break flow from the primary side will continue after SI flow is stoppeduntil RCS and ruptured SG pressures

equalize, an excess amount of inventory is needed to ensurethat the pressurizer level remains on span. The excess amount required depends on the RCSpressure, and reduces to zero when the RCS pressure equals the pressure in the ruptured SG.The analyses assumed that 5 minutes elapsed from the time the cooldown was terminated untilthe depressurization was initiated.

The RCS depressurization is performed using normalpressurizer spray if the RCPs are running.

Because offsite power was assumed to be lost at thetime of RT, the RCPs were not running and thus normal pressurizer spray was not available.

Therefore, the depressurization was modeled using a pressurizer PORV.The RCS depressurization is continued until any of the following three conditions in the WCGSEOPs for SGTR (using setpoints without adverse environment) are satisfied:

RCS pressure is lessthan the ruptured SG pressure and pressurizer level is greater than 6 percent, pressurizer level isgreater than 75 percent, or RCS subcooling is less than required to address the subcooling uncertainty.

5. Terminate SI to Stop Primary to Secondary Break FlowThe previous actions will have established adequate RCS subcooling, a secondary side heat sink,and sufficient RCS inventory to ensure that the SI flow is no longer needed. When these actionshave been completed, the SI flow must be stopped to terminate primary to secondary break flow.The analyses assumed that 4 minutes elapsed from the time the depressurization was terminated until SI could be stopped.

SI can be stopped provided the following conditions in the WCGSEOPs for SGTR (using setpoints without adverse environment) are satisfied:

RCS pressure isstable or rising, pressurizer level is greater than 6 percent, RCS subcooling is greater thanrequired to address the subcooling uncertainty, and a secondary heat sink is confirmed.

After SI termination, the analyses do not model specific actions leading to break flowtermination, consistent with the Reference 2 method. The primary to secondary break flowcontinues after the SI flow is stopped until the RCS and ruptured SG pressures equalize.

The total time required to complete the recovery operations consists of both operator action time andsystem, or plant, response time. For instance, the time for each of the major recovery operations (i.e., RCScooldown) is primarily due to the time required for the system response, whereas the operator action timeis reflected by the time required for the operator to perform the intermediate action steps.The operator action times to isolate AFW flow to the ruptured SG, to isolate the MSIV on the rupturedSG, to initiate RCS cooldown, to initiate RCS depressurization, and to terminate SI were developed forthe design basis analyses.

WCGS has determined the corresponding operator action times to performthese operations.

The operator actions and the corresponding operator action times used for the analysesare summarized in Table 2.7.2-5.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3202.7.2.1.2.6 Acceptance CriteriaThe analyses were performed to demonstrate that the secondary side of the ruptured SG did notcompletely fill with water. The available secondary side volume of a single SG is 5852 ft3.Margin tooverfill is demonstrated, provided the transient-calculated SG secondary side water volume is less thanthis value. No credit is taken for the volume of the nozzle or any steam piping.2.7.2.1.3 Description of Analyses and Evaluations The RETRAN analysis for the limiting margin to overfill case is described below. The limiting case withrespect to margin to SG overfill considered operation at the minimum operating temperature (575.0°F),

with the minimum main FW temperature (400.07F),

the maximum SGTP level (10 percent),

low AFWenthalpy, low SI enthalpy, low decay heat (1979-2a),

and the failure of the MDAFW control valve toclose. The sequences of events for these transients are presented in Table 2.7.2-6.Following the tube rupture, water flowed from the primary into the secondary side of the ruptured SGbecause the primary pressure is greater than the SG pressure.

In response to this loss of coolant,pressurizer level decreased (Figure 2.7.2-1).

The RCS pressure (represented by the pressurizer pressure) also decreased (Figure 2.7.2-2) as the steam bubble in the pressurizer expanded.

As the RCS pressuredecreased due to the continued primary to secondary break flow, an automatic RT occurred on an OTATtrip signal.After RT, core power rapidly decreased to decay heat levels. The turbine stop valves closed and steamflow to the turbine was terminated.

The steam dump system is designed to actuate following RT to limitthe increase in secondary

pressure, but the steam dump valves remained closed due to the loss ofcondenser vacuum resulting from the assumed LOOP at the time of RT. Thus, the energy transfer from theprimary system caused the secondary side pressure to increase rapidly after RT (Figure 2.7.2-3),

until theSG ARVs lifted to dissipate the energy. As a result of the assumed LOOP, main FW flow was assumed tobe terminated and AFW flow was assumed to be automatically initiated following RT.The RCS pressure and pressurizer level continued to decrease after RT as energy transfer to the secondary system shrank the primary coolant and the tube rupture break flow continued to deplete primaryinventory.

The decrease in RCS inventory resulted in a low pressurizer pressure SI signal. The SI flowincreased the RCS inventory and the RCS pressure trended toward the equilibrium value, where the SIflow rate would equal the break flow rate.TDAFW flow to the ruptured SG was isolated at 368 seconds, MDAFW flow to the ruptured SG wasisolated at 398 seconds, and the ruptured SG MSIV was closed at 625 seconds.

The ruptured SG level waswell above the level required for identification and isolation by these times as a conservatively high SGlevel was assumed in the analysis.

Cooldown of the RCS was initiated 23 minutes after RT. It was therefore assumed that the ARVs on threeintact SGs were opened for the RCS cooldown at 1525 seconds (Figure 2.7.2-6).

The cooldown wascontinued until the cooldown termination temperature obtained from WCGS EOPs was reached.

Whenthis condition was satisfied, the operator closed the ARVs to terminate the cooldown.

This cooldownensured that there would be adequate subcooling in the RCS after the subsequent depressurization of theWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-321RCS to the ruptured SG pressure.

The reduction in the intact SG pressure required to accomplish thecooldown is shown in Figure 2.7.2-3.

The pressurizer level and RCS pressure also decreased during thiscooldown process due to shrinkage of the RCS (Figures 2.7.2-1 and 2.7.2-2).

The ARVs on the intact SGs that were used for the cooldown also automatically opened as necessary tomaintain the prescribed RCS temperature to ensure that subcooling was maintained.

When the ARVs wereopened, the increased energy transfer from the RCS to the secondary system also aided in thedepressurization of the RCS to the ruptured SG pressure after the SI flow was terminated.

After termination of the cooldown, a 5-minute operator action time was imposed prior to the RCSdepressurization.

In these analyses, the RCS depressurization was terminated when the RCS pressure wasreduced to less than the ruptured SG pressure and the pressurizer level was above the required value,because there was adequate subcooling margin and the high pressurizer level setpoint was not reached.The RCS depressurization is shown in Figure 2.7.2-2.

The depressurization reduced the break flow(Figure 2.7.2-4) and increased SI flow to refill the pressurizer (Figure 2.7.2-1).

After termination of the depressurization, a 4-minute operator action time was imposed prior to SItermination.

The SI flow was terminated at this time because the requirements for SI termination weresatisfied.

(RCS subcooling was greater than the required allowance for subcooling uncertainty, minimumAFW flow was available or at least one intact SG level was in the narrow range, the RCS pressure wasstable or increasing, and the pressurizer level was greater than the required value.) After SI termination, the RCS pressure began to decrease (Figure 2.7.2-2).

Break flow was terminated at 4132 seconds.2.7.2.1.4 ResultsThe primary to secondary break flow rate throughout the recovery operations is presented inFigure 2.7.2-4.

The water volume in the ruptured SG is presented as a function of time in Figure 2.7.2-5.The ruptured loop RCS temperature is presented in Figure 2.7.2-7.

The intact loops RCS temperature ispresented in Figure 2.7.2-8.

The peak ruptured SG water volume is 5789 ft3 resulting in 63 ft3 of marginto overfill.

Therefore, it is concluded that overfill of the ruptured SG will not occur for a design basisSGTR for WCGS.2.7.2.2 Conclusion It is concluded that overfill of the ruptured SG causing water to pass through the main steam relief valveswill not occur for a design basis SGTR for WCGS.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3222.7.2.3 References

1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.2. WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,"

August 1987.3. Supplement I to WCAP-10698-P-A, "Evaluation of Offsite Radiation Doses for a SteamGenerator Tube Rupture Accident,"

March 1986.4. NSAL-07-1 1, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology,"

November 2007.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-323Table 2.7.2-1 AFW Flows for Design Basis SGTR Analyses MDAFW Failure, All AFW PumpsOperating SG Pressure AFW Flow to Ruptured SG AFW Flow to Intact SGs(psig) (gpm) (gpm/SG)1125 684.76 320.001000 787.93 332.67900 862.25 367.13800 931.00 398.64700 995.51 428.37Table 2.7.2-2 AFW Flows for Design Basis SGTR Analyses MDAFW Failure, TDAFW Pump Stopped,MDAFW Pumps Operating SG Pressure AFW Flow to Ruptured SG AFW Flow to Intact SGs(psig) (gpm) (gpm/SG)1125 440.50 320.001000 500.80 332.67900 544.60 367.13800 585.50 398.64700 623.90 428.37Table 2.7.2-3 AFW Flows for Design Basis SGTR Analyses MDAFW Failure, Ruptured SG Isolated, TDAFW Pump Stopped, MDAFW Pumps Operating During CooldownSG Pressure AFW Flow to Ruptured SG(psig) (gpm)1125 320.001000 332.67900 367.13800 398.64700 428.37WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-324Table 2.7.2-4 SI Flows for Design Basis SGTR AnalysesPressure (psig) Total Injection Flow Rate (gpm)900 12821000 12241100 11651200 11001300 10291400 9511500 8581600 7361700 5381800 5141900 4892000 4632100 4362200 4062235 3952300 3742400 3403000 340WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-325Table 2.7.2-5 Operator Action Times for Design Basis SGTR Margin to Overfill AnalysesAction TimeOperator action time to isolate TDAFW flow to 15% NRSruptured SG")Operator action time to isolate MDAFW flow to AFW control valve single failure -30 seconds fromruptured SG"' TDAFW isolation SG ARV single failure -coincident with TDAFWisolation Operator action time to isolate MSIV on ruptured SG 8 minutes from RTOperator action time to initiate cooldown 23 minutes from RTCooldown Calculated by RETRANOperator action time to initiate depressurization 5 minutes from end of cooldownDepressurization Calculated by RETRANOperator action time to terminate SI following 4 minutes firom end of depressurization depressurization Pressure equalization Calculated by RETRANNote:I. Isolation is assumed to occur no earlier than 2 minutes after RT.Table 2.7.2-6 Sequence of Events for Limiting Margin to Overfill AnalysesEvent Time (seconds)

SGTR 100RT (OTAT) and LOOP 145AFW Initiated 145SI Actuated 304TDAFW Flow to Ruptured SG Isolated 368MDAFW Flow to Ruptured SG Isolated 398Ruptured SG MSIV Closed 625RCS Cooldown Initiated 1525RCS Cooldown Terminated 2149RCS Depressurization Initiated 2449RCS Depressurization Terminated 2481SI Terminated 2721Break Flow Terminated 4132WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3261IU-C)_0 1000 2000 300o 4000Time (s)5000Figure 2.7.2-1 Pressurizer Level -Margin to Overfill AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-327WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-327C/')C)0 1000 2000 3000 4000Time (s)5000Figure 2.7.2-2 Pressurizer Pressure

-Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-328WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-328Intact SGsR...Rup tured SG2500-2000-o 1500-c_C !)0-1000-0 1000 2000 3000 4000Time (s)5000Figure 2.7.2-3 Secondary Pressure

-Margin to Overfill AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-329WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-329504030200ElCD10-I0 1000 2000 3000 4000Time (s)5000Figure 2.7.2-4 Primary to Secondary Break Flow -Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-330WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-330Av i clab I eRuptured SI60005000-4000-ILD3000-2000-1000-0010002000Time7000(S)40005000Figure 2.7.2-5 SG Water Volumes -Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-331WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-33 1Ruptured SGTotal IntactSGsQ90)Ct)co0 1000 2 o00 3000 4000Time (s)5000Figure 2.7.2-6 SG Steam Releases

-Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTLNGHOUSE NON-PROPRIETARY CLASS 32-332WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 32Hot Leg InCo d Leg II eetn Ie t700600500-40)cB400-:73000)a-G~)K-~A\ ,~'N300200-100-01000200003000040005000Time (s)Figure 2.7.2-7 Ruptured Loop RCS Temperature

-Margin to Overfill AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-333Hot Leg InIetCo d Leg Inlet700-600500WS400-C)500-Cr,C)0 1000 2000 3000 4000Time (s)5000Figure 2.7.2-8 Intact Loops RCS Temperature

-Margin to Overfill AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3342.7.3 Steam Generator Tube Rupture -Input to Dose (USAR Section 15.6.3)2.7.3.1 Technical Evaluation 2.7.3.1.1 Introduction The major hazard associated with an SGTR event is the radiological consequences resulting from thetransfer of radioactive primary coolant to the secondary side of the ruptured SG and subsequent release ofradioactivity to the atmosphere.

A T/H SGTR analysis was performed to determine the mass releases foruse in calculating the radiological consequences, assuming the limiting single failure relative toradiological consequences without ruptured SG overfill.

Section 2.7.2 confirmed that ruptured SG overfilldid not occur.The SGTR T/H transient analysis was performed using the RETRAN computer program (Reference 1)following the methodology developed in Reference 2 and its Supplement 1 (Reference 3). The plantresponse to the event was modeled using conservative assumptions of break size and location, condenser availability, and initial secondary water mass. The analyses include the simulation of the operator actionsfor recovery from an SGTR based on the WCGS EOPs, which are based on the Westinghouse OwnersGroup Emergency Response Guidelines.

A detailed SGTR T/H analysis was performed for the time period from the SGTR until the primary andsecondary pressures equalized (break flow termination).

In the T/H analysis, the primary to secondary break flow and the steam releases to the atmosphere from the ruptured and intact SGs were calculated foruse in determining the activity released to the atmosphere.

The mass releases were calculated with theRETRAN computer code from the initiation of the event until break flow termination.

For the time periodfrom break flow termination until all releases are terminated, steam releases from the intact and rupturedSGs were determined from a mass and energy balance.The SGTR T/H analysis supports operation at a core power up to 3637 MWt. The analysis supports a fullpower RCS Tavg operating range from 570.7°F to 588.4°F, and a main Tfeed range from 400'F to 448.6°F,with up to 10 percent of the SG tubes plugged.2.7.3.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe T/H analyses, which determined the mass releases for the radiological consequences

analyses, modeled the plant operating at the higher end of the Ta,,g range. A higher operating temperature results inincreased steaming from the ruptured SG and a higher fraction of the break flow flashing to steam insidethe ruptured SG. The analyses also assumed that the plant was operating at the high end of the Tfeed range.This results in increased steaming from the ruptured SG. A SG tube plugging level of 10 percent wasmodeled in the analyses because this results in a higher fraction of the break flow flashing to steam insidethe ruptured SG.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3352.7.3.1.2.1 Design Basis AccidentThe design basis accident modeled was a double-ended break of one SG tube located at the top of the tubesheet on the outlet (CL) side of the SG. The location of the break on the cold side of the SG results inhigher primary to secondary break flow than a break on the hot side of the SG, as determined byReference

2. However, the break flow flashing fraction was conservatively calculated, assuming that allof the break flow comes from the HL side of the SG. The combination of these conservative assumptions results in a very conservative calculation of the radiological consequences.

It was also assumed thatLOOP occurred at the time of RT, and the highest worth control rod assembly was assumed to be stuck inits fully withdrawn position at RT. Due to the assumed LOOP, the condenser was not available for steamreleases once the reactor was tripped.

Consequently, after RT, steam was released to the atmosphere through the SG ARVs.2.7.3.1.2.2 Single Failure Consideration Based on Reference 3, the most limiting single failure with respect to radiological consequences (assuming no overfill) is a failed-open ARV on the ruptured SG. Failure of this ARV causes anuncontrolled depressurization of the SG, which increases primary to secondary break flow and the steamrelease to the atmosphere.

The lower secondary pressure also results in a higher break flow flashingfraction.

Pressure in the ruptured SG will remain below that in the primary system until the failed ARVcan be isolated, and recovery actions completed.

2.7.3.1.2.3 Conservative Assumptions This section includes a discussion of the methods and assumptions used to analyze the SGTR event and tocalculate the mass released, the sequence of events during the recovery operations, and the calculated results.Most of the assumptions used for the margin to overfill analyses are also conservative for the radiological consequences analyses.

The major differences in the assumptions that were used for the RETRANanalyses for radiological consequences compared to those used in the margin to overfill analyses arediscussed below.1. SG Secondary MassA low secondary mass is conservative for the dose analyses because it promotes steam releasefrom the ruptured SG. A low secondary mass also results in a lower ruptured SG pressure whenthe ruptured SG ARV is failed open. This was considered in the confirmation that the pressure didnot decrease below 275 psig as noted in the operator action time discussion below.2. Decay Heat and NSAL-07-11 As noted in NSAL 11 (Reference 4), SGTR T/H analyses for input to the radiological consequences analyses have no competing effects with respect to decay heat. Higher decay resultsin increased steam releases from the ruptured SG and a longer cooldown, leading to a later breakflow termination.

These effects are conservative for the SGTR radiological consequences WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-336calculation, and thus, lower decay heat was not considered.

Similarly, the maximum AFW and SIenthalpies were used. The following changes were made to the related assumptions used in themargin to overfill analyses:

-The 1971+20%

ANS decay heat model specified by Reference 2 was used for theseanalyses.

-Maximum AFW enthalpy is conservative consistent with Reference

2. For this analysis, themaximum AFW enthalpy of 96.0 Btu/lbm was modeled.-Maximum SI enthalpy is conservative consistent with Reference
2. For this analysis, themaximum SI enthalpy of 73.91 Btu/lbm was modeled.3. Flashing FractionWhen calculating the fraction of break flow that flashes to steam, 100 percent of the break flowwas assumed to come from the HL side of the break. Because the tube rupture flow actuallyconsists of flow from the HL and CL sides of the SG, the temperature of the combined flow willbe less than the HL temperature and the flashing fraction would be correspondingly lower. Thus,this assumption is conservative.

2.7.3.1.2.4 Plant InputThe significant WCGS input is the same as modeled in the margin to overfill analyses except for the AFWflow. It was assumed that the minimum AFW flow (285 gpm/SG) was delivered to the SGs following RTand LOOP with a maximum delay (60 seconds).

The maximum purge volume (138.4 ft3) was modeled todelay delivery of cold AFW to the SGs and maximize steam release.

Flow to the ruptured SG continued atthis rate until it was isolated by the operators.

Flow to the intact SGs was throttled to maintain the levelbelow 50 percent NRS.2.7.3.1.2.5 Operator Action TimesThe major operator actions required for the recovery from an SGTR are discussed in Section 2.7.2.1.2.5, and the operator action times used for the analyses are presented in Table 2.7.3-1.

With the exception ofthe time to isolate AFW flow to the ruptured SG, the operator action times assumed for the margin tooverfill analyses were also used for the radiological consequences analyses.

Earlier AFW isolation resultsin higher releases, so it was assumed that AFW flow to the ruptured SG was isolated when level in the SGreached the WCGS required level (but not before 8 minutes because earlier isolation is considered unrealistic).

Assuming a minimum of 8 minutes from event initiation until AFW isolation used in theinput to dose analyses is not a critical operator action time, and does not impose a requirement on theoperators.

This time constraint was included to avoid unrealistic AFW isolation times.For the radiological consequences

analyses, the ARV on the ruptured SG was assumed to fail open at thetime the ruptured SG is isolated.

Before proceeding with the recovery operations, the failed-open ARV onthe ruptured SG was assumed to be isolated by locally closing the associated block valve. An operator canlocally close the block valve for the ARV on the ruptured SG within 30 minutes after the failure.

Thus, itWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-337was assumed that the ruptured SG ARV was isolated at 30 minutes after the valve is assumed to fail open.After the ruptured SG ARV was isolated, an additional delay time of 10 minutes (Table 2.7.3-1) wasassumed before initiation of the RCS cooldown.

The cooldown was performed using the ARVs on allthree of the intact SGs. The cooldown target temperature was selected based on the ruptured SG pressure.

As specified in the WCGS EOPs for SGTR, this pressure must be above 275 psig.2.7.3.1.2.6 Mass Release Calculations The mass releases were determined for use in evaluating the offsite and control room radiological consequences of the SGTR using the methodology of Reference

3. The steam releases from the rupturedand intact SGs, and primary to secondary break flow into the ruptured SG and the associated flashingfraction, were determined for the period from accident initiation until break flow termination and frombreak flow termination to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the accident.

In the RETRAN analyses, the SGTR recovery actions in the WCGS EOPs were simulated until thetermination of primary to secondary break flow. After the primary to secondary break flow is terminated, the operators will continue the SGTR recovery actions.

The plant is then cooled and depressurized to coldshutdown conditions.

In accordance with the methodology in Reference 3, it was assumed that theoperators perform the post-SGTR cooldown using steam dump to the atmosphere.

This method results ina conservative evaluation of the long-term releases for use in the radiological consequences analysescompared to the other cooldown methods in the WCGS EOPs. This procedure for depressurizing theruptured SG was assumed even though the RETRAN analyses performed to calculate releases up untilbreak flow termination assumed ruptured SG ARV isolation.

The high level actions for the post-SGTR cooldown method using steam dump are discussed below.I. Prepare for Cooldown to Cold ShutdownThe initial steps to prepare for cooldown to cold shutdown will be continued if they have notalready been completed.

A few additional steps are also performed prior to initiating cooldown.

These include isolating the CL SI accumulators to prevent unnecessary injection, energizing pressurizer heaters as necessary to saturate the pressurizer water and to provide for better pressurecontrol, and ensuring shutdown margin in the event of a potential boron dilution due to in-leakage from the ruptured SG.2. Cooldown RCS to RHRS Temperature The RCS is cooled by releasing steam from the intact SGs similar to a normal cooldown.

Becauseall immediate safety concerns have been resolved, the cooldown rate should be maintained lessthan the maximum allowable rate of 100lF/hr.

The preferred means for cooling the RCS is viasteam dump to the condenser, because this minimizes the radiological releases and conserves FWsupply. The ARVs on the intact SGs can also be used if steam dump to the condenser isunavailable.

Because a LOOP is assumed, it is assumed that the cooldown is performed usingsteam dump to the atmosphere via the ARVs on the intact SGs. When the RHRS operating temperature is reached, the cooldown is stopped until RCS pressure can also be decreased.

Thisensures that pressure/temperature limits will not be exceeded.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3383. Depressurize RCS to RHRS PressureWhen the cooldown to RHRS temperature is completed, the pressure in the ruptured SG isdecreased by releasing steam from the ruptured SG. It was assumed that the ruptured SG isdepressurized by releasing steam via the ARV. As the ruptured SG pressure is reduced, the RCSpressure is maintained equal to the pressure in the ruptured SG in order to prevent excessive in-leakage of secondary side water or additional primary to secondary break flow. Although normalpressurizer spray is the preferred means of RCS pressure

control, auxiliary spray or a pressurizer PORV can be used to control RCS pressure if pressurizer spray is not available.
4. Cooldown to Cold ShutdownWhen RCS temperature and pressure have been reduced to the RHRS in-service values, RHRScooling is initiated to complete the cooldown to cold shutdown.

When cold shutdown conditions are achieved, the pressurizer can be cooled to terminate the event.2.7.3.1.2.7 Acceptance CriteriaThe analyses were performed to calculate the mass transfer data for input to the radiological consequences analyses.

As such, no acceptance criteria are defined.

The results of the analyses were usedas input to the radiological consequences analyses.

2.7.3.1.3 Description of Analyses and Evaluations The RETRAN results for the limiting input to dose analysis are described below. The limiting case withrespect to the input to dose considered operation at the maximum operating temperature (588.40F), withthe maximum main Tfeed (448.60F), the maximum SGTP level (10 percent),

and the failure of the ARV onthe ruptured SG in the full open position when the operator closes the MSIV. The sequences of events forthese transients are presented in Table 2.7.3-2.Following the tube rupture, water flowed from the primary into the secondary side of the ruptured SGbecause the primary pressure was greater than the SG pressure.

In response to this loss of coolant,pressurizer level decreased (Figure 2.7.3-1).

The RCS pressure (represented by the pressurizer pressure) also decreased (Figure 2.7.3-2) as the steam bubble in the pressurizer expanded.

As the RCS pressuredecreased due to the continued primary to secondary break flow, automatic RT occurred on an OTAT tripsignal.After RT, core power rapidly decreased to decay heat levels. The turbine stop valves closed and steamflow to the turbine was terminated.

The steam dump system is designed to actuate following RT to limitthe increase in secondary

pressure, but the steam dump valves remained closed due to the loss ofcondenser vacuum resulting from the assumed LOOP at the time of RT. Thus, the energy transfer from theprimary system caused the secondary side pressure to increase rapidly after RT (Figure 2.7.3-3) until theSG ARVs lift to dissipate the energy (Figure 2.7.3-5).

As a result of the assumed LOOP, main FW flowwas assumed to be terminated and AFW flow was assumed to be automatically initiated following RT.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-339The RCS pressure and pressurizer level continued to decrease after RT as energy transfer to the secondary system shrunk the RCS and the tube rupture break flow continued to deplete primary inventory.

Thedecrease in RCS inventory resulted in a low pressurizer pressure SI signal. The SI flow increased the RCSinventory and the RCS pressure trended toward the equilibrium value where the SI flow rate would equalthe break flow rate.AFW flow to the ruptured SG was isolated when the ruptured SG level reached 6 percent NRS, and theruptured SG MSIV was closed at 8 minutes after RT.The ruptured SG ARV was assumed to fail open when the MSIV was closed at the time the ruptured SGlevel reached 6 percent NRS. The failure caused the ruptured SG to depressurize

rapidly, which resultedin an increase in primary to secondary break flow. The depressurization of the ruptured SG increased thebreak flow and energy transfer from primary to secondary, which resulted in RCS pressure andtemperature decreasing more rapidly than in the margin to overfill analyses.

The ruptured SGdepressurization caused a cooldown in the intact SGs loops. The operators identified that the ruptured SGARV had failed open and closed the associated block valve 30 minutes after the failure.

Once the rupturedSG ARV block valve was closed, the ruptured SG pressure began to increase (Figure 2.7.3-3).

Theruptured SG pressure was confirmed to be above 275 psig at all times in the transient.

This was alsoconfirmed for transients run with less limiting mass transfer

results, but greater ruptured SG pressurereductions.

The lowest ruptured SG pressure for all cases analyzed was greater than 360 psia.After the block valve for the ruptured SG ARV was closed, a 10-minute operator action time was imposedprior to initiating the cooldown.

The ARVs on all three of the intact SGs were opened at approximately 60 minutes for the RCS cooldown (Figure 2.7.3-5).

The depressurization of the ruptured SG due to thefailed-open ARV affected the RCS cooldown target temperature.

The target temperature was determined based upon the pressure in the ruptured SG at the time the cooldown was initiated.

The cooldown wascontinued until the cooldown termination temperature obtained from WCGS EOPs was reached.

Whenthis condition was satisfied, the operators closed the ARVs to terminate the cooldown.

The cooldownensured that there would be adequate subcooling in the RCS after the subsequent depressurization of theRCS to the ruptured SG pressure.

The reduction in the intact SG pressure required to accomplish thecooldown is shown in Figure 2.7.3-3.

The pressurizer level and RCS pressure also decreased during thiscooldown process due to shrinkage of the RCS (Figure 2.7.3-1).

The ARVs on the intact SGs also automatically opened to maintain the prescribed RCS temperature toensure that subcooling was maintained.

When the ARVs were opened, the increased energy transfer fromthe RCS to the secondary system also aided in the depressurization of the RCS to the ruptured SGpressure after the SI flow was terminated.

After termination of the cooldown, a 5-minute operator action time was imposed prior to the RCSdepressurization.

In these analyses, the RCS depressurization was terminated when the RCS pressure wasreduced to less than the ruptured SG pressure and the pressurizer level was above the required value,because there was adequate subcooling margin and the high pressurizer level setpoint was not reached.The RCS depressurization is shown in Figure 2.7.3-2.

The depressurization reduced the break flow(Figure 2.7.3-4) and increased SI flow to refill the pressurizer (Figure 2.7.3-1).

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-340After termination of the depressurization, a 4-minute operator action time was imposed prior to SItermination.

The SI flow was terminated at this time because the requirements for SI termination weresatisfied.

(RCS subcooling was greater than the required allowance for subcooling uncertainty, minimumAFW flow was available or at least one intact SG level was in the narrow range, the RCS pressure wasstable or increasing, and the pressurizer level was greater than the required value.) After SI termination, the RCS pressure began to decrease (Figure 2.7.3-2).

2.7.3.1.3.1 Calculation of Mass ReleasesThe operator actions for the SGTR recovery up to the termination of primary to secondary break flowwere simulated in the RETRAN analyses.

Thus, the steam releases from the ruptured and intact SGs alongwith the break flow into the ruptured SG were determined from the RETRAN results for the period fromthe initiation of the accident until the break flow was terminated.

Following the termination of break flow, it was assumed that the RCS and intact SG conditions weremaintained stable until the cooldown to cold shutdown was initiated.

The ARVs for the intact SGs werethen assumed to be used to start to cool down the RCS to the RHRS operating temperature of 350'F, atthe maximum allowable cooldown rate of 100°F/hr.

The RCS cooldown was assumed to continue until the RHRS operating temperature of 350'F wasreached.

Depressurization of the ruptured SG was then assumed to be performed immediately following the completion of the RCS cooldown.

The ruptured SG was assumed to be depressurized to the RHRSoperating pressure (using a bounding value of 375 psia) via steam release from the ruptured SG ARV. Thismaximizes the steam release from the ruptured SG to the atmosphere, which is conservative for theevaluation of the radiological consequences.

The RCS pressure was also assumed to be reducedconcurrently as the ruptured SG is depressurized.

It was assumed that the RCS cooldown anddepressurization to RHRS operating conditions were completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the accident.

Thesteam releases from break flow termination to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> were determined for the intact SGs from a massand energy balance using the RCS and intact SG conditions at break flow termination and at the RHRSin-service conditions.

The steam released from the ruptured SG from break flow termination to 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />swas determined based on a mass and energy balance for the ruptured SG using the conditions at the timeof break flow termination and saturated conditions at the RHRS operating pressure.

After 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, it was assumed that further plant cooldown to cold shutdown as well as LTC was providedby the RHRS. Therefore, the steam releases to the atmosphere were terminated at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.2.7.3.1.4 Results2.7.3.1.4.1 Results of RETRAN AnalysesThe primary to secondary break flow rate throughout the recovery operations is presented inFigure 2.7.3-4.

The break flow flashing fraction was calculated using the ruptured HL loop temperature (Figure 2.7.3-6).

The intact HL loop temperature is presented in Figure 2.7.3-7.

The flashing fraction ispresented in Figure 2.7.3-8.

The integrated flashed break flow is presented in Figure 2.7.3-9.

The rupturedSG ARV steam release is presented in Figure 2.7.3-5.

The ruptured SG fluid mass is shown inFigure 2.7.3-10 and ruptured SG water volume is shown in Figure 2.7.3-11.

WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3412.7.3.1.4.2 Mass Release ResultsThe mass release calculations were performed using the methodology discussed in Section 2.7.3.1.3.1.

For the time period from initiation of the accident until break flow termination, the releases weredetermined from the RETRAN results for the time prior to RT and following RT. Because the condenser was in service until RT, any radioactivity released to the atmosphere prior to RT would be through thecondenser vacuum exhaust.

After RT, the releases to the atmosphere were assumed to be via the SGARVs.The transfer and release data are presented in Tables 2.7.3-3 and 2.7.3-4.2.7.3.2 Conclusion The analyses performed to calculate the mass transfer data for input to the radiological consequences analyses were completed and the data were tabulated for the limiting cases. The results of the analyseswere used as input to the radiological consequences analyses.

2.7.3.3 References

1. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

April 1999.2. WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,"

August 1987.3. Supplement I to WCAP-10698-P-A, "Evaluation of Offsite Radiation Doses for a SteamGenerator Tube Rupture Accident,"

March 1986.4. NSAL-07-1 1, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology,"

November 2007.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-342WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-342Table 2.7.3-1 Operator Action Times for Design Basis SGTR T/H AnalysesAction TimeOperator action time to isolate TDAFW flow to ruptured SG(" 6% NRSOperator action time to isolate MDAFW flow to ruptured SGt1) Coincident with TDAFW isolation Operator action time to isolate MSIV on ruptured SG 8 minutes from RTOperator action time to identify and isolate the failed-open ARV 30 minutes from AFW flow isolation Operator action time to initiate cooldown 10 minutes from closure of the failed-open SG ARVCooldown Calculated by RETRANOperator action time to initiate depressurization 5 minutes from end of cooldownDepressurization Calculated by RETRANOperator action time to terminate SI following depressurization 4 minutes from end of depressurization Pressure equalization Calculated by RETRANNote:I. Isolation is assumed to occur no earlier than 2 minutes after RT.Table 2.7.3-2 Sequence of Events for Limiting Input to Radiological Consequences AnalysesEvent Time (seconds)

SGTR 100RT (OTAT) and LOOP 152AFW Actuated 212SI Actuated 425Ruptured SG MSIV Closed 632AFW Flow to Ruptured SG Isolated 1202Ruptured SG ARV Fails Open 1202Ruptured SG ARV Block Valve Closed 3002RCS Cooldown Initiated 3602Break Flow Flashing Terminated 3946RCS Cooldown Terminated 4955RCS Depressurization Initiated 5255RCS Depressurization Terminated 5377SI Terminated 5616Break Flow Terminated 7627Time RHRS Takes Over Cooling 43,300WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-343Table 2.7.3-3 Break Flow and Flashed Break FlowTotal Break Flow during Total Flashed Break FlowStart of Period (sec) End of Period (sec) Period (Ibm) during Period (Ibm)0 152 2424 399152 1202 41,847 26381202 3002 92,563 12,2113002 3602 32,388 23963602 3946 17,119 5513946 5255 55,9305255 7627 37,4247627 43,300Table 2.7.3-4 Intact and Ruptured SG Steam Flow to Atmosphere Total Intact SGs Steam Total Ruptured SG SteamFlow to Atmosphere Flow to Atmosphere Start of Period (sec) End of Period (sec) during Period (Ibm) during Period (Ibm)0 152 511,500 171,000152 1202 63,525 24,9721202 3002 0 136,2283002 3602 0 03602 3946 85,734 03946 5255 118,909 05255 7627 89,233 07627 43,300 1,496,300 2300Note:Pre-trip steam releases are through the condenser.

WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3441CiC)GC)CJ0 2000 4000 6000Time (s)8000Figure 2.7.3-1 Pressurizer Level -Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-345WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3452400220020001800-V~)16001400.1200'1000 I0 2000 4000 6000 8000Time (s)Figure 2.7.3-2 Pressurizer Pressure

-Input to Radiological Consequences AnalysisWCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-346Intac t SGs.u Rup t ured SG1) [ý ()njVUU2000-1500-cn1000I "\\I -I I I500-0-02000 4000 6000Time (s)8000Figure 2.7.3-3 Secondary Pressure

-Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-347WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3476050-40--IC-)LEU~)co~C)20-10-0-I I II I II I II I-10I I I I I I I I I I I02000 4000 6000Time (s)8000Figure 2.7.3-4 Primary to Secondary Break Flow -Input to Radiological Consequences AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-348WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-348Ruptured SG....Total Intact SGs1600-14001200-(.9n 1000-E800-0C-nU6o 600-0 2000 4000 6000Time (s)8000Figure 2.7.3-5SG Steam Releases

-Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-349H Hot Leg InletCold Leg Inlet700600-500-0L)0z)co0)H-400-300-200100-'N.........................................................

~-'-'.7I, / AI I II I II I IU02200014000Time (s)060008000Figure 2.7.3-6Ruptured Loop HL and CL Temperatures

-Input to Radiological Consequences AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-350WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-350Hot Leq InIetCo d Leg Inlet-7AF .I L060;U50orLF-400-.-0 ---.' ---- -_-\ , \0-0-0SI I302010U020004000Time (s)60008000Figure 2.7.3-7Intact Loop HL and CL Temperatures

-Input to Radiological Consequences AnalysisWCAP-I 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-351WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-35 10.20.1500(90.10.5E-010 2000 4000 6000Time (s)8000Figure 2.7.3-8 Break Flow Flashing Fraction

-Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3522000015000-10000-K//rCoCO0I5000-/I II I II I II I I002000 4000 6000Time (s)8000Figure 2.7.3-9 Integrated Flashed Break Flow -Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-353EC-)C-n0 2000 4000 6000Time (s)8000Figure 2.7.3-10 Ruptured SG Fluid Mass -Input to Radiological Consequences AnalysisWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-354WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-354AvoR u p eRuptured S(Or.,MnUUUU5000-4000-3~000-(GDO2000-1000-I I II I I020004000Time (s)60008000Figure 2.7.3-11 Ruptured SG Water Volume -Input to Radiological Consequences AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3552.7.4 Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breakswithin the Reactor Coolant Pressure Boundary (USAR Section 15.6.5)2.7.4.1 Post-LOCA Subcriticality 2.7.4.1.1 Technical Evaluation 2.7.4.1.1.1 Introduction Post-LOCA subcriticality sump boron calculations were performed in support of the TM CDSA Program.The methodology used to demonstrate compliance with the requirements of 10 CFR 50.46(b) isdocumented in WCAP-8339 (Reference 1). Reference 1 states that the core will remain subcritical post-LOCA by borated water from the various injected ECCS water sources.

Post-LOCA sump boroncalculations demonstrate the core will remain subcritical upon entering, and during, the sumprecirculation phase of ECCS injection.

Containment sump boron concentration calculations are used todevelop a core reactivity limit that is confirmed as part of the Westinghouse RSE Methodology (Reference 2).2.7.4.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe major input parameters and assumptions used in the sump boron calculations are given inTable 2.7.4-1.

The sump boron concentration model is based on the following assumptions:

The calculation of the sump mixed mean boron concentration assumes minimum mass and minimumboron concentrations for significant boron sources and maximum mass and minimum boron concentration for significant dilution sources.Boron is mixed uniformly in the sump. The post-LOCA sump inventory is made up of constituents thatare equally likely to return to the containment sump; that is, selective holdup in containment is neglected.

The sump mixed mean boron concentration is calculated as a function of the pre-trip RCS conditions.

There are no specific acceptance criteria when calculating the post-LOCA sump boron concentration.

Theresulting sump boron concentration, which is calculated as a function of the pre-LOCA RCS boronconcentration, is reviewed for each cycle-specific core design to confirm that adequate boron exists tomaintain subcriticality in the long-term post-LOCA.

2.7.4.1.1.3 Description of Analyses and Evaluations A post-LOCA subcriticality boron limit curve was developed using Westinghouse methodology.

Providedthat the cycle-specific maximum critical boron concentration remains below the post-LOCA sump boronconcentration limit curve (for all rods out, no Xenon, 68°F-212°F),

the core will remain subcritical post-LOCA and the only heat generation will be that due to the remaining long-lived radioactivity.

Thiscriterion will be evaluated on a cycle-specific basis in accordance with the RSE Methodology (Reference 2).WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3562.7.4.1.1.4 ResultsA post-LOCA subcriticality boron limit curve was developed for the TM CDSA Program.

The post-LOCA subcriticality boron limit curve is shown in Figure 2.7.4-1.2.7.4.1.2 Conclusion A post-LOCA subcriticality sump boron curve was generated.

The methodology used to generate thecurve aids in demonstrating compliance with 10 CFR 50.46(b).

The post-LOCA subcriticality sump boroncurve will be tracked on a cycle-specific basis using the Westinghouse RSE Methodology and will aid indemonstrating continued compliance with 10 CFR 50.46(b).

2.7.4.1.3 References

1. WCAP-8339, "Westinghouse Emergency Core Cooling System Evaluation Model -Summary,"

June 1974.2. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

July 1985.2.7.4.2 Post-LOCA Long-term Cooling2.7.4.2.1 Technical Evaluation 2.7.4.2.1.1 Introduction A post-LOCA LTC analysis was performed for the methodology transition program.

There are twoaspects to an LTC analysis:

6Boric acid precipitation control (BAPC)Long-term decay heat removal (DHR)This analysis satisfies the requirements of 10 CFR 50.46(b),

Item (5). The 10 CFR 50 GDC contribute tosupporting the conclusions that the following requirements are met:(5) Long-term cooling.

After any calculated successful initial operation of the ECCS, thecalculated core temperature shall be maintained at an acceptably low value and decay heatshall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.The injection and sump recirculation ECCS modes are described in USAR Section 6.3: Emergency CoreCooling System.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3572.7.4.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaLong-Term CoolingThe major inputs to the boric acid precipitation calculation include core power assumptions, assumptions for boron concentrations, and water volume/mass assumptions for significant contributors to thecontainment sump. The input parameters used in the WCGS methodology transition boric acidprecipitation calculations are given in Table 2.7.4-2.

The accumulator maximum boron concentration utilized in the safety analysis is 2500 ppm and the RWST maximum boron concentration utilized in thesafety analysis is 2500 ppm.The boric acid precipitation model is based on the following assumptions:

meets USNRC guidance aspresented in Reference 1, and is consistent with the interim methodology reported in Reference 2.Additional detailed input assumptions are given as follows:The boric acid concentration in the core region was computed over time by considering the effectof core voiding on liquid mixing volume.0 The boric acid concentration limit is the experimentally determined boric acid solubility limit asreported in Reference 3 and summarized in Table 2.7.4-3 and Figure 2.7.4-2.

For large breaks,containment back pressure is not credited and the RCS is assumed atmospheric.

The boric acidsolubility limit credits an increased boiling point of 21 8°F (boiling point of saturated boric acidsolution under atmospheric conditions).

For break sizes where the RCS pressure might remainelevated (or instances where RCS depressurization is not complete),

the boric acid solubility limitunder atmospheric conditions is assumed.* The liquid mixing volume used in the calculation includes 50 percent of the lower plenum asjustified in Reference 2 and Reference 4.* For SBLOCA scenarios, this analysis does not assume a specific start time forcooldown/depressurization emergency procedures.

In reality, it is anticipated that operators willbegin cooldown/depressurization within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the initiation of the event.& The effect of containment sump pH additives on increasing the boric acid solubility limit is notcredited.

0 The boric acid concentration of the makeup containment sump coolant during recirculation is acalculated mixed mean boron concentration.

The calculation of the sump mixed mean boronconcentration assumes maximum mass and maximum boron concentrations for significant boronsources, and minimum mass and maximum boron concentrations for significant dilution sources.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-358USNRC requirements pertaining to the decay heat generation rate for both boric acidaccumulation and decay heat removal (1971 ANS Standard for an infinite operating time with20 percent uncertainty) was considered when performing the boric acid precipitation calculations.

The assumed core power includes a multiplier to address uncertainty as identified by Section L.Aof 10 CFR 50, Appendix K.ECCS recirculation flows are evaluated by comparing the limiting single-failure minimum SIpump flows to the flows necessary to dilute the core and replace core boil-off, thus keeping thecore quenched and amenable to cooling.The acceptance criteria for the LTC analysis are demonstrated by the capability to keep the core cool aftera LOCA and by calculating a time to initiate the BAPC plan with methods, plant design assumptions, andoperating parameters that are consistent with the interim methodology reported in Reference 2.2.7.4.2.1.3 Description of Analyses and Evaluations The LTC phase of the accident begins at the transfer to CL recirculation.

Prior to sump recirculation, corecooling is addressed for the full spectrum of break sizes by the LBLOCA and SBLOCA analyses areas.This satisfies the 10 CFR 50.46 acceptance criteria pertaining to PCT, maximum cladding oxidation, andmaximum hydrogen generation.

DHR checks are performed for the transfer to CL recirculation at transient times based upon full ECCSinjection for the injection phase. Full ECCS injection for the injection phase consists of two residual heatremoval (RHR) pumps (for low head injection),

two SI pumps (for intermediate head injection),

twocentrifugal charging pumps (CCP) (for high head injection),

and two containment spray pumps.Maximizing the flow in the RWST drain down calculation conservatively bounds entry to CLrecirculation.

The earliest entry to CL recirculation for the LBLOCA scenario was conservatively assumed to be 13 minutes.

The earliest entry to CL recirculation for the SBLOCA scenario wasconservatively assumed to be 25 minutes.The adequacy of the CL recirculation flow is checked at the earliest entry to CL recirculation.

Minimumflows are generated assuming the failure of a diesel generator.

One RHR pump takes suction from thecontainment sump, while one IHSI pump and one CCP pump take suction from the RHR pump discharge.

The RHR pump, SI pump, and CCP inject to all four CLs. The spilling line is assumed to be atatmospheric pressure to conservatively minimize the ECCS available for core cooling.The latest acceptable time to enter HL recirculation is determined by the calculated incipient boric acidprecipitation time. Consistent with regulator expectations, the earliest acceptable time to enter HLrecirculation is determined by subtracting 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from the latest acceptable HL recirculation time. Thelatest acceptable HL recirculation time is determined from the calculated incipient boric acid precipitation time. An HL recirculation window provides the operators 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of margin to establish HL recirculation.

The adequacy of the earliest acceptable time to prevent core uncovery and to effectively remove the decayheat generated in the core is then confirmed.

The earliest acceptable time to transfer to HL recirculation isalso confirmed against the entrainment threshold.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-359The transfer to the HL recirculation procedure instructs the operators to transfer the discharge of the SIpump from three CLs to two HLs. The SI pump is stopped to perform this transfer.

The limiting flows forperforming HL recirculation flow checks with respect to DHR occur when the HL recirculation transferprocess is complete.

This is due to the RHR pump providing boosting to the SI pump.The limiting scenario for post-LOCA LTC for HL recirculation consists of one RHR pump taking suctionfrom the containment sump and discharging to all four CLs. The CCP continues to take suction from theRHR pump discharge and inject to all four CLs. The SI pump takes suction from the RHR pumpdischarge and injects to two of the four HLs. The adequacy to effectively remove decay heat at the earlyentry to HL recirculation was checked for both a CL and HL break. The adequacy of the HL flow to haltand reverse the concentration of boric acid in the core was checked at the late entry to HL recirculation for a CL break. An HL break is not a concern with respect to the concentration of boric acid in the core.This is due to CL recirculation flow always being present and providing a forward flushing path.For small breaks, emergency procedures instruct operators to take action to depressurize and cool downthe RCS. Although this depressurization and cooldown process typically begins within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after theevent, the LTC analysis makes no specific assumptions regarding time to depressurize.

However, the post-LOCA LTC safety analyses do assume the rate of cooldown is limited by the operating procedures to100lF/hr.

Depressurization to 120 psia (the threshold for boric acid precipitation concerns) may occurbefore or after the prescribed HL switchover (HLSO) time.2.7.4.2.1.4 ResultsAn incipient boric acid concentration time was calculated.

This value is conservatively rounded down tothe nearest half hour to determine the latest acceptable time to complete the transfer to HL recirculation.

The latest acceptable time to complete the transfer to HL recirculation was determined to be 7.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />sfrom the initiation of the accident.

Figures 2.7.4-3 and 2.7.4-4 show the buildup of boric acid in the corealong with the impacts of the HL dilution flow at the latest entry to HL recirculation.

It is shown that theHL recirculation flow with one CL spilling is adequate to halt and reverse the concentration of boric acidin the core.Consistent with regulator expectations relative to the interim methodology, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of margin wasprovided to the operators to complete the steps necessary to transfer from CL recirculation to HLrecirculation.

The adequacy of the ECCS to effectively remove decay heat at an early entry to HLrecirculation of 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from the initiation of the accident was checked.

HL recirculation flow wasshown to be adequate to effectively remove decay heat for a CL break, whereas CL recirculation flow wasshown to be adequate to effectively remove decay heat for an HL break. The earliest HL recirculation time of 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is well after the entrainment threshold.

Calculations were performed for a condition where HL dilution flow is not established until 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sfrom the initiation of the accident.

This demonstrates the effectiveness of HL dilution flow for thescenario where the RCS remains at an elevated pressure for an extended period. Figure 2.7.4-5 shows theboric acid concentration in the core with the RCS at 120 psia for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> assuming no SG heat removal,no dilution flow, and no benefit of reduced steaming due to SI subcooling.

At 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the boric acidconcentration is still below the boric acid solubility limit at the saturation temperature of concentrated boric acid associated with 120 psia.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-360Figure 2.7.4-5 shows HL flow at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with the RCS at saturation conditions.

The RCS is then cooled(with corresponding depressurization) at the maximum cooldown rate of 1 00°F/hr.

It is shown that thecore boric acid concentration is still maintained below the incipient boric acid precipitation limit at thesaturation temperature of concentrated boric acid at the associated pressure.

A higher core region pressure has two significant effects on the calculation of the incipient precipitation time. A higher pressure would decrease core voiding and increase the available mixing volume. With nocredit for subcooling, a higher pressure would increase the core boil-off due to the heat of vaporization decreasing with increasing pressure and thus increase the rate of concentration of boric acid in the core.Loop seal refilling would be significant to the calculations only if the loop seal closure was sustained.

However, neither LOCA ECCS evaluation models nor observations during the Rig-of-Safety Assessment tests (Reference
5) predict sustained loop seal closure, but instead predict cyclic loop seal refilling andclearing.

Cyclic loop seal refilling/clearing would promote mixing in the vessel by forcing liquid from thecore region to the lower plenum and downcomer.

Effective mixing resulting from this type of oscillatory behavior was observed in the modified VEERA test facility (Reference 6).In summary, the WCGS post-LOCA BAPC calculations used a conservative methodology to establish a6.5 to 7.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> timeframe to realign the ECCS to provide flushing flow to the HLs. Flushing flow to theHLs provides effective core dilution to halt and reverse the concentration of boric acid in the core prior toreaching the boric acid precipitation limit. This realignment addresses the requirements of10 CFR 50.46(b),

Item (5) LTC. ECCS flows during sump recirculation were shown to be sufficient toremove decay heat after a LOCA.The post-LOCA LTC analyses for the methodology transition are applicable with the following modification to the Emergency Procedures (EMGs):The modification of the transfer to an HL recirculation time of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to an early initiation time of6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and a latest completion time of 7.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />s2.7.4.2.2 Conclusion A post-LOCA LTC analysis was completed.

The capability to keep the core cool in the long-term postLOCA was shown and compliance with 10 CFR 50.46(b),

Item (5) was demonstrated.

A BAPC plan wasestablished to keep the core cool post-LOCA and demonstrate compliance with 10 CFR 50.46(b),

Item (5).WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3612.7.4.2.3 References

1. U.S. NRC letter dated November 23, 2005, D.S. Collins to J.A. Gresham, "Clarification of NRCLetter Dated August 1, 2005, Suspension of NRC Approval for Use of Westinghouse TopicalReport CENPD-254-P,

'Post-LOCA Long-Term Cooling Model,' Due to Discovery ofNon-Conservative Modeling Assumptions During Calculations Audit (TAC NO. MB 1365),"(U.S. NRC ADAMS Accession Number ML052930272).

2. Letter dated October 3, 2006, "Slides for the Summary of August 23, 2006 Meeting with thePressurized Water Reactor Owners Group (PWROG) to Discuss the Status of Program toEstablish Consistent Criteria for Post Loss-of-Coolant (LOCA) Calculations,"

(U.S. NRCADAMS Accession Number ML062720565).

3. WCAP-1570, "Literature Values for Selected Chemical/Physical Properties of Aqueous BoricAcid Solutions,"

May 1960.4. Supplement W3FI-2005-0007 dated February 5, 2005, "Supplement to Amendment RequestNPF-38-249, Extended Power Uprate, Waterford Steam Electric

Station, Unit 3," (U.S. NRCADAMS Accession Number ML050400463).
5. NSD-NRC-97-5092, "Core Uncovery Due to Loop Seal Re-Plugging During Post-LOCA Recovery,"

March 1997.6. Tuunanen, J.; Tuomisto, H.; Raussi, P., "Experimental and Analytical Studies of Boric AcidConcentrations in a VVER-440 Reactor During the Long-Term Cooling Period ofLoss-of-Coolant Accidents,"

Nuclear Engineering and Design, Vol. 148, July 1994, pgs. 217-231.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-362Table 2.7.4-1 Subcriticality Analysis Input Parameters Parameter Current ValueRWST Boron Concentration, Minimum (ppm) 2400RWST Delivered Volume, Minimum (gallons) 236,993RWST Temperature, Maximum ('F) 120Accumulator Boron Concentration, Minimum (ppm) 2300Accumulator Liquid Volume, Minimum (gallons) 6122Number of Accumulators 4Accumulator Tank Temperature, Maximum ('F) 120Table 2.7.4-2 LTC Analysis Input Parameters Parameter Current ValueAnalyzed Core Power (MWt) 3565Analyzed Core Power Uncertainty

(%) 2.0Decay Heat Standard 1971 ANS, Infinite Operation, plus 20%(10 CFR 50 Appendix K)H3B03 Solubility Limit (wt %) See Table 2.7.4-3RWST Boron Concentration, Maximum (ppm) 2500RWST Delivered Volume, Maximum (gallons) 419,000RWST Temperature, Minimum ('F) 37Accumulator Boron Concentration, Maximum (ppm) 2500Accumulator Liquid Volume, Maximum (gallons) 26,376Accumulator Tank Temperature, Minimum ('F) 50WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-363Table 2.7.4-3 Boric Acid Solution Solubility Limit DataSolubility Solubility Temperature g H3BO3/100 g of Temperature g H3BO3/100 g ofOC (OF) Solution H20 °C (OF) Solution H20P = I Atmosphere 75 (167)0(32) 2.70 80(176)5(41) 3.14 85(185)10(50) 3.51 90(194)15 (59) 4.17 95(203)20(68) 4.65 100(212)25 (77) 5.43 103.3 (217.9)30 (86) 6.34 P = PSAT35 (95) 7.19 107.8 (226.0)40(104) 8.17 117.1 (242.8)45 (113) 9.32 126.7 (260.1)50 (122) 10.23 136.3 (277.3)55(131) 11.54 143.3 (289.9)60(140) 12.97 151.5 (304.7)65 (149) 14.42 159.4 (318.9)70 (158) 15.57 171 (339.8) = Congruent Melting of H3B03WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3642,2502,1500OP0doan0.E0LeC2,0501,9501,8501,7502,0001,6500 250 500 750 1,000 1,250 1,500 1,750RCS Boron Concentration, Peak Xe (ppm)Figure 2.7.4-1 Post-LOCA Subcriticality Boron Limit CurveWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-365WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-365706050:~4030Saturated Boric Acid SolutionBoiling Point, 218 *FP=PATM P=PSAT..20100050100150200250300350Temperature

(*F)Figure 2.7.4-2 Boric Acid Solubility LimitWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-366Boric Acid Concentration (wt .%)NO HL DILUTION FLOW475 GPM OF HOT LEG DILUTION FLOWBORIC ACID SOLUBILITY LIMIT 29.27Mass Flow Rate (Ibm/sec)

CORE BOILOFFHL SI FLOWWT%000M-0nU)Cr?(D0 2 4 6 8 10Time (hr)Figure 2.7.4-3LBLOCA Boric Acid Concentration Analysis

-Vessel Boric Acid Concentration, Boil-off, and Flushing Flow versus TimeWCAP- 17658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-367WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-367Boric AMass FI50-40--30-.0C-)C- _0o201-cocid Concentration (wi .t- )NO HL DILUTION FLOW475 GPM OF HOT LEG DILUATION FLOWBORIC ACID SOLUBILITY LIMIT 29.27 WT%ow Rate (Ibm/sec)

CORE BOILOFFHL SI FLOW0CD-nCD0 2 4 6Time (hr)10Figure 2.7.4-4SBLOCA Boric Acid Concentration Analysis

-Vessel Boric Acid Concentration, Boil-off, and Flushing Flow versus TimeWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-368Boric Acid Concentration (wt.I %)NO HL DILUTION FLOW450 GPM OF HL DILUTION FLOWBORIC ACID SOL LIMIT W/ 1OOF/HR COOLDOWNTemperature (F )TEMP W/ 1OOF/HR COOLDOWN10080--r350--- --- --- --+\iI\,iI""S IVSI0C-)00300250200-15060-40-----------------------------------------------------------__4CD-0C'DCD20-5'-I-10050I I I9-n5)111213Time (hr)1415Figure 2.7.4-5 Core Dilution at 12 Hours for SBLOCA Pressure HangupWCAP- 17658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3692.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM (USAR SECTION 15.8)2.8.1 Technical Evaluation 2.8.1.1 Introduction The Final ATWS Rule, 10 CFR 50.62(c)(1)

(Reference 1), requires the incorporation of a diverse (fromthe RT system) actuation of the AFW system and turbine trip for Westinghouse-designed plants. Theinstallation of the USNRC-approved AMSAC satisfies this Final ATWS Rule. However, it must also bedemonstrated that the deterministic ATWS analyses that form the basis for this rule and the AMSACdesign remain valid for the plant. This is typically done by confirming that the analyses documented in NS-TMA-2182 (Reference

2) remain valid or by performing new deterministic analyses for theproposed plant state.For the WCGS, the LOL and LONF ATWS events were analyzed to ensure that the analytical basis for theFinal ATWS Rule continues to be met. The LOL and LONF ATWS events are the two most limiting RCSoverpressure transients reported in NS-TMA-2182.

The objective is to show that the ATWS pressure limitof 3200 psig is met for at least 95 percent of the cycle, and therefore the analytical basis for the FinalATWS Rule continues to be met.2.8.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe LOL and LONF ATWS analyses for the WCGS used a plant-specific ATWS model consistent withthe methodology described in NS-TMA-2182.

The following analysis assumptions were used:The nominal and initial conditions were set consistent with the design parameters for an NSSSpower of 3651 MWt.* Westinghouse Model F SG characteristics were used.Consistent with the analysis basis for the Final ATWS Rule (NS-TMA-2182):

-TDF is assumed, no uncertainties are applied to the initial power, RCS average temperature or RCS pressure.

-0 percent SGTP is assumed.

0 percent SGTP is more limiting (that is, results in a higherpeak RCS pressure) for ATWS events.-Control rod insertion was not assumed.-100 percent pressurizer PORV capacity was assumed.WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-370Turbine trip and AFW actuation are modeled to occur at plant-specific times after eventinitiation, using the WCGS AMSAC setpoint and delays.A 25 second AMSAC response time was assumed.

This delay time is added to the time atwhich the SG water mass reaches a mass equivalent to the water level at the AMSAC lowSG water level setpoint of 12 percent of span. The AFW initiation time was determined byadding an additional 60-second delay to account for the time to get the AFW pumps up tospeed, sensor delays, and logic delays. The turbine trip initiation time was determined byadding an additional 5-second delay.A WCGS best-estimate AFW flow of 1200 gpm was assumed.A WCGS-specific MTC of-8 pcm/°F was modeled to bound 95 percent of the cycle. This value isconsistent with that assumed in generic ATWS analyses (Reference 2). The ATWS MTC limit isconfirmed each cycle as part of the reload process.To remain consistent with the basis of the Final ATWS Rule and the supporting analyses documented inNS-TMA-2182,.

the peak RCS pressure reached in the WCGS ATWS evaluations should not exceed theASME B&PV Code, Service Level C stress limit criterion of 3200 psig. This value corresponds to themaximum allowable pressure for the weakest component in the RPV (the nozzle safe end).2.8.1.3 Description of Analyses and Evaluations The LONF and LOL ATWS events were analyzed based on a conservative NSSS power of 3651 MWt.The LOFTRAN computer code (Reference

3) was used to perform the WCGS ATWS analyses, consistent with the analysis basis for the Final ATWS Rule.2.8.1.4 ResultsTo remain consistent with the basis of the Final ATWS Rule (10 CFR 50.62), the peak RCS pressurecalculated in both the LOL and the LONF ATWS analyses shall be less than 3200 psig (or 3215 psia). Thecalculated peak RCS pressure obtained for the LOL and LONF ATWS analyses is 2897.9 psia and3129.0 psia, respectively.

The time sequence of events is documented in Table 2.8.1-1 for the LOL ATWSand in Table 2.8.1-2 for the LONF ATWS. Key transient parameters are shown in Figures 2.8. 1-1through 2.8.1-8 for the LOL ATWS and in Figures 2.8.1-9 through 2.8.1-16 for the LONF ATWS. Basedon these results, it has been demonstrated that the analytical basis for the Final ATWS Rule continues tobe met for operation of the WCGS at an NSSS power level as high as 3651 MWt.WCAP-17658-NP August 2013Licensing Report Revision 0

WESUNGHOUSE NON-PROMETARY CLASS 32-3712.8.2 Conclusion The information related to ATWS has been reviewed and it was concluded that it has adequately accounted for the WCGS plant-specific effects on ATWS. The evaluation has demonstrated that theAMSAC continues to meet the requirements of 10 CFR 50.62. The evaluation has shown that the plant isnot required by 10 CFR 50.62 to have a diverse scram system. Additionally, the evaluation has shown thatthe ATWS pressure limit of 3200 psig will be met for at least 95 percent of the cycle. The MTC assumedin this analysis will continue to be checked on a cycle-specific basis. Therefore, the WCGS is acceptable with respect to ATWS.2.8.3 References

1. 10 CFR 50.62 and Supplementary Information
Package, "Requirements for Reduction of Riskfrom ATWS Events for Light Water-Cooled Nuclear Power Plants."2. NS-TMA-2182, "ATWS Submittal,"

December 1979.3. WCAP-7907-P-A.,

"LOFTRAN Code Description,"

April 1984.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-372Table 2.8.1-1 LOL ATWS Time Sequence of EventsEvent Time (seconds)

Turbine trip occurs 1.0Loss of FW flow initiated 4.0Peak RCS pressure reached [2897.9 psia] 104.7AFW initiated 131.0Table 2.8.1-2 LONF ATWS Time Sequence of EventsEvent Time (seconds)

Loss of FW flow initiated 4.0Turbine trip occurs 61.0Peak RCS pressure reached [3129.0 psia] 90.2AFW initiated 116.0WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-373WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3731.20.80.6<D 0.40.201.2o 1x< 0.8t0.60.40C-) n.90 50 100 150 200 250Time (s)Figure 2.8.1-1 Nuclear Power versus Time for LOL ATWS3000 50 100 150 200 250Time (s)Figure 2.8.1-2 Core Heat Flux versus Time for LOL ATWS300WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-374WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-374CL.Co0 50 100 150 200 250Time (s)Figure 2.8.1-3 RCS Pressure versus Time for LOL ATWS3001-1Cl)M00 50 100 150 200 250Time (s)Figure 2.8.14 Pressurizer Water Volume versus Time for LOL ATWS300WCAP-1 7658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-375WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-375LL0~F-Q)0 50 100 150 200 250Time (s)Figure 2.8.1-5 Vessel Inlet Temperature versus Time for LOL ATWS3000C-)V.)0 50 100 150 200 250Time (s)300Figure 2.8.1-6 RCS Flow versus Time for LOL ATWSWCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-376WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-376C)C/)CDV)C-DC/")0 50 100 150 200 250Time (s)Figure 2.8.1-7 SG Pressure versus Time for LOL ATWS3000 50 100 150 200 250Time (s)Figure 2.8.1-8 SG Mass versus Time for LOL ATWS300WCAP- 17658-NP August 2013WCAP- 17658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-377WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-377i0::00~':1)0 50 100 150 200 250Time (s)Figure 2.8.1-9 Nuclear Power versus Time for LONF ATWS3000 50 100 150 200 250Time (s)Figure 2.8.1-10 Core Heat Flux versus Time for LONF ATWS300WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-378WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-378a,-C/)C/)C--)0 50 100 150 200 250 300Time (s)Figure 2.8.1-11 RCS Pressure versus Time for LONF ATWS2,EC')0 50 100 150 200 250Time (s)Figure 2.8.1-12 Pressurizer Water Volume versus Time for LONF ATWS300WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-379EC',a)0 50 100 150 200 250 300Time (s)Figure 2.8.1-13 Vessel Inlet Temperature versus Time for LONF ATWS0c-?O-0050 100 150 200Time (s)Figure 2.8.1-14 RCS Flow versus Time for LONF ATWS250300WCAP- 17658-NPLicensing ReportAugust 2013Revision 0I WESTINGHOUSE NON-PROPRIETARY CLASS 32-380WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-380CLCnCoa,C-oCo0 50 100 150 200 250Time (s)Figure 2.8.1-15 SG Pressure versus Time for LONF ATWS3000 50 100 150 200 250Time (s)Figure 2.8.1-16 SG Mass versus Time for LONF ATWS300WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3812.9 RADIOLOGICAL DOSESThe WCGS radiological consequences analyses have been performed to follow Regulatory Guide (RG) 1.183 (Reference

1) which provides guidance on the application of Alternative SourceTerm (AST) methodology, as allowed by 10 CFR 50.67. The AST methodology is being used to calculate the offsite, control room, and Technical Support Center radiological consequences for WCGS to supportthe CDSA TM Program.

The following accidents are analyzed:

0 MSLB* LOAC* Locked rotor0 Rod ejection* Letdown line break* SGTR* LOCA0 Waste gas decay tank failure* Liquid waste tank failure0 Fuel handling accident (FHA)Detailed discussion of the input parameters, assumptions, event descriptions, acceptance

criteria, analysisresults, and conclusions for each accident is presented in Section 4.3 of Enclosure VI of this LAR.References
1. Regulatory Guide 1. 183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,"

July 2000.2.10 INSTRUMENT UNCERTAINTIES 2.10.1 Reactor Trip System/Engineered Safety Feature Actuation Setpoint Uncertainties 2.10.1.1 Introduction and Background The TS Reactor Trip System (RTS)/Engineered Safety Feature Actuation System (ESFAS) and Loss ofPower Diesel Generator Start setpoints have been reviewed, and TS changes have been identified consistent with the Westinghouse setpoint methodology defined in WCAP- 17746-P "Westinghouse Setpoint Methodology as Applied to the Wolf Creek Generating Station,"

Revision 0 (Reference 1). Forthe WCGS TSs, the Allowable Values have been removed and the Nominal Trip Setpoints, which are theWestinghouse defined Limiting Safety System Setting required by 10 CFR 50.36, have been added.Modified footnotes in TSTF-493 Revision 4 Option A have been added to the TS.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3822.10.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe RTS/ESFAS and Loss of Power Diesel Generator Start uncertainty calculations for WCGS wereperformed based on WCGS specific instrumentation, plant calibration procedures, and verified designinputs. The acceptance criterion for the RTS/ESFAS and Loss of Power Diesel Generator Start setpoints isthat the calculated margin, defined as Total Allowance

-Channel Statistical Allowance (defined inReference 1), is >_ 0 % of span.2.10.1.3 Description of Analyses and Evaluations The setpoint analysis uses the square root sum of the squares technique to combine the uncertainty components of an instrument channel in an appropriate combination of those components, or groups ofcomponents, that are statistically independent.

Those uncertainties that are statistically dependent arearithmetically summed to produce statistically independent groups that are then statistically combined.

The methodology used for WCGS is defined in Reference

1. Uncertainty calculations were performed forthe RTS/ESFAS and the Loss of Power Diesel Generator Start functions based on the design inputparameters and the Nominal Trip Setpoints noted in the revised TS Tables 3.3.1-1 and 3.3.2-1 andLCO 3.3.5 and confirmed that the acceptance criterion was satisfied for each protection function.

Thecalculations for the RTS/ESFAS and Loss of Power Diesel Generator Start functions are provided inWCAP-17602-P "Westinghouse Setpoint Calculations for the Wolf Creek Generating Station Control,Protection and Indication Systems,"

Revision 0 (Reference 3).2.10.1.4 ResultsThe results of the setpoint calculations for each RTS/ESFAS and Loss of Power Diesel Generator Startfunction are provided in Reference

3. The Westinghouse setpoint calculations are based on using theNominal Trip Setpoint as the Limiting Safety System Setting.

For WCGS TS Tables 3.3.1-1 and 3.3.2-1and LCO 3.3.5, the Allowable Values are to be removed and the Nominal Trip Setpoints are to be added.This is consistent with the Westinghouse definition of the Limiting Safety System Setting required in 10CFR 50.36 and the Westinghouse setpoint methodology.

The footnotes defined in TSTF-493 Revision 4Option A are to be added to the TS. The Allowable Values for WCGS are to be replaced with As Left andAs Found operability criteria consistent with TSTF-493 Revision 4, Option A requirements.

The As Leftand As Found operability criteria for the RTS/ESFAS and Loss of Power Diesel Generator Startinstrumentation have been defined in Reference 3 and are consistent with the Westinghouse setpointmethodology defined in Reference 1.2.10.1.5 Conclusions All RTS/ESFAS functions meet the acceptance criterion.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTfNGHOUSE NON-PROPRIETARY CLASS 32-3832.10.1.6 References

1. WCAP- 1 7746-P, Revision 0, "Westinghouse Setpoint Methodology as Applied to the Wolf CreekGenerating Station."
2. TSTF-493, Revision 4 "Clarify Application of Setpoint Methodology for LSSS Functions,"

April 2010.3. WCAP- 1 7602-P, Revision 0, "Westinghouse Setpoint Calculations for the Wolf Creek Generating Station Control, Protection and Indication Systems."

2.10.2 Initial Condition Uncertainties 2.10.2.1 Introduction and Background Initial condition uncertainties are conservative steady-state instrumentation measurement uncertainties that are applied to nominal parameter values to obtain conservative initial conditions for use in the safety(accident) analyses.

The following initial condition parameter uncertainties were calculated for WCGS foruse in the accident analyses to assess the safety analyses acceptability.

  • Daily Power Measurement
  • RCS Flow Measurement
  • Pressurizer Pressure Control* Tavg Control* SG Level Control* Pressurizer Level ControlThe results of the uncertainty calculations are presented in WCAP- 17602-P "Westinghouse SetpointCalculations for the Wolf Creek Generating Station Control, Protection and Indication Systems,"

Revision0 (Reference 1). The final uncertainty calculations confirm that the initial condition values utilized in thesafety analysis are bounding.

2.10.2.2 Input Parameters, Assumptions, and Acceptance CriteriaThe initial condition uncertainty calculations for WCGS were performed based on WCGS specificinstrumentation, plant calibration procedures, and verified design inputs. The acceptance criteria for theinitial condition uncertainties are that the calculated uncertainty must be less than or equal to theuncertainty values used in the safety analyses.

2.10.2.3 Description of Analyses and Evaluations The initial condition uncertainty analysis uses the square root sum of the squares technique to combinethe uncertainty components of an instrument channel in an appropriate combination of those components, or groups of components, that are statistically independent.

Those uncertainties that are statistically dependent are arithmetically summed to produce statistically independent groups that are then statistically WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-384combined.

The Westinghouse setpoint methodology is defined in WCAP- 17746-P "Westinghouse Setpoint Methodology as Applied to the Wolf Creek Generating Station,"

Revision 0 (Reference 2).2.10.2.4 ResultsThe results of this analysis are summarized in Reference 1.2.10.2.5 Conclusions As demonstrated by the calculations provided in Reference 1, the acceptance criteria have been satisfied.

2.10.2.6 References

1. WCAP-17602-P, Revision 0, "Westinghouse Setpoint Calculations for the Wolf Creek Generating Station Control, Protection and Indication Systems."
2. WCAP- 1 7746-P, Revision 0, "Westinghouse Setpoint Methodology as Applied to the Wolf CreekGenerating Station."

2.11 CONTROL SYSTEMS ANALYSIS2.11.1 NSSS Pressure Control Component Sizing (USAR Sections 5.4, 7.7, & 10.4.4)2.11.1.1 Technical Evaluation 2.11.1.1.1 Introduction The following pressure control components were evaluated for the TM Program.

The purpose of thisevaluation is to ensure that the NSSS pressure control system component installed capacities are adequateand meet the plant design basis sizing requirements.

  • Pressurizer PORVs* Pressurizer spray valves* Pressurizer heaters* Steam dump valvesThe analyses were performed to envelop the window of operating conditions, which include thefull-power Ta,,g window, the full-power Tfeed window, and 0 to 10 percent average SGTP levels. Theanalyses utilized the potential MUR power level and conditions; the analysis results therefore, bound thecurrent power level conditions.

The pressure control components are described in the USAR, Section 5.4 ("Component and Subsystem Design"),

Section 7.7 ("Control Systems not Required for Safety"),

and Section 10.4.4 ("Turbine BypassSystem").

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3852.11.1.1.2 Input Parameters, Assumptions, and Acceptance CriteriaPressurizer PORVsThe pressurizer PORV sizing analysis was performed at the MUR operating conditions shown inSection 1.1. The analysis at MUR power conditions bounds current power conditions.

The pressurizer PORV sizing analysis was performed to confirm that the installed PORV capacity is adequate to meet theapplicable sizing criteria for the TM Program.

The analysis included the following key input parameters and assumptions:

  • The design basis transient is modeled as a 50 percent step load rejection from full power.* The analysis is performed at a full power Tavg of 588.4°F (high Tavg), 0 percent SGTP, and a Tfeedof 448.60F, which bounds all other normal operating conditions.

0 The analysis is best estimate and best estimate conditions are assumed except a 4°F Tavguncertainty and 0.6 percent power uncertainty were applied for conservatism.

  • The secondary side water mass was reduced by 10 percent for conservatism.

0 The initial pressurizer pressure is at the nominal pressure of 2250 psia.* The initial pressurizer water level is at the nominal setpoint.

  • There are a total of two pressurizer PORVs, each with a rated capacity of 2 10,000 lbm/hr at2335 psig.0 The NSSS control systems (rod, pressurizer
pressure, pressurizer level, SG level, and steam dumpcontrol systems) are assumed to be operational and functioning as designed.
  • The pressurizer PORV sizing analysis is completed using best estimate BOL fuel reactivity data.BOL parameters have lower differential rod worths and least negative MTC; thus, using BOLparameters yields conservative results and bounds the entire fuel cycle.Acceptance CriteriaThe Westinghouse sizing basis for the pressurizer PORVs is to prevent the pressurizer pressure fromreaching the high pressurizer pressure RT setpoint during the design basis large load rejection transient.

This design basis large load rejection is defined as a 50 percent step load reduction from full power. Thesizing criterion is conservatively met if the maximum pressurizer insurge during the transient is less thanthe total capacity of the PORVs. The sizing basis for the PORVs is documented inUSAR Sections 5.4.13.1 and 7.7.1.5 and is consistent with the Westinghouse sizing basis.WCAP-17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-386Pressurizer Spray ValvesThe pressurizer spray valves sizing analysis was performed at the MUR power conditions shown inSection 1.1. The analysis at MUR power conditions bounds current power conditions.

The pressurizer spray valves sizing analysis was performed to confirm that the installed spray valves capacity is adequateto meet the applicable sizing criteria for the TM Program.

The analysis included the key input parameters and assumptions listed below:0 The design basis transient is modeled as a 10 percent step load decrease from full power.* The analysis is perforned at a full power Tavg of 588.4°F (high Tavg), 0 percent SGTP, and a T'eedof 448.6°F, which bounds all other normal operating conditions.

  • The analysis is best estimate and best estimate conditions are assumed except a 4°F Taguncertainty and 0.6 percent power uncertainty were applied for conservatism.
  • The secondary side water mass was reduced by 10 percent for conservatism.
  • The initial pressurizer pressure is at the nominal pressure of 2250 psia.0 The initial pressurizer water level is at the nominal setpoint applicable to the full power Tavgoperating conditions.
  • There are two pressurizer spray valves with a combined total capacity of 896 gpm.* The NSSS control systems (rod, pressurizer level, pressurizer
pressure, and SG level) areassumed to be operational and functioning as designed.

The steam dump system is not actuatedfor load changes less than 10 percent; therefore, steam dump is not modeled for this analysis.

  • The pressurizer spray valve sizing analysis is completed using best estimate BOL fuel reactivity data. BOL parameters have lower differential rod worths and least negative MTC; thus, usingBOL parameters yields conservative results and bounds the entire cycle.Acceptance CriteriaThe Westinghouse sizing basis for the pressurizer spray valves is to prevent the pressurizer pressure fromreaching the pressurizer PORV actuation setpoint of 2335 psig (2350 psia) for the design basis 10 percentstep load decrease transient.

The sizing basis for the spray valves is documented in USARSection 5.4.10.3.4 and is consistent with the Westinghouse sizing basis.Pressurizer HeatersThe evaluation included the following key input parameters and assumptions:

  • The total backup heater design capacity is 1384 kW and the total proportional heater designcapacity is 416 kW. This provides a total heater design capacity of 1800 kW.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-387The currently installed backup and proportional heater capacities are 1315 kW and 347 kW,respectively.

The total pressurizer internal volume is 1800.0 ft3.Acceptance CriteriaThe pressurizer heater capacity should be sufficient to 1) provide heat during cold plant startup,2) regulate pressure to avoid reaching applicable RT and ESFAS setpoints during Condition I transients, and 3) counteract the steady-state heat loss that occurs within the pressurizer to maintain steady stateoperating pressure.

The design basis pressurizer heater capacity to meet these requirements is 1 kW ofheater capacity per cubic foot of pressurizer volume. Additionally, limiting condition for operation (LCO)3.4.9.b requires two groups of backup heaters to be operable with the capacity of each group > 150 kW toensure the normal operating pressure can be maintained after accounting for heat losses.Steam Dump ValvesThe steam dump valves sizing analysis was performed at the MUR power conditions shown inSection 1.1. The analysis at MUR power conditions bounds current power conditions.

The steam dumpvalves sizing analysis was performed to confirm that the installed steam dump system capacity isadequate to meet the applicable sizing criteria for the TM Program.Two design basis transients were analyzed for the steam dump capacity sizing: the 50 percent loadrejection from full power transient and the plant trip from full power transient.

The 50 percent loadrejection transient was modeled as a 50 percent step rejection in turbine load from full power and theplant trip was modeled as a turbine trip followed by a RT from full power. The 50 percent load rejection transient was analyzed as part of the margin to trip analysis and the details of the input parameters andassumptions for this transient are defined in Section 2.11.2. The analysis of the plant trip transient included the following key input parameters and assumptions:

0 A two second delay is conservatively assumed for RT on turbine trip.a The analysis is performed at a full power Tavg of 588.4°F (high Tavg), 0 percent SGTP, and a Tfeedof 448.6°F, which bounds all other normal operating conditions.

  • The analysis is best estimate and best estimate conditions are assumed except a 4°F Tavguncertainty and 0.6 percent power uncertainty were applied for conservatism.
  • The secondary side water mass was reduced by 10 percent for conservatism.

0 The SG PORVs are not modeled in the analysis.

  • The NSSS control systems (rod, pressurizer level, pressurizer
pressure, and SG level) areassumed to be operational and functioning as designed.

The steam dump system is operating inthe RT controller mode.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-388Acceptance CriteriaThe Westinghouse sizing basis of the steam dump system is to enable the plant to survive a 50 percentload rejection without generating a RT or challenging the MSSV, and by being able to survive a turbinetrip with a RT without challenging the MSSV actuation setpoint of 1185 psig (1200 psia). The sizing basisfor the steam dump system is documented in USAR Section 10.4.4.1 "Power Generation Design BasisThree" and is consistent with the Westinghouse sizing basis.2.11.1.1.3 Description of Analyses and Evaluations Transients for the pressure control component sizing analysis were analyzed using the LOFTRANcomputer code (Reference 1). This computer code is a system-level program code that models the overallNSSS, including detailed modeling for control and protection systems.

A LOFTRAN computer modelwas developed for the WCGS. The key input parameters and assumptions for each transient are given inSection 2.11.1.1.2.

No computer analyses were performed specifically for the pressurizer heater sizing evaluation; however,the operational and margin to trip analyses in Section 2.11.2 were performed using the current heatercapacity and show that an adequate transient response is maintained with the reduced pressurizer heatercapacity.

Additionally, the impact of the reduced heater capacity is evaluated in Section 2.11.1.1.4.

2.11.1.1.4 ResultsPressurizer PORVsThe results of the analysis show a maximum pressurizer pressure of 2351 psia, which is less than the highpressurizer pressure RT setpoint of 2400 psia. The results of the analysis also show that the maximumpressurizer insurge flow rate is 152,900 lbm/hr compared to the installed PORV capacity of420,000 lbrm/hr.The calculated peak pressurizer pressure on a design basis 50 percent load rejection was less than the highpressure RT setpoint.

Therefore, the PORVs have sufficient relief capacity to avoid a RT on highpressurizer pressure for the design basis load rejection.

Similarly, it was shown that the required PORVcapacity (i.e., the pressurizer insurge) during the transient was less than the total installed capacity.

ThePORVs are therefore adequately sized for the TM program.Pressurizer Spray ValvesThe results of the 10 percent step load decrease from full power show a maximum pressurizer pressure of2330 psia, which is less than the pressurizer PORV actuation setpoint of 2350 psia. Because the peakpressurizer pressure was less than the PORV actuation setpoint of 2350 psia, the total installed sprayvalves capacity of 896 gpm is adequate to avoid actuation of the pressurizer PORV during a 10 percentstep load decrease from full power transient.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-389Pressurizer HeatersThe total design heater capacity of 1800 kW meets the design criteria of 1 kW/ ft3 of pressurizer volume;however, as discussed in Section 2.11.1.1.2, the current heater capacities are 1315 kW for the backupheaters and 347 kW for the proportional

heaters, for a total heater capacity of 1662 kW. Based on thecurrent capacity of the heaters, the criterion of one kilowatt per one cubic foot is not met. Therefore, thesizing evaluation was performed for a total heater capacity of 1662 kW to confirm that the plant responseto the design basis transients would remain acceptable with the reduced heater capacity.

Design basis transients resulting in pressurizer insurges/outsurges such as loadings/unloadings, loadrejections, and RTs show pressurizer pressure changes that are too rapid for the pressurizer heaters tosignificantly influence.

In addition, past analyses have demonstrated only small differences in themaximum/minimum pressurizer pressure when it is assumed that a fraction of the pressurizer heaters areout of service.

Analyses have demonstrated that a reduced heater capacity results in increased times forplant heatup. A reduction in pressurizer heater capacity of this magnitude is acceptable for transient mitigation based on the results of the operational transient analysis in Section 2.11.2. The currentproportional heater capacity of 347 kW is greater than the 300 kW specified in LCO 3.4.9.b; therefore, theproportional heaters are capable of maintaining the normal operating pressure as designed.

It is concluded that a total heater capacity of 1662 kW is acceptable and this conclusion is furtherdemonstrated by the results of the operational transient analyses documented in Section 2.11.2.Steam Dump ValvesThe results of the turbine trip followed by a RT from full power analysis show a maximum SG pressure of1158 psia, which provides 42 psi of margin to the first MSSV lift setpoint of 1200 psia. The results of the50 percent load rejection analysis discussed in more detail in Section 2.11.2 show that acceptable marginis maintained to all applicable RT setpoints, and the first MSSV lift setpoint is not exceeded.

Because the SG pressure was less than the lowest MSSV actuation setpoint of 1200 psia for both designbasis transients and no RT setpoints are reached during the 50 percent load rejection transient, the totalinstalled capacity steam dump system is adequate for the TM Program.2.11.1.2 Conclusions Pressure control component sizing analyses for the pressurizer PORV, pressurizer spray valves,pressurizer

heaters, and steam dump valves were performed using Westinghouse methodology as part ofthe WCGS TM Program.

The results of these analyses showed that the installed capacities of thesecomponents at WCGS are adequate at the current and MUR power levels.2.11.1.3 References

1. WCAP-7907-P-A, "LOFTRAN Code Description,"

April 1984.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3902.11.2 Operational Analysis and Margin to Trip (USAR Section 7.7)2.11.2.1 Technical Evaluation 2.11.2.1.1 Introduction The following transients were analyzed or evaluated for the TM Program.

These transients are listed inUSAR Section 7.7.2 ("Analysis"),

as design basis operational transients that the plant control systemsdescribed in USAR Section 7.7.1 should regulate without actuation of plant safety systems.* 5 percent per minute unit loading and unloading (Condition I)* 10 percent step load decrease (Condition I)* 10 percent step load increase (Condition I)0 Large load rejection (Condition I)Additionally, the turbine trip without a RT transient from the P-9 permissive setpoint was analyzed todemonstrate that adequate margin is maintained to the pressurizer PORVs actuation setpoint(NUREG-0737, item II.K.3.10).

The transients were analyzed to envelop the window of operating conditions that include the full-power Tavg window, the full-power Tfeed window, and 0 to 10 percent average SGTP levels. The analyses wereperformed at the potential MUR power uprate conditions and the results of the analyses bound the currentpower level conditions.

2.11.2.1.2 Input Parameters, Assumptions, and Acceptance CriteriaThe analyses included the following key input parameters and assumptions:

The analyses were performed at the MUR operating conditions shown in Section 1.1. Theanalyses were based on the MUR NSSS power level of 3651 MWt and bound the current NSSSpower level of 3579 MWt. Additionally, the analyses cover a full-power Tag range from 570.71Fto 588.4'F, full-power Tfeed range from 400.00F to 448.60F, and average SGTP levels between0 percent and 10 percent.The plant operational analysis is a best-estimate analysis; therefore, the plant conditions are at thenominal values and instrument uncertainties are not applied.

However, a 0.6 percent powerallowance was applied for conservatism.

All NSSS control systems (reactor, pressurizer

pressure, pressurizer level, SG level, andsteam dump) are assumed to be operational and functioning as designed.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTrNGHOUSE NON-PROPRIETARY CLASS 32-391Best-estimate reactor kinetics data (such as MTC, Doppler power defect, and control rod worth)were used as input to the analysis.

BOL core conditions were used, which are bounding for theentire cycle.The analyses were performed using the currently installed control system settings and the RPSsettings as discussed in the WCAPs referenced in Section 2.10.Acceptance CriteriaThe acceptance criteria for the NSSS control systems are based on GDC 13, which requires thatinstrumentation and control systems be provided to monitor variables and systems over their anticipated ranges during normal operation as well as anticipated operational occurrences, and maintain thesevariables and systems within prescribed operating ranges.There should be adequate operating margin to the relevant RTS and ESFAS setpoints during andfollowing the Condition I (normal operating) transients.

All control system responses should be smoothand stable without diverging oscillations.

In addition to the limiting RTS and ESFAS setpoints, the10 percent step load decrease transient should not challenge the pressurizer PORVs or MSSV liftsetpoints.

The turbine trip without a RT from the P-9 setpoint analysis is performed to demonstrate that adequatemargin is maintained to the pressurizer PORVs actuation setpoint of 2350 psia. The results of this analysisare used to demonstrate that the requirements of NUREG-0737, item II.K.3.10 (Reference

1) are satisfied.

Although it is not a design requirement of the turbine trip without a RT from the P-9 permissive transient, it is desirable to ensure that the MSSVs do not lift during this transient.

This is demonstrated bymaintaining adequate margin to the first MSSV lift setting of 1185 psig (1200 psia).2.11.2.1.3 Description of Analyses and Evaluations The transients were analyzed using the LOFTRAN computer code (Reference 2). This computer code is asystem-level program code that models the overall NSSS, including detailed modeling for control andprotection systems.

A LOFTRAN computer model was developed for the WCGS. The key inputparameters and assumptions for the analyses are given in subsection 2.11.2.1.2.

5 Percent per Minute Unit Loading and Unloading The 5 percent per minute loading and unloading transients are not limiting transients and are enveloped by the 10 percent step load increase and decrease transients, respectively.

Therefore, no specific analyseswere performed for the 5 percent per minute loading and unloading transients.

10 Percent Step Load DecreaseThis transient was analyzed as a step decrease in turbine load from 100 to 90 percent power which boundslower power levels. The primary control systems that mitigate this transient are the reactor control systemand pressurizer pressure control system. The steam dump system is blocked on load decreases less than10 percent.

The 10 percent step load decrease transient should not result in challenges to the pressurizer WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-392PORVs actuation setpoint and the maximum steam pressure should not challenge the first MSSV liftsetpoint.

10 Percent Step Load IncreaseThis transient was analyzed as a step increase in turbine load from 90 to 100 percent power, which boundslower power levels. The primary control systems that mitigate this transient are the reactor control systemand pressurizer pressure control system. The 10 percent step load increase transient should not result in anautomatic actuation of an RTS or ESFAS function.

The limiting RTS and ESFAS functions are the highneutron flux, low pressurizer

pressure, OTAT, OPAT, and low steam line pressure RTS and ESFASsetpoints.

Large Load Rejection The large load rejection is the most severe operational transient and was analyzed as a step decrease inturbine load from 100 to 50 percent power, which bounds lower power levels. The primary controlsystems that mitigate this transient are the reactor control system, pressurizer pressure control system, andsteam dump control system. The steam dump control system maintains the RCS temperature within thecontrol range until a new equilibrium condition is reached.

The RTS functions that are most limiting forthis transient are the OTAT, OPAT, low pressurizer

pressure, and high pressurizer pressure setpoints.

Turbine Trip without Reactor Trip from P-9 SetpointThis transient was analyzed as a step change in steam flow from the P-9 setpoint of 50 to 0 percent power.A turbine trip from 50 percent power bounds all lower power levels. The analysis is performed todemonstrate that the pressurizer PORVs do not lift following a turbine trip without a RT transient from theP-9 permissive setpoint to address NUREG-0737, Item II.K.3.10 (Reference 1). The analysis is best-estimate and credits the reactor control system, pressurizer pressure control system, and steam dumpcontrol systems.2.11.2.1.4 ResultsThe results of the analyses show that the current control system setpoints and RPS settings, as discussed in the WCAPs referenced in Section 2.10, are acceptable for the TM Program and enable the plant tosatisfy the requirements of the design basis operational transients.

Ten Percent Step Load DecreaseFollowing the 10 percent step load decrease, the secondary side steam pressure and temperature initially

increase, resulting in a Tavg and pressurizer pressure increase.

The control system automatically inserts thecontrol rods to restore Tavg to the programmed value. Pressurizer spray restores the pressurizer pressure tothe nominal value.Based on the results, a 10 percent step load decrease transient can be accommodated without challenging the pressurizer PORVs setpoint of 2350 psia and the first MSSV setpoint of 1200 psia. The maximumpressurizer pressure was 2326 psia and the maximum steam line pressure was 1060 psia. This resulted inWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-393minimum margins of 24 psi to the pressurizer PORVs and 140 psi to the nominal MSSV actuation setpoints.

The results indicated that no RTS or ESFAS setpoints were challenged and the control systemresponses were smooth and not oscillatory.

Therefore, the 10 percent step load decrease transient can besuccessfully accommodated for the TM Program.Ten Percent Step Load IncreaseFollowing the 10 percent step load increase, the secondary side steam pressure and temperature initially decrease and the Tavg and pressure also initially decrease.

The control system automatically withdraws thecontrol rods to restore Tavg to the programmed value. Pressurizer heaters restore the pressurizer pressureto the nominal value.The 50 seconds/5 seconds lead/lag compensated steam pressure reached a minimum of 668.2 psia, whichis above the low steam line pressure SI actuation setpoint of 630 psia. Therefore,

.the criterion of notchallenging the low steam line pressure SI on a design basis 10 percent step load increase transient is met.The minimum pressurizer pressure was 2211 psia, which is greater than the low pressure RTS setpoint of1955 psia. The power range neutron flux reached a maximum value of 104.4 percent, which is less thanthe RTS setpoint of 109 percent.

Acceptable margins of 9.68 and 5.0 percent were maintained to theOTAT and OPAT RTS setpoints, respectively.

The results indicated that no RTS or ESFAS setpoints werechallenged and the control system response was stable and not oscillatory.

Therefore, the 10 percent stepload increase transient can be successfully accommodated for the TM Program.Large Load Rejection Following the large load rejection, the secondary side steam pressure and temperature initially

increase, resulting in a Tavg and pressurizer pressure increase.

The steam dump valves open to mitigate the RCStemperature increase and the reactor control system automatically inserts the control rods to decreasereactor power and restore Tavg to the programmed value. The steam dump valves modulate closed as theplant is brought to a new equilibrium condition.

Pressurizer spray and relief valves prevent the pressurefrom reaching the high pressurizer pressure RTS setpoint.

Based on the results, a 50 percent step load rejection can be sustained over the range of operating conditions.

Minimum margins of 6.02 and 8.88 percent of nominal AT were maintained to the OTAT andOPAT RT setpoints, respectively, which is acceptable.

The pressurizer PORVs open and limit the pressureto less than the high pressurizer pressure RTS setpoint of 2400 psia. Following the opening of thepressurizer PORVs, the minimum pressurizer pressure of 2139 psia maintains adequate margin to the lowpressurizer pressure RTS setpoint of 1955 psia.The results indicated that no RTS or ESFAS setpoints were challenged and the control system responsewas stable and not oscillatory.

Therefore, the large load rejection transient can be successfully accommodated for the TM Program.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-394Turbine Trip without Reactor Trip from the P-9 Permissive Following the turbine trip, the secondary side steam pressure and temperature initially

increase, resulting in a Tavg and pressurizer pressure increase.

The steam dump valves open to mitigate the RCS temperature increase and the reactor control system automatically inserts the control rods to decrease reactor powerand Tavg. Pressurizer spray prevents the pressure from reaching the pressurizer PORVs actuation setpoint.

This transient is modeled as a step change in steam flow from the P-9 setpoint power level to 0 percentpower. The analysis was performed from the current P-9 permissive setpoint of 50 percent RTP andcovered the range of operating conditions.

The current P-9 setpoint was found to be acceptable at conditions corresponding to the high full powerTavg of 588.4°F;

however, the results indicated that the pressurizer PORVs actuation setpoint waschallenged for the case corresponding to low full- power Ta,,g (i.e., 570.7°F) conditions.

Further analysesshowed that acceptable results are obtained for the current P-9 setpoint if the full-power Tavg is restricted to no lower than 575°F.Therefore, the current P-9 setpoint of 50 percent RTP is acceptable for plant operation over a restricted full-power Tavg window of 575°F to 588.4°F.

The limiting full-power Tag for the current P-9 setpoint is5750F and the pressurizer pressure increases to a maximum pressure of 2339 psia, which results inacceptable margin of approximately 11 psi to the PORVs actuation setpoint.

The case corresponding tohigh Tavg (588.4°F) conditions resulted in a maximum pressurizer pressure of 2302 psia, which is wellbelow the pressurizer PORVs actuation setpoint.

The maximum SG pressure for all cases analyzed is1071 psia, which is well below the first MSSV lift setting of 1200 psia. The analyses indicate the controlsystem response was smooth during the transient with no oscillatory response noted. Therefore, theturbine trip without RT transient from the P-9 permissive setpoint of 50 percent RTP can be successfully accommodated for the TM Program over a restricted full-power Tavg window of 575°F to 588.4°F.WCNOC will administratively limit full-power Tavg to greater than or equal to 575°F.2.11.2.1.5 Conclusions Plant operational margin to trip analyses were performed using Westinghouse methodology as part of theWCGS TM Program.

The results of these analyses conclude that the plant response is acceptable andsufficient margin exists to the relevant RTS and ESFAS setpoints during the design basis operational transients as described in USAR 7.7.2 at the WCGS for the current and MUR power levels. The results ofthe analyses show that the current control system setpoints and RPS settings, as discussed in the WCAPsreferenced in Section 2.10, are acceptable for the TM Program and enable the plant to satisfy therequirements of the design basis operational transients.

The current P-9 permissive setpoint of 50 percent RTP is acceptable for a restricted full-power Tavgwindow between 575°F and 588.4°F.

The results show that a turbine trip from the P-9 setpoint will notchallenge the pressurizer PORVs actuation setpoint with all NSSS control systems operable in theautomatic mode; therefore, the requirements of NUREG-0737, item II.K.3.10 are satisfied.

The WCGSwill administratively limit full-power Tavg to greater than or equal to 575°F.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3952.11.2.1.6 References

1. NUREG-0737, "Clarification of TMI Action Plan Requirements,"

Item II.K.3.10, ProposedAnticipatory Trip Modification, October 1980.2. WCAP-7907, "LOFTRAN Code Description,"

April 1984.2.12 THERMAL AND HYDRAULIC DESIGN2.12.1 Introduction This section describes the T/H analysis performed to support operation of the WCGS with a corecontaining 17x17 RFA-2 fuel at the nominal conditions described in Table 2.12-1.2.12.2 Input Parameters and Acceptance CriteriaFor the purposes of the WCGS methodology transition, bounding fuel-related safety and designparameters have been chosen. These bounding parameters have been used in the analyses discussed in thissection.

Table 2.12-1 summarizes the current T/H design parameters used in the DNB analyses.

Thelimiting direction for these parameters for DNB is shown in Table 2.12-2. The core inlet temperature usedin the DNB analyses is based on the upper bound of the RCS temperature range for conditions corresponding to a conservatively higher core power.The current licensing basis for T/H design for the WCGS includes the prevention of DNB on the limitingfuel rod with a 95-percent probability at a 95-percent confidence level (95/95) and criteria to ensure fuelcladding integrity.

The licensing basis is documented in USAR Section 4.4, Thermal and Hydraulic Design. The DNB analysis for the methodology transition is based on this licensing basis whileincorporating a conservatively higher core power. The analysis addresses DNB performance and theeffects of fuel rod bow, bypass flow and lower plenum flow anomalies.

2.12.2.1 Design Basis and Methodology The T/H DNB analysis of the fuel at the WCGS is based on the RTDP (Reference

1) and the WRB-2DNB correlation (Reference
2) using the Westinghouse version of the VIPRE-01 subchannel analysiscode (Reference 3). The STDP is used when RTDP is not applicable.

For analyses that are outside of therange of applicability of the WRB-2 correlation, a W-3 alternative DNB correlation (ABB-NV or WLOP)is used (Reference 4). The RTDP methodology is applicable to accidents that initiate from normaloperating conditions whereas the STDP methodology is typically applied to events that are initiated fromshutdown conditions.

The WRB-2 correlation is used for analysis of fuel regions above the first mixingvane grid whereas the ABB-NV correlation is used for analysis of fuel regions below the first mixingvane grid. The WLOP correlation is used when the coolant conditions are outside the range ofapplicability of the WRB-2 and ABB-NV correlations.

Specific methodologies and correlations forspecific events are identified in other portions of Section 2.0 of this report. The analyses demonstrate thatthe 95/95 DNB design basis is met for the core in operation at the maximum analyzed core power inTable 2.12-1.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESUNGHOUSE NON-PROPRIETARY CLASS 32-3962.12.2.1.1 Subchannel Analysis CodeThe Westinghouse version of the VIPRE-01 code (VIPRE, Reference

3) is used for DNBR calculations.

The use of VIPRE for the methodology transition is in full compliance with the conditions specified in theUSNRC SER in WCAP-14565-P-A (Reference

3) for THINC and FACTRAN replacement.

See AppendixA of this WCAP for code applicability.

2.12.2.1.2 DNB Methodology With the RTDP methodology, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, computer codes, and DNB correlation predictions are statistically considered to obtain the overall DNB uncertainty factors.

Proprietary DNBR sensitivity

factors, which are used todevelop DNB uncertainty
factors, are calculated over ranges of conditions that bound the events for whichRTDP methodology is applied.

Based on the DNB uncertainty

factors, RTDP design limit DNBR valuesare determined such that there is a 95-percent probability with a 95-percent confidence level that DNBwill not occur on the most limiting fuel rod during normal operation, operational transients, or transient conditions arising from faults of moderate frequency.

The uncertainties included in the overall DNB uncertainty factor are:* nuclear enthalpy rise hot channel factor, (FNAH)0 enthalpy rise engineering hot channel factor, (FEAH)* uncertainties in the DNB correlations and the computer codes* vessel coolant flow* effective core flow fraction (1 -bypass flow fraction)

  • core thermal power* coolant temperature
  • system pressureInstrumentation uncertainties in core thermal power, RCS flow, pressure, and inlet temperature weretaken into account for the methodology transition.

Only the random portion of each plant operating parameter uncertainty is included in the statistical combination for RTDP. Any adverse instrumentation bias is treated either as a DNBR penalty or a direct analysis input.In addition to the above considerations for uncertainties, DNBR margin is retained by performing thesafety analyses to DNBR limits higher than the design limit DNBR values. Sufficient DNBR margin isconservatively maintained in the safety analysis DNBR limits to offset penalties for rod bow, lowerplenum flow anomalies, and plant instrumentation biases and to provide flexibility in design andoperation of the plant. Table 2.12-3 provides a summary of the DNBR margin and penalties applicable atnominal conditions.

The STDP is used for those analyses where RTDP is not applicable.

The DNBR limit for STDP is theappropriate DNB correlation limit increased by sufficient margin to offset the applicable DNBR penalties.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3972.12.2.1.3 DNB Correlations and LimitsThe WRB-2 DNB correlation is based entirely on rod bundle data and takes credit for the significant improvements in DNB performance due to the mixing vane grid effects.

USNRC acceptance of a95/95 WRB-2 correlation DNBR limit of 1.17 for 17x17 RFA-2 fuel is documented in References 2, 3,and 5.For the methodology transition, the WRB-2 RTDP design limit DNBR is 1.24 for the 17x 17 RFA-2 fuel atthe WCGS. The RTDP DNB analyses are performed to a higher DNBR limit referred to as the SALDNBR that includes additional margin. For cases in which WRB-2 is not applicable, the W-3 alternative correlations (ABB-NV or WLOP) are used as approved in Reference 4.For events in which STDP is used, the 95/95 correlation DNBR limits are 1. 17 for WRB-2, 1.13 forABB-NV, and 1.18 for WLOP.The reactor core is designed to meet the following limiting T/H criteria:

There is at least a 95-percent probability, at a 95-percent confidence level, that DNB will notoccur during any anticipated normal operating condition, operational transients, or any condition of moderate frequency.

Fuel melting will not occur during any anticipated normal operating condition, operational transients, or any conditions of moderate frequency.

Thermo-hydrodynamic instabilities will not occur during any anticipated normal operating condition, operational transients, or any conditions of moderate frequency.

DNBR is defined as the ratio of the heat flux causing DNB at a particular location in the core, aspredicted by a DNB correlation, to the actual heat flux at the same location.

Analytical assurance thatDNB will not occur is provided by showing the calculated DNBR to be higher than the 95/95 limit DNBRfor all conditions of normal operation, operational transients, and transient conditions of moderatefrequency.

The Design Limit DNBR is calculated by using the USNRC-approved RTDP methodology (Reference 1). Meeting the Design Limit assures compliance with the aforementioned DNB criteria.

A SAL DNBR, which is higher than the Design Limit DNBR, is conservatively used in safety analyses toprovide DNBR margin to offset the effect of rod bow, lower plenum flow anomalies, and plantinstrumentation biases and to provide flexibility in the design and operation of the plant.The RCS lower plenum anomaly is applicable to the WCGS. The probable cause of the flow anomaly isan unsteady, vortex flow disturbance in the reactor vessel lower plenum. The vortex restricts flow into thecore in the perturbed region and causes coolant temperature increases in the affected fuel assemblies.

Thehigher coolant temperatures then depress the local neutron fluxes due to reactivity feedback.

The flowdisturbance in the reactor vessel lower plenum also increases the overall hydraulic resistance of thereactor, and thus decreases the flow rate to all loops. The effect of the flow anomaly is a DNBR penalty,which is offset by available margin.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3982.12.3 Description of Analyses and Evaluations For the WCGS methodology transition, a DNB re-analysis was required to define new core limits, axialoffset limits, and Condition II and IV accident acceptability.

The core limits, axial offset limits, anddropped rod limit lines are generated based on the SAL DNBR and a design FN H limit of 1.65. This FNAHlimit incorporates all applicable uncertainties, including a measurement uncertainty (Reference 6), and isadjusted for the power level using the following equation:

FAN, = 1.65 * [1 + 0.3(1 -P)]where P is the fraction of full power.Various DNB analyses that were performed in support of the WCGS methodology transition are described below; all analyses were performed at a conservatively high core power. The descriptions belowsupplement the write-up already provided in previous sections.

2.12.3.1 Core Thermal LimitsThe core thermal limits are required for the generation of the OTAT and OPAT trip setpoints.

The corethermal limits define the loci of points of thermal power, primary system pressure, and coolant inlettemperature that satisfy the following criteria:

  • The minimum DNBR is not less than the SAL DNBR.* The hot channel exit quality is not greater than the upper limit of the quality range of the DNBcorrelation (adjusted for the analysis-specific quality uncertainty).
  • Vessel Th., < Tst to ensure that the difference between Thor and Tcold remains proportional to power.For the transition to Westinghouse methods and to support operation at a conservatively higher power,new core thermal limits were generated for the 17x 17 RFA-2 fuel at the WCGS. The DNB-limited portionof the core thermal limits was generated with the VIPRE code using the WRB-2 DNB correlation and theRTDP methodology.

2.12.3.2 Axial Offset LimitsThe axial offset limits are used to reduce the core DNB limit lines to account for the effect of adverseaxial power distributions that are more limiting for DNB than the axial power shape used to generate thecore thermal limits. For the transition to Westinghouse methods and to support operation at aconservatively higher power, new axial offset limits were generated for the 17x 17 RFA-2 fuel at theWCGS. The axial offset limits were generated with the VIPRE code using the RTDP methodology.

Forthe DNB analysis of axial power distributions that were limiting in the fuel region above the first mixingvane grid, the WRB-2 DNB correlation was used. For the DNB analysis of axial power distributions thatwere limiting in the fuel region below the first mixing vane grid, the ABB-NV DNB correlation was used.The axial offset limits were used to define the f(AI) reset function in the OTAT reactor trip function suchthat the DNB design criterion is met for accidents terminated by the OTAT reactor trip function.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-3992.12.3.3 Loss of FlowThe DNB analysis of the loss-of-flow accident was performed using RTDP for three different cases,including partial loss of flow, complete loss of flow, and UF. Each case was checked to ensure that thelimiting scenario was identified.

The effect of fuel temperatures is included in the analysis of this event.The complete loss of flow case results in the lowest minimum DNBR. The minimum DNBRs calculated for each of the three cases are greater than the SAL, thereby demonstrating compliance with the DNBdesign criterion for this event.DNB analysis was also performed to confirm that the DNB criterion was met for low flow conditions supporting the P-8 setpoint.

2.12.3.4 Locked RotorThe locked rotor accident is classified as a Condition IV event. To calculate the radiation release as aconsequence of the accident, calculations are performed using RTDP to quantify the inventory of rods thatwould experience DNB. Any rods in DNB are conservatively presumed to fail. For the WCGS, theanalysis indicates that there would be less than 1.0 percent rods in DNB due to the locked rotor accident.

The radiological consequences analysis conservatively assumes 5 percent of the fuel rods have failed andshows that the site dose limits are met.The Locked Rotor PCT analysis is performed using STDP and the VIPRE code. The acceptance criterion for this analysis is to demonstrate that the PCT is less than 2700'F. The PCT analysis for the WCGSsatisfied this acceptance criterion, thus, confirming that the fuel melt limit for ZIRLO (2700'F) highperformance fuel cladding materials is met.2.12.3.5 RCCA Drop/Misoperation This section supplements the methodology discussion in Section 2.5.3, "Control Rod Misoperation."

The USNRC-approved Westinghouse analysis methods in Reference 7 were used for analyzing the RCCAdrop event. The dropped rod limit lines were generated to define the loci of points that would result in theRTDP SAL DNBR for a wide range of core conditions (inlet temperature, power, and pressure).

Per themethodology described in Reference 8, these lines are used to verify that the DNB design basis is meteach cycle.The maximum allowable FNAH limit for RCCA misalignment was calculated using RTDP methodology.

This is the value of FNAH at normal operating conditions that results in a minimum DNBR equal to theRTDP SAL DNBR.The limits provided for the RCCA drop and RCCA misalignment events were used to confirm that theDNB design basis was met for the Wolf Creek methodology transition.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-4002.12.3.6 Steam Line Break AccidentThe event descriptions for the HZP and HFP events have been provided in Sections 2.2.5.1 and 2.2.5.2respectively.

SLB cases were analyzed at both HZP and HFP conditions.

For each of these cases, the appropriate methodology was applied.For the HFP cases, the RTDP methodology and the WRB-2 correlation are used. The DNB analysisshowed that the minimum DNBR values were above the SAL, thereby demonstrating that the DNBRdesign basis was met.For the HZP cases, STDP and the WLOP DNB correlation were applied.

The DNBR limit is thecorrelation limit increased by a small amount to account for any DNB penalties applicable at theseconditions.

The analysis showed that the minimum DNBR was greater than the DNBR limit, therebydemonstrating that the DNBR design basis was met.2.12,3.7 Feedwater Malfunction The HZP FWM event is analyzed using the same method that is used for the HZP SLB analysis.

For theWCGS methodology transition, Nuclear Design analyses indicated that the HZP FWM event wasbounded by the HZP SLB event. The DNBR design basis was met for the HZP SLB event, therebyindicating that the DNBR design basis was met for the HZP FWM event at the WCGS.2.12.3.8 Uncontrolled Rod Cluster Control Assembly Withdrawal from Subcritical Because the event is initiated from HZP conditions, the analysis for the uncontrolled RCCA withdrawal from subcritical accident is based on the STDP methodology.

Results and additional information arecontained in Section 2.5.1.This transient results in a power excursion and a bottom-skewed power shape due to the withdrawal of arod bank. A conservative accident-specific power shape was applied.

Two DNBR calculations arerequired for this accident.

The ABB-NV correlation is applied for fuel assembly spans below the firstmixing vane grid. The WRB-2 correlation is applied for spans above the first mixing vane grid. For theSTDP application, the DNBR limits applied are the correlation limits for ABB-NV and WRB-2, increased by any applicable DNBR penalties.

The results of the calculations showed that the calculated DNBRvalues remain above the respective DNBR limits, thereby demonstrating that the DNB design basis ismet.2.12.3.9 Rod Withdrawal at PowerA detailed DNB analysis of the rod withdrawal at power (RWAP) event was performed using the RTDPmethodology.

Statepoints for the limiting case (a low reactivity insertion rate case initiated from 10percent power, as described in Section 2.5.2, "Uncontrolled Rod Cluster Control Assembly BankWithdrawal at Power") were analyzed using the VIPRE code. The DNB design basis was met withmargin. Furthermore, in order to provide WCGS with adequate future flexibility in the design andWCAP- 17658-NP August 2013Licensing Report Revision 0

WESTfNGHOUSE NON-PROPRIETARY CLASS 32-401loperation of the plant relative to DNBR margin, the following credits were applied to the DNB analysisfor this particular event:* a reduced thimble bypass flow (since WCGS is currently operating with TPI), and* an increased MMF of 376,000 gpm, since WCNOC decided to raise the MMF value by5000 gpm.2.12.3.10 Bypass FlowTwo different bypass flow rates are used in the T/H design analysis.

The thermal design bypass flow is theconservatively high core bypass flow used in conjunction with the TDF in power capability analyses thatuse standard (non-statistical) methods.

The best estimate bypass flow is the core bypass flow that wouldbe expected using nominal values for dimensions and operating parameters that affect bypass flowwithout applying uncertainty factors.

The best estimate bypass flow is used in conjunction with the vesselMMF for power capability analyses using the RTDP design procedures.

As discussed inSection 2.12.2.1.2, for RTDP, the bypass flow uncertainty is included in the statistical combination for theRTDP design limit DNBR.2.12.3.11 Effects of Fuel Rod Bow on DNBRRod bow can occur between mid-grids, reducing the spacing between adjacent fuel rods and reducing themargin to DNB. Rod bow must be accounted for in the DNB safety analysis of Condition I andCondition II events. Westinghouse has conducted tests to determine the impact of rod bow on DNBperformance.

The testing and subsequent analyses are documented in References 9 and 10.The rod bow penalties are included in the DNBR margin summary shown in Table 2.12-3. In the spanscontaining IFM grids in the 17x17 RFA-2 fuel, no rod bow penalty is necessary due to the short spacingbetween grids. The maximum rod bow penalty accounted for in the safety analysis is a function ofassembly average bumup (References 9 and 10). Credit may also be taken for the effect of F AHburndown due to the decrease in fissionable isotopes and the buildup of fission products (Reference 11).2.12.3.12 License Renewal Impact Evaluation A review of T/H design for impact on plant license renewal evaluations was not necessary becausecontinued applicability of the safety analysis for the Westinghouse 17x 17 RFA-2 fuiel assemblies isre-evaluated during the RE process for each reload cycle. The reload design methodology includesevaluation of the reload core key safety parameters that comprise input to the safety evaluation for eachreload cycle.2.12.4 ResultsAnalyses described in the previous sections show that the DNB design basis is met for the WCGSmethodology transition.

The DNBR limits and margin summary are listed in Table 2.12-3. Cycle specificevaluation is to be performed in accordance with Reference 8.WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-4022.12.5 Conclusion WCNOC has reviewed the analyses related to the effects of the proposed methodology transition on theT/H design of the core and the RCS. WCNOC concludes that the analyses demonstrated that the design(1) has been accomplished using acceptable analytical

methods, (2) is equivalent to proven designs,(3) provides acceptable margins of safety from conditions that would lead to fuel damage during normalreactor operation and AOOs, and (4) is not susceptible to thenno-hydrodynamic instability.

WCNOCfurther concludes that the analyses have adequately accounted for the effects of the proposedmethodology transition on the hydraulic loads on the core and RCS components.

Based on this, WCNOCconcludes that the T/H design will continue to meet the requirements of GDCs 6 and 7 following implementation of the proposed methodology transition.

Therefore, WCNOC finds the proposedmethodology transition acceptable with respect to T/H design.2.12.6 References

1. WCAP- 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.2. WCAP-10444-P-A, "Reference Core Report -VANTAGE 5 Fuel Assembly,"

September 1985.3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.4. WCAP-14565-P-A Addendum 2-P-A, "Extended Application of ABB-NV Correlation andModified ABB-NV Correlation WLOP for PWR Low Pressure Applications,"

April 2008.5. LTR-NRC-02-55, "Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design,Revision 1," November 2002.6. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties,"

June 1988.7. WCAP-1 1394-P-A, "Methodology for the Analysis of the Dropped Rod Event," January 1990.8. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

July 1985.9. WCAP-8691-RI "Fuel Rod Bow Evaluation,"

July 1979.10. Letter from Rahe, E. P., Jr. (Westinghouse) to Miller, J. R. (USNRC),

"Partial Response toRequest Number 1 for Additional Information on WCAP-8691, Revision 1," NS-EPR-2515, October 9, 1981; and Letter from Rahe, E. P., Jr. (Westinghouse) to Miller, J. R. (USNRC),"Remaining Response to Request Number 1 for Additional Information on WCAP-8691, Revision 1," NS-EPR-2572, March 16, 1982.11. Letter from Berlinger, C. (USNRC) to Rahe, E. P., Jr. (Westinghouse),

"Request for Reduction inFuel Assembly Burnup Limit for Calculation of Maximum Rod Bow Penalty,"

NS-NRC-85-3901 NRC Response, June 18, 1986.WCAP-1 7658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-403Table 2.12-1 T/H Design Parameters Comparison Methods Transition T/H Design Parameters Current Design Value Analysis ValueReactor Core Heat Output, MWt 3565 3637Reactor Core Heat Output, 106 BTU/hr 12,164 12,410Heat Generated in Fuel, % 97.4 97.4Core Exit Pressure,

Nominal, psia 2270 2270Pressurizer Pressure.,
Nominal, psia 2250 2250Radial Power Distribution, FNAH I1 1.65[ 1+0.3(1-P)]

1.65[ 1+0.3(1 -P)]HFP Nominal Coolant Conditions (uncertainties and biases not included)

Vessel TDF Rate (including bypass)106 Ibm/hr 134.7 134.9gpm 361,200 361,200Core Flow Rate (excluding bypass)12)106 lbm/hr 123.4 123.6gpm 330,859 330,859Core Flow Area, ft2 51.08 51.08Core Inlet Mass Velocity, 106 Ibm/hr-ft 2 2.416 2.419Nominal Vessel/Core Inlet Temperature, IF 555.8 555.2Vessel Average Temperature, IF 588.4 588.4Core Average Temperature.,

F 593.2 593.4Vessel Outlet Temperature, IF 621.0 621.7Core Outlet Temperature, IF 626.2 627.0Average Temperature Rise in Vessel, IF 65.2 66.5Average Temperature Rise in Core, OF 70.4 71.8Heat TransferActive Heat Transfer Surface Area, ft2 59,742 59,742Average Heat Flux, BTU/hr-ft 2 198,315 202,326Average Linear Power, kW/ft 5.691 5.806Peak Linear Power for Normal Operation, 3" kW/ft 14.23 14.52WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-404Table 2.12-1 T/H Design Parameters Comparison (cont.)Methods Transition T/H Design Parameters Current Design Value Analysis ValuePeak Linear Power for Prevention of Centerline Melt, 22.4 22.4kW/ftPressure Drop Across Core, psi(4) 28.7 28.7Notes:Thermal Power1. P=Rated Thermal Power2. A design bypass flow of 8.4 percent was used.3. Based on maximum FQ of 2.5.4. The core pressure drop calculations are based on the same best estimate flows for 3565 MWt and 3637 MWt and full coresof 17x 17 RFA-2 fuel.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-405Table 2.12-2 Limiting Parameter Direction for DNBParameter Limiting Direction for DNBF NAH, nuclear enthalpy rise hot-channel factor maximumHeat generated in fuel (%) maximumReactor core heat output (MWt) maximumHeat flux (BTU/hr-ft

2) maximumVessel/core inlet temperature

('F) maximumCore pressure (psia) minimumPressurizer pressure (psia) minimumTDF for non-RTDP analyses (gpm) minimumMMF for RTDP analyses (gpm) minimumBypass flow maximumWCAP-l 7658-NP August 2013Licensing Report Revision 0WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 32-406Table 2.12-3 RTDP DNBR Margin SummaryCurrent Operation with 17x17 RFA-2 fuel(3565 MWt)t)DNB Correlation WRB-2DNBR Correlation Limit 1.17DNBR Design Limit(2) 1.24Total DNBR Penalties (due to rod bow, instrumentation biases, 13.6%lower plenum flow anomaly and RWAP)Total DNBR Margin(3)

> 13.6%Notes:1. DNBR analyses for the WCGS methodology transition were performed at a bounding power level of 3637 MWt.2. Design limit DNBR calculations are based on the measurement uncertainties and the sensitivity to changes in theparameters.

3. DNBR margin is the margin that exists between the SAL and the Design Limit DNBRs.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-1APPENDIX ASAFETY EVALUATION REPORT COMPLIANCE A.1 SAFETY EVALUATION REPORT COMPLIANCE INTRODUCTION This Appendix is a summary of USNRC-approved codes used in the LR. This appendix addresses compliance with the limitations, restrictions, and conditions specified in the approving safety evaluation of the applicable codes.Table A. 1-1 presents an overview of the SER by codes. For each SER, the applicable report subsections and Appendix A subsections are listed.Table A.1-1 Safety Evaluation Report Compliance SummaryLimitation, Licensing Topical Report Restriction, Report Appendix ANo. Subject (Reference)

Code(s) Condition Section Section1. Non-LOCA WCAP-7908-A FACTRAN Yes 2.5.1 A.2Thermal (Reference A. 1-1) 2.5.6Transients

2. Non-LOCA WCAP-14882-P-A RETRAN Yes 2.2.1 A.3Safety Analysis (Reference A. 1-2) 2.2.22.2.32.2.42.2.5.12.2.5.22.3.12.3.22.3.32.3.42.4.12.4.22.5.22.6.12.6.22.7.13. Non-LOCA WCAP-7907-P-A LOFTRAN Yes 2.5.2 A.4Safety Analysis (Reference A. 1-3) 2.5.32.84. Non-LOCA WCAP-1 1397-P-A RTDP Yes 2.12 A.6Thermal / Reference A. 1-5Hydraulics
5. Neutron Kinetics WCAP-7979-P-A TWINKLE None for 2.5.1 Not(Reference A. 1-4) Non-LOCA

2.5.6 Applicable

Transient AnalysisWCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-2Table A.1-1 Safety Evaluation Report Compliance Summary(cont.)Limitation, Licensing Topical Report Restriction, Report Appendix ANo. Subject (Reference)

Code(s) Condition Section Section6. Multi- WCAP-10965-P-A ANC None for 2.2.2 Notdimensional (Reference A. 1-6) Non-LOCA

2.2.4 Applicable

Neutronics Transient 2.2.5Analysis 2.5.37. Non-LOCA WCAP- 14565-P-A VIPRE Yes 2.2.2 A.5Thermal / (Reference A. 1-7) 2.2.4Hydraulics 2.2.52.4.12.4.22.5.12.5.22.5.32.128. Steam Generator WCAP- 10698-P-A RETRAN None for 2.7.2 NotTube Rupture (Reference A.1-8) Steam 2.7.3 Applicable Generator WCAP-14882-P-A Tube A.3(Reference A. 1-2) RuptureReferences A. 1-1 WCAP-7908-A, "FACTRAN

-A FORTRAN IV Code for Thermal Transients in a UO2Fuel Rod," H. G. Hargrove, December 1989.A. 1-2 WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

D. S. Huegel, et al., April 1999.A.1-3 WCAP-7907-P-A, "LOFTRAN Code Description,"

T. W. T. Burnett, et al., April 1984.A. 1-4 WCAP-7979-P-A, "TWINKLE

-A Multi-Dimensional Neutron Kinetics Computer Code,"D. H. Risher, Jr. and R. F. Barry, January 1975.A. 1-5 WCAP-I 1397-P-A, "Revised Thermal Design Procedure,"

A. J. Friedland and S. Ray, April 1989.A. 1-6 WCAP- 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," Y. S. Liu,et al., September 1986.A.1-7 WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

Y. X. Sung, et al., October 1999.A. 1-8 WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill,"

R. N. Lewis, et al., August 1987.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-3WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3A.2 FACTRAN FOR NON-LOCA THERMAL TRANSIENTS Table A.2-1 FACTRAN for Non-LOCA Thermal Transients Limitations, Restrictions, and Conditions

1. "The fuel volume-averaged temperature or surface temperature can be chosen at a desired value whichincludes conservatisms reviewed and approved by the USNRC."Justification The FACTRAN code was used in the analyses of the following two transients for the WCGS:Uncontrolled RCCA Withdrawal from a Subcritical Condition (USAR Section 15.4.1) and RCCA Ejection(USAR Section 15.4.8).

Initial fuel temperatures used as FACTRAN input in the RCCA Ejection analysiswere calculated using the USNRC-approved PAD 4.0 computer code as described in WCAP-15063-P-A Revision 1 (Reference A.2-1). As indicated in WCAP-15063-P-A Revision 1., the USNRC has approvedthe method of determining uncertainties for PAD 4.0 fuel temperatures.

2. "Table 2 presents the guidelines used to select initial temperatures."

Justification In summary, Table 2 of the SER specifies that the initial fuel temperatures assumed in the FACTRANanalyses of the following transients should be "High" and include uncertainties:

Loss of Flow, LockedRotor, and Rod Ejection.

As discussed above, fuel temperatures were used as input to the FACTRAN codein the RCCA Ejection analysis for the WCGS. The assumed fuel temperatures, which were calculated using the PAD 4.0 computer code (Reference A.2-1), include uncertainties and are conservatively high.FACTRAN was not used in the Loss of Flow and Locked Rotor analyses.

3. "The gap heat transfer coefficient may be held at the initial constant value or can be varied as afunction of time as specified in the input."Justification The gap heat transfer coefficients applied in the FACTRAN analyses are consistent with SER Table 2. Forthe RCCA Withdrawal from a Subcritical Condition transient, the gap heat transfer coefficient is kept at aconservative constant value throughout the transient; a high constant value is assumed to maximize thepeak heat flux (for DNB concerns) and a low constant value is assumed to maximize fuel temperatures.

For the RCCA Ejection transient, the initial gap heat transfer coefficient is based on the predicted initialfuel surface temperature, and is ramped rapidly to a very high value at the beginning of the transient tosimulate clad collapse onto the fuel pellet.4. "...the Bishop-Sandberg-Tong correlation is sufficiently conservative and can be used in theFA CTRAN code. It should be cautioned that since these correlations are applicable for local conditions only, it is necessary to use input to the FA CTRAN code which reflects the local conditions.

If the inputvalues reflecting average conditions are used, there must be sufficient conservatism in the input valuesto make the overall method conservative."

Justification Local conditions related to temperature, heat flux, peaking factors and channel information were input toFACTRAN for each of the two transients analyzed for the WCGS: RCCA Withdrawal from a Subcritical Condition (USAR Section 15.4.1) and RCCA Ejection (USAR Section 15.4.8).

Therefore, additional justification is not required.

WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-4Table A.2-1 FACTRAN for Non-LOCA Thermal Transients (cont.)Limitations, Restrictions, and Conditions

5. "The fuel rod is divided into a number of concentric rings. The maximum number of rings used torepresent the fuel is 10. Based on our audit calculations we require that the minimum of 6 should beused in the analyses."

Justification At least 6 concentric rings were assumed in FACTRAN for each of the two transients analyzed for theWCGS: RCCA Withdrawal from a Subcritical Condition (USAR Section 15.4.1) and RCCA Ejection(USAR Section 15.4.8).

Therefore, additional justification is not required.

6. "Although time-independent mechanical behavior (e.g., thermal expansion, elastic deformation) of thecladding are considered in FACTRAN, time-dependent mechanical behavior (e.g., plastic deformation) is not considered in the code....for those events in which the FACTRAN code is applied (see Table 1),significant time-dependent deformation of the cladding is not expected to occur due to the shortduration of these events or low cladding temperatures involved (where DNBR Limits apply), or the gapheat transfer coefficient is adjusted to a high value to simulate clad collapse onto the fuel pellet."Justification The two transients that were analyzed with FACTRAN for the WCGS (RCCA Withdrawal from aSubcritical Condition (USAR Section 15.4.1) and RCCA Ejection (USAR Section 15.4.8))

are included inthe list of transients provided in Table 1 of the SER; each of these transients is of short duration.

For theRCCA Withdrawal from a Subcritical Condition transient, relatively low cladding temperatures areinvolved, and the gap heat transfer coefficient is kept constant throughout the transient.

For the RCCAEjection transient, a high gap heat transfer coefficient is applied to simulate clad collapse onto the fuelpellet. The gap heat transfer coefficients applied in the FACTRAN analyses are consistent with SERTable 2.7. "The one group diffusion theory model in tire FACTRAN code slightly overestimates at beginning oflife (BOL) and underestimates at end of life (EOL) the magnitude offlux depression in the fuel whencompared to the LASER code predictions for the same fuel enrichment.

The LASER code uses transport theory. There is a difference of about 3 percent in the flux depression calculated using these two codes.When JT(centerline)

-T(Surface)l is on the order of 3000F, which can occur at the hot spot, thedifference between the two codes will give an error of 100lF. When the fuel surface temperature isfixved, this will result in a 100VF lower prediction of the centerline temperature in FA CTRAN. We haveindicated this apparent nonconservatism to Westinghouse.

In the letter NS-TMA-2026, datedJanuary 12, 1979, Westinghouse proposed to incorporate the LASER-calculated power distribution shapes in FACTRAN to eliminate this non-conservatism.

We find the use of the LASER-calculated power distribution in the FACTRAN code acceptable."

Justification The condition of concern (T(centerline)

-T(surface) on the order of 3,000°F) is expected for transients that reach, or come close to, the fuel melt temperature.

As this applies only to the RCCA ejection transient, the LASER-calculated power distributions were used in the FACTRAN analysis of the RCCA ejectiontransient for the WCGS.Reference A.2-1 WCAP- 15063-P-A (Proprietary) and WCAP- 15064-NP-A (Non-Proprietary),

Revision 1 (withErrata) "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0),"July 2000.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRJETARY CLASS 3A-5A.3 RETRAN FOR NON-LOCA SAFETY ANALYSISTable A.3-1 RETRAN for Non-LOCA Safety AnalysisLimitations, Restrictions, and Conditions I. "The transients and accidents that Westinghouse proposes to analyze with RETRAN are listed in thisSER (Table 1) and the NRC staff review of RE TRAN usage by Westinghouse was limited to this set. Useof the code for other analytical purposes will require additional justification."

Justification The transients listed in Table I of the SER are:* Feedwater system malfunctions

" Excessive increase in steam flow" Inadvertent opening of a steam generator relief or safety valve* Steam line break* Loss of external load/turbine trip* Loss of offsite power* Loss of normal feedwater flow* Feedwater line rupture* Loss of forced reactor coolant flow* Locked reactor coolant pump rotor/sheared shaft* Control rod cluster withdrawal at power* Dropped control rod cluster/dropped control bank* Inadvertent increase in coolant inventory

15. 1.1 and 15.1.2),* Excessive increase in secondary steam flow (USAR Section 15.1.3),* Inadvertent opening of a steam generator atmospheric relief or safety valve(USAR Section 15.1.4),* Steam system piping failure (steam line break) (USAR Section 15.1.5),* Loss of external electrical load/turbine trip (USAR Sections 15.2.2, 15.2.3, 15.2.4, and 15.2.5),* Loss of non-emergency alternating current (AC) power to the station auxiliaries (loss of offsitepower) (USAR Section 15.2.6)," Loss of normal feedwater flow (USAR Section 15.2.7)," Feedwater system pipe break (feedwater line rupture)

(USAR Section 15.2.8)," Loss of forced reactor coolant flow (USAR Sections 15.3.1 and 15.3.2),* Locked reactor coolant pump rotor/shaft break (USAR Sections 15.3.3 and 15.3.4),* Uncontrolled RCCA bank withdrawal at power (USAR Section 15.4.2),* Inadvertent operation of the ECCS (increase in coolant inventory)

(USAR Section 15.5.1),* CVCS malfunction that increases reactor coolant inventory (USAR Section 15.5.2),* Inadvertent opening of a pressurizer safety or relief valve (USAR Section 15.6. 1),* Steam generator tube rupture (USAR Section 15.6.3).As each transient analyzed for the WCGS using RETRAN matches one of the transients listed in Table 1of the SER, additional justification is not required.

WCAP- 17658-NP August 2013Licensing Report Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-6Table A.3-1 RETRAN for Non-LOCA Safety Analysis(cont.)Limitations, Restrictions, and Conditions

2. "WCAP-14882 describes modeling of Westinghouse designed 4-, 3, and 2-loop plants of the type thatare currently operating.

Use of the code to analyze other designs, including the Westinghouse AP600,will require additional justification."

Justification The WCGS consists of one 4-loop Westinghouse-designed unit that was "currently operating" at the timethe SER was written (February 11, 1999). Therefore, additional justification is not required.

3. "Conservative safety analyses using RETRAN are dependent on the selection of conservative input.Acceptable methodology for developing plant-specific input is discussed in WCAP-14882 and inReference 14 [WCAP-92 72-P-Al.

Licensing applications using RETRAN should include the source ofand justification for the input data used in the analysis."

Justification The input data used in the RETRAN analyses performed by Westinghouse came from both WCNOC andWestinghouse sources.

Assurance that the RETRAN input data is conservative for the WCGS is providedvia Westinghouse's use of transient-specific analysis guidance documents.

Each analysis guidancedocument provides a description of the subject transient, a discussion of the plant protection systems thatare expected to function, a list of the applicable event acceptance

criteria, a list of the analysis inputassumptions, e.g., directions of conservatism for initial condition values, a detailed description of thetransient model development method, and a discussion of the expected transient analysis results.

Based onthe analysis guidance documents, conservative plant-specific input values were requested and collected from the responsible WCNOC and Westinghouse sources.

Consistent with the Westinghouse ReloadEvaluation Methodology described in WCAP-9272 (Reference A.3-1), the safety analysis input valuesused in the WCGS analyses were selected to conservatively bound the values expected in subsequent operating cycles.Reference A.3-1 WCAP-9272-P-A (Proprietary) and WCAP-9273-NP-A (Non-Proprietary),

"Westinghouse Reload Safety Evaluation Methodology,"

July 1985.WCAP-17658-NP Licensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-7A.4 LOFTRAN FOR NON-LOCA SAFETY ANALYSISTable A.4-1 LOFTRAN for Non-LOCA Safety AnalysisLimitations, Restrictions, and Conditions

1. "LOFTRAN is used to simulate plant response to many of the postulated events reported in Chapter 15of PSARs and FSARs, to simulate anticipated transients without scram, for equipment sizing studies,and to define mass/energy releases for containment pressure analysis.

The Chapter 15 events analyzedwith LOFTRAN are:* Feedwater System Malfunction

  • Excessive Increase in Steam Flow" Inadvertent Opening of a Steam Generator Relief or Safety Valve* Steamline Break* Loss of External Load* Loss of Offsite Power* Loss of Normal Feedwater
  • Feedwater Line Rupture* Loss of Forced Reactor Coolant Flow* Locked Pump Rotor* Rod Withdrawal at Power* Rod Drop* Startup of an Inactive Pump* Inadvertent ECCS Actuation
  • Inadvertent Opening of a Pressurizer Relief or Safety ValveThis review is limited to the use of LOFTRANfor the licensee safety analyses of the Chapter 15 eventslisted above, and for a steam generator tube rupture..."

Justification For the WCGS, the LOFTRAN code was used in the analysis of the uncontrolled RCCA bank withdrawal at power transient (USAR Section 15.4.2),

in the analysis of the dropped rod transient (USAR Section 15.4.3),

and in the analysis of the anticipated transients without scram(USAR Section 15.8). As each of these transients match one of the transients listed in the SER, additional justification is not required.

WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-8WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-8A.5 VIPRE FOR NON-LOCA THERMAL/HYDRAULICS Table A.5-1 VIPRIE for Non-LOCA Thermal/Hydraulics Limitations, Restrictions, and Conditions 1."Selection of the appropriate CHF correlation, DNBR limit, engineered hot channel factors forenthalpy rise and other fuel-dependent parameters for a specific plant application should be justified with each submittal."

Justification The WRB-2 correlation with a 95/95 correlation limit of 1.17 was used in the DNB analyses for the WCGS17x 17 RFA-2 fuel. The use of the WRB-2 DNB correlation was approved in WCAP- 10444-P-A (Reference A.5-2). Applicability of the WRB-2 to 17x17 RFA-2 fuel was established through the FuelCriterion Evaluation Process (FCEP) in LTR-NRC-02-55 (Reference A.5-3). For conditions where WRB-2is not applicable, analyses were performed using approved secondary CHF correlations (such as ABB-NVand WLOP) in compliance with the SER conditions licensed for use in the VIPRE code.(WCAP-14565-P-A and its Addendum 2-P-A, Reference A.5-4).The use of the plant specific hot channel factors and other fuel dependent parameters in the DNB analysisfor the WCGS 17x 17 RFA-2 fuel were justified using the same methodologies as for previously approvedsafety evaluations of other Westinghouse four-loop plants using the same fuel design.2. "Reactor core boundary conditions determined using other computer codes are generally input intoVIPRE for reactor transient analyses.

These inputs include core inlet coolant flow and enthalpy, coreaverage power, power shape and nuclear peaking factors.

These inputs should be justified asconservative for each use of VIPRE."Justification The core boundary conditions for the VIPRE calculations for the 17x 17 RFA-2 fuel are all generated fromUSNRC-approved codes and analysis methodologies.

Conservative reactor core boundary conditions werejustified for use as input to VIPRE. Continued applicability of the input assumptions is verified on acycle-by-cycle basis using the Westinghouse reload methodology described in WCAP-9272-P-A (Reference A.5-1).3. "The NRC Staffs generic SER for VIPRE set requirements for use of new CHF correlations withVIPRE. Westinghouse has met these requirements for using WRB-1, WRB-2 and WRB-2Mcorrelations.

The DNBR limit for WRB-1 and WRB-2 is 1.17. The WRB-2M correlation has a DNBRlimit of 1.14. Use of other CHF correlations rot currently included in VIPRE will require additional justification."

Justification As discussed in response to Condition 1, the WRB-2 correlation with a limit of 1.17 was used as theprimary correlation in the DNB analyses of 17x 17 RFA-2 fuel for WCGS. For conditions where theWRB-2 is not applicable, analyses were performed using approved secondary CHF correlations licensedfor the VIPRE code in Reference A.5-4.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-9WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-9Table A.5-1 VIPRE for Non-LOCA Thermal/Hydraulics (cont.)Limitations, Restrictions, and Conditions

4. "Westinghouse proposes to use the VIPRE code to evaluate fuel performance folow#ving postulated design-basis accidents, including beyond-CHF heat transfer conditions.

These evaluations arenecessary to evaluate tihe extent of core damage and to ensure that the core maintains a coolablegeometry in the evaluation of certain accident scenarios.

The NRC Staffs generic review of VIPRE didnot extend to post CHF calculations.

VIPRE does not model the time-dependent physical changes thatmay occur within the fuel rods at elevated temperatures.

Westinghouse proposes to use conservative input in order to account for these effects.

The NRC Staff requires that appropriate justification besubmitted with each usage of VIPRE in the post-CHF region to ensure that conservative results areobtained."

Justification For application to Wolf Creek safety analysis, the use of VIPRE in the post-critical heat flux region islimited to the PCT calculation for the locked rotor transient.

The calculation demonstrated that the PCT inthe reactor core is well below the allowable limit to prevent clad embrittlement.

VIPRE modeling of thefuel rod is consistent with the model described in WCAP-14565-P-A (Reference A.5-4) and included thefollowing conservative assumptions:

  • Film boiling was calculated using the Bishop-Sandberg-Tong correlation,
  • The Baker-Just correlation accounted for heat generation in fuel cladding due to zirconium-water reaction.

Conservative results were further ensured with the following input:* Fuel rod input based on the maximum fuel temperature at the given power,* The hot spot power factor was equal to or greater than the design linear heat rate,* Uncertainties were applied to the initial operating conditions in the limiting direction.

References A.5-1 WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

S. L. Davidson (Ed.),July 1985.A.5-2 WCAP-10444-P-A, "Reference Core Report -VANTAGE 5 Fuel Assembly,"

S. L. Davidson(Editor),

September 1985.A.5-3 LTR-NRC-02-55, "Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design,Revision 1," November 2002.A.5-4 WCAP- 14565-P-A Addendum 2-P-A, "Extended Application of ABB-NV Correlation andModified ABB-NV Correlation WLOP for PWR Low Pressure Applications,"

A. Leidich,et. al., April 2008.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-10A.6 REVISED THERMAL DESIGN PROCEDURE FOR NON-LOCA THERMALHYDRAULICS Table A.6-1 Revised Thermal Design Procedure for Non-LOCA Thermal Hydraulics Limitations, Restrictions, and Conditions

1. "Sensitivity factors for a particular plant and their ranges of applicability should be included in theSafety Analysis Report or reload submittal.

Justification Sensitivity factors were evaluated using the WRB-2 and ABB-NV correlations and the VIPRE code forparameter values applicable to the 17x 17 RFA-2 fuel at conditions corresponding to a conservatively higher nominal core power of 3637 MWt. These sensitivity factors were used to determine the RTDPdesign limit DNBR values which are to be included in the WCGS USAR.2. "Anj' changes in DNB correlation, THINC-IV correlations, orparameter values listed in Table 3-1 ofWCAP-11397 outside of previously demonstrated acceptable ranges require re-evaluation of thesensitivity factors and of the use of Equation (2-3) of the topical report."Justification Because the VIPRE code was used to replace the THINC-IV code, sensitivity factors were evaluated forusing the VIPRE code. VIPRE has been demonstrated to be equivalent to the THINC-IV code inWCAP-14565-P-A (Reference A.6-1). See the response to condition 3 for a discussion of the use ofEquation (2-3) of the topical report. Evaluations using both WRB-2 and ABB-NV correlations were donein compliance with the methodology described in WCAP- 11397-P-A (Reference A.6-2).3. "If the sensitivity factors are changed as a result of correlation changes or changes in the application oruse of the THINC code, then the use of an uncertainty allowance for application of Equation (2-3) mustbe re-evaluated and the linearity assumption made to obtain Equation (2-17) of the topical report mustbe validated.

Justification Equation (2-3) of WCAP-I 1397-P-A (Reference A.6-2) and the linearity approximation made to obtainEquation (2-17) were confirmed to be valid for the WCGS using the combination of the VIPRE code andthe WRB-2 and ABB-NV correlations, at conditions corresponding to a conservatively higher nominalcore power of 3637 MWt.4. "Variances and distributions for input parameters must be justified on a plant-by-plant basis untilgeneric approval is obtained."

Justification The plant specific variances and distributions were justified for use at conditions corresponding to aconservatively higher nominal core power of 3637 MWt and are presented in Section 2.12 of the LR.5. "Nominal initial condition assumptions apply only to DNBR analyses using RTDP. Other analyses, such as overpressure calculations, require the appropriate conservative initial condition assumptions."

Justification Nominal conditions were only applied to the DNBR analyses which used RTDP.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3A-11Table A.6-1 Revised Thermal Design Procedure for Non-LOCA Thermal Hydraulics (cont.)Limitations, Restrictions, and Conditions

6. "Nominal conditions chosen for use in analyses should bound all permitted methods of plant operation.

Justification Bounding nominal conditions corresponding to a conservatively higher nominal core power of 3637 MWtwere used in the DNBR analyses using RTDP, consistent with the current methods of plant operation atWCGS.7. "The code uncertainties specified in Table 3-1 (of WCAP-1139 7-P) (+/- 4 percent for THINC-IV and+/- 1 percent for transients) must be included in the DNBR analyses using R TDP."Justification The code uncertainties specified in Table 3-1 of WCAP-1 1397-P-A (Reference A.6-2) remainedunchanged and were included in the DNBR analyses using RTDP. The THFNC-IV uncertainty was appliedto VIPRE, based on the equivalence of the VIPRE model approved in WCAP-14565-P-A to THINC-IV.

References A.6-1 WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water ReactorNon-LOCA Thermal-Hydraulic Safety Analysis,"

Y. X. Sung, et al., October 1999.A.6-2 WCAP-1 1397-P-A, "Revised Thermal Design Procedure,"

Friedland, A. J. and Ray, S.,April 1999.WCAP- 17658-NPLicensing ReportAugust 2013Revision 0

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