ML22285A125

Attachment 8 to Enclosure 1: ANP-3969NP, Callaway Non-LOCA Summary Report, Revision 2, Dated October 2022

ML22285A125
Person / Time
Site:	Callaway
Issue date:	10/12/2022
From:	Ameren Missouri, Framatome, Union Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML22285A115	List: ML22285A116 ML22285A117 ML22285A118 ML22285A119 ML22285A120 ML22285A122 ML22285A123 ML22285A124 ML22285A125 ML22285A130 ML22285A131 ULNRC-06768, Attachment 4 to Enclosure 1: Framatome Affidavits for Withholding from Public Disclosure
References
ULNRC-06768 ANP-3969NP, Rev. 2
Download: ML22285A125 (190)

Text

Attachment 8 to to ULNRC-06768 Page 1 of 190 ATTACHMENT 8 NON-PROPRIETARY VERSION OF NON-LOCA

SUMMARY

REPORT The following pages provide the non-proprietary version of the technical summary reports provided by Framatome supporting this license amendment request.

ANP-3969NP, "Callaway Non-LOCA Summary Report,"

Revision 2, dated October 2022

[NON-PROPRIETARY REPORT]

189 pages follow this cover sheet

ANP-3969NP Callaway Non-LOCA Summary Revision 2 Report October 2022 (c) 2022 Framatome Inc.

ANP-3969NP Revision 2 Copyright © 2022 Framatome Inc.

All Rights Reserved FRAMATOME TRADEMARKS ARTEMIS, COBRA-FLX, COPERNIC, GAIA, GRIP, HMP, M5, ORFEO, Q12, and S-RELAP5 are trademarks or registered trademarks of Framatome or its affiliates in the US or other countries.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page i Nature of Changes Revision Section(s) or Item No Page(s) Description and Justification 1 1 Section 3.9.2 Added a methodology change for the COBRA-FLX topical report (Reference 12) 2 1 Section 5.5.3 Updated several MDNBR and Peak LHR values in Table 5-8 3 2 All Updated Proprietary notations throughout document.

4 2 Section 5.29 Added missing commas in first sentence of first paragraph.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page ii Contents Page

1.0 INTRODUCTION

.................................................................................................. 1 2.0

SUMMARY

OF RESULTS .................................................................................... 1 3.0 CHAPTER 15 GENERAL DESCRIPTION ............................................................ 1 3.1 Initial Conditions ........................................................................................ 1 3.2 Component Setpoints and Capacities ........................................................ 3 3.3 Plant Operational Modes ........................................................................... 3 3.4 RPS and ESFAS Functions ....................................................................... 3 3.5 Fuel Mechanical Design............................................................................. 4 3.6 Peaking Factors ......................................................................................... 4 3.7 Reactivity Coefficients................................................................................ 5 3.8 RCCA Insertion Characteristics ................................................................. 5 3.9 Analysis Methodologies ............................................................................. 5 3.9.1 Methodology Description ................................................................. 5 3.9.2 Methodology Changes .................................................................. 11 3.10 Computer Codes ...................................................................................... 13 3.11 Event Classification ................................................................................. 16 4.0 CHAPTER 15 DISPOSITION OF EVENTS ........................................................ 28 5.0 CHAPTER 15 EVENT ANALYSES..................................................................... 32 5.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (FSAR SP 15.1.1) ............................................. 32 5.1.1 Event Description .......................................................................... 32 5.1.2 Method of Analysis ........................................................................ 33 5.1.3 Results .......................................................................................... 35 5.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow (FSAR SP 15.1.2) .......................................................... 41 5.2.1 Event Description .......................................................................... 41 5.2.2 Method of Analysis ........................................................................ 42 5.2.3 Results .......................................................................................... 45 5.3 Excessive Increase in Secondary Steam Flow (FSAR SP 15.1.3)...................................................................................................... 59 5.3.1 Event Description .......................................................................... 59 5.3.2 Method of Analysis ........................................................................ 60 5.3.3 Results .......................................................................................... 62

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page iii 5.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve (FSAR SP 15.1.4) .......................................................................... 68 5.5 Steam System Piping Failure (FSAR SP 15.1.5) ..................................... 68 5.5.1 Event Description .......................................................................... 68 5.5.2 Method of Analysis ........................................................................ 70 5.5.3 Results .......................................................................................... 73 5.6 Steam Line Break with Coincidental RCCA Withdrawal at Power (FSAR SP 15.1.5.5) ...................................................................... 89 5.7 Steam System Piping Failure at Full Power (FSAR SP 15.1.5.6)................................................................................................... 89 5.7.1 Event Description .......................................................................... 89 5.7.2 Method of Analysis ........................................................................ 90 5.7.3 Results .......................................................................................... 92 5.8 Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow (FSAR SP 15.2.1) .......................... 103 5.9 Loss of External Electrical Load (FSAR SP 15.2.2) ............................... 103 5.10 Turbine Trip (FSAR SP 15.2.3) .............................................................. 103 5.11 Inadvertent Closure of Main Steam Isolation Valves (FSAR SP 15.2.4) .............................................................................................. 104 5.12 Loss of Condenser Vacuum and Other Events Resulting in Turbine Trip (FSAR SP 15.2.5) .............................................................. 105 5.13 Loss of Nonemergency AC Power to the Plant Auxiliaries (FSAR SP 15.2.6) .................................................................................. 105 5.14 Loss of Normal Feedwater Flow (FSAR SP 15.2.7) ............................... 105 5.15 Feedwater System Pipe Break (FSAR SP 15.2.8) ................................. 106 5.16 Partial Loss of Forced Reactor Coolant Flow (FSAR SP 15.3.1).................................................................................................... 106 5.17 Complete Loss of Forced Reactor Coolant Flow (FSAR SP 15.3.2).................................................................................................... 107 5.17.1 Event Description ........................................................................ 107 5.17.2 Method of Analysis ...................................................................... 108 5.17.3 Results ........................................................................................ 110 5.18 Reactor Coolant Pump Shaft Seizure (Locked Rotor) (FSAR SP 15.3.3) .............................................................................................. 116 5.18.1 Event Description ........................................................................ 116 5.18.2 Method of Analysis ...................................................................... 116 5.18.3 Results ........................................................................................ 119

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page iv 5.19 Reactor Coolant Pump Shaft Break (FSAR SP 15.3.4) ......................... 124 5.20 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition (FSAR SP 15.4.1)................................... 124 5.20.1 Event Description ........................................................................ 124 5.20.2 Method of Analysis ...................................................................... 125 5.20.3 Results ........................................................................................ 128 5.21 Uncontrolled RCCA Bank Withdrawal at Power (FSAR SP 15.4.2).................................................................................................... 133 5.21.1 Event Description ........................................................................ 133 5.21.2 Method of Analysis ...................................................................... 134 5.21.3 Results ........................................................................................ 135 5.22 RCCA Misoperation (System Malfunction or Operator Error)

(FSAR SP 15.4.3) .................................................................................. 150 5.22.1 Event Description ........................................................................ 150 5.22.2 Method of Analysis ...................................................................... 151 5.22.3 Results ........................................................................................ 153 5.23 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (FSAR SP 15.4.4)............................................................. 159 5.24 A Malfunction of Failure of the Flow Controller in a Boiling Water Reactor Loop that Results in an Increased Reactor Coolant Flow Rate (FSAR SP 15.4.5) .................................................... 159 5.25 Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant (FSAR SP 15.4.6) ....................................................... 159 5.25.1 Event Description ........................................................................ 159 5.25.2 Analysis Method .......................................................................... 160 5.25.3 Results ........................................................................................ 161 5.26 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (FSAR SP 15.4.7) ..................................................... 161 5.27 Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSAR SP 15.4.8).................................................................. 162 5.28 Inadvertent Operation of the Emergency Core Cooling System During Power Operation (FSAR SP 15.5.1) .............................. 162 5.29 Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory (FSAR SP 15.5.2) ....................... 163 5.30 A Number of BWR Transients (FSAR SP 15.5.3) .................................. 163 5.31 Inadvertent Opening of a Pressurizer Safety or Relief Valve (FSAR SP 15.6.1) .................................................................................. 163

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page v 5.31.1 Event Description ........................................................................ 163 5.31.2 Method of Analysis ...................................................................... 164 5.31.3 Results ........................................................................................ 165 5.32 Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundary that Penetrate Containment (FSAR SP 15.6.2) .................................................................................. 171 5.33 Steam Generator Tube Failure (FSAR SP 15.6.3)................................. 171 5.34 Spectrum of Boiling Water Reactor Steam System Piping Failures Outside of Containment (FSAR SP 15.6.4) .............................. 171 5.35 Loss-of-Coolant Accidents Resulting from a Spectrum of Postulation Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR SP 15.6.5)................................................... 172 5.36 A Number of BWR Transients (FSAR SP 15.6.6) .................................. 172 5.37 Radioactive Release from a Subsystem or Component (FSAR SP 15.7) ..................................................................................... 172 5.38 Anticipated Transient Without Scram (FSAR SP 15.8) .......................... 172

6.0 REFERENCES

................................................................................................. 173

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page vi List of Tables Page Table 3-1 Summary of Initial Conditions and Computer Codes Used......................... 17 Table 3-2 Key Component Setpoints and Capacities ................................................. 21 Table 3-3 Plant Operational Modes ............................................................................ 22 Table 3-4 RPS Trip Setpoints and Response Times .................................................. 23 Table 3-5 Credited RPS Trip Functions ...................................................................... 24 Table 3-6 ESFAS Setpoints and Response Times ..................................................... 25 Table 3-7 Fuel Design Parameters ............................................................................. 25 Table 3-8 Core Power Distribution Parameters .......................................................... 26 Table 3-9 Reactivity Parameters ................................................................................ 26 Table 3-10 ORFEO-GAIA and ORFEO-NMGRID Design Limits .................................. 26 Table 3-11 Event Classification and Acceptance Criteria ............................................. 27 Table 4-1 Event Disposition Summary of Results....................................................... 29 Table 5-1 Decrease in Feedwater Temperature: Sequence of Events ....................... 36 Table 5-2 Decrease in Feedwater Temperature: Results ........................................... 36 Table 5-3 Increase in Feedwater Flow: Sequence of Events...................................... 46 Table 5-4 Increase in Feedwater Flow: Results.......................................................... 47 Table 5-5 Increase in Steam Flow: Sequence of Events ............................................ 63 Table 5-6 Increase in Steam Flow: Results ................................................................ 63 Table 5-7 Steam System Piping Failure: Sequence of Events ................................... 75 Table 5-8 Steam System Piping Failure: Results ....................................................... 76 Table 5-9 Steam System Piping Failure at Full Power: Sequence of Events............. 93 Table 5-10 Steam System Piping Failure at Full Power: Results.................................. 93 Table 5-11 Complete Loss of Forced Reactor Coolant Flow: Sequence of Events .................................................................................................... 111 Table 5-12 Complete Loss of Forced Reactor Coolant Flow: Results ........................ 111 Table 5-13 Reactor Coolant Pump Shaft Seizure: Sequence of Events ..................... 120 Table 5-14 Reactor Coolant Pump Shaft Seizure: Results ......................................... 120 Table 5-15 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Sequence of Events ..................................... 129

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page vii Table 5-16 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Results .......................................................... 129 Table 5-17 Uncontrolled RCCA Bank Withdrawal at Power: Sequence of Events .... 137 Table 5-18 Uncontrolled RCCA Bank Withdrawal at Power: Results ......................... 138 Table 5-19 RCCA Drop: Sequence of Events............................................................. 155 Table 5-20 RCCA Drop: Results................................................................................. 155 Table 5-21 Single RCCA Withdrawal: Results ............................................................ 155 Table 5-22 Inadvertent Boron Dilution: Results .......................................................... 161 Table 5-23 Inadvertent Opening of a Pressurizer Safety Valve: Sequence of Events .................................................................................................... 167 Table 5-24 Inadvertent Opening of a Pressurizer Safety Valve: Results .................... 167

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page viii List of Figures Page Figure 5-1 Decrease in Feedwater Temperature: Reactor Power ............................... 37 Figure 5-2 Decrease in Feedwater Temperature: Core Average Heat Flux ................. 38 Figure 5-3 Decrease in Feedwater Temperature: Pressurizer Pressure ...................... 39 Figure 5-4 Decrease in Feedwater Temperature: Reactor Coolant System Temperatures .......................................................................................... 40 Figure 5-5 Increase in Feedwater Flow (HFP): Core Power ........................................ 48 Figure 5-6 Increase in Feedwater Flow (HFP): Core Average Heat Flux ..................... 49 Figure 5-7 Increase in Feedwater Flow (HFP): Pressurizer Pressure.......................... 50 Figure 5-8 Increase in Feedwater Flow (HFP): Reactor Coolant Loop Hot Leg to Cold Leg Temperature Difference............................................................ 51 Figure 5-9 Increase in Feedwater Flow (HFP): Vessel Average Temperature............. 52 Figure 5-10 Increase in Feedwater Flow (HZP, Multi-Loop): Core Power ................... 53 Figure 5-11 Increase in Feedwater Flow (HZP, Multi-Loop): Core Average Heat Flux .......................................................................................................... 54 Figure 5-12 Increase in Feedwater Flow (HZP, Multi-Loop): Core Reactivity ............. 55 Figure 5-13 Increase in Feedwater Flow (HZP, Multi-Loop): Vessel Average Temperature ............................................................................................ 56 Figure 5-14 Increase in Feedwater Flow (HZP, Multi-Loop): Pressurizer Pressure .................................................................................................. 57 Figure 5-15 Increase in Feedwater Flow (HZP, Multi-Loop): Pressurizer Liquid Volume..................................................................................................... 58 Figure 5-16 Increase in Steam Flow: Core Power ...................................................... 64 Figure 5-17 Increase in Steam Flow: Pressurizer Pressure ........................................ 65 Figure 5-18 Increase in Steam Flow: Pressurizer Liquid Level ................................... 66 Figure 5-19 Increase in Steam Flow: Reactor Coolant System Temperatures ........... 67 Figure 5-20 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Power ...................................................... 77 Figure 5-21 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Average Heat Flux .................................. 78 Figure 5-22 Steam System Piping Failure (Mode 3, Mode 3, Maximum Safety Injection, Offsite Power Available): Pressurizer Pressure ........................ 79

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page ix Figure 5-23 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Pressurizer Liquid Volume ............................... 80 Figure 5-24 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Cold Leg Temperatures ................................... 81 Figure 5-25 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Loop Average Temperatures ........................... 82 Figure 5-26 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Reactivity ................................................ 83 Figure 5-27 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Boron Concentration ............................... 84 Figure 5-28 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Main and Auxiliary Feedwater Flow Rates ....... 85 Figure 5-29 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Steam Flow Rates ........................................... 86 Figure 5-30 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Steam Generator Pressures ............................ 87 Figure 5-31 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Flow Rate ................................................ 88 Figure 5-32 Steam System Piping Failure (HFP, 0.55 ft2 Break): Core Power and Core Average Heat Flux ................................................................... 94 Figure 5-33 Steam System Piping Failure (HFP, 0.55 ft2 Break): Pressurizer Pressure .................................................................................................. 95 Figure 5-34 Steam System Piping Failure (HFP, 0.55 ft2 Break): Pressurizer Liquid Volume .......................................................................................... 96 Figure 5-35 Steam System Piping Failure (HFP, 0.55 ft2 Break): Core Inlet Temperatures .......................................................................................... 97 Figure 5-36 Steam System Piping Failure (HFP, 0.55 ft2 Break): Reactor Coolant System Average Temperatures .................................................. 98 Figure 5-37 Steam System Piping Failure (HFP, 0.55 ft2 Break): Main Feedwater Flow Rates ............................................................................. 99 Figure 5-38 Steam System Piping Failure (HFP, 0.55 ft2 Break): Steam Generator Flow Rates ............................................................................ 100 Figure 5-39 Steam System Piping Failure (HFP, 0.55 ft2 Break): Steam Generator Pressures.............................................................................. 101 Figure 5-40 Steam System Piping Failure (HFP, 0.55 ft2 Break): Break Flow Rates ..................................................................................................... 102

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page x Figure 5-41 Complete Loss of Forced Reactor Coolant Flow: Reactor Coolant System Flow Rate .................................................................................. 112 Figure 5-42 Complete Loss of Forced Reactor Coolant Flow: Core Power............... 113 Figure 5-43 Complete Loss of Forced Reactor Coolant Flow: Pressurizer Pressure ................................................................................................ 114 Figure 5-44 Complete Loss of Forced Reactor Coolant Flow: Core Average Heat Flux ............................................................................................... 115 Figure 5-45 Reactor Coolant Pump Shaft Seizure: Reactor Coolant System Flow Rate............................................................................................... 121 Figure 5-46 Reactor Coolant Pump Shaft Seizure: Pressurizer Pressure ................. 122 Figure 5-47 Reactor Coolant Pump Shaft Seizure: Core Power and Core Average Heat Flux ................................................................................. 123 Figure 5-48 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Core Power ................................................... 130 Figure 5-49 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Core Average Heat Flux ................................ 131 Figure 5-50 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Fuel Centerline Temperature ........................ 132 Figure 5-51 Uncontrolled RCCA Bank Withdrawal (HFP, BOC, 1.9 pcm/sec):

Core Power and Core Average Heat Flux.............................................. 139 Figure 5-52 Uncontrolled RCCA Bank Withdrawal (HFP, BOC, 1.9 pcm/sec):

Pressurizer Pressure ............................................................................. 140 Figure 5-53 Uncontrolled RCCA Bank Withdrawal (HFP, BOC, 1.9 pcm/sec):

Pressurizer Liquid Level......................................................................... 141 Figure 5-54 Uncontrolled RCCA Bank Withdrawal (HFP, BOC, 1.9 pcm/sec):

Reactor Coolant System Temperatures ................................................. 142 Figure 5-55 Uncontrolled RCCA Bank Withdrawal (HFP, EOC, 23.7 pcm/sec):

Core Power and Core Average Heat Flux.............................................. 143 Figure 5-56 Uncontrolled RCCA Bank Withdrawal (HFP, EOC, 23.7 pcm/sec):

Pressurizer Pressure ............................................................................. 144 Figure 5-57 Uncontrolled RCCA Bank Withdrawal (HFP, EOC, 23.7 pcm/sec):

Pressurizer Liquid Level......................................................................... 145 Figure 5-58 Uncontrolled RCCA Bank Withdrawal (HFP, EOC, 23.7 pcm/sec):

Reactor Coolant System Temperatures ................................................. 146 Figure 5-59 Uncontrolled RCCA Bank Withdrawal (HFP): DNBR ............................. 147 Figure 5-60 Uncontrolled RCCA Bank Withdrawal (60% RTP): DNBR ..................... 148

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page xi Figure 5-61 Uncontrolled RCCA Bank Withdrawal (10% RTP): DNBR ..................... 149 Figure 5-62 RCCA Drop (25 pcm): Core Power ........................................................ 156 Figure 5-63 RCCA Drop (25 pcm): Reactor Coolant System Temperatures ............. 157 Figure 5-64 RCCA Drop (25 pcm): Pressurizer Pressure ......................................... 158 Figure 5-65 Inadvertent Opening of a Pressurizer Safety Valve: Core Power and Core Average Heat Flux ................................................................. 168 Figure 5-66 Inadvertent Opening of a Pressurizer Safety Valve: Pressurizer Pressure ................................................................................................ 169 Figure 5-67 Inadvertent Opening of a Pressurizer Safety Valve: Reactor Coolant System Temperatures .............................................................. 170

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page xii Nomenclature Acronym Definition AC Alternating Current AFW Auxiliary Feedwater ANS American Nuclear Society AOR Analysis of Record BDMS Boron Dilution Mitigation System BOC Beginning of Cycle BWR Boiling Water Reactor CCP Centrifugal Charging Pump CHF Critical Heat Flux CVCS Chemical and Volume Control System DNB Departure from Nucleate Boiling DNBR Departure from Nucleate Boiling Ratio DTC Doppler Temperature Coefficient ECCS Emergency Core Cooling System EOC End of Cycle ESFAS Engineered Safety Feature Actuation System FCM Fuel Centerline Melt FSAR SP Final Safety Analysis Report (Standard Plant)

HFP Hot Full Power HMP High Mechanical Performance HZP Hot Zero Power IGM Intermediate GAIA Mixer LAR License Amendment Request LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LOOP Loss of Offsite Power MDNBR Minimum Departure from Nucleate Boiling Ratio MFW Main Feedwater MSIV Main Steam Isolation Valve MSLB Main Steam Line Break MSSV Main Steam Safety Valve MTC Moderator Temperature Coefficient NR Narrow Range NRC Nuclear Regulatory Commission OPT Overpower Delta Temperature OTT Overtemperature Delta Temperature PLHGR Peak Linear Heat Generation Rate

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page xiii Acronym Definition PORV Power Operated Relief Valve PWR Pressurized Water Reactor RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System RPS Reactor Protection System RTP Rated Thermal Power SAFDL Specified Acceptable Fuel Design Limit SER Safety Evaluation Report SG Steam Generator TCD Thermal Conductivity Degradation TCV Turbine Control Valve TS/COLR Technical Specifications/Core Operating Limits Report VQP Vendor Qualification Program

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 1

1.0 INTRODUCTION

The purpose of this document is to provide Chapter 15 non-loss of coolant accident (LOCA) input to the license amendment request (LAR) in support of the vendor qualification program (VQP) for the use of Framatome fuel in Callaway. A summary of the Chapter 15 non-LOCA transient analyses and evaluations is contained herein.

The rod cluster control assembly (RCCA) ejection event (FSAR SP 15.4.8) analysis using Framatomes NRC-approved AREA methodology (Reference 14) is addressed in the LAR documentation.

Section 2.0 summarizes the results of the analyses and evaluations performed. Section 3.0 presents the initial conditions, plant operating conditions, trip setpoints, equipment capability, analysis methods and computer codes. Section 4.0 discusses the Chapter 15 disposition of events supporting the VQP. Section 5.0 presents a summary of the non-LOCA event analyses performed for the VQP.

2.0

SUMMARY

OF RESULTS Chapter 15 non-LOCA event analyses were performed in support of licensing the Framatome GAIA 17x17 fuel design for Callaway. Departure from nucleate boiling (DNB) and fuel centerline melt (FCM) specified acceptable fuel design limits (SAFDL) were shown to be met as well as other event-specific criteria such as the time-to-criticality for the inadvertent boron dilution event. Events that challenge system-related design limits, e.g., system overpressure, are not significantly impacted by the Framatome fuel design and, therefore, the existing analyses of record (AOR) remain applicable.

3.0 CHAPTER 15 GENERAL DESCRIPTION 3.1 Initial Conditions Key operating parameters for the Callaway plant that are considered in transient analyses include:

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 2 x Rated core power level is 3565 MWt.

x Framatomes GAIA 17x17 fuel design for Westinghouse-type pressurized water reactors (PWR).

x Nominal hot full power (HFP) average reactor coolant temperature is 588.4°F.

x Nominal hot zero power (HZP) reactor vessel average temperature is 556.8°F.

x Technical Specification minimum reactor coolant system (RCS) flow rate is 374,400 gpm.

x Maximum steam generator (SG) tube plugging is 5%.

x Nominal pressurizer pressure is 2235 psig.

x Technical Specifications/Core Operating Limits Report (TS/COLR) radial peaking factor (F+) limit is 1.65 at HFP.

x TS/COLR local peaking factor limit (FQ) is 2.5.

x Maximum core bypass flow is 8.6%.

In accordance with the Reference 1 methodology, initial conditions include treatment of measurement uncertainties. For the transient analyses, uncertainties are deterministically applied in an additive fashion to ensure a conservative analysis, e.g.,

the uncertainty for core power is conservatively biased and Technical Specification minimum flow is modeled. Nominal values of pressurizer pressure and reactor vessel average temperature are used in the transient analyses to predict more realistic protective system responses. Initial condition measurement uncertainties are either treated deterministically or statistically in the DNB calculations. The system related uncertainties included in the safety analyses are:

x Power measurement uncertainty is +/-2% rated thermal power (RTP).

x Pressurizer pressure measurement uncertainty and deadband is +30/-60 psi.

x Reactor vessel average temperature measurement uncertainty and allowance for steady-state fluctuations is +4.3/-3.5°F.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 3 x RCS flow rate is assumed to be the thermal design flow and no steady-state errors are applied.

Table 3-1 summarizes the principal initial conditions, computer codes used, DNB correlations, and thermal hydraulic methods. Other event specific initial conditions are given in those sections describing the event. The level of SG tube plugging assumed for each transient is listed in Table 3-1.

3.2 Component Setpoints and Capacities Table 3-2 shows the key component setpoints and capacities supported by the analyses. In accordance with the approved Reference 1 methodology, only safety grade equipment is credited to mitigate the consequences of an event. Non-safety grade equipment is modeled when, under normal operation, event consequences are exacerbated. For example, the pressurizer power operated relief valves (PORVs) and pressurizer spray system are assumed operable while the pressurizer heaters are assumed inoperable for departure from nucleate boiling ratio (DNBR) transient events where suppressing primary side pressure is conservative.

3.3 Plant Operational Modes Table 3-3 lists the plant operational modes, consistent with the plant Technical Specifications (Reference 2), which are supported by the Framatome non-LOCA safety analyses.

3.4 RPS and ESFAS Functions Plant Technical Specifications define the available reactor protection system (RPS) and engineered safety feature actuation system (ESFAS) functions. Table 3-4 and Table 3-5 list the RPS functions and Table 3-6 list the ESFAS functions credited in the non-LOCA transient analyses. Uncertainties and response times associated with each of these functions are also given in Table 3-4 and Table 3-6. The setpoints and response times modeled in the transient analyses are conservatively applied to provide bounding simulations of the plant response. To the extent that the RPS and ESFAS are

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 4 credited in the event analyses, the setpoints have been verified to adequately protect plant operation with Framatome fuel.

3.5 Fuel Mechanical Design The Framatome fuel design incorporates GAIA grid spacers, intermediate GAIA mixing (IGM) grid spacers, high mechanical performance (HMP) lower and upper spacer grids, and a GRIP lower tie plate. The fuel rod contains M5 cladding material and the guide tubes and instrument tube use Q12 cladding material. Key fuel design parameters that affect the analyzed transient events are summarized in Table 3-7.

3.6 Peaking Factors The power distribution limits are shown in Table 3-8. Current TS/COLR radial peaking factor (F+) and local peaking factor (FQ) limits are supported by the analyses. The F+

is important for transients that are analyzed to assess DNB concerns. For events like uncontrolled RCCA bank withdrawal from HZP that challenge FCM, a hot spot model is used with event specific power peaking factors to calculate peak fuel centerline temperature which is compared to the fuel melt temperature. For other events that evolve more slowly, power peaking factors are combined to determine peak linear heat generation rate (PLHGR) which is compared to a linear heat generation rate (LHGR) corresponding to the FCM temperature. Power peaking factors are also used as input to assess DNB margin. The power peaking factors mentioned above may be event specific (if required) to account for asymmetries in the reactor core response resulting from the event. An engineering factor of 3% is included in the analyses to account for manufacturing tolerances.

The continued applicability of the power distributions input to the DNB and peak LHGR calculations are verified each reload.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 5 3.7 Reactivity Coefficients Transient response of the reactor core is dependent on reactivity feedback effects, in particular the moderator and Doppler feedback. Depending on the event-specific characteristics, e.g., RCS heatup or cooldown, conservatism dictates the use of either maximum or minimum reactivity coefficient values. Justification for the use of the reactivity coefficient values is treated on an event-specific basis. Table 3-9 presents the key core kinetics parameters and reactivity feedback coefficients supported by the transient analyses. The current TS/COLR limits on moderator temperature coefficient (MTC) are supported. The Doppler reactivity coefficients include biases according to the approved Reference 1 methodology with additional conservatism to bound potential cycle-to-cycle changes. The continued applicability of the transient analyses is verified for each fuel reload.

3.8 RCCA Insertion Characteristics A time delay for the trip breakers open and the RCCAs to start to insert into the core is accounted for in the RPS trip delays given in Table 3-4. The delay includes the time required to process the trip signal and for the magnetic flux of the RCCA holding coils to decay sufficiently to release the RCCAs. The maximum Technical Specification time for the RCCAs to reach the entrance of the guide tube dashpot is 2.7 seconds.

For events initiated from HFP conditions, a conservative minimum HFP scram worth is used which accounts for the most reactive RCCA being fully withdrawn. For events initiated from HZP and part-power conditions, the scram worth is set to the TS/COLR minimum shutdown margin requirement (i.e., 1300 pcm). The shutdown margin requirements are verified for each reload cycle.

3.9 Analysis Methodologies 3.9.1 Methodology Description The approved methodology for evaluating non-LOCA transients is described in Reference 1. For each non-LOCA transient event analysis, the nodalization, chosen

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 6 parameters, conservative input and sensitivity studies are reviewed for applicability to the Framatome VQP in compliance with the safety evaluation report (SER) for Revision 0 of the non-LOCA topical report (Reference 1).

x The nodalization used for the calculations supporting the Framatome VQP is specific to Callaway and is consistent with the Reference 1 methodology.

x The parameters and equipment states are chosen to provide a conservative estimate of the challenge to the acceptance criteria. The biasing and assumptions for key input parameters are consistent with or conservative to the approved Reference 1 methodology.

x The S-RELAP5 code assessments in Reference 1 validated the ability of the code to predict the response of the primary and secondary systems to non-LOCA transients and accidents. No additional model sensitivity studies are needed for this application.

The following changes are made in the non-LOCA system transient analyses and downstream analyses; however, these changes are within the scope of the approved Reference 1 methodology.

x The RODEX2 code is replaced with the COPERNIC code (Reference 7) for the purpose of generating the fuel thermal-conductivity, heat capacity and fuel pellet-to-clad gap coefficient inputs for the average core and hot spot models in the S-RELAP5 code. This change is made to explicitly account for the effects of thermal conductivity degradation (TCD). The properties from the COPERNIC code are developed for beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions in accordance with Reference 1. The COPERNIC fuel properties and gap coefficients are conservatively implemented in the S-RELAP5 model as approved in Reference 1.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 7 As stated in Section 5.4 of Reference 1, there is flexibility in the methodology to accommodate vendor and reactor type differences, as well as different approaches to various aspects of MSLB analysis, such as reactivity feedback and mixing within the reactor pressure vessel. Therefore, the above changes are consistent with and permissible within the Reference 1 methodology.

x The COBRA-FLX code can be used in place of the XCOBRA-IIIC code for any DNB analyses as indicated in the SER in Reference 12. This flexibility is exercised in the HZP increase in feedwater flow and post-scram MSLB events in Section 5.2 and 5.5, respectively. As allowed by Reference 12, future cycle specific analysis may utilize the COBRA-FLX code in place of the XCOBRA-IIIC code for any DNB analysis as long as the appropriate critical heat flux (CHF) correlation design limit is chosen.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 8 The DNB and FCM analyses for the Callaway VQP are performed in accordance with the approved methodology described in Reference 1 and the approved methodology for performing statistical DNB and FCM analyses in Reference 5. As specified in the SER conditions, the statistical treatment of specific variables analyzed are consistent with Reference 5 with one addition. The core inlet temperature has been justified to be treated statistically for statistical DNB analyses. The statistical DNB and FCM analyses described in Reference 5 do not preclude the deterministic treatment of any, or all, of the uncertainty parameters. The approved methodology for performing DNB calculations using the XCOBRA-IIIC code in a mixed core is described in Reference 4.

A mixed core penalty of 2% is applied to the CHF correlation limit in accordance with the Reference 4 SER for DNB analyses performed.

DNB calculations performed with the XCOBRA-IIIC code are in accordance with Reference 8. The SER for the Reference 8 topical report states that the use of XCOBRA-IIIC is limited to the snapshot mode. Thus, minimum departure from nucleate boiling ratio (MDNBR) calculations are performed using a steady-state XCOBRA-IIIC model with core boundary conditions at the time of MDNBR from the S-RELAP5 transient analyses.

DNB calculations performed with COBRA-FLX are in accordance with Reference 12 using the same methodology as XCOBRA-IIIC (Reference 8). The solver options and empirical models approved in the limitations and conditions of Reference 12 are selected for COBRA-FLX analyses. Additionally, the fuel rod model and post-CHF heat transfer models within COBRA-FLX are not used. To be consistent with the methodology used for XCOBRA-IIIC, MDNBR calculations are performed using a steady-state COBRA-FLX model with core boundary conditions at the time of MDNBR from the S-RELAP5 transient analyses. COBRA-FLX can be used in place of XCOBRA-IIIC for any DNB analyses as indicated in the SER in Reference 12.

The DNB calculations are performed utilizing the Nuclear Regulatory Commission (NRC)-approved ORFEO-NMGRID and ORFEO-GAIA CHF correlations described in

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 9 the Reference 6 topical report. The fuel design parameters for Framatomes 17x17 GAIA assembly are within the applicable range for the ORFEO-NMGRID and ORFEO-GAIA CHF correlations. The Framatome fuel transition operating conditions are within the applicable range of coolant conditions for the ORFEO-NMGIRD and ORFEO-GAIA correlations. The conditions specified in the Reference 6 SER are addressed below:

1) The inlet subcooling is always greater than 0 degrees for DNB analyses.

2) ORFEO-NMGRID is not used for analyses in the [ ]

subregion.

3) Analyses where ORFEO-NMGRID or ORFEO-GAIA are used in the low-quality region (e.g., post-scram MSLB) are confirmed to be non-limiting in the low-quality region. Limiting in this context is defined as the scenario in which the event is approaching the design limit.

The SER for Reference 6 states that any application deviation from the modeling options or the use of COBRA-FLX would require re-validation similar to the validation provided in the Topical Report. The ORFEO-GAIA and ORFEO-NMGRID CHF correlations were incorporated into XCOBRA-IIIC and were re-validated using a similar validation methodology as Reference 6. The validated design limits for ORFEO-GAIA and ORFEO-NMGRID in XCOBRA-IIIC are provided in Table 3-10. A mixed core penalty of 2% is applied to the CHF correlation limit in accordance with Reference 4 for transition analysis performed with XCOBRA-IIIC. The mixed core penalty is only required for transition cycles containing hydraulically dissimilar fuel assembly types.

The DNB calculations with COBRA-FLX are performed utilizing the NRC-approved ORFEO-NMGRID and ORFEO-GAIA CHF correlations described in the Reference 6 topical report. The SER for Reference 6 states that any application deviation from the modeling options or the use of COBRA-FLX would require re-validation similar to the validation provided in the Topical Report. The application of the ORFEO-NMGRID CHF correlation was extended to the PV-Solver using COBRA-FLX and was validated in full compliance with the SER limitation indicated above. The design limit for the Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 10 ORFEO-NMGRID CHF correlation in COBRA-FLX using the PV-Solver is shown in Table 3-10. This extension was performed to support the post-scram MSLB with a loss of offsite power analysis (Section 5.5); however, this does not preclude using the ORFEO-NMGRID CHF correlation with the PV-Solver for other events where appropriate. Since COBRA-FLX is replacing XCOBRA-IIIC for certain applications supporting the Framatome VQP, the same mixed core penalty from Reference 4 is applied to transition cycles containing hydraulically dissimilar fuel assembly types in COBRA-FLX.

Protection against FCM is expressed as a limit on LHGR allowed in the core. A FCM limit is established for UO2 fuel rods such that, FCM is precluded for all fuel rod types.

This methodology is described in Reference 5, Appendix A. The FCM limit is verified each reload. No restrictions, limitations, and/or conditions are identified in the SER for Reference 5, Appendix A relative to the FCM limit calculation.

The quantity of fuel failures includes the effects of DNB propagation as defined in Reference 13, Supplement 5. This method is typically applied conservatively by assuming a single rod failure propagates to the entire assembly. No restrictions, limitations, and/or conditions are identified in the SER for Reference 13, Supplement 5 relative to DNB propagation.

Reference 11 incorporates M5 cladding properties into the S-RELAP5 based non-LOCA methodology. No restrictions or requirements are identified in the SER for the Reference 11 methodology relative to its application to S-RELAP5 non-LOCA analyses.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 11 3.9.2 Methodology Changes The method used for the non-LOCA system transient analyses and downstream analyses differ from that in the approved Reference 1 topical report as described below.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 12 x The ARTEMIS code (References 9 and 10) is used to calculate the radial and axial power distribution and the reactivity verification for the HZP increase in feedwater flow (Section 5.2) and post-scram MSLB (Section 5.5) events. The ARTEMIS and COBRA-FLX codes (Reference 12) are internally coupled. The ARTEMIS code calculates the power distribution and core reactivity under the imposed conditions within the core and feeds that information to the COBRA-FLX code. The COBRA-FLX code calculates moderator densities and temperatures to transfer back to the ARTEMIS code. This iteration is performed until both codes converge. The coupling process to obtain the power distribution and core reactivity is similar to that described in Reference 1, Section 5.4.4.1.

The method used for the calculation of the LHGR that corresponds to the fuel centerline melt temperature limits differs from that in the approved Reference 5 topical report as described below.

x RODEX2 is replaced with COPERNIC (Reference 7) for the purpose of generating FCM limits for each rod type. This change is made to explicitly account for the effects of TCD. No other modifications are made to the Reference 5, Appendix A methodology.

The method used for COBRA-FLX (standalone and within ARTEMIS) differs from that in the approved Reference 12 topical report as described below.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 13 3.10 Computer Codes Descriptions of the computer codes used in the safety analyses are provided below.

Table 3-1 lists the principal computer codes used in each of the non-LOCA analyses.

x S-RELAP5 - The S-RELAP5 code is a Framatome modification of the RELAP5/MOD2 code. The S-RELAP5 code is used for simulation of the system response to transient events. Control volumes and junctions are defined which describe the major components in the primary and secondary systems that are important for the event being analyzed. The S-RELAP5 hydrodynamic model is a two-dimensional, transient, two-fluid model for flow of a two-phase steam-water mixture. The S-RELAP5 code uses a six-equation model for the hydraulic solutions. These equations include two-phase continuity equations, two-phase momentum equations, and two-phase energy equations. The six-equation model also allows both non-homogeneous and non-equilibrium situations encountered in reactor problems to be modeled.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 14 The S-RELAP5 code is used to calculate neutron power, fuel thermal response, surface heat transport, and fluid conditions (such as coolant flow rates, temperatures, and pressures). Core boundary conditions calculated by the S-RELAP5 code are used in subsequent MDNBR and PLHGR calculations. The S-RELAP5 model also contains a hot spot used to calculate fuel centerline temperatures for events that are initially mitigated by Doppler reactivity feedback followed by reactor trip.

x COPERNIC - The COPERNIC code (Reference 7) performs thermal-mechanical calculations for a fuel rod under normal operating conditions. The code incorporates models to describe the thermal-hydraulic condition of the fuel rod in a flow channel; the gas release, swelling, densification and cracking in the pellet; the gap conductance; the radial thermal conduction; the free volume and gas pressure internal to the fuel rod; the fuel and cladding deformations; and the cladding corrosion. The code has been extensively benchmarked, and its predictive capabilities are correlated over a wide range of conditions applicable to light water reactor fuel conditions.

The COPERNIC code accounts for TCD with increasing rod exposure. To account for the effects of TCD in the S-RELAP5 simulations, the COPERNIC code is used to generate the fuel thermal-conductivity, heat capacity and fuel pellet-to-clad gap coefficient inputs for the average core and hot spot models.

The properties from the COPERNIC code are developed in accordance with Reference 1, and the COPERNIC code replaces the RODEX2 code for this purpose in the approved topical report. The COPERNIC fuel properties and gap coefficients are conservatively implemented as approved in Reference 1.

Average core and hot spot fuel properties and gap coefficients conservatively bound UO2 and Gd2-O3 bearing fuel rods.

The COPERNIC code is also used to establish the fuel centerline melt LHGR limit as a function of exposure.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 15 x XCOBRA-IIIC - The XCOBRA-IIIC code (Reference 8) is a steady-state thermal-hydraulics code that calculates the axial and radial flow and enthalpy distributions within assemblies and sub-channels for non-LOCA events. When used in conjunction with core boundary conditions from the S-RELAP5 transient analysis and the ORFEO-GAIA and ORFEO-NMGRID correlations (Reference 6), the XCOBRA-IIIC code also calculates the corresponding MDNBR. MDNBR calculations are performed in a two-step process. Calculations are first performed on a core-wide basis to calculate the axially varying flow and enthalpy distributions in the peak powered fuel assembly. Next, these flow and enthalpy boundary conditions are applied to a sub-channel model of the peak powered assembly to determine the local conditions for the calculation of MDNBR.

x COBRA-FLX - The COBRA-FLX code (Reference 12) is a steady-state and transient thermal-hydraulics code. The COBRA-FLX code calculates the axial and lateral flow, pressure, and enthalpy distribution within assemblies and sub-channels. The COBRA-FLX code can calculate MDNBR when used in conjunction with boundary conditions provided from S-RELAP5 transient analysis and the ORFEO-GAIA and ORFEO-NMGRID CHF correlations (Reference 6).

The COBRA-FLX code is also used as the thermal-hydraulic module for the core simulator ARTEMIS within the ARCADIA code package for coupled neutronic and thermal-hydraulic analysis (References 9 and 10).

x ARTEMIS - Framatomes PWR neutronics methodology uses the NRC approved advanced code package, ARCADIA. This system is built around the lattice code APOLLO2-A, the 3D nodal core simulator ARTEMIS, and the 3D core thermal-hydraulics code COBRA-FLX. The COBRA-FLX code acts as the thermal-hydraulics solver within the ARTEMIS code. The ARTEMIS code is a coarse-mesh core simulator code using 2-group cross section data provided by the APOLLO2-A code. The ARTEMIS code models the core in three dimensions and all calculations are performed in a full core configuration, explicitly capturing asymmetries. The ARTEMIS code calculates reactor core reactivity, nodal power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 16 distribution, pin power distribution, in-core, and excore detector responses and can be used to simulate fuel shuffling, insertion, and discharge. NRC approval for the ARCADIA code system is provided in References 9 and 10.

3.11 Event Classification The Chapter 15 non-LOCA events are organized according to the characteristics of the transients. Each event is placed in one of four categories adopted by the American Nuclear Society (ANS) (Reference 3). Event acceptance criteria are defined based on the frequency of occurrence and event consequences. Table 3-11 summarizes the non-LOCA event classifications and acceptance criteria for Callaway.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 17 Table 3-1 Summary of Initial Conditions and Computer Codes Used Principal Moderator Doppler Statistical DNB FSAR SP Event Computer Feedback Feedback DNB Correlation Codes Used (pcm/°F) (pcm/°F) Method S-RELAP5 15.1.1 Decrease in feedwater temperature -47.9 -1.23 ORFEO-GAIA No XCOBRA-IIIC S-RELAP5 Increase in feedwater flow (HFP) -1.23 ORFEO-GAIA XCOBRA-IIIC 15.1.2 S-RELAP5 -47.9 No Function of ORFEO-Increase in feedwater flow (HZP) COBRA-FLX power NMGRID ARTEMIS S-RELAP5 15.1.3 Excessive increase in steam flow -47.9 -1.23 ORFEO-GAIA No XCOBRA-IIIC Steam system piping failure (HFP) S-RELAP5 Function of fuel ORFEO-15.1.5 COBRA-FLX -47.9 No Steam system piping failure (HZP) temp. or power NMGRID ARTEMIS S-RELAP5 Function of Function of fuel 15.1.5.6 Steam system piping failure at full power moderator ORFEO-GAIA Yes XCOBRA-IIIC temp.

density Complete loss of forced reactor coolant S-RELAP5 15.3.2 0.0 -0.91 ORFEO-GAIA No flow XCOBRA-IIIC S-RELAP5 15.3.3 RCP shaft seizure 0.0 -0.91 ORFEO-GAIA Yes XCOBRA-IIIC

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 18 Core Reactor Reactor Pressurizer Pressurizer Feedwater SG Tube Thermal Coolant Coolant FSAR SP Event Pressure Liq. Level Temp. Plugging Power Flow T-Avg (psig) (% span) (°F) (%)

(% RTP) (gpm) (°F)

Decrease in feedwater 15.1.1 102 374,400 588.4 2235 60 446 0 temperature Increase in feedwater 102 588.4 60 390 flow (HFP) 15.1.2 374,400 2235 0 Increase in feedwater HZP 556.8 25 100 flow (HZP)

Excessive increase in 15.1.3 102 374,400 588.4 2235 60 390 0 steam flow Steam system piping 100 588.4 60 390 failure (HFP) 15.1.5 374,400 2235 0 Steam system piping HZP 556.8 25 100 failure (HZP)

Steam system piping 15.1.5.6 102 374,400 588.4 2235 60 390 0 failure at full power Complete loss of forced 15.3.2 102 374,400 588.4 2235 60 446 5 reactor coolant flow 15.3.3 RCP shaft seizure 102 374,400 588.4 2235 60 446 5

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 19 Principal Moderator Doppler Statistical DNB FSAR SP Event Computer Feedback Feedback DNB Correlation Codes Used (pcm/°F) (pcm/°F) Method Uncontrolled RCCA bank withdrawal from S-RELAP5 Function of fuel 15.4.1 5.0 ORFEO-GAIA No a subcritical or low power startup condition XCOBRA-IIIC temp.

Uncontrolled RCCA bank withdrawal at power (HFP)

Uncontrolled RCCA bank withdrawal at S-RELAP5 15.4.2 Spectrum Spectrum ORFEO-GAIA Yes power (60%) XCOBRA-IIIC Uncontrolled RCCA bank withdrawal at power (10%)

S-RELAP5 15.4.3 RCCA drop -47.9 -2.9 ORFEO-GAIA No XCOBRA-IIIC S-RELAP5 15.4.3 Single RCCA rod withdrawal Spectrum Spectrum ORFEO-GAIA No XCOBRA-IIIC S-RELAP5 Function of Inadvertent opening of a pressurizer 15.6.1 moderator -0.91 ORFEO-GAIA No safety or relief valve XCOBRA-IIIC density

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 20 Core Reactor Reactor Pressurizer Pressurizer Feedwater SG Tube Thermal Coolant Coolant FSAR SP Event Pressure Liq. Level Temp. Plugging Power Flow T-Avg (psig) (% span) (°F) (%)

(% RTP) (gpm) (°F)

Uncontrolled RCCA bank withdrawal from a 15.4.1 HZP 187,200 556.8 2235 25 100 5 subcritical or low power startup condition Uncontrolled RCCA bank withdrawal at 102 588.4 60 446 power (HFP)

Uncontrolled RCCA 15.4.2 bank withdrawal at 62 374,400 575.8 2235 46 414 5 power (60%)

Uncontrolled RCCA bank withdrawal at 12 560.0 28.5 337.5 power (10%)

15.4.3 RCCA drop 102 374,400 588.4 2235 60 446 5 Single RCCA rod 15.4.3 102 374,400 588.4 2235 60 446 5 withdrawal Inadvertent opening of 15.6.1 a pressurizer safety or 102 374,400 588.4 2235 60 446 5 relief valve

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 21 Table 3-2 Key Component Setpoints and Capacities Nominal Setpoint Item Total Capacity Setpoint Tolerance 1,260,0000 lbm/hr at 2485 Pressurizer safety 2460 psig +/-2% psig valves (3 valves) 420,000 lbm/hr at 2335 psig Pressurizer PORVs 2335 psig +/-30.46 psi (2 valves)

Main steam safety (5 valves per steam line) valves x MSSV-1 1185 psig 893,160 lbm/hr at 1221 psig x MSSV-2 1197 psig 902,096 lbm/hr at 1233 psig x MSSV-3 1210 psig +3%, -1% 911,779 lbm/hr at 1246 psig x MSSV-4 1222 psig 920,715 lbm/hr at 1259 psig x MSSV-5 1234 psig 929,652 lbm/hr at 1271 psig 2260 psig Pressurizer spray (start)

--- 900 gpm (2 valves) valves 2310 psig (full on)

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 22 Table 3-3 Plant Operational Modes Average Reactor Reactivity  % Rated Thermal Operational Mode Coolant Temperature Condition (keff) Power (Note 1)

(°F)

1. Power Operation 0.99 > 5% N/A

2. Startup 0.99 5% N/A

3. Hot Standby < 0.99 N/A 350

4. Hot Shutdown < 0.99 N/A 350 > Tavg > 200 (Note 2)

5. Cold Shutdown < 0.99 N/A 200 (Note 2)

6. Refueling N/A N/A N/A (Note 3)

Notes:

1. Excluding decay heat.

2. At least 53 of 54 reactor vessel head closure bolts fully tensioned.

3. Two or more reactor vessel head closure bolts less than fully tensioned.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 23 Table 3-4 RPS Trip Setpoints and Response Times Analytical Response Time Trip Allowable Trip Setpoint Value (Note 1)

Power range neutron flux 0.5 sec x High setting 112.3% RTP 118% RTP x Low setting 28.3% RTP 35% RTP OvertHPSHUDWXUH7 Note 4 x K1 1.226 1.355 x K2 0.019/°F same x K3 0.0011/psi same x  0 sec same x  0 sec same x  4 sec same x  27 sec same x  4 sec same x  2 sec same OverpRZHU7 Note 4 x K4 1.1073 1.1665 x

0.02/°F (incr. Tavg)

K5 same 0.0/°F (decr. Tavg) x 0.0015/°F (Tavg > Tavg at 100% RTP)

K6 same 0.0/°F (Tavg 7avg at 100% RTP) x  VDPHDV277 same x  10 sec same Reactor coolant flow - low

 87% 1.0 sec (Note 2)

RCP underfrequency +] 57 Hz 0.6 sec RCP undervoltage 9$& Note 3 1.5 sec Pressurizer pressure 2.0 sec x High SVLJ 2420 psig x Low SVLJ 1845 psig Steam generator level - low 2.0 sec x Adverse containment

15VSDQ 0% NR span environment x Normal containment

15VSDQ 0% NR span environment Notes:

1. The total delay to trip is defined as the time delay from the time that trip conditions are reached to the time the rods are free and begin to fall.

2. Percent of flow with four RCPs operating.

3. Trip is assumed to occur at event initiation.

4. Trip delay accounts for a resistance temperature detector lag time of 6 seconds and 2 seconds trip processing time.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 24 Table 3-5 Credited RPS Trip Functions FSAR SP Description Credited RPS Trips Event Power range neutron flux (high setting) 15.1.1 Decrease in feedwater temperature 277 237 Low-low SG level Power range neutron flux (high setting) 15.1.2 Increase in feedwater flow 277 237 Low pressurizer pressure Power range neutron flux (high setting) 15.1.3 Excessive increase in secondary steam flow 277 237 237 15.1.5.6 Steam system piping failure at full power Safety injection (low steam line pressure)

RCP undervoltage 15.3.2 Complete loss of forced reactor coolant flow Low RCS flow 15.3.3 RCP shaft seizure Low RCS flow Uncontrolled RCCA bank withdrawal from a subcritical or Power range neutron flux (low 15.4.1 low-power startup condition setting)

Power range neutron flux (high setting) 15.4.2 Uncontrolled RCCA bank withdrawal at power OTT 237 15.4.3 Dropped RCCA/RCCA bank Low pressurizer pressure Inadvertent opening of a pressurizer safety or relief valve OTT 15.6.1 Low pressurizer pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 25 Table 3-6 ESFAS Setpoints and Response Times Allowable Analytical Response Time Actuation Setpoint Setpoint (sec) 2.0 sec Steam line pressure - low SVLJ 458 psig (Note 1)

Pressurizer pressure - low SVLJ 1700 psig 2.0 sec Steam generator water level - 15VSDQ 100% NR span 2.0 sec high-high Notes:

1. 7LPHFRQVWDQWVXVHGLQWKHOHDGODJFRQWUROOHUDUHVHFRQGVDQGVHFRQGV

Table 3-7 Fuel Design Parameters Parameter Value Fuel assembly array 17 x 17 Number of fuel rods per assembly 264 Guide tubes per assembly 24 Instrument tubes per assembly 1 Fuel rod pitch, inches 0.496 Fuel pellet outside diameter, inches 0.3225 Clad inside diameter, inches 0.329 Clad outside diameter, inches 0.374 Heated fuel length, inches 144.0 Number of spacers x High mechanical performance 2 x GAIA mixing grid 6 x Intermediate GAIA mixing grid 3

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 26 Table 3-8 Core Power Distribution Parameters Parameter Value F+ (TS/COLR limit at HFP) 1.65 FQ (TS/COLR limit) 2.50 Engineering tolerance 3%

Table 3-9 Reactivity Parameters Parameter BOC EOC MTC (TS/COLR limits), pcm/°F Note 1 -47.9 DTC, pcm/°F (bounding analysis value range) (Note 2) -0.91 -2.9 Delayed neutron fraction 0.006279 0.005267 Notes:

1. TS/COLR limit of +5 pcm/°F for core power 573and 0 pcm/°F for core power >

70% RTP.

2. Doppler temperature coefficients include bias required by the Reference 1 methodology.

Table 3-10 ORFEO-GAIA and ORFEO-NMGRID Design Limits Subchannel Thermal-Hydraulics Code Design Limit ORFEO-GAIA = 1.12 XCOBRA-IIIC ORFEO-NMGRID = 1.15 COBRA-FLX (PV Solver) ORFEO-NMGRID = 1.18

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 27 Table 3-11 Event Classification and Acceptance Criteria ANS Classification Acceptance Criteria x Condition I events are considered from the point of view of Condition I: Normal affecting the consequences of fault conditions (Conditions II, III and operation and IV events). In this regard, fault conditions are generally based on a operational transients conservative set of initial conditions corresponding to adverse conditions which can occur during Condition I operation x Pressures in reactor coolant and main steam systems should be less than 110% of design values.

x Fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded by assuring that the minimum Condition II: Faults of calculated DNBR does not exceed the applicable limits of the moderate frequency DNBR correlation being used.

x Fuel centerline temperatures do not exceed the melting point.

x The event should not generate a more serious plant condition without other faults occurring independently.

x A small fraction of fuel failures may occur.

x Radiological consequences should be a small fraction of 10 CFR Condition III:

100 guidelines.

x Infrequent faults The event should not generate a limiting fault or result in the consequential loss of the reactor coolant or containment barriers.

x Radiological consequences should not exceed 10 CFR 100 guidelines.

x Condition IV: Limiting faults The event should not cause a consequential loss of the required functions of systems needed to cope with the reactor coolant and containment systems.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 28 4.0 CHAPTER 15 DISPOSITION OF EVENTS A summary of the Chapter 15 non-LOCA disposition of events is provided in Table 4-1.

Each event is categorized with respect to its potential consequences according to the categories discussed in Section 3.11. For the VQP, detailed analyses are performed for the non-LOCA events that potentially challenge the DNBR and FCM SAFDLs, as well as event specific criteria such as time-to-criticality for boron dilution.

Several Callaway Chapter 15 non-LOCA events are affected by the transition to Framatome GAIA fuel, specifically because of changes in thermal-hydraulic performance and neutronics inputs to the safety analyses. On the other hand, events and/or sub-events whose key parameters are plant related system responses, e.g., core power, decay heat, auxiliary feedwater (AFW) capability, offsite power availability, safety injection and/or charging capability, safety valve performance, etc., rather than the fuel design parameters, are not analyzed for the fuel transition and remain bounded by the current AOR.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 29 Table 4-1 Event Disposition Summary of Results FSAR SP ANS Description Disposition Event Condition 15.1 Increase in Heat Removal by the Secondary System 15.1.1 Decrease in feedwater temperature II SAFDLs reanalyzed 15.1.2 Increase in feedwater flow II SAFDLs reanalyzed Excessive increase in secondary 15.1.3 II SAFDLs reanalyzed steam flow Inadvertent opening of a steam 15.1.4 II Bounded generator relief or safety valve 15.1.5 Steam system piping failure III/IV SAFDLs reanalyzed Steam Line break with coincidental 15.1.5.5 --- Not applicable RCCA withdrawal at power Steam system piping failure at full 15.1.5.6 III/IV SAFDLs reanalyzed power 15.2 Decrease in Heat Removal by the Secondary System Steam pressure regulator malfunction 15.2.1 or failure that results in decreasing --- Not applicable steam flow 15.2.2 Loss of external electrical load II Bounded 15.2.3 Turbine trip II Bounded 15.2.4 Inadvertent closure of MSIVs II Bounded Loss of condenser vacuum and other 15.2.5 II Bounded events resulting in turbine trip Loss of nonemergency AC power to 15.2.6 II Bounded the plant auxiliaries 5.2.7 Loss of normal feedwater flow II Bounded 15.2.8 Feedwater system pipe break IV Bounded 15.3 Decrease in Reactor Coolant System Flowrate Partial loss of forced reactor coolant 15.3.1 II Bounded flow Complete loss of forced reactor 15.3.2 III SAFDLs reanalyzed coolant flow 15.3.3 RCP shaft seizure IV SAFDLs reanalyzed 15.3.4 RCP shaft break IV Bounded

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 30 FSAR SP ANS Description Disposition Event Condition 15.4 Reactivity and Power Distribution Anomalies Uncontrolled RCCA bank withdrawal 15.4.1 from a subcritical or low-power II SAFDLs reanalyzed startup condition Uncontrolled RCCA bank withdrawal 15.4.2 II SAFDLs reanalyzed at power RCCA misoperation x Dropped RCCA /RCCA bank II SAFDLs reanalyzed x

15.4.3 Statically misaligned RCCA II Bounded x Withdrawal of a single RCCA III SAFDLs reanalyzed Startup of an inactive RCP at an 15.4.4 --- Not applicable incorrect temperature Malfunction or failure of the flow controller in a BWR loop that results 15.4.5 --- Not applicable in an increased reactor coolant flowrate CVCS malfunction that results in a SAFDLs bounded 15.4.6 decrease in the boron concentration II in the reactor coolant Time-to-criticality analyzed Inadvertent loading and operation of a Evaluation to be provided 15.4.7 III fuel assembly in an improper position by Licensee Fuel-related criteria 15.4.8 Spectrum of RCCA ejection accidents IV reanalyzed but is outside scope of this document 15.5 Increase in Reactor Coolant Inventory Inadvertent operation of the ECCS 15.5.1 II Bounded during power operation CVCS malfunction that increases 15.5.2 II Bounded reactor coolant inventory 15.5.3 Number of BWR transients --- Not applicable 15.6 Decrease in Reactor Coolant Inventory Inadvertent opening of a pressurizer 15.6.1 II/III SAFDLs reanalyzed safety or relief valve Break in instrument line or other lines 15.6.2 from reactor coolant pressure II Bounded boundary that penetrate containment 15.6.3 Steam generator tube failure IV Bounded Spectrum of BWR steam system 15.6.4 --- Not applicable piping failures outside of containment

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 31 FSAR SP ANS Description Disposition Event Condition 15.6.5 Loss-of-coolant accidents Outside scope of this III/IV document 15.6.6 Number of BWR transients --- Not applicable 15.7 Radioactive Release from a Subsystem or Component 15.7.1 Radioactive gas waste system leak or III failure 15.7.2 Radioactive liquid waste system leak Assessment of radiological III or failure doses is outside the scope 15.7.3 Postulated radioactive release due to of this document IV liquid tank failures 15.7.4 Fuel handling accident IV 15.8 Anticipated Transients Without Scram 15.8 Anticipated transients without scram --- Bounded

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 32 5.0 CHAPTER 15 EVENT ANALYSES 5.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature (FSAR SP 15.1.1) 5.1.1 Event Description Reductions in feedwater temperature cause an increase in core power by decreasing the reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary plant and the RCS. The overpower/overtemperature protection (high neutron flux, overtemperature delta temperature (277) and overpower delta temperature (OP7) trips) prevents any power increase which could lead to a MDNBR less than the safety analyses limit value. FCM is also precluded.

A reduction in feedwater temperature may be caused by spurious opening of the low-pressure feedwater heater bypass valve. Subsequently, there is a sudden reduction in feedwater temperature to the inlet of the high-pressure heaters. This increases extraction steam flow to the heaters and could lead to isolation of the heater string followed by a heater drain pump trip. At power, this scenario results in a reduction in feedwater temperature to the SGs and creates a greater load demand on the RCS.

With the plant at no-load conditions, the addition of cold feedwater may cause a decrease in RCS temperature, and a reactivity insertion from negative moderator feedback. However, the rate of energy change is reduced as load and feedwater flow decrease, so the no-load transient is less severe than the HFP case.

The net effect on the RCS due to a reduction in feedwater temperature is like the effect of increasing secondary steam flow, i.e., the reactor will reach a new equilibrium condition at a power level corresponding to the energy extraction rate of the secondary side unless terminated by a reactor trip.

A decrease in normal feedwater temperature is classified as an ANS Condition II event.

See Section 3.11 for a discussion of the event classifications and acceptance criteria.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 33 The protection available to mitigate the consequences of this event is discussed in Section 3.4 and listed in Table 3-4.

5.1.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. This transient is analyzed at HFP conditions with manual rod control since automatic rod control is disabled at Callaway. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6).

The following assumptions are made to conservatively predict the consequences of this event:

x The event is initiated from HFP and the power measurement uncertainty is included in the initial S-RELAP5 power level.

x Due to the cooldown caused by the decrease in main feedwater (MFW) temperature, the power increase at EOC conditions will occur more rapidly and be more severe than at BOC. Since automatic rod control has been disabled at Callaway, only EOC is analyzed. The TS/COLR most-negative MTC limit of -47.9 pcm/°F and an EOC Doppler coefficient result in a conservatively bounding increase in reactor power as the moderator cools down and the fuel heats up.

x EOC Doppler temperature coefficient (DTC) is biased less negative in accordance with the Reference 1 topical report.

x Initial RCS average temperature is the upper bound nominal value for HFP conditions. Temperature measurement uncertainties are applied in the DNBR calculations.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 34 x Reactor trip occurs on the power range neutron flux. The analytical RPS trip setpoint used in the S-RELAP5 calculation includes instrument and setpoint uncertainties. An excore detector decalibration factor for the high neutron flux trip function is conservatively set to [ ] to account for neutron attenuation in the reactor vessel downcomer. Maximum reactor trip response times include delays for trip signal actuation and scram system holding coil release.

x Four reactor coolant pumps (RCP) are modeled to be in operation consistent with Mode 1. RCS flow is modeled as the Technical Specification minimum flow rate.

x Nominal pressurizer pressure is used in the S-RELAP5 calculation. The measurement uncertainty is applied in the DNBR calculations.

x EOC fuel pellet to cladding gap conductance is conservatively modeled to maximize rod surface heat flux.

x EOC average core fuel thermal-mechanical properties are modeled and account for the effect of TCD.

x SG tube plugging is assumed to be zero to minimize the resistance to heat transfer across the SG tubes during the transient, which in turn will maximize the cooldown of the RCS and maximize the positive reactivity insertion.

x Feedwater temperature control is assumed to malfunction resulting in a step decrease to 280°F from the nominal feedwater temperature of 446°F.

x The feedwater flow into the SGs is terminated by actuation of a safety injection signal, from low pressurizer pressure, which closes all feedwater isolation valves.

Main feedwater isolation may also occur on high-high SG level ESFAS trip. The main feedwater isolation valve stroke time is maximized.

x Since the systems designed to mitigate this event are redundant, there is no single active failure that will adversely affect the consequences of the event.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 35 5.1.3 Results The inadvertent opening of the low-pressure feedwater heater bypass valve causes a reduction in feedwater temperature that increases the thermal load on the primary system. The maximum reduction in feedwater temperature is 166°F, resulting in a significant increase in heat load on the primary system.

The HFP case with EOC reactivity feedback coefficients and manual rod control is performed as it results in the lowest DNBR. Automatic rod control is not considered because this function is disabled at Callaway.

The reactor is tripped by high neutron flux signal and MDNBR occurs at about this time.

The addition of the cooler feedwater is terminated once a safety injection signal occurs from a low pressurizer pressure signal which causes all feedwater isolation valves to be automatically closed.

The sequence of events is presented in Table 5-1. Transient results presented in Figure 5-1 through Figure 5-4 show the parameters of interest for this event. Figure 5-1 shows the reactor power as a function of time and Figure 5-2 shows the core power based on rod surface heat flux. Figure 5-3 and Figure 5-4 show pressurizer pressure and RCS temperatures, respectively.

Table 5-2 presents the MDNBR and peak LHGR for this event. The results demonstrate that acceptance criteria are met.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 36 Table 5-1 Decrease in Feedwater Temperature: Sequence of Events Case Event Time (sec.)

Loss of feedwater heaters occurs 0 Reactor Trip setpoint reached 31.8 HFP, EOC, Reactor Trip on high neutron flux w/ delays 32.4 manual rod Turbine trip on reactor trip 32.4 control Minimum DNBR occurs 32.5 Safety injection signal causes MFW isolation signal 87.1 MFW isolation valves fully closed (flow is terminated) 105.0 Table 5-2 Decrease in Feedwater Temperature: Results Criterion Result Limit MDNBR 1.440 1.142 PLHGR (kW/ft) 19.5 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 37 Figure 5-1 Decrease in Feedwater Temperature: Reactor Power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 38 Figure 5-2 Decrease in Feedwater Temperature: Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 39 Figure 5-3 Decrease in Feedwater Temperature: Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 40 Figure 5-4 Decrease in Feedwater Temperature: Reactor Coolant System Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 41 5.2 Feedwater System Malfunctions that Result in an Increase in Feedwater Flow (FSAR SP 15.1.2) 5.2.1 Event Description Addition of excessive feedwater causes an increase in core power by decreasing the reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary plant and of the RCS. The overpower/overtemperature protection (high neutron flux, 277 and OP7WULSV SUHYHQWVDQ\SRZHULQFUHDVHZKLFKFRXOGOHDGWRD

MDNBR less than the safety analysis limit value. None of these trips are actuated in the event analyses.

An example of excessive feedwater flow is a full opening of one or more feedwater control valves due to a feedwater control system malfunction or an operator error. At power, this excess flow causes a greater load demand on the RCS due to increased subcooling in the SG. With the plant at no-load conditions, the addition of cold feedwater may cause a decrease in the RCS temperature and thus a positive reactivity insertion from the negative moderator coefficient.

Continuous addition of excessive feedwater is prevented by the SG high-high level trip, which closes the feedwater isolation valves. This signal also closes the operable feedwater control valves, feedwater pump discharge valves, and trips the main feedwater pumps; however, of this equipment, only the feedwater control valves are part of the primary success path for event mitigation. Turbine trip on SG high-high level is not directly credited in this analysis.

An increase in normal feedwater flow is classified as an ANS Condition II event, fault of moderate frequency. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

Plant systems and equipment, which are available to mitigate the effects of the event, are discussed in Section 3.4 and listed in Table 3-4.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 42 5.2.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. This transient is analyzed at HFP and HZP conditions.

Manual rod control is assumed since automatic rod control is disabled at Callaway. For the HFP cases, the core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8),

which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6). For the HZP cases, the core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the COBRA-FLX code (Reference 8), which is used to calculate the MDNBR using the ORFEO-NMGRID CHF correlation (Reference 6), and the ARTEMIS code (References 9 and 10) is used to calculate power distribution information and kinetics parameters.

The following assumptions are made to conservatively predict the consequences of this event:

x The event is initiated from HFP and HZP initial power levels. Power measurement uncertainty is included in the S-RELAP5 analyses initiated from HFP because higher power is conservative for DNBR calculations. The initial power level for the HZP case is set at 10-9 of rated thermal power, which leads to the maximum power excursion during the event.

x Due to the cooldown caused by the increase in MFW flow rate, the power increase at EOC conditions occurs more rapidly and more severely than at BOC.

Since automatic rod control has been disabled at Callaway, only EOC is analyzed. The TS/COLR most-negative MTC limit of -47.9 pcm/°F and biasing the nominal EOC Doppler feedback less negative result in a conservatively bounding increase in reactor power as the moderator cools down and the fuel heats up.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 43 EOC Doppler reactivity or DTC is biased less negative in accordance with the Reference 1 topical report.

x Initial RCS average temperature modeled in S-RELAP5 is the upper bound nominal value for the power level analyzed. Temperature measurement uncertainties are applied in the DNBR calculations.

x Reactor trip may RFFXURQHLWKHUWKHSRZHUUDQJHQHXWURQIOX[277DQG237 trip functions. Reactor trip may also occur on the low-low SG level trip which results from MFW isolation on a high-high SG level ESFAS trip. The analytical RPS trip setpoints used in the S-RELAP5 calculation include instrument and setpoint uncertainties. An excore detector decalibration factor that accounts for the effect of changes in reactor vessel downcomer density on the indicated excore detector power level is conservatively set to [ ]. 7KH277

I , DQG237I , IXQFWLRQVDUHFRQVHUYDWLYHO\LJQRUHG Maximum reactor trip response times include delays for trip signal actuation and scram system holding coil release.

x Four RCPs are modeled to be in operation consistent with Mode 1. RCS flow is modeled as the Technical Specification minimum flow rate.

x Nominal pressurizer pressure is used in the S-RELAP5 calculation. The measurement uncertainty is applied in the DNBR calculations.

x Fuel pellet to cladding gap conductances are conservatively modeled to maximize rod surface heat flux. Time-in-cycle is considered when setting gap conductance.

x Average core fuel thermal-mechanical properties are modeled considering time-in-cycle and account for the effect of TCD.

x SG tube plugging is set to zero to minimize the resistance to heat transfer across the SG tubes during the transient, which in turn maximizes the cooldown of the RCS and positive reactivity insertion.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 44 x For HFP, one feedwater control valve is assumed to malfunction resulting in a step increase to 190% of nominal feedwater flow to one SG. An additional case with all feedwater control valves assumed to malfunction resulting in a step increase to 162.9% of nominal feedwater flow to each of the four SGs is also analyzed.

x For the HZP case, one feedwater control valve is assumed to malfunction resulting in an increase in flow to one SG from zero to 143% of the nominal full load value for one SG. An additional HZP case with all feedwater control valves assumed to malfunction resulting in a step increase from zero to 161.3% of nominal feedwater flow to each of the four SGs is analyzed.

x Initial MFW temperature is 390°F for the HFP case consistent with normal plant conditions. A HFP sensitivity case evaluates a maximum MFW temperature of 446°F. For the HZP case, initial MFW temperature is 100°F.

x Main feedwater isolation occurs on a high-high SG level ESFAS trip.

x The main feedwater isolation valve stroke time is 15 seconds after a 2 second signal delay.

x The turbine control valve (TCV) is modeled to remain at or above its initial value until a reactor trip signal is reached.

Since the systems designed to mitigate this event are redundant, there is no single active failure that will adversely affect the consequences of the event.

Normal reactor control systems and engineered safety systems are not required to function. The reactor protection system may function to trip the reactor due to either SG low-low level or an overpower condition. No single active failure will prevent operation of the reactor protection system.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 45 5.2.3 Results The sequence of events is presented in Table 5-3. Transient results presented in Figure 5-5 to Figure 5-9 IRUWKH+)3FDVHVVKRZWKHLQFUHDVHLQQXFOHDUSRZHUDQG7

associated with the increased thermal load on the reactor.

The asymmetric HFP case with EOC reactivity feedback coefficients, manual rod control, and the lowest feedwater temperature condition results in the lowest DNBR.

When the SG water level in the faulted loop reaches the high-high level setpoint (100%

of narrow range (NR) span), feedwater is isolated which prevents continuous addition of feedwater. Following feedwater isolation, the reactor will be automatically tripped when the low-low SG level trip setpoint (0% of NR span) is reached. After reactor trip, the plant approaches a stabilized condition. Standard plant shutdown procedures may then be followed to further cool down the plant.

Since the power level rises during the excessive feedwater flow event, the fuel temperatures also rise until after reactor trip occurs. The core heat flux lags slightly behind the neutron flux response. The peak neutron flux does not exceed 118% of its nominal value (i.e., the analytical high neutron flux trip setpoint, see Table 3-4). The peak fuel temperature thus remains well below the fuel melting temperature.

Transient results for the HZP power case in which all feedwater control valves are assumed to malfunction resulting in a step increase from zero to 161.3% of nominal feedwater flow to each of the four SGs are presented in Figure 5-10 to Figure 5-15.

The transient results show that DNB does not occur at any time during the excessive feedwater flow event; thus, the ability of the primary coolant to remove heat from the fuel rod is not reduced. MDNBR and peak LHGR results are shown in Table 5-4.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 46 Table 5-3 Increase in Feedwater Flow: Sequence of Events Case Event Time (sec)

All main and bypass feedwater control valves fail fully 0.0 open Minimum DNBR occurs 63.5 HFP, EOC, High-high steam generator water level setpoint reached 72.3 symmetric Feedwater isolation complete ~90 Low-low steam generator water level setpoint reached 169.8 Rods begin to drop 171.8 Turbine trip occurs due to reactor trip 171.8 One main feedwater control valve fails fully open 0.0 Minimum DNBR occurs 43.0 High-high steam generator water level setpoint reached 63.7 HFP, EOC, Feedwater isolation complete ~80 asymmetric Low-low steam generator water level setpoint reached 117.4 Rods begin to drop 119.4 Turbine trip occurs due to reactor trip 119.4 All bypass feedwater control valves fail fully open 0.0 Safety injection signal on low pressurizer pressure 25.1 HZP, EOC, High-high steam generator water level setpoint reached 53.4 symmetric Feedwater isolation complete ~70 Minimum DNBR occurs 86.8 All bypass feedwater control valves fail fully open 0.0 HZP, EOC, High-high steam generator water level setpoint reached 53.4 asymmetric Feedwater isolation complete ~70.5 Minimum DNBR occurs 70.8

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 47 Table 5-4 Increase in Feedwater Flow: Results Case Criterion Result Limit MDNBR 1.927 1.142 HFP, EOC, symmetric PLHGR (kW/ft) 17.6 [ ]

MDNBR 1.806 1.142 HFP, EOC, asymmetric Bounded by HFP PLHGR (kW/ft) symmetric case [ ]

MDNBR 3.299 1.173 HZP, EOC, symmetric PLHGR (kW/ft) 11.79 [ ]

MDNBR 3.680 1.173 HZP, EOC, asymmetric PLHGR (kW/ft) 13.11 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 48 Figure 5-5 Increase in Feedwater Flow (HFP): Core Power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 49 Figure 5-6 Increase in Feedwater Flow (HFP): Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 50 Figure 5-7 Increase in Feedwater Flow (HFP): Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 51 Figure 5-8 Increase in Feedwater Flow (HFP): Reactor Coolant Loop Hot Leg to Cold Leg Temperature Difference

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 52 Figure 5-9 Increase in Feedwater Flow (HFP): Vessel Average Temperature

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 53 Figure 5-10 Increase in Feedwater Flow (HZP, Multi-Loop): Core Power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 54 Figure 5-11 Increase in Feedwater Flow (HZP, Multi-Loop): Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 55 Figure 5-12 Increase in Feedwater Flow (HZP, Multi-Loop): Core Reactivity

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 56 Figure 5-13 Increase in Feedwater Flow (HZP, Multi-Loop): Vessel Average Temperature

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 57 Figure 5-14 Increase in Feedwater Flow (HZP, Multi-Loop): Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 58 Figure 5-15 Increase in Feedwater Flow (HZP, Multi-Loop): Pressurizer Liquid Volume

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 59 5.3 Excessive Increase in Secondary Steam Flow (FSAR SP 15.1.3) 5.3.1 Event Description This event is modeled to be initiated by a 10% step increase in steam demand. This increase in steam demand is within the limits which the reactor is designed to accommodate; therefore, a reactor trip is not expected to occur as a result of this event.

Steam flow increases greater than 10% are analyzed in Section 5.5 and Section 5.7.

The increased steam demand may be initiated by the operator, system demand, or equipment malfunction in the turbine bypass control or turbine speed control. An interlock in the steam dump control which blocks the opening of the steam dump valves unless a large turbine load decrease or a turbine trip has occurred. Therefore, the previously described equipment malfunction can result in either the rapid opening of the turbine control or bypass valves.

As a result of the increase in steam flow, the feedwater regulating valves open to increase the feedwater flow to match the new steam demand and maintain SG water level. In response to the increased steam flow, the secondary system pressure decreases, resulting in an increase in the primary-to-secondary heat transfer rate. The primary side SG outlet temperature decreases due to the enhanced heat removal.

Consequently, the primary system core average temperature decreases and the primary system fluid contracts, resulting in an outsurge of fluid from the pressurizer.

The pressurizer level and pressure decrease as fluid is drained from the pressurizer due to RCS volumetric contraction.

The effect of this cooldown on the core power level depends upon the MTC and the state of the rod control system. For Callaway, where automatic control rod withdrawal capability is disabled, a negative MTC will increase core power as the coolant temperature decreases with the reactor system reaching a new steady state condition at a power level which is consistent with the increased heat removal rate. A positive MTC, on the other hand, would decrease the core power level and not challenge the acceptance criteria.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 60 The RPS is assumed to be operable; however, as described above, a reactor trip is not typically generated for this event. Protection against an excessive load increase event is SURYLGHGE\WKHIROORZLQJUHDFWRUSURWHFWLRQVLJQDOV237277DQGSRZHUUDQJH

high neutron flux (high setting).

An excessive load increase is classified as an ANS Condition II. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

5.3.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. This transient is analyzed at HFP conditions with manual rod control since automatic rod control is disabled. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6).

Based on historical precedence, this event does not lead to a serious challenge to the acceptance criteria and a reactor trip is not typically generated.

The following assumptions are made to conservatively predict the consequences of this event:

x The event is initiated from HFP and the power measurement uncertainty is included in the initial S-RELAP5 power level. Higher power is conservative for DNBR calculations.

x Due to the cooldown caused by the increase in steam flow, the power increase at EOC conditions occurs more rapidly and more severely than at BOC. Since automatic rod control has been disabled at Callaway, only EOC is analyzed. The TS/COLR most-negative MTC limit of -47.9 pcm/°F and EOC DTC results in a conservatively bounding increase in reactor power as the moderator cools down and the fuel heats up. The EOC DTC is biased less negative in accordance with

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 61 the Reference 1 topical report.

x Initial RCS average temperature is the upper bound nominal value for the power level analyzed. Maximum temperature is conservative for DNBR calculations.

Temperature measurement uncertainties are applied in the DNBR calculations.

x Reactor trip functions available in the analysis are the power range neutron flux, OT7DQGOP77KHDQDO\WLFDO536WULSVHWSRLQWVXVHGLQWKH6-RELAP5 calculation include instrument and setpoint uncertainties. The OT7I , and OP7I , IXQFWLRQVDUHFRnservatively ignored. Maximum reactor trip response times include delays for trip signal actuation and scram system holding coil release.

x Four RCPs are modeled to be in operation consistent with Mode 1. RCS flow is modeled as the Technical Specification minimum flow rate. Minimizing RCS flow rate is conservative with respect to DNBR.

x Nominal pressurizer pressure is used in the S-RELAP5 calculation. The measurement uncertainty is applied in the DNBR calculations.

x HFP average core fuel pellet to cladding gap conductance is conservatively modeled to maximize rod surface heat flux and minimize the mitigating effect of Doppler reactivity on transient core power. Time-in-cycle is considered when setting gap conductance.

x Average core fuel thermal-mechanical properties are modeled considering time-in-cycle and account for the effect of TCD.

x SG tube plugging is assumed to be zero to minimize the resistance to heat transfer across the SG tubes during the transient, which in turn will maximize the cooldown of the RCS and maximize the positive reactivity insertion.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 62 x Normal reactor control systems and engineered safety systems are not required to function. The RPS is assumed to be operable; however, reactor trip is not encountered due to the uncertainty allowances assumed in the setpoints. No single active failure will prevent the RPS from performing its intended function.

5.3.3 Results The sequence of events is presented in Table 5-5. Figure 5-16 to Figure 5-19 illustrate the transient response with the reactor in the manual control mode. Reactor trip does not occur, and the plant rapidly reaches a stabilized condition at the higher power level corresponding to the increase in steam flow.

Table 5-6 presents the MDNBR and PLHGR results for this event. The results demonstrate that acceptance criteria are met.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 63 Table 5-5 Increase in Steam Flow: Sequence of Events Case Event Time (sec)

HFP, EOC, 10% step load increase 0.0 manual rod control Equilibrium conditions reached ~200 Table 5-6 Increase in Steam Flow: Results Criterion Result Limit MDNBR 1.733 1.142 PLHGR (kW/ft) 18.3 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 64 Figure 5-16 Increase in Steam Flow: Core Power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 65 Figure 5-17 Increase in Steam Flow: Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 66 Figure 5-18 Increase in Steam Flow: Pressurizer Liquid Level

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 67 Figure 5-19 Increase in Steam Flow: Reactor Coolant System Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 68 5.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve (FSAR SP 15.1.4)

The most severe core conditions resulting from an accidental depressurization of the main steam system are associated with an inadvertent opening of a single steam dump, relief, or safety valve.

The steam release from this event results in an initial increase in steam flow which decreases during the event as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a negative MTC, the cooldown results in an insertion of positive reactivity.

This event is like the steam system piping failure event documented in Section 5.5.

Since the more severe steam system piping failure event meets Condition II criteria, this event does not require reanalysis for the Framatome VQP.

5.5 Steam System Piping Failure (FSAR SP 15.1.5) 5.5.1 Event Description This section describes an analysis of the post-scram phase of a steam system piping failure or main steam line break (MSLB) event. Consequences of this event prior to reactor scram are presented in Section 5.7.

The steam release arising from a rupture of a main steam line results in an initial increase in steam flow that decreases during the event as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure.

In the presence of a negative MTC, the cooldown results in an insertion of positive reactivity. If the most reactive RCCA is assumed stuck in its fully withdrawn position after reactor trip, there is possibility that the core will become critical and return to power. A return to power following a steam line rupture is a potential problem mainly because of the high power peaking factors which exist, assuming the most reactive RCCA to be stuck in its fully withdrawn position. The peak core power excursion is ultimately mitigated by the boric acid solution delivered by the safety injection system.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 69 A major steam line rupture is classified as an ANS Condition IV event. Minor secondary system pipe breaks are classified as Condition III events. Effects of minor secondary system pipe breaks are bounded by the analysis presented in this section. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

A major rupture of a steam line, the most limiting cooldown transient, is analyzed at HFP, HZP with no decay heat, and HZP with assumptions to bound Mode 3 and no decay heat . Decay heat retards the cooldown, thereby reducing the return to power.

During startup or shutdown evolutions when safety injection on low pressurizer pressure or low steam line pressure is blocked and steam line isolation on low steam line pressure is blocked below the P-11 permissive, the high negative steam line pressure rate signal is enabled by P-11 to provide steam line isolation. It should be noted that steam line isolation can also be provided by a containment pressure (High-2) signal for breaks inside containment or by manual actions performed in accordance with established procedures.

The following functions provide the protection for a steam line rupture:

x Safety injection system actuation from any of the following:

i Two out of three low steam line pressure signals in any one loop i Two out of four low pressurizer pressure signals i Two out of three High-1 containment pressure signals x The overpower reactor trips (neutron flux and OPT) and the reactor trip occurring in conjunction with receipt of a safety injection signal x Redundant isolation of the main feedwater lines Sustained high feedwater flow causes additional cooldown. Therefore, in addition to the normal control action which closes the main feedwater valves following reactor trip, a safety injection signal rapidly closes all feedwater isolation valves. This signal also trips the main feedwater pumps, closes the

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 70 pump discharge valves, and closes the feedwater control valves. The success of this analysis is not predicated on the operation of the main feedwater pumps nor the pump discharge valves. The feedwater control valves are the primary success path equipment for event mitigation of secondary side pipe ruptures.

x Trip of the main steam isolation valves on:

i Safety injection system actuation derived from two out of three low steam line pressure signals in any one loop (above P-11 permissive) i Two out of three High-2 containment pressure signals i Two out of three high negative steam line pressure rate signals in any one loop (used only during cooldown and heatup operations, below P-11 with Tavg greater than 400°F)

Isolation valves are provided in each steam line. For breaks downstream of the isolation valves, closure of all valves would completely terminate the blowdown. For any break, in any location, no more than one SG would experience an uncontrolled blowdown, even if one of the isolation valves fails to close. These valves are assumed to fully close within 17 seconds upon receipt of a steam line isolation signal following a large break in a steam line. The 17 seconds includes a 2 second signal processing delay assumption. In the Mode 3 analysis, these valves are assumed to fully close within 62 seconds upon receipt of a steam line isolation signal following a large break in a steam line. The 62 seconds includes a 2 second signal processing delay assumption.

The effective throat area of the SG integral flow restrictors is 1.39 ft2, which is considerably less than the main steam pipe area; thus, the integral flow restrictors also serve to limit the maximum steam flow for a break at any location.

5.5.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. Core asymmetry is modeled by dividing the core into a

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 71 sector adjacent to the affected loop and a sector adjacent to the unaffected loops. The stuck-rod region is modeled as a separate region in the affected sector. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the COBRA-FLX code (Reference 8), which is used to calculate the MDNBR using the ORFEO-NMGRID CHF correlation (Reference 6), and the ARTEMIS code (References 9 and 10) is used to calculate power distribution information and kinetics parameters.

The following assumptions are made to conservatively predict the consequences of this event:

x The event is modeled to initiate from HFP and HZP initial power levels, as well as a Mode 3 analysis initiated from HZP level.

x Initial RCS average temperature is the maximum value without uncertainties corresponding to the initial power level.

x Nominal pressurizer pressure is used in the S-RELAP5 calculation.

x RCS flow is modeled as the Technical Specification minimum flow rate.

x Fuel pellet to cladding gap conductance is conservatively modeled to maximize rod surface heat flux. Time-in-cycle is considered when setting gap conductance.

x Average core fuel thermal-mechanical properties are modeled considering time-in-cycle and account for the effect of TCD.

x SG tube plugging is zero to minimize the resistance to heat transfer across the SG tubes during the transient.

x Since the SGs are provided with an integral flow restrictor with a 1.39 ft² throat area, any rupture with a break area greater than 1.39 ft² would have the same effect during the transient as the 1.39 ft² break. Therefore, the maximum effective break size analyzed is 1.39 ft².

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 72 x Moderator reactivity feedback is based on the TS/COLR most-negative MTC limit of -47.9 pcm/°F.

x MFW temperature is set to the lower bound for the power level analyzed to maximize the cooldown.

x Reactor trip is assumed to occur at time zero of the event. The delay associated with reactor trip is assumed to be zero for conservatism.

x In cases in which loss of offsite power is assumed, it occurs simultaneously with initiation of the safety injection signal at time zero of the event when the break occurs. Loss of offsite power causes the RCPs to coastdown.

x Maximum AFW flow rate and minimum AFW temperature are assumed to maximize the cooldown. AFW is assumed to begin at time zero of the event.

The delay associated with AFW initiation is assumed to be zero. The AFW temperature is [ ]. The AFW sweep out volume is assumed to be zero for conservatism. AFW is assumed to be terminated by operator action at 10 minutes.

x Doppler reactivity feedback model accounts for the effects of core power level and fuel temperature in accordance with the approved methodology.

x Upon initiation of the break with MFW control in automatic, the MFW flow rate increases to a maximum value to compensate for the increase in steam flow rate.

The MFW flow continues until isolation occurs. In the HFP cases, the MFW flow is modeled to increase to a maximum flow rate of 24,000 gpm per pump. In the HZP cases, the MFW flow increases to a maximum flow rate corresponding to the value associated with the HFP steady state condition.

x Minimum safety injection flow rate and minimum boron concentration are assumed to minimize the negative reactivity addition of boron. The Mode 3 analysis considers a case with minimum safety injection flow rate and a case with maximum safety injection flow rate.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 73 x Main steam isolation valve stroke time is 15 seconds after a 2 second signal delay for signal processing. In the Mode 3 cases, the main steam isolation valve stroke time is 60 seconds after a 2 second signal delay.

x Main feedwater isolation valve stroke time is 15 seconds after a 2 second signal delay. In the Mode 3 cases, the main feedwater isolation valve stroke time is 90 seconds after a 2 second signal delay.

x After generation of the safety injection signal, the time required for borated safety injection flow to reach the core with offsite power available is 27 seconds plus the time to sweep out the unborated water from the emergency core cooling system (ECCS) piping before reaching the core. In the cases without offsite power available, an additional 12 second delay is assumed to start the emergency diesel generators.

x In cases with minimum safety injection flow rate, the most limiting single failure is the failure of one safety injection pump which reduces the injection flow and delays the mitigation of the post-trip power excursion from the negative reactivity insertion of the boron in the safety injection flow. In cases with maximum safety injection flow rate, no single failure is identified which adversely affects the results of the analysis.

5.5.3 Results The sequence of events for the limiting PLHGR case is shown in Table 5-7. Figure 5-20 through Figure 5-31 show the RCS transient and core heat flux following a complete severance rupture of a main steam line pipe rupture at HZP initial conditions. Offsite power is assumed to be available so that full reactor coolant flow exists. The transient shown assumes an uncontrolled steam release from only one SG. Should the core be critical at HZP when the rupture occurs, the initiation of safety injection by low steam line pressure trips the reactor. Steam release from more than one SG is prevented by automatic trip of the isolation valves in the steam lines by low steam line pressure signals, high-high containment pressure signals, or by high negative steam line

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 74 pressure rate signals. Even with the failure of one valve, release is limited by main steam isolation valve closure for the other intact SGs while the one faulted SG blows down.

As shown in Figure 5-20, the core attains criticality with the RCCAs inserted (with the design shutdown assuming one stuck RCCA) shortly before boron solution enters the RCS. The continued addition of boron results in a peak core power significantly lower than the nominal HFP value.

The calculation assumes that the boric acid is mixed with, and diluted by, the water flowing in the RCS prior to entering the reactor core. The concentration after mixing depends upon the relative flow rates in the RCS and from the ECCS centrifugal charging pump (CCP). The variation of mass flow rate in the RCS due to water density changes is included in the calculation, as is the variation of flow rate from the CCP due to changes in the RCS pressure.

Note that zero decay heat is assumed in the analysis. However, following a steam line break only one SG blows down completely. Thus, the remaining SGs are still available for the dissipation of decay heat after the initial transient is over.

Table 5-8 presents the MDNBR and PLHGR results for this event. The results demonstrate that acceptance criteria are met.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 75 Table 5-7 Steam System Piping Failure: Sequence of Events Case Event Time (sec.)

Steam line ruptures 0.0 AFW initiated 0.0 Turbine stop valve closes 0.0 Scram rod full insertion 0.1 Steam line isolation signal 2.0 Mode 3, EOC, Low pressurizer pressure safety injection setpoint reached 15.0 maximum safety Criticality attained ~15.4 injection, Safety injection actuation 29.0 offsite power Boron reactivity reaches core ~47.4 available Peak core average power ~59.4 MSIV closure complete 62.0 Minimum DNBR occurs 78.0 MFW isolation complete 92.0 AFW isolated by operator action 600.0

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 76 Table 5-8 Steam System Piping Failure: Results Case Criterion Result Limit HFP, EOC, offsite power MDNBR 3.212 1.173 available PLHGR (kW/ft) 15.00 [ ]

HFP, EOC, loss of offsite MDNBR 3.609 1.204 power PLHGR (kW/ft) 10.50 [ ]

HZP, EOC, offsite power MDNBR 2.816 1.173 available PLHGR (kW/ft) 15.32 [ ]

HZP, EOC, loss of offsite MDNBR 2.322 1.204 power PLHGR (kW/ft) 14.99 [ ]

Mode 3, EOC, minimum MDNBR 2.674 1.173 safety injection, offsite power available PLHGR (kW/ft) 14.40 [ ]

Mode 3, EOC, maximum MDNBR 2.505 1.173 safety injection, offsite power available PLHGR (kW/ft) 16.10 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 77 Figure 5-20 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 78 Figure 5-21 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 79 Figure 5-22 Steam System Piping Failure (Mode 3, Mode 3, Maximum Safety Injection, Offsite Power Available): Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 80 Figure 5-23 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Pressurizer Liquid Volume

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 81 Figure 5-24 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Cold Leg Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 82 Figure 5-25 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Loop Average Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 83 Figure 5-26 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Reactivity

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 84 Figure 5-27 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Boron Concentration

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 85 Figure 5-28 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Main and Auxiliary Feedwater Flow Rates

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 86 Figure 5-29 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Steam Flow Rates

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 87 Figure 5-30 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Steam Generator Pressures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 88 Figure 5-31 Steam System Piping Failure (Mode 3, Maximum Safety Injection, Offsite Power Available): Core Flow Rate

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 89 5.6 Steam Line Break with Coincidental RCCA Withdrawal at Power (FSAR SP 15.1.5.5)

Not applicable to Callaway since the automatic rod control system has been disabled.

5.7 Steam System Piping Failure at Full Power (FSAR SP 15.1.5.6) 5.7.1 Event Description This section describes an analysis of the pre-scram phase of a MSLB event.

Consequences after reactor scram are presented in Section 5.5.

This event is defined as an increase in steam flow from one or more SGs caused by an inadvertent opening of a secondary side valve or a rupture of the main steam piping.

Increased steam flow from the SGs causes an increase in the heat extraction rate from the RCS, resulting in a reduction of primary coolant temperature and pressure. With negative moderator temperature and Doppler fuel temperature reactivity feedback, the core power will inherently seek a level bounded by the steam load demand, assuming no intervention of control, protection, or engineered safeguards systems. The rate at which core power approaches equilibrium with the secondary load is greatest when the reactivity feedback is the most negative, which corresponds to EOC. Thus, in the absence of any protective actions, a reactor power level dictated by steam flow rate could be established.

Breaks of various sizes are postulated to occur in the steam line upstream of the main steam isolation valve. The larger break sizes generate reactor trips on the low steam OLQHSUHVVXUH(6)$6IXQFWLRQZKLOHVPDOOHUEUHDNVWULSRQWKH237UHDFWRUWULS

function. The most limiting break size is the largest break case that results in a reactor WULSRQWKH237UHDFWRUWULSIXQFWLRQ

A range of MTC values is considered up to and including the most-negative TS/COLR MTC of -47.9 pcm/°F. In all cases, the most-negative MTC results in the lowest DNBR for each break size.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 90 HFP cases are analyzed to demonstrate that the applicable acceptance criteria are satisfied. Depending upon the break size, the event is a Condition III or IV event; however, more restrictive Condition II criteria are conservatively applied for all break sizes analyzed.

5.7.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6).

The following assumptions are made to conservatively predict the consequences of this event:

x The event is initiated from HFP and the power measurement uncertainty is included in the initial S-RELAP5 power level. Higher power is conservative for DNBR and FCM calculations.

x Initial RCS average temperature is the upper bound nominal value. Maximum temperature is conservative for DNBR calculations. Temperature measurement uncertainties are not applied in the S-RELAP5 analysis but are applied in the DNBR calculations.

x Four RCPs are modeled to be in operation consistent with Mode 1. RCS flow is modeled as the Technical Specification minimum flow rate. Minimizing RCS flow rate is conservative with respect to DNBR.

x Nominal pressurizer pressure is used in the S-RELAP5 calculation. The measurement uncertainty is applied in the DNBR calculations.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 91 x Fuel pellet to cladding gap conductance is conservatively modeled to maximize rod surface heat flux. Time-in-cycle is considered when setting gap conductance.

x Average core fuel thermal-mechanical properties are modeled considering time-in-cycle and account for the effect of TCD.

x SG tube plugging is zero to minimize the resistance to heat transfer across the SG tube during the transient, which in turn will maximize the cooldown of the RCS and maximize the positive reactivity insertion.

x A range of moderator feedback extending to the TS/COLR most-negative MTC limit of -47.9 pcm/°F is analyzed.

x Reactor trip occurs on an 237RPS trip or an ESFAS signal from low steam line pressure. The analytical trip setpoints used in the S-RELAP5 calculation include instrument and setpoint uncertainties, and harsh containment conditions to cover breaks inside containment. 7KH237I , IXQFWLRQLVFRQVHUYDWLYHO\

ignored. Maximum reactor trip response times include delays for trip signal actuation and scram system holding coil release.

x A range of break sizes is considered, ranging from breaks smaller than or equivalent to the inadvertent opening of a steam system valve to a hypothetical double-ended rupture of a main steam line. Since the SGs have integral flow restrictors with a 1.39 ft² throat area, i.e., no steam line upstream of the flow restrictors, any rupture with a break area greater than 1.39 ft² would have the same effect during the transient as the 1.39 ft² break. Even though the integral flow restrictor limits the effective break area to 1.39 ft², larger breaks, up to a double-ended guillotine break in the steam line are analyzed; however, there is little change in event results for break sizes greater than the size of the integral flow restrictors.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 92 x Loss of offsite power is assumed to occur at reactor trip for all cases. Coastdown of the RCPs is assumed to occur at reactor trip plus trip signal delay time, not including control rod drive mechanism holding coil delay.

x [

], and the initial core bypass flow is set to the design maximum.

x Since the systems designed to mitigate this event are redundant, there is no single active failure that will adversely affect the consequences of the event.

5.7.3 Results Table 5-9 gives the time sequence of events for the limiting break size of 0.55 ft2. The transient responses for this event are shown in Figure 5-32 through Figure 5-40. The RCS cooldown, in conjunction with negative moderator feedback, causes the reactivity to increase. In response to the increasing reactivity, the reactor power and heat flux (Figure 5-32 LQFUHDVHXQWLOWKH237signal occurs and the reactor trips.

Table 5-10 presents the MDNBR and PLHGR results for this event. The results demonstrate that acceptance criteria are met.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 93 Table 5-9 Steam System Piping Failure at Full Power: Sequence of Events Case Event Time (sec.)

Break / transient initiation 0 237WULSRFFXUV 21.3 HFP, 0.55 ft2 Rods begin to move 23.4 break Maximum core heat flux 23.4 Minimum DNBR reached 23.4 Table 5-10 Steam System Piping Failure at Full Power: Results Criterion Result Limit MDNBR 1.163 1.142 PLHGR (kW/ft) 21.5 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 94 Figure 5-32 Steam System Piping Failure (HFP, 0.55 ft2 Break): Core Power and Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 95 Figure 5-33 Steam System Piping Failure (HFP, 0.55 ft2 Break): Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 96 Figure 5-34 Steam System Piping Failure (HFP, 0.55 ft2 Break): Pressurizer Liquid Volume

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 97 Figure 5-35 Steam System Piping Failure (HFP, 0.55 ft2 Break): Core Inlet Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 98 Figure 5-36 Steam System Piping Failure (HFP, 0.55 ft2 Break): Reactor Coolant System Average Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 99 Figure 5-37 Steam System Piping Failure (HFP, 0.55 ft2 Break): Main Feedwater Flow Rates

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 100 Figure 5-38 Steam System Piping Failure (HFP, 0.55 ft2 Break): Steam Generator Flow Rates

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 101 Figure 5-39 Steam System Piping Failure (HFP, 0.55 ft2 Break): Steam Generator Pressures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 102 Figure 5-40 Steam System Piping Failure (HFP, 0.55 ft2 Break): Break Flow Rates

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 103 5.8 Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow (FSAR SP 15.2.1)

There are no steam pressure regulators in Callaway whose failure or malfunction could cause a steam flow transient.

5.9 Loss of External Electrical Load (FSAR SP 15.2.2)

A major load loss can result from loss of external electrical load due to some electrical system disturbance. Offsite alternating current (AC) power remains available to operate plant components. For a loss of external electrical load without subsequent turbine trip, no direct reactor trip signal would be generated, and the plant would be expected to trip from the RPS if a safety limit is approached. A loss of external load event results in a nuclear steam supply system transient that is bounded by the turbine trip event.

No aspect of Framatome fuel affects the relative severity of the loss of electrical load event compared to the turbine trip event. Therefore, this event does not require reanalysis for the Framatome VQP.

5.10 Turbine Trip (FSAR SP 15.2.3)

For a turbine trip event, the reactor trips directly (unless below approximately 50%

power) from a signal derived from the turbine stop emergency trip fluid pressure and turbine stop valves, though credit is not taken for a direct reactor trip from the turbine trip. Turbine stop valves close rapidly on loss of trip fluid pressure actuated by one of several possible turbine trip signals. Upon initiation of stop valve closure, steam flow to the turbine stops abruptly. A slightly more severe transient than the loss of electrical load event occurs for the turbine trip event due to a more rapid loss of steam flow caused by the more rapid valve closure of the turbine stop valves versus closure of the turbine control valves in the loss of electrical load event. A turbine trip is more limiting than a loss of external load, a loss of condenser vacuum, and other events which result in a turbine trip.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 104 The key parameters driving the extent of the system heatup for these events are unrelated to the fuel type. That is, for example, the closing speed of the turbine stop or control valves, main feedwater response, the level of SG tube plugging, RCP performance, and the RPS. The differences between these analyses and their relative severity are due to different assumptions in the availability and performance of plant systems because of the specifics of the event which are not affected by a change in the fuel type or DNB correlation. No aspect of Framatome fuel affects the relative severity of the loss of electrical load, loss of condenser vacuum, or other events which result in turbine trip, compared to the turbine trip event.

The MDNBR for this event is bounded by other more severe events and does not require reanalysis for the Framatome VQP.

There is no change in the plant configuration, operating parameters, or RPS or ESFAS functions associated with Framatome fuel. Global reactivity feedback is not a significant parameter for this event. No aspect of Framatome fuel affects the power mismatch between the primary and secondary systems. Therefore, Framatome fuel does not affect the severity of the challenge to overpressure criteria; reanalysis is not required for the Framatome VQP.

5.11 Inadvertent Closure of Main Steam Isolation Valves (FSAR SP 15.2.4)

Inadvertent closure of the main steam isolation valves results in a turbine trip with no credit taken for the turbine bypass system. Framatome fuel does not impact the relative severity of this event compared to the turbine trip event, and therefore reanalysis for the Framatome VQP is not required.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 105 5.12 Loss of Condenser Vacuum and Other Events Resulting in Turbine Trip (FSAR SP 15.2.5)

Loss of condenser vacuum is one of the events that can cause a turbine trip. A loss of condenser vacuum precludes the use of steam dump to the condenser; however, since steam dump is assumed to be unavailable in the turbine trip analysis, no additional adverse effects result if the turbine trip is caused by loss of condenser vacuum.

Therefore, the conclusions contained in Section 5.10 apply to the loss of the condenser vacuum. Framatome fuel does not impact the relative severity of this event compared to the turbine trip event, and reanalysis for the Framatome VQP is not required.

5.13 Loss of Nonemergency AC Power to the Plant Auxiliaries (FSAR SP 15.2.6)

A complete loss of nonemergency AC power may result in the loss of all power to the plant auxiliaries, i.e., the RCPs, condensate pumps, etc. The loss of power may be caused by a complete loss of the offsite grid accompanied by a turbine-generator trip at the plant or by a loss of the onsite AC distribution system.

The consequences of this event primarily depend on initial operating conditions, plant-related systems and capacities, and decay heat. Framatome fuel does not significantly impact any of these controlling parameters; thus, this event does not require reanalysis to support the Framatome VQP.

5.14 Loss of Normal Feedwater Flow (FSAR SP 15.2.7)

A loss of normal feedwater flow, caused by pump failures, valve malfunctions, or loss of offsite AC power or feedwater control system failure, results in a reduction in the capability of the secondary system to remove the heat generated in the reactor core. If an alternative supply of feedwater is not supplied to the plant, core residual heat following reactor trip would heat the primary system water to the point where water relief from the pressurizer would occur, resulting in a substantial loss of water from the RCS.

Since the plant is tripped well before the SG heat transfer capability is reduced, the primary system variables never approach a DNB condition.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 106 The consequences of this event primarily depend on initial operating conditions, plant-related systems and capacities, and decay heat. Framatome fuel does not significantly impact any of these controlling parameters; thus, this event does not require reanalysis for the Framatome VQP.

5.15 Feedwater System Pipe Break (FSAR SP 15.2.8)

A major feedwater line rupture is defined as a break in a feedwater line large enough to prevent the addition of sufficient feedwater to the SGs to maintain shell-side fluid inventory in the SGs. A break located upstream of the feedwater line check valve would affect the nuclear steam supply system only as a loss of feedwater which is covered by the evaluations in Section 5.13 and Section 5.14. If the break is postulated in a feedwater line between the check valve and the SG, fluid from the SG may also be discharged through the break. Depending upon the size of the break and the plant operating conditions at the time of the break, the event could cause an RCS cooldown which is evaluated in Section 5.5 and Section 5.7. A break between the check valve and SG could preclude the subsequent addition of auxiliary feedwater to the affected SG, which could cause RCS heatup and overpressurization of the RCS or loss of hot leg subcooling due to failure to remove decay heat.

The consequences of this event primarily depend on initial operating conditions, plant-related systems and capacities, and decay heat. Framatome fuel does not significantly impact any of these controlling parameters; thus, this event does not require reanalysis for the Framatome VQP.

5.16 Partial Loss of Forced Reactor Coolant Flow (FSAR SP 15.3.1)

A partial loss of forced reactor coolant flow transient can result from a mechanical or electrical failure in an RCP or from a fault in the power supply to the pump or pumps supplied by an RCP bus. If the reactor is at power at the time of the event, the immediate effect of the partial loss of forced reactor coolant flow is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor does not trip promptly.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 107 A sensitivity study demonstrates the analysis in Section 5.17 bounds a partial loss of flow event since power is lost to four RCPs and Condition II criteria are met for the more severe complete loss of forced reactor coolant flow event.

5.17 Complete Loss of Forced Reactor Coolant Flow (FSAR SP 15.3.2) 5.17.1 Event Description A complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to all RCPs. If the reactor is at power at the time of the event, the immediate effect of loss of coolant flow is a rapid increase in the coolant temperature.

This increase could result in a DNB with subsequent fuel damage if the reactor is not tripped promptly.

This event is classified as an ANS Condition III event. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

The following signals provide the necessary protection against a complete loss of flow transient:

x RCP power supply undervoltage x RCP underfrequency x Low reactor coolant loop flow The reactor trip on RCP undervoltage is provided to protect against conditions which can cause a loss of voltage to all RCPs, i.e., loss of AC power to plant auxiliaries. This function is blocked below 10% power (Permissive P-7).

The reactor trip on RCP underfrequency is provided to trip the reactor for an underfrequency condition resulting from frequency disturbances on the power grid.

The reactor trip on low primary coolant loop flow is provided to protect against loss of flow conditions which affect only one reactor coolant loop. This function is generated by two out of three low flow signals per reactor coolant loop.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 108 5.17.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6).

The complete loss of flow transient is analyzed for a loss of power to all four RCPs with four loops in operation. A single case is analyzed at BOC HFP initial conditions, maximum vessel average temperature, and minimum Technical Specification RCS flow rate.

The following assumptions are made to conservatively predict the consequences of this event:

x This event is initiated from HFP and the power measurement uncertainty is included in the initial core power level modeled in the S-RELAP5 analysis.

Maximum power is conservative for DNBR calculations.

x All four RCPs are assumed to be in operation and the initial flow is set to the Technical Specification minimum flow. Minimizing RCS flow rate is conservative for DNBR calculations.

x Upper bound reactor vessel average temperature is used which is conservative for DNBR calculations. Temperature measurement uncertainty is applied in the DNBR calculations.

x Nominal initial pressurizer pressure is used in S-RELAP5 calculation. The measurement uncertainty is applied in the DNBR calculations.

x Core bypass flow rate is set to the maximum value.

x Reactor trip is initiated by RCP power supply undervoltage. This reactor trip is modeled to occur when the event begins. If the loss of all RCPs is due to a loss

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 109 of offsite power, scram reactivity insertion of the rods does not occur directly on the loss of power, but after actuation of the RPS and the appropriate reactor trip delay.

x A conservative scram reactivity table is used. Delayed scram reactivity insertion maintains power at a higher level while the core flow coastdown proceeds, which is conservative for DNBR calculations. Minimum HFP scram worth is used which assumes the most reactive rod is stuck out of core.

x BOC reactivity feedback parameters are used. Since this event involves an increase in core coolant temperatures, the event is modeled to occur at BOC with a most-positive TS/COLR MTC limit of 0.0 pcm/°F consistent with the HFP initial power level. A conservative bias is applied to the DTC to comply with the Reference 1 methodology.

x Fuel pellet to cladding gap conductance is conservatively modeled to maximize rod surface heat flux. Time-in-cycle is considered when setting gap conductance.

x Average core fuel thermal-mechanical properties are modeled considering time-in-cycle and account for the effect of TCD.

x Maximum SG tube plugging is assumed to increase the RCS flow resistance and hence, to increase the flow coastdown rate.

x Isolation of main feedwater has essentially no effect on the event; however, isolation of main feedwater at the beginning of the transient is simulated to bound the loss of AC power event (Section 5.13).

x Operation of the pressurizer sprays and PORVs are conservative for the DNBR calculations to minimize any RCS pressure increase. The PORVs are powered from the station batteries, and therefore the PORVs can operate following a loss of offsite power. The PORV lift setpoint is conservatively modeled. The driving head of the sprays from the RCPs is lost following loss of the RCPs. For conservatism, operation of the pressurizer sprays and PORVs is modeled in the

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 110 S-RELAP5 analysis.

x Since the systems designed to mitigate this event, specifically the RPS, are redundant, there is no single active failure that will adversely affect the consequences of the event.

5.17.3 Results The sequence of events is presented in Table 5-11. Figure 5-41 through Figure 5-44 show the transient response for the loss of power to all RCPs with four loops in operation. The reactor is assumed to be tripped on an undervoltage signal. Table 5-12 presents the MDNBR and peak LHGR for this event. The results demonstrate that the more restrictive Condition II acceptance criteria are met for this Condition III event.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 111 Table 5-11 Complete Loss of Forced Reactor Coolant Flow: Sequence of Events Case Event Time (sec.)

All operating pumps lose power and begin coasting down 0.0 RCP undervoltage trip setpoint reached 0.0 HFP, BOC Control rod insertion begins 1.5 Minimum DNBR occurs 2.6 Table 5-12 Complete Loss of Forced Reactor Coolant Flow: Results Criterion Result Limit MDNBR 1.592 1.142 PLHGR (kW/ft) 18.1 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 112 Figure 5-41 Complete Loss of Forced Reactor Coolant Flow: Reactor Coolant System Flow Rate

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 113 Figure 5-42 Complete Loss of Forced Reactor Coolant Flow: Core Power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 114 Figure 5-43 Complete Loss of Forced Reactor Coolant Flow: Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 115 Figure 5-44 Complete Loss of Forced Reactor Coolant Flow: Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 116 5.18 Reactor Coolant Pump Shaft Seizure (Locked Rotor) (FSAR SP 15.3.3) 5.18.1 Event Description The event postulated is an instantaneous seizure of an RCP rotor. Flow through the affected reactor coolant loop is rapidly reduced, leading to an initiation of a reactor trip on a low flow signal.

Following initiation of the reactor trip, 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; first, because the reduced flow results in a decreased tube side heat transfer coefficient, and then, because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon reactor trip). The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the SGs, causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steam volume, actuates the automatic spray system, and opens the PORVs, in that sequence.

A free spinning RCP impeller is assumed in the analysis upon loop flow reversal to address the RCP shaft break event (Section 5.19). This allows higher reverse flow through the loop with the locked RCP rotor, which is conservative for DNB calculations since the flow to the core is reduced.

This event is classified as an ANS Condition IV event. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

5.18.2 Method of Analysis Detailed analyses are performed with an approved non-LOCA methodology (Reference

1) with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8),

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 117 which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6).

Two cases are analyzed at BOC HFP initial conditions, maximum reactor vessel average temperature, and minimum Technical Specification RCS flow rate: one case assumes a loss of offsite power (LOOP) and the other assumes offsite power remains available.

The following assumptions are made to conservatively predict the consequences of this event:

x A free spinning RCP impeller is modeled in the faulted RCP to address the RCP shaft break event (Section 5.19). This assumption allows higher reverse flow in the affected loop, which is conservative for DNB calculations.

x This event is initiated from HFP and the power measurement uncertainty is included in the initial core power level modeled in the S-RELAP5 analysis.

Maximum power is conservative for DNBR calculations.

x All four RCPs are assumed to be in operation and the initial flow is set to the Technical Specification minimum flow. Minimizing RCS flow rate is conservative for DNBR calculations and does not affect the time of reactor trip on low RCS flow.

x Upper bound reactor vessel average temperature is used which is conservative for DNBR calculations. Temperature measurement uncertainty is applied in the DNBR calculations.

x Nominal initial pressurizer pressure is used in S-RELAP5 calculation. The measurement uncertainty is applied in the DNBR calculations.

x Reactor trip is initiated by the low reactor coolant loop flow trip with a conservatively biased setpoint and response time.

x A conservative scram reactivity table is used. Delayed scram reactivity insertion maintains power at a higher level while the core flow decreases, which is

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 118 conservative for DNBR calculations. Minimum HFP scram worth is used which assumes the most reactive rod is stuck out of core.

x BOC reactivity feedback parameters are used. Since this event involves an increase in core coolant temperatures, the event is modeled to occur at BOC with a most-positive TS/COLR MTC limit of 0.0 pcm/°F consistent with the HFP initial power level. A conservative bias is applied to the DTC to comply with the Reference 1 methodology.

x Fuel pellet to cladding gap conductance is conservatively modeled to maximize rod surface heat flux. Time-in-cycle is considered when setting gap conductance.

x Average core fuel thermal-mechanical properties are modeled considering time-in-cycle and account for the effect of TCD.

x Maximum SG tube plugging is assumed to increase the RCS flow resistance and hence, to increase the flow coastdown rate.

x A case is analyzed assuming LOOP occurs on reactor trip without delay. A sensitivity case is also analyzed assuming offsite power remains available.

x Isolation of main feedwater has essentially no effect on this event; however, complete isolation of main feedwater with no coastdown at the time of LOOP is simulated. Isolation of main feedwater is not modeled in the sensitivity study with offsite power available.

x Operation of the pressurizer sprays and PORVs are conservative for the DNBR calculations to minimize any RCS pressure increase. Therefore, operation of the pressurizer sprays and PORVs is modeled in the S-RELAP5 analysis. The PORV lift setpoint is the nominal value minus tolerance. The PORVs are powered from the station batteries, and therefore the PORVs can operate following LOOP. The driving head of the sprays from the RCPs is lost following LOOP as the pumps coastdown. In the offsite power available case, sprays are available for the duration of the event.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 119 x Since the systems designed to mitigate this event, specifically the RPS, are redundant, there is no single active failure that will adversely affect the consequences of the event.

5.18.3 Results The sequence of events for the limiting case is shown in Table 5-13. Figure 5-45 through Figure 5-47 show the system response to an instantaneous seizure of an RCP rotor. Flow through the faulted RCS loop is rapidly reduced, leading to an initiation of a reactor trip on a low flow signal. The remaining unaffected RCPs continue to coast down, and natural circulation flow is eventually established. With the reactor tripped, a stable plant condition is eventually attained. Normal plant shutdown may then proceed.

Table 5-14 presents the MDNBR and peak LHGR for this event with a loss of offsite power, which was the limiting case. The results demonstrate that acceptance criteria are met.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 120 Table 5-13 Reactor Coolant Pump Shaft Seizure: Sequence of Events Case Event Time (sec.)

Rotor on one RCP locks 0.0 HFP, BOC, Low flow trip setpoint reached 0.1 loss of offsite power Control rod insertion begins 1.1 Minimum DNBR occurs 2.3 Table 5-14 Reactor Coolant Pump Shaft Seizure: Results Criterion Result Limit MDNBR 1.204 1.142 PLHGR (kW/ft) 18.1 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 121 Figure 5-45 Reactor Coolant Pump Shaft Seizure: Reactor Coolant System Flow Rate

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 122 Figure 5-46 Reactor Coolant Pump Shaft Seizure: Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 123 Figure 5-47 Reactor Coolant Pump Shaft Seizure: Core Power and Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 124 5.19 Reactor Coolant Pump Shaft Break (FSAR SP 15.3.4)

The event is postulated as an instantaneous failure of an RCP shaft. Flow through the affected reactor coolant loop is rapidly reduced, though the initial rate of reduction of coolant flow is greater for the RCP rotor seizure (locked rotor) event. A free spinning pump impeller is assumed in the faulted RCP for the locked rotor analysis (Section 5.18) to address higher reverse flows that are characteristic of this event. Thus, the consequences of an RCP shaft break are bounded by the analytical assumptions made for the locked rotor event analysis and no explicit analysis is required for the Framatome VQP.

5.20 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition (FSAR SP 15.4.1) 5.20.1 Event Description An RCCA withdrawal event is initiated by an addition of reactivity to the reactor core caused by the uncontrolled withdrawal of a sequential pair of RCCA banks resulting in a core power excursion. Such a transient could be caused by a malfunction of the rod control system or operator error. This event could occur with the reactor subcritical, at HZP, or at-power. This section addresses the event initiated from low power or startup conditions. Section 5.21 documents the analysis of this event from at-power conditions.

Although the reactor is typically brought to power from a subcritical condition by means of RCCA bank withdrawal, initial startup procedures with a clean core call for boron dilution. The maximum rate of reactivity increase in the case of boron dilution is less than that assumed in this analysis for an RCCA bank withdrawal.

The maximum reactivity insertion rate analyzed in the detailed plant analysis is that occurring with the simultaneous withdrawal of the combination of two sequential control banks having the maximum combined worth, moving together at maximum speed.

Although the analysis results reflect the inadvertent withdrawal of overlapping control banks from a low power condition, this event could be initiated by the inadvertent withdrawal of a shutdown bank from a subcritical condition (Mode 3). As such,

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 125 Technical Specification operability requirements are imposed on the power range high neutron flux (low setting) reactor trip function whenever all RCS cold leg temperatures are )

This event is classified as an ANS Condition II event. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

The neutron flux response to a continuous reactivity insertion is characterized by a very fast rise mitigated by negative Doppler reactivity feedback. This self-limitation of the power excursion is of primary importance since it limits the power to a tolerable level during the delay time for protective action. Should a continuous RCCA withdrawal event occur, the transient will be terminated by the following RPS signals:

x Source range high neutron flux reactor trip x Intermediate range high neutron flux reactor trip x Power range high neutron flux reactor trip (low setting) x Power range high neutron flux reactor trip (high setting) x High neutron flux rate reactor trip 5.20.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6). A hot spot model in the S-RELAP5 code is employed to calculate the peak fuel centerline temperature and evaluate margin to fuel melt.

The following assumptions are made to conservatively predict the consequences of this event:

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 126 x The event is initiated from 10-9 RTP which is below the power level expected for any shutdown condition. Starting the transient from this power level, together with a maximum reactivity insertion rate, results in the highest core power.

x Since the peak power excursion is strongly dependent on the negative Doppler reactivity feedback, BOC conditions conservatively minimizes the absolute magnitude of the Doppler feedback and maximizes core power. BOC fuel temperature dependent Doppler reactivity is biased less negative in accordance with the Reference 1 topical report.

x Moderator reactivity feedback is not important for this event since the time for heat to transfer from the fuel to the moderator is much longer than the time for Doppler reactivity to mitigate the core power excursion and reactor scram to occur. Nonetheless, the most-positive TS/COLR MTC limit of 5.0 pcm/°F for HZP conditions is conservatively modeled.

x The highest initial RCS temperature for Modes 2 and 3 is modeled. The S-RELAP5 calculation uses a nominal temperature without adjustment for measurement uncertainty. The measurement uncertainty is applied in the DNBR calculations.

x Two RCPs are modeled to be in operation consistent with Mode 3. Flow for operation with two RCPs is based on tripping two RCPs from a steady-state condition with four RCPs operating at the Technical Specification minimum flow rate. Backflow though the idle RCS loops is included. Total pump heat is representative of two operating RCPs.

x Nominal pressurizer pressure is used in the S-RELAP5 calculation. The measurement uncertainty is applied in the DNBR calculations.

x Reactor trip is initiated by a power range high neutron flux (low setting) signal.

Instrument and setpoint uncertainties are included. Maximum reactor trip response times include delays for trip signal actuation and scram system holding coil release.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 127 x A bounding reactivity insertion rate is modeled by using a maximum rod withdrawal speed combined with a maximum differential bank worth. Reactivity is inserted assuming a constant insertion rate without regard to total bank worth.

Using a constant maximum insertion rate is more conservative than adding reactivity according to the actual control bank differential worth curves since the potential for power overshoot is increased and the total inserted bank worth to the time of reactor trip is maximized.

x Maximum FQ peaking factors are used to maximize the hot spot peak fuel centerline temperature. Uncertainty factors are applied to the FQ that is used in the S-RELAP5 model. The peak fuel centerline temperature and fuel melt limit bounds UO2 and UO2-Gd2O3 fuel pellets.

x Fuel pellet to cladding gap conductances are conservatively modeled. A bounding maximum gap conductance for the average core at BOC is modeled to minimize the mitigating effect of Doppler reactivity on transient core power and maximize rod surface heat flux. A hot spot model, used to calculate peak fuel centerline temperatures, [

]

x BOC average core and hot spot fuel thermal-mechanical properties account for the effect of TCD and bound all GAIA fuel types (i.e., UO2 only and UO2-Gd2O3).

x This analysis assumes 5% steam generator tube plugging (SGTP). This event is dominated by the very rapid increase in power that occurs from HZP, and the event is terminated by Doppler feedback and reactor scram before there is any significant response from the secondary system. Therefore, this event is not sensitive to the level of SGTP. However, using higher SGTP level is in the conservative direction.

x The uncertainty associated with the opening and closing of the pressurizer PORV is considered and the PORV flow rate is conservatively biased.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 128 x Since the systems designed to mitigate this event are redundant, there is no single active failure that will adversely affect the consequences of the event.

5.20.3 Results The sequence of events is shown in Table 5-15. Figure 5-48 through Figure 5-50 show the transient behavior for the uncontrolled RCCA bank withdrawal event, with the event terminated by reactor trip at 35% of nominal power.

The reactivity insertion rate used is equal to that calculated for the two highest worth sequential control banks, both assumed to be in their highest incremental worth region.

Figure 5-48 shows the core average power response. The energy release and the fuel temperatures are relatively small. The thermal flux response, of interest for DNB considerations, is shown in Figure 5-49. The beneficial effect of the inherent thermal lag in the fuel is shown by a peak heat flux significantly less than the HFP nominal value.

Figure 5-50 shows the response of the hot spot fuel centerline temperature. The hot spot fuel centerline temperature increases to a value lower than the fuel melt limit. The MDNBR remains above the limit value. With the reactor tripped, the plant returns to a stable condition. The plant may subsequently be cooled down further following normal plant shutdown procedures.

Table 5-16 presents the MDNBR and peak fuel centerline temperature results for this event. The results demonstrate that acceptance criteria are met.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 129 Table 5-15 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Sequence of Events Case Event Time (sec.)

Initiation of uncontrolled rod withdrawal from 10-9 power 0.0 Power range high neutron flux low setpoint reached 7.82 Peak nuclear power occurs 7.92 HZP, BOC, two Rods begin to fall into core 8.32 operating RCPs Minimum DNBR occurs 9.5 Peak heat flux occurs 10.0 Peak hot spot fuel centerline temperature occurs 11.7 Table 5-16 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Results Criterion Result Limit MDNBR 1.371 1.142 Peak fuel centerline temperature (°F) 2862.4 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 130 Figure 5-48 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Core Power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 131 Figure 5-49 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 132 Figure 5-50 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low-Power Startup Condition: Fuel Centerline Temperature

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 133 5.21 Uncontrolled RCCA Bank Withdrawal at Power (FSAR SP 15.4.2) 5.21.1 Event Description Uncontrolled RCCA bank withdrawal at-power results in an increase in the core heat flux. Since the heat extraction from the SGs lags core power generation until the SG pressure reaches the relief or safety valve setpoint, there is a net increase in the reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise could eventually result in DNB.

Therefore, to avert damage to the fuel clad, the RPS is designed to terminate any such transient before the MDNBR falls below the safety analysis limit value.

This event is classified as an ANS Condition II event. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

The primary automatic features of the RPS which prevent core damage following the postulated event include the following:

x Power range neutron flux (high setting) x OT7 x OP7 The following RPS trips, relevant to this event, are not credited:

x High pressurizer pressure x High pressurizer water level x Power range neutron flux, high positive rate In addition to the above listed reactor trips, there are control rod withdrawal blocks that discontinue manual control rod withdrawal.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 134 5.21.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6).

The following assumptions are made to conservatively predict the consequences of this event:

x This event is initiated from three different initial core power levels: HFP, 60%

RTP and 10% RTP. Power measurement uncertainty is included in the initial core power levels modeled in S-RELAP5 analyses.

x For BOC cases, the most-positive power-dependent TS/COLR MTC limits of 5.0 SFP)IRUFRUHSRZHU573DQGSFP)IRUFRUHSRZHU!573

are modeled. A BOC DTC is used which represents a conservatively small (in absolute magnitude) value based on time-in-cycle. The BOC DTC is biased less negative in accordance with the Reference 1 topical report.

x For EOC cases, the most-negative TS/COLR MTC limit of -47.9 pcm/°F is modeled. An EOC DTC is used which represents a large (in absolute magnitude) value based on time-in-cycle. The EOC DTC is biased more negative in accordance with the Reference 1 topical report.

x Upper bound nominal power-dependent initial RCS average temperatures for Mode 1 are modeled to predict conservative input to the subsequent DNBR calculations. The S-RELAP5 calculation uses nominal temperatures without adjustment for measurement uncertainty. The measurement uncertainty is applied in the DNBR calculations.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 135 x Four RCPs are modeled to be in operation consistent with Mode 1. RCS flow is modeled as the Technical Specification minimum flow rate. Minimizing RCS flow rate is conservative with respect to DNBR.

x Nominal pressurizer pressure is used in the S-RELAP5 calculation. The measurement uncertainty is applied in the DNBR calculations.

x Reactor trip is initiated by the power range high neutron flux (high setting) trip, OTT trip or OPT trip. The analytical RPS trip setpoints used in the S-RELAP5 calculation include instrument and setpoint uncertainties. Maximum trip response times include delays for trip signal actuation and scram system holding coil release. The OTT I , and OP7I , functions are conservatively ignored.

x Bounding ranges of reactivity insertion rates are modeled to account for maximum rod withdrawal speed combined with maximum differential bank worth.

Since this event bounds the boron dilution event (Section 5.25) in Mode 1, reactivity insertion rates associated with a boron dilution event are defined to ensure they are within the analyzed range.

x Fuel pellet to cladding gap conductances are conservatively modeled to maximize rod surface heat flux. Time-in-cycle is considered when setting gap conductances.

x Average core fuel thermal-mechanical properties are modeled considering time-in-cycle and account for the effect of TCD.

x Since the systems designed to mitigate this event are redundant, there is no single active failure that will adversely affect the consequences of the event.

5.21.3 Results The sequence of events and figures for the limiting cases initiated from HFP are representative of the transient response for cases initiated at 10% RTP and 60% RTP.

The sequence of events for the limiting cases initiated from HFP is shown in Table 5-17.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 136 Figure 5-51 through Figure 5-54 show the transient response for a most limiting RCCA withdrawal event starting from HFP BOC conditions. The transient response for a most limiting RCCA withdrawal from HFP EOC conditions is shown in Figure 5-55 to Figure 5-58.

Figure 5-59 shows the DNBR as a function of reactivity insertion rate from initial HFP operation for minimum (BOC) and maximum (EOC) reactivity feedback. Two RPS trip functions provide protection over the whole range of reactivity insertion rates. These are the high neutron flux and OT7trips. The MDNBR is never less than the safety analysis limit values.

Figure 5-60 and Figure 5-61 show the minimum DNBR as a function of reactivity insertion rate for RCCA withdrawal events starting at 60% and 10% RTP, respectively, for minimum (BOC) and maximum (EOC) reactivity feedback. The results are like those for the HFP case, except as the initial power is decreased and the range over which the OT7WULSLVHIIHFWLYHLVLQFUHDVHG In neither case does the MDNBR fall below the safety analysis limit values.

Limiting MDNBR and PLHGR results for each case are given in Table 5-18. The overall limiting case for both MDNBR and PLHGR initiates from 10% RTP with BOC kinetics.

The results for all cases demonstrate that acceptance criteria are met.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 137 Table 5-17 Uncontrolled RCCA Bank Withdrawal at Power: Sequence of Events Case Event Time (sec.)

Initiation of uncontrolled RCCA withdrawal at the most 0

limiting reactivity insertion rate (1.9 pcm/sec)

HFP, BOC OTT reactor trip setpoint reached 49.8 Rods begin to fall into core 51.8 Minimum DNBR occurs 52.1 Initiation of uncontrolled RCCA withdrawal at the most 0

limiting reactivity insertion rate (23.7 pcm/sec)

HFP, EOC 277reactor trip setpoint reached 33.7 Rods begin to fall into core 35.7 Minimum DNBR occurs 35.7

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 138 Table 5-18 Uncontrolled RCCA Bank Withdrawal at Power: Results Limiting Case Criterion Result (Note 1) Limit MDNBR 1.301 1.142 100% RTP BOC PLHGR (kW/ft) Note 2 [ ]

MDNBR 1.316 1.142 100% RTP EOC PLHGR (kW/ft) Note 2 [ ]

MDNBR 1.245 1.142 60% RTP BOC PLHGR (kW/ft) Note 2 [ ]

MDNBR 1.348 1.142 60% RTP EOC PLHGR (kW/ft) Note 2 [ ]

MDNBR 1.228 1.142 10% RTP BOC PLHGR (kW/ft) 21.8 [ ]

MDNBR 1.204 1.142 10% RTP EOC PLHGR (kW/ft) Note 2 [ ]

Notes:

1. The MDNBR for 10% BOC was generated using statistical methods, all other MDNBRs are generated using deterministic methods.

2. Bounded by the value for 10% BOC since the overall power for the 10% BOC case is higher than the other cases.

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 139 Figure 5-51 Uncontrolled RCCA Bank Withdrawal (HFP, BOC, 1.9 pcm/sec): Core Power and Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 140 Figure 5-52 Uncontrolled RCCA Bank Withdrawal (HFP, BOC, 1.9 pcm/sec):

Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 141 Figure 5-53 Uncontrolled RCCA Bank Withdrawal (HFP, BOC, 1.9 pcm/sec):

Pressurizer Liquid Level

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 142 Figure 5-54 Uncontrolled RCCA Bank Withdrawal (HFP, BOC, 1.9 pcm/sec):

Reactor Coolant System Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 143 Figure 5-55 Uncontrolled RCCA Bank Withdrawal (HFP, EOC, 23.7 pcm/sec):

Core Power and Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 144 Figure 5-56 Uncontrolled RCCA Bank Withdrawal (HFP, EOC, 23.7 pcm/sec):

Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 145 Figure 5-57 Uncontrolled RCCA Bank Withdrawal (HFP, EOC, 23.7 pcm/sec):

Pressurizer Liquid Level

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 146 Figure 5-58 Uncontrolled RCCA Bank Withdrawal (HFP, EOC, 23.7 pcm/sec):

Reactor Coolant System Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 147 Figure 5-59 Uncontrolled RCCA Bank Withdrawal (HFP): DNBR

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 148 Figure 5-60 Uncontrolled RCCA Bank Withdrawal (60% RTP): DNBR

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 149 Figure 5-61 Uncontrolled RCCA Bank Withdrawal (10% RTP): DNBR

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 150 5.22 RCCA Misoperation (System Malfunction or Operator Error) (FSAR SP 15.4.3) 5.22.1 Event Description RCCA misoperation events include:

x One or more dropped RCCAs within the same group x A dropped RCCA bank x Statically misaligned RCCA x Withdrawal of a single RCCA The dropped RCCA, dropped RCCA bank, and statically misaligned RCCA events are classified as ANS Condition II events. However, the single RCCA withdrawal event is classified as an ANS Condition III event. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

A dropped RCCA or RCCA bank is detected by:

x Sudden drop in the core power level as seen by the nuclear instrumentation system.

x Asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples.

x Rod at bottom signal.

x Rod deviation alarm (control rods only).

x Rod position indication.

Misaligned RCCAs are detected by:

x Asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples.

x Rod deviation alarm (control rods only).

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 151 x Rod position indication.

Withdrawal of a single RCCA results in both positive reactivity insertion tending to increase core power and an increase in local power density in the core area associated with the RCCA. Automatic protection for this event is provided by the OT7UHDFWRUWULS

Due to the increase in local power density, it is not possible in all cases to provide assurance that the core safety limits will not be violated.

Plant systems and equipment which are available to mitigate the effects of the various RCCA misoperations are discussed in Section 3.4 and listed in Table 3-4.

No single active failure in any of these systems or equipment will adversely affect the consequences of the event.

5.22.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9.

Dropped RCCA / RCCA Bank The S-RELAP5 code is used to calculate the system response. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6). Conservative radial peaking augmentation factors are used to bound core power redistribution resulting from this event.

The following assumptions are made to conservatively predict the consequences of this event:

x This event is initiated from HFP. Power measurement uncertainty is included in the initial core power level modeled in the S-RELAP5 calculations.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 152 x Most-negative TS/COLR MTC limit of -47.9 pcm/°F is modeled to maximize core power levels after the RCCA drop. An EOC DTC is used which represents a large (in absolute magnitude) value based on time-in-cycle. The EOC DTC is biased more negative in accordance with Reference 1.

x An upper bound nominal initial RCS average temperature for Mode 1 HFP is modeled to predict conservative input to the subsequent DNBR calculations. The S-RELAP5 calculation uses a nominal temperature without adjustment for measurement uncertainty. The measurement uncertainty is applied in the DNBR calculations.

x The automatic rod control system is disabled; therefore, it is not modeled in this analysis.

x A bounding spectrum of dropped RCCA worth is modeled to provide a range of core thermal hydraulic conditions for subsequent SAFDL analyses. A range of dropped RCCA worths from 25 to 2000 pcm is analyzed to allow determination of the most limiting combination of augmented radial peaking and core boundary conditions. The range of dropped worth bounds the worth of dropped RCCAs and RCCA banks.

x Four RCPs are modeled to be in operation consistent with Mode 1. RCS flow is modeled as the Technical Specification minimum flow rate. Minimizing RCS flow rate is conservative with respect to DNBR.

x Core average fuel pellet to cladding gap conductance is conservatively modeled to maximize rod surface heat flux. Core average gap conductance is biased high to produce a conservative prediction of Doppler feedback during the event.

x Average core fuel thermal-mechanical properties are modeled considering time-in-cycle and account for the effect of TCD.

x The turbine control valve is modeled in automatic mode to exacerbate RCS cooldown leading to higher core power levels resulting from positive moderator feedback.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 153 x Since the systems designed to mitigate this event are redundant, there is no single active failure that will adversely affect the consequences of the event.

Statically misaligned RCCA The steady-state initial conditions are input to the XCOBRA-IIIC code (Reference 8),

which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6). A conservative maximum radial peaking augmentation factor is determined considering bounding allowances for RCCA misalignment along with xenon redistribution.

Withdrawal of a single RCCA The overall system response for a single RCCA withdrawal is identical to that for an uncontrolled RCCA bank withdrawal (Section 5.21). The difference is in the local peaking in the region of the single withdrawn RCCA that is not present if the entire bank is withdrawn. Therefore, the MDNBR calculation for the most limiting HFP case from the RCCA bank withdrawal analysis is reevaluated with a conservative radial peaking augmentation factor. The core fluid boundary conditions and average rod surface heat flux are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6).

5.22.3 Results Dropped RCCA / RCCA Bank A bounding spectrum of dropped RCCA worth is modeled to provide a range of core thermal-hydraulic conditions for subsequent SAFDL analyses. The core is not adversely affected during this period when the RCCA is dropping since power is decreasing rapidly. Following a dropped RCCA, in manual rod control, the plant will establish a new equilibrium condition due to positive reactivity feedback predominantly from changes to the moderator temperature. For dropped RCCA banks, the return to

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 154 power will be less due to the greater worth of the entire bank or a reactor trip will be initiated.

The sequence of events is similar for all cases except for the largest worth (absolute) dropped banks which initiate a reactor trip. As such, a representative case (i.e., 25 pcm dropped worth) is presented in Table 5-19. Figure 5-62 provides the core power response which shows that core power reaches an equilibrium condition from reactivity feedback. Figure 5-63 and Figure 5-64 show the RCS temperature and pressurizer pressure responses, respectively. Following plant stabilization, normal rod retrieval or shutdown procedures are followed. The operator may manually retrieve the RCCA by following approved operating procedures.

Table 5-20 presents the MDNBR and PLHGR results for the limiting case for this event.

The results demonstrate that acceptance criteria are met.

Statically misaligned RCCA The consequences of this event are bounded by the dropped RCCA analysis; therefore, the acceptance criteria are met.

Withdrawal of a single RCCA Table 5-21 presents the MDNBR and PLHGR results for this event. The results demonstrate that acceptance criteria are met.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 155 Table 5-19 RCCA Drop: Sequence of Events Case Event Time (sec.)

RCCA drop initiates 0.0 HFP, 25 pcm Minimum core power 4.4 Maximum return-to-power 300.0 Table 5-20 RCCA Drop: Results Criterion Result Limit MDNBR 1.146 1.142 PLHGR 20.8 [ ]

Table 5-21 Single RCCA Withdrawal: Results Criterion Result Limit MDNBR 1.202 1.142 PLHGR 18.4 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 156 Figure 5-62 RCCA Drop (25 pcm): Core Power

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 157 Figure 5-63 RCCA Drop (25 pcm): Reactor Coolant System Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 158 Figure 5-64 RCCA Drop (25 pcm): Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 159 5.23 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (FSAR SP 15.4.4)

The plant Technical Specifications do not permit operation in Modes 1 and 2 with fewer than four reactor coolant loops operating.

5.24 A Malfunction of Failure of the Flow Controller in a Boiling Water Reactor Loop that Results in an Increased Reactor Coolant Flow Rate (FSAR SP 15.4.5)

This event is not applicable to Callaway.

5.25 Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant (FSAR SP 15.4.6) 5.25.1 Event Description A boron dilution event is caused by a malfunction or inadvertent operation of the chemical and volume control system (CVCS) that results in the reduction of the boron concentration in the RCS. The reduction of the boron concentration causes a positive reactivity insertion which could increase core power and challenge DNBR and FCM.

x In Mode 1, the results of the boron dilution event are bounded by the range of reactivity insertion rates considered for the uncontrolled bank withdrawal event (Section 5.21). Therefore, the Mode 1 boron dilution event does not require reanalysis with Framatome fuel.

x In Modes 2 through 5 the event is analyzed to assess the adequacy of allowed operator response times (Mode 2) or the boron dilution mitigation system (BDMS) (Modes 3, 4, and 5) to prevent core re-criticality. The time required for a return to power is based upon the dilution flow rate, the mixing volume, temperature, pressure, the initial boron concentration, and initial shutdown margin. Modes 2 through 5 do not involve system transient calculations but the time to re-criticality is analyzed for the VQP.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 160 x Mode 6 does not require analysis because an uncontrolled boron dilution event will not occur during this mode. Inadvertent dilution via unborated water sources is prevented by administrative controls described in the plant Technical Specifications (Reference 2) Section 3.9.2 which isolates the RCS from potential sources of unborated water.

The boron dilution event is classified as an ANS Condition II event.

5.25.2 Analysis Method The boron dilution event is analyzed with the approved Reference 1 methodology. The instantaneous mixing model is applicable when at least one RCP is operating and assumes the unborated water is instantaneously mixed with the entire water volume in the RCS. Because no operations are permitted that would initiate a boron dilution without at least one RCP operating, this mixing model can be used exclusively.

For Modes 2-5 the critical boron concentration is calculated. The instantaneous mixing model is then used to determine the time necessary for the RCS to dilute from the initial boron concentration to the critical boron concentration. In addition, for Modes 3-5, the concentration at which the BDMS would initiate is calculated, and the instantaneous mixing model is used to calculate the time for the RCS to dilute to the BDMS initiation concentration.

Consistent with the current licensing basis, the acceptance criteria for this event are:

x Cold shutdown, hot shutdown, hot standby (Modes 5, 4, 3): The BDMS, with associated delays, must activate before the time of loss of shutdown margin.

x Startup, and power operation (Modes 2, and 1): If operator action is required to terminate the transient, a minimum time interval of 40 minutes (Mode 2) or 34 minutes (Mode 1) must be available from the time of initiation of the dilution and the time of loss of shutdown margin.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 161 Consistent with the current licensing basis, the following assumptions are made to conservatively predict the consequences of this event:

x At least one RCP is running and therefore the instantaneous mixing model can be used.

x A boron dilution event cannot occur in Mode 6.

x Dilution events that progress too slowly to trigger the BDMS actions are assumed to be bounded by the more severe dilutions considered explicitly.

5.25.3 Results The results of the boron dilution analysis in Table 5-22 show that there is adequate time for the operator or BDMS to terminate the source of dilution flow during all modes of operation. Boron dilution during power operation is bounded by the analysis presented in Section 5.21.

Table 5-22 Inadvertent Boron Dilution: Results Critical Time - BDMS Initiation Time Delay Allowance Margin Mode (minutes) (minutes) (minutes) 5 19.5 6.6 12.9 4 13.2 4.5 8.7 3 12.6 4.5 8.1 Critical Time Response Time Margin Mode (minutes) (minutes) (minutes) 2 41.5 40 1.5 5.26 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (FSAR SP 15.4.7)

Evaluation of this event will be provided by the Licensee.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 162 5.27 Spectrum of Rod Cluster Control Assembly Ejection Accidents (FSAR SP 15.4.8)

The analysis of this event for the Framatome VQP is outside the scope of this document.

5.28 Inadvertent Operation of the Emergency Core Cooling System During Power Operation (FSAR SP 15.5.1)

Inadvertent operation of the ECCS at power could be caused by operator error or a false electrical actuation signal. Following the actuation signal, the ECCS CCPs inject borated water into the cold leg of each loop. The safety injection pumps also start automatically but provide no flow when the RCS is at normal pressure.

If the RPS does not produce an immediate reactor trip on a spurious safety injection signal, the reactor will experience a negative reactivity excursion due to the injection of the borated water. The addition of negative reactivity causes a decrease in core power and core temperature with subsequent decrease in RCS pressure with the overall response resulting in an increase in the margin to DNB. Subsequently, reactor trip will occur on low pressurizer pressure or manual reactor trip, and this event poses no challenge to the SAFDLs. No aspect of Framatome fuel will significantly affect the injection flow rates or reactivity insertion of the ECCS, and the system transient response to the inadvertent operation of the ECCS regarding the SAFDLs is unaffected by Framatome fuel. Therefore, SAFDL analyses are not required for the Framatome VQP.

The event must also maintain reactor coolant and main steam systems below 110% of the design pressures, and not generate a more serious plant condition without other faults occurring independently. No aspect of Framatome fuel will significantly affect the controlling parameters for this aspect of the event, i.e., initial conditions, system setpoints and capacities, or operator action times. The challenge to these criteria is not impacted by Framatome fuel; thus, this event does not require reanalysis for the Framatome VQP.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 163 5.29 Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory (FSAR SP 15.5.2)

An increase in reactor coolant inventory may involve injected fluid, which is either unborated water, higher boron concentration than the RCS, or at boron concentrations that are the same as the RCS boron concentration. An increase in reactor coolant inventory which results from the addition of cold, unborated water to the RCS is discussed in Section 5.25. An increase in reactor coolant inventory which results from the injection of highly borated water into the RCS is discussed in Section 5.28.

If the injected boron concentration is nearly the same as the RCS concentration, this event is not a reactivity event. Core power and RCS temperature change very little during the event because the CVCS malfunction event is not causing changes in core reactivity. Therefore, this event does not challenge the SAFDLs and does not require reanalysis for the Framatome VQP.

5.30 A Number of BWR Transients (FSAR SP 15.5.3)

These events are not applicable to Callaway.

5.31 Inadvertent Opening of a Pressurizer Safety or Relief Valve (FSAR SP 15.6.1) 5.31.1 Event Description An accidental depressurization of the RCS could occur from an inadvertent opening of a pressurizer relief or safety valve. Since a safety valve is sized to relieve approximately twice the steam flow rate of a relief valve and will therefore allow a much more rapid depressurization upon opening, the most severe core conditions resulting from an accidental depressurization of the RCS are associated with an inadvertent opening of a pressurizer safety valve.

Initially, the event results in a rapidly decreasing RCS pressure until this pressure reaches a value corresponding to the hot leg saturation pressure. At this time, the pressure decrease is slowed considerably. The pressure continues to decrease

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 164 throughout the transient. Pressurizer level increases initially due to expansion caused by depressurization and then decreases following reactor trip. Reactor power remains effectively constant until control rods are inserted. The reactor may be tripped by the 277DQGORZSUHVVXrizer pressure RPS signals.

An inadvertent opening of a pressurizer relief valve is classified as an ANS Condition II event. The failure of a pressurizer safety valve is classified as an ANS Condition III event. The analysis performed conservatively bounds the more limiting failure while still applying the more restrictive Condition II acceptance criterion of ensuring the DNB design basis is met. See Section 3.11 for a discussion of the event classifications and acceptance criteria.

5.31.2 Method of Analysis Detailed analyses are performed with the approved non-LOCA methodology given in Reference 1 with the changes described in Section 3.9. The S-RELAP5 code is used to calculate the system response. The core fluid boundary conditions and average rod surface heat flux from the S-RELAP5 calculation are input to the XCOBRA-IIIC code (Reference 8), which is used to calculate the MDNBR using the ORFEO-GAIA CHF correlation (Reference 6).

The criterion related to the DNB SAFDL is evaluated. Since this event does not exhibit a significant power increase, FCM is not challenged. This event also does not challenge the RCS and main steam system pressure boundary limits. The potential to generate a more serious plant condition is generally related to overfilling the pressurizer; however, as the change in fuel does not affect the important phenomena for pressurizer overfill, this criterion is not analyzed for the VQP.

The following assumptions are incorporated in the analysis to give conservative results in the prediction of the consequences of this event:

x This event is initiated from HFP. Power measurement uncertainty is included in the initial core power level modeled in the S-RELAP5 calculations.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 165 x An upper bound nominal initial RCS average temperature for Mode 1 (HFP) is modeled to predict conservative input to the subsequent DNBR calculations. The S-RELAP5 calculation uses a nominal temperature without adjustment for measurement uncertainty. The measurement uncertainty is applied in the DNBR calculations.

x RCS flow is modeled as the Technical Specification minimum flow rate.

Minimizing RCS flow rate is conservative with respect to DNBR.

x Moderator density reactivity feedback is modeled based on the most-positive TS/COLR MTC limit at HFP of 0.0 pcm/°F.

x The BOC HFP DTC is biased less negative in accordance with Reference 1 topical report.

x The maximum flow rate of the failed open pressurizer safety valve is assumed to be 20% above the rated flow rate to cover the American Society of Mechanical Engineers requirement that the rated flow of a safety valve is reduced by 10%

from flow tests.

x Since the systems designed to mitigate this event are redundant, there is no single active failure that will prevent operation of the reactor protection system.

x The automatic rod control system is disabled; therefore, it is not modeled in this analysis.

5.31.3 Results The sequence of events is shown in Table 5-23. Figure 5-65 through Figure 5-67 show the transient response for an inadvertent opening of a pressurizer safety valve. The system response, shown in Figure 5-65, illustrates the core power and rod surface heat flux responses. Nuclear power remains near the initial value until reactor trip occurs when the OTT setpoint is reached. The rate of depressurization in the pressurizer is relatively constant during the event as shown in Figure 5-66. RCS temperatures are given in Figure 5-67.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 166 Table 5-24 presents the MDNBR and peak LHGR for this event. The results demonstrate that acceptance criteria are met.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 167 Table 5-23 Inadvertent Opening of a Pressurizer Safety Valve: Sequence of Events Case Event Time (sec.)

Safety valve opens fully 0.0 OTT reactor trip setpoint reached 33.5 HFP, BOC Rods begin to drop 35.6 Minimum DNBR occurs 35.7 Table 5-24 Inadvertent Opening of a Pressurizer Safety Valve: Results Criterion Result Limit MDNBR 1.463 1.142 PLHGR (kW/ft) 18.3 [ ]

Criteria (c) and (d) defined in associated affidavit for this document apply to bracketed material on this page.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 168 Figure 5-65 Inadvertent Opening of a Pressurizer Safety Valve: Core Power and Core Average Heat Flux

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 169 Figure 5-66 Inadvertent Opening of a Pressurizer Safety Valve: Pressurizer Pressure

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 170 Figure 5-67 Inadvertent Opening of a Pressurizer Safety Valve: Reactor Coolant System Temperatures

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 171 5.32 Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundary that Penetrate Containment (FSAR SP 15.6.2)

A radiological dose analysis is performed for this event, but a system transient response analysis is not performed. Therefore, no system transient response is required for the Framatome VQP.

5.33 Steam Generator Tube Failure (FSAR SP 15.6.3)

This event is the complete severance of a single SG tube. The transient is assumed to take place at HFP with the reactor coolant contaminated with fission products corresponding to continuous operation with a limited number of defective fuel rods. The event leads to an increase in contamination of the secondary system due to leakage of radioactive coolant from the RCS. In the event of a coincident LOOP or failure of the condenser steam dump system, discharge of activity to the atmosphere could take place via the SG PORVs or main steam safety valves.

The DNBR response due to the depressurization of the RCS from the ruptured tube is less severe than the inadvertent opening of a pressurizer safety or relief valve documented in Section 5.31. Therefore, a DNB analysis is not performed for this event.

The current AOR contains a system transient response analysis of this event for SG overfill and input to the radiological dose analysis. The consequences of this event primarily depend on the break flow rate, secondary side relief setpoints and capacity, charging and safety injection flow rates, and operator actions. Framatome fuel does not impact any of these controlling parameters; thus, this event does not require reanalysis for the Framatome VQP.

5.34 Spectrum of Boiling Water Reactor Steam System Piping Failures Outside of Containment (FSAR SP 15.6.4)

This event is not applicable to Callaway.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 172 5.35 Loss-of-Coolant Accidents Resulting from a Spectrum of Postulation Piping Breaks Within the Reactor Coolant Pressure Boundary (FSAR SP 15.6.5)

LOCA events are outside the scope of this document.

5.36 A Number of BWR Transients (FSAR SP 15.6.6)

These events are not applicable to Callaway.

5.37 Radioactive Release from a Subsystem or Component (FSAR SP 15.7)

This event can be caused by any of the following events:

x Radioactive gas waste system leak or failure (ANS Condition III).

x Radioactive liquid waste system leak or failure (ANS Condition III).

x Postulated radioactive release due to liquid tank failures (ANS Condition IV).

x Fuel handling accident (ANS Condition IV).

Assessment of radiological doses is outside the scope of this document.

5.38 Anticipated Transient Without Scram (FSAR SP 15.8)

The effects of anticipated transients without scram are not considered as part of the design basis for transients analyzed in Chapter 15. Therefore, the Framatome VQP does not affect the results of this analysis.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 173

6.0 REFERENCES

1. EMF-2310(P)(A), Revision 1, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, May 2004.

2. Callaway Plant Technical Specifications, Amendment 221.

3. ANSI-N18.2, Nuclear Safety Criteria for the Design of Stationary PWR Plants, 1973.

4. XN-NF-82-21(P)(A), Revision 1, Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, August 1983.

5. EMF-92-081(P)(A), Revision 1, Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors, February 2000.

6. ANP-10341(P)(A), Revision 0, The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations, September 2018.

7. BAW-10231P-A Revision 1, COPERNIC Fuel Rod Design Computer Code, January 2004.

8. XN-75-21(P)(A), Revision 2, XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady State and Transient Core Operation, March 1985.

9. ANP-10297P-A, Revision 0, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results, February 2013.

10. ANP-10297, Revision 0, Supplement 1PA, Revision 1, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results, December 2020.

11. BAW-10240(P)(A), Revision 0, Incorporation of M5TM Properties in Framatome ANP Approved Methods, May 2004.

12. ANP-10311P-A Revision 1, COBRA-FLX: A Core Thermal-Hydraulic Analysis Code, October 2017.

Framatome Inc. ANP-3969NP Revision 2 Callaway Non-LOCA Summary Report Page 174

13. XN-NF-82-06(P)(A), Revision 1, Supplement 2, 4, and 5, Qualification of Exxon Nuclear Fuel for Extended Burnup, October 1986.

14. ANP-10338P-A, Revision 0, AREATM - ARCADIA Rod Ejection Accident, December 2017.