ML20099D803
| ML20099D803 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 07/31/1992 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | |
| Shared Package | |
| ML20099D802 | List: |
| References | |
| DCTR-10, NUDOCS 9208070093 | |
| Download: ML20099D803 (530) | |
Text
{{#Wiki_filter:E_ 1 SYSTEM 80+ SHUTDOWN RISK EVALUATION REPORT DCTR 10 + JULY 31, 1992 ABB-COMBUSTION ENGINEERING NUCLEAR POWER SYSTEMS WINDS 0P,, CONNECTICUT ESA*1883R$$8o8502 A PDR
LEGAL NOTICE T!!IS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY ABB COMBUSTION ENGINEERING. NEITHER ABB-COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF: l A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, vs COMPLETENESS, OR .USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS; OR B.- ASSUMES ANY LIABILITIES W1TH RESPECT TO THE USE OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS-DISCLOSED IN THIS REPORT. i e
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I ABSTRACT In engineering the System 80+ Standard Plant Design, ABB-Combustion . Engineering recognized the significance of addressing safety.during shutdown operations. System 80+ is engineered with features that enhance Jshutdown safety: 1) by deliberate system engineering, equipment specification 'and plant- arrangements for shutdown operation, 2) by mode dependent control logic that assists and limits _ operations, 3).by instrumentation, displays and' alarms that clearly portray plant status _ in each' mode . and 4!-by thorough procedural guidance and Technical Specifications that address important shutdown evolutions. This report presents these features and evaluates them in the context of.the specific shutdown issues
~
identified Dy_the NRC. The report fulfills the ABB-Combustion Engineering. commitments to the NRC _to 1) provide shutdown information in support of.the System 80+ Design Certification and 2)_ provide responses to specific-RAI's on shutdown operations.
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77a.wp (9212) bh DEFINITIONS The following definitions of terms are en. ployed throughout-this report. DEFINITIONS AVAILABLE (AVAILABILITY): The status of a system, structure or component that is in service or can be placed in service in a FUNCTIONAL or OPERABLE state by immediate manual or automatic actuation. CONTAINMENT CLOSURE: The action to secure primary (PWR) or secondary (BWR) containment and its associated structures, systems, and components as a FUNCTIONAL barrier to fission product release under existing plant conditions. CONTINGENCY PLAN: An approved plan of compensatory actions: o To maintain DEFENSE IN DEPTH by alternate means when pre-outage planning reveals that specified systems, structures
- or components will be unavailable; i
o To restore DEFENSE IN DEPTH when system AVAILABILITY drops below th0 planned DEFENSE IN DEPTH during the outage; o To minimize likelihood of a less of KEY SAFETY FUNCTIONS during HIGHER RISK EVOLUTIONS. DECAY HEAT REMOVAL CAPABILITY: The ability to maintain reactor coolant system (RCS) temperature and pressure, and spent fuel pool (SFP) temperature below specified limits following a shutdown. i DEFENSE IN DEPTH: For the purpose of managing risk during shutdown, defense in depth is the concept of; l o Providing systems, structures and components to ensure l backup of KEY SAFETY FUNCTIONS using redundant, alternate
- j. or diverse methods; o Planning and scheduling outage activities in a manner that i optimizes safety system AVAILABILITY; o Providing administrative controls that support and/or supplement the above elements.
DEFUELED: All fuel assemblies have been removed from the reactor vessel and placed in the spent fuel pool or other storage facility.
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! 77c.wp(9212)bh l DIVISION: One or more trains that share a cummon component, e.g. AC power. Divisions are the highest level of separation and independence. ESTIMATED CRITICAL POSITION (ECP): A calculated set of reactor conditions and/or parameters that define a critical reactor state. 4 (Keff = 1.0) l FUNCTIONAL (FUNCTIONALITY): The ability of a system or component to perform its intended service with considerations that applicable technical specification requirements or licensing / design basis assumptions may not be maintained. HIGHER RISK EVOLUTIONS: Outage activities, plant configurations or conditions during shutdown where the plant is more susceptible to an event causing the loss of a key safety function. INVENTORY CONTROL: Measures established to ensure that irradiated fuel remains covered with coolant to maintain heat transfer and shielding requirements. KEY BAFETY FUNCTIONS: During shutdown, they are decay heat removal, inventory control, power availability, reactivity control, and containment. MID-LOOP: PWR condition with fuel in the reactor vessel and level below the top of the hot legs at their junction to the reactor vessel. MODE: The reactor cperating state defined by reactivity, power level and coolant temperature, as follows: MODES MODE REACTIVITY % RATED COOLANT CONDITION, K.fr THERMAL POWER TEMPERATURE, 'F 1 2 0.99 >5 2 350 2 2 0.99 55 2 350 3 < 0.99 NA 2 350 4 < 0.99 NA 350 > Ty. > 210 5 < 0.99 NA $ 210 6* s 0.95 NA $ 135 Fuel in the reactor vessel with one or more of the vessel head closure studs less than fully tensioned or with the head removed.
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l 77a.wp(9212)bh l-l OPERABLE: The ability of _ a system to perform its specified i- function with all applicable technical specifications requirements I catisfied. REACTIVITY CONTROL: Measures established .to precluded inadvertent dilutions, criticalities, power excursions or losses of shutdown margin, and to predict and monitor core behavior. REDUCED INVENTORY: PNR condition with fuel in the reactor vessel L and level lower than three feet below the reactor vessel flange. RISK MANAGEMENT: Integrated process of assessing and reducing the likelihcod and/or consequences of an adverse event. SHUTDOWN: Plant status when the reactor core is subcritical and a ! startup-is not in progress. STARTUP: Plant status commencing with activities to neatup the RCS above 200 degrees-F and to bring the reactor core to a critical condition and up to 5% of rated thermal power. l TRAIN: A set of safety related components that perform a safety ! function. Trains performing redundant functions are physically, i electrically, and mechanically separated to the extent necessary to ; insure independent performance of its safety function. 1 l l l l l i
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' LIST OF ACRONYMS ACC -
Advanced Control Complex ACC - Advanced Control Complex ADV - Atmosphere Dump Valve . ALWR - Advanced Light Water Reactor ! ARO - All Rods Out ! BAST - Boric Acid Storage Tank CC_ - Component Cooling CCW - Component Cooling Water System CD - Condensate CDF -
-Core Damage Frequency CEA -
Control Element Assembly CETL - Core-Exit Thermocouple CI - Containment. Isolation CPC - Core-Protection Calculator CRO - Control Room Operator CS - Containment Spray CSAS - Containment-Spray Actuation Signal
. CSS -
Containment Spray System CVCS - Chencial and volume control System CW - Circulating Water DG - Diesel Generator DBA - Design Basis Accident DBE - Design Basis Event DEHLS - Double Ended Hot Leg Slot DESLS- -- Double Ended Suction Leg Slot DF - Decontainment Factor DHR - Decay Heat Removal DIAS - Discrete Indication Alarm System DLS - Diesel Load Sequence dP - Pressure Differential DVI - Direct Vessel Injection EAB - Exclusion AreaEBoundary , -ECCS - EmergencyLCore Cooling System-EDS. - Electrical Distribution System L EF - Emergency Feedwater EFAS - Emergency Feedwater Actuation Signal , EFW - Emergency Fcedwater L EPRI - Electric Power Research-Institute
- FW -
Feed Water L GIS - Generated Iodine Spike l HACT - Head Area Cable Tray Assembly HCR -
-Human Congnitive Ecliability HJTC -
Heated. Junction Thermocouple HPSI. - High Pressure Safety Injection HVT - Holdup Volume Tank I&C - Instrumentation & Control IA - Instrument Air IBD - Inadvertent Boron dilution ICI - In-Core Instrument
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770.wp(9212)bh INPO - Institute for Nuclear Power Operations IPSO -- Integrated Process Status Overview IRWST - In-containment Refueling Water Storage Tank LCO - Limiting Condition for Operation LER - Licensco Event Report LOCA - Loss of Coolant Accidents LOCV - Loss of Condoser Vacuum LOP - Loss of Offsite Power LTOP - Low Temperature Overpressure Protection MFW - Main Feedwater MMI - Man-Machine Interface MPC - Maximum Permissible Concentration MSIV - Main Steam Isolation Valve NPSH - Net Positive Suction Head NPSHA - Net Positive Suction Head Available OSI - Operational Support Information P&ID - Piping and Instrumentation Diagram PCT - Peak Clad Temperature PNS-Bus - Normal Permanent Non-Safety Bus RAI - Requests for Additional Information RC - Reactor Coolant RCP - Reactor Coolant Pump RCS - Reactor Coolant System REM - Realistic Evaluation Model RHR - Residual Heat Removal RTD - Hot Leg Resistance Temperature Detector SC - Shutdown Cooling SCS - Shutdown Cooling System SD - Safety Depressurization SDS - Safety Depressurization System SG - Steam Generator SGTR - Steam Generator Tube Rupture SI - Safety Injection SIS - Safety Injection System SIT - Safety Injection Tank SRP - Standard Review Plan SUFW - Start Up Feedwater SW - Sersice Water TB -
'i 17e Bypass TS -
Te n1Nal Specification URD - UtiAlt} Requirements Document
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_ . = . _ _ . .. . ~ . . . _ =. _ 1716.wp(0212)bh: IARLE OF CONTENTS SECTIOli SUBJECT PN;E 10. ABSTRACT i i DEFINITIONS 11 LIST OF ACRONYMS y TABLE OF CONTENTS vii LIST OF TABLES xx LIST OF FIGURES xxiii
1.0 INTRODUCTION
1-1 1.1 PURPOSI 1-1 1.2 SCOPE 1-1
1.3 BACKGROUND
1-1
? 4 SYSTEM 80+ FEATURES 1-2 1
2.0 SHUTDOWN RISK ISSUES 2-1 2.1 PROCEDURES 2-1 ] l 2.1.1 ISSUE 2.1-1 l 2.1.2 ACCEPTANCE CRITERIA 2.1-1 2.1.3 DISCUSSION 2.1-1 2.1.4 RESOLUTION 2.1-2 2.2 TECHNICAL SPECIFICATION I JMPROVEMENTS 2.2-1 1 2.2.1 ISSUE 2.2-1 2.2.2 ACCEPTANCE CRITERIA 2.2-1 2.2.3 DISCUSSION 2.2-1 j 2.2.4 RESOLUTION 2.2-2 l l t
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770.wp(9212)bh TABLE OF CONTENTS (Cont'd) SECTION SUBJECT PNE N3. 2.3- REDUCED INVENTORY OPERATION AND GL 88-17 FIXES 2.3-1 2.3.1 ISSUE 2.3-1 2.3.2 ACCEPTANCE CRITERIA 2.3-1 2.3.3 DISCUSSION 2.3-2 2.3.3.1 Instrumentation for Shutdown Operations 2.3-2 2.3.3.2 SCS Desian 2.3-3 2.3.3.3 Steam Generator Nqgzle Dam Intecrity 2.3-4 2.3.3.4 ALTERNATE INVENTORY ADDITIONS AND DHR METHODS 2.3-6 2.3.3.5 OPERATIONS 2.3-8 2.3.4 RESOLUTION 2.3-9 2.4 LOSS OF DECAY HEAT REMOVAL CAPABILITY 2.4-1 2.4.1 ISSUE 2.4-1 2.4.2 ACCEPTANCE CRITERIA 2.4-1 2.4.3 DISCUSSION 2.4-1 2.4.3.1 Shutdown Event Initiation and Analyses 2.4-2 2.4.3.1.1- Introduction 2.4-2 l2.4.3.1.2 Resistance to Initiators 2.4-3 2.4.3.1.3 Recovery from Initiators 2.4-5 l 2.4.3.1.3.1 Group I Initiators 2.4-6 2.4.3.1.3.1.1 Recovery During Mode 5 2.4-6 2.4.3.1.3.1.2 Recovery During Mode 6 2.4-7 2.4.3.1.3.2 Group II Initiators 2.4-7 2.4.3.1.3.2.1 Recovery During Mode 5 2.4-8 2.4.3.1.3.2.2 Recovery During Mode 6 2.4-8
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TABLE-OP CONTENTS (Cont'd) i SECTION SUBJECT PIGE 10. 2.4.3.1.3.2.3 Group III Initiators 2.4-8 2.4.3.1.3.2.4 Group IV Initiators 2.4-8 2.4.3.1.4 Recovery Based on Plant Configuration 2.4-8 2.4.3.1.5 Conclusions 2.4-9 2.4.3.2 System 80+ AC Power Availability 2.4-25 2.4.3.2.1 Introduction 2.4-25 2.4.3.2.2 Discussion 2.4-25 2.4.3.2.3 Conclusion 2.4-26 2.4.3.3 System 80+ Diesel Generator Availability 2.4-27 2.4.3.3.1 Introduction 2.4-27 2.4.3.3.2 Discussion 2.4-27 2.4.3.3.3 Conclusion 2.4-29 2.4.4 RESOLUTION 2.4-29 2.5 -PRIMARY / SECONDARY CONTAINMENT CAPABILITY AND SOURCE TERM 2.5-1
- 2. Sol ISSUE 2.5-1 2.5.2 ACCEPTANCE CRITERIA 2.5-2 2.5.3 DISCUSSION 2.5-2 2.5.3.1 Problem Formulation 2.5-2 2.5.3.2 Containment Intecrity 2.5-2 2.5.3.2.1 Integrity Requirements 2.5-3 2.5.3.2.1.1 Modes 1-4 2.5-3 2.5.3.2.1.2 -Mode 5 2.5-3 '
2.5.3.2.1.2.1. .RCS Level.Above Reduced Inventory 2.5-3 4 2.5.3.2.1.2.2 RCS Level Below Reduced Inventory 2.5-3 2.5.3.2.1.3 Mode 6 2.5-4 2.5.3.2.2- System 80+ Containment Features 2.5-4 2.5.3.2.2.1- Building Arrangement and Ventilation 2.5-4 2.5.3.2.2.2 Personnel Locks 2.5-4 2.5.3.2.2.3 Equipment Hatch 2.5-5 2.5.3.2.2.4 Penetrations 2.5-6 2.5.3.3- Events Analyzed 2.5-6
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2 770.vp(9212)bh IABLE OF CONTENTS (Cont'd) SECTION SUBJECT = PN3E 10. 2.5.3.4. Acceptance Criteria 2.5-7 2.5.3.4.1 Radiation Limits. 2.5-7 2.5.3.4.1.1 Loss of Shutdown Cooling; Site Boundary Limits 2.5-7 2.5.3.4.1.2. Loss of Coolent Accidents; Site Boundary Limtis 2.5-8 2.5.3.4.1.3 Limits on Utility Personnel -2.5-8 ' 2.5.3.4.1.3.1 Air Borne Radiation -2.5-8
-2.5.3.4.1.3.2 Whole Body Radiation 2.5-9 2.5.3.4.2 Temperature Limits 2.5-9 2.5.3.5 Analysis 2.5-10 2.5.3.5.1 Thermodynamic' Conditions 2.5-10 2.5.3.5.2 Radiation Release 2.5-11 2.5.3.5.3 Results 2.5-12 2.5.3.5.3.1 Mode 5:- Loss-of-Shutdown Cooling
. at Reduced Inventory 2.5 2.5.3.5.3.1.1 Site Boundaries 2.5-13 2.5.3.5.3.1.2 Utility Personnel 2.5-13 2.5.3.5.3.2 Mode 5: LOCA 2.5-14
-2.'5.3.5.3.2.1 Site Boundaries 2.5-14 2.5.3.5.3.2.'2 Utility-Personnel 2.5-14 2.5.3.5.3.3 Mode 6: Refueling; Inventory Boil-Off 2.5-34 2.5.3.5.3.3.1. Site Boundary Limits 2.5-15
-2.5.3.5.3.3.2 Utility Personnel 2.5-15 2.5.4 RESOLUTION 2.5-15 2.5.4.1 Mode 5; Reduced Inventory (Loss of Shutdown Coolina) 2.5-16 2.5.4.2- Mode 5; Full Inventory (LOCA) 2.5-16 l
-l 2.5.4.3 ' Mode 6 Refueliner Confiauration (Inventerv Boil Off) 2.5-16 .)
i 2.6 RAPID BORON DILUTION 2.6-1 : 2.6.1. ISSUE 2.6-1 2.6.2 ACCEPTANCE CRITERIA 2.6-1 l _x_
77a.wp(9212)bh TABLE OF CONTENTS (Cont'd) SECTION SUBJECT P/GE 10. 2.6.3 DISCUSSION 2.6-2 2.6.3.1 Identification of Dilution Sources 2.6-2 2.6.3.2 Event Analyzed 2.6-2 2.6.3.3 Mathematical Model 2.6-4 2.6.3.4 Results 2.6-4 2.6.3.5 Conclusion 2.6-4 2.6.4 RESOLUTION 2.6-4 2.7 FIRE PROTECTION 2.7-1 2.7.1 ISSUE 2.7-1 2.7.2 ACCEPTANCE CRITERIA 2.7-1 2./.3 DISCUSSION 2.7-1 2.7.3.1 Mitication of Fire Consecuences 2.7-1 2.7.3.2 Detection and Suppression of Fires 2.7-2 2.7.3.3 Prevention of Fires 2.7-3 2.7.4 RESOLUTION 2.7-4 2.8 INSTRUMENTATION 2.8-1 2.8.1 ISSUE 2.8-1 2.8.2 ACCEPTANCE CRITERIA 2.8-2 2.3.3 DISCUSSION 2.8-3 2.i>.3.1 Instrumentation Desion Basis 2.8-3 2.8.3.2 Jnstrumentation Description 2.8-4 2.8.3.2.1 Level 2.8-4 2.8.3.2.2 Temperature 2.8-6
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770.wp(9212)bh TABLE OF COlRQQS (Cont'd) -SECTION EUBJECT ' PAGE lot 2.8.3.2.3 Shutdown Cooling System Performance 2.8-6 2.8.3.2.4 Quality Assurance 2.8-7 2.8.3.2.5 Display and Monitoring capability 2.8-8 2.8.3.2.5.1 IPSO 2.8-8 2.8.3.2.5.2 Alarm Tiles and Associated Alarm Moosages 2.8-10 2.8.3.2.5.3 Discrete Indicators 2.8-11 2.8.3.2.-5.4- CRT Display-Pages 2.IF12 2.8.3.2.5.5 Component and Process Control Indicators 2.8-12 2.8.3.2.5.6 NUPLEX 80+ Alarm Characteristics 2.8-13 2.8.4 RESOLUTION 2.8-14 _2.9 ECCS RECIRCULATION CAPABILITY 2.9-1 2.9.1 ISSUE 2.9-1 2.9.2 ACCEPTANCE CRITERIA 2.9-1 2.9.3 DISCUSSION 2.9-2 2.9.4 RESOLUTION 2.9-4 12 . 1 0- _ EFFECTS OF'PWR UPPER INTERNALS 2.10+1 2.10.1 ISSUE 2.10-1 2.10.2 ACCEPTANCE CRITERIA 2.10-1
- 2.10.3 DISCUSSION 2.1CF1 2.10.4 RESOLUTION 2.10H3 2.11' FUEL' HANDLING AND HEAVY LOADS 2.11-1 2.11.1 ISSUE 2. U.-1 2.11.2 ACCEPTANCE CRITERIA 2 . U.-1
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77c.wp(9212)bh i TABLE OF CONTENTS (Cont'd) r SECTION . SUBJECT gg 2.11.3 DISCUSSION ?.11-1 2.11.4 RESOLUTION 2.11 -4 2.12 POTENTIAL FOR DRAINING THE REACTOR COOLANT SYSTfd 2.12-1 2.12.1 ISSUE 2.12-1 2.12.2 ACCEPTANCE CRITERIA 2.12-1 2.12.2.1 Prevention Criteria 2.12-1 F 2.12.2.2 Detection Criteria 2.12-1 2.12.2.3 Mitication Criteria 2.12-2 2.12.3 DISCUSSION. 2.12-2 2.12.3.1 Potential Drain Paths Directly from the Reactor Coolant Syste_m 2.12-4 2.12.3.2 Pot'ential Drainaae Paths Throuch Interfacina Systems 2.12-7
-2.12.3.2.1 Potential Drainage Paths from the.RCS Through the SCS 2.12.7 2.12.'3.2.2 Potential Drainage Paths from the RCS to the SIS /SCS 2.12.10 2.12.3.2.3 Potential. Drainage Paths-from the RCS to CVCS ? U.10 2.12.3.2.4-' Potential Drainage Paths from the RCS tc the SS 2.12-10 2.12.4 RESOLUTION 2.12-11 2.-13 ELOODING AND SPILLS- 2.13-1 2.13.1 ISSUE 2.13-1
'2.13.2 ACCEPTANCE CRITERIA 2.13-1 2.13.3 DISCUSSION 2.13-1 L
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770.wp(9212)bh -- TABLE OF CONTENTH (Cont'd) ' SECTION _ SUBJECT RN2 NO. 2.13.4 RESOLUTION 2.33 -2 3.0 PROBABILISTIC RISK ASSESSMENT 3-1 3.1 PRA
SUMMARY
AND CONCLUSION 3-1 ' 3.2- PRA INTRODUCTION 3-4 3.3 INITIATING EVENT FREOUENCIES 3-9 3.4 ACCIDENT SEQUENCES 3-12 3.4.1- LOSS OF DHR, MODE 4 3-12 3.4.2- LOSS OF DHR, MODE 5,6, REFUELING CAVITY EMPTY 3-19 3.4.3 LOSS OF DHR, MODE 5, REDUCED INVENTORY 3-22 3.4.4 LOSS OF DHR, MODE 6, REFUELING CAVITY FLOODED 3-25 3.4.5 LOCA, MODE 4, PRESSURE ABOVE 500 PSIG 3-27 3.4.6 LOCA, MODES 5, 6 (IRWST FULL) 3-31 3.4.7 LOCA OUTSIDE CONTAINMENT, MODES 6 3-33 3.4.8 LOCA INSIDE CONTAINMENT,_ MODE 6
-(REFUELING CAVITY FULL) 3-35 3.4.9 LOSS OF OFICITE POWER 3-36 3.4.10 CRITICALITY EVENTS 3-36 l 3.5 -RADIOLOGICAL CONSEOUENCES 3-57 h 4.0 APPLICABILITY OF' CHAPTER 15 ANALYSES 4-1 4.0.1 EORMAT AND CONTENT 4-1
[. 4.1 INCREASE IN HEAT REMOVAL BY THE L SECONDARY SYSTEM' 4.1-1 4.
1.0 INTRODUCTION
4.1-1 1
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. .. . - . - . . - - - ~ .__ .~ .. 770.wp(9212)bh TABLE OF_fETRIE (Cont'd) 1 SECTION SUBJECT p/ GEE 4.1.1 DECREASE IN FEEDWATER TEMPERATURE 4.1-1 , 4.1.2 INCREASE IN FEEDWATER FLOW 4.1-1 4.1.3 INCREASE MAIN STEAM FLOW 4.1-4 4.1.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE 4.1-4 4.1.5 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT 4.1-5 4.2 DICREASE IN HEAT. REMOV3JtR Fli10NDAR'I SYSTEM 4.2-1 4.
2.0 INTRODUCTION
4.2-1 4.2.1 LOSS OF EXTERNAL LOAD 4.2-1 4.2.2 TURBINE TRIP 4.2-1 4.2.3 LOSS OF CONDENSER VACUUM (LOCV) 4.2-1 4.2.4 MAIN STEAM ISOLATION VALVE CLOSURE 4.2-2 4.2.5 STEAM PRESSURE REGULATOR FAILURE 4.2-2 4.-2.6 LOSS OF NON-EMERGENCY AC POWER ' TO TIIE STATION 4.2-3 4.2.7 LOSS OF NORMAL FEEDWATER FLOW 4.2-3 4.2.8 -FEEDWATER SYSTEM PIPE BREAKS 4.2-3 4.3 p.ECREASE IN REACTOR COOL _ ANT FLOW RATE 4.3-1 4.3.0- INTRODUCTION 4.> 4.3.1 TOTAL' LOSS OF REACTOR COOLANT FLOW 4.3-1
- 4. 3. 2 - FLOW CONTROLLER MALFUNCTION CAUSING FLOW COASTDOWN 4.3-2
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770.wp(9212)bh f T TABLE OF-CONTENTS (Cont'd)
~SECTION SUBJECT PN3E 10.
4.3.3 SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER 4.3-I 4.3~.4 REACTOR COOLANT PUMP SHAFT BREAK WIT)! LOSS OF OFFSITE POWER 4.3-2 1 4.4- REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER 4.4-1 4.4.O INTRODUCTION 4.4-1 4.4.1 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FROM SUBCRITICAL OR LOW POWER CONDITIONS 4.4-1 4.4.2 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL-AT POWER 4.4-3 4.4.3 SINGLE CONTROL ELEMENT ASSEMBLY DROP 4.4-3 , 4.4.4 STARTUP OF AN INACTIVE REACTOR COOLANT PUMP 4.4 4.4.5 FLOW CONTROLLER MALFUNCTION CAUSING AN INCREASE IN BWR CORE FLOW RATE 4.4-3 4.4.6 _ INADVERTENT DEBORATION 4.4-3 4.4.7 INADVERTENT LOADING OF A FUEL ASSEMBLY INTO THE IMPROPER POSITION 4.4-3 4.4 N ' ' CONTROL ELEMENT ASSEMBLY (CEA) EJECTION 4 . 4 -3 4.5 INCREASE'IN RCS INVENTORY 4.'5-1 4.
5.0 INTRODUCTION
4.5-1 4.5.1 INADVERTENT OPERATION OF THE ECCS 4.5-1 4.5.2 CVCS MALFUNCTION-PRESSURIZER LEVEL CONTROL SYSTEM MALFUNCTION WITH LOSS OF OFFSITE POWER 4.5-1 4.6 DECREASE IN REACTOR COOLANT SYSTEM INVENTORY 4.6-1
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770.wp(9212)bh TABLE OF CONTENTS (Cont'd) SICTION SUBJECT PAGE 10. 4.
6.0 INTRODUCTION
4.6-1 4.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY / RELIEF VALVE 4.6-1 4.6.2 DOUBLE-ENDED BREAK OF A LETDOWN LINE OUTSIDE CONTAINMENT 4.6-1 4.6,3 STEAM GENERATOR TUBE RUPTURE 4.6-1 4.6.4 RACIOLOGICAL CONSEQUENCE OF MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR) 4.6-5 4.6.5 LOSS--OF-COOLAhI ACCIDENT 4.6-5 k_ 4.7 RADIOACTIVE MATERIAL RELEASE FROM A S_ilDSYSTEM OR COMPONENT 4.7-1 4.
7.0 INTRODUCTION
4.7.1 4.7.1 RADIOACTIVE GAS WASTE SYSTEM FAILURE 4.7.1 4.7.2 RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE 4.7.1 4.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID CONTAINING TANK FAILURES 4.7.1 4.7.4 FUEL HANDLING ACCIDENT 4.7.1 + 4.7.5 SPENT FUEL CASK DROP ACCIDENTS ( 4.7.1 5.0 APPLICABILITY OF CHAPTER 6 LOCA ANALYSES TO LOWER MODES OF OPERATION 5-1 5.1 ISSUE 5-1 5.2 ACCEPTANCE CRITERIA 5-1 5.3 DISCUSSION 5-1 3 5.3.1 DESChTPTION OF LOCA SCENARIO 5-4
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770.wp(9212)bh l TABLE OF CONTENTS (Cont'd) SECTION- SUBJECT PN3E 10. 3 5.3.2 SELECTION OF REVERSE PLANT PARAMETERS . AND CONDITIONS FOR MODE 4 ANALYSIS 5-5 5.3.3 ANALYSIS COMPUTER CODES 5-5 5.3.4 LOCA ANALYSIS FOR MODE 4 5-6 5.3.5 RESULTS OF LOCA ANALYSIS FOR MODE 4 5-6 5.3.5.1 Results of LOCA Cases With No SI Deliverv 5-7 5.3.5.2 Initiation of SI Delivery Within Ten Minutes After Break 5-8 5.4 BESOLUTIOt{ 5-9 6.0 . APPLICABILITY OF CHAPTER 6 CONTAINMENT ANALYSES 6-1
6.1 INTRODUCTION
6-1 62 -LOSS OF COOLANT ACCIDENTS (LOCAsi 6-1 6.3 MAIN STEAM LINE BREAKS (MSLBs) 6-2 6.4- INADVERTENT OPFRATION OF CONTAINMENT HEAT REMOVAL SYSTEMS 60 " 6.5- CONCLUSION 6-3 7.0 SYSTEM 804 DESIGN FEATURES FOR
-SIMPLICITY OF SHUTDOWN OPERATIONS 7-1
7.1 INTRODUCTION
7-1 7.2 DISCUSSION 7-1 7.2.1 TECHNICAL SPECIFICATIONS FOR REDUCED INVENTORY 7-1 7.2.2 SHUTDOWN COOLING SYSTEM 7-1 7.2.3 CONTAINMENT SPRAY SYSTEM 7-2 7.2.4- COMPONENT CQOLING WATER SYSTEM 7-2
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' 77a.wp (9212) bh TABLE OF CONTENTS (Cont'd)
' SECTIOl{ SUBJECT PAGE NO.
7_ . 2 . 5 - -STATION SERVICE WATER SYSTEM 7-3 7.2.6 -ELECTRICAL DISTRIBUTION SYSTEM 7-3
-7.2.7 NUPLEX 80+ ADVANCED CONTROL COMPLEX 7-3 7.2.8 REDUCED INVENTORY INSTRUMENTATION 7-5 7.2.9 CONTAINMENT 7-6 7.3 . CONCLUSION 7-7
-8.O CONCLUSIONS 8-1
- 9. 0- REFERENCES 9-1 APPENDICES
-APPENDIX A RESPONSES TO REQUESTS FOR A-1 ADDITIONAL INFORMATION APPENDIX ~B CESSAR-DC CP.ANGES B-1 APPENDIX C PROCEDURE GUIDANCE TO SUPPORT REDUCED RCS INVENTORY OPERATIONS C-1 APPENDIX D- TECHNICAL SPECIFICATION MARKUPS l- OF LCO AND APPLICABILITY L FOR-SHUTDOWN OPERATIONS D-1 l
APPENDIX E TECHNICAL SPECIFICATIONS FOR REDUCED RCS INVENTORY OPERATIONS E-1 L
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' 77c.wp(9212)bh ,5 ?
LIST OF TABLES i TABLE TITLE PME10. 1-1 SHUTDOWN EVENT CATEGORIES AND SYSTEM 80+ FEATURES FOR PREVENTION, DETECTION AND MITIGATION 1-4 1-2 SHUTDOWN-EVENTS AND SYSTEM 80+ PREVENTION, DETLCTION AND MITIGATION FEATURES 1-9
~2.1-1
SUMMARY
OF PROCEDURAL GUIDANCE 2 ELATED TO SHUTDOWN OPERATIONS 2.1-3 2.2-1 SYSTEM 80+ TECHNICAL SPECIFICATIONS MODIFICATIONS RELATED TO IMPROVED SHUTDOWN OPERATIONS 2.2-3 2'4-1 .
SUMMARY
.0F SYSTEM 80+ SCS DESIGN FEATURES THAT INCREASE IMMUNITY AGAINST INITIATORS 2.4-3 2.4-2 SCS INSTRUMENTATION 2.4-11 2.4-3 TERMINATION POINT (11 SHEETS, POINTS 1 THROUGH 11) 2.4-14 2.5' 1 CONTAINMENT OPENINGS 2.5-17 2.5-2 MASS-ENERGY RELEASE FOR MODE 5 LOSS
.OF SHUTDOWN' COOLING 2.5-17
- 2.5-3 MASS-ENERGY RELEASE FOR MODE 5 LOCA 2.5-18
-2.5-4' MASS-ENERGY. RELEASE FOR MODE 6 INVENTORY EOIL OFF 2.5-19 2.5-5 PROTECTION FACTORS AND EQUIVALENT MPC 2.5-19
- 2. 6-1: POSSIBLE FLOW PATHS OF NON-BORATED WATER' 2.6-5 2.6-2 RAPID BORON DILUTION ANALYSIS -
ASSUMPTIONS.AND INITIAL CONDITIONS 2.6-7 2.8-1 REDUCED INVENTORY INSTRUMENTATION PACKAGE 2.8-14 2.10-1 HYDRAULIC FLOW RESISTANCE DATA FOR
- l. SYSTEM 80+ 2.10-3 l
wigg .
770.wp(9212)bh'- LIST OF TABLES TABLE- TITLE PAGE 10. 2.12-1 FACTORS WHICH AFFECT THE RISK ASSOCIATED WITH AN INDICATOR 2.12-3 2.12-2 POTENTIAL-DRAIN PATHS DIRECTLY FROM THE REACTOR' COOLANT SYSTEM 2.12-5 2.12-3: GROUPING OF PRIMARY COOLANT DRAINAGE PATHS FROM THE SCS 2.12-9 I 3.1-1
SUMMARY
OF BRANCH POINT FAILURE RATES 3-2
,,1-2 FREQUENCY 0" CORE DAMAGE FROM 3-3 SHUTDOWN E\ 'S 3.1-3 COMPARISON OF CORE DAMAGE 3-3 FREQUENCIES 3.2-1 PLANT STATES AND INITIATING 3-8 EVENTS 3.3-1 OBSERVED FREQUENCIES OF DHR i EVENTS IN PWRs 3-11 3.3-2 INITIATING EVENT FREQUENCY FOR PLANT STATES 3-11 3.5-1 FRACTIONAL CORE RADIOACTIVITY FOLLOWING SHUTDOWN 3-58 - 3. 5-2 ~ FRACTIONAL DELAY OF SELECT RADIOACTIVE NUCLIDE GROUPS FOLLOWING SHUTDOWN 3-58 '4'.0-1 INITIAL CONDITIONS 4-2 .4.0-2 REACTOR VESSEL INLET COOLANT FLOW RATE 4-3 5 SYSTEM 80+ LOCA ANALYSIS, MODE 4 INITIAL CONDITIONS 5-11
- 6-1 ESFAS INSTRUMENTATION 6-5 6-2 CASES-ANALYZED 6-6 6-3 _ INITIAL CONDITIONS FOR LOCA INITIATED FROM ZERO POWER 6-7
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l 770.wp(9212)bh LIST OF TABLES TABLE TITLE PACE 10. 4 ACCIDENT CHRONOLOGY FOR LOCA INITIATED FROM ZERO POWER 6-8 6-5 INITIAL CONDITIONS FOR MSLB INITIATED-FROM MODE 5 6-9 , 6-6 ACCIDENT CHRONOLOGY FOR MSLB INITIATED FROM MODE 5 6-10 W h
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770,wp(9212)bh LIST OF FIGURES FIGURE SUBJECT PAGE 10.
-g 2.4-1 PLANT STATES AND TERMINATION POINTS FOR RESTORATION OF DHR 2.4-10 2.4-2 SAFETY INJECTION PIPING AND INSTRUMENTATION DIAGRAMS (SHEET 1 OF-3) 2.4-31 2.4-3 SAFETY INJECTAON PIPING AND INSTRUMENTATION DIAGRAMS (SHEET 2 OF 3) 2.4-32 2.4-4 SAFETY INJECTION PIPING AND INSTRUMENTATION DIAGRAMS (SHEET 3 OF 3) 2.4.33 2.4 SYSTEM 80+ ELECTRICAL DISTRIBUTION 2.4-34 2.4-6 DIESEL LOAD SEQUENCER - SIMPLIFIED LOGIC DIAGRAM 2.4-35 2.5-1A PERSONNEL LOCK; LEVEL 115 2.5-20 2.5-1B EQUIPMENT HATCH; LEVEL 146 2.5 2.5-2 EVENT TIMES FOR CONTAINMENT CLOSURE 2.5-22 2.5-3 LIMITS OF CONTAINMENT CLOSURE TIME 2.5-23 2.5-4A MODE 5 SDC, CONTAINMENT TEMP VS TIME 2.5-24 2.5-4B MODE 5-SDC, WHOLE BODY DOSE VS TIME 2.5-25 l
t 2.5-4C MODE 5 SDC, THYROID DOSE VS-TIME 2.5-26 L 2.5-4D MODE 5 SDC, MFC VS TIME 2.5-27 l ! 2.5-4E MODE 5 SDC, UTILITY PERSONNEL WHOLE 2.5-28
-BODY DOSE VS TIME l
2.5-5A MODE 5 LOCA, CONTAINMENT TEMPERATURE VS TIME 2.5 3.5-5B MODE 5 LOCA, WHOLE BODY DOSE VS TIME 2.5-30 2.5-5C MODE 5 LOCA, THYROID DOSE VS TIME 2.5-31 2.5-5D MODE 5 LO.5, MPC VS TIME 2.5-32
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770.wp(9212)bh L. LIST OF FIGURES TIGURE SUBJECT PAGE 10.
- 2. !f -5E MODE 5 LOCA, UTILITY PERSONNEL WHOLE BODY DOSE VS TIME 2.5-33 2.5-6A MODE 6 BOIL OFF, CONTAINMENT TEMP VS TIME 2.5-34 2.5-6B MODE 6 BOIL OFF, WHOLE BODY DOSE VS TIME 2.5-35 8
2.5-6C MODE 6 BOIL OFF, THYROID DOSE VS TIME 2.5-36 2.5-6D MODE 6 BOIL OFF, MPC VS TIME 2.5-37 2.5-6E MODE 6 BOIL OFF, UTILITY PERSONNEL WHOLE BODY DOSE VS TIME 2.5-38 1 2.7-1 NUCLEAR ISLAND FIRE BARRIER LOCATIONS, PLANT AT ELEVATION SO+0 2.7-6 " 2.7-2 NUCLEAR ISLAND FIRE BARRIER LOCATIONS, ' PLANT AT ELEVATION 70+0 2.7-7 2.8-1 REACTOR COOLANT SYSTEM ELEVATIONS RELATED TO SHUTDOWN COOLING OPERATIONS 2.8-18 2.,8-2 DIFFERENTIAL PRESSURE INSTRUMENT TAP LOCATIONS - SCHEMATIC 2.8-19 2.8-3 SCHEMATIC REPRESENTATION OF THE INADEQUATE CORE COOLING, HJTC PROBES 2.8-20 2.8-4 SCHEMATIC REPRESENTATION OF THE NARROW-RANGE HEATED JUNCTION
, THERMOCOUPLE PROSES 2.8-21 2.9-1 LOCATIONS OF SAFETY INJECTION SYSTEM 8 SUCTION IN IRWST 2.9-6 2.9-2 LOCATION OF WING WALL DEBRIS SCREEN ASSEMBLIES 2.9-7 2.9-3 LOCATIONS OF TRASH RACK AND SPILLWAY FOR IRWST AND HVT 2.9-8
'c 1
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=
I 770.wp(9212)bh ' LIST OF FIGURES FIGURE -SUBJECT PAGE 10. 2.10-1L RESULTS OF. MODE 6 NATURAL C3 'CULATION ANALYSIS OF SYSTEM 80+ 2.10-4 2.11-1 CONTAINMENT BUILDING LOAD HANDLING PATHS 2.11-5 2.11-2 FUEL HANDLING BUILDING LOAD HANDLING r P C IS 2.11-6 2.12 POTENTIAL RCS INTERFACING SYSTEM DRAINAGE PATHS FROM SCS TRAIN NO. 1 2.12-13 2.12-2 POTENTIAL RCS INTERFACING SYSTEM DRAINAGE PATHS FROM SIS /SCS DVI NOZZLE 1A 2.12.18 2.12-3 POTENTIAL RCS INTERFACING SYSTEM DRAINAGE PATHS FROM CVCS 2.12-20 2.12 -POTENTIAL RCS INTERFACING SYSTEM DRAINAGE PATHS FROM PSS 2.12-24 2.13-1 NUCLEAR ISLAND DETAILED ARRANGEMENT PLAN AT EL. 50+0 2.13-3 2.13-2 NUCLEAR ISLAND DETAILED ARRANGEMENT ,, PLAN AT EL. 70+0 2.13-4 3 . 4-l' EVENT TREE FOR LOSS OF DHR, MODE 4 3-38 , 3.4-2 OPERATOR FAILS TO RESTORE THE OPERATING TRAIN OF THE SHUTDOWN COOLING SYSTEM 3-39 3.4-3 OPERATOR FAILS TO START THE REDUNDANT TRAIN OF THE SHUTDOWN COOLING SYSTEM 3-40 3.4-4 OPERATOR FAILS TO MAKE USE OF THE STEAM GENERATOR" AS EMERGENCY DECAY HEAT REMOVA3 3-41 3.4-5 OPERATO.? r.4LS TO INITIATE FEED AND BLEED 'wvLING 3-42 3.4-6 OPERATOR FAILS TO INJECT INVENTORY FROM THE BORIC ACID STORAGE TANK TO THE RCS. 3-43
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*/70,wp(9212)bh '
LIST OF FIGUPES FIGURE SUBJECT PAGE lot ,
! . 4 * +7 EVENT THREE FOR LOSS OF DHR, MODE 5, 6 REFUELING CAN EMPTY 3-44 i
3.4-8 OPERATOR FAILS TO INITIATE FEED AND BLEED COOLING 3-45 3.4-9 EVENT TREE FOR LOSS OF DHR, MODE 5 REDUCED INVENTORY 3-46 3.4-10 OPERATOR FAILS TO INITIATE MAKEUP UTILIZING THE SHUTDOWN COOLING SYSTEM 3-47 3.4-11 EVENT TREE FOR LOSS OF DHR, MODE 6 REFUELING CAVITY FULL 3-48 3.4-12 EVENT TREE FOR LOCA, MODE 4 3-49 ; 3.4-13 OPERATOR FAILS TO CONTINUE TO MAKE USE OF STEAM GENERATORS AS EMERGENCY DECAY HEAT REMOVAL 3-50 3.4-14 OPERATOR FAILS TO INITIATL SAFETY DEPRESSURIZATION 3-51 3.4-15 EVENT TREE FOR LOCA MODES 5, 6 (IN CONTAINMENT LOCA) 3-52 3.4-16 OPERATOR FAILS TO MANUALLY INITIATE SAFETY INJECTION A7""ER THE L'VACUATION OF CONTAINMENT 3-53 3.4-17 EVENT TREE FOT. ' OCA, MODE 6 LOCA OUTSIDE CONTAluMENT 3-54 3.4-18 OPERATOR FAILS TO ISOLATE "'HE LEAK AT THE CONTAINMENT BOUNDARs 3-55 3.4-19 EVENT TRE1 FOR LOCA, MODE 6, LOCA IN CONTAINTMENT 3-56 4.0-1 INITIAL CONDITIONS FOR PRESSURE AND TEMPERATURE IN MODES 3 THROUGH 6 4-4 4:1-5 MAXIMUM POWER VS STEAM LINE BREAK AREA 4.1-8
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==- --
770.wp(9212)bh LIST OF FIGURES-FIGURE EUBJECT PNIE 10. 5-1 CIAD SURFACE TEMPERATURE, SAFETY INJECTION, INITIATED AT 10 MINUTES, DVI LINE BREAK 5-12 , 5-2 PRESSURIZER PRESSURE, NO SAFETY INJECTION, DVI LINE BREAK 5-13 5-3 REACTOR VESSEL MIXTURE ilEIGliT, NO SAYETY INJECTION, DVI LINE BREAK 5-14 b-4 CLAD SURFACE TEMPERATURE, NO SAFETY INJECTIONS, DVI LINE BREAK 5-15 5-5 REACTOR VI:;3SEL MIXTURE HEIGliT, THREE EREAK LOCATIONS 5-16 , 5-6 REACTOR VESSEL MIXTURE !!EIGliT, T11REE BREAK LOCATIONS COMPARED 5-17 5-7 COMPARISON OF HOT SPOT CLADDING TEMPERATURE, DVI LINE BREAK, WITH AND WITHOUT SAFETY INJECTION 5-18 6-1 CONTAINMENT PRESSURE VS. TIME FOR LOCA FROM ZERO POWER 6-11 6-2 CONTAINMENT ATHOSPilERE TEMPERATURE VS. TIME FOR LOCA FROM ZERO POWER 6-12 6-3 CONTAINMENT PRESSURE VS. TIME FOR MSLB FROM-MODE 5 6-13 6-4 CONTAINMENT ATMOSPHERE VS. TIME FOR MSLB FROM MODE 5 6-14
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e s k THIS PAGE IliTENTIONALLY BLANK ! I i 1 l l l t I i l I
770.wp(9212)bh i l l 1.0 HIRQHECILQ)] 1 1.1 PURPo1E This report presents features of the System 80+ design which address the issues of shutdown risk. It further evaluates these features with respect to their ability to reduce and/or mitigate ' the consequences of this risk. It fulfills the comn.itment made to the NRC by ABB-Combustion Engineering (ABB-CE) in Reference 1 to submit shutdown risk information in support of the System 80+ Design Certification. 1.2 f.C313 Sections 2.1 through 2.13 present detailed discussions on the specific shutdown issues. Following the detailed discussions of these shutdown risk issues, the report provides a probabilistic risk assessment in Section 3.0. This is followed in Sections 4.0, 5.0 and 6.0 by an evaluation of the applicability of the analyses in CESSAR-DC Chapters 6 and 15 to LOCA and accident events that are initiated from shotdown modes. Section 7.0 evaluates the features of System 80+ that simplify shutdown operations and thereby reduce the potential for initiating shutdown events. Conclusions of this report are provided in Section 8.0. The scope of the information presented was discussed with the NRC at a presentation by ABB-CE in Rockville, Maryland on December 18, 1991 and is outlined by the ABB-CE slides enclosed with the NRC minutes of the meeting in Reference 2.
- The report also addresses the RAI's from the NRC staf f on CESSAR-DC that pertain to shutdown risk. Appendix A of this report lists the RAI's and provides either the response or a referral to sections of the report which encompass the response to each RAI.
1.3 BACKGROUNR In Generic Letter No. 88-17 (Reference 4) the NRC issued recommendations to all holdera of licenses for PWR's to implement certain " expeditious actions" before operating their plants in a reduced inventory condition and to implement, as soon as practical,
" program enhancements" concerning operations during shutdown cooling. The objective was to prevent the reoccurrence of events that had occurred and that had the potential for core damage and/or release of radiation. In NUREG-14 4 9, NRC staff evaluations of shutdown operations indicate that recommendations have been l implemented and/or are underway at operatjng plants.
l 1.4 SYSTEM 80+ FEATUREH l In this section, a comparison is made between the characteristics of past events and the System 80+ design features. The categories 1 1-1 (
. - _ _ _ _ _ . _ _ _ _ _ _ __m____ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _
'77c.wp(9212)bh I i
of shutdown events at operating plants are those used by the NRC in , Chapter 2 of NUREG-1449 with little modification. These categorie.s are mostly the same as the issues identified by Secy-91-283 and presented by ABB-CE at the December 18, 1991 meeting with the NRC , (Reference 2). Each category encompasses a group of similar events ! that have in common the type of event initiator. Ultimately, if left unmitigated by automatic or manual actions, all events might t eventually lead to over heating and/or physical-damage to fuel with consequent radiation release, but each scenario sequence may differ. Depending upon the importance placed on eacn step in a ' sequence, the same events could be grouped differently. For example, the NUREG-1449 category, " Loss of Shutdown Cooling", includes the issues listed in Reference 2 as 1) Mid-Loop operation, !
- 2) Loss of Decay Heat Removal Capability and 3) Ef fect of PWR Upper Internals.
The categories employed in this section to group past events encompass (and for some categories are identical to) the issues which are listed by Reference 2 and which are presented in detail in this report with a fow exceptions. The exceptions apply _to postulated LOCA events initiated at high pressure and other significant events initiated at high pressure for which we do_not , have actual experience because they have not occurred in operating i plants. They exist only as analyses for use as guidance to avoid the physical event and therefore are not included in the categories ; of past avents. L Past events are grouped into the following ten categories: Loss of shutdown cooling Loss of electrical power Loss of reactor coolant Containment integrity Overpressurization Flooding and spills Boron and reactivity events Fire protection Heavy loads and fuel handling Mode change events ! For each past event placed into a category an initiator is identified. The plant design objective is to prevent the occurrence of the event initiator, but realistically, absolute prevention is impracticable and may be impossible. A combination of prevention and mitigatia, is employed in the System 80+ design. Table 1-1 provides an overview of the System 80+ features that avoid core damage during shutdown operating modes. It lists the ten shutdown event categories and for each category it lists event initiators for past events. These initiators are presented in a generic fashion; each initiator representing many cpecific events l 1-2
_ _ .. - . _ ~ _ _ _ _ _ _ - _ _ . _ _.-. _ _ .._. _ ._._.._ ____ 770 v '921'>)bh l lJ s it have occurred. For each initiator, the features of the System , t 40* iesign that are available to prevent occurrence of an initiator ' { a'.a/or to mitigate the consequences of an initiator are listed. , Table 1-2 providos a list of specific past events initiated from ! ohutdow.'. modes. Events were selected to include all ten event categories-and all typos of event initietors, but not all similar i significant events that have occurred. Soveral information sources were utilized to compile this list. They include events listed in 1 i NUREC-1449 which wore taken from the 1990 AEOD report (Reference 5) ' and which occurred mortly betwoon January 1988 and July 1990 with . some additional events. For events since July 1990, ABB-CE I searched the INpo database for LER's using a soloction of keywords portinout to the ton event categories and to shutdown operation. Various other INpo and NSAC documents were also reviewed for significant ovent reports dating from 1976 to 1990. Events in Table 1-2 are grouped into the ten categories given above. For each specific event, the features of the System 80+ design that apply to prevent and mitigate the event are indicated. A rev10w of this table serves as a design review of the System 80+ ! capabilities to avoid core damage and/or significant radiation i release during shutdown modes. The design features are discussed in more detail in the following Sections of this report. I i 1-3
77b(9132)da/1 TAFLE 1-J SHtfDOWW EVENT CATEC091ES AND SYSTEM 80+ FE4tt*ES FOR PeEVENTION. CETECTIOu sup WITICAf tow EVEwt CATECORY EVENT th!TIATOR SYSTE= 8J+ FEATUaES FOR setyEgTIou, DETECTION AND WITIGaf f04 1.) Loss of Shutdown Cooling SCS flow loss by p>Jmo suction vortex. A) Mid-loop level meAimited by locating SCS suction piping at the bottom of the hot teg. I
- 8) Mard piped venting for SCS pupps relieves gas binding more 1 quickly and conveniently. There are loop seats in the suction lines. ,
C) One SCS suction line from each het leg provides SCS j redtridancy with separation of pump suction sources. . D) Contairunent spray pumps inted. b2Le with SCS pupps I
, provide redundant capacity and ar=y take suction from IRWST to l ref ht t RCS a d to mitigate gas binding. l t
inaccurate mid-loop levet leadirg to E) With hemo on, reactor vesset levet monitoring system level ! suction vottex. Indications from vesset f*ead to a ?.evel below thet required for SCS operation. Levet indication is accurate for Intended t use. F) Core emit theriascoa.ptes monitor coolant tesperature down to 100*F prier to withdrawal of CETs prio? to fuel shuffling. ' The RTDs and SCS te w atures are accurate during SCS operation. I G) With head off, tevet indication near hot les elevation is provided by high resolution instruamts. N) SCS perforvierre ponitored on each of 2 SCS gumps by pisp petor current, flow rate, discharge pressure and suction i pressure. Possible SCS flow variance with occay heat to minimfre potentist for verteming during mid-loop. - Loss of flow white head of f, wper 1) Internals design limits coctant flew from cavity to core, internals in vessel and cavity however, high availability of SCS systwo and/or backw a i
- fIcoded teeds ta core hestup. assures forced convectlon.
Various low level and loss of ##R J) ton-shared Sts system allows SCS asintenance and testing ! events. dtring Modes 1-4 prior to cold shutdown, increasing evaltabitity in moder 5 and 6. K) Att SCS valves are motor operated, preventing faltures on tess of air if electro-pnetsistic operators were used. i r
- _ _ _ _ _ ___ - - - , >n,--. . - . - -
p -
'--c - . . . - - - . , - _
77b(9132)da/2 TA8t.E 1-1 (Continued) Sw!JTDouw EVENT CATEGMTES AWD SYSTE" 80+ FEsTus[S F0e Pa[y[mitow. OETECTios AC MITIGAf f 04 EVENT CATEGORY EVEst Im!TfATOR SYSTE" 50+ FE ATUeES FOR Pa[VE4f f 04. OETECTf 04 AC a[TfGAf f 04 Loss of Shutd:mri Cooling L) shutcbun specific controt roca displays, tech specs, and 1.) procedural guidance reduces likelihood of persomet errors. (Continued) m) Inadvertent errors are redwed ard early cperator evatuation of f ailures is irgroved Ly 1.) IPSU overview display with critical function and system status specific to shut h modes, 2.1 CRT displays with system linews and i-mm ; status and 3.) storms that are b e ; on plant mode and equi e t status. N) Prevention of ina@ropriate outcratic vtlons frue persorvwt errors by shutchun specific control logic (e.g., reiiove autoctosure interlocks from SCS suctior valves.) C) CCW availability is increased by 2 redsdart Divisions, each with two pm and heat enchangers. P) Seevice water availability is increased by 2 tedsdant Divisions, each with two pups. C) Each SCS Division has four potentist sources of AC power for increased avsiIabititv. Ewipnmt f aiture and/or inadvertent A) Alternate AC gas turbine provides thled on-site power ; 2.) Loss of Electric Power source. 4 personnet error leading to loss of power and shutdown cooling B) Two switchyard interfaces provide flemibility. C) Shutdawn spacific tech specs and procedral guidance redre likelihood of personnet errors. D) A reserve transformer provides an atternate sissAy to the safety bus if the noemet source twiit auxiliary transformer) is de-energized. E) Each safety division has a dedicated diesel generator. F) no equipnent is shared between diesels. G) No e w ipment is shared with another unit.
77bt9132)da/3 TA8tf T-1 (Continued) S$UTDOM EVENT CaTEG0o!ES aED STSTEm SO+ FEatvets Feet peEVEutt04. CETECTION AND #f f!CATICut EVEsT CATEGoof EVENT imiT1ATOR ST5tE9" 30+ FEAPJeEs FOR PeEVEeTION. O*'ECTION Am3 p!TIGATION 3.) t.oss of Reactor Cootent From shutdown morte, coulpwnt failure A) Inadvertent errors are reduced eruf certy operator evatustion and/or personnet error leads to toss of failures is improved by 1.) IPSC overview disp 4ey with of coolant, usual (y through systems critical fssiction and system status specific to shutdonet connected to RCS. modes 2.) CRT displays with systee lincies end i m ,4 status and 3.) alorses that are deperwirnt on plant mode end equippumt status. Inadvertent RPV pressurization =Alle 8) Removet of repressuciter menmey will not allow significent connected systems are open causing RW bead pressuritation. Thus, instruments are not af fected. coolant level drop in vesset. C) In-ccre instr e seat table evolutions are prohibited by procedurst guidance while vesset heed is on and mid-loop evolutions are in progrtss, preventing seet les&s. D) Coctant loss via RCP & ring seet maintenance reduced by puso ispetter weight creating sesl Cavitw draining esposes fuel baing E) Cavity draining limited by reinforced poot seet between transferred. vesset ftsage and cavity floor. F) Conteirvaent toyeut p* everts totet draining if seet faits. 4.) Conteirveent integrity loss of shutdown caoling and/or loss A) Tech spec requires hatch and all penetrations closed during cf reactor coolant results in core mid-loop evothtions. Conteirunent configuration and size boiling requiring rapid contairvaent attow more outage ottivities mittii t contairewnt, re=utting closure to prevent radiciogical in less time without containment integrity. release. E) Re&ndancy in SCS system, ef ectric power simply and support systems toge*bec with increased instrsamentatfort reduce liketihood of an initiating event progressing to bolting. Persorviet ectors result in opaming C) shutdown specific contret room displays, ted spacs, and pothways f rom containraent to procedural guidance reduce likelihood of peesonnet error *. etmosphere during shutdown evolutions. S.) overpressurization Insdvertent high pressure safety A) SCs system retlef vetves sized for ammisman safety injection injection octurtion at low tiquid flow. te v rature pressurires RCS and SCS system. 5.) o w essurization (Continued) s) RCs is vented through the pressu-frer menwey.
77b(9132)de/4 - TABLE 1-1 (Continued) ss"JTDOWN EVEWT CATECORIES Ae7 SYSTEM 8% FEATUtES FOR MEVEsitos. WTECTrow map of 71GATfou t J EVENT CATEGoeY EVEw? IstTIATOR $T3 Test 85 FEAft1rts FOR PREVEsTIos MTECTityt Ac ofTIGAfftpf C) ting forged reector vesset belttine and vesset metecial provide adfitionet asesin to pressurf ted therent eheck. 6-)
. F!ooding and spitts Urrsit etted cootent flow from cpened A) Inedvertent errors are reduceo and early opr*ator evetustion systems, typically conbined with of failures is Imroved by 1.) IPSO overview display with other inadvertent and/or poorly c*4tical function and system status specific to shutdone+
planned evolutions, floods essential modM 2.) Ctf displays with system linewpe 'ad caugsanent equipment. status and 3.) stores that are Gaa on plant made and j equipment status. a
- 8) Shutdown specific controt roam displays, tech specs, and procedural guidance reduce liketthood of persomet errys.
C) Plant terout, including separation of redsident dtvisions, limits damege that may occur to affected division. to commtestestion betmeen divisions, including piping, electrical, NVAC, floor drains, etc. 4 7.) Soron and Rasctivity Events various CVCS misoperations and A) Shutdown specific control roomtdisplays, tech specs, and tsicetitssted source rev neutron procedret guide ce renze tikelihood of lessreper operation. monitors cause approach to l criticality. CVCS misopeestion causes boron 8) Pr=cipitatien prevented by design that limits berm ' 4 dilution or potentist boron concentration to below cold piecipitetton concentration in precipitation. sest borated coolant lines, etiminating **ed foe suost heet
- tracing, i C) Boron diluties eterm provides advanced morning. l 1 8.) Fire prot =ction During shutdeun evoluttes, use of A) Plant toyout ord fire barriers seperate redsident divisions carbustible weterials pluz ignition and systems to limit potentist fire demPge.
sources such es tecesorary pcuee lines intresses potentist for fire damege ; to essentiei systems. i 8) 4 Codaustible materiets are limited in specific fire contret ( orees. I 9.) Mee r Leeds and Fuel Mandling Inedequate design end/or surveiltence A) shutdoun specific guidance timits pathueys for heavy lifts. I cf 1ifting dewices causes potentiet ' i damege to fuel or essentist , ewipnent. I i.
.t
77b(9132}Ge/5 i TABLE 1-1 (Contifund) T*N EVENT CATEGo*1ES 4i0 STsTE8 80+ FEATL*ES for P*EVEWTION. CETECTir4r sup #1TIGaitou EVEst CATEcce? EVEw? Iw!TIATOR SYSTEu 80* FEAft*E3 FtR PeEVEwTION. trETErTros Amo w*T!Gatitur
- 3) Plant arrow m nimizes potentist for deess'ngr drocs. .
d) Proven desien for fwet, core . am _ .; and fuet bendting machine min' mites potentist fuet drop. 10.) Pode Change Events operator and/or procedurst errors A) shutdows specific centret roas displeys, tech specs, and attow mode changes without satisfyin, pNscedtsrat guidance redxe liketthood of persomet errors. , estry requiremects.
- 8) Insdvertent errors are reduced and eer!y eperator evetustion of feitures is issroved by 1.) IPs0 emiew displey with criticet function and system status specific to shutdtwi moe s 2.) CRT disptsys with system tirmaps and cm ~ .;
, status eruf 3.) alerns t!*et are dependert en plant mode end equipment status. i k I 4
* - - % se
i. F k 5
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!r
^
R. .E. : _s-5#
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77c.wp(9212)bh 2.0 SHUTDOWN RTSK ISSUES Sections 2,1 through 2.13 present detailed evaluations of specific shutdown issues that were identified at the December 18, 1991 meeting with- the NRC and that are listed in Reference 2. Each section is subdivided into four subsections. The first subsection states the issue consistent with the interpretation and evaluation in NUREG-14 4 9 and the appropriate RAIs. The second subsection lists the acceptance criteria that are employed to evaluate the System 80+ design to prevent and/or mitigate unacceptable consequences related to each shutdown issue. The third subsection discusses the postulated plant scenarios, the analyses and the evaluations considered by to assure that the shutdown issue is adequately addressed. Finally, the fourth subsection states how the issue is resolved by the System 80+ design. Depending upon cach issue and its significance in evaluating System 80+, the content of these subsections varies. Where appropriate, reference is made to RAI's on the issue, both outstanding and previously submitted. Appendix A contains the responses to all the RAI's. l I i 1 l i 2-1
77a.wp(9212)bh 2.1 EB_QQEDURES 2.1.1 ISSUE The operational guidance provided by the plant designer to the owner / operator might not be suf ficient to insure that procedures to avoid, detect, mitigate, and/or recover from abnormal events initiated from shutdown operations can be developed by the plant owner / operator. 2.1.2 ACCEPTANCE CRITERIA The operational guidance provided by the plant designer to the owner /operatcr shall be sufficient to properly utilize design features that are available to detect, mitigate and assist recovery from abnormal events initiat3d during shutdown operations. 2.1.'3- DISCUSSION The System 80+ design incorporates advanced features which promote safer and simpler plant operation. The features include redundancy and diversity of components and systems, dedicated and/or permanently aligned systems, and an advanced informa, an / stem which better informs the operations staff of plant s t a tus. , potential adverse system interactions, and available recovery paths if an abnormal event occurs. These features also contribute to improved operability and maintainability that should significantly reduce the initiating situations that have contributed to increased shutdown risk. - The - plant owner / operator is responsible for preparing detailed procedures for normal, abnormal, and emergency operations using guidance developed by the plant designer and plant site specific information. The plant designer's guidance generally is in the form of suggested operational sequences that preserve the safety bases of the design. Since shutdown operations must be intimately connected to an outage strategy, specific _ procedures cannot be imposed by the plant designer to cover the array of possible ; shutdown events. However, the plant designer can provide guides which instruct the plant owner / operator in the use of design features which can detect, mitigate, and assist recovery from abnormal events that can occur during shutdown operations. The operational guidance contained in this report will be provided to -the plant owner / operator through the plant designer's Operational Support Information (OSI) program. The intent of the OSI program is to insure that features of the System 80+ design are effectively utilized in the operation of the plant as specified in the plant owner / operator's operations program. The OSI program also provides a formal means to transfer design related bases for operations, regulatory operational commitments and related 2.1-1
770.wp(9212)bh information that is typically provided to and required by a plant license applicant. The OSI program is staged to develop more detailed information as the plant design matures and is expected to be completed during the construction and pre-operational phases of a plant project. . The OSI program integrates information from various interrelated areas, e.g.,_ the Maintenance plan, the Reliability Assurance Program and as-procured equipment characteristics, to insure the owner / operator can efficiently operate-the plant within the design bases. A summary of general o p e r a t i o n a l'. g u i d a n c e related to shutdown operations is provided in Table 2.1-1. Specific details are contained in the appropriate sections of this report. An outline of the operational guidance developed to support RCS reduced inventory operations is provided in Appendix C. This guidance together with supporting information from this report is sufficient for a plant owner / operator to develop an operational guideline for reduced RCS inventory operations. The development of a detailed procedure by the plant ' owner / operator would require specific equipment characteristics of procured components - and results of the pre-operational testing (CESSAR DC Chapter 14) to determine system performancc values. As an example to support mid-loop operation, the shutdown coolant system flows, suction line vortexing characteristics, level instrumentation calibration, among others, would be measured during plant startup. This data would be used to verify performance as well as to provide operational data. Specific testing requirements for shutdown oriented instrumentation will be added to CESSAR DC Chapter 14 in an update amendment. 2.1.4 RESOLUTION The issue of procedures for shutdown operation is resolved for System 80+ by providing operational guidance to address use of advanced design features to detect, mitigatt, and assist recovery from abnormal events initiated from shutdown operations. l l l l l l-l 2.1-2
~
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l77a.wp(9212)bhY <
^
TABLE 2.1 ..
SUMMARY
OF PROCEDURAL GUIDANCE RELATED TO BRUTDOWN OPERATIONS-i
. TOPIC -PROCEDURAL GUIDANCE. REFERENCE TO REPORT SECTION:
1 Unplanned Draining of the Prevention 2.12.1, 2.12.2.1(3),12.12A3,"
' Reactor-Coolant , 2.12.3.2.1,.2.12.4 l Administratively Control Major' Potential'Draindown' Paths Identified'for Shutdown Modes LIdentification '
-Monitor Instrumentation for 2.3.3.1, 2.12.4, 2.8 RCS. Level, Inventory and- Table 2.8-1.
Temperature controls
- a. . Refueling _ pool level.
b; Containment'and subsphere-sump levels.
- c. Level indicators'and alarms: EDT,'RDT, IRWST,-
HVT,'VCT4
- d. RCS: operational leakage' ,
(Tech. Spec. Surveillance). e.. RCS' level indicators and. alarms. ~
- 1) Pressurizer' level
- 2) RV Delta-P: level instruments
- f. Pressurizer! pressure i
v 77a.wp(9212)bh' , 4 TABLE 2.1-1
SUMMARY
OF-PROCEDURAL GUIDANCE RELATED TO SHUTDOWN OPERATIONS-i TOPIC PROCEDURAL GUIDANCE REFERENCE TO REPORT'8ECTION'- Unplanned Draining of the Prevention 2.12.1, , 2.12. 2.1(3 ) , L 2.12.3, .. ' Reactor Coolant . 2.12.3.2.1, 2.12.4. 'i Administratively. Control ~ Major' Potential Draindown
~
Paths Identified for Shutdown Modes . 3' Identification ' t Monitor Instrumentation for 2.3.3.1, 2.12.4, 2.8 RCS Level, Inventory.and Table'2.8-1 . Temperature Controls t a.- Refueling pool level.- ,
- b. Containment and subsphere '
sump levels.
- c. Level indicators and alarms: EDT, RDT, IRWST, HVT,'VCT. '
- d. RCS operational leakage
. -(Tech. Spec.
Surveillance).
- e. RCS level indicators and-alarms.
1)' Pressurizer level
- 2) RV. Delta-P-level instruments 1
- f. Pressurizer pressure
. _ _ _ _ _ - . _ _ _ _ __ _.-._ _ _ _m. . _ . _ _ - - _ .
77a.wp(9212)bh information that is typically provided to and required by a plant license applicant. The OSI program is staged to develop more detailed information as the plant design matures and is expected to be completed during the construction and pre-operational phases of a plant project. The OSI program integrates information from various interrolated areas, e.g., the Maintenance Plan, the Reliability Assurance Program and as-procured equipment characteristics, to insure the owner / operator can efficiently operate the plant within the design bases. A summary of general operational guidance related to shutdown operations is provided in Table 2.1-1. Specific details are contained in the appropriate sections of this report. An outline of the operational guidance developed to support RCS reduced inventory operations is provided in Appendix C. This guidance together with supporting information from this report is sufficient for a plant owner / operator to develop an operational guideline for reduced RCS inventory operations. The development of a detailed procedure by the plant owner / operator would require specific equipment characteristics of procured components and results of the pre-operational testing (CESSAR DC Chapter 14) to determine system performance values. As an example to support mid-loop operation, the shutdown coolant system flows, suction line vortexing characteristics, level instrumentation calibration, among others, would be measured during plant startup. This data would be used to verify performance as well as to provide operational data. Specific testing requirements for shutdown oriented instrumentation will be added to CESSAR DC Chapter 14 in an update amendment. 2.1.4 RESOLUTION-The issue of procedures for shutdown operation is resolved for System 80+ by providing operational guidance to address use of advanced design features to detect, mitigate, and assist recovery from abnormal events initiated from shutdown operations. I 2.1-2
77a.wp(9212)bh ' ' TABLE 2.1-1.(Cont'd)
SUMMARY
OF PROCEDURAL GUIDANCE RELATED TO SHUTDOWN OPERATIONS TOPIC PROCEDURAL GUIDANCE REFERENCE TO REPORT SECTION . 1
. g. RCS temperature' 2.8, 2.3.3.1
- 1) Core Exit-
.Thermocouples~
- 2) Resistance . .
Temperature _ Detectors
- 3) Shutdown Cooling.
-a) SG Parameters ,
b) Shutdown ~ Cooling System. y MITIGATION (Immediate 2.3.3.4, 2.12.3, 2.12.4, Operator Action) 2.12.2.3(2) Appendix C ., Identify leakage path.
- Isolate leakage path. ,
Make up losses. a.- Safety Injection b.. SDC via IRWST Containment spray from "
~
c. IRWST viaSDC. lines . d.. Charging pumps . ! e . -- BAST !
- f. Safety Injection Tanks 3
l 1
-F
, , , ., -,.m _ , . . , -
, - - -- -- - m- , - -
L7.7a.wp (9212 ) bh - 3 TABLE 2.1-1.-(Cont'd)- BUMMARY'OF PROCEDURAL' GUIDANCE RELATED TO BHUTDOWN_ OPERATIONS TOPIC PROCEDURAL GUIDANCE' -REFERENCE'TO-REPORT SECTIOW
. Heavy Loads Restrictions specifled..fcr: 2.11.3
- 1) Drop of. transported a. ' Lift Height'
~
equipment. b.- Travel Directions
- 2) Drop of fuel bundle c. Systems Lineup.(Specified 3): Refueling pool' seal ~ in.CESSAR DC, Chapter 9l Integrity.. .
Lnd" Plant Designer's. , 4); Loads over ICI table.- " Heavy Load Guides") Outaae Maintenance. ' Strategy for Shutdown .2.4.3.2.'2 Operations
.a. Define operating and Appendix C l operational ~ divisions.
'b. Limit maintenance ,
activities to' components i , 'and systems not included I
-in a).
V
'h b
?
e
- w - w -
- , , - g , V y -' .m
77a.wp'(9212) bh;L i v t TABLE 2.1-1-(Cont'd) _
SUMMARY
OF PROCEDURAL ~ GUIDANCE RELATED TO SHUTDOWN OPERATIONS TOPIC ' PROCEDURAL GUIDAUCE.
. REFERENCE TO REPORT SECTION' Fire Protection , Administratively. require fire 2.7.3.2,'and 2.7;3.3 protection. systems to1 remain:
- operable in shutdown modes.-
. Procedurally _ Control::
- a. ' Colabustible materials s ,
'b.' Housekeeping-
- c. Ilot work ~
Pre-Fire' Plan. 2.7.3.2 a.- Outline fire fighting ; strategie's
'b. Monitor status of-fire barriers Y
i i A
, ,L, ,. _, , - - ,, c .,s - + , -
.77a.wp (9212 ) bh.:; ^
TABLE ' 2.'1-1.(Cont'd)
SUMMARY
OF' PROCEDURAL' GUIDANCE RELATED TO SHUTDOWN OPERATIONS TOPIC '; PROCEDURAL GUIDANCE REFERENCE TO REPORT SECTION-
'RCS' Cooling Using Feed..and 'RCS Pressurized
- Bleed (other systems not.
! available) ' 1. - .. Start SI-pump. .
- .2. 4 ; 3 .1. 3 .1.1
-2.
" Reduce pressure through. 2.4.3.1.3.2.1 Safety Depressurization)
- System '. (SDS)..,. venting to -
'IRWST.: . -(Maintain
..subcooled. temperatures.in'
' ECS) '.
- 3. Secure.operatingLRCPs'(if-applicable) 1
_4 . Cycle SI feed and SDS bleed.to reduce.RCS pressure and. temperature.
- 5. . When depressurized, open
'SDS-and Run SI-continuously.
- 6. Align SDC heat exchanger. "
for IRWST~ cooling. 4
- 7. -Restore Normal SDC systems.
77a'.wp (9212 ) bh ) - , t '
- TABLE 2.1-1L(Cont'd),
I
SUMMARY
OF PROCEDURAL GUIDANCE RELATED TO SHUTDOWN OPERATIONS' TOPIC [ PROCEDURAL' GUIDANCE. ~ BEFERENCE TO REPORT SECTION -
' 'RCS'Depressurized:
1., Start'SI
- 2. .Open.SDS. '.
Secure,RCP's:(if RCS.not.
.^
3.- . vented)- , 4. Align SDC heat' exchanger-
.for!IR1:ST Cooling.
- 5. Restore normalLSDC:-
. Systems-SG Tube' Rupture. Include-in Emergency' Table 2.6-1 Section'C(a).
Procedure Guides.a requirement to maintain a
-positive primary to secondary.
pressure > differential. Lockout of main feedwater ? Administratively lockout main 4.1.1 and 4.1.2 pumps in shutdown: modes ~ 'feedwater pumps if with RTCDs closed.- subcritical'.
+
b R { r, , - - , , - - e , , - , v ,., , ,
77a.wp(9212)bh-2.2 TECHNICAL SPIRFJCATION IMPROVEMENTS 2.2.1 ISSUE When a plant is operated within the limiting conditions for operation provided by the technical specifications, the consequences of design basis events should be bounded by the results of the safety analyses. However, limiting conditions for operation developed for power operation might not be sufficient to insure that the consequences of events initiated from shutdown modes are bounded by the analyses. Technical specification should [ include the necessary limiting conditions for operation that are applicable to shutdown modes. 2.2.2 ACCEPTANCE CRITERIA Technical specification shall insure that when the plant is operated within the limiting conditions for operation applicable to the mode of operation, consequences of design basis events shall be bounded by the results of safety analyses for that mode. 2.2.3 DISCUSSION The System 80+ design incorporates advanced features which promote safer operation and greater margins to operating limits. The features include redundancy and diversity of components and systems and an advanced information system which better informs the operations staff of plant status, potential adverse system interactions, and the recovery paths if an abnormal event occurs. These features also contribute to improved operability and maintainability that should significantly reduce the initiating situations that have contributed to increased shutdown risk. [ One objective of the plant designer is to reduce the operational constraints that limit the plant owner's flexibility to operate the plant as efficiently as possible. Another objective is to formally impose the operational constraints required to insure the plant remains within analyzed bounds for operation through the initial set of technical specifications. Overly restrictive technical specifications especially for shutdown modes may unnecessarily complicate operations and may increase risks by prolonging the shutdown period and adding to staff stress. The objective of this assessment of shutdown risk for the System 80+ relative to technical specifications is to modify existing technical specifications to the extent necessary to address event initiators not fully covered by analysis of the traditional design basis events. A summary of the proposed technical specification modifications is provided in Table 2.2-1, with a markup of the LCO and APPLICABILITY ; sections of the present technical specifications, as applicable, in 2.2-1
t 77a ._wp (9212 ) bh1 t
; Appendix D. -In. additioni a new -. technical: specification section applicableLtoHreduced RCS inventory operations in--Mode's 5 and 6
~
(with fuel in the= core) is presented in Appendix E. .
.The_ technical specification modifications-and additions reflect:
~
a.- the,added~ redundancy and diversity _of the System 80+: design that allows these modifications without affecting operational ~ flexibility; b.-. analysis of events initiat'ed during shutdown _ operations;.
- c. assessment of the risk .of operating in these plant-configurations for extended periods, e.g. , refueling, unplanned maintenance.
.=- - The ~ rationale for i the ' technical specification modifications ~ is
,. contained _-in the appropriate sections of this report. 2.2.4 RESOLUTION-
~
The issue of shutdown specific technical specifications _is resolved for_ System 80+ by: mc.difications L and additions to the technical-specifications- based : upon safety analyses performed for Modes 2 through-6. .These modifications and additions provide' additional-ensurance that:the consequences of transients and accidents which
- might. occur-during shut down modes-of operation are_less: limiting than those given in Chapter'.6Tand 15 of CESSAR-DC. The proposed
-technical'-specification modifications - and additions will- be reviewed against-_the shutdown.PRA--_ findings and a change package
;will:be: proposed for a future-CESSAR-DC amendment..
l i L g 1 l 2 . 2.-2
l
~ s l77a.wp(9d12)bh TABLE 2.2-1" SYSTEM 80+ TECHNICAL' SPECIFICATIONS. MODIFICATIONS
.RELATED TO IMPROVED SHUTDOWN OPERATIONS Tech- Number Type of Item ' Spec and Title' Chancel 'LCO Bases
- 1. 3.1.1 Shutdown' Revision- Change mode' 1 Extended applicability.
Margin (SDM) ' applicability $500*F
>210*F l 2. 3.1.2- SDM s 210*F- Replace Add Kn-1 and'ECP. : Provide protectionLfor ejected requirements. CEA and CEA group withdrawl1in-shutdown modes.
- 3. 3.1.10 SDM Test. _New Allow CEDMs Testing in Provide exceptions i to Test Exemption'for ' Modes'4.and.5- Operability of CEDES. . Movement CEDMs Testing of only one CEA-at 9t time is allowed.
- 4. 3.3.1 RPS . Revision Specify the. modes of Provide Reactor _ Trip function for; Instrumen- ' applicability in Table' . Steam Line Break (SLB).in tation 3.3.1-1. Extend SG ' Shutdown Modes.
Pressure-Low to Mode 3
- -and RC Plow-Low to Modes 3, 4, and 5 when the CEAs can be moved.
- 5. 3.3.5 - Core Revision Extend Operability to Provides. Reactor Trip Function Protection Modes 3, 4, and.5 when for unplanned CEA Group; Calculators CEAs can be moved. withdrawal.
__- 3 m ._ - m e- g --+m .' gr g. p gg - ..m
77a.wp(9212)bh TABLE 2.2-1 (Cont'd) BYSTEM 80+ TECHNICAL SPECIFICATIONS MODIFICATIONS RELATED TO IMPROVED SHUTDOWN OPERATIONS Tech Number Type of Item Spec and Title Chance LCO Boses
- 6. 3.3.10 ESFAS Instru- Revision Add Mode 4 to CSAS Insures availability of automatic mentation- Mode applicability in CSAS for pitigation of LOCA event Automatic Table 3.3.12-1 in Shutdown Mode 4.
Actuation
- 7. 3.4.11 RCS P/T Revision Add minimum Pressure Provide a SIAS for SLB and other Limits Restriction RCS increased heat removal events Temperatures between initiated in this temperature 483*F and 543*F. regime.
- 8. 3.5.3 LTOP Revision Change restriction on Two SI divisions required number of SI pumps operable in applicable modes.
cperable to 2.
- 9. 3.5.4 Safety Revision Extend requirements Required for RCS inventory makeup Injection for 2 SI divisions to for LOCA events in lower System all of Modes 4, 5, and operating modes.
6.
- 10. 3.5.4 IRWST Revision a. Extend operability a. For compatibility with 3.5.3.
requirements to Modes 5 and 6.
- b. Presently stated in Bases,
- b. Specify maximum water temperature gg, at 110*F.
- 77a.wp(9213)bhL
-TABLE 2.2-1 (Cont'd). ,
i SYSTEM'80+ TECHNICAL SPECIFICATIONS h0DIFICATIONS , RELATED TO: IMPROVED SHUTDOWN OPERATIONS Tech Number Type of Item Spec and. Title Chance LCO Bases
- 11. 3. 8. 2. AC Sources . Revision' Require 1 circuit Provide additional-backup AC-(Shutdown) between the offsite. power source.
transmission network to each onsite Class IE distribution system ; in Modes 5 and 6.
- 12. 3.3.15 Boron New. ~Both baron dilution Provide addf.tlonal protection'for.
Dilution' . alarms shall be- prevention of an~ inadvertent Alarm operable in: Modes 3, boron'dilutionLof the RCS. 4, 5, and 6..
- 13. 3.3.14 Accident Add Radiation . Required for SG tube rupture Monitoring Monitoring detection in shutdown modes.
Instrumen- Instrumentation for tation (AMI)
- a. -SG Liquid Blowdown
- b. Steam Line
- c. AirLEjectors
- d. Stack '
to Table.3.3.14-1
- 14. 3.1.6 Shutdown-CEA Add special test Clarify applicability and STE's.
' Insertion execp'<. ions and Limits- applicability to onlyL l critical conditions.
i
. ~- - . .
3
'G
~ 77a".wp ( 9212 ) bh :
N
..u.
+
h
-TABLE 2.2-1-(Cont'd);
SYSTEM 80+: TECHNICAL __ SPECIFICATIONS MODIFICATIONS
- -RELATED TO IMPROVED SHUTDOWN OPERATIONS Tech Number!., ; Type of Item Spec and Title Chance LCO Bases 15, 3.1.7 Regulating ,Same as Item 14. Sames as Item 14.
.CIUi Insertion .
Timits-
- 16. 3.3.12 'ESFAS Add Mode 4 to-CSAS Insures availability offmanual Instrumentati Mode' Applicability'in CSAS fon mitigation of LOCA' event on. Manual Table'3.3.12-l' in shutdcwn mode'4.
Actuation
- 17. '3.4.11 .LTOP . Delete requirement for!. 2 required-for Shutdovn,-see LCO' SI pumps. 2_5.3. . LTOP sizingLincreased to avoid PTS.
- 18. 3.8.5 DC Sources Clarify LCO to provide Prevents loss of operaille D/G due (Shutdown) most reliable line up. to reaintenance.
- 19. 3.8.8- Distribution Clarify LCO to provide ' Prevents loss of opera: ale' D/G due
-Systems mostLreliable line up. to maintenance.
} (Shutdowni
- 20. .3.9.4 Shutdown Require Additional SDC 1 Allows increased reliability for Cooling- division'to be decay heat renoval.
i -(Refueling operable. Operations) I i4
. , - . * * -, . . r
- - - . -- . .. - . -. - . . ~ .
770.wp(9212)bh 2.3 REDUCED INVENTORY OPERATION AND GL 88-17 FIXES 2.3.1 ISSUE I The NRC has voiced increasing concern over the safety of operations during plant shutdowns. Plant events which have occurred in the i industry have highlighted the need for 7 close examination of
. operations during reduced inventory conditions in the reactor coolant system. Following the Diablo Canyon incident, the NRC published Generic Letter 88-17, which required that holders of cperating licenses or construction permits address a number of deficiencies in order to enhance the safety of shutdown operations and reduce the risk to the public. Specific areas of concern include:
- 1. instrumentation which would greatly improve the operator's monitoring capability during reduced inventory operations,
- 2. the availability of existing equipment for use in mitigating a loss of SOS or loss of RCS inventory, l 3. nozzle dam installation procedures which would ensure a vent l
pathway is available so that RCS pressurization can be minimized if shutdown cooling is lost.
- 4. alternate ways to add inventory to keep the core covered should SCS be lost,
- l. 5. administrative procedures that would avoid RCS perturbations l
during reduced inventory operations, and
- 6. containment closure issues.
The NRC has specified that programmed enhancements should accomplish a comprehensive improvement in the plant's ar'lity to cope with shutdown operations. The NRC asserted that plants are not well designed for reduced inventory operations, that procedures are incomplete for shutdown cooling recovery or alternate actions-and that mitigating features may not be available under shutdown conditions. Therefore, the NRC has recommended that licensees implement means to prevent accident initiation, to monitor a progression that may lead to core damage and to evaluate consequences and, where needed, to provida-mitigation. 2.3.2 ACCEPTANCE CRITERIA The System 80+ design shall reflect a comprehensive consideration of shutdown and lower power risk, by adequately addressing all GL 88-17 recommendations and other issues relevant to reduced j inventory, especially in the areas of instrumentation, technical specifications, procedurec, equipment availability and analyses. 2.3-1 gt.
77a.wp(9212)bh 2.3.3 DISCUSSION During plant shutdowns, certain maintenance and testing activities require a draindown of the RCS to a partially filled condition. : Normal maintenance activities include the replacement of RCP seals and journal bearings. A testing activity requiring RCS-draindown is the Technical Specification for inservice inspection of the steam generator tubes. The use of nozzle dams during maintanance and testing activities minimizes the time during which the rcd must be operated in a partially filled condition. To minimize operating time at mid-loop level, nozzle dans are installed on the st-sm generators and the RCS is reflooded to continue maintenance ad testing.
~
While the RCS coolant level is lowered to within the hot leg, the risk of loosing shutdown cooling is increased due to the possibility of_ vortexing at the SCS suction line interface with the hot leg. In the worst scenario, subsequent to vortexing in the SCS quction line, a large percentage of air is entrained into the SCS sustion piping and the SCS pump performance is degraded or r Jerrupted. If SCS operation is not reestablished, core boiling and pressurization can produce very rapid core uncovery, sometimes in as little as 15 to 20 minutes. This phenomenon, and the high probability of it occurring, prompted the NRC to issue the recommendations of GL 88-17. System 80+ design features result in practical and significant benefits during reduced inventory operations. These design features are outlined in Sections 2.3.3.1 through 2.3.3.5 which follow. Details of the capabilities of these System 80+ design features to enhance safety during reduced inventory conditions and ' of the analytical bases for changes to Technical Specifications and , procedure guidance to the owner / operator are presented in other sections of this report. 2.3.3.1 Instrumentation for Bhutdown operations Diverse, accurate, and redundant instrumentation (including control room CRT displays) give continuous system status and provide the operations staff with precise information- to monitor reduced inventory operations and to respond to-loss of shutdown cooling events, should they_ occur. Detailed information on reduced inventory instrumentation is included in Section 2.8 of this # Report. Analyses form the basis for instrument design and calibration so as to assure correct _ instrument operability during reduced inventory states. Phenomena which can affect instrumentation operation are considered in the recommended use of instrument types for various scenarios. Instrumentation availability _ during reduced inventory is assured via the plant Technical Specifications that are provided in Section 2.2 of this report. 2.3-2 k
I 77a.wp(9212)bh
)
l 1 1 A general description of the types of instrumentation which are used for-reduced inventory is outlined below.
- 1. Redundant'and_ independent wide and narrow range level sensors '
are provided _ for continuous monitoring of RCS level during draindown operations. The level indicators provide monitoring l capability from the pre-drain down normal level in the pressurizer to a point -lower than that required for SCS ' operation. The level indicators are calibrated for low temperature operation and they provide a high degree of accuracy. Indication in the main control room and low-low, low, high and high-high level alarms are provided. The wide range level instruments cover draining from the , pressurizer to below the bottom of the hot leg and are available with the head on and off the vessel. The narrow range level instruments cover reduced inventory operations, and are also available-with head on and off the vessel. The narrow range instruments are accurate for measuring level within the hot leg. During a draindown, level monitoring would be transitioned from the wide range level instruments to the narrow range instruments when the greatest degree of accuracy _ is required during operations with level within the hot leg region.
- 2. Several independent diverse temperature measurements representative of core exit temperature are provided during reduced inventory operations. Temperature indication is available when the head is located both on and off the vessel.
Since temperature is valuable in guiding SCS restoration actions and in monitoring the success of recovery actions, alarm setpoints are based on integrated response times necessary to support SCS recovery, event mitigation, time to boil, and containment closure.
- 3. SCS operation monitoring instrumentation is provided that assures precise knowledge of the status of the operating SCS-loop; including pressure, temperature, flow and pump performance indicators.
2.3.3.2 SCS Desion The functional design of the Shutdown Cooling System (SCS) is substantially complete for System 80+. Design features that improve SCS performance during- shutdown operation are detailed below.
- 1. The System 80+ SCS suction lines do not contain any loop seals.
An improved suction piping layout allows self venting. 2.3-3
778.wp(9212)bh Entrained air _ travels back up to the hot leg without the possibility of being trapped anywhere in the SCS suction line. This feature allows the SCS pumps to be restarted without requiring complicated venting procedures, assuring an expedited reflood of the sh"tdown cooling pump saction.
- 2. The-two SCS . suction lines are independent and redundant to each other. Problems associated with a specific _suctio.. line would not limit the other shutdown cooling train from being operated, after level recovery (if necessary), for continued decay heat removal.
- 3. The two containment spray system pumps are interchangeable with the SCS pumps and are riesigned to back up the SCS pumps in the event of a non-electrical pump failure. Thus, there are four pumps available for shutdown cooling, provided support systems are available. Plant Technical Specifications will assure pump availability during shutdown operations.
- 4. There are no interlocks on the shutdown cooling suction piping which have the _ potential for disturbing shutdown cooling.
Although previous designs (e.g. , System 80) included interlocks to isolate the SCS in the event of an unanticipated RCS pressurization during shutdown cooling, this interlock has been
-deleted from the System 80+ design per the EPRI ALWR Utility Requirements Document. This reasonably reduces the likelihood of losses of SCS.
2.3.3.3 Steam Generator Nozzle Dam Intecrity The System 80+ design addresses the NRC concern for preventing significant pressurization in the upper plenum of the reactor vessel during core - boiling scenarios. System 80+ procedural guidance recommend, a nozzle dam installation and removal sequence by which the nozzle dams-are installed in the cold legs first. After the cold leg nozzle dams are installed, the hot leg nozzle dams can be-installed. Likewise, when removing the nozzle dams, the hot leg nozzle dams are removed first, subsequently cold leg nozzle dams can be removed. This installation and removal procedure _ will maximize the _ time that the steam generators are available for reflux boiling in the case of a loss of decay heat removal, and minimize the time that both hot legs- are simultaneously blocked by nozzle dams. This installation procedure requiree that the pressurizer manway is opened so that a hot side vent pathway exists prior to blocking both RCS hot legs with nozzle dams. In the _ System 80+ design, the ability of the RCS to withstand pressurization during reduced inventory operations with the nozzle dams installed is ' limited by the design pressure of the steam generator hot and cold leg nozzle dams. Based on field hydrostatic 2.3-4
770.wp(9212)bh tests performed on nozzle dams, a conservative value of 40 psia is assumed for this pressure limit. In order to assure that the nozzle dan design pressure is not exceeded during reduced inventory operations with boiling conditions in the reactor vessel, the system 80+ design includes a requirement will be imposed to establish a mid loop vent pathway via the pressurizer manway before operating in reduced inventory. When the manway is opened to the containment atmosphere, it provides suf ficient venting capacity to prevent RCS pressurization and subsequent nozzle dam failure. The pressurizer manway vent pathway is of sufficient capacity to prevent core uncovery due solely to pressurization of the hot side resulting from boiling in the core coolant. The pressurizer manway will be closed except during normal RCS draindown activities (see Section 2.1) . During a normal draindown to reduced inventory operations, when pressurizer level decreases to a preestablished setpoint, the RCS vent pathway will be aligned by opening the pressurizer manway. Following refueling operations, RCS integrity will not be reestablished until the RCS coolant level reaches the pressurizer. Only at that point will the manway be reinstalled. This mid loop vent alignment allows sufficient venting of the RCS to the pressurizer cubicle should SCS-be lost, resulting in onset of core boiling. Analyses have indicated that the pressurizer manway opened and relieving to the pressurizer cubicle will be sufficient for venting the RCS during RCS boiling and preventing steam generator nozzle dam failure. An acceptable, conservative RCS equilibrium pressure which is below the assumed steam generator nozzle dam design pressure has been calculated to occur 4 days post shutdown. Therefore, the earliest time after shutdown (from full power) for operating at mid loop level is recommended as 4 days. Based on industry operational data, a reasonable minimum RCS cooldown from Mode 1, followed by a draindown from normal RCS level to mid loop, can be-performed-in approximately 4.5-days. Therefore, the 4 day requirement does not impact the achievable start time for nozzle dam installation. -Additionally, it provides the necessary degree of-protection required for loss of decay heat removal scenarios. This data _ is to be incorporated into guidance - for the owner / operator _ to employ when planning' outage evolutions. Additionally, procedural guidance regarding the earliest time after
-shutdown for entry to reduced _ inventory operations is provided in Section 2.2 of this report. Such restrictions are-implemented to minimize _ the consequences of a loss of shutdown cooling event during reduced inventory operations.
The specified time after shutdown is based on the following analytical results: decay heat vs. time after shutdown 2.3-5
-77a.wp(9312)bh the resultant RCS heat up rate assuming a total loss of decay heat removal the consequential maximum RCS steam pressure for Mode 5, reached by boiling RCS inventory.
Power history (unit specific, cycle specific) decay heat loada vs. time plots could be used to calculate lower decay heat loads at 4 days, thus allowing entry into mid loop operation and nozzle dam l installation earlier during shutdown. In fact, the use of the pressurizer manway as the vent pathway yields an acceptable equilibrium pressure (below steam generator . nozzle dam design pressure) at 2 days post shutdown. This capability would be further supported by higher nozzle dam pressure limits than assumed in the analysis. Although this meets EPRI ALWR requirements for l refueling scheduling, it is not recommended that the plant enter ! mid loop level operations before 4 days. Complying with procedural guidance which requires mid loop operation no earlier than 4 days adds margin to the relationship between RCS pressure and nozzle dan design pressure. Furthermore, the entry time for mid loop level reasonably limits the makeup flow required to match boil-off to within several makeap schemes available in the System l 80+ design (see Section 2.3.3.4). It fixes the time to boil (assuming an initial RCS temperature of 150 deg F) to greater than 15 minutes, which lengthens the time available for loss of shutdown cooling system (SCS) mitigation actions. 2.3.3.4 Alternate Inventory Additions and DHR Methods The effective management of time end efforts is crucial to coping
-with a loss of shutdswn cooling. Awareness of time constraints provides information that is:useful in deciding how to allocate effort. -If shutdown cooling cannot be restored within the time to core uncovery, getting a source of water lined up to keep the core covered becomes a first priority. Inventory makeup directly extends the margin of safety prior to uncovering the core.
Successful coping with a loss of shutdown cooling would include performing the steps outlined in Section 2.4 of this report. One of the last measures specified in that section includes adding makeup __to the RCS to replenish boil off. This is thought of as a last resort measure in the Sy:= tem 80+ design due to the multiple success paths available to restore decay heat removal, and the time available to take corrective actions. However, as required by Generic Letter 88-17, sufficient existing equipment should be maintained in an operable or available status so as to mitigate a loss of RCS inventory should core boiling or an uncontrolled and significant inventory loss occur. Generic Letter 88-17 also recommends that the water addition rate capable of being provided by each of the means should be at least suf ficient to keep the . ore covered. Finally, Generic Letter 88-17 states that the 2.3-6
77a.wp(9212)bh path of water addition must assure that iakeup flow does not bypass the reactor vessel before exiting any opening in RCS. For the System 80+ design, at least 2 available means of adding inventory to the RCS will be available whenever the RCS is in a reduced inventory condition. Operating guidance is provided to specify the source of makeup water, the means of providing makeup to the RCS, and the recommended implementation strategy. The guidance will designate makeup pathways that ensure that makeup water does not bypass the reactor vessel. The water addition rate capable of being provided should be at least sufficient to makeup for the boil off rate. This keeps the core covered and provides an adequate degree of protection for loss of decay heat removal scenarios. With the earliest nozzle dam installation occurring at 4 days post trip, the decay heat present would require approximately 135 gpm of makeup flow to compensate for boil off. This value is based on the pressurizer manway being opened, venting steam to the pressurizer cubicle. The steaming rate, and therefore, the required makeup rate, w311 reduce beyond 4 days post trip. The exact makeup rate may require adjustment based on actual, as-built conditions. For Mode 5, reduced inventory operations, a shutdown cooling pump, the containment spray pumps, or safety injection pumps will be utilized as described in Section 2.4 to provide pumped makeup (for boil off) should SCS be lost. Procedural guidance will caution operators on using a containment spray pump in the same loop as the affected shutdown cooling pump, especially if the shutdown cooling pump has been lost due to air entrainment/ pump cavitation. The makeup pump (shutdown cooling, containment spray or safety injection)_ will be aligned to the in-containment refueling water storage tank .(IRWST) as the preferred source for makeup. An alternate source for borated vater is the boric acid storage tank (BAST). For Modes 5 and 6 the charging system via a charging pump (or alternatively, a boric tcid makeup pump) will be utilized to provide pumped makeup should all methods of decay heat and inventory replenishment delineated above be lost. The pump chosen will be aligned to the BAST. If no method of pumped inventory addition is available, a source for gravity feed inventory addition is via the-safety injection tanks (SITS). This is applicabic in Modes 5 and 6, and is considered only as a last resort. This method of gravity feed will only be implemented if SCS is lost along with all other means of supplying water to the RCS, and RCS boiling is occurring. If at least 2 tanks available provides assurance that over 20,000 gallons of water are available for discharge. Inventory addition from 2 2.3-7 ,
1 C' 77e.wp(9212)bh l l l SITS would assure approximately a 5 1/2 ft, rise in RCS water level and would make up for approximately 3 hours of core boil off. Use.of the- shutdown cooling pumps, containment spray pumps, or safety injection pumps ensures that makeup flow will not bypass the core regardless of postulated openings in the RCS. Flow delivery will be through the direct vessel injection (DVI) nozzles. Even in the case of -a DVI line break, sufficient inventory will be added to make up for core boil off. Use of the charging pump or boric acid makeup pump also ensures that makeup flow will not bypass the core, since it will be injected via the cold leg and enter the core via the normal charging path. 2.3.3.5 Operations Procedures and Technical Specifications necessary to support the program are identified and will be implemer,ted into the plant design. Procedural guidance for the conduct of mid-loop draindowns is provided to assure -that no testing or maintenance activity adversely affects the NSSS during mid-loop operations. Guidance will be provided that assures that testing and maintenance
-activities performed during reduced inventory avoid operations that deliberately lead to-perturbations An the RCS and all supporting systems necessary to maintain the RCS in'a stable condition. These operations include (but are not limited to):
RCS drain operations shutdown cooling testing and maintainance activities reactor coolant gas vent system testing and maintenance component cooling water testing and maintenance withdrawal of the incore instrumentation for refueling safety injection system testing and maintenance personnel communications system perturbations in-core instrument seal table evolutions while the reactor vessel head is on and mid-loop operations are in progress. Avoiding RCS and support system perturbations assures that adequate operating, operable, and/or available equipment of high reliability is provided for cooling the RCS and for avoiding a loss of RCS cooling. These actions also maintain suf ficient existing equipment in an operable or available status so as to mitigate a loss of SCS
.or a loss of RCS inventory, should either occur. Adequate communications are essential to activities related to the RCS or systems necessary to- maintain the RCS in a stable, controlled condition.
Due to the Diablo Canyon incident and other industry events, the requirements for evacuating personnel from the containment building, closing of the containment building e ment hatch and containment air lock doors, and isolation of pc stions looding outside containment were evaluated based on ti boil and time L 2.3-8
770.vp(9212)bh to core uncovery criteria. A description of the containment closure conditions referred to, alcng with a description of containmer.t closure design features, is contained in Section 6.0 of this report. 2.3.4 RESOLUTION f The resolution of the re.duced inventory issue on Systwa 80+ is complete. Resolut1on is comprised of the results of the analyses outlined above, related evaluations in Section 2.4 on availability of decay heat removal, Technical Specifications in Section 2.2 anc. procedural guidance in Section 2.1. The System 80+ design reflects a comprehensive consideration of shutdown and low power risk by adequately addressing all Generic Letter 88-17 recommendations and other issues relevant to reduced inventory. t l l l 2.3-9
___. _ . ~ __ _ __ - _ _ _ _ . _ . . _ _ __._ _ ____ _ __ _ _ __--___ _ . _ .__ _ 77a.wp(9212)bh 2.4 LOSJ fLDfiQAL1[DT REMOVAL CAPADILIf] , 2.4.1 ISSUE Events that have occurred at operating plants der nnstrate the vulnerability during shutdown Modos to loss of decay heat removal. The variety of maintenance activities taking place at shutdown ' combined with the possible system and equipment interactions that may occur lead to many conceivablo scenarios for experiencing a loss of decay heat removal. Three dominant design objectives have evolved from the emphasis placed on prevention of shutdown events: t
- 1. provide reduadant Shutdown Cooling System capacity and idantify alternate decay heat removal capability. ,
- 2. provide instrumentation to effectively monitor shutdown .
operations, including critical plant configurations such as mid-loop.
- 3. provide flexible redundancy in AC power. l The System 80+ features that address these issues are presented :
below in the context of demonstrating an integrated design capable of avoiding unacceptable consequences from the entire spectrum of potential event scenarios. 2.4.2 ACCEPTANCE CRITERIA All e.ent scenarios may be characterized by initiation, detection, mitigation and consequence. To measure the success of the integrated response of System 80+ to events initiated from Modes 2 through 6, two criteria related to the potential for radiological release are adopted here. Significant release can only occur from fuel cladding rupture resulting from heatup after the coolant level drops below the top of the active core. Therefore, the first acceptance criterion is that there shall be no fuel cladding f ailure resulting from postulated events, excluding LOCA, initiated from Modes 2 through 6. The second critorion is that the radiological exposure of the public to events resulting in the loss of decay heat removal- shall be limited to a fraction of the 10CFR100 limits that is specified in sections of this report where applicable. 2.4.3 DISCUSSION In this section, an evaluation is presented of the System 80+ features that are designed to prevent violation of the above criterion. Section 2.4.3.1 examines events and event initiatorc which potentially result in the losc of shutdown cooling leading to boiling. Causes of past events considered include mid-loop opera
- ion, power failure and operator error. Appropriate Technical 2.4-1
77a.wp(9212)bh Specification limitations and procedural guidance are identified by the analyses and ara provided in Sections 2.1 and 2.2. Section 2. 4. 3. 2 presents the features of System 80+ which help prevent a loss of decay heat removal due to the loss of AC power. This is one of the specific concerns identifi>*d in NUREG-1410. The discussion in this section is directly related to the means of coping with a loss of decay heat remeval. This concern is also identified in NUREG-1410 and evaluated in Soction 2.4.3.1. Section 2.4.3.3 presents the features of System 80+ that help assure the availability of the diesel gonarator. This issue was also identified in NUREG-1410. Availability of the diesel generater has been a significant factor in numerous past events. Taken together, these sections demonstrate the integrated capability of the System 80+ to prevent and mitigato a loss of decay heat removal to ensure that the acceptance criteria are not i violated. 2.4.3.1 ghu_tdown Ky_gnt IniMpMon_aAULn.alyAqa
" 4.3.1.1 Introduction This section examines events which could result in a loss of the Shutdown Cooling Systein (SCS) due to various initiators (events which challenge the SCS, such as a loss of power, inadvertent closure of a valve in the pump suction line and air ingestion in the pump suction) under various plant configurations and modes of operation and the ways these events can be prevented, detected and mitigated.
The discussion is structured into three parts. The first part focuses on design features which improve the SCS's resistance to initiators. The emphasl.s here is on hardware design. The second part assumes a loss of the SCS, regardless of the initiator, and discusses the ability of System 80+ to recover from the event. ' Here too, the emphasis is primarily on hardware design. The third cart recognizes the limitations of hardware design as a response to initiatort and the need to demonstrate that adequate redundancy is provided m cover all possible plant configurations. This will include the plant's ability to cope with a loss of DHR. The emphasis here is on operator actions, operating procedures and technical specifications in the context of the various plant configurations which can exist in Modes 4, 5 and 6. 2 2.4-2
77a.wp(9212)bh 2.4.3.1.2 Resistance to Initiators Design improvements have boon made to the System 80+ SCS that reduce the likelihood for a loss of DHR. This is, in part, the result of applying a "beyond singic failure criteria" design philosophy to improve the SCS's ability to withstand a wide range of initiators, including a loss of power, equipment failure, control system failure and operator error. The major design features attributed to the SCS's increased resistance are summarized in Table 2.4-1. The SCS Piping and Instrumentation Diagrams are shown in Figure 2.4-2 through 2.4-4. -n._,-, TABLE 2.4-1 E.UMMARY__OF SYSTEM 80+ B_QJ_DISIGN FEAIUREE THAT INCREASE RESISTANCE AGAINST INITIATORS SCS Nozzle at bottom of hot leg. Dedicated DHR function. Independent suction lines for each train. Elimination of auto-closure interlocks in suction valves. Elimination of cross train communication. Increased system design pressure. Improved flow control. Improved protection against pump excessive flow conditions. Flexibility to reduce flowrates to maximize NPSHA. Improved RCS level instrumentation at mid-loop. Instrumentation to indicate incipient pump cavitation. Elimination of loop seals in suction lines. Improved AC power reliability. The most important feature that was added for the System 80+ SCS is its dedication to the DHR function. No portions of the SCS are included in the Emergency Core Cooling System (ECCS) as has been done with past designs. This means the SCS components do not double for components credited in other safety systems during Modes 1 through 4. This single design change allows various SCS improvements and simplifications including the ability to perform routine maintenance outside of shutdown cooling modes (5 and 6). The system is comprised of two !dentical, redundant, and totally separate trains each capable of performing the required DHR function. Dedicated heat exchangers have been provided in each SCS train. Previous designs used.a single heat exchanger for both the SCS and the conta e nment spray system (CSS), and as a result this required system realignment as the plant moved through modes 4 to
- 5. However, the System 80+ design eliminates these manual actions 2.4-3
i 77a.wp(9212)bh l l t which decreast the potential for operator error during Modes 5 and i G. Each SCS train has independent auction lines irom the RCS hot-legs. [ There are no cross-connections between SCS trains. Direct vessel : injection (DVI) introduced for System 80+ and the dedicated DilR ! function of the SCS have snabled each train to be separate. This i al owed simplifications in the arrangements resulting in greater i protection for each pump from suction lino failures due to air ingestion and discharge line failurns reculting from pump to pump 1 interaction. l Interlocks are provided on the SCS suction isolation valves to i prevent theno valves from being opened when RCS pressure is above the SCS entry pressure. Those interlocks are enabled when RCS l pressure is slightly above the shutdown cooling initiation pressure. All interlocks to close these valves, such as closure on i high pressure, have been eliminated. During an overpressurization ! transient, the SCS will he maintained active to continue DHR. The i SCS is designed to -mitigate these events with low temperature i overpressure protection (LTOP) using spring-loaded relief valves ; and by an increased SCS design pressure. ; System flow control and protection from pump overspeed has been ' improved. System flow control is accomplished using valving and fixed resistance orifices in each train (orifices are not shown in Figure 2.4-1). The .orificca are sized to limit the maximum flowrate from the SCS pumps and adjustmentn to shutdown cooling ' flowrate to match decay heat levels is accomplished by modulating valves. This design philosophy not only minimizes seat wear due to high fluid velocitios resulting -from throttling, but also prevents , pump excessive flow conditions, operating procedures for the SCS during reduced inventory operation provide minimum flowrates necessary to perform DHR as a function of time af ter shutdown. SCS flowrate will be decreased as the cooling requirements decrease from lower decay heat levels. The lower shutdown cooling flowrate increases the not positive suction head available (NPSHA) to the SCS pumps. This provides greater operational margin for the RCS during midloop when SCS NPSHA is at , a minimum and the potential for cavitation is at a maximum. ' The ability to' accurately measure and provide the RCS fluid levels has been the cause of many incidents resulting in the loss of DHR.
. system 80+-has made many improvements in the instrumentation for measuring- the liquid level in the RCS and data display to - the operator in the control room. Further discussion on this topic appears in sections 2.3 and 2.8 of this report.
Improvements have been made in the instrumentation of the SCS to provide the operator with more information about critical points in 2,4-4
=
77c.wp(9212)bh # the system. The intent is to provide the operator with detailed system parameters so appropriate actions can be taken bo.3re the loss of DilR occurs. If a loss of the SCS does occur these parameters will aid in the correct and timely evaluation of the initiator thus decreasing SCS recovery time. Major new instruments which have becn included in the System 80+ design are auction and discharge pressure indicators and SCS pump motor current indication. These instruments are all indicated in the main control room. Suction piping arrangements have been sirnplitled and improved. Several incidents have been attributed to the presence of loop seals in the suction lines that allow air to collect and lead to the reduction of !!PSHA and air binding. System 80+ arrangements for the suction lines do not have loop seals and thereby nnhance the ability of the pumps tc survive low NPSHA conditions. Improvements in AC power reliability are discussed in sub-section 2.4.3.2. The System 80+ SCS design features presented above, and summarized in Table 2. 4-1, provide a way to minimize a loss of the SCS. These design features also address initiators which are known to have defeated Di!R systems in currently operating plants. A summary of these initiators and corresponding SCS deign features are provided in Tables 1-1 and 1-2. 2.4.3.1.3 Recovery from Initiators Recognizing that some initiators may defeat the SCS, the System 80+ design will require that both SCS trains and one division of AC power be operable during Modes 5 and 6. This allows safety injection and containment spray equipment in the redundant division to undergo maintenance activities as necessary. Table 1-2 provides a detailed listing of events that have resulted in the loss of shutdown cooling. Table 1-1 summarizes design features incorporated into System 80+ to prevent, detect and mitigate the effects of the events listed in Table 1-2. Consequently, a detailed listing of all potential initiators will not be provided in this section. Instead, initiators that result in the loss of DHR are categorized into four groups. This categorization is structured primarily to simplify the discussion but may also aid in constructing diagnostic loss of SCS procedures. These groups relate the initiator to a location in the system with respect to the SC pump. The instrumentation provided for monitoring the pump's performance identify whether the failure is in the suction line, discharge line, the pump itself, or a power failure. With proper diagnostic information from these groups, the operator can perform appropriate recovery actions to restore DHR. Table 2.4-2 identifies the groups, some representative initiators 2.4-5
__ --_ _ _ ._ _ _ _ _ _ _ _ _ _>_.... ___ _._ ~._ _.._ _ ._ _ 77a.wp(9313)bh i in cach group, a brief description of the event and the I instrumentation available to detect the event. The discussion that follows examinos how DHR can be recovered using this information assuming a loss of a SCS train. 2.4.3.1.3.1 Group I Initiators Group I initiators includo a failure in the suction sido of the SC pump. Suction lino initiators are the most common during the f midloop operation. These would includo air ingestion, inadvertent i closure of a valvo in the suction lino, failure of a relief valve ; to closo, laakago from the system and procedural errors. The i result of any of those initiators is to reduce the NPSHA for the SC pump. Information provided to the operator in the control room for ; detecting and diaanosing those ovents include various alarms and indichtors. The SCS includes an alarm for a low flow condition during shutdown cooling. This will be the initial indication of a , possible suction lino it.itiator since its set point is above the onset of cavitation. The-typical alarm used for SCS flow is set to indicate a drop in flow from the design value of 5000 gpm to approximately 3000 9pm. This, in conjunction with a low suction pressure, fluctuating motor current and near normal discharge pressure (during the onset of cavitation), will confirm a Group I ; initiator (SC punp suction). 2.4.3.1.3.1.1 Recovery During Mode 5 The equipment available to recover from this i.11tiator depends on the modo of operation and includes the redundant SCS train, one of two containment spray (CS) and two of four safety injection (SI) pumps. (Techvacal Specifications will require that two AC sources will be available to each division of class IE AC power during reduced inventory operationu in Modes 5 and 6. Son section 2.2 of this report.) The CS pumps are identical to the SC pumps and provido a redundant sourco for DHR flow during Modes 5 and 6. The SI pumps provide a viable sourco of DHR flow in Modo 5 as their capacity will match the reduced decay heat generation rate. If the redundant SCS train cannot be used to recover from a Group I initiator during Mode 5, the CS pumps can be used to re-establish , inventory control and DHR. The CS pumps can be aligned to take i suction from either the RCS hot-logs or the In-containment Refueling Water Storage Tank _(IRWST). During a Group I initiator, however, the CS pumps, which are normally aligned to the IRWST, can i be used to re-establish inventory' control by injecting IRWST water into the reactor vessel through the DVI nozzles (see Figure 2.4-1) . This alignment can also provide DHR using the SCS heat exchanger. This response requires operator action to open an SDS valve RCS to - allow water to circulate through the RCS to effect DHR (by opening 2.4-6 t _ . _ - _ . . . _ . _ . _ . . . _ . _ _ _ - - _ - _ . _ . _ _ . - _ ._ ~ _ . _
77 a .wp (9312) bh l l a Safety Depressurization System (SDS) valvo), the manual opening I of one (normally locked closed) cross-connect valvo and the actuation of the CS pump from the control room. Onco the event is terminated, either the original SCS train or the redundant train can be activated to resumo DilR. If a group I initiator produces a loss of coolant accident outside of containment, the operator has over 24 hours to jdentify and mitigate the ovont based on CS pump flow to match bolloff rate assuming the source of water is the IRWST. If the CS pump is not functional, the S1 pumps can be used to re-establish inventory control by injecting IRWST water into the reactor vossol. DlIR would then be performed by either the redundant SCS train af ter RCS level is recovered, by " break" flow, if the initiator provided an opening in the system, or by food-and-blood by opening a SDS valvo. In tho extremo, a single SI pump can provide sufficient flow to match boil-off thereby extending operator response time to identify the initiator and terminate tho l ovent. j 2.4.3.1.3.1.2 Recovery During Mode 6 l The equipment available to recover from this initiator during Mode ' 6 includes the redundant SCS train and possibly the opposito division's CS pump. j In this modo, the primary recovery system will be the redundant SCS train. Ilowever, if it is not available, then a CS pump may be aligned to take suction from the RCS hot-leg and discharge into the DVI nozzles. The success of this action is dependant on the particular Group I initiator since the operablo CS pump must use the same RCS-suction as the defonted SCS train (see Figure 2.4-1) and the opposite division's CS pump may be inoperable due to maintenanco. 2.4.3.1.3.2 Group II Initiators Group II initiators include a failure in the discharge side of the SC pump. Dischargo line initiators include inadvertent losure or opening of a valvc and the inadvertent actuation or leakage from a relief valvo. The result of Group II initiators is to change the SCS system resictanco curve. The pump will respond in accordanco , with its characteristic curve. Specifically, for the closure or a valvo in the discharge lino, the system resistance will-increase resulting in a decrease in DliR-flow and power consumption with a concurrent increase in discharge head. pump by-pass lines provent pump operation at shutof f. For the inadvertent opening of a valvo, there will be a reduction in system resistance which will produce an increase in flow and a decrease in power consumption at a lower pump head. \ 2.4-7
- -- -= . - . -
77c.wp(9212)bh Information provided to the operator in the control room to detect and diagnose those events include t';e name alarms and indicators discussed in connection with Group I initiators (sub-section 2.4.3.1.3.1). The instrumentation critical to identifying a Group II initiator includes the SC pump dischargo pressure, flowrate and - motor curront indication (con Tablo 2.4-2). 2.4.3.1.3.2.1 Recovery During Mode 5 , The equipment available to recover from this initiator during Mode 5 include the redundant SCS train, ono CS pump and two SI pumps. In this modo, the primary recovery system will be the redundant SCS train as the potential to lose the other SCS pump is unlikely. For the low probability case where the redundant SCS train cannot bo , used, the CS pump may be used to re-establish inventory control and DHR. If the CS pump is not functional, the SI pumps can be used to re-establish inventory control. DilR then would be performed by i cither the redundant SCS train after RCS level has been recovered, by break flow or by feed-and-bleed. 2.4.3.1.3.2.2 Recovery During Mode 6 The oculpment available to recover from this initiator during Mode 6 include the redundant SCS train and possibly the CS pump. The success of using the CS pump is dependent on the specific Group II initiator since the CS pump shares injection lines with the defeated SCS train. 2.4.3.1.3.3 Group III Initiators Group III initiators include a mechanical failure of the SC pump. Table 2.4-2 shows some examples. Recovery from Group III initiatora include activating the redundant SCS train or aligning the CS pump. 2.4.3.1.3.4 Group IV Initiators Group IV initiators are due to a loss of AC power. Recovery from Group IV initiators include automatic actuation of battery power, actuation of the combustion turbine or, if the loss of power is local to the train, activating the redundant SCS train or CS pump. System 80+ AC power availability is discussed in section 2.4.3.2. 2.4.3.1.4 Recovery Based on Plant configuration i Section 2.4.3.1.3 provided a general discussion regarding the recovery from initiators. This section examines the recovery from initiators for several specific plant configurations and modes. This analysis illustrates the capability of System 80+ to recover from losses of DHR and identifies new procedural requirements and 2.4-8
77a.wp(9212)bh technical specifications to minimize such losses and facilitato mitigativo actions, i Figuro 2.4-1 f acilitates the identification of soveral major plant configurations of interest for shutdown risk. The terminat!on points shown in the figure relate to Tablo 2.4-3 which provides an analysis of the configuration identifying possible initiators, l curront applicable technical specifications, now technical specifications, additional procedural requirements, alternativo support equipment and systems available to mitigato lossos of DHR and recovery actions from initiators. 2.4.3.1.5 Conclusions The System 00+ SCS design features provide the necessary redunda. icy, flexibility, and diversity to reduce the liPollhood of losing decay heat removal due to a loss of the SCS. The features of the design, the Technical Specifications, and the proceduro guidanco allow shutdown activition within certain limits and provido operational guidance for system flexibility and assurance that a loss of decay heat removal is unlikely. l. , 2.4-9 I l
p b O @
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W i m 9 - m rE a a )2 .ts i 9 l e a ggfd o g"g , og 4 E. j a U ' W
- ~
g R 3 Y o b i b w Figure f jfffg / d.u PLANT STATES AND TERMINATION POINTS FOR RESTORAn'lON OF DHR
, , . . . ..,_r ,~-, -.n . * * - ~ ^ * - " ' ~ ' " - . - . - - - . .- - -- -----~~ ~
77a.wp(9212)bh TA8tE 2.4-2 SCS INSTRtMENTATION lut TI ATCNP ITFM FESUtT TWf CAftWS/ At Ames
! - FAtttRE IN THE SUCTION t!NE 1kADVERTENT SIGNAL S1-651, 6533, 655 LOSS OF Co0 LING FLOW LOW FLOW Attsu FT-3C2 & F1-305 CLCSES MOTOR OR FLUCTUATING DISCNARE PRESSURE' OPERATED VALVE SI-652, 654, 666 CURRENT FLUCTUATIONS LOW SUCTION *RESSURE POSITION ICICATION CN VALVE CPERATORS OPERATOR ERROR IN SI-106 & S!-107 toss OF Co0 LING FLOW LOW FLOW ARM FI-302 & FI-305 ClostkG SCS SUCTION FLUCTUATING DISCMAFGE PRESSURE P-3C2 & P-3C5 IS0 TAT!ON VALVE CURRENT FLUCTUAffovS I-306 & 1-3C7 LOW SUCTION PutSSURE P-300 & P-301 POSITION IC ICATION ON WALVE CPERATORS LOW RCS LEVEL PLMPS WILL CAWITATE LOW FLOW ARM RESULTING IN VORTEM RESULTING IN LOSS OF FLUCTUATING DISCMARGE PRESSURE FORMATION AND AIR COOLING FLDW CURRENT FLUCTUAft045 ENTRAINMENT LOW SUCTION PRESSURE RCS LEVEL II - FAtttNE IN TNL JISCnARGE tlWE IkADVERTENT SIGNAL $1-310 & 312, 601 STSTEM PES!!TANCE LOW FLOW ARM F1-302 & FI-305 CLOSES MOTOR OR INCREASES CAUSING THE F..UCTUATING DISCPARGE PRESSURE P-3C2 & P-305 OPERATED VALVE S1-311 & 313, 600 PLMP TO OPERATE WE8R CURRENT FLUCTUATIOeiS P-300 & P-301 SMUTOFF LOW SUCTION PRESSURE I-306 & 1-307 RCS LEVEL OPERATOR ERROR tu St-579, 578 STSTEM PESISTANCE LOW FLOW AFM Fi-302 & FI-305 CLOSING SCS INCREASES CAUSING TME FLUCTUATING DISCPARGE PRESSURE P-302 & P-305 DISCdARGE ISOLATION PLP*P 70 OPERATE WEAR CURRENT FLUCTUAflDNS P-300 & P-301 VALVE SNUTOFF LOW SUCTION PRESSURE I-3G6 & t-307 RCS LEVEL i
I
77a.wp(9212)bh TABLE 2.4-2 SCS INSTRtBECNTATION
!NITI A7087 ITEM RCSULT INDICATMS/Alamr5 II - FAf ttRE IN TME DISCMAPCE LINE (CONT)
INADVERTENT SICMAL $1-690, 691 SYSTEM RESISTANCE LOW FLOW INDICATICW F1-302 & Ft-305 OPENS MOTOR ODERATED DEtaEASES CAUSING PUMP LOW DISCMARGE PRESSURE P-302 & P-305 . VALVE TO EXCEED ITS RUNOUT SUCTION PRESSURE CECREASES P-300 & P-301 FLOW. ALSO, FLOW SPLIT INCREASED POWER C045U=PTION 1-306 & 3-307 MAT CAUSE RCS TO HERT POSITION ICICATION ON VALVE OPERATORS UP AS LESS FLOW WILL SC DELIVERED. VALVES IN THE 1RWST S1-315,693 & 301 PLPEPS WILL DRAIN ThE Lt2UID LEVEL INSTR FOR MIDLOOP CPERATION SEE SECTION 1.3 TEST PATM NOT CLOSED OR RCS INVENTORY INTO TME RAPID DECREASE IN RCS PRESSURE FOLLOWING COMPLETION S1-304,686 & 300 IRWST TM*00GM THE TEST RAPID DECREASE IN Rf- EVEL OF SC FULL FLOW TEST PATM TMEN LOSE SUCTION LIQUID LEVEL ALAR *3 th IME IRWST L-350, L-351 ! AS THE FLUID IN inE HCT TEPP ICICAT10N IN TME 1RWST T-350. T-351 LEG DROPS AFTER 1EVEI rECREASES BELOW M!ut00P LOW FLOW ALARM F1-302 & F1-305 EORMAL CISCMAK0E PRESSURE P-302 & P-305 CURRENT FLUCTUAfl0NS I-306 & I-307 LOJ SUCTIC4 PRESSURE P-300 & P-301 SCS AC'. . 'E TR AI N IS SI-4SO & 455 TwE SCS NAS BEEN 4 OETECTION UNTIL RCS INVENTORT DECREASES USED TO FILL St-454 & 455 DESICALD TO SUPPORT DECRDSES BELOW THE MIDLOOP. REFUELMG POOL EPRI REQUIREMENT 4.3.1.2 FOR REFUELING LOW FLOW ALARM FI-302 & FI-305 i POOL (RFPS FILL. fF THE NORMAL DISCMARGE PRESSURE P-302 & P-305 RFP IS FILLED USING T W CURRENT FLUCTUATIONS P-3'M & P-301 TRAth PERFOR4tNG SC, LOW SUCTION PRESSURE I-30S & 1-307 THEN THE RCS INVENTORT COULD PE TRANSFERED TO THE RTP. INAJVE# TENT C80$$ 51-430, 431 LOSS OF CCCLANT FLOW LOW RCS LEVEL , CO MECT To f*E CONTAIN"ENT SPRAT SYSTEM i i
+- -
\'
i 5 0 361
- 0C 133 F - -
PP 2 C50 300
- 33 1 - -
FPP s w R A L A
/
st f O T A .E C R I UE SR W SU T ES RS PE R MEP RG ARN AO MI N O WCT LSC I T F I U A DS 2 W 4 T N OOO E LWW 2 M U E_ R t T s S A N T I W O S L C F S T N A L O O C F T C L U S S S E O R L M E T I h _ b
) P
_ 2 M U 1 P _ E 2 B 9 _ 9 f U t L ( R t I p O_ T A F A F A w I
- T
. T F a I N
A H 7 I ! S 7 f
,~ . . . _ _ . . . . _ . . . , _ . - - . _ . _ _ -
770.wp(9212)bh i i l TABLE 2.4-3 TERMINATJON POINT 1 Plant configuration Modes 4, 5 or 6 Initiators Loss of power current Applicable Technical specifications Lco 3.8.1 - 3.8.8 New Technical specification
-Requirements See section 2.2
+
New~ Procedural
. Requirements See section 2.1 Recovery i From Initiators See section 2.4.3.2 and 2.4.3.3.
6 2.4-14
r
. 770.wp (9212) bh - i f
TABLE 2.4-3 IRBlf1 NATION POINT 2 Plant Mode 4. Ccafiguration RCS cold leg temperature less then 317'F. ' IRWST Full. Initiatore Group I-IV. RCS line breaP., ; current Applicable- , Technical _ specifications LCO 3.4.6 , Two RCS loops or two SCS trains or any combination of these to be operable.. One- i RCS loop or SCS train to be in operation, j LCO 3.5.1 , _Four- SIT's operable when pressurizer-pressure is greater than 900 psia. 1 LCO 3.5.4 ! IRWST operable. LCO 3.6.6 Two CSS trains operable. ,. LCO 3.8.2 AC Power (Shutdown)
- New Technical
-specification -
Requirements- Two SI: pumps operable. ; New Procedural Requirements- Hone Alternative support _ Equipment / systems. Nono required. Recovery . . From Initiators DHR will be provided by sources other_ than + the SCS when the RCS pressure is above (500] psig. During these conditions,.the ECCS will , be' operable. The; SIS will - be available-by automatic actuation-down to RCS pressures of 400 poig (SIAS cutout pressure) and manually to 317' F. The: CSS is operable through out Mode 4. Below ' [500] psig Group I - IV initiators- can be nitigated- per section 2.4.3.1. 2.4-15
. a .. . .-___.~__.___._....__.___.,__..___a.. _..._.__.__._-..._.,__..n_.-_-.
W-770.wp(9212)bh i TABLE 2.4-3 IKIO(INATION POINT 3 Plant . Mode 4. Configuration RCS temperature greater than 317'F. IRWST full. Initiators Group I - III (for RCS pressure less then (500) psig). ; Group IV RCS line break. current Applicable Technical specifications -LCO.3.4.6 i Two RCS loops or two SCS trains or any combination of these to be operable. One , RCS loop or SCS train to be in operation. LCO 3.5.1 Four SIT's operable when pressurizer pressure is greater than 900 psia. LCO 3.5.3 Two SIS trains operable. - LCO 3.5.4 IRWST operable. [ LCO 3.6.6 Two CSS trains operable. New' Technical-specification i
. Requirements None
-New Procedural Requirements None
. Alternative' support Equipment / systems None required.
Recovery From Initiators DHR will be : provided by sources other than ; the SCS when the RCS pressure is.above (5001 psig. During these conditions, the ECCS will be-operable. =The SIS will- be available by-automatic actuation down to RCS pressures of ; 400 psig (SIAS cutout pressure)-and manually to 317' F. The CSS is operable through out: Mode 4. Below -psig Group I -IV initiators can b(e 500)itigated m per section 2.4.3.1. i I 2.4-16
77c.wp(9212)bh TABLE 2.4-3 TERMINATION POINT 4 Plan Cont guration Mode 5 RCS in reduced inventory. Nozzle dams installed. IRWST full. Initiators I-IV. Grou$1nebreak. RCS Current Applicable Technical specifications LCO 3.4.8 Two SC trains operable. One SCS-train operat ng. LCO 3.4.11 LTOP operable with a maximum of two SIS train operable. LCO 3.8.2 AC Power (Shutdown) Specification Requirements One CS pump operable. Two SI pumps operable. Midloop vent operable. New Procedural Requirements SC Msintain the minimum shutdown required cooling for D(HR.) flow rate near I Alternative Support Equipment / Systems Pumps...
- Charging pump.
Boric Acid Make-up pump. Tanks...
- Safety Injection Tanks Boric Acict Storage Tank (SITgT) (BA '
Recovery . From Initiators . Regain inventory control... SC, CS or SI pumps can be used to inject
-IRWST water into the RCS to reonin water level. If these pumps are not functional inventory co trol can. be established using a charg ng pump or boric acid make-up pump byyin ecting BAST water into the RCS.
Regain DHR capability... DHR can -be regai ed by using the , redundant SCS tra n once level ,s ' recovered. If the redundant SC pump ..s ' not functional DHR can -. be - established usin the CS p.- If the CS .s not func ional can be -esta hed by feed and bleed using SI pumps and opening the SDS valves
- Time To Boil Approximately 10-15 minutes.
I l
- 2.4'17 I
l
- - ~ m., , . . - . . , , , _ . . , . - . ~ . ~ . . . . - , . . - . . ~ , < - - . . . . - - - - . , . - - - . - ~ . - - - - - - - - - . - . - ~ - - - - - - - - - - - - - - '
770.wp(9212)bh-i t IARLE 2.4-3 ; IERKIFATIoM POINT 5 f Plant Configuration Mode 5. ! RCS in reduced inventory. Nozzle dams not installed. RCS closed IRWST full.(mid loop vent or RCP seals). a Initiators Group I-IV. - RCS J.ine break. I current Ap c Technical.pli_able , specifications LCO 3.4.8 Ivo SC One SCS train operat["ng. trains operable. LCO 3.4.11 s LTOP operable with a maximum of one SIS - train operable. New Technical specLfication Requ:.rements- ! Ono CS pump. operable. ' Two SI pumps operable. Midloop vent operable. New Procedural Requirements SC Maintnin the minimum shutdown required cooling for D(HR.) flow rate near Alternative support Equipment / Systems Pumps... Bhi!A!idMko-uppump. Tanks... Safety Inj ection Tanks ' Boric-Acia Storage Tank (SIT).(BAST) Steam Generators- t Recovery From Initiators Regain inventory control... SC CS or SI. pumps can be used to i.dect IREST
'evel.
water into the RCS to recain water If these pumps are not funct:.onal nventory control can be establ:.shed
-pump using bycharging injecting pumpBAST or boric acidinto water make-up the v ACS.
Regaan DHR capability... DHR can be regained by using the redundant SCS train. If the redundant SC pump is not operable DHR can be reoained using the cs pump. If the CS pump Is not functional DHR can be established . then by feed i initiallb and blee byusing reflux SI boiling,d pumps an opening the SDS valves. l 2.4-18 5
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- , 770.wp (9212 ) bh --
IhBLp a.4-2 IKEMINATION POINT 6 Plan Cont guration Mode 5.
-RCS in reduced inventor .
Nozzle dams not install d. RCS open (manway)' IRWST fuli. Initiators I-IV. Grou$1nebreak. RCS-Current Applicable Technical
. specifications LCO 3.4.8-
'Two SC trains operable. One SCS train o erat ng.
LCO 3 4.11 LTOP operable with a maximum of two SIS trains operable. LCO 3.8.2 AC Power (Shutdown)
-specification Requirements One CS pump operable.
Two SI pumps operable. Midloop vent operable. New Procedural Requirraents Maintain shutdown cooling ) flow rate near the minimum required for . Alternative support
. Equipment / systems pumps...-
Bhf!A!idMke-uppump. Tanks... Safety In"ection Tanks Boric Aci6 Storago Tank (SIT)(BAST) Steam Generators Recovery From Init iators Regain inventory control... SC CS SC or SI-pumps can be used to in ect ,IRWST water into the RCS to regain wa er level. If these pumps- are not functional -inve7 tory control can be-
.e:s ablished-using char pump or boric a ac d-make-u pu y1 in BAST water nto the R . s can also e used.
Rega n DHR capability...- , DHR can be regained by using .the redundant-SCS train. If the redundant SC 4 pump is not operable DHR can be regained " usin the CS p. IF the CS- -la not func onal . can - be est hed by i reflux boiling, - or feed - and bleed using- ~ CS or SI pumps and utilizing the open i pressurizer manway. ; f L' L 2.4-19 i
- , ,..s_ . _ , _ . . ...-,.~.__.a,,._a__,._a.,..,____-.-..._.,...- ,. _ ,,_ -. _ ,___ _,_,;, _ ,._.. n . . .._.2
770,wp(9212) bh TABLE 2 1-3,. TERMINATION POINT 7 Plant Mode 5.' Configuration RCS not in reduced-inventory. Nozzle dans installed. IRWST full.
' Initiators I-IV.
GrouSinebreak. RCS current Applicable Technical specifications LCO 3.4.8 Two SCS-trains operable. One SCS train operating. LCO 3.4.11 i LTOP operable with a maximum of one SIS ; train operable. i New Technical spec:.fication
~ Requ:.rements Two SI pumps operable. .New Procedural Requirements One CS pump available.
Midloop vent operable. , Aitornative support Equipment / systems Pumps... CS pump.. SI pump. 1!A!ihbke-uppump. Tanks... SafetyIn"ectionTanks(SIT Boric Acill Storage Tank (bas[T) Recovery From Initiators Regain DHR capability... DHR 'can be regained by using the redundant SCS train. If the redundant SC pump is not functional DHR can.be regained using the . CS pump. If the CS :
-is not DHR can be pump lished by functional
..estab feed and bleed - usino SI . --
pumps or utilizing the open pressurlzer manway.. l l 2.4-20
'87c.wp ( 9212 ) bh TADLE 2.4-3 TIRMINATION POINT 8 Plant Configuration Mode 5.
RCS water level above reduced inventory. Nozzle dams not installed. IRWST full. Initiators Group I-IV. RCS lino break. Current Applicable Technical specifications LCO 3.4.8 Two SCS trains operable. One SCS train operating. LCO 3.4.11 LTOP operable with a maximum of one SIS train operable. New Technical Specification Requirements Two SI Pumps operable. New Procedural Requirements One CS pump available. Alternative Support Equipment / Systems pumps... CS pump. SI pump, charging pump. Boric Acid Hake-up pump. Tankr... Safety Ind cction Tanks Boric Aci6 Storago Tank (SIT)T) (BAS Steam Generators Recovery From Initiators Regain DilR capability... DHR can be regained by using the redundant SCS train. If the redundant SC pump is not functional, DHR can be regained using the CS pum If the CS pump is not functional,,p. DHR can be established by reflux bolling, or feed and blood using SI pumps and opening the SDS valves.
-n >>. . . .
2.4-21 I
770.wp(9212)bh ; RRLE 2.4-3 RBJiD(ATION. POINT 9 ; Plant Mode 6 Configuration Refuelina pool empty
- IRWST Full Initiators Group 1-IV -
LOCA current rpplicable : Technics 1 ' specifications LCO 3.9.5 Two SCS trains operable. One SCS train operating. '
-New Technical .
specification Two SI pumps operable. ; Requirements' New Procedural Requirements None, Alternative bupport Equipment / systems one CS pump available. , Recoverv . From Initiators Regain DilR capability... DliR can be regained using the redundant SCS train. -If the SCS . pump ns not functional, D11R can be regained ustno the CS pump and SCS heat exchanger. Il the CS pump is not functional Di!R can be established using feed and bleed since the IRWST is not fully drained. i l' + l l 2.4-22 ,
'A n ,-r--+,-.mnw - ,,n ---.venow,
i 770.wp(9212)bh TA llLI L b i _1 T.EEt11 NATION POINT 12 Plant Mode 6 Configuration Refueling pool filled. Reactor vessel head off. Upper internals in place. IRWST empty. Initiators Group I-IV RCS line break. Current Applicable Technical Specifications LCO 3.9.4 For high water level one SCS train operable and in operati,on. LLO 3.9.5 For low water level two SCS trains operable and one in op,eration. New Technical specification Requirements Two SCS pumps operable. New Procedural Requirements One CS pump available. Alternative support Equipment / Systems Instrumentation.4 . Refueling Pool water level indication in addition to high and low level alarm. Pumps... Charging pumps. Boric acid make-up pumps. Tanks... Boric Acid Storage Tank (BAST) Recovery From Initiators Regain DHR DHRcan capability.*'ned be regai by using the redundant SCS train. If the redundant SCS pump is not functional DHR can be established by either passi,ve or active means as described in section 2.10.3. If DHR has been defeated due to an inter-system LOCA DHR can be regained by matching bo 'il-of f using the charging pumps or boric acid make-up pumps injecting BAST water. 2.4-23
I 770.wp(9212)bh TABLE 2.4-3 TERMINATION POINT 11 Plant Mode 6 Configuration Refueling pool filled. Reactor vessel head off. Upper internals removed. IRWST empty. Initiators Group I-IV RCS line break. Current Applicable Technical specifications LCO.3.9.4 For high water level one SCS train operable and in operati,on.
-LCO 3.9.5 For low water level two SCS trains operable and one in op,eration.
New Technicr' specificatit-Requirements Two SCS pumps operable. New Procedural Requirements One CS pump available. Altarnative' support Equipment / systems Instrumentation... Refueling Pool water level indication in addition to high and low level alarm. Pumps... Charging pumps. Boric acid.make-up pumps. Tanks... Boric-Acid Storage Tank (BAST)
-Recovery From Init iators- Regain DHR capability...
DHR can be regained by using the redundant SCS train. If the redundant SCS pump is not functional, DHR can be established by feed and bleed.- If DHR has been defeated due to an inter-system LOCA DHR can be regained by matching boil-off using the charging pumps or boric ' acid make-up pumps injecting BAST water. 2.4-24
770.wp(9212)bh 2.4.3.2 S_ya_tefa 8 0+ AQ._ht. w er..Rel.1 AhiliLY , t 2.4.3.2.1 Introduction This section presents the System 80+ features that increase the availability of electrical power to supply the Class lE buses and the capability to restore power if the electrical source is interrupted. The electrical distribution system provides redundant and diverse sources of power to the Class 1E buses during shutdown modes and reduced in.antory in the reactor coolant system and provides redundancy a 1 flexibility to insure re-energizing the Class 1E buses is post .ble if power is interrupted. 2.4.3.2.2 Discussion Electrical power sources need to be carefully managed during shutdown operations to maintain a desired level of safety. This is especially true during reduced inventory operations. Reduced inventory requires heightened awareness to manage the risks of maintaining an electrical source to the Class IE buses and of insuring an alternate source is available. The potential for a complete loss of decay heat removal due to the loss of electrical power is lowered when the electrical supply requirements for shutdown modes and reduced inventory are managed properly. The management and operation of these electrical sources will be guided by Technical Specifications for shutdown operations and reduced inventory. Technical Specifications will bo written to identify the minimum acceptable electrical distribution system alignments for operating in shutdown modes and reduced inventory. The operation of the electrical distribution system during shutdown modes and reduced inventory can be guided by procedures for normal alignments and for aligning alternate electrical sources if normal sources are interrupted. The electrical distribution system design will provide flexibility and redundancy to allow for the management of competing priorities during shutdown. These competing priorities include the need to perform maintenance en electrical system equipnent versus the need to have electrical sources available to provide power to the Class 1E buses. The System 80+ electrical system design (see Figure 2.4-5) provides the redundancy and flexibility to insure the risks associated with shutdown modes and re med inventory operations are lowered to acceptable levels. J is is accomplished by providing two I independent divisions of AC Elect *;ical Power. Each division has two 4.16 KV Safety Buses with three sources of electrical power. 2.4-25
77a.wp(9212)bh These three sources are:
- 1. Normal-Permanent Hon-Safety Bus (PNS-Bus),
- 2. Alternate-Reserve Transformer and
- 3. Emergency-Diesel Generator.
The normal sout.e (PNS-Bus) of power to the Safety Bus has three i sources of electrical power. The three sources are: (1) Normal - The division related Unit Auxiliary Trancformer (UAT) being powered from Switchyard Interface I through the Unit Main Transformer (UMT), (?) Alternate - The division related Reserve Transformer being set ved from Switchyard Interface II, and (3) Backup - the Combustion Turbine. Thereforo, the Class 1E-Safety Buses have the potential to be fed from four different ultimate sources during shutdown modes and reduced inventory operations. These sources are:
- 1. Switchyard Interface I,
- 2. Switchyard Interface II,
- 3. Diesel Generator, and
- 4. Combustion Turbine.
This distribution system provides the shutdown manageme team with the flexibility to perform shutdown activities on .
*ource of power to a division 4.16 VV Safety Bus and still mL :ain other diverse sources of reliable electrical power to the 4.1 KV Safety Bus.
Along with the electrical system design features, the System 80+ Technical Specifications include shutdown modes and reduced inventory operation Limiting Conditions for Operations (LCOs) . The LCOs provide minimum acceptab?.e electrical distribt' tion aligr.ments. Guidance is also provided by procedure to the operation staff to insure available source alignments are identified whenever shutdown activities are in progress. Add.tiona) procedural guidance is provided for aligning any r.vailable source (s) to the S3fety Buu(es) if power to the bus (es) le interrupted. The procedirre guidance and Technical Specificatiens are provided in Soutions 2.1 and 2.2 of this report. 2.4.3.2.3 Conc 1tision The System 80+ electrical distribution system design features provide the necessary redunl incy, flexibility, and diversity to reduca the likelihood of losi
- decay heat removal due to a loss of electrical pcwer. The feat ares of the design, the Technical Specifications, and the procedure guidance allow shutdown activities within certain limits and provide operational guidance for system flexibility and assurance that a loss of the decay heat removal is extremely unlikely.
2.4-26 i l
77a.wp(9323)bh 2.4.3.3 33 stem 80+. Diesel Generator Avati1 Ability t 2.4.3.3.1 Introduction The availability of the Dionel Generator and the Diesel Loading Sequencer to automatically start and load during shutdown modes of operation is one of the issues identified in NUREG-1410. The availability of the Diesel Generator instrumentation and control system to provide reliable indications and automatic trip signals for Diesel Generator protection during emergency operation (e.g. automatic start while in shutdown modes); and the availability of adequate information and indications to identify, diagnose, and correct Diesel Generator operational problems are significant to the overall maintenance of duay heat removal as presented in Section 2.4.3.1. The Diesel Generator (DG) and Diesel Load Sequencer (DLS) provide emergency power to the Class 1E buses during shutdown modes of operation with the same methods used during power modes of operation. The Instrumentation and control (I&C) system for the DG provides signals to start the diesel for omorgency operation, applicable protective trips to prevent or limit damage to the DG at all times and DG status to the Control Room and to the local control panel. This status includes trip signals (alarms, indications and recordings), parameter indications, and alarms for abnormal parameters. Also, controls for starting, stopping, , synchronizing, and loading the DG cre provided in the control Room and at the local control panel. 2.4.3.3.2 Discussion The Diesel Generator (DG) and Diesel Load Sequencer (DLS) need to maintain a conr41 stent means of operation independent of the plant operation coniition. This ensures the operating staff is not required to learn different operating senemes and therefore reduces potentici human error. The System 80+ DG and DLS provides this simplicity of operation. The DG is the amergency source of power to the Class 1E bus. The DG and- the DLS are available for operation during shutdown conditions unless undergoing maintenance. The Class 1E buses are monitored for undervoltage and degraded voltage conditions. If either condition is sensed, the DG is started and the DLS is initiated (see Figure 2.4-6 copied from CESSAR-DC Figure 7. 3-5) . For a loss of power to the Class 1E bus,. the response of the DG and DLS is not dependent un plant operational modes. Therefore, the response of the System 80+ equipment provides the operator with the same parameters and indication to be monitored whether shutdown or operated at power. This design characteristic provides a basis for , consistency in operating procedures and operator training. This ' eliminates the necessity of two sets of procedures dependent on l l l l 2.4-27
. .- . - - - _ . -- . - _ - ~ ~ . ,
77a.wp(9212)bh plant operating conditions. It also eliminates extra required training for the operation staff. (Detail on the Emergency Diesel Generators can be found in CESSAR-DC Section 8.3.1.1.4). The DG-I&C system needs to ensure the diesel is'protecter during all modes of operation, llowever, certain protective tript med to be bypassed during emergency operation. The System 80+ Diesel Generator . protection system provides automatic trips to prevent or limit damago to the DG. The protection trips provided during emergency operation are:
- 1. Engine Overspeed,
- 2. Generator Differential Protection,
- 3. Low-Low Lube Oil. Pressure, and
- 4. Generator Voltage - Controlled Overcurrent.
These trips are provided in accordance with Reg. Guide 1.9 Position
- 7. _ All other trips are bypassed during eme rgency operation. (See CESSAR-DC Section 8.3.1.1.4.4 for a complete description of trips bypassed during emergency operation) . The protection circuitry is dependent on the initiating signal and not dependent en plant operational modes. The sensing -of an undervoltage or degraded voltage condition during shutdoen causes an automatic DG start, activates the protective circuitry, and bypasses all non-emergency trips. This circuitry allows for consistency in the operational response to an emergency start of the DG independent of plant operating mode.
i The I&C system needs to ensure the operator lE informed of the DG's , operational status. This status includes parameter indications and alarms. 'The I&C systems need to provide controls to allow the operator to sta*t and load the diesel te provide pcwer to-the Class 1E buses. This status and control scheme needs to be provided locally and in the con *.rol room. The System 80+ coM e room is designated as the Nuplex 80+ Advanced-control Ccrtplex (hCC) . The Nuplex 80+ ACC presents the operator.with the information and controls necessary to complete any- tasks ~ identified in a te sk analysis proceus. The task analysis for_DG operationfidentit 2 the parameters, alarms, and controls required ' to - opuatc DG from the Nuplex 80+ ACC. This identified statue and ;ntrol scheme is presented to the Control
'm Operator on the .ectrkal Distribution Auxiliary Console.
presentation of this information is accomplished in accordance
~
a structural ar.d hierarchial termat discussed in CESSAR-DC Se-clon 18.7.1. This formatting provides the operator with parameter displays,-a] arm status, alarm categorization,-and alarm wior. tty. . This meth,J of information presentation provides the 2ntio) Room Operator (CRO) with the tools necessary to monitor
- .nd/or diagnose DG status.
2.4-28
770.wp(9212)bh The System 80+ local control panels for the DG provides the plant Equipment Operator with the same information and controls as is available to the CRO. The DG status information and control scheme on the local control canal utilizes the same Man-Machine Interface (MMI) features used i he Nuplex 80+ ACC. These features meet the System 80+ human f act. ors standards and guidelines. 2.4.3.3.3 Conclusion The Systu 80+ Diesel Generator instrumentation and control systems design features provide starting signals for the DG and DLS initiation and protective trip signals for DG emergency. operation and provide DG status information to the control room and local control panel which allows the operator to operate, monitor, and diagnose DG and DLS operation. These features of the System 80+ design enhance the operator's interface with the emergency equipment and reduces Je potential of human error. 2.4.4 RESOLUTION The issue regarding vulnerability during shutdown modes to a loss' of-decay heat removal (DHR) is resolved for System 80+ by the design features for the Shutdown Cooling System (SCS), instrumentation and controls, electrical power distribution system, nrw 'echnical specifications and procedure guidance described in t: ; od other sections of this report. These features demonstrate 'e uced potential for significant radiological releases from n+1 C iddine failure due to postulated events and radiological .- ' ow s froc a loss of DHR due to loss of SCS events. In p u L ur 'ar, features of the SCS and electrical distribution system p.e ide the necessary redundancy, flexibility and diversity to si9 c '.icantly reduce the likelihood of losing DHR. I l 2.4-29
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177e.wp(9212)bh 2.5 PRIMARY / SECONDARY CONTAINMXFT CAPADILITY AND SOURCE TERM m 2.5.1 ISSUE This section addresses the ability of the containment to protect the public from the consequences of a release of radiation during the time the containment is oper . This issue is related to events initiated in Mode 5 or 6 which have the potential for radiological release. The events which will be
- considered are the loss of decay heat removal capability initiated by either a loss of shutdown cooling or by a loss of coolant caused by either operator error or a pipe break.
Following a loss of decay heat removal not the result of a' pipe break, a radiological release from an open containment can occur when the time for the core to reach saturation is less than the time to restore RCS cooling and, failing this, the additional time to evacuate, close and isolate the containment. The time for the coolant to reach saturation is a function of plant conditions at the time the event is-initiated. Time to restore includes the time to detect that decay heat removal has'been-lost plus the time to restore either shutdown cooling or initiate alternate means of cooling. Time to detect depends on the instrumentation available to detect that Primary System cooling has been lost. The time to restore decay heat removal depends _on the j available systems and procedures. Once Primary System cooling has been lost measures must be taken to
- evacuate and' seal' the containment before the system begins to boil. - The time to close and isolate the containment depends on:
Design, operation,_ condition and status of equipment to close penetrations, equipment hatches and personnel air-locks, l Procedures for routing material and lines through these ! openings, Training of personnel Conditions of pressure, temperature and radiation within the containment as the core uncovers. l 2.5-1
77a.wp(9212)bh 2.5.2 ACCEPTANCE CRITERIA The following acceptance criteria apply to the issue addressed in this section:
- 1. Radiological exposure of the public to any event resulting in a loss of decay heat removal shall be limited to a small fraction of the limits stated in 10CFR100,
- 2. Radiological exposure to the public to any event resulting in a pipe break shall be limited to the limit stated in 10CFR100.
2.5.3 DISCUSSION 2.5.3.1 Eroblem FormulalioJ At issue is the radiological exposure of the public,for events leading to a release of radiation, while the containment is open. The release will depend on the events considered, the containment integrity; e.g. access areas open, and technical opecifications and procedures for closing the containment. Conservative assumptions are made in the analysis that leads to the time in which the containment must be closed such that radiation levels for personnel inside the containment and at the site boundaries- are within limits of applicable acceptance criteria. Containment closure depends on conditions of pressure and temperature within the containment following the start of the event. These will influence the time to achieve containment closure: pressure through the dose release rate outside containment, -and temperature through the limits on work time within
.the containment.
The results are used to support recommended changes to technical specifications and/or procedures. 2.5.3.2 _qsntAimment Intectrity The integrity of the containment vessel is to ensure that the release of any radioactivity does not exceed the limits established in 10CFR100. Containment integrity is maintained in accordance with Technical Specifications (TS) in Modes 1, 2 , 3, 4 and 5, with reduced inventory, and Mode 6, with reduced inventory or core alterations. In Modes 1, 2, 3 and 4 the containment is required to-be operable per TS 3.6.1 (Containment). Integrity exists when the items defined in the Definitions.Section of the Technical Specification are satisfied. Additional Technical Specifications for containment personnel locks and containment isolation valves provide specific 2.5-2
77c.wp (9212)bh actions and surveillance requirements to ensure containment integrity is not compromised. In Mode 5, with the Reactor Coolant System (RCS) in reduced inventory, and Moda 6 during core alteration, or reduced inventory, containment integrity is maintained in accordance with TS 3.9.3 and 3.10.5 (Containment Penetrations). 2.5.3.2.1 Integrity Requirements 2.5.3.2.1.1 Modes 1-4 Maintaining containment integrity in Modes 1-4 is accomplished by ensuring compliance with Technical Specifications. Prior to entry into Mode 4 from Mode 5, all surveillance requirements of TS 3.6.1 (Containment), TS 3.6.2 (Containment Personnel Locks) and TS 3.6.3 (Containcent Isolation Valves) are verified complete, in accordance with the applicable procedures. 2.5.3.2.1.2 Mode 5 Mode 5 is divided into two operational conditions:
- 1. RCS Level above reduced inventory
- 2. RCS Level below reduced inventory Entry into or out of these operational conditions is controlled by procedures and Technical Specifications and require verification by the-Senior Reactor Operator.
2.5.3.2.1.2.1 -RCS Level Above Reduced Inventory There are no Technical Specification requirements on containment integrity in Mode 5 when not in reduced inventory. Therefore
. proceeding from Mode 4 to Mode 5 does not require compliance with Technical Specitications dealing with containment integrity. It is
, -during this mode of operation that equipment for maintenance and ! refueling-outages, and support personnel, are moved into and out of containment through the one equipment hatch and two personnel locks. Also during this mode the surveillance testing of containment penetrations is completed and verified in accordance with site specific procedures. 2.5.3.2.1.2.2 RCS Level Below Reduced Inventory In Mode 5 the RCS may be drained to facilitate installation of the steam generator nozzle dams as well as other maintenance. Draining the RCS to a reduced inventory level (>3 feet below the reactor flange) requires (through the Containment Penetrations Technical Specifications) monitoring for any leakage of radiation through the 2.5-3
77a.wp(9212)bh penetrations. To maintain containment integrity, the equipment hatch and one of the two doors on each of the personnel locks must be closed during core alterations. 2.5.3.2.1.3 Mode 6 The potential for fuel handling accidents in Mode 6 establish the requirement that cor.tainment integrity be maintained. Thus the equipment hatch and one of the two doors on each of the personnel locks must be closed during core alterations. Entry from Mode 5 to Mode 6 may require verification of containment penetration status. Since Mode 5 has two possible operational states, containment configuration must be satisfied and verified by the Senior Reactor Operator. Entry into Mode 6 from Mode 5 reduced inventory operation requires monitoring of containment penetrations for radiation leakage. Entry into Mods 6 from Mode 5, not at reduced inventory, requires verification of penetration status. 2.5.3.2.2 System 80+ Containment Features 2.5.3.2.2.1 Building Arrangement and Ventilation The _ containment openings are surrounded by the Nuclear Annex Building. Therefore, there are no direct openings to the outside environment, and all leakage and air flow from' conte'nment openings (perconnel locks, equipment hatch, open penetrations) is exhausted into the Nuclear Annex. I The Nuclear Annex Ventilation System (a non-safety grade system)
. draws air from various points in the Nuclear Annex and exhausts to the unit vent. If high radiation levels are detected by the system radiation monitor, the exhaust flow automatically aligns to a filter train. The filter train consists of particulate filters and carbon absorbers to remove radioactive material prior to exhausting into the unit vent.
2.5.3.2.2.2 Personnel Locks The personnel locks allow passage of the work force into and out of the containment during all modes of operation. System 80+ has two personnel locks; one at elevation 115+6 (Figure 2.5-1A), one at elevation 146. Each personnel lock is a right circular cylinder approximately 10 feet in diameter with a door at both ends. The locks form part of the containment pressure boundary. Therefore, closure and sealing of the locks prevents leakage of radioactive material. The design and testing of the personnel locks ensures its ability to withstand pressures in excess of the maximum pressura following I~ ! 2.5-4 l
L77c.wp (9212)bh ~ containment DBA. Closure of a single door assures containment integrity. Each of the doors contains doubic seals and local leakage rate testing capability to provide pressure integrity. To offect a leak tight seal,-the personnel lock design uses pressure seated doors. Any leakage passes into the Nuclear Annex. Each personnel lock is provided with limit switches on both doora that provide control room indication of door position. The doors are interlocked, to prevent simultaneous opening thus compromising containment integrity during Modes 1-4. -The normal alignment of the personnel locks during the various modes of operation is listed in Table 2.5-1. In Mode 5 with inventory greater than the reduced level (<3 feet below the flange) both personnel locks can be only opened during an outage when it is necessary to transfer equipment into and out of containment. Closure can be initiated by dispatching personnel from the control room if containment integrity needs to be restored. Closure of both doors can be accomplished uitnin 10 minutes. 2.5.3.2.2.3- Equipment Hatch The containment equipment hatch provides a means for moving large equipment and components into and out of containment. On System 80+ =the hatch is 22' feet in diameter and located on the 146 elevation (Figure 2.5.-1B). Normal alignment of the equipment hatch during the modes of operation is listed in Table 2.5-1. -The hatch, when closed, is part of the containment pressure -boundary. Sealing is by means of a double seal which is Type B leak rate tested in.accordance with 10CFR50, Appendix J, prior to entry into Mode 4. The equipment hatch is removed following cleanup of containment : atmosphere and entry into Mode 5, at full inventory. The hatch moves . horizontally - on a: rail . systam. This design allows the hatch to be moved, allowing equipment to be transferred in and out of containment without interference._ The rail system is designed to minimize hatch movement thus reduce closure time. The rail system utilizes a AC powered trolley. This AC power is from a 1E bus which is_normally supplied from offsite power through the Unit Auxiliary Transformers. On loss of offsite power, power can be' supplied from the Reserve Transformer, Emergency Diesel Generator, or the combustion Turbine. In the event of the failure of all power sources the trolley system is designed to be operated manually. 2.5-5
1 77a.wp (9212)bh l Before proceeding to Mode 5, at reduced inventory, or Mode 6, reduced inventory, or core alterations, the equipment hatch must be closed. With or without AC power closure time is less than one hour. After being set in place, the hatch is bolted. Technical Specifications require that in Modes 1-4 all bolts be in place and tightened. In Mode 5, with reduced inventory, and Mode 6, Tech Specs require that (fourj bolts be in place and tightened. This minimum number of bolts is sufficient to secure the hatch so that no visible gap can be seen between the seals and sealing surface. The hatch is designed to be pressure seated. Thus any increase in pressure inside the containment will act to seal the haten. In addition any radiation leakage will be into the Nuclear Annex. 2.5.3.2.2.4 Penetrations There are [100) fluid system penetrations in the containment vessel. Each penetration is provided with a means of isolation by the Containment Isolation System (CESSAR-DC, Section 6.2.4). Procedures, to meet Technical Specification surveillance requirements, are provided for maintaining proper valve alignment to ensure containment integrity, prior to entry into Mode 4, Mode 5 (at reduced inventory) or Mode 6, reduced inventory of core alterations. In Mode 5, with the RCS level greater than reduced inventory, these penetrations are leak tested in accordance with 10CFR50, Appendix J. Mis-alignment of the valves can result in leakage paths limited by size of these normally small diameter (<.75 inches) valves. 2.5.3.3 Events Analyzed The radiological release is a function of the mass of coolant entering.the containment either as subcooled liquid or steam. In mode 5 this release can be the result of either a loss of shutdown cooling (Section 2.4) or a LOCA (Section 5.0). The release of radiation in Mode 5, at reduced 'nventory, can be the result of the events discussed in Section 2.4 leading to a loss of shutdown cooling. The most conservative assumption is events in which the system is in a reduced inventory condition resulting in the minimum time for the coolant to reach saturation. Per EPRI ALWR outage guidelines, the earliest a plant will enter Mode 5, at reduced inventory, is 50 hours after shutdown. For events leading to loss of shutdown cooling the minimum time to reach saturation was 10.5 minutes. Core uncovery was reached 55 minutes after saturation. The mass-energy releases for this event are listed in Table-2.5-2. 2.5-6
77c.wp (9212)bh For the LOCAs discussed in Section 5.3.5.1, a break initiated from in the DVI fullofinventory,or line 0.4 ft less results in a time to reach saturation of 1.7 minutes, a time to the initiate core uncovery of 7.17 minutes and an additional time of 16 minutes for the peak clad temperature to reach 2200*F. The mass-energy releases are listed in Table 2.5-3. Mode 6 events are all assumed to result from loss of shutdown cooling and subsequent heating of the coolant to saturation. Per the EPRI guidelines for ALWR outages the earliest a plant will start refueling is 86 hours after shutdown. Decay heat consistent with this time was assumed in calculating mass-energy releases in Hodo 6 (Table-2.5-4). The time to reach saturation, assuming an initial temperature of 135'F, with this decay heat rate is 14.6 minutes with an additional 125 minutes to core uncovery. 2.5.3.4 Acceptance Criteria Per Section 2.5.2 Acceptance Criteria are based on limits to radiological exposure to the public stated in 10CFR100. However, radiation exposure and containment temperature will affect the ability of utility personnel to close containment within an acceptable time. Acceptance criteria are stated to meet site boundary limits of 10CFR100 and for utility personnel, based on utility guidelines. 2.5.3.4.1 Radiation Limits Two limits on radiation levels will be considered;1imits on exposure to utility personnel working in-the containment and site boundary limits for release from the containment. These limits are used to eFbablish the time at which the containment must be closed to prevent exceeding either the off-site or in-containment acceptance -linits. 2.5.3.4.1.1 Loss of Bhutdown Cooling; Bite Boundary Limits The acceptance criteria in Section 2.5.2 for Loss of Shutdown Cooling refer to-limits based on a fraction of the whole body dose (25 rem) and thyroid dose (300 rem) mandated in 10CFR100. The fraction selected is related to the event probability per reactor year,.or the event frequency. The fraction is taken as 10% of the integrated whole body and thyroid doses stated in 10CFR100 for two hour exposure; Whole Body dose < .10 (25 rem) = 2.5 rem Thyroid dose < .10 (300 rem ) = 30 rem 2.5-7
3 E 77G.wp (9212)bh 2.5.3.4.1.2 Loss of Coolant Accidents; Site Boundary Limits Par the acceptance criteria in Section 2.5.2 exposure to the public to any event resulting in a pipe break are the limits stated in 10CFR100; Whole Body dose < 25 rem Thyroid dose < 300 rem 2.5.3.4.1.3 Limits on Utility Personnel Routine containment closure does not require special precautions outside of routine radiation work permits which provides workers with instructions on clothing required and the radiation exposuro of the work areas. In addition the atmosphere both inside and outside containment ic continually monitored for radiation levels. Maximum levels are established to limit exposure to workers both inside and outside containment and yet permit work to continue to mitigate the effects of any accident and close the containment. 2.5.3.4.1.3.1 Air Borne Radiation Equipment hatch installation during an accident situation requires added precautions to protect workers from both internal and external exposure while work is being performed and the containment is being closed. Workers are protected fron external contamination by being required to wear anti-contamination clothing. If airborne radiation levels in the work area reach or exceed .25 MPC workers are required to don one of the.following types of breathing apparatus; forced flow respirator supplied from a breathing air system, self-contained breathing apparatus, full face cartridge respirator. Work times for workers varies, depending on prior exposure history. Quarterly exposure of 520 MPCs have been established as the limit at which a worker will be required to exit containment. Each type of self contained breathing unit has associated with it a protection factor which reduces internal exposure, per the following
~
relationship; Received MPC
- Hour = (Measured MPC* time)/ Protection Factor Protection factors and the equivalent; measured MPCs used for determining the maximum time for utility personnel in containment are listed in Table 2.5-5.
Radiation levels are measured locally. Thus times at which exposure reaches unacceptable limits will vary with location within containment. The acceptance limit serves as an indication of the l time limits based on radiation exposure. 1 I L 2.5-8 l
1770.wp 1(9212)bh l
)
l
-2.5.3.4.1.3.2' Whole Body-Radiation '
LAreas' outside and inside containment are continually monitored for radiation. Maximum whole body dose of.2.5 rems, 10% the 10CFR100 limits, is permitted tc allow work to-continue to mitigate consequences of the accident. 2.513.4.2 Temperature Limits ' A combination of-environmental (temperature, relative humidity), and work related-(type of; clothing, type of work)-factors influence the-work time duration within containment, in addition to radiation-exposure., The temperature, as radiation exposure,-varies with location in containment. Thus, wcrk times based on temperature limits will also vary. The acceptance criteria is not absolute but serves:as an indicator of the work time based on average-containment conditions. Work 1 times:will also be_ influenced by_the type of protective clothing. :- NUREG 1449 notes an upper limit on temperature of 160*F to. avoid: burning _the lungs. -However, a self contained breathing pack =will provide air at a breathable temperature for a longer '
. period 1of time.-
3 Guidelines for the limits on the time in which work can be performed ' cin'high; temperature humid environments-are established in EPRI-NP4453-LRI. These " Stay Times" are based-on an average, or
- global, wet bulb temperature,. adjusted for-type of' clothing-(work
- clothing without or with vapor barrier) isnd. type of-work L(light,1 moderate or.. heavy). At the containmcnt initial conditions of 100"F and:50% relative humidity,Jassuming no protective clothing, ,
imaximum work-times for moderate work is longer than-two hours. WhereasTwith protective clothing ~the time _is reduced to about 60 minutes, lAcceptance limits on temperature will be assumed based on the-maximum-times'neededito closed containment in each of the modes and' Levents considered; ForoMode SLloss of shutdown cooling at reduced' inventory, and-Mode 6 with loss:of" inventory?due to boil-off, a minimum time needed to close the; personnel hatches of 10 minutes of moderate work is
. assumed.
For Mode 5 LOCA, a minimum time of 60 minutes of moderate work ja assumed. l 2.5-9 l l-
p
; 770.wp y (9212 ) bh --
e 2.5.3.5- __ Analysis
' Calculations were done to-predict pressure, temperature and activity-
-within'the containment for the Mode 5'and 6 events discussed above.
These calculations were donelin two_ steps.-
- 2.5.3.5.1 Thermodynamic Conditions' ,
, Ailumped parameter nodal model.isiused to. predict pressure, temperature versus time for;a given rate of mass and energy flow
'into the containment. The model includes. provisions for varying, as j 'a functien'of tine,-flow-areas open~tolthe ambient.
Mass and energy are' conserved in the vapor spaca for both-condensible and_non-condensibles components. The vapor space model- ,
. assumes:
1s." Complete-and instantaneous mixing of-the flows, 2.- Quasi-static _ equilibrium for temperature,
-3.-'Alliconstituents;are uniformly distributed, 4 .- . Dalton's-law applicable to find total pressure; containment-pressures-is--thelsum'of.the partial pressures of the-
- non-condensibles Land vapor pressure of the steam, 5.- Ideal _ Gas 11aw: applicable for determining partial pressures of f 'the.non-condensibles (air),
_.6. Steam partial pressures. determined from steam tables, for-both saturated and superheated conditions.
- The: code modelsLa_-sump;for the! accumulation of condensed-water, allowing for flashing of.the water based on(the temperature,of the
~
+
waterfexceedingLthe saturation temperature,? based on containment
-pressure..
1 Heat losses to passive?(walls)_ and active _(fan-coolers) heat sinks are includedt as options. - The method has-been validated'by comparison 1with the more detailed, multi-node code, CONTRANS;used for_the-detailed' containment analysis
.shown in/CESSAR-DC, Chapter 6. ;The main limitation, as compared to tCONTRANS, isLin the representation of passive heat sinks by,a slab-geometry, rather than_the more representative, multi-node model.
lThis resultsiin't.n-over-prediction of heat removal;resulting?in lower'containmentxtemperatures. However, the presentianalysis-
-assumes no passive or_ active heat sinks, resulting'in the maximum-values and_ rates of; change-of temperature and pressurefwith time' inside the containment.
'The,following nominal-initial conditions are assumed in the
- calculations; l
-2.5-10
L77a.wp'(9212)bh ! 1 Containment volume = 3,377,000 ?t' Pressure = 14.7 psia Temperature = 100'F Relative humidity = 50% Based on values for the open areas, the analysis predicts thermodynamic conditions inside the containment versus time for the mass flow rate and enthalpy release for the Modes 5 and 6 events (Section 2.5.3.3). The calculation starts (calculation time =0) at the time at which the coolant reaches saturation. 2.5.3.5.2 Radiation Release The amounts of activity inside and leaving the containment as a function of time are computed using results of the analysis for the ' thermodynamic variables in the containment; e.g., containment steam mass, pressure, temperature, integrated mass flow into the containment and integrated steam and air flow leaving the containment. The procedure calculates, for a given RCS activity:
- 1. Curies present in the containment,
- 2. Curies input to the containment,
- 3. Integrated Curies into the containment,
~4. Curies discharged out of the containment, -5. Integrated Curies discharged out of the containment. The model assumes perfect mixing.in the containment of the incoming RCS activity without assuming any benefit from decontamination factors (DF=1). 'The amount of exiting activity is based on the volumetric discharge of the air-steam mixture. The assumption of no heat removal in the thermodynamic calculation results in high values of pressure versus time, resulting in a conservatively high mass rate of flow though the open areas. In determining atmospheric concentrations a two hour EAB dispersion factor of 4.97x10" was used '(CESSAR-DC) . Atmospheric releases were calculated assuming no mitigating effect of the filters in the Annex building surrounding the containment. In calculating radiation limits for utility personnel a maximum control room atmospheric dispersion factor of 2.0 x10" was used (CESSAR-DC). RCS specific activity for the events analyzed were taken as the Technical Specification limits of a gross activity of (100/ E) 2.5-11
77a.wp (9212)bh-microcurie / gram and a dose equivalent I-131 specific activity of 1 microcurie / gram. 2.5.3.5.3 Results The calculation for activity, pressure and temperature starts at the time the coolant reaches saturation and continues, assuming no recovery, through core uncovery. The time limit of when containment must be closed is a function of when acceptance criteria for radiation exposure to the public are met. The time available for closing the containment (closure window ) is the difference between the acceptance criteria time limit and the time when the event is detected. The earliest time for detection (see Section 2.8 for detection-methods),- resulting in the maximum closure window, is at the initiation of the event (Figure 2.5-2). The latest time the event will be detected is assumed to be when the coolant reaches saturation. This results in the minimum closure window for closing the containment. Closure times herein are conservatively based on the minimum closure window. The dose rates and containment temperature are functions of the open area of the equipment and personnel penetrations and how long they
~
are open. Per Section 2.5.3.2, in Mode 5, with a full inventory, both equipment and personnel penetrations are open, while in Mode 5, with reduced inventory, and for Mode 6 reduced inventory or core alterations, one door in each of the two personnel locks must be closed but not sealed. For Mode 5 analysis for the LOCA, assumes full inventory,thus equipment hatches (380 ft ) and personnel locks (60 f tz) are assumed 2 open,to maximize release to ambient. Mode 5,' Loss of Shutdown Cooling, at reduced inventory fluid level is at the mid-plane of the hot-legs. For Mode 6 the fluid level is assumed at the level of the upper flange. This analysis l conservatively assumes that both personnel locks (60 f ta) are open. The thermodynamic and radiation analyses were done assuming that the equipment hatch and personnel lock areas remain open throughout the calculation. Times at which the open areas must be closed, based on the exposure limits in' Sections 2.5.3.4.1.3 and 2.5.3.4.1.2, are shown schematically in Figure 2.5-3: 2.5-12 i
770.wp1(9212)bh' d on
- 1. For:LOCA, limiting-site boundary doses-to; < 25 rem whole body, <
300 rem thyroid for-a'two hour period following start of the release. , 2.- For11oss shutdown' cooling,- limiting site' boundary doses-to; < 2.5 rem whole Triy, <-30 rem thyroid for a two hour period following the , t. a a s e . , i
-The time available-for_ utility personnel to complete containment closula'are based on the limits:in Sections 2.5.3.4.1.3 and 2.5.3.4.2;
- 1. Limiting dosefto utility: personnel within containment of 520 5 MPCs, '
- 12. : Limiting utility _ personnel to a whole body dose < 2.5 rems, ,
- 3. ? Temperature-limits inside containment based on minimum work time of 10 minutes of-moderate-work for-.. Loss of Shutdown Cooling events and'60 minutes of moderate work-for LOCA, t 2.5.3.5.3.1 Mode 5: Loss-of-shutdown' Cooling at Reduced LInventory The mass-energy release for this case is listed in Table 2.5-2 (Mode
'5,~ Loss of_ Shutdown Cooling). Results are shown in Figure 2.5-4.
- 2 . 5. 3 . 5 . 3 .1.1- Site Boundaries The.integratedctwo hour-doses for both whole body and tyroid exposures _are below:the' acceptance levels; Whole Body (rem)1= .216 < 2.5; Thyroid jem) = 4.54 < 30 g Thusuthe minimum closure window based on protection to the public is' over two hours.
- 2.5.3.5.3.1.2 Utility Personnel
.The-MPC. level at which breathing protection must be used is reached Lalmost immediately. The_ minimum closure window, based on the time z equivalent MPC level for full-face cartridge protection, is about:89 minutes. . With the next level.of protection (forced air) the MPC level after'two hours is,_ Maximum MPC'after two hours =15.85 < 520-The1whole body rem levels after two hours,for utility. personnel, are-
~below-the acceptance levels..
Whole-Body Dose (rems) = .087 < 2.5 i 2.5-13
770.wp (9212)bh Containment temperature after about two hours is about 180'F. However at-the ten minute limit needed to close the personnel locks -the temperature is 209'F. Thic temperature is lower than the EPRI temperature and humidity (120'F, 50% RH) limit needed for 10 minutes of moderate work, providing work time of about 30 minutes in which to close.the personnel hatches. 2.5.3.5.3.2 Mode 5: LOCA The mass-energy release for this case is listed in Table 2.5-3 (Mode 5, LOCA). Results are shown in Figures 2.5-5. 2.5.3.5.3.2.1 Site Boundaries The whole body and tyroid integrated doses after two hours are both below the acceptance limits for a LOCA; Whole Body (rem) = 1.35 < 25 Thyroid (rem) = 97.8 < 300 The-minimum closure window, based on site boundary limita is greater than'two hours. 2.5.3.5.3.2.2 Utility Personnel The lower equivalent MPC level for full face cartridge is reached almost insediately. However the maximum MPC levels with either air -supplied or self-contained unit are-not attained during the two hour period. Thus the minimum closure time vindow is greater than two hours. Maximum MPC after 2 hours = 173.5 < 520 The rem 1cvels for utility personnel are below the acceptance levels. Maximum whole Body Dose after 2 hours = .54 < 2.5 rems containment temperature within the first 10 minutes of the LOCA rises to about 170'F and then, due to the decrease in mass flow from the break and increased flow out of containment, decreases to an equilibrium value of 130*F. Based on the EPRI guidelines minimum closure time window to close the containment would be less than the time of one hour to close the equipraent hatch. 2.5.3.5,3.3 Mode 6: Refueling; Inventory Boil-Off The mass-energy releases for this case is listed in and Table 2.5-4 (Mode 6, Boil Off of Inventory). Results are shown in Figure 2.5-6. 2.5-14
77a.wp (9212)bh 2.5.3.5.3.3.1 Site Boundary Limits The two hour integrated release is below the_ acceptance limits for both whole body and thyroid exposures. Whole Body (rem)'= .227 < 2.5 Thyroid (rem) = 4.8 < 30 Thus the minimum closure window based on exposure to the public is in excess of two hours. 2.5.3.5.3.3.2 Utility Personhol The MPC level at which breathing protection must be used is reached almost immediately. The minimum closure window, based on the lowest level of protection (face cartridge), is reached in about 86 minutes. The exposure limit with the next level of protection (Forced Flow) is, after two hours, Maximum MPC after two hours = 16.3 < 520 The whole body levels for utility personnel are below the acceptance level. Maximum whole body dose after two hours = .G91 < 2.5 rems At the 10 minute point, the time set to close the personnel hatches, containment temperature is about 110*F. The EPRI guidelines allows for 30 minutes of moderate work at thic temperature. 2.5.4 RESOLUTION At' issue is_ release of radiation to the public during the time the l containment is open in Modes 5 and 6 Events in Mode-5 or 6 considered in which radiation may be released include loss of shutdown cooling or loss of coolant. l L The time in which-the containment closure must be completed is based on dose limits set by 10CFR100 and the rate of release for the events considered.. L The time to close the containment depends on: design, operation, condition and status of equipment; procedures and training of personnel in closing the containment. Conditions of pressure, temperature and radiation within the' containment influence the time available_for utility personnel to close the containment. 2.5-15
77c.wp (9212)bh-l l i 2.5.4.1 Modti 5; ReduceA_Loventory (Loss of Shutdown Coolina) i Results of the analysis show that the integrated two hour whole body ' and thyroid doses are below the acceptance limits of 2.5 rem (whole body) and 30 rem (thyroid). Radiation and temperature within the containment are within limits to provide a 10 minute closure window needed by utility personnel to close the containment. 2.5.4.2 Mode 5; Full Inventory (LOCA) . The analysis indicates that the integrated two hour wholo body and thyroid doses are below the acceptance limits of 25 rem (whole body) and 300 rem (thyroid). Internal radiation levels for utility personnel are within acceptable limits provided either air supplied or self-contained breathing units are used. Whole body dose is also within acceptance limits. Temperature within the containment are at levels that could limit the time available for utility personnel to close the containment within one hour. The use of multiple crews would help but the temperature levels could permit only short work intervals per crew. The following should be considered to either decrease the time needed to close and seal the equipment hatch and/or lengthen the work time within containment:
- 1. The use of fan coolers to decrease the maximum temperatures thus permitting longer work periods,
- 2. Improved design of the equipment hatch to reduce the time needed to close the containment,
- 3. A review of procedures and training with the objective of reducing the time needed to close the containment following a LOCA.
2.5.4.3 Mode 6: Refuelina Conficuration (Inventory Boil Off) Results of the analysis show that the integrated two hour whole body and thyroid doses are below the acceptance limits of 2.5 rem (whole body) and 30 rem (thyroid). Radiation and temperature within the containment are within the limits necessary for the 10 minutes needed by utility personnel to close the personnel hatches in the containment. 2.5-16
..- . . . , . ~ - - . -. - . . - . = =. - . . -.. ~ .. . .- . .- . - - -- . - . . ,
b 777a.wp(9212)bh i
- .'L' I TABLE 2.5-1 CONTAI]DLE.NT OPENINGS -
-Openina- . M9A Normal Status ft: Modes'l-4 Mode 5 Mode-S Mode > Reduce' < Reduced .
. Inventory Inventory Equipment 380 Closed Open. . Closed Closed
-Hatch !
'I
-Personnel 30/ Lock Closed- Open. 1 door: 1 door
. Locks (2)= closed closeJ per lock per lock TABLE 2.5-2 F
, Haps-ENERGY RELEASE FOR MODE 5 LOSS OF SHUTDOWN COOLING.
-Time Mass Flow- Enthalpy (sec) -flbm/sec) (B tu / lbm_1, 0 0 0
.025 13.8 1162.8 1000 13.8 1162.8 1
3000 13.8 1162.8 3300 Time to reach core-uncovery-10,000- 13.8 11G2.8 i .
'2 i
I -. l .. li l l' l- 2.5-17 4
L 770.wp(9212)bh l e TABLE 2.5-3 MASS-ENERGY RE_ LEASE FOtt MODE 5 LOCA Time Mass Flow Enthalpy (sec) LLbm/sec) 1 Btu /lbm.1 O O O 6.2 2128.8 267.9 100.1 1896.8 278.5 200.1 1383.5 292.3 { 300.1 1284.3 294.4
'400.1 841.3 308.9 430 Time to reach core uncovery 500.1 151.2 551.6 600.1 119.9 639.6 g 700.1 96.5 707.8 1100.2 98.4 430.4 1190 Time for Fuel to Exceed 2200 F 1500.2 129.3 380.5 1900 134 383.4 2300 133.9 385.6 1 .
2500 133.4 386.2 _ 3000 132 386.1 3500 131 384.1 4000 130.5 381 10,000 130.5 381 2.5-18
-I
- 77a.wp'(9212 ) bh1 -
TABLE 2.5-4 MASS-ENERGY RELEASE-FOR MQREL_(.. INVENTORY-BOIL-OFF Time Mass Flow Enthalpy . fsec) (Ibn/sec) ,[B tu, Libs) - 0 0 0 1000- 14.39 1150.5 . 4000 -14.39- 1150.5 6000 14.39 1150.5
-/500 Time-to~ reach core uncovery ;
10,000 14.39 1150.5 TABLE 2.5-5 PJtOTECTION FACTORS AND EOUIVALENT MPC 2 Type of-Breathing' Unit Protection Factor Equiv. MPC* Full Face Cartridge-50 2.6 x 10^4-Air Supplied-Forced Flow 2000 1.04 x 10^6 Self-Contained Breathing. 10,000 5.2 x 10^6
* . Based exposure limit of 520 MPC i-o 4
2.5-19
i s s 1
.I, it i
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770.wp(9212)bh l 2.6 BAPID BORON DILUTION 2.6.1 ISHUE The issues of the rapid boron dilution can be broken down luto three categories as follows:
- 1. The introduction of deborated water into the RCS via Shotdown Cooling System (SCS), which flows into the RCS through the Direct Vessel Injection (DVI) lines, during maintenance of inline components. ,
- 2. Introduction of a water slug into the RCS during startup or refueling operations, including a specific example from NUREG-1449 (Reference 3). In that example, a loss of offsite power han occurred and the charging pumps are returned on line, powered by the Emergency Diesel Generators. If the plant were in startup mode - i.e., deboration in progress - the charging purps could continue to operato causing a " slug" of unborated water to collect in the lower plenum of the reacter vossal.
If it is then assumed that offsitu powcr is restored and the RCP's are restarted, thea a water slug of deborated water can be injected into the core.
- 3. A potential boron dilution resulting from inleakage from the secondary side of a steam generator during a SGTR event.
All the above issues will be addressed in the discussion and resolution sections of this report. 2.6.2 ACCEPTANCE ORITEhIA The acceptance criteria for the rapid boron dilution event should be consistent with the acceptance criteria that are necessary to neot the relevant requirements of GDC 10, 15 and 26. Specifically, these criteria are as follows:
- 1. Pressure in the reactor coolant and main steam systems should be naintained below the RCS P/T limits (see Figure 3.4.3-1 of Technical Specification 3.4.3) or below 110% of the design value, whichever is less.
- 2. Fuel cladding integrity shall be maintained by ensuring that the minimum DNDR remains above the 95/95 DNBR limit for PWRs and CPR remains above the MCPR safety limit for BWRs based on
-acceptable correlations (see-SRP Section 4.4).
- 3. An incident of moderate frequency erould not generate a mnre serious plant condition without other faults occurring -
independently. 2.6-1
77a.wp(9212)bh l
- 4. An incident of moderate frequency in combination with any )
sing'o activo component failure, or single operator error, ' shall be considered and is an ovoit for which an estimato of I the number 6f potential fuel failures shall be provided for radiological doso calculations. For such accidents, the number of fuel failures must be assumed for all rods for which the DNBR or CPR falls below those values cited above for cladding integrity unless it can be shown, based on an acceptable fuel damage model (soo SRP Sectioli 4.2), that fewer f ailures occur. There shall be no loss of function of any fission product barrier other then the fuel cladding. The above criteria are the same requirements as the acceptance critoria for the Inadvortent Boron Dilution (IBD) event as stated in NUhtG-0800 Section 15.4.6, Reference 6, with the exception of Item 5. This critoria states that the available operator action time be 30 minutos for an IBD event during refualing conditions and 15 minutes for startup, cold shutdown and power operation. This requiremont is not applicable to a " rapid" boron dilution event. 2.6.3 DISCUBSION i 2.6.3.1 Melitificatipn of Di.111 tion Ecurces A study was performed tu identify possible flow paths of non-borated water which could potentially result in a water slug being injected into the RCS which subsequently finds its way into the core. The results of this study are presented in Table 2.6-1. The study concluded that considering restrictions on operations (see Section 2.1), the only source of non-borated water is the DVI linos. The maximum slug volume was determined to be 60 FT. 3 The , study also considered the issues identified in Section 2.6.1. The conclusion as shown in the resolutions of Table 2.6-1 is that for the System 80+ design, the scenarios defined by these issues do not result in a potential source of a non-borated water slug. 2.6.3.2 Event AnalyzpJ As mentioned in Section 2.6.3.1, the only credible acurce of an unborated water slug is the DVI lines, the volume of which is a 3 maximum of 60ft. This event was thus analyzed to determine the impact on the core and reactor coolant system. Table 2.6-2 list the assumptions and initial conditions used in the analysis. The water slug was assumed to be injected into the reactor vessel via the DVI lines at the maximum flowrate of the 4 high pressure - safety injection (HPSI) pumps. After the water slug was injected, a reactor coolant pump was assumed to start in order to instantaneously flush the water slug through the reactor vessel system and into the core. 2.6-2 _. . _ _ . _ _ _ _ _ _ . _ . - _ _ _ _ _ _. _ . _ -- - ~ _ - . _ _ _ - - - _ - - .
770.wp(9212)bh Two cases were analyzed. The first case assumed the plant to be in Mode 5 reduced inventory with a boron concentration resulting in a K-offectivu of 0,99 with all rods out (Ano) of the core. This configuration results in the least amount of water with the highest boron concentration prior to the injection of the water slug. The second case likewise assumed reduced inventory, and a boron concentration acsociated with ARO conditions. Ilowever, Mode 3 conditions were used. This resulted in a reduced fluid density and increased boron concentration. The purpose of the second case was to bound conditions in Mode 3, t and 6 (without reduced inventory) by maximizing the change in reactivity resulting from a larger differential in temperature and boron levels bntween the DVI injection water and the RCS water. The change -in the boron concentration, and RCS temperature, resulting from an unborated cold water slug being injected into ARO boron conditions was quantified in terms of reactivity units and compared with the margin to criticality available. In Modes 3, 4, and 5 this valoo will be equivalent to the required shutdown margin, since the reactor trip breakers will be open. This is required per the technical specification associated with the low flow trip, which requires at least two reactor coolant pumps be in operation if the reactor trip breakers aro closed. The above two cases bound the consequencer in Modes 3 through 5, since reduced inventory conditions were assumed, boron concentrations much higher than would be expected to occur with 6.5% Ak/k shutdown margin were used to conservatively calculate changes in boron concentration; in addition, saturated RCS conditions were also us ed .. thus resulting in the minimum initial RCS fluid mass. Although the Mode 6 boron concentrations would be larger than those utilized in the analyses, the slug which reaches the core will have a higher boron concentration than that which resulted from the mode 3 analysis due to the higher initial RCS boron concentration. Since the minimum value of the boron concentration for the Mode 3 case was considerably higher than the critical boron concentration with all rods out in Mode 6, the result of the Mode 3 analysis verifies the acceptability of the Mode 6 case. 2.6.3.3 Mathematical Model The above scenarios were modeled utilizing a computational Fluid Dynamics (CFD) software package utilizing a 2-D model. 2.6.3.4 Results The results of the above analysis demonstrated that with a rapid injection of an unborated water slug of 60 FT into the reactor 3 coolant system, in conjunction with the operational constraints as stated in the Technical Specification identified in section 2.6.3.2 2.6-3
77c.wp(9212)bh that the maximum positive reactivity addition for both cases 1 and 2 is lors than 2%. This is significantly less than the available shutdown margin of greater than or equal to 6.5%. For Mode 6 the maximum positive reactivity insertions will result in a K-effective from criticality. 2.6.3.5 Conclum.ign The analysis confirmed that the Acceptance Critoria, as stated in Section 2.6.2 has been mot. The core remained substantially subcritical, thus RCS pressure and DNBR Limits were not violated. 2.6.4 RESOLUTION The design of the System 80+ plant minimizes the possibility of a Rapid Boron 911ution event. Analyses have shown that the core remainc suberitical when the maximum credible " ster slug is aflushed" through the RCS. The concern of a slug bu.ig produced by the charging pumps following a loss of offsite power is not credible in the System 80+ design. The issue of a slug forming as a result of a SGTR event will be prohibited by the Emergency Proceduro Guides. In summary, the issue of a rapid boron dilution event for System 80+ can be considered resolved. 2.6-4
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Table 2.6-2 Rapid Boron Dilutions Analysis-Assumptions and Initial l Conditions l l Parameters ,qonditions Case 1 Case 2
- 1. RCS Liquid Volume Mode 5, Reduced Mode 3, Reduced Inventory Inventory (mid-loop) (mid-loop)
- 2. RCS Temperature 210*F' 572*F l
- 3. RCS Pressure atmospheric 1250 psia
- 4. CEA Configuration N-1 N-1 '
- 5. RCS Boron 1% suberitical 1% subcritical Concentration assuming ARO assuming ARO
- 6. Available Shutdown 7/6.5% AK/K 7/6.5% AK/K Hargin
- 7. Water Slug Volume, 60 FT 8 , 0 ppm 60 FT 2 , O ppm Boron Concentration
- 8. Water. Slug Injection 4 HPSI Pumps, 4 HPSI Pumps, Method Maximum Flow Maximum Flow
- 9. . Water Slug 40*F 40*F-Temperature
- 10. Single Failure * *
*No single failure will impact the event consequences.
I 2.6-8
770.wp(9212)bh 2.7 EIRE _PROTEC.II.QH 2.7.1 ISSUE The risk of fire during shutdown operations is higher than when the plant is-in power operation. This increase in risk is due to the presence of transient combustibles and ignition sources such as welding, grinding, and cutting operations necessary to support shutdown maintenance activities. Another risk in the reduced level of firo protection for systems such as the shutdown cooling and fuel pool cooling systems when the plant is in a shutdown mode, ' resulting in a higher susceptibility of failure due to fire. 2.7.2 ACCEPTANCE CRITERIA A defense in depth philosophy shall be employed in the des.'gn of the fire protection system in order to reduce the overall shutdown , risk due to fire. The elements in this defense in depth philosophy ! are:
- 1. Prevent a fire from occurring,
- 2. Promptly detect and suppress a fire,
- 3. Mitigate the consequences of a fire.
The fire protection features shall be independent from other features or systems which are routinely taken out of service during shutdown modes of operation. 2.7.3 DISCUSSION For clarity the three elements of the defense in depth philosophy outlined above will be discussed in reverse order. Only Division 1 of a system is discussed; Division 2 is identical to Division 1. 2.7.3.1 Hitication of Fire Conp_ecuences DIVISIONAL SEPARATION Shutdown Cooling System components for each division are completely separated from each other with 3-hour rated fire barriers with no communicating openings (see -CESSAR-DC Figure 9.5.1-2_ reproduced ' here as Figure-2.7-1). All penetrations within these barriers are scaled with assemblies that are qualified to maintain the integrity of the 3-hour rating. This assures that a fire involving oae division of Shutdown Cooling Systen coryonents will rot af fect the redundant division. 2.7-1
1 l 77a.wp(9312)bh I 1 INTERDIVISIONAL BEPARATION Within each division, the containment spray pump and the shutdown cooling pump can be interchanged with each other. These pumps can be used interchangeably with valve manipulttions guided by approved l procedures. For each division, the shatdown cooling pump is separated from the containment spray pump with 3-hour rated _ fire , barriers and 3-hour rated fire doors for openings. The valve which _ allows switchover from one pump to the other is located in a
-separate fire area. This will enable operators to make the ;
switchover without being exposed to a fire involving either the ' containment Spray or Shutdown Cooling Systems. Finally, the containment spray pump is powered from a safety bus separate from the shutdown cooling pump. The safety buses are separated from et.ch other with 3-hour rated fire walls. For example, the Division 1 Safety Bus A is located in Fire Area 65 and the Division 1 Saf ety Bus C is located in Fire Area 70 (see CESSAR-DC Figure 9.5.1-3 reproduced here as Figure 2.7-2). This interdivisional mechanical and electrical separation assures j
~
the operating of shutdown cooling can be maintained if a fire occurs concurrent with the redundant division being out of service. 2.7.3.2 Detection and Suppr.eju! Lion ofdires DETECTION Fire Area 38 contains the Division 1 containment spray pump and heat exchanger and Fire Area 41 contains the shutdown cooling pump and heat exchanger. These areas were evaluated during the recently completed System 80+ Fire Hazards Assessment. This assessment considared the fixed and transient combust.ible loads in these areas and the importance of the components to plant shutdown. Both areas will be equipped with full area coverage ceiling mounted-ionization smoke detectors. These detectors provide an early warning alarm at the central fire alarm console in the event of a fire. Detector location and spacing is based on engineering analysis to optimize detector effectiveness. This analysis will be referenced in the System 80+ Fire Hazards Analysis to be completed later in the design process. The detection system is highly reliable and will be kept in service at all times, even during shutdown modos of operation. SUPPRESSION The System 80+ Fire Hazards Assessnent concludes that a fixed automatic suppression in the form of automatic sprinklers is not warranted. This is due to the minimal combustible loadings in these areas. This will be verified later by engineering analysis, which is similar to the analysis for detector layout and location, 2.7-2
. _ , , ._ - ~_ __ _ .,._. _ _. _ _ _ _ _ _ _ _ _ _ - . _ . . _ _ _ _ _ _ _ . , _ . _ . _ _ .
77c.wp(9212)bh 1 and will be referenced in the System 80+ Fire Hazards Analysis to be completed by the plant designer before operations. Portable fire extinguishers and fixed manual fire hose stations provide manual fire fighting capability. The fire hoses are supplied from a dedicated fire protection water supply. Because of the fire barrier arrangement discussed previously, manual fire fighting activities can be accomplished without exposing either the l l redundant division equipment or interdivisional equipment to the effects of smoke or hot gases from a fire. MANUAL FIRE FIGHTING A fully trained and equipped on-site fire brighde would provide fire fighting activities for the System 80+. (See CESSAR-DC Section 9. 5.1. 9. 3. ) The brigade would be thoroughly familiar with the plant layout and will conduct sufficient fire drills and fire pre-planning to ef fectively control and suppress any credible fire. A documented pre-fire plan which outlines the necessary fire fighting strategies, will be prepared prior to plant start-up. MAIN 7AINED LEVEL OF FIRE PROTECTION The System 80+ fire protection system is not degraded or reduced during plant shutdown. There will ba no reason to breach the fire boundaries, interrupt the detection system, or impair the fire hose (standpipe) system. All of these features are provided specifically for fire protection and are not shared with or dependent on any other systems or features. 2.7.3.3 PreventionJf Fires Provention is the most important element in the defense in depth philosophy. When this element is successful there is no need to employ the other elements. To facilitate the implementation of this element, work place procedures and guidelines will be established by the owner-operator based on guidance provided by the plant designer. Procedural guidance would include control of combustibles, housekeeping, and control of hotwork. The preparation of these procedures will consider those areas in which a fire during shutdown modes of operation could pose a risk. The . procedures will include requirements to reduce the risk of fire ignition during shutdown. For example, the control of combustibles procedure may establish a maximum amount and configuration of combustible materials that may be left unattended in any of these areas. This will not be based soluly on an arbitrary " good engineering practico" approach, but will consider the amount of combustibles necessary to result in a fire that cou ld cause unacceptable damage. The control of hotwork and houcekeeping procedures will be developed by the owner-operat and implemented 2.7-3
._c
l t 77a.wp(9212)bh so as to not place unnecessary restrictions on shutdown maintenance i activities, yet will provide a high level of fire prevention. l 2.7.4 RESOLUTION , The fire protection features provided by the System 80+ design are ', consistent with the acceptance criteria outlined in Section 2.7.2. , These features will significantly reduce the risk due to fire ' during shutdown operation to an acceptable level. The combination ! of fire protection features resulting from employing Ko defense in depth philosophy will minimize the potential for fito damage to systems required for shutdown operations. This issue has been resolved by the design features of System 80+. I r 9 i i b r I
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_ _ _ _ ___.m . ._. 77e.wp (9212)bh 2.8 INSTRUMENTATLQH 2.8.1 ISSUE' Over the past several years, industry and regulatary concern with a-loss of shutdown cooling has increased. Despite an emphasis on improved shutdown procedures, the frequency of some incidents has not been reduced, particularly for losses of shutdown cooling during mid-loop operations. Furthermore, the effects of a loss of shutdown cooling are more serious than originally realized. The Nuclear Regulatory Commission (NRC)_ has requested responses to several design issues related to Nuclear Steam Supply Systems (NSSS) operations while on shutdown cooling; specifically during reduced inventory operations. Operators have, in many cases, had difficulty in determining plant parameters and equipment status during depressurized, shutdown conditions. This is due to the amount and quality of information available being marginally adequate or inadequate for prevention, recognition and mitigation of abnormal conditions in a timely manner. In particular, this information includes the reactor coolant system water level, reactor core exit temperature, and performance of decay heat removal systems. Losses of shutdown cooling can be partially attributed to misleading, inaccurate, or erroneous vessel level indication, particularly when vessel coolant level is lowered to within the hot leg between the level required for steam generator nozzle dam installation and the level required to prevent vortexing in the shutdown cooling suction line. Refer to Figure 2.8-1. Providing an adequate fluid level in the hot leg above the level at which vortexing occurs will ensure that the shutdown cooling fluid will not entrain air. This scenario-has been a contributor to the loss of shutdown cooling due to pump cavitation. The NRC has recommended that advanced reactor designs include an enhanced instrumentation package which assures:
- 1. that reduced inventory operations can be accurately and continuously' measured. For example, accurate instrumentation can establish reactor coolant level anytime during the draindown process. Accurate level measurement can assist in differentiating between the anticipated dynamic effects of the draindown process and additional, unintended. inventory losses; and
- 2. that a loss of decay heat removal event during reduced inventory operations can be readily detected. This ensures o
a timely response to a loss of shutdown cooling event. The ( instrumentation should " provide reliable indication'of j parameters that describe the state of the Reactor Coolant i l 2.8-1 l' L _ _ ___ _
770.wp (9212)bh System (RCS) and tio performance of systems normally used to cool the RCS for both normal and accident conditions" (Reference 4). The NRC has specified that instrumentation for reduced inventory conditions should provide both visible and audible indications of abnormal conditions in reactor vessel temperature and level, and decay heat removal system performance. 2.8.2 ACCEPTANCE CRITERIA The instrumentation provided for reduced inventory operations in the System 80+ design will reduce the safety risks associated with shutdown modes of operation. Instrumentation will be provided to avoid causing or contributing to a loss of shutdown cooling at reducad inventory conditions, and to aid in correctly interpreting a loss of shutdown cooling, should one occur. The following recommendations are taken from Enclosure 2 to Reference 4:
"At a minimum, provide the following in the Control Room (CR):
- 1. two independent RCS level indications when the reactor vessel (RV) head is on the vessel
- 2. at lear.t two independent temperature measurements representative of the core exit whenever the RV head is located on the top of the RV (we [NRC] suggest that temperature indications be provided at all times)
- 3. the capability of continuously monitoring decay heat removal (DHR) system performance whenever a DHR system is being used for cooling the RCS
- 4. visible and audible indications of abnormal conditions in temperature, level and DHR system performance."
Also, Enclosure 2 of Reference 4 includes NRC concerns and suggestions on meeting these recommencations. These include, for example:
"1. We suggest that licensees investigate ways to provide
[ accurate) temperature [ measurements] even if the head is removed, particularly if a lowered RCS inventory condition exists.
- 2. We expect sufficient information [be provided] to the operators that an approaching [DHR system) malfunction is clearly indicated.
2.8-2
-770.wp (9212)bh'
- 3. We expect both audible alarms and a panel indication when conditions exist which jeopardize continued operation of a DHR system, as well as when DHR is lost.
- 4. The low limit of level indication must be below the level necessary for operation of the DHR system.- Level information is necessary under loss of DHR conditions since it provides an indication of core coverage and ... of the ,
time to core uncovery. It is also useful in mitigating the l loss of DHR accident." i Section 2.8.3.2 of this report contains the description of the System 80+ instrumentation package for reduced inventory operations, ! including: the monitored paramotors, , instrumentation ranges and accuracies, alarm setpoints,- instrument availability, display and monitoring capability, and-quality assurance. A summary of the system 80+ design features which meet each of the above mentioned NRC recommendations for instrumentation are provided in the following. 2.8.3 DISCUSSION 2.8.'3. 1 Instrumentation Desian Basis To effectively monitor the draindown process to mid-loop via System 80+ enhanced instrumentation, information obtained from plant analysesiforms the basis for the instrument's design requirements. Instrumentation-specified for reduced inventory operations istbased on analyses ~in the following areau:- operations from a solid plant to mid-loop conditions (which define dynamic draindown. characteristics) ; instrumentation features which will reduce the likelihood of operator error during shutdown operation; possible ways in which shutdown cooling can be lost while the plant is_in a reduced inventory condition; 2.8-3
1 772.wp (9212)bh
-flow dynamics of the shutdown cooling system (SCS),
including those which contribute to vortexing; the plant response to losses of shutdown cooling, due to various initiators, including RCS thermal hydraulic effects and manometric effects;-and mitigation planning aimed at the reinitiation of shutdown cooling, delaying the onset of boiling, and delaying core uncovery. The design goals of the instrumentation package are to provide: prevention - enhanced monitoring capabilities for prevention of a complete loss of SCS operation, and mitigation - the timely response to a loss of SCS. These' goals have been achieved with the design features of the system 80+ instrumentation described in the following. 2.8.3.2 Instrumentation Description Table 2.8-1 describes the instrumentation package for reduced-inventory operations included in the System 80+ design. Additional de : ails are provided below. 2.8.3.2.1 Level Four. unique sets of instruments are provided for the measurement of levol-during RCS draindown and reduced inventory operations. These instruments make up the refueling water level indication system. The first set of instruments is a pair:of wide-range, dP-based level sensors. These sensors are provided to measure level between the pressurizer and the-junction of each SCS suction line with the RCS during draindown operations. Another pair of dP-based level sensors is utilized to determine RCS water level once'it is within the reactor vessel. These narrow-range level sensors function to measure level between the direct vessel injection (DVI) nozzle and the junction of the SCS suction lines with the RCS. One wide-range and one narrow-range dP instrument are connected to each SCS suction line. Separate lower level taps are provided for each instrument. See Figure 2.8-2. Because of the location of the upper level-taps, each of these dP instruments will operate with, or without, the reactor vessel head in place. In addition to the dP-based instruments described above, two heated-junction thermocouple (HJTC) systems will also be available for ! reactor vessel level measurement during Mode 5 reduced inventory l l l' 2.8-4
77Q.wp (9313)bh operations. The first system displays the output from the two inadequate-core cooling probes which are located inside the reactor vessel. The range of these probes extends from the reactor vessel
- head to the fuel alignment plate (See Figure 2.8-3). The measurement of RCS water level via these probes is limited only to those periods when the reactor vessel head is installed.
A second HJTC system provides narrow-range level indication for mid-loop operations via measurement of reactor vessel water level in the hot leg region. This system displays the output from two HJTC probes specifically designed with thermocouples clustered in the hot leg region _(see Figu'e 2.8-4). The benefit of this design is that it permits very accurate measurement when the reactor vessel is in the hot legs.
~
The HJTC systems compensate for'the flow gradient across the core associated with the operation of only one SCS suction line. The HJTC instruments are located in areas of the core which minimize the effect of the core outlet nozzles. The HJTC sensors have an accuracy and response time consistent with the maximum draindown rate of the RCS. The_HJTCs are designed so that instrument signal and power are transmitted on individual electrical conductors. Failure of one HJTC sensor will not result in a loss of signal from the remaining sensors. The measurement of RCS water level via these probes is limited to those periods when the reactor vessel head is installed. The use of both wide-range and narrow-range dP instruments, and two
. pairs of HJTC probes for refueling water level monitoring provides highly reliable, redundant, and independent indication of reactor vessel water level. Overlapping instrument ranges provide continuous draindown measurement from the pressurizer to a level below that necessary for SCS operation. Since this level instrumentation is independent, common mode misoperatica, or failures due to dynamic effects, will not be masked.
Each independent level instrument provides a suitable measurement, and is accurate, for its intended range of use. For mid-loop operations, the narrow-range HJTC probes provide accurate level measurement to within one inch of vessel level. This is critical since there is a very narrow margin between the RCS water level necessary for nozzle dam installation, and that required to prevent SCS pump cavitation. The refueling water level instrumentation is _ displayed and alarmed in the control room because of its importance to plant safety. 2.8-5
- 5
77c.wp (9212)bh L 2.8.3.2,2 Temperature Several instruments are available for continuous temperature measurements during reduced inventory operations with the reactor vessel head on. These include: core exit thermocouples (CETs), shutdown cooling heat exchanger inlet and return line temperature sensors, hot leg resistance temperature detectors (RTDs), and refueling water level instruments temperature sensor (HJTC probe only). All provide representative indications of the core exit temperature when the shutdown cooling system is operational. If the shutdown cooling system is lost, the CETs, hot leg RTDs, and refueling water level instruments temperature sensors (HJTC) input are available to track the response to the loss of shutdown cooling or the approach to boiling. Per Enclosure 2 to Reference 4, temperature measurement is provided with the reactor vessel head off. The temperature instruments operable during this mode are the hot leg resistance temperature detectors and, prior _to fuel shuffle, the CETs. Core exit fluid temperature can be measured through the use of hot leg RTDs as long as the SCS is operable. Each RCS hot leg has a total of five RTDs which are located in the hot leg at the junction of the SCS suction nozzle. In relation to the hot leg horizontal centerline, two RTDs are located above the centerline, one is at the centerline, and two are below the centerline. Only the lowermost two in each hot leg
.will provide input to the temperature reading for mid-loop . operations, since they will be the only ones in full contact with reactor coolant. The lowest probes penctrate the internal diameter of the hot leg pipe at approximately 10" below the midloop fluid level, thus assuring accurate readings are provided.
All temperature sensors will have associated alarms in the control room to be used as aids in determining the response to a loss of shutdown cooling and tracking the approach to boiling. Awareness of time constraints operator via training provides information that is useful.for deciding how to allocate effort. 2.8.3.2.3 Shutdown Cooling System Performance As stated in Enclosure 2 to Reference 4, sufficient information will be available to the control room operator to indicate an approaching shutdown cooling system malfunction. Indications of sufficient pump suction pressure and possible vortexing include unsteady pump 2.8-6
- 1 17 7e.wp-(9212)bh current (as indicated by SCS/ containment spray system (CSS) pump motor current), loss or reduction in shutdown cooling flow (as indicated by the shutdown cooling system flowrate), insufficient -pump NPSH.(as indicated by the pump suction pressure sensor), or indication of rising RCS 3evel (as water is displaced by the air and vapor in the shutdown cooling system). If a pump gives indications of air ingestion or cavitation, alarms will prompt the operator to stop the pump immediately. As detailed in Section 2.8.3.2.5, shutdown cooling panel displays will include valve lineup information for critical shutdown cooling flowpaths.
2.8.3.2.4 Quality Assurance The following instruments are designated as safety related and therefore within the' scope of environmental qualification and quality assurance. core exit thermocouples
-hot leg resistance temperature detectors refueling water level temperature sensor (unheated thermocouple) refueling water level instrument (ICCI heated junction thermocouple based design) shutdown cooling flowmeter shutdown cooling heat exchanger inlet and return liao temperature sensors
-shutdown cooling valve position indicators The-sarety related designation of these instruments is a consequence of their' required functions in other plant modes of operation, including for some, inadequate core cooling. The CENP Quality Assurance Program designates items which are safety-related as Quality Class 1 equipment, and therefore, are subject to the highest level of quality activity.
Enclosure 2 to Reference 4 states: " ...we will accept the following for resolving the items identified in the letter: ..... (2) reliable equipment in lieu of the comparable safety grade classification ...." The CENP Quality Assurance Program designates items which are not safety-related but nevertheless require a high level of quality activity, as Quality Class 2 equipment. In this case,:where reliable and accurate instrumentation is required for reduced RCS inventory conditions, designating the instrument as Quality Class-2 requires that a quality program be implemented that assures.that quality is commensurate with intended use. In the 2.8-7
-77c.wp (9212)bh' procurement of the instrumentation, appropriate technical requirements and quality requirements are specified in the purchase order to this end. The following list of Quality Class 2 instruments identified on Table 2.8-1 are classified as non safety-related:
g refueling water level indicator (wide and narrow range dP design),- refueling water level indicator (clustered HJTC design), shutdown cooling pump suction and discharge pressure sensors, and SCS pump /CS pump ammeter. 2.8.3.2.5 Display and Monitoring Capability
-Details of the NUPLEX 80+ Advanced Control Complex Information presentation and panel layout evaluation are described in CESSAR-DC Section 18.7. In addition to the following summary, refer to Section 18.7 for detailed or supplementary explanation of control room information presentation.
The operator obtains plant information from a number of sources in the NUPLEX 80+ control room, which include:
- 1. A large plant overview status board known as the Integrated Process Statur Overview (IPSO),
2.. Alarm tiles and associated alarm messages,
- 3. Discrete indicators which provide frequently used and important information,
- 4. CRT display formats containing essentjally all power plant information, and
- 5. Component and process control indicators.
There are a number of NUPLEX 80+ design features in 1 through 5 above that specifically implement indications, alarms, ano displays applicable to depressurized, shutdown conditions. They are described in the following sections. 2.8.3.2.5.1 Integrated Process Status Overview (IPSO) IPSO is used for quickly assessing overall plant status, organizing operational concerns, and establishing priorities for operator action. Information provided on the IPSO display includes: 2.B-8
770.wp (9212)bh -1. Major system andicomponent statuses shown on an overview schematic which are representative of the current operating heat transport systems,
- 2. Alarms to aid the operator in quickly identifying the location of important status information, 3.- Deviations from control setpoints and . identification of
-improving or degrading trends to improve the operator's awareness of plant conditions, and
- 4. Key representative parameters (e.g., RCS temperature and reactor vessel level).
Alarm windows are provided for plant critical functions: Reactivity Control - Electrical Generation
- Core Heat Removal -
Heat Rejection
- RCS Heat Removal -
Containment Environment Control RCS Inventory Cont. - Containment Isolation RCS Pressure Cont. - Radiological Emissions Control Steam /Feedwater Conversion *-
*For power production only Nuplex 80+ alarms are mode-dependent and equipment dependent to ensure their validity for different operational conditions. - For all modes, including shutdown and refueling conditions, individual sensed process parameter values-and alarm states are used to determine critical function alarms, either directly or as processed by an algorithm that-uses more than one (1) process parameter input.
In either case, - the operator quickly is made aware of the affected critical function (s). For example, a high core exit temperature alarm state would be used as an input to the Core Heat Removal critical safety function alarm during a loss of-shutdown cooling. The systems represented'on IPSO are the major heat transport pathways and systems that are required to support the heat transport process. These systems include those that require availability monitoring per Regulatory Guide-1.47, and~all major success paths that support the Plant Critical Functions. The following systems.have dynamic operating status representations on IPSO. Their identifying descriptors on the IPSO display are shown below: CC - Component cooling water CD - Condensate CI - Containment isolation CS - Containment spray 2.8-9
177c.vp (9212) bh CW - Circulating water EF - Emergency feedwater FW - Feedwater IA - Instrument air SC - Shutdown cooling RC - Reactor coolant 'SI - Safety injection SW - Service water TB - Turbine bypass SD - Safety Depressurization , System information presented on IPSO includes system operational ' status, any change in operational status (i.e., active to inactive, or inactive to active) and the existence of alarms associated with < the system. Alarm information on systems helps to directly inform l __an operator about possible underlying causes of critical function alarms. The IPSO display, as well as all display pages, is also available at any data processing system CRT, which includes control room panels, the control room supervisor's desk, assistant operator workstations, and the technical support station. 2.8.3.2.5.2' Alarm Tiles and Associated Alarm Messages Alarm tiles are displayed on electroluminescent flat panel displays
'in the Discrete Indication Alarm System (DIAS). These tiles are functionally grouped and located on the appropriate control room panel. Shutdown cooling system alarm tiles are located on the Engineered Safety Features-panel. This panel includes the controls for Safety Injection, the Safety Injection Tanks, Shutdown Cooling, Reactor: Cavity-Flood, Safety Depressurization, Emergency Feedwater, Containment Spray, IRWST, and Containment Isolation. Individual alarm inputs to the shutdown cooling alarm tiles include (for each train):
low shutdown cooling pump header pressure low shutdown cooling flow high shutdown cooling heat exchanger outlet temperature shutdown cooling pump motor current deviation -In addition, this panel will have a tile for RCS conditions, with individual inputs for shutdown, depressurized conditions: low RCS water level' high core exit temperature low refueling cavity level 2.8-10
77s.wp (9212)bh-Tc_ ensure alarm validity, all NUPLEX_80+ alarms _are mode and equipment status dependent, and signal validation of inputs is done where multiple signals of the same process parameter exist. These features eliminate nuisance alarms and help ensure a true " dark board" when alarms do not exist. These features enhance operator diagnosis of alarms when they do exist. When alarm tiles in DIAS are acknowledged, the operator is presented with a DIAS display with alarm messages showing which of the alarm tile inputs caused the alarm. 2.8.3.2.5.3 Discrete Indicators Discrete indicators are provided on the NUPLEX 80+ control room workstations to provide the operator with information that (1) is frequently used.to assess system level performance, and (2) allows continued operation if the Data Processing System should become unavailable. Discrete' indicators use validated process parameter inputs where multiple process parameter measurements exist, and include trend information for routine monitoring, and diagnosis of abnormal conditions. Where analog data is composed of different . ranges of information, DIAS automatically shifts to the appropriate range, and indicates to the operator that a-range change has occurred.- Discrete-Indicator displays to support shutdown cooling for key parameters are on the Engineered Safety Features panel. These incluce: Shutdown Coolina System (ner train) inlet temperature outlet temperature pump header pressure flow heat exchanger inlet temperature heat exchanger outlet temperature pump motor current Reactor Coolant System pressurizer level reactor coolant system level 2.8-11
77a.wp (9212)bh l l pressure core exit temperature refueling cavity level 2.8.3.2.5.4 CRT Display Pages CRT display pages contain, in a structured hierarchy, all the System 80+ plant information that is available to the operator. The CRT pages are useful for information presentation because they allow graphic-layouts of plant processes in formats that are consistent with the aperator's visualization of the plant. In addition, CRT formats are designed to aid operational activities of the plant by providing trends, categorized listings, messages, operational ' prompts, as well as alert the operator to abnormal processes. The IPSO display page forms the apex of the NUPLEX 80+ CRT display page hierarchy. Three levels exist below IPSO: general monitoring, system / component control, d6t. ail / diagnostic. Each level of the hitrarchy provides an information content designed to satisfy particular operational needs. The CRT displays are provided by the Data Processing System (DPS). Any display page is available at any CRT. Operator acknowledgement of CRT alarms also acknowledges the same alarm in DIAS (and vice versa). The CRT alarm actuation message indicates the cause of the alarm, similar to DIAS. i ' The shutdown cooling system will be shown on a Level 2 display, with more detailed information on two Level 3 displays, one per shutdown cooling train. These displays will include all necessary information to clearly describe the status and performance of the system. This includes system mimic, component activity (e.g., on/off or open/ closed) component controllability (e.g., key valves locked open or closed), system parameters (e.g., temperature, level), and system / component alarms. The Level 2 display will include reactor coolant system level and core exit temperature to integrate the shutdown cooling and RCS status for this display. The RCS is also presented on a separate Level 2 display. 2.8.3.2.5.5 Component and Process Control Indicators NUPLEX 80+ component control features (e.g., actuation / switches / controls) provide the primary method by which the operator actuates equipment and systems. The shutdown cooling system controls are functionally grouped within a system mimic on the Engineered Safety Features panel. At that panel, shutdown cooling system control is l integrated with DIAS alarm tiles important to shutdown cooling"and f 2.8-12 I
77a.wp(9?l2)bh 4 with CRT display of the shutdown cooling system. Controls, alorms and CRT displays for other systems applicable to shutdown 1 operations, such as component cooling water and safety injection, are available at that panel as well. 2.8.3.2.5.6 NUPLEX 80+ Alarm Characteristics There are a number of special features in the design of the : UPLEX 80+ alarm system _that support operator diagnosis of alarm conditions and that would be particularly supportive of depressurized, shutdown operations. These are:
- 1. Mod 6 and Equipment Status Dependency
- 2. Audible Alarr Information
- 3. Stop Flash Feature
- 4. Operator Established Alarms
- 5. Operator Aids In addition, the categorization of all alarms is considered in the bases for alarm display location. For instance, alarms that indicate approach to potential equipment damage, but do not affect critical function or - success _ path status, are presented only on alarm tiles. These would not be included as input to alarms shown on IPSC.
A key feature to aid operator navigation in the CRT display page hierarchy also includes alarm categorization to assist the operator. This feature, the " display page menu", is on each CRT display page. The menu indicates alarms exist in various sectors of the hierarchy, and depending on the sector e the operator can distinguish between lower level alarms that are critical function or success path related, and those that are not (e.g., personnel 4 hazard, or equipment damage) . By displaying critical function alarms, success path alarms,
- personnel hazard alarms and equipment damage alarms on unique display locations, the operator can rapidly' determine the type and relative significance of alarms. For example, an Inventory Control critical function alarm, without a concurrent Volume Control success path performance alarm immediately suggests that inventory may be decreasing due to a non-success-path cause, such as a coolant leak, in which case indication for IRWST level, and containment temperature, pressure and humidity would be immediately checked by the operator. Similar distinctions can be made by the operator for single or multiple alarm conditions to assist the 2.8-13
77a.sp(9212)bh operator in quickly establishing needs and priorities for operator action. 2.8.4 RESOLUTION The issue of instrumentation for shutdown operation is resolved on System 80+ by the instrumentation and control room displays described in the previous sections of this report. This instrumentation will meet or exceed the recommendations of Generic Letter 88-17, and will significantly reduce risk associated with operations during shutdown, particularly when the reacter is in a reduced inventory condition, as long as prior to the start of draining the reduced inventory in9trumentation is placed into operation. The NUPLEX 80+ Advanced Control Room Complex provides an overview display, indicators, CRT displays, and alarms that meet the
-acceptance criteria in Section 2.8.2. Indication and alarms are provided on discrete indicatora, alarm tile windows and CRTs for RCS level and temperature. In addition, shutdown cooling system status and performance is monitored on CRTs. Shutdown cooling system performance is alarmed on IPSO, alarm tile windows and on CRTs. Also, all alarms are processed for their individual effect on plant critical functions such as reactivity control, core heat removal and RCS heat removal.
l l 2.8-14 l L
770.wp (9212)bh TABJ.E 2. 8-1 REDUCED INVENTORY INSTRUMENTATION PACKAGE Monitored Instnammt Instnamt Indication and fitesseter lype Emc t ion __ Larste Atare 1 oration gommmmt a RCS Water level Refueling Water Cont iruous, Vide Ranges two Control Room, Highty reliable. Levet Irdication redundant wide inst runent s, with low and Meets WRC System (dP based range RC$ water each with a low tow level requirement for design) level indication tap at hot les/ alarms, water Level during draindown SCS suction measurement to a operations, line interface, point Lower than reference leg that required at top of for SC$ pressurizer. operation. RC$ Water Level Refueling Water independent Top of the Control Room, Redundant axist Level Indication continuous tevet vessel down to with low ard strings of System (ICCI indication in the the fuel low-low level thermocot.pt es MJTC system) reactor vessel, alignment starms. from the vessel plate. 5ead to the fuel aligrvnent plate. System provides excellent accuracy and continuous measurement. RCS Watsr Level Refueling Water Irdependent, Top of the Control Room, Red edant axial Levet Indicating cont irmous, vessel down to with low-low, strings of System narrow range level the fuel tow, high and thermocouples (clustered HJTC indication in the aligreent high-high level from the vessel design) reactor vessel. plate. alarms. heed to the fuel allgment plate. This instninent is different from the ICCI HJTC system discussed above in that thermocouples are clustered in-l the het leg l region to provioe greater inst runent accuracy t<1a). RCS Water Level Refueling Water Continuous, karrow range: Control Room, Highly rettable Level Indication redundant narror two with low and for mid loop (dP based range level instrunents, low-low level operations, design) indication during each with a tap alarms. Meets NRC reduced inventory at SCS suction requirement for operations. tine / hot leg water level interface, measurement to a reference leg point tower than at DVI nozzle. required for SCS cperation. l l l 2.8-15 l
77c.wp (9212)bh LAnt E 2.8-1 (Contirued) R[ptKIp 1MMORJ LipiPts O TAilow PAC (Ari 14anitored Instruent Irmtrument Indication and ter m Tee Furrt ion Perse hjpralocation gamemt s RC$ Tecperature Ctis Measures Optimited for Cor. trol Room, Tracks approach ( thermocouple terrperature of SCS and with alarms at to boiling. design) coolant emiting refueling high and high- Tenperature core, modes. Will high indication measure t ecperature, provided even boiling, when head is off Apprenimate vessel. Not range 50 250 evallable during des F. f a t shuffling. Availability will be maalmited. RCS Tenperature Refueling water Continuous. Optimited for C mtrol Room Indicates actual level probe irdependent SCS and with alarms at vessel water (heated junction temperature refueling high and high- tewperature, thermocouple measurement inside modes. Witt high Tracks approach based design) the vesset, measure tecperat res. to boiling, bolling. Approximate range 100 - 250 deg F. Hot Leg temperature Resistance Measures core exit optimized for control Room, Tewperature Tevperature tecperature in the shutdown alarms at high indication is l Detectors (RTDs) hot leg at both operations. temperature. affected by loss SC5 suction line Approximate of shutdown regions. range 50 - 250 cooling flow, Redundant RTDs deg F. Since flow by provided in each the RfDs will hot Leg. not occur. SCS Flowrate flowmeter Decay heat removal Bounds SC5 purp Control Room, One tocated in system flow range, includes low each SC5 return performance. flow alarm, line to the RCS, Can be used to measure CSP flow if C$Ps are used for SCS. SCS Purp/CS Pwp Pressure sensor Measures 0 to system Control Room one instrunent Discharge Pressure individual purp design with low located at the discharge pressure, pressure alarm, discharge to pressures. each purp. Identifies individual pwp status. SCS Pwp/CS Purp Annet er Measure current 0 to system Control Room, confirms pwp Motor Current drawn by pwp design pressure alarms with status motor. preset drop in (individual pw p Fluctuations show current. air entrairvnent) air entrainment, independent of pressure and flow indicators. 2.8-16
77a.wp (9212)bh l l l 1 l JfAalt 2.8-1 (Contirsed) REDUCfD inTNTCRT leSTPUMt al Afitw McKAM Manitored Ins t rtment Instruunt Indication ar:1 Peresset er Type bection m o elarmlocatioD E0'emmt s SCS Pupp/CS Pwp Pressure sena;r Measure pwp 0 to system Centrol Room one instrunent suction pressure suction pressure design with low located at the in each pw p. pressure. pressure alarm. suction of each pwp. Identifies individual punp utatus. SCMX Inlet and beturn Tecperature heasures 40 - 372 deg. Control Room Ternperature Line Ternperature sensor tecperature in the F. alarms at high irdication only suction and t epperatur e, evallable when i i discharge lines of SCS is the shutdown operational, cooling heat exchanger. SCS Valve Position Valve position Status of valve Open/ closed /thr Control Room Will provide Indication indication positions in the ottled position information of open/ closed or SCS. indication. system lineup throttled. status and available flowpaths. 2.8-17
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. SCHEMATIC 2.8-2
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SCHEMATIC REPRESENTATION OF THE jfd / INADEQUATE CORE COOLING HJTC PROBES 2.8-3
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4 77a.wp(9212)bh o 2 '. 9 ECCS RECIRCULATION CA_RABR Uj( 2.9.1 lbsUE, The issue. is'. the potential for loss of flow to the Containment Spray-_.(CS) and Safety Injection -(SI) pumps during accident conditions. System flow could be inhibited by a number of factors. These factors include: i
- 1. Hydraulic effects, such as air ingestion and vortex formation. .
- 2. Debris in the-IRWST resulting from maintenance activities or .
deterioration - of insulation from actuation of containment sprays or from LOCA consequences, or
- 3. The combined-effects of items (1) and (2).
2.9.2 ACCEPTANCE CRITERIA The design of th'e System 80+.Incontainment Refueling Water Storage Tank _(IRWST) and Holdup. Volume Tank (HVT) and their ' associated d e b r i s - b l o c k i n g .- d e v i c e s . s h a l l comply with the requirements. of iGeneral-; Design Criterion 35 of Title 10, Code of Federal
-Regulations,-Part 50,-Appendix A.- Design-Criterion 35 requires
-_that " ... suitable containment capabilities shall be provided to ,
assure that.... the system's safety function can be accomplished."
'To satisfy this requirement, the IRWST and HVT are designed'to provide a clean and_'eliable r source of water to the SI pumps for .
long-term reci:.aulation. The containment is designed to direct
? containment spray water and emergency core cooling water to the HVT and then to the IRWST.
The _ SIS : shall. meet E the = acceptance- criteria specified in USNRC ; Standard Review Plan.rection16,3,- Emergency Core Cooling System,
' Revision 2. In particular,_Section 6.3 of the SRP addresses the
-availability :of -an-. adequate. source _of water :for -the SIS.
L Acceptance ' criteria pertaining to the design of -the containment emergency sumps are provided in SRP section 6.2.2, Containment Heat
-Removal Systems,- Revision 4. These. criteria-address the drainage-of containment spray water and-emergency core cooling water to the recirculation _ suction - points - (sumps) and the screen ass.emblies
- surrounding these suction points. Regulatory Guide 1.82. Water
. Sources for Long-term Recirculation Cooling Following a Loss-of-Coolant' Accida.nt, Rev. '1, provides the guidelines for the design.of the?IRWST and the HVT, and the oesign of.the screens associated with these tanks.- Technical-considerations related to this' issue are _ detailed in NUkEG-0897,-Containment Emergency' Sump Performance, Revision'1.
2.9-1 L
. , , , - ., av.-
77a .wp (9213) bh 2.9.3 DISCUSSION Water introduced into the System 80+ containment from a RCS break or from containment sprays drains into the Holdup Volume Tank (HVT). This tank serves the purpose of the " containment sump." The Holdup Volume Tank is therefore the low collection point in containment. The contents of this tank are directed to the IRW5T through the two IRWST spillways (see Figure 2.9-3). The IRWST serves as the single water source of long-term recirculation for emergency core cooling and containment heat removal. With the System 80+ design, it should be noted that the IRWST does not serve as the containment sump; this tank specifically serves as a storage tank for refueling water, a clean and reliable source of water for Safety Injection, and a heat sink for condensing steam discharged from the pressurizer. The arrangement of the IRWST within containment meets the multi-sump requirement of Reg. Guide 1.82, Water Sources for Long-term Recirculation- Cooling Following a LOCA. . The general plant arrangement separates redundant trains of the SIS and the CSS. The divisional boundary provides complete separation between divisions and effectively creates two identical support buildings. The result is a plant arrangement with two SI pumps and one CS pump in each division. Within each division, the two SI trains (and each CS train) are separated by a quadrant wall to isolate the trains from each other to the maximum extent practical. Each of the four L SI pumps has its own suction connection to the IRWST (see Figure 2.9-1) and each of the two CS pumps shares one of these four connections. Following an accident, water introduced into containment drains to l the Holdup Volume Tank. Debris that may exist in containment may i' be transported to the HVT with this fluid, Debris greater than 1.5 l inches diameter is prevented from entering the HVT by a vertical trash rack, which is located at the entrance to the HVT (see Figure 2.9-3). The vertical trash rack is greater than six feet high and more than forty feet long. A debris curb exists at the base of this trash rack-to prevent high density debris that may be swept along the floor by fluid flow toward the HVT from reaching the trash rack. The vertical orientation of the trash rack will help impede the deposition of debris buildup on the screen surface. Particles that are smaller than the trash rack mesh will enter the Holdup Volume Tank.
-The Holdup Volume Tank is designed to function as a solids trap to help prevent debris from entering the IRWST. High density debris that makes its way through the trash rack will accumulate in the ' bottom of this tank. The IRWST spillways are located at a hign enough elevation to assure that much of the higher-deIsity debris (and debris that tends to sink slowly) wit' settle to the bottom of 2.9-2
770.wp(9212)bh I the llVT bMore spilling over into the IRWST. Dobris that remains i in suspension will make its way to the IRWST spillways. The spillways are.shown in Figure 2.9-3. Screenn are not present in those s- '11 ways to assure uniteter,upted flow to the IRWST. l The finu debris that is introduced into the IRWS? is provented from ontoring the SIS suction }. ping by a debris screen. These scroons are located at each end of the f our wing walls that serve as supports for the reactor coolant pumps (sco Figures 2.9-1 and 2.9- , 2). These wing wall assemblies extend from the IRWST floor to the maximum IRWST water level, assuring that all dobris will be filtered before reaching the SIS suction lines. The screen - assemblies completely enclose the suction lin; & by running from the i end of each wing wall t,a the side walls of the iloldup Volume Tan); or the primary-shield walls, as applicablo. The wing-wall screens havo the capability of removing particles greater than 0.09 inches i diameter This screen size is consistent with the screens used on currently operating units. Tho wing-wall screens are the final barrior to debris before the SIS suction lines. j Diockage of the dobris screens is a major concern with respect to recirculation. The System 80+ screens have a vertical orientation to prevent debris fro,a settling on the screon surf acos. This helps
'in keeping the acroons cicar. The design considered the types and quantities of insulation used for the System 80+ components, sinco post-LOCA deterioration of this insulation is the major potential source of debris in containment. The location of insulation with respect to 4 IIVT and IRWST as well as the possible location cf !
breaks hav Iso been corsidered. The offectivo areas of the screens hav. .een determined according-to the guidelines provided in Appendix . to Regulatory Guide 1.82, Guidelines for Review of l' Sump Design and Water Source 'or Emergency core coo;ing. The. debris screens ~have been designed to withstand the vibratory motion of a seismic event without loss of structural integrity. , Each screen is capable of withstanding loads impoced by postulated i missiles as well as loads due to pressure head difforentials. Consideration has also been given to the materials used for the debris scree s. Materials have been selected to avoid-degradation j during perisds of inactivity (i.e., no submergence), and during ( -periods in which tho screens are partially-or fully submerged. Each - screen- used in the System 80+ design is provided with an access opening to allow for inspection of=the racks or screens.
.The screens will be visually examined periodically to detect any corrosion or satructural degradation during refueling outage ~
periods. As seen in figure 2.9-1, the suction lines are located
- within the confines of the wing vall away from the IRWST spargers.
This wall dewign helps isolate-the suction lines from the open sections of the IRWST, where most of the maintenance activities 2.9-3 l
77a.wp(9212)bh i will be pnrformed. The fine wing-wall screen will filter any trash generated from this type of activity. In the event that maintent. ace is needed within the wing walls and near the ECCS suction inlets, pesmanent box-like screens over the auction niping will protect these lines. Long-term return of spray water from upper Invel elevations is not dependent on individual piping runs or spillways. Multiple passive spillways are provided to route water back to the lloldup Volume Tank. Major openings such as hatches and stairwells are also available to return water to the screened entrance to the llVT. protection against air ingestion by SIS pumps is also a major concern with respect to recirculation and has been considered in the System 30+ design. The location and size of the suction lines in the IRWST have been chosen such that air entrainment is minimized, pump air ingestion analysis is based on minimum submergence, maximum Froude number, and maximum pipe velocities. The available surface area used in determining the design coolant velocity has been calculated conservatively to account for blockage that may result as per Reg. Guide 1.82, Appendix A, Guidelines for Review of Sump Design and Water Sources f or Emergency Core Cooling. The minimum water level in the IRWST has been conservatively calculated to be 75+6 (eleve' ion). This water level allows for sufficient NpSil for the C ntainment Spray punpa and Safety Injection pumps operating at runout flow. A concorvative margin has been provided between the elevation of the suction piping opening and this minimum water level to minimize the pos.sibility of air ingestion. Applying the parameters of the IRWST to the equations in Reg. Guide 1.82, Appendix A, yields zero air ingestion at normal pump flowrates and less than 2% air ingestion at pump runout flowrates. The IRWST suction lines are also provided with vortex suppressors to aid in minimizing air ingestion by the SIS _ pumps. The guidelines in Appondix A of Reg. Guide 1.82 regarding the design of these vortex suppressors have been considered. During normal full power operation, it is possible to perform a full flow test of the SIS and CSS pumps while taking suction from the IRWST and returning to the IRWST via a recirculation line (see SIS p& ids in CESSAR-DC, Figures 6. 3. 2-1 A, B, C) . This testing can verify the satisfactory hydraulic performance of the IRWST by running the pumps at runout flow. 2.9.4 RESOLUTION The design of the System 80+ IRWST and ilVT assures that a clean and reliable source of borated water is available for ECCS recirculation. The arrangement of the IRWST within the System 80+ containment of fers advantages over conventional sumps. Like sumps, the tank serves as the single source of water for SIC and CSS pump recirculation, but the protection afforded the SIS pumps against 2.9-4
77c.wp(9212)bh h debris ingestion or blockage is significantly greater than in I current designs. First, w#or in containment draining back to the i IRWST must pass through a largo trash rack before entering the HVT. i The HVT sorvos as an effective solids trap for high density debris. Lower density debris that makes its way into the IRWST via the IRWST spillways encounters debris screens that filter fine ' particles from the SIS suction inlets. Each of the four SIS pumps have separate IRWST suction lines and each of the two CSS pumps tt.kes suction from one of those four lines. Box screens at all four suction lines provido a final trap. Multiple spillways are available to return water from the upper containment olevations to the IRWST. The drain pathways are fully rodundant to assure recirculation capability. The location of the suction -inlets within the IRWST provide additional protection against suction inlet damage and/or blockage. Consideration has been given to IRWST hydraulic performance, the generation of potential debris or) associated effects (including debris screen blockago), and the preservation of SIS pump NPSH during post-LOCA conditions in the overall design. The parformance
.of the design is deemed acceptable with respect to these
[ considerations. This' issue has been resolved by design features of System 80+. 2.9-5
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l 77c.wp (9212)bh l 2.10 EZEXQlj_QP PWR UPPER INTERNALS 2.10.1 ISSUE Evonts witt. the potential for loss of Decay Heat Removal (DHR) have initiated from plant configurations with the reactor vessel head removed, the refueling pool filled with water and the reactor upper internals still in place. Under these conditicns, the reactor vossol upper internals may provide sufficient hydraulic resistance to natural circulation flow betwoon the refueling pool and the reactor coro to inhibit, or even provent, the refueling pool water from cooling the core under circumstances wher forced convection DKR has been lost. 2.10.2 ACCEPTANCE CRITERIA When the reactor vossol head is off and the coro and upper internals are in the vessel, any one or more of the following conditions shall be satisfied:
- 1. Demonstration by analysis that the time to boil exceeds the time required to evacuato and establish containment integrity; and/or
- 2. Demonstration by analysis that either a natural or forced circulation flow path, with or without heat exchangers, can be established to perform DHR for a sufficiently long period of time to allow plant operators to terminate the event.
2.10.3 DIDCUSSION In mode 6 configurations with the vensel head removed, loss of shutdown cooling events could be of concern if the uppar internals inhibit natural circulation cooling of the core via-the heat sink in the upper refueling pool. An analysis to predict the extent of natural circulation flow through the Upper Guido Structure (UGS) is described below. Resolution of this issue is based on the results of that analysis. Hydraulic flow resistance data are provided in Table 2.10-1. These were utilized in an analysis to ascertain the possibility of natural circulation flow communication thrcugh the upper internals between the core and refueling pool for a Moda 6 configuration with the vessel head removed. The entry flow path through access holes in the core support barrel flange to the upper downcomer from the refueling pool was shown to admit negligible flow in view of its high hydraulic resistance. Equal flow areas for the down-flow and up-flow were assumed through the core and upper internals with turnarouna at the base of the active core. An iterative algorithm, based on the principles of cor.servation of mass, momentum and energy, was used to determine the t 2.10-1
77a.wp (9212)bh transient natural circulation flow, after loss of shutdown cooling, and the time to saturation at the core outlet. Sensitivity of the results to magnitude of the upper internals flow resistance was also determined. Several key assumptions made in the analysis include one-dimensional flow with no transverse acmentum and energy exchange, no conduction loss through the vessel, no convection from the surface of the refueling pool and no heat storage within the bounds of the upper internals. The decay heat generation was conservatively assumed constant, characteristic of two days after shutdown, with an additional 10 percent uncertainty. No credit was taken for the available need developed at the core exit due to the forty foot elevation difference. Other simplifications included a uniform - refueling pool temperature together with equal metal and fluid temperatures in each core flow zone that actually exhibit a linear variation axially. This analysis indicates that natural circulation flow mak6s the water in the upper refueling pool available for cooling the core and significantly extends one time required for the core outlet temperature to reach saturation under these circumstances. Fluid temperature versus time variation results as a function of initial coolant temperature are rhown in Figure 2.10-1. The curvus in the figure for the maximum Mode 6 initial coolant temperature of 135'F indicate that the core outlet temperature would not reach saturation i until 35 minutes after the loss of decay heat removal. The , calculated natural circulation flow rate increases throughout this time period with a value of about one percent of the rated design core flow rate at one minute after loss of DHR rising to 1.5 percent - where the outlet temperature reaches saturation at 35.5 minutes. For an initial coolant temperature of 100'F, the time to saturation j increases significantly to 125.5 minutes. A sensitivity evaluation shows that it would take only a few minutes to reach saturation at the core outlet in the complete absence of ; flow communication with the refueling pool. Another analysis indicates an additional time period of at least one hour to core uncovery after saturation is reached throughout the entire core volume in addition to the core outlet. These results make it possible to conclude that the presence of the upper internals does not prevent natural circulation communication with the refueling pool. _ Depending upon the initial fluid I temperhture, the time-to-boil will exceed 35 minutes and may be- r several hours in duration. Ample time, therefore, exists for the implementation of operating procedures to deal with the restoration of forced circulation decay heat removal and/or to begin the process of containment closure, which requires 10 minutes when in Mode 6 configurations. 2.10-2
770.wp (9212)bh IABLE 2.10-1 NYDRAULIC FLON_BKSISJANCE DATA FOR SYSTEM 80+ Dimensionless Cross-sectional ; Recion flow ResistalLqt_E M w Area - Ftz ___ Core. Support Barrel 1823.0 2.949 i Flange Entry l Upper Downcomer 0.17 40.53 Lower Ocwncerer 2.644 33.53: Inlet Plenum 14.0 115.6 Cote 12 77 60.8 - Fuel Alignment Plate 0.69 20.6 Upper' Guide structure 1.246 3.34 2.10.4 RE80LUTION The results discussed above indicate the availability of a natural . circulation flow path through the upper internals in the vessel in the-event of a Icss of both SCS trains during Mode 6 when the i refueling pool-ic. full. The heat removal effected by transference i of core-generated decay heatLto the upper refueling pool would allow
- plant. operators;a. minimum time period of at least-35 minutes to terminate the. event before the core outlet temperature reached i saturation. Substantially more time would be available before possible uncovery of--.the core. This time period is of sufficient-duration that the_ issue can be considered resolved on the basis of both of the previously stated acceptance critoria. It_is clearly l longer than the minimum Mode 6 containment closure time of 10
. minutes and sufficiently long enough to allow plant operators to
- restore shutdown cooling, i
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77a.wp(9212)bh l l 1 2.11 EHKL_1[h)3QLJ NG AHD_}i EhV Y._ LQADR 2.11.1 ISSUE Questions have boon raised regarding the potential for damage to fuel and safety related equipment due to dropping of heavy loads during plant shutdown. Related issues involve the transport of heavy loads within the reactor containment building and the spent fuel-building. These include dropping the reactor vessel closure head and internals, dropping the head area cable trays (11 ACTS ) , accidental release of a fuel assembly, and movement of the spent i fuel storage cask. Drop accidents involving primary NSSS piping are not considered, since by design piping is routed beneath the reactor refueling pool. 2.11.2 ACCEPTANCE CRITERIA Fuel and safety related equipment shall not be subject to damage that may adversely effect public health. Also, fuel assemblies located within the reactor or within storage racks shall remain
- subcritical during and following postulated load drop accidents.
2.11.3 DISCUSSION The transport of heavy loads within the containment building and the spent fuel building is controlled by integrating relevant design characteristics for the building and the handling equipment. Plant layout, equipment design and handling procedures are chosen to insure that heavy loads are restricted to preassigned travel zones. Equipment interlocks and procedures are also used to insure that load transport is accomplished in a predictable manner. Specific issues associated with the transport of heavy loads within the containment building include movement of th. reactor vessel closure head, the reactor internals, the 11 ACTS, and individual fuel assemblies. Special measures are taken to safeguard these operations and mitigate the consequences of postulated load drop accidents. Procedural guidance for raising the reactor closure head, as provided-in CESSAR-DC Section 9.1.4.2.3.3, specifies that the fuel transfer tube valve be closed and that the pool water level follow the vertical movement of the closure head as it is raised from the _ reactor. This _ insures that __ the _ containment building remains isolated from the spent fuel pool building during transport of the closure head. Also, by isolating the containment building from the spent fuel building: the spent fuel pocl_ is- protected against drain down that might occar as a result of a postulated drop accident.
-2.11-1
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77a.wp(9313)bh i Evaluations have been performed which demonstrate that a postulated head drop, f rom its specified maximum lif t height, onto the roactor vessel will not result in a significant risk to public safety. . Though the reactor vesael and internals may sustain damage, the reactor vessel sill remain filled and the fuel will remain covered i and in a suberitical configuration. Evaluations of the reactor . inturnals demonstrate that a drop accident involving the internals ! would be less severe than the postulated head drop accident. Travel paths for the closure head and the internals, leading from the reactor vessel-to the respective storage stands, are arranged so that the transported loads do not pass directly over the-ICI i seal table (Refer to Tigure 2.11-1). If it is postulated that : portions of these structures do impact the seal table, seal ! housings and guide tubing above the seal table, the resulting l damage would be localized to these components. Under these
- conditions, the water level within the vessel will' remain at the flange level. In addition should the reactor cavity pool seal be ,
d6maged to the extent that there will be significant pool drainage, > the reactor vessel will romain filled. The refue2ing machine is structurally designed to withstand the ef fects of design basis seismic motions. - In addition, this machine i is provided with interlocks which restrict machine movements to permissible zones as well as lock the fuel grapple in place. The ' refueling machine is designed to transport one fuel assembly at a time between the reactor core and the fuel transfer system. It is also designed to transport CEA rod and ICI aisposal containers between an intermediato storage rack and the fuel transfer system. The grapple for the disposal containers is the same design as the one used for fuel assemblies. The refueling machine is designed so that it can not pass over the top of.the 1CI seal table. This precludes the possibility of load drop accident involving _a fuel assembly falling onto the ICI seal table. Also, during normal refueling operations the travel path is . restricted so that it passes over the reactor cavity pool seal at predetermined locations. 'As a minimum, the pool seal is designed so that it will withstand without leakage a postulated fuel drop accident in these zones. If, for other postulated reasons, there is significant drainage of the pool, it is possible to rapidly lower a fuel assembly on the refueling machine grapple to-an elevation which insures that it remains submersed in water. The assembly may be inserted into the reactor vessel or lowered into the deep 4 end of the refueling pool, adjacent to the fuel transfer system. x The head area cable tray assembly (HACTS), which is used to route power and signal lines away from the reactor vessel closure head, i is designed to be handled by t' - auxiliary: hoist on the polar crane. The cable trays are ra . vertically by this hoist from 2.11-2
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77c.wp(9212)bh f s the installed position over the reactor vessel and moved to a storage position on top of either of the two steam generator walls. The trays are handled by four separate slings that are fastened to specially designed lift fixtures on the E.ructural frame of the l MACTS. All lifting components are designed in accordance with the criteria of_NUREG-0612, Control of Heavy Loads at Nuclear Power plants. Prior to movement of the cable trays, the reactor must be in the shutdown mode and depressurized. During a postulated load drop accident involving the impact of the HACTS onto the reactor vessel, the maximum impact energy is estimated to be about twenty percent of that associated with the i reactor vessel closure head drop. ,. Though the reactor vessel may be damaged, the level of damage to the vessel and its support would be less severe than that associated with dropping the reactor vessel head. Furthermore, t since the- cable tray assembly is not a rigid structure, an ' appreciabic fraction of the impact energy would be dissipated ' during plastic deformation of the llACTS itself. For the postulated dropped cable tray accident, it is most probable that the HACTS would impact the closure head lift rig and the CEDM ' pressure-housings. These structures are likely to be permanently deformed by bending ar1/or buckling. In some instances the pressure housings may alao leak. The extent of damage, however, would not be suf ficient to cause the reactor vessel to drain down, nor to adversely affect core criticality. As with the containment building, special consideration is given to the transport of heavy loads and fuel assemblies in the fuel building. _ Also restrictions regarding the transport of heavy loads i over fuel storage racks, and movement of the fuel shipping cask apply (Refer to Figure 2.11-2). Transport of the fuel shipping cask within the spent fuel building is accomplished using a special high capacity hoist. The cask is transported using a staggered lift from the wash down area to the laydown -area, where fuel loading takes place. This is done to limit the maximum drop heityht for the respective regions. In each case the floorc and walls have been designed to withstand a , postulated cask drop accident. The spent fuel pool is connected to the cask laydown area by a
-gate, which will be closed to isolate the-two zones during cask movement. The elevation of the gate is specified so that fuel
- located in the spent fuel storage racks would remain submerged following a postulated pool drain down through the gate.
The hoist iused for transport of the fuel shipping cask is mechanically interlocked to prevent travel over the spent fuel 2.11-3 _ --- _ _ , __ _ ._~ _ _ . - -_ - . - -
77a.wp(9212)bh pool. This interlock prevents the possibility of inadvertent movement of heavy loads over the spent fuel storage racks. New fuel enters the fuel building through a designated unloading area. It is handled and transported to new fuel storage racks by an intermediate capacity hoist. The lift height of the hoist is rescricted to limit the maximum drop height of the fuel and tool onto the new fuel storage racks. Like the fuel storage cask hoist, this hoist is also mechanically interlocked to prevent travel over the spent fuel pool. The fuel handling machine used in the fuel building is similar in design to the refueling machine. It is structurally designed to withstand seismic excitations. Also, it is provided with - interlocks to control the movement of fuel within the pool. Both the new fuel and spent fuel storage racks are designed to withstand impact energies associated with postulated fuel drop accidents. They are designed to limit damage to the stored fuel and to maintain it in a subcritical configuration. Plant operating procedures also restrict the transport of loads over the fuel storage areas so that they do not exceed the design requirements for the storage racks. The consequences of dropping a spent fuel assembly in the spent fuel pool have been evaluated and the results presented in CESSAR-DC, Section 15.7.4. It has been shown that a postulated accident of this type would not present a risk to public health. 5 2.11.4 RESOLUTION The issue of fuel handling and heavy loads is resolved f or System - 80+ by the equipment design and building layout which satisfy applicable criteria and provide physical limitations to rovement _ and by administrative limitations in Chapter 9 of CESSAR-DC. The design and layout of the System 80+ plant incorporates features that have proven to be successful on other plants. Appropriate ir.provements have been introduced which provide an additional margin of safety during handling operations. 2.11-4
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l 77c.wp(9212)bh 2.12 Fj2TINTI AL FOR DEAU11HG_1HE_REAclRR_C00LhNT SlplEM 2.12.1 ISSUE The issue is the risk of losing primary coolant from the reactor coolant system during Modes 2 through 6 (shutdown, hot standby, hot shutdown, cold shutdown, and refueling). The safety significance of draining the coolant from the reactor coolant system during shutdown is that such an event can directly lead to voiding in the core and eventual cord damage. The draining of the reactor coolant system may also lead to a loss of decay heat removal cooling capability which in turn could lead to core uncovery. 2.12.2 ACCEPTANCE CRITERIA The criteria utilized to evaluate the adequacy of the System 80+ design with respect to the potential for draining the RCS are prevention, detection and mitigation. Provention is the preferred criteria but in some instances detection and mitigation are to be provided. 2.12.2.1 Er_ovention Critetia
- 1. The design shall prevent or inhibit the draining through the use of isolation valves, interlocks, and system alignment restrictions during the various modes of plant operation.
- 2. The design shall minimize the potential for component f ailure, inadvertent action, or human / operator error to result in the rapid draining of the reactor vessel. Redundant components shall be provided as appropriate. The design shall provide instrumentation, overview displays, and alarms to clearly supply the operator with equipment status specific to shutdown modes.
- 3. Technical specifications and procedural guidance shall be provided to the plant owner / operator to assist in identifying plant conditions and configurations in Modes 2-6 that could result in a potential primary coolant drainage event.
2.12.2.2 Detection Criterig The design shall have the capability to detect and monitor a drainage event. Drainage pathways that could lead to a loss of decay heat removal or core uncovery shall be considered. Appropriate instrumentation, displays and alarms shall be provided. The adequacy of System 80+ instrumentation to detect drainage events shall be confirmed and additional instrumentation shall be provided as necessary. 2.12-1
77c.wp(9212)bh 2.12.2.3 tiiligation critelin
- 1. The design shall have the capability to mitigate the loss of primary coolant from the reactor including isolation of a drain path and the ability to provide a source and path for sufficient make-up.
- 2. Technical specifications and procedural guidance shall be provided to identify potential make-up water injection sources and paths in the event a drainage path does occur. Recovery actions shall be specified.
2.12.3 DISCUSSION Primary coolant can drain from the reactor vessel due to a path directly from the Reactor Coolant System (RCS) or by way of paths through systems interfacing with the RCS. Plant -modes of operation characterize the potential for draining the reactor coolant system (the alignments and conditions, such as pressure, temperature and flow which can exist within and between the RCS and interfacing systems) and the rate at which such draining can occur. The causes of RCS draining (initiators) can be categorized into two groups. The first gre"p includes components and equipment that fail to operate as intended. This could, for example, result from equipment malfunction (e.g., stuck open relief valve). The second group includes operator error such as misoperation of valves or pumps. There are several. key factors that affect the probability and the consequence (i.e., the risk) associated with an initiator. Consideration of these factors can aid .in the development of procedures for prevention, detection, identification, and mitigation or termination of such events. These factors are the plant configuration, the ability to respond to the event, and the characterization of the initiator. These are summarized in Table 2.12-1. l l l 2.12-2
770.wp(9212)bh TABLE 2.12-1 FACTORS WHICH_ AFFECT THE RIRE_ hag _oCIAIXD WITH AN INITIATOR Plant Configuration... System alignment with the RCS. Initial water level in the RCS. Availability of mitigating systems. Use of temporary seals (e.g., nozzle dams) Ability to Respond... Detection of a draining event. Identification of the initiator. Termination or mitigation of the event. Characterization of the Initiator... Probability of the initiator. Rate of drainage from the RCS. Plant configuration is most notably characterized by system alignments with the RCS during various modes of shutdown operation. These alignments define the possible drain paths from the reactor vessel. The initial water level in the RCS influences operator response time. The use of temporary seals in the RCS (e.g. , nozzle dams) and interfacing systems during maintenance and refueling activities can either be the source of an initiator (e.g., if the dam fails) or preclude mitigating actions by defeating injection paths (e.g., if an in place dam blocks circulation). Plant configuration is also characterized by the availability of mitigating systems. Mitigating systems must have a sufficient source of borated water and the ability to deliver this water to the reactor vessel at a rate greater than or equal to the rate at I which water is draining from the RCS. Finally, maintenance l- activities must not preclude mitigating systems from being used to l respond to the event. l Another factor that can affect the risk associated with an initiator is the ability to quickly determine that a loss of ! coolant is occurring (detection) and the source of the loss (identification). This defines instrumentation and control l requiremants to respond to draining events. l l The analysis presented in this section was conducted ,y examining design drawings and piping and instrumentation diagrams for System 80+ to define potential drain paths from the reactor vessel and to 2.12-3
--~ ~ . . - - . - - - . _ - . - - - - - - , - . - - - - - - .
77a.wp(9212)bh t consider these drain paths in the context of the factors shown in Table 2.12-1. A potential drain path is any opening in the RCS (seal, manway) or any interfacing system piping path that can take ; primary coolant away from the RCS. Many openings and interfacing systems are designed specifically to allow fluid to leave the RCS or interfacing system (e.g. , relief valves) or to circulate primary coolant for normal letdown, purification, charging (CVCS), sampling (SS) activities or shutdown cooling (SCS). A shutdown risk drni: A th results in an unplanned loss of primary coolant from the R , .Ith the accompanying lowering of the water level - in- the react a vessel. The path could be short (seal leakage) or relatively long (via piping of an interf acing system) . . The driving head for the draining flow will depend on the relative , conditions between the coolant in the reactor vessel and the conditions at the end of the drain path. This includes RCS pressure, elevation head, pump head, and backpressure. For purposes of this analysis, a shutdown risk drain path is categorized as either major or minor. A major drain path is one that, based on analysis for certain specified plant configurations, could result in a rapid drain flowrate. The preferred recovery , would be to isolate the drainage source before the RCS water level reaches the break level-and to add makeup to the RCS. It is , possible, however, that such a major path could drain'the RCS to the bottom of the hot leg elevation too rapidly before the operator could take mitigative action. Such an occurrence could result in a loss of SCS due to insufficient water level. The identification > of major drain path does not imply that such a path is likely or probable for System 80+, but only that it'is possibla and that its potential consequence requires _special attention be directed to prevention and recovery procedures. Minor drain paths are those that result in a drain flowrate that can be compensated for using i available make-up systems, or are otherwise insignificant.
- The identification of a minor drain path requires no further [
action. The identification of a major drain path requires specific j procedural guidance to aid tie operator in avoiding this path. It
- also results in providing the operator the means and. guidance to recover from such an event. The means to detect, mitigate and recover from such postulated major drainage events are those described in Sections 2.3, 2.4, and 2.8 of this report. ,
1 2.12.3.1 _ Potential Drain Paths Directit from t.he Reactor l CoolaAt System Potential drainage paths directly from the Reactor Coolant System are associated with the reactor coolant pumps, the steam generators, the In-Core Instrument Seals and the reactor cavity seal. Note that reactor cavity seal leakage is not an actual RCS leakage but it'is a form of inventory loss during Mode 6. Table 2.12-4
770.wp(9212)bh l i 2.12-2 identifies those paths which are major (i.e., have the potential to rapidly drain the RCS to a critical water level) and minor (i.e., can be controlled by available systems). l l l TARLE 2.12-2 POTENTIAL DRAIN PATHS DIREQTAX IROM THE REACTQR_QAQLANT SYSTJlLM Major Drain-Paths Steam Generator Nozzle Dam Failure Steam Generator Manway Opening Minor Drain Paths Reactor Coolant Pump Seal Leakage ICI Seal Table / Housing-Leakage Steam Generator Tube Rupture Reactor Cavity Seal Leakage 1 l l j 2.12-5 o l
77a.wp(9212)bh 2.12.3.1.1 Major Drain Paths Directly from RCS The failure of temporary steam generator dams and the opening or leakage through steam generator manways could lead to a rapid loss of primary coolant perhaps to an RCS water level at the bottom of the hot legs. A discussion of steam generator nozzle dam integrity { is presented in Section 2.3.3.3 of this report. The System 80+ design includes a requirement to establish a mid loop vent pathway before operating in reduced inventory. When opened to the containment atmoshphere, it provides sufficient venting capacity to prevent RCS pressuritations and subsequent dam failure. Guidance to address the safety and risk aspects of nozzle dam installation __ timing is presented in Section 2.3.3.3 of this report. procedural guidance will preclude manway openings that would lead System 80+ { to a loss of reactor coolant. 2.12.3.1.2 Hinor Drain Paths Directly from the RCS There are three stages of RCP seals for each System 80+ RCP and each of the three seals are capable of operating at full RCS pressure. However, the reactor coolant pump can potentially suf fer some leakage from pump seals during any mode. Seal leakage is detected by a RCP bleed-off flow alarm (F-156, -166, -176, -186). The source of leakage can be identified to a specific pump. Drainage from RCP r,cals would be manageabic and can be compensated by available systems. 1 The System 80+ In-Core Instrumentation (ICI) system design employs instrument tubes that terminate in the refueliny cavity at an elevation several feet above the reactor vessel flange. The System 80+ design does not employ temporary thimble tube seals. Even so, there has been concern expressed in Section 6.7.2 of NUREG-1449, (DRAFT) that evolutions could exist that would provide a potentially significant flow path between the bottom of the reactor vessel and the top of the seal table, particularly if the RCS is pressurized. Table evolutions are prohibited by procedural guidance (see also Section 2.3.3.5 of this report) while the vessel head is on and mid-loop evolutions are in progress, thus, preventing seal leaks. An evaluation of heavy load handling for System 80+ relative to ICI seal table in discussed in Section 2.11 of this report. Travel paths for the closure head and internals from the reactor vessel to the respective storage stands are arranged so that the transported loads do not pass directly over the ICI seal table. This precludes the possibility of a load drop accident falling on the ICI table. A steam generator can leak primary coolant through tube failure during any mode when the RCS is pressurized. An evaluation of steam generator tube rupture initiated in a shutdown mode is presented in Section 4.6.3 of this report. The leaket from a a 2.12-6
770.wp(9212)bh ruptured tube during shutdown is minor relative to reaching a RCS water level at the bottom of the hot legs. Reactor cavity seal failure could be postulated during Mode 6 when the refueling pool is full. Although such a drainage could bo a concern relative to any f uel being transported, the drainage would be self limiting to level of the reactor vessel flange and the SCS 1 will be uninterrupted. I l 2.12.3.2 Potential DralDace k aihs Throuch Interf aginc 8ystema Potential RCS interfacing system drainage paths from the Shutdown ) Cooling System (SCS), the Safety Injection System (SIS), the Chemical and Volume Control System (CVCS) and the Sampling System . (SS) are illustrated in Figures 2.12-1, 2.12-2, 2.12-3 and 2.12-4. l The originating points of the drainage paths are location on the Reactor Coolant System. The paths include piping, valves, pumps, heat exchangers and orifices. The greek symbol 4 represents the piping outer diameter. The terms IC and OC are used to show whether a path segment is inside containment or outside containment, respectively. The " letter-number" number designation in the boxes generally refers to system valves which are of the motor operated, manually operated, check or relief type. The letter "F" followed by a number referos to a flow measuring device. Those paths represented by the darker / bold lines identify major drain paths. The end points of these paths are various tanks or syctoms assumed to be open for maintenance. An assessment of these paths was made relating potential flow rates and the time it would take for the RCS water level to reach the bottom of the hot-legs. An assessment of.the relative probability, of single versus multiple valve misalignments or failures was not considered at this time. All paths were considered possible, even though it is recognized, multiple misalignments are less likely. 2.12.3.2.1 Potential Drainage Paths From Tt? RCS Through the SCB (SCS Trains From RCS Hot Leg i Through DVI Nozzle 1A and From RCS Hot' Leg 2 Through DVI Nozzle 1B) This discussion is specific to train 1 but applies to train 2 as well. This overall path represents one of the two normal shutdown cooling trains. As long as this path is isolated from the RCS during Modes 2, 3 and part of 4 loss of coolant from the RCS through the SCS is prevented. While the SCS is in operation during part of Mode 4 and during Modes 5 and 6 there exists potential paths from the SCS through which the primary coolant passing through the SCS could be drained. The main initiators to open a drain path would be an overpressurization leading to the lifting of a relief valve (which 2.12-7
77a . wp ( 9'212 ) bh ; i fails to rescat) and/or operator error opening a valve or series of valves to an open system. A valve failure is another possible initiator. An open system is a system which has boon drained for maintenanco activities. l I The SCS suction piping contains tuo motor operated isolation valves (SI-651 and SI-653) in series insido the containment (soo Figuro 2.12-1). Those valvos are closed during Modos 1, 2 and 3. In ' addition, there is also a motor operated valve (SI-655) outsido the containment. All throo valves can be operated from the control , room and have position indication in the control room. An alarm exists to notify-the operator if the two motorized valves inside containment are not fully closed coincident with high RCS pressure. Operator actions require that the RCS be-depressurized below the maximum pressure for SCS operation in order to clear the permissivo i SCS interlock. Therefore, drainage from the RCS through the SCS to , interfacing systems in the normal flow path direction is provented during Modos 2 and 3. Flow will occur through the SCS train during part of Mode 4 and throughout Modes 5 and 6. The potential drain paths from the SCS-as shown in Figure 2.12-1 are summarized in Table 2.12-3. The 30 paths of Figuro 2.12-1 have been broken down into groups of similar drain path characteristics. ! 9 l t l' l l l l l 2.12-8 , l,.___.- _ - _ . . _ . , _ ___._,__.i_____..__-._,._._,__._._
-_ - . _ . . -m . _
, _ - - _ _ - , . ~ . _ _- .. - ._- ._
770.wp(9212)bh l l TA R R_2412-3 GROUPING OF PillHARY COOLANT D MIyh0E PATHS FROM Tl[E_E_CJ Kinor Drain Patha Group Paths A. Thermal Relief Valve Discharge 1, 3, 18, 19, 21, 22, 25, 27, 28 B. Paths to the Sampling System 6, 12 C. Paths to the CVCS 5, 8, 9, 15, 17
.D. Paths to the CCWS 10, 11, 14 E. Paths to SIT #1 23 F. Paths to IRWST, RDT, and EDT 24, 26, 29, 30 Through Manual Valves and Small Piping Maior Dggin Patha G. Paths Thrcagh Motor Operated Valves 4, 16 and Large Piping to the IRWST H. SCS Suction Relief Valve Discharge 2 I. Paths Through large Manual valves 7, 13 to CS Pump Suction and Discharge J. Path to the Refueling Pool 20 Group A through F paths are minor paths and include relief valves, paths through small piping (1/2 in - 3 in O.D.) to assumed "open" systems, Sc pump seal leakage, SC mini-ficw heat exchanger tube leakage, SC heat exchanger tube leakage and a postulated path to the SITS.
The paths of groups A through F, if they were to occur and lead to a loss of primary coolant, -could be mitigated with available equipment and mcke up sources during Modes 4, 5 and 6. The discharge would be slow enough such that a loss of primary coolant to the bottem of the hot legs should not occur before-detection and mitigation have been accomplished. No new design features, technical specifications or procedural guidance are identified f or l the paths in these groups. l l l t 2.12-9
. , ~ . - .
770.wp(9212)bh Groups G through J represent major paths that, if established, could result in a rapid loss of reactor coolant during Modes 5 and '
- 6. Depending on the availability of equipment and systems to perform mitigative action during these modes such a rapid discharge may lead to an RCS water level at the bottom of the hot legs with a concurrent loss of both SCS trains. Procedural guidance to aid the operator in addressino these paths is specified in Section 2.1. 4 The recovery action for this scenario is discussed in Section !
2.4.3.1. 2.12.3.2.2 Potential Drainage Paths from the RCS to the SIS /SCS (From DVI Nczzles 1A, 2A, 1D and 2D) The potential . drain paths from the RCS through the SIS /SCS, originating at the DVI nozzles, are shown in Figure 2.12-2. The potential drain paths presented in Figure 2.12-2 are " reverse" paths relative to the normal safety injection flow direction. As a result, in order to assume RCS drainaga from the DVI nozzles, it is postulated in this analysis th ", multiple failures of check valves occur. The DVI nozzles in System 80& are located several (4-5) feet above the hot leg and cold legs. Therefore, loss of coolant through a DVI nozzle would be self limiting. The level of primary coolant would stabilize at the level of the DVI nozzle, and not result in a loss of shutdown cooling. No new design features, technical specifications or procedural guidance have been identified for these paths associated with flow from DVI nozzles. 2.12.3.2.3 Potential Drainage Paths from the RCS to CVCS The potential drain paths from the RCS through the CVCS are shown in Figure 2.12-3. The potential drain paths presented in Figure 2.12-3 are the normal letdown, charging, RCP seal injection, RCP seal leak off, RCP seal bleed off and drain paths associated with the CVCS design. A major opening in the letdown or charging line needs to occer for any appreciable drainage to occur. The paths defined for the CVCS, if established, would be managesble with available make up sources during Modes 2, 3, 4, 5 or 6. This discharge would be slow enough such that a loss of primary coolant to the hot leg level should not occur before detection and mitigation have been accomplished. No - new design features, technical specification or procedural guidance are identified for paths associated with the CVCS. 2.12.3.2.4 Potential Drainage Paths from the RCD to the SS The potential drain paths from the RCF through the SS are shown in Figure 2.12-4. The potential drain paths presented in Figure 2.12-2.32-10
e - - 770.wp (9212) bh 4 are normal sampling paths. A major opening in the sampling lines would need to occur for a not loss of primary coolant to occur. The paths defined for the SS, if they were to occur, would be manageable with available make up sources during Modes 2, 3, 4, 5 or 6. The discharge would be slow enough such that a loss of primary coolant to the hot leg level should not occur before detection and mitigation have been accomplished. No new design features, technical specifications or procedural requirements are identified for paths acociated with the SS. 2.12.4 RESOLUTIO? The shutdown risk issue of the potential for draining the System 80+ RCS is resolved primarily by design features, technical specifications and procedural guidance to prevent a drainage event from occurring-and to allow the operator to recover in a timely manner if such an event occurs. The vast majority of potential paths reviewed were judged to be minor such that tne drain flow rate can be compenssted using available detection--and mitigating systems or are otherwise insignificant. System 80+ design features, technical specifi-cations and procedural guidance are sufficient for such paths. An examination of the potential drainage paths for various System
.80+ plant arrangements and operating configurations has provided candidate paths, that if atsumed to be opened, could lead to a rapid loss of primary coolant. The candidate paths primarily involve those opened by misoperation or misalignment of one or multiple valves by the operator. The importance of such potential major drainage paths to a shutdown risk scenario has led- to
_ procedural guidance (see Section 2.1) being specified to aid the operator in addressing these paths The issue af potential RCS drain paths is ultimately resolved, from a core uncovery prevention perspective, by the use of System 80' design features to detect, mitigate and recover from postulated loss of shutdown cooling events as described in Section 2.4 of this report. { f 2.12-11
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2.13 : FLOODING AND SPILLA 2.13.1 ISSUE
< Essential systems may be at higher risk'_for failure due to flooding and spills-during shutdown because of-the varied and interrelated-maintenance : activities that may' be in progress simultaneously.
Past events =have involved,'-forsexample,' spills-from the component cooling _ water _ system, service water- system, condensers,_ and refueling pool' seals. ..The issue' addressed here_is the potential for loss _of decay heat removal as a consequence 'of spills. iand'
- internal- flooding - that- may disable components of the shutdown ,
cooling system. 2 '.13 . 2 - ACCEPTANCE CRITERIA The flood protection design;will provide' separation of redundant equipment' to ensure decay heat, removal- (DHR) systems availability
- and capability = are not precluded due to flooding and spills.
2.13.3E ' DISCUSSION-The flood: protection '. provided - insures a boundary, ' of separation between redundant DHR1 systems. The separation includes' components and structures-to' prevent the migration of' water. Preventing the migration'of: water eliminates the potential for rendering redundant DHR equipmentLinoperable. ThecSystem180+ ~ design _provides separation and flood barriers to 4 prevent'the floed:of. redundant equipment. The design: features a
' divisional separation. This divisional separation is a wall in the
. Nuclear Annex and the' Reactor! Building _Sub-Sphere. The. wall forms a' barrier between the' Division 1Eand the Division 2 mechanical and electrical; equipment.- This wall contains' no unsealed penetrations below the 70'. elevation level. This wall is~along. column-linetl7
- -(see CESSAR-DC Figures 1.2-4 and 1.2-5, reproduced hereias' Figures
. 2 ; 13-1~ ' and 2.13-2). _ Additional ' separation ' of _the divisions -is.
'provided'by the floor drain systems.; The: sumps and floor drains located- in tho - Nuclear Annex :and 'the Reactor Building Sub-Sphere 9
' are -divisionally separated.- This ' design feature--prevents the "
migration of floodwater from one division to_the other through the
. floor drains.
t The Systems:80+ design utilizes flood' doors to provide separation within the same division. 'In the' Reactor Building Sub-Sphere,1the flood doors provide Equadrant separation, therefore equipment is' ' protected-from floods.within'the same' division. Flood doors-also , provide protection for Reactor. Building Sub-Sphere Quadrants A' and..
- B from flooding outside the sub-sphere. This protects-the Shutdown -
-Cooling ~ Systems from floods.that could. occur in the Nuclear Annex
, and: . migrate into the sub-sphere. Flood _ doors also prov ic'e
)
2.13-1 _ N* _ , _, ___- . - - _ _ ,
l 77a.wp(9212)bh protection for the Vital Electrical Equipment located in the t Nuclear Annex on elevation 50' (see Figure 2.13-1). The System 80+ does not have any raw water systems inside the Nuclear Annex or Reactor Building. This design prsvides a significant contribution to flood protection because the flood sources are finite. Two significant sources of water are the Component Cooling Water System and the Emergency Feedwater Storage Tanks. Emptying the entire volume of water contained in a division s.f either of these systems will not flood above the 70' elevation. Therefore, no migration of water to the other division or to other protected areas (e.g., electrical equipment) will occur due to the flood. This ensures the redundant systems and equipment located in _ the other division are available for decay heat removal. > 2.13.4 RESOLUTION The System 80+ flood protection design features are consistent with acceptance criteria outlined above in section 2.13.2. These features resolve the issue of flooding and spills during shutdown operations on Systen 80+ by providing separation of redundant equipment required for decay heat removal. This separation provides the availability of DHR when a flood has occurred within the Nuclear Annex or Reactor Building Sub-Sphere. 2.13-2
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SUMMARY
AND CONCLUSIONS An assesoment of the risk associated with internal events during shutdown modes of operation was perfcfmed. The scope of this assessment included internal events during Mode 4 through 6 operation. Event trees were developed and quantified for loss of decay heat removal (DHR) and loss of coolant inventory for Modes 4 through 6. Loss of offsite power and criticality events were not quantified af ter an initial review showed these events to be a small contributor to risk. In quantifying the core damage frequency (CDF), emphasis was placed on the human error probabilities (HEPs) because earlier studies have shown that- these dominate the shutdown risk. The system failure rates were either taken from the System 80+ PRA or estimated using simplified fault trees. Table 3.1-1 summarizes the operator error rates,-mechanical failure rates and event tree branch failure rates that are developed in Section 3.4. The results from the Human Reliability Analysis reflect engineering judgement and assumptions associated with current operations practice. As further design and operations detail become available, the analysis will be updated as part of the PRA maintenance program to reflect a more accurate assessment of the impact of human error at shutdown. It is believed, however, that the analysis presented here reflects the correct level of dominance in certain events with respect to risk and that further analysis will provide more accuracy, with respect to the models, but should not affect the results of the dominant sequences. Table 3.1-2 summarizes the contributors to core damage frequency by modes and -initiating _ event. Mode 6 LOCA represents 69% of the risk. The leading contributor is a LOCA outside containment witn failure to isolate while the refueling cavity is full. Loss of DHR in Mode 6 while the refueling cavity is full is the second leading contributor to risk (17%). In both sequences, the IRWST is empty and feed and bleed is not available. If one of the two SCS trains is not restored, then core damage will result. Table 3.1-3 compares the CDF for shutdown events with those for power operation. The CDF 'for internal events during shutdown modes is 15% larger than for internal events at power and smaller in size than at power external events. The EPRI goal is to have CDF less that 1.0E-5/ry. The CDF from shutiown is significantly less than this goal and the sum of the CDF is also less than the EPRI goal. , This PRA contributed to technical specification and instrumentation changes. One example of such a change -was a technical specification change requiring two SIS trains be available in Modes 5 and 6 when the IRWST is available. 3-1
-. ~. . _ _ _,
I L77a.wp(9212)bh 1
' Table 3.1-1
SUMMARY
OF.JRANCH POINT' FAILURE RATES Operator: Mechanical Total Description Error Failure REP' INCL. INCL 1.0E-02 Repair during 81 hour coolant boil-off CS2 INCL INCL 3.1E-03 Containment Spray system (11of 2) DP 3.0E -8.0E-04 3.8E-03 Depressurize with SDS MUI 3.1E-03 1.0E-03 4.1E-03 Make Up Inventory OI S.0E-02 4.0E-02 9.0E-02 Operat:' Isolates leak OR 0.84 u0.16 1.6E-01 Operator Restarts SCS train i
*1.0E-03 ,
OS1 2.8E-03 2.0E-02. 2.3E Operator Starts standby SCS train (1 of 1) j . OS2 2.8E-03 2.0E-03 4.8E-03 Operator Starts 1 of 2 SCS trains SGHR '3.4E-02 .7.0E-04 3.5E-02 Operator starts SG Heat Removal' SGCOM 3.4E-05 INCL 3.4E-05 Conmission error in continued use of SG SIP 2 2.3E-3 7.0E-04 3.0E-03 Mangally start 1 of 2 SIS SCSFB 3.1E-03 1.8E-03 4 . 9 E,-03 ManU+lly use SCS for Feed and Bleed SIFB2 2.7E-03 1.5E-03 4.2E-03 Mant. ally use SIS for Feed and Bleed (1 of 2) BOC l - 3. 0 E-03 0.1 1.0E-01 Boll-Off using CVCS _; OIC 8.2E-03 8.0E-04' 9.0E-03 Operator Isolates leak at Containment - 3-2
)
770.wp(9212)bh TABLE 3.1-2 EBEOUENCY OF_ .RE DAMAGE FROM BHUTDOWN EVENTS LOSS OF DHR LOCA M9 2 M M TOTAL 4 2.3E-9 2.3E-8 2.5E-8 5 4.1E-10 5.4E-9 5.8E-9 5 REDUCED 6.5E-8 2.0E-8 8.5E-8 INVENTORY 6 1.3E-7 5.3E-7 6.6E-7 TOTAL 2.0E-7 5.7E-7 7.7E-7 TABLE 3.1-3 COMPARISON OF CORE DAMAGE FREOUENCIES POWER OPERATION, INTERNAL EVENTS
- 6.7E-7 POWER OPERATION, EXTERNAL EVENTS
- 1.2E-6 SHUTDOWN, INTERNAL EVENTS 7.7E-7
- FROM SYSTEM 80+ PRA7 3-3
77a.wp(9212)bh 3.2 PRA INTRODUCTION Until recently, emphasis has been on the safety of power plants during power operation. This was due to the f act that the plant is in this configuration most of the time and the core power, decay heat rate and fission product inventory are highest at this time. The System 80+ PRA 7 documents the risk associated with the operation of this ALWR at normal power. More recently, people have been investigating the risk of plants during shutdown and refueling. During these modes of operation, the plant has lower decay heat generation rates and fission product inventory. The plant configuration is not as well defined as in full power operation because of the maintenance and testing that is being per' armed. This section estimates the risk of operation of this plant in Modes 4 through 6. Mode 1 through 3 are covered in Ref. 7. The awareness of the risks of plant operation during refueling and maintenance outages developed slowly. Althau;b ao core damage has occurred during reactor outages, a f ew sw e; .E Lave occurred which were precursors to more severe accidents, one of the first events to increase the awareness of risks in outages was the loss of the refueling cavity seal at Connecticut Yankee on August 21, 1984. In this event, 200,000 gallons of water quickly spilled into the containment, draining the refueling cavity. This event would have been more difficult to handle if the refueling had actually started. The seal failure had not been considered in the safety analysis. Another event which increased the awareness of the risks during outages was the Vcgtle 1 event on March 20, 1990. The event started with a truck backing into equipment in a switch yard, causing a loss of power to the first auxiliary transformer. The second auxiliary transformer was out for maintenance. One of the diesels was also out for maintenance. The second kept tripping due to an instrument failure. This combination of. maintenance activities and f ailures led to a station blackout and loss of Decay Heat Removal (DHR). Under normal conditions, with the vessel filled or with the refueling cavity flooded, the operator would have many hours to restore DHR. In this incident, the plant was in mid-loop operation and the primary inventory was greatly reduced. When DHR was restored in 41 minutes, the primary coolant temperature had risen 46 deg F to 136 deg F. This incident demonstrates the unusual plant configurations that can exist during an outage and the risks associated with maintenance activities and mid-loop operation. During 1991, there were a rash of incidents during shutdown. After four events occurred within six days in March, the NRC issued Information Notice No. 91-22 describing these events. One plant 3-4
77a .'wp (9212 ) bh had two incidents during the same outage. There were at least seven events where loss of DHR occurred in 1991. All these events increased the awareness of the risks during outages. The NRC requested additional information from the ALWR participants pertaining to shutdown risk. The EPRI ALWR Requirements Document then was modified to include'a risk assessment for shutdown modes. This analysis satisfies that requirement. This cnalysis uses a simplified event tree approach to estimate the core damage frequency (level' 1 part of a PRA), with only a qualitative
-evaluation of the radiological releases.
The first step in the analysis was the identification - of _ _the initiating events of potential interest. This was done by first defining-the plant conditions (in terms of physical parameters such as = temperature, pressure, and inventory for the RCS) that will exist for different plant evolutions. For each of these operating conditions, general category of initiating events were then defined. These initiating events were small LOCAs, Loss of Decay Ht:at Removal (DHR), Loss of AC, and Boron Dilution. The frequencies for these events were determined by operational history. For each plant condition and initiating event, the plant and operator. response was estimated based on the advanced instrumentation, procedures, technical specifications, and safety systems employed in the System 80+ design. The plant states ~and operator response were modeled and quantified using simplified
- event trees. The unavailability of each system was estimated using
- simplified ' assessments and adaptation of models developed for power operations. Care was taken in estimating the reliability of human actions. Earlier studies found that operator actions are one of the dominant factors in this' analysis.
This analysis was performed with the insight obtained from the previous PRAs. In 1981, NSAC-84 looked at the shutdown risk at Zion. This study concluded that failures during reduced inventory
; operation accounted for 61% of the Core-Damage Frequency (CDF).
operator' actions were required in almost all sequences. operator
. failure to determine the proper actions to restore DHR accounted for 56% of the total CDF. Loss of DHR also accounted for 56% of the CDF. NUREG/CR-5015 tended to confirm the findings in NSAC-84.
The Seabrook Shutdown PRA concluded that 82% of the CDF was due to loss of DHR and 71% was from reduced inventory operation. The study also showed that early health risks were dominated by LOCAs with the containment open. The NRC's Shutdown PRA for Surry (as summarized in NUREG-1449) showed the importance of plant specifics such as the controls of the ADVs, and the response to Generic Letter 88-17. The insights gained from reviewing these PRAs helped in analyzing the System 80+ plant. 3-5
77a.wp(9212)bh This 'inalysis is an extension of the CESSAR-DC System 80+ PRA. The System 80+ PRA used the standard " sell event tree /large fault tree" approach used by most of the industry. A detailed discussion of the method employed is given in Section 2.0 of Reference 7. This section describes how that procedure was modified. The modifications were suggested by EPRI'. The scope of the Shutdown PRA is to evaluate the core damage frequency (level 1 PRA) using a simplified assessment. Only a qualitative evaluation of tha radiological consequences are presented. Event trees were developed for the accident sequences for each plant state in Modes 4 through 6 given in Table 3.2-1. The plant states were developed based on what equipment was available to mitigate the event. Mode 3 was not considered in this study because the plant configuration is very similar to Mode 1 and 2 and therefore, the effects are enveloped in the Mode 1 analysis. The event trees are based on the transient studies, for shutdown events described in Reference 4. Assumptions about operating or maintenance actions are based on current plant practices or improvements suggested in recent shutdown studies (e.g. NUREG-1449) and are listed in the event tree element descriptions. This will facilitate verification when the plant is complete and operating. The event trees are supported by fault trees and Human Reliability Analysis (HRA) models. This analysis is for a simplified assessment for the operating conditions other than power operation conditions to ensure that there are no unrecognized or unacceptable sources of risk associated with these conditions. This assessment is limited in scope because of the limited level of operational detail that is available at this time and because many requirements aimed at addressing problems at existing operating plants are currently being developed. The first step in the analysis is the identification of the initiating events of potential interest and the estimation of the frequencies for these events. A search of the earlier shutdown risk assessmente, LER data bane, and NPRDS was performed. The initiating event frequencies were proportioned to the plant states based on the fraction of the outage time spent in each state. The plant states represent plant modes, or part of a plant mode, which have certain physical parameters and equipment availability (both safety and operational). For each plant state and initiating event, a functional assessment of the plant response is made. A team of engineers, including specialists in thermal-hydraulics, operator error analysis, plant design, and outage management, were assembled. The team developed the plant response to each transient, including the sequence of events, available instrumentation and alarms, human response, and system response. The times to core damage were also estimated. 3-6 4
--_______-_______.__.__________.___________-_.________.______-____________________._________.____._________________s
77a.wp(9312)bh Emphases was on the human performance,. which has been shown in past studies - to be very. important in natisfactorily assessing the shutdown risk. The event troes wert developed with participation of the full team. The. branch points of. the event trees were evaluated using simplified fault trees for the responding systems. Human error rates were evaluated using Human Reliability Analysis models. These models are similar to event trees and model the operators actions sequentiality, including recovery. - The models also include operator interaction with - the available instrumentation. This detailed analyris of operator actions is used because earlier studies showed that operator actions dominated the analyses. Best estimate assumptions were made about the operator's environment. Also, little use has been made of such time based models as the human cognitive reliability model. This is because more data would be needed to evaluate the performance impact of the redesigned control room on the time lines associated with certain tasks. Consequently, consideration for the impact of time available to perform a task, and other factors utilized by the cognitive reliability model, have been dealt with by judicious application of certain performance shaping factors defined in the System 80+ PRA. These are: Availability of necessary indication Accuracy of indication Training Workload Annunciated Stress Level of Experience
-Quality / availability of procedures Ergonomic design of display / control Each of the abov2 factors can contribute a factor by which the Human Error Probability (HEP) can be modified, given the judgement of the analyst. This modification provides - a value that more accurately reflects the probability of human error.
The estimated unavailability of plant systems is based on an adaptation of models developed for power operation. Earlier studies have shown that the core damage frequencies were dominated by operator error. Therefore, less detail is _ provided in the system analysis. The assumptions used for each system are presented in the discussion of each event tree branch point. i l l l l 3-7 : I
'770.wp(9212)bh i
i TABLE 3.2-1 ELANT STATED AND INITIATING EVENTS Plant States MODE 4, normal inventory MODE 5, normal inventory MODE 5, Reduced inventory MODE 6, Refueling cavity empty MODE 6, Refueling cavity full Initi_a_ tina Events LOCA, inside containment LOCA, outside containment Loss of DHR Loss of Offsite Power Boron Dilution 3-8
770.wp(9212)bh 3.3 114ITI ATIE9_EYEllT_IREQUI11CIEA A scarch of relevant literature and databases was conducted in order to assons the frequency of occurrence for the specified events actually experienced at presaurized water reactors during plant shutdown pornods. The initiating events identified and examined consisted oft Boron and Reactivity Events; Loss of Electrical Power;
~
Loss of Reactor Coolant; Loss of Docay lleat Removal The IllPO Licensco Event Reports database constitutos the most rollable source of documentation for licensen svents, and was searched for the above listed events. Tau literature search included direct examinations of the following: A. NUREG-1449, " Shutdown and Low-?ower Operation at Commercial Nuclear Power Plants in the United States (Dratt)", pubashed 2/92 B. INPO 91-007, " Selected Significant Operating Experience Report
. Recommendatisns 1980-1990", published 9/91 C. NSAC-52, " Residual llont Removal Experience Review and Safety Analysis, Pressurized Water Reactors", published 1/83 D. FI L-1344, "?WR Low Power and Shutdown Accident Frequencies Programs, Phase 1A -
Coarso Screening Analysis, 11/13/91, Prepared for NRC" E. NRC Information Notico "<-55, incuad 8/31/90 F. NRC Information Notico 91-22, issued 3/19/91 The documento listed above were based to a largo extent en examinatioMs of LERs, but also prc,vided documentation for a numboc of events not documented in the
- LER database. Additionally, the documents above aleo drew 1 xmation from other literature, including:
G.- NSAC-144, " Loss of Of fsitt . r At U.S. Nuclear Power Plants, All Years Through 1988", pu bshed 1989 3-9
77c.wp(9212)bh 11 . 11UREG/ CR-4 550, Volume 1, Rev 1, SAllD86-2084, Jan 1990 J. Sequenco coding and Search System, Office of Analysis and Evaluation of nperational Data, U.S. liuclear Regulatory Commission. In total, the literaturo and database provido a reasonably completo sourco of data for calculating the ovent floquencios, llovertholoss in ordor to further en9ure that such frequenclos have boon calculo 'd conservatively, total reactor years have been calculated baned on the earliest dato recorded for each event, as summarized in Tablo 3.3-1. The frequency for each initiating event was distributed over all , the plant states based on the fraction of time that the plant is in each configuration. A 23 day refueling schedule was developed for the System 80+. This schedule was divided into each plant stato listed in Table 3.2-1. The number of hours in 11odos 4 and 5 was then doubled to account for forced outages. The first column of Table 3.3-2 gives the fraction of time (given as 1) that the plant in in each of thona states. The frequencios for loss of DilR and LOCA are then distributed over the various plant states based on the percent of time in each plant stato. This distribution assumos that the human errors and mechanical failures that cause the initial event are random in naturo. 3-10
77a.vp(9212)bh ThDLE'3.3-1 QREXILYIQ_.EMQUIllCIIQ_.QF_ DER EVENTS IN PWRs Number Earliest of Reactor Years Frequency (per Ry3nt Ryant RRt_g Events .(193 note *1 I.gA9 tor-yr) , Doron/Roactivity 03/14/82 12 747.3 .02 Loss of 06/07/73 52 1412.3 .04 Electrical Power Loss of Reactor 11/14/79 51 924.7 .06
- Coolant Loss of-Shutdown 09/02/76 232 1165.3 .20 -
Cooling
- note total roactor-years (PWR) between "Earliesst Event Date" and 01/01/92 IARkE _3. 3-1 INITIATU(G EVENT FREQUEliCY FOR PLANT _SIATIE FREQUENCIES, /RY PLANT % llOURS LOSS OF .4CA ETATE IN STATE DHR MODE 4 21 0.042 0.013 MODE 5 27 0.054 0.01t f MODE 5 18 0.026 0. 0M I i REDUCED IrfVENTORY NODE 6 16 0.032 0.009
- j. REFUELING CAVITY NOT FLOODED MODE 6 18 0.036 0.011 i REFUELING CAVITY FLOODED SUM 100 0.20 0.06 1
i 3-11 , 1 l
_ _ -- _ _ _ . _ _ _.=. _ -_-___._.._-.__._~.__m___ 77a.wp(9213)bh 3.4 ARCIDKHT_SIQUEN. CIA For nach of the accident acquencos censidered in i.his study, an i ovent tree was developed and ovaluated. Tne event tree starts with tho initiat.ing event frequency developed i' in Section 3.3 and describes the soquence of events as dtscreto events which are completed either successfully or not. Each specific sequenco leads to either coro damage or a safe plant condition. The following subsections discuss cach event tree and their branch points. Similar branch points are used in many of tho event trees. Such ! ovent troos are described in some detail the first time they are used and referred to in later uses. ' The event tree branches are composed of an operator error and a system mechanical failure. Sinco earlier shutdown risk studies have identified operator error as a significant contributor to risk, emphanos has boon placed on operator error. The iluman Rollability Analysis (llRA)- trees are also given and discussed in this section at the branch point where they are first used. System fault trees for hardware f ailuros are not presented in this report, but are procented in the System 80+ PRA. 7 [ 3.4.1 LOSS OF DHR, MODE 4 Figure 3.4-1 g ive.s c.he ovent tree for loss of DHR during Mode 4 operation. The following paragraphs discuss cach branch poir t of i the event tree. The operator has apprcximately 2 hours to restoro Dl!R" before core damage, assuming that the loss of DHR occurs at the beginning of sne outage (a conservative assumption). In this modo, two SIS trains are available and the IRWST is full. It is also assumed that the SGs are still available. LDHRM4 The initiating event frequency for loss of SCS while in mode 4 is i 0.042/ry. The derivation for this frequency ic given in Section 3.3. OR (Onorator RestoLCD1 In 16% of the Di!R losses, equipment failures were identified as the 3 cause of loss of DHR. It is assumed-no operator action will restore these mechanical failures in the short term in the 1 operating SCS train. In the remaining cases, the SCS train la lost because of some operator error and can be restored by the operator taking some 3-12.
_ _ _ . . _.__.____._____.-.._.___.__._m..___._..___._._._ 770.wp(9212)bh manual action. The operator has the ability to detect loss of the SCS with flow alarms, CCW alarms, temperature roadout, pump suction ; and dischargo roadout, and pump current roadout. The operator will have procedures and training in the operation of the SCS system and i will be using it at regular intervals. The performance of the SCS oporating train may bo lost because of , some operator error that causes the pumps to tri: but does no ! sirjnificant damage to the system. In this caso the operator may be ' able to restart the operating SCS train by restarting the pump. In order to achieve this, the operator must identify that there has boon a loss of shutdown cooling (task A), restart the pump (task D), adjust the SDC flow control valvo(task c) and adjust the SDC temperature controller as necessary (task D, see Figuro 3.4-2). The operator can identify the loss of SDC by SDC alarms, SDC flow ' alarms, Component cooling water alarms, pump suction and dischargo , indications, pump component indications and Decay heat removal ! critical function alarms. Once the identification has boon I achieved, the process of restart will procond according to the available lower mode ful.ctional recovery procedures. Quantification Task llEP PSF (specific) Sourco l (NUREG-
-- CR-1278)
A 0.0001 /5 ( control 20-23 room orgonomice) D 0.001 + 0:0005/5 /5 ( control 13-3 room orgonomics) i C 0.001 + 0.0005 /5 ( control 13-3
-room ergonomics) >
D 0.001 + 0.0005 /5 ( control 13-13 ' Ioom I orgonomics) HEP .= 0.0001/5 + 0.0011/5 + 0.0015/5 + 0.0015/5 '
= 1.0 x 10~8 The total failuro rate for this branch point is the sum of the mechanical failure rate (0.16) and recovery from operator error.
It is dominated by the mechanical failures and is: OR = 0.16 3-13 i l
770.wp(9212)bh 051 (Operatoi Starts 1 of I trains) -- If the operator can't start the first SCS train, he will go to ctart the second SCS train. The technical specifications requires that the train be available. It must be manually started. This event tree branch includos both operator failuro to start the train and the probability of mechanical failure of the train to start and run. The CSS pump can be aligned to substituto for the SCS pump. This action is included in the OSI failure rato. The mechanical failuro for a single trein is 2.0E-2. dince the technical specifications require that a rodundant train of the Shutdown cooling system.be available, the operator may be able to start this manually if re-starting the original train was unsucconsful. Assuming that the train is available and operable,_the operator would need to identify the nood-to make use of the redundant train (task A), start the SDC pump (task B), ensure that the SDC flow control valvo is controlling the SDC flow (task C), ensure that the SDC temperature controller is performing correctly (task D-see Figure 3.4-3). QuantificatioD Task HEP PSF (specific) Sourco (NUREG-CR-1278) A 0.0001 +0.01 /5 ( control 20-23 room 20-7 orgonomics) B 0.001 + 0.0005/5 /5 ( control 13-3 room orgonomics) C 0.001 + 0.0005 /5 (: control 13-3 room orgonomics) D O.001 + 0.0005 /5 ( control 13-13 room crongmics) HEP = 0.0101/5 + 0.0011/5 + 0.0015/5 + 0.0015/5
= 2.8 x 10'3 The total failure rato is the sum of the mechanical and operator f ailure ratos and is 2.3E-2. It is dominated by mechanical failure.
3-14 - - . -. . - . - -. - . - - - - . - . , . . . - - , - - , , _,- - .L _ -, . . - , -
1 778.wp(9212)bh SGilR (Steam Generator lleat_Ermoval) The operator normally usos tho SGs for heat removal in Hodos 1 through 3 and will try to use them attor failure to restore the SCS trains. By Technical Specification, he must have a SG availablo if 2nd SCS train is out of servico, 11 0 only has electric feodwater pumps available. In many casos the SGs are filled witn water when
. laid up, and in this caso, the operator has additional timo to start the foodwater pumps. The secondary sido can be vented through the ADVs.
The operator normally uses the steam gonorators during modos 1 through 3 and will be required to attempt to use them af ter failure to restore the SDC trains. Technical specifications require that at least one steam generator be available if the 2" train of SDC is out of servico. In order to utilize this method of decay heat removal, the operator neods to identify the nood to uso a steam generator (task A), start the electric omergency foodwater pumps (task B), when the steam generator prassure rison to 950 psia to vont the ADV (task C monitor, task D actuate ADVs), and monitor the decay heat removal (task E, 000 Figure 3.4-4). The indications available to the operator to identify the need for this method include SDC alarms, component indications for the availability of the SDC system and critical function decay heat removal alarms. The actions associated with this method can all be achieved from the control room. These actions are adequately covered in the lower modo functional recovery procedures. Egntification Task EEP PSF Source (specific) (NUREG-CR-1278) , A_ 0.0001 +0.01 x2 (stress)- 20-23 20-7 B (0.001 + 0.0005/5)x2 x2 (stress) 13-3 C 0.001 x2 (stress) 13-3 D 0.001 + 0.001 + x2 (stress) 13-13 0.0005 E O.001 x2 (stress) 20-10
!!EP = 0.0101*2 + 0.0022*2 + 0.001*2 + 0.0025*2 +0.001*2
= 3. 4 x 10-2 '
3-15
-.;^...-~,,. - _ - - - -- - - . . . -
770,wp(9312)bh-The mechanical failure to start and run 1 of 2 electric feedwater pumps was estimated as 7.0E-4. For this brench point, the total failure rate is 3.5E-2 and is dominated by operator error. SIFB2 (Safety Iniection system Feed add Bleed, 1 of 2 traingl The operator has the option of going to a food and blood operation. Two SI trains are available in tho division with the Diesel also available. One of the two pumps will supply sufficient flow. In addition, one of the two charging pumps could also be used but was not modeled hero. The primary system must be vented with the Dafety Depression System (SDS). The failure rato for SIFB2 is the sum of-the failure of the operator to initiato feed and bleed and mechanical failures. While in modo 4 or__ lower, if the operator looses shutdown cooling he will' attempt to satisfy decay heat removal. Assuming that all also fails, the operator would attempt to initiato food and bloed cooling. To initiato food and blood cooling the operator must first identify that food and blood cooling is required (tank A) . Then ensure that the RCPs are tripped (task B) and ensure that the line-up from the IRWST through the SI purps to the DVI line is availablo (task C). Since RCS pressure is below the shutoff head of- the Safety Injection pumps already, safety _ injection initiation would need to be achioved manually (task D) . Then they must open the Safety Depressurization Valves (task E) and the Safety Depressurization Block Valvos (task F) to allow circulation. In Mode 5, reduced inventory operation, opening the SDS Valvos is not necessary because the pressurizer manway is remcVed. Finally the operator would monitor flow-to ensure the correct operation of the systen (task G - see Figure 3.4-5). To be successful, Food and Blood Cooling must be initiated at or before the time at which the primary safety valves lift. All actions can be performed from the control room. Procedures would be available in the lower mode functionni recovery procedure for loss of secondary side heat removal. Inclusion of this event represents failure to initiate feed and Bleed cooling in mode 4. Ouantification Task HEP PSF (specific) Source (NUREG-CR-1278) A 0.0001+0.01 /5 ( control 20-23 room 20-7 ergonomics 3-16 L
-w--- -- , --e * + e er --,-,,-,w.,-.,w..gmy--,.-- we,------,3- -ww. ..-e- ,-r *. .-- , . - . . - . . - - -w e.- - -
. - --_ . . - . -- ._. ~ . _ _ - . - - _ - . . - - . - . . _ - - - - _ ~ -. ,_
77c.wp(9212)bh 1 B 0.05*0.05 /5 ( control 20-22 room ergonomics C 0.05*0.05*0.0 *0 /5 ( control 20-22
.0$ room ergonomics D 0.001*0.001 /5 ( control 20-12 room ergonomics E 0.001*0.001 /5 ( control 20-12 room ergonomics F 0.001*0.001 /5 ( control 20-12 room orgonomics G 0.001 /5 ( control 13-3 room orgonomics HEP =
0.0101/5 + 0.0025/5 +0. 05'/ 5 + (0.001*0.001;/5 + (0.001*0.001)/5 + (0.001*0.001)/5 + 0.001/5
= 2.7 x 10*3 The mechanical failures consist of f ailure of both SI trains (7.0E-
- 4) and the SDS (8.0E-4). Total failure for feed and blood is:
SIFB2 = 4.2E-3 and du dominated by operator error. BOC (Boil-Off usina CVCS) Using either 1 of 2 charging pumps or the boric acid makeup pump as a backup, water can be added directly to the reactor vessel from tho. Boric Acid Storage Tank (BAST). There are approximately 80,000 gal. of borated water available and by matching the boil-off rate, the water in the BAST can be used to cool the core for approximately 12 hrs. There is an additional 20,000 gal of water available in two of the four SITS that could be added to the core through gravity feed and give the operators an additional 3 hours. It is assumed that if BOC is successful, DHR is restored in the approximate 12 hours during-boil-off. BOC failure rate consists of two failure rates, failure to feed with the CVCS and failure to restore some DHR capability during the approximate 12 hours. During a Loss of Coolant Accident in the lower modes, making up inventory loss may be achieved via a different method than at 3-17 3
. .,. . . . . - . . ~ , , , , - - . , - - , - . . - .-,------.v , ,e , -
1 1 77a.wp(9212)bh l l l power. It is possibic to makeup inventory from the Boric Acid Storage tank through the CVCS and the charging pumps. This requires that a line up be established from the BAST to the i suction line for a charging pump, then the charging pump can ) deli /cr the inventory to the RCS. ' The actions on the part of the operator would be to identify the need for makeup (task A), open the gravity feed lino from the BAST to the charging pumps i.e. valvo Cll-532 (task B) and Cil-536 (task C), ensure that one of the two charging pumps is operating (task D), and monitor the makeup (task E - see Figure 3.4-6). All actions can be performed from the control room. Procedures would be available in the lower mode functional recovery procedure for loss of secondary side heat renoval. Inclur, ion of this event represents a failure to makeup inventory coolant from the CVCS. 3-18
- l
77c.wp(9212)bh l l Q1gntification Task HEP PSP (specific) Source (NUREG-CR-1278) A 0.0001+0.01 /5 ( control 20-23 room 20-7 , ergonomics B 0.001 + 0.0005 /5 ( control 13-3 room ergonomics C 0.001 + 0.0005 /5 ( control 13-3 room ergonomics lD 0.001 + 0.0005/5 /5 ( control 13-3 room Orgonomics E 0.001 /5 ( control 13-3 room ergonomics HEP = 0.0101/5 + 0.0015/5 + 0.0015/5 + 0.0011/5 + 0.001/5
= 3.0 x 10
In -this scoping study, we did not explicitly analyze the probability of restoring DHR in approximately 12 hours, but ausumed a failure rate of-0.1. The operator has the options of restoring any of 4 SI trains, 2 SCS trains or 2 CSS trains. He could also find additional water for his BAST and continue boil-off. As an example of such repairs, af ter the severe Brown's Ferry cable fire (March,1975), operators successfully restored DHR using temporary jumpers. BOC failure rate is dominated by the probability of repair and BOC = 0.1. Figure 3.4-1 gives the event tree with the branch failure rates discussed above. In this event tree, one sequence leads to core i damage with a frequency cf 2.3E-9. 3.4.2 LOSS OF DHR, MODES 5,6, REFUELING CAVITY EMPTY Figure 3.4-7 gives the event tree for loss of DHR during Mode 5 operation when the coolant level is not reduced and during Mode 6 operation while the IRWST is still ful.1. During this time, two SIS trains and one CSS train is available. It is assumed that the SGs 3-19
77c.wp(9212)bh are not available and the system is being cooled with one SCS, and one in standby. This event tree is similar to the previous tree. The operator has approximately 2.2 hours to restore DHR assuming that the ucquence occurs when the plant first enters Mode 5 (highest decay heat) and not in later Mode 5 or 6 operations. The following paragraphs discuss nach branch point of the event tree. DHRCE (Decay. Heat Removal loss with refuelina Cavity Erotv) The initiating event frequency for loss of SCS while in Mode 5 is 0.054/ry. In Mode 6, with the refueling cavity empty, the frequency is 0.032. The derivation for this frequencies is given in Section 3.3. The total frequency is the sum, DHRCE = 0.086 . QR (Operator Restorest The failure rate for the operator to restore the SCS train is (see Section 3.4.1): OR = 1.6E-1 . OS1 (Operator Starts) The failure rate for the operator to start the second SCS train is (see Section 3.4.1) : OS1 = 2.3E-2 .
,STFB2 (Safetv_Iniection train Feed and Bleed)
The failure rate for the operator to start the feed and bleed operation is 4.2E-3 and discussed in Section 3.4.1. SCSFB (Shutdown Coolina System for Feed and Bleed) If the SIS-fails, the reactor inventory and DHR can be maintained by using the SCS in a feed and bleed mode. The SCS can be aligned to the IRWST and used to inject water. Heat removal is through the SCS coolers. If the SCS pumps fail, the CSS pumps can be aligned to substitute for the SCS pumps. While in mode 4 or lower and the operator attempts to satisfy decay heat removal, assuming that all else fails, the operator would attempt to initiate feed and bleed cooling. Also given that the safety injection pumps may be unavailable it is possible to use the SCS pumps for this purpose. 3-20
77a .wp (921::) bh To initiate food and biced cooling using the SCS, the operator must first identify that feed and bloed cooling is required (task A). Then ensure that the RCPs are tripped (task B) and the cross-connect valve SI-110 is opened to ensure the line-up from the IRWST is available (task C). This must then be checked by another operator, who could fail in either of two ways; tt.sk D - the 2nd operator makes a checking error and then changes the valve position to ref)oct his incorrect belief of the correct position, and task D'- a simple checking error with a report to the control room that the valve change was successful. The operators would need to start j the Shutdown Cooling Pumps manually (task E). Then they must open the Safety Depressurization Valves (task F) and the Safety l Depressurization-Block Valves (task g) to allow circulation. In l Mode 5, reduced inventory operation, the pressurAzer manway is ; removed and opening the SDS valves is not necessary. Finally the operator would monitor flow to ensure the correct operation of the system (task.H - see Figure 3.4-8). . To be successful Feed and t Blood Cooling must be initiated at or before the time at which the ' primary safety valves lift. All actions, except one, can be performed from the control room. Procedures would be available in the lower mode functional recovery procedure for loss of secondary side heat removal. Inclusion of this event represents failure to initiate feed and biced using the SCS and the IRWST. Quantification 1 Task HEP PSF (specific) Source (NUkEG-CR-1278) A 0.0001+0.01 /5 ( control 20-23 room ergonomics B (0.05)2 /5 ( control 20-22 room ergonomics C 0.001 /5 ( control 20-6 room ergonomics D 0.1*0.001
/5 ( control room ergonomics ,
D' O.1 3-21 ,-s 9.-m-+-- - .am,- w. y .-,.yy,,..,---r-te,.,,,y. , m my.en,mypmy-g u gy pra sg y.-g-,,,.9p. s g. ,y 7.,, ,yymess.,,ygw, ---&m,p 3,,--g9 a. --- c--9-a.,--
-p .- 4r*--.a-
. . = _ =_ - ._ - - - - - . - _ - - . - _ - - - . 770.wp(9212)bh l E 0.001+0.001 /5 ( control room ergonomics F 0.001*0.001 /5 ( control room-ergonomics G 0.001*0.001 /5 ( cor' trol room
~
Orgonomics , l F O.001 /5 ( control 20-12 ! room ergonomics HEP = a + b + cd' + ce + d + 0 + cf + f + eg + g + ch + h
=
0.0101/5 + 0.0025/5 + 0.0001/5 + 0.0001/5 + 0.000002/5 + 0.002/r + 0.000000001/5 + 0.000001/5 + 0.000000001/5 + 0.000001/5 + 0.000001/5 + 0.001/5
= 3.1 x 10*3 The mechanical failures for SCSFB are the mechanical failures for SCS injection (MUI) and the mechanical failure.of the SDS (DP).
SDS is needed to return the coolant to the IRWST tank and not to depressurize the system. Tho total failure rate is: SCSFB = 4.9E-3. BOC fBoil-Off usina the Cvec system) As discussed in Section 3.4.1, the failure rate to successfully cool the coro and restore DHR.is 0.1. Figure 3.4-7 gives the event tree with the branch failure rates disdussed above. In this event tree, one sequence leads to core damage with a frequency of 6.5E-10. 3.4.3 LOSB OF DHR, MODE 5, REDUCED INVENTORY Reduced inventory is defined as having the primary coolant level.3 f t. below the flange or lower. Experience developed with operating
' plants and results of earlier Shutdown PRAs has shown that becauce of the short time for coolant boiling given loss of DHR (approximately 15 minutes assuming mid loop operation), the plant has technical specifications and operational guidance for operation in this mode. Figure 3.4-9 gives the event tree for loss of DHR 3-22
_ _ __ ,=. , . . _ ,
770.wp(9212)bh during Mode 5 operation when the coolant level is reduced. This event. tree is also used to evaluate LOCAs during reduced inventory because the operator would treat both events the same way. For this event, the operator has approxinately 1.5 hours to restore DliRas before core damage assuming that the loss of D11R occurs at the first time in the outage that the plant is operated at reduced inventcry (a conservative assumption). The following paragraphs discuss each branch point of the event tree. DHR5R (loss of Decay Heat Renoval in mode 5. Reduced invent;pryl The initiating event frequency for loss of SCS while in mode 5 is 0.047/ry. This frequency includes both loss of D11R (0.036) and the frequency for LOCAs in this mode (0.011) . LOCAs were included here because the operator would perceive them as a loss of DiiR. The derivation for this frequency is given in Section 3.3. OR (Onerator Reststrial, The failure rate for the operator to restore the SCS train, discussed in Section 3.4.1, is: OR = 1.6E-1 . MUI (Make Un Inventorvi For reduced inventory events, it is assumed that the operator must diagnose and increaso the coolant level before starting the second SCS train. We assumed the operator would use the SCS system for
-this task and align it to the IRWST. lie could also use the CVCS sistem, but this was not explicitly modeled. inclusion of the CVCS make up would not significantly reduce the f ailure rate because the failure rate is dominated oy operator error.
While in the lower modes, making up inventory loss may be achieved via a-different method than at power. It is possible to makeup inventory from IRWST through the SC system. This requires that a line up be established from the IRWST to thc suction line for a shutdown cooling pump, then the SCl pump can deliver the inven*ory to the RCS. The actions on the part of the operator would be to identify the need for makeup - (task A) , Ensure the RCPs are tripped (task B), locally open the cross-connect valve (task C), have this checked by another operator (task- D - checking failure and reposition of valve to wrong position based on incorrect checking and task D' - checking failure and report to control room that change was successful ), and start the pump (task E - see Figure 3.4-10). 3-23
, , . ,. .~
77c.wp(9212)bh Most actions can be performed from the control room. Procedures would be available in the lower mode functional recovery procedure for loss of secondary side heat removal. Inclusion of this event represents a failure to makeup inventory coolant from the IRWST through the SCS. QuantificAt1QD Task HEP PST(specific) iource (NUREG- _, CR-1278) A 0.01+0.0001 /5 ( control 20-23 room 20-2 ergonomics) B. 0.05 x 0.05 /5 ( control 20-12 room ergonomics) C 0.001 /5 ( local 20-12 ergonomics) D (0.1 x 0.001) 13-13 D' O.1 13-13 E 0.001+0,.001 /5 ( control 20-12 roota ergonomics) HEP = a + b + cd' + ce + d + e
=
0.0101/5 + 0.0025/5 + 0.0001/5 4 0.0000002/5 + 0.0001 + 0.002/5
= 3.1 x 10'3 The mechanical failure of the SCS to make up inventory in 1.0E-3.
The total failure rate in: MUI = 4.1E-3, and is predominately operator error. OS1 (Ocerator StartM, ' The failure rate for the operator to start the second SCS train is ' 2.3E-2 and discussed in Section 3.4.1. SJf32 (Safety Inigq1.jpn trairl reed and Bleed. 1 of 2 traj,Its). In reduced inventory mode, suggested new technical specifications require that two SIS train be available. As noted in Section 3-24 l l
770.wp(9212)bh l 3.4.1, in this mode, the pressurizer manway has ocen removed and opening the SDn valves is not necessary. The fillure rate ist S'FB2 = 4.2E-3 . BOC (Boil-off usina the Cycs avstemi As discussed in Section 3.4.1, the failure rate to successfully ! cool the core using boil-off and restore DHR is 0.1. Figure 3.4-9 givos the event tree with the branch failure rates discussed above. In this event treo, two sequences lead to core damage and the total CDF is 8.5E-8. 3.4.4 LOSS OF DHR, MODE 6, REFUELING CANAL FLOODED At this time in the refueling, the refueling cavity and canal is flooded. .The IRWST is empty and in general, not available for any make up, safety injection, or feed and bleed operation. The steam generators also are not available for decay heat removal (assumed isolated with nozzle dams) . Because of the lower decay heat and large volume of water in the refueling cavity, the operator has 81 hours to restore DHR before core damage". Figure 3.4-11 gives the event tree for this event. DHRCF (Decay Heat _Removql_with the Refueliro Cavity filledl TheLinitiating event frequency for loss of SCS while in Mode 6, cavity flooded,'is 0.036/ry. The derivation for this frequency is given in Section 3.3. OR (Oneratgr Restores) The failure rate for the operator to restore the SCS train is OR = 1.6E-1 and is discussed in Section 3.4.1. OS1 (Onorator Starts) The failure rate for the operator to start the second SCS train is-2.3E-2 and is discussed in Section 3.4.1. 3-25
77a.wp(9212)bh REP (Renairi llaving failed to restore or start either SCS train in a timely manner, the operators now have approximately 81 hours to repair or restore D}iR, or replenish the water in the refueling cavity. In ! this scoping study, we did not explicitly analyze the probability , of restoring DilR in approximately 81 hours but assumed a failure ' rate of 0.01. The operator has the options, of rostoring any of the 2 SCS trains or 2 CSS trains. llo could also refill the '.kWST and . uso any 1 of 4 SIS trains that he could make operational in a feed and blood modo. Finally, ho could also find additional water for his refueling cavity and continue boil-off. DOC fBoil-Off tsina the CVCS) As discussed in Section 3.4.1, the failure rate to successfully cool the caro and restoro DliR is 0.1. Figure 3._4-11 gives the event tree with the branch failure rates discussed above. In this event tree, one sequence leads to coro ! damago and the CDF is 1.3E-7. This is the recond largest sequence in the study. It roprosents a sequence where there are only 2 l trains for a DilR available (two SCS trains) and no capability to uro any other paths (SIFB2, SCSFB). i 6 6 h p 3-26
. - _ , _ _ _ . . _ _ _ . - _ , , _ _ . . . . 1... ,_ ,_._, ,_ _.,- _ _,---.,__,_____ . __.___ __ _.
_,,1
77a.vp(9212)bh 3.4.5 LOCA, MODE 4, PRESSURE ABOVE 500 PSIG ; Modo 4 is hot shutdown and includes the period of timo that the primary system is ooing cooled and depressurized (or heated and pressurized). In mode 4, the coolant temperature varies from 350'F to 210'F. Above 500 psig, tha SCS system .s not available but the SIS, and SITS are. This plant configuration is very Jimilar to higher modos and are modoled here because it represents a mode whero the SGs are being used and depressurization is required to get onto SCS or a feed and bleed mode. It is asLumed that the temperaturo la abovo 317'F, and the pressure is abovo 500 poig. Two CSS trains, the IRHST, two RCS loops, and two SIS trains are availhble. The SCS trains are available only after depressurization. 'Iho low pressure and temperature part of Mode 4 is represented by the Modo 5 normal inventory event tree described in the next section. The Mode 4 tree is given in Figure 3.4-12. - M4LOCA , The frequency of a LOCA.in Mode 4 is 0.013/ry (Section 3.3). t 01 (Onerator Isolates leak)
' Most of the LOCAs in shutdown modes are caused by operator errors in aligning valves and other operations and are quickly corrected. '
Others are caused by mechanical failures that can not be easily isolated. 1 Given that most of errcrs are caused by cperators inadvertentl*/ opening valves or creating leak paths, it is assumed that most of these occurrences are caught either by the operator who causod it or by operator check work already performed. With this assumption, it is reasonable to assume that the type of checking being performed would be associated with some kind of aaerting f actors associated with the type of situation i.e sound of rushing water when none should be present, visual indication of the leak etc. i For the : purposes of this analysis, the quant.ification of this factor will maka use of the value- for opecial one-of-a-kind checking with alerting ' factors. This is because of the unique nature of most of the activities associated with the lower modes and the concept that this will be the type of error the operator will be attuned to. 5 f 3-27
77c.wp(9212)bh
,Qupntification Task HEP PST(specific) Source (NUREG-CR-1278)
A 0.05 20-22 HEP = 0.05 For leaks caused by mechanical failures it was assumed that 4% could not be isolated and are sufficiently large to be true LOCAs. The total failure to isolate is the sum of the mechanical and operator failuroc and is : OI = 0.09. ji]P2 (Safety Iniection Pumns) In Mode 4, and at theca pressures, two SI trains are available and are automatically actuated. Operator action to initiate SI is not required and the failure rate for both trains to start and run is 7.0E-4. The IRWST is available as a water supply. CS2 ( Cont a inment_jpf ty_l, To remove decay heat trom the containment, one of two containment spray trains must start and run. Failure of CS is 3.10E-3. SGCOM (Steam Generato1 A rator Commission Errori The operator was using the SGs before this event started and, af ter isolating the leak, would continue to use the SGs for heat removal. This branch element represents failure of the operator to continue using this DHR path. Typically the method employed to cooldown the RCS using the steam generators in the lower modes is to fill the steam generator using the emergency feedwater system then switch off and let this boil away.. Since the operator action involves checking the steam generator level-and initiating the emergency feed this operator error has two components. Task A - failure to initiate steam generator heat removal (SGHR) and task B - operator fails to check for the initiation ( see Figure 3.4-13). 3-28 i i __m_. _ - . - - - - - - - - - - - - - - - - - - - -
Q .Wp(9212)bh quantification Task IIEP PSP Source . (specific) (NUREG- { CR-1278) l A 3.4 x 10'8 see SGliR see SGIIR B 0.01 20-22 l IIEP = 3.4 x 10-8 x 0.01 SGCOM = IIEP= 3.4 x 10** DP (DeProcsurizationi If the SIS fails or heat can not be removed through the SGs, than the operator must depressurize the primary system in preparation to use the SCS or CSS to add inventory and remove decay heat. The operator must manually actuate the Safety Depressurization System
. (SDS).
In order to use the shutdown cooling. pumps or containment spray pumps to add inventory the operator must deprensurize the system. This is achieved by utilizing the safety depressurization system. In order to achieve depressurization the operator must identify the need for depressurization using the SDS (task A), open the safety depressurization valves (task B), open the safety depressurization block valves (task C), and monitor the flow through the system to ensure that it is operating properly (task D - see Figure 3.4-14) . These actions can be performed from the~ control room and would be part of the lower modo functional recovery procedure. L i 1 4 3-29 p-* yp gr g y- .w,,., yr- , .g .,y%-w- y .,e59,yogp - p w-, -q-a .%--w-,---5 m-pr y.-i+.,-
l 77c .wp (9212) bh Ouantification Task IIEP PSF (specific) Source (NUREG-CR-1278) A 0.01+0.0001 /5 ( control 20-23 room 20-2 orgonor cs) B 0.001 x 0.001 x2 ( stress) 20-12 C 0.001x 0.001 x2 ( stress) 20-12 D 0.001 13-13 HEP = 0.0101/5 + (0.000001)*2 + (0 000001)*2 +0.001
= 3.0 x 10
Hochanical failure of 1 cf 2 MOVs in the SDS to open is 8.0E-4. The total failure to depressurize is: DP = 3.0E-3 and is dominated by operator error. OS2 (Onorator Starts one of 2 SCS trains) Having depressurized the system, the operator can now start one of two SCS systems. Since the plant was pressurized, both trains are assumed available. The operator error in starting 1 of 2 SCS i trains is assumed to be the same as that for starting 1 SCS trcin and is 2.8E-3 (Section 3.4.1). Mechanical failure of both SCS trains is 2.0E-3 and the total failure rate is: OS2 = 4.8E-3. SIFB2 (Safety Iniection to Feed and Blead, L_gf 2 trains) Assuming that the normal DHR paths can not be used and the operator has depressurized the system, the operatur can now go to a feed and bleed operation. This operation is not available if the SIS had failed earlier in the event sequence. The failure rate for this branch is: SIFB2 = 4.2E-3 and is discussed in Section 3.4.1. Figure 3.4-12 gives the event tree with the branch failure rates discussed above. In this event tree, eight sequences lead to core damage and the total CDF is 2.3E-8. Half of the sequences and almost all of the risk is caused by f ailure to depressurize so that the SCS or feed and bleed trains can be employed. 3-30
I 770.wp(9212)bh ! 3.4.6 LOCA, MODES 5, 6 (IRWST FULL) In Mode 5, and some part of Mode 6, the IRWST is available for safety injection and two SIS trains are available for both injection and food and bloed operation. Figure 3.5-15 gives the event tree for this sequenco. The starting point for this event is that the plant is depressurized and one SCS train is in operation. The SGs are i assumed to be isolated. Two SIS trains are available for manual actuation in Mode 5 and Mode 6 with the IRWST full. One CS pump is also available for injection or feed and blood. L561R (LOCA Mode 5 and 6, IRWST Available) The frequency for this event is c.025.- It consists of Mode 5 LOCAs (0.016) and Mode 6 LOCAs with the IRWST full (0.009). LOCAs occurring during Mode 5 mig loop operation would be diagnosed as loss of DilR by the operator and are addressed in Section 3.4.3. OI (Onerator Isolates leak)_ The failure of the operator to isolate the leak is discussed in Section 3.4.5 and ist OI = 9.0E-2 . SIP 2 fSafety Inigction Pump. 1 of 2) In Mode 5 and 6 with the IRWST full, two SI trains and the IEWST are available. SIS would be manually actuated. While in lower modes, the system pressure would be lower than the Safety Injection setpoint, and since procedures require the override of the ESFAS actuation for normal progression below, it would be necessary for the operators to. manually initiate Safety ' Injection. The actions requircJ 3r this task are that the operator must identify the need fc; safety injection actuation (task A), ensure the correct line up for the required Safety Injection pump (task B), then start the Safety Injection pump (task C, see rigure 3.4-16). All of these actions can be performed from the control room, and are governed by lower mode emergency procedures. Inclusion of this event in the fault tree represents the failure of the operator to manually initiate Safety Injection. 3-31
770.wp(9212)bh t Ouantification Tack llEP PSF (specific) Source (NUREG- l CR-1278) _ A 0.0001+0.01 /5 ( control 20-23 room ergonomic s B (0.01): 20-22 c 0.001 + 0.0005/5 /5 ( control 20-12 room orgonomic . s t HEP = 0.0101/5 + 0.0002 + 0.0011/5
= 2.3 x 10~3 The mechanical ,*.0E-4.
failure rate for one of two trains of SIS to start and run is The operator error rate is 2.3E-3 and the total failure rate is: SIP 2 = 3.0E-3 and is dominated by operator error. OS1 (Onerator Starts one SCS traini llaving isolated the leak, the operator can now start one of the SCS systems. It is assumed that the leak caused the other train to be unavailable. The failure rate is: 051 = 2.3E-2 and discussed in Section 3.4.1. SIFB2 (Safety Injection to Feed and Bleed) If the leak isolation was successful but starting the SCS train was not successful, then the operator can continue to cool the core by operating one of two SI pumps after having opened the SDS to ret *2rn the fluid to the IRWST. This failure rate SIFB2 = 4.2E-3 and is discussed in Section 3.4.1. SCSFB (Shutdown Coolina System for Feed and Bleed) If the SIS fails, the reactor inventory and DHR can be maintained by-using the SCS in a feed and bleed mode.- The use of the SC pump is included in the SCS fault tree. This branch point failure rate 3-32
.--_ . _- ., =- . -_. . . .- _ - _ .
77c.wp(9212)bh I was evaluated in Section 3.4.2 and is: SCSFB = 4.9E-3. BOC__fBoll-Off usina the CVCS) i As discussed in Section 3. 4.1, the failure rate to successfully cool the core and restore DHR is 0.1. Figure 3.4-15 gives the event tree with the branch failure rates discussed above. In this event tree, three sequencen lead to core damage and with total CDF being 8.5E-9. The dominant sequence is failure to isolate the leak, followed by failure of the SIS and SCS to feed and bleed. 3 4.7 LOCA OUTSIDE CONTAINHENT, MODE 6 The LOCA outsido containment (Figure 3.4-17) is an interfacing type LOCA where the coolant inventory is lost and in Mode 6 with the refueling cavity filled, the IRWST is not available. This means that the SIS is not available and no feed and bleed path is available (SIFB2 or SCSFB). L60C (LOCA Mode 6 Outside Containment) The frequency for this event is 0.0055 assuming that half of the Mode 6 LOCAs occurred inside containment and the other half occur outside. 01 (Ocerator isolates leak) The failure of the operator to isolete the leak is discussed in Section 3.4.5 and is: OI = 9.0E-2. OIC (Ocerator Isolates at containment interface) If the leak was not initially isolated locally (OI), the operator will try to isolate the leak at the containment penetrations using 1 of 2 MOVs that are in series. During a Loss of Coolant Accident in the lower modes and once the makeup issue has been addressed, the operators' primary concern will be isolating the leak that is causing the flow outside containment. This does not necessarily mean identifying the leak only dealing with the flow path to outside containment. In order L 3-33 !=
i l 77a.wp(9212)bh to mitigate this, the operator would need to actuato a close signal to olther of the two containment isolation valves. The actions performed to achieve isolation at the containment would be, to identify the need to isolate onco core decay heat removal has been satisfied (task A), actuate the controls to closo valve 1 (task B), and closo Valvo 2 (task C), then verify the success of this approach by monitoring the RCS level indication (task D, see Figure 3.4-18). All actions can be performed from the control room. Procedures would be available in the lower mode functional recovery procedure for loss of secondary side heat removal. Quantification Task llEP PSF (specific) Source (liUREG-CR-127 8 ) - A 0.0001+0.01 /5 ( control 20-23 room 20-7 ergonomics B 0.001 + 0.0005 x2 ( stress),, 13-3 C 0.001 + 0.0005 x2 ( stress) 13-3 D 0.001 /5 ( control 13-3 room ergonomics llEP = 0.0101/5 + 0.0015x2 + 0.0015x2 + 0.001/5
= 8.2 x-10**
The mechanical failure of 1 of 2 MOVs was estimated to be 8.0E-4 and the total failuro rate is: OIC = 9.0E-3 and is dominated by operator error. psi (Operator Starts one SCS train) liaving isolated the leak, the operator can now start one of the SCS systems. It is assumed that the leak caused the other train to be unavailable.- The failure rate is: OS1 = 2.3E-2 and is discussed in Section 3.4.1. REP (Repair)
- At this point in the sequence, the operator has successfully isolated the leak but now has a loss of DliR event, with neither of the SCS trains working and no -feed and bleed capability. As 3-34 i
. _ = =_ = = = ]
l' 77c.wp(9212)bh discussed in Section 2.4.4, f ailure to restore either SCS train vac estimated ast REP = 0.01. BQq__(1}pihQf f usino the CVCS systoq). As discussed in Section 3.4.1, the failure rate to successfully cool the core and restore DilR is 0.1. F_gure 3.4-17 git /es the event tree w#th the branch failure rates discussed above. In this event tree, two sequences lead to core damage and the total CDF of 4.6E-7. The dominant sequence in this event (and this study) is the LOCA outside containment that is not isolated. The second sequence represents a loss of DilR and requires operator action to repair one of the two SCS trains. 3.4.8 LOCA INSIDE CONTAINHENT, MODE 6 (REFUELING CAVITY FULL) In Mode 6, with the refueling cavity filled, if the LOCA occurs inside the containment, the coolant will drain back into the IRWST and be available for injection or feed and bleed using any of the SCS pumpe or the CS pump in the make-up (MUI) mode. The coolant also becomes available for a forl and bleed mode using the SCS and CS trains. Figure 3.4-19 gives the event tree for this sequence. The rtarting point for this event is that the plant depressurized and one SCS train in operation. The SGs are assumed to be isolated. No SIS trains are available. One CS pump cnd both SCS pumps are available for injection after the coolant drains back into the IRWST. L6IC Loca Mode 6 Inside Containnellt The frequency for this event is 0.0055. This assumes that half of the Mode 6 LOCAs occurs inside containment. QI (onerator Isalates leak) The failure of the operator to isolate the leak is discussed in Section 3.4.5 and is: OI = 9.0E-2 . HUI (Make Un Invejltqtyl _ In Mode 6 with the IRWST empty, SIS is not available but injection is possible with SCS and CS pumps (Secticn 3.4.3): MUI = 4. 3 E-3 . 3-35
770.wp(9212)bh R5.1 (ODeratqr__G1 Rts one SCS LEAiD1 The failuru rato is: Os1 = 2.3E-2 and is discussed in Section I 3.4.1. S.CSFB (ShutdQWn Coolina System for Feed and B1qgA). If-the SIS fails, the reactor inventory and DliR can be maintained by using the SCS or the CS pump in a feed and bleed modo. This branch point failuro rate was ovaluated in Section 3.4.2 and is: SCSFB = 4.9E-3. JipC (Boll-Of f unjna the Cven synter11 As discussed in Section 3.4.1, the failure rato to successfully cool the core and restore DiiR is 0.1. Figure ?.4-19 givos*the event tree with the branch failure rates discussed above. In this event troe, throo acquences lead to core damage and with total CDF being 6.3E-8. The dominant sequence is failure to restore the second SCS train (including the CS pump) including feed and bleed. 3.4.9 LOSS OF OFFSITE POWER In the previous Shutdown Risk Studies summarized in NUREG 1449*3 loss of offsito power (LOOP) was not a major contributor to risk. In the Seabrook Shutdown PRA (May, 1988), LOOP contributed 64 of the core damage risk. This sequence was only qualitatively studied in this analysis because of the low frequency for IDOP for the System 80+ plant. Section 2.4.3.2 describes AC power reliability. The System 80+ has the capability to draw from four different AC sources including a combustion turbine. With two independant incoming sources and switchyards, the EPRI ALWR Utility Requirements Document gives the probability of losing both switchyard as 7.7E-3. - When this is coupled with failure of 1 Diesel and f ailure of the combustion turbine, the risk becomes very small. Additionally, in Section 3.3 it was estimated that the refueling cavity is flooded for 29% of the time the plant is in Modes 3 through 6. During this period, a long recovery time exists I for restoration of AC power. l l -3.4.10 CRITICALITY EVENTS In the previous Shutdown Risk Studies summarized in NUREG 1449 8 reactivity events were not a major contributor to risk. Boron dilution could be either a rapid injection of deborated water or a slow dilution. Section 2.6 discusses rapid boron dilution events and concludes that the reactor would stay subcritical, even for the}}