ML20095L504
| ML20095L504 | |
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|---|---|
| Site: | 05200002 |
| Issue date: | 04/30/1992 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
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{{#Wiki_filter:_ _ - _ _ _ _ _ _ . _ __ ._ . - _ _ SYSTEM 80+
.t SHUTDOWN RISK EVALUATION ,
REPORT DCTR 10 , f APRIL 30, 1992 (DRAFT) ABB-COMBUSTION ENGINEERING NUCLEV W ER SYSTEMS WIND' ,x, CONNECTICUT l
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LEGAL NOTICE THIS REP 0RT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY ABB' COMBUSTION ENGINEERING. NEITH14 ABB COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALc 4 A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILI'IY, WIT!! RESPECT TO THE ACCURACY, COMPLETENESS, OR.USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METilOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS; OR B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OR FOR DAMACES RESULTING F140M THE USE OF, ANY INFORMATION, APPARATUS, ME HOD OR PROCESS DISCLOSED IN THIS REPORT.
770.wp (9075) bh ' ABSTRACT In engineering the System 80+ Standard plant Design, ABB recognized the significance of addressing safety during shutdown operations. System 80+ is engineered with features that enhance shutdown-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 '
, by thorough procedural guidance and Technical Specifications that andress important shutdown evolutions. This report presents these features and evaluates them in the context of the specific shutdown issues identified by the NRC. The report fulfills the ABB :ommitments 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.
77c.wp (9075) bh DEFINITIONS The-following-definitions of terms are employed throughout this report. [This section will be provided in the June 15, 1992 updated submittal of this report.)
. .. ~ . _ . . '77c.wp(9075)bh TABLE OF CONTENTS SECTION SUBJECT PAQB NO.
__ ABSTRACT i DEFINITIONS 11 TABLE OF CONTENTS iii LIST OF TABLES LIST OF FIGURES
1.0 INTRODUCTION
1-1 1.1 PURPOSE 1-1 1.2 SCOPE 1-1
1.3 BACKGROUND
1-1 1.4 SYSTEM 80+ FEATURES 1-2 2.0 SHUTDOWN RISK ISSUES 1 2.1 PROCEDURES 2-1 2.1.1 ISSUE 2-1 2.1.2 ACCEPTANCE CRITERIA 2-1 2,1 3- . DISCUSSION 2-1 2.1.4- RESOLUTION 2-2 2.2 TECHNICAL SPECIFICATION IMPROVEMENTS 2.2-1 2.2.1 ISSUE 2.2-1 2.2.2 ACCEPTANCE CRITERIA 2.2-1 2.2.3 DISCUSSION 2.2-1 2.2.4 RESOLUTION 2.2-2 2 . 3 -- BEDUCED INVENTOR'! OPCRATIOli kND GL 88-17 FIXES 2.3 -111-
I 77a.wp(9075)bh l TABLE OF CONTENTS (Cont'd) SECTION' SUBJECT PNE ND. 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 Nozzle Dam Intecrity ?.3-4 2.3.3.4 ALTERNATE INVENTORY ADDITIONS
-AND DHR METHODS 2.3-5 2.3.3.5 OPERATIONE 2.3-6 2.3.4- RESOLUTION 2.3-7 2.4 LOSS OF DECAY HEAT REMOVAL CAPABILTTY 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 Analyseg 2.4-2 2.4.3.2 System 80+ AC Power Availability 2.4-2 2.4.3.2.1 Introduction 2.4-2 2.4.3.2.2' Discussion 2.4-2 -
2.4.3.2.3 Conclusion 2.4-4 2.4.3.3 System 80+ Diesel Generator Availability 2.4-4 2.4.3.3.1 Introduction 2.4-4 2.4.3.3.2 Discussion 2.4-5 2.4.~3.3.3 Conclusion 2.4-6 l-l l u
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77c.itp ( 907 5 ) bh - TABLE OF CONTENTS (Cont'd) SECTION SUBJECT pffigh 2.5 PRIMARY / SECONDARY CONTAINMENT CAPABILITY AND SOURCE TERM 2.5-1 4 2.5.1 ISSUE 2.5-1 2.5.2 ACCEPTANCE CRITERIA 2.5-1 2.5.3 DISCUSSION 2.5-2 2.5.4 RESOLUTION 2.5-2 2.6 RAPID BORON DILUTION 2.6-1 2.6.1 ISSUE 2.6-1 2.6.2 ACCEPTANCE CRITERIA 2.6-1 2.6.3 DISCUSSION 2.6-2 2.6.4 RESOLUTION 2.6-2 2.7 flRE PROTECTION 2.7-1 , 2.7.1 IS$UE' 2.7-1 2.7.2 -ACCEPTANCE CRITERIA 2.7-1 2.7.3- DISCUSSION 2.7-1 2.7.3.1 Mitication of Fire Consecuences 2.7-1 2.7.3.2 Detection and Sucoression 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 2.8.3 DISCUSSION 2,8-3
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"'70.wp ( 9 07 5) bh TABLE OF CONTENTS (Cont'd) SECTION SUBJECT PICE 10. 2.8.3.1 Instrumentation Desian Basis 2.8-3 2.8.3.2 Instrumentation Description 2.8-4 2.8.3.2.1 Level 2.8-4 2.8.3.2.2 Temperature 2.8-5 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 Messages 2.8-10 2.8.3.2.5.3 Discrete Indicators 2.8-11 2.8.3.2.5.4 CRT Display Pages 2.8-11 2.8.3.2.5.5 Component and Process Control Indicators 2.8-12 2.8.3.2.5.6 NUPLEX 80+ Alarm Characteristics 2.8-12 2.8.4 RESOLUTION 2.8-13
-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 2.10 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 2.10-1 2.10.4 RESOLUTION 2.10-2 2.11 FUEL HANDLING AND HEAVY LOADS 2.11-1 2.11.1 ISSUE 2.11-1
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77c.wp (9075) bh TABLE OF CONTENTS (Cont'd) SECTION SUBJECT PICE NO. 2.11.2 ACCEPTANCE CRITERIA 2.11-1 2.11.3 DISCUSSION- 2.11-1 2.11.4 RESOLUTION 2.11-4
-2.12 POTENTIAL FOR DRAINING THE REACTOR VESSEL 2.12-1 2.12.1 ISSUE- 2.12-1 2.12.2 . ACCEPTANCE CRITERIA 2.12 2.12.2.1 Prevention Criteria 2.12-1 '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 2.12.4 RESOLUTION 2.12-2 2.13. FLOODING 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 2.13.4 RESOLUTION 2.13-2 3.0 PROBABILISTIC RISK ASSESSMENTS 3-1
3.1 INTRODUCTION
. 3-1 4.0- APPLICABILITY OF CHAPTER 15 ANALYSES 4-1 4.0.1 FORMAT AND CONTENT 4-1 4.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 4-1 4.1.O INTRODUCTION 4-1
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1770.wp(9075)bh-TABLE OF CONTENIH (Cont'd) SECTIOr{ SUBJECT PAGE 10. 4.1.1- DECREASE IN FEEDWATER TEMPERATURE 4-1 4.1.2 INCREASE IN FEEDWATER FLOW 4-2 4.1.3 INCREASE MAIN STEAM FLOW 4-2 4.1.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE 4-2 4.1.5 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT 4-2 4.2 DECREASE IN HEAT REMOVAL BY SECONDARY SYSTEM 4-2 4.
2.0 INTRODUCTION
4-2 4.2.1 LOSS OF EXTERNAL LOAD 4-3 4.2.2 TURBINE TRIP 4-3 4.2.3 LOSS OF CONDENSER VACUUM (LOCV) 4-3 4.2.4 MAIN STEAM ISOLATION VALVE CLOSURE 4-3 4.2.5 STEAM PRESSURE REGULATOR FAILURE 4-3 4.2.6 LOSS OF NON-EMERGENCY AC POWER TO THE STATION 4-4 4.2.7 LOSS OF NORMAL FEEDWATER FLOW 4-4 4.2.8 FEEDWATER SYSTEM PIPE BREAKS 4-4 i 4.3 DECREASE IN REACTOR COOLANT FLOW RATE 4-4 4.
3.0 INTRODUCTION
4-4 l 4.3.1 TOTAL LOSS OF REACTOR COOLANT FLOW- 4-5 i
'4.3.2 FLOW CONTROLLER MALFUNCTION l CAUSING FLOW COASTDOWN 4-5
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77c.wp(9075)bh TABLE OF CONTENTS (Cont'd) SECTION SUBJECT PNQs 10. 4.3.3 SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER 4 4.3.4 REACTOR COOLANT PUMP' SHAFT BREAK WITH-LOSS OF OFFSITE POWER 4-6 4.4 REACTOR COOLANT PUMP SHAFT BREAK WITH LOSS OF OFFSITE POWER 4-6 4.
4.0 INTRODUCTION
4-6 4.4.1 UNCONTROLLED CONTROL ELEMENT , ASSEMBLY WITHDRAWAL FROM SUBCRITICAL OR LOW POWER CONDITIONS 4-6 4.4.2 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL AT POWER 4-6 4.4.3D SINGLE CONTROL ELEMENT ASSEMBLY DROP 4-7 4.4.4 STARTUP OF AN INACTIVE REACTOR COOLANT PUMP 4-7 4.4.5 FLOW CONTROLLER MALFUNCTION CAUSING AN INCREASE IN BWR CORE FLOW RATE 4-7 4.4,6 INADVERTENT DEBORATION 4-7 4.4.7 INADVERTENT LOADING OF A FUEL ASSEMBLY INTO THE IMPROPER POSITION 4-7 4.4.8 CONTROL ELEMENT ASSEMBLY (CEA) EJECTION 4-7 4.5 INCREASE IN RCS INVENTORY 4-7 4.
5.0 INTRODUCTION
4-7 4.5.1 INADVERTENT OPERATION-OF THE ECCS 4-8 4.5.2 CVCS MALFUNCTION-PRESSURIZER LEVEL CONTROL SYSTEM MALFUNCTION WITH LOSS OF OFFSITE POWER 4-9
.4.6 DECREASE IN REACTOR COOLANT SYSTEM INVENTORi 4-10
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770,Wp(9075)bh TABLE OF CONTENTS (Cont'd) SECTION SUBJECT PN3E 10. 4.
6.0 INTRODUCTION
4-10 4.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY / RELIEF VALVE 4-10 4.6.2 DOUBLE-ENDED BREAK OF A LETDOWN LINE OUTSIDE CONTAINMENT 4-11 4.6.3 STEAM GENERATOR TUBE RUPTURE 4-11 4.6.4 RADIOLOGICAL CONSEQUENCE OF MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR) 4-11 4.6.5 LOSS-OF-COOLANT ACCIDENT 4-11 4.7 RADIOACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 4-11 4.
7.0 INTRODUCTION
4-11 4.7.1 RADIOACTIVE GAS WASTE SYSTEM FAILURE 4-11 4.7.2 RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE 4-11 4.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID CONTAINING TANK FAILURES 4-12 4.7.4 FUEL HANDLING ACCIDENT 4-12 4.7.5 SPENT FUEL CASK DROP ACCIDENTS 4-12 5.0 APPLICABILITY OF CHAPTER 6 LOCA ANALYSES 5-1 5.1 ISSUE 5-1 5.2 ACCEPTANCE CRITERIA 5-1 5.3 DISCUSSION 5-1 5.
3.1 DESCRIPTION
OF LOCA SCENARIO 5-5 5.3.2 -SELECTION OF REVERSE PLANT PARAMETERS AND CONDITIONS FOR MODE 4 ANALYSIS 5-5 _x_
770.wp(9075)bh TABLE OF CONTENTS (Cont'd) SECTION SUBJECT PA22 NO. 5.3.3 ANALYSIS COMPUTER CODES 5-6 5.3.4 LOCA ANALYSIS FOR MODE 4 5-6 5.3.5.1 Results of LOCA Cases With No ECCS Delivery for More than 10 Minutes 5-7 5.3.5.2 Influence of Restorina 1 HPSI Pumo Not At The Broken DVI Line 5-7 5.4 RESOLUTION 6.0 AEELICABILITY OF CHAPTER 6 CONTAINMENT ANALYSES 6-1
6.1 INTRODUCTION
6-1 6.2 LOSS OF COOLANT ACCIDENTS (LOCAs) 6-1 6.3 MAIN STEAM LINE BREAKS (MSLBS) 6-2 6.4 INADVERTENT OPERATION OF CONTAINMENT HHEAT REMOVAL SYSTEMS 6-3
6.5 CONCLUSION
6-3 7.0 SYSTEM 80+ 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 COOLING WATER SYSTEM 7-2 i
7.2.5 STATION SERVICE WATER SYSTEM 7-3 7.2.6 ELECTRICAL DISTRIBUTION SYSTEM 7-3 i
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77c.wp(9075)bh TABLE OF CONTENTS _(Cont'd) SECTION SUBJECT PAQE No. 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
7-6 8.0 C.ONCLUSIONS 8-1
- 9. 0 - REFERENCES 9-1 APPENDICES APPENDIX A RESPONSES TO REQUESTS FOR A-1 ADDITIONAL INFORMATION
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'770.wp(9075)bb LIST OF FIGURES-FIGURE' SUBJECT PNm3 No. 2.4-1 SYSTEM 80+ ELECTRICAL DISTRIBUTION 2.4-2 DIESEL LOAD SEQUENCER - SIMPLIFIED LOGIC DIAGRAM 2.7-1 NUCLEAR ISLAND FIRE BARRIER LOCATIONS, PLANT AT ELEVATION 50+0 2.7-2 NUCLEAR ISLAND FIRE _ BARRIER LOCATIONS, PLANT AT ELEVATION 70+0 2.8-2 REACTOR COOLANT SYSTEM ELEVATIONS RELATED TO SHUTDOWN COOLING OPERATIONS 2.9-1 LOCATIONS OF SAFETY INJECTION SYSTEM SUCTION IN IRWST 2.9-2 LOCATION OF WING WALL DEBRIS SCREEN ASSEMBLIES 2.9-3 LOCATIONS OF TRASH RACK, DEBRIS SCREEN AND SPILLWAY FOR IRWST AND HVT 2.11-1 CONTAINMENT BJILDING LOAD HANDLING-PATHS 2.11-2. FUEL HANDLING BUILDING. LOAD HANDLING PATHS 2.13 NUCLEAR ISLAND DETAILED ARRANGEMENT PLAN AT EL. 50+0 2.13-2 NUCLEAR ISLAND DETAILED ARRANGEMENT PLAN AT EI . 70+0 6-1 CONTAINMENT PRESGURE VS. TIME FOR LOCA FROM ZERO POWER 6-2 CONTAINMENT ATMOSPHERE TEMPERATURE VS. TIME FOR LOCA FROM ZERO POWER 6-3 CONTAINMENT PRESSURE VS. TIME FOR MSLB FROM MODE 5 6-4 CONTAINMENT ATMOSPHERE VS. TIME FOR MSLB FROM MODE 5
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77c.wp(9075)bh LIST OF TABLES TABLE TITLE ENTE NO. 1-1 SHUTDOWN EVENT CATEGORIES AND SYSTEM 80+ FEATURES FOR PREVENTION, DETECTION AND MITIGATION 1-2 SHUTDOWN EVENTS AND SYSTEM 80+ PREVENTION, DETECTION AND MITIGATION FEATURES 2.8-1 REDUCED INVENTORY INSTRUMENTATION PACKAGE 6-1 ESFAS INSTRUMENTATION 6-2 CASES ANALYZED - 6-3 INITIAL CONDITIONS FOR LOCA INITIATED FROM ZERO POWER
'6-4 ACCIDENT CHRONOLOGY FOR LOCA INITIATED FROM ZERO POWER 6-5 INITIAL CONDITIONS FOR MSLB INITIATED FROM MODE 5 6-6 ACCIDENT CHRONOLOGY FOR MCLB INITIATED FROM MODE 5 l
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77mewp(9075)bh l l
1.0 INTRODUCTION
1.1 PURPOSE 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 commitment made to the NRC by ABB in Reference 1 to submit shutdown risk information in support of the System 80+ Design Certification. 1.2 SCOPE Sections 2.1 ttrough 2.13 present detailed discussions on the specific shutdown issues. Following tne detailed discussiona 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 if to LOCA and accident events that are initiated from shutdown 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 in Rockville, Maryland on December 18, 1991 and is outlined by the ABB slides enclosed with the NRC minutes of the meeting in Reference 2. The report also addresses the RAI's from the NRC staff 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 wnich encompass the response to each RAI.
1.3 BACKGROUND
In Generic Letter No. 88-17 (Reference 4) the NRC issued recommendations to all holders 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 implemented and/or are underway at operating plants. ABB reviewed NRC evaluations and recommendations and searched the events which NRC considered significant. Generic characteristics of.these events were compared with features of System 80+ that can prevent or mitigate the events. In this way ABB has demonstrated that the System 80+ design satisfies the intent of the ALWR progra.m 1-1
77c.wp(9075)bh to benefit from past PWR experience and that the specific NRC concerns have been addressed in the design. 1.4 SYSTEM 80+ FEATURES In this section, a comparison is made between the characteristics of past events and the System 80+ design features. The categories of shutdown events at operating plants are those used by the NRC in Chapter 2 of NUREG-1449 with little modification. These categories are mostly the same as the issues identified by Secy-91-283 and presented by ABB 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 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 pla. cad on each 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) Effect 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 few 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 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 events. 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 han ming 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 1-2
77a.wp (9071i) bh - prevention is impracticable and may be impossible. A comb'. ation of prevention and mitigation' is employed in the System 80+ asign.
- Table- 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 specific events that have occurred. For each initiator, the features of the System 80+' design that are available to prevent occurrence of an initiator and/or to mitigate the consequences of an initiator are listed.
Table 1-2'provides a list of specific past events initiated from shutdown modes. ' Events were selected to include all ten event categories.and all-types of event initiators, but not all similar significant events that have occurred. Several information sources were utilized to compile this list. They include events listed _in NUREG-1449 which were taken from the 1990 AEOD report (Reference 5) and which occurred mostly between January 1988 and July 1990 with some additional events. For events since July.1990, ABB searcheit the INPO database for LER's using a selection of keywords pertinen'; to the ten event categories.and to shutdown operation. Various other-INPO and NSAC documents were also reviewed for significanc event 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 review of this table serves as a_ design review of the System 80+ capabilitie's to avoid core damage and/or significant radiation release during shutdown modes. The design features are_ discussed in more detail in the following Sections of this report, 4 ' l: L 1-3 l
77b(9132)da/1 TABt.E 1-1 SHUTDOWN EVENT CATECORIES AND SYSTEM 80+ FEATURES FOR PREVENTION. DETECTION AND MITICAT10er EVENT CATEGORY EVENT INif f ATOR SYSTEM 80+ FEATURES FOR PREVENT!Ott DETECTION AND MITICAT!ON 1.)' Loss of Shutdown Cooling SCS flow toss by pu m suction vortex. A) Mid-loop levst maximized by locating SCS suction piping at the bottom of the hot teg. B) Hard piped venting for SCS ptrps relieves gas binding more - quickly and conveniently. C) One SCS suction line from each hot leg provides SCS redundarty with separation of ptsy suction sources. D) Containment spray pupps identical to SCS pumps provide redtsident capacity and sey take suction from IRv5T to refitt RCS and to mitigate gas binding. Inaccurate mid-tocp levet leading to E) With head on, reactor vesset levet s-2nitoring system levet suction vortex. Indications from vessel head to a levet below that regJired for SCS operation. Levet indication is accurate for intended use. f) Core exit thermocouples nonitor coolant temperature down to 100*F pricr to withdrawal of CETs daring fuel shuf fling. The RTDs and SCS tetyeratures are accurate dJring SCS operation. G) With head off, tevet indication near hot leg elevation is provided by high resolution instrtsnents. N) SCS performance sonitored on each of 2 SCS ptsys by punp motor current, flow rate, discharge pressure and suctio.s pressure. Possible SCS flow variance with decay heat to minimize potential for vorteming dring mid-loop. Loss of flow white head off, t4per t) Internals design limits coolant flow from cavity to core. internals in vesset and cavity Nigh availability of SCS system and/or backups assures forced f Loaded leads to core heattp. convection. Various low levet and loss of RHR J) Non-shared SCS system at tows SCS maintenance and testing events. daring Modes 1-4 prior to cold shutdoasn, increasing availability in Modes 5 aruf 6. K) Att SCS vet us are setor operated, preventing failures on toss of air if electro-pneunatic operators were used. L) Shutdown specific tech specs and procedural guidance reduces likelihood of persomet errors.
77b(9132)da/2 TABLE 1-1 (Continued) SHUTDOWN EVENT CATECORIES AND SYSTEM 80+ FEATURES FOR PREVENil0N. DETECTION AND MITICATION EVENT CATEGORY EVENT INITIATOR STSTEM 80+ FE ATURES FOR PREVENTION. DETECTION AND MITIGATION 1.) Loss of Shutdown Cooling M) Inadvertent errors are redwed and early operator evaluation-(Continued) of faltures is improved by 1.) IPSO overview display with critical fmetion and system statua specific to shutdown modes, 2.) CRT displays with system lineups and conponent status and 3.) storms that are dependent on plant mode and equipment status. N) Prevention of inerproprista automatic actions from persomet errors by shutdam specific controt logic (e.g., remove autoctosure interlocks from SCS suction valves.)
- 0) CCW availability is increased by 2 redJndant Divisions, nch with two punpa and heat exchangers.
P) Service water availability is increased by 2 Redandant Divisions, each with two ptsps. Q) Each SCS Division has four potential sources of AC power for increased availability. 2.) Loss of Electric Power Equipment failure and/or inadvertent A) Alternate AC gas turbine provides third on-site power personnet error leading to toss of source. power ard shutdown cooling . B) Two switchyard interf aces provide fienibit ity. C) Shutdown specific tech specs and procedural guidance reduce likelihood of personnel er va. D) Nof1nal power from safety transformer to saiety buses does not depend on any non-safety conponents. E) Each safety division has a dedicated diesel generator. F) No equipment is shared between diesets. G) No equipnent is shared with another unit. 3.) Loss of Reactor Coolant Frra shutdown mode, egaipneet failure A) Inadvertent errors are reduced and early operator evaluation and/or personnet error leads to loss of Silures is leproved by 1.) IPSO overview display with of coolant, usually through systems critical function and system status specific to shutdown connected to RCS. codes 2.) CAT displays with system lineups and conponent status and 3.) stares that are depen $ent on plant mode and egJi f rient status, i
77tg9132)da/3 TABLE 1-1 (Continued) SHUTDOWN EVENT CATEGORIES AND SYSTEM 80+ FEATURES FOR P9EVENTION. DETECT 104 AN9 MIT!GATION EVENT INITIATOR _ SYSTEM 80+ FEATURES FOR PREVENTION. DETECT!ON AND MITICATION EVENT CATEGORT 3.) Loss of Reactor Coolant (Continued) Inadverteni RPV pressuritation while B) Mid loop vent will not attow significant EV head connected systems are open causing pressurization. Thou, instruments are not af fected. coolant levet drop in vesset. C) In-core instrunent seat table evolutions are prohibited ty p-oce&ral guidance iAile vesset head is on and mid loop evolutter- tre in progress, preventing seat teaks. D) Coolant losa via 2CP during sea'. maintena2ce reduced by punp lavetter weight creating seat. E) Cavity draining limited by reinforced pool seat between Cavity draining exposes fuel being vesset flange and cavity floor. transferred. f) Containment layout prevents total draining if seal faits. A) Tech spec requires hatch and att penetrations closed during Contairnent Irtegrity Loss of shutdown cooling and/or loss mid-loop evolutions. All temporary lines pass through 4.) of reacter coolant results in core penetrations, not through the open hatch. Containment boiling requiring rapid cc3taireent conf)Juration and size attow acre outage activities uithin closure to prevent radiological containment, resultine in less time without containment release. integrity.
- 8) Ret 1Jndancy en SCS system, electric pouer supply and stopcrt systems together with increased instrumentation reduce likelihood of an initiating event progressing to botting.
C) Shutdcun specific tech specs and proceiral guidance reduce Persornet errors result in opening likelihood of personnel errors. pathways f rom contefrnent to stinos @ ere during shutdown evolutions. A) SCS system relief valves sized fo^ menisua safety injectio+- Ir advertent high pressure safety
- 5. Overpressurizatior; liquid flow.
injectior; actuation at low tenperatura pressurizes RCS and SCS system. B) RCS is vented through two vents charing Modes 5 aruf 6.
w/ F7b(9133)da/4 TA8tE 1-1 (Continued) SHUTOOWN EVENT CATEGORIES AND STSTEM 80+ FEATURES FOR PREVENTION. DETECTION AND MITIGATION EVENT INITIATOR SYSTEM 80+ FE AfitRES FOR PREVENf!ON. DETECTION AND MITIGATION EVENT CATEGORY C) Ring foeged reactor vesset belttine and vesset materlat 5.) overpressurization (Continued) provide additional margin to pressurized thermal shock. uncontrotted coolant flow from opened A) Inadvertent errors are reduced and early operator evaluation 6.) Flooding and Spftts of failures is improved by 1.) IPSO cverview display with systems, typically caMned with critical function and system status specific to shutdown other inadvertent and/or poorly modes 2.) CRT displays with system lineups and corponent planned evolutions, floods essential status and 3.) stanns that are dependent on plant made and equipment. equipment status. B) Shutdonni specific tech specs and procedural guidance redxe likelihood of personnet errors. C) Ptvit layout, including separation of redJndant divisions, limits c.amage that may occur to affected division. No comnunication between divisions, including piping, electricat, HVAC, floor drains, etc. Various C1/CS M soperations and A) Shutdown specific tech specs and procedural guidance redace 7.) Boron and Reactivity Events uncalibrated source range neutron t(keti%od of inproper operation. s,,ailton c use agproach to criticality. l CVCS-misoperat m a w.es boron B) Precipitation prevented by detign that limits boron dilution A poterial how concentration to below cold precipitation concentration in most borated coolant lines, eliminating need for most heat precipitatino, tracing. C) Boron dilutica alarm provides advanced warning. During shutdown evolutine'.S, use of ;) Plant layout and fire barriers separate red m dant divisions 8.) Fire Protection conbustible materists pf e ine.itico ano systems to limit potential fire damage. sources such as tyrary p er Iines increases potential for , ice derar,e to essential systems.
- 8) KorAnastible materials are timited in specific fire controt areas.
l Heavy Loads and fuel Handling Inadequate design and/or surveillance J) Shutdom specific guidance limits pathways for heavy lif ts. 9.) of 1ifting devices causes potent (a1 damage to fuel or essential equipment. j
.77b(9132)da/5'-
.i TABLE 1-1 (Continued)..
- SHUTDOWN EVENT CATECORIES AND SYSTEM 80+ FEATURES FOR PREVENTION. OETECTION AND MIT! Gait 0No EVENT CATEGORY ' EVENT INITIATOR- SYSTEM 80+ FEATURES FOR PREVENTION.- DETECTION Aae MITIGATION- 11 9.) Heavy Loads and Fuel Handling-' ' 8 ) '. Plant arrangement minimizes potential for % ing drops.
(Continued) *
.. c) . Proven design for fuet,' core arrangement and fuel handling machine minimizes potential fuel. drop.
~
10.) Mode Change Events operator and/or procedural errors . A) Shutdoun specific. tech specs and procedarat guldence redace . +
' atlow mode changes without satisfying - , likelihood of personnet errors.
entry re pirements. 8)' Inadvertent errors are reduced and early operator evaluation of faltures is improved by 1.) IPSO cverview display with
. critical function and system status snecific.to shutdown sudes 2.) CRT displays with system linetes and component status and 3.) alarms that are dependent on plant mode and equipment status.
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77 c.vp ( 9075) bh 2.0 SHUTDOWN RISK ISSUER 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-1449 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 AB3 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 each 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. 2.1 PROCEDHFIE 2.1.1 ISBUE The operational guidance provided by the plant designer to the owner /operater might not be sufficient to insure that procedures to avoid, detect, mitigate, and/or recover frc; abnormal events initiated from shutdown operations can be developed by the plant owner / operator. 2a1.2 ACCEPTANCE CRITERIA The operational guidance provided by the plant designer to the owner / operator shall be sufficient to properly utilize decign features that are available to detect, mitigate and assist recovery from abnormal events initiaced 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 information system which better informs the operations staff of plant status, 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. I 2-1 l
77a.wp(9075)bh 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. Discussion and explanation of the procedural guidance that is appropriate to the shutdown modes and that will be provided to the owner operator will be presented in the June 15, 1992 updated submittal of this report. 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 chutdown operations. The information that will be provided to the owner / operator as operational guidance during shutdown modes will be presented in the June 15, 1992 updated submittal of this report. l l 2-2
770.wp(9075)bh 2.2 TECHNICAL BPEQJFICATION IMPROVE)igl{Il 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 safsty 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 b. nded by the analyses. Technical specification should include the necessary limiting conditions for operation that are applicable to shutdown modes. 2.2.2 ACCEPTANCE CRITEP.M 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 DISCUS 810N The System 80+ design incorporates advdaced features which promote safer operation and greater margins to operating limits. The fonteres include redundancy and diversity of components and systems and an advanced information system which better intorms the operations staff of plant status, potential adverse system interactions, and available recovery paths if an abnormal event occurs. These features also contribute ^o 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 ef ficiently as possible. Another objective is to formally i impose the operational constraints required to insure the plant remains within analyzed bounds far operation through the initial set of technical specifications. Overly restrictive tecnnical specifications especially for shutdown moder 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 traditicnal design basis ever.ts . Discussion and explanation of proposed changes or additions to the technical specifications will be presented in the June 15, 1992 updated submittal of this report. 2.2-1
770.wp(9075)bh 2.2.4 RESOLUTION The issue of shutdown specific technical specifications is resolved for System 80+ by modifications and additions to the technical specifications based upon safety analyses performed for modes 2 through 6. These changes will be presented in the June 15, 1992 updated submittal of this report. W e 2.2-2
770.wp(9075)bh 2.3 REDUCED INVENTORY OERRATION AND GL 88-17 FIlgg 2.3.1 ISSUE The NRC has voiced increasing concern over the safety of operations during plant shutdowns. Plant events which have occurred in the industry have highlighted the need for a 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 operating 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. containment closure issues,
- 2. instrumentation which would greatly improve the operator's monitoring capability during reduced inventory operations,
- 3. alternate ways to add inventory to keep the core covered should SCS be lost,
- 4. administrative procedures that would avoid RCS perturbations during reduced inventory operations, and
- 5. nozzle dam installation procedures which would ensure a vent pathway is available so that RCS pressurization can be minimized if shutdown cooling is lost.
The NRC has specified that programmed enhancements should accomplie5 a comprehensive improvement in the plant's ability 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 feacures may not be available undo.r shutdown conditions. Therefore licensees should implement means to prevent accident initiation, to monitor a progression that may lead to core damage and to evaluate consequences and, where needed, to provide mitigation. 2.3.2 ACCEPTANCE CRITERIA The System 80+ design shall reflect a comprehensive consideration , of chutdown and lower power risk, by adequately addressing all GL l 88 recommendations and other issues- relevant to reduced inventory, especially in the areas of instrumentation, technical specifications, procedures, equipment availability and analyses. I 2.3-1 l e
770.Vp(9075)bh l I 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 maintenance and testing activities minimizes the time during which the RCS must be operated in a partially filled condition. To minimize operating time at mid-loop level, nozzle dams are installed on the steam generators and the RCS is reflooded tc continue maintenance and 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 interf ace with the hot leg. In the worst scenario, subsequent to vortexing in the SCS suction line, a large percentage of air is entrained into the SCS suction piping, shutdown cooling pump cavitation occurs, and the SCS pump is damaged or lost entirely. 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. The System 80+ design features will realize 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 /oporator will be presented in the June 15, 1992 updated submittal of this report. 2.3.3.1 Inptrumentation for Shutdown operations Diverse, accurate, and redundant instrumentation, including control room CRT displays, gives continuous system status, and provide the operations staff with precise information to respond to loss of shutdown cooling events, should they occur (See also Section 2.8 of this Report). Phenomena which can affect instrumentation operation are considered. Analyses form the basis for instrument design, calibration, and operation so as to assure correct instrument operability during reduced inventory states. Instrumentation availability during shutdown will be assured vie the plant Technical Specifications that willbe provided in Section 2.2 of the l final submittal of this report. The types of instrumentation used are cutlined below. i 2.3-2
77a . wp (9075) bh
- 1. Independent wife and narrow range level sensors are provided for continuous monitoring of RCS level during draindown operations. The level indicators provide monitoring 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 and high level alarms are provided. The wide range level instrument covers draining from the pressurizer to the bottom of the hot leg and will be available with the head on and off the vessel. The narrow range level , instrument covers reduced inventory operations. Two instrument ! types are being considered for level indication during reduced I inventory operations. The final instrument design will be I precented in the June 15, 1992 submittal. Either of the two- ! instrument types being considered is accurate for measuring level within the hot leg. During a draindown, level monitoring would be transitioned from the wide range level instrument to the narrow range instrument 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 indications.
2.3.3.2 SCS Desien The functional design of the Shutdown Cooling System (SCS) is substantially complete for System 80+. Some of the design features that improve the performance during shutdown operation are listed below. These features will be evaluated in the context of the results pending from activities described in Section 2.4 on decay heat removal and other sections of this report and will be l presented in the final submittal of this report.
- 1. The System 80+ SCS suction lines do not contain any loop seals.
An improved suction piping layout allows self venting (in as much as possible). Entrained air travels back up to the hot l-l 2.3-3 l
. - - - . - _ . . . . - . . - - _ . _ _ . , . . , . , . _ - . . , , , . , - . , . - - - - - , _ = -
--.-.-_- - - - - - _ - - - - . - - - . - ~ . - - . - - - . - - _ . . 77e.wp(9075)bh 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 flooded suction conditions at the shutdown cooling pump.
- 2. The two SCS suction lines are independent and redundant to each ;
other. Problems associated with a specific suction line would not limit the otP 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 designed 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 Nossle Dan Intecrity The System'90+ design addresses the NRC concern for preventing pressurization in the upper plenum of the reactor vessel during core boiling scenarios. 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 nozzle dams.- Based on field hydrostatic tests performed on nozzle dams, a conservative value of 40 psia is assumed for this pressure limit. In order to assure that the nozzle dam design pressure is not exceeded during reduced inventory operations .with boiling conditions in the reactor . vessel, the System.80+. design includes two blind-flanged connections off the primary safety valve inlet piping to be used as mid loop operation vent pathways. These connections, when opened to the containment atmosphere, provide sufficient vent capacity to prevent RCS pressurization and subsequent nozzle dam failure. These vents will , be opened prior to draining the RCS- for - reduced inventory operations and will not be closed until af ter the final RCS refill. Analyses have shown .that two 6 inch vents connected to_- the pressurizer safety valve piping and relieving to the pressurizer cubicle will'be sufficient for venting the RCS during RCS boiling. 2.3-4
- a. .- --.w.--
77a.wp(9075)bh These vents will allow an equilibrium pressure of approximately 70% of nozzle dam design pressure at 4 days after shutdown. The final analysis will provide details that will define the equilibrium pressure versus days af ter shutdown for the conservative case where no steam generators are available for reflux boiling. This data will be incorporated into guidance for the owner / operator to employ when planning outage evolutions. The guidance will address the safety and risk impacts of nozzle dam installation timing and will be provided in Section 2.1 of the final submittal of this report. Based on operating plant data, the time required to achieve the mid-loop condition from full power operation varies, depending on planned outage activities. Typically, a minimum for four days is required to cool down and then drain down the RCS from a full power condition. As a result of the analyses performed for Section 2.4 of this report, a Technical Specification regarding the earliest time after shutdown for entry to reduced inventory operations will be provided in Section 2.2 of the final submittal of this report. Such restrictions could minimize the consequences of a loss of shutdown cooling event during reduced inventory operations. l If multiple operator errors result in significant reactor vessel head pressurization, vessel water could be displaced )ato the cold legs. In operating plants, this situation has beer postulated to have it's greatest negative ef fect during RCP seal cnangeout, since it can lead to inventory loss thru the RCP. The System 80+ RCP design includes a shaft stop seal which uses the weight of the shaft and impeller to effect a seal between the RCP shaft and l casing when uncoupled from the motor during seal replacement. During seal replacement, when RCS fluid is present in the cold leg, any displaced inventory (and subsequent spillage) caused from manometric effects due to vessel head pressurization is minimized. 2.3.3.4 Alternate Inventory Additions and DHR Methods The effective management of time and efforts is crucial to coping with a loss of shutdown cooling. Awareness of time constraints provides information that is useful in deciding how to allocato l effort. If shutdown cooling cannot be restoreC within the time to l core uncovery, getting a source of water lined up to keep the core j covered becomes a first priority. Inventory makeup directly extends the margin of safety prior to uncovering the core. The recommended plan for coping with a loss of shutdown cooling includes the following steps: l 1. verify the proper functioning of level indications
- 2. verify reactor coolant pressure boundary integrity 2.3-5
77c.wp(9075)bh
- 3. establish the availability of makeup sources
- 4. if no makeup sources are available, remove RCS decay heat via reflux boiling, if steam generators are not isolated.
The System 80+ design ensures that adequate operabic equipment is available for mitigation of the effects of a total, sustained loss of shutdown cooling. Mitigating actions, such as operating redundant shutdown cooling pumps and containment spray pumps, are being developed and will be provided in the final submittal of this report. Safety injection system pumps will be employed for those scenarios where RCS pressure is too high for gravity feed. 2.3.3.5 OperatiqAs_ Procedural and Technical Specification changes necessary to support the program will be identified and implemented into the plant design basis. Procedural guidance for the conduct of mid-loop draindowns will be provided to assure that no testing or maintenance activity adversely affects the NSSS during mid-loop operations. This guidance will be reported in Section 2.1 of the final submittal of this report. Restrictions on maintenance and testing activities while the RCS is in a mid-loop or a reduced inventory condition are provided in plant Technical Specifications. The outage schedule is derived from Technical Specification Requirements, necessary plant maintenance and testing activities, and plant and equipment conditions required for these activities. Procedural guidance will specify that the plant operator suspend any activities which could alter the NSSS state or result in losing shutdown cooling during reduced inventory operations. This guidance will be provided in Section 2.1 of the final submittal of this report. Due to the Diablo Canyon incident and other industry events, operations in Mode 5 will be evaluated, including the rmquirements for evacuation of personnel from the containment building, closing
-f the containment equipment hatch and containment air lock doors, and isolation of penetrations to the outside atmosphere, as appropriate, based on time to boil and time to core uncovery criteria. A description of the containment closure conditions referred to, along with a description of containment closure design features, is contained in other Sections of this report. This evaluation and description will be included in Section 2.5 of the final submittal of this report.
Appropriate operating and emergency response guidance directs the operator in the proper conduct of reduced inventory operations. Training guidance aids in operator detection of abnormal conditions, and mitigation sequences. This guidance will be provided in Section 2.1 of the final submittal of this report. 2.3-6 '
770 vp(9075) bh 2.3.4 RESOLUTION The resolution of the reduced inventory issue on System 80+ will comprise the results of the analyses outlined above, related evaluations in Section 2.4 en availability of decay heat removal, Technical Specifications in Section 2.2 and procedural guidance in Section 2.1. These will be integrated and focused on tne reduced inventory issue in the June 15, 1992 updated submittal of this report. i i l l l 2.3-7
77 a .wp (907 5) bh 2.4 LOSS OF DICAY_]iKAT REMOVAL CAPABILITY 2.4.1 ISSUE Events that have occurred at operating plants demonstrate the vulnerability during shutdown Modes to loss of decay heat removal. The variety of maintenance activities taking place at shutdown i combined with the possible system and equipment interactions that l may occur lead to many conceivable scenarios for experiencing a I loss of decay heat removal. Three dominant design objectives have evolved from the emphasis placed on provention of shutdown events:
- 1. Provide redundant Shutdown Cooling System capacity and identify 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.
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 event 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 ri pture resulting from heatup af ter the coolant level drops below tc .s 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 Ndes 2 through 6. The second criterion 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 will be specified in the final submittal of this report. 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 initiators which potentially result in loss of shutdown cooling leading to boiling. Causes of past events considered include mid-loop operation, power failure and operator error. Thermal-hydraulic i 2.4-1 _ . _ __ - _ _ . _ _ _ ,_ - _ _ . _ _ _ _ . _ _ ~ - -
- - _. . . ._. ~ _ _ . -- - --- .- _ - _ - - _ -
77a.Up(9075)bh l l analyses specific to the System 80+ configurations co.1 firm the ! flexibility afforded the operator to mitigate these events. Appropriate Technical Specification limitations and procedural l guidance are identified by the analyses and will be provided in the I final submittal of this report. 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 identified in NUREG-1410. The discussion in this section is directly related to the means of coping with a loss of decay heat removal. This concern is also identified in NUREG-1410 and evaluated in Section 2.4.3.1. Section 2.4.3.3 presents the features of System 80+ that help assure the availability of the diesel generator. This issue was also identified in NUREG-1410. Availability of the diesel generator has boon a significant factor in numerous past events. Taken together, these sections demonstrate the integrated capability of the System 80+ to prevent and mit'. gate a loss of decay heat removal to ensure th:t the acceptance criteria aru not violated. 2.4.3.1 8hutdown Ry_e.At__ Initiation and Am!Lly_g_e!La This section Vill present event initiators and the means of prevention and mitigation on System 80+. Thermal-hydraulic analyses will evaluate scenario time sequences and identify available action times. Results will be presented in the June 15, 1992 updated submittal of this report. 2.4.3.2 System _80+ AC Power R3Jiability 2.4.3.2.1 Introduction i This section presents the System 80+ features that assure the availability of electrical power to supply the Class 1E 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 inventory in the reactor coolant system and i provides redundancy and flexibility to insure re-energizing the ! Class 1E buses is poscible if power is interrupted. l 2.4.3.2.2 Discussion 1 Electrical power sources need to be carefully managed dur'ng 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 1E buses and of 2.4-2
770.Up(9075) bh insuring an alternate sourco is available. The potential for a complete loss of decay heat removal due to the loss of electrical power is lowered when the electr4 cal 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 be written to identify the minimum acceptable electrical distribution system alignments for operating in shutdown modes and reduced inventory. The operation of the electr'al distribution system during shutdown modes and reduce-1 inventory a e Le guided by procedc as for normal slignments and for aligning niternate electrical sources if normal sources are interrupted. The electrical distribution system design will provide . xibility and redundancy to allow for the management of competing priorities during shutdown. These competing priorities include the need to perform maintenance on electrical system equipment versus tne need to have electrical sources available to provide power to the Class 1E buses. The System 80+ electrical system design (see Figure 2.4-1) provides the redundancy and flexibility to insure the risks associated with shutdown modes and reduced inventory operations are lowered to acceptable levels. This is accomplished by providing two independent divisions of AC Electrical Power. Each division has two 4.16 KW Safety Buses with tnree sources of electrical power. Two.of these are Class 1E sources. The two Class 1E sources are: (1) Normal - the Safety System Transformer being powered from the Switchyard Interface II, and (2) Emergency - the Diesel Generator. The third source is a backup from the Permanent Non-Safety (PNS) Bus. This backup cource (PNS-Bus) of power to the Safety Bus has three j sources of electrical power. The three sources are: (1) Normal - Th3 division related Unit Auxiliary Transformer (UAT) being powered from Switchyard Interface I through the Unit Main Transformer (UMT), (2) Alternate - The opposite divisions UAT being powered from Switchyard Interface I through the UMT, and (3) Backup - the Combustion Turbine. Therefore, the Class 1E Safety Buses have the potential to be fed
- from four different ultimate sources during shutdown modes and l reduced inventory operations. These sources are
- 1. Switcnyard Interface I,
- 2. Switchyard Interface II,
- 3. Dict.el Gcnerator, and l
- 4. Combustion Turbine.
2.4-3
77a.wp(9075)bh l This distribution system provides the shutdown management team with the flexibility to perform shutdown activities on one source of power to a division 4.16 KV Safety Bus and still maintain three diverse sources of reliable electrical power to the 4.16 KV Safety Bus. Along with the electrical system design features, the System 80+ Technical Specifications include shutdem modes and reduced inventory operation Limiting Conditions for Operations (LCOs) . The LCOs provide minirum acceptable electrical distribution alignments. Guidance is also provided by procedure to the operation staff to insure available source alignments are identified whenever shutdown activities are in progress. Additional procedural guidance is provided for aligning any ava'lable source (s) to the Safety Bus (es) if power to the bus (es) is interrupted. The procedure guidance and Technical Specifications will be provided in Sections 2.1 and 2.2 of the final submittal of this report. -2.4.3.2.3 Conclusion The System 30+ electrical distribution system design features provide the necessarf redundancy, flexibility, and diversity to reduce the likelihood of losing decay heat removal due to a loss of electrical power. The features 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 unlikely. 2.4.3.3 System 80+ Diesel Generator Availability 2.4.3.3.1 Introduction The availability of the Diesel Generator and the Diesel Loading Sequencer to automatically start and load during shutdown modes of cperation is one of the issues identified in NUREG-1410. The availability of Diesel Generator instrumentation and control system to provide reliable indications and automatic trip signals for Diecel 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 decay heat removal as presented in Section 2.4.3.1. The Diesel Generator (D/G) 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 D/G provides signals to start the diesel for emergency operation, applicable protective trips to prevent or limit damage to the D/G 2.4-4
77a.wp(9075)bh at all times and D/G 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 D/G are provided in the Control Room and at the local control panel. 2.4.3.3.2 Discussica The Diese' Generator (D/G) and Diesel Load Sequencer need t c, maintain a consistent means of operation independent of the plant operation condition. This ensures the operating staff is not required to harn different operating schemes and therefora reduces potent.- , E error. The % stem a 1 and DLS provides this simplicity of operation. The '4 - trgency source of power to the Class 1E bus. The D/G cr- are available for operation during shutdown condi undezgoing maintenance. The Class 1E buses are monit ? . dervoltage and degraded voltage conditions. If either ,n is sensed, the D/G is started and the DLS is initiatem s ee attached CESSAR-DC Figure 7.3-5). For a loss of power to the Class 1E bus, the response of the D/G and DLS is not dependent on 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 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 D/G I&C system needs to ensure the diesel is protected during all modes of operation. However, certain protective trips need to be bypassed during_ emergency operation. The System 80+ Diesel Generator protectior. system provides automatic trips to prevent or limit damage to the C/G. 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 emergency 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 on plant I
l 2.4-5
770.wp(907S)bh operational modes. The sensing of an undervoltage or degraded voltage condition during shutdown causes an automatic D/G 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 D/G independent of plant operating node. The I&C system needs to ensure the operator is informed of the D/G's operational status. This status includes parameter indications and alarms. The I&C systems need to provide controls to allow the operator to start and load the diesel to provide power to the Class lE buses. This status and control scheme needs to be provided locally and in the control room. The System 80+ control room is designated as the Nuplex 80+ Advanced Control Complex (ACC). The Nuplex 80+ ACC presents the operator with the information and controls necessary to complete any tasks identified in a task analysis process. The task analysis for D/G operation identifies the parameters, alarms, and controls required to operate the D/G from the Nuplex 80+ ACC, This identified status and control scheme is presented to the Control Room Operator on the Electrical Distribution Auxiliary Console. The presentation of this information is accomplished in accordance with a structural and hierarchial format discussed in CESSAR-DC Section 18.7.1. This formatting provides the operator with parameter displays, alarm status, alarm categorization, and alarm priority. This method of information presentation provides the Control Room Operator (CRO) with the tools necessary to monitor and/or diagnose D/G status. The System 80+ local control panels for the D/G provides the Plant Equipment Operator with the same information and controls as is available to the CRO. The D/G status information and control scheme on the local control panel utilizes the same Man-Machine Interface (MMI) features used in the Nuplex 80+ ACC. These features meet the System 80+ human factors standards and guidelinen. 2.4.3.3.3 Conclusion The System 80+ Diesel Generator instrumentation and control systems design features provide starting signals for the D/G and DLS initiation and protective trip signals for D/G emergency operation and provide D/G status information to the control room and local control panel which allows the operator to operate, monitor, and diagnose D/G and DLS operation. These features of the System 80+ design enhance the operator's interface with the emergency equipment and reduces the potential of human error. 2.4-6
,'ti-k l PREFERRED SWITCHYARD INTERFACE 18 l l' REFERRED SWITCHYT.RD MTERFACE I l l PREFERRED SWITCHYARD MTERFACE u v SAFETY SYSTEMS . UNIT MAIN uvuvuv viu uvuvuu' UNIT MAIN SAFETY SYSTEMSuu g ^ ^ TRANSFORMER TRANSFORMER """"^^ ' CEO ^ ^^ ^ ^ ^ TRANSFORMER - ' TRANSFORMER ^"
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1770.wp (907 5) bh 2.5: PRIMARY /8ECONDARY CONTAINMENT CAPABILITY AND SOURCE IEBM >
.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 open.
ThisDissue is related to events initiated in Modes 5 or 6 which have the potential _ for -radiological release. The events considered are the loss of decay heat removal capability initiated by either-a loss of shutdown cooling or by a loss of coolant accident caused by:either a pipe break or operator error. Following a loss of decay heat removal that is 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 additicnal time to evacuate, close and isolate _the containment. The time for the coolant to reach saturation is a function of plant conditions at the sine the event is initiated.
4 Time to bestore includes the time to detect that decay heat removal has-.beed lost plus the time.either to restore shutdown cooling or to initiate alternative 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 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 oc: I
- Design, operation, condition and status of equipment to close-penetrations, equipment-hatches-and personnel air-locks;
-* Procedures for routing material and lines through these-openings; e : Training-of_ personnel; and
- Conditions of pressure, temperature and radiation within the containment as--the core uncovers.
- 2.5.2, ACCEPTANCE CRITERIA The following acceptance criteria apply to the issue addressed in this_section:-
l 2.5-1
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- it " ?
77a.wp(9075)bh ) l l l l
' 1. = Radiological exposure of the public to_any event rGsulting in a loss ofJdecay heat removal, shall be limited to a fraction of - - the .10CFR100 limits . that - will - be specified in the final submittal'of this reoort.
2 .- Radiological exposure of-the public to'any event resulting'in a pipe break shall-be limited to the limit stated in 10CFR100, 2.5.3 DISCUSSION - Evaluations of postulated initiators, event progression and timing,
- mitigating:. equipment, operator actions and source terms are - in.
-progress and - - will be _ presented in the June 15,- 1992 updated.
submittal =of this report.
'I 2.5.4 RESOLUTION The resolution of the open containment radiological release issue for System 80+-.will be presented in the June 15, 1992 updated
- submittal of-this report.
2.5-2
- _ _ _ _ _ _ __..y..~ _ _ . . . _ ,
77a.wp(9075)bh 5
- 2.6 RAPID BORON DILUTIONS 2.6.14 ISSUE The ' issues - of = the ~ rapid boron dilution can be broken down into three categories as_follows: ~
- 1. The introduction of deborated water into the RCS via Shutdown 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 has occurred _ and the . charging pumps are returned on -line, powered by the Emergency Diesel Generators. If the plant were in startup node - 1.e., deboration in progress - the charging , pumps could continue to operate causing a " slug" of unborated water to collect in the lower-plenum of thefreactor vessel. If it.is-thenLassumed that offsite power is restored and the. RCP's are restarted,-then a water slug of deborated water can- ' be. injected into the core.
- 3. A potential' boron dilution resulting-.from inleaksge from the secondary side of a steam generator during a SGT: event.
. All' the - above issues will .be addressed in the _Iscussion and resolution sections of;this report.
-- 2.6.2 ACCEPTANCE CRITERIA The acceptance criteria _for the rapid-boron dilution event should be' consistent with the acceptance criteria that are necessary to j meet the relevant requirements of-GDC 10, 15 and 2 6. Specifically, these criteria are as follows: :
1.. ' Pressure in the reactor coolant and main steam systems should
-be maintained below the RCS P/T limits (see Figure 3.4.3-1 of- '
Technical - Specification 3.4.3). or below 1110%_ of the design value, whichever l's :less. -
~2 -- . Fuel cladding integrity _shall be-maintained by ensuring that the minimum DNBR remains above the'95/95;DNBR limit for PWRs
-and CPR remains atove thecMCPR safety limit for BWRs based on
-acceptable. correlations (see SRP'Section-4.4).
- 3. -An incident.of moderate frequency shoul'd not generate a more serious- plant -condition without other faults- occurring independently.
l-L 2.6-1
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- 4. An incident of moderate frequency in combination with any single - active component failure, or single operator error, shall-be considered and is an event for which an estimate of the number of potential fuel failures shall be provided for radiological dose calculetions. For such accidents, the number of fuel failures must be assumed for all rods for which rie DNBR or CPR falls below those values cited above for cladding integrity unless it can be shown, based on an acceptable fuel damage model (see SRP Section 4.2) , that fewer failures occur.
There shall be no loss of function of any fission product barrier other than the fuel cladding. The above criteria are the same requirements as the acceptance criteria for the Inadvertent Boron Dilution (IBD) event as stated in NUREG-0800 Section 15.4.6, Reference 6, with the exception of Item 5. This criteria states that the available operator action time be 30 minutes for an IBD event during refueling conditions and 15 minutes for startup, cold shutdown and power operation. This requirement is not applicable to a " rapid" boron dilution event. 2.6.3 DISCUSSION Analyses of the rapid baron dilution event initiated from a shutdown mode are in progress and vill be presented in the June 15, 1992 updated submittal of this report. 2.6.9 RESOLUTION The resolution of the rapid boron dilution issue will be presented in the June 15, 1992 updated submittal of this report. l i 2.6-2
- 77Q.wp(9075)bh.-
2.7 FIRE. PROTECTION 2.7.1 Iss0E' 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 is the reduced level of fire 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 susceptability of failure due to fire. 2.7.2 ACCEPTANCE CRITERIA A defense in depth philosophy shall be employed in the design of the fire protection system in order to reduce the overall shutdown risk due to fire. The elements in tnis defense in depth philosophy are: s 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- Mitication of Fire Consecuences 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 sealed with assemblies that are qualified to maintain the integrity of-the 3-hour rating. This assures - that a fire involving one division of Shutdown Cooling System components will not effect the redundant division.
l 2.7-1
770.wp(9075)bh INTERDIVISIONAL SEPARATION 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 manipulations guided by approved procedures. For each division, the shutdown 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 each 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 Safety 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 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 SupDression of Fires DETECTION Fire Area 38 contains the Division 1 containment spray pump and l heat exchanger and Fire Area 41 contains the shutdown cooling pump and heat exchanger. These areas were evaluates during the recently completed System 80+ Fire Hazards Assessment. This assessment considered the fixed and transient combustible loads in these areas and the importance of the componentn to plant shutdown. Both areas will be equipped with full area coverage ceiling mounted lonizati 1 l 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 l design process. l l The detection system is highly reliable and will be kept in service at all times, even during shutdown modes of operation, i SUPPRESSION l The System 80+ Fire Hazards Assessment concludes that a fixed I automatic suppression in the form of automatic sprinklers is not l warranted. This is due to the minimal combustible loadings in j these areas. This will be verified later by engineering analysis, j 2.7-2
6a. e m,. p - a , 4a 4AsA J a,J2_.._e*sm- 12. _k,+s4 E.h 77a.wp(9075)bh 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 acconplished without exposing either the redundant M vision equipment or interdivisional equipment to the ef fects of s.noke or hot gases f rom a fire. MANUAL FIRE FIGHTING A fully trained and equipped on-site fire brigade 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. MAINTAINED LEVEL OF FIRE PROTECTION The System 80+ fire protection system is not degraded or reduced during plant shutdown. There will be 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 Prevention of Fires Prevention 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 sole]y on an arbitrary " good l engineering practice" approach, but will consider the amour. of combustibles necessary to result in a fire that could cause unacceptable damage. The control 'of hotwork and housekeeping procedures will be developed by the owner-operator and implemented 2.7-3
i 77a.wp(9075)bh-so as to not place unnecessary restrictions on shutdown maintenance activities, yet_will provide a high level of fire prevention. 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 reduce the risk due to fire during shutdown operation to an acceptable level. The combination of fire protection featares resulting from employing the defense 4.n depth philosophy will minimize the potential for fire damage to systems required for shutdown operations. This issue has been resolved by the design features of system 80+. 2.7-4
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77a.Tip(9075)bh 2.8 INSTRUMENTATION 2.8.1 ISSUE Over the past several years, industry and regulatory 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 ralated to Nuclear Steam Supply Systems (NSSS) operations while on shutdown cooliag; 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 differentiat:ng between the anticipated dynamic effects of the drainc.own 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 a timely response to a loss of shutdown cooling event. The 1.istrumentation should " provide reliable indication of parameters that describe the state of the 2.8-1
. 77a .wp(9075) bh .
-Reactor. Coolant System (RCS) and the performance _of systems: normally used to cool the RCS for both normal 'and s
2 accident conditions"-(Reference-4). The NRC has'specified that instrumentation for reduced _ inventory
-conditions should provide both visible and audible indic.ations of 1
abnormal conditions:in reactor vessel temperature and level, and decay heat removalLsystem performance. P 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 reduced inventory conditions, and to aid in correctly interpreting-a-loss'of shutdown cooling,-should one occur. The following recommen'dations are takan - frca Enclosure 2 to
' Reference l4:
"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 least - 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 l-
- 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 recommendations. These include, for example: y "1. We' suggest that _ licensees- investigate ways to provide l [ accurate) temperature [ measurements] even if the head is l removed, particularly if a lowered RCS inventory condition L , exists. g 2. We expect sufficient information [be provided) to the l operators that an approaching [DHR system] malfunction is ! clearly indicated. 2.8-2
, , - . ~ . - - , . - - - . . - .
-. - - ~_- -. -. .- . . -.
- 770.wp(9075)bh We expect both-audible. alarms and a panel indication when
~
3. conditions exist which jeopardize continued operation of a DHR system,-as:well as when DHR is lost.- 4.5 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 DNR onditions '
; since it provides an_ indication.of core coverage and ...
of the - time to core uncovery. It is also useful in
- mitigating the loss of DHR accident."
. Section _2. 8. 3. 2 of this report contains the description of the System-_ 80+ instrumentation package _ _ for - reduced inventory operations, including:- "
tha' monitored parameters, 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 will be included in-the final submittal of this report. 2.3.3 DISCUSSION 2.8.3.1 Instrumentation Desian Basis-
- To effectively monitor the draindown process-to mid-loop via System 80+- enhanced instrumentationi information obtained from plant '~ analyses forms the basis for-the instrument's design requirements.
Instrumentation specified for reduced inventory operations is based on analyses lin;the following areas: operatilons . from_ ~a solid- plant to mid-loop conditions (wnich 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 5
77c.wp(9075)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 thommal hydraulic effects and manometric effects; and mitigetion 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 cannot be achieved without a complete understanding of plant behavior during reduced inventory operations. 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 detail will be provided in the June 15, 1992 updated submittal of this report. 2.8.3.2.1 Level Two diverse instruments types will be provided for the measurement of level during reduced inventory operations. These instruments will make up the refueling water level indication system. A wide range dP based level sensor will be provided which functions to meauure level between the pressurizer and the junction of the shutdown cooling suction line on the hot leg. A narrow range system will be provided which measures RCS coolant level during reduced inventory operations. Two instrument types are being considered for level indication during reduced inventory operations. Although complete details of the final instrument design will be presented in the June 15, 1992 submittal, brief descriptions of each of the instrument types are provided below for information. The choice of instrument type does not impact the evaluations presented in this report, given that the measurement function will be provided in the plant. A narrow range heated junction thermocouple based system is being considered which uses input from four separate heated junction thermocouple probes for accurate measurement of level in the 2.8-4
l 7 7 a . wp(9075) bh i reactor vessel and hot leg region. The heated junction thermocouple system combines the output from two inadequate core cooling heated junction thermocouples and two heated- junction ' thermocouple probes designed specifically for the hot leg region of measurement (see Figure 2.8-2). The range of the inadequate core cooling heated junction thermocouple probe extends from the reactor vessel head to the fuel alignment plate. The range of the clustered heated junction thermocouple probe emphasizes the hot leg region'. Thus, two control room displays are obtained from combining the output from one inadequate core cooling HJTC and one clustered HJTC probe. Thus, the operator is provided with two continuous, independent RCS level indications in the control room for reactor vessel level. The heated junction thermocouple system compensates for the flow gradient across the core associated with the operation of only one shutdown cooling suction line. The instruments are located in areas of the core which minimize the effect of core outlet nozzle exit ef fects. The instrument sensors have an accuracy and response time consistent with the maximum drain down rate of the RCS. The instruments -are designed so that the signal and power are transmitted on individual electrical conductors. Failure ot one sensor will not result in the loss of signal from the remaining sensors. The measurements of RCS water level obtained via these probes are limited to those periods when the reactor vessel head is installed. A narrow range dP based level sensor system is being considered which functions to measure level in the hot leg region. The lower tap is located at the hot leg /SCS suction line interface. The l reference leg for this system will be located either at a direct vessel injection nozzle or will be opened to atmosphere. The measurement of RCS water level obtained via the dP level sensor is available when the head is on or off the vessel, t The use of both a wide range pressure differential (dP) instrument and either two pairs of heated junction thermocouple (HJTC) based instruments or a narrow range dP based instrument for refueling water level monitoring provides highly reliable, redundant, and independent indications of vessel water level. Instrument ranges provide a continual'draindown measurement from the pressurizer to a level below that required for SCS operation. Since the level
-instrumentation is independent, common mode misoperation or failures due to dynamic effects will not be masked.
Each independent level instrument provides a suitable measurement L and is accurate for it's intended range of use. For mid-loop operations, the HJTC based level probe provides accurate level measurements to within less than 1 inch of vessel level. This is key, since there is a very narrow margin between manway flooding (or nozzle dam installation) and SCS pump cavitation. The level 1 2.8-5
. . _ _ _ __ _ -_ _ .~
77a.wp(9075)bh instrumentation is alarmed in the Control Room to reflect this tight operational tolerance in the hot leg. 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 instrument temperature sensor (HJTC system only). All provide representative indications of the core exit temperature when the shutdown cooling syatem is operational. If the shutdown cooling system is lost, the CETs, hot leg RTDs, and reduced inventory water level thermocouple sensor 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, while fuel is not being shuffled, the CETs. The 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 penetrate l the internal diameter of the hot leg pipe at approximately 10" l 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 less of shutdown cooling and tracking the approach to boiling. Awareness l' of time constraints provides information that is useful for deciding how to allocate effort. l l 2.8-6
77 a .wp (907 5) bh 2.8.3.2.3 Shutdown Cooling System Performance As stated in Enclosure 2 to Reference 4, sufficient information will be available to the contro) room operator to indicate an approaching shutdown cooling system malfunction. Indications of sufficient pump suction pressure and possible vortexing include unsteady pump 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 or rising RCS level !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 indicator (ICCI heated junction thermocouple based design) shutdown cooling flowmeter shutdown cooling heat exchanger inlet and return line temperature sensors
- shutdown cooling valve position indicators The safety related designation of these instruments is a consequence of their required functions in other plant modes of operation,~ including for some, inadequate core ccoling. 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. 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 1
1 1 2.8-7 t
77a.wp(9075)bh l quality program be implemented that assures that quality is commensurate with intended use. In the procurement of the instrumentation, appropriate technical requirements cnd quality requirements are specified in the purchase order to this end. 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 .... " Therefore, the following list of Quality Class 2 .'nstruments identified on Table 2.8-1 are classified as non saf ety-re lated: refueling water level indicator (wide and narrow range dP design), refueling water level indicator (clustered heated junction thermocouple based design), shutdown cooling pump suction and discharge pressure eensors, 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 Status Overview (IPSO),
.2. Alarm tiles and associated alarm messages,
- 3. Discrete indicators which provide frequently used and important information,
- 4. CRT display formats containing essentially 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, and displays applicable- to depressurized, shutdown conditions. They are described in the following sections. 2.8-8
77a.wp(9075)bh' 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 dCtion. Information provided on the IPSO display includes:
- 1. Major system and component 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 loc ation 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 td equipment dependent to ensure their validity for different rational conditions. For all modes, including shutdown and ref ulng conditions, individual sensed proce"s parameter values an alarm states are used to determine crit.ical function alarms, either directly or as processed by an algorit.im 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. 2.8-9
7"ic , wp (907 3) bh The following systems have dynamic operating status representations I on IPSO. Their identifying descriptors on the IPSO display are shown below: I CC - Component cooling water , CD - Condensate l CI - Containment isolation CS - Containment spray CW - Circulating water EF - Emergency feedwater FW - Feodwater IA - Instrument air , SC - Shutdown cooling ) RC - Reactor coolant I SI - Safety injection SW - Service water TB - Turbin) bypass SD - Safcty Depressurization System information presented on IPSO includos 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 an operator about possibic 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 Massages Alarm tiles are displayed on electroluminescent ilat panel displays in the Discrete Indication Alarm System (DIAS). These tiles are functionally grouped and located on the appropriate control room panel. Shitdown cooling system alarm tiles are located on the Engineared Saf ety 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 incluce (for each train): los shutdown cooling pump header pressure - low shutd Wn cooling flow high shutdown cooling heat exchanger outlet temperature shutdown cooling pump motor o rent deviation 2.8-10
. ~ . - . _ . - - . . - .. -. -. . . . - _.
770.wp(1075) M r In addition, this panal will nave a tile for RCS conditions, with individual inputs for shutdown, depressurized conditions:
- low RCS water love; high core exit temperature low refueling cavity level To 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 f*atures eliminate nuicance alarms and help ensure a true " dark bcard" 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 accmrements exist, and include trend information for routine monif or .g, and diagnosis of abnormal conditions. Where analog data v aposed of dif ferent ranges of information, DIAS autor.atically ^.W,s to the appropriate range, and indicates to the operator n>t a range change has occurred. Discrete Indicator displays to support shutdown cooling for key parameters are on the Engineered Safety Features panel. These include: Shutdown coolina System (ner train)
- inlet temperature outlet temperature
. pump header pressure flow heat exchanger inlet temperature heat exchanger outlet temperaturc 2.8-11
770.wp(9073)bh pump mctor current Reactor coolant SysteJD pressurizer level reactor coolant system level pressure
- core exit temperature - refueling cavity level 2.8.3.2.5.* 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 concistent with the operator'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 operatcr 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, detail / diagnostic. Each level of the hierarchy 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. Opera +or 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 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 diLplays 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 integt ste the shutdown cooling and RCS status for this display. The RCS is also presented on a separate Level 2 display. 2.8-12
77a.wp(9075)bh
.t.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 equipinent 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 integrated with DIAS alarm tiles important to shutdown cooling and with CRT display of the shutdown cooling system. Controls, alarms and CRT displays for other systems applicable to shutdown 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 NUPLEX 80+ alarm system that support operator diagnosis of alarm conditions and that would be particularly supportiv: of depressurized, shutdown eperations. Tnese are:
- 1. Mode and Equipment Status Dependency
- 2. Audible Alarm Information
- 3. Stop Flash Feature
- 4. Operator Established Ala-as
- 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 p a t..i status, are presented only on alarm tiles. These would not be included as input to alarms shown on IPSO. 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, the operator can distinguish between lower level alarms that are critical function or success path related, and those that are not (e.g., personnel hazard, or equipment damage). 2.8.4 RESOLUTION Tne issue of instrumentation for shutdown operation is resolved cn System 80+ by the instrumentation and control room displays described in the previous sections of this report. This 2.8-13 (
- ~ _ _ . ...__ _ . _ _ - _ _ -._ _ _ - ____ _ ._.. _ _ - _ _ _ _ ._ __.. _ - -
77a.wp(9075)bh-instrumentation will ucct or oxceed th9 recommendations of ueneric Letter 88-17, and till signilicantly reduct. tisk associated with operations during shutdown, particularly-when the reactor is in a reduced inventory condition, as long as prior to the start of l draining the reduced inventory instrumentation is placed into l operation. i The NU?'2X 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 a'e r ! provided on discrete indicators, 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 indi.vidual effect on plant critical functions such as reactivity cr. rol, core heat removal and RCS heat removal. , l r 2.8-14
. . _ _ - . . _ _ _ _ . _ . ~ , . .._.._. . . - . . _ , . _ , _ _ . . . -
77a.wp(9075)bh Table 2.8-1 Reduced Inventory Instrumentation Package Monitored Inst rumnt irr t runent Indication and P er e"e t e r Type f unc t ion Range Alarm Locetton Certwnts RCS We".ec level Refueling We'er Continuous wide Wide Range; Control Roca, mighly reliable. Level Indication range RC$ water .ap at hot leg / with low level Meets hRC System (dP based level indication GCl suction eterm, requitecent f or cesign) during draindown line inter 4 ace, water tevel operations. reference leg nessurement to a at top of point lower than pressuriger, that required for St$ operetton. RCS Water level Ref uellr's Water I rdererdent , top of the Control Roum i located near Levet trdlS* ion continuous narrow vessel to the with low and each hot Leg. System (hesied range level f uel al t gruent high level EffectivJ when Junction indication in the plate. storms. $C$ suction le thermnceuple het leg region. eligF 4 to based design) eith> t leg. 2 Axi. +1ngs of the,ne* couples. One spans from the vesset heed to the fuel ellen-nent plate. The other contains therwocouples clustered over hot leg region. System provides excellent occuracy and continuous measuremtnt. RCS Water Level Refueling Water Continucus narrow harrow rangel Control Rocun Highly tellable Level Indication range level top at $C$ with low and for mid loop (dP based irdication during suction line/ high level operations, design) reduced inventory het leg alarms. Meets WRC operations. Interface, requirement for ref erence leg water level at DVI nottle measurement to a or open to point lower than atmosphere, required for $C$ operation. RCS Teeperature CEis Measures optimited for Control Room Tracks approach (thermocouple ten erature of $C$ and with alarms et to bolling. design) coolant exiting refueling high and high. Tenpera turn core, modes. Will high tenperature indication measure provided even belling. when head is off A mroximate vessel, hot range 100 250 avaltable during deg F. fuel shuffling. Availability will be maximited. 2.8-15
770.wp(9075)bh Table 2.51 (Continued) Redxed Inventory Instrw entation Package Moni t or ed Inst rwent Inst rwent Indication ard Paranet er type f unc t ion Range Alarm Location Convent s RC$ tenperature Refueling water Continuous, Optimited for Control Room Indicates actual Livet probe frdependent SC$ and with storms at vessel water (heated jmetion tecperature refueling high and high* tecperature, thermoc ouple measurecent inside modes. Witt high Tracks approach based design) the vessel. nessure terperatures. to boiling. boiling. Approximate range 100 250 des F. Not Leg seryttature Resistance Measures cure eAlt OptimiF 1 for Control Room, Terperature Tent 4rature terperature in the shutdown alarms at high Indication is Detectors (RfDs) hot leg at both operations. t ecpe r a t ur e. affected by less
$C$ saction line Approntmate of shutdown regions, range $0 250 cooling flew, Redundant RfDs deg F. since flow by provided in each the RfDs will hot leg. not occur. $C$ Flowrote F lownet er Decay heat removal tourds SC$ pwp Control Room, One located in syste a flow range. InctWes low each $c$ return performante, flow alarm. Line to the RC$.
Can be used to measure C$P fIow if C$Ps are used for $CL
$!$Pwp/C$Pwp Pressure sensor Measures 0 to system control Room one instr ment Discharge Pressure individual pro design with low located at the discharge pressure, pressure. discharge to pressures. each pwp.
Identifies irdivictual pwp status. SC$ Pwp/CS Pe p Anunet er Measure current 0 to design Control Room, Confirms pwp Motor Current drawn by pwp storms with status motor, preset drop in (individaal pwp Fluctuatices show current, air entrairvrent) air entrairvnent. Independent of pressure ard flow irdicators.
$C$ Pwp/CS Pwp Pressure sensor Measure prp 0 to system Control Room One inst rwent su.tlon pressure suction pressure design with low located at the in each pwp. pressure, pressure alarm, suction of each pwp.
identifies individual pwp status.
$CNx Intet and teturn Temperature Measures 40 392 deg. Contrtt koom fenperature Line fenpereture sensor tewerature in the F. alarms at high irdication only suction and tecperature avaltable uhen discharge lines of SC$ is the shutdown operational, cooling heat exchanger.
2.8-16
77a.wp(9075)bh , Table 2.8 1 (Continued) keduced inventory Instrtsnentation Package
- Monitored Inst rwent Ins t h.anent Irdicatien and Parametei type Function Range Alarm Location Cwrentt SCS Valve Position Valve position Status of valve Open/tlosed/the control toori Witt provide t indication ottled position Indication positions in the information of cpen/ closed or SC$. Indication, system lineup throttled. status and o' ellebte flowpaths.
4 2.8-17 c
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= Figure jfd / REACTOR COOLANT SYSTEM ELEVATIONS RELATED TO SHUTDOWN COOLING OPERATIONS 2.8 1
)
I CLUSTEREO I TH2RMOCOUPLE $
\ PROBE REACTOR
\ ;
ICCl HJTC PROBE VESSEL Ng - FLANGE '. ' ! UPPER HEAD I ( SPANNING REGION - f DRAINDOWN
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/ STRUCTURE SUPPORT PLATE
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- SPANNING THE
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. s s s -
/ FUEL AllGNMENT PLATE
- , -.p-l EX AMPLE OF SENSOR POSITIONING. NOT INTENDED TO BE TO SCALE (ONE OF TWO INSTRUMENT PAIRS SHOWN) SCHEMATIC REPRESENTATION OF THE NARROW RANGE' IU " J ga f HEATED JUNCTION THERMOCOUPLE BASED REFUELING WATER LEVEL INDICATION SCHEME 2.8 2 r - % s -.2..-we-- ...,_m- -,3xww.- - 3, . . - - _ , --wrv,,-,---,,m. .7+i, .. v,, _ - . . , . , . , .
77a.wp(9075)bh 2.9 ffQS__ RECIRCULATION CAPAJ11LTX 2.9.1 ISSUE The issue is the potential for loss of flow to the Containment i Spray (CS) and Gafety Injection (SI) pumps during accident conditions. Syctem flow could be inhiLited by a number of f actors. These factors includet
- 1. Hydraulic effects, such as air ingestion and vortex formation.
- 2. Dobris 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 CRITF.RIA The design of the System 00+ Incontainment Refueling Water Storage Tank (IRWST) and Holdup Volume Tank (HVT) and their associated debris-blocking devices shall comply with the requirements of General 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 reliable source of water to the SI pumps for long-term recirculation. 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 the acceptance criceria specified in USNRC Standard Review Plan section 6.3, 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. 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 assemblies surrounding these suction points. Regulatory Guide 1.82. Water sources-for Long-term Recirculation Cooling Following a Loss-of- ' Coolant Accident, Rev.1, provides the guidelines for the design of the IRWST and the HVT, and the design of the screens associated with these tanks. Technical considerations related to this issue are detailed in NUREG-0897, Containment Emergency Sump Performance, l Revision 1. l 2.9-1
770.wp(9075)bh 2.9.3 DISCUSSION Watsr introduced into the System 80+ containment from a RCS break c,r 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 IRWST 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 containtrent 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 kafety 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 So,'rces for Long-tern Recirculation Cooling Following a LOCA. .he general plant arrangement separates redundant trains of the S23 and the CSS. Tho ' divisior-1 boundary provides complete separation between divisions and effectiveJu creates two identical support buildings. The result is a p1d.e arrangement with two SI pumps and one CS pump in each division. Within each division, the two SI trains (and each CS train) cre separated by a quadrant wall to isolate the trains from each other to the maximum extent practical. Each of the four SI pumps has its own suction connection to the IRWST (see Figure 2.9-1) and each cf the two CS pumps rehares one of these four connections. Following an accident, water introduced into contains 7t drains to the Holdup Volume Tank. Debris that may exist in containment may be transported to the HVT with this fluid. Debris greater than 1.5
- inches diameter is prevented from enter'.ng 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.
l The Holdup Volume Tank is designed to function as a solids trap to I help prevent debris from entering the IRWST. High density debris l that makes its way through the trash rack will accumulate in the i bottom of this tank. The IRWST spillways are located at a high l enough elevation to assure that much of the higher-density debris (and debris that tends to sink slowly) will settle to the bottom of l l 2.9-2 l
-. -- -- .. .-.- -_ _- - _ _ - _ ___ - - =__._ - - . - 77a wp(9075) bh the HVT before spilling over into the IRWST. Debris that remains in suspension will make its way to the IRWST spillways. The spillways are shown in Figure 2.9-3. Screens are not present in these spillways to assure uninterrupted flow to the IRWST. The fine debris that is introduced into the IRWST is prevented from entering the SIS suction piping by a debris screen. These screens are located at each end of the four wing walls that serve as supports for the reactor coolant pumps (see Figures 2.)-1 and 2.9-2). These wing wall assemblies extend from the IRWST floor to the maximum 1RWST water level, assuring that all debris will be i filtered before reaching the SIS suction lines. The screen assemblies conpletely enclose the suction lines by running from the end of each wing wall to the side walls of the Holdup Volume Tank or the primary shield walls, as applicable. The wing-wall screens have the capability of removing particles greater than 0.09 inches diameter. This screen size is consistent with the screens used on currently operating units. The wing-wall screens are the final barrier to debris before the SIS suction lines. Blockage of the debris screens is a major concern with respect to recirculation. The System 80+ screens have a vertical orientation to prevent debris from settling on the screen surfaces. This helps in keeping the screens clear. The design considered the types and quantities of insulation used for the System 80+ components, since post-LOCA deterioration of this insulation is the major potential source of debris in containment. The location of insulation with respect to the HVT and IRWST as well as the possible location of breaks have also been considered, The effective areas of the screenn have been determined according to the guidelines provided in Appendix A to Regulatory Guide 1.82, Guidelines for Review of Sump Design and Water Sources for Emergency Core Cooling. The debris screens have been designed to withstand the vibratory motion of a seismic-event without loss of structural ir.tegrity. Each screen is capable of withstanding loads imposed by postulated missiles as well as loads due to pressure head differentials. Consideration has also been given to the materials used for the debris screens. Materials have been selected to avoid degradation during periods of inactivity (i.e., no submergence), and during periods in which the 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. Tho screens will be visually examined periodically to octect any corrosion or structural degradation during refuelin; outage periods. As seen in figure 2.9-1, the suction lines are located
'ithin the confines of the wing wall away from the IRWST spargers.
This wa.1 1 design helps isolate the suction lines from the open sections of the IRWST, where most of the maintenance activities 2.9-3
= -- . . . - - . - - . - - - - _ - - -
l 77a4wp(9075)bh l l will be performed. The fine wing-wall screen will filter any trash i generated from this type of activity. In the event that maintenance is needed within the wing walls and near the ECCS suction inlets, permanent box-like screens over the suction piping will protect these lines. Long-term return of spray water from upper level elevations is not dependent on individual piping runs or spillways Multiple passive l spillways are provided to route water back to the Holdup Volume Tank. Major openings such as hatches and stairwells are also available to return water to the screened entrance to the HVT. Protection against air ingestion by SIS pumps is also a major concern with respect to recirculation and has been considered in the System 80+ 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 for Emergency Core Cooling. The minimum water level in the IRWST has been conservatively calculated to be 75+6 (elevation). This water level allows for sufficient NPSH for the Containment Spray pumps and Safety Injection pumps operating at runout flow. A conservative margin has been provided between the elevation of the suction piping opening and this minimum water level to minimize the possibility 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 Appendix 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 with the minimum IRWST water level. 2.C.4 RESOLUTION The design of the System 80+ IRWST and HVT 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 SIS and CSS pump l 2.9-4
77 a .vp (907 5) bh recirculation, but the protection afforded the SIS pumps against debris ingestion or blockage is significantly greater than in current designs. First, water in containment draining back to the IRWST must pass through a large trash rack before entering the HVT. The HVT serves as an ef fective solids trap f or 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 takes suction from one of these four lines. Box screens at all four suction lines provide a final trap. Multiple spillways are available to return water from the upper containment elevations to thn IRWST. The drain pathways are fully - redundant 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 and associated effects (including debris screen blockage), and the preservation of SIS pump NPSH during post-LOCA conditions in the overall design. The performance e of the design is deemed acceptable with respect to these considerations. 4 This issue has been resolved by design features of System 80+. 2.9-5
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_ _. - - _ _ -_. -_. . _ _ _ . _ _ _ ~ - __ 7 7 a . wp (907 5) bh 2.10 E1]ICTS OF PWR UPPER INTERNJLLS 2.10.1 ISSUE Events with tu<t pNential for loss of Decay Heat Removal (DilR) 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 conditions, the reactor vessel upper internals may provide suf ficient hydraulic resistance to natural circulation flow between the refueling pool and the reactor core to inhibit, or even prevent, the refueling pool water from cooling the core under circumstances when forced convection DHR has been lost. 2.10.2 ACCEPTANCE CRITERIA When the reactor vessel head is off and the core 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 evacuate and establish containment integrity; and
- 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 DISCUSSION An evaluation of the hydraulic flow resistance through the upper internals, of the time to boil and of the characteristics of the SDC SysteN will be presented in the June 15, 1992 updated submittal of this report. 2.10.4 RESOLUTION resolution of loss of DHR while the upper internals are in the asel will be presented in the June 15, 1992 updated submittal of a report. 2.10-1
772.Vp(9075)bh l 2.11 ZPJk_HMD11NG AND HEAVY LOAQR 2.11.1 ISSUE Questions have been 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 [HACTS), accidental release of a fuel assembly, and movement of the spent 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 Fuol and safety related equipment shall nct be subject to damage that may adversely offect 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 arc 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 the reactor vessel closure head, the reactor internals, the HACTS, 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 pool is protected against drain down that might occur as a result of a postulated drop accident. 2.11-1
77a . wp (907 5) bh Evaluations have been performed which demonstrate that a postulated head drop, from its specified maximum lif t height, onto the reactor vessel will not result in a significant risk to public safety. , Though the reactor vessel and internals may sustain damage, the i reactor vessel will remain filled and the fuel will remain covered and in a - suberitical configuration. Evaluations of the reactor i internals demonstrate that a drop accident involving the internals j would be less severe than the postulated head drop accident. j 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 seal table (Refer to Figure 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 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 damaged to the extent that there will be significant pool drainage, the reactor vessel will remain filled.
The refueling machine is structurally designed to withstand the af fects of design basis seismic motions. In addition, this machine 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 disposal containers between an intermediate 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 at can not pass over the top of the ICI seal table. This precludes the possibility of load drop accident involving a fuel assembly falling onto tne ICI seal table. Also, during normal refueling operations the travel path is restricted so that it psses 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 in 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 end of the refueling pool, adjacent to the fuel transfer
- system.
The head area cable tray assembly (HACTS), which is used to route power and signal lines away from the reactor vessel closure head, is designed to be handled by the auxiliary hoist on the polar crane. The cable trays are raised vertically by this hoist from l 2.11-2
77a.wp(9075)bh 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 clings that are fastened to specially designed lift fixtures on tha structural frame of the HACTS. All lifting components are designed in accordance with the critoria of NUREG-0612, Control of Heavy Loads at Nuclear Power plants. Prior to movement of the cable traye, 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 a of
.the 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, since the cable tray assembly -is not a rigid structure, an appreciable fraction of the impact energy would be dissipated during plastic deformation of the HACTS 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 and/or buckling. In some instances the pressure housings may also leak. The extent of damage, however, would not be sufficient 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 over fuel storage ~ racks, and movement of the fuel shipping cast apply (defer 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 height for the respective regions. In each case the floors 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 used for transport of the fuel shipping cask is mechanically interlocked to prevent travel over the spent fuel 2.11-3
77c.wp(9075)bh pool. This interlock prevents the possibility of inadvertent movement of heavy loads ovr *he spent fuel storage racks. I New fuel enters the fuel bul ng through a designated unloading area. It is handled and trans,arted to new fuel storage racks by an intermediate capacity hoist. The lift height of the hoist is restricted 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 i accidents. They are designed to limit damage to the storrd fuel and to maintain it in a suberitical configuration. Plant operating l 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 i 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. 2.11.4 RESOLUTION The issue of fucl handling and heavy loads is resolved for System - 80+ by the equipment design and building layout which satisfy applicable _ criteria and provide physical limitations to movement and by administrative limitations in Chapter 9 of CESSAR-DC. 1 l 2.11-4 y -- ---s -- y
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77c.wp(9075)bh 2.12 POTENTIAL FOR DRAINING THE REAQIgR..VEg.gg 2.12.1 ISSUE The issue is the risk of losing primary coolant from he reactor vessel during modes 2 through 6 (startup/ 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 core damage. The amount of damage would, among other factors, depend on the degree to which the core is uncovered I and the decay heat level of the core at the time of uncovery. The draining of the reactor coolant system could 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 reactor vessel are prevention, detection and mitigation. Prevention is the preferred criteria but in some instances (when draining potential cannot be eliminated) detection and mitigation are to be provided. 2.12.2.1 Prevention Criteria
- 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 failure, inadvertent action, or human / operator error to result in the 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 Criteria The design shall have the capability to detect, monitor and locate identifica drainage paths that would occur in a time frame required to prevent a loss of decay heat removal or core uncovery. Appropriate instrumentation, displays and alarms shall be provided. t 2.12-1
77c.wp(9075)bh 1 2.12.2.3 liftication Criteria
- 1. The design shall have the capability to mitigate the loss of I primary coolant from the reactor including shutting-off of a drain path and the ability to provide the source and path for sufficient make-up.
- 2. Technical specifications and procedural guidance shall be prcvided to identify potential make-up water injection sources and paths in the event a drainage path does occur. Recovery actions shall be specified.
7 ) DISCUSSION The primary coolant can potentially drain from the reactor via paths directly from the Reactor Coolant System (RCS) or via paths through systems interfacing the RCS. The plant mode of operation affects the potential for draining the reactor coolant system. This task is focused on shutdown risk. Shutdown risk encompasses operation when the reactor is subcritical or in transition between suberiticality and power operation up to five percent rated thermal power; i.e., for the System 80+ design between Mode 6 and Mode 2 where the modes of operation are defined in CESSAR-DC Chapter 16, Table 1.1.1. The plant modo prescribes and/or allows certain alignments and conditions (pressure, temperature and flow) within and between the RCS and interfacing systems. The causes (initiators) of reactor vessel draining events are grouped into the following categories:
- 1. Components / equipment that fail to operate as intended. This could result from squipment malfunction (e.g., stuck open relief valve), or overpressurization (e.g., seal rupture or-safety valve lift).
- 2. Human (operator) error such as misoperation of valves or pumps leading to a loss of coolant directly from the RCS or through systems connected to the RCS.
There are several factors that affect the likelihood and/or the ultimate consequence of a loss of primary coolant during shutdown. The nt configrration is one such factor. Plant configuration inclum the following:
- 1. The alignu nt, during various modes of shutdown operation, of systems with the RCS and other systems. This affects possible paths for primary coolant flow from the reactor vessel.
- 2. The use of temporary seals in the RCS and interfacing systems during maintenance and refueling activities. The temporary seals include for example, nozzle dams. The failure of such 2.12-2
77c.wp(9075)bh 1 seals can open a path for leakage or close-off a mitigation path.
- 3. The elevations of water within the reactor vessel. Midloop operation reduces margin for recovery after initiation of a loss of primary coolant event relative to a more conservative water inventory.
/. The availability of mitigating systems. Assuming some drain path is opened, allowing primary coolant to et. cape, there must be sufficient make-up available and paths available to transport this make-up to replenish the water in the reactor vessel. The means to close the drainage path is also needed.
It must be assured that maintenance activities do not take required mitigating systems out of service-. Another factor that can affect the potential fo:e and consequence of a lose of primary coolant event during shutdown is the ability to quickly determine that a loss is occurring and the source of the loss. This would include temperature, presnure and water level monitoring. Such monitoring can also provide operators with information that could reduce the likelihood of failure. An evaluation of the Syste.m 80+ plant arrangements and prop (' sed operating configurations will complete this discussion and will be presented in the June 15, 1992 updated submittal of this report. 2.12.4 RE80LUTION The resolution of the reactor vessel draining issue will be presented in the June 15, 1992 updated submittal of this report. l t 2.12-3
77a . wp (907 5) bh 2.13 FLOODING AND SPILL _S_ 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, for example, 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 and internal flooding that may disable components of the shutdown cooling system. 2.13.2 ACCEPTANCE CRITERIA The flood procection design will provide separation of redundant equipment to ensure decay heat removal (DHR) systems availability and capability are not precludad due to flooding and spills. 2.13.3 DISCUSSION The flood protection provided insures a boundary of separation between redundant DHR 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 equipment inoperable. The System 80+ design provides separation and flood barriers to prevent the flood 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 1 and the Division 2 mechanical and electrical equipment.- This wall contains no unsealed penetrations below the 70' elevation level, This wall is along column line '.7 (see CESSAR-DC Figures 1.2-4 and 1,2-5, reproduced here as Figures 2.13-1 and 2.13-2). Additional separation of the divisions is provided by the floor drain systems. The sumps and floor r ins located in the Nuclear Annex and the Reactor Building Sub-6,...ere are di-risionally separated. This design feature prevents the migration of floodvater from one division to the other through the floor drains. The Systems 80+ design utilizes flood doors to provide separation within the same division. In the Reactor Building Sub-Sphere, the flood doors provide quadrant separation, therefore equipment is protected fr'om floods within the same division. Flood doors also provido 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 provide 2.13-1
77a.wp(9075)bh I protection for the Vital Electrical Equipment located in the 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 provides 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 of 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 di. vision are available for decay heat removal. 2.13.4 RESOLUTION The System 80+ flood protection design features are consistent witD acceptence criteria outlined above in Section 2.13.2. These features resolve the issue of flooding and spills during shutdown operations on System 80+ by providing separation of redundant equipment required for decay heat removal. This separation provides the availability of DHR wnen a flood has occurred within the Nuclear Annex or Reactor Building Sub-Sphere. 2.13-2
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77a.wp(9075)bh i 1.O PAQpl%ILISTIC RISK'%BRIHMENTS
- ' 3 .1 ' INTRODUCTION Emphasis has previously.been.on the safety of power plants during power operation. This:was due to the fact that the plant is in this configuration mont-of the time and the core power, decay heat-rate, and fission product inventory are highest at this time. The
' System 80+ PRA (Reference.7)_ documnnts the risk associated with the operation of this. ALER 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 rates _ and fission product inventory. The plant configuration is not as well defined as in full power-operation because of-the: maintenance and testing tnat is being_ performed.
Thisireport docyments the risk _of operation of this plant in Modes Modes 1 and 2 art covered in Reference 7. 3 through 6.
-The awareness of the risks of plant operation during refueling and ma2ntenance outage developed slowly. During refueling the plants have low decay heat and also have a large primary coolant inventory with the refueling cavity flooded. The emphasis in the SARs was on
;-operational events _ and operator training - also emphasis power operation 1 modes .- Although no core damage has occurred during reactor 1 outages, a few events have occurred which were precursors to more-severe accidents.. One of the first events to increase the
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- 'l awareness of riska 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
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difficult _to-handle if the-refueling had actually started. The
-seal failure had not been-. considered in the. safety analysis.
%nother event which increasedLthe awareness of the risks during.
outages _ was the- Vogtle ~ l' 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 waL out for maintenance. Une of the diesels wa's also out for maintenance and - the~ second failed to start. This combination of maintenance activities and failures let
-to a station -blackout and loss of Residual Heat Removal ( RHR) ~.
Under normal conditions, with the vessel filled . or with- the refueling cavity flooded, .the operator would have many hours to ress. ore RHR. In this incident, the plant was in mid-loop operation and the primary inventory ' was great-ly reduced. When RHR was restored in 41 minutes, the primary coolant temperatura had risen
'46.deg F to 13 6 - deg ' F. .This incident demonstrates the: unusual plant configurations that-can exist during an oatage and the risks associated with maintenance activities and mid-loop operation.
3-1
77c.wp(9075)bh During 1991, there was 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 -had two incidents during the same outage. There were at least seven events where loss of RHR occurred in 1991. All these events increased the awareness of the risks during outages. The NRC requested additional information (Reference 8) from the ALWR participants pertaining to shutdown risk. The EPRI ALWR Utility Requirements Document thus was modified to include a risk assessment for shutdown modes (Reference 9). This analysis satisfies that requirement. This analysis uses a simplified event tree approach to estimate the core damage frequency (level 1 PRA) with only a simplified evaluation of the release frequencies and magnitudes. 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, genera' categories of initiating events were then defined. These initiating events wera small LOCAs, Loss of RHR, Loss of AC, and Boro 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 usi..g 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 Shutdown PRA was performed with the inc.ight obtained from the previous PRAs. In 1981, NSAC-84 looked at the risk associated at Zion during outages. 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 RHR accounted for 56% of the total CDF. Loss of RHR 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 CDP was due to loss of RHR 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 ADVG, and 3-2
77a.wp(9075)bh the response to. Generic Letter 88-17. The insights gained from reviewing these PRAs helped in analyzing the System 80+ plant. Section 3.2 contains a discussion of the methodology. The initiating event evaluation is evaluated in Section 3.3. The accident sequence determination is given in Section 3.4. With the data in Section 3.5 and the system analysis in Section 3.6, the accident sequences are quantified in Section 3.7. The consequence analysis is given in Section 3.8. Sections 3.2 and 3.3 will be provided in the June 15, 1992 updated submittal of this report. The remaining sections will be provided in the final July 31, 1992 updated submittal of this report. 3-3
770.wp(9075)bh 4.0 APPLICABILITY OF CHAPTER 15 ANALYBES 4.0.1- FORMAT AND CONTENT The purpose of this section is to present evaluations which confirm the applicability of the analyses presented in Chapter 15 of CESSAR-DC to shutdown Modes (Mode :c subcritical or Modes 3 through
- 6) for the System 80+ Standard Design.
The approach used in the docunentation of the events in Chapter 15 of CESSAR-DC is to present the results for the events with the most adverse consequences. As a result, reference is most frequently to events postulated to occur in Mode 1 or Mode 2 critical. Only in certain cases .which intrinsically involve shutdown Modes (e.g. startup of an. inactive reactor coolant pump) are shutdown Modes stressed. The purpose of this section is to ensure that all possible Modes have been treated in the documentation for Chapter 15 events. The following sections have been organized to parallel the sections of Chapter 15 of CESSAR-DC. For example, Section 4.1.1 treats the same group of initiating events that are documented in Section 15.1.1 of CESSAR-DC. 4.1 INCREASE IN HEAT REMOVAL BY THE SECONDARX_8J8 TEM 4.
1.0 INTRODUCTION
The purpose of this section is to present evaluations which confirm that ali increase in heat removal by the secondary system events postulated to be initiated in a shutdown Mode have acceptable consequences for the System 80+ Standard Design. Section 15.1 of CESSAR-DC documents results which show that all increase in heat -emoval by the secondary system events have acceptable consequmces if they are explicitly postulated to occur in Mode 1 or Mode 2 critical for the System 80+ Standard Design. This is demonstrated for steam system piping failures inside and outside containment (Section 15.1.5) by analyses for both Mode 1 and Mode 2 critical. The choice of initial conditions for the other analyses of Section 15.1 to minimize the transient DNBR for any operating condition ensures that the results presented bound those for all of Mode 1 and Mode 2 critical. 4.1.1 DECR2ABE IN FEEDWATER TEMPERATURE Evaluation of decrease in feedwater temperature events initiated in a shutdown Mode is in progress. It is expected that the conse-quences of these events will be bounded by the most adverse conse-quences presented for the events in Section 15.1.4 of CESSAR-DC. 4-1
--- . . . . - - - - . _ . ~ . .- .. -. -
~
- . 77a.wp'(9075)bh l The-results of this evaluation are to be submitted to the NRC in the_ June 15,-1992 updated submittal of this report, w
4.1.2 INCREASE IN FEEDWATER FLOW Evaluation of increat.e in feedwater flow events initiated in a shutdown Mode is in progress. It is expected that the consequences of these events will be bounded by the most adverse consequences presented for-the events in Section 15.1.4 of CESSAR-DC. The results of this evaluation will be presented in the June 15, 1992, updated submittal of this report. 4.1.3 INCREASED MAIN STEAM FLOW As noted in Section 15.1.3 of CESSAR-DC, the steam flow due to an increased main steam flow event is the same as-(or less than) that due to an inadvertent opening of a steam generator relief or safety valve event. Further, there are n* other dif ferences between these events which af fect their consequences. Therefore, the conclusions of . Section '4.1.4 'of this report apply also to . an increased main steam flow event. 4.1.4 INADVERTENT. OPENING OF-A STEAM GENERATOR RELIEF OR SAFETY VALVE Evaluation of inadvertent opening of a steam generator relief or safety valve events initiated in a shutdown Mode is in progress. It is ' expected that the consequences of these events will be bounded by the.most adverse consequences' presented for the events
-in Section 15.1.4 of CESSAR-DC. The results of this evalu
- ion will be presented an the June 15, 1992 updated submittal of this report.
4.1.5 -STEAM SYSTEM-PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT Evaluation of steam system piping failures' inside and outside containment,' initiated in a shutdown Mode, is in progress. It is expected that the consequences of these events will be bounded by the most adverse consequences presented for the events in Section 15.1. 5 CESSAR-DC. The results of this evaluation will be presented in the June 15, 1992 updated submittal of this report, c 4-2 t . f
- _ . _ _._ . . _ ~
770.wp(9075)bh l {' 4.2 DECREhBE IN HEAT REMOVAL DY SECONDARY SYSTEM 4.
2.0 INTRODUCTION
The purpose of thiis section is to present evaluations which confirm that all decrease in heat removal by the secondary system events postulated to be initiated in a shutdown Mode have acceptable consequences for the System 80+ Standard Design. Section 15.2 of CESSAR-DC documents results which show that all decrease in heat removal by the secondary system events have acceptable consequences if they are explicitly postulated to occur in Mode 1 or, in genezil, in Mode 2 critical for the System 80+ Standard Design. If there is a question as to whether au event has already been considered in Mode 2 critical, however, this Mode is also included in the evaluation. The focus of the evaluations presented in this section is on ensuring that the peak primary and secondary pressures are less than 110% of their design pressures, that the pressure-temperature limits for brittle fracture of tne reactor coolant system (RCS) are not violated, and that fuel integrity is maintained for the Svents considered. Fuel performance, as measured by the departure from nucleate boiling ratio (DNBR), is used for verification of fuel integrity. 4.2.1 LOSS OF EXTERNAL LOAD Since the turbine is not on line, a loss of external load is not possible in a shutdown Mode. 4.2.2 TURBINE TRIP Since the turbine is not on line, a turbine trip is not possible in a shutdown Mode. 4.2.3 LOSS Or CCNDENBER VACUUM (LOCV) 1 Evaluation of loss of condenser vacuum (LOCV) events initiated in - a shutdown Mode is in progress. It is expected that the conse-quences of these events will be acceptable. The results of this evaluation will be presented in the June 15, 1992 updated submittal of this report. 4.2.4 MAIN STEAM ISOL.. TION VALVE CLOSURE The comparison between nain steam isolation valve (MSIV) closure and LOCV events made in Section 15. 2.4 of CESSAR-DC applies for shutdown Modes, also. The evaluation of the LOCV event presented in Section 4.2.3 assumes a faster reduction in steam flow rate than would result from MSIV closure. The consequences cf the MSIV l 3
H L 7 7a ~.wp ( 9 07 5 ) bh - a j l I closure ' eventi are,- therefore, no more adverse in shutdown Mcies than those?for the:LOCV presented in Section 4.2.3. 4.2.5 STEAM PRESSURE REGULATOR FAILURE This event does.not apply to the SYSTEM 80+-Standard Design and is,
.therefore, not evaluated here.
4 4.2.6 -LOSS OF:NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES The results-of tha loss of non-emergency AC power to the station auxiliaries' (LOAC) event are the same as those for the loss of reactor coolant flow event presented in Section 4.3.1. 4.2'.7. LOSS OF NORMAL FEEDWATER FLOW A postulated loss of normal feedwater flow (startup feedwater flow)_ during:a shutdown Mode would be less adverse than the.LOCV event. The~ analysis assumptions for the LOCV event result in termination : of steam flow, as well-as termination of startup feedwater flow, causing a-~more L-evere decrease in heat removal by:the secondary D i system.: .The consequences of the loss of normal feedwater flow are, Etherefore, bounded - by the consequences of the - LOCV event for shutdown. Modes.
~
_4.2.8 -FEEDWATER SYSTEM PIPE BREAKS Depending on the break size and location'and the response of the feedwater_-system, the effect of a postulated feedwater system pipe break can vary from a heatup to-a cooldown'of.the-RCS, Based on the same arguments given in Section .15. 2. 8.1 - of CESSAR-DC, the heatup event -is considered in this section.- The caoldown potential.
'would . be worse for a steam line break, which -is- discussed in Section 4.1'.5; A- heatup event- is mitigated - by the pressurizer
. safety valves _or'theLshutdown cooling (SCS) relief _ valves when.RCS-
-_ temperatures a r e -- above or- below, respectively, the ~LTOP ,
L enabls./ disable temperatures, the main steam safety valves, and the emergency _feedwater (EFW) system. o ~_A feedwater system pipe break postulated to occur-in a shutdown l Mode would result in less Lsevere consequences than the event
-documented:in Section 15.2.8 of-CESSAR-DC dueLto the lower initial reactor power-level. DNB is not of concern due to the low initial-core J power levels,_ in addition _ to' the _ pressurization _ following
~
' event initiation.- Further,.the'-EFW system is= capable of removing-decay. heat _ event with -only one EFW pump in operation. The dominant =
w factor which determines peak primary-and secondary pressures is the magnitud cC-the energy mismatch between.the primary and secondary systems. This mismatch is very.much less for events postulated to occur in shutdown Modes than for the event of Section 15.2.8 of l: , 4-4 I I' . a -_ _. .
7 7 a .wp (907 5) bh CESSAR-DC. There is, therefore, no approach to 110% of steam generator design pressure, since the maximum pressure will be much less than that for the full power event. For the same reason there is no approach to the criterion of 110% of RCS design pressure when RCS temperatures are above the LTOP enable / disable temperatures. For RCS temperatures below the LTOP enable / disable temperatures the design of the SCS relief valves ensures that the pressure-temperature limits for brittle fracture of the RCS are not violated. 4.3 DECREASE IN REACTOR COOLANT FLOW RATE 4.
3.0 INTRODUCTION
The purpose of this section is to present evaluations which confirm that all decrease in reactor coolant flow rate events postulated to be initiated in a shutdown Mode have acceptable consequences for the System 80+ Standard Design. Section 15.3 of CESSAR-DC documents results which show that all decrease in reactor coolant flow rate events have acceptable consequences if they are explicitly postulated to occur in Mode 1 or Mode 2 critical for the System 80+ Standard Design. This is demonstrated for total loss of reactor coolant flow (Section 15.3.1) by explicit analyses. The choice of initial conditions for the other analyses of Section 15.3 to minimize the transient DNBR for any operating condition ensures that the resalts presented bound those for all of Mode 1 and Mode 2 critical. Decrease in reactor coolant flow rate in Modes 4 through 6 when the shutdown cooling system (SCS) is being used for decay heat removal is addressed in Section 2.4 as integral to the evaluation of loss of decay heat removal capability. Therefore this section addresses e vents postulate d to be initiated in Mode 2 subcritical, Mode 3 or Modes 4 and 5 when the SDS is not being used. 4.3.1 TOTAL LOSS OF REACTOR COOLANT FLOW Evaluation of the factors affecting the consequences of the total loss of reactor coolant flow event shows that if this event is r postulated to be initiated in shutdown Modes, the results are less adverse than those of the CESSAR-DC Chapter 15.3 full power event. Loss of offsite power is the postulated initiating event for the total loss of reactor coolant flow event in Modes 1 or 2. All i systems available to mitigate the Mode 1 transient are available in Mode 2 subcritical. The initial conditions for Mode 2 subcritical include four pumps operating, temperature and pressure identical to l Mode 1, a low power level, and total energy stored in the reactor core much less than at full pouer. Therefore, a very large margin to DNB exists at the initiation of the event and the heat to be 4-5
77 a . wp ( 9 07 5) bh removed during the event is much less than for an event initiateo at full power. The minimun DNBR for this event is substantially higher for Mode 2 than that for the full power case. In addition, the initiatia.g event could be postulated to be loss of power to any operating RCPs in Mode 3 or in Modes 4 or 5 when the SCS is not being used. Natural circulation is, hoaever, suf ficient for the removal of decay heat in these Modes. Thus no approach to DNB would occur. The consequences of the Chapter 15.3 full power event are therefore more adverse than an event in these Modes. The full power four pump loss of flow transient produces RCS and steam generator pressures which are less than 110% of their design values. Transients postulate to be initiated in Modes 2, 3 or 4 when RCS temperatures are above the LTOP enable / disable temperatures, and for which the rate of heat production is orders of magnitude below full power values, would therefore yield even greater margins to the design pressure values. For transients postulated to be initiated in Modes 4 or 5 with RCS temperatures below the LTOP enable / disable temperatures, the design of the SDC relief valves ensures that the pressure-temperature limits for brittle fracture are not violated. 4.3.2 FLOW CONTROLLER MALFUNCTION CAUSING FLOW CUASTDOWN This event is categorized as a Boiling Water Reactor event in SRP 15.3.2 and is, ti.erefore, not evaluated here. 4.3.3 SINGLE REACTOR COOLADT PUMP ROTOR SEIZURE WITH LOSE OF OFFSITE POWER The major parameter of concern for the single reactor coolant pump rotor seizure with loss of offsite power event documented in Section 15.3.3 of CESSAR-DC is the minimum hot channel DNBR. This is minimized by higher power conditions. Lower power Modes woulf _herefore, not produce any conditions which are more adverse than those presented in CESSAR-DC. The second parameter of concern is the peak RCS pressure attained. The full power single reactor coolant pump rotor seizure with loss of offsite power event produces RCS pressures which are less than 110% of their design values. Transients postulated to be initiated in Modes . 2, 3, or 4 when RCS temperatures are above the LTOP enable / disable temperatures, and for which the rate of heat production is orders of magnitude below full power values, would therefore yield even greater margins to the design pressure values. For transients postulated to be initiated in Modes 4 or 5 with RCS temperatures below the LTOP enable / disable temperatures the design of the SCS relief valves ensures that the pressure-temperature limits for brittle fracture of the RCS are not violated. 4-6
77a.wp(9075)bh In addition, at these low power levels a concurrent turbine trip which results in a loss of offsite power is not an issue since the turbine is not in operation below 5% power. 4.3.4 REACTOR COOLANT PUMP SKAFT BREAK WITH LOSS OF OFFSITE POWER Since a postulated reactor coolant pump shaft break (SB) transient results in a less rapid flow coastdown than a rotor r3izure (RS) event, the results of the SB event are bounded by those of the evaluation of Section 4.3.3. 4.4 REACTIVITY AND POWER DISTItIBUTION ANOMALIES 4.
4.0 INTRODUCTION
The purpose of this section is to present evaluations which confirm - that all reactivity and power distribution anomaly events postula*' ed to be initiated in a shutdown Mode have acceptable consequen e . fer the System 80+ Standard Design. l
/
4.4.1 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FRC>a SCPCRITICAL OR LOW POWER CONDITIONS Analyses are in progress to demonstrate that the consequences of an uncontrolled CEA withdrawal initiated from a shutdown Mode is acceptable. It is expected that the consequences of these analyses will be acceptable. The results of these analyses will be presented in the June 15, 1992 updated submittal of this report. 4.4.2 UNCONTROLLED CONTROL ELEMENT ASSP.MBLY WITHDRAWAL AT POWER This event is not an issue for shutdcwn Modes since its intent is to examine high power operation only. 4.4.3 SINGLE CONTROL ELEMENT ASSEMBLY DROP A postulated single control element assembly drop at power is analyzed for approach to the DNBR limit in CESSAR-DC, Section 15.4.3. For Mode 2 subcritical through the other subcritical - Modes, a dropped rod only adds more negative reactivity to an already suberitical core and is, therefore, much less adverse than the full power event documented in CESSAR-DC. 4.4.4 STARTUP OF AN INACTIVE REACTOR COOLANT PLdP Chapter 15.4.4 of CESSAR-DC presents analysis of startup of an inactive reactor coolant pump events which show that events postulated to be initiated in Modes 3 through 6 have acceptable 4-7
I 77c.wp(9075)bh consequences for the System 80+ Standard Design. (Operation with less than i RCPs is not permitted in Modes 1 or 2.) 4.4.5 FLOW CONTROLLER MALFUNCTION CAUSING AN INCREASE IN BWR CORE FLOW RATE This event is categorized as a Boiling Water Reactor event in SRP 15.4.5 and is, therefore, not evaluated here. 4.4.6 INADVERTENT DEBORATION Analysci of inadvertent deboration events in shutdown Modes have been presented in Chapter 15 of CESSAR-DC. An additional evalua-tion which considers rapid deboration is presented in Section 2.6 of this report. 4.4.7 INADVERTENT LOADING OF A FUEL ASSEMBLY INTO THE IMPROPER POSITION This event has been evaluated in Section 15.4.7 of CESSAR-DC and is not mode dependent. 4.4.8 CONTROL ELEMENT ASSEMBLY (CEA) EJECTION The core remains subcritical for CEA ejection events postulated to be initiated in a shutdown Mode. The technical specification on K, .1 (see Section 2.2) requires that the highest worth CEA be excluded from the subcriticality calculation for Modes 2 through 5. Thus, even if the highest worth CEA were to be assumed to be ejected from the core during shutdown, the core would remain subcritical. The full power event of Section 15.4.8 of CESSAR-DC is, therefore, the limiting CEA ejection event. 4.5 INCREASE IN RCS INVENTORY 4.
5.0 INTRODUCTION
The purpose of this section is to present evaluations which demonstrate that all increase in RCS inventory events postulated to be initiated in a shutdown Mode have acceptable consequences for the System 80+ Standard Design. Section 15.5 of CESSAR-DC documents results which show that all increase in RCS inventory events have acceptable consequences if they are explicitly postulated to occur in Mode 1 or, in general, in Mode 2 critical for the System 80+ Standard Design. If there is a question as to whether an event has already been considered in Mode 2 critical, however, this Mode is also included in the evaluation. 4-8 ; l
77a.wp(9075)bh K L The focus of the evaluations presented in this section is on ensurir.3 that the peak primary pressure is less than 110% of design
- pressure and that. the pressure-temperature limits for brittle fractura of the RCS are not violated. Fuel performance, as measured by the departure from nucleate boiling ratio (DNBR), is used for verification of fuel integrity. Peak secondary pressure is also evaluated, as necessary, to ensure it remains less than 110% of its design pressure.
. 4.5.1 INADVERTENT OPERATION OF THE ECCS EE
@' 1 The evaluation presented in Section 15.5.1 of CESSAR-DC establishes
*i t.L % postulated inadvertent operation of the ECCS would have
, or;~c.able consequences for any Mode for the System 80+ Standard i 60 an, pn d 2.5.2 CVC8 MALFUNCTION-PRE 88URIZER LEVEL CONTROL SYSTEM MALFUNCTION WITH LOSS OF OFFSITE POWER Peak RCS pressure due to a postulated malfunctior./ actuation of a charging pump in Modes 2 through 4 (before LTOP is active), is well within 110% of design pressure. Further, the pressure-temperature limits for brittle fracture of the RCS are not challenged for a postulated malfunction / actuation of a charging pump in Modes 4 thr3 ugh 6 (when LTOP is active or when the reactor vessel head is off).
In Modes 2 cubcritical through 4, before shutdown cooling and LTOP are placed in service, the peak RCS pressure is limited by the pressurizer safety valves. Since only decay heat ex3sts under these conditions, the peak pressures during the CVCS malfunction event would - be substantially less severe than the Mode 1 case described in CESSAR-DC. It should be noted that the centrifugal' y charging pumps are protected from runout at low RCS pressures by design. In Modes 4, 5, and 6 when-LTOP is active (or in Mode 6 with the reactor vessel head of f and LTOP not act ive), the CVCS malfunction event would be less limiting than the inadvertent SIS actuation described in Section 4.5.1 above, due to the much lower flow from one charging pump versus the four SI pumps. 4.6 DigREASE IN REACTOR COOLANT SYSTEM LNVENTORY 4.
6.0 INTRODUCTION
The purpose of this section is to present evaluations which demonstrate that all decrease in RCS inventory events postulated to be initiated in a shutdown Mcde have acceptable consequences for the System 80+ Standard Design. 4-9
-77a.wp(9075)bh Section 15.6 of CESSAR-3C documents results which show that all decrease in RCS inventory events have acceptable consequences if they are explicitly postulated-to occur in Mode 1 or, in general, in Mode 2 critical for the System 80+ Ctandard Design. If there is a question as to whether an event has already been considered in Mode 2 critical, however, this Mode is also included in the evaluation.
Fuel performance, as measured by the departure from nucleate boiling ratio (DNBR), is used for verification of fuel integrity. 4.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY / RELIEF VALVE The LOCA evaluations presented in Section 4.6.5 of this report show that the inadvertent opening of a pressure safety / relief valve is at a non-limiting location for LOCAs postulated to occur in shutdown Modes. The inadvertent opening of a pressurizer safety valve as described in the Standard Review Plan 15.6.1 is, therefore, a non-limiting event in the Safety Injection System evaluations. 4.6.2 DOUBLE-ENDED BREAK OF A LETDOWN LINE OUTEIDE CONTAINMENT Evaluation of double-ended break of a letdown line outside containment event initiated in a shutdown Mode is in progress. It is expected that the consequences of this event will be bounded by the most adverse consequences presented for the events in Section 15.6.2 of CESSAR-DC. The results of this evaluation will be presented in the June 15, 1992 updated submittal of this report. 4.6.3 STEAM GENERATOR TUBE RUPTURE Evaluation of steam generator tube rupture events initiated in a shutdown Mode is in progress. It is expected that the consequences of these events will be bounded by the most adverse consequences presented for the events in Section 15.6.3 of CESSAR-DC. The results of this evaluation will be presented in the June 15, 1992 updated submittal of this report. 4.6.4 RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR) The radiological consequences of main steam line failure outside l containment (BWR) do not apply to the System 80+ Standard Design and are, therefore, not evaluated here. l 4-10
' 77a .wp ( 9075) bh 4.6.5 LOSS-OF-COOLANT ACCIDENT Consequences of a .LOCA initiated from Modes 2 through 4 are bounded by the 1 consequences reported for a LOCA from Mode 1 since the containment : spray and annulus. ventilation systems which are available in Mode 1 are also available in Modes 2 through 4.
Evaluation of the conseque.,:es of.a LOCA initiated from Modes 5 and 6 is in progress and will be presented in the June 15, 1992 updated submittal of this report. 4.7 RADIOACTIVE MATJRIAL RELEASE FROM A SUBSYSTEM OR COMPONENT 4.
7.0 INTRODUCTION
The purpose of this section is to present evaluations which confirm that all radioactive material release from a subsystem or component events postulated to:be initiated in a shutdown Mode have accept- , able consequences for the System 80+ Standard Design.
. 4 . 7 .1 - RADIOACTIVE GAS WASTE SYSTEM FAILURE This section of'the Standard Review Plan has been deleted (Refer-ence 26 of Section 15.0) 4.7.2 RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE This section of the Standard Review Plan has been deleted (Refer-ence 26.of Section 15.0) 4.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID CONTAINING TANK FAILURES This event has been evaluated in Section 15.7.3 of CESSAR-DC and is not Mode dependent.
4.7.4 FUEL HANDLIA ACCIDENT This event has been evaluat.d in'Section 15.7.4 of CESSAR-DC and is
-not Mode dependent.
- 4. 7. 5 - ~ SPENT FUEL CASK DROP ACCIDENTS This event has been evaluated in Section 15.7.4 of. CESSAR-DC and is not Mode dependent.
4-11
77a.wp(9075)bh 5.0 APPLICMILITY OF CHAPTER 6 LOCUNALYSES TO SHUTDOWN MODES _ 5.1 ISSUE A majority of the analyses of the Loss-of-Coolant Accident and all criteria associated with the accident have focused on scenarios starting from 102% of rated thermal power as prescribed in 10CFR50, Appendix K. In this section, scenarios initiated from other than full power (or 102% in the case of a LOCA) will be addressed to demonstrate that the analyses performed in CESSAR-DC are the bounding cases for all modes of operation. This section will identify the scenarios anticipated for non-Mode 1 operation, and where required the acceptance criteria for the dif ferent scenarios. 5.2 ACCEPTANCE CRITERIA The acceptance criteria fer lower operating mode LOCAs is established to be the same as those for higher operating mode LOCAs (Reference 19): ,
- 1. Peak clad temperature < 2200*F,
- 2. < 17% peak local clad oxidation,
- 3. < 1% core wide oxidation,
- 4. Maintain coolable geometry, and
- 5. Maintain long-term cooling.
5.3 DISCUSSION The types of LOCAs and corresponding break sizes in lower modes are as follows: because of the lower pressures in lower modes, only traditionally "small" break LOCAs are considered. The following lower mode scenarios are considered: most adverse misoperation of valves, the likelihood of cross train maintenance errors, and the possibility of rupture of a tributary pipe. The limiting size and location for a "small" break LOCA for Modes 3 and 4 is a direct vessel injection (DVI) line discharge leg (0.4 square feet) . A DVI line break at this location would minimize injection flow into the RCS. Modes 5 and 6 consider a loss of decay heat removal (DHR) resulting from a break in the bottom of the shutdown cooling system l suction or a lower head instrument line (.003 square feet). A discussion of the equipment which would be available to mitigate the consequences of a LOCA in each mode of operation, the ef fectiveness of that equipment under the conditions for that mode, and the applicability of analyses to that mode are as follows: . Mode 1: In Mode 1, RCS temperature is >350*F and power is >5%. The CESSAR-DC analysis for LOCAs in this mode concludes that all criteria are met. 5-1
77a.wp(9075)bh. Mode 2: In Mode 2, RCS temperature is >350*F and power is <5%. -The equipment available in Mode 1 is also available in Mode 2. Mode 2 LOCA results are bounded by Mode 1 because the decay heat and stored ene gy, which are of concern to meeting LOCA criteria, are proportioral .to the lower power level in Mode 2. Mode 3: In Mode _3, RCS temperature is >350'F and k.n is <.99. For the case where RCS pressure is >900 psia, the equipment available in Mode 1 (SIT, 4 SI pumps) is available. Because the stored energy and decay heat is lower, this parameter space in Mode 3 is bounded by Mode 1 for LOCA. For the case where pressure is <900 psia, 4 SI pumps are available. Although at a higher temperature than Mode 4, the parameter space in Mode 3 is bounded by the Mode 4 analysis because Mode 4 does not credit automatic SI actuation. Furthermore, despite the fact that SITS would not be available when the pressure is <900 psia, this Mode 3 space is bounded by the Mode _1 analysis because there is less RCS pressure to drive the LOCA leak flow. Mode 4: RCS temperature is <350* F. Four subspaces are considered:
- 1. Conditions:
pressure >900 psia and temperature >317'F. SITS and 2 HP pumps on automatic.
Conclusions:
Conclusions will be presented in the June 15, 1992 updated submittal of this report.
- 2. Conditions:
Pressure < 900 psia and temperature >317'F: 2 HP pumpt on automatic.
Conclusions:
Results for these conditions are bounded by the results for Subspace 4 below because 2 HP pumps are available to accommodate the higher temperature.
- 3. Conditions:
Pressure > 900 psia and temperature < 317'F.
Conclusions:
Results for these conditions are bounded by the results for Subspace 4 below because SITS are available to accommodate the higher pressure.
- 4. Conditions:-
Pressure < 900 psia and temperature < 317*F.
Conclusions:
(snclusions will be presented in the June 15, 1992 updated submittal of_this report. Mode 5: RCS temperature $210*F. An analysis was performed which assumed the maximum Technical Specifications cooldown rate of 100*F/hr. Additional information will be presented in the June 15, 1992 updated submittal of this report. 5-2
l 770.wp (9 07 5) bh Mode 6: This information will be presented in the June 15, 1992 updated submittal of this report. Subsection 5.3.1 describes the primary system boundary conditions and event scenario for the postulated LOCA used in this study. Subsection 5.3.1 also compares and contrasts this lower operating mode event scenario to the conservative design basis licensing LOCA event normally associated with ECCS evaluation analyses. Subsections 5.3.2 and 5.3.3 describe the System 80+ plant parameters and conditions for the analysis and the computer modes and analysis methods used in the LOCA calculations. Subsection 5.3.4 describes the results of an analysis of a limiting LOCA during Mode 4. These calculations show that hot fuel rod conditions remain in compliance with ECCS Acceptance Criteria during an - 'Imed 10 minute time delay for operator action without SI pump ..ilability. Furthermore, the analysis for the postulated LOCA shows that availability of 1 SI pump at the 10 minute mark maintains the hot rod cladding temperatures in compliance with the ECCS Acceptance Criteria. 5.
3.1 DESCRIPTION
OF LOCA SCENARIO Following powered operation of the NSSS, cooldown proceeds at the Technical Specifications maximum rate of 100*F/ hour (the maximum cooldown rate shown in Reference 18) . Therefore, a primary coolant temperature of 317'F could be reached as early as 2.4 hours after shutdown. The equipment available to mitigate LOCAs from lower modes is basically safety injection tanks (SIT) and safety injection (SI) pumps. SITS are available for pressures >900 psia. SI pumps are
~
automatically actuated for temperatures >317*F. If a postulated LOCA transient were to occur at pressures and temperatures slightly below these conditions (i.e., pressure <900 psia, temperature <317*F, and time >2.4 hours), the LOCA would be significantly less dynamic than a design-basis LOCA transient from full power operating conditions (i.e., 2250 psia, ~600*F and full fission power-and associated decay heat) . Factors which would significantly mitigate the potential and consequences of a lower operating mode LOCA compared to a full power LOCA are (1) lower initial primary system pressure which would limit the internal forces on the piping and the duration of blowdown, (2) lower coolant flow rate out of a postulated break and slower depressurization rate which would reduce inventory loss and flashing rate, and (3) lower decay heat levels which would lessen the core boiloff rate. 5-3 l
- . . - - - - .~. . - - - _ - - - _ - -.
77a Wp(9075)bh Based on ~these factors, a _ postulated lower operating mode LOCA followed by, _ if necessary, timely operator action to initiate safety injection flow would be expected to be much less severe.than a-- LOCA from full power conditions. The most severe lower operating mode ~LOCA scenario would occur ~for (1) the largest potential pipe break, (2) after-the:most rapid possible cooldown from full power, (3) after reducing temperature slightly below which no HPSI pumps are - required to be on _ automatic, (4) after reducing pressure slightly below which SITS are not available, and (S) the longest expected time-for mitigating operator action. The largest and most harmful potential pipe break, based on consideration of mechanistic breaks accounting for the system energy at the postulated cooling conditions,- would be a-significant leak in one of the direct vessel injection (DVI) lines of the reactor coolant system corresponding to the flow area of the'DVI'line. This DVI break size of 0.4 ft was chosen which envelopes the size and limiting location of all traditional-small break LOCAs. Decay heat levels based on a time period of 2.4 hours after shutdown is assumed. No operator action for 10 minutes is also assumed. Forced circulation through the core during_ lower mode operation would tend to prolong the time of adequate core cooling during a postulated LOCA; therefore, for conservatism, the RCPs are tripped before the start of the event. An aggressive- cooldown. of the secondary side of the steam generators .would considerably benefit the RCS heat removal process for 'a postulated LOCA; ~ - theref ore , for conservatism, the steam generator secondary sides are isolated for the event. 5.3.2 SELECTION OF REFERENCE PLANT PARAMETEPc AND CONDITIONS FOR MODE 4 ANALYSIS A . limiting set of values is . selected from among the plant parameters- and conditions. Additional information will be presented.in the June 15, 1992 updated submittal of this report. 5.3.3- ANALYSIS COMPUTER CODES l This task required the selection of realistic inputs such as decay heat to provide as much realism in representation of the system transient response durjng a LOCA as possible. An adaptation of the Realistic Evaluation-Model'(REM) for s n ll break LOCA was selected for, - this reason. - This model is a second-generation small break LOCA evaluation model intended to replace the 1974'EM currently i used for ECCS licensing calculations. . Topical reports describing l- the REM were submitted to- the NRC starting in 1988 and are
. currently under review (see References 12 through 16).
For design basis LOCA calculations from plant initial primary pressures of-2250 psia, the largest break size apalyzed using the CEFLASH-4AS-code has historically been a 0.5 ft break. For the S-4
l 77c.wp(9075)bh I 0.4 square foot DVI line lower mode LOCA analysis, the REM version of the CEFLASH-4AS code was used with realistic decay heat. The REM version of the PARCH code was used for calculating hot rod haatup (3eferences 12 through 16). The PARCH base deck included a realistic decay heat. 5.3.4 LOCA ANALYSIS FOR Mode 4 The LOCA analysis examines the hot rod heatup response during a LOCA with the ECCS Acceptance Criteria of peak cladding temperature and peak local cladding oxidation. The objective of this analysis is to determine if the calculated response of the hot rod remains in compliance with the ECCS Acceptance Criteria during the 10 minute time frame before operator actions may be credited. A bounding analysis, starting at 2.4 hours after shutdown, allowed the LOCA to proceed without safety injection. Primary coolant inventory is assumed to be lost through an opening in the direct vessel injection line at the vessel penetration to che upper annulua. When the two-phase level falls to the top x 4 'te active core, there is a reduction in core cooling and the f % i rods begin a heatup driven by the core decay heat power and at higher temperatures by the heat added from cladding oxidation. As cladding temperatures increase, fuel rod swelling and rupture may occur. For this analysis, the PARCH code is used for calculating hot rod heatup. The hot rod in the core is initialized by the PARCH code with 900 psia and 317'F primary coolant conditions. A limiting axial power shape is assumed for the LOCA. At 8500 seconds (-2.4 hrs), the coolant inventory and, consequently, core cooling is reduced. Figure 5.2-1 shows the resulting hot rod heatup calculation. A realistic model for decay heat is used, which is the 1979 ANS Standard 5.1 plus two sigma uncertainty plus actinide decay. 5.3.5.1* Besults of LOCA Case with No ECCS Delivery for More than 10 Minutes 5.3.5.2* Influence of Restorina 1 HPSI PUMP Not at the Broken DVI Line 5.4* RESOLUTION
- These sections will be provided in the June 15, 1992 updated submittal of this report.
5-5 l
77 a .wp (9075) bh 6.0 APPLICABILITY OF CHAPTER 6 CONTAINMENT ANALYSES
6.1 INTRODUCTION
CESSAR-DC Section 6.2.1.1 discusses containment functional design. A series of loss of coolant ace'idents (LOCAc) and main steam line breaks (MSBLs) were analyzed to letermine the resulting containment pressure and temperature for comparison with the containment design pressure and the equipment environmental qualification envelope. The highest containment pressures and temperatures occur when the
-NSSS stored energy is maximized and the containment heat removal capability is minimized. . Maximum RCS stored energy occurs when the plant - is at 102 percent power. Maximum' steam generator stored energy occurs at 0 percent power (hot standby). Consistent with the-Standard Review Plan (SRP), a series of LOCAs were analyzed at 102 percent power and a series of MSLBs were analyzed at powers from 102 percent to O percent. The o percent MSLB with the failure of a containment spray train is the design basis event (DBE) for System'80+. these cases are presented in CESSAR-DC.
In going from Mode 2 to 6, r'ored energy is removed from the NSSS. If the safeguards features available in Modes 1 and 2 were available through Modo 6, there would be no question that the DBE identified in CESSAR-DC was limiting. Cowever, some safeguards equipment is removed from service at lower modes. Table 6-1 lists the availability of safeguards equipment credited in the containment analyses as a function of operating mode. Since the Main _ Steam Isolation Signal (MSIS) and Containment Spray Actuation Signal (CSAS) may-be removed from service in Modes 5 and 6, an evaluation of LOCAs and MSLBs in these modes must be made. In addition, a~LOCA initiated from zero power (Mode 2) was analyzed. This report discusses the results _ of these evaluations. The results show that the events presented in CESSAR-DC remain limiting for both ' containment pressure and equipment environmental qualification. Table 6-1 presents a - list of cases that were considered. 6.2 LOSS OF COOLANT ACCIDENTS-(LOCAs) Section 6.2.1.1 _of _CESSAR-DC presents the results of the
-containment pressure and temperature analysis for a series of. hot leg, suction leg, and discharge leg LOCAs. Consistent with SRP 6.2 . 1.3', the cases were' based on an initial power level of 102 percent. The limiting LOCA in terms of containment peak pressure is the Double Ended Hot Leg Slot (DEHLS) break.
As power-level decreases from 102-percent, the hot leg coolant and core-temperatures decrease. At the same time, the mass of coolant and pressure on the steam generator secondary sides increases. As a result, primary side stored energy decreases and secondary side energy increases. To show that the cases presented in CESSAR-DC l 6-1
770.wp(9075)bh are limiting, a O percent power LOCA has been analyzed. Although the case in CESSAR-DC which produced the highest LOCA containment peak pressure was the DEHLS, the Double Ended Suction Leg Slot (DESLS) break with minimum safety injection (most limiting cold leg break) was selected for this ..nalysis because the effect of the increased secondary inventory at no load has more impact on cold leg breaks. Table 6-3 lists the assumptions and initial conditions for this case. This case is actually-a Mode 2 case. With the RCS and SG coolant at f56 F, this represents the case with the most NSSS energy of any Mode 2 case. Table 6-4 provides the chronology of events table. The containment peak pressure for this case is 42.0 psig compared to 44.96 psig for the equivalent case at 102 percent power. Results are shown in Figures 6-1 and 6-2. For Modes 3 and 4, the NSSS stored energy is less than for the case analyzed above. Containment spray is still available so that LOCAs for these modes would produce lower containment pressures and tcmperatures. For Modes 5 and 6, the Containment Spray Actuation Signal (CSAS) may not be activated. As a result, should a LOCA occur, the containment sprays would not be available; however, since the RCS coolant temperature for Mode 5 is less than 210 F, a LOCA in Modes 5 and 6 would not result in containment pressurization. Since the LOCAs described in Chapter 6.2.1 of CESSAR-DC are more severe than a LOCA during shutdown modes as far as ccntainment pressurization is concerned, the annulus transient described in Chapter 6.2.1.8 of CESSAR-DC bounds all modes. 6.3 MAIN STEAM LINE BREAKS (MBLBs) In CESSAR-DC, MSLBs were analyzed at 102, 50, 20, and 0 percent prewer, representing Modes 1 and 2. The cases were analyzed with aither the failure of an MSIV oc the loss of a containment spray train. The O percent power case with the loss of a containment spray train produced a containment peak pressure of 48.34 psig. This pressura was the highest of any LOCA or MSLE and this case is the containment DBE. The O percent cases analyzed in CESSAR-DC are Mode 2 cases. Sinco they were based on SG pressures of 1100 psia and RCS and SG temperatures of 556 F, they represent the cases with the most stored energy for Mode 2. Mode 2 cases with less stored energy would be less limiting. In Mode 3 and 4, the NSSS stored energy is less than Modes 1 and 2. As shown in Table 6-1, main steam isolation and containment spray are still available in these modes. Therefore, a MSLB during these modes would not be more limiting than a MSLB during Mode 1 or Mode 2. In Mode 5 or 6, main steam isolation and containment spray may not l -be available. On the other hand, the RCS coolant temperature in 6-2
._ -. .. . - . _ -_ . _ - . - . . -- - _ _ . - .- . ~ . _ - _ -
77a.wp(9075)bh l 1 I l Mode 5 will be less than 210 F. If the SG coolant temperature was also at 210 F, no containment pressurization would occur following a MSLB. An analysis has been performed conservatively assuming that the SG coolant was still at 350 F following shutdown cooling of the RCS to 210 F. Table 6-5 lists the assumptions and initial conditions. Table 6-6 lists the chronology of events. The containmcnt peak pressure for this case is 13.4 psig, well below the CESSAR-DC DBE result. Results for this Mode 5 MELB are shown on Figures 6-3 and 6-4. 3.4 INADVERTENT OPERATION OF CONTAINMENT HEAT REMOVAL SYSTEM 8 During shutdown the containment is purged using either tha low or high volume containment purgo. An inadvertent actuation of the spray system with the containment purgo valves opor vill result in an insignificant decrease in the containment i nta nal, pressure. The parameters affecting the negative containment pressure are the containment atmosphere initial conditions and the spray water teuperature. The source of the spray water for the System 80+ design is the IRWST. Figure 3. 5. 4.1 of Chapter 16 of CESSAR-DC which specifies acceptable IRWST temperatures for a range of containmen' atmosphere temperatures is based on the properties of steam and is not mode dependent. Current TecN.ical Specifications brate that Figure 3.5.4.1 applies to Modus 1 through 4. Section 2.2 of the June 15, 1992 updated submittal of this report will expand the applicability of the figure to include Modes 5 and 6. 6.5 CONOLuuION CESSAR-Oc Section 6.2.1 provides containment analyses to support the establishment of the containment design pressure and temperature and an envelops for equipment e:Nironmental qualification. A spectrum of primary and secondary line breaks were analyzed. With the exception of the o purcent power MSLB l cases (Mode 2), all of the cases analyzed were for Mode 1. As discussed in the sections above, NSSS stcred energy decreases in going from Mode 2 to Mode 6. Safeguards equipmont important to , containment analyses (containment spray and main steam isolation) are available in all modes with the possible exception of Modes 5 and 6. The analyses above show that by the time the plant is in Modes'5 or 6, th+ NSSS energy has been reduced to the point where if a postulated gcimary or secondary line break were to occur, the 6-3
, . - . . .-.r- _ . .
w . - . , . . ,,a
770.vp(9075) bh l resulting containment pressure arid temperature would be less severe than those presented in CESSAR-DC even with containment spray and main steam isolation unavailable. This result ensures also that ; the annulus transient presented in CESSAR-DC bounds all Modes. The inadvertent containment spray act% Lion event presented in CESSAR-DC Section 6.2.1 is used to determine the maximum external containment danign pressure. The analysis presented is not mode dependent. -l I
770.Up(9075)bh TABLE 6-1 E8FAS INSTRUMENTATION l SIGNAL APPLICABLE I Modos ! CSAS 1,2,3,4 (Note 1) MSIS 1,2,3,4 Noto 1: Table 3.3.10-1 of CESSAR-DC Chapter 16 does not show CSAS applicable for Mode 4. Sectiot: 2.2 of the June 15, 1992 updated submittal of this report will expand the applicability of CSAS to include Mode 4.
)
i 1 ( l 6-5 i
. -. = _ . . .-. __ - - . _ _ -
770.wp(9075)bh ) 1 IABLE 6-2 CASES ANALYZED MODE LOCA MSLB 1 102% Power DEHLS, 102%, 50%, 20% DESLS, DEDLS CESSAR-DC CESSAR-DC 2 0% DESLS 0% Current Report CESSAR-DC 3 Analysis Not Required Analysis Not Required Note 1 Note 1 4 Analysis Not Required Analysis Not Required Note 1 Note 1 5 Analysis Jot Required Caae Analyzed With RCS T Since RCS T < 210 F At 210 F and SG T At 350 F Current Report 6 Analysis Not Required Analysis Not Required Since Since RCS T < 135 F Less Limiting Than Mode 5 Note it Less NSSS stored energy than Modes 1 and 2. Same ESF available as in Modes 1 and 2. 6-5
77c.wp(9075)bh i TABLE 6-3 INITIAL CONDITIONS FOR LOCA INITIATED FROM ZERO POWER ParamJitt YAble Reactor Coolant System Average Coolant Temperature, F 556 Containment Initial conditions are consistent with CESSAR-DC Chapter 6, Table 6.2.1-18. l l l l l l 6-6
'77c.wp (9 07 5) bh TABLE 6-4 ACCIDENT CHRONOLOGY FOR LOCA INITIATED FROM 2ERO POWEB Time,sec Ey.ent 0.00 Break occurs 18.00 Start Safety Injection Tank Injection 21.98 Peak Containment Pressure before End of Blowdown 22.00 End of Blowdown 25.90 Downcomer Full 71.00 Containment Spray Injection 71.80 Peak Containment Pressure Subsequent to End of Blowdown 100.00 Safety Injection Tank Empty 108.30 End of Reflood 202.57 End of Post Reflood 6-7
77e.wp(9075)bh TABLE 6-5 INITIAL CONDITIONS FOR MSLD INITIATED FROM MODE 5 Paragglel Value Reactor Coolant System Average Coolant Temperature, F 210 Steam Generator Secondary Pressure, psia 132.8 Steam Generator Secondary Temperature, F 350 Containment Initial conditions are consistent with CESSAR-DC Chapter 6, Table 6.2.1-1C. l r i l 6-8
770.Vp(9070)bh TABL2 6-6 7171 DENT CHBONOLOGY FOR MSLD INITIATED FROM HODE 5 T}Ab.319 Evant 0.00 Break occurs 148.71 Peak Containment Temperature Peak Containment Pressure flote MSIS and Containment Sprays are assumed not to be available in this mode 4 1 t l 6-9 .
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.=. _ ._
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l l l l l l l ! l l l l l l 0 40 - 80 120 R0 200 20 280 1 WE,ECONDS i m Figure dh / [u LONTAINMENT. PRESSURE vs. TIME FOR LOCA FROM ZERO POWER 61 l
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. 220 -
u ,
$210-g 200 -
u j -190-J 15 - u '
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120 - 110 i i i r i i i i i i i i i i 0 40 80 120 1(0 200 240 2"u - INE, SECONDS Figure
.$M /
i CONTAINMENT ATMOSPHERE TEMPERATURE vs. TIME FOR LOCA FROM ZERO POWER F ^ Fif-*f PA*+- W p
14 ! 13 - !
.l 12 -
ll~ f 0 . I 5 10 - a U g_ . E t
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2- , 1-0 i. i i i i ;i i i , , , , , i 0 02 - 0.4 0.6 - 0.8 1 12 1,4 (Ihouds) x steos: i m Figure CONTAINMENT PRESSURE vs. TIME
-$N .FOR MSLB FROM MODE S 63
' . ,m,, ._ . - . _ , - ~ . . _ _ _ _ . , , , , _ . , , . . _ . , _ , . _ _ - - . - _ . _ , .
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- 1. 180 -
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WE, TCWDS m Figure
-d /d u CONTAINMENT ATMOSPHERE vs. TIME FOR MSLB FROM MODE 5 64
_ _ _. .. _ _ _ _ _ _. ________.___.m ...__..___.m
-77a.wp(9075)bh ;
7.0 SYSTEM 80+ Dl8IGN FEATUREg FOR DIMPLICITY OF SHUTDOWM _ Q.P_ERATIONS ;
7.1 INTRODUCTION
The System 80+ evolutionary ALWR design-takes maximum benefit from prior design and- operating experience. - It is an objective that ; this benefit _ be evident in design features that aid outage planning, reduce operator stress and simplify operator training.
-Previous scctions of this ~ report describe and evaluate features of System 80+_that will improve the overall shutdown operations. In the following discussion, the outage benefits accruing from these design features are presented. ,
.7.2 DISCUSSION There are several features of the System 80+ design which will aid
. in - the management of an outage. Many of these same features ,
relieve some of the stresses and pressures placed on the licensed operators. These: features are presented in-Sections 7.2.1 through 7.2.9. 7.2.1 TECHNICAL SPECIFICATIONS FOR REDUCED INVENTORY [ Technical Specifications for the Jystem 80+ design specify restrictions on operation in Reduced Inventory. Reduced Inventory
-is determined by water level in the Reactor Coolant System (RCS).
This_ plant operational condition is defined as "RCS level greater than 3 feet below the reactor vessel flange". These specificatien. provide guidance-to the operations staff and the outage management team. - This guidance ensures that equipment assumed to be available
-for accident mitigation is operable. This aids the outage planners by identifying _ equipment .on _
which_ _ maintenance cannot Le accomplished during Reduced -Inventory. This planning process i reduces the chance for. removal of components relied upon . for i detection and response to accidents. These specifications provide-the Senior _ Licensed Operator a clear standard for determining minimal equipment availability. This_will alleviate some of the . stress placed on the operator to make the final judgement as to' D whether equipment is required to be operable for Reduced Inventory L vperations. . 7.2.2 SHUTDOWN COOLING SYdTEM l' Ine Shutdown Cooling. System (SCS) design provides two divisions for decay heat removal capability. . _These two divisions are completely redundant, that is,ithey share no components or equipment. This , redundancy providos the operatur with a' standby decay' heat removal system if any component fails to perform its function. This assurance of a standby system availability reduces some of the , 7-1 l- _ _ _ _ . . __~ .. . _ _ _ _ . _ - _ _ _ . . . _ _. _ _ __. ~ . . -._ . - . , _ - _ . . . . _ - - _
-m_ m m. A-i_ e.A. a m- 4 44 As..hma -44 A 4 2 4. F-e Ju---- -3i h.-4 E.-,Am,4A--X.M_Eu4--.b4A-M4A4E.I 5A24++---h--4-h+a ---L-.A 4-444L4u--i1.rM-= A 77a.wp(9075)bh pressures and stresses placed on the operator when no standby system is available. ,
Another feature of the SCS is that it is not a subsystem of the Emergency Core Cooling System. Therefore, the SCS is not required to be operable in Modes 1 through 4. This feature of System 80+ design increases the availability of SCS during Mode 5, Mode, 6 and Reduced Inventory. In addition, this feature aids the outage planning team by allowing maintenance and repair of SCS components and equipment to be accomplished during power operations. This feature eliminatts the necessity of finding a window during the outage to allow vork on decay heat removal equipment. Besides the features discussed above, the SCS is . designed to provide faster venting of the pumps. . If SCS pumps become vapor bound due to misoperation, venting is required. The vent piping for the pumps is hard piped and directed to 11e floor drain sumps. This allows the plant equipment operator to quickly vent the pump without attaching vent rigs. These vent rigs waste valuable time if recovery from a loss of decay heat removal is required. , 7.2.3 CONTAINMENT SPRAY SYSTEM The Containment Spray Syst' t (CSS) design provides two divisions or egripment which can be utilized as an alternate decay heat removal flow path. The CSS pumpa are interchangeable with the SCS pumps. This feature provides the operations staff with increased flexibility in the area of forced circulation. Therefore, this alternate alignment for forced coolant flow during shutdown conditions reduces stresses placed on the operators since it increases the redundancy and therefore reliability of the System 80+ decay heat removal capability. With these alternate pumps, the operators have assurances that redundant equipment is available. 7.2.4 COMPONENT COOLING WATER SYSTEM The Component Cooling Water System (CCW) design has two redundant divisions. Each of the two divisions contains two pumps and two heat exchangers. This interdivisional redundancy of syt, tem components provides flexibility for the management of the maintenance outage. Therefore, major components (e.g., pumps and heat exchangers) requiring maintenance can be removed from service without affecting the availability or reliability of the interdivisional equipment. This enhancement of system design provides the outage planner with options to facilitate easier outage scheduling.. 7-2
77a.wp(9075)bh 7.2.5 STATION 3ERVICE WATER SYSTEM The Station Service Water System (SSW) design has two redundant divisions. Each of the two divisions contains two pumps. This interdivisional redundancy of the pumps provide flexibility for outage planning. The outage planner can schedule a pump for maintenance without affecting the availability or reliability of the interdivisional equipment nor the redundant division equipment. 7.2.6 ELECTRICAL DISTRIBUTION SYSTEM The Electrical Distribution System (EDS) design has two redundant safety divisions. (Refer to Figure 2.4-1.) Each division is capable of being powered from four separate and diverse sources. These sources include:
- 1. Switch)ard Interface I,
- 2. Switchyard Intsrface II.
- 3. Diesel Generators, and
- 4. Combustion Turbine.
The EDS provides the outage planner with the flexibility to emove a source of power for maintenance and still maintain other reliable sources to the safety buses. Therefore, required maintenance activities are scheduled wi'hout reducing the reliable sources of power to unacceptable levels. These same features provide the licensed operator with alternate sources to which safety buses can be aligned. The operator is aware of these approved alternate alignments through procedures and training. Therefore, the stress placed upon the operator to align to any available sourca regardless of guli.ance is reduced. In addition, operator training is facilitated by the procedural guidance. Another feature of the EDS design is the use of 4 safety buses, 2 per Division. Th3 1E loads are uvenly distributed on the buses to ensure redundancy of system components. For example, each bus powers 1 of 4 Component Cooling Water pumps. This feature provides f'exibility for the outage planner. One bus can be removed from service for maintensnce and redundant componentt still have a power supply availabic. 7.2.7 NUPLEX 80+ ADVANCED CONTROL COMPLEX The NUPLEX 80+ Advanced Control Complex (ACC) is an integral part of the System 80+ design. The design goals of NUPLEX 90+ include the integration of NSSS and balance of plant systems into a unified i control complex, reduction of human errors that af fr at plant safety l and improving the reliability of the man-machine interface through redundancy, segmentation and diversity. 7-3 1
77a . wp (9075) bh Control room information provided by NUPLEX 80+ is consistent with operator information requirements when performing operational tasks during plant evolutions or responding to unexpected conditions. The operator can obtain information from a number of sources in the NUPLEX 80+ ACC. These sources include: A large plant overview status board known as the Integrated Process Status Overview (IPSO).
- Alarm tiles and associated message windows.
Discrete indicators provida frequently used and importent information. CRT displays containing essentially all power plant information. More detailed information on the NUPLEX 80+ ACC can be found in Chapter 18 of System 80+ CESSAR-DC. The NUPLEX 80+ ACC utilizes the same parameter conventions for the indicators and alarms for shutdown operations as required for power operations. The features of each which simplify operator training, aid in outage planning and reduce operator stress are described below.
- IPSO - IPSO providts'.ho operators, especially Senior Reactor Operators with an overview of plant status during shutdown conditions. Thia overview allows the operators to view system status during outage activities. Having an overview of the nian* will reduce uncertainty of the availability of required
- M e systems. This knowledge of the availability reduces the
.n ' placed on the operators by uncertainties.
- Tl u is -
Mode and equipment dependent alarms are a special cM. ore of NUPLEX 80+ ACC. This- feature eliminates the alarming of alarms not applicable to the current mode, operating conditions, or equipment status. A large amount of maintenance activities involved with the outage af fect control room alarms. The mode and equipment dependent alarms eliminate operator response to nuisance alarms caused by authorized work being performed. Outage work may be planned, alarms disabled and unnecessary investigation by operators into these alarms eliminated.
- Discrete Indicators -
Discrete Indicators provjde several simplifying attributes for operators. Automatic ranging scales on Discrete Indicators allow accurato indication over the entire range of system design with the same indicator. Using the same indicc r ur for all conditions of operation, including shutdown, avoids confusion for the operators. This feature 7-4
._ ~ _ _ _ . _ , _ - _ . . .-_ - - - _.
I 3 1 7 7 a . wp ( 9 07 5) bh allows training to utilize the same indicators on a simulator. It also eliminates the utilization of indicators solely for shutdown. Discrete Indicators receive multiple channel input i signals to be displayed on one indicator. These signals are 1 validated and provided the operator with reliable indication l oven if some channels are removed from service. Individual l chanr.el inputs to the Discrete Indicators ainem to alert the operator when one has been removed fvna .orvice. This allows the operator to check the status of information provided. Using validated displays reduces stress on the operator by climinating doubt of instrument availability and accuracy.
- CRT Displays - c'T displays for the NUPLEX 80+ ACC are arranged in a structured . rmation hierarchy. This structure provides the operator information consistent with operational needs.
Levels of display infGrmation start with IPSO and continue through detailed plant information. A feature of the CRT display is graphic representation of systems. This reinforces system layout training and leads to better understanding by the operator. It provides consistency for the operators which reduces stress. Color representation of valves to indicate operable / inoperable status gives the operator the information to determine flovpath status. This feature also provides status of maintenal."a in progress on important valves in the plant. 7.2.8 REDUCED INVENTORY INSTRUMENTATION The System 80+ design provides the instrumentation to insure the Contro Room Operator (CRO) is informed of the decay heat removal system performance and the reactor coolant system level and temperature. The instrumentation includes:
- 1. Reactor Coolant System Level,
- Heated Junction Thermocouple e dP
- 2. Reactor Coolant Temperature,
- Core Exit Thermoccuple
- Heated Junction Thermocouple
- 3. Shutdown Cooling System Flow,
- 4. Shutdown Cooling System Pressure,
- 5. Shutdown Cooling System Temperature, and
- 6. Shutdown Cooling Pump Motor Current.
See Section 2.8, Instruinentation, of this report for a discussion of this instrumentation. The instrumentation is coupled with the Nuplex 80+ Advanced control Complex (ACC) to provide the CRO with indications and alarms to monitor reduced inventory operations. (See Section 7.2.7 above for a description of Nuplex 80+ ACC features). 7-5
l 770.wp(9075)bh s.0 CONCLUSIONS This section will be provided with the July 31, 1992 updated submittal of this report.
+
8-1 x.
77c.wp(9075)bh
9.0 REFERENCES
- 1. Letter LD-92-038, C. B. Brinkman (ABB) to D. M. Crutchfield (NRC) dated March 25, 1992.
- 2. Memo Letter, " Summary of Meeting Held on December 18, 1991 Regarding Shutdown Ris' " , T. V. Wambach (NRC), dated January 30, 1992.
- 3. NUREG-1449, "Shutdo'in and Low-Power Operation at Commercial Nuclear Power Plants in the United Stnas", Draf t Report dated February, 1992.
- 4. USNPC Generic Letter No. 88-17, " Loss of Decay Heat Removal",
dated October 17, 1988.
- 5. USNRC AEOD Special Report, " Review of Operating Events occurring During Hot and Cold Shutdown and Refueling", dated December 4, 1990.
- 6. NUREG-OPOO, USNRC Standard Review Plan, Revision 1, July,1981.
- 7. Jaquith, R.E., et.al., "Probabilistic S.isk Assessment for the System 80+ Standard Design", Combustion Engineering Inc. , DCTR-RS-02, January, 1991.
- 8. Letter from Jan.es H. Wilson (NRC) to E. E. Kintner (EPRI) dated September 5, 1991.
- 9. Brockhold, G. (EPRI) , Memo to ALWR Utility Steering Committee, ALWR-92-18, January 15, 1992,
- 10. (Later)
- 11. ANSI /ANS-58.8-1984.
- 12. CEN-373-P, Volume 1, " Realistic Small Break LOCA Evaluation Model, Calculational Models," April 1988.
Volume 2, "\pplication of Evaluation Model," December 1938. Volume 2 SupplemLnt 1-P, " Application of Evaluation Model to Calvert Cliffs Units 1&2, September 1989. V(.lume 3, " Computer Program Input and Output Description," December 1988,
- 13. Letter LD-88-030, " Submittal of Realistic SBLOCA Evaluation i
Model,' A. E. Scherer (ABB CE) to J .- A. Norberg (NRC), April 27, 1988. L , l- 9-1 l w - - _. .- .- - _. .. - - .. - - _ .
.- _ _ . _ _ . _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . = . . _ _ . _ . . - . _ . . _ _ . _ _ _
77a.wp(9075)bh t
- 14. Letter LD-88-155, " Submittal of Volumes 2 and 3 of Combustion Engineering's Realistic Small Break Losn-Of-Coolant-Accident Evaluatiott Model, " A. E. Scherer (ABB CE) to J. A. Norberg (NRC),. December 9, 1988.
- 15. Letter LD-89-001 " Addendum to Volume 3 of Combustion Engineering's Realistic Small Break Loss-of-Coolant Accident Evaluation Model," A. E. Scherer (ABB CE) to J. A. Norberg (NRC), January 11, 1989.
- 16. Letter LD-89-099, "Supplomont i to Volumo 2 of Combustion Engineering's Realistic Small Break LOCA Evaluation Model," A.
E. cherer (ABB CE) to J. A. Norberg (NRC) , August 28, 1989.
- 17. ANSI-N18.2-1973. ,
- 18. " System 80+, CESSAR-DC," Technical Specification, 3.4.3.
19.- 10CFR50. 4 6.
- 20. CENPD-138, Supplement 2-P, " PARCH, A FORTRAN-IV ' Digital Program to Evaluate Pool Boiling, Axial Rod anca Coolant Heatup,"
January, 1977. 9-2 l
. - , __, , .__ _ -- - . - - - - - - - - - - - - - - - ~ ~ - - - - -
1 77a.wp(9075)bh l APPENDIX A RESPONSES-TO REQUESTS FOR ADDITIONAL INFORMATION e
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77a.vp(9075)bh
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APPENDIX A l TABLE OF CONTENTS Section Pace No. l A1.0 Introduction A-1 ) i A2.0 Requests-for Additional Information A-2 i Question 440.14* Question 440.16a Question 440.16b Question 440.16c Question 440.16d Question 440.16e* Question 440.16f Question 440.16g
-Question 440.16h Question 440.161 Question 440.16j- -
Question 440.35
-Question 440.36*
Question 440.37* Questiort 440.49* Question 440.54* Question 440.70* Question.440.86f Question 440.91 Question 440.109*
' Question 440.129
-Question 440.130 Question 440.131 Question 440.132 Question 440.133
. Question 440.134
-Question. 440.135.
. Question 440.136.
Question 440.137.
. Question 440.139*
Question ~ 440.140 Question 440.141~ Question' 440.142 Question-440.143 Question 440.144-
-Question- 440.145
. Question 440.146
. Question 440.147 Question 440.148
. Question' 440.149 Question 440.150-
- .. - . .. . - . . . ._. _.= ....__ __ - - _
77c.wp(9075)bh APPENDIX A (Continued) TABLE OF CONTENTS Section Pace No. A2.0 Requests for Additional Information (Continued) Question 440.151 Question 410.54* Question 410.65* Question 410.66* Question 410.72* Question 410.73* Question 410.103g*
-Question 410.104d Question 410.107a*
Question 410.107b Question- 410.107c Question 410.107d Question 720.98 Question.722.94 Question 280.1*
- Response was previously submitted and is reproduced in Section A2.
A3.0 References l
77a.wp(9075)bh P A
1.0 INTRODUCTION
Letters from the NRC that transmitted Requests for A'Jditional Information (RAIs) on shutdown related topics are listed in Section 3.0 of this Appendix as References A ' *hrough A-6. ABB provided individual responses to some of t: '1Als in References A-7 through A-13. These responses are ,duced in this Appendix. The remaining RAIs related to chutdown ..sk were listed in the ABB letter, Reference A-14 along with a commitment to provide responses via this report. These responses are also given in this Appendix, where the response may refer to the content of applicable' sections of thiP report. The combination of the specific responses provided in the Appendix along with the overall report content fulfills all outstanding commitments to provide shutdown risk information in support of the CESSAR-DC review process.
-A2.0 RAIs The following-- pages - provide the -usual format for question and response to the RAIs. Those not included at this time will be submitted in the June 15, 1992 or the July 31, 1992 updated submittals scheduled for this report.
i ev e ,a
77a.Vp(9075)bh Ouestion 440.16(i): Safety analysis reports (SARs) typically concentrate on power ! operation when consideration is given to many of the potential operational transients. The recent experience from the events in , operating reactors indicated that further evaluation for plant operation at lower modes may be required. Hence, it may be prudent to address non-pow.r operation in more depth than has been traditional. What plans exist, if any, with respect tc this topic and the System 80+ program? Response 440.16fi): The System 80+ program encompassed _an extensive evaluation to assess the vulnerability of the System 80+ design to various transients. The results of this evaluation are provided in the System 80+ Shutdown Risk Evaluation Report. Sections 2.4, 2.5, 2.6, 4.0, 5.0, and 6.0 of this' report present transient analyses covering the extent of events of concern for shutdown operations. Sections 2.4 through 2.6- concentrate on the loss of decay heat removal and rapid boron dilution events. Section 4.0 represents the evaluation of the impact of CESSAR-DC Chapter 15 events occurring during shutdown modes of operation. Section 5.0 represents-the evaluation of the impact of the CESSAR-DC Section 6.3 loss of coolant accidents occurring during shutdown modes of operation. Section 6.0 evaluates the impact of the CESSAR-DC Section 6.2 events on the containment response when these events are initiated from shutdown modes of operation. In addition, to supplement the above evaluations Section 3.0 of the Systou 80+ Shutdown Risk Evaluation Report presents a probabilistic , Risk Asuessment which covers various event sequences which can
- occur during shutdown operations.
\ This response and the mentioned report sections fulfill the commitment in Reference A-14 relevant to this RAI. l l w +-m-
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- Question 440.49 Provide a discussion of the procedures and plant systems used to take the plant from normal operating conditions to cold ;
shutdown conditions. This discussion snould include, heat I removal, depressurization, flow circulation, and reactivity control. i
- Resnonse 4402 11 f l
The principal systems utilized in taking the plant from Mode 1, Power Operation, to Moce 5, Cold Shutdown are: ! Roactor Coolant System 4 Feedwater System Feedwater Control System Reactivity Control System Boron Control System Chemical & volume Control System Sutdown Cooling System ensurizer Level Control System steam Bypass Control System Pressurizer Pressure Control System Liquid and Gaseous Waste Management Systems Main Steam System
- Condensate System Reactivity control capability is discussed in Sections
- 7. 7.1.1.1; 7. 7.1.1. 7 ; 9. 3. 4.1. 3. 31 9. 3. 4. 2.1 (last paragraph) ,
and 9.3.4.2.3C. Power is reduced by increasing the boron concentration in the RCS to reduce k-effective to 5 0.99. At low power the rods are inserted. The operator borates to the cold shutdown boron concentration consistent with the Technical Specifications prior to the beginning of cooldown. This ' margin is maintained throughout cooldown by making up shrinkage volume by means of the CVCS with water at the cold shutdown margin boren concentration. Cooldown is ef fected by the systems described, and techniques discussed in Sections 5.4.7.2.6.A; 9.3.4.2.3C; 7.7.1.1.2.1; 7.7.1.1.2.2; 7.7.1.1.4; 7.7.1.1.5 and 10.4.7.2.4. Additionally, the following precautions, limits and techniques are utilized during cooldown:
- 1. T;.a reactor coolant pumps continue to run until they are manually tripped o Four RCPs shall not be operated below app roximately 500*F o During cooldown RCP 1A and 1B shall be running to maintain pressurizer spray capability until they are required to be shutdown for some other reason
- r ,
, _ ~- . _ .- ._
'Resoonse 440.49'(continued) o The RCPs shall not be operated when the system o pressure is below cavitation or seal operation limits
. The RCS pressure is maintained at 2250 psia until
'cooldown is initiated.
- 3. Pressurizer pressure and level controls are placed in aanual mode at the beginning of cooldown, and power to
. taters is reduced.
flow is maintained throughout cooldown by RCPs a Icr shutdown cooling system pumps. a hubble in the pressurizer is maintained as long as assible. C. .1me Centrol Tank (V C) gas space is vented to reduce P 1.:sion gas and hydrogen gas prior to cooldown.
- 7. Letdown flow is directed, as required, to the gas stripper to re;mve dissolved nas.
- 8. Initially, heat is removed-from RCS by dumping steam:
o Steam may be dumped to the condensers through the Steam Bypass System, or to the atmosphere through the Atmospheric Dump vaJves (ADV). o Fccc cori.r >l is in manual during cooldown using the start", pump and manual control valves. o The STIS setpoint is adjusted to 200 psi below existing steam pressure as cooldown progresses.
- 9. As RCS water cools, pressure is decreased by- manually adjusting pressulizer spray to cool the vapor space.
Prassuriza.r pressure is controlled such that saturation margin limit is not exceeded, and such as to comply with the pressure-temperaturc curves specified for the plant.
- 10. As pressurizer pressure decreases the SIAS & CIAS setpoints are decreased to 400 psi below existing pressurizer pressure.
- 11. RCS cooldown rate shall be maintained within rechnical Specifications (TS) at all times during cooldown.
- 12. Pressurizer water temperature should c.s c eed PCS water temperature.by no more than 350*F and no less then 50*F whenaver there is a bubble in the pressurizer.
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4-ResDonse 440.49-(continued)
- 13. Auxiliary pressurizer spray is utilized to reduce i pressuriErsr pressure whenyver normal spray is inadequato
.or not available.
- 14. When tne pressurizer pressure is approximately 400 psia and the RCS temperature decreases to 350'F, cooldown is transferred to the Chutdown Cooling System (SCS).
Cooldown from this point is fully described in Section 5.4.7.2.6A. Steaming and feed may be terminated.
*This response was previously transmitted by Reference A-12 l
l l l = { 1 l l -
1 Ouestion 440.70 Describe the means provided for ECCS pump protection including instrumentation and alarms availab'e to indicate degradation of ECCS purp performance. The staff's position is that suitable means should bo provided to alert the operator promptly to possible degradation of ECCS pump performance. All instrumentation associated with mo"itoring the ECCS Jump performance should be operable without 'fsite power, and should be able to detect cor.ditions of low discharge flow.
*Ersoonse 440.*FJ The following instrumentation is used to determine Safety Injection and Shutdown Cooling pump performance. This instrumentation is shown in Figures 6.3.2-1A, B, and C and described in Section 6.3.5.3 of CESSAR-DC.
Channel Epnction Control Room Features P-302, 305 .SCS Pump Discharge Indication Preswere P-306, 307, SI Pump Discharge Pressure Indication 308, 309 P-319, 329, SI Line Pressure Indicution 339, 349 Alarm (High) P-390, 391 SI Hot Leg Injection Indication Pressure Alarm (High) F-302, 305 SCS Lin6 Flow Indication Alarm (Low) F-306, 307, SI Pump Discharge Flow Indication 308, 309- Alarm (Low) F,-311,- 3 21, DVI Nozzle Injection Line Indication 331,.341 Flow
- :F-390, 391 SI Hot Leg Injection Flow Indicatior. ,
The normal power supplies for the above instrument 0 tion are the 120-VAC vital I&C buses, which are powered from either the Class 1E 480 VAC buses or the 125 VDC buses through inverters (See CESSAR-DC Figure 8.3.2-2). If offsite power is lost, the Class 1E 480 VAC buses may be powered via the 4.16 Kt' safety buses by the emergency diesel generators (See CESSAR-DC Figure 8.3.1-1) or by the alternate AC power source (gas turbine).
Testing to confirm that SI and SCS pump performance is within specification is included in the Safety Injection System Test sections of Preoperational Tests, Section 14.2.12.1 of CESSAR-DC. In addition, Technical Specification 3.5.2 (CESSAR-DC Section 16.'3.2) provides rerlirements for testing safety injection flowrates.
*This response was previously transmitted by Reference A-12 # e m 6
e f 4
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77a.wp(9075)bh' ) l l Ouestion 440,91: Most of the Chapter-15 and 6.3.3 (LOCA) analyses are performed based on the event being initiated at full power operation. The staff requires that C-E provide an assessment on the consequences of the transients and accidenta initiated at low power levels or lower modes of plant operation such ao shutdown operations. This is required to demonstrate that tne analyses performed in CESSAR-DC are the bounding cases for all modes of plant operation. Response 440.91: This assessment is provided in the System 80+ Shutdown Risk Evaluation Report. Section 4.0 of the report contains the assessment of CESSAR-DC Chap' " 15 events and Section 5.0 contains the assessment of CESSAR-DC Section 6.3.3 events. Also see the response to RAI 440.16(j). The response and the mentioned report sections fulfill the commitment in Reference A-14 relevant to this RAI. l l
779.wp(9075)bh-l Ouestion- 440.109 (15.6.3)
- Provide the results of an analysis for the potential boron dilution event during the recovering phase fcllowing a SGTR when backfill fron the secondary system through the ruptured steam generator occ i : red.
- Response 440.109 The System 80+ Emergency Procedure Guides will include steps to l prevent backfill from the secon.lary system through the ruptured steam generator by maintaining a positive pressure difference between the primary and secondary systems. (See Secticn 2.1 of the System 80+ Shutdown Risk Evaluation Report.) Therefore, boron dilution should not occur and has not been analyzed. A further note is made that backfill is not necessary to prevent overfilling of the larger System 80+ steam generator as the result of a SGTR event.
O
*This response was previously transmitted by Reference A-11.
Changes have been made as noted by the bars in the margin. L l I l l l l l l t l l' - _
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I
- Ouestlon 440.139--
As mentioned in.RAI 440.109,. discuss the potential for boron dilution:daring the recovering phase following a:SGTR when
- backfi2.'O from-the : secondary system- through the ruptured S/G occura. This analysis should also.be provided:in. support of GSI-
~
.22, CESSAR-DC Section,15.4.6, etc. . . .
--* Respong.e 440.139 The issue of'a " potential" boron dilution-resulting from a-SGTR accident war addressed in the response to RAI 440.109, 4
5
*This response was previously transmitted by Reference A-11 O
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77a.wp(9075)bh Ouestion 440.148: Will there by any maintenance-activities for the System 80+ that will require isolation ' of IRWST pump suction inlets (or allow foreign material in the sump with potential for blockage)? If so, this would preclude operation of safety systems. What guidance can be provided to minimize this potential risk? Have TSs been provided limiting such maintenance activities? Response 440.148:* No maintenance activities that will require isolation of the IRWST pump suction inlets are possible because the inlets will be submerged during all modes of operation. Mai1Menance in the IRWST is only possible during moda 6, when IRWST inventory has been transferred to the refueling pool. During refueling operations, the Shutdown Cooling System pumps utilize the IRWST ECCS suction connectior.s to fill the refueling cavity. Due to NPSH and vortexing considerations, the suction inlets are sufficiently submerged to protect the SCS pumps while the pumps are in opert * . While ma_ntenance is being performed in the IRWST, the possibility exists for foreign material to accumulate in the tank. The System 80+ design includes provisions to prevent this debris from entering or blocking-the ECCS suction lines. The suction lines are isolated from areas of high maintenance (i.e. , away from the IRWST spargers) to decrease the possibility that debris will reach the inlets. ! Large vertical screens capable of filtering particles greater than 0.09 inches diameter are located within the IRWST to effectively block - debris. Should maintenance be required in areas where maintenance-generated trash would not be filtered by the debris screens, a mesh " cage" that completely surrounds the euction inlets - would- prevent debris from entering the lines. A- complete discussion of this issue is provided in section 2.9.3 of the Shutdown Risk Evaluation Program Report. Using procedural guidance provided by the plant designer in Section 2.1 of the final submittal of the System 80+ Shutdown Risk Evaluation Report,_::he owner-operator will develop plant specific procedures that require that maintenance-generated trash be removed l from the IRWST before refi] ling the tank. l
- This reponse supplements this report and together they fulfill the commitment to respond in Reference A-14.
l l l l l
~ i ggestion 41QJ.i: The safety evaluattu of both the new and spent fuel storage areas includes an evaluation of the effects of dropping a fuel assembly and-its handling teo11from a height of two feet above the storage rack. Provide. the following additional information in accordance with SRP 9.1.2, Item III.2.e guidance: Verify that the drop of any allowed lighter loads at a greater height does not result in.a higher potential enenay than a fue1~ arsembly and its handling tool dropped from its normal operating elevation. Perform an evaludion of this in accordance with SRP 9.1.4 guidance.
- Response 410.54:
. The spent fuel racks have been evaluated ar.d the results show that the rack L will be less than .95 under the following postulated accident conditi,o,ns:
(1) Drop of a fuel assembly handling tool from its maximum lift height ever the fuel racks. (2) Drop of a fuel 3: .embly and the handling tool froc, their maximum
-lift height ove the fuel racks.
(3) Drop of other items, such as a failed fuel canister with a fuel assembly, from their maximum lift' height over the fuel racks.
*This response was previously transmitted by Reference A-13
~ . . - . .- - -
770.wp (9075) bh Ouestion 410.54 The . response to RAI 410.54 is incomplete. Please provide the values for the maximum lifting height assumed for each case analyzed.
- Response 410.54 Lift heights and grapple weights are not known prior to final procurement of the refuuling machine used in the fuel building.
System 80+ spent fuel racks can absorb an impact energy of 93,100 inch-lbs without exceeding the rack K,rc criteria. The refueling aquipment will be designed so that the maximum impact energy resulting from a dropped fuel assembly and handling tool, a dropped handling tool or-any other dropped fuel handling related load from their maximum respective lift heights will not exceed the energy absorbing capacity of the fuel rack while maintaining K.tr criteria.
-In addition,-the owner-operator, using procedural guidance provided by the plant designer in Section-2.1 of the final submittal of the System 80+ Shutdown Risk Evaluation Report, will develop administrative controls to limit the sizc. and lif t height of any-other non-fuel handling loads that are carried over the fuel racks such that this maximuu impact energy is r.ot exceeded.
*This response was ' previo 1 sly transmitted by Reference A-13.
Changes have been made as noted by the bars in the margin. i t i
, ,. ~. - _. _ _ _ ,
L Ouestion 410'.64 vnji.have identified the different' torage densities for' regions I ar.d 11-of the-spent fuel pool. (50% and 75 , respectively) infyour submittal. Provide ~ pertinent information concerning:the design criteria and anticipated controls;to be? implemented for the storage of spent fuel assemblies-in the. above regions.
- Resconse 410.64 luth Region =I and II storage areas.are designed to. accommodate 'uel essemblies with initial enrichment up.to 5 weight percent U-235. ' Region I.has no restriction on burnup history of stored fuel assemblies. Region II-is restricted for storage of fuel having a minimum cumulative burnup which is dependent-on the initial enrichment for each fuel assembly. The
. burnup versus enrichmer.t curve is internally documented._ This restriction on fuel storage in Region II will be imposed by administrative controls _ developed and implemented by the Owner-Operator; TheLfollowing will be added in Section-9.1.2.2.2: "A fuel assembly may.
~
be stored in Region"Il only if it has the minimum burnup required for an Jassembly of:its initial enrichment. The Owner-0perator will develop and implement administrative controls to_ permit storing _ a fuelfassembly in Region'II- only if it meets established burnup versus initial enrichment requirements."
*This response was previously transmitted by Reference A-8 \
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ai .. J jg Question 410.65 You have statedLin: Section19.1.2.3.1.3 that one of the accidents considered ~in the design of _the spent fuel pool storage racks is a fuel-
' assembly and its handling tool " falling into a blocked-off fuel storage cavity." Supply additional-information concerning the mechanical blocking assemblies to: allow determinaticn of the extent of penetration of a tuel assembly into~ a blocked cavity. ~
- Response 410.657 -
The spent febl racks provide- storage for 363. fuel assemblies in Region I (50% density; :.nd 544 fuel assemblies in Region II (75% density). The -
. restricted. rack-cells contain cell blockers which prevent the. placement of a fuel assembly =into the rostricted cells. The racks have been analyzed based on a postulated accident cendition of a. fuel asserably fully insertedrinto a restricted cell. Taking pool btron concentrations into consideration, the results show that the rack k,,, is'less thar. 95.
-The cell blockers cannot be inadvertently removed once installed as -
special tooling is required to' unlock and remove them from the spent fuel storage racks.
~
*This' response was previously transmitted by Reference A,8 l
L
-. . .-,,-, , c.-,.
Ouestion 41Q dQ
*aur su'bmittal doet not provide information-concerning the handling of-heavy loads in the vicinity of the spent fuel pool. Provide an evaluation of the capability of the spent fuel loading pit to withstand a dropped heavy load. The evaluation ,chould include a shipping cask drop without brecch of the pit area or loss of spent fuel pool water. *Resoonse 410.66 The spent fuel cask laydown area is separated from the spent fuel pool by a gate and a structurally reinforced concrete wall. The gate is closed, sealed, and locked during all cask handling operations. The floor in the laydown area has been designed to withstand the impact of a shipping cask dropped from a height of 30 feet without breaching the integrity of the floor plate.
Any small water loss as a resul+ 9 local damage to the laydown area wall liner cannot be communicated cu the spent fuel pool due to the closed gate and the integrity of the independent spent fuel pool liner. Damage to the gate is prevented during cask handling by stops on the bridge crane rail that limit cask travel and by the recessed gate design. Design features to address the spent fuel cask drop accident are summarized in CESSAR-DC Section 15.7.5, Amendment H.
*This response was previously transmitted by Reference A-8 i
l l I I
Ouestion 410.72 Provide the fuel building layout drawings which show the (1) overhead heavy load paths and (2) safety-related equipment locations in the vicinii f if those paths susceptible to dnmage by failure of electrical interlocks, swinging of the load, or other mechanisms for causing damage.
- Response 410.72 The containment polar crane, the cask handling crane, .nd the fuel handling crane are designed to prevent the drop of a heavy load such as the reactor vessel head and the spent fuel shipping cask. In addition, predetermined load paths for major iifts (see figures 9.1-19 and 9.1-20),
operator training, and regular crane maintenance minimize the possibility of load mishandling. Limit switches, electrical interlocks and mechanical interlocks prevent improper crane operation which might result in a fuel handling accident. This is also discussed in Section 3.1.4.2.1.7. The spent fuel cask l handling hoist is restricted from movement over the new and spent fuel storage areas when the fue! racks contain fuel assemblies. The new fuel handling hoist is restricted from movement over the spent fuel storage arca when the spent fuel racks contain fuel assemblies. i In accordance with the regulatory position of Regulatory Guide 1.13 and i General Design criteria 61 of Appendix A to 10 CFR 50, the hoists are also restricted frra passing over the spent fuel pool cooling system or ESF systems which could be damaged by dropping the load. Set points for the hoist interlocks are set to preclude falling or tipping of the loads into the fuel' storage areas. Typically, administrative controls prepared by the Owner-Operator preclude movement of heavy loads within the containment building pool l when the refueling machine contains a fuel assembly. During heavy load ' movement, the fuel transfer tube vai,e is closed to avoid water level , changes in the fuel building during postulated accident conditions such ! as dropping the heavy load on.the reactor vessel pool seal. THe first sentence of the last paragraph of Section S.I.4.3.1 has been , modified to state: " Administrative evtrols prepared by the Own**- Operator..."
*This response was previously transmitted by Reference A-C r
l
Ouestion'410.73 Provide containment-layout drawings showing the reactor vessel head storage location, the upper guide structure storage stand, the load paths L from t .3 reactor to those locations, and safety-related equipment in the
-vicinity of the load paths susceptible to damage by load handling accidents.
- Resoonse 410.73 T Figure 9.1-19 depicts the load paths of the reactor vessel closure head, the core-support barrel (CSU), and the upper guide structure (UGS) from the reactor vessel to their respective storage areas during the refueling-outage. --The__ designated lead path for each component passes over the reactor! vessel: flange. An analysis has shown that in the event the reactor vessel head or the internals are dropped on the reactor vessel, the core will be maintained in a coolable condition.
Typically, operating ,cocedures prepared by_ the Owner-Operator control the' lift height of the UGS to minimize its clearance with the pool floor and the polar crane is positioned to in ure direct travel from the reactor vessel to the UGS storage area. Additionally, the ICI holding frame is--installed over the seal -table at the operating floor level during fuel handling. This prevents the UGS from moving over the seal table.__ Therefore,1 operating procedures preclude the UGS free being lifted above the seal table or being closer than approximately five feet to the seal table, thereby making seal: table damage a remote possibility.
' However, under a postulated load drop on the seal table, the' seal table would fail resulting la containment pool draindown to the reactor vessel flange area. Since ICI tubes are only restrained laterally and not vertically, the tubes would be bent down to the level of _the reactor vessel flange. Any tube failure, therefore, would in all likelihood be
-at or near-the reactor vessel-flange level which would result _ in a water level within the reactor vessel similar to that prior to reactor vessel head removal.. The accident condition would not be.any more severe than that analyzed for the reactor vessel head drop on-the reactor vessel-flange.
There are no other unprotected safety-related components within the load paths of the reactor vessel head, UGS and CSB. An unprotected component is defined as a' component that is not protected by the pool walls and/or operating . floor. The transfer tube valve will be closed during these handling evolutions L to preclude. water level changes in the fuel build:ng. The following sentence will be added to the end of the first paragraph of
.Section 9.1.4.3.1: "The Owner-Operator's operating procedures will con _ trol the load paths and _ height of the reactor vessel closure head, the
- core support barrel and the upper guide structure above the pool floor.
.* This response was' previous 1v transmit ted by Reference A- 8 I=
l I; J
77a.wp(9075)bh l Qug,stion- 410.103 l l
- a. Section 9.1.1.1 states compliance with the " intent" of I Regulatory Guide 1.13 as a design basis. Considering that Regulatory Guide 1.13 pertains to spent fuel storage, explain what parts of the Guide, and to what extent ~, are met by the new fuel storage Cesign.
- b. -Section 3.1.1.3.3 states that "new fuel storage racks and facilities are qualified as Seismic Category I." Identify the
" facilities" which are so qualified.
- c. Section 9.1.1.2 does not provide sufficient descriptive information on features illustrated in the figures. For
-instance, what is the' function of "L" insert slots and-boxes?
How are the " coll blockers" attached to the structure? What is the equipment in the "new fuel inspection area"? What is their seismic classification?
- d. The new fuel storage capacity changed from 166 in Amendment E
-to 121 'in Amendment I. What is the design basis for the storage capacity of the system?
- e. According to SRP Section 9.1.1, Lhe design of the new fuel storage facility is acceptable if the integrated design is in accordance with, among other criteria, General Design Criteria 61 and 62 of 10 CFR 50, Appendix A. Specific criteria ne'cessary to meet the requirements of GDC 61 and 62 are ANS 57.1 and ANS 57.3 as they relate to the prevention of criticality and to-the aspects of radiological design. Provide information on the extent of compliance of the design to ANS 57.1 and ANS 57.3.
, f._ According to SRP Section 9.1.1, design calculations should show - [ that the storage racks and the anchorages can withstand the L maximum- uplif t forces availcble from the - - lif ting - devices l- . without an increase in k re. A statement in the Safety Analysis that; excessive forces cannot be applied due to the design is acceptable if justification is provided. l-
- g. -It-is the position of the Plant Systems Branch that the vaults and racks of the new fuel storage. facility are to be designed
-to preclude damage from dropped heavy objects. Provide-the design features included in the design which either. preclude the ft11 of heavy objects onto the racks or preclude damage from a drop of the load with the maximum. potential energy.
- h. Reference to -Section 9.1.1.3.1.2.D in Section 9.1.1.3.1.1, regarding potential moderators such as fire extinguishing aerosols, appears to be in error. Should it be 9.1.1 '
.1.2.C?
77a;wp(9075) bh ^
- i. According to SRP Section 9.1.1, the failure of non-seismic Category I systems or structures located in the vicinity of the new fuel storage racks should not cause an increase in k,tg beyond the maximum allowable. Provide analysis that this condition is met or include in your application a commitment to the above condition as a design criterion.
- Response 410.103
- a. Although Regulatory Guide 1.13 pertains to the design of spent fuel storage rackc, it is also used for the design of the new fuel racks. The appliccole portions of the Regulatory Guide that are met are defined by Paragraphs 9. . . .1. A and 9.1.1.1.C.
- b. The "fa :llities" associated with new fuel storage consist of the storage vault and the rack restraint system. The seismic category of other building components associated with handling fuel-assemblies is noted in Table 3.2-1. (see response to NRC RAI 210.1)
- c. The L-insert slots are provided in the wall of the fuel rack cavity (box) to permit the L-insert to be locked to the fuel cavity by its locking tab after it has been installed. The design of the locking tab and slot is such that the L-inserts can be-remotely removed from the fuel racks, if required.
The cell blockers are installed in the fuel racks before the fuel assemblies are placed in the fuel rack and before the pool is flooded. The design is basically two concentric tubes with end restraints that limit the engagement of the tubes in the rack cavity wall (to avoid protrusion into an adjacent fuel rank cavity). The tubes are collapsed, installed into the fuel recA cavity,_ expanded into the holes in the fuel rack cavity hall, then locked together with a captured pin. In this manner zh? cell blockers are positively locked to the fuel racks but can be remotely removed if desired.
*This response was previously transmitted by Reference A-13.
Changes have been made as noted by the bars in the margin. L
77a.wp(9075)bh Odestion 410.107
- a. Evaluate the structural design features of the refueling cavity water = seal _ that would preclude a leak or failure from occurring. Include the possibility of a fuel assembly or other structure dropping on the seal.
- b. If a seal failure / leak occurred, determine the time to lower a fuel assembly below the reactor vessel flange level before unacceptable dose rates from a lowered water level above spent fuel in the reactor core.
- c. For a postulated seal failure / leak, evaluate containment dose rates from a lowered level above spent fuel in reactor core.
- d. For a postulated seal failure / leak, evaluate the following parameters: makeup capacity, emergency procedures, fully loaded spent fuel pool thermal-hydraulle and dose effects including dose rate to someone trying to manually close the transfer tube valve to hydraulically isolate the spent fuel pool from the leak, time to cladding damage without operator action.
Specifically provide the maximum allowable time to isolate the spent fuel pool from the transfer tube and refueling pcol before there are unacceptably high dose rates in the spent fuel pool area and inedequate spent fuel pool cooling due to the level dropping below the minimum NPSH requirement above the elevation of the pool cooling suction inlet piping.
*ReJoonse 410.107
- a. The refueling pool seal is designed to be installed in one piece between the reactor vessel flange and the pool floor.
All fabrication welds will be liquid penetrant inspected prior to installation to ensure adequacy. After the seal assembly has been set in place, it will be permanently attached to the reactor. vessel flange and to an embedment-plate in the pool floor. Penetrations in the seal plate for ventilation and - access to the ex-core instrumentation will be covered by bolead access hatches equipped with double - seals when the poo as flooded. The annulus between the seals will be pressure tested after the hatches have been installed to determine the sealing adequacy.
*This response was previously transmitted by Reference A-13.
Changes have been made as_noted by the bars in the margin.
77t.wp(9075)bh The pool _ seal is designed to withstand OBE displacements without le u age. The pool seal is design'ed to limit potentiel leakage resulting from SSE displacements,. Pool seal inspection will be required as part of the post seismic recovery
-procedure. The-pool seal is also designed to a ccommod ".t.e ,
without leakage, relative displacements between the pool floor and the reactor vessel due to normal plant operation. During refueling operations with the pool flooded, the heavy lift components that pass over the pool seal are the reactor vessel head, the upper guide structure with its lift rig, and the upper guide structure lift rig. Administrative controls as provided in CESSAR-DC Section 9.1.4 require that prior to transfer of heavy loads over the pool seal, the fuel transfer tube valve or the gate between the fuel building transfer system canal and the spent fuel pool shall be closed. This is done to preclude any change to the spent fuel pool water level during a postulated heavy load drop on the pool seal which may result in containment pool draindown. In addition, administrative controls as provided in CESSAR-DC Section 9.1.4 ' preclude the movement of heavy loads over the pool seal if the refueling machine contains a fuel assembly. The refueling machine is designated seismic category II so that it will not fall on the pool seal during the seismic accelerations. The maximum clearance between the bottom of the refueling machine and the top of the pool seal is less than two inches to minimize the h. pact energy for the postulated accident condition of a dropped fuel assembly on the pool seal. The pool seal has been designed so that it will not leak as a result of this impact load.
- b. It has been determined that a 24 square inch opening in the pool seal will result in pool draindown to the reactor vessel flange level in approximately 4 hours without additional water being added to the pool. It has also been determined that present plant systems are capable of mair,taining the pool water level in the event there is a 24 square inch opening in the pool seal. A fuel assembly can be lowered below the reactor vessel flange level from the fully withdrawn position within
-3 minutes,
- c. With the water at the reactor vessel flange level, the radiat"mi level at the pool seal area as a result of the fuel assemblies within the core will not be significantly greater than that with the reactor vessel head in place. The exposed CEA extension shafts will result in an increase in the overall radiation level. However, if it is necessary to do maintenance on the reactor vessel pool seal, temporary shielding can be placed around the extension shafts to reduce the radiation levels in the work area.
g
- 7.7 a. wp ( 9 07 5) bh.
.d. The responses to parts a, b, and c of this RAI address the
-concerns.of this question.
i o i l-l, t
Ouestion 280.1 Provide the fire protection analysis and/or interface requirements to ensure that safe shutdown can be achieved, assuming that all equipment in any one fire area will be rendered inoperable by fire and that re-entry into the-fire area for repairs and operator actions is not possible with i exception of the control room. For the control room, provide the fire protection analyses and/or interface requirements having an independent i alternative shutdown capability that is physically and electrically i independent of the control room. Also, provide the fire protection requirements for redundant shutdown systems in the reactor containment building that will ensure, as much as practicable, that one shutdown division will be free of fire damage. Additionally, also ensure that smoke, hot gases, or the fire suppressant will not migrate into other fire areas to the extent that they could adversely affect safe shutdown capabilities, including operator actions.
- Response 280.1 A fire protection analysis of each fire area is conducted as part of the Fire Hazards Analysis. The System 80+ design basis, as stated in CESSAR-DC Section 9.5.1 (as revised in Amendment I), is t assure the ability to achieve Safe Shutdown following fire in any fire area outside of containment. This includes loss of all equipment in any given area and effects of electrical interaction which may disable equipment outside of l the imediate area. The plant is arranged so that Safe Cold Shutdown can be achieved following fire in any area outside of containment without need for repairs or extraordinary operator action. Emergency shutdown from outside the control room is described in CESSAR-0C Sections 7.4.1.1.10 and 7.4.2.5. Outside of containment, redundant divisions of safety related equipment are separated by three hour fire rated boundaries. In the control complex (and most locati'ns in the Nuclear Annex) redundant safety related divisions of protective electrical channels are separated by three hour fire rated barriers so that loss of all equipment in these areas would not affect either division of safety related equipment required to achieve cold shutdown. Inside containment Engineering Analysis conducted as part of the Fire Hazard Analysis assure that fire at any location which can disable more than one channel of cold shutdown equipment will not affect the ability to achieve cold shutdown using equipment which would not be affected by fire at that location.
l- Smoke control is recognized as an important element of the Plant Fire Protection esign features. In the subsphere area, containment, Fuel Pool building, Reactor Annex and Diesel Genernor building the HVAC l System has smoke control capability by allowing any area to be purged l with 100% outside air. In the control complex, dedicated smoke exhaust fans are provided for the control room and TSC. In addition, a smoke exhaust system is provided for each channel of safety related equipment. A connection to the normal HVAC system intake is used for fresh air supply. Smoke detectors are installed in return air ducts to alarm and annunciate in the control room. Smoke dampers are arranged for remote operation from the control room. L The System 80+ design does not have connectionc (door or ventilation l openings) between redundant safety-related divisions. This further mitigates the possibility that smoke and products of combustion of fire suppression agents will affect redundant safety-related eqvpment.
*This response was previously transmitted by Reference A-8
F 1 77a.wp(9075)bh A
3.0 REFERENCES
A- Letter, Reactor Systems Branch RAIs, T. V. Wambach (NRC) to E. H. Kennedy (C-E) , dated December 24, 1990 A-2 Letter, Reacter Systems Branch RAIs. T. V. Wambach (NRC) to E. H. Kennedy (C-E) , dated January 31, 1991 A-3 Letter, Reactor Systems Branch RAIs, T. V. Wambach (NRC) to E. H. Kennedy (C-E), dated May 13, 1991 A-4 Letter, Reactor Systems Branch RAIs, T. V. Wambach (NRC) to E. H. Kennedy (C-E), dated August 21, 1991 A-5 Letter, Plant Systems Branch RAIs, T. V. Wambach (NRC) to E. H. Kennedy (C-E) , dated October 10, 1991 A-6 Lettec Risk Assessment Branch RAIs, T. V. Wambach (NRC) to E. a. Kennedy (C-E) , dated October 30, 1991 A-7 Letter LD-91-013, E. H. Kennedy (C-E) to USNRC, dated March 15, 1991. 1 A-8 Letter LD-91-014, E. H. Kennedy (C-E) to USNRC, dated March 26, 1991. A-9 Letter LD-91-019, E. H. Kennedy (C-E) to ,$dRC, dated May 6, 1991. A-10 Letter LD-91-024, E. H. Kennedy (C-E) to USNRC, cSted May 16, 1991. A-11 Letter LD-91-062, E. H. Kennedy (C-E) to USNRC, dated L November 27, 1991. A-12 Letter LD-91-071, E. H. Kennedy (C-E) to USNRC, dated December 24, 1991. A-13 Letter LD-92-017, C. B. Brinkman (C-E) to USNRC, dated l February 12, 1992. I A-14 Letter LD-92-008, E. H. Kennedy (C-E) to USNRC, dated January 29, 1992.. m._ .}}