ML20153D495
| ML20153D495 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/31/1988 |
| From: | Campbell D, Dycus F, Ellison B, Farquaharson J, Flanagan G, Guthrie V, Kirchner J, Kirkman J, Paula H JBF ASSOCIATES, INC., OAK RIDGE NATIONAL LABORATORY |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-B-0825, CON-FIN-B-825 NUREG-CR-5021, NUREG-CR-5021-V02, NUREG-CR-5021-V2, ORNL-TM-10604, NUDOCS 8805090170 | |
| Download: ML20153D495 (224) | |
Text
{{#Wiki_filter:- _ _ NUREG/CR 5021 Vol. 2 ORNL/TM-10604/V2 l OAK RIDGE NATIONAL
? LABORATORY.
i USER'S GUIDE FOR PRISIM ARKANSAS NUCLEAR ONE - UNIT 1 , Volume 2 - Program for Regulators H. M. Paula D. J. Campbell I V. H. Guthrie B. C. Ellison J. R. Kirchner F. M. Dycus i J. O. Kirkman J. A. Farquharson G. F. Flanagan i , i Prepared for the . U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research ; Under interagency Agreement DOE 40-550-75 I. 3 I i , j i ~. : I ! t t i I I (PERATED BY - MAP.M MARIETTA ENERGY SYSTEMS. INC. ; ! FDA THE UNITED STATES ! - DEPARTMENT W ENERGY 8805090170 800331 ! 1 PDR NUREG ; 1 CR-5021 R PDR;.
. '3 _ _ _ _ . _ . , - . - _ . _ . _,
NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. e l l Avalable from i Supenntendent of Documents . U.S. Goverrrnent Printing Offee Post Offee Box 37082 Washington, D C. 20013-7982 e and National Techncal information Serwce Spnngfeld, VA 22161
NUREG/CR-5021 Vol. 2 ORNL/IM-IO604/V2 Distribution Category RG Engineerin6 Physics and Mathematics Division USER'S GUIDE FUR PRISIM ARKANSAS NUCLEAR ONE - UNIT 1 Volume 2 Program for Regulators D. J . Campbell
- H. M. Paula*
e V. H. Guthrie* B. C. Ellison* J. R. Kirchner* F. M. Dycus* J. Q. Kirkman
- J . n. Farquharson*
G. F. Flanagan NRC Monitor: J. C. Glynn Severe Accident Issues Branch Program Manager: G. F. Flanagan Oak Ridge National Laboratory Principal Investigators: G. F. Flanagan Oak Ridge National Laboratory D. J. Campbell JBF Associates, Inc. Prepared for the
. U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Under Interagency Agreement DOE 40 550 75 ,
NRC FIN No. B0825 Date Published: March 1988
*JBF Associates, Inc.
Knoxville, TN 37932 e Oak Ridge National Laboratory
, Oak Ridge, Tennessee 37831 operated by MARTIN MARIETTA ENERCY SYSTEMS for the DEPARTMENT OF ENERGY under Contract No. DE-AC05-840R21400 i
r 9
iii o e ABSTRACT This user's guide is a two-volume document designed to teach NRC inspectors and NRC regulators how to access probabilistic risk assessment information from the two Plant Risk Status Information Management System (PRISIM) programs developed for Arkansas Nuclear One - Unit One (ANO-1). Volume 1 describes, how the PRA information available in Version 1.0 of PRISIM is useful for planning inspections. Using PRISIM, inspectors can quickly access PRA information and use that information to update risk analysis results, reflecting a plant's status at any particular time. Voluma . 2 describes how the PRA information available in Version 2.0 of PRISIM is useful as an evaluation tool for , regulatory activities. Using PRISIM, regulators can both access PRA information and modify the information to assess the impact these changes may nave on plant safety. Each volume is a stand-alons document, e 0 0
v CONTENTS Section ZASR ABSTRACT........................................................ iii LIST OF FICURES................................................. vii LIST OF TABLES................ ................................. x PREFACE......................................................... xi e
- 1. INTRODUCTION......................... ...................... 1 L
- 2. PRISIM USER CONTR0LS........................................ 3
- 2.1 Controls Associated with Accessing Information in the PRISIM Modules......................................... 3 t
, 2.1.1 Controls Associated with the PRISIM Control l Screen......................................... 3 2.1.2 Controls Associated with the PRISIM DBM Routine 6 i
2.1.3 controls Associated with the PRISIM PRA Data Tables......................................... 8 2.1.3.1 Component Failure Event and Operator Failure Event Tables.................. 12 2.1.3.2 Initiating Events Table............... 16 2.1.4 Controls Associated with the PRISIM PRA Update Routine........................................ 19
. 2.2 Controls Associated with Obtaining Hard Copies of ,
PRISIM Screens.. ...................................... 20 ,
- 3. CATECORIES OF SAFETY-REIATED INFORMATION (AND SAMPLE
- COMPUTER SESSIONS).......................................... 23 3.1 Updated PRA Information................................ 23 3.1.1 Component Failure Events....................... 25 ,
1 3.1.2 Initiating Events.............................. 34 3.1.3 Operator Action Events......................... 40 3.2 Preprocessed Information............................... 50 i 3.2.1 Dominant Accident Sequences.................... 50 3.2.2 Safety.Related Systems......................... 55 , 3.2.3 Safety Related Subsystems...................... 60 3.2.4 Safety-Related Components...................... 60 3.2.5 Support System Interfaces...................... 66
. 3.2.6 Operator Actions............................... 75 l 3.2.7 Component Failure Data......................... 80 ; 3.2.8 System Testing / Surveillance.................... 80 ,
l 3.2.9 Limiting Conditions for Operation.............. 87
- 3.2.10 System Comparisons for Selected B&W Plants..... 87 i
r f 6 6 L
vi CONTENTS (continued) Section f.aE2 APPENDIX A - OBTAINING HARD COPIES OF CRAPHICS SCREENS PROVIDED BY PRISIM.......................................... A1 APPENDIX B - DIAGRAM OF THE PRISIM INFORMATION HIERARCHY........ B-1 ' APPENDIX C - QUANTIFICATION OF FAILURE EVENTS................... C-1 e. O e e
vii LIST OF FIGURES F1rure f.a&R 2.1 PRISIM Interfaces.......................................... 4 2.2 PRISIM Control Screen (Initial Entry into the Program)..... 5 2.3 PRISIM Control Screen (Following a Session)................ 5 2.4 Sample Text Menu........................................... 9 2.5 Sample Schematic Menu...................................... 9 e 2.6 Component Failure Event Data Table......................... 13 2.7 Initiating Events Data Table.... .......................... 17 2.8 PRISIM Control Screen (Picture Number Check)............... 22 3.1 Type 1 o f Safe ty Related Information. . . . . . . . . . . . . . . . . . . . . . . . 28 3.2 Options Available for Updating PRA Information. ........... 28 3.3 Paths Available for Accessing Component Failure Event Probability Data........................................... 29 3.4 Safety-Related Systems for Which Schematics Are Available.. 29 3.5 Emergency Feedwater System Schematic....................... 30 3.6 EFS Turbine-Driven Pump P7A Fails (FRA Data Table before Data Modification)......................................... 30 3.7 EFS Turbine Driven Pump P7A Fails (PRA Data Table after Data Modification)......................................... 32 3.8 Emergency Feedwater System Schematic....................... 32 3.9 Updated PRA Results................ ....................... 33 3.10 Ranking of Failure Events.................................. 33
- 3.11 Updated PRA Results........................................ 35 3.12 PRISIM Lontrol Screen...................................... 35
. 3.13 Emergency Feedwater System Schematic.................... .. 36 3.14 Safety-Related Systems for Which Schematics Are Available.. 36 3.15 Emergency AC Electrical System Schematic................... 37 3.16 Diesel Generator 1 Fafis (PRA Data Table before Data
- Modification).............................................. 37 3.17 Diesel Generator 1 Fails (PRA Data Table after Data Modi-fication).................................................. 38 3.18 Emergency AC Electrical System Schematic................... 38 3.19 Updated PRA Results........................................ 39 3.20 Ranking of Core Damage Scenarios........................... 39 3.21 Types of Safety-Related Information. . . . . . . . . . . . . . . . . . . . . . . . 41 3.22 Options Available for Updating PRA Information.... . ..... 41 3.23 Selection List for Dominant Accident Sequence Initiating Events (FRA Data Table before Data Modification)........... 42 3.24 Selection List for Dominant Accident Sequence Initiating
- Events (PRA Data Table after Data Modification)............ 42 3.25 Options Available for Updating PRA Information............. 43 3.26 Updated PRA Results......................... .............. 43 3.27 Ranking of Core Damage Scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . , 44
. 3.28 Types of Safety Related Information........................ 46 3.29 Options Available for Updating PRA Information... ......... 46 3.30 Operator Errors W5ose Probabilities May Be Changed......... 47 l
viii LIST OF FIGURES (continued) figure Pano 3.31 Failure to Initiate HPIS (PRA Data Table before Data Modification)............. ................................ 47 3.32 Failure to Initiate HPIS (PRA Data Table after Data Modi-fication)............................................ ... . 48 3.33 Operator Errors Whose Probabilities May Be Changed......... 48 . 3.34 Updated PRA Results........................................ 49 3.35 Types of Safety-Related Information. . . . . . . . . . . . . . . . . . . . . . . . 52 3.36 Categories of Preprocessed PRA Information................ 52 3.37 Dominant Accident Sequence List....................... .... 53
- 3.38 Description of Accident Sequence #2. . . . . . . . . . . . . . . . . . . . . . . . 53 3.39 Dominant Scenarios for Accident Sequence #2................ 54 3.40 Recovery Information for Scenario 2-1, Accident Sequence #2 56 3.41 Types of Safety-Related Information........................ 58 3.42 Categories of Preprocessed PRA Information................. 58 3.43 Information Available for Safety-Related Systems........... 59 l 3.44 Front Line Systems Ranked by Risk Reduction Icportance..... 61
! 3.45 Types of Safety-Related Information........................ 62 l 3.46 Categories of Preprocessed PRA Information................. 62 3.47 Information Available for Safety-Related Subsystems........ 63 3.48 Subsystems Grouped by Risk Significance Importance......... 63 3.49 Types cf Safety-Related Information.................... . . 67 3.50 Categories of Preprocessed PRA Information................. 67 3.51 Information Available for Safety-Related Components........ 68 , 3.52 Options Available for Selecting Out-of-Service Components. . 68 j 3.53 Safety Related Systems for Which Component Information Is t Available.................................................. 69 3.54 Emergency Feedwater System Schematic............ ..... .... 69 * , 3.55 Available Information Based on Pump P7A Being Out of Service 70 l 3.56 Most Likely Single Failure Events for the Emergency Feed- ! water System when Pump P7A Is Out of Service. . . . . . . . . . . . . . . . 70 3.57 Support System Interfaces for Emergency Feedwater Pump P78.. 71 3.58 Type s o f S a f e ty-Rela te d Info rma tion. . . . . . . . . . . . . . . . . . . . . . . . 73 3.59 Categories of Preprocessed PRA Information.......... ....... 73 3.60 Types of Support System Interface Information Available. ... 74 3.61 Support Systems for Which Information is Available........ , 74 3.62 Emergency AC Power System Component / Function List........... 76 3.53 Components Dependent on 480 V AC Bus B5.. ..... ............ 76 3.64 Types of Safety Related Information......................... 77 3.65 Categories of Preprocessed PRA Information. . . . . . . . . . . . . . . .. 77
- 3.66 Information Available for Operator Actions.................. 78 l 3.67 Operator Recovery Actions Grouped by Risk Significance Importance...................................... .......... 78 ,
3.68 Recovery Information for Restoring EFS Motor Driven Pump... 81 3.69 lypes of Safety Related It .ormation.......... ..... ........ 32 3.70 Categories of Preprocessed PRA Information.................. 82
ix 4 LIST OF FIGURES 1 (continued) Firure fagg, 1-3.71 Component Failure Data Available............................ 83 i 3.72 Component Categories for- Which Plant Specific LERs Are Available................................................... 83 l
. 3.73 LERs Involving Turbine Driven Pumps......................... 84 3.74 Types of Safety Related Information.................. ...... 85 3.75 Categories of Preprocessed PRA Information.................. 85 !
3.76 Safety Related Systems for Which Information Associated with
- Periodic Testing / Surveillance Requirements Is Available..... 86 .
3.77 Periodic Testing / Surveillance Requirements--High Pressure I nj e c t i o n S y s t e m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 3.78 Types of Safety Related Information......................... 89 i 3.79 Categories of Preprocessed PRA Information.................. 89
+ 3.80 Systems for Which Limiting Conditions for Operation Infor- i mation Is Available......................................... 90
[ 3.81 Emergency Feedwater Initiation and Control System Available l . Information Regarding Limiting Conditions for Operation..... 91
+ 3.82 EFICS Main Steam Line Isolation Channels Limiting Conditions for Operation............................................... 91 ;
.j 92
- 3.83 Required Operator Action.................................... i 3.84 Types of Safety-Related Information......................... 94 ,
l Categories of Preprocessed PRA Information.................. 94 3.85 ! 3.86 System Configuration Comparisons Which Are Available for l' i Selected Babcock and Wilcox P1 ants.......................... 95 3.87 High Pressure Injection System Configurations Used in i Selected Babcock and Wilcox P1 ants.......................... 95 I . B.1 Diagram of the PRISIM Information Hierarchy................. B 2 ' ! C.1 Unavailability Function for Components Periodically Tested.. C 4 . i ' C.2 Unavailability Function for the Example Pump................ C 7 i I l 1 i 1 t i e l i I I i l 1 l l l 4 l
X LIST OF TABLES Table Pa r.e 2.1 Control Screen Keyboard Controls ........................... 7 2.2 DBM Routine Keyboard Controls .............................. 10 2.3 PRA Data Table Keyboard Controls ........................... 11 2.4 PRA Update Routine Keyboard Controls ....................... 21 3.1 Summary of Subevent Input Va1ues............................ 26 l 3.2 Measures of Importance and Their Interpretation............ 57 l l l l l l 1 e I i l I i l I l l 1 t l l ! l l
xi PREFACE Volume 2 is the user's guide for Version 2.0 of the Plant Risk Status i Information Management System (PRISIM) designed for the Arkansas Nuclear i One - Unit 1 (ANO 1) Nuclear Power Plant. PRISIM is a computer program that operates on a personal computer and makes probabilistic risk assessment (FRA) information and information from other safety studies useful to regulators as an evaluation tool.
- This version of PRISIM is based on a Level 1 PRA of ANO-) performed by JBF Associates Inc. The Interim Reliability Evaluation Program (IREP) 4 study of ANO-1 was used to the extent possible as a basis for the PRA.
However, changes in the plant design and operating procedures, as well as improvements in the state of the art of PRA methodology, led to significantly different results for the new PRA and at times precluded use of the IREP. The PRA results in PRISIM are based on core damage frequency. PRISIM does not incorporate the results of assessments of plant damage, containment responses, or public health consequences. Only point estimate PRA results are presented in PRISIM; uncertainty bounds are not included. Many of the results presented in PRISIM do not reflect the effects of operator "ecovery actions. Recovery actions are actions operators can take to maintain core cooling during accidents involving failures of the designed emergency cooling systems. These actions include restoring
. offsite power, restoring failed equipment to service, or providing temporary support services to safety-related equipment during accident conditions. The effects that recovery actions would have on plant risk are presented separately in PRISIM, This guide is not an attempt to explain all possible uses of PRISIM, to provide the results that would be obtained for all possible applications, or to interpret how results provided by PRISIM can affect plant risk. The purposes of this guide are to teach NRC regulators how to access PRA-related information and to define, in very general terms, how this information is useful, i
1 e i s
*G. J. Kolb et al., _ Interim Reliability Evaluation Program-Analysis of the Arkansas Nuclear One - Unit 1 Nuclear Power Plant, U.S.
Nuclear Regulatory Commission, NUREG/CR-2787, Washington, DC, June 1982. t L
USER'S CUIDE IOR PRISIM ARKANSAS NUCLEAR ONE - UNIT 1 Volume 2 Program for Regulators
- 1. INTRODUCTION
. This guide describes the use of the Plant Risk Status Information Management System (PRISIM), a personal computer program designed for the Arkansas Nuclear One - Unit 1 Nuclear Power Plant (ANO 1), that will help you gain rapid access to PRA-related information useful for
- regulatory activities. PRISIM is a user-friendly and decision oriented program designed to run on an IBM PC/XT or AT with a 640 K memory, a math coprocessor, a 20 megabyte hard disk, an Enhanced Graphics Adapter, and an Enhanced Graphics Display. It is compatible with DOS Version 2.1 and any 1=ter versions of DOS. The program is designed with protective features that prevent the user from damaging the program or destroying the data base.
Using PRISIM for regulatory activities offers several advantages over other media that can be used to present PRA related information. The following are two of the main advantages:
- 1. PRISIM will allow you to quickly access needed information without having to sort through a lot of irrelevant information.
- 2. PRISIM can be used to update safety analysis results, reflecting, for example, a change in technical specifications.
. To operate PRISIM, you do not need to be an expert in the use of PRA methodology; the program presents information without using PRA jargon.
Also, operating the program does not require previous computer experience. The information offered by PRISIM can be easily accessed by making selections from a series of menus on the screen. There are two major portions of the PRISIM computer program that provide you with useful PRA-related information: the Data Base Management (DBM) Routine and the PRA Update Routine. The PRISIM data base contains preprocessed information that is useful regardless of the plant's status. Examples of preprocessed information available from PRISIM include system piping and instrument diagrams, summaries of selected plant licensee event reports (LERc), information on equipment l realignment for testing or maintenance, and information on operator recovery actions. The PRA update routine allows you to (1) specify the plant's status (operating configuration) at any particular time and/or (2) modify the PRA parameters and receive updated PRA information. This portion of the , program permits evaluating the impact on plant safety associated with l
2 changes made (1) in the PRA assumptions, (2) in the Technical Specifications, or (3) in the operating status of the plant (e.g., planned shutdown of a specific component or system because of testing or maintenance). This guide provides the novice user with all the information necessary for successful operation of PRISIM. If you are using PRISIM for the first time, you should first scan through Section 2 to become familiar with the keyboard controls you will use to access PRISIM information. Next, read Section 3 carefully to gain an understanding of the types of , information available in PRISIM and how this information can be accessed. We also recommend you work through the sample computer sessions in Section 3 at your keyboard to become more proficient with the use of your keyboard controls. . Three appendixes to this guide provide additional information on the use of the PRISIM computer program. Appendix A explains how to obtain hard copies of the graphics screens provided by PRISIM. Appendix B details the information provided in this version of the PRISIM program. Appendix C describes the models used to quantify failure events in the ASO 1 PRA; these same models are used in PRISIM when PRA parameters are changed. e 0
3
- 2. PRISIM USER CONTROIS ,
The PRISIM program consists of four distinct modules: (1) the control screen, (2) the DBM routine, (3) the PRA t.ata tables, and (4) the PRA update routine. You enter the program through the control screen and, from that point, enter the DBM routine. The DBM routine allows you to view preprocessed PRA information valuable for monitoring plant safety. It also allows you to access the PRA data tables in order to modify PRA input information and then access PRISIM's PRA update routine to a receive information on the plant's safety status when specific PRA modifications are considered. Interfaces associated with the four modules of the PRISIM program are illustrated in Figure 2.1. e During any PRISIM session (that period of time between entry into the prograa and escape from the program) and while in any of the four PRISIM modules, you crin control the information provided by PRISIM through the use of several function keys (user controls) on the computer keyboard. Because it is not feasible to explain in detail all of the actual steps (and hence the user controls) that would be involved in all of the various applications of PRISIM. this section of the user's guide provides an overview of PRISIM user controls, giving you an idea of how you can "move about" in the PRISIM program. 2.1 Controls Associated with Accessinn Info rma t ion in the PRISIM Hodules 2.1.1 Contro1r Associated with the PRISIM Control Screen The first screen presented by PRISIM is the control screen (Figure 2.2). It offers three options (Begin, Escape, and Load) when you first enter the pcogram. Pressing the letter "B" on your computer keyboard
. allows you to begin your new session, which will take you into the DBM routine. Pressing the letter "L" allows you to call up and load (make available for playback) a previously saved session. Pressing the "Esc" key takes you out of the program.
A slightly different version of the control screen (Figure 2.3) appears any time you load a previously saved session or end a PRISIM session; ' at this time, the control screen provides three additional user control options. Pressing the letter "R" allows you to resume the current session. Pressing the letter "P" lets you play back the current session. (And pressing the "+" or " " key allows you to decrease or increase the speed at which the session screens are played back.) Pressing the letter "S" lets you save the current sassion. The "Resume" option will allow you to do two things. First, the "Resume" option will return you to the program anytime you exit the program to obtain a hard copy of a PRISIM screen. (The use of the "Resume" control for this purpose is discussed in detail in Section 2.2 of this guide.) Second, it will let you resume a session after making PRA changes. That is, after you i, ave changed PRA input parameters and obtained updated PRA results, you must return to the , i I
l 90 TURN To l -> CON TROL toAo
->g stsgoN upoa TE PRA j scet[N l
I -> PRA > UPD A TE -
! REWS , ROU TINE '
V:Ew T l BE GrN L>l DC9 FED _p _ SEs90N l RESULTs CON TROL DA A BAK l wE w SCRED4 -> Mm.J AGEMEN T -- > ME PROCEs5E D CHANGE ROU TINE l -> COUPONENT __ l Ptsuvc . rAr l mon,tuut g,o rts j scssoN a l csANcc PRA f,* I l 3
-> PRA > DATA > FAtuRE l sa vt D'7^
l TABLES l PROBAPAirES scssoN l l cnuc,t l _, PL Av Ei Ac<
-> unAtNc -
sEs90N l rRtaucNcits Figure 2.1 PRISIM Interfaces
. . . e
~
( 3 Iyte Binder RUN TIME DATA BASE CONit0LLER Copyright, Eastern Technical Engineering Corp.,1985 87 Rev 2.5 seriet 8 000006 LicPROTO TYPE C0WTt0L $CtEEN 8 egfn new session L oed Session ESC to exit a e Brede Supported 50, Pg 1 ( ) Figure 2,2 PRISIM Control Screen (Initial Entry into the Program) [ \ tyte Binder EUN TIME DATA BASE CONTROLLER
- Copyright, Eastern Technical Engineering Corp.1985 87 Rev 2.5 Sertal # 000006 Lle:PROTO 1fPE C0NTt0L $CREEN 8 egin new session L oed Session 5 ave Session t estre this Session P Lay Back Current Session +- Playback Speed 1 Sec.
ESC to exit Brede Supported Green Override
$1, Pg 2,P 35 Y )
l Figure 2.3 PRISIM Control Screen (Following a Session) l
6 control screen. Then the "Resume" option will send you back to the updated PRA information menu (while retaining in memory the modified PRA data). You can then make additional data changes. If you want to reenter the program with the PRA data reset to their nominal values, use the "Begin" option. The "Playback" option will let you replay the DBM routine screens viewed in the current session in the same order in which they were originally viewed. PRA update routine screens, however, will not be displayed. As previously mentioned, you can select the speed at which " the screens are played back using the "+" key or the key to decrease or increase the screen viewing time. The viewing time options are rapid scan (less than one second per screen); l,2, 3,4, 5, 10 , 15 , 20 , 25 , or 30 second scan; and "manual step" scan (a scan that is controlled by the user with the return, or enter, key. While you have the freedom to select the speed at which you view screens using the "Playback" option, once you have selected this option you cannot stop or again change the speed of the playback; you must view every screen before you can change the screen viewing time. The "Save" option will let you store on disk the current session for future playback. This option will not save changes made to PRA input parameters, nor will it save FRA vpdate screens. PRISIM will ask you to identify the session to be saved using a unique, one digit alphanumeric identifier (e.g., 1, 2, 3 . . . , a, b, c ...) and a password so that this session may be loaded at some future date. If you specify an identifier that has already been used, PRISIM will give you the option of writing over the previously saved session or saving . the current session using a different identifier. Table 2.1 provides definitions of the PRISIM keyboard controls associated with the PRISIM Control Screen. Use this table for quick
- reference to the controls discussed in this section.
2.1.2 controls Associated with the PRISIM DBM Routine PRISIM's DBM routine provides rapid access to several types of preprocessed information. To get to the information of interest in the DBM routine, you use a cursor, which is controlled by the "t" and "4" keys, to make a series of selections from menus shown on the screen. When the curser is positioned on the selection of interest, you sittply press the return key and the PRISIM data base manager will find and quickly display the appropriate screen. , Three function keys allow you to backtrack in the DBM routine- These functions are: PREVIOUS - Pressing the "F2" key calls up the previously displayed screen IAST BRANCH POINT - Pressing the "F3" key calls up the last screen
- that displayed multiple menu options I
l E _--
7 Table 2.1 control Screen Keyboard Controls Keyboard Symbol Function of Key i e B Begin New Session - Enters the program and begins a new
- session Esc Escape - Exits the program L Load Session - Calls up and loads a previously saved session P Playback - Plays back the current session R Resume Resumes a session in which PRA data was changed or returns to the program after checking a picture number to obtain a hard copy S Save Session - Saves the current session Increaces the playback speed
) i
+ Decreases the playback speed O
e G I
8 MASTER - Pressing the "F1" key calls up the PRISIM Master Menu
- PRISIM uses both text menu screens (Figure 2.4) and schematic meau screens (Figure 2.5) to help you obtain PRA related information.
Schematic menus are included in the program to give you an alternate way of selecting components of interest. On a schematic menu, you select the component for which you want to obtain information by moving the carsor until the component is highlighted on the system schematic . menu and then pressing the return key. On a text menu, you obtain the desired information by moving the cursor to the desired selection and then pressing the return key, Table 2.2 provides definitions of the PRISIM keyboard controls associated with the PRISIM DBM routine. Use this table for quick reference to the controls discussed in this section. 2.1.3 Controls Associated with the PRISIM PRA Data Tables Two basic types of PRA data tables are included i t, this version of PRISIM. These data tables allow you to alter quantitative PRA input values in order to determine the effect these alterations have on plant safety and to receive other information based on the updated plant status. The purpose of this section is to familiarize you with the PRA data table controls available. Section 2.1.3.1 describes the controls used to modify component failure , events and operator failure events in the FRA data tables. These tables can be accessed from the PRISIM DBM routine by cursor selection from system schematics, component lists, and operator failure event lists.
- Section 2.1.3.2 describes the controls used to modify accident sequence initiating event frequencies. Initiating events are disruptions to the normal operation of the plant that cause a rapid shutdown of the plant, or a need to trip the plant, thus challenging the safety systems designed to remove the decay heat being generated by the reactor core.
This data table can be accessed by cursor selections from a PRISIM DBM routine menu. Table 2.3 provides definitions of the PRISIM keyboard controls associated with PRISIM PRA data tables. Use this table for quick reference to the controls discussed in this section. , ,
- l 1
*The Master Menu is the first selection menu presented to the PRISIM user upon entry to the DBM routine, 1 I
i 1 I I 1 l
9 ( h CATEGORIES OF PkiPROCESSED PRA INFORMATION.
- 1. Deafnant core damage sequences
- 2. Safety related systems
- 3. Safety related subsystems
- 4. Safety related ompenents
- 5. Support system interfaces
- 6. Operator actions
- 7. Conponent f alture data
- 8. Syst. testing / surveillance
- 9. Limiting' conditions for operation
- 10. S) em cornparisons for selected B&W plants N ]
Figure 2.4 Sample Text Menu ( ) EVE 8iCO.CY AC ELECTRtCAL SYSTCU SCHEW AN A *' @ "' , , ,
~=r +> - > ., L ., ', .;. 6 +> =r 43 4 4 5 4 .4 t
i) t t i) t
? t) t t) t) 6 i i
i i i I I / -/ 22 1,
, .s -- as uu f
+ 6> e
) -
f f f f [ f[ 2 [ .,,
.3 r,
.ll ll j.f 7f~~h ._ / 4 , , ,
?
cm. _ i 2=llll u,. L . ,
,) * > .. !L.
)
9 L
)* .
g s, situ utw t..o ci en e N } l i i Figure 2.5 Sample Schematic Menu
1 10 l l Table 2.2 DBM Routine Keyboard Controls ; 1 l
)
Keyboard i Symbol Function of Key l
= l l
F1 Master - Calls up the Master Henu a F2 Previous - Calls up the previously displayed screen l l 1 F3 Last Branch Point - Calls up the last screen that displayed l multiple menu options l Esc Escape Calls up the PRISIM control screen 1 Prtsc, Prints the currently viewed text screen l Shiftb d Return - Selects the indicated menu option t Moves the cursor in the "reverse" directivn i
- l 4
Moves the cursor in the "forward" direction , l I 1 l aThe Master Menu is the first selection menu presented when you enter the program. ' b 1 These keys must be depressed simultaneously when you want to print l a hard copy of a text screen. [l 1 l e e
11 Table 2,3 PRA Data Table Keyboard Controls Keyboard Symbol Function of Key F2 Previous - Returns you to the schematic or selection menu in the DBM routine from where you entered the data table Prints the currently viewed text screen PrtSc.a Shift Home Moves the cursor to the first entry in a data table End Moves the cursor to the last entry in a data table PgUp Returns you to the previous page of entries in a data table PgDn Returns you to the next page of entries in a data table 4 Returns an entry to its data base value b Sequences through model types c Se gces through options for setting the EVENT PROBABILITY va Ctrl, W a Resets all data in the PRA data table to their data base values
~
t Moves the cursor up t Scrolls up the list of entries if the cursor is at the top of the screen 4 Moves the cursor down 4 Scrolls down the list of entries if the cursor is at the bottom of the screen
- Moves the cursor to the right
- Moves the cursor to the left aThese two keys must be depressed simultaneously.
b This key function is applicable when the cursor is positioned in any PRA data table field except MODEL and ENENT PROBABILI U fields. cThis key function is applicable when the cursor is positioned in the MODEL field of a PRA data table, d inis key function is a an EVENT PROBABILITY field, pplicable when the cursor is positioned in l i l i
12 2.1.3.1 Component Failure Event and Operator Failure Event ~ Tables This section describes -the user controls associated with the PRA data tables used to alter failure pri babilities ' associated with component failure events and operator failu'.e events (hereinafter called events). , Each event is presented in a r,eparate PRA data . table along with the descriptions and failure data for each of the subevents associated with the event selected. A subevent is one of the specific component failure modes that can cause the occurrence of the event. Figure 2.6 - " is an example of a PRA data 1 table for . a ' typical event with several associated subevents (component failure modes). As shown in the figure, each subevent has an associated probability'of
- failure (P). The failure probabilities of the subevents are summed c.ogether (OR logic) to obtain the probability of failure for the event selected. For each subevent, the probability value is either assigned (e.g., subevent . "EFS Motor Driven Pump P7B fails to start" on demand) or derived from a model (e.g., subevent "EFS Motor-Driven Pump P78 fails to run" after a successful start) . Three models are available (Fault Exposure Time (FET), Test Interval (TI), or Downtime (DT]).
Each model uses a separate equation to combine failure data specified for a subevent and calculate a subevent probability value. In order ~to calculate a subevent probability the user must specify a RATE value (failure rate), a TIME value (downtime, test interval, or mission time), and an appropriate MODEL option. Appendix C describes the models that are available for quantifying the probability of each subevent occurring. . User controls associated with the PRA data tables can be classified into two groups. The first set of user controls are those coerols , used to move the cursor within the PRA data table. The second s.t.t are
- those controls used to alter or modify values in the PRA data table '
fields. The cursor can be moved to any PRA data table field by depressing the appropriate "t, 4 , * , o r *" keys. (Note that only the t text field associated with the User Defined Subevent can be accessed ; with the cursor.) In addition, depressing the "Home" key will move the , cursor to the failure probability field associated with the first subevent listed in the PRA data table; depressing the "End" key will - move the cursor to the EVENT PROBABILITY field contained on thu last line of the table. 5 Some PRA data tables require several screen pages to display the entire- , list of associated subevents. The "PgDn," "PgUp," and scroll features are available for moving the cursor from one screen page to another or from one subevent to the preceding or to the next subevent in the list. Depressing the "PgDn" key will replace the current display of subevents a with the next page of subevents. Depressing the "PgUp" key will replace the current display of subevents with the previous page of subevents. I
i 13 ( h EFS MOTOR e*tVEN PUMP P7B FAILS RATE . TIME , SUBEVENT DESCRIPTION (/HR) -(HR). MODEL P EFS Motor Driven Ptarp P7B f alls to 1.0E 03 .l start
- EFS Motor-Driven Punp P78 f alls to 3.0E 05 4.0E+00 TET 1.2E 04 ,
run j EFS Motor Driven Pum P78 maintenance 3.1E 05 7.0E+00 DT 2.2E 04
,a EFS Motor Driven Purp P78 test 0.0E 00 cable A311 faits open circuit 3.0E 06 7.2E+02 fl 1.1E 03 Circu;t Breaker A311 falls to 1.0E 03 transfer (F2]
- PREV!r'US T HORE T EVENT PROBABILITY = 5.4E 03 y Figure 2.6 Component Failure Event Data Table
[ ) EFS MOTOR DRIVEN PUMP P7B FAILS e ' RATE TIME SUBEVENT DESCRIPil0N (/HR) (HR) MODEL P Circuit Breaker A311 control 2.0E 03 circuit faits User Defined Subevent i l t I (
- EVENT PROBABjlliy = 5.4E 03
( (F21 - PREVIOJS A MORE A j Figure 2.6 (continued) l ,
l 14 The scroll feature allows you to scan through the list of subevents in a PRA data table, one subevent at a time, in either the up or down direction. As stated earlier, depressing the "4" key moves the cursor, line b; line, down the displayed page of subevents. Once the cursor is at the bottom of the displayed list of subevents, continu'.ig to depress the "4" key will allow you to scroll downward through the remaining subevents in the list, one subevent at a time. Depressing the "t" key coves the cursor up the displayed page of subevents. Once the cursor is at the top of the displayed list of subevents, continuing to depress the "t" key will allow you to scroll upward, one subevent at . ; a time, through those subevents preceding the displayed list in the PRA l data table, , l Data contained in the PRA data tables can be easily altered or modified
- by using a combination of the cursor controls previously described and normal keyboard entries. The PRA data table fields that contain i numerical values (RATE, TIME, or P) can be changed by executing the I following steps:
1
- 1. Position the cursor on the selected data field by using the cursor controls described previously. I 1
- 2. Use the keyboard to enter a new numerical value in the field.
- 3. Move the cursor to a new data field or exit the data table to enter the new numerical value in the data base.
)
i Note that although numerical values may be entered in a variety of . formats from the keyboard, they will always be displayed in scientific notation with two significant figures. Also, numerical values that are inapprcpriate, or meanir.gless , will produce an error message when you attempt to enter these, values.
- l The PRA data table field that contains the MODEL types available for l l
calculating subevent failure probabilities can be changed by performing the following steps:
- 1. "osition the cursor on the selected MODEL field by using the cursor controls described previously.
- 2. Find the model desired (FET, TI, DT) by depressing the "RETURN" key on the keyboard.
- 3. Move the cursor to a new data field or exit the data table to enter the new model in the data base.
If RATE, TIME, and MODEL values exist for a subevent, the PRA data , table will automatically calculate the subevent failure probability based on these selections. The automatic calculation feature is terminated (overridden) any time the user directly enters a value in
15 the subevent failure probability field as described above. In this-case, the RATE, TIME, and MODEL fields associated with the subevent become blank. The EVENT PROBABILITY field at the bottom of each PRA data table. display is a special field, Normally, the value in . this field is automatically, and continuously, . calculated by summing the currently displayed .subevent failure probabilities -in the table. However, you may override the automatic summing feature by executing the following
- steps:
l'
-1. Position the cursor on the EVENT PROBABILITY field by' using the cursor controls described previously ("END" key) . Note that the l cursor is positioned on this field when a data table . is first entered. ;
- 2. Se' et a value of 1.0 (basic event failed) or 0.0 (basic event
..rfectly working) by depressing the return key on the keyboard, or enter a numerical value of your choice by using the input keys on the keyboard.
- 3. Exit the data table or strike any cursor control key to enter the new EVENT PROBABILITY value in the data base.
9 When the EVENT PROP ABILITY field is removed from the automatic summing mode (by direct entry of an EVENT PROBABILITY value) all data fields 4 associated with the e tbevents in the data table become blank so that
. misleading information is not presented. Also, access to the subevent f data fields in a table is prohibited as long as the EVENT PROBABILITY remains in the override (non summing) mode. (To leave the override mode, press the return key.) Inappropriate, or meaningless, numerical i values that are directly specified for the EVENT PROBABILITY field will produce an error message when entry is attempted.
As described in Section 2.1.1, a user's control associated with the PRISIM control screen can be used to reset all previously altered PRA N data to their nominal (original) data base values (i.e., the "Begin" option can be used to reset all altered PRA data to their nominal : value). If you do not use.the "Begin" command, the last changes you : have made to the PRA data will be retained as the current data base value and will remain so until the PRISIM session is terminated. If you are modifying data within a PRA data table and make an error, PRISIM provides you with the capability to reverse those modifications. This data restoration capability can be utilized for all data
- associated with a PRA data table (event) or for data associated with a specific subevent in the table. PRA data associated with an individual
. subevent -sy be restored to their current data base values by executing the following steps:
- 1. Position the cursor on any field associated with the subevent of interest.
- .- - -_ . - . - ~ . . - . _ . . _
e 16
- 2. Depress the return key.
- 3. Exit the data table or strike any cursor control key to restore the '
indicated subevent's PRA data. All PRA data associated with an. event may'be restored'to their current data base values by simultaneously depressing-the control key and the return key. The cursor need not be in a specific posicion to utilize the "control / return" command. Since modifications made in a PRA data table are automatically stored in the current data base.when you exit . the table, the data restoration capabilities described above must be utilized before' leaving the PRA data table. The last subevent in each data table is left undesignated so that you can supply a description and data for a subevent of your own choosing. The cursor allows you to change the descriptive text of this undesignated subevent (by moving the cursor to this field and typing
-the desired description) and permits you to change or specify RATE data, TIME data, MODEL values, and probability data in the same manner as previously described.
The "F2" (PREVIOUS) function allows you to exit the data table and returns you to the schematic or selection list in the DBM routine from ; which the failure event choice was originally made. 7 2.1.3.2 Initiating Events Table i This section describes the user controls associated with the PRA data , i table used to alter the accident sequence initiating event frequencies, ' The format for this table is very similar to that described in Section 2.1.3.1. Figure 2.7 is the data table for initiating events. As shown , in Figure 2.7, each initiating event has a description and an * , associated frequency value. Similar to the user controls for the failure event PRA data tables, the user controls for the PRA data table for initiating events can be l categorized as controls associated with moving the cursor and controls used to alter or modify initiating evert frequency values. As with the failure event data tables, the cursor can be moved to any initiating event frequency field by depressing the "t" or "4" keyc. In_ addition, depressing the "Home" key will move the cursor to the frequency field associated with the first initiating event listed in the PRA data j table. Depressing the "End" key will move the cursor to the frequency , field associated with the last initiating event listed in the PRA data table. Because of the number of initiating events, the initiating events PRA . data table is displayed on two screen pages. The "PgDn," "PgUp," and-scroll features are available for - repositioning the cursor from one i screen page to the other or from individual events to the preceding or i to the next event; Depressing the "PgDn" key will replace the current i
)
17 ', f 3 SELECTION LIST FOR DCMINANT ACCIDENT SEQUENCE INITIATING EVENTS FREQUENCY INITIATING EVENT DESCRIFTION (/YR) Small Small LOCA (.38" < D 5 1.2") 2.0E 02-Small LOCA (1.2" < D 5 1.66") 3.1E 04 Small LOCA (1.M" < 0 5 4") 3.8E 04 Medita LOCA (4" < 0 5 10") 1.7E 04
'* Large LOCA (10" < D 5 13.5") 1.2E 05 Large LOCA (D > 13.5") 7.5E 05 Loss of Power Conversion System Transient 1.1E+00 Loss of Offsite Power Transient 1.4E 01 LF2] PREVIOUS THOREY j Figure.2.7 Initiating Events Data Table f
SELECil0N LIST FOR DOMINANT , ACCIDENT SEQUENCE INITI ATING EVENTS FREQUENCY INITIATING EVENT DESCRIPTION (/YR) Loss of offsite Power Transient Longer than 8 Hours 2.0E 02 Transient with Main feedwater System Initially Available 6.7E+00 Loss of Service Water System Transient 1.0E 02
}
4 LF2] FREVIOUS A MORE A j Figure 2.7 (continued) l
l l 18 display of initiating events with the second page of initiating events i and will move the cursor to-the first initiating event field on the new .! page. Depressing the "PgUp" key will replace the current display of initiating events with the previous page of initiating events and ' will move the cursor to the first initiating' event frequency field on the new page. l The scroll feature allows you to scan through one event at a time in i the list of initiating events in the PRA data table for initiating events. As stated earlier, depressing the "4" key moves the cursor, , line by line, down the displayed page of initiating events. Once the cursor is at the bottom of the displayed list of initiating events, 1 continuing to depress the "1" key will allow you to scroll-downward one l initiating event at a time, through the remaining-initiating events in -
- the PRA data table. Depressing the "t" key moves the cursor up the disp layed list of initiating events. Once the cursor is at the top of the displayed list of initiating events,-continuing to depress the "t" key will allow you to scroll upward, one initiating event at a time, through.those initiating events preceding the displayed list in the PRA data table.
Frequency values contained in the PRA data table for initiating events can be easily altered or modified by using a combination of the cursor controls previously described and standard keyboard entries. A frequency value associated with a specific initiating event can be changed by executing the following steps:
- 1. Position the cursor on the selected frequency field by using the cursor controls previously described. *
- 2. Use the keyboard to enter a new numerical value in the field.
- 3. Move the cursor to a new frequency field or exit the data table to enter the new nuraerical value in the data base.
Note that although numerical values may be entered in ; a variety of l formats from the keyboard, they will always be displayed in scientific { notation with two significant digits. Also, numerical values that are ! inappropriate, or meaningless, will produce an error message when you i attempt to enter these values. If you are modifying frequency values within the PRA data table for initiating events and make an error, PRISIM provides you with the capability to reverse those modifications. This data restoration ' capability can be utilized for all initiating event frequency values or for individual initiating event frequency values. A frequency value associated with an individual initiating event may be restored to its , current data base value by executing the following steps:
19
- 1. Position the cursor on the frequency' field associated with the initiating event of interest.
- 2. Depress the return key.
- 3. Exit the data table or strike any cursor control key to restore the ,
indicated initiating event's frequency to its original value.
=- All initiating event frequency values may be restored to their current.
data base values by simultaneously depressing the control key and the return key. The cursor need not be in a specific position to utilize the "control / return" command. Since modifications made in a PRA data
- ' table are automatically stored in the current data base when you exit the table, the data restoration capabilities described above must be utilized before leaving the PRA data table.
The "F2" (PREVIOUSi function allows you to exit the initiating events PRA data table and returns you to the selection menu in ' the DBM routine. 2.1.4 Controls Associated with the PRISIM PRA Update Routine ; Using the PRA data tables, you can specify a plant condition using any , combination of altered component failure event probabilities, altered ' initiating event frequencies, or altered operator failure event ,
. probabilities. PRISIM's PRA update routine allows you to determine the ,
effect on plant safety resulting from changes to these PRA parameters and to receive other information based on these changes.
- Once you have made all desired changes in the PRA data tables, you will select an option in the PRISIM DBM routine to enter the PRA update 4
routine. The routine will then calculate (in approximately 7 seconds) and display a significance factor, which is the factor of increase that reflects t% altered PRA input values for the instantaneous core damage frequency. The same screen that displayed the plant's significance factor will i give you the option of returning to the control screen or seeing additional information based on the - plant's status. Specifically, PRISIM's PRA update routine provides the following types of information, updated to reflect the current plant status: l ,
- 1. A list of failure events ranked according to their relative f contributions to the instantaneous core dcmage frequency.
l e
- 2. A list of the most liksly scenarios (minimal cut sets) for core damage.
I i i l
~
l _ _ _ _. .- ~_ . _ _ . _ , _ - , _ , _ ~ . - ._ . _ . . , , . . _ _ .
20
- 3. A
- Benefit of ' restoration" option, which shows the factor of decrease in the instantaneous core damage frequency that would be achieved as each disabled comrenent is returned to service.
You select the information you want to obtain from the selection menu or return to the control screen by- moving the cursor to the desired position and prassing the return key. You can scroll up or down event or scenario importance lists using the "t" - or " 4 " keys. The "Home" key may be pressed to return to the first item on the list. Pressing the "Esc" key when you are in the PRA update routine will . return you to the selection menu so you can either request other safety information or return to the PRISIM control screen. Table 2.4 provides definitions of the PRISIM keyboard controls ' associated with the PRA Update Routine. Use this table for quick reference to the controls discussed in this section. 2.2 Controls Associated with Obtaininn Hard Copies of PRISIM Screens If you want to obtain a hard copy of. a text screen while in either the DBM or the interactive portion of the program, simultaneously-pressing the "PrtSc" and "Shift" keys will command your printer to print a copy of the screen being viewed. If you want to obtain a hard copy of a graphics screen being viewed while you are in the DBM routine, you rust use the "Esc" key to first return to the PRISIM control screen. The control screen will appear, and at the bottom of the screen a "picture number" will be displayed. (See Figure 2.8, which is an ,, example of a control screen called up for a schematic with an assigned picture number of 9006.) Using the picture number displayed on the control screen, you can then locate the appropriate schematic in Appendix A and make a photocopy of the schematic. Anytime you leave - the PRISIM DBM routine to check a picture number and obt O 1 hard copy of a graphics screen, selecting the "Resume" option on the control screen will take you back to that graphics screen. At that point, you may continue your PRISIM session. 9 a
., - - . - - e , , . , - -,,_,,,.,e v. , - - - - - = - , - - - , -,n,
21 Table 2.4 PRA Update Routine Keyboard Controls Keyboard Symbol Function of Key s Esc Escape - Returns you to the Selection Menu (Updated PRA
- Resuits)
PrtSc, Prints the currently viewed text screen Shifta Home Returns you to the first of the "Ranking of failure avents" or "Ranking of core damage scenarios" list d Return - Selects the indicated menu option t Moves the cursor in the "reverse" direction t Scrolls up if viewing the "Ranking of failure events" or "Ranking of failure core damage scenarios" list 4 Moves the cursor in the "forward" direction 1 Scrolls down if viewing the "Ranking of falibre events" or a "Ranking of core damage scenarios" list aThese two keys must be depressed simultaneously when you want to print a hard copy of a text screen. e e
22 ( ) Byte Binder RUN TIME DATA BASE CON 1 ROLLER Copyright, Eastern Technical Engineering Corp.1985 87 Rev 2.5 Serial at 000006 Lic PF.0TO TYPE CONTROL SCREEN 8 egin new session L vad Session S ave Session R esune this Session P lay Back Current Session +- Playbeck Speed : 1 Sec. ESC to exit Brade Supported Green Override
$29, Pg 458,P 9006
( ) Figure 2.8 PRISIM Control Screen (Picture Nurnber Check) 6 I I l t 0 e I l l
)
_-=_. __ -
23
- 3. CATECORIES OF SAFETY-REIATED INFORMATION (AND' SAMPLE COMPUTER SESSIONS)
PRISIM provides two general categories of information: (1) updated PRA information (information that has been updated via the PRA update routine and that reflects ANO-l's status when nominal PRA data is changed) and (2) preprocessed information (information that has been de;ived using the results of the ANO-1 PRA and that is meaningful regardless of the plant's safety status). The following sections . define the specific types of information provided by PRISIM and illustrate, through sample computer sessions, how you can obtain the information useful for making decisions that affect plant safety. Specifically, Section 3.1 describes the information provided via the
- PRA update routine and provides instructions on how to obtain this information; Section 3.2 describes 10 types of preprocessed information and provides instructions on how to access this information.
3.1 Uodated PRA Information Changes in plant configuration can greatly affect the safety status of a nuclear power plant. PRISIM's flexibility enables you to specify . equipment that is out of service in order to determine the effect these conditions have on plant safety and to receive other information based on the updated plant status. This capability allows you to account for components that have failed or are removed from service for testing or maintenance. Additionally, PRISIM can be used to perform sensitivity studies by altering any combination of PRA input values. You are able to specify a case to analyze asing any combination of out of-service components, altered component failure probabilities, altered initiating event frequencies, and altered operator failure probabilities. . PRISIM's PRA update routine will aid you in evaluating the safety impact associated with proposed regulatory changes. For example, a request for a one-time extension of an allowed outage time for c specific componeat can be evaluated by determining the significance factor associated with taking that component out of service. This significance factor can be considered the worst-case limit for the proposed ch ange . If permanent Technical Specification changes are proposed, the impact on plant safety could be determined by using the PRA data tables to modify the component's allowed outage time or surveillance test interval and then calculating the new significance factor based on these changes. Other areas in which potential changes can be assessed using PRISIM include certain plant design changes, operating procedure changes, and initiating event frequency changes. Once you have made all of the desired changes using the PRA data
, tables, you enter the PRA update routine of PRISIM in order to receive plant safety status information. Using the PRA configuration you input to the program, the routine will calculate and display (in approximately 7 seconds) a significance factor, which is the factor of increase in the instantaneous core damage frequency that is due to the
24 specified changes. The significance factor is a useful index for quickly evaluating the safety implications of a plant's status. In addition to this significance factor, PRISIM's interactive routine will provide three other types of information updated to reflect the current plant status:
- 1. A list of failure events ranked according to their relative contributions to the instantaneous core damage frequency.
- 2. A list of the most important failure scenarios (minimal cut sets) for core damage.
- 3. A "Benefit of restoration" option, which shows the factor of decrease in the instantaneous core damage frequency that would be achieved when each disabled component is returned to service.
The first screen you will cce in the interactive routine provides the significance factor arid options for obtaining the other types of information. You will notice that a "> " symbol appears to the left of an event whose probability er frequency value has been changed and that an out-of-service event S dimly displayed. The screen has the following format: UPDATED PRA RESULTS XX IS Tile NEW SIGNIFICANCE FACTOR WITil FAILURE DATA CilANGES FOR: Out-of-service equipment > Changed component failure event * > Changed initiating event > Changed operator failure event MENU FOR ADDITIONAL INFORMATION Rankir.g of failure events Benefit of restoration Ranking of core damage scenarios Return to control screen You obtain the additional information desired or return to the control screen by moving the cursor to highlight the desired information and pressing the return key. , Selecting the "Ranking of failure events" option will result in the display of all events ranked according to their relative contributiont to the instantaneous core damage frequency. This list indicates the e i relative importances of cll components. I i 1 i
l i 4 25 l i Selecting the "Ranking of core damage scenarios" option will result in the display of the plant's most likely failure mode given the specified plant condition. You can view additional failure modes, ranked according to their relative contribution to the instantaneous core damage frequency, by pressing the return key. Selecting the "Benefit of restoration" option will result in a ranking of all out-of-service events specified by the user according to the benefit of restoring each to service. This information will help you
- identify the out-of-service equipment that should be repaired first.
PRISIM contains a reduced core damage frequency equation. Not all safety-related components at ANO-1 have been accounted for in the ~* reduced equation; some components that you may want to specify as being out of service cannot be specified as being out of service. Generally, failures of the components that have not been accounted for in the equation will not significantly affect the core damage frequency. However, if several components are simultaneously out of service and cannot be specified as such because of the limitations of the program, the PRISIM interactive routine will not provide a reliable indication of the plant's safety status. Under these circumstances, the PRA update routine should not be used. Sections 3.1.1, 3.1.2, and 3.1.3 present more detailed descriptions and sample sessions to show how the PRISIM PRA data tables and the PRA update routine can be used to monitor the safety status of a tant configuration. 3.1.1 Component Failure Events The core damage frequency is estimated by a PRA, a 10 3 model that
- combines the frequencies of several initiating enents with the probabilities of subsequent basic failure events in order to estimate the risk associated with operating a nuclear power plant. Basic failure evcnts are categorized as either component failure events or operator failure events. Each basic failure event is composed of several subevents. Numerical values that represent each subevent's failure rate, che ensuing downtime, and the resulting failure probability are used as PRA input values.
PRISIM offers you the capability of changing these subevent input values and determining the effect of these changes on plant risk. For , example, you may want to account for components that have failed or are !
- removed from service for testing or maintenance. Or, you may wish to perform a quick sensitivity study to determine the effect of varying the failure rate of a component.
Table 3.1 summarizes the subevent values that can be changed, the field restrictions on these values, and how the input values combine to form subevent and event probabilities. Appendix C describes the methods , used to quantify failure events in the ANO-1 PRA study. An understanding of these methods is essential if you plan to alter the
1 1 Table 3.1 Summary of Subevent Input Values l 1 PRA Input Display Notes Value Designator Restrictions i l Failure rate RATE Two sifnificant digits, The value specified must be a non-negative 3 (per hour) scientific notation (e.g., number not greater than 1.0 to pass the X.XEiW) error checks i i Test interval, TIME Two significant digits, The value specified must be a non-negative downtime, or scientific notation (e.g., number to pass the error checks mission time X.XE+W)~ (hours) $ PRA model used MODEL Model selection from DT, When a MODEL is specified a RATE value i FET, or TI and a TIME value must also be specified to pass the error checks. ; Subevent P Two significant digits, The value calculated or specified must be probability (no scientific notation (e.g... between 0.0 and 1.0 to pass the error units) X.XEiW) checks Event EVENT Two significant digits, The value calculated or specified must be probability (no PROBABILITY scientific notation (e.g., between 0.0'and 1.0 to pass the error units) X.XEiW) checks -
- t
! *
- e , , ,
_.____________ _ _ __ _ - . - .- --- , . . ~ _ _ . _
1 1 27 subevent probability models. Section 2.1.3.1 de sc ribes the user controls associated with the PRA basic event data tables. The following sample session illustrates how you can utilize the PRA data tables to alter the event probabilities associated with selected component failure events and then, using PRISIM's PRA update routine, determine the resulting impact on plant safety. Sample Session In this session, you wish to determine the impact on plant safety of increasing, by a factor of two, the probability that Pump P7A, the turbine-driven pump for the Emergency Feedwater System, fails to start.
- You use PRISIM's PRA update routine to assess the safety implications of this change in the failure probability associated with Pump P7A.
When first entering the PRISIM program, a PRISIM control screen similar to the one displayed in Figure 2.2 will appear on the screen. As described in Section 2.1.1, you should press the letter "B" to begin a PRISIM s e s s ion. The next screen that appears, the PRISIM welcome screen, identifies the plant name and the version of the PRISIM program that you are entering. Hove the cursor to "BEGIN" (by pressing the "t" or "4" key) and then press the return key. You are now in the DBM routine of the PRISIM program. The displayed screen is called the Master Menu (Figure 3.1). It offers the option of obtaining updated PRA information or preprocessed PRA information. For this sample secsion you should move the cursor to the "Updated PRA information"
. option by pressing the "t" or "4" key and then pressing the return key.
This cursor selection process is illustrated in Figure 3.1 by the arrowhead located to the left of the selected menu option.
- After choosing to obtain updated PRA information, you are presented with a list of available options (Figure 3.2). You select "Change component failure probabilities" (by moving the cursor and then pressing the return key) and are presented two paths available for selecting component failure probabilities to be changed (Figure 3.3).
If you select the schematics option, as in Figure 3.3, you will be presented with a list of safety-related systems for which schematics are available (Figure 3.4). Having selected the schematic for the Emergency Feedwater System, you will next see the screen featured in Figure 3.5. You specify Pump P7A as being the component of interest by positioning the cursor on this component and then pressing the return key. The PRISIM program transfers you tc the PRA data tabic associated with the component failure event "EFS Turbine-Driven Pump P7A Fails" (Figure
. 3.6). You are presented with a listing of subevent descriptions and probability values associated with failure of Pump P7A. The PRA data table format is as described in Section 2.1.3.1. After reviewing this list, you determine that the subevent of interest is the first subevent, "EFS Turbine Driven Pump P7A fails to start." You wish to increase the subevent's nominal failure probability of 3.0E 03 by a l
l
l 28 l [ TYPES OF SAFETY RELATED INFORMATION b 1. UPd ated PRA information
- 2. Preprocessed PRA Infor1 nation I
4 I e N ) Figure 3.1 Types of Safety-Related Information l l f [ \ OPi10NS AVAILABLE FOR UPDATING PRA INFORMATION N 1. Change cceponent f atture event probabilities .
- 2. Change operator failure event probabilities
- 3. Change initiating event frequencies 4 Update significance f actor and imortarce rankings l
( j Figure 3.2 options Available for Updating PRA Inforrnation
29 ( h PATHS AVAILABLE FOR ACCESSING COMPONENT FAILURE EVENT PROBABILITY DATA
- 1. Safety related component lists
> 2. Safety related system schematics 4
( ) Figure 3.3 Paths Available for Accessing Component Failure Event Probability Data
~
r 3 SAFETY RELATED SYSTEMS FOR WHICH SCHEMATICS ARE AVAILABLE
, 1. North Battery and Switchgear Rooms Emergency Coolity Tystem
- 2. South Battery and Switchgear Roems Emergency Cooling System
- 3. Core Flooding System
- 4. DC Power System
- 5. Emergency AC Power System
- 6. Emergency Feedwater Initiation and Control System W 7. Emergency Feedwater System
- 8. Engineered Safeguards Actuation System 8 9. High Pressure Injection System
- 10. High Pressure injection System: Room and Lube oil Cooling (CONT!NINDi
( ) ' l Figure 3.4 Safety Related Systems for k'hich Schematics Are Available l
30 ( D EVERGENCY FEEDWATER SYSTEM SCHEuATIC rm o e rm> _emnA muA. .o~
, E'# E#
m won p p 0 m ,7 'w a con 3_E 7 g c=h= d' <m.*
;b ro ec.a y u%, G,o e
- m. ., / ~ ~ ~ ~
ma y u oh moos - - . can .w --
- ,0 em,. em. ~ 4 c,,
m ;)~===
'^~ cmn
+v1 m=20 C}.,*Nt% em
~'A rn,, -
- ,m
-_~,
u cma m., n g
$ *' 'A ._)
- 4. m.um -
c52* 'c$'
,0,CGt 3 SYSTEM UENU END OF &#UT
( ) Figure 3.5 Emergency Feedwater System Schematic ( y EFS TUR8thE* DRIVEN PUMP P7A FAILS RATE TIME SU8 EVENT DESCRIPTION (/NR) (HR) MCCEL P EFS Turbine Driven Ptsp P7A falls 3,0E.03 to start EFS Turbine Driven Ptep P7A 3.0E 05 4.0E+00 falls to run FET 1.2E 04 EFS Turbine Driven Ptrp P7A 3.1E 05 7.0E+00 Of maintenance 2.2E 04 EFS Turbine Dristo Ptrp P7A test EFS Turbine Driven Ptrp P7A regulator falls to operate User Defined Subevent
- n PREVIOUS
( [F2] EVENT PROBASILITT = 3.3E 03 y Figure 3.6 EFS Turbine Driven Pump P7A Fails (PRA Data Table before Data Modification)
31 . l i factor of 2 (i.e., make it 6.0E-03). Using the user controls associated with the PRA data table as described in Section 2.1.3.1, you can use the following procedure to make the. desired change (see Figures 3.6 and 3.7):
- 1. Move the cursor to the probability value for the subevent "failure to start" (3.0E-03).
- 2. Enter the new probability value (6.0E 03).
- 3. Press the return key.
The updated PRA data table (l'igure 3.7) presents the new subevent ~ and
- event probabilities as 6.0E-03 and 6.3E 03, respectively. Note that the PRA data table automatically calculates the new event probability.
You have the option to make changes to other subevents while in the PRA data table. For this sample session, you end the data modification process by pressing the F2 function key. This option returns you to the PRISIM program's schematic or selection list from which the component failure event choice was originally made. For this sample session, you return to the Emergency Feedwater System schematic (Figure 3.8). 4 You have increased, by a factor of 2, the probability of Emergency Feedwater Pump P7A's failing to start. You now have the option to continue to make PRA data changes or to calculate the change in plant . J safety associated with the PRA data modifications already made. As t
, indicated in Figure 3.8, you choose the "END OF INPUT" option, which ,
automatically transfers you to the PRA update routine. The next screen that appears (Figure 3.9) provides the significance factor (factor of increase in the instantaneous core desage frequency) associated with
- the specified PRA configuration, and it offers options to return to the control screen or to receive the following updated safety information:
(1) a ranking of failure events, (2) a ranking of core damage ;
.cenarios, and (3) a ranking of benefit of restoration for out-of- .
service components. As indicated in the figure, the core darnage frequency is not significantly affected when the probability of : Emergency Feedwater Pump P7A's failing to start is increased by a factor of 2. ; i If you want to see a ranking of events, move the cursor to that option (as in Figure 3.9), press the return key, and a screen will appear that provides a list of all basic events ranked according to their relative contributions to the instantaneous core damage frequency (Figure 3.10). Noting that the diesel generators are ranked fourth and fifth, you decide to determine the additional impact on plant safety of taking r Diesel Generator, 1 out of service (i.e., setting the probability of failure of Diesel Generator i equal to 1.0). 7 l i l l t m , . - __ _ _ - _ - - _ _ _ _ , __ _ _ . - . ._. . - . _
I 32 i I [ D EFS TURBINE DRIVEN PUMP P7A FAILS RATE TIME SUBEVENT DESCRIPfl0N (/HR) (HR) MODEL P EFS Turbine Driven Ptrp P7A f alls 6.0E 03 to start EFS Turbine Driven Ptry P7A 3.0E 05 4.0E+00 FET 1.2E 04 4 falls to run EFS Turbine Oriven Ptep P7A 3.1E 05 7.0E+00 DT 2.2E 04 maintenance EFS Turbine Driven Ptrp P7A test EFS Turbine Driven Ptrp P7A regulator falls to operate User Defined Stbevent (F2] PREVIOUS EVENT PROBABILITY = 6.3E 03 j Figure 3.7 EFS Turbine-Driven Pump P7A Fails (PRA Data Table , after Data Modification) ! curactscv retouTta sysitu scauanc a._ m:,_.
&'_a #
co . u . - a_" m&,, p Cw WA Cm sev CT7444 { 64 gg a p>- b" "] -Ov y e" 'u' .B, >- vk[ ,
'tv2.M 'Uta'A
~ ~ '
g p m.as , my u
- O,.- b;7* 1 l _A cs '
n p cun 7+w. u
~v:r- E - @ s - Q l';fd ecuse v ca
=,
Uma JuJr ****** cu,, m
-*% =a ,,,, ry .
T'I J 4 ,,,,_.:, -6 cup xst
+
_AT2 srsitu voiv > eo a uur . N } Figure 3.8 Ernergency Feedwater System Schematic F l t
33 [ UPDATED PRA RESULTS 1.017 IS THE NEW $1GNIFICANCE FACTOR WITH FAILURE DATA CHANCES FOR:
*EFS Turbine oriven Pwp P7A Falls '
r A MENU FOR ADDITIONAL INFORMATION
> Ranking of failure events Benefit of restoration Ranking of core damage scenarios Return to control screen j Figure 3.9 Updated PRA Results r 3 RANKING OF FAILURE EVENTS
, FAILURE EVENT IMPORTANCE PZR Safety Relief Valve Falls to Reclose (EFS Available) 3.2E 01 Battery and Switchgear Room Chilled Water System Train 8 Falls 1.8E 01 Battery and Switchgear Room Chilled Water System Train A Falls 1.8E 01 Olesel Generator 1 Falls 1.5E 01 olesel Generator 2 Falls 1.4E 01 PZR Safety Relief Valve Falls to Rectose (LOSP with EFS Available) 1.3E 01 ICVS ! solation Valve CV3820 Falls to Close 1.1E 01 ACVS isolation Yalve CY3643 Falls to Close 1.1E 01 Independent Failure of the Power Conversion System 8.1E 02 Aligned Stan ty Maketp Pw p P36C Produces low Flow 7.3E 02 EFS Ptrp P7A Turbine Steam Relief Valve PSV6602 Falls open 5.7E 02 South Switchgear Roore 100 Unit Cooler WC2B Falls 4.6E 02 North Switchgear Room 99 Unit Cooler WC20 Falls 4.5E 02 BWST outlet Header Isolation Valve CV14088 Falls to open 4.0E 02 e SWST Outlet Header Isolation Valve Cv1407A Falls to open 3.5E 02 DG1 Heat Exchanger E20A Exhibits Low Service Water Flow 3.5E 02
*EFS Turbine Oriven Ptro P7A Falls 3.5E 02 i DG2 Heat Exchanger E208 Exhibits low service Water Flow 3.2E 02
>ESCtoreturntotheSelectionMenu. j Figure 3.10 Ranking of Failure Events
34 To do this, you must first press the "Esc" key on your keyboard (as instructed by the screen in Figure 3.10), which will return you to the additional information menu (Figure 3.11). Selecting the "Return to Control Screen" option on this screen will then return you to the PRISIM control screen (Figure 3.12). To ensure your previous PRA data changes are retained, press the "R" key on your keyboard to resume the current session. PRISIM will then display the last system schematic presented in the DBM routine (Figure 3.13), and you will return to the system menu and choose the "Emergency AC Power System" (Figure 3.14), Using this schematic (Figure 3.15), you can specify Diesel Generator 1 , as the component of interest. The PRA data table associated with Diesel Generator 1 appears on the screen (Figure 3.16). The cursor position is automatically placed on = the event probability value. In order to specify Diesel Generator 1 as being out of service, you press the return key until the basic event probability value changes to 1.0 (Figure 3.17). As before, you end the data modification process by pressing the "F2" function key. This returns you to the Emergency AC Electrical System schematic. Selecting the "END OF INPUT" option (Figure 3.18), you will next see a screen with a new significance factor that represents the relative increase in cere damage frequency based on Emergency Feedwater Pump P7A's failure to start probability being increased by a factor of 2 and on Diesel Generator 1 being out c.f service (Figure 3.19). In this sample session, you select the "Ranking of core damage scenarios" option and are presented with the most likely failure scenario under the specified plant conditions (Figure 3.20). You may scroll through additional scenarios by pressing the "4" key. , 3.1.2 Initiating Events Accident sequence initiating events are those disruptions to the normal + operation of the plant that cause a rapid shutdown of the plant, or the need to trip the plant, thus challenging the safety systems designed to remove the decay heat being generated by the reactor core. PRISIM offers the user the capability of changing these initiating event frequencies in order to determine tbn effect the changes have on plant safety. For example, you may wish to perform a quick sensitivity study to determine the effect of varying the frequency of a loss of offsite power transient. You may also determine the effect on plant safety of coinbining altered initiating event frequencies with altered basic event probabilities. i Section 2.1.3.2 describes the user controls associated with the initiating events PRA data table. The following sample session l illustrates how you can change an initiating event frequency to , j determine the impact the change will have on plant safety. j
35 ( h UPCATED PRA RESULTS i i 1.017 !$ THE NEW SIGNIFICANCE FACTOR WITH FAILURE DATA CHANGES FOR: ! PEFS Turbine Driven Ptrp P7A Faits d MENU FOR AD0!TIONAL INFORMATION Ranking of failure events Benefit of restoration Ranking of core dage scenarios > Return to control screen j ( Figure 3.11 Updated PRA Results t [ ) I Byte Blnder RUN TIME DATA BASE CONTR01.LER
, Copyright, Eastern Technical Engineering Corp.1985 87 Rev 2.5 Serial # 000006 Lic PROTO TYPE CONTROL $CREEN ;
o 8 eqin new session L oad Session S ave Session i ! R esune this Session 7 lay Back Current Session +- Playback Speed : 1 Sec. ESC to exit Brade supported Green Override L SS, Pg 451,P 9010 ( ) - t Figure 3.12 PRISIM Control Screen ; f I
36
?
[ EMERGENCY FEE 0 WATER SYSTEM SCHEM ATIC
._ A _ A_.
cw. cuo. ,o. b wi gp o u_g,__ b
= =
r,
"" .w ma ,,, ma e.
nsa u <%aA-N Tjy"Q
#mur v ,
y lamm =
, , en.y -
~ '
cv2. Au n can. b en= r L_. a4.,, ew
'." u n,
oJM.,-4Y' I em
':com E .,m, g,
a es, 6 mw3 v cc.*
-3, , 3 <b-*4 - ** cu., n
- , cvr2 ,v.3 >-
'1'.'l'h-.;
~
%. wun
- cs? ""
_ genes *
> sysTEu uCNu to or mmt
( ) Figure 3.13 Emergency Feedwater System Schematic ( ) SAFETY RELATED SYSTEMS FOR WHICH SCHEMATICS ARE AVAILABLE
- 1. North Battery and switchgear Rxms Erergency Cooling System *
- 2. South Battery and Switchgear Rooms Emergency Cooling System
- 3. Core Flooding System 4 DC Power System h 5. Erergency AC Power System
- 6. Erergency Feedwater Initiation and Control System
- 7. Energency Feedwater System
- 8. Engineered Safeguards Actuation System
- 9. High Pressure injection System *
- 10. High Pressure Injection System: Room and Lube Oil Cooling (CONTINUED)
( ) Figure 3.14 Safety Related Systems for Vhich Schematics Are Available i
37 i f (VEPCENCi AC ELICTR1 CAL SYSitW SC>tEW ATIC i a , , n I
" 0**
4o3=t 3 t). &
) a.
s) .. t>
"O" i, f) t) t t f) t
" t + +) } }
i i i i i < .. < I Le ""a "y , T) s 4> - .
- 7>=
4 ,. b ') b b J' .,
)."
s - s q~p s) -. llll 2 f 7r se _ , 2 =llll
- m. ,
~
- r. % , 5, 5, ,,
r1 a SysitM WENU (ND Or input ( ) Figure 3.15 Emergency AC Electrical System schematic r 3 DIESEL GENERATOR 1 FAILS
=
RATE TIME SUBEVENT DESCR'* TION (/NR) (HR) MODEL P Diesel Generator 1 falis to start 2.5E 02 > I i Diesel Generator 1 fails to run 3.0E 04 2.4E+01 FET 7.2E 03 1 Diesei Generator 1 maintenance 6.0E 05 2.5E+01 DT 1.5F 03 Olete! Generator 1 test 1.4E 02 8.0E 02 FET 1.2E 04 l
- Time Delay Relay fails short 5.0E 05 circuit (i; c8 5) tircuit c A308 fai.s '1 1.0E 03 close
* ' US - A 'T EVENT PR08 ABILITY = 3.9E 02 2?
m _ .. y ; Fig re > . rator 1 Fails (FRA Data Table before l M .tication) l t 1
,= ,
38 ( D O!ESEL CENERATOR 1 FAILS RATE TIME SUBEVENT DESCRIPTION (/MR) (MR) MCcEL P Diesel Generator i falls to start i Oletel Generator 1 f alls to run
- Diesel Generator 1 mainte unce
+
Diesel Generator 1 test Time Delay Relay falls short circuit (1 of 5) Circuit Breaker A308 falls to close T MORE T ( [F2) PREVIOUS EVENT PROBARILITY = 1.0E+00 y Figure 3.17 Diesel Generator 1 Fails (PRA Data Table after Data Modification) ( . tuf RCENCY AC ELECTBCAL SYSTEW SO= o .') ...
",%F
) ) ) ) ) )
i i i i _me y an
"
- me
+ ,) = ,>= T 3
., 4) 6 + 4 ,
t t t t t) t)
",>' _ ' 7'> j lll12, r, 2,t11 11
.: e _ .=, l i ,
m.".' o.
, ) , , ..n. L .
c.2 srsitu w;
> Eso ce mi .
}
{ Figure 3.18 Emergency AC Electrical Systern Schematic j i
39 ( UPDATED PRA RESULTS h 4.9 IS THE NEW $1GNIFICANCE FACTOR WITN FAILURE DATA CHAhCES FOR:
*EFS Turbire Driven Pupp P7A Falls Diesel Cemrator 1 Faits O
MEWU FOR ADDITIONAL INFORMATION Ranking of-failure events Benefit of restoration
> Ranking of core damage scenarios Return to control screen j Figure 3.19 Updated PRA Results t 1
e f D ' RAhKING OF CORE DAMAGE SCENARIOS l SCENAtto # 1 FRACTION OF CMF 3.8E 0. ' Loss of Offsite Power Transient Longer than 8 Hours AND Diesel Generator 1 Falls AND Diesel Generator 2 Falls . . + i e ESC to return to the Selection Menu. y [ i Figure 3.20 Ranking of Core Dr.:nage Scenarios 4 , i a i i 1
40 Sample Session In this sample session, you wi s'c to determine the impact on plant safety of increasing, by a factor of 2, the initiating event frequency associated with the loss of the Power Conversion System transient. First, you enter the PRISIM program and are presented with the Master Menu (Figure 3.21). For this session, you select "Updated PRA information" and are then offered four options (Figure 3.22). You select the option associated with initiating events and Figure 3.23 appears on the screen. - Figure 3.23 is the initiating events PRA data table. The table provides a listing of all ac:ident sequence initiating events and associated frequency values in units of events / year. You may scroll through the complete list of initiating events by using either the "4" key or "PgDn" key. Upon scanning the list, you determine that the initiating event of interest is "Loss of Power C"nversion System 1ransient," whic h has a frequency of 1.1 events / year. You move the cursor to highlight this value, type in the new value (2.2), and press the return key (Figure 3.24). For this sample session, you select the PREVIOUS option and are returned to the "Options Available for Updating PRA Information" menu (Figure 3.25.) You have successfully increased the "Loss of Power Conversion System Transient" initiating event frequency by a factor of
- 2. You now have the option to continue to make PRA changes or to ,
calculate the change in plant safety associated with the initiating event modification already made. As indicated in figure 3.25, you choose to calculate the new significance factor (Figure 3.26). As indicated in the figure, the core damage frequency is increased by a approximately 20% when the loss of power conversion system transient frequency is increased by a factor of 2. With the indicated PRA data modification, the most likely accident scenario involves the modified initiating event (Figure 3.27). 3.1.3 Operator Action Events Planned operator action events are those actions that are to be carried out by the operators following the occurrence of an initiating event that causes or requires a reactor shutdown. PRISIM offers the user the capability of changing the failure probabilities associated with planned operator action events in order to determine the e f fe c t a change has on plant safety. For example, you may wish to determine the eifect of eliminating credit for certain p:anned operator actions (i.e., setting the probability of failure equal to 1.0 for these . selected operator action events).
i 1 41 [ \ ' TYPES OF SAFETY RELATED INFORMAfl0N l
> 1. Updated PRA infamation
- 2. Preprocessed PRA information I
e ( ) Figuru 3.21 Types of Safety Related Infonsation [ D OPTIONS AVAILABLE FOR UPDATING PRA INFORMAfl0W
. 1. Che..ge ccrponent f ailure event probabilities
- 2. Change operator failure event probebilities
> 3. Change initiating event frequencies
- 4. Update significance factor ard leportance rankings 4
1 ( ) .I Figure 3.22 Options Available for Updating PRA Information ! i a b
42 ( T , SELECTION List FOR DOMINANT ACCIDENT SEQUENCE IN!!! ATING EVEhts FREQUENCY INITIATING EVENT DESCRIPfl0N (/YR) SanL L Smell LOCA (.38" < 0 51.2") 2.0E 02 Smet t LOCA (1.2" < D s 1.M") 3.1E 04 Smelt LOCA (1.M" < 0 5 4") 3.8E 04
- Nedlue LOCA (4" < 0 f 10") 1..*E 04 Large LOCA (10" < 0 s 13.5") 1.2 05 ,
Large LOCA (D > 13.5") 7.5E 05 Loss of Power Conversion Systera '.ransient 1.1E+00 Loss of Of fsite Power Transient 1.4E 01 PREVIOUS T HORET ( (F2] y Figure 3.23 Selection List for Dominant Accident Sequence Initiating Events (PRA Data Table before Data Hodification) g - 3 . SELECTION LlST FOR DOMlW4NT ACCIDENT SEQUENCE INiflAfikG EVENTS FREQUENCY INITIAflNG I VENT CESCRIPf!0N (/YR) Smal1 SmalL LOC 4 (.38" < 0 s 1.2") 2.0E.02 Small LOCA (1.2" < 0 s 1.M") 3.1E 04 Small LOCA (1.M" < D 5 4") 3.8E 04 Medius LOCA (4" < D 510') 1.7E 04 Large LOCA (10" < 0 s 13.5") 1.2E 05 Large LOCA (D > 13.5") 7.5E 05 Loss of Pcver Conversion System Transient 2.2E+00 " Loss of offsite Power Transient 1.4E 01 ( (F2]
- PREVIOUS T MORET
- j I.
Figure 3.24 Selection List for Dominant Accident Sequence
! Initiating Events (PRA Data Table after Data Hodification)
43 [ D OPTIONS AVAILABLE FOR UPDAT!WG PRA INFORMATION
- 1. .'hange corponent f ailure event probablLities
- 2. Chgm e operator failure event probabilities
- 3. Change initiating event frequencies
> 4. Update significance factor and importance rankings e
J Figure 3.25 Options Available for Updating PRA Information
- r 3 VF0ATED PRA RESULTS 1.23 15 THE NEW $1GNIFICANCE FACTOR WitM FAILURE DATA CHANGE $ 102:
e
- Loss of Power Conversion System Transient I
MEN'J FOR ADDITIONAL IhFORMAt!ON Ranking of falture events Benefit of restoration
. ( > Ranking of core damage scenarios Return to control screen Figure 3.26 Updated PRA Results
44 1 I f RANKING OF CORE DAMAGE SCENARIOS SCENARIO F 1 FRACTIN OF CMF 8.2E 02 6 1
- Loss of Power Conversion System fransient AND PZR Safety Relief Valve Falls to Rectose (EFS Avaltable)
A'io Battery and Switchgear Room Chitled Water System Train 8 Falls j i l s ,. 1 > l ( ESC to retu*n to the Selection Menu. Figure 3.27 Ranking of Core Damage Scenarios - i a I 9 i i l l J
) I i
1 i f 1 1 I J G ( 1 l 1 i
)
i i 45 3
- i 1 '
1 Section 2.1.3.1 describes the user controls associated with - the operator action event PRA data tables. The following sample session ~ l illustrates how you can change an operator action ~ event probability to t determine the impact that change would have on plant safety, a S =nle Session ; i In this sample session, you wish to determine the impact on plant j j { safety of increasing the failure probability associated with a planned i
- operator action. The operator action you are interested in is [
"Operator fails to initiate the High Pressure Injection System (HPIS) !
You wish to increase the failure probability by j during a small LOCA. " a factor of 20. r As with the previous sample session, you enter the PRISIM program ! I. through the Master Menu (Figure 3.28) and select the "Updated PRA information" option. You are then presented with a list of options l ] available for changing PRA data (Figure 3.29). As indicated in the figure, you choose to "Change operator failure event . r probabilities." The next screen is a listing of operator actions whose probabilities ! may be changed (Figure 3.30). You select the event of interest end are l then presented with the PRA data table associated with the operator l 4 failure event "Failure to Initiate HPIS" (Figure 3.31). The format of , i I this PRA data table is functionally identical to the component failure event PRA data table described in Section 3.1.1. ! You wish to increase the nominal event failure probability (1.0E-04) by ; a factor of 20 (new value - 2.0E 03). Using the keyboard controls i described in Section 2.1.3.1, move the cursor to the probability value
- for the operator action (1.0E 04), type in the new probability value !
(2.0E-03), and press the return key. : The updated PRA data table (Figure 3.32) now contains the new event ; probabilities (2.0E 03). For this sample session, you end the data i modification process by selecting the "F2" key. You are then returned l to the operator failure event menu (Figure 3.33). l You now have the option to make changes to other operator failure , i events, to make changes to component failure ar.d/or initiating events, ! or to calculate the change in plant safety associated with the PRA data [ As indicated in Figure 3.33, you choose the j modification already made. j l
"END OF INPUT" option, which automatically transfers yoa to the PRA ;
- update routine. The significance factor associated with tt 3 data modification is then presented (Figure 3.34). ,
a u i i
46 [ 3 TYPES OF SAFETY RELATED INFORMAfl0N h l. Updated PRA information 4
- 2. Preprocessed PRA Information i
s
% )
Figure 3.28 Types of Safety Eelated Information CPfl0NS AVAILABLE FOR UPDAflWG PRA INFORMATION
- 1. Change component, f ailure event probabilitle. ,
> 2. Change operator f atture event probablLitles
- 3. Change initiating event frequencies
- 4. Update significance factor and leportance rankings l
1 L )
- Figure 3.29 Options Available for Updating PRA Information i
l i i l
47 ( ) OPERATOR ERRORS WCSE PR08 ABILITIES MAY BE CHANGED
> 1. Operator falls to initiate the HPl$ during a small LOCA
- 2. Operator falls to close Stop Check Valve MU13 on a low levet signal from the makew tank
- 3. Operator f alls to align makew pumps to
- take suction from the LPl$ during a small LOCA
- 4. Operator falls tt switch the decay heat removal ptsp suctions to the step Wring a small LOCA
# Operator falls to establish feed and bleed cooling 5.
(CONTINUEC) OPfl0NS MENU END OF lhPUT N Y Figure 3.30 Operator Errors Whose Probabilities May Be Changed l J i FAILURE 10 INITIATE HPIS : 1 g e RATE TIME , i SUBEVENT DESCRIPTION (/HR) (HR) MODEL P operator falls to initiate HPIS & ring 1.0E 04 small LOCAs User Defined Subevent e I ( (f2) PREV 10JS EVEhT PROBABILITY = 1.0E 04 j Figure 3.31 Failure to Initiate HPIS (FRA Data Table before Data Modification) i i
J 48 I . FAILeRE TO l'ITIATE J HPl3 RATE TIME
$UBEVENT DESCRIPT!0N UHR) (HR) KCEL P Operator falls to initiate HPIS during 2.0E 03 I small LOCAs User Defined Subevent
- l
.1 A
l l (F2] PREVICUS EVENT PROBABILITY a 2.0E-03 y 1 Figure 3.32 Failure to Initiate HPIS (PRA Data Table after Data Modification) l C$ERATOR ERRORS WN"$E PRO 6 ABILITIES MAY BE CHANGED j= f *
- 1. Operator f alls to initiate the HPIS during a small LOCA
- 2. Operator falls to close stop-Check Valve Mu13 on a low level signal from the makeup tank
- 3. Operator f alls to align maketp ptrps to 4
take suction from the LPIS during a small LOCA .; I j 4. Operator falls to swltch the decay heat removal I prp suctions to the strp d.scing a small LOCA
- 5. Op rator fails to establish feed and bleed cooling s
(CONiIWUED) l
- (
opilows MEwu > END Or ikPut I l k ) I i 1 Figure 3.33 Operator Errors k' hose Probabilities May Be Changed i n i 1
49 [ UPDATED PAA RESULTS
* .i 1.091 IS THE h8W SIGNIFICANCE FACTOR WITN FAILURE DATA CHANGES FOR: -
I
>Fa((ure to Initiate NPIS I
4- ; 1 i
2. Preprocessed PRA information s
( _ ) Figure 3.35 Types of Safety Related Information ( CATECORl[$ C; PREPROCES$10 PRA thPC6MAfl0N D
> 1. Donnnant core damage segaences e
- 2. Safety related systems
- 3. Safety estated s h ystems
- 4. Safety related cepponents
- 5. S e rt system interfaces
- 6. Operator actions 7 Corponent falture data
- 8. System testing / surveillance
- 9. Limiting conditions for operation
- 10, System cosperisons for selected B&W plante
( ) Figure 3.36 Categories of Freprocessed PRA Information
P 53 ( 3 DOMI MNT ACCIDENT SEQUCMCE SELECil0N LIST
% OF CORE MELT % OF CORE MELT INITIATING SAFETY RELATED FREQUENCY FREQUENCY ltANK EVENT SYSTEM FAILURES (W/0 RECOVERY) (W/ RECOVERY) 1 Loss of Fower SRVR NPRS 21.0 7.4 Conversion Systee
* >2 Small Small LOCA HPIS 16.0 17.0 (Eq. dia 5 1.2a) 3 Loss of offsite HPIS 14.0 27.3 power longer than
, 8 hours
& Loss of offsite StVR MPRS 12.0 3.8 poWP (CONTINUED) i
( ) . r l
, Figure 3.37 Dominant Accident Sequence List I i
DESCRIPfl0N OF ACCIDENT SEQUENCE #2 , i
, INITIATING EVENT This seg ance is initiated Ly a (pture of a Reettor Coolant System (RCS) pwp seat or a r@ture in the RCS piping in the range of eqJiv- '
atent diameters .3Sa < dia 5 1.2". Neither type of rwture directly causes any safety related equipment f altures. SYttEM FAILUEES The RCS rwture is followed by a failure of att Nigh Pressure Injec.
; tion System (HPIS) pg trains to provide flow to the RCS. With no recovery actions cunsidered, this accident sequence contributes 16%
to core melt f req >ency. I ! I
> DETAILED SCENARIO INFORMATION FOR THIS SEQUENCE RECOVERY INFORMAfioh FOR THis SEQUENCE 4 l
( ) l Tigure 3.38 Description of Accident Sequence #2 l l
~ . __
54 ( ) Dm!NANT SCENARIOS FOR ACCIDENT $E0VENCE #2 Scenario 2 1 described below contributes 11% to the sequence frequency. The initiator is a "Smell Small LOCA .38 in. < Eq. die 51.2 In." HPl$ FAILURE Failure of the Chilled Water Unit for cooling
$witchgear Room 100
- resulting in toss of oc power to Pwps P36A and P368 ANO fatture of the Chitted Water Unit for
- cooling Switchgear Room
! 99 resulting in Loss of oc power to Pwp P36C
> RECOVERY adFORMATICW FOR TNIS SCENAitIO
SUMMARY
OF SCENARIOS NEXT SCENARIO N Y Figure 3.39 Dominant Scenarios for Accident Sequence #2 D 1 4 D i i i 4 1 ___________________\
J 55 also include options for accessing recovery information for. the ! selected scenario, for accessing a detailed description of the next a scenario, and for accessing a summary of all scenarios for the selected
- accident . sequence. In this example, you ' choose . to see recovery information for ' the selected scenario. - This information will be displayed on a screen like that illustrated by Figure 3.40. !
At any time during the selection of these options, you can backtrack to
- the previous screen, backtrack to the last branch point' '(the last l
. screen that displayed multiple menu options), or return to the Master Menu by using the keyboard controls (described in Section 2) for these !
options. J
- 3.2.2 Safety-Related Systems PRISIM's data base per.vides importance measures for safety-related l systems--measures that may be used to make decisions about where to .
apply resources to maintain or improve plant safety. [ PRISIM provides' three types of risk importance measures for safety- i related systems. J e Risk reduction e Risk sensitivity ] a ! e Risk significance j Table 3.2 defines *hese three importance measures.
. The measures of importance in this version of PRISIM are based on core - damage !
frequency. j i l If you command PRISIM to present either ti.e risk redvetion or the risk !
- sensitivity importance mensure, you will be given a list of systems 1 ranked according to a numerical value that. appears with each system on i l the screen. If, on the other hand, you select the risk significance ;
} importance measure, a list of systems grouped according to high, ! j moderate, and lo + risk significance will appear. No numerical values 3 l appear on the screen with these systems, and there is no implied ; j ranking within any of the three groups. i i The following is a description of how you could access some of the l l information on safety related systems available in PRISIM. l l
- 1 f Samole Session {
l l , Assuma you want a list of ANO-1 safety-related-systems ranked according [ to their risk reduction importance measures. You enter the j preprocessed informa'io, category of PRISIM (Figure 3.41) and select I
- the "Safe ty-relater .jstems" option (Figure 3.42). The subsequent 1 screen (Figure 3.43) offers you the option of seeing systems listed according to the risk reduction, risk sensitivity, or risk significance i
i ) 3
l 56 I [ ) I RECOVtkY INFORMAT!DW FOR $NWAR10 21, ACCIDENT SEQUENCE #2 I s Recovery from this scenerlo is achieved by restoring the room coollne capability of either Chltled Water Unit VCN43 or VCM4A, or both, to pre
- i' vent a toss of Rooms 100 and 99 cooling, respectively. Room cooling '
capability is regelned by either restoring the chitted water mit or by manually starting portable fans. In either case, this is a local recov. ; ery action. The time evettebte for restoring cooling capabl(Ity without i irreversible danese is etwt 70 airwtes. Recovery decreases the eett-l meted fregency of this scenerlo by a factor of 100. . , l l l i ( ) $ Figure 3.40 Recovery Information for Scenario 21 Accident I Sequence #2 ! b l i i i
- l t
l e
57 Table 3.2 Measures of Importance and Their Interpretation Importance Measure Interpretation l Risk Reduction This measure is the likelihood the equipment would be failed and would contribute to a core damage event if a core damage event were to occur Risk Sensitivity This measure is the rate of change (or sensitivity) of the frequency of core damage events with changes in the probability of the equipment failure Risk Significance This measure is a combination of the risk reduction and risk sensitivity measures. Like the tisk reduction measure, it emphasizes the equipment's contribution to risk. But it also
. gives additional weight to equipment with high risk sensitivity ireportance .
i e P i l l i 1
I l 58 [ D TYPES OF SAFETY RELATED lbf0RMAfl0N
- 1. Updated PRA informattoi
> 2. Preprocessed PRA information l l
a l k ) Figure 3.41 Types of Safety Related Information {
. t
( ) ' CATEGORIES OF PREPROCESSED PRA INFORMATION .. 1 l 1 Dominant core damage segaences
- l h 2. Safety relettd systems !
d
- 3. safety retsted subsystems 4 Safety related cceponents l 5. Support system interfaces
].
- 6. Operator actions :
l T. Corponent f ailure data
- 8. System testing /survelttance
.i 9. Limiting corditions for operation J
- 10. System coparisons for selected B&W plants k , )
4 i Figure 3.42 Categories of Preprocessed PRA Information 1 4
i 59 [ D l INF0aMAi!ON AVAILAllt FOR SAFETY ttLATED SYSTEMS i
> 1. Risk redxtion leportance i
- 2. Risk sensitivity importance r
- 3. Risk significance isportance ;
t l 4 i ; r s
; k )
i i ' Figure 3.43 Information Available for Safety Related Systems " i 4 r l 1 . / 4 i i 1 i
- 9 I
i i 4 L s i & d e l a 9
l 60 l importance measures. You select the "Risk reduction importance" option l and a ranking of systems according to their risk reduction importance measures will appear on the next screen (Figure 3.44). 3.2.3 Safety-Related Subsystems The PRISIM data base provides the same three types of information on ! safety-related subsystems as it provides on safety related systems; that is, it provides rankings according to the risk reduction and risk i sensitivity importance measures and provides groupings according to . risk significance importance metsures. The following sample session ' illustrates how you can access some of the information on safety-related subsystems available in PRISIM. l Sample Session l Suppose you want a list of ANO-1 safety-related subsystems grouped according to their risk significance importance measures. You enter the preprocessed information category of PRISIM (Figure 3.45) and select the "Safety-related subsystems" option (Figure 3.46). The subsequent screen that appears (Figure 3.47) offers you the option of seeing subsystems listed according to the risk reduction, risk sensitivity, or risk significance importance measures. You select the "Risk significance importance" option and then a screen vill appear that presents the grouping of subsystems according to their risk significance importance measures (Figure 3.48). 3.2.4 Safety-Related Components , PRISIM also provides importance information on components (i.e., the . l three importance measures defined for safety-related systems and l ! subsystems). And it provides five other types of information that
- are relevant when a particular component is out of service.
I 1. Single active component failures (e.g., the failure of a motor- I operated valve to close, or the failure of a pump to operate) that I will disable the system when the specified component is out of l service. l l 2. Information on support system interfaces (for single active l component failures described above), i l
- 3. Schematics that show the equipment realignment required for a , l component being taken out of service for testing.
- 4. Schematics that show the equipment realignment required for a component being taken out of service for maintenance. ,
- 5. Plant-specific LERs that describe previous failures of components similar to the component that has been specified as being out of service.
l l l
i l 61 l [ D FRCNT LINE SYSTEMS RANKED BY RISK REDUCTION IMPORTANCE ' RISK REDUCTION , I SYSTEM IMPORTANCE High Pressure injection System 0.8 Pressuelter safety Relief System 0.5 f Emergency Fee & ster Systee 0.09 Power Conversion System 0.08 Low Pressure Injection System 0.04 Reactor Protection System 0.008 Core Flood Systee 0.0001 Reactor Building Cooling System a Reactor Building Sprey System a N ) Figure 3.44 Front Line Systems Ranked by Risk Reduction Importance 4 e l
62
)
r 3 TYPES OF SAFETY RELATED INFORMATION
- 1. Updated PRA Information
> 2. Preprocessed PRA information
( ) Figure 3.45 Types of Safety-Related Information f .)
- CATEGORIES OF PREPROCESSED PRA INFORMAfl0N
- 1. Dominant core damage sequences e
- 2. Safety related systems
> 3. Safety related subsystems 4 Safety related corponents
- 5. Support system interfaces
- 6. Operator actions
- 7. Cw ponent fatture dats
- 8. System testing /survelttance 9 Limiting conditions for operation *
- 10. System corparisons for selected 88W plants
( ) Figure 3.46 Categories of Preprocessed PRA Information
- 9. . .
63
? \
f INFORMATION AVAILABLE FOR SAFETY RELATED SUBSYSTEMS
- 1. Risk reduction leportance t 2. Risk sensitivity leportance
> 3. Risk significance importance
) l ( ) Figure 3.47 Information Available for Safety-Related Subsystems l
,- 3 SUS $YSTEMS GR. >ED BY RISK SIGNIFICANCE IMPORTANCE l
- MIGH RISK SIGNIFICANCE Emergcuy Feedwater System Turbine Driven Ptap Train Creen AC Power Green DC Power High Pressure Injection Stan ty Pttp (P36C)
Lew Pressure Injection Train A low Pressure Injection Train R North Battery and Switchgear Eoom Cooling System Chilled
, Water Unit (Train A) horth Switchgear Room (Roem 99) Cooling
, > (CONTINUED)
\ }
Figure 3.48 Subsystems Grouped by Risk Significance Importance m )
64
- ( D i SUBSv$TEMS CROUPED BY RISK SIGNIFICANCE IMPORTANCE (CONTINUED)
L MIGN RISK SIGNIFICANCE (continued) Pressurfter Sa.^ety Relief Valves fall to open Pressurizer Safety Relief Valves' fall to reclose Reactor Protection System Channel A Reactor Protection System Chamel B ! Reactor Protection System Chnwwl C Reactor Protection System Chavwl D
- Red AC Power Service Water System Loop 1 .
Service Water System Loop 2
> (CONilWED)
N Y Figure 3.48 (continued) _ e [ \ SUS $YSTEMS CRCUPED BY RISK SIGNIFICANCE IMPORTANCE (CONTINVED) NICat RISK SIGNIFICANCE (continued)
- South Battery and Switchgear Room Cooling System Chilled Water Unit (Train B) i South Switchgear kocwn (Room 100) Cooling l
MODERATE RISK SIGNIFICANCE Emergency feedwater System Motor Driven Pwp Train Engineered Safeguards Actuation System Digital Channel 1 Engineered Safeguards Actuation System Digital 1 (corren equipnent for Digital Channels 1, 3, 5, 7,
- f ard 9) j h (CONilWED) '
( ) i i i Figure 3.48 (continued) i
65 ( ) SU85YSTEMS GROUPED BY RISK $1GNIFICANCE IMPORTANCE (CONTINUED) MODERATE RISK SIGNIFICANCE (continued) Migh Pressure Injecticv1 Normally Operating and Aligned Stanty Pwps (P368 and P36A) Red DC Power I LOW RISK $1CNIFICANCE Core Flood Systee Train A a Core Flood System Train B J
- Emergency Feedwater Initiattun and Control System Control / Vector Trains A and D (Motor Driven Pwp Train)
Emergency Feedwater Initiation anc witrol System Control / Vector Trains B and C (Turbine Driven Pwp frain)
> (CONTINUED)
( ) Figure 3.48 (continued) f ) SUBSYSTEMS CROUPED BY RISK $!CNIFICANCE IMPORTANCE (CONTINUED)
, LOW RISK SIGNIFICANCE (continued)
Emergency Feedwater Initiation and Control System Trip Train A (starts both EFS pw ps) Emergency Feedwater Initiation and Control System Trip Train B (starts EFS Turbine Driven Pwp) Engineered Safeguards Actuation System Digital Chamel 2 Engineered Safeguards Actuation System Digital Channel 3 Engineered Safeguards Actuation System Digital Channel 4 Engineered Safeguards Actuation System Digital 2 (ccmon equipnent for Digital Channels 2, 4, 6, 8, and 10) 1 I ( _ _M Figure 3.48 (continued) i i l l
66 The following sample session illustrates how you can access some of this safety related component information using PRISIM. Snacle Sessi<m Assume you want to see PRA related information that is relevant when a particular component, Emergency Feedwater Pump P7A, is out of service. As in previous PRISIM sessions, you are first presented the option of obtaining updated PRA information or preprocessed information (Figure 3.49). You select "Preprocessed Information" and are presented a list . of preprocessed information categories addressed in PRISIM (Figure 3.50). You then choose "Safety related components" (as indicated in Figure 3.50), and a menu of information categories for components appears on the next screen (Figure 3.51). Since you are interested . in seeing information that is relevant when Pump P7A is out of service, you select the "Information relevant when a particular component is out of service" option. The next screen that appears (Figure 3.52) will be a menu screen with two options for specifying the out-of service component. You select the schematics option, as indicated in the figure, and a new screen will appear that lists the safety-related system schematics that are available for use in specifying a component out of service (Figure 3.53). On this screen, you indicate you want to see a schematic of the Emergency Feedwater System. The next screen that appears will be a schematic of that system (Figure 3.54). . You can now specify the out of service component by moving the cursor (using the "t" or "4" keys on your keyboard) to Pump P7A's position on the schematic and then pressing the return key. The next screen that , appears (Figure 3.55) lists the information that is available based on Pump P7A being out of service. If you elect to first receive information on the most likely single failures for the Emergency Feedwater System when Pump P7A is out of service, you can command . PRISIM to provide this information by positioning the cursor on this entry on the screen and pressing the return key. PRISIM will then provide the screen illustrated in Figure 3.56. This new screen will let you see the most likely single failures, and it will give you the option of seeing additional single failures involving support system equipment that will fail the system. Moving the cursor to the desired component on the screen and pressing the return key will command PRISIM to provide this additional failure information. (Figure 3.57 shows the support system equipment for Pump P7B.) This screen format is discussed in Section 3.2.5. 3.2.5 Support System Interfaces
- i A fif th type of preprocessed information in PRISIM's data base relates to the support services (power, service water, etc.) required by safety ,
system components. Information on support system interfaces can help you verify the operability of equipment. For example, if you know one train of a safety system is to be isolated for maintenance on a component, you would want to first verify the operability of the 1 i
.~. 8 67 1
( h j ffPts 07 SAFETY RELATED INf0dMAfl0N ;
- 1. Updated PRA information
> 2. Preprocessed PRA information l
( ) Figure 3.49 Types of Safety Related Information f 3 CAftGORit$ CF PREPROCES$tD PRA INFORMATION
- 1. Dominant core damage sequences
- 2. Safety related systems
- 3. Safety related ednyttems
>4 Safety related corponents
- 5. Support system interf aces
- 6. Operator actions
- 7. Corpenent f ailure cate
- 8. System testing / surveillance
- 9. Limiting conditions for operation
- 10. System coeparisons for selected 81W plants I
( ) Figure 3.50 Categories of Preprocessed PRA Information
1 68 i ( ) INFORMAfl04 AVAILA8tE FOR SAFEff RELATED COMPOWENTS l
- 1. alsk reduction isportance
- 2. Risk sensitivity leportance
- 3. Risk significance isportance b 4. Information relevant when a H rticular corponent is out of service P a
a 1 Y ) Figure 3.51 Information Available for Safety Related 4 Components ?, , d ( 3 ,; OPi!ONS AVAILABLE FOR SELECilhG OUT OF*StRV!CE COMPOWENTS 1 1 Corponent Lista '
- h 2. Schematics i
1 4 4 I s i i t ( j t Figure 3.52 Options Available for Selecting Out of Service j Components i i i
69 i [ I
$AFETY RELATED SYSTEMS FOR WlCM CCMPohENT INFORMATION l$ AVAILABLE l > 1. Emergency Feed.ater Systers
- 2. Migh Pressure injection System j 3. LowPressureInjectionSystem k l
a 4. Reactor Building Cooling System .
- r l 5. Reacto* Building spray System a 4
l 1 4 - I 1 a i 9 t i 1-( ) i i l i 4 q Figure 3.53 Safety-Related Systems for Vhich Component i Information Is Available i i
- i. .
tocaco.cv recuna sysxw sc4w nc 3 l _ 1
, . , o , ..;., ;
m * .* u 44
. - g,_-c.>- .u. . .s , , .m i c.u , un.. . us ,
- m. m m.m_.
s:.r u
- ><- w i
'E y,p~AmbytQ ou21 3
3 4 [ _s ",x>._ v
,;.. ~
o,,,., -
. ,-- - ~ . ~ . ,.
- l
, c ,m, u?>
D ~ :}-- -
~
?
t g ,
- 4 a - .4..
cm o w->.- .-- o. .
%.3 _ . . t,,
l n " pe.-.- My, ! c== 4 w:emo
- , n
- a. +4 pbw!> con.,
++ ', j ' ,' ccu tE ,m,,,.,., :
e,, t j B7 4 3,9 U Cr'9 8 j - *;', g 4 9a4 3 'l '"5 (" < ,', . , n2 , n
. ~-, ,- .
4 de eqw *"
* * ,5 J
' ,. n_w , 5, a + !
4 _- i ' ( ) i i i Figure 3.54 Er.ergency Feedwater System Schematic : i 4 i a >
- . - , m , _ . , , _, . - , . . - - - - , . . ,7,, , . - r
1 I 1 l 1 I 70
)
I i ( ) i AVAILABLE INFORMAfl0N BASED j 04 PUMP P7A BEthG OUT OF SERVICE l J
> 1. Most likely single failures for the Entreeney j Feedwater Systee 3
- 2. Equipnent reallgrunent for corrective maintenance on Ptep P7A l
- 3. Equipment reellenment for testing Ptap P7A .
A,
- 4. Historical causes of f ailure for Ptsp P7A and similar ptsps at AWO 1 l
1 N )
. 1 l
I \ Figure 3.55 Available Information Based on Pump P7A Being Out of Service j I i l
^
( 3 ! MOST LIKELY $1NGLE FAILUtt EVENTS FOR TPE i EMERGENCY FEEDWATER SYliEN WHEN PLMP PTA IS OUT OF SERVICE e i > 1. Ptro PTB f alls to operate <
- 2. Condensate Stpply Valve CV2800 (Ptep P78) f alls closed i
l i I select: a ccrponent to see additional single failures involving
.iuppor t systes eg.Ipaent.
J ( ) i , 1 Figure 3.56 Most Likely Single Failure Events for the
, Emergency Feedwater System when Pump P7A Is Out of Service
1 1 i 71 t ( \ I SUPPORT SYSTEM INTERFACES FOR EM(RGENCY FECDWATER PLMP P78 i Pu9 PTB ' j 4160 V AC dus A3 . Roca 100 Cooling 480 V AC MCC 851 Room 100 Cooling (Repeat) 480 V AC MCC 452 Room 100 CootIng (Repeat) Service Water Loop 1 4160 V AC Bus A3 (Repeat) 125 V DC Panet 011 Roca 109 Cooling ESAS Chamel 1 4 120 V AP Panet R$1 480 V AC MCC 856 480 V AC MCC 851 EFICS Channel A Initiation 120 V AC Panet R$1
- 125 V DC Bus 001
( ) Figure 3.57 Support Systern Interfaces for Emergency Feedwater Pump P7B O J 1 J l e I ' I; f ! l l I l ! l
72 redundant train. PRISIM will provide you the support system interface information (i.e., all the support components or functions that must be available for the components in the redundant train to be operable, including some that may not be obvious) necessary to make this verification. PRISIM only considers support functions that can slone fail either (1) a component of a front line safety system or (2) another support function that will, in turn, fail a front line safety system component. In the format used to display this information, each level of - indentation corresponds to a separate tier of support service. For example, consider this hypothetical sequence: Pump 100
- Service Water Loop X 480 V AC MCC B51 Room 100 Cooling 125 V DC Panel D11 4160 V AC Bus C5
( Here, Pump 100 is served directly by Service Water Loop X and Bus C5. Service Water Loop y is served by MCC B51 and Panel Dll. MCC B51, in turn, requires Room 100 Cooling for continued operation. Failure of any one of these supporc functions will fail Pump 100. The appearance of "(Repeat)" following an entry indicates that the support for that component or function has been developed "higher up" in the sequence and is therefore not shown again. The development . starts where the same component or function is first listed. If no support development is shown for a component or function that is not followed by "(Repeat)," no single support system failures exist for l that item.
- l l PRISIM alro provides support system interface information in a "reverse l logic sequence." You can identify a support component or function and l PRISIM will supply a table that !dentifies the front line safety l equipmer.t (and support equipment) depetident on the support component or function. The following example illustrates how you can use PRISIM to obtain support system interface information in the "reverse logic sequence."
l Sample Session In this session, assume you want to determine which components are dependent on a specific Emergency AC Power System Component (480 V AC Bus B5). You enter PRISIM through the preprocessed information ' category (Figure 3.58) and select the support system interface r , l information category indicated in Figure 3.59. On the subsequent l l screen you choose the option that will provide lists of components ) I served by a particular support system component or function (Figure j 3.60). The next screen presents the support systems for which ' information is available (Figure 3.61). You choose the "Emergency AC
73 [ TYPES OF SAFEff RELATED INFORMAfl0N , i
- 1. Updated PRA information j
> 2. Preprocessed PRA information .a i
( ) i Figure 3.56 Types of Safety-Related Information l r [ D CATEGORIES OF PREPROCESSED FRA INFSRMAfl0N
- 1. Dominant core damage sequences
- 2. Safety related systems
- 3. Safety related siksystems
- 4. Safety related corponents
> 5. S m et system interfaces
- 6. Operator actions j 7. Corponent f ailure data
- 8. System testing /survelttance l 9. Limiting conditions for operation l
- 10. System cocparisons for selected B&W plants l
( ) Figure 3.59 Categories of Preprocessed PRA Infortnation , i 1
74 ( D TYPts OF SUPPCdti SY$ FEM INTERFACE INFORMATION AVAILABLE
- 1. Support system interfaces for front line safety system ccaponents and selected swet system copponents
> 2. Lists of conponents served by a particular support systea component or function
( ) Figure 3.60 Types of Support System Interface Information Available [ ) r SUPPORT SYSTEMS FOR VNICM INFORMATION !$ AVAILABLE I
- 1. Battery and Switchgear Emergency Cooling Systees *
- 2. DC Power System b 3. Emergency AC Power System 3 4. trergency feedwater Initiation and Control System J 5. Engineered safeguards Actuation System
- 6. Service Water System k , ,
( ) i Figure 3.61 Support Systems for Which Information Is Available i l I
75 Power System" and are presented with a system component / function list (Figure 3.62). You select the component of interest ("480 V AC Bus B5"), and a table that lists the components dependent on 480 V AC Bus B5 appears on the screen (Figure 3.63). 3.2.6 Operator Actions PRISIM provides importance measures of operator actions, measures that may be used to make decisions concerning where to apply resources to a limit or reduce risk. Importance measures are provided for two types of operator actions: planned operator actions and operator recovery actions. In general, planned operator actions ranked by PRISIM are those actions that are established in ANO-1 Emergency Operating
- Procedure 1202.01 (but not specifically called "planned" actions) and that are to be carried out by the operator following the occurrence of an initiating event that causes or requires a reactor shutdown. (These operator actions are also the ones whose probability of failure can be modified by the user as described in Section 3.1.3.) Operator recovery actions presented in PRISIM are those actions taken by an operator that mitigate the effects of an accident following an initiating event and the ensuing safety-related system or component failures. The recovery actions are not explicitly established in Ernergency Operating Procedure 1202.01. (Three operator actions that are called for by Emergency Operating Procedure 1202.01 are treated as recovery actions: [1]
recovery of some Service Water System faults following a degraded power condition, [2] verification and correction of failures of the Emergency Safeguards Actuation System to actuate safety-related systems and
. components, and [3] manual actuation of the Reactor Protection System.)
PRISIM ranks planned operator actions according to their risk sensitivity and risk reduction importance measures and groups thern
. according to risk significance. Operator recovery actions are also ranked and grouped according to their appropriate importance measures.
The following sample session illustrates how to access operator action information (in this case, information on operator recovery actions) and provides the results that are obtained using this session. Sample Session In this session, assume that you want PRA-related information associated with operator recovery actions that could be taken following a failure of the Emergency Feedwater System (EFS) motor driven pump. As in previous sessions, you enter the program through the preprocessed information category (Figure 3.64) and then select the desired category of information ("Operator actions") illustrated in Figure 3.65. On the next screen you select "Operator recovery actions grouped by risk significance importance" (Figure 3.66) to obtain a list of operator recovery actions grouped according to high, moderate, and low risk significance (Figure 3.67).
76 [ EMERGENCY AC POWER SYSTEM COMPONENT / FUNCTION LIST
- 1. 4160 V AC Bus A3 7. 4160 V AC Bus A4
> 2. 480 V AC Bus 85 8. 480 Y AC Bus to
- 3. 480 Y AC MCC 851 9. 480 V AC MCC B61
- 4. 480 V AC MCC 352 10. 480 V AC MCC 862
- 5. 480 V AC MCC B53 11, 480 V AC MCC B56
- 6. 120 V AC Panet R$1 12. 120 V AC Panet R$2 e
l ( ) Figure 3.62 Ernergency AC Power System Cornponent/ Function List ( ) COMPONENTS DEPENDENT ON 480 V AC Bus 85 ,
- 1. Reactor Building Cooling Fan VSF1A e
- 2. Reactor Building Cooling Fan VSF1B
- 3. 480 V AC MCC 851
! 4 480 V AC MCC 852 i
\ * ( ) l l Figure 3.63 Components Dependent on 480 V AC Bus 35 l l 4
77 : l l l f D TYPES OF SAFEff RELATED INFORMAfl0N j
- 1. Updated PRA information
> 2. Preprocessed PRA Information e
e ( ) Figure 3.64 Types of Safety Related Information a ( ) tATECORIES OF PREPROCESSED PRA INFORMAfl0N
+ 1. Dominant core damage sequences
- 2. Safety related systems
- 3. Safety related s h ystems j 4. Safety related corponents
- 5. S e rt system interfaces 4
2
> 6. Operator actions
- 7. Corponent f ailure data
- 8. System testing / surveillance
- 9. Limiting conditions for operation J
- 10. System corporisons for selected B&W plants e
( ) Figure 3.65 Categories of Preprocessed PRA Information . 1 F
78 ( ) INFORMATION AVAILABLE FOR OPERATOR ACTIONS
- 1. Planned operator responses ranked by risk redxtion importance
- 2. Planned operator responses ranked by risk sensitivity leportance
- 3. Plamed operator responses grouped by risk significance importance
- 4 Operator recovery actions ranked by risk reduction isportance
- 5. Operator recovery actions ranked by risk sensitivity leportance
> 6. Operator recovery actions groi. ped by risk significance leportance i
a Q ) Figure 3.66 Information Available for Operator Actions ( ) OPERATOR RECOVEtt ACTIONS CRCUPED BY tlSK SIGNIFICANCE IMPotfANCE MICM tlSK SIGNIFICANCE * , testore Chitted Water Unit VCM4A Restore Chilled Water Unit VCM4B l Restore Cleset Generator 1 following loss of offsite power that parsista for longer than 8 hours Restore Dieset Generator 2 following loss of offsite power that persists for longer than 8 hours testore offsite power Restore room cooling to Maketp P g s l Restore sus Loop 1 on loss of flow che to e flow diversion to the Auxillery Cooling Water System (ACVS)
> (CONTINutD) k j Figure 3.67 Operator Recovery Actions Grouped by Risk Significance Importance j
79 l f A l OPERATOR RtCovtRY ACTIONS GROUPED BY RISE SIGNIFICANCE IlePORTANCE (CONilNUED) l hlCN RISE $1GNIFICANCE (continued) Restore SWS Loop 1 on loss of flow che to a flow diversion to the Intermediate Cooling Water System (ICWS) ! F 1 140DERATE RISE SIGNIFICANCE e saanuelty control EFS turbine driven Puip Train B on Loes of green de l 5 P0"'r ! i Open R8 S g Isolation valve CV1405 t j p Open RI Suip Isolation Velve Cv1404 Restore cooling to Room 99 i Restore cooling to Room 100 i
> Restore EFS motor driven pusp ,
(CONTINUED) ; ( ) I I I Figure 3.67 (continued) t l 4 j i .1 ; l, 1 h j , a I d 1 i j ! 1 l 2 4 j. i i I 2 + l 1 i 1 i j l i
80 You can now obtain information on the operator recovery action of ir.terest by moving the cursor to the "Restore EFS motor-driven pump" entry and pressing the return key. The new screen that appears (Figure 3.68) provides the detailed recovery information. 3.2.7 Component Failure Data Two types of information on component failures are provided by PRISIM. First, there are summaries of ANO 1 LERs, listed by component type. Second, there are comparisons of plant specific failure data and e industry averaged failure data for plant equipment. The following sample session illustrates how to access component failure data (in this case LERs) and provides information typical of A the results you would obtain if you were interested in component failure data. . Sample Session Assume you want to obtain LER information on turbine driven pumps. You obtain this information via the preprocessed information category (Figure 3.69). You then select the "Component failure data" category of preprocessed information (Figure 3.70). You will be presented a text trenu that allows selecting either LER information or comparisons of plant data with industry averaged data (Figure 3.71). Since you are interested in obtaining LER information, you select this category and then the conponent of interest (turbine driven pumps) from the next menu that appears (Figure 3.72). Summaries of LERs involving the . plant's turbine driven pumps will then appear on another screen (Figure 3.73). (Note that only the first of a sequence of screens is shown.) 3.2.8 System Testing / Surveillance
- PRISIM provides information associated with periodic testing /
surveillance requirements for safety related systems, This information is a summary of the integral and component t e s t i ng/ surveillance requirements as outlined in the ANO 1 Technical Specifications. The following sample session illustrates how to access system testing / aurveillance data. This session also provides information typical of the results you would obtain if you were interested in s ys t e rn testing / surveillance information. Sample Session Assume you want to obtain testing /surycillance requirements for the High Pressure Injection System (HPIS), You enter the program via the , preprocessed information category (Figure 3.74) and then select the j desired category of preprocessed information (Figure 3.75). On the next screen you are presented the systems for which testing / i surveillance information i .s available (Figure 3.76). You choose the ' HPIS and are first presented with a screen that presents the integral
l 4 81 j l f 3 REC 0Yttf INFCRMATION FCft AtlTORING EFS MOTOR DtlVEN PUMP This recovery action consists of restoring the Ef t mtor driven pwp to establish flow In EFS Train A. The tocation of the recovery action de-perds y the specific cause of the pwp f ailure:
$PECIFIC CAU$t CF EFS MOTOR *DtlVEN LOCATION OF RECOV!tf PUMP P75 FAILUtt ACT'ON PTB falls to start or run Unrecoverable
} g , P78 unavailable due to corrective Unrecoverable l maintenance , a PTB Discharge Check Valve plugs unrecoverable Power Cable falls open circuit Unrecoverable Circuit 3reaker (C8) falls to tr6 Local fer
> (CONilWUED) k ) l I
Figurs 3,68 Recovery Info rina tion for Restoring EFS Motor- } Driven Pump i l f 3 REC 0%Itf thF0tMAfl0W FOR tt$ tot!NG EFS MOTOR OtlVEN PUMP (CONilWUED) I
. $PECIFIC CAU$t 0F EF$ MOTOR *DtlVEN LOCAfl0N OF RECOVitf PUMP PTB FAILutt ACTION CB Control Circuit falls Control Roce j The time avaltable for restoring the EF$ m tor driven pw p without ; frreversible damage Is about 20 minutes. Recovery, where it takes i
place, decreases the estimated f reQJenCy of the scenarios by a f actor of 1.8. This recovery action is taken in scenarlos 15 2 and 15 3. 1 i i r j
- l a
1 : ) { i .l ; e ( ) [ i 1 j Figure 3,68 (continued) i d i d i I l ! I i'1 ! 1 i l
82 ! [ D > 4 TYPES OF SAFEff* BELATED INF0eMATION
- 1. Updated PAA information
> 2. Preprocessed PRA information
- i P
1 I ( ) r Figure 3.69 Types of Safety Related Information ] ( 3 ' CATEGORIES OF PREPROCESSED PRA INFORMATION j 1. Ocainant core damage segsences . I
< 2. Safety *related systems
- 3. Safety related sdarstems l ,
- 4. Safety related c a ts
- 5. S w rt system interfaces ;
- 6. Operator actions 4
, > T. Ccoponent f atture data
- 8. System testing / surveillance j 9. Limiting tenditions for operation *
- 10. System ccrparisons for selected B&W plants i
1, N ) l 1 : Figure 3.70 Categories of Preprocessed PRA Information l l i i t
l 83 i
, [
COMPONENT FAILUtt CATA AVAILAh! l l > 1. A listing of plant LERs
- 2. A conparison of plant date with Irdntry averaged date d
1
- i l
l I .i i e I 3 l c I 3.; ( ) i i Figure 3.71 Corsponent Failure Data Available r
}
i h [ ' I COMPONENT CAttG0ttt$ 704 VHICW PLANT *$PECIFIC Ltte Ast AVAILABLE ;
}
. 1. Batteries
- 2. Buses i
- 3. Check Valves ,
I 4. Circuit Breakers t , t
- 5. Olesel Generators
- 6. Fans T. Motor Driven Ptrps ;
- 8. Motor Operated Yalves
- 9. PnetretIc/HydrsuliC V4Lves
! 10. Relief Valves j 1 :
, W (CONTlWTO)
( ) , i I i ! j Figure 3.72 Cornponent Categories for Which Plant. Specific LERst Are Available t t 1 r n e_.____ _ , . - , __
u l 84 f i ( ' 3 CCse0NENT CAf tGotits FOR WICW PLANT stECIFIC Lits AAt AVAILABLE (CONTlWUtD) 4
> 11. Turbine Oriven Pwes e
i r 1 i ( j
! Figure 3.72 (continued) ,
I L Lits lWVOLvihG TUettht 0tivtu PUMPS . LER f Daft $YstEM/ COMP 0htWT
- DESCalPfl0N l
85 008 08/19/85 Main feedwater/ PIB trt@ed on over-Turbine *0 riven Pwp speed dJe to the f ei t a PIB ute of a drive shaft j j cowling to the speed , control governor shaft ( 85 007 08/11/85 Main Feedwater/ PIB tripped dJe to a i Turbine Oriven Pwp high thrust bearing war PIB IMica tion 1
! 85 001 01/07/85 Emergercy Feedwater/ PTA failed to develop :
Turbint Driven Pwp the required flow ard i 1 ! { PTA discharge daring test. l ing. The determined . j < cause was low steam flow to the turbine driver, - Alch was the result of Inadegante modification j (CONTlWUED)
\ A 1
i s I Figure 3.73 LERs Involving Turbine. Driven Ptmps ' d 1
d l l 1 85 l t a ! [ ffPts of SAFiff RtLATED lhF0eMil0N ' l 1. Updated PtA Information !
- > 2. Preprocessed FRA information ,
! i i l
.i ;
f i f 1 i ( ) : i
! I l , Figure 3.74 Types of Safety Related Infortnation 4
t I h a r ( ) CAftGotit$ of PREPt0CES$t0 PRA thf0RMAfl0N !
, 1. Ocninant core damage segences
- 2. Safety related systems
]
- 3. Safety related s h ystems .
] 4 Safety related eteponents 1
- 5. tert system interf aces
] 5
- 6. Operator actions <
l
- 7. Cceponent f at ture data j l
3 i
> 8. System testing /survelltance !
I
- 9 Limiting conditions for operation I
l 1
- 10. System ccrparisons for selected $1W plants i i
( ) J l t ! i ! Figure 3.75 Categories of Preprocessed PRA Information I I i j l l
i 86- ) ) t I ,I r 0 ( ) ! SAFEffettLAft0 sistgm8 Pot WICM ltFoteuflou Alt 0CIAfte ' ) Wifu piticett itsflWG/9URytILLANCE titulttagets is AVAILA4LE EntRGiWCY COOLING SYSites I h 1. Nigh Pressure Injection Systes ! [ a i
- 2. Low Pressure injectlen System
- 3. Core Floodtg System 4 toergency feedseter System RfActot DVIL0thG C00LlWG SY$ttet
- 1. teactor Sulldine Cooline Systen '
- 2. Reactor Building Spray Systes
)
3 (Cout!wuto)
- 4( ) I i <
i rigure 3.76 Safety Related Systems for Which Infornation { Associated with Periodic Te s ting / Surveillance i Requirements Is Available , i < I f i ! i i ; ! l + 4 1 4 4 i I I ( , l l 1 ) ! i , I I ) I
i I 87 i I system testing / surveillance requirements (Figure 3.77) and then a screen that presents the component testing / surveillance requirements (Figure 3.77 [ continued)). ] 3.2.9 Limiting Conditions for Operation This PRISIM option provides information on the limiting conditions for i operation (LCOs) for safety related systems and descriptions of the 6 operator actions required if the LCOs are not met. If you select one r
. of the safety related instrumentation systems (the Engineered l' Safeguards Actuation System or the Emergency Feedwater Initiation and Control System), you are presented with a menu that contains the :
various functions of the chosen instrumentation system. From this
. menu, you can select system functions of interest to see LCOs related ;
to those functions. I The following sample session illustrates how to access LCO information and provides information typical of the results you would obtain if you ' were interested in LCOs. Samole Session For this example, assume you want to obtain LCO information associated with the Emergency Feedvater Initiation and Control System. , Specifically, you desire the LCOs associated with monitoring the Main Steam Line Isolation function. You enter PRISIM via the preprocessed j information category (Figure 3.78) awd then select the desired category of preprocessed information (Figure 3.79). You are then presented with a list. of safety related systems for which LCO information f.s available (Figure 3.80). After selecting the Energency Feedwater Initiation and , control System (EFICS), you will see the screen presented in Figure ;
. 3.81.
j As indicated in Figure 3.81, you request LCO information concerned with l the Main Steam Line Isolation function. You are then presented with a ! ! summary of the LCOs associated with this function (Figure 3,82). Note I that you may return to the LCO System Menu or the EFICS Function Menu [ j if you desire. For this case, you choose to review a summary of the j
; operator actions required if the minimum specified conditions cannot be ;
I met (Figure 3.83). 3.2.10 Systen Comparisons for Selected B&W Plants , The PRISIM data base contains system configuration comparisons for ) j selected Babcock and Wilcox (B&W) plants. The information was taken t from an Accident Sequence Evaluation Program report entitled ASEP Plant j Survey and Initial Plant Grouoine Letter Report and dated December 22, i 1983. B6W plants that were modeled in this program have been included [ in the system comparisons. (Except for ANO 1, JBF Associates has not j independently verified any of this plant specific information.) [ I t
l 1 88 i 1 i ( D PttlCCIC TIstlhC/$UtvilLLAWCE Rt0UlttMEuf 8**MIGN Pitt$Utt INJECfl04 $Y$f tM i i luf tceAL fElf thG/$URYtlLLAhCE ! Cexe every 18 smanths the hP18 shall be tested to demonstrate it le operable. A test signal allt be applied to demonstrate the l actuation of the systee for emergency core cooling operation. I the test wlLL be ccesidered satisf actory if control board Indicotton . verlfles that ett appropriate sup breakers have opened or closed ard ELL velves have completed their travet. ; e I l 1
> (CC%flWUt0) '
( ) ! 1 Figure 3.77 Periodic Testing / Surveillance Requirerer , -h'Sh Pressure Injection Systern i i r 3 Pitt0Dic filflWC/$URytlLLAttt I AtoulttMENTS *klGM 7 tit $Uet INJECfl0N SY$f tM (CONilWUED) I I
- i COMP 0hthf filfth3/$UtytlLLAktt I
PtMPS i The NPil pwpe shall be tested gaerterly to verify proper start-Lo *M operation. Acceptable performance will te iMicated if l the pep starts ard operates for il minutes *M the discharge pressure aM flow are m' thin a 101L of acceptable limits. j VALvis l All t$f valves in the NPil ard all associated t$F valvet in the
$WS designed to furstion in the event of a LOCA shall be tested quarterly. Acceptable performance will be indicated if appropriate eetton is trdicated soon actuaticn. .
(ttfutu TO MW'J1 ( ) Figure 3.77 (continued)
89 (
~
h TYPES OF SAFETY RELATED INFORMAfl0N
- 1. Updated PRA information
> 2. Preprocessed PRA information .e 3
( ) Figure 3.78 Types of Safety-Related Information [ CATEGORIES OF PREPROCESSED PRA INFORMATION .
- 1. Dominant core damage sequences
- 2. Safety related systems
- 3. Safety related subsystems
- 4. Safety related ccrponents
- 5. Support system interfaces
- 6. Operator actions
- 7. Corrponent f ailure date
- 8. System testing /surveltlance
> 9. Limiting conditions for operation
- 10. System emparisons for selected B&W plants
( ) Figure 3.79 Categories of Preprocessed PRA Infortnation
l l l l 90 l l [ \ \ l SYSTEMS FOR WHICH LIMITING CON 0!T10NS l FOR OFERATION INFORMATION IS AVAILABLE
)
- 1. Reactor Coolant System
- 2. Emergency Core Cooling, Reactor Building Cooling, and Reactoe '
Building Spray Systems ; l
- 3. Steam and Power Conversion System
- 4. Reactor Protection System l
- NOTE
- This section contains selected LCO's which have been ,
abstracted from the ANO 1 Technical Specifications , t and formatted for use as reference information in 1 the PRISIM program. However, issues related to LC0's should alwa/s be resolved by consulting the ANO 1 Technical Specifications.
> (CONTINUED) k )
l Figure 3.80 Systems for Which Limiting conditions for ; I Operation Information Is Available ( h l SYSTEMS FOR WHICH LIMITING CONDITIONS FOR OPERATION INFORMATION !$ AVAILABLE (CONTINUCD) ) j
- 5. Engineered Safeguards Actuation Syste,
> 6. Emergency Feedwater Initiation and Control System
- 7. Auxiliary Electrical Systems
- 8. Other Safety Related Systems
- NOTE
- This section contains selected LCO's which have been abstracted from the ANO 1 Technical Specifications and formatted for use as reference information in the PRISIM program. However, issues related to LCO's should always be resolved by consulting the ANO 1 L =
Technical Specifications. i
.J Q )
Figure 3.80 (continued)
91 [ D EMERGENCY FEEDWATER INITIATION AND CONTROL SYSTEM AVAILABLE INFORMATION REGARDING LIMITING CONDITIONS FOR OPERATION SYSTEM FUNCTION Emergency feedwater Initiation
> Main Steam Line Isolation 4
LCO SYSTEM MENU ( ) Figure 3.81 Emergency Feedwater Initiation and Control System Available Information Regarding Limiting Conditions for Operation ( ) EFICS MAIN STEAM LINE ISOLATION CHANNELS LIMITING CONDITIONS FOR OPERATION e NUMBER CHANNELS MINIMUM MINIMUM OF FOR STSTEM OPERABLE DEGREE OF FUNCTIONAL UNIT CHANNELS TRIP CHANNELS REDUNDANCY Manual isolation 2/SG 1/SG 2/SG 1 Low pressure steam generator 4/SG 2/SG 2/SG 1
=
h REQUIRED ACTION LCO SYSTEM MENU EFICS FUNCTION MENU ( ) ' Figure 3.82 EFICS Main Steam Line Isolation Channels Limiting Conditions for Operation 1 i
92 ( REQUIRrJ OPEk/ TOR ACTION )I
- 1. Initiate a shutdown using norrnal operating instructions and place the reactor in the hot shutdown coMition within 12 hours.
- 2. Isotation on low steam generator pressure function may be bypassed below 750 psig 0T50 pressure. This bypass is automatically removed when the pressure exceeds 750 pels.
S t LCO SYSTEM MENU EFICS FUWCTION MENU ( ) Figure 3.83 Required Operator Action 9 9 4
- - , - , - . . ~ . . , . , - , - ~ , , , . . _ , - , . . , , . - - ~ - , - - . , r -. + - , - - , - - - . - -,-
1 93 1 l l This safety related information category of PRISIM provides a brief summary of the similarities and differences in several safety-related systems contained in the selected B&W plants. The comparisons will be helpful in evaluating the impact potential regulatory changes may have on the various B&W plant configurations. The following sample session illustrates how to access this information category and provides information typical of the results you would obtain if you were interested in system comparisons of B&W plants. Sample Sess_i_oa For this example, you want to determine if there are any operating ' plants that have the same high pressure inj ection system (HPIS) configuration as ANO-1. As with previous examples, you enter the program and make a selection from the Master Menu (Figure 3.84). In this specific case, you choose the "System comparisons for selected B&W plants" category of preprocessed information (Figure 3.85). You are then presented with a menu screen that allows you to select the system configuration of interest (Figure 3.86). As shown in the figure, you choose the HPIS option cnd are presented with a summary of the first design group (Figure 3.87). You continue through the HPIS design groups (Figure 3.87) until you find the group which contains ANO-1. From the last screen, you discover that ANO-1 and Oconee Units 1, 2, 6 3 have similar HPIS configurations. This completes the description and presentation of sample sessions of safety-related information provided by the PRISIM program. The
- appendixes that follow provide additional information on the use of the PRISIM computer program. Appendix A explains how to obtain copies of the graphics screens provided by PRISIM. Appendix B details the information provided in this version of the PRISIM program. Appendix C describes the models used to quantify failure events in the ANO-1 PRA study. (These same models are used when PRA parameters are changed.)
l
. I e
96 (' h TYPES OF SAFETY RELATED INFORMATION
- 1. Updated PRA Information
> 2. Preprocessed PRA information 3
( ) Figure 3.84 Types of Safety Related Information ( ) ' CATEGORIES OF PREPROCESSED PRA INFORMATION
- 1. Deninant core danage sequences
- 2. Safety related systems
- 3. Safety related subsystems
- 4. Safety related cceponents
- 5. S w rt system interfaces
- 6. Operator actions
- 7. Cceponent f ailure data
- 8. System testing / surveillance
- 9. Limiting conditions for operation *
> 10. System corparisons for selected B&V plants k 1 Figure 3.85 Categories of Preprocessed PRA Information
l l I l 95 ) [ D SYSTEM CONFIGURATION COMPARISONS WHICH ARE I AVAILABLE FOR SELECTED BABCOCK AND WILCOX PLANTS
- 1. Containment configurations
> 2. High Pressure Injection System configurations
- 3. Power Operated Relief Valve configurations
- 4. Emergency Feedwater System configurations
- 5. Electric Power System configurations
,
- NOTE
- The information in this section was taken from an Accident Sequence Evaluation Program report entitled "ASEP Plant Survey and Initial Plant Grouping Letter Report" dated Decenber 22, l983. B&W plants that were modelled in this program have been included in the system conparisons.
Except for ANO 1, however, JBF Associates has not indepen-dently verified any of this plant specific information. ( ) Figure 3.86 System Configuration Comparisons Which Are ' Available for Selected Babcock and Wilcox Plants [ D HIGH PRESSURE INJECTION SYSTEM CONFIGURATIONS USED IN SELECTED BABC0CK AND WILCOX PLANTS e DESIGN HPIS APPLICABLE GROUP CONFIGURATION PLAKTS 1 Three motor driven ptmps WNP 1 & 4 Cecinon suction for two Crystal River 3 motor driven ptrps separate suction for third motor driven ptro Normally closed suction valves o Normally closed discharge valves
, W (CONTINUED)
- Information not verified by JBF Associates
- j Figure 3.87 liigh Pressure Injection System Configurations Used in Selected Babcock and Wilcox Plants
l 96 I 1 [ HICH PRESSURE INJECTION SYSTEM CONFIGURATIONS USED IN SELECTED BABCOCK AND WILCOX PLANTS (CONTINUED) DESIGN HPIS APPLICABLE CRCUP CONFIGURATION PLANTS 2 Three motor driven ptrps Bellefonte 1 & 2 Comon suction for two TMI 1 motor driven ptmps - l Separate suction for third motor driven ptmp , 1 Normally open suction 9 valves Normally open discharge l j valves
> (CONTINUED)
- Information not verffled by JBF Associates
- j Figure 3.87 (continued) l
\
r 3 ' NICH PRESSURE INJECTION SYSTEM 1 CONFIGURATIONS USED IN SELECTED BABCOCK AND WILCOX PLANTS (CONTINUED) l DESIGN HPl$ APPLICABLE CROUP CONFIGURATION PLANTS 3 Three motor driven ptrps Oconee 1, 2, & 3 Comon suction for all ANO 1 three motor driven ptrps Normally closed suction valves Normally closed discharge valves o (CONTINUED) *
- Information not verified by J8F Associates *
( j Figure 3.87 (continued)
AI APPENDIX A OBTAINING IIARD COPIES OF CRAPilICS SCREEFS l PROVIDED BY PRISIM i 1 m To obtain a hard copy of a graphics screen provided by PRISIM, you must make a photocopy of the appropriate plot of the screen image (schematic) provided in this appendix. You can quickly find the plot of the desired e screen by bringing the screen up on your monitor and pressing the "Esc" key. The control screen will appear, and, at the bottom of the screen, the picture number of the desired sc ,en will be displayed. Use this picture number to find the location of .he appropriate schematic in this appendix. Schematics are arranged according to ascending picture number, beginning with 9002 and ending with 9146, with intentional gaps to allow for the inclusion of other schematics at a later date. a. 9 1 l
LOW PRESSURE INJECT ON SYSTEM SCHEMABC A REACTOR DH10 - swAss B VESSEL DH10 JL BYPASS
- O -
FROM DH8A = [4 <
- =
Mg)i( W l,. FROM DH88 FROM FROM HTX 37 CFT CFT BYPASS J J T2B T2A TO TO
).(
m 4 , E35A ll HP ' TO RB - - P34A SPRAY)k o /' ?4 k CV1405 > 1 Pq CV1410 j . CV1 A k FROM RCS _ Sn 5, 52 Yy HOT LEG
' Cb050 Cb404 TO TO BWST gp HU W 3 7 BYPASS TO RB SPRAY BYPASS E358 j d 4L AL P34B I .
i CV1400 CV1429 CV1406 ! TO BWST 4 [4 Q i TO HP < CM 415 l /L HTX BYPASS l
+ * . * , ,
4 HIGH PRESSURE INJECTION SYSTEM SCHEMATIC A A To RCs JL JL _.._...._._. COLD LEGS h DWST T3 f _ l LO. h b g b
-s m
CV1407 y CV1408 y TO LP TO LP o )( )( )( )( )( AND R8 W AND RB W :> o SPRAY PUMPS SPRAY PUMPS w
]y CV1220 y CV1228 C CV1227 y CV1219
! g A
. T g "Jm
' , _.FROM LP COOLER I
Q E35A DISCHARGE J , P36B k ""
- CV1235 C 1r 'M Q k
- <
TANK T4 j TO RCS PUMP 3 l P36C SEAL HEADERS ;hL ,
;Lh J2 m FROM LP COOLER E358 DISCHARCE SYSTEM MENU END OF INPUT
LOW PRESSURE INJECBON SYSTEM SCHEMABC A REACTOR DH10 - BYP^SS BWST VESSEL DH10 A BYPASS FROM DH8A : y't 5 FRCM '/t- FRCM FROM DHS8 CFT T2B
'/t3>( CFT T2A B SS TO
^ E L
TO y E35A A gp # TO R8 - P34A SPRAY b
)( >(
$ i 5 b CV1405 I y
g CV1401 CV1428 g ^gg> g p 9, 9, CV1410 9,
$ ' CV1414 SUMP /
, FRou RCS . '
Y y hot ec ' CG o'50 CE4o4 TO TO BWST gp HTX HTX BYPASS E35B BYPASS TO RB SPRAY A A P348 q C 4C0 CV1429 TO BWST e k < TO HP c CV1415 A HTX BYPASS SYSTEM MENU END OF INPUT
LOW PRESSURE HEAT EXCHANGER E35A SECONDARY COOLING SCHEMATIC E35A FROM ROOM COOLER WCI A. ROOM COOLER WC18 L O COOLER E50A AND L 0 COOLER E47A FROM e OTHER g SWS >* u LOADS , TO EstERGENCY {CCv3822 >S % CV3823 POND TO OTHER SWS < LOADS CV3824h FROM SWS LOOP 1 V TO DISCHARGE FLUME SYSTEM MENU END OF INPUT
l LOW PRESSURE HEAT EXCHANGER E35B SECONDARY COOLING SCHEMATIC i E350 i g m FROM ROOM COOLER WCI A. ROOM COOLER WC18 L 0 COOLER E50A ANO L O COOLER E47A FROM e OTHER m g Sws co LOADS 1f ' - O MOCY JL B CV3821
- CV"38'23 PONO TO OTHER SWS
LOADS CV3824h ' JL FROM SWS LOOP 2 P TO DISCHARCE FLUME i 6 4 SYSTEM MENU END OF INPUT
. _ _ _ . - - _ - . _ . - - - - - - . - - --- e ~ -e - - -
T U P N I F O 9 D 1 N 4 E 2 V C T B O F 2 C T c N T P N E L E R M H E a, IN P , M O n () ,_ AS R TO F NM OT CA M'
- r. X 4
_-- e R O T E L x C S AE EV S R y W(,_ Vs T N E C E R E P I I T MH NP A IASa A TO f O I) M R NJ F E OT H CA C S " M $ T A 4 E T AT F 2 C T
\-
5 S 1 U Y 4 N 2 E S V M C G M N I E T D S O Y
- O S L
F E R O C go
. _m._ _ _ _ ..m. . . _ . _ _ _ _ .
t l l J EMERGENCY FEEDWATER SYSTEM SCHEMATIC l FROM V y FROM g 9 SERVICE SERVICE 4 h TO WATER WATER CV2676 CV2668 ATM LOOP 1 Q Loop 2 CV3850 CV3851 ! -H- CV2646 STEAM p7g CV2800 7 CEN M j '/d , l CV2670 ' i E24A e' TO CONDENSATE V CV2645 j TRANSFER PUMPS CV2667 CV2803 =
- CV2627 g<
Sv2663 CV2626 s CS281 4 O
>4 PSV-6602 CS280 QCST g CV2613 b SYSTEM MENU END OFINPUT t 4 S O O O
. . . , e e FUNCTIONAL BLOCK DIAGRAM OF THE REACTOR PROTECTION SYSTEM MO4 RC TDAP NI - A KA1 low RC PRESS &-A A MagPaE55 [ _- ^ BISTABLE kA2 KB1 *[ C s
- _c,-A STRING KA3 KC1 "N rouet/PuuPS PP-A "
& A y g
^ A 2&4 gyAD*,, _" , xA4 _ _
KD1 , qBREAxER rtronarca rw - ^ 1 OF 10 gg1 gg aw rr-A SECONDARY BUS 1 MT - o K81 l t.r - e - -
} REACTOR e ii Ill '
e-e xe2 te - a BISTABLE 0-- - UN N JL JL E2 r2
<r - o STRING x83 KC7 - J
- rP - e O 8 m
8 xa4 KD2 2 OF 4
", - ', CAE CAE oms m -a 1 OF 10 c2 m o tr - a BRK E4 - _. F4 O C' g
r nr - c xci c2 r C I BRx {E REGULATING REGULATING F} Y e O m _,r PO6 PO 6 [' '- C BISTABLE xc2 K03 RE , ACTOR v tc - c iss SUPPUES SUPPUES ce-c STRING xC3 um - " ~ re - L rr-c C
} c l xc4 2 OF 4
_e-c m KO3 } . D2 [ N-C 1 OF 10 tr - c BRK D1
,, - o xo, 2-c - - -
sC 87 a v-o } cr on h[f PP - o dN )o3 c} O -
~ ~ ~
e- 4 1 OF 10 0 RELAY [ _- l SAFETY ROD GitGUPS REGULATING ROO GROUPS SYSTEM MENU END OF INPUT 1
. . _ _ _ _ _ _ _ . - _ _ _ . _ _ . _ _ _ _ _ _ - _ - _ . _ - _ _. _ -- .-. -- -_-- -.. - - - _ - - - - - - _ - - -. .~. -- - -
- - - , . ~ . . ,
BLOCK DIAGRAM OF THE ENGINEERED SAFEGUARDS ACTUATION SYSTEM RC 'm PRESSURE ANALOG SUBSYSTEM RS ' m NO. 1 _ PRESSURE 3 C CHANNEL 1 DIGITAL SUBSYSTEM CHANNEL 3 NO. 1 p] CHANNEL 5 POWER SUPPLY CHANNEL 7 RC T _ syu b C CHANNEL 9 e PRESSURE ANALOG SUBSYSTEM >
" RB 'm NO.2 suu L PRESSURE i C CHANNEL 2
,< DIGITAL
'm SUBSYSTEM CHANNEL 4
'} NO.2 P-
>C CHANNEL 6
- CHANNEL 8 POWER e PPLY c: 5 C CHANNEL 10 PRESSURE ANALOG SUBSYSTEM R8 7 'm NO. 3 PRESSURE SYSTEM MENU END OF INPUT l
l - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ .
l. SERVICE WATER SYSTEM PUMP TRAINS SCHEMATIC TO INTERMEDIATE COOUNG WATER SYSTDA O CV3811 LOOP 2 = % M P4C ASSUMED j LOADS CV3641 RUNNING CV3642 h CV3640 e AUX 1UARY P48 :>
' O COOUNG 4 k M ASSUMED r
- l WATER SYSTEM CV3643 RUNNING h CV3644 1
CV3646 CV3645 ASSUMED k M I LOOP 1 O W LOADS STANDBY CV3820 TO INTERMEDIATE ' V CoouMG WATER SYSTEM l
' END OFINPUT SYSTEM MENU i
4
._______m. __r-'- - . . , . -. -----=vT-,3,-ww-- - - -r w-- -- er - e w e, - . - - wr --- - -,,-ww-
PIPE SEGMENT SCHEMATIC FOR SERVICE WATER COOLING OF DIESEL GENERATOR DG1 l l I COOLER E20A l FROM e OTHER m c - sws >
$ DG 1 LOADS CV3823 ,'.
p TO w CV3806IO 4 > EMERGENCY TO OuiER POND SWS 1 O CV3824 LOADS n y FROM TO SWS LOOP 1 DISCHARGE FLUME SYSTEM MENU END OF INPUT
PIPE SEGMENT SCHEMATIC FOR SERVICE WATER COOLING OF DIESEL GENERATOR DG2 COOLER E208 l FROM >
" OmER W
SWS DG 2 LOADS CV3823 CV3807h 9> y4 ERGENCY POND TO OTHER 824 LO DS u y FROM TO SWS LOOP 2 D!SCHARGE FLUME SYSTEM MENU END OF INPUT
= . . . . .
IndNI .30 ON3 DN3n E M E0A 10A mowm Aniva (' oN('i oN (' I (' mouvm Aniv8 t00 998 C00 990 SGGN1 SGH3ANI tZA ZZA 'd j
'# CIA ttA C98 CS8 (I _ . _ _ _
(I I (I '" T (I I Z90 198 19 8 ISO (4 (9 (p 9 4 Y Y Y Y (Y (Y g 98 4 s0 , < on (4 U g(4 4 oN C 4 g *
- mm
- mm
;q l i T Y Y Y Y T c c c c c c t Y Y Y Y Y 83 mod ON
+
90t
+
ON
+ - t BOCevt 83%Od 31s33o 01 (4 (4 (4 ON(4ON(4 {4 31s33o 01
** ~
Z oo Y ro Y 011VW3HOS W31SAS lVOl813373 DV ADN3083W3
a- m s-. s a n-- A 4 m b 10dNI .30 ON3 ON3N W31SAS 190 NOH3 198 MOH3 198 n0HJ 198 NOH3 l l I I EvH tvH (I (I (I (I LEO 110 ESH - tSH CSH ISH ON OY A DEL DY A OZ1 y DY A OZ1 OY A OZ1 ON wnNYn V
"I ZZA 43rA -v Ct[b tt[b C C CCC v C CC C C
{ ZOO 100 { C C i (i "C 00 A SZ1 ' 30 A SZll l00 A SEE 30 A Stt __ n 2, ir00 C00 000
~~
OY A 087 OY A 09t OY A 09t Z 900 100 Z
~
(l 190 (l 9S0 ([ ICO ~ Ol1VW3HOS W31SAS 83 mod 30
l l NORTH BATTERY AND SYaTCHGEAR ROOMS EMERGENCY COOLING SYSTEM SCHEMATIC - ROOMS 99, 95. AND 149 PVC6051 I AUXIUARY BLDG. T 4 k INSTRUMENT AIR k k< CHILL WATER 4- M h $mCV6034 UNIT TO SWS VCH4A g ^ DISCHARGE m l FROM S $ LOOP 2 k FREON UNIT l FOR ROOM 99 ROOM 95 ROOM 149 NORTH BATTERY ROOM NORTH NORTH NORTH I VE18 M CHGEAR BATTERY ELECTRICAL ROOM ROOM ROOM SYSTEM MENU END OF INPUT l
SOUTH BATTERY AND SWITCHGEAR ROOMS EMERGENCY COOLING SYSTEM SCHEMATIC - ROOMS 100,109. AND 104 l PVC6050 AuXIUARY DLDG. k 1NSTRUMENT AIR CHILL WATER 4 "M h
~
$mCV6036 UNIT vCH4e TO SWS g 4 DISCHARGE
=
FROM SWS e LOOP 1 0
* ?
C FREON UNIT FOR ROOM 100 ROOM 109 ROOM 104 SOUTH BATTERY ROOM sOum sOum soum VEIA s%1TCHCEAR DATTE.RY ELECTRICAL ROOM ROOM ROOM l l SYSTEM MENU END OF INPUT l l
.m.__ .__m-._ _ . _ _ . . _ . _ . . . _ _ _ _ . m. . _ _ _ _ _ . . . _ . _ - . m -- -- -
HIGH PRESSURE INJECTION SYSTEM SCHEMATIC A h TO RCS & h -^^^^^^^^-
' COLD LEGS BWST l T3
~
B '" B t B Q B
O }j{ X X X )( SPRAY PUMPS l' O ' SPRAY PUMPS j h CV1228 O CV1227 y CV1220 pCV1219 - - i P36A I
% ' T
% FROM LP COOLER
%m ' E35A DISCHARGE 3r Ak' P36B CV1235 % 'g 'dT % e FROM MAKEUP TANK T4
- 1
] 1P
'4 TO RCS PUMP # P36C I "k
SEAL HEADERS
* 'A .) R ' a m m FROM LP COOLER 4
' ' ' E358 DISCHARGE
REACTOR BUILDING SPRAY INJECTION SYSTEM SCHEMATIC BWST T3 g I I LP \\ TO LPS TO HPS REC 1RCULATION JL JL P358 d ! gr-X-> 4 k l , M CV2400 l CV14088 b 84 8<: ? CW406 CW 415 C TO LPS g -------- P35A CV1407A mr vi-x d
- I .p M y ?<
/
)f TO HPS CV1405 CV1414 c____________________ . _ _ _ _ _
EMERGENCY FEEDWATER SYSTEM SCHEMATIC FROM V y FROM SERVICE SERVICE g'
> TO WATER WATER CV2676 CV2668 ATu LOOP 1
_h CV3850 CV3851 LOOP 2
-H- CV2646 80 STEAM P78
--t 7
GEN Vq M
/ E24A CV2670
/
/ TO CONDENSATE V CV2645 / TRANSFER PUMPS CV2667 O CV2803 y e
Cv2627 C ,,,,,,, 8 m O' O
" ~~ ~~
Cb619 CV2618 ATM. A -O Cv2806 VM SEAM P7A 's ' W CEN W ' TO/FROM CV2617 CV2647 CV2802 g UNIT 2 CST l CV2663 V CV2548 Sv2663 s CS281 CV2526 4 CS280 OCST
>4 PSV-6602 C h 3 v2e13 l
l
REACTOR BUILOING COOLING SYSTEM SCHEMATIC RADIApon CAMPER ^'^au ^~o FROM Sppy cv747D INTERLOCK W1 A 0 RB > ATMOSPHERE
' VMW 3 g; N
\
CV3812 RADIATlO." FROM 8; CTOR g SW = r3 814R LOOP 1 I CV3814 FROM ePOD CV7471 VSF1D gg . ATMOSPHERE o N e \/ SUPPLY TO e L.> FROM DAMPER BPDC
> A:R
"'t"uu
=
RB
, VSFIC 0 cg2 ATMOSPHERE 3
f CV3813 CV3815 FROM S, RADIATION SW = # F7 % >9' DETECTOR LOOP 2 815R DAMPEg GPDD cv747s FROM RB - N g vSrso 0 s ATMOSPHERE D 0 RADIATION ALARM AND INTERLOCK
g_..._________. - _ _ . . . _ _ _ . _ . _ - _ . . __ _- ._ - t SERVICE WATER SYSTEM PUMP TRAINS SCHEMATIC TO INTERMEDIATE i COOUNG WATER SYSTEM f , TO SWS LOOP 2 LOADS
= k % Z P4C ASSUMED CV3641 RUNNING
,I CV3642
\
.CV3640 ,,
e AUXIUARY > O COOUNG -c % M P48 A N ED pj ! WATER SYSTEM CV3643 CV3644 RUNNING P4A TO SWS M ASSUMED LOOP 1 LOADS
.IN r-STANDBY CV3820 c
TO INTERMEDIATE Y COOUNG WATER SYSTEM i 4 4 i I 6 D g Y
ZOA 10A N30WO AH2L1Y8 ( oN(I oN(I ( H30WW AH311Y8 tou CSG C00 SH31N3AN! 9SS SH2183AN! tZA ZZA lJ g
'd CIA ttA I C98 CS8 I I
Z98 I "" T I I 4 198 y p p g LS8 ZS9 c C c c = Y_e- Y Y Y c+ se c+f a c+ on4 c+ oN c+ c+ a
.x g ,
6
.x g . > ->
I c+Tc+hc+T c+'c+h c+Y agdo1 c+ "os3 c+ at c+ oN oN c t A tvt oN c, noe c 31 83;gdo1 ZY ' tY Z oo Y t so Y 011VW3HOS W31SAS 7VO!810313 DV ADN3083W3
168 MORF 16B MORF 158 MORF 15B MORF I 520 51D I I I g g g 2AR 1AR l 12D 11D 2SR 4SR 3SR 1SR ON CA V 021 CA V 021 CA V 021 CA V 021 O LAUNAu O N___ m rI p _42Y m 22Y _O M l, 11Y 1
) )) ) h.
)' -
) )) ) )'
20D f 10D f
) ) ) os )
-- CD V 521 CD V 521l lCD V 521 CD V 521 l CL CL '
40D 50D 30D 4 CA V 084 CA V 084 CA V 084 z sod 700 r
)I 15 8 )l 65D )l 15D --
CITAMEHCS METSYS REWOP CD
EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1219 To RCs Coto Ltcs MU1232 ((hijih{ THESE VALVES ARE MU1212 CLO5ED TO REMOVE TO INJECTION , g - VALVE CV1219 FROM UNE A MU1224 SERV;CE AND ARE
. . . . . . OPENED TO RETURN TO INJECTION n , :::g:f IT TO SERV.CE UNE B ;:MUt223I!!
MU1213 480 V AC MCC B51 MU1211 $ ),( CB 5152 ) ) C8 5151 X y e L - CV1220
][C D][ CV1213 I
..' h i T c oR$EEN
.. .)!:'49.39A!i!5 MU19^
TANK P36A ii.tU.2.
- i. . . .S.ji!. ..
uu25 1I (CONTINUED)
i EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1219 (CONTINUED) l l TO RCS COLD LECS l
A A MU1232 MU1231 l I MU1212 TO INJECT 10f4 , g -
l UNE A uuj224 TO INJECTION
w l l u cv1220 g}cv1219
{C i
RACKED OUT TO REMOVE % M MU20A MU19A b i T
[OR M VALVE CV1219 FROM TANX SERV!CE At:0 LEFT 6A UU26 OPEN Iti STAT DDY TO RETURN IT TO SER%1CE MU25 Y
TO TRAIN 8 l
l
EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1220 TO RCS COLD LEGS MU1232 4 MU1231 i MU1212 TO IMJECTICN ____
~
UNE A g ypygy4;. TO INJECTION UNE B
'_ <s q M51223 l MU1213 480 V AC uCC 051 MU1211 e 8 O
- k CD 5152 ,) ) CD 5151 );( Y l l "
CV1220 fC D{CY1219 1 i FROM BWST THESE VALVES ARE CLOSED TO REMOW I
..$ bh '
f.MU20.A3j MU19A LV T e OR WAXEUP TANK YALVE CV1220 .90u ;igyffi![..C P36A SERMCC AND ARE ' ' ' ' ' ' ' ~ - ' OPENED TO RERJRN IT TO SERvCE MU25 h lf (CONTINUED)
EQUIPMENT REAUGNMENT FOR CORRECTIVE MA!NTENANCE ON VALVE CV1220 (CONTINUED) TO RCS cot.D LEGS MU1232 4 MU1231 -{ RACKED OUT TO REuCVE MU1212 VALVE CV1220 FROM
~
TO INJECT;0N - - . . SERMCE AND LEFT UNE A MU1224 OPOJ IN STANDBY TO RETURN IT TO SERMCE TO INJECT 10N _ o g UNE B MU1223
- ~
M1213 e 480 V AC MCC B51 MU1211 e i k [ CGM $(ji; iQi[@515! X
=
i i CV1220 {C D{CV1219 1 1 FROM BWST M 'A MU20A MU19 A
' T : OR MAXEUP TANK
^
uu26 VU25 TO TRA2J D
. . e * . .
a - e . . . EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1227 - TO RCS COLD LEGS uu1234 uu1233
-{
g 1 TO INJECnOH
" TO INJECDON uu1224 $
450 V AC MCC B61 MU1214 g.J1215 I
}< CB 6152 l) ,) CD 6151 3.(
- CLOSED TO REuOVE D
7 CV1227 VA4.VE CV1227 FROu Y 8
" CV122S {C SERV.CE AND ARE @
[ ND O REW TO TRAlu 8 si TO SERV.CE MU24 f-{ l I uu23f4 FROu BWST i . . .:- - -:. I 0 : . :-' g i T EOR uAKEUP ii:'u :12ccgi% .... uvisc W P36C Tux (CONTINUED)
l l l l EQUIPMEtJT REALIGtH.'Et!T FOR CORRECTIVE MAttJTEN ANCE ON VALVE Cv1227 (COilTINUED) To RCs COto LEcs d y uU1233 MU1234 p l l g ,_ as ; TO IN,:ECnON MU1224 LINE D TO IN JEC T->
- 480 V AC uCC B61 l
MU1215 \ 5. MU1214
) NbNbb hk.N)bk.k 3
$ CV1228 {C O{CV1227 Y
- A _ _ _ _ , _ . _ _ _ _ ! $
I i ! uu24 fi
; I RACKED OUT TO REuOVE VALVE CV1227 FROu MU23fi I FROM Bust SER' ACE AND LEFT OPEN IN STAND 8Y
_ . . . . _ _ _ . _ . _ _ I
% sj MU20C MU19C I T OR M REUP TANK TO RETURN IT TO SER-ACE P36C
= , e
- a a
EQU!PMENT REALIGNMEtJT FOR CORRECTIVE MAINTEN ANCE ON VALVE CV1228 TO RCS COLD LECS A A MU:234 1 uu1233
-~
. . . b' ' ' ': .
1 TO ENJECTiori _ LINE 'O , TO It3JECDOr3
- i uJ122+ i: I UNE C C 480 V AC uCC B61 udi21s, y uu1214 M CB 6152 ) ) C'e (151 )(
l THESE VALES ARE CLCSED TO REuGVE CV1228[O D CV1227 VALVE CV1228 FROM SERVICE Ar;D ARE l l e OPENED TO RETURN 10 N4 8 2 o 4 si TO SERV:CE D w I "
; MU24 fl i i MU23f4 I
i FROM 8WST I
' bM_ -s/
i:yd)CCi!! MU19C I T
- OR WAXEUP TANX P36C (CONTINUEC)
~
EOUlPMD:T REALIGNMENT FOR CORRECTIVE M AtMTEN ANCE ON VALVE CV1228 (CONTINUED) TO RCS COLD LEGS u.J1234 y uu1233 f g h , TO L'JJECnori p;;1224 I UNEO To L.iggcnon uu1223 WC yg,215 450 V AC uCC B61 y,g j4 l I
){ !!CbNUkN. Mk 61N. ){
Cv122S{G OfCV1227 1 i ou i TO TR.*.IN D y " 1 . i w I w uu24 RACKED CUT TO REuovE VALVE CV122S FPCM i f'i s SER ACE AND LETT i ? OPEN IN STANC8Y uu23 ] TD RETURN IT TO SER' ACE gggy g.A3T 2
" ' T Z < OR MAKEUP uu2cC uus 9C W TANK P36C
EOU!PMENT REALIGNMENT FOR CORRECTIVE MA!NTEN ANCE ON VALVE CV1400 To REACTOR VESSEL A A Tat!S VALVE IS CLOSED TO REuoVE VALVE CV1400 rRCu SERVCE AND IS OPENED TO RETURN IT TO SERMCE. DH18 \_ CH138 480 V AC UCC 361 4100 V AC BUS A4 i i CS 6161 ) CB A405 ) $ >: xl ;, V1 00 C 29 L 8 U l f-l Dd1B l
' Y Y- :- TO B'#ST DHB8 DH10 PG TR AIN A (CONTINUED)
EQU!PMENT REALIGNMENT FOR CORRECTI'E M A;NTEN ANCE ON VALVE CV1400 (CONTINUED) TO REACTCR VESSEL RACKED OUT TO REuCVE g i VALVE CV14GO FROu g t SERV.CE AND LEfT
, . OPEN IN $!AND9Y
+ .L TO RETURN IT TO CH;33 SERV.CE.
1 DHIS 450 V AC L*CC 861 4160 V AC BUS A4 e j I l w *
!!.CS.3(E.C (l [cajf4[o@j:yy e
X k' - n o l
!. ! . .__ __ _ _ M X
~
*hg T Q FRCu EWST C'4400fCV1429 CH3B CH2B l BwSB 48 E359
-{ OHIB l
Y --M > TO BWST y DH83 DH10 g TRAIN A
EQUIPMENT REAllGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1401 To REACTOR ESSEL A 1 THIS VALW IS CLOSED TO REMOVE VALE CV1401 FROM SERMCE AND IS OPENED TO DH17 DH13A 480 V AC MCC B51 4160 V AC DUS A3 o e e CS 51114 ) CD A305 ) d, u, N aa Cv140, Cvice m oHu oHu ,,,, m kmqu =r
^
fdDH1A Y M > TO BWST y DHSA DH10 TO LP15 TRAIN D (CONTINUED)
l l EOUlPMENT REAUGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1401 (CONTINUED) l TO REACTGR l NESSEL d A RACKED OUT TO REMOVE VALVE CV1401 FROM SERMCE AND LEFT CPEN IN STAND 0Y k_ DH17 \_ DH13A TO RETURN IT TO SERwCE. 480 V AC MCC B51 4160 V AC BUS A3 e
- ce.m m 4:. 2. :ca:Asc3::.......--..n......: m o
X X l ?: CV1401 CV1428 ~ % DH3A DH2A m raj>g;st o
^ l DWBA E35A 1 j hDH1A M 74 > TO BWST y DHSA DH10 TO LP!S TRAIN 8
o e e . O e EOUlPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1428 TO REACTOR VESSEL A A THIS VALVE IS CLOSED TO REMOVE VALVE CV1428 FROM SERMCE AND IS OPENED TO REWRN IT TO ERM1
,\DH17 DH13A 450 V AC MCC 051 4160 V AC DUS A3 o >
ao
- C8 51114 ) CD A305 ) ,
X X h h, , -:: i- FROM BWST OR W MP CV1401 CV1428 0
^ f:fDwtg-2' E35A V j j DH1A Y M = TO DWST y DH8A DH10 TO LPts TRALN D (CONTINUED)
l 1 i EQUlPMENT REAUGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1428 (CONTINUED) l TO REACTOR WSSEL 1 F. RACKED OUT TO REuGVE V*1VE CV1428 FROM SERVICE AND LETT OPEN IN STANDBY TO RETUR,4 IT TO
\CH17\DH13A SERMCE.
480 V AC MCC B51 4160 V AC BUS A3 o j:C035ii.ts,Ql! 1 NM$!@ji; L m X X
>< RC" 8 k g WMP CV1401 CVl428 DHJA DH2A Bw3A
^
E35A 1r j j DH1A
>< >4 : TO DWST y DHSA DH10 TO LPIS TRA:N O
a o . . + e EC'JIPMENT REAllGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV1429 TO REACTOR NESSEL A A THIS VALNE IS CLOSED TO REMO\E VALSE CV1429 FRCu SERMCE AND 15 OPENED TO RETURN IT TO SERVCE.
,\ DHIS ,\ DH138 450 V AC MCC B61 4160 V AC BUS A4 o
CD 6161 CD A405 ') l) ' x x 4 5I CV1400
! k.' m up CV3429 DH38 DH2D gjj3M5!i
-{ DH1B M C 5 TO BWST oHaa DH10 ,[6 TRAIN A (CONTINUED)
EQUIPMENT REAUGNMENT FOR CORRECTIVE MA!NTENANCE ON VALVE CV1429 (CONTINUED) TO l l REACTOR l NTSSEL l 1 A RACKED OUT TO REuC\E VALSE CV1429 FRCM LERVCE AND LEFT OPO4 IN STANDSY kDH18\DH130 TO RETURN IT TO SERwCE. 483 V AC MCC D61 4160 V AC SUS A4 o > (Cu s!sr :) #m5%% b X X b OR Cv> 4 cvim V oazo e P348 l one fiDH1B _m DH88 a DH10
= ToomT y TO LP!5 TRA24 A l
l
i A 41 l 6. mg E5 v v . 8!
.c &
, hd D 4 O Nb si
- R 8 l 'M o n
>2 R !Bo
~~
dE" 9 o 8 U!k 4 8 WUb QG Nd
-b- >yOh 6i
$gn>IR O n O I 3
- t s a o m
> g*
m A A ..x. O o e ... f. W > t : a at . f c.$ ss <
> 8 A_+5 sa 5 mZ 8 o OO s up
-n On g w ') -
o HO # C ZZ W< 2 g 22 *
\-m3 m
ZW OH DZ =
- wn m O
%* W '
1 H 3 2 >0 A Z . Wo m
~w 6 i
em am o v OO 1 i WO 9090 l l
ECU!PMENT REAUGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2400 (CONTINUED) RACKED OUT TO REMOVE VA1.NE CV2400 FROM SERVICE AND LEFT OPEN IN STANDDY O# TO RETURN IT TO SERV.CE A I 14 omo 11oss 1 480 V AC MCC B61 4160 V AC BUS A4 f CB)17h I. 4 BS3 ![h bh) N y L u p e l T _ FRCM B'A"ai W 3AU es45 Cv2400 USta awtl8 BW38 ' OR SUup P350 14B528 Y TO LPts TRA N A
>bw T
Y2 BWU uS o RR FC A 5 p7 W B E V A O j6BW M EEI RC N T 3 A O MR S T ETRU U B D E SER C ) TdA53 ' SM OOO A 4 P LRT CF V 0 II D 3 SiE 9 0 A I 0 N . H 6 4E E D 1 4 D C E2PC L VO A CSRM l hI V S EI I LD S E HANO TVAT 0 1 3 1 H S 0 T D D S
. 4 W i l 2 BA f (
V O ), C T A, 1 S A E E 8; 2 S V I B SD L 1 5 I A 0 {l yL: P N V C OR A C 1 TT RN M
)
0 h2V 4 OO C A F 7 1 C E V 1 T C 3 NN 0 8 A EA MN 8 4 C jSD 4 NE GT 1 UN 1 1 I 1 AA . EM ) D R E E U T V N NI T I E T o C N ME PR I O C UR ( QO EC
$ie
i A 66 l n. ma 88 EB y f-f.
\A* .
q e m r O 8 td N b h._ > : - P E 3 i. . .! O o q 'g 8 -M-O o wy - CQ o s44 & Q ,, a f' G
< a y
> Hu -
m o5 E
- A_ $ g=-
o a y"sas8 e 8 s
~
sO 32"s W n
~ %uG g 2z "@K o > ~ C~
W a <z R 5"t o
,nS[
b8g a
\'Ia az ' e$w .c .t Ol N L n" ><
4 b > h h tJ
~: 0 - u) .
b (t am OO G1 0 9093
. s o . . .
EQUlPMENT REALIGNMENT FOR CORRECTIVE MA'NTEN ANCE ON VALVE CV2613 P7A FRCH CST OR TO EFS l , SERtiCE WATER y TRAW 8 Q SUPPLY CY2666 I (LC.) i 125 V DC MCC D15 450 V AC MCC B52 M A 1 FROu 50 *A~ C3 72-1512 ) C8 5241 ,) THESE VALVES ARE o o dD .: ' .'. CLCSED AND THDR BREAKERS RACKED CUT TO REuovE VALNT y b7663 : . G 613 BCM SER%E THE VALVES ARE CPENE3 AND THDR CV2613 , BREAKERS LEFT CPEN
- >4 - 1 IN STAND 3Y TO C ! RETURN IT TO 48C V AC WCC B62 SER%il CB 72-2512 )
Ca 6241 ) 125 V DC UCC 025 .i:Mi:.
*/ _ [2 (__
= FROM SC ~B~
?C/2$17:
(CONTINUED)
EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2613 (CONTINUED)
"7^
mou cst OR SI,o % L"/17 *^" gg, (LC.) I i 125 V DC WCC 015 450 V AC WCC B52
+ - rRou sc A-CB 72-1512 3 CU 5241 )
8 A e u M ^' gg - CV2667 . o q6]3 Ti 480 V AC UCC B62 jiC0My25iMd CS 6241 ) RACKED OUT TO REMOVE VALVE CV2613 FRou SERMCE 125 V DC UCC D25 AND tffT OPG31N STAND 9Y TO RETURN IT TO SERV!CE W >g4 : FROJ SG *B* CV2617
. y .
-f . .
i
~* A s , , ,
1
- EQUIPMENT REAUGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2620 l
DC WS M2 125 V DC WCC Dt5 CB 72-1514 G W -06
.+:5 +:+
TO '* " ROM 3g .8* M e
+.......-i. 4N NM CV2620 P7A FW55A i ' ~ ' ' ' ' y-'::g- * '-
480 V AC WCC B63 e CB 6335 FW57 [- o m CV2626 1I > TO CST g-TO csr W FW59 125 V DC SUS MA1 - g ygg g THOR BREAsqRS RADED OUT TO N VALW CV2620 FROM N THE VALE S Ang i
#I-" hI ... ..
..9 CPDED AND THDR BREA8ERS m
- 493 CPDE M STAND 8Y TO RERJRN IT
- Wn.ji ,.
7o a J L 4
"0" "A" ^
(CONT 1NUED) _ _ . _ _ . _ _ _ _ _ _ _ . , _ _ _ . _ . _ . . . - - . - - - . _ . ~ _ - . . _ _ . . _ _ _ _ . _ _ . _ - - - - ~ _ . _ _ . . _ - - _ . . _ . - . _ _ _ . . . _ _ . _ . _ _ - _ - _
1 EQUlPMENT REAUGNMENT FOR I CORRECTIVE MAINTENANCE ON VALVE CV2620 (CONTINUED) l RACxEO DJT TO REuOst 125 V DC uCC 015 l VA_\E CY2620 FROM SERVZ AND LEFT OPEN IN STANODY TO RE"JRN IT TO SERVZ TO = FR N CV2647
- 5. ,O,
< .A v AMD PUUP P7A TW138 CV2520 480 V AC WCC B63
~
FW57 [ CD 6335 y D CV2525 I e TO CST T ", 125 V DC BUS RA1 \ FW5ED l Ce aAl-Z c, O4gCY2545 u FROM TRA:N A i l l
- n. . _
-4 5 , e a EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2626 125 V DC MCC D15 C8 72-1514 TO .j m FROM CY2647 SG ,B, ' AND PUMP P7A FW138 jc',26%i 480 V AC MCC 863 CB 6335 ( ,
CV2626 U O TO CST b e DV59 125 V DC BUS RAI WESE VALVES ARE CLOSED AND BOR BREAKERS RACKED OUT TO REMOVE VALVE CV2626 FROM CB RA1-06 SERVICE. mt VALVES ARE
* ::' ~: :: . . . . . . . . OPENED AND WDR BREAKERS LEFT
.[O :jfy@(jAji OPEN IN STANDBY TO RETURN IT
- o. < . . , TO SERVICE.
d FROM TRAIN A (CONTINUED)
l l EQUlPMENT REAUGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2626 (CONTINUED) l RACKED OUT TO REMOVE 125 V DC uCC 015 VALVE CV2626 FROM SERN*CE AND LEFT OPEN IN STANDBY TO RETURN IT TO SERMCE C8 72-1514 ( l TO '_ .j _ FROM CV2647 SG "O" pwgjB CV2620 ' AND PUuP P7A 450 V AC MCC B63 i
!dsb[6M(,(;i:,
I Y CV2626 o TO CST y
$ TO CST FwS9 O 125 V DC BUS RA1 \FW568 I
C8 RA1-06 ( 0 CV2648 a FRCu TRA:N A "O f f b &
_ . . ~ _
- - _. ~ .
6 o o q l G l EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2627 480 V AC NCC B61 I ' CD 6141 TO _
.j m FROM CV2646 SG 'A" '
pggjA [ gqi j: ' AND PUMP P78 l 125 V DC MCC ' _ ~ CD72-1522(l CV2627 U
- TO CST i
8 TO CST
- ma ~
125 V DC BUS RA2 - THESE VALVES ARE CLOSED AND THUR BREAKERS RACKED OUT TO l REMOVE VALVE CV2627 FROM S RVICE. THE VALVES ARE
!; r OPENED AND 1 HOR BREAKERS LEFT C8RA2-06(l ....
$ifG:i[:Vg;:CV2n45; OPEN IN STAND 8Y TO RETURN IT 4
. TO SERV.CE.
o FROM TRAIN B (CONTINUED)
8 6 7 4 P 6 7 2 P VM CU P u oD RN FA T S
) C D 1 O T
E 6 8 U C N C 0 I T u 7 6 N C ( 2 O A V C 7 5 C V M 2 6 8 5 4 6 8 ( 0 2 5 8 B V W C N 7 4 C C F C' ! A 2 V R 4 6 O \- 1 2 C u V 5 ?. C R C 1 D A R F C S E C U V A u !! B L 1 C A ,W D C ( F 2 8 D V O V 2 5 V 6 0 E T 5 5 l p,FW - RN CY MB 2 f s. 5 2 2 A OO D _' 1 2 1 R F E REN 7 i, S E V O SA T E '8 T C TC "A 'C; O MM S C M " TG NN E C R RI NR OG EA O F NS E TS MN T 7E NE T 2P O T GT U 26O T I L N I OVTICF D AA E ELR EN . EM R KV CLD T U AANE E RVAR TV NI E TC . ME P R IUR QO EC
~$
.IilII '
>.w" 5N 4
T 6 2 A RI T N UvVDN Oc HEN I E TPR R RD A DE H KL EWDN OU V T TE S AV A EFR E LO V DR E TCAED E TE
< LN C AA SVRNS RO ERY!V V
DEM PE8R 1 6 6 1 EEKE OKDE B B SSAR ANS EOE EEA C C HLRO RRTO C C TCBT ABST MM a ) ) l 1 C C 2 C 8 A A 5 1 5 1 8 5 6 1 7 6 V V 6 1 2 1 5 - D 6 B 0 0 2 2 C 8 8 B D 7 C 8 4 4 C C C .i" z.i2 '- C D M O' C C ) .S" M C '
) A . N';*' ;
C ; W D V .i S 2 C V 0 - P 3 8 - MO 5 6 1 4 6 1 OO 2 2 0 5 RL F 1 V 8 8 C 2 3 I g 1 V C V C M A ' EM TE ST ] dA 2 51 A 5 4 MY OS S [Y N , ' 7 P R S C 0 6 ! U C 2 M l) S M ) V 3 A 6 C C D D 2 C 6 0 0
- V V 2
5 5 6 1 A'F 1 1 2 W 2 E D V A 5 5 2 V C C T S R 2 2 1 1 7 L C C 8 8 A A M ) 2 A O C C. V C R T S N I D 2 A 1 S E
.!. 9 R RN V 5 1
U D
. 7 ..
45 8 , f. l T OO 5 2 C
) .6 24 6 M A5._ ki A ,.~. O F 2 v T TC E 1 7 B
D V 6 0
'.CV v.2 c.!iC N ' Q,ff.h.
C A : NN 5 2 A EA 2 1 R A'5 5 W T MN D F T S C NE C S C GT - O T L IN O I
- 7 T . AA < ~B 5
W EM R 0C F ) D 2S E 6 E TV NI OVN 2D U TCA N E T
. C E ME N P R I
O UR C OO ( EC o oN l lll1
EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2645 (CONTINUED) 125 V DC MCC D15 FROM STEAM 125 V DC MCC D25 SYSTEM
$$ii7255{${ii U jQ]CS.571@y3{fl CV2663 b C CV2613 125 V DC BUS RA2 480 V AC MCC 861 CB RA2-06 ') TO CST
') CB 6175 h
M b ' ' V2620 1 M :
, AND SC ~B* FW55A CV2647 FW10A CST CV2802 o ^
- CV2645 S CV2506 Y
t - g7 :N$.@.5bN$-)U 5M!iN bh.!:1). # Y - 125 V DC BUS RA2 - 480 V AC MCC B61 TO CST M2 125 V DC MCC D15 480 V AC MCC B61 TO CST M FW38 CB 72-1522 ) CB 6185 CV2627frE CV3851 O . O DOW J L VALVE CV2645 FROM SERVICE AND LEFT OPEN P n tN STANDBY TO RETURN TO TRAIN A FROM SW LOOP 2
. . . 4
?u*
1 1 5 5 E B 8 VE C C L RN u C C RA I V AEP oT MM E R S ) HE SO E TV E V Fc l 3' C C 4 OL VF TL 9 A A 9 A _' 1 V V 1 D N M E V AELV 1 5 5 5 AR SE B B 0 0 8 8 a DO E RH C 4 4 C E T TH ET C ! s S C OT LUE E R
.KA N M ) p C O C RU E
DR I VB TE C A 3 7 y s W S 1 R A CEEER KSHTO V 1 5 ' y ' y aM P
. 0 - ]
S T 8 B ] OO E N AM RO O . 4 C s 0 5 RL F A SR N BC YE L o 8 VR FA DI V a 2 3 v E6D N 3 v C EK4E SA6N AR T ES A C EE2E S 5 HRVP N O TBCOIT 0 0 1 3 C A
)
1 3V 7P8 A R 6 B 1 S C V 3 U C A i' B M ) 0 6 D l) C C 1 C B 6 D A 5 4 0 3 jdW 1 - V V 3 6 1 F A 5 0 R 2 8 s p T 1 4 C S 8 B 1 C C A O , N I 6 R T .." 7 A
. 4 S ,h:.: [
- - R T
6 4, U ," 6 U .. 8 t9 6 B O 2 V C
)
6E~ 2 . 2 6 2 T C O 0 6 C.sv ,N, N v C V - 4
, E 1 ^.i..
5 A 6 - V 2 R j5 T L 1
. S A 8 C
N T S C V 1 - C O T 6 O RN 0 0 6 T OO C W F E C M
$'" F w
) ?" -
TC NN C A 1 4 N" Y" EA V 1 6 N'" 5 MN 0 S NE 8 C GT 4 ILIN
- AA _ )
EM R 0 E E U T V O N NI T T I T E C
. N ME O -
P R I C - UR t ( QO EC
*y i
- ' ' ! ! , ll
l l l EQUIPMENT REALIGNMENT FOR l CORRECTIVE M AINTENANCE ON VALVE CV2646 (CONTINUED) RACKED OUT TO REuCNE VAL \E CV2646 FROM SERMCE ( AND LEFT CPEN IN STANDBY
'O RETURN IT TO SERMCE l
480 V AC MCC B61 125 v DC BUS RA1 4160 V AC BUS A3 480 V AC MCC B51 C8 6141 ) i[$0[.RI,'1}c{;f[ TO CST (C8l!M1[1 Ii CS 5173 ) TO SC "A" m
' j
}~ je i T s _ FROM I
e CV2670 FW36A CV2646 FW100 CV2800 g " Cv26483 Cv2eo3Xc g CD RA1-06 ,) 'ik%T93
; )]j'
- I I 125 v DC sus RA1 -
480 V AC MCC 851 To CST N56 -[ 480 v AC uCC 063 - 480 v AC uCC B51 TO CST e'/r-W59 C8 6335 ) CB 5194 ) I Cv2626 Cvasso][C t a TO TRA!N B FROM SW LOOP 1
EOUl? MENT REAUGNMENT FOR CORRECTIVE M AINTEN ANCE ON VALVE CV2647 125V DC MCC 015 FROM STEAM 125V DC MCC 025 j SYSTEu CD 72-1512 ) ) CB 72-2512 I I al CV2663 I Ic I CV2613 125 V DC MCC 015 125 V DC BUS RA2 480V AC MCC 861 I f CS 72-1514 ') C8 RA2-C6 ) TO CST h CB 6175 ) Ic e * . be- .! FW55A j[CV2620;fj
- X Sb r ~
CV2647 FW10A
' ~ - ' -
^ !!h230)f]
$ ! kddT$i: -- CV2806 C > 1 ; CS RA2-06 l) f CO 6181 I) N I I I 125 V DC BUS RA2 - 480V AC MCC B61 TO CST FW559 SW13 125 V DC BUS D15 480V AC MCC B61 THESE VALVES ARE CLOSED TO CST % AND THE:R BREAKERS FW58 i C8 72-1522 ) CB 6185 ) RACKED OUT TO REMOVE VALE { l CV2647 FROM SERV!CE. THE Cv2627 CV3851 [O---- VALVES ARE OPENED AND THE i EREAKERS LEFT OPEN :N y STANDBY TO RETURN THE VALVE TO SERV CE. TO TRA:N A FROM SW (CONTINUED)
- , i Yu*
1 1 6 6 B B C C I!, C C [ )-j MM j C C 5 N- 1 C [85 A A V V 8 1 6 [i 0 0 6 5 .N B f 8 8 B 2 D D C d [ 4 4 C . C 2 _ M 0
-E- ) 8 2 _
E
- C A 5 V ,
C 7 C W D V 1 S 2 0 8 C - C V 5 3 1 84 8 C [
] nM P O 2 6 6 1 OO 1 2 0 5 RL F
V 8 8 C 2 3 - V
$c C V
C
) 2 51 D ^ A E R D U 0 l
S U S U N m ) B B ) TI ti a l C C 2 N 3 A 6 D D 2 6 0 O 6 M0FW1 - V V 5 1 C M 2 2 5 5 - ( D V A 2 21 2 C T R 7 C S 1 7 2 C 8 8 A
-" f.3 4 E A O C C R N
- _
T h 6 C I - D I- -
\- A 2
^S S
U 7 + RT _ V W B 4 5 8 7 C s!- 6 4 2 A'F 5 2 O 2 m C ! 2 6 W 6 T E 1 - M V D QI V 2 C V C N 2 V C V L <M. 5 2
$A A 5
=
A ZW5F 1 N T RN V E 5 k -
'C T
S S C O T OO CY 1 7 O D T F M0 5 W E E R 0C E C F TC VEN C 0 O S TR AM M ) 2 NN M M SE C 6 2 EA EON S R D 4 V MN RI O O F NT V 1 5 C NE T 7E T 1 GT T 4PI 5 2 I L N U 26O N 1 2 7 O VTR I AA D CFU B < .
~
EM R ET E ELE C
"B KV R ~
E CLD TV AANO G RVAT S NI T E C . ME P R I UR QO EC
.8w i
i ' i
EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2648 mESE VALVES ARE CLOSED AND THDR BREAKERS RACKED OUT TO REMOVE VALVE CV2648 FROM SERMCE. THE VALVES ARE OPENED AND THOR BREAKERS LE.FT OPEN IN STANOBY TO RETURN IT TO SERMCE. 125 V DC BUS RA1 4160 V AC BUS A3 480 V AC MCC 851 CB RA1-06 ) TO CST CB A311 ) CB 5173 )
. . . .. ik c OM CV2670 t 'd! 'A l , . . . . . . , , ,,
AND SC "A* FW56Aij.CV2646j FW108 jiCpfl00:j CV2648 C CV2803]C :> CB RA1-06 ) CB 5193 ) $
" 125 V DC BUS RA1 -
480 V AC MCC B51 TO CST N6B[- 480 V AC MCC B63 - 480 V AC MCC 851 TO CST M Fw59 CB 6335 ) CB 5194 ) CV3850) h U JL. TO TRAIN B FROM SW LOOP 1 (CONTINUED) g.-
EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2648 (CONTINUED) . _ _ _ RACKED CUT TO REMONE vat \E CV2648 FROM SERVCE AND LEFT CPEN IN STAND 8Y TO RETURN IT TO SERMCE 125 V DC BUS RA1 4160 V AC BUS A3 480 V AC MCC D51 C8 RA1-C6 TO CST (C8Nt[id! C8 5173 ) TO CV2670 < /. A X
- FROM e AND SG 'A- FW56A CV2646 FW100 l CV2800 CV2648 C CV2803fC FW60[
Y - 125 V DC SUS RAI - 480 V AC MCC B51 TO CST FW568[ 480 V AC MCC B63 - 480 V AC MCC B51 TO CST W i-FW59 CD 6335 ') C8 5194 ) CV2626 CV38%O (C t a TO TRAIN D FROM SW LOOP 1
EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTEN ANCE ON VALVE CV2663 P7A PROM CST OR TRAF 8 3p y I (LC.) l 125 V DC MCC D15 480 V AC MCC B52
'A I FROM SG 'A" CB 72-1512 ) CB 5241 ) ?
O THESE VALVES ARE $
- ::: ] CLOSED AND THEIR
! : N:: 'f- 1 BREAKERS RACKED
_h_ CV2663 iiih(i'6
" " ' ~ ' '
yi: OUT TO REMOVE VALVE CV2663 FROM
~ ~~
SERVICE. THE VALVES ARE OPENED CV2613 AND THEIR BREAKERS
~ ~~
LEFT OPEN IN STANDBY TO RETURN IT TO 480 V AC MCC B62 SERVICE. C8 72-2512 ) CB 6241 ) 125 V CC MCC 025 ...17:. (CONTINUED) A
- FROM SG ~B"
$sv2st7::b*46
- i
EQUIPMENT REAUGNMENT FOR CORRECTl'E MAINTENANCE ON VALVE CV2663 (CONTINUED) P7A FRCM CST OR NB 1 SERMCE WAU y4 SUPPLY 8 CV2666 (LC.) l l 125 V DC MCC D15 480 V Ac uCC 052 A FROM SG 'A'
- ca!Z2.sts.023-?j Cs 5241 ')
o w q 5 ,, [ M-CV2667 CV2663 $ m g6J3 W RACKED OUT TO REMOW 480 V AC MCC B62 VALVE CV2653 FRCPJ SERMCE AND LETT OPEN IN STANDBY CB 72-2512 ) TO RETURN IT TO SERVCE C8 6241 ') 125 v DC uCC 025
'A >g; FRou sc 'a' CV2617
> e" T
F O ET 9 N D T LI A T SN . 8 UM3 RR _ 7 DO KEKT EU P E SDFRAE MP OM OE LK CC 7 0 AERR RU 6 SBOT FP A 2vR E ER R V . _ AS CAlD Y B S R E VED N eve DATE . 1 A EKLmN V AA _
.A SC L EV R AR M _
S U j: j+ 6: V E BEEDI VCE NR E B 4.o. 6:: S ROMNNS EOMREE C ( 2+: HEEPPO NTRSOOT D 6 V+ C: j+ V 0 A 5 1 6 . 2 1 A R j5W F T B S C C Y 1 O 6 T B C
. 0 C 0 7
7 M 6 = 6 C C 2 2 A V :y. 1 C 7 :. .. V V 4 1 2 6 8 5
- 6
- f. .
B C 0 6 2 5 :2
- V N
e 8 B V W : C F :C.
- I A
E 4 C ::. R V \- y:y.1T _ L sO.. M A O V 5 1 2 A R F 0 R RN C S . OO AM C U B F E jI
.W 1
C ( C ( l TC F D 2 8 D 6 NN V 2 5 /* W5 V 0 EA 5 1 ' F 5 2 MN 2 1 2 2 1 A R NE _' 7 . GT B T 8 C . e UIN ~^
~
C OS TC AA OG ) E R M TS D E E U T V NI N E T I T M EC N PI R O UR C ( QO EC or
EO'JIPMENT REALIGNMENT FOR CORRECT!VE MAINTENANCE ON VALVE C\'2670 (CONTINUED) 480 V AC MCC B61 125 V DC DOS RA1 UI$I C$670 FREM SERY.CE AND LEFT OPEN IN STANDBY TO RETURN IT TO SERV.CC :.- . '-
.: CB RA1-06 (
[3 O ,,, _
,A _ _.~
~ -y X e FROM
^ FW13A CV2670 FW56A CV2646 PUMP P78 125 V DC uCC D15 FW60 ~
CB 72-1522(
$ 3 CV2627 y C +
TO CST E T " C5. FW58
. m !
i FW558 125 V DC Bus RA2 C8 RA2-06 ( L___ . .- OfCv2645-FROu TRAIN D
, , a * . .
l l . e
- 4
- i-4 5
!~ 1 EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2803 [ THESE VALVES ARE CLOSED AND THOR 1 BREAKERS RACKED OUT TO REMOVE VALNE CV2503 FROM SERMCE. THE VALVES ARE OPENED AND THDR BREAKERS i ,' LEFT OPEN IN STANDBY TO RETURN IT TO SERMCE. ( 125 V DC DUS RAI 4160 V AC DUS A3 480 V AC MCC B51 CD rat-06 ,) TO CST C8 A311 ) CD 5173 ) ' TO
^
~
CV2670 < M,.. M l ..'.3..... O i AND SG *A" FW56A),'QV2846 FW108 ,d ?V2800j ST hn CV2803 C8 rat-06 FW60[ C8 5193 ) .$ 125 V DC BUS RA1 - 480 V AC MCC B51 f TO CST N6B [ 480 V AC MCC B63 480 V AC MCC B51
' TO CST M FW59 C8 G335 ) C8 5194 ) ,
Cv2626) Cv3850][C Y JL TO TRA.N B FROM SW LOOP 1 l (CONTINUED) 4 ).
l EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2803 (CONTINUED) RACKED OUT TO REMOVE VALVE CV2803 FROM SER\1CE AND LEFT OPEN IN STANDSY TO RETURN IT TO SER\1CE 125 V 3C BUS RA1 4160 V AC BUS A3 480 V AC MCC B51 l C8 RAl-06 ) TO CST gjyjj:j C8 5173 ') N TO
' T c
CV2670 '/ 'A CST i e AND SG "A" FW56A CV2646 FW100 CV2800 CV'548 C CV2803 (C FW60[ U - 125 V DC BUS RA1 480 V AC MCC B51 TO CST NSB [- 480 V AC MCC B63 480 V AC MCC 851 I TO CST M . . ,, . . . , . , . . FW59 C8 6335 ). :CS!:.41M)::j ( CV2626h CV3850]f u u TO TRAN B FROM SW LOOP 1 6 l 8 D & * , g
EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON VALVE CV2806 125 V DC MCC D15 FROM STEAM 125 V DC MCC D25 SYSTEM
] THESE VALVES ARE CB 72-1512 ) U ) CB 72-2512 CLOSED AND THDR BREAKERS RACKED OUT
-+
. ?3Fi 1.1 TO REMOVE VALVE Cv2806 i i!?M6 bi/ [{pjgyg; g
- -^
FROM SERVICE. THE VALVES ARE OPENED AND THO3 125 V DC BUS RA2 480 V AC WCC B61 BREAKERS LEFT OPEN IN N STANDBY TO RETURN IT TO SERVICE. CB RA2-06 ) TO CST , J CB 6175 )
' ' OM CV2620 < J A
d CST. AND SC ~B" FW55A:CV2647L FW10A h2502ii
^
CVM45! CV2806 C
* ~
FW57[ ~t Y - 125 V DC BUS RA2 480 V AC WCC B61 WB[ TO CST 125 V DC MCC D15 480 V AC MCC B61 To CST -+-tM-FW55 C8 72-1522 ) CB 6185 )
; Cv2s27Xc Cvaa51 ])
y a TO TRAIN A FROM SW LOOP 2 (CONTINUED) 1 i
EQUlPMENT REALIGNMENT FOR CORRECTIVE MAINTEN ANCE ON VALVE CV2806 125 V DC MCC D15 f ROM STEAM 125 V DC MCC D25 SYSTEM THESE VALVES ARE CO 72-1512 ) V ,) CD 72-2512 CLOSED AND THEIR
..I I BREAKERS RACKED OUT
[' i. TO REuoVE VALVE CV2806 FROu SERMCE. THE VALES i'Cvicest i m Cv2nd ~:i ARE OPENED #4D THOR 125 V DC BUS RA2 480 V AC MCC B61 BREAKERS LEFT OPEN IN N STANDBY TO RETURN IT TO SERMCE. CB RA2-06 ) TO CST - CB 6175 )
.. . A I '
W620 e '/ A
'U. <
CST AND SG "D" FW55A C."26 47 - FW10A MV22102)
^
[chjdh 'yn CV2806][C > FW57 "CB RA2-06 ) C8 6181 ) $ I - 125 V DC BUS RA2 - 480 V AC uCC B61 TO CST Fw350[ 125 v DC uCC Dis 480 v AC uCC B61 TO CST W r-Fw58 CB 72-1522 ) CD 6185 ) Cv2627 C Cva851 ][C v m TO TRAIN A FROu SW LOOP 2 (CONTINUED)
e a s , e t i I i EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON PUMP P7A mEel VALSES ARE CLOSED 125 V DC MCC 015 FROM STEAM 125 V DC MCC D25 AND THEIR BREAKERS ARE RACKED SYSTEM OUT TO REMOVE PUMP P7A FROM SERV:CE. THE VALVES ARE OPENED CD 72-1512 ) V ') CD 72-2512 AND mOR BREAKERS LEFT OPEN IN STANDBY TO RETURN IT } 3' I 1 TO MCE. CV2663 CV2813 i 125 V DC BUS RA2 480 V AC MCC 061 CB RA2-06 ) TO CST - CB 6175 ')
' T M
620 : M. I: ..
< CST AND SG ~B* FW55AIfW2dDj FW10A Mj$0%i x ; e g
- : ..: :9.-::::-
ICV 2$45;y[:, ;. - p7A CV2806 3 j y"""""
. l C8 RA2-06 l) CB 6181 ) $
j U - 125 V DC BUS RA2 - 480V V AC MCC B61 TO CST NM -- 125 V DC MCC D15 - 480 V AC MCC B61 TO CST M FW58 CD 72-1522 ) CB 6185 CV2627 CV3851 y ' U JL , TO TRAIN A FROM SW LOOP 2 (CONTINUED)
I EQU!PMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON PUMP P7A (CONTINUED) 125 V DC MCC D15 FROu STEAM 125 V DC MCC D25 SYSTEu I.Nk.kNhMkM.l. U l QL 1r 1 j r E g
- []Mi!:14.k.YN.$!!
THESE BREAXERS ARE RACKED OUT TO REuOVE 4 CV2663 CV2613 PUMP P7A FROM SERMCE AND LEFT OPEN IN STANDBY 125 V DC BOS RA2 480 V AC MCC B61 TO RETURN IT TO SERMCE. N
) TO CST
)
i TO . f CV2620 < A A s : F
. AND SC *B- ,
FW55A CV2647 DV10A CV2802 CV2645h CV2806 > g.3, ca 842-o6 ) CB 6i81 ) o I U 125 V DC BUS RA2 4SOV AC kCC B61 TO CST FW558 k- SW13 k-480V AC MCC B61 - 480V AC WCC B61 TO CST 4-IA > Fw58 C8 72-1522 ) C8 6185 ) CV2627 - CV3851 i " a TO TRA:N A FROM SW (CONTINUED) m2 _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ -.-.-,.m m ,- - . - . _ . . ,,,.,_m_ - . . - + , __- ~. . . . - ..
e e e . a e EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON PUMP P78 mESE vatvES ARE CtOsEo ANo moR BREAKERS RACKED OUT TO REMOVE PUMP P78 FROM SOtMCE. THE VALVES ARE OPENED AND THE BREAKERS LEFT OPEN IN STANDBY TO RETURN THE PUMP TO SERMCE. 125 V DC BUS RA1 4160 V AC DUS A3 480 V AC MCC B51 CD RA1-06 ) TO CST CD A311 ) CD 5173 5
' T 670 < '
/
CST AND SG "A" FW5SAjd/g6fk ., FW108 li,$i/2iN)dj ilChN. E i CV2803 C C) FW60[ h Y - 125 V DC BUS RA1 - 480 V AC MCC B51 TO CST N8[ 480 V AC MCC B63 - 480 V AC MCC 851 TO CST M FW59 ) CB 5194 CV2626h CV3850](C U d TO TRA:N B FROM SW i LOOP 1 (CONTINUED) ?
f EQUIPMENT REAUGNMENT FOR , CORRECTIVE MAINTENANCE ON PUMP P78 (CONT!NUED) t RACKED OUT TO REMOVE ,' PUMP P78 FROM SERMCE i AND LEFT OPEN IN STAND 8Y TO RETURN IT TO SERMCE u a
- 125 V DC BUS RA1 4160 V AC BUS A3 480 V AC MCC B51 C8 RA1-06 ') TO CST iMjf3,1ji)jji CB 5173' u l CV2670 1 A
'A ' T ! C AND SG *A* FW56A CV2646 FW100 CV2800 3 CV2648 C CV2803 h
U - C8 RA1-06 FWSO CB 5193 ) b N l Y - 125 V DC BUS rat TO CST NG[ 480 V AC MCC B63 - 480 V AC MCC B51 480 V AC WCC B51 s h 4 TO CST M ] M9 ) CB 5194 . CV2626 CV3850 TO TRA34 8 FROM SW LOOP 1 I i g g h
- - = _ - - _ - . . . -, .- - , . - . ~ - - . - - , . --. . - . .~. - . _ _ --_ . _ _
__=_ _ _ . _ .
. . _ - . . . _ _ m _. .._.__m... _ .. _ _ _ _
e e e . *
- EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON PUMP P7A (CONTINUED) 125 V DC MCC D15 FROM STEAM 125 V DC MCC D25 SYSTEM I U CD 72-1512 ) ) C8 72-2512
' ' RACKED OUT TO REMOVE CV2663
%CV2613 " ^
AND LEFT OPEN IN STANDBY 125 V DC BUS RA2 480 V AC MCC BG1 TO RETURN IT TO SERVICE. s CD RA2-06 ) TO CST , C8 6175 ) 620 t M b Z < AND SG D* Fw35A CV2647 FW10A CV2802 CV2645 S CV2806 h y
~
m,y Cs -2 d :..ian a d s Y 125 V DC BUS RA2 480V V AC MCC B61 TO CST N k- - 125 V DC MCC D15 - 480 V AC MCC B61 TO CST M Fw58 CB 72-1522 C8 6185 CV2627h CV3551 h U h TO TRAIN A FROM 3W t.OOP 2
EQUIPMENT REAUGNMEN f FOR CORRECTIVE MAINTENANCE ON PUMP P34A TO REACTOR
'.ESSEL u A THIS VALVE 15 CLOSED TO REMO%E PUMP P34A FROM SERMCE AND IS OPENED TO
,\_ DH17 ,\_ DH13A RETURN IT TO SERMCE.
430 V AC MCC 851 4160 V AC BUS A3 o CD 51114 ') CD A305 ) 3 x x > h CV1401 k CV1428 j A _T sgi- FROM BWST S "P DH3A DH2A khMf8'
- A [
E35A P34A fiDHIA b DHBA M - TO BWST r DH10 TRAN O (CONTINUED)
EQUIPMENT REALIGNMENT FOR CORRECT!'E MAINTENANCE ON PUMP P34A (CONTINUED) 1 l TO REACTOR VESSEL l A A RACKED OUT TO REi.!OVE l PUMP P34A FROM SERV!CE AND LEFT OPEN IN STAND 8Y
\_ DH17 \ DH13A TO RETURN IT TO SERV:CE.
l 480 V AC MCC 051 4160 V AC OUS A3 j@j5tt.15iiid ICD. h50.$ji!lQiji C e ) Y y N CV1401 CV1428 DH3A DH2A
% [g*"y$'
DW8A E35A 1P J DH1A Y >< > TO BWST y DH5A DH10 TO LP:S TRA:N D l l l l l
A-76 O " hh ma 4 S, Ed ve 3 28 o O d o$ b. ,, $ db 2Oh , W24-3V8& u;5e bN p!$ _
. cs:-
D4 m Z m Q) a g a c5 r Q - O m O *
< g -
> g 9 o , m
$ % O 8 5 g a i5 e
e da ' s+
~
R n O Q.
- o. m "
, , ~
z 2 o G o bd h d"$
- n. u u
TZ m OO S g{ g 'w HO
> s ZZ g -
w< R 8 2Z zw E O Z x * < rt eed e W2 m ~<0 w 5 -
" 0 t
H >o 1 LM X- w o kU -O z G
- i
\
$_.id E Z , am O OO o v l wO 9125
o e e e n e EQUlPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON PUMP P348 (CONTINUED) TO REACTOR VESSEL 4 ;L RACKED OUT TO REuoNE PUMP P340 FROM SERMCE AND LEFT OPEN IN STANDOY
\DH18,\DH130 TO RETURN IT TO SERMCE.
480 V AC MCC B61 4160 V AC 0U5 A4 l , fea:sien ml isaT49em; C > e >c x L h CV1400 k CY1429 DH30 OH2S 3 gm BW88 FROM BWST E350 i DH10 M 74 : TO BWST y DHBO DH10 TO Lots TRAIN A
5 T S WP DU U MS O R FO R ,
' i"i*
"- C S C MT ORI A TES N D R
,8 EM UT 3 SO E A ORR LF S
U C AT O B E5 C ) 3dA53 ' R3D APE A 4 P SPEN V 0 EMP 3 ;I V L U O E. C 9 0 A AP H 6 V EM R D 1 4 D ER C EVA E 4 SO S IfI EMD HENO _ TRAT 0 A 1 H 3 S 5 T D B 3 S P W B If l 4 f P M O T : '!iA-.- t U 3. A
}- 2 P 3
[t- S B 1 SB RN 5 ! 8 4 fl P LL N OO C O A F C TT R E 1 Q '0 M TC ) 4 C NN A 1 7 m N C EA V 1 MN 0 5 NE 8 B C GT 4 LI N I B AA . EM R I I I I
)
D E E T NI V U E T N C I T . M E N P R I O UR C QO ( EC e
i l A-79 w to N $ H or p: w$sg 3 Mw n
- n.
m :s E 3E O s 8" - sea kO s'0O. 34 U O"Hh n f a w CL Jo w&
- 5%
43 o
- 20. s <
m
\.4 s o
n U f C .... m o 5 $ O w < ....' _ , n D > 5?ri -- Z cn o p :t e !Q..i Z ih! - o -.-. m _ e o v
< o
~ n nn o- h a.
= +-- A--._-_..A c
2 ~ D < 1 hG l O N g ez -
= ao oo t i. -~ ~ b ' '~ + *I u e g<
sU y - 2z n v N On g? Id < - N u 2Z 2y
> g o
OH e g O_2 Y hia w1 W -. m 2 F td sO
~
- 0. W e ae oo WU 9128
A 80 A wa jb te dO
- 1r
. . .:.: y .
gi y i-am-
, %[:j Mdg Ssa nOw e Ogg
+
X.m 'S m d6mS 4 wn m 428 8 m m~5 nr O ._O~. - 4 O a3OW 4 L 4 A U
> > MS m O O
4 uYE5 mO m I O w30 0 g. I w '/ O F-EMr l CD 3 m o I o g O m g' 1 s4 l ww _ _ -.. A .__. _. __ O. O
?. :.:. .
3 :: J.-,M:i :: m . Q. 'I::.. . . . I.!! I4 m I' $U k U _.. 4 !>am 5,j Ox lij 9 O ** b H O n 1rG ZZ U - Ojg 4 L2] < E u 2Z > U Z 12) o , O I-* $ ~$ l -J Z n , m -t _< a Itj 7 m M - O W 2 U H 3 l Z p> 2 l Q $ A
- I tj Z aM o am o v
OO tt1 0 9129
A 81 Wsd ds ~ h S
,z o ma RE5" w Ek wo 3r 8CD4t" 8'S$
09e= 4258 f 8
= ,
m c W O 4
-?
o)&
- n.
O e o f:i M Z I
- I' ':i:
p 7 0i y -h-O e b M to s o o n 0 1 ' $+-b b v a R e 3 1 p$ O N
$ mm xZ 5 52 b2 OO '
o o w y ' " ga n ZZ 31:a , w< W - t 8 22 #
- ZW @ m OH a2 v
\$=
W5 Z =
=
Zp D W 2y se oe OO WO 9130
A.82 b lI
; ~;
!.k!.
"8
-)g @8 g<fe v -^ -
- n _ _ meme s
- Wa E 5s
' 8 U$@8,5
>YN*O os
< *$ W55
- w=n j5 n .. Q Q.
Q. .hw!.! 2 g Q. Ir ZZ OO ti. W bO ZZ W =< 2z. ZW OH U 2 t< ' W2 m O W H$ D Z 3mo P -
-- W z bZ o De O v
OO WO 9131
EQUlPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON PUMP P36A (CONTINUED) 4160 V AC DUS A3 RACKED OUT TO REMOVE ggg}.)::' PUuP P36A FROM SERV.CC AND LEFT OPEN IN STAND 0Y TO RETURN IT TO SERVICC 1 M MU20A
'A Mll19A k
MU18A $ P36A y b
A 84 9 lf
- /
n_ . , 2 e n '.]D8
@8 o
5
~
U a de0o
= -
> a a, h
- 8 9010M o
6 80
>63-m V-s a h
M ... ' :: 5 G.
!: t
- b 4
3: Q. l' ZZ OO
- h. W bU ZZ W<
eZ ZW O& OZ << a W2
^
l Hy 8 D Z SnO I p . , 4 W 7 , b% o U l DT " l OO WU 9133 l e
o , , , . . EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTENANCE ON PUMP P368 (CONTINUED) 4160 V AC DUS A3 RACKED OUT TO REMONE iggg---
- - - g:;.)-.; PUMP P368 FROM SERVICE AND LEFT OPEN IN STANDBY TO RETURN IT TO *.E.RVICE
< k MU208
'/
MU193 I T k : e MU188 [ e P3GS > a w
T l l l 1 EQUIPMENT REALIGNMENT FOR l CORRECTIVE MAINTENANCE ON PUMP P36C l l l l 4160 V AC BUS A4 l C8 A406 ) i ...... ........- o 5 .CY. M C ,.. si [ P36C y w a m l THESE VALVES ARE CLOSED TO REMOVE PUMP P36C FROM l SERMCE AND ARE OPENED TO RETURN IT TO SERMCE 1 i l l l (CONENUED) l l l l l l l l l
- s . . ,
o , . . . . EQUIPMENT REALIGNMENT FOR CORRECTIVE MAINTEN ANCE ON PUMP P36C (CONTINUED) 4160 V AC BUS A4 RACKED OUT TO REuoVE PUuP P36C FROM SERV.CE
- !QQjA,$,$- - - . - :- - - d,-
AND LEFT OPEN IN STANDBY TO RETURN IT TO SERV.CE
= h *
/ l N =
L!U20C MU19C P35C V :JU18C E b
l : EQUIPMENT REAUGNMENT FOR TESTING PUMP P7A , 125 V DC MCC 015 FROM STEAM 125 V DC WCC D25 SYSTEu- . CB 72-1512 U ) CB 72-2512 To TEST PUMP P7A AND ARE I F 1 r CPENED TO RETURN IT TO SERV.CC. CV2663 CV2613 125 E'DC BUS RA2 480 V AC MCC B61
) TO CST ,
N a62,, gg TW55A !,EV2645
. , s _
= r-AND SG *B" FWlOA CV2802 >
w .. + :.y .w. .:. p7A q p e l O..;.; .5 4 L
~5) - ~}
CB RA2-06 ) G 6181 FW57 ,k. 125 v oC sus RA2 - 480V AC WCC 961 To cst rwS58 \- swi3 480V AC WCC B61 - 480V AC WCC B61 TO CST M RSS Cs 72-1522 ) Ca sia5 CV2627 CV3851h) v n TO TRAIN A FROM SW LOOP 2
& b W O O
. . . , o a EQUIPMENT REAUGNMENT FOR TESTING PUMP P78 THESE VALVES ARE CLOSED 10 TEST FUMP P78 AND ARE OPEHED 10 RETURN IT TO SERMCE.
125 V DC BUS RA1 4160 V AC BUS A3 480 V AC MCC B51 CD RA1-OG ) TO CST CD A311 ) CD 5173 ) i T 670 -c A
$.. 'M . <
AND SC *A* FW56AIAV2646? FW108 h CV2800 MiSM8f CV2803][G ,
' CD RA1-06 ) CB 5193 )
- I -
125 V DC BUS RA1 480 V AC WCC B51 TO CST rw3cs / 480 v AC uCC ass **" b- 480 v AC uCC B51 Ca 6335 ) CD 5194 ) cv262s ! cv3850{C Y a TO TRAIN B FROM SW LOOP 1
A-90 s O. ma 8" E5 I. .
. m.
B t 7 _m r5 5 43 85 h <o a : m
~
g l
> g g ys@~2d
# S 2 = l 5 R 8 g
{y@l " g is s Wa ggo n , Q. . ....... $ ikyo:i A5ji
- a 4 ::-{ ;
Q. O o W:i
- I Z e :.
Q \] N V N: (/) 5 :; l c h --
' :-.-]~;;5!
s - x 2 ,, u? l @ W 4s i s Z 2 E W <
- 2 O O I ZI $ l 0'
g -e A
.XJ,
< d s , W h N E a: ay a H
- A -X-Z W s 2
Q: o O W 9139
A 91 w A au b E5 0 dB im m< U M & 5I g e o
-5 w5wO D m r m
k '* '
" g gt- > s s*
- ( 2 @ d 8 I
>>ON 9" p Wydg 5 u w n h ESkS F 0
n 5 1
.:: ai e b :!, !
2 . - . :. 3 S
' O O . . .
- Z e ;: ' gii 9 7 tn .o .. f. .i.! k 8
^
u.
< a 3150 W o Z o @
w e v 2 0 z 5 o '
] ce
- M X-
- < o EI
- w $,Mo r Z W O
s 4 - W. X-Z w 2 o O w 9140
EQUIPMENT REALIGNMENT FOR TESTING PUMP P35A TO 8A3T A
..f.
GNy!(' THESE VA!.VES ARE OPEr;ED f, LDH 10 hi TO TEST PUMP P35A AND g .- 450 V AC MCC 051 "' 4160 V AC BUS A3 SE bhiN55 CB 5171 l) C8 A304 ) S a g M g i 3 g. g raou swsT " 183I BS4A cv24ci
-:. +
BSir W P35A BW5A B W5A OR SuuP
.!lih.l;. . . . . . . .ss2A; v
TO LP15 TRAIN B (CONTINUED) s - s > , ,
, =
EQUIPMENT REAUCNMENT FOR TESTING PUMP P35A (CONTINUED) TO BWST R ACKED OUT TO TEST PUUP P35A AND LEFT f' I l OPD.* IN STANBY TO RETURN IT TO SERV.CE f4DH10 fiDH9 l 480 V AC L4CC 051 ' 4160 V AC BUS A3 iC n n7f::),:'. . .; CS A304 ) m l Y e l "
,j y ' Q > '
I T j g _ FRou BWST OR SUMP UI A BS4A CV2401 BSI A DW6'A BW5A P35A flBS7A j _ I TO LPIS TR A:N B
EQUIPMENT REAL;GNMENT FOR TESTING PUMP P358 TO BWST d sh : . . flDH10 :: D$ji THESE VALVES ARE OPENED j "- TO TEST PUMP P358 AND 480 V AC MCC B61 ARE CLOSED TO RETURN IT 4160 V AC BUS A4 TO SERVICE. w ff
" ~ "
Nk!! > CB 6171 ) CB A404 ) U ', ' A h l __gp( B548 CV2400 8518 I T ,j BW68 g ; FROM BWST OR SUMP BW58 l.........
.:- - PJ58 fi N523Y F-t TO LP:S TRANA (CONTINUED)
= D . ,
- 9 EQU!PMENT REAUGNMENT FOR TESTING PUMP P358 (CONTINUED)
TO DW5T A RACKED OUT TO TEST PUu? P358 AND LEFT OPEN 1N STAN8Y TO RETURN IT TO SERV.CE 4 DH10 1 DH9 i 450 V AC WCC BS1 4160 V AC BUS A4
'4 BS3 (CUNNU;7! C8 404 .)
0 > E a w v A g , i s ,j g _ rRou owsr Uli Bs4B Cv2410 usia W P358 ows's owse ' OR SUMP 4 8520 l v TO LP!S TRAIN A
MAKEUP TANK AND SEAL RETURN LINE SCHEMATIC FOR THE H!GH PRESSURE INJECTION SYSTEM E26A k FROM U I LETDOA?J , COOLERS ! FROu CONTROLLED y ; BEm&F W l
^^-
' EACH RC PUuP e TO HP 5 "d$$ONMU13 s *Mr 14 M ,
SEAL RETURN COOLERS h CV1274
; Y HEADER MU12 i l,
MAKE UP TANK { i E268 b Y U - SYSTEM MENU END OF INPUT
. - , o
7 HlCH PRESSURE PUMP COOLING SCHEMATIC: ROOM AND LUBE OIL
~
SW LOOP 1 LO. COOLEH E39A g CV3808 b W k WC7A e ROCu 54 FN4 P360 (NORMAL 1Y OFF) b LO. COOLER E398 e 7 m CV3809
- E W E d WC78 ROOM 55 FAN h
P36C ( M RWMC) R b LO. Com E39C OTHER l 3" Cv38to FROMSb Lo m M $ toOP 2
\U CV382I EMERGENCY RO u 56 FM Cv3824 G POND (NCRMALLY GFF)
I DISCHARGE FUJME SYST~c.M MENU END OFINPUT
-m..-mw-.a_..---.-m_--e-.a - --AA.4__ m..m -aeasaw-aa-..m a a<m._.ae a a. , w a .ui>_A_ame- -s - +#a-m sM-m,e ae.m,.._ a.u-6a. Eac2. .4 m aaE--m-Am-w_as-... . _ms- 4. 4 aJ _.
n , 6 4 4 i i J 4 i j 1 I. -r i, , l l .i i 1 A 1 r p t i s I 4 i i I
- l. 4 I b
a : i i l i i - i I < i i l i r 1 1 4 h 4 I l
.- ---_.- -,I
B-1 APPENDIX B DIAGRAM OF Tile PRISIM INFORMATION llIERARCl!Y This appendix contains a diagram of the PRISIM program information hierarchy. The diagram was developed as a visual aid to help you obtain
- PRA related information from PRISIM. The diagram is a user oriented flow chart that details the information provided by PRISIM and shows the series of steps you must take to obtain this information. Diagram interfaces are denoted by identifying triangles.
4 l 9 i
)
B-2 F PRISlM CON TROL + - - SCREEN
.- 3 MASTER MENU 3 1L UPDATED PREPROCESSED PRA INFORMATION INFORMATION v v A B ,
Figure B.1 Diagram of the PRISIM Information liierarchy l
B-3 UPoAMD A PnA WC5esATOd O 4 1 L 4
*** mmoc *mt upoAs Eo.4PtsDIf N TIA20 CPERA M PRA
'g sofs r,Aum usu.fs ir 1F iP iP seuencm ena sutrefoi searH:mem
& DAfA & FACTOR tmf fASLE Emf I
i, ir ir enA unAmo pnA nueme DATA DENT EDif DATA --4 & TAALL PUNCTOf TAEK E M f5 1P 1r so m ent r openAfon no e er FAA.Unt E W T FAdLU81E H CINE DAasAm EDT A DENTEDit gggggag naernow rumenom
' 1r gEMPlf A
H ol' A mtsf0 mates
- newsw to
--e ccm7ska.
sessun ir n , l - j e Figure B.1 (continued) l l 1 J
B-4 l . i j PaEPaoct Mo MrclWAfpq 1 _H I comANT SAM TY- myy.mg ACODENT 4LAfs0 gygggg C . NWeems sWftWs i 1P 1P 1r RANKh4 & ACONNfs BY 5 sWW m4:W W e
& Coat DAWAGE EPORI*# RPSIANM PatCL4NCY M W8MJ t
l I l 1r l DEfAAAD 4Asem0 w AAsesse &- otsOWP W W ,, sufgus gy gggggg gy [ EAQ4 ACQOENT gm MDUC 6 N gg ggg I E h asPMfANm I
! ' - -papdKhG W maanrma y ' ] O ,g sW?this ST orIAa2D smNeo e m stNe w n wo=An0N ran armimer ~
4 =me:WftWs
= sENswn av 1 MLICfED ACODENT E f *==
MOUENG l , et0VPh4 W
! + emnimusso eCANg av 90Wh4 &
i mCownvwo ~- nemi-~ ,=manysus av 4 rom an2C?to = sesecusa : some ] meantam 1 I i wxt
! -* semo >
J
; s I l
- RAAd AAY
- -s w i j somos !
meeway wo
-g FOR E12CitD j ACOODef N OutMCE i .
t b 4 . , 1 J e i , Figure B 1 (continued) l l f i l i t
, -- - - ~ , . , .
t i - B-5 l, I I
; C (
j T l l _,. , _ v SArt fy. 5JPPORf m,g, RELAft0 SWfud ACNDel COWPChCNTS N MRFA38 g 6 1 y i PuTF ,
, I I 1 I l l M. mANuso m m asauctos er !
CCWPONENf , , t_hMm f Ahm e kPORfANCE 4 utNiJ SELECDON SELic;n,0N gjppg 7;ggW p,
& RApotD By i
COWP0NENT CohPONENT/
?
j n)NQMm km EDeS W fy l _ . hrCRf aNm e J y I 9 KW TAA.! W gAgg y Em - , q COWPONENis av ssPPOR fac peupes av ;
'~b , COWPONEN TS COWPONENn r , :SM RE*RJCDCN ggg g W SGNFICANCE
! I (R IWPom f AN2 RJNCEONL
. troupem . I l
- amA3 I RApueD av
( ~ amne w ~
' I ** NW88 l ~
- COWPCNENTS SY -"I A""- 6
) { ESC Rf*pRT SENSfhAfY E i [ mA 4 i naseso av
, et smertWTV l _ @CRJPING & . 1 tre f assa t
_ COWPCsetNTS BY Sk SCN1mCAN GRA SNhrtD St y
* -eno mesecu.ca ! asPaktamm
[ M ORWAt0N I RELIVANY FOR St hoh N ] Af0 a ,D,AR, - 3e _ i .l SEllCfl0N OF 0 COMPONEN f a IPOR-Memed 0>retar mespeese I ORA-Operetar Roeowy Aettene L t i G WO$f L*tLY DCLE FaluJPES FOR APTICTED SWTEWS , i j G IOJAttNT REAUCNWENT FOR WAJNTENANCE OF StilC113 COWPONENT f G E0JPWENT REARQedtNT FOR TESTING SE1.lCTED CChePONENT S M5TCfECAL CAUSES OF FALURE FOR SElfC?to CCWPONENT
- #C SWLAM COUPONENTS Af MS M,ANY 1
1 1 i ! Figure B.1 (continued) 1 i k l j 1 1 i
B-6 t a o , I ,, omecwt mm Lasses m = co w Amesma f AA.L84 M*19es/ 00lERNBe8 Ptat s cim OA1A se'= Ptst cPmAncu seupunn I usSH cr u ir i,
~0 PuMT t#to sase n sasenan suenan
; e e e sase= yl cxnecsent CAMMY 97 { ,,
f we ******* m.:e_n .. m m
= annummie.
= = =. ,,o.. = n amLseno .omecsemi asts cousmaus een mm OtRPcBest (Pemam38 OObreAAhet 9'ama --
carmeem u
-* 7 u ve DATA R. ANT v.
cWERAfgR ' N M lheta r we not wt l some num
--4 PER>t1RY ,
ASAN unen num
-, ocur=,
A SAE l i i i i Figure B.1 (continued) I l i l
C-1 APPENDIX C i QUANTIFICATION OF FAIIDRE EVENTS c.1 Introduction This appendix describes the input parameters and models used to quantify failure events in the PRISIM program. Applications of the parameters and _ models to specific technical specification evaluations are described. An understanding of the PRISIM models is essential for users sho plan to modify PRA input data through the program's PRA data tablos, a ictions C.2 and C.3 describe the input parameter and mo' del options vailable in the PRISIM PRA data table and discuss applications of these aptions to various situations. Section C.4 describes an example of the PRA data table models applied to a problem involving specification of a probability of failure on demand. C.2 Event Probability Model The PRISIM PRA data tables contain a tabular listing of all the events included in the program's reduced core damage frequency equation. These events are unique combinations of subevents that were created to simplify the fault tree analysis performed for the ANO-1 PRA study. All events are independent with respect to the subevents they contain; that is, a particular subevent may appear in only one event. Furthermore, the events are defined as logical "0R" combinations of their respective subevents so that the occurrence of any subevent within an event will cause the event to occur. The subevents contained in PRISIM include failures of specific components or piece-parts. component outages that are due to maintenance or testing, and common cause events. These subevents reflect the maximum level of detail considered in the fault tree analysis. In some instances, subevents with a failure probability of zero have been listed in the PRA data tables. These subevents are included in the tables to indicate that they were considered in the detailed analysis of an event but their contributions to the event's failure probability were negligible. Once you have selected a given event through the DBM routine, all subevents associated with that event are displayed along with the input parameters of the reliability models that were used to calculate each subevent's failure probability. The PRA data table allows the user to modify these input parameters and models as described in Section C.3. Since the events in the PRA data table are defined as logical "0R" combinations of their respective subevents, we can determine the probability of an event in terms of subevent probabilities using the expression:
. - . _ . . .- . . _ . ~, - -
C-2 P E 5 SE
- i- 3 ij where P = the probability.of event i,.
E g P SE
= the probability of subevent j associated with event i.
q e Therefore, after you have . completed changing subevent failure probabilities , . the PRA data . table will automatically sum all subevent probabilities associated with an event and ' calculate the event probability. The new event probability is then used in all calculations . performed during a PRISIM session. C.3 Subevent Parameters /Models C.3.1 Failure Rate (RATE) Several of the reliability models included in the PRISIM PRA data table + require you to assign a subevent, or component, failure rate. In general, a component's time-dependent failure rate, A(t), is defined > such that the product A(t)dt is the probability that the component fails in the incremental time interval t to t + dt, given that it survives up until time t. Since a component failure rate is usually constant throughout most of its service life, the reliability models used-in the PRISIM PRA data table assume constant failure rates that.are effectively average values experienced over a period of time for similar components. ' When assigning failure rates to subevents within the PRA data table, you should keep in mind that most components have different failure rates when they are operating than when they are in standby . and that the , t appropriate values should be assigned according to the subevent definition and model used. C.3.2 Test Interval (TI) : The test interval model in the PRA data table. allows you to model the unavailability of components whose failures are not announced but are ; detected through periodic testing. For these types of failures, the unavailebility of the component during the time interval between tests, T, is equal to a(t) e s(t - To) [(t - To)s T) (C.2) I where As - standby failure rate, i To - time of the most recent test. l, _. _ _ - . _ _ . ._ _ - . . _ - _ - - - - -- ~ - -
C-3 l This equation assumes that the periodic testing- will always detect an existing failure. If we assume that testing and repair are performed instantaneously, then the unavailability of the component with time can be described by the periodic function shown in Figure C.1. Note that l for small values of At (<0.1) the function 1 - e-At is approximately ; equal to At. If we now assume that the demand on the component is equally likely to occur at any point within the test interval, then the average unavailability for the component is obtained by computing the integral ' average unavailability over one test interval, T. Thus
, To+T a(t) de
- To n avg - To+T 1.0 dt '
To To+T 1-e s(t-To) dt To
* ~
- 8 (unannounced failure.*). (C.3)
Equation C.3 is the reliability model used in the PRA data table test interval option. With the test interval capabilities in the PRA data tables, you can , investigate the risk impact associated with proposed changes in
, surveillance tsst intervals specified in the plant Technical Specifications. Af ter you enter a value in the TIME column of the PRA data table (which is equal to the new test interval proposed for the component), specify an appropriate standby failure rate in the RATE column, and select the TI model, the PRA data table will automatically '
calculate the unavailability using equation C 3. PRISIM will then update the FRA to determine the effect of the new test interval. C.3.3 Downtime (DT) The downtime model allows you to model the unavailability of components , whose failures are alarmed or otherwise announced to the operators. If i a component is repairable, and we assume the length of time to repair ! and restore the component to service after failure is exponentially : distributed, then the unavailability is given by
-(A + 1/rg)t D (C.4) a (t) - {1 - e ]
i f i _ _ _ _ ..-_ _ _, _. .. _ _ , _ . _ _ - . ~ _ . , _ - . _ , _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ ,
I 1.0 ------------------------------ 76 Test Component Intervo!. T Uncvoilability i(t)
) _- A ,T ---------
l ! o-i s. e l l 1-e -A ,t l 0 T 2T 3T l Tme I Figure C.1 Unavailability Function for Components Periodically Tested 1 I i l b -4 g g m -. . m,. ,- . . v- ,3-,--, --- - - -,. - w - '__m -- _ ___i__--__
C-5. where to is the average restoration time. When t > 3rp, the unavailability approaches an asymptote and equation C.4 is approximated by A
-a rD .
(C.5) 1, , Equation C.5 is the reliability model used in the PRA data table b downtime option. With the downtice capabilities in the PRA data tables, you can investigate the effect on plant safety that would occur if permanent
- changes are made to the allowed outage times specified for certain components. Once you have determined that the component you are investigating will meet the downtime model assumption concerning announced failures, you may set a time in the TIME column of the PRA data table that is equal to the new allowed outage time proposed for the component, specify an appropriate operating failure rate in the RATE column, and select the DT model. When these three data entries are complete, the PRA data table will automatically use equation C.5 to calculate the unavailability. B It should be noted that the downtime model is also the appropriate model to use for calculating the average unavailability that is due to scheduled testing or planned maintenance. In this case, you assign a failure rate equal to the frequency of scheduled testing or maintenance and a downtime equal to the average time interval the component will be '
out of service for the testing or maintenance. Once these values are assigned, the PRA data table automatically uses equation C.5 to calculate the unevailability that is due to planned outages.
?
C.3.4 Fault Exposure Time (FET) The fault exposure time model allows you to model the unreliability of components that must function for a specified length of time when called upon or the unavailability of components that are nonrepairable. , If the unreliability of a component, f(t), is defined as the probability that the component will fail one or more times during a specified time l interval, given that it is initially working, then it can be expressed i as: l fm
- r-1-e- (C.6) 4 where A has been assumed to be a constant operating failure rate for the component and rm is the mission time.
In a similar fashion, if a component is nonrepairable its unavailability becomes the probability that the component has failed ! prior to time t, which is simply the unreliability. Therefore ! l l l
C-6
~
3(t) - r(t) e (nonrepairable case). (C.7) Equation C.6 (or, equivalently, C.7) is the reliability model used in the FRA data table fault exposure time model. The only difference is l the interpretation you specify and place on the FET. After you assign a mission time or functional time interval in the TIME column of the PRA data table, specify an appropriate operating failure i rate in the RATE column, and select the FET model, the PRA data table .J i (using equation C.6 or C.7) automatically calculates the probability l that the component has failed before the end of the specified time ' interval. 3 C.3.5 Direct Assignment of Failure Probability (p) The PRA data table provides the option to assign a subevent failure probability directly, without utilizing any of the reliability model options previously described. This option is used for quantifying subevents involving human errors and for quantifying subevents describing a probability of failure on demand. Other failure models of more complexity can be used for individual components external to PRISIM, and the results can be put in directly using this option. For an example of a more complex model, see the following section. C.4 Example Problem - Probability of Failure on Demand , The probability of failure on demand can consider both the probability 1 I the component fails while in standby and the unavailability of the component because of planned testing. The following example shows how ' these factors ! are accounted for in calculating the probability and how this problem can be handled in the PRA data table of the PRISIM program. Consider a pump that is tested monthly (every 730 hours), with the test requiring 2 hours to complete. There is no scheduled maintenance for this pu p. 5x10g/hr. We will assume that the failure rate for failure to start is The unavailability function for this example is plotted in Figure C.2. Assuming that a demand on the pump is equally likely to occur at any time, the probability of failure on demand is equal to the integral average unavailability for one cycle. That is ,
.T a(t) dt
=0 f Pg - ,3 1.0 dt
=0
A e , , - y 5
+
i'
! 1,0 ____________ ___________ ___
pd =$ S%$ t r 1
; Component Unovailability
&(t) -2 3.57x10 ------------ -------------- ------ .
o i i i 2 -s u
-5x10 t ! 1-e i
I a O 730 1460
- ume t (hours) t i
Figure C.2 Unavailability Function for the Example Pump l i
+
i 4 .4
, . - - - - - - - . . , - , - , - . . . . - - - _ _ _ _ _ - _ =
C-8 1 728 7 _ ,-5x10-5e)de+f728 30 730 }o 1.0 de 1 - 728 + ,-(5 x 10-5)(728) 1 1 g + g (2) 5 x 10-5 5 x'10-5 .
- 2.07 x 10-2' (C.8) where the first integral represents the integral average unavailability for an unannounced failure detected through testing, and the second integral represents the known unavailability that is due to the planned testing. a This problem can be handled using the PRISIM PRA data table in two separate ways. First, if you have hand-calculated the probability of failure on demand as described by equation C.8, you can enter this value directly in the subevent failure probability column, p, of the PRA data table as described in Section C.3.5.
Alternatively, you may choose to work the example problem by making two separate subevent entries in the PRA data table. The portion of the problem describing the unavailability for an unannounced failure ' detected through testing would be entered as a subevent failure to start) with a standby failure rate (RATE) equal to 5 x 10-5(/hr, a test interval of 730 hours entered in the TIME column, and the test interval (TI) cption displayed in the MODEL column. The PRA data table test 3 interval model (equation C.3) would yield the following:
-AsT+ -(5 x 10-5)(730) -5 s T-1 e + (5 x 10 )(730) -1 i a -
AT s (5 x 10-5) (730)
- 1.8 x 10 2 (C.9)
Next, the known unavailability that is due to testing would be entered as another subevent (out of service for test) with a failure rate (RATE) i equal to the expected test frequency of once per 730 hours, a time value (TIME) equal to the expected outage time for testing of 2 hours, and the downtime (DT) option displayed in the MODEL column. The PRA data table ; downtime model (equation C.5) would yield u
* ~
'D ~
730 (2) 1+Ar D 1 + h (2) f
- 2.7 x 10-3 (c.10) l l
C-9 Finally, since the PRA data table sums the subevent failure probabilities automatically to obtain the event probability (adds C.9 and C.10), we would have the following approximation to equation C.8: Pg (FRA data table) - 1.8 x 10-2 + 2.7 x 10-3 (C.ll)
- 2.1 x 10-2, t
b 9
\
l s j
1 l l l NOTICE When you no longer need this report, please return it to G. F. Flanagan, Engineering Physics and Mathematics Division, Bldg. 6025, MS 09W, Oak Ridge National Laboratory, P.O. Box X, Oak Ridge, Tennessee 37831. 'j Thank you. s e i e i 1 1
NUREG/CR-5021 Vol. 2 ORNL/TM-10604/V2 Distribution Category RG INTERNAL DISTRIBUTION
- 1. J. R. Buchanan 14. L. C. Oakes 2-4. G. F. Flanagan 15 D. L. Selby k 5. J. K. Ingersoll 16. A. Zucker
- 6. J . E. Jones , Jr. 17. J. J. Dorning (Consultant)
- 7. H. E. Knee 18. G. H Golub (Consultant)
- 8. T. S. Kress 19. R. M. Haralick (Consultant)
- 9. F. C. Maienschein 20. D. Steiner (Consultant)
- 10. A. P. Malinauskas 21. Central Research Library
- 11. J. R. Merriman 22. .-12 Document Ref. Section
- 12. D. L. Moses 23. Laboratory Records ORNL, RC
- 13. F. R. Mynatt 24. ORNL Patent Office 25-27. EP&MD Reports Office EXTERNAL DISTRIBUTION
- 28. Office of the As.,istant Manager for Energy Research and Development, DOE-ORO, Oak Ridge, TN 3'830, 29-36. JBF Associates, Inc., 1000 Technology Park Center, a Knoxville, TN 37932:
D. J . Campbell V. H. Guthrie F. M. Dycus J. R. Kirchner
, B. C. Ellison J. Q. Kirkman J. A. Farquharson H. M. Paula 37-44. U.S. Nuclear Regulatory Commission, Washington, DC 20555:
G. R. Burdick J . C. Glynn A. J. Buslik J. Murphy E. J . Butcher R. C. Robinson F. Congel M. L. Wohl
- 45. K. G. Murphy, U.S. Nuclear Regulatory Commission, 631 Park Avenue, King of Prussia, PA 19406
- 46. R. Walker, U.S. Nuclear Regulatory Commission, 101 Marietta ,
St., Suite 3100, Atlanta, CA 30303. !
- 47. J. E. Gagliardo, U.S. Nuclear Regulatory Commission, 611 l' 6 Ryan Plaza Dr., Suite 1000, Arlington, TX 76011.
- 48. W. D. Johnson, P.O. Box 2090, Russellville, AK 72801.
- 49. D. F. Kirsch, U.S. Nuclear Regulatory Commission, 1450 g Maria Lane , Suite 210, Walnut Creek, CA 94595.
50 51. Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN. 52 251. Given distribution as shown in NRC Category RG, Systems and Reliability Reports.
.u s covowsm nwnsc omc i i e u . e.: o e i e e l
NAC FC1. att u s ~ucts Aa af Gutafoav cowwisseo.s I he *Oa , %b v8.a s. #w .* OC *** V8' ** . *8 8ar'
'*/c'O on' NUREC/CR-5021, Vol. 2 LlOGRAPHIC DATA SHEET mi. am lit .%$taWit:0%5 0, f at alvt all ORNL/TM-10604/V2 f
)
3 tirkt .sa sw.ritit 3tt.vt.6.%s User's Guide for PRISIM Arkansas Nuclear One - Unit 1, Volume 2,PrograkforRegulators N( a o., e . *oa r go.et e t t o n .o%, . .t.a
. .or .c. .s. g December f 1987 D. J. Campbell, V. H. Ch,thrie , J. R. Kirchner, . o*,saiao',,vio J. Q. Kirkman, H. M. Paub, B. C. Ellison, voa "'"
F. M. Dycus, J. A. FarquhArson, and G. F. Flanagan Januaryj7 l 1988 t .........,.o......,.,,....~o....00...,.<. . . _ ..., ,. - ..- A
- N JBF Assoeiates, Inc. a Knoxville, TN 37932 i .,,o<[,Ar.....
f (N jf
,40825
. . . . o s so. . o c . . . ,, , . . ., , . . . ,0 . . . o . . . . ,, ,,- , . c , ,
y ... o. . . -, Oak Ridge National Laboratory g P.O. Box X ' 7 Unc Oak Ridge, TN 37831 1, ,,8 * " a'oo c o' ' a'o ~ " "".
\
g
,, ,2.....s,...~o .,
\
y N)V L T . . . . a , a x .. ,i This itser's guide is a two-volume document d igned to teach NRC inspectors and NRC regulators how to access probabilistic riskfisigssment ~f information from the two Plant Risk Status Information Management temi(PRISIM) programs developed for Arkansas Nuclear One - Unit One (ANO-1), olume'} describes how the PRA information available in Version 1.0 of PRISIM is use ul for planning inspections. Using PRISIM, inspectors can quickly access PRA infopp'ation and Ope that information to update risk
. analysis results, reflecting a plant'ststatus at an), particular time. Volume 2 describes how the PRA information aval'lable in Version 2.0 of PRISIM is useful as an evaluationtoolforregulatoryactigties. Using PRISJM, regulators can both access PRA information and modify the infgrmation to assess the impact these changes may have on plant safety. Each volume is afstand-alone document.t
.46 4
# 3 y '.
o % j' Jr i,- f i
, . c : . . ,. , , , . . . . . . . . . - n . , ; . , . 7 4 q,. .c;, .
, l Risk Analysi a
\ v,
't . SE ve f ' C t.$,' 8 eC . t.c g ,
'a 4 ..ves 5
. u m i 5:*t,s u o 's v5 .
Unc. 3,,....., g i Unc.
, p. . . o. . . a ,
d i 224
...s.
l
12- 0 5.,55 0;7 B 9 7 7 1 1AN1RG g;.0A1M-43M {' $9hu GT HR-PDR fiUREG
- W-537 OC 20555 WASHINGT0f b
i s}}