Text

Safety Parameter Display Sys Safety Analysis
	ML20087A949
Person / Time
Site:	Cooper
Issue date:	03/01/1984
From:	Lobner P; SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	;
Shared Package
ML20087A940	List: A840006, Forwards Three Repts Re Facility Safety Parameter Display Sys (Spds),Per Requirements of Suppl 1 to NUREG-0737. SPDS Will Be Installed During Upcoming Recirculation Piping Replacement Outage; ML20087A949; ML20087A964; ML20087A998;
References
	503-8500000-76, TAC-51232, NUDOCS 8403080268
	Download: ML20087A949 (72)
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AGREEMENT NO. 83A-C5 1-323-05-766-XX NEBRASKA PUBLIC POWER DISTRICT Plant Management information System Cooper Nuclear Station SAFETY PARAMETER DISPLAY SYSTEM SAFETY ANALYSIS DOCUMENT NO. 503-8500000-76 MARCH 1,1984 Technical Review Peter Lobner Date Author Technical Reviewer Date Documentation manager I 0 < dh Date /dl/ i Configuration Manager Date O. A. Manager Date Date / h i Principal Investigator W Date Division Manager 8 SCIENCE APPLICATIONS,INC. da 2109 W. Clinton Avenue, Suite 803,Huntsville, AL 35805 * (205) 533-5900

NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION SAFETY PARAMETER DISPLAY SYSTEM SAFETY ANALYSIS March 1, 1984 l P. R. Lobner D. J. Wester Staff Scientist Project Manager Science Applications, Inc. Science Aoplications, Inc. 1200 Prospect Street 2109 West Clinton Avenue P.O. Box 2351 Suite 800 La Jolla, CA 92038 Huntsville, AL 35805 619/456-6264 205/533-5900 t --,_-_.-s __,vw, ,--,__..,__v_-

503-8500000-76 3/1/84 TABLE OF CONTENTS Szction Page 1 INTRODUCTION........................ 1 1.1 Purpose of the SPDS...........,...... 1 1.2 Users of the SPDS................... 2 1.3 Scope of this SPDS Safety Analysis 3 2 OVERVIEW OF THE CNS SPDS.................. 5 2.1 Level 1 Displays 6 2.2 Level 2 Displays 7 2.3 Level 3 Displays 9 3 PLANT INFORMATION TO SUPPORT THE BWR OWNER'S GROUP EPGs... 23 3.1 Overview of the BWROG EPGs 23 3.2 Uses of Plant Information in the BWROG EPGs...... 26 3.3 Specific Plant Information Required by the BWROG EPGs. 25 4 PLANT INFORMATION TO SUPPORT DETERMINATION OF SAFETY FUNCTION STATUS...................... 46 t 5 DISPLAY HIERARCHY AND PLANT INFORMATION CONTENT OF THE t i 4 BWP. 0WNER'S GROUP GRAPHIC DISPLAY SYSTEM (GDS)....... 49 i 6

SUMMARY

OF INFORMATION NEEDS................ 53 7 PLANT VARIABLES THAT ARE EXPECTED TO BE AVAILABLE ON THE CNS SPDS........................ 60 8 CONTROL ROOM SOURCES FOR IDENTIFIED PLANT INFORMATION THAT IS NOT EXPECTED TO BE AVAILABLE ON THE CNS SPDS....... e4 9 CONCLUSIONS 66 10 REFERENCES......................... 67 f i

~ 503-8500000-76 3/1/84 LIST OF FIGURES l Figure Page 1-1 Overview of Safety Analysis of the Cooper Nuclear Station Safety Parameter Display System (SPDS)........... 4 2-1 Functions of the CNS Plant Management Infonnation System (PMIS)........................... 11 2-2 Expected Hierarchy of SPDS Displays for the Cooper N ucl ea r S ta ti o n....................... 12 2-3 Expected CRT Layout for the Cooper Nuclear Station SPDS... 13 ?.-4 Heat Capacity Temperature Limit 14 14 2-5 Heat Capacity Level Limit 2-6 Suppression Pool Load Limit 15 2-7 Pressure Suppression Pressure Limit 15 2-8 Primary Containment Pressure Limit............. 16 2-9 Primary Containment Design Pressure 16 2-10 Drywell Spray Initiation Pressure Limit 17 2-11 Hydrogen and Oxygen Deflagration Overpressure Limits.... 18 2-12 Pump Net Positive Suction Head (NPSH) Limits........ 19 2-13 Boron Injection Initiation Temperature Limit........ 20 2-14 RPV Saturation Temperature Limit.............. 20 2-15 Maximum Core Uncovery Time Limit.............. 21 2-16 Alternate Shutdown Cooling and RPV Flooding Pressure Limits. 21 2-17 Alternate RPV Flooding Pressure Limit 22 2 *8 RPV Pressure / Level Status Matrix.............. 22 3-1 Overview of the Organization of the BWROG Emergency Procedure Gui deli nes.................... ES 5-1 Hierarchy of Displays in the BWROG Graphic Display System 51 ii

503-8500000-76 3/1/84 LIST OF TABLES Table Page 3-1 Sumary of Relationships Among Major Elements of the BWROG EPGs............................ 29 3-2 Sumary of Conditions Requiring Entry into the EPGs and/or Transition Between Elements of the EPGs 30 3-3 Summary of EPG Contingency Entry Conditions 37 3 Use of Multiple-Parameter Limits in the EPGs........ 40 3-5 Plant Information Required to Generate Multi-Parameter Graphic Displays...................... 42 3-6 Sumary of EPG Plant Information Requirements 44 4-1 Approximate Relationship Between NUREG-0737, Supplement 1 Safety Functions and the Regulatory Guide 1.97 Variables.. 47 4-2 Sumary of Plant Infonnation Needed to Determine Status of S a fe ty Fu n :ti o n s,...................... 48 5-1 Sumary of Plart Inforn:ation Displayed on the BWROG Graphic D i s pl ay Sy s tem....................... 52 6-1 Relatienship Between NUREG-0737, Supplement 1 Safety Functions and SWROG Emergency Procedure Guidelines (EPGs) 55 6-2 Compirison of EPG and Safety Function Data Requirements 56 6-3 Reg. Guide 1.97 Variables Related to NUREG-0737, Supplement 1 Safety Functions but not Referenced in BWROG EPGs............................ 58 6-4 BWROG Grapnic Display System Variables not Referenced in BWROG EPGs......................... 59 7-1 Plant Variables Expected to be Monitored by Cooper Nuclear Station SPDS 61 8-1 Sources of Identified Plant Information not Expected to be Monitored by the Cooper Nuclear Station SPDS........ 65 iii f L

503-8500000-76 3/1/84 1. INTRODUCTION 1 l l i A Safety Parameter Display System (SPDS) is one of the facilities and systems required by the Nuclear Regulatory Commission to provide improved emergency response to accidents at commercial nuclear power plants. The SPOS is a control room enhancement that is related to the following I other initiatives aimed at improving emergency response capabilities: o Detailed control room design review ~ e Regulatory Guide 1.97 application to Emergency Response Facilities e Upgraded Emergency Operating Procedures e Emergency Response Facilities (ERFs) e Operating staff training This safety analysis has been prepared in compliance with the requirements of NUREG-0737, Supplement 1 (Ref.1) to describe the basis on which parameters selected for display on the Cooper Nuclear Station (CNS) SPDS are sufficient to assess the safety status of the plant for a wide range of events, including potentially severe accidents. 1.1 PURPOSE OF THE SPOS The principal purpose and function of the SPDS is to aid the Cooper Nuclear Station control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective actions by operators to avoid a degraded core. As implemented at CNS, the SPDS will accomplish its pur-pose by providing control room personnel with concise displays related to the following plant safety functions which have been identified in NUREG-0737, Supp.1 (Ref.1): e Reactivity control e Reactor core cooling and heat removal from the primary system o Reactor coolant system integrity 1

503-8500000-76 3/1/84 e Containment conditions e Radioactivity control An additional purpose of the CNS SPDS is to support the implementation of symptem-oriented emergency operating procedures (EOPs) that will be developed from the BWR Owner's Group (BWROG) Emergency Procedure Guidelines (EPGs) (Ref. 2). This is consistent with the NRC recommendation that the SPDS, E0Ps and other control room initiatives "should be integrated with respect to the overall enhancement of operator ability to comprehend plant conditions and cope with emergencies" (Ref.1). 1.2 USERS (Y THE SPDS At CNS, the users of the SPDS will include the following members of the control room staff: e Reactor operator e Shift supervisor e Shift technical. advisor There will be no single person assigned as the primary user of the SPDS. All potential users of the SPDS will be trained to perform their functions irrespective of SPDS availability. 1.3 SCOPE OF THIS SPDS SAFETY ANALYSIS As required by NUREG-0737, Supp.1 (Ref.1), this safety analysis describes the basis on which the parameters selected for display on the CNS SPDS are sufficient to assess the safety status of the plant for a wide range of events, including potentially severe accidents. The CNS SPDS is a subsystem of a larger Plant Management Information System (PMIS). This safety analysis addresses only to the SPDS portion of the PMIS.

503-8500000-76 3/1/84 The selection of plant information for display on the CNS SPDS is driven by the stated purpose of the SPDS: e to provide information on plant safety functions e to provide information in support of symptom-oriented emergency operating procedures (EOPs) NUREG-0737, Supp.1 and Regulatory Guide 1.97 (Ref s.1, 3) are the techni-cal bases f.-* identifying key plant information related to specific safety functions. The BWR Owner's Group EPGs (Ref. 2) are the technical bases for identifying information needed to support the implementation of symptom-oriented E0Ps. The EPGs are particularly useful as a tool for establishing realistic information needs of the operator for assessing the safety status of-the plant for a wide range of events, incluaing potentially severe acci-dents. Needless to say, the SPDS will not be a single source for all of i the important plant information identified by the above procedures. To the extent practical, the identified plant information will be partitioned among a logically related hierarchy of SPDS displays based on prior work by the BWROG (Refs. 4, 5). Sources in the control room will be identified for important information that will not be included on the SPDS. In summary, the variables selected for display on the CNS SPDS will be determined and justified based on: (a) a review of the actual plant information needed by control room operators to assess safety function status and implement symptom-oriented E0Ps, (b) compatibility of specific variables with a logical display hierarchy, and (c) availability of informa-tion from other control room sources. This approach to SPDS safety analysis is illustrated in Figure 1-1, and is believed to be fully responsive to the requirements for an SPOS safety analysis as stated in NUREG-0737, Supp.1 (Ref. 1). It is expected that the specific variables and displays provided on the CNS SPDS will be revised in the future, if necessary to improve integration of the SPDS with plant-specific E0Ps and operator training. I

O 'r' 8? .8 o 10ENilFY 1(EY o' PL ANT INF ORM ATiON REQUIRf010 SUPPORT BWROG EPGs REL AIE NUREG s737, y SUPPLEMENT I SAf E TY SUMMARIZE FUNCTIONS TO INFORMATION j l REG.Gul0E NEEDS AND USES l 1.97 PARAMETERS f l RELATE IDENTIF Y NEEDED u nEviEw swROG INFORMarl0N INFORMAriON T0 sE 3,,,,, c, y,,, GR APHIC DISPL AY E DSTO WOS W PRWIDM FROM SASEllNE SPOS SYSTEM I H OL DISPLAYS FOR CNS HEtHARCHY ROOM SOURCES Figure 1-1. Overview of Safety Analysis of the Cooper fluclear Station Safety w Parameter Display System (SPDS). 3 s CD s-

503-8500000-76 3/1/84 1 2. OVERVIEW 0F THE CNS SPDS The CNS SPDS is a subsystem of an integrated computer system called the Flant Management Information System (PMIS). The PMIS is com-prised of: (a) a modular, intelligent, multiplexed, front-end data acquisi-tion subsystem, (b) redundant preprocessors, (c) modern, high-speed, real-ti me, mul ti-user, multi-tasking central processors coupled with operator-interactive sof tware,, and (d) color graphic display equipment. The PMIS incorporates the following functions: (a) all functions of the existing GE/ PAC 4020 Plant Process Computer, (b) the Safety Parameter Display System, (c) ability to characterize and predict radiological plumes, (d) the func-tions of the transient recording and analysis system, and (e) additional plant management systems. The PMIS will provide improvements in the ability of the plant operators and support staff to determine the status of the plant, avoid abnormal events, and react promptly to recover from adverse conditions. Human-factored CRT displays will ' assist the operator in assessing the plant status and will guide him in the response to abnormal plant conditions. Appropriate sample rates and on-line, long-term data storage and retrieval capabilities are provided to support post-transient analysis and core performance calculations. An overview of the various PMIS functions, including the SPDS is shown in Figure 2-1. The information presented to the control room operators via the SPDS is expected to be structured into a three-level hierarchy of color grapk' _isplays as shown in Figure 2-2. This is similar to the approach taken oy the BWR Owner's Group in the development and simulator testing of their Graphic Display System (see Section 5). The basic CRT screen format to be used in the CNS SPDS is shown in Figure 2-3. A general description of the expected CNS SPDS display hier-archy follows. The final set of variables and displays provided on the CNS SPDS will be revised in the future, if necessary to improve integration of the SPDS with plant-specific E0Ps and operator training. 5

503-8500000-76 3/1/84 2.1 LEVEL 1 DISPLAYS (Plant Overview) A key feature of the CRT layout shown in Figure 2-3 is the SPDS Status Area (SSA) which includes five rectangular blocks, or safety function indicators that show, at a glance, the current status of the following five safety functions: e Reactivity control e Core cooling (and heat removal from the priraary system) e Coolant system integritv e Containment integrity e Radioactive release These are the same safety functions identified by the NRC in NUREG-0737, Supp.1 (Ref.1). An alternative being considered is to replace the five safety function indicators with "EPG Entry Condition Indicators" that re-flect a better integration of the SPDS with emergency operating procedures. The relationships between the safety functions and the EPG entry conditions are described in Section 6. A Level 2 display provides more detailed information on plant variables related to each safety function (or EPG entry condition). Each block in the SSA is color coded to indicate status of the safety function, and is controlled by the plant variables in the corresponding Level 2 dis-play. For example, during normal power operation, all safety function indicators will be green. A block will change to display the out-of-limits color of the most severe alarm status of any variable contained in the corresponding Level 2 display. The Level 1 display for the CNS SPDS is actually SPDS Status Area, which appears in the same location on all SPDS displays. Regardless of whicn SPDS display is being viewed, the operator is constantly appraised of the current status of plant safety functions. To provide additional infor-mation of an overview nature, the following plant variables are expected to be displayed in the General and Graphic Display Area (GGDA, see Figure 2-3) of the Level 1 display: 6

503-8500000-76 3/1/84 e Average Power Range Monitor (APRM, average) e Reactor Pressure Vessel (RPV) water level e RPV pressure e Drywell pressure These variables will be displayed as bar graphs, current values and rates-of-change. 2.2 LEVEL 2 DISPLAYS (Saf.ety Functions) The Level 2 displays consist of a bar chart display and a time (trend) plot display of variables related to each of the safety functions identified previously. In addition, the Level 1 safety function indicators appear in the SPDS Status Area of every Level 2 display. The plant variables that are expected to be displayed in the CNS Level 2 displays include the following: e Reactivity control APRM (individual) Intermediate Range Monitor (IRM, average) IRM range IRM position Source Range Monitor (SRM, average) SRM position Scram demand status All-rods-in status Suppression pool temperature (average) e Reactor core cooling and heat removal from primary system RPY water level (narrow, wide, refueling, and fuel zone ranges) RPV pressure Drywell temperature (local, near cold reference leg instru-ment vertical runs) e Reactor coolant system integrity RPV pressure Drywell pressure Main steam isolation valve (MSIV) isolation demand (Group 1) status 7

503-8500000-76 3/1/84 Safety / relief valve (SRV) position Drywell sump collection rate (sump pump flow rate) Containment activity (area radiation) e Containment conditions Primary containment conditions (drywell) Drywell pressure Drywell temperature Drywell hydrogen concentration

Drywell oxygen concentration Containment isolation demand status (Groups 1 to 7)

Primary containment conditions (suppression chamber) Suppression chamber (torus) pressure Supprcision pool (torus) water temperature (average) Suppression pool (torus) water level Suppression chamber (torus) hydrogen concentration

Suppression chamber (torus) oxygen concentration Containment isolation demand status (Groups 1 to 7)

Secondary containment (reactor building) conditions Secondary containment differential pressure Secondary containment area temperature alarm status

Secondary containment HVAC exhaust radiation level alarm status
Secondary containment area radiation level Secondary containment floor drain sump alam status
e Radioactivity control Offsite radioactivity release rate from plant release points Elevai;ca elease point (ERP) effluents Augmented off-gas ( A0G) and radioactive waste (RW) building effluents Reactor building effluent Turbine building effluent Steam jet air ejector (SJAE) monitors
These variables are not presently available on the CNS PMIS, but may be added in the future. At that time, these variables will be available for display on the SPOS.

8

11 503-8500000-76 3/1/84 i 2.3 LEVEL 3 DISPLAYS (EOP Support) The Level 3 displays consist of graphic plots that display the proximity of the plant to multiple-parameter limits curves or decision points that are specified in the BWROG Emergency Procedure Guidelines (Ref. 2). In addition, the Level 1 safety function indicators appear in the SPDS status area of every Level 3 display. The following limit curves defined in the BWROG EPGs are expected to be presented as Level 3 displays in the CNS SPDS: e Suppression pool heat capacity temperature limit e Suppression pool heat capacity level limit e Suppression pool load limit e Containment pressure limits Pressure suppression pressure limit Primary containment pressure limit Primary containment design pressure e Drywell spray initiation pressure limit Drywell hydrogen and oxygen deflagration overpressure limits

e e

Suppression ch' amber hydrogen and oxygen deflagration overpressure limits

e Pump net positive suction head (NPSH) limits (RHR, LPCS, HPCI, and RCIC pumps) e Baron injection initiation temperature limit e

RPV saturation temperature limit e Maximum core uncovery time limit a RPV flooding and alternate shutdown cooling limits ** Maximum / minimum alternate shutdown cooling RPV pressure Minimum RPV flooding pressure Minimum SRV reopening pressure e Alternate RPV flooding pressure limit ** e RPV pressure / level status matrix See note on page b.

- These limit curves are not illustrated in the BWROG EPGs, however the bccis for the example multi parameter linits shown in Figures 2-16 and 2-17 are specified in the EPGs.

9

503-8500000-76 3/1/84 Examples of these limit curves, abstracted from the BEGG EPGs (Ref. 2) are shown in Figures 2-4 to 2-18. The CNS Level 3 displays will be plant-specific versions of the applicable generic limit curves, therefore, the final set of the CNS Level 3 SPDS displays may be different than the set shown iti this section. e i 10

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CURRENT DATE AND' OPERATOR COMM. AREA (OCA) TIME AREA (CDTA) i xi s I44 PD [ Yf f fSYSTEM ALARM IIl IIllilli bbIll 1 l1l i lbl i lRI l lIl I l l ll l l i lb i l]lJ[l l 1 l Wl l l l l l i ll y, l hlI ll I llW ulb l, AREA (SA A) g ~ ~ ~ 1 ~. ~ ~ 8 ~ g g - - ~ .-GENEHAL AND GRAPillC DISPLAY AREA (GGDA) ~ ~ ~ (40x 85) g ~ ~ ~ ~ 1 L COLOR OUN AREA F(COA) l ~ r s e n h k e na b b r=- i s t u m u k u p u r I qqquq q q q q r[ l W a d d i t u t h gI' - SPDS STATUS g q.q.an h mWen l I cool.an;. CWIAMENT

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%l AREA (SSA) REACTIVITY . COHE l SYSTEW INTEGRITY, yi GRI 4 RREASE g . COOUNG F HN Ibu f d II I JI l lh I l lh I l l I I I I I I GI I I } Ib g l l i g i l l i b i b u t a bi l l 0 l pFUNCTION MEY AREA R (pggy q ini rain. u ?? Figure 2-3. Expected CRT Layout for the Cooper Nuclear Station SPDS. M $. a >p 9...;S.9.:;.s_. j[ W. :;&... ;,.p ;; f....-n, 3.,9..yg;._.nqqk 'g#*.p..;7" 3 g 1 4 / ** 4 7 - p

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503-850C000-76 3/1/84 .[18 160 <, b'~ h g Rat Ca:acity Te;eratart Limit g122 t i o 0 135 240 logo DT Prsamu., (;akg) Figure 2-4. Heat Capacity Temperature Limit. 312'2" ", w ~ / 'Haat Capacity Level Limit 5 / 5 8'2" cf 8; E b Ji 5 0 27 YU 6T HC Uhere AT = H st Capacity Te=perature Li: it =inus EC suppression pool te perature Figure 2-5. Heat Capacity Level Limit. 14

503-8500000-76 3/1/84 20 -- T E a b

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Suppresssion Pool / o Lcad Limi t E / S

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400 1200 RPV Pressure (psig) Figure 2-6. Suppression Pool Load Limit. ^ cn t 'l / 5

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g Pressure Suppression Pressure ~ m U 5. a 0 0 12.2 17 34.5 Frimary Cor.tcir. ment L'ater Level (f t.) igure 2-7. Pressure Suppression Pressure Limit. 15

503-8500000-76 3/1/84 ^ ee i W/ .oj Pri=sry Cc=tei =ent g ."rescura Li=it c .3: Sg 0 0 12.5 34.5 Pri=ary Contaic=ent Vater Level (ft.) Figure 2-8. Primary Containment Pressure Limit. ^n g 56 f / 2 45.5- .o j ? i=sry Contaf-mut u Design Pntssuro E a 0 0 12.5 34.5 Pri=r.ry C:,etai====t Cater bel (f t.) Figure 2-9. Primary Contt.inment Design Pressure. 16

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EI E4 \\ Drywell Spray Initiation 5 Pressure Limit 100 ?' 10 20 30 40 50 60 Drywell Pressure (psig) l l Figure 2-10. Drywell Spray Initiation Pressure Limit. 1 j 17

503-8500000-76 3/1/84 / //// / // // / 15 Hydrogen Def l agr e.t i on Overpressure Limit Hydrogen 10 -- [ C=ncenteatten (.,. ) 5 a O O 5 10 15 20 25 Drywell P r e s t.u r e (psig) // / / / / / 15 O>: y g en Deflagration Overpressuro Limit 0::yg en 10 / Concentratien / ( */. ) j 5 O O 5 to 15 20 25 Drywell Pr e s r.ur e (pcig) Fi gura 2-11. Hydrogen and Oxygen Deflagration Overpressure Limits.*

See note on page 8.

18

I 503-8500000-76 3/1/84 240-10 psig Eo .psig ou 220 s

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"" 200 - RHR NPSH Limit 2 4 6 8 10 RHR Pump Flow (X1000 gpm) ),0 psig 240 - oC 5 psig N 'N -= 220 w' 0" 0 psig a s-uo saIU e* LPCS NPSH Limit 200-t 2 4 6 LPCS Pump Flow (X1000 gpm) Figure 2-12. Pump Net Positive Suction Head (NPSH) Limits. 19

503-8500000-76 3/1/84 //// / ' !!// / Ecron / 130 Injection Initiation Suppression Pool 120 [ Temperature Temperature [ ("F) / 110 -- 100 0 1 '2 3 4 5 6 7 Reactor Power (%) Figure 2-13. Baron Injection Initiation Temperature Limit. 550 \\\\ N N 8-

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.j* z il i j g. i. .t. i .a_. I i i I i l I I l,il Si b 8 i 0 1 min. 10 min. 1 hr. 10 hr. 100 hr. Time After Reactor Shutdown. Figure 2-15. Maximum Core Uncovery Time Limit. Paximum Alternate Shutdown Cooling RPV Pressure 172 gsd

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E pt-STO,6 *4 5je ] 94. 9,\\" - 9g 9,e5 s 9 og 77 g\\*- gott a-3o- { 50 - 0 z's s'a Suppression Chamber Pressure (psig) Figure 2-16. Alternate Shutdown Cooling and RPV Floeding Pressure Limits. 21

503-8500000-76 3/1/84 I i 1000 - l I 1 I l 7. I i 1 I e i e j 500 - Minimus Alternate g RPY Flooding Pressure SE 400 - 300 - l 200 - I 100 - 0 0 1 2 3 4 5 6 7 8 9 10 11 12 Number of Open sRVs Figure 2-17. Al, ternate RPV Flooding Pressure Limit. RPV PRESSURI REGION [425 psig]l [100 pc'g]2 EIGH INTERMEDIATE LOW INCRIASING Ea h DECR~ASING Figure 2-18. RPV Pressure / Level Status Matrix. 22

503-8500000-76 3/1/84 3. PLANT INFORMATION TO SUPPORT THE BVR OWNER'S GROUP EPGs 3.1 OVERVIEW 0F THE BWROG EPGs The BWROG EPGs (Ref. 2) are generic symptomatic guidelines from which plant-specific, symptomatic, emergency operating procedures will be developed for the Cooper Nuclear Station. The BWROG EPGs describe in generic terms how a BWR plant should respond to emergencies and events which may degrade into emergencies. Fron a normal power operating state, " entry conditions" have been specified for implementing each of the fol-lowing guidelines: e Reactor Pres:;ure Vessel (RPV) Control Guideline o Primary Containment Control Guideline e Secondary Containment Control Guideline e Radioactivity Release Control Guideline i The RPV Control Guideline describes procedures for maintaining adequate core cooling, shutting down the reactor, and cooling down the RPV to cold shutdown conditions. This guideline is entered for any of the following reasons: (a) low RPV water level, (b) high RPV pressure, (c) high drywell pressure, (d) a condition which requires MSIY isolation has occur-red, or (e) whenever a condition which requires reactor scram exists and j reactor power is above the average power range monitor (APRM) downscale trip l or cannot be determined. There are three sub-elements of the RPV Control Guideline which are implemented concurrently when any entry condition listed above occurs. These three sub-elements are: i l e RPV level control (RC/L) e RPV pressure control (RC/P) l e Reactor power control (RC/Q) l l l l 23 m o y

503-8500000-76 3/1/84 The Primary Containment Control Guideline describes procedures for maintaining primary containment integrity and protecting equipment in the primary containment. This guideline is entered whenever suppression pool temperature, drywell temperature, containment temperature, drywell pressure, suppression pool water level, or primary containment hydrogen concentration is above its high operating limit, or suppression pool water level is below its low operating limit. There are six sut'-elements of the Primary Contain-ment Control Guideline which are implemented concurrently when any entry condition listed above occurs. The,se six sub-elements are: o Suppression pool temperature control (SP/T) e Drywell temperature control (DW/T) e Containment temperature control (CN/T, applicable only to plants with a Mark III containment) e Primary containment pressure control (PC/P) e Suppression pool water level control (SP/L) e Hydrogen and oxygen concentration control (PC/H) It should be noted that the Cooper Nuclear Station has a Mark I containment, therefore, the containment temperature control sub-element is not appli-cable. The Secondary Containment Control Guideline describes procedure for protecting equipment in the secondary containment, l i.m i ti n g radioactivity release to the secondary containment, and either maintaining secondary. containment integrity or limiting radioactivity release from the secondary containment. This guideline is entered whenever a secondary containment temperature, radiation level, or water level is above its maxi-mum normal operating value or secondary containment differential pressure j reaches zero (i.e., no longer at a lower-than-atmospheric pressure). There j are three sub-elements of the Secondary Containment Control Guideline which are implemented concurrently when any entry condition listed above occurs. l These three sub-elements are: I l e Secondary containment temperature control (SC/T) e Secondary containment radiation level control (SC/R) e Secondary containment water level control (SC/L) i 24

503-8500000-76 3/1/84 The Radioactivity Release Control Guideline describes procedures for limiting radioactivity release into areas outside the primary and secondary containments. This guideline is entered whenever the offsite radioactivity release rate is above that which requires an Alert condition to be declared. There are no sub-elements of the Radioactivity Release Control Guideline. In addition to the four basic gui.delines described above, the following seven contingency guidelines are incluced in the BWROG EPGs: e Level restoration (Contingency 1) e Emergency RPV depressurization (Contingency 2) e Steam cooling (Contingency 3) e Core cooling without level restoration (Contingency 4) e Alternate shutdown cooling (Contingency 5) e RPV flooding (Contingency 6) e Level / power control (Contingency 7) These contingencies provide for a graduated response to abnormal events and emergency conditions based on the seveiity of the initiating event and the success of safety systems in mitigating the initiating event. Conditions for entering these contingencies from the four basic guidelines or from another contingency are specified in the BWROG EPGs. Ultimately, control is re-turned to the four basic guidelines or to other plant procedures for cool-down to a cold shutdown condition. The relationships that exist between normal power operations, the - four basic guidelines, f.he seven contingencies, and a cold shutdown condi-tion are illustrated in Figure 3-1. In this figure, the following conven-tions are used: OTHER BLOCKS l l l IDENTIFIER $ 0F y OTsER. Lout i j OF N 10ENTIFIER OF BLOCK (XX) TR ANSFERS TO OTHER BLOCKS 25

503-8500000-76 3/1/84 A more detailed summary of these relationships can be found in Table 3-1. 3.2 USES OF PLANT INFORMATION IN THE BWROG EPGs Plant information required by the BWROG EPGs is used in the following four basic ways: e to determine the existence of an EPG entry condition from a normal 1-power operating plant state .o - ; to determine the need to transfer to, or implement concurrently, a , ~7C, different part of the EPGs (or other plant procedures) after ~ initial entry into the EPGs e to establish proximity to multiple-parameter limit curves or ' decision points specified in the EPGs - ~ 'e to support step-by-step implementation of a specific part of the EPGs ~ The first three uses. listed above are candidates for being supported by the SPDS, and will be examined in greater detail in this section. Support of step-by-step-procedures is beyond the scope of the SPDS, and these information requirements will not be discussed further. 3.3 ~ SPECIFIC. PLANT I" FOP.MATION REQUIRED BY THE BWROG EPGs 3.3.1 Entry Conditions ~ and all A tnorough review of the BWROG EPGs was conducted, o l ', conditions were identified fdr: (a') entry into the EPGs from a normal power c y operating state. or (b) transferring to, or concurrently implementing a .different part of 'the EPGs following initial EPG entry (see Table 3-1). Results of thiI. review bre summarized'in Table 3-2, including reference to specilic ' paragraphs in the EFGs which establish each entry condition. Also ~ j jncluded' in ; Table >3-2 is a listing of the specific plant information nee,1ed ~ to determine the existence. of each identified entry condition. l , The EPG entry conditions from a normal power operating state are 'i 'ra'ther straightforward, hovaver; s: ore complex decisions are often required ~~ l to determi' rte the need to f avoke contingencies that are a part of the EPGs. a: n.,- 26 a

503-8500000-76 3/1/84 -To ' illustrate this point, all contingency entry conditions were abstracted from Table 3-2 and reorganized-in Table 3-3 to concisely define the plant conditions requiring each contingency to be invoked, related plant informa-tion requirements, and reference to the specific paragraph in the EPGs which establish each entry condition. It should be obvious from Table 3-3 that Contingencies #2 and #6 both have a relatively large set of diverse entry conditions that could be difficult to ascertain from " conventional" control room instrumentation and operator aids. 3.3.2 Multiole-Parameter Graohic Plots As discussed in Section 2.3, the BWROG EPGs specify a variety of multiple-parameter limit curves and decision points that determine the need for important operator actions in response to abnormal or emergency plant conditions. Proximity of the plant to these limits is difficult to deter-mine from conventional control room instr: mentation, therefore, these limit curves and decision points are candidates for support by the SPDS. Use of the multiple-parameter limits in the EPGs is summarized in Table 3-4. Plant information required to generate graphic displays containing these multiple-parameter limits is summarized in Table 3-5. 3.3.3 Summary of EPG Plant Information Requirements t The specific plant variabler required by the BWROG EPGs are sum-marized in Table 3-6. This table also indicates how each var,iable is used in the generic EPGs. The actual information requirements to support the CNS plant-specific E0Ps being developed from the EPGs may be different than the listing in Table 3 6. 27 L

503-8500000-76 3/1/84 40 AanAL POWS A CPtAAT1048 I I f I I W"/\\'h%'*/\\'/* Y \\'7 Y PttuAAY SEC040AAf AA010ACfivsfY APWComiAOL C04TA.4ta647 C04fAleast47 AELIA&E IAC3 C04fAOL CONTROL C04 TROL . x: i C: i A Al I i i i i i l i I I I I g" MM TW YY V h NyNyhh0yhh N h N NyNc/h\\b yNc/hNy

.v- =I L..t.

E Aete., I,E _ .t.I,0Wt. _.,000,40 A ttT0 Aafl04 APW 0tPetSEU AllAft04 C00040 LtytL CONTA0L C00ue0 (Can iC1) (C2) IC3 At370 A ATIO N (C73 ggy (Can +M+M

+

+ + M MM ~ < I I M cot 01Nuf00W4 C040lft04 1801 Figure 3-1. Overview of the Organization of the BWROG E.mergency Procedure Guidelines. 28

503-8500000-76 3/1/84 Table 3-1. Sumary of Relationships Among Major Elements of the BWROG EPGs. To e,,,,,,,,,,,, y,, C.is Proc. C4 l 05 l C6 C7 4C/Q aC/L

C/P (2)

C1 C2 C3 seector Po=or Control (eC/0) (24) avector Pressure vessel (nPv) Py L.,el Control (RC/L) I (23) Control Guideline APV Pressure Control (RC/P) t I I i -(23.4) Sugeressten Pool Temperature g g g' Control (17/T) Orywell fasserature Control I X (OW/T) Centatement Tempeture g y Contml @ /T)6 Prime Containment Contre Guideline Peteery Containment I (2c,4) Pressure Control (PC/P) Sucoression Pool WateP r (23) Level Control (5P/L) nyerogen and Caygen I (2d) Concentration Control (PC/M) Secondary Containment remoerature Control (sC/t) I 8 I 1econdary Containment Secondary Cont 41'usent g g g g Contr,1 Guideline W4ter Level Control ($C/L) Secondary Containment g 3 g g Paetation Level Control ($C/2) Radioactive Release g Control Guideltne (RW) Contingency #!* I r x t I X Level Restoration (C1) Contingency et, Emergency g g APV Coeressur12ation (C2) Contingency 83. Steae Cooltag (C3) Contingency #d. Core Coolin g Contingency Without Level 8 storetton (gCS) Guidelines 0 Conttagency *S. Alternate Shutdeum Coeling (CS) (3) Contingency 86 X X I RPV Flooding (CS) Contingency v7. Level / g g ( g} Power Control (C71 (1) lased on *0 raft Essegency Procedure Guideltnes.' Aevtsten 3G. S'eR Mers Group. Movegeer IC.1983 %otes: (2) The B.ROG EPGs reference tre folloutng clant orecedures: (a) reactor scree crocrourt. (b) orocedure for cooldown to cold smutoown. (c) proceoure for contatenent venting. (d) otner system operating or secoling crocecures (3) Not soolicaole to Cooper Nuclear Statien $PC5. agolicaole only to !WR olants eitn a Mart !!! contatenent. 29

Table 3-2. Summary of Conditions Requiring Entry into the EPGs and/or Transitien Batwe:n Elements-of the EPGs. u. Ow e Information Required to Determine From lo Entry Condition Entstente of Entry Conditions Q Oo Moraal RPV Control RPW water level < low level scram setpoint (el2.5") RPV water level O I N Normal RPV Control RPV pressure > RPV high pressure scram setpoint RPV pressure cn Nrmal RPV Control Dr pell pressure > drywell high pressure scram Drywell pressure setpoint Normal RPV Control A condition occurs which requires PtSIV isolation: MSIV isolation demand (Group I) status - RPV water level low (-145.5") - h in steam line radiation high - Main steam line tunnel leak detection - high temperature - min steam line flow rate high - Main steam line pressure low - min condenser vacutes low Normal RPV Control A condition occurs which requires reactor scram Reactor power (APRM). scram demand status (listed below) and reactor power is above APRM downscale trip setpoint or cannot be determined: W - Drywell pressure high a - RPV pressure high - kPV level low (*12.5") - Reactor power level high (APRM, flow biased) - Nin steam line radiation high - Scram dise.harge volume water level high - MSIV isolation - Turbine stop valve closure when power >301 - Turbine control valve fast closure when power >301 - APRM inoperable - Reactor mo h switch in shutdown Normal Primary Containment Suppression pool temperature > most limiting Suppression pool temperature Control suppression pool temperature LCO hrmal Primary Containment Drywell temperature > drywell temperature 100 or Drywell temperature Control manimum normal operating temperature, whichever is higher Normal Primary Containment Containment temperature > containment temperature Containment temperature (Mark Ill containments only) Control 100 y w Normal Primary Containment Drywell pressure > drywell high pressure scram Drywell pressure g Control setpoint w Normal Primary Cantainment Suppression paol water level > maximum water leve? Suppression pool water level Control LCO

Table' 3-2. (Continued). m O u e from To Cntry Condition Information Required to Deterwine m Catstence of Entry Conditions t o 8 Normal Primary Containment Suppression pool water level < minimum water level Suppression pool water level Q l Control LCO e N Normal Primary Containment Primary containment hydrogen concentration > high Drywell hydrogen concentration, suppression chamber Control hydrogen alarm setpoint hydrogen contentration Normal Secondary Secondary containment dif ferential pressure at or Secondary containment differential pressure Containment Control alieve 0 inches of water Normal $econdary Any secondary cor.tainment area temperature > maalaus Secondary containment area temperatures Containment Control normal operating temperature Normal . Secondary Any secondary containment HV E cooler differential Secondary containment HVAC cooler differential j l Containment Control temperature > maximum normal operating delta T temperatures 1 Normal Secondary Any secondary containment HVAC exhaust radiation SeconJary containment HVAC enhaust radiation levels i Containment Control level > maileum normal operating radiation level Normal Secondary Any secondary containment area radiation level > Secondary containment area radiation levels Containment Control manimum normal operating radiation level Normal Secondary Any secondary containment floor drain sump water Secondary containment floor drain sump water level d Containment Control level > maalmum normal operating water level i Normal Secondsry Any. secondary containment area water level > Secondary containment area water levels Containment Control maalmum normal operating water level Normal Radioactivity Offsite radioactivity release rate > rate which Offsite radioactivity release rate from plant Release Centrol requires an Alert condition to be decleared release points j RC/L Contingency #1 RPV water level cannot be maintained above tar RPW water level l (RC/L-2) RC/L Contingency #5 Alternateshutdowncoolingrequired(RC/L-2) Normal shutdown cooling system status, control rod f j position [ HC/L Contingency #6 RPV water level carnot be deternt'ned (RC/L-1) RPW water level l RC/L Contingency #6 RPV flooding is required for other reasons (RC/L-1) RPW water level, RPV pressure, drywell temperature, suppression chamber pressure, suppression pool j water level, SRW position to RC/L Contingency H Baron injection is required (RC/L-1) Reactor power, suppression pool temperature D (conditional) m RC/l Procedure for Cool-When cooldown procedure is entered from RC/P { down to Cold Shutdown (RC/L-3) g i 1 em~-

Table 3-?, (Continued), on C) LJ frne Tc. Entry Condition nformation Required to Determine 0$ Esistence of Entry Conditions c3 C3 C3 RC/P Contingency #2 Emergency RPV depressuritation or RPV flooding SRV position RPV water level and level trend, RPV c$ required and less than a SRVs are open (x is the pressure.and pressure trend, suppression pool 8, number of SRVs dedicated to ADS) (RC/P) temperature, delta T heat capacity (calculated), en suppression chamber pressure, suppression pool water level, drywell temperature (local), drywell hydrogen and oxygen concentrations, serandary containment area temperature, area radiation level, area water level,.nd floor drain sump level, of fsite radioactivity release rate RPV coolant injection system status, contair. ment temperature (Mark III containments only), RPV isolation status RC/P Contingency #1 Steam cooling is required (RC/P-l) RPV water level and level trend, RPV pressure, RPV coolant injection system status RC/P Contingency #5 RPV cooldown is resuired but cannot be accomplished Normal shutdown cooling system status, control rod and all control rods are inserted beyond the maximum position subcritical banked withdrawal position (RC/P-4) RC/P Contingency #6 RPV floodin'g is required and at least x SRVs are RPV water level RPV pressure, drywell temperature La open (x is the number of SRVs dedicated to ADS) (local), suppression chant er pressure, suppression D4 (RC/P) pool water level. SRV position, containment temperature (Mark III containments only) RC/P Procedure for Cool-As a normal transition at the completion of RC/P (conditional) d down to Cold Shutdown (RC/P-5) RC/P Reactor Coolant As part of RC/P (RC/P-2) (conditional) Sampling Procedure RC/Q Reactor Stram As part of RC/Q (RC/Q-4.3) (conditional) Procedure SP/T RPV Control Suppression pool temperature cannot be maintained Suppression pool temperature, RPV pressure (concurrent) below the lleat Capacity Temperature Limit (SP/T-4) SP/T Contingency #2 Suppression pool temperature and RPV pressure cannot Suppression pool temperature, RPV pressure be maintained below the Heat Capacity Temperature Limit (SP/T-4) DW/T Contingency #2 Drywell temperature cannot be maintained below Drywell temperature Od (concurrent) maximum temperature for which ADS is quellfled, or 3 drywell design temperature, whichever is lower $0 { DW/ T-3) I DW/ T Contingency #6 Drywell temperature near the cold reference leg Drywell temperature (local) (concurrent) instrument vertical runs reaches the RPV saturation temperature (DW/T-2) r

Table 3-2. (Continued). Us ow Information Requir'ed to Determine h from to Entry Condition Entstence of [etry Conditions y o CN/I Contingency #2 Nrk III. containment ter:perature Cannot be. Not applicable to Cuoper Nuclear Station SPDS - o (concurrent) maintained below containment design temperature O (CN/I.3) k m CN/I Contingency #6 Nrk Ill containment temperature near the cold Not applicable to Cooper Nuclear Station SPD$ reference leg instrument vertical runs reaches the RPV saturation temperature (CN/i-4) PC/P Contingency #2 Suppression chamber pressure cannot be maintained Suppression chamber pressure, suppression pool below the Pressure Suppression Pressure tielt water level _ (PC/P-4) PC/P Contingergy #6 Suppression chamber pressu.re cannot be maintained Suppression chamber pressure, suppression pool below the Primary Containment Design Pressure water level (PC/P-5) PC/P Procedure for Suppression chamber pressure enceeds the Primary Suppression chanter pressure, suppression pool Containment Venting Containment Pressure Limit (PC/P 4 ) water level PC/P Containment Pressure As part of PC/P (PC/P-l) (conditional) Control, 58GT and W Drywell Purge System W Operating Procedures SP/L Contingency f2 Suppression pool water level cannot be maintained Suppression pool water level, suppression pool (concurrent) above the Heat Capacity level Limit (SP/L-2) temperature SP/L Contingency 82 Suppression pool water level and RPV pressure cannot Suppression pool water level, RPV pressure (concurrent) be maintained below the Suppression Pool load Limit (SP/L-3.1) SP/L Suppression Pool As part of the SP/L (SP/L-l) (conditional) 4 Water Samp1tng Procedure PC/H Contingency 82 Drywell hydrogen concentration and oxygen concentra-Drywell hydrogen concentration, drywell oxygen (concurrent) tion reach their lowest respective concentrations concentration which can support a deflagration (PC/H-2.4) PC/H Containment As part of PC/H (PC/H-1.1, PC/H-4.1) (conditional) Atmosphere Sampling Procedure w SC/T RPV Control A primary system is discharging into a secondary kPV water level, secondary containment area ( (concurrent) contaltunent area, but before any area temperature te6perature RPV isolation statas 03 reaches its maalmum safe operating temperature (SC/T 4) t

Table 3-2, (Continued). mow eao Information Required to Determine g from To Cntry Condition Caistence of Entry Conditions ooo SC/T Contingency #2 A primary system is discharging into a secondary RPV water level, sewondary containment area tempera-f containment area and an area temperature exceeds its ture RPV isolation status N l maximum safe operating temperature in more than one ch area (SC/T-5) SC/R RPV Control A primary system is discharging-into a secondary RPV water level, secondary containment area (concurrent) containment area, but before any area reaches its radiation level, RPV isolation status maximum safe operating radiation level (SC/R-2) $C/R Contingency #2 A primary system is discharging into a secondary RPV water level, secondary containment area containment area and an area radiation level enceads radiattun level, RPV isolation status its maximum said operating radiation level in more. than one area (SC/R-3) SC/L RPV Control A primary system is discharg?ng into a secondary itPV water level, secondary containment floor drain (concutrent) containment area, t>ut before any floor drain sump sump level, secondary contaireent area water level, l or area water level reaches its maximum safe RPV isolation status operating water level (SC/L-2) 4 SC/L Contingency #2 A primary system is discharging into a secondary RPV water level, secondary containment floor drain 4 w containment area and a floor drain sump or area sump level, secondary containment area water level. So water level exceeds its ac.imum safe operating water RPV isolation status level in more than one area (SC/L-3) RR Contingency #2 Offsite radioactivity release rate approaches or Offsite radioactivity release rate, RPV water exceeds the release rate which requires a General level RPV isolation status i Caergency to be declared and a primary system is discharging into an area outside the pristry and l secondary containments (RR-2) Contingency #1 RPV Control RPV water level trend increasing and RPV pressure RPV water level trend, RPV pressure high (Cl-4) Contingency #I RPV Control RPV water level trend increasing, RPV pressure RPV water level trend, RPV pressure and pressure i intermediate, and RPV pressure trend decreasing trend l (Cl-5)- Contingency J1 RPV Control RPV water level trend increasing, RPV pressure RPV water level trend, RPV pressure and pressure intermediate. HPCI and RCIC are not available and trend HPCI status, RCIC status RPV pressure trend is not increasing (CI-5) w Contingency #1 RPV Control RPV water level trend increasing, RPV pressure RPV water level and water level trend, RPV pressure N Intermediate and RPV water level restored to low ( l level scram setpoint (Cl-5) as b fontingency #1 RPV Control RPV water level trend increasing, and RPV pressure RPV water level trend, RPV pressure low (Cl-6)

Table 3-2. (Continued). u1 OW Information Required to Determine From To Entry Condition Entstence of Entry Conditions ogoo Contingency #1 Contingency #2 RPV water level trend increasing. RPV pressure RPV water level trend. RPV pressure anJ pressure o intermediate. RPV pressure trend increasing and trend. HPCI status. RCIC status ( HPCI and RCIC are not available,(Cl-5) cn Contingency #1 Contingency #2 Same as above followed by RPV pressure trend RPV water level trend. RPV pressure and pressure then to RPV Control decreasing (Cl-5) trend, llPCI status. RCIC status Contingency #1 Contingency #2 RPV water level trend increasing. 2PY pressure low RPV water level trend. Rev pressure and pressure and RPV pressure trend increasing (Cl-6) trend Contingency #1 Contingency #2. Same as atsove followed by RPV p. essure trend RPV water level trend. RPV pressure and pressure then to RPV Control decreasing (Cl-6) trend Contingency #1 Contingency #2 RPV water l'evel trend decreasing. RPV pressure high RPV water level and water level trend. RPV or intermediate, and RPV water level drops to TAF pressure (Cl-7) Contingency #1 Contingency #3 Same as above, but with no system, injection sub-RPV water level and water level trend. RPV pressure, system er alternate injection subsystem lined up injection and alternate injection system status with at least one pump running (CI-7) u ut Contingency #1 Contingency #2 RPV water level trend decreasing. RPV pressure low. RPV water level trend RPV pressure and pressure RPV pressure trend increasing (Cl-8) trend Contingency #1 Contingency #4 RPV water level trend decreasing. RPV pressure low. RPV water level and water level trend. RFV and RPV water level drops to TAF (Cl-8) pressure Contingency #1 Contlugency #6 RPV wa'ter level cannot be determined (Cl) RPV water level Contingency #1 ContinJency #6 RPV flooding is required for other reasons (Cl) RPV water level. RPV pressure, drywell temperature (local), suppression chamber pressure, suppression pool water level. SkV position, containment temper-ature ( N rk lli containments only) Contingency #1 Contingency #7 Boron injection is required (Cl) Reactor power suppression pool temperature Contingency #2 Contingency #6 RPV flooding required (C2-1.3) RPV water. level. RPV pressure, drywell temperature (local). suppression chant er pressure. suppression pool water level. SRV position, containment temperature (Mark lli containments only) Contingency #2 RC/P RPV depressurized using Contingency #2 (C2-2) (conditional) W w Contingency #3 Contingency #2 Any system, injection subsystem or alternate injection system and alternate injection system g injection subsystem becomes available (af ter status .p initially being unavailable) and is lined up with at least one pump running (C3-1)

Table 3-2. (Continued). On C3 LJ O rena To Entry Condition Information Required to Determine O$ Esistence of Entry Conditions (j. C3 Contingency #3 Contingency #2 Imergency RPV depressurtration required (C3-1) 1RV position. RPV water level and level trend. RPV C) C3 pressure and pressure trend, suppression pool 8, temperature, delta T heat capacity (calculated), en suppression chamber pressure suppressica pool water level, drywell temperature (local). drywell hydrogen and osygen concentrations secondary containment area temperature, area radiation level, area water level and floor drain sump level, of fsite radio-activity release rate. RPV (colant injection system status, containment temperature (Mark til contain-ment only). RPV isolation status Contingency #3 Contingency #2 RPV pressure drops below minimum single SRV steam HP# pressure cooling Pressure (C3-1) Contingency #4 RC/L RPV water level is restored to TAF (C4-3) RPV water level Contingency #5 Procedure for Cool-At conclusion of Contingency #5 (C5-1) (conditional) Jown to Cold Shutdown {j$ Contingency #6 Centingency #1 and RPV water level can again be determined, thus RPV RPV water level RC/P (concurrent) flooding is no longer requireJ (C6-2.1) Contingency #6 RC/L end RC/n Suppression chaseer pressure can again be maintained Suppression chamt;er pressure suppression pool water (concurrent) below the Primary Containment Design Pressure (C6-6) level Contingency #1 Coatingency #2 RPV water level cannot be. maintained above TAF (C1-2)RPV water level Contingency #1 Contingency #5 Alternate shutdown cooling is required (C1-3) Normal shutdown cooling system status, control rod position Contingency #1 Contingency #6 RPV water level cannot be determined (C1) RPV water level Contingency #1 Continger,cy #6 RPV flooding' is required for other reasons (C1) RPV water level. RPV pressure drywell temperature (local), suppression chamber pressure, suppression pool water level. SRV position, containment temperature (Mark III containment only) Contingency #1 Procedure for Cool-When cooldown procedure is entered from RC/P (C1-4) (conditional) .fown to Cold Shutdown LJ bd 00 Pm

Tabl e 3-3. Sununary of EPG Contingency Entry Conditions. E w Composite of Information Required to IPG Contingency Contingency Intry Conditions Determine Inistence of Intry Condition oo Contingency fl. RPV water level cannot be maintained above TAT RFV water level ? Level Restoration (RC/L-2) Ncn Contingency #2, Suppression pool temperature and RPV pressure cannot RPV water level Emergency kPV be restored and maintained below the Heat Capacity pepressurization Temperature Limit (SP/T-4) RPV water level trend Drywell temperature cannot be meentained below maxi-RPV pressure num temperature for which ADS is qualified, or dry-welldesigntemperature,whichevorislower(DW/i-3) RPV pressure trend Containment temperature cannot be maintained below Suppression pool temperature containment design temperature (Nrk Ill contain-ments only. CN/i-3) Delta T heat capacity (calculated) Suppression chamber pressure cannot be maintained Drywell temperature (local) below the Pressure Suppression Pressure Limit (PC/P-4) Suppression chamber pressure w Suppression pool water level cannot be maintained Suppression pool water level N above the Heat Capacity Level Limit (SP/L-2) Drywell hydrogen concentration Suppression pool water level and RPV pressure Cannot be restored and maintained below the DryWell oxygen CcinCentration Suppression Pool Load Limit (SP/L-3.2) Secondary containment area temperature Drywell nydrogen concentration and oxygen concen-tration reach their lowest respective concen-Secondary containment area radiation level trations which can support a deflagration (PC/H-2.4) Secondary containment floor drain sump level A primary system is discharging into a secondary Secondary containment area water level containment area and an area temperature exceeds its maximum safe operating temperature in unre Offsite radioactivity release rate than one area (SC/T-5) SRV position A primary system is discharging into a secondary containment area and an area radiation level RPV coolant injection system status e u eeds its maximum safe operating radiation level in more than one area (SC/R-3) Containacnt temperature (krk Ill containments W only) ) N A primary system is discharging into a secondary containment area and a floor drain sump or area RPV isolation status C,D water level esceeds its maximum safe operating level in more than one area (SC/L-3)

moc gooo?wm g%wD4 a a n oo ti t died no uC Ny r t nn I fo _ Ne c n ne it s la f _e en i mre D-te ) d eun i m se t d - t l s l ) l dd dda e ih nso ee e 7 e en ene b t n ye l een l vr l v l v pa ) pat n g ot ns e s rh err eep ee l ee p 1 p ,s it n d o ii C eaat vt pe vl vl rC vl o,es - o h m-e r e V e e( e V r)ti3 r)V ( c si e l rVa l rP l rt l rP dea C der cn oX nd( lS ea o( aerd) eP eR enF eR l orpi2 rtRC rt rtiA rt lbrn lb jc l ri s - ea I ead ea T ead eaeag eae n f n t wn vlt n vll il e) puatR t w .C t wn t wr pq uR a eR a a a oo a a eil ei eig e V pP 3 s aedo( wVt wV wV t wV l aal n l an nv Po/ n rn Pad P, Ph P, v b n vi oe R C o e aas VRt n VRw VRgs VRw Varau Vas rl eR 3 it e R .l R h o R .l) R s i r P o rr( P o P ol th et P .d a P o P.i p Rsm br oa) e i acdrn R mu (e S mmap ri e a : () vI () () r () 8 ( ( hr u F r C f e F ed F e-eevm em t nsD d n l n ewann dAeP dA r dAr dArl dtt au dti da OVA o s l ai eTt t eT u) eI ul eT uC ess p esn ew iR b C aec a t n i t s6 t se t s( t yys t yi t ) tSo et eot neid nes - nesv nes nssee nsm nV3 a t a T y l adt n ev n evel evee eveg e bmn e eP - zX r er no moea morC morl morn mnuoo mnw mR7 i d t r eic ebr e b p( ebp e b pi eosc eoo) e C rne n eb l au, l a l a r l a s li et li 1 ld( uat E ys gy p sg p Vg p V e p Va pt nbs ptle-pn sha c vl gd nrs imdPn mdPt imdPe mco a mcb3 maF stic t aonr mdsn eime ie C i A e y ieti a i eei eRi i eR a err i jt el js( ,T rsd n.s n w n.c g c nhc tV r5 t nr t nP t n isnct snp s) pse n i a si. si n eyrn sipa si e trcao i a e) i age i a gV i a gi i est ii oe iL e eed jya rr /v dl n aeece 1 nPc-1 nic 1 niR 1 nid 3 ons 3 odu 1Co s i oh gs s

iRnl
i sn
i s
i sn
nibh
n s
Rb VdV t

it ri u.i C aai aad aae ut es a P nR ( yme ynen ymer yh .si yh r e yn ras i n d edd yn o asn n c gd c rd c ra c rt ct m w ct ur cid C rdE sa nenne necn nec nec nien nisp n e y .f e i ebiel ebne eb e, ebee ewt op ews edn cdo t camr nt at a nti r g de g dr g si u g eg gei ne eel y g srb g t nt u nF yt nI rn nra err i xrea ioe l it nne int t it nns tT ee tT l t un rub od oda ods iAscd i A pi ii t gi e seet m t nrei t nni f nsi nncra nner nned nnee n yjn n V o nql eqn f reye oanuv oaru oare oarr oonni ooP o oea meu OoGsp Ccisa Cct s CCt m Cct p Ct ail CtR c Crn Ern 6 y _ c. ne 2 g n _ i _ ) t yd n ce o nu C en igi G t P t n no I

o. (C t

'g d

u, Table 3-3. (Contiriued). O" /n8 Oo Composite of Information Required to O (PG Contingency Contingency Intry Conditions Dete mine [mistence of intry Conditior; N Contingency f3, - Contingency #1 is implemented (RPV level cannot be RPV water

level Steam Cooling maintained above TAF), RPV level trend decreasing, RPV pressure high or intermediate, RPV water level RPV water level trend drops &a TAF, and no system, injection subsystem or alternate injection subsystem is lined up with RPV pressure at least one PUNp running (Cl-II RPV coolant injection system status Contingency #4,

- Contingency #1 is implemented (RPV water l'evel can-RPV water level Core Couting not be maintained above TAF), RPV water level trend Without Level decreasing, RPV pressure low, and RPV water level RPV water level trend ites toration drops to IAF (Cl-8) RPV pressure Contingency #5, - RPV cooldown is required but cannot be accomplished Control rod position w Alternate Shut-and all control rods are inserted beyond the maximian W down Cooling subcritical banked withdrawal position (RC/P 4) Normal shutdown cooling system status Contingency #6, - RPV water level cannot be determined (RC/L-1) RPV water level RPV Flooding - Drywell temperature (or Mark Ill containment terperature) RPV pressure near the cold leg instrument vertical aans reaches HPV saturation temperature (DW/i-2, CN/i-4) Drywell temperature (local) - Suppression chauber pressure cannot be shst'ntained Suppression chamber pressure below the Primary Containment Design Pressure (PC/P-5) Suppression pool water level - kPV flooding is required (for one of the reasons SRV position i listed above), and at least X SRVs are open (X is the nuaber of $RVs dedicated to ADS)(RC/P) Containment temperature (Mark III containments only) Contingency #1, - Born injection is required (RC/L-l) Reactor power (APRMs level / Power y Cnntrol (i.e., reactor cannot be shutdown before suppression Suppression pool temperat6re pool temperature reaches the borow injection initia-tion tenverature)(RC/Q-4) D

503-8500000-76 3/1/84 i w w 8 l e a M l M y i I f e w f f y I f p w t es C 8 l o m W w 6 I I w* f 3 e a B l t ~ C e a I l . g" i I L w a 8 8 u I C-e w d t D i t w e l D w M to 9 9 L U = g .T. D. = = t = = = = C a I w O k a E d 3 re d a'. n w I = = b e M = A 3 A g s T U w w = x x e Q w 4 t o 7" w l l %s.m

w w

s ee =f M W uw l c 1 W 3 l E .=s m g E u W t 4 O a m u l c e t m I D I f U

m i = t i Y e u t I M i W E I u w I g D e E h4 e. W w I l o e ~ 8to 5 t o - T

es we j

a e - $e" w-o m== = 9 E* E E h - em 6 I *** ee .g o E2W = a W= W w a aw== o w 2

I*

a.e)3,s . g.o-b. wt g o ~5 - 6 Se o-s- e -w C. .a g,. 3 g g g g g g , g w, we a-a w t.e e o 6 om em e a w. a =.- e o =,

e. -

4w-m m. ew o-e o-egm-w. -w . ec .e W-we El o s-ez g-

== = -- -. c w ee am -* -a = w w me -e

e e&e
w 2 6)e W w w 2 C ow -eee es e a

eme = we .e l w o s-ee c-ee -2 ww e96 = =

w o - t x =>

= c. om= c=2 o ew== twoe

w g a-e-

oe o-o s.

== =., e ** h # w h.

e. es w e.e me

-=- a e ae s. e -em w e w. a. o -C a em && 48 l a 3 t 9 6 6 w h hw WD, es e.e I,a= e -o a= o n ew

==w = = = = we um = 6 - tsme wW w +8 C.

en C4 em &c e en

-e ar en 6 u ( e== wC& w& E E z.- a w s oa i e ee ae e se- -6 e a, s-

s.,. t =s.

a., s-cc o

sms - a

r. c.

sa - u e. M.. u - g w a - . = = -* -e e

== m-c =v

== - ww ww w e., ww m. w = = = = = > e-3.* I.m,A w eW e== Z e. en M b E e8 E, C"o G. b& C. 4 A& to V = e O4 3 eCC Z =8 40 t .a n.._,

cn - O . Table 3-4. (Continued). w g, tn O O O Of Nor Emergency Precedure Guidelines Contingencies male S. Parameter timit t rGst)') Defined in SWROG RC/Q RC/L RC/P $P/T DW/T Cll/T PC/P SP/l PC/M SC/T SC/L SC/R R4 Cl C2 C3 C4 C5 C6 CF Curve (2) Minimum reopening pressure g Alternate shutdown cooling pressure limits (3) X a Minimumgflooding pressure g Minimum alternate W flooding pressure X X g RPVpressure/legel status matria 48 X i Mutes: (1) Based on "Draf t Energency Procedure Guidelines." Revision 3G, BWR Onners Group. Nonder 10. 1983. (2) Not applicable to Cooper Nuclear Station SPDS. Applicable only to 8WR plants with a N rk Ill contaltunent. (3) These limit curves are not illustrated in the BWROG IPGs. but the ag.plicable limits are defined. (4) the decision matria for Step Cl-J of the EPGs provides a sumanary of comples decision-making rules. w N e N 00 Em f

_ _ _ ~... _,.. _ _ m. 3I i l2 617 R717 lie;' e{ E } ;E E j % :E O f EE! ?? 7 2r 33 3]i 2-

-: E t; ? ?

F: la I % 1 % E E E,. O '

T*a

. C2 07 ~*;* " ; ; ; -" ~

r

~ :: 2

~*
g

.- F

,; B _,.,4 '; E 7 ri: 2 T -s

-i*

3-72 7

7 ' r ". * *. F4 F 2 8 :

8 T.F 7 ))

1,; ?.7 E* 1 ",l 8 r r _ f.

i o g cr r-t=

t-3 7,3 -I _ r _.

== = r, a :. 2*2 Es w 1 32 ~1~ E n -s s _n =n i E,T = T S tri i

l;

,3 ? Suppresston pool .l y 4 water temperature W J T Delta T heat 3 capacity (calculated) 1 Suppression pool s = = = = -9 I mater level O 7 Suppression champer B = = = 2 pressure e Sappression enameer o 3 t hycrogen concentration 3s' z Sucpression chamber tb oxygen concentration j 1 HPCI pump flow = 9 i rate rn 3 cL i ROIC pump flow g rate ( C I LPOS pump flow 2 m = a tv l rate o ,o en LPCI (RHR) pump R c s, 8 flow rate 2 r* O ={ Drywellpressure [ l e Drywell temp near cold ref leg instrument 1, s i o i Deywell hydrogen l concentration tu 3 (D Drywell oxygen p y concentration 6 rp ~$ s t Reactor power K 3 1 (APRM) e 5 Time f*os shutcown to V 3 onset of Core uncovery { n s= = = ApY pressure g RPY pressure trend "Ue c8 M RPY level m 1 RPV level trend I 1 e 1-e SRv Position i I i i 18/1/E 94-00000SB-EOS

EI =-e = = -< 4- - ,s; eer 2 e2 ,-2*-Feee = - g " S i, ! i 2 3 '? ~ =i s e -Ej t}90:

d ji O': .

M It 8 E! F2 E!'Cf 7. 2 T T-51 8 t

m8 3.0;-*-

? f,2. ;*='T,;=' - 7, e.=.. -,o er e

E 1 Y.O 3,

E r i *E t B 1 0 *1 j

3 2

F.

" 3 ~:- 31 6 2 j. ." t' - 3' t. = i :l.- -2' Ork 4 5 2

Ei e.
T**Ia 8

T' =.r 2

I ".

F 34 J* si 4 o, w = m. -. g ;. 8 = g i Y.,* O O O S a crcssion pool mater temperature R e, v v =. n ) 9 .4 g Leita. heat 4. c. a a capacity (calculates) 2A - * * -e ir 22 7 9f Su:oression pool 7. mater level '~* a 2l%e 8 Suppression enanter I~;~

'r,%

i : e pressure -,F 2T Suceression enameer = o =- E*O nyorogen concentration ,2 u. o l -4 8 Superession cheeser

OJB O. T T2 t

osygen concer.tration C" s w -o -. c (D . %. 2 MPCI pump flow - o w .m 'O cJ EEP y y i 3;l0 pump flow y (J) 2 pa g E f-I ratt 1

PT; I,e LPC5 pumc flos 1

.-4 8 rate n. 2 2 E t. m* O o -= 5- = 7 w -2* 2

LPCI (RHR) pump p

n 3*2* i flow rate 2

5

- e g# t c f* Dry =41Ipressure t5 i g t a y 2[ O e - e. e i E Drywell tema near col 6 ~ 'E p i ref leg instrument 1, a.. .%EB72 1 Drywell hydrogen j C** 1 concentration b a a 35L EfR 2

I

i,E Drywell oxygen p

I. concentration a o a e E*Ia F E 8 Reactor power a t e e c = 1 f (Aper} . e e c. e e e c Time from srutcown to u R c, j - * * - onset of core uncevery 0 i a~ " i RPV pressure E * ** i !3 2' RPV pressure trend E I m g I ppV level e 1 l lR8V level trent g SRY Position f

8/l/E 94-0000098-E09

503-8300000-76 3/1/84 Table 3-6. Summary of EPG Plant Informaticn Requirements. Cata Use fg EdROG EPGstl) I t ~* UE !.2 0

=

t' 1 0 T3 Plant vartseles Associated with

g b ri BWRCG EPGs(1)

E*

q 2

d om do RPV Water Level I X X RPV Water Level Trend X I RPV Pressure X X X RPV Pressure Trend X X MSIV ! solation Comand (Group 1) Status g Scram Comand Status X Reactor Power (APeM) X X Succression Pool Temperature X X X Delta T Heat Capacity (Calculated) X X Sucoression Pool Water Level X X X Suooression Chammer Pressure X X Suppression Chammer Hydrogen Ccncentration y Suppression Chamoer Oxygen Concentration g Crywell Pressure X X Crywell Temperaturt (average) I Crywell Temperat:.re (local) X X Containment Teacerature (average) (2) Containment Temperature (local) (2) Crywell Hydrogen Concentration X X X Crywell Oxygen Concentration I X da

503-8500000-76 3/1/84 Tabl e 3-6. (Continued) Cata Use 19 BWROGEPGstl) .I..t Uk !3 .a D: 28 t' 3: I" T3 Plant variables Associated with

D ri SWROG EPGs(1) 23 31 ni e ww uw Secondary Contairment Of fferential y

Pressure Sacendary Containment Area X X Temperature Secondary Containment HVAC Cooler g Differential Temperature Secondary Containment HVAC Exhaust y Radiation Level Secondary Containment Area X X Radiation Level Secondary Containment Floor Orain X X Sump Water Level Secondary Containment Area Water y y Level Offsite Radioactivity Release g g Rate HFCI Pump Flow Rate X RCIC Pump - Flew Rate X LPCS Pump Flow Rate X LPCI (RHR) Pump Flow Rate X SRV Position X X Control Rod Position X RPV Coolant Injection System y Status l Normal Shutdown Cooling System X Status Time from Shutdown to Onset g of Core Uncovery (1) Sased on *0 raft E.wrgency Pmcedure kicelines." tevision 23. !WR Owner's Notes: Group. Noveveer 10. 1983. (2) Not applicaele to Coocer Nuclear Statien. 1.colicaole only to SWR atants nita a v. ark ::t containment. l 45 ._ =

503-8500000-76 3/1/84 4. PLANT INFORMATION TO SUPPORT DETERMINATION OF SAFETY FUNCTION STATUS As described in Section 1.1, the status of the following five tafety functions should be displayed by the SPDS: e Reactivity control e Reactor core cooling and heat removal from the primary system o Reactor coolant system integrity e Containment conditions e Radioactivity control Regulatory Guide 1.97 (Ref. 3) lists plant variables that can be related to these safety functions; however there is not a one-to-one correspondence between the Regulatory Guide 1.97 variables and the NUREG-0737, Supp.1 (Ref.1) safety functions. The approximate relationship.between these variables and safety functions is shown in Table 4-1. Based on Table 4-1, Table 4-21ists the Regulatory Guide 1.97 ariables that are related to the NUREG-0737. Supp. I safety functions. It v should be noted that several of the listed variables actually represent information derived from analysis of manual samples rather than from on-line instrumentation systems. Information derived from analysis of manual samples is very static data in comparison to information derived from on-line instrumentation systems. This type of " static" data is incompatible with the real-time, on-line, interactive nature of the SPDS, and can adequately be presented on other media (i.e., log sheets). l l 1 46

Lv i 2

o n m :o

,* m m

o Tm2 1

,o D o tis: to s s to ro o@ C j + m= mam w c. ,m

s. - m fo fe,n

,n ,n ,*,ce m a M o D "r:0 o -a.

O owa q

Es Eo s n s s -ms o g E R E. o !"n ? EPO E. n i m s w to,o , z w s to c w .~ eo er n m-a to n .,m. w n se me o -m

c. w s m 3,+e :n-

,, s, w<n a mec o x== o= n s m=o w to .a w o ,e, oso e to - e ro o ewe e m s,= to n a s. E,- j o m.. r --so 4 c s o- ,+ > i os.e = = - - - e o o ua sm. . =s - cr o n s - s e er co o m-- to w m to ea 4 ,o -=mm%u-a _,. 1 mo - c.r. m o-ox s

c. o -s a s

3 n 3 e.G =

Reactivity 5 L n.,o.ce ,w x = c+ s-- 2 Control e s,+ c. s c :o ,+ o . ro - to a s n y o w sm E,E ,wnm = Core a -m.= -m -e. E O F 2 E 'E Cooling ? E =- = c aC 4 C, n

%
9 =. M% 2. % E Maintaining Reactor

3 ,d 4 A o"** ro os n aC ,Eoo*Q Coolant System i x 3 . ~ _,. E E * Integrity h y" - m -- E m c -c assows-w m,to e s a e m. Maintaining < ro m. to o,,+ Containment eae=e x x

s. m e m e e n to w a n to,+ s.-

Integrity a r3 . Q,# E,o.$ g E3 =- = to m to s w c to 2 Fuel Cladding w to - cu x m m to o,er <= w m c=, m3 e= e~ n,.a v o = llc m = m p m e e m " 3 3 3 S w co Reactor Coolant o. 6

- EoD<

PressJre Boundary f y-p 7"*% 2. w n e,, n, e =. a er ~ E* 2 E Containment E E x C *E13 3 E 4 Em, 1 = er m Containment - smeo a %7*1 O Radia tion E c,. o a $,+ e ^ a, < = cog

moo, Area s

o -., - s "' E S E O Radiation =w Gi-3 2 w E

E ", %

Airborne Radioactive 5 G 23%o Materials Released x E, 3 2 " From Plant m a c. ~ e+ wc-o a n E

SS-Environs Radiation 1

's .,i,+E. O and Radioactivity E,; s to.o - n 8 e m. s, to, - > m aL =, + o, c* x -* M Q. ro .N. Meteorology to o to - C-x m n, w e w

a. n.,

n,. n o mc 'E t' w c. E Accident Sampling o .o., ?o E O Capability pg/t/t 94-00000SS-EOS

503-8500000-76 3/1/84 Table 4-2. Summary of Plant Information Needed to Determine Status of Safety functions. H;AEG-0737. Swoo. I Safety Functiors(3) t 39 18= n g-r y: p. 22 w 2 x2 Reg L ide 1.37 Yariables Type w dw aw % e a av Neutron Flux (AP*y, $RM) 5 I Control Rod Pos1tfon 9 X RC3 Soluole Soron Concentration g 3 (grao sample) RPV Level 8 X RPV Pressure S.C X a Crywell Pressure B,C X Crywell Sump Level S.C X Primary Centainment Pressure B.C X Primary Containment Isolatien 8 Yalve Position (Excluding Check I Yalves) Radf oactivity Concentration or Radiation Level in Circulating C X Primary Coolant Analysis of Primary Coolant c g (gamma spectrum) Primary Containment Area y Radiation $uopression Pool Water Level C I Containment Hydrogen Concentration C X Crywell Mydrogen Concentration C X Containment Oxygen Concentration C X Nywell Crygen Concentratton C t Contain ent Effluent Radica:tivity C X (noole gas) Noble Gas, particulate and Malogen g e a tivity at plant Vents c Vent Flen Rate Notes : (1) USNRC Regulatory Guide 1.97, Revision 2. " Instrumentation for Light Water-Cooled Nuclear Pcwer plants to Assess Plant and Environs Conditions During ano Followirg an Accident.* Cecemeer 1980. as modified by NUPEG-0737, Suco.1. (2) As deff ned in Peg. Guide 1.97. Type 9 variables rovide infor ation to indicate whetner specific safety functions are teing cerformed. Type C variables provide information to indicate the potential for being breached, or the actual breach of the barriers to fission produc* release. 7yce E variables crovide infor a-tion for use in determining the magnitude of tre release of racicactive materials and continually assessing such releases. (3) NUREG-0737, Suco.1 (Generic Letter 32 33) "Pecaire-ents for E ergency Aesconse Capability." U.S. Nuclear Pegulatory Comission, race-ter 17. '952. 48

503-8500000-76 3/1/34 5. DISPLAY HIERARCHY AND PLANT INFORMATION CONTENT OF THE BWR OWNER'S GROUP GRAPHIC DISPLAY SYSTEM (GDS) 1 The BWROG GDS is conceptually similar to the Cooper Nuclear Station SPDS. The primary function of the GDS is to aid the control room operators in detecting abnormal operating conditions, assessing the safety status of the plant, executing corrective actions and monitoring plant response (Ref. 4). The information presented to the control rocm operators via the GDS is structured into a three-level hierarchy of color graphic displays as shown in Figure 5-1. The Level 1 display is a plant overview; the Level 2 displays present information related to individual safety functions, and the Level 3 displays present specific limit curves specified in the BWROG Emergency Procedure Guidelines (EPGs) and other supplementary displays recommended by the BWROG. A summary of the GDS displays and their associated plant data content is presented in Table 5-1. A two-phase evaluation of the BWROG GDS was sponsored by the Department of Energy througn Sandia National Laboratories. The first phase of this evaluation was a dynamic screening by BWR plant personnel of prototype displays driven by " canned" simulator data (Ref. 5). This screening process, and a separate, limited human factors review identified improvements which were incorporated into the displays used in the second phase of this program. The second phase was an evaluation of the GDS displays ia a BWR control room simulator (Ref. 4). Findings of the simulator evaluation of the GDS, relevant to SPDS parameter selection and display content are the following-I e Much display content is pl ant-s peci fic, nece s si ta ti ng customization of GDS displays for individual plant applications. e The GDS should display emergency operating procedure entry l conditions. e There should be a clear indication of GDS malfunction. 49

503-8500000-76 3/1/84 It should be noted that the GDS Level 3 displays were based on an early version of the BWROG EPGs. The Cooper Nuclear Station SPDS will adopt a display hierarchy similar to that of the GDS, but will have a more exten-sive set of Level 3 displays that are tailored to provide information speci-fically required by the current version of the BWROG EPGs (Ref. 2), and later by the plant-specific, symptom-oriented emergency operating procedures which will be derived from the BWROG EPGs. l l l l t 50 -=

II L3 ,['(,y DVERVIEW mo Y o8 oo BAR BAR SAR GAR SAR m d b d b d b d L d b LEVEL 2 ORE COOLANTSYS CONTAMNT RADIDACTIVE REACTIVITY BtSPLAYS COOLING INTEGRITY 84TEGRITY RE LE ASE 1 r 1 r < r 5 i r un TREND TREND TRING TREND TREND RPV LEVEL RPV LEVEL HEA T CAPACITY SUPP. POOL IW100 VS. (FUEL ZONE) TEM. LMIT LO AO LMIT PRESSURE TRENO LEVEL 3 DISPLAYS RPV SATUR ATION HE AT CAPACITY CONTAINMENT (fuel ZO El VS. PRESSURE I SPRAY LIMITS t R CD j Figure 5-1. liierarchy of Displays in the BWROG Graphic Display System. Se l i l

503-8500000-76 3/1/84 J 16m61 9901 toog

cons

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503-8500000-76 3/1/84 6.

SUMMARY

OF PLANT INFORMATION NEEDS In Sections 3 and 4, information needs were defined for supporting the implementation of the BWROG Emergency Procedure Guidelines and determining safety function status. The information needs of the EPGs are user-oriented in that this information allows control room operators to determine the need for specific operator actions in response to actual plant conditions. In contrast, knowledge of safety function status is more of an abstract engineering concern, and is not as closely related to specific operator actions as the EPG entry conditions described in Section 3. To aid in integrating the symptom-oriented emergency operating procedures with the SPOS, it is important to define the relationships that exist between the EPG entry conditions and safety functions. These relationships are summarized in Table 6-1, which was developed from information in Section 3 and 4. It should be very clear from Table 6-1 that the EPGs and the safety functions will define different bases for aggregating plant data in SPOS displays. For example, plant data related to the reactivity control safety function encompasses only a portion of the plant data related to the RPY control entry condition. It is reasonable, however, to relate the reactivity control safety function with the reactor power control (RC/Q) sub-element of the RPV control EPG. Relatively high reactor power following l a scram demand is a type of condition that would indicate a problem uniquely ~ associated with the reactivity control safety function. As required by the EPGs, the existence of this (or any other) RPV centrol entry condition will concurrently invoke reactor power control (RC/Q), RPV water level control (RC/L), and RPV pressure control (RC/P). This example serves to emphasize l th&t there is not a 1-to-1 correspondence between EPG entry conditions and safety functions. This lack of correspondence will require tradeoffs when j defining the SPDS display characteristics and data aggregation so that the displays can support the operator in both E0P implementation and definition of safety function status. [ A comparison of the plant variables associated with the BWROG EPGs and the safety functions is presented in Tables 6-2 and 6-3. Common data l 53

503-85.00000-76 3/1/84 i 5 \\ requirements and data requirements unique to the EPGs can be identified in Table 6-2. Data requirements unique. to the safety functions are listed in Table 6-3. For completeness, Tab'le 6-4 lists plant data used in the BWROG Graphic Displ'ay System, but not clearly related to the current version of the EPGs. e 1 I l l / 4 ,g.,.. 7..

503-8500000-76 3/1/84 Table 6-1. Relationship Between NUREG-0737, Supplement 1, Safety Functions and BWROG Emergency Procedure Guidelines (EPGs). III Generic Symptomatic EPGs(2) Major Sub-elements of EPGsI) Safety Functions Reactivity Control Reactor Power Control (RC/Q) Reactor Core Cooling and Heat Removal From the Reactor Pressure Yessel (RPV) RPV Level Control (RC/L) Primary System Control Guideline Reactor Coolant System RPV Pressure Contro* (RC/P) Integrity Suppression Pool Temperature Control (SP/T) Drywell Temperature Control (CW/T) ContainmentTegrature Control (CN/T) Primary Containment Control Guideline Primary Containment Pressure Control (PC/P) SuppressionPookWater Containment Conditions Level Control (SP/L) Hydrogen and Oxygen , Concentration Control (PC/H) Secondary Containment Temperature Control (SC/T) Secondary Containment Secondary Containment Control Guideline Water Level Control (SC/L) Secondary Containment Radiation Level Control (SC/R) Radioactive Release Radioactivity Control Control Guideline (RR) Notes: (1) Safety functions listed in Section 4.1 of NUREG-0737, Supplement 1 (Generic letter 82 33), " Requirements for Emergency Response Capability," U.S. Nuclear Regulatory Comission, December 17, 1982. (2) Based on " Draft Emergency Procedure Guidelines," Revision 3G, BWR Owners Group November 10, 1983. (3) Not applicable to Cooper Nuclear Station. Applicable only to BWR plants witn a Mark I!! containment. 55

503-8500000-76 3/1/84 Table 6-2. Comparison of EPG and Safety Function Data Requirements. NUREG-0737. Suco. I Sa fety Cata Use 19 Functions and Relateo se BWPCG EPGstl) Guide 1.97(ariables(2.3 ) t I G u 28 2 x x x: $~ F-x2 ei er b a EE 3 ?O $2 5% $5 Y $5 57 Plant Varigbles Associated with

E Ot Oi

3

.h 2E St BWROG EPGstl) 2*

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~Z 8 % *w 0 88 78 wd dw do av um a vu av RPV Water Level 1 X X X RPV Water Level Trend X X RPV Pressure X X X X X RPV' Pressure Trend X X MS!V Isolation Cemand (Group 1) Status y Scram Cemand Status X Reactor Power (APRM) X X (5) Suppression Pool Tamperature X X X Celta T Heat Capacity (Calculated) X X Suppression Pool Water Level X X X X Suppression Chamber Pressure X X X Suppression Chamber Hydrogen Concentration y y i Suppression Chamber Oxygen Concentration x x Crywell Pressure X X X X Crywell Temperature (average) X Drywell Temperature (local) X X Containnent Temperature (average) (g) Containment Temperature (local) (4) Drywell Hydrogen Concentration X X X X Crywell Oxygen Concentration X X X 56

503-8500000-76 3/1/84 Table 6-2. (Continued). NUSEG-0737, Suco.1 Sa fety Cata Use in Functions and Related Reg BWROG EPGstl) Guide 1.97 variables (2.Jj t ? 23 G if u 3 2. x m: 5" Ta =E M7 CE b 2 2 3_ &a

2..

_". ag 3 _3 O.. Plant Varigbles Associated with O& Ci O3 3;

3 BWROG EPGsL1) 2*

g. ~ g a' m, O

g i*

do C -a dw .a ww .w Secondary Containment Differential Pressure y Secendary Containment Area X X Temperature Secondary Containment HVAC Cooler x Differential Temperature Secondary Containment HVAC Exhaust g g g Radiation Level Secondary Containment Aree X X Radiation Level 7econdary Contairment Floor Crain X X Sump Water Level Secondary Containment Area Water x x Level I Offsite Radioactivity Release Rate x x g y HPCI Pump Flow Rate X e RCIC Pump Flow Rate X LPCS Pump Flow Rate X LPCI (RHR) Pump Flow Rate X SRV Position X X Control Rod Position X X RPV Coolant Injection Syster' x Status Normal Shutdown Cooling System l Status 7 i Time from Shutdown to Onset x of Core Uncovery Notes: (1) Based on " Craft E.-ergency Procedure Galdelines," Revision 3G, BWR Owner's Group, November 10, 1983. (2) Safety functions listed in Section 4.1 of NUREG 0737, Sucolement I (Generic Latter 82 33), " Rec::f rements for Emergency Response Capability," U.S. Nuclear Regulatory Comission, Cecember 17, 1982. l (3) Based on Reg. Guice 1.97, Revision 2. Decereer 1980, as redified by Section 6 of NUREG-0737, Sucolerent 1. (4) Not app 1tcable to Cooper Nuclear Station. Aeolicable only to B'.iR l plants witn a.' dart III containrent. -(5) /.PRM and SRM neutron flux. 57 o t

503-8500000-76 3/1/84 Table 6-3. Reg. Guide 1.97 Variables Related to NUREG-0737, by BWROG EPGstl.2,y) Functions but not Referenced Supplement 1 Safeta, NUREG-0737, Supp. 1 Safety Functions d' a-I: Q. O h r. d 3 ?3 E, 5E D 8= 3 !!3 ;- 2 "5 5 %d re %4 E" 22 %4 v, Plant Variables Ed 6& M dd EO Drywell sump level X Primary containment X isolation valve position Primary containment area X radiation level RCS soluble boron concentration (grab sample) Radioactivity concentration or radiation level in X circulating primary coolant l Analysis of primary coolant X l (gamma spectrum) 1 Notes: (1) Based on " Draft Emergency Procedure Guide-l lines," Revision 3G, BWR Owners Group, November 10, 1983. (2) Safety functions listed in Section 4.1 of NUREG-0737, Supplement 1 (Generic Letter 82-33), " Requirements for Emergency Response Capability," U.S. Nuclear Regu-latory Comission, December 17, 1982. l (3) Based on Reg. Guide 1.97, Revision 2, i December 1080, as modified by Section 6 of L NUREG-0737, Supplement 1. l 58

503-8500000-76 3/1/84 Table 6-4. BWROG Graohic Display System Variables Not Referenced in BWROG EPGs(l 2)- Use In GDS E E E J2 22 .2 C3 C1 C3 ~ n 5 13 13 Plant Variable 3 3 3 IRM power X IRM position X IRM range X SRM count rate X SRM position X Core flow X ~ .Drywell floor sump X pump status Containment isolation group signals sent X i MSIV position 'X Containment activity gas X iodine X - particulate X Notes: (1) Based on " Draft Emergency Procedure Guidelines," Revision 3G, BWR Owner's Group, November 10, 1983 (2) Based on ALO-1019, " Simulator Evalua-tion of Boiling Water Reactor Owner's Group (BWROG) Graphic Display System (GDS)," Sandia National Laboratories, May 1983 59

503-8500000-76 3/1/84 7. PLANT VARIABLES THAT ARE EXPECTED TO BE AVAILABLE ON THE CNS SPDS The plant variables that are expected to be available on the CNS SPOS were derived from the safety function and EPG information needs defined in Section 3 to 6. A summary of the selected plant variables is presented in Table 7-1, along with an indication of the basis for selecting each variable. The hierarchy of SPDS displays for the CNS SPDS is expected to be based on the BWROG Graphic Display System described in Section 5. The planned display hierarchy for the CNS SPDS, and the initial partitioning of plant variables among the displays in this hierarchy has been discussed previously, in Section 2. It is expected that the specific variables and displays provided on the CNS SPDS will be revised-in the future, as neces-sary to improve integration of the SPDS with the plant-specific E0Ps and operator training, i l l I l l l 60 i . ~ -....

503-8500000-76 3/1/84 Table 7-1. Plant Variables Expected to be Monitored by Cooper Nuclear Station SPDS. Technical Basis Variable Currently Available Safety On CMS Verfables EPGs Function Remarks PMIS Average power range monitor (APRM) X X Yes power level Intermediate range monitor (!RM) For continuity, used in Yes log power BWROG GOS IRM range Necessary for interpreting yes IRM log power reading IRM position Necessary for interpreting Yes IRM log power reading Source range monitor (SRM) X Yes count rate SRM position Necessary for interpreting Yes SRM count rate All-rods-in status In lieu of individual Yes control rod position Yes Scram demand status X Reactor pressure vessel (RPV) water level - narrow range (0* to +60*) X X Yes - wide range (-150 to +60*) X X Yes -{' - refueling range (0* to +400*) X X (1) t. - fuel zone range (-100* to +200*) X X Yes RPV pressure X X Yes RPV isolation desend status - Group 1 (MSIV) I X in lieu of isolation Yes valve position 3 - Group 2 to 7 X In lieu of isolation Yes valve position Yes Safety / Relief Valve (SRV) position X Yes High pressure Coolant irjection (HPCI) ' X pump flow rate Yes Reactor core isolation cooling (RCIC) pump flow rate X Yes Low pressure core spray (LPCS) - pump flow rate X Yes Low pressure coolant injection (LPCI. or RHR) pump flow rate X Notes: (1) Possible future addition to CNS PMIS. Variables will be available on the SPDS if they are added to the PMIS. 61

503-8500000-76 3/1/84 Ta'> l e 7-1. (Continued). Tecnnical Basts Variabla Currently Available Sa fety % CNS Yariables E PGs Function e marts e p3g5 Drywell pressure X X Yes Drywell temperature average X X (3) local (individual) X yes U) Drywell hydrogen concentration X X Orywell oxygen concentratton X X Yes Drywell sump collection rate (sump pump flow rate) X In Iteu of sump level Yes Containmnt activity (area radiation) X Yes Suppression chamber (toris) pressure X X Yes Suppression pool (torus) water temperature - average X X (3) delta T heat casacity (calculated) X (2) Suppression pool (torus)' water level X-X Yes Suppression chamoer (torus) hydrogen concentration X X (1) Suppression chamber (torus) oxygen cor. centration X X Yes Secondary containment (reactor building) differential pressure X Yes; Secondary contaten=nt (reactor butiding) X Alarm status in lieu of (1) area temperature alarm status analog value Secondary contafnment (reactor butiding) X Alarm status in lieu of (1) MVAC exhaust radiation level alarm status analog value Secondary containment (reactor butiding) X Yes area radiation level Secondary containment (reactor butiding) floor drain sump water level and X Alarm status in lieu of . (1) torus area water level monitor analog value Notes: (1) Possible future addition to CNS PM!$. Variables will be available on the SPOS if teey are added to the PMIS. (2) Calculated by $POS as the difference between suppression pool heat capacity te m erature Ilmit and suppression pool average temperature. (J) Calculated from nultiple points on PMIS for this variable. 62

503-8500000-76 3/1/84 l Table 7-1. (Continued). Technical Basis Verfable Currently Available Safety On CNS Vertables (PGs function Re w rts PMIS Offstte radioactive release rate from plant release points X X elevated release point (ERP) effluents Yes Augmented off. gas (A0G) and radmaste (Rid) building effluent Yes Reactor trailding effluent Yes - Turcine building effluent Yes - Steam f,et air ejector ($JAE) monitors Yes l l 1 i i 63 i

503-85000b0-76 3/1/84 8. CONTROL ROOM SOURCES FOR IDENTIFIED PLANT INFORMATION THAT IS NOT EXPECTED TO BE AVAILABLE IN THE CNS SPDS Some of the plant variables identified in Sections 3 to 6 are not expected to be available on the CNS SPDS. Alternate control room sources for this information, or justifications for not monitoring this information are identified in Table 8-1. Adequate sources in the control room are available for all of the plant variables listed in Table 8-1 that apply to the Cooper Nuclear Station. l l l-l I 64 l

503-8500000-76 3/1/84 Table 8-1. Sources of fdentified Infonnation Not Expected to be Monitored by the Cooper Nuclear Station SPDS. Reference to variable OG Sa fety Remarks Variable III Functions (2' GDS ) EPGs Secondary Containment HVAC X No HVAC coolers in main HVAC Cooler Differential Temperature system serving secondary containeer, Control Rod Position X X Core mimic including control rod position information is available on control room panel 9-5. "All-rods-in" status is on SPOS. Pump and valve status is RPV Coolant Injection System X Status displayed on control room pan-1 9-3. Pump and valve status is dis-Normal. Shutdown Cooling y System Status played on control room panel 9-3. Time from Shutdown to Onset can be determined by operator y of Core Uncovery from reactivity and reactor water level trends on SPOS. This variable is not ex. plicitly calculated. A containment and RPV isolation Primary Containment Isolation y Valve Position mimic is available or. control room panel 9-3. The SPDS does monitor group isolation deraand status. RCS Soluble Baron Concentration Manual sample. This relatively y (grab sample) static data is available on control room log sheets. I Radioactivity Concentration or Manual sample. This relatively Radiation Level in Circulating X static data is available on Primary Coolant control room tog sheets. Manual samole. This relatively Analysis of Primary Coolant y (gamma spectrum) static data is available on control room log sheets. Core Flow X Not related to EPGs or safety functions. Avaliable on control room panel 9-5. MSIV Position X A containnent and RPV isolation mimic is available on control room panel 9-3. The SPDS does monitor Group 1 (MSIV) iso-lation demand status. [ Containment Activity X Not related to EPGs or safety functions. Available from (gas,todine,particulatal analysis of grab samoles. SPOS i does monitor containment area radiation level. Notes: (1) See Section 3. l (2) See Section 4 (3) See Section 5. l 65

503-8500000-76 3/1/84 9. CONCLUSIONS This safety analysis defines in detail the technical basis for initial selection of the plant variables which will be displayed on the Cooper Nuclear Station SPDS. The technical bases sited in this analysis are two-fold: o Regulatory Guide 1.97 (Ref. 3) provides the basis for identifying the plant variables needed to assess the status of safety functions identified in NUREG-0737, Supp.1 (Ref.1) for a wide range of events,,which include symptoms of severe accidents. o The BWROG Emergency Procedures Guidelines (Ref. 2) provide the basis for identifying specific information needs of control room operators during abaarmal and emergency conditions, in determining the safety status of the plant, and in assessing the need for corrective actions by the operators to avoid a degraded core. Completeness of the set of variables selected for display on the CNS SPOS is dependent on the ; adequacy of the sources referenced above. It is expect,ed that the specific variables and displays provided on the CNS SPDS will be revised in the future, if nei:essary to improve integration of the SPDS with l the plant-specific EOPs and operator traini.ng. The information contained in this safety analysis is comprehensive I and should support-the' detailed design of the CNS SPOS displays. In addition, sufficient information is provided to support an independent review and audit of the selection process undertaken in this safety analysis to establish SPDS dtta requirements. This document should, therefore, be j fully responsive to the requirements for an SPDS safety analysis, as established in NUREG-0737, Supp. 1 (Ref. 1). I j 66 i t. =,_ , - -. ~,

503-8E00000-76 3/1/84 10. REFERENCES 1. NUREG-0737, Supplement 1 (Generic Letter 82-33), " Requirements for Emergency kesponse Capability," U.S. Nuclear Regulatory Commission, December 17, 1982. 2. " Draft Emergency Procedure Guidelines," Revision 3G, BWR Owner's Group, November 10, 1983. 3. USNRC Regulatory Guide 1.97, Revision 2, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980. 4. Mullee, G. R. and Aburomia, M.M., " Simulator Evaluation of Boiling Water Reactor Owner's Group (BWROG) Graphic Display System (GDS)," ALO-1019, Sandia National Laboratories, May 1983. 5.

Buckley, D.W.,

Lobner, P.R.,

Hope, E.,

and Roy, G., "BWR Graphics Display System Dynamic Screening Program," ALO-1003, SAI-1381-364LJ, Science Applications, Inc., February 1982. 67 _}}

v t e Letter Sequence Other
TAC:51232, (Open)
Initiation Request Acceptance... Administration Meeting Questions Answers Results Approval... Other: A840006, Forwards Three Repts Re Facility Safety Parameter Display Sys (Spds),Per Requirements of Suppl 1 to NUREG-0737. SPDS Will Be Installed During Upcoming Recirculation Piping Replacement Outage, ML20087A949, ML20087A964, ML20087A998, ML20099D468, ML20099D484, ML20155G453, ML20155G463, ML20199C843, ML20199C854, ML20212E169, NLS8400254, Forwards Schedule for Submittal of Addl Safety Parameter Display Sys Info,Per 840803 Request, NLS8500045, Forwards Addl Info Re Spds,Per 841005 Commitment.Finalized List of SPDS Parameters,Isolation Test Procedure for Fiber Optics & Description of Displays for SPDS Encl, NLS8500319, Advises That Addl SPDS Info Will Not Be Submitted by Date Specified in Re Verification & Validation Testing.Start of 1,000 H Test Scheduled for 860115.Info Will Be Submitted by 860501, NLS8600153, Forwards Results of 1,000 H Test After SPDS Implementation for Confirmatory Review, Documentation That Info Displayed on SPDS Readily Perceived & Does Not Mislead Operators.Validation Test Plans & Results Also Encl, NLS8600212, Forwards Rev C to Failure Mode & Effects Analyses & Rev 0 to Final Verification & Validation Rept for Nebraska Public Power District Plant Mgt Info Sys
MONTHYEAR1986-03-12 [Table View]

ML20087A949

Text

SUMMARY

=

SUMMARY

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