ML20126J247

Research Info Ltr 98:submits Results of Research on LWR Status Monitoring During Accident Conditions

ML20126J247
Person / Time
Issue date:	08/18/1980
From:	Murley T NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Harold Denton, Minogue R, Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE), Office of Nuclear Reactor Regulation, NRC OFFICE OF STANDARDS DEVELOPMENT
References
RIL-098, RIL-98, NUDOCS 8104230867
Download: ML20126J247 (22)

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UNITED sTATEa

f. g NUCLEAR REGULATORY COMMISSION

s ;y Q.[(./3 g 1 t WASHINGTON, D. C. 20555 /

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/ i i MEMORANDUM FOR: Harold R. Denton, Director ,

Office of Nuclear Reactor Regulation  !

t Robert Minogue, Director Office of Standards Development .

Victor Stello, Director Office of Inspection and Enforcement g

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FROM: Thomas E. Murley, Acting Director v.,  !

Office of Nuclear Regulatory Research

SUBJECT:

RESEARCH INFORMATION LETTER # 98 ,

LIGHT WATER REACTOR STATUS MONITORING DURING ACC! DENT CON :TIONS This memorandum transmits the results of conpleted research describing an improved method for analyzing accident sequences. The method is demonstrated by applying it to detemine the operator's information needs during accidents. The results are relevant to the revision of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following '

i an Accident." Appendix A summarizes the results and Appendix B is the detailed documentation on which this Research Infomation Letter is i based'. .

1 1.0 Introduction l The accident at Three Mile Island in March 1979, and the results of l subsequent investigations have reemphasized the importance of reactor operators and the role they play in detemining the level of safety associated with nuclear power. At the same time, the adequacy of some longstanding regulatory approaches to safety, such as design basis events and the single failure criterion, is being questioned. Al ternate methods, some employing insights from probabilistic risk assessment, are being proposed to broaden our perspectives on reactor safety.

This research introduces an analytical approach which could make signifi-cant contributions to accident analysis. As an illustration, the approach is used to identify the necessary and sufficient set of light water reactor instrumentation needed by analyzing the appropriate operator response to specific plant states associated with risk significant 8104280 JY 7

l 3 j-i Multiple Addressees l l

j accident sequences. The resultant set of measurable parameters is  :

compared to the list of such parameters in Regulatory Guide 1.97 "Instru- I mentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant ,

and"Envimns During and Following an Accident."

Criterion 13, " Instrumentation and Control," of Appendix A. " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, " Domestic l

Licensing of Production and Utilization Facilities," includes a require- i ment that instrumentation be provided to monitor variables and systems - j over their anticipated ranges for accident conditions as appropriate to ensure adequate safety.

Criterion 19, "Contml Room," of Appendix A to 10 CFR Part 50 includes a l requirement that a control room be provided from which actions can be

-taken to maintain the nuclear power unit in a safe condition under '

i accident conditions, including loss-of-coolant accidents, and that equipment, including the necessary instrumentaion, at appropriate locations outside the control room be provided with a design capability '

for prompt hot shutdown of the reactor.

Criterion 64, " Monitoring Racicactivity Releases," of Appendix A to 10 i CFR Part 50 includes a requirement that means be provided for monitoring the reactor containment atmosphere, spaces containing components for  ;

recirculation of loss-of-coolant accident fluid, effluent discharge I paths, and the plant environs for radioactivity that may be released  !

from postulated accidents. ,

Regulatory Guide 1.97 describes a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation to conitor plant' variables and systems during and following an accident '

in a light-water-cooled nuclear power plant. The most recent version of the guide (Revision 2 dated June 1980) contains a list of variables to be measured together with the associated measurement range and purpose for the measurement. The design criteria (e.g., qualification and display requirements) for the associated instrumentation are also identi-fied. This list was assembled by surveying the NRC staff and by reviewing 3 accident response procedures involving preplanned manual actions during l design basis events. Interactions among the staff, the Advisory Comittee on Reactor Safeguards, licensees, applicants, vendors, and other interested ,l members of the public have resulted in modifications to the original list. For the most part, it is a product of engineering judgment based on past experience and on the perceptions of individuals as to the

' significance of particular parameters and the impacts of implementation.

N research described herein developed a more systematic approach to I Ining instrumentation requirements. The application of the technique 8l to confinn the reasonableness of the list generated via engineering

.ent. It also identifies, however, differences whose significance ,

i lN Id be reviewed by the staff.  :

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Multiple Addressees 2.0 Discussion The analysis reported here is based on two observations concerning the enhancement of operator capabilities: .

1. The operator's capability to diagnose and respond correctly to accident conditions is sensitive to the amount and quality of infonnation available to him through the plant instrumentation.

Accordingly, one of the primary objectives of this analysis was to deternine systematically the necessary and sufficient set of plant instrumentation which would satisfy the operator's infomational needs during accident conditions.

2. While there exist many diverse aspects of the general operator / plant interface problem, any efficacious changes to present designs and/or procedures must be based upon a foundation consisting of a thorough understanding of the plant response to accident events and a careful delineation of the specific responsibilities of the operator as the accident sequence progresses. Therefore, an additional objective of this analysis was to develop such a foundation upon which both this and additional analyses concerning enhanced operator capability could be perfomed.

The technical approach used in this analysis to accomplish the objectives outlined above was based on evaluating appropriate operator response in a logical progression of events. This approach can be succinctly summarized by addressing three fundamental questions.

1. What actions can (or must) the operator take in response to the accident condition?

2. What infomation is required by the operator to take this action? .

3. What instrumentation is necessary and sufficient to provide this

infomation?

By translating the general objectives into these three interrelated questions, the analysis could be performed systematically, increasing assurance that important operator infomational needs will not be overlooked. ,

The approach is diagrammed in Figure 1. The seven accident sequences analyzed were detemined to dominate risk in the previous risk analyses from which they were selected. All sequences involved system failures in excess of the single failure criterion.

.:h sequence the physical response of the plant is defined in tenns i asurable parameters. The time-dependent variations and the interrelationships *

.ese parameters generate an " accident signature," a uniquely characteristic

' .y which can be used to evaluate the status of the plant.

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Determine the necessary Identefy tocats 3avolvsag ~ Pee Plan d- _

aeJ 5effR6**t Para- _

Oper atier Ateen evners to Orse rnbe Status of the Plaat t>tene the Persetel Orftme the Phsst(al enenorwaa A. sos e nteJ .ith 5elett irDef taat Phenomens assalated rey States sa l apwt* 1 _%. ~+

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Figure 1. Flow chart of technical approach used to detennine operator Infonnation needs and instrumentation requirements. Dotted lines imply possible feedback in an iterative process.

_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _-_m. _ _ _ , e- iu-af 7 --

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Multiple Addressees The development of the event trees began with the trees as they appeared t in the original risk analyses. The events in each sequence which involved operator action were identified and in some cases broken down into addi-tional events in order to highlight individual operator tasks. In addition, the sequences were expanded (events added to the event tree) to include additional operator actions which could be performed to l prevent core melt, but were not taken credit for in the original analysis.

These additional events included " repair events," where the operator is given the opportunity to attempt to restore or replace a particular function, and " delay events," where the operator is called upon to delay an inevitable melt as long as possible or to perform some other consequence mitigating action. The result of these efforts was an " operator action event tree" which identified success paths and which logically displayed the role of the operator throughout the progression of the accident.

Figure 2 presents a simple example of such a tree developed for interfacing systems LOCA (V) sequence of WASH-1400.

Once the event logic and physical response of the plant are established, it is relatively straightforward to identify the key operator actions and the operator's informatien requirements. This is done by characterizing the status of the plant on ea:h branch of the tree and associated appropriate actions in terms of physically measurable parameters. Table I summari:es this information for the t-sequence.

Prior to presenting the results, it is important to point out that this work represents a first-of-a-kind study conducted over a short time

. period. As such,there are limitations involved and refinements to be made in the analysis. These are delineated in Appendix B. Section 5.

3.0 Results The results of this study pertinent to the revision of Regulatory Guide 1.97 are summarized in Appendix A. The table lists the variables derived from the-analysis, indicates the significance of each, and identifies those not contained in the Revision 2 to Regulatory Guide 1.97. This study yields results which compare quite favorably with Revision 2, '

despite the major variations in technical approach. There are, however, some specific differences worth noting, especially PWR reactor vessel water level, containment sump water temperature, process parameters e associated with the low pressure injection system, and positions of various valves. I Speaking nore generally, this research introduces some important new

-aots and technical approaches which, if properly developed and

. could make significant contributions to accident analysis. It ,

izes the perceptions of the operator, the needs for information ne alternative successful actions one might take given various

's inations of component failures. Beyond determining instrumentation

.irements, the methods have important implications with respect to l

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Operator Operator Operator Lnsures RPS long-Term

Delays isolates Melt Rupture lleat Rennival '

i Operation .

O.u t_co.ine A no nw21t G

• O's designate key plant states i a late melt (D

C late nw21t e,Ye5

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- [1 no nelt OI t

E late eelt F early melt h G early nelt l

Figure 2. Interfacing Systems LOCA Operator Action Event Tree i i b -. ~ ~

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s I. Sumary cf Key Operator Actions and Infonnation RIquirements for V-Sequenc2

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APPROPRIATE OPERATOR INFORMATION REQUIRED PLANT 5(ATE DESCRIPTION OF INFORMATION REQUIRED AC1!DN FOLLOWING TO TAKE (See F1 pre 2 ) PLANT STATE TO IDENTIFY PLANT STATC STATE IDENTIFICATION APPROPRIATE ACTION

@ Rupture of check e RCS P T Prepare for actions See states @ @ @

l valves results in LPIS e Pressurizer water illustrated in Fig.4.8 and@

l overpressure and level i rupture e Containment P.T.R

- e Aux. Building T R

. e LPIS P.T.R.F l

j @ Reactor scram; decay Control Rod Position Initiate core melt RCS P;T.

' power level; RCS pres- Neutron flux delay actions and Vessel water level sure rapidly decreas- isola tion HPIS flow ing to HPIS actuation Accunulator fler level Accumulator Tank level LPIS flow from RWST CSIS flow from RWST  ?

RWST level Isolation valve (s) posi tion Reactor not scramed; Control Rod Position Monitor approach to Primary system

@ radiation level power level above Neutron Flux cladding failure; capacity of HPIS to RCS P T initiate consequence Aux. Building R remove heat; core melt mitigation systems assumed to follow

@ Minimum sufficient RCS P.T Initiate (or continue) Isolation valve (s) flow from HPIS to keep Vessel water level isolation actions position core covered and RWST level prevent melt LPIS flow from RWST CSIS flow from RWST

@ Either insufficient Sameas@ Same as@ Sameas@

HPIS flow or excessive draw on RWST

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Table I. (Continued) ,

APPROPRIATE OPERATOR INFORMATION REQUIRED PLANT STATE DESCRIPTION OF INFORMATION REQUIRED ACTION FOLLOWING TO TAKE (See Figure 2) PLANT STATE 10 IDENTIFY PLANT STATE STATE IDENTIFICATION APPROPRIATE ACTION LOCA successfully Isolation valve Initiate long-tem RCS r',T

@ heat removal Vessel water level isolated before core position Steam generator water melt occurs RCS P ':

LPIS flow level Pressurizer water lew1 Auxiliary FW flow CST level Reactor power level Isolation fails af ter Sameas@ Monitor approach to Primary system

@ core melt and initiate radiation level

• delaying action core melt occurs when RWST consequence mitigation RWST level &

Aux. Building R depleted actions

@ Isolation fails; no Same as @ Same as @ Same as @

delaying action has occurred; core melt occurs more quickly than 4a

@ Long-tem heat removal RCS P.T established Steam gen. level Aux. FW flow Long-tern heat removal RCS P.T Initiate consequence

@ Steam gen. level mitigation systems not established; no corrective action Aux. FW flow possible i

P = Pressurt T = Temperature R = Radiation Level F = Flow Ra te l

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Multiple Addressees developing energency procedures, generating training simulator exercises, and designing operational aids, including computeri:ed diagnostic systems.

Therefore, the methodology itself, as described previously in Section 2.0, should be viewed as a major result of this research.

4.0 Recommendations The following recommendations are made with respect to the results reported here:

1. The regulatory and standards development staffs should review the concepts and technical approach described in Appendix B and advise the research staff as to the value and validity of these technioues, areas for their improvenent, and suggested topics for their application.

Assuming the methods are deemed promising, the regulatory staff may also want to encourage licensees and applicants to apply them to their own facilities.

2. The regulatory and standards development staffs should review these results and assimilate them into the technical basisAppropriate for decisions relative to the revision of Regulatory Guide 1.97.

considerations should be given to the limitations of the study which generated these results.

In the meantime, RES is continuing this research. Additional accident sequences are being analyzed as is a broader spectrum of reactor designs.

Furthermore, the development of best-estimate codes to calculate the physical response of plant systems during accidents continues to provide updated information on which to base these analyses.

The RES technical contact for this work is Raymond DiSalvo.

/

Thomas E. Murl , eting Director Office of Nucle , egulatory Research

Enclosures:

1. Appendix A: Summary of Yariables Identified in Sequence Evaluations

2. Appendix B: LWR Status Monitoring , .

During Accident Conditions (NUREG/CR-1440) l; S

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APPENDIX A ..

APPENDIX A - Summary of Variables identified in Sequence Evaluations Major Purpose for Indicated PWit Accident iequence Measured f Y $C Sg ilF TML/TftB Coassents' Vertable 2 saae as V same as V Provides primary Indication Control Itod

• Verification of scram Saee as V of successful scram

Position i Indicates shutdown margin;

• Verification of scram Sane as V Same as V Same as V fleutron Flun important af ter initial j f ailure to scram; might be unreliable under velding comit tlons RC5 Pressure

• Diagnosis of Initiat-

• Identification of same as 52C

• Indication of tran- -

Ing LOCA event initiating small stent initiator

• Determination of need break

• Indication of in-for and effectiveness *Deteretnation of tegrity of primary of ECl need for and ef- system eProvides, along with tectiveness of *Provides, along RC5 tesperature, de- ICI and ECR with RC5 temperature, gree of subcoollag

• Provides, along degree of suttooling 3 e

• Indication of break with RC5 temper- "

Isolation ature, degree of l

sutguoling i

RCS Temperature

• Provides, along with Same as V Same as V *Prowldes, along Measurements of both hot RC5 pressure, degree with RCS pressure, and cold leg temperatures of sulxualing degree of sut(oolleg useful for natural cir-

• Indicator of natural culation circulation f Pressurizer Lbvel

• Indication of initiat-

• Indication of Same as 57 0

• Indication of int-Ing event initiating event tlating event

• 8pdication of isolat-

• Diagnosis of stre ton of break and location of  ;

break i

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sumary of variables _IdentlIled_irtSitquencn_ Evaluations Major Purpose for Indicated PWR Accident Semsence .

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I' 5C Sg itr TML/lbt s' Coassents

(*r 2 Pressuriter Relief Valve position,

• Verification of Other parameters designed ,

discharge line prenurtrer relief to indicate RC5 Integrity ,

flow, or drain can be used as back-op to -

valve reclosure tank level these direct Indications ,

Vessel Water level

• Indication of need for eindication of taut 6 t.e.d := a.,. Geese 1.97.

and effectiveness of initiattag event Sane as 5,0' eindh atton of int- Other thermodyr.amic parameters I

[Cl sindication of tlating event (e.g. RC5 pressure and tem- .

eindication of 150- need for and *Veillication of~re- perature) can t,e used for lation of break effectiveness of lirl valve closure e.ast accident conditions.

ECl and success of msin- f urther analysis is reestred talning adequate to determine if these para-liquiJ Inventor y meters are sufficient fer all - '

Significant accident condt-tions i ,

Frimary System

• Indication of approach Sane as V Sane es V Sane as V un-line timely measurements are necessary; system should Radletion Level to core melt reneln operable under all oAssesssent of entent accident conditions including !

of core damage fol.

loulng restoration containment Isolation ,

of e. ore cooling i

Boron Concentrat-

• Indication of shut- Same as V Same as V same as V Could be useful back-up ll lon doom margin accident progresses to con- ,

ditions uhlch make neutron l fluu monitors unreliable l I

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Sumary of Variables Identified'in Sequence Evaluations

.. Major Purpose for Indicated PWit Accident Sequence

• I SC Sg Hf IML/IML9' Comuments r 2 Containment

• Diagnosis of Initiat. = 0tagnosis of s ofagnosis of =Verificatfon of Pressure Ing LOCA initiating brean initiating relief valve etndication of break reclosure CSIS failure. =Provides. In

• Indication of repair of C515 (umbtnation (ontainment in-and effectiveness with sum tege t ty of C5RS water teihp-eProvides. In crature in-ctednination with ditation of ~

sump water teg - adequate NP5H esature. In- for ELR pu w s dication of

• Indication of adettuate leP5tl for containent ECR pumos. Integrity eindication of eindicatiews of contalrunent in- CSR5 failure p tegrity or effective-to ness overtfles contain- Sames as 5 2C S*** ** S 2C

' Containment lsolation Valve ment isolation to Position preclude trans-port of radle-active material through contain-ment penetrations O

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. _ . _ - - _ . . - . . _ _ . - - - - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - _ - a - - -~. _ _ _ - _ _

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Sunenari of Variables Identified in Sequence Evaluations Major Purimse for indicated tvit Accident Sceluence Measured SC Si lir mt/ Ins / Coments

' i verlebte

• 0fagnnsis verleication of Containment humidity can -

• Diagnosis of Initiat- *0tagnosis of of e be used as a highly tellable Containment initiating break initiating relief valve Teaperature Ing LOCA reclosure backup to containment

• Indication of firea k

• fn.flcation of pressure and tengerature (515 failure. to indicate primary system repair of CSIS. OH5 falluse

• or effectiveness er effective- Integrity of C5R5 nen t Serves as backup to con- 1 eDiagnosis of Inttlet- Same as V Same as V tale ent pressurg and Containment ^

Radiation Level ing LOCA teeversture for indication of le n of primary boundary integrity eindiqate ab- Can also be used as indicator ,

Containment Susy

• Indicate avall- of inttleting break * ,

ability of water sense of tieter level for (CR and C5R5 coolant flow l 3 between uliper

• and Itwr l convartwent j and suct en-i, fut restor-atton ni flim

{

Not included in lleg. Guide ,

sin conjunction Samw' as 52C Containment Suay l.97 lieter lesperature with contain-went pressure.

Indicates ade- [

quate NP5tl for C5R$ and (CR l pump operation 0

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Sununary of Variables identified in Sequence Evaluations .,,

MJJor Purpnse for Indicated IWft Accident Seepsence I f Me. .. itt/TMLS Cossacets I

IC 2 5,Ilf Verteble

• ladhat8on of Isot specifically identitled tipper Contalmuent major cause in lleg. Gelde 1.97 tot only

. [tg artment fer ICL5 applicable to plants with Water level and e cc le c ula t ion stallarly designed contain-Drain Valve (be- fatture sent drain systein tween upper and

• Inditatton of lower compart-sents) position repair and ecslorattim of flow

• Indication of cap- ' Indication of. Wu5C 7

• Indication of Intfl- .

Steam Generator ating transient ability of long term feedwater system Level alnification of per-decay heat renovel per f ormance fn aance of aus-litary system Steam Generator '* Indication of capability eindication of Sawc as $2E *I"dI'*II'" 'I P

fo,mance of feedwales Pressure of long ters decay liest feedwater systein performance system

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reseval *lrullcation of cap-

• Indication of setoralary systein atillity of using Inte9eIty e twwlensate psargss (IHt )

' Steam Generator

• Indicattois of 5ame as $ 2C 5d"* 8550 2 Safety / Relief setorafary systen. -

+

integrity Valve Positions

• Indication of fatti- Pune discharge pressere 1 Main Feedwater ator, sectess of (not tr<luded on Reg. Guide flow repair. or uttitra- 1.97) could be used as tion of condensate techup indication and gamms (for lit) assist in specifying cause of failure for IML 1

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Slimary of Variables Ident.ified in Sequence Evaluations , ,,

Majer Purpose for indlcated I'WR Act! dent Sequence a IU I g3II Ili/IME S Co'unents I

'Vart...e 2 Pump discharge pressure Ausillary Feed-

• Indication of adequate eindication of 5ame as s7t

• Indication of oAfnts could be used as bedupt water flow to steam adequate flow to f altare and c eter-teater Flow mination of se- flow control valve positions generators f or long steam generators to enhance test storation could be useful in de-term decay heat reenval termining cause of AfWS remva l failure and in regulation of restowed AfW5

'! *Potentially useful tent included in Peg. Guide Condensate Pump in diagnosis of 1.97 Flow or Discharge initiating event Pressure *InJIcation of effectiveness of using condensate pungis to supply feed-water to steam gen-erators for some IML initiators 3=

5tese Supply to

• 0lagnosis of AfW not included in Iteg. Guide &

failure cause and 1.97 AfW turbine sutisequent sepalr driven puq same as V Passive system; Indirect Accumulator Tank *lndicate injection Indication of performance level. flow rate, after initiator can be obtained from other and/or isolation parameters valve position

• Indication of ability Sane as V Same as V Some as V Condensate Stor-age Tank Level to use AfW as heat removal system 59 g -

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l l Sumary of Variables identified in sequence Evaluations M.6jor Purpose for Indicated I'WR Accident Se.luence 5C 181L/1 HLB' Coassents I' 5,ltf

$"r 2 Refueling Water eindication of avall- *lemilcation of Saw as S jC ability of water for evallability of Storage Tank Level ECl water for 101

• Determination of op-timum use of RW51 water supply in tore melt delaying actions Sam as 5 2C Pump discharge pressure HPlh finw eindicates success of

• Verification of can be used as badup (Cf for core melt (Cl operation Indication of systest delay actions following int-flator operation LPIS pressure. temperature.

LPIS pressure.

• Diagnosis of initlet- and radiation level not tesqperature. Ing event (dif ferent- included in Reg. Guide 1.91 late from other events 3*

radiation level.

and/or flow with similar RC5 re- O sponse)

• Indication of isolation of break

• Determination of break location LPIS Isolation

• Indication of success flot teocluded in Reg. Guide valve position of isolation 1.97

• Indica tion rump distliarge pressure can Containment Spray

• Indication of need to sindication of be used as tekup indication isolate system for failure of CSIS of operation flow (including and subsequent (ontainent of system oseration CSIS and C5RS) delaying actions repair heat eemoval i

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tg Summary of Variables identified in Sequence Evaluations .

Major Purpose for Indicated INR Accident Seeruence I

- Van , Y $C 2 5,Ilf IMt/TMLs' cosaments same as V Saast- .s V Pimp discharge pressure can RitR Flow elswitcation of system same as V te used as backup indica-operation for long tion of system operation term heat reseval Same as V saae as V Mot specifically included Positions of key eindication of capebil- Seer as V in Reg. Guide 1.97 valves in safety Ity of systems to related systems operate when called 6 (ItPIS, LPIS, upon C515 [5RS, eDiagnosis of failure i OIRS, RilR)

Component Cooling

• Indication of same as SyC Water flow in effectiveem s CitHS heat en- of contairwent '

cliangers cooling using C5RS Sane as V Sane as V 3 i Component Cooling eindication of effect- Same as V e

, Water Flow to iveness of long-term CD RitRS liest Ea- heat removal changes a.. ins.,,e ties ,1, . ,,,.

Auxillary Bulld- eDiagnosis of inttlat- **t *** l.&.e s e n. , cotee i.97 Ing leaperature ing event or Radiation

• Determination of level successful isolattose of break i

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Sunnary of Variables identified'in Sequence Evaluations

. Major Purpose for Indicated pWR Accident Se<suence

• IC 5 gilF IML/IMtn' Cosaments I

r e  ? -

• Indication ut only applicable to plants Containment with sixh a system aualliary heat tlw anuent removal fan dis- of contain-tharge flow >= a t (tm l in's which is leIng pee-I n seil and t he v entu i re-no nts fin f $fl5 Sane as V S, sV

• Indication of safety 5tatus of Class- *Verlitcation of safety system availability lE power supplies system avallel*llitF eDiagnosis of cause to key safety que AfW5 failure system cosmonents Status of leon- b Class-l[ Power Sane.as y eindtration of In-Supplies

• Verification of same as V Illating event for available power souece Itu tt' and deter-mination of re-stnestlon a

t t

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Sunmary of Variables identified in Sequence Evaluatfof1s -

Eh Major Purpose for Indicated tut Accident Settuence Measured it Vertable Control lied Position

• Indication of f ailure of automatic scram, and success /f ailure of manual insertion of rods Itcutron Flus

• Indication of f ailure to scram and deterinination of elfcct of manual shutdown actions RCS Pressure *Detetinination of effect of delayed scram .

• Meed for and etfectiveness of itPCI 4

et f fectiveness of long tera cooling

• $econdary Indication of reactor shutdown

• Indication of ef fectiveness of core cooling (in tombination with RCS tocation of instruments not RCS Temperature yet determined; cure exit ,

pressure)

. teeversture (as listed in 3 Reg. Guide 1.97) does not e seem to be best location. y Intended for those accident conditluns where coolant level measurement might te espected to be unreliable Vessel teater

• Indication of inttleting transient event level

• Indication of water Inventory 4

• Determination of need for and effectiveness of emergency core cooling

• Determination of tehen to secure 11Pl5 and vely on RLIC Iur long terie cooling M5lW should automatically Main Steam Flow

• Indication of inttlator close following the in-isolation

• Determination el smtential core cooling proccetures Ittating loss of feedwater Position transient event

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i l . h s' Sunnary of Variables identifled in Sequence Evaluations 14ajor Purpose for Indicated IWit Accident Sequence COMMEMIS Meas W , TC Verteble .

Safety / Relief

• Indication of efrect of delayed shutdown Valve Positions

• Indication of potential etfectiveness c' manual shutdown using 5tCS in Primary System

• Indication of primary boundary integrity (including ADS)

Radiation Level *Infor1setion for monitoring of core melt in Coolant

• Indication of amount of core damage Containment

• Indication of integrity of primary pressure l>oundary . __

Pressure

• Indication of containment integrity e

Contalement

• Indication of integrity of primary pressure tmundary Temperature

• Indication of containment integrity Containment eindication of Integrity of primary pressure (namdary ,

Radiation Level p a a

Suppression Pool eindication of primary coolant boundary Integrity level eindication of availability of water for itR Suppression Pool eindication of ability of tooling system to ptst water iepperature Boron Tanit level

• Indication of Boron injection for shutdown SLCS flow or

• Indication of system operation pump discharge pressure e

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[ I, Sumary of Variables identified in Sequence Evaluations 'v .

Major Purpose for Indicated IWit Acclitent Setsuence COPM(IIIS Pim. -

• 1C Verlabee

• Determinat ton of eifecttveness of sunual shutd. no using 51ts; feut incInded in Reg. Guide Boron Concentrat- 1.97. Could be useful ton ~ Indication of shutdown surgin taskup under accident (onditions uhtch male neuteon flun onmitors less seliable Feeduater flow

• Indication of inttleting event Feeduster pump

• Indication and diagnosis of cause of initiatos -

discharge pressure current to punes.

or controller position teot specifically included RCIC valve pos- *[nsure availability of system in Reg. Guide 1.97

' itions

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Stesen flow to

• Indication of adequate flow to ensure system operation y RCIC turbine .

RCIC flow or pump eindication of successful system operation or (ause of failure discharge pressure ,

itPCS valve pos-

• Ensure availability of system flot specifically included in Reg. Guide 1.97 itions 1905 flow. pump

• Indication of successful system weration or cause of f ailure discharge pres-sure, or current to pumps -

- , . . _ ,, , ..n- r . , - , - , , -,nov -- - no- e- e s -n s , -+, - - , m ,ve.- ,-.-r,- --.-..-e -w - -m-~ v ,-,,-- - --w - - - - - - . - - - - - - . - - - - . - - - -

/ / .'

Sumary of Variables identified in Sequence Evaluations

. Major Purpose for Indicated IMM Accident Sequence COMMtItIS Measured it Variable .

flot incitekel in Reg.'

RHR valve pos.

• Allow startup ni system acid subsequent opee at<>r t eieit e til of f liu h ide 1.91 ition (valves required for pre-tearning and flushing and flow control valves) liftit heat en- *Information necessary for manual startup and indication of sutisequent changer inlet / system pes forinance outlet teagiera-tore tr5W valve eindication of availability of system position I

If5W flow or e indication of system operation g pungs discharge pressure e

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