ML19347C741
| ML19347C741 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/30/1980 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML19262F356 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM NUDOCS 8101050330 | |
| Download: ML19347C741 (200) | |
Text
DUQUESNE LIGHT COMPANY 3eaver Valley Power Station Response To NUREG 0737 Non-Proprietary Version i
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i 12/30/80 l 810.105O N 1
Pagn 1 I:e= I.A.l.1 Shift Technical Advisor Insinefe~mfmanMilesis
- 1. Tr*d g that neers the lessens-learned requirements shall be ce=pleted by January 1,1981 or by the -d=e de fuel leading license is issued, whichever is later.
NUREG-0573 - I.esscus learned. Short-:er= recc==enda-dons. Section 2.2.1.6 s ates:
"In addi:1cn to a bachelor's degree or ecuivalent in a science or engineering discipline, :he shif: technical advisor shculd have seecific training in de respense and de analysis of de particular plan: for transients and accidents. Shift technical adviscrs shculd also receive training in de structures , systa=s , and ce=-
penen: design and laycut cf de plan:, including traf ning in the func:1cus and capabilicies of instru-nents:1cn and centrols in the centrol :ce=."
Shortly af ter de interi= Shif t Technical Advisors were *md, a five (5) veek :raindng progrs= vas condue:ed. Prio :o assign =ent :o shift, each interi= Shif: Technical Advisor had satisfactorily cc=pleted de progra=. In addi: ion to classrec= verk, the interi= Shif: Technical Advisor : raining progra= included rec (2) veeks a: the Surry si=ula:or where each Shift :achnical advisor was required to func:1cn during various scenarios involving ::ansient and accident condi:icus. All U CAT 3 1:e=s were ec=pleted November, 1980.
Se de: ails of de classroc= and si=ulator : raining given are as follevs:
Classrec= Training - Three ~4eeks - A 3V?S
'Jeek 1 Ace =ic Physics
- Nuclear Physics
+ Reac:or Physics
+ Reactor Kinetics
+ P'4R Core Physics
'Jeeks 2 and 3 SV?S Pri=ary Syste=
3V?S Primary Au: ciliary Sys:e=s
- 3V?S Engineered Safeguards Features 3VPS Instru=enca:1cn and Control Systems 3V?S .%jor Secondary Syste=s
- 3V?S E=ergency Plan
Page 2 I es I.A.l.1 Shif: Technical Advisor (centinued)
Si=ulator Training - Twe '4eeks - A: Surry
- Nor=al Operations
- Abnor=al Operaticus E=ergency Operations e Role Playing as STAS The si=ulator portion of de progrs= provides de STAS w1:h sixty (60) hours of intense sd 'M or :r * *ng.
I. No later dan Janua:f 1, 1981, all licensees of opera:ing reac: ors shal' provide dis cffice vid a description of : heir icng-:er= S*A pregra=, including qualificatien, selection cri:eria, training plans ,
and plans, if any, for :he eventual phase-out of de program.
Dese-detion of -le STA Training The objec:ive of the Shif t Technical Advisor Program is :: provide de required academic =aterial necessary for the STA's role. Iha:
is, to provide transient and accident assessment advice :o de Shif t Supervisor and perform evaluation of Shif t to Shif t cperations to ensure dey are conducted in a safe m er.
The entire progras is =cdulariced wi:h each :odule satisfying existing regula:iccs for STA crai d"g vi din its academic area.
The program censists of five (5) =cdules:
Module I Funda=entals Training 17 weeks Module II Plan Systa=s Instrue:1on 14 weeks Module III Transient and Acciden: 2 weeks i
Assess =ent Module I7 Manage =en:/ Supervisory 1 week Training Modula V S4 'ta:cr 2 weeks MODUI.Z ! -
Module I consists of a detailed 17-week lecture series on de funda-nen:als of Nuclear Pcwer Genera:1ca. It includes 7 weeks of Generic and 3eaver 7 alley specific A:c=ic, Nuclear and ?WR core physics, as well as a weeks covering Ther=cdyna=1cs, Fluid Dyna =1cs and Heat Transfer. he re=ainder of the =cdule presents Nuclear Field Engineering concepts, i.e.; Inorganic Chemistry and Corresion; Nuclear Materials; Zlectrical ""heory; Nuclear Instrumentation and Control; and Radia:icn Protection and Control.
i Page 3 I:e= I.A.1.1 Shift Technical Advisor (continued)
Se sedule is presented by qualified Engineers in their appropriate field of exper ise. Lesson format would be to first review assi .ned Math and Science knowledge and den :o present new =aterial, er suring de nr.cessary basis exists. Each lessen also includes a real lant reference, as well as several problem solving sessions. This 'odule is presented at the college level to de STA, =eaning de info.mation provided will build upon his existing kacwledge. Qui :es will be administered weekly and sn-ng exa=s ad=inistered at approximately l 4-week intervals broughout the module. The successful completion of dis Funda=entals Trn'n h g will provide the new STA de theoretical foundation necessary to access plant transients and accident situations.
MCDULI II Modu.'e !! presents the details of de design, functional arrangement and op atien of the 3eaver Valley Plant Systa=s. This training is necessary to ensure that the meaning and significance of instrument readings and the effect of control actions are known. Experienced operator training engineers or licensed instructors will present the sacerial. Qui::es will be ad=inistered weekly and su= mary exams will be ad-d d stered at approx 1=ately 4-week intervals.
The purpose of this nodule is to take de college graduate, with lit,.le or no knewledge of PWR system specifics, and give his suff.cient back-grcund for perfor=ance of his plant analysis function. 721s is acccmplished by presenting flev paths, major components, Instrumentation and Ccutrols, operations and te<-hnical specificatiens related to each system listed in the attachment.
i MODULI III This base kncvled,. 3TA has received is integrated in Module III.
i Transient and accide t analysis are discussed on a plant specific basis.
Ihis area is vital to de STA's role in the control roc =, in dat he nust be capable of recognizing problem areas which =ay be masked by nor=al transient indicaricas. Therefore, de assessnent function is stressed in this =odule, w1:h all subjects discussed being related to indications fres "Real" plant cenditions. In addi:1cn to identi-fi.. tion of problems, preacribed operator responses and alternative damage ni:1 gating actions are discussed. This Moduit is presented by experienced SRO licensed instructors.
The purpose of this training is to broaden the ability for prcept l recogni:1cn of and response to unusual events and to recognize that real accidents a-= d '-dated and acec=panied by unusual events. An exam is p;cvided at the end of the 2-week progras to monitor the STA's understanding in this important area.
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Pags 4 I:a= I.A.1.1 Shift Technical Advisor (continued)
MCDUTI IV - MANAGEMENT AND SUPERVISORY SKILLS his sedule focuses on -die need for the Shif Technical Advisor to acquire a =acroscopic viewpoint of his role as supervisor and advisor.
Due to de high specialized cachnical cc=petence require.d for :his posi ica, little i=portance is placed on de interassociation role de STA will play in the control reos.
This week lowers the magnifying glass on the hu=an factor, the need for the STA :o recognize his capaci:7 in the :eam efforr required for safe, effective reactor opers ions.
The sedule ce==ences wi:h a look at de broad base skills necessary for de supervisory and advisory role. This is followed by an in-dep h analysis of de tools available anc those required for achieving com-pe:ency in de STA's function.
E::phasis is also placed on de stress factor and the inevitabili:y of having :o make decisions under 1:. Also interpersonal ec==ccications and chain of ce==and is weighed strongly during :he presentation.
~he theory and practice of the supervisory role is reinforced throughout the week with heavy student participation in actual role playing and decision aking in case studies presen:ed.
MCDULE V - S~MLUTCR The s1=ulator porrion of the progras prov. des rigorous training in nor=al, abnor=al and accident cenditions. The progras is 2 weeks in length and provides 60 si=ulator and 20 classroo= hours.
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Page 5 I:e= I.A.l.1 Shif: Technical Advisor (centinued) 3es"er Vallev/OIPC Cemearison "Overall evaluation of the 3eaver Valley STA Progra= with the DGO standard indicates all subject =aterial w1 hin de recc== ended education and training I
program are covered in required detail.
Assu=ed prerequisites beycad a Eddh Scheel Dipic=a (6.1.1) and college level funda=en:al educa:1cn training in nathe=a:ics are not presented since selec-
- 1cn cri:eria requires an engineering degree for the STA position. *he
- -a d dng college level subjects are presen:ed in Module I of the 3eaver 7 alley STA ?rogra=.
Plan: specific academics are censolidated vidin the funda=entals ::sining area ensuring de $*A receives his plant specifics while learning de fun-da=en:als of nuclear plant design. This no: culy prevides be:ter trad"d g, but reduces .he to:al instrue:ce contac hours necessary. The Ei?O Standard requires 520 !.C.H. (Instruc:or Cen:act Ecurs) in de funda=entals area and II0 I.C.H. in the plant specific acade=1c area. Our progra= waives 90 of the hours in funda= ental =athematics and utill:es 510 I.C.H. for the re=ainder of de fundamench and plan: specific training. All subject =aterial is presented :o provide the STA de kncwledge to neet his objective.
Manage =en:/ Supervisory Skills (6.2) are presented as recc== ended. A '40 I.C.H.
program in Mcdule III of :he 3eaver Valley STA course is provided. In addi-
- 1:n Oc DGO cu lines, 3eaver Valley includes organising and planning, influencing and guiding, and nanaging in the future, essential skills and
- case studies.
The Plant Systa=s Instrue:ica (6.3), Ad d dstrative Centrols (6.I.) , and General Operating ?rocedures (6.3), are presenced in Module II of de progrs=.
The ec=bina:icn of instructor centac: hour requirement frc= DGO is 310. The 3eaver Valley Mcdule II Plant Systa=s ins: rue:1cn p cvides 336 I.C.H. In l addi:icn :o CIPO cutline, 3eaver Valley Module II includes red posi:1cn
- indica ica, pressurizer pressure and level control, fire pro
- ecticn, fuel handling, spen: fuel systa=s, plan: cc=puter, :echnical specifica:icns, startup, she down, running precedures , and operatices ad d-d r,tra:17e procedures.
Specific opera:1cns ad d Mstrative guidance is included da 3eaver Valley Module II. Securi:/ and access centrol are part of de new empicyee orden:a: ion, emergency plant is covered under dat progra=, radiological control is included in 3eaver Valley Module I. Duties, responsibili:ies, and plant status are included in Beaver Valley Mcdule III.
l The Transient, Acciden: Analysis and E=ergency Procedure (6.6), is provided by Module III of de progra=. The required 30 I.C.H. appeared inadequa:e for this i=portant area; -herefore, Module III of :he 3eaver Valley STA
?rogra= provides 60 I.C.H.
Paga 6 I:e= 1.A.1.1 Shift Technical Advisor 3eaver Vallev/Ci?O Ccccarisen
- he 3eaver 7 alley Si=ulator progra= is 30 hcurs - 60 s1=ulator centac: ac 20 classroc= hours. This is a carcially available progra=, designec by si=ulator operators, to meet industry needs by providing a well-balanced progrs= w1:h de S'"A duties and responsibilities in =ind.
Annual Requalification:
The 3eaver Valley S 3's will be scheduled :o at:end selecied secticus if de Licensed Operator Requalification Progra=. S.is ccuples vid the 30-hour SIA si=ula:or package, exceeding Ei?O guidelines.
[ Cualifica:icn and Selection Cri: aria
- 1. Degree (or equivalent) in Engineering or Physical Sciences which includes a =ajority of the following discipline areas =
of study:
(a) Mathe=atics (b) Che=1stry (c) Physics (d) S:rength of Materials (e) Thermcdyna=1:s (f) Fluid Dyna =ics (g) Hea: Transfer (h) I.'ectronics (1) bectrical Pcwer Equip =er.:
(j) 'iuclear Funda=en:als (k) Radiological Centrols
- 2. S.e Shift Technical Advisor shall have a =1.94m m of 13 =cuths of nuclear pcuer plant experience, at least two Ocnths of which shall be at an cpera:ing nuclear plant.
(a) A ~ = v' '" of six scads of :his experience may be obtained in the military er at a productica nuclear plan and should be evaluated on a case-by-case basis.
(b) A =axi=u= of three =cn:hs of systems and operations training
=ay be applied Ocward dese experien:e require =ents.
(c) A: least 12 =cnths of his experience shall be a: the s:ation at which the posi:~.on is to be filled. This =ay be waived in part when :wo essentially identical plants
! are involved.
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- 3. De=enstra:ed ability to work with others and provide cencise, clear ec==unicaticus.
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Page 7 Item I.A.l.1 Shif Technical Advisor 3eaver Vallev/ INFO Cc=carison
- Heurs equivalent to college requirements for degree or within NRC educational guidelines. Supplemental training to be provided in areas where significan: deficiencier exist based on INFO recc= mended
" college level education" requirements. Tc.e required education or tr,1,4"g require =ents can be waived if an accept:ble level of knew-ledge is demonstrated by evaluation or thr:: ugh comprehensive anninations.
Plans for Phase-out of the STA Progras
- 4e have not yet developed plans for the eventual phase-out of the program.
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Page 1 I:em I.C.1
- Guidece fot the Evaluation and Development of Proc 1dures for Transients and Accidents ne '4estinghouse Owner's Grcup vill submi
- by January 1,1981, a detailed description of our program :o comply vi:h de requirements of Item I.C l.
The program vill identify Cvners Group previcus submi::als :o the NRC, which we believe vill cceprise the bulk of the respense. Addi:1cnal effor rem 2 ired to obtain full compliance with this 1:em (vich proposed schedules for completion) vill also be identified, as discussed with the NRC en November 12, 1980.
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Training for Mitizaring Core Da= ace I:e= 1 - Deveico Training Progrs=
Se 3eaver 7 alley Training Progrs= censists of six (,6) 4-hour lectures covering incere instr =entatien, execre nuclear instr.:=entatien, vi:al instrcen:atien, pri=ary che=1stry resul:s vid core da= age, radia:1 n teni:cring systa= and gas generation. Shif t TechMcal Advisor and opera:ing perscnnel frc= de Sta:icn Superin:endent, drough the opers:ica chain :c de licensed operaters, shall receive all :he training. Super 71 sors and
- achnicians in Radeon, Che=istry and I&C shall be scheduled for selected
'.ac cures .
Due :o NRC changes addressed in E' REG--0737, -le Jriginal 3eaver 7 alley Plans of having had the p cgrs= ccepleted pric: :: Septa =ber, 1981 have been re-evaluated. ? cgrs= scheduling censis:en vi h Operation / Main:enance a:affing vill no: per=1: progra= cc=pletien un:il Dece=ber, 1981.
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Page 1 Item II.E.1.2 Auxillahr Feedwater System Aut=utic Initiation and Flcw Indication The reviews and documenta:1cn required :s be submi::ed in response to this 1:am are not completa at thi.s ti:ie. A 3.pplemental response addressing the docu::lentation requiramants of this 1:am will be submitted by February 1,
, 1981.
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Page 1 r.e=
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u . s F.=erzenc'r ?cwer Surolr for Pressurizer lieaters ne inf or=ation required by this 1:e= is ac: yet fully ec= piled.
An additional sub=1::21 will be ade by February 1,1981.
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i Paga 1 7:e= II.E.4.2 Containment Isolation Decendabiliev
?csiticu (5)
The containment setpoint pressure dat iniziates contatment isolation
- f or nonessential penetrations =ust be reduced to the =ini=um ec=patible i
vid normal operating conditions.
3eaver Valley has been designed to use a suba:=cspheric contad an: syste=.
Technical Specti1 cation 3.6.1.4 (page 3/4 6-6) requires tha: containmen:
internal air partial pressure shall be =aintained greater dan S.9 ?SIA and less than a range of values related to river water and refueling water storage :ank (RWST) ta=peratures which is set for:h in Figure 3.6-1, a : ached hereto. Our operational experience has been : hat we have not had difficulties =eeting de require =ents set forth in his Technical Specification.
3asically, the -=vd- m allewable containment air partial pressure under
-le coldest river water and RWS! temperatures is 4.2 ?SI less dan ac=inal at=ospheric pressure, whereas the setpoint for isola:1on of nonessential systa=s is set at a concal =ent pressure of 1.5 PSI above nc=inal at=cs-J pheric pressure. The difference between the highest allevable centainment -
pressure and the setpoint of courn4 ment isolation (?hase A)/ Safety Injection .
is 3.7 ?SI.
While we agree that de setpoint for isolation of nonessential syste=s in con:nd--an: should be the -d 4 m setpoint cc=patible vid nor=al ope ;cing ,
conditicus, we do not believe da: it is appropriate o reduce the se point en subatsospheric con
- ents below de nor=al range of at=ospheric pressure plus reascnable instru=ent error and drif t. Our concerns are based upon de fac: da: to reduce de centain=en: isolation setpoint, a hardware change
=ust be =ade to lockcut or bypass de concal =ent isolation signal which provides the reactor trip function, safety ' injection functien and isolation of nonessential syste=s during dese codes of plant operatien where the I
cen~=4--ant is per:10:ed Oc be operated at atmospheric pressure. The use of dis bypass feature :o per i: :his type of operation will create the possibili:7 dat a =aFanc:fon or error will occur which could prevent these i=portant safety fanctions frc= occurring during accident conditions when cperating in =cdes where these protective features shculd be in service.
'Jnder these circu= stances, we do not believe da 1: is appropria:e to I
reduce de existing contain=ent isolation setpoint frc= its present value.
Posi:icn (6)
The centainment purge supply and exhaust inside and cutsice contain=ent-isola:1cn da=cers (VS-D-3-3A, 33, 5A and 53) have deir linestar:ars de-energized, their breakers locked in de "0FF" posi:1cn and de =anual operating =echanis=s chained and locked to prevent opera:1cn. The
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Page 2 Ita= II.E.'. 2 Centainmen: Isolati=n Cependability (centinued) i courainment purge damper pcsiticus will be verified every 31 days by
- use of an Operating Surveillance Test (CST 1.47.2) which providas for the verification da
- the da=pers are closed by energi:ing -le lines:arters for the dampers, observing da: the green posi:1on indica:ica ligh: is illum4,=ced, and then de-energizing the lines:ar:ers. Dampers 75-D-5-3A and 5A vill be obserred visually every 31 days :o assure tha: the dampers are clesed and tha: the manual opera crs are chained and locked.
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Page 1 I:em II.F.1 Accident Monitoring This 1:em requires a submi::21 if the syst.es to be installed deviates from
.he posizion as set forth in NU G 0737. Duquesne Light C.:cpany placed a purchase order on December 17, 1979 for a S?!NG-4 system which we believe fully neers the posi: ion as set forth in NUREG 0737. The NUREG requires
. hat licensees have available for review the final design descript. ion of
- he as-buil: system, including piping and instru=en: diagra=s and either (1) a description of procedures for system operazion and calibration, or (2) copies of procedures for syst-2m operation and calibration. These design documents are not expected to be available for review until July 1, 1981 and
- he procedure documen:s are not expec:ed to be available until September 30, 19'1. Equip =en: delivery is scheduled for July,1981 and installation vill be completed by January 1, 1982.
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I:e= II.F.2 Report of Details for the Planned Instr =entaticn Sys:e=
for Menitorinst of Inadecuate Core Cooling INTRCEUCTION This report detailing de planned instr =entation for senitoring of Inadequa:e Core Cooling (ICC) has been prepared and sub=10:ed for NRC review in accordance wi:n NUREG 0737, Nove=ber 1980 "Clarifica 1cn of IMI Ac:1cn Plan Require =ents," I:e= II.F.2. This repor has been prepared frc= a '4estinghouse generic report on de "'Jestinghcuse Reactor Vessel Level Instr =entatica Systa= for Monitoring Inadequate Core Cooling, December 1980.
3ACKGRCUND 3e NRC has established requirenents (1:e=s I.C.1 and II.F.2 of SUREG 0737),
to provide de reac:cr opera:cr with instr =entation, procedures, and
opersecr action is required :c cool the core before da= age occurs.
Ihe report has been organized to address the specific NRC report contant ite=s in -J.e sequence listed in Ita= II.F.2 of NUREG 0737. The su==ary report, "'4estinghcuse Reac:or vessel Level Instr =entation Syste= for Mcnitoring Inadequate Core Ccoling" (Reference 3), prepared en a generic basis has been a:: ached for ec=ple:eness and reference :o specific see:1ons for supple =enting infor=arion is =ade where appropria:e.
RE?CRT
- 1. JESCRIPTICN OF PRCPCSED FINAL SYSTE4 1.1 ?ba 1 design descriptica of additional instr =entation and displays.
1.1.1 General Description 3e reactoi vessel level instru=entatien syste= (RVLIS) uses differential pressure (d/p) neasuring devices to =easure vessel level or relative void cen:ent of the circulating primary coolanc systa= fluid. The syste= is redundant and includes aucc=atic cc=pensation for potential e:perature varia:1cns of de i pulse -
lines. Essential infor=ation is displaye 1 in the sain centrol roc = in a for= directly useable by the operator.
The functions perfor=ed by de RVLIS are:
Assist in detecting the presence of a gas bubble or void in the reactor vessel
Itc= II.F.2 Pagn 2 Assist in detecting de approach to ICC Indicate de for=atien of a void in de RCS during forced flew conditicus.
1.1.2 Hardware Descriptica 1.1. 2.1 Differential Pressure Measure =ents The R7LIS (Tigure 4-1, Ref. 3) utilizes two sets of three d/p cells. These cells =easure de pressure drop fro = :he botto=
of the reactor vessel to de top of the vessel, and frc= the hot legs to the top of the vessel. This d/p =easuring systa= utill:es cells of diffaring ranges :o cover differen: iicw cehaviors wi:h and wident pu=p operatica as discussed belcw:
- 1. Reac:or Vessel - Upper Range (A?,)
The d/p cell AP, shown in Figure 4-1, Ref. 3, provides a seasure=ent of reactor vessel level above the hot leg pipe when de reactor coolant pu=p (RC?) in the loop with de hot leg eccuection is not opsracing.
- 2. Raactor Vessel - Narrow Range (A? )
b This =easure=en: provides an indicatien of reactor vessel level fr== the bot:c= of :he reac:c: vessel to de top of de reac:or during natural circula::.on conditions.
- 3. Reactor Vessel - Wide Range (AP )
This instru=ent provides an indication of reactor core and in:ernals pressure drop for any ec=bination of operating RC?s. Cc=parison of de neasured pressure d:cp vid de nor=al, single-phase pressure drop will provide an approxi= ate indication of de relative void centent or densi:7 of de circulating fluid. ~his instru=en: will moni:or coolant conditions on a continuing basis during forced ficw conditions.
To provide de required accuracy for level =easure=ent, temperature measure =ents of the i= pulse lines are provided. These =easure=ents ,
together with the existing reactor coolant ce=perature =easurements and wide range RCS pressure, are e= ployed to ec=pensate the d/p trans=it:er cutputs for differences in syste= density and reference leg densi:y, particularly during de change in the environ =ent inside de cen:ai=ent structure following an accident.
The d/p cells are located outside of de contai=ent to eli=inata de large reduction (approxi=ately 15 percent) of neasure=ent accuracy associa:ed with the change in de contai=ent enviren=ent (temperature, pressure, radiation) during an accident. The cells are also located cutside of con:ai=ent so that systa= cperation including calibratien, cell replace =ent, reference leg checks, and filling is =ade easier.
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I::= II.F.2 Pagn 3 1.1. 2. 2 Sysca= Layout A sche =atic of .he syste= layou: for the RVLIS is shewn in Figure 4-2, Ref. 3. here are four RCS penetrations for the cell reference lines; one reactor head connection ac the reactor vessel head vent pipe, one connec:1cn to an incere instra=ent condui: at the seal table, and connec:1cus into the side of rao RCS hot leg pipes.
The pressure sensing lines extending fr = de RCS pene rations vill be a ce=bina:1:n of 3/4 inch Schedule le0 piping and 3/8 inch
- ubing and will include a 3/4 inch =anual isola:1cn valve. These lines connec: to six sealed capillary i= pulse lines (rao at the reac:or head, reo at the seal cable and cne a: each hot leg) which trans=i: de pressure measure =ents to the d/p trans=1::ers located curside de ccc: sin =ent building. The capillary i= pulse lines are sealed at de RCS end with a sensor bellows which serves as a hydraulic coupling for de pressure =easure=ent. The i= pulse lines extend fr== the sensor believs drough de conta1==en: vall :o hydraulic isola: ors, which also provide hydraulic coupling as well as a seal and isolation of the lines. The capillary tubing extends frc= the hydraulic isolators to :he d/p trans=1::ers, where instru=ent valves are provided for isolation and bypass.
Figure 4-3, Ref. 3, 2.s an elevation plan of a typical plant showing de routing of de i= pulse lines' . The i= pulse ifnes frc= de vessel head connectica =ust be :=uted upward out of de refueling canal to de operating deck, then radially cward de seal table and then :o the contain=ent penetration. The cennection to de bot:c= of the reactor vessel is =ade drough an incors detec:cr condui: which is capped with a T c=nnec:1cn at de seal able. The i= pulse line frc= this connection is routed axially and radially to join with de head connection line in reuting to che pene: rations. 51=11a:17, the hot leg connec:Lon i= pulse lines are rou:ed cward de seal table / penetration routing of de other reo connec:icts.
The impulse lines located inside the contain=ent building will be exposed to the contain=en: te=perature increase during a LCCA or E13. Since the vertical runs of i= pulse lines for= de reference leg for de d/p measurement, the change in densi:7 due to the a cident temperature change must be taken into account in the vessel level deter =ination. Therefore, a strap-on RTD is located
- on each vertical run of separately routed i= pulse lines to determine I the i= pulse line temperature and correct the reference leg density
! centributien to the d/p measure =ent. Temperature =easure=ents
! are not required where all three i= pulse lines of an instra=ent train are routed :cgether. Based on de studies of a nu=ber of representative plant ar angements, a
- m of 7 independent l vertical runs =ust be =easured to adequately ec=pensate for density changes.
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Ite: II.F.2 Pagn 4 1.1.2.3 Microprocessor (RVLIS) and Displays The =icroprocessor RVLIS indications include equivalent reactor vessel level on redundant flat panels vid alphanumeric displays provided for control roo= installation in addition :o having dis infor=ation available for display at the =icroprocessor chassis.
RVLIS is configured as reo protec:1on sets, in separated sections of a single instr:nent rack. The block diagrs= of the RVLIO using
=icroprocessor equip =ent is shown in Figure 4 This diagra= shows dat in addi:1on :o de reactor vessel level (d/p) ::ansmitter input, dere are also te=perature compensating signals, reactor pump :"nning s.atus inpu:s, and RCS para =eter inputs to each chassis of the :wo redundant sets. Se output of each set will be :o displays and to a recorder, as well as an output for a serial data link. A general display arrange =ent is shown in Figure 4-3, Ref. 3 and exa=ples of infor=ation displays of Vessel Level and Vessel Level Trend are provided as Figures 4-3a and 3b, Ref. 3.
- he =1crepor:essor syste= inpu:s are as follows. (See Ref. 3, Sec. 4.2.2.1 for a = ore cotalace description of dese inputs.)
- 1. Differential Pressure Trans=1:rers - The three d/p trans=1: ers per se: are used :o nessure the d/p's between the three pressure .ap points on de pri=ary systa=.
- 2. Reference Leg Te=cerature RC - The reference leg te=pera:ure RCs are used to =easure the ta=perature rf the coolant in the capillary tube reference legs. his is used to co=pute de densi n of de reference leg fluid. The gpical arrange =ent of
-hase
- RCs is shewn in Figure 4-6, Raf. 3.
- 3. Hot ieg Te=cerature - Ilther existing or new wide range het leg ta=perature sensors are used to =easure -le coolant ta=perature. This :e=perature is used :o calculate coolant densin.
- 4. Wide Ranze Reactor Cociant pressure - Either existing wide range pressure sensors or new pressure sensors will be used to =easure reac:or coolant pressure. *he pressure is used :o I
calculate reactor coolant densi n.
The block diagra= of the co=pensation functions is shown in Figure 4-7, Ref. 3.
- 3. Digital Incuts,- The reactor coolant pu=p status signals indicate whether or no: pu=ps are running. Recognising that hydraulic isolators are provided on each i= pulse line for contain=ent isolation purposes, each hydraulic isolator has 11=1: swi:ches :o indicata they have reached the W : of travel.
I tcm II. F . 2 Pago 5
- 6. Censt:v C=censation Svstem - To provide the required accuracy for vessel level =easure=en:, te=perature =easurement of the i= pulse lines are provided. Sese =easure=ents , together with the existing reac:ce coolant :e=perature =easure=ents and wide range RC3 prasaure, are employed :o ccer-rma 24 d/p transducer outputs for differences in system density and reference 'eg densiry, particularly during the change in de envirea=ent inside the contain=en: structure folicwing an acciden:.
- 1. 2 ::etailed description of exis:ing 1:stru=enta:1cn syste=s (including para =eter ranges and displays) pertinen: to ICC consideraticas.
1.2.1 '41de Range Reactor Coolan: Pressure - present :.nstru=en:ation is available for detW ng general RCS pressure trends during the ICC event. The expecter' ace;. racy following ICC eve =:s is such . hat dis instr =ent. cannut be used fer precise deter =inations of de pressure required to assure onset of icw head safety injec:1cn ficw to the RCS.
- 1. 2. 2 Pressurizer Pressure and Level - condi:icus in de pressurizer will generally lie curside the ranges of these instr =ents during an ICC event in a *4estinghouse ?'4R. Pressuri:er pressure and level are not used for deter =ining nitigatica acticus :o be taken during ICC.
- 1. 2. 3 Au* ary Feedwater Flcw - present instr =entation is ava11able for assuring the sufficiency of =akeup vater ficu [
- o de steam genera:crs during an ICC even:.
1.2.4 Core Subccoling - does not provide useable infor=ation during and ICC condi:1cn. '4111 indica:e superheat conditions in core coolant. "4111 help indicate the approach to ICC by shcwing saturatien conditions. Refer :o Table 3.1, Ref. 3 for infor=ation on sukooling =enitor.
1.2.5 Stea=line Pressure - present instru=enta ion is available for deter =ining heat sir.k availability and heat removal capabilig during ICC =1:igatica actions.
1.2.6 Stea= Generater Level - present instru=entatica is available for de ar=ining de availability of a heat sink for de RCS during an ICC condition.
1.2.7 'Jide Range Resistance Temperature Detectors - present instr =en:ation is available in deter =ining trends of recovery acticus but =ay not be available in deter =ining the onse: of ICC cendt:icns for all break si:es.
j Item II.F.2 !
Pags 6 1.2.3 Core Exit Ther=ccouples - present inst = mentation is j 3 available in determining both the existance of ICC and ,
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the trends of recoverf actions. :
l 1.3 Description of Planned Modifications to the Existing Inst =enta:1cn Systems
- Current plant inst = mentaticu is considered adequate to determine i
! heat sink availability, :o detect the onset of ICC, and to detect
( the effectiveness of mitigation acticus following the onset of an I j ICC event. Therefore no changes to existing plant inscrumentation 1 is planned other than connecting wiring to selected instrumentation signals. Core Exit Thermocouples are being evaluated per Attachment j I Criteria which may resul: in recommended changes to the existing '
3 system. Results of this evaluation and planned changes will be :
i submitted by 7/1/31.
- 2. DESIGN ANALYSIS (including various instruments to monitor water level !
and availaole test data :o suppor: de design) l 2.1 Design Analysis 2.1.1 Parameters Critical to ICC The analysis provided in Raferences 1 and 2 delineates those parameters
] crt:1 cal for the detection of and-the necessary mitigation actions for the recovery from an ICC condition.
To briefly summarize those paramatars, ICC is detec:ed by aider high core ext: thermocouple ce=peratures or by a low reactor vessel level indication (core uncovery) in conjunc:1on with core ext:
thermocouples. Mitigation actions consist of depressurizing the j reactor coolant system (RCS) to permit injection of accumula:or water and/or :o establish low head safety injection flow.. The RCS
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is 1:self depressurized.by depressurizing the steam generator secondarf side. Critical parameters at this point are steam generator pressures and wide range RCS loop temperatures. Once low
( head safety injec:Lon flow is established, transfer out of the ICC
- procedure can be nada when core exit thermoccupla temperatures are reduced and
- he reae:or vessel level gauge indicates a level above
- de top of the core.
With de exception of reactor vessel level, all paramarers are
, monitored by currently existing instrumentation. The RVLIS is provided to persi: a more continuous indication of the approach
- o ICC. ,
j To s cluate the various instruments to =eni:or water level, a report by the LCFT Experimental Measurements 3 ranch was reviewed which
- described and discussed the perfomance to date of. (1) Liquid Level j Transducers-(LLT), (2) Clad Thermoccupies, (3) Fluid Thermocouples ,
(4) Self Powered Neutron Detector (SPND), (3) Differential Pressure Transducer, (6) External Neutron Detectors and (7) Heated Differential
- Thermocouples Liquid Level Detector (LLD).
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_ . _ ,. .... _ , _ _ _ _ _ . _ _ _ . _ - . . _ _ _ . - . . ~ . _ ___
- . _. . - - _- . - . _ - . - -. - --. -. ._ __--_---. ___ -= -- _ ._ _ - . _ . .-
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Item II.F.2 4
Page 7 I: was concluded that either the Westinghouse RVLIS or possibly de Heated Differential Ther:occuple were the most premising level 4
detector systems alternatives. Ecuever, in evaluation of c=amarcial availab111:7 all systems other than de Westinghcuse RVI,IS vere i
considered too developmental to warrant any investment if Duquesne I.ight Ccmpany were to meet the original installation date of 1/1/81
, or :he presen: instanation da:e of 1/1/82.
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! 2.2 Test Data Supporting the Design l *4estinghouse Forest Hills Test Facili:7 A breadboard ins:allation consis:ing of one train of a AVI.!S was instaned and testad at :he Westinghouse Fores: Hins Tes: Fac111:7
, The systa= consisted of a fun single ::ain of RVLIS hydraulic ccmponents (sanser assemblies, hydraulic isolators, isolation and bypass valves and d/p transmi::ers) connected to a simulated reactor vessel. Process connecticus were made :o simulate de reactor head,
- het leg and seal table connections. Cap 1Hary
- ubing which in one >
sensing line simulated the mdm:m expected length (400 feet) was used :o connect the sensor assemblies to the hydraulic isola:crs and all joints were welded. Connec: ions between de hydraulic ,
isolators, valves and transmitters utilized compression ft :ings in scst cases. Resistance temperature detectors, special large volume sensor bellows and volume displacers inside de hydraulic isolator assemblies which are nor=any part of a RVLIS ins allation were not included in the ins:anation since eleva:ed temperature testing was not included in the program.
i The hydraulic isola:cr assemblies and transmi::ars were scunced at an elevation slightly belcw de simulated seal :able eleva:1on.
l The objectives of the test were as fo11cws:
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! 1. Obtain instana:1cu, filling and maintenance experience i
l 2. Prove and establish fining peccedures for ini:ial fining and system nintenance.
, 3. F.stablish calibratica and fluid inventory maintenance procedures for shutdown and normal opera:1cn conditions.
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4 Prove icng :erm integri:7 of hydraulic components
- 3. Verify and quantify fluid ::ansfer and =akeup requiaments
- associated with insert =en
- valve opera:1ca.
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- 6. Verify leak test procedures for field use i
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Item II.F.2 Paga 3 Reactor Vessel Simulator:
The reactor vessel simulator consisted of a 40 foot long 2 inch diameter stainless steel pipe with taps at the top, side and bottom to st=ulate the reactor head, hot leg and incore detector thi=ble conduit penetration at the bottom of the vessel. Tubing (0.375 inch dia=eter) was used to connect this icwer tap to the sensor at the simulated seal table elevation and the hot leg sensor to the head connection was simulated by 1 inch tubing which connected the sensor to the vessel.
The reactor vessel s1=ulator was designed for a pressure rating of 1400 psig to comply with local stored energy and safety code considerations.
Installation:
The system was installed in the high bay test area of the Wes. 2ghouse Forest Hills Test Facility by Westinghcuse personnel under the supervision of Forest Eills Test Engineering. All local safety codes were considered in the construction.
Filling Operation:
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Ite ;I.F.2 Page 9 c _,.
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- 3. A DESCRIPTION OF ADDITIONAL TEST PROGRAMS TO BE CO?OUCTED FOR EVALUATION, QUALIFICATION AND CALI3 RATION A variety of Westinghouse test programs are in progress or will be carried out to study the Static and dynamic perfor=ance of the RVLIS at two Westinghouse test facilities, and to calibrate the syste= over a range of nor=al operating conditions at each reactor plant where the system is installed. These progrs=s, which supplement the vendors' tests of hydraulic and electrical co=ponents, will provide the appropriate verification of the system response to accident conditions as well as the appropriate procedures for proper operation, =aintenance and calibration of the equip =ent. A description of these progra=s is presented in the following section:
3.1 Westinghouse Forest Hills Test Facility Testing at the Forest Hills Facility as described in Section 2.2 above is ongoing in continuing support of evaluation of the RVLIS perfor=ance.
3.2 Se=iscale Tests In order to study the transient response of the RVLIS during a small-break 10CA and other accident conditions, the hydraulic ce=ponents of the RVLIS have been installed at the Se=iscale Test Facility in Idaho. Vessel level measure =ents will be obtained during the current seniscale test program series which runs from December 1980 to March 1982. The tests scheduled to be co=pleted by July 1981 are expected to provide the desired transient response verification; additional data will be obtaired from the tests scheduled for co=pletion by November 1981.
The Semiscale Test Facility is a model of a 4-Loop pressurized water reactor coolant system with elevation di=ensions essentially equal to the di=ensions of a full-size syste=. *he reactor vessel contains an electricelly heated fuel asse=bly consiscing of 25 fuel rods with a heated length of 12 feet. Two reactor coolant loops are provided, each having a pu=p and a steam generator with a full height tube bundle. One loop models the loop containing
I:em II.F.2 Page 10
- he pipe break, which can be located at any point in the loop.
The other loop nodels the :hree intac: loops. A bicwdown tank t conec:s and cools the fluid discharged from the pipe break during the simulated accident. Over 200 pressure, sempe 4sure, flew, level and fluid density inscrments are instaned in the reac:or vessel and loops to record :he fluid conditions droughout
! a test run. Test results are compared with predicticus for verification of computer code models of the transient performance.
The Westinghouse level measurements obtained during a test run will be compared with data obtained from existing instrmentation instaued on the semiscale reactor vessel. The semiscale facility has two methods of measuring de level or fluid densi:y: d/p
=easurements are obtained over 11 vertical spans on :he reactor vessel :o determine level wi:hin each span, and gam =a densitometers j are insulled at 12 elevations on the reactor vessel to deter =1ne
! :he fluid density at each elevation. *his data establishes a fluid density profile within the vessel under any-operating condi:ica, and :his information will be ecmpared with :he data obtained frem the Westinghouse level inscrmentation. Other j semiscale fac111:7 instraents (loop flcws and fluid densities j when pumps are operating, and pressure and temperatures for an j cases) win provide supplemental information for interpretation of the test facility fluid condi:1ons and level measurement.
Specific tests included in the semiscale test program during which Westinghouse RVLIS measurements will be obtained are as fo nows:
) 1. Miscellaneous steady state and transient :ests w1:h pu=ps f on and off, to calibrace test facility heat losses.
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- 2. Small-break LOCA test with equivalent of a 4 inch pipe break.
- 3. Repeat of small-break LOCA test with test facility modiided to si=ula:e a plant w1:h upper head injection (L'dI) .
- 4. Several natural convection. ests covering subccoled and saturated coolant condi: ions and various void contents.
- 5. Tests to simulate a station blackout with discharge through relief valves.
- 6. Simulation of the St. Lucie cooldewn incident.
3.3 Plant Startup Calibration h During the plant startup, subsequent to installing :he R7LIS, a test program will be carried out to confirm the system t calibration. The program win cover normal operating conditions and will provide a reference for comparison with a potential accident condi:1cn. The elements of the program are described belcw:
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I: m II.F.2 Pags 11
- 1. During refilling and venting of the reactor vessel, =easurements of all 5 d/p ::ans=1::ers would be ce= pared to confirm identical level indications.
- 2. During plan: heatup with all reac:cr coolan: pu=ps running, measure =ents would be obtained frc= the vide range d/p trans=1::ers t to confirm or correct the :e:perature cc=pensation provided in de system electronics. Se camperature ce=pensa: ion, based on a bes: estimate of the flew and pressure drop variation during s:artup, corrects the ::ansmitter cu:put so that de centrol board indication is =aintained at 100 percent over the entire opera:ing temperature range.
- 3. At hot standby, =easure=ents would be cbtained f c= all trans-
=1:ters with different ecmbinations of reactor coolant pu=ps operating, to p cvide the reference da:a for ce=parisen vi:h accident condi:icns. Tor any pump operating conditica, the i reference data, represents de nor=al cendi:1ca, '_.e. , vich a !
t water-solid system. A reduced d/p during an accident would be an indica:1cn of voids in de reac:ce vessel.
4 At hot s:andby, seasure=ents veuld be obtained from de reference leg RTDs, to confirm or correct reference leg temperature ec= pen-sation provided in the system elec:renics.
- 4 EVALUATICN OF THE ICC INSTRUME'iT STS"EM FOR CCNFORMANCE TO NUREG 073"7 I ITEM II.F.2, INCLUDING AITAC'-DfENT 1 AND APPENDIX 3 The RVL!3 is being designed on a generic basis to neet the majori
- 7 of the I:e= II.F.2, At:sch=en: 1 and Appendix 3 requiresen s as is l
indicated in Sectica 2.4 and Table 4.1 of Reference 3. Similarly, Table 3.1 of Reference 3, provides info:=ation required on the subccoling
=cnitor. Ecwever, an evaluation specific to 3V-1 of all compenents of the equipment consti:uting de Inadequate Core Cooling De:armination system is not expec:ed to be cc=pleted until 4/1/81. A repor to the NRC indicating confor=ance to the cri:eria or justifications of why cri:eria no: =et will be sub=1 :ed by 4/15/31.
- 5. DESCRIPTICN OF CCMPUTER FUNCTIONS ASSCCIA'_D '4 I~~d ICC MONITORING AND FUNCTIONAL SPECIIICATICNS FCR SOF-'JARE (ircluding addressing accredundant cc:pu:er reliabill:y) l 5.1 RVLIS Cc=puter Functions 5.1.1 Vessel Liquid Densi:y Calculation I
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Item II.F.2 Page 12 5.1.2 Vessel Vapor Phase Density Calculatica
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.J 3.1.3 Vessel Level Calculatiot
_l Proprie tary 3.1.4 Pump Flow d/p Calculation r-Proprietary 6- _
The lower of the two calculated d/p corrrctions is divided into the seasure d/p. The result is the percent of expected d/p and should read 100 percent with all pumps operating and no circulating voids.
5.1.5 Scaling of Displayed Values Each of the three d/p =easure=ents after the preceding calculations shall be scaled to read in percent. With the vessel full of water and no pumps running, the outputs of AP, and aPb should read 100 percent.
5.1.6 Disabled Inputs -
Any inputs can be disabled by the operator. This action is under the control of a keyswitch on the front panel of the main ec=putational unit and causes the processor to disregard the analog input for that variable.
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Itm II.F.2 Page 13 5.2 Func:1 w. ?pecifica:1ces for Sof vare he cu:put of each transmi::er is compensa:ed for the densi:/
difference between de fluid in :he reactor vessel and de fluid i- .2 reference leg at de initial ambient temperature. The s tapansatica is based on a vide range hot leg temperature seasurement i
or a wide range system pressure measurement, whichever results in de highes: value of wa:er density, and, derefore, de icwest value of indicated level. Compensation based on temperature is applied when the system is subccoled, and eccpensation based on pressure (saturated conditions) is applied if superheat exists
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- the hot leg ta=perature =easurament poin:.
The outpu: of each ::ansmi::er is also ec=pensa:ed for de densi:7 I
difference between de fluid in the reference leg during an accident with elevated :emperature in the centainment and de fluid in the reference leg at de ini:ial ambient :emperature. The compensation is based en te=pera:ure measurements on the vertical secticus of de reference leg.
The corrected ::ansmi::er outputs are shown on a digital display installed en de centrol board, one statement for 'each measuremenc in each train. A three pen recorder is also provided on de centrol board to record de level or relative d/p and to display trends
) in the =easure=ents. The display would also indica:e which reactor coolant pumps are operating, and which level measurements are invalid due to pump operatier..
During normal plan: hea:up or hot standby cpera:ica with all reac:or coolant pu=ps operacing, the wide range d/p display would indicate 100 percen: en the display, an indication that the system is water-solid. If less than all pumps are operating, the display would indicate a lower d/p (deter ined during the plant s:artup test program) that would also be an indication of a wa:er-solid system. .
- 41 6 pumps operating, de narrow range and upper range displays wculd indicata off-scale.
If all pumps are shut dcwn, a: any te=perature, the narrow range and upper range displays veuld indicate 100 percent, an indication that the vessel is full. The wide range d/p display would indicate about 33 percent of the span of the display, which would be the value (determined during the test program) corresponding to a full vessel with pu=ps shut down.
In the event of a LCCA where coolant pressure has decreased to
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a prede:er=1:ied se: point, existing emergency precedures would require shutdewn of all reac:or coolant-pumps. In these cases, a level will event tally be established in the reactor vessel and indicated en all of the displays. The plant operator wculd monitor the displays and :he recorder to determine de trend in fluid
= ass or level in de vessel, and confirm tha: the ECCS is adequately ecmpensating for de accident conditions to prevent ICC.
., - - - . . _ , . ~ _ - . - . _ _ - . . - -- _ . _ _ _ _ . - _ _ . , _ . . _ . ~ . - _. .
Itsm II.F.2 Pags 14 Future precedures nay rcquire operation of one or = ore pu=ps for recovery frc= certain :ypes of accidents. When pu=ps are operating while voids are developing in the systes, de pu=ps will circulate the water and steam as an essentially ho=cgeneous sixture. In these cases, ther vill be no discernible level in :he reactor vessel.
A decrease in the measured d/p compared to the nor a1 operating value vill be an ndication of voids in the system, and a continuously decreasing d/p v1. indicate that de void content is increasing, that nass is bei: g lost frem the system. An increasing d/p will indicate that . nass conten: is increasing, that the ECCS is effectively restoring the syste= nass centen:.
- 6. CURRE'iT JCHEDULE ?CR I'ISTE' ATION, TESTING AND CALI3R.CTON (including centingencies)
Installatica of the Westinghcuse R7LIS is planned during the November 1981 Refueling ou: age. Systes calibra:1cn is plant.ed af ter refueling during Cold Cperations. System Testing is planned during Hot Operations.
- 7. GUIDELI'?ES FOR USE OF THE ADDI~ICNAL I'ISTRLMENTATIO'* AND ANALYSES USED TO DEVELOP T9ESE PROCEDURES l
2 7.1 Reference Owners Group Procedures 3ased on the analyses defined in See:1cus 1.3 and 4.5 of the Reference 3, Westinghouse report, Westinghouse and the Westinghouse
, Owners Group have developed a Reference Emergency Operating
! Instruction to address recovery frce ICC cenditicus caused by I a small-break LCCA withcut high head safety injection. This l instruction has been trans=1:ted to the NRC via Westinghouse Owners Group Letter, OG-44, dated Nove=ber 10, 1980. It should be noted that this instruction was developed on a generic basis as technical reference for i=ple=enting plant specific precedures, and sust Lc milared to meet plant specific needs. Please refer to I:en I.C.1 for a schedule f ar plant specific submi::als to be made.
7.2 Sample Transient I The response of :he vessel level indicaticus, other ICC r instru=en:atica and system response during these ICC events l and recovery actions are described in Refarences 1 and 2.
l I
l
. - . . - - =_ . . -- . -- _. . - _ . _ - .
I:cm II.F.2 Paga 15 i
I 3. SUMvARY CF KI? CPERATOR ACTION INS *RUC* IONS FOR THE CURRENT EMERCENCY
?RCCD URES FOR ICC AND DESCRI?TICN CF ECW TdESE PRCCEDURES WILL 3E I
MCDLED WEEN TdE FINAL MONITORING SYS*IM IS IMPLEME'CED Ihe curren: emergency procedures have been developed from the guidelines for emergency opera:1ag procedures prepared by the Westinghouse CWers 4
Group and subst::ed previously to de NRC. *he curren: guidelines address use of de subecoling neter, bu: do no: include specific 4
reierence :o a reactor vessel level ins:rument at this eine. The specifica:1cn of use of the reactor vessel level instrument wi.1 be addressed in fu:ure respenses to I:em I.C.1.
- 9. DESCRI?CCN AND SCEDULE CCMMIT*ME'C FCR ADDI IONAL SUEMICALS 'u7,3 ARE NEDED IC SUP?CRT rdE ACCE?TA3ILI*? CF 'F.E PRCPCSD ?~NAL INSTRUME'iTAIION SYS*IM AND EMERGE' ICY ?RCCD URES ?CR ICC 9.1 Core Ex1: Iher:occuples Ivaluatica of Core Exi: *hermoccuple in 3V?S Uni: 1 :o de criteria in A::ach=en I will be performed and raccucended changes planned will be submi::ed to che NRC 7/1/S1.
9.2 ICC Ins:: =en Syste=
1 Evalua:1ca Report of th( ICC Instru=en: System confer =ance :o NUREG 0737 Ite= II.F.2 Criteria (including A::ach=en: 1 and Appendix 3) will be sub=1 :ed to de NRC by 4/15/S1.
4 e
i.
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_.-,..,....,..,,._,..,4 m.. , y y _ _ _ . . ,.._.-,_v . < _ . - - - - < - , , , - . -
Item II.F.2 Page 16 REFUENCIS
- 1. Tho=eson, C. M.-, et al., " Inadequate Core Cooling Studies of Scenarios with Feedwater Available, Using the NOTRLW Cceputer Code," WCA?-9753
(?roprietary) and *4CA?-0754 (Non-Proprie: arf), July 1980.
- 2. Mark, R. H. , et al. , " Inadequate Core Cooling Studies of Scenarios with Feedwater Available f or UEI Plants, Using the NOGUM? Ccmputer Code," WCA?-9762 (?roprietarf) and WCA?-9763 (Non-Proprietarf),
June 1980.
- 3. Westinghouse Su=marf Reporr, " Westinghouse Reactor Vessel Level Instrumentation Systes for Moni:oring Inadequate Core Cooling I (Microprocessor System)" (?roprietarf) and (Non-Proprietarf),
December, 1980 (at: ached).
i f
e 3
4 I
- _ - - , ...,e . . . * - , _ . _ - , , . , - . - , , . , - , - - .,y e,-% r .,__.-%-. , ,., , - .. -. .-.w,. ----4w- ,- , - ,.--.. , ---- - - ,-r.-,,e-
t Fage 1
- l i
Item II.K.3.1 and 2 ,
1 1
Installation and Testing of Automatic ?cwer-Operated Relief Valve Isolatien System and Report on Overall Safe:7 Effect of Power-Operated Relief Valve Isolation Svstem The '4estinghouse Owners Group is in the process of developing a repor:
(including historical valve failure rate da:a and docu=entation of actions
- aken since the DiI-2 event :o decrease che probabili:7 of a stuck-opec PORV) :o address the NRC concerns of I:es II.K.3.2. However, due to the ti=e-consu=ing processing of da:a gathering, breakdown, and evaluation,
- his reporr is scheduled for submi::al to the NRC on March 1,1981. As required by the NRC, :his report will be used to support a decision on :he necessi:7 of incorporating an automatic POR7 isola:1on system as specified in Task Action I:em II.K.3.1.
T l
I l
l
. . . _ , - - ~ , . . _ _ . . , . , . . . , _ . . . . _ _ _ _ .
- _ = ~
Page 1 I:en II.K.3.3 Recor:ing SV and RV Failures and Cga11enges
, Duquesne Light Company conmi:s to the 'ollowing:
1 a) Future failures of a relief valve :o close shall be reported promptly
- o :he NRC.
b) Future challenges :o he relief valves shall se doc :mented in the annual report.
j c) Future failures of a safety valve to close shall be repor ed to the NRC.
l d) Fu:ure challenges :o the safety valves shall be docu=enced in :he annual report.
37 " relief valves," we La:arpre: that the pressurizer pilot opera:ed relief valves are intended.
3y " safety valves," we interpre: :ha: the pressuri er code safety valves are in: ended.
'Je have avamined the nain:enance history of the power operated relief valves
! L: stalled on :he pressurizer a: 3eaver Valley. The following defective opera-
- 1ons are reported:
Data Failure Description 4/26/76 Failed :o Rese Incorrse: Valve Spring Tension (pre-op :sst) 5/02/76 Pre =ature Lif: Defec:1ve Sov 11/16/76 Failed :o Open Defective SOV There have been no naloperations of the code safe:7 valves.
l 'Je have investiga:ed the operating history (challenges) and the following Table represen:s the resul:s of the research effort:
Table of ?CRV Actuations Date Autonatic/ Manual Ocerations and Cause Actuations 06/13/76 Reactor trip / loss of nain feedwater 6 06/15/76 Synchroni:a: ion of uni:/ steam du=p nalfunction 7 06/17/76 Rnergency shutdown /{PT-RC-403] leak 2 06/26/76 Reac:or : rip / feed reg valve failure 2 06/23/76 Reacecr : rip / steam dump =alfune:1on 1 11/05/76 Reactor crip/s:eas dump nalfunction 1 12/16/76 Synchronization / steam du=p =alfunction 4 12/27/76 Reactor : rip /EHC =alfunction 1 04/20/77 Reactor crip/ SIS MSI7 closure 2 05/05/77 Reac:or trip /MSIV flutter 2-
Page 2 2:as II.K.3.3 Recorring SV and RV Failures and Challenges (ccatinued)
Table of ?ORV Actuaticus (continued) ge Automatic / Manual Oeers: ions and Cause Actuaticas 05/10/77 SIS /High steam flow S2S 2 11/16/77 inergency shu:down/ noise in genera:or 1 11/27/77 Synchronisation/EHC/ steam du=p =alfunc:1on 1 01/14/73 Emergency shutdown / loss of CCR to RC? 1A 1 07/28/73 31ackout/nain transforner failure (=anually opened) 6 01/03/79 SIS /reac:or trip /MSI7 closure 1 01/13/79 E=ergency load reduction / SIS /s:e.as du=p nalianc:1on 1 01/2S/79 E=ergency load /redue:1on/ loss of CC2 :o RC? 1C 1 09/20/79 Reactor trip / SIS /14 Inver:er failure 3 11/17/30 Reac:or : rip /5IS/ low steam pressure bistable trip 1 Testing Date Tes: Procedure Actuations 04/24/76 3VT 1.1 - 4.6.3 1 04/25/76 3VT 1.1 - 4.6.12 3 04/26/76 3VT 1.1 - 4.6.12 3 04/27/76 Operations Tes 2 04/2S/76 3VT 1.1 - 4.6.12 1 05/30/76 3VT 1.1 - 9.4.1 3 09/17/76 3VT 1.1 - 10.4.3 1 10/27/76 BVT 1.1 - 9.4.0 2 03/17/77 3VT 1.1 - 9.4.4 1 06/06/77 3VT 1.1 - 9.4.4 1 07/10/77 3VT 1.1 - 9.4.4 1 07/16/77 3VT 1,1 - 9.4.3 2 10/29/80 Acoustic =onitor checkout 6 11/06/30 PORV acous:ic noni:or checks 1 11/19/30 Checkout pressurizer pressure control 1 T0nI.S_
Au:cmatic opera:1ons: 40 Manual operations: 6 Testing: 10 GRAND TOTAL: 75 There have been no challenges :o the pressuriser code safety valves.
, Page 1 i
ITEi II.K.3.17 Report on Cutages of E=ergenev Core Cooling Svste=s The attached Table I lists cutages of ECC equip =en: and ESF equip =en: at 3eaver Valley ?cwer Station beginning w1:h the pre-operational test period
- in early 1976 through Dece=ber, 1980. This
- able lists equipmen: identi-ficaricas, the dates of de cutage f ce uch outage duratica can be determined and lisa the cause of de t._: age (if known) and de corrective action taken. The table is arranged to grcup de cata by system. The' cutages listed have been :aken frc= records of License Event Reports (LER's) and Maintenance 'Jork Requests (5A's) . For our ::acking purposes, de LIR and 5A nu=bers have been included in Table I. Since under 3eaver Valley Uni: 1 Technical Specifications 1: is possible :o re=ove cer:ala pieces of ECCS or IS? equipment f c= service vi:hout entering the Ac:ica S tata=ent in technical specifications, =any of the outages had not been reported ria LIR since reporting was required only if an ac.ica statement was entered.
4 In hose instances vnere an acticu statement was entered, the LZR nu=ber has been noted and an asterisk has been placed by de equipment mark number.
'4 hen using this data for de purpose of structuring :echnical specificacicas or developing outage statistics, one =ust censider dat the extra redundancy designed in:o de planc has a profound influence in Ge real ava11ab111:7 of adequate safeguards and also influences indiree:1y the speed at which equip-nent is returned :o ser rice under condi:1cus where fully redundant systems are available in spi:e of the outage of similar items of equip =ent. It should be noted dat de nu=ber of instances where the action sta:ecent has been entered b.'e to equipment outages is very such s= aller than the
- otal number of instances of cu rges of equip =ent of all categories.
Outages for de purpose of :esting have not been listed in the attached Tar I since :esting is considered to represent culy a ninor portien of
- hs .otal outage -J.:.e and de data is not in a form that lends itself to redue:1cn in:o the reporting for=at used here.
Table II lists the varicus abbrevia:icns and sy=bols used in Table !.
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'A3LI II Ecuic=en: Kev Chaoter 7 - CVC9 SYS~Ei CH-?-1A Eigh Head Charging Pu=p 1A Eigh Head Charging Pu=p 13 i CE-?-D CH-P-lc High Head Charging ?*. p 1C MOV-CH-il5D Mo:or operated valve - volume centrol tank cutiet isola:1cn MOV-CH-115E Motor operated valve - charging pu=o .=uction Chac ter 11 - SAFE ? DUECTICN SYSTE{
1 SI-?-1A Lcw Head Safety Injec:ica Pu=p 1A
, SI-P-13 Law Head Safc:7 Injec:1cn Pu=p 13 I SI-?-1C Lew Head Safe:y Injection Pu p 1C SI-IK-2 3oron Injection Tank
- MOV-SI-8623 Motor operated valve - icw head SI-P-13 suction MOV-SI-8643 Motor opera
- ed valve - 13 lcw head SI to reac:ce coolane cold legs Chapter 13 - CCNTAINMEIT DEPRESSL*RI:LC'CN SYS*Ei QS-?-1A Quench Spray ?u=p 1A QS-?-13 Quench Spray Pu=o 13 RS-?-1A Inside Recire Spray Pu=p 1A RS-P-13 Inside Recire Spray ?u=p 13
- l. RS-P-2A Outside Recire Spray Pu=p 1A 3S-P-23 Outside Recire Spray ?u=p 13 I
Chacter 36 - EMERGENCY DIESEL GE'IERATORS II-EG-1 E=ergency Diesel Genera:or 1 II-EG ' ~ ergency Diesel Genera:Or 2
- Indica:es i:es which exceeded 11:1:ing condition for opera:1on and required entering an ac:ica s:sta=ent in 3V?S Iech Specs.
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Page 1 I:e= III.D.3.3 In-plant Radiation Monitoring A: presen:, 3eaver Valley ?cwer Station has f car: =cun:ed battery or AC cperated Radeco air sa=plers set aside for e=ergency operations and located in a locked cage en de Turbine Deck. The air sa=clers are equipped w1:h holders f:r particulate fil:er papers and charcoal or silver :eolite car-tridges. Ouring an e=ergency, silver :eolice wculd be used :o preclude the buildup and re:en:1:n of ncble gases. This also eli=inates the need f or purging de cartridges prior :o evaluation. Each cart is also equipped vi:h a portable radia ica dose rate meter for reasuring area radia:1cn levels m .:untered by e=ergency personnel and a portable count race ze:er equipped vid an EIC EP-210 probe, for use in easuring and rapidly eval-uating iodine ac:ivig present cu the silver :eoli:e cartridges (reference 37?S-RCi Chapter 3, R.P. 7. 4) . The car: ridges =ay be evaluated at the site of sa=pling or re=cved to an unaffected area for evaluation with any HP-210 equipped ins:rment. ~he cartridges =ay also be fur _her evalua:ed at 3eaver Valley's on-site counting facilig using de =ulti-channel analy:er or sent :o another laboratory facilig for quantita:ive analysis.
In addi:ica, :esting and technical evaluation has been ec=oleted for use of a single channel analyzer (SCA) for iodine activi g deter =inatica on silver :eolice cartridges. 3eaver 7 alley ?cwer Station has procured 4 E!C Stabill:ed Assay Meters (SAM-2's), each in ec=bination with a Oigital Ra:e Mul:1 plier (DRM) and hocked to an EIC RD-22 scintilla:icn probe. 3ecause of the SCA's sensi:1vi:y to environ =en:a1 effects and physical abuse, they will not be ecunted en the car:s w1:h .he air sa_plers, but are accessible in de plant for i==edia:e use during an e=ergency. The RD-22 probes will be =ounted in codified TA LS-6 counting shields to increase sensi:1vi:/ and provide accurate replication of resul:s.
Use of de shiel(s provides an MDA of 0.1 - 0.2 aci (!-131 equiv.) and an MDC, for a 10 f:' air sa= ole, of 4.4 - 3.7*, of 1 MFC (I - 131 egrAv.). The SCA's vill operate only en AC power because of proble=s associa:ed with their use when cperscad on batte n pcwer (i.e., re-setup of de instru=ent prior to each use). The SCA's may also be transported :o any off-si:e location da:
is provided w1:h 110 7AC power for evaluation of iodine activig of de silver :eoli:e cartridges. A 4' x 4' x 4' lead cave in the coun:ing rec =
can be used :o reduce background effects if the above-=entioned councing shields p cve to be inadequate. The abilin to _ove the equip =en: off-si:e resolves any proble=s which sigh: be associated w1:h high background condi-tions arising fro acciden: situations.
We plan :o have procedures available for use of the equip =en: and training cc=cleted for Radiatica Technicians assigned to shif: du g by 1/1/81.
Ecwever Duquesne Ligh: Cc=pany believes tha: :he present HP-210 reched of evalua:ing iodine activig on air sa=ples mee:s de require =ents 1 cesed by NUREG-0737, III.D.3.3, Position 1, and the use of de SCA's will only enhance presen: =enitoring capabilities.
Page 1 1
ITZM III.D.3.4 Control Roce Habitabiliev l
1 We have perfor=ed the evaluation se: forth in Item III.D.3.4 as set for:h I in NRC letter dated May 7,1980 and the following is the data and infor-sation specified as being required in Attachment No.1. "Informa:1on for Cont:cl Room Habitabili:7 Evalua:1on:"
l I:em No. 1 - Control Roce Mode of Coeration See Attach =ent i f cm Chapter 44A of the 3V?S Operating Manual at: ached hereto.
Itam No. 2 - Control Roce Characteristics
! a. The control roe = at: volu=e is 220,000 f:3
- b. There are no separate con :cl room area :enes for nor=al and emergency operation since the control room has its own i ventilation system.
I c. See Figure 44A-1 attached hereto.
- d. The infiltration rate for the control rocs is essentially zero since the control roem is at a positive pressure of 1/3" W.G.
(Refer to Technical Specifica:1cus 3/4.7.7.)
- e. The HEPA and Char: cal absorbers efficiencies for.75-FL-3 and 2 are 99.97% and 95%.
- f. 205 ft. is the closes distance f:cm the control room air intake to the edge of contain=en:. (Refer :o a:: ached drawing il700-RY-13-5, 1 Station Arrangamant.)
- g. See attached drawing ll700-RY-13-5, Station Arrangement. Note scale is 1" - 80' since this drawing is a 50% redue:1on.
4
- h. The control rocm is enclosed by a 2 f:. :encrete wall and ceiling for radiation and =issile protection. There.is a shield door to cover a temocrar/ ope. rating door. Nor=al entrance into the cont:cl
- room is through evo doors at right angles from each other. The control rocs at
- intake is protected by 2' thick concrete structure i (doghouse). There are no other penetrations available for direct
! strea=ing of radiation into the control room. The integrated radiation dose assessments for the Service Building (Con:rol Room) are as follows:
- 40 yr. dose - 2.63 x 10 Rads.
l 4 o month post ,
LCCA dose - 1.08 x 10' Rads .
Total (40 yr + 6 =enthsl i - 3. 71 x 10 ~ Rads .
These assessments vera cade using present pipe break criteria.
i
Fage 2 I:en III.D.3.4 Cen:rol Rcce Habicabili:7 (continued)
- 1. Ccuerol roc = da=per closure ti=e en receiving an isolation signal is 17 seconds plus air seal inflation ti=e (approx 1=ately 10 seconds) .
Refer to Di 20,372 dated 4/4/30. These da=pers are 43" CD butterfly valves with inflatable seats for approximately zero leakage.
- j. Local in-line chlorine detec:crs are installed. Refer to RD-2D, Flow Diagram - Secondary Plant A/C and 0.M. Chapter 44A for addi-tional information.
- k. F.ight self-contained breaching apparatuses are located in the con:rol rec =.
- 1. There is a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> emergency bottled air supply to pressuri:e the control reem af:cr a radia:1cn or chlorine accident. Sere is also an additional 6-8 hour breathing air supply into the centrol IcC3.
- n. ?=erzenev Food Sucolv: Twenty (20) cases courMn'ng 24 =eals (canned) each are ava11abl. onsite.
potable Water Sucolv: The 37?S-EPP indicates that makeup :o the de=estic water storage tank will be isolated during emergencies.
The dc=estic water storage tank holds up to 15,322 gallens.
- n. Centrol rocm is normally staffed with five persons and eight persens in an e=crgency. (Refer to See:1cn 6.0 of the Technical Specifications.)
- o. The potassium iodide drug supply in centrol rocm is 1400 doses of 130 =g KI per dose.
Ice = No. 3 - Onsite Storate of Chlorine and Other Hazardous Che=icals
- a. The folicwing hazardous chemicals are scored ensite.
Chlorine - up :o eigh: 1-ten (2,000 lbs/ ten) containers Am=onia -diree 16-gallon con:ainers Carbon dioxide - one 10-ton and ene 5-ten (2,000 lbs/ con) canks.
- b. Listed belcw are the hazardous chemicals and their closest distance :o the control air intake:
Closes: Distance to the Control Che=ical Roc = (Meters)
Chlorine 15 2 A==cnia 13 1 Carben Dioxide 38
Page 3 I:c.m III.D.3.4 Centrol Room Habi:abilig (continued)
Item No. 4 - Offsite Manufacturing, Storage and Transporta:1on Facilities
! of Harardous Chemicals
- a. Arco-Polycars located 4.5 northeast of the ccuerol roca air intake (Pot:er Tcwnship, PA) is the culy =anuf acturer and storage f acilig of hazardcus che=icals within a five =ile radius of 3V-1. Table No. 1 lists the che=icals and a=ouv.s stored at the Arco-Poly =er facili:7
- b. Table No. 2 lists the hazardous rail :raffic dat passes wi din 648 =eters (2160 f:) of :he control rocm air in:ska. The average container quanti g can be deter =ined by dividing the nu=ber of carloads.
- c. The hazardous barge traffic dat passes wi:hin approxi=ately 1900 f:
of the control rocm air intake will be provided as seen as it is received frcm de Army Corps of F.ngineers,
- d. Le hazardous truck traffie that passes by the centrol recm is censidered negligible by DLC. Arco-Poly =er truck traffic uses FA Rcute 13 and 60 (Expressway). The rail and possibly the barge traffic are the =cre l' ' ting cases for transportation of hazardous
- che=icals by Seaver Valley Uni: No. 1.
Ite: No. 5 - Technical Seecifications Seaver Valley Control Roca has a chlorine detection system and an e=ergency filtration system. Refer to Technical Specifications 3/4 (3-49) and 3/4 (7-16).
Modifica:1cus required :o =eet de habi.abili g require =ents specified by this item are as folicws:
- a. A gas nenitoring system (two =cnitors) using spectrc=eters with a response ti=e less than 10 seconds is required. The list of chemicals : hat need to be =onitored based upon the transportation survey is set forth in Table III attached here:o.
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- b. A new air intake, cc==on to de control reces of both 3V-1 and 3V-2 with two rad 1ation =onitors and the above referenced gas
=eni:oring system shculd be installed.
A schedule for construction of these =odifications _is i=cossible to firmly l deter =ine at this ti=e. Ele =ents of that schedule consist, however, of an apprcximate 2 year lead 1=e for delivery of the gas =eni:oring system =eeting de require =ents of this ite and an additional 4 to 6 =onths installation, checkout and testing ti=e. Cn that basis, a reascnable schedule for satisfying the require =ents of this ite= appears to be July,1983.
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a _a _. _. 2..g_, ,a . . _. .a.. .a ._u_. . =a_a . s t.,,,,,. _, 7
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a c r..a..a - s e.a. . ::. .a,.
..a .
. .c .
vo:x. 2.ac: :: ,.a sets ::c:2i: =a : .,_:w :g equi;=en::
- 1. Manual isola i:= da=per
.. c .. ._ 2. . . a. _
- 3. 5:02 : :pe:2:4d s= cite is=per
. Met == :pe:2:4d firs e=e:gency bypass da=pers le :::e a and 5 w -mi isola:ie d2=per anci bvpass da=per a:s c:mmon.
Izhaus: air f::m :: es 1 -1:: ugh i is d 2wn is.: i dir. a dual ::ce ** ,us: s owns::sa: :i ducts, ead of whid ::: aiss a so:c: c. ers:ed s=oka .ame. er. the : ce i exhaus: duct s=oka da=per is a single he:4d :sgista: whid exhaus s -le Coc:::1 Area 7entilation Equi;=en: ?.cc= (Sce 5) . '"his single exhaus :sgister is p :v.ded vi:h a :occ: Ope:2:ed s=oke daspe . All of
.u
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._...u". ".. .y e _ -2 .d_.. g . . _- a __ 3.2 d
[ '.7.* 7. -40 A ] .d . ' e.
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. u s e - .o ,__a g .: a.., s , .., g.
- uscarpar (.175-E-la3] allev -la ::=p:sssor-condenser uni:s .c be dow- fe:
Air disch2:gsd
=a : -' ce while still =ai: a4-d g adequa:4 :sdn=dssc7 f::= -d e s::: air fan (175 7-40A] passes .h:: ugh i s :sspective :stu:=
air is: discharge da=per (da=cer is : pen vb-- f2: is Ope:2:=g) and is a..d ._" e _- ~scha.g=.d. ...m d3 '
.e s f s ..=.: ... 2 -esr"..--.
u... a _ow.-. .-,e_-'..d. . exhaus: it=per asi .he series 2: seged =c::: Ope:2:md cu:dec: exhaus u.u._. ..._ .:, .,. . a_, yes r. ,.m.3_ca v _ ,.c- a-s - .-
,.d :- . .as .... a. .., s. .s. u.a e w. .3 ....
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e .~. =... . 1c, a w- _ . , =.1 c: , .c e.. .- . s ,c. .T , . .. e.s..a u,,a 2_. . ..g a. aa=pe s. . ...o =a. , . , 4._f .,. res
,, e _.2 .. ..s ..c_
a . 2 . a.- . _ r . .. are p:c rided downst:sa: f -de red = dss: :scimenla:ica dampers.
.a u.,ga .,.. . . . n, ges a. .
.x. .- . . a ,. .g 1 g ga.u.n_a.sa.a..1.a n a.3 n . 2,.,7 ,,..,a.a.e
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c:. .. :..- e y a .ua. - - - --
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ful'. :apaci:7, 0 ::c f:eo: c cpressor-cendenser units (175-E-/.A :: a3i
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'~
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- ircula = g pu=ps (175-?-3A and 33] are p: vided and ve d : '::c,iucti: ,
2.."4 - " =-_' - a s s a c ' a w.= s^.
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s andby vill pu=pdevr. eve:7 30 '-":es :n suction p::ssurt. hese :vo pu=pdcwn -" ' g relays a:s installed 2: :5.e be:::m of ::e 3S?. An add. i cal ::n::01 fea .n=e arev.ded . c : sis.2 :s ".e : spec-ive su ply /-v-'ust
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- cid ': ave as adverse effec : .se ::c::el ::cm ambie:: air. his bo:: led supply is es.g=ed d to release 500 sci:s of air di se:17 :a :he cen::cl
-e 4, .W 4a- e. , .a. .
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-i=ers vi'.1 s ar: ::c::el ::cm e=ergenc7 supply fans U.75 r -41.i and 113]
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sucnon sys em, or a pist:n d:.ven 3:.: ::mpresso r .s] v::.:: -axes t.< w ., ::,,... w es.,*
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- ange ce ;csi i:cs ei ne suppli air ds=per, ---'us: 21: ia=p er , and ':c e
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s1=.. .,... .:2.a.. -.. se da. c. e .-s . "=z. ..- . . <.._- .- ' 2 _ _ . ..-. 2 ..= . #. . . " -. sys em vc:.le also ::=pensa:ing i:: ce danged cu: side 21: suppl 7 and sanaus: d'.:w :s:es. ':ur :g se sr~er seascc, lici:s 2:e =pesed :
-se amcu .. .c. . . u. .s_4 .;.e a_4 _. .- . 5 4 .w 4.s 4.... . c.c s 4 ...: .
.w. . s y s.,. . . sese e._....
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. ... , . .v = . ".e -va...2.-add
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- mpresser-c cde=se :i: wci:.t wc ua, .a. :ssu : =
c:c 4. .:g
- n en:
tempers ::s condi.i ns abcvs ,.e .aesig=, va.3 ua :. ::: anc pcssure everses:ing ed -le :-~' g air c::dizio"- g ::= presser-c cde=se: ::. .. Cpersti:cs accve -l e "Nc =:21 Candi:d,0cs ...""a;a.o-). ." a:s permi.:ed dur :g
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f ur.. a
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le..: . A1.1 doc s is:: the C c::cl A:32 2:s fi ned vd.-d wea =erseal :7;e ;ssle:s :: id+: -le on: leakage of air i::m =e a:=2, ,.: s *d d 4 : g -l e :aseep air
- squireme :s f:: ;:sssur_:2.i::.
A empo:2:.-r vall espable of wi-us s g :::::el ::cs p:sssuri:2:isc p:sss t:s is iss:.alled be:ves: Uni: 1 and Uni: 2 ::c::al ::ces. j e.as t 3a.-. n e .. .. ,.. . 4. .. . ., .e 7s. .-
-'c
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-l e s:::2ge a:sa, au ma ic de:ectics, isclatics and pressurizan:c we sd a-la
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4
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~%:se seis=i:2117 . qualified de*.ect::s (,.~.A ~S-101.A,3,Cl a:s.neins.slied i
^
serv . s
-te<ent ,.a u:n qu:.pme . eleva nac
,eem s a: de iu :.
3 u,4._,.24 . g 4
.w e . 4., ._e :7 ,e . .a.e s e...s . 4 .. ae 42=p 2 e .. -.. . w . .. ..,t. ...... ..cm air intake.
~
- .ac d e:ect:: has i s sv: sensi:g lire oc :te downs:: sam side ci (~S-0 1A ,131 de is 2ke da=pe.:s . The sampled air is d:2wn -2:: ugh 4
. s.a_. e.s,u .t ..i .we,. . ,..;..,
- -a d.'scea g=.A :..- ue.". 4. . .- .- . s ee . .47=- --- " ' a _.. - .
detects: and : star:ed :: -de ai: 12:ake duc 2. a pois: ice.cs::eam sd -le sample pcia . a " '. g.".
'^..=".=e.a.s,
"~,c c ^ e " = c ...' c . o .: - a .. .. . o .# .
.ie ..t.3
.. . w4 .
w
.... ac.e. s-" ,', .-.
- c. ' *. c- c _=.. . .1 - '. c ' .
- c u'.
~he cu: side Air ::2ke Dartpers (75-D-40-1A,13] receive :1:se sig:21s.
g.e... .w. ds=.e.-s ..ac" ."a- ..oseA. 'L _' a d . swd.. ".-s, ' sc ~.d..d va .i v.s
' i i ... . - . .
- 21. res4 .u.e --,se.s
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.a y:_. u azz.,. . a. a
.a
. sca.: . s.
,cs"..'- . .. ... ='..imi..
. 1..akag*.
~he cu: side Air x aus: Dampers (75-D .'.0-10-10] receive :1:se s'.;: sis and are sealed in a similar :2: e: as described above. 00c:::1 .icem r- ..e.. .g . . . , .: t r.,.., . 2 s t7 r.c ..g , s t..s. 'i ..
....... .. . e s i. . a .' s .
_4 .J 4* .
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Attac".en: 1
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. v e- n.ts.
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_ u -. ,. . . ..N ( .... _. . _. ._. e. 3, we. bl lq>S4 . .- lTL n sc., e c:.d valves (s....u -v S .m,..,.s
. . .: ....,m,. .:2=
. . .s. . . . , as.a. (:u. .<. -
,..n.en , s.e , ,-.... .
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4
'.,"'3 2.-. .....-,__-A :; ... +. ._" #__' ', s u--ry .,
. s ::,4c. aa : ;:swas : ..e ;:ss surs ::::::., vs.ves t r,.< v: .,., A :. ,2
- s
",.ese. ..valves v 11 --- =cdu12.e : =a n:21: app::x=a el2 50 .si-4
- se :: :w= g:
3ct:ied air suppl 7 ::i; val es (~7-175-101.L3,C,J,Il are :pened
. 3 _a . ., ..s s ur . s".. , ' _' .ad. _~- .u,_
d
,y,y
.. . .e 'e-ene.3.__-d s o .' .=...- d. .'.
valves (307-175-101.b3,0,J,I]. -.ese solencid valves :smais de-energ_:ed ;;:v. ding ;;sssure as :easured 57 ; sssure switches (?S-175-105A; :_2] :ssains bel w 200 psig. If ; sssure ecceeds
._w .., , ,. o ,-s-g, a . _ . . s ., _.. : .; 2.s .j_t ....,-_..
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;:st.=sa i: su .:=es (I?-175-10 .-l.1,13] vai= cver-ide -se :anual posi:_:e- g sv.::1 and alse 01:se .he er.:2us . 5412 :=g ds=;er.
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39 7 iS 229 ti t ul ata0 A 2 49164u1 C AL C lutteetti AL 11 9%1bguS C LS t utt fit i AL iS 229 HIOLAhu A 1 fS 249 til0LA40 A 9 195 - - - 4916%uu C At C lutt LAnuluE 6 311
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+91d#57 Soulun sellHIIE UN 229 tiluL At*8) A S 200
, 49tulb4 S IHule l lutt til]HAIE OH 249 Hlut.AhD A 16 79S
$91b850 lihC Ahhutelutt telldilE Ort 229 pl ut At40 A 1 - . . . - - - - - . 9 0 --- --
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L = LUCAL. ua uHlutt Al t hes e I = l t util n a l ise6 e Ar.0 A a IHou __ et Al . tulL. IYPL uf 10 45 SICC seuhiit i, DLSCulPilote Of CuttHuully Cout ftut tsuuf #4AttE IHAf f lC "- " CAHLu ADS = 2/9 Hjul Afsu A 12 101 t'h 49214 b ) out.AlelC PhuSPHAll HlAluHLeuHGAtelC PHuS= /gy H l ut. Alep A 1 39 - -- - - s' AH Als l ittu Ate tL ilat , Sully Ps 4 9 214 ts i 95 delu9 Pu t StateuuS L luulu ele.u.S. lPolSuhuuS l'u 2/9 enlul AteD A bl5
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v.. - - . w .v. l S .,- H.UARDOUS CHIMICAIS TO BE MCNITORED T Anhydrous .4tmenia .-
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- - . 1. .
~
-; . 2. Butadiene Inhibited -
- 3. Vinyl Chloride
~ '
- 4. Chlorine
- 3. Ethylene oxide
- 6. Carben Dicxide
- 7. Ethyl Acrylate Inhibited
- 3. Acetone
- 9. Toluene
. . . 10. Hydrofluoric. Acid ( Anhydrous)
. . ~ . . ..
- 11. Methyl Methacrylate-Menemer Inhibited
. . .a.. ... -
. . . .. .12. Stvrene Mcncmer Inhibited
- 13. Petroleum Naptha .
- 14. Hv. drogen Peroxide Sclution
- 15. Ca5bolic Acid (Phenol) ,. .
' % r~- 16. Phosphorous Oxychloride
- 17. Srcmine .
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- 19. Dimethyl Ither
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- 20. P ropylene ' Cxide
- Cces nct include barc.e traffic.d.ata.
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l i l WIS"'NG40LSi CMSS 3 l l t i
SUMMARY
REPORT WESTINGHOUSE REACTOR VESSEL LE'IEL INSTRUMENTATION SYSTEM FCR MONITORING INADEQUATE CORE COOLING (MICROPROCESSOR SYSTEM) l '0ecemoer 1980 i t 1
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TABLE OF CONTENTS 1.0 INTRCDUCTION 1.1 NRC Requirements 1.2 Definition of ICC 1.3 Condition or Events Which Describe the Approacn to ICC 2.0 FUNCTIONAL REQUIREMENTS 2.1 Parameter Critical to ICC 2.2 Instrumentation Accuracies, Ranges, and Time Response 2.3 Qualification Requirenents 2.4 Codes and Standards 3.0 ICC INSTRUMENTATION IDENTIFICATION 4.0 RVLIS - SYSTEM DESCRIPTION 4.1 General Description 4.2 Detailed System Description 4.2.1 Hardware Description 4.2.1.1 Differential Pressure Measurements 4.2.1.2 System Layout 4.2.2 Microprocessor System 4.2.2.1 Inputs 4.2.2.2 Density Compensation System 4.2.2.3 Plant Operator Interface and Displays 4.2.2.3.1 Display Functions for Remote Control Board 4.2.3 Resistance Temperature Detectors 4.2.4 RVLIS Valves 4.2.5 Transmitters, Hydraulic Isolators, and Sensors 4.3 Test Programs 4.3.1 Forest Hills 4.3.2 Semiscale Tests 4.3.3 Plant Startup Calibration 7583A MP
_ .-_. ~ - .- .. - . . . .- . _. . .. . . _ ] i f i i i TABLE OF CCNTENTS (Continued) j 4.4 Operating Performance ! 4.5 RVLIS Analysis 4.5.1 Transients Investigated 4.5.2 Observations of the Stuoy
'4.5.3 Conclusions l
5.0 GUIDELINES FOR THE USE OF ICC INSTRUMENTATION 5.1 Reference Westinghouse Owners Group Procedure 5.2. Sample Transient
6.0 REFERENCES
b i l' f I i f f
- 7i324 .MP
- _ - _ , . _. - - - ~ . . . . . , . . , . ~ . - - m...-. _ ,_. ,,-.. . _ . . . , - = . . , - - - -
1 l l LIST OF TABLE 3 71 ole 3.1 Information Requirea on :ne Core Succooling Monitor Tacle 4.1 Canaliance witn Regulatcry Guice 1.97 Orsft 2, Rev. 2 6/4/S0 Tacle 4.2 Transients Investigated 75alA 1
l LIST OF FIGURES Figure 1-1 Reactor Vessei Level Instrument System Figure 1-2 Process Connection Senematic, Train A Figure 4-3 Typical Plant Arrangement for RVLIS Figure 1 4 Reactor Vessel Level Instrument System 31ock Diagram (One Set of Two Redundant Resets Shown) 1 Figure 4-5 Remote Display Module (Control Board) Figure 4-5a Vessel Level Summary Display Figure 4-5b Vessel Level Trend Display Figure 4-6 Typical Plant Arrangement for RVLIS Figure 4-7 Block Diagram of Compensation Function Figure 4-7a Simplified Schematic of Density Compensation System Figure 4-8 Surface Type Clamp-On Resistance Temperature Cetector Figure 4-9 HEL3 Simulation Profile Figure 1-10 ITT Sarton Hydraulic Isolator Figure 4-11 ITT Barton "High Volume" Sensor Sellows Check Valve Figure 4-12 Case A 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip with l Reactor Trip, RYLIS Reading and Vessel Mixture Level 7683A MP e v - , ,r r w r p , ., +w,n , , , , ow-x-- m - . - - - ,-. + r --, n - - ' -,r--
l Figure 4-13 Case A 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip with Reactor Trip, Void Fraction Figure 4-14 Case B 3-Loop Plant, 3 Inch Cold Leg 3reak, Pump Trip at 750 Sec:ncs, Wide Range Reading i Figure 4-15 Case 3 3-Leop Plant, 3 Inch Cold Leg Break, Pump Trip at 750 Seconds, RVLIS Reading and Mixture Level Figure 4-16 Case 3 3-Loop Plant, 3 Inch Cold Leg Break, Pumo Trip at 750 Seconds, Void Fraction. Figure 4-17 Case B 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip at 750 Seconds, Cold Leg Mass Flowrate (L3/Sec) Figure 4-18 Case C 2.5 Incn Pressurizer Break, No. SI, RVLIS Reading and Mixture Level. Figure 4-19 Case C 2.5 Inch Pressurizer 3reak, No. SI, Void Fraction Figure 4-20 Case 01 Incn Cold Leg Break, ICC Case, RVLIS Reading and Mixture Level. Figure 4-21 Case 01 Inch Cold Leg Break, ICC Case, Mixture Level, l RVLIS Reading and Measured Inventory. Figure 1-22 Case 01 Inch Cold Leg 3reak, ICC Case, RVLIS Reading and Mixture Level. ! Figure 4-23 Case 0 1 Inch Cold Leg Break, ICC Case, Void Fraction i 7683A MP
1.0- INTROBSCTION 1.1 13C 4E0BIRE.*ENTS The NRC nas estaclisned recuirements (items I.C.1 and II.F.2 of NUREG-0737, ' Clarification of TMI Action Alan Requirements') to provide the reacter operater with instrumentation, procedures, 2nd training neces-sary to readily recogni:e and implement actions to correct or avoid conditions of inadequate core cooling (ICC). I i Uncer certain plant accident conditions, the potential exists for the l fonnation of voids in the reactor coolant system (RCS). Under these conditions, it would be advantagecus for the reactor operator to monitor the nater level in the reactor vessel or the approximate void content curing forced circulation conditions in arcer to assist him in subse-cuent actions. Therefore, a reactor vessel level instrumentation system (RVLIS) has been incorporated to provide readings of vessel level which can be used by the operator. Vessel level as measured by the RVLIS is the collapsed liquid level in the vessel. The RVLIS provices a relatively simple and straight-forward means to
- monitor the vessel level. This instrumentation system neither replaces l
any existing system nor couples with any safety system; however, it does act to provide additional information to the operator during accident conditions. The RVLIS utilizes differ'ential pressure (d/p) measuring devices to indicate relative vessel level or relative void content of the circulating primary coolant system fluid. 1.2 BEFINITION-GF ICE ICC as cefined in References 1 and 2, is a high temocrature condition in tne core such that operator action is required to cool the ccre before damage occurs. l 1-! 755'a
i i a t
'. 3 CONDITIONS CR EVENTS WHICH CESCRISE THE APCROACH TO ICC f
i ?!e w micus f silure that would lead to ICC curing 4 sma11-break
' 0CA, althougn nignly unrealistic since multiple f ailures are requireo, is tne loss of all hign pressure safety injection. The accroach to ICC
- ncitions anc :ne analyses for this event sequence are proviced in l References 1 anc 2.
i r i f } l i l l t i 4 s l l I j 1-2 i l i
- - , . . . - . _r,.y ., , e,,-,--,,,,.m , . . . . , . , _ , , , , , , , . . , . . . . , , _ , , , , , , , . . _ , . , . . , _ . . . . . . _ . , , , , , , . . . ~ . . , _ . , .._...,...r.,,..-,. ,- . . ,., _ . . _
2:0 FUNCTIONAt REOUIREMENTS, 2.1 84RAMETIRS CRITICAL TO !EC
. The analysis proviced in References 1 and 2 delineates those parameters critical for the detection of and the necessary mitigation actions for the recovery frem an ICC condition.
To triefly summarize those parameters ICC is detected by either high core exit thermoccuole temperatures or by a low reactor vessel level indication (core uncovery) in conjunction with core exit thermoccuole indic ation s. Mitigation actions consist of depressurizing the reactor coolant system (RCS) to permit injeetion of accumulator water and/or to estaolish low head safety injection flow. The RCS is itself depressurized by depressurizing the steam generator recondary side. Critical parameters at this point are steam generator pressures and wide range RCS loop temperatures. Once low head safety injection flow is established, transfer out of the ICC procedure can be made when core exit thermoccuole temoeratures are reduced and the reactor vessel level l ! gauge indicates a level above the too of the core. With the exception of reactor vessel level, all parameters are monitored by currently existing instrumentation. 2.2 INSTRUMENTATION ACCERACIES; RAN6ES; ANS TIME RESPONSE Accursey An accuracy of 6 percent is required on all three types of reactor vessel level instruments. This should be a statistical ecmcination of all uncertainties including those due to environmental effects (if any) ! on instrumentation. For the upper range instrument, this corresponds to in allowable ceviation of about + 1 foot elevation head. This will give the operator a good estimate of the steam or gas volume in the uocer nead during a situation in anien the head vent aculd ae employed. For the narrew range instrument this corresponds to an allowaole deviation 2-1 7531A -
I l Of accut ; 2.5 feet elevation head. This is recuired to: 1) =covide ! acecuate margin against inacvertant use of the ICC ope-sting guideline l (EI CI-1, see Section 5.1), 2) assure that the vessel level reading can
- e reascnaoly used to sia in :ne detecticn of the onset of ICC conci-tiens, 3) :erive useful information reguarding vessel level benavice
- uring tne vessel refill : erica of a LOCA transient.
Rance The vice range instrument will cover the full range of ex:ected differ-ential pressures with all reactor coolant sumos running. The maximum span of the wide range htstrument will change with the numcer of pumos operating. The coerator must be aware of the maximum scan for a given number of ccerating pumos. Both the narrow range and the upper range instrument indications should be set to indicate that the vessel is full with the pumos tripped. Time Rescense The d/p instrument response time shall not exceed 10 seconds. This time delay is defined as the time required for the display instrument to reach the micooint of a 50 percent step input d/c change. 2.3 CUALIFICATION RECUIREMENTS Environmer. cal qualification of the RVLIS shall verify that the system i ecuipment will meet, on a continuing basis, the performance recuirements cetermined to be necessary for achieving the system recuirements as presented acove. Verification must include confirmation that those portions of RVLIS equipment whicn are within the containment will oper-ate curing and subsequent to the conditions and events fer which the system is recuired to be ocerational. Verification will include deter-mination that the system is sufficiently accurate during this time to meet its cesign basis. The system post-accident environment cualified
'ife ecuirement fer electrical equipment inside containment is 120 days 2-2
f ollowing certain postulatec events. The electrical equipment that is it allec utstce of containment neea not aeet a 4ualifiea life for m extencea perioc of time providing replacement or calibration Onecks can se mace in snort enougn time connensurate 4ttn tne reliaoility goals of tne recuncant system. For tne resistance temoerature detectors (RTDs) envircnmental recuiremelts for serv 1ce witnin the containment, refer to l 5ection 4.2.3. Electrical equipment insica containment snall se instal-led sucn that it is renoved from areas were hign energy pipe creaks or pipe anio could cause failure. The d/p transmitters and electronic l processing equionant srall be locatec in a low amoient rac14 tion area. The RVLIS sensing transmitters and associated electronic processing equipment snall :e located in an arta wnose temperature ru1ge is netween a0 and 120*F with 0 to 95 r? cent amoient relative numidity. Normal operating environment for transmitter locations snall se cetween 60 and 80*F and 0 to 50 percent relative humidity. The instrumentation snall me qualified to assure tnat it continues to operate and reaa witnin ene remaired accuracy following but no't necessarily curing a safe u1utdown 2arthquak e. Qualification of the electronic equipment and reactor ves-sel level sensing transmitters applies to and includes the cnannel' iso. lation device or anere interface witn a computer is involved, the input suffer. The location of the electronic isolation cevice or input buffer snould ce sucn :nat it is accessible for maintenance curing accicent conditions. 2.4 CODES AND STANDARDS The RVLIS is in conformance witn tne following Coces and Stancaros: Regulations GDC 1 quality Standards and Recorcs GDC 2 Design Bases for Protection Against Natural Phenomena GDC 4 Environmental and Missile Design Bases 2-3 l l I emme = i
I 3CC 13 Instrumentation anc Control GCC 15 Containment 3esign 3CC 13 Inspection :na Testing of Electric Pcwer Systems 3CC 19 Control Room 3CC 24 Seoarction of Protection and Control Systems GCC 30 Cuality of Reactor "colant Pressure Bouncary GCC 31 Fracture Prevention of Reactor Coolant Pressure 3cuncary GCC 37. Inspection of Reacter Coolant Pressure 3cuncary GDC 50 Containment Cesign Basis GDC 55 Reactor Coolant Pressure Soundary Penetrating Containment GCC 56 Primary Containment Isolation 10CFR50, Accendix 3, "Cuality Assurance Criteria for Nuclear Power 81 ants anc Fuel Reoracessing Plants" Incostry Standards IEEE-308-1971, "IEEE Standard Criteria fer Class IE Electric Systems fer Nuclear Power Generating Stations" IEEE-323-1971, "IEEE Trial-Use Standard: General Guide for Qualifying Class 1 Electric Equiement for Nuclear Power Generating Statiens"* IEEE-338-1971, "IEEE Standard Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems" IEEE-344-1971, " Guide for Seismic Qualification of Class 1E Equio'nent for Nuclear Power Generating Stations"'* IEEE-321-1977, aIEEE Standard Criteria for Incecendence of Class IE ! Ecuioment anc Circuits" AEFE 3PVC, Section III, Class 2 Nuclear Pcwer Plant Ccmconents ( l ' " For certain specific clants, IEEE-323-1974 is apolicable.
-- For certain specific clants, IEEE-344-1975 is acolicaole.
I 2-4
4 NSI 331.1.0, 1967 including sedenca througn and including 6/30/71,
*0:ce 'cr Pressure Picing", including nuclear c0ce cases wnere acclic3 Die Regulatory Suices t
l 4.3. 1.11 Instrument Lines Penetrating Primary Reacter Containment 4.G. 1.22 Deriodic Testing of Protection System Actuation Functions A.G. 1.75 Physical Independence of Electric Systems , [ . l I i
^
2-5
'921)
!--- -_. - , . - - , - . . . . . - _ . - . . . - . _ _ - . - . . . . . _ . , .- , . - . - _ . = _ . _ , . _ - _ .
3.0 ICC INSTRUMENTAT:CN IDENTIFICATION Acecuate instrutnentation is necessary to ciagnose tne approacn to ICC inc o cetermine tne effectiveness of :ne mitigation actions taxen. Curing :ne ::recaration f ne ICC operating instructions, consiceration
<as given .o tne scecuacy of curren: trstrumentation Ina :ne cenefits
- erivaDie from :ne accition of new ins trumentation. The folicwing is a list of existing instrumentation consicered (refer to tne FSAR for 1etails) anc conclusions derived:
- 1. Current Instrumentation
- a. WIDE RANGE REACTCR CCOLANT PRESSURE - present instrumentation is availaole for cetermining general RCS pressure trenos curing tne ICC event. The expectec accuracy following ICC events is such that inis instrument cannot ce used for precise determina-tions of tne pressure required to assure onset of low nead safety injection flow to the RCS.
- b. PRESSURIZER PRESSURE AND LEVEL - conditions in the pressurizer will generally lie outside tne ranges of these instruments during an ICC event in a Westingnouse PWR. Pressurizer pres-sure and level are not used for determining mitigation actions to be taken during ICC.
- c. AUXILIARY FEEDWATER F1.0W - present instrumentation is availaole for assuring ne sufficiency of maxeup water ficw to.One steam generators curing an ICC event.
- d. WIDE ANGE RESISTANCE TEMPERATURE DETECTORS - present instru-mentation is availaole in determining trenas of recovery actions but may not ce availaDie in cetermining :ne onset of ICC conditions for all creax sizes.
- e. CORE E.(IT T'r4ERMCCOUPL~S - present instrumentation is availaole in cetermining cotn :ne exisance of ICC and :ne trends of recovery acticns.
3-1 d
l l 1
- f. CCRE SUBCCOLING - does not provide useable infonnation during an ICC conditien. Will indicate superheat conoitions in core coolant. Will help indicate the approach to ICC by showiag saturation conditions. Since the core subcooling monitors may not be described in the FSAR, refer to Table 3.1 for informa-tion.
- g. STEAMLINE PRESSURE - present instrumentation is available for determining heat sink availability and heat removal capability during ICC mitigation actions.
- h. STEAM GENERATOR LEVEL - present instrumentation is available for determining the availability of a heat sink for the RCS during an ICC condition.
- 2. New Instrumentation
- a. REACTCR VESSEL LEVEL - provides an indication of the approach to ICC and confirms the achievement of adequate core cooling when level in the reactor vessel is restored.
To sunnarize the above considerations, current plant instrumentation is adequate to determine heat sink availability, to detect the onset of ! ICC, and to detect the effectiveness of mitigation actions following the onset of an ICC event. The RVLIS is provided to permit a more continuous indication of the approach to ICC. . 3-2 7531:
- f. CCRE SUSCCOLING - does not provide useacle information curing an ICC concition. Will indicate superneat concitions in core
- colant. Will help incicate the accroacn *o ICC by snowing saturition conditions. Since the core succooling monitors may not :e described in the .:5AR, refer to Table 3.1 for informa-tion.
- g. STEAMLINE' PRESSURE - present instrumentation is available for determining heat sink availabilty and heat removal capability curing ICC mitigation actions.
- h. STEAM GENERATCR LEVEL - present instrurentation is available for determining the availability of a heat sink for the RCS curing an ICC condition.
- 2. New !nstr: mentation
- a. REACTOR VESSEL LEVEL - provides an indication of the approach to ICC and confirms the acnievement of acecuate core cooling when level in the reactor vessel is restored.
To sumnarize the aerve considerations, current plant instrumentation is adequate to determine heat sink availability, to detect the onset of l ICC, and to detr.ct the effectiveness of mitigation actions following the onset of an ICC event. The RVLIS is provided to permit a more continuous indication of the sacroach to ICC. I 1 l 1 l l i 32 7521A 1
_ _ _ - _ . = - - - - . _ - _- . TABLE 3.1 INFCRMAT:CN REQUIRED CN THE CCRE SUSCCOLING .MONITCR 31solay Infomation Disclayed (T-Tsat, Tsat, P-Psat succooled press,etc.) 7-Tsat superneat 01 splay Type (analog, cigital, CRT) Analog and digital Continuous or on Demand Analog - continuous Digital - on comand Single or Recundant 01 splay Redundant Location of Display User supplied Al ams Caution - 258F subccoled for RTD Alam - O'F suecooled (include setpoints) 15'F succooled for T/C for RTD and T/C l Overall Uncertainty (:F, psi) Digital - 4eF for T/C; 3'F for RTO Analog - 5'F for T/C; 5'F for RTD Range of Calibrated region - 1000 psi subcooled to 2000'F superneat 01 splay Overall - never offscale Qualifications None at present* Calculator Type (process computer, dedicated digital Oedicated digital or analog calc.) If process computer is used, specify availacility N/A (percent of time) Single or Redundant Calculators Redundant , Selection Logic (highest T., lowest press) Highest T for RTD or T/C; Lowest ? Qualifications None at present ! Calculational Technique (steam tables, Functional fit - functional fit, ranges) ancient to critical point
- The display is currently uncergoing seismic qualification testing cy Westingnouse wien will confom to IEEE-344-1971. This infomation will only ::e proviced at the s;:ecific recuest of One custmer and af ter accropriate installation enecks have ceen made to verify One acclicamility of tnis qualificatien.
i 2-3
,,,gr-~<e- , y-,,- r- - - - -- .p,, y~ - -- ,.:, ---i _- - , - . , -
TM Li 2,5 (Cantinued) Incut RTD, T/C and Tref Temcerature (RIDS er T/Cs) RTD - 2 not ind 2 cold Temeerature (numcer of sensces and locations) leg per cnannel l T/C - 8 per enannel Range of Temoersture Senscrs RTD 700 F T/C 1550*F 1 (calibration unit ' range 0 - 2300*F) Uncertainty
- of Tencerature Senscrs ( F at 12) User supplied User supplied Qualifications Pressure (specify instrumerm used) User supplied 2 wide range - Loco Pressure (numcer of sensors and locations) L narrow range -
Pressurizer Wide range 3000 psi Range of Pressure Sensors Narrow range - 1700 - 2500 psi Uncertainty ** of Pressure Senscrs (psi at 13) User supplied user supplied Qualifications Backuo Cacability Availantlity of Temo and Press Availability of Steam Taeles etc. Precedures
** Uncertainties must address conditions of forced flew and natural circul ation 3-4
4.0 REAC~CR VESSEL LEVEL INS RUMENTATON SYS EM - SYSTEM DESGIPTION 4.1 3ENERAL DESGIcTION ie eactor vessel level instrumentation system (RVLIS) uses oifferen-tial pressure ( /p) .easuring cevices to measure vessel level cr rela-tive void content of the circulating primary coolant system fluid. The system is redundant and incluces automatic compensation for potential temcerature variations of the imoulse lines. Essential information is displayed in the main control room in a fcrm directly useaDie by the oper ator. The functions performed by the RVLIS are:
- 1. Assist in cetecting the presence of a gas buccle or void in the reactor vessel
- 2. Assist in detecting the approacn to ICC
- 3. Indicate the formation of a void in the RCS during forced flow conditions.
4.2 OETAILED SYSTEM CESGIPTION 4.2.1 HARDWARE DESGIFTION 4.2.1.1 Differential pressure Measurements The RVLIS (Figure 4-1) utilizes :no sets of tnree d/p cells. These cells measure the pressure crop from the sottom of the reactor vessel to the top of the vessel, and from the hot legs to tne top of the vessel. This d/p measuring system utilizes cells of differing ranges to cover ! different ficw Denaviors with ano without pump oCerstion as discussed celow: ,
)
l 1 4-1
- 1. Reactor Vessel - Uccer Range (aP3 )
The clo cell AP snown in Figure 1-1 provices a measurement of 3 reactor vessel level accve tne not leg pice unen the reactor cool-ant pumo (RCP) in tne loco witn the not leg connection is not operating.
- 2. Reactor vessel - Nar-ow Range (APb)
This measurement provides an indication of reactor vessel level from the bottom of the reactor vessel to the top of the reactor during natural circulation conditions.
- 3. Reactor Vesre Wice Range (APe )
This instrument provices an indication of reactor core and inter-nals pressure drop for any comoination of operating RCPs. Can-parison of the measured pressure drop with the normal, single-pnase pressure drop will provide in approximate indication of the relative void content or censity of the circulating fluid. This instrunent will monitor coolant conditions on a continuing basis during forced flow conditions. To provide the required accuracy for level measurement, temperature measurements of the impulse lines are provided. These measurements, togetner with tne existing reactor coolant temperature measurements and vice range RCS pressure, are enoloyed to compensate the d/p transmitter cutputs for differences in system density and reference leg density, particularly during the cnange in the environment inside the containment structure folicwing an accident. The d/p cells are located outside of the containment to eliminate the large reduction (approximately 15 percent) of measurement accuracy asso- - ciated with the cnange in the containment environment (temoerature, ressure, radiation) during an accident. The cells are also located outsice of containment so that system operation including calibration, cell replacement, reference leg cnecks, and filling is made easier. 1-2 l
d.2.1.2 Svstem Lsvout A schematic of the system layout for the RVLIS is shown in Figure 4-2. There are four RCS penetrations for the cell reference lines; one reac-tar head connection at a scare penetration near the center of the head or the reactor vessel head vent pipe, one connection to an incore instrument conduit at the seal table, and connections into the side of two RCS hot leg pipes. The pressure sensing lines extending from the RCS penetrations will be a canbination of 3/4 inch Schedule 160 piping and 3/8 inch tubing and will include a 3/4 inch manual isolation valve as descrioed in Section 4.2.4 These lines connect to six sealed capillary impulse lines (two at the reactor head, two at the seal table and one at each hot leg) which transmit the pressure measurements to the d/p transmitters located outside the containment building. The capillary impulse lines are sealed at the RCS end with a sensor bellows which serves as a hydraulic coupling for the pressure measurement. The impulse lines extend from the sensor bellows through the containment wall to hydraulic isolators, which also provide hydraulic coupling as well as a seal and isolation of the lines. The capillary tubing extends fran the hydraulic isolators to the d/p transmitters, where instrument valves are provided for isolation and bypass. Figure 4-3 is an elevation plan of a typical plant showing the routing of the impulse lines The impulse lines fran the vessel head connection must be rcuted upward out of the refueliny canal to the operating deck, then radially toward the seal table and tien to the containment penetra-tien. The connection to the bottom of tne reactor vessel is made through an incere detector conduit which is tapped with a T connection at the seal table. The impulse line from this connection is routed axially and radially to join with the head connection line in reuting to the penetrations. Similarly, the hot leg connection impulse lines are routec toward the seal table / penetration routing of the other two con-necticns.
- 4-3 7551A
l i The imoulse lines located insice the containment ouilding will oe exacsed to the containment temperature increase during a LOCA or HEL3. Since the vertical runs of impulse lines form the reference leg for the c/; measurement, the change is density due to tne accicent temperature cnange must be ta<en into account in the vessel level ceterminatien. Therefore, a strap-on RTO is located on eacn vartical run of separately routed impulse lines to determine the impulse line temperature and cor-rect the reference leg density contribution to tne d/p measurement. Temperature measurements are not required where all tnree impulse lines of an instrument train are routed together. Based on the stucies of a nuccer of representative plant arrangements, a maximum of 7 independent vertical runs cust be measured to adequately compensate for density cn anges . 4.2.2 MItRCPTCCESSCR RVI.IS The microprocessoF RVLIS includes equivalent reactor vessel level. indications on redundant flat panels with alphanumeric displays provided for control room installation in addition to having this information availacle for display at the microprocessor chassis. RVLIS is configured as two protection sets, in certain installations in separated sections of a single instrument rack and in other installations in two separated instrument racks. The envelope of an instrument rack occupies a,c a space at the base of , The blocx diagram of the RVLIS using microprocessor equipment is shown in Figure 4 4 This diagram snows that in addition to the reactor vessel level (d/p) transmitter input, there are also temperature compensating signals, reactor pumo running status inputs, and RCS parameter inputs to each chassis of the two redundant sets. The output of each set will be to displays and to a recorder, as well as an output for a serial data link. A'Ieneral 3 display arrangement is shown to Figure 4-5. Conformance with Regulatory Guide 1.97 for the processor display system is given in Tac'c a.l. 44 MP
'532
4.2.2.1 RVLIS Inouts The microprocessor system inputs are as follows. If existing unquali-fied inouts are used, isolation as required will be provided by the
- wner.
Differential Pressure Transmitters The tnree d/o transmitters per set are used to measure the d/ps between the three pressure tap points on the primary system, as discussed below:
,a,c l.
The direction of this transmitter's output is full scale (20 ma) with the vessel full and zero scale (4 ma) with the vessel emptied to the hot leg tap. These endpoints are nominal and are for low coolant temperatures. If no pumps are operating, APa gives an indication of level in the region above the hot leg. If the pump is running in the loop with tne hot leg connection, this indication will be invalid and most likely off-scale. The reading would. be flagged as " invalid" under these conditions. The effect on the indication from the pump not running in this loop, but running in other loops, is less than 10 percent of the range. _.a,c 2.
~
7583A MP
I ad) gives an '* -':ation Of reac::" .Es se' 'iveI 49eq "O Jm0s 1re running. *#
- ne Or 9ere cum 3s a#e "."94rg, a?) .sil' .6 Off-scale 100
- i *eaCing inv ali .
The sense of :7e 13 cu ;ut is s.cn :11: 1-20 ma signai is a 3
- nominally full sessel anc a 2 ma signal is f
- r a ocminally emoty vessel. a,c
~
1 j 3. i 1 - . . The secse Of :ae a? out:ut is :nat 20 ma represents all oumos running inc ma is smo:y vessel. Witn all pumas -unni,g and no voic fraction, ,e APc snoul: reac 100 percent a :e*0 cower. The rea:ing at fall ocwer is slign:ly higner. ! Reference Leg Temcera ;re RTD The reference leg temperature RTDs are used :s measare One temoerature i of the coolant in :ne capillary tuce reference legs. ' This is used to ! compute the density Of ne reference leg fl;i . A typical arrangenen: of tne reference leg tem:ers:;re 370s is snown in Figure 4-6. i The conversion of RT3 resistance to tem era ;re srall :0 /er ne temcera-ture range of 32 to 250*.:. 1 l l ' The RTDs are 100 com sl atinum four wire RT3s as s,:we 'n Figure 1-8. Hot Leg Temoerature i
'i:ner exis*ing QF s'a aide r ange no leg' :e :e"at.."a ss9s;rs 1"e 'Jsed to measure the 0olar.: *e oerature. ~ 1 is SF:e"1 .r5 is ,se: :: OllCM-late coolant densi y.
c.6 vp 7 4. s: . a
--+u *= ,- ,, . - . - ,, y - - - ~ . . .
2 :e lance Teacter :colan Dressure I':ner existing wi:e range cressure sensors or new pressure sensors will
- s ased to measure reactor coolant pressure. The pressure is used to
- alculate reactor c:olant density.
Tne block diagram of the compensation functions is shown in Figure 4-7. 21: ital Incuts
~ e reactor coolant pumo status signals indicate whether or not cumos are running. Reccgnizing that hydraulic isolators are provided on each M:ulse line for containment isolation purposes, each hydraulic isolator als limit switches to indicate they have reached the limit of travel.
2.2.2.2 Density Ccmoensation System , T: orovide the required accuracy for vessel level measurement, tem-
- erature measurements of the impulse lines are provided. These sasurements, together with the existing reactor coolant temperature easurements and wide range RCS pressure, are employed to comoensate the
- /p transducer outputs for differences ir. system density and reference leg density, particularly during the enange in the environment inside tne containment structure following an accident. A simplified schematic Of the density ccmoensation system is shcwn in Figure 4-7a. The d/p cells are located outside the containment.
e reference leg fluid density calculation must cover a range of 32 to 153cF. The fluid is assumed to be cr.,mpressed liquid water at 1200
- sia.
Iach of the three d/o measurements will have density corrections fecm
- ertain temperature measurements. Some of these will have a positive
- rr:: tion and scme negative depending on the orientation of the imoulse
? ine wlere tne temperature is being measured.
i.T l
. _ - . . _ ~ . - - - - . . - . -_-. - _- - . _ .. - - - _. . - .- _ . .-- .. .. . . . - . .
.j ! 2 Vessel Licuic Density Calculation i
. a,c s
l il 4 i 1 - 1 ' t Vessel lacor Phase Censity Calculation ' 2 a,c F { i l l Vessel level Calculation
.,a,c ,
Pump Flow clo Calculation
. a,c 1
1 1
. I 4-8 7533 A .4p
..a.c
~ The icwer of tne two calculated d/p corrections is divided into the measured d/p. The result is the percent of expected d/p and should read 100 percent with all pumps operating and no circulating voids. Scalino of Disolayed Values Each of the three d/p measurements after the preceding calculations snall be scaled to read in percent. With the vessel full of water and no pumps running, the outputs of APa and aPb should read 100 percent. 4.2.2.3 Plant Ooerator Interface and Disolays Information displayed to the operator for the RVLIS is intended to be unamoiguous and reliable to minimize the potential for operator error or misinterpretation. The redundant control board displays provide the following information:
, .m a,c 1.
, _ d.C 2.
m *
- . a,c 3.
~ ~
a.c vo
_ _ . - - - __ _. . _- -= - .-. -. . 1 i All signals are inout to a microprocessor-cased data analysis system. The control rocm Jisplay format atilizes an alpnanumeric display located remotely from the comoutational system. . Raoundant displays are provided for the two sets. Level information based on all tnree d/p measurements is presented. Correction for refer- , ence leg densities is automatic. Any error conditions such as out-of-
. range sensors or hydraulic isolators are automatically displayed on the affected measurements.
There are two display sneets for reactor vessel level: the first is a summary sneet, and the second is a trending of the three vessel level l indications, j . a,c i 4.2.2.3.1 Display Functions for Remote' Control Board The prime display unit for the vessel level monitor is tne 8 line, 32 character per line alphanumeric display which is located in the control board remote from the main processing unit.. Vessel Level Monitor Summary Disolay Figures 4-5, 4-5a and 4-5b give example displays. General arrangement is shown en Figure 4-5. The vessel level summary display is snown on Figure 4-5a. The following is a description of the display.
- 1. The first line gives the title of the display as shown. The use of the underbar feature delineates this line from the rest of the display.
- 2. The second line gives column headings as snown. Again, the use of the underbar clarifies the display.
~
76SIA MP
- 3. The thirc line gives the measured and normally expected values from tne AP3 measurement. The first field gives the title, the second gives the measured level, the third gives the normal value for tne current status, anc the last field gives the validity status and is blank under normal conditions.
4 The fourth line gives the aP measurement 3 results using the same format as in line 3.
- 5. The fifth line gives the AP measurement e
results using the same format as in line 3. The use of underbar in line 5 delineates tnis line from the next.
- 6. The sixth line gives the status of the pumps as seen by the unit.
The running pumps are identified. 7-8. The seventh line and eighth line are normally left blank and are reserved for hydraulic isolator limit switch indicators, out of range sensors and operator disabled sensors. Trend Disolay The trend display for the vessel level monitor shall use the format snown in Figure 4-56. Disolays on Main Processing fjnit The one-line forty character alphanumeric display on the front panel of the main processing unit is used to display individual sensor inputs. The sensor is selected with a two digit thumbwheel switch. The following information is to be given for each sensor:
- 1. Sensor identification
- 2. Input signal level
- 3. Incut signal converted to engineering units
- 4. Status of sensor input MP 2-10a
,,-m - --
Jisaciec :nouts any incuts can ce disaoled oy the coerator. This action is under the tontrol of a keyswiten on tne front panel of tne main comoutational unit and tauses one crocessor to disregard the analog input for that variaole. 4.2.3 RESISTANCE TEMPERATURE DETECTCRS (RTD) The resistance temperature detectors (RTO) associated with tne RVLIS are utili:ed to cocain a temperature signal for fluid filled instrument lines inside containment during normal and post-accident ooeration. The teoperature measurement for all vertical instrument lines is used to correct the vessel level indication for density cnanges associated with tne environmental temperature cnange. The RTD assemoly is a totally enclosed and hermetically sealed strap-on device consisting of a :nermal element, extension caole and termination cable as indicated in Figure 4-8. The sensitive portion of the device is mounted in a removable adacter assemoly whicn is designed to conform to tne surf ace of the tubing or oiping being monitored. The materials are all selected to be compatible with the normal and post-accident env ironmen t. Randomly selected samples from the controlled (material, i.anuf acturing, etc.) production lot will be qualified by type testing. Qualification testing will consist of thermal aging, irradiation, seis-mic testing and testing under simulation hign energy line break environ-mental conditions. For the qualified life requirements, see Section 2.3. The specific qualification requirements for the RTDs are as fol-lows:
- 1. @
The thermal aging test will consist of operating the detectors in a nign tempersture environment: either 200'.: for 528 hours or per otner similar arrhenius temcerature/ time relationshio. 4-11 MP 7533A -
- 2. Radiatien The detectors sna11 se irradiated to a total integrated oose (TID) of 1.2 x 103 rads gamma raciation using a Cc60 source at a minimum rate of 2.0 x 100 rads / hour and a maximum rate of 2.5 x 100 racs/ hour. Any externally exposed organic materials sna11 :e evaluated or tested to 9 x 108 rads TID beta raciation. The energy of the beta particle shall be 6 PEV for the first 10 Mrad, 3 MEV for 340 Mrad and 1 MEV for 150 Mrad.
- 3. Seismic The detectors will be tested using a biaxial seismic simulation.
The detectors shall be mounted to simulate a plant installation and will be energized througnout the test. 4 Hign Energy tine 3reak Simulation The detectors shall be tested in a saturated steen environment using the temoerature/ pressure curve shown in Figure 4 9, Caustic spray, consisting of 2500 ppm boric acid dissolved in water and adjusted to a pH 10.7 at 25*C by sodium hydroxide, shall be appifed during the first 24 hours. The test units will be energized throughout the test. The RTD device is designed to operate over a temperature range of
-58 to 500*F (the normal temperature range is 50 to 120*F).
4.2.4 REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM VALVES Two types of valves are supplied for the RVLIS. The root valve (3/4 778) is an ASME Class 2, stainless steel, globe valve. The basic func-tion of the valve is to isolate the instrumentation from the RCS.The other valve (1/4 x 28 ID), is an instrumentation-type valve. It is a manually actuated ball valve used to provice isolation in the fully 4-12 753;A
.. .- . -- - .. _ _ - - - . _ _ - ._ =- . _ - - - . - __. . . .
i i s
- icsed :csition. The valve is nermetically sealed ino utilizes s cacx-
'ess :esign to eliminate the possibility Of fluid leskage cast the stem to the atmosenere.
4.2.5 TRANSMITTERS, HYORAULIC ISCLATCRS, AND SENSCRS Differential 2resscre Transmitters The d/o transmitters are a seismically qualified design as used in i numerous other plant acolicaticns. In the RVLIS application, accuracy considerations dictate a protected environment, censecuently trans-mitters are rated for 40 to 1300F and 10 4 rad TID. Several special recuirements for these transmitters are as follows:
- 1. Must withstand long term overloads of up to 300 percent with minimal effect on calibration.
- 2. Hign range and bi-directional units required for pumo head measure-ments.
1
- 3. Must displace minimal volumes of fluid in normal and overrange oper-ating modes.
The first two requirements are related to the vernier characteristic of the pumos off level measuremer.ts and the wide range measurements, j l respectively. The third is related to the limited driving displacement of the nycraulic isciator wnen preserving margins for pressure and ther-mal expansion effects in the coupling fluids. The d/p transmitters are rated 30C0 psig working pressure and all units are tested to 4500 psig. Internal valving also provides overrange ratings to full werking pressure. l l 1-13 6 75alA l
dycelulic !selater - _ !.C 4 Hign Yolcme Senser _ , a,C
= -
.t 14 7521A
_ a,c 4.3 TEST PROGRAMS A variety of test pecgrams are in progress or will be carried cut to study the static and dynamic performance of the RVLIS at two test facil-ities, and to calibrate the system over a range of normal ocerating conditions at each reactor plant wnere the system is installed. These programs, which supplement the vendors' tests of hydraulic and electrical components, will provide the appropriate verification of the system rescanse to accident conditions as well as the appropriate precedures for proper operation, maintenance and calibration of the ecuipment. A description of these programs is presented in the follcwing section: 4.3.1 Forest Hills A breacboard installation consisting o< one train of a RVLIS was instal-led and tested at the Westinghouse Forest Hills Test Facild'y. The system consisted of a full single train of RVLIS hydraulic ccmoonents (sensor assemolies, hydraulic isolators, isolation and bypass valves and d/p transmitters) connected to a simulated reactor vessel. Process connections were made to simulate the reactor head, hot leg and seal table connections. Capillary tubing which in one sensing line simulated the maximum expected length (400 feet) was used to connect the sensor assemblies to the hydraulic isolators and all joints were welded. Con-nections between the hydraulic isolators, valves and transmitters util-i:ed comoression fittings in most cases. Resistance temoerature detec-tors, special large volume sensor bellows and volume disclacers insice a-15
- ne a <craulic isolator assemolies wnica are normally part of a RVLIS nstailaticn were not included in the installatien since elevated tem-
- erstare testing was not includec in tne program.
~he ycenulic isolator assemolies and transmitters were mounted at an elevation sligntly celow the simulated seal taole eleration. The Ocjectives of tne test were as follows:
- 1. *btain install 4cion, filling and maintenance experience
- 2. 8"ove and establish filling procedures for initial filling and system maintenance.
- 3. Establish calibration and fluid inventory maintenance procedures for shutdown and normal operation conditions.
4 prove long term integr'ity of hydraulic components
- 5. Verify and quantify fluid transfer and makeup requirements asso-ciated with instrument valve oper ation.
- 6. Verify leak test procedures for field use Reactor Yessel Simulator The reactor vessel simulator consisted of a 40 foot long 2 inen diameter stainless steel pipe with tags at the top, side and bottom to simulate the reactor head, hot leg and incore detector thimole conduit penetra-tion at the bottom of the vessel. Tucing (0.375 inch diameter) was used to connect this icwer tag to the sensor at the simulated seal table elevation and the hot leg sensor to the head connection was simulated by 1 incn tubing which connected the sensor to the vessel.
The eactor vessel simulator was designed for a pressure rating of 1400 osig to como1y with local stored energy and safety coce considerations. 1 16 753 3
Ins 311atien The sys:am nas installed in ene high bay test area of the Westinghouse
.: crest Mills Test Facility by Westingnouse personnel under the supervi-sion of Fores Hills Test Engineering. All local safety codes were
, censidered in the construction. i i. Filling Goeration F _ a.c r l I i, I e i t I I e . ( 4-17 s531A 1
,ac 4.3.2 SEMISCALE TESTS In orter to study the transient response of the RYLIS during a small-becak LOCA and other accident conditions, the hydraulic components of the RVLIS have been installed at the Semiscale Test Facility in Idaho. Vessel level measurements wil1 be obtained during the current semiscale test program series which runs from December 1980 to Marca 1982. The test scheduled to be completed by July 1981 are expected to provide the desired transient response verification; additional data will be obtained from the tests scheduled for completion by Noventer 1981.
The Semiscale Test Facility is a model of a 4-Loop pressurized water reactor coolant system with elevation dimensions essentially equal to the dimensions of a full-size system. The reactor vessel contains an electrically heated fuel assembly consisting of 25 fuel rods with a heated length of 12 feet. Two reactor coolant loops are provided, each having a pumo and a steam generator with a full Meight tube bundle. One loco models the loop containing the pipe break, which can be located at any point in the loop. The other loop models the three intact locos. A blowdcwn tank collects and cools the fluid discharged from the pipe ) break during the simulated accident. Over 300 pressure, tenperature, fl ow, level and fluid density instruments are installed in the reactor i vessel and loops to record the fluid conditions throughout a test rui.. f Test results are compared with predictions for verificatien v~ comouter code accels of the transient performance. 1-18 7531A
The Westingnouse level measurements cotained curing a test run will :e 0:mcarec *ita cata cotained frem existing instrumentation installed n the semiscai4 coactor vessel. The somsca4 facility nas tw metnocs of l 7easuring the level or fluid censity: /: measurements are obtainec Over il vertical spans on the reactor vessel to determine level witnin eacn saan, anc ga.ima censitemeters are installec at 12 elevations cn the aeactor vessel to cetermine the fluia density at each elevaticn. This l cata estaclishes a fluid density profile within the vessel under any operating condition, and this information will be compared with the data Octained frem the Westingnouse level instrumentatien. Other semiscale f acility instruments (lcoo flows and fluid densities *nen pumos are Ocerating, and pressure and temperatures for all cases) will provide supplemental information for interpretation of the test facility fluid l cenaitions and the level measurement. Specific tests included in the semiscale test program during which Westingneese RVLIS measurerents will be obtained are as folicws:
- 1. Miscellaneous ste.ady state and transient tests with pumps on and off, to calibrate test facility heat losses.
- 2. Emall-breck LCCA test with equivalent of a 4 inch pipe break.
- 3. Repeat of small-break LOCA test with test facility mcdifiec to simu-late a plant with ucper head injection (UHI).
4 Several natural convection tests covering succcoled and saturated coolant canditions and various void centents.
- 5. Tests to simulate a statien blackout with discharge througn relief I va%es.
l
- 6. Simulation of the St. Lucie cooldewn incident.
l l i 4-19 l 7551A
4.3.3 PLANT !TARTUP CALISRATION Quring the olant startuo, subsequent to installing the RVLIS, a test rogram will te carriec out to confirm the system calibration. The orogram will cover normal operating conditions and will provide a reference for ccmoarison with a potential accident condition. The ele-ments of the program are described below:
- 1. During refilling and venting of the reactor vessel, measurements of all 5 o/p transmitters would be comoared to confirm identical level indications.
- 2. During plant heatup with all reactor coolant pumps running, measure-ments would be obtained fran the wide range d/p transmitters to confirm or correct the temperature compensation provided in the system electronics. The temperature compensation, based on a best estimate of the flow and pressure drop variation during startup, corrects the transmitter output so that the control board indication is maintained at 100 percent over the entire operating temperature range.
- 3. At hot standby, measurements would be obtained from all transmitters with different combinations of reactor coolant pumos operating, to provide the reference data for comparison with accident conditions.
For any pump operating condition, the reference data, represents the ! normal condition, i.e., with a water-solid system. A reduced d/p during an accident would be an indication of voids in the reactor vessel. 4 At hat standby, measurements would be obtained from the reference leg RTDs, to confirm or correct reference leg temperature compensa-tion provided in the system electronics, a-20 l 7ECA
1 l a.a CPERATING 2ERFORMANCE Eaca train of tre IVL!S is caciole of monit; ring ;pplant mass in the vessel fecm normal coerstion to a conoiticn of complete uncovery of the e* actor core. This capacility is provided by the three d/p transmit-ters, eacn transmitter covering a specific range of operating conci-tions. The tnree instrument ranges provice overlac so that the measurement can :e octained from more than one meter under most accicent conditions. Cacabilities of each of the measurements are described below:
- 1. Reactor Vessel - Upper Range The transmitter span covers the distance from the hot leg piping connection to the top of the reactor vessel. With the reactor c:ol-ant pump shut down in the loop with the hot leg connection, the transmitter output is an indication of the level in the upper plenum or uoper head of the reactor vessel. The measurement will provide an accurate indication for guidance when operating the reactor ves-sel head vent. The neasurement will also provide a confinnation that the level is above the het leg noztles.
When the pump in the loop with the hot leg connection is Operating, the d/p would be greater than the transmitter span, and the trans-mitter output would be disregarded.
- 2. Reactor Vessel - Narrcw Range The transmitter span covers the total heignt of the reactor vessel.
With pumps shut down, the transmitter output is an indication of the collapsed water level, i.e., as if the steam bubbles had been separ-ated from the water volume. The ac;ual water level is slightly nigher than the indicated water level since there will be some cuan-tity of steam bubbles in the water volume. Therefore, the RVLIS orovides a conservative indication of the level effective for ade-cuate core cooling. l l 1 1-21 i
?E51A 1
when reactor coolant cumos are coerating, the d/p would be greater than the transmitter scan, and the transmitter output noulo be dis-regarced. l l 3. Reactor Vessel - Wice Range The transmitter scan covert the entire range of interest, from all pumps operating with a water-solid system to a completely empty reactor vessel and therefore, covers the measurement spans of the other two instruments. Any recuction in d/p canpared to the normal operating condition is an indication of voids in the vessel. The l reactor coolant pumos will circulate the water and steam as an essentially homogeneous mixture, so there would be no distinct water level in the vessel. When pumos are not operacing, the transmitter cutout is an additional indication of the level in the vessel, sun-plementing the indications from the other instruments. The output of each transmitter is compensated for the density difference between the fluid in the reactor vessel and the fluid in the reference leg at the initial ambient temperature. The compensation is based on a wide range hot leg temperature measurement or a wide range system pres-sure measurement, wnichever results in the hignest value of water den-sity, and, therefore, the icwest value of indicated level. Compensation based on temperature is acplied when the system is subcooled, and com-pensation based on pressure (saturated conditions) is applied if super-heat exists at the hot leg temperature measurement point. The output of each transmitter is also compensated for the density dif-ference between the fluid in the reference leg during an accidant with elevated temperature in the containment and the fluid in the reference leg at the initial ambient temperature. The compensation is based on temperature measurements on the vertical sections of the reference leg. The corrected transmitter outputs are displayed on meters installed on the control boaro, one meter for each measurement in each train. A . three-cen recorcer is also proviced on the control board to record the 1-22 7531A
levei or relative d/p and to display trends in the measurements, an incicater lignt instaliec unoer the upcer range level meter would pro-vice an incication if :ne sure in tne 1sco with tne hat leg connection is :cerating, anc therefore in indication that the off-scale reading on
- ne e:er snould be disregarcea.
Dur'9; normal clant heatuo or hot stancy oceration with all reactor coolant puros operating, the nice range d/p meter would indicate ICO l percent on the meter, an indication that the system is water-solid. If 1 l less than all pumps are operating, the meter would indicate a icwer d/D i (cetermined during the plant startup test program) that would also be an indication of a water-solid system. With pumos operating,.the narrow l range and upper range meters would indicate off-scale. If all pumps are shut down, at any temoerature, the narrow range and upper range meters would indicate 100 percent, an indication that the vessel is full. The wide range d/p meter would indicate aceut 32 per-cent of the span of the meter, wnicn would be the value (determined during the test program) corresponding to a full vessel with pumos shut dcwn. In the event of a LOCA where coolant pressure has decreased to a prede-termined setpoint, existing emergency procecures would recuire shutdown i of all reactor coolant pumps. In these cases, a level will eventually ce established in the reactor vessel and indicated on all of the meters. The plant operator would monitor the meters and the recorder to determine the trend in fluid mass or level in the vessel, and confirm that the ECCS is adequately compensating for tne accident conditions to orevent ICC. Future procedures may require operation of one or more pumos for recov-ery from certain types of accidents. When pumos are operating wnile voics are developing in the system, the pumps will circulate the water and steam as an essentially homogeneous mixture. In these cases, there will be no ciscernible level in the reactor vessel. A decrease in the ( 4-23
$5 o - ,-n- , - - , . , ., - - -
measurec Up comparea to tne nor-nal operating value *TTl be an indica-tion of voids in tne system, and a continuously decreasing d/p will j indicate that the void content is increasing, that mass is being lost l fran :ne system. An increasing d/p will indicate that the mass content I is increasing, tnat the ECCS is effectively restoring the system mass content. 4.5 RVLIS ANALYSIS In order to evaluate the usefulness of the RVLIS during tne approach to ICC, it was cecided to determine the response of the RVLIS under a variety of fluid conditions. The RVLIS response was analytically deter-mined for a nuncer of small breat transients. The response was deter-mined by calcuiating the pressure difference between the upper head and lower plenum and converting this to an equivalent vessel head in feet. (Note that RVLIS indications will actually be represented by percent of span) Saturation density at the fluid temperature in the upper plenum was used for this conversion. This approximates the calibration that will be used for the RVLIS. This indication corresponds to the RVLIS configurstion used for non-UHI pl ants. The conclusions of the study are expected to be the same for the UHI configuration. The indication of the upper span (hot leg to upperhead) is not included in tais analysis. The upper span indication will be used for head venting operations and will not be used to indi-cate the approach to ICC. When the reactor coolant pumps are not operating, the RVLIS reading will be indicated on the narrow range scale ranging from zero to the height of the vessel. A full scale reading (100 percent of span) is indicated wnen the vessel is full of water. This reading represents the equiva-lent collapsed liquid level in the vessel wnica is a conservative indi-caticn of the approach to ICC. The RVLIS inoicitica can alert the ocerster that a condition of ICC is being approached and the existance of ICC can be verified oy checking the core exit thermocouples. When
- ne reactor coolant pumos are operating the narrow range RVLIS meter will be pegged at full scale. .
a-24
'ahen ,ne reactor c :olant Dumos are cDerating, One IVLIS reading 4ill 3e incicatec cn :ne nice range scale .nica reacs frem 0 to 100 percent.
he 100 cer:ent reacing corresponcs :o a full vessel witn all of :ne aumos in aceration. di:n :ne ;umos -unning :ne ;VLIS reacing is an incication of :ne sota fraction of :ne vessel mixture. As tne voic content of :ne vessel mix-ture increases, the censity cecreases and :ne RVLIS reaaing will decrease due to the reduction in static head and frictional pressure croo. The latter effec: will ce ennanceo ay cegracation in reac:ce coolant puma performance. 'ahen :nis reading drops to approximately 33 percent, tnere will also ce an indication on :ne narrow range scale. This fraction approximately corresponds to a vessel mass at wnich nould just cover :ne core if tne pumps were tripped. Four small-creak transients under a variety of conditions are ciscussed in :ne next section. Three of these cases ere cotained from aFLASH analyses and the otner was ootainea fran the ICC analysis using NOTRL'MP . A description of :nese codes can ce found in References 1
- nrougn 5 in Section 5.0.
The transients includeo in this report are listec Taole 4.2 wnien gives i a crief description Of :ne transient, tne plant type, and tne model usec l for :ne analysis. A aiscussion of each transient is provided in :ne next section. Figures 4-12 througn 4-23 provice plots of vessel eno-anase mixture level, RVLIS nar-ow range reading, mixture and vessel voic fraction, and for Case 3 with pumos running, RVLIS wide range reacing and cold leg mass flowrate.. The reo-pnase mixture level pict:ed is :nat anicn was precic:ed sy One codes for the mixture :e..;nt below the upper support plate. Water in the upper nead is not reflectec in this plot. The RVLIS reading :nat would ce seen is slottec on tne same figure for ease of comparison. The vcid fraction plots are for the core ana upper plenum fluta l volumes. The lixture soic fraction includes :ne volume selcw .ne two I : nase mixture level anile :ne Octal voic fraction also incluces tne steam space acove :ne mixture level. 1-25 l l
4.5.1 Transients Investicated ( Case A The initiating event fer tnis transient is a 3 incn break in the :ald leg. Af ter the break opens, the system depress;.rizes rapidly to the steam generator secondary safety valve setpoint. Consistent witn tne FSAR assumptions, the reacter coolant pumps are assumed to trip early in I the transient *nen the reactor trips, l The system pressure hangs up at the secondary setpoint until the loop seal uaalugs at approximately 550 seconds, allowing steam to flow out the break and the depressurization continues. The cere uncovers wnile the loop seal is craining then recovers wnen the loco seal unplugs. The core then begins to uncover again as more mass is being lost through the l break than ~is being replaced by safety injection. The core begins to l recover at about 1500 seconds wnen the accumulators begin to inject. . This transient does not represent a condition that would lead to ICC but it does represent a break size in the range that would be most procable if a small-break did occur. The response of the RVLIS for typical con-ditions for wnich it would be used can be investigated with this tran-sient. After the reacter coolant pumps trip the RVLIS reading drops rapidly to the narrow range scale. It f alls until the pressure dron due to flow becomes insignificant compared to the static nead of the fluid in the vessel. The first dip in the RVLIS reading is due to the behavior of the upper head. When the upper nead starts to drain it benaves like a pressurizer. The pressure in the upper head renains nign until the mixture level drops to
, belcw the too of th'e guide tube wnere steam is allowed to flow frcm the l
l a-25 l l
uoeer head to the uocer plenum. When this Occurs the uocer head pres-sure :ecreases - therecy increasing the vessel d/p - ano the RVLIS
- ct.g p =ccra!y m~. ~u % , ~y, ajs onencrenen is more prevalent for large-break sizes and the effect will be Of 3rief duration for breaks in this range. Further nore, the ICC guicelines recuire verification Of the RVLIS reading througn the use of tne c:re exit thermoccuoles. During this onenemenon, the core exit thermoccuoles would read saturation temperature. Therefore, this early phenomena in the upper head will not cause a false indication of ICC.
When the vessel begins to drain during the 1000 seal uncovery the RVLIS reading trends in the same direction as the vessel level. The RVLIS reading remains below the vessel mixture level and is therefore a con-servative indication. When the vessel mixture level increases after the loop seal unplugs the RVLIS reacing follows it. Then, RVLIS readings continue to follow the vessel mixture level throughout the transient while underpredicting the actual two-phase level. The wider difference between the RVLIS level and the two-phase level later in the transient is cue to the system being at a lower pressure which allcws more bubbles to exist in the mixture. Case 3 This case is the same as case A except it was assumed that the reactor coolant pumos continued to operate until 750 seccnds. If the reactor ecolant pump trio criteria is folicwed the pumos would be tripped much earlier in tne transient. This case is, however, instructive in oeter-mining the RVLIS response when the pumos are running. After the break opens, the system depressurizes rapidly to the secondary safety valve setpoint, and then begins a period of very slow depressuri-zation. During this time the upper portions of the system drain. Due to the reactor coolant pumo operation, the two-phase mixture in the vessel remains at the hot leg elevation, althouen the void fracticn of the mixture continues to increase. 4-27 i 7531A
12 N cennde 2ha JyjLram Aas minad to the maint that steam can be vented through the break and the system begins to depressurize more rapidly. The pumos are also tripped at this time resulting in a col-lapse of the mixture in the vessel and the core uncovers. The vessel continues to drain until the accmulators inject at about 1000 seconds to recover the core. There is a subsequent uncovery which will be ended when the pressure is low enough for the safety infection to make up for mass lost through the break. During the early portion of the; transient the wide range RVLIS reading drops f airly snothly from 100 percent to about 20 percent, which is due to the decreasing mass in the vessel and the decreasing pressure drop as the pump performance is degraded. The plot of cold leg mass flowrate is indicative of the pumo degradation. The oscillations in this plot are due to alternate steam and two-phase flow predicated by WFLASH. When the flow through the pump becomes mostly steam, the increasing void fraction of the vessel mixture becomes the predominant f actor in the decreasing RVLIS reading. RCP operation keeps the steam and water mixed enough that the mixture level does not f all below the hot legs, although the mixture void frac-tion is increasing during this time. This loss of inventory is indi-cated by the continued drop in the RVLIS reading. When the p e ps trip, the steam and water in the mixture separate and there is a rapid decrease in the core mixture level and mixture void fraction although the vessel void fraction continues to rise. The f act that mass is being redistributed rather than lost is seen in the RVLIS reading - there is little change in the reading (compared to the change in level) fran 750 seconds to the time that the accumulators cone on. The prolonged reactor coolant pump operation has caused the 10wncomer to drain 50 that when the accumulators cane on the cold accumulater water condenses steam in the downcomer causing a local depressurization. The downcaner pressure is then temporarily lower than the upper head pres-sure cue to ine.rtia and the RVLIS reading becomes temporarily negative. 4-25 7531;
s
~his perico of er-ktic inaicaricn is artef (one or two minutes). The crassure 4111 equilicrate anc :ne RVLIS 4111 resume following :ne vessel m xture level. This pnenemenon nas c nly seen caservec anen :ne accumu-
'.ators in;ect onen tne ocwncomer is nignty voicec. There is no apparent discrepancy curing accumulator injection anen :nere is a significant amount of water in tne cowncomer. It is celievec :nat :nis effect is exaggeratec by ne modeling tecnnicues used in WFLASH (wnica utilize a nomogenous equilibrium assumotions at :ne accuaulator injection loca-tien). For :ne remainder of :ne transient :ne RVLIS reading follows :ne vessel level closely.
Case C The initiating event for tnis transient is ne opening of :ne pressur-izer pcwer coerated relief valves (PORVs). The reactor coolant cunos and ne reactor trip early in the :ransient on a low pressurizer pres-sure signal consistent with FSAR assumptions. Auxiliary feecwater is availaole in this case out, no pumced safety injection is assumed. The pressurizer mixture level rises to tne top of the pressurizer early in :ne transient and stays at tnis level througnout r:ost of the tran-sient. The flow througn the PCRVs alternates setween steam and twopnase mixture wnile the cressure in the system crops rapidly to :ne steam generatcr seconcary safety valve setpoint. The pressure nangs up at this value until ne upper portion of the system nas crained anc :nen continues to cecrease. When the upper portions of the primary system i (excluaing One pressurizer) nave crainec :ne vessel mixture level cegins to cecrease and continues until tne core comoletely uncovers. The RVLIS reading crops rapidly to the narrow range span after the reac-I
- or coolant pumps are tripped. When the vessel level reacnes :ne not leg elevation the calculated RVLIS readings begin to oscilate cue to the modelling used in WFLASH. In WFLASH, tne not legs are connectec to ne I
vessel by ::oint contact connections. This modalling tecnnique causes
- ne not leg flow to alternate cetween steam ano two pnase flow. The oscillitary cenavior of :ne calculateo RVLIS reading continues anile :ne 1-29
level remains at the het legs. The average calculatec value during this cerico of time shows that the RVLIS reading is a conservative indication of the mixture level. i When the vessel mixture begins to cecrease, the RVLIS reading decreases as nell. The RVLIS continues to underpredict the two-onase mixture level and to follcw the trend. Case 0 l l his case is one of the transients investigated for the ICC study using NOTRUMP. A more detailed discussion of this transient can be found in Reference 1. The RVLIS reading is below the vessel mixture leve.1 throughout most of the transient and is therefore a conservative indication. The RVLIS reading follows the same trend as the vessel mixture level except for early in the transient when the mixture void fraction is fluctuating. ( Included in the plots for this case is a canparison of the mass inven-tory in the core and upper plenum regions to the RVLIS reading. This comparison shows that the RVLIS reading also corresponds very well with the relative vessel sass inventory. Also included is a comparison for the UNI and non-UHI RVLIS configurations. For the UHI RVLIS configura-tion, the pressure difference is measured from the hot leg to the lower plenum rather than the upper head to lower plenum. This plot shows a very good comparison between the two systems, indicating that either will give a useful indication. 4.5.2 Observations Of ,T*a Study The RVLIS will provide useful information for breaks in the system ranging from small leaks to breaks in the limiting small-break range. For breaks in this range, the system conditions will change at a slew enougn rate that the operator will be able to use the RVLIS information as a basis for some action. l 4-30 7531A
- ce larger breaks, the response of the RVLIS will ce more erratic, due to racid cressure cnanges in the vessel, in the early sortion of the l 1
- ' - - N N " rocci :g mill = useful f or acci+ 4 ~' 2 M daa' ' -ecovery, wnen other corroborative inoications of ICC could also te coserved.
Very few instances nave been identified where the RVLIS may give an imibiguous indication. These include a break in the upper head, accumu-lator injection into a hignly voided downcomer, periods of time when the uocer head behaves like a pressurizer, upper plenum injection, and peri-ods of void redistribution. ; A break in the upper head may cause a much lower pressure to exist in the upper head compared to the rest of the RCS. Because of this the pressure difference between the lower plenum and the upper head is much larger than is seen for an equivalent vessel level wnen the break is lccated elsewnere in the system. The reading, in fact, : y never reach the narrow range scale. If the narrow range reading remains at full scale and the wide range reading is greater than that reading which , would indicate a full vessel with the reactor coolant pumps tripped, a break in the upper head is indicated. This situation should not cause a problem in detecting ICC because of the parallel logic for the " kick-out" to the ICC procedures. If the RV!.IS indication is erroneous due to a break in the reactor vessel uoper head, the operator will begin fol-lowing the ICC procedure if the selected core exit thermoccuoles read i 1200*F, 1 l l This situation only exists, however, when the break discharge is large l enough to cause a large d/o through the flow paths connecting the upper head to the rest of the system. These flow paths become the limiting factor in the depressurization rate. Thisanalysisisapplicable50allWestinghousePWRplants, including those plants with upper plenum injection (UPI). The normal condition for continuous UPI occurs only with the operation of the low head safety injection pumps, wnich does not occur until a pressure of under 200 psi a-31 7531A
e n:w 71ctmtge n rct uws..cu m - mu m uwru rmrss :T w mm incication. The increased c/p due to the ficw blockage will be small during natural circulation. The RVLIS will continue to follow the treno in vessel level. When the reacter ccolant pumos are coerating, flew clocxage is not excected to occur unless the pumos had previously been tricced and are being restarted after an ICC situation already exists. If flow blockage were present when the pumps were running the RVLIS indication woula still be useful and, although the indication would be somewnat higher, would centinue to follow the trend in vessel inventory. 4.5.3 Conclusiens
- 1. With the RCPs tripped, the Westinghouse RVLIS will result in an underpredicted indication of vessel level while providing an unanbi-guous indication of the mass in the vessel. The Westinghouse RVLIS will also measure the vessel level trend reasonably well.
- 2. With the RCPs tripped, it is feasible to determine a setpoint for the RVLIS to warn the operator that the system is approaching an uncovered cere.
- 3. The RVLIS should be used along with the core exit thermocouples to detect ICC. i 4 With the RCPs running, the RYLIS is an indication of the mass in the vessel.
- 5. When the RCPs are running, and the RVLIS reading drops to the narrow range scale, there is significant voiding in the yessel and the core would just be covered if the pumos were tripped.
- 5. A break of sufficient size in the upper head could cause the RVLIS to give an ambiguous indication of vessel mass. The core exit thermoccuoles, newever, will provide an indication of ICC if appro-priate.
1-33 7531A l l
i
- 7. 4ccumulat:r iqjection wnen the dcwnc mer is hi_gni,y voided could result in a temcorarily erratic indication.
1
- 3. The RVLIS may significantly under;refict the vessel mass anile the
'iuid in the uccer head is flasning. However, use of the core exit thermoccuoles aill preclude a premature entry to the ICC peccedures.
- 9. Rapid void redistributions will not be detected by the RVLIS.
f t l 1-34 7531A
TABLE 4.1 CONFORMANCE WITH REGULATORY GUIDE 1.97, ORAFT 2, REY. 2 (6/4/80) FOR THE MICROPROCESSOR DISPLAY SYSTEM Seismic qualification Yes i Single failure criteria Yes Environmental quailification Yes
'[IEEE-323-1971 applicability]
Power Source Class 1E Quality Assurance Yes 10CFREO Appendix 3 applicability Disolay type and method Vertical scale voltage processed in addition to a l i recorder Yes Unioue identification Periodic Testing Yes l [
* :n scme cases IEEE-323-1974 is applicable.
4 35 MP
'552A
TABLE 4.2 TRANSIENTS INVESTIGATED CASE ptANT DESCRIPTION A 3 loco 3 inch cold leg break - FSAR assumotions'; WFLASH 2775 MWt 3 3 loop 3 inch cold leg break - RCPs trip at 750 seconds - 2775 MWt otherwise, FSAR assunctions; WFLASH C 4 loco 2.5 inch break in top of pressurizer - no UHI - no UHI type pumped safety injection - pumps not running; 3411 MWt WFLASH 0 4 loop 1 inch cold leg break - no high head safety Ncn-UHI injection; NOTRUMP 3411 MWt I I
'RCPs trippec at reactor trip, minimum pumped safety injection is availacle, . minimum auxiliary feedwater is available.
4-36 7521A
17917 1 i I l SPARE HEAD PENETRATION 8N ). _ cQUTLIT w ,. -g REACTOR l Cone I wovtAatz I DETECTOR CONCUlf - _ n NN A TRAIN 3 1 Figure 1.l Reactor Vessel Level Instrument System i l l I
179172 l l
- HYDRAULIC 13CLATOR (TYPtCAL)
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! TABLE 4 CCNTAINMENT ' MALL t
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17917-5 F _a.- 1 1 l Figure 4 2 Reactor '/essel Level Instrwent System 31cck Diagram (One Iet of Two 3.ecundan: Resets Shcwn)
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.i i i 4 1 1 HEACTOH VESSEL LEVEL
SUMMARY
i VALUE NORMAL STATUS PLENUM LEVEL 73% 100% ALAHM i i VESSEL LEVEL 47%* 0% INV A LID FLOW llEAD >110%*# 100% OFF SCA PUMPS HUNNING: 41, 42, 43, 44 ISOLATOR A L AHMS: LI3
# DIS AB LE D: T3 T il l 5
v iigiare 4-54 VeShel I.evel Ste:uitary (lisplay
1 i i i REACTOR VESSEL LEVEL TREND T I A4 E PLENUM VESSEL F'L O W MIN LEVEL LEVEL llEAD 00 73% 47%I >110% OS i
-16 78% 49% I 98%
-30 79% 62% I 97%
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4
-60 97% 99% I 99%
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Figure 4-Sb Vessel Level Trend Displey
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i l l l 9 17317 7 3.b.c 9 i i i l i
- J Figure 1-7 Slack Diagram of Ccmcensation Function
179178 1,c Figure L7a Simplified Schematic of Density Compensation System
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- a tie nu l g 206 -
I" I I I I 1 I I I I O 10 3 6 0 10 3 b 20 24 16 SEC MIN MIN SEC MIN MIN MIN llO9H DAY T;ME
*IBME DE1 WEEN TEMPERATURE THANSIENIS MUST BE AT LEAST ONE ilOUH OH UNill TEST UNIIS HEluitN fu A STEADY STATE OU1PUI. IlME ABOVE 340F MUST BE FIVE MINUTES OH LESS.
7, 2 v ilEll! Simulation Profile g Fi 3are 4-9
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! ! l l l l l l l 0.0 0 2tio 600 760 1000 1260 1600 1760 2000 22b0 2btM1 11ME (SECONDS) i I
Figure 4-13 Case A 3-Loop Plant, 3 Inch Cold Leg Break, Pump Trip ;., ! with Reactor Trip, Vold fraction 5, i i
1T317 74 5 9 5 c
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. igure J.-15 Case 3 3-L::co Plant, 3 Inch Cold Leg 3reak, Pumo Trio at 750 Sec::ncs, RVLIS Reading and Mixture Level l
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TIME (SECONDS)
~
g figure 4-17 Case B 3-Loop Plant. 3 Indi Cold Leg Break. Pump Irlp at
?
750 Seconds. Cold Leg Itass Flowrate (LB/Sec) -1
I 17317 3 a .b .3 j , 40 ,
, 1 35 1
t 1 I l 30 - i i 25 - g 20 - u. I 15 l l l l 10 i 5 I l I I I l l l i O 0 400 800 1200 1600 m 2400 2800! TIME (SECONCS)
- igure 1-18 Case C 2.5 Inc:s Pressuri:er Break, No SI, RVLIS Reading 1
' and Mixture Level . 1 i i
l v, 17917 29 2
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17917 37 a. i
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E k N a 0 O e I 0 C r 6 7 E B S ( g E e L M I d T l o C . h l 0 c e _ 0 n v 0 i e _ 6 l _ 1 e D r u e t n s a i C H 2 0 2 I 0 - 6 4 2 e _ r - u g i F 0 0 6 0 6 0 6 1 0 1 6 e 4 3 3 2 2 . Uutss - l 1;
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88j TOTAL VOID FHACTION [g , g
--- MIXTURE VOID FHACTION f [ l i in 0.7 -
g l l 0 0 l l
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~~
0.1 - #~ VI f i LI e 14 I I I I i w 0 2500 5000 7600 10000 12500 1b000 TIME (SECONDS) r.
"I Y,
figure 4--21 Case 1) I inch Cohl leg fireak, ICC Case, Vold fractioin "
5.0 GUIDEt!NES FOR THE USE OF !EC !NST9BMENTATION 5.1 REFERENCE OhMERS 3R00P PROCEDURES 3ased on the analyses defined in See:4cns 1.3 and 4.5 of this report, Westingneuse ano the Westinghouse Owners Group have developed a Refer-ence Emergency Operating Instruction to address recovery frem ICC condi-tions caused by a small-break LOCA without hign head safety injection. This instruction has been transmitters to the NRC via Westinghouse Cwners Group Letter, OG 44, dated November 10, 1980. It should be noted that this instruction was developed on a generic basis as a technical reference for implementing plant specific procedures, and must be tailored to meet plant specific needs. 5.2 SAMPtE TRANSIENT The response of the vessel level indications, other ICC instrumentation and system response during these ICC events and recovery actions are cescribed in References 1 and 2. 5-1. 7551A
6.0 REFERENCES
- 1. Thomoson, C. M., et al., "Inacequate Core Cooling Stuc1es of Scenaries witn Feecwater Availaole, using the NOTRUMP Comcuter Coce,* '4AP-9753 (Proprietary) and WCAP-9754 (Non-Proprietary), July 1980.
- 2. Mark, R. H., et al., " Inadequate Core Cooling Studies of Scenarios with Feecwater Available for UHI Plants, Using the NOTRUW Computer Code," WCAP-9762 (Proprietary) and WCAP-9763 (Non-Proprietary), June 1980.
- 3. " Report on Small 3reak Accidents for Westingnouse Nuclear Steam Supply System," WCAP-9600 (Proprietary) and WCAP-9601 (Non-Pro-prietary), June 1979.
4 Esposito, V. J. , Kesavan, K. , and Maul, 8. A. , "WFL5SH - A FORTRAN-IV Cascuter Program for $1mulation of Transients in a Multi-Loop PWR," '4AP-0200, Revision 2 (Proprietary) and WCAP-8251, i Revision 1 (Non-Proprietary), July 1974. ! 5. Skwarek, R., Johnson, W., and Meyer, P., "Westingnouse Emergency Core Cooling System Small Break October 1975 Model," WCAP-8970 (Pro-l prietary) and WCAP-8971 (Non-Proprietary), April 1977.
- 6. ' Analysis of Delayed Reactor Coolant Pump Trio Ouring Small Loss of l Coolant Accident for Westinghouse NSSS," WCAP-9584 (Proprietary) and WCAP-9585 (Non-Proprietary), August 1979.
5-1
- :}}