ML17345B347

Administrative Procedure 0103.16, Duties & Responsibilities of Shift Technical Advisor

ML17345B347
Person / Time
Site:	Turkey Point
Issue date:	07/29/1982
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17345B345	List: L-83-555, Forwards Response to Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events ML17345B346 ML17345B347 ML17345B348
References
0103.16, 103.16, NUDOCS 8311160156
Download: ML17345B347 (102)

Text

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ACi<Ii!ISTRATi'lE ?'l CROUP.

3 33 o j5 s PAGE 2

nUTIES '"~ g/",(5 I> ~> IT!<<nF HE SHI.'lSIC-'L "O'I SC

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'lUREG U737

3.4 Oefinitions

3 f>>

Sh'.f evhn ca!

-'c'".'sor -

Sn

>>nci /id"al with a bache!or'5 equivalent ln a scientific or ngineering 0:scioline who trained ana

@vali iec in accoraance

'/i:h administrative 0307.

>>'ei>>rea or nas

'een

?race"ur4 3.4.2 Pl an>

'hnoma!

"c urronce or

= ent nanna'iant ope. atlna canal::ons.

ol r lo bu "5"

/

si"nificant events category.

an unol annec "eoarture fr om

.nl 5 map or ma J not

." sul la

'P a~1+

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eric

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=.U oes cnsib!1! ties:

ine

.echni c 1

eoar"..-..ent Sucerv i scr has ove. -! i

. 4socns'bi!

-v =or staffi c, imolement'ng, coor"..'nating, eva!uating, ana review na -he S.~ -rogr1m.

he

.echn;cs'.

St a=~

ra.'nina

-'cm>>nis:rative

?rocecure 33C..

oor.natal ls

/ esccrsib!=

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A o ~ 4 Eacn STD sna!'e r4soonsible for performing tn4 ac".'aent Sssessment "oer.

.'na exoer'.'ence assessment func:ions cescr.'"e" in paravraons 8.2

~ v \\

Snc anc

'le S.v i 5 r4soons,',ol e

.o the

.erhn'.ca!

".eoar.-..,ent

. "evv.'scr.

".i

. '.".~

o

~or,. a feac:or o i ant cond: t ons,

-ne 5

~ ~

sna

~

a+'/: se ne

'uc. ea; Suoerv'sor

',.'IPS'i ir. -he ".ontrol

~oem or echn>> "a.

5 occr-.

"entev, a I>> '+

A

% ~ V

.ne on-shif.

STA group will be comoose~

of oersonne!

ass;"ne"

-." the "1

Technical 5 aff.

The Technical repar-.ment Su"e. v.'scr has

=.".e "o".'cn

<<s-i anina ST-'hi <.s.

rur'na

~e

".-.--"-.-':;.-..e "ev r45OOns O l e

~ r'ev. OMance Of nose unC f5 1CI, a

'/

SC

,O ',.ec lc 1 Stl a5 a5sl +neo b)l le 4vnni 1 i

>Joe/ v sc>>

~

ant of

~ V

.ne S.v posit'.on is requlrea to oe

-..annea

=u.:nc oowe.

"e'r--'.on stSv=

os

".: stanc"'y, anc hct sn.tccwn.

AQ)II)IISTRATIV PROC aUR 0103

~

6

. AGE 3

OU) IES Al<P RES?ONS)3ILITI"=S OF THE SHIFT T"=CHt<ICAL ADVISOR 11/5/0

'ererences lIUREG 9573 "::1.'ess-ns Reed!rmenaatiors'.

'arned ask.

For ce Stat s

-. per=

=nor:-,

r.-..

6.2 USNRC letter of Septemoer 13,

197g, "Fol l "w up Ac lans Resu) lrg fr"m.he ARC Staff Reviews Regaraing tne Tnr e.'4lile Islana Unit 2 Acc aent" 6.3 US;)RC:nrar

.. at? Cn Acti Ce 3CvO6 ana Suppi ement

.'(O.

'=

".')Ot ? fi Cat iC1 Si gn? ficant vents at Operating

?ower Reactor Faci it;es" 6.4 10 CFR ZO.'Q3 ".'iotificat?on of.ncicenzs" 6.5 FPL Let er of nvece!liber 27,

1979, Rev i sed l~arch 7,

'.930,

. rcm A.

0 ~

Sc.".?i dt to J.

K. Hays, "Outies ana Resoonsioilities of '(uclear Plant Suoe. risors and Snif:.ecnnical Acvisors"

,IUREG O737 Rec=ras anc "cti=i"aticns:

lone 3 ~ 0 Iristr'Jc

~ ons:

8 ':".ualif;caticns

<<nc

". ai1i

~~:

r

~

~

)v ~ a

~ v Tne ST-'. shal)l

".ave a oacne)or's "egree or e" iva!ent

<<qg'~r

)<<

vI

~

~

~ Q v) <<4 v

~

~I ~

l'l <<Scl e1tl; 3.'..2

.ne S

A snail "e ra?ac in:

~ ne response ana ana iys is of:1e p:.ant for

=.- ns:ants anc acciaents.

Ce ails of the desi"n,

.unction, r; 1"amen

<<nc er~

, ns olant

systems, inciuaing

ne caoaollities of ;nstr ~entat.on

<<nc controls in:he C"ntr"l Rcc....

8.1.3 S)A initial quaiification and re-aualifica:lon shai) =e:er:,f'iea writinc by t1e Tec1nical Starf Tra ning coorcinatcr,

.ec.".ni "eoar:went SuoervlSor anc:ne

?'ant

'!an-ce.

-.'.uc.e<<r;-on

-".;.,= <<t

.ua

) 1 f1 cat? ons.

r<< Ir n 3 o 2

>cc'e 1 Asses s,ent 3.2. '.

.ne primary concern

,or

<<Caompl?Snea

=ur'. ng of-.-n

e rmi nate -r asi(

. e

tne, sarety Qy pr Qv. "; 10 orbal events ill i gate:le S'A iS O

?I "V'Ce

<<n

?nce~e1C81 4r

.ne,

.urkeg

?oint

?'n:.

ai=cnostic"suooort oe. =:.;"ns per ;n anc 5y advising t1e

.')PS on ac=::ns onsecue1c s

QT s~ucn eve 1'ts.

ple

ADMINISTRATI'/E PROC DURE 0103.16, PAGE A'lO RESP liS'.B,LI '"S OF THE SHIFT 7

HHICAI AD'I'.SOR 11l 6/81 8.2.2 ne role of tne ST@ is to serve in an acvisory capacity only, and 1ot

~

~

T t

i to'assume any command

."r contr" r.nc i:ns.

8.2.3 To acc'moi:sh tnis furctlon, "ne 5 n 5'louie -nysicaliy ema'.n

~i:.".in aO area wnich wiii allow nim

-o oe avallaoie to tl e Contr@i

Rccm, prereraoly immediately, out at the most wltnin tan minutes.

8 + 2 t+

8.2.5 Tne S

A snouid iemai1 in tne Control Rocm during the course of:he acciaent i1 Qr er to assess vi:ai cora parameters to ensure sa;ety of the reactor.

i-;e snould not oecome invol ied witn administrative and pnone c

1 ling duties.

off-shi.t STA wiil ac as liason "etween tne Control Rocm ana;SC when tne TSC is ac:ivatad ana manned.

e ~

~ I

ne 5"..

';S l eSOOnSiOle f".r assurl1C:lar.

AotlTlcatl on 5 cop as oos s i o 1 a el.d in a'.

1

".cc rrence of ary ".f ne foi assist nc :ne ?1'r Manage.

.'luc ear is mace to tne

.'<RC "oera:ions Canter as cases

~i:nin one 1cur oy a eonone

-..".e towing 5'~nl. 'nt eve155

~

'ny event requiring ini:iation o= the licensee's amer",ency plan or any sec.ion o-the pian.

2.

Tne exceeding of any Technical Specirication Safety

.'mit.

3.

~y event tna: results in.ne nuclear power plant not 'eing

n

~

ContrOIiea Or eXpeC=aa COOCl:ian ~r...ia accreting Or Snut

~Own.

rV ac 1$ r Qraatans ne saf af'V sita personnei, or tne sec rlty 11C. JQ. llo 115 ACe5

~

Sao

he 1uc'.a=r power.'":."r s "ec i'a i 1uc iear 1a r.' i, mo c saoo" ge.

0 ~

Any even.

reauiring nitiation or shut"cwn of:he nuclear power plant in accordance with

ecnnical SoecIfica:

on

.mi:;n=

. Conai:ions for Ooerat.'on.

6.

Pers "nnei error or procedure i

inadequacy

~ni ch, during noma i OOerar lOnS, anti Ci pataa OOerat i Ora i "C" r"enCeS,

".r aCCi cent COACltlOAS, OraVentS;r Cauld praVen:

Oy l

ai.

OT tne SaT ty funC lOA OT thase 5

~ "C Jr 5, Sys

ems, ana components important to safety that are neeaea to,'.,'nutcown tne reactor safely anc maint in i: in a

sare s1u:"own c"na =lcn or

',2) remove 145'icudl

.'lea't ro I I owing a

r sn" own I

limit tne reiaasa of racioacc'.ve maseral:o acceota" ie '.eveis;r reauCe

..le patent;ai far SuC1 reieaSe.

'ny ava1 E1gineee en Sys em.

l'as J 1

. 1c

',lanua 1

9r au file '

ac

'a 1

afaty

FeatJreS,

'1Ctuaing the ReaCtar P~".tact."n v ~

Any accicental, Jnplannea, or unccntroiiea racoact ve re!e.se.

,:lOnal eXOeCtan rai eaSaS

-. "".m

-.a!l.tananCe

.r

t.".e.

"oerational ac:ivl:ies are not incivaeaj.

APZ[S[STRATt'/E PROCEDURE O1O3 r.>i PAGE GUT) =S A'<[i RESPQNS [3 ['

'ES OF 7'"'.E SI'FT, EC:":N[CAL AOV f SOR

'/5/Sl 9.

Any fatality Or SerlauS lngury OCCurring On tne Site ana reauirlng transport to n off-site

-. ciil g

.Cr treatment,

[0.

Any serious perscnnei

.-"l-active "cntamlration re""',ring extensive on-site aecontaminat; n or outslce ass s ance.

i.l.

Any event meeting the criteri a of LO CFR Para.

29. '03

=or notific tion.

'.2.

Strikes of operating 4mpioyees or sec"ri:J guards, or honorirlg cf picket lines oy t1ese ~pioyees.

13

[T one or more

= lS

( rea r pnone extensions is fauna:o "e

inoperaoie, the

.'tRC Cperations Center must be notified <<it1in One nour oy ne oest availaoie means (an ooer able

=.NS 4x-. ns cn, corrrrrercial

eiepnone, cisoatcner pncre

".o Miami, relay

-." 'I""

etc.>

4 ~

he S,A ma r "e

r4ielvea cf 1'.5 CC',uen 55 ssm resoonsioi'.ities Qniy "y a "uallfiea S.A.

B.3 Coe. sting Excerience Assessment Func:ion:

8 ~ 3 ~

8.3.2 The plant operating exnerience assessment func:ion Is ver-.ormec as

>oint effor-.

"e"<<een memcers of :1e plant

.ecnni"ai

.= ar=,

anc

~Power Resour'ces

.'Iuc'ear S- -.-,,

and tne STA.

me

'oiiowing are "uiceii1es Tor tne STAin tne performance;f 1

. 4saons ill1 1 lle5:

.'IG:=:

Cec wnere

nOted, i 1i <<i al l ina

"." i na.; Cata ev'. 4 tne oeiow aocuments is nct re'u're".

1

<< ~

Coerator l.ogs - During coani zant of eaui pnent 1 eaaS, Snift tur1Over procress

~

eaCn Shift,:1e STA Snaula r4v:

w anC Ce cut of servl Ce,

".roerS are -: Sc"nnec=ea sne ts

~

survei 11r'ce no soec 1

asti 2 ~

ce.lsee

-ve1t

<<e "cr'"5

'Rs;;

-u i:et. ns,

".ars, Not ces; ana pertinent NRC or u-. lity assessments of :ce.

experience.

ine STA wi 1 I review the out"ut o; ne Cpe. aticnal Moerience

.-eeaback Program ara "t1er inform t;Dn 5 =l 4r=ec

=.'ne Tec1nlcai

.epar.ment Supe.visor.

3.

jumper and ulsccnr e :ec

.eacs 1g.neer Supe. v,sor sna'

~4v aisconnec:

ieaa or caoies and

.'nuclear Plant u"erv scr as to eaui ment ana piant parame ers.

~ ie c1 sn

<<4,>>

~.4i..-

g 4 5

or>>v ce roi <<~ence nCw

-.1e ro~ueSt "r -"e 4

ll<<

Q 1 0

l ~

I 4 ~

Tne STA snail

",onitor the RCS

'e x

e ascertain i;

".ne scurce "f :ne

<<CS

eaxa=e

=ne invostlgation of :ne RCS 'eatage sour"e leaKage exce ds

'3.5 gpm.

nc4

5

<nc<<n or =5:5-nct

<nc<<n

<<nen

ACM[81STRATIVE PROCiOURE O'.G3.16, PAGc.

6 QL'Tj>ware o..

C"erat!".ns surveiliance es:;ng on his shif: ard any unsatisfac=cry resu's.

8.4.2 Reac or/Turbine Trios and Transients-The STA shoul d irrnedi ately assess al 1 reac.or trips and transients occurring on nis shi.t witn regar" to safety.;nis assessment should include (but not limited :o):

(1)

(2)

(3)

~

1 Sequence of ="vents Causes Plant

Response

Ccrrec.ive ac ion -aken to ensure plant safetv

'I;o!aticrs of.ac.".nice!

Soec.'fioat'.'ons anc/"r Sa=ety Prccedural tnadequacies/"-cui"ment "ut of Service I

e

~ "I i

.ne S.~

snoul 0 commence prepar'nc a

repcr:~

cn

=1!

Jf "5 'a i r

aonor...a events cccurrina on n.s sh.ft as scen as

". ac-.cab'.e af:er nis accident assessment dut.'es are completed.

This repcrt should be drafted by the STA or by another membe.

of the Technical Staf'. within 2< hours of.he event.

The

"cs wl n

STA should collec-cco es of all pertlne>>t strip

charts, da.a

sneets, etc.,

-.".cr.: :he ccntrol rocl"..

and i..c'.""e hem h

5 f'eyer

~

! he

~eccrt snoul "

': nc! uce enouar..'nf"r..a:

. on

. n the

-.".'. ', ow ng a>> eas to anabl e C.

S a nersorne Q JrCers ard he DVen and provide feedback if appropriate.

.':ne fcl lowing crite..a shou d

'e used wnen assembling per:nent;nfcrma"'.on acout the event:

~ Il(f>> 5~ ~ 'TQQ

~

Record tne init al cond:::"ns.

nis shcu'" 'nc'"ce

=ne ;.".:

nu.-..ber,

date,

-.'.';..e when he even:

>egad, arc

-."e r.'=

"cwe.

'evel wnen :he event

-egan.

ne uni= was

".ct a=

a s:eacy state power operation, given an explanat'on of =he status.

record the ef.ec:

cn :he ;n.'

a ocwe.

-ec c:icn cr ;n'-.

snut Gwn resul s

ncte

-..".e

.-.,e.ncd

.Sec a

., >ve

."'s

manual rip, aut"mat;c trip etc.'eccr" the

i...e

=na pcwen neduction was ini:iated and the tine linen

~ot shu==cwn cr desired recucea power

!evel was reacnec.

urino

~ecovenv record

ne t

-..e of startuo or 'cac incr as

."o:e

.".e ::"e when full pcwer was reached.

men:

2 :r".v."es a for.. ="r renorting -"'an-

-'bnor...al

.ccu1

'e"."Ps'

anQl".1NlSIRanTi'IE PRCCEQl.aAE O',03. jga PAGE 7

QU aES

/Q RESP'Q~(S f~;1 f T DIES QF THE cjl (cT kriss, iCL n y ', SCP

/2a/82 llf E\\lgaaw

,':are any

'<ncwn c"mpcnen

,"lures ano ne

-,a'.!ure

-oce, availaoi e.

Nhen available, note oossible procedural and oersonnel

errors, "esign/construction proolems and he plant department involved.

E HQO QF 01SCCVE2y State how ne prob'lem was d.scovered routine testing, visua:

obse.vation, department made the discovery

,"~0'tc.)

p s es e paLasG e a ~

~

~ aaC oQR$ a& 5 (routing,.es:inc, etc.j

".o=e wnich ol s,

RCQ's, Ma;ntenan cn-art ces pr"vice ava laole i,".==r.atior.

corcerning an I uncontrol release.

'emote

'.-,'ea?=n phys cs Sn;. =

Supe. visor on " :y case aaoit cnal incor.,ation l's reui red.

PERSG'ieEL INVGL'/ED eri

.'!ake note o

the cperaticns

",rouo =hat was on sh:-;: "urina the event, in case accit:onal incor

.,a icr. or clar.-.ic.-ion o; events is r ouireo.

a ~anyqrve tg

~ aa~ gras(

"escr'.be to the

..ulies:

extent

."ossi'"le wra:

"".r-ec-.:I~

ac:icns

<<en tacan at :ne:ine n-.

=ne ven '.

ACtllNaSTRATi'IE PPOCE"URE 0103

~ '.6, PAGE 8 QUTl=S.>"C ?ESP(,<Sl? )l rT1ES OF T'r'.E S'r'fFT TEr~i.lrL. iy. SC?

7/29/82 Shi f=

Cata from stri p

charts, logs, "ata

streets, et.,

shcul d be presented in a

concise anc usable fcr..

nis

.-.ay if.elude repictting data, iaoe',irg cata, c"..".oin;nc dat

-n a s:rc'.e."it 4tC ~

SIvIp 1 ec SyS eIvv )raw n 5

srCu '

'Le C

Ced in understanding tne even:.

3.

Radiological 'Ccnditfons -

The STA should be cognizant (on a shift basis) of he radiclogical conditions existing in the

piart, including any gasecus or. liquid re'.eases in progress.

4.

Generai Plant Conditions

,ne S

r'.

should be ale~

tc ard responsive toward any of the following:

Plant Efficiency Maintenance ttms

":ousa~ee4p ir.g

?rccecures "uality.assurance

?roolens noticed in any cf -nese areas snoulc be bf ought to

.he at=ention o

the appr"priate Cepar=".ent head and "uc'.ear

?'.ant Shi,- Superrisor.

mover and Reli4f:

ra

~

V ~ V ~

The of -goinc c reit.;cns

c anc shall STA sncul d f evi ew apci i cab 1 e pl ant

"'cement s and cetera.'ne sny of.-ncr...al srsten ccndit crs cr trends "ut a

"mover checklis-

'. ttachr..ent I

V ~ V ~ ~

ine off goinccv STA snail pass on to,he r4', 4r'ng

"--;'he 1~ sr-s

tus, e~pnas'.'zing system o.. -nol...a icrs arC =rencs

-ss-s

progress, etc.

V ~ V ~ V tcul f

he plant as "rac:i"al afta.

4 gtl ~ 0 v0 e nv v Q e y st 1 tlg should be conducted bv the relieving ST~

as scen aSSuing the Sni. t

~ n Or"e..

Yel

~.J Ccf'o.

"araneters and sta us.

V

~

~

~

ecn', rc cf 4ac.l work week I Incav

.,c

~

c 1

~ c

~ 4" 4

~ l'V

~

VV'n;f:

tufncver Sheets with:ne exceoticn cf tne

"..cs=

ent:.-.re4

'.;.'ays shoulld be forwarded to the SIR E..c neer Supervisor

-.or review and d',secsticn.

V ~ V ~ V

'n "e evefl c

5'cxness pef'sofia I

>.".Ier en(y, cr o.".er "rcc '. 4rI wn;:."

f4c I uces tne S-,>>

frcrI sssvrll no I

p: e

~,.

h' he~

s sh'f:,

tne S,i Eng neer Supe.v sor shculc be nfc~ed..ne of-.---cine snot i c af

. ange a

I 1 ef 'as i

I e e~or4 a-v:nc.

SrC..C "e ar~argec :" fil; :ne S.~,

pCS'.:

Cn ~i:.". n s: 'esS:

m I'lCI 'I S.

ADM[f<I STRATt'lE P ROC =DUP E 0163. 16, PAGc 9

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.";;D RESPOND.S!B.'L.";1=-S CF T~E S~:FT TZC~riIC;.."CV:S".P 7/29/>?

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ACHI'IISTRATIV PRCCE"URE 0103.16, PAGE 14 DUTIES A'lD RESPONDS!SILITIES OF THE SHIFT TECHrlICAL ADY.'SCR 7/29/82 rai ons ~hi.t on duty "urino =vent 'n=.,'r':

~ ne following pers"nnel were directly invo'lve" ~ith the event ard may be a" le supplement tne information in this report if clari ication is needed:

.'5.

~aaitional Caments:

1.2 POST-TRIP REVIEW - DATA AND INFORMATIONCAPABILITY 1.

Capability for assessing sequence of events (on-off indications)

~Res ense:

Brief description of equipment (e.g. plant computer, dedicat'ed

computer, strip chart)

Eciuioment The equipment utilized as the primary source of information for assessing the post trip review is the plant computer.

This computer is referred to as DDPS or Digital Data Processing System.

This equipment consists of a Data General Nova 800 central processor with 32 kilo bytes of internal memory.

Mass storage is performed by an ampex megastore with 750 kilo bytes of magnetic core memory.

In addition, a Diablo 1 mega byte disc drive provides backup mass storage and program loading capability.

rinputs to the central processor are serviced by four (0) digital input/output controllers, two per nuclear unit and four (4) wide range analog input systems also two per unit.

This peripheral equipment was manufactured by Computer Products.

uence of Events 2)

Sequence of Events inputs feed the central processor through the digital input/output controllers and output on the control room line printer which is shared between both units.

This output is nearly instantaneous, reflecting the real time events as they occur.

Two hundred and sixty-five (265) onmff indications can be supported on each nuclear unit by this hard~are.

Two hundred and thirtyweven (237) are active on Unit 3 and two

'undred and thirty-eight (238) are active on Unit 0.

PC/Ms and controlled P%'Os necessitate frequent additions and reassignments of input channels.

Parameters Monitored The Sequence of Events parameters that are monitored by the plant computer are listed on the attached pages marked, "Digital Channels".

All of the digital input channels on the computer are responded to by the control room line printer.

In addition to the digital inputs, the computer software also displays events of internal nature.

These are listed in the digital channels'ages with an "X"prefix on the channel number.

3)

Time Discrimination Between Events Printed events are time tagged with a time discrimination of.Ol seconds.

That is, events occurring within the same

.01 second will share the same time tag.

Format for displaying data and information In addition to a time tag, the Sequence of Events channel number is displayed.

Alarm status, that is, alarm or clear and the channel description (event) are also printed.

Allof this information is printed on a single line situated in the left half of the printed page to indicate Unit 3 events or situated to the right half of the page for Unit 0.

5)

C-pability for retention of data'and information 6)

The hard copy sequence of events print-out is administratively routed to Document Control for storage following the review of the trip (O.P.

0208.1); there is no software retention of sequence of events.

Power source(s) 4e.g., Class IE, Non-Class IE, non-interruptable)

The power source for the plant computer supporting both sequence of events and analog variables is a dedicated inverter.

This inverter is powered from the DC Bus. It is non-Class IE and its output feeds through a static switch.

This switch senses power interruptions and switches input from a second source (Breaker 00530 on the Unit 0 A Bus).

This switching may also be performed manually for required maintenance on the inverter or DC breaker.

The plant computer's central processor is programmed to enable recovery and resume operation following power interruptions.

Static switch operation will only interfere with computer function for a short period of time.

2.

Capability for assessing

" the time history of analog variables needed to determine the cause of unscheduled reactor shutdowns, and the functioning of safety-related equipment.

~Res onse:

I)

Brief description of equipment (e.g., plant computer, dedicated

computer, strip charts)

Covered in the response to l.l.

2)

Parameters monitored, sampling rate, and basis for selecting parameters and sampling rate The analog variables which are monitored by the plant computer are inputs to the

%'ide Range Analog Input Systems (WRAIS).

The WRAIS has reference voltage inputs which provide for automatic scale ranging and the reference voltage inputs may be displayed as a quick indication of analog to digital conversion errors.

There are two Wide Range Analog Input

,Systems per unit with each one accepting 128 input channels.,

Many other analog parameters are monitored by strip chart recorders.

There are to be forty-four (00) recorders providing retention of data and information.

There is an attached list of analog variables which are monitored by the plant computer.

This list includes the DDPS channel

number, sampling rate, and signal name.

Another list includes the analog variables which are displayed and retained by strip chart recorders.

The basis for selecting parameters monitored on DDPS was the consensus of qualified personnel engaged in the original development of the specifications for the Digital Data Processing System.

Along with the monitored parameters determination, the sampling rates were determined.

"The main considerations being the maximum expected rate of change of the parameter and also data base size limitations.

Assignment of parameters monitored and sampling rates is available for

,inspection by all qualified plant personnel by the distribution of a reference document "DDPS Commands Summary".

Reassignment of parameters monitored and sampling rates has been in support of PC/Ms and controlled P WOs.

3)

Duration of time history (minutes before trip and minutes after trip)

A program of the plant computer called "Post Trip Review" maintains a

data base for each unit.

The Post Trip Review Program along. with the data base allows for retention and recall of data and information irom five (5) minutes prior to a trip until three (3) minutes following a trip.

Information can only be recalled one channel at a time.

The format for displaying data is as follows:

The output is displayed on the applicable unit's CRT and if a hard copy is desired, a single keyboard key initiates the line printer output.

The output is printed on the single line printer in the dual unit control room.

The right side of the page is used for Unit 0 and

the left side of the page is used for Unit 3. Individual channel requests are separated by a row of asterisks

( ). A header line follows in the form:

UNIT // PTR CHANVEL /P///I The time tagged data lines follow the header line.

The time tag displays the number of seconds before the trip time, preceeded by a minus (-) sign or no sign with seconds following a trip.

The time tag refers to the'first sampled record following the time tag.

Three more records" follow on the same line, displayed with their analog readings in E format.

In this

manner, every fourth reading has a time tag and the total number of readings depends upon the sampling rate.

0)

Format for displaying data including scale (readability) of time histories Covered in the response to 2.3.

5)

Capability for retention of data, Information and physical evidence (both hardware and software)

The DDPS Post Trip Review data base is retained in magnetic core

'memory.

This mass storage memory device retains its contents in the event of a total interruption of power.

The data base itself is only lost'hen it is released by a keyboard command by reactor control operators prior to unit startup.

The capability exists for copying the post trip review data base on a removable magnetic disc.

This, however, has not been found to be necessary and would require removing the computer from service.

Storage of pertinent post trip review printouts is covered in Operating Procedure 0208.1.

It is routed through the Operations Supervisor and stored in Document Control.

6)

Power source(s) (e.g., Class IE, non-Class IE, non-interruptable)

Covered in the response to 1.6 3.

Other data and information provided to assess the cause of unscheduled reactor shutdowns.

Resoonse:

Other available data present for the operating staff to evaluate the causes o.

a reactor trip include trend recorders, bistable status lights and the plant annunciation system.

The following recorders provide real time trends for evaluation:

Reactor Power:

NR-*-45 Nuclear Power Recorder NR-~-46 Overpower Recorder NR-*-47 Overpower Recorder Reactor Coolant System:

TR-*-420 RCP Status Recorder FR-~-154B RCP Seal Leak-off Recorder - Hi Range TR-+-423 Loop A Temp Recorder TR-~-424 Loop B Temp Recorder TR-*-425 Loop C

Temp Recorder PR-*-444 Pressurizer Pressure Recorder

Secondary System:

TR- -454 A S/G Level Flow and Feed TR- -464 B S/G Level Flow and Feed TR- -474 C S/G Level Flow and Feed TR-~-444 Turbine Valve Position TR- -441 Turbine Vibration Recorder These recorders allow hard copy of the transient.

Real time instrument channel indications are located on the console, vertical panels and on the Nuclear Instrument Cabinets.

These provide tne operator with the status of the total plant.

Schedule for any planned changes to existing data and information capability.

~Res once:

In approximately one year, a new central processor will replace the DDPS Data General Nova.

At about the same time, an additional processing system will carry out the Safety Assessment System (SAS) functions.

Along with these modifications, the existing sequence of events channels on DDPS wiO remain essentially unchanged but many more sequence of events parameters will be added with the SAS.

The channels remaining in DDPS will become much more accessible with the use of four (4) Post Trip Review software files.

In this implementation, the operations "Release Post Trip Review" command will no longer be necessary.

The software retention of data willbe much improved.

The exact implementation of retrieval of Post Trip Review data from the Safety Assessment System has not been finalized at the time.

Data and information will be stored'n magnetic tape but format of output, that is, printed on Hne printer or played back in a simulated real time display has not been decided.

SE "l:ENCE-OF-EVE.'ITS F"NCTICN The first function of the DDPS is to sense, and report on the 'ine

printer, any of a defined set of discrete "events".

An event

-...ay be the result of a contact change-of-state, an operator

command, or detection of an unusual condition by the program.

=ach event is identified by a unique event

number, Associated with each recorded event is a current state, which is either "alarm" or "clear".

The event numbers which may be reported are:

001-256 One of a possible 256 contacts has changed state from open to closed, or vice-versa.

257-264 One of a possible 8 AC voltages has dropped below 90 V r.m.s.

265 The crystal oscillator which provides a time base for computer operations has failed, and a back-up oscil-lator has taken over.

C001-C265 One of the above inputs has been either inhibited from further event reporting, or re-enabled for reporting after having been previously inhibited.

A001-A256 One of the possible 256 analog inputs has been either inhibited from further alarm checking, or re-enabled for checking after having been previously inhibited.

X001-X006 Some unusual condition has been detected by the program.

An English-language text string is associa,ted with each event number.

Each event is reported as a single printed line containing the event

number, the associated text label, and the work "alarm" or "c'ear",

indicating the current status of the input.

SOE SCAN Once every 10 milliseconds all the digital inputs corresponcing to events 1-:56 above are examined for a change of state.

For each input "hich has changed state s'nce the previous scan, an event

's reported.

The assignment of an "alarm" or.'-'Clear" indmat.ion to an "open" or "closed" contact state is made.i'ndependently for. each discrete input.

The collection of these associations forms a part of the program's permanent data base.

The DDPS I/O LIST, lists for each event number the corresponding discriptive label and the physical input state interpreted oy the program as an "alarm" condition.

If a given input changes state too frequently in a short period of time, it will automatically be inhibited from further'eport-ing.

In such a case, a "C"-type event will be recorded, with state

~ "clear".

To re-enable the input for event detection, an ENABLE CCI corn...and must be issued from the CRT texminal.

EXP BLE/INHIBIT CCI Each time a CCI (Contact Closure Input) is inhibited from, or re-enabled for, SOE reporting, an event is generated.

The Report consists of:

The letter "C" followed by the input number (1-256) of the affected digital input.

The identifying text belonging to the addressed SOE

~ number.

The word "Alarm" if the input is enabled for reporting, or "Clear" if reporting is inhibited.

ENABLE/INHIBITA/D Each tine an A/D (Analog-to-Digital Converter) input is inhibited from, or enabled for, limit checking an event is generated, The report consists of:

The letter "A" followed by the channel number (1-256) of the affected A/D input.

The identifying text belonging to the addressed A/D input.

The word "Alarm" if the input is enabled for limit checking, or "clear" if checking is inhibited.

MISCELLANEOUS EVENTS A few special conditions will cause "events" to be recorded.

Each such event is assigned a number beginning with the letter "X" and a self-explanatory text label.

Although the word "alarm" or "clear" is printed, this state indication has no significance for these events.

The following is a complete catalog of "X" type events:

X001 X002 EVENT BUFFER FULL.

One or more detected events have gone unreported.

(Because too many events occurred within a very short time span.

)

PTR FILE FROZEN.

The Post-Trip Review file (an 8 minute history of A/D data) has been locked, and may be examined with with the Post-Trip Review program.

X003 PTR'ILE RELEASED. The Post-Trip Review file has been freed for data gathering.

In this state, the file contents are not subject to examination.

X004 A/D FAILURE, CH 1-128.

The A/D converter containing the low-numbered channels has failed to deliver data when requested.

X005 A/D FAILURE, CH 129-256.

The A/D converter containing the high-numbered channels has malfunctioned.

X006 DATE-TlitE CHANGE.

This event is generated whenever the program's time base is altered (by operator command).

ANALOG DATA INPUT The DDPS can handle up to 256 channels (per unit) of analog inputs, coming into two multiplexed analog-to-digital converters.

Each input is identified by a channel number (1-256),

and has associated with it an English-language label and several items'of control information.

For purposes of discussion, it is convenient to recognize the follow-ing input groups:

Plant-monitor channels.

(Approximately 130 channesl.)

Each..

input in this group is sampled either once a second, every ten seconds, or once a minute.

Every value input is written to the Post-Trip Review file on disc, unless that file happens to be frozen.

Reactor Core Analysis channels.

(Approximately 70 channels.)

These inputs are applicable only to flux mapping, and are sampled only upon specific request.

A/D Calibration channels.

(16 channels)

There are reference voltage inputs used for self-calibration of the A/D converters.

They are sampled once every 20 minutes.

Unimplemented and Spare channels.

(Approximately 40 channels.)

These channels are never scanned.

DATA RECORDING A circulating disc file, the Post Trip Review File, is used to maintain an eight-minute history of all readings taI en of A/D channels in the "plant monitor" category.

Each reading is written to the disc as it is received, along with the channel number and a time tag.

Data in this file may be selectively examined by means of the Post-Trip Review Program described in Se'ction 3.4.11.

The PTR file is locked i.e.,

no more data are written to it=

three minutes after closure of the "Freeze PTR" CCI is detected.

At the moment the contact closure is detected, a blinking message "PTR FREEZE PENDING" is flashed on the CRT screen.

When the file is actually locked three minutes later, an event is logged on the line printer.

The file is unlocked--i.e.,

data output resumes when either:

1)

The command "RLS" is entered on the CRT terminal, or 2)

The "Release PTR" CCI changes from open to closed.

whenever the file is released after having been frozen, it is "wiped clean" I.e., data gathered prior to the freeze are completely lost, even if the file is ic~ediately re-frozen.

This is not true, however, if a "RLS" command is issued after the CCI trip has occurred but before it has been frozen.

In this case, the impending lock-up is simply cancelled, and the file retains a full eight minutes worth of data.

~

~

TURKEY POINT PLANT DDPS SCAN RATES TABLE 1 A/D CH!/

SIGNAL SCAN RATE 9

10 14 15 16 17 18 19 20 21 22 23 24 25 26 27 30 0 MV REF 2

MV REF 8 MV REF 32 MV REF 128 'iV REF 512 MV REF 2048 MV REF 8192 NV REF CHARGING FLOW F-122 CHARGING PRESS P-121 VCT LEVEL L-115 VCT TAP T-116 VTC PRESS P-117 RC FLOW LP A CH1 RC FLOW LP A CH2 RC FLOW LP A CH3 T AVE LP A CONTROL T AVE LP A PROT OVER TAP SP LP A OVERPOWER SP LP A DELTA T LP A CONTROL DELTA T LP A PROT RC FLOW LP B CH1 RC FLOW LP B CH2 RC FLOW LP B CH3 T AVE LP B CONTROL T AVE LP B PROT OVERTEMP SP LP B

OVERPOWER SP LP B

DELTA T LP B CONTROL 20 MIN 20 MIN 20 MIN 20 iiIN

-20 MIN 20 MIN 20 MIN 20 MIN 1 iiIN 1 iMIN 1 'MIN 1 MIN 1 MIN 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC

A/D CH SIGNAL SCAN RATE 31 32 33 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 DELTA T LP B

PROT RC FLOW LP C CH1 RC FLOW LP C

CH2 RC FLOW LP C CH3 T AVE LP C CONTROL T AVE LP C PROT OVERTOP SP LP C

OVERPOWER SP LP C

DELTA T LP C CONTROL DELTA T LP C PROT PZR LEVEL CHAN 1 PZR LEVEL CHAN 2 PZR LEVEL CHAN 3 PZR LEVEL WIDE RANGE PZR PRESS CH1 PZR PRESS CH2 PZR PRESS CH3 PZR PRESS LOOP 444 PZR PRESS LOOP 445 TURB FIRST STS PR CH3 TURB FIRST STG PR CH4 PZR LIQUID TPP PZR STEAM TEMP PZR SURGE LINE TGQ'ZR SPRAY TEMP LP B

PZR SPRAY TBQ'P C

ROD BANK A POSITION ROD BANK B POSITION RAD BANK C POSITION ROD BANK D POSITION RCS WIDE RANGE PRESS AUCTIONEERED T AVE AUCTIONEERED DELTA T T REF FEED FLOW LOOP A CH3 FEED FLOR LOOP A CH4 10 SEC.

10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC

.10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 1

SEC 1

SEC 1

SEC 10 SEC 10 SEC 1

SEC 1

SEC 1 MIN 1 MIN 1 MIN 1 MIN 1 MIN 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC

A/D CH 8

SIGNAL SCAN RATE 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 8S 86 87 88 89 90 91 92 93 94 95 96 97 98 SREAM FLOW LOOP A CH3 STEAM FLOW LOOP A CH4 STEAM GEN A LEVEL CHl STEAM GEN A LEVEL CH2 STEAMa GEN A LEVEL CH3 STEAlf GEN A LEVEL WR STBN GEN A PRESS CH2 STEAM GEM A PRESS CH3 STEA f GEN A PRESS CH4 STEAM HDR PRESS CH2 STEAM HDR PRESS CH3 STEAM HDR PRESS CH4 STEAM GEN BLOWDOWN FLOW FEED FLOW LOOP B CH3 FEED FLOW LOOP B CH4 STEAM FLOW LOOP B CH3 STEAM FLOW LOOP B CH4 STEAM GEN B LEVEL CHl STEAM FEN B LEVEL CH2 STEAM GEN B LEVEL CH3 STEAM GEN B LEVEL WR STEAM GEM B PRESS CH2 STEAM GEN B PRESS CH 3 STEAM GEN B PRESS CH4 FEED FLOW LOOP C CH3 FEED FLOW LOOP CH CH4 STEAM FLOW LOOP C CH3 STEAM FLOW LOOP C CH4 STEAM FEN C LEVEL CH1 STEAM FEN C LEVEL CH2 STEAM GEN C LEVEL CH3 STEAM FEN CL PRESS CH2 10 SEC 10 SEC 10 SEC 10 10

..10 SEC SEC SEC 10 SEC 10" sEc 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 1 MIN 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC 10 SEC

A/D CHd SIGNAL SCAN.

RATE 99 100 101 102 103 104 105 106 107 108 109 110 112 113 114 115 116 117 118 120 121 122 123 124 125 126 127 128 219 130 131 132 STEAM GEM C

PRESS CH3 STEA~1 GEN C

PRESS CH4 STER)

GEN C LEVEL WR CONTAINMENT SUMP LEVEL (SPARE)

(SPARE)

ABS CONDENSER PRESS CONTAINMENT PRESS WR CONTAINMENT PRESS WR N41 DET A CURRENT N41 DET B CURRENT N42 DET A CURRENT N42 DET B CURRENT N41 X POWER N42 X POWER N43 X POWER N44 X POWER N31 LEVEL N32 LEVEL N35 LEVEL N36 LEVEL FE 476 DIFF PRESSURE FEEDWATER PRESSURE STEAM PRESSURE LOOP A FE 486 DIFF PRESSURE STEAM PRESSURE LOOP B

FE 496 DIFF PRESSURE STEAM PRESSURE LOOP C

TURBINE CONTROL OIL PRESS TOTAL POWER - NUCLEAR 0 MV REF 2

MV REF 8 MV REF 32 MV REF 10 SEC 10 SEC 10 SEC 1 MIN SPECIAL SPECIAL 1 ~lIN 1 MIN 1 MIN 1

SEC 1

SEC 1

SEC 1

SEC 1

SEC 1

SEC 1

SEC 1

SEC 1 MIN 1 MIN 1 MIN 1 MIN 1 MIN 1 MIN 1 MIN 1 Mlh 1 MIN 1 MIN

~ ~

.1 MIN 1

SEC 1 MIN 20 MIN 20 MIN 20 MIN 20 MIN

0 A/D CH /I S ICiVAL SCAiV RATE 133 134 135 136 137 138 139 140 li1 142 143 144 148 149 150 151 152 128 MV REF 512

>iV REF 2048 NV REF 8192 MV REF GEN ~fECAWATT, REC GEN MEGAWATT IND.

T COLD LP A T HOT LP A T COLD LP B

T HOT LP B

T COLD LP C

T HOT LP C

FEEDWATER TEMP 1 FEEDWATER TBQ' GEN MEGAVARS IND N43 DET A CURRENT N43 DET B CURRENT N44 DET A CURRENT

.N44 DET B CURRENT D.D.P.S.

TC REF JCT 20.fIN 20 MIN 20 MIN 20 ifIN 1

SEC 1

SEC 1 MIN 1."fIN 1 MIN 1 MIN 1 ifIN 1 MIN 1 ifIN 1 MIN 1

SEC 1

SEC 1

SEC 1

SEC 1

SEC 1 MIN

DIGITALCHANNELS DDPS CH8'IGNALNAME 001 002 003 000 005 006 007 008 009 010 Oll 012 013 014 015 016 017 01S 019 020 021 022 023 020 025 026 CONTAINMEVTSPRAYS HI CONT. PR 10% CH 1

HI CONT. PR 10% CH 2 HI CONT. PR 1096 CH 3 HI CONT. PR 50% CH 1

HI CONT. PR 50',% CH 2 HI CONT. PR 50'5 CH 3 MAINSTM ISOL VLV A CL MAINSTiVl ISOL VLV B CL MAIiVSTM ISOL VLV C CL RPI ROD BOTTOM LO T AVE TO SI LOOP A LO T AVE TO SI LOOP B LO T AVE TO SI LOOP C RC LOW FLOW LOOP A CH 1

RC LOW FLOW LOOP A CH 2 RC LOW FLOW LOOP A CH 3 METALIMPACT MONITOR (UNIT 4 ONLY)

OVERPOWER DELTA T LP A OVERTEMP DELTA T LP A RC PUMP A OFF RC LOW FLOW LOOP B CH 1

RC LOW FLOW LOOP B CH 2 RC LOW FLOW LOOP B CH 3 OVERPOWER DELTA T LP B OVERTEMP DELTA T LP B R6:0

DIGITALCHANNELS DDPS CH 8 SIGNAL NAME 027 028 029 030 031 032 033 030 035 036 037 038 039 000 001 002 003 000 005 006 008 049 050 RC PUMP B OFF RC LOW FLOW LOOP C CH 1

RC LOW FLOW LOOP C CH 2 RC LOW FLOW LOOP C CH 3 OVERPOWER DELTA T LP C OVERTEMP DELTA T LP C RC PUMP C OFF PZR LO PRESS TRIP CH 1

P2R LO PRESS TRIP CH 2 PZR LO PRESS TRIP CH 3 P2R HI PRESS TRIP CH 1

PZR HI PRESS TRIP CH 2 PZR HI PRESS TRIP CH 3 PZR HI LVLTRIP CH 1

P2R HI LVLTRIP CH 2 PZR Hl LVLTRIP CH 3 PZR LO PRESS TO Sl CH 1

PZR LO PRESS TO SI CH 2 PZR LO PRESS TO Sl CH 3 PZR LO LVI. TO SI CH 1

PZR LO LVL.TO SI CH 2 PZR LO LVL TO SI CH 3 PZR LO LVL AND PRESS SI Sl UNBLOCKED DIGITALCHANNELS DDPS CH8'IGNALNAME 051 052 053 054 055 056 057 058 1

059 060 061 062 063 064 065 066 067 068 069 070 071 072 073 074 STEAMLINE SI UNBLOCKED AUTO SI MANUALSI PZR PR >2000 CH 1

PZR PR >2000 CH 2 PZR PR >2000 CH 3 Sl BLOCK MAiV PB SI UNBLOCK XtAN PB SI MANUALPB RT 1 AiVD2 RELAYS RT 3 AND 4 RELAYS RT 5 AND 6 RELAYS RT 7 AND 8 RELAYS RT 9 AND 10 RELAYS N 31 HI LEVEL TRIP N 32 HI LEVEL TRIP SOURCE RANGE iVIS BLOCKED N 35 HI LEVEL TRIP N 36 HI LEVEL TRIP INT RANGE NIS BLOCKED 41 LOW RANGE TRIP iV 42 LOW RAiVGE TRIP V 43 LOW RAiVGE TRIP N 44 LOW RANGE TRIP NIS POWER RANGE BLOCKED R6:4

DIGITALCHANNELS DDPS CH 0 SIGNAL NAME 076 077 018 079 080 081 082 083 086 087 088 089 990 091 092 093 090 095 096 097 099 100 101 N 41 HI RANGE TRIP iV 42 HI RAiVGE TRIP N 43 HI RAiVGE TRIP iV 04 HI RAiVGE TRIP POWER ABOVE P 10 START UVTD RELAY TEST CONT ISOL PHASE A CONT ISOL PHASE B GENERATOR LEADS BACK-UP RELAY TRIP GENERATOR BACK-UP DISTAiVCE RELAY TRIP PC006 A1X ABOVE 10%

PC407 E1X ABOVE 10%

REVERSE POWER RELAY TRIP GENERATOR OVEREXCITATIONRELAY TRIP UVTD-RX TRIP LOW 0 KV UVAKVBUS A CH 1

UVAKVBUS A CH 2 UV-4KV BUS B CH 1

UVAKVBUS B CH 2 iRIS ROD DROP TURB LOAD LIMITRUNBACK DELE' TURBINE RUNBACK MANUALPB RX TRIP CONS MAiVUALPB RX TRIP VP B

/

i&l~iFREQ TO RCP A UNDERFREQ TO RCP B AiVD C R6:0

DIGITALCHANNEI.S DDPS CH8 SIGNAL NAME 102 103 104 105 106 107 IOS 109 110 l)2 113 115 116 117 119 120 121 122 123 120 125 126 WARN-GEV STATOR DELTA TEMP Hl STEAM GEV A LO LEVEL CH 1

STEAM GEV A LO LEVEL CH 2 STEAiM GEN A STM > FW CH 0 STEAM GEN A STM > FW CH 0 STM GEN A HI LVL TRIP CH 1

STM GEiN A Hl LVLTRIP CH 2 STM GEN A HI LVLTRIP CH 3 STEAM GEN A FW ISOLATION STM GEN A LOLO LEVEL CH 1

STM GEN A LOLO LEVEL CH 2 STM GEN A LOLO LEVEL CH 3 STEAM LINE A HI DlFF CH 2 STEAiM LINE A HI DIFF CH 3 STEAM LINE A HI DIFF CH 0 STEAM LINE A HI DIFF Sl STEAM LINE A HI FLOW CH 3 STEAM LINE A HI FLOW CH ~

STEAM LINE HI FLOW 2/3 STEAM LINE A LO PR TO SI LO T AVE OR STM PR TO SI MAINSTM ISOL LOOP A MAINSTM ISOL MAiV PB A STM GEN B LO LEVEL CH 1

STM GEN B LO LEVEL CH 2

DIGITALCHANNELS DDPS CH 0 SIGNAL NAME 127 128 129 130 131 132 133 13$

135 136 13?

138 139 100 142 103 105 107 148 149 150 151 STEAM CEN B STM > FW CH 3 STEA Vl GEN B STM > FW CH 0 STM GEN B HI LVLTRIP CH 1

STM GEN B HI LVLTRIP CH 2 STM GEN B Hl LVLTRIP CH 3 STM GEN B FW ISOLATION STM GEN B LOLO LVL CH 1

STM GEV B LOLO LVL CH 2 STM GEN B LOLO LVLCH 3 STEAM LINE B HI DIFF CH 2 STEAiV LINE B Hl DIFF CH 3 STEAM LINE B HI DIFF CH 0 STM LINE B Hl DIFF SI STM LINE B HI FLOW CH 3 STM LINE B HI FLOW CH 0 STM LINE B LO PR TO SI MAINSTM ISOL LOOP B MAINSTM ISOL MAN PB B STEAM GEN C LO LEVEL CH 1

STEAiul GEN C LO LEVEL CH 2 STEAM GEN C STM > FW CH 3 STEAM GEV C STM > FW CH 0 STM GEN C Hl LVLTRIP CH 1-.

STM GEN C HI LVL TRIP CH 2 STM GEN HI LVL TRIP CH 3 R6:0 DIGITALCHANNFLS DDPS CH 8 SIGNAL NAME 152 153 154 155 156 157 159 160 161 162 163 164 165 166 167 168 169 170 171 172 173 174 175 STEAM GEN C FW ISOLATION STEAM GEN C LOLO LVL CH 1

STEAM GEN C LOLO LVL CH 2 STEAM GEN C LOLO LVL CH 3 STEAM LINE C HI DIFF CH 2 STEAM LINE C Hl DIFF CH 3 STEAM LINE C HI DIFF CH 4 STEAM LINE C Hl DIFF SI STEAM LINE C HI FLOW CH 3 STEAM L!NE C Hl FLOW CH 4 STEAM LINE C LO PR TO Sl MAINSTM ISOL LOOP C MAINSTM ISOL MAiN PB C AUTO STOP OIL < 45 PSI UNDERFREQ LO RELAY OPER TURB AUTO STOP OIL CH 1

TURB AUTO STOP OIL CH 2 TURB AUTO STOP OIL CH 3 TURB AUTO STOP OIL 2/3 LEFT STOP VLV CLOSED RIGHT STOP VLV CLOSED TURB STOP VLVS CLOSED INT RANGE > P 6 POWER RANGE 1 > P S R6:4

DIGITALCHANNELS DDPS CH /P SIGNAL NAME 1?6 l?S 179 1S0 ISI 182 183 185 186 187 188 189 190 191 192 193 194 195 196 197 199 200 POWER RANGE 1 > P 10 INT RANGE 2 > P 6 POWER RAiVGE 2 > P 8 POWER RANGE 2 > P 10 POWER RANGE 3 > P 8 POWER RAiVGE 3 > P 10 POWER RANGE 4 > P 8 POWER RAiVGE 0 > P 10 POWER ABOVE P S RX TRIP BYPASS BKR A OPEN (FUTURE)

POWER ABOVE P 7 TIME CHECK TURB TRIP MAiN PB CONS RX TRIP BKR A OPEN RX TRIP BKR B OPEN THRUST BRG TRIP ARMED RX TRIP BYPASS BKR B OPEV (FUTURE)

GEN BKR A OPEiV GEN BKR B OPEiV OVERPOWER DELTA T TRIP OVERTEMP DELTA T TRIP P2R HI-LO PR TRIP 2/3 PZR HI LVL TRIP 2/3 iVIS POWER RANGE LO TRIP TURB OVERSPEED OIL TRIP

-1 1-

DIGITALCHANNELS DDPS CH8'IGNALNAME 201 202 203 200 205 206 207 208 209 210 211 212 213 210 215 216 217 218 219 220 221 222 223 220 THRUST BRG TURB TRIP LOW VACUUMTURB TRIP LOW BRG OIL TURB TRIP AUX XFMR DIFF'L TRIP A SG FEED PUMP OFF EXHST HD HI TEMP TRIP TURB AiVTIMOTORINC TRIP AUX XFMR BRKR A OPEN AUX XFMR BRKR B OPEN REHEAT OR INTERCEPT CL GEN iVEG SEQUENT TRIP GEN LOSS OF FIELD TRIP OVEREXCMATIONTRIP GEN DIFFERENTIAL TRIP GEN GROUND TRIP MAINXFMR DIFFEREiVTIAL MAIN XFMR GROUND SU XFMR DIFFERENTIAL RC LO FLOW LOOP A TRIP RC LO FLOW LOOP B TRIP RC LO FLOW LOOP C TRIP iVIS POWER RANCE HI TRIP STM.ZEN A LO FLO AND LVL. ~

STM GEiV A LOLO LVLTRIP

DIGITALCHANNELS DDPS CH0 SIGNAL NAME 225 226 227 228 229 230 231 232 233 230 235 236 237 238 239 200 241 202 203 STM GEN B LO FLO AiVDLVL STM GEN B LOLO LEVEL TRIP STM GEN C LO FLO AND LVL STM GEN C LOLO LVLTRIP AUX TRANSFORMER GROUND (FUTURE)

GENERATOR LOCKOUT RELAY TURBINE OVERSPEED PROT B SG FEED PUMP OFF MAINTRANS FAULT PRESS (FUTURE)

AUX TRAiVS FAULT PRESSURE (FUTURE)

S.U. BKR-A OR EMERG. S.U. BKR (FUTURE)

STARTUP TRANS BKR B (FUTURE)

S,U. TRANS FAULT PRESS (FUTURE)

S.U. TRAiVS GROUND (FUTURE)

SPARE SPARE C BUSS TRANS SUDDEN FAULT PRESSURE (FUTI'F C BUSS TRANS LOCKOUT (FUTURE)

C BUSS TRANS HI-TEMP (FUTURE) 1C BUSS LOCKOUT (FUTURE) 0 205 246 247 209 250 BKR 4ACOI (FUTURE)

BKR OAC16 (FUTURE)

SPARE SPARE SPARE SPARE e

~

R6 0.

DIGITALCHANNELS DDPS CH 0 SIGNAL NAME 251 252 253 250 255 256 257 25S 259 260 26l 262 263 260 265 X001 X002 X003 X000 X005 X006 X007 XOOS X009 Xolo SPARE SPARE SPARE SPARE SPARE SPARE INST AC BUS 3POS DOWN INVERTER POS SPARE DOWN INST AC BUS 3P07 DOWiV INVERTER P07 SPARE DOWN INST AC BUS 3P06 DOWN INVERTER P06 SPARE DOWN INST AC BUS 3P09 DOWN INVERTER P09 SPARE DOWiV A CLOCK FAILURE EVEiVTBUFFER FULL PTR FiLE FRO. EN PTR FILE RELEASED A/D FAILURE CH 1-12S A/D FAILURE CH 129-256 DATE TIME CHAiVCE MIDS INACTIVE SPARE SPARE SPARE

DIGITALCHANNELS DDPS CH 8 SIGNAL NAME XOI I XOI I X013 XOI!4 X015 X016 X017 SPARE SPARE SPARE SPARE SPARE SPARE SPARE

SIGNAL NAME T RECORDERS

,E

.'E Feed Flow Level, Wide Range Levels

.ls, Bearings and Stator Windings) tture Setpoint and Cooler Inlet and Outlet Temperatures earing Yibration Expansion and Thrust Bearing Positions ser Temperatures sd Cooling Water icy Containment Cooler aperatures tctlvity

DDPS CH 0 POST TRIP REVIEW ANALOGVARIABLES SCAN RATE SIGNAL NAME 009 010 Oll 012 013 014 015 016; 017 01S 019 020 021 022 023 024 025 026 I

027 02S 029 1 min I min 1 min 1 min I min 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec CHARGING FLOW F-122 CHARGING PRESSURE P-121 VCT LEVEL L-115 VCT TEMP T-116 VCT PRESS P-117 RC FLOW LP A CH 1

RC FLOW LP A CH 2 RC FLOW LP A CH 3 T AVE LP A CONTROL T AVE LP A PROT OVER TEMP SP LP A OVERPOWER SP LP A DELTA T LP A CONTROL DELTA T LP A PROT

'I RC FLOW LP B CH I RC FLOW LP B CH 2 RC FLOW LP B CH 3 T AVE LP B CONTROL T AVE LP B PROT OVERTEMP SP LP B OVERPOWER SP LP B

DDPS CH 8 POST TRIP REVIEW ANALOGVARIABLES SCAN RATE SIGNAL NAME 030 03l 032 033 034 035 036

037, 038 039 040 041 043 045 046 047 048 049 050 051 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec IO sec 10 sec 10 sec 10 sec 10 sec I sec I sec I sec 10 sec 10 sec I sec I sec DELTA T LP B CONTROL DELTA T LP B PROT RC FLOW LP C CH I RC FLOW LP C CH 2 RC FLOW LP C CH 3 T AVE LP C CONTROL' AVE LP C PROT OVERTEMP SP LP C OVERPOWER SP LP C DELTA T LP C CONTROL DELTA T LP C PROT PZR LEVEL CHAN I PZR LEVEL CHAN 2 PZR LEVEL CHAN 3 PZR LEVEL WIDE RANGE PZR PRESS CH I PZR PRESS CH 2 PZR PRESS CH 3 PZR PRESS LOOP 444 PZR PRESS LOOP 445 TURB FIRST STG PR CH 3 TURB FIRST STG PR CH 4

~ ~

POST TRIP REVIEW ANALOGVARIABLES DDPS CH 0 052 055 056 05?

058 059 060 061 062 063 060 065 066 067 068 069 070 071 072 073 Q7$

075 SCAN RATE I min I min I min I min I min 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec IO sec 10 sec 10 sec 10 sec IQ sec SIGNAL NAME PZR LIQUIDTEMP PZR STEAM TEMP PZR SURGE LINE TEMP PZR SPRAY TEMP LP B PZR SPRAY TEMP LP C ROD BANK A POSITION ROD BAiVKB POSITION ROD BAiilKC POSITION ROD BANK D POSITION RCS NARROW RAiVGE PRESS AUCTIONEERED T AVE AUCTIONEERED DELTA T T REF FEED FLOW LOOP A CH 3 FEED FLOW LOOP A CH 0 STEAM FLOW LOOP A CH 3 STEAM FLOW LOOP A CH 0 STEAM GEV A LEVEL CH I STEAM GEN A LEVEL CH 2 STEAM GEV A LEVEL CH 3 STEAM GEN A LEVEL WR STEAM GEN A PRESS CH 2 STEAM GEN A PRESS CH 3 STEAM GEN A PRESS CH 0 R6:0

-I9-

POST TRIP REVIEW hNALOC VARIABLES DDPS CH 8 076 077 078 079 080 081 082 083 084 085 086 087 088 089 090 091 092 093 090 SCAN RATE 10 sec 10 sec 10 sec 1 min 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec SICNAL NhME STEAM HDR PRES5 CH 2 STEAM HDR PRE55 CH 3 STEAM HDR PRESS CH 0 STEAM CEN BLOWDOWN FLOW FEED FLOW LOOP B CH 3 FEED FLOW LOOP B CH 0 STEAM FLOW LOOP B CH 3 STEAM FLOW LOOP B CH 0 STEAM GEiV B LEVEL CH 1

STEAM CEN B LEVEL CH 2 STEAM GEN B LEVEL CH 3 STEAM GEN B LEVEL WR STEAM GEN B PRESS CH 2 STEAM GEN B PRE55 CH 3 STEAM GEV B PRESS CH 4 FEED FLOW LOOP C CH 3 FEED FLOW LOOP C CH 0 STEAM FLOW LOOP C CH 3 STEAM FLOW LOOP C CH ~

R6:0

DDPS CH 0 POST TRIP REVIEW ANALOG VARIABLES SCAN RATE SIGNAL NAME 095 096 097 098 099 100 101 102

/

103 104 105 106 108 109 110 112 113 114 115 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec 10 sec I min.

10 sec I min I 'min I min I min I sec I sec I sec I sec I sec I sec I sec I sec STEAM GEN C LEVEL CH I STEAM GEN C LEVEL CH 2 STEAM GEN C LEVEL CH 3 STEAM GEiV C PRESS CH 2 STEAM GEiV C PRESS CH 3 STEAM GEiV C PRESS CH 4 STEAM GEN C LEVEL WIDE RANGE CONTAINMENTSUMP LEVEL SPARE INTAKECOOLING WATER TEMP ABS CONDENSER PRESS O.O.S.

O.O.S.

iV41 DET A CURRENT iV41 DET B CURRENT N42 DET A CURREVT N42 DET B CURRENT iV41 '5 POWER N42 8 POWER N43 6 POWER iV44 5 POWER R6:4

DDPS CH d POST TRIP REVIEW ANALOGVARIABLES SCAN RATE SIGNAL NAME 116 117 118 119 120 121 122 123 124 125 126 127 128 137 138 139 140 141 142 143 144 145 146 I min I min

.'I min I min I min I min I min I min I min I min I min I sec I min I sec I sec I min I min I min I min

1. min I min I min I min iV31 LEVEL iV32 LEVEL iV35 LEVEL iV36 LEVEL FE 476 DIFF PRESSURE FEEDWATER PRESSURE STEAM PRESSURE LOOP A FE 486 DIFF PRESSURE STEAM PRESSURE LOOP B FE 496 DIFF PRESSURE STEAM PRESSURE LOOP C TURBINE CONTROL OIL'RESSURE TOTAL POWER-NUCLEAR GEN MEGAWATT, REC GEN MEGAWATT, IND T COLD LP A T HOTLPA T COLD LP B T HOTLPB T COLD LP C T HOTLPC FEEDWATER TEMP I FEEDWATER TEMP 2

Screening of Orders Entered All orders entered are screened by the Westinghouse District Sales Offices, Distributors, Engineering Service Offices, or Repair Plants to.

determine whether or not safety-related nuclear equipment, parts or services are involved.

Our nuclear utilities have been informed that they must ensure that procurement documents issued by them or by their sub-contractors or agents to Mestinghouse clearly indicate if the equipment, parts or services are for a nuclear plant and ff so whether or not they are safety-related if they expect Mestinghouse to ensure that applicable HRC regulatory requirements are met.

If an order is received for equipment, parts or services for a nuclear plant without any indication of whether or not it is safety-related, then the customer w$ 'll be asked the fo11owtng questton:

Does the customer have any responsibilty under 10 CFR Part 21 for reporting defects, as defined in that regulation, which exist in the equipment, parts or services at the time of delivery or which may subsequently occur?

If the answer to this question is yes, the order is entered through an approved order entry channel, s>

If the answer to this question is no, normal commercial practices apply.

A proved Order Entry Channels All orders for equipment parts identified by the purchaser or through screening for safety-related applications are processed through the Mater Reactor Divisions (MRD). All safety-related service orders will be placed with divisions who are on an approved supplier list for the particular services requested.

If no approved supplier is listed for requested

services, the order will be processed through WRD.

gualification and Trainin NRC regulations place the responsibility for ensuring that regulatory requirements are met in the procurement of safety-related equipment, parts or services on the utility.

The utility is required to provide any special requirements to. those providing safety-related equipment, parts or services.

Westinghouse will ask for any special requirements or instructions from the customer when safety-related equipment, parts or services are involved and to ensure that the work done conforms to those spec ia I requirements or instruct ion s.

?nformin Customers - Instruction Books Instruction books are provided by divisions supplying equipment or parts.

Such instruction books include information necessary for proper and safe installation, operation, maintenance and repair in ordinary commercial applications of, such equipment and parts.

Westinghouse divisions providing services are to obtain copies of appropriate instruction books for the equipment and parts which they service from the divisions which originally supplied the equipment or parts.

WRD is to include in its equipment specifications and purchase orders for equipment and parts appropriate requirements for copies of instruction books for fts customers and for its own use in developing recommenda'tions to fts customers for installation, operation, maintenance and repair of the equipment and parts in nuclear applications.

Instruction books for safety-related equipment are subject to the quality assurance requirements of 10 CFR Part 50 Appendix B and are to be included on the quality assurance release which must accompany any delivery of safety-related equipment or parts.

Substantive errors discovered in instruction books for safety-related equipment after delivery to the customer are subject to the reporting requirements of IO CFR Part 2I.

Correction of errors may be accomplished by MRD Technical Bulletins in lieu of revising the instruction books or,.

in cases in which MRD is not involved, by supplementary information provided directly to its customers by the involved division.

!nformin Customers - Technical Bulletins

!nformation supplementing or revising the instruction books or other similat materials necessary for proper and safe installation, operation, maintenance or repair of MRD-supplied equipment or parts in nuclear applications. is provided by WRD in the form of Technical Bulletins.

Preparation of such Technical Bulletins has been centralized within MRD and is subject to the same design control process which applies to the equipment, parts and services to which they relate.

Technical Bulletins which are safety-related are so identified.

AII Technical Bulletins will be transmitted by MRD to every Westinghouse NSSS customer, domestic and international, and such other MRD customers as are affected.

Responsibility for this distribution has been centralized for customers with operating plants and for customers for plants not yet in operation in order to ensure that all Technical Bulletins are promptly distributed.

Customers have been requested to provide the necessary distribution lists for their organizations.

AII distributions of safety-related Technical Bulletins are now accompanied by a return receipt.

The return rece'pts are pre-addr essed to a central point in MRD for recording all Technical Bulletins transmitted and their status.

Technical Bulletins for which receipt is not acknowledged with a reasonable time are retransmitted.

A list of current Technica1 Bu11etins and Data Letters will be prepared and transmitted to all customers periodically but not less frequently tpan once per year.

This transmittal is to be in the form of a Technica1 Bu11etin and is to be transmitted in the same manner as any other Technical Bulletin.

Appendix A is Bulletin tsSD-TB-83-05 which gives the index of currently valid Technical Bulletins.

FPL Control of Vendor Maintenance Work FPL guality Procedures 4.1 (Control of Requisitions and the Issuance of Purchase Orders to Spare Parts, Replacement Items and Services) and 4.4 Review of Procurement Documents for Items and Services Other Than Spare Parts) provide

'a system to assure that the appropriate technical and quality d equirements are placed upon suppliers uho provide material, equipment and services for operating nuclear po~er plants.

Safety-related Purchase Orders for services are issued only to suppliers whose guality Assurance Program and implementing procedures have been evaluated and approved by the FPL guality Assurance Department.

Vendor maintenance work on reactor trip systems

. components must comply with applicable plant procedures.

Vendor implementing procedures prepared in accordance with the vendors gA program, wnich must be approved by FPL gA, may also be used.

vendor procedures used to perform on-site work must be approved by the Plant Nuclear Safety Committee.

2.1 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP SYSTEM COMPONENTS)

Position Licensees and applicants shall confirm that all components whose functioning is requi red to trip the reactor're identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related acti vities, including maintenance, work orders, and parts replacement.

In addition, for these components, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information is complete, current and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures.,

Vendors of these components should be contacted and an interface established.

Mhere vendors cannot be identified, have gone out of business, or will not supply the information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reactor trip system reliability.

The vendor interface program shall include periodic communication with vendors to

'ssure that all applicable information has been received.

The program should use a system of positive feedback with vendors for mailings containing technical information.

This could be accomplished by licensee acknowledgement for receipt of technical mailings.

The program shall also define the interface and division of responsibilities among the licensees and 4he nuclear and nonnuclear divisions of their vendors that provide service on reactor trip system components to assure that requisite control of and applicable instructions for maintenance work are provided.

Response

FPL has determined that at Turkey Point Units 3 and 4 components whose functioning is required to trip the reactor are included in systems which are treated as safety related for plant activities, such as maintenance, work orders and parts replacement.

Future changes or modifications of these systems will be reviewed by FPL engineering to ensure that the correct safety classification is made.

A description of the. information handling systems for component classification is provided in our response to action 2.2. 1.

In addition, a continuing program exists to receive and review vendor information.

It is the policy of Westinghouse Electric Corporation to be a reliable supplier of equipment,

parts, and services needed by FPL for use in nuclear power plants.

Orders for such equi pment, parts ana services for safety-related applications are to be given the special attention which is required by applicable regulatory requirements as

~~11 as commercial practices.

The following information summarizes pertinent parts of this Mestinghouse policy on the 1/customer interface:

Definition of Safet -Related E ui ent, Parts and Services Equi pment ~~

and services are safety-related if the utility customer

f5r whom the equipment, parts or services are intended indicates that he "has responsibility under Nuclear Regulatory Commission Regulation 10 CFR Part 21 for reporting any defects, as defined in that regulation, in the equipment, parts or services as delivered to him or which may subsequently occur.

EgUIPHENT CLASSIFICATION AND VENDOR INTERFACE (PROGRANS FOR ALL SAFETf-RELATED COMPONENTS}

Position C

Licensees and applicants shall submit, for staff review, a description of their programs for safety-related*

equipment classification and vendor interface as described below:

1.

For equipment classification, licensees and applicants shall describe their program for ensuring that all components of safety-related systems necessary for accomplishing required safety functions are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including 'maintenance, work orders and replacement parts.

This description shall include:

1.

The criteria for identifying components as safety-related within systems currently classified as safety-related.

This shall not be interpreted to require changes in safety classi7ication at the systems level.

2.

A description of the information handling system used to identify safety-related components (e.g., computerized equipment list) and the methods used for its development and validation.

3.

A description of the process by which station personnel use this information handling system to determine that an activity is safety-related and what procedures for maintenance, surveillance, parts replacement and other activities defined in the introduction to 10 CFR 50, Appendix B, apply to safety-related components.

4.

A description of the management controls utilized to verify

~.

.that the procedures for preparation, validation and routine utilization of the information handling system have been followed.

5.

A demonstration that appropri'ate design verification and qualification testing is specified for procurement of safety-related components.

The specifications shall include qualification testing for expected safety service conditions and provide support for the licensees'eceipt of testing documentatiOn to support the limits of life recommended by the supplier.

6.

Licensees and applicants need only to submit for staff review the equipment classification program for safety-related

'omponents.

Although not required to be submitted for staff review, your equipment classification program should also include the broader class of structures,

systems, and components important to safety required by GDC-1 (defined in 10 CFR Part 50, Appendix A, "General Design Criteria, Introduction" ).

2 ~

For vendor interface, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor information for safety-related components is complete, current and controlled throughout the life of their plants, and appr'opriately referenced or incorporated in plant instructions and procedures.

Vendors of safety-related equi pment should be contacted and an interface established.

Where vendors cannot be identified, have gone out of busines, or will not supply information, the licensee or applicant shall assure that sufficient'attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reliability commensurate wi th

.fts safety function (GQC-1).

The program shall be closely coupled with action 2.2.1 above (equipment qualification)

~

The program shall include periodic communication with vendors to assure that all applicable information has been received.

The program snould use a system of positive feedback with vendors for mailings containing technical information.

This could be accomplished by licensee acknowledgment for receipt'of technical mailings.

It shall also define the interface and division of responsibilities among the licensee and the nuclear and nonnuclear divisions of their vendors that provide service on safety-related equipment to assure that reqQisite control of and applicable instructions for maintenance work on safety-related equipment are provided.

Response

1.1 and 1.2 t'uality Procedure 2.7 (Identification of Safety-Related and Nuclear safety-Related gA Required Structures, Systems Components and Services) describes the FPL system for identifying nuclear safety-related structures,

systems, components.

A copy of thi s gP is attached.

The Turkey Point g-List is a system level document which has been generated apd maintained in accordance with this gP.

A program to increase tne specificity of the g-List is now underway.

Our initial contractor has prepared a detailed safety-related component list for the CVCS system, and other systems are expected to be completed by a vendor yet to be selected.

A draft of the general guidelines to be used for evaluating the quality classification of systems and structures for Turkey Point Units 3 5 4 is attached as an example of effort that is currently underway.

The final guidelines are planned to be provided with the contract to the vendor when selected.

The final product of this effort is expected to be a component level safety-related listing and drawings to designate g-boundaries.

1.5 Administrative procedure 0103.4 (In Plant Equipment Clearance Orders) requires that the equipment involved in a plant clearance be checked to determine if safety-related.

Administrative Procedure 0190.15 (Plant projects-Approval, Implementation and Regulatory Requirements) provides controls for safety-related Plant Changes/Modifications.

Administrative Procedure 0190.19 (Control of Maintenance on Nuclear Safety-Related and Fire Protection Systems) provides guidelines for procedural requirements on maintenance of nuclear safety-related and fire protection

systems, components and equipment.

The safety-related determination for the above procedures are made by reference to the Turkey Point g-List.

In addition the FPL guality Procedures as contained in the FPL guality Assurance Manual are followed to ensure that procedures for maintenance, surveillance, parts replacement and other activities defined in the introduction to 10 CFR 50, Appendix 8 apply to safety-related components.

The most recent revision to the FPL 'Topical guality, Assurance Report was found to be acceptable by the NRC on September 7,

1983.

Like all aspects of the t}A Program, procedures for preparation, validation and routine utilization of the safety-related identification documents are the subject of routine guality Assurance audits.

guality Procedure 4.1 provides a system to assure that the appropriate technical and quality requirements are placed upon suppliers who provide

material, equipment and services for FPL nuclear plants.

A copy of this guality Procedure is attached.

1.6 No response is required.

2.0 FPL is a member of the Nuclear Utility Task Action Committee (NUTAC) on Generic Letter 83-28, Section 2.2.2 formed on September 1, 1983, for the specific purpose of defining an appropriate vendor interface program.

At present, we intend to incorporate the results of the NUTAC.

Our schedule for submission of our program description was provided in our letter L-83-480, dated September 7, 1983.

This report is scheduled for submittal by February 29, 1984.

GENERAL GUIDELINES FOR Q-BOUNDARIES:

The following general guidelines shall be used in evaluating the Quality Classification of Systems and Structures for Tuzkey Point Units 3 5 4.

1.

Nuclear Safety Related a)

Nuclear Safety Related is defined as those structures, systems or components which are necessary to assure:

i) the integrity of the reactor coolant pressure boundary; ii) the capability to shut down the reactor and maintain it i'n a safe shutdovn condition; or iii) the capability to prevent or mitigate the consequences of accidents vhich could result in potential off-site exposures comparable to the guideline exposures of 10 CFR Part 100.

b)

Reactor coolant pressure boundary means all those pressure-containing components such as pressure

vessels, piping, pumps and valves Mhich. are:

i) part of the reactor coolant system or, ii) connected to the reactor coolant system up to and-including any and all of the folloving:

1.

The outermost containment isolation valve in system piping vhich penetrates primary reactor containment, 2.

The second of ~o valves normally closed during normal reactor operation in system piping vhich does not penetrate primary reactor containment, c) 3.

The reactor coolant system safety and relief valves.

A 1085 of capability to achieve a required safety function cited in Paragraph l(a) means the loss of a safety function to the extent that there is a major reduction in the degree of protection provided to public health and safety.

A major reduction in the degree of protection means:

i) loss of an item, in conjunction with a single fai$ure, results in the inability to perform the nuclear safety-related function utilizing equipment normally designed for the purpose Mithout the availability of offsite pover, or,

ii) a partial or total loss of function of a radioactive confinement system that compromises its ability to contain radioactive materials such that offsite exposures approach 25 rem whole body or 300 rem thyzoid, or iii) a partial or total loss of function of a radioactive confinement system that results in an operator exposure appro'aching 25 rem whole body or 300 rem thyroid during the performance of an operator action normally required to achieve a

safety function cited in Paragraph l(a) assuming an exposure period of 20 minutes.

d)

Safe shutdown means the reactor is in the hot shutdown condition, i.e., it is subcritical by an amount greater than or equal to a margin specified in the Technical Specifications and T

is above 540'F.

avg o e),.

h single failure means an occurrence which zesults in the loss of capability of a component to perform its intended function.

Multiple failures resulting from a single occurrence are considered to be a single failure.

hll items classified nuclear safety related shall be considered Q and shall meet the quality assurance requirements of the FPL QA Program.

Safety Related Design.'eature Safety Related Design Featuze is defined as those structures, systems or components which are not nuclear safety related and are in one or more of the following categories:

a)

Equipment, components and structures designed to meet seismic requirements or whose failure could:

i) damage safety-related equipment such that the equipment

~ould be prevented from performing its safety function, oz ii) resulted in releases exceeding the exposure guidelines of Part C} below.

b)

Fire protection equipment i) required to protect nuclear safety related equipment, or ii) whose failure could result in water damage to nuclear safety related equipment which could prevent the equipment from

• performing its safety function.

DRAPT c) h paztfal or total loss of function of a radioactive confinement system that results fn an accidental, unplanned, or uncontrolled release of radioactfvfty exceeding an:

i) exposure to an individual in an unrestricted area in any period of one calendar year of 0.5 zem ~hole body, oz ii) exposure to an fndfvidual in a restricted area of 5 zen while performing normal operator actions required to achieve and maintain safe shutdown conditions.

d)

Equipment whose failure under normal operating conditions or an anticipated transient. results fn i) exceeding a safety Ifmft specified in the Tcchnical Specifications, or ii) initiation of an FShR Design Basis hccident, or fif) results in thc reactor coolant system not being in a controlled or design condition while operating or shutdown.

e)

Equipment, components, or structures required to be opezable by the Technical Specifications 0

f)

Instrumentation that is essential to monitor emergency reactor

shutdown, containment isolation, reactor core cooling, and containmcnt and reactor heat removal, or otherwise are essential fn preventing release of radioactive material to the environment -.

which would exceed the guidelines of Part c) above.

Items classified as safety related design feature will be considered "Q" but need not meet the full quality assurance requirements of 10 CFR 50, hppendix B.

hpplicable quality assurance requirements. for items important to safety shall be specified in the FPL QA Program.

'3.

Functional and Seismic Classification:

For Class 1 systems functional boundazy The Q-level for the guidelfnes of Items and structures (seismic Category I) the entire and suppozt svstem shall be considezed "Q".

systems and structures shall be based on the.

1 and 2 above.

For determining Q-level for boundary areas and supports, wheneve~

nuclear safety related system has both Class 1 and Class 3 poftions, the "Q" functiunal boundary shall in most cases be taken td t5e first normally closed valve or valve capable of automatic closure.. The pipe suppozt system shall be considered nuclear safety related to the first anchor downstream of the boundary valve for stress analysed piping.

The piping downstream of the boundary valve to the seismic a~~hall'e considered important to safety since any physical modifications to this piping could affect the seismic analysis.

For 2" and smaller piping that is not stress

analyzed, the "Q" functional boundary ends at the isolation valve.

The 2" and smaller piping beyond the "Q" boundary must be supported per the small pipe installation manual M-18.

For systems evaluated for high energy line break (Florida Pover Light's response to hZC dated February 26, 1973), +here any support system or vhip restraint vhose failure could damage safety-related equipment, the support or vhip restraint shall be considered important to safety.

Containment Penetrations:

Mechanical and electrical containment penetrations for any system, whether essential or non-essential, shall be considered safety-related since they are required to keep accident doses less than the guidelines of Item l.a)iii).

FSAR Section 6.6 and Table 6.6-1 shall be the basis for determining containment mechanical penetration boundaries.

The pipe support system shall follov the same criteria ss listed in Paragraph 3.

Instrumentation:

For Instrumentation vithin the "Q" boundaries of a system, any portion of the instrument vhich forms a part of the system boundary shall be considered "Q" and shall meet the same "Q" requirements as the portion of the system in vhich it is located.

Hovever, the associated circuitry need not necessarily be "Q".

The associated circuitry shall be "Q" only if it performs a nuclear safety related function or is important to safety. per Paragraphs 1 and 2. If the instrument output function is only for normal operation and has no safety significance, then the electrical circuitry shall be considered non+.

An instrument that is required to perform an active function (initiation of safeguard equipment) is considered nuclear safety related.

Instrumentation used to monitor plant conditions and vhich serve as a basis for operator action vill be considered important to safety.

A reviev vill be performed to determine vhich instruments are required by the operator to achieve or maintain a safe shutdown.

Paver Actuated Valves:

For po~er operated valves vithin "Q" boundaries, the valve, operator, solenoid valve'nd associated circuitry shall be considered

".Q" and shall meet the same "Q" requirements as the portion of the svstem in vhich it is located.

Turkey Point's Instrument Air System is non-safety related and not important to safety, therefore, the air supply line is non-Q.

Air actuators shall be considered the same level of "Q" as the portion of the system in vhich they are located since they are required to ensure that the valve fails in the safe position.

There -are a few exceptions to the above criteria and they are as follows:

1 Air Reservoirs Some air operated valves are required to operate a minimum number of times after an accident, and in these

cases, they are provided with air reservoirs.

These reservoirs and associated tubing, controls, valves, etc., shall be considered the same "Q" level as the valve.

B.

Nitro en Back-u Su l - Some air operated valves and instrumentation are required to operate on an extended basis after an accident and in these cases a nitzogen bottle back-up supply is provided.

The Nitrogen Back-up Supply System shall be considered the same "Q" level as the component it serves.

Other Devices (Heat Exchangers,

'Pumps,

Fans, MOVS, etc.):

All components such as heat exchangers,

pumps, and other vessels that are within the "Q" functional boundaries and which are requized for the functions listed in Items 1 and 2 are considered "Q" and sh 11 meet the same "Q" requirements as the portion of the system in which they are located.

8.

Electrical Power Supplies:

Power supplies to any nuclear safety related equipment or equipment important to safety which aie required for the functions listed in Items 1 and 2 shall be considered "Q" and shall meet the same "Q"

requirements as the portion of the system to vhich they supply power.

9.

Structures:

The containment, auxiliary building, control building, diesel generator building, radwaste facility building, intake structure, portions of the tu'rbine building, and spent fuel pool building support equipment that is safety-zelated or important to safety.

These structures shal'eet the same "Q" requirements as the systems they support..

Anv structuze that could interact with nuclear safety-related equipment shall be considered important to safety.

10.

Piping and Valve Classification:

The design codes for piping and valves and how they were app'ied at Turkev Point requires some clarification.

At the time Turkev Point as designed, ASM'II Code for piping was not in use.

Therefore,

@SEE I and NSI B31.1 were used in the design with additional commitments such as seismic design and non-destructive testing (NDT) as indicated in FSAR Appendix 5A and Table 6.2-3.

The ASIDE I and ANSI B31.1 Codes are not, at present, applicable for piping used in safety systems oi nuclear power plants.

Specification of ANSI B31.1,. at present, is not

sufficient for safety-related

systems, since the required documenta-tion for seismic and non-destructive testing is beyond the scope of Code.

For most of the retrofit vork vhich Bechtel has designed und II un er Job Number 5177, the term upgraded B31.1" has been applied to safet related piping and valves, in order to extend the scope of ANSI B31,1 a ety to include the additional FSAR commitments.

In essence, "upgraded B31.1" generally follovs the requirements of the applicable class of pipe contained in ASME III and in some cases, the actual equipment purchased in ASME III since vendors do not recognise the meaning of "upgraded B31.1".

However, the PC/M makes no commitment to ASME III.

The equipment is installed per Upgraded B31.1.

POST-MAINTENANCETESTING (REACTOR TRIP SYSTEM COMPONENTS l.

Licensees and applicants shall submit the results of their review of test and maintenance procedures and Technical Specifications to assure that post-maintenance operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

2.

Licensees and applicants shall submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.

3.

Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing Technical Specifications which can be demonstrated to degrade rather than enhance safety.

Appropriate changes to these test requirements, with supporting justification, shall be submitted'or staff approvaL

- Resoonses:

Post maintenance testing of the reactor trip system components is governed by MP 0707.10 and AP 190.19.

MP 0707.10 covers the maintenance and lubrication of the reactor trip breakers.

It has been verified to be in accordance with all manufacturer's recommendations concerning lubrication and testing of the trip breakers and their shunt and undervoltage trip devices.

It is performed at 6 month intervals with a check of the manual pushbutton trips conducted each refueling.

QC Department surveillance checks have been instituted to ensure that maintenance is performed at the correct intervals.

AP 190.19 covers the performance of maintenance and operability testing after maintenance.

The reactor trip system including the reactor trip breakers is classified as nuclear safety related under the Turkey Point Q-List.

Any maintenance performed on these systems requires that the appropriate Maintenance Superintendent specify retest requirements and that these retest requirements ensure that the component is operable upon completion of maintenance.

A review of Technical Specifications did not identify any post maintenance test requirements that would degrade rather than enhance safety.

POST-MAINTENANCE TESTING (ALLOTHER SAFETY RELATED COMPONENTS)

S 1.

Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and Technical Specifications review to assure that post-maintenance operability testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

Resoonse:

Testing of safety related equipment to asce tain its proper post maintenance funct'oning is already required by Section 3.2 of Administrative Procedure 0190.19, "Control of Maintenance on Nuclear Safety Related and Fire Protection Systems".

Additionally, Sections 3.4.3 and 3.4.4 of Administrative Procedure 0103.4, "in-Plant Clearance Orders" requires verification of operability, and surveillance testing of all Technical Specification systems or components prior to return to service after a clearance has been issued.

Representatives of each Turkey Point Maintenance Department were contacted to discuss how the above referenced requirements are being met.

A.

The Mechanical Maintenance Department utilizes Administrative Procedure 0}90.28, "Mechanical Test Control (Post Maintenance)" to select and document the tests which are conducted on specific safety related equipment following specific maintenance work.

B.

The Electrical Department has incorporated the post maintenance test requirements and documentations into specific maintenance procedures such as MP 0729, "Safety Related Motor Operated Valves (MOV) Motor Maintenance".

In cases where the maintenance work on safety related electrical equipment is not covered by a specific procedure the post maintenance test requirements are included in the "Work Description" block of the PWO and documented in the "journeyman's Work Report" on the back of the canary copy of the PWO.

C.

The IkC Department performs the necessary calibration checks on safety related equipment after repair or replacement and documents this work on the calibration forms and PWO.

.At the completion of maintenance work on safety related equipment, the PWO and supporting documentation is reviewed by a Q.C.

inspector IAW AP 0190.19 and these records are retained in accordance with AP 0}90.}4, "Document Control and Quality Assurance Records".

In order to provide the documentation requested by the VRC and to further strengthen the Administrative control on post maintenance testing of safety related equipment, a

change was made to AP }90.1'9.and.~as submitted for review.

Licensees and applicants shall submit the results of their cneck of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications were required.

Response

Vendor recommendations for HSSS equipment are reviewed and acted on by means of Westinghouse Technical Bulletins which are entered into the FPL Operating Experience Feedback Program for tracking and implementation.

Information supplementing or revising the instruction books or other similar materials necessary for proper and safe installation, operation, maintenance or repair of WRD-supplied equipment or parts in nuclear appf ications is provided by WRD in the form of Technical Bulletins.

Preparation of such Technical Bulletins has been centralized within WRD and is subject to the same design control process which applies to the equipment, parts and services to which they relate.

Technical Bulletins which are safety-related are so identified.

AII Technical Bulletins will be transmitted by WRD to every Westinghouse NSSS customer, domestic and international, and such other WRD customers as are affected.

Responsibility for this distributio'n has been centralized for customers with operating plants and for customers for plants not yet in operation in or'der to ensure that all Technical Bulletins are promptly distributed.

Customers have been requested to provide the necessary distribution lists for their organizations.

All distributions of safety-related Technical Bulletins are.now accompanied by a return receipt.

The return receipts are pre-addressed to a central point in MRD for recording all Technical Bulletins transmitted and their status.

Technical Bulletins for which receipt is not acknowledged with a reasonable time are retransmitted.

A list of current Technical Bulletins and Data Letters will be prepared and transmitted to all customers periodically but not less frequently than once per year.

This transmittal is to be in the form of a Technical Bulletin and is to be transmitted in the same manner as any other Technical Bulletin.

In addition, as participants in the IflPO SEE-IN Program, significant events occurring throughout the nuclear industry and important vendor information items are entered into the FPL Operating Experience Feedback Program.

This program provides additional assurance that test and maintenance items which have caused problems at other plants will be reviewed for applicability to Turkey Point Units 3 and 4 and dispositioned.

3.

Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing Technical Specifications which are perceived'o degrade rather than enhance safety.

Appropriate changes to these test requirements, with supporting justification, shall be submitted for staf f approval.

~Res ense:

There were no post maintenance test requirements identified by the Maintenance Department representatives contacted that were perceived to degrade rather than enhance safety.

There ~ere.

however.

3 oeriodic surveillance reauirements that could "e perceived as degrading safety.

They are:

A.

The Containment Spray Pump periodic test per OP 4000.1 requires shutting the manual discharge valve of the pump being, tested.

This effectively takes the pump out of service for the duration of the test.

B.: As part of the test of the auxiliary feedwater pumps during periods of single unit operation, the 2 pumps not under test are manually tripped effectively removing them from service while the third pump is running for test.

C.

The monthly battery charge imposes a higher than normal voltage on the DC buss and results in increased failure rates for the normally energized NBFD relays in the reactor protection system.

Additionally, the Tech.

Specs.

requirement for a monthly battery charge conflicts with the manufacturer's recommendation to charge only when needed, as indicated by voltage and gravity readings.

In cases A and B above, an operator is available throughout the test to" return the pumps'to service if needed.

A Tech. Spec. change request was submitted in 1977 to test these pumps in accordance with Section XI "In Service Testing" of ASME Code.

V/hen approved, this change will require pump testings every quarter versus the current monthly frequency.

This willreduce the frequency at which these systems are put into this degraded mode

REACTOR TRIP SYSTEM RELIABILITY (VENDOR-RELATED HODIFICATIONS)

Position All vendor-recommended reactor trip breaker modifications shall be reviewed to verify that either:

(1) each modification has, in'act, been implemented; or (2) a written evaluation of tne tecnnical reasons for not implementing a modification exists.

For example, the modifications recommended by Westinghouse in NCD-Elect-18 for the DB-50 breakers and a March 31, 1983, letter for the DS-416 breakers shall be implemented or a justification for not implanenting shall be made available.

Hodifications not previously made shall be incorporated or a written evaluation shall be provided.

Response

Westinghouse records indicate that the modifications recommended in NCU-Elec-18 for 0850 switchgear have been implemented for the Turkey Point Units 3 and 4 reactor trip breakers.

The Westinghouse Owners Group has recommended a method to visually inspect for reconfirmation of implementation.

This inspection was conducted for the Unit 3 trip and bypass breakers and the Unit 4 bypass breakers and confirmed the presence of post 1972 units (modified units).

The two other UVTA modifications identified by Westinghouse in the Harcn 31, 1983 letter and in NSD-TS-75-2 of February 20, 1955 do not apply.

0.2 REACTOR TRIP SYSTEM RELIABILITY(PREVENTATIVE MAINTENANCE AND SURVEILLANCEPROGRAM FOR REACTOR TRIP BREAKERS) l.

A planned program of periodic maintenance, including lubrication, housekeeping, and other items recommended by the equipment supplier.

2.

Trending of parameters affecting operation and measured during testing to forecast degradation of operability.

3.

Life testing of the breakers (including the trip attachments) on an acceptable sample size.

0.

Periodic replacement of breakers or components consistent with demonstrated life cycles.

~Res onses:

This item is accomplished under MP '0707.10, OP 1000.2, and ONOP 0208.1.

MP 0707.10 covers

cleaning, lubrication and testing of the breakers.

Every 6

months,

~ each circuit breaker is cleaned, lubricated as

required, checked for alignment, checked for proper operation of shunt and undervoltage trip devices, and tested on the bench and in-place.

Every refueling overhaul operability of the shunt trip is checked via the manual trip pushbuttons.

No specific trending of r'eactor trip breaker opening times is performed.

However, OP 1000.2 performed monthly and ONOP 0208.1 (Reactor Shutdown Resulting From a Trip) require that breaker opening times. of greater than 120 msec be reported to the Assistant Superintendant-Electrical Maintenance in order to schedule preventive action.

This opening time is 70% of the maximum permissible and will ensure maintenance is conducted prior to opening time specifications being exceeded.

Life cycle testing of the shunt trip attachment and the undervoltage trip attachment of the reactor trip switchgear is being conducted by Westinghouse for the Westinghouse Owners Group.

This program is aimed toward establishing the service life of these devices, and substantiating periodic test requirements with proper maintenance.

The results of"this program will be factored into maintenance, replacement and qualification programs.

The test program is scheduled for completion in the second quarter of 1984.

~ ~

4.3 REACTOR TRIP SYSTEM RELIABILITY (AUTOMATIC ACTUATION OF SHUNT TRIP ATTACHMENT FOR WESTINGHOUSE AND B8W PLANTS) position Westinghouse and B&W reactors shall be modified by providing'utomatic reactor trip system actuation of the breaker shunt trip attacnments.

The shunt trip attachment shall be considered safety related (Class IE).

Response

A detailed generic design package for -incorporation of an automatic shunt trip feature into various Westinghouse Reactor Protection Systems has been developed under WOG sponsorship.

The complete generic design package of the automatic shunt trip modification was submitted to NRC on June 14, 1983. J. J.

Sheppard, Chairman of WOG by letter OG-101.

FPL is in the process of requesting that our Architect Engineer begin preliminary design work for the automatic shunt trip modification for Turkey Point Units 3

& 4.

The generic design provided by the WOG and plant specific recommendations will be considered.

A proposed design will be submitted to NRC review by May 15, 1983.

As part of the submittal, a response to the thirteen NRC concerns in the SER to the WOG design will be provided.

A schedule for impl.ementation of the design change will also be provided at that time.

.4 REACTOR TRIP SYSTEM RELIABILITY (IMPROVEMENTS IN MAINTENANCE AND TEST PROCEDURES FOR B&W PLANTS)

This item is not applicable to Turkey Point Units 3

& 4.

e ~

4.5 REACTOR TRIP SYSTEM RELIABILITY(SYSTEM FUNCTIONALTESTING) 1.

The diverse trip features to be tested include the breaker undervoltage and shunt trip features on Westinghouse, BOW and GE plants; the circuitry used for power interruption with the silicon controlled rectifiers on BRW-plants; and the scram pilot valve and backup scram valves (including all initiating circuitry) on GE plants.

30 Plants not currently designed to permit periodic on-line testing shall justify not making modifications to permit such testing.

Alternatives to on-line testing proposed by licensees'will be considered where special circumstances exist and where the objective of high reliability can be met in another way.

Existing intervals for on-line functional testing required by Technical Specifications shall be reviewed to determine that the intervals are consistent with achieving high reactor trip system availability when accounting for considerations such as:

l.

uncertainties in component failure rates 2.

~ uncertainty in common mode failure rates 3.

reduced redundancy during testing 4.

operator errors during testing 5.

component "wearwut" caused by the testing Licensees currently not performing periodic on-line testing shall determine appropriate test intervah as described above.

Changes to existing required intervals for on-line testing as well as the intervals to be determined by licensees currently not performing on-lint testing shall be justified by information on the sensitivity of reactor trip system availability to parameters such as the test intervals, component failure rates, and common mode failure rates.

Responses Turkey Point currently has a complete on-line system test procedure which is performed monthly.

This test checks the reactor trip breakers (via the undervoltage trip device),

the reactor trip relays and their associ'ated 'logic relays, and a calibration check of the protective system inputs.

The current shunt trip device and undervoltage device are functionally bench tested every 6 months.

The current shunt trip device is operationally checked at each refueling outage.

There is currently no provision for on<<line testing of the shunt trip devices on the reactor trip breakers.

It is only operated using the manual trip pushbuttons, all automatic trips utilize the undervoltage device.

The shunt device is bench tested every 6 months and checked via the manual pushbutton each refueling.

Provisions for on-line testing of the shunt device will be considered more formally when the PC/M for inclusion of the shunt device in the automatic trip system is designed.

As noted above in Section 4.3, the WOG generic design package for tne automatic shunt trip includes an installation for on-line surveillance testing of the UVTA and automatic shunt trip that provided indepenaent verification of each attachment.

The Westinghouse Owners Group in January 1983, submitted WCAP-10271 to the NRC for review.

MCAP-10271, "Evaluation of Surveillance Frequencies and out of Service Times for the Reactor protection Instrumentation System" documents and evaluation of the impact on RPS unavailability of current and extended surveillance intervals.

The WCAP considers common mode failure, operator error, reduced redundancy during testing and equipment bypass.

WCAP-10271 also considers correlative effects on plant peration and safety including the manpower expenditure associated with surveillance, the number of inadvertent trips which occur during testing and the distraction from plant monitoring on the part of the control room operator and shift supervisor associated with testing.

Supplement 1 to MCAP-10271 whicn was submitted to the NRC on October 4, 1983 is an extension of the evaluation and provides a discussion of component wearout caused by testing.

The NRC review of WCAP-10271 to date has resulted in a request for additional information the NRC felt necessary to complete the review.

Information that was submitted to the NRC in response to that request includes an overall evaluation of the

impact, on plant safety of RPS surveillance, a discussion of the uncertainty of failure rates and common mode failure and more detail concerning the impact of surveillance intervals on RPS unavailability.

MCAP-10271, Supplement 1,

and the information provided to the NRC in defense of WCAP-10271 provides in a comprehensive form the information requested by item 4.5.3.

The conclusion of MCAP-10271 and Supplement 1 is that although RPS unavailability is increased less frequent testing of RPS components is warranted and will result in an improvement in overall plant safety and equipment reliability.