ML19224C144

Responds to IE Bulletin 79-05A.Forwards Info Relevant to Design of BWRs & Events Surrounding TMI Incident

ML19224C144
Person / Time
Site:	Crystal River, Crane
Issue date:	05/12/1979
From:	Stewart W Florida Power Corp
To:	O'Reilly J NRC Office of Inspection & Enforcement (IE Region II)
References
NUDOCS 7906290146
Download: ML19224C144 (18)

Text

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Florida W. P. 51 F W Al1 T, Oll t I (T 01 t PL1WL 14 PilOl1LeC1 TON y

April 12,1979 Mr. J. P. O'Reilly, Director U.S. Nuclear Regulatory Commission Office of !nspection and Enforcement 101 Marietta Street, Suite 3100 Atlanta, GA 30303

Subject:

iver Unit No. 3 (rtuclear.)

3 M 02 OccKet no.

Overating License No. DPR-72 IE Bulletin 79-05A

Dear Mr. O'Reilly:

Enclosed please find our response to Items 6, 7, 8, 9, 10, 11, and 12 of IE Bulletin 79-05A.

hould there be any questions, please contact us.

Very truly yours, FLORIDA P0..'EP CORPCRATIONO -

< = - -

Lh.

C QA 0c 1 )

W.P. Stewa rt ECSekcM03 024 At t acht.:ent s File:

2-0-3-a-3 5<s s u Q 256 191 79062gg 309-s YL f

General Ottice 320s Tniny.tosin street scem. P O Box 14042. St. Petersoug. Fiono4 33733 813-866-5151 l

4

4 isolation initiatior design and

• ITE 1 6.

Review the containment all changes necessary to procedures, ard prepare and implement

~

h isolation does containment isolation of all lines w ose degrade core cooling capability upon automatic initiation cause not of safety injection.

The a t tached Table S-4, Reactor Building Isolation Valve isolation RESPONSE 6.

Information, contains a listing of all containment if it This tabic indicates for each valve CR #3.

valves at pressure), its presently receives an ES signal (4'psig R.B.

accident position.

normal position and its post if they have reviewed the valves in this table to determine We 1 actuation and/or 4 psig could be isolated by automatic E:

Pressure or if they are ret uired for core cooling and R. B.

therefore should not be isolated on automatic IIPI actuatier..

If the valves listed receive no ES s'ignal and they are normally closed and remain closed following the accident

onditions, no further action is required.

column added on the right of the table indicates The first which valves are needed for core cooling capability and will The second col'umn be isolated by automatic HP1 acutation.

further detail not identifies which valves are being reviewed in for the addition of an auto-close isolation signal ba. ed on This review will insure tha t automatic HPI actua tion.

isolation of these valves on automatic HP1 actuation will not alter the results of any previous saf ety analysis and will insure compatability with the existing CR F3 design.

md*

256 192

TABLt 3-4 Sh'e t 1 of 4 p rirTr.a wit 01% Ivt Attr< vat te tvinrrtw Location Line Method

.g est e 44 Marm:1 Past

,,,4 ( ;,.

y 4,',, b g,4 - e -

Flow Valve Referred Valve

Stae, et p, e a s.e t Valve Position Acctleat H 91 fee 6ms.hac.

{cre (s el. a Service 9 v o rm Dirtetten A f rf,t.

to e,s.

g e_

yo.

Aceuittom Steaal Fostrfoa 1-iteset a P. a l t s..

Fenetretten Ae.

ICP. 301 5tese Llose P.S out 1

Outstje Isalation 26 Spring Open tee C I : =rJ 1

.IY CL g

Hain 5tcas 105, 1C6 C 1,< J)

S C.,

))k OutatJe Chc(k 14 be 4 3 3, 41t re=Jwe<=r Lance FW Ie 1

Catstte Check 6

Fa Cl 4 w iJ 103, l')9 bL

h.

• to 2.5.

5A la 11 OutstJe Cate 3

19 CL..cd Ho

( t a.. A 110 seswice A ir su p pl y 113 soeply te A.s.

IA le Il C telde Cate 2

IrJ Closed Ee C eseJ

_I L

_ Ij D, lit in:ssu ent Air C l as. J}

,f !f.

j r,?

Isolde Bestestly 48 f.t3 15 Closed Tee F-.: g e G.,p 1 f AW

!=

3 outende B..s t e s t l y 45 Att 15 Cl uc4 Tee (l e s. t) 113 p eac c ur E.11Jira 116 E.%. le.k Fate Tces tR Cut 4

f e elJe 511nd Tiense 1

C l o.. J Ib-

- I Ie, oo t. 4 Ciae a

tu C i..e s n.

C t. s...

j f(,

.j f -

act i

t..

,.t a t io.

' isc J InstJe Cf6cc k

)

44 B s ac t *r F a ll4 8..g LW

n 3

outs 11e Cate 1

WG ts o,,..

Le i t.. J

!!I D ota. 62. a < r to C a o,. 4'

~ [ } ('

~N, Cl e.5 J ts.

OutatJe C it e NW re tte rication Line tR In 1

Out sid e Cete 8

IlU

'6 Class e

4e C. S..

!!L 3.5. Leaa p a r e Tsis t gg g,

04:314e Case 8

trW

'* t.e se d Ea Ct. [

-~

(* m (J a

Outtlde Cete S

H'J Closed He lit t. 3. L e a ', F.t e Ts a t Clos.

it)

Depress 6s teat aan L:se Outelde Cate

)

W Closed be e.J test Accid 4 mt l'rdregen Pogge IR Out 6

353, 313 S ptly CF In 1

0 tette Cl#e 1

Air ES Clemed Tee C l a s.

-Y,' '

II O fantdo Check 1

Fe o,.4 l'11, 126 Case Fle*Jans Tase 3CF, 367 k'ater so ;1y to P.S.

Cot oute Je Cate 2%

A6r t$

Cpea fee C1e s e 1]

~ (J llc to 4

Outelde Cete li Air 11 Open Tee Cl as e d)

~~

2C4, 346 InJestetal Cooler Cavity Caaltas C1 eAll velvee tiltis ele:ttle sa:ct operecere ete slee equipped with tendwheele.

ECTsa pcter to F:.ut e 9-L f e r ef e:ce s',tir eviatice id ntif ic ot ts..s.

Nm a

y w

E

, a c

b u

4e...

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p rprTon kit n r. " 15.*l> T ION V Al % / I N FC F PtA T I M Short 2 of 6 Lc c at ion Ling

'te t t.c4 ag pe.g i f, f a rr al Fe s t g, 4, g g) p,l gg g

.e detretten Flow Valve Referred valve

Sise, et pr es s eg

t. lea Po sit ion A< r :

a (or d ; M hfI I4 t Y u ** k i C '

• • e.

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o 2.5 he In Acraitton 5!- e) to t?n Ira'est: 21

• --it s

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re s t:

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Atr (W ri Y. s t1i.

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1 Cut 9

Out.ife C1de 1%

tri ris.4

o

s. t.-

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No tl 17, 335

s ticci n sep,ly to eat.:Je C.i t e 1%

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trJ

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Att

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Ops, Yes f l.i.

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^^2. 356,

?:wc le ar Ses y st e s

. 326 Comitr.g to M.

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o. c 1 Yes Ci. v }

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CesstJe C te 3

Air ES Op

• Yes C!' s 6) pg 11 C<.a * :r s s o CT

't's su o.t outstJe Cate 3

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g a,,ed $.e Igh r *I '

Laretten 116 :

Hethe*

dlppgf6 h e ral "a.t j

F1w Valve Referred valve 51 c, of henact Vilve Positten Ar e !.f es t Mr g

G %

$< rv tr e 5*m fi -c:efon

1. rret.

to 3.9 Tm

'n.

Ae-it 'r e steral Pa**t'ea T +11e n 'ca f.it'.

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i.settation

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J g_

I ns t.te Ditt.-sfly 43 tri E5 Cla=s' Yes t'i r, g e

-y

,{

1;i re.seter tillding t rge Tshtunt

.ut Cut 23

( u t side tutterfly Air CS Clowed tes C? m s

\\

i O.M

'. d etterfly S

Air Resote npen yes op. m

)).1.

)ss,St$.

Nelear service.

St.f35}f j Covatng to R.4.

SW In 13 Outside t

M P*aual 8

[)-

Cut Outelde Sutterfly C

Air De.cte Cycn Tes Cpea f$

71 *- factrculation fan "g

_e Coolcro L n a.1 F($

b.

s

>.., ;g. :p C1~.:) l No

_ t.5 _ ?

6 R.C. Dre;n Tri.k Instde C l o',e 3

!" )

E5 Cl a sed Yes 14(,

Ci a !.i 42 Oi t 1

Cu:stde e t apt r s::a 3

Air 15 Cloud Yes Cle i.J J LC. Pusp seals Inside Cl o'ne 1

D:3 T5 Ops a Yes Cic =eh (1...sJ l s.'

j !P f O'ildetoff l

• b

• • J Out 26 Inside C l o 'o c 1

TNO Es C;en Ven

'h f

). C Istide f.. ct.e 1

F2.0 15 Orea Tce Cle a. d

'i lastdo C che 1

C;0 E5 cres Yes C h.. J

' ea Outside C ota 1

Air

!5 Cres Yes C 1. u c, I g I

l0 p*

i of f i.

5 condary Drain 411 F,_

fram steam Cererator Y.5 Out 9

Cosside C1che 3

air 25 Closed Yes C ;u...:

I

• f.

e b atal r.:.k-up to S

,Tssctor Ccelant $ y s t ern MU In 27 instJe Check 2's.

Cpen

\\f

'd' '.'

o i.

-ifs F.

.< d itsr.t tre sur e Cutstee Cted.

2's Fr'J ts c14eed ves v

In % t;on Ostside Cate 2's D:3 E5 C p.- n Yes C14 sed g

3 Frersvriter eni Reactor 4)$ s h

'9

.,' 'e

-., '. Coolaat Sa ple L has CA CA 28 Insife Clote 3/8 D'O t$

Clas?d Tcs Cta.-

' N.

l 0

"; O l [

b Ins 1Je C:ebe 3/8 0.0 S

C l o,c 4 tes C!c <J Ir.s td e C1:bo 3/8 C;J

, f5 C;osel Ten C t e,.t *

)

p' C;a..:JJ Cutstie C1che 1

Air E5 C'.

sed Yes

[ 441'

!a:ple 09tstJe C1cte 1

Air 15 Cren YJ3 C1c cJ

~

S 140 Steam Cenera:or CA Out 29 Inside Clote 3/8 DJ 15 0 p c r.

Tcs Ctesgj O

a PODR DWi@i6l

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4

!!E:' 7.

For manual valves or manually-operated motor-driven valves which could defeat or comprem$se the flow of auxiliary feedwater to the steam genciators prepare and implement s

procedures which:

a.

require that. such valves be locked in their cotreet position, er b.

require c;her sim'11ar positive position contevis.

RCS PO!!SC 7.

The minual valves which could defeat. er c amornaise the flow of auxiliary (emergency) f eedwater to the steam Aenerators,

i.e.,

canual valves in the principle auxiliary (emergency) feedwater piping, are already included in the Lock >d Valve List Procedure, SP-381, to ensure that they will be locked in their correct positi ns.

These valves include CDV-103 & 104 and EFV-23, 24 & 36 The manually-operated motor-driven valves in the principle auxiliary (cmergency) f eedwater piping flow path will have their remote ind ica to rs in the control rmn cheeked daily to ensure that they are in their correct positicas by revision to SP-300, Operating Daily Suiveill;ince Log-Thesu val ve s a re EFV-1, 2, 3, 4, 7, 8, 11, 14, 32 & 33 and FWV-161

& 162.

Because of other safety analyses, ii is not practical to lock these valves.

e e

6 w

256 197

ITC! 8.

Prepare and implement immediately procedures which assure that two independent steam generator auxiliary feedwater flow paths, each with 100% flow capacity, are operable at any time when heat removal frcun the primary system is through the steam generators.

When two independent 100% capacity flow paths are not available, the capacity shall be. restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant shall be placed in a cooliny, n. ode which does not rely on steam generators for cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

When at least one 100% capacity flew path is not available, the reactor shall be made subcritIcal within one hour and the facility placed in a shutdown cooling made which does not rely on steam genera tors for cooling within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or at the maximum safe shutdown rate.

RESPONSE 8.

The following flow paths are for the auxiliary (emergency) feedwater to the steam generators from the steam-driven eme rgency f eedwater punp (EFP-2):

"A" OTSG - (1) Piping through EFV-5, 8, 11, 17, FWV-162, 44 (2) Piping through EFV-5s 8, 11, 17, FWV-157, 40, 35, 44

• B" OTSG - (1) Piping through EFV-5, 8, 32, It, FWV-161, 43 (2) Piping through EFV-5, 8, 32, 18, FWV-158, 39, 34, 43 The following flow paths are for the auxil tary (caergency) feedwater to the steam generators from the motor-driven emergency feedwater pump (EFP-1):

'A-OTSG - (1) Piping through EFV-6, 7, 14, 15, FWV-162, 44 (2) Piping through EFV-6, 7, 14, 15, FWV-157, 40, 35, 44

• B" OTSG - (1) Piping through EFV-6, 7, 33, 16, FWV-162, 44 (2) Piping through f.F7-6, 7, 33, 16, FWV-157, 40, 3S, 44 SP-349, f.me rgenc y Feed wa te r Sys t em Ope rab i l i ty Deuons t ra t l on,

has been revised to implement the above 1 Low paths.

A tech-nical specification change request will be submitted to include the Action Statements of Item 3 into Speci fi-cation 3. 7.1. 2.

Until that license amendnent is issued, these Act ion Statements will be administratively enf orced.

256 198 I

4

IT C 9.

(This item revises iten 6 of IE Bulletin 79-05. )

Review your operating nodes and procedures for all systems designed to transf er potentially radioactive gases and liquids out of the primary containment to assure tnat undesired pumping of radioactive liquids and gases will not occur inadvertently.

In pa rt icula r, ensur e that such an occurrence would not he caused by the reset t ing, of ene,ineer ed sa fe ty feat ures instrumentation.

Li s t all such

.y tems and indicato Whether interlocks exist to prevent transfer when high a.

radiation indication exists, and b.

Whether such systems are isolated by the containment isolation signal.

RESPONSE 9.

In response to the question pertainicg to systems capable of pumping racioactive gases or liquids from the containnent, the following list summarizes those systems in the containment that fall into this classification.

1.

Itain.: cam 2.

Decay Heat Return Line 3.

Reactor Coolant Pur.ip Seal Return tiakeup line f rom Reactor Coolant Systpm (letdown 4.

line) 5.

RF1-A6 (Containment atmosphere radiation monitor) 6.

Reactor Coolant Drain Tank (liquid) 7.

Reactor Building Pu'rge 8.

Reactor Building Sump to Waste Disposal 9.

Gaseous Waste Disposal from Pressurizer, Steam Generator, Reactor Coolant Pumps and Reactor Coolant Drain Tank (WDV-405, 406) 10.

RB Sump Recirc. thru Decay Heat 11.

RC Drain Tank Vent (WDV-60, 61)

All systems in the containment were reviewed and all systems and components that are part of the Reactor Building isolation and cooling system were reviewed.

The 11 items listed above are those that fit the question and are listed in the enclosed table with the type of actuation that occurs during an emergency.

Three of the systems above, RB purge, RB soap, and the Fu and

S rupture matrix systems, have special conditions that can automatically change state other than actions per fo rmed by the Engineered Sa fegua rds RBl&C systen.

256 199

1.

R3 Sump - The Reac tor thiilding sump contains 2 pumps.

One pump is automat ically tur ted on and off by sump level on at Clevation 85. 5 f t. (Plant Datum) and off at 85 f t.

I t' the first pump fails or the increase in level is too great, the second pump autouatically start:4 at Elevation 86. 5 f t.

The system becomes inoperable during Reactor Building isolation and cooling actuation since WDV-3 and WDV-4 are cloacd, which prevents the pump (s) from starting. Work is progressing to modify this system to be interlocked'wlth the auton itic high pressure injection signal to assure against inad ve r t en t pumping of the sump contents when high radiation might be present.

2.

The RB purge system is inittaced by the operator by procedure and is tripped off by exceeding a pre-set radiation level prior to the release.

As shown in the enclosure, all four exhaust dam pe r - close when the Icvel is reached, and purge ope rat Lon mus t he reini t iated by the opera tor.

This system would cause no inadvertent pumping of radioactive gases during an accident.

3.

Each steam generator has two independent matrices (A &.B) for redundancy of isolation capability.

Isolation actuation is initiated only on the af fected steam generator.

Matrix logic is such that up'on steam line pressure decrease to 600 psig, isolation is actuated to close and "open inhibit" Main Steam Iaolation Valves and Main Feedwater Valves on*the affected steam generator.

This "open inhibit" remains in effect until steam line pressure increases above 725 psig or the matrix is manually by passed.

Manpal by pass is automatically reset above 725 psig steam line pressure.

After isolation actuation and the steam pressure returns to above 72 5 psig, or nanual by pass, all valves wtIL return to their normal control capability.

This provides automatic reactor coolant heat removal through the steam generator.

In addition, the operator has the capability to manually control feedwater through the Integrated Control System (ICS) or directly through control of Emergency Feedwater Ey pass Valves to each steam genera:or.

Modifications are also being reviewed to allow the WC drain tank pump and all waste gas connections to the RC system (pressurizer, steam penera tor, RCP and RC drain tank) to be a ut ena t ic al ly shut down on a high pressure injection signal.

This would assure that during a pos s ible s i t ua t ion of high radiation in the containnent atmosphere, operator action would be required te relieve the reactor systems to the waste systems.

256 200 4

4 w=

In the meantime, Short Term Instruction 79-26 has been issued whereby if a Reactor Building high radiation signal or on the unexpected loss of reactor coolant, the following will be pe r fo rmed :

1.

Place the control handics for the RS suur pumps in the pull lock position.

2.

Place the RC drain. tank pump controller in the

<>l f pos'. tion.

3.

Close either WuV-405 or WDV-406 (Waste cas licader) 4 Close either WDV-60 or WD'J-61 (RC drain tank vent line)

All other systems listed could not inadvertently pump radioactive fluids without operator action, or proper ES signal.

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S I'.TI'O I NT OR I ll rL R -

SYS fill OPER ATIOri SYS fif! 1:;01.Al l'1) 8Y AUTOtlATIC SETII!iG Of f'

!.OCK FOR ACTilATIOff Af t ECTEI) liY RAl> I ATIOtt

!. p s i r. R !! l'H l.S S U R E S if f1AI. WilllOlli Ol'4 H A l o.,

OR ISOLATIO!J I.EVEl.S - Yl.S o r 110 SIGMAL - YES o r !!O ACTIO!1 - YES OR !;O SY ST E!!

tio tJo No VI.

Dil return lino Ope ra to r -on t rol l eil

( lillV -'6, lillV-41 ), (#2) flo auto li. <ic t t on 711.

flain St eam (#1)

Isola tes !!S isola-tJo No See Writeup 13 tion valves FW block valves & Emergency, fu tilock valves on (600 psig flS pressure k

No tio No L

Vill.

RB sump recirc. thru Operator controlled

<lec ay be.it (#10)

No auto functinn 9

e a

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s i 1~L....

Review and nndify as necessary your maintenance and test procedures t; ensure that they require.:

Verification, by inspection, of the operability of a.

redundant safety-related systems prior to the removal of any safety-related system from service.

b.

Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing.

c.

A means of notifying involved reac tor operat ing pe rsonnel whenever a safety-related system is removed f run and returned te servtee.

R:5? OSSE 10. The surveillance procedures at Crystal River 3 presently require the verification of the onerability of redundant safety-related systems prior'to one removal of any saf e ty-related system from service.

This verification incluJes either the performance of the appropriate surveillarces on the redundant saf ety-related system or the confi rmation that the east surveillance on the redundant saf ety system would not exceed the ' technical specification interval for performance until af ter the safety system is esticated to be back in service.

All systems at Crystal River.3 are removed and returned to service using approved proced*ures that include the verification of operability when a syctem is returned to service pe r AI-600, Conduct of lbintenance.

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If an approved procedure does not exist for returning a system to se rvic e, such a' procedure is required before that system removed from service.

Reactor operating personnel.are notified whenever a saf ety-related system is removed f rom and returned o service as fellows:

1.

It is the reactor operating personnel that perfo rm the switching and tagging of plant equipment so they know when it is removed from and returned to service.

2.

Any components of the Engineered Safegua rd System that are out of se rvice are noted on the Plant Status thia rd which is reviewed by each operator as they come on shif t.

l 256 204

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4 ITC1 11.

All ope rating and maintenance pe rsonnel should 'oe made aware of the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident.

RESPONSE 11. Training classes for the opera tors at CR #3 were condar t ed by FFC on April 4,

5 and 6, 1979 and again by the NRC for ou opera to rs on April 7, 8 and 9, 1979.

The purpose of these training classes was to make the operating personnel at CR #3 aware of the extreme seriousness and resultant consequences of the simultaneous blocking of both auxiliary feedwate. tecins.

premature te n,inat ion of the llPI system, tripping vil reactor coolant pumps during loss of '

-iwa te r

• rans ie nt s, and to revic the sequence of events. c TM I - 2.

Training classes for the maintenance personnel at CR I/ 3 were cunpleted on April 11, 1 T/ 9. ' Again, the purpose of the=~

classes is to stress the seriousness of the events that occurred at U11-2 and to emphasize that proper clearance procedures fo r pumps and valves must be followed in accordance with CP-115, In-Plant Equipment Clearance and Switching O rd e r s.

These classes will also stress the fact that, in accordance with standard operating procedures, only operating personnel are pe rmi t t ed to alter the status of pumps or the positions of valves.

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a 256 203 2

4 ITO: 12.

Review your prompt repo rt'ir.g procedures for NRC notification to assure very early notification of serious events.

RESPONSE 12. The prompt reporting procedures for NRC notification to assure very early notification of serious events have been reviewed and the results of this review are discussed below.

Referring to Section 2.22 of Al-500, Conduct of Operations, various operating situations are addressed in which the Operations Superintendent and/or the individual on call is promptly advised by the ' Duty Shift Supervisor through verbal communication.

These events are:

a)

Reactor Trip b)

In ad ve rt en t (radioactivity bearing) liqu'd or gaseous releases.

c)

Major equipment failure or malf unction (includes all sa f egua rds equipment),

d)

Unex plained reactivity changes.

e)

Loss of offsite power.

f)

Employee injury or radiation overexposure.

g).

Accidents occurring on plant property (except minar injury).

h)

Events requiring reports within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Region I office.

(Technical Specifications 6.6 and 6.9) i)

Turbine Trip j)

Load Restric tions The Shift Supervisor shall note in his log when he notified the Ope rations Superintendent. or pe rson on call.

The Duty Shift Supervisor notif ication implicitly includes obtaining advice, assistance, and/or direction for those everts 4r situations which are classified as emergencies.

The Duty D.if t Supe rvisor must refer to EM-207, Reporting Require-ments or Emergencies.01-207, Section 2.6 defines reporting reqnt rements on Emerge.icies.

C1-2 07, Section 2.6 defines requirements for Class A, B or C radiation emergencies.

Emergencies involving radioactive releases or property damage are reported pursuant to 10 CFR-20.403 and 20.405.

Both documents require immediate notification to the appropriate regional NRC Inspection and Enforcement Of fice by telephone and/or telegraph.

Section 2.6.1 of el-100 states that this notification be c.ade by the Emergency Coordinator.

Per Section 2.1 of Cl-100, t he Emergency Coordinato r pon it ion is to be filled by the Nuclear Plant ?lanager or by any of t ho se designated in advance by ldm.

Dur ing of f hours, the Duty Shift Supervisor shall ac t as Tempo ra ry Eme rgency Coo rdina tor.

'06 C

i E't-2 n 7, Keporting Requirements on 1:.-e rr e nc ies,,; l v-the Emergeacy Coordinator the res po ns ib ili t y for assuri ng pr:cedural it.;pl eme n t a t io n.

Aga in, f o r 21 ass A, B or C emergencies, Section 7.4, Notification of Regula tory Agencies, is addressed.

D!-2 G o, Emergency Plan Roster and Not;fication, provides a list of names, addresses and phone nnubcr3 of those organi ations and personnel that must be contacted during emergency situa tions.

The reporting of events which occur contrar: to Techn!-

F pe c i f ic a t io ns is covered in Section o..,

Reportable Occurrence Action, and 6.9, Reporting Ktquirements.

Any event falling within the nine listed areas addressing situations for prompt notification, required a report :o the NRC within twenty-four (24) hours by telephone and ccnfirmed by written communication.

• 1he Operations Superintendent and/or the man on call will also be advised of these ecents.

CP-1:1, Procedure for Documenting the Reporting and Review of

o nconf o rcing Ope ra tions, establishes tne mechanism by whicle nonconformi'ng operations are documented and brought to the attention of specific personnel for pro pt evaluatinn or reportability, and reported to plant managerent.

S.etion 1. 2 defines a nonconforming operation. Witnin the definition, reference is made to any incident described in 10 CFR 20.403, 20.405, and also E!!-207.

Se c~t io n 3.2.4 i Initin1 f.va l ua t tu n :

Prompt Reportable, requires the Duty Sh.ft Su pe rv i so r, in the abseace of the Technical Support En,i nee r or Nuc l ea r Technical Specifications Coordinator, to de te nnine it the occurrence requires prompt notification offsite to NP.C, by the Nuclear Plant ?!anager.

If such is the case, he is to con'.act the Plant 'bnage r or the man on call immed iately.

It is felt that adequate procedural control exists to report a ny incident or event which f alls within the guidelines of the Eme rgency Procedures, Class A, B or C radiation emergency, 10 CTR 20-403, and 20-405, and Technical Specifications.

We feel all other incidences are adequa t el-covered by Nuclear c

Plan: :ianager responsibility and those responsibilities o.ated by the Nuclear delegated to certain individuals as desi c

Plant ':anager in letter file 3-0-2, da ted 21 August 1978, Delegation of Authority.

256 207

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• 3l. Sg. Ferr. 'a INTEROFFI CE CO RRESPO NDENCE

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l Crystal River Nuclear Plant CR #3 (OfflCI)

(a All C30(1 3cIJtCT:

Delegation of Authority i

Nuclear Plant Manager 10: Wino = It May Concern saft:

August 11, !978 3-0-2 s letter supercedes the letter dated.innuary 1, 1977, as required I

by paragraph 6.1.1 of the Technical F ccification f or Crystal River f

e Unit 3.

I hereby. during ny absence, delegate to the Technical Support Superintendent, Paul F. McKec, opera tions Superintendent.

William R. Nichols, and' Maintenance S u p e r in t e r.d en t, G a r, F.. We s t a f e r,

in tha t order, the responsibility for decisions regarding overall f acility ope ration, including signature authority.

The purpose of this letter of delegation is to perpetuate facility operation durin;; =y absence.

Mr. McKee, Mr. Nichels, and Mr. Westafer should atte=pt to contact ne prior to acting in =y behalf on any signif-icant decision or authorication.

/

When utilicing this signature authority, the signer will sign his own 1

na=e followed by /for Guy P. Beatty, Jr.

t

. m/JY &

o Guy F/ Bea tty, Jr. /

/

g Nuclear Plant Manager RLA/tb sc:

W.

P. Stewart P.

F. YcKcc R. Nichols h N;c

,.7,c,

t..

3. R. Westafer i

J. Cooper I

4

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256 208 1

,