ML20134K570

Declares IAW 10CFRS2.1233,enclosed Paperwork Factual & Based on Procedures & Documents from Plant

ML20134K570
Person / Time
Site:	Turkey Point, 05520726
Issue date:	12/30/1996
From:	Tetrick R AFFILIATION NOT ASSIGNED
To:	Lam P Atomic Safety and Licensing Board Panel
References
CON-#197-18164 SP, NUDOCS 9702140095
Download: ML20134K570 (47)

Text

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DOCKET NUMBER

! BYPRODUCTS 6 5- 7 c 72 6 -J4 [

, DOCKETED l TO: Administrative Judge USNRC l Peter S. Lam '

Special Assistant Atomic Safety and Licensing Board 97 FE812 P4 :19 l U.S. Nuclear Regulatory Commission l Washington, DC 20555 0FFICE OF SECRETARY DOCMETING & SERVICE i BRANCH  ;

Sir:

In accordaace with 10 CFR S 2.1233 I am declaring that the accompanying paperwork is factual to the best of my knowledge and based on procedures and documents from FP&L's Turkey Point  ;

Plant.

The paperwork consist of the following: ,

1. A copy of the original request and documentation.

2. A copy of the denial dated September 12, 1996.

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3. A copy of my reply to the September 12, 1996 letter.

/ e Ra ph L. Tetrick 18990 SW 270 Street Homestead, FL 33031 Docket No. 55-20726 "Hs

/ AA m

y;. MY COMMMMON # 0C 808200

< D0'UES: Sepumher 27,2000 Bauhd11su Nutry PetAc undswears I

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9702140095 961230 ADOCK 05000250 PDR G PDR g)

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September 25, 1996 j i

i TO: Secretary of the Commission U.S. Nuclear Regulatory Commission Washington D.C. 20555

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Dear Sir:

.396 - am requesting a Per the letter dated September 12, hearing IAW 1C CFR 2.103 (b) (2' .

Enclosed you will find: j and d::urentation.

1. A copy of the original request l .2, is r6.

eeptember

2. A copy of the denz.a. cate

3. My reply 0 the at:te dated ~_e::er. t sincerely,  !

Ra!.rh L. e: rick

. g. .:. .:. g ~6~ '. . . v. o. o. ~. f mores: sad, FL 3303_ ~ -

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September 25, 1996 l f

d TO: Assistant General Counsel for Hearings and Enforcement OfficeNuclear of the General Regulatory Commission Counsel U.S. i Washington D.C. 20555 ,

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Dear Sir:

, ~996 : am requesting a

]

i Per the letter dated Septer.ber 12, hearing IAW 10 CFR 2.'.03 ibis 2'.

l Enclosed you will find:

and d :urentatien.

1. A copy of the origina. request 1

.2, 1996.

2. A ccpy of the denia_ dated September the ab:ve dated '.e:ter.

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3. My rep'.y.::

Sincere _y, ,

i

. . a '_ p h L . Tetrick ,,.

_, . .:. a ...s s .c.

Herestead, FL 33'31

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. 5=.mn-t.a te e

. _ _ . - , _ . . . _ _ . . _ 7.

1996 applicant In response to the letter dated September 12, contends the following:

EXAM QUESTION 863 Plant conditiens:

-Preparati:ns are being made for refueling cperations.-T valve open.

-Alarm annu:. cia: Ors H-1/1, SFP LO LEVEL and G-9/5 CNTMT SUMP HI LEVE' are in alarm.

r. mediate action in Which one of the f: 1;cwing is the required response te these ::ndi icns?

surp '. eve". reccrder A. Verify alar-s by check:ng containten and spent fue; level indication.

B. Scund the ::ntaineen: evacuation aliarm.

C. ~ nit : st e ::n a nment ventilati:n is:la:icn.

venti.ati:n ;s:lati n.

D. Init:ste ::ntr:1 r ::

ANSWER: S reac;or operat rs The NRC analysis and conclusion contends thatanalyze alarms and and senior rese :: Operaters are expe::ed ::

determine the appr:priate course of action based upon specific plant condit:ons and indications.

Applicant' centends that performing an acticn based sotely onthe annunciation alone is not SFP LOW LEVE*_ and OONTAINMENT SUMP SEAL FAILURE" The annunciators should be verified by additional supportive failure.

inf or-sti:n to preclude the possibility of ann fer all alarms the ARP shall"A" answer be verify 3-ARP-097.CF.

consulted. Applicantstates thattherefore contends that alarms is also a :orrect answer. .

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t EXAM QUESTION #84 l

Which one of the following is the basis for step 1, ." VERIFY j

. REACTOR TRIP", of FR-S.1, RESFONSE TO NUCLEAR POWER l GENERATION /ATWS? '

A. To ensure that only decay heat and reactor c:clant pumps  !

are adding heat to the RCS, l

B. To ensure shutdown marg:.n isstandby. within teihri:C l specifications limits for h:: )

J Orre::ive acti:n if To alert the operator :: take f urther C.

the reacter is not tripped.

reacter prc:ective f eatures D. To verify that all aute-at:.:

have f unc::ened as des:gned.

ANSWER: A E-0 step One and FR-S.1 step The NRC contends that the bas:s f:

One are the same and that there is Only One answer.

a :. .v... . - ..- .a

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a... ....

..e arp.. .. .

gives puidance to rest:re CRtripped T: CAL in SAFETYE*

PROCEDL*RE And that it FUNCTIONS. S!nce the reacter was ver fied n :to FR-S.; ~4here the operate tripped. Because FR-S.;

step one you are sentrods he:ause the reacter is not c:ntends that the basis to insert is a FRF and gives guidance the applicantensure only decay heat is added and for FR-S.1 is twofold, (1) t: asks corrective ac  ::ns. Therefore the applicant answer.

(2) To direct that answer "C" also be accepted as a corre :

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EXAM QUESTION #96 Please review this questien as stated in the ori- nal request.

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i THe following question was discovered The app'.icant to bewishes wrong for af ter thethe first l request for review was sent.

question to also be considered.

EXAM QUESTION #90 l REDUCE ~ INVENTORY j When OPERATIONS, draining the RCS using 3-OP-04;..:,the reactor vessel head and l' vented to containment atmosphere.

s Which cne cf the fe;;cwing describes the ef fe :spath en reactorprevided?

is not l vessel level indica:icn if an adequate vent (Assume the reference leg remains fu'.'. .

l A vacuu- in the res ". cops wi'.'. result .n level indicati n A.

j beins 1:wer than a::ua; leve'.s

! A vacuur in the .rcs locps wi'.; ~

result in ;evel indicati:n B.

• 2 eve s.

j

' being higher t' nan actua' j . cops wi.; resu;: in ; eve;

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p:si::ve pressure in the r:s ;evi.s .

ind::a . n 'e s:.ng '.:wer :har a:: a'.

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fer

.he level instruments au:::atica'.ly c: ,rensate

-pes;;;ve or negative pressure .

.2 ANS'4ER: A n RE;UCED ~NVE!. TORY OPERATI2NS , Fage 25,

REFERENCE:

3-0?-04~. 9, Section 5.2.2.3 Cautzen l

i E.O. 3 0F LP-6902121 i l

t The assumption that Turkey the reference leg remains full makes thisthe dra n

j Point question invalid. A:has dry reference ".egs. This conditi:n is verified bythis q 2

0-FMI-041.110. Applicant requests thr: t 4 ]

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l 1 Florida Power & Light Company Turkey Point Nuclear Plant "C WWi l t va s 35 y

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I EE Asi, E$E 0-PMI-041.110 )

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Title:

RCS Drain Down Level Calibration Safety Related Procedure Maintenance Responsible Department:

316/9 6 Revision ApprovalDate:

8/29/00

. Periodic Review Due:

RTS 954620. 95-0144 PCAf 95-150 OTSC 0450-95 i

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page

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Proteovre No.. Proceeurs Tee 11 _

Appro.si oste ,

8/34">5 O.PMI.041.110 RCS Drain Down Level Calibration _

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INITlAL5

[K'O VERIF 6.2 Level Transmitter LT.6421/23 Calibration Check i

• C AUTION Care shall be taken not to break the neck seal between the sensor module the electronics housing.

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. . . . . . . . . . . 3 NOTES J

. i 4 I j Transmitter output Test terminais clocan are tocated inside the transmitter houssog.

l. To gasn access. housing suce cover odentif:ed as Terminal side (see namenta,er l

must be removed i l t i I ,

.

• Zero and Span adjustment screws are accessicle extemally and are locatec l)

I behind the transmitter name plate.

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! l The transmitter output oncreases wnth clockwuse ro*ation of the adjustment screw. I

.

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. O-rings shallbe replaced of housing cover os removed. .

l I l Level Transmitters LT-6421:23 are located inside containment on the l 14 fo h outside the Oro-wall. *

t l 6.........-................J s' up rssan I 6.2.1 Shut transmitter high. low. and equalizing valves at 3 valse manifold.

6.2.2 Open transmitter sent valves slowly to release any trapped pressure.

. .-. . . . .-. . .g 1 . . .

NOTES l I i -

Transmitter and umpulse line loquod contents should be collected in acc b i with RWP requirements. I i l Equalizing valve should never be opened as low siae is dry. .

l s........................J 6.2.3 Remose caps from test Ottings.

6.2.4 Allow both sides of transmitter to dra2n.

6.2.5 Connect pressure source and test gauge to high pressure test fitting.

6.2.6 Close vent salse on high pressure side.

COev

Page.

Procoowre No . Proceoura Tam Aspeova Cate 2/22/96 0 PMI 041.110 RCS Drain Down Level Calibration 4

INITfALS

CK'D VERIF ST tait it4433 6.3.7 Install cover and hand tighten.

6.3.8 Attach special transmitter cover tool to torque wrench.

4 INDEPENDENT VERIFICATION POINT

' Independent Verifier shall:

'

• Venty proper torque appbcation on Step 6.3.9.

• Initialappropnate space on Data Sheet.

,6.3.9 Torque cover to 200 in lb.

Acceptance Cntena:190 in lb to 210 in Ib

, 6.3.10 Record O nng part number and attach QC tag to PWO.

6.3.11 Disconnect and remose pressure test set.

1. Replace defectis e transmitter test fittings (i .e .. Swagelot s.

required.

6.3.12 Reinstall test fitting caps and tighten properly.

6.3.13 Reinstall vent satse caps finger tight do not torque at this time.

6.4 Placme Level Transmitter in Service 6.4.1 Verify that Operations has es'. .blished a vent path through the pressurizer.

6.4.2 Remove the cap and connect a hose or place a poly bag to catch any fluid from the dry leg low point drain valve below the transmitter three valve manifold.

6.4.3 Open the dry leg low point drain below the transmitter three.way valve.

I 6.4.4 Place a poly bag to catch any fluid from the three-valse t manifold on test tee downstream of PRZR Safety Valse I

• 551 A (*-551B) Loop Drain '-545 A P 546A).

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Page haceeee No.: Procaoura f eie t5 Aspreooce 2/22/96 0.PMI.041.110 RCS Drain Down Level Calibration d

INITI ALS CK'D VFRIF LT W31 gt4433 I

6.4.5 Disconnect dry leg tubing at B valve on three valve manifold I

or remove test 'T" cap downstream of

• 545A (*-546A).

. . ._._. ._. ._. 3

. NOTE I I j If test tee us used, have Operations close 'B' valve on three-valve manifold.

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6.4.6 Connect a source of dry nitrogen or instniment air to the I test tee on dry leg tubing on the 58 foot level and blow down  !

to the drain vahe on the 14 foot level until all moisture I is removed from the line.

i 6.4.7 If test tee was not used. disconnect the blowdown connection I from the dry leg tubmg.

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NOTE i I i

g If test tee os used, have Operations close "B' valve on miee-valve manifold.

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I 6.4.S Place a poly bag to catch any Guid below the pipe cap or I drain vahe downstream of *-545 A (*-546A).

I 6.4.9 Remove the pipe cap or open the drain vahe downstream of I 545 A (* 546A) and drain any moisture.

I 6.4.10 Replace the pipe cap or close the drain vahe downstream of I

• -545 A (* 546A).

i 6.4.11 If test tee was not used, connect a source of dry nitrogen I

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or instrument air to the B valve on three-vah e mamfold I downstream of

• 545A l' 546A).

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1. Blow into the pressurizer until all moisture is removed from I the line.

I 6.4.12 Disconnect the blowdown connection.

6.4.13 Reconnect the dry leg tubing to the B vahe connection or I replace test 'T" cap on the three salve mamfold.

6.4.14 Venfy closed equalizing valve on three-s ahe manifold at transmitter.

6.4.15 Remove cap from transmitter low side sent.

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Proc.aure re.

[Pfoc.ourm asero..i o.i.

2/22/96 5 0.PMI.041.110 RCS Drain Down Level Calibration ^i

) I N I T I A 1., S 1

CK'D VERIF '

J LT .821 4f4423 I._._._._._..._._._._._..._._._._._._._._._._.I NOTE ,

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! i Dry nitrogen or instrument air enay be used to assist in $ anting the low side ofI the I transmitter. Direct al flow down kom transmotter wat to low point draon. ,

1._._._._._._._._._._._._._._._._._._._._._._._.J 6.4.16 Slowly open low side isolation valve and allow transmitter to drain.

1 6.4.17 Close the dry leg low point drain valve below the transmitter three valve manifold.

a 6.4.18 Remove the hose if used. and replace 'he cap on the low pomt drain 4 valve.

6.4.19 Replace cap on transmitter low side s ent.

)

6.4.20 Remose cap on high side test tee above transmitter three.vahe i manifold. Install a length of hose from the test tee into a poly bag.

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,4 I._._......._...-...._.-._....._.I NOTE .

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• It may be flecessary to D!eed several ga :ons of fluta before all att is removed frorr, the I j i line.

i._._._._._._._._._._._._._..._._._._._._._._.J I 6.4.21 Slowly open the high side isolation valve and drain RCS fluid 1 l into poly bag until all air is removed from line. Open the vah e I l fully to obtam the maximum flow rate.

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" 6.4.22 Close the high side isolation valve.

6.4.23 Remove the hose and replace the test tee cap.

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6.4.24 Remose the transmitter high side vent cap.

! 6.4.25 Slowly open the high side isolation valve and fill the transmitter.

6.4.26 Close the high side isolation valve.

6.4.27 Replace cap on transmitter high side vent.

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4 Pages Prw eure w: reoc.avra re.;

aser aone:

' Reduced Inventory Operations 6/8/95 3 OP-041.9 I

l INITI ALS EEp VERIF 5.2.1 (Cont'd) l I

6. Both RHR Pump Discharge Isolation valves have been throttled to limit maximum RHR System flow to 3200 gpm.

a. RHR Pump A Disch Isol,3 754A

b. RHR Pump B Disch Isol,3 754B i

7. RHR Hz Bypass Flow valve, FCV-3 605, has been adjusted to maintain between 3100 and 3200 gpm RHR flow.

8. Verify one Source Range Nuclear Instrument audible count rate is on in the Control Room when fuel is in the Reactor Vessel.

5.2.2 Procedure Steps  !

1. Station an operator at Drain Down Level Indicator Hose.

LI-3-6422 and verify direct communication with Control Room in order to commence logging level every 15 minutes l using Attachment 1. [ Commitment - Steps 2.3.6 and 2.3.8}

2. Commence logging reduce inventory parameters using Attachment 2.

3. Place Letdown Diversion Valve, TCV 3-143 to DIVERT.

C A U TIO N S e

MCS levelindication may be lower than actuallevelduring RCS draining unless large vent paths are provided, e RCS levelindication is connected to Loop A intermediate leg. At high RHR flow rates, the indicated level will be different than actuallevel at the RHR hot leg

-- suction. Refer to Enclosure 2 forminimum require d RCS levelindiention.

-- 4. Verify open RHR Letdown Stop,3-205B.

5. Open RHR LTDN to CVCS, HCV-3142.

6. Throttle Low Pressure LTDN Controller, PCV-3-145 as necessary to maintain RCS Drain Down Level indication on LIS-3 6421 within 4.0 percent of LIS-3 6423 during RCS draining. [ Commitment Step 2.3.8) e ,1 2185 vm F C 3 DC

4 Page 9 08/01/91 3-BD-EOP-E-0 REACTOR TRIP OR SAFETY INJECTION 1

1 BASIS DOCUMENT PTN Procedure Step: 1 j

WOG Procedure Step: 1 BASISt the only bet bei ts added to Reactor trip must be verified to ensure that The safeguards

! the RCS is from decay heat and reactor coolantIfpump systems that protect Neat.the plan th+ reactor

! only decay heat and pump heat are being added to the RCS.a trans cannot be tripped, j POWER CENERATION/ATVS , to deal with ATWS conditions.

STEP DEVIATION FROM WOG GUIDELINE:

TYPE DESCRIPTION 8

The rod bottom lights are checked to be ON vice LIT to conform with plant specific terminology.

a transition to FR S.I. RESPONSE TO 1 The RNO was chan6ed so that NUCLEAR POWER GENERATION /ATVS, will only be made if the criteria fr ~

the Critical Safety function status tree for suberiticality is satisfied. ,

constitutes a reactor trip, and eliminates This changethe need for the complies withoperator the intent i

to make a decision under stress.

of the RSO column provided in ERC Feedback item DW 88-033.

9 The WOG guidelines require initiation of Critical Safety FunctionThe RSO status tree monitoring whenever exiting E-0.for performancethe of this task s provide procedural guidance requirements is eliminated.

need to memorize User's Guide PLANT SPEC-IFIC SETPOINTSt (EOP Setpoint P.2) 54 - Reactor power level just in the Power Range.

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08/01/91 3-BD-EOP-FR-5.1 RESPONSE TO NUCLEAR POWER GENERATION /ATWS 1

BASIS DOCUMENT l

PTN Procedure Step: 1 WOG Procedure Step: 1 ,

l I _ BASIS:

Reactor trip must be verified to ensure that the only heat being added to the RCS is from decay heat and reactor coolant pump heat. The safeguards f during accidents are designed assuming that systems that protect the plant are being added to the RCS. If the reactor only decay heat andthen pump heat cannot be tripped, the control rods should be manually inserted into l the core in order to decrease reactor power.

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STEP DEVIATION FROM WOG GUIDELINE

1 TYPE DESCRIPTION 8

The rod bottom lights are checked to be ON vice LIT to conferm with l

j plant specific terminology.

f PLANT SPECIFIC SETPOINTS:

4 N/A i

P e.

proo.aw. m 8 -

P Pro w. ha.: %m l

6G96 Control Room Annunciator Response _

3 ARP-097.CR 1

5.0 SUBSEOUENT ACTIONS

.....-.1IOTSC g.

ANNUNCMTOR RESPONSE QUlOEUNES lI

• I

j Unit ANPSNPS M be made My swwQizant of aE Annunciators Iat al

. 1)

I limes (whether hey han denrod or are lodred ht). l

, I l Upon recept of an annundstor, take immecGate corrective actions II as I

. 2) RCO-i I necessary, infonnng ANPS of any correcun adens. I

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. 3) Daily Annunciator Response Procedure Usage: '

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(-

• For expected alarms such as l&C working in Racks. *I actual l ARP's is not required. g(

1 g-For common or frequent alarms (WBP, Blender Dekbon) use of*Ithe ARP i lI l required for the first annunciation on the particular shift for theIIday.

Subsequent annunciation does not require ARP consultaton 1

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• For A2 cther alarms the ARP SHALL be consulted as well as II any other i

=

• 1 0520-96 '

I applicable procedures.  :

l._._._._._._._._._._._._._._._._..._._._._._._.J 5.1 Annunciator on Panel A l

Perform Appropriate Attachment 1. Page 15 k 5.1.1 5.2 Annunciator on Panel B 5.2.1 Perform Appropriate Attachment 2. Page 69 5.3 Annunciator on Panel C Perform Appropriate Attachment 3. Page 123 5.3.1 5.4 Annunciator on Panel D Perform Appropriate Attachment 4. Page 177 5.4.1 L 5.5 Annunciator on Panel E 5.5.1 Perform Appropriate Attachment 5. Page 231 5.6 Annunciator on Panel F 5.6.1 Perform Appropriate Atuchment 6. Page 285 5.7 Annunciator on Panel G Perform Appropriate Attachment 7. Page 339 5.7.1 L

umaso

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Page Procoowre idle. 8 _

Pro.coute No .

%.i ose.

Emergency and Off Normal Operating 8/23/95 Procedure Usage 0 ADM 211 i

" 4.0 DEFINITIONS 4.1 Action Verbs All action verbs used in two-column format procedures are defined in 0 ADM-101. Procedure Writer's Guide.

4.2 Critical Safetv Function  !

4 An activity which serves to protect the integrity of one or more of the barriers against radiation release.  !

4.3 Emercenev Operatine Procedures (EOPs)

Plant procedures that specify the operator actions cause plant required parameters to mitig to exceed 3 consequences of transients and accidents that engineered safety features setpoints. ar i reactor protection system setpoints. The EOP network consists of all Optimal other appropnate technical limits.

Recovery Procedures and Funct.on Restoration Procedurer 4,

4.4 Faulted 1 Refers to any steam generator with an unisolable leak in it's secondary pressure boundary of sufficient size to require Safety injection.

j 4.5 Functional Restoration Procedures iFRPs) l Safety Function challenges.

which respond to Critical Those procedures  ;

i Guidance is provided to restore the Critical SafetyofFunction the severity to a and the challenge satis '

condition. Typically, actions are based on R ese procedures are !

may not correspond to " good opera:icnal practice". l i

identified by the procedure identifier F or FR. ,

1 4.6 Local (Locailv) f An action performed by an operator outside the Control Room.

f 4.7 ' Manual blanuallv)

His does not

• An action performed by the operator in the Control Room.

include automatic actions, which take place without operator intervention.

4.8 Optimal Recoverv Procedures (ORPs)

R ose procedures which provide guidar.ce Typically, to recover the correspond actions plant in th efficient manner to a safe and stableThese end state.

procedures are identified by the to " good operational practice" procedure identifiers E, ES, and ECA.

ces,e::n

July 30, 1996 TO: Director f Division of Reactor Controls and Human Factors Office U.

of Nuclear Reactor Regulations. Nuclear Regulatory comm l

Dear Sir:

1996 I am requesting Per the letter sent to me on July 19, Enclosed you ,

an informal review of my written examination. l will find copies of the questions I wish to be reviewed {

along with supporting documentation.

Sincerely, Ralph L. Tetrick 18990 SW 270 Street Homestead, FL 33031 Docket No. 55-20726 4

==

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SRO QUESTION 24 Which ONE of the following describes the Spent Fuel Pool Cooling (SFPC) system basic operation and connections to the Spent Puel i l

Pool (SFP)?

The SFPC pumps normally take a suction on the:

ANSWER:

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a. "High" line near the top of the SFP and inchdischarge siphon breakthrough hole a line 1 foot below the top with a 1/2 6 inches below the water level.

REFERENCE:

SD-041, Fuel Pool Cooling, Purification and Ventilation System' '

page 16. E.O. OF LP 6902141 COMMENT:

(1) the discharge Answer (A) is partically incorrect because, inches below nominal water level a line is routed 10 and (2) the middle of the pool (le 20 feet from top and bottom) inches below the siphon break is 14 inches. (1) the "High" Answer (C) is partically incorrect because, 1/2 feet below the nominal water suction line is approximately 3 ,

level and (2) the discharge line is as stated above. I Both answers A and C are equally correct because they indicate l the suction is from the high line and that there is a siphon  ;

break in the top of the discharge line. An answer of A or C l indicates the operator is aware of the design requirement to l prevent inadvertant draining of the SFP.

RECOMHENDATION:

Accept answer c as an additional correct answer sence both a and c are partically incorrect.

C. "High" line 1 foot below the top of the SFP and discharge through a line at thr bottom of the SFP with a 1/2 inch siphon break hole 6 inches below the water level.

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. I 5D 041 l

01/14/93 l Page 16 FUEL POOL COOLING, PURIFICATION AND VENTILATION SYSTEM in the piping on the discharge of the Goulds SFP  ;

There is a tr.armal expansion loop  !

cooling pump to accommodate thermal sti esses due to pool boiling at 212*F.

Spent Fuel Pool Cooling Puups Refer to Ficure 10. A&8 Three SFP cooling pumps are provided, A, 8 and emergency. 125 ft. TOH.

pumps are . horizontal centrifugal pumps rated for 2300 gpm at Additionally, both pumps are powered from LC C (breaker One 0309) viais manual switch for pumptrans panel P-16. There are two switches located on this panel; and the other switch is for pump 8. They are interlocked such that only one switca l

The pumps are located in the SFP heat exchanger room and are can be closed at a time.

I controlled locally.

It is used only when the l The third pump, emergency SFP cooling pump is also provided.

Power for this pump is provided by a receptacle

SFP cooling pumps are not available. When the emergency pump is used, the SFP i in the cask wash area new fuel room. The Emergency Spent Fuel Pit Pump Motor is not purification loop is bypassed. If its use is required t*e 480V AC power source.

normally connected to a permanent temporary local motor starter / disconnect stand and attached cables reeds to be movec to outside of the Spent fuel Pump Room and the load side cable connected to tre The temporary motor starter / disconnect stand is Emergency Spent Fuel Pump Motor. The temporary normally stored in the new fuel storage room when not in use.

motor / starter disconnect may be connected to provide standby operations a discretion.

The SFP cooiing pumps can take a suction on the SFP through the high suct The high suction line penetrates the SFP near (796) or the low suction valve (797).The low suction line penetrates at a l the top and terminates. Complete of the fuel assemblies and extends downward in this line is preventedtobyalmost a normally thelocked botto siphon draining of the pit by a break (797). There are no closed valve located at the same elevation as the penetration other connections provided on the SFP. The cooling loop discharge line pene and extends straight down towards the stored SFP at approximately l' below the top fuel.

A 1/2" hole is drilled in the discharge lire at approximately 6" below the l

water surf ace, it acts as a siphon breaker.

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SRO QUESTION 63 Plant conditions:

- Preparations are being made for refueling operations

- The refueling cavity is filled with the transfer gate valve open.

, - Alarm annunciators H-1/1, SFP I4 LEVEL and G-9/5, CNTMT l

SUMP HI LEVEL are in alarm.

Which ONE of the following is the required IMMEDIATE ACTION in

" response to these conditions?

ANSNER:

b. - Sound the containment evacuation alarm.

REFERENCE:

' 3-ONOP-033.2, Refueling Cavity Seal Failure, page 5

. E.O. 6 of LP-69C2144 COMMENT:

Annunciator H-1/1 is a entry condition for 3-ONOP-033.1The im (Attachment 1).

verify the alarm is valid. ,

Additionally, the RCO is required to respond to alarms per2). Alar 0-ADM-219 ' Attachment as priority 3 (BLUE) alarms requiring prompt (not immediate) i action. ? a specified operator actions for both alarms per ARP-097.G is to verify the alarms (ie containment Seesump level 3 Attachment recorder a.nd spent fuel pit level indication) .

for the AU actions.

RECOMMENOA"" CN:

Accept a m A as an additional correct answer.

l recorder A. - Ver ff alarms by checking containment sump leve and spent fra; level indication.

1

• = .. ,,.ws 1m.

seem s.i o.o Spent Fuel Pit (SFP) 10/21/92 .

! Cooling System Malfunction l 4

3 ONOP.033.1 kNbNuc \ l.7%E i c50 1.0 PURPOSE 1.1 This procedure provides instructions for response to off no the S pent Fuel Pit (SFP) and the SFP Cooling System includ

^4 4 Level, and High Temperature.

i 2.0 SYMPTOMS 2.1 Annunciators f

2.1.1 H 1/1,SFP LO LEVEL i

2.1.2 H 1/2,SFP HITEMP i

~

2.1.3 H 1/3, SFP HI LEVEL 2.2 Indications (VPB) or by local visual 2.2.1 High/ low SFP. as indicated on LI.3 651 inspect SFP (normallevel is 56 - 10"- 57'2")

2.2.2 Low SFP Cooling Pump Discharge Pressure (PI 3-651B; 651A) j 2.2.3 SFP Filters High .iP (DPI 3-151 A. B, C),

AP across filters should be <10 psid 2.2.4 Low SFP Skimmer Pump Discharge Pressure (PI-3-671 A) i 2.2.5 SFP Skimmer Pump Filters High AP (DPI-3-150A, B, C). -

aP across filters should be <10 psid 2.2.6 SFP Demineralizers High AP (INLET PI-3-655 A - OUTLE (N/A if RWST is on Recire through the SFP Demin)

AP across Demin should be <35 psid 3.0 AL* TOM ATIC ACTIONS .-. . . ...-.-.-... .-.g Q

r.-. l

. -m l Skrs for $FP pps are locatedinside Umt 3 $FP Mx Room.

l l

a. 3A 5FP Pp skt 39212A '

I k.

38 $FP Pp Skt 392128 (Power supply to 3NP212 panelis I fed i from 8kr 30309) J L.-._. . .-. .

l 1.1 Possible SFP Cooling Pump Breaker trip on overload.

3.2 Possible SFP Skimmer Pump Breaker trip on overload. (Bkr 3 4.0 IMMEDIATE ACTIONS

- J; 4.1 Verify annunciated alarm is valid.

Pfootmas N PmcachesT h:

Awove o =:

Annunciator Response Procedure Usage 3/12/96 0 ADM 219 hPCamre Nr 2 (9RGE 1ofP) 3.0 RESPONSIBILITIES 3.1 Nuclear Plant Supervisor (NPS) - De NPS shall provide technical guidance for event maugauon when ARPs are in effect.

Assistant Nuclear Plant Supervisor (ANPS)

- The ANPS should direct the 3.2 detailed event miugauon strategy for the affected unit unless otherwise directed by the NPS when ARPs are in effect.

Nuclear Watch Engineer (NWE) - De NWE should direct non licensed 3.3 operators. if necessary, to determme the cause of the alarm condition and th performance of corrective actions when ARPs are in effect.

3.4 Affected Unit Reactor Control Oxrator SCO) - De affected unit RCO is responsible for the following when ARPs are in effect:

on color code priority and plant Respond to alarms based 3.4.1 conditions.

Reading the ARP in effect and performing the event mitigation 1

>- 3.4.2 strategy for alarms received in the Control Room.

3.4.3 Transition to the appropriate procedures if required by the ARP.

3.4.4 Inform the unit ANPS of abnormal alarm conditions.

operators when the alarm 3.4.5 Coordinate actions with non-licensed condition occurs at local annunciator panel in the field.

3.5 Non-affected Unit Reactor Control Operator - The non-affected unit RCO should mamtam the non-affected unit to a safe condition which does not threaten the event mitigation strategy on the affected unit when ARPs are in effect.

3.6 n ird Licensed Or,erator - ne third licensed ope ator should assist the affected umt(s) RCO m performance of the event mitigation strategy when tj

~

~ 'ARPs are in effect.

Non Licensed Operator (NLO) - ne NLO is responsible for the following 3.7 l wben ARPs are in effect on either unit:

3.7.1 Read the ARP in effect and perform the event mitigation strategy for alarms received at local annunciator panels.

3.'.2 loform the affected unit RCO of the alarm condition.

3.~.3 Performing actions requested from the affected unit RCO or NWE to correct the alarm conditions in the Control Room.

' 2 ste : x

g.,,,

. y, p a.a,. teu emsans rm.-

Approsas Dese:

N12/96 Annunciator Response Procedure Usage 0-ADM 219 MervntrtT 2 (7. CV P-1 e

4.0 DEFINITIONS 4.1 Annunciator Response Procedures ( ARPs)

Plant procedures that specify the operator actions required to mitig consequences of transients that cause plant parameters to exc setpoints.

4.2 Local (Locally)

An action performed by an operator outside the Control Room.

4.3 Manual (Manuallv)

This does not An action performed by the operator in the Control Room.

include automatic. actions which taxe place without operator intervention.

4.4 _Pnontv 1 (Radt Nuclear Saferv a potential challenge to

'Ihese alarms require imtnediate response and reDect These alarms mclude.

plant safety and require protecdve systems te activate.

Si and Reactor / Turbine / Gen Trips.

4.5 Priority 2 (Yellowr Power Production Availability These alarms require irnmediate response and reDect a challenge t equipment or systems that may affect continued plant availability or ti Immediate response to these alarms would be deferred only if a l

recos ery. Failure to properly respond to a Prionty were required by a Prionty I alarm.

2 condition may lead to or contnbute to a higher level condition.

4.6 Pnonty 3 (Bluet investment Protection information that, if alarms require prompt response and provide Prompt action to These unanended, may result in a threat to higher level actions. ii ii this level of alarm may reduce the consequences of the problem by m n m

. equipment damage or material waste.

4.7 Pnontv 4(Whitet Status /Information response and reflect equipment status.

T kse alarms require non-priority transitions, or conditions to be corrected, but do not threaten the unit stnctly "Informauon Only" items, they may availability. Because they are not items are deferred in the face of higher w arrsnt operator action. Priority 4 pnenty items.

4.8 Trnsition A diange from one place to another in the procedures. either from ancoer step or from one procedure to another procedure.

z sG(tc.= x

t

• .e. 3,3 nu u .

i wa w . NI18I3 Control Room Annunciator Response 3.ARP.097.CR H 1.1  :

l INVEST >IENT PROTIciiON BLUE l

~nueve t Nof5( ~ SFP

.rdE-j i{

y LO LEVEL gggg 7

3 -

~

(,'ih]e lop 2.J-5 ii}

l l -

l 6  :

3 s

~

1 234 5678 9 SETPOIh"rS:

56'10" DEVICES:

Level actuator at north end of SFP LT 3 651 OPERATOR 1.

ACTIONS: Verify alarm by checking the following:

a. LI 3 651(VPB)

2. Corrective actions:

a. Dispatch (1) operator to check: Spent fuel level indication (2) Locallevel at the SFP. Power to LT-3 651 (LP 50. Bk (3)

b. Refer Ein a refueling to 3 ONOP-033.1.

configuration SPENT ONOP with FUEL the 033.2.Refue SFPPIT ng avity ~

tran(

c.

terminate refueling operations and refer to 3-S .

d.

required actions. Refer to TS 3.9 for additional actions.

NOTE 1 1

-'. 3 of I

if $FP techng has to be securoet, monstar SFP tempe I

l. e 3 09-033. nou+y neoctor Engmeenng.

1 l* If annunesatoris 00$.reforto0 ADM.214. .........a

....................................... l misalignmenu CAUSES: Actuallow levelin SFP (Evaporation leakage, or SFP sys 1.

2. Less of power to LT 3 651

3. Instrumentation failure .

REFERENCES:

1. FPL DWG 5613 M 3033 Sh 1

2. Tech. Spec. Sectics 3.9

. i .3 :- :: m

, .. 7 1 .

i ,

f Pep 391 -

! J. - 7* /

Dveseensre ses.: a inew l

8/8/92

.' C'ontrol Room Annunciator Response <

a i

3.ARP.097.CR 4 ~~6 3/5 j DCa.si mNTPRQTECTION '

BLUE f in inssums,n , 7 G45 h0 '54 CNTMT 3

l j 3r:10 SUMP E MWL 1

I I 2

Il 3

_____ .bys cmp, z.n .

6 1 234 56789 SETPOINTS:

DEVICES:

30" R 1418 (unit 4 VPA)

OPERATOR ACTIONS:  :

I 2,

1. Verify a.

alarm by checking the following:CNTMT sump record i l

DDPSA102 3. \

2. Corrective actions: uinment sump pumps.

a. Verify proper operation of the con

b. Pump Monitor c,own the sump as required.

RCS parameters for indications ofleak. if applicable. ,

c. Perform 3.OSP.041.1 to determine the RCS leak rate. if applica  !

d.

,._._._._._._._._._._._ y ._._._._._._  !

11annunewtor us oos. refer to 0.AOM-2 tt

. ~

t.w_._._._._._._._._._._._._._._._._._._._._._.;

l CAUSES:

1. RCS leak. l

2. Instrument malfunction.

REFERENCES:

1. FPL DWG 5610.M 12 e 15/JGr,ar

. e i

SRO QUESTION 84

" Verify Which ONE of the following is the basis for step 1, Response to Nuc Reactor Trip", of FR-S.1, ATWS7 ANSWER:

a.

- To ensure that only decay heat and reactor coolant pumps are adding heat to the RCS.

REFERENCE:

page 3-BD-EOP-FR-S.1, Response to Nuclear Power Generation /ATWS,

8. E.O. 6 of LP-6902346 COMMENT:

A review of the corresponding Step 1 of 3-EOP-E-O and 3-EOP-FR-S.1 with respect to reducing reactor power indicates Rods are manually inserted in FR-S.1 but not a difference.

E-0. While the basis documents for both procedures discussonly Basis Docume decay heat and reactor coolant pump heat,BD-EOP-FR-S.1 (ie manually insert disc action control ifrods) the reacto is not tripped.(see attached basis documents and p RECOMMENDATION:

Accept answer C as an additional correct answer.

i if C.

- To alert the operator to take further corrective act on the reactor is NOT tripped.

l I

l

)

2 .

• e: 7 era w . s.. er.c w o itto: w v 6 ces.: ,

06/22/95 l l REACTOR TRIP OR SAFETY INJECTION 3-EOP-E-0 l

l RESPONSE NOT O8TAINED STEP ACTION / EXPECTED RESPON5E 7..........................................................................,

!!QIES i

i

' i l i i i e Steps 1 through it are IMME01 ATE ACTION steps.

! i i Foldout page shall be monitored throughout this procedure. i e

i .........................

a iL.................................................

^

, Manuaily trip reactor. It reactor l

1 verify Reactor Trip: power is greater than 5% QR intermediate range power is NOT

• Rod bottom lights - ON stable or decreasing, TH{!! perform

the following:

o Reactor trip and bypass breakers - OPEN a. Direct operator to monitor Critica' Safety Functions using e Rod position indicators - AT 3-EOP-r-0. CRITICAL SAFETY ZERO FUNCTION STATUS TREES.

• Neutron flux - DECREASING b. Go to 3-EOP-FR-5.1. RESPONSE T NUCLEAR POWER GENERATION /ATWS, Step 1.

l l

I 4

Page 9 08/01/91 3-BD-EOP-E-0 REACTOR TRIP OR SAFETY INJECTION 4 BASIS DOCUMENT I

PTN Procedure Step: 1 WOG Procedure Step: 1 85SI5:

Reactor trip must be verified to ensure that the only heat being added to The safeguards the RCS is from decay heat and reactor coolant pump heat.

systems that protect the plant during accidents are designedIfassuming that the reactor only decay heat and pump heat are being added to the RCS.

cannot be tripped, a transition is made to FR-S.I. RESPONSE TO NUCM POWER GENERATION /ATUS, to deal with ATUS conditions.

STEP DEVIATION FROM WOG GUIDELINE:

.TIPE DESCRIPTION 8

The rod bottom lights are checked to be ON vice LIT to conform with plant specific terminology.

1 The RNO was changed so that a transition to FR S.1, RESPONSE TO NUCi. EAR POWER GENERATION /AWS. will only be made if the criteria from -

the Critical Safety Function status tree for suberiticality is satisfied.

This provides the operator with a clear definition of what  ;

constitutes a reactor trip, and eliminates the need for the operator This change complies with the intent to make a decision under stress.

of the RNO column provided in ERG Feedback item DW-88-033.

9 The WOG guidelines require initiation of Critical Safety Function status tree monitoring whenever exiting E-0. The RNO was modified the to provide procedural guidance for performance of this task so that need to memorize User's Guide requirements is eliminated.

. . PLANT SPECIFIC SETPOINTS:

54 - Reactor power level just in the Power Range. (EOP Setpoint P.2)

e.m 5 eroc a r r m e: l ercr.%r. ca. Approvst Date:

03/30/95 3-EOP-FR-5.1 RESPONSE TO NUCLEAR POWER GENERATION /ATWS RESPONSE NOT OBTAINED STEP ACTION / EXPECTED RESPONSE

(......... ... ..... ........... .......................

NOTE

. .......... ...... 3 i

j i i i i Steps 1 through 2 are IMMEDIATE ACT10tl steps. i a

iL......... . . ... ... . ..................... ......... .........- =. ... .

Manually trip reactor. ir reactor 1 verify Reactor Trip: will NOT trip. THEN manually Rod bottom lights - ON insert control rods.

1

. Reactor trip and bypass i breakers - OPEN

• Rod position incicators - AT ZEP0

. Neutron flux - DECREA5!%

2 verify Turbine Trip: i

a. Manually trip turbine. If '

l

a. All turbine stop salves - CLOSED turbine will N_0T trip, T_ HEN manually run back turbine.

~

IF steam *Iow to turoine causes 3conteci' ed RCS cocidown. THEN close main steamline isolation and bypass valves.

b. Close MSR Main Steam 5.oply . Close main steamline isolation anc bypass valves.

Stop MOVs Remove timing cam to close ai ZERO  :.

c. Reheater timing cam - timing salves. IF any timing valve can NOT be closed. THEN close main steamline isolation and bypass valves.

Manuallj close MSR purge

d. MSR Purge Stea- Valves - CLOSED salves. If any MSR purge salve car. NOT be closed. THEN ciose main steamline isolation in; cypas s nives.

anuailj cren steam supply valves.

3 Check AFW Pumps - ALL RUNNING

Page 8 i

08/01/91 3-80-EOP-FR-5.1 RESPONSE TO NUCLEAR POWER GENERATION /ATWS

! l BASIS DOCUMKNT PTN Procedure Step: 1 WOG Procedure Step: 1 l

i 8 ASIS:

the only heat being added to 1

Reactor trip must be verified to ensure that The safeguards l

the RCS is from decay heat and reactor coolant pump heat.during accidents the plant systems that protect and pump heat are being added to the RCS.

If the reactor

• only decay heat cannot be tripped, then the control rods should be manually inserted into the core in order to decrease reactor power.

! STEP DEVIATION FROM WOG GUIDELINE:

TYPE DESCRIPTION 8 The rod bottom lights are checked to be ON vice LIT to conform with plant specific terminology.

i

) PLANT SPECIFIC SETPOINTS:

l N/A i

r i

i e

4 I

]

, . s SRO QUESTION 96 Which ONE of the following is the lowest level of position responsible for ensuring entries are made in the Technical Specification Related Equipment Out-Of-Service Index?

ANSWER:

b. Assistant Nuclear Plant Supervisor i

1

REFERENCE:

0-ADM-213, page 10 1

COMMENT:

0-ADM-200, Conduct of Operations states that the Nuclear W Engineer (NWE)of the Control Room command function". By relieving the A the NWE becoming theassumes lowest all responsiblities of the ANPS thereby EOOS index. Additionally FP&L Training Dept, see similar test question on our Contractor Exam (Question 88,(ie accepted ANPS attached) and ruled similar as this request NPS since ANPS can relieve the NP3) .

RECOMMENDATION:

i Accept answer d as an additional correct answer.

d. - Nuclear Watch Engineer 4

o w.

' peger.

Pieneen %: 7. r -- re.: I8 Asswove ome:

Conduct of Operations 2/22/96 0-ADM 200 h

3.4.6 Review and approve all unit Plant W ork Orders. pnor to work commencing and ensuring the NWE and RCO are aware of all work outside of the Control Room.

I 3.4.7 Maintain the equi pment out-of-service book in accordance with 0-ADM-213. Technical Specification Related Equipment and i i

Risk Significant SSC Out-of-Service Logbook 3.4.8 Coordinate the on shift training oflicensed operators.

Maintaining a thorough knowledge and understanding of the

3.4.9 following

1. The duties and responsibilities of the ANPS required by the facility operating licenses.

2. Conditions and limitations contamed in the facility operstmg licenses and Technical Specincations.

3. Operating procedures for the nuclear units.

4. Plants' status at all times. [ Commitment - Step 2.3.3]

Speci6 cation Limiting i 3.4.10 Notifying the NPS when any Technical Condition for Operation is entered.

I 3.4.11 Notifying the NPS when any Risk Significant SSC is removed 1 from service. I Nuclear Watch Engineer (NWE) - One Nuclear Watch Engineer will be 3.5 assigned to assist the NPS m coordinating the acuvities of Licensed ana Non. Licensed personnel during routine. complicated, or infrequent evolutions.

The NWE reports to the NPS and is responsible for:

3.5.1 Performing duties assigned by the NPS or his designee for each unit-

~

activities of the Control Room with other

' ' 3.5.2 Coordinating the operations and plant personnel to achieve safe. reliable, and efficient unit operation as directed by the NPS/ANPS.

3.5.3 Supervising and coordinating the operation of plant equipment and systems when assigned by the NPS.

3.5.4 Acting as the Fire Brigade Chief or Shift Communicator. when assigned. but not both.

3.5.5 Routinely relieving the ANPS of the Control Room command In an function to enable the ANPS to leave the Control Room.

emergency. funcuon as the NPS if required.

a vess.u a

- = - . - - . _

s * *

QUESTION 88 RO & SRO Which one of the following is correct regarding who has control and responsibility for the issuance of ICCS keys?

a. NPS

b. NPS/ANPS

c. NPS/ANPS/NWE

d. NPS/ANPS/NWE/RCO I

ANSWER: A (3.4/3.6)

~

REFERENCE:

294001K1.05 ADM-205, Section 10, Key Control Key control, clearly statesissuance, ICCS are under ADM-205 section S.4, nothing the control of the NPS. The question asks about i else. NO CHANGES REQUIRED.

final decision made to QC reviewer recommendation considered and accept A or B as correct because of numerus exceptions to the NPS issue requirement. ANPS may be acceptable for issue of ICCS keys under certain conditions. ADM-205 does not have any provisions for the RCO or NWE ever issuing ICCS keys.

3 o

J a*