ML18005B109

Submits Reactor Operator Written Exam Comments,Including Ability for Training Personnel to Review Exam in Advance & Accepting Answers a or D for Question 2.21 Re Loss of Reactor Coolant

ML18005B109
Person / Time
Site:	Harris
Issue date:	08/24/1989
From:	Richey R CAROLINA POWER & LIGHT CO.
To:	Brockman K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
CON-NRC-624 HTU-89-254, NUDOCS 8910200125
Download: ML18005B109 (24)

Text

ACCELERATED DlQBUTION DEMONSTROJON SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8910200125 DOC.DATE: 89/08/24 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION RICHEY,R.B. Carolina Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION BROCKMAN,K.E. Region 2, Ofc of the Director

SUBJECT:

Submits reactor operator written exam comments.

'DISTRIBUTION CODE: IE42D COPIES RECEIVED:LTR J ENCL SIZE:

TITLE: Operator Licensing Examination -Reports f NOTES:Application for permit renewal filed. 05000400 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 BECKER,D 1 1 INTERNAL: ACRS 2 2 AEOD/DS P/TPAB 1 1 NRR SHANKMAN,S 1 1 NRR/DLPQ/HFB 10 1 1

/J3~/OLB 10 1 1 NUDOCS-ABSTRACT 1 1 REG FILE 02 1 1 RGN2 .FILE 01 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 NOIR 'ZO ALL RIDS" RECIPIENTS:

PIZASE HELP US K) REXQCE %ASTE.'GMZACT 'IHE DXUMEÃZ CDNZROL DESK, RXN Pl-37 (EXT. 20079) K) EZZHINREB YOUR MQ93 PKR DISX%G33UTICN LISTS PQR DOCXREÃZS YOU iv TOTAL NUMBER OF COPIES REQUIRED: L R 4 ENCL 14

ENCLOSURE 3 CIA@I Carolina Power & Light Company Harris Training Unit Post Office Box 165 New Hill, North Carolina 27562 August 24, 1989 FILE: NTS-3501 SERIAL: HTU-89-254 Mr. Kenneth E. Brockman US NRC - Region II 101 Marietta St. NW Atlanta, GA 30323

SUBJECT:

RO NRC Written Exam Comments NRC-624

Dear Mr. Brockman:

On August 21, 1989, Shearon Harris Nuclear Power Plant received NRC written RO examinations. The examination comments are submitted by CP&L.

Copies of reference mat'erial are included where indicated.

Should you need any explanations or additional reference material, please do not hesitate to contact the SHNPP Manager Training, Mr. A. W. Powell, at (919) 362-2618.

R. B. Richey Manager Harris Nucl ar Project HWS/tpw Attachments h

cc: Mr. W. H. Bradford (NRC-SHNPP) bcct L. H. Martin A. Baxter (Shaw, Pittman, Potts & Trowbridge)

C. Carmichael (2)

A. B. Cutter G. L. Forehand S. McManus C. H. Moseley D~ L. Tibbitts Day File Document Services g2+t ) 1 "' ~,"./t la'wg'+

AitCtt=i-'.>'-<<'04'-'<'

g ppr,

AUGUST 21, 1989 NRC EXAM RO EXAM

GENERAL COMMENT

S The ability for utility training personnel to review the exam in advance is a worthwhile practice as evidenced by the relatively small number of comments made on the following pages. This is a welcome change to past NRC operator examining practice and. should be continued for future e,.ams.

Three of the plant,-wide generic responsibilities mult'ple-choice questions required .the examinees to make ine distinctions in wording between the correct answer and one or more of the d'stractors. These sub"leties go on beyond the knowledge requ'ed for a licensed =eactor operato~ in that, were -he quest=on to arise in an ope~ at'na environment, the ope!. at': 1. wou ' consult pl ocedures available in th: Control Room. Questions 3.38 (all chcices), 3. 9 (choices b & c), and

",.42 {choices a,b,6 c) apply.

Mo.~- deta=..led comme!';ts are not=d on the follow'ng pages.

NRC QUESTION 2.14 The control room has recieyed a RH-11 alarm for the Tank Area Drain Transfer Pump Monitor. The Tank Area Drain Transfer Pumps are aligned to the storm drain system. According to AOP-008, "Accidental Release of Liauid Waste", which one of the following is NOT .one of the automatic actions which would occur as a result of an accidental effluent release?

a. The isolation valve to the oil separator closes.

b. 1E'D-109, "Tank Area Drain Pump Discharge to Storm Drains",

closes if the leak is in the Tank Building.

c. The isolatio:: valve to divert t!:e flow to the Secondary Wast= Holdup Tan) opens ~

d. Tank Area D airs Transfer Pumps shut off ANSWER 2.14 (1.00)

'PGL COMMENT Page 4 of AOP"008 (attached) shows that the automatic action for a>> =-la.m on the Tank Area Dra'n Transfer Pump Moritor is that the

-.ank Area Drain Transfe - Pump "shuts off h= s ~olm dLain s .'stem ..he c the~

if it was discharging to actions 11sted undeL Automatic Ac,'"ions on 'hat page are or a>> alarm of the Turb'e Buildirg Drain f

Honitor vnlv and are not applicable in the situatxon posed by the.

question. As'"urrentl; worded, choices a, b', and c a' a l orrect

nce nc:.'.- cf -hem cc =u in this case.

RECOMMENDATION Eith=r acce. t answers a, b, o -

c as cor. ect or delete the ques=ion since three c the four choices are co=rect.

CLUES~Ctl 2. i4 OS1 TURBINE BUILDING DRAIN OR TANK AREA LEAK nP-SECTION 1.0 1.0 SYMPTOMS

1. Local radiation alarm on Tank Area Drain Transfer Pump Monitor.

2. RM-ll alarm for Tank Area Drain Transfer Pump Monitor.

3. "REFUELING WATER STORAGE LOW LEVEL" alarm. ALB 4-3-4 4~ "REFUELING WATER STORAGE TANK 2/4 LOW-LOW LEVEL" alarm. ALB 4-3-3 a

i

5. "REFUELING WATER STORAGE TANK LOW-LOW 2 LEVEL" alarm.

ALB 4-3"2 6., "REFUELING WATER STORAGE TANK EMPTY" alarm. ALB 4-3-5

7. RWST or RMWST Level Indication Decreasing Abnormally.

8. "RWMU STORAGE TANK MINIMUM - HIGH LEVEL" alarm.

ALB 8-1-.4 Verbal notification to the. Control Room.

2.0 AUTOMATIC ACTIONS

l. Turbine Building Drain Monitors 3528 Pe N

~.- ~

Sf-i '. Isolation Valve to the Oil Separator shuts and the

~ a ( isolation valve to divert the flow to the Secondary I ='Pl ', Waste Holdup Tanks opens; 2~ Tank Area Drain Transfer Pumps Monitors 3530 Drain Transfer Pumps shut off if they are discharging to the storm drain system (continue to operate if they are lined up to the FD system).

3 ' OPERATOR ACTION 3.1 Immediate Action None

3. 2 Follow-U Action NOTE: A radiological effluent release may require the R initiation of the SHNPP Emergency Plan. Refer to PEP-101 and enter point X on the EAL Network.

AOP-008 Rev. 4 Page 4 of 10

NRC QUESTION 2 '1 EOP-EPP-020, "SG R With Loss of Reactor Coolant: Subcooled Recovery", contains a CAUTION which reads: "Steps to depressurize the RCS an terminate SI should be performed as quickly as possible after, the coo'ldown has been initiated...", Which one of the following is the'urpose of this CAUTION?

a. Mi!1imize possible pressurized thermal'hock of the reactor vessel.

tubes'ossible

b. Minlm1ze plessul1zed the!.li'al shock o the S/G "nsul. e he RCF mznimum seal delta-p is ma'ta 'ed.

d. Minimize the potential or S/G over ill.

ANSWER 2.21

a. (1.00)

'PGL COMMENT Althou-h t'-~e correc" answer spec1f ied is a q; o" e o'he CAUTION

p. 1cr to step ".4:;f EPP-020, i provides onlv hal= o'he reason o" per Glm1!>' a 'pid coo down and dep=essi!1 iza 1on 'llnde1 these condi t:ons. Th= <i'>"s' nghc llse Owne1 s Gl ouu E!iie1 gency Pesponse Guicel" ne Background Documents provzde additiona 1nformataon. In the step c: script1on f ol ECA-3. 1 step 10 ( at. aci;ed } wh1ch corresponds to SHNPP EPF-020 step 14(also attached.), the first sentence of the basis for that step provides a more complete rationale or quic); cooldown and depressurizatlon. The two !.easons found there are to minim'ze both leakage of reactor coolant and radiological releases from the ruptured steam generator. The reasons provided here are directly related to the. steam generator overfil'oncern in that a continuing leak of coolant into the steam generator will overfill it and lead to a radiological release. This release starts due to increasing steam generator pressure caused by 1ncreasing level and 's e!'acerbated b-. a release of st=am/water mi::.tur: with th= over.ill condition.

RECOMMENDATION A-cep. e"'=he -

cho1ces a. o d. 'a= acceptable espo!.se."-.

AVESTA tod 2,.21 C~e i)

SGTR 'MITH LOSS OF REACTOR COOLANT: SUBCOOLED RECOVERY Instructions Res onse Not Obtained CAUTION Steps to depressurize the RCS and terminate SI should be performed as quickly as possible after the cooldoMn has been initiated to minimize possible pressurized thermal shock of the reactor vessel.

14. Initiate RCS Cooldown To Cold Shutdown:

a. Check SGs - AT LEAST ONE a. Cooldown using any of the INTACT SG AVAILABLE following (listed in order of preference):

1) RHR system

2) IF CNMT conditions are normal, THEN dump steam using faulted SG.

3) Consult TSC AND determine Mhether to dump steam using faulted OR ruptured SG.

GO TO Step 15.

b. Dump steam from intact SGs using any of the following (listed in order of preference):

1) Condenser steam dump

2) SG PORV$

3) Locally operate SG PORVs using OP-126, "MAIN STEAM, EXTRACTION STEAM AND STEAM DUMP SYSTEM

4) TDAFW pump

5) MSIV before seat drains EPP-020 Rev. 3 Page 18 of 55

STEP DESCRIPTION TA8LE FOR ECA-3.1 STEP 10 Correl a44 S +u EPPGZo Sfgp k4 STEP: Initiate RCS Cooldown To Cold Shutdown PURPOSE:. To begin or continue a controlled RCS cooldown to cold shutdown temperature as quickly as feasible in order to minimize total leakage of reactor coolant BASIS:

The RCS must be cooled and'depressurized to. cold shutdown conditions as quicklyg as possible to minimize both leakage of reactor coolant. and radiological eleases from the ruptured steam generator. This step establishes a 100'F/hr cooldown rate, which balances the need for rapid RCS cooldown with the concern of pressurized thermal shock of the reactor vessel. The preferred method is, steam release from the intact steam generators to the condenser since this conserves feedwater supply and minimizes radiological releases. If steam dump to the condenser is unavailable, atmospheric steam releases via the intact steam generator PORVs provides an alternative means of cooling the RCS. In the unlikely event that no intact steam generator is available, one must select either a faulted steam generator or ruptured steam generator to cool the RCS until the RHR System can support further cooldown t'o cold shutdown.

I ACTIONS; 0 Determine if RCS cooldown rate is less than 100 F/HR 0 Use RHR system 0 Dump steam to the condenser from intact SGs 0 Dump steam from intact SG PORVs 0 Dump steam from intact SGs by other plant specific means 0 Control feed flow to faulted SG to cooldown RCS 0 Dump steam from ruptured SG INSTRUMENTATION:

o RCS hot leg temperature indication

. o Wide range RCS cold leg temperature indication o Condenser status indications o Steam dump valves position indication o SG PORVs position indication o RHR System indication ECA-3. 1 HP-Rev. 1 0110V:1b

NRC QUESTION 3.29 Match the Digital Rod Position Indication System (DRPIS) alarms= in COLUMN A with their correct function in COLUMN B. (each alarm has only one correct function)

COLUMN A-ALARMS COLUMN B-FUNCTIONS

a. "PPI Rod Deviation" 1. Indicates banks B,C,D on bottom with bank A above 6 steps, "RPI Urgent'. Indicates + or 12 step deviation between rod and bank demand position ~

c. "One Pod A" Bottom" 3. Indicates ei ror or f ailure f rom both data cabinets.

'PI Non u> gent 4. Indicates dropped rod or improper bank sequence.

5. Indicates + or - 12 step deviation between any 2 rods -'n a bank.

6. Ind'cates con.lict between two deviat Eon cax ds ~

ANSWER 3.29

a. 2 (0.5 each)

b. 3

c. 4 6

CP&L COMMENT

'There are two rod deviate.on alarms fed by DR?I. One is, a local rod deviation alarm on the DRPZ panel on AEP-1. The attached table 6.6 fr'om SD-104, "Rod Control System", shows that this alarm is driven bv either a shutdown rod at or below 210 step.=., or any two rods in control bank d'ffer b'reater than or equal to 12 steps. A second alarm fed by DRPI'is the "Rod Dev/Seq N1S Power Range Tilts" alarm on Ma'n Control Board ALB 13 window S-5. The attached pages rom RODCS-LP-3.1 indicate this alarm actuates on rod out of sequence,

+/- 12 step deviation between D".PI and bank demand, +/-12 step deviation between any two rods in a bank, or shutdown rod below 218 steps. Note that both functions 2. and 5. from the question are included fo this second deviation alarm. Since the quest on does ~

not spec'-y which DPP d=iven rod deviat'.on alarm is under e:.a..inst on, then either func-io>> 2. ox =, sat'factorily answers th& qu=s 3.0n.

CP&L COMMENT TO NRC QUESTION 3.29 (contd)

RECOMMENDATION Accept either answers 2. or 5. for part a. of this question,

OS4 0 g VC.Q C lOh) 8-2-9 cs-.

~<<)

Table 6.6 SD-104 Rod Control System D.R.P.I. Alarms Alarm Location Cause RPI Urgent Alarm ALB-13

• 1 ~ A and B Data Failure, or (6-1) 2. A and B Data differs by more than one bit, or

3. Combined A and B Data is binary 39 or greatere RPI Non-Urgent Alarm ALB-13

• 1. Data A Failure, or (6-2) 2. Data B Failure, or

3. Central Control or Rod Deviation card removed, or

4. Rod Deviation Cards differ

'in output.

Two or More Rods at Bottom ALB-13

• 1., Combined A and B Data is (See, Note 1) (7-3) zero, or ~

d/C

~

2. Urgent Alarm, for 2 or s/i more rods.

One .Rod at Bottom ALB-13

• 1.. Combined A and B Data is g/C (See Note 1) (7-4) zero, or

2. Urgent Alarm, for one or

)

F/l more rods.

Rod Deviation Local DRPI ~ 1 ~ Shutdown rod at or below Display 210 steps, or 2~ Any two rods in control

,bank differ by greater than or equal to 12 steps.

General Warning Local DRPI 1. Data A Failure, or Display 2. Data B Failure, or 3~ Urgent Alarm.'

inc Local DRPI Display also Input also seat to ERFZS for computer generat-ed alarm (Eee ED-163)g Note 1: Rod Bottom alarms are blocked by the DRPI logic when rods are withdrawn in their normal sequence during a reactor startup. During A/C normal startup, the alarms will clear when all shutdown bank and control bank A rods are off the bottom, even though control bank B, s/I C, and D rods remain on the bottom. However, rod bottom alarms will occur if a withdrawn rod drops or if the normal withdrawal sequence is not followed.

SD-104 Rev. 5 Page 28 of 47

G}UEsi<oLJ 3. z9a.

RODCS-LP-3.1 Cv ~<<)

KEY AIDS 2.5.3 Interlocks A. Control Bank D rod withdrawal limit (C-ll)

1. From P-A converter

2. Bank D a't 220 steps

3. Stops pulses to bank D out

4. Automatic only B. Central Control Card if

l. Automatic disconnect one card is not in agreement with the other two

2. Generates "Central Control Failure" alarm 2.5.4 Alarms 2,.5.4.1 Di ital Rod Position Indication S stem Alarms A. Computer Alarm Rod Dev/Seq NIS Power Range Tilts

l. On ALB-13

2. Causes

a. Rod out of sequence

b. + 12 step deviation Rod Bank demand position

c. + 12 step deviation between any two rods in bank

d. Shutdown rod below 218 steps

3. + 25 steps deviations between bank demand and DRPI with rods B. "RPI Rod Deviation" alarm p

l. ALB-13 v 2. + 12 step deviation between rod and bank demand position 18 of 25

Qu EST ion B.Z9~.

C~e S)

RODCS-LP-3.1 KEY A!DS

3. LED's 1, R and 2 indicate

a. Position of shutdown rod below deviation limit

b. Deviations between rods in control bank exceed limit

c. Urgent Alarm has been generated C. "One Rod at Bottom" alarm (< 20 steps)

1. ALB-13

2. Causes

a. One rod on bottom (< .20 steps)

b. One bank on bottom and.next sequential bank is not

3. Defeated when banks B, C, D on bottom and bank A above six steps

4. Indicates

a. Dropped rod

b. Improper bank sequence D. "Two or More Rods at Bottom" alarm (< 20 steps)

l. ALB-13

2. Two or more rods of one bank on bottom

3. Defeated if banks B, C, D on bottom with bank A above six steps E. "RPI Urgent Alarm" annunciator

=-1: ALB-13

2. Error or failure from both data cabinets

3. "One Rod at Bottom" alarm on ALB-13

4. "Urgent 1, 2, 3" alarm flashing on display panel

a. Error in both A and B data from Data cabinets A and B 19 of 25

NRC QUESTION 3.31 List the four (4) sources of control signals which are used by the SGLCS to positioh the feedwater flow control valves. (0 '.5 each)

ANSWER 3.31

1. Steam generator level (0

2. Feedwater flow

'5 each)

3. Steam flow

4. Power demand level or reference level (first stage turbine pressure or reactor power) ~

CP&L COMMENT The above answer omitted another source of a signal for use by SGLCS. Steam pressure is used for density compensation 'of the steam flow signal. Th~s 'is shown in the text of SD-126.02 on page 9 (attached) and in figure 7.2 of the same SD. Since the question asked or signal sources, then steam pressure is an acceptable respons=. It actually has more impact on SGLCS operation than first stage turbine pressure since we operate with a constant programmed steam generator water level of 66'.".

RECOMMENDATION Accept steam alternate answer. This results in having "o l'st presSure as four answers an per the question statement out of five possible al"ernatives for an acceptable response.

QgCSltog 3 Qi C~e ~)

OS3 4.1 Normal Operations (continued)

The summation of the pressure and momentum effects is shown in Figure 7.8. The SGWLCS will act to maintain SG level at 66Z level after these transients. The SGWLCS is designed to reduce the size and length of the level oscillations. Thus, normal level oscillations are within the bounds of reactor trip at the low-low SG level and turbine trip at the high-high SG level.

During normal operation (15<<100X power) the SGWLCS provides automatic SG level control by using a three element controller which senses steam flow, feed flow, and SG level (Figure 7.2).

The ste flow sed and corrected E densit b a steam summer. A flow error signal is produced by subtracting the feed flow signal from the steam Elow signal ~ The flow error signal.

goes to a proportional plus integral (PI) controller.

The level program receives an input from the turbine impulse pressure transmitter. The impulse pressure signaL is multiplied by zero and 66 percent is then added to give a constant Level Program of 66 percent regardless of power level (Level program ~

zero x Pim + 66 percent).

Actual SG level is sent through 'a lag circuit to dampen out natural., oscillations in the level signal. A level error signal i.s produced by subtracting actual level from program level.

The level error signal is sent to the PI controller. This PI controll'er allows level to dominate over the flow error signal.

It also eliminates steady state level and flow errors. The level and flow error signals are added to produce a total error signal. This total error signal is the output of the PI controller when it is in AUTO. The output of the controller can be manually controlled by the operator. The output then goes to the IP converter to position the FW control valve.

4.2 Start-U and Shutdown 0 erations Start-up and shutdown operation is considered low power operation, so the feedwater bypass control. system (see Figure 7.3) will be used. It will give stable level control. from 0-25Z reactor power. Normally the change over, point to and from the main Eeedwater control system will be at about 15Z power.

This system shares the level program and level detectors with main FW Control System. In place of a flow error signal, the FW Bypass Control System uses nuclear power channel N-44 as an input of anticipated steam demand. This arrangement is used because the flow signals are unstable at low power levels and transient response.

it gives better SD-126.02 Rev. 1 Page 9 of 30

Cl I

SIEAI1PIIESS lRIISIIVCOIP l Pl PFIOOIIM

~

AClUALLEVfL LEVEL ARbQX Fl LEVEL EIIAOR

$ 1I1 fLOW FLOW GIN%

AMO 0

oo

$ 1EAFI 684AATOO Ll Q

0 P.l. GMOOLLER fffO PL&PS NAWS ~ VEMF F'WIV FCV - IE EEIWAIEFI COHIIIOL VH VE SD I 26.2 t WIV - FlEOWAIFR ISLA lKFI VH VE M h IN FEED% hTER OONTROL FIGURE 7.2

NRC QUESTION 3.35 List all the signals that will generate a containment ventilation isolation signal.

ANSWER 3.35

1. High radiation (on 2/4 containment ventilation isolation lllon 1 to 1 s ) ( 0 ~ 5 )

Anv SI sig:lal (0. 5)

(low steam line pressure)

( low PZR pl essul e )

(h'h containment pressure)

(illanual "n1t1at 1on )

3. Manual phase A or B containment isolation actuation (0.5)

CP&L COMMENT SHNPP has no manual conta1>>ment phase B isolation. Phase B 1solat ion can b initiated by manually actuating containment spray.

Documentat'oil of ths s in SD-103 section 4.3.6 (attached) which

~

describes how the Containment; Ventilation Isolation S'gnal (CVIS) is actuat=-d. It 1s assumed that the part of the NPC answer key in pa .entheses is not reauired for full credit s'nce the 2/4 logic men ioned, 1n alswex 'tern 1. is not asked for by the auestion 8 tatemen i ~

RECOMMENDATION Delete "...or B containment isolation actuation" from answer item Add new answer item 4. "Manual containment spray actuat,ion signa'". Each response should be worth 0.375 vice 0.5.

a.van(ou S.S~

(p~e ~)

OS4 4.3.2 Main Isolation"'(MSIS) (Fi ure 7.28) (continued)

'zovided on MCB. Main steam isolation is addressed in SD-l26.

The steamline isolation signal also enables auxiliary feedwater isolation.

4.3.3 Auxiliar Feedwater Isolation (Fi ure 7.29)

Auxiliaz'y feedwater isolation is derived from the main steamline isolation coincident with a high steam line differential pressure (hP). When low steamline pressure is detected in one loop, that loop's auxiliary feedwater line is isolated.

4.3.4 Main Feedwater Isolation (MFIS) (Fi ure 7.24)

A Main feedwater isolation consists of closure of the main feedwater valves, the feedwater isolation valves, the feedwater bypass valves, and tripping of the main feedwater pumps and the-turbine. The MFIS is derived from an SI signal or a two out of four Hi-Hi steam generator signal (P-14) Reset of the bypass

~

valves and main feedwater isolation valves is accomplished by the Train A and Train B reset switches on the MCB. Closure of only the main feedwater valves will automatically occur when the reactor is tripped (P-4) and a two-out-of-three low Tavg signal occurs.

4.3.5 Containment Isolation SA (T) (Fi ure 7.30)

Containment isolation SA results from either a safety injection signal (Figure 7.27) or from manual actuation of either of two MCB switches. Reset of the T signal is accomplished by actuation of the Train A and Train B Containment Isolation 8A Reset MCB switches. See SD-114 for more discussion.

4.3.6 Containment Ventilation Isolation (CVIS) (Fi ure 7.30)

Containment ventilation isolation results from a 5A containment isolation signal, safety injection signal (Figure 7.27), from a Manual Containment spray actuation signal (Figure 7.31), or a high radiation signal fzom the Radiation Monitoring System containment radioactivity detectors (RM-1CR<<3561A, B, C, & D). The CVIS signal can be reset by the Train A and Train B Containment Ventilation Isolation Reset switches on the MCB. See SD-114 for more discussion. (See Section 4 ')

4.3.7 Containment S ra (CSAS) (Fi ure 7.31)

Containment Spray results from either a two out of four, Hi-3 containment pressure, or simultaneous manual actuation of either group of two MCB switches. The CSAS signal can be reset by the Train A and Tz'ain B Containment Spray Reset switches on the MCB.

It is noted that the manual spray actuation controls also initiate Containment Ventilation Isolation and Containment Isolation Phase B. Reference SD-112.

SD-103 Rev. 4 Page 27 of 83

(p o

SAF ETY ItuECTION SIGNAL (FIG. 7.27)

CONTAINMENT ISOLATION 9 A p

ACT. ACT ~f CONTAINt1ENT .

CHI1T ISOL 'ROt1 RADIOACTIVITY O'A RESET DETECTORS l1ANUALSPRAY ACTUATION (FIG. 7.3 I) RC RC 356I A

fRC 356I C

AX RC 356 I 3561 S 8 S'8~

CONTAINMENT R L VENl ILATION RESET ISOLATION MANUALRESET 2/4 R L SS I SEC CONTAINMENT ISOLATION )t(A 9

C CONTAINt'IENT VENTILATION ISOLATION SD 103 CON1 A I NMENT ISOLATION OA f(n 0

FIGURE 7.30 5c uj tQ

ENCLOSURE 4 NRC RESOLUTION OF FACILITY COMMENTS RO Examination guestion 2.14 NRC resolution: Comment accepted. guestion deleted from exam.

Section and total point values adjusted accordingly.

guestion 2.21 NRC.resolution: Comment not accepted. The facility comment correctly references the basis for the initiation of RCS cooldown to Cold Shutdown.

The question asked for the basis of the CAUTION preceding this step. The CAUTION addresses the reason for terminating the SI as quickly as possible "after the cooldown has been initiated". By looking at the entire Step Description for ECA-3. 1 step 10, (facility only referenced the first page) the KNOWLEDGE portion describes verbatim the reason for this CAUTION. It should also be noted that this question was written by the facility representative during the exam pre-review.

guestion 3.29 (a)

NRC resolution; Comment not accepted. Each alarm in the question is clearly identified by the use of quotation marks indicating the actual wording of each alarm. The alarm referenced in the facility corrment, which would make 85 a correct response for (a), is the "Rod Dev/Seq NIS Power Range Tilts" alarm. The question asked for the "RPI Rod Deviation" alarm.

guestion 3.31 NRC resolution: Comment partially accepted. As shown in the reference provided the P. I. controller in the SGLCS takes input from two comparators: Flow error (compensated steam flow - feed flow) and Level Error (actual level - program level). Since steam pressure is one parameter that is used to determine compensated steam flow, it will 'not be counted as an incorrect answer. However, to receive full credit all signals as mentioned in the answer key must be listed.

guestion 3.35 NRC resolution: Comment accepted. Answer was taken from CVS-LP-3.0 p.23 which is in error. Answer key changed to reflect facility comment. Each response will be worth 0.375 vice 0.5.

ENCLOSURE 5 SIMULATOR FACILITY REPORT Facility Licensee: Carolina Power and Light Company Facility Docket No.: 50-400 Operating Tests Administered On: August 22 - 24, 1989 During the conduct of the simulator portion of the operating tests, the following items were observed:

ITEM DESCRIPTION Main Feed It was not possible to override the automatic trips Pumps associated with the Main Feed Pumps.

During a scenario in which both Main Feed Pumps were lost, the simulator went into a loss of all AC unexpectedly. This pr'oblem was reviewed and corrected while the exam team was still on site.

On two occasions the simulator froze during the running of the exam scenarios.

This effected the exams in that it detracted from the sense of realism.