Text

Forwards Exam Outline for NRC Exam to Be Administered Wk of 960819 at Plant
	ML20147C259
Person / Time
Site:	Robinson
Issue date:	06/17/1996
From:	Natale T; CAROLINA POWER & LIGHT CO.
To:	Steiner P; NRC
Shared Package
ML20147C236	List: IR 05000261/1996301; ML20147C259;
References
	NUDOCS 9702060052
	Download: ML20147C259 (440)
	v • d • e

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1 To: Paul Steiner

. Date: 06/17/ %

From: Thomas Natale

Subject:

Test outline submittal and product delivery As required by the pilot guidance, enclosed you will find the test outline for the NRC exam to be administered the week of August 19,1996 at H.B. Robinson. If there are any questions regarding the test outline, please contact me at your earliest convenience.

In addition to the test outline, I would like to provide you with a schedule of expected delivery l dates for the exam products to support the pilot guidance due date of July 19,1996. '

- June 21 - 35 questions /1 scenario /10 JPM's and associated questions June 27 - 45 questions / 2 scenarios /10 JPM's and associated questions July 10 - 50 questions / 2 scenarios /10 JPM's and associated questions The above deliveries will be draft products to support the final exam submittal date of July 14, 1996. Please provide any comments on these draft exam products at your convenience to support the final exam submittal date. Your assistance, teamwork, and communications with our -

examination team will be the key to a successful exam.

/ ) s \

l Superlrite (ent tOperations Training 09WSod 64^ h /f 9702060052 970124 NMtgI cef ]eA ) b'Mi -

PDR ADOCK 05000261 V PDR

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ES-301 SCENARIO EVENTS ES-301-3 SIMULATION FACILITY: H. B. Robinson SCENARIO NO.: DSS-QQ1 EXAMINERS: __, APPLICANTS :

INITIAL CONDITIONS: The Unit is at 100% oower. The following eauioment is out of service:"B" Charging Pumo for an oil change (OOS for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months /back in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). SDAFW oumo for steam inlet line leak (OOS for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months /back in 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months ). "A" CCW oumo for excessive vibrations (oumo uncoupled).

TURNOVER: Maintain current olant conditions. Track eauioment out of service and orecare for post maintenangs -

testing to return eauinment to service. Boron concentration 875 com. CBD at 218 steos. cauillibrium xenon.

EVENT MALF. EVENT EVENT NO. NO. TYPE DESCRIPTION MFI SIS 01 A C SI Failure to Auto Initiate MFI SIS 01B RFI CFWO83 C "A" and "B" AFW pumps auto start failure RFI CFWO84 MFI CCWlBC C Trip of running CCW pump after Path-1 entry point C i 1 CORDS RCS I FT-416 fails low (RCS Loop Flow Transmitter) f FT:416 !

2 MFI CRF03A ROD R Dropped Rodfrurbine Rtmback G-3 3 MFI RCS09A C 100 gpm RCS Leak 4 i

N, R Power reduction due to excessive RCS leakage.

4 MFI RCS01 A M LOCA

(N) Normal, (R) Reactivity, (I) Instrument, (C) Component, (M) Major Developed By: Approved By: t Examiner: ChiefExaminer:

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SCENARIO DESCRIPTION (DSS-003) l After shift turnover and allowing the crew to walkdown the board, the initiating event will be failure of a Loop 1 flow transmitter FT-416.

When the transmitter has been removed from service 1 control rod will fall into the core initiating a turbine runback. After the plant is stabilized, an RCS leak develops which will require a plant shutdown.

The leak will escalate to a LOCA requiring a manual reactor / turbine I

) trip and will result in a manual or automatic Safety Injection signal.

PATH-1 will be entered and followed to mitigate the accident. Safety

injection will fail to automatically initiate requiring manual action.

The MDAFW pumps will fail to auto-start requiring identification and manual actuation. A failure of the running CCW Pump will require the use of EPP-015 due to insufficient Supplement D components.

The scenario should progress until EPP-015 entry is directed. The exercise may be terminated at any time at the evaluator (s) discretion after EPP-015 entry.

ES-301 SCENARIO EVENTS ES-301-3 SIMULATION FACILITY: II.B. Robinson SCENARIO NO.: DSS-005 EXAMINERS: APPLICANTS :

INITIAL CONDITIONS: The Unit is at 100% cower. The followine eauioment is out of service:"B" MDAFW for motor reolacement (OOS for 8 hrs /no oroiected return time). IIVII-l for vibration concern:

(OOS for I hr/back in 3 hrs).

TURNOVER: Maintain current olant conditions. Boron concentration 1017 com. CBD at 218 stecs.

cauillibrium xenon.

EVENT MALF. EVENT EVENT NO. NO. TYPE DESCRIPTION MFI EDG4A C "A" and "B"IIIISI Pump auto start failure RFI CFW83 C "A" MDAFW auto start failure RFI CFW85, C VI-8A, B, C fails to auto open 86 and 87 1 CORD 1 PT-445 Fails high (pressurizer control channel)

PT:445 2 MFI SGN2B M 100 gpm steam generator tube leak 3 MFI CND2 C Condenser air inleakage 4 CND02 300 C, R Condenser Air inleak (Rapid load reduction) (Leak increases after initial actions taken) 5 MFI EPS13 C Startup transformer relay malfunction !

6 SGN0lli M,C "B" S/G safety valve failure af ter entry into EPP-4

(N) Nomial, (R) Reactivity, (1) Instrument, (C) Component. (M) Major ,

l Developed By: Approved By:

Examiner: Chief Examiner:

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SCENARIO DESCRIPTION (DSS-005)

After shift turnover and allowing the crew to walkdown the board, a pressurizer pressure control channel PT-445 will fail high. After the plant has been stabilized and appropriate procedure actions have been completed, a 100 gpm tube leak will occur on the "B" SG requiring a plant shutdown due to excessive leakage. A condenser vacuum leak will escalate into the need to trip the turbine due to a loss of l condenser vacuum. The crew will follow Path-1 to EPP-4. While !

performing EPP-004, a Startup Transformer relay malfunction will j occur. Following plant stabilization from the transformer failure, a 1 S/G safety valve will open on "B" SG, causing a safety injection l which will require re-entering Path-1 with a subsequent transition to l Path-2. Safety Injection pumps, MDAFW pumps, and the Steam i Driven AFW pump valves will fail to start /open automatically, !

requiring the operator's action. The final plant conditions will be a faulted / ruptured steam generator. The scenario may be terminated at l the evaluators discretion following transition to EPP-17. l i

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ES-301 SCENARIO EVENTS ES-301-3 SIMULATION FACILITY: II.B. Robinson SCENARIO NO. DSS-008 EXAMINERS: APPLICANTS :

INITIAL CONDITIONS: The Unit is at 100% nower. The following equinment is out of service: "B" EDG (OOS for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months /back in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). "B" MDAFW Pumn (OOS for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months /bumn motor grounded). "B" Serv ice Water booster numn SWBP (OOS for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . motor grounds)"B" S/G PORV has a small gasket leak . "A" S/G tube leakage is 0.1 gpm.

TURNOVER: Commence a normal niant shutdown to renair"B" S/G PORV. Boron Concentration 1017 nom. CBD at 218 steps.

EVENT MALF. EVENT EVENT NO. TYPE DESCRIPTION MFI RPS01 A R Failure of both reactor trip breakers to open.

MFI RPS01B RFI SGN023, C Failure of all MSIV's to auto close 024 and 025 MFI MSS 03C C Failure of the "C" MSIV to manually close i

1 N, R Plant Shutdown l

2 MF1 NIS12A 1, R N-41 Fails low (control power fuse) 3 ORP XN36105 C Override of a seismic alarm to cause an earthquake 1

4 MFI CFWO29 C 2000 gpm leak on the bottom of the CST due to the earthquake.

5 MFI TURB18 C EH pump common suction line leak 6 MF1 MSS 09 M,R Steam break on the 72" header (common steam line)

, with an ATWS.

(N) Normal, (R) Reactivity, (I) Instrument, (C) Component, (M) Major Developed By: Approved By:

Examiner: ChiefExaminer:

SCENARIO DESCRIPTION (DSS-008)

After shift turnover and allowing the crew to walkdown the board, the i crew will initiate a normal plant shutdown due to the S/G PORV.

Following initiation of the plant shutdown, a blown fuse in the control circuit of PR NI-41A will initiate a Turbine Runback. When the plant is stabilized, a seismic event will occur. The seismic event causes an unisolable leak at the bottom of the CST that will be found

! by the makeup water treatment AO. Use of AOP-021 (Seismic j Disturbances) and Technical Specifications should result in a plant

shutdown being directed. During the subsequent plant shutdown an

EH leak will result in a Turbine / Reactor Trip signal with a failure of l the reactor to trip (ATWS). A Steam Break on the 72" header will j develop immediately after the turbine / reactor trip and will be

'; compounded by an automatic close failure of all MSIV's and a stuck open MSIV on "C" MS line. The scenario should progress through EPP-11 (Faulted S/G Isolation); The scenario may be terminated at j the evaluators discretion following transition to EPP-7.

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ES-301 SCENARIO EVENTS ES-301-3 SIMULATION FACILITY: SCENARIO NO: . DSS-009 EXAMINERS: APPLICANTS :

l INITIAL CONDITIONS: The Unit is at 100% nower. The followine cauinment is out of service: HVH-1 out j founc*or renlacement. will be back this shift. "A" EDG out for covernor renair (OOS for 2 davs/back ;

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TURNOVER: You have been instructed to maintain current plant conditions. Boron concentration 1011 j nom. CBD at 218 stens. CV nressure reliefin orocress IAW OP-921. section 6.l.

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EVENT MALF. EVENT EVENT l NO. NO. TYPE DESCRIPTION C Prevents auto closure of V12-10 and V12-11 on the R11/R12 alarm.

CORDS PI:953 C Auto spray actuation failure.

CORDS PI:955 RFI RIIR009 C RIIR-764 out of position closed. l SHUTRHR764 ;

MFI CFW0lC C SD AFW Pump trips on auto start.

l 1 ORP AA085A I PC-444J partial failure causes spray valves to open.

PC444J 60%

2 MFI RCS013B C RCP "B" #1 Seal failure 3 MF1 RCS016B C RCP "B" high vibrations 4 MFI RCS02B C, R RCP "B" trips on overcurrent prior to operator action to trip the RCP.

5 MFI RCS09A C 300 gpm RCS leak ramped over 480 seconds 6 MFI RCS01 A M Large Break LOCA

(N) Normal, (R) Reactivity, (I) Instrument, (C) Component, (M) Major Developed By: Approved By:

Examiner: ChiefExaminer:

SCENARIO DESCRIPTION (DSS-009)

After shift turnover and allowing the crew to walkdown the board, the first event will be a failure of PC-444J which causes a PZR spray valve to open slowly and continuously until the controller is shifted to Manual. PZR pressure will decrease due to the excess spray flow requiring prompt operator action. When the plant has been stabilized, "B" Reactor Coolant Pump will develop high vibrations. These vibrations will cause seal leakoff flows and pump bearing temperatures to increase, indicating a severe problem with the RCP.

The RCP vibrations will increase as the crew attempts to decrease power to remove the pump from service. As power is decreased, RCS leakage will increase until the RCP shaft binds resulting in an overcurrent trip of the pump and subsequent Loss of Flow trip followed by a LBLOCA in the affected loop. The CV spray will fail to operate automatically requiring operator identification and manual actuation. RHR flow does not occur during large break LOCA due to valve RHR-764 being shut. The operating crew will investigate and have the valve re-opened. The LOCA will require entry into PATH-1 and eventually transition to FRP-P.1. The scenario may be terminated at the discretion of the evaluators any time after FRP-P.1 has been implemented.

I ES-301 SCENARIO EVENTS ES-301-3 i SIMULATION FACILITY: II. B. Robinson SCENARIO NO: . DSS-038 EXAMINERS:_ APPLICANTS :

INITIAL CONDITIONS: The Unit is at 20% nower. The followine eauioment is out of service: HVH-1 out for motor renlacement. will be back this shift. "B" EDG out for high chromates in cooline water. will return this shift MOV-350 out for limit switch renair. back in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Pressurizer level channel 461 out due to being out of tolerance.

TURNOVER: You have been instructed to increase nower to 50%. Boron concentration 1017 onm.

CBD at XXX stens.

EVENT MALF. EVENT EVENT NO. NO. TYPE DESCRIPTION RFI SWS56 C Both Sevice Water Booster pumps fail to auto start.

RFI SWS57 RFI RIIR013 C Both RIIR pumps fail to auto start.

RFI RIIR014 1 CORD I Failure high of PT-484, S/G "B" steam pressure.

PT:484 2 CORD I Failure low of LT-459, Pressurizer level channel.

LT:459 3 MFI CVC01.A C Fa!!ure of LCV-460A and 460B to reopen.

MFI CVC01B 4 N Reduce power to take unit offline due to inability to take channel LT-459 OOS without tripping unit.

5 MFI PRSO4C M SBLOCA due to failure of Pressurizer relief valve causing steam space LOCA.

(N) Normal, (R) Reactivity, (I) Instrument, (C) Component, (M) Major Developed By: Approved By:

Examiner: Chief Examiner:

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l SCENARIO DESCRIPTION (DSS-038) l l

i After shift turnover and allowing the crew to walkdown the board, the l

. initiating event will be a failure of steam pressure transmitter PT-484 I high on "B" S/G. This will require manual feed control of the "B" S/G )

main FRV to restore S/G level. After the plant has been stabalized and the appropriate actions taken to remove the channel from service, a i l failure of pressurizer level channel LT-459 low will occur. On l restoration of the normal letdown flowpath, the crew will not be able to open LCV-460A and 460B which will require the crew to initiate excess letdown to maintain pressurizer level. The crew will not be able to remove the instrument from service due to LT-461 being out of service. This will require a plant shutdown in accordance with Technical Specifications. After excess letdown has been established and a power reduction has been initiated, a pressurizer relief valve will fail open resulting in a steam space SBLOCA. The crew should trip the reactor and initiate safety injection and carry out the actions of PATH-1. The Service Water Booster pumps and RHR pumps will fail to automatically start and will require manual operator action to start. The exercise may be terminated at the discretion of the evaluators after entry into EPP-8.

ES-301 INDIVIDUAL WALK-THROUGH TEST OUTLINE FORM ES-301-2 JPM Simulator / Control Room /In-Plant Set # 1 EXAMINATION LEVEL (Circle One): RO / SRO (1) / SRO (U)

FACILITY: H. B. ROBINSON WEEK OF EXAMINATION:

EXAMINER'S NAME (PRINT):

SYSTEM /JPM SAFETY FUNCTION PLANNED FOLLOWUP QUESTIONS K/A/G // IMPORTANCE // DESCRIPTION

1. CR-033, Perform Boration of the 1 004000 Gen.7 // 3.0/3.3 // Loss of Aux Panel DC RCS -1AW GP-007 SIM 004000 K1.16 // 3.3/3.5 // Clearance Reg'd on "A" BAST

2. CR-005, Re-establish Letdown Flow II 004010.A2.04 // 3.6/4.2 // PCV-145 fails /etTeets on CVCS IAW OP-301 SIM 073000.K3.01// 3.6/4.2 // R-17 alarm while est. LTDN 3 CR-035, PZR Pressure Control III 010000.K1.01// 3.9/4.1// Basis of PZR Low Pressure Trip Malfunction IAW AOP-019 (PSA) SIM (Alt. Path) 010000.K1.06 // 2.9/3.1 // Why can't use Aux Spray if normal LTDN isolated at power

4. CR-030, Loss of ResidualIIcat IV 005000.A 1.06 // 2.7/3.1 // In CSD, reduced inv., how long 1 Removal (Shutdown Cooling)IAW SIM until core uncovered AOP-020 -'

005000.A2.04 // 2.9/2.9 // Close iICV-758, Open FCV-605, how efTect RilR Flow

5. CR-009, Remove Power Range IX 015000 K4.09 // 2.8/3.1//N-44 OOS, What etTect if pull Channel N-44 from service IAW MCR instrument power fuses OWP-011 015000.K4.03 // 3.1/3.3 // Purpose of power mismatch ckt.

6. CR-095, Drain the PRT IAW OP- VI 007000.A4.02 // 2.2/2.2 // Loss of Inst. Air while draining

??? (NEW) MCR 007000.K 1.01 // 2.9/3.1 // Loss of DC bus "B" while draining

7. CR- 066, Respond to a Loss of CCW X 008000.A1.01 // 2.8/2.9 // Ilow verify CCW flow to RCP to the RCP motor coolers IAW MCR AOP-014 008000. A4.01 // 3.3/3.1 // FC V-626 open, Ilow verify flow to RCP thermal barrier fl. IP-055, Ali Fn Deepwell to AFW V 061000.A3.01// 4.1/4.2 // AFW pump Starting Duty req'a 4

' IAW OP-402 (NEW)(PSA) PLANT 0616en 05 // 3.3/4.0 // TS for AFW

9. IP-050, Energize Pressurizer IIcaters VII 010000.K4.02 // 3.0/3.4 // litt response if PZR level dec.

from the Emergency Busses IAW PLANT below 14.4%

EPP-21 (Al. Path)(RCA)(S/D) 010000.K1.03 // 3.6/3.7 // What is purpose of the PZR llTR BKR ann switch

10. IP-053, Turbine Building Operator Vill 061000.K4.01// 3.9/4.2 //fS for CST PLANT I Actions IAW AOP-004 (S/D) 061000 K4.02 // 4.5/4.6 // Auto Start Signals fbr MD AFW l

DEVELOPED BY: APPROVED BY:

EXAMINER: ClIIEF EXAMINER:

ES-301 Administrative Topics Outline ES-301-1 Set #1 Examination Level (Circle One): SRO Facility: Week of Examination:

Examiner's Name (print):

Administrative Describe method of evaluation:

Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Shin Stafling What do the oncoming and offgoing CRSSs discuss What review actions are completed by the oncoming ROs Temporary How would the Shin Superintendent know that a Temporary q Mods Modification existed that prevented heating up above 200 degrees while preparing for a plant heatup?

What is the duration of the temporary mods A.2 Plant JPM-CR-033 radngs JPM-IP-057 A.3 Use of Covered in RCA entry Radiation Instrument Covered in RCA entry 3

! A.4 Lines of Covered in Simulator Scenario i j Authority Covered in Simulator Scenario l Developed By: Approved By:

! i Examiner: ChiefExaminer: l l

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ES-301 INDIVIDUAL WALK-THROUGH TEST OUTLINE FORM ES-301-2 JPM Simulator / Control Room /In-Plant Set # 2 EXAMINATION LEVEL (Circle One): RO / SRO (I) / SRO (U)

FACILITY: H. B. ROBINSON WEEK OF EXAMINATION:

EXAMINER'S NAME (PRINT):

SYSTEM /JPM SAFETY FUNCTION PLANNED FOLLOWUP QUESTIONS K/A/G // IMPORTANCE // DESCRIPTION

1. CR-044, Withdrawing Control Rod 1 001000.K1.04 // 3.2/3.4// Withdrawing CR S steps Shutdown Bank A LAW GP-005 SIM 001000.K5.29 // 3.7/3.9 // Maintain Temp. during rod withdrawal

2. CR-074, Re-establish Letdown Flow 11 004010 A4.02 // 3.1/3.6 // Charging System Response IAW OP-301 (Alt. Path) SIM 004000.A2.07 // 3.4/3.7// Des.11U after a Phase A isol.

3 CR-082, Depressurize RCS using III 000038.EK3.06 // 4.2/4.5 // Options for SGTR Cooldown PORV's IAW Path-2 (Alt. Path) SIM (PSA)(S/D) 000038.EK3.01// 4.1/4.3 // Basis for SGTR Cooldown

4. CR-041, Respond to a RCP Seal IV 004000.Kl .04// 3.4/3.8 // Flowpaths for Seal Injection Malfunction IAW AOP-018 SIM 004000.A2.05 // 4 0/4.3 // Seal Leakoffind. during #2 seal

5. CR-045, Transfer from the Bypass to V 059000.K4.19 // 3.2/3.4 // Auto closure of FRV's the Main Feedwater Regulating MCR Valves IAW GP-005 (low power) 059000.A3.06 // 3.2/3.3 // FWI signal inputs / actions

6. CR-096, Fill the PRT IAW OP- VI 007000.A4.09 // 2.5/2.7 // Inadvertent RCS M/U (NEW) MCR 007000.K4.02 // 2.6/2.9 // Primary wtr pump starting duty

7. CR- 010, Place a Reactor Protection IX 012000.K4.01// 3.7/4.0 // Prot & Cont func of sw positions Channelin the Tripped Conditie - MCR IAW OWP-030 000027.EA 1.01 // 4.0/3.9 // Lvl chan. Fail during NC Cooldown

8. IP.051 Peifonn Electrical Operator VII 067 GEN // 3.8/4.0 // Criteria for entering DSP's Actions of DSP-002 (Turbine PLANT Building)(Alt. Path)(low power) 000067.EA2.16 // 3.3/4.0 // Est.\ Control AFW flow to S/G l

9. IP , Perform Subsequent Actions X 076 GEN 9 // 3.0/3.0 // Restoring SW Pressure of AOP-022 in the Auxiliary PLANT Building (PSA)(RCA)(NEW) 076000.K1.16 // 3.6/3.8 // SWBP oper. W/ north hdr isol.

10. IP-033, Establish Emergency XI 008000 K4.01 // 3.1/3.3 // operation of CCW Pumps Cooling to the Spent Fuel Pit Heat PLANT Exchanger IAW OP-306 (RCA) 033 Gen 05 // 2.4/3.2 // TS associated w/SFP DEVELOPED BY: APPROVED BY:

EXAMINER: _

ClIIEF EXAMINER:

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l ES-301 Administrative Topics Outline ES-301-1 i Set #2 l Examination Level (Circle One): SRO Facility: Week of Examination:

Examiner's Name (print):

Administrative Describe method of evaluation:

Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Shin Staffmg What are the requirements for the STA if they are not in the control room What are the requirements if the shin complement is less than the minimum shin complement Short Term Explain purpose /use of the " Operation's Directive Book" Information How long do night orders remain in effect A.2 Plant IPM-CR-041 l JPM-CR-045 A.3 Use of Covered in RCA entry Radiation Instruments Covered in RCA entry A.4 Lines of Covered in Simulator Scenarios Authority Covered in Simulator Scenarios Developed By: Approved By: __

l Examiner: ChiefExaminer:

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ES-301 INDIVIDUAL WALK-TIIROUGH TEST OUTLINE FORM ES-301-2 JPM Simulator / Control Room /In-Plant Set # 3 EXAMINATION LEVEL (Circle One): RO / SRO (I) / SRO (U)

FACILITY: H. B. ROBINSON WEEK OF EXAMINATION:

EXAMINER'S NAME (PRINT):

SYSTEM /JPM SAFETY FUNCTION PLANNED FOLLOWUP QUESTIONS K/A/S // IMPORTANCE // DESCRIPTION

1. CR-047, Perform 13oron 1 004000. A2.01// 4.2/4.3 // What prot. features prevent Concentration Dilution of the RCS SIM overpower condition & core damage of continuous dilution IAW OP-301 with no operator action 004000.K1.06 // 3.1/3.1// LT-115 fails, afTect on AutoM/U

2. CR-086, Depressurize RCS LAW III 000038 EK3.01// 4.1/4.3 // Why must cooldown RCS prior Path-2, PORV Failure (Alt. Path) SIM to depressurizing during a SOTR (PSA)(S/D) 010000.A4.03 // 4.0/3.8 //What are the press. interlocks assoc. with the PORV's 3 CR-057, Respond to a Loss of RCP IV 003000. A2.01// 3.5/3.9 // Des. all norm flowpaths thru SealInjection IAW AOP-018 SIM RCP 003000, A3.01 // 3.3/3.2 // Auto system response

4. CR-023 Transfer Auxiliary Loads VII 062000.A4.07 // 3.7/4.2 // Interlocks for 52nd12 from Auxiliary to Startup SIM Transformer OP-603 062000.K1.04 // 3.7/4.2 // Nonnal at power 4KV lineup

5. CR-062, Remove Source Range Vill 015000.K6.02 // 2.6/2.9 // During S/D lose IR compens.

Channel N-3 I from Service IAW MCR Volts to N-35 OWP-011 015000 K4.01// 3.1/3.3 // P-6 Interlocks for SR

6. CR-098, Initiate Containment Spray VI 026 Gen 05 // 4.2/4.2//1 CSP OOS, What equip. by TS (NEW)(S/D) MCR 026000 K1.01// 4.2/4.2 // DitT. In actions if Man. or AUTO

7. CR- 097 Establish Excess Letdown 11 004010.A2.05 // 4.1/4.3 //1 low Phase A efTect ability to (Alt. Path), (NEW) MCR maintain Ex. LTDN temp.

004000 K1.01// 3.6/4.0 // Expectd alarms w/No Ltdn avail.

8. IP-002, Shift Auxiliary Feedwater V 076000 K4.01 // 2.5/2.9 // Turbine 13uilding Isolation Pump Suction to Service Water IAW PLANT OP-402 (PSA)(RCA) 022000 Kl.01// 3.5/3.7 //13as. for SW Booster Pmp Oper.

9. IP-052, Perform Subsequent Actions IX 000067.EA2.04 // 3.1/4.3 // local control / operation of SW of AOP-022 in the Auxiliary PLANT pumps Building (Alt. Path)(RCA) 076000 K3.07 // 3.7/3.9 // Conditions for AUTO isolation

10. IP-035, Manually Actuate Ilalon X 086000 K4.05 // 3.0/3.4// Des. basis of the 1Ialon 130I sys.

Suppression System for E-1/2 Room PLANT IAW OP-804 & LocalInst. 086000 A3.02 // 2.9/3.3 // What signals req'd ihr Auto act.

DEVELOPED BY: APPROVED BY:

EXAMINER: ClIIEF EXAMINER:

ES-301 Administrative Topics Outline ES-301-1 Set #3 Examination Level (Circle One): RO I Facility: Week of Examination:

Examiner's Name (punt):

Administrative Describe method of evaluation:

Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Shift Stafling What is the minimum shift complement !

Describe the responsibilities of the CRSS Key Control What are the required actions if a controlled key is lost What is required to revise the key inventory A.2 Plant JPM-CR-023 l Drawings JPM-CR-062 A.3 Use of Covered in RCA ently Radiation Instruments Covered in RCA ently A.4 Emergency Describe purpose / activities of OSC Facilities Location of alternate TSC/ EOF i

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Developed By: Approved By:

Examiner: ChiefExaminer:

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Knowledge and Abilities Record Form PLANT-WIDE GENERIC RESPONSIBILITIES PWR - Senior Reactor Operator Check if 194000 Statement Rating Included K/A #

/ Kl.01 Knowledge of how to conduct and verify valve 3.7 lineups

/ Kl.02 Knowledge of tagging and clearance procedures 4.1

/ Kl.03 Knowledge of 10CFR20 and related facility 3.4 radiation control requirements

/ Kl.04 Knowledge of facility ALARA prs - 3.5 Kl.05 Knowledge of facility requirements for controlling 3.4*

access to vital / control areas K1.06 Knowledge of safety procedures related to rotating 3.4

equipment

/ Kl.07 Knowledge of safety procedt related to electrical 3.7*

equipment Kl.08 Knowledge of safety procedures related to high 3.4 temperature Kl.09 Knowledge of safety procedures related to high 3.4 pressure Kl.10 Knowledge of safety procedures related to caustic 3.3 solutions Kl.11 Knowledge of safety procedures related to chlorine 3.5

Kl.12 Knowledge of safety procedures related to noise 2.9 Kl.13 Knowledge of safety procedures related to oxygen- 3.6 deficient environment

/ Kl.14 Knowledge of safety procedures related to confined 3.6 spaces Kl.15 Knowledge of safety procedures related to 3.8*

hydrogen

/ K1.16 Knowledge of facility protection requirements, 4.2*

including fire brigade and portable fire-fighting equipment usage Kl.17 Knowledge of the equipment rotation schedu'es and 2.5 the reasoning behind the rotation procedure

/ Al.01 Ability to obtain and verify control procedure copy 3.4 A1.02 Ability to execute procedural steps 3.9

/ A1.03 Ability to locate and use procedures and station 3.4 directives related to shift staffing and activities Al.04 Ability to operate the plant phone, paging system, 3.2 and two-way radio Al.05 Ability to make accurate, clear, and concise verbal 3.8 reports Al.06 Ability to maintain accurate, clear and concise logs, 3.4 records, status boards and reports A1.07 Ability to obtain and interpret station electrical and 3.2 mechanical drawings A1.08 Ability to obtain and interpret station reference 3.1 material such as graphs, nomographs, and tables which contain system performance data

/ A1.09 Ability to coordinate personnel activities inside the 3.9*

control room

/ A1.10 Ability to coordinate personnel activities outside the 3.9

control room

/ Al.11 Ability to direct personnel activities inside the 4. l

control room

/ A1.12 Ability to direct personnel activities outside the 4. l

control room

/ A1.13 Ability to locate control room switches, controls, 4.1 and indications, and to determine that they are correctly reflecting the desired plant lineup A1.14 Ability to maintain primary and secondary plant 2.9 chemistry within allowable limits '

Al.15 Ability to use plant computer to obtain and evaluate 3.4 parametric information on system and component status

/ A1.16 Ability to take actions called for in the Facility 4.4

Emergency Plan, including (if required) supponing or acting as the Emergency Coordinator

Knowledge and Abilities Record Form PLANT SYSTEMS PWR - Senior Reactor Operator - 40%

Plant Specific Priorities System # K/A # K/A Topic Rating Group I Plant Systems - 19%

001 Control Rod Drive System 025 Ice Condenser System 003 Reactor Coolant Pump System 026 Containmm Spray System 004 Chemical and Volume Control System 056 Condensate System 013 Engineered Safety Features Actuation System 059 Main Feedwater System 014 Rod Position Indication System 061 Auxiliary / Emergency Feedwater System 015 Nuclear instrumentation System 063 DC Electrical Distribution 017 In-Core Temperature Monitor System 068 Liquid Radwaste System 022 Containment Cooling System 071 Waste Gas Disposal system 072 Area Radiation Monitoring System <

l System # K/A # K/A Topic Rating 014000 Kl.01 Knowledge of the physical connections and/or cause- 3.2 */3.6 effect relationships between the RPIS and the CRDs ,

system.

022000 K2.0i Knowledge of the power supplies to the following: 3.0/3.1 Containment Cooling Fans 015000 K3.01 Knowledge of the effect that a loss or malfunction of the 3.9/4.3 NIS will have on RPS.

061000 K4.02 Knowledge of AFW design features and/or interlocks 4.5/4.6 which provide for the following: AFW automatic start upon loss of MFW pump, S/G level, blackout, or SI.

001000 A3.01 Ability to monitor automatic operation of the CRDS, 4.1/4.0 including: Reactor Power.

003000 K6.02 Knowledge of the effect of a loss or malfunction on the 2.7/3.1 following will have on the RCP's: RCP seals and seal water supply.

022000 A1.01 Ability to predict and/or monitor changes in parameters 3.6/3.7 (to prevent exceeding design limits) assocalated with operating the CCS controls including: containment temperature.

I 004000 A2.07 Ability to predict the impacts of the following 3.4/3.7 malfunctions or operations on the CVCS, and based ability on those predictions, use procedures to correct, l control, or mitigate the consequences of those malfunctions or operations, isolation of l letdown / makeup. l 013000 A3.02 Ability to monitor automatic operation of the ESFAS 4.1/4.2 including: Operation of actuated equipment 1 063000 A4.02 Ability to manually operate and/or monitor in the control 2.8/2.9 room: Battery voltage indicator. l 004000 G0.07 Knowledge of purpose and function of major system 3.3/3.3 )

components and controls. '

1 061000 Kl.07 Knowledge of the physical connections and/or cause- 3.6/3.8 I effect relationships between the AFW and the )

Emergency Water source. j 022000 K3.01 Knowledge of the effect that a loss or malfunction of the 2.9/3.2 )

CCS will have on Containment equipment subject to i damage by high or low temperature, humidity, and pressure.

013000 A2.02 Ability to predict the impacts of the following 4.3/4.5 malfunctions or operations on the ESFAS, and based ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, excess steam demand.

015000 A3.02 Ability to monitor automatic operation of the NIS, 3.7/3.9 including annunciator and alarm signals.

001000 K4.03 Knowledge of CRDS design features and/or interlocks 3.5/3.8 which proside for the following: Rod Control Logic.

015020 K5.08 Knowledge of the following theoretical concepts as they 2.9/3.4 apply to the NIS, enthalpy.

017020 K6.01 Knowledge of the effect of a loss or malfunction on the 2.7/3.0 following will have on the NNIS: Sensors and detectors 003000 Al.09 Ability to predict and/or monitor changes in parameters 2.8/2.8 (to prevent exceeding design limits) assocaiated with operating the RCP controls including: seal flow and D/P

Group Il Plant Systems - 17%

002 Reactor Coolant System 035 Steam Generator System 006 Emergency Core Cooling System 039 Main and Reheat Steam System 010 Pressurizer Pressure Control System 055 Condenser Air Removal System 011 Pressurizer Level Control System 062 AC Electrical Distribution 012 Reactor Protection System 064 Emergency Diesel Generator Sys 016 Non-Nuclear Instrumentation System 073 Process Radiation Monitoring Sys 027 Containment Iodine Removal System 075 Circulating Water System 028 Hydrogen Recombiner and Purge 079 Station Air System Control System 086 Fire Protection System 029 Containment Purge System 103 Containment System 033 Spent Fuel Pool Cooling System 034 Fuel Handling Equipment System System # K/A # K/A Topic Rating 064000 Kl.01 Knowledge of the physical connections and/or cause- 4.1/4.4 effect relationships between the EDG and the AC Distribution syst m.

062000 K2.01 Knowledge of the power supplies to the following: 3.3/3.4 Major system Loads.

006000 K3.02 Knowledge of the effect that a loss or malfunction of the 4.3/4.4 ECCS will have on the following: Fuel.

016000 K4.03 Knowledge of NNIS design features and/or interlocks 2.8/2.9 l which provide for the following: Input to control systems. l 006000 K5.06 Knowledge of the operational implications of the 3.5/3.9 following concepts as they apply to the ECCS: l Relationship between ECCS flow and RCS pressure. j 012000 K6.03 Knowledge of the effect of a loss or malfunction on the 3.1/3.5 following will have on the RPS: Trip Logic Circuits. ,

039000 Al.05 Ability to predict and/or monitor changes in parameters 3.2/3.3 (to prevent exceeding design limits) assocaiated with operating the MRSS controls including: RCS Tavg 011000 A2.10 Ability to predict the impacts of the following 3.4/3.6 i malfunctions or operations on the PZR LCS, and based ability on those predictions, use procedures to correct, control, or mitigate the consequences of those 1 malfunctions or operations: Failure of a PZR level instrument high i 029000 A3.01 Ability to monitor automatic operation of the 3.8/4.0 Containment Purge system, including: CPS isolation. :

010000 A4.03 Ability to manually operate and/or monitor in the control 4.0/3.8 room: PORV and Block valves.

033000' G0.06 Knowledge of the bases in Technical Specifications for 2.7/3.1 limiting conditions for operations and safety limits.

035000 Kl.01 Knowledge of the physical connections and/or cause- 4.2/4.5 ,

effect relationships between the S/Gs and the l MFW/AFW systems.

012000 K2.01 Knowledge of the power supplies to the following: RPS 3.3/3.7 channels, components, and interconnections.

062000 K3.01 Knowledge of the effect that a loss or malfunction of the 3.5/3.9 l AC distribution system will have on the following: Major system loads.

064000 K4.02 Knowledge of EDG design features and/or interlocks 3.9/4.2 I which provide for the following: Trips for EDG while operating (normal or emergency).

073000 K5.01 Knowledge of the operational implications of the 2.5/3.0 following concepts as they apply to the PRM: Radiation theory, including sources, types, units, and effects.

010000 K6.03 Knowledge of the effect of a loss or malfunction on the 3.2/3.6 following will have on the PZR PCS: PZR sprays and heaters.

Group III Plant Systems - 4%

005- Resdiual Heat Removal System 041 Steam Dump System /furbine 007 Pressurizer Relief Tank / Quench Tank Bypass Control System 045 Main Turbine Generator System !

008 Component Cooling Water System 076 Service Water System l 078 Instrument Air System l System # K/A # K/A Topic Rating 078000 Kl.04 Knowledge of the physical connections and/or cause- 2.6/2.9 effect relationships between the IAs and the Cooling water to compressor systems.

078000 K3.02 Knowledge of the effect that a loss or malfunction of the 3.4/3.6 IA system will have on the following: Systems having pneumatic valves and controls.

005000 K4.07 Knowledge of RHRs design features and/or interlocks 3.2/3.5 j which provide for the following: Overpressure ;

Mitigation system. '

008000 K4.01 Knowledge of CCWS design features and/or interlocks 3.1/3.3 which provide for the following: Auto start of Stby pmp.

j

Knowledge and Abilities Record Form EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator - 43%

Plant Specific Priorties E/A # K/A # K/A Topic Rating l Group I Emergency and Abnormal Plant evolutions - 24%

000001 Continuous Rod Withdrawl 000055 Loss of Offsite and Onsite Power 000003 Dropped Control Rod 000057 Loss of Vital AC Electrical Instrument 000005 Inoperable / Stuck Control Rod Bus 000011 Large Break LOCA 000059 Accidental Liquid Radioactive Waste 000015 RCP Motor Malfunction Release 000024 Emergency Boration 000067 Plant Fire On Site 000026 Loss of Component Cooling 000068 Control Room Evacuation Water 000069 Loss of Containment Integrity 000029 Anticipated Transient Without 000074 Inadequate Core Cooling Scram 000076 High Reactor Coolant Activity 000040 Steam Line Rupture 000051 Loss of Condenser Vacuum E/A # K/A # K/A Topic Rating 000011 EKl.01 Knowledge of the ormational implications of the 4.1/4.4 following concepts as they apply to Continuous Rod Withdrawl: Prompt criticality 000068 K2.07 Knowledge of the interrelations between the Control 3.3/3.4 Room Evacuation and the following components: EDG 000005 EK3.06 Knowledge of the bases or reasons for the following: 3.9/4.2 Actions contained in the EOP for inoperable / stuck control rod.

000024 A1.05 Ability to operate and monitor the following: Performance 3.1/3.2 of the letdown system during emergency boration.

000051 EA2.02 Ability to determine or interpret: Conditions requiring 3.9/4.1 reactor and/or turbine trip.

l 000055 G0.12 Ability to utilize symptom based procedures. 3.9/4.0 ;

1 000040 EKl .01 Knowledge of the following theoretical concepts as they 4.1/4.4 apply to the steam line mpture emergency task:

Consequences of PTS. i l

l l

1

1 000068 EK3.18 Knowledge of the bases or reasons for the following: 4.2/4.5 Actions contained in the EOP for control room evacuation emergency task.

000015 EA 1.06 Aoility to operate and monitor the following: CCWS 3.1/2.9 000026 EA2.06 Ability to determine or interpret: The length of time after 2.8/3.1 1 the loss of CCW flow to a component before that i

component may be damaged. -

000051 G0.11 Ability to recognize abnormal indications for system 2.7/2.9 )

l operating parameters which are entry level conditions for emergency and abnormal operating procedures.

000029 K1.03 Knowledge of the following theroetical concepts as they 3.6/3.8 apply to the ATWS emergency task: Effects of boron on reactivity.

000076 EK3.06 Knowledge of the bases or reasons for the following: 3.2/3.8 Actions contained in the EOP for high reactor coolant activity.

000015 Al.22 Ability to operate and monitor the following: RCP Seal 4.0/4.2 failure / malfunction 000026 A2.01 Ability to determine or interpret: Location of a leak in the 2.9/3.5 CCWS.

000067 G0.01 Knowledge of system status criteria which require the 3.6/4 0 notification of plant personnel.

000055 EK3.01 Knowledge of the bases or reasons for the following: 2.7/3.4 Length of time for which the battery capacity is designed.

000069 EA'. 01 Ability to determine or interpret: Loss of containment 3.7/4.3 integrity.

000069 A2.02 Ability to determine or interpret: Verification of automatic 3.9/4.4 and manual means of restoring integrity.

000040 G0.11 Ability to recognize abnormal indications for system 4.1/4.3 operating parameters which are entry level conditions for emergency and abnormal operating procedures.

000011 K3.15 Knowledge of the bases or reasons for the following: 4.3/4.4 Criteria for shifting to recirculation mode.

000057 A 1.16 Ability to operate and monitor the following: Manual 3.5/3.5 control of components for which automatic control is lost.

1 l

000057 EA2.19 Ability to determine or interpret: The plant automatic 4.0/4.3 actions that will occur on the loss of a vital AC electrical i instrument bus.

l l

000055 G0. I 1 Ability to recognize abnormal indications for system 4.1/4.1 operating parameters which are entry level conditions for emergency and abnormal operating procedures.

Group 11 Emergency and Abnormal Plant evolutions - 16%

000007 Reactor Trip 000037 Steam Genreator Tube Leak 000008 Pressurizer Vapor Space 000038 Steam Generator Tube Rupture Accident 000054 Loss of Main Feedwater 000009 Small Break LOCA 000058 Loss of DC Power 000022 Loss of Reactor Coolant 000060 Accidental Gaseous-Waste Release Makeup 000061 Area Radiation Monitoring System 000025 Loss of Residual Heat Removal 000065 Loss ofInstrument Air 000027 Pressurizer Pressure Control System Malfunction 000032 Loss of Source Range Nuclear instrumentation 000033 Loss ofIntermediate Range Nuclear Instnimentation E/A # K/A # K/A Topic Rating 000009 EKl.01 Knowledge of the following theoretical concepts as they 4.2/4.7 apply t6 the small break LOCA emergency task: Natural circulation and cooling, including reflux boiling.

000058 K3.01 Knowledge of the bases or reasons for the following: Use 3.4/3.7 of DC control power by EDG's.

000025 EAl.02 Ability to operate and monitor the following: RCS/RHR 3.8/3.9 cooldown rate.

000058 A2.03 Ability to determine or interpret: DC loads lost: impact on 3.5/3.9 ability to operate and monitor plant systems.

000054 G0.11 Ability to recognize abnormal indications for system 3.4/3.3 operating parameters which are entry level conditions for emergency and abnormal operating procedures.

000007 EAl.03 Ability to operate and monitor the following: RCS 4.2/4.1 pressure and temperature.

000022 EK3.02 Knowledge of the bases or reasons for the following: 3.5/3.8 Actions contained in SOP's and EOP's for RCP's, loss of makeup, loss of charging, and abnormal charging.

000022 A1.08 Ability to operate and monitor the folk wing: VCT level. 3.4/3.3 000061 EA2.05 Ability to determine or interpret: Need for area evacuation 3.5/4.2 000032 G0.10 Ability to perform without reference to procedures those 2.9/3.1 actions that require immediate operation of system components or controls.

1 000037 K3.07 Knowledge of the bascs or reasons for the following: 4.2/4.4 Actions contained in EOP for S/G tube leak.

000038 Al.01 Ability to operate and monitor the following S/G levels, 4.5/4.4 for abnormalincrease in any S/G.

000032 EA2.06 Ability to determine or interpret: Confirmation of reactor 3.9/4.1 trip.

000007 G0.12 Ability to utilize symptom based procedures. 3.8/3.9 000065 EK3.03 Knowledge of the bases or reasons for the following: 2.9/3.4 Knowing effects on pint operation ofisolating certain equipment from instrument air.

000008 Al.08 Ability to operate and monitor the following: PRT level, 3.8/3.8 pressure, and temperature.

Group Ill Emergency and Abnormal Plant Evolutions - 3%

000028 Pressurizer Level Malfunction 000056 Loss of Offsite Power 000036 Fuel Handling Incident 000028 EA2.08 Ability to determine or interpret: PZR level as a function 3.1/3.5 of power level.

000036 G0.07 Ability to explain and apply all system limits and 3.2/3.5 precautions.

000028 A2.02 Ability to determine or interpret: PZR level as a function 3.4/3.8 of power level or Tave, including interpretation of malfunction.

l t

i Knowledge and Abilities Record Form '

PLANT-WIDE GENERIC RESPONSIBILITIES PWR - Reactor Operator i Check if 294000 Statement Rating :

Included K/A # 1

/ Kl.01 Knowledge of how to conduct and verify valve 3.6 lineups 1

/ K1.02 Knowledge of tagging and clearance procedures 3.7 I

/ Kl.03 Knowledge of 10CFR20 and related facility 2.8 !

radiation control requirements

/ Kl.04 Knowledge of faility ALARA program 3.3 Kl.05 Knowledge of facility requirements for controlling 3.1 access to vital / control areas K1.06 Knowledge of safety procedures related to rotating 3.4 equipment

/ Kl.07 Knowledge of safety procedures related to electrical 3.6*

equipment !

Kl.08 Knowledge of safety procedures related to high 3.5*

temperature l

i K1.09 Knowledge of safety procedures related to high 3.4 pressure l Kl.10 Knowledge of safety procedures related to caustic 3.0*

solutions Kl .11 Knowledge of safety procedures related to chlorine 3.4

l Kl.12 Knowledge of safety procedures related to noise 2.6*

Kl.13 Knowledge of safety procedures related to oxygen- 3.3

i deficient environment

/ Kl.14 Knowledge of safety procedures related to confined 3.3 spaces Kl.15 Knowledge of safety procedures related to 3.4 hydrogen

/ Kl.16 Knowledge of facility protection requirements, 3.5 including fire brigade and portable fire-fighting equipment usage Kl.17 Knowledge of the equipment rotation schedules and 2.1 the reasoning behind the rotation procedure

Al.01 Ability to obtain and verify control procedure copy 3.3 Al.02 Ability to execute procedural steps 4. l

A1.03 Ability to locate and use procedures and station 2.5 directives related to shill stafling and activities Al.04 Ability to operate the plant phone, paging system, 3.0 and two-way radio A1.05 Ability to make accurate, clear, and co<icise verbal 3.6 reports Al.06 Ability to maintain accurate, clear and concise logs, 3.4 records, status boards and reports Al.07 Ability to obtain and interpret station electrical and 2.5 mechanical drawings A1.08 Ability to obtain and interpret station reference 2.6 material such as graphs, nomographs, and tables which contain system performance data Al.09 Ability to coordinate personnel activities inside the 2.7 control room

/ Al.10 Ability to coordinate personnel activities outside the 2.9 control room

/ Al.11 Ability to direct personnel activities inside the 2.8 control room

/ A1.12 Ability to direct personnel activities outside the 3.1 control room

/ Al.13 Ability to locate control room switches, controls, 4.3

and indications, and to determine that they are correctly reflecting the desired plant lineup A1.14 Ability to maintain primary and secondary plant 23*

chemistry within allowable limits A1.15 Ability to use plant computer to obtain and evaluate 3.1 parametric information on system and component j status !

A1.16 Ability to take actions called for in the Facility 3.1 Emergency Plan, including (if required) supporting or acting as the Emergency Coordinator l

Knowledge and Abilities Record Form PLANT SYSTEMS PWR - Reactor Operator - 51%

Plant Specific Priorities System # K/A # K/A Topic Rating Group I Plant Systems - 23%

001 Control Rod Drive System 056 Condensate System 003 Reactor Coolant Pump System 059 Main Feedwater System 004 Chemical and Volume Control System 061 Auxiliary / Emergency Feedwater 013 Engineered Safety Features Actuation System System 015 Nuclear instrumentation System 068 Liquid Radwaste System 017 In-Core Temperature Monitor System 071 Waste Gas Disposal System 022 Containment Cooling System System 072 Area Radiation Monitoring 025 Ice Condenser System System System # K/A # K/A Topic Rating 072000 Kl.04 Knowledge of the physical connections and/or cause 3.3/3.5 effect relationships between the ARM system and the following systems: Control room ventilation.

013000 Kl.01 Knowledge of the physical connections and/or cause 4.2/4.4 effect relationships between the ESFAS and the following systems: Initiation signals for ESF logic circuits.

022000 K2.01 Knowledge of the power supplies to the following: 3.0/3.1 Containment Cooling Fans 015000 K3.01 Knowledge of the effect that a loss or malfunction of the 3.9/4.3 NIS will have on RPS.

061000 K4.02 Knowledge of AFW design features and/or interlocks 4.5/4.6 which provide for the following: AFW automatic start upon loss of MFW pump, S/G level, blackout, or SI.

004000 K5.01 Knowledge of the cperational implications of the 2.7/3.3 following concepts as they apply to the CVCS:

importance if oxygen control in the RCS.

003000 K6.02 Knowledge of the effect of a loss or malfunction on the 2.7/3.1 following will have on the RCP's: RCP seals and seal water supply.

022000 Al.01 Ability to predict and/or monitor changes in parameters 3.6/3.7 (to prevent exceeding design limits) assocaiated with operating the CCS controls including: containment temperature.

004000 A2.07 Ability to predict the impacts of the following 3.4/3.7 malfunctions or operations on the CVCS, and based ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, isolation of letdown / makeup.

013000 A3.02 Ability to monitor automatic operation of the ESFAS including: Operation of actuated equipment 004010 A4.02 Ability to manually operate and/or monitor in the control 3.6/3.1 room: Letdown isolation and flow control valves.

004000 G0.07 Knowledge of purpose and function of major system 3.3/3.3 components and controls.

061000 Kl .07 Knowledge of the physical connections and/or cause- 3.6/3.8 effect relationships between the AFW and the Emergency Water source.

022000 K3.01 Knowledge of the effect that a loss or malfunction of the 2.9/3.2 CCS will have on Containment equipment subject to damage by high or low temperature, humidity, and pressure.

013000 A2.02 Ability to predict the impacts of the following 4.3/4.5 malfunctions or operations on the ESFAS, and based ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, excess steam demand.

015000 A3.02 Ability to monitor automatic operation of the NIS, 3.7/3.9 including annunciator and alarm signals.

013000 K4.10 Knowledge of ESFAS design features and/or interloc'ks 3.3/3.7 which provide for the following: Safeguards equipment j control reset. I 015020 K5.08 Knowledge of the following theoretical concepts as they 2.9/3.4 l

apply to the NIS, enthalpy.

017020 K6.01 Knowledge of the effect of a loss or malfunction on the 2.7/3.0 following will have on the NNIS: Sensors and detectors 003000 A1.09 Ability to predict and/or monitor changes in parameters 2.8/2.8 (to prevent exceeding design limits) assocaiated with operating the RCP controls including: seal flow and D/P l

071000 A4.13 Ability to manually operate and/or monitor in the control 3.0/3.1

, room: Recovery from automatic termination of gas i release due to PRM system alarm.

015000 K4.05 Knowledge of NIS design features and/or interlocks 4.3/4.5 which provide for the following: Reactor trip.

001000 K1.03 Knowledge of the physical connections and/or cause- 3.4/3.6 efTect relationships between the CRDS and the CRDM.

Group II Plant Systems - 20%

002 Reactor Coolant System 035 Steam Generator System 006 Emergency Core Cooling System 039 hiain and Reheat Steam System 010 Pressurizer Pressure Control System 055 Condenser Air Removal System 011 Pressurizer Level Control System 062 AC Electrical Distribution 012 Reactor Protection System 063 DC Electrical Distribution 014 Rod Position Indication System 064 Emergency Diesel Generator Sys 016 Non-Nuclear Instrumentation System 073 Process Radiation MonitoringSys

026 Containment Spray System 075 Circulating Water System 029 Containment Purge System 079 Station Air System 033 Spent Fuel Pool Cooling System 086 Fire Protection System System # K/A # K/A Topic Rating 4

064000 Kl.01 Knowledge of the physical connections and/or cause- 4.1/4.4 effect relationships between the EDG and the AC Distribution system.

4

! 062000 K2.01 Knowledge of the power supplies to the following: 3.3/3.4 Major system Loads.

006000 K3.02 Knowledge of the effect that a loss or malfunction of the 4.3/4.4 ECCS will have on the following: Fuel.

016000 K4 03 Knowledge of NNIS design features and/or interlocks 2.8/2.9 which provide for the following: Input to control systems.

006000 K5.06 Knowledge of the operational implications of the 3.5/3.9 following concepts as they apply to the ECCS:

Relationship between ECCS flow and RCS pressure.

012000 K6.03 Knowledge of the effect of a loss or malfunction on the 3.1/3.5

{ following will have on the RPS: Trip Logic Circuits.

, 039000 A1.05 Ability to predict and/or moniMr changes in parameters 3.2/3.3 (to prevent exceeding design limits) assocaiated with operating the MRSS controls including: RCS Tavg d

062000 A2.04 Ability to predict the impacts of the following 3.1/3.4 malfunctions or operations on the AC distribution system, and based ability on those predictions, use i

procedures to correct, control, or mitigate the consequences of those malfunctions or operations, efTect on plant of de-energizing a bus.

029000 A3.01 Ability to monitor automatic operation of the 3.8/4.0 Containment Purge system, including: CPS isolation.

010000 A4.03 Ability to manually operate and/or monitor in the control 4.0/3.8 room: PORV and Block valves.

006000 G0.05 Knowledge oflimiting conditions for operations and 3.5/4.2 safety limits.

035000 Kl.01 Knowledge of the physical connections and/or cause- 4.2/4.5 effect relationships between the S/Gs and the MFW/AFW systems.

012000 K2.01 Knowledge of the power supplies to the following: RPS 3.3/3.7 channels, components, and interconnections.

062000 K3.01 Knowledge of the efTect that a loss or malfunction of the 3.5/3.9 AC distribution system will have on the following: Major system loads.

064000 K4.02 Knowledge of EDG design features and/or interlocks 3.9/4.2 which provide for the following: Trips for EDG while operating (normal or emergency).

073000 K5.01 Knowledge of the operational implications of the 2.5/3.0 following concepts as they apply to the PRM: Radiation theory, including sources, types, units, and effects.

010000 K6.03 Knowledge of the effect of a loss or malfunction on the 3.2/3.6 following will have on the PZR PCS: PZR sprays and heaters.

010000 A1.07 Ability to predict and/or monitor changes in parameters 3.7/3.7 (to prevent exceeding design limits) assocaiated with operating the PZR PCS controls including: RCS pressure.

035000 A2.04 Ability to predict the impacts of the following 3.6/3.8 malfunctions or operations on the S/Gs, and based ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, steam flow / feed flow ,

mismatch.

006000 A3.03 Ability to monitor automatic operation of the ECCS 4.1/4.1 system, including: ESFAS-operated valves.

Group III Plant Systems - 8%

1 005 Resdiual heat Removal System 034 Fuel handling Equipment System 007 Pressurizer Relief Tank / Quench 041 Steam Dump System Tank System 045 Main Turbine Generator System !

008 Component Cooling Water System 076 Service Water System 027 Containment lodine Removal System 078 Instrument Air System l

028 Hydrogen Recombiner and Purge 103 Containment System l

Control System '

1 System # K/A # K/A Topic Rating l 078000 K1.04 Knowledge of the physical connections and/or cause- 2.6/2.9 effect relationships between the IAs and the Cooling water to compressor systems.

034000 K4.03 Knowledge of design features and/or interlocks which 2.6/3.3 provide for the following: Overload protection.

076000 Kl.16 Knowledge of the physical connections and/or cause- 2.7/3.1 effect relationships between the SWs and the ESF '

system.

028000 Al.01 Ability to predict and/or monitor changes in parameters 3.4/3.8 l (to prevent exceeding design limits) assocalated with operating the HRPS controls including: Hydrogen concentration.

005000 A2.04 Ability to predict the impacts of the following 2.9/2.9 malfunctions or operations on the RHRs, and based ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations, RHR valve malfunction 008010 A3.03 Ability to monitor automatic operation of the CCW 2.9/3.2 system, including: Requirements on and for the CCWS for different conditions of the power plant.

041020 A4.08 Ability to manually operate and/or monitor in the control 3.0/3.1 room: Steam Dump valves.

005000 G0.10 Ability to explain and apply all system limits and 3.3/3.5 precautions.

l Knowledge and Abilities Record Form EMERGENCY PLANT EVOLUTIONS PWR - Reactor Operator - 36%

Plant Specific Priorties l l

E/A # K/A # K/A Topic Rating l 1

l i

Group I Emergency and Abnormal Plant evolutions - 16% 1 000005 Inoperable / Stuck Control Rod 000055 Loss of Offsite and Onsite Power 000015 RCP Motor Malfunction 000057 Loss of Vital AC Electrical Instrument 000024 Emergency Boration Bus 000026 Loss of Component Cooling 000067 Plant Fire On Site l Water 000068 Control Room Evacuation 000027 Pressurizer Pressure Control 000069 Loss of Containment Integrity System Malfunction 000074 Inadequate Core Cooling 000040 Steam Line Rupture 000076 High Reactor Coolant Activity 000051 Loss of Condenser Vacuum E/A # K/A # K/A Topic Rating 000074 EK3.07 Knowledge of the bases or reasons for the following 4.0/4.4 response as they apply to the Inadequate Core Cooling:

Starting up emergency feedwater and RCP's.

000068 K2.07 Knowledge of the following components: EDG 3.3/3.4 000005 EK3.06 Knowledge of the bases or reasons for the following: 3.9/4.2 Actions contained in the EOP for inoperable / stuck control rod.

000024 A1.05 Ability to operate and monitor the following: Performance 3.1/3.2 of the letdown system during emergency boration.

000051 EA2.02 Ability to determine or interpret: Conditions requiring 3.9/4.1 reactor and/or turbine trip.

000055 G0.12 Ability to utilize symptom based procedures. 3.9/4.0 000040 EKl.01 Knowledge of the following theoretical concepts as they 4.1/4.4 apply to the steam line rupture emergency task:

Consequences of PTS.

000068 EK3.18 Knowledge of the bases or reasons for the following: 4.2/4.5 Actions contained in the EOP for control room evacuation emergency task.

000015 EA1.06 Ability to operate and monitor the following: CCWS 3.1/2.9 000026 EA2.06 Ability to determine or interpret: The length of time after 2.8/3 1 i the loss of CCW flow to a component before that component may be damaged.

000051 G0.11 Ability to recognize abnormal indications for system 2.7/2.9 operating parameters which are entry level conditions for emergency and abnormal operating procedures. l 000069 K3.01 Knowledge of the bases or reasons for the following 3.8/4.2 i response as they apply to Loss of Containment Integrity: )

Guidance contained in EOP for loss of containment integrity. l 000076 EK3.06 Knowledge of the bases or reasons for the following: 3.2/3.8 l

Actions contained in the EOP for high reactor coolant !

activity. l 000015 Al.22 Ability to operate and monitor the following as they apply 4.0/4.2 )

to RCP malfunctions: RCP seal failure / malfunction 000026 A2.01 Ability to determine or interpret: Location of a leak in the 2.9/3.5 l CCWS.

000067 G0.01 Knowledge of system status criteria which require the 3.6/4.0 notification of plant personnel.

Group II Emergency and Abnormal Plant evolutions - 17% :

000001 Continuous Rod Withdrawl 000033 Loss ofIntermediate Range 000003 Dropped Control Rod Nuclear Instrumentation 000007 Reactor Trip 000037 Steam Genreator Tube Leak 000008 Pressurizer Vapor Space 000038 Steam Generator Tube Rupture !

Accident 000054 Loss of Main Feedwater 000009 Small Break LOCA 000058 Loss of DC Power 000011 Large Break LOCA 000059 Accidental Liquid Radioactive Waste 000022 Loss of Reactor Coolant Release 000025 Loss of Residual Heat Removal 000060 Accidental Gaseous-Waste Release 000029 Anticipated Transient Without Makeup Scram 000061 Area Radiation Monitoring System 000032 Loss of Source Range Nuclear Instrumentation 000009 EKl.01 Knowledge of the following theoretical concepts as they 4.2/4.7 apply to the small break LOCA emergency task: Natural circulation and cooling, including reflux boiling.

000058 K3.01 Knowledge of the bases or reasons for the following: Use 3.4/3.7 of DC control power by EDG's.

000025 EA1.02 Ability to operate and monitor the following: RCS/RHr 3.8/3.9 cooldown rate.

i

000058 A2.03 Ability to determine or interpret: DC loads lost: impact n 3.5/3.9 ability to operate and monitor plant systems.

000054 G0.11 Ability to recognize abnormal indications for system 3.4/3.3 operating parameters which are entiy level conditions for emergency and abnormal operating procedures.

000029 K1.03 Knowledge of the operational implications of the 3.6/3.8 following concepts as they apply to the ATWS: Effects of boron on reactivity.

000022 EK3.02 Knowledge of the bases or reasons for the following: 3.5/3.8 Actions contained in SOP's and EOP's for RCP's, loss of makeup, loss of charging, and abnormal charging.

000022 Al.08 Ability to operate and monitor the following: VCT level. 3.4/3.3 000061 EA2.05 Ability to determine or interpret: Need for area evacuation 3.5/4.2 000032 G0.10 Ability to perform without reference to procedures those 2.9/3.1 actions that require immediate operation of system components or controls.

000037 K3.07 Knowledge of the bases or reasons for the following: 4.2/4.4 Actions contained in EOP for S/G tube leak.

000038 Al .01 Ability to operate and monitor the following: S/G levels, 4.5/4.4 for abnormalincrease in any S/G.

000032 EA2.06 Ability to determine or interpret: Confirmation of reactor 3.9/4.1 trip.

000007 G0.12 Ability to utilize symptom based procedures. 3.8/3.9 000008 Al.08 Ability to operate and monitor the following: PRT level, 3.8/3.8 pressure, and temperature.

000008 EA2.20 Ability to determine or interpret the following as they 3.4/3.6 apply to,the PZR vapor space accident: The effect of an open PORV on code safety, based on observation of plant parameters.

000003 G0.07 Ability to explain and apply all system limits and 3.4/3.6 precautions.

Group III Emergency and Abnormal Plant Evolutions - 3%

000028 Pressurizer Level Malfunction 000056 Loss of Offsite Power 000036 FuelIIandling Incident 000065 Loss ofInstrument Air 000065 EK3.03 Knowledge of the bases or reasons for the following 2.9/3.4 responses as they apply to the Loss ofInstrument air:

Knowing effects on plant operation ofisolating certain equipment from instrument air.

000028 EA2.08 Ability to dctermine or interprct: PZR level as a function 3.1/3.5 of power level.

000036 G0.07 Ability to explain and apply all system limits and 3.2/3.5 precautions.

s Question ID RO hxam # SRO Exam #

i PLP-030-06-01 1 1

, OMM-001-03-004 2 2 4

OMM-005-03-006 3 3

! 10CFR20-003 4 4 AP-0000-001 5 5 OMM-005-09 6 6 )

i PLP-012-14-001 7 7 OMM-002-03-001 8 8 PLP0-0-06-001 9 9 OMM-001-??-03-004 ,o 10 PLP-015-03-001 ,, 11 OMM-008-03-001 3 12 MOD-018-09-001 ,J6 13 AP-030-06-001 ff 14 ADMIN-002 d 15 OMM-001-03-002 13 16 PLP-007-001 17 RPI-09-001 18 CVHVAC-04-002 15 19 NI-14-005 16 20 AFW-09-001 17 21 RDCNT-05-005 22 AOP-018-03-002 19 23 CV-09-001 20 24 CVCS-09-005 21 25 ESF-09-002 26 DC-09-001 27

.m_ KE - .__._2._ _ , a L,_

e W

t P

4 D lea Loa ird

, e

~

kW ha l~ s m w j

4 l

i I

I i

l l

l

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Question ID RO Exam # SRO Exam #

CVCS-09-004 24 28 OP-402-06-001 25 29 CVliVAC-04-003 26 30 ESF-04-002 31 31 NI-10-003 32 32 RDCNT-05-002 33 FMP-007-06-001 28 34 ICCM-14-002 29 35 OP-101-05 001 30 36 CCW-14-003 37 37 AFW-14-002 38 38 SI-09-001 39 39 AMSAC-09-002 40 40 ESF-09-004 41 41 AMS AC-09-001 42 42 h1S-09-002 43 43 RCS-09-005 44 CSS-09-001 44 45 OP-006-001 46 46 SFPC-02-002 47 FW-09-002 48 48 RPS-09-004 49 49 VAC-05-002 50 50 EDG-09-001 51 51 RM-09-001 52 52 PZR-09-005 53 53 AIR-08-001 57 54

]

Question ID RO Exam # SRO Exam #

AOP-017-05-001 55 RHR-10-005 56 CCW-10-002 57 EPP-008-05-002 58 EDG-14-003 66 59 AOP-001 06-002 67 60 AOP-002-06-002 68 61 AOP-012-06-001 69 62 EPP-001-14-003 70 63 FRP-P.1-06-001 71 64 i

AOP-004-06-001 72 65 '

AOP-018-03-003 73 66 AOP-014-05-001 74 67 AOP-012-09 001 75 68 FRP-S.1-10-001 86 69 CVCS-10-001 77 70

. AOP-018-05-001 78 71 1

AOP-014-09-001 79 72 FP-001-05-001 80 73 I

EPP-001-14-004 74 a

i AOP-023-06-001 75 AOP-023-09-001 76 l FRP-H.1-14-002 77 i

j EPP-009-05-001 78 AOP-024-14-002 79 1

AOP-024-10-002 80 i

l OMM-022-14-002 81 i

i Question ID RO Exam # SRO Exam #

PATil-1-09-001 81 82 EPP-026/27-14-001 82 83 l AOP-033-06-001 83 84 DC-14-002 84 85 MCD 99-003 85 86 l PATil-1-05-001 87 OP-105-06-001 87 88 CVCS-14-002 88 89 FP-001-06-001 89 90 NI-09-003 90 91 PATH-2-03-003 91 92 PATil-2-03-002 92 93 NI-14-006 93 94 OMM-022-14-003 94 95 1

IA-14-001 98 96 PZR-09-002 95 97 RCS-10-002 99 98 FH-14-001 100 99 l

l PZR-14-004 100 l

l l

ESF-05-002 14

! CVCS-09-007 18 ESF-04-001 22 CVCS-14-001 23 ESF-09-001 27 RM-14-003 33 WD-09-001 34 l

l i

l 0

Question ID RO Exam # SRO Exam # 1 DC-14-003 35 I NI-10-002 36 AOP-024-10-001 44 TS3.3.1.2F-001 47 ,

PZR-10 001 54 l SG-14-002 55 )

1 ESF-09-003 56 Fli-09-001 58 1

SW-10-001 59 l CSS-09-004 60 RllR-10-002 61 CCW-10-001 62 SD-09-001 63 OP-201-06-001 64 FRP-C.I 05-002 65 FRP-J.1-10-002 76 PZR-09-007 96 RDCNT-09-001 97 J

e

., 95-2 NRC EXAM - SENIOR REACTOR OPERATOR 1

1. PLP-030-06 001 Given the following plant conditions:

l

A throttle valve is to be set 6 turns open from the full closed position l 1

1

Independent Verification is required I l

1

The operator sent to position the valve has a trainee with him )

1 Which ONE (1) of the following describes how to verify the position of this valve in l accordance with PLP-030, independent Verification?

l A. llave the Trainee position the valve and the Operator concurrently verify the correct I position.

B. 1 lave the Operator position the valve and the Trainee independently verify the correct position.

C. The Operator or Trainee should position the valve and another Operator should independently verify the position.

vD. The Operator or Trainee should position the valve and another Operator should concurrently verify position.

d

'7 K/A 294001.Kl.013.7 , A "/4 PLP-030 I

New Question l

4 I

Question 1 of 100 i

95-2 NRC EXAM - REACTOR OPERATOR '

i

1. PLP-030-06 001 Given the following plant conditions: ;

i a

e A throttle valve is to be set 6 turns open from the full closed position

Independent Verification is required

The operator sent to position the valve has a trainee with him Which ONE (1) of the following describes how to verify the position of this valve in accordance with PLP-030, independent Verification?

A. 11 ave the Trainee position the valve and the Operator concurrently verify the correct position.

B. 11 ave the Operator position the valve and the Trainee independently verify the correct position.

C. The Operator or Trainee should position the valve and another Operator should independently verify the position.

vD. The Operator or Trainee should position the valve and another Operator should concurrently verify position.

d K/A 294001.Kl.013.7 PLP-030 New Question Question 1 of 100

i 5.9 (C:ncinued) !

l

2) Personnel shall be aware of effects that mechanical valve operator !

O slack will have'on determining the number of valve handwheel turns required to position a valve and make the necessary compensation to !

allow for the slack.

3) The operator shall pay particular attention to the type of valve verified. Some verification techniques may not be appropriate for a particular type, make or model of valve due to the physical construction of the valve.

4) locked valves positions are verified the same as manual valves.

When the operation of locked valve is required to determine the position AND the locked valve requires a second position

. verification,-Concurrent Verification is preferable to having both

. . persons independently operate the valve, since the second valve operation would effectively nullify the first and, therefore, serve no purpose.

5) Remotely operated or automatic operating valves - Verification of position shall be either direct observation at the valve or observation of local or remote position indicators.

6) Throttled valves -

Verification of position shall be direct observance while the valve repositioning is performed.

7) When the operation of a throttled valve is necessary to determine its position, Concurrent Verification is preferable to having both persons independently operate the valve, since the second valve operation would effectively nullify the first and, therefore, serve no purpose.

8) Breakers, switches, wires, sliding links - Verification of position shall be direct observation.

9) Blank flanges, pipe plugs or caps - Verification of position shall be direct observation.

O PLP-030 Rev. 8 Page 10 of 23

^ ~ ~

6.1 (C:ntinued) rN 6.1.3 Concurrent Verification (INPO OP-214) f )

V 1. When performing Concurrent Verification, both the positioner and the verifier have the responsibility to perform the following:

1) Prior to manipulation of the component:

a. Determine the required component and position from the applicable document.

b. Determine the correct component is to be manipulated.

2) After manipulation of the component, determine the correct component is left in the required position.

2. Concurrent Verification should be performed for the following:

1) . Positioning a throttle velve to a specific position when the valve

. . has no accurate and discernable position indicator.

2) Performing position verification of locked valves which require a second position verification.

3) Specific evolutions where an improper positioning of a component has a high probability of resulting in an immediate plant trip, Safety System actuation DE could result in an immediate threat to safe and reliable plant operation. Examples of such evolutions are:

Installing jumpers

Lifting leads

Fuse removal l

Placing Bistable switches in the TRIP position when the parameter (s) associated with the Bistable switches has failed.

4) Consideration should be given, on a case by case basis, to performing Concurrent Verification when taking the initial actions for removing a component from eervice.

O PLP-030 Rev. 8 Page 14 of 23

,. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

2. OMM-001-03 004 Given the following plant conditions:

Tte plant is in hot shutdown due to a forced outage e A section of feedwater piping has been replaced and is ready for hydrostatic testing Which ONE (1) of the following describes what is used to mark components to be used for a hydrostatic test?

l A. Hydrostatic Test Boundary tags.

B. Line Clearance tags.

C. Red " Men at Work" tags.

@. Orange " Caution" tags, d

K/A 194001.Kl.02 3.7/4.1 #

OMM-001-9, pg 6, 5.2.1.2 Modified Question Question 2 of 100

95-2 NRC EXAM - REACTOR OPERATOR

2. OMM-001-03 004 Given the following plant conditions:

The plant is in hot shutdown due to a forced outage

A section of feedwater piping has been replaced and is ready for hydrostatic testing Which ONE (1) of the following describes what is used to mark components to be used for a hydrostatic test?

A. Hydrostatic Test Boundary tags.

B. Line Clearance tags.

C. Red " Men at Work" tags.

@. Orange " Caution" tags.

d i

K/A 194001.K1.02 3.7/4.1 l OMM-001-9, pg 6, 5.2.1.2 i i

Modified Question Question 2 of 100

1 l

5.2.1 (Continued)

CAUTION There is a potential for boron dilution if a hydrostatic test boundary component were to leak, allowing dilution of the RCS. (SOER 94-2)

- - - - a;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;

2. Caution Tags should be used for system hydrostatic tests. Each component, required to be in a specific position for the test, shall have a properly completed Caution Tag QB Cap attached.

3. Caution Tags should be used to identify instruments that are Out of Calibration. The Caution Tag should state: "This gauge has been found to be Out of Calibration. Readings from this gauge can only be used for trending purposes until it has been recalibrated".

4. If a Caution Tag is required for a component controlled from the RTGB, a Caution Cap should be placed on the control switch. If a local control also exists, a properly completed Caution Tag shall be attached.

5. If a condition exists such that component / system status indication is different from that expected for normal operation 6 E it is HOT practical to use a tag or cap, a Caution Dot should be used to identify the discrepancy. This does not include items which are covered by Clearance Information Tags (CITs) used in OMM-005.

OMM-001-9 Rev. 2 Page 6 of 17

j# 95-2 NRC EXAM - SENIOR REACTOR OPERATOR 1

3. OMM-005-03 006 Given the following paint conditions:

,

A clearance is in place for maintenance i e Maintenance is complete and is ready for testing Which ONE (1) of the following sta:ements describes a requirement in the process for remova of clearance tags for testing (clearance modification)?

\ ,g 30 + '- l A. Boundary additions do not require notifying all personnel signed on the clearance. 1 B. The clearance holder can operate equiment inside the clearance boundary without a j

release for test even if tagged.

j W

W. All personnel signed on the clearance is responsible for understanding the scope of the i change.

D. Tags may be temporarily lifted, as long as the test is completed within the same day not to exceed 2 shifts.

c K/A 294001.Kl.02 4.1 OMM-005 New Question I

1 l

f p30 p i Question 3 of 100

)

a

1 95-2 NRC 2XAM - REACTOR OPERATOR f

3. OMM-005-03 006 !

Given the following paint conditions:

A clearance is in place for maintenance i

i e Maintenance is complete and is ready for testing Which ONE (1) of the following statements describes a requirement in the process for removal of clearance tags for testing (clearance modification)? l A. Boundary additions do not require notifying all personnel signed on the clearance.

B. The clearance holder can operate equiment inside the clearance boundary without a t

release for test even if tagged.

W. All personnel signed on the clearance is responsible for understanding the scope of the change.

i

D. Tags may be temporarily lifted, as long as the test is completed within the same day not I

to exceed 2 shifts.

c K/A 294001.Kl.02 4.1 OMM-005 New Question l

l i

l J

v Question 3 of 100 l

r

J j'

! 8.5 Clairinca M:dificati:n3 i

] NOTE: IF the clearance modification results in. cancellation of ALL tags, THEN j the Clearance should be cancelled unless permis,sion is obtainsd from j the SSO or CRSS. (ACR 91-061) 1 1

8.5.1 Clearance boundaries may be changed if approved by all clearance i holders.

8.5.2 Clearance Boundary Changes will be processed as follows:

] "

1. The clearance holder requesting the change ensures a Clearance l Boundary Change / Temporary Tag Lift Request is submitted.

i j- 2. The clearance holder requesting the change is responsible for j understanding the scope of the change. If the boundary change '

8 involves only adding tags to an existing clearance, the requestor need not notify all the clearance holders of the change.

l 3. If the boundary change involves removing tags, the requestor ;

j shall notify and receive permission from the other clearance ;

j holders to make the change.

lg 4. Work groups shall stop related work as determined necessary ,

j until the boundary change is completed. If a group's work is not affected, their work may continue while the boundary change is j performed. ,

l 5. The WCC is responsible for approving and implementing all boundary changes.

1

j. 6. The clearance holder requesting the change then accepts the

{ boundary change.

i 7. If the boundary change caused other work to be put on hold, the 1 requestor is responsible for initiating another boundary change to j restore the clearance boundary and notifying the other workers

when the clearance boundary is restored.

I i 8.5.3 A tag may be temporarily lifted for a short duration. The purpose of

] a temporary tag lift is to allow a tag to be lifted to perform testing

] such as bumping a motor for rotation. The duration of the j temporary lift should not exceed the duration of the Operations i shift. -

J

!O i

j OMM-OO5 REV.32 Pass 29 of 40 1

M .

) 8.5 Clerr:nco M:dificati:n3 i

i j 8.5.4 Temporary Tag Lifts are processed the same as a boundary change.

i 8.5.5 When the need for the Temporary Tag Lift no longer exists, the 4

requestor will notify Operations.

[ 8.5.6 Operations will reposition the device and rehang the tag as required

. by the clearance tag sheet.

4 8.5.7 Ground Tag modifications are processed the same manner as a

boundary change. Grounds may be temporarily lifted using a l Temporary Tag Lift.

j 8.5.8 Work items may be added to a clearance. Additions require two ,

reviews to verify that the clearance boundary is adequate for the

! work item.

t 8.6 Training and Qualification l j 1 I g 8.6.1 Only Qualified Clearance Holders may hold clearances on Plant '

instrumentation and equipment. The requirements for being a Qualified Clearance Holder are as follows:

,f -

All personnel requesting to hold clearances shr.il attend classroom training on RNP Clearance Procedures and have their name added to I i the approved QCHL.

j -

The list of Qualified Clearance Holders (QCHL) is maintained on the j clearance computer. This list shall be r~eferr'eo to by Operations j personnel prior to issuing a clearance to any personnel.

A copy of the QCHL may be placed in the Local Clearance Book for i times when the clearance computer becomes unavailable.

Requalification of Qualified Clearance Holders shall be conducted by the Training Department on a Biennial basis.

[ -

Additions to the OCHL will be made by WCC when notified by l supervisors of individuals desiring to be added to the.QCHL with

proof of satisfactory training via a Training Report or class roster.

(CR 95-02373) i Deletions to the OCHL will be made by the WCC when notified of I any training expirations QB personnelleaving the company.

l .(CR 95-02373)

Corrections to the QCHL will also be made by the WCC when ,

discrepancies are identified during the performance of OST-913. !

} '

l LO

! OMM-005 REV.32 Page 30 of 40 1

i .

. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR 1

1 4.10CFR20 003 Which ONE (1) of the following completes the statement ...

p u d ,1 b [ Derived Air Concentration hour (DAC hour) is the product of the concentration of ra

@" smaterial in air c.d the time of exposure to that radionuclide, in hours. DAC i hours represents one Annual Limit on Intake (All), which is a committed effective dose !

equivalent of rem /.

A. 2000,2.

vs. 2000,5.

C. 5000,2.

D. 5000, 5.

i b l l

I

K/A 194001.Kl.03 2.8/3.4 '

l 10CFR20, pg 20-2,3 Modified Question l t

l g% E m sR 1 *$

un to va k

, ,,, u cecclvesf 200 DAC h o vr4 clv a~ .%3 ; ;c,, es ef r ? f r G ? "'I b? J ( O n><* it e l 95^EC c N d e t/ Cl0/C t! ' N s lf , f

~

j Obj _ WOOT1 9 ,

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.T-e-c- 2 ,. e n

/ I ,, , s l

l l Question 4 of 100

95-2 NRC EXAM - REACTOR OPERATOR 1

4 i

1 4.10CFR20 003 Which ONE (1) of the following completes the statement ...

-l i Derived Air Concentration hour (DAC hour) is the product of the concentration of radioactive ,

material in air and the time of exposure to that radionuclide, in hours. DAC i l hours represents one Annual Limit on Intake (ALI), which is a committed effective dose

equivalent of rems.

i A. 2000,2.

4. 2000,5.

C. 5000, 2.

D. 5000,5.

b K/A 194001.K1.03 2.8/3.4 10CFR20, pg 20-2, 3 Modified Question l

I I

1 Question 4 of 100

y M 20 o stale 4RDS POR N AGANdST RADIATION

,.e 130.1888 Delhimenew .._nts in relation t) state of Senisse 30.5-384E1 As used in this part:

_,,. the econoodce of (Renuned M FR 190a] Absorbeddose means the energy -.w.-. -te in relation to benefits to imparted by ionising radiation per unit the public health and safety, and other Appendis A to NRI 28402 mass of irrediated material The unite of soc 6etal and socioeconomic g w y pgigogg absorbed dose are the red and the grey considerations. and in relation to (Gy), utikaation of nuclear energy and k', AppeedisBse N Rl Meet Act meses the Atomic Energy Act of ha====d materiale in the public interest.

(Removed M FR N00.]

1964 (42 U.S.C. 2011 et seq.), as AnnuallimitonintoAe(AU)means g amended. 3 the derivedlimit for the amount of Activity is the rete of disintegration g radioactive material taken into the body Appendia C to NRI 28402 (transformation) or decay of radioactive g of an adult worker by inhalation or

[bdemsnaud 59 FR 1900.] materialThe unite of activity are the ingestion in a year. ALIis the smaller curie (Cf) and the becquerel (Bq). E value ofintake of a given radionuclide Appendis D eo MRI.20402 Adult means an individual 18 or more in a year by the rehrence man that years of age. would result in a cc mmitted effective (kmoved M FR N0a]

' Altborne radioactive moderial means does equivalent of f rems 10.05 Sv) or a radioactive material dispersed in the air connutted dose equiulent of 80 rems 1

3 in the form of duets, fumes, particulates. (0.5 Sv) to any individual organ or g asists, vapore, or geoes, tissue. (ALI values forintake by Suhport A- Generalm Air

- E oom, borne radioactivity arco means aingestion and by inhalation of selected t enclosure, or area in which I se.iest purpose E airborne radioactive materials, .

(e) The regulatione in this part composed wholly or partly of licensed establish standarde for protection material exist in concentrations--

against lontains radiation resulting from (1)In excess of the derived air

activities conducted under beenees concentretione(DACs) epeciAed in

! issued by the Nuclear Regulatory appendix B, to il 30.1001-30.3401, or Commission.These regulations are (2) To such a degree that an individual issued under the Atomic Energy Act of present in the area without roepiratory 1864. as amended, and the Energy protective equipment could exceed.

Reorganization Act of1e74. as amended.

during the hours an individual is present

) (b)It is the purpose of the regulations in a week, an intake of 0.8 percent of the in this part to control the receipt.

, possession, use transfer. and dmpoeal 6anuallimit on intake (AU) or 12 DAC-hours.

of hcensed material by any bcensee in

' ALAllA (acronym for "as low as is such a manner that the total does to en individual (including doses resulting reasonably achievable") means making every reasonable effort to maintain imm hcensed and unlicensed g radioactive material and from radiation exposures to radiation se far below the a

g sources other than background dose limits in this part as is practical radiation) does not exceed the consistent with the purpose for which standards for protection against the licensed activity is undertaken, radiation prescribed in the regulations in taking into account the state of (N

I this part. flowever, nothing in this part technology, the economics of I shall be construed as hmitmg actions V that may be necessary to protect health

,and safsty.

030.teeg seeps.

The regulations in this part apply to persons hcensed by the '%==wion to roomtve, poseoas, use, transler, or dispose of byproduct, source, or special nuclear meterial or to operste e production or utilisation facility under

= parts 30 through36,39.40,50,80,et, f 70, or 72 of this chapter, and in g accordance with 10 CPR 78.60 to a persons required to obtain a cert 18cate 3 of cannpliance or an approveo compliance plan under part 76 of this che . 'Ilie limits in this part do not ap y to dosee due to backgrouw!

re tion, to exposure of patients to radiation for the purpose of medical diagnosis or therapy, or to voluntary participation in medical seeeerch Programs, March 31,1905 (reset) 20-1 O

I

_ .~..- _ ._ _ _ _ _ _ ___ _ _ _ _ _ _ _ .. _ .__._ _ _ . _ _ . _ _ _ _ _

2duelp ~

i 20.1003 j . .. . FMIT 30 o STANDMIDS POR PROTECTION MBAINST RADIATION

! rahwam are given in Table 1. the body organo er tissess that = -

Columme 1 and 2. of appendix 5 to gushty facter.and elimber neemary 4 irradiated and the comunitted dose mods factors et thelocation of i ll 30. test-soJe01). equivelset to these organs or tissues Asc4groundredsorion means interest. units of dose equivalent

} (Hom = Iw,H,m). are the som and sievert (Sv).

re&ation imm cosmic sources: naturally ~

occurring ra&oac6ve materiale, Controllederse means an area.

j including redon (except as a deony outside of a restricted area but inside Dosisuerypeesseernesasas

, product of soures or special nuclear the site boundary, access to which can h ladleidualeressentanties that !

l meterial) and global fallout as it existe be heited by the liconese for any = peessessandouelusteeindividual I

,eassa. E la sederto i in the environment from the teenne of nuclear explaatve devices. ** ' . _f Declosedpregnant avemon means a ; does dehvased )

women who has volunts infenned to the equipment.

4 redsstion" does not include ts&ation her employer. In writing, her

{ from source, bypmduct. or I pregnancy and the settmated dele of Ef)hetin dose equivoient(He)le the j nucient matensie regulated the conceptaon Co====t. Deep. dose evolveJont (He), whicii sum of the products of the dose j #soassey(radiobioasesyl means the equivalent to the organ or tassue (H,)

j dowrmanshon of kinde, guensues or opphee to external -n ' f _ and the weigh factors (w,) '

concentratione, and. in some cases, die exposure. le the does equivais'at at a applicable to of the body organs or

i locations of re&oactive materialin the tisone depth of1 sm (1000 mg/ce'). tissume that are irradietod (He = l human body, whether by direct Department means the Department of Zwer).

(in mo counting) er by Energy ambikhed by b Deparnum M Embryo / fetus sneens the developens ,

analysis and evaluation of materiale Energy Orgnalsetion Act (pub. L e641, human organism frosi conception until I g excreted or removed from the h=a 91 Sm ass. 42 UAC. 72 %) 2 h b tune of birth.

j extent that the P--

body. t. or its duly Entrance oraccesepoint means any 4 Appsduct materiale authensed represe'nlatives, exercises location through which an individual (1) Any radioactive material (except functions formerly vested in the UA could gain acem to radietion areas or ,

j special nuclear meterial) yielded la, or Atomic Enessy f*===Aa=Aan its to redsoective matenals. This includes .

'" e pam W the Chairman, membre, adhcore, and entry or exit portals of sufficient esse to hon n g componente and transioned to the U.S.

","*" '"8'y. ifmPecun oWir

producing or utilising special nuclear Energy

', material: and Emposure mamaa being exposed to thereof pursuant to sectione los (b). (c). ioni i (2) The tallings or wastes produced by ,,, sing

,g,g,radaation or to redsoective E the extraction or concentration of gand (d) of the Ensegy Roosgenisation

$ uranium or thonum from ore processed Extesno/due m tk ponson d g 'Act of 1974 [ pub. L 88-488, as Stat.1333 the does equivalent received from prunasily for its source material content. at 1237. 42 UAC. 3814) and including discrete surface wastes 8'8'enoiened to the Mary of Emergy rah, bon omen outside h body.

a resulting frose uranium solution pursuasuo section 301(a) of die g, ,emity sneane hand. elbow. arm j extraction sees. U ' Department of Enessy Organise6en Act below the elbow. foot. knee. or les

' are below the knee.

l bodass ed by these sofution (pub. I. es.41. 91 Stat ses at 877-878. 42 g g ye dose equivojent opphee to the l extraction operetiene do not constituk UAC. 7181). l i " byproduct meterial" within this E extetaal exposure of the lone of the eye i i D'rivadoitconcentration (DACI and le taken as the does equivalent et a definition. means the ==hanon of a given Close (or kas class orinholation en&onuchde in air which,if breethed taeene depth of 0.3 centuneter(300 mg/

4 cm8).

close) means a classification scheme for the reistence man for a yur Genemlly applicehle environmental

! inhaled meterial according to its rete of constions t i 2- clearance froes the rodsofian saandoide nenne standarde regnen of meters of issued by the Environmental protection {

i the lung. Materiale are c as D. air per hour). -he in an intaks af ana ,

l W. or Y. which appines to a Agency (epa) under the authonty of the of AIJ. DAC values are given in Table L i

clearance half times: for Class (Days) Ciiliiian 3. of appendix 5 to ll 30.1001-Atomic Energy Act of tees. as amended. I 3

ofless than to days, for Claes W thatimposelimits on radiation I 20.3e01.

exposures or levene, or concentratione or (Weeks) from 10 to 100 doye. and for Derrvedairconcentst/Mour(DAC.

i quanut6es of rad 6eacuve metanal. in the Claes Y (Years) of greater than 100 doye. hour)is the product of the concestration ,

i Co//ective dose to the sum of the of radioactive anatorial in air (expressed pen,ral environment outside the indsvidual doses received in a given

_ '; oflocanone under the' as a fraction or multiple of the denved period of time by a specified population control of persone possese6ag or using l . air concentretion for each re&onuclide) resoective material.

from exposure to e specified source of and the time of expoews to that

! radiation. Government egency means any 1

redsonuclide,in hours. A heensee may executive department. ea=='aanon.

Commisalon means the Nucleer take 2.000 DAC-hours to represent one i independent estabhehment. corporebon Regulatory Commission or its duly AI1 equivalent to a committed effective

! authorised representatives, whouy or partly owned by the United dose equivalent of 6 rene (0.0s Sv). States of Asnenca.which le an i

Comicitied dose equivalent (%,) Dose or sodiotion dose is a generic

( instrumentahty of the United States. or means the dose equivalent to orgsas or term thatt means absorbed does. does any hoard, bureau. &vio6an. service. l tissues of reference (T) that will be equivalent. effective dose equivelsat, olhos. officer, authority. administration. !

received from an antake of rednect we committed does equivalent. committed

) material by an individual during the 80 effecuve dose equivalent, or total or other estabholument in the executive I year penod following the intake, branch of the Government.

effective does equivalent, se defined in Committed effective dose equivalent other paragraphs of this section. Gro (See i 30.100t}. ,

! rodsotson osso sneans an eres. '

(%m)is the sum of the producte of the Dose equirofent(%) means the weighting factors applicable to each of accessible to individua's. in which

' product of the absorbed does in tissue, radiation levels could result in en j individual receiving a dose equivalent in I excess of 0.1 tem (1 a5v)in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 30 l

30,3 DesentherSONS poseg i

1 I

k

).

J 1' ,

I 1

e l

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

5. AP-0000 001

[Given the following plant conditions:

\

You are on shift as an extra operator d) '

A Russian delegation is on-site to visit our station i

You have been assigned to conduct a tour of the Auxiliary Building with one of the 1 delegation members Which ONE (1) of the following states the minimum qualification that you must have to ;

escort 4he individual into the Radiation Control Area (RCA)? l an i

VA. You must have completed Plant Access Training AND be badged.

B. You must have Plant Access Training AND receive a documented brief by HP.

C. You must have completed authorized escort training within the last year.

D. You must be a permanent staff employee at the plant and completed Plant Access Training within the last year.

a l K/A 294001.Kl.04 3.3 7 ye s k/A l l SP- 007, pg 7 of 20, 6.1.2.1 New Question 1

(*>E n 0 M l c o o c i) [] E m o a c W#

Question 5 of 100 l-

'~

95-2 NRC EXAM - REACTOR OPERATOR

5. AP-0000 001 Given the following plant conditions:

You are on shift as an extra operator i

A Russian delegation is on-site to visit our station

You have been assigned to conduct a tour of the Auxiliary Building with one of the delegation members l

1 1 Which ONE (1) of the following states the minimum qualification that you must have to

escortthe individual into the Radiation Control Area (RCA)?

VA. You must have completed Plant Access Training AND be badged.

1 B. You must have Plant Access Training AND receive a documented brief by HP.

C. You must have completed authorized escort training within the last year.

I D.

You must be a permanent staff employee at the plant and completed Plant Access Training within the last year.

a K/A 294001.Kl.04 3.3 SP- 007, pg 7 of 20 , 6.1.2.1 New Question Question 5 of 100

6.0 PROCEDURE (Continued) 6.1.2 Authorized Escort 6.1.2.1 Individuals with a need to enter protected and/or vital areas who have not obtained proper personnel screening or completed the required training and are not listed on the Authorized Access List must be escorted by an authorized escort. Under special circumstances (restrooms, fenced construction areas) ,

wherein the escorted personnel cannot exit the area and is under l some' control, the escort need not remain with the escortee. An l escortee may not be allowed to leave the area without an escort.

All badged personnel who have successfully completed Plant Access Training are qualified to perform duties of an escort.

6.1.2.2 Normally, not more than five (5) persons will be escorted by a single authorized escort; however, in those cases where an escort can effectively supervise more than five individuals and l with the specific approval by Plant General Manager or designee, more than five individuals may be escorted by a single escort.

6.1.3 Continual Behavior Observation Program (CBOP)- AFFECT ON ACCESS h

G 6.1.3.1 All workers, while on CP&L property and when performing or reporting to work, are expected to be fit for duty and not under the influence of drugs, alcohol, or other controlled substances that could adversely affect the worker's job performance or the health and safety of other workers or the public. Each individual is responsible for self-assessment of personal fitness for duty to ensure the performance of work in a safe, reliable, and competent manner.

Each individual granted unescorted access to CP&L's nuclear plants and Emergency Operations Facility (EOF) and Technical Support Center (TSC) responders shall be subject to a Continual Behavioral Observation Program (CBOP) in accordance with NRC Regulatory Guide 5.66, Access Authorization Program for Nuclear Power Planto and 10CFR26, Fitness for Duty Programs.

O i

-SP-007 Rev. 45 Page 7 of 20

,. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

6. OMM-005-09 001 Given the following plant conditions:

. A clearance was placed in effect to work on a pump

The maintenance is complete e The clearance is ready for removal Which ONE (1) of the following describes the proper sequence for removal of clearance tags?

.d. Vent and drain tags are removed first.

4. Pumps are venteb reaker tags are removed. Nov W E

.C. Discharge valve tags are removed before the suction valve tags, fo o' 097Dt" l ,

0 D. Grounding device tags remain in place until the equipment breaker is returned to service. l po n, ~

ht se e b f,{veeve l i

oct - ~c' K/A 194001.Kl.07 3.6/3.7 !

OMM-005, rv 32, pg 23,5.2.15. I thru 8 l Modified Question !

i l

l i

Question 6 of 100

. 95-2 NRC EXAM - REACTOR OPERATOR

6. OMM-005-09 001 1 1

Given the following plant conditions: j e A clearance was placed in effect to work on a pump e The maintenance is complete

The clearance is ready for removal Which ONE (1) of the following describes the proper sequence for removal of clearance tags?

A. Vent and drain tags are removed first.

W. Pumps are vented after breaker tags are removed. 1 i

C. Discharge valve tags are removed before the suction valve tags.

D. Grounding device tags remain in place until the equipment breaker is returned to service, b i K/A 194001.Kl.07 3.6/3.7 l OMM-005, rv 32, pg 23,5.2.15. I thru 8 )

Modified Question l l

Question 6 of 100 l

5.2 Clearance Tagging Requirements 5.2.15 For systems where a pump or fan was affected by the clearance, O the following sequence of events are to be performed to prevent equipment damage: ;

i

1. Remove tags from handwheels of all vent and drain valves and j reposition as required.

'2. Remove tags frorn handwheels of all other manual isolation valves !

and reposition valves as required. For pumps, open the suction I valve before opening the discharge valve. ;

3. Remove tags on power sources to valves and restore the power supply as required. !

i

4. Remove tags from valve control switches and reposition valves as required.

5. Vent pumps as necessary to remove air from casing.

6. Remove the' tag from the power source for the pump / fan prime j mover. Restore the power source as directed by restoration lineup. ;

7. , Remove the tag on pump / fan control switch and reposition the .

switch as required.

8. Deviations from the above sequence are only allowed for safety, i ALARA or the deviation would not impact personnel or equipment safety considerations. ;

5.2.16 When a clearance significantly reduces or isolates flow through a ,

pump, the pump motor breaker shall be tagged in the deenergized l position. Additionally, a Red Cap or tag shall be placed on the control switch. (ACR 92-352) 5.2.17 Systems which normally operate at temperatures and pressures above ambient should be vented and drained if possible for the work ;

to be performed. Vents and drains used to depressurize a system' i should be tagged in the OPEN position unless operation of valve (s) is to be performed by maintenance personnel during the repair i activity. In this case, the valve (s) restoration will be documented on the clearance.

O OMM-003 - REV.32 Page 23 of 40

~ *

. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR l

7. PLP-012-14 001 Given the following plant conditions: )

l

A non-permit required confined space exists 1

l l

Welding is required to be performed inside i

1 i

Which ONE (1) of the following describes the correct requirements to enter the confined space?

i i

A. Atmospheric survey ONLY is required.

B. Inform the job supervisor of the entry and obtain a hot work permit. -y a/;. ra v! ?

l W. Reclassify the area as permit required. l D. Atmospheric survey and hot work permit is all that is required for entry. + 7e vE 7 c

i K/A 194001.Kl.14 3.3/3.6 PLP-012 Modified Question l

na LEVEL 2 l

l 4

l i

Question 7 of 100

I

. . 95-2 NRC EXAM - REACTOR OPERATOR l

l 7. PLP-012-14 001 l Given the following plant conditions:

i e A non-permit required confined space exists

Welding is required to be performed inside l ,

l Which ONE (1) of the following describes the correct requirements to enter the confined

. space? l l

l l .A . Atmospheric survey ONLY is required.

B. Inform the job supervisor of the entry and obtain a hot work permit. l W. Reclassify the area as permit required. ,

1 D. Atmospheric survey and hot work permit is all that is required for entry.

I C l

l K/A 194001.Kl.14 3.3/3.6 PLP-012

, Modified Question I

f i

I Question 7 of 100 l

l

. _ ._ .._ _ , __. _ ,~. _ . _. . . _ __ __.

5.0 PROCEDURE (Continued) )

I G 6) The job supervisor should contact the Fire Protection Technical Aide (FPTA) when ready for the initial survey. (Section 4) The FPTA will complete this per

- Section 5.2 and will document the results on the Entry Certificate.

7) Section 5 will be N/A for Non-Permit Entry Certificate. If atmospheric or other hazards are discovered, this permit will need to be upgraded to an Alternate or Permit Required Entry.

8) When there are changes in the use or configuration of I

a non-permit confined space that might increase the hazard to entrants, the supervisor shall consult wi t h the Evaluator and reevaluate that space and, if necessary reclassify it as a permit-required confined space.

9) Sections 6, 7, and 8 can be n.arked N/A.

10) List any special required safety equipment, if d

applicable, in Section 9.

11) An Attendant is not required for a Non-Permit Required Confined Space.

12) Complete Section 10 by listing the acceptable entry conditions. Refer to Attachment 6.4 for examples of 4 acceptable and prohibited entry conditions.

13) Complete Section 11 by verifying that the area is Jafe for entry.

14) Section 12, 13 and 14 can be marked N/A.

15) List in Section 15 any additional permits that have been issued to authorize work (Hot Work Permits, etc.)

or any other safety information.

16) A copy of the Confined Space Entry Certificate shall be posted at each entrance to the Confined Space.

O PLP-012 Rev. 7 Page 23 of 42

h 1

5.0 PROCEDURE (Continued) n i 17) ' Painting, solvent cleaning, and Hot Work will cause i Non-Permit Required Confined Spaces to become Alternate Permit or Permit Required Confined Spaces.

I Upgrading shall be required when this occurs. ,

18) Radiological Conditions in an Non-Permit required-4 confined space will not cause the space to become an

~

4 Alternate Permit or Permit-Required space, provided l that they are identified and controlled with a RWP.

5.5 Comoleting An Alternate Permit Entry Certificate

1. The Entry Evaluator and job supervisor shall complete' the Alternate Permit Entry Certificate as follows: (Non-Applicable sections can be N/A): ;

1) The job supervisor should obtain a Certificate number from the Work Control Group and verify that the Entry Certificate is the latest revision against the Revision Status List.

V '

2) Painting, solvent cleaning and Hot Work may cause Alternate Permit Required Confined Spaces' to become Permit Required Confined Spaces. Upgrading shall be required when this occurs.

3) The job supervisor should complete Section 1 of the l Certificate by listing the duration of the job. This must be the time to complete the assigned task.

4) The job supervisor should complete Section 2 by listing the exact location, plan't system, component and reason for entry.

5) The evaluator should verify and list in Section 3 the methods used to isolate or eliminate the hazards.

6) An Attendant is not required by regulation for an Alternate Required Confined Space, but may be invoked by the Evaluator for a specific job.

O PLP-012 Rev. 7 Page 24 of 42

.+'

5.0 PROCEDURE (Continued) 5.8 Additional Safety Precautions

1. When working in a Confined Space, the following precautions shall be observed:

1) When providing fresh air, at least 100 cfm of fresh air for each occupant should be provided. If oxyacetylene cutting, welding, or other operations that use oxygen are performed, additional air should be supplied. During welding operations, OSHA Standards require a minimum of 2,000 cfm per occupant.

2) Smoking shall not be permitted in enclosures.

3) Emergency lights should be explosive proof and or the low voltage type.

6.0 ATTACHMENTS 6.1 Areas Posted As Confined Spaces 6.2 Operations That May Create Atmospheric Hazards 6.3 Confined Space Entry Certificate 6.4 Examples Of Acceptable And Prohibited Entry Conditions For Confined Spaces 6.5 Confined Space Access Log Continuation Sheets 6.6 Atmospheric Monitoring Continuation Sheet 6.7 TLV Concentration of Selected Air Contaminants O PLP-012 Rev. 7 Page 29 of 42

. 1 95-2 NRC EXAM - SENIOR REACTOR OPERATOR l l 9, PLP-00-06 001 ,

o d Givnj the following plant conditions:

ad1B gcN(g 4

g

The plant is in cold shutdown in preparations for a refueling outage l

0 J i4f

D' g/}. e The Outage schedule is being reviewed by the oncoming shift 4

do g- d .

z4 M Which ONE (1) of the following describes " Shutdown Safety Functions" as defined in PLP-055, Outage Risk Assessment and Control? l i

A. Radioactive release control and Decay Heat Removal capability

4. Inventory Control and Containment I l

C. Containment Access control and Power Availability l l

D. Contingency Planning and RCS Pressure control l K/A 294001.K1.16 4.2 y NE W ff/A PLP-55, Def 4.9 New Question i

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Sil //IF1Csucc oF ~T .1-1/j,

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Question 9 of 100

/

l 1

. I 95-2 NRC EXAM - REACTOR OPERATOR

9. PLP-00-06 001 l Givne the following plant conditions:

The plant is in cold shutdown in preparations for a refueling outage

The Outage schedule is being reviewed by the oncoming shift Which ONE (1) of the following describes " Shutdown Safety Functions" as defined in PLP-055, Outage Risk Assessment and Control?

A. Radioactive release control and Decay Heat Removal capability -l

4. Inventory Control and Containment C. Containment Access control and Power Availability D. Contingency Planning and RCS Pressure control b

K/A 294001.Kl.16 4.2 PLP-55, Def 4.9 New Question 1

Question 9 of 100

Attichmsnt 10.1 PIga 5 of 5 Definitions SCOPE CHANGES: Changes to the outage scope, after completion of the pre-

, outage Independent Outage Risk Assessment, that potentially affect shutdown

{

safety functions described in section 8.3. The following are examples of outage. )

scope changes: '

Emergent work added to the outage schedule or worklist. See step 8.4.3.3 for j scope changes which are not risk significant. l

Added scope activities resulting from scheduled inspections, tests, or corrective / preventative maintenance. See step 8.4.3.3 for scope changes which are not risk significant.

l Moving activities from one schedule work window to another. See step l 8.4.3.3 for scope changes which are not risk significant.

Adding, changing, or deleting schedule network logic ties between specific work windows or between a specific window and another level of the network. , , , .

SHUTDOWN SAFETY FUNCTION: Functions required to ensure nuclear safety during 9 shutdown. These functions are:

Decay Heat Removal capability

Inventory Control

Power Availability a Reactivity Control of Core

RCS Pressure Control

Containment r

( i PLP-055 REV.7 Page 47 of 60

a 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

10. OMM-001-??-03 004 Given the following conditions:

A procedure is in progress when a later revision requiring immediate implementation is distributed Which one (1) of the following describes how you can determine what changes were made in the latest revision?

VA. Only the revised sections will have revision bars.

B. Only the revised pages will be provided for an immediate implementation.

C. The entire procedure will have to be re-performed so changes do not require identification.

D. The revised procedure text is placed in brackets.

a 294001. A1.01 3.4 ->NEW II/A AP-022, rv 21, pf 10 of 63, substep 8)

New Question S r mfL i Question 10 of 100

.- . - . - -... .- . . _ . - - - - - . _ . . . - . - - . . - - . _ . . ~ . . . - . - . .

5.3.4 (Continued)

,.' l

4) When revising a document to delete requirements, a review of all references, as well i as the document's historical fle, should be performed to allow appropriate

- consideration of past revisions which may have incorporated significant commitments ;

or resolutions to problems or requirements (INPO Recommendation 2.11 A-1).

5) incorporate any outstanding temporary changes intended to be made permanent. l

6) . Multiple DCFs may be resolved by one revision as follows:

a. Select one DCF as the LEAD DCF. ]

a) All signatures should be recorded on this DCF.

b) The Lead DCF shall be completely pi -:::::1 as described in this procedure.

c) The training, offective date, and approval / voiding apply to the entire

]

revision

b. Annotate on the LEAD DCF on Page 1 that it is the

LEAD DCF for DCF #(s)

(list all Secondary DCF #s)". ,

c. All secondary DCFs shall remain with the LEAD DCF throughout the process j a) Strike out the review page using a diagonal line and annotate the review signoff page "See LEAD DCF # XX-X-)000(.* !

l b) ' The secondary DCFs may be volded if all information is combined on the primary DCF form.

c) No L .r entries are required on these DCFs. I

7) Any permanently approved revision to a page, editorial, administrative, or otherwise, requires the page revision number to advance to the revision number of the overall document.

8) Text, tables, and data revisions shall be annotated by placing a black vertical bar, called the Rev. Bar, in the adjacent right margin. The following exceptions apply:

a. Material that has only beeri shifted in position on the page or from page to page due to addition or deletion of text should not be barred. ,

b. The page/ revision line should not be barred. l'

c. For sketches in non-procedure documents, the changed area should be clouded and the cloud tagged with the revision #.

d. Revisions that change an entire procedure such that Rev. Bars would not benefit the user may be annotated on the List of Effective Pages with a i statement describing the complete revision and the lack of Rev. Bars. This !

statement would be deleted in the subsequent revision !

Example: The body of this procedure has been revised in its entirety. !

Revision Bars are not used to mark changes to the body of the procedure.

Rev. Bars are used only to mark changes to the Attachments.

I i

AP-022 Rev.21 Page 10 of 63

. l l,." '5.3.4 (Continued)

9) Any markup that was provided by the sponsor should continue to accompany the ]

{ DCF throughout the process. The markups should be destroyed when the DCF is l

NM.

10) A copy of each affected page of the document which is being revised should be attached for the convenience of reviewers, and marked in bold letters *0LD". Such pages are not required to be filed with the permanent record of the DCF.

i

! E i e Pen-and-ink changes are normally NOT incorporated into the electronic media POM, and require

!. _ additional controls as described in Steps 5.3.5. through 5.3.6.

! e The Penend-ink process shall NOT be used to bypass required formatting requirements IAW applicable writers' guides if adequate pise-g time is available.

l

e Pen-and-ink changes DO NOT relieve the sponsor or approver from any requirement of this procedure '

unless specified in Steps 5.3.5. through 5.3.6. ;

5. Pen-and-ink changes may be performed in circumstances where Temporary Changes are not allowed, time available is critical, or word processing is impractical, as follows: !

1) Process the DCF. Safety Review, and changes as prescribed in this procedure.

2) Using black ink capable of distinct reproduction and legible handwriting, make the required changes to the document, including page revision numbers.

3) If the change requires the addition of pages due to length, then typed pages may be added provided that it is obvious that the new pages are part of the ponend-ink change (pages with letter designations are acceptable, e.g., '9a, 9b, 9c*).

4) Submit the cover page and list of effective pages (LEP) to Word Pis ;rg to change the revision number on the cover sheet and the revision number for each revised page to the next sequential revision number.

5) The ponend-ink changes to the body of the document may be reviewed, recommended and approved as is.

6) Document Services shall process and distribute the change normally. j

6. When it is practical or desirable to electronically process the approved ponend-ink change (s),

the responsible group shall initiate a new DCF and process the document as follows:

1) if no additional changes to the document are to be made, then annotate in Section ll, C:::-4"=. of Change, that this change incorporates approved pen-and-ink changes AND list the DCF numbers to be incorporated.

a. The DCF for a package that incorporates previously approved ponend-ink changes requires only a CHECKER review in Section Ill.

AP.022 Rev.21 Page 11 of 63 j

, . 95-2 NRC EXAM - SENIOR REACTOR OPERATOR lc i

t i *

11. PLP-015-03 001 Which one (1) of the following is the MAXIMUM allowable number of hours that an operator may work at the RTGB position in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period and the' approval required? ... ,

i,,, 4 I

Time Allowed dooroval Re,auired .

l \\ l A. 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Superintenden Shift Operations l

4. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Plant General Manager C. 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months Superintendent Sh t Operations D. 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months Plant General Manager b

V

)), o 8d ,3 l_ K/A 194001.A1.03 (2.5/3.4) [, j c e /se l PLP-015, Pg.13. ,

! Modified Question l i i !

l l

l

\ l Question 11 of 100 I

_ _ _ _ _ _ _ . ___ __. __ _._ ._ .. _ _ _ _ _ . _ _ . - - _ - ~ . _ ~ - _ _ . _ _ _ - . _ _ _ _ ___

5.0 PROCEDURE (Continued) 5.2 Technical Specifications requirements set forth in detail in Section 6,2.3.b state that administrative procedures shall be developed and implemented to limit the working hours of Plant Staff who perform safety related functions. This procedure applies to the following job categories for individuals on-shift, performing safety-related work activities:

. all licensed Operators, Auxiliary Oparators, RC Technicians, EC Technicians, IEC Technicians, Electricians, Mechanics, and their First Line Supervisors. First Line Supervisors are defined as those individuals who direct safety-related work activities of

,the above personnel. All other job categories are exempt. This information is intended to clarify and expand upon the requirements defined within definitions 4.1.1 and 4.1.2, and represents Robinson's interpretation and application of the available regulatory guidance. (ACR 93-211) 5.3 Reauirements '

Enough' plant operating personnel should be employed to maintain adequate. shift coverage without routine heavy use of overtime. i The objective is to have operating personnel work 4_ normal shift, based on their work schedule, while the plant is operating. However, in the event that unforeseen problems I require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, ,

or major plant modifications, on a temporary basis, the following guidelines shall be followed (Reference Technical Specification 6.2.3): '

l. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time, i

PLP-015 Rev. 1 Page 12 of 18

5.0 PROCEDURE (Continued)

2. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in

& any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven day period, all excluding shift turnover time.

3. _A break of at least eight hours should be allowed between work periods, including shift turnover time.

4. Except during extended shutdown periods, the'use of overtime should be considered on an individual basis and not for the entire staff on a shift.

These requirements shall apply to the plant staff who perform safety-related work activities (e.g., Senior Reactor Operators, Reactor Operators, Auxiliary Operators, Health Physicists, and key Maintenance personnel).

It should also be noted that Generic Letter No. 82-12 encourages procedures that would allow licensed operators at the controls ;

to be periodically relieved and assigned to other duties away from the control board during their tours of duty.

In accordance with an NRC internal memorandum, dated September 17, 1992, " Health Physics Position: Clarification of Nuclear Power Plant Staff Working Hours," the 7-day week period should not be treated as a fixed, one-week period (e.g.,

Saturday through Friday) such that the 7-day window is reset at the end of the week. Rather, this 7-day window should be treated as ADZ rolling 7-day period.

PLP-015 Rev. 1 Page 13 of 18 l

t 1

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

12. OMM-008-03 001 Given the following conditions:

An oncoming RTGB Control Operator is preparing to take the shift from the offgoing (u j)_ operator.

.Q The plant is in cold shutdown. g NE 6***Cr ra 17E N Which one (1) of the following describes the correct actions for turning overin evolution in

- progress?

4 A. The evolution must always be finished prior to completing turnover.- m rosnB4 f l ho(L B. Turnover may proceed without special considerations when in cold shutdown.

W. Superintendent Shift Operations approval is ne. <ary to turn over an evolution that is still in progress.

D. Turnover may proceed providing the operators work in parallel on the evolution for a minimum of 10 minutes.

'U TM/J fvgg pg ,c c

K/A 194001.A1.09 2.7/3.9*

OMM-001-12, Rev. 3, page 11. middle paragraph j Modified Question j

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j DIS TR A C TC R 3 fooR j l

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Question 12 of 100

)

.i 6 5.0 (Continu2d) 5.5 RO and BOP Shift Relief Instructions

1. The oncoming RO shall complete and prepare the applicable portions of Attachment 6.6. I

2. The oncoming RO shall perform the following:

e review CD's Narrative Log and other logs taken by the RO e review Status Board

. walkdown the RTGB

3. The oncoming and offgoing R0s should discuss the following:

e current status of plant equipment E9Il Evolutions in progress will be turned over only with the approval of the SSO.

e evolutions / maintenance in progress e evolutions / maintenance planned for the next shift e any LCO Action Statements in effect e outstanding OUPs, EIR, waste releases e major clearances and new caution tags e abnormal equipment performance trends or behavior E99TE Significant addits snal information discussed during shift turnover should be recorded on the shift turnover checklist as an aid to the oncoming RO during the shift.

e any other information that would be of assistance to the oncoming RO Page 11 of 77 OMM-001 12 Rev. 3

s . . .

95-2 NRC EXAM '- SENIOR REACTOR OPERATOR j 13. MOD-018-09 001 !

/Given the following plant conditions:

g / 1 4d l :

M gt l a You are stationed in the Work Control Center i lh e Several Maintenance personnel bring work packages for your review f

1 e No permanent design changes are required for any of the packages 1 k i Which ONE (1) of the following describes an activity that requires issuing a Temporary Modification IAW MOD-018, Temporary Modifications?

l i

A. Jumpers installed during a calibration procedure.

5 *#

.- - B. Drain hoses to be installed to drain a pump casing for maintenance.

p? C p 'j d .< - C.

i Extension cord used to power a portable sump pump in the Aux. Building.

4 Demineralized water hose connection to the Primary Air Compressor Cooler.

I d Wo @

,,9 cc w ; 1 r n estn/e c/tt ?

K/A 194001.A1.10 2.9/3.9 MOD-018 Modified Question h6 L EJEL]

f 7)o { q y c nyay gg5 m cc oovi ojo wc w L A +'# T "/M Question 13 of 100 4

95-2 NRC EXAM - REACTOR OPERATOR I

10. MOD-018-09 001 l I Given the following plant conditions:

l

You are stationed in the Work Control Center

Several Maintenance personnel bring work packages for your review

No permanent design changes are reqLired for any of the packages

] Which ONE (1) of the following describes an activity that requires issuing a Temporary Modification I AW MOD-018, Temporary Modifications?

A. Jumpers installed during a calibration procedure.

B. Drain hoses to be installed to drain a pump casing for maintenance.

C. Extension cord used to power a portable sump pump in the Aux. Building. i

@. Demineralized water hose connection to the Primary Ali Compressor Cooler. -

d K/A 194001. A1.10 2.9/3.9 MOD-018 !

Modified Question l

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. Question 10 of 100

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- 1 4.0 GENERAL (Continuso)

O l

I 4.2 Antilicability l

i 4.2.1 Temporary Modifications (TMs) apply to projects not requiring a i permanent design change. Major distinctions of a TM include:

e No Proposal, Plan, or Plant Review Group presentation is appropriate.

No changes are allowed to the Unit 2 or ISFSI Technic'al Specifications or that create an unreviewed safety question.

e No extensive walkdown or turnover review process is needed since the field work is to be of a simple nature.

]

e An exception list is not permitted because of the smaller range

. of activities and the additional implementation and restoration requirements of the "h process.

4.2.2 If larger scope projects are needed on a temporary .tsis, they should be done by a standard modification, and restored by another standard ,

1 modification. '

4.2.3 The following are examples of suitable uses for TMs:

O

Disabled Annunciator Alarm, Valve, or Other Component l

i e Lifted Lead e Electrical Jumper / Temporary Power Feed e Pulled Circuit Card e Temporary Setting Change e Mechanical Jumper (Temporary Piping or Flexible Hose) e Blank Flange e Structural Support ;

e Enclosure or Partition e Leak Repair i Non-code repairs involvin5 the use of on-line leak sealing should have the Mod package reviewed in accordance with TMM-031 )

in addition to this procedure.

Removal of Component For Repair Without System Clearance

Installation of Unmonitored Test Equipment MOD-018 Rev. 13 Page 5 of 21

4.0 m (Continuad)

O e Non-code repair of. code components. (Use of a TM is limited by the requirements of Generic Letter 90-05, including prior NRC approval of repair.)

4.2.4 TMs are not required.in these specific instances:

e When standard troubleshooting or calibration techniques are used which may require jumpers or lifted leads and the system condition is being monitored.

. When extension cords, drain hoses, welding leads, air hoses for tools, DI water or deepwell water supply connections, service air connections, etc. are used in the intended fashion.

. When the activity is fully covered by an approved procedure or j modification.

i k

4.3 Resoonsibilities This section is a summary that contains generalities. Refer to Section 5 of this procedure for specific requirements.

4.3.1 Engineering

Upon receiving a request for a TM (from Operations, Maintenance, etc.), Engineering evaluates need and requests the appropriate RESS Discipline Unit (s) prepare the TM (unless otherwise agreed). Engineering will designate the appropriate individual to serve as the TM Coordinator.

4.3.2 RESS Discipline Unit (s): Prepares design for TM, unless otherwise agreed.

4.3.3 TM Coordinator (TMC): Assumes TM coordination duties and ensures TM ,

tags have been properly hung.

4.3.4 Operations

Hangs tags, maintains a log of TM Tags (ATTACHMENT 6.5),

and ensures that the Plant is not placed in an operating condition for which the TM is not analyzed.

4.3.5 Document Control: Maintains TM Log (ATTACHMENT 6.1), provides copies of the TM and TM Log, and assembles the Closeout Package.

() MOD-018 Rev. 14 Page 6 of 21 ;

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.- . --- .. . - . . . . _ . - - - . . _ . _ . _ = . . - .

95-2 NRC EXAM - SENIOR REACTOR OPERATOR l

l

14. AP-030-06 001

[Given the following plant conditions:

\

g9 f ) . The plant was at 100% power f*

A LBLOCA with significant fuel damage has occurred

. A General Emergency has occurred -

,[ 1

. f. '

Which ONE (1) of the following describes the. approval permission and ' radiation levels )

required to administer KI Io' site personnel? - -

- jy"

Ypu v Moi A A Ap

,/

l LEnc '

VA. Radiation Control Director - 25 Rem /s ~

i B.

D e gy b g *c

~

7* N47* * ' d i Radiation Control Director - 10 Rem C. Site Emergency Coordinator - 25 Rem D. Site Emergency Coordinator - 10 Rem '

a h I 1, F no s fM~' .

O 3 EC pq W;g Ro n ees K/A 294001.A1.11 4.1 y g.u r, ;., r / c/4 EPSPA03 g

New Question A l.Ik 1

t l

l Question 14 of 100

95-2 NRC EXAM - REACTOR OPERATOR i i

i l

' 11. AP-030-06 001 l

Given the following plant conditions:

The plant was at 100% power !

l

A LBLOCA with significant fuel damage has occurred i

i e A General Emergency has occurred j Which ONE (1) of the following describes the approval permission and radiation levels l

required to administer KI to site personnel?

i VA. Radiation Control Director - 25 Rem l

B. Radiation Control Director - 10 Rem C. Site Emergency Coordinator - 25 Rem D. Site Emergency Coordinator - 10 Rem i

a ;

K/A 294001.A1.11 4.1 EPSPA03 i New Question l 1

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Question 11 of 100

l 8.3.1 PURPOSE i

1. To proy.ide guidance for the administration of potassium iodide (Kl).

O j ' 8.3.2 RESPONSIBILITIES j -1. The Site Emergency Coordinator (SEC) will be responsible for

i 4

a. Advisirig off sita authorities whenever it is estimated that

! releases may be of such levels that administration of Kl may be l_ a9propriate.

The actual decision to administer Kl will be made by the i officials of the State.

1

b. Approving the administration of Kl to CP&L employees and any

j. contractors where CP&L is responsible.

2. The Radiological Control Manager / Director will be responsible for: ;

1

a. Determining the need for administration of Kl to CP&L and I contractor employees who are deemed subject to a thyroid committed dose (CDE) of greater than or equal to 25 Rem. I

b. Advising the SEC or the Emergency Response Manager (ERM) of any recommendation to administer Kl. i

3. The E&RC Unit is responsible for maintaining Klinventory as part of the emergency kit inventory.

8.3.3 INSTRUCTIONS

1. The Radiological Control Director (RCD) and the Radiological Control Manager (RCM) shall coordinate to ensure the following steps are completed, i

a. The RCD will provide recommendations for in plant personnel and the RCM will provide recommendations for off site personnel.

I

2. Direct removal of unnecessary personnel from areas of high radioactive iodine concentration as practical.

l O EPSPA-03 REVISION O Page 3-4 of 3-8

I I 8.3.3 INSTRUCTIONS i j 3. The RCp shall provide a recommendation to the SEC regarding the administration of Ki to CP&L and contract personnel.

! a. Included in this recommendation shall be the projected doses without the administration of Kl.

l 4. The RCM shdil advise the ERM of a recommendation for off site j authorities to consider in their decision to administer Kl for non- !

! CP&L personnel. l j

i a. Included in this recommendation shall be the projected doses j without the administration of Kl.

'b. Use of Kl shall be recommended when actual or projected 4

. thyroid committed dose equivalent is greater than or equal to i 25 Rem.

1 i 5. If Administration of Klis authorized by the SEC or ERM for

personnel on site or out in the field then perform the following

3

. a. Administer Kl prior to or immediately after inhalation I

See Attachment 8.3.5.1, Graph of Percentage of Blockable Dose from lodine-131,

b. Only one single dose (approximately 130 mg) shall be administered per day.

I

!

Kl shall not be administered more than 10 days without i the authorization of a doctor.

[ 6. The E&RC Team Leader and the Environmental Monitoring Team i Leader shall perform the following:

l j a. Prior to administering KI, advise each individual that use is l voluntary.

i

b. Ask if the individual has ever had an allergic reaction to iodine.

}

Do not administer Kl to any individual who is allergic to ;

3 iodine.

I c. Do not issue any KI which has passed the expiration date i f on the bottle.

!O l

EPSPA-03 REVISION O Page 3-5 of 3-8 I .

4 E

8.3.3 INSTRUCTIONS

] d. A'dminister one single dose of KI to each potentially affected j CP&L or contractor employee.

Kl is located in the emergency kits in the TSC, EOF, OSC, 3

Back up OSC, Control Room, Byerly and Wilson Hospitals and the Environmental Monitoring Emergency Kits. I i

, e. Record the following information on the Attachment 8.3.5.2, j Potassium lodide Log. l l

Name and Social Security Number of the individual; j

Lot number and expiration date of KIissued; and

!

Name, Social Security Number and reason for individuals i . choosing not to take Kl.

i

! 7. Ensure that allindividuals using Kl receive a follow up bioassay as applicable.

t 8.3.4 RECORDS O N/A i

8.3.5 ATTACHMENTS !

l

\

8.3.5.1 Graph of Percentage of Blockable Dose from lodine-131 l 8.3.5.2 Potassium lodide Log EPSPA-03 REVISION O Page 3-6 of 3-8 W - -e

. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

15. ADMIN 002 Given the following plant conditions;

An Operational Surveillance Test (OST) on "A" Si pump is in progress e The pump is placed on recirculation IAW the OST Which ONE (1) of the following describes the maximum amount of time the pump can be on recirculation and the basis for the time limit?

VA. 30 minutes, to avoid damage from overheating.

B. 15 minutes, to limit the time the pump is inoperable.

C. 30 minutes, to allow for sufficient warm-up to normal operating temperatures for consistent, meaningful data.

D. 15 minutes, to ensure sufficient run time to check for proper temperature and vibration.

a K/A 194001.A1.12 3.1/4.1 a NI% to / r e d /e-~.v e M */c (# f OST-151-1, page 7 of 30, P&L #3

/ Modified Question

\

M o o C tm ~ G- 10 3. w ! 3. y r>L i., E c u

)

Question 15 of 100

)

C _

. 95-2 NRC EXAM - REACTOR OPERATOR

12. ADMIN 002 Given the following plant conditions; e An Operational Surveillance Test (OST) on "A" SI pump is in progress

. The pump is placed on recirculation IAW the OST l

Which ONE (1) of the following describes the maximum amount of time the pump can be on recirculation and the basis for the time limit?

l l

VA. 30 minutes, to avoid damage from overheating.

B. 15 minutes, to limit the time the pump is inoperable.

C. 30 minutes, to allow for sufficient warm-up to normal operating temperatures for consistent, meaningful data.

D. 15 minutes, to ensure sufficient run time to check for proper temperature and vibration. ;

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a i K/A 194001.A1.12 3.1/4.1 OST-151-1, page 7 of 30, P&L #3 l Modified Question l

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i Question 12 of 100

~

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.

I 4.0 PRECAUTIONS AND LIMITATIONS

, 1. To prevent overpressurization of the Refueling Water Purification System, the Refueling Water Purification Pump shall be shutdown and SFPC-8055, RWST RET, closed.

2. 'Immediately following the start of an Safety Injection Pump "A",

recirculation flow shall be checked. If recirculation flow is less than 30 gym, the SI Pump shall be stopped immediately and declared inoperable, i

3. Safety Injection Pump "A" shall be run for a minimum of 15 minutes, but shall not exceed 30 minutes while on mini-flow recirculation.

l 4 If the starting limitations stated in the Precautions and l Limitations Section of OP-202 are exceeded, motor. damage can occur due to motor overheating. (ACR 92-325) i

5. Any steps that are not applicable shall be marked N/A and the reason

/ documented in the Comments section of Attachment 8.4.

V)

6. The performance of this OST shall be coordinated with other plant i evolutions such that the minimum equipment operability requirements of the Technical Specifications are met.

7. The principles of ALARA shall be used in planning and performing work and operations in the Radiation Control Area, i

8. If OST-152 1 is being performed concurrently, the applicable sections of OST-152-1 shall be referenced prior to stroking each valve.

9. This procedure has been screened IAW PLP-037 criteria and determined not applicable (N/A) to PLP-037.

O OST-151-1 Rev. O Page 7 of 30

~ 95-2 NRC EXAM - SENIOR REACTOR OPERATOR 1

16.,OMM-001-03 002

' Given the following plant conditions:

e'$hssure instrument on the RTGB with indication of 1000-1500 psig currently reads b

1400 psig.

Which ONE (1)' of the following is the MAXIMUM administratively allowed pressure deviation on redundant instrumentation?

> fro w wy4 7 A. 1% of full scale f g , fg B .- 2% of full scale /% x Ifos /

C. 4% of full scale f ;; , reo II

@. [% of full scale f ?. ' 13 # "

d.

/

K/A 194001 All3 (4.3/4.1)

I OMM-00l-11, Pg. I1, 5.2.3.4.

pg ,, a f Modified Question Dbar ~

u n t>t n n o n 7 RLF b* tsNr HElf USE fl E A L VA L L/E3 I

Question 16 of 100

(

95-2 NRC EXAM - REACTOR OPERATOR

13. OMM-001-03 002 Given the following plant conditions:

3

A pressure instrument on the RTGB with indication of 1000-1500 psig currently reads

1400 psig.

Which ONE (1) of the following is the MA.XIMUM administratively allowed pressure

deviation on redundant instrumentation?

, A. 1% of full scale B. 2% of full scale C. 4% of full scale

@. 5% of full scale d

K/A 194001 All3 (4.3/4.1)

OMM-001-11, Pg.11, 5.2.3.4.

Modified Question l

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I Question 13 of 100

5.2.3 (C:ntinusd)

4. The following guidance does M apply to Steam Flow instruments when O Turbine load is less than 20% ME only applies to RTCB instrumentation.

It is M intended to apply to instrumentation of different types >

sensing the same parameter (i.e. , a direct reading local pressure gauge shall M be compared against a pressure transmitter driving a remote meter): (CR 95-02700)

1) A deviation greater than 5% of full scale is used as the DOS limit.

The instrument shall be declared OOS.

2) Attachment 6.1 provides values.

5. The Steam Flow loops are pressure compensated to account for density changes in the steam that occur from low-load to full-load operation.

The calibration of the loops are set up for optimum performance at 100%

of rated flow. As a result, the performance of the loops at low flows is not as accurate, and can be operable but may not meet the 5%

tolerance. Based on this information, the following should be used for a channel check of the redundant Steam Flow instruments: (CR 95-02700)

1) IE Turbine load is less than or equal to 204, M e Steam Flow should be greater than 90% of indicated Feed Flow.

M.Q i e The tolerance between redundant Steam Flows should be within lit of full scale.

2) IE Turbine load is greater than 204, M the tolerance between redundant Steam Flow instruments should be within St.

3) IE it appears that the 5% tolerance will not be met prior to exceeding 20% Turbine load, M Engineering should be contacted to evaluate the specific situation. '

O OMM-001-11 Rev. 2 Page 11 of 45

m i _ 64 4 g m- A-.,---4 -m e- A. - s- 24--:-a- J w-w- S WM &a AL--2-A 3 = a-WA-L24 mu&,, -e -AJ'S -- ---~ - - + e n- k -- -

95-2 NRC EXAM - SENIOR REACTOR OPERATOR >

i

.v r:c.,

17. PLP-007 001 r Which one (1) of the following is the H. B. Robinson Administrative TEDE exposure limit for emergency workers who are attempting to protect valuable property?

I A. 1 REM

B. 5 REM

< . 10 REM D. 25 REM l I

c K/A 194001.A1.16 3.1/4.4 PLP-007, Rv 34 Modified Question l l

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Question 17 of 100 t

i 5.0 EL&H (Continu:d)

The west ~a cess road and the east access road will be used as appropriate to depart from the site as advised, and evacuation from the 10-mile EPZ will be by way of appropriate evacuation routes .

identified in Figures 5.1.1-2 and the Annual CP&L Safety Information.

5.4.4.3 centrol of Persennel Radiation Tvnemures 4

Although an emergency situation transcends the normal requirements for limiting exposures to ionizing radiation, guideline levels are '

established in EPOSC-04, " Emergency Work Control," for exposures 4 that may be acceptable in emergencies. The maximum TEDE received by

)_ any worker should not exceed established regulatory limits (see 5.4.4.3.1). Every reasonable effort will be used to ensure that an emergency is handled in such a manner that no worker exceeds these limits. This also includes the following personnel: assessment groups, first aid, personnel decentamination, ambulance service, and medical personnel.

The administration of radioprotective drugs to CP&L personnel and contractor employees may also be useful in mitigating the consequences of inhalation of radioactive materials during an emergency.

s Procedures for the administration of radioprotective drugs to CP&L and contractor employees are described in EPSPA-03, " Administration of Potassium Iodide".

Decision-making is based on conditions at the time of an emergency and should always consider the probable effects of an exposure prior to allowing any individual to be exposed to radiation levels exceeding the established occupational limits. The probable high radiation acute exposure effects are:

e no to so nam in i dav - no physiological changes are likely to l be observed. I l

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PLP-007 Rev. 34 Page 90 of 226 I

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T 5.0 EL&H (crntinu d)

~

e 50 tt 100 R , in 1 dav - no impairment likely but some

physiological changes, including possible temporary blood changes, may occur. Medical observations would be required after exposure.

' e 100 en 100 nam in 1 dav - some physical impairment possible.

Some lethal exposures possible. i The following subsections describe the criteria to be considered for life-saving and facility protection actions. j i

5.4.4.3.1 Lifesavinn A-tions in emergency situations that require personnel to search for and remove injured persons or entry to prevent conditions that would probably injure numbers of people, a clanned dose shall not exceed i l

limits as outlined below: !

Dose Limit i Ram Trnri Activity c5ndition {

5 All 10 Protecting valuable property Lower dose not practicable l 25 Lifesaving or protection Lower dose not of large populations practicable

>25 Lifesaving or protection of only on a voluntary large populations basis to persons fully aware of the risks involved 8

Doses to the lens of the eye should be limited to three times the stated TEDE value and doses to any other organ (including skin and body extremities) should be limited to ten times the stated TEDE value.

PLP-007 Rev. 34 Page 91 of 226

~

l- 95-2 NRC EXAM - SENIOR REACTOR OPERATOR i.

). I8. RPI-09 001 Which one (1) of the following describes the signal that is input into the Rod Insertion Limit Computer?

J VA. The Pulse to Analog converter output. b

B. The Bank Overlap Unit output.

( C.

3 The Tref signal.

( D. Power Range Average power.

, u-a

] I K/A 014000.Kl.01 .2/3.6 0ffII CRDS l RDNCT, RPI l

) New Question J

1 e

i Question 18 of 100

LESSON BODY KEY AIDS

c. Each step corresponds to a 5/8 inch movement of a group of rods when the slave cycler completes a movement

d. Rezeroed by pushing rod control startup push button on control board

e. Can be adjusted by a thuinbwheel switch located under the cover

~ N'OTE: Simulator has digital indicators to reduce wear and tear D. PULSE TO ANALOG CONVERTERS RPI-TP-5, 6 OBJ. #5. 7

1. One for each control bank

2. Located on a single chassis in RPI cabinet #2, in Unit 2 cable spread room

3. Receive up or down pulses from the slave cycler in the rod control system !

l 1

4. P/A converter converts these pulses to a DC analog signal l

5. This signal is sent to rod bottom bypass bistable, to rod 1 insertion limit circuit (TR-409 and alarms), and to local !

indication ;

RPI REY. O PAGE 10 OF 23 '

l

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LESSON BODY KEY AIDS 4

6. A five position selector switch allows display of OBJ.#8 ,

position of the selected bank l i

7. A manual / auto spring loaded switch allows a specific OBJ. #8 '

P/A converter position to be adjusted if the switch is held in the manual mode E. ROD BOTTOM BISTABLE OBJ. #7

1. Provides indication, control, and protection functions in the l event of a dropped rod i l

2. Bistable setpoint is adjustable and normally set at 20 steps from bottom of travel

3. Output relay driven by the bistable operates the following alarms and controls

a. Red rod bottom light on control board (located under individual RPI)

b. Rod bottom rod drop annunciator (APP-005-F2)

c. Turbine runback to 70%, load reduced at 200% per minute

d. Block automatic rod withdrawal RPI REY. O PAGE 11 OF 23

i 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

)

19. CVHVAC-04 002 Given the foliowing plant conditions:

e "A" EDG is out of service 1

Reactor Trip and Si have occurred e Ten minutes later the Startup Transformer is lost ,

Which ONE (1) of the following describes which components are affected by the Loss of Startup Transformer?

.VA. I and HVH-2, Service Water Pump "/ '

B./ HVH-1 and HVH-2, RHR Pump "B" 1

CI HVH-1 and HVH-3, CCW Pump "C" D.' HVH-1 and HVH-3, St Pump "A"

\

a K/A 022000.K2.01 3.0/3.1 EDP-002, pg i1 Modified Question P)Emotti rpj(c gf?aceYro u l

Question 19 of 100

95-2 NRC EXAM - REACTOR OPERATOR 1

l

15. CVHVAC-04 002 Given the following plant conditions:

1

"A" EDG is out of service

Reactor Trip and SI have occurred

Ten minutes later the Startup Transformer is lost Which ONE (1) of the following describes which components are affected by the Loss of Startup Transformer?

VA. HVH-1 and HVH-2, Service Water Pump "A" B. HVH-1 and HVH-2, RHR Pump "B" I C. HVH-1 and HVH-3, CCW Pump "C"

,- D. HVH-1 and HVH-3, SI Pump "A" l

K/A 022000.K2.01 3.0/3.1 EDP-002, pg 11 Modified Question l

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Question 15 of 100 l

. _- .- . .- ..- -_-- - -~~-.-.-. -- .-._ ._ .-- _

4 j i

j 7.0 480V-El Section 7.0

} Page 1 of 2 I

lO 480V-El l POWER SUPPLY: NORMAL - 4160V BUS 2 (52/13) LOCATION: E-1/E-2 ROOM i

, CMPT LOAD TITLE CWD BKR NO. EDBS LOAD TAG NO. NO. EDBS NO.

17A PT'S & METERING EQUIPMENT (*)

4 N/A N/A N/A ,

i 17B EMERGENCY DIESEL GENERATOR A TO 480V

/ BUS E-1 890 52/17B 480V-El 18A PT'S & METERING EQUIPMENT (**) N/A N/A N/A jsB STATION SERVICE TRANSFORMER 2F TO 480V 892 52/188 BUS E-1 480V-El 19A CV SPRAY PUMP A 287 52/19A CV-SPRAY-PMP-A 19B CV RECIRC FAN, HVH-1 511

~

HVH-1 52/19B O J9C SERVICE WATER PUMP B SW-FMP-B 832 52/19C j0A AUX FEEDWATER PUMP A 651 52/20A AW- FMP- A -

fB SERVICE WATER PUMP A SW-FMP-A 831 52/208 20C CV RECIRC FAN, HVH-2 512 HVH-2 52/20C 21A FEED TO MCC-5 (NORM POWER) & MCC-16 1187 52/21A MCC-5, MCC-16 218 CHARGING PUMP B 162B 52/21B CHC-PMP-B 21C SAFETY INJECTION PUMP A 237 52/21C SI-FMP-A 22A RESIDUAL HEAT REMOVAL PUMP A -

214 52/22A RHR-PMP-A 22B 480V BUS E-1 SUPPLY TO Sl PUMP B 891 52/22B 480V-Ei, E2 22C COMPONENT COOLING WATER PUMP B 205 52/22C CCW-PMP-B l i EDP-002 Rev. 5 Page 11 of 14

a 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

20. NI-14 005 i Given the following plant conditions:

4

The unit is at 100% power l

l

An instrument technician inadvertently pulls ONE control power fuse for NI Channel 42.

l

5 seconds later, he realized the error and replaces the control power fuse.

1

) Which ONE (1) of the following describes the actions that will occu from the following actions? d.,e vs e ra m #

A. No actions will occur providing the other fuse was not pulled, i 4. . A 9 second turbine runback will occur then stop. !

C. A continuous turbine runback will occur causing a reactor trip.

l D. A limit runback will occur until the fuses are reinstalled. cre b

[ i y

un 1:l on<

2 K/A 015000K301 3.9/4.3 WiH ' f g

'/""

NI-LP pg 53-55 pp , g , % 6is 4 $IrJ Modified Question chthN g [LoSb[

g 1 a f fe e , L {'

l Question 20 of 100

. . l l 95-2 NRC EXAM - REACTOR OPERATOR

.I l.

16. NI-14 005 j Given the following plant conditions:

)

The unit is at 100% power

An instrument technician inadvertently pulls ONE control power fuse for NI Channel 42.

5 seconds later, he realized the error and replaces the control power fuse. j Which ONE (1) of the following describes the actions that will occur from the following actions?

A. No actions will occur providing the other fuse was not pulled. ;

4. A 9 second turbine runback will occur then stop.

C. A continuous turbine runback will occur causing a reactor trip.

D. A limit runback will occur until the fuses are reinstalled. ,

b-K/A 015000K301 3.9/4.3 NI-LP pg 53-55 Modified Question Question 16 of 100

l LESSON BODY KEY AIDS V

bistable.

Q. What will happen if the fuses are reinstalled within the runback period?

A. The runback will continue I

c. Runback Analysis - runback in progmss and Control Power Fuses are then removed (1) Will continue normally (have no effect)

Plant > 70% Turbine Power - Turbine Reference Runback will continue for 9 seconds until TDRX relay times out - Then it will stop f s

Turbine Limit Runback - will continue until one

( (1) of the turbine power 70% bistables clears and then the mnback will Stop (2) Explanation

The removal of the control power fuses causes all the channel output tmnsformers to deenergize, however the runback bistable is already deenergized which has already deenergized the mnback transformer

The deenergized mnback transformer has deenergized the relay that causes the runback, therefore the removal of the contrul power fuses will have no effect on the runback that is in progress (3) What will happen if the control power fuses are reinstalled while the runback is in progress?

No effect

\ /

NI Rev.1 Page 53 of 65

. *

j j

LESSON BCDY KEY AIDS O) m

When the control power fuses are reinstalled, the i channel output transformers will reenergize with !

the exception of the output transfonner for the runback circuit will remain dcenergized due to the !

signal from the runback bistable which is tripped. I 4

d. Runback Analysis - No nmback in progress and

. instrument power fuses are removed )

(1) / r.mback will be initiated

Turbine reference runback will operate for 9 seconds as explained above (Will opemte for 9 seconds from any power level) i

Tutbine limit runbad will operate as described

above

e. Runbher Analysis - No runback in progress and control power fuses are removed 4

(1) A runback will be initiated

Turbine reference runback will operate for 9

, seconds as explained above (Will operate for 9 seconds from any power level)

Turbine limit runback will operate as described 4

above O. What will happen if the control power fuses are reinstalled during the ,

runback?

I A. The runback will Stop. The runback.

stops because the channel output transformers are reengerized and the runback relay reenergizes and stope the runback 4

e i N1 Rev.1 Page 54 of 65

4 i

e LESSON BODY KEY AIDS

!O

f. Runback Analysis - Runback held in RESET and

Control Power Fuses are removed (1) A mnback will be initiated

)

The removal of the contml power fuses causes the

. channel output transformers to deenergize which

! will initiate a runback 2

l

Reinstalling the control power fuses during the

runback will stop the runback because the channel aitput transformers are reenergized when the fuse

! is reinstalled l

l g. Runback Analysis - Runback RESET and Instrument i Power Fuses are removed i,

j (1) A runback will be initiated

! G

The removal of the instrument power fuses causes lV i

the rod drop bistable to deenergize which will initiate a runback i

)

Reinstalling the instrument power fuses during the l runback will stop the runback because the rod drop i bistable will reenergize and reset automatically

} because in the RESET position the reset of the

bistable is automatic

I

!- 4 4

1

. 3. CONTROL POWER AND INSTRUMENT POWER l j FUSES WITH NIS CHANNEL IN ROD DROP BYPASS f- a. Control Power Fuses are pulled i 1

, (1) The removal of the control power fuses causes the channel output transformers to deenergize which

, will initiate a runback and the runback will not be blocked because the removal of the control power fuses will cause the runback relay to deenergize O.

N1 Rev.1 Page 55 of 65

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

21. AFW-09 001 Which one (1) of the following plant events requires the LEAST amount of Auxiliary feedwater to maintain an adequate Heat Sink as required by the EOP network?

A. ATWS event .

\

vs. Large Break Loss of Coolant \'

C. Loss of Main Feedwater -

D. Loss of All AC Power b

K/A 061000.K4.02 4.5/4.6 -h ,f hw I#"" "id p ~ J </r &

, . New Question n- ,,,i, it s 7

)

/ A = 600 gpm (FRP-S.1)

/ B= 300 gpm (Path-1) ,

/ C=Not required AOP stem asks for "EOP". AFW not used in AOP only in EOPf l

/ D = > 300 gpm (EPP-001)

! No r ele-coc 6, ,t, de-v e ~ , v . ~.s ,

vhe., }o (b a c Y"M l"'

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3 g y_of n c..c+;,,1 i=~ R CJ hed f e s alen, i

l l

Question 21 of 100 i

95-2 NRC EXAM - REACTOR OPERATOR

17. AFW-09 001 Which one (1) of the following plant events requires the LEAST amount of Auxiliary feedwater to maintain an adequate Heat Sink as required by the EOP network?

A. ATWS event vB. Large Break Loss of Coolant I l

C. Loss of Main Feedwater I D. Loss of All AC Power i

b I K/A 061000.K4.02 4.5/4.6 l New Question A = 600 gpm (FRP-S.1)

B= 300 gpm (Path-1).

C=Not required AOP stem asks for "EOP". AFW not used in AOP only in EOP D = > 300 gpm (EPP-001) l I

l I

I i

Question 17 of 100

,a

' 95 2 NRC EXAM - SENIOR REACTOR OPERATOR

'l l

22. RDCNT-05 005 l Which one (1) of the following describes the purpose of the Variable Gain Unit in the Reactor Control Unit (Automatic Rod Control System)?

3 ~ bd l A. Increase the output for large error signal to increase response. crf b I B. Decrease the output for lower power levels to compensate for higher rod worth.

C. Decrease the output for large error signal to decrease response.

i v'D. Increase the output for lower power levels to compensate for lower rod worth. l d

K/A 001000. A3.01 4.1/4.0 SD-007 New Question

't a ,, elt , >% c! vo " Eis h 9

l I

$ l Question 22 of 100 j l

LESSON BODY KEY AIDS oT%+T* c

=Tavg 2

o Median signal selector selects median signal Q: If"A" loop controls Tw fails low, what will auto rod signal do? Ifit fails high?

A: Won't change because MSS selects median signal

)

o T,,, signal is passed through a lead / lag compensation unit and a filter to increase the

, efTect of the signal and filter out noise o T,,,is compared to T urin the summing unit to give a total temperature error signal j

)

Power Mismatch Signal o Provides fast response to load changes and provides control stability I

o Q, compared to Q,, in the rate compensator l (Anticipates temp change due to power mismatch) 1 o No output for constant mismatch o Output is developed when the rate of changes of Q,, and Q, are different RDCNT Rev.I Page 14 of 47

4 i LESSON BODY KEY AIDS o If rate of change not different there is no output o Temperature unit provides fine adjustment at

, this point to bring Tave to Tref Q: If N-44 fails high, what will be the response of rod control and the overall effect on T.,, and PZR level?

4

.A: Rods will automatically insert, while N-44 moving high

and a rate of change exists. When N-44 is at full value, power mismatch circuit drops out and rods stop moving.

T,,, will drop to' a lower value and stabilize, and PZR 4 level will follow T.,,. High flux signal (103%) will

prevent rods from withdrawing to restore T,y,'

NOTE: Outward Rod motion has been disabled also.

i o Non-linear gain (K ) modifies i the Qmism.ica (power mismatch) signal to yield Toi,,,,cs 1

l 0 Tmi ,,,,c3 = (K i) Qmi,,,,cn (power mismatch)

, o K =i 0.3 F/%, when Qmi, ics 51Yo i

o K =i 1.5 F/%, when Q,,,,,,cn >1%

o Variable gain unit enables adequate low power control and stable high power operation which minimizes overshooi e Variable gain setpoint:

Qi, < 50% = 2.0 l

RDCNT Rev.1 Page 15 of 47

, r

i LESSON BODY KEY AIDS Q,, at 100% = 1.0

o The gains used above were selected high enough '

i to enhance system response yet low enough to I avoid overcompensation and oscillations ,

e Summing Unit l o Three input signals are summed here j i

i- o Total error is applied to the rod speed 5 programmer i

e Rod Speed Program TP-RDCNT-4

{ o Converts the total temperature error signal to

desired direction and rod speed signal for the logic cabinet i

o Deadband - amount ofinput signal needed before j any automatic rod motion will occur o Input signal must be outside dead band of +0.5 or -2.5 F before rod motion is called for o If only signal is temperature error from power ;

mismatch ekt, it still has to exceed deadband (ie, with T,,, = T,.r , rods would move in with a +0.5 temp error from power mismatch) o T. , must be at least 0.5*F above T,.r before automatic rod control will insert rods o Lockup (band width)is the amount of signal '

decrease necessary prior to automatic signal ;

clearing (0.5 F below/above dead band setpoint) j

. o Prevents rod cycling when limit is reached RDCNT Rev.I Page 16 of 47

95-2 NRC EXAM - SENIOR REACTOR OPERATOR 3 .

i

23. AOP-018-03 002 3 Given the following plant conditions

A failure of the #1 seal on RCP "A" has occurred.

The reactor and RCP "A" have been tripped.

i Which ONE (1) of the following is the reason for the 90 second delay in closing CVC-303A, i Seal Water Leakoff valve, after tripping RCP "A"?

h'gafY

. b s A. To avoid damaging the motor lower guide (radial) bearing. I E

9 j

[ ( B. To maintain adequate backpressure on #2 seal.

g qQ 1 5

! 3 3e ( C. To maintain #1 seal DeltaP greater than 210 psid.

QM f ( s j @. To ensure that the RCP has sufficient coastdown time come to a complete stop. l l e

i i

9 K/A 003000.K6.02 (2.7/3.1) '

AOP-18, Step 33.

Westinghouse Technical Bulletin,41171, RCP Seal Failure Response, Pg.67P.

Modified Question 1

i 4

1 Question 23 of 100 1

95-2 NRC EXAM - REACTOR OPERATOR

19. AOP-018-03 002 Given the following plant conditions:

A failure of the #1 seal on RCP "A" has occurred.

The reactor and RCP "A" have been tripped.

l Which ONE (1) of the following is the reason for the 90 second delay in closing CVC-303A, l Seal Water Leakoff valve, after tripping RCP "A"?

A. To avoid damaging the motor lower guide (radial) bearing.

l B. To maintain adequate backpressure on #2 seal.

C. To maintain #1 seal DeltaP greater than 210 psid.

l

@. To ensure that the RCP has sufficient coastdown time to come to a complete stop.

d-

\

l K/A 003000.K6.02 (2.7/3.1) l AOP-18, Step 33. ;

Westinghouse Technical Bulletin,41171, RCP Seal Failure Response, Pg.67P.

l Modified Question j

1 l-

]

i 1

! i l

l l

i I l

(

e i

Question 19 of 100 j l

, Rev. 8 AOP-018 REACTOR C001 ANT PUMP ABNORMAL CONDITIONS Page 22 of 42 4

STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED SECTION B REACTOR C001 ANT PUMP SEAL FAILURE (Page 10 of 13)

CAUTION Do N2I restart a RCP that has been tripped until the cause of the seal

-malfunction has been determined and corrected.

30. Check RCP(s) B QE C - RUNNINC Go To Step 33.

31. Check RCP B - RUNNING Place PCV-455A, PZR SPRAY 444G, Controller to MAN 6ED adjust controller output to ZERO.

32. Check RCP C - RUNNINC Perform the following:

a. Place PCV-455B, PZR SPRAY

[-s) 444H, Controller to MAN AND

\- / adjust controller output to ZERO.

b. Maintain PZR level be'; ween 30% and 40% to provide adequate PZR spray.

33. Check Time Elapsed Since EEXH at least 90 seconds have Stopping The Affected RCP(s) - elapsed since tripping the CREATER THAN 90 SECONDS affected RCP(s), IHXH Go To Step 34 34 Close Seal Leakoff Valve (s) For Affected RCP(s):

RCP VALVE A CVC-303A B CVC 303B C CVC-303C

j BASIS DOCUMENT, REACTOR COO 1 ANT PUMP ABNORMAL CbNDITIONS i

,. Section B (Continu2d) jfm Description i

j 28 When at power, tripping a Reactor Coolant Pump will result in S/G shrink as 3

a result of loss of Primary System coolant flow through the S/G. This step instructs the Operator to take manual control of the S/G 1evel(s) prior to tripping the Reactor Coolant Pump.

3 29 This step provides instruction to trip the affected RCP.

l l

C30 Tripping the affected RCP may reduce the severity of the condition that was the reason for initially tripping the pump, which might lead -the Operator to believe that the RCP could be returned to service. This caution reminds the Operptor that the RCP should not be restarted until the cause of the event is corrected.

l 30-32 Tripping RCP B or C will affcet Pressurizer Spray flow. If both RCP B and C are running and one of the pumps is tripped, the PZR Spray valve en the loop

() for the tripped pump must be closed to prevent diverting spray flow from the operating loop to the non-operating loop. These steps will provide the diagnostics and actions to assure that the idle loop spray valve is closed 1 if its pump is stopped while the other loop remains running.

33 Westinghouse requires closing #1 Seal Leakoff Valve if the RCP was tripped due to RCP #1 or #2 Seal failure. However, the valve should not be closed until the RCP has stopped rotating. This step acts as a hold point in.the procedure until the RCP has been stopped for at least 90 seconds which assures that the pump will be stopped. [Ref: Westinghouse RCP Manual 728-621-13, Par. 6.1.6.1, 6.1.6.2 )

l 1

/~'i V

AOP-018-BD Rev. 8 Page 21 of 33

95-2 NRC EXAM - SENIOR REACTOR OPERATOR 4

24. CV-09 001 Given the following plant conditions:

fe,

, ,d , p u e- f V #

Containment temperature is 122 F <

A design basis LOCA occurs obe,s s co-ll4i<-'

Which ONE (1) of the following describes the potential effect of Containment 1emperature-being-greater 4han-1-20*F at the onset of the LOCA?

A. Instrumentation inside the CV are not qualified at, initial temperatures > 120 F.

5 wecc O .vr ? -> B. CV temperature may exceed its design limit.

<. CV pressure may exceed its design limit.

j O'"~'d i

D. CV level instrumentation will be innaccurate. C o c

K/A 022000. A1.01 3.6/3.7

/

j FSAR Table 6.2.1-4

- 75 7 '

New Question {

A EF d . < >4 ) gr o H*" A 1~ I% th e o He,, u .,1 ceu.

Question 24 of 100

- .= -. .. .- - - . . - _ - . - - . _ - ~ _ . _ _ . _ _ - _ - -

95-2 NRC EXAM - REACTOR OPERATOR l

l

20. CV-09 001 l Given the following plant conditions:

Containment temperature is 122 F

A design basis LOCA occurs Which ONE (1) of the following describes the potential effect of Containment temperature being greater than 120 F at the onset of the LOCA?

A. Instrumentation inside the CV are not qualified at initial temperatures > 120 F.

B. CV temperature may exceed its design limit.

W. CV pressure may exceed its design limit.

D. CV level instrumentation will be innaccurate.

c K/A 022000.A1.01 3.6/3.7 l FSAR Table 6.2.1-4 New Question i

1 l

F Question 20 of 100

HBR 2

/ UPDATED FSAR I t l 1

Table 6.2.1-4 I Initial Conditions for the Main Steam Line Break l Containment Analysis I

Reactor Coolant and Secondary Systems HZP 102% ,

Reactor rower Level, Mwt 0 2346 Core Inlet Temperature, F 547 547 Primary Flow, lbm/hr 104,926 97.3E6 !

i AFW Flow Rate, gpm 1325 1325 I l

Steam Generator Fluid Mass, Ibm 137,294 94,503 Steam Flow Rate, lbm/hr 0 10.3E6 Steam Pressure, psia 1021 800 Containment Pressure, psia 15.7 Temperature, F 120.0 Relative Humidity, t (peak pressure) 0.0 (peak temperature) 100.0 Total Free Volume, cu. ft. 1.95E6 Outside Temperature, F 90.0 l Service Water Temperature, F 95.0 Spray Water Temperature, F 100.0 Spray Setpoint, psig 20.0 Spray Delay Time, sec. 55.0 Fan Cooler Setpoint, psig 5.0 Fan Cooler Delay Time, sse 69.0 0

6.2.1-18 Amendment No. 10

HBR 2 UPDATED FSAR The done liner, being uninsulated, feels the thermal effects of the accident O immediately and its temperature follows but lags the containment steam tempe ratu re.

The temperature of the liner is a time dependent variable for both portions.

The loads resulting f rom the thermal expansion of the liner af ter an accident a re shown on Figure 3.8.1-29.

The accidcat condition will create only a very slight increase in the concrete temperature on the insulated portion of the wall, as illustrated by Figure 3.8.1-30.

The accident condition creates a high skin temperature on the concrete behind the uninsulated liner; however, af ter 10,000 seconds, when the containment temperature and pressure are rapidly decreasing, the total depth of the wall which has felt any inc rease in temperature is about 9 in. in the 2 f t 6 in.

wall thickness. This is illustrated by Figure 3.8.1-31.

Because accident temperature rise in the conc rete is minimal, increases in temperature of the conc rete behind both the insulated and uninsulated liner were not considered.

e) Operating thermal expansion stresses caused by normal thermal gradients across the containment wall and done were analyzed. Three extreme conditions were considered.

For all three an "as constructed" temperature for the liner and concrete was assumed at 60 F. The three operating conditions analyzed were:

1) Summer operation:

(a) Operating temperature inside containment: 1200 F (b) Exterior sustained conc rete temperature: 90 F ,

,2) Winter operation (hot) .

(a) Operating temperature inside containment: 120 F (b) Exterior sustained conc rete temperature s 20 F

3) Winter operation (cold):

(a) Operating temperature inside containment: 500 F (b) Exterior sustained conc rete temperature: 20 F In all cases a straight line temperature gradient was used through the wall of the containment.

The loads resulting f rom the thermal operating conditions are shown on Figures 3.8.1-32 and 3.8.1-33.

f) Uplif t due to buoyant forces will be created by the displacement of ground water by the structure. The ground water elevation was assumed to be

( at grade level for conse rvatism. The buoyant force was computed as the weight of the displaced water assumed at a specific weight of 62.4 pcf.

3.8.1-13

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95-2 NRC EXAM - SENIOR REACTOR OPERATOR

25. CVCS-09 005 Given the following plant conditions:

7*

p i, ni l' "

PZR level has decreased to < 14%

The operator takes the control switch to close for CVC-200A, i.etdown Isolation valve, and nothing happens.

Which ONE (1) of the following events will cause this valve to close?

A. Letdown Isolation valve CVC-204A or B starts to close.

B. Letdown Isolation valve LCV-460A/B starts to close.

C. Phase B Containment Isolation.

@. Loss of Instrument Air to Containment. -

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Question 25 of 100

95-2 NRC EXAM - REACTOR OPERATOR

21. CVCS-09 005 Given the following plant conditions:

PZR level has decreased to < 14%

The operator takes the control switch to close for CVC-200A, Letdown isolation valve, and nothing happens _

Which ONE (1) of the following events will cause this valve to close? l l

A. Letdown Isolation valve CVC-204A or B starts to close.

B. Letdown isolation valve LCV-460A/B starts to close.

1 C. Phase B Containment Isolation. <

@. Loss of Instrument Air to Containment.

d 4

K/A 004000.A2.07 3.4/3.7 CVCS LP pg 13,14 New Question Question 21 of 100

i LESSON BODY KEY AIDS l O Charging ~ 130'F to 490'F @ 60 gpm l

4. Reduces heat loss to the CCW System thereby improving system thermal efficiency

5. Letdown temperature indication on RTGB (TE-140) OBJ. #7

a. LP LTDN LN HI TEMP (APP-001-B6),380*F FIGURE-13

~ Q:' Letdown flows through: the ? shel1 side or the tube side'of the regenerative heat exchanger?

A: Shel1~ side O

C. LETDOWN ORIFICES OBJ. #4,5 FIGURE-2

1. Regulates letdown flow

2. Basically a plate with a hole drilled in the center

3. Two 60 gpm orifices and one 45 gpm orifice

4. Isolated by orifice isolation valves CVC-200A (45 gpm OBJ. #4,5 orifice) and CVC-200B, C (60 gpm orifices)

a. Air operated valves OBJ. #9

b. Fail closed O. CVCS Rev.2 Page 13 of 73

~

LESSON BODY KEY AIDS O c. Shut on containment isolation signal phase A

d. Needle valves throttle instmment air to ensure that the 200 valves will open slower than CVC-204A & B, which helps PCV-145 to regulate pressure and helps prevent 4 overpressure when placing letdown in service i e. Valves may also be operated at the dedicated shutdown

panel (DSP)in the charging pump room (1) Local / remote switch for each valve (2) SHUTDOWN EQUIP IN LOCAL, APP-036-J6 FIGURE-15 alarm (3) OPEN/CLOSE switch and light for each valve (4) DSP discussed in later lesson plan D. LETDOWN ORIFICE PRESSURE RELIEF VALVES OBJ. #4,5 FIGURE-2

1. CVC-203 A & B, relieve to pressurizer relief tank

2. Set at 500 psig & 630 psig, respectively

3. Staggered setpoints prevent relief valve chattering from low reliefflowrate

4. RTD on discharge indicates valve (s) open or leaking OBJ. #9

a. LP LTDN RELIEF HI TEMP, (APP-001-E6),200*F FIGURE-12 O

CVCS Rev.2 Page 14 of 73

95-2 NRC EXAM - SENIOR REACTOR OPERATOR b26. ESF-04 001 Given the following paint conditions:

\tb e The piant is at 100% power

All systems are aligned for normal operation

(

A small (<.2 gpm) RCS leak is identified in the CV Which ONE (1) of the following describes thtrsignal which will result in a Containment Ventilation Is'olation? q VA. Low pressurizer pressure SI or R-11/12 in alarm.

B. High pressurizer level reactor trip and R-ll/12 in alarm.

C. Reactor Trip and High radiation alarm on R-14C, Plant Vent Radiation Monitor.

1 D. Manual actuation of Phase B and R-14C, Plant Vent Radiation Monitor in alarm.

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K/A 013000. A3.02 N i Logic dwg 5379-2759 sht 8 of 18 i l

New Question l

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Question 26 of 100

95-2 NRC EXAM - REACTOR OPERATOR

22. ESF-04 001 Given the following paint conditions:

The plant is at 100% power

All systems are aligned for normal operation

A small (<.2 gpm) RCS leak is identified in the CV Which ONE (1) of the following describes the signal which will result in a Containment Ventilation Isolation?

VA. Low pressurizer pressure SI or R-11/12 in alarm.

B. High pressurizer level reactor trip and R-11/12 in alarm.

C. Reactor Trip and High radiation alarm on R-14C, Plant Vent Radiation Monitor.

D. Manual actuation of Phase B and R-14C, Plant Vent Radiation Monitor in alarm.

i a l K/A 013000. A3.02 Logic dwg 5379-2759 sht 8 of 18 New Question i l

1 l

l Question 22 of 100

95-2 NRC EXAM - SENIOR REACTOR OPERATOR V

i 27. DC-09 001 Given the following plant condition /:

, e DC Bus voltage decreases to < 123 volts

( Which one (1) of the following describes the undervoltage relay response to the above

! '-eenp>

A. It automatically shifts the associated battery to the spare charger.

4

4. It actuates the BATT A/B LO VOLT annunicator for the applicable bus.

C. It trips the charger to separate the battery from the faulted supply.

1 l D. It actuates the DC GROUND annunciator for the applicable bus.

a b

4 4

K/A 063000.A4.02 2.8/2.9

! APP-036-D3,4 l

New Question 4

1 i d

Question 27 of 100

- -.~ - - - . - - . . . - . - . - - ~ ~ . . .. .. - . - - - . - - -

t

!

APP-036-D3 L AL6EH BATT A/B LO VOLT l

l AUTOMATIC ACTIONS

! 'l . None Applicable CAUSE

1. Battery Charger Misalignment

2. Battery Charger AC Supply Breaker Open

3. Battery Charger failure i OBSERVATIONS ,

-1. ERFIS Points:

1) APV3022A, DC MCC-A VOLT

2) APV3023A, DC MCC-B VOLT ,

! i

3) APV3024A, DC MCC-A CURRENT

4) APV3025A, DC MCC-B CURRENT

2. Local indications on battery chargers l

ACTIONS

1. Check Battery voltage immediately.

2. Reenergize in service battery charger or place standby battery charger in l 1ervice IAW OP-601, DC Supply System.

l 3. Initiate a Work Request to direct I&C Maintenance to perform MST-903, l

Station Battery Charger - Monthly for the Battery with low voltage (Reference 5). 1 l

DEVICE /SETPOINTS j i 1

l. Low voltage Alarm /123 VDC (Reset at 126.5 VDC) ;

i j

REFERENCES l

1. TECH SPECS 3.7.1

2. OP-601, DC Supply System

3. MST-903, Station Battery Charger - Monthly j 4. CWD, B-190628, Sheet 955, Cable M

5. Memo - APP-009-40, "A-B Battery Low Volts" Revision, Robinson File No. 5235, Serial: RNPD/91-2874 APP-036 Rev. 20' Page 30 of 91

- . - . - ~ . - _... _-

-., i e. - _ _

a APP-036-D4 .

l AIABli BATT C UNDER VOLT-AUTOMATIC ACTIONS i

1. None Applicable CAUSE

'.. Low Voltage on Battery "C" i

OBSERVATIONS

1. Local Battery and Charger indications ,

i t

AG77.Qtui ;

1. Dispatch an operator to perform the following:

1) Check Battery voltage.

2) t, heck Battery charger.

3) Verify the battery charger etH2 inverter in service IAW OP-601, DC Supply System.

DEVICE /SETPOINTS

1. Ten Voltage Alarm /123 VDC (Reset at 126.5 VDC) i i

REFERENCES OP-601, DC Supply System i 1.

2. CWD B-190628, Sheet 1345, Cable A I

l i

APP-036~ Rev. 20 p.g. 31 og 91

t s .

95-2 NRC EXAM - SENIOR REACTOR OPERATOR i 28. CVCS-09 004 Given the following plant conditions:

i j e Pressurizer level transmitter LT-459 is selected for control

.

The reference leg for LT-459 develops a slow / leak. "I* i '

Which ONE (1) of the following describes the instrumentation and plant response to this leak?

t 4

i LI-459 LI-460 I PZR LVL PZR LVL' VCT Indication Indication I evel i

b 't j

/ \ j

! VA. Increases # Dec'reases / Increases '

b B. Decreases Increa{ses increases I i

\

C. Increases / Increases \ Decreases

! D. Decreases Iner/'

D Decreases d

eases \ e l a 7

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1 Question 28 of 100 ,

_ _ . _. . _ _ _ . ~ _ . _ _ .. ..

g 95-2 NRC EXAM - REACTOR OPERATOR

~

24. CVCS-09 004 Given the following plant conditions:

i i

]

Pressurizer level transmitter LT-459 is selected for control

+ The reference leg for LT-459 develops a slow leak.

l, Which ONE (1) of the following describes the instrumentation and plant response to this leak?

LI-459 LI-460. I PZR LVL PZR LVL VCT j Indication Indication Level i

VA. Increases Decreases increases B. Decreases increases increases P

C Increases Increases Decreases D. Decreases Increases Decreases a

K/A 004000.G0.07 (3.3/3.3)

SD-021 DWG 5379-1971 Modified Question Question 24 of 100

. \

., 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

(, o e o Y o h C'

29. OP-402-06 001 r Given the following plant conditions:

?

'}

The CST and Deepwell pumps are not available fl* " /) f rv m's '

i

The Service Water system is supplying the AFW pumps

- the Which ONE (1) of the following describes limih%nd the basis for operating with Service Water aligned to the AFW pumps? ' j

?

A. 600 gpm total, prevent pump runout.

B. 900 gpm total, maximum that service water can support. 3 vW O 4 A ##$

C. 345 gpm MDAFW, adequate NPSH, '

..- -~

@., ' 325 gpm MDAFW, '

prevent tripping the pump on overcurrent.

d 4'/ 5 / Jg-f 7 K/A M1000.K1.07 (3.6/3.8)

OP-402, Pg. 31. -y pg jg

- Modified Question

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Question 29 of 100 1

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_ . _ . _ . . _ _ _ _ . _ . _ _ _ . . _ _ _ . _ ~ . _ . _ , _ _ _

l, 95-2 NRC EXAM . REACTOR OPERATOR

. 25. OP-402-06 001 - t Given the following plant conditions:

l

The CST and Deepwell pumps are not available )

i !

l

The Service Water system is supplying the AFW pumps

Which ONE (1) of the following describes limits and the basis for operating with Service -

Water aligned to the AFW pumps?

I

.A . 600 gpm total, prevent pump runout. i l B. 900 gpm total, maximum that service water can support.

C. 345 gpm MDAFW, adequate NPSH.

@. 325 gpm MDAFW, prevent tripping the pump on overcurrent.

d l K/A 061000.Kl.07 (3.6/3.8)

OP-402, Pg. 31. ;

Modified Question l

l Question 25 of 100 a

r Ssetion 8.1 Page 2 of 7 8.1.2 (Continued) INIT VEgl HOIE The following list is the maximum TOTAL allowable feed flow rates for various {

pump combinations. These flow rates when added to the 90 gpm seal leak off flow

]

and 165 gpm recire flow for the SDAFW Pump or 60 gpm recirc flow for each MDAFW pumps will prevent exceeding 600 gpa total. The Service Water System is designed to supply a maximum of 600 gpm as a backup source of water to AFW. ,

l 2 MDAFW pumps (not to exceed 325 gpa/ pump) 480 gpm SDAFW pump (only) 345 gpm 1 MDAFW pump 325 gpm (The 325 gym / pump limitation is to provent tripping the pump on overcurrent.)

3. Based on the above limitations, start AFW Pumps as follows:

1) H SDAFW Pump is to be used, IEEN perform the following:

a. Remove cap from AFW-7, SDAFW PUMP SUCTION VENT.

b. Open AFW-7.

c. }@EE a solid stream of water issues, IH._EH close AFW-7.

d. Install cap on AFW-7. *

e. Start the SDAFV Pump as follows:

a) lE desired, IEEE open Vl-8A, STEAM SHUTOFF.

b) IE desired, IHEN open Vl-88, STEAM SHUT 0FF.

c) IF desired, IEEE open V1-8C, STEAM SHUT 0FF.

(

OP-402 Rev. 39 Page 24 <

r

7 l

t Ssetien 8.1 i

Page 2 of 7 8.1.2 (Continued) H H yJJ Q E91E The following list is the maximum TOTAL allowable feed flow rates for various I pump combinations. These flow rates when added to the 90 gpa seal leak off flow and 165 gpm recire flow for the SDAW Pump or 60 gpa recirc flow for each MDAFW '

pumps will prevent exceeding 600 gpm total. The Service Water System is designed to supply a maximum of 600 gpm as a backup source of water to AW.

2 MDAW pumps (not to exceed 325 gps / pump) 480 gpm SDAW pump (only) 345 gpm 1 MDA W pump 325 gpm (The 325 gpa/ pump limitation is to prevent tripping the pump on overcurrent.)

3. Based on the above limitations, start A W Pumps as follows:

1) H SDAW Pump is to be used, HEN perform the following: :

a. Remove cap from AW-7, SDAW PUMP SUCTION VENT.

b. Open AW-7.

c. ]@EE a solid stream of water issues, HEN clo- .W - 7 . '

1

d. Install cap on AW-7. '

e. Start the SDA W Pump as follows:

a) H desired, HEN open V1-BA, STEAM SHUT 0FF.

b) H desired, H EN open Vl-8B, STEAM SHUT 0FF.

c) H desired, H EN open V1-8C, STEAM SHUTOFF. __

9 l

I OP-402 Rev. 39 Page 24 r

95-2 NRC EXAM hENh0R REACTOR OPERATOR

30. CVHVAC-04 003 Given Ihe following plant conditions:

. \

c[

The p'la,nt was at 100% power

/ \

/

A design bayis LOCA occurred f5NE Which ONE (1) f the A

o\ describes thev' following basis bf the Containment Spray System?

\

(

Designed to ......

\

A. deliver minimum requir\ed flow within 30 seconds of the initiating signal.

'\

4. deliver minimum required flow within 60 seconds of the initiating signal.

, ,Nel? C. \

prevent contamment pressure trom exceeding 20 psig.

\

D. prevent containment temperature ffom exceeding 190*F.

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_ _ _ . _ _ _ _ . _ . . - _ _ . _ _ . . _ . . _ _ . _ . . . _ . _ . . . ~ _ _ . _ . _ - _ . _ _ . _ . _ . . _ . _ . . _ _

95-2 NRC EXAM - REACTOR OPERATOR l 1

i I

26. CVHVAC-04 003 Given the following plant conditions:

The plant was at 100% power

A design basis LOCA occurred Which ONE (1) of the following describes the basis of the Containment Spray System?

Designed to ......

A. deliver minimum required flow within 30 seconds of the initiating signal.

4. deliver minimum required flow within 60 seconds of the initiating signal. !

C. prevent containment pressure from exceeding 20 psig. ;

D.' prevent containment temperature from exceeding 190*F. I l

l b ;

i K/A 022000.K3.01 2.9/3.2 !

FSAR 6.5.2.2 l New Question l

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Question 26 of 100 i.

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HBR 2 l UPDATED FSAR i I

6.5.2.2 System Design Adequate containment iodine removal capability is provided by the Containment d

Spray System shown in Figure 6.2.2-1. The components of this system are aligned !

into two subsystems. Each subsystem contains a pump, associated valving, and I spray headers independently capable of delivering one-half of the total required I flow of 2322 gpm. If one train is inoperable, the minimum delivered flow is, therefore, 1161 gpm. This system operates in two sequential modes:

a) Spray f rom the refueling water storage tank into the entire containment atmosphere using the containment spray pumps. During this mode, the contents of the spray additive tank (soditsa hydroxide) are mixed into the spray stream to !

provide adequate iodine removal from the containment atmosphere by a washing ,

action. I b) Recirculation of water f rom the containment sump is provided by the diversion of a portion of the recirculation flow from the discharge of the residual heat removal heat exchangers to the suction of the spray pumps af ter injection from the refueling water storage tank has been terminated.

The principal components of the Containment Spray System are two pumps, one spray additive tank, spray ring headers and sozzles, and the necessary piping and valves. The containment spray pumps and the spray additive tank are located in the Auxiliary Building. The spray pumps take suction directly from the l refueling water storage tank.

The Containment Spray System also utilizes the two residual heat removal piraps, two residual heat exchangers, and associated valves and piping of the Safety Injection System (SIS) for the long-term recirculation phase of containment cooling and iodine removal.

During spray injection, approximately 80 gpm of pump discharge flow is diverted f rom the spray pump discharge through the spray eductors. The liquid from the tank then mixes with the liquid entering the suction of the pumps via the !

eductors. The pH of the resulting solution is suitable for the removal o'f I iodine f rom the containment atmosphere (refer to Section 6.1.1.2).

During spray recirculation operation, the water is screened through a 7/32 in, i mesh before leaving the containment sump.

l The spray nozzles are stainless steel and have a 3/8 in. diameter orifice. The spray nozzles, of the ramp bottom design, are not subject to clogging by particles less than 1/4 in. in maximum dimension. Since particles larger than 7/32 in. in dimension are screened out of the spray recirculation flow, as ;

indicated above, the spray nozzles are ef fectively protected against clogging )

and are capable of producing a mean drop size of approximately 1000 microns in '

diameter with the spray pump operating at design conditions and the l containment at design pressure. The nozzles are connected to six ring headero i located within the dome of the Containment Building. The lowest ring header is located at Elevation 372.3 f t and the highest ring header is located at Elevation 412.1 ft. There are 116 Spraco Model 1713 nozzles distributed on the six headers. l l

)

6.5.2-2 l

I i

BR 2 i UPDATED FSAR J

$ The nossles and headers are so oriented as to ensure adequate spray coverage 4

of the containment volume. .

The procedure for the change-over from the injection mode to the recirculation j

, mode of operation is described in Section 6.2.2.

i l i All associated components, piping, structures, and power supplies of the i

! Containment Spray System are described in Section 6.2.2, with the exception of 4 f the spray additive tank and eductor, which are described in Section 6.1.1.2. )

4

6.5.2.3 Design Evaluation j 6.5.2.3.1 Theoretical Background By virtue of the large surface are, provided by the liquid droplets, the cpray system affords an excellent means of absorbing the soluble components from the i

gas phase of the fluid inside containment following a postulated accident. If j

! the solubility of the component is sufficiently high, the rate of absorption -

is limited only by the mass transfer rate of the absorbing species through the l

gas film. In the case of fission product iodine vapor, the removal process,

controlled only by gas film resistance, would permit the absorption by sprays j to proceed with a removal half-life of approximately two minutes, as will be

shown later.

l l j The removal process will be gas film limited if the vapor component has a l sufficiently high solubility in the spray drop solution, or in other terms, if the partition coefficient for the vapor component between liquid and gas j phases is high. The partition coefficient P as reported by Eggleton (Reference 6.5.2-1), is on the order of 10 5 r,an per liter liquid / gram per g

liter gas for iodine distributed between an aqueous solution at pH ~ 9 at 1000C and the vapor phase. With this value, and assuming a liquid film coefficient, V ,tof 5.25 x 10-3 cm/sec (Reference 6.5.2-2), and a gas film l coefficient, V of about 7 cm/sec (Reference 6.5.2-3), the overall mass i transfercoeffhc,ient, V T, for the iodine removal process is obtained as j follows:

1 ,

1

,#*Y 1 ,1+ 1

= 0.14 + 0.0019 L 105 5.25 x 10~3 i

i Vg = 6.9 cm/sec To obtain the advantages of the high partition coefficient which results in a j high absorption rate and nearly complete removal of 12 at equilibrium, the

! chemistry of the spray solution is modified by adding NaOH, raising the pH to approximately 9.3. According to the known behavior of elemental icdine in highly dilute solutions, the hydrolysis reaction

~

1 2 + OH" HIO + I proceeds nearly to completion (Reference 6.5.2-4) for pH > 8. The iodine form is highly soluble, and HIO readily disproportionates to 105 to the non-volatile and soluble iodate (105) forat 3HIO + HIO3 + 2HI 6.5.2-3

- 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

\

'\

31. ESF-04 002 Which one (1) of the following describes the Engineered Safeguards System signal that is desigqed specifically to provide protection for a Steam Break inside the Containment?

\

A. Lo\w Pressurizer Pressure. l

4. Hig Steamline Delta P.

\

C. High S. team Line Flow with Low Tavg.

.\

D. liigh Steam Line Flow with Low Steam Line Pressure.

\

b \

\

K/A 013000. A2.02, 4.3/4.5

\

SD-006

\

Design Basis document s 81.hnp pg 11,12,27

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3 CD 3 /2 Question 31 of 100

'Documsnt o. DBD/R8703B/SD25

,- Page No. 30, Revision 1

1.2.2.13 Reactor Safeguards and Protection System (RSPS) 1 The RSPS provides output signals to automatically close the MSIVs, as necessary, to mitigate the consequences of certain accident conditions. (Reference 6.5)

T 1.2.2.14 Steam Generator Blowdown and Wel Layup System (SGB&WLS)

I

! The SGB&WLS provides a means to drain the secondary side of the 2 SGs. (Reference 6.33, Sheet 1) i I

1 l The SGB&WLS helps protect the SGs from corrosion during normal, transient, and inactive operational periods (Reference 6.119) 1.2.2.15 Turbine and Centrols System (T&CS) )

The T&CS provides output signals to automatically open Valves DV-1, DV-2, DV-3 and DV-4, in conjunction with a turbine trip, to drain the steam supply lines to the HP Turbine.

l (References 6.132; 6.133; 6.134; and 6.135) 1.3 SPECIFIC SYSTEM TRANSIENT RESPONSE FUNCTIONS The MSS is not an ESF system; however, it does provide steam line isolation, which is a protective function initiated by the RSPS.

Steam line isolation prevents excessive cooldown of the RCS by limiting the steam release resulting from a MSLB. Excessive cooldown adversely affects RCS capability in two ways. First, the reduced temperature and pressure of the reactor coolant changes its moderator properties, resulting in increased reactivity. Second, the material integrity of the reactor I vessel and RCS piping could be adversely affected due to the resulting pressurized thermal shock.

l

l 95-2 NRC EXAM - SENIOR REACTOR OPERATOR l

31. ESF-04 002 Given the following plant conditions:

. RCS pressure is 1875 psig and decreasing

Steam Generator pressures are:

A = 400 psig and decreasing B = 980 psig

C = 970 psig

(

Tavg is 537*F and decreasing l l l

Steam Flows are:

l A = lx106lbm/hr l

! B = zero C = zero Which ONE (1) of the following ESF signals are designed to provide protection for the above l conditions? 1 1

A. Low Pressurizer Pressure. U "

l 4. High Steamline Delta P. !

C. High Steam Line Flow with Low Tavg. 3 'a v i L D. High Steam Line Flow with Low Steam Line Pressure. !

b l K/A 013000.A2.02 4.3/4.5 ggg yj g.g y , ,3 j l SD-006 3 TS Basis pg 3.5-4 i

New Question 4

i Question 31 of 100 4

, 95-2 NRC EXAM - REACTOR OPERATOR 4

31. ESF-04 002 Given the following plant conditions:

RCS pressure is 1875 psig and decreasing i i

Steam Generator pressures are: I

I

] A = 400 psig and decreasing

. B = 980 psig i

C = 970 psig ;

Tavg is 537 F and decreasing

Steam Flows are:

A = lx106lbm/hr B = zero C = zero Which ONE (1) of the following ESF signals are designed to provide protection for the above conditions?

A. Low Pressurizer Pressure.

vB. High Steamline Delta P.

C. High Steam Line Flow with Low Tavg.

D. High Steam Line Flow with Low Steam Line Pressure.

b K/A 013000.A2.02 4.3/4.5 SD-006 TS Basis pg 3.5-4 New Question Question 31 of 100

! IAE15.

Operational Safety Instrumentation Instrumentation has been provided to sense accident conditions and to initiate operation of the Engineered Safety Features."3 4 Safety Iniection System Actuation -

! Protection against a Loss-of-Coolant or Steam Break accident is brought about i

by automatic actuation of the Safety Injection System which provides emergency i cooling and reduction of reactivity.

The Loss-of-Coolant Accident is characterized by depressurization of the

Reactor Coolant System and rapid loss of reactor coolant to the containment.

The Engineered Safety Faatures have been designed to sense these effects of j the Loss-of-Coolant AcciJent by detecting low pressurizer pressure and

generate signals actuating the SIS active phase. -

1

! The SIS active phase is also actuated by a high containment pres ~sure signal

{ (Hi-Level) brought about by less of high enthalpy coolant ~ to.the containment.

i This actuation signal acts as a backup to the low pressurizer pressure signal l

~

actuation of the SIS and also adds diversity to protection against loss of cool ant.

Signals are also provided to actuate the SIS upon sensing the effects of a steam line break accident. Therefore, SIS ~ actuation following a steam line break is designed to occur upon sensing high differential steam pressure

~

between the steam header and steam generator line or upon sensing high steam line flow in coincidence with low reactor coolant average temperature or low steam line pressure. -

The increase in'the extraction of RCS heat following a steam line break results in reactor coolant temperature and pressure reduction. For this reason, protection against a steam line break accident is also provided by low pressurizer pressure signals actuating safety injection.

l l

1 l

i l

~

3.5-4 Amendment No. 85

Protecticn is also provided f[r a stsam lin: bre::k in the c:ntain ent by actuatien of SIS up:n high etntainment pressure.

SIS actuation injects highly borated fluid into the Reactor Coolant System in order to counter the reactivity insertion brought about by cooldown of the reactor coolant which occurs during a steam line break accident.

Containment Soray - -

. l TheEngineeredSafetyFeaturesalsoinitiatecontainmentsprayup$niensinga high containment pressure siCnal (Hi-Hi Level). The containment spray acts to reduce containment pressure in the event of a loss of coolant or steam line break accident inside the containment, in order to reduce containment pressure. The containment spray cools the containment directly and limits'the release of fission products by absorbing iodine should it be released to the containment.

Containment spray is designed to be actuated at a higher containment pressure (approximately 50% of design containment pressure) than the SIS (10% of design). Since spurious actuation of containment spray is to be. avoided, it is initiated only on coincidence of Hi-Hi Level containment pressure sen2ed by l both of the two sets of containment pressure signals provided for its i actuation. -

Steam Line Isolation i

Steam line isolation signals are initiated-by the Engineered Safety Features closing all steam line stop valves. In the event of a steam line break, this action prevents continuous, uncontrolled steam release from more than one '

steam generator by isolating the steam lines on high containment pressure (Hi-Hi-Level) or high steam line flow. Protection is afforded for breaks inside or outside the gontainment even when it is assumed that there is a single failure in the steam line isolation system.

r .

t

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3.5-5 Amendment No. 85

Feedwater Line Isolation

The feedwater lines are isolated up: n actuation of the Safety Injection System in order to prevent excessive ecoldown of the reactor coolant system. This mitigates the effects of an accident such as a steam break which, in itself,

causes excessive coolant temperature cooldown.

' Feedwater line isolation also reduces the consequences of a steam line break .

inside the containment, by stopping the entry of feedwater. -

Se'ttina limits

a. The Hi-Level containment pressure limit is set at about 10% of design containment pressure. Initiation of Safety Inj'ection protects against Loss-of-Coolant:23 or steam line break ' accidents 4

as discussed in the safety analysis.

.b. The Hi-Hi Level containment pressure limit is set at about 50% of design containment pressure. Initiation of Containment Spra i Steam Line Isolation protects against large Loss-of-toolant'y or and steam line break accidents,'* as discussed in'the safety analysis.

c. The pressurizer low pressure limit is set substantially below 1 system operating pressure limits. However it is sufficiently high

to protect against a Loss-of-Coolant Accident as shown in the safety analysis.: : ,_

d. The steam line high differential pressure limit is set in the event of a large steam line break accident, as shown in the safety analysi s . ,

e. The high steam line flow limit is set at approximately 40% of the steam flow from no load to 20% and at 110% of full steam flow at full load, with the steam flow differential pressure measurement linearly programmed between i .

5

~

3.5-6 Amendment No. 85

7 - - .-

20% leed tnd 1005 lord in crder to protect against large steam line break accidents.'" The coincident low T setting limit for SIS and st:ar.i line isolation initiaticn is se,t below its hot shutdown value. The coincident steam line pressure setting limit is set below the full load operating pressure. The safety analysis shows that these settings provide protection in the event of a large steam line break.*

Instrument Operatina Conditions -

During plant operations, the complete instrumentation systems will normally be

. in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not comprised, however, by continuing operation

, with certain instrumentation channels out of service since provisions were i

made for this in the plant design. This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with ' sufficient redundancy j to provide the capability for channel calibration and test at power.

Exceptions are tackup channels such as reactor coolant pump breakers. The removal of one trip channel on process control equipment is accomplished by placing that channel bistable in a tripped mode; e.g., a two-out-of-three circuit becomes a one-out-of-two circuit. The nuclear instrumentation system channels are not intentionally placed in i tripped mode since the test signal is superimposed on the normal detector signal to test at power. Testing of the NIS power range channel requires (a) bypassing the Dropped Rod protection from NIS, for the channel being tested, (b) defeating the AT/T protection CHANNEL SET that is being fed from the NIS channel, and (c) defeiting the i power mismatch section of T , control channels when the : w ropriate NIS channel is being tested. However, the Rod Position System and remaining NIS channels still provide the dropped-rod protection. Testing does not trip the system unless a trip condition exists in a concurrent channel.

~

3.5-7 Amendment No. 85

Document No. DBD/R87038/SD25 Page No. 30, Revision 1 1.2.2.13 Reactor Safeguards and Protection System (RSPS) j 1

The RSPS provides output signals to automatically close the j MSIVs, as recessary, to mitigate the consequences of certain '

I accident conditions. (Reference 6.5) )

1.2.2.14 Steam Generator Blowdown and Wet Layup System (SGB&WLS) l The SGB&WLS provides a means to drain the secondary side of the SGs. (Reference 6.33, Sheet 1)

The SGB&WLS helps protect the SGs from corrosion during normal,.

transient, and inactive operational periods. (Reference 6.119) 1.2.2.15 Turbine and controls System (T&cs)

The T&CS provides output signals to automatically open Valves DV-1, DV-2, DV-3 and DV-4, in conjunction with a turbine trip, ;

to drain the steam supply lines to the HP Turbine.

(References 6.132; 6.133; 6.134; and 6.135) 1.3 SPECIFIC SYSTEM TRANSIENT RESPONSE FUNCTIONS The MSS is not an ESF system; however, it does provide steam )

line isolation, which is a protective function initiated by the RSPS.

1 Steam line isolation prevents excessive cooldown of the RCS by limiting the steam release resulting from a MSLB. Excessive cooldown adversely affects RCS capability in two ways. First,

'the reduced temperature and pressure of the reactor coolant changes its moderatot properties, resulting in increased reactivity. Second, the material integrity of the reactor vessel and RCS piping could be adversely affected due to the resulting pressurized thermal shock.

4 l

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! 4 4 l 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

32. NI-10 003 Given the following plant conditions:

. The unit is Cold Shutdown t

l

N-3' counts = 100 cps ,

1

! l

N-32 counts = 105 cps j e The RO hears a significant increase in audio count rate and observes N-31 at 400 cps .

g>* I l Which ONE (1) of the following describes the plant response to this increase in count rate?

I l p.w m-d ,

A. Reactor trip breakers open iftlosed7 B. Initiates Phase "A" isolation.

l {. Containment evacuation alarm.

l D. Isolate Containment purge system. ,

A v4 e

c I

K/A 015000.A3.02 3.7/3.9 NI-LP-2, Rev. 5, NI-TP-2.7. Obj. 3.

APP-005-Cl l Modified Question l

l i

s.

i Question 32 of 100 i

l i

95-2 NRC EXAM - REACTOR OPERATOR

32. NI-10 003.

Given the following plant conditions:

! 4 l

The unit is Cold Shutdown l

I l

l

N-31 counts = 100 cps

e N-32 counts = 105 cps

The RO hears a significant increase in audio count rate and observes N-31 at 400 cps l Which ONE (1) of the following describes the plant response to this increase in count rate? )'

l A. Reactor trip breakers open if closed.

I l

B. Initiates Phase "A" isolation.

{. Containment evacuation alarm. l l

D. Isolate Containment purge system.

l c

l K/A 015000.A3.02 3.7/3.9 NI-LP-2, Rev. 5, NI-TP-2.7. Obj. 3.

APP-005-Cl Modified Question l

l i i l

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Question 32 of 100 .

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- i APP 005-Cl Pags 1 of 2 i

i ALARM !

i SR HI FLUX AT SHUTDOWN AUTOMATIC ACTIONS 2

1. Alarm in Containment Vessel ,

. CAUSE

1. Decrease in RCS Boron Concentration l 2. Xenon Decay i

1

3. Reactor Coolant Temp. changes 3 4. Failure to block prior to startup

5. Electrically' induced " noise" i

i OBSERVATIONS

1. Source Range NI ,

i ACTIONS

1. IE a valid High Flux level is indicated, IHEN evacuate Containment.

I

2. Insert any Rods withdrawn.

i. 3. Stop fuel movements. I

! 4 Borate the RCS- I

, 5. Block alarm.

6. IE EQI due to known core reactivity change, IEEE notify Reactor

! Engineering to evaluate the potential need to perform EST-001. (ACR 93-00198)

I DEVICE /SETPOINTS

1. N-31,~ or N-32/ 3 times the Shutdown Countrate with all rods inserted.

(CR 95-00294)

POSSIBLE PLANT EFFECTS

1. Approach to Criticality

. 2. Possible entry into Tech. Spec LCO Action l

APP-005 Rev. 14 Page 15 of 42

o 95-2 NRC EXAM - SENIOR REACTOR OPIRATOR

33. RDCNT-05 Out:

WHICH ONE of the following is NOT reset by the Rod Control System Startup Push )

Button?

A. Bank Overlap Counter. l B. Group Step Counters.

C. Slave Cycler Counters.

4. Individual Rod Position Indication. l 1

d K/A 001000K403 (3.5/3.8)

SD-007, Pg. I1,12.

RDCNT, Obj. 9, Pg. 8 & 20.

1

( va W

Question 33 of 100 i

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l LESSON BODY KEY AIDS

m. Rod Drive Mechanism

n. Auto Rod Defeat Pushbutton l

II. COMPONENT DESCRIPTION Show Items on Mock-up l

A. SYSTEM ARRANGEMENT / SIMPLIFIED 1 DRAWING l

i

1. CONTROL PANEL l

a. Contains the following items: OBJ. #4,5,7 1

i (1) Step Counters o Displays demand signal o Three digit o Add-subtract o 12 total o 4 shutdowns 0 .8 control groups o Thumb wheel (2) In-out lamps e Indicate direction of called-for motion via manual / automatic (3) Startup Push Button e Used to reset the following:

RDCNT Rev.1 Page 8 of 47

. .. .. ~. - - - . _ . . - - . -. -. .. . .

Y LESSON BODY KEY AIDS o Group step counters (RTGB) '

a l o Master cycler counter (Logic) o All slave cycler counters (Logic) o Bank overlap counter (Logic) i o All internal memories and alsrms (urgent)

(Logic) o Pulse-to-analog converters (RPI)

(4) Alarm Reset Push Button e Resets urgent alarms only ;

e Located RTGB OBJ. #7 4

(5) Demand Speed Indicator 1

e The current command signal from Tavg control or

any preset potentiometers is displayed in

, steps / minute of demanded rod speed 4

(6) Auto Rod Defeat Pushbutton s Prevents auto rod movement when moving the rod bank selector switch through the AUTO position 2.RCCA OBJ. #4.5

a. The control rod is actually a rod control cluster component assembly (RCCA) 1 1

(1) The drive rod (144 inch) can be raised or lowered by energizing three electromagnetic jacks (coils)

(2) The 45 RCCA are divided into four control banks and two shutdown banks RDCNT Rev.I Page 9 of 47

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95-2 NRC EXAM - SENIOR REACTOR OPERATOR !

3

34. FMP-007-06 001 Given the following plant conditions: l Data for the power range nuclear instruments is as follows:

d I

N41 N42 N43 N44 l i

l Upper actual reading 65 mA 69 mA 70 mA 67 mA !

l Upper 100% current 104 mA 112 mA 112 mA 108 mA ;

J

Lower actual reading 63 mA 68 mA 66 mA 64 mA I

( Lower 100% current 1% mA 110 mA 112 mA 108 mA j 1

j Which ONE (1) of the following is the quadrant power tilt ratio?

4 l

t i i A. 1.0151 I f

B. 1.0213 1-

4. 1.0275 :

i l

D. 1.0312 I

c

, K/A 015020.AK5.08 (2.9/3.4) 2 FMP-007, " Quadrant Power Tilt", ATT. 7.1.

j' Modified Question 1.0317

. 1.0147 Av yJ4e o tu-,o- is<**<-/,o,

{ $1.0606 1.0468 c~

1.03845 avg ='l.0213h [ + edc/ f

'J c17 'M twam l

I I

Question 34 of 100 i

~E4 *-4,- .4, 14--45-4--Am M +ma- -.M J4-Ju4J.d.-----a 4 94se4 J.--4. --4m-----a+ ha hl A 95-2 NRC EXAM - REACTOR OPERATOR

28. FMP-007-06 001 Given the following plant conditions: ,

Data for the power range nuclear instruments is as follows:

N41 N42 N43 N44 Upper actual reading 65 mA 69 mA 70 mA 67 mA Upper 100% current 104 mA 112 mA 112 mA 108 mA Lower actual reading 63 mA 68 mA 66 mA 64 mA Lower 100% current 106 mA 110 mA 112 mA 108 mA i

Which ONE (1) of the following is the quadrant power tilt ratio? j I

A. 1.0151 B. 1.0213. l l

W. 1.0275 l D. 1.0312 e

i K/A 015020. AK5.08 (2.9/3.4)

FMP-007, " Quadrant Power Tilt", ATT. 7.1, Modified Question 1.0317 1.0147 1.0606 1.0468 1.03845 avg = 1.0213 ;

I Question 28 of 100

\ >

. . - . ~. -. _ .

. . ATTACHMENT 7.1 Paga 1 cf 1 OUADRANT POWER TILT This revision is the latest revision available and has been verified against the Revision Status List.

(Print)

Name Signature Date Date:

CO: 07-19 SS:

CO: 19-07 SS:

DETECTOR CURRENT NOPJOJIZED FRACTION QUADRANT POWER TIME CHANNEL TILT RATIO UPPER IDWER UPPER IDVER N41 UPPER N42 IAWER N43 POWER COMMENTS N44 AVERAGE

~

N41 UPPER N42 LOWER ]

I N43 POWER COMMENTS N44 AVERAGE

~~

N41 UPPER 1

N42 IDWER j N43 POWER COMMENTS ;

N44 AVERAGE N41 UPPER N42 IDWER N43 POWER COMMENTS N44 AVERAGE l l

FMP-007 Rev. 5 Page 11 of 11

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95-2 NRC EXAM - SENIOR REACTOR OPERATOR

35. CCM-14 002 Given the following plant conditions:

i e The plant is in Hot Shutdown vMh bd *

"B" RCP is running i'

"A" and "C" RCP's are secured Which ONE (1) of the following represents the expected Reactor Level Monitoring indications when only the "A" RCP is running? pm]

A. Dynamic Head = 64%, Full range level = "RCP ON"

4. Dynamic Head = 43%, Full range level = "RCP ON" C. Dynamic Head = 64%, Full range leve'. = 120M Why llo D. Dynamic Head = 43%, Full range level = '120 l05 b

K/A 017020K601 (2.7/3.0)

OP-307, pg 10 New Question Question 35 of 100 1

_ . _ , . - . _ _ _ . - . _ _ _ _ _ . _ _ ~ _ _ . - _ _ . - . _ _ . _ . _ _ . _ _ . _ . _ _ _ . _ _ _ . . _ . . - _ _ _ .

i t 95-2 NRC EXAM - REACTOR OPERATOR I 4 9 4

29. ICCM-14 002 3 Given the following plant conditions: e i'

L

The plant is in Hot Shutdown i ,

j e "B" RCP is runnmg :

) i l- * "A" and "C" RCP's are secured ,

Which ONE (1) of the following represents the expected Reactor Level Monitoring indications when only the "A" RCP is running?

1 A. Dynamic Head = 64%, Full range level = "RCP ON" "

4. Dynamic Head = 43%, Full range level = "RCP ON"

]

. C. Dynamic Head = 64%, Full range level = 120% l l

i D. Dynamic Head = 43%, Full range level = 120% -

b K/A 017020K601 (2.7/3.0) j OP-307, pg 10 i New Question i

i i

Question 29 of 100 1

6.1.2.2 (Continued)

(3

\ l v

EXPECTED RVLIS INDICATIONS FOR FULL VESSA Plant Conditions Full Range Upper Range Dynamic Head (Heatun) LI-511AB (BB) LT-511AA (BA) LI-511AC (BC),

No RCP running 108% 108% All RCPs Off 1 RCP running RCP On RCP On 43%"

2 RCPs running RCP On RCP On 64%

3 RCPs running RCP On RCP On 102%

Full Power RCP On RCP On 110%

3. RVLIS Sensor Display Pace KVLIS Sensor Display Page shows a general layout of the Reactor Vessel Level Instrumentation System. The display also shows the condition of the hydraulic isolators and the current RCS

-s temperature and pressure.

\_s/

4. RVLIS Trend Pare RVLIS Trend Page maintains a record of vessel level conditions for the preceding 30-minute period. A trend is displayed for both Full Range and Dynamic Head indications. l 6.2 Core Exit Thermocoucle/Subcooline Monitor i

i l

6.2.1 Initial Conditions l

1. The Inadequate Core Cooling Monitor has been placed in service in accordance with Section 5.0.

r"~N w- Y OP-307 Rev. 6 Page 10 of 17

. + ,

95-2 NRC EXAM - SENIOR REACTOR OPERATOR 4

36. OP-101-05 001

/ iven G the following plant conditions:

g

FRP-P.1 has been entered

[.

Conditions for starting a RCP are being verified i Which ONE (1) of the following describes the basis for maintaining a MINIMUM of 210 psid l on the RCP seals during RCP startup and operation? I l

l I

VA. Ensures the #1 seal has proper separation between surfaces.

l

- I Prevents the #13Ef4eal from becoming a " floating" seal. 1 W'

B.

, l 4 l

. C. Ensures adequate backpressure is maintained to the #2 RCP seal, 1 3

l D. Prevents the loss of adequate seal cooling flow from the RCS. 7

-o ;

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K/A 003000A109 (2.8/2.8)

OP-101, RCS and RCP Startup and Operation, Step 4.2.1.9. )

Modified Question 1 Is(f d j dog,,'r gF E

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Qu'.:stion 36 of 100 i

l j 95-2.NRC EXAM - REACTOR OPERATOR

30. OP-101-05 001

, . Given the following plant conditions: i i a FRP-P.1 has been entered i

Conditions for starting a RCP are being verified

!. Which ONE (1) of the following describes the basis for maintaining a MINIMUM of 210 psid

. on the RCP seals during RCP startup and operation?

l f,- VA. Ensures the #1 seal has proper separation between surfaces.

B. Prevents the #1 RCP seal from becoming a " floating" seal.

t

] C. Ensures adequate backpressure is maintained to the #2 RCP seal.

D. Prevents the loss of adequate seal cooling flow from the RCS.

I j a l

.f i

KiA 003000A109 (2.8/2.8)

OP-101, RCS and RCP Startup and Operation, Step 4.2.1.9.

} Modified Question l 1

J 4

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i 1

a 1

Question 30 of 100

)

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. - - . _ - . - . - - . - - . . .. - - - . _ - ~ ~ . . . . . - . . - . . . - - . - . . . . . . . . - - . . - - . . _

I i

4.2.1 -(Continued) ]

7. The RCPs ma,'; be operated without seal water provided either of the following criteria is met:

~

1) Reactor Coolant' temperature is less than 150*F.

j i +

e

2) .RCP seal leakage-rate is 5 gpm or less and at least 25 gpm of I -

Component Cooling Water at an inlet temperature less than i-t- 115'F is flowing through the Thermal Barrier Cooling Coil. i i

i a

8. Injection Water inlet temperature to an RCP should not exceed j

[. 130*F. .

I l

9. A differential pressure of greater than 210 psid should be i i

a '

j- maintained across No. 1 seal. This condition should be met if i- the RCS pressure is at least 325 psig, however,.RCP operation may continue if RCS pressure drops below 325 psig after pump start as j i long as No. 1 seal dp of 210 paid is maintained. If decrease of ;

i No. 1 seal dp to below 210 paid-is' expected, the RCP should be i tripped'as soon as practical to minimize' coast down time with

[ 1ess than 210 psid. -

l 10. The No. 1 Seal Bypass Valve is used when RCS pressure is less

- than 1000 psig, to prevent the RCP pump bearing temperature and l the No. 1 seal leakoff temperature from reaching alarm levels.

Prior to opening CVC-307, PRI SEAL BYP ISO, the following conditions shall all be satisfied:

i

1) RCS pressure is between 100 and 1000 psig.

2) All three No. 1 Seal Leakoff valves (CVC-303A, B, C) are a !

open.

3) Any No. 1 seal leakoff flow rate is less than 1 gym.

Seal injection flow rate to each RCP is greater than o gpm, 4)

OP-101- Rev. 30 Page 13 of 65

.. ~ . . . . _ ~ - . - - . _ . - . - _ _ . - . - = . . - - - _ . - -

1

. j 95-2 NRC EXAM - SENIOR REACTOR OPERATOR l l

! 37. CCW-14 003

~

/Given the following plant conditions:

/} r>"^ I grDLY

">c n n

!

The unit is at 100% power g h.) L e "A" EDG output breaker is closed for testing Which ONE (1) of the following describes the AUTOMATIC pump start that veill be I PREVENTED when the "A" Diesel Generator output breaker is closed?

! A. CCW pump "C" on low header pressure.

4. CCW Pump "B" on low header pressure.

4

.C. SD Aux Feed Water Pump "A" from low S/G level.

! D. MD Aux Feed Water Pump "A" from low S/G level.

. b pa pels ed l

l K/A 064000.K1.01 (4.1/4.4) ;

i

' Dwg 5379-3368 sht 13 of 18 l

^f F' '

OP-604 P&L

Modified Question c or ty j (oM l (*? > C J 'Ins 1 p E n ctd h f)O REE l r ,

i !

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l Question 37 of 100 l 1

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- . - - - - . . . - . - - . - . - . - . - . - . - - . - . - - . - _ . ~ . - . . - - . _ - -

p I

95-2 NRC EXAM - REACTOR OPERATOR 1>

i~ ;

1

37. CCW-14 003 I

! Given the following plant conditions:

j

The unit is at 100% power l * "A" EDG output breaker is closed for testing I

l j Which ONE (1) of the following describes the AUTOMATIC pump start that will be l PREVENTED when the "A" Diesel Generator output breaker is closed?

1 i

A. CCW pump "C" on low header pressure. ;

! 4. CCW Pump "B" on low header pressure.

1

i. C.. SD Aux Feed Water Pump "A" from low S/G level.

1

~ i l

i D. ' MD Aux Feed Water Pump "A" from low S/G level. ,

i b K/A 064000.K1.01 (4.1/4.4) .

. Dwg 5379-3368 sht 13 of 18 OP-604 P&L- l Modified Question l

)

1 I I L

4 Question 37 of 100

. _ _ . . _ . . _ .. ..__m __ .._.._._ _. ._ . _ . _ - . . . . . _ . . . _ . . _.. _ ._

is

95-2 NRC EXAM - SENIOR REACTOR OPERATOR I

j 38. AFW-14 002 q ,, c

).

Which ONE (1) of the following. describes the-loads lost if power was removed from MCC-6?

i A. RC-536, PZR PORV Block Valve; VI-8B, SDAFW Pump Steam Isolation.

B. RC-536, PZR PORV Block Valve; VI-8A, SDAFW Pump Steam Isolation.

I C. RC-535, PZR PORV Block Valve; VI-8A, SDAFW Pump Steam Isolation.

VD. RC-535, PZR PORV Block Valve; VI-8B, SDAFW Pump Steam Isolation.

d K/A 062000.K2.01 (3.3/3.4)

EDP-003, Pg 32 Modified Question R ;> c oss e J.

Question 38 of 100

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_ _ ._..__ _ . _ _ . _ . _ _ _ _ _ _ _ _ . . _ _ _ . _ _ _ _ . _ _ . . _ . - _ . . _ _ _ . . = . _ . . _ . _

95-2 NRC EXAM - REACTOR OPERATOR i

! 38. AFW-14 002 Which ONE (1) of the following describes the loads lost if power was removed from MCC-6? :

l A. RC-536, PZR PORV Block Valve; Vi-88, SDAFW Pump Steam isolation. :

i

B. RC-536, PZR PORV Block Valve; VI-8A, SDAFW Pump Steam Isolation.

C. RC-535, PZR PORV Block Valve; VI-8A, SDAFW Pump Steam Isolation.

l v'D . RC-535, PZR PORV Block Valve; VI-8B, SDAFW Pump Steam isolation. !

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K/A 062000.K2.01 (3.3/3.4)

EDP-003, Pg 32

Modified Question <

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l Question 38 of 100 l

7.0 MCC-6 Section 7.0 Page 1 of 7 Q' MCC-6

POWER SUPPLY: 480V BUS E-2 (52/23C) LOCATION: E-1/E-2 ROOM CMPT LOAD TITLE CWD BKR NO. LOAD EDBS TAG NO. NO. EDBS NO.

1M CURRENT LIMITING REACTOR 1188 N/A CLRX/MCC-6 2BL FEED TO MCC-9 N/A 52/MCC-6(2BL)

MCC-9 2BR EDG B AUX PANEL 950 52/MCC-6(2BR)

DG-B-AUX-PNL 2D MDAFW PUMP ROOM RECIRC FAN, HVH-7A 555 52/MCC-6(2D)

HVH-7A 4 2FL FEED TO INSTRUMENT BUS 4 N/A 52/MCC-6(2FL)

INST-4 2FR TEED TO St LOAD SHED RELAY 1371 52/MCC-6(2FR SI23X1

, p 2M 30 KVA TRANSFORMER FOR MCC-9 N/A N/A V VT/MCC-9 i 3B EDG B AIR COMPRESSOR 951 s2/MCC.6(3B)

DG-B-AIR-CMP 3D EDG B ROOM SUPPLY FAN, HVS-5 558 52/MCC-6(3D)

Hvs-5 3G INSTRUMENT AIR COMPRESSOR B 588 52/MCC-6(3G)

IA-CMP-B 3M AUX BUILDING EXHAUST FAN, HVE-2B 541 52/MCC-6(3M) l HVE-2B 4B N/A 52/MCC-6(48)

BLANK (PARTIAL)

L N/A a

4DL 0.30 KVA TRANSFORMER FOR H'/S-5 558 N/A CPT/HVS 3 _

4DR 0.30 KVA TRANSFORMER FOR HVE-17 560 N/A CPT/HVE-17 EDP-003 Rev. 6 Pagn 26 of 72

I l l l l l, 7.0 (Continuad) Section 7.0 l Page 2 of 7 iO j MCC-6 l POWER SUPPLY: 480V BUS E-2 (52/23C) LOCATION: E-1/E-2 ROOM l CMPT LOAD TITLE CWD BKR l NO. LOAD EDBS TAG NO. NO. EDBS NO.

4F BLANK N/A N/A l

l N/A _

4G N/A N/A

{ PLANK djA l.

! 4M CONTROL ROD DRIVE MECHANISM COOLING 517 52/MCC-6(4M)

! FAN, HVH-5B l HVH-5B j 5B BORIC ACID TANK B HEATERS 189 52/MCC-6(55)

BA-TNK-B-HTR-A. B i i

j 5D N/A 52/MCC-6(5D)

BLANK (PARTIAL) ~

l N/A j 5F SI PUMP ROOM RECIRC FAN, HVH-6B 552 52/Mcc 6(5F) i HVH-6B i J

t j SH AUX BUILDING EXHAUST CHARCOAL FILTER 543 52/MCC-6(5H) l BOOSTER FAN, HVE-5B f HVE-5B i

j SK SPARE N/A 52/MCC-6(SK) l N/A 4

! SM SPARE N/A 52/MCC-6(5M) l N/A ,

l .

j. 6BL SPARE N/A 52/Mcc-6(6BL) i N/A 6BR SPARE N/A 52/MCC-6(6BR[

N/A 6D RHR PUMP A PIT RECIRC FAN, HVH-8B 554 52/MCC-6(6D)

HVH-8B ,

a 6E BLANK N/A N/A -

N/A O

.EDP 003 Rev. 6 Page 27 of 72

7.0 (Continu2d) Sectf.on 7.0 ;

Page 3 of 7 '

Oi U MCC-6 jPOWER SdPPLY: LOCATION: E-1/E-2 ROOM 480V

. -BUS

- _ , E _2 (52/23C)

CMPT LOAD. TITLE CVD BKR NO. LOAD ED.BS TAG NO. NO. EDBS NO.

6F BLANK N/A N/A N/A 6J RHR-7528, RHR PUMP B SUCTION 211 52/MCC-6(6J)

RHR-752B

~

6M l BLANK N/A N/A N/A 7C 839A 52/MCC-6(7C) 1 V6-12D SERVICE WATER DISCHARGE (NORMAL POWER)

V6-12D

~

7F EDG B ROOM EXHAUST FAN, HVE-17 560 52/MCC-6(7F)

HVE-17 7J RC-535, PRESSURIZER PORV PCV-456 BLOCK 121 52/MCC-6(7J)

RC 535

( 7K 0.30 KVA TRANSFORMER FOR SI LOAD SHED 1371 N/A RELAY CPT/SI23X1 7ML FEED TO PF 48 N/A 52/MCC-6(7ML) ,

PP.48 7MR SI LOAD SHED RELAY 1371 N/A SI23X1 8c V6-12C, SW PUMP DISCHARGE HEADER NORTH 838 52/MCC-6(8C)

CROSS CONNECT V6-12C 8F FCV-626, RCP THERMAL BARRIER OUTLET 234 52/Mc':- 6 ( 8F)

ISOLATION FCV-626 8J RC-536, PRESSURIZER PORV PCV-455C BLOCK 122 52/MCC-6(8J) < .

RC-536 O

EDP-003 Rev. 6 Page 28 of 72

f 4

i 7.0 (continued) Section 7.0 q Page 4 of 7 (D

's s' l

MCC-6 i i POWER SUPPLY: 480V BUS E-2 (52/23c) 14 CATION: E-1/E-2 ROOM CMPT LOAD TITLE CWD BKR NO. LOAD EDBS TAG NO. NO. EDBS NO.

an - _

8M RHR-751, RHR PUMP SUCTION FROM RCS 213 52/MCc-6(8M)

RRR-751 9c BOF ACID TRANSFER PUMP B 192 52/Mcc-6(9C)

BA - X1 1:R - PMP - B 9F CC-716B, RCP COOLING WATER INLET 232 52/Mcc-6(9F)

ISOLATION CC-716B 9J CC-730, RCP BEARING COOLING WATER OUTLET 233 52/Mcc.6(9J)

ISOLATION cc-730 9M SI-866A, S1 PUMP DISCHARGE HOT LEG 241 52/MCC-6(9M)

INJECTION

. SI-866A

( ,i 10A BLANK N/A N/A N/A loc SDAFW PUMP AUX OIL PUMP 634 52/Mcc-6(10c)

SDAFWP-0IL-PMP 10F CC-7498, RHR HEAT EXCHANGER B COOLING 219 52/Mcc-6(10F)

WATER OUTLET CC-749B 10J SI-865B, ACCUMULATOR B DISCHARGE 283 52/Mcc-6(10J)

SI-865B 10M SI-8678, BIT INLET 244 52/MCC-6(10M)

SI-867B lic N/A 52/MCC-6(11C)

SLANK (PARTIAL)

N/A 11F Sl-880B, CV SPRAY PUMP A DISCHARGE 289 52/Mcc-6(11F)

SI-88CB T

I y.) .

EDP-003 Rev. 6 Page 29 of 72

7.0 (continuad) Saction 7.0 Page 5 of 7 l U MCC-6 POWER SUPPLY: 480V BUS E-2 (52/23C) LOCATION: E-1/E-2 ROOM l CMPT LOAD TITLE cWD BKR NO. LOAD EDBS TAG NO. NO. EDBS NO.

11J SI-8643, RWST DISCHARGE 236 52/Mcc-6(11J)

' l SI-864B 1 11M SI-8628, RHR LOOP RWST ISOLATION 249 52/Mcc-6(11M) l SI-862B '

12c 31-8448, CV SPRAY PUMP B SUCTION 294 52/Mcc-6(12c) ,

SI-844B l 12F SI-8800, CV SPRAY PUMP B DISCHARGE 292 52/Mcc-6(12F)

SI-880D ,

1 12J RHR-7448, RHR LOOP TO RCS COLD LEG 221 52/Mcc-6(12J)

RHR-744B 12M SI-863B, RHR PUMP B DISCH.ARGE TO St PUMPS 281 52/Mcc-6(12M)

SUCTION

. SI-863B

() 13c RHR-7598, RHR HEAT EXCHANGER B OUTLET 217 52/Mcc-6(13c)

RHR-759B l

13F SI-8458, SPRAY ADDITIVE TANK DISCHARGE 296 52/MCC-6(13F)

SI-845B 13J SI-8608, CV SUMP RECIRC SUCTION 267 52/Mcc-6(13J)

SI-860B 13M SI-870B, BIT OUTLET COLD LEG INJECTION 245 52/Mcc-6(13M) ;

SI-870B l 14c CC-832, CCW MAKE UP FROM PRIMARY WATER 203 52/Mcc-6(14c) cc-832 14F LCV-1150, VOLUME CONTROL TANK DISCHARGE 160 52/McC-6(14F) l LCV-115c 14J SI-8618, CV SUMP SUCTION 269 52/MCC-6(14J), l SI-8,61B i

l EDP-003 Rev, 6 Page 30 of 72

~

7.0 (Continusd) Section 7.0 Page 6 of 7

/%

-Q MCC-6 POWER SUPPLY: 480V BUS E-2 (52/23C) LOCATION: E-1/E-2 ROOM CMPT LOAD TITLE CwD BKR NO. LOAD EDBS TAG UO. NO. EDBS NO.

14M BATTERY CHARGER B-1 956 52/MCC-6(14M)

BAT-CHRGR-B-1 l 15C SI-878B, Si PUMPS B & C DISCHARGE CROSS 258 52/MCC-6(15C) l- CONNECT

! SI-878B 15EL , BLANK N/A N/A N/A i

1 15ER BATTERY CHARGER B 956 52/MCC-6(15ER) l

BAT-CHRGR-B 15F BLANK N/A N/A ;

N/A l 15J FP-258, RCP SPRINKLER & CV FIRE HOSE 751 52/MCC-6(15J)

ISOLATION O,

FP-258 15M FP-249, ELECTRICAL PENETRATION SPRINKLER 749 52/MCC-6(15M)

ISOLATION FP-249 16C N/A 52/MCC-6(16C)

BLANK (PARTIAL)

N/A 16EL FEED iO PP 27 N/A 52/MCC-6(16EL)

PP-27 16ER ;

FEED TO LP-29 (NORMAL POWER) N/A 52/MCC-6(16ER) !

LP-29 -!

l I

16CL SPARE N/A 52/MCC-6(16GL)

N/A 16GR BLANK (PARTIAL) .

N/A 52/MCC-6(16CR). j N/A ,

16JL BORIC ACID HEAT TRACE PRIMARY PANEL N/A 52/MCC-6(16JL)

l. BA-HT-PRI-PNL l

l l0 I

i i

EDP-003 Rev. 6 Page 31 of 72 i

,, 7.0 (continusd) Section 7.0 Page 7 of 1 C'

MCC-6 POWER SUPPLY: 480V BUS E-2 (52/23C) LOCATION: E-1/E. ROOM cMPT LOAD TITLE cWD BKR NO. LOAD EDBS TAG NO. NO. EDBS NO, j ,

16JR BLANK N/A N/A N/A 16M V1-88, SDAFW PUMP STEAM ISOI.ATION 632A 52/MCC-6(16M) l MS-V1 8B 17B ! FUEL HANDLING BUILDING LOWER LEVEL 86 52/Mcc-6(17B)

! EXHAUST GAS, R-20 R-20 17D EDG FUEL OIL TRANSFER PUMP B 953 52/Mcc-6(17D)

DG-FO-XFER FMP-B 17F CV IODINE REMOVAL UNIT, HVE-4 522 l 52/Mcc-6(17F)

HVE-4 17J V6-34D, CV RECIRC COOLER HVH-4 SW OUTLET 508 52/Mcc-6(17J)

V6-34D O. 17M V6-34C, CV RECIRC COOLER HVH-3 SW OUTLET 506 52/Mcc-6(17M)

V6-34C 18c vs-33C, CV RECIRC COOLER HVH-3 SW INLET 502 52/Mcc-6(18c)

V6-33C 18F V6-33D, CV RECIRC COOLER HVH-4 SW INLET 503 52/Mcc-6(18F)

V6-33D 18J V6-33F, CV RECIRC COOLER HVH-2 SW 505 52/Mcc-6(18J)

SELECTIVE INLET V6-33F IBM V1-8C, SDAFW PUMP STEAM ISOLATION 633A 52/Mcc-6(18M)

MS-vi-8c h

O EDP-003 Rev. 6 Page 32 of 72

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

39. SI-09 001 Which ONE (1) of the following describes a design requirement for the Emergency Core Cooling System (ECCS) following a Loss of Coolant Accident?

,p sc A. Full ECCS operation will keep the fuel centerline temperature from exceeding 4500*F.

B. One train of ECCS operation will maintain normal core geometry. "#'

p W ( <. ECCS operation will maintain clad temperature less than or equal to 2200 F. l f, d ' )

D. ECCS operation will limit the H 2generated by Zirconium-Water reaction to less than p., f 17% of the total possible, ygec /

c

/ Ws" 4a K/A 006000.K3.02 4.3/4.4 SD-002 FSAR ,

New Question i

Question 39 of 100

95-2 NRC EXAM - REACTOR OPERATOR

39. SI-09 001

^

Which ONE (1) of the following describes a design requirement for the Emergency Core

Cooling System (ECCS) following a Loss of Coolant Accident?

A. Full ECCS operation will keep the fuel centerline temperature from exceeding 4500 F.

B. One train of ECCS operation will maintain normal core geometry.

<. ECCS operation will maintain clad temperature less than or equal to 2200 F.

l D. ECCS operation will limit the H generated by Zirconium-Water reaction to less than 2

17% of the total possible.

, C K/A 006000.K3.02 4.3/4.4 l SD-002 1

! FSAR New Question r

1 Question 39 of 100

t

! BR 2

, UPDATED FSAR 1

1

! Response:

Adequate emergency core cooling is provided by the SIS (which constitutes the

. Emergency Core Cooling System) whose components operate in three modes. These i modes are delineated as passive accumulator injection, active SI .and residual i heat removal recirculation. i The primary purpose of the SIS is to automatically deliver cooling water to the i i

reactor core in the event of a LOCA. This limits the fuel clad temperature and !

thereby ensures that the core will remain intact and in place, with its heat transfer geometry preserved. This protection is afforded for: -

a) All pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant loop, assuming unobstructed discharge f rom both ends b) A loss of coolant associated with the rod ejection accident, and c) A steam generator tube rupture. )

i The basic design criteria for LOCA evaluations are:

a) The cladding temperature is to be less that:

1) The melting temperature of Zircaloy-4

2) The temperature at which gross core geometry distortion, including clad fragmentation may be expected b) The total core metal-water reaction will be limited to less than one pe rcent.

Thus the core geometry is retained to such an extent that effective cooling of the core is not impaired. '

For any rupture of a steam pipe and the associated uncontrolled heat removal f rom the core, the SIS adds shutdow reactivity so that with a stuck rod, no .

offsite power and minimum engineered safety features, there is no consequential damage to the RCS and the core remains in place and intact.

Redundancy and segregation of instrumentation and components is incorporated to asgure that postulated malfunctions will not impair the ability of the system to meet the design objectives. The system is effective in the event of loss of normal plant auxiliary power coincident with the loss of coolant, and can accommodate the failure of any single component or instrument channel to respond actively in the system. During the recirculation phase of a LOCA, the system can accor.modate a loss of any part of the flow path since back up alternative flow path capability is provided.

3.1.2.45 Inspection of Emergency Core Cooling System Criterion: Design provisions shall, where practical, be made to facilitate physical parts of the Emergency Core Cooling System, including reactor vessel internals and water injection nozzles. (CDC 45) 3.1.2-30

i . . , > _

i LESSON BODY KEY AIDS f, '

2) Post Accident Hydrogen Recombiner System 4

] 8. Emergency Core Cooling Systems function 4

1 i a. Provide core cooling and shutdown capability following:

1) Ioss of coolant accident i

2) Rod cluster control assembly (RCCA) ejection

3) Steam or feedwater system break accident

. 4) Steam generator tube rupture i . . :

b.10 CFR 50.46, BCCS acceptance criteria, applied to

accidents in 10 CFR 50, Appendix K

, 1) Peak cladding temperature will not exceed 2200'F .

! l

! 2) Cladding oxidation will not exceed 17 percent of the j total cladding thickness i 3) Hydrogen generation (due to zirc- water reaction) will not exceed 1 percent of the hydrogen generated if all the zirconium surrounding the fuel reacted 1

i 4) Core remains in a coolable geometry f .

l 5) Ieng-term cooling capability will be maintained j c. ECCS consist of:

1

1) Intermediate head injection (SI pumps)

, 2) Iow head injection (RHR pumps) i i .

! 3) Refueling water storage tank i

4) Accumulatorinjection 1

4

, McD2 Rev.I 1

Pase 10 of 42 i

l

95-2 NRC EXAM - SENIOR REACTOR OPERATOR w

I

{(

j

40. AMSAC-09 002 Given the llowingpa[nt conditions:

' on a

. The unit i t f/v :

power increasing load to 100%

*Noe piment is out of service i

Which ONE (1) of the following describes the MINIMUM setpoints and coincidence required j for the ATWS mitigation system to actuate?

1't A. 2/4 Power Reage NI's greater than 40%, AND 1/3 S/G narrow range level transmitters

! less than 16%.

z l B. J/2 Turbine first stage pressures greater than 40% turbine load, AND 1/3 : 'G narrow i range level transmitters less than 11 %.

W. 2/2 Turbine first stage pressures greater than 40% turbine load, AND 2/3 S/G narrow range level transmitters less than 11 %.

x .. - . . . . . .

D. (2/4 Power Range NI's greater than'40%,' ANDyS/G narrow range level transmitters .

less than 16%. I c

gut [.y A 1e tor h*l 727 Nn K/A 016000.K4.03 (2.8/2.9)

SD-062, Pg. 4.

H.B. Robinson Dwg HBR211259 Modified Question mgo m 4 ro yyp ay ircn nh W G "'"

PT-e t E - rt z(

Question 40 of 100

i 95-2 NRC EXAM - REACTOR OPERATOR i

40. AMSAC-09 002 Given the following plant conditions: i e The unit is at 30% power ,ncreasing load to 100%

. No equpiment is out of service Which ONE (1) of the following describes the MINIMUM setpoints and coincidence required for the ATWS mitigation system to actuate?

A. 2/4 Pe ver hnge NI's greater than 40%, AND 1/3 S/G narrow range level transmitters less than 16%.

B. 1/2 Turbine first stage pressures greater than 40% turbine load, AND 1/3 S/G narrow ;

range level transmitters less than 11 %.

<. 2/2 Turbine first stage pressures greater than 40% turbine load, AND 2/3 S/G narrow i range level transmitters less than 11 %. j D. 2/4 Power Range NI's greater than 40%, AND 2/3 S/G narrow range level transmitters less than 16%. ;

c J

K/A 016000.K4.03 (2.8/2.9)

SD-062, Pg. 4. [

H.B. Robinson Dwg HBR211259 Modified Question ,

i l

l l

l I

f Question 40 of 100 er

l l

n

%/

AMSAC LOGIC DRAWING l

LT-474 Li 485 LT-496 PT-446 PT-447 I

2/3 2/2

< I I '5 360 second @

T.D. Relay

,~) O AW

/ 'i h

(_./ ' '

RTG8 BfPASS 25 second T.D. Relay (h -- %c TURBlNE TRjP/AFw START l

,f-q ss l

\

AMSAC-TP-5

7 _.

1 l 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

! 41. ESF-09 004 !

Given the following plant conditions:

. A reactor trip and safety injection have occurred l

S [ * "B" Si pump started automatically

i A s, 4~)

i .

sp.

g.61 g o (, e "A" SI pump had to be manually started Od s.3 '

' [ .3 . Which ONE (1) of the following describes the shutoff head and maximum flow this pump configuration? \' -of---

Maximum Flow Shutoff Head p2- '# < l y A. 600 gpm each 1600 psig i i

4. 600 gpm each 1500 psig l C. 375 gpm total 1600 psig D. 375 gpm total 1500 psig b

l K/A 006000.K5.06 3.5/3.9 i SD-002. page 23. SI-LP, pg 12 Modified Question o y) 7 O gb E (1 ITA^IO $ EV- ;

i l

l l

i l

l

! l l

r Question 41 of 100 i l

1

l 95-2 NRC EXAM - REACTOR OPERATOR l

41. ESF-09 004 -

Given the following plant conditions:

A reactor trip and safety injection have occurred

"B" SI pump started automatically.

"A" SI pump had to be manually started Which ONE (1) of the following describes the shutoff head and maximum flow this pump configuration?

Maximum Flow Shutoff Head A. 600 gpm each 1600 psig

4. 600 gpm each 1500 psig C. 375 gpm total 1600 psig I

D. 375 gpm total 1500 psig

]

b K/A 006000.K5.06 3.5/3.9 SD-002 page 23. SI-LP, pg 12 Modified Question l

Question 41 of 100

(

LESSON BODY. KEY AIDS

c. Centrifugal

d. Shuteff head 1500 psig i

e. Flow l 375 gpm each - design (1)

(2) 600 gpm each - maximum

f. Staning duty limitations of 3 starts in an hour unless ACR 92-00325 1 of the last 2 starts resulted in at least a 15 minute mn Maximum of 8 starts in one day

g. Contauls - Two Position Switch (RTGB)

(1) STOP l

(2) Un-marked (Spring return to middle)

I (3) START

2. USES

a. Inject borated water to RCS during safeguards

b. Fill accumulators

c. Slow fill of refueling cavity OBJ. #9

3. Area cooling fans started when pumps are staned HVH-6A & 6B 1

l l 4. Seals cooled by Component Cooling Water l

4 51 Rev.1 Page 12 of 43 l

l I

, 95-2 NRC EXAM -SENIOR REACTOR OPERATOR 1 l ' 42. AMSAC-09 001 i Given the following plant conditions: I l

The unit is at 100% power 1

l

All system are aligned for nomal full power operation l Which ONE (1) of the following describes the turbine trip provided by AMSAC? j VA. Actuates 20AST, auto stop oil trip relay, to trip the turbine. l 1

B. Actuates 20ET, Reactor Protection System relay, to trip the turbine. !

d

[ C. Actuates the thrust bearing trip device to trip the turbine.

(ht y, ,

D. Actuates the 86P, Generator Lockout Relay, to trip the turbine. l a

i K/A 012000.K6.03 (3.1/3.5)

H.B. Robinson Dwg 53793695 AMSAC, Obj 9/10.

Modified Question I !

l l

l Question 42 of 100 l

. 95-2 NRC EXAM - REACTOR OPERATOR

.g

'42. AMSAC-09 001

, Given the following plant conditions: -

The unit is at 100% power

. )

. All system are aligned for nomal full power operation 4

4 Which ONE (1) of the following describes the turbine trip provided by AMSAC? l 1

VA. Actuates 20AST, auto stop oil trip relay, to trip the turbine.

B. Actuates 20ET, Reactor Protection System relay, to trip the turbine.

C. Actuates the thrust bearing trip device to trip the turbine.

D. Actuates the 86P, Generater Lockout Relay, to trip the turbine.

a K/A 012000.K6.03 (3.1/3.5)

H.B. Robinson Dwg 53793695 AMSAC, Obj 9/10. !

Modified Question i

I i

Question 42 of 100

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j 95-2 NRC EXAM - SENIOR REACTOR OPERATOR i

43. MS-09 002

- Given teh following plant conditions:

.. <

The unit is at 100% power

!h

. No equipment is out of service Which ONE (1) of the following describes why there are four safety valves on each steamline

_incenaer ene er tw?

Withfour valves;- .. .

- geot VA. total steam flow from a loss of load can be relieved. # JM H ,( ,

B. 120% of total steam flow can be relieved preventingthe overpreutrrizatirnrufTiiew ,

secondar y i vd. failae of one valve to open will still prevent lifting the PZR PORV's following a turbine trip.

I. failwe of one valve to close will still allow for a controlled shutdown of the plant, a

K/A 039000.A1.05 3.2/3.3 SD-025 New Question MEM Ai y f w o Q>

316m l r

Question 43 of 100 1

95-2 NRC EXAM - REACTOR OPERATOR l

43. MS-09 002 Given teh follawing plant conditic.ns: ,

.The unit is at 100% power l

l

No equipment is out of service

- Which ONE (1) of the following describes why there are four safety valves on each steamline instead of one or two?

l With four valves, ...

VA. total steam flow from a loss of load can be relieved.

l B. 120% of total steam Dow can be relieved preventing the overprresurization of the I secondary.

t C. failure of one valve to open will still prevent lifting the P7.R PORV's following a turbine trip.

J D. failure of one valve to close will still allow for a controlled shutdown of the plant.

E a

K/A 039000.A1.05 3.2/3.3 SD-025 New Question i

l I

l l

Question 43 of 100 i

Rath * '

j A reactor shutdown from power requires removal ef core decay heat. Immediate l

!. decay heat removal requirements are normally satisfied by the steam bypass to j

! the condenser. 'Therefore, core decay heat can be continuously dissipated via ;

, the steam bypass to the condenser as feedwater in the steam generator is j l converted to steam by heat absorption. Normally, th* capability to return :

! feedwater flow to the steam generators is provided broperation of the turbine .

l l

cycle feedwater system. -

..- )

F The twelve main steam safety valves have a total combinad capability of 7

, .1.022 x 10 lbs/hr. The total full power steam flow is 1.011 x 10 7 lbs/hr.; j i therefore, twelve (12) main steam safety valves will be able to relieve the 1 total . steam flow if necessary."' Following a loss of load,. which represents . l

the worst transient, steam flows are below the total capacity of the 12 safety l l valves. Therefore, over-pressurization of the seccndary system is not possible. l In the unlikely event of complete loss of turbine-generator and offsite electrical power to the plant, decay heat removal would continue ~.to be assured by the availability of either the steam-driven auxiliary Teedwater pump or one 4

of the two motor-driven auxiliary stsam generator feedwater pumps operated from the diesel generators and steam discharge to the atmosphere via the main steam safety valves and atmospheric relief valves. One motor-driven auxiliary ;

. feedwater !

the plant. gapThecanauxiliary supply feedwater sufficient feedwater for removal system essential of are features decay heat from those l

features that provide auxiliary feedwater~ flow to two out of three steam I i generators consistent with auxiliary feedwater pump operability. In order to l provide a high degree of reliability all three auxiliary feedwater pumps will i be operable prior to exceeding 350*F. The minimum amount of water in the !

j condensate storage tank is the amount needed for at least two hours operation ;

1 at hot standby conditions. If the outage is more than two hours, deep well or

! Lake Robinson water may be used.

{ An unlimited supply is available from the lake via either leg of the plant

- Service Water System for an indefinite time period.

i o

i i

I j .

3.4-3 Amendment No. M,102

m_. _._. . _ _ . _ _ ,_ _ . _ . - . . . _ . _ _ . _ . . . . _ _ . ~ , . . _ . _ . _ _ - .__._..-..__..._,_m

, e >

4.

Document No. DBD/R8'/038/SD25 Page No. 50, Revision 1 '

4.5.3.2 Ebasco Specification CPL-R2-IN-4, i.e., Reference 6'.'178,'and -

1 Ebasco Purchase Order NY-434135, i.e., Reference 6.179 contain ,

the origina1' design information and requirements for the MSR

tube bundle vent Valves to the condensers. .

i

'4. 6 SAFETY VALVES 5 t

i

4. 6,,1 SG Safety Valves (SV1-1A, 15, 1C, 2A, 2B, 2C, 3A, 3B, 3C, 4A,

{.

}- 4B & 4C) i i l j) Safety'Related Functions r- 4.6.1.1 ' The SG safety' valves shall limit the pressure rise that may I l

! occur'following a loss of load, as required. A loss of load l l

b? <

represents the worst case transient. The SG safety valves were l- designed to provide a total flow capacity equal to the steam generation rate at maximum calculated conditions. (References i 6.1,. Sections 10.2 and 10.3; 6.53; and 6.169).

1

.4.6.1.2 The SG safety valves shall remain closed, as required, to ]

provide containment isolation for Penetrations P-7 through P-9.

.The SG safety valves function as the second barrier for these penetrations, with the first barrier being a closed system inside containment. For containment isolation requirements )

1 associated with these valves, see Reference 6.104. (Reference 6.104) i 4.6.1.30 The maximum flow capacity of a single SG safety valve shall'be l limited to prevent excessive RCS cooldown in the event of a failure of the valve in the open position. (Reference 6.46) 4.6.1.4 In the original design, the SG safety valves were solely ;

credited as the means by which the RCS maintained itself near nominal no load conditions. The valves, in conjunction with other credited equipment, assured the plant could achieve and maintain hot shutdown following a control room evacuation.

(Reference 6.1, Section 7.7.5)

.- 95-2 NRC EXAM - SENIOR REACTOR OPERATOR i J

44. RCS-09 005 Which, ONE (1) of the following describes the response to the controlling channel of Pressurizer level failing HIGH during full power? Assume NO operator actions.

N

\

VA. Reacto'r twill trip on low pressure.

\

B. Charging' pump speed increases.

\

.l '

C. Letdown iso \lates and remains isolated. l

'I D. All pressurize \r heaters trip and reenergize when level returns above 14.4 %.

a \,

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K/A 011000.A2.10 3.4/ 6 SD-059 \

Modified Question

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Question 44 of 100 j

.- ..- -- . . - - . . .-. . - - - -. - .-.- - ~ .-

i

! SD 059 ,_

PRES 5URIZER SYSTEM

! 5.1.3 Pressurizer PORV Contml (PZR-Figure 8 & PZR-Figum 13) l t I The Pmssurizer PORVs have two modes of control, Normsl and Low Temperature Overpressure Protection (LTOPP). In normal mode the PORVs have a permissive of

2000 psig to open in Autome. tic. This " permissive" is supplied by the protection l

channels meeting a 2/3 logic . As stated before PCV-456 receives its signal from FT-

445 set at 2335 psig and PCV-455C receives its signal from PC-444A at +100 psi j which is nominally 2335 psig also. When the key switch for OVERPRESSURE i

PRCITECTION on the RTGB is place in the LOW PRESSURE position (one switch for I

each PORV) the input to each PORV is shift:.d to the LTOPP contmller.

l Sil.4 Low Temperature Overpressure Protection Control (LTOPP) (PZR-Figure 13) i LTOPP contml is required to be activated when the RCS is cooled down below 360*F l

i to minimize Pressurized Thermal Shock (P.S..) concems. The LTOPP controller uses i

i, the lowest of TE-410, TE-420 and TE-430 to determine RCS temperature and pressure as sensed by PT-500 and FT-501. The lift setpoint is variable based upon auctioneered low RCS temperature. At the highest RCS temperature that LTOPP is required to be f

activated, 360*F, the pressure setpoint is 400 psig. 'Ihe setpoint of the Comparators

! PC502 and PC503 are adjusted downward as RCS temperature is decreased.

i

! There is one alarm associated with each channel of LTOPP. It actuates for 3 reasons:

s l (1) RCS temperature is <360*F and LTOPP is not " armed" low Pressure not selected 4 on the key switch for OVERPRESSURE PROTECTION, (2) The PORV has received j and actuation signal based upon current pressure and temperature or (3) the associated i Block valve is shut.

i

5.1.5 Pressurizer level Control (PZR-Figure 12)

! Pressurizer level is controlled by controlling charging pump speed. The level is i

4 PZR Page 26 of 36 Revision 0 l INFORMATION USE ONLY 4

4

,x - , - - , - - - -

____._._...e__ _ _ _ _ .._.. _

i SD-059 PRESSURIZER SYSTEM i

programmed to ramp up as Tavg increases by LC-459G. This maintains approximately constant mass in the RCS as Tavg is incmased and the coolant in the RCS ' expands.

l

]

i Level program is 22.2% at Tavg of 547'F and 53.3% st Tavg of 575.4*F. j I

t There are 3 Pressurizer level channels LT-459, LT-460 and LT-461. II-459G the -

Pmssurizer level controller is normally fed by level channel LT.459 but can be replaced l by LT461 with a selector switch on the RTGB. The output of LC-459G is then fed to l

the charging pump speed controllers to control speed of the charging pump if their I

contrellers are selected to Auto.

i s .

i If Pressurizer level inemases 5% above pmgram LC-459D will turn on the backup l heaters and sound an annunciator for High level Heaters on.

On Pressurizer low level of 14.4%, proportional and backup heaters are deenergized and letdown is isolated by shutting LCV-460A & B if zespective control switches are in f

I auto. LC-459 and the LC-460, the low level bistables, are normally supplied by LT-459 i

! and LT-460 igevely but either can be replaced by LT-461 with a selector switch on the RTGB.

l L

3 It has been noted by checking the Control Wiring Diagrams that LC459 will ordy tum off the backup heaters that are selected to Automatic where LC-460 will turn off the

)

i backup heaten in Automatic or Manual. ' The only time this would have any bearing l would be in the event of an instrument failure. If the channel feeding LC-459, usually

! LT-459, were to fait low the proportional heaters and any backup heaters in Automatia !

would deenergize and any backup heater in manual would remain energized. )

l 5.1.6 Pressurizer level Control Setpoints i

1. Level program as function of Tavg l

i- (TM-459)

! for Tavg 547'F 22.2% of level span 1

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i PZR Page 27 of 36 Revision 0 l INFORMATION USE ONLY l i

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SD-059 PRESSURIZER SYSTEM l

i

for T., 575.4*F 53.3% oflevel span 4 (Pmgram in linear fmm 547'F to 575.4*F)

I.ow limit 22.2 % of level span

}

High limit 53.3% of level span

a

2. I.ow-Low Level Heater Cutout

. (LC-459C, LC-460C) 14.4.% of level span l- 3. Level Contmlier j (LC-459F) 10% charging pump Pmportional gain speed /% level dniation

) Reset time constant 430 seconds

4. I.etdown Valve Isolation 14.4% of level span l'

j 5. Back-up Heaters on +5% of programmed level i.

i 6.0 SYSTEM OPERATION

) 6.1 Normal Opemtion i

i Insurge of RCS Coolant - produced by increase in Tavg. An insurge of coolant will reovce volume of the steam bubble causing an increase in the temperature and pressure of the steam. The steam space or bubble becomes superheated and some minor condensation occurs at surface and on walls. i The increased pressure causes the spray valve to open which cools and condenses a part ,.

1 of the steam bubble, thereby reducing pressure.

. i i

The increase in level will energize backup heaters if the level increases to 5 % above :

I program. 3 l

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PZR Page 28 of 36 Revision 0 ; )

INFORMATION USE ONLY

]

4- .

SD-059 PRESSURIZER SYSTEM Outsurge of RCS Coolant l

i i' t t An outsorge of RCS coolant will incmase the volume of the steam bubble, which will l'

! - cause water to flash to steam, limiting the pressure decmase.

i, j

ii Proportionai heaters will be full on to limit pressure dectuase. If p!vssure decrease is large enough, the backup heaters will energize.

! 6.2 Infrequent Operations 3

I 6.2.1 RCS Heatup j i I' i

j The pressurizer is initially water solid with pressure being controlled by charging and j

. letdown from RHR. A bubble is fornwd by energizing the heaters, heating the system )

to saturation, then -Ms.g to pruper level. Bubble formation is indicated by an increase l l

in letdown flov when the Pressurizer reaches saturation. "1he rate of heating the Pressurizer is limited by Technical Specifications.

I f '6 ..2.2 RCS Cooldown i l During an RCS and Pressurizer cooldown a bubble is maintained until at least 350'F, !

]

i at which time the bubble may be collapsed whh cool water from CVCS after placing the !

!' RCS on RHR. The bubble must be collapsed prior to cooling the RCS below 150*F. !

4 ,

. The cooldown rate of the Pressurizer is limited by Technical Specifications. !

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I 6.3 Abnormal Operation

'1he Preuurizer automatically mponds to all abnormal conditions via automatic controls j or safety valves.

PZR P6ge 29 of 36 Revision 0 i INFORMATION USE ONLY

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1 4

95-2 NRC EXAM - SENIOR REACTOR OPERATOR !

l i

$ 45. GSS-09 001 l

. Given the following plant conditions:

\

g . An Si has occurred l

All systems operated as designed i

! Which ONE (1) of the following describes die complete action (s) a Safety Injection signal will have on the Containment Systems? c ff

~ u ns oo r / p, c ~S sc{

V' 7Aq.',7'f',nv-c l

A. Close Phase A andhalyes/ Start HVH-1 thru 4.

.. W. Close Phase A valves, Start HVH-1 thru 4. -> 04LI #'" O C. Close Phase B vaives, Stop CV purge.

4 D. Close Phase A and B velves, Stop CV purge.

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l KA 029000.A3.01 3.8/4.0 Dwg 5379-2759 l Dwg 5379-3367 sht 14 of 18

Modified Question .

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. Question 45 of 100 i

- -. -- . . _ _ . _ . ~ . - - - .-.- - ... -- -. -, ._ .

~

95-2 NRC EXAM - REACTOR OPERATOR

, 45. CSS-09 001 Given the following plant conditions: i 4

An SI has occurred

All systems operat:d as designed i Which 'ONE (1) of the following describes the complete action (s) a Safety injection signal will

~

have on the Containment Systems? i l

l A. Close Phase A and B valves, Start HVH-1 thru 4.

4. Close Phase A valves, Start HVH-1 thru 4.

I C. Close Phase B valves, Stop CV purge.

D. Close Phase A and B valves, Stop CV purge.

b l

KA 029000. A3.01 3.8/4.0 Dwg 5379-2759 i Dwg 5379-3367 sht 14 of a ;

Modified Question j a

I Question 45 of 100

95-2 NRC EXAM - SENIOR REACTOR OPERATOR i

I i 46. OP-006 001

'Given the following platit conditions:

i x '

+ The plant is in cold shutdown l VVin <> ]

Low Temperature Overpressure Protecti on (LTOPP) is in service !

! /

i Which ONE (1) of the following would prevent PCV-456, PZR PORV, from automatically ;

opening on an overpressure signal from the LTOPP system? !

1 i

A. RTGB RC-535, PORY Block, control switch in CLOSE.

W. Pressurier PORV Overpressure Protection key switch in NORMA l O ,

- 1 i

C. TE-410, WR Tcold, failed HIGH. )

i D. PT-501, PZR Press, failed LOW :

.1 b l K/A 010000A403 '(4.0/3.8) l OP-006, Att 9.2, PORV Logic. 3; g f j I

Modified Question v, g4 l fiY ffsI$ n y,9 l

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Question 46 of 100 l J

l 95-2 NRC EXAM - REACTOR OPERATOR l 46. OP-006 001 l Given the following plant conditions:

l l

The plant is in cold shutdown

Low Temperature Overpressure Protection (LTOPP) is in service Which ONE (1) of the following.would prevent PCV-456, PZR PORV, from automatically

opening on an overpressure signal from the LTOPP system?

A. RTGB RC-535, PORV Block, control switch in CLOSE.

4. P ssurizer PORY Overpressure Protection key switch in NORMAL.

C. TE-410, WR Tcold, failed HIGH.

D. PT-501, PZR Press, failed LOW b

K/A 010000A403 (4.0/3.8)

OP-006, Att 9.2, PORV Logic.

Modified Question Question 46 of 100

INFORMATION USE ATTACHMENT 9.2 Page 1 of 1 LOGIC DIAGRAM LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM i

l !

I es ,

I I -s n u u

j

v v TW502 L0s of l T LO' 0' 3 l

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. 0W502 l '/"* l atT POINT 350 op QW503 l 'a"r"w" ,

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'*UN((I C502 TC503 s'e UTcM 3

NORM PC502 lN ss NORM l t! 3 P 503

-I

{_. ._]

OL Jur_ Jur_ . I.J L OPERAff ARM ARM __0PERATE

[ " $ " "'

f Mi pre @ Al ceen closco closeo rWAMMM ell I

,,, ,,,,,, a re ewR Re6:e, l -[g l l -

l lctese CPIN l l0 PIN CLO5E l l

ir h l u v u u l ,

l lPCV-456 l APP 003 A3 App.co3.Ap h ll PCV 455C l

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l OP-006 Rev. 22 Page 38 of 40

. _ . . _ _ . _ _ . _ - _ . . . . . . . ~ _ . _ . _ . _ . . . __ . _ . . . _ _ _ . . . . _ . . _ _ .

, , 95-2 NRC EXAM - SENIOR REACTOR OPERATOR i

47. SFPC-02 002 I Which one (1) of the following is the basis for the MINIMUM Spent Fuel Pool temperature of 68'F7 A. Boraflex liner adhesion decreases below this temperature.

! B. SFP demineralizer efficiency decreases below this temperature. l i

vC. SFP criticality analysis is invalid below this temperature. l l

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D. SFP radiation monitor calibration is invalid below this temperature. l

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K/A 0330000006 (2.1/3.1)

OP-910, Spent Fuel Pool Cooling and Purification System Pg. 7, Precaution 7.

1 SFP, Obj. 9,12,& 13, i Modified Question l

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l Question 47 of 100

. . _ . _ _4

4.0 -(Continued)

6. Do not allow t;ne Spent Fuel Filter to exceed 25 psid. The filter f _

should be changed when Differential Pressure is between 20 and 25 psid.

t

7. Do not allow the SFP temperature to drop below 68'F. The Spent Fuel Pit criticality analysis is based on a minimum SFP temperature of 68'F. If SFF temperature decreases to less than 68'F, the analysis is no longer valid.(SER 90-17) 8 .. This procedure has been screened in accordance with PLP-037 criteria and determined not applicable (N/A) to PLP-037.

3 9. If the starting limitations stated below are exceeded, SFP PUMP motor damage can occur due to motor overheating:(ACR 92-325)

Maximum number of starts per hour is 5.

Minimum time between starts is 2 minutes.

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i OP-910 Rev. 15 Page 8 of 46

)

, o 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

\

i 48. FW-09 002 Given the following plant conditions:

The' unit is at 50% power 2

Main'Feedwater Regulating valves are in AUTO

\

Which ONE (I) of the following describes e' signal that will NOT shut the Main Feedwater l Regulating valves,in 5 seconds?

1 s

A. Reactor Tr!p with Lo Tavg vB. Phase A Isolation ;3 g

C. Safety Injection 1,1 l D. S/G Hi-Hi l evel

) b K/A 035000.Kl.01 4.2/4.5 Logic dwg 5379-2761 sht 10 of 18 Modified Question l 5 seconds comes from logic dwg i

ll0 nEr 4

i 7)h f

a i

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. Question 48 of 100 1

o .

l 95-2 NRC EXAM - REACTOR OPERATOR

48. FW-09 002 Given the following plant conditions:

The unit is at 50% power e Main Feedwater Regulating valves are in AUTO Which ONE (1) of the following describes the signal that will NOT shut the Main Feedwater Regulating valves in 5 seconds?

A. Reactor Trip with Lo Tavg

4. Phase A isolation C. Safety injection D. S/G Hi-Hi Level )

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l K/A 035000.K1.01 4.2/4.5 Logic dwg 5379-2761 sht 10 of 18 Modified Question 5 seconds comes from logic dwg 4

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l Question 48 of 100 i ,

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

49. RPS-09 004 Which ONE (1) of the following describes the control power supply for Reactor Trip Breaker

"A" and Reactor Trip Bypass Breaker "A"?

Reactor Trin " A" Reactor Trip Bvoass " A" A. "A" 125 VDC Dist. Panel "A" 125 VDC Dist. Panel

4. "A" 125 VDC Dist. Panel "B" 125 VDC Dist. Panel j C. "B" 125 VDC Dist. Panel "A" 125 VDC Dist. Panel D. "B" 125 VDC Dist. Panel "B" 125 VDC Dist. Panel I

KA 012000.K2.01 3.3/3.7 EDP-004 Modified Question

! F) k mo,y j l

4 a

Question 49 of 100 P/J

95-2 NRC FX AM - REACTOR OPERATOR

49. RPS-09 004 Which ONE (1) of the following describes the control power supply for Reactor Trip Breaker -

"A" and Reactor Trip Bypass Breaker "A"?

Reactor Trip " A" Reactor Trio Bvoass " A" A. "A" 125 VDC Dist. Panel "A" 125 VDC Dist. Panel-

4. "A" 125 VDC Dist. Panel "B" 125 VDC Dist. Panel C. "B" 125 VDC Dist. Panel "A" 125 VDC Dist. Panel D. "B" 125 VDC Dist. Panel "B" 125 VDC Dist. Panel b

KA 012000.K2.01 3.3/3.7 EDP-004 Modified Question J

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Question 49 of 100 J

2.0 DISTRIBUTION PANEL "A" ,

Location: On 125V DC MCC "A" Power Supply: 125V DC MCC "A"-

Circuit Lggd 1 480V Switchgear No. E-1 l i

2 4160V Switchgear Busses 1 & 2 3 Hydrogen Control Panel 4 ,

480V Switchgear Busses 1 & 2A 5 Lighting Panel LP-33 +

6 125V DC Power Panel "A-1" 7 Startup Transformer Motor Operated Disconnects 8 Diesel Generator "A" Exciter 9 Inverter "C"

! 10 Reactor Trip Breaker "A" & Reactor Trip Bypass ,

Breaker "B"

! 11 Inverter "A" t I

i 12 Rod Drive M G Set "A" l

13 Main Generator Exciter Field Breaker 14 Gas Stripper Control Cabinet "A" 15 Generator Lockout Relay 86P 16 Aux. Panel "D-C" Fuse Panel ,

17 Auxiliary Transformer Annunciator 18 Reactor Protection Train "A" 19 Spare f 20 Safeguards Train "A" 21 Spore 22 Turbine Auto Trip

! 23 Startup Transformer Annunciator 24 Diesel Generator "A" Control Power i

EDP-004 R:;. 4 Page 5 of 14 l

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5.0 DISTRIBUTION PANEL "B" l

Location: On 125V DC MCC "B" !

Power Supply: 125V DC MCC "B" Circuit LEA $

i.

1 480V Switchgear No. E-2 l 2 4160V Switchgear Busses 3 & 4 3 4160V Breaker Test Panel

! 4 480V.Switchgear Busses 2B & 3 5 125V DC MCC "B-A" l

l 6 Sample Panel 7 Spare 8 Diesel Generator "B" Exciter 9 Reactor Trip Breaker "B" & Reactor Trip Bypasp -

! Breaker "A"

! 10 Annunciator Panel (RTGB) t 11 Waste Disposal Panel j 12 Diesel Generator "B" Control Power l

l 13 Turbine Emergency Trip 14 Cas Stripper Panel "B" 15 Gas Analyzer Panel '

16 Aux. Panel "G-C" Fuse Panel i 17 Generater Lockout Relay 86 BU 18 Reactr,r Protection Train "B" l

19 Reverse Current Valves 20 Safeguards Train "B" 21 Drumming Room Control Panel 22 Distribution Panel "B-1" '

23 Stee.m Driven AW Pump Control Power 24 Rod Drive M-G Set "B" i

t EDP-004 Rev. 4 Page 8 of 14 i

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l 95-2 NRC EXAM - SENIOR REACTOR OPERATOR !

i

50. VAC-05 002 l Given the following plant conditions:

. The plant is in hot shutdown l 1

l

A partia. loss of AC power has occurred e The operating crew has diagnosed a loss of 4KV bus 1 Which ONE (1) of the following describes the plant equipment that is affected by the power loss?

VA. Circulating Water Pump "A", Station Service Tranformer 2B B. Main Feewater Pump "A", Circulating Water Pump "B" C. Station Service Tranformer 2D, Heater Drain Pump "B" D. Reactor Coolant Pump "A", Main Feewater Pump "A" a l i

062000.K3.01 3.5/3.9 i EDP-001 l New Question i

O T oo n g of f I19 ft 6 j'O w L fl NPtLY Q v Ej r, o us, l

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Question 50 of 100

{'!J

4 l*

95-2 NRC EXAM - REACTOR OPERATOR

50. VAC-05 002 Given the following plant conditions:

. The plant is in hot shutdown

. A partial loss of AC power has occurred e The operating crew has diagnosed a loss of 4KV bus 1 Which ONE (1) of the following describes the plant equipment that is affected by the power loss?

VA. Circulating Water Pump "A", Station Service Tranformer 2B B. Main Feewater Pump "A", Circulating Water Pump "B" C. Station Service Tranformer 2D, l{ eater Drain Pump "B" D. Reactor Coolant Pump "A", Main Feewater Pump "A" a

%2000.K3.01 3.5/3.9 EDP-001 New Question Question 50 of 100

1.0 4160V AC Bus No. 1 Location: 4160V Switchgear Room Power Supply: As per RTGB Line Up LgAdi: Cubiele Breaker M Reactor Coolant Pump "A" 1 52/1 109 Circulating Water Pump "A" 2 52/2 811 Steam Generator Feedwater Pump "A" 3 52/3 615 Station Service Transformer No. 28 4 52/4 933 Heater Drain Pump "A" 5 52/5 625 Condensate Pump "A' 6 52/6 605 Incoming Line - Unit Aux. Transformer No. 2 7 52/7 926 pts and Fan Equipment 8 N/A 948 pts and Fan Equipment and Metering 9 W/A 948 Bus 1 and 2 Tie Breaker 10 52/10 928 s

EDP-001 Rev. 2 Page 4 of 8

2.0 - 4160V BUSS NO. 2 -

Location: 4160V Switchgear Room Power Supply: As per RTGB Line Up Lgada: Cubiele Breaker g!D pts and Fan Equipment and Metering 11 N/A 948 Incopt tg Line - Startup Transformer No. 2 12 52/12 927 2A ano 'T Station Service Tr ansformers 13 52/13 932 Reactor Coolant Pump "C" 14 52/14' 105 l

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. - - . . . .~ . . - ~ . - . -- ..

i 3.0 4160V AC BUSS NO. 3 i

l Location: 4160V Switchgear Room i

l Power Supply: As per RTGB Line Up

! Loads: Cubicle Breaker Q]E!

2C and 2G Station Service Transformers 15 52/15 934 pts and Fan Equipment 16 N/A 949 Incoming Line - Startup Transformer No. 2 17 52/17 929B pts and Fan Equipment 18 N/A 949 l

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I EDP-001 Rev. 2 Page 6 of 8

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4.0 ' 4160V AC BUSS NO. 4 location: 4160V Switchgear Room Power Supply: As per RTGB Line Up Loads: Cubicle Breaker M Bus No. 3 and 4 Tie Breaker 19 52/19 931 Incoming Line - Unit Aux. Transformer No. 2 20 52/20 930 pts and Fan Equipment ,

21 N/A 949 Condensate Water Pump "B" 22 52/22 606 Circulating Water Pump "B" 23 52/23 813 4160V Bus 5 Feeder 24 52/24 1344 Heater Drain Pump "B" 25 52/25 626 Steam Generator Feedwater Pump "B" 26 52/26 620 Reactor Coolant Pump "B" 27 52/27 101 2D Station Service Transforuer 28 52/28 1041 l

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EDP-001 Rev. 2 Page 7 of 8 4

5.0 4160V AC BUSS NO. S Location: Turbine Bldg., 1st Level, ;

Grid location 3B Power Supply: As per RTGB Line Up L9.AAA: Cubicle Breaker gg Incoming Line - 4160V Bus No. 4, cubical 24, cts and Metering 29 N/A 1344 pts and Control Power Transformer 30 N/A N/A l SPARE 31 52/31 N/A Station Service Transformer No. 2E 32 52/32 1399

Circulating Wacer Pump "C" 33 52/33 815 SPARE 34 52/34 N/A I

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EDP-001 Rev. 2 Page 8 of 8 ;

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95-2 NRC EXAM - SENIOR REACTOR OPERATOR

51. EDG-09 001 Given the following plant conditions:

4

. The plant is at 100% power e A loss of E-1 results in an AUTO start of "A" EDG Which ONE (1) of the following describes the EDG trip signahthat remain / act an emergency start?

] A.' Low Oil Pressure and Overspeed i L. < D t

B. High.Wa.ter Temperature and Reverse Power i

C. Low Water Pressure and Overspeed l

@. Reverse Power and Overspeed

! d i

K/A 064000.K4.02 3.9/4.2 i EDG LP

New Question l

l Question 51 of 100

95-2 NRC EXAM - REACTOR OPERATOR

51. EDG-09 001 Given the following plant conditions:

The plant is at 100% power

-

A loss of E-1 results in an AUTO start of "A" EDG ,

Which ONE (1) of the following describes the EDG trip signal that remains active following an emergency start?

A. Low Oil Pressure and Overspeed B. High Water Temperature and Reverse Power C. Low Water Pressure and Overspeed J

@. Reverse Power and Overspeed d

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l K/A 064000.K4.02 3.9/4.2 l EDG LP New Question 1

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Question 51 of 100

LESSON BODY KEY AIDS

8. Starting Air System

9. D/G Ventilation 1

4

10. Emergency Diesel Generator A and B RPM Measurement Enclosure B. EMERGENCY DIESEL GENERATOR

1. DIESEL ENGINES - (Two) 12443 cubic inches each

a. Provides the driving force for the Diesel Generator OBJ. #4

! b. Fairbanks-Morse fuelinjected, turbo charged and air started i

!

c. 12 cylinder opposed piston engine rated at 3600 HP i

d. Standby status at all times .

e. Auto-start on safeguards signal or undervoltage (UV) on either E-! or E-2 P

(1) UV on E-1 and E-2 diesels start and assume load of El or E2

, (2) SI signal- start only

f. Manually started from RTGB or locally

g. Designed to reach rated speed and energize bus within 10 seconds EDG Rev.1 Psge 7 of 68

m _ . _ . _ . . . . . . . . - . _ . ~ . _ _ _ . . _ . - . . _ _ . . _ _ - _. . _ . - _ _ . . . . _ _ _ . . . . . .

LESSO'N BODY KEY AIDS i

h. Engine Trips (1) Over speed 1940 RPM (Within the range of OBJ. #9 l 1030-1052 RPM)

Cannot be defeated

Trips fuel rack o Fuel racks must be reset by moving fuel rack trip l reset lever to its full counterclockwise position l NOTE: Trips (2) thru (6) diesel energize the governor l shutdown solenoid which repositions fuel l racks to "no-fuel" position (Zero Fuel l Position) l l (2) Hi crankcase pressure + 0.5" H2 O

Normally defeated (3) Low lube oil pressure - 18 psi

Normally defeated l

(4) Lowjacket water pressure- 12 psi l

Normally defeated l (5) High jacket water temperature - 205"F

Nonnally defeated (6) Manual

Requires resetting fuel racks

Cannot be defeated l

l

EDG A/B GROUND (APP-010-F1/F4) l Requires Manual Trip TP-EDG-13 I

I

! EDG Rev.1 Page 8 of 68 l

l I

1 i

LESSON BODY KEY AIDS (7) 10 Second Overcrank OBJ. #9 i e < 200 RPM 10 seconds after receiving start signal.

l (Shuts the air start solenoids causing diesel to stop) ,

l :

I e Cannot be defeated j e Closes air stan solenoids in 10 seconds to conserve :

! O i- (8) Start failure i ,

4 j e Normally defeated i

e Closes air start solenoids and runs governor to "0" speed 1 i

e EDG <200 RPM after 10 sec.

l i. Trips Defeat Switch OBJ. #8 l'

(1) Located on diesel control panel (2) Key operated f

Trips defeated / trips in service (3) Normally in trips defeated position (4) For OST operation-trips in service e NOTE: All trips controlled by the Trips Defeat i

Switch are bypassed for 20 seconds to allow engine to start *

(5) The following trips are normally bypassed to insure i that in an emergency situation the D/G's will start under all conditions. They are reset by depressing the alarm reset push button.

l e High crankcase pressure - crankcase is evacuated by j turbo-blower suction, (not operating on idle engine)

, I i

EDG Rev.1 Page 9 of 68 j .

i 4

LESSON BODY KEY AMS 4

Low lube oli pressure - pressure provided by engine driven pump which must come up to speed l

{

Lowjacket water pressure - water pressure also

provided by attached pump

High jacket water temp - high jacket water l temperatures could exist after engine shutdown due to no flow would clear when flow was initiated J

(6) While running the EDG (per OP-604) with the TP-EDG-13 trips defeated if a condition occurs that I i would automatically trip the Diesel, the EDG should be trippedimmediately:

1 4

Coolant Temperature High - 205 *F

,

Crankcase Pressure + 0.5 inches H 2O .

l

Lube Oil Low Pressure- 18 psig

Coolant Low Pressure - 12 psig ,

(7) EDG A/B TROUBLE (APP-010-A2/A3) OBJ. #9

2. GENERATOR O BJ.#4,5 1

a. Provides power to the Emergency Busses E? and E2 'when '

needed b.' Rated at 2500 KW,900 RPM,480VAC,3 phase

]. c. Directly coupled to engine lower crank shaft

d. Exciter - regulator - provides and controls the current to field winding to maintain generator voltage within the

regulation band frem no load to full load EDG Rev.1 Page 10 of 68 4

i LESSON BODY KEY AIDS ;

2 (1) Normal power- DC Bus OBJ. #6

EDG "A" - 125 VDC MCC "A" l

e EDG "B" - 125 VDC MCC "B" i

(2) Backup power - Bank of batteries in D/G rooms (3) EDG Field Flash TP-EDG-2

The low speed relay energizes a K-1 relay, at 200

RPM, via normally closed #5 Relay contact ("b" contact from stopping relay) 4 l * 'When energized the K1 relay opens the K1 contact .

("b" contact) [when closed the K1 contact grounds the field] l l

i' e The K-1 relay also closes a K1 contact ("a" contact) which allows the K2 relay to apply field flashing l voltage to the generator o K3 relay is then energized when AC current is ;

! sensed on the output of the generator and opens a '

"b" contact deenergizing the K2 relay which

! removes the field flashing circuitry i

4 e The generator is then self exciting as long as the l Low Speed Relay (LSR) is kept energized j

~

.

When an Auto or Manual diesel shutdown signalis generated and " Stopping Relay" is energized o 'Re " stopping relay" energizes the governor shutdown solenoid and opens the #5 relay contact o The K1 relay is deenergized which closes the K1 contact, shorting out the field and the generator output goes to zero

/ EDG Rev.1 Page 11 of 68

.. . - - . =- . . . .

LESSON BODY KEY AIDS e If an Auto start comes in with the stopping relay 4

energind the field will still re-flash because the 4ax contact (in parallel with #5 - relay contact) will shut and thus energize K1 relay (the 4ax relay is energized via 4a relay "b" contact) o The 4a relay is deer-y ed by auto Stan/R"tGB switch o The 4a relay is deenergized by Start / Diesel switch

e. Generatortrips -

OBJ. #9 (1) Reverse power (2) Overpower e After tripping on reverse power or overpower, the

" FLAG' on the associated relay must be reset to remove trip signal from breaker (3) Circuit breaker amptector (overcurrent) 4

The overcurrent breaker trip will be indicated by

both the red and green breaker indicating lights being energized

After tdpping on overcurrent the breaker must be reset. Tais is accomplished by:

o Placing the RTGB control switch to " Trip" or o Depressing the mechanical tdp button on the front of the circuit breaker o Note: The redbreakeindication Kgkt willgo oct signaling that "NESET" has been uccomplished (4) "Emer Gen A/B mtto trip" annunciator TP-EDG-13 (APP-010-A1) OBJ. #9

, EDG Rev.1 Page 12 of68

i . ,' ,

95-2 NRC EXAM - SENIOR ' REACTOR OPERATOR

. 52. RM-09 001 j Given the following plant conditions

1 l ,

A liquid waste release is in progress ,.9,,s/ ,

, u' i e R-18, Liquid Waste Disposa! System Effluents, fails #j;i 4

f Which ONE (1) of the following describes the required actions in accordance with technical j Specifications? >

1 Sin ,

< VA. The release may continue with sampling and calculations. 'po J

f j B. Releases may continue for up to 30 days while repairs are in progress with no additional

! actions.

l / 'C. Terminate the release, no further releases are allowed until R-18 is repaired.

! '., F 2 D. Terminate the release, and inform the NRC and local officials of the release within 24 (w hours.

,i i a 1 1,

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' K/A 073000.K5.01 2.5/3,0 e7 R A p ( A r r e d T //6 DR Y i 1

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A Question 52 of 100 t

I L.

95-2 NRC EXAM - REACTOR OPERATOR i r i

52. RM-09 001 ,

Given the following plant conditions

A liquid waste release is in progress e R-18, Liquid Waste Disposal System Effluents, fails i

Which ONE (1) of the following describes the required actions in accordance with technical i

} Specifications?

VA. The release may continue with sampling and calculations.

B. Releases may continue for up to 30 days while repairs are in progress with no additional actions. i C. Terminate the release, no further releases are allowed until R-18 is repaired.

D. Terminate the release, and inform the NRC and local officials of the release within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i a ,

i l

l K/A 073000.K5.01 2.5/3.0 SD-019 !

New Question l

I i

Question 52 of 100

. . . ,. - - _. . . . . . - . - - ..-.. - - - . ~ - . ~ . - - . - . . . - . - . - . - - - . - . - . - . . . - .

TABLE 3.5-6 RADI0 ACTIVE LIQUID EFFLUENT FO!11TORING INSTRUMENTATION Release Pathway / Instrumentation MCO' Required Action

1. Liquid Radwaste Effluent '

Discharge Line .

a. Monitor (RMS-18) provides l With the number of channels operable less than the MCO requiregents:

automatic termination of .

release upon exceeding a. Exert best efforts to return the instruments to operable status.within ,l alarm / trip setpoint 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent' Release Report why the inoperability was not- .

corrected in a timely manner in accordance with Specification 6.9.1.d and,

b. Effluent releases via this pathway may continue provided that prior to initiating a release:

l. Two independent samples are analyzed in accordance with the Surveillance Requirements of Specification 3.9.1.1 and; ,

2. Two members of the facility staff 2ndependently verify the release l rate calculations and the discharge line valving. e

b. Flow rate measurement 1 With the number of channels operable less than the MCO requirement:

device

a. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not' ' '

corrected in a timely manner in accordance with Specification 6.9.1.d ;

and, . ,

i

~

b. Effluent releases.via this pathway may be continued, provided that the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump performance curves' generated "in situ"'and tank volumes may be used to estimate flow. '

'MCO - Minimum Channels Operable 3.5-20 Amendment No. 86, 103 L aw

._ . _ . . . .. __ ___ _._- ~. .. _ . . . _ _ - .. _ _ . . . . . _ . . . ..

i 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

53. PZR-09 005

Given the following plant conditions:

t i.) i e The plant is at 100% power TeJ '

q e.

RCS pressure is at 2210 #7 i Which ONE (1) of the following describes the status of the Pressurizer heaters and spray valves?

i j PZR Control groun PZR B/U Groun A PZR B/U Group B E7J Sprav Valves

! A. ON ON ON CLOSED

! B. ON ON OFF CLOSED l l C. OFF ON ON OPEN 4

@. ON OFF ON . CLOSED 1

) d I b

i K/A 010000.K6.03 3.2/3.6 l SD@ PFL hede.r. 4 tue t WI com New Question 1' 4

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I Question 53 of 100

i 95-2 NRC EXAM - REACTOR OPERATOR .

l

53. PZR-09 005 ,

j Given the following plant conditions:

i i i'

.

The plant is at 100% power

RCS pressure is at 2210 -

Which ONE (1) of the following describes the status of the Pressurizer heaters and spray valves?

4

) PZR Corittel group PZR B/U Group A PZR B/U Grouo B PZR Splay Valves 2

l A. ON ON ON Cl.OSED e

i B. ON ON OFF CLOSED j C. OFF ON ON OPEN 4A ON OFF ON CLOSED d

a K/A 010000.K6.03 3.2/3.6 SD-59 New Question Question 53 of 100

_ . _ _ _ - . _ _ . . . _ _ _ . _ . ~ . . _ _ . _ _

4 i

[ .

.c -

95-2 NRC EXAM - SENIOR REACTOR OPERATOR t i

i- !

54. AIR-08 001 Given the following plant conditions:

i

!;

The plant is at 50% power i

i i e The Primary Air Compressor (PAC) is in service j t

Which ONE (1) of the following describes the basis for changing the flow path of service

{ water cooling through the primary air compressor intercooler and water jacket? l s.

j- A. A series flow path is provided to ensure adequate cooling with lake water temperature

. > 85 F.

. s

) r77g B. A' parallel Dow path is provided to prevent over-cooling with lake water temperature

,/ < 85 *F.

p .4 4 ,

l ", (. e . e c,gt 'C. A ;cparallel flow path is used to maintain the inlet service water temperature to the i cylinders HIGHER than the incoming air temperature with lake water temperature j > 85 F.

u .

i

@. . A series flow path is used to maintain the inlet service water temperature to the cylinders i lilGHER than the incoming air temperature with lake water temperature..< 85'F.

l d 1 l

i K/A 078000 Kl.04 2.6/2.9

!' OP-903, 905.

Modified Question ,.

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,r, I i Question 54 of 100

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.,- 95 2 NRC EX AM - REACTOR OPERATOR o , j i

'57. AlR-08 001

Given th'e following plant conditions: l e The plant is at 50% power .

t j'

The Primary Air Compressor (PAC) is in service '

j Which ONE (1) of the following describes' the basis for changin;, the flow path of service water cooling through the primary air compressor intercooler an i water jacket? l 1

t A. A series flow path i.s provided to ensure adequate cooling with i..Pe water temperature !

j > 85 F. !

i 1

B. A parallel flow path is provided to prevent over-couting with lake water temperature 1

< 85

F.

i l C. A parallel flow path is used to maintain the inlet service water temperature to the j cylinders HIGHER than the incoming air temperature with lake water temperature ,

> 85 F.

4 ~

e @. A series flow path is used to maintain the' inlet service water temperature to the cylinders

! HIGHER than the incoming air temperature with lake water temperature < 85 F.

s

.d- l l

i K/A 078000.Kl.04 2.6/2.9 1

OP-903, 905 Modified Question F

i l

1 1

i Question 57 of 100 l

4 - _. - ., _,

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CONTINUOUS USE Section 8.6 Page 1 of 2 ,

. t 8,0 (Continued) INIT r

'8.6 Changing Cooline Water Flow Path to the Primary Air Compressor I ;

8.6.1 Initial conditions

1. This revision is the latest revision available and has been verified against the Revision Status List.

(Print) __

Name Signature Date !

I 8.6.2 Instructions I

1. JJE SW inlet temperature is less than 85'F, TjiQi perform the following to line up for series flow:

1) Close SW-613, PRIMARY AIR COMPRESSOR JACKET COOLER BYPASS.

2) Close SW-767, INTERCOOLER BYPASS TO PRIMARY AIR COMPRESSOR LP CY1.INDER.

3) Open SW-768 INTERCOOLER DISCHARCE TO PRIyYAIRCOMPRESSORLPCYLINDER.

1 l

I i'

O OP-903 Rev. 55 Page 42 of 120

.. __._-..._..__._.__._...._.~.-__._.-.____..__._.____._m_. _ . _ . _ _ . . ._

C Section 8.6 Page 2 of 2

!. . 8.6.2 (Continued) INJ1 i

2. .LE SW inlet temperature is greater than or equal i to 85'F, Ilijj;N perform the following to line. up

for parallel flow

, 1). Open SW-613, PRIMARY AIR COMPRESSOR JACKET COOLER BYPASS.

2). Open SW-767, INTERCOOLER BYPASS TO j PRIMARY AIR COMPRESSOR LP CYLINDER. i 3). Close SW-768, INTERCOOLER DISCHARGE TO f

j PRIMARY AIR COMPRESSOR LP CYLINDER, ,

Initials Name (Print) Date Performed By: ]

Approved By:

Superintendent Shift Operations Date

.O- ,-

H

' l i

i 5

I' OP 903 Rev. 55 Page 43 of 120 n

, _ - , . . - , . - . , . _ _ _ l

4

. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR l

55. AOP-017-05 001 i

Given the following plant conditions:

The unit is at 100% power I

A total loss of instrument air occurs

No Operator actions are taken

Which ONE (1) of the following describes the primary side plant response to this loss of air

pressure?

4 i ,

4 A. The spray valves fail ope resulting in an SI actuation. ;

a j 4. the charging pump becomes air bound.

VCT level decreases to zeroh C. Seal injection flow increas ue to normal charging path being isolated.

D. Letdown line flashin > due to loss of Non-Regenerative Heat exchanger temperature l control. I J l b

K/A 078000.K3.02 3.4/3'6 AOP-017, Attachment I pg 1 1 New question AY O m7 rt N L 4 7EF vt Q u G1oe ~

l 1HE Tce t a r"C

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Question 55 of 100

.o & J-s &9nJEhu,A - ebkm k4 1- ~M,,2a-- <M4e hlb-4 6 ,"A+- 4\34 4 A e b = h- 4 4A B A--A4 6 er,W-4 -e64 a A.- m Rsv. 17 2

A0P-017 IDSS OF INSTRUMENT AIR Page 35 of 58 4

!' INFORMATION USE ATTACHMENT 1

! FAJOR COMPONENTS AFFECTED BY LOSS OF IA

{ (Page 1 of 5)

1. Chemical and Volume Control System Components FAIL POSITION

a. FCV-113A, BA TO BLENDER OPEN

, b. FCV-114A, PW TO BLENDER

CLOSED

c. FCV-113B, BLENDED MU TO CHG SUCTION CLOSED

., d. FCV-114B, BLENDED MU TO VCT CLOSED i

e. HCV-105, BORIC ACID TK B RECIRC CLOSED l f. HCV-110, BORIC ACID TK A RECIRC CLOSED i

g. LCV-115A, VCT/HLDP TK DIV FAILS TO VCT 4

h. LCV-115B, EMERG MU TO CHG SUCTION CLOSED

1. CVC-310A, LOOP 1 HOT LEG CHG OPEN J . CVC-310B, IDOP 2 COLD LEC CHG OPEN f k. CVC 311, AUX PZR SPRAY CLOSED

1. LCV-460 A & B, LTDN LINE STOPS CLOSED

m. CVC-200 A, B & C, LTDN ORIFICES . CLOSED

n. CVC-204 A & B, LTDN LINE IS0s CLOSED

o. TCV-143, VCT/DEMIN DIV FAILS TO VCT

p. TCV-144, NON-REG HX OUTLET TEMP CONTROL OPEN

q. PCV-145, LETDOWN PRESSURE PCV OPEN

)

. r,, HCV-121, CHARGING FIhW OPEN

! s. CHARGING PUMP SPEED CONTROL FULL SPEED

t. CVC-303 A, B & C, SEAL LEAKOFFS OPEN

u. HCV-137, EXCESS LTDN FLOW CLOSED O' (CONTINUED NEXT PAGE)

R3v. 17 AOP-017 LOSS OF INSTRUMENT AIR ,

k

, Pega 36 of 58

'O' INFORMATION USE ATTACHMENT 1 MAJOR COMPONENTS AFFECTED BY LOSS OF IA (Page 2 of 5)

1. (CONIINUED)

CLOSED

v. CVC-387, EXCESS LTDN STOP

w. CVC-389, EXCESS LTDN DIV FAILS TO VCT

x. CVC-307,-PRI SEAL BYP ISO CLOSED FORCED LOW

y. CVCS ICT LEVEL CONTROLLERS

2. Component Cooling Water System Components FAIL POSITION

a. CC-739, CCW IROM EXCESS LTDN EX CLOSED

3. Containment Ventilation System Components FAIL POSITION CLOSED

a. CV VENTILATION IS01ATION VALVES 4 Feedwater and Condensate System Components FAIL POSITION CLOSED

a. FEED REG VALVES

b. FEED REG BYPASS VALVES CIDSED

c. LCV-1417A, HOTWELL LEVEL CONTROL VALVE OPEN

d. LCV-1530A, HEATER DRAIN TANK LEVEL CONTROL VAINE AS IS'

e. LCV-1530B, HEATER DRAIN PUMPS SUCTION DUMP TO CONDENSER OPEN

5. Instrument Air System Components FAIL POSITI('l

a. PCV-1716, INSTRUMENT AIR ISO TO CV CLOSED l

6. Isolation Valve Seal Water System Components FAIL POSITION

a. PCV-1922 A & B, IVSW AUTO HEADER ISOLs OPEN

7. Main Steam System Components FAIL POSITION

a. MAIN STEAM ISOLATION VALVES CLOSED

b. STEAM LINE P0kVs' CIDSED l

l l

l

. Rsv. 17 AOP-017 IDSS OF INSTRUMENT AIk !

Page 37 of 58

')

t (d .INFORMATION USE l ATTACHMENT 1 l

MAJOR COMPONENTS AFFECTED BY IDSS OF IA (Page 3 of 5) ;

8. Primary Sample System Components FAIL POSITION [

f a. PS-956 A through H, PRIMARY SAMPLE IS01ATIONS CLOSED i 9 Radiation Monitoring System Components

FAIL POSITION

a. RMS-1,2,3 & 4, R-11/R-12 ISOL VALVES CIDSED

10. Reactor Coolant System Components FAIL POSITION t i

a. PCV-455 A & B, PZR SPRAYS CLOSED

b. RC-544, RV F1ANGE LEAKOFF OPEN i

c. RC-516 & 553 PRT TO GAS ANALYZER CLOSED f

d. RC-519 A & B, PW TO CV IS0s CLOSED l 0

e. RC-550, PRT NITROGEN SUPPLY CLOSED

11. Residual Heat Removal System Components FAIL POSITION I

a. FCV-605 RHR HX BYPASS FIDW CIDSED

b. HCV-142, PURIFICATION FIDW CIDSED

c. HCV-7.- 'lR HX DISCH FLOW CLOSED

, l l 12. Safety Injection System Components FAIL POSITION l

a. SI-855, ACC NITROCEN ISO CLOSED l

b. SI<856 A & B, SI PUMP RECIRCS OPEN !

c. SI-850 A, B & C, SI ACCUMULATOR TESTS CLOSED l

d. SI-850 D, E & F, COLD LEG INJ TESTS CIDSED

e. . SI-851 A, B & C, SI ACCUEULATOR MAKEUPS CIDSED

f. S1-852 A, B & C, SI ACCUMUTATOR DFAINS CLOSED

g.51-853 A, B & C, SI ACCUMUIATOR VENTS CLOSED I

t

. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR l

l

56. RHR-10 005 l Which one (1) of the following describes an interlock in the RHR system hich is designed to i

prevent overpressurizing the lines to the RWST?)

l A. RiiR-750, RHR Hot Leg Suction Isolation, caniwt be opened if RCS pressure on PT-403 exceeds 535 psig.

B. RHR-751, RHR Hot Leg Suction Isolation, cannot be opened if SI-862A, RHR Suction from RWST, is open.

W. RHR-750, RHR Hot Leg Suction Isolation, cannot be' opened if SI-863A, RHR Hx Outlet to RWST, is open.

D. RHR-751, RHR Hot Leg Suction isolation, cannot be opened if RCS pressure on PT-403 exceeds 535 psig.

I c

l K/A 005000.K4.07 3.2/3.5 !

OP-201 New Question l

pp gre mW-l 7

Kysse>d i l

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i l

Question 56 of 100 l

I l

I m ,

, 4.0 PRECAUTIONS AND LIMITATIONS

1. Reactor Coolant System temperature and pressure shall be less than 350*F and 375 psig before the Residual Heat Removal System is put in service, and the RHR system will be removed from service before RCS pressure and temperature are raised above these values.

2. To prevent boiling the CCW liquid contained in an RHR HX, CCW flow should not be isolated to an RHR HX when the temperature of the RHR System is greater than 200*F. (CR 95-00565)

3. Neither RHR-750 nor RHR-751 will open unless the following conditions are satisfied:

The breakers for SI-862A and B are closed.

The breakers for SI-863A and B are closed.

e The control power switches for SI-862A and B are in NORMAL, e The control power switches for SI-863A and B are in NORMAL.

e Valves SI-862A and B are closed.

Valves SI-863A and B are closed.

e RCS pressure is less than 465 psig.

4. SI-862A & B, and SI-863A & B are interlocked so they cannot be opened unless the RHR loop pressure is less than 210 psig.

5. When the Residual Heat Removal System is providing Core Cooling AND' seal injection flow is desired to maintain a positive AP across the Thermal Barrier of the Reactor Coolant Pumps, letdown flow through HCV-142 and PCV-145 should be maintained to provide makeup to the VCT.

6. When running RHR Pumps with SI-863A and/or SI-863B open, RHR-744A and RHR-744B should be closed to prevent excessive RHR pump runout.

tu 6

- r4

\ "'] w.n y _ n l $yo ,

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OP-201 Rev. 28 Page 7 of 59

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

57. CCW-10 002 -

Given the following plant conditions:

i e The plant is at 100% power ,

d

A Safety Injection and Containment Spray occur

A Loss of Off-Site power occurs

, Which ONE (1) of the following describes the response of the Component Cooling Water

System?

l VA. Neither B'Sr C CCW pumps start automatically and can NOT be started manually from j 4

the control board. !

B. Neither B C CCW pumps wiH start automatically but can started manually from the j control board.

j C. Only Pump B will auto start after the Diesel Generator breaker closes.

! D. Only Pump A will auto start proyiding the D.S. bus is energized.

, a i

)

K/A 008000.K4.01 3.1/3.3 Dwg 5379-3368 sht 13 of I8 Modified Question I

No , e 4 h cleck i

Question 57 of 100

p 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

.t ,

t 4

58. EPP-008-05 002 t Given the following plant conditions:

4

You have just entered EPP-008, " Post LOCA Cooldown and Depressurization"

You are confirming that Natural Circulation exists Which one (1) of the following conditions wuld bWrovide indication of natural I

, circulation?

W A. The AT across the S/G's are 10*F and slowly decreasing.

i 4

- B. S/G Pressures are slowly increasing. /

! C. Core exit TC's are slowly increasing.'

M'T PLN/6 v'D.

~

RCS subcooling based on core exit TC's is 40 y incre (

d I

i 000011.EKl.01 4.1/4.4 EPP-008 Supplement E New Question utsi davi A c, Question 58 of 100

e

.s

Rev. 15 EPP Supplements SUPPLEMENTS Page 24 of 71 STEP -

INSTRUCTIONS RESPONSE NOT OBTAINED COlfrINUOUS USE Surolement E Magglal Circulation Verification (Page 1 of 1)

1. Check Natural Circula. ion Status Increase dumping steam.

. As Follows:

s'

Steam pressure - STABLE OR DECREASING 3.

1 M

RCS subcooling - GREATER THAN 35'F [55 *F]

i i

j

Core exit T/Cs - STABLE OR DECREASING 1

. A152

RCS hot leg temperatures -

l STABLE OR DECREASING i

~

AtH2 i

4

RCS cold leg temperatures -

i TRENDING TO SATURATION TEMPERATURE FOR STEAM

! PRESSURE lI

2. Return To Procedure And Step In Effect 2

- END -

L i

y

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

59. EDG-14 003 Given the following plant conditions:

a The control room has been evacuated due to toxic gas.

. The Control Roorn Supervisor has sent an operator to start the "B" EDG.

h w e An operator is stationed at the "B" Diesel Generator control panel.

The Diesel is ready for loading.

Which ONE (1) of the following describes theJ4AX1MUKT~ rating for the "B" EDG?

VA. 2750 KW for two hours.

B. 2750 KW for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. 2500 KW for one hour.

D. 2500 KW for 30 minutes.

Ss, f

2 , 7 <> o.e nu, a

no F'o- fL. an, g;,ld.l j K/A 000068.K2.07 (3.3/3.4)

OP-604, P&L #7 ro> *- j Modified Question 1

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i l

l Question 59 of 100

l

!* 95-2 NRC EXAM - REACTOR OPERATOR I

66. EDG-14 003 Given the following plant conditions:

The control roorn has been evacuated due to toxic gas.

The Control Room Supervisor has sent an operator to start the "B" EDG.

An operator is stationed at the "B" Diesel Generator control panel.

t

The Diesel is ready for loading. ;

Which ONE (1) of the following describes the MAXIMUM rating for the "B" EDG7 VA. 2750 KW for two hours.

B. 2750 KW for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. 2500 KW for one hour. -

I D. 2500 KW for 30 minutes.

a K/A 000068.K2.07 (3.3/3.4)

OP-604, P&L #7 Modified Question l

Questica 66 of 100 t

l t- -_ , -.

~ _ . - . . . - . - . _. - .. - . . - . _ _ . = . . - _ . - . _ . . . - _ _ - . . , . ~ . _ . - - _ - - . _ . _

t 4.0= (Continued) i .. .

l- 5. When the Imbe Oil Filter pressure differential exceeds 10 paid.

f cartridge replacement is required.

!' ~

i '6. When the Fuel Oil Filter pressure. differential exceeds 10 psid, l j cartridge-replacement is required. '

7. Diesel Generator loads shall E exceed ratings of 2,500:KW.for 1

[

continuous operation.

'2750'KW'shall g be exceeded.

~

4 e

e Operation at 2750 KW for more than 2: hours within a124 hour !

!~ period shall M occur.

i-

. e 4,000 amps on the Generator shall M be exceeded.

I

j. 8. %e Diesel Generators should not be set for automat.ic start after l draining the EDG Fuel Oil System. Always perform a manual start to ensure the EDG Fuel Oil System is not airbound afror refilling.

. 9. When the Diesel is running and the Trips Defeat Key Switch is in

{

the TRIPS DEFEATED position; the Diesel should be manually tripped if a condition exis'es that would automatically trip the Diesel.

-These conditions are: (Rail 92R0044) ]

e Coolant Temperature High 205'F i e Crankcase Pressure + 0.5 inches H O 2 q e Lube Oil Low Pressure 18 psig e . Coolant Low Pressure 12 psi 5

10. Synchroscopes will be left OFF unless in use for synchronizing to prevent damaging by inadvertent energizing of two synchrorcopes.

i l

i

+ _. .

l OP-604- Rev. 34 Page 12 of 154

, . ~ , . . . - .

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

60. AOP-001-06 002 l Given the following plant conditions: i l Ab b e The unit is shutdown # conducting rod testing following a refueling outage ;

l

ROD P6 in Control Bank "B" is at 228 steps

! i e The RO inadvertantly. pulls the rod out vice inserting-it r directed--

Ly f u . ,) 1 % c n :1c4 i

Which ONE (1) of the following describes the potential problern associated with demanding j RCC(s) to step past 228 steps? i I

l n A. The control rod reactivity effect goes to zero so inward motion will not initially add

' l negative reactivity.

0 l i

B. The control rod may move out of the rod guide tuoc and misalign preventing reinsertion into the assembly. l l . !

W. The staionary gripper may fail and result in a misaligned, partially inserted or dropped )

rod.

l D. The Pulse-to-Analog converter will stop counting at 228 steps and actual rod position will

- be unknown.

c pcpow < !/ dof Fott 3rona K/A 000005.EK3.06 3.9/4.2 N w Q stion b# NO r

_f fAnupc3 ff5y f c p a d. l d b lI If a C 70 Q R L t b 1 t t) i Question 60 of 100 L

l

- - - . - . . . . -. . - . . - . . . - - . . - . - . . - - - . - _- ~ ... - -. -. .

95-2 NRC EXAM - REACTOR OPERATOR i

i i

S

. 67. AOP-001-06 002 i Given the following plant conditions: i i

e The unit is shutdown conducting rod testing following a refueling ourage

ROD P6 in Control Bank "B" is at 228 steps I = The RO inadvertantly pulls the rod cut vice inserting it as directed 1 g .Which ONE (1) of the following describes the potential problem associated with demanding

RCC(s) to step past 228 steps? .

$ l A. The control rod reactivity effect goes to zero so inward motion will not initially add l

negative reactivity. ,

B. The control rod may move out of the rod guide tube and misalign preventing reinsertion j into the assembly.

l 1

vC. The staionary gripper may fail and result in a misaligned, partially inserted or dropped rod.

D. The Pulse-to-Analog converter will stop counting at 228 steps and actual rod pesition will be unknown.

c.

K/A 000005.EK3.06 3.9/4 2 l AOP-001 !

New Question l

1 l

l i

1 1

Question 67 of 100 ;

Rsv. 10 A0F 001 MALFUNCTION OF REACTOR CONTROL SYSTEM Page 15 of 73 STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED l SECTION A l

DROPPED ROD l (Pay,e 6 of 23)

14. Check GEN VARS Indication - Adjust reactive load using the WITHIN LIMITS FOR CURRENT PLANT VOLTAGE ADJUSTER.

CONDITIONS

{ 15. Check REGULATOR BALANCE - Adjust REGLATOR BALANCE as

BETREEN +2 TO -2 follows

f

Decrease by momentarily placing the FIELD CURRENT

- ADJUSTER to RAISE.

2B e Increase by momentarily placing the FIELD CURRENT ADJUSTER to LOWER.

- - - - k**********************************************************

CAUTION Equipment repairs or manipulations to correct the cause of the dropped rod prior to procedural direction could inadvertently withdraw the dropped rod.

- www**w******************w***********************************************

16. Notify Reactor Engineering AND I&C Personnel To Perform The Following:

a. Verify.the status of the dropped rod

b. Investigate the cause of the dropped rod

c. Determine appropriate recovery actions l

i 95-2 NRC EXAM - SENIOR REACTOR OPERATOR l

I (4 61. AOP-002-06 002 i i 1 Given the following plant conditions: j l

l Initial plant power was at 100%

l !

A runback to 70% occurred

! APP-005-C5, Rod Banks A/B/C/D Lo-l.o Limit is received 1

l Which ONE (1) of the following cescribes the correct operator response to restore control rod i position? )

l A. Place rods in Manual and immediately withdraw rods to clear the alarm. !

B. Emergency Borate using MOV-350, "BA to Charging Pmp Suct.", to restore temperature j within 15 minutes.

f C. Emergency Borate IAW OP-301, " Chemical and Volume Control", and withdraw control rods to restore rod position within 30 minutes. i 7 l

@. Borate using OP-301, " Chemical and Volume Control", and withdraw control rods to

_ restore rod position within I hour.

d .

I l

K/A 000024.A1.05 (3.1/3.2) ,

APP-005-C5 OP-301, Sect. 6.2.3 ;

Modified Question I gg ;n [o H e ,

, , ,; I 1.e b

y 3 0 r , '-

C 0 1 p in I '-

I Question 61 of 100

- . - . . . . - - . - - . . . - . - ~ . . . . . . . . . - - . - - - . . . . .. .. _ .~. - . _

95-2 NRC EXAM . REACTOR OPERATOR-i i.

, 68. AOP-002-06 002-Given the following plant conditions:

Initial plant power was at 100%

l l 'A runback to '70% occurred APP-005-C5, Rod Banks A/B/C/D Lo-Lo Limit is received E Which ONE (1) of the following describes the correct operator response to restore control tod i position?

1 A. Place rods in Manual and immediately withdraw rods to clear the alarm.

l ,

4

B. Emergency Borate using MOV-350, "BA to Charging Pmp Suct.", to restore temperature j within 15 minutes. ,

! C. Emergency Borate IAW OP-301, " Chemical and Volume Control", and withdraw contro!

rods to restore rod position within 30 minutes..

L

. W. Borate using OP-301, " Chemical and Volume. Control",' and withdraw control rods to ,

restore rod position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

]

.d -

l l

l K/A 000024.A1.05 (3.1/3.2) t APP-005-C5 l OP-301, Sect. 6.2.3 i Modified Question !

4

)

1

). )

l i l I

l j- !

i .

I i 4

4 ,

L Questian 68 of 100 I y ;

__ ,e

~

APP-003-C5 Pegn 1 of 1

'A M '

ROD BANKS. A/B/C/D ID-ID LIMIT *** WILL REFIASH ***

b*

AUTOMATIC At'TIONS

1. None Applicable PAUSE

1. Excessive dilution of Boron Concentration

2. Malfunction of Automatic Rod Control System I

3. Instrument Failure p 4. Plant transient requiring deep rod insertion.

. 5. OST-011 in progress (expected alarm) 1 OBSERVATIONS I

1. Control Bank RFI and Bank Step Counters )

2. Power Range NI

3. RCS Tavg l

4. RCS Makeup Flow l j' 69ILO.Ei

. 1. II OST.011 is in progress THEN verify alarm clears when red bank is i i returned to initial position.

2. II en RCS. dilution is in progress, IKEN STOP the dilution, i

3. Borate the RCS using OP-301, Chemical and Volume Control System, while simultaneously withdrawing Control Rods to clear the alarm within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months I

l of receipt. l 4; E a malfunction of Makeup Control is indicated, THEN refer To AOP-003. j

5. E a malfunction of Reactor Control is indicated, TREN refer To AOP-001. I

6. Apply Technical Specification 3.10.1.3.

3. DEVICE /SETPOINTS

1. Rod Banks A and B TC-409D, E/less than or equal to 210 steps 1

2. Rod Banks C and D., TC 409F, L/ refer to Curve 1.9 in the Curve Book.

0 E,_SSIBLE PIANT EFFECTS

} 1. Lo.ds of required Shutdown Margin

2. Entry into Tech. Spec. LCO Action REFERENCES

1. Plant C2rve 1.9 A and B j

2. OST-Oll, Rod Cluster Control Exercise & Rod Position Indication.

I i

/

V APP-005 Rev. 14 Page 20 of 42 i

~^

. ., pug &3M32?@6-% -

I

' ,. . (g.n N.e.u.w%.Ryere.4mWaeye$V"/f4W~D *p

_ .2.#/ s , a;c.s..: ;. j e. y p s p . n y g eb p.6ft'A Srction 6.

.Mer fG ~9 "M f

/

Pag 3 6 ef 12 +

6.2.2 (Cmtinued) ph i 20. H a chemical analysis et the RCS Boron concentration indicates an

, N.,.) .

appreciable deviation between actual RCS Boron concentration and the RCS Automatic Makeup Boron concentration, ,T,tig adjust the Automatic Makeup be:on concentration in accordance with Section

, 6.2.1.

6.2.3 Boration

1. Verify the Boric Acid Tanks contain, sufficient 20,000 to 22,500 ppa boric acid solution to permit boration without reducing the tanks total liquid volume below the 3080 gallons required to achieve cold a

shutdown conditions.

2. Verify Attachment 9.1 has been completed to the extent necessary to support operation of the RCS Makeup System.

1

' j N I Changes in RCS boron concentratien that are required as a result of normal i 3 :

leakage makeup, fuel burnup, load changes, or xenon transients following load changes are considered normal operational changes to RCS Boron concentration.

t

3. E the boration is going to be performed for reasons OTHER than normal operational changes to RCS Boron concentration THEN obtain permission from the Superintendent Shift Operations or SCO.

4

4. Verify the most recent RCS boron sample analysis is available.

5. Determine the rate and magnitude of the RCS boron concentration l change required to accomplish the desired reactivity chan5e.

6. Estimate the total voluue of boric acid required from the proper i boration nomograph.

7. Place the RCS MAKEUP MODE selector switch in BORATE.

8. Place controller FCV-ll3A, BORIC ACID FLOW, in MAN.

9. Set the BORIC ACID TOTALIZER, YIC-ll3, to the desired quantity as follows: .

1. Depress BUTTON "A".

2. Depress "CLR" BUTTON.

3. Key in the desired quantity AfiD deprersa the "ENT" BUTTON.

I OP-301 Rev. 53 Page 30 of 157 i

l i

Retion 6.2.3 (Continued) j The following step will open FCV-113A, BA TO BLENDER, and FCV-113B, BLENDED MU TO ]

CHG SUCT, and will start a Boric Acid Putsp.

10. Place the RCS MAKEUP SYSTEM switch in START.

11. Manually adjust controller FCV-113A, BORIC ACID FLOW. using the i

increase and decrease pushbuttons to establish the desired Boric

- Acid flow. -

12. IE any of the following conditions occur, M stop the boration by placing the RCS MAKEUP SYSTEM switch in STOP:

Rod motion is blocked.

i e Rod motion is in the wrong direction.

Two increases 4

Suberitical Count Rate increases by a factor of two.

The desired condition is exceeded.

13. M the desired amount of Boric Acid has been added to the RCS, l M verify the following:

p,

FCV-113A,'BA TO BLENDER, closes

(

FCV-113B, BLENDED MU TO ChG SUCT, closes

The BORIC ACID PUhP stops j e The RCS MAKEUP SYSTEM is off I

14. IE desired, M flush the Boric Acid flow path IAW section 6.2.4. ll

. 15. Record, in the Control Operator Log, the total amount of Boric Acid added during the boration operation as indicated by BORIC ACID i TOTALIZER, YIC-ll3. j

16. Return the RCS Makeup System to automatic operation by performing fl b the following:

l

1) Place controller FCV-113A, BORIC ACID FLOW, in AUT0. ;

} 2) Place RCS MAKEUP MODE selector switch in AUTO. ,l 1

I i

< i l

l O 1 U '

i OP-301 Rev. 53 Page 31 of 157

, ' wppjym ya. g je u g 4.4 -,

~~

~4

~-

2.~

<* 7. ,._l,

- .?.S..;f.J .f.fs'%_

LN vg' .s?f*

. ;..,;4. :w.

. . . . . 4Mt.".,.&&. .-. .,qy, .yw,:.*%.. ~'

S:ction 6.2

" ~ '

l l P:ga 8 of 12

[ 6.2.3.16 (centinued)

}O M

< ******; ;. ;. ;. ;. ; :. ;. ;. ; :. ; ; .t+**** * :. ;. :. ;. ;. ; :. ;. ;. ; ;. ;. ;. ;. ;. ; :. ;. ; :. ;. . ****

i

, .W2Il9.li f Failure to position the RCS MAKEUP SYSTEM switch to the START position will i

prewnt the Boron concentration control system from operating properly.

- - - - ww*******************a********************************************

i a

j- 3) Place RCS MAKEUP SYSTEM switch in START.

4) Depress the "CLR" BUTTON to reset BORIC ACID TOTALIZER, i YIC-113 to zero.

i l '- !!QIK l e At least 10 minutes should be allowed for mixing before samples are taken.

4 Additional time should be allowed depending on magnitude or the concentration l

3 change and accuracy demanded.

Changes in RCS boron concentration that are required as a result of normal leakage makeup, fuel burnup, or xenon transients following load changes are considered normal operational changes to RCS Boron concentration.

17. J.E the horation was performed i'or reasons OTHER than a nonnal operational change to RCS Boron concentration QE a sample is desired, TllEli request chemical analysis of RCS boron concentration.

18. IF the desired rod position QE boron concentration is ]!QI achieved, IliEl! repeat Steps 6.2.3.5 through 6.2.3.17. l t 19. lE a chemical analysis of the RCS Boron concentration indicates an appreciable deviation between actual RCS Boron concentration and the RCS Automatic Makeup Boron concentration, IllE!! adjust the Automatic Makeup Boron concentration in accordance with Section 6.2.1.

i l

i i

1 l

l OP-301 Rev. 53 Page 32 of 157 l

] 95-2 NRC EXAM - SENIOR REACTOR OPERATOR i ,

Q. AOP-012-06 001

\

l

' hich ONE (1) of the following describes the condenser vacuum valuephat will result in an !

a omatic turbine trip?

A. 24. , inches lig

'S. 20 in esHg -> 4% b L e) , v / Le d' ;

, M> .

! J' [C. 5.5 inches ekpressure

J +

D. 8 inches backptyssure b \s i -

K/A 000051.EA2.02 3.9/4.1 7 i AOP-012, Step 16 APP-008-85 l

New Ouestion !

A=5.5" backpressure in vacuum ,

C=5.5" backpressure from step 13 j D=8 is trip for oil pressure and number 'e close to actual vacuum trip setpoint I ogg osa p3o9 c i, ; I . r , 1 ~.:<. 8 to n f.2 1"-l.- r i. ;;.

p,E nc a r LEvCL j

l I

N - f e9 Q !

l Question 62 of 100

-. - - . . - . - . - . . - . - . . . . - - . - . - . - . - . - . . . . ~ - - - . . - . . - - . - . . . . .

95-2 NRC EXAM '- REACTOR OPERATOR l .

,, 69. AOP-012-06 001

- Which ONE (1) of the following describes the condenser vacuum values that will result in an

l automatic turbine trip?

i j-

! ~ A. 24.5 inches Hg i i

i l I

j 'v'B. 20 inches Hg -l e

t' ..

C. 5.5 inches backpressure D. 8 inches backpressure

f. ,

i b i

l i.

K/A 000051.EA2.02 3.9/4.1 i AOP-012, Step 16 !

!. . APP-008-B5 i x

, New Question j

! A=5.5" backpressure in vacuum '

!- C=5.5" backpressure from step 13 I D =8 is trip for oil pressure and number is close to actual vacuum trip setpoint i

t i.

t t

J i

l' e

Question 69 of 100

PARTIAL IDSS OF CONDENSER VACUUM OR CIRCUIATING Rw. 8 l AOP-012 WATER PUMP TRIP Page G of 25 t

STEP -

INSTRUCTIONS RESPONSE NOT OBTAINED i

13. Check Condenser Back Pressure On Perform the following:

PI-1310 M PI-1311 - LESS THAN

, 5.5 INCHES Hg ABS a. Reduce Turbine load as

, necessary to maintain j Condenser back pressure less

than 5.5 inches Hg abs. ;

b. Notify Ioad Dispatcher of the load reduction.

i

c. IE Condenser backpressure can HQI be maintained less than

5.5 inches Hg abs, "DiEH begin j plant shutdown using GP-006,

{ Normal Plant Shutdown From :

. Power Operations To Hot )

Shutdown, while continuing i

, with this procedure.

1

14. Verify The Following Valves - !

j C1DSED !

1 1 2 \

f a MS 70A, VACUUM BREAKER VALVE 1 l

i e MS-708, VACUUM BREAKER VALVE 2 l

I 15. Verify All Available Circulating j Water Pumps - RUNNING

16. Check Condenser Back Pressure On Go To Step 19.

j PI-1312 M PI-1313 - GREATER i THAN 10 INCHES Hg ABS '

i

, 17. Check REACTOR TRIP FROM TURB Perform the following:

s BIDCK P-7 Status Light -

! ILLUMINATED a. Trip the Reactor. !

i

b. Go To Fath-l. ]

l A

= i

1 j

APP-008-B5 j

j

.

ALARM )

J l

[ h CONDENSER LO VACUUM TURB TRIP *** NO REFLASH ***

V AUTOMATIC ACTIONS

l. Turbine TRIP

2. One minute after Turbine TRIP: I

1) Generator LOCKOUT i 2) ' Automatic Bus Transfer CAUSE

1. Loss of vacuum Pump 4. Loss of Circulating Water l 2. High Hotwell level due to 5. Condensate Pump Suction Line

, failure of Hotwell Level Expansion Joint ruptured l Controls 6. Excessive Steam Dump ;

3 ~. Loss of Gland Seal Steam i l

I OBSERVATIONS

1. Turbine Valve Positions

2. Condenser Vacuum i O_

ACTIONS

1. If below P-7, Refer to AOP-007, Turbine Trip Without Reactor Trip Below P-7.

i

2. If above P-7, Go to EOP Network.

)

l i

DEVICE /SETPOINTS

.l. 63-AST-1/45 psig (Low Auto Stop Oil Pressure)

AND

2. h3-LYS (SW-1)/19.7 in. Hg, Vac. (10.3 in Hg Abs)

POSSIBLE PLANT EFFECTS

1. Reactor Trip if above P-7

2. Plant Shutdown REFERENCES

1. PATH-1, EOP Network g-~g 2. AOP-007, Turbine Trip Pithout Reactor Trip Below P-7 k, / 3. CWD B-190628, Sheet 711, cable H APP-008 Rev. 10 Page 18 of 55

a * ;

95-2 NRC EXAM - SENIOR REACTOR OPERATOR l l

63. EPP-001-14 003 Given the following plant conditions:

4

The plant was at 100% power i N

A loss of all AC power has occurred.

Emergency Bus E2 was energized by the "B" EDG 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the loss of offsite power.

Offsite power was restored to the plant 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after it was lost.

Which ONE (1) of the following conditions would require the restart of an RCP during -

recovery from the loss of all AC power event?

[

e j A. During implementation of EPP-2, " Loss Of All AC Power Recovery Without SI j Required".

. B. During implementation of EPP-3, " Loss Of All AC Power Recovery With SI Required".

C. When procedurally directed by action steps of EPP-5, " Natural Circulation Cooldown".

@. When procedurally directed by action steps of FRP-C.1, " Response to Inadequate Core Cooling".

j d l gJE JrP6 - @ 44 c o '

[

K/A 000055.G0.12 (3.9/4.u) 'y p p e r y , ,.

WOG/ ERG EOP Background Document, RCP Restart following a Loss of All AC Power.

Modified Question

\1ypt doe 3 DH5 TEST l 4

S IA 10] g ALon n p a

Question 63 of 100

95-2 NRC EXAM - REACTOR OPERATOR 4

4

70. EPP-001-14 003 Given the following plant conditions:

The plant was at 100% power J

A loss of all AC power has occurred.

4

Emergency Bus E2 was energized by the "B" EDG 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the loss of offsite power.

J

Offsite power was restored to the plant 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after it was lost.

Which ONE (1) of the following conditions would require the restart of an RCP during recovery from the loss of all AC power event?

A. During implementation of EPP-2, " Loss Of All AC Power Recovery Without SI .

Required". !

B. During implementation of EPP-3, " Loss Of All AC Power Recovery With SI Required". ;

C. When procedurally directed by action steps of EPP-5, " Natural Circulation Cooldown".

I i @. When procedurally directed by action steps of FRP-C.1, " Response to inadequate Core

Cooling".

d K/A 000055.G0.12 (3.9/4.0)

WOG/ ERG EOP Background Document, RCP Restart following a Loss of All AC F,wer.

Modified Question 1 I

i I

Question 70 of 100

RCP Start Followino Restoration of Seal System Coolino folicwing restoration of RCP seal cooling, the RCP should not be started prior

{ to a complete RCP status evaluation in order to minimize the potential for damage to the RCP. The status evaluation should be performed consistent with the limitations and requirements in the plant specific RCP Instruction and Operating Manual. In general, the status evaluation should consist of the following three elements:

i

1. Upor restoration of RCP seal cooling, manually rotate the pump in accordance with the instruction book limitations and requirements. -

2. Restart the RCP, monitoring vibration and seal leakage rate:

4 4

a. If seal leakage and pump vibration are acceptable, use the RCP for forced circulation cooldown of the plant.

b. If seal leakage or pump vibration is excessive, trip the pump.

3. Upon achieving cold shutdown conditions, disassemble the RCP seals and inspect all seal components including 0-rings, channel seals, No.1 i seal faceplates, etc. '

If an RCP cannot be manually rotated due to containment access limitations or restrictions, the plant should be taken to cold shutdown conditions under natural circulation to permit pump disassembly and visual inspection as part of the status evaluation. An RCP should not be started without this status evaluation since seal misalignment or crud blockage could ' ggravate RCP seal

~

a damage, potentially propagating into RCP seal damage and increased seal leakage flow. However, there is an exception to this requirement for a status evaluation prior to RCP start. The RCP should be started even without a status

___p evaluation if an extreme (red level) or severe (orange level) challenge to a Critical Safety Function is diagnosed via Status Tree monitoring and the operator is instructed to start an RCP in the associated Function Restoration Guideline. Under these conditions, the RCP support systems should be restored to as near normal conditions as possible and the RCP started.

ECA-0.0 9 LP-Rev. IB 0084V

2.2 Transient Analysis Descriotion The response of the Westinghouse nuclear steam supply system (NSSS) to the loss of all ac power initiating event has been analyzed to identify the behavior of important NSSS process variables. This analysis has included computer analyses in which a limited set of potential RCP leak flows, steam generator

, depressurization (cooldown) rates, etc. have been evaluated. In the event of an actual ac power loss at a plant the response of these variables could vary from the conditions assumed in the analyses, and consequently, could alter the quantitative results from those presented herein. Therefore, the primary importance of the analytical information presented in this section is with regard to the phenomena described and not with the absolute magnitude of plant responses. In order to define more clearly these limitations, the important

! assumptions and simplifications used in the analyses are described below.

k RCP Seal leakaae f in all the analyses presented in this section, RCP seal leakage has been '

modeled as a break-in the RCS cold leg at the pump discharge. The break size was selected such that at the noninal operating conditions of the cold leg the l

break flow would equal a pre-selected value of seal leakage rate. i s ;

As RCS conditions changed during the analyses the break flow also changed as dictated by critical flow correlations for subcooled and saturated water. This

~

approach is expected to produce a realistic representation of the response of an RCP seal leak particularly should the seals be degraded by overheating and subsequent erosion of the sealing surfaces, extrusion of 0-rings, etc.

Variations from an actual leak situation are expected to be due only to the l likely overprediction of leak rates characteristic of critical flow l correlations. This deviation is in a conservative direction, however, and in no way would change the qualitative phenomena illustrated by the analyses.

.A more important assumption made in the analyses with regard to pump leaks was that the leak was assumed to start at the pre-selected rate as soon as ac power is lost. No time delay was included to simulate the effects of gradual seal degradation. In an actual loss of ac power event, the seal degradation phenomenon, if it occurs at all, could extend over periods of time lasting from several minutes to hours.

ECA-0.0 10 LP-Rev. IB 0084V

]

The most likely result of this gradual degradation would be a slow increase in pump leakage from only several gallons per minute per pump at the time of reactor trip to potentially larger leak rates at some point later in time.

Unfortunately, all possible scenarios of seal degradation and maximum potential leak rates'could not be analyzed. Therefore, a conservative approach of assuming immediate seal failure and large leak rate was used, and a variety of leak rates up to the maximum estimated leakage were analyzed to i provide bounding cases for all situations.

Manually Controlled Cooldown of the RCS ,

Part of the phenomenology illustrated in analyses of this section relate to the effects of cooling the RCS using steam generator power operated relief valves (PORV) and the turbine-driven auxiliary feedwater (AFW) pump. In the event of a total loss of ac power, control of these systems will likely require manual and possibly local control of both the PORVs and the AFW pump.

Therefore, the dynamics of the cooldown will not be as orderly as would normally be the case with automatic control systems in service. However, in l the analyses no attempt was made to characterize the effects of manual / local j system operation. All cooldown sequences were assumed to proceed orderly at a l constant cooldown rate of approximately 100 F/hr in the steam generater l secondary. In practice such an orderly cooldown may not occur because of (1)

~

the slow' action / feedback loop between the operator in the control room ,

watching plant instrumentation and~the people at the equipment stations operating the pump and valves, and (2) the possible limitations on steaming rates that could result from a requirement to maintain adequate water levels in the steam generators (e.g. narrow range level indication in at least one .j intact steam generator to ensure a secondary heat sink) with only the turbine-driven AFW pump running. As with the situation on RCP leakage, this limitation on the analyses will not affect the qualitative nature of the result but only the absolute magnitudes. The qualitative conclusions regarding the benefits of plant cooldown, therefore, remain unchanged.

I, ECA-0.0 11 LP-Rev. 1 0084V 4 4 y _

I As long as the ultimate limits are observed, faster cooldown rates than those analyzed will be more beneficial and, conversely, slower ones less beneficial.

Decay Heat Power The level of decay heat following any reactor trip will depend on the immediately preceding power history as well as the total burnup histo y of the f core. Again it was impossible to analyze all possible decay heat situations that might exist following a loss of all ac power; however, a set of assumptions was msde that should be representative of a majority of the .

situations that might be exper'ienced and should be conservative for most others. This was done by performing all the analyses using decay heat data representative of the standard data presented in ANSI standard :

ANSI /ANS-5.1-1979 assuming long term reactor operation at full power. No !

uncertaintiss have been included in order to obtain best estimate analyses.

Thus, the results of the analyses presented here should be reasonably i representative of post trip core response fur burnup periods beyond 150-200 days of operation at full power. Trips that occur earlier in core life or i during periods of extended operation at part power would exhibit lower decay heat and therefore would be bounded by the analyses that have been done.

Lower decay heat levels for these situations would result in a reduction in ;

the amount of RCP seal leakage because of a more rapid depressurization of the RCS. Lower decay heat values would also lessen the difficulties of manual plant cooldown using the steam generator PORVs and the turbine-driven AFW ;

pump.

I i

ECA-0.0 12 LP-Rev. 1 0084V

.. . ~. - .- , - . - - . . ~ . . . - . - . . . . - - . _ . . ..

3. . .

I 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

64. FRP-P.1-06 001 l

/ Given the following plant conditions:

The unit is operrting at 100% power l 3

All systems are aligned for normal operations l

\b \

An event occurred in the CV that raised CV pressure to 18 psig l

Path-1 immediate actions were performed

{

( ,

N

The STA has been directed to reset SPDS and monitor CSFST's i

! Which ONE (1) of the following describes an event that could result in a Pressurized Thermal l Shock concern?

l i

i i A. ATWS due to l'oss of all Main Feedwater. j W. Stearn Break inside of Containment.

C. Large Break LOCA. -> TNE l D. Design basis SGTR.

l r a d GSI NC 1 l b n te, i Fa nn t:

K/A 000040.EKl.01 4.1/4.4 FRP-P.1 Step 13 '

i New Question i

Question 64 of 100

y.. _ _ _ .

. i 4

j 95-2 NRC EXAM - REACTOR OPERATOR

~

' I 4

71. FRP-P.1-M 001 '

i Given the following plant conditions: I b

l

The unit is operating at 100% power 1 i y

All systems are aligned for normal operations l l

l

An event occurred in the CV that raised CV pressure to 18 psig

Path-1 immediate actions were performed

The STA has been directed to reset SPDS and monitor CSFST's i Which ONE (1) of the following describes an event that could result in a Pressurized Thermal Shock concern?

i I E i A. ATWS due to loss of all Main Feedwater. !

vB. Steam Break inside of Containment.

]

i C. Large Break LOCA.

i D. Design basis SGTR. l b j i

K/A 000040.EKl.01 4.1/4.4 FRP-P.1 Step 13 ,

New Question !

-i i

Question 71 of 100

l Rav. 9 l FRP-P.1 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK Page 4 of 21 i

STEP -

INSTRUCTIONS RESPONSE NOT OBTAINED

l. Determine If RCS Cooldown Is Due To A Large Break IDCA As Follows:

a. Check both of the following a. Go To Step 2.

conditions exist:

RCS pressure - LESS THAN '

275 PSIG l l

a

RHR flow on FI-605 -

l GREATER THAN 1200 GPM i r

l l

l

b. Reset SPDS E return to procedure and step in effect 1

2. Check RCS Cold Leg Temperature - Go To Step 9.

l DECREASING

3. Attempt To Stop RCS Cooldown As Follows:

l t

a. Verify STEAM LINE PORVs -

l CLOSED i

b. Verify COND DUMPS - CIDSED l

c. Check RHR System - ALIGNED c. Go To Step 4.

FOR CORE COOLING

d. Stop cooldown from RHR Systee l

l l

l l :

I l

l

l 0 95-2 NRC EXAM - SENIOR REACTOR OPERATOR b

_I

\

65, AOP-004-06 001

\

Given the following plant conditions:

1

\\

f The Control Room was evacuated IAW AOP-004, " Control Room Evacuation" -

i

!

Allqhree charging pumps were started 4 times in the last hour for testing l

The A xiliary Building operator has secured A and B Charging pumps

\

t

"C" Char \ging pump just tripped Which ONE (1) of the following correctly describes charging pump status with respect to

starting du:y? \

' \

i vA. No Chargingcanpump be started\ without exceeding the allowed starting duty limitations.

r

\

e a (~\ B. Only "A" Charging Pump can be restarted without exceeding the allowed starting duty J

4

' Bg[6 t 4 limitations. \

g \

C. Only "B" Charging Pump can'he restarted without exceeding the allowed starting duty i limitations. \ l 2'

\

D. Only "C" Charging Pump can be re\ started without exceeding the allowed starting duty limitations.

i a \

,- \,-

i

'p K/A 000068.EK3.18 4.2/4.5 l'" 9 3 M*

AOP-004, Step 13 Caution g 3c6 'Q T' )

w e - 9 42 M . ,' ;; , *

,,3 i New Question p 3 7 y, c

g e i,, 9 " a

,g in-g 4 a ""

y\ &sa c M

QWW Question 65 of 100

. . . - . . . . . . - - - . . . . - - - . . . . - . - . . _ . _ . - . _ . - - . - . . ... _ . _ ..~ . .

I g 95-2 NRC EXAM - REACTOR OPERATOR 0 ,

l 72. AOP-004-06 001 t

Given the following plant conditions:

l e The Control Room was evacuated IAW AOP-004, " Control Room Evacuation" ~ l l-

All three charging pumps were started 4 times in the last hour for testing ,

2 l

The Auxiliary Building operator has secured A and B Charging pumps l

l ' * "C" Charging pump just tripped

! 1 i

Which ONE (1) of the following correctly describes charging pump status with respect to '

starting duty?

l l VA. No Charging pump can be started without exceeding the allowed starting duty limitations.

B. Only "A" Charging Pump can be restarted without exceeding the allowed starting duty i

limitations.

l C. Only "B" Charging Pump can be restarted without exceeding the allowed starting duty

!. limitations.

'D. Only "C" Charging Pump can be restarted without exceeding the allowed starting duty ' )

limitations.

a i

K/A 000068.EK3.18 4.2/4.5 l

AOP-004, Step 13 Caution l New Question

, l t

l l i l

L

?

i Question 72 of 100 l

1

.- ,, ,_ ~ , , _ . . -,- -. ;-

Rsv. 6

, AOP-004 CONTROL .100M INACCESSIBILITY Page 12 of 30 STEP -

INSTRUCTIONS RESPONSE NOT OBTAINED CONTINUQUS USE ATTACHMENT 1 ale UJ N IDE (Page 6 of 9)

- - - - :. ; :. ; :. ;. ;. ;. ;. ;. :. ;. ;. :. ;. ;. ;. ;. ;. :. ;. ;. ;. ;. ;. ;. :. ;. ;. ***************+++**********

.QAUTION Starting duty limitations allow four Charging Pump starts per hour and require a minimum of five minutes between starts.

- - - - . ;. u :. ;. :. ; ;. ;. ; :. :. . ;. ;. :. :. ;. ;. ;. ;. . ; :. ;. ; :. ;. . ;. ;. :. :. ;. ;. ;. :. ;. ;. ;, :. ;. ;. ;. ;. :. ;. ;. ;. ;. ;. *+++++++++++

13. Control PZR Level From The '

Charging Pump Room As Follows:

a. Verify the following transfer switches - 7.N IDCAL

CHARGING PUMP A TRANSFER SW

CHARGING PUMP B TRANSFER SW l

l

CHARGING PUMP C TRANSFER SW

b. Verify only one Charging l Pump - RUNNING

c. Decrease running Charging Pump speed as follows:

1) Place the selector switch or. CHARGING PUMP SPEED CONTROLLER to MAN

2) Turn the Speed Control Knob counter-clockwise to decrease Charging Pump speed to minimum

d. }Qi]lE PZR level is greater than 706, 2}i]lE stop the

running Charging Pump i

k (CONTINUED NEXT PAGE)

i .. 1 I

Rzv. 6 i j AOP-004 CONTROL ROOM INACCESSIBILITY Page 12 of 30 p

STEP -

INSTRUCTIONS i RESPONSE NOT OBTAlliED e a =

CONTINUOUS USE Al'IACIDENT 1 -i l AUXILIARY BUILDING OPERATOR

. i 2

(Page 6 of 9)

. ;. ; ; :. ; ; ; :. ;. :. :. ;. ; :. ;******************* ******************r; :. ; ;. ;. ;. : ;. ; ; :. ;. ;. *****

CAUTION i

! Starting duty limitations allow four Charging Pump starts per hour and i

- require a minimum of five minutes between starts. ]

- - - - +****************************************************** !

I

13. Control PZR Level From The Charging Pump Room As Follows: 1 l

a. Verify the following transfer switches - IN LOCAL

CHARGING PUMP A TRANSFER SW

CHARGING PUMP B TRANSFER SW ,.

a CHARGING PUMP C TRANSFER SW

b. Verify only one Charging - ,

Pump - RUNNIIIG

c. Decrease running Charging Pump speed as follows:

1) Place the selector switch on CHARGING PUMP SPEED CONTROLLER to MAN

2) Turn the Speed Control

Knob counter-clockwise to ,

decrease Charging Pump -

speed to minimum

d. E PZR level is greater than 70%, M stop the

running Charging Pump P

d (CONTINUED NEXT PAGE)

- - . . ~ . . ._ .. - .. ~

l l l 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

- 1 l

i 66. AOP-018-03 003 '

l- Given the following plant conditions:

A failure of an RCP #1 seal has occurred

A loss of seal injection has occurred l j i

e FCV-626, Therm Bar Flow Cont Viv, valve closes

You are directed to have FCV-626 de-energized and opened locally I. Which ONE (1) of the following describes the contingency actions to be taken?

An Operator must......

L VA. stand by locally at FCV-626 to reclose it if needed due to a Phase B Isolation.

B. stand by MCC-5 to re-energize the valve if needed due to a Phase B Isolation.

l C. split the CCW headers to reduce the possible spread of contamination from a leaking !

thermal barrier. I D. cross-connect the CCW headers and close FCV-626 in the event of a Phase A Isolation. '

l a

)

K/A 000015.EA1.06 3.1/2.9 .

l AOP-014, AOP-018 New Question l

I l

4 l

Question 66 of 100 l

l __ _ _-

l 95-2 NRC EXAM - REACTOR OPERATOR i 1

' 73. AOP-018-03 003 )

l Given the following plant conditions:

i

A failure of an RCP #1 seal has occurred l 1

A loss of seal injection has occurred a FCV-626, Therm Bar Flow Cont Viv, valve closes l

l . You are directed to have FCV-626 de-energized and opened locally )

i Which ONE (1) of the following describes the contingency actions to be taken? j l

An Operator must......

i-VA. stand by locally at FCV-626 to reclose it if needed due to a Phase f5 Isolation. 1

( B. stand by MCC-5 to re-energize the valve if needed due to a Phase B isolation. j C. split the CCW headers to reduce the possible spread of contamination from a leaking thermal barrier.

D. cross-connect the CCW headers and close FCV-626 in the event of a Phase A Isolation..

a K/A 000015.EA1.06 3.1/2.9

AOP-014, AOP-018 New Question l

Question 73 of 100 1

l

, s v

Rev. 8 l . AOP-018 REACTOR COOLANT PUMP ABNORMAL CONDITIONS

;r g .Page 38 of 42

~ \+ E l'

CONTINUOUS USE.

l ATTACHMENT 2 I

RESTORATION OF RCP SEAL COOLING

-- (Page 3 of 7)

6. (CONTINUED)

b. Check FCV-626, THERM BAR FLOW b. Establish CCW to RCP Thermal CONTROL - OPEN Barrier as follows:

1) Open Breaker FCV-626, RCP THERMAL BARRIER OUTLET ISOLATION, at MCC-6 (CMPT-8F).

2) Establish communications between control Room personnel and an Operator stationed at FCV-626 in Pipe Alley.

(,/ 3) Direct operator at FCV-626 \

to close valve, if either g of the following events

( occur while FCV-626 is I de-energized:

\

Containment Isolation -\

l Phase B signal

\ \

OR Recieve indication of k

( any RCP Thermal

\ Barrier Cooler failure. (

4) In Pipe Alley, slowly throttle open FCV-626, THERM BARRIER OUT ISOL, to minimize the cooldown rate on the following:

l RCP Bearing i

AND

()

RCP #1 Seal leakoff (CONTINUED NEXT PAGE)

BASIS DOCUMENT, REACTOR COOLANT PUMP ABNORMAL CONDITIONS ATTACHMENT 2 (Continued) l 1

, ,y Step Description i

6 Flow is re-established to the RCP Thermal Barrier slowly (if required) in this step.

. Step 6.a checks position of CC-735, CC-71GA, and CC-716B. If any of these i valves are closed and cannot be opened from the RTGB, it is assumed that a l

failure has occurred that would inhibit proper operation of RCP Thermal Barrier cooling. The RNO provides transition around steps to re-establish i

CCW flow to the RCP Thermal Barriers if any of the valves listed is closed i 1

.and cannot be opened remotely.

Step 6.b checks FCV-626 open. If FCV-626 is closed, RNO 6.b provides instruction to slowly manually open the valve in an attempt to ccntrol ;

1 cooldown rate of the RCP Bearing and RCP #1 Seal Leakoff. '

RNO 6.b also provides direction to station an Operator to close FCV-626 l

Breaker in the event that a Phase B or Thermal Barrier leak occur. This is done because opening the breaker removes the automatic operation of the l

, ,. valve. In order to ensure containment Integrity exists and protection

(,,/ against a LOCA caused by RCP Thermal Barrier breach exists, the breaker must j

. be closed to perform automatic functions.

I 7 If re-establishing CCW flow to the RCP Thermal Barriers resulted in a cooldown rate in excess of l'F per minute, controlling RCP Seal Injection flow will not reduce cooldown rate. This step checks the cooldown rate. If j i cooldown rate exceeds 1*F per minute, the RNO provides instruction to allow i

the affected RCP to cool down below limits and then provides transition to i

steps re-establishing normal Seal Injection flow. l l

B If RCP Seal Water Flow Control valves have not been closed due to exceeding temperature limits, RCP Seal Injection may be re-established without controlling cooldown. This step checks status of RCP Seal Water Flow Control valves and bypasses steps to control cooldown if not required.

fO)

\m/

AOP-018-BD Rev. 8 Page 30 of 33

. .- . - ~ . - .. .~ _.. . .. - . - . - - -.- .- -.. - - . .

P Rev. 8 '

!' ' A0P- 01,8 REACTOR COOLANT PUMP ABNORMAL CONDITIONS Page 24 of 42 '

/s t (n, --

STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED SECTION B REACTOR COOIANT PUMP SEAL FAILURE (Page 12 of 13) ,

4

, 39. Check At Least One Of The Perform the following:

Following - INTACT:

Affected RCP #1 Seal i b. Refer to Technical og Specifications for any 1

applicable LCOs.

t

Affected RCP #2 Seal i c. Return to Procedure and Step l in effect.

!' 40. Check Affected Reactor Coolant Co To Attachment 2, Restoration Pump Parameters As Follows: Of RCP Seal Cooling.

RCP Bearing temperature -

LESS THAN 225*F lo

RCP #1 Seal Leakoff temperature - LESS THAN 235'F

41. Restore CCW To RCP Thermal

Barriers As Follows: J j a. Verify the following valves -

l OPEN:

1 4

CC-735, THERM BAR OUT ISO j

!

CC-716A, CCW TO RCP

CC-716B, CCW TO RCP l

b. Open FCV-626, THERM BAR FLOW CONT Valve

42. Check APP 001-B5, RCP HIGH VIB Go To Section C, High Reactor Alarm - EXTINGUISHED Coolant Pump Vibration.

43. Implement The EALs I

.- . - - _ . - . _ ~ ,_ .. -. - --- - -

BASIS DOCUMENT, REACTOR C001 ANT PUMP ABNORMAL CONDITIONS l Ssetien B (Continusd) l t Sten Description 1

l 34 This step closes the Seal Leakoff Valve for the affected RCP. l 35 Normal operation of the RCP Seals is checked in this step as a final check prior to returning to the normal operating procedures. If required, the RNO j

, provides instructions to restore RCP Seal Injection to normal operating I parameters. If HIC-121 is required to be throttled, the Operator aust i

monitor Charging Pump discharge pressure and maintain below 2500 psig to prevent lifting the Charging Pump Discharge Safety valves. (CR 95-1752) !

N36 This note reminds the Operator of conditions that may have caused FCV-626 to l close.

36 Status of FCV-626 is checked in this step. Position of this valve is j important because the Thermal Barrier provides RCP Seal cooling in the event of a loss of Seal Injection. Additionally, if a RCP #1 Seal has failed, b)

%.J there may be flow from the RCS up through the Thermal Barrier even if Seal Injection exists. If FCV-626 is open, the-RNO provides transition around steps to re-establish Thermal Barrier cooling water.

37 The Thermal Barrier Coolers are checked intact prior to opening FCV-626 to assure that a possible leak path is not reinitiated. Multiple indications should be used for this determination since R-17 alone can not be relied on for positive indication. Calculations have shown that as much as 2500 gallons of RCS would be needed prior to R-17 reaching the alarm setpoint.

The operator must be use judgment in this determination. The indications of a Thermal Barrier Cooling coil rupture are:

. Increasing CCW Surge Tank level

. High CCW flow from the Thermal Barriers

. Possible increase in R-17 indication

. Decreasing RCS inventory If a failure of a Thermal Barrier has occurred AOP-014 must be performed which will restore cooling to the intact RCFs 'ihile isolating the faulted.

O g . .

AOP-018-BD Rev. 9 Page 22 of 33

BASISDOCUMENT,REACTORCOO1ANTPUMPABNORMALCObDITIONS I

Section B (Continued)

F.tS.2 ppscriotion 4 38 If Containment Isolation Phase B signal exists, FCV-626 should remain closed j I

, to ensure' Phase B integrity. This stop checks Phase B status prior to opening FCV-626. The RNO provides transition around steps to re-estkblish CCW flow if Containment Isolation Phase.B exists. When the Phase B signal clears, the RNO provides direction to re-establish CCW flow to the RCP l Thermal Barriers.

39 RCP #1 or #2 Seal must be intact on all RCPs in order to effectively l 5

complete the remainder of this section. If neither RCP Seal is intact, this l procedure is exf.ted. Depending on the size of the leak rate either the RCS leakage AOP or the E0P network would be entered. Also, this procedure will not effectively restore RCP Seals or Seal Cooling to a Reactor Coolant Pung, with failed seals.

1 I

,, 40 This step checks to see if RCP temperature limits were exceeded. This is done to prevent restoring Thermal Barrier cooling in an uncontrolled manner to an affected RCP when FCV.626 is opened.

If required, RCP Seal Cooling is restored via transition to Attachment 2. I 41 This step restores CCW flow to the RCP Thermal Barriers, i

1 1 42 Problems with the RCP Seals could result in increased RCP vibration. This step checks RCP vibration alarm and provides tr$ncition to Section C if i vibration is in alarm.

i l 43 This is a standard step in the procedurs network to evaluate the need to implement the EALs. [NRC cominnent 92R0391]

1 44 This is a standard step in the procedure network to check Technical l Specifications for any applicable LCOs.

7 45 This is a standard etep in the procedure network to return to procedure and 8

( step in effect.

I d

AOP-018-BD Rev. 8 Page 23 of 33 l

l 4

?

+

j 95-2 NRC EXAM - SENIOR REACTOR OPERATOR ,

i i i :

j. .

!' 67. AOP-014-05 001 .

Given the following plant conditions: i e The unit is at 100% power l

A loss of CCW occurs i Which ONE (1) of the following describes how long AOP-014, Component Cooling Water

System Malfunction, allows operation of a charging pump without CCVJ ;mling?

i i-A. I minute a

! 4. 5 minutes C. 20 minutes l D. 30 minutes j.. < ,

. i

! K/A 000026.EA2.05 2.8/3.1 !

! AOP-014, Step 8 1 j New Question l 1-1 Jnt:nca f *EVEL 1

i I

.t 1

] Question 67 of 100

l !

l 95-2 NRC EXAM - REACTOR OPERATOR l.

i # l l 74. AOP-014-05 001 i .

i Given the following plant conditions: ;

1

'

The unit is at 100% power j

A loss of CCW occurs l l Which ONE (1) of the following describes how long AOP-014, Component Cooling Water

System Malfunction, allows operation of a charging pump without CCW cooling? !

l i l .

t l

i L A. 1 minute

.4. 5 minutes C. 20 minutes

! D. 30 minutes !

l i b l K/A 000026.EA2 06 2.8/3.1 AOP-014, Step 8 New Question i

r i

l 3

J 1

)

l l

i f l 1

i' 1

Question 74 of 100 p

-,no - - . . . . -

._._._.._..._.._____7-_.___._._._.._-.__

, Rev, 12 J

-I

, ADP-014 CCMPONENT' COOLING WATER SYSTEM' MALFUNCTION- f Page 35 of 103 5 STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED ;

SECTION C l

(~ CCW PUMP DISCLARGE PRESSURE IBM l

l (Page 4 of 7) l l 6, Verify Letdown Isolated As l 'Follows:

I

CVC-460A & B, LTDN LINE STOPS - CLOSED l

HIC-142, PURIFICATION 11DW -

! SET TO Ot.

CVC-387, EXCESS LTDN STOP -

l CLOSED i

7. Determine If Charging Pump (s)

Should Be Stopped As Follows:

i,

a. Check Charging Pumps - ANY a. Go To Step 10.

PUMP RUNNING

.A s b. Check'RCS temperature - b. Stop all Charging Pumps.

GREATER THAN 150"F Go To Step 10.

i

8. Establish Alternate Cooling To Charging Pumps As Follows:

a. Dispatch an operator to l perform Attachment 1

b. Stop all but one Charging Pump

c. Rotate operation of Charging Pumps at five minute intervals until Attachment 1 i

.ts complete as follows:

1) Start a non-running pump.

and note the time of start

2) Stop the previously running pump i

i

~

BASIS DOCUMENT, COMPONENT COOLING WATER SYSTEM MALFUNCTION Section C (Continued) .

rm S_ tan Description t

7 If the RCS is above 150'F a Charging Pump must be maintained for Seal Inj ection. This step checks to see if any are running or the RCS is greater than 150*F. If less than 150*F the punpr are stopped due to the lack of cooling water. If none are running the step is bypassed.

8 The pumps are alternated every 5 minutes !f one must be maintained running due to manufacturer recommendations. Emergency cooling is lined up via an attachment to provide cooling.

9 While starting and stopping Charging Pumps, the Operator thould observe Seal Injection flow by checking Thernal Barrier indicotions. During this period no CCW cooling to the Theinal Barrier exists, therefore the only seal cool;*.ng is provided by intermittent seal injection. (CR 95-1752, see f

Section A; step 34 for additional information) f~hi Chemistry is notified to stop sampling due to the lack of cooling to the 10

. G

' t heat exchangers, i 11 This steps determines if cooling must be supplied to the SFP Heat Exchanger, j The step checks temperature above the high temperature alarm setpoint, then I commences efforts to establish cooling. It is not expected that the SFP will reach elevated temperatures for an extended period of time. This procedure assumes that refueling efforts are NOT in progress. Should those efforts be underway, plant procedures are in place to more rapidly establish i emergency cooling in the event that normal cooling is lost.

12 This step acts as a hold point until a CCW Pump is started. At this point all actions to restart a pump have been directed and emergency cooling is e

established where possible.

V AOP 014-BD Rev. 12 Page 24 of 33

95-2 NRC EXAM - SENIOR REACTOR OPERATOR !

,- i l

l

i ,

68. AOP-012-09 001 Which ONE (1) of the following indications would lead the RTGB operator to diagnose and

, enter AOP-012, " Partial Loss of Condenser Vacuum or Circulating Water Pump Trip?

4 3

j VA. Decreasing generator electrical output j B. Decreasing turbine exhaust hood temperature

]

l C. Increasing steam pressure to Gland Sealing Steam 4

, g D. Increasing circulating water flow through the condenser huY j T I J 5 "

l

,W 7 i

K/A 000051 G0.11 2.7/2.9 i AOP-012' l APP-008-A5 '

! Modified Question B= exhaust hood temps would increase ;

C=would tend to increase vacuum i- -l D=would increase vacuum i 1

5([ en s6 Doe l Low t Eu EL i-4 4 4

4 i

I 4

r Question 68 of 100 i

.. 95-2 NRC EXAM - REACTOR OPERATOR ,

-75. AOP-012-09 001 i

j Which ONE (1) of the following indications would lead the RTGB operator to diagnose and 4

enter AOP-012, " Partial Loss of Condenser Vacuum or Circulating Water Pump Trip?

., VA. Decreasing generator electrical output

B. Decreasing turbine exhaust hood temperature '

1 i

! C. Increasing steam pressure to Gland Sealing Steam 1

D. Increasing circulating water flow through the condenser ;

1

, a 1

1 i

K/A 000051.GO.1I 2.7/2.9 .l AOP-012

APP-008-A5 Modified Question f

B= exhaust hood temps would increase C=would tend to increase vacuum

D=would increase vacuum I I i t

i i

2 1

l l

Question 75 of 100

,. - _ . . - - . . . . - . . . - . . . . _ - . . . . - . ~ . - . . . . . . . - . - . . - . . - - . . - - . . . . .

PARTIAL IDSS OF CONDENSER VACUUM OR CIRCUIATING Rsv. 8 AOP-012 i WATER PUMP TRIP Paga 3 of 25 1

U :

i

1.0 FURPOSE 4

This procedure provides instructions for a partial loss of Condenser vacuum or a Circulatir.g Water Pump trip.

i 2.0 ENTRY CONDITI.QE This procedure is~ entered whenever a partial loss of Condenser vacuum or a Circulating Water Pump trip occurs while the Turbine is on line.

i

(

I i

)

i s

4 4

i ,

8 l3 i

e 1

1 I .

I I

a-3 i

i i

l 1

l D

(

1

RASIO DOCUMENT, PARTIAL IDSS OF CONDENSER VACUUM OR CIRCUIATING WATER PUMP TRIP

, DISCUSSIDH:

O The dbjective of this procedure is to determine and correct the cause of a loss of Condenser vacuum, including a Circulating Water Pump trip. This procedure is j entered whenever a partial loss of Condenser vacuum, or a Circulating Water Pump trip occurs while the Turbine is on line.

The procsdure first checks for a Circulating Water Pump Trip and takes action for this event. A trip of a running CW Pump is the prevalent reason for~a loss of f

l Condenser Vacuum. If a pump has not tripped the procedure investigates other ;

possible reasons for a loss of vacuum. Included with these actions are a power l reduction in order to maintain back preseure,less than the manufacturer recommendations. An attachment is included to assist the local operator in j troubleshooting potential causes of the low vacuum. l l

i l l

INDIVIDUAL STEP DESCRIPTION:

O

j_t.gp t ppserietion

! i i

N1 This note informs the operator that steps 1, 2 and 3 are Immediate Actions.

These actions are immediate actions in order to preclude a trip. If the standby pump can be started soon enough vacuum will not deteriorate to the point requiring a trip.

I 1 This immediate Action step checks for a tripped Circulating Water Pump. If one of the Circulating Water Pumps tripped, flow may be insufficient to remove the heat from the Condenser, resulting in an increase in Condenser back pressure (decreasing vacuum). Also, if a running pump has tripped with the discharge valve open, CW flow will short-cycle the condenser from the running pumps.

l l

If no Circulating Water Pumps have tripped, the RNO bypasses the actions associated with a tripped pump, and sends the operator to the procedure step where loss of Condenser vacuum is addressed.

!O

! AOP-012-BD Rev 8 Page 3 of 13 i

L_-._. _ - -

,. - - - r

a APP-008-A5-

. ALARM

,. CONDENSER LO VACUUM

[N

]

AUTOMATIC ACTIONS

1. None Applicable 2

CAUSE

1. Loss of Vacuum Pump 4. Loss of Circulating Water

2. High level due to failure of 5. Excer.sive Steam Dump Hotwell Level Control Valve 6. Inleakage of Air into Condenser

3. Loss of Gland Seal Steam OBSERVATIONS Decreasing Genera r Ou 5. Low Gland Seal Steam Pressure

2. Increasing Exhaust Hood Temp (PI-4004)

3. Vacuum Pump Breaker Indicating 6. Low Cire. Water Pressure Lights 7. Increasing Condensate Water 1

4. Condenser Hotwell Level (LI-1417A) Temperature gs ,

( ACTIONS

1. Refer to AOP-012, Partial Loss of Condenser Vacuum or Circulating Water Pu'mp Trip.

DEVICE /SETPOINTS

1. 63-LVS (SW-2 ) /21 in. Hg. vac (9 in. Hg Abs)

POSSIBLE PLANT EFFECTS

1. Turbine Trip ;

2. Reactor Trip if above P-7 PEFERENCES l'. AOP-012, Partial Loss of Condenser Vacuum or Circulating Water Pump Trip

2. CWD B-190628, Sheet 855, cable B

(T .

(,,,) i APP-008 Rev. 13 Page 8 of 55

. . . - .. -._ _ -- _- .. -- - . . - - . . = -

o c.

95-2 NRC EXAM - SENIOR REACTOR OPERATOR 4

69. FRP-S.1-10 001 Given the following plant conditions:

4 i e The operators are using FRP-S.1, " Response to Nuclear Power Generation /ATWS", to respond to an ATWS

The Turbine is tripped 1

e MOV-350, BA to Charging Pmp Suct, has failed to open Which ONE (1) of the following describes the actions that the operator is required to perform?

A. Open FCV-ll3A, Blended MU to VCT, and FCV-Il4A, PW to Blender B. Open FCV-ll3B, Blended MU to CHG SUCT, and FCV-Il48, Blended MU to VCT

<. Open LCV-IISB, Emerg MU to Chg Suct, and Close LCV-Il5C, VCT Outlet D. Open LCV-115C, VCT Outlet, and Close LCV-115B, Emerg MU to Chg Suct c ,

4l,c,, el i c d l('

7"

jeen ,{ Ben s- Ek '

K/A 000029.K1.03 3.6/3.8

FRP-S.1 step 4 Modified Question fib W t 'l E FRF Ct No fgovato7 Question 69 of 100 L,.

~~

l

,' 95-2 NRC EXAM - REACTOR OPERATOR

86. FRP-S.1-10 001

~

Given the following plant conditions:

)

The operators are using FRP-S.1, " Response to Nuclear Power Generation /ATWS", to respond to an ATWS

The Turbine is tripped

MOV-350, BA to Charging Pmp Suct, has failed to open I

Which ONE (1) of the following describes the actions that the operator is required to perform?

4 A. Open FCV-ll3A, Blended MU to VCT, and FCV-114A, PW to Blender B. Open FCV-113B, Blended MU to CHG SUCT, and FCV-Il48, Blended MU to VCT i

<. Open LCV-Il5B, Emerg MU to Chg Suct, and Close LCV-Il5C, VCT Outlet

~

l D. Open LCV-Il5C, VCT Outlet, and Close LCV-Il5B, Emerg MU to Chg Suct c

} K/A 000029.K1.03 3.6/3.8 FRP-S.1 step 4

. Modified Question 1

i 4 Question 86 of 100

~ . . - . . - . - . . - . - . . . . _ . .-. - . . - . _ - . - . . - . _ . - - .

Rsv. 9 FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION /ATWS Page 6 of 17 4

STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED

4. Initiate Emergency Boration Of l RCS As Follows:

f a. Verify Two Charging Pumps - l

RUNNING AT FULL SPEED l

^

l b. Verify Boric Acid Pump b. Perform the following:

aligned for blend - RUNNING

1) Open one of the following valves: i i I

LCV-115B, EMERG MU TO I q CHG SUCT l

CVC-358, RWST TO l CHARGING PUMP SUCTION

(locally) l 4

l 2) Close LCV-115C, VCT OUTLET.

3) Go Ta Step 4.e.

c. Verify MOV-350, BA To c. Perform the following:

CHARGING PMP SUCT - OPEN

1) Open one of the following valves:

LCV-115B, FRERG MU TO CHG SUCT E

CVC-358, RWST TO CHARGING' PUMP SUCTION (locally)

2) Close LCV-1150, VCT OUTLET.

3) Go To Step 4.e.

(CONTINUED NEXT PAGE) i

- - - .. - -- . .. ~. . - - - ..

95-2 NRC EXAM - SENIOR REACTOR OPERATOR f

4

70. CVCS-10 001 4

Given the following plant conditions:

R 4 and R-9 alarms are received 1'

APP-036 and AOP-005, Radiation Monitoring System, are entered 4

j. Which ONE (1) of the following describes the appropriate operator response to the above conditions?

1 A. Start ilVE-5A and 58, Verify 120 gpm letdown in service. p I

,' vB. Start HVE-5 A or 5B, Verify one letdown orifice in service.

C. Start HVE-19A or 19B, Verify 120 gpm letdown in service.

4 D. Start HVE-19A or 19B, Verify one letdown orifice in service.

b K/A 000078.EK3.06 3.I/3.8 n c s do a ct ffMb' AOP-005 1 i New Question b'Q ce l-( W 0 R e AOf fitf ;

Question 70 of 100

95-2 NRC EXAM - REACTOR OPERATOR .

.. i y

77. CVCS-10 001 Given the following plant conditions:

R-4 and R-9 alarms are received t

APP-036 and AOP-005, Radiation Monitoring System, are entered i

Which ONE (1) of the following describes the appropriate operator response to the above ,

conditions?

A. Start HVE-5A and 58, Verify 120 gpm letdown in service. f

4. Start HVE-5A or 5B, Verify one letdown orifice in service.

C. Start HVE-19A or 19B, Verify 120 gpm letdown in :..rvice. l D. Start HVE-19A or 198, Verify one letdown orifice in service. ;

b .

l l K/A 000078.EK3.06 3.2/3.8 ;

AOP-005 ;

New Question :

i t

I I ?

l i l

Question 77 of 100 i i i __ ,,i

_ - . . .. . ._ _ . . . ~ . - . _ - . _ . _- .- . - - . - _ . - ._ . . _ . _

4 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

.c * .

71. AOP-018-05 001 Given the following plant conditions:

An RCS heatup is in progress.

i e RCS pressure is 375 psig.

RCP "B" was started four hours ago.

O e The RO has drained the RCP "B" standpipe six times since the pump was started. :

! i

e RCP "B" frame vibration is two mils.

4

. e #1 Seal leakoff flow is 0.26 gpm.

I e #2 seal leak off flow 0.4 gpm ;

i e #1 seal D/P is 275 psid. !

Which ONE (1) of the following describes the cause of the above indications? I i

A. #1 seal is not fully seated.

4. #2 seal is not fully seated.

C. CVC-307, PRI SEAL BYP ISOL, is not closed.

D. VCT high pressure.

b K/A 000015.A1.22 (4.0/4.2) d f;- r o- G ' ' / -> O APP-001-C5.

AOP-018, Step 16, Pg.16.

Modified Question fj b f j u r. . p 0 7HE t? ) f b 15 l

l Question 71 of 100

)

95-2 NRC EXAM - REACTOR OPERATOR

78. AOP-018-05 001 Given the following plant conditions:

lt f

. An RCS heatup is in progress. ;

e RCS pressure is 375 psig.

e e RCP "B" was started four hours ago.

The RO has drained the RCP "B" standpipe six times since the pump was started.

RCP "B" frame vibration is two mils.

i e #1 Seal leakoff flow is 0.26 gpm. !

1 e #2 seal leak off flow 0.4 gpm 4

e #1 seal D/P is 275 psid. ;

Which ONE (1) of the following describes the cause of the above indications? !

A. #1 seal is not fully seated.

4. #2 seal is not fully seated.

C. CVC-307, PRI SEAL BYP ISOL, is not closed. i D. VCT high pressure. !

L b i i

K/A 000015.A1.22 (4.0/4.2)

APP-001-C5. ;

AOP-018, Step 16, Pg.16.

Modified Question s

Question 78 of 100 u -_ _

APP-001-C5 AMBH RCP STANDPIPE HI/14 LVL *** WILL REFIASH ***

l f

( AUTOMATIC ACTIONS '

i f 1. None Applicable !

l l CAUSE '

i l 1. Failure of Reactor Coolant Pump Seals

2. No. 2 Seal Hanging Open

3. Excessive Leakage into RCDT

l l OBSERVATIONS

1. 2 x 2 Status Light Panel to determine affected Pump I

2. Seal Leak Off on affected Pump i 3. Affected Pump Vibration and Bearing Temp !

4. Thermal Barrier AP on affected Pump

5. RCP No. 1 Seal AP !

6. RCDT level and pressure ACTIONS

1. IE level is Low, IEEE fill the Standpipe. i

2. IE level is High, IEEE check for rising level in Reactor Coolant Drain Tank.

3. If seal failure is indicated, IEEE refer to AOP-018.

)

DEVICE /SETPOINTS

1. LC-406A, LC-407A, LC-408A/1 ft. above normal l

2. LC-406B, LC 407B, LC-408B/1 ft. below normal REFERENCES

1. TECH SPECS 3.1.1.1

2. AOP-018, Reactor Coolant Pump Abnormal Conditions

3. Flow Diagram, 5379-1971, Sh 2

4. CWD B-190628, Sh 104, Cable J l

i l I l

?

APP-001 Rev. 14 Page 25 of 53 i t

~

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

72. AOP-014-09 001 Given the following plant conditions:

The unit is at 75% power e CCW Surge Tank level is increasing e No makeup to the CCW surge tank is in progress e R-17, CCW Surge Tank Radiation Monitor, indication is Stable l

Which ONE (1) of the following describes the leak that would provide the above indications?

VA. S/G Blowdown sample heat exchanger B. Excess L/D heat exchanger C. Air Ejector Condenser D. RHR heat exchanger a

K/A 000026.A2.01 2.9/3.5 Answers A and C are the only non-primary systems C= pressure is significantly lower than CCW pressure CCW Flow diagram 5379-376 Modified Question f4 l il & T, c e. n<a m> ovo r i

l Question 72 of 100

p 95-2 NRC EXAM - SENIOR REACTOR OPERATOR ,.

73. FP-001-05 001 Given the following plant conditions:

i e A plant fire has occurred inside the RCA y)0 rf) . The Fire Brigade Team Leader has determined that the source is within a locked High Radiation Area (LHRA)

('

k._ e The Radiation Control Fire Support (RCFS) qualified person is not available i Which ONE (1) of the following additialisl requirements must be met before the Fire Brigade Team Leader can direct the Fire Team to enter the LHRA and combat the fire?

A. Security Supervisor at the access point I

B. Radiation Control Supervisor approval W. A licensed operator must accompany the team with a survey meter i l D. A non-RCFS qualified technician must accompany the team with a survey meter i 3 MAf c

e,os. I

i 1

, K/A 000067.G0.01 3.6/4.0

FP-001

. Modified Question l

1 l

1 Question 73 of 100

95-2 NRC EXAM - REACTOR OPERATOR

.y

I l; 80. FP-001-05 001 i Given the following plant conditions: ;

A plant fire has occurred inside the RCA

The Fire Brigade Team Leader has determined that the source is within a Locked liigh !

Radiation Area (LHRA) i l

The Radiation Control Fire Support (RCFS) qualified person is not available

! Which ONE (1) of the following additional requirements must be met before the Fire Brigade i i-Team Leader can direct the Fire Team to enter the LHRA and combat the fire?

1 2 -

i A. Security Supervisor at the access point l

B. Radiation Control Supervisor approval I W. A licensed operator must accompany the team with a survey meter D. A non-RCFS qualified technician must accompany the team with a survey meter c

K/A 000067.GO.01 3.6/4.0 FP-001 Modified Question i

i i

i l

) lI Question 80 of 100

- .. ~ . . - _ - - _ . = . - _ . - - - . . - . - . - _ - . .. - .. . =. - -

6.0 (Continued) 6.10 Radiation Control (RC) Personnel E9.IE i RC personnel who are not RC Fire Supporr (RCFS) qualified can perform RC l functions as requested by the Team Leader or Shift Supervisor, but they shall not i l

enter a room or fire area where there is smoke or fire until the ALL CLEAR is l determined.

1. Support the fire brigade in a Radiation Control Area as follows:

1). A RCFS qualified person will respond to the fire area in full turnout gear and SCBA with a radiation key and survey meter.

2) If needed access locked radiation control doors and accompany the i \

fire brigade into the area and perform RC functions. l 4 i

)

3) If a RCFS qualified person is not available within a reasonable j time, the Fire Brigade Team Leader can direct the fire brigade to 1 enter a locked high radiation area but the following shall be performed:

A licensed operator with a survey meter must enter with fire

brigade personnel.

Advise RC personnel of this entry.

4) RC personnel who are not RCFS qualified should respond as follows:

Report to the fire brigade team leader at the command post.

Advise and support the fire brigade and team leader on RC matters as appropricte.

i FP-001 Rev. 24 Page 18 of 28

n 95-2 NRC EXAM - SENIOR REACTOR OPERATOR ,

l 74. EPP-001-14 004 I Given the following plant conditions:

l

A complete loss of AC power has occurred '

i I

e Batteries are at maximum load i

Which one (1) of the following describes the DC system response to the above conditions?

Loss of DC will occur ..... ;

l l l j A. if battery chargers are not restarted within 30 minutes.

4. if battery chargers are not restarted within 60 minutes.

C. if inverters are not restarted within 30 minutes.

D. . if inverters are not restarted within 60 minutes. j b l, i l K/A 000055.EK3.01 2.7/3.4 EPP-001, pg 26 New Question i

g ggg LEJJ i M'# O !O /

g f if p t rJ co RRii 7 l l

l i

l Questicn 74 of 100 I

i

Rav. 20 j,

EPP-1 LOSS OF ALL AC POWER Page 26 of 51 STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED CAUTION

If the loads placed on the energized Emergency Buses E-1 or E-2 exceed the capacity of the power source, the supply breaker will trip.

-

A loss of DC power may occur if the DC busses are at maximum load and j the battery chargers are not restarted within 60 minutes of a loss of

all AC power.

i

43. Perform Ihe Following:

a. Check Smergency Bus E-1 a. Go To Step 43.b.

status - ENERGIZED i

! 1) Check MCC-5 status - 1) Locally verify CLOSED the

ENERGIZED FROM THE DS BUS FEED TO MCC-5 (NORM POWER) i

& MCC-16 (CMPT-21A) at s

Emergency Bus E-1.

2) Locally verify.the

' following breakers at MCC-5 . CLOSED:

BATTERY CHARGER A-1 (CMPT-58) 1 1

INSTRUMENT AIR

COMPRESSOR A (CMPT-7M)

'

FEED TO LP-29 (ALT l POWER) (CMPT-10CR) ,

l

BATTERY CHARGER A (CMPT-11BR)

FEED TO MCC-10 (CMPT-17FR)-

4 q

l 1

j (CONTINUED NEXT PAGE) 4 4

95-2 NRC EXAM - SENIOR REACTOR OPERATOR i

4 l

, 75. AOP-023-06 001 Given the following plant conditions

l-o i

. The unit is in hot shutdown prior to refuelmg ;

i i e Refueling preparations are in progress 3 Which ONE (1) of the following describes the condition which would cause a loss of l' Containment Integrity? l l

g,4 vA. The personnel air lock doors are closed but not sealed. ,

i B. An automatic contamment isolation valve has failed closed and has been deactivated.

1 3

i C. A containment isolation valve has failed its stroke test and been isolated by a manual l valve.

D. A manual valve has been found closed but does not have a locking device on it, as the i procedure requires.

5 a

i 1

1 K/A 000069.EA%01 3.7/4.3 I' AOP-023, Step 18 i' New Question g g,y 10 #.

$E E c, ig pr tosn ooor CLo** ' E*

Question 75 of 100

, Rev. 7 I AOP-023 LOSS OF CONTAINMENT INTEGRITY i

Page 9 of 10 i

=

t l

STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED l , 18. Check CV Integrity Established Notify maintenance to expedite As Follows:

repair of component, to

. re-establish containment '

j

The equipment hatch - integrity.

j PROPERLY CLOSED M SEALED e

E

At least one of the

, personnel airlock doors -

PROPERLY CLOSED AND SEALED I i AN_D

)

All non-automatic CV

isolation valves, not

! required for normal i i

operation - CLOSED j l

] AND !

All CV isolation blind I flanges - PROPERLY INSTALLED j WHERE REQUIRED 4 AN_D

,

All automatic CV isolation '

i trip valves, required to be closed during accident j conditions are as follows: - l 1

OPERABLE f

l

9E i

1

INOPERABLI, BUT ACTIONS i DESCRIBED IN TECHNICAL I

, SPECIFICATION 3.6.3 ARE i ACCOMPLISHED WITHIN THE

. SPECIFIED TIME REQUIREMENTS l

I

All manual valves qualifying I

as automatic CV isolation valves - CLOSED

(

l . 95-2 NRC EXAM - SENIOR REACTOR OPERATOR l .

i 6

76. AOP-023-09 001 ,

Given the following plant conditions:

l A loss of Containment Integrity has occurred ,

Shutdown Margin is greater than 1.77% Delta K/K l

Which ONE (1) of the following describes the evolution that is NOT allowed to continue?

VA. Refueling -y o Jf er V e//ow f i'

B. Rod Drop Timimg Test C. Control Rod Exercise Test i

D. Fully withdrawing Shutdown Bank B l

a !

CIO }.i /3 5 4 ""/,/ms, m en nr of reh . .s7, K/A 000069. A2jd,,2 3.9/4.4 - V c i 3 '. d .'* '

AOP-023, pg 4 New Question 1

l i

t ,

I Question 76 of 100

- . - . . . . . - - . ~ . - - .--

Rav. 7

, AOP-023 IDSS OF CONTAINMENT INTEGRITY Page 4 of 10 STEP -

INSTRUCTIONS . RESPONSE NOT DETAINED

1. Implement Technical Specification 3.0 LCO

2. Check Shutdown Margin - GREATER Stop all positive reactivity THAN @ EQUAL TO 1.77% AK/K changes made by the following, until CV integrity E shutdown margin is re-established:

l

Boron dilution

- Rod drive motion Go To Step 4. >

l 3. Perform The Following, Until CV Integrity Is'Re-established:

i a. Stop all positive reactivity changes made by rod drive l motion, except during any of l the following evolutions:

Rod drop timing test

Rod drive mechanism I

timing test

Control Rod exercise test

Withdrawing S/D BANK A

Withdrawing S/D BANK S AEQ all Control Rods to less than or equal to 5 steps '

4. Check Refueling - IN PROGRESS Go To Step 6.

i.O

Rsv., 7

, AOP-023 IDSS OF-CONTAINMENT INTEGRITY .

. Pegs 5 of 10 )

l $ i j /'7% !

l - ( ,) --

STEP --

INSTRUCIIONS RESPONSE NOT OBTAINED 1, j o 3 l .

l

5. Notify Refueling Personnel To Perform The Following:

a. Pl' ace any_ fuel assembly in

. transit.in one of the l following locacions

l L

Original Core location

)

i !

l

Upender j l

Storage location approved l l by FMP-019 Fuel and Insert Shuffle

b. Place any core upper or lower internals in transit in one of the following locations: ;

Core (preferred)

Storage location in transfer canal I

l c. Stop all refuelin5 Operations l i

6. Check Service Water Leak Inside Co To Step 15.

Containment IDENIIFIED i

f l

l l

l

. l I

i

~

I i j i 1 l

.- . .. . . - - . - - -- t

i 95-2 NRC EXAM - SENIOR REACTOR OPERATOR 0

i

77. FRP-H.1-14 002 Given the following plant conditions:

The plant has experienced a nonisolable main steam line break.

The operators are implementing actions of EPP-16, " Uncontrolled Depressurization of All S/Gs".

All S/G Narrow Range levels are offscale LOW S/G pressures are 450 psig and decreasing slowly Feed flow has been reduced to 80 gpm to each S/G by the operators per EPP-16 guidance.

Which ONE (1) of the following describes when the operators would be required to TRANSITION to, and IMPLEMENT actions contained in FRP-H.1, " Loss of Heat Sink" to restore S/G levels?

A. If 10% Narrow Range level cannot be restored in at least one S/G.

B. If 5 % Narrow Range level cannot be restored in all S/Gs.

C. If "A" and "B" AFW pumps are not available.

vD. If a total feed flow capability of 300 gpm is not available.

i d ),c, ,2 , cen n te ca,.,s 'AW FRP-HI I

K/A 000040.GO.I1 (4.1/4.3) s+eO -

FRP-H.1, Loss of Heat Sink, Step 1 Modified Question f

r i

i j

Question 77 of 100 t

J

--- . _ - _ - - . _ . ~ . .- -. . . . . - ~ . . -.

Rsv. 9 FRP-H.1 RESPONSE TO IDSS OF SECONDARY HEAT SINK Page 4 of 40 -

6 STEP -

INSTRUCTIONS RESPONSE NOT OBTAINED

- - - - w*************

CAUTION Feed flow is not re-established to any faulted S/G if an intact S/G is i available, i

~

1. Check Total Feed Flow - LESS Go To Step 3.

THAN 300 GPM DUE TO OPERATOR ACTION I

2. Reset SPDS And Return To

- Procedure And Step In Effect i i

3. Determine If Secondary Heat Sink i I

Is Required As Follows:

f

a. Check RCS pressure - GREATER a. Reset SPDS and Go To PATH-1, ,

THAN ANY NON-FAULTED S/G Entry Point C. '

PRESSURE

),

b. Check RCS temperature - b. Perform the following: i GREATER THAN 350*F [310*F] l

1) Place RHR System in service using Supplement I.

l

2) M adequate cooling with RHR is established, THEN reset SPDS and return to procedure and step in effect.

4. Check Any Two S/G Wide Range IE any two S/G Wide Range levels Levels - LESS THAN 26% [37%) decrease to less than 26% [374),

M perform Steps 4.a and 4.b.

a. Stop all RCPs

b. Observe CAUTION prior to 1 Step 18 and Go To Step 18 l l

l

( 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

78. EPP-009-05 001 Given the following plant conditions:

A large break LOCA has occurred.

The Control Room Shift Supervisor (CRSS) has just entered EPP-9, " Transfer To Cold Leg Recirculation."

The STA informs the CRSS of an ORANGE path on FRP-C.1, " Response To Inadequate Core Cooling."

Which ONE (1) of the following describes when the CRSS should transition from EPP-9 to FRP-C. l?

N "'?

, ; ,h r e

. cwmu a* t'e n '% " I " -

en f r f7p,,

6# 9 A. Immediately upon notification of the Orange Path on FRP-C.l.

B. Immediately after verification of greater than 1200 gpm RilR recirculation flow."

C. When FRP-C.1 RED path entry conditions are met.

vD. When directed to transition to " Procedure and Step In Effect" in EPP-9.

d t_,s;s., ,i ~oi; /~

K/A 000011.K3.15 (4.3/4.4) - c. , :t .'5 i~ JAN1A3 7" ' ' "' r EPP-009 Modified Question i G7 T. ! c.

Question 78 or 100 b

d

, Rav. 17 l< EPP-9 TRANSFER TO COLD LEG RECIRCULATION 1

] Page 4 of 28 j i >

i j

STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED 4

i: 1. Perform Steps 1 Through 27 Without Delay I

2. Foldouts Are Not Applicable l 4 During The Performance Of This i Procedure j l

3. Do EgI Implement Functional 1 Restoration Procedures Prior To i Completion Of This Procedure l

4 lE At Any Time It Is Determined

,' That At Least One Flow Path From The CV Sump To The RCS Can EQI i I

Be Established Or At Least j 354 INCHES Will NOT Be Available l In The CV Sump, THEN Co To j EPP-15,' Loss Of Emergency i Coolant Recirculation ,

i I

5. Verify S1 PUMP RECIRC Valves -

. CLOSED l

SI-856A

SI-856B i .

'I l

l 1

l

95-2 NRC EXAM - SENIOR REACTOR OPERATOR i

79. AOP-024-14 002 Given the following plant conditions:

The unit is at 100% power

A loss of MCC-6 has just occurred

~

Which ONE (1) of the following describes the response of the feedwater control valves?

q<c

l. VA. Automatic and Manual controly lost.

B. Only Automatic control is lost.

C. Orly Manual control is lost.

D. Only RTGB valve position indication is lost.

a

! .jg y 4 4sk t eso c o n i. , I K/A 000057.A1.06 3.5/3.5) .s p g l

l AOP-024, ATT 4, Pg.1

, q ~y b# jkI t

Modified Question 1

i i

i l

l Question 79 of 100 fIS 1

. = . - . . ~ . . _ - .-.. _ _ . . _ . .- .

R'2v. 6 a AOP-024 LOSS OF INSTRUMENT BUS Prge 53 of 93 2

O V CONTINUOUS USE ATTACHMENT 5 EXTENDED LOSS OF INSTRUMENT BUS 4 (AND 9)

(Page 1 of 4) 4 EQIE

{ The following control functions will be affected until Instrument Busses 4 and 9 are restored:

l FRV A, B, & C Auto & Manual Control (Locked up)

W Isolation Circuit (Reactor Trip with Low Tavg)

OT & OP AT Circuitry (Five percent interval runbacks energized)

All automatic PZR Pressure Controllers Automatic PZR Level Controllers Charging Pump A, B, & C Automatic Controls LC-ll2, Automatic VCT Level Control RCS Letdown /PZR Heater Controls i Steam Dump Pressure Control mode CCW Pump Auto-Start Circuit (starts all three pun,ps)

! R-ll/12 Vacuum Pump (RMS-1, 2, 3, & 4 close)

[ PT-447. Turbine First Stsge Pressure

I i

l. Control PZR Spray Valves in manual.

, 2. Verify Turbino First Stage Pressure selected to PT-446 position. I i

i a

W 4

)

f

. f

%) 1

___ mange. -' -

$ 95-2 NRC EXAM - SENIO.R REACTOR OPERATOR I

80. AOP-024-10 002 Given the following plant conditions:

The plant is water solid on RHR l

All systems are in automatic i

j.

Instrument Bus 3 is lo.st Which ONE (1) of the following describes the actions that will occur? ;

i j A. A power range rate trip signal will occur.

4. "C" Charging pump controller will lock-up, if running.

4 I C. PZR PORV 455B will automatically open.

4 l

4. l D. FCV-626, Therm Bar Flow Cont Viv, valve closes.

b 1

4.

i K/A 000057.EA2.19 4.0/4.3 .3 4 - f 4., L/ t/

AOP-024, Attachment 4 note i New Question i

I gA, t . t ot :

< , ,, y 4 ,, , p4 1

j i

l.

Question 80 of 100

_ . _ _ . ~ _ ... _ _ ... _ ... _ _ ._. . _ _ _ _ .. _ _ ...-. . .._.. ._ _ _ . _ __._.. _ _ _. _ .

. . i Rsv. 6

-AOP-024 IDSS OF INSTRUMENT BUS

Page 50 of 93 CONTINUOUS USE A

ATTACHMENT 4 l EXTENDED IDSS OF INSTRUMENT BUS 3 (AND 8) ,

i- )'

i (Page.l'of.3)

S The following control functions / indications will be lost until Instrument Bus 3 and 8 are restored:

PT-446 Turbine 1st Stage Pressure FCV-114A, PW to Blender (locked up full open)

FRV A, B, & C Automatic Control PCV-455B PORV (Indication only)

Safeguards Train B Sequencer i FCV-1425 (AFW PUMP B inoperable)

Charging Pump C Controller, SC-153A (locks up)

ICCM - Channel II i Steam Dump Steam Pressure Control l RMS Racks 2 & 3 and R-32B

~

PT-137 Excess Letdown Pressure Indication O S/G A PORV Control TCV-1447 and TCV-1448 Exhaust Hood Spray Valves j

Solenoids for R-ll/12 Skid (fail closed) i l

1. Place Turbine First Stage Pressure Selector Switch to PT-447 ;

position.

F.Q.II In the event that the Plant experiences a trip due to-difficulty in !

maintaining all S/Cs in manual level control, feed flow to the S/Cs will be accomplished via the AW Pumps 9E FRV Bypass Valves.

2. Continue to operate FRVs A, B, & C in MAN.

3. Contact Operations Staff for. availability of a dedicated FRV watch.

_ - -. . - -- .. . - . - . . . . - . . . .- - ~ . . . -

95-2 NRC EXAM - SENIOR RI' ACTOR OPERATOR

!' 81. OMM-022-14 002 Given the following plant conditions:

/-% y e E p oj/r>

3 tr a n1 A loss of all AC power ha(occurred

) During recovery, the STA reports the status of the CSFST's as follows:

Subcriticality -

GREEN i

Core Cooling - -RED- N d 4 c'

Heat Sink -

RED i

Integrity -

GREEN d

Containment -

ORANGE

+

Inventory -

YELLOW Which one (1) of the following describes the procedure that should be in effect to respond to d

this event? ,

1 s

i A. FRP-C.1, " Response to inadequate Core Cooling" B. FRP-J.1, " Response to High Containment Pressure" 1

C. FRP-H.1, " Response to Loss of Secondary Heat Sink"

@. EPP-1, " Loss of All .AC Power" d

K/A 000055.G0.I1 4.1/4.1 OMM-022, rv8, pg 20,4th paragraph Modified Question i

Question 81 of 100

. . - . . _ . _ _ _ . ~ . _ _ _ . _ . _ . _ _ _ . _ _ _ _ . _ _ . _ . _ _ . _ . _ . _ . _ . . .- _

T

5.2.2 (Continued)

If any ORANGE terminus is encountered, the operator is expected to i monitor all of the remaining trees and if no RED-condition is l encountered, suspend any PATH or EPP in progress and perform the FRP i required by the ORANGE terminus. l If during the performance of an ORANGE-condition FRP, any RED-condition or higher priority ORANGE-condition arises, then the RED or higher j priority ORANGE-condition is to be addressed first, and the original i ORANGE-condition FRP suspended, i

h

Once an FRP is entered due to a RED or ORANGE-condition, that FRP is.

! performed to comoletion, unless preempted by some higher priority 4

l condition. It is expected that the actions in the FRP will clear the RED i <

4 or ORANGE-condition before all the Operator actions are complete.

1 i However, the FRPs should be performed to the point of the defined 3

transition to a specific EOP or to the " procedure and step in effect."

Due to the special situations imposed by a loss of all AC power, an FRP l in progress should be immediately suspended if entry into EPP-1, Loss of 1

All AC Power, is required. Because none of the electrically powered' safeguards equipment used to restore Critical Safety Functions will be operable, further performance of FRP steps would be pointless. Also, during the performance of EPP-009, Transfer To Cold Leg Recirculation, entry to the FRPs will not be performed. This is due to the time critical nature of Cold Leg Switchover.

Status Tree monitoring should be continuous if any ORANGE or RED-condition is found to exist. If no condition more serious than YELLOW is encountered, monitoring frequency may be reduced to 10-20 minutes, unless a significant change in plant status occurs.

A YELLOW terminus does not require immediate Operator attention.

Frequently, it is indicative of an off-normal and/or temporary condition which will be restored to normal status by actions already in progress.

In other cases, the YELLOW-condition might provide an early indication of

'a developing RED or ORANGE-condition.

QMM- .022- Rev. 8 Page 20 of 43

, , , 95-2 NRC EXAM - SENIOR REACTOR OPERATOR 4

i

82. PATH-1-09 001 Given the following plant conditions:

~

.

A small break LOCA has occurred

SI pumps fail to start l

RCS Hot Legs and the Reactor Vessel Head have voided j i

e RCS Pressure is 775 psig

RCP's are tripped in accordance with the EOP network l l

Assume that all other ECCS equipment operates as required 3

l Which ONE (1) of the following describes the current method of cooling the wre?

4 A. Break flow is the only core cooling method available.

4. Break flow and reflux flow are providing core cooling. I C. Natural Circulation is the primary core cooling mechanism. w -/ r A'07 e

D. - No core cooling mechanism exists at the present time.

b 000009.EK1.01 4.2/4.7 WOG Background on core cooling mechanisms MCD-6, pg 10,11 New Question f( EF P O Cw't - A ff L T I

i i

Question 82 of 100

.' 95-2 NRC EXAM - REACTOR OPERATOR

81. PATH-1-09 001 Given the following plant conditions:

A small break LOCA has occurred ;

SI pumps fail to start

RCS Hot Legs and the Reactor Vessel Head have voided e RCS Pressure is 775 psig e RCP's are tripped in accordance with the EOP network e Assume that all other ECCS equipment operates as required Which ONE (1) of the following describes the current method of cooling the core? !

1 1

I A. Break How is the only core cooling method available. I vB. Break flow and reflux flow are providing core cooling.

C. Natural Circulation is the primary core cooling mechanism. !

D. No core cooling mechanism exists at the present time.

b l 000009.EK 1.01 4.2/4.7 WOG Background on core cooling mechanisms MCD-6, pg 10,11

. New Question Question 81 of 100

, LESSON BODY KEY AIDS 4

a. Depends upon heat removal relation to heat pmduction l

{ b. During LOCA - inventory loss dmins the pressurizer i

c. Without pressurizer, RCS pressure decreases

~

d.' Saturation occurs, creating steam in vessel j e. Vessel acts as pressuri7er

_ f. Vessel stiam temperature contmls RCS pressure 3

i 1) Heat in (heaters) - decay heat l 2) Heat out -- break removes heat I

i 3) Decay heat > break removal -- RCS pressure j increases

4) Break removal > decay heat - RCS pressure decreases

g. Lsrge LOCA

1) High heat out

2) Rapid RCS pressure decrease l

3) Break removal exceeds decay heat

h. Very small breaks

1) Normal charging controls inventory

2) No pressurizer drain

3) Normal pressure control

i. Small break (1 to 6 inches)

1) Exceeds normal charging capability l

2) Pressurizer empties

3) RCS pressure control to the vessel steam MCD6 Rev.O Page 10 of 44 i

LESSON BODY KEY AIDS

4) Break size - determines rate of prerare decreases until saturation

5) Pressure stable at saturation

6) Break size - length of stabilization a) Larger size - short length b) Small size -- long length h j. Small-si7e r, mall breaks

1) Pressure tends to increase - Jg.ay heat > break rr,moval

, 1) Saturated RCS temperature and pressure increase

3) SG temperature and pressure increase

4) Safety valves open l

5) Limit SG and RCS pressure and temperature

6) Stable pressure -- excess decay heat out atmosphere

5. Summary

, a. Small-break T.OCAs L

b. RCS stabilizes pressure

1) Relationship between decay heat production and f

2) Break flow heat removal

c. RCS pressure limits ECCS flow (

1) ECCS flow < break flow l

2) RCS inventory decrease i i.

d. Inventory decrease and vessel liquid flash MCD6 Rev.O I Page 11 of 44 l 5

j 95-2 NRC EXAM - SENIOR REACTOR OPERATOR i

83. EPP-026/27-14 001 .

1 Given die following plant conditions: ;

l e The unit is in Hot Shutdown l l

l e SUT is supplying all 4KV buses l e A severe short on the "B" DC Bus has caused the loss of that bus Which ONE (1) of the following describes the response of the emergency diesel generators (EDG's)?

wuc de7 A. "A" EDG starts, "B" EDG air start solenoids fait closed preventing start.

4. "A" EDG remains available, "B" EDG starts but the exciterSeldwill not flash 2nd the output breaker will not close.

C. Both EDG's will auto start, "B" EDG will not load due toSUT already powering-E-3,-

"A" EDG output breaker will close onto the bus.

D. Both EDG's will auto start, "A" EDG will not load due4o-SUT already. powering E Ar-

"B" EDG output breaker will close onto bus. ;

i b

j!r' 7^ E Tw ( E ij:w K/A 000058.K3.01 3.4/3.7 EPP-26/27, Attachment 1 b 0 N'r /dr TivC /A',

l New Question

]

I 4

)

Question 83 of 100

.=- . .. - . - . . . . - . _ . - . . . . . - - ~ . -.-. ..-.--- - .-..- .

i a '

, 95-2 NRC EXAM - REACTOR OPERATOR i

l

82. EPP-026/27-14 001 Given the following plant conditions:

i j e The unit is in Hot Shutdown s

SUT is supplying all 4KV buses

!

A severe short on the "B" DC Bus has caused the loss of that bus j 4

Which ONE (1) of the following describes the response of the emergency diesel generators

(EDG's)? 4 3

a ,

i A. "A" EDG starts, "B" EDG air start solenoids fail closed preventing start.

1

4. "A" EDG remains available, "B" EDG starts but the exciter field will not flash and the l output breaker will not close, d

C. Both EDG's will auto start, "B" EDG will not load due to SUT already powering E-2, ;

"A" EDG output breaker will close onto the bus.

l i

j D. Both EDG's will auto start, "A" EDG will not load due to SUT already powering E-1, j "B" EDG output breaker will close onto bus.

b i

i K/A 000058.K3.01 3.4/3.7 i

EPP-26/27, Attachment 1 New Question Question 82 of 100 t_ -,_

1 Rsv. 2 LOSS OF DC BUS *B"

.~' EPP-27 l

'+ Page 16 of 27 i

i i- INFORMATION USE ,

ATTACHMENT 1 ,

M.UOR EFFECTS / IDAD LIST ,

(Page 1 of 4)

Maior Effects:

t Reactor Will trip due to loss of power to 52/RTB undervoltage coil.

Turbine Will trip via 20/AST from Rx Trip-(20/ET has lost power).

Generator Will receive lockout signal. However, 86P cannot open OCB 52/8 & 52/9 due to the loss of their control power. ,

This causes a Breaker Failure scheme which trips OCB 52/3,-

52/6, 52/7, 52/12 and the downstream breakers on the r Darlington SCPSA line. The Exciter Field Breaker will open.

4KV Busses 1 & 2 If initially on SUT, nothing will happen. If initially on :

UAT, the busses will auto-transfer due to the Rx Trip. l i

In either case, 4KV busses 1 and 2 and all downstream busses and equipment will remain' energized. ,

4KV Bus 3 Will remain energized on the SUT. 4KV Bus 3 and 480V Bus 3 ;

will lose DC Control Power (including a loss of protective relaying).

4KV Buss'es 4 & 5 4KV Bus 4 will try to auto-transfer to Bus 3 but cannot due to the loss of DC Control Power. Thus, 4KV Busses 4 & 5 and all downstream busses and equipment will deenergize. ,

4KV Bus 4 and 480V Bus 4 will lose DC Control Power (including a loss of protective relaying)'. Control Power l (and protective relaying) will remain for 4KV Bus 5 and ;

480V Bus 5. ,

1 Emergency Bus E-1 Will remain energized. SST 2F will lose cooling fans. ;

Emergency Bus E-2 Will remain energized on the SUT but will-lose DC Control !

Power (including a loss of protective relaying). SST 2G-will lose cooling fans.

DS Bus Will remain energized with Control Power available. ;

EDG A Remains available, if needed. f EDG.B Auto-starts due to loss of power to air start solenoids .j but will not field flash and output breaker will not close.

I t

~

i-

l

'o' 95-2 NRC EXAM - SENIOR REACTOR OPERATOR i

84. AOP-033-06 001 ,

Given the following plant conditions:

The plant is at -50" on the RCS standpipes l

l . The running RHR pump is lost due to an RCS leak l

! e The leak is stopped and RCS level is now -70" l LC Which ONE (1) of the following.parametersgifies-conditons necessary to start the RHR

! pump?

t t

i

A. Level is adequate to restart the RHR pump.

B. Level must be increased to a minimum of-58" prior to restart. W fc M **'

C. Level is adequate, but pump must be vented prior to restart.

@. Level must be increased to a minimum of -68" prior to restart.

d K/A 000025.EA1.02 3.8/3.9 AOP-020, Step 33 New Question l

I i

Question 84 of 100

o 95-2 NRC EXAM - REACTOR OPERATOR J

. 83. AOP-033-06 001 -

I Given the following plant conditions:

I

The plant is at -50" on the RCS standpipes '

The running RHR pump is lost due to an RCS leak l

The leak is stopped and RCS level is now -70" Which ONE (1) of the following parameters satifies conditons necessary to start the RHR i pump?

1 A. Level is adequate to restart the RHR pump.

B. Level must be increased to a minimum of-58" prior to restart.

C. Level is adequate, but pump must be vented prior to restart.

@. I evel must be increased to a minimum of -68" prior to restart.

d K/A 000025.EA1.02 3.8/3.9 AOP-020, Step 33 New Question Question 83 of 100

. . _ . . ~ _ _ _ . . . . _ - _ _ _ _ . - . _ ..._ _ . ___ __ _ . - _ . _ _ . . __._ . _ _ _ _ _ _ _ - - _

l *

. Rav. 15 1 AOP-020. IDSS OF RESIDUAL HEAT REMOVAL (SHUTDOWN COOLING) .

Page 20 of 98 STEP -

INSTRUCTIONS RESPONSE NOT OBTAINED 1

l

33. Establish Conditions To Start An )

RHR Pump As Follows: l

a. Check RCS level.- EQUAL TO QE a. Evaluate Core Exit !

l ABOVEi-68 INCHES-(704.RVLIS Thermocouple (CETC) readings FULL RANCE) for determination of further l recovery actions

l 1) H CETCs indicate greater l than or equal to 200*F, M Go To Attachment 3. ,

2) H CETCs indicate less than to 200*F @ SI flow has been established, M Go To Attachment 3.

3) H CETCs indicate less than 200*F M SI flow has l HQI been established, M j Co To Step 31.

b. Check RHR trains - ALL THE b. Perform the following:

FOLIDWING SATISFIED FOR AT LEAST ONE TRAIN 1) Continue attempts to make an RHR train available.

Train intact

2) Evaluate Core Exit

Power available to RHR Thermocouple (CETC)

Pump readings for determination ,

of further recovery j

CCW available ,

actions: ;

RHR Pump vented as a) H CETCs indicate necessary greater than or equal ;

to 200*F, M Go To '

FCV-605, RHR HX BYPASS, Attachment3.

operable b) H CETCS indicate less >

HCV-758, RHR HX OUTLET T0 than 200 *F M SI flow COLD LEGS, operable has been established, M Go To Attachment 3.

c) H CETCs indicate less than 200*F M SI flow '

has EDI beeh established, M Go To i Step 33.a. r 1

(CONTINUED NEXT PAGE) i

t t

. Rav. 15 i

AOP-020 IDSS OF RESIDUAL HEAT REMOVAL (SHUTDOWN COOLING)

Page 21 of 98 l

STEP -

INSTRUCTIONS RESPONSE NOT OBTAINED l 33. (CONTINUED)

c. Verify OPEN the following J

valves: 3

RHR-750, RHR IDOP SUPPLY

i

RHR-751, RHR 1h0F SUPPLY

!

RHR-744A, RHR COLD LEG INJ

RHR-744B, RHR COLD LEG INJ d

1 I i f

4

, i

} I

- \

l l

_ _. _ _ _- m . _.

~

. Pav. 15 A0P-020 IASS OF RESIDUAL HEAT REHOVAL (SHUTDOWN COOLING)

Page 22 of 98 l -

STEP -

INSTRUCTIONS RESPONSE NOT OBTAINED l ******;i.;;.;;**;;;;;;****************************w* ;;;;;;;****************

l CAUTION i

RCS level may decrease due to void collapse when an RHR Pump is starm..e if RCS temperature is greater than 200*F.

i

34. Determine If Additional Hakeup Is Required To Compensate Fcr Void Collapse:

a. Check Core Exit Thermocouples a. Go To Step 35.

- GREATER THAN 200*F

b. Check Pressurizer level - b. Go To Step 35.

OFFSCALE IAW

c. Check SI Cold Leg 0.1 Hot Leg c. Perform the following:

injection - ESTABLISHED

1) Establish Hot Leg injection as follows:

a) Verify OPEN the following valves:

SI-856A @_D SI-856B, SI PUHP RECIRC

SI-864A AND SI-864B, RUST DISCH

= SI-878A AND SI-878B, SI DISCH CROSS CONN

SI-869, SI HOT LEG HDR b) Start one SI Pump.

c) Using Key #31 ,0_R 32, open SI-866A, IDOP 3 HOT LEG INJ, Of SI-866B, IDDP 2 HOT I.EG INJ.

d) Observe flow on FI-940.

(CONTINUED NEXT PAGE)

s 95-2 NRC EXAM - SENIOR REACTOR OPERATOR 1,

! 85. DC-14 002 l' Given the following plant conditions:

The reactor has tripped due to a turbine trip

.

Upon investigation you note that both 230Ks generator output breakers have failed to -

open Which ONE (1) of the following describes the failure that caus these breakers to fail to open? .,

ArL !~8< & c 7,gg T

i A. Loss of "A" DC Bus NAv- ' '

_&st J.n,' j .)

l , ,

1 i

s-l W. Loss of "B" DC Bus

. C. Loss of "C" DC Bus I

D. - 1.oss of "DS" Distribution Panel "B" 1

[ b l

K/A 000058.A2.03 3.5/3.9 EDP-004, pg 7 l EDP-007, pg 17 Modified Question i'

i 1

4 i

l I

l 1

I Question 85 of 100 i

. . . _ _ . . ~ . _ . _ _ _ . _ ._ . . _ _ - . . _ . - _ . _ _ _ . _ . _ . . . . _ . _ _ . _ . _ . . . _ . _ . . .

t L

l ,

95-2 NRC EXAM - REACTOR OPERATOR :

84. DC-14 002 Given the following plant conditions: l e The reactor has tripped due to a turbine trip -

i.

e Upon investigation you note that both 230Kv generator output breakers have failed to open l Which ONE (1) of the following describes the failure that caused both of these breakers to fail L to open?

t

A. Loss of "A" DC Bus ;

I l'

I- vB. Loss of "B" DC Bus C. Loss of "C" DC Bus i D. Loss of "DS" Distribution Panel "B" b

i I

l K/A 000058.A2.03 3.5/3.9 EDP-004, pg 7

EDP-007, pg 17 Modified Question l

i i

1

~

1 l

L Question 84 of 100 !

l 6.

9.0 Power Panel PP-25 Szetion 9.0

Pego 1 of 1 PP-25 i

POWER SUPPLY: 125V DC'MCC "B" IACATION: 230KV SWITCHYARD pircuit M LWC 1 SPARE l 2 SPARE l

3 SPARE 4 SPARE 5 Cenerator 0.C.B. 52-8 (south) C-2923-A 6 Cenerator O.C.B. 52-9 (north' , C 2924-A 7 SPARE 8 SPARE 9 SPARE 10 SPARE 11 SPARE 12 SPARE 13 SPARE 14 SPARE 15 SPARE ,

16 SPARE 17 SPARE 18 SPARE 1

l l

I l

EDP-007 Rev 11 Page 17 of 52

95-2 NRC EXAM - SENIOR REACTOR OPERATOR k 86. MCD-99 003

\ Given the following plant conditions:

\

The unit is at 100% power

\

e Loss of Main Feedwater has occurred N

Which ObE,(1) of the following describes the response of the RCS temperature and press prior to an automatic reactor trip (assuming no operator action)?

\

N - I) v u VA. RCS temperatureINCREASES s and pressure INCREASES.

N 0 B. RCS temperature DECREASES and pressure INCREASES. 6 C. RCS temperature INCREA(ES and pressure DECREASES.

\ '.k g g\@(V

\

D. RCS temperature DECREASESNnd pressure DECREASES.

a :

K/A 000054.G0.I1 3.4/3.3 FRP-H.1, WOG Figure 1 periods 1 & 2 pg 6 & 7 Modified Question I

l I

l Question 86 of 100

._ m ._. . . . . .

l 95-2 NRC EXAM - REACTOR OPERATOR '!

l 6 I f 85. MCD-99 003 i

Given the following plant conditions: l

. i

. The unit is at 100% power

e Loss of Main Feedwater has occurred !

Which ONE (1) of the following describes the response of the RCS temperature and pressure

. prior to an automatic reactor trip (assuming no operator action)? j i

- i

-)

4 VA. RCS temperature INCREASES and pressure INCREASES.

i i

! B. RCS temperature DECREASES and pressure INCREASES. l C. RCS temperature INCREASES and pressure DECREASES.

]

D. RCS temperature DECREASES and pressure DECREASES.

l a -

1- i 1 '\

K/A 000054.G0.ll~ 3.4/3.3

{

j FRP-H.1, WOG Figure 1 periods .1 & 2 pg 6 & 7 4

Modified Question !

i i

i .

f i

i Question 85 of 100

i

. 1

! s l 2.1 Loss of All Feedwater Event from a Power Condition Without Operator Action i' . 1 i

i In order to better explain the restoration actions for a loss of heat sink, a l discussion will first be presented for the situation where no oper,rtor actions l t

are taken. To aid in tha understanding of the loss of all feedwater transient l l without operator action, transient results for RCS pressure from the WFLASH- !

l code (Reference 1) have been divided into periods (see Figure 1). The phenomena controlling system behavior during the varicus periods will be- ;

discussed along with the implications on the eventual restoration of a heat 1 j

sink or establishing a heat remosal path.

i period 1 l

1 l

l The loss of all feedwater transient from a power condition will begin with a loss of all main feedwater. The steam generator water levels will rapidly l decrease since steam will still be flowing to the turbine without being i replaced by feedwater. The secondary pressure and secondary fluid temperature will increase as the cooling effect of the subcooled feedwater is lost. The reduction in primary-to-secondary heat transfer rate, caused by the reduction in primary-to-secondary temperature difference and partial uncovery of the ]

steam generator tubes on the secondary side, will result in an RCS pressure and !

temperature increase. Thus, the RCS fluid will be forced to absorb some of the full core power due to degraded secondary side conditions which have reduced the heat transfer capability of the steam generators. The resultant ;

temperature increase will swell the RCS flaid causing a surge into the pressurizer, raising its level. During this period (Period 1) it is possible that the pressurizer PORVs could open to relieve the increasing pressure. Th' i s is more likely to happen in the absence of pressurizer spray operation.

However, the opening of the pressurizer PORVs will also be a function of the

reactor trip time and the amount of subcooled feedwater inventory present in

the steam generator secondaries at the start of the transient. Therefore, whether a PORV will open during this period cannot be generally predicted due to dependence on plant conditions.

l FR-H.1 5 LP-Rev. 1 0190V E_________.__

Om } ,

~ ::o, ,

== -

Figure 1. RCS PRESSURE FOR LOSS OF HEAT SINK DUE TO LOSS OF ALL FEEDWATER !

RCS Pressure (PSIG) 2soo

-w L -

2000 - -

= 3 : 4 == 5 : 6

<a 1500 -

Period 1 & 2 t

i 1000 -

[

i 500 -

I I I

o 0 25 50 75 100 ,

Time (Minutes) 90$titNIG03FDf t i

i i

I

~

i The initial RCS pressurization and heatup will be terminated when the reactor and turbine trip on either the coincident signals of feed flow-steam flow 1

mismatch and low steam generator level, or from a low-low steam generator level signal depending on plant design features. This early, short-lived phase is characterized as a power pressurization since the core remains at full power

! for about 16 seconds after main feedwater is lost. The reactor trip time is l l

' 1 dependent upon the signal source for reactor trip. Thus, a plant utilizing a '

low-low SG 1evel trip setpoint will trip at about 45 seconds.

Period 2 After the reactor trip, the RCS pressure and hot leg temperature will

immediately drop due to the reduction in core power. The pressurizer surge l

line mass flow rate will reverse and mass will flow out the pressurizer reducing the level as the RCS fluid cools and shrinks. Although not as rapidly as before the reactor trip, the steam generator levels will continue to fall as !

steam continues to be generated and relieved through either the condenser steam l dump system, steam generator PORVs or the steam generator safety valves.

Makeup by the AFV pumps is assumed to be unavailable, j!

period 3 l

1 The initial RCS depressurization after reactor trip gives way to a quasi-steady state period characterized by core decay heat energy removal through the steam generators. As secondary side mass is depleted through the condenser steam dumps, steam generator PORVs or steam generator safety valves, the steam generators will slowly dry out. During this period the RCS pressure and temperature will be relatively constant as the steam generator level continues to decrease and more of the steam generator tube heat transfer area uncovers.

There will still be sufficient secondary heat removal capability, even with a portion of the tubes uncovered, to maintain relatively stable RCS conditions for pressure, temperature and pressurizer level.

l l

t l

FR-H.1 7 LP-Rev. 1 0190V l

A i

Period 4

]

When most of the tube bundle is uncovered, the primary-to-secondary heat transfer l rate will degrade enough such that the RCS will begin to hest up. The heatup .

l rate will increase as the steam generators approach dryout. Fluid swell due to

! the heatup will be reflected in an increasing pressurtzer level and RCS pressure.

The RCS pressure is expected to increase to the pressurizer PORV setpoint at

this time. Figure I shows the Period 4 pressure rise terminating in PORV 1 opening at 1845* seconds (dashed line at 31 minutes), which occurs substantially

! (5 minutes) before steam generator dryout, j g , The actual opening time for ths PORVs can vary widely based on. pressurizer ;

conditions. The time of opening will be a function of whether or not the PORVs opened during Period 1. Also, if PORVs opened during Period 1, the duration of -l the Period 1 opening will impact the duration of PORV opening during Period 4. ]

The concentration of non-condensibles in the pressurizer will also impact PORV !

opening time during Period 4. However, the pressurizer PORVs will open during Period 4 sometime before steam generator dryout. l Period 5 Once the steam generators dry out, the secondary will no longer be capable of' RCS heat removal and virtually all the core decay heat will go into raising the RCS fluid taimperature through heat absorption. The pressurizer PORVs will cycle open and closed about their setpoints to relieve enough mass to maintain RCS pressure stable.

This opening time was calculated by the LOFTRAN code (Reference 2) which uses nonequilibrium mndelling assumptions for the pressurizer. The WFLASH code does not show PORV opening until the PRZR is water solid due to equilibrium modelling assumptions in tne PRZR.

FR-H.1 8 LP-Rev. 1 0190V

o

- 95-2 NRC EXAM - SENIOR REACTOR OPERATOR v,

i a

a

87. PATH-1-05 001 Given the following plant conditions:

A reactor trip from 100% power has occurred s

Immediate Actions are complete

. '

EPP-004, " Reactor Trip Response" is in effect

Foldout A is in effect 4

1

RCS Subcooling is 22 F Which one (1) of the following describes the actions you should take IAW EPP-004? !

I A. Im. mediate trip all RCP's.

y ;

B. If PZR level cannot be maintained > 10%, THEN initiate SI. l i

! C. Immediately trip all RCP's, then actuate SI.

i

. vD. Actuate SI and return to Path-l.

' d 1

4 ,, y ad /"a f K/A 000007.EA1.03 4.2/4.1 Foldout A

! New Question C /L i.

4 a

5 E c @'c Question 87 of 100

. - - ~- . . - , - . - . . . . . - . . . - . - . - . . . - -

~

EPP-Foldouta Rev. 17 FOLDOUTS Page 4 of 15 i

CONTINUOUS USE -

J 1

FOLD 0D i

(Page 1 of 4) i

1. RCP TRIP CRITERIA l

H 10,I}] Q conditions below are met, M stop all RCPs:

L

S1 Pumps - AT LEAST ONE RUNNING RCS Subcooling - LESS THAN 35'F (55*F]

i

, 2. SI ACTUATION CRITERIA t H EITHER condition below occurs, IEEN Actuate SI and Go To !

PATH-1, Entry Point A: l 1

RCS Subcooling - LESS THAN 35*F (55'F]

PZR Level - CAN EQI BE MAINTAINED GREATER THAN 10% [26%) i

3. AFU SUPPLY SWITCHOVER CRITERIA 1

1 4

H CST level decreases to less than 10%, IEEN switch to backup water supply using OP-402, Auxiliary Feedwater System.

4. EMERGENCY COOLING WATER SWITCHOVER CRITERIA '

B E normal cooling is lost to any of the following components, IHEN establish emergency cooling water using the referenced procedure:

i Charging Pump 011 Coolers - Use Attachment 1 of AOP-014, Component Cooling Water System Malfunction.

S1 Pump Thrust Bearing - Use Attachment 1 of AOP-022, Loss of

Service Water, j

s-

i MDA W Pr.mps - Use Attachment 2 of AOP-022, Loss of Service Water.

5 1

i

. _ . . _ ___. . _ . _ - ._. _ _._ _ _ _ , . . _ _ _ - - . ~ _ . . ~ . . _ _ . . . . _ . _ _

^

I 95-2 NRC EXAM - SENIOR REACTOR OPERATOR -)

.- 1 1 e :

)

I

, 88. OP.-105-06 001 ;

{ Given the following plant conditions:

l

. The unit is at 100% power

[ ,

]

e The running charging pump trips on overcurrent

)

The remaining charging pumps will not start 4

A load decrease at the maximum allowable rate is started ,

! l

, .

During the load decrease you receive the LO-LO Rod Insertion limit alarm ' I i l l Which ONE (1) of the following describes the actions you should take?

. 3 a

^

A. Stop the load decrease and emergency borate to clear the alarm.

[ .B . Stop the load decrease and reduce turbine load until th alarm clears.

t c s w9 ? s p; c TJ

~

i j C. Continue the load decrease at the maximum rate and emergency borate.

41 i 4). Continue the load decrease at the maximum allowable rate. / rd sa d

l K/A 000022.EK3.02 3.5/3.8 APP-005-C5 New Question fLi DotJ p7 SJffo U Question 88 of 100

I 95-2 NRC EXAM - REACTOR OPERATOR I

87. OP-105 06 001 >

Given the following plant conditions:

The unit is at 100% power i

The running charging pump trips on overcurrent i

i ,

.

The remaining charging pumps will not start >

A load decrease at the maximum allowable rate is started

!~

During the load decrease you receive the LO-LO Rod Insertion limit alarm j Which ONE (1) of the following describes the actions you should take? l 1

l i

A. Stop the load decrease and emergency borate to clear the alarm.

B. Stop the load decrease and reduce turbine load until the alarm clears. )

C. Continue the load decrease at the maximum rate and emergency borate.

@. Continue the load decrease at the maximum allowable rate.

d K/A 000022.EK3.02 3.5/3.8 APP-005-C5 New Question i

l i

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Question 87 of 100

- . - . . - . - .~.- - - . - . - . . ~ . . . . . . . . - - - - - . - - - . - - . - . - .-

j. 4' m APP-005-C5 !

Ptgs 1 of 1 U ., l 1 AleM ,

ROD BANKS ' A/B/C/D M-M LIMIT *** WILL REFIASH *** j j

AUTOMATIC ACTIONS

1. None Applicable ;

CAUSE ;

1. Excessive dilution.of Boron-Concentration j

2. Malfunction of Automatic Rod Control System ,

. i 3.

Instrument Failure

4. Plant transisnt requfIls deep rod insertion. ,

5. OST-Oli in progress (expected alarm) i OBSERVATIONS ,

-1. Control Bank RPI and Bank Step Counters l

2. Power Range NI l

3. RCS Tavg i

4. RCS Makeup Flow ACTIONS !

1. H OST-Oli is in progress, H E verify alarm clears when rod bank is i

returned to initial position.

' 2. E an RCS dilution is in progress, H E STOP the dilution. 1

- 3. Borate the RCS using OP-301, Chemical and Volume Control System, while l simultaneously withdrawing Control Rods to clear the alarm within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months l of receipt. ;

4. E a malfunction of Makeup Control is indicated, THEN refer To AOP-003, j

5. E a malfunction of Reactor Control is indicated, H E refer To AOP-001. I

6. Apply Technical Specification 3.10.1.3.

DEVICE /SETPOINTS

1. Rod Banks A and B, TC-409D, E/less than or equal to 210 steps f

2. Rod Banks C and D, TC-409F, L/ refer to Curve 1.9 in the Curve Book.

POSSIBLE PLANT EFFECH

1. Loss of required Shutdown Margin

2. Entry into Tech. Spec. LCO Action REFERENCES

1. Plant Curve 1.9 A and B.

2. OST-Oll, Rod Cluster Control Exercise & Rod Position Indication, i

APP-005 Rev. 14 Page 20 of 42

95-2 NRC EXAM - SENIOR REACTOR OPERATOR a

89. CVCS-14 002 3

Given the following plant conditions:

?

VCT level is 20" I

Level transmitter LT-ll5 fails HIGH

]

The Hagan rack switch is in the NORMAL positica l i ;

No other plant transients are in progress i

). Which ONE (1) of the following explains the CVCS system response (assume no operator ;

4 actions)?

t VA. Actual level will decrease due to LCV-Il5A diverting to the HUT's.

p. r t'a ' - 'c t c B. Actual level will increase due to continuous auto make-up signal from LT-115. !

C. Actual level will decrease and LT-ll2 will initiate an automatic make-up.

)i D. Actual level will increase until 46% then LCV-Il5A will open to divert letdown flow. 1 e

a l

i 1 i

K/A 000022.A1.08 (3.4/3.3)

AOP-003, pg 5 caution Modified Question

)I 4 t

i 3

Question 89 of 100

+

95-2 NRC EXAM - REACTOR OPERATOR

88. CVCS-14 002 Given the following plant conditions:

e VCT level is 20"

. Level transmitter LT-ll5 fails HIGH

+ The Hagan rack switch is in the NORMAL position e No other plant transients are in progress Which ONE (1) of the following explains the CVCS system response (assume no operator actions)?

VA. Actual level will decrease due to LCV-Il5A diverting to the HUT's.

B. Actual level will increase due to continuous auto make-up signal from LT-115.

C. Actual level will decrease and LT-112 will initiate an automatic make-up.

D. Actual level will increase until 46% then LCV-115A will open to divert letdown flow.

a K/A 000022. A1.08 (3.4/3.3)

AOP-003, pg 5 caution Modified Question Question 88 of 100

I

, Rsv. 3

, A0P-003 MALFUNCTION OF REACTOR MAKEUP CONTROL ,

Page 5 of 41 1 1 i i 1 I d STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED

- - - - ted;;;;;;;;;;**;;;;;;*******

l CAUTION With no operator action, LT-115 failed high will result in the loss of Charging Pump suction, i

a ************************************;;;;;;;;;;A***************************

M 4 The selection of LT-ll2 in the Hagan Rack will return indicated level

to LI-115 on the RTGB.

3. Stabilize The RCS Makeup System As Follows:

a. Place LCV-115A, VCT/HLDP TK DIV, Control Switch to VCT

b. Obtain Hagan Racks Key number 1 10 i

l c. Place the selector switch in i the bottom of Hagan Rack 19 to LT-ll2 i 1

d. Check selsetor switch in d. WHEN switch is selected to Hagan Rack 19 - SELECTED TO LT-112, THEN Go To Step 3.e.

LT-ll2

e. Place the LCV-115A Control Switch to AUTO 1

f. Contact I&C to repair failed channel l

g. Go To Step 10 i l

4. Check LT-ll5 - FAILED LOW Go To Step 6.

l

. . - . - . _ ~ _ - .. . - _ - - - - . . .. . .-. . . . . . .

..- 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

90. FP-001-06 001 Given the following plant conditions:

There is hot work in progress in the Auxiliary Building in the vicinity of the Charging Pump room with a continuous fire watch posted 1

e The hot work is completed at 1635 Which ONE (1) of the following describes the actions the fire watch should take?

Remain in the area for ....

A. 10 minutes and return paperwork to the WCC SRd.

W. ~'

l 30 minutes and return paperwork to the WCC SRO.1 C. 10 minutes and return paperwork to the CRSS.

D. 30 minutes and return paperwork to the CRSS. ,

p~ . f 3 /

\ q al b

1 l 7 , A a al 7 '

l K/A 000061.EA2.05 3.5/4.2 i FP-004, Step 6.1.5.3 New Question

I I l

c s e ' 'c .

l l

I

! i I

l l

Question 90 of 100

+

95-2 NRC EXAM - REACTOR OPERATOR t

89. FP-001-06 001 Given the following plant conditions:

There is hot work in progress in the Auxiliary Building in the vicinity of the Charging Pump room with a continuous fire watch posted

The hot work is completed at 1635 i

l Which ONE (1) of the following describes the actions the fire watch should take?

i Remain in the area for ....

1 A. 10 minutes and return paperwork to the WCC SRO. ;

I l

4. 30 minutes and return paperwork to the WCC SRO. !

t l l

C. 10 minutes and return paperwork to the CRSS. 1 l

D. 30 minutes and return paperwork to the CRSS.

i b .

J 4

K/A 000061.EA2.05 3.5/4.2 ,

i FP-004, Step 6.1.5.3 I New Question 1

1 I

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Question 89 of 100

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I Section 6.1 r Page 2 of 3 e

6.1 1 (Continued) )

. t'

5. Prior to commencement of Pot Work, during the work, and at completion of the job the Fire Watch and Lead Person shall:

1) Verify that any special instructions or accepted deviations of the Hot Work Permit have been met and that no fire hazards exist which would prohibit commencement of work. l

2) Throughout the duration of the work, regularly inspect for protection of combustibles from ignition sources, introduction of new fire hazards, comp 1*ance with posted housekeeping requirements, and evidence of fire.

3) Upon completion of the Hot Work remain in the area for at l 1 east thirty minutes and verify that affected items have sufficiently cooled to no longer present a fire hazard and I that all housekeeping provisions have been met. Satisfactory results of this inspection will be indicated by signing the

" Work Completed" Section 4 portion of the Hot Work Permit and teturning it to the Work Control Center Senior Reactor l Operator. If a fire watch is not required by Fire Protection, the' Lead Person will perform a thorough inspection to ensure, prior to leaving, that affected items have sufficiently cooled to no longer present a fire hazard and will sign Section 4.

6. If a break becomes necessary at any time within the duration of the watch reqaest, the cognizant Lead Person is to provide a qualified relief when required.

7. Request the Lead Person of the area or work activity to correct any observed noncompliance with established controls for 7 transient combustibles and housekeeping, and instructions of the Hot Work termit. If the Lead Person is not available or does not correct identified concerns, the Hot Work Job shall be stopped and promptly notify Fire Protection, the Superintendent Shift Operations, or the Work Group Supervisor. ,

FP-004 Rev. 7 Page 8 of 14 l

.-. . -. _. -- -.. - - . - . . - . - . . _ - _ = . - . _ - . . - . . - . ._. . . . - .

}

. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR j 91. Given the following plant conditions:

l e Plant is in refueling.

Power Range N-44 is out of service in accordance with the OWP.

. Core reload is in progress.

Souice Range N-31 indicates 3 cps.

. Source Range N-32 indicates 4 cps.

Which ONE (1) of the following should be performed if Power Range N-41 fails HIGH?

a VA. ly suspend all refueling operations or positive reactivity changes.

m /

B. Continue Core Reload, Place N-31 and N-32 Level Trip Bypass Switch in " Bypass".

C. Continue Core Reload, Emergency Borate the RCS to 1950 ppm.

D uspend Core Reload, Place N-51 A and N-52A in service and recommence core reload.

I a j

hK/A 000032.GO.10 0.9/3.1)

OP-002, Precaution 4.9 Modified Question i

l l

Question 91 of 100

_. _ _ .. _ _ _ . _ _ _ . _ ._ _ .-_.. . _ _ . _ _..._. _ . _ _ _ _ ~ . -.. _

' 95-2 NRC EXAM - SENIOR REACTOR OPERATOR f l

92. PATil-2-03 003 j Given the following plant conditions: I

S/G "B" has a S/G tube leak. ,

- e S/Gs "A" and "C" are intact with narrow range levels at 45 %.

Which ONE (1) of the following is the basis for maintaining greater than 10% narrow range level in the "B" steam generator?

A. Reduces the probability of a larger tube rupture by avoiding thermal stresses associated with tube uncovering.

B. Maintains inventory in the ruptured S/G in the event it is required for subsequent plant cooldown.

<. Maintains the thermal stratification layer in the steam generator to assist in RCS/SG pressure equalization.

D. Reduces the atmospheric releases if a code safety valve opened.

l c

K/A 000037.K3.07 (4.2/4.4)

/

Background Information, Path 2.

Modified Question

(, 0 0 D 1

I1 I

Question 92 of 100

- ~ , _ . --._)

4

.' 95-2 NRC EXAM - REACTOR OPERATOR ,

, t 5

91. PATH-2-03 003 Given the following plant conditions: I

S/G "B" has a S/G tube leak.

S/Gs "A" and "C" are intact with narrow range levels at 45%. j Which ONE (1) ef.the following is the basis for maintaining greater than 10% narrow range ,

level in the "B" steam generator?

\

A. Reduces the probability of a larger tube rupture by avoiding thermal stresses associated with tube uncovering.

1 B. Maintains inventory in the ruptured S/G in the event it is required for subsequent plant cooldown.

I

<. Maintains the thermal stratification layer in the steam generator to assist in RCS/SG pressure equalization.

D. Reduces the atmospheric releases if a code safety valve opened. l c (

l l

K/A 000037.K3.07 (4.2/4.4) l Background information, Path 2. l Modified Question l

I i

.I i

l l

Question 91 of 100 1

i

l

\

~

?

STEP DESCRIPTION TABLE FOR E-3 STEP 4 1

I STEP:. Check Ruptured SG(s) Level l PURPOSE: o To reduce feed flow to the ruptured steam generators to minimize the potential for steam generator overfill o To establish and maintain a water level in the ruptured steam generators above the top of the U-tubes in order to promote thermal stratification to prevent ruptured steam generator depressurization BASIS:

1 Following a steam generator tube rupture, primary-to-secondary leakage into the affected steam generator will exceed steam flow and lead to an accumulation of water in the steam generator. Feed flow will increase the rate of accumulation and reduce the time at which steam generator overfill i would occur. Hence, feed flow to the ruptured steam generator should be I minimized.

It is also important to maintain the water level in the ruptured steam generator above the top of the U-tubes. When the primary system is cooled in subsequent steps, the steam generator tubes in the ruptured steam generator will approach the temperature of the reactor coolant, particularly if reactor coolant pumps continue to run. If the steam space in the ruptured steam generator expands to contact these colder tubes, condensation will occur which )

would decrease the ruptured steam generator pressure. As previously '

demonstrated (see Step 3), this would reduce the reactor coolant subcooling )

margin and/or increase primary-to-secondary leakage, possibly delaying SI termination or causing SI reinitiation. Consequently, the water level must be maintained above the top of the tubes to insulate the steam space. In addition to insulating the steam space, this ensures a secondary side heat sink in the event that no intact steam generator is available and also provides protection against misdiagnosis of the ruptured steam generator due l to an imbalance of feed flow. '

ACTIONS:

o Check ruptured SG narrow range level o Close feed flow control valves l

INSTRUMENTATION: I 1

o SG narrow range level indication o Feed flow indication l o Feed flow control valves position indicat-ion.

l

! 1 l- E-3 58 LP-Rev. 1A 0202V:1b

STEP DESCRIPTION TABLE FOR ES-3.3 STEP 8 :

l 1

111P: Check Ruptured SG(s) Narrow Range Level - GREATER THAN (27)% [(28)%

FOR ADVERSE CONTAINMENT]

l PURPOSE: o. To maintain ruptured SG tubes covered o To refill the ruptured SG with feed flow to aid in cooling the RCS and ruptured SG I

A B_ASll: j i

When level is in the narrow range, the steam region in the ruptured SG is insulated from colder water in the U-tubes region by a layer of warmer I water. Consequently, pressure is maintained in the ruptured steam generator when the RCS is cooled by the intact SGs. Steam flow from the ruptured SG ;

will decrease the water level. If the U-tubes uncover, the ruptured SG !

pressure could rapidly decrease due to condensation of steam on the cooler I

surface of the U-tubes. This rapid depressurization could reinitiate break )

4 flow and may result in a loss of RCS pressure control or SI reinitiation. :

Thus, in order to maintain the U-tubes covered, feed flow to the ruptured SG may be required. This feed flow will also reduce radiological releases by diluting the contaminate secondary side water and minimizing steam release.

However, flow should be initiated slowly to avoid a rapid decrease in SG pressure due to condensation of steam by cold feed flow. In some cases, SG pressure may increase as feed flow compresses the steam bubble. If this ,

occurs, feed flow should be stopped prior to lifting the PORV or safety valve l on the ruptured steam generator. i ACTIONS: l

! o Refill ruptured SG using feed flow o Monitor ruptured SG pressure and level 1

I INSTRUMENTATION: l o SG narrow range level indications '

o Feed flow indication o SG pressure indication

, I 4

[ ES-3.3 40 LP-Rev. IB I l 0100V

= 95-2 NRC EXAM - SENIOR REACTOR OPERATOR ]

] :

i J

7 i 93. PATH-2-03 002

/Given 'the following plant conditions: !

t

The plant was at 100% power

reactor trip and safety hijection occurred due to a S/G Tube Rupture

Path- has been completed and a transition to Path-2 was made

RCP's ar till operating l v :

Which ONE (1) of e following describes the RCP trip criteria while responding to a SGTR using Path 27 RCPs should be tripped.... ;

I

/ \

A. ONLY when the RCP trip cr ria are met DURING the cooldown and depressurization. ;

4. during Path 2 ONLY when the o tor is specifically directed to trip the RCPs by a !

PATH 2 procedural step. '

l C. during Path 2 ONLY if the RCP trip criter are met AFTER completion of the cooldown j and depressurization. )

i D. anytime the RCP trip criteria are met. i i

I b i l

K/A 000038A101 (4.5/4.4) -+ A6 frfy 1o n 4:1.- J/1 /e ch fe., S6Te I

Path 2. l Modified Question 90 rtY \

l hew Question 93 of 100

. . . - - . - - - .- - . - ~ . . _ . - .. - -- . . _ . . . . . - . - -

95-2 NRC EXAM - REACTOR OPERATOR

92. PATH-2-03 002 Given the following plant conditions:

The plant was at 100% power

A reactor trip and safety injection occurred due to a S/G Tube Rupture !

a Path-1 has been completed and a transition to Path-2 was made

. RCP's are still operating i Which ONE (1) of the following describes the RCP trip criteria while respending to a SGTR using Path 27 I

RCPs should be tripped.... i A. ONLY when the RCP trip criteria are met DURING the cooldown and depressurization.

4. during Path 2 ONLY when the operator is specifically directed to trip the RCPs by a PATH 2 procedural step.

C. during Path 2 ONLY if the RCP trip criteria are met AFTER completion of the cooldown and depressurization.

D. anytime the RCP trip criteria are met.

l b {

K/A 000038A101 (4.5/4.4)

Path 2.

Modified Question l

l Question 92 of 100

~

95-2 NRC EXAM - SENIOR REACTOR OPERATOR I5lV*'

l

94. NI-14 006 Given the following plant conditions:

1

The unit is hot shutdown with the Reactor Trip breakers closed

Source Range Channel N-31 is behaving erractically

N-31 was removed from service using the OWP e After being removed frora service the control power fuses blow Which ONE (1) of the following describes (the actions that will occu'?l?

~e y[Qj;~r~Gqw \

w " ; a " e cy a ti .L o M A. Nothing, the channel is removed from service.

B. A high flux at shutdown alarm will occur and can only be cleared by placing high flux at shutdown in block.

W. The reactor trip breakers will open due to a high flux trip signal from channel N-31.

i --

D. A source range high flux trip signal will simunciat]but the trip breakers will remain shut.

" -~

7 ,

c

./ \

K/A 000032.EA2.06 3.9/4.1 -9 A 6; I 7 ry 1 Logic dwg sht 3 of 18 ccodb c e "f*4 1-iff New Question

&3 rE l

t l

l l

l Question 94 of 100

i 95-2 NRC EXAM - REACTOR OPERATOR

. 1 i

93. NI-14 006 '

Given the following plant conditions

i i

, e The unit is hot shutdown with the Reactor Trip breakers closed

. I

Source Range Channel N-31 is behaving erractic. ally

= W-31 was removed from service using the OWP

[

i.'

l

After being removed from service the control power fuses blow l l Which ONE (1) of the following describes the actions that will occur?

! I j A. Nothing, the channel is removed from service.

t

B. A high flux at shutdown alarm will occur and can only be cleared by placing high flux at i shutdown in block.

l 4 W. The reactor trip breakers will open due to a high flux trip signal from channel N-31. j D. A source range high flux trip signal will annunciate, but the trip breakers will remain shut.

C K/A 000032.EA2.06 3.9/4.1 Logic dwg sht 3 of 18 New Question Question 93 of 100

- . _ _ __. ~. _ __ _ _ . . _ . _ _ _ - _ _ _ _ __ . . . _ _ _ _ _ _ . . . -

n

~

,- 95-2 NRC EXAM - SENIOR REAC'iOR OPERATOR

95. OMM-022-14 003 -

t Given the following plant conditions:

\

e The plant was operating at 100% power

!~

h e A reactor trip and Si have occurred

/

Path-1 has directed that' Supplement D, " Emergency Recirculation Equipment" be verified -

7 ,

f -

Which ONE (1) of the following 41escribesBthewiens permitted during performance O o' Supplement D, " Emergency Recirculation Equipment"?

A. Restoring flowpath from containment sump to RHR. 3 ve/-e g,,.c,,/3 f.,,

B. Aligning flowpath from RI-IR pumps to the SI pumps.

t 6 -u %' y' W. Restoring control power to SI valves controlled from the RTGB.

D. Aligning containment spray pumps to containment sump.

c

$4,;g fonc'; i' * ~

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K/A 000007.G0.12 3.8/3.9 . q n f/4(/< 1 OMM-022, rv8, pg 15, after bullets Modified Question Question 95 of 100

95-2 NRC EXAM - REACTOR OPERATOR

94. OMM-022-14 003 Given the following plant conditions:

Ihe plant was operating at 100% power

A reactor trip and Si have occurred

e Path-1 has directed that Supplement D, " Emergency Recirculation Equipment" be verified Which ONE (1) of the following describes the actions permitted during performance of Supplement D, " Emergency Recirculation Equipment"?

)'

\

A. Restoring flowpath from containment sump to RHR.

B. Aligning Howpath from RHR pumps to the SI pumps.

W. Restoring control power to Si valves controlled from the RTGB.

D. Aligning containment spray pumps to containment sump.

c K/A 000007.G0.12 3.8/3.9 OMM-022, rv8, pg 15, after bullets Modified Question Question 94 of 100

.. . - - . . . - . - - - . . -- .. .- ~.-

.~

5.1.3 (Continued)

Supplement D is a listing of valves and components which must be available for Cold Lag Recirculation. PATH-1 has a step which asks if Supplement D components are available. This means that Supplement D is to be reviewed to ensure that the valves or components listed are capable of being repositioned when the transition has been made to EPP-9, Transfer to Cold Leg Recirculation. It should be noted that all Supplement D components are not required to be capable of being repositioned. As a minimum the following are required:

e One flowpath from the CV sump to the RHR Pumps.

e One flowpath from the RHR Pumps to the SI Pumps.

e One flowpath from the required pumps to the core.

e Pumps as specified in the Supplement.

When referenced by PATH-1, Supplement D should HQI be used as permission to realign the valves included on that Supplement. It is acceptable, however, to restore control power to SI valves on the RTGB.

Supplement E contains parameters to be monitored to verify that natural circulation flow exist-s. This' allows Operations the option of performing a natural circulation cooldown in accordance with EPP-5 or maintaining current plant conditions while on natural circulation using Supplement E.

Supplement F provides a listing of Emergency Diesel Generator loads in kilowatts. The information in the Supplement allows the Operator to determine if he has adequate capacity on the EDG to load-additional components or if loads must be stripped from the bus prior to loading any additional components. The kw values listed in the Supplement are based on the maximum horsepower rating of each component listed. Therefore a pump that is running at minimum speed on recirculation would not provide the full value specified in the Supplement wt n stripped from the bus.

QMM4022- Rev. 8 Page 15 of 43

. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR

%. IA-14 001 Given the following plant conditions:

The unit is at 100% power

PCV-1716 IIA to CV, is inadvertantly closed [

r Which ONE (1) of the following describes the actions that will automatically occur asair lr

~~

pressure-is-tost-irrcont iinme1H?'

A. CVC-310A, Loop 1 Hot Leg Charging, fails closed. V:T4 t

B. PCV-455A and B, PZR Spray, fails open. S f der t C. CVC-204A, Charging line isolation, fails open. _ _ .-m

@. CVC-460A & B,' Letdown isolation, fails shut L d $0 y [,,wyA 1 ##87 Tygt FA ll M'"T N'%- .__ _ __...-

]

K/A 000065.EK3.03 2.9/3.4 AOP-017, Attachment 1 pg 12 Modified Question l

P)[ t:10R Y l e_

l wb, n 61 mk See feJr

<eya-s I

l l

1 l

l Question 96 of 100 l 1

l ,

95-2 NRC EXAM - REACTOR OPERATOR 4

I ' 98. IA-14 001 -

[ Given the following plant conditions: !

l

The unit it, at 100% power

PCV-1716, IA' to CV, is inadvertantly closed i Which ONE (1) of the following describes the actions that will automatically occur as air pressure is lost in containment?

]

1.

e

A. CVC-310A, Loop i Hot Leg Charging, fails closed.

. B. PCV-455A and B, PZR Spray, fails open.

e i

C. CVC-204A, Charging line isolation, fails open.

@. CVC-460A & B. Letdown isolation, fails shut.

i t

d i-

! K/A 000065.EK3.03 2.9/3.4 AOP-017, Attachment 1 pg 12 l Modified Question i

i 4

4 N

Y 4

4 1

Question 98 of 100 i

..- . . - . - . . . . . - . .-... - ~.. - -. ..... - - . .- ~.-- . - - . - . _ _

4.

i Rsv. 17 i : ,, , AOP-017 LOSS OF INSTRUMElft AIR

' Page 35 of 58

! q'

! l INFORMATION USE s

EPJACHMENT 1 ,

MAJOR COMPONENTS AFFECTED BY IDSS OF IA (Page 1 of 5)

{

, 1.- Chemical and Volume Control System Components FAIL POSITION i

t a. FCV-113A,'BA.TO BLENDER OPEN !

p

b. FCV-114A, PW TO PLENDER

CLOSED

{ c. FCV-113B, BLENDED MU TO CHG SUCTION CLOSED 1 <

4" d. FCV-114B, BLENDI;D MU TO VCT CLOSED l F e. HCV-105, BORIC ACID TK B RECIRC CLOSED s

i'

f. HCV-110, BORIC ACID TK A RECIRC CLOSED l

! j

g. ICV-115A, VCT/HLDP TK LJ FAILS TO VCT l h. LCV-1155, EMERG MU TO CHC SUCTION CLOSED L

1. CVC-310a, LOOP 1 HOT LEG CHG OPEN. ,

, ,j. CVC-310B, LOOP 2 COLD LEG CHG OPEN

k. CVC-311, AUX PZR SPRAY CLOSED l'

1. LCV;460 A & B, LTDN LINE STOPS CLOSED

[

m. CVC-200'A, B & C, LTDN ORIFICES CLOSED f4

n. CVC-204 A & D,=LTDN LINE ISon CLOSED.

c. TGV-143, VCT/DEMIN DIV FAILS TO VCT

p. TCV-144, NON-REG HX OUTLET TEMP CONTROL OPEN

q. PCV-145, LETDOWN PRisSSURE PCV OPEN i

i r. HCV-121, CHARGING FIDW OPEN

s. CHARGING PUMP SPEED CONTROL FULL SPEED l- t. CVC 303 A, B & C, SEAL LEAKOFFS OPEN

u. HCV-137,' EXCESS LTDN FLOW CIDSED

}_ (CONTINUED NEN.T PACE) r _. - --

d m -.

Rav. 17

.' A0P-017 LOSS OF INSTRUMENT AIR Peg 2 36 of 58 i

, INFORMATION USE ATTACHMENT 1 MAJOR COMPONENTS AFFECTED BY LOSS OF IA

{

(Page 2 of 5)

, 1. (CONTINUED)

v. CVC-387 EXCESS LTDN STOP CLOSED I

w. CVC-389, EXCESS LTDN DIV

FAILS TO VCT

'4

x. CVC-307. PRI SEAL BYP ISO CLOSED

y. CVCS HUT LEVEL CONTROLLERS FORCED LOW

2. Component Cooling Water System Components FAIL POSITION

a. CC-739, CCW FROM EXCESS LTDN HX CIASED

3. Containment Ventilation System Components FAIL POSITION

a. CV VENTILATION IS01ATION VALVES CLOSED

4. Feedwater and Condensate System Components FAIL POSITION

a. FEED REG VALVES CLOSED

b. FEED REG BYPASS VALVES CIDSED

c. LCV-1417A, HOTWELL LEVEL c0NTROL VALVE OPEN .

d. LCV-1530A, HEATER DRAIN TANK LEVEL CONTROL VALVE AS IS

e. LCV-1530B, HEATER DRAIN PUMPS SUCTION DUMP TO CONDENSER OPEN

5. Instrument Air System Components FAIL POSITION

a. PCV-1716, INSTRUMENT AIR ISO TO CV CLOSED

6. Isolation Valve Seal Water System Components FAIL POSITION

a. PCV-1922 A & B, IVSW AUTO HEADER ISOLs OPEN

7. Main Steam System Components FAIL POSITION

a. MAIN STEAM IS01ATION VALVES CLOSED

b. STEAM LINE PORVs CIDSED

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

/

97. PZR-09 002 Given the following plant conditions:

/

e Th unit was at 100% power

/ ,

A pressurizer safety valve is PEN. )

iu -

Tailpipe temperature is 0F I l

Which ONE (1) of the 11owing is the expected PRT pressure?

l 1

VA. 34 psig B. 48 ps' C. psig

. 62 psig l a I l

l K/A 000008A108 (3.8/3.8)

Steam Tables Modified Question i

r e wr A ^3 On n- e

/

1 l

l Question 97 of 100

E

, , 95-2 NRC EXAM - REACTOR OPERATOR p

95. PZR-09 002 ,

Given the following plant conditions:

The unit was at 100% power

A pressurizer safety valve is OPEN.

Tailpipe temperature is 280 F Which ONE (1) of the following is the expected PRT pressure?

VA. 34 psig B. 48 psig C. 50 psig D. 62 psig a I K/A 000008A108 (3.8/3.8) l Steam Tables -

Modified Question 1

)

l l

l 1

l

)

i Question 95 of 100 ;

~

95-2 NRC EXAM - SENIOR REACTOR OPERATOR i

. 98. RCS-10 002 Given the following plant conditions:

Tavg = 560 F (all loops) ]

1 .y

Reactor Power = 41 % M OR1 1

Tref = 558 F \ orw I 1.,

e ( f c in

\

U,o

Mw = 460 gross ,/ l l

Which ONE (1) of the following describes the value of Pressurizer Program Level based on ~

1 the above indications?

l f , (;j c o d c E P T!od!

B. 33 % \ %J gggrE b GY W5 b

C. 35 % OA3 '

p y T f. A C 7

- @. 37 % i t

d 4

i K/A 000028.EA2.08 3.1/3.5 SD-059 New Question

, No S ? I, i.

i a

Question 98 of 100 ,,

i

.- 95-2 NRC EXAM - REACTOR OPERATOR l

l

99. RCS-10 002

Given the following plant conditions:

j e Tavg = 560 F (all loops) 4 1

Reactor Power = 41 % l

Tref = 558 F

', I 4

l

Mw = 460 gross Which ONE (1) of the following describes the value of Pressurizer Program Level based on ;

the above indications? I A. 30 %

i B. 33 % !

C. 35 %

@. 37 % I

, d i

I I 4

K/A 000028.EA2.08 3.1/3.5 i' SD-059 !

4 New Question 4

4 i

l l

Question 99 of 100 ,,

95-2 NRC EXAM - SENIOR REACTOR OPERATOR

99. FH-14 001 Given the following plant conditions:

The unit is at 50% power

An\ operator performing rounds in the ISFSI area determines that an air inlet on a ,

Horiz' ntal Storage Module, (HSM), is blocked

The blockhge will require maintenance assisstance to remove ,, a

< u ew v Which ONE (1) of'the following describes the design basis time available before fuel damage

" t is- expected to occur (under worst case conditions) and procedural actions t * ' the HSM?

A. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Force ater Qi ulation i vB.

/

/g/ pi 7 [~ J f M ' !

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , Forced ir kculation C.

k\

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , Forced Water Circulation 1 7 g/g/

O ! / ^'

D. 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , Forced' r Orculation d /## #

/ \

7 rb3 } IM C l b

K/A 000036.G0.07 3.2/3.5 AOP-028 Modified Question l

l p

\

17 foRTAN7 To

Erlo 4/2 ( l

,- g m; \{\

TH y'os HM De W To

t. 09 v/ 7 l

l Question 99 of 100

95-2 NRC EXAM - REACTOR OPERATOR :

100. FH-14 001 '

Given the following plant conditions:

1

The unit is at 50% power

An operator performing rounds in the ISFSI area determines that an air inlet on a Horizontal Storage Module, (HSM), is blocked

The blockage will require maintenance assisstance to remove Which ONE (1) of the following describes the design basis time available before fuel damage is expected to occur (under worst case conditions) and procedural actions to cool the HSM7 A. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Forced Water Circulation

4. 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , Forced Air Circulation C. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , Forced Water Circulation D. 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , Forced Air Circulation b

K/A 000036.G0.07 3.2/3.5 AOP-028 Modified Question Question 100 of 100

1 Rsv. 3 A0P-028 ISFSI ABNORMAL EVENTS

Page 5 of 16 STEP - INSTRUCTIONS RESPONSE NOT OBTAINED SECTION A 1

JLOCKAGE OF THE HSM DRAINS. AIR INLETS. OR AIR OUTLETS i

(Page 2 of 4)

7. Notify The Following Personnel:

I

Plant General Manager

Manager - Operations i 8. Implement The EALs i

E

! The ISFSI system design certifies that no fuel damage will occur for l at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months following a complete blockage, on the hottest days, j with the fuel at its maximum allowable heat generation rate. The i following step is designed to extend that time by performing a 4 calculation using the actual existing conditions.

1 l 9. Notify Technical Support .

4 Personnel To Perform The

l

! Following:

i Conduct A Fuel Cladding

Integrity Evaluation

, (calculation of time to i

damage) i

Evaluate the need for alternate methods of DSC

Cooling (forced air

' circulation of the HSM, etc)

10. Notify Maintenance To Determine Alternate Methods of Blockage Removal

11. Initiate Alternate Methods Of l Cooling As Directed By Technical j Support ,

12. Initiate Alternate Methods Of Blockage Removal As Determined ;

By Maintenance ;

. 95-2 NRC EXAM - SENIOR REACTOR OPERATOR 4 100. PZR-14 004 Given the following plant conditions:

The plant is operating at 100% steady state conditions

! The reference leg for LT-459, Pre:surizer Level, ruptures All plant systems are in their automatic / normal alignment 4

Assuming NO operator actions are taken

Which one (1) of the following describes the automatic actions that will occur? ;

A. Reactor will trip on high pressurizer level.

(4. Charging flow will decrease to mimmum. l 1

( C. Charging flow will increase to maximum. ;

I D. PZR pressure will increase and cycle between the PORV lift and reseat setpoints. ;

b K/A 000028.A2.02 3.4/3.8 SD-059, PZR LP Modified Question m- a 3" 0 e4 g F' "- 7

$ 9 o'g7 d.sh

[lG N b rean +e f Question 100 of 100

- .- - - - . . . - - . - . . . - - . - - . - - ~ . . . _ _ _ - - - - . - . - -

1 4

SD-059 PRESSURIZER SYSTEM i

5.1.3 Pressurizer PORV Contml (PZR-Figure 8 & PZR-Figure 13) i i The Pressurizer PORVs have two modes of contml,' Normal and low Temperature j Ove: pressure Protection (LTOPP). In normal mode the PORVs have a permissive of

2000 psig to_ open in Automatic. 'Ihis " permissive" is supplied by the protection channels meeting a 2/3 logic . As stated before PCV-456 receives its signal from FT-l 445 set at 2335 psig and PCV-455C receives its signal fmm FC-444A at +100 psi

] which is nominally 2335 psig also. When the key switch for OVERPRESSURE

! PROTECTION on the RTGB is place in the LOW PRESSURE position (one switch for j each PORV) the input to each PORV is shifted to the LTOPP contmller.

l ' 5.1.4 low Temperature Overpressure Pmtection Contml (LTOPP) (PZR-Figure 13) i, LTOPP control is required to be activated when the RCS is cooled down below 360*F i to minimize Pressurized Thennal Shock (P.S..) concems. The LTOPP controller uses the lowest of TE-410, TE-420 and TE-430 to determine RCS temperature and pressure

) as sensed by FT-500 and FT-501. The lift setpoint is variable based upon auctioneered low RCS temperature. At the highest RCS temperature that LTOPP is required to be activated, 360*F, the pressure setpoint is 400 psig. The setpoint of the Comparators PC502 and PC503 are adjusted downward as RCS temperature is decreased.

There is one alann associated with each channel of LTOPP. It actuates for 3 reasons:

(1) RCS temperature is <360*F and LTOPP is not " armed" low Pressure not selected on the key switch for OVERPRESSURE PROTECTION, (2) The PORV has received .

and actuation signal based upon current pressure and temperature or (3) the associated Block valve is shut. l 5.1.5 Pressurizer Level Control (PZR-Figure 12)

Pressurizer level is controlled by controlling charging pump speed. The level is I

PZR Page 26 of 36 Revision 0 INFORMATION USE ONLY

i SD-059 PRESSURIZER SYSTEM ,

programmed to ramp up as Tavg increases by II-459G. This maintains approximately constant mass in the RCS as Tavg is increased and the coolant in the RCS expands.

j I.evel program is 22.2% at Tavg of 547'F and 53.3% at Tavg of 575.4*F. :

l j There are 3 Pressurizer level channels LT-459, LT-460 and LT-461. LC-459G the !

l Pressurizer level controller is normally fed by level channel LT-459 but can be replaced

! by LT-461 with a selector switch on the RTGB. The output of LC-459G is then fed to j the charging pump speed controllers to control speed of the charging pump if their f controllers are selected to Auto. j l

4 l If Pressurizer level increases 5% above pmgram LC-459D will turn on the backup

- heaters and sound an annunciator for High level Heaters on.

On Pressurizer low level of 14.4%, proportional and backup heaters are deenergized and

! letdown is isolated by shutting LCV-460A & B if respective control switches are in j auto. II-459 and the LC-460, the low level bistables, are normally supplied by LT-459 i and LT-460 r@vely but either can be replaced by LT-461 with a selector switch on the RTGB. i I

It has been noted by checking the Control Wiring Diagiums that II-459 will only tum off the backup heaters that are selected to Automatic where 1f-460 will turn off the i backup heaters in Automatic or Manual. The only time this would have any bearing would be in the event of an instrument failure. If the channel feeding LC-459, usually ,

LT-459, were to fail low the proportional heaters and any backup heaters in Automatic would deenergize and any backup heater in manual would remain energized. I 5.1.6 Pressurizer Level Control Setpoints l

1. I.evel program as function of Tavg !

(TM-459) i for Tavg 547'F 22.2 % of level span j 1

PZR Page 27 of 36 Revision 0 l

INFORMATION USE ONLY l

SD-059 PRESSURIZER SYSTEM

=

for T,575.4*F 53.3% oflevel span

, (Program in linear from 547'F to 575.4*F) i I.ow limit 22.2% of level span High limit 53.3 % of level span i i

2. I.ow-low Level Heater Cutout

] (LC-459C, LC-460C) 14.4.% oflevel span 4

3. Ievel Controller (LC-459F) 10% charging pump l Proportional gain speed /% level deviation Reset time constant 430 seconds

4. I.etdown Valve Isolation 14.4% of level span

5. Back-up Heaters on +5% of programmed level l

6.0 SYSTEM OPERATION l 6.1 Normal Operation Insurge of RCS Coolant - produced by increase in Tavg. An insurge of coolant will reduce volume of the steam bubble causing an increase in the temperature and pressure of the steam. The steam space or bubble becomes superheated and some minor condensation occurs at surface and on walls.

4 The increased pressure causes the spray valve to open which cools and condenses a part of the steam bubble, thereby reducing pressure.

The increase in level will energize backup heaters if the level increases to 5% above program.

PZR Page 28 of 36. Revision 0 INFORMATION USE ONLY

. . - - -~ _ - - - . .- .-

a l

1 SD-059 PRESSURIZER SYSTEM i

l i Outsurge of RCS Coolant !

l An outsurge of RCS coolant will increase the volume of the steam bubble, which will .

cause water to flash to steam, limiting the pressure decrease.

l 1

l Pmportional beaters will be full on to limit pressure decrease. If pressure decrease is !

large enough, the backup heaters will energize.

l 6.2 Infrequent Operations ;

! 6.2.1 RCS Heatup j The pressurizer is initially water solid with pressure being contmiled by charging and ladown from RHR. A bubble is fcnned by energizing the heaters, heating the system .

i to saturation, then draining to proper level. Bubble formation is indicated by an increase I l in letdown flow when the Pressurizer reaches saturation. The rate of heating the .

Pressurizer is limited by Technical Specifications.

6.2.2 RCS Cooldown

During an RCS and Pressurizer cooldown a bubble is maintained until at least 350*F, j at which time the bubble may be collapsed with cool water from CVCS after placing the

RCS on RHR. The bubble must be collapsed prior to cooling the RCS below 150*F.

The cooldown rate of the Pressurizer is limited by Technical Specifications.

l i,

6.3 Abnormal Operation l

h The Pressurizer antam=411y repands to all abnormal conditions via automatic controls 4 or safety valves.

e I

PZR Page 29 of 36 Revision 0 INFORMATION USE ONLY l

_ _ _ I

! 95-2 NRC EXAM - REACTOR OPERATOR 1 ;

i l 14. Given the following plant conditions: >

l l

The unit is at 50% power

All cystems are aligned for normal operation f i

l

A transient occurrs that results in Control Room Ventilation shifting into Pressurization Mode Which ONE (1) of the following describes the signal that would directly cause this transfer?

VA. Si actuation. l B. Manual Phase "A" actuation. ,

l C. Manual Phase "B" actuation.

D. R-11, Containment Radiation high alarm.

a K/A 013000.K1.01 (4.2/4.4)

Logic dwg 5379-2759 sht 8 of 18 Modified Question i

Question 14 of 100

l 95-2 NRC EXAM - REACTOR OPERATOR f l

23. CVCS-14 001 Given the following plant conditions:

The plant is at 100% power

All systems are aligned for normal operation l

Which ONE (1) of the following describes a condition that would cause LCV-460A and B, ;

Letdown Stop Valves, to CLOSE with their control switches in AUTO? !

VA. Pressurizer level at 14 %.

B. Containment isolation Phase A ACTUATED.

C. Pressurizer level 5% below program.

D. Closing the letdown orifice isolation valves, CVC-200A, B and C.

a l

l K/A 004010.A4.02 (3.6/3.1) ]

Logic dwg 5379-3693 sht 17 of I8 i Modified Question

/[ Q 1 6 1

\ / M p6 .pn ;

ch '

y 4

l

\

t l

! Question 23 of 100 i

95-2 NRC EXAM - REACTOR OPERATOR 1 .- j 1

l l l

27. ESF-09 001 Given that 10 minutes ago the following events occurred:

An MSIV inadvertently closed causing secondary safeties to lift and a reactor trip signal i to be generated.

The reactor trip breakers failed to open from the control board and locally. 1 The operators tripped the reactor by locally tripping the MG sets.

An Si occurred prior to the secondary safeties reseating.

The operators inadvertently secured AFW Pump "A" from the RTGB.

RCS pressure is now 1650 psig and increasing slowly.

SI has NOT been reset.

Which ONE (1) of the following must occur to restart the "A" AFW from the RTGB?

A. Pressurizer pressure must increase to greater than the Si setpoint.

l B. The AFW pump breaker must be racked out and then re-installed. I C. The reactor trip breakers must be manually opened tc' actuate the feedwater isolation signal.

@. The control power fuses must be removed, then reinstalled in the "A" AFW pump breaker.

d l

1 S/D-006, Pg.13. () l K/A 013000.K4.10 (3.3/3.7) i Modified Question l

1 l

Question 27 of 100 l

l

^

l

l l

5.0 OPERATION (Continued) l CAUIIDH l UNTIL THE SAFEGUARD SIGNAL IS MANUALLY RESET, ANY SAFEGUARD EQUIPMENT STOPPED FROM RTGB CANNOT BE RESTARTED WITHOUT REMOVING THE "0NTROL POWER FUSES AT THE BREAKER AND REINSTALLING THEM. THIS IS DUE TO l ANTI-PUMP FEATURE IN THE BREAKERS. !

The Phase "A" Containment isolation, Containment Ventilation isolation, and Feedwater isolation must be reset individually after the SI signal is reset or cleared, j i

1 5.2 Sorav Actuation When a containment Spray signal is generated, it will perform the l following:

1. A. Automatic Signal:

Steamline Isolation - Shuts all three Main Steam isolation valves.

B. Manual Signal:

- e C.V. Ventilation Isolation - Shuts C.V. Purge Valves, Pressure Relief Valves, and Vacuum Relief Valves.

2. Spray Actuation - It starts both Containment Spray Pumps, and opens valves 880A, 880B, 880C, 880D, 845A, and 845B. This will deliver borated water with sodium hydroxide to the CV atmosphere to depressurize and remove free iodine.

3. Phase "B" Containment Isolation - This signal will further isolate the containment by shuttin5 containment isolation valves as follows:

CC-716A CC to R.C.P. "A", "B",."C" and C.R.D. Cool Isol e CC-716B CC to R.C.P. "A", "B", "C" and C.R.D. Cool Isol l

i e CVC-381 RCP Seal Water Return SD-006 Rev. 16 Page 13 of 18

95-2 NRC EXAM - REACTOR OPERATOR i

I l 33. Given the following list of channel numbers and noun names:

1 R-1 Control Room R-9 Letdown Line Area i

R-2 CV Low Range Monitor R-11 CV Air / Stack Part.

, R-3 Health Physics Work Area R-12 CV Air / Stack Gas l R-4 Charging Pump Room R-18 Liquid Waste Disposal

! R-5 Spent Fuel Building R-32A CV High Range i R-6 Sampling Room R-32B CV High Range i R-7 CV Incore Inst. R-33 Monitor Buil ding Area j R-8 Drumming Station i i !

Which ONE (1) of the following describes ONLY those radiation monitor channel (s) that have i control functions? J l

l VA. R-1.

B. R-1 and R-5. .

, ~ ~ .

l C. R-1, R-2, R-l l, R-12. .

O .

D. R-1, R-2, R-11, R-18.

{ sggs ,p a ! ),p w

K/A 072000.Kl.04 3.3/3.5 I

, l

, SD-019, Rev.1, page 8. . RM Obj. 3. ;

! Modified Question <

l \5 CW5w l !

/ 1

[ CVhN l \

s i l Question 33 of 100 l

i

95-2 NRC EXAM - REACTOR OPERATOR

34. WD-09 001 Given the following plant conditions:

A waste gas release is in progress e R-14 alarms and shuts RCV-014, Waste Gas Decay Tank Release valve Which one (1) of the following describes the actions necessary to re-open RCV-014?

A. RCV-014 will re-open when the high radiation alarm clears. !

I B. RCV-014 will not re-open until the control switch is placed in close and then open.

W. RCV-014 will not re-open until the valve has been ran down to the cleared position.

D. RCV-014 will not re-open until its inlet valve is closed and reopened.

c K/A 071000.A4.13 3.0/3.1 WD pg 51,52 New question 1

l i

4 Question 34 of 100 l

J

r .

LESSON BODY KEY AIDS i*

i l*

l (3) Overpressure protection provided by 150 psig relief valve '

(4) Connected to plant nitrogen system for purging

)

l m. Gas decay tanks are sampled and analyzed prior to release e Verifies sufficient decay of radioactive nuclides to meet 10CFR20 mquirements e Provides a record of activity released 1

n. Gas release valves (PCV-1040) (RCV-014) !

l (1) Air Operated Valves l

l (2) Both must be open to allow release to the plant j stack

-l i

(3) Both located in GDT room l

(4) PCV-1040 '

e Used to maintain 5 psid across RCV-014 l o This ensures a constant release rate regardless of tank pressure ]

(5) RCV-014 OBJ.#9 i

e Throttle valve used to control the release from the Gas Decay Tanks l e Setpoint is controlled by Curve 4.2 in the Plant Curve Book l e Automatically closes on RMS-14C alarm WD Rev.1 Page 51 of 75 l /

t _.

, LESSON BODY KEY AIDS l.

o RCV-014 cannot be opened until alarm clears and valve run down to " cleared" position

o. Tank contents released by process monitor R-14C (1) Monitors release (2) High activity levels measured by R-14C will cause RCV-014 to shut

^

p. Gas Analyzer (1) During normal operation, oxygen and hydrogen concentration in each gas source is determined by ,

the gas analyzer operating in auto e Hydrogen concentration varies from tank to tank

No significant oxygen expected (2) Records results on Westronics 3000 color chart recorder (Each point is a different color) and warns ;

operator of hazardous conditions (Hi/Lo alarm for Hydrogen and Oxygen (3) Two cabinets, one samples and analyzes, the other l selects analysis point, records alarms l (4) Three modes of operation i 1

e Auto - sequentially samples, analyzes and records M each point, cycle time is approximately 5 minutes e Manual - only one point sampled at a time - alarms ,

inoperative ;

l e Calibrated - usee nitrogen 1

' o Nitrogen is used to zero the analyzer and as a

! reference gas

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WD Rev.l Page 52 of 75 i .

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. 'q l'

j 95-2 NRC EXAM - REACTOR OPERATOR

35. DC 14 003' [ )

l Which one (1) of the following desc es the battery charger that has the capacity to carry its

!- associated DC bus if its battery is isconnected from the bus?

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l VA. A-1 l

i B. A l'

I L 'C. B !

L D C i a 1 l.

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t i-K/A 063000.G0.10. 3.1/3.2 OP.601, P&L - !

l DC LP Rc/ 2 pg 12 .l l Modified Question l

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, Question 35 of 100 1.

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4.0 PRECAUTIONS AND LIMITATIONS l

l 1. One battery charger for each battery shall be in service so that the l

l batteries will always be at full charge in anticipation of a l

l Loss-of-AC incident.

1) Battery Chargers "A" and "B" gre considered the lead battery chargers and shall be in service when available.

2) The time Battery Chargers "A" and "B" are in the STANDBY mode shall be minimized to avoid instability resulting from their not being connected to their associated station battery.

2. The DC distribution system for station battery "A" and "B" shall be maintained operable.

3. The Battery Room Ventilation Fans "A" and/or "B" shall be in service. l 4 Caution should be exercised when working near the Batteries. Rubber gloves, apron and face shields should be worn when handling acid.

5. NO SMOKING, BURNING OR WELDING is permitted in the battery rooms.

6. Each time a circuit is energized, the applicable Battery Charger l

("A" or "A-1", "B" or "B-1", or "C") should be checked for ground indication.

7. All normal precautions pertaining to energized electrical gear shall be observed.

8. Inverter input voltage trip setpoints are 100 VDC and 140 VDC for Inverters "A" and "B" and 105 VDC and 140VDC for Inverter "C".

9. In the event that it becomes uscessary to disconnect a Station Battery from its DC Bus, only Chargers A-1 and B-1 have the l

capability to stand alone carrying its associated DC Bus.

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l OP-601 Rev 16 Page 5 of 51

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, 95-2 NRC EXAM - REACTOR OPERATOR l

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! 36. NI-10 002 l Which one (1) of the following describes the protection signal that is taken credit for by the Updated Final Safety Analysis Report (UFSAR) to protect the reactor core during a reactor

- startup accident?

A. Source Range liigh Flux Trip.

B. Intermediate Range liigh Flux Trip. .

<. Power Range liigh Flux, Low Setpoint Trip.

I D. Power Range liigh Flux, liigh Setpoint Trip. ,

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K/A 015000.K4.05 4.3/4.5 ;

FSAR Section' 7.2 pg 7.2.17 New Question i qs1 I cg)$

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l Question 36 of 100 l -- - - . - --

. _ . . . - - - . - - - . - -. . . - - . - . - . . - . ... - ~ . . - . - -

HBR 8 UPDATED FSAR 1 .

J

! (Appendix A to the operating license) . The trip settings remain more j restrictive than the core safety limits, and are used in the protection system

to provide suitable margin for measurement and instrument errors.

J Adequate margins exist between the ==w4=um nominal steady state operating point (which includes allowance for temperature, calorimetric, and pressure errors) and required trip points to preclude a spurious plant trip dttring design transients.

1

' Completion of Protective Action 7.2.1.1.1 J Where operating requirements necessitate automatic or manual bypass of a -

j protective function, the design is such that the bypass is removed j automatically whenever permissive conditions are not met. Devices used to j achieve automatic removal of the bypass of a protective function are part of the protective system and are designed in accordance with the criteria of this section.

4 i 7.2.1.1.2 Reactor Trips i

j The reactor trips are as follows:

j a) Hinh Nuclear Flux (Source Ranae) Trin - This circuit trips the reactor j when one of the two source range channels reads above the trip setpoint. His 1 j trip, which provides protection during reactor startup, can be manually j bypassed when one of two intermediate range channals reads above the P6

, setpoint value and is autcoatically reinstated when both intermediate range channels decrease below this value (P6). His trip is also bypassed by two

out of four high power range signals (P10). The trip can also be reinstated

} below P10 by an administrative action requiring coincident manual actuation.

! The trip point is set between the source range cutoff power level and the l' maximum source range power level.

b) Mich Nuclear Flux (Intermediate Rance) Trio - his circuit trips the ,.

j reactor when one out of the two intermediate range channals reads above the j trip setpoint. This trip, which provides protection during reactor startup, j can be manually bypassed if two out of four power range channels are above approximately 10 percent (P10) : Three out of four channels below this value-automatically reinstate the trip. The intermediate channels (including j detectors) are separate frem the power range channels.

J c) Hiah Nuelaar Fltar (Power n n,.= ) Trin - This circuit trips the reactor j when two of the four power range channels read anove the trip setpoint. There j are two independent trip settings, a high and a low setting. The high trip i setting provides protection during normal power operation. p setting,

! which provijppa ro*=e*4an during startup, can be manually bypassed when two cut of the four power range channels read above approximately 10 percent power (P10). Three out of the four channels below 10 percent automatically reinstate the trip. The high setting is always active.

7.2.1-2 Amendment No. '10 l

.. _ .. . . . . .. - . - . - . _ ~ . - . . . . . . - . = .-. - - - - . - . . . - _ . . . - --

1

!

HBR 2 UPDATED FSAR l d) Overtemperature AT Triy - The purpose of this trip is to protect the

' ' core against DNS. This circuit trips the reactor on coincidence of two out of the three signals, with one_ set of temperature measurements per loop. The setpoint for this reactor. trip is continuously calculated for each loop by solving the following equationt (l'+'t S AT -K g ) (T-T') + K L

setpoint - < ATO (K1 2 '(1 + t 28 3 (P - P') - f(AI)) 3 i

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1 7.2.1-2a Amendment No. 3 I

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~ .- -. -- ~ - . - . - -- . . . - . . - . ~ . - .. -.. . - .- ..-. - ... -.. - - ~. .. . . ~ -

l-l 95-2 NRC EXAM - REACTOR OPERATOR l

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44. Given the following plant conditions.

1 l

s The plant is at 100% power -l l

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A loss of instrument Bus 3 has occurred Which ONE (1) of the following describes the expected response to the above conditions?

)

A. Turbine runback from rod bottom bistable energizing. 1

4. Turbine runback from NIS rod drop signal. !

i C. Condenser Steam Dump will shift to pressure mode.

' l i

D. Diesel Generator "B" will auto start on undervoltage.

b K/A 062000.A2.04 3.1/3.4 ,

'l AOP-024, Attachment 1 New Question l

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Question 44 of 100

\

95-2 NRC EXAM - REACTOR OPERATOR 4

47. TS-3.3.1.2F 001 l Given the following plant conditions:

The plant ha; been at 100% power for 30 days.

All systems are in aur.omatic and operating as expected.

In assuming the morning shift at 0700 AM, the turnover included that the "A" accumulator isolation valve is energized to accommodate maintenance and testing.

I hour into the shift, the RO reports a review of the logs indicates the accumulator isolation valve was initially energized at 0400 AM and has been energized continuous;y since.

Which ONE (1) of the following describe compliance with Technical Specifications (TS) concerning the report?

A. The TS limit was violated at 0600 AM.

W. The TS limit has just now been reached.

C. The TS limit will not be affected for another 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D. The TS limit will riot be affected for another 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

b K/A 006000G005 3.5/4.2 TS 3.3.1.2(f), page 3.34.

Modified Question

)

Question 47 of 100

.. . - -- __ - . =. . _. . ._

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1. Power cperation with 1 css than three loops in s:rvice is prohibited.

3.3.1.2 During power operation, the requirements of 3.3.1.1 may be modified to allow any one of the following components to be j inoperable. If the system is not restored to meet the l requirements of 3.3.1.1 within the time period specified, the I reactor shall be placed in the hot shutdown condition utilizing l normal operating procedures. If the requirements of 3.3.1.1 are .

i not satisfied within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed in the cold shutdown condition utilizing normal operating procedures.

a. One accumulator may be isolated or otherwise inoperable relative to the requirements of 3.3.1.1.b for a period not to exceed four hours.

l b. If one safety injection pump becomes inoperable during normal reactor operation, the reactor may remain in l

operation for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l c. If one residual heat removal pump becomes inoperable during normal reactor operation, the reactor may remain in operation for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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l 3.3-3 Amendment No. 97, 446, 449, 446, 153 h

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d. If cne residual h at cxchang:r becomes in:perable during normal reactor operation, the reactor may remain in operation for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

! e. If any one flow path including valves of the safety l injection or residual heat removal system is found to be

! inoperable during normal reactor operation, the reactor may remain in operation for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l The hot leg injection paths of the Safety Injection System, .

l including valves, are not subject to the requirements of this specification.

f. Power or air supply may be restored to any valve referenced l in 3.3.1.1.g. and 3.3.1.1.h. for the purpose of valve testing or maintenance providing ao more than one valve has power restored and provided that testing and maintenance is completed and power removed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months except for accumulator isolation valves (MOV 865 A,B,&C) which will have this time period limited to four hours.

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3.3-4 Amendment No. 97, 44+, 446, 153

95-2 NRC EXAM - REACTOR OPERATOR i i

I

54. Given the following plant conditions: j

Reactor Power is 98% and stable l

All systems are in automatic and operating normally

Pressurizer pressure transmitter PT-444 fails to 2370 psig Which ONE (1) of the following describes the system response to this event?

l A. All spray valves open, all pressurizer heaters energize. I B. All spray valves close, all pressurizer heaters energize.

W. Power operated relief valve PCV-455C opens, all spray valves open, all pressurizer heaters deenergize.

i D. Power operated relief valve PCV-456 open, all spray valves open, all pressurizer heaters deenergize.

c K/A 010000.A1.07 3.7/3.7 SD-059, page 16. PZR LP Modified Question i i

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Question 54 of 100 i

95-2 NRC EXAh! - REACTOR OPERATOR

55. Given the following plant conditions:

The unit is at 100% power l

l

Steam Flow channel input to S/G Water Level Control has just failed high

No Operator action is taken Which ONE (1) of the following describes the plant response to this event?

The affected S/G level will.....

A. decrease until a LO-LO Level trip occurs.

4. increase until a FW isolation and a turbine trip occurs.

C. increase until the flow error signal off-sets the level error signal.

D. decrease until the level error signal off-sets the flow error signal.

b K/A 035000.A2.04 3.6/3.8 Imgic Dwg 5379-2758 Sht 1 of 18 New Question l

Question 55 of 100 i

95-2 NRC EXAM - REACTOR OPERATOR

56. Given the following plant conditions:

A LBLOCA has occurred e The operators are transitioning to cold leg recirculation

RilR-860A had cold water in the bonnet and has not heated up causing the valve disc to seat tightly Which ONE (1) of the following describes this thermodynamic problem associated with RHR-860A7 A. Stem Thermal Binding.

B. Disc Thermal Binding.

vC. Liquid Entrapment.

D. Bonnet Depressurization.

c K/A 006000.A3.03 4.1/4.1 SO R4-7 pg 5,6 Modified testion

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lh o g\

Sg' l

l Question 56 of 100 l

95-2 NRC EXAM - REACTOR OPERATOR

58. Given the following plant conditions:

A refueling outage is in progress

. The manipulator is being used to move fuel in the CV i Which ONE (1) of the following describes the " Upper Slow Zone" manipulator Crane ;

interlock? ,

VA. operable only in the core area with the gripper engaged. l B. operable only in the core area with the gripper disengaged.

C. operable only in the fuel transfer area with the gripper engaged. :

D. operable only in the fuel transfer area with the gripper disengaged. !

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i K/A 034000.K4.03 2.6/3.3 SD-8, pg 25

' FHP-001 New Question f

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t Question 58 of 100

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$5-2 NRC EXAM - REACTOR OPERATOR :

f l 59. SW-10 001

. Given the following plant conditions:

l A reactor trip / turbine trip has occurred.

l North Service Water IIcader pressure is 25 psig. i South Service Water 11eader pressure is 35 psig.

1 Service Water licader pressures have been at these values for 2 minutes.

The following valve noun names apply: l V6-16A North lidr Supply Turbine Bldg Cooling Water. !

V6-168 South lidr Supply Turbine Bldg Cooling Water. ,

V6-16C Turbine Bldg Cooling Water Isolation. !

Which ONE (1) of the following describes the response of the Service Water System to these conditions. !

f l

l A. Only valve V6-16A will close.

B. Only valve V6-16C will close.

<. Valves V6-16A and V6-16C will close.

D. Valves V6-16A, V6-168, and V6-16C will close.

c v' C l K/A 076000.Kl.16 2.7/3.1 SD-004, Rev. 22, page 10. !

SW LP, Modified Question l ,

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! Question 59 of 100

g 4.0 (Continued) 4

11. Coordinated Chlorination (coordinating the periodic operation of j ,4 an isolated. portion of the SW system with routine total system

$sl chlorination) should be performed on the following components when normal flow is stopped or the component and associated piping will remain in a wet, unchlorinated " stand by" condition for greater than 7 days (contact Engineering for assistance if 1

required.):

s A & B AFW Motor Driven Pump Lube 011 Coolers and associated P iPi ng

A & B Emergency Diesel Generator Heat Exchangers and associated piping

, i

12. For HVH-1, HVH-2, HVH-3, and HVH-4 to be OPERABLE when RCS !

i temperature is greater than or equal to 200*F, a flow rate of at )

least 800 gpm is required through each HVH unit, fan and motor coolers combined, as indicated on FI-1698A, FI-1698B, FI-1698C, and FI-1698D. )

13. Valves V6-16A, V6-16B, and V6-16C have automatic closure features to isolate the Turbine Building.

Valve V6-16A will close if PSL-1616A reaches 31 psig decreasing for 60 seconds with any Turbine Trip signal present.

Valve V6-16B will close if PSL-1684A reaches 31 psig decreasing for 60 seconds with any Turbine Trip signal .,

present.

Valve V6-16C will close if PSL-1616B or PSL-1684B reaches 31 psig decreasing for 60 seconds with any Turbine Trip signal present.

Keylock switches are provided to inhibit the auto-closure feature for V6-16A, V6-16B, or V6-16C for maintenance, testing, or when the unit is in cold shutdown.

. 1 OP-903 Rev. 56 Page 11 of 123

Y 4.0 (Continued) l

$ EDIE l This information is provided locally by Instructional Aid 96-OP-15.

i 14'. Throttling of SW-739 and SW-740 (CCW HEAT EXCHANGER OUTLET VALVES) shall be limited as follows:

Valve 1 ccW Hx 2 CCW Hx l 29L1 POS 1 .

l SW-739 10 1/2 turns '

9 1/4 turns SW-740 11 1/2 turns 8 3/4 turns This is to prevent exceeding the capability of 2 SW pumps during a 2 SW pump accident response. These valve positions will -!

permit approximately 5000 gym flow with 2 heat exchangers in operation (POS 1) and 10,000 gpa flow with I heat exchanger in )

operation (POS 2). (Reference SP-895, Rev. 1 and SP-1010).

I

15. The principles of AIARA shall be used in planning and performing I 1

work and operations in the Radiation Control Area. '

16. Venting the Service Water system or components will be done at the direction of the Superintendent Shift Operations. Venting should be considered any time flow is secured or a section is drained.

17. Isolation of Service Water flow through HVH-6A, HVH-6B, HVH-7A, HVH-7B, HVH-8A, or HVH-8B will make the fan inoperable and make the pump with the same train electrical supply inoperable. As an example, isolation of Service Water flow through HVH-8B makes RHR Pump "B" inoperable 98 isolation of Service Water flow through HVH-7B makes MDAFW Pump "A" inoperable.

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OP-903 Rev. 56 Page 12 of 123

., 95-2 NRC EXAM - REACTOR OPERATOR

60. CSS-09 004 Given the following plant conditions:

A LOCA inside containment has occurred Which ONE (1) of the following describes the action to be taken by the operator if hydrogen concentration exceeds 4% while the Post Accident Hydroben Recombiner (PAHR) is in operation? ,

l A. Stop the PAHR as an explosion may occur.

B. Throttle the inlet valve to reduce the flow rate and ensure complete combustion.

W. Begin nitrogen flow to dilute the hydrogen concentration.

D. Reduce power to the electric heaters to reduce the reaction chamber temperature.

c K/A 028000.Al.01 l OP-922, Sect 5.1.2 i SD-048, pg 4-6 New Question 1

Question 60 of 100

1 l

Sacticn 5.1-Page 7 of 9 1

5.1.2 (Continued) JNJT l

FQT3 The H2 Recombiner Heater will be energized, the Containment Gas Blower will start, and the Heat Exchanger Fan will start when the START pushbutton is depressed.

Containment gas will begin circulating from Containment through the H 2 Recombiner and returned to Containment.

11. Press and hold the START pushbutton for 10 seconds.

neu Startup time req'uires approximately one and one half hours. After the temperature of the gas in the reaction chamber' reaches 1350*F, the H 2 Recombiner System operates automatically, with the heater power gradually changing to maintain the same preset reaction chamber temperature.

12. Check H 2 Recombiner Flow rate is greater than or equal to 5.0 inches H 20,

13. Containment H 2 Concentration greater than 4%. YES / NO (Circle one)

OP-922 Rev. 8 Page 15 of 6'

, Szction~5.1 ,

Pegs 8 of 9 5.1.2 (Continued) IEII -

l l i L , 14. II Hydrogen Concentration is in excess of'44. THEN perform j l '

the following: *

1) Calculate the approximate Nitrogen dilution required ;

using the-following formulas:

l a. SCFM - 177.4 .23 _3_gL 1/2 - l Tt l

SCFM is the flow rate at the inlet nozzle

Pi ,is absolute pressure at the flow meter throat i in paia -(determined from installed instrumentation)-

e M4 is read from FI-l in inches

. .

Tg is flow meter throat temperature

b. SCFM - SCPM x 3.04 - Na flow rate @ STP I tH 2 l e tH 2is the decimal equivalent of percent H 2 e

Na flow rate @ STP is in SCFM N2 dilution flow required - SCFM @ STP )

. i

'l EQIE Hydrogen recombination begins at 1100'F. By placing Nitrogen dilution in service at 900*F, Hydrogen combustion is prevented when temperature reaches 1100*F,

2) HHEN TIC-4 reaches 900*F IEEE throttle open NS-26 Nitrogen Supply Valve, until FI-1077 indicates required ;

Nitrogen dilution flow.

(Located on West Wall across from MCC-18) i l

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OP-922 Rev. 8 Page 16 of 63 l---n v ,- - - - - ,.ne, - - ,,w.. --, -- , ,g--

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, S3cticn 5.1 Pega 9 of 9 5.1.2 (Continued) E l

) . 15. Monitor H2 Recombiner heat-up at TIC-4 until the temperature reaches 1350*F and stabilizes.

! EQIE j When TIC-4 stabilizes,,the Hydrogen concentration of the H2 Recombiner influent should be checked periodically to determine the need for Nitrogen dilution.

Quicker and more efficient recombination will occur if Nitrogen is periodically 1

reduced in p: oportion to Hydrogen concentration reduction over the length of time l the H2 Recombiner is operating.

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l 16: -Complete Attachment 9.2 every 30 minutes.

} Initials Name (Print) h

) Performed By:

j 3

l Approved By:

5 Shifi~ Supervisor Date i

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1 OP-922 Rev. 6 Page 17 of 63

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1.0 GENERAL DESCRIPTION 1.1 System Purpose 4

Following a loss of coolant accident (LOCA), hydrogen gas may be j

' generated inside the containment by reactions such as radiolysis of j aqueous solutions in the sump and core, zirconium metal with wate.r, j and corrosion of materials of construction. The post accident

{ hydrogen recombiner system provides the capability to limit the

[ buildup of hydrogen inside containment to a level less 'than 3.5% by

) volume.

, 1.2 System Description 5

j The post accident hydrogen recombiner consists of two portable,

l. skid-mounted packages, the recombiner and the power / control cabinet.

{ The recombiner will be shared with the Duke Power Company and stored f at the Oconee Nuclear Power station. It will be transported to H.B. j Robinson, Unit 2, when needed. Permanently installed piping, valves and power supplies will provide for rapid hookup of the hydrogen recombiner. The recombiner piping ties into the post accident l containment venting system (PACVS) downstream of the containment isolation valves and utilizes the PACVS containment penetration piping.as the supply and return points for the containment gases. A lay down area in the fuel handling building has been provided for the recombiner. The supply line to the hydrogen recombiner system

, is tied to the PACVS containment air exhaust "A" between valves V12-15 and V12-16. Isolation valve V12-61 is provided in the supply

line and is located on the second floor (elevation 246'-0") of the -

Auxiliary Building just to the left of the R-11 and R-12 cabinet, which is an accessible area after a major accident.

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. I SD-048 Rev. O Page 4 of 12

j

! l l 1.0 GENERAL DESCRIPTION (Continued) 1J All supply' piping is insulated to minimize condensation should the i

i air inside containment be saturated. A low point manual drain l l valve, V12-65, for the supply piping is located in the south l

[ domineraliser hallway at elevation 226'-0" of the Auxiliary l Building. Depending on containment air temperature at time of recombiner operation, the condensation may need to be drained every

] 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the recombiner is run continuously. Reduced flow of supply gas to the recombiner, as indicated by the control cabinet j flow rate indicator, would indicate possible excess condensation in

{ the supply line.

The gas return from the recombiner is through the PACVS air exhaust "B", between valves V12-19 and V12-20, with isolation valve V12-63 near V12-61 on elevation 246'-0" of the Auxiliary Building. There is a low point manual drain valve, V12-66, for the return line.

V12-66 is located in Pipe Alley, Plant nitrogen is provided for the supply line to the recombiner.

The nitrogen will be used to' purge the radioactive gases from~the recombiner and piping to allow maintenance accessibility. Also, ;

nitrogen will be used to dilute the incoming gases to the recombiner if a concentration of hydrogen greater than 4% is present in the

. C.V. A flowmeter and throttling valve are installed in the nitrogen supply line to provide manually controlled dilution adjustments.

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I SD-048 Rev. O Page 5 of 12 !

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1.0 GENERAL DESCRIPTION (Continued)

All piping and valves in the hydrogen recombiner supply piping are designed for seismic loads, since the areas in which the pipe is i

routed will not be radiologically accessible for repairs during post

accident conditions.

The hydrogen recombiner will be located behind a solid concrete block shield wall. The shield wall will be constructed prior to

recombiner operations and will be used to protect plant personnel 4

who must operate the.,recombiner.

i

} The hydrogen recombiner skid consists of_an enclosed gas blower, a

heater section, a reaction chamber, an airblast gas cooler, and a j centrifugal fan. Inlet and outlet piping is connected through

! flexible hoses to the containment gas supply and return piping. The ,

j power and control cabinet consists of components integrated into a j system which activates, controls, and indicates the operation of the recombiner. Power for the recombiner is supplied from MCC No. 1.

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{ A ventilation system is provided to exhaust hot air from the-j recombiner gas cooling section to the fuel handling building HVAC d

[.

system. The ductwork is permanently installed except at the j

5 recombiner interface where flexible duct is utilized.

I

All valves installed for the recombiner piping will be aligned in t the normally closed position. In the event of a design basis

! accident, proper valve alignment will be made depending on the PACVS service condition.

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SD-048 Rev. O Page 6 of,12 4

95-2 NRC EXAM - REACTOR OPERATOR 1

i

61. RHR-10 002 Given the following plant conditions:

The plant is shutdown and RilR is in service j

A complete loss of instrument air has occurred l

l Which one (1) of the following describes the response of the RHR System?

1 VA. RHR HCV-758, RHR loop temperature control valve, fails CLOSED.

B. RHR FCV-605, RHR loop flow control valve fails OPEN.

C. RHR-744A and B, RHR discharge to RCS' isolation valves, fail OPEN. l l

D. SI-860A and B, RHR pump suction from containment sump valves fail CLOSED. l a

K/A 005000. A2.04 2.9/2.9 AOP-017, pg 37 ;

Modified Question i

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Question 61 of 100

i R: v. 17 i -'- AOP-017 LOSS OF INSTRUMENT AIR Page 37 of 58 i,

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INFORMATION USE :

ATTACHMENT 1 MAJOR COMPONENTS AFFECTED BY LOSS OF IA.

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(Page 3 of 5) l l ,

8. Primary Sample System Components FAIL POSITION i

, a. PS-956 A through H, PRIMARY SAMPLE IS01ATIONS CLOSED T

9. Radiation Monitoring System Components

FAIL POSITION 3'

a. RMS-1,2,3 & 4, R-11/R-12 ISOL VALVES CLOSED 1

[ 10. Reacto'r Coolant System Components FAIL POSITION 1

$ a. PCV-455 A & B, PZR SPRAYS CIDSED

~

b. RC-544, RV FIANCE LEAROFF OPEN

j. c. RC-516 & 553, PRT TO GAS ANALYZER CIASED 1 d. RC-519 A & B, PW TO CV IS0s CLOSED l -

l e. RC-550, PRT NITROGEN SUPPLY CIASED J

.11. Residual Heat Removal System Components FAIL POSITION

a. FCV-605, RHR HX BYPASS FIDW CLOSED I

b. HCV-142, PURIFICATION FLOW CLOSED

c. HCV-758, RHR HK DISCH FLOW CIASED

12. Safety Injection System Components FAIL POSITION

a. SI-855, ACC NITROGEN ISO CLOSED

b. SI-856 A & B, SI PUMP RECIRCS OPEN

c. SI-850 A, B & C, SI ACCUhTIATOR TESTS CIASED

d. SI-850 D, E & F, COLD LEG INJ TESTS CIASED

e. SI-851 A, B & C, SI ACCUMULATOR MAKEUPS CLOSED i

f. SI-852 A, B & C, SI ACCUMUIATOR DRAINS CLOSED

g. SI-853 A, B & C, SI ACCUMUIATOR VENTS CLOSED

I 95-2 NRC EXAM - REACTOR OPERATOR l

l 62. Given the following plant conditions:

l l

e The unit is in Cold Shutdown preparing for heatup l l

One CCW pump is operable

One CCW heat exchanger is operable

Which ONE (1) of the following describes the number of ADDITIONAL pumps and heat j exchangers required to go above 200 F? l 1

A. ONE pump, No additional heat exchangers.

I B. TWO pumps, ONE heat exchanger. l C. No additional pumps, ONE heat exchanger.

i

@. ONE pump, ONE heat exchanger.

d K/A 008010. A3.03 2.9/3.2 GP-003, pg 24 l New Question I

I i

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i I

Question 62 of 100

_ . . . _ . - . - . _ _ ~ . _ . . _ . . . . . _ . _ . _ . _ . _ _ _ - - _ _ _ _ - _ _ . . _ . . - . . _ _

i

. 95-2 NRC EXAM - REACTOR OPERATOR 4 .

e

?

63. SD-09 001 Given the following plant conditions: '

The steam dumps are in the Tavg Mode of control ;

. i j PT-446. Turbine first stage pressure, has just failed HIGH No operator actions have been taken Which one (1) of the following describes how this failure affects the Steam Dump System?

A. The steam dumps will not arm.

B. The system can be armed, but will not open on a turbine trip. !

4. The system can be armed, but will not open on a load rejection.

D. The steam dumps will arm and remain armed until manually reset.

c ,

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K/A 041020.A4.08 3.0/3.1 Steam Dump, Pg Modified Question l

Question 63 of 100

l i ,

i LESSON BODY KEY AIDS (2) When primary power > secondary power (ie, Tavg

> Tref), steam dumps open to increase secondary power until rods can lower primary power to match secondary power.

b. Steam pressure mode GBJ.#5 (1) Steam pressure used to control steam dump position.

(2) Controlling steam pressure when shutdown controls RCS temperature.

I. COMPONENT DESCRIPTION A. POWER OPERATED RELIEF VALVES

1. PURPOSE OBJ. #4

a. Provide means for plant cooldown by steam discharge to atmosphere if condenser steam dumps not available.

2. DESIGN OBJ. #5

a. 8 inch, air operated globe vrJve

b. Backup nitrogen supply (1) Instrument air not available (2) N2 can be aligned per AOP-017, Loss ofIA or using Dedicated Shutdown Procedures (D.C.Ps)

c. Flow = 580,000 lbm/hr each @ 790 psi

d. Total flow = 1,740,000 lbm/hr @ 790 psi SD Rev.1 Page 9 of 24 1

, LESSON BODY KEY AIDS :

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4

3. ANNUNCIATORS OEJ. #9 !

SD-Figure-9 I 1

a. APP-021-B3, PORY RV-1-1 STEAM DUMP DEFEAT j l

(1) Steam dump signal to PORV RV-1-1 defeated l I

(2) Also defeats PORV operation by respective RTGB l controller -

l (3) NORMAIJDEFEAT switch on secondary control panelin DEFEAT position '

l

b. APP-021-B4, PORY RV-1-2 STEAM DUMP DEFEAT 1

(1) Steam dump signal to PORV RV-1-2 defeated j (2) Also defeats PORV operation by respective RTGB i controller (3) NORMAUDEFEAT switch on secondary control panelin DEFEAT position

c. APP-021-C4, FORV RV-1-3 STEAM DUMP DEFEAT ;

(1) Steam dump signal to PORV RV-1-2 defeated (2) Also defeats PORV operation by respective RTGB q controller (3) NORMAUDEFEAT switch on secondary cor. trol !

panelin DEFEAT position B. STEAM DUMP VALVES l

1. PURPOSE OBJ. #4 I i

1

a. Provide means for plant cooldown by steam discharge to the main condenser i

l 1

SD Rev.1 Page 10 of 24 i

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LESSON BODY KEY AIDS b Normal method of controlling temperature or power level during startup and low power conditions 1

2. DESIGN OBJ. #5 t

a. 8 inch, nitrogen operated globe valve SD-Figure-3

! b. Instmment air backup supply l

l l

II. SYSTEM INDICATIONS !

A. INSTRUMENTATION OBJ. #9

l. l

1. ANNUNCIATORS

a. APP-006-F5, STEAM DUMP ARMED SD-Figure-8 I

(1) Normally indicates nitrogen supplied to steam )

dump valves !

(2) Can be due to Tavg - Tref difference, system SD-Figure-2 would not actually be armed III. CONTROL FUNCTIONS AND INTERLOCKS OBJ. #9 A. POWER OPERATED RELIEF VALVES

1. NORMALLY OPERATED BY RESPECTIVE CONTROLLERS

a. Normal setpoint ~30 psig above steam pressure with Tavg of 547 F (1005 psig) -

b. Cabinet on mezzanine level I

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SD Rev.1 Page 11 of 24

f LESSON BODY KEY AIDS

c. "A" PORV, RVl-1 controller PIC-477

d. "B" PORV, RV-1-2 controller PIC-487

e. "C" PORV, RV-1-3 controller PIC-497

f. Local manual using thumbwheel to adjust pressure As directed by EOPs 4- and DSPs

{ (1) Increasing pressure opens valve d

} (2) Decreasing pressure closes valve

g. Automatic control using pots on RTGB 4

h. Controller displays setpoint t

. (1) When actual pressure increases to setpoint PORV

throttles open to relieve pressure.

l 2. In event ofload rejection > 70%, steam dump controls take i over control P

B. TRANSMITTERS O.BJ.#9 i

. ., 1. MAIN STEAM PRESSURE i a. Four pressure transmitters (PT-1310,464,466 and 468) .

located on 72 inch header.

(1) PT-1310 indication on RTGB only (2) PT-464,466 and 468 provide safeguards high steam line differential actuation (3) PT-464 also used for Steam Dump control Il l

SD Rev.1 Page 12 of 24

LESSON BODY KEY AIDS

2. TURBINE IST STAGE PRESSURE TRANSMITTERS SD-Figure-6

a. Proportional to reactor power

b. PT-446 - reference temperature signal to Steam Dumps OBJ. #7

c. PT-447 - sudden loss ofload bistable to Steam Dumps OBJ. #7 C. CONTROL SWITCHES

1. STEAM DUMP MODE SWITCH SD-Figure-4 Show on mock-up

a. Three position (1) RESET, spring return to T-AVG OBJ. #7, 8

Resets logic controlling N 2supply to valves, i removes arming signal following loss ofload or turbine trip.

(2) T-AVG OBJ. #7, 8

Places Steam Dumps in automatic under control of !

Tavg j l

(3) STEAM PRESS OBJ. #7, 8

Places Steam Dumps under control of steam header pressure with adjustable setpoint.

o 0-1400 psig range o 10 turn pot = 140 psig/ turn i

o Setpoint = desired pressure /140 I

SD Rev.1 Page 13 of 24

, _ .. - =. . ..-. . _ - ..

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l LESSON BODY KEY AIDS l 2. STEAM DUMP CONTROL SWITCH Show on mock-up

a. Three position (1) OFF (Reset Tavg Bypass) OBJ. #7, 8

Resets Steam Dumps placing Tavg interlock in service

!

Leaving switch in this position disables Steam Dumps except PORVs.

(2) ON (3) BYPASS T-AVG INTLK, spring return to ON OBJ. #7, 8

Bypasses low Tavg interlock (2/3 channels < 543 "F). Allows opening Bank 1 Steam Dump valves for j controlled cooldown.

l D. BISTABLES I

1. SUDDEN LOSS OF LOAD BISTABLES SD-Figure-2 l l

a. Input from PT-447, Turbine 1st stage impulse pressure j

b.

Purpose:

Provide anning signal for Steam Dumps OBJ.#7 )

I (1) Opens solenoid valve in nitrogen supply SD-Figure-3 ;

(2) For PORVs, shifts control from normal to steam dump controller, if turbine not tripped (3) Does not cause valves to open, only arm

Valves open in response to controller or trip-open bistables (see below)

c. Setpoints (1) PM-447A arms 3 dumps at 15% of fullload 1

i SD Rev.1 Page 14 of 24

'~

l l l :

LESSON BODY KEY AIDS i l :

l (2) PM-447B arms 2 dumps at 35% of fullload l i

l (3) Condenser dumps can only be armed if condenser

! available 1

l l e At least one cire water pump nmning i

e Sufficient vacuum o >l9.7 " Hg Vac i

o PSHH-1338, PSHH-1339 (A & B) l 1

(4) PM447D arms PORVs at 70% offullload e Turbine not tripped

2. TPJP OPEN BISTABLES SD-Figure-5 l

a.

Purpose:

Trips open Steam Dumps if: OBJ. #7 (1) Steam dumps armed (2) Load rejection large enough

b. Uses Median high Tavg I

c. Three-way solenoid valve bypasses positioner, applying SD-Figure-2 N 2direct to Steam Dumps and PORVs.

d. Turbine not tripped (ie. Load Rejection)

(1) Tavg - Tref = 12.1 'F, Bank 1, three valves j (2) Tavg - Tref = 16.6 *F, Bank 2, two valves l l

l (3) Tavg - Tref = 32.5 *F, Bank 4, three PORVs l

, 1 I

f I

SD Rev.1 Page 15 of 24

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5 !

95-2 NRC EXAM - REACTOR OPERATOR 1

T

64. OP-201-06 001.

Given the following plant conditions:

4 The plant is in refueling i

No core alterations are in progress l Which one (.3) of the following describes the conditions under which the in service RHR pump l can continue to be operated without component cooling water to the seal coolers per the RHR operating procedures?

i Pump may be operated..... I i'

t A. for only one (1) hour.

f B. indefinitely with flow through ONi Y the heatup recirculation line.

i C. until RCS temperature exceeds 200 F.

v'D . until pump discharge temperature exceeds 135 F.

i d

i-L. ,

K/A 005000G010 (3.3/3.5)

OP-201, Pg. 7, Precaution 9.

! Modified Question !

M

)

1, 1-i l

l Question 64 of 100 T- r -

-.- .- .. - - . ~ . . - - . _ -....=. _ . _,. _ - = - - .. ~ - .. . -...-. ~

S*

4.0 pgCAUTIONS AND LIMITATIONS i, -

i .

Reactor Coolant System 'cemperature and pressure shall be less than j' .,

r 1.

j 330*F and 375 psig before the Residual Heat Removal System is put in service, and the RHR system will be removed from service before RCS~

l pressure aad temperature are raised above these values.

l 2. To prevent boiling the CCW. liquid contained in an RHR HK, CCW flow ,

j abould not be isolated to.an RHR HX when the temperature of the RHR l-System is greater than 200*F. (CR 95-00565) f 3. Neither RHR-750 nor RHR-751 will open unless the following

) conditions are satisfied:

l e .The breakers for SI-862A and B are closed.

2

'* The breakers for SI-863A and B are closed.

I e The control power switches for SI-862A and B are in NORMAL.

i- *

The control. power' switches for SI-863A and B are in NORMAL.

f-

Valves. SI-862A and B 'are closed.

e Valves SI-863A and B are closed.

i

j. e RCS pressure is less than 465 psig.

4. SI-862A & B, and SI-863A & B are interlocked so they cannot be opened unless the RHR loop pressure is less than 210 psig.

i l'

5. When tihe Residual Heat Removal System is providing Core Cooling &HD l

I seal injection flow is desired to maintain a positive AP across the

Therme.1 Barrier of the Reactor Coolant Pumps, letdown flow through :

i HCV-142 and PCV-145 should be maintained to provide makeup to the t

i VCT.

4

6. When running RHR Pwups with SI-853A and/or SI-863B open, RHR-744A and RHR 744B should be cloisd to prevent excessive RHR pump runout.

OP-201 Rev'. 28 Page 7 of 59

?S

9

!. ' f,' 4.0* '(Continu d):

j

i. . .

)

l-

7. When RHR-757C or 2HR-757D is closed, 3,350 gym flow, indicated on .,

! FI i605, with one RHF pump running or 6,700 gpm flow with two RHR l J

l pumps running shall not be exceeded, except as allowed / required by

.F Approved. test proce'dures for which flowrates on FI-605 may be as high as 3800 gpm for one pump or.7600 gpa for two pumps. l

. 1

i. . .

j

8. When both RHR-757C and RHR-757D.are open, 3750 gpm total per running l

[ pump as read from FI-605, FI-608A and FI-608B shall not be exceeded, L s except as' allowed / required by approved test procedures for which l total flowrates may be as high as 4200 gpm for one pump.or 8400 gpm for two pumps. l l

l' .

I

! 9. If CCW is not available to the ped pump seal coolers, the RHR pumps i' i j shall not be operated with pump discharge temperature greater than ;

i 135'F. With CCW available to the RHR pump seal coolers there is no time limit for running a single pump with flow only through the heatup recirculation line. .It will be necessary to rotate the RHR pumps to avoid-exceeding the 50*F AT limit between RHR loops as stated in GP-007. l l

i

10. RHR pump flowrates of less than 2,800 gpm have been shown to !

increase pressure and flow fluctuations and should be avnidad when j plant conditions permit. This does not apply during recirculation ]

operation. I (ACR 91-078)

11. The principles o'f AIARA shall be used in planning and performing i work and operations in the Radiation Control Area. {

12. With no flow in the RHR system,.an RHR Pump should not be started with FCV.605 in automatic. This could allow runout of the pump before FCV-605 could respond to control flow.

13. This procedure has been screened IAU PLP-037 criceria and determined not applicable to'PLP-037.

I OP-201 Rev.:28 Page 8 of 59

l 95-2 NRC EXAM - REACTOR OPERATOR i.

S

65. Given the following plant conditions: 1

e You have entered FRP-C.1, " Response to Inadequate Core Cooling"

Core Exit thermocouples are > 1200"F and increasing i

! + You are directed by procedure to start RCP's I i

-Which ONE (1) of the following describes RCP operation once they have been started?

l

, The RCP's should be ......

i l' !

vA. secured when at least 2 hot leg temperatures are < 350 F.

B. secured when at least 2 cold leg temperatures are < 350 F.

i j C. left running because securing them at this point has not been analyzed.

i

! D. left running so as not to risk losing power due to starting and stopping the pumps in this 5

condition. !

! l 3 a i

?

K/A 000074.EK3.07 4.0/4.4 ,

j FRP-C.1 l i New Question i

b= cold legs not used j c=has been analyzed I d=no damage should be incurred FRP-C.1 Admin section of WOG l l

Question 65 of 100

95-2 NRC EXAM - REACTOR OPERATOR

76. Given the following plant conditions:

The plant was at 100% power for 350 days

The plant tripped due to a Large Break LOCA

The crew has entered FRP-J.1, " Response to High Containment Pressure." on an URANGE PATH

+ Containment temperature is 195 F.

Containment radiation level is five (5) R/HR.

Containment hydrogen concentration is 7.0%.

Containment pressure is 7 psig.

Which ONE (1) of the following describes the hydrogen concentration readings?

A. Actual concentrations are HIGHER than indicated due to the steam content in Containment.

B. Actual concentrations are LOWER than indicated because hydrogen is not flammable in a steam environment.

W. All hydrogen measurements are referenced in dry air even if steam exists.

D. Hydrogen flammability is non-conservative when based on using hydrogen concentrations in dry air.

d K/A 000069.K3.01 (3.8/4.2)

FRP-J.1, " Response to High Containment Pressure," Pg. 7. WOG basis for step Modified Question Question 76 of 100

95-2 NRC EXAM - REACTOR OPERATOR l

1 q 79. AOP-014-09 001 '

Given the following plant conditions

The unit is at 75 % power ;

CCW Surge Tank level is Increasing i e No makeup to the CCW surge tank is in progress

R-17, CCW Surge Tank Radiation Monitor, indication is Stable

i. :

Which ONE (1) of the following describes the leak that would provide the above indications?

i

VA. S/G Blowdown sample heat exchanger B. Eacess L/D heat exchanger l C. Air Ejector Condenser l

D. RHR heat exchanger a

l I

K/A 000026)A2.01 2.9/3.5 Answers A and C are the only non-primary systems C= pressure is significantly lower than CCW pressure CW.J3ogdiagram 5379-376 Modified Ouestion s

g vg)

/

tb Question 79 of 100

95-2 NRC EXAM - REACTOR OPERATOR

%. Given the following plant conditions:

RCS pressure = 900 psig

PRT Pressure = 15 psig

CV Pressure = 10 psig

PZR level = 50%

Whicli ONE (1) of the following temperature indications on TE-469, PZR Safety Valve Discharge Temperature, would indicate that the associated PZR PORV was open?

A. 300 F v8. 320*F i C. 340 F D. 360 F b

K/A 000008.EA2.20 3.4/3.6 Steam tables New Question

>f Question 96 of 100

l 95-2 NRC EXAM - REACTOR OPERATOR

.! r

97. RDCNT-09 001 Given the following plant conditions:

A Shutdown Bank "B" rod has fallen into the core.

A dropped rod recovery is in progress per AOP-001, " Malfunction of Reactor Control System", Section A, " Dropped Rod".

The APP-005-E2, " ROD CONT SYSTEM URGENT FAILURE", alarm actuates just after recovery commences.

Which ONE (1) of the following is the source of the URGENT' failure alarm?

A. Pulser failure detector.

B. Phase failure detector.

C. Logic error detector. O[

v0. Multiplexing error detector.

y@Yf '

d I Y /

' K/A 000003G007 (3.4/3.6) Q RDCNT, Gbj. 8, Pg. 31. q'[}( bUsk p[

AOP-001, " Malfunction of Reactor Control System", Section A, Pg.28.

SD-007, Pg 17,18.

Modified Question Question 97 of 100

=

. Rsv. 10 AOP-001 MALFUNCTION OF REACTOR CONTROL SYSTEM

! Pcgs 28 of 73 STEP --

INSTRUCTIONS RESPONSE NOT OBTAINED SECTION A 1

DROPPED ROD (Page 19 of 23) j

= l

APP-005-E2, ROD CONT SYSTEM URGENT FAILURE, will illuminate when i the rod is moved due to all Lift Coil Disconnect Switches being I

4 off in the unaffected group.

TECH SPEC 3.10.1.5 re'stricts operation above 70s power when rods

are misaligned greater than TECH SPEC limits.

t

47. Align The Affected Rod As j Follows:

4 a. Depress 6ED hold the AUTO ROD i 2 DEFEAT Pushbutton

! I i

b. Select the affected bank with 5

the ROD BANK SELECTOR Switch

c. Release the AUTO ROD DEFEAT Pushbutton I d. Withdraw the rod at the specified rate of rod withdrawal until the Group Step Counter has returned to the position recorded in

~

Step 41 1

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ML20147C259

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Subject:

4.3.1 Engineering

4.3.4 Operations

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4.3.1 Engineering

4.3.4 Operations

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