ML041890336

Draft RO & SRO Written

ML041890336
Person / Time
Site:	Oyster Creek
Issue date:	04/19/2004
From:	Hackenberg J AmerGen Energy Co
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
50-219/04-301 50-219/04-301
Download: ML041890336 (152)

Text

R D QUESTION #I The plant is operating in a normal full power lineup when annunciator S-3-d SBO SYSTEM DC LOST activates concurrently with a number of other annunciators which confirm a loss of DC power.

As the BOP operator, you identify that you 4160 V busses I A and I C .

The loss of which of the following 125 VDC busses would account for this indication?

A) DC Distribution Center A B) DC Distribution Center B C) DC Distribution Centers A and B D) DC Distribution Center C ANSWER:D EXPLANATION:

This is specified in attachment ABN-3200.13C-1. A licensed operator would be expected to diagnose the extent of the 125 VDC problem by observing 4160V TECHNICAL REFERENCE(S): Loss of DC Distribution Center C Procedure

- (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # - 1 -

WA # 295004lAA2.02 Importance Rating 3.5 -

KIA Topic

Description:

Ability to determine and/or interpret the extent of partial or complete loss of DC power Question Source: Bank #

Modified Bank # X (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Modified from a DAEC exam

QUESTION #2 What plant design features mitigate the consequences of a release of Radioactive Iodine during a fuel handling accident in the fuel pool.

A. Twenty five feet of water coverage over the fuel plus automatic start of SBGT on high radiation sensed on the 119 elevation of the Reactor Building.

B. Twenty five feet of water coverage over the fuel plus automatic start of SBGT on high radiation sensed on the Reactor Building Ventilation Exhaust line.

C. Twenty feet of water coverage over the fuel plus automatic start of SBGT on high radiation sensed on the 119 elevation of the Reactor Building.

D. Twenty feet of water coverage over the fuel plus automatic start of SBGT on high radiation sensed on the Reactor Building Ventilation Exhaust line.

ANSWER: A EXPLANATION:

At least 23 feet of water level are required to scrub sufficient iodine from released gases (RG 1.25). The auto start of SGTS ensures additional scrubbing of iodine as well as an elevated release. The two, combined, mitigate the consequences (dose to the public) from a fuel handling accident. The Hi Rad on the vent duct is not credited for the refueling accident.

TECHNICAL REFERENCE(S): UFSAR 9.1.2.2.1: RG 1.25 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # J- -

Group # 2 -

WA # 295033/EK2.04 Importance Rating 3.9 -

KIA Topic

Description:

Knowledge of the interrelations between High Secondary Containment Ventilation Radiation Levels and the Standby Gas Treatment System Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #3 The reactor has scrammed but the scram cannot be reset. WHICH ONE of the following explains why the CRD Charging Water Supply Valve V-15-52 is required to be closed?

A) To minimize the cold water injection into the bottom head of the vessel.

B) To prevent damaging the CRD accumulators C) To prevent the CRD pumps from reaching run out conditions.

D) To stop input to Scram Discharge Volume and allow resetting SCRAM.

ANSWER: A EXPLANAT1ON:

This was added to Step 3.12 of 2000-ABN-3200.01 based on SER-5-93 TECHNICAL REFERENCE(S): Step 3.12 of 2000-ABN-3200.01 Reactor Scram (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # 1 -

WA # 295006/AK2.03 Importance Rating 3.7 -

KIA Topic

Description:

Knowledge of the interrelations between SCRAM and the CRD hydraulic system Question Source: Bank # x Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #4 In procedure 205.0, Reactor Refueling, the operator is directed to IMMEDIATELY EVACUATE the area if radiation levels begin to increase following a drop of a fuel assembly from the refueling equipment. What is the basis for evacuating?

A. The direct radiation from the dropped fuel bundle could cause overexposure B. The dropped fuel bundle may breach the cavity seal and drain the cavity.

C. The dropped bundle may release radioactive gasses which will cause over-exposure.

D. The dropped bundle may create a criticality event in the reactor.

ANSWER: C EXPLANATION:

This is the classic fuel handling accident in which the fission product gasses contained within the fuel cladding are released and bubble up through the water. Personnel remaining over the cavity can be exposed to doses approximating 10CFRI 00 limits. The approximately 25 feet of water provide adequate shielding from direct radiation and allow time for action to restore level.

Criticality is avoided by refueling interlocks and refueling patterns, though answer D could be considered a plausible distractor for someone unfamiliar with refueling interlocks and core design.

TECHNICAL REFERENCE(S): RG 1.25: Procedure 205.0 DCI 5.0 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # - 2 -

WA # 295034/EK1.01 Importance Rating 3.8 -

WA Topic

Description:

Knowledge of the operational implications of Personnel Protection as it applies to Secondary Containment High Radiation Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #5 Given the following plant conditions:

e Reactor scrammed from high drywell pressure I O minutes ago.

e Drywell leakage = zero 0

No drywell venting in progress e Drywell temperature has increased from 120F to 240F since the scram e The spread between the GEMAC Narrow Range and Yatway level instruments has increased approximately 10" in the last 10 minutes.

rn All drywell pressure indicators are stable at 3.0 psig rn Suppression pool level constant What can be determined with regard to drywell pressure/pressure indication?

A. Actual drywell pressure is as indicated 3.0 psig and decreasing B. Drywell pressure instruments have failed and pressure is increasing.

C. Drywell pressure is as indicated; 3.0 psig and increasing D. Drywell temperature instruments have failed and pressure is stable.

ANSWER: B EXPLANATION:

Since the temperature has increased 100 degrees since the scram (which occurred at2.9 psig) the pressure should have increased significantly. Since no drywell cooling is available the temperature and pressure will continue to increase. The fact that the spread between the wide range and narrow range has increased should be an indication that the 100 F rise is actual since the WR are affected more than the NR for a given drywell temperature change.

TECHNICAL REFERENCE(S): EOP Bases 2-21 and 2-22 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Out1ine Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

KIA # 295012/AA2.02 Importance Rating 3.9 -

WA Topic

Description:

Ability to determine and/or interpret drywell pressure as it applies to high drywell temperature.

Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #6 The plant is in normal full power operation with no LCOs when the annunciator activates for EMRV OPEN (8-3-9). Operators confirm that a EMRV is partially open using temperature readings at Panel IF/2F. All attempts to shut the EMRV have failed.

The following conditions exist at this time:

Containment Spray has been started in the Torus cooling mode at maximum capacity.

Torus average water temperature was increasing slowly but has STABLIZED at 85F since Torus cooling was placed in service.

Identify the correct actions to address the above conditions.

A. SCRAM the reactor in accordance with Procedure ABN-3200.01,

8. Perform a controlled power reduction per 202. Iand shutdown per 203.

C. for up to 7 days and then a controlled power reduction per 202.1 and shutdown per 203.

D. for up to 14 days and then a controlled power reduction per 202.1 and shutdown per 203.

ANSWER: A EXPLANAT1ON:

RO would be expected to know that a scram is directed if SRV cannot be closed.

Plausible since temp has stablized, but ABN directs immediate scram.

Plausible since temp has stablized, but ABN directs immediate scram.

Plausible since temp has stablized, but ABN directs immediate scram TECHNICAL REFERENCE(S): 2000-ABN-3200.40 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # -2 -

WA # 295013/AK3.02 Importance Rating 3.6 -

KIA Topic

Description:

Knowledge of the reasons for limiting heat additions as it applies to High Suppression Pool Temperature Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # 7 While core loading is in progress when is the Control Room Licensed Operator (CRO) required to complete a communication with the Refueling SRO?

A. Commencement and completion of each shift.

6. Commencement and completion of each move.

C. Whenever a bundle enters the fuel pool.

D. Whenever a bundle leaves the fuel pool.

ANSWER: B EXPLANATION:

TECHNICAL REFERENCE(S): Procedure 205.0, Section 7.3 (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Examination Out1ine Cross-reference: Level RO SRO Tier ## G -

Group ## 2 -

KIA ## 2.2.26 Importance Rating 2.5 -

tVA Topic

Description:

Knowledge of refueling administrative requirements.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 55.43 x Comment: This is specified as a RO duty in Procedure 205.0.

QUESTION #8 Following a control room evacuation, condensate transfer makeup to the shell side of the A Isolation Condenser is established by...

A. Starting the 1-2 Condensate Transfer Pump at the Remote Shutdown Panel (RSP) and then controlling the makeup valve (V-I 1-36) at Local Shutdown Panel (LSP-I 832)

B. Starting the 1-2 Condensate Transfer Pump at Local Shutdown Panel (LSP-1B32) and then controlling the makeup valve (V-I 1-36) at the Remote Shutdown Panel (RSP)

C. Starting the 1-2 Condensate Transfer Pump at Local Shutdown Panel (LSP-I B32 and then controlling the makeup valve (V-I 1-36) locally in the Rx Building, el 95.

D. Starting the 1-2 Condensate Transfer Pump at the Remote Shutdown Panel (RSP) and then controlling the makeup valve (V-I 1-36) locally in the Rx Building, el 95.

ANSWER: C EXPLANATION:

Per page E3-1 of ABN-3200.30 the A IC makeup valve is controlled locally. Only the B has makeup valve control at RSP.. There are no controls for the 1-2 Condensate Transfer Pump on the RSP TECHNICAL REFERENCE(S): 2000-ABN-3200.30 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01) 02029 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 -

Group # 1 -

KIA # 295016/K2.02 Importance Rating 4.0 -

KIA Topic

Description:

Knowledge of the interrelations between the local control stations and Control Room Abandonment Question Source: Bank #

Modified Bank # X (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X 55.43 Comments: Changed from MSlV positions to IC control

QUESTION #9 The Standby Liquid Control (SLC) injection line inside the reactor vessel has broken off. The reactor is at 100% power.

What indication do you have that this occurred?

A. SLC FLOW ON alarm (G-l-b) will illuminate.

B. Core Differential Pressure indication will show a significant decrease.

C. Core Differential Pressure indication will show a significant increase.

D. Fuel Zone level B instrumentation will show a significant decrease.

ANSWER: B EXPLANATION:

With the injection pipe broken there will be no indication of core differential pressure and the reading should approach 0. The injection line is not used for B Fuel Zone Level indication.

TECHNICAL REFERENCE(S): GE Systems Technoloav Manual. ~a 7.4-3

- (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 21 1000/K3.03 Importance Rating 2.6 -

WA Topic

Description:

Knowledge of the effect that a loss or malfunction of the Standby Liquid Control system will have on core plate differential pressure indication.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # I O The reactor has just scrammed from 100% power, all rods are at 00 and you are following ABN-3200.01, "Reactor Scram" procedure. Step 3.4.2 has you "Insert SRM and IRM detectors."

What is the response of the SRM recorder readings as you insert the SRM detectors? Also, what is the reason for the readings you observe?

A. The SRM readings will decrease because the detectors are moving to a lower flux area.

B. The SRM readings will increase because the detectors are moving to a higher flux area C. The SRM readings will remain the same because the flux remains constant with the reactor shutdown.

D. The SRM readings will decrease because the flux is decreasing exponentially with the reactor shutdown.

ANSWER: B EXPLANATION:

Fundamental knowledge TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # -

WA # 215004/K5.03 Importance Rating 2.8 -

KIA Topic

Description:

Knowledge of the operational implications of changing detector position as it applies to the Source Range Monitor (SRM) System.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #I 1 The following conditions exist:

b RBCCW to the drywell coolers isolated on low-low-low RPV water level b

The RPV water level has been restored to >53 inches.

0 The Unit Supervisor has entered EMG-3200.02 Primary Containment Control based on drywell temperature exceeding 150 degrees F b

The Unit Supervisor has directed you to operate all available drywell coolers.

b All RBCCW isolation valve switches are in their normal lineup.

Describe how to re-open the RBCCW isolation valves to the drywell coolers.

A. Place the isolation valve switches in CLOSE, then return the switches to OPEN.

B. Place the isolation valve switches in CLOSE, press the ISOLATION LOGIC RESET pushbuttons (2), then return the switches to OPEN.

C. Place the isolation valve switches in OPEN D. Press the ISOLATION LOGIC RESET pushbuttons (2)

ANSWER: D EXPLANATION:

With the valve position switches in their normal lineup, it is only necessary to press the ISOLATION LOGIC RESET pushbuttons (2) on panel IF/2F. The valves will automatically OPEN as long as the isolation signal has been reset.

TECHNICAL REFERENCE(S): OPS-3024.21 PQ 9 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination OutIine Cross-reference: Level RO SRO Tier # -2 -

Group # - I -

KIA # 223002lA4.03 Importance Rating 3.6 -

KIA Topic

Description:

Ability to manually operate and/or monitor reset system isolations in the control room.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # I 2 While operating at 100% power the following occur:

A loss of RBCCW CCW FLOW LO A and CCW FLOW LOW B alarms have occurred on panel E reviews ABN-3200.19, RBCCW Failure Response directs you to Scram the Reactor in accordance with Procedure ABN-3200.01 ,Reactor Scram.

What is the basis for a Reactor Scram under these conditions?

A. Imminent loss of all Reactor Recirculation Pumps B. Imminent loss of all Drywell Coolers C. Imminent loss of Reactor Water Cleanup System.

D. Imminent loss of Steam Tunnel Recirc Fan Cooling.

ANSWER: A EXPLANATION:

Although all of the statements could occur it is the imminent loss of the recirc pumps (within 60 seconds) that prompts the operator to Scram.

TECHNICAL REFERENCE(S): RBCCW Lesson Plan Dq 9 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01~00061 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 -

Group # 1 -

WA # 295018/AK2.02 Importance Rating 3.4 -

KIA Topic

Description:

Knowledge of the interrelations between partial or complete loss of component cooling water and plant operations.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # I 3 Given the following plant conditions:

The air compressors are in operation per Procedure 334 Instrument and Service Air System

1. 3 Compressor is Lead compressor 0

1. 2 Compressor is Lag compressor 0

A rupture of a 2 InstrumentAir line occurs Annunciator CONTROL AIR PRESS LO illuminates and remains illuminated 0

Current air pressure trend is steady one minute after the line ruptures All compressors are functioning per design basis Describe the response of the three air compressors to this loss of InstrumentAir assuming no operator action A. #3 compressor will run loaded continuously, #2 compressor will start at90 psig and run for I O minutes, #I compressor will start but run unloaded.

B. #3 compressor will run loaded continuously, #2 compressor will start at90 psig and run for 10 minutes, #I compressor will not automatically start.

C. #3 compressor will run loaded continuously, #2 compressor will start at90 psig and run unloaded, #I compressor will start but run unloaded.

D. #3 compressor will run loaded continuously, #2 compressor will start at90 psig and run continuously, #I compressor will not automatically start.

ANSWER: D EXPLANATION:

The 2 line rupture is large enough to equal the output of both the Lead and Lag compressors running loaded continuously and pressure is not increasing above 75 psig Therefore, the Lag compressor will not shutdown after 10 minutes. The #1 Compressor will not start since it would be lined up in STANDBY. #I Compressor will need to be manually started.

TECHNICAL REFERENCE@): Procedure 334 Daaes 6,7,and 16 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # -

Group # - 1 -

WA # 295019/K3.02 Importance Rating 3.5 -

KIA Topic

Description:

Knowledge of the reasons for the response of Standby Air Compressor Operation during partial or complete loss of Instrument Air

Question Source: Bank #

Modified Bank # X (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # I 4 Given the following parameters, determine what action, if any, is required to ensure the shutdown cooling mode of operation is maintained.

Reactor has been shutdown for five hours C Shutdown Cooling Pump just tripped Shutdown Cooling loops A and B are inservice at full flow A and B RBCCW pumps are operating with 1500 GPM flow through A&B Shutdown Cooling Heat Exchangers Reactor Pressure is at 100 psig and slowly rising A reactor recirculation pump is operating Reactor water level is normal (GEMAC)

A. No action is required since the design basis for Shutdown Cooling is satisfied by two loops.

B. Increase RBCCW flow before reactor pressure exceeds 125 psig C. Increase reactor water level to be consistent with Condition 1 D. Re-Start C shutdown Cooling loop ASAP.

ANSWER: D EXPLANATION:

Rising pressure requires additional heat sink. Need an additional S/D Cooling loop. RBCCW flow is already maxed out and cant be increased further. Water level increase will not significantly increase heat sink.

TECHNICAL REFERENCE(S): Procedure 305 Dqs 18, 20,31,32 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01) 000042 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # - 1 -

WA # 295021/2.1.22 Importance Rating 2.8 -

KIA Topic

Description:

Ability to determine Mode of Operation during Loss of Shutdown Cooling Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 55.43 x

Comments: This K&A does have a RO learning objective QUESTION # I 5 After a transient initiates, the following parameter values are noted on the control room panels:

Drywell pressure 4.5 psig rising.

Drywell air temperature 140F rising.

0 Torus pressure 4.9 psig rising Torus water temperature 82F stable Which of the following is indicated?

A. A high pressure discharge into the torus airspace.

6. A safety valve has opened and closed.

C. The containment is functioning normally following a water break LOCA.

D. A high pressure discharge into the drywell and at least one torus to drywell vacuum breaker is open.

ANSWER: A EXPLANATION:

Since the torus pressure is higher than the drywell pressure the discharge is into the torus.

Since torus temperature is stable it can be concluded that the discharge is to the airspace and not the torus water. The discharge is passing through a vacuum breaker to the drywell.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 -

Group # - I -

KIA # 295024/EA2.04 importance Rating 3.9 -

KIA Topic

Description:

Ability to determine and/or interpret Suppression Chamber Pressure as it applies to High Drywell Pressure.

Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # 6I Following a valid scram signal at 100% power the following plant conditions exist:

e Mode switch in SHUTDOWN e All RODS are at MSlVs are closed EMRVs are cycling at their setpoint Reactor Power 40%

What malfunction or malfunctions have occurred?

A. SCRAM and ARI have failed B. SCRAM and ATWS have failed C. ARI and ATWS have failed D. SCRAM, ARI and ATWS have failed ANSWER: A EXPLANATION:

With all rods at 48 and reactor pressure above EMRV setpoint ARI should have initiated, but did not. ATWS has (obviously) worked (and tripped the recirc pumps) since power would be above 40% if it had failed.

TECHNICAL REFERENCE(S): Procedure ABN-3200.01. Reactor Scram

- (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 -

Group # 1 -

KIA # 295025/EK3.06 Importance Rating 4.2 -

KIA Topic

Description:

Knowledge of the reasons for the response of Alternate Rod Insertion as it applies to High Reactor Pressure Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # I 7 Given the following plant conditions:

e Reactor scrammed on MSlV closure 10 minutes ago (manually closed due to high radiation)

All control rods are at 00 One IC is in service, the other is inoperable and unavailable EMRVs are being manually opened as necessary for pressure control Feedwater is providing reactor water makeup Reactor Pressure is between 950 psig and 1000 psig Reactor water level is at +I 00" and steady Torus temperature is 94F and slowly rising Drywell pressure I.5psig and steady Drywell bulk temperature is 151 degrees F What system(s) must be initiated to control plant parameters per EOPs A) DrywelVTorus Sprays and Drywell Cooling B) ADS and Drywell Cooling C) Torus Cooling and Drywell Cooling D) Bypass valves and Drywell Cooling ANSWER: C With heat being added by EMRVs torus cooling will exceed 95 degrees F. Primary Containment Control requires that torus temperature be maintained below 95 degrees. Drywell/torus sprays are not required until 12 psig. Other choices have no affect on torus temperature.

EXPLANATION:

TECHNICAL REFERENCE(S): EMG-3200.02 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning 0bjective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier ## 1 -

Group # 1 -

KIA ## 295026lEK3.02 Importance Rating 3.9 -

KIA Topic

Description:

Knowledge of the reasons for the response of suppression pool (torus) cooling as it applies to Suppression Pool (Torus) High Water Temperature Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge

Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #I8 Given the following plant conditions:

Mode switch is SHUTDOWN All rods are at 48 Reactor power is at 100%

MSlVs are closed Both ICs are operating All SRVs are cycling Torus temperature is 125F Drywell pressure is 0.9 psig What system(s) must be operated to return suppression pool temperature to within Technical Specifications limits?

A) Containment Spray in torus cooling, ARI and SLC B) Containment Spray in torus spray, ARI and SLC C) Reactor Recirculation (trip pumps), ARI and CRD D) Reactor Recirculation (trip pumps), ARI and CRD ANSWER: A EXPLANATION:

A combination of heat removal from the torus as well as reducing reactor power to ~ 2 %is required to return torus temperature to <95 degrees F. Torus spray cannot be used below I.O Psig TECHNICAL REFERENCEtS): EMG-3200-02 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # 1 -

KIA # 29502612.4.21 Importance Rating 3.7 -

KIA Topic

Description:

Knowledge of the parameters and logic used to assess the status of safety functions (as they relate to Suppression Pool High Water Temperature)

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X

Comments:

QUESTION #I9 Given the following conditions:

0 A Loss of Offsite Power has occurred 0

Reactor is at rated temperature and pressure 0 entry condition for EMG-3200-02, Primary Containment Control 0

0 You are operating 0

The CRS asks the question can bulk drywell temperature be maintained below 150

. degrees F .

Your response is NO.

What is the basis for this response?

A. Chilled water is not available.

B. Drywell cooler fans have shut down.

C. RBCCW isolation valves have closed.

D. Rated capacity of 5 drywell coolers is inadequate.

ANSWER: C EXPLANATION:

RBCCW isolation occurs with Lo-Lo water level and High Drywell Pressure. Without RBCCW there is no heat sink for drywell cooling and temperature cannot be reduced. The RBCCW isolation must be cleared or bypassed (Support Procedure 27) this is done if/when the answer is NO.

TECHNICAL REFERENCE(S): Primarv Containment Lesson Plan DCI 13; EOP-2 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning Objective: (04)07346 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - I -

Group # 1 -

WA # 295028/EA1.02 Importance Rating 3.9 -

KIA Topic

Description:

Ability to operate and/or monitor the drywell ventilation system as it applies to high drywell temperature Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X

55.43 Comments:

QUESTION #20 The reactor is operating at 100% power with A , B , C and D recirc loops in operation. The brushes on the Atach generator fail and the tach generator output signal goes to 0. What is the effect of this failure on reactor power over the next 60 seconds?

A. Reactor Power will remain constant since the scoop tube will lock up.

B. Reactor Power will drop slightly as the TOCL shifts to DCC Y.

C. Reactor Power will initially start a ramp decrease, then will have a step decreaseas the A MG set trips on undervoltage.

D. ramp increase, ANSWER: C EXPLANATION:

Flow will initially wind down but then the loop will trip on undervoltage due to effect on voltage regulator.

TECHNICAL REFERENCE(S): Recirc Flow Control Lesson Plan DCI 5 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (04) 00214 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # - 2 -

WA # 202002/K3.02 Importance Rating 4.0 -

KIA Topic

Description:

Knowledge of the effect that a loss or malfunction of the Recirculation Flow Control System will have on Reactor Power Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #21 Given the following plant conditions:

Reactor at 100% power RBHVS operating normally Hi area temperature alarm occurs at the 75' Precoat Tank Hi area temperature at 75' Demin Vault S

A) Monitor for Maximum Safe Radiation and shutdown theRBHVS system before it exceeds Max Safe (1000 MR/hr).

B) Monitor the 75' area temperatures and shutdown theRBHVS system before either reaches max safe (225 degrees F).

C) Monitor the RBHVS ductwork temperature to determine if it exceeds the design basis temperature of 299 degrees F.

D) Initiate Emergency Depressurization because two areas in the reactor building have exceed maximum safe operating temperature.

ANSWER: C EXPLANATION:

Step 3.2.3 in Procedure 329. Operator should take action before auto isolation. For answer A the operator may assume the temperature is from a RWCU leak TECHNICAL REFERENCE(S): Procedure 329 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning Objective: (01/04) 10447 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # 2 -

WA # 295032/EA1.03 Importance Rating 3.7 -

WA Topic

Description:

Ability to operate and/or monitor secondary containment ventilation as it applies to high secondary containment area temperature Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis

10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #22 System Operating Procedure for the Reactor Building HVAC (Procedure 329) has the operator start the exhaust fan first (on system startup) and shut down last (during system shutdown).

What is the reason for the fan sequence?

A. Prevent collapse of exhaust ductwork B. Prevent collapse of the supply ductwork C. Prevent SGTS from automatically starting D. Prevent positive pressure and possible siding failure ANSWER: D EXPLANATION:

This is extrapolated from two sources: the lesson plan pages 15 and 16 says the reason for the fan sequence is to prevent creation of positive pressure; page 6 specifies that Rx building relief (from internal pressure greater than .2 psig) is provided by buckling of girts.

TECHNICAL REFERENCEtS): Secondary ContaoinmenffSGTS Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01/04)10447 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group ## - 2 -

WA # 295035/EK3.01 Importance Rating 2.8 -

K/A Topic

Description:

Knowledge of the reasons for blow out panel operation as it applies to Secondary Containment High Differential Pressure Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge - X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments: The OC Reactor Building design provides structural girts that will, preferrentially, yield during an overpressure condition. When these girts yield, the sheet metal siding seams will open and allow for release of the overpressure condition. This is comparable in function to the more standard blowout panel design which uses grooved rivets on selected panels to preferrentially blow out.

QUESTION #23 Given the following plant conditions:

0 0

0 Main Steam Line Radiation Monitors all at approximately 550 mr/hr 0 Stack Effluent HI alarm 0

Reactor Bldg Vent Radiation at 8 mr/hr 0

RCS activity at 90% of TS limit B IC isolated for maintenance Significanthisible packing leak from A IC outboard steam isolation valve 0

No leaks in the A IC tube bundle What action would result in having the greatest reduction in the thyroid damage for the public?

A. Close A IC outboard steam isolation valve B. Reduce reactor power until stack effluent HI alarm clears C. Start SGTS and shutdown Reactor Building HVAC D. Close AIC vent valve ANSWER: C EXPLANATION:

Starting SGTS is the only action that will remove radioactive iodine being released from the steam leak. The AOG will remove all iodine TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning 0bjective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # -

Group # 1 -

WA # 295038/EK1.01 Importance Rating 2.5 -

KIA Topic

Description:

Knowledge of the operational implications of the biological effects of radioactive ingestion as it applies to Off Gas Release rate.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X

55.43 Comments:

QUESTION #24 During startup preparations for the Shutdown Cooling System per procedure 305, Shutdown Cooling System Operation, the operator is directed to ensure the RBCCW surge tank level is low in the allowable band (Prerequisite 4.1.8)

What is the basis for this prerequisite?

A. This will ensure rapid automatic makeup should a leak develop in the RBCCW Heat Exchanger.

B. This will ensure rapid response (manual isolation)

C. This will ensure a leak of RCS into the RBCCW System is readily detected D. This will prevent overflowing the RBCCW Surge Tank from the heatup of RBCCW fluid ANSWER: D EXPLANATION:

D is specified as the basis in the lesson plan. may be a true answers but are not the basis for the step.

TECHNICAL REFERENCE(S): Shutdown Coolinq Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (01) 07232 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 205000/K1.05 Importance Rating 3.1 -

KIA Topic

Description:

Knowledge of the physical connections and/or cause-effect relationships between the Shutdown Cooling System and the Component Cooling System Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #25 Following a loss of offsite power, the crew has initiated EMG-3200.01A RPV Control-No ATWS A. SPARGER 1 DP HI, SYSTEM 1 FLOW PERMISSIVE, BSTR PUMP N C OL B. SPARGER 1 DP HI, SYSTEM 2 FLOW PERMISSIVE, BSTR PUMP C/D OL C. SPARGER 2 DP HI, SYSTEM 2 FLOW PERMISSIVE, BSTR PUMP N B OL D. SPARGER 2 DP HI, SYSTEM 2 FLOW PERMISSIVE, BSTR PUMP C/D OL ANSWER: B EXPLANATION:

For A, C and D the sparger dp alarm indicates the that has the flow permissive faulted and the flow m TECHNICA L REFERENCE(S) : (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 209001/A2. 05 Importance Rating 3.3 -

KIA Topic

Description:

Ability to predict the impacts of Core Spray Line Break on the Low Pressure Core Spray System and based on those predictions use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #26 Given the following conditions have existed for 3 minutes:

Loss of offsite power has occurred Drywell pressure at or above 5 psig Reactor Vessel level at or below 60 TAF Reactor Pressure > 950 psig All EMRVs are closed Both ICs have initiated ADS BYPASSED and ADS TIMER START annunciators are NOT lit All four AUTOSTART DISABLED annunciators on Panel B are annunciated All emergency buses are energized by the diesel generators 0

All DC buses are energized What is the configuration of the Core Spray Booster Pumps?

A. All Core Spray Booster Pumps are running B. NO Core Spray Booster Pumps are running C. Only Aand B Core Spray Booster Pumps are running D. Only IC and ID Core Spray Booster Pumps are running ANSWER: B EXPLANATION:

With LO LO RPV Level and HI Drywell Pressure signals in for 3 minutes ADS should have hat A

and B priority pumps) will not start until the Core Spray Pumps start.

TECHNICAL REFERENCE(S): Core Sprav Lesson Plan (Attach if not previously provided) icants during examination:

Learning 0bjective: (01) 00302 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

KIA # 209001/K1.05 Importance Rating 3.7 -

KIA Topic

Description:

Knowledge of the physical connections and/or cause-effect relationships between Core Spray System and ADS.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #27 A plant startup is in progress with the following conditions:

0 Mode switch is in STARTUP.

0 The SRMs are fully inserted and reading approximately 20 CPM.

At step 6.18 of Procedure 201, Plant Startup you are required to perform the following:

VERIFY operability of the ROD WORTH MINIMIZER by confirming the RWM is not bypassed, 0

Selecting Control Rod 30-51 and verifying a SELECT ERROR is received.

The status of Control Rod 30-51 is:

A. In the first group of rods to be withdrawn.

B.

C. Already withdrawn beyond position 02 D. Inoperable with low pressure in the Scram Accumulator.

ANSWER: B EXPLANATION:

None of the other choices will result in a SELECT ERROR.

TECHNICAL REFERENCE(S): RWM Lesson Plan pci 14 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

K/A # 2.2.1 Importance Rating 3.7 -

KIA Topic

Description:

Ability to perform pre-startup procedures for the facility/including operating those controls associated with plant equipment that could affect reactivity.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #28 Given the following:

Reactor has just scrammed following 6 months at 100% power MSlVs have closed Both ICs are operable Coincident with the SCRAM there was a loss of DC-D What is the status of ICs one minute after the SCRAM?

A. Ais In Service , B is Standby B. Ais Standby, B is In Service C. Ais In Service, B is In Service D. Ais Standby, B is Standby ANSWER: B EXPLANATION: Loss of DC-D causes isolation of the A IC. B will be on in response to rising reactor pressure considering the MSlVs have closed TECHNICAL REFERENCE(S): Isolation Condenser Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 207000/K2.02 Importance Rating 3.5 -

KIA Topic

Description:

Knowledge of electrical power supplies for initiation logic of Isolation Condensers Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #29 The plant is at 100% power when a partial loss of TBCCW occurs. What action can you take that will have the greatest impact on maximizing the utilization of the available TBCCW flow while minimizing the impact on the reactor power?

A. Shutdown Control Room air conditioners system A.

B. Reduce the Turbine-Generator load.

C. Trip one operating Feed Pump and one Condensate Pump D. Isolate Feed and Main Steam samples.

ANSWER: B EXPLANATION:

The largest heat load on TBCCW is Hydrogen Cooling. Other large heat loads include Generator Phase Bus Duct cooler, Stator Winding and Turbine Lube Oil. These are all directly related to Turbine-Generator Load. By reducing turbine load all of these heat loads are diminished. If the load limit is used the reactor can be maintained at or near 100% power, thus minimizing its impact. Reducing recirc pump speeds will actually increase heat load on TBCCW since there is more "slippage" in the fluid couplings. Tripping FeedICondensate pumps will result in reducing reactor power.

TECHNICAL REFERENCE(S): TBCCW Lesson Plan, Dq 8 and 9 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01108209 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # 1 -

WA # 295018/AK1.01 Importance Rating 3.5 -

KIA Topic

Description:

Knowledge of the operational implications of the effects on componentkystem operations as it applies to Partial or Complete Loss of Component Cooling Water Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #30 Following a SCRAM from 10% power (Mode switch in STARTUP), one of the IRMs is left in Range 8 during performance of Reactor Scram procedure. All other Neutron Monitors are operating normally.

What is causing the Withdraw Block?

A. IRM downscale and NOT in Range 1

6. IRM downscale and NOT in Range 2 or below C. APRM Downscale with corresponding IRM downscale D. IRMHI ANSWER: A EXPLANATION:

With the Mode switch

• in STARTUP the IRMs would not be withdrawn so answer D is incorrect. With range left in 8 and reactor scrammed there would be no IRM HI. The APRM downscale rod block is in RUN.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Examination 0utline Cross-reference: Level RO SRO Tier # 2 -

Group # -

WA # 215003/K4.01 Importance Rating 3.7 -

WA Topic

Description:

Knowledge of IRM design features and/or interlocks which provide rod withdrawal blocks Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #31 A reactor startup is in progress with the following conditions:

0 The previous shift had increased power from 60% to 80%.

e Your shift just increased power from 80% to 90% power when all operable Main Steam Line Rad Monitors are verified to read greater than 550 mr/hr but less than 800 mr/hr.

0 Within one minute area radiation monitors in the vicinity of the Moisture Separators experience at least a doubling of their readings.

0 The Off Gas monitors are observed to be unaffected.

What actions are required to mitigate the consequences of these abnormal conditions?

A. SCRAM the reactor in accordance with ABN-3200.01 , Reactor Scram and CLOSE MSlVs and IC Vents.

B. Reduce Hydrogen Injection flow to between 5 and 6 scfm and monitor the effect on Main Steam Line Activity for the next 10 minutes.

C. Direct Chemistry to sample Off Gas (Technical Specification 4.6.E) and request guidance from Reactor Engineering.

D. Start plant shutdown in accordance with Procedure 203, Plant Shutdown ANSWER: B EXPLANATION:

Since the radiation increase is not observed in the off Gas System it can be concluded that the radiation is from an isotope with a very short half life. Hydrogen Injection produces, almost exclusively, N16 with a half life of approximately 7 seconds. Per Plant Startup procedure Hydrogen Injection is increased in 10% power increments and , given that there has been no increase in Off Gas activity it is reasonable to assume the increase is attributed to hydrogen Injection. The reference procedure has the operator first reduce Hydrogen Injection before taking other action. If the MSL radiation decreases in response to reducing Hydrogen Injection then only monitoring of Off Gas is required. The other actions are specified if the HI MSL condition is attributed to fuel failure.

TECHNICAL REFERENCE(S): ABN-26 Rev 0 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination-OutlineCross-reference: Level RO SRO Tier # -2 -

Group # 2 -

WA # 272000/A2.07 Importance Rating 2.6 -

WA Topic

Description:

Ability to predict the impacts of Hydrogen Injection on the Radiation Monitoring System and based on those predictions use procedures to correct, control or mitigate the consequences of these abnormal conditions Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments: A potential discrepancy between the ABN and the normal procedure was noted. If the ABN is revised, this question may need revision before administering.

QUESTION #32 The following conditions exist:

The plant is at 100% power The AOG is in service.

e A swap of the recombiners has just been completed with A placed in service.

e The following alarms come in in rapid succession: RECMBNR A H2 HI, OFF GAS TEMP HI, AOG FLOW LOW.

e This is followed by an automatic Off Gas Isolation as indicated by annunciator OFF GAS ISOL ACT II.

What condition caused the noted alarms and indications?

A. Inadequate warmup of the on-coming recombiner B. Explosion and fire in the Off Gas System C. Loss of steam flow to the Steam Jet Air Ejectors D. Excessive flow in the Hydrogen Injection System ANSWER: B EXPLANATION:

As noted in the alarm response IOF-1-b the Off Gas isolation is intended to protect against an off gas explosion. The alarms an indications are classic signs of an off gas explosion and fire as denoted in the reference procedures.

TECHNICAL REFERENCE(S): Annunciator 1OF-I -b; 2000-ABN-3200-23 and 25 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

KIA # 271OOO/A3.03 Importance Rating 2.8 -

KIA Topic

Description:

Ability to monitor automatic operations of the Off Gas System including system temperatures.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #33 A reactor startup is in progress in accordance with "Plant Startup" procedure. The following conditions exist:

Reactor power at approximately 20% with 3 to 4 bypass valves open.

The Turbine Generator is at 1800 RPM and ready to be sychronized to the grid.

When the turbine is initially loaded, and bypass valves closed how do the Condensate System "sucker" and "dumper" (hotwell level control) valves respond over the succeeding 30 minutes?

Assume reactor power is maintained constant for the 30 minutes following the initial turbine loading.

A. The small "sucker" valve opens initially but within approximately 10 minutes it closes.

From 10 minutes to 30 minutes the small "dumper" valve periodically cycles open and closed.

B. The small "dumper" valve opens initially but within approximately 10 minutes it closes.

From I O minutes to 30 minutes the small "sucker" valve periodically cycles open and closed.

C. The large "sucker" valve opens initially but within approximately 10 minutes it closes.

From 10 minutes to 30 minutes the large "dumper" valve periodically cycles open and closed.

D. The large "dumper" valve opens initially but within approximately I O minutes it closes.

From I O minutes to 30 minutes the large "sucker" valve periodically cycles open and closed.

ANSWER: A EXPLANATION:

Initially the steam flow is diverted from the condenser to the turbine and fills up some of the feed heaters, therefore the small "sucker" valve opens in response to a decreasing hotwell level.

After conditions stabilize (in approximately 10 minutes) the small sucker valve closes. Over the next 20 minutes the small "dumper" valve opens in response to increasing hotwell level as the cascading drains on the feedwater heaters open. Additionally, the hotwell level control is set up with a bias to reject water from the hotwell as a means to keep the CST in circulation. The large valves are not required for this level change.

TECHNICAL REFERENCEtS): Condensate/Feedwater Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

K/A # 256000/A3.06

Importance Rating 3.0 -

KIA Topic

Description:

Ability to monitor automatic operations of the Reactor condensate System including hotwell level Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #34 The plant is at 100% power with the following conditions:

Both Emergency Diesel Generators (EDGs) are in Standby.

0 Following a CNTRL DC LO/LOST annunciator actuation on EDG 1, you are dispatched to the EDGs to determine the status.

0 You find that the battery charger for EDG 1 output voltage is 110 volts.

0 All breaker position indication at Panel 9XF are normal.

What impact does this have on EDG operation?

A. The EDG is operable since the EDG can start with only one crank of the engine.

B. The EDG is inoperable due to the impact on the DC Turbocharger lube oil pump.

C. The EDG is operable since there is adequate breaker control power D. The EDG is inoperable since there is inadequate starting power.

ANSWER: D EXPLANATION:

Nominal voltage is 117.6 V DC. With voltage at 110 V it is likely to impact the ability of the battery to crank the engine. It does not satisfy surveillance requirements 4.7.B.1 .c (112 V). DC Turbocharger is a backup to the AC pump. Breaker control will be adequate down to approximately 105 V (coils rated at 90 V).

TECHNICAL REFERENCE(S): EDG Lesson Plan: TS 4.7.B.1 .c: annunciator T-2-b (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01)00800 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 264000/K6.01 Importance Rating 3.8 -

WA Topic

Description:

Knowledge of the effect that a loss or malfunction of the starting air (Starting Motors/Batteries) would have on the Emergency Diesel Generators Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43

Comments: Used Starting Motors (DC) vs Starting Air since the OC EDGs do not have starting air.

QUESTION #35 There is an auto trip of the containment spray pumps if drywell pressure is 0.6 psig in the DW spray mode. What is the purpose of this trip?

A. To preclude exceeding NPSH limits of the Containment Spray Pump(s).

B. To prevent opening the Torus to Reactor Building Vacuum Breaker(s).

C. To prevent opening the Torus to Drywell vacuum breaker(s).

D. To provide a differential pressure in the containment spray heat exchanger(s).

ANSWER: B EXPLANATION:

Opening the Torus to Reactor Building Vacuum Breaker(s) will introduce oxygen to the containment, which is required to be inerted.

TEC HNICA L REFERENCE(S): Containment Spray Lesson Plan pg 7 of 19 (Attach if not previously provided)

Proposed references t o be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

WA # 226001/K5.06 Importance Rating 2.6 -

KIA Topic

Description:

Knowledge of the operational implications of vacuum breaker operation on the Containment Spray System Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Used RHR K&A even though OC has a Containment Spray System, not a mode of RHR as in later BWRs. This was discussed with Fred Guenther and he agreed this is the right thing to do

QUESTION #36 The Reactor is at 100% power with 4 recirculation pumps in service. The shutdown pump is in the IDLE configuration. A single additional recirculation pump trips.

What is the direct impact on reactor powedreactor operation from the recirculation pump trip and what action/actions must be taken to satisfy license conditions?

Assume the tripped pump is placed in the IDLE configuration per 2000-ABN-3200.02 "Recirculation Pump Trip" A. The safety limit for power/flow has been exceeded. Operation in this configuration is not permitted. A reactor shutdown must commence immediately.

B. The LCO for Recirculation Loop Operability cannot be satisfied. Operation in this configuration is not permitted. A reactor shutdown must commence immediately.

C. Reactor power will initially decrease to approximately 7 5%. Reactor power can be raised up to 95% by a combination of rod pulls and recirc flow increases D.

ANSWER: D EXPLANAT1ON:

Per TS 3.3.3.F power operation up to 90% is allowed with 3 recirc pumps running. A combination of rod pulls and flow increases will be necessary to restore power to 90% while satisfying core thermal limits.

TECHNICAL REFERENCE(S): Reactor Recirculation lesson plan : TS 3.3.F

- (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01) 10441 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

WA # 202001/K3.03 Importance Rating 3.9 -

KIA Topic

Description:

Knowledge of the effect that a loss or malfunction of the Reactor Recirculation system will have on reactor power.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43

Comments:

QUESTION #37 Following a rod drifting in, the RWM will relatch. RMCSwill locate the highest completed step that meets the following criteria:

A. LESS THAN three insert errors and AT LEAST two rods are withdrawn past the insert limit.

B. NO insert errors and AT LEAST one rod is withdrawn past the insert limit C. LESS THAN three insert errors and AT LEAST one rod withdrawn past the insert limit.

D. NO insert errors and NO withdraw errors ANSWER: C EXPLANATION:

Obtained from OC Training as a bank question used previously.

TECHNICAL REFERENCE(S): RWM Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01) 10446 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # - 2 -

WA # 201006lK4.06 Importance Rating 3.2 -

KIA Topic

Description:

Knowledge of Rod Worth Minimizer design feature(s) that permit correction of out of sequence rod positions Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #38 During a reactor startup the Unit Supervisor directs you to use continuous rod withdrawal.

Describe the design feature of the Control Rod Hydraulic System/RMCS that provides for continuous rod withdrawal.

A. The ROD OUT switch bypasses the rod out permissive relay in RMCS and allows for continuous application of drive flow.

B. The NOTCH OVERRIDE switch bypasses the rod out permissive relay in RMCS and allows for continuous application of drive flow.

C. The ROD OUT switch stops the automatic sequence timer in RMCS and allows for continuous application of drive flow.

D. The NOTCH OVERRIDE switch stops the automatic sequence timer in the RMCS and allows for continuous application of drive flow.

ANSWER: D EXPLANATION:

When the notch override position is selected in conjunction with selecting withdraw on the RMCS the timer is prevented from continuing into the settle cycle. This causes thendriveout signal to remain and drive water pressure is maintained on the top of the drive piston. The ROD OUT-NOTCH switch does not affect the SE-TTLE cycle.

TECHNICAL REFERENCE(S): RMCS lesson Dlan (PCI 9 of 20)

(Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01) 00726 (As available)

Examination OutIine Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

KIA # 201002/K4.05 Importance Rating 3.3 -

KIA Topic

Description:

Knowledge that the notch override has on rod withdrawal.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #39 Following a loss of offsite power incident the CRD system continues to provide a high pressure makeup water source for the Reactor.

During this mode of operation what concerns do you have regarding the CRD system operation.

A. CRD pump suction pressure will increase which may cause the discharge relief to lift.

B. CRD pump suction pressure will decrease which may cause pump trip.

C. Oxygenated water supplied directly from the CST may induce stress corrosion.

D. The minimum flow back to the CST will be warmer and affect CRD pump NPSH.

ANSWER: C EXPLANAT1ON:

The CRD pump is normally provided de-oxygenated water from the water quality line in the Condensate System (downstream of Condensate Demineralizers). With a loss of offsite power Condensate Pumps are lost and the CRD suction flow comes directly from the CST which has oxygen-rich water. The suction pressure should not significantly change. There will be no change to the CRD pump NPSH even though the suction is changed to the CST.

TECHNICAL REFERENCE(S): CRD Hvdraulic lesson plan(Pn 14 of 47)

(Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (04)00004 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

WA # 201001/K6.02 Importance Rating 3.0 -

KIA Topic

Description:

Knowledge of the affect that a loss or malfunction of the Condensate Storage Tank will have on the CRD Hydraulic system.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #40 Given the following plant conditions:

Reactor is in COLD SHUTDOWN following a forced outage 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> ago Reactor head is still installed Head vent is open All Reactor Recirculation Pumps are shutdown for maintenance Two loops of Shutdown Cooling are in service at 3000 GPM each Shutdown Cooling inlet temperatures are being monitored at 212 degrees F. b) Raise Reactor level >185"TAF D. a) There is a leak in RBCCW system. b) Shutdown RBCCW system ANSWER: C EXPLANATION:

These conditions replicate the conditions noted in LER 93-002. This is the basis for Precaution 5.2.1 in Procedure 305. The other answers may have some credibility but would not result in all the observed conditions; specifically they would not raise head temperature.

TECHNICAL REFERENCE(S): Shutdown Coolina System Lesson Plan page 20and 21 Procedure 305. pas 30 and 31 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None

Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 205000/A2.09 Importance Rating 3.6 -

KIA Topic

Description:

Ability to (a) predict the impact of Reactor Low Level on Shutdown Cooling System operation and (b) based on those predictions use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: The decrease in SDC flow can be attributed to the increase in SDC inlet temperature.

QUESTION ##41 The reactor is starting up following a forced outage to repair a RBCCW piping leak in the drywell. Both RBCCW pumps are operating with both pumps indicating approximately 2000 GPM. Restoration of RBCCW flow to the drywell is in progress and the following sequence occurs:

b An EO has been dispatched to the reactor building to "maintain local Drywell RBCCW D/P between 18.4 and 19.6 during the flow restoration" a Restoration of RBCCW to the drywell is completed and the startup proceeds b

During the next shift, when Shutdown Cooling is secured, the 1-1 RBCCW pump is shutdown.

b Wdhin approximately two minutes of the 1-1 pump shutdown all five low flow alarms associated with RBCCW flow to the recirc pump seals come in.

b RBCCW pump 1-1 is restarted and all five low flow alarms clear.

b When asked for a local dp reading the EO reports "it is approximately 19 psid".

What was the cause of the low flow alarms when only 1-2 pump is operating?

A. The 1-2 RBCCW pump is failing or has failed.

B. Debris from the piping repair has clogged the recirc pump seal coolers C. Drywell Outlet Valve (V-5-709) is throttled too far closed for single pump operation.

D. The local dp indicator is faulty and is reading too low.

ANSWER: C EXPLANATION:

Per procedure, the operator would throttle V-5-709 to maintain dp at approximately 19 psig when both pumps were running. When one pump is shutdown the flow to the drywell (and the dp) decreased. 2000 GPM is the expected flow from a "healthy" pump during two pump operation. Clogging of the coolers would cause the dp to increase.

TECHNICAL REFERENCE(S): CAP 02000-0855, Procedure 309.2 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None-Learning 0bjective: (01100065 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 -

Group # - 1 -

WA #

400000/K3.01 Importance Rating 2.9 -

KIA Topic

Description:

Knowledge of the affect that the loss or malfunction of CCWS would have on loads cooled by ccws.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: There is no step in 309.2 to "reset" V-5-709 when going from two pump to one pump operation.

QUESTION #42 With the unit at 100% power during a particularly cold, snowy night a RCVRZANSTR AIR PRESS L O alarm actuates and you are dispatched to implement ABN-3200.35 Instrument Air System Failure. You have completed all Operator Actions through manually isolating the cross tie with Service Air (V-6S-2). Pressure is now stable at 76 psig.

What are the most significant affects of your actions?

A. Main Steam Isolation valves have closed; two or more control rods have drifted in.

B. Feedwater Control Valves have locked up; CST Makeup Valves have failed open C. Instrument air will be contaminated with water vapor; subsequent control problems may develop.

D. Drywell supply and exhaust dampers have closed; Reactor Building to Torus vacuum breakers have opened ANSWER: A EXPLANATION:

As long as the air pressure is maintained >75 psig there should be no direct equipment problems associated with instrument air. However, continued bypass of air dryers will cause the instrument air to saturate with moisture. Moisture laden air will result in equipment failures; particularly outside equipment.

T ECHNICA L REFERENCE(S) : (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

KIA # 300000/K4.02 Importance Rating 3.0 -

KIA Topic

Description:

Knowledge of design features or interlocks in the Instrument Air System which provide for cross over to other air systems.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X IO CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #43 A Station Blackout occurred and DC Bus C has inadequate voltage to close associated 4160 V breakers from the control room. Offsite power has been restored and you are dispatched to 4160 Bus A to close 4160 V breakers locally in accordance with Loss of Distribution Center C procedure.

How is this accomplished and what safety/equipment concerns exist with this mode of operation?

A. Pull control fuses and use test block to CLOSE. Concern is for not being able to synchronize locally B. Ensure control fuses are installed and use test block to CLOSE. Concern is the potential loss of auto trip capability C. Ensure closing spring is charged and depress local mechanical CLOSE pushbutton.

Concern is the potential loss of auto trip capability.

D. Ensure closing spring is charged and depress local mechanical CLOSE pushbutton.

Concern is for not being able to synchronize locally.

ANSWER: C EXPLANATION:

With a loss of the DC bus the trip coil would not energize to trip the breaker even if a faulted condition existed on the bus. A severe fault on the bus could cause the breaker to catastrophically fail and severely injure anyone standing in front of the breaker.

There is no concern for synchronizing since the 4160 buses are dead. The test switch will not close the breaker if it is racked in.

TECHNICAL REFERENCE(S): Loss of Distribution Center C procedure (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # -

WA # 262001/A4.03 Importance Rating 3.2 -

WA Topic

Description:

AC Electrical Distribution; Local operation of breakers.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X

55.43 Comments:

QUESTION ##44 The reactor is at 2% power following a refueling outage. The drywell temperature is 125 degrees F. EMRVs are being cycled to satisfy Technical Specification 4.4.8.1 Valve Operability (attached).

Describe the indications, on any given EMFU, that demonstrate the valve is OPERABLE A. Pre-stroke tailpipe temperature 125 degrees F, with EMFW switch in OPEN tailpipe temperature is 500 degrees F, with EMF# switch in AUTO tailpipe temperature slowly returns to 125 degrees F.

B. Pre-stroke tailpipe temperature 125 degrees F, with E M W switch in OPEN tailpipe temperature is 330 degrees F, with E M W switch in AUTO tailpipe temperature slowly returns to 125 degrees F.

C. Pre-stroke tailpipe temperature 200 degrees F, with EMKV switch in OPEN tailpipe temperature is 330 degrees F, with EMFU switch in AUTO tailpipe temperature immediately returns to 200 degrees F.

D. Pre-stroke tailpipe temperature 200 degrees F, with EMFW switch in OPEN tailpipe temperature is 500 degrees F, with EMF# switch in AUTO tailpipe temperature slowly returns to 200 degrees F.

ANSWER: B EXPLANATION:

The back pressure on the tailpipe will result in a temperature approximately half way between RCS saturation (540 F) and 212 F. A temperature of 330 F was confirmed by OC Training. If the valve is not leaking, the tailpipe temperature should equilibrate with the drywell. The return (following opening) will be slow (several hours).

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: TS 4.4.B.1 Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 -

Group # 1 -

K/A # 239002/5.04 Importance Rating 3.3 -

KIA Topic

Description:

Knowledge of the operational implications of tailpipetemperatures as they apply to ReliefSafety valves.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge

Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 x 55.43 Comments:

QUESTION #45 In Attachment ABN-3200.01-2 of the Reactor Scram procedure the operator is directed to Open EMRVs in sequence: A, D, B, C, E . This same sequence is specified in EOP Support Procedure 12 Alternate Pressure Control System-EMRVs. The basis for the EMRVs sequence is:

A. To minimize the possibility of having a stuck open EMRV.

B. To minimize the total energy transmitted to the torus.

C. To equalize the thermal loading on the EMRVs D. To equalize the thermal loading in the torus and tailpipes.

ANSWER: D EXPLANATl0N:

The EMRV tailpipes are arranged around the torus. By alternating the EMRVs the discharge of steam is equalized throughout the torus. This prevents one area of the torus from becoming overheated. Answers B and C are each thermodynamically accurate, but are not the basis for alternating the EMRVs. Answer A is wrong since alternating the EMWs may actually increase the probability of having one stick open.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Examination 0utIine Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

KIA # 239002/2.4.6 Importance Rating 3. I -

KIA Topic

Description:

Knowledge symptom based EOP mitigation strategies Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #46 Following a Loss of Offsite Power and Reactor SCRAM from 100% power the following conditions exist:

0 All control rods are still at 0 Reactor water level at 0 inches.

0 Reactor power is currently less than 50%.

0 All other systems function as designed 0

All appropriate operator actions have been completed.

Answer the following:

a) How will ADS system operation affect plant conditions over the next 10 minutes?

b) What action is taken to mitigate these affects, if any?

A. ADS will not initiate since EOPs require it is inhibiting. Reactor power will remain stable.

Injection from CRD flow will continue. An additional CRD pump may be started to maintain level at 0 inches.

B. ADS will not initiate since the drywell coolers will preclude automatic ADS from Hi Hi Drywell pressure. Reactor power will slowly decrease with level decrease. Emergency RPV Depressurization is required.

C. ADS will automatically initiate and this will allow ECCS to rapidly refill the reactor.

Reactor power will increase above 100%. The ADS timers must be bypassed before ADS occurs.

D. ADS will automatically initiate but ECCS will be unable to refill the reactor due to cavitation of the Core Spray pumps. Alternate injection systems must be initiated to restore reactor water level.

ANSWER: A EXPLANATION:

The operators should have transitioned to RPV Control- with A i w S . The second step in the Levelpower leg directs the operator to Prevent initiations and isolations... ADS is bypassed to prevent uncontrolled introduction of water which would result in a severe power excursion.

TECHNICAL REFERENCE(S): EOPs (Attach if not previously provided)

Proposed references t o be provided t o applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # -

KIA # 218000/A2.06 Importance Rating 4.2 -

KIA Topic

Description:

Ability to predict the impact of ADS initiation signals present on the ADS system and based on those predictions to correct or mitigate the consequences of those abnormal conditions.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #47 Given the following conditions:

Immediately following a loss of all offsite power you are the reactor operator and observe one control rod at position 48 with the remaining control rods at 00.

0 Ten seconds later both emergency buses are energized from Diesel Generators (EDGs).

None of the LPRM downscale lights are lit.

When IRMs are selected the recorder readings do not change.

Are these observations expected or unexpected? Provide the basis for your answer A. Unexpected condition since LPRMsIAPRMs are powered from EDGs.

B. Expected condition since an ATWS has occured and reactor is not shutdown.

C. Unexpected condition since IRMs are also powered from EDGs D. Expected condition since RPS buses are deenergized during LOOP.

ANSWER: D EXPLANATION: LPRMsIAPRMs (and associated downscale lights) are powered from RPS. The RPS EPA relays will trip the RPS buses on low voltage. The APRMs and IRMs will lose signal and be downscale. The APRMARM recorders will, maintain power and read downscale on either range (IRM/APRM). With only one rod not full in the reactor is shutdown.

TECHNICAL REFERENCE(S): Neutron Monitorina Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # -2 -

Group # 1 -

WA # 215005/K6.01 Importance Rating 3.7 -

KIA Topic

Description:

Knowledge of the effect that a loss or malfunction of the RPS will have on the APRM/LPRM Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Information on recorder power obtained from OC training.

QUESTION #48 The reactor has just scrammed from MSlV closure following > 100 days at full power operation The foillowing conditions exist:

Both ICs have auto initiated and are operating.

All EMRVs are operable.

A, B and E reactor recirc pumps have tripped.

How will these conditions affect the reactor vessel design limits during the first hour following the scram assuming no operator action?

The RPV design limit for...

A. Maximum pressure will be exceeded B. Cooldown rate will be exceeded.

C. Minimum Pressurization Temperature will be exceeded.

D. Differential temperature will be exceeded.

ANSWER: B EXPLANATlON:

The full capacity of both ICs is adequate to exceed 100 degrees F per hour (estimate> 200F/hr) during the first hour. However maximum design pressure and Minimum Pressurization Temperature are not exceeded. Since the reactor was at 100% power there is at least one reactor recirc pump running and this will preclude exceeding any differential temperature limits.

Minimum temperature attainable with ICs is 212 degrees F. At this temperature a pressure of 200 psig is allowed by minimum pressurization curves.

TECHNICAL REFERENCE(S): IC Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 207000/K1.01 Importance Rating 3.8 -

KIA Topic

Description:

Cause/effect relationship between ICs and Reactor Vessel.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New x Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # 49 Startup of the rotary inverter has been accomplished per Procedure 339, Vital Power System.

The following conditions exist:

0 The start selector is in the AUTO-RUN position 0

The amber AC DRIVE ON indicating light is energized The AC generators field has been flashed Which of the following is an indication that the rotary inverter is malfunctioning?

A. Generator frequency is 60 Hz B. AC generator output voltage is 208 V C. DC DRIVE ON indicating light is energized D. AC DRIVE ON indicating light is energized ANSWER: C EXPLANATION:

The DC drive is not energized during a normal start (AUTO-RUN).

TECHNICAL REFERENCE(S): Procedure 339, Section 11.O (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

WA # 262002/2.1.30 Importance Rating 3.9 -

KIA Topic

Description:

Ability to locate and operate [UPS] components including local control Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #50 The reactor is at 100% power with the B CRD pump out of service for maintenance and cant be restarted for at least one hour. The following alarms come in at Annunciator Panel Hover a period of approximately 5 minutes:

0 PUMP A OL 0 CHARG WTR PRESS LO 0 ACCUMULATOR PRESS LO/LEVEL HI What are your actions in response to these alarms?

A. Immediately SCRAM the reactor in accordance with ABN-3200.01 B. Insert rods having an accumulator alarm as specified in OPS-3024.08 C. Immediately restart the A CRD Pump as specified in OPS-3024.08.

D. SCRAM per ABN-3200.01 when two or more rods have accumulator alarms.

ANSWER: D EXPLANATION:

This is specified in the CHARG WTR PRESS LO annunciator procedure. It is not necessary to immediately SCRAM the reactor. It is not appropriate to restart a CRD pump with an OL alarm until the pump is checked TECHNICAL REFERENCE(S): Annunciator procedures for Panel H (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # G -

Group # 4 -

K/A # 2.4.50 Importance Rating 3.3 KIA Topic

Description:

Ability to verify system alarm setpoints and operate controls-identified in the alarm response manual Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New x Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #51 The plant is at 20% power with reactor pressure in its normal band. Per Procedure 201 the EPR is set at I010 psig. The turbine trips and all RPS functions actuate correctly. What is the reactor pressure and power trend after steady state is achieved after the turbine trip?

A. Pressure and power are increasing B. Pressure and power are decreasing C. Pressure and power are stable D. Pressure and power are oscillating ANSWER: C EXPLANATION:

The reactor will not scram with power ~ 4 0 % .Although pressure and power may, initially, increase following the turbine trip once steady state is achieved the bypass valves will take control and maintain a stable pressure and power level.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # -

WA # 295005/AK1.01 Importance Rating 4.0 -

WA Topic

Description:

Knowledge of the operational implications of the pressure effects on reactor power Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments: Pressure and power will initially increase as the EPR opens the Bypass Valves, but should stabilize within a minute or so.

QUESTION #52 Following a SCRAM the Reactor Power is > 2% and at least one EMRV is cycling. Which of the following is the reason for terminating and preventing all injection into the RPV except Boron Systems and CRD during this condition.

A. Promotes a lower reactor water level which, in turn, will reduce reactor power through increased voiding B. Prevents thermal stressing the fuel cladding, thereby preventing fuel damage.

C. Prevents diluting SLC,allowing it to more effectively insert negative reactivity into the core.

D. Promotes a more uniform reactor moderator temperature which will prevent power oscillations from occurring ANSWER: A EXPLANATION:

The increased void fraction adds negative reactivity to the reactor. Other distractors are factual statements but are not applied to the terminating and preventing injection step of EMRG-3200.01B.

TECHNICAL REFERENCE(S): EOP Bases Daae 1B-19 and 1B-20.

(Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # 1 -

WA # 295037/EK1.02 Importance Rating 4.4 -

KIA Topic

Description:

Knowledge of the operational implications of reactor water level on reactor power as they apply to Scram Condition Present and Reactor Power Above APRM Downscale or Unknown.

Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #53 The following conditions exist:

A half scram exists on RPS A due to APRM testing A fire caused a loss of RPS Bus B and a full scram The testing was stopped and APRMs were returned to normal You are following ABN-3200.01, Reactor Scram Procedure and have been directed to reset the scram Per Step 3.1 1.1 the SDV HI-HI LVL SCRAM switch was placed in BYPASS When may the scram be reset?

A. After RPS Bus B is energized B. Immediately C. After the dump test switch is placed in Isolate.

D. After the SDV vent and drain valves are opened ANSWER: A EXPLANATION:

The only scram signal remaining after the APRMs are normal is to reset the B RPS bus TECHNICAL REFERENCE(S): ABN-32100.01 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

KIA # 212000/K3.05 Importance Rating 3.7 -

KIA Topic

Description:

Knowledge of the affect that a loss or malfunction of the Reactor Protection System will have on the RPS logic channels Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: This is a Pilgrim Bank question from 1998

QUESTION #54 The crew has entered EMG-3200.OlA RPV Control-No A N V S following a SCRAM and Loss of Offsite Power. The following conditions exist:

One IC failed to initiate.

0 Reactor pressure is >I 045 psig.

0 Entry conditions for Primary Containment Control have NOT been satisfied.

0 The Unit Supervisor is in the Pressure Control leg of EMG-3200.01A and has indicated he intends to use EMRVs to augment pressure control until the second IC can be manually initiated .

0 Unit Supervisor requests that you confirm Torus Level is above 90.

What is the bases for the 90 torus level?

A. This is the Torus Load Limit at 1045 psig reactor pressure B. This is the Heat Capacity Temperature Limit at 1045 psig reactor pressure C. This level is required to provide submergence of the EMRV tailpipes D. This level is required to ensure adequate NPSH for the Core Spray Pumps.

ANSWER: C EXPLANATION:

90 inches is the lowest torus level that will ensure submergence of the EMRV discharge and will ensure the steam being released is condensed in the torus. With entry conditions of Primary Containment Control not satisfied the remaining answers are incorrect.

TECHNICAL REFERENCE(S): EPG basis page 1A-53 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning Objective: (As available)

Examination 0utline Cross-reference: Level RO SRO Tier # 1 -

Group # 1 -

WA # 295030/EK2.08 Importance Rating 3.5 -

KIA Topic

Description:

Knowledge of the interrelations between Low Suppression Pool (Torus) Water Level and SRV discharge submergence.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X

55.43 Comments:

QUESTION # 55 Given the following conditions:

The reactor is at 100% power when a Large Break Loss of Coolant Accident coincident with a Loss of Offsite Power has occurred.

EDG 1 started but EDG 2 failed to start.

All ECCS equipment functioned as designed.

b No operator action has been taken other than verifying automatic actions.

What pumps are running 30 seconds after the LOCA?

A. ACore Spray Pump, ACore Spray Booster Pump, ACRD Pump B. B Core Spray Pump, B Core Spray Booster Pump, B CRD Pump C. ACore Spray Pump, ACore Spray Booster Pump D. B Core Spray Pump, B Core Spray Booster Pump ANSWER: C EXPLANATION:

Only Aequipment is powered from # I EDG. The CRD pump will not start for 60 seconds.

TECHNICAL REFERENCE(S): Procedure 341, Dacle 12 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning 0bjective: J01)00814 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 -

Group # 3 -

WA # 264000/K1.07 Importance Rating 2.9 -

KIA Topic

Description:

Knowledge of the physical connections and/or cause effect relationships between the Emergency Diesel Generators and the ECCS Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION # 56 Given the following conditions:

A LOCA has occurred.

The reactor is shutdown.

All isolations were successful ESW Pumps 52 A&B have failed to start 0 There are no entry conditions for the Secondary Containment Control Procedure.

The Unit Supervisor has directed you to use both Containment Spray Systems in the Drywell Spray mode since the drywell pressure has exceeded 12 psig.

You hear the STA advise the Unit Supervisor that the Heat Capacity Temperature Limit (HCTL) is predicted to be reached in 30 minutes.

If you continue to use Containment Spray in the current mode what operational implications, if any, will there be for the Containment Spray Pumps?

A. No operational implications since the Containment Spray Pumps are designed to operate in a Post-LOCA environment B. No operational implications since the action taken upon reaching the HCTL are intended to maintain the Containment Spray Pumps functional.

C. If containment pressure is reduced concurrent with elevated torus temperatures the Containment Spray Pumps may cavitate.

D. Containment Spray must be throttled as containment pressure is lowered to prevent pump runout.

ANSWER: C EXPLANATION:

This is the intention of the Caution in the Torus Water Temperature leg of EMG-3200.02 TECHNICAL REFERENCE(S): EMG-3200.02 Cautions (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs without entry conditions Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # -2 -

Group # 2 -

KIA # 219000/K5.02 Importance Rating 2.7 -

KIA Topic

Description:

Knowledge of the operational implications of Pump Cavitation as it applies to RHWLPCI (Containment Spray): Torus Cooling Mode Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X 55.43 Comments: The operator should know that he needs to transfer from the drywell spray mode to the torus cooling mode with the System illoop of Containment Spray to resolve the NPSH issue.

QUESTION #57 Given the following plant conditions:

Reactor Water Level at 86 TAF and decreasing Reactor Pressure 950 psig and stable for at least 100 seconds Drywell Pressure ~ 3 . 0psig and stable All Core SprayKOre Spray Booster Pumps function per design in response to ECCS Signals How many Core Spray/Core Spray Booster pumps are expected to be running for the stated conditions?

A. All Core Spray and Core Spray Booster Pumps even though Reactor Pressure is 950 psig.

B. None of the Core Spray or Core Spray Booster Pumps with Reactor Pressure at 950 psig.

C. Aand B Core Spray as well as Aand B Core Spray Booster Pumps even though

~ 3 . 0psig in the drywell.

D. C and D Core Spray as well as C and ID Core Spray Booster Pumps since ~ 3 . 0 psig in the drywell..

ANSWER: C EXPLANATION:

A&B pumps are priority and will start on Lo Lo water level even though drywell pressure remains<3.0 psig. C&D pumps will remain OFF as long as the priority pumps start.

TECHNICAL REFERENCE(S): Core Sprav Lesson Plan DQS 9-12 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning 0bjective: (04) 07283 (As available)

Examination OutIine Cross-reference: Level RO SRO Tier # -1 -

Group # 1 -

WA # 295031/EA1.03 Importance Rating 4.4 -

WA Topic

Description:

Ability to operate and/or monitor the Low Pressure Core Spray System as it applies to Reactor Low Water Level.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X

55.43 Comments:

QUESTION #58 In Procedure 120.4 General Response to Fires there is a precaution that reads: Care must be exercised in the use of water for fire fighting on Elevation 119 when the new fuel storage contains new fuel. What is the basis for this precaution?

A. Avoid stress corrosion cracking of the fuel cladding.

B. Avoid foreign material intrusion on the new fuel.

C. Avoid criticality of the fuel in the new storage vault.

D. Avoid stress corrosion cracking of the new fuel vault.

ANSWER: C EXPLANATION:

The new fuel vault has not been analyzed to remain subcritical with a water spray environment on new fuel. The other choices may be concerns but are not the basis for the precaution.

TECHNICAL REFERENCE(S): UFSAR 9.1.1.3 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # -1 -

Group # - 1 -

KIA # 6000000/AK3.04 Importance Rating 2.8 -

K/A Topic

Description:

Knowledge of the reasons for the actions contained in the abnormal procedure for Plant Fire on Site Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge x Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #59 The reactor is at 100% power with Advanced O f fGas (AOG) in service. The following abnormal conditions/alarms occur over the next 30 minutes:

a WTR REM MOIST HI alarm (IOXF) illuminates 0

STACK EFFLUENT HI alarm (1OF) illuminates The Radwaste Operator reports local readings of moisture and temperature into the first charcoal bed are increasing a Stack Activity is increasing What is causing these abnormal conditions/alarms A. Excessive flow from Steam Jet Air Ejectors (SJAE).

B. Blown loop seal on SJAE Condenser.

C. Malfunction in the AOG water removal system.

D. AOG Bypass (v-7-31) has opened.

ANSWER: C EXPLANATION:

The AOG charcoal relies on very dry air ( below 0 degrees F dew point) to function properly. If the water removal system malfunctions then moist air will contaminate the charcoal and reduce it ability to adsorb the krypton and xenon exhausted by the SJAE. As the charcoal becomes more contaminated with moisture the outlet activity (and Stack Activity) will increase.

TECHNICAL REFERENCE(S): AOG Lesson Plan PCI27 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: 30401 (01) 003 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 1 -

Group # 2 -

WA # 295017lAK2.03 Importance Rating 3.3 -

KIA Topic

Description:

Knowledge of the interrelations between High Off-Site Release Rate and the Off Gas system.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Moisture cLcontamination of AOG Charcoal will cause an increase in stack

gas activity. Therefore, the question relates High Off-Site release rate and the Off Gas System. Applicant must combine two facts: dew point affect on charcoal and relationship to increased stack gas release rate.1

QUESTION #60 A General Emergency has been declared at Oyster Creek. You have been designated by the Unit Supervisor to make off site notifications. What agency (agencies) must be notified within 15 minutes?

A. New Jersey State Police B. New Jersey State Police and Ocean County C. New Jersey State Police, Ocean County, Lacey Township, and Ocean Township D. New Jersey State Police, Ocean County, Lacey Township, Ocean Township and NRC ANSWER: C EXPLANATION:

Per Step 5.2 of EPIP-OC-03, Emergency Notification procedure TECHNICAL REFERENCE(S): EPIP-OC-03 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 3 -

Group # 4 -

KIA # 2.4.29 Importance Rating 3.3 -

K/A Topic

Description:

Knowledge of the emergency plan Question Source: Bank ##

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X 55.43 x Comments: RO is used as communicator until TSC is activated.

QUESTION #61 A plant startup is in progress using Procedure 201 Plant Startup. The following conditions exist:

0 The reactor pressure at 800 psig using the bypass valves for pressure control 0 Mode switch in STARTUP 0

IRM A is near the top of Range 9 0

All other IRMs are near the bottom of Range 9 What action, if any, is required, according to Procedure 201 to continue with a power increase?

I A. No action is required since a single IRM will not cause any trips.

B. Raise the pressure above 850 psig before going to Range 10 on any IRM.

C. Place the Mode switch in RUN before going to Range I O on any IRM.

D. Bypass A IRM and continue power increase until all IRMs are at top of Range 9.

ANSWER: B EXPLANATION:

Any IRM in Range 10 with mode switch in STARTUP and pressure below 850 psig will cause MSlVs to go closed. Although bypassing the A IRM will avoid the MSlV closure there is no direction in Procedure 201 to do so. Going to RUN with pressure below 850 will also cause MSlV closure.

TECHNICAL REFERENCE(S): Procedure 201, Precaution 4.9 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # -

Group # 1 -

WA # 2. I.23 Importance Rating 3.9 -

KIA Topic

Description:

Ability to perform specific system and integrated plant procedures during different modes of plant operation.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: See 55.41 (IO). .

QUESTION #62 You recall noting that each of the Squib valves in the Standby Liquid Control System has two firing coils as you attempt to Initiate System I by placing the keylock switch to FIRE SYS 1 position. When the SQUIBS light for System 1 does NOT come on you attempt to initiate System 2 but the PUMP ON light for System 2 is not on. You then return the keylock switch to FIRE SYS 1 position and note the PUMP ON light for System 1 is on.

When you look again to see the condition of the System 1 Squib valve, you would expect to see the following:

A. Continuity for System 1 is at llOO.

B. Squib light for System Iis now on.

C. Continuity for System 1 is at 60.

D. Squib light for System1 is still NOT on.

ANSWER: D EXPLANATION:

Although each Squib valve has two explosive circuits they are not fired from different circuits. Placing the keylock switch back in FIRE SYS 1 will not develop a second signal to fire the second circuit.

TECHNICAL REFERENCE(S): Procedure 304 DC~S8-12 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # -1 -

WA # 21 1000/K2.02 Importance Rating 3.1 -

WA Topic

Description:

Knowledge of electrical power supplies to the explosive valves Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: The stem involves two failures, SYS 1 Squib valve does not fire and SYS 2 pump does not start. This will give a total loss of function

QUESTION # 63 The plant is at 100% power with the following conditions:

It is a particularly hot Sunday in July and the switchgear room temperature is approaching 104 degrees F.

The Unit Supervisor directs you to place a portable fan in the switchgear room until engineering can evaluate the situation (on Monday).

What administrative requirements, if any, must be completed prior to operating the portable fan in the switchgear room.

A. Control and issue as a Special Condition Tag (SCT) per Procedure OP-MA-109-101, Clearance and Tagging B. Control and issue as a Temporary Procedure Change (TPC) per AD-AA-101 Processing of Procedures.

C. Control and issue as a Temporary Configuration Change Package (TCCP) per CC-AA-I12, Temporary Configuration Changes D. No administrative control necessary since plugging into 120 V circuits is an exclusion to TCCP requirements in Procedure CC-AA-112.

ANSWER: C EXPLANATION:

Must use a TCCP if the fan is used to replace or augment permanent plant HVAC equipment TECHNICAL REFERENCE(S): CA-AA-112, Attachment 2, pq 2 of 3 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # & -

Group # -2 -

WA # 2.2.1 1 Importance Rating 2.5 -

KIA Topic

Description:

Knowledge of the process for controlling temporary changes.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 x Comments: Applicant must combine two facts: fan represents a chance to HVAC and that this change requires a TCCP

QUESTION # 64 You are about to sign on to an RWP to operate a valve in the Radwaste Building and notice there is no requirement to wear a respirator. You recall from your tour this morning that the area you will enter is posted as a High Airborne Activity Area. You ask the HP tech about this and are told it is consistent with ALARA to not wear a respirator.

why would not wearing a respirator in a High Airborne Activity Area be consistent with ALARA?

A. The respirator is only effective on particulates so there will be no difference in TEDE.

B. The respirator will limit your vision and may be a safety hazard in "tight" spaces.

C. Wearing a respirator may increase your stay time and actually increase TEDE.

D. The Airborne Activity is short lived and the respirator will have no impact on ALARA.

ANSWER: C EXPLANAT1ON:

TEDE includes internal and external dose. If the HP calculates TEDE to be lower without a respirator then that is consistent with ALARA.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # G -

Group # - 3 -

WA # 2.3.2 Importance Rating 2.5 -

WA Topic

Description:

Knowledge of the facility ALARA program.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 x Comments: This goes beyond what is taught in GET.

QUESTION # 65 The reactor is at 100% power and a drywell entry is planned for tomorrow. Night orders specifies that your shift purge the primary containment with air using Procedure 312.9, "Primary Containment Control.

During the Unit Supervisor brief you are reminded to NOT open the Drywell and Torus valves together.

What is the basis for the restriction on opening both the torus and drywell valves simultaneously?

Having both sets of valves open simultaneously:

A. Does not satisfy the definition for Primary Containment Integrity.

B. Creates a pathway to bypass the Torus to Drywell Vacuum Breakers C. Is prohibited by Technical Specifications when in the RUN mode.

D. Will leave pockets of nitrogen in the lower elevations of the drywell.

ANSWER: B EXPLANATION:

This is specified in Precaution 7.2.6 of Procedure 312.9, "Primary Containment Control".

TECHNICAL REFERENCE(S): Precaution 7.2.6 of Procedure 312.9 (Attach if not prkiously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - -

Group # - -

WA # 2.3.9 Importance Rating 2.5 -

KIA Topic

Description:

Knowledge of the process for performing a containment purge Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

QUESTION #66 Given the following plant conditions:

e A plant startup is in progress per Procedure 201, Plant Startup e Reactor pressure is 200 psig e Reactor pressure control/heatup is utilizing MPR e Heatup rate is 10 degrees F each 10 minutes e Reactor Water Cleanup (RWCU) system isolates from faulty HELB relays.

What additional action would be necessary to continue the startup/heatup to the point of bringing the Turbine on line?

A. Repair the faulty HELB relays and re-start RWCU.

B. Obtain required reactor samples from A Recirc loop.

C. Adjust the heat balance to account for the lack of RWCU input.

D. Increase the frequency of recording vessel temperatures on Attachment 201-1 1.

ANSWER: B EXPLANATlON:

RWCU is not required to be in operation to support the continued startup, however periodic reactor samples must be taken. Chemistry will not change very quickly at this low power level.

Chemistry can be monitored via sample from AReactor Recirc Loop. Heat balance is not an issue with power level up to 25%. Vessel temperature monitoring is not affected by RWCU.

TECHNICAL REFERENCE(S): RWCU Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01) 00243 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 2 -

WA # 204000/K1.01 Importance Rating 3.1 -

K/A Topic

Description:

Knowledge of the physical connections and/or cause-effect between the Reactor Water Cleanup System and the Reactor Vessel Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments: Applicant must combine how isolation affects sample and where alternate

sample is obtained.

QUESTION # 67 A Station Blackout has occurred and preliminary information indicates it will be approximately two hours until any AC will be available from the Forked River CTs. The following is a table of voltage readings on the available DC buses:

DC Bus # At SBO SBO + 15 SBO + 30 SBO + 60 minutes minutes minutes 1 I 122v 120v I 115V 120v 122v Answer the following:

a) Upon re-energization of the Forked River CTs what 4160 V buses can be reenergized from the control room.

b)Why can they be reenergized?

Note: Assume that after the first minute the discharge rate is constant. 4160 V breaker closing coils are rated at 90 V.

A. No 4160 V buses can be reenergized since adequate DC voltage will not be available.

B. All 4160 V buses can be reenergized since DC is not required to close the breakers from the control room.

C. 4160 Buses A and C can be reenergized since the associated DC Bus voltage will provide adequate control power.

D. 4160 Buses B and D can be reenergized since the associated DC Bus voltage will provide adequate control power.

ANSWER:

C.

EXPLANATION:

4160 Buses A and C are supplied DC from Bus C which will be above 105 V. Even though DC Bus B is above 9OV, line losses will reduce this to less than 90 V at the closing coils on the associated 4160 Buses. With available coil voltage on 4160 V Bus B c 90 V the incoming breakers will not be able to be closed.

TECHNICAL REFERENCE(S): Loss of DC Distribution Center Procedures

- (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 263000/A4.03 Importance Rating 2.7 KIA Topic

Description:

Ability to manually operate and/or monitor the battery dischaze rate in the control room.

Question Source: Bank ##

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 X Comments:

QUESTION #68 The plant is in a startup at 30% power with the following conditions in the RBCCW system:

e Reactor Building Cooling Water (RBCCW) pump 1-1 is operating e RBCCW Pump 1-2 is in standby A Loss of Offsite Power (LOOP) occurs and the following plant conditions exist:

e Reactor water level remains between 150 and 160 e No entry conditions for EMG-3200.02, Primary Containment Control have been satisfied e Both diesel generators start and load as designed for the existing conditions e No manual operator actions are taken for the first five minutes e RBCCW breaker control switch in NORMAL e RBCCW system temperature is rising (Panel 13R)

Five minutes after the LOOP you are directed to complete appropriate sections of RBCCW Failure Response Procedure. What manual actions, on the RBCCW system, are required to maintain safe shutdown conditions?

A. Confirm RBCCW pressure (panel 13R) is greater than 75 psig (I pump operating)

B. Confirm RBCCW pressure (panel 13R) is greater than 90 psig (2 pumps operating)

C. Place RBCCW breaker control in BYPASS and manually start RBCCW pump 1-1 D. Place RBCCW breaker control in BYPASS and manually start RBCCW pump 1-1 and 1-2 ANSWER: B EXPLANATION:

With no LOCA signal both RBCCW pumps will auto start in 166 seconds. This is verified by observing >90 psig at panel 13R. As long as pressure is >90 psig no action in the control room is required. With both pumps running, transition to Shutdown Cooling can be made with no further changes to RBCCW, however the Shift Manager may elect to shutdown one RBCCW Pump.

TECHNICAL REFERENCE(S): ABN-19, RBCCW Lesson Plan DCI 7 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01)00068 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

KIA # 400000/A4.01

Importance Rating 3.1 -

KIA Topic

Description:

Ability to manually operate and/or monitor RBCCW indications and control in the control room.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: ABN-19 recently revised. OPS-3024.21 has been deleted. Ensure action to verify 90 psig is contained in a procedure (alarm response?) and note in Technical References

QUESTION # 69 The reactor is at 100% power with the following plant conditions:

Both Diesel Generators are OPERABLE Absorption chamber water volume is 81,000 cu ft b SGTS is OPERABLE b Reactor Building Differential Pressure is -0.20 water b

APLHGR is 91% of limits given in Specification 3.10.A 0

Which of the following systems/components are inoperable?

A. Primary Containment and Secondary Containment.

B. Primary Containment, Secondary Containment and Core Spray.

C. Primary Containment, Core Spray and Containment Spray.

D. Primary Containment, and Core Spray.

ANSWER: C EXPLANATION:

The ~82,000Cu Ft of water affects OPERABILITY of Primary Containment, Core Spray and Containment Spray. The APLHGR only affects the time the plant may remain running in this condition. Secondary Containment parameters satisfy TS requirements (-0.25 is not in definition for Secondary Containments).

TECHNICAL REFERENCE(S): TS 3.4 and 3.5 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # -

Group ## -1 -

K/A ## 2.1.12 Importance Rating 2.9 -

KIA Topic

Description:

Ability to apply Technical Specifications for a system Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 55.43 x Comments: ROs should know entry conditions for LCOs

QUESTION #70 The "B" Standby Gas Treatment (SGTS) train is near the end of a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run with the following conditions:

Reactor Building HVAC is shutdown.

You notice on the log sheets that the second absolute filter differential pressure (dip) has increased from 1.3"WG to 2.6"WG over the last ahours.

0 Total train d/p has increased from 4.3"WG to 5.6"WG. The Reactor Building differential pressure decreased from .30' to .25'.

What other change(@ , if any, in SGTS configuration and/or parameters would be observed locally?

A. None; a change of I.3"WG filter d/p alone would affect building differential pressure.

B. The condensate drain line is plugged and the prefilter is saturated with moisture.

C. The electric heating coils have failed and the charcoal is saturated with moisture.

D. The flowrate of the " B SGTS system (as read on the "A" panel) has decreased.

ANSWER: D EXPLANATION:

A doubling of filter differential pressure would have an effect on system flow. It is the decrease in flow that has caused the decrease in building DIP. Since only the d/p across the second absolute filter was effected, then moisture/condensate is not a problem.

TECHNICAL REFERENCE(S): Secondary Containment/SGTS Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01)10445 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 -

Group ## - 1 -

WA # 261OOO/AI.04 Importance Rating 3.0 -

KIA Topic

Description:

Ability to predict andlor monitor changes in parameters associated with operating the Standby Gas Treatment System controls including Secondary Containment differential pressure.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43

Comments: Per discussion with Greg Young the limit for filter DIP is 2.6". This is a fairly clogged state.

QUESTION #71 The following plant conditions exist:

The reactor power has just been increased to 40% power Turbine-Generator is on the line at approximately 200 MWE A malfunction causes a bypass valve to fully open FLOW MISMATCH alarm (J-7-a) annunciates shortly after the bypass valve (BPV) opens Answer the following:

a) Is FLOW MISMATCH an expected alarm for the stated conditions?

b) What is the operational significance of this alarm at 40% power?

A. No this is not expected. The significance is that a steam line break has occurred in the Turbine Building.

B. No this is not expected. The significance is that extraction steam has isolated from feedwater heaters.

C. Yes this is expected. The significance is that extraction steam has isolated from feedwater heaters.

D. Yes this is expected. The significance is that Turbine Anticipatory Scrams have been bypassed.

ANSWER: D EXPLANATION:

This is an expected alarm since BPVs will pass approximately 10% steam flow. The alarm is set at 7%. The steam going through the BPVs bypass first stage turbine and will not be counted as power for the 40% trip setpoint. With no other alarms in it should not be assumed that there may be a steam line break. Although the BPVs will have some impact on extraction steam flow it is not the reason the alarm is actuated.

TECHNICAL REFERENCE(S): Alarm Response J-7-a FLOW MISMATCH (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier ## 2 -

Group # - 2 -

KIA ## 239001/A1.09

~~

Importance Rating 3.5 -

KIA Topic

Description:

Ability to predict and/or monitor changes in parameters associated with the Main and Reheat Steam System controls including the Main Steam Flow.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Applicant must be able to relate at least two facts to obtain correct response:

What causes FLOW MISMATCH alarm and the fact that BPVs will bypass the turbine anticipatory trip. The FLOW MISMATCH alarm is usually attributed to a steam line break.

QUESTION #72 The reactor is at 100% power with the following conditions:

0 Three feed pumps are operating 0

Reactor water level control is in automatic set to maintain 1501 0 Two condensate pumps trip 0

Reactor Scrams on low water level 0

Lowest reactor water level observed is 128l 0

Highest post-scram level observed is 145" You follow the Operator Action in ABN-3200.01 , "Reactor Scram" and observe reactor water level has stabilized at 130" three minutes after the Scram.

Specify if this is an expected or unexpected water level and the basis for your answer.

A. Expected since the MFRV block valve is closed, per 3200.01 after reactor level restores I

above 140".

B. Unexpected since the water level should recover to the level of the feedwater level setpoint.

C. Expected since the feedwater level setpoint setback has automatically reduced the setpoint to 130" D. Unexpected since one feedwater pump would be running and that should provide adequate makeup through LFRV.

ANSWER: C EXPLANATION: The setpoint setdown is at 130' and would affect MFRVs and LFRV. All feed pumps should NOT trip on loss of 2 condensate pumps. Even if all feedpumps tripped, one would be restarted to restore level.

TECHNICAL REFERENCE(S): Feedwater Level Control Lesson Plan, Dase 12 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: 10453 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # J- -

KIA# . 259002/A3.06 Importance Rating 3.0 KIA Topic

Description:

Ability to monitor automatic operations of the Reactor waterGvel Control System including reactor water level setpoint setdown following a reactor scram Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION #73 The plant is at 100% power with the following conditions:

0 A Core Spray Loop is inoperable 0

The Reactor Building Operator calls you and indicates that a Tech Spec locked open valve on the B Core Spray Loop is missing the lock and he is unsure if the valve is open The Shift Manager is in the switchyard with the Security Supervisor investigating a possible tampering issue.

Your response to the information received from the Reactor Building Operator is as follows:

A. Have him confirm the valve is full open, to avoid a plant shutdown, then notify the Shift Manager and Security of possible tampering. Continue Power Operation.

B. Instruct the Reactor Building Operator to leave the valve as found, then notify the Shift Manager and Security of possible tampering. Start reactor shutdown.

C. Have him re-install the lock, then immediately notify the Shift Manager and Security. Start reactor shutdown.

D. Instruct the Reactor Building Operator to leave the valve as found, then notify the Shift Manager and Security of possible tampering. Continue Power Operation.

ANSWER: B EXPLANATION:

The position of the valve should not be changed to allow security investigation of the possible tampering. The RO has the responsibility to shutdown the plant if a situation exists which jeoparadizes the health and safety of the public. With the locked open valve position indeterminate the B Core Spray loop is inoperable. With both core spray loops inoperable the plant must be placed in the COLD CONDITION.

TECHNICAL REFERENCE(S): Conduct of Operations, m s 14. 15 and 16

- (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 3 -

Group # 1 -

WA # 2.1.2 Importance Rating 3.0 -

K/A Topic

Description:

Knowledge of operator responsibilities during all modes of operation Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X

Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: Licensee should confirm the stated conditions would prompt an RO to initiate a shutdown without specific direction from the Shift Manager.

QUESTION #74 The plant is at 100% Power with Service Water pump 1-1 out of service. The following alarms/indications occur almost simultaneously:

MN BRKR I D TRIP (T-I-e), BUS I D VOLTS LO (T-3-e), BUS I D VOLTS LO LO (T-4-e), MN BKR 1D 86 LKOUT TRIP(T-2-e).

Which equipment is available to ensure Safe Shutdown should a Loss of Offsite Power occur?

A) AIsolation Condenser, RBCCW Pump 1-1, CRD Pump NC08A, Shutdown Cooling Pump NU02B and Service Water Pump 1-2.

B) B Isolation Condenser, RBCCW Pump 1-1, CRD Pump NC08A, Shutdown Cooling Pump NU02B and ESW Pump 52B.

C) A Isolation Condenser, RBCCW Pump 1-1, CRD Pump NC08A, Shutdown Cooling Pump NUO2A and Service Water Pump 1-2.

D) B Isolation Condenser, RBCCW Pump 1-1, CRD Pump NC08A, Shutdown Cooling Pump NU02A and ESW Pump 52C.

ANSWER: B EXPLANATION:

All equipment on the 4160V D bus will be lost. Since Service Water Pump 1-1 is OOC, RBCCW will rely on ESW Pump 52B for a heat sink. Either Aor B Isolation Condenser is available to allow cooldown to the initiation of Shutdown Cooling. ESW Pump 52C is powered from the D 4160V bus.

TECHNICAL REFERENCE(S): OPS-3024.1OA. pas 37,38 and 44: RBCCW Lesson Plan PQ 9 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # 2 -

Group # 1 -

WA # 295003/A1.03 Importance Rating 4.4 -

KIA Topic

Description:

Ability to operate and/or monitor systems necessary to assure safe plant shutdown as they apply to partial or complete loss of AC power.

Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments: From INPO bank (Dresden exam)

QUESTION #75 Given the following configuration of the Reactor Recirculating system and reactor power at 75%:

0 Recirc pump A is shutdown and its suction valve and discharge bypass are closed 0

Recirc pump B is shutdown and its suction and discharge valves are open 0

No further valve movement on loop A or B is possible 0

Loops C, D, and E are operating normally What action, if any, must be taken to satisfy Technical Specifications (TS) for the conditions specified?

A. No action is required since the plant can operate up to 90% power during three loop operation.

B. APLHGR TS (3.1.A) must be verified to be at or below 98% of the APLHGR limits.

C. The circuit breaker for recirc pump B must be opened and defeated from operation.

D. Immediately initiate an orderly shutdown with mode switch in REFUEL or SHUTDOWN in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ANSWER: D EXPLANATION:

Per lesson plan page 9, an IDLE loop has the discharge valve closed and the discharge bypass and suction valves open. An ISOLATED loop has all three valves closed. The current configuration does not satisfy either and therefore TS 3.3.F.1 and 3.3.F.2 are not satisfied. Per TS 3.3.F.3, an orderly shutdown must be initiated immediately.

TECHNICAL REFERENCE(S): TS 3.3.F; Reactor Recirculation Lesson Plan

- (Attach if not previously provided)

Proposed references to be provided to applicants during examination: TS 3.3.F Learning Objective: (01) 10451 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 -

Group # 1 -

WA # 295001/2.1.33 Importance Rating 3.4 -

KIA Topic

Description:

Ability to recognize indications for (Reactor Recirculation) system operating parameters which are entry-level conditions for Technical Specifications Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X

I O CFR Part 55 Content: 55.41 55.43 x Comments: There is a specific RO Learning Objective for the Reactor Recirculation system to be able to utilize TS and be able to explain actions. This question was given a preliminary review by Greg Young with no objections raised.

QUESTION # SRO-1 The plant is operating at rated power on the 100% Rod Line with the following conditions:

0 A malfunction results in all Reactor Recirculation pumps running to minimum speed.

Core Flow is determined to be to the left of the buffer zone.

APRMs are stable.

Which one of the following identifies the required action?

A. Manually scram the reactor per ABN 3200.01 B. Raise recirculation flow or insert control rods.

C. Commence a normal reactor shutdown.

D. Confirm core flow using core plate differential pressure.

ANSWER: B EXPLANATION:

A reactor scram is only required if power oscillations are occurring. The procedure calls for exiting Exclusion Region by using rods or flow.

TECHNICAL REFERENCE(S): ABN-3200.03, Step 3.1.7 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group # - 1 KIA # 295001/AAZ.OI Importance Rating - 3.8 KIA Topic

Description:

Ability to determine and/or interpret the Power to Flow Map as it applies to Partial or Complete Loss of Forced Core Recirculation.

Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 55.43 (bI(5)

Comments: INPO Bank; NMP-I 01/18/98 simply changed to OC terminology Applicant must chose the correct procedure to follow for given conditions (ABN-3200.03). Considered memory since applicant only has to recall Step 3.1.7 to answer question

QUESTION # SRO-2 The reactor was operating at 94% at the end of cycle just prior to a refueling outage when a reactor scram occurred from a turbine trip. The following conditions exist:

Mode switch is in SHUTDOWN You initially enter RPV Control- No ATWS The scram has not been reset.

All but one of the control rods are at position 02 and the remaining rod has no numerical position indication Red SCRAM lights are illuminated on the full core display SCRAM solenoid lights are extinguished on panels 4F, 6R and 7R All LPRM downscale lights are lit SLC has NOT been initiated From the above indications what EOP is required to be implemented at this time. Also, provide a basis for this action.

A. RPV Control-No ATWS. With only one rod, potentially, not inserted beyond 02 the reactor will remain subcritical under all conditions.

B. RPV Control-No ATWS. Power level is below decay heat levels with no boron required to maintain the reactor subcritical.

C. RPV Control- With ATWS. At this point the last control rod could be at any position from 00 to 48 which is not at or beyond position 02 D. RPV Control- With ATWS. At this point it cannot be stated that the reactor will remain subcritical under all conditions.

ANSWER: D EXPLANATION:

Without green backlight on the last rod its position cannot be determined to be inserted at or beyond position 02. Additionally, the single rod out criteria is not satisfied for immediately determining the reactor will remain subcritical under all conditions. Procedure RPV Control-No ATWS requires you to exit and enter RPV Control- With ATWS.

TECHNICAL REFERENCE(S): Step 2.3.1 of ABN-3200.01, Reactor Manual Control Lesson Plan, page 3, RPV Control- With ATWS Procedure (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01) 00078 (As available)

Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1

K/A # 295006/AA2.02 Importance Rating - 4.4 K/A Topic

Description:

Ability to determine control rod position as it applies to SCRAM.

Question Source: Bank #

Modified Bank # X (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 55.43 (b)(5)

Comments: INPO Bank; Grand Gulf 03/27/98 changed stem and answer selection. Considered memory level since entry conditions are memorized.

QUESTION # SRO-3 A loss of all drywell cooling has occurred and you have entered Primary Containment Control, EMG-3200.02 when the drywell temperature entry conditions are exceeded.

The following conditions exist:

0 All attempts to restore drywell cooling have failed.

0 You enter RPV Control and re-enter EMG-3200.02 when drywell pressure exceeds the respective entry condition@).

0 When you direct the RO to vent the containment per support procedure 31 the STA notifies you that the drywell temperature is approaching 200 degrees F.

0 No other entry conditions for Primary Containment Control or RPV control exist at this point.

Answer the following:

Are you required to vent the containment? Also, provide a basis for this action.

A. Yes, reduction of drywell pressure is the most important strategy at this point.

B. Yes, venting the drywell will also result in reduction in drywell temperature.

C. No, venting the drywell will result in exceeding the Containment Spray Initiation Limit.

D. No, venting the drywell may cause a inadequate NPSH for the Containment Spray Pumps.

ANSWER: C EXPLANATION:

With drywell temperature above 200 degrees F any reduction in drywell pressure will result in exceeding CSIL. Once CSlL is exceeded the plant will, ultimately require EMERGENCY DEPRESSURIZATION since sprays cannot be initiated. The order should be changed to spray the containment before CSlL is exceeded. With no other entry conditions present, NPSH should not be a problem for Containment Spray Pumps.

TECHNICAL REFERENCE(S): EOPs (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier ## - - 1 Group # - 1 KIA ## 295024/EA2.02 Importance Rating - 4.0

K/A Topic

Description:

Ability to determine and/or interpret drywell temperature as it applies to High Drywell Pressure.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 55.43(b)(5)

Comments: Applicant must select the appropriate flow path in Containment Control. With a LOCA, drywell pressure will increase with temperature, however in the loss of drywell cooling with no LOCA temperature increase will be greater than pressure increase, hense venting the containment is NOT the right thing to do.

QUESTION # SRO-4 Following an Anticipated Transient Without Scram (ATWS) from 100% power, you transition to EMG-3200.01B RPV CONTROL-WITH ATWS. IF power level cannot be determined to be below 2%, this procedure directs you to lower RPV water level until reactor water level is below 30 in.

Answer the following:

a) How do you quickly determine power is <2%?

b) What is the EOP basis for the 30 inches of water level?

A. c 2% power will be indicated by LPRM Downscale lights lit. 30 inches reactor water level will minimize subcooling and power oscillations.

B. c 2% power will be indicated by LPRM Downscale lights lit. 30 inches reactor water level will minimize Keff of the core via neutron leakage.

C. c 2% power will be indicated by all EMRVs closed. 30 inches reactor water level will minimize chugging in EMRV tailpipes.

D. c 2% power will be indicated by IRMs below range 6. 30 inches reactor water level will ensure HCTL will not be exceeded if ED is needed.

ANSWER: A EXPLANATION:

The fastest way to determine <2% is using APRM downscale lights since other indicators may be slower responding or may be oscillating. IRMs will not be inserted at 100% power and their usefullness will be delayed by several minutes (until they are inserted). Each of the other answers are partially correct, but do not reflect the true bases for chosing 30. Although the remaining choices are identified in the EOP, only A provides the basis for lowering reactor water level to 30 inches. It is expected that neutron leakage would be the dominant factor at TAF (O),not at 30.

TECHNICAL REFERENCE(S): EOP Bases 1B-21 and 1B-22 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - I Group # - 1 K/A # 295031/EA2.02 Importance Rating - 4.2 KIA Topic

Description:

Ability to determine and/or interpret reactor power as it applies to Reactor Low Water Level Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X

55.43 55.43(bM Comments: Two independent facts do not require analysis to derive correct answer.

QUESTION # SRO-5 The Reactor is at 100% power when a partial loss of Circulating Water occurs. COND VAC LO 25 INCHES alarm illuminates. Vacuum is decreasing at 1" HG/minute. Describe the Operator Actions specified in ABN-3200-14, "Loss of Condenser Vacuum" necessary to preclude automatic opening of an EMRV.

A. Immediately SCRAM and control pressure by manually opening EMRVs in sequence.

B. Reduce Power to maintain vacuum =.22"HG. SCRAM before vacuum reaches 22" C. Immediately SCRAM and control pressure using Isolation Condensers (IC).

D. Reduce power to maintain vacuum >I 0"HG. Initiate IC before vacuum reaches 10".

ANSWER: B EXPLANATION:

As long as vacuum is maintained>22" all reactor power will be transmitted to the circulating water. If the turbine trips, an EMRV may, momentarily, open. Therefore, manual SCRAM should precede 22". There are 3 minutes available so that power reduction can be made before a SCRAM is necessary. Actions A, C and D although effective are not specified in the ABN.

TECHNICAL REFERENCE(S): ABN-3200.14, Section 3.0 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group # - 2 KIA # 295002/2.4.3 1 importance Rating - 3.4 KIA Topic

Description:

Knowledge of annunciators alarms and indicationdand use of the response instructions related to Loss of Main Condenser Vacuum Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 55.43(b)(5)

Comments: The alarm directs entry into ABN. Knowledge of ABN actions beyond "immediate Actions are an SRO-level function.

QUESTION # SRO-6 The reactor is at 100% power when the following conditions are reported by the control room operators:

e Reactor has scrammed on APRM Hi-Hi e SCRAM CONTACTOR OPEN (G-l-c) is illuminated e No other SCRAM annunciators are illuminated All rods are at 00 Reactor recirc pumps A, B and E have tripped LOGIC TRAIN I (C-1-a) and LOGIC TRAIN II (C-2-a) alarms are illuminated What condition initiated the high flux condition? Also, what is the expected plant response to the stated conditions?

A. High Reactor Pressure caused the high flux condition. C and D recirc pumps should have tripped but they did not.

B. High Reactor Pressure caused the high flux condition. Reactor should have scrammed on Hi RPV pressure but it did not.

C. Malfunction in the Recirc Flow Control caused the high flux condition. C and D recirc pumps should have tripped but they did not.

D. Malfunction in the Recirc Flow Control caused the high flux condition. Reactor should have scrammed on Loss of Flow signal but it did not.

ANSWER: B EXPLANATION:

With LOGIC TRAIN I (C-l-a) and LOGIC TRAIN II (C-2-a) alarms illuminated and Reactor recirc pumps A, B and E tripped this would indicate the reactor pressure had exceeded 1045 psig and should have scrammed on Hi reactor pressure, but it did not. Although initiation of the Isolation Condensers (LOGIC TRAIN I and LOGIC TRAIN II alarms are illuminated) is known to have caused a high flux scram in the past the specified conditions indicate a high pressure condition preceded the Isolation Condenser actuation. The increase in flux from the high pressure condition is 1.5 seconds faster than the initiation of ICs. There is no indication from the given conditions that the recirc flow control failed.

TECHNICAL REFERENCE(S): Reactor Scram Procedure: Alarm Responses (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group ## - 2 KIA # 295014iAA2.03

Importance Rating - 4.3 KIA Topic

Description:

Ability to determine and/or interpret the cause of reactivity addition as it applies to lnadvertant Reactivity Addition Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 55.43(bN6)

Comments: Considered COMP since this is equivalent to a partial post-trip analysis

QUESTION # SRO-7 It is a particularly cold January night. The Turbine Building Operator calls you up to let you know that the 4160 V switchgear room temperature is abnormally cool with a local room thermometer reading only about 40 degrees F.

What immediate action(s) are required?

A. You will initiate a Temporary Configuration Change Package (TCCP) and direct the operator to install a portable heater in the room.

B. You will initiate an Action Request to have the HVAC unit cleaned and/or adjusted as necessary to maintain normal room temperature.

C. You will, conservatively, declare the 4160 Switchgear Room Fire Suppression System inoperable and assign a continuous Fire Watch in the room.

D. You will initiate a plant shutdown to the COLD SHUTDOWN CONDITION while attempting to resolve any HVAC problems.

ANSWER: D EXPLANATION:

With a 4160 V room temperature below 50 degrees F the C battery must be declared inoperable. With the C battery inoperable, TS 3.7.B requires The reactor shall be placed in the COLD SHUTDOWN CONDITION.... Although the US may take other actions, the TS requirements must be initiated immediately and take precedence over other actions.

TECHNICAL REFERENCE(S): DC Distribution Lesson Plan page 4: TS pages 3.7-1 and 4.7-1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01) 10445 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group # - 1 WA # 295004/2.1.33 Importance Rating - 4.0 KIA Topic

Description:

Ability to recognize indications of DC system operating parameters which are entry level conditions for Technical Specifications.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 55.43 55.43(b)(2) and (3)

Comments: The applicant must combine at least three facts to obtain the right answer.

Other equipment may be affected but that is not germane to the question.

QUESTION SRO #8 Given the following conditions:

Plant is shutdown for a refueling outage.

Power to 4160V buses is being supplied from the Startup Transformers EDG #2 is out of service for PMs The I Diesel Fire Pump is out of service for PMs Pond Pump 1 is in manual supplying the fire water system Pond Pump 2 is in Automatic A fire develops in the BStartup transformer and the deluge system actuates The 2 Diesel Fire Pump fails to start.

A security guard reports that flames are engulfing the B Startup Transformer but that no water spray is visible In accordance with Procedure 333, Plant Fire Protection System what action must be taken to extinguish the fire at the B Startup Transformer within I O minutes?

A. Call the Forked River Fire Department and request backup fire protection (Le. fire PUmPer)

B. Manually start Pond Pump No. 2 per Section 4.0 of Procedure 333.

C. Power up 4160V Bus 1B from the SBO transformer and energize both Pond Pumps.

D. Place the Redundant Fire Protection Water System in service.

ANSWER: D EXPLANATION: Loss of the B Startup Transformer (in combination with EDG #2 out of service) will result in both Pond Pumps losing power. Although A & C could be done, the time-frame will be too long to be effective in extinguishing the fire within 10 minutes. The response of an off site volunteer fire department is likely to take 15 minutes just to get on site with the pumper.

Energizing the SBO transformer requires coordination with Jersey Central and is expected to take up to one hour. Placing the Redundant Fire Protection System in service (answer D) is the first action listed in Procedure 333 if both diesel fire pumps become inoperable TECHNICAL REFERENCE(S): Procedure 333 PU 48 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 Group # - 2 K/A # 286000/A2.03

Importance Rating - 3.0 KIA Topic

Description:

Ability to (a) predict the impacts of AC distribution failure and (b) based on those predictions use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 55.43 55.43(b)(5)

Comments:

QUESTION # SRO-9 A Plant Startup is in progress per procedure 201. The following condition exist:

0 Power is in the IRM range with reactor coolant temperature approaching 212 F.

0 The Reactor Operator is in control of rod pulls The Reactor Operator is controlling Reactor Water Level 0 The BOP Operator is preparing the secondary plant for introduction of steam.

0 ACRD Pump is running.

0 You observe RX LVL HVLO alarm initiated.

0 Water level on Panel 5F/6F is 145.

For this plant configuration, what actions are needed to clear the RX LVL HVLO alarm condition?

A. Start a second CRD pump and maximize flow B. Throttle down on Reactor Water Cleanup blowdown C. Reduce CRD pump flow to minimum.

D. Increase Reactor Water Cleanup blowdown ANSWER: B EXPLANATION:

In this configuration the Reactor Water is swelling from the heat addition and RWCU is being used to blowdown to the hotwell. CRD flow should already be at minimum. The most appropriate action is to reduce the blowdown flow. The addition of any outside water source would complicate level control until steaming to the secondary plant commences.

TECHNICAL REFERENCE(S): Reactor Water CleanuD Lesson Plan (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: 20401(01)007 (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group # - 2 KIA # 295009/AA2.03 Importance Rating - 2.9 KIA Topic

Description:

Ability to determine and/or interpret Reactor Water Cleanup blowdown rate as it applies to Low Reactor Water Level Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X

10 CFR Part 55 Content: 55.41 X 55.43 55.43(b)(5)

Comments:

QUESTION # SRO-I 0 The plant is at 100% power when a failure of RBCCW occurs. You direct the operators to complete the required actions in ABN-19, RBCCW Failure Response. The following sequence of events occurs:

One of the operators advises you he is venting the drywell through V-23-21 and V-23-22.

The second operator scrams the reactor and trips all Reactor Recirculation Pumps FLOW HVMN STM LINE AREA TEMP HI-HI alarm illuminates NU47 REACTOR BLDG EQUIP DRAIN TK HIGH TEMP alarm illuminates DW TEMP HI alarm illuminates DW PRESS HVLO alarm illuminates TORUS BULK TEMP HI-HI alarm annunciates The second operator advises you all rods are inserted at or beyond position 02.

Steam and Feedwater systems function as designed and as specified in Reactor Scram Procedure No other containment-related or reactor-related alarms are illuminated Assuming parameters remain at the alarms setpoints, what EOP(s) entry conditions, if any, are currently satisfied by the above sequence of events?

A. RPV CONTROL-NO A W S , SECONDARY CONTAINMENT CONTROL, PRIMARY CONTAINMENT CONTROL B. PRIMARY CONTAINMENT CONTROL C. SECONDARY CONTAINMENT CONTROL D. No EOP Entry conditions are currently satisfied by the above sequence ANSWER: B EXPLANATION:

With all rods in beyond 02 and steam/feedwater responding as designed level should be above 138 so no entry conditions in RPV Control exist. Drywell TEMP and Press alarms are below the entry conditions for Containment Control. The FLOW HVMN STM LINE AREA TEMP HI-HI alarm and NU-47 REACTOR BLDG EQUIP DRAIN TK HIGH TEMP alarm are not entry conditions for Secondary Containment Control. The setpoint for TORUS BULK TEMP HI-HI is 109 degrees F which is above the entry condition for Primary Containment Control, therefore B is the correct answer.

TECHNICA L REFERENCE(S) : EOP,s Alarm Responces (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOPs Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group ## - 2

WA # 295012/2.4.4 Importance Rating - 4.3 KIA Topic

Description:

Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for Emergency and Abnormal Operating Procedures [Related to High Drywell Temperature]

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 X 55.43 55.43(bM)

Comments: Considered COMP since the applicant will have to consider all the stated conditions individually and in concert to determine that he does not have any EOP entry condition satisfied. Requires the knowledge of setpoints for the stated alarms. Although none of the above is usually avoided in this case it makes the question more challenging.

QUESTION # SRO-I 1 A reactor startup is in progress using Procedure 201, Plant Startup. The following plant conditions exist:

Power has risen on IRM 18 from I O / 40 on Range 3 to 40/125 on Range 4 in 40 seconds No rod motion is in progress ROD BLOCKS come in on all remaining IRMs (which are still on Range 3)

Based on the above conditions you (as US) should direct the RO to:

A. Uprange the remaining IRMs to clear ROD BLOCKS so rod withdrawal can resume B. Insert rods to reduce power below the ROD BLOCK setpoint for the remaining IRMs C. Uprange remaining IRMs to Range 4; allow power rise to continue at the current rate.

D. Uprange remaining IRMs to Range 4, then Insert rods to slow down the power rise ANSWER: D EXPLANATION: The reactor period should be calculated at approximately 29 seconds which is shorter that specified in Precaution/Limitation 4.1 of Procedure 201 and must be slowed down by rod insertion, however the IRMs should first be upranged or a SCRAM will occur shortlly. The period is already too short so rod pulls are not in order (A). Since a startup is in progress it is not intended to reduce power (B).

TECHNICAL REFERENCE(S): Procedure 201, Precaution 4.1 and pages 19 and 20 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 Group # - 1 WA # 215003/A2.04 Importance Rating - 3.8 KIA Topic

Description:

Ability to (a) predict the impacts of Upscale or downscale trips on the IRM and (b) based on those predictions use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations.

Question Source: Bank #

Modified Bank # X (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 55.43(b)(5)

Comments: From INPO bank Dresden 2 02/15/2001. Modified due to different WA (2.1.7)

QUESTION # SRO-12 At noon on April 1, 2004 the plant is at 80% power with three reactor recirc pumps operating (NGOI-A, C and E). No LCOs are in effect at this time. At 12:05 PM the following conditions occur on the AC distribution system:

a The following alarms annunciate: MN BRKR 1B TRIP, MN BRKR 1B 86 LKOUT TRIP, BUS 1B UV, S I B BRKR TRIP, S I B BRKR OL TRIP/BRKR PERM OPN 4160V BUS IB voltmeter is reading downscale a 4160V BUS IA voltmeter is reading 4160 volts EDG No. 2 has started and has energized 4160V Bus 1D Security reports that Startup Transformer SB deluge system is discharging on the transformer.

All other switchyard equipment is available for use.

The operators quickly respond to the 1B Bus alarms and indications (using OPS-3024.lOa) and stabilize the plant within the design capability of the remaining energized systems and components. All applicable Technical Specification ACTION statements are satisfied.

Answer the following:

a. What is the maximum power level sustainable with the AC distribution configuration as it exists at 12:05 PM.

b. How long can the conditions existing at 12:05 PM be allowed to continue.

A. The plant would have scrammed from the transient. The existing conditions can be maintained indefinitely.

B. The plant could be run at approximately 33% power. The existing conditions can be maintained for 7 days.

C. The plant could be run at approximately 50% power. The existing conditions can be maintained for 7 days.

D. The plant could be run at approximately 33% power. The reactor must be placed in the COLD SHUTDOWN CONDITION.

ANSWER: B EXPLANATION: The limiting configuration with Bus IB deenergized is the condensate/feedwater system which will have only one condensate and one feedwater pump running. The remaining 4160V equipment (fed from Bus ?A) will sustain over 50% power. A half SCRAM will occur, but it can be reset after the #2 EDG loaded onto Bus 1D. It is expected that Bus IB would be energized from the Station Blackout Transformer within about one hour. TS allows operation in this configuration for 7 days.

TECHNICAL REFERENCE(S): OPS-3024.1 Oa; TS 3.7 (Attach if not pr&iously provided)

Proposed references to be provided to applicants during examination: None Learning 0bjective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 Group # - 1 KIA # 262001/2.1.7 Importance Rating - 4.4 KIA Topic

Description:

Ability to evaluate plant performance and make operational judgements related to AC Electrical Distribution based on operating characteristics/reactor behavior/ and instrument interpretation.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 55.43 55.43(b)(5)and (2)

Comments: Sustainable power level must be validated by licensee.

QUESTION # SRO-I 3 The following plant conditions exist:

0 Plant is at 87% power 0

1. I Emergency Diesel Generator (EDG) is out of service for PMs 0

1. 2 EDG is being run to satisfy surveillance requirements 0

1. 2 EDG Crankcase Pressure readings for the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> have increased steadily 0

The System Engineer indicates the engine will have to be shutdown prematurely and the engine declared inoperable if the crankcase pressure reaches the trip setpoint.

0 The Maintenance Engineer indicates he can maintain the EDG Operable by adjusting the crankcase pressure setpoint. He indicates it must not be important since it is bypassed during a Fast Start.

Shortly before the end of your shift a 5% setpoint change (increase) and 50.59 screen is approved and the night orders indicate the EDG should continue to run until the surveillance requirement is satisfied.

Based on your understanding of EDG operation, is continued operation of the #2 EDG permissable? What is the basis for this action?

A. Yes. The Maintenance Engineer is correct regarding the trip being bypassed during a fast start. Continued operation will have no deleterious effects on the EDG.

B. Yes. A 5% increase in crankcase pressure is well within the design pressure of the crankcase (designed for 110% above atmospheric pressure).

C. No. A 5% increase in crankcase pressure is at the design pressure of the crankcase (designed for 105% above atmospheric pressure).

D. No. Increasing crankcase pressure is indicative of a serious mechanical failure in the engine. The engine must be shutdown before reaching a positive pressure condition.

ANSWER: D EXPLANATION: Increasing crankcase pressure is indicative of excessive piston or piston ring wear. If the crankcase pressure reaches above atmospheric it is indicative of impending catatrophic failure of one or more pistonshings.

TECHNICAL REFERENCE(S): EDG Lesson Plan Dane 12 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (01)00811, (01) 00791 (As available)

Examination 0utI ine Cross-reference: Level RO SRO Tier # - 2 Group # - 1 KIA # 26400012.4.48 Importance Rating - 3.8

KIA Topic

Description:

Ability to interpret control room indications to verify the status and operation of system/and understand how operator actions and directives affect plant and system conditions [EDGs]

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 55.43 55.43(b)(5)

Comments: This replicates an actual event at a plant I worked at. When the piston failed they could hear the bang over a mile away.

QUESTION # SRO-14 Core loading has started in the North East Quadrant with the following conditions:

0 SRM 24 is reading ICPS 0

All other SRMs are operable and reading >2 CPS Repair of the drywell equipment drain sump pump is in progress 0

One of the criticality monitors reads downscale and is declared inoperable What action must be taken in order to continue fuel movement into the North East Quadrant?

A. Calibrate SRM 24 in accordance with Procedure 620.3.006 B. Evacuate all personnel from the drywell C. Complete Procedure 205.62 Refueling Bridge Check-Off D. Calibrate SRMs 21, 22 and 23 in accordance with Procedure 620.3.006 ANSWER: B EXPLANAT1ON:

The only requirement for the stated conditions is to evacuate personnel from the drywell. The 21 and 22 SRMs are all that is required to be operable to move fuel in the North East Quadrant TECHNICAL REFERENCE(S): Procedure 205, Reactor Refueling, pages 7, 8 & E5-1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning 0bjective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - G Group # - 2 WA # 2.2.26 Importance Rating - 3.7 WA Topic

Description:

Knowledge of refueling administrative requirements.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 55.43 55.43(b)(6)

Comments: The applicant must relate two facts: 24 SRM is not required in NE quadrant and that the criticality monitor relates to personnel radiation protection, not reactor reactivity.

QUESTION # SRO-15 The following conditions exist:

The reactor is shutdown; mode switch is in REFUEL.

A plant startup is planned for later today.

The ASLC has been inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to replace a defective SQUIB trigger assembly The Liquid Poison System Functional test described in Procedure 612.4.002, Liquid Poison System Functional Test has been performed within the last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

0 The trigger assembly for V-19-0044 has been replaced.

Post maintenance testing of the SLC system consisted of completing Attachment 2400-SMM-3209.02-7, Liquid Poison System Explosive Valve Maintenance.

A copy of the completed attachment is attached 0

Except for ASLC all LCOs and surveillance requirements are satisfied Based on the maintenance performed what additional maintenance and/or post-maintenance testing, if any, is necessary to ensure Operability of A SLC?

A. Re-perform The Liquid Poison System Functional test described in Procedure 612.4.002, Liquid Poison System Functional Test.

B. Perform the Standby Liquid Control Pump and Valve Operability and In-Service Test, Procedure 612.4.001.

C. Replace the V-19-0044 with a new valve and re-perform Attachment 2400-SMM-3209.02-7 with the new assembly.

D. No additional maintenance or testing is required prior to declaring the A SLC Operable.

ANSWER: C EXPLANATION: With the circuit readings documented on Attachment2400-SMM-3209.02-7 and given the Bridgewire circuit configuration denoted on page E6-1 the SRO should conclude the trigger assembly is in an incorrect configuration and needs to be replaced before the ASLC can be considered Operable. The remaining distractors would not be required if the circuit test was OK. The remaining distractors would not reveal the fact that an incorrect valve configuration existed.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: SMM-3219.02 ~q 3.0 and pa E6-I Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 3 Group # - 2 K/A # 2.2.21

Importance Rating - 3.5 KIA Topic

Description:

Knowledge of pre and post maintenance operability requirements Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 55.43 55.43(b)(5)&(3)

Comments: This actually happened at another BWR in the 1980's and the plant ran for an entire cycle with two non-functioning squib valves installed.

QUESTION #SRO-I 6 The following plant conditions exist:

0 Reactor has scrammed from 100% power due to Lo Lo Reactor Water Level 0

You have entered Primary Containment Control based on Drywell Pressure 0

You re-enter Primary Containment Control based on Drywell Temperature 0

Reactor pressure is being controlled with EMRVs 0

The only available high pressure water source is A CRD pump 0

Drywell temperature has increased to 350 degrees F 0

GEMAC Wide Range Level indication is 65 inches and slowly increasing 0

A and B Fuel Zone instruments have failed due to power supply problems 0

C and D Fuel Zone instruments are full scale high The STA recommends exiting RPV Control and entering RPV Flooding.

Answer the following:

a) Do you agree with the STAs recommendation?

b) What is the basis for your position?

A. No. Indicated level is 65 inches and as long as level is being maintained above 61 inches we stay in RPV Control.

B. No. We must enter Emergency Depressurization because level cannot be maintained above 61 inches.

C. Yes. Flooding is the only way we can decrease drywell temperature since we have exceeded the CSlL D. Yes. With the current drywell temperature and level instrument readings we have NO reliable indication of reactor water level.

ANSWER: D EXPLANATION:

TECHNICAL REFERENCE(S): EOP Users Guide Dalqes 2-21 and 2-22 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EOP Flow Charts Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group # - 1 K/A # 295028/2.4.20 Importance Rating - 4.0 KIA Topic

Description:

Knowledge of operational implications of EOP warnings/cautions and notes related to High Drywell Temperature Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 55.43(b)(5)

Comments: Not a direct look up. Must understand temperatureAeve1 interactions to be able to answer.

QUESTION # SRO-I 7 At the end of a refueling outage following core re-load you are making a Shift Manager tour of the refuel floor and observe one central core cell is not symmetric with adjacent core cells. It is not clear what is different, but it doesnt look like the cells surrounding it. You recall that Reactor Engineering has verified the final fuel cell configuration in the core.

In accordance with Procedure 205, Reactor Refueling you immediately notify Reactor Engineering of the apparent discrepancy.

What is the operational significance, if any, of the symmetry issue you observed.

A. No operational significance since cell symmetry is only important during movement of fuel in the outage.

B. Operationally significant since a misoriented bundle could prevent proper seating of the steam separator.

C. Operationally significant since a misoriented fuel bundle can affect CRD scram times and Core Operational Limits.

D. Operationally significant since a mis-oriented bundle can affect total core flow and core differential pressure.

ANSWER: C EXPLANATION:

Regardless of core verification responsibility, SROs must be aware of bundle orientation requirements and ensure core loading has been done correctly. Because of the different pin loadings (% U235lburnable poison, etc.) Orientation can have significant impact on power operation. With the fuel assembly buttons misoriented there may be some interaction with the CRD. However, mis-orientation will have no effect on core flow or dp.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 Group # - 2 WA # 290002/2.2.32 Importance Rating - 3.3 KIA Topic

Description:

Knowledge of the effects of alterations on core configuration as it relates to Reactor Vessel lnternals Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 55.43 55.43(b)(7)

Comments:

QUESTION #SRO-18 As stated, in part, in TS 3.7 Bases The general objective is to assure an adequate supply of power with at least one active and one standby source of power available for operation of equipment required for safe shutdown, to maintain the plant in a safe shutdown condition and to operate the required engineered safety feature following an accident.

The plant is at 100% power with the following conditions:

Two 230 KV lines are fully operational One 34.5 KV line is fully operational Two Diesel Generators are OPERABLE All in-plant AC buses and panels are energized All other incoming lines are inoperable A lightning hit takes out the fully operational 34.5 KV line but has no effect on either 230 KV line All in-plant AC buses and panels remain energized Can reactor operation continue with this AC configuration? Provide a basis for your answer.

A. Yes, operation may continue since the two 230 KV lines provide an active and standby source of power that satisfies the general objective B. Yes, operation may continue since the two 230 KV lines plus the two diesel generators provide at least one active and one standby source of power.

C. No, operation may not continue since the two 230 KV lines are on the same set of towers and count as only one source of power.

D. No, operation may not continue since one of the two 230 KV lines does not have adequate capacity for all equipment required for safe shutdown ANSWER: C EXPLANATION: This is explicitely stated in the second paragraph of the TS Bases. The EDGs are a separate TS and not counted as an active or standby AC Source.

TECHNICAL REFERENCE(S): TS 3.7 Bases (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 2 Group ## - 1 WA # 26200112.2.25 Importance Rating - 3.7 KIA Topic

Description:

Knowledge of bases in Technical Specifications for limiting

conditions of operations and Safety Limits related to AC Electrical Distribution Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 55.43 55.43(b)(2)

Comments:

QUESTION #SRO-19 The following conditions exist on the refueling floor:

Refueling is in progress in the reactor 0

A fuel bundle has been grappled in the core and grapple engagement has been visibly verified.

ENGAGED light is on.

The bundle is being lifted out of the core "SLACK CABLE" light on the refuel bridge is extinguished Just as the fuel bundle clears the upper grid a failure in the air system directs full system air pressure to the grapple and bubbles can be seen rising from the grapple head.

What design feature@) prevent a refueling accident under these conditions?

A. The grapple uses air pressure to grapple the bundle. Full air pressure just engages the grapple tighter and precludes release.

B. Refueling interlocks prevent disengaging the bundle until a slack cable is sensed by a load cell on the refueling bridge.

C. The grapple has a mechanical "hook" that prevents the grapple from disengaging the bundle as long as weight is on the grapple.

D. The Boundary Zone Computer "Fuel Hoist Interlock" prevent disengaging the bundle if the hoist position is not seated in a storage location.

ANSWER: C EXPLANATION:

There are no electrical interlocks to prevent a refueling accident if a failure occurs in the air system, however the mechanical "hook" on the grapple prevents disengagement once weight is on the grapple.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group # - 1 WA # 29501912.2.27 Importance Rating - 3.5 KIA Topic

Description:

Knowledge of the refueling process [as it relates to Partial or Total loss of Inst. Air]

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge

Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 55.43 55.43(b)(7)

Comments: SRO needs to know that grapple depressurizes to engage as well as the fact that no refueling interlocks are applicable to the stated conditions.

QUESTION # SRO-20 One of the license conditions for Oyster Creek is that the fire protection program be maintained.

One of the issues that must be addressed by the fire protection program is hot shorts.

Answer the following regarding the fire protection program at OC:

a. What is a hot short?

b. As described in ProcedureABN-3200.29 Response to Fire what action must be taken in the control room to protect safe shutdown during a fire on R.B. 51 Elevation (RB-FZ-ID) as described in Special Instruction # I A. A hot short is a condition where molten debris causes the + and - cables of a particular circuit to fuse. For a fire on El 51 the A Isolation Condenser valves must be overridden by placing the individual control switches to the desired position.

B. A hot short condition is where insulation on cabling is destroyed and an external source of 125 V DC causes electrical equipment to spuriously operate. For a fire on the 51 elevation the disable switch for one or more EMRV must be placed in DISABLE.

C. A hot short condition is where insulation on cabling is destroyed and an external source of 125 V DC causes electrical equipment to spuriously operate. For a fire on El 51 the AIsolation Condenser valves must be overridden by placing the individual control switches to the desired position.

D. A hot short is a condition where molten debris causes the + and - cables of a particular circuit to fuse. For a fire on the 51 elevation the disable switch for one or more EMRV must be placed in DISABLE.

ANSWER: B EXPLANATION:

This is described in Special Instruction No. I.The A IC is not affected by a fire on el 51 TECHNICAL REFERENCE(S): ABN-3200.29, Dq E1-5 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 1 Group # - I WA # 600000/2.4.27 Importance Rating - 3.5 KIA Topic

Description:

Knowledge of fire in the plant procedure related to Plant Fire on Site Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 55.43 55.43(bM)

Comments:

QUESTION #SRO-21 The plant is at 100% power and you are notified that the Sewage Lift Pump High Radiation Alarm has been received at the SAS.

Answer the following with regard to the Sewage Lift Pump High Radiation Alarm:

a. What action is required by chemistry personnel

b. When would you implement IMP-I 300.01 Classification of Emergency Conditions A. Chemistry will obtain a sample of the sewage at the sewage collection pit pumps and confirm the sewage collection pit pumps trip at 850 CPM indicated. Implement IMP-1300.01 if sample readings are 10% above background.

B. Chemistry will verify the validity of the high radiation signal using equipment in RAGEMS building before obtaining grab sample. Implement IMP-1300.01 if grab sample is >IO%

above IOCFR20 limits.

C. Chemistry will obtain a sample of the sewage at the sewage collection pit pumps and confirm the sewage collection pit pumps trip at 850 CPM indicated. Implement IMP-1300.01 if sample readings are 20% above background D. Chemistry will verify the validity of the high radiation signal using equipment in RAGEMS building before obtaining grab sample. Implement IMP-1300.01 if grab sample is >20%

above 10CFR20 limits.

ANSWER: D EXPLANATION:

This is specified in ABN-3200.28.

TECHNICAL REFERENCE(S): ABN-3200.28,~ns4 and 7 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 3 Group # - 4 WA # 2.4.36 Importance Rating - 2.8 KIA Topic

Description:

Knowledge of chemistry/health physics tasks during emergency operations Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 55.43-55.43 (b)(5)

Comments: Two independent facts relating to the high radiation do not require analysis to arrive at the correct answer

QUESTION # SRO-22 One hour has elapsed since a steam line break occurred in the Turbine Building. The transient has caused fuel damage a reactor scram and manual closure of the MSIVs. Following the transient the following conditions exist:

All rods reached 00 on the SCRAM Torus temperature is 96 degrees F There is indication of 50,000 Ibs/hr flow on the "C"main steam line flow instrument RPV level is 60" TAF and slowly increasing from a low point of 30" TAF RPV pressure is 760 psig and dropping slowly Security calls and informs you that steam can be seen issuing from, the Turbine Building Chemistry sampling results of reactor coolant are not in yet but the accompanying HP reported that the sample bottle was 5 WHR when the chemist left the sample station Iodine release is 50 uCl/sec An HP calls from Route 9 bridge and reports 700 mREM/hr TEDE at his location Classify the event.

A. General Emergency B. Site Area Emergency C. Alert D. Unusual Event ANSWER: B EXPLANATION: The key factor is the last data from the HP which satisfies Site Area Emergency.

The remaining indications all satisfy UE or ALERT.

TECHNICAL REFERENCE(S): .

Procedure EPIP-OC-01 Classification of Emergency Conditions. Appendix 1 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: EPIP-OC-01 Appendix I Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 3 Group # - 4 WA # 2.4.41 Importance Rating - 4.1 K/A Topic

Description:

Knowledge of the emergency action level thresholds and classifications Question Source: Bank # X Modified Bank ## (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X I O CFR Part 55 Content: 55.41 55.43 55.43(b)(5) and (4)

Comments: INPO bank Susquehanna 9/30/99. Changed units and terminology to OC specific.

QUESTION #SRO-23 The plant is in normal full power operation with no LCOs on April 1, 2004 when massive grid instabilities result in the loss of offsite power for the foreseeable future. The plant responds as designed including both Standby Diesel Generatorswhich have started and loaded to their respective buses. The following conditions exist as of noon on April 1, 2004:

b Diesel fuel oil delivery is uncertain due to infrastructure problems b

The Standby Diesel Generator Fuel Tank is at 14,500 gallons The heating boiler tank has 16,500 gallons of available fuel b

No other sources of diesel fuel are available on site b

The heating boilers are shutdown for maintenance The System Dispatcher has asked how long you can continue in this mode and still be able to re-start the plant once power is restored?

For your answer assume two diesels continue to run at the consumption rate specified in Amendment 18. Round off you answer to the nearest day.

A. Threedays B. Fourdays C. Fivedays D. Sevendays ANSWER: B EXPLANATION: Per TS Bases the rate is 12,410 gallons for three days. The 16, 500 gallons in the heating boiler tank will last three days and 23+ hours (16,546 gallons for four days) with the Standby Diesel Generator Fuel Tank maintained above its TS minimum level of 14,000 gallons.

TECHNICAL REFERENCE(S): TS 3.7.C and TS Bases for TS 3.7 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 3 Group # - 1 WA # 2.1.33 Importance Rating - 4.0 WA Topic

Description:

Ability to recognize indications for system operating parameters which are entry conditions for Technical Specifications

Question Source: Bank # X Modified Bank # (Note changes or attached parent)

New Question Cognitive Level: Memory or Fundamental Knowledge Comprehensive or Analysis X 10 CFR Part 55 Content: 55.41 55.43 55.43(b)(2)

Comments: Used INPO bank question from Duane Arnold 5/25/99. Made values and terminology consistent with OC Tech Specs

QUESTION #SRO-24 The plant is operating at 100% power with five reactor recirc pumps operating when one reactor recirc pump trips. The following conditions exist:

0 Reactor power is at 82%

All Operator actions of ABN-3200.02 Recirculation Pump Trip have been completed and continued operation is permissible with the four remaining pumps The Shift Manager directs you to make the necessary notifications Considering the above conditions, what personnel, at a minimum, are required to be notified by procedure?

A. Operations Manager, Plant Manager and on-shift chemist B. System Owner/Dispatcher, Reactor Engineer, on-shift chemist C. Operations Manager, System OwnedDispatcher, Reactor Engineer D. System OwnedDispatcher,Plant Manager and on-shift chemist ANSWER: B EXPLANATION: The power decrease has exceeded 289.5 MWT and the chemist must be notified to take a sample. Other personnel are specified in ABN-3200.02 TECHNICAL REFERENCE(S): -ABN-3200.02 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 3 Group # - 1 WA # 2.1.14 Importance Rating - 3.3 KIA Topic

Description:

Knowledge of system status criteria which requires notification of plant personnel Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 55.43 55.43(b)(5)and (2)

Comments:

QUESTION #SRO-25 A drywell entry must be made in order to inspect for increased unidentified leakage. A plant shutdown is in progress. The following conditions exist:

0 Reactor Power is 90% and decreasing 0

Purging of the drywell with air is in progress in accordance with Procedure 312.9, Primary Containment Control.

0 The Chemistry Department indicated that the Stack Gas Activity should not exceed 900 CPM, based on their sample 0 DRYWELL VENT-PURGE INTERLOCK BYPASS switch is in the BYPASS position (Panel 12XR)

Venting is via the Reactor Building Ventilation System 0

Stack gas activity is at 1100 CPS and slowly increasing Your direction to the operator(s) controlling the purge in accordance with Procedure 312.9 is that they are required to:

A. Decrease the purge flow until stack gas activity decreases below 900 CPM B. Confirm stack release rate with RAGEMS and then decrease purge flow rate.

C. Secure the primary containment purge by closing V-28-17 and V-28-18.

D. Shift the purge to go through the Standby Gas Treatment System ANSWER: C EXPLANATION: This is specified in Step 7.3.2.6 of Procedure 312.9. The other distractors, though possible mitigation strategies, are not specified actions.

TECHNICAL REFERENCE(S): Section 7.0 of Procedure 312.9 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: Section 7.1 and 7.2 of Procedure 312.9 Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 3 Group # - 3 WA # 2.3.9 Importance Rating - 3.4 KIA Topic

Description:

Knowledge of the process for performing a Containment Purge Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X

Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 55.43 55.43(b)(4)

Comments: