ML042750523

Final Written (RO & SRO) (Folder 3)

ML042750523
Person / Time
Site:	Susquehanna
Issue date:	08/09/2004
From:	Roush K PPL Generation
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
50-387/04-302, 50-388/04-302
Download: ML042750523 (163)

Text

SSES LOC 20 NRC Exam u

1.

Unit 2 is at 50% power and 90 Mlbm/hr Total Core Flow when the " B Recirc MG Set Drive Motor Breaker trips open due to a spurious signal.

Assuming no operator actions, when the plant is stable, what will final reactor power level and total core flow be?

A.

40% Power Total Core Flow 72 Mlbm/hr B.

40% Power Total Core Flow 45 Mlbm/hr C.

30% Power Total Core Flow 45 Mlbm/hr D.

30% Power Total Core Flow 72 Mlbm/hr Question Data C

30% Power Total Core Flow 45 Mlbmlhr

-d ExplanatiodJustification:

A.

B.

Assumes same proportional drop in flow as two to one pump at 30% speed (lines on current map). If the candidate assumes the same proportional drop then in flow on the proper rod pattern line then this choice will be chosen Flow drops approximately in half, power drops more than 10%. If the candidate properly evaluates the change in flow but assumes the same proportional drop in power as shown on the curves for 1 and 2 pump operation at 30% speed then this answer will be chosen correct answer, Reduction in power will follow current power line. It will drop to approximately half of its current value while the power drops significantly but by less than half.

The flow will drop much more based on experience with pump trips. If the candidate properly evaluates reactor power but believes that the same proportional drop as the 30% lines on the figure occurs then this answer will be chosen C.

D.

Sys#

System Category KA Statement 295001 Partial or Complete Loss Ability to determine and/or interpret the following as Neutron monitoring of Forced Core Flow Circulation FORCED CORE FLOW CIRCULATION:

they apply to PARTIAL OR COMPLETE LOSS OF WA#

295001.~~z.02 WA Importance 3.113.2 (RO/SRO)

Exam Level RO References provided to Candidate NDAP-QA-0338 Power Technical

References:

NDAP-QA.0338 to Flow Map Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.10 I 43.5 I 45.13 Objective:

3233 Plot power and core flow conditions and Task:

00.m.0 Implement Appropriate determine appropriate actions IAW the power to flow map.

Management & Controls 47 Portions Of Reactivity Program LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOCZONRCForm

.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

2.

Unit 1 is operating at 100% power when the Main Generator trips on differential current.

Based on the Main Generator trip on differential current, what are the reasons for theAuxiliary Bus (BUS 1 1 A & 1 1 B) transfer schemes?

A.

Auxiliary Bus Fast Transfer Protect ESS Transformer Auxiliary Bus Slow Transfer Protect Startup Transformer B.

Auxiliary Bus Fast Transfer Keep Loads running Auxiliary Bus Slow Transfer Protect ESS Transformer C.

Auxiliary Bus Fast Transfer Keep Loads running Auxiliary Bus Slow Transfer Backup to Fast transfer D.

Auxiliary Bus Fast Transfer Protect ESS Transformer Auxiliary Bus Slow Transfer Backup to Fast transfer Question Data C

Auxiliary Bus Fast Transfer Keep Loads running Auxiliary Bus Slow Transfer Backup to Fast transfer ExplanationNustification:

A.

8.

C.

-u' Fast transfer occurs first to keep loads running and slow transfer is the backup If the candidate reverses which is the primary and which is the backup this answer may be chosen.

Slow transfer backs up fast transfer, Load Shed protects the ESS transformers If the candidate misapplies the load shed purpose then this answer will be chosen.

correct answer, Auxiliary Bus Fast Transfer Provide power to the auxiliary buses without interruption, in the event that the normal power is lost Auxiliary Bus Slow Transfer Provide a second closure signal to Breakers 1A10104 "TIE BUS TO BUS 11A" and 1A10204 'TIE BUS TO BUS 11 B" Transformers.

Loads on the bus must be manually restarted.

Fast transfer is designed to keep the loads running. The load shed protects the ESS transformers. If the candidate confuses these purposes, this answer will be chosen

0.

Sys#

System Category KA Statement 295003 Partial or Complete Loss of AC. Power Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF AC. POWER:

Manual and auto bus transfer WA#

295003.~~3.01 WA Importance 3.313.5 Exam Level RO (RO/SRO)

References provided to Candidate None Technical

References:

TM-OP-003 Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.5 145.6 Objective:

1497 Describe the operation of a 13.8 W Auxiliary Bus Task:

03.0P.O Shift Auxiliary Buses Between Fast Transfer resulting from the Fault Trip of the 03 Generator Output And Startup Main Generator.

Buses LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCF0rm.d0~

Printed on 07128104

SSES LOC 20 NRC Exam

3.

A Unit 2 startup is in progress. The reactor is criticalland power is in the source range with an infinite period.

The I D673 +24 VDC Battery Charger fails, resulting in no output from the charger.

Assuming no operator action, how will the plant respond?

A.

Power will be immediately lost to the +24 VDC Bus supplying Process Rad Monitors, A and C SRMs and A/C/E/G IRMs. A half scram will immediately occur.

B.

The +24 VDC Bus will be energized by the -24 VDC Battery Charger 1 D674. No loss of power will occur.

C.

The +24VDC Bus will be initially energized by the battery. As the battery discharges, power will be lost to Process Rad Monitors, A and C SRMs and A/C/E/G IRMs. A half scram will occur.

D.

The +24 VDC Bus will be initially energized by the battery. As the battery discharges, the -24 VDC Battery Charger 1 D674 will continue to charge the battery, thus maintaining power to the bus.

L a

Question Data C

The +24VDC bus will be initially energized by the battery. As the battery discharges, power will be lost to Process Rad Monitors, A & C SRMs and PJCIEIG IRMs. A half scram will occur.

ExplanationlJustification:

A.

B.

C.

D.

Battery will carry the loads as it discharges If the candidate does not understand that the battery directly ties to the bus and will carry it, this answer will be chosen.

The -24VDC charger cannot carry the +24VDC bus. If the candidate believes that either of the chargers can carry the entire bus, this answer will be chosen.

correct answer, The battery will initially carry the load but will discharge without the charger.

The -24VDC charger will not charge the +24VDC battery The two busses each have their own battery and charger The charger cannot carry the other buss battery. If the candidate believes that this is possible then this answer will be chosen.

Sys#

System Category KA Statement 295004 Partial or Complete Loss Knowledge of the interrelations between PARTIAL OR Battery charger WA#

295004.~~2.01 WA Importance 3.113.1 Exam Level RO of D.C. Power COMPLETE LOSS OF D.C. POWER and the following:

(ROISRO)

References provided to Candidate None Technical

References:

~M-0~4175 Question Source:

New Susquehanna, 81412004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.7.45.8 Objective:

1448 Explain how key parameters of the 24 VDC Task:

75.ON.

Implement Loss Of 24V DC System and the Plant will respond to failure of a Battery Charger.

003 Bus LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

4.

Unit 1 is operating at 100% power when a turbine trip due to high vibration occurs. All systems have responded as designed for the turbine trip and reactor scram. As the turbine coasts down, vibrations become more severe. After consulting with the System Engineer, it is decided that the turbine must be immediately stopped. The turbine is rotating at 1,080 rpm and slowing. To STOP the turbine, ON-193-002, "MAIN TURBINE TRIP", directs the operator to perform which one of the following?

A.

Stop both EHC Pumps.

B.

Open CDSR VAC BKR HV-I0742A, B, C Valves.

C.

Close the Inboard (HV-141-F022A-D) and Outboard (HV-14 I

-F028A-D) MS IVS.

D.

Secure the steam jet air ejector Question Data B

Open CDSR VAC BKR HV-I0742A,B,C valves.

ExplanationlJustifiction:

A.

B.

C.

The turbine stop and control valves are already closed due to the trip. If the candidate does not understand this, this answer will be chosen.

correct answer, Directed by procedure to admit air and increase friction to stop the turbine. Friction is needed to slow the turbine.

Breaking vacuum adds air creating friction with the blades slowing the turbine Performed by procedure if the turbine does not slow down following the trip due to failure of the TSVs and TCVs. Since all systems functioned as designed this action will have no impact on turbine speed. If the candidate does not understand this, this answer will be chosen.

Not directed by procedure and will result in a slower loss of vacuum. This will result in a slow loss of vacuum and will not quickly stop the turbine. If the candidate does not understand this, this answer will be chosen..

-\\.-

D.

Sys#

System Category 295005 Main Turbine Generator Emergency Procedures and Plan Trip KIA#

295005.2.4.1 I WA Importance 3.413.6 (ROISRO)

Exam Level KA Statement Knowledge of abnormal condition procedures.

RO References provided to Candidate None Technical

References:

0~-193-002 Question Source:

New Susquehanna. 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.10 143.5 I 45.13 Objective:

3295 Respond to a main turbine trip Task:

43.0~.

003 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.d0~

Printed on 07/28/04

SSES LOC 20 NRC Exam

.~ -.-/-

5.

Unit 2 is performing a Reactor Startup with Main Turbine Shell warming in progress, and reactor pressure at 500 psig.

The Main Turbine Warming Demand ramps from 40% to 100%.

Assume no operator actions, what will be the plant response to this failure?

A.

Turbine Trip occurs.

6.

Reactor Scram occurs.

C.

Rod Block Monitor bypasses.

D.

Rod Worth Minimizer bypasses.

Question Data B

Reactor Scram occurs ExplanationNustification:

A.

B.

There is no turbine trip under these conditions. If the candidate does not recognize that there is no turbine trip for these conditions then this answer will be chosen.

correct answer, The reactor scram on TCV and TSV closure is bypassed until Reactor power reaches 30% as measured by first stage turbine pressure. Too much warming steam can raise the pressure to the point where the scram is unbypassed. Since the TSV are closed at this point a scram will result.

The REM is activated when reactor power is sensed as above 30% by the reference APRM. If the candidate does not understand where REM gets its power signal from, then this answer may be chosen.

The RWM is auto bypassed at 20% power as sensed by steam flow using feedwater flow. The student must understand where RWM receives its power signal for bypass.

---

./'

C.

D.

Sys#

System Category KA Statement 295005 Main Turbine Generator Knowledge of the interrelations between MAIN RPS WA#

295005.~~2.01 WA Importance 3.813.9 Exam Level RO Trip TURBINE GENERATOR TRIP and the following:

( RO/S RO)

References provided to Candidate None Technical

References:

TM-OP-093 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.7.45.8 Objective:

2746 Control first stage shell pressure during shell Task:

OO.GO.

Implement Appropriate warming.

01 3 portions of Plant Startup, Heatup, and Power Operation LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L,

6.

Unit 1 was coasting down with reactor power at 94%, when a reactor scram occurred. During the scram, there was a failure of several rods to fully insert.

The present plant conditions are as follows;

- EO-I 00-1 13, "LEVEUPOWER CONTROL" has been exited.

- Main Steam Isolation Valves Closed.

- Reactor Water Level being maintained using CRD and RWCU with letdown to Radwaste.

- One Control Rod indicating Position 48; remaining Control Rods full in.

- Reactor pressure 150 psig and cooling down.

- Preparations are in progress for placing Shutdown Cooling in service.

Describe the implications on Core Reactivity if the plant continues the evolutions described.

A.

Core Reactivity will go positive unless Xenon decay is countered.

B.

Core Reactivity will remain negative unless the Control Rod remaining at Position 48 is the most reactive rod.

C.

Core Reactivity will go positive unless feed and bleed of CRD and RWCU is terminated.

D.

Core Reactivity will remain negative unless the plant is cooled down to less than 68 O F.

Question Data D

Core Reactivity will remain negative unless the plant is cooled down to less than 68 deg F.

ExplanationlJustification:

A.

B.

C.

D.

The control rods will maintain the reactor shutdown Xe free. Xe decay will add positive reactivity but the reactor will remain shutdown. This answer assumes Xe decay will add negative activity. If the candidate does not understand the designed shutdown margin this answer may be chosen The core is designed to remain shutdown with the most reactive rod full out. If the candidate believes that the Rx can go critical on one rod then this answer may be chosen.

Unlike PWRs there is no need under normal conditions to maintain the reactor shutdown using boron. Feed and bleed would reduce any dissolved boron. But there is none. If the candidate does not understand that no boron would have been injected under these circumstances then this answer may be chosen correct answer, The core is designed to remain shutdown with all control rods full in except the single most reactive rod full out and the core cold (68 degrees F) and Xe free Sys#

System Category KA Statement 295006 SCRAM Knowledge of the operational implications of the Reactivity control WA#

295006.~~1.03 WA Importance 3.714.0 Exam Level RO following concepts as they apply to SCRAM:

( RO/S RO)

References provided to Candidate EOPS Technical

References:

EO-100-113 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.8 to 41.10 Question Source:

New Susquehanna, 8\\4/2004 Level Of Difficulty: (1-5) 3 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

5455 Explain the difference between Hot Shutdown Task:

OO.EO.0 Implement LevellPower 31 Control L.-

Boron Weight (HSBW) and Cold Shutdown Boron Weight (CSBW), including their importance in maintaining reactor shutdown during accident conditions.

LOC 20 As Given H:\\ExarnBank\\MergeDo~W-OC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 1.'

7.

Units 1 and 2 Control Rooms have entered ON-lOO(200)-009, "CONTROL ROOM EVACUATION". Unit I Control Room Operators were unable to scram the reactor prior to exiting.

After the appropriate actions of ON-I 00-009, "CONTROL ROOM EVACUATION", have been completed, the procedure directs the operators to determine that the reactor is scrammed by observing scram valve position on a few HCUs. If an HCU has properly scrammed, the operator will observe which one of the following?

A.

Scram Inlet Valve indication UP Scram Outlet Valve indication UP B.

Scram Inlet Valve indication UP Scram Outlet Valve indication DOWN C.

Scram Inlet Valve indication DOWN Scram Outlet Valve indication UP D.

Scram Inlet Valve indication DOWN Scram Outlet Valve indication DOWN

'i-Question Data A

Scram Inlet Valve indication UP Scram Outlet Valve indication UP ExplanationNustification:

A.

correct answer, UP indicates that the valves are open. The valves are held closed by air pressure and open on a scram. The distracters represent the other possible combinations. The ON specifically directs this to be done in order to verify the reactor has scrammed.

8.

The outlet is closed C.

The inlet is closed D.

Both the inlet and outlet are closed.

Sys#

System Category KA Statement 295016 Control Room Ability to determine andlor interpret the following as Reactor power WA#

295016.A42.01 WA Importance 4.114.1 Exam Level RO Abandonment they apply to CONTROL ROOM ABANDONMENT (RO/SRO)

References provided to Candidate None Technical

References:

ON-io0-009 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.10 143.5 I 45.13

\\-

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2413 Describe the operation and controls of the following components in the CRD Hydraulic System:

a. Suction Filter Strainers

c. CRDPumps Tie Line

e. Minimum Flow Line Filters

g. Flow Transmitter Flow

i. Charging Pressure Restricting Orifices

k. HCUs Valves

m. Flow Controller Valves

0. Drive Flow Transmitter Pressure

q. Directional Control Valves Water Pressure Valve

s. Cooling Water Flow Water Pressure

u. Scram Valves Valves

w. Scram Discharge Volume and Drain Valves

y. SDV Level Switches Accumulators aa. Lights and Alarms Rod Insertion Valves

b. Suction

d. Unit Cross-

f. Drive Water

h. System

j. Flow I. Flow Control

n. Stabilizing

p. Drive Flow

r. Drive

t. Cooling

v. Scram Pilot

x. SDVVent

z. Scram bb. Alternate Task:

55.0P.O Perform Hydraulic Removal 03 From Service Of A HCU LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

'v

8.

Unit 1 is at 100% power with 'B' RBCCW Pump out of service for overhaul. All other systems are in a normal lineup.

The RBCCW System Engineer reports that analysis of pump performance data indicates that the 'A' RBCCW Pump impeller has degraded.

Continued impeller degradation will cause RBCCW flow and pressure to decrease and will result in which one of the following?

A.

The 'A' RBCCW Pump will trip on low flow causing a complete loss of RBCCW.

B.

Reactor Building Chilled Water Chillers will swap.

C.

Reactor Recirculation Pump 'A' and '6' Motor Winding Temperatures will increase causing a trip.

D.

RWCU System will isolate on high temperature.

Question Data D

RWCU system will isolate on high temperature.

ExplanatiodJustification:

k B.

C.

The RBCCW pumps do not have a low flow trip. Other systems, such as RWCU, do have a trip on low flow. The candidate may incorrectly believe that RBCCW has a low flow trip.

The swap is for loads supplied by RBCW to swap to RBCCW to maintain cooling. There is no auto swap of normal RBCCW loads to RBCW. If the candidate is confused over what the swap does, this answer may be chosen.

The normal supply to the RR Pump motor winding coolers is RBCW. RBCCW is the alternate supply. If the candidate is confused as to the normal cooling supply to the Rx Recirc pumps, this answer may be chosen. Motor bearing oil temperatures would increase and eventually cause a trip ca correct answer, RBCCW supplies the RBCCW heat exchangers. As pump pressure and flow drop, the heat exchanger will receive less cooling. The system can isolate on high temperature.

\\-.-/

D.

Sys#

System Category KA Statement 295018 Partial or Complete Loss Knowledge of the operational implications of the Effects on componentlsystem following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER of Component Cooling Water operations WA#

295018.~~1.oi WA Importance 3.513.6 (RO/SRO)

Exam Level RO References provided to Candidate None Technical

References:

TM-OP-014 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.8 to 41.10 Objective:

1675 Predict how key parameters of the Reactor Task:

14.ON.

Implement Loss Of Reactor Question Source:

New Susquehanna, 81412004 Level Of Difficulty: (1 -5) 3 Building Closed Cooling Water system and the plant will respond to failure of the following components or controls.

003 Building Closed Cooling Water

a. RBCCW Pumps

b. Temperature Controller

c. Containment Isolation Valves L

d. System piping LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07128104

SSES LOC 20 NRC Exam

.--- 9.

Unit I is initially at 100% power, and the Instrument and Service Air Systems are in a normal lineup with Instrument Air Compressor 1 K107A out of service for maintenance.

An Instrument Air System leak slightly greater than the available Instrument Air Compressor's capacity has occurred.

As Instrument Air pressure degrades, what will be the automatic response of the Instrument Air System?

A.

At 75 psig, both the AR-124-AO1, "INSTRUMENT AIR LOOP A LO PRESSURE", and AR-124-BOI, "INSTRUMENT AIR HEADER LO PRESSURE", alarms will be received.

The SERVICE AIR CROSSTIE TO INSTRUMENT AIR (PCV-12560) will close to isolate service air.

B.

At 80 psig, AR-124-AO1, "INSTRUMENT AIR LOOP A LO PRESSURE", alarm will be received and the SERVICE AIR CROSSTIE TO INSTRUMENT AIR (PCV-12560) will close to isolate Service Air.

C.

At 75 psig, both the AR-124-AOI, "INSTRUMENT AIR LOOP A LO PRESSURE, and AR-124-BOI, "INSTRUMENT AIR HEADER LO PRESSURE", alarms will be received.

The SERVICE AIR CROSSTIE TO INSTRUMENT AIR (PCV-12560) will open and restore pressure approximately 90 psig.

\\-

D.

At 95 psig, the SERVICE AIR CROSSTIE TO INSTRUMENT AIR (PCV-12560) will open and maintain pressure approximately 90 psig.

Question Data D

At 95 psig, the ERVICE AIR CROSSTIE TO INSTRUMENT AIR (PCV-12560) will open and maintain pressure approximately 90 psig.

ExplanationlJustification:

A.

6.

C.

D.

The crosstie does not close to isolate service air but opens to supply IA with SA. This combines the misconceptions of the other distracters The crosstie does not close to isolate service air but opens to supply IA with SA. If the operator does not understand the relationship, since at some plants service air is supplied by instrument air, then this answer may be selected The crosstie will open at 95 psig and maintain pressure without any alarms. The alarms come in at 7% and 70#. The operator may believe that the alarms come in at the same time as the auto actions.

correct answer, The Crosstie valve will begin to open at 95 psig and will be full open at 90 psig and will maintain pressure at 90 psig Sys#

System Category KA Statement 295019 Partial or Complete Loss of Instrument Air Ability to operate andlor monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR WA##

295019.~~i.02 WA Importance 3.313.1 Exam Level RO Instrument air system valves:

PlantSpecific (RO/S RO)

References provided to Candidate None Technical

References:

ON-118401 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.7 I 45.6 L-LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-Objective:

1769 List the signals and setpoints that cause the Task:

18.ON.

Implement Loss Of Instrument Air System to automatically initiate, 003 Instrument Air isolate, and trip.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-e' IO.

The reactor is shut down with one loop of shutdown cooling in use, and NO Recirculation Pumps running.

How would the Control Room Operators respond if Reactor Vessel Level decreased from

+50 inches to +4 inches?

A.

Enter EO-I 00-1 02, "RPV CONTROL", verify alarm response for AR-101 -B17, "RX WATER HI-LO LEVEL"; verify shutdown cooling remains in service.

B.

Enter ON-149-001, "LOSS OF SHUTDOWN COOLING", maximize RHR keepfill, check the shutdown cooling suction inboard and outboard isolation valves isolate; manually trip the operating RHR Pump.

C. Enter EO-I 00-102, "RPV CONTROL", maximize CRD and check the shutdown cooling suction inboard and outboard isolation valves isolate; verify the operating RHR Pump trips.

D.

Verify alarm response for AR-109-C09, "HV-151-FOO6NC AND HV-151-FO07A OPEN DRAINS RX VESSEL", verify the shutdown cooling suction inboard and outboard isolation valves isolate, verify the operating RHR Pump trips; verify remaining RHR Pumps auto start.

Question Data C

Enter EO-100-102, "RPV CONTROL", maximize CRD and check the shutdown cooling suction inboard and outboard isolation valves isolate, verify the operating RHR pump trips.

ExplanationlJustification:

A.

B.

C.

D.

No suction path, running pump will trip. The RHR shutdown cooling suction valves close on low RPV level to isolate a potential leakage path. The candidate may not recognize the isolation setpoint of shutdown cooling has been reached.

No suction path running pump will trip. Core spray pumps and the RHR pumps controlled from the RSP have no trip on closed valves in the suction path. The candidate may incorrectly apply this.

correct answer, running RHR pump will trip due to loss of suction path., Entry into EO required due to level less than +13 inches F015 valve auto opens and RHR pumps will auto start at -129. This occurs on low RPV level but the candidate may not recognize that the setpoint has not been reached.

Sys#

System Category KA Statement 295021 Loss of Shutdown Cooling Emergency Procedures and Plan Knowledge of low power I shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

wA#

295021.2.4.9 WA Importance 3.313.9 Exam Level RO (RO/SRO)

References provided to Candidate EOPs Technical References :

TM-OP-049 Question Source:

Exam Bank Susquehanna, 8/1/2003 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.10 143.5 I 45.13 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2067 List the interlocks for the following components Task:

49.ON.

Implement Loss Of RHR 003 Shutdown Cooling Mode and controls of the RHR System, including setpoints:

L-

a. RHRPumps

b. Heat Exchanger Bypass Valve

c. Injection Valves

d. Suppression Pool Test and Spray Valves

e. Drywell Spray Valves

f. Radwaste valves

g. Pump Suction and Shutdown Cooling Valves

h. Minimum Flow Control Valves LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

11.

Fuel Handling operations are in progress. The following conditions exist:

Mode switch in REFUEL.

Fuel Grapple OPEN.

Fuel Grapple full up.

One (1) control rod selected and fully withdrawn.

Refueling Platform currently located over the Spent Fuel Pool.

For these conditions, which of the following interlocks will be enforced, and why?

A.

Selecting a second control rod will cause a rod block; prevents inadvertent criticality.

B.

Attempting to move the Refueling Platform near or over the Core will not be permitted; prevents platform operator overexposure.

C.

Attempting to move the Refueling Platform near or over the Core will not be permitted; prevents inadvertent criticality.

D.

Selecting a second control rod will cause a rod block; prevents platform operator overexposure.

Question Data A

Selecting a second control rod will cause a rod block; prevents inadvertent criticality.

ExplanationIJustification:

A.

B.

C.

D.

With the mode switch in REFUEL only one control rod can be withdrawn Motion of the bridge is permitted since it is unloaded. If the grapple was loaded then motion toward the core would be blocked. The candidate may believe that all motion over or near the core is automatically prevented with a single control rod withdrawn.

Motion of the bridge is permitted since it is unloaded If the grapple was loaded then motion toward the core would be blocked. The candidate may believe that all motion over or near the core is automatically prevented with a single control rod withdrawn.

The rod block is designed to prevent inadvertent criticality. An inadvertent criticality on the refuel floor would result in a high exposure to personnel on the refuel floor but this is not the basis for the interlock. The candidate may not understand the basis for the interlock.

There are many operational restrictions based on preventing operational exposure.

Sys#

System Category UA Statement 295023 Refueling Accidents Knowledge of the reasons for the following Interlocks associated with fuel WA#

295023.~~3.02 WA Importance 3.413.8 Exam Level RO responses as they apply to REFUELING ACCIDENTS:

handling equipment (RO/S RO)

References provided to Candidate None Technical

References:

TM-oP-081~

Question Source:

Exam Bank FikPatrick 1,7/1/2003 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.5 145.6 Objective:

2672 State the refueling interlocks associated with the Task:

81.AD.O Perform Refueling Operations following:

05

a. Refueling Platform forward or reverse motion

b. Refuel Switches # I and #2

c. Fuel Grapple Hoist motion

d. Control Rods insertedlwithdrawn LOC 20 As Given H:\\ExamBank\\MergeDoc\\LOC2ONRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

.\\,-'

12.

EO-100-1 14, "RPV FLOODING" is currently being executed, and the following conditions exist:

- All control rods at 00.

- All ADS Valves are open.

- RPV pressure 98 psig and steady.

- Drywell pressure 15 psig and increasing slowly.

- Suppression Chamber pressure 18 psig and increasing slowly.

- Condensate and both Loops of Core Spray are injecting to the RPV.

- Suppression Pool level is 22 feet and rising slowly.

What action, if any, is the operator required to perform at this time, and what is the reason for this action or inaction?

L:

A.

B.

C.

D.

No action is required at this time. The operator will close Non-ADS SRVs to maintain RPV pressure at least 81 psig above Drywell pressure. This ensures sufficient pressure to restore RPV level indication by allowing the reference legs to refill and avoid reference leg flashing.

Increase RPV injection to maintain RPV pressure at least 81 psig above Suppression Chamber pressure. This differential pressure ensures sufficient steam flow through the SRVs to remove all core decay heat and maintain adequate core cooling.

No action is required at this time. The operator will be preparing to increase RPV injection to maintain RPV pressure at least 81 psig above Drywell pressure. This differential pressure ensures sufficient steam flow through the SRVs to remove all decay heat from the core and maintain adequate core cooling.

Close SRVs until four are open to maintain RPV pressure at least 81 psig above Suppression Chamber pressure. This ensures sufficient pressure to restore RPV level indication by allowing the reference legs to refill and avoid reference leg flashing.

Question Data B

Increase RPV injection to maintain RPV Dressure at least 81 DsiQ above Sumression Chamber Dressure. This differential pressure ensures sufficient steam flow through the SRVs to remove all core decay heat and ma'intain adequate core cooling.

ExplanationlJustification:

A.

6.

C.

D.

Closing SRVs is not permitted by the Bases. This is an obvious but incorrect method of trying to increase pressure Correct Answer. This is the Minimum RPV Flooding Pressure. It ensures that there is sufficient steam flow through 4 SRVs to remove the decay heat generated 10 minutes after the reactor scram The MRFP is based on Rx to Suppression Chamber DP and is based on adequate core cooling. The candidate may use drywell pressure given its ubiquitous presence in the EOPs The MRFP is based on Rx to Suppression Chamber DP and closing SRVs is not permitted by the Bases This is an obvious but incorrect method of trying to increase pressure.

Sys#

System Category KA Statement 295024 High Drywell Pressure Knowledge of the reasons for the following RPV flooding responses as they apply to HIGH DRYWELL 1-PRESSURE:

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L

KIA#

295024.~~3.05 WA Importance 3.513.8 (RO/SRO)

Exam Level RO References provided to Candidate EOPs Technical

References:

EO-OOO-114 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 145.6 Objective:

5473 Describe core cooling by RPV Flooding.

Task:

OO.EO.0 Implement RPV Flooding 32 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForrn.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

13. Twenty minutes after a scram, an MSlV Isolation occurs and the operator controls RPV pressure using the SRVs. To aid in pressure control, RClC is placed in pressure control CST to CST with the RClC Controller in AUTO. RClC speed is observed to oscillate.

Which of the following explains the RClC speed oscillations while in RPV pressure control?

A.

In pressure control, the RClC Controller attempts to maintain a constant speed, but cannot respond fast enough to maintain speed as the RPV pressure changes.

B.

Using RClC for pressure control is inherently less stable than using it for level control due to the lower pressure in the CST compared to the RPV. The greater instability is seen as an increase in oscillations.

C.

The comparatively small CST volume results in the RClC suction and discharge points in the CST being close together, and at high flow the turbulence causes oscillations.

D.

Pressure control using SRVs results in swings in RPV pressure as the SRVs open and close. This requires the RClC speed to change as the controller maintains a constant flow.

Question Data D

Pressure control using SRVs results in swings in RPV pressure as the SRVs open and close. This requires the RClC speed to change as the controller maintains a constant flow.

ExplanationlJustification:

A.

6.

C.

D.

In auto the controller maintains flow and in manual it maintains speed. The candidate may reverse the methods of RClC control.

The controller is equally stable in the pressure and level control modes. The candidate may believe that the operation of RClC in other than its design function of injecting to the core is less stable The CST has a relatively small volume compared to the SP but it does not result in oscillations. The operator may accept that the smaller flow volume results in suction/discharge interaction SRV pressure control causes pressure to cycle as the SRVs open and close. The RClC controller in auto will maintain a constant flow but with turbine inlet pressure varying the control valve position as thus speed as the steam pressure changes.

Sys#

System Category KA Statement 295025 High Reactor Pressure Ability to operate andlor monitor the following as RCIC: Plant-Specific KIA#

295025.~~1.05 KIA importance 3.713.7 Exam Level RO they apply to HIGH REACTOR PRESSURE:

(RO/SRO)

References provided to Candidate None Technical

References:

TM-OP-050 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.7 145.6 Objective:

2011 Determine if RCIC and Plant response is Task:

50.0P.O Perform AutomaticlManual appropriate for any combination of 10 Startup Of RClC System

a. System mode of operation

b. Plant conditions

c. Key parameter indications LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\-

14.

A Main Generator trip and an Aux Bus load shed occurred while Unit 2 was operating at 100%

power.

The following conditions exist;

- A failure to scram has occurred.

- Initial Reactor power was 4%.

- SRVs are cycling to maintain Reactor Pressure.

- Drywell pressure is 18 psig.

- Drywell Temperature is 149 OF.

- Suppression pool water level is 32 feet.

- Suppression Pool Temperature is 155 O F.

WHICH of the following states the required operator action and the basis?

A.

Perform a normal cooldown using the Main Turbine Bypass Valves to prevent using SRVs so the impulse load on the Suppression Pool does not exceed design loads.

B.

Reduce reactor pressure using HPCl and RClC to prevent using SRVs so the impulse load on the Suppression Pool does not exceed design loads.

u C.

Perform an emergency cooldown using the Main Turbine Bypass Valves to prevent containment overpressure following RPV depressurization.

D.

Perform a rapid depressurization using the SRVs to ensure primary containment vent valve opening pressure will not be exceeded following RPV depressurization.

Question Data D

Perform a rapid depressurization using the SRVs to ensure primary containment vent valve opening pressure will not be exceeded following RPV depressurization.

ExplanatiorVJustification:

A.

B.

C.

D.

HCTL has been exceeded, the operator is required to RD and must use the SRVs. Main Condenser not available due to the load shed and loss of circ water. If the candidate does not recognize this then use of the main condenser is an attractive option.

Operation of HPCl and RClC with Suppression Pool Temperatures greater than 140 deg F will damage the equipment. If the candidate does not recognize the application of the caution, HPCl and RClC could be used to reduce pressure.

HCTL has been exceeded, the operator is required to RD and must use the SRVs. Main Condenser is not available due to the load shed and loss of circ water. If the candidate does not recognize this then use of the main condenser is an attractive option.

HCTL has been exceeded and depressurization is required to ensure vent valve opening is assured.

Sys#

System Category KA Statement 295026 Suppression Pool High Knowledge of the operational implications of the Steam condensation following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

Water Temperature WA#

295026.~~1.02 WA Importance m 3. 8 (RO/SRO)

Exam Level RO References provided to Candidate EOPs Technical

References:

EO-100-103 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.8 to 41.10 Question Source:

NRC Exam Hope Creek Unit 1,8/10/1998 Level Of Difficulty: (1-5) 4

x.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.d0~

Printed on 07/28/04

SSES LOC 20 NRC Exam

\\ -

Objective:

2598 For each Symptom Based EOP:

Explain the basis for each step.

Task:

OO.EO.0 Implement RPV Flooding 32 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCPONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

15.

Unit 2 has experienced a small steam line break in the Drywell with plant equipment responding as designed. Drywell Temperature is 280 O F, and drywell sprays are unavailable.

RPV pressure is 900 psig and stable.

At this time, with no operation of SRVs, AR-210-EO1, "MAIN STEAM SRV LEAKING, alarm is received on Panel 2C601, AR-210-EO2, "MAIN STEAM DIV 1 SRV OPEN", alarm has NOT been received, AR-210-E03, "MAIN STEAM DIV 2 SRV OPEN", alarm has NOT been received.

What is the cause for the AR-210-EO1, "MAIN STEAM SRV LEAKING", alarm?

A.

The SRVs are leaking due to exceeding the SRV's ambient temperature Environmental Qualification (EQ).

6.

High Ambient Temperature has heated the tailpipe to the alarm setpoint.

C.

The tailpipe temperature sensor has failed due to exceeding the instrument's Environmental Qualification (EQ).

D.

ADS has initiated automatically, resulting is steam flow through the valve.

~

L Question Data B

High Ambient Temperature has heated the tailpipe to the alarm setpoint.

ExplanationlJustification:

A.

The valves are designed for max drywell temp of 340 deg F. IF the valves were leaking with significant flow the MAIN STEAM DIV 1 SRV OPEN (AR-210-E02) and MAIN STEAM DIV 2 SRV OPEN (AR-210-E03) alarms would have been received due to input from the acoustic monitors. If the candidate does not understand the expected drywell conditions and valve operability this answer may be chosen.

The alarm setpoint is 250 deg F. The high drywell temperature has raised the sensor to its setpoint.

The sensors are designed for max drywell temp of 340 deg F. Since the steam temperatures in the tailpipe can be above 300deg F (from Mollier Diagram) the sensors are designed for higher temperatures than in this situation. The candidate may not understand the expected temperature and chose this answer.

If all systems operated as designed, low pressure ECCS would reflood the vessel and ADS will not initiate. If the valve was open with flow then the acoustic alarms would also be in. If the candidate does not understand the ADS logic and the expected indications, this answer may be chosen.

B.

C.

D.

Sys#

System Category KA Statement 295028 High Drywell Temperature Ability to operate andlor monitor the following as ADS K/A#

295028.~~1.0s WA Importance 3.713.7 Exam Level RO they apply to HIGH DRYWELL TEMPERATURE:

(RO/SRO)

References provided to Candidate None Technical

References:

AR-21 0-001 Question Source:

New Susquehanna. 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 I 45.6 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2092 Describe the operation of the following Task:

83.ON.

implement Stuck Open w

components and controls of the Automatic 003 Safety-Relief Valve Depressurization System:

a. Timer Reset Switches

b. ADS Inhibit Switches

c. SRV Switches

d. Man lnit Switches

e. Local ADS Keyswitches

f. Acoustic Monitors

g. SRVs

h. Vacuum Breakers

i. T-Quenchers

j.

k. Tailpipe Temp Recorders I. Lights and Alarms Air Accumulators ( I O and 42 gallons)

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20N RCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam c

16.

Unit I is at 100% power.

'A Loop RHR, is in Suppression Pool cooling using 'C' Pump ( I P202C),

RHR is in Suppression Pool Cooling to support SO-I 50-002, "QUARTERLY RClC FLOW VERIFICATION."

RClC is in-service for SO-I 50-002, "QUARTERLY RClC FLOW VERIFICATION."

The Reactor Building Operator sent to investigate a Reactor Building Sump Alarm reports both Reactor Building Sump Pumps running.

In addition, the following alarm is received:

- AR-III-EO2, "SUPPRESSION POOL DIV 1 LO LEVEL."

Rad Waste Control Room Reports high influent to Radwaste Collection Tanks Control Room Indications:

- Drywell Pressure 0.5 psig Constant

- Suppression Chamber pressure 0.4 psig Constant

- Suppression Pool Level 22 feet Lowering from 22.5 feet

- Suppression Pool Temperature 84 OF Rising from 80 OF

- RHR Loop 'A' Flow indication

- NO other Control Room Alarms

- All other Control Room indications are constant and consistent with 100% power.

10,500 GPM Rising from 9,000 GPM The operator is required to enter EO-I 00-1 03, "PRIMARY CONTAINMENT CONTROL", and:

A.

Close RClC MIN FLOW VLV TO SUPP POOL, FV-149-FO19.

Adjust RHR flow to < I 0,000 GPM.

B.

Close RHR LOOP A CROSSTIE, HV-151-FOIOA.

Adjust RHR flow to <10,000 GPM.

C.

Shut down RCIC.

Adjust RHR flow to 40,000 GPM.

D.

Shut down 'C' RHR Pump.

Shut down RCIC.

Question Data D

Shutdown 'C' RHR pump Shutdown RClC ExplanationlJustification:

A.

This action is for the RClC min flow valve open that would cause SP level to increase as CST water is sent to the SP This would not explain the increase in RHR flow and would not increase flow to radwaste. If the candidate misanalyzes the situation, this answer may be chosen.

If the RHR loop was pumping directly to radwaste, which these actions a intended to stop, there would not be the water flowing rapidly into the sump. If the operator misinterprets the indications and does not realize that RHR is not pumping directly to radwaste or does not recognize the impact of these steps, this answer may be chosen.

B.

L.-

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam C.

RClC suction open from the SP would cause SP level to drop but not cause high sump flows. The RHR flow change would not be explained by a changed in RClC parameters and will not be fixed by shutting down RCIC. If the candidate does not recognize an RHR break then this answer may be chosen.

The SP low level and the increase in flow to radwaste indicates a break in pipeway. The RClC system is only discharging steam to the SP. It is running CST to CST.

'L..'

D.

Sys#

System Category 295030 Low Suppression Pool Emergency Procedures and Plan Water Level W M 295030.2.4.48 WA Importance 3.513.8 (RO/SRO)

Exam Level KA Statement 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.

RO References provided to Candidate EOPS Technical

References:

EO-100-io3 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

43.5 145.12 Objective:

5491 Identify the problems and corrective actions Task:

20.ON.

Implement Flooding In associated with low suppression pool level.

006 Reactor Building LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

17.

Unit 2 is in Mode 4 with the following conditions:

- A RHR Pump running in Shutdown Cooling mode.

- Both Recirculation MG Sets out of service.

- RWCU isolated.

- CRD Pumps shut down.

- Condensate and Feedwater out of service.

- A TOTAL loss of Shutdown Cooling occurs with vessel water level at +40 inches.

Following the loss of shutdown cooling, RPV stratification would be indicated by the following trends:

A.

Bottom Head temperature remaining constant.

Recirc Loop temperatures increasing.

Reactor Vessel metal temperature at vessel head increasing.

B.

Bottom Head temperature increasing.

Recirc Loop temperatures increasing.

Reactor Vessel metal temperature at beltline increasing.

C.

Bottom Head temperature increasing.

Recirc Loop temperatures remaining constant.

CRD Mechanism High temperature alarming.

D.

Bottom Head temperature remaining constant.

Recirc Loop temperatures remaining constant.

Reactor Vessel metal temperature at the beltline increasing.

Question Data D

Bottom Head temperature remaining constant.

Recirc Loop temperatures remaining constant Reactor vessel metal temperature at the beltline increasing ExplanatiodJustification:

k B.

C.

D.

There is no flow in the recirc loop so temperature will not increase in the recirc loop. If the candidate does not recognize this, this answer may be chosen.

Bottom head temperature will not increase since there is no flow in the region with CRD, Recirc and RWCU out of service. Temps are also lower since the reactor is in Mode 4 (cold shutdown) There is no flow in the recirc loop so temperature will not increase in the recirc loop. If the candidate does not recognize the loss of flow into and through the bottom head, this answer may be chosen.

Bottom head temperature will not increase since there is no flow in the region with CRD, Recirc and RWCU out of service. CRD mechanisms will not alarm since there is no hot water to heat them. If the candidate does not understand the impact of the loss of flow on the parameters, this answer may be chosen based on the belief the bottom head will heat up but since the reactor is in mode 4 the water is cold and hot water will not flow down to replace it..

correct answer, indicated reactor pressure will increase due to decay heat raising the temperature of the coolant in the immediate core area. Bottom head temperature remains constant since there is no flow in and around the bottom head with RWCU and CRD out of service. Recirc loop temperatures will not change since there is no natural circulation out side of the core barrel due to no flow path through the steam separators. Vessel beltline temperature will rise due to radiant and conductive heating from the core region.

Sys#

System Category UA Statement LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOCPONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 295031 Reactor Low Water Level Knowledge of the operational implications of the Natural circulation: Plant-

-\\

following concepts as they apply to REACTOR LOW Specific WATER LEVEL WA#

295031.~~1.02 WA Importance 3.814.1 Exam Level RO (RO/S RO)

References provided to Candidate None Technical

References:

ON-2494011SOER 8202 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.8 to 41.10 Objective:

3410 Determine if RHR and plant response is Task:

49.ON.

Implement Loss Of RHR appropriate when shutdown cooling is lost.

003 Shutdown Cooling Mode Question Source:

Exam Bank Susquehanna, 8/1/2003 Level Of Difficulty: (1-5) 2 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L

18.

Cold Shutdown Boron Weight is DEFINED as the amount of boron necessary to maintain the reactor shutdown under the following:

A.

Water level is at the high-level trip setpoint (+54 in.),

Xenon free, and Shutdown cooling and RWCU are in service with the filters isolated.

6. Water level is between -1 I O and -60 inches, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> shutdown Xenon, and Shutdown cooling and RWCU are in service with the filters isolated.

C.

Water level is at the high-level trip setpoint (+54 in.),

72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> shutdown Xenon, and Shutdown cooling and RWCU are NOT in service.

D.

Water level is between -1 10 and -60 inches, Xenon free, and Shutdown cooling and RWCU are NOT in service.

Question Data A

Water level is at the high level trip setpoint (+54 in.),

Xenon free, and Shutdown cooling and RWCU are in service with the filters isolated.

ExplanationIJustification:

A.

B.

C.

D.

This is the definition. High water level has greater dilution of boron than low level, Xenon adds negative reactivityso Xenon free is more reactive, and systems are in service as RWCU would remove boron water level is too low and corresponds to the level band in EO-100-113 step LQ/L 6, Xenon concentration is not correct and RWCU is not in service. If the candidate does not understand the definition this answer may be chosen Xenon concentration is too high. If the candidate does not understand that it is defined for Xe free conditions, this answer may be chosen.

Water level is too low. If the candidate does not understand the definition this answer may be chosen Sys#

System Category KA Statement 295037 SCRAM Condition Present Knowledge of the reasons for the following Cold shutdown boron weight:

and Reactor Power Above APRM Downscale or Unknown DOWNSCALE OR UNKNOWN:

responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM PlantSpecific WA#

295037.EK3.05 WA Importance 3.213.7 (ROE RO)

Exam Level RO References provided to Candidate None Technical

References:

EO-loo-113 Question Source:

Modified Grand Gulf 1, 3/27/1998 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Memory 10 CFR Part 55 Content:

41.5 I 45.6 Objective:

121 5 Define andlor discuss the operational implications Task:

OO.EO.0 Implement LevellPower of the following terms for the Standby Liquid 31 Control Control System:

a. Hot Shutdown boron weight

b. Cold Shutdown boron weight b

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

.\\

19.

A leaking valve has wetted down the 'A' SBGT Charcoal Bed. If a refuel handling accident were to occur at this time, when the ' A SBGT Train is initiated, which of the following release rates would be much higher than design?

A.

Transuranics B.

Particulates C.

Noble Gases D.

Iodine Question Data D

Iodine ExplanationlJustification:

A.

6.

C.

D.

Transuranics are released as a result of fuel failure but do not go airborne and remain in solution. If the candidate believes transuranics will be released as a gas, this answer may be chosen.

The HEPA filter removes the particulates and will continue to do so. If the candidate believes that the charcoal holds up the particulate this answer may be chosen.

The Noble gases have very little hold up and would not be significantly changed. If the candidate confuses the function of charcoal in the offgas system with this charcoal, this answer may be chosen.

The charcoal bed is the primary component for removing iodine and water diables iodine removal.

Sys#

System Category KA Statement 295038 High OffSite Release Rate Knowledge of the interrelations between HIGH OFF-Plant ventilation systems SITE RELEASE RATE and the following:

KIA#

295038.~~2.03 WA Importance 3.613.8 (RO/SRO)

Exam Level RO References provided to Candidate None Technical

References:

TM-OP-070 Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.7,45.8 Objective:

1288 Describe the operation of the following Task:

1994 components and controls of the Offgas Recombiner System.

a. Hydrogen Recombiner

b. Charcoal Adsorbers

c. Isolation Valves

d. Recombiner Condenser

e. Preheater

f. H2 Analyzer

g. Pretreatment Rad Monitor

h. Condensate Cooler

i. Control Switches

j. LocallRemote panels LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCPONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-

20.

The following Simplex Alarm is received.

FIRE SUP X222-23 ALM TIME: 0300 DATE: 08/14/04 02-656 WPSI 11 CNDNSR Which of the following would be the plant response for the given Simplex Alarm:

A.

AR-036-B01, "PUMP (Fire) IS OPERATING", alarm will be received, and AR-036-B05, "ENGINE RUNNING", alarm will be received, and Input to Radwaste Collection Tanks will increase.

B.

AR-036-B01, "PUMP (Fire) IS OPERATING", alarm will be received, and AR-036-B05, "ENGINE RUNNING", alarm will be received, and HVI 61 50 Condenser Area Transfer Sump Isolation Valve closes.

C.

High flow from FSH12201A ( FSH FOR WPS-1 I 1 UNIT 1 TB CDSR AREA ) and WPS-111 OS&Y SUPPLY VALVE via ZS-12201A Not Full open, and HV16150 Condenser Area Transfer Sump Isolation Valve closes.

D.

High flow from FSH12201A ( FSH FOR WPS-1 I 1 UNIT 1 TB CDSR AREA ) and WPS-11 I OS&Y SUPPLY VALVE via ZS-l2201A Not Full open, and Input to Radwaste Collection Tanks will increase.

Question Data B

AR-036-B01, "PUMP (Fire) IS OPERATING" alarm will be received, and AR-036-BO5, "ENGINE RUNNING" alarm will be received and HVI 6150 Condenser Area Transfer Sump Isolation Valve closes.

ExplanatiodJustification:

A.

B.

C.

The candidate may believe that there will be an increase in flow to the radwaste, but the sump isolates as part of the fire response.

Thus there will be no sudden inrush of fire protection water to radwaste.

The motor driven firepump and the diesel driven firepumps will have started bringing in the two alarms The high flow from the flow switch is expected on fire suppression initiation in the area. The supply valve is normally open. A trouble alarm will result if the valve is not full open. The valve is a manual valve and will be open. The candidate may believe the valve operates on an initiation signal.

The candidate may believe that there will be an increase in flow to the radwaste, but the sump isolates as part of the fire response.

Thus there will be no sudden inrush of fire protection water to radwaste.

D.

Sys#

System Category KA Statement 600000 Plant Fire On Site Ability to operate andlor monitor the following as Fire alarm W M

~OOOOO.AAI.O~

WA Importance 3.013.0 Exam Level RO they apply to PLANT FIRE ON SITE:

(RO/S RO)

References provided to Candidate None Technical

References:

~ ~ 4 P - 0 0 1 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental I O CFR Part 55 Content:

41.7 I 45.6 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2291 Describe the operation of the following Task:

13.ON.

Fire components and controls of the Fire Protection System:

a. Diesel Fire Pump

b. Jockey Pump

c. Halon System

d. C02 System

e. SIMPLEX Panel

f. Electric Fire Pump

g. Preaction Sprinkler

h. Dry Pipe Sprinklers

i. Deluge System

j. Total Flooding C02

k. Manual Spurt C02

1. Total Flooding Halon

m. Manual C02

n. Wet Pipe Sprinkler

0. Status Lights and Alarms 003 Y

c LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

'V'

21.

A Unit I startup was in progress, with Main Condenser vacuum at I

.5" Hg Absolute and Reactor power at 22%.

- 'A' RFP feeding; 'B' RFP in recirc.

- Feedwater Level Control in Automatic.

- Main Generator output breaker closed.

- Extraction Steam in service.

Main Condenser vacuum has decayed to 10.0" Hg Absolute and is presently steady.

Assuming NO operator action in response to the lower vacuum, Reactor power will:

A.

decrease to 0% due to the Reactor Feed Pumps tripped and reactor scram.

B.

decrease to 0% due to the turbine trip and reactor scram.

C. increase to greater than 22% due to a loss of feed water heating.

D.

increase to greater than 22% due to the increase in Condensate temperature.

Question Data C

increase to greater than 22% due to a loss of feed water heating.

ExplanationIJustification:

A.

B.

C.

No loss of RFPs on loss of vacuum until 17.4" Hg Vacuum. The candidate may believe that the RFPs tripped and the reactor scrammed on low RPV water level. Vacuum has degraded but not to the point that RFPs will trip No scram on turbine trip, bypassed at power ~30%.

The candidate may not recognize that the scram due to a turbine trip is bypassed at less than 30% power Correct answer, Main Turbine trips at 8.2" Hg Absolute (21.7" Hg Vacuum) which will cause a loss of feedwater heating with the loss of extraction steam causing reactor power to rise. The reactor scram on a turbine trip is bypassed at e30 % power so no reactor scram. The loss of condenser vacuum Off Normal would be implemented and the off normal implements the Scram procedure for scram imminent actions.

The loss of condenser vacuum will cause condensate temperature to increase due to the change in condenser pressure (92 to162 degrees F). Increased condensate temperature will cause power to go down. Candidate may not recognize that the condensate temperature change will make power go down.

D.

Sys#

System Category KA Statement 295002 Loss of Main Condenser WA#

295002.~~2.02 KIA Importance 3.213.3 Exam Level RO Ability to determine andlor interpret the following as they apply to LOSS OF MAN CONDENSER VACUUM:

Reactor power: PlantSpecific Vacuum (ROE RO)

References provided to Candidate None Technical

References:

0~-143-001 Question Source:

Modified Susquehanna, 8/1/2003 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.10 143.5 I Objective:

3592 Respond to a loss of main condenser vacuum.

Task:

43.ON.

Implement Loss of Main 45.13 003 Condenser Vacuum LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam I---

22.

Unit 1 is operating at 100% power when the MAIN STEAM LINE 'A' FLOW, (FTC321 NOOZA),

Flow Transmitter develops a diaphragm leak causing delta P to equalize.

How will the plant respond to the 'A' Steam Flow Transmitter failure?

A.

Indicated Total Steam flow will be higher than Indicated Total Feed flow.

Actual Total Steam flow will be lower than actual Total Feed flow.

Water level rises to the High-Level Turbine trip setpoint.

B.

Indicated Total Steam flow will be higher than Indicated Total Feed flow.

RFP Speed initially Rises, and then returns to original speed.

Water level increases to a new higher level below High-Level Turbine trip setpoint.

C.

Indicated Total Steam flow will be lower than Indicated Total Feed flow.

Actual Total Steam flow will be greater than actual Total Feed flow.

Water level lowers to the scram setpoint.

D.

Indicated Total Steam flow will be lower than Indicated Total Feed flow.

RFP Speed initially Lowers, and then returns to original speed.

Water level lowers to a new lower level above scram setpoint.

x.

Question Data D

Indicated Total Steam flow will be lower than Indicated Total Feed flow.

RFP Speed initially Lowers and then returns to original speed.

Water level lowers to a new lower level above scram setpoint.

ExplanationlJustification:

A.

8.

C.

0 dp results in no flow so indicated steam flow will be less than indicated feed flow The candidate may believe 0 DP is high flow 0 dp results in no flow so indicated steam flow will be less than indicated feed flow. The candidate may believe 0 DP is high flow As water level lowers due to flow error, a level error develops (low level) that increases pump speed, Level stabilizes before the scram setpoint. The plant can survive a loss of a steam flow transmitter (25% of the total) but cannot survive a feed flow transmitter loss (50% of the total). The candidate may believe that the plant will scram.

0 dp corresponds to no flow. This results in total steam flow less than total feed flow. FWC reduces pump speed to try to match feed flow to steam flow. As pump speed drops, RPV level will begin to drop creating a level error that tries to speed up the feed pump.

The plant will stabilize at some point where the pump is at the original speed and at a lower level such that the flow error is balanced by a level error.

D.

Sys#

System Category KA Statement 295009 W M 295009.~~2.02 WA Importance 3.6/3.7 Exam Level RO Low Reactor Water Level Ability to determine andlor interpret the following as they apply to LOW REACTOR WATER LEVEL:

Steam flowlfeedflow mismatch (RO/S RO)

References provided to Candidate None Technical

References:

ON-145401 Question Source:

New Susquehanna, 8l42004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.10143.51 45.1 3 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

i-Objective:

SSES LOC 20 NRC Exam 1820 Predict how key parameters of the Reactor Task:

45.ON.

Implement Reactor Water Feedwater System, and the Plant will respond to failure of the following components or controls.

007 Level Anomaly

a. Master Controller

b. Individual RFP MIA Station

c. RPT Trip Pushbutton

d. Low Load Valve Controller

e. Hydraulic Jack

f. RFP Minimum Flow Controller

g. Motor Speed Changer

h. Level N B Selector Switch

i. SIU Bypass Valve Controller

j. Feedwater Controller

k. Steam Flow

1.

Feed Flow

m. RPV Level

n. One Elementrrhree Element Select Switch LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCFo~.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\Y

23. Unit 2 is operating at 100% power when an inadvertent MSlV isolation signal is initiated, causing the MSIVs to go closed.

Assuming no operator action, how and why will RPV level respond to this event?

A.

RPV Level will cycle between approximately -40 inches and +60 inches as the SRVs control pressure, and HPCl and RClC feed the reactor.

B.

RPV Level will cycle between approximately +20 inches and +50 inches as the SRVs control pressure, and the reactor feed pumps feed the reactor.

C.

RPV Level will slowly lower to -129 as SRVs cycle to maintain pressure, and level will not recover.

D.

RPV Level will cycle between approximately -40 inches and +60 inches, as HPCl and RClC control pressure and feed the reactor.

Question Data A

RPV Level will cycle between approximately -40 inches and +60 inches as the SRVs control pressure and HPCl and RClC feed the reactor.

Explanation/Justification:

A.

B.

C.

D.

Initially level will shrink, the setpoints are constant with RCIC. HPCl does not initiate until -38 inches.

RFPs not available due to MSlV closure. If RFPs were available, RPV level would cycle in this manner with SRVs controlling pressure and RFPs maintaining RPV level. The candidate may not realize that the MSlVs are closed and the RFPs are not available.

HPCI and or RCIC will maintain level so level will not drop much below -38 inches.

With no operator action, HPCl and RClC will not control pressure effectively. Left alone HPCl and RClC will overfeed and trip. The only means of pressure control left art that point will be SRVs. The candidate may believe that HPCl and RClC can automatically operate to control level and pressure. The systems will cyde between the upper and lower limits but will trip on high level and not control pressure while tripped.

Sys#

System Category UA Statement 295020 Inadvertent Containment Knowledge of the reasons for the following Reactor water level response Isolation responses as they apply to INADVERTENT CONTAINMENT ISOLATION:

WA#

295020.~~3.05 WA Importance 3.813.9 Exam Level RO (ROB RO)

References provided to Candidate None Technical

References:

FSAR 15.2, TM-OP-O~O Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 I 45.6 Objective:

2041 Describe how key parameters and components Task:

52.0P.O Perform Recovery From HPCl respond to automatic and manual initiation, isolation, or tripping of the HPCI System.

05 System Isolation Or Turbine Trip LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCF0rm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

24.

A small Loss of Coolant Accident (LOCA) has occurred on Unit I, and Suppression Pool level has been rising steadily and is presently at 25 feet.

- Suppression Pool temperature is 140 O F.

- RPV pressure is 900 psig.

EO-I 00-1 03, "PRIMARY CONTAINMENT CONTROL", step SP/L-11 directs the operator to ensure HPCl and RClC are running.

What is the basis for this step?

A.

To prevent a swap from the CST suction to the Suppression Pool B.

To reduce the amount of heat being added to the Suppression Pool C.

To prevent adverse consequences of an automatic start D.

To reduce RPV pressure and avoid violating the Heat Capacity Temperature Limit Question Data C

To prevent adverse consequences of an automatic start.

ExplanatiodJustification:

A.

B.

C.

Does not prevent a swap. Suctions can automatically swap to the suppression pool on low CST level or high SP level. The candidate may believe that manual start will prevent an auto suction swap.

Operating the systems will reduce the amount of heat to the SP as compared to the SRVs but not as helpful an BPVs. The candidate may believe that this is the reason. It is not the basis at this time.

The HPCI and RCIC turbine exhaust lines begin to flood at a suppression pool water level above 26. If either were to auto start with a flooded exhaust line, there is no guarantee that the systems would remain functional. Therefore, both HPCl and RClC are ensured to be running when pool level reaches 26. If the turbines are running, continued operation with levels above 26 will not result in adverse consequences. Adding heat to the suppression pool from HPCI and RCIC steam turbines is acceptable. Adding water to the suppression pool if HPCI and RClC are operating with minimum flow valves open is acceptable. If HPCl or RClC subsequently trip, restart is acceptable if the system is needed for adequate core cooling or pressure control.

can reduce RPV pressure but will not help HCTL since it also adds mass (Exhaust Steam) and heat to the SP. Reducing pressure helps one aspect of the curve but does not help with the approach to the HCTL.. The candidate may believe that this will help with HCTL.

D.

Sys#

System Category KA Statement 295029 High Suppression Pool Knowledge of the interrelations between HIGH HPCI: Plant-Specific Water Level SUPPRESSION POOL WATER LEVEL and the following:

WA#

295029.~~2.02 WA Importance 3.413.6 Exam Level RO (ROERO)

References provided to Candidate EOPs Technical

References:

EO-100-103 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.7,45.a LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\-

Objective:

284 Predict how ECCS Systems will be affected by Task:

59.ON.

Implement Primary any of the following Primary Containment 007 Containment Water Level parameters:

Anomaly

a. Suppression Pool Low Level

b. Suppression Pool High Level

c. Suppression Pool High Temperature

d. DW High Pressure

e. DW High Temperature

\\-

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCFonn.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

25.

Unit I has a Small line break LOCA in progress with the following conditions;

- Reactor Shutdown.

- Reactor Pressure 1,000 psig.

- HPCl is throttled to 500 gpm and is maintaining Reactor level at 35 inches.

- MSIVs closed.

- Drywell Pressure 17 psig.

- Drywell Temperature 235 O F.

- Suppression Pool Level 23.5 feet.

- Suppression Pool temperature 108 OF.

- RClC is available and not running.

- ADS 00s.

The following alarm and indications are received:

- AR-114-E05, "HPCI LEAK DETECTION HI TEMPlHI DlFF TEMP."

- HPCl Equipment Area inlet, outlet -1 N028A, 1 N029A reading 47 OF slowly increasing.

- HPCl Equipment Area inlet, outlet -1 N028B, 1 N029B reading 48 OF slowly increasing.

- HPCl Equipment Area - I N024A reading 122 OF slowly increasing.

- HPCl Equipment Area - 1 N024B reading 121 O F slowly increasing.

Determine the operator actions required and the basis for the actions.

A.

HPCl should continue to operate, since it is the only ECCS available to maintain adequate core cooling B.

Start RClC and isolate the HPCl Steam supply, which terminates heat addition from high-energy systems.

C.

Open all Main Turbine Bypass Valves in anticipation of Rapid Depressurization, to reduce the energy being discharged to the secondary containment.

D.

Open six SRVs to rapidly depressurize the reactor, to reduce decay heat levels and the energy that the RPV discharges to the secondary containment.

Question Data B

Start RCIC and isolate the HPCl Steam supply, which terminates heat addition from high-energy systems.

ExplanatiodJustification:

A.

B.

C.

D.

HPCl should be shut down to terminate the energy addition. RClC although not an ECCS is capable of providing high pressure water. Candidate must identify HPCl as source of the leak and determine if HPCl is needed to cany out mitigation strategies.

Max normal temperatures have been exceeded. The break is in the HPCl System, since RCIC, is available it should be started and HPCl Shutdown. ADS, and RFPs are not available. Thus, per SCTT-4 HPCl should be shutdown.

Directed when two areas have exceeded the Max Safe Temperature. If the candidate misreads the table and believes that 2 areas have not just exceeded the max normal but have exceeded the rnax safe, this is an option Directed when two areas have exceeded the Max Safe Temperature If the candidate misreads the table and believes that 2 areas have not just exceeded the max normal but have exceeded the max safe, this is an option

'U Sys#

System Category KA Statement LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 295032 High Secondary Knowledge of the reasons for the following Isolating affected systems

-v Containment Area responses as they apply to HIGH SECONDARY Tern perat u re CONTAINMENT AREA TEMPERATURE:

WA#

295032.~~3.03 WA Importance 3.813.9 (RO/SRO)

Exam Level RO References provided to Candidate EOPs Technical

References:

EO-100-104 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

Objective:

1276 Predict how each supported System will be Task:

OO.EO.0 Implement Secondary Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 2 41.5 I 45.6 affected by any of the following Secondary 28 Containment Control Containment System failures.

Supported System:

a. Reactor Building

b. ECCSlSafety System Components

c. Emergency Switchgear Rooms

d. Pipe Tunnels

e. Drywell Cooling System Failure:

a. Loss of Chiller Pumps

b. HVA Lockout Relays

c. Room Cooler

d. Recirculation Fans

e. Reactor Recirculation Pumps

f. Direct Expansion Chiller Units

g. High Energy System Break

h. High Radiation

i. High Temperature

j. High Level

k. Abnormal DIP LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCFom.d0~

Printed on 07/28/04

SSES LOC 20 NRC Exam

26. A refueling accident has occurred on the Refuel Floor causing the Refuel Floor Wall Exhaust Radiation level to reach 22 mR/hr.

f For the Control Room Emergency Outside Air Supply System (CREOASS), which one of the following actions is required by the Control Room operator?

A.

Verify automatic start in the Recirculation/lsolation mode.

B.

Verify automatic start in the Pressurization/Filtration mode.

C.

Manually start the system in the Pressurization/Filtration mode.

D.

Manually start the system in the Recirculatiotdlsolation mode.

Question Data B

Verify automatic start in the PressurizatiotVFiltration mode.

ExplanatiorVJustification:

A.

Does not start in the Recirculation Mode. The candidate may not know if the CREOASS will automatically or manually start and what mode of operation the system should be in. Thus the candidate is offered the combinations of manual and auto starts and pressurization and recirculation modes.

High Refuel Floor Wall Exhaust rad will automatically start in the Pressurization mode.

B.

C.

Starts automatically.

D.

Starts automatically Sys#

System Category KA Statement 295033 High Secondary Ability to operate andlor monitor the following as Control room ventilation:

Containment Area they apply to HIGH SECONDARY CONTAINMENT Plant-Specific Radiation Levels AREA RADIATION LEVELS:

K/A#

295~33.~~1.08 WA Importance 3.613.8 Exam Level RO (ROISRO)

References provided to Candidate None Technical

References:

TM-OP-OJO Question Source:

New Susquehanna, 0l412004 Level Of Difficulty: (1-5) 3 Question Cognitive Level :

Memory 10 CFR Part 55 Content:

41.7 145.6 Objective:

1965 List the interlocks and setpoints associated with Task:

34.0n.0 the following Control Structure 06 HVAC components:

a. CS Chiller

b. Contaminated Filter Exhaust Fan

c. CS Chilled Water Pump

d. Control Room KitchenlToilet Fan

e. CS Heating and Ventilation Fan

f. Computer Room Fan

g. Control Room Fans

h. CREOASS Fans

i. Access Control Area Toilet Exhaust Fan

k. Access Control General Area Exhaust Fan LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-

27. Unit 2 has declared an emergency situation as a result of an unisolable primary to secondary containment leak. A cooldown is in progress with the MSlVs closed.

Standby Gas Treatment System is in service with the following parameters;

- Secondary Containment delta P: -.31 WG

- SGTS SPING Noble Gas: 1.0 E6 uc/min

- OSCAR whole body dose readings: 0.05 mrem/hr A failure of a siding panel on the Refuel Floor occurs, causing Secondary Containment delta P to indicate 0 WG.

With the change in Secondary Containment delta P, how will the SPING readings relate to the offsite release rate, and how will OSCAR whole body dose readings respond?

A.

SBGT SPING Noble Gas - Representative of Total Offsite Release OSCAR whole body dose readings - No Change B.

SBGT SPING Noble Gas - Not representative of Total Offsite Release OSCAR whole body dose readings - No Change C.

SBGT SPING Noble Gas - Not representative of Total Offsite Release OSCAR whole body dose readings - Increase

-\\..-

D.

SBGT SPING Noble Gas - Representative of Total Offsite Release OSCAR whole body dose readings - Increase Question Data C

SBGT SPlNG Noble Gas - Not rewesentative of Total Offsite Release OSCAR whole body dose readings - Increase Explanation1Justification:

A.

B.

C.

D.

If the operator does not understand the operation of SBGT and that the loss of a panel will bypass SBGT then may believe that SPlNG is still representative.

The OSCAR readings will increase as the release increases due to the siding failure. If the operator does not understand the operation of OSCAR, then may believe that OSCAR will not see the increased release.

flow will bypass SBGT and thus it will not be representative of the actual release The OSCAR readings will increase as the release increases due to the siding failure.

If the operator does not understand the operation of SBGT and that the loss of a panel will bypass SBGT then may believe that SPlNG is still representative.

Sys#

System Category KA Statement 295035 Secondary Containment Ability to determine andlor interpret the following as Off-site release rate: Plant-High Differential Pressure they apply to SECONDARY CONTAINMENT HIGH Specific DIFFERENTIAL PRESSURE:

KIA#

295035.~~2.02 WA Importance 2.814.1 Exam Level RO (ROISRO)

References provided to Candidate None Technical

References:

TM-OP-070 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3

'v Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.8 to 41.10 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

1266 Define andlor discuss the operational implications Task:

34.ON.

Implement Loss Of Reactor of the following terms:

006 Building Differential Pressure

a. Radiation Releases

b. Secondary Containment Integrity

t. Equipment Qualificaton

d. Protection of PubliclPersonnel LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam V

28.

Unit 1 is in Mode I, and Electrical Maintenance has requested permission to remove the Limitorque Cover for the RHR LOOP A INJECTION FLOW CONTROL VLV, HV-151 F017A.

The request is to visually monitor limit switch operation, while the valve is stroked for the SO-I 49-AO5, QUARTERLY RHR LOOP A VALVE EXERCISING.

Prior to stroking and monitoring the RHR LOOP A INJECTION FLOW CONTROL VLV, HV-151 FO17A, what LCO and/or TRM entries must be entered?

A.

Technical Specification 3.5.1 Condition A Technical Requirements 3.8.2.1 and 3.8.2.2 Condition A B.

Technical Specification 3.5.1 Condition B Technical Requirements 3.8.2.2 Condition A C.

Technical Specification 3.5.1 Condition A Technical Requirements 3.8.2.1 Condition A D.

Technical Specification 3.5.1 Condition B Technical Requirements Condition 3.8.2.1 Condition A Question Data C

Technical Specification 3.5.1 Condition A Technical Requirements 3.8.2.1 Condition A

\\-

Explanation/Justification:

A.

B.

Valve is listed in TRM 3.8.2.1 not TRM 3.8.2.2. If the candidate does not check the tables at the end of each TS, this answer may be chosen.

Condition B applies if one RHR Pump is inop. In this case it is due to the valve being stroked and both pumps in a LPCl loop being inoperable and thus Condition A (other than condition B) applies If the candidate misinterprets TS 3.8.2.1, this answer may be chosen.

Valve is listed in TRM 3.8.2.1 not TRM 3.8.2.2 LPCl is inop due to cover removed To manually stroke the valve the thermal overload is unbypassed, This valve is listed in TRM 3.8.2.1 Condition B applies if an RHR Pump is inop. In this case it is due to the valve being stroked and thus Condition A (other than condition B) applies. If the candidate misinterprets TS 3.8.2.1, this answer may be chosen.

C.

D.

Sys#

System Category 203000 RHWLPCI: Injection Mode Equipment Control (Plant Specific)

KIA#

203000.2.2.24 WA Importance 2.613.8 (RO/SRO)

Exam Level KA Statement Ability to analyze the affect of maintenance activities on LCO status.

RO References provided to Candidate TS & TRM Technical

References:

TS 3.5.1 8 3.8.2 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

43.2 145.13 Objective:

3137 Determine if a component or system is required to Task:

OO.TS.0 Apply Technical Specification be operable per Technical Specifications.

03 (TS) And Technical Requirements Manual (TRM)

Requirements LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

29.

The plant is in MODE 4.

While operating RHR Loop 'B using the ' B Pump in the Shutdown Cooling mode, the following alarm is received:

AR-113-CO9, HV-I51-F006B/D AND HV-151-FO07B OPEN DRAIN RX VESSEL The following indications are observed:

- HV-151-F007B, RHR PUMPS B&D DISCHARGE MIN FLOW VLVTO SUPP POOL indicates OPEN WITH Red Light LIT, Amber Light OUT.

- HV-151-F006BJ RHR SHUTDOWN COOLING RHR PUMP B SUCTION VLV indicates OPEN WITH Red Light LIT, Amber Light OUT.

- PI-El 1-1 R606B1, INLET PRESSURE TO HX B indicates pressure oscillating from -50 to

-300 psig.

- FI-Ell-1 R603B, RHR LOOP B indicates flow oscillating from zero to 100 gpm.

- Pll5194B, RHR PUMP B SUCTION PRESSURE after being valved in, indicates full minimum to maximum deflection on the gauge.

Explain what has occurred in the 'B RHR System, what immediate operator actions are required; what procedure will be used to correct, control, or mitigate the consequences of the observed conditions?

A.

The disc for the HV151 F006B RHR SHUTDOWN COOLING RHR PUMP B SUCTION VLV has come off the stem, OPEN HV151F006D, use OP-149-002 RHR SHUTDOWN COOLING, TRANSFER OPERATING PUMPS WHILE IN SDC to start ID' RHR pump.

B.

The disc for the HV151 F006B RHR SHUTDOWN COOLING RHR PUMP B SUCTION VLV has come off the stem, trip the ' B RHR pump and enter ON-149-001, LOSS OF RHR SHUTDOWN COOLING MODE.

C. HV-151-FO17 B RHR INJ FLOW CTL has inadvertently closed, use OP-149-002 RHR SHUTDOWN COOLING, TRANSFER OPERATING PUMPS WHILE IN SDC to start ID' RHR pump, stop the 'B' RHR pump.

D.

HV-151-FOI 7 B RHR INJ FLOW CTL has inadvertently closed, trip the 'B' RHR pump and enter ON-149-001, LOSS OF RHR SHUTDOWN COOLING MODE, start 'Dl RHR Pump-Question Data B

The disc for the HV151F006B RHR SHUTDOWN COOLING RHR PUMP B SUCTION VLV has come off the stem, trip the

'B' RHR pump and enter ON-149-001, LOSS OF RHR SHUTDOWN COOLING MODE.

ExplanatiodJustification:

A.

The 'B' pump needs to be tripped as soon as possible to prevent impeller damage due to heating caused by no suction path.

Tripping the '6 RHR pump will immediately require entry into the loss of shutdown cooling procedure. The candidate may analyze the situation properly and recognize D RHR pump is available but not recognize that the ON must be entered.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam B.

correct, the indications show the min flow valve open, open suction path from reactor but min flow valve alarm and drain-down path to suppression pool alarm. Min flow valve is open on low flow as indicated by the discharge flow indicating low. The drain down alarm is based on a suction path from the reactor F006 open with min flow valve open which is the present indication. The indication of suction pressure low and low flow is indicative of a loss of suction path. A loss of suction path requires the pump to be tripped since the pump trip is from the limit switch on the F006 valve.

The situation presented is a loss of suction path as indicated by suction pressure. Closure of the injection valve would produce similar indications with low flow causing the rnin flow valve to open, but not the low suction pressure and deviating discharge flow. If the candidate rnisanalyzes the situation, this answer may be chosen.

The situation presented is a loss of suction path as indicated by suction pressure. Closure of the injection valve muld produce similar indications with low flow causing the min flow valve to open, but not the low suction pressure and deviating discharge flow. If the candidate misanalyzes the situation, this answer may be chosen.

L.

C.

D.

Sys#

System Category 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

Ability to monitor automatic operations of the SHUTDOWN COOLING SYSTEMIMODE including:

WA#

205000.~3.03 WA Importance 333.3 Exam Level References provided to Candidate None Technical Referen (RO/SRO)

KA Statement Lights and alarms RO es:

AR-II~-OOI, ON-I~~-OO Question Source:

Modified Perry 1,3/1/2002 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 I 45.7 Objective:

2067 List the interlocks for the following components Task:

49.ON.

and controls of the RHR System, including setpoints:

003

a. RHRPumps

b. Heat Exchanger Bypass Valve

c. Injection Valves

d. Suppression Pool Test and Spray Valves

e. Drywell Spray Valves

f. Radwaste valves

g. Pump Suction and Shutdown Cooling Valves

h. Minimum Flow Control Valves LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-Y

30. Following a valid HPCl initiation due to high Drywell pressure on Unit 2, HPCl turbine tripped on high water level. Water level has dropped to +IO inches with Drywell pressure at 12 psig, and the Unit Supervisor has directed that HPCI be restarted.

When the HPCl HI WTR LVL TRIP RESET HS-E41-1S25 pushbutton is depressed and no other operator actions are taken, how will the HPCl Turbine respond, and why?

The HPCl Turbine will:

A.

attain rated speed. When the reset pushbutton is depressed, the HPCl TURB STEAM SUPPLY VLV, HVl55F001 will open, and the ramp generator will initially control speed.

When the ramp generator has increased turbine speed so that system flow is equal to the flow controller setpoint, then the flow controller will control speed.

B.

attain rated speed. When the reset pushbutton is depressed, the HPCl TURB STOP VLV, FV15612 will open, and the ramp generator will initially control speed. When the ramp generator has increased turbine speed so that system flow is equal to the flow controller setpoint, then the flow controller will control speed.

C.

trip on overspeed. When the reset pushbutton is depressed the HPCl TURB STOP VLV, FV15612 will open, and flow control will not react fast enough to prevent overspeed.

D.

trip on overspeed. When the reset pushbutton is depressed the HPCl TURB STEAM SUPPLY VLV, HV155F001 will open, and flow control will not react fast enough to prevent overspeed.

Question Data B

attain rated speed. When the reset pushbutton is depressed, the HPCl TURB STOP VLV, FV15612 will oDen and the ramD generator will initially control speed. When the ramp generator has increased turbine speed so that system flow is equal to the flow controller setpoint then the flow controller will control speed.

ExplanationlJustification:

A.

B.

C.

Fool valve is open and does not open on reset. The ramp generator is driven by the opening of N15612. If the operator does not know the position of Fool and HV15612 after high water level or what initiates the ramp generator, then this answer may be chosen.

Stop valve closed on high RPV level, reopening the stop valve causes the ramp generator to control the start and prevent an overspeed.

The ramp generator will be in control and will limit speed on the start to prevent an overspeed. When the stop valve begins to open a ramp generator produces a control signal that causes the turbine to increase speed in a control manner to prevent it from overspeeding. If the candidate does not understand the initiation of the ramp generator, then this answer may be chosen.

The ramp generator will be in control and will limit speed on the start to prevent an overspeed When the stop valve begins to open a ramp generator produces a control signal that causes the turbine to increase speed in a control manner to prevent it from overspeeding. If the candidate does not understand the initiation of the ramp generator, then this answer may be chosen.

D.

Sys#

System Category KA Statement 206000 High Pressure Coolant Knowledge of HIGH PRESSURE COOLANT Turbine speed control: BWR-2, Injection System INJECTION SYSTEM design feature(s) andlor 3,4 interlocks which provide for the following:

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

206000.K4.11 SSES LOC 20 NRC Exam KIA Importance 3.413.5 (ROE RO)

Exam Level RO References provided to Candidate None Technical

References:

TM-OP-052 Question Source:

Modified Peach Bottom 2,9/1311999 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 Objective:

2914 Predict how key parameters of the system and the Task:

52.0P.O Perform Recovery From HPCI plant will respond to manipulating the following controls for HPCI.

Trip

a. MOV OL bypass switches

b. Flow controller

c. Manual initiation pushbutton

d. Initiation signal reset pushbutton

e. System valve switches

f. Condenser vacuum pump

g. Turbine trip plb 05 System Isolation Or Turbine LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-.

31.

Unit 1 has experienced a Loss of Coolant Accident outside the primary containment.

Present plant conditions:

- Reactor level

-70 inches

- Reactor Pressure 350 psig

- Drywell Pressure 0.5 psig

- Drywell temperature 130 O F

- Offsite power available.

The Unit Supervisor has directed manual initiation of both loops of Core Spray.

What actions are required to fulfill the directive and what is the plant response?

A.

B.

C.

D.

Arm and Depress either CORE SPRAY LOOP MAN INlT HS-E211 S16A or CORE SPRAY LOOP MAN INlT HS-E211 S16B pushbutton.

All four Core Spray Pumps start 10.5 seconds after D/Gs start.

A and C D/Gs start.

Unit 2 A and B Core Spray Pumps trip if running.

Arm and Depress both CORE SPRAY LOOP A and B MAN INlT HS-E211S16A and B pushbuttons.

All four Core Spray Pumps start 10.5 seconds after D/Gs start.

A and C D/Gs start.

Unit 2 A and B Core Spray Pumps trip if running.

Arm and Depress either CORE SPRAY LOOP MAN INIT HS-E21 I S16A or CORE SPRAY LOOP MAN INlT HS-E21IS16B pushbutton.

All four Core Spray pumps start after 15 seconds.

A and B D/Gs start.

Unit 2 A & C Core Spray pumps trip if running.

Arm and Depress both CORE SPRAY LOOP A and B MAN INlT HS-E211 S16A and B pushbuttons.

All four Core Spray pumps start after 15 seconds.

A & B D/Gs start.

Unit 2 A and C Core Spray Pumps trip if running.

Question Data D

Arm & Depress both CORE SPRAY LOOP A & B MAN INlT HS-E211S16A & B push buttons.

All four Core Spray pumps start after 15 seconds.

A & 0 DIGS start.

Unit 2 A 8, C Core Spray pumps trip if running.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam ExplanatiodJustification:

A.

L--

The 10.5 second Core Spray pump start times apply if on emergency power. Both push buttons must be armed and depressed per procedure. Pushing only one button will initiate only one CS loop and one EDG the Busses are on normal power and thus will start in about 15 sec. The candidate may confuse the start times. The candidate may also confuse which EDGs will start.

The 10.5 second Core Spray pump start times apply if on emergency power (EDG). The Busses are on normal power and thus will start in about 15 sec. The candidate may confuse the start times. The candidate may also confuse which EDGs will start.

Both push buttons must be armed and depressed per procedure. Pushing only one button will initiate only one CS loop and one EDG The candidate may not understand the difference between pushing one or both PB, and chose this answer.

Per procedure, both division I and 2 diesels start (A&B) and with offsite power available pumps will start in apprx 15 seconds.

6.

C.

D.

Sys#

System KA Statement Category 209001 Low Pressure Core Spray System control room:

Ability to manually operate andlor monitor in the Manual initiation controls K/A#

209001.~4.05 WA Importance 3.813.6 (RO/SRO)

Exam Level RO References provided to Candidate None Technical

References:

0~-151-001 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 145.5 to 45.8 Objective:

2082 Describe the logic schemes used for automatic Task:

51.0P.O Perform Manual Operation Of and manual initiation, isolation, and tripping of the Core Spray System.

02 Core Spray System LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

--=

32. The reactor is operating at 50% power. The SBLC injection line, pipe within a pipe, has completely broken off on the inside of the reactor vessel at the penetration.

Which of the following describes a consequence of this failure?

A.

B.

C.

D.

Jet Pump flow will indicate zero.

RWCU Bottom Head Drain Flow indication will fail downscale.

CRD Drive Water flow indication will fail downscale.

Core Spray Loop Break Detection differential pressure will go positive.

Question Data D

Core Spray Loop Break Detection differential pressure will go positive.

ExplanationIJustification:

A.

B.

C.

D.

JPs use the below core plate pressure and this should not change. If the candidate believes that JPs use above core plate then the jet pump DP would go down RWCU uses the below core plate pressure and this should not change since only the below core plate pressure is now sensed If the candidate does not understand what pressure RWCU uses, this answer may be chosen.

The CRD System uses the above core plate pressure for DPs but not for flow. The candidate may believe that the CRD system use the above core plate tap for a flow element DP The DP detector is normally negative and indicates a break when it becomes positive. The break will change the HP side of the detector from above core plate pressure to the higher below core plate pressure. This will make the indication go positive and bring in the alarm..

Sys#

System Category KA Statement 211000 Standby Liquid Control Knowledge of the effect that a loss or malfunction of on following:

Core spray line break detection System the STANDBY LIQUID CONTROL SYSTEM will have system: PlantSpecific KIA#

211000.~3.02 WA Importance 3.013.2 (RO/SRO)

Exam Level RO References provided to Candidate None Technical

References:

TM-OP-OIO Question Source:

Modified LaSalle 1,1112012000 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.7 I 45.4 Objective:

1480 Predict how RPV Instrumentation parameters and Task:

components will be affected by a failure of any of the following support systems.

a. Primary Containment

b. RPV (coolant leak)

c. SBLC Injection Sparger

d. AC Distribution

e. DC Distribution LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

.--- 33. Unit I is in Mode 2 when the high voltage power supply for IRM E is lost.

- The RPS Shorting links are installed.

What is the plant response for the power loss to IRM E?

An IRM INOP trip will occur and produce:

A.

B.

C.

D.

a Rod Block. No RPS Scram signal will be produced.

a full reactor scram and a Rod Block.

a reactor scram signal in the A RPS logic and a Rod Block.

a reactor scram signal in the B RPS logic and a Rod Block.

Question Data C

a reactor scram signal in the A RPS logic and a Rod Block ExplanationfJustification:

A.

B.

A half scram will be produced on the IRM failure unless the IRM is bypassed or the reactor is in Mode 1. If the candidate confuses the IRM inop with the SRM inop this answer may be chosen.

This will occur if the shorting links are removed. The shorting links place the neutron monitoring system in non-coincident mode where a single NMS trip in RPS will give a full scram If the candidate does not understand the impact of the shorting links being removed v installed, this answer may be chosen HV low is an INOP trip and results in a half scram and a rod block IRM E feeds the A RPS logic There will be no impact on the RPS B logic. If the candidate confuses the RPS divisions, this answer may be chosen.

C.

0.

Sys#

System Category KA Statement 212000 Reactor Protection System Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR PROTECTION SYSTEM:

RPS sensor inputs KIA#

212000.~6.05 KIA Importance 3.513.8 (ROB RO)

Exam Level RO References provided to Candidate None Technical

References:

~ ~ - 0 p - 0 7 8 ~

Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.7 I 45.7 Objective:

2338 Predict how the following IRM failures affect each Task:

78.AR.0 Respond To Nuclear supported system.

06 Instrument Failure IRM

a. IRM INOP (inoperative) or UPSCALE

b. IRM DOWNSCALE LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\--

34.

A reactor scram has occurred due to a turbine trip. All immediate operator actions have been completed. All control rods are at 00, and RPV level has been stabilized at +35 inches.

In order to RESET the reactor scram, which of the following annunciators MUST be illuminated?

A.

AR-104-FO3, "DISCHARGE VOLUME HI WATER LEVEL TRIP BYPASS AR-104-HO1, "NEUTRON MON AID DISPLAYED" B.

AR-104-H01, "NEUTRON MON AID DISPLAYED" AR-103-AO3, "RPS MAN SCRAM CHANNEL AllA2 SWITCH ARMED AR-104-AO3, "RPS MAN SCRAM CHANNEL B1/B2 SWITCH ARMED" C.

AR-104-CO3, "MODE SWITCH SHUTDOWN SCRAM BYPASS AR-103-FO2, "RPS CHAN AI/A2 SCRAM DSCH VOL HI WTR LEVEL TRIP" AR-104-FO2, "RPS CHAN B1/B2 SCRAM DSCH VOL HI WTR LEVEL TRIP" D.

AR-104-C03, "MODE SWITCH SHUTDOWN SCRAM BYPASS AR-104-FO3, "DISCHARGE VOLUME HI WATER LEVEL TRIP BYPASS Question Data D

AR-104-C03, "MODE SWITCH SHUTDOWN SCRAM BYPASS AR-104-FO3, "DISCHARGE VOLUME HI WATER LEVEL TRIP BYPASS"

--~

ExplanationlJustification:

A.

B.

C.

D.

Neutron Mon Aid not required. If the candidate does not understand the purpose of the NEUTRON MONITORING AID alarm this answer may be chosen.

Man scram switches do not have to be armed to reset the scram. The neutron aid alarm is not necessary.. If the candidate does not understand the purpose of the NEUTRON MONITORING AID and the manual switches armed alarm this answer may be chosen.

The disch volume alarms will be in but are not required to be in for reset. The candidate may confuse present with required to be in.

The logic requires the mode switch to have been in shutdown for 10 seconds to ensure a complete scram and the operator must bypass the SDV high level scram Sys#

System Category KA Statement 212000 Reactor Protection Ability to monitor automatic operations of the System status lights and System REACTOR PROTECTION SYSTEM including:

alarms W M 212000.~3.04 WA Importance 3.913.8 (RO/S RO)

Exam Level RO References provided to Candidate None Technical

References:

~ ~ - 0 p - 0 5 8 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental I O CFR Part 55 Content:

41.7 145.7 Objective:

2485 List the interlocks and setpoints associated with Task:

58.0P.O Perform Reset Of Scram And the following Reactor Protection System 03 ARI components:

a. Scram Reset Switch

b. Reactor Mode Switch

c. Scram Discharge Volume (SDV) High Level Bypass Switch LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-\\.

35.

Unit 1 is in Mode 2 with the following conditions:

-- Control Rod withdrawals are in progress.

-- IRM Channel 'A' indication is on Range 8 reading 75/125 and rising.

-- The PCOM is UNABLE to range up IRM Channel 'A' to Range 9 (the IRM remains on Range 8).

As reactor power continues to rise, which one of the following describes the response of IRM Channel 'A' with no operator action, AND what actions are required to continue the startup?

A.

An IRM control rod block and half scram trip signal would be generated.

Withdraw the IRM Channel 'A' Detector to maintain indication between 25/125 and 75/125.

B.

Only an IRM control rod block trip signal would be generated.

Bypass IRM Channel 'A' on Panel 1 C651.

C.

No automatic IRM protective actions would occur.

No additional actions are required.

D.

An IRM control rod block and half scram trip signal would be generated.

Bypass IRM Channel 'A' on Panel 1 C651.

~~

~

Question Data D

An IRM control rod block and 1/2 scram trip signal would be generated.

Bypass IRM Channel 'A' on panel 1C651.

ExpIanatiodJustification:

A.

There is no procedural guidance to withdraw the IRM detector to maintain indication between 25/125 and 75/125 due to a valid failure. This will also produce a rod block since the detector will not be full in with the Rx Mode switch not in RUN. The candidate may recognize that this would prevent a trip but not realize that it is not permitted by procedure and would produce a rod block.

As power increases it will first reach the rod block setpoint 108/125 and will continue to increase to the scram setpoint 120/125. The operator may believe that power would not continue to rise, but the stem says that power will increase.

A rod block and half scram are produced. The IRM must be bypassed. The candidate may believe the Rx Mode switch is in mode 1 which will bypass the IRM scram and rod block Correct answer. IRM rod block and 112 scram trip signals are generated due to the reactor mode switch in startup and IRM on Range 8.

B.

C.

D.

Sys#

System Category 215003 Intermediate Range Equipment Control Monitor (IRM) System WA#

215003.2.2.2 WA Importance 4.013.5 References provided to Candidate None Question Source:

Exam Bank Susquehanna, 6/1/2003 Question Cognitive Level:

Analysis

( RO/S RO)

.L KA Statement Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

Exam Level RO Technical

References:

TM-0~4788 Level Of Difficulty: (1 -5) 3 45.2 10 CFR Part 55 Content:

LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2339 List the signals and setpoints that cause the IRMs Task:

78.AR.O Respond To Nuclear to initiate automatic trip, interlock and alarm functions.

06 Instrument Failure IRM LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\u

36. A Unit 1 startup is in progress following a refueling outage with IRMs and SRMs fully inserted.

The neutron monitoring shorting links are installed. The reactor is at the point of adding heat.

IRMs A,C,D,E, and F are on Range 8. IRMs B,G, and H are on Range 7.

Describe the plant response if the "A" SRM high voltage power supply failed low:

A.

Only a control rod block would occur.

B.

Both a control rod block and a half reactor scram would occur.

C.

D.

No automatic actions would occur.

A full scram would occur.

3uestionData A

only a control rod block would occur ExplanationlJustification:

A.

B.

SRM INOP due to HV Low produces a rod block but does not input to RPS.

SRM INOP due to HV Low produces a rod block but does not input to RPS. This is unlike the lRMs or APRMs where a detector hop is a scram signal The candidate may confuse the impact of an IRM inop with the shorting links installed with this situation and choose this answer.

If the IRM G was on range 8 this would be a correct answer since the SRM INOP function would be bypassed. When all of the Div I IRMS are on range 8, the rod block signals from the associated SRMS (A&C) are blocked. The candidate may believe the IRMs are at a level to bypass the SRM The scram signal from the SRMs is armed by having the shorting links removed but the SRM INOP signal does not produce a scram.

The candidate may confuse the impact of shorting links installed and removed and the response to an IRM inop with the SRM inop and choose this answer.

C.

-u.

D.

Sys#

System Category KA Statement 215004 Source Range Monitor Knowledge of the effect that a loss or malfunction of RPS the SOURCE RANGE MONITOR (SRM) SYSTEM will have on following:

(SRM) System WA#

215004.K3.01 WA Importance 3.413.4

( RO/S RO)

Exam Level RO References provided to Candidate None Technical

References:

TM-OP-O~~A Question Source:

Exam Bank Hatch 1,3/14/1997 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.7 I 45.4 Objective:

1334 Predict how the following SRM failures affect each Task:

78.AR.0 Respond To Nuclear supported alarm.

05 Instrument Failure SRM

a. SRM INOP (inoperative) or DOWNSCALE

b. SRMUPSCALE

c. SRM SHORT (or FAST) PERIOD LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\-

37. Unit I is at 90% power, normal plant conditions when a single LPRM input to the " A APRM is bypassed. All other LPRMs inputting to "A" APRM are not bypassed. The LPRM bypassed is the highest LPRM input to the APRM.

When the LPRM is bypassed, how will the " A APRM indication change, and why?

A.

B.

C.

APRM indicated output will remain the same. Bypassing the LPRM reduces the input signal to the APRM and reduces the number of LPRMs seen by the count circuit that increases the output signal.

APRM indicated output will remain the same. Bypassing the LPRM does not change the actual core power. Since the actual core power has not changed, the average power across the core will not change and the output will be the same.

APRM indicated output will decrease. Bypassing the LPRM reduces the signal input to the APRM by more than the average value. The count circuit sees fewer inputs, and causes an increase in the output signal, but not enough of an increase to return the APRM output to its initial value.

APRM indicated output will decrease. Bypassing any LPRM reduces the signal input to the APRM, and any reduction in input causes a reduction in the output signal, since the gain is constant until the next calibration.

Question Data C

APRM indicated output will decrease. Bypassing the LPRM reduces the signal input to the APRM by more than the average value. The count circuit sees fewer inputs and causes an increase in the output signal but not enough of an increase to retum the APRM output to its initial value.

ExplanationlJustification:

A.

B.

C.

The candidate must realize that bypassing the highest value causes the average value to drop and will result in a lower output.

If the candidate believes that the APRM indicates average power without regard for the APRM inputs then this answer may be chosen correct answer. The LPRM input bypassed will lower the input by more than the average. The count circuit will cause an increase in the average output but the count circuit will not boost the signal enough to compensate for losing the highest signal. If the lowest signal input was lost the APRM signal would increase.

The candidate must recognize that depending on the value of the LPRM bypassed, the APRM signal may increase, decrease or stay the same.

D.

Sys#

System Category KA Statement 21 5005 Average Power Range MonitorlLocal Power control room:

functioning/ operability Range Monitor System Ability to manually operate andlor monitor in the Verification of proper W M 215005.~4.06 WA Importance 3.613.8

( RO/S RO)

Exam Level RO References provided to Candidate None Technical

References:

TM-OP-078D Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.7 145.5 to Question Source:

Exam Bank Susquehanna, 6/1/2002 Level Of Difficulty: (1-5) 3 45.8 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\-

Objective:

2355 Predict how the LPRMs and supported systems Task:

78.AR.0 Respond To Nuclear respond to manipulating the following LPRM and related controls:

07 Instrument Failure APRM

a. LPRM Card Function Switch

b. LPRMlAPRM Meter Switches to select LPRM inputs

c. LPRM Trip Reset pushbutton LOC 20 As Given H:\\ExamBank\\MergeDoc\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-L

38.

Unit 2 RClC is maintaining water level constant at +30 inches in manual as the result of a loss of feedwater and a reactor scram.

The RClC Controller is set for 500 gpm and indicates 55% signal output.

The RClC speed indicates 3,500 rpm.

The Unit Supervisor has given the direction to cool down at 90 O F per hour using SRVs.

Assuming no operator adjustments to the RClC System, what will be the response of RClC turbine speed and RPV water level as the plant is cooled down from 500 to 400 O F ?

A.

RClC speed will decrease.

RPV water level will decrease.

B.

RClC speed will remain essentially constant.

RPV water level will decrease.

C.

RClC speed will increase.

RPV water level will increase.

D.

RClC speed will remain essentially constant.

RPV water level will increase.

Question Data D

RCIC speed will remain essentially constant.

RPV water level will increase ExplanationfJustification:

A.

B.

C.

In manual the operator has set pump speed, thus pump speed will remain constant. If the candidate thinks the turbine is in flow control this answer will be chosen As RPV pressure drops due to the cooldown, the RCIC pump will pump more water, increasing flow to the vessel, and raising water level. If the candidate does not understand the pump curves this answer will be chosen In manual the operator has set pump speed, thus pump speed will remain constant. If the candidate believes that the turbine is in flow control and considers only the drop in steam pressure to the turbine and not the drop in discharge pressure, this answer will be chosen In manual the operator has set pump speed, thus pump speed will remain constant. As RPV pressure drops due to the cooldown, the RClC pump will pump more water, increasing flow to the vessel, and raising water level D.

Sys#

System Category KA Statement 217000 Reactor Core Isolation Ability to manually operate andlor monitor in the Reactor water level W M 217000.~4.05 KIA Importance 4

Exam Level RO Cooling System (RCIC) control room:

(RO/S RO)

References provided to Candidate None Technical

References:

TM-OP-050 Question Source:

New Susquehanna, 81412004 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.7 145.5 to 45.8 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

L.'

Objective:

SSES LOC 20 NRC Exam 2013 Predict how key parameters of the Reactor Core Task:

50.0P.O Perform AutomaticlManual Isolation System and the Plant will respond to failure of the following RClC components or controls:

10 Startup Of RClC System

a. Flow detector

b. Steam line pressure sensor

c. Flow Controller

d. Steam flow detector

e. System valves

f. Barometric Vacuum Pump

g. Lubelcontrol oil

h. Room Cooler

i. Exhaust diaphragm

j. Ramp Generator k.Speed Control Circuit LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L.--

39. Which one of the following RClC features is designed to minimize water hammer?

A.

RClC Exhaust Line Vacuum Breakers [RCIC TURB EXH VAC BKR TEST CKV, 149-F063 and 149-FO641 B.

RClC Exhaust Check Valve [RCIC TURB EXH CKV TO SUPP POOL, 149-FO401 C.

RClC Rupture Diaphragms [RCIC TURBINE EXHAUST VENT, PSE-E51-1 DO01 and PSE-E51-1 DO011 D.

Seven-second delay in opening STEAM TO RClC TURBINE, HV-150-FO45 when the valve reaches 40% open Question Data A

RClC Exhaust line vacuum breakers [RCIC TURB EXH VAC BKR TEST CKV, 149-FO63 and 149-FO641 Explanation/Justification:

A.

B.

C.

2 inch line allows Suppression Chamber air into the RClC exhaust piping as the steam in the line condenses The check valve acts as primary containment isolation. In the event of an exhaust line break it prevents release of suppression chamber material out of the primary containment. The candidate may confuse this valve with the vacuum breakers The rupture discs prevent over pressurization and do not operate until 150 and 165 psig. These rupture discs protect the exhaust piping from damage but from over pressurization, not water hammer. The candidate may confuse the purposes and choose this answer.

Designed to prevent turbine overspeed. This does occur but does not minimize water hammer. The candidate may confuse the purposes and choose this answer.

L--

D.

Sys#

System Category KA Statement 217000 Reactor Core Isolation Knowledge of REACTOR CORE ISOLATION COOLING Prevent water hammer: Plant-which provide for the following:

Cooling System (RCIC)

SYSTEM (RCIC) design feature@) and/or interlocks Specific WA#

217000.~4.oi WA Importance 2.812.8 Exam Level RO (ROISRO)

References provided to Candidate None Technical

References:

TM-OP-050 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Memory 10 CFR Part 55 Content:

41.7 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam i.-

Objective:

2007 Describe the operation of the following Task:

50.0P.O Perform Normal Setup Of components and controls of the RClC System:

06 RClC System For Automatic Response Alignment

a.

Trip Pushbutton

b.

Manual Initiation P/B

c.

Flow Controller

d.

Manual Isolation P/B

e.

System valves

f.

g. MOV OL Bypass Switch

h. Isolation Reset Key Switches

i. RClC Turbine

j.

RClCPump

k. Lights and alarms

1.

Lube Oil Cooler

m. Barometric Condenser

n. Flow detector

0. Ramp Generator

p. Woodward Governor

q. Speed Control Circuit High Water Level Trip Reset PIB LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

'L...

40.

With Unit 1 at 100% power, an ADS SRV is manually opened for 15 seconds.

What will be the status of the following alarms when the valve is closed, and what will the SRV/ADS Temperature Recorder (TR-B21-1 R614) at Panel IC614 read for the affected valve?

A.

AR-1 10-EO1, 'I MAIN STEAM SRV LEAKING1 AR-110-E02, 'I MAIN STEAM DIV 1 SRV OPEN" AR-110-E03, "MAIN STEAM DIV 2 SRV OPEN" Alarm clears Remains in alarm Remains in alarm TR-B21-1 R614 250 O F

6.

AR-1 IO-EOI, 'I MAIN STEAM SRV LEAKING" AR-110-E02, I' MAIN STEAM DIV 1 SRV OPEN" AR-110-E03, "MAIN STEAM DIV 2 SRV OPEN" Remains in alarm Alarm clears Alarm clears TR-621-1 R614 340 O F C. AR-1 10-EO1, 'I MAIN STEAM SRV LEAKING" AR-I IO-E02, 'I MAIN STEAM DIV 1 SRV OPEN" AR-110-E03, "MAIN STEAM DIV 2 SRV OPEN" Remains in alarm Alarm clears Alarm clears TR-B21-1 R614 280 O F D.

AR-1 IO-EOI, I' MAIN STEAM SRV LEAKING" Alarm clears AR-110-E02, I' MAIN STEAM DIV 1 SRV OPEN" AR-110-E03, "MAIN STEAM DIV 2 SRV OPEN" Remains in alarm Remains in alarm TR-B21-lR614 540 O F Question Data C

AR-1 IO-EOI, " MAIN STEAM SRV LEAKING" Remains in alarm AR-110-E02, 'I MAIN STEAM DIV 1 SRV OPEN" Alarm clears AR-110-E03, "MAIN STEAM DIV 2 SRV OPEN" Alarm clears TR-B21-1R614 280 deg F ExplanationNustification:

A.

B.

C.

The leaking alarm will remain in since it is due to temperature and it takes some time for the tailpipe to cool down. The open alarms will clear as soon as flow stops.

Using the Mollier diagram for a constant enthalpy throttling process, temperature will be about 280 degrees F If the candidate uses the Mollier diagram and stops at the steam dome line (before becoming superheated) 340 will be chosen.

The leaking alarm will remain due to the heating of the tailpipe from SRV operation. Since there is no flow once the valve is closed, the open alarms will clear. Using the Mollier diagram for a constant enthalpy throttling process, temperature will be about 280 degrees F The leaking alarm will remain in since it is due to temperature and it takes some time for the tailpipe to cool down. The open alarms will clear as soon as flow stops. Using the Mollier diagram for a constant enthalpy throttling process, temperature will be about 280 degrees F D.

Sys#

System Category KA Statement LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 218000 Automatic Ability to predict andlor monitor changes in ADS valve tail pipe i-Depressurization System parameters associated with operating the tern peratures AUTOMATIC DEPRESSURIZATION SYSTEM controls including:

KIA##

ZI~OOO.AI.OI WA Importance 3.413.6 (RO/SRO)

Exam Level RO References provided to Candidate Steam TableslMollier Technical

References:

AR-110-E01, EO2, E03 diagram Question Source:

New Susquehanna. 8/1/2004 Level Of Difficulty: (1 -5) 2 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.5 145.5 Objective:

2092 Describe the operation of the following Task:

83.ON.

Implement Stuck Open components and controls of the Automatic 003 Safety-Relief Valve Depressurization System:

a. Timer Reset Switches

b. ADS Inhibit Switches

c. SRV Switches

d. Man lnit Switches

e. Local ADS Keyswitches

f. Acoustic Monitors

g. SRVs

h. Vacuum Breakers

i. T-Quenchers

j. Air Accumulators ( I O and 42 gallons)

k. Tailpipe Temp Recorders

1.

Lights and Alarms LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCFom.d0~

Printed on 07/28/04

SSES LOC 20 NRC Exam

--e

41.

Unit 2 was at 100% power, with Reactor Engineering running TIP traces when a loss of Feedwater occurred, causing an automatic start of RCIC. Water level did not drop to the auto start of HPCI.

The Reactor Engineer reports that a TIP trace was in progress, and when the reactor scrammed, the TIP stopped at mid-core.

What actions are required for the situation reported by the Reactor Engineer?

A.

Manually withdraw the TIP and close the Ball Valve.

B.

Manually withdraw the TIP. No other action required.

C.

Manually withdraw the TIP; ensure when the CHECK IN-SHIELD light ILLUMINATES the Shear Valve auto closes.

D.

Momentarily depress CONT ISOLATN PUSH TO RESET pushbutton, and verify TIP automatically withdraws.

Question Data A

Manually withdraw the TIP and close the Ball Valve.

L-Explanation/Justification:

A.

B.

C.

D.

Since RPV level dropped below 13" (RCIC auto initiates at -30, HPCl at -38),

the TIP should have withdrawn and isolated. The isolation failed so manual action should be taken to withdraw the TIP and isolate it.

Since RPV level dropped below 13" (RCIC auto initiates at -30, HPCI at -38),

the TIP should have withdrawn and isolated. The candidate may believe that this is all that is required (no isolation valve dosure) and choose this answer.

Shear valve must be manually fired and would only be fired if the TIP could not be isolated and indications were of leakage from the tube. The candidate may believe that the shear valve will auto close, This will reset the isolation, not cause the isolation to occur. The candidate may confuse the reset and initiation pushbuttons.

Sys#

System Category KA Statement 223002 Primary Containment Knowledge of the effect that a loss or malfunction of Traversing in-core probe Isolation SystemlNuclear the PClSlNSSSS will have on following:

system Steam Supply Shut-Off W M 223002.K3.21 WA Importance 2.6/2.7 Exam Level RO (ROE RO)

References provided to Candidate None Technical

References:

OP-178-001 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 1 45.4 Objective:

2320 Predict how TIP System key parameters and Task:

78.0P.O components will be affected by a failure of any of 02 the following support systems.

a. AC Distribution

b. Containment Instrument Gas (CIG) System

c. DC Distribution

d. Instrument Air System

e. Primary Containment Isolation System (PCIS)

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam W

42.

Which one of the following describes the purpose and function of the Primary Containment Isolation System (PCIS)?

PCIS uses:

A.

normally de-energized dual-bus logic (two independent divisions). Deliberate operator action is required to reset an isolation, and no single failure will prevent the system from performing its intended functions.

B.

normally de-energized dual-bus logic (two independent divisions). No operator action is required to reset an isolation, and no single failure will prevent the system from performing, or cause the system to perform its intended functions.

C.

normally energized dual-bus logic (two independent divisions). Deliberate operator action is required to reset an isolation, and no single failure will prevent the system from performing, or cause the system to perform its intended functions.

D.

normally energized dual-bus logic (two independent divisions). Deliberate operator action is required to reset an isolation, and no single failure will prevent the system from performing its intended functions.

Question Data D

normally energized dual-bus logic (two independent divisions). Deliberate operator action is reauired to reset an isolation and no single failure will prevent the system from performing its intended functions.

Explanation/Justification:

A.

B.

C.

D.

Correct answer Normally energized logic, isolations occur if power is lost. The candidate may believe that it is normally deenergized such as the ECCS logic and choose this answer..

Normally energized logic, isolations occur if power is lost The candidate may believe that it is normally deenergized such as the ECCS logic and choose this answer..

Single failure may cause system actuation. The candidate may confuse RPS with PCIS. A single failure will not cause a scram but can cause an isolation. This answer may be chosen.

Sys#

System Category KA Statement Isolation SystemlNuclear andlor function.

Steam Supply Shut-Off 223002 Primary Containment Conduct of Operations Knowledge of system purpose WA#

223002.2.1.27 WA Importance 2.812.9 Exam Level RO (RO/SRO)

References provided to Candidate None Technical

References:

TM-OP-059B Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (I

-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.7 Objective:

21 17 Describe the logic schemes used for automatic Task:

59.ON.

Implement Containment and manual initiation, isolation, and tripping of 006 Isolation the Primary Containment Isolation System.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

43.

With the unit operating at 100% power, a turbine trip with no bypasses occurs coincident with a loss of 1 D614 and 1 D624. All other systems work as designed.

Explain the Safety Relief Valve and Reactor Vessel pressure response for the Turbine trip transient.

A.

No relief function of SRVs will operate.

Safety Function of SRVs will operate.

Design pressure of 1,250 psig will be exceeded.

Safety Limit of 1,375 psig steam dome pressure will NOT be exceeded.

B.

Relief function of Division 1 SRVs will operate.

Safety Function of SRVs will operate.

Design pressure of 1,250 psig will NOT be exceeded.

Safety Limit of 1,375 psig steam dome pressure will NOT be exceeded.

C. No relief function of SRVs will operate.

Safety Function of SRVs will operate.

Design pressure of 1,250 psig will be exceeded.

Safety Limit of 1,325 psig steam dome pressure will NOT be exceeded.

D.

Relief function of Division 2 SRVs will operate.

Safety Function of SRVs will NOT need to function.

Design pressure of 1,250 psig will NOT be exceeded.

Safety Limit of 1,325 psig steam dome pressure will NOT be exceeded.

Question Data C

No relief function of SRVs will operate.

Safety Function of SRVs will operate.

Design pressure of 1250 psig will be exceeded.

Safety Limit of 1325 psig steam dome pressure will NOT be exceeded.

Explanation/Justification:

A.

B.

Safety limit is 1325. The limit of 1325 is based on 1375 at the bottom head with 5Opsi allowed for the height of water in the vessel.

The candidate The candidate may confuse the actual safety limit #with the basis # and choose this answer..

Relief function will not operate due to a loss of power and design pressure will be exceeded and safety limit is 1325. The limit of 1325 is based on 1375 at the bottom head with 50 psi allowed for the height of water in the vessel. The candidate may confuse the actual safety limit # with the basis #and may believe that SRVs may still have power to one division and choose this answer..

Loss of the DC buses removes power from the solenoids used to operate the SRVs in the relief function. Overpressure protection will thus depend on the safety function. In the safety function the design pressure will be exceeded but the safety limit of 1325 psig steam dome pressure will not be exceeded.

Relief function will not operate due to a loss of power and design pressure will be exceeded per the analysis. The candidate may believe the design limit will not be exceeded and may believe that SRVs may still have power to one division and choose this answer..

C.

D.

Sys#

System Category KA Statement 239002 Relief/Safety Valves KIA#

239002.~3.02 WA Importance 4.2l4.4 Exam Level RO Knowledge of the effect that a loss or malfunction of the RELlEFlSAFETY VALVES will have on following:

Reactor over pressurization (RO/SRO)

Technical

References:

FSAR 5.2 L-References provided to Candidate None LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 I 45.4 Objective:

2100 Determine if SRWADS and Plant response is Task:

83.0P.O Perform Manual Operation Of appropriate for any combination of 01 ADS

a. System mode of operation

b. Plant conditions

c. Key parameter indications LOC 20 As Given H:\\ExarnBan k\\MergeDocs\\LOCPONRCForrn.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

44.

The plant is operating at 100% power. A 150 psig Containment Instrument Gas (CIG) supply line shears just inside the containment, upstream of the check valve.

What impact will this have on the operation of the SRVs?

All automatic actions occur as designed. No operator action is taken.

A.

All SRVs will operate manually and in the Relief Mode.

Three ADS Valves are not impacted by the event, and three ADS Valves are limited to the gas stored in the individual ADS SRV Accumulators for ADS actuation.

B.

All SRVs will operate manually and in the Relief Mode.

All of the ADS Valves are limited to the gas stored in the individual ADS SRV Accumulators for ADS actuation.

C.

All SRVs in both the manual and Relief Modes are limited to the gas stored in the individual SRV Accumulators for actuation.

None of the ADS Valves are affected for ADS actuation.

D.

All SRVs will operate manually and in the Relief Mode.

Three ADS Valves are not impacted by the event, and three ADS Valves are limited to the gas stored in thirteen ADS Backup Bottles and the individual ADS SRV Accumulators for ADS actuation.

Question Data A

All SRVs will operate manually and in the Relief Mode.

Three ADS valves are not impacted by the event and three ADS valves are limited to the gas stored in the individual ADS SRV accumulators for ADS actuation.

ExplanationlJustification:

A.

6.

C.

D.

Each of the two 150# headers supplies 3 ADS valves for the ADS function. This break will affect 3 ads valves but will not affect the other valves in the line since the solenoid valves Manual and relief functions are supplied by the 90# header If the candidate believes that a single 150# header all ADS valve functions, this will be selected If the candidate believes that the 150# header supplies manual and relief operation, this will be selected The backup bottles are outside primary containment and thus removed from the valves by the break. The candidate may not know where the bottle tie into the valve supply and chose this answer.

Sys#

System Category KA Statement 239002 RelieflSafety Valves Knowledge of the effect that a loss or malfunction of the following will have on the RELIEF/SAFETY VALVES:

Air (Nitrogen) supply: Plant-Specific WA# 239002.~6.02 WA Importance 3.413.5 Exam Level RO (RO/SRO)

References provided to Candidate None Technical

References:

EO-100-1 13 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 145.7 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2103 Predict how SRVlADS key parameters and Task:

25.ON.

implement Loss Of

-L components will be affected by a failure of any of 003 Containment Instrument Gas the following support systems:

a. CIG

b. DC Distribution

c. RHR

d. Corespray

e. RPV instruments

f. Primary Containment instrumentation

g. AC Distribution

h. Suppression Pool LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 1.-

45. The Unit is operating at 100% power, when the Total Feedwater Flow Summer output signal fails low.

Explain how the plant will respond to the Total Feedwater Summer failure.

Feed pump speed initially:

A.

Rises The plant will stabilize with the reactor feed pumps controlling level at a higher than original level, and the main turbine in service.

B.

Lowers The plant will stabilize with the reactor feed pumps controlling level at a lower than original level, and the main turbine in service.

C.

Rises The reactor will scram on the main turbine trip, and the plant will stabilize with HPCl and RClC controlling level and the Bypass Valves controlling pressure.

D.

Lowers The reactor will scram on low RPV level, and the plant will stabilize with HPCl and RClC controlling level and the Bypass Valves controlling pressure.

Question Data C

Rises The reactor will scram on the main turbine trip and the plant will stabilize with HPCl and RClC controlling level and the Bypass Valves controlling pressure.

ExplanationlJustification:

A.

The FWC system sees steam greater than feed and increases pump speed the system cannot find a stable point before the feedpumps and main turbine trip on high level. If the candidate believes that the plant will stabilize with offsetting level and flow errors, this will be chosen.

Speed initially increases since the steam flow greater than feed flow indicates that level will begin to drop. FWLC anticipates that drop and increases RFP speed to increase flow and minimize the level fluctuation. The candidate may misanalyze the FWLC response and the direction of RPV level and choose this answer.

The FWC system sees steam greater than feed and increases pump speed the system cannot find a stable point before the feedpumps and main turbine trip on high level. Level will drop until HPCl and RClC initiate.

Speed initially increases since the steam flow greater than feed flow indicates that level will begin to drop. FWLC anticipates that drop and increases RFP speed to increase flow and minimize the level fluctuation. The candidate may misanalyze the FWLC response and the direction of RPV level and choose this answer.

B.

C.

D.

Sys#

System Category KA Statement 259002 Reactor Water Level Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM:

Reactor feedwater flow input Control System KIA#

259002.~6.04 KIA Importance 3.113.1 Exam Level RO (ROERO)

References provided to Candidate None Technical

References:

TM-OP-045 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 145.7

-L-.---

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L.

Objective:

1820 Predict how key parameters of the Reactor Task:

45.ON.

Implement Reactor Water Feedwater System, and the Plant will respond to failure of the following components or controls.

007 Level Anomaly

a. Master Controller

b. Individual RFP M/A Station

c. RPT Trip Pushbutton

d. Low Load Valve Controller

e. Hydraulic Jack

f. RFP Minimum Flow Controller

g. Motor Speed Changer

h. Level N B Selector Switch

i. S/U Bypass Valve Controller

j. Feedwater Controller

k. Steam Flow

1.

Feed Flow

m. RPV Level

n. One Elementrrhree Element Select Switch LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

46.

Unit 1 is operating at 25% power with a Drywell vent evolution in progress IAW OP-173-003, "PRIMARY CONTAINMENT NITROGEN MAKEUP AND VENTING."

Unit 2 is in Mode 4 when a High Drywell pressure of 2.0 psig occurs in the Unit 2 Drywell.

For the conditions provided:

A. The SGTS fan in service for the Unit 1 Drywell vent will:

B. How will the Unit 1 Drywell vent evolution be affected?

C. What Unit 1 operator action, if any, is required for the Drywell vent evolution?

A.

A. Continue to run.

B. The Unit 1 DW Vent line up will not change.

C. The operator must manually secure the Drywell venting lineup.

B.

A. Continue to run.

B. The Unit 1 DW Vent line up will isolate.

C. No further realignment of the Drywell venting lineup is required at this time.

C.

A. Trip.

B. The Unit 1 DW Vent line up will not change.

C. The operator must manually secure the Drywell venting lineup.

D.

A. Trip.

B. The Unit 1 DW Vent line up will isolate.

C. The operator must manually start the STGS fan that was in service to support Unit 2.

~

_ _ _ _ ~

Question Data A

A. Continue to run B. The Unit 1 DW Vent line up will not change.

C. The operator must manually secure the Drywell venting lineup.

ExplanatiodJustification:

A.

B.

C.

D.

Fan continues to run and receives an auto start signal. The vent lineup dampers would auto close on high drywell pressure for unit 1 but not for the other unit.

The vent lineup dampers would auto close on high drywell pressure for unit 1 but not for the other unit. The candidate may believe that the system completely realigns Fan continues to run and receives an auto start signal. The candidate may believe that the fans will trip to minimize the possibility of release and choose this answer.

Fan continues to run and receives an auto start signal. The vent lineup dampers would auto close on high drywell pressure for unit 1 but not for the other unit. The candidate may believe that the fans will trip and the dampers close to minimize the possibility of release and choose this answer.

Category KA Statement Sys#

System 261000 Standby Gas Treatment Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

High containment pressure System i

LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam WA##

261000.~2.11 WA Importance 3.213.3 Exam Level RO (ROISRO) i d

References provided to Candidate None Technical

References:

TM-OP-070 Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 I 45.6 Objective:

1951 Describe the operation of the following Task:

70.E0.0 Implement Secondary components and controls associated with the 03 Containment HVAC Standby Gas Treatment System.

IsolationlSGTS Initiation

a. Outside Air Flow Element

b. Outside Air Dampers

c. Reactor Building Differential Pressure SensordController

d. SGT System Filter Train

e. SBT System Exhaust Fans

f. Filter Train Crosstie Damper

g. Exhaust Flow Element

h. SGT System Fan Inlet Dampers

i. Reactor Building Recirculation Plenum Dampers

j. Drywell and Suppression Pool Burp and Purge Dampers

k. SBT System Exhaust Fan Switch

1.

Reactor Building Recirculation Plenum Flow Element LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 1-

47.

A LOCA is in progress on Unit 1 I and reactor water level has decreased to -1 29 inches.

Offsite power is available.

When will each of the RHR Pumps start?

A.

'A' RHR Pump starts immediately upon reaching -129 inches.

'8' RHR Pump starts immediately upon reaching -1 29 inches.

'C' RHR Pump starts seven seconds after reaching -1 29 inches.

'D' RHR Pump starts seven seconds after reaching -1 29 inches.

B.

All RHR Pumps start three seconds after reaching -129 inches.

C.

'A' RHR Pump starts immediately upon reaching -129 inches.

IB' RHR Pump starts immediately upon reaching -129 inches.

'C' RHR Pump starts three seconds after reaching -129 inches.

ID' RHR Pump starts three seconds after reaching -129 inches.

D.

All RHR Pumps start immediately upon reaching -129 inches.

Question Data A

'A' RHR pump starts immediately upon reaching -129"

'6 RHR pump starts immediately upon reaching -129"

'C' RHR pump starts 7 seconds after reaching -129"

'D' RHR pump starts 7 seconds after reaching -129" ExplanationlJustification:

A.

B.

C.

D.

correct answer, with offsite power available normally A and B start immediately, C and D seven second time delay This is the sequence is the sequence if the diesels supply power. The candidate may confuse the sequences This is a mix of the correct sequence with the no offsite power start times. The candidate may recognize that all the diesels should not start at once but use the wrong time delay.

This is the simultaneous start of the pumps when on the diesels with the A and 6 start times The candidate may believe that the RHR pumps all start simultaneously and use the A and B RHR Pump start times.

Sys#

System Category KA Statement 262001 A.C. Electrical Distribution Ability to monitor automatic operations of the AX.

Load sequencing WA#

262001.~3.04 WA Importance 3.413.6 Exam Level RO ELECTRICAL DISTRIBUTION including:

(RO/SRO)

References provided to Candidate None Technical

References:

TM-OP-049 Question SOU~Ce:

Exam Bank Susquehanna, 8i112002 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Memo@

10 CFR Part 55 Content:

41.7 145.7 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2146 Describe the operation of the following Task:

05.0P.O Perform ESS Bus 01 IA(B)(C)(D)(IAZOI) And Load Center 1B210(20)(30)(40)

Scheduled Outage L-components and controls of the 4.16 KV1480 VAC ESS Systems.

a. ESS Bus Supply Breaker

b. ESS Transformer Cooling C. UV Load Shedding

d. Protective Relays

e. ESS 480 VAC Load Center Transformer

f. Swing Bus Automatic Transfer Switch

g. Diesel Generator Voltage Restraint

h. ESS Bus Lockout Relay Protection

i. Lateral Pull Switches

j. Synchroscope

k. ESS Load Center Transformer Cooling

1.

RPS MG Set (480 V) Transfer Switch

m. Diesel Generator Automatic Transfer

n. Load Shedding Feeder Breaker Overcurrent Switches LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-v-

48.

1 E Instrument AC UPS Distribution 1 D130 supplies power to 120 VAC Distribution Panels 1Y128 and OY301.

IMMEDIATELY after the preferred source to 1 E Instrument AC UPS Distribution 1 D130 (1B246) is lost, what will be the configuration for ID130 to supply power to 1Y128 and OY301?

A.

Maintenance Backup stepped down to 120 VAC through the Maintenance Switch.

B.

Alternate source stepped down to 120 VAC through the Static Switch.

C.

Alternate source rectified and inverted to 120 VAC.

D.

The UPS dedicated battery is inverted to 120 VAC.

Question Data D

The UPS dedicated battery is inverted to IPOVAC ExplanationNustification:

A.

B.

C.

D.

Maintenance backup is used for maintenance and is a manual transfer. It is not the first supply to carry the inverter on a loss of the normal feed. The candidate may believe that the maintenance backup is the first alternative power supply and choose this answer.

Battery is first alternate power supply, until dead then 18226 is the power supply through the static switch. The candidate may believe that the static switch is the first alternative power supply and choose this answer.

The alternate source is stepped down and does not go through the rectifier and inverters the normal supply does The candidate may believe that the alternate supply is also rectified and choose this answer.

Correct answer, battery, UPS being supplied from the dedicated battery can carry the distribution panels for approximately 20 minutes. When the external batterys output decreases to less than 210 VDC, the Static Transfer switch operates to supply alternate power to the distribution panels, and alternate supply is via 480-20811 20 V step down transformers P -

Sys#

System Category KA Statement 262002 Uninterruptable Power Knowledge of UNINTERRUPTABLE POWER SUPPLY Transfer from preferred power Supply (AC.lD.C.)

(A.C.lD.C.) design feature@) andlor interlocks which provide for the following:

to alternate power supplies WA#

262002.~4.01 KIA Importance 3.113.4 Exam Level RO

( RO/S RO)

References provided to Candidate None Technical

References:

~ ~ - 0 ~ 4 1 7 Question Source:

Exam Bank Susquehanna, 81112003 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.7 Objective:

1467 Predict how the failure of the following support Task:

17.ON.

Implement Loss Of systems may impact the Instrument AC System.

003 Instrument Bus

a. Loss of preferred source to the Non-Class I E UPS Panel

b. Loss of the alternate source to the Non-Class 1E UPS Panel

c. Loss of the preferred source to the Vital AC or Computer UPS Panel

d. Loss of the alternate source to the Vital AC or Computer UPS Panel

e. Loss of the alternate source to SPDS UPS

f. Loss of the preferred source to the SPDS UPS

g. Loss of the inverter to the 1(2)Y115 and 1(2)Y125 Distribution Buses LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L'

49.

The AR-029-C03, "CS BAT RMS EXH FAN FAILED" alarm is received, and NO automatic actions occur.

Which one of the following describes the operator actions required, and the reason for the action?

A.

WITHIN three hours ENSURE the mode switch for 125 VDC AND 250 VDC Battery.

Chargers are in FLOAT position.

Establish firewatches.

OPEN the doors for Control Room cabinets.

Minimize Temperature rise for equipment operability.

B.

WITHIN three hours ENSURE the mode switch for 125 VDC AND 250 VDC Battery.

Chargers are in FLOAT position.

Open Battery Room doors.

Establish firewatches.

Minimize flammable hydrogen gas concentrations.

C.

OPEN the doors for Control Room cabinets.

Open breakers to DE-ENERGIZE A/C lighting.

DE-ENERGIZE non-essential equipment; such as ovens, printers, coffee pots, microwaves, spare PC terminals, etc.

Minimize Temperature rise for equipment operability.

D.

Open Battery Room doors.

Establish firewatches.

Open breakers to DE-ENERGIZE N C lighting.

Minimize flammable hydrogen gas concentrations.

Question Data B

WITHIN 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ENSURE the mode switch for 125 VDC AND 250 VDC Battery Chargers are in FLOAT position:

Open Battery Room doors Establish fire watches Minimize flammable hydrogen gas concentrations ExpIanationlJustification:

A.

Opening the door to the control room cabinets is part of the actions for loss of the floor cooling system, not loss of battery room ventilation. Both are covered in the same ON. The primary concern is hydrogen buildup in the room. The candidate may misuse the procedure and draw the wrong conclusion on the primary concern and choose this answer.

As directed by ON-030-001, the switches are placed in FLOAT to minimize hydrogen production. The doors are opened to help air flow and since the battery room doors are fire doors fire watches must be established.

Opening the door to the control room cabinets is part of the actions for loss of the floor cooling system, not loss of battery room ventilation. Both are covered in the same ON. The primary concern is hydrogen buildup in the room. he candidate may misuse the procedure and draw the wrong conclusion on the primary concem and choose this answer.

Deenergizing AC lighting is part of the actions for loss of the Control Room Floor Cooling System which is part of the same procedure but not required for the battery room fan failure. he candidate may misuse the procedure and choose this answer.

B.

C.

.- ~

D.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-Sys#

System Category KA Statement 263000 D.C. Electrical Distribution Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of ventilation during charging KIA#

263000.~2.02 KIA Importance 2.612.9 Exam Level RO (ROE RO)

References provided to Candidate AR-029-CO3.ON-030-Technical

References:

AR-029-C03,0N-030-001 001 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.5 I 45.6 Objective:

1430 Predict how key parameters of the 125 VDC Task:

02.ON.

Implement Loss Of 125v DC System and the Plant will respond to failure of the following components or controls.

a. Battery Chargers

c. Load Centers

d. Breakers 005 Bus 1 D610 (1 D620.1 D630, 1 D640)

b. Battery Banks LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\.

50. After maintenance, the 'A D/G is being restored per Section 2.16 DIESEL GENERATOR A(B)(C)(D)(E) REMOVAL FROM AND RESTORATION TO STANDBY AUTOMATIC OPERATION ALIGNMENT Of OP-024-001, "DIESEL GENERATORS."

Which indications will be observed in the Main Control Room as the 'A' D/G is placed in STANDBY AUTOMATIC mode?

A.

When the CONTROL MODE SELECT Switch on Panel OC521A is placed in REMOTE, simultaneously the Panel OC653 Diesel Generator 'A READY TO RUN white indicating light ILLUMINATES, and AR-OI5-Gl0, "DG A NOT IN AUTO" Alarm CLEARS.

B.

When the FUEL OIL TRANSFER PUMP OP-514A Switch is placed in AUTO, Panel OC653 Diesel Generator 'A READY TO RUN white indicating light ILLUMINATES, and When the CONTROL MODE SELECT Switch on Panel OC521A is placed in REMOTE, AR-015-GI 0, "DG A NOT IN AUTO" Alarm CLEARS.

C.

When the GOVERNOR MODE SELECTOR Switch is placed in ISOCH and the VOLT MODE SEL Switch is placed in AUTO, Panel OC653 Diesel Generator 'A READY TO RUN white indicating light ILLUMINATES, and When the CONTROL MODE SELECT Switch on Panel OC521A is placed in REMOTE, AR-Ol5-Gl0, "DG A NOT IN AUTO Alarm CLEARS.

When the GOVERNOR MODE SELECTOR Switch is placed in DROOP and the VOLT MODE SEL Switch is placed in AUTO, Panel OC653 Diesel Generator 'A' READY TO RUN white indicating light ILLUMINATES, and When the CONTROL MODE SELECT Switch on Panel OC521A is placed in REMOTE, AR-015-G10, "DG A NOT IN AUTO, Alarm CLEARS.

\\-

D.

Question Data A

When the CONTROL MODE SELECT switch on Panel OC521A is placed in REMOTE, simultaneously the Panel OC653 Diesel Generator 'A' READY TO RUN white indicating light ILLUMINATES, and AR-Ol5-GlO. "DG A NOT IN AUTO Alarm CLEARS.

ExplanatiodJustification:

A.

The diesel is in AUTO (NOT IN AUTO cleared) with the ready to run light illuminated, The mode selector in DROOP and the voltage regulator in auto. The Diesel is ready to auto start and will use isochronous speed control for an emergency start and droop for a manual start. The ready to run light illuminates and the alarm clears when the MODE SELECT SWITCH is placed in AUTO The Fuel Oil Transfer pump is the last switch noted in the Section 2.16 as being placed in auto. With all of the auxiliaries in auto the candidate may believe that the READY TO RUN light will illuminate The GOVENOR MODE SELECTOR and VOLT REG SEL switches are frequently discussed and important to the proper operation of the diesel. The candidate may believe that proper alignment of these switches will be indicated in the control room The GOVENOR MODE SELECTOR and VOLT REG SEL switches are frequently discussed and important to the proper operation of the diesel. The candidate may believe that proper alignment of these switches will be indicated in the control room

6.

C.

D.

Sys#

System Category KA Statement 264000 Emergency Generators Ability to manually operate andlor monitor in the Transfer of emergency control (DiesellJet) control room:

between manual and automatic LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam u

WA#

264000.~4.03 UIA Importance 3.213.4 Exam Level RO (ROE RO)

References provided to Candidate None Technical

References:

TM-OP-001,ON-024-001 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 145.5 to Objective:

2025 Describe the following logic schemes used for Task:

24.0P.O Perform Auto Operation Of Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 45.8 automatic and manual initiation and tripping of 05 Diesel Generator

- Permissives and logic for Emergency Start

- Permissives for Normal Start

- Emergency Trips

- DIG and Output Breaker response, during testing, to a LOCA or LOOP signal

- Bus Loading Sequence following a LOOP andlor LOCA Signal the Diesel Generators.

A(B)(C)(D)(E)

LOC 20 As Given H:\\xamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

51.

While operating at 75% power, a large leak develops at the point where the TBCCW Head Tank (IT1 15) piping joins the pump suction causing level to drop.

The Instrument Air Compressors are in a normal line up with the 'A' in LEAD and the 'B' in STANDBY.

Will the Instrument 'A' Air Compressor trip as a result of the TBCCW leak, and if so, what will be the first signal that will trip the compressor?

A.

The 'A' Instrument Air Compressor will not be impacted by the failure.

8.

The 'A' Instrument Air Compressor will trip. The first trip signal will be Low Cooling Water Pressure.

C.

The 'A' Instrument Air Compressor will trip. The first trip signal will be High Discharge Air Temperature.

D.

The 'A Instrument Air Compressor will trip. The first trip signal will be High Cooling Water Tem pe ra t ure.

\\-

Question Data B

The 'A' Instrument Air Compressor will trip. The first trip signal will be Low Cooling Water Pressure.

ExplanationlJustification:

A.

B.

C.

D.

The Instrument Air Compressor will lose cooling.

Low Water Pressure: 20 psig sensed by PSL-12504NB INST AIR COMP A CLG WTR IN will come in first as flow drops. The high temperatures will be generated after the cooling flow is lost.

High Discharge Air Temperature: 320 degrees F sensed by TSH-12503NB INST AIR COMPRESSOR A OUTLET. A temp control valve set at 118 degrees F will increase cooling water flow to maintain air temp. Will not be the first trip on a loss of cooling flow.

High Water Temperature: 150 degrees F sensed by TSH-12504AIB INST AIR COMP A CLG WTR OUT. A temp control valve set at 118 degrees F will increase cooling water flow to maintain air temp and the flow increase will minimize the cooling water temp increase. Will not be the first trip on a loss of cooling flow.

~

~~

Sys#

System Category KA Statement 300000 Instrument Air System Knowledge of the connections andlor cause effect Cooling water to compressor (IAS) relationships between INSTRUMENT AIR SYSTEM and the following:

KIA##

~OOOOO.KI.O~

WA Importance 2.812.9 Exam Level RO (RO/SRO)

References provided to Candidate None Technical

References:

TM-OP-OI 8

Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.2t041.91 Objective:

1769 List the signals and setpoints that cause the Task:

18.ON.

Implement Loss Of Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 45.7 to 45.8 Instrument Air System to automatically initiate, 003 Instrument Air isolate, and trip.

LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCFo~.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

52.

Unit 1 is at 100% power with the 'A' and 'B' Instrument Air Dryer Skids in service.

Turbine Building NPO reports a failure of the dryer transfer valves, and Instrument Air is being vented to atmosphere.

What impact will the failure of the 'A and '6 Skids have on the Instrument Air System, and what actions will be required?

Instrument Air System pressure will:

A.

remain constant while the operator verifies the IA TO SA CROSSTIE (PCV-12560) is automatically supplying air.

B.

lower until the operator manually places Dryer Skid 'C' in service.

C.

lower until the operator manually opens the PCV-12560 BYPASS (125143).

D.

remain constant while the operator verifies the Unit 1 - Unit 2 Crosstie is automatically supplying air.

Question Data B

Lower until the operator manually places Dryer Skid 'C' in service.

Explanation/Justification:

A. Air pressure will drop. The Service Air Crosstie is upstream of the dryers and is thus blocked by the dryer failure. If the candidate is unsure of the location of the crosstie and believes that it can supply air without going through the dryers then this answer may be chosen.

Correct answer. The operator must place the E& F dryers in service to maintain instrument air Air pressure will continue to drop. The Service Air Crosstie is upstream of the dryers and is thus blocked by the dryer failure. May believe that PCV-12560 is one way IA to SA and the bypass must be opened to allow SA to supply IA The unit IA cross tie must be manually opened. The candidate may believe that the IA supply from Unit 2 will automatically supply Unit 1 IA and choose this answer.

B.

C.

D.

Sys#

System Category KA Statement 300000 Instrument Air System Ability to (a) predict the impacts of the following on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Air dryer and filter

([AS) the INSTRUMENT AIR SYSTEM and (b) based on malfunctions WA#

3ooooo.~2.o1 WA Importance 2.9/2.8 Exam Level RO (ROB RO)

References provided to Candidate None Technical

References:

TM-OP-018, LA-I 14-803 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 I 45.6 Objective:

1769 List the signals and setpoints that cause the Task:

IRON.

Implement Loss Of Instrument Air System to automatically initiate, 003 Instrument Air isolate, and trip.

\\--

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

53. Unit 1 is at 100% power.

A RWCU Pump trip and system isolation was preceded by the following two alarms:

- AR-101-B01, "RWCU FILTER INLET HI TEMP

- AR-101-AOI, "RWCU FILTER INLET HI TEMP I S 0 Which of the following will cause these alarms and the RWCU System response?

A.

Low purge flow to the RWCU Pump causing the pump to heat up and resulting in a pump trip and system heating due to a loss of RWCU System flow B.

Low flow in Reactor Building Chilled Water causing isolation of RBCCW to RWCU non-regenerative heat exchanger C.

An electrical fault trip of the in service RBCCW Pump followed by an auto start of the standby RBCCW Pump on header low pressure D.

Insufficient RWCU blowdown flow to main condenser or Liquid Radwaste Question Data B

Low flow in Reactor Building Chilled Water causing isolation of RBCCW to RWCU non-regenerative heat exchanger.

Explanation/Justification:

A.

B.

C.

D.

The high temperature isolation came first and caused the pump trip and the pump will not overheat as long as RBCCW is available. If the candidate believes that RBCCW is required for pump operation then this answer may be chosen is correct. A low RBCW flow signal with a 13 second time delay will cause isolation of RBCCW to RWCU non-regenerative heat exchanger.

auto start of the standby pump should not lead to high temperature isolation of RWCU. If the candidate believes that this transient would cause a RWCU high temp and isolation this answer may be chosen excessive blowdown flow can lead to high temperature isolation, not insufficient flow. If the candidate confuses excessive flow with insufficient flow, this answer may be chosen.

Sys#

System Category KA Statement 400000 Component Cooling Water Ability to predict andlor monitor changes in CCW flow rate System (CCWS) parameters associated with operating the CCWS controls including:

WA#

400000.A1.01 WA Importance 2.812.8 Exam Level RO (ROISRO)

References provided to Candidate None Technical

References:

OP-161-OOI,ON-I~~-~OI Question Source:

Modified Susquehanna, 6/1/2003 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 I 45.5 Objective:

1358 Determine a course of action to mitigate or Task:

340NO Implement Loss Of Reactor correct an off-normal situation.

05 Building Chilled Water LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCPONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

54.

Which one of the following describes the power supply for the "A" Scram Solenoid Valves?

A.

B.

C.

D.

Division 2 RPS with power normally from I B227 and alternately from 1 B253 Division 2 RPS with power normally from 1 B227 and alternately from 1 B261 Division 1 RPS with power normally from 1 B217 and alternately from I B261 Division I RPS with ower normally from 1821 7 and alternately from 1 B253 Question Data D

Division I RPS with power normally from 16217 and alternately from 18253 ExpIanationlJustification:

k B.

C.

D.

Correct A Solenoids are powered by Division I RPS. If the operator is mistaken on the power supply for the A RPS solenoids, believing they are powered from Div II not Div I, this answer may be chosen.

A Solenoids are powered by Division I RPS. If the operator is mistaken on the power supply for the A RPS solenoids, believing they are powered from Div II not Div I, this answer may be chosen.

The alternate power supply for the bus is 16253. If the candidate confuses the alternate power supply for RPS A with that for RPS 6 this answer may be chosen.

KA Statement Category Sys#

System 201001 Control Rod Drive Knowledge of electrical power supplies to the Scram valve solenoids Hydraulic System following:

WA#

201001.~2.02 WA Importance 3.613.7 (RO/S RO)

Exam Level RO References provided to Candidate None Technical

References:

TM-OP-058 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Memory 10 CFR Part 55 Content:

41.7 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2413 Describe the operation and controls of the following components in the CRD Hydraulic System:

a. Suction Filter Strainers

c. CRDPumps Tie Line

e. Minimum Flow Line Filters

g. Flow Transmitter Flow

i. Charging Pressure Restricting Orifices

k. HCUS Valves

m. Flow Controller Valves

0. Drive Flow Transmitter Pressure

q. Directional Control Valves Water Pressure Valve

s. Cooling Water Flow Water Pressure

u. Scram Valves Valves

w. Scram Discharge Volume and Drain Valves

y. SDV Level Switches Accumulators aa. Lights and Alarms Rod Insertion Valves

b. Suction

d. Unit Cross-

f. Drive Water

h. System

j. Flow

1.

Flow Control

n. Stabilizing

p. Drive Flow

r. Drive

t. Cooling

v. Scram Pilot

x. SDVVent

z. Scram bb. Alternate Task:

55.ON.

Implement Loss Of CRD 014 System Flow LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

55.

The PCOP is preparing to place the CRD Flow Controller, FC-C12-I R600, in automatic in accordance with OP-I 55-001, "CONTROL ROD DRIVE HYDRAULIC SYSTEM."

The following parameters represent the stable CRD System conditions prior to placing the Flow Controller to automatic;

- Flow Control Station Total Water Flow FI-I ROI 9 - 63 gpm.

- DRIVE WATER DlFF PRESSURE PDI-C12-I R602 - 250 psid.

- COOLING WATER DlFF PRESSURE PDI-C12-I R603 - 20 psid.

- CRD Flow Controller, FC-C12-I R600 Meter indications are as given on the attached diagram.

If the PCOM takes CRD Flow Controller, FC-C12-1 R600, from MANUAL to AUTO, what will be the change (faster, slower or the same) in the CRD speeds for normal control rod motion, and why?

CRD speed for normal notching of a control rod will be:

A.

the same, since control rod speeds are set by adjusting needle valves in the flow from the below piston area of the CRD.

B.

Slower, due to lower drive header pressure when the flow control valve closes.

C.

the same, since the pressure control valve is set to maintain a constant pressure.

D.

Faster, due to higher drive header pressure when the flow control valve opens.

Question Data D

faster due to higher drive header pressure when the flow control valve opens ExplanatiodJustification:

A.

The needle valves adjust speed for a SET PRESSURE, not for varying system pressure conditions. The candidate may remember the purpose of the needle valves is to set speed but not recognize that this is not controlling for this situation, this answer may be chosen.

if the candidate does not understand that the deviation meter indicates that flow will increase then the candidate will believe that flow will decrease The pressure control valve has been positioned to establish the current pressures. The valve does not change position. There is less of a pressure drop across the flow control valve and thus higher pressure downstream and an increase in DP. This increases the DP across the control rods and will cause the rod to move faster. If the candidate believes the PCV will automatically change position to maintain pressure, this answer may be chosen.

The downscale indicator marks a mismatch between the setpoint and the actual value with the demanded value being greater than the current flow. When the controller is placed in AUTO, the flow control valve will open to increase flow. The pressure control valve has been positioned to establish the current pressures. The valve does not change position. There is less of a pressure drop across the flow control valve and thus higher pressure downstream and an increase in DP. This increases the DP across the control rods and will cause the rod to move faster B.

C.

D.

Sys#

System Category KA Statement 201003 Control Rod and Drive Ability to predict andlor monitor changes in Reactor power Mechanism parameters associated with operating the CONTROL ROD AND DRIVE MECHANISM controls including:

LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-w#

201003.A1.01 KIA Importance 3.713.8 (ROB RO)

References provided to Candidate Question Source:

New Susquehanna, 8/4/2004 Question Cognitive Level:

Analysis Objective:

2413 Describe the operation and controls of the CRD flow controller diagram following components in the CRD Hydraulic System:

Exam Level RO Technical

References:

OP-155-001 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

Task:

55.ON.

Implement Control Rod 41.5 145.5 01 5 Problems

a. Suction Filter Strainers

c. CRDPumps Tie Line

e. Minimum Flow Line Filters

g. Flow Transmitter Flow

i.

Charging Pressure Restricting Orifices

k. HCUs Valves

m. Flow Controller Valves

0. Drive Flow Transmitter Pressure

q. Directional Control Valves Water Pressure Valve

s. Cooling Water Flow Water Pressure

u. Scram Valves Valves

w. Scram Discharge Volume and Drain Valves

y. SDV Level Switches Accumulators aa. Lights and Alarms Rod Insertion Valves

b. Suction

d. Unit Cross-

f. Drive Water

h. System

j. Flow

1.

Flow Control

n. Stabilizing

p. Drive Flow

r. Drive

t. Cooling

v. Scram Pilot

x. SDVVent

z. Scram bb. Alternate LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-*I

56.

Unit 1 control rods are being withdrawn for power ascension. Core flow is 5570, and reactor power is 54% when the following alarms are received;

- AR-103-C04, "RBM UPSCALE OR INOP ROD BLOCK"

- AR-I 04-H03, "ROD OUT BLOCK To continue the startup, the operator must:

A.

increase recirculation flow.

B.

C.

bypass APRM "A."

D.

depress the PUSH TO SETUP pushbutton.

bypass APRM Flow Unit "A."

Question Data B

Depress the PUSH TO SETUP Pushbutton ExplanationlJustification:

A.

Once the setpoints are calculated they do not change as core flow changes but only recalculated when w null sequence is initiated.

As recirc flow is increased, power is also increased and the slope of the rod block monitor lines roughly parallels the rod pattern lines. As the setpoint increases, the power also increases at the same rate and thus the alarm does not clear and allow significant room for rod withdrawal. If the candidate does not recognize this, this answer may be chosen.

AR-104-H03 will come in when the RBM is giving a rod block due to reaching the setpoint. To continue, the procedure directs using push to setup Bypassing the reference APRM will cause the RBM to initialize and would recalculate the setpoints. A is not the reference APRM. If the candidate does not recognize this, this answer may be chosen.

Bypassing the flow unit does not initiate a new null sequence to recalculate the trip setpoints. If the candidate does not recognize this, this answer may be chosen.

-\\--

B.

C.

D.

Sys#

System Category KA Statement 215002 Rod Block Monitor Knowledge of the operational implications of the Trip reference selection: Plant-System following concepts as they apply to ROD BLOCK Specific MONITOR SYSTEM:

WA#

215002.K5.01 KIA Importance 2.612.8 Exam Level RO (RO/SRO)

References provided to Candidate AR-103-001, AR-104-Technical

References:

AR-103-001, AR-IO~-OOI,

~ t t 001, Att C NDAP-338 C NDAP-338 Question Source:

New Susquehanna. 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.5 I 45.3 Objective:

3019 Predict how key parameters of the Rod Block Task:

78.GO.

Operate the RBMs Monitor system and the plant will respond to manipulating the set up switches for the RBMs.

01 3 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-.-/ 57. A Reactor Water Cleanup System small suction line break has occurred inside the containment. The leak has been isolated by operator actions, and the following parameters are indicated several minutes after the line break:

- All rods fully inserted.

- Drywell pressure is 8 psig and slowly lowering.

- RClC is controlling RWL +I3 to +54 inches.

- Reactor pressure is 940 psig and slowly lowering.

- SPOTMOS temperature is 89 O F and slowly increasing.

- Suppression Chamber pressure is 12 psig and slowly increasing.

Which of the below statements explain@) the Containment and Drywell Ventilation Systems response for a Reactor Water Cleanup System suction line break in the Containment?

Containment pressures indicate:

A.

a failed SRV tailpipe in the Suppression Chamber.

B.

Primary Containment Vacuum Reliefs have failed to open.

C.

Suppression Chamber atmosphere composed entirely of nitrogen.

D.

a failed downcomer and failure of pressure suppression.

Question Data B

Primary Containment Vacuum Reliefs have failed to open.

ExplanationlJustification:

A.

A failed SRV tailpipe in the Suppression Chamber will cause pressure to rise in the Suppression Chamber until the chamber pressure is approximately 1 pound greater than Drywell pressure at which time the Suppression Chamber to Drywell vacuum breakers will relieve back to the drywell.

correct, The break in the RWCU line has occurred adding energy to the containment raising pressure in the Suppression Pool and the Drywell. The immediate analyzed response for a loss of coolant in the containment has Drywell pressure greater than Suppression Pool pressure due to the height of water in the Suppression Pool that the steam/vapor must displace to pass through the downcomers. When energy input to the Drywell is terminated, the steam/vapor mixture in the containment will condense and pressure will lower. Non-condensables that had been displaced into the Suppression Pool from the Drywell will be relieved back into the drywell through the vacuum breakers. For the indications provided the vacuum breakers are not opening therefore the atmosphere in the Suppression Chamber is not venting back into the Drywell. With RClC in service heat is still being added to the Suppression pool causing temperature and pressure to rise.

Nitrogen in the suppression chamber may confuse the candidates due to the extended write up in the EOP basis. From the EOP basis, the value of 13 psig is the lowest suppression chamber pressure which can occur when 95% of the non-condensables (N2) in the drywell have been transferred to the suppression chamber. This non-condensable concentration limit is established to preclude chugging-the cyclic condensation of steam at the downcomer openings of the drywell vents.

A failed downcomer in the Suppression Chamber will cause pressure to rise in the Suppression Chamber until the chamber pressure is approximately 1 pound greater than Drywell pressure at which time the Suppression Chamber to Drywell vacuum breakers will relieve back to the drywell.

B.

C.

D.

Sys#

System Category KA Statement 223001 Primary Containment Ability to monitor automatic operations of the ContainmenUdrywell response System and Auxiliaries PRIMARY CONTAINMENT SYSTEM AND during LOCA AUXILIARIES including:

L-.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOCZONRCFonn.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\.-

WA#

223001.~3.04 WA Importance 4.214.3 Exam Level RO (RO/SRO)

References provided to Candidate None Technical

References:

FSAR Sect 6.2 Question Source:

New Susquehanna, 8l112003 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 145.7 Objective:

266 Explain the sequence of events and flowpaths Task:

00.~0.0 that occur within the Primary Containment during a DBA LOCA. In your discussion include:

- Drywell

- Suppression Pool

- Downcomers

- Suppression Chamber

-Vacuum Breakers 27 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L,.

58. Unit 1 was operating at 100% power with all systems in normal alignment. A LOCA occurred inside containment; all systems responded as designed.

Current plant status is as follows:

- Drywell pressure

- RPV Pressure

- RPV level -

- HPCl and RClC controlling level and pressure 14 psig and slowly rising 600 psig and stable 30 inches and stable While attempting to initiate Containment spray, using the "A" Loop of RHR, the LOCA ISOLATION MANUAL OVERRIDE [HS-El 1-1 S17AI Keylock Switch breaks and cannot be moved to the OVERRIDE position.

What impact will this have on the ability to initiate Containment spray AND what actions will be required to address the high Drywell pressure condition.

A.

The Containment spray flowpath from BOTH the "A" Loop of RHR AND the "A" RHR Service Water Loop cannot be aligned.

Align "B" Loop of RHR for Containment spray.

B.

The Containment spray flowpath from the "A" Loop of RHR cannot be aligned.

Align "A" Loop of RHR service water for Containment spray.

C.

The Containment spray flowpath from BOTH the "A" Loop of RHR AND the "B" Loop of RHR cannot be aligned.

Align " A Loop of RHR service water for Containment spray.

D.

BOTH the "A" and" B loops of RHR AND BOTH the "A" and " B Loops of RHR service water cannot be aligned for Containment spray flowpath.

Vent the Drywell using a flowpath determined by the TSC.

Question Data A

The Containment spray flowpath from BOTH the "A" loop of RHR AND the "A" RHR service water loo^ cannot be alianed.

Align "B" loop of RHR for Containment spray.

ExplanationNustification:

A.

B.

C.

D.

The containment spray valves receive an isolation signal during a LOCA which must be bypassed using this switch.

RHR ' B is still available to spray the drywell A RHR Service Water relies on the same valves as A RHR for drywell spray. If the candidate does not recognize this, this answer may be chosen.

A RHR and A RHRSW use the same injection valves. If the candidate believes that the RHR systems share valve logic and that RHRSW does not share an injection flow path with RHR loop A This only impacts the A loop since RHR is divisionalize. If the candidate believes that one isolation valve in each division then this will be chosen Sys#

System L.

Category KA Statement LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 226001 RHRILPCI: Containment Spray System Mode Ability to (a) predict the impacts of the following on the RHRILPCI: CONTAINMENT SPRAY SYSTEM MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Valve logic failure WA#

226001.~2.13 KIA Importance 2.812.9 Exam Level RO (RO/SRO)

References provided to Candidate None Technical

References:

OP-149404 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 145.6 Objective:

2070 Discuss selected Industry Events, NRC concerns, Task:

49.0P.O Perform RHR Operation In and Plant operating history, concerning the RHR 05 Containment Suppression System.

Pool Spray Mode LOC 20 As Given H:\\ExarnBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam v

59. Unit 1 and Unit 2 are both operating at 100% power with all systems in normal alignment.

The feeder breaker to the ESS Transformer 101 inadvertently OPENS due to breaker failure.

All automatic actions occur as designed.

Assuming no operator actions, what will be the power supply to the 1A and 1 D RHR Pumps when the automatic actions have been completed?

The 1A RHR Pump will be powered from (1) and the 1D RHR will be powered from (2)

A.

(1 ) Startup Bus 20 (2) Startup Bus 20 B.

(I)

Startup Bus 20 (2) "D" Diesel Generator C.

(1) "A" Diesel Generator (2) "D" Diesel Generator D.

(1) "A" Diesel Generator (2) Startup Bus 20 Question Data A

(1) Startup Bus 20 (2) Startup Bus 20 ExplanationlJustification:

k Correct answer. The normal power supply to "A" RHR pump is Startup Bus 10 and the normal power supply to "D" RHR pump is Startup Bus 20. When the feeder breaker to the ESS Transformer 101 inadvertently OPENS, the power supply to the 1A201 bus will fast transfer to Startup Bus 20. Since the D RHR pumps is normally being powered from Startup Bus 20. both pumps will now be powered from Startup Bus 20.

Since Startup Bus 10 is the backup power to the D RHR pump, the candidates will choose this distractor if they believe this loss of backup power will result in the D Diesel generator starting and sequencing.

Candidates will choose this distractor if they believe both the fast bus transfers are onto the diesel generators instead of Startup Bus

20.

Candidates will choose this distractor if they believe bus 1A201 will transfer to the diesel generator and recognize that the Startup Bus 20 power to the D RHR is unaffected.

B.

C.

D.

Sys#

System Category KA Statement 230000 RHWLPCI:

Knowledge of electrical power supplies to the Pumps ToruslSuppression Pool following:

Spray Mode WA#

230000.K2.02 WA Importance 2.812.9 (RO/SRO)

Exam Level RO References provided to Candidate None Technical

References:

m-isooi Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 2 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Question Cognitive Level:

Memory 10 CFR Part 55 Content:

41.7 Objective:

195 Describe the operation of the following Task:

49.ON.

Implement Loss Of RHR 003 Shutdown Cooling Mode L

components and controls of the Residual Heat Removal System.

a. RHRPumps

b. Pump Suction and Shutdown Cooling Valves

c. Hx Bypass Valve Cooling Isolation Valves

e. Dry Well Spray Valves

f. SPTestand Spray Valves

g. Radwaste valves

h. Injection Valves

i. Min-Flow rates

j. F015AS/D ControllReset

k. Manual linitiation Push Button 1. Logic Power Test Switch

m. LOCA Override

n. MOV in Test Switch

0. Testable Check Valve Control Valve

q. RHR Heat Exchanger

d. Shutdown

p. RHRSW Flow LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

60. The plant is operating at full power with the EHC System 'A' Pressure Control Pressure Regulator in control.

Describe the plant response if the 'A' Pressure Control Pressure Regulator output signal FAILS LOW.

A.

The 'A' Pressure Control Pressure Regulator will open the Turbine Control and Bypass Valves.

Reactor pressure will lower.

Main Steam flow will go to zero.

B.

The 'B' Pressure Control Pressure Regulator will take control.

Reactor pressure will increase approximately 3 psi and remain at the new value.

Main Steam flow will initially lower and remain at the lower value.

C.

The 'B' Pressure Control Pressure Regulator will take control.

Reactor pressure will increase approximately 3 psi and remain at the new value.

Main Steam flow will initially lower then return to approximately the original flow value.

D.

The 'A' Pressure Control Pressure Regulator will open the Turbine Control Valves.

Reactor pressure will decrease.

Reactor will Scram on high main steamline flow.

Question Data C

The ' B Pressure Control Pressure Regulator will take control.

Reactor pressure will increase approximately 3 psi and remain at the new value.

Main Steam flow will initially lower then retum to approximately the original flow value.

ExplanationlJustification:

A.

6.

C.

This reflects a misunderstanding of how the regulator responds to a lower pressure.

Steam flow returns to the previous value since steam flow has to equal reactor power.

The pressure regulator signals go through a gate such that the one calling for the most opening is in control. When the pressure regulator fails low it attempts to close the turbine control valves to maintain pressure. As pressure increases the B pressure regulator which is set 3 psi higher senses the increase in pressure and takes over. Initially the A regulator caused the TCVs to close, lowering steam pressure, but when the B regulator takes over it will reestablish a stable pressure 3 psi higher but with the same steam flow.

This reflects a misunderstanding of how the regulator responds to a lower pressure.

D.

Sys#

System Category KA Statement 241000 Reactormurbine Pressure Ability to predict andlor monitor changes in Reactor pressure Regulating System parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including:

KIA#

z41ooo.~1.o1 WA Importance 3.913.8 Exam Level RO (ROE RO)

References provided to Candidate None Technical

References:

TM-OP-093 Question Source:

Modified Fermi 2,12/1111995 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 I 45.5 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

1643 Describe how the supported systems will be Task:

93.ON.

Implement Turbine EHC 005 Malfunction L-affected by any of the following EHC Pressure Control and Logic system failures:

a. Pressure Regulator Failure

b. Speed Control Failure

c. Load Control Failure

d. Bypass Control Failure LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

61.

Unit 1 is at 100% power. I&C has bypassed the 'A RPS Channel inputs from the 'A and 'C' ETS Pressure Switches (PSL-C72-1 N005A and C) to allow testing the switches as part of SI-183-314, "18-MONTH CALIBRATION, MAIN TURBINE CONTROL VALVE FAST CLOSURE."

While this test is in progress, the suction strainer of the in-service EHC Pump begins to clog, and EHC pressure begins to slowly drop.

IF the standby EHC Pump fails to automatically start and the EHC pressure trend continues, HOW will the plant respond to these conditions?

Assume NO operator actions are taken.

The turbine will trip on either low ETS pressure at:

A.

1,100 psig, or low FAS pressure at 800 psig.

The Reactor will scram on Control Valve Position after the turbine trip.

B.

800 psig, or low FAS pressure at 1,100 psig.

The Reactor will scram on Turbine Stop Valve Position after the turbine trip.

C.

800 psig, or low FAS pressure at 1,100 psig.

The Reactor will Not scram; a half scram signal will be received.

D.

1,100 psig, or low FAS pressure at 800 psig.

The Reactor will Not scram; NO half scram signals will be received.

~~

Question Data B

800 psig, or low FAS pressure at 1100 psig.

The Reactor will scram on Turbine Stop Valve Position after the turbine trip.

ExplanationlJustification:

A.

B.

C.

D.

If the candidate mixes the ETS and FAS setpoints and believes the Control valve position inputs to RPS, then this will be selected as the correct answer.

The turbine trip on low pressure will still occur on either low ETS or FAS pressure. The TSV position will cause a scram. Since the ETS pressure switches for the A and C TCVs are bypassed only a half scram (B RPS) will be generated by the TCV closure.

If candidate believes the bypassing of the A RPS channel has blocked the A RPS scram capability, then this will be selected as the correct answer.

If the candidate does not know that the Turbine Stop valve input to RPS is still active, and mixes the ETS and FAS setpoints, then this will be selected as the correct answer.

Sys#

System Category KA Statement 245000 Main Turbine Generator Knowledge of the effect that a loss or malfunction of the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS will have on following:

Reactor protection system and Auxiliary Systems LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam KIA#

245000.~3.07 WA Importance 3.613.7 (ROISRO)

L Exam Level RO References provided to Candidate None Technical

References:

TM-OP-093 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 145.4 Objective:

1604 Describe the operation of the following Main Task:

93.ON.

Implement Main Turbine Trip Turbine components and controls:

006

a. Turbine Components 1.) HP Turbine 2.) LP Turbine 3.) Turning Gear 4.) Moisture Separators 5.) Turbine Bearings 6.) Atmospheric Relief Diaphragm 1.) Low Speed Switch 2.) Manual Mechanical Trip 3.) Main (Shaft) Oil Pump 4.) Permanent Magnet Generator (PMG) 5.) Speed Sensor 6.) Lockout Valve 7.) Oil Trip Valve 8.) Master Trip Solenoid Valve 9.) Extraction Air Relay Dump Valve

b. Front Standard IO.) Mechanical Trip ValvelRemote Operating Reset Mechanism

c. Turbine Valves I.) Main Stop 2.) Turbine Control 3.) Bypass 4.) Combined Intermediate 5.) Extraction Non-Return 6.) Extraction Relay Dumps 7.) Exhaust Hood Spray

d. Instrumentation and Detectors 1.) Valve Position 2.) Bearing Temperature 3.) Vibration 4.) Thrust Bearing Wear 5.) Eccentricity 6.) Differential Expansion 7.) Shell Temperature 8.) Exhaust Hood Spray Pressure 9.) Exhaust Hood Temperature IO.) Shaft Ground 11.) Main Bearing Oil Pump Discharge Pressure 12.) Bearing Header Pressure LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\.-

62.

An event has occurred on the Unit I Refuel Floor with the following Radiation Monitor indications:

- Railroad Access Shaft 0.7 mrlhr

- Refuel Floor Wall Exh 18 mr/hr.

- Refuel Floor High Exh 25 mr/hr.

Which of the following automatic actions will occur as a result of the radiation levels indicated?

A.

Reactor Building HVAC Zones 1 and 3 ISOLATE.

Reactor Building HVAC recirculation INITIATES.

B.

Reactor Building HVAC Zone 3 only ISOLATES.

Reactor Building HVAC recirculation INITIATES.

C.

Reactor Building HVAC Zone 3 only ISOLATES.

Reactor Building HVAC recirculation does NOT initiate.

D.

Reactor Building HVAC Zones 2 and 3 ISOLATE.

Reactor Building HVAC recirculation INITIATES.

Question Data B

Reactor Building HVAC Zone Ill only ISOLATES.

Reactor Building HVAC recirculation INITIATES.

ExplanationlJustification:

A.

6.

C.

D.

Zone 1 does not isolate. Candidate may believe that another zone will isolate and choose this answer.

Refuel Floor Exhaust is high and will cause a Zone 111 isolation only. It also starts the Rx Bldg. recirc and SBGT and CREOASS Zone 111 isolates and Reactor building HVAC recirculates Candidate may believe that the ventilation system does not go into recirculation and choose this answer.

Only Zone 111 isolates. Candidate may believe that another zone will isolate and choose this answer.

Sys#

System Category KA Statement 272000 Radiation Monitoring Knowledge of the physical connections andlor cause-Reactor building ventilation MONITORING SYSTEM and the following:

System effect relationships between RADIATION system: PlantSpecific WA#

272000.~i.06 WA Importance 3.213.3 Exam Level RO (ROISRO)

References provided to Candidate None Technical

References:

TM-OP-034 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.2 to 41.9 I 45.7 to 45.8 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

1197 Describe the support function the Process Task:

79.0P.O Operate Process Radiation 03 Monitoring System L-Radiation Monitoring System operation provides to the following Plant Systems:

a. Reactor Protection System

b. PClS

c. Reactor Building Ventilation

d. Control Room Ventilation

e. Liquid Radwaste System

f. Offgas System

g. Standby Gas Treatment System

h. Condenser Air Removal System LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-v

63.

One of the four thermostats monitoring temperature for the C02 Fire Protection System in the Unit 1 Lower Relay Room has lost continuity.

How and why will this failure affect the automatic C02 initiation logic?

A.

If a second sensor in the Lower Relay Room detects a high temperature or loses continuity, the coincident sensor logic will be completed and the C02 initiation sequence will begin.

B.

As soon as the sensor loses continuity, the C02 initiation sequence for the Lower Relay Room will start, and C02 initiation will occur thirty seconds later.

C.

If a fire occurs in the portion of the room protected by the sensor without continuity, the Con System will not automatically initiate, but can be manually initiated.

D.

The fire protection circuit for the Lower Relay Room will not automatically function, and the COz System can only be manually initiated.

Question Data C

If a fire occurs in the portion of the room protected by the sensor without continuity, the C02 system will not automatically initiate, but can be manually initiated.

i d ExplanatiodJustification:

A.

B.

C.

D.

If the candidate confuses the C02 logic (single sensor to initiate) with the Halon logic (Coincident sensors required) then this answer will be selected If the candidate believes that the loss of continuity is the same as a fire signal then this answer will be chosen.

The Detector Circuit Trouble light on the Electrical Control Cabinet will be illuminated and the protection for the area is lost.

The other sensors in the room are still operable and can automatically initiate CO2. If the candidate believes that a single failure inops the entire system (detectors wired in series) this answer will be chosen.

Sys#

System Category KA Statement 286000 Fire Protection System Knowledge of the effect that a loss or malfunction of the FIRE PROTECTION SYSTEM will have on following:

The ability to detect fires WA#

286000.~3.01 KIA importance 3.213.4 (ROE RO)

Exam Level RO References provided to Candidate None Technical

References:

TM-OP-013 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 145.4 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

2291 Describe the operation of the following Task:

13.0P.O Operate Fire Protection components and controls of the Fire Protection 01 System System:

a. Diesel Fire Pump

b. Jockey Pump

c. Halon System

d. C02 System

e. SIMPLEX Panel

f. Electric Fire Pump

g. Preaction Sprinkler

h. Dry Pipe Sprinklers

i. Deluge System

j. Total Flooding C02

k. Manual Spurt C02

1. Total Flooding Halon

m. Manual CO2

n. Wet Pipe Sprinkler

0. Status Lights and Alarms

\\4 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\-

64.

The plant is operating with both Units at full power. The 'A' Reactor Building Zone I Exhaust Fan (1V205A) is in service, with the 'B' Reactor Building Zone I Exhaust Fan (1V205B) in stand by.

The Pressure Element (PE-I7501A), sensing the outside air pressure for the Reactor Building Zone 1 Exhaust Fan (1V205A and B), fails downscale.

How will the Reactor Building Zone 1 Exhaust Fans and Fan Discharge Dampers (PDD-17578NB) respond?

A.

The 'A' Exhaust Fan will continue to run, and the 'B' Exhaust Fan will automatically start.

The 'A' Fan Discharge Damper will be full open, and the 'B Fan Discharge Damper will open fully.

B.

The 'A' Exhaust Fan will continue to run, and the 'B' Exhaust Fan will remain in standby.

The 'A' Fan Discharge Damper will open fully, and the 'B' Fan Discharge Damper will remain closed.

C. The 'A' Exhaust Fan will trip, and the 'B Exhaust Fan will automatically start. The 'A' Fan Discharge Damper will close, and the 'B Fan Discharge Damper will modulate to control building pressure.

D.

The 'A' Exhaust Fan will trip, and the 'B' Exhaust Fan will not automatically start. The 'A' Fan Discharge Damper will close, and the 'B Fan Discharge Damper will remain closed.

~

~~

Question Data B

The 'A' Exhaust Fan will continue to run and the 'B' Exhaust Fan will remain in standby. The 'A' Fan Discharge Damper will open fully and the '6' Fan Discharge Damper will remain closed.

ExplanationlJustification:

A.

B.

C.

D.

If the candidate believes that the exhaust fans will operate at full capacity in an effort to create a negative pressure this answer will be chosen The PE failure will cause PDC17581 to see a high Reactor Building pressure. This will cause the discharge damper to open fully to try to maintain a negative sensed pressure. The standby fan will not start.

If the candidate believes that the A exhaust fan will be perceived as failed causing the 6 fan to start this answer will be chosen If the candidate believes that the fans will operate in a manner similar to the turbine building ventilation fans, this answer will be chosen Sys#

System Category KA Statement 288000 Plant Ventilation Systems Knowledge of the operational implications of the Differential pressure control following concepts as they apply to PLANT VENTILATION SYSTEMS:

WA#

288000.~5.02 WA Importance 3.213.4 Exam Level RO (RO/SRO)

References provided to Candidate None Technical

References:

TM-OP-033 Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis I O CFR Part 55 Content:

41.7 I 45.4 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC2ONRCForm.doc Printed on 07128104

SSES LOC 20 NRC Exam c

Objective:

2762 Predict how key parameters of the Turbine Task:

Building HVAC system and the plant will respond to manipulating the following controls for the system.

a. MG set vent fan switch

b. MG set vent fan auto switch LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

.L-

65. A Main Steam Line Break has occurred in the Reactor Building Steam Tunnel, and the A Steam Line Isolation Valves have failed to fully close. Security reports a steam plume rising above the Control Structure and a steam plume rising from the Unit 1 CST area.

Explain whether or not the Primary Containment is intact and why.

Explain whether or not the Secondary Containment is intact and why.

A.

Primary Containment is NOT intact because RB Main Steam Tunnel to atmosphere Blowout Panel lifted.

Secondary Containment is NOT intact because RB Main Steam Tunnel to Turbine Building Blowout Panel lifted.

B.

Primary Containment is intact; no breach in Primary Containment.

Secondary Containment is NOT intact because MSlVs not closed.

C.

Primary Containment is NOT intact because MSIVs not closed.

Secondary Containment is NOT intact because Main Steam Tunnel Blowout Panel is lifted.

D.

Primary Containment is intact; no breach in Primary Containment.

Secondary Containment is intact because Backdraft Isolation Dampers closed.

Question Data C

Primary Containment is NOT intact because MSlVs not closed.

Secondary Containment is NOT intact because Main Steam Tunnel Blow out panel is tiffed.

ExplanatiodJustification:

A.

8.

C.

RB steam tunnel to atmosphere panel is not part of primary containment. This choice has the proper determinations but for the wrong reason for primary containment.

Failure of the A MSlVs to close is a failure of primary containment isolation and thus of primary containment. MSlVs are part of primary containment not secondary containment.

The MSlVs are part of the primary containment. The plume above the control structure indicates the rupture is in the Rx Bldg. steam tunnel.

The Unit 1 CST area plume indicates that the main steam tunnel blowout panel has lifted and thus secondary containment is not intact The MSlVs are part of the primary containment. The plume above the control structure indicates the rupture is in the Rx Bldg. steam tunnel.

The Unit 1 CST area plume indicates that the main steam tunnel blowout panel has lifted and thus secondary containment is not intact D.

Sys#

System Category KA Statement 290001 Secondary Containment Knowledge of the effect that a loss or malfunction of the following will have on the SECONDARY CONTAINMENT Primary containment system WA#

290001.~6.04 KIA Importance 3.~4.1 Exam Level RO

( ROlS RO)

References provided to Candidate None Technical

References:

TM-OP-034, TS 83.6.1.3 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.7 I 45.7 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.d0~

Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

1994 Describe how the Standby Gas Treatment System Task:

70.AR.O Implement High Radiation In key parameters and components respond to 03 Reactor Building automatic and manual initiation, isolation, or tripping within the System.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

66.

A Loss of Coolant Accident has occurred on Unit I with Reactor Pressure Vessel water level being restored by Condensate. Water level is transitioning from the Fuel Zone level to Wide Range level.

The following plant conditions exist:

- All ADS Valves open.

- All Core Spray and RHR Pumps are running.

- All Diesel Generators are running unloaded.

- Main Generator Lockouts reset.

-Wide Range Level

- Reactor Vessel pressure

-140 inches and slowly rising.

450 psig and lowering.

Given direction to close RHR INJ FLOW CTL HV-151-FO17A and B from the Unit Supervisor, what prerequisite actions (if any) need to be taken before closing the RHR Injection Valves?

A.

Verify Reactor Water Level on Fuel Zone is upscale.

Verify Reactor Water Level on Wide Range is on-scale.

Close Valves when water level >-I 29 inches.

6.

No prerequisite actions required.

Close Valves now.

C. Verify RPV pressure not decreasing.

Verify RPV pressure 81 psid > Suppression Chamber pressure.

Close Valves.

D.

Verify injection with both A and B Core Spray Loops.

Close Valves.

Question Data A

Verify Reactor Water Level on Fuel Zone is upscale Verify Reactor Water Level on Wide Range is on-scale.

Close Valves when water level s-129 inches ExplanationNustification:

A. Correct answer, Per OP-AD-001, section 7.1 Adequate core cooling assured by at least two (2) independent indications. Adequate core cooling defined as Core Submergence or Steam Cooling With Injection Of Makeup Water To The RPV or Steam Cooling Without Injection (Blowdown Cooling) of Makeup Water to the RPV with Indicated RPV Water Level at or above the Minimum Zero Injection RPV Water Level (-205 inches):

No check of two independent level indications as required by procedure so adequate core cooling has not been verified and the valves cannot be closed.

No check of ADS/SRVs open which would satisfy adequate core cooling by Steam Cooling With Injection Of Makeup Water To The RPV method and since adequate core cooling is not verified the injection valves cannot be closed.

Injection with core spray alone is no longer sufficient for adequate core cooling since adequate core cooling is not verified the injection valves cannot be closed.

B.

C.

D.

Sys#

System None Generic Category Conduct of Operations KA Statement Knowledge of conduct of operations requirements.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

wA#

2.1.1 SSES LOC 20 NRC Exam WA Importance 3.713.8 (RO/SRO)

Exam Level RO References provided to Candidate None Technical

References:

OP-AD401 Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41. I O I 45.1 3 Objective:

4481 STATE, from memory, the guidelines for Task:

OO.AD.1 Implement Appropriate overriding Emergency Core Cooling System 31 Portions Of Operations (ECCS) Controls. This ability will be demonstrated in simulator course.

Standards For System and Equipment Operation LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForrn.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

67.

What is the function and the purpose of the seven-minute timer in the ADS logic?

The seven-minute timer functions in place of:

A.

high drywell pressure, in the event of a LOCA outside the containment.

B.

the ECCS confirmatory low water level permissive to back up the level signal in the event of an instrument failure C.

high drywell pressure, to back up high drywell pressure signal in the event of LOCA inside containment.

D.

the -129 inches in the A I and B1 Logic Trains to back up the low level signal for a break that depressurizes the reactor with out lowering level to Level 1.

Question Data A

high drywell pressure, in the event of a LOCA outside the containment.

ExplanatiodJustification:

A.

6.

C.

D.

Break outside containment that does not depressurize the reactor will require HPCl but will not increase drywell pressure Other ADS system logic is the backup for this failure not the 7 minute timer Break inside containment of sufficient size to require HPCl will pressurize the Drywell and ADS will initiate prior to the 7 minute No need to ADS if core is adequately cooled and the core is adequately cooled if level is maintained above -161..

timer..

x Sys#

System None Generic wA#

2.1.28 Category Conduct of Operations KIA Importance 3.2f3.3 (RO/SRO)

Exam Level KA Statement Knowledge of the purpose and function of major system components and controls.

RO References provided to Candidate none Technical

References:

TM-0~483 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Memory 10 CFR Part 55 Content:

41.7 Objective:

2102 Predict how key parameters of the SRV/ADS and Task:

83.0P.O Perform Manual Operation Of the Plant will respond to failure of the following 01 ADS components or controls:

a.

SRVs

b.

ADSTimers

c.

Tailpipe Temperature

d.

Acoustic Monitors

e.

Vacuum Breakers

f.

Initiation Logic LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

68.

The reactor is operating a 100% power, with all systems in normal alignment. AR-103-H6, "CRD ACCUMULATOR TROUBLE" alarm is received. The Reactor Building NPO reports that HCU 32-27 Accumulator has a high water level and the water cannot be drained.

What actions are required for this situation?

A.

Check Control Rod associated with accumulator alarm is at designated position, and hydraulically disarm HCU.

B.

Verify the control rod associated with the inoperable accumulator is fully inserted.

AND Declare the associated control rod inoperable within one hour.

C.

Declare the associated control rod scram time "slow" within eight hours.

OR Declare the associated control rod inoperable within eight hours.

D.

If accumulator alarm is a fully withdrawn control rod, electrically disarm HCU and EXERCISE scram accumulator piston in accordance with OP-155-001.

Question Data C

Declare the associated control rod scram time "slow" within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

OR Declare the associated control rod inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

ExplanationlJustification:

A.

6.

C.

D.

Control rod is required to be fully inserted prior to being disarmed.

This is the action if Rx pressure is less than 900 psig. The candidate should recognize that Rx pressure is greater than 900 psig.

The Accumulator cannot be drained, therefore it cannot be kept free of water and must be declared inop. The associated control rod must be declared slow or inoperable.

HCU is hydraulically isolated to exercise the accumulator piston, and control rod is required to be fully inserted prior to being disarmed.

Sys#

System Category KA Statement None Generic Conduct of Operations Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

WA#

2.1.33 WA Importance 3.414.0 Exam Level RO (ROISRO)

References provided to Candidate TRM & TS Technical

References:

TRM 3.1.4, TS 3.1.5 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

43.2 143.3 I 45.3 LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\.-

Objective:

3355 Locate the procedures to use when the following Task:

55.ON.

Implement Control Rod CRD problems occur during a plant startup:

01 5 Problems

a. The CRD pump trips

b. An individual control rod scrams

c. An accumulator fault occurs

d. A rod position detector string fails

e. A control fails its coupling check LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

69.

The plant is at 100% power. During the performance of SO-134-001, "QUARTERLY ZONE I ISOLATION DAMPER TIMING", the closing times are recorded as follows:

- First Damper tested (HD-17576A) is recorded at seven seconds.

- Second Damper tested (HD-17576B) is recorded at 11 seconds.

- Remaining Dampers (HD-17524AI HD-I7524B, HD-I7586A, HD-17586B) have NOT been tested.

With the acceptable closure stroke times for all dampers established at 10 seconds, which one of the following describes the required action per NDAP-QA-0722, "SURVEILLANCE TESTING PROGRAM?"

A.

Inform the Unit Supervisor that the second Damper (HD-17576B) is inoperable, and place the test on hold.

B.

Document that the second Damper (HD-I7576B) is inoperable, and continue the test.

C. Restroke the second Damper (HD-I7576B), and if the time is acceptable, then continue the test.

D.

Restroke the second Damper (HD-l7576B), and if the time is above 10 seconds, then place the test on hold.

Question Data A

Inform the Unit Supervisor that the second damper (HD-17576B) is inoperable and place the test on hold.

ExplanatiodJustification:

A.

B.

C.

D.

Per SO-134-001 and NDAP-QA-0722, if the stroke time is above the acceptance criteria immediately notify shift management. The test will be placed on hold until an investigation can be performed for the affected damper.

The test must be stopped and placed on hold until an investigation can be performed for the affected damper.

Re-stroke of the valve is not permitted for these isolation dampers. This would not be a valid test of the damper's operability and would be preconditioning.

Re-stroke of the damper is not permitted for these isolation dampers. This would be preconditioning of the damper.

Sys#

System None Generic Category Equipment Control KA Statement Knowledge of surveillance procedures.

WA Importance 3.013.4 (RO/SRO)

Exam Level RO References provided to Candidate None Technical

References:

so-I 34-001, NDAP-QA-O~~~

Question Source:

Exam Bank Susquehanna, 8/1/2001 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.10 145.13 Objective:

1284 Determine if the Secondary Containment System Task:

OO.TS.0 Ensure Plant Operates In or a component is required to be operable per Technical Specificatoins.

Operating License, Technical 01 Accordance With The Specifications (TS), and Technical Requirements Manual (TRM)

LOG 20 As Given H:\\ExamBank\\MergeDoc\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam c

70.

Maintenance has requested that Instrument Air be lined up for use in the 'A' DIG Room, but the isolation valve for the Instrument air drop in the room has a PINK Status Control Tag with CLOSED for the valve position and a comments section statement that says; "The D/G System Engineer found the valve packing leaking when the valve is OPEN."

In accordance with NDAP-QA-0302, "SYSTEM STATUS AND EQUIPMENT CONTROL", can the valve be opened to support Maintenance? Why or Why NOT?

A.

No, the valve cannot be operated. The valve must remain in the closed position until the Status Control Tag is removed.

B.

Yes, the valve may be opened IF Operations Supervision gives permission to do so.

C.

Yes, the valve may be opened IF the D/G System Engineer gives permission to do so.

D.

No, the valve cannot be operated. Operating components with Status Control Tags applied is ONLY allowed for personnel safety concerns.

Question Data 6

Yes the valve may be opened IF Operations Supervision gives permission to do so.

ExplanationlJustification:

A.

6.

C.

D.

May be operated with Operations concurrence.

correct answer Per NDAP-QA-0302 section 6.5.6 Operations Supervision must concur, NOT the individual that applied the tag.

Operations supervision may give permission to operate the valve for any reason not solely for personnel safety.

i Sys#

System Category KA Statement None Generic Equipment Control Knowledge of tagging and WA#

2.2.13 WA Importance 3.613.8 Exam Level RO clearance procedures.

(RO/S RO)

References provided to Candidate None Technical

References:

NDAP-QA-0302 Question Source:

Modified Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

41.10145.13 Objective:

936 Utilizing Procedure NDAP-QA-0302, System Task:

OO.AD.0 Implement Appropriate And Equipment Control Status and Equipment Control, discuss 38 Portions Of System Status administrative controls and requirements placed upon equipment component manipulations, where the component is not within a Permit boundary, and there are no personnel protection requirements.

\\-

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-.-- 71.

Reactor power is 80% and being returned to 100% power following special testing and a control rod sequence exchange. The following alarms are received in the Control Room.

- AR-OI5-DO4, "STACK MONITORING SYS OC63O/OC677 HI-HI RADIATION"

- AR-OI5-EO4, "STACK MONITORING SYS OC63O/OC677 HI RADIATION" Further investigation reveals that Turbine Building Exhaust Radiation (Point #5) is the cause of the alarm, and Offgas flow is 75% of the value it was before the power increase began.

What actions are required for this situation?

A.

B.

Shutdown Radwaste Ventilation.

C.

D.

Isolate the Primary Coolant Degasifier.

Start the Common Offgas Recombiner and shut down the Unit 1 Offgas Recombiner.

Isolate the open pair of OFFGAS DELAY LINE DRAIN VLVS.

Question Data D

Isolate the open pair of OFFGAS DELAY LINE DRAIN VLVS ExplanationlJustification:

A.

B.

C.

D.

incorrect - this action would be appropriate ONLY after chemistry sampled the degasifiers and determined that they are the source of the high radiation.

incorrect - appropriate if Radwaste is believed to be the source, but this is not consistent with the indications given the drop in offgas flow.

incorrect - with the drop in offgas flow, a candidate may believe the source of the problem to be with the recombiners, in which case shutting down the ineffective recombiner and starting the common recombiner would be appropriate.

Per step 3.2 this is the probable cause given the drop in offgas flow Sys#

System None Generic WA#

2.3.11 Category Radiological Controls WA Importance 2.713.2 (RO/SRO)

Exam Level KA Statement Ability to control radiation releases.

RO References provided to Candidate o~mo-001 Technical

References:

0~470-001 Question Source:

Modified Susquehanna 1,81112002 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

45.9 145.10 Objective:

2682 For the STA Only:

Task:

OO.EO.0 Implement Radioactivity 29 Release Control Utilize Control Room prints to assist the Shift Supervisor.

LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

1.

72. A valve line-up needs to be performed for Reactor Building Sample Station Manual Isolation Valves from Recirc. 143F059, RX WTR SMPL LINE TO HV143F019 is located in a high radiation area where the dose rate is approximately 250 mr/hr general area. It is estimated that the line-up will take about I O minutes.

For this situation, what two verification options are available, and why?

A.

B.

C.

D.

Valves are normally open, do not fulfill any requirement of a Safety Related System; therefore are EXCLUDED from Verification requirements.

OR Concurrent Verification, which minimizes dose, and meets requirements.

OMIT the verification requirement due to whole body exposure greater than 10 mrem to conduct the verification.

OR System Test Verification; no additional dose will be accumulated.

Concurrent Verification minimizes dose, and meets requirements.

OR System Test Verification; no additional dose will be accumulated.

Restoration Self-verification, since valves are not safety related.

OR OMIT the verification requirement due to whole body exposure greater than 10 mrem to conduct the verification.

Question Data B

OMIT the verification requirement due to whole body exposure greater than 10 mrem to conduct the verification.

OR System Test Verification, no additional dose will be accumulated.

ExplanationlJustification:

A.

B.

There is no justification for omitting verification due to normal position and concurrent verification does not significantly reduce dose.

correct, per OP-AD-002. The candidate needs to verify using the attached Check Off List (COL) that a second verification is required. Per STANDARDS FOR SHIFT OPERATIONS, OP-AD-002 a dose accumulation of 10 mrem or greater is sufficient justification to NOT perform a second verification. Per the OP-AD-002 a System Test Verification may be performed in lieu of second verification.

Concurrent verification does not significantly reduce dose.

Restoration self verification is not permitted.

C.

D.

Sys#

System Category KA Statement None Generic Radiological Controls Knowledge of 10 CFR 20 and related facility radiation control requirements.

wA#

2.3.1 WA Importance 2.6/3.0 Exam Level RO

( RO/S RO)

References provided to Candidate c~-16440110 Technical

References:

OPAD402 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3

-1.-

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.12143.41 45.9 145.10 Objective:

491 Discuss the requirements for verification of Task:

OO.AD.l Implement Appropriate operating activities.

31 Portions Of Operations Standards For System and Equipment Operation LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

'i 73. Within the power leg of EO-100-1 13, "LEVEL/POWER CONTROL" there is a shifting of priorities concerning starting Standby Liquid Control (SLC), depending on reactor power.

What is the basis for starting SLC immediately if power is >5%?

A.

B.

C.

D.

Enables Main Turbine to be tripped earlier.

Prevent positive reactivity addition due to boron dilution and temperature reduction.

Complete SLC injection before RWCU is needed for pressure control.

Preclude power oscillations, and ensure the plant remains in a controlled state.

Question Data D

Preclude power oscillations and ensure the plant remains in a controlled state.

ExplanationNustification:

A.

B.

Tripping of the Main Turbine is not included in the ATWS mitigation strategy but could be confused with tripping of the Reactor Recirc pumps for power reduction.

Positive reactivity due to boron dilution and temperature reduction is referring to possible initiation of ADS causing dilution of boron and introduction of cold water. The distracter wording which refers to ADS is part of the EOP basis discussion covering the rapid insertion of Boron and could be confused.

RWCU is not considered for pressure control if Boron is injected.

correct answer, EOP Technical Bases states: "If initial ATWS power was greater than 5%, then a relatively large number of control rods have failed to insert. The seriousness of this condition requires immediate injection of boron to positively tetminate the ATWS event."

Early boron injection has the following benefits:

Stop or prevent large-magnitude Limit Cycle Oscillations which can lead to core damage.

Limit fuel damage from uneven flux patterns that could result from partial rod inserts.

Protect the primary containment from excessive heat input.

C.

D.

L-'

Sys#

System Category KA Statement None Generic Emergency Procedures and Plan Knowledge of the bases for prioritizing safety functions during abnormallernergency operations.

WA#

2.4.22 WA Importance 3.014.0 Exam Level RO (RO/SRO)

References provided to Candidate None Technical

References:

Eo-100-1 13 Question Source:

Modified Clinton 1,7/23/2001 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

43.5 145.12 Objective:

2682 For the STA Only:

Task:

OO.EO.0 Implement Level/Power 31 Control Utilize Control Room prints to assist the Shift Supervisor.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCPON RCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

74.

Unit is at 100% power, when a report of smoke inside the 'C' RFP Turbine Room is received, and the Fire Brigade is activated.

What is the minimum required number and position names of the on-shift Operations personnel that will respond to the Fire Brigade activation?

What gear is the Fire Brigade REQUIRED to bring to the command post?

A.

One Fire Brigade Leader who need not be fully dressed out and Four Non-Licensed Operators (NLOs) in Full Dress-out Gear, with radios, flashlight, and carrying SCBA.

B.

One Fire Brigade Leader who need not be fully dressed out and Two Non-Licensed Operators (NLOs) in Full Dress-out Gear, with radios, flashlight, and other equipment as directed by Fire Brigade Leader.

C. One Fire Brigade leader in Full Dress-out Gear and Three Non-Licensed Operators NLOs dressed as directed by Fire Brigade with radios, flashlight, and other equipment as directed by Fire Brigade Leader.

D.

One Fire Brigade leader in Full Dress-out Gear and Two Non-Licensed Operators (NLOs) dressed as directed by Fire Brigade Leader; equipment as directed by Control Room.

Question Data B

1 - Fire Brigade leader who need not be fully dressed out and 2 - Non-Licensed Operators (NLOs) in Full Dress out Gear, with radios, flashlight, and other equipment as directed by Fire Brigade Leader.

Explanation/Justification:

A.

B.

C.

D.

Only 2 NPOs are part of the fire brigade and SCBA is not always necessary. In the past there were more NPOs assigned to fire brigade.

The Ops fire brigade leader does not have to be fully dressed out but sets up the command post. The NPOs are fully dressed out and bring additional equipment as directed.

The Ops fire brigade leader does not have to be fully dressed out but sets up the command post.

The Ops fire brigade leader does not have to be fully dressed out but sets up the command post and the fire brigade members are fully dressed out with additional equipment as necessary..

Sys#

System Category

~

KA Statement None Generic Emergency Procedures and Plan Knowledge of facility protection requirements including fire brigade and portable fire fighting equipment usage.

W M 2.4.26 WA Importance 2.913.3 Exam Level RO (ROB RO)

References provided to Candidate None Technical

References:

NDAP-QA-0445 Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

43.5 145.12 LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\-- Objective:

5085 Identify the impact of the Fire Protection System's Task:

13.PM.

Implement Inventory Of operability on Technical Requirements Manual.

002 Turbine Deck Fire Brigade Shed LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-

75. Unit 1 was at 100% power when grid instabilities result in activation of EHC power load unbalance protection circuitry. After the reactor scram, the Unit is stabilized with the following conditions:

- Reactor water level + 5 inches.

- Reactor pressure -955 psig controlled with Bypass Valves.

- RPV bottom head drain temperature is 380 O F.

Which of the following is correct water level band for the above conditions?

A.

Maintain level -30 to +5 inches.

B.

Restore level +I3 to +30 inches.

C. Restore and maintain level +45 to +54 inches.

D.

Restore level +I3 to +54 inches.

Question Data B

Restore level +I3 to +30 inches, ExplanationlJustification:

A.

B.

no bases for level band, -30 inches is initiation setpoint for RClC correct, candidate needs to recognize in stem that a turbine trip occurred, causing a trip of the recirc pumps resulting in no forced core circulation. The stated level is the correct level band with stratification and no recirc pump in service. Must determine the delta T between bottom head temp and saturation temp using steam tables, recognize delta T is greater than 145 deg F and from memory know the proper level band for control of +I3 to +30 inches.

level band when delta T is equal to or less than 145 deg F level band requires at least one recirc pump in service also is level band directed by EO-100-102 C.

D.

Sys#

System Category KA Statement None Generic Emergency Procedures and Plan 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.

wA## 2.4.48 KIA Importance 3.~3.8 Exam Level RO (RO/S RO)

References provided to Candidate Steam Tables, ON-loo-Technical

References:

ON-100-1 01 101 Question Source:

Modified Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

43.5 145.12 Objective:

1364 Explain the reasons for steps contained in an off-Task:

OOONO Implement Scram normal procedure.

18 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam u

76.

Unit 1 is operating at 100% power when the Main Turbine trips due to false High Reactor Vessel Level signals.

After the Turbine trip, the Control Room Operators report the following conditions and alarms:

-Auxiliary Buses 11A (1A101) and I I B (1A102) transferred to Startup Bus 10 (OA103).

- The Main Generator Sync breaker (1 RIOI) - OPEN.

- The 230 kV Switchyard Breakers, 3W (Generator 1 West) and 3T (Generator 1 East) -

- Main Generator Exciter Field Breaker - OPEN.

OPEN.

- AR-106-A08, GEN LOCKOUT RELAYS TRIP.

- AR-106-E08, GEN ANTI MOTORING TRIP.

What actions must be directed as a result of the above information?

A.

Enter ON-198-004, UNIT 1 MAIN GENERATOR UNABLE TO DISCONNECT FROM GRID AFTER A TURBINE TRIP.

CONTACT the Scranton System Operator to block open 230 kV Breakers, 3W and 3T.

Verify AR-106-E08, GEN ANTI MOTORING TRIP cleared after 30 seconds.

-v B.

Enter ON-I 00-1 01, SCRAM and ON-003-001, LOSS OF STARTUP BUS I O.

CONTACT Transmission Control Center (TCC) to investigate the cause of the 3W and 3T 230 kV Breaker trip and reclose.

Re-energize Auxiliary Buses 11A and 1 I B.

C.

Enter EO-I 00-1 02, RPV LEVEL CONTROL and ON-I 04-201, LOSS OF 4 kV ESS BUSES I A and IC.

CONTACT Transmission Control Center (TCC) to re-energize Auxiliary Buses I 1A and IIB.

Verify A & C D/Gs running with cooling water.

D.

Enter ON-I 00-1 01, SCRAM and ON-I 93-002, MAIN TURBINE TRIP.

CONTACT the Scranton System Operator to investigate the cause of the 3W and 3T 230 kV Breaker trip and reclose.

Restart Reactor Recirculation Pumps.

Question Data 0

'F Enter ON-100-101, SCRAM and ON-193-002, MAIN TURBINE TRIP CONTACT the Scranton System Operator to investigate the cause of the 3W and 3T 230 kV breaker trip and reclose, Restart Reactor Recirculation Pumps.

ExplanatiodJustification:

A.

B.

C.

The main generator has separated from the grid (the Main Generator Sync breaker is open). The operator should not enter ON-198-004. If the candidate does not recognize this, this answer may be chosen.

The startup bus has not been lost and the Aux buses are still energized. Scranton not the TCC should be contacted to operate the switchyard breakers. If the candidate does not recognize this, this answer may be chosen.

Power was not lost to the Aux buses or the ESS buses. The power supplies to the aux buses has transferred, but power is automatically restored. If the candidate does not recognize this, this answer may be chosen.

L4 LOC 20 As Given H:\\ExamBank\\MergeDo~s\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam D.

Correct answer. ON-100-101 and ON-193-002 should be entered simultaneously. Scranton should be contacted to operate the switchyard breakers. The RR pumps tripped on EOC-RPT and should be restarted per procedure for forced circulation through the core.

c _

Sys#

System Category KA Statement 295005 Main Turbine Generator Ability to determine and/or interpret the following as Electrical distribution status KIA#

2 9 5 0 0 ~. ~ ~ 2. 0 8 WA Importance 3.213.3 Exam Level SRO Trip they apply to MAN TURBINE GENERATOR TRIP:

(ROB RO)

References provided to Candidate None Technical

References:

ON-100-101 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis I O CFR Part 55 Content:

41.10 143.5 I 45.13 Objective:

1166 Determine if the Main Generator and Plant Task:

98.~0.

response is appropriate for any combination of

a. System mode of operation

b. Plant conditions

c. Key parameter indications 003 LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC2ONRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-

77. Unit 1 and Unit 2 have entered ON-I00 (200)-009, "CONTROL ROOM EVACUATION." Unit 1 Reactor pressure was -900 psig upon leaving the Control Room,and one and one-half hours later Reactor pressure is -100 psig.

What are the next actions to perform at the Unit 1 Remote Shutdown Panel?

A.

Be in Mode 3 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> THEN Close any open SRVs AND Align RHR Loop 'A' for Shutdown Cooling.

B.

Close any open SRVs AND Raise RPV level to 90 - 100 inches AND Align RHR Loop 'B' for Shutdown Cooling.

C.

Remain at current pressure for at least 30 minutes THEN Cooldown to 90 psig AND Raise RPV level to 90 - 100 inches.

D.

Raise RPV level to 90 - 100 inches THEN Be in Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND Align RHR Loop 'B' for Shutdown Cooling.

i--

Question Data C

Remain at current pressure for at least 30 minutes THEN Cooldown to 90 psig AND Raise RPV level to 90 - 100 inches ExplanationlJustification:

A.

B.

C.

Being in Mode 3 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is incorrect, the time is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The other listed steps in the distracter are correct.

The A loop of RHR is the preferred loop to be placed in service for shutdown cooling. Close SRVs and raise level are appropriate actions for given plant conditions.

correct answer, The cooldown rate has been exceeded and Tech Specs Section 3.4.10 RCS Pressure and Temperature (PK) Limits has to be entered. A.l Restore parameter(s) to within limits in 30 minutes is the action required. The procedure provides direction to get into shutdown cooling which is to cooldown to 90 psig, raise level and place A loop of RHR is SDC.

Being in Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is incorrect, the time is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The A loop of RHR is the preferred loop to be placed in service for shutdown cooling.

D.

Sys#

System Category KA Statement 295016 Control Room Ability to determine andlor interpret the following as Cooldown rate Abandonment they apply to CONTROL ROOM ABANDONMENT WA#

295016.~~2.06 WA Importance 3.313.5 (RO/SRO)

Exam Level References provided to Candidate ON-100-009, TS 3.4.10, Technical

References:

0~-100-009, TS 3.4.10, Steam Tables Steam tables Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.10 143.5 I 45.1 3 Objective:

3472 Direct the crew when evacuating the main control Task:

0O.ON.

Implement Plant Shutdown room.

025 From Outside Control Room LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-

78. Unit 1 is in MODE 4 when a complete loss of instrument air occurs.

What Technical Specification or Technical Requirement must be entered for the loss of Instrument Air?

A.

B.

C.

D.

TS 3.1.7, Standby Liquid Control System TS 3.7.2, Emergency Service Water System TRO 3.1 1.2.5, Ventilation Exhaust Treatment System.

TS 3.5.2, Residual Heat Removal System Question Data C

TRO 3.1 1.2.5 Ventilation Exhaust Treatment System.

ExplanationlJustification:

A.

6.

C.

D.

SLC is not inoperable since there is an alternate method for determining level in the daily surveillance.

ESW can not be lined up to TBCCW or RBCCW since there is a total loss of instrument air and the cross tie valves (ESW to TBCCW OR RBCCW) fail closed Supplying TBCCW or RBCCW is not a safety function of ESW correct answer, Loss of Turbine Building and Reactor Building HVAC filtered exhaust require entry into the ventilation exhaust treatment system.

RHR injectionispray subsystems depend on motor operated valves. Lass of air will not impact TS 3.5.2

\\-

Sys#

System Category KA Statement 295019 Partial or Complete Loss of Instrument Air Ability to determine andlor interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

(see AK2.1-AKZ.19)

Status of safety-related instrument air system loads WA#

295019.~~2.02 WA Importance 3.613.7 (RO/SRO)

Exam Level SRO References provided to Candidate Tech Spec, Tech Reg Technical

References:

ON-I 18-001 Question Source:

Modified Susquehanna, 8142004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.10 143.3 I 45.13 Objective:

4081 Identify the impact of system operability on Task:

18.ON.

Implement Loss of Technical Specifications.

003 Instrument Air LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam G

79. Unit I experiences a total loss of Shutdown Cooling with the Reactor in Mode 4 and Reactor Water level at 35 inches.

Based on the given conditions. which one of the following operator actions will you direct with respect to the Reactor Recirculation System and the basis for these actions?

A.

If Reactor Recirculation Pump is not in service, do NOT start a pump, open four SRVs.

Uses Suppression Pool as heat sink, creates natural circulation, and prevents entry into E-Plan.

6.

Check that a Reactor Recirculation Pump is in service, starting a pump if none is in service and a start is permitted. This prevents thermal stratification that may result in surface boiling and reactor pressurization.

C.

Check that a Reactor Recirculation Pump is in service, starting a pump if none is in service and a start is permitted. This meets the LCO for Shutdown Cooling - Cold Shutdown.

D.

If Reactor Recirculation Pump is not in service, do NOT start a pump, but raise level to 40 inches. If a pump is in service, this meets the LCO for Shutdown Cooling - Cold Shutdown; otherwise verify alternate means of decay heat removal.

~

Question Data B

Check that a reactor recirculation pump is in service, starting a pump if none is in service and a start is permitted. This prevents thermal stratification that may result in surface boiling and reactor pressurization.

ExplanationlJustification:

A.

B.

C.

D.

If a pump is not in service, one should be started. Using SRVs is referenced in the procedure if secondary containment established and temperature continues to rise BUT E-Plan entry will be required.

Circulation is necessary to keep the water in the core from boiling and pressurizing the vessel. This could cause a mode change without meeting the proper requirements. SRO level requiring indepth knowledge of the bases.

A pump in operation provides circulation, but does not fully comply with the LCO. The operator must still verify alternate means of decay heat removal.

If a pump is not in service, one should be started. A pump in operation provides circulation, but does not fully comply with the LCO.

The operator must still verify alternate means of decay heat removal. Water level is required to be raised above 45 inches not 40.

Sys#

System Category KA Statement 295021 WA#

295021.AA2.07 WA Importance 2.9j3.1 Exam Level SRO Loss of Shutdown Cooling Ability to determine andlor interpret the following as they apply to LOSS OF SHUTDOWN COOLING:

Reactor recirculation flow (ROE RO)

References provided to Candidate None Technical

References:

ON-149-001 Question Source:

New Susquehanna, 8142004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.10 143.5 I Objective:

3412 Predict how RHR parameters and components will Task:

49.ON.

Implement Loss Of RHR 003 Shutdown Cooling Mode 45.13 be affected by a loss of RPS when RHR is operating in shutdown cooling.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

80.

A spent fuel bundle being removed from the Reactor Vessel has been dropped in the Fuel Pool onto another fuel bundle just removed from the rReactor Vessel. The general area dose rates on the Refuel Foor have exceeded 1,000 mrem/hr.

The 'A' Channel of the Refuel Floor High Exhaust Duct monitor and the Refuel Floor Wall Exhaust Duct Monitor do NOT respond to the increased radiation levels.

A. The radioactive release caused by the dropped fuel bundle will be:

B. What initial Emergency Classification will be required by the dropped fuel bundle?

A.

A. less than 10CFR100 (REACTOR SITE CRITERIA) limits.

B. ALERT B.

A. greater than FSAR Safety Analysis values.

B. UNUSUAL EVENT C.

A. less than 1 OCFR20 (STANDARDS FOR PROTECTION AGAINST B. ALERT RADIATI0N)limits.

D.

A. less than 1 OCFR50.72 IMMEDIATE NOTIFICATION REQUIREMENTS FOR OPERATING NUCLEAR POWER REACTORS limits.

B. UNUSUAL EVENT Question Data A

A. less than 10CFR100 (REACTOR SITE CRITERIA) limits.

B. ALERT ExplanationlJustification:

A.

correct answer, See TS bases 3.3.6.2, 3,4,5,6,7 Rad monitors and associated auto action are plant design to prevent exceeding IOCFRIOO limits which is basis for FSAR. One channel will cause initiation which will isolate 2-111 HVAC and start SBGT minimizing non-treated radioactive release. The normal value of radiation on the refuel floor as read on the Area Radiation Monitors and the Process Radiation Monitors is less than 1 rnremhr. 1000 mrern/hr as given in the stem is ALERT per RA3.

The rad monitors will limit release to < (10CFR100 limits).

The rad monitors are not designed to limit release to < (IOCFRZO limits).

The rad monitors are not designed to limit release to < (10CFR50.72 reporting requirements).

is ALERT per RA3.

B.

C.

D.

1000 mremlhr as given in the stem is ALERT per RA3.

1000 mremlhr as given in the stem is ALERT per RA3.

1000 mremlhr as given in the stem Sys#

System Category KA Statement 295023 Refueling Accidents Ability to determine andlor interpret the following as Entry conditions of emergency plan WA#

295023.~~2.05 KIA Importance 3.2/4.6 Exam Level SRO they apply to REFUELING ACCIDENTS:

(RO/S RO)

References provided to Candidate EALS Technical

References:

TS 63.3.6.2 Question Source:

New Susquehanna, 81412004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.10143.51 Objective:

3787 Evaluate plant conditions during a seismic event Task:

OO.EP.0 Classify The Emergency As 45.1 3 01 Conditions Indicate.

and classify the emergency as an Unusual Event.

LOC 20 As Given H:\\ExamBank\\MergeDoc\\LOC2ONRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\--

81.

When preparing to spray the containment, you encounter an EOP instruction that states:

LIMIT FLOW TO BETWEEN 1000 AND 2800 GPM FOR FIRST 30 SECONDS What is the purpose of this instruction, and how will you direct the operator to control spray flow during this 30-second timeframe?

A.

This direction minimizes thermal shock of the drywell spray piping. During a LOCA, the drywell spray piping will become heated, and rapid initiation of drywell sprays may damage the piping. Slowly filling the piping for the first 30 seconds cools the piping.

Throttle HV-151-FO16A, DRYWELL SPRAY OB IS0 B.

This direction prevents damaging water hammer from occurring in the spray piping.

During a LOCA, the RHR Drywell spray piping may empty due to boiling. Slowly filling the piping for the first 30 seconds minimizes water hammer.

Throttle HV-151 -F027A SUPP POOL SPRAY CTL C.

This direction prevents excessive pressure drop in the drywell due to evaporative cooling upon the initiation of drywell spray. After 30 seconds there is sufficient vapor to produce conductive cooling.

Throttle HV-151-F016A, DRYWELL SPRAY OB IS0 D.

This direction will prevent causing the pump to operate at pump runout and damaging the pump motor. Limiting flow 30 seconds provides sufficient time for the system valves to respond and limit total pump flow.

Throttle HV-151 -F027A SUPP POOL SPRAY CTL Question Data C

This direction prevents excessive pressure drop in the drywell due to evaporative cooling upon the initiation of drywell spray.

After 30 seconds there is sufficient vaporto produce conductive cooling.

Throttle HV-151-FO16A. DRYWELL SPRAY OB IS0 ExplanationlJustification:

A.

B.

C.

D.

Thermal shock is not a concern Water Hammer is not a concern correct answer, Analysis has shown that by limiting flow there will not be excessive cooling in the drywell.

Pump runout is not a concern -

Sys#

System Category KA Statement 295024 High Drywell Pressure Conduct of Operations Ability to explain and apply system limits and precautions.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

-r K/A#

295024.2.1.32 WA Importance 3.413.8 (RO/SRO)

References provided to Candidate None Question Source:

New Susquehanna, 8/4/2004 Question Cognitive Level:

Fundamental Objective:

2598 For each Symptom Based EOP:

Explain the basis for each step.

Exam Level SRO Technical

References:

EO-100-103 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

Task:

OO.EO.0 Implement Primary 41.10143.21 45.12 27 Containment Control LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam v

82.

Unit 1 at 100% power when an automatic scram occurred.

The following parameters exist:

- All Control Rods inserted.

- Reactor pressure maintained 965 to 1,087 psig.

- Reactor Water Level 10 inches and lowering.

- Main Condenser Vacuum at 22.9" Hg Absolute (7" Hg Vacuum) and degrading.

- Drywell Pressure 21 psig and slowly lowering.

- Drywell Instrument Run Temperature 255 OF.

- Suppression Pool Temperature 121 OF and rising.

- Suppression Pool level at 36' 2" and rising.

- RCIC is not operating.

What failure could have caused the transient, and what actions will be required as a result of the above parameters?

A.

A steam line break between the Inboard and Outboard MSlVs has occurred.

Prevent uncontrolled condensate injection.

Rapidly lower Reactor Pressure using Bypass Valves.

Maximize Suppression Pool Cooling.

Initiate Drywell Sprays.

\\..

B.

A feedwater line break inside containment has occurred.

Control Water level between 13 and 54 inches using RCIC.

Initiate Drywell Sprays.

Lower Suppression Pool Level to less than 25 feet.

Rapidly lower Reactor Pressure using Bypass Valves.

C.

A feedwater line break inside containment has occurred.

Control Water level between 13 and 54 inches using CRD and SLC.

Maximize Suppression Pool Cooling.

Lower Suppression Pool Level to less than 25 feet.

Enter EO-I 00-1 12, "RAPID DEPRESSURIZATION."

D.

A steam line break between the Inboard and Outboard MSlVs has occurred.

Reset Main Generator Lockouts.

Control Water level between 13 and 54 inches using Condensate.

Enter EO-I 00-1 12, "RAPID DEPRESSURIZATION."

Initiate Drywell Sprays.

Lower Suppression Pool Level to less than 25 feet.

Question Data C

A feedwater line break inside containment has occurred.

Control Water level between 13 and 54 inches using CRD and SLC, Maximize Suppression Pool Cooling, Lower Suppression Pool Level to less than 25 feet, Enter EO-100-1 12, "RAPID DEPRESSURIZATION" LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForrn.doc Printed on 07/28/04

SSES LOC 20 NRC Exam ExplanationlJustification:

A.

Such a main steam line break will not add such an amount of water to the suppression pool. Rapid depressurization required due to exceeding the HCTL curve which takes precedence over anticipation of using bypass valves. If a mainsteam line break has occurred the MSlVs will be closed and Load shed on LOCA will causes the loss of Circ Water pumps. Without Circ Water the vacuum will degrade and loss of vacuum will cause bypass valve dosure.

RCIC cannot be started since SP level is above 26 feet. Rapid depressurization required due to exceeding the HCTL curve which takes precedence over anticipation of using bypass valves correct answer, A feedwater break will change containment parameters without reducing reactor pressure. Rapid depressurization required due to exceeding the HCTL curve which takes precedence. EO-I02 directs using system to maintain level. RCIC is an option but since level is above 26' in the SP RClC cannot be started. SP/T-2 directs maximizing SP Cooling,SP/T-7 directs ED on HCTL violation. SPIL-IO requires SP level to be reduced to <25 Condensate can only supply approximately 500 psig discharge pressure. A steam line break would not increase suppression pool level by this amount.

B.

C.

D.

Sys#

System Category KA Statement 295026 Suppression Pool High Ability to determine andlor interpret the following as Reactor pressure Water Temperature they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

KIA#

295026.EA2.03 WA Importance 3.914.0 Exam Level SRO (ROE RO)

References provided to Candidate EOPS Technical

References:

EO-100-103 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.1 0 I 43.5 I Objective:

2598 For each Symptom Based EOP:

Task:

OO.EO.0 Implement Primary 45.13 Explain the basis for each step.

27 Containment Control LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

83.

Unit I is at I00 % power when a SIMPLEX Panel Alarm from Photo Electric Detectors indicate a fire in Fire Zone 1-4G. The Field Unit Supervisor (FUS) is dispatched, and reports that the Steam Tunnel is full of smoke emanating from the valve motor of MAIN STEAM LINE DRAIN TO CONDENSER, HV-141 F021, and appears to be confined to the valve motor. The FUS requests that the breaker to the valve be opened.

Twenty minutes later the FUS reports that the valve is still on fire from what appears to be grease running down the motor onto the grating.

What actions are required for the given situation?

A.

Enter EO-I 00-1 04, "SECONDARY CONTAINMENT CONTROL."

Activate the Fire Brigade.

Open Breaker 1 B216112.

Classify situation as an ALERT.

Make One Hour ENS Notification.

B.

Enter ON-01 3-001 "RESPONSE TO FIRE."

Initiate Fire Pre-Plan for Reactor Building Steam Tunnel.

Open Breaker 1 B112122.

Classify situation as an UNUSUAL EVENT.

Make Four-Hour ENS Notification.

~.

C.

Enter EO-I 00-1 04, "SECONDARY CONTAINMENT CONTROL. I' Initiate Fire Pre-Plan for Turbine Building Steam Tunnel.

Open Breaker 1 B112122.

Classify situation as an ALERT.

Make Four-Hour ENS Notification.

D.

Enter ON-013-001, "RESPONSE TO FIRE."

Activate the Fire Brigade.

Open Breaker 1 B216112.

Classify situation as an UNUSUAL EVENT.

Make One-Hour ENS Notification.

Question Data D

Enter ON-013-001, "RESPONSE TO FIRE" Activate the Fire Brigade Open breaker 1B216112 Classify situation as an UNUSUAL EVENT Make One Hour ENS notification ExplanationlJustification:

A.

B.

C.

Secondary Containment Control would be entered on hi temperature not on a smoke detector or actual fire. Initially classified as a UE not an ALERT.

Requires a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report. Wrong Breaker and notification is required to be made within an hour. The breaker is the incorrect breaker as indicated by the IBIXXXXX where the 1 represents the load as being in the Turbine Building.

Secondary Containment Control would be entered on hi temperature not on a smoke detector or actual fire. Initially classified as a UE not an ALERT. The location of the valve on fire is in the Reactor Building Steam tunnel not the Turbine Building. The breaker is the incorrect breaker as indicated by the IBIXXXM where the I represents the load as being in the Turbine Building.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20N RCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam D.

correct answer, Entry into the Site Fire Protection procedure will key, activating the fire brigade and key entry into the Eplan which requires a classification for a Fire greater than 15 minutes as a UE and requires a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report. The breaker is the correct breaker as indicated by the 1B2xxxxX where the 2 represents the load as being in the Reactor Building. Classification based on OU4.

L.

Sys#

System Category KA Statement 600000 Plant Fire On Site Emergency Procedures and Plan Knowledge of fire protection WA#

600000.2.4.25 WA Importance 2.913.4 Exam Level SRO procedures.

(ROE RO)

References provided to Candidate EOPs, EP Technical

References:

ON-~I~~OI. EP Classification Matrix, Classification Matrix, NDAP-NDAP-QA-0720 QA-0720 Question Source:

New Susquehanna, 81412004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.10 145.13 Objective:

4294 STATE the actions required in response to a Fire.

Task:

13.ON.

Fire 003

\\--

LOC 20 As Given H:\\ExamBank\\MergeDoc\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L

84.

Unit 2 is operating at 100% power when power is observed to increase and stabilize at 100.4%.

What could have initiated the change?

Is this a Reactivity Anomaly requiring entry into Technical Specifications?

A.

A '6' Moisture Separator Relief Valve failing Open.

YES, a Technical Specification entry is required.

B.

A rise in Main Condenser absolute pressure.

NO, a Technical Specification entry is NOT required.

Reactor Feed Pump 'B' Min-Fow Valve failing partially Open.

YES, a Technical Specification entry is required.

Feedwater Heater 5C Emergency Dump Valve failing Open.

NO, a Technical Specification entry is NOT required.

C.

D.

Question Data D

Feedwater Heater 5C Emergency Dump Valve failing Open.

NO, a Technical Specification entry is NOT required.

ExplanationlJustification:

A.

B.

C.

D.

Moisture separator relief valve opening will rob the LP turbines of some flow and reduce feedwater heating. This is not a reactivity anomaly since it is not equal to 1% reactivity and is explained by the loss of feedwater heating.

The increase in condenser pressure will cause feedwater temperature to increase not decrease. Higher feedwater temperature will cause reactor power to decrease. This is not a reactivity anomaly since it is not equal to 1% reactivity Min flow valve will recycle some of the feedwater flow back to the condenser. This will require the feed pumps to pump more water.

This is not a reactivity anomaly since it is not equal to 1% reactivity correct answer, Opening the Emergency Dump Valve will rob the heater string of heating due to a loss of the cascading drains. The colder water will cause an increase. This is not a reactivity anomaly since it is not equal to 1% reactivity and is explained by the loss of feedwater heating.

Sys#

System Category KA Statement 295014 Inadvertent Reactivity.

Ability to determine andlor interpret the following as Cause of reactivity addition WA#

295014.~~2.03 WA Importance 4.014.3 Exam Level SRO Addition they apply to INADVERTENT REACTIVITY ADDITION:

(RO/SRO)

References provided to Candidate None Technical

References:

ON-I 56-001 Question Source:

New Susquehanna, 81412004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

41.10 143.5 I Objective:

3234 Respond to an unexplained reactivity change.

Task:

56.ON.

Implement Unexplained 45.13 L

003 Reactivity Change LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L

85. A Reactor Scram with an MSlV isolation has occurred from 100% power, and RClC is being used for level control when the following alarms and indications are received:

- All RClC Room Riley indicators in ALARM and rising.

- RClC Inboard, Outboard, and Bypass, Steam isolation Valves closed.

- RClC Steam Header pressure 900 psig constant.

- RClC INJECTION VALVE HV149FOl3, open.

- RClC Pump suction pressure upscale constant.

- SIMPLEX Panel audible alarm.

- AR-101-B05, 'RX BLDG AREA PANEL IC605 HI RADIATION."

- AR-036-BO1, "PUMP IS OPERATING."

- AR-OI6-Gl5, "FIRE PROTECTION PANEL OC650 SYSTEM TROUBLE."

What transient has occurred, and what actions are required?

A.

Feedwater discharging at a break at RClC suction.

Start all Unit Coolers.

Shut RClC INJECTION VALVE.

Monitor RClC Room temperatures lowering.

B.

Break in a steam line discharging in RClC Room.

Trip RCIC.

Start all Unit Coolers.

Monitor RClC Room temperatures lowering.

C.

Break in a steam line discharging in RClC Room.

Monitor SIMPLEX Panel.

Start all Unit Coolers.

Monitor RClC Room temperatures lowering.

D.

Feedwater discharging at a break at RClC suction or fire in RClC Room.

Initiate MANUAL RClC Isolation.

Start all Unit Coolers.

Monitor RClC Room temperatures lowering.

Question Data A

Feedwater discharging at a break at RClC suction.

Start all Unit Coolers Shut RClC INJECTION VALVE Monitor RClC Room temperatures lowering ExplanatiodJustification:

k correct answer, A break has occurred in the RClC room. The steam pressure at 900 psig in the steam piping eliminates the possibility of a steam line break. A fire would not produce a high suction pressure. A RClC suction line break with feedwater flowing back through RClC to the suction would release hot water and steam into the RClC Room. Actions per EO would increase cooling, isolate the leak by closing the injection valve which must be done manually.

Not a steam line break due to pressure in the steam piping Not a steam line break due to pressure in the steam piping Manual initiation will not isolate the leak. It will close the steam valves but will not close the RClC water flow path which is the source of the leak.

6.

C.

D.

L.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam L-Sys#

System Category KA Statement Containment Area steps.

Temperature 295032 High Secondary Conduct of Operations Ability to execute procedure WA#

295032.2.1.20 WA Importance 4.3i4.2 (RO/SRO)

Exam Level SRO References provided to Candidate EOPS Technical

References:

EO-100-104 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.10 i 43.5 I 45.1 2 Objective:

2644 Coordinate the execution of Symptom-based and Task:

OO.EO.0 Implement Secondary Event-based EOPs. {SRO Only}

28 Containment Control LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

86.

Unit I is starting up, and at 55% power, when the following alarm is received;

- AR-109-HO8, "RHR LOOP A PUMP ROOM FLOODED" Shortly after the room flooded alarm, the following alarm is received;

- AR-11 1-E02, "SUPPRESSION POOL DIV 1 LO LEVEL

The operator dispatched to investigate reports the watertight door can not be opened, and attempts to isolate the leak are unsuccessful.

What is the cause of the above indications, and what are the required actions?

A.

RHR Pumps 'A' or IC' or HPCl Suppression Pool Suction line break.

Monitor Suppression Pool Level.

Scram Reactor and reduce Pressure by opening all Main Turbine Bypass Valves.

Declare an UNUSUAL EVENT.

B.

RHR Pump 'A' Suppression Pool Suction line break.

Enter ON-I 69-002, "REACTOR BUILDING FLOODING."

Scram Reactor and reduce Reactor Pressure by opening all Main Turbine Bypass Valves.

Enter Technical Specification 3.6.2.2.

C.

RHR Pumps 'A' or 'C' Suppression Pool Suction line break.

Monitor Suppression Pool Level.

Enter EO-100-1 02, "RPV CONTROL" and scram reactor.

Declare an UNUSUAL EVENT.

D.

RClC Suppression Pool Suction line break.

Enter ON-169-002, "REACTOR BUILDING FLOODING."

Use RClC on Min Flow.

Enter Technical Specification 3.6.2.2.

Question Data C

RHR Pump 'A' or 'C' Suppression Pool Suction line break Monitor Suppression P&l Level Enter EO-100-102, "RPV CONTROL" and scram reactor Declare an UNUSUAL EVENT ExplanatiodJustification:

k No need to use BPVs to reduce Rx pressure. Rx pressure is not driving the leak. From given conditions would not be able to determine one pump or the other. SP level will stabilize at 16'. Must not be able to keep above 12" before reducing pressure by opening all BPVs in anticipation of RD No need to use BPVs to reduce Rx pressure. Rx pressure is not driving the leak. From given conditions would not be able to determine one pump or the other. SP level will stabilize at 16'. Must not be able to keep above 12" before reducing pressure by opening all BPVs in anticipation of RD correct answer, Suppression pool leak into the reactor building through the RHR piping has occurred. Entry into EO-100-103 is required. The operator should monitor level and before level reaches 17' enter RPV Control EO-100-102. Classification based on OU3 B.

C.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam D.

RClC suction from the Suppression Pool is not located in the RHR pump room.

Sys#

System Category KA Statement 295036 Secondary Containment Ability to determine andlor interpret the following as they apply to SECONDARY CONTAINMENT HIGH Cause of the high water level High SumplArea Water Level SUMPIAREA WATER LEVEL:

KIA#

295036.~~2.03 WA Importance 3.413.8 (RO/SRO)

Exam Level SRO References provided to Candidate EOPS Technical

References:

EO-I 00-103 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.I 0 I 43.5 I 45.13 Objective:

2621 Determine and interpret the following:

Task:

0O.EO.O Implement Primary 27 Containment Control

a. Reactor power

b. Reactor water level

c. Reactor vessel pressure

d. Suppression Pool water temperature

e. Suppression Pool water level

f. Drywell and Suppression Chamber pressure

g. Drywell and Suppression Chamber HZ02 concentrations

h. Drywell air temperature

i. Containment radiation

j. Reactor Building area temperature and delta-T

k. Zone 3 HVAC exhaust radiation I. Reactor Building or SGTS SPING release rate

m. Area radiation

n. Zone I or 3 HVAC de1ta-P

0. Secondary Containment water level

p. Offsite radiation releas LOC 20 As Given H :\\ExarnBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\-

87. Thirty-six hours ago, a Loss of Coolant Accident occurred on Unit 2. The plant has been cooling down slowly due to indicated fuel damage with Containment High Range Radiation Monitors reading approximately 1,500 R/hr.

The following plant conditions exist:

- All Control Rods fully inserted.

-Water level at +45 inches using RCIC.

- Reactor Pressure at 400 psig controlled with SRVs.

- Containment Pressure at 11 psig.

- Drywell H2 concentration 3.3%.

- Drywell O2 concentration 5.1 %.

- Drywell Recombiners in service.

- Suppression Chamber Pressure at 6 psig.

- Suppression Pool Temperature at 129 O F.

- Unable to sample Suppression Chamber H2 concentration.

- Suppression Chamber O2 concentration 4.8%.

- Unable to place Suppression Chamber Recombiners in service.

- RHR in Suppression Pool Cooling and Suppression Pool Spray.

What actions need to be implemented for the given plant conditions?

i A.

Continue to operate the available recombiners and monitor Drywell and Suppression Chamber H2/02 concentrations to verify a decrease in H2 and O2 levels.

Notify Chemistry to determine concentration of H2/02 gases.

B.

Notify Chemistry to determine concentration of H2/02 gases.

Notify TSC to troubleshoot and start Suppression Chamber Recombiners.

C.

Continue to operate the available recombiners and monitor Drywell and Suppression Chamber H2/02 concentrations to verify a decrease in H2 and O2 levels.

Contact TSC to enter EP-DS-001, "CONTAINMENT COMBUSTIBLE GAS CONTROL."

D.

Shut Down All Recombiners and Shut Down All DW Coolers.

Contact TSC to enter EP-DS-001, "CONTAINMENT COMBUSTIBLE GAS CONTROL."

Question Data D

Shut Down All Recombiners and Shut Down All DW Coolers Contact TSC to enter EP-DS-001, "CONTAINMENT COMBUSTIBLE GAS CONTROL" ExplanatiodJustification:

A.

6.

C.

D.

Procedure requires shutdown of recombiners (since H2 concentration in the Sup Chamber cannot be determined)

Procedure requires shutdown of recombiners (since H2 concentration in the Sup Chamber cannot be determined) not startup as given in this answer.

Procedure requires shutdown of recombiners (since H2 concentration in the Sup Chamber cannot be determined) which is not provided in this answer.

correct answer, Since H2 concentrations in the Supp Chamber cannot be determined, step PUG-4 requires shutting down all recombiners, fans, and coolers. With H2 >2%, step PUG-6 requires contacting the TSC.

Sys#

System Category KA Statement LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForrn.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 500000 High Containment Ability to determine andlor interpret the following as Combustible limits for drywell they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS:

L.

Hydrogen Concentration WA#

~ O O O O O. E A ~. O ~

WA Importance 3.313.8 Exam Level SRO (RO/S RO)

References provided to Candidate EOPS Technical

References:

EO-100-103 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.10 143.5 I Objective:

341 Predict how the Primary Containment Instrument Task:

OO.EO.0 Implement Primary 45.13 27 Containment Control will be affected by a loss of normal or alternate power.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20N RCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

=---

88.

Unit I is in Mode 2 with preparations for a Reactor Startup in progress. SO-153-004, "QUARTERLY SBLC FLOW VERIFICATION", was being performed when the STANDBY LIQ CONTROL PUMP 'A' (1 P208A) Breaker tripped open due to a faulty breaker.

How will Control Room indications respond to the breaker tripping, and what actions are required?

A.

Only 'A' Squib Valve Continuity Light will be lost.

Amber/Red light indication for 'A' SLC Pump will be lost.

Restore SLC System to service within seven days.

B.

Both Squib Valve Continuity Lights remain illuminated.

Amber/Red light indication for 'A' SLC Pump will be lost.

Restore SLC System to service within seven days.

C.

Both Squib Valve Continuity Lights remain illuminated.

Amber/Red light indication for 'A' SLC Pump will remain lit.

Restore the pump breaker 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or input a trip to PClS for the RWCU System.

D.

Only 'A' Squib Valve Continuity Light will be lost.

Amber/Red light indication for 'A' SLC Pump will remain lit.

Restore the pump breaker within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or input a trip to PClS for the RWCU System.

Question Data B

Both Squib Valve Continuity Lights remain illuminated AmberIRed light indication for 'k SLC pump will be lost.

Restore SLC System to service within 7 days Explanation/Justification:

A.

B.

C.

D.

Squib continuity lights remain on. The candidate may believe that both pump and squib power are lost correct answer, Running indication is lost. Entry into TS 3.1.7 requires restoration in 7 days.

Red light for the pump is off. PClS actions not required. The candidate may believe that PClS TS actions are required.

Red light for the pump is off. PClS actions not required. The candidate may believe that PClS TS actions are required.

Sys#

System Category KA Statement 21 1000 Standby Liquid Control Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

AC. power failures System WA#

2110oo.~2.03 WA Importance 3.213.4 Exam Level SRO (ROE RO)

References provided to Candidate TS 3.1.7, TS 3.3.6.1 Technical

References:

AR-107601107601 TS 3.1.7, TS 3.3.6.1 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 145.6 I 55.43 LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Objective:

1209 When presented with a set of Plant conditions, Task:

53.0P.O Perform Initiation Of Standby 02 Liquid Control System L.'

determine if the Standby Liquid Control System and Plant response is appropriate for any combination of

a. System mode of operation

b. Plant Conditions

c. Key parameter indications LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

'1-

89.

Mode Switch in startup on Unit 2 with the following conditions:

- Reactor Power 10% on APRMs.

- Reactor Pressure 950 psig.

- Reactor Level 36 inches.

- Two Bypass Valves open.

- MAIN STEAM LINE A OB ISO, HV141F028A closed.

For these conditions, which alarm will clear when the Mode Switch is taken to the RUN position?

What plant response will occur, and what actions must be taken if the alarm DOES NOT clear?

A.

B.

C.

D.

AR-111-B04, "MAIN CONDENSER LO VAC TRIP A/C BYPASS1 A Half Scram If the alarm does not clear, investigate the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low RPS functions operability in Technical Specifications.

AR-104-D03, MSIV CLOSURE BYPASS" A Half Scram If the alarm does not clear, investigate the Main Steam Isolation Valve -Closure RPS function operability in Technical Specifications.

AR-104-DO3, "MSIV CLOSURE BYPASS" No Plant response.

If the alarm does not clear, investigate the Reactor Mode Switch - Shutdown Position RPS function operability in Technical Specifications.

AR-111-804, "MAIN CONDENSER LO VAC TRIP A/C BYPASS" No Plant response.

If the alarm does not clear, investigate the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low RPS functions operability in Technical Specifications.

Question Data C

AR-104-DO3, "MSIV CLOSURE BYPASS" No Plant response If the alarm does not clear, investigate the Reactor Mode Switch - Shutdown Position RPS function Operability in Technical Specifications.

Explanation/Justification:

-L-'

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam A.

This alarm should clear when the mode switch is taken to run even if the keylock switches are still in bypass. No half scram will occur since there is no low vacuum condition. This is a turbine trip and the RPS functions on turbine control valves have not been challenged.

MSlV closure is a scram signal but requires two MSlVs to give at least a half scram. The Scram on MSlV closure has not been challenged.

correct answer, The alarm should be automatically unbypassed when the mode switch is taken to RUN. If it does not clear, no response will occur since it requires 2 MSlVs to give at least a half scram. The fact that it did not unbypass calls into question the operation of the mode switch.

This alarm should clear when the mode switch is taken to run even if the keylock switches are still in bypass. This is a turbine trip and the RPS functions on turbine control valves have not been challenged.

B.

C.

D.

Sys#

System Category KA Statement 212000 Reactor Protection Ability to (a) predict the impacts of the following on System the REACTOR PROTECTION SYSTEM; and (b) based position on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Changing mode switch WA#

212000.~2.16 WA Importance 4.014.1 (ROE RO)

Exam Level SRO References provided to Candidate TS 3.3.1.1 Technical

References:

AR-104-001 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 I 45.6 /

55.43 Objective:

2495 Predict how the Reactor Protection System and Task:

58.SO.O Perform 24 Month Reactor supported systems respond to the following 02 Mode Switch Shutdown failures:

Position Functional Check

a. RPSMGSet

b. PA Breakers c Backup Scram Valves

d. Scram Pilot Valves

e. Scram Reset Switch

f. Reactor Mode Switch

g. SDV Vent and Drain Valves (fail to close/open)

h. SDV Vent and Drain Pilot Valves

i. SDV High Level Bypass Switch LOC 20 As Given H :\\ExamBan k\\MergeDocs\\LOC20NRCFom.doc Printed on 07/28/04

SSES LOC 20 NRC Exam A.

This alarm should clear when the mode switch is taken to run even if the keylock switches are still in bypass. No half scram will occur since there is no low vacuum condition. This is a turbine trip and the RPS functions on turbine control valves have not been challenged.

MSlV closure is a scram signal but requires two MSlVs to give at least a half scram. The Scram on MSlV closure has not been challenged.

correct answer, The alarm should be automatically unbypassed when the mode switch is taken to RUN. If it does not clear, no response will occur since it requires 2 MSlVs to give at least a half scram. The fact that it did not unbypass calls into question the operation of the mode switch.

This alarm should clear when the mode switch is taken to run even if the keylock switches are still in bypass. This is a turbine trip and the RPS functions on turbine control valves have not been challenged.

\\-

B.

C.

D.

Sys#

System Category KA Statement 212000 Reactor Protection Ability to (a) predict the impacts of the following on System the REACTOR PROTECTION SYSTEM; and (b) based position on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Changing mode switch WA#

212000.~2.16 WA Importance 4.014.1 Exam Level SRO (RO/SRO)

References provided to Candidate TS 3.3.1.I Technical

References:

AR-IO~-OOI Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 I 45.6 Objective:

2495 Predict how the Reactor Protection System and Task:

58.SO.O Perform 24 Month Reactor supported systems respond to the following 02 Mode Switch Shutdown failures:

Position Functional Check

a. RPSMGSet

b. EPA Breakers c Backup Scram Valves

d. Scram Pilot Valves

e. Scram Reset Switch

f. Reactor Mode Switch

g. SDV Vent and Drain Valves (fail to closelopen)

h. SDV Vent and Drain Pilot Valves

i. SDV High Level Bypass Switch LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

90.

Unit I is in a refueling outage. Maintenance has signed off of the clearance for the replacementhesting of SRVs. Operations is performing the restoration valve line-upCL-I 25-001 2, "UNIT I CONTAINMENT INSTR GAS SYSTEM MECHANICAL", as outlined in the Outage Schedule in preparation for Post-Maintenance Testing.

The operator performing the valve line-up calls the Control Room to report a leak atPSV-141 F013G Solenoid SV-14113G3 (See attached). If the valve line-up is not completed, the Post-Maintenance Testing scheduled to start in the morning will be delayed.

What actions will be taken for the leaking solenoid?

A.

B.

C.

D.

1. Classify leaking solenoid as Emergent Work; contact Outage Management IAWOP-

2. Isolate the leak, and apply a Status Control Tag IAW NDAP-QA-0302, "SYSTEMSTATUS AND EQUIPMENT CONTROL."

3. Place CIG CL on hold, and initiate "ABORTED EVOLUTION CONTROL LOG

4. Declare six SRVs inoperable. Initiate tracking an LCO entry due to inoperableSRVs.

AD-002, "STANDARDS FOR SHIFT OPERATIONS."

FORMI'IAW OP-AD-002, "STANDARDS FOR SHIFT OPERATIONS."

1. Reapply the original clearance IAW NDAP-QA-0322, "ENERGY CONTROL PROCESS."

2. Contact Outage Management IAW OP-AD-002, "STANDARDS FOR SHIFT OPERATIONS."

3. Place CIG CL on hold, and initiate "ABORTED EVOLUTION CONTROL LOG

4. The inoperable ADS Valve function must be restored in 14 days.

FORMI'IAW OP-AD-002, "STANDARDS FOR SHIFT OPERATIONS1'

1. Classify leaking solenoid as Emergent Work; contact Outage Management IAWOP-

2. Isolate the leak and apply a Status Control Tag IAW NDAP-QA-0302, "SYSTEM STATUS AND EQUIPMENT CONTROL."

3. Complete remainder of the CIG Valve Line-up.

4. No Technical Specification LCO must be entered or tracked.

AD-002, "STANDARDS FOR SHIFT OPERATIONS."

1. Reapply the original clearance IAW NDAP-QA-0322, "ENERGY CONTROLPROCESS."

2. Initiate AR-CR to document As-Found Condition of solenoid.

3. Isolate leaking solenoid, document in alternate position IAW OP-AD-092, "CHECK-OFF LIST PROGRAM."

4. Initiate tracking an LCO entry due to one SRV inoperable.

Question Data C

1. Classify leaking solenoid as Emergent Work, contact Outage Management IAW OP-AD-002, "STANDARDS FOR SHIFT OPERATIONS".

2. Isolate the leak and apply a Status Control Tag, IAW NDAP-QA-0302 "SYSTEM STATUS AND EQUIPMENT CONTROL".

3. Complete remainder of the CIG Valve Line up.

4. No Technical Specification LCO must be entered or tracked.

LOC 20 As Given H:\\ExarnBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07128104

~-

SSES LOC 20 NRC Exam ExplanationlJustification:

A.

=---

There is no reason to stop the rest of the lineup and complete the papework for an aborted evolution if the candidate believes this, this answer may be chosen. The TS for SRV operability does not apply now since the Rx is not in Modes 1,2, or 3, but even if the Rx was in one of those modes, the TS would be NA since it requires the safety function only which is not impacted by the leak and isolation Once the original clearance has been removed it cannot be reapplied. A new clearance would be needed. If the operator believes that is permitted, this answer may be chosen. Management should be notified but there is no reason to stop the rest of the lineup and complete the paperwork for an aborted evolution. The leak on the valve does not impact ADS and reactor pressure is less than 150#

thus ADS is not affected and even if it was the TS does not apply to these conditions correct answer, This fits the definition of emergent work and should be handled as emergent work. The leak must be isolated and since there is no procedure controlling the valve position to isolate the leak, a status control tag should be used. The remainder of the lineup may be completed since it is not impacted by the leak.

No TS entry or tracking is required since the leak is on the solenoid for the relief function. The ADS and Safety functions are not impacted and those functions are the only ones required by TS.

Once the original clearance has been removed it cannot be reapplied. A new clearance would be needed. If the operator believes that is permitted, this answer may be chosen. An AR would be written. The valve out of its normal position would be controlled by a status control tag. The SRV TS does not apply since it is only the safety function. Also to isolate the leak, the supply to 6 valves would be cut off.

B.

C.

D.

_c_

Sys#

System Category KA Statement 239002 RelieWSafety Valves Equipment Control Knowledge of the process for managing maintenance activities during shutdown operations.

WA#

239002.2.2.18 WA Importance 2.313.6 (RO/SRO)

Exam Level SRO References provided to Candidate M-141 Sht I, TS 3.5.1 Technical

References:

NDAP-QA-D~OP, M-141 sht I Question Source:

New Susquehanna, 81412004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

43.5 145.13 Objective:

894 Use NDAP-QA-0302, "System Status and Task:

0o.m.0 Implement Appropriate And Equipment Control Equipment Control" to:

38 Portions Of System Status

a. Identify responsibilities for the following positions

- Work GroupMlorker

- Maintenance LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

91.

A step needs to be added to SO-024-001, "MONTHLY DIESEL GENERATOR OPERABILITY TEST", at step "5.1 2 1, Sample Diesel Fuel Oil Day Tank OT528A (B) (C) (D) for accumulated water as follows: 'I to put a turbidity meter on the sample connection."

A Procedure Change Process Form (PCAF) is initiated.

Which of the below steps must be performed before the PCAF can be issued?

A.

Select NOT Administrative change; Individual making change signs Initiator and Reviewer Block; Have change reviewed by knowledgeable individual; Determine if cross-disciplinary review required; Obtain PORC Approval; Obtain approval signature.

B.

Select Administrative change; Individual making change signs Initiator and Reviewer Block; Obtain approval signature.

C.

Select NOT Administrative change; The individual making the change signs the Initiator Block; Determine if a I OCFR50.59 Evaluation is required; Have change reviewed by knowledgeable individual; Determine if cross-disciplinary review required; Obtain approval signature.

D.

Select Administrative change; Determine if a 1 OCFR50.59 Evaluation is required; Individual making change signs Initiator Block; Obtain approval signature.

Question Data C

Select NOT Administrative change, The individual making the change signs the Initiator Block, Determine if a 10CFR50.59 evaluation is required, Have change reviewed by knowledgeable individual, Determine if cross-disciplinary review required, Obtain approval signature.

ExplanatiodJustification:

A

6.

C.

D.

These steps are not correct as there is no determination of the need for a 50.59 review and no need for PORC approval prior to issuance, If the candidate believes that PORC approval is required this answer may be chosen.

This does not meet the definition of an administrative change since the actual process is being changed by adding an action. If the candidate believes that this meets the criteria for an administrative change this answer may be selected.

correct answer, This represents the steps necessary to approve a PCAF changing the actual process in the procedure.

This does not meet the definition of an administrative change since the actual process is being changed by adding an action. If the candidate believes that this meets the criteria for an administrative change this answer may be selected.

Sys#

System Category 262001 A.C. Electrical Distribution Equipment Control WA#

262001.2.2.6 WA Importance 2.313.3 References provided to Candidate None Question Source:

New Susquehanna, 81412004 Question Cognitive Level:

Comprehension Objective:

397 Initiate procedure changes.

(RO/S RO)

KA Statement Knowledge of the process for making changes in procedures as described in the safety analysis report.

Exam Level SRO Technical References :

NDAP-QA-0002 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

Task:

00.AD.O Implement Nuclear 43.3 145.13 28 Department Procedure Program LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam I.

92.

Unit 1 is at 100% power with 'A' CRD Pump out of service for impeller replacement.

Investigation of TBCCW Head Tank Low Level Alarm revealed the 'B' CRD Pump, Cooling Water Relief Valve (PSV-11411 B) Line broken off.

After the leak is isolated, which of the following operator actions are required?

What notifications would be required by NDAP-QA-0720, "STATION REPORT MATRIX AND REPORTABI LlTY EVALUATION GUIDANCE?"

A.

Immediately scram the reactor and trip the Reactor Recirculation Pumps. The RWCU System may remain in service.

ONE HOUR ENS NOTIFICATIONS B.

Scram the reactor within 20 minutes of the second accumulator alarm on low accumulator pressure. The Reactor Recirculation Pumps and RWCU System may remain in service.

FOUR HOUR ENS NOTIFICATIONS C.

Scram the reactor within 20 minutes of the second accumulator alarm on low accumulator pressure. The Reactor Recirculation Pumps and RWCU System may remain in service.

ONE HOUR ENS NOTIFICATIONS D.

Immediately scram the reactor and trip the Reactor Recirculation Pumps. The RWCU System must be secured.

FOUR HOUR ENS NOTIFICATIONS Question Data B

Scram the reactor within 20 minutes of the second accumulator alarm on low accumulator pressure. The Reactor Recirculation Pumps and RWCU system may remain in service.

FOUR HOUR ENS NOTIFICATIONS Explanation/Justification:

A.

B.

If the candidate believes that the loss of seal flow to the RR pumps requires the pumps to be shutdown this answer will be chosen.

Also must believe a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report required correct answer, Cooling water is lost to the in Service CRD pump with no standby pump available requiring entry into loss of CRD.

Since reactor pressure is greater than 900 psig, TS allows 20 minutes to recover CRD. Given this set of circumstances, CRD will not be recovered in time. RR and CRD can continue to operate as long as RBCCW is in service.

If the candidate believes that a one hour report is required, then will choose this answer.

If the candidate believes that the loss of seal flow to the RR pumps requires the pumps to be shutdown this answer will be chosen,

C.

D.

Sys#

System Category KA Statement LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 201001 Control Rod Drive Hydraulic System Ability to (a) predict the impacts of the following on the CONTROL ROD DRIVE HYDRAULIC SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Low cooling water flow L-WA#

2oio01.~2.13 KIA Importance 2.712.8 Exam Level SRO (ROE RO)

References provided to Candidate M-114 Sh I, NDAP-QA-Technical

References:

ON-I 55-007, NDAP-QA-0720 0720 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 145.6 Objective:

535 Determine and locate the appropriate procedure Task:

55.ON.

Implement Loss Of CRD to use for noamrl, abnormal and emergency 014 System Flow Control Rod Drive Hydraulics System operations.

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

93. Unit I is operating at 100 % power with the following alarms acknowledged on 1C651:

- AR 101-H16, "RX LEVEL XMTR C TURBINE TRIPS BYPASSED"

- AR 103-COI, "RX VESSEL LO LEVEL TRIP The alarms are caused by I&C performing SI-I 80-305, "QUARTERLY CALIBRATION OF REACTOR VESSEL WATER LEVEL CHANNEL LIS-B21-1 N024A" at the 1 COO4 Instrument Rack.

I&C reports that LIS-B21-1 N024A, "RX WTR LVL 3 AUTO SCRAM TRIP LOGIC AI" Barton DP Cell Diaphragm has failed while performing the calibration check.

If I&C were to return the DP Cell to service, what would be the status of AR 103-COI, "RX VESSEL LO LEVEL TRIP" alarm?

What Technical Specification actions would be required based on this failure?

1 A.

B.

C.

D.

Alarm would remain in ALARM.

No actions required.

Alarm would be CLEAR.

Place channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Restore the channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or declare RClC inoperable.

Alarm would be CLEAR.

Be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Restore the channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or declare RClC inoperable.

Alarm would remain in ALARM.

Place channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Question Data B

Alarm would be CLEAR.

Place channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Restore the channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or declare RClC inoperable.

ExplanationNustification:

A.

If the candidate believes that 0 dp corresponds to low level then this answer would be chosen. Since the alarm condition will be in and thus the channeled tripped, based on the misconception above, the candidate may believe no further action needed.

LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam B.

correct answer, Failure of the diaphragm results in 0 dp which corresponds to a high level. When the DP cell is placed in service, the high level sensed by the failed cell would cause the low level alarm to clear.

Per TS 3.3.1.1 Condition A The channel or trip system must be placed in trip in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Alarm would clear. If the candidate mistakenly uses Table 3.3.1.I-1 and follows the conditions referenced from D.l column then this is the proper action. D.l is only applicable if action A, B, or C is not met If the candidate believes that 0 dp corresponds to low level then this answer would be chosen. This is the proper TS action.

c C.

D.

~~

~

Sys#

System Category KA Statement 21 6000 Nuclear Boiler Ability to (a) predict the impacts of the following on based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Detector diaphragm failure or Instrumentation the NUCLEAR BOILER INSTRUMENTATION; and (b) leakage W M 216000.~2.04 WA Importance 2.913.0 Exam Level SRO (RO/SRO)

References provided to Candidate TS 3.3.1.1 Technical

References:

SI-180-305, TS 3.3.1.1 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.5 I 45.6 Objective:

5499 Describe the reactor vessel level instrumentation Task:

system at SSES.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

\\.-

94. The plant was operating at 20% power. Chemistry reported to the Main Control Room the following chemistry parameters:

- Reactor pH 8.8

- Reactor Water conductivity

- Reactor Water chlorides 11 micromhos/cm 150 ppb Six hours later with the plant in Mode 2, Chemistry reports the following:

- Reactor pH 6.5

- Reactor Water conductivity

- Reactor Water chlorides 0.9 micromhos/cm 150 ppb Which one of the following actions is appropriate for these plant conditions?

A.

Be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

B.

Be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

C.

Stay in Mode 2, and restore chlorides to within limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

L-D.

Restore chlorides to within limits within 66 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br />, and verify that the cumulative time exceeding the limit is less than or equal to 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> in the past year.

Question Data B

Be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Explanation/Justification:

A.

B.

This is Condition E If the candidate does not recognize that 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> have already passed, this answer may be chosen.

correct answer, Initially conductivity is too high (greater than 1.0) and pH is too high (above 8.6). Chlorides are within spec

(<200ppb). Since conductivity is greater than 10, TRM 3.4.1 Condition E applies. Since the event has been in progress for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> are left to reach Mode 3 and 30 Hours are left to reach Mode 4 This is the action for CI out of spec in Mode 2 (TRM 3.4.1 Conditions F and G). If the candidate does not recognize Condition E applies, this answer may be chosen.

This is TRM 3.4.1, Condition B less the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> already used. If the candidate does not recognize that Condition B does not apply, this answer may be chosen.

C.

D.

Sys #

System Category KA Statement None Generic Conduct of Operations Knowledge of conditions and limitations in the facility license.

wA## 2.1.10 KIA Importance 2.713.9 Exam Level SRO (ROB RO)

References provided to Candidate TRM 3.4.1 Technical

References:

TRM 3.4.1 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 4 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

43.1 145.13 Objective:

2099 Determine if a component or system is required to Task:

0O.ON.

Implement Chemistry L--

be operable, per Technical Specification.

020 Anomaly LOC 20 As Given H:\\ExarnBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam c

95. Units 1 and 2 are at 100% power.

'A' Diesel Generator is out of service. Maintenance activities are on-going.

Mechanical Maintenance has requested permission to perform the feed and bleed portion of the work instructions. The feed and bleed will be done on the Jacket Water Stand Pipe by opening the demin water supply, filling the Stand Pipe to the high level, closing the demin water supply and opening the Standpipe Drain to the low level alarm and closing the drain.

Can Maintenance perform the requested actions?

If yes, what additional requirements must be met? If no, why not?

A.

Maintenance can perform the requested activities if Ops Supervision has released the appropriate Work Instructions and equipment is CAUTION Tagged.

B.

Maintenance can perform the requested activities if Ops Supervision has released the appropriate Work Instructions and equipment is restored prior to the end of the workers shift.

C.

Maintenance CANNOT perform the requested activities Only Operations personnel are allowed to manipulate plant equipment.

D.

Maintenance CANNOT perform the requested activities unless Operations personnel are present to oversee the manipulations.

Question Data Maintenance can perform the requested activities if Ops Supervision has released the appropriate Work Instructions and equipment is restored prior to the end of the workers shift.

B ExplanationlJustification:

A.

B.

C.

D.

The procedure states that operations will determine the need for status control tags. If the candidate confuses this with a requirement for caution tags, then this answer will be chosen.

correct answer, Per NDAP-QA-0302 Section 6.6 with operations permission and an appropriate instruction. Operations must be notified if the work is not complete at the end of the shift.

As a general practice, only operations is allowed to operate plant components, but some exceptions are allowed. If the operator does not recognize that this is a permitted exception, this answer may be chosen.

As a general practice, only operations is allowed to operate plant components, but some exceptions are allowed. The presence of operations personnel is not required as part of the exception. If the operator does not recognize that this is a permitted exception, this answer may be chosen.

Sys#

System None Generic Category Equipment Control KA Statement Knowledge of the process for managing maintenance activities during power operations.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam Iv KIA#

2.2.1'1 KIA Importance 2.313.5 Exam Level SRO (RO/S RO)

References provided to Candidate None Technical

References:

NDAP-QA-O~OZ Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

43.5 145.13 Objective:

1039 Utilizing Procedure NDAP-QA-0302, "System Task:

OO.AD.0 Implement Appropriate 38 Portions Of System Status And Equipment Control Status and Equipment Control", define the assigned responsibilities by position, as they relate to the Energy Control Process for the SSES System Status and Equipment Control Program.

These positions shall include Operations.

\\-

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20N RCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

'L

96.

The following work has been performed on the HPCl System with the Unit at 100% power:

The HPCl PUMP TURBINE AUXILIARY OIL PUMP I P213 has been overhauled and rebuilt, The HPCl VACUUM TANK DRAIN CONDENSATE PUMP1 P215 has been overhauled and rebuilt.

1. Which of the following Post-Maintenance Testing actvities will be completed by the responsible work group to close out the work packages?

2. Which of the following Operational Testing actvities will be entered on the RLWO Surveillance Testing Screen by Operations Supervision?

a) Component calibration of the HPCl Lube Oil System Oil Pressure and Temperature b) HPCl piping is filled with water from the Pump Discharge Valve to the Injection Valve.

c) Mechanical alignment of HPCl Aux Oil Pump.

d) HPCl Subsystem manual, power-operated, and automatic valves in the flowpath are in e) Motor bump starts to check rotation of the HPCl VACUUM TANK DRAIN CONDENSATE f) Observation of running HPCl VACUUM TANK DRAIN CONDENSATE PUMP 1 P215 for g) The HPCl System can develop a flowrate equal to or greater than 5,000 gpm against a Indicators.

the correct position.

PUMP, lP215.

unusual noise, temperature, or vibration.

system head corresponding to reactor pressure.

L I.

Post-Maintenance Testing

2. Operational Testing Question Data A

a), c), e), 9 ExplanationlJustification:

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam A.

correct answer, the post maintenance testing items listed in the stem are examples taken from the "SYSTEM STATUS AND EQUIPMENT CONTROL" procedure NDAP-QA-0302. The operational testing items listed in the stem are Tech Spec surveillances for the HPCl system.

a) Post Maintenance Testing - Component calibration of the HPCl Lube Oil System Oil Pressure and Temperature indicators.

b) Operational Testing - HPCl piping is filled with water from the pump discharge valve to the injection valve.

c) Post Maintenance Testing - Mechanical alignment of HPCl Aux Oil Pump.

d) Operational Testing - HPCl subsystem manual, power operated, and automatic valves in the flow path, are in the correct e) Post Maintenance Testing - Motor bump starts to check rotation of HPCl Aux Oil Pump.

9 Post Maintenance Testing - Observation of running HPCl Aux Oil Pump for unusual noise, temperature, or vibration.

g) Operational Testing - The HPCl pump can develop a flow rate greater than or equal to 5000 gpm against a system head position.

corresponding to reactor pressure.

b) is a HPCl surveillance not post maintenance testing.

d) is a HPCl surveillance not post maintenance testing.

b) is a HPCl surveillance not post maintenance testing.

d) is a HPCl surveillance not post maintenance testing.

B.

C.

D.

Sys#

System None Generic WA#

2.2.21 Category Equipment Control WA Importance 2.313.5 (ROISRO)

References provided to Candidate None Question Source:

New Susquehanna, 8/4/2004 Question Cognitive Level:

Comprehension Objective:

894 Use NDAP-QA-0302, "System Status and Equipment Control" to:

a. Identify responsibilities for the following positions

- Work GroupNVorker

- Maintenance KA Statement Knowledge of pre and post maintenance operability requirements.

Exam Level SRO Technical

References:

NDAP-QA-0302 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

Task:

00.m.0 Implement Appropriate And Equipment Control 43.2 38 Portions Of System Status LOC 20 As Given H:\\ExarnBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

97. With Unit I at 100% power, a special entry is being planned into the RWCU Backwash Receiving Room to operate a valve. A broken reach rod requires operation of the valve in a very high radiation area. The job is being planned with portable shielding to minimize the dose.

To enter the room, operate the valve, and exit the room, a Whole Body dose of 3,600 mrem will be received, and a shallow dose equivalent of 48,000 mrem to the hands is expected.

An operator with 120 mrem TEDE for the year, plus 400 mrem received from a medical procedure, has volunteered to perform the job.

Using NDAP-QA-0625, "Personnel Radiation Monitoring Program", determine if the volunteer may enter the RWCU Backwash Receiving Tank Room to perform the job.

If yes, state what papenvork must be completed to authorize the extension.

If no, state the reason that the extension is not allowed.

The volunteer:

A.

may NOT be authorized to perform the entire job. The expected dose to the hands will exceed the extremities limit, and an extension CANNOT be authorized.

L-B.

may NOT be authorized to perform the entire job. The expected TEDE combined with the current dose will be greater than 4,000 mrem limit, and an extension CANNOT be authorized.

C.

may be authorized to perform the entire job. A single Form NDAP-QA-0625-1, DOSE EXTENSION FORM, must be completed to authorize the TEDE and extremity dose limit extension.

D.

may be authorized to perform the entire job. Two Forms NDAP-QA-0625-1, DOSE EXTENSION FORM, must be completed, one to authorize the TEDE extension, and the second to authorize the extremity dose limit extension.

Question Data A

may NOT be authorized to perform the entire job. The expected dose to the hands will exceed the extremities limit and an extension CANNOT be authorized.

ExplanatiodJustification:

A.

correct answer, Per NDAP-QA-0625 section 6.2.1 the max dose for extremities is 40000 mrem. Section 6.3.4 states that exceeding ant of the admin limits other than TEDE should be considered using the guidance for PSE. Section 3.4.5 prohibits PSE during normal operations, thus an extension for the extremities dose cannot be authorized.

The expected TEDE with current dose will be less than 4000 mrem. If the candidate believes that medical dose is included in occupational dose TEDE the this answer will be chosen.

If the candidate believes that the extension may be granted, then this answer may be chosen.

If the candidate believes that the extension may be granted but that separate authorization of each limit is required, then this answer may be chosen.

B.

C.

D.

Sys#

System 1-Category KA Statement LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOCZONRCForm.doc Printed on 07/28/04

L SSES LOC 20 NRC Exam None Generic Radiological Controls WA#

2.3.4 KIA Importance 2.513.1 References provided to Candidate NDAP-QA-0625 Question Source:

New Susquehanna, 8/4/2004 Question Cognitive Level:

Analysis (RO/SRO)

Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

Exam Level SRO Technical

References:

NDAP-00-0625 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

43.4 I 45.1 0 Objective:

4347 DESCRIBE the access and control requirements Task:

for:

a. Radiation Areas

b. High and Very High Radiation Areas LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

.Ld'

98.

Unit 1 was at 100% power when a MSlV isolation occurred.

- Drywell pressure is currently at 1.2 psig.

- Drywell temperature is 146 OF.

- Suppression Chamber is currently at 1.7 psig.

- Suppression Pool Temperature is 98 O F.

What directions will be given for controlling Containment pressure?

A.

Purge Suppression Chamber in accordance with EP-DS-004, "PRIMARY CONTAINMENT AND RPV VENTING."

Notify Chemistry to Obtain/Analyze Noble Gas sample prior to venting.

B.

Vent the Drywell in accordance with EP-DS-004, "PRIMARY CONTAINMENT AND RPV VENT1 NG."

Monitor for SPING Alarms, and stop venting if in alarm.

C.

Purge Drywell using the "A" and/or " B SBGT Train@) in accordance with EO-100-103, "PRIMARY CONTAINMENT CONTROL", step PC/P-1.

Monitor for SPING Alarms, and stop venting if in alarm.

L' D.

Vent the Suppression Chamber using the "A" and/or "BI SBGT Train@) in accordance with OP-173-003, "PRIMARY CONTAINMENT NITROGEN MAKEUP AND VENTING."

Notify Chemistry to Obtain/Analyze Noble Gas sample prior to venting.

Question Data D

Vent the Suppression Chamber using the "A" and/or " B SBGT train(s) in accordance with OP-173-003, "PRIMARY CONTAINMENT NITROGEN MAKEUP AND VENTING".

Notify Chemistry to ObtainIAnalyze noble gas sample prior to venting.

ExplanationlJustification:

A.

B.

C.

The candidate may misread the EOPs and believe that guidance from the TSC is needed to vent. EP-DS-004 is used when venting to maintain the containment below 65 psig. It is not needed at this time.

The candidate may misread the EOPs and believe that guidance from the TSC is needed to vent. EP-DS-004 is used when venting to maintain the containment below 65 psig. It is not needed at this time.

Use of the EOPs is not applicable since containment pressure is less than 1.72. This is a viable distracter since the candidate may believe the EOPs are applicable and the candidate does not understand the difference of venting and purging. Also when venting either or both SBGTs may be used but when purging only 1 SBGT may be used.

correct answer, OP-173-003 directs venting the Suppression Chamber. Either or both trains of SBGT may be used. Since SRVs have opened a noble gas sample is required.

D.

Sys#

System None Generic Category Radiological Controls WA#

2.3.8 WA Importance 2.313.2 References provided to Candidate EOPS Question Source:

New Susquehanna, 8/4/2004 Question Cognitive Level:

Analysis (RO/SRO) 1-LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOCPONRCForm.doc Printed on 07/28/04 KA Statement Knowledge of the process for performing a planned gaseous radioactive release.

Exam Level SRO Technical

References:

EO-100-103, o~-i73-003 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

43.4 145.10

SSES LOC 20 NRC Exam Objective:

2644 Coordinate the execution of Symptom-based and Task:

73.0P.O Perform Venting Drywell Event-based EOPs. {SRO Only}

01 LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam

99.

Unit I is operating at 100% power when the following alarms are received:

- TBCCW PUMP A MOTOR TRIP (AR-123-G01)

- TBCCW PUMPS DISCHARGE HEADER LO PRESS (AR-123-G03)

The following alarms are received after the TBCCW alarms:

- AR-016-H16, FUEL POOL FlLT DEMIN PANEL OC207 SYSTEM TROUBLE (AR-OI6-Hl6)

- AR-I 06-Dl 5, TURB BLDG SMPL STATION 1 C132 SYSTEM TROUBLE

- PlCSY Hi Temperature Alarm 250 OF on CONDENSATE PUMP MOTOR STATOR For the second group of alarms, which alarm(s) islare a result of the initial TBCCW Alarms, and what actions are to be taken?

A.

B.

C.

D.

AR-OI6-HI6 and AR-106-D15 Enter ON-I 15-001, "LOSS OF TURBINE BUILDING CLOSED COOLING WATER."

Dispatch Operator to OC207 and monitor Instrument Air Pressure.

PlCSY Hi Temperature Alarm Reduce Reactor Power in accordance with Reactor Engineering instructions in CRC Book.

Remove Condensate Pump from service IAW OP-144-001, "CONDENSATE SYSTEM."

PlCSY Hi Temperature Alarm and AR-106-Dl5 Enter ON-I 15-001, "LOSS OF TURBINE BUILDING CLOSED COOLING WATER."

Dispatch Operator to Condensate Pumps to check Motor Cooling.

PlCSY Hi Temperature Alarm Reduce Reactor Power in accordance with Reactor Engineering.

Closely Monitor RFP SUCT PRESS PR-10609.

Remove Reactor Feed Pump from service in accordance with OP-145-001, "RFP."

Question Data A

AR-016-Hl6 and AR-106-DI5 Enter ON-115-001, "LOSS OF TURBINE BUILDING CLOSED COOLING WATER Dispatch Operator to OC207 and monitor Instrument Air Pressure ExplanationlJustification:

A.

correct answer, The indication provided with the alarm of low discharge pressure indicates that the standby pump did not start or was slow starting indicating a total loss of TBCCW flow. A total loss of TBCCW flow or low pressure will cause a trip of the Instrument and Service Air Compressors which will be first noted by control room low instrument air pressure alarms and the OC207 trouble alarm.

TBCCW does not cool the Condensate pump motor, but does cool the pump motor bearings. A PlCSY alarm would be expected from one of the motor bearings not from the stator. Candidate must recognize the difference in TBCCW cooling loads.

B.

i--

LOC 20 As Given H:\\ExamBan k\\MergeDocs\\LOC20NRCForrn.doc Printed on 07/28/04

SSES LOC 20 NRC Exam C.

D.

TBCCW does not cool the Condensate pump motor, but does cool the pump motor bearings. A PICSY alarm would be expected from one of the motor bearings not from the stator. Candidate must recognize the difference in TBCCW cooling loads.

TBCCW does not cool the Condensate pump motor, but does cool the pump motor bearings. A PlCSY alarm would be expected from one of the motor bearings not from the stator. Candidate must recognize the difference in TBCCW cooling loads.

L-Sys#

System Category KA Statement None Generic Emergency Procedures and Plan Ability to prioritize and interpret the significance of each annunciator or alarm.

wA#

2.4.45 WA Importance 3.313.6 (ROB RO)

Exam Level References provided to Candidate None Technical

References:

ON-I 15401 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

43.5 145.3 I 45.12 Objective:

1766 Describe the support function the following Plant Task:

15.ON.

Implement Loss Of Turbine Systems provide for Instrument Air System 003 Building Closed Cooling operation.

Water

a. AC Distribution

b. TBCCW System

c. Service Air System

\\--

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC20NRCForm.doc Printed on 07/28/04

L, 100

'L SSES LOC 20 NRC Exam With Unit 1 at 100% power, I&C reports that investigation of the AR-111-D02, "CONTAINMENT ACCIDENT RANGE DIV 1 HIGH RADIATION", Upscale Alarm is a result of the same circuit board failure which occurred to the Division 2 Channel two months ago. The Division 2 Channel repair is waiting on a circuit board to arrive from the vendor.

What actions are required to be taken for the second failure?

A.

Notify the Emergency Plan Duty Planner of the lost input to the Emergency Response Data System (ERDS), and within Eight hours, notify the NRC via the ENS that both instruments are out of service.

B.

Restore one of the two channels to operable status within seven days or Prepare and submit within 14 days a written report for both instruments out of service, outlining alternate method of monitoring, cause for inoperability and plans, and schedule for restoring instruments per Specification 5.6.7.

C.

Restore one of the two channels to operable status within seven days or Be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

and Prepare and submit within 14 days a written report for both instruments out of service, outlining alternate method of monitoring, cause for inoperability, and plans and schedule for restoring instruments per Specification 5.6.7.

D.

Notify the Supervisor Nuclear Emergency Planning that NO parameters for Emergency Classification of containment High radiation are available, and within eight hours, notify the NRC via the ENS that both instruments are out of service.

Question Data B

Restore one of the two channels to operable status within 7 days or Prepare and submit within 14 days a written report for both instruments out of service outlining alternate method of monitoring, cause for inoperability and plans and schedule for restoring instruments per Specification 5.6.7.

ExplanationlJustification:

A.

B.

C.

D.

This is not required and does not satisfy TS. The instruments do supply an input to ERDS and based on this the candidate may choose this answer.

correct answer, Per TS 3.3.3.1 Condition C if both instruments are inoperable then 7 days are allowed for restoring one instrument. If that is not met Condition D is entered which for this function results in entry into 5.6.7 and a second report must be submitted.

If the candidate misreads TS for any other function with both instruments out, this is the required action.

This is not required and does not satisfy TS. The output of the detectors is used for emergency classification. The candidate may choose the answer based on E Plan training.

Sys#

System None Generic Category Emergency Procedures and Plan KA Statement Ability to identify postaccident instrumentation.

LOC 20 As Given H:\\ExamBank\\MergeDocs\\LOC2ONRCForm.doc Printed on 07/28/04

SSES LOC 20 NRC Exam 2-W M 2.4.3 WA Importance 3.513.8 Exam Level SRO (RO/SRO)

References provided to Candidate TS Technical

References:

TS 3.3.3.1 Question Source:

New Susquehanna, 8/4/2004 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

41.6 I 45.4 Objective:

3882 Identify the impact of Area Radiation Monitoring Task:

79.0P.O CRM Operation from CR panel System operability on Technical Specifications.

06 IC693 (2C693)

LOC 20 As Given HAExarnBan k\\MergeDocs\\LOC2ONRCForrn.doc Printed on 07/28/04