ML051780437

CGS-05-2005 - Final - Written Exam

ML051780437
Person / Time
Site:	Columbia
Issue date:	06/27/2005
From:	Gody A T Operations Branch IV
To:	Parrish J V Energy Northwest
References
50-397/05-301, CGS-05-2005 50-397/05-301
Download: ML051780437 (175)

Text

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 1

EXAM KEY 06/2005 Page 1 of 75

EX05023 The plant was operating at 99% power when a loss of battery charger C1-1 occurred. Battery voltage is 92VDC and going down.

Which of the following is correct concerning this condition?

A. RHR-P-2A and LPCS-P-1 will not be available during a LOCA.

B. RHR-P-2B and RHR-P-2C will not be available during a LOCA.

C. TG-EOP-1 Main Turbine Emergency O il Pump will not be available if needed.

D. RFT-EOP-1A RFP Turbine Emergency O il Pump will not be available if needed.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 295004 AA1.02 - Ab ility to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF DC POWER: System necessary to assure safe plant shutdown IMP 3.8

REFERENCE:

ABN-ELEC-125VDC rev. 1 page 13, SD000188 rev. 7, pages 31 & 32

SOURCE: BANK QUESTION - MODIFIED -

T1, GP1 LO: 5262, 5263

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Charger C1-1 feeds the div. 1 battery, B1-1. This is a 125 VDC battery. C and D are incorrect because they are 250VDC loads. B is incorrect because

it is a div 2, 125VDC load. A is correct as the loss of div 1, 125VDC will

cause the loss of ability to operate div 1 breakers. COMMENTS: Changed may to will in all four distracters.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 2

EXAM KEY 06/2005 Page 2 of 75

EX05024 A plant startup is in progress. Reactor pressure has just been increased to 950 psig and is stable.

A leak in the CAS System has resulted in a reducti on of CAS pressure to 90 psig and going down.

A complete loss of air is imminent.

Which of the following describes the effect of this loss on the Feedwater System?

A. Feedwater heater level control valves will fail as is.

B. Feedwater heater level control valves will fail closed.

C. RFW-FCV-10A & B will fail as is.

D. RFW-FCV-10A & B will fail closed.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 295019 AK2.03 - Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Reactor feedwater IMP 3.5

REFERENCE:

SD000205 rev. 9 page 21 & SD000157 rev. 12 page 19

SOURCE: NEW QUESTION -

LO: 7605, 5400

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: On a complete loss of CAS, feedwater heater LCVs fail open. A and B are incorrect. RFW-FCV-10A & 10B fail as is. D is incorrect. C is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 3

EXAM KEY 06/2005 Page 3 of 75

EX02025 How does boron injection affect reactor power during ATWS conditions?

A. The magnitude of power oscillations dur ing ATWS conditions is reduced by the initiation of the SLC System as the bor on concentration in the core increases.

B. The magnitude of power oscillations dur ing ATWS conditions is reduced by the initiation of the SLC System as inlet subcooling increases.

C. Core power goes down because core void fraction goes down as the concentration of boron in the core goes up.

D. Core power goes down because moderator density goes up as the concentration of boron in the core goes up.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 295037EK1.03 - Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR

POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Boron affects of

reactor power IMP 4.2

REFERENCE:

PPM 5.0.10 Flowchart Training Manual rev 8, pages 187 & 188

SOURCE: NEW QUESTION -

T1, GP1

LO: 8086

RATING: L4

ATTACHMENT: NONE

JUSTIFICATION: B is incorrect because an increase in core inlet subcooling causes increases in the occurrences of core power oscillations. C is incorrect because a

decrease in void fraction causes power to increase. D is incorrect because

an increase in moderator density causes a power increase. A is correct as stated in the reference. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 4

EXAM KEY 06/2005 Page 4 of 75

EX02026 Which of the following lists EOP actions that mi tigate off-site doses due to release of reactor coolant in the secondary containment?

A. Isolate primary systems leaki ng into secondary containment.

Operate available Reactor Building Ventilation.

Isolate Standby Gas Treatment System.

B. Isolate primary systems leaki ng into secondary containment Shut down the reactor.

Operate available Reactor Building Ventilation.

C. Isolate Standby Gas Treatment System.

Shut down the reactor.

Emergency depressurize the reactor.

D. Isolate primary systems leaki ng into secondary containment.

Shut down the reactor.

Emergency depressurize the reactor.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 295038 EK3.02 - - Knowledge of the reasons for the following responses as they apply to HIGH OFF SITE RELEASE RATE: System isolations IMP 3.9

REFERENCE:

PPM 5.0.10 rev. 8 pages 312 - 314

SOURCE: BANK QUESTION LR00884 - Slightly modified for clarity -

T1, GP1 LO: 8460

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: PPM 5.0.10 specifically states the actions in D as methods to mitigate rad release. Neither operating Reactor Building Ventilation nor isolating SGT reduces release of radioactivity fr om the secondary containment. D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 5

EXAM KEY 06/2005 Page 5 of 75

EX05027 The plant was operating at 99% power when a scram occurred due to a LOCA concurrent with a lockout on the Startup Transformer. Drywell te mperature is going up rapidly. The CRS has directed you to "Maintain drywell temperature bel ow 135°F with available drywell cooling. Upon investigation, you discover all drywell cooling fans have stopped.

Which of the following conditions caused these fans to trip?

A. Reactor level -68 inches.

B. Drywell pressure +2.05 psig.

C. RB Exhaust Plenum 19 mr/hr.

D. Momentary loss of voltage on their respective buses.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 295012 AK2.02 - Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Drywell cooling IMP 3.6

REFERENCE:

EWD 23E-001 (typical)

SOURCE: NEW QUESTION -

T1, GP2 LO: 5639 H4 RATING:

ATTACHMENT: NONE

JUSTIFICATION: The FAZ signals do not trip the drywell cooling fans. Due to the arrangement of the control switch contac ts, the fan stops on loss of voltage and will not restart until the CS is taken to the START position. D is correct. COMMENTS: Changed signals to conditions in stem.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 6

EXAM KEY 06/2005 Page 6 of 75

EX05028 The plant was operating at 100% power when an ATWS occurred coincident with a loss of MC-7B.

Suppression pool temperature is now 115°F and going up.

Which of the following is the current configuration of the SLC System?

A. SLC-P-1A on, SLC-V-1A (suction) open, SLC-V-1B (suction) closed, SLC-V-4A (squib valve) actuated, and SLC-V-4B not actuated.

B. SLC-P-1B on, SLC-V-1A (suction) cl osed, SLC-V-1B (suction) open, SLC-V-4A not actuated, and SLC-V-4B (squib valve) actuated.

C. SLC-P-1A on, SLC-V-1A /1B (sucti on) open, and SLC-V-4A/4B (squib valves) actuated.

D. SLC-P-1B on, SLC-V-1A/1B (sucti on) open, and SLC-V-4A/4B (squib valves) actuated.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 211000 K6.03 - Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY LIQUID CONTROL SYSTEM: AC power IMP 3.2

REFERENCE:

SD000172 SLC rev. 10, page 19

SOURCE: BANK QUESTION LX00530 - Slightly modified -

T2, GP1 LO: 5931

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: MC-7B powers the Div 1 equi pment, SLC-P-1A, V-1A and V4A. MC-8B powers Div2 equipment, SLC-P-1B, V-1B and V4B. The loss of MC-7B only

allows Div 2 equipment to operate. B is correct. COMMENTS: Changed rating to H3, revised distracters A and B to include other SLC valves COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 7

EXAM KEY 06/2005 Page 7 of 75

EX05029 APRM A, C, and E just failed downscale.

Which of the following caused these indications?

A loss of-

A. MC-8A B. 125 VDC division 1 C. 24 VDC distribution panel DP-SO-A D. Reactor protection system (RPS) bus A

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 215005 K2.02 - Knowledge of the electrical power supplies to the following: APRM channels IMP 2.6

REFERENCE:

SD000149 ARPM rev. 10 page 31

SOURCE: BANK QUESTION - Slightly modified LO00391 -

T2, GP1 LO: 5095

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: The power supply for APRM A, C, & E is RPS A as stated n the systems text. D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 8

EXAM KEY 06/2005 Page 8 of 75

EX05030 The plant was operating at 89% power when a trans ient occurred. The CRS has directed the CRO to open the 7 ADS SRVs by arming and depressing the A and C Logic Channel pushbuttons.

When the CRO pushes the pushbuttons, the 7 ADS SRVs open immediately. All 7 ADS SRVs close immediately upon release of the pushbuttons by the CRO.

Which of the following is correct concerning these conditions?

A. The Division 1 Inhibit switch is in the INHIBIT position.

B. The Division 2 Inhibit switch is in the INHIBIT position.

C. RHR-P-2A is not running.

D. RHR-P-2C is not running.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 218000 K5.01 - Knowledge of the operational implications of the following as they apply to ADS SYSTEM: ADS logic operation IMP 3.8

REFERENCE:

SD000186 ADS rev. 10 page 4

SOURCE: BANK QUESTION - LO01235 - 2001 NRC Exam -

T2, GP1 LO: 5073

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: With all ADS logic made up (all auto contacts made up and the 105 second timer timed out) and the INHIBIT Switc hes in inhibit, there is no auto initiation. If all ADS logic is made up and the Arm and Depress logic

pushbuttons are pushed with t he INHIBIT Switches in inhibit, the valves open. When the pushbutton is released, the valves close. A is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 9

EXAM KEY 06/2005 Page 9 of 75

EX05031 Which of the following functions would be affected by a trip of BKR S-1 when the plant is in Mode 4, with TR-B and DG-1 out of service?

A. High Pressure Core Spray B. RHR-C LPCI Injection C. RHR-A Suppression Pool Spray D. RHR-B Head Spray

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 230000 K2.02 - Knowledge of the electrical power supplies to the following: Pumps IMP 2.8

REFERENCE:

SD000198 RHR rev. 11, pages 5 and 49

SOURCE: NEW QUESTION -

T2, GP2 LO: 5058

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: BKR S-1 fees SM-7 via SM-1.

Suppression pool spray could be performed by RHR-P-2A and could be affected by the loss of S-1. C is correct. High

and RHR-C pumps are both powered from a different bus. Head spray is on RHR loop B and would not be affected. COMMENTS: Changed could to would in stem. Added RHR specific loops to C and D.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 10

EXAM KEY 06/2005 Page 10 of 75

EX05101 Columbia Generating Station is in a refueli ng outage. A full core offload is in progress.

Select the statement below that describes a condition that would prevent moving the refueling bridge over the core.

A. Placing the reactor mode switch in START/HOT STBY with the main hoist grapple is loaded.

B. A single control rod is withdrawn and the ma in hoist grapple is partially lowered.

C. Placing the reactor mode switch in ST ART/HOT STBY with a control rod selected.

D. Selecting a single control rod while a fuel bundle is loaded on the main hoist grapple.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 234000 K4.01 - Knowledge of FUEL HANDLING EQUIPMENT design features and/or interlocks which provide for the following: Prevention of core alterations during control rod movement IMP 3.3

REFERENCE:

SD0000207

SOURCE: BANK QUESTION - Modified LO00933 -

T2, GP2 LO: 5359

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: A is incorrect because the main hoist being loaded is not a prerequisite for the interlock of the mode switch in the START/HOT STBY position. B and D are incorrect because by themselves, would not prevent bridge movement

over the core. C is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 11

EXAM KEY 06/2005 Page 11 of 75

EX05102 The plant is operating at full power when a leak in the CAS system develops. In response to lowering CAS system pressure, CAS-C-1A starts.

Which of the following is correct?

A. If started from STANDBY, CAS-C-1A will run for 1800 seconds and then stop.

B. CAS-C-1A unloaded light will illuminate for a maximum of 1800 seconds.

C. CAS-C-1A will run for 1800 seconds after CAS pressure returns to normal.

D. CAS-C-1A will run until CAS pressure returns to normal and then stop.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 2.1.24 Ability to obtain and interpret station electrical and mechanical drawings. IMP 2.8

REFERENCE:

EWD-78E-001

SOURCE: NEW QUESTION -

T3 LO: 4047

RATING: H3

ATTACHMENT: EWD-78E-001

JUSTIFICATION: In standby, CAS-C-1A will run fo r 30 minutes after the auto start on low pressure clears. C is correct. A is incorrect because the compressor runs

until the low pressure clears plus 1800 seconds. B is incorrect because the

unload light will be illuminated for as long as CAS-PS-REG/1A is picked up.

D is incorrect because CAS-C-1A runs for 1800 seconds after the pressure

returns to normal. COMMENTS: Changed 10 minutes to 30 minut es in justification statement COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 12

EXAM KEY 06/2005 Page 12 of 75

EX05034 In accordance with the precauti ons and limitations associated wit h SOP-RHR-SDC, which of the following is an acceptable method of providing adequat e core flow and also states the reason for assuring adequate core flow?

A. RPV level GT 60"; to promote natural ci rculation and limit thermal stratification.

B. One Shutdown Cooling Loop in service with RPV level GT +60"; to ensure a minimum RPV piping temperature of GT 70°F.

C. One RRC Pump in operation; to ens ure RPV Head temperature remains GT 80°F. D. One RRC Pump and One SDC Loop in operation; to ensure RPV metal temperature remains LT 200°F.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 2.1.32 Ability to explain and apply system limits and precautions IMP 3.4

REFERENCE:

SOP-RHR-SDC P&L 4.20 and 4.21 Pages 7 and 8

SOURCE: NEW QUESTION -

T3 LO: 6486

RATING: H2

ATTACHMENT: NONE

JUSTIFICATION: As stated in the P&Ls , the corre ct answer is RPV level GT 60" to promote natural circulation and limit therma l stratification. A is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 13

EXAM KEY 06/2005 Page 13 of 75

EX05035 A valve in a high-high rad area has to be closed to prevent uncovering the core.

Which of the following is the maximum administrative TEDE limit for an individual to complete this task?

A. 5 rem TEDE B. 10 rem TEDE C. 15 rem TEDE D. 20 rem TEDE

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 2.3.2 Knowledge of the facility ALARA program IMP 2.5

REFERENCE:

GEN-RPP-07 rev. 3, page 8

SOURCE: BANK QUESTION - LO00351 -

T3 LO: 6016

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: As stated in the procedure, t he maximum administrative dose for equipment or health and safety of the public is 10 rem TEDE. B is correct. COMMENTS: Changed to a BANK QUESTION - not modified.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 14

EXAM KEY 06/2005 Page 14 of 75

EX02036 The reactor was operating at 98% power when a scram occurred. The CRO noted the following conditions:

Reactor pressure 935 psig, down fast Reactor level + 3 inches, down slow Bypass Valves 100% open Generator load 572 MWe, down fast Turbine throttle valves Closed Governor valves Closed

Which of the following is correct concerning subsequent Bypass Valves response?

The Bypass Valves remain full open...

A. until feedwater returns reactor level to GT +13 inches.

B. until intercept valves close.

C. until generator load is less than 25%

with a 3-5 second time delay.

D. then close after a 20 second time delay following the turbine trip.

ANSWER: C QUESTION TYPE: RO KA # & KA VALUE: 295005AA2.03 - Ab ility to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Turbine valve position IMP 3.1

REFERENCE:

SD000129 Main Turbine rev. 9, page 20

SOURCE: BANK QUESTION - 99 NRC Exam - ex99019 - RO T1, G1 LO: 5562 RATING: H3 ATTACHMENT: NONE JUSTIFICATION: C is correct because the DEH opens the BPVs following a scram until main generator load is less than 25% with a 3-5 second time delay. COMMENTS: Added until to B.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 15

EXAM KEY 06/2005 Page 15 of 75

EX02037 Which of the following is correct concerning t he basis for the High Drywell Pressure scram?

A. The allowable value is selected to ensure that, for transients involving MSIV isolations, energy discharged to the cont ainment is at its lowest level.

B. The allowable value is selected to ens ure that, for transients involving LOCAs initiation of low pressure ECCS at RPV level 1 will not be required.

C. Minimizes the possibility of exc eeding ASME code stresses on the primary containment during transients.

D. Minimizes the possibility of fuel dam age and reduces the amount of energy being added to the coolant and containment.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 295024 EK3.06 - Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Reactor Scram IMP 4.0

REFERENCE:

TS Bases for 3.3.1.1 RPS Instrumentation

SOURCE: NEW QUESTION -

T1, GP1 LO: 5949

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: D is correct as stated in the TS basis for the LCO. A, B, and C are all incorrect combinations of other bases for RPS instrumentation. COMMENTS: Changed rating to H3 from L3.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 16

EXAM KEY 06/2005 Page 16 of 75

EX05038 The plant was scrammed and EOPs were entered due to increasing temperature from a leak in the drywell.

To which of the following panels should CRO 3 respond first?

A. P840, Bd. A B. P800, Bd. C C. P602 D. P601 ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 295028 2.4.13 - Knowledge of crew roles and responsib ilities during EOP flowchart use IMP 3.3

REFERENCE:

PPM 1.3.1 rev. 67, page 35

SOURCE: NEW QUESTION -

T1, GP1 LO: 6088

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: PPM directs that CRO 3 be dire cted to P601 during the initial stages of a casualty. D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 17

EXAM KEY 06/2005 Page 17 of 75

EX05103 Columbia Generating Station is starting up followi ng a refueling outage. Reactor power is currently 45%. CRO1 notes that reactor power is slowly going up. Scanning his panel he notices that RRC-P-1A speed is slowly rising.

Which of the following is correct for this condition?

A. Immediately SCRAM the Reactor and then stop RRC-P-1A by opening E-CB-RRA or by depressing the STOP pushbutton.

B. Attempt to stop the speed increase of RRC-P-1A, and if unsuccessful, stop RRC-P-1A by opening E-CB-RRA or by depressing the STOP pushbutton.

C. Immediately stop RRC-P-1A by opening E-CB-RRA or by depressing the STOP pushbutton.

D. Attempt to stop the speed increase of RRC-P-1A, and if unsuccessful, SCRAM the Reactor.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 295014 AA1.02 - Ab ility to operate and or monitor the following as they apply to INADVERTENT REACTIVITY ADDIT ION: Recirculation Flow Control system IMP 3.6

REFERENCE:

ABN-POWER rev. 4, page 3

SOURCE: NEW QUESTION -

T1, GP2 LO: 6747

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: The immediate actions for a RRC Flow Control System Failure if speed is increasing for one pumps is to attempt to control the speed and if speed

cannot be controlled, stop the affected pum

p. The procedure also states that the preferred method for stopping the pum p is to depress the STOP P/B or open E-CB-RRA. B is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 18

EXAM KEY 06/2005 Page 18 of 75

EX05040 The plant was operating at 100% power when a transi ent occurred. Available drywell sprays were put into service per PPM 5.2.1, Primary Containment Control, due to high containment hydrogen and oxygen levels.

After a period of time of spraying the wetwell and the drywell, the following conditions now exist:

Reactor level is -186 inches and stable Reactor pressure is 75 psig and down slow

Drywell pressure is 1.6 psig and steady

Drywell temperature is 155°F

Drywell and Wetwell hydrogen is 7%

Drywell and Wetwell oxygen is 0.4%

Suppression pool temperature is 101°F

SM-7 has a lockout due to overcurrent relay actuation

Which of the following is the appropria te use of the identified RHR system?

Utilize-..

A. RHR-P-2B for RPV injection due to low RPV level.

B. RHR-P-2A to cool the suppression pool due to high suppression pool temperature.

C. RHR-P-2B to spray the drywell due to high containment hydrogen and oxygen concentrations.

D. RHR-P-2A to spray the drywell due to high drywell temperature.

ANSWER: A QUESTION TYPE: RO KA # & KA VALUE: 500000 EA1.06 - Ab ility to operate and or monitor the following as they apply to HIGH CONTAINMENT HYDROGEN CONTROL: Drywell sprays IMP 3.3

REFERENCE:

PPM 5.2.1, PPM 5.1.1

SOURCE: NEW QUESTION -

T1, GP2 LO:

RATING: H3 ATTACHMENT: NONE JUSTIFICATION: Reactor level at -186 inches r equires that all available ECCS be used for RPV injection. With SM-7 OOS, that leaves only Div. 2 ECCS. B and D are incorrect because they are Div 1 and C is incorrect because the 2B pump is

required for RPV injection. A is correct. COMMENTS: Rewrote per NRC comments.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 19

EXAM KEY 06/2005 Page 19 of 75

EX05041 The plant was operating at 99% power when a LO CA occurred. The following conditions exist:

Reactor level

Reactor pressure

Suppression pool level

Wetwell pressure

-280 inches and up slow

14 psig

+5 inches and stable

12 psig All ECCS pumps are running on minimum flow.

Which of the following operator actions will cause suppression pool level to go down and stabilize at a substantially lower level?

Start-A. RPV Head spray with RHR-P-2A B. Drywell spray with RHR-P-2B C. RPV injection with RHR-P-2C D. Wetwell spray with RHR-P-2B

ANSWER: C QUESTION TYPE: RO KA # & KA VALUE: 203000 A1.05 - Ab ility to predict and/or moni tor changes in parameters associated with operating the RHR/LP CI: INJECTION MODE controls including: Suppression pool level IMP 3.8

REFERENCE:

SD000198 rev. 11, fig. 1

SOURCE: NEW QUESTION -

T2, GP1 LO: 5774 RATING: H2 ATTACHMENT: NONE JUSTIFICATION: A is incorrect because the head spray line is on RHR-B. B and D are incorrect because these flow paths take suction from the wetwell and put the water right back into the suppression pool. C is correct because the pump

takes suction from the SP and refills the core shroud back up to the -210

inch level (2/3 core height), which reduces the level in the SP until water

spills out of the core shroud and back into the SP. COMMENTS: Changed reactor pressure from 10 to 14 psig and -125 to -280 in stem COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 20

EXAM KEY 06/2005 Page 20 of 75

EX05042 A transient occurred that has caused reactor level to be -140 inches for the last 115 seconds.

No SRVs are open at this point of the transient.

Concurrently with this transient, a loss of SM-8 occurred due to an overcurrent. Reactor pressure is currently 485 psig. All other plant equipment operated as expected.

LPCS-V

A. opens when ADS auto initiates and reduces reactor pressure to less than the setpoint.

B. opens after the ADS DIV 1 manual initiation pushbutton is pushed and reactor pressure is reduced to less than the setpoint.

C. opens after the ADS DIV 2 manual initiation pushbutton is pushed and reactor pressure is reduced to less than the setpoint.

D. cannot be opened because ADS will not reduc e reactor pressure with a timer failure.

ANSWER: B QUESTION TYPE: RO KA # & KA VALUE: 209001 K6.11 - Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM: ADS IMP 3.6

REFERENCE:

SD000192 LPCS rev. 10, page 5 & SD000186 ADS rev. 10, page 4&5

SOURCE: NEW QUESTION -

T2, GP1 LO: 5484b, 5071

RATING: H3 ATTACHMENT: NONE JUSTIFICATION: From the conditions given, Div 2 ECCS pumps are not running and Div 2 ADS logic is not made up. Div 1 ADS will not reduce pressure without manual action because the 105s timer should be timed out, but no valves

are open. A and C are incorrect. Since reactor pressure is 485 psig, above

the valve permissive at 470 psig, the valve is not currently open. D is

incorrect. B is correct because the only way the valves will open under

these conditions is to push the Div 1 pushbutton. The valves stay open

while the button is depressed. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 21

EXAM KEY 06/2005 Page 21 of 75

EX05043 A plant startup is in progress with reactor power on range 6 of the IRMs. Annunciator 603.A7 drop 3-5 IRM MONITORS UPSCALE then illuminates. No RPS actuations have occurred.

Which of the following indications caused this alarm?

IRM-A at-

A. 100/125 of scale.

B. 105/125 of scale.

C. 115/125 of scale.

D. 120/125 of scale.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 215003 A3.02 - Ab ility to monitor automatic oper ations of the IRM SYSTEM including: Annunciation and alarm signals IMP 3.3

REFERENCE:

PPM 4.603.A7.3-5 rev. 30

SOURCE: NEW QUESTION -

T2, GP1 LO: 5459

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: As per the references, the rod block is at 108/125 of scale. Neither A nor B would cause a rod block. And since the stem states that there have been no RPS actuations, D is incorrect. C is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 22

EXAM KEY 06/2005 Page 22 of 75

EX05044 Which of the following concerning operation of the Safety Relief Valves is correct?

A. If gas pressure from the CIA is los t, the seven ADS SRVs have an additional 42-gallon accumulator to allow for one manual actuation against maximum drywell

pressure with 0 psig reactor pressure.

B. If gas pressure from the CIA is los t, all eighteen SRVs have an additional 10-gallon accumulator to allow for one manual actuation against maximum drywell

pressure with 0 psig reactor pressure.

C. All eighteen SRVs can be opened from panel H13-P601 by operation of the associated control switch which energizes ei ther the "A" or the "B" solenoid pilot valve. D. Position indication for MS-RV-3D, 5B and 5C at the Alternate Remote Shutdown Panel is driven by Linear Variable Diffe rential Transducers (LVDT) mounted on the stem.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 239002 K4.09 - Knowledge of RELIEF/SAFETY VAL VES design features and/or interlocks which provide for t he following: Manual opening of the SRV IMP 3.7

REFERENCE:

LO000128 Page 4-9

SOURCE: NEW QUESTION -

T2, GP1 LO: 5528

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: The 10 gallon tank is sized for one SRV actuation with RPV/P at 1000 psig and normal DW pressure. H13-P601 operates the "C" solenoid pilot valves.

Position indication at the ARSP is dependent on control switch position not

LVDT. A is correct as per Systems Text. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 23

EXAM KEY 06/2005 Page 23 of 75

EX05104 At the beginning of the fuel shuffle sequence, S ource Range Monitor count rate readings are recorded.

Which of the following indications would require stopping of the fuel bundle insertion, close observation of subcritical multiplication behavior while slowly lowering the bundle, and require concurrence of the Reactivity Manager to resume the fuel shuffle?

A. If any SRM count rate exceeds 350 counts/second.

B. If an unexpected two doublings of the average SRM count rate occurs.

C. If a doubling of any single SRM count rate occurs.

D. If an unexpected two doublings of any single SRM count rate occurs.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 259002 2.2.27 - Knowledge of the refueling process IMP 2.6

REFERENCE:

PPM 6.3.2 Rev. 17 page 18 of 32

SOURCE: NEW QUESTION -

T2, GP1 LO: 7700

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: Per PPM 6.3.2, if an unexpect ed doubling in the average SRM count rate occurs, or two doublings of any singl e SRM count rate occurs, then bundle insertion is stopped and slowly completed. It takes the Reactivity Managers

concurrence to resume the shuffle. D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 24

EXAM KEY 06/2005 Page 24 of 75

EX05046 The plant is operating at 100% power when a loss of MC-7A occurs.

What effect will this have on the operation of IN-1?

A. The inverter is powered from the normal DC power supply, S1-2.

B. The inverter is powered from the normal DC power supply, S2-1.

C. The Static Switch auto transfers to the backup AC supply, MC-7B.

D. The Static Switch auto transfers to the backup AC supply, MC-8A.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 262001 K3.04 - Knowledge of the effect that a loss or malfunction of the AC ELECTRICAL DISTRIBUTION will have on the following: UPS IMP 3.1

REFERENCE:

SD000194 rev. 9, page 4

SOURCE: BANK QUESTION - 99 NRC EXAM Ex99060 -

T2, GP1 LO: 5896

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: The normal AC source for IN-1 is MC-7A. The parallel source is fed from S2-1, 250 VDC. When the primary AC source is lost, the auctioneering

diode allows 250 VDC from S2-1 to power the inverter. B is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 25

EXAM KEY 06/2005 Page 25 of 75

EX05047 A fault has caused the complete loss of DC Bus S1-1.

Which of the effects would this loss have on plant equipment?

Loss of control power to....

A. Bkr 3-8 B. Bkr 1-7 C. Bkr 8-3 D. Bkr 7-1

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 263000 A4.01 - Ab ility to manually operate and/or m onitor in the control room: Major breakers and control power fuses IMP 3.3

REFERENCE:

SD000188 rev. 7, pages 24 & 25

SOURCE: BANK QUESTION - slightly modified 99 NRC Exam ex99084 -

T2, GP1 LO: 5065

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: Bus S1-1 powers division 1 equi pment/breaker indication. Only D is a Division 1 load and is correct. COMMENTS: Changed rating to L2 from H2.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 26

EXAM KEY 06/2005 Page 26 of 75

EX05048 The plant is operating at 100% power. You hav e just closed RCC-V-6 with the control switch.

Which of the following loads will NOT experience a loss of RCC flow?

A. EDR-HX-2 RB Equipment Drain Heat Exchanger B. FPC-HX-1A Fuel Pool Cooling Heat Exchangers C. RWCU-P-1A Motor Coolers D. RRC-P-1A Reactor Recirculation Pump

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 400000 A1.01 - Ab ility to predict and/or moni tor changes in parameters associated with operating the CCWS cont rols including: CCW flow rate IMP 2.8

REFERENCE:

SD000196 rev. 10, page 18

SOURCE: NEW QUESTION -

T2, GP1 LO: 7668

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: RCC-V-6 isolates all equipment outside of the drywe ll from the RCC system and causes all flow to be diverted into the drywell. The only load listed in the

drywell Is D. It is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 27

EXAM KEY 06/2005 Page 27 of 75

EX05049 Which of the following RWCU component s/design features is utilized to maximize plant efficiency?

A. CRD purge flow B. RWCU pump cooling heat exchangers C. Regenerative heat exchangers D. Non-regenerative heat exchangers

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 204000 K4.06 - Knowledge of RWCU design features and/or interlocks which provide for the following: Maximize plant efficiency IMP 2.6

REFERENCE:

SD000190 rev. 11, page 5

SOURCE: NEW QUESTION -

T2, GP2 LO: 5034d

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: Specifically, the Regen HXs are designed to increase overall plant efficiency by using return water to cool reactor wa ter prior to the demineralizers. C is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 28

EXAM KEY 06/2005 Page 28 of 75

EX05050 To which of the following systems does the Rod Wort h Minimizer input rod blocks for application of control rod movement restrictions?

A. RMCS, Reactor Manual Control System B. RPIS, Rod Position Information System C. RSCS, Rod Sequence Control System D. RBM, Rod Block Monitor System

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 201002 K1.05 - Knowledge of the physical connections and/or cause-effect relationships between RMCS and the following: RWM IMP 3.4

REFERENCE:

SD000154 RWM, rev. 11, page 5

SOURCE: NEW QUESTION -

T2, GP2 LO: 5916

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: The RWM inputs control rod bl ocks to the RDCS/RMCS, which imposes the rod movement restrictions. A is correct. The RWM does not interface with

RSCS and RBM. The RPIS inputs to the RWM, not from it. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 29

EXAM KEY 06/2005 Page 29 of 75

EX05113 Columbia is starting up following a refueling out age. Reactor power is currently 75%. CRO1 is raising power with recirculation flow at a rate of 10 Mwe per minute. During the past few minutes, the Reactor Operator notes that R PV level is slowly trending downward.

Which of the following would explain this unexpected drop in RPV level?

A. The Master Level M/A Station, RFW-LIC-600, has failed as is.

B. RFW-P-1A Speed Controller, RFW-SC-1A, failed to the MDEM Mode.

C. The Startup Valve M/A Station, RFW-LIC-620, has failed as is.

D. RFW-P-1B Speed Controller, RFW-SC

-1B, has failed to the MDVP Mode.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 259001K1.08 - Knowledge of the physical and/or cause-effect relationship between REACTOR FEEDWATER SYSTEM and the following: Reactor water level control IMP 3.6

REFERENCE:

SD000157 Pg 7 and 17

SOURCE: NEW QUESTION -

T2, GP2 LO: 5394 and 5400

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: The stem indicates two RFW pum ps are in operation. If B or D were to occur, the opposite RFW pump would co mpensate for the malfunction thus B and D are incorrect. C is incorrect because at this power level the startup

controller is not in service. The Master controller is in service and if it fails as

is, as power is raised, RPV level drops. A is the correct answer. COMMENTS: Question replaced after un-secure email sent indicating KA used for this question.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 30

EXAM KEY 06/2005 Page 30 of 75

EX05052 The plant is operating at 100% power. The normal c ooling water supply for Fuel Pool Cooling has been lost and is not available.

Under these conditions, which of the following syst ems can be operated from the control room for fuel pool temperature control?

A. RHR Fuel Pool Cooling Assist B. Reactor Closed Cooling Water C. Standby Service Water D. Plant Service Water

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 233000 A4.05 - Ab ility to manually operate and/or m onitor in the control room: Pool temperature IMP 2.5

REFERENCE:

SD000202 FPC rev. 11 page 15

SOURCE: NEW QUESTION -

T2, GP2 LO: 8931

RATING: H2

ATTACHMENT: NONE

JUSTIFICATION: RCC is the normal cooling water and is unavailable. TSW cools RCC, which is unavailable, which makes both B and D incorrect. RHR fuel pool assist

needs a spool piece installed in the plant and is incorrect. SW is the backup

system operated from the control room. C is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 31

EXAM KEY 06/2005 Page 31 of 75

EX05053 The plant is operating at 85% power when several alarms are received on back panels. Upon investigation, you note that the Reactor Bu ilding exhaust fan tripped off and Reactor Building pressure increased to approximately 4 inches of H 2 O. You then noted that Reactor Building pressure drops and stabilized at 0 inches H 2 O.

Which of the following explains these indications?

A. ROA-AD-5 Alternate Relief Damper opened to reduce pressure to 0 inches H 2 O. B. ROA-V-1 & 2 and REA-V-1 & 2 closed for pressure control.

C. The Reactor Building supply fan tripped due to high pressure.

D. The standby Reactor Building exhaust fans started due to high pressure.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 290001 K4.02 - Knowledge of SECONDARY CONTAINMENT design features and/or interlocks which provide for t he following: Protection against over pressurization IMP 2.8

REFERENCE:

SD000183 RB HVAC rev. 9, pages 8 & 16

SOURCE: NEW QUESTION -

T2, GP2 LO: 5680, 5681, 5677b

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Both REA and ROA fans trip at 4 in ches of water for pressure control. C is correct. A is incorrect because ROA-AD-5 will close by 3.5 inches of water.

B is incorrect because these valves do not close on high pressure in the

reactor building. D is incorrect becaus e the standby fan auto starts only if the running fan remains on. COMMENTS: Changed rating to H3 from L3. Revised stem COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 32

EXAM KEY 06/2005 Page 32 of 75

EX05054 The plant was operating at 98% power when a Stati on Blackout occurred. The following conditions exist:

4 control rods failed to insert fully Reactor Level -172 inches

Which of the following actions is required by Tech Specs?

A. Initiate action within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to rest ore level to greater than -129 inches.

B. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, restore reactor level to greater than + 13 inches and insert all insertable control rods.

C. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, restore reactor leve l greater than -161 inches and insert all insertable control rods.

D. Initiate action within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to rest ore level to greater than -129 inches.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 2.2.22 Knowledge of limiting conditions for operations and safety limits IMP 3.4

REFERENCE:

TS 2.1

SOURCE: BANK QUESTION - EX00005 -

T3 LO: 6934

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Reactor level at -172 inches is a sa fety limit violation. C is the correct action reactor water level. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 33

EXAM KEY 06/2005 Page 33 of 75

EX05116 FIXED TYPO IN REFERENCE Which of the following statements is correct c oncerning a loss of Control Room Annunciation?

A. If a total loss of Control Room annunc iation exists and has been verified, immediately insert a manual scram.

B. If a total loss of Control Room annunciation exists and has been verified, immediately commence a controlled shutdown.

C. If annunciation is lost to H13-P602 and H 13-P603 coincident with a control rod drift, perform the immediate actions associated with ABN-ROD.

D. If a loss of Control Room annunciation occurs on H13-P800, H13-P820 and H13-P840 coincident with the performance of a RCIC surveillance, the surveillance may continue.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 2.4.32 Knowledge of operator response to loss of all annunciators IMP 3.3

REFERENCE:

AND-ABN-ANNUN

SOURCE: NEW QUESTION -

T3 LO: 5262

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: A is incorrect because ABN-ANNUN states that a reactor scram is not required as a response to a total loss of annunciation. B is incorrect because

ABN-ANNUN states that a controll ed shutdown should be made after consultations between the Shift Manager and Plant Management. C is

correct as ABN-ANNUN states to suspend operations not essential to safe

plant operation. D is incorrect because ABN-ANNUN states to suspend all

surveillance testing. COMMENTS: Re-written due to KA not matching previous question.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 34

EXAM KEY 06/2005 Page 34 of 75

EX05056 The reactor was operating at 100% power when a scram occurred. Following the scram the RO notes the UPSCL NEUT FIRST light is illuminated on P608.

Which of the following caused the scram?

A. Loss of both RPS MG Sets.

B. Loss of containment cooling.

C. Trip of both feed pumps.

D. Main turbine trip.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 295006AK2.06 - Knowledge of the interrelations between SCRAM and the following: Reactor Power IMP 4.2

REFERENCE:

SD000149 rev/ 10, pages 11 & 12

SOURCE: BANK QUESTION -

T1, GP1 LO: 5089

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: The loss of RPS, containment cooling and both feed pumps cause a scram but no large power spike. Loss of the main turbine causes a larger pressure/power spike, which results in an upscale neutron trip before the

thermal trip occurs (due to the time constant associated with the thermal trip). COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 35

EXAM KEY 06/2005 Page 35 of 75

EX05057 The control room has been evacuated to the Remo te Shutdown Panel due to a fire. RHR-P-2B was in Shutdown Cooling, but the normal discharge flowpath has been lost.

Which of the following is an acceptable fl ow path for Alternate Shutdown Cooling?

RHR-P-2B in operation discharging through-

A. RHR-V-42B LPCI Injection.

B. RHR-V-53B SDC return.

C. RHR-V-115 Containment Flooding.

D. RHR-V-23 Head Spray.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 295021 2.4.35 - Knowledge of local operator aux iliary operator tasks during emergency operations including system geography and system implications. IMP 3.3

REFERENCE:

ABN-CR-EVAC rev. 7, page 55

SOURCE: NEW QUESTION -

T1, GP1 LO: 5574i

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: ABN-CR-EVAC directs the use of RHR-V-42B if the 53B valve is unavailable for injection. A is correct. COMMENTS: Changed distracter C fr om RHR-V-24B Full Flow Test.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 36

EXAM KEY 06/2005 Page 36 of 75

EX05058 PPM 5.2.1 Primary Containment Control has been entered due to low suppression pool water level.

Which of the following systems is used to add wate r to the suppression pool from the Condensate Storage Tanks under these conditions?

A. RHR-B B. HPCS C. RHR-C D. RCIC ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 295030 EA1.06 - Ab ility to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Condensate storage and transfer (make up to the suppression pool) IMP 3.4

REFERENCE:

PPM 5.5.23 rev. 4

SOURCE: NEW QUESTION -

T1, GP1 LO: NO LO

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: PPM 5.5.23 gives direction to us e HPCS to fill the suppression pool from the CSTs. B is correct. COMMENTS: Changed A from LPCS and C from RHR.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 37

EXAM KEY 06/2005 Page 37 of 75

EX05059 The plant is operating at 100% power for the past week. Control Room logs are being taken. The operator notes that reactor pressure is three psig hi gher than it was the last time logs were taken.

Which of the following could have caused a higher reactor pressure?

A. The in-service DEH pressure contro ller "CONT A OUTPUT" signal slowly failing low. B. The in-service DEH pressure contro ller "CONT A OUTPUT" signal slowly failing high. C. A short voltage dip on US-PP which resulted in a shift to PRESS SET PT MANUAL. D. Both DEH pressure "CONT A OUTPUT

" and "CONT B OUTPUT" have failed and DEH has shifted to BPV MANUAL.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 295007AK3.06 - Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESS URE: Reactor/turbine pressure regulating system IMP 3.7

REFERENCE:

SD000146 DEH, rev. 8, page 9, 39 and dwg 4D.

SOURCE: NEW QUESTION -

T1, GP2 LO: 5286

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: As stated in the reference, t he HSS selects the controller with the highest output. The standby controller is biased with 3 psi so as the in-service

controller fails low, the HSS will transfer control to the standby controller and

reactor pressure will have gone up by the 3 psi bias. A is correct. B would

result in a lower reactor pressure and C and D do not result in a RPV

pressure change. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 38

EXAM KEY 06/2005 Page 38 of 75

EX05060 The plant was operating at 100% power when a transi ent occurred which resulted in a scram. The following indications now exist:

Narrow Range

Wide Range

Fuel Zone

Upset Range

Drywell Temperature

Reactor Pressure

0 inches

-148 inches and stable

-129 inches and stable

0 inches 165° F and up slow

480 psig and down slow Which of the following is correct concerning these indications?

The RO should report level is-

A. 0 inches B. -148 inches C. -129 inches D. not able to be determined

ANSWER: C QUESTION TYPE: RO KA # & KA VALUE: 295009 AA2.01 - Ab ility to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL: Reactor water level IMP 4.2

REFERENCE:

PPM 5.1.1 rev. 16, caution 1 and RPV Saturation Curve.

SOURCE: NEW QUESTION -

T1, GP2 LO: 8491 RATING: H3 ATTACHMENT: YES - PPM 5.1.1 rev. 16, caution 1 and RPV Saturation Curve. JUSTIFICATION: It is given in the stem that indications are stable and conditions are less than the sat curve, which indicates there are no inop instrument issues. The WR, NR, and UR are below the MUL, which makes A, B, & C incorrect. The Fuel

Zone is on scale and operable. C is correct. COMMENTS: Changed conditions to indications in stem.

Changed stem to: The RO should report RPV level is-" from "Actual RPV level is-."

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 39

EXAM KEY 06/2005 Page 39 of 75

EX05061 The plant was operating at 25% power when a trans ient caused a scram. The following conditions now exist:

Reactor level

Reactor pressure

Drywell pressure

+14 inches

129 psig 1.62 psig

Which of the following is interl ocked closed/prevented from opening?

A. RHR-V-24B Full Flow Test B. RHR-V-53B SDC Return C. RHR-V-27B Suppression Pool Spray D. RHR-V-42B LPCI Injection

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 205000 K1.01 - Knowledge of the physical connections and/or cause-effect relationships between SHUTDOWN COOLING SYSTEM and the following:

Reactor pressure IMP 3.6

REFERENCE:

SD000198 rev. 11, page 30

SOURCE: NEW QUESTION -

T2, GP1 LO: 5780

RATING: H2

ATTACHMENT: NONE

JUSTIFICATION: Of the signals listed, only t he reactor pressure signal causes an RHR isolation/interlock. This high-pressure interlock prevents the SDC section of piping from being over pressured. The LPCI piping is also protected from over pressurization, but the setpoint is 470 psig. B is the correct answer. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 40

EXAM KEY 06/2005 Page 40 of 75

EX05062 Which of the following systems does RCIC share a comm on injection line for its return to the RPV?

A. HPCS B. LPCS C. RHR-A D. RHR-B ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 217000 K1.05 - Knowledge of the physical connections and/or cause-effect relationships between RCIC SYSTEM and the following: RHR system IMP 2.6

REFERENCE:

SD000180 RCIC rev. 12, page 32 SD000198 rev. 11, figure 1G

SOURCE: NEW QUESTION -

T2, GP1 LO: 5774, 5726

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: The RHR-B system injects through the RCIC head spray line and tap is between RCIC-V-13 and RCIC-V-65. D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 41

EXAM KEY 06/2005 Page 41 of 75

EX05063 A normal plant startup is in progress with r eactor power at approximately 1000 counts on the source range. SRM-B is bypassed for maintenance.

During work on SRM-B, the mode switch for SRM-D is inadvertently placed in Standby position.

What effect does this have on the startup?

A. A 1/2 scram and an upscale trip on SRM-D.

B. A 1/2 scram and an Inop trip on SRM-D.

C. A rod out block and an upscale trip on SRM-D.

D. A rod out block and an Inop trip on SRM-D.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 215004 A1.06 - Ab ility to predict and/or moni tor changes in parameters associated with operating the SRM cont rols including: Lights and alarms IMP 3.1

REFERENCE:

SD000132 SRM rev. 10, page 26

SOURCE: NEW QUESTION -

T2, GP1 LO: 5942

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Placing the mode switch in Standby causes an Inop trip on SRM-D. A and C are incorrect. An Inop trip on an SRM causes a rod block and not a scram.

D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 42

EXAM KEY 06/2005 Page 42 of 75

EX05064 The plant is operating at 100% power when a loss of control power to RCC-P-1A occurs.

There will be an increase in-.

A. CRD pump lube oil temperature.

B. CRD pump motor temperature.

C. drywell air temperature.

D. drywell (EDR-HX-1) Equipment Drain sump temperature.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 201001 K6.06 - Knowledge of the effect that a loss or malfunction of the following will have on the CRDH SYSTEM: CCW system IMP 2.8

REFERENCE:

SD000142 CRD rev. 12, page 37

SOURCE: NEW QUESTION -

T2, GP2 LO: 5706

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: The loss of control power to RCC-P-1A closes RCC-V-6. This closure causes a loss of cooling to loads external to the containment. The loss of

cooling to CRD results in the increasi ng temperature of the lube oil for the CRD Pumps. A is correct, B is incorrect. C and D are both incorrect

because these coolers do not lose RCC flow. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 43

EXAM KEY 06/2005 Page 43 of 75

EX05106 The plant has been shutdown following a LOCA wit h significant core damage. Drywell Hydrogen concentration is now 10% and Oxygen concentration is at 7%.

In preparation for spraying the drywell, the CRS has directed the RO to stop the drywell recirculation fans per the EOPs. However, one recirc ulation fan fails to stop due to a control switch malfunction.

Which of the following may occur as a result?

A. Damage to the operating recirculation fan due to water from drywell spray.

B. Inaccurate Drywell Hydrogen indication du e to circulation from the running fan.

C. Damage to the operating recirculat ion fan due to elevated pressures in containment.

D. Hydrogen combustion in the drywell, igni ted by the operating recirculation fan.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 223001 K3.04 - Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM A ND AUXILIARIES will have on the following: Drywell hydrogen gas concentration IMP 3.3

REFERENCE:

PPM 5.0.10 rev 9 pg 290 and 298

SOURCE: NEW QUESTION -

T2, GP2 LO: 8426

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Per PPM 5.0.10, the drywell re circulation fans are secured to eliminate potential ignition sources. D is correct. COMMENTS: Rewritten per NRC comment COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 44

EXAM KEY 06/2005 Page 44 of 75

EX05066 The plant is operating at 100% power. FP-P-2A is r unning in support of hydrant testing. Fire main pressure is 140 psi. A loss of power to MC-5N o ccurs and FP-P-2A trips. Fire main pressure is reading 120 psi and trending down quickly. Shortly t here after, the FIRE MAIN PRESSURE LOW annunciator illuminates. The ARP directs the operator to check the status of the fire pumps and start them as required.

Which of the following is correct for these conditions?

A. At 120 psig, FP-P-2B Electric pump will start after a 10 second time delay. If pressure continues to drop, at 110 psig, FP-P-1 Diesel pump starts and FP-P-110 Diesel pump will start after a 30 second time delay.

B. At 100 psig, FP-P-2B Electric pump will start. FP-P-1 Diesel pump and FP-P-110 Diesel pump will start when fire main pressure drops to 90 psig.

C. At 110 psig, FP-P-2B Electric pump and FP-P-1 Diesel pump will start. If pressure continues to drop, FP-P-110 Die sel pump will start at 100 psig.

D. At 110 psig, FP-P-2B Electric pump will start after a 10 second time delay and FP-P-1 diesel pump will start after a 15 se cond time delay. If pressure drops to 100 psig, FP-P-110 Diesel pump will start after a 30 second time delay.

ANSWER: D QUESTION TYPE: RO KA # & KA VALUE: 286000A2.11 - Ab ility to predict the impacts of the following on the FIRE PROTECTION SYSTEM and based on those predictions, use procedures to

correct, control or mitigate the consequences of those abnormal conditions or operations: Pump trips IMP 3.1

REFERENCE:

PPM 2.8.7 Rev. 37 page 6

SOURCE: NEW QUESTION -

T2, GP2 LO: 5377 RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: Per PPM 2.8.7 FP-P-2B and FP-P-1 start at 110 psig after a 10 and 15 sec.

TD. FP-P-110 starts at 100 psig after a 30 second time delay. D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 45

EXAM KEY 06/2005 Page 45 of 75

EX05067 The reactor is operating at 99% power, with HPCS out of service, when a loss of DP-S1-1A occurs.

Shortly thereafter, a loss of feedwat er initiates a reactor scram. Reactor level is -63" and up slow.

Which of the following describes an action taken based on these conditions?

A. Verify that IN-2 transferred to the alternate AC source.

B. Go to P628 to operate all SRVs for pressure control.

C. Go to P631 to operate the ADS SRVs for pressure control.

D. Verify that IN-1 transferred to the alternate AC source.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 2.1.30 Ability to locate and operate components / including local controls IMP 3.9

REFERENCE:

LO000128 MS rev. 8, pages 6 & 7

SOURCE: BANK QUESTION - 98 NRC Exam EX98110 -

T3 LO: 5262

RATING: H4

ATTACHMENT: NONE

JUSTIFICATION: A is incorrect because IN-2 is powered from DP-S1-2. B is incorrect because only ADS SRVs can be operated from P628, not all. D is incorrect

because IN-1 is powered from 250 VDC not 125 VDC. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 46

EXAM KEY 06/2005 Page 46 of 75

EX05068 The reactor is in MODE 5. The following conditions exist:

SRM-A SRM-B SRM-C SRM-D MODE SWITCH

Reactor Coolant Temperature

Reactor Head and Internals OOS

2.8 counts

per second (S/N 9.3/1)

2.5 counts

per second (S/N 10.7/1)

OOS REFUEL 108°F Removed It is desired to start a full core off load from the vessel at this time.

Irradiated fuel movement .....

A. may not start at this time because there are no operable SRMs.

B. may not start at this time because reac tor coolant temperature is greater than 100°F. C. may only start in either t he quadrant with the B or C SRM.

D. may start in any quadrant in the reactor vessel.

ANSWER: A QUESTION TYPE: RO KA # & KA VALUE: 2.2.30 Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling area / comm unications with fuel storage facility /

system operated from the control room in support of fueling operations/ and supporting instrumentation IMP 3.5

REFERENCE:

TS 3.3.1.2 pages 3.3.1.2-1 through 6 SOURCE: BANK QUESTION - 99 NRC Exam EX99009 -

T3 LO: 10298

RATING: H3 ATTACHMENT: NONE JUSTIFICATION: The TS requires at leas 2 operable SRMs before core alts can start. In this instance, with B and C LT 3 cps, they are not operable and no fuel movement is allowed. A is the correct answer. COMMENTS: Changed Attachments to NONE.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 47

EXAM KEY 06/2005 Page 47 of 75

EX05069 The plant was operating at 98% power when a LOCA occurred. Following the LOCA, LPCS-P-1 is secured, then LPCS PUMP DISCH PRESS HIGH/LO annunciator illuminates. Reactor pressure is 290 psig and going down.

Which of the following is correct if LPCS-P-1 is restarted under these conditions?

A. LPCS-P-1 starts but trips due to overcurrent from excessive flow.

B. LPCS-P-1 starts but does not inject at this pressure.

C. The discharge piping could break resulting in Reactor Building flooding and a reduction in suppression pool level.

D. The discharge piping could break in cont ainment resulting in a reduction in suppression pool level.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 209001K5.05 K5.05 - Knowledge of the operational implications of the following as they apply to LOW PRESSURE CORE SPRAY SYSTEM: System venting IMP 2.5

REFERENCE:

PPM 4.601.A3 drop 5-3 rev. 14

SOURCE: NEW QUESTION -

T2, GP1 LO: 7447

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: A is incorrect because the excessi ve flow will not cause an overcurrent. B is incorrect because LPCS is injecting at this pressure. D is incorrect because

a break in containment would cause suppression pool level to remain static.

C is correct as stated in the procedure. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 48

EXAM KEY 06/2005 Page 48 of 75

EX05070 The reactor was at 100% power when a loss of feedwater caused an auto start of HPCS and RCIC. HPCS-V-4 (injection valve) has been closed to stop HPCS flow. RCIC is injecting into the RPV at 600 gpm. A high suppression pool level annunciator is received. Suppression pool level is

+2 inches and increasing.

Which of the following is the reason for the above indications?

A. HPCS on minimum flow from the CS Ts causes suppression pool level to increase.

B. RCIC steam exhaust to the suppression pool causes a temperature increase and a false increasing level indication.

C. HPCS on minimum flow causes air ent rainment in the suppression pool and a false indicated level increase.

D. RCIC on minimum flow from the CSTs ca uses suppression pool level to increase.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 209002 K1.02 - Knowledge of the physical connections and/or cause-effect relationships between HIGH PRESSURE CORE SPRAY SYSTEM and the following: Suppression pool IMP 3.5

REFERENCE:

SD000174 HPCS rev. 10, page 4

SOURCE: BANK QUESTION - 98 NRC Exam EX98044 -

T2, GP1 LO: 5421

RATING: H2

ATTACHMENT: NONE

JUSTIFICATION: B is incorrect because a temper ature increase from RCIC would not give a false high level indication. C is incorrect because the level upset from

minimum flow would not cause a steadily increasing level. D is incorrect

because the min flow valve is closed when RCIC is injecting into the vessel. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 49

EXAM KEY 06/2005 Page 49 of 75

EX05107 Columbia Generating Station is oper ating at 100% power. Due to a r upture in its control oil system, RFW-P-1A trips.

Without any operator action, which of the fo llowing explains the resultant plant response?

A. At +13 inches the reactor scrams. RPV level drops to approximately -40 inches and returns to approximately +18 inches. Reactor power drops with a stable -80

second period.

B. At +13 inches the reactor scrams. At -50 inches, HPCS and RCIC start, recover RPV level and cycle between +54 inches and -50 inches. Reactor power drops

with a stable -80 second period.

C. At +31.5 inches both RRC pumps run back to 51 Hz. Reactor power decreases to approximately 78% power.

D. At +31.5 inches both RRC pumps run back to 30 Hz. Reactor power decreases to approximately 68% power.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 212000 A4.05 - Ab ility to manually operate and/or m onitor in the control room: Reactor power IMP 4.3

REFERENCE:

Simulator; SD000184 Rev. 14 page 21

SOURCE: NEW QUESTION -

T1, GP1 LO: 7670, 9683

RATING: H4

ATTACHMENT: NONE

JUSTIFICATION: A trip of one RFW pump does NOT result in RPV level dropping to the scram setpoint. RRC-P-1A and 1B runback to 30 Hz following a trip of a RFP and a

+31inch level input. Power drops to about 50% and returns to about 68%. D

is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 50

EXAM KEY 06/2005 Page 50 of 75

EX05108 Which of the following abnormal procedures direct s arming and depressing the NS4 MSIV Isolation Logic A, B, C, and D pushbuttons?

A. ABN-RPS B. ABN-CR-EVAC C. ABN-FAZ D. ABN-CAS

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 223002 2.4.11 - Knowledge of abnormal condition procedures IMP 3.4

REFERENCE:

SD000173 NS4 rev. 10, page 12; ABN-CR-EVAC Rev. 7 page 6

SOURCE: NEW QUESTION -

T2, GP1 LO: 6889

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Depressing the NS4 MSIV Isolation Logic pushbuttons is one of the immediate operator actions associ ated with ABN-CR-EVAC. B is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 51

EXAM KEY 06/2005 Page 51 of 75

EX05073 HPCS was in a normal standby lineup when annunciator HPCS SUCTION SWITCHOVER CST LEVEL LOW illuminated. CST level is 1 foot 6 inches and going down.

Which of the following is directed by the ARP?

Verify-A. HPCS-V-15 Suppression Pool Suction auto closed HPCS-V-1 CST Suction auto closed B. HPCS-V-15 Suppression Pool Suction auto opened HPCS-V-1 CST Suction auto opened C. HPCS-V-15 Suppression Pool Suction auto opened HPCS-V-1 CST Suction auto closed D. HPCS-V-15 Suppression Pool Suction auto closed HPCS-V-1 CST Suction auto opened

ANSWER: C QUESTION TYPE: RO KA # & KA VALUE: 209002A2.13 A2.13 - Ab ility to predict the impacts of the following on the HPCS and based on those predictions, use procedures to correct, control or

mitigate the consequences of those abno rmal conditions or operations: Low condensate storage tank level IMP 3.4

REFERENCE:

4.601.A1 drop 5-6 HPCS SUCTION SWITCHOVER CST LEVEL LOW SD000174 rev. 10, pages 8-10 & 19 SOURCE: NEW QUESTION -

T2, GP1 LO: 5429 RATING: H3 ATTACHMENT: NONE JUSTIFICATION: As stated in the procedure and the interlock section of the systems text, only answer C is correct. COMMENTS: Removed HPCS-V-10 and HP CS-V-11 from each distractor.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 52

EXAM KEY 06/2005 Page 52 of 75

EX05074 A transient has resulted in a LOCA. Reactor level has been -130 inches for the last 60 seconds.

All plant equipment is operating as designed.

The CRO pushes the Div. 1 Reactor Vessel Low Level/Timer Seal-In pushbutton.

Which of the following is correct concerning these conditions?

A. Only Div.2 ADS SRVs open in 105 seconds to reduce reactor pressure.

B. Only Div.2 ADS SRVs open in 45 seconds to reduce reactor pressure.

C. All ADS SRVs open in 105 seconds to reduce reactor pressure.

D. All ADS SRVs open in 45 seconds to reduce reactor pressure.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 218000A1.04 - Ab ility to predict and/or moni tor changes in parameters associated with operating the ADS controls including: Reactor Pressure IMP 4.1

REFERENCE:

SD000186 ADS rev. 10, page 4

SOURCE: NEW QUESTION -

T2, GP1 LO: 5073

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: The low level/timer seal in pushbutton interrupts the time and restarts the timing sequence to 105 seconds. In this case, however, only the Div. 1 timer

is reset - Div. 2 continues. When the Div.2 timer times out in 45 seconds, the Div. 2 solenoids energize and open all 7 ADS valves. D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 53

EXAM KEY 06/2005 Page 53 of 75

EX05075 While moving a spent fuel bundle from the co re during a refueling outage, the fuel bundle accidentally strikes the edge of the cattle chute while moving into the spent fuel pool. The refuel bridge phone talker then reports seeing bubbles st reaming to the surface from the bundle.

What actions are required to be performed immediately?

A. Stop the fuel movement and evacuate all personnel from the refuel floor.

B. Stop the fuel movement and all personnel go to the RB 606' HP control point for further assistance.

C. Continue the fuel movem ent until the bundle can be place in the correct location in the spent fuel rack.

D. Move the bundle back into the reactor cavity and lower the fuel bundle into the RPV as far as possible to maximize shielding.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 295023AA1.05 - Ab ility to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: Fuel transfer system IMP 2.8

REFERENCE:

ABN-FUEL-HAND rev. 2, page 2

SOURCE: BANK QUESTION - 99 NRC Exam EX99081 -

T1, GP1 LO: 6897

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: Answer A is the only response that matches the immediate actions of the procedure. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 54

EXAM KEY 06/2005 Page 54 of 75

EX05076 The plant was operating at 100% power, when a DE H malfunction caused reactor pressure to go up to 1050 psig.

Which of the following is correct for this condition?

Reactor power-

A. goes down, RPV level is controlled at a new higher level by feedwater level control. B. goes down, reactor scram and level is controlled by FWLC setpoint setdown.

C. goes up, feedwater level control re turns RPV level to the normal range.

D. goes up, RPV level is controlled at a new lower level by feedwater level control.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 295025 EA2.02 - Ab ility to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor power IMP 4.2

REFERENCE:

SD000161 RPS rev. 12, page 12 BWR GFES Reactor Theory, page 52

SOURCE: NEW QUESTION -

T1, GP1 LO: 7271

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: When pressure goes up, voids decrease, power goes up and level is returned to the normal band by feedwater. A and B are both incorrect

because power goes up, not down. D is incorrect because FWLC returns

RPV level to the normal range and not a lower level. C is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 55

EXAM KEY 06/2005 Page 55 of 75

EX05077 The plant is operating at 100% power when UPS E-IN-1 TRIPPED annunciator illuminates. Local investigation indicates that IN-1 has tripped from over voltage.

The ARP refers you to ABN-ELEC-INV, which di rects you to verify which of the following?

A. The static switch has auto transferr ed to Alternate AC input from MC-7A.

B. The static switch has auto transferr ed to Alternate AC input from MC-7F.

C. The Kirk Key has been manually transferred to Alternate AC input from MC-8A.

D. The Kirk Key has been manually transferred to Alternate AC input from MC-8F.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 262002A2.02 - Ab ility to predict the impacts of the following on the UPS and based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations: Over voltage IMP 2.5

REFERENCE:

ABN-ELEC-INV rev. 1, page 2 4.800.

C1 drop 6-4, SD000194 UPS rev. 9, pages 3-6 & Fig. 3

SOURCE: NEW QUESTION -

T2, GP1 LO: 5896

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: The alternate power supply for t he Kirk Key interlock is from MC-7A so C &

D are incorrect. The static switch Al ternate AC input is from MC-7F which makes A incorrect. MC-7A is the bypass source. MC-7F is the alternate AC

input for IN-1, B is correct. COMMENTS: Completed ju stification statement.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 56

EXAM KEY 06/2005 Page 56 of 75

EX05078 The plant was operating at 23% power when annunc iator MAIN GEN EXCITER FIELD BKR TRIP illuminated.

Which of the following is correct for this condition?

A. Enter ABN-BKR-FAULT.

B. Enter ABN-GENERATOR.

C. Enter PPM 3.3.1 Reactor scram.

D. Enter PPM 5.1.1 RPV Control.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 262002A2.01 - Ab ility to predict the impacts of the following on the AC ELECTRICAL DISTRIBUTION and based on those predictions, use

procedures to correct, control or mitigate the consequences of those abnormal

conditions or operations: Turbine/generator trip IMP 3.4

REFERENCE:

4.800.C4 drop 8-2

SOURCE: NEW QUESTION -

T2, GP1 LO: 5520

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: When the field breaker trips, the main generator and turbine trip. Since reactor power was less than 25%, there is no reactor scram. A, C, and D are incorrect. B is the correct answer. COMMENTS: Revised A from "The main tu rbine remains in operation as before".

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 57

EXAM KEY 06/2005 Page 57 of 75

EX05079 Which of the following describes the reason St andby Gas Treatment auto starts on the high radiation signal from Reactor Building Ventilation?

Standby Gas Treatment...

A. recirculates and filters reactor buildi ng atmosphere during accident conditions to allow personnel entry.

B. maintains a negative pressure in the reac tor building of at least 0.5 inches during accident conditions.

C. limits the release of radioactive mate rial within the guidelines of 10CFR100 during accident conditions.

D. provides controlled air movement from areas of potentially low radiation to areas of potentially high radiation.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 295034 EK1.02 - Knowledge of the operational implications of the following concepts as they apply to SECO NDARY CONTAINMENT VENTILATION HIGH RADIATION: Personnel protection IMP 3.8

REFERENCE:

SD000144 SGT rev. 12, page 3

SOURCE: BANK QUESTION - MODIFIED - 99 NRC Exam ex99054

- T1, GP2

LO: 5821

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: By definition, SGT reduces the discharge to the atmosphere to less than the limits of 10CFR100. C is correct. COMMENTS: Reworded distract ors B & D to be credible.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 58

EXAM KEY 06/2005 Page 58 of 75

EX05080 Which of the following requires the operator to have a procedure in hand?

When responding to a transient and he is-

A. taking manual control of the FW level controller to prevent exceeding level 8.

B. performing the immediate actions of PPM 3.3.1 Reactor Scram.

C. starting SW-P-1A following a failure to auto start.

D. starting RHR in suppression pool cooling.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 2.1.1 Knowledge of Conduct of Operations Requirements IMP 3.7

REFERENCE:

OI-9 rev. 2, page 29

SOURCE: NEW QUESTION -

T3 LO: 6060

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: As stated in OI-9, A, B, and C are specifically exempted from have a procedure present. Only D would require the use of a procedure and is

correct. COMMENTS: Changed stem from "you are" to "he is".

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 59

EXAM KEY 06/2005 Page 59 of 75

EX05081 The plant is operating at 99% power. RHR-P-2B was in operation for surveillance when a fault caused an overcurrent on RHR-P-2B. The overcurrent caused a lockout on BKR 8-3 and a loss of SM-8.

Which of the following is correct for these conditions?

A. The lockout on BKR 8-3 auto resets afte r the breaker for RHR-P-2B is racked out.

B. The lockout relay must be manually reset at BKR 8-3 after RHR-P-2B is racked out. C. SM-8 can be manually repowered after RHR-P-2B is racked out.

D. SM-8 will repower automatically after RHR-P-2B is racked out.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 2.4.48 Ability to interpret contro l room indications to verify the status and operation of system/and understand how operator actions and directives affect plant and system conditions IMP 3.5

REFERENCE:

SD000182 AC, rev. 13, page 20

SOURCE: BANK QUESTION - 03 NRC Exam EX03030 -

T3 LO: 5049d

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Prior to the reset of an 86 dev ice the fault must be cleared from the bus.

The 86 then can be manually reset and the breaker can be close manually.

B is correct. COMMENTS: Changed B from "The lockout rela y must be manually reset at BKR 8-3 before the bus can be repowered.".

Changed 'as soon as' in all distractors to 'after' to reduce sense of urgency.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 60

EXAM KEY 06/2005 Page 60 of 75

EX05109 Columbia is operating at 100% power. Due to a t agging error the N2/5 breaker is manually tripped open.

Which of the following explains the plant response to this tagging error?

A. The reactor scrams at +13 inches. RCI C and HPCS auto start at -50 inches and returns RPV level to +54 inches.

B. RPV water level initially drops to LT +31.5 inches. Both Reactor Recirc pumps runback to 51 Hz. RPV level slowly returns back to normal.

C. The reactor scrams at +13 inches. Feedwater system returns reactor water level back to normal.

D. RPV water level initially goes up but remains less than +54 inches. Feedwater system returns reactor water level back to normal.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 295001 AK3.01 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Reactor water level response IMP 3.4

REFERENCE:

Simulator

SOURCE: NEW QUESTION -

T1, GP1 LO: 5023

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: With the trip of 1 RRC pump, co re voiding increases which results in an immediate power reduction and an increase in indicated level. This makes

D the only correct answer. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 61

EXAM KEY 06/2005 Page 61 of 75

EX05083 The plant is in Mode 4 with RHR-P-2B in operation in Shutdown Cooling.

Which of the following would cause the loss of RHR-P-2B?

A. Reactor pressure 115 psig B. Reactor level +10 inches C. Drywell pressure 1.96 psig D. RHR SDC flow equal to 5.3 psid

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: AA1.02 - Ability to operate and/or monitor the following as they apply to LOSS OF SHUTDOWN COOLING: RHR/shutdown cooling

REFERENCE:

SD000173 NS4, rev. 10, pages 5, 12, &21

SOURCE: NEW QUESTION -

T1, GP1 LO: 5597

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: Of the signals given, only B c auses the loss of RHR-P-2B and the loss of shutdown cooling. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 62

EXAM KEY 06/2005 Page 62 of 75

EX05084 The plant is operating at 100% power with DG-1 in operation for a surveillance. A fire alarm has been received in the DG-1 room. Subsequent to this annunciator, the standby fire pumps start to sequence on.

What are the required immediat e actions for these conditions?

A. Sound the Alerting Tone for 5 seconds Announce the location of the fire

Repeat the above two steps B. Sound the Alerting Tone for 5 seconds Announce evacuation of the Diesel Generator building Repeat the above two steps C. Evacuate the Diese l Generator building Direct the Fire Brigade response using the ROLM PA

Notify the Hanford Fire Department D. Send an operator to verify the fire is real Announce the location

Notify the Hanford Fire Department

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 600000 AA2.03 - Ab ility to determine and/or interpret the following as they apply to PLANT FIRE ON SITE: Fire Alarm

REFERENCE:

ABN-FIRE rev. 7, page 2

SOURCE: NEW QUESTION -

T1, GP1 LO: 6902

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: As stated in the procedure, only the actions in A are the correct immediate actions for the indications given. COMMENTS: Revised B to include 'announce the..'.

Changed 'as' to 'has' in the stem.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 63

EXAM KEY 06/2005 Page 63 of 75

EX05110 Which of the following is prevented by the perform ance of the PC Gas leg of PPM 5.2.1 Primary Containment Control?

A. Exceeding a maximum hydrogen concentration of 6%.

B. An uncontrolled release of radioactivity to the environment.

C. Exceeding a maximum oxygen concentration of 5%.

D. A failure of the drywell downcomers.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 2.3.11 Ability to control radiation release IMP 2.7

REFERENCE:

PPM 5.0.10 rev. 8, page 277

SOURCE: BANK QUESTION - MODIFIED - 02 NRC Exam EX02074 -

T3 LO: 8425

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: PPM 5.0.10 states the reason/basis for the PC Gas control leg of PPM 5.2.1 is to prevent the uncontrolled release of radioactivity to the environment. B is correct. COMMENTS: Revised distractors A and C to be credible.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 64

EXAM KEY 06/2005 Page 64 of 75

EX05086 During a Control Room Evacuation, transfer swit ches are repositioned at the Remote Shutdown Panel to the EMERG position to separate equipment control circuits exposed to the control room fire from Safe Shutdown control signals initiated from-A. DG-1 Local Control Panel DIV 1 Switchgear Cubicles B. DG-2 Local Control Panel DIV 2 Switchgear Cubicles C. DG-1 Local Control Panel DIV 2 Switchgear Cubicles D. DG-2 Local Control Panel DIV 1 Switchgear Cubicles

ANSWER: B QUESTION TYPE: RO KA # & KA VALUE: 295016 AK2.03 - Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:

Local control stations. IMP 4.0

REFERENCE:

ABN-CR-EVAC bases 7.2

SOURCE: NEW QUESTION LO: 5886 RATING: L2 ATTACHMENT: NONE JUSTIFICATION: B is the only choice that includes all of the locations where safe shutdown control signals may be initiated. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 65

EXAM KEY 06/2005 Page 65 of 75

EX05111 With the plant operating at full power, an RCC leak inside containment occurs. Due to a failure of the operator being able to vent containment, the reactor is scrammed when drywell pressure reached 1.5 psig. Just after the manual scram, the automatic scram at 1.68 psig drywell pressure happens.

Which of the following automatic actions mitigate the effects of this leak?

A. RCC-V-6 Radwaste/Reactor building supply closes.

B. All RCC pumps trip on low surge tank level.

C. All RCC containment is olation valves close.

D. All RCC drywell cooling fan inlet isolation valves close.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 295018 AA1.03 - Ability to operat e and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Affected systems so as to isolate damaged portions. IMP 3.3

REFERENCE:

SD000196 RCC rev. 10, pages 8-10

SOURCE: NEW QUESTION LO: 5707

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: At 1.68 psig drywell pressure the RCC containment isolation valves close. A is incorrect because RCC-V-6 closes with LT 2 RCC pumps running. B is

incorrect because there is no trip of the RCC pumps on low surge tank level.

D is not correct because the isolation valves auto open on a fan start and auto close on a fan trip, but the f ans have no automatic start/stop. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 66

EXAM KEY 06/2005 Page 66 of 75

EX05088 Columbia was operating a full power when a series of events occurred that left the plant in an ATWS condition. PPM 5.1.2 was entered from PPM 5.1.1. Per PPM 5.1.2 Power Leg, Boron injection is required before wetw ell temperature exceeds 110°F.

Which of the following is the reason for boron inje ction before wetwell temperature reaches 110°F?

110°F is---..

A. the wetwell temperature at which Tec hnical Specifications requires a reactor scram. B. the wetwell temperature where no dam age to the RCIC system would occur if operation is required to support boron injection.

C. the highest wetwell temperature at wh ich initiation of boron injection will permit injection of the Hot Shutdown Boron We ight of boron before drywell pressure exceeds the Primary Containment Pressure Limit.

D. the highest wetwell temperature at wh ich initiation of boron injection will permit injection of the Cold Shutdown Boron Wei ght of boron before wetwell temperature exceeds Heat Capacity Temperature Limit.

ANSWER: A QUESTION TYPE: RO KA # & KA VALUE: 295026 EK3.04 - Knowledge of the reasons for the following as they apply to HIGH SUPPRESSION POOL WATE R TEMPERATURE: SLC injection IMP 3.7

REFERENCE:

PPM 5.0.10 Pg 189 of 318

SOURCE: NEW QUESTION LO: 8086 RATING: L2 ATTACHMENT: NONE JUSTIFICATION: BIIT is defined as: the wetwell te mperature at which Tec hnical Specifications would require a reactor scram or the highest wetwell temperature at which initiation of boron injection will permit injection of the Hot Shutdown Boron Weight of boron before wetwell tem perature exceeds Heat Capacity Temperature Limit. Only A is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 67

EXAM KEY 06/2005 Page 67 of 75

EX05112 A loss of all high pressure feed sources has occurr ed at Columbia. As RPV level drops, PPM 5.1.1, RPV Control, establishes adequate core cooling at different RPV levels. Per PPM 5.1.1, adequate core cooling can be defined four ways:

Core Submergence

Steam Cooling with Injection

Steam Cooling without injection

Spray Cooling

Which of the following is the reason Spra y Cooling can provide adequate core cooling?

A. RPV level is maintained GT -201 inches with HPCS or LPCS flow GT 6000 gpm which maintains clad temperature less than 1500°F.

B. RPV level is maintained GT -201 inches with RHR-A, B, or C combined flow GT 6000 gpm which maintains clad temperature less than 1800°F.

C. RPV level is maintained GT -210 inches with RHR-A, B, or C combined flow GT 6000 gpm which maintains clad temperature less than 2200°F.

D. RPV level is maintained GT -210 inches with HPCS or LPCS flow GT 6000 gpm which maintains clad temperature less than 2200°F.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 295031 EK3.03 - Knowledge of the reason for the following responses as they apply to REACTOR LOW WATER LEVEL: Spray Cooling IMP 4.1

REFERENCE:

PPM 5.0.10

SOURCE: NEW QUESTION LO: 8018

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: PPM 5.0.10 defines Spray cooli ng as GT -210" with HPCS or LPCS flow at GT 6000 gpm. Clad temperature will not exceed 2200°F. D is correct. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 68

EXAM KEY 06/2005 Page 68 of 75

EX05090 Columbia Generating Station is operating in MODE 1 at 95% power. Due to a fault on the bus, a loss of power to S1-2 occurs.

Which of the following describes t he operational impact of a loss of S1-2 on the diesel generators?

A. DG-1 cannot be started locally or from the control room.

B. DG-2 cannot be started locally or from the control room.

C. DG-3 could be started locally but not from the control room.

D. DG-2 could be started locally but not from the control room.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 264000 K1.02 - Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY DG S and the following: DC electrical distribution. IMP 2.9

REFERENCE:

SD000200 PG 26

SOURCE: NEW QUESTION LO: 7653

RATING: H2

ATTACHMENT: NONE

JUSTIFICATION: Per SD000188, DG-2 cannot be start ed locally or from the control room. B is the correct answer. The loss of S1-2 has no effect on the other diesels. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 69

EXAM KEY 06/2005 Page 69 of 75

EX05091 Columbia is in the process of a startup follo wing a refueling outage with power being supplied from TR-S. Preparations for starting the first RRC pump, RRC-P-1A, have been completed. The Channel Selector Switch for Channel 1A1 is in t he OFF position and the Channel Selector Switch for Channel 1A2 is in the ON position.

The Reactor Operator momentarily depre sses the ASD START pushbutton for RRC-P-1A.

Which of the following describes the expected pump start sequence?

A. ASD Channel 1A2 "READY" light immediat ely illuminates. RRC-P-1A then starts and ramps to approximately 400 RPM which correlates to 15 Hz.

B. RRC-P-1A starts immediately. Pump speed ramps to approximately 150 RPM which correlates to 15 Hz.

C. RRC-P-1A starts immediately. Pump speed ramps to approximately 450 RPM which correlates to 15 Hz.

D. ASD Channel 1A2 "READY" light immediat ely illuminates. After a five second time delay RRC-P-1A starts and ramps to 450 RPM which correlates to 15 Hz.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 202001 A3.02 - Ability to monitor automatic operation of the RECIRCULATION SYSTEM including: Pump start sequence. IMP 3.1

REFERENCE:

SOP-RRC-START

SOURCE: NEW QUESTION LO: 9681

RATING: H2

ATTACHMENT: NONE

JUSTIFICATION: Per SOP-RRC-START, the "READ Y' light should already be illuminated and the RRC pump immediately starts and increases speed to approximately 450 RPM (15 Hz). COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 70

EXAM KEY 06/2005 Page 70 of 75

EX05092 Due to a primary coolant pressure boundary leak, the CRS, ordered a reactor scram prior to the automatic scram signal being generated.

Wetwell pressure has steadily increased and is currently 15 psig. All systems functioned as designed in response to the event.

Which of the following is correct concerning t he spraying of containment at this pressure?

A. If drywell sprays are placed on the 'B' RHR loop, indicated flow will be approximately 8500 GPM.

B. If wetwell sprays are placed on the 'B' RHR loop, indicated flow will be approximately 1000 GPM.

C. If a single loop of RHR was utilized for bot h wetwell and drywell sprays, indicated flow would be approximately 4200 GPM.

D. If drywell sprays are placed on the 'A' RHR loop, indicated flow will be approximately 6500 GPM.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 226001 A3.03 - Ability to m onitor automatic operations of THE CONTAINMENT SPRAY SYSTEM MODE incl uding: System flow IMP 2.8

REFERENCE:

PLANT SIMULATOR

SOURCE: NEW QUESTION LO: 5774, 5777

RATING: H2

ATTACHMENT: NONE

JUSTIFICATION: Pump run out is at approx imately 8000 GPM. Wetwell spray flow is approximately 500 GPM and Drywell spray flow is approximately 6500 gpm

on the Plant Simulator COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 71

EXAM KEY 06/2005 Page 71 of 75

EX05093 Columbia is operating at 100% power when the on line after filter for the CAS system becomes clogged. Control air pressure starts to drop.

If efforts are unsuccessful at stopping the pressure decrease, which of the following describes the CAS system response and procedural guidance given to mitigate the consequences of this event?

A. The standby CAS air compressor starts at 100 psig. ABN-CAS directs inserting a manual scram if any outboard MSIV starts to close.

B. The Service Air Compressor, SA-C-1, fully loads at 105 psig. ABN-CAS directs that if any inboard MSIV has closed, pl ace its control switch in the CLOSED position.

C. The Control Air Desiccant Dryer Bypass Valve, CAS-PCV-1, closes at 75 psig.

ABN-CAS directs that if two or more r ods start to drift, manually scram the reactor. D. The Service Air Header Isolation Valv e, SA-PCV-2, opens at 80 psig. ABN-CAS directs that if a complete loss of ai r is apparent, manually scram the reactor.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 300000A2.01 - Ability to predict the impacts of the following on the INSTRUMENT AIR SYSTEM and based on those predictions, use

procedures to correct, control or mitigate the consequences of those

abnormal conditions or operations: Air drye r and filter malfunctions IMP 2.9

REFERENCE:

ABN-CAS Rev 5 page 2 and 3

SOURCE: NEW QUESTION LO: 5878

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: The standby Comp. starts at 100 psig, CAS-PCV-1 opens at 75 psig. The SA Comp. is always fully loaded when running. SA-PCV-2 closes to 80 psig.

ABN-CAS directs inserting a scram if any outboard MSIV starts to close COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 72

EXAM KEY 06/2005 Page 72 of 75

EX05094 With Columbia operating in MODE 1, RHR-P-2B is running to lower suppression pool level. SW-P-1B is operating in support of the evolution. D ue to a problem with the feedwater level control system, the CRS orders a reactor scram prior to t he +13" automatic scram. When the Main Turbine trips, TR-S develops an overcurrent condition and a lockout is generated. SM-7 and SM-8 are re-energized from the Backup Transformer.

All systems operate as designed to this event.

Which of the following describes the response of the Service Water System to this event?

When power is restored,-

A. SW-P-1B immediately re-starts.

SW-V-2B and SW-V-12B remain open.

B. after a 20 second time delay, and after SW-V-2B is fully closed, SW-P-1B re-starts. When SW-P-1B starts, SW-V-2B re-opens.

C. after a 20 second time delay, and afte r SW-V-2B and SW-V-12B are both fully closed, SW-P-1B re-starts. D. SW-V-12B fully closes. When SW-V-12B is fully closed, SW-P-1B re-starts and SW-V-12B re-opens.

ANSWER: B QUESTION TYPE: RO KA # & KA VALUE: 295003 2.1.24 - Ability to obtai n and interpret station electrical and mechanical drawings IMP 2.8

REFERENCE:

EWD-58E-004

SOURCE: NEW QUESTION LO: 4046 RATING: H3 ATTACHMENT: EWD-58E-004 JUSTIFICATION: As per the EWD, SW-V-2B has to be closed before SW-P-1B will restart.

There is also a 20 second time delay. SW-V-12B strokes closed but stops

mid-stroke when SW-P-1B starts and then goes full open. If SW-V-12B were

to go full closed, SW-P-1B would trip. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 73

EXAM KEY 06/2005 Page 73 of 75

EX05095 SGT A has been placed in operation for a surveillance.

Which of the following describes the operation of strip heater, SGT-ESH-1A?

A. SGT-TS-1A21/1A41 open at 125° F and a llow the heaters to automatically cycle back on at 110° F.

B. SGT-RLY-ESH1A11/1A31 close at 125° F and allow the heaters to automatically cycle back on at 110° F.

C. SGT-TS-1A11/1A31 and SGT-TC-1A1/1A2 cause the heaters to cycle on and off between 90° F and 110° F.

D. SGT-TS-1A11/1A31 cause the heater s to cycle on at 110° F and SGT-TS-1A21/1A41 cause the heater to cycle off at 125° F.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 261000 2.1.24 - Ability to obtai n and interpret station electrical and mechanical drawings. SGT System. IMP 2.8

REFERENCE:

EWD-39E-004

SOURCE: NEW QUESTION LO: 4047

RATING: H3

ATTACHMENT: EWD-39E

-004 and EWD-46E-157

JUSTIFICATION: Per the EWD SGT-TS-1A11, 1A31 open if temp rises above 110°F and de-energize the heater. SGT-TC-1A1/1A1 are closed when the temp is less

than 90° F which causes the heaters to cycle off and on between 90° and

110° F. If temperature increases to GT 125° F, SGT-RLY-ESH1A (CR1) de-energizes, which prevent s the operation of the heaters until the reset is pushed. Only C is correct. COMMENTS: Changed 'place' to 'placed' in the stem.

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 74

EXAM KEY 06/2005 Page 74 of 75

EX05096 The pressure input from the NBI system to the Main Steam Relief Valves has been lost.

What effect does this loss of pressure input from the NBI system have on the operation of the SRVs? A. The Safety and the Relief mode of operation would remain unaffected.

B. Only the Safety mode of operation would be affected.

C. Only the Relief mode of operation would be affected.

D. Both the Safety and Relief m ode of operation would be affected.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 239002.K6.01 - Knowledge of the e ffect that a loss or malfunction of the following will have on the RELIEF/SAFETY VALVES: NBI pressure indication IMP 3.2

REFERENCE:

LO000128

SOURCE: NEW QUESTION LO: 5527; 7638j

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: NBI provides pre ssure input to the 18 SRVs for operation of the SRVs in the RELIEF mode of operation. The SAFET Y mode of operation is actuated directly by the force exerted upon the valve spring by reactor pressure as

remains unaffected by a loss of the pressure input. COMMENTS:

COLUMBIA RO WRITTEN EXAM RETAKE QUESTION # 75

EXAM KEY 06/2005 Page 75 of 75

EX05097 The plant was operating at 25% power when annunciator RHR A PUMP ROOM WATER LEVEL HIGH illuminated. PPM 5.5.27 has been perform ed and the operator repor ts RHR-P-2A Room water level as 47 inches.

Which of the following is correct for these conditions?

A. RHR-P-2A may not be able to perform it s intended safety function if required.

B. RHR-P-2A may not be operated under any conditions.

C. RHR-P-2A may only be started one time under these conditions.

D. RHR-P-2A may only be started w hen directed by the Shift Manager.

ANSWER: A QUESTION TYPE: RO

KA # & KA VALUE: 295036 EA2.01 - Ab ility to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL:

Operability of components within the affected area IMP 3.0

REFERENCE:

PPM 5.3.1 Sec Cont Control, rev.13 PPM 5.0.10 rev. 8, pages 78 & 301

SOURCE: NEW QUESTION -

T1, GP2 LO: 8040

RATING: H3

ATTACHMENT: YES - Table 25 from PPM 5.3.1

JUSTIFICATION: The level given in the stem is above the Max Safe Operating Level in Table

25. PPM 5.0.10 states the equipm ent may not be able to perform its intended safety function with water level greater than max safe. A is correct. COMMENTS: Changed D from "There are no operati onal implications with water at this time.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 1

EXAM KEY 06/2005 Page 1 of 25

EX05001 The plant was operating at 99% power when a fire occurred in the control room that required an immediate evacuation. The immediate actions were performed and the following conditions now exist:

Reactor Power is 6% and steady 2 SRVs are cycling open and closed Reactor level is -15 inches and down slow Drywell pressure is 1.83 psig and up slow Suppression Pool level is +3 inches and up slow

Which of the following procedures ta kes precedence under these conditions?

A. PPM 5.1.1 RPV Control B. PPM 5.1.2 RPV Control ATWS C. ABN-CR-EVAC D. PPM 5.2.1 Primary Containment Control

ANSWER: C QUESTION TYPE: SRO

KA # & KA VALUE: 295016AA2.01 - Ab ility to determine and/or interpret the following a they apply to CONTROL ROOM ABANDONMENT: Reactor Power 55.43.5 IMP 4.2

REFERENCE:

PPM 13.1.1 rev. 33, pages 21

& 36 ABN-CR-EVAC rev. 7, pages 6 &7

SOURCE: MODIFIED QUESTION 2002 exam -

SRO Tier 1 GP 1

LO: 6105

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Conditions indicate an ATWS that would require an entry into both PPM 5.1.1 and 5.1.2. There are also entry conditions for PPM 5.2.1. However, a control room evacuation is required because of the fire, therefore ABN-CR-EVAC takes precedence as stated on the procedure. C is correct. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 2

EXAM KEY 06/2005 Page 2 of 25

EX05002 The plant was operating at 99% power when a trans ient occurred which resulted in an offsite release. The Standby Gas Treatment system is in operation. QEDPS indicates the following dose at the site area boundary:

TEDE dose rate 2 mrem/hr CEDE thyroid dose rate 159 mrem/hr

Which of the following is correct for these conditions?

The release originates from the-

A. turbine building and PPM 5.4.1 Radioactivi ty Release Control entry is required.

B. turbine building but PPM 5.4.1 Radioactivity Release Control is not required.

C. reactor building and PPM 5.4.1 Radioactivi ty Release Control entry is required.

D. reactor building but PPM 5.4.1 Radioactivity Release Control is not required.

ANSWER: A QUESTION TYPE: SRO KA # & KA VALUE: 295038 EA2.04 - Ab ility to determine and/or interpret the following a they apply to HIGH OFF-SITE RELEASE RATE: S ource of off-site release 55.43.5 IMP 4.5

REFERENCE:

PPM 13.1.1 rev. 33 page 37

SOURCE: NEW QUESTION -

SRO T1, GP 1 LO: 5821, 8017

RATING: H3 ATTACHMENT: YES - PPM 13.1.1 re

v. 33 page 37, and tables 4 & 5 JUSTIFICATION: As stated in the description and purpose of SGT, radioactive iodines are filtered out in the HEPA filt ers. A high CEDE dose rate indicates that either SGT is not in operation or the release is from a different source. This makes C and D incorrect. The CEDE dose rate is high enough for an ALERT

classification and requires an entry into PPM 5.4.1. A is the correct answer. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 3

EXAM KEY 06/2005 Page 3 of 25

EX05003 The plant was operating at 90% power with HPCS out of service for a motor replacement when a loss of both SM-1 & SM-2 occurred and a failure of RCIC to start.

Assume no operator actions are taken for this transient.

Which of the following is the most restrictive time notification required for these conditions?

A. PPM 1.10.1 requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report due to a reactor scram.

B. PPM 1.10.1 requires an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report due to RCIC failure.

C. GIH-9.1.3 requires that the CEO be notified in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. GIH-9.1.3 requires that t he CEO be notified in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ANSWER: C QUESTION TYPE: SRO

KA # & KA VALUE: 295009 2.1.14 - Knowledge of system status criteria which require the notification of plant personnel 55.43.5 IMP 3.3

REFERENCE:

PPM 1.10.1 rev. 26, page 10 GIH-9.1.3 rev. 0

SOURCE: NEW QUESTION -

T1, GP2 LO: 6011

RATING: H4

ATTACHMENT: YES - PPM 1.10.1 rev. 26, page 10, & 11 GIH-9.1.3 rev. 0

JUSTIFICATION: A is incorrect because the scram requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report. B is incorrect because it would require an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notific ation. D is incorrect because GIH

9.1.3 requires

a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notificati on of the CEO on an emergency classification GT UE. C is correct. In this scenario a SAE would be declared

due to RPV level less that -161" as both feed pumps trip on low suction and so there is no HP feed in scenario.

COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 4

EXAM KEY 06/2005 Page 4 of 25

EX05004 The plant is operating at 99% power with the LPCS System Operability Test, OSP-LPCS/IST-Q702, in progress. Surveillance te st results reveal the following:

LPCS-FCV-11 opens at 679 gpm

LPCS-FCV-11 closes at 1148 gpm

All other plant equipment is operating as required.

Which of the following is correct for these conditions?

LPCS -

A. must be declared inoperable immediately.

B. must be declared inoperable and the minimu m flow function must be restored to operable in 7 days.

C. minimum flow function must be declared inoperable immediately.

D. minimum flow function must be restored to operable in 7 days.

ANSWER: D QUESTION TYPE: SRO KA # & KA VALUE: 209001 2.2.24 - Ab ility to analyze the affect of ma intenance activities on LCO status 55.43.2 IMP 3.8

REFERENCE:

TS 3.3.5.1 and bases.

SOURCE: NEW QUESITON -

T2, GP 1 LO: 6925 RATING: H3 ATTACHMENT: YES - TS 3.3.5.

1, Table 3.3.5.1-1 page 1 JUSTIFICATION: According to table 3.3.5.1-1, the LPCS minimum flow function is out of spec.

3.3.5.1.d gives the direction to restor e the channel to operable in 7 days.

Since the redundant feature ECCS is operable, the LPCS system does not have to be declared inop. D is the correct answer. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 5

EXAM KEY 06/2005 Page 5 of 25

EX05005 The plant was operating at 99% power when a trans ient occurred. The operating crew took all immediate actions. The following conditions exist 10 minutes following the transient:

Reactor level -66 inches and down slow Reactor pressure 1096 psig and steady MSIVs closed MS-RV-1A & 1B open Suppression pool temp 88° F Drywell pressure 1.58 psig and stable

Which of the following procedures should have been entered to mitigate the transient?

A. PPM 5.1.1 RPV Control and PPM 5.

3.1 Secondary

Containment Control B. PPM 5.1.2 RPV Control ATWS and PPM

5.3.1 Secondary

Containment Control C. PPM 5.1.1 RPV Control and PPM 5.

2.1 Primary

Containment Control D. PPM 5.1.2 RPV Control ATWS and PPM

5.2.1 Primary

C ontainment Control

ANSWER: B QUESTION TYPE: SRO KA # & KA VALUE: 295015AA2.01 - Ab ility to determine and/or interpret the following a they apply to INCOMPLETE SCRAM: Reactor power 55.43.5 IMP 4.3

REFERENCE:

PPM 5.1.2 RPV Control ATWS and PPM 5.3.1 Secondary Containment Control entry conditions, LO000128 rev. 8, Main Steam System SOURCE: NEW QUESTION -

T1, GP 2 LO: 8017 RATING: H3 ATTACHMENT: NONE JUSTIFICATION: The conditions given indicate approximately 12% steam flow 10 minutes following the transient. This is indicative of an ATWS. A and C are

incorrect. With reactor level at -66 inches, RB ventilation supply and

exhaust fans have tripped which results in an entry into PPM 5.3.1 from hi Sec. Cont. Pressure. There are no entry conditions at this time for PPM

5.2.1. B is correct. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 6

EXAM KEY 06/2005 Page 6 of 25

EX05006 During performance of OSP-ELEC-M702 DG-2 Mont hly Operability Test, the equipment operator notifies you that annunciator 3.3 on E-CP-DG/C P2, ENG. 1 LUBE OIL LEVEL LOW is illuminated and there is 8 inches of oil above the LOW ma rk on the Engine 1 lube oil sump dipstick.

Which of the following is correct for these conditions?

A. Declare DG-2 inoperable immediately.

B. Restore the lube oil inventory to GT 165 gallons in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

C. Restore the lube oil inventory to GT 283 gallons in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

D. Restore the lube oil inventory to GT 330 gallons in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

ANSWER: D QUESTION TYPE: SRO

KA # & KA VALUE: 264000 2.1.33 - Ab ility to recognize indications for system operating parameters, which are entry-level conditi ons for technical specifications. 55.43.2 IMP 4.0

REFERENCE:

OSP-ELEC-M702 rev. 20, page 13, 4.DG2 drop 3-3 rev. 10, page 13, and TS 3.8.3 and bases

SOURCE: NEW QUESTION -

T2, GP1 LO: 10305

RATING: H3

ATTACHMENT: YES - 4.DG2 drop 3-3 rev. 10, page 13, and TS 3.8.3

JUSTIFICATION: The indications given put lube oil level between 283 gal and 330 gal. This requires the lube oil be returned to GT 330 gal in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. D is correct. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 7

EXAM KEY 06/2005 Page 7 of 25

EX00022 The plant was operating at 98% pow er when the following occurred:

SGT started

CSP/CEP isolated

CN makeups isolated

CR and TSC Emerg Filtration starts and aligns to remote air intakes

RB Emerg Room Coolers start

RB Lighting quenches

RB HVAC isolates and fans trip

The plant remains operating at power following the initiations.

Which of the following is correct concerning these conditions?

Enter-A. PPM 5.1.1 RPV Control and PPM 5.1.2 RPV Control ATWS B. PPM 5.1.2 RPV Control ATWS and PPM

5.2.1 Primary

C ontainment Control C. PPM 5.3.1 Secondary Cont ainment Control and ABN-FAZ D. PPM 5.4.1 Radioactivity Release Control and ABN-FAZ

ANSWER: C QUESTION TYPE: SRO KA # & KA VALUE: 2.4.7 Knowledge of how the event based emergency/abnormal operating procedures are used in conjunction with the symptom-based EOPs 55.43.5 IMP 3.8

REFERENCE:

SD000173 NS4 rev. 10 pages 5, 19, 20 & entry conditions for EOPs/ABN-FAZ SOURCE: BANK QUESTION -

T3 LO: 6914 RATING: H3 ATTACHMENT: NONE JUSTIFICATION: A and B are incorrect because there is no indication given that an ATWS occurred. -50" and 1.68 psig in the drywell would cause these conditions, they would also cause a scram. D is incorrect because the signal causing

the indications is a high rad in the ex haust plenum, there is no indication of an offsite release. C is correct because the isolation of RB HVAC causes a

high pressure entry into PPM 5.3.1 and an ABN-FAZ entry. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 8

EXAM KEY 06/2005 Page 8 of 25

EX05008 The plant was operating at 99% power when a DEH l eak resulted in a DEH pressure reduction to 1100 psig.

Which of the following is correct?

A. A reactor scram occurs because, under these conditions, control rod insertion may not initially add enough negative reactivity to overcome the positive reactivity

added by the pressure increase from a turbine trip.

B. A reactor scram occurs because, under these conditions, control rod insertion initially adds positive reactivity late in core life that must be compensated for by the trip of both recirc pumps.

C. A reactor scram occurs because, under these conditions, Recirc Pumps must be tripped late in core life to minimize the e ffect of all control rods withdrawn to the full out position and prevent exceeding the LHGR.

D. The plant continues to operate at 99% power.

ANSWER: A QUESTION TYPE: SRO

KA # & KA VALUE: 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits 55.43.2 IMP 3.7

REFERENCE:

TS 3.3.4.1 and Bases, RPS Systems Text SD000161 rev. 12, page 12

SOURCE: BANK QUESTION -

T3 LO: 6925, 5949

RATING: H3

ATTACHMENT: None

JUSTIFICATION: Since the DEH pressure given is less than the RPS scram setpoint, D is incorrect. The correct basis as in TS is stated in A. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 9

EXAM KEY 06/2005 Page 9 of 25

EX05115 If a fire has been reported and confirmed at one of t he following locations, to which location should the CRS have the Hanford Fire Departm ent respond to and not the Fire Brigade?

A. Main Transformer Yard within the Protected Area B. ISFSI within the Protected Area C. Service Water Pump House 1A D. ASD Building

ANSWER: B QUESTION TYPE: SRO

KA # & KA VALUE: 2.4.27 Knowledge of fire in the plant procedure 55.43.2 IMP 3.5

REFERENCE:

ABN-FIRE pg 2 and 15

SOURCE: NEW QUESTION -

T3 LO: 6783

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Per ABN-FIRE, the CRS/Shift Manager directs the Fire Brigade to respond to areas included in answer A, C, and D.

If a fire exists in ISFSI in the protected area, the HFD is contacted. COMMENTS: Re-wrote question per NRC comment that previous was not an SRO type question.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 10

EXAM KEY 06/2005 Page 10 of 25

EX05009 The plant was operating at 23% power when the Ma in Turbine tripped and a fuel bundle is dropped in the spent fuel pool during loading of a spent fuel container for ISFSI.

The following conditions exist:

Reactor level is 25 inches and stable

SGT has auto initiated

RB HVAC has isolated

Drywell pressure is 1.1 psig

Wetwell level is -1.9 inches

Drywell temperature is 129 °F

Which of the following actions is correct?

Enter-..

A. PPM 5.1.1 RPV Control and PPM 3.3.1 Reactor Scram.

B. PPM 5.1.1 RPV Control and PPM 5.

2.1 Primary

Contai nment Control.

C. PPM 3.3.1 Reactor Scram and PPM 5.

3.1 Secondary

Containment Control.

D. PPM 5.3.1 Secondary Cont ainment Control and immediately evacuate the refuel floor of all personnel.

ANSWER: D QUESTION TYPE: SRO KA # & KA VALUE: 295023 2.4.1 - Knowledge of EOP entry conditions and immediate action steps 55.43.5 IMP 4.6

REFERENCE:

ABN-FUEL-HAND rev. 2, page 2 PPM 5.0.10 rev. 8, page 294 SOURCE: NEW QUESITON -

T1, GP1 LO: 6897, 8017

RATING: H3 ATTACHMENT: NONE JUSTIFICATION: Since the reactor was at 23%

power, there was no scram from the turbine trip. There is no entry given for PPM 3.

3.1. Therefore, A-C are incorrect.

The immediate action for a dropped fuel bundle is to evacuate the refuel

floor and 5.3.1 must be entered because of the high exhaust plenum rad as indicated by the auto SGT start. D is correct. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 11

EXAM KEY 06/2005 Page 11 of 25

EX05114 Columbia is in day 15 of a 25 day refueling outage. The fuel shuffle is approximately 60%

completed and currently underway. OPS 4 repor ts that Standby Service Water spray pond temperatures are both 79°F.

Which of the following is correct?

A. Enter TS LCO 3.0.3 immediately.

B. Restore control room AC subsyst em to operable status in 30 days.

C. Suspend core alterations immediately.

D. No actions are required, LCO applicability not met.

ANSWER: C QUESTION TYPE: SRO KA # & KA VALUE: 295018 2.1.12 - Partial or Complete loss of CCW. Ability to apply technical specifications for a system. IMP 4.0

REFERENCE:

TS 3.7.4 and bases and TS 3.7.1

SOURCE: NEW QUESTION -

T1, GP1 LO: 5226 RATING: H2 ATTACHMENT: YES - TS 3.7.4 and Bases; TS 3.7.1 JUSTIFICATION: TS 3.7.4 bases states that SW and the UHS are part of the OPERABILITY requirements for CR HVAC to be operable.

TS 3.7.1 requires the UHS temp to be LE 77°F. A is incorrect because TS 3.0.3 is not applicable. B is

incorrect because both CR subsystems are inoperable. D is incorrect because TS 3.7.1 is applicable per TS 3.7.4 bases. C is correct. COMMENTS: Rewritten due to un-secure email sent.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 12

EXAM KEY 06/2005 Page 12 of 25

EX05011 Which of the following is NOT a basis for the low suppression pool water level Tech Spec?

Low suppression pool water level could result in-

A. less energy absorption and higher suppr ession pool temperatures following a DBA LOCA.

B. inadequate makeup water source for ECCS systems required following a DBA LOCA. C. excessive clearing loads from SRV Tailpipe pipes during subsequent SRV actuations.

D. inadequate steam condensation from SRV quenchers during subsequent SRV actuations.

ANSWER: C QUESTION TYPE: SRO

KA # & KA VALUE: 295030 2.4.21 - Knowledge of the parameters and logic used to assess the status of safety functions including:

Containment conditions 55.43.2 IMP 4.3

REFERENCE:

TS 3.6.2.2 Basis

SOURCE: NEW QUESTION -

T1, GP1 LO: 6925

RATING: L3

ATTACHMENT: NONE

JUSTIFICATION: A, B, and D are all bases for the low suppression pool water level LCO and are incorrect. C is correct. It is a basis for the high suppression pool level. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 13

EXAM KEY 06/2005 Page 13 of 25

EX05100 Which of the following would not require a 50.59 screening to take place?

A. A plant modification made to the FPC H eat Exchanger service water outlet valve that allows it to be throttleable.

B. Installation of a jumper around the so lenoid for ROA-V-1 which will prevent ROA-V-1 from closing on a loss of power.

C. Placing a portable heater in the SM-7 Switchgear Room during abnormally cold weather conditions to maintain operability.

D. Partial disassembly of DG-2 HVAC ducting to support repair of failed damper which is scheduled to be completed in one week.

ANSWER: D QUESTION TYPE: SRO

KA # & KA VALUE: 262001 2.2.8 - Knowledge of the process for determining if the proposed change /test/or experiment involves an unreviewed safety question 55.43.3

IMP 3.3

REFERENCE:

10CFR50.59 Resource Manual Pg 42; SWP-LIC-02 Rev. 4 page 3

SOURCE: NEW QUESTION -

T2, GP1 LO: NONE

RATING: H3

ATTACHMENT: NONE

JUSTIFICATION: Disassembly of DG-2 ducting to support maintenance activities that are LT 90 days in duration do not require 50.59 review. The other three choices all require 50.59 review per references. D is correct. COMMENTS: Rewritten per NRC comment.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 14

EXAM KEY 06/2005 Page 14 of 25

EX05013 The plant is operating at 32% power with contro l rod 30-31 selected. One of 4 operable A level LPRMs and all of the operable C level LPRM s feeding the RBM system fail downscale.

Which of the following is correct?

A. The failure of these LPRMs does not inop the RBM system.

B. All control rod movement must be suspended immediately except by scram.

C. RBM-A is not required to be operable until 35% power.

D. RBM-A has to be restored to operable status in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ANSWER: D QUESTION TYPE: SRO

KA # & KA VALUE: 215002 2.1.12 Ab ility to apply technical specifications for a system. 55.43.2 IMP 4.0

REFERENCE:

SD000166 RBM rev. 10, page 5 and 8 TS 3.3.2.1

SOURCE: NEW QUESTION -

T2, GP2 LO: 5690, 7667

RATING: H3

ATTACHMENT: YES - TS 3.3.2.1

JUSTIFICATION: Center control rods provide inputs to each RBM from 8 LPRMs. Less than 1/2 the inputs causes a Rod Block and inops the RBM. A is incorrect. The

RBM is required to be operable at 30% power so C is incorrect. The required

action for one inop RBM at power is stated in D. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 15

EXAM KEY 06/2005 Page 15 of 25

EX05014 The plant was operating at 99% power with OG-RI S-601A Post Treatment Rad Monitor, failed upscale. A failure on PP-8A-A results in a loss of power and a downscale on OG-RIS-601B Post

Treatment Rad Monitor. All other plant equipment operates as expected.

Assuming no operator actions, which of the following is correct?

A. PPM 5.2.1 Primary C ontainment Control will not be entered.

B. PPM 5.1.1 RPV Contro l will be entered for RPV level, pressure, and power control directions.

C. The plant continues to operate at a new lower power level due to increased backpressure.

D. Bypass flow would isolate and all OG flow is directed through the Offgas absorbers.

ANSWER: B QUESTION TYPE: SRO

KA # & KA VALUE: 271000A2.08 - Ab ility to predict the impacts of the following on the OFF GAS System; and based on those predictions, use procedures to correct, control, or mitigate the consequences of those abno rmal conditions or operations: AC Dist failures 55.43.5 IMP 2.7

REFERENCE:

LO000187 Off Gas rev. 10, pages 24-25 SOURCE: NEW QUESTION -

T2, GP2 LO: 5621 RATING: H2 ATTACHMENT: NONE JUSTIFICATION: The loss describe causes an isolat ion of the Off Gas system. This results in increasing backpressure until the main turbine trips. C is incorrect. The

reactor scrams and an entry will be made into PPM 5.1.1 on reactor level. B is correct. Vacuum continues to decrease until the MSIVs shut which

causes suppression pool temp to increase and an entry into PPM 5.2.1. A is

incorrect. With both rad monitors de-energized Offgas flow is isolated by

OG-V-60 closing. D is incorrect. This system response is correct if just a hi

radiation signal is received. COMMENTS: Re-wrote distractor D to be credible.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 16

EXAM KEY 06/2005 Page 16 of 25

EX05015 The plant was operating at 100% power with all pl ant equipment operating as required. A transient occurred which caused the following indication.

No operator actions have been performed.

Reactor power 92% and stable Reactor level 36 inches and stable Core flow 100 mlbm/hr and stable RRC Pump Speed 60 hz RRC Loop A Flow 43,000 GPM RRC Loop B Flow 44,000 GPM JP Loop A Flow 43E6 lbm/hr JP Loop B Flow 56E6 lbm/hr

Several feedwater heater level annunciators illuminated and cleared.

Which of the following actions is required?

A. Declare the loop with the lower flow to be "not in operation" in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. 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 />.

C. ABN-RRC-LOSS.

D. PPM 3.3.1 Reactor Scram.

ANSWER: B QUESTION TYPE: SRO KA # & KA VALUE: 290002 2.1.7 - Ab ility to evaluate plant performance and make operational judgments based on operation characteristics / reactor behavior / and instrument interpretation 55.43.2 IMP 4.4

REFERENCE:

ABN-POWER rev. 4, page 2

SOURCE: NEW QUESTION -

T2, GP2 LO: 6727 RATING: H3 ATTACHMENT: YES - TS 3.4.2 JUSTIFICATION: The indications given are from a failure of the #1 and #2 jet pumps. This is an entry condition for ABN-POWER and does not require an entry into any

of the distracters. B is the only correct answer.

COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 17

EXAM KEY 06/2005 Page 17 of 25

EX05016 FIXED TYPO ON ANSWER KEY Maintenance is replacing the old refuel floor jib crane (1250 pounds) and wants to transport it over the spent fuel pool in order to remove it. The wate r level in the spent fuel pool is at the 605 feet elevation.

What are the limitations of this lift?

The load can be lifted to a maximum of -

A. 6 feet above the refuel floor.

B. 5 feet above the refuel floor.

C. 4 feet above the refuel floor.

D. 3 feet above the refuel floor.

ANSWER: C D QUESTION TYPE: SRO

KA # & KA VALUE:

2.2.26 Knowledge of refuel administrative requirements 55.43.7 IMP 3.7

REFERENCE:

LCS 1.9.2

SOURCE: BANK QUESTION -

T3 LO: 5362

RATING: H2

ATTACHMENT: LCS 1.9.

2 and Figure 1.9.2-1

JUSTIFICATION: The 1250 pound load can be lifted a maximum of 4 feet above the level of the spent fuel pool. Since the 606 elevat ion is 1 foot above the water level, the correct answer is 3 feet above the floor level. C D is correct.

COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 18

EXAM KEY 06/2005 Page 18 of 25

EX05017 The plant is operating at 22% power in preparat ion for a refueling outage. A containment purge has just been initiated to de-inert the containment.

Which of the following is correct for this condition?

A. It is acceptable to use SGT with i noperable heaters for the purge because the heaters have no effect on the ability of the train used for the purge to perform its function.

B. The SGT train used for containment purge is inoperable due to the controller being placed in manual and no core alterations, operations with the potential for

draining the reactor vessel, or movem ent of irradiated fuel is allowed.

C. The SGT train used for containment pur ge is inoperable due to the potential for rapid over pressurization prior to closur e of the containment isolation valves following a LOCA.

D. At this power level, it is correct to use both trains of SGT for the containment purge because there is no postulated acci dent that can damage the Standby Gas Treatment System.

ANSWER: C QUESTION TYPE: SRO KA # & KA VALUE: 2.3.9 Knowledge of the proce ss for performing a containment purge 55.43.4 IMP 3.4

REFERENCE:

ODCM 6.2.2.

6 and bases, TS 3.6.4.3 SOURCE: NEW QUESTION -

T3 LO: 9498 RATING: H3 ATTACHMENT: YES - ODCM 6.2.2.6 and TS 3.6.4.3 JUSTIFICATION: The plant is in Mode 1, which requires that the purge be through 1 train of SGT for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The train used for the purge, must be operable, which makes A incorrect. The direction is to use only 1 train at this power

level, which makes D incorrect. With 1 train of SGT operable, there is a 7

day limit to return SGT to operable. Core alts, etc are allowed with only 1

train inop. B is incorrect. The basis for this RFO states that the train used

for the purge is inoperable due to possible over pressurization during a

LOCA. C is correct. COMMENTS: Took ODCM bases out of attachment COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 19

EXAM KEY 06/2005 Page 19 of 25

EX05018 The plant is operating at 99% power with a stuck open SRV. Suppression pool temperature is now 111° F and going up.

Which of the following is the appropriate procedure path used under these conditions?

Entry into-..

A. PPM 5.2.1 Primary Cont ainment Control which requires entry into PPM 5.1.1 RPV Control which requires entry into PPM 3.3.1 Reactor Scram.

B. PPM 5.1.1 RPV Control and entry into PPM 5.2.1 Primary C ontainment Control which requires entry into PPM 5.1.3 Emergency RPV Depressurization.

C. PPM 5.1.1 RPV Control and entry in to ABN-SRV and entry into PPM 5.2.1 Primary Containment Control which r equires entry into PPM 5.1.3 Emergency RPV Depressurization.

D. PPM 5.2.1 Primary Cont ainment Control which requires entry into PPM 5.1.1 RPV Control which requires entry into PPM 5.1.3 Emergency RPV Depressurization.

ANSWER: A QUESTION TYPE: SRO

KA # & KA VALUE: 295026 2.4.16 - Knowledge of EOP implementat ion hierarchy and coordination with other support procedures 55.43.5 IMP 4.0

REFERENCE:

PPM 5.1.1, 5.2.1, 5.1.3, 3.3.1, and ABN-SRV SOURCE: NEW QUESTION -

T1, GP1 LO: 8017 RATING: H3

ATTACHMENT: YES - 5.2.1 WW temp leg, PPM 5.1.1 Power leg. HCTL curves.

JUSTIFICATION: The conditions given require EO P entries in PPM 5.2.

1 Primary Containment Control, which directs that PPM 5.1.1 RPV Control be entered before 110° F.

5.1.1 directs

exiting to 3.3.1 with all c ontrol rods in. There are no conditions given which require an ED and ABN-SRV actions are superceded by the

directions given in the EOPs . A is correct. COMMENTS: Rewritten per NRC comments.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 20

EXAM KEY 06/2005 Page 20 of 25

EX05019 The plant was operating at 75% power when a tr ansient occurred that resulted in a loss of feedwater. Control rod 30-31 did not fully inser

t. None of HPCS, CRD, and RCIC could be started and level has been returned by low pressure systems to the normal operating band following

Emergency Depressurization as directed by t he EOPs. All other plant systems operated as expected.

Which of the following is correct concerning these conditions?

As the Emergency Director, declare a(n)-

A. Unusual Event.

B. Alert.

C. Site Area Emergency.

D. General Emergency.

ANSWER: C QUESTION TYPE: SRO

KA # & KA VALUE: 295031 2.4.38 - Ability to take actions called for in the facility emergency plan

/ including supporting or acting as emergency director 55.43.5 IMP 4.0

REFERENCE:

PPM 13.1.1

SOURCE: NEW QUESTION -

T1, GP1 LO: 6131

RATING: H3

ATTACHMENT: YES - PPM 13.1.1 pages 14 and 15

JUSTIFICATION: The loss of high pressure system s results in reactor level being reduced to TAF and ED to allow LP systems to inject. One control rod not inserting by

itself does not require entry into t he ATWS procedure. Therefore, the correct answer is C SAE was declared due to 2.1.S.1. COMMENTS: Revised stem to specifically sa y we ED'ed to eliminate having reduced RPV pressure to feed with Condensate pumps which may have prevented RPV/L from dropping below TAF and getting a different answer.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 21

EXAM KEY 06/2005 Page 21 of 25

EX05020 A control rod withdrawal is in progress for a plant start up. Power is currently 250,000 counts on SRM A and C. SRM B and D indicate 95,000 counts.

Which of the following is correct for this condition?

A. Enter PPM 4.603.A7.3-4 SCRAM SYSTEM A.

B. Enter PPM 4.603.A7.2-7 ROD OUT BLOCK.

C. Enter PPM 3.3.1 Reactor Scram.

D. Control rod withdr awal can continue.

ANSWER: B QUESTION TYPE: SRO

KA # & KA VALUE: 215004A2.04 - Ab ility to predict the impacts of the following on the SRM System; and based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Upscale and downscale trips 55.43.5 IMP 3.7

REFERENCE:

SD000132 rev 10, page 26

SOURCE: NEW QUESTON -

T2, GP1 LO: 5795

RATING: H2

ATTACHMENT: NONE

JUSTIFICATION: The count level given for both A and C is above the scram setpoint on 2E5 counts. Since the shorting links are inst alled for a normal startup, the scram is not in affect. A and C are incorrect. The indications are above the rod

block setpoint of 1.0E5 counts so D is incorrect. Since the rod block setpoint has been exceeded, B is correct. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 22

EXAM KEY 06/2005 Page 22 of 25

EX05021 The plant is operating at 21% power following an outage. During the outage maintenance was performed on 6 SRVs with setpoints of 1165 psig and 1175 psig. These SRVs have not yet been

declared operable.

Which of the following is correct concerning these conditions?

A. Action must be taken within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the plant in Mode 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and Mode 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.

B. The Mode Switch must be plac ed in SHUTDOWN immediately.

C. Power can be increased to GT 25% as long as 2 of the inoperable SRVs are declared operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of exceeding 25% power.

D. Power can be increased to 24% without declaring the SRVS operable.

ANSWER: D QUESTION TYPE: SRO

KA # & KA VALUE: 239002 2.2.21 - Knowledge of pre and post maintenance operab ility requirements 55.43.2 IMP 3.5

REFERENCE:

TS 3.4.3, 3.4.4, 3.0.4

SOURCE: NEW QUESTON -

T2, GP1 LO: 10306

RATING: H3

ATTACHMENT: YES - TS 3.

4.3, 3.4.4, 3.0.4

JUSTIFICATION: We have 18 SRVs and TS 3.4.3 requires only 12 to be operable for the safety function at GT 25% power. Howe ver, 3.4.3 requires that at least 2 be operable in the lowest 2 lift groups.

The question states that there are no operable SRVs in the lowest 2 lift groups.

TS 3.0.4 states that the specified condition of the operability must be allowed for an indefinite time for 3.0.4 to be utilized to go above 25% power (the condition is limited for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in this

case). There is no 3.0.3 condition giv en in the question. Therefore, only D is correct. . COMMENTS: Revised justification from 3.4.4 to 3.4.3 in second sentence.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 23

EXAM KEY 06/2005 Page 23 of 25

EX05022 According to the Columbia Generating Station Fa cility Operating License, what is the maximum licensed power level?

A. 3323 mwt B. 3386 mwt C. 3423 mwt D. 3486 mwt

ANSWER: D QUESTION TYPE: SRO

KA # & KA VALUE: 2.1.10 Knowledge of conditions and limitations in the fac ility license. 55.43.1 IMP 3.9

REFERENCE:

Columbia License

SOURCE: NEW QEUSTION -

T3 LO: 10296

RATING: L2

ATTACHMENT: NONE

JUSTIFICATION: According to the facility operating license, 3486 is the max license power level. A is the old max level and B/

C are incorrect combinations of the correct numbers. COMMENTS: This question was added after deletion of the old 05022 to more evenly balance the exam to the requirements of 10CFR 55.43. The outline will be updated to reflect this change.

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 24

EXAM KEY 06/2005 Page 24 of 25

EX00100 The plant is operating at 97% power with a dischar ge from the Waste Collector Tank to the Circ.

Water Blowdown line underway. An annunciator is re ceived in the control room indicating that Process Rad Monitor FDR-RIS-606 (Radw aste Effluent) has failed downscale.

Which of the following is correct concerning these conditions?

A. The discharge may continue for up to 30 days provided grab samples are collected and analyzed for radioactivity of at least 10

-7 micro curie/ml, at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. The discharge may continue for up to 30 days provided that the discharge flow rate is estimated at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the release.

C. Stop the discharge. The discharge may continue when 2 independent samples have been analyzed and 2 technically qualifi ed members of the plant staff have independently verified the release calculations and the discharge valve lineup.

D. Stop the discharge. The discharge ma y continue when a temporary monitor has been installed and the monitor calibration has been verified by analysis of 2

independent batch samples.

ANSWER: C QUESTION TYPE: SRO KA # & KA VALUE: 2.3.3 Knowledge of SRO responsibilities for auxiliary system that are outside the control room (e.g. / waste di sposal and handling systems) 55.43.4 IMP 2.9

REFERENCE:

ODCM 6.1.

1 table 6.1.1.1-1 SOURCE: BANK QUESTION from 2000 NRC exam - Slightly modified for clarification -

T3 LO: 7721 5650

RATING: H3 ATTACHMENT: YES - ODCM 6.1.1 tabl e 6.1.1.1-1, PPM 4.602.A5.6-6 JUSTIFICATION: A and B are incorrect because they both allow the discharge to continue and the actions given are for the SW monito rs and for the flow rate monitor of Rad Waste. D is incorrect because t here is no action allowing the use of a temporary monitor in the place of FDR-RIS-606. C is correct. This is the action given in the ODCM. COMMENTS:

COLUMBIA SRO WRITTEN EXAM RETAKE QUESTION # 25

EXAM KEY 06/2005 Page 25 of 25

EX05007 The plant is shutdown in Mode 5 with a complete core offload in progress. The outside air temperature is 110° F. Ops 2 reports that t he general area temperature in the 471 RB west end (around E-SH-10) has been 105° F for the last 75 minutes. The temperature is due to maintenance on the ventilation system in the local ar ea. All other plant equipment is operating normally.

Which of the following actions is correct for this condition?

A. Immediately suspend movement of irr adiated fuel in the secondary containment, core alterations, and initiate actions to suspend operations with a potential for draining the reactor vessel.

B. The actions required can be delayed for a maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> due to the temperature increase resulti ng from a maintenance issue.

C. Initiate action to restore the area to wit hin the limits of Condition B in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

D. Initiate action to restore the area to wit hin the limits of Condition C in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ANSWER: A QUESTION TYPE: SRO KA # & KA VALUE: 295032EA2.02 - Ab ility to determine and/or interpret the following a they apply to HIGH SECONDARY CONTAINMEN T AREA TEMPERATURE: Equipment Operability 55.43.2 IMP 3.5

REFERENCE:

LCS 1.7.1 and TS 3.6.4.3

SOURCE: NEW QUESTION -

T1, GP2 LO: 9540 RATING: H4 ATTACHMENT: YES - LCS 1.7.1 pages 3, 8, 15, &16 and TS 3.6.4.3 pages 3 JUSTIFICATION: This high temperature for 75 minutes causes SGT div 1 and 2 to be inop under these conditions. The extra allotted 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> does not apply when

Condition C is exceeded or the Condition B is exceeded for maintenance

activities. B is incorrect. C and D are both incorrect because the allotted

time would be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. A is correct per TS 3.6.4.3 COMMENTS: