ML15167A138

2015-04 Draft Written Exams

ML15167A138
Person / Time
Site:	Columbia
Issue date:	05/05/2015
From:	Vincent Gaddy Operations Branch IV
To:
References
Download: ML15167A138 (138)

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COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-01 RHR-P-2A is currently running and injecting into the vessel at rated flow. A relay failure causes RHR-V-4A, the pumps suction valve, to close.

Which of the following is correct?

RHR-P-2A ..

A. continues to operate. To prevent the pumps continued operation, the ARP directs RHR-P-2As control switch be momentarily placed to STOP.

B. continues to operate. To prevent the pumps continued operation, the ARP directs RHR-P-2As control switch be held in STOP and the control power fuses be removed.

C. stops as soon as dual indication on RHR-V-4A occurs. The ARP directs preventing RHR-P-2A from starting. RHR-P-2A cannot be restarted until RHR-V-4A is fully opened.

D. stops when RHR-V-4A is fully closed. The ARP directs preventing RHR-P-2A from starting. RHR-P-2A cannot be restarted until RHR-V-4A is fully opened.

ANSWER: C KA # & KA VALUE: Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures 203000 A2.03 (RO 3.2) (CFR: 41.5) Tier 2 / Group 1

REFERENCE:

SD000198 Page 10; 4.601.A4 6-1 SOURCE: New LO: 11586 LOK / LOD: F/3 HANDOUT: None JUSTIFICATION: A (incorrect): RHR-P-2A will trip when valve is not full open. This is a credible distractor because LPCS-P-1 does not have a suction valve interlock and would keep running even with the suction valve full closed.

B (incorrect): See A C (correct): RHR-P-2A trips as soon as dual indication on RHR-V-4A occurs. The ARP directs disabling the pumps starting capability.

D (incorrect): See A

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-02 During operation of RHR B in Shutdown Cooling, if the control switch for RHR-V-3B, the RHR Heat Exchanger Shell Side Outlet valve, is closed and total system flow drops to 100 gpm less than it was originally, which of the following is correct?

Over the next thirty minutes, RPV temperature will.

A. increase and RPV level will decrease.

B. increase and RPV level will increase.

C. decrease and RPV level will decrease.

D. decrease and RPV level will increase.

ANSWER: B KA # & KA VALUE: Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MORE) controls including: Reactor temperatures (moderator, vessel, flange) 205000 A1.06 (RO 3.7) (CFR 41.5) Tier 2 / Group 1

REFERENCE:

SD000198 Fig 1B; Page 39 SOURCE: NEW LO: 7728 LOK / LOD: H/3 HANDOUT: None JUSTIFICATION: A (incorrect) See B B (correct) Throttling RHR-V-3B closed reduces RHR SDC flow. This causes RPV temperature to increase and RPV level to increase.

C (incorrect) See B D (incorrect) See B

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-03 Which of the following design features help to maintain adequate NPSH for the LPCS pump?

1. LPCS pump elevation in reference to the Suppression Pool

2. LPCS Water Leg Pump keeping piping filled with water

3. LPCS suction strainer size A. 1 and 2 B. 2 and 3 C. 1 and 3 D. 1 only ANSWER: C KA # & KA VALUE: Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following: Adequate pump net positive suction head 209001 K4.06 (2.6) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

FSAR 6.3.2.2.3 SOURCE: NEW LO: 11586a LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: Per the CGS FSAR The LPCS pump is located in the reactor building sufficiently below the water level in the suppression pool to ensure a flooded pump suction and to meet pump NPSH requirements with the containment at atmospheric pressure and post-accident debris entrained on the beds of the suction strainers.

The water Leg pump maintains discharge piping filled to prevent damage from fluid hammer.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-04 Initially HPCS and RCIC are operating following a reactor scram both taking a suction from the Suppression Pool with the following conditions:

RCIC flowrate is 650 GPM HPCS flowrate is 7000 GPM RPV pressure is 850 psig steady RPV level -205 inches and up slow Debris partially plugs the HPCS suction strainer resulting in the following conditions:

RCIC flowrate is 650 GPM HPCS flowrate is 5500 GPM RPV pressure is 850 psig steady RPV level -203 inches and steady Can adequate core cooling be assured under these conditions and why or why not?

A. Yes. RPV level is being maintained above level required for spray cooling with HPCS and RCIC providing sufficient flow to ensure adequate core cooling.

B. Yes. RPV level is above level required for spray cooling and steam cooling with injection will ensure adequate core cooling.

C. No. HPCS is not providing an adequate amount flow for spray cooling, and RPV level is below the level for steam cooling with injection to ensure adequate core cooling.

D. No. HPCS and RCIC are providing sufficient spray flow, but RPV level is below level required to ensure adequate core cooling.

ANSWER: C KA # & KA VALUE: Knowledge of the effect that a loss or malfunction of the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) will have on following: Adequate Core Cooling 209002 K3.03 (RO 3.9) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

PPM 5.0.10 Pages 17 and 18 SOURCE: NEW LO: 8018 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): HPCS or LPCS by themselves must provide greater than 6000 GPM for spray cooling to be effective.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION B (incorrect): RPV level must be greater than -183 inches with injection to ensure adequate core cooling.

C (correct): HPCS is not providing adequate spray flow and RPV level must be greater than -183 inches with injection to ensure adequate core cooling.

D (incorrect): HPCS or LPCS by themselves must provide greater than 6000 GPM for spray cooling to be effective.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-05 Which of the following describes the effect on the High Pressure Core Spray system components if MC-4A loses power just prior to a HPCS low RPV level initiation signal occurring?

HPCS-P-1..

A. starts. HPCS-V-4 (Injection valve) and HPCS-V-12 (Min flow valve) remain closed due to a loss of power. HPCS-P-3 (Water leg pump) loses power and stops.

B. starts. HPCS-V-4 (Injection valve) opens to inject to the vessel. HPCS-V-12 (Min flow valve) opens and closes based on injection flow. HPCS-P-3 (Water leg pump) loses power and stops.

C. does not start. HPCS-V-4 (Injection valve) and HPCS-V-12 (Min flow valve) remain closed due to the loss of power. HPCS-P-3 (Water leg pump) stops due to the loss of power.

D. does not start. HPCS-V-4 (Injection valve) and HPCS-V-12 (Min flow valve) remain closed due to the loss of power. HPCS-P-3 (Water leg pump) continues to operate.

ANSWER: A KA # & KA VALUE: Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS): Electrical power. 209002 K6.01 (RO 3.6) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000174 Page 21 SOURCE: BANK LO: 5431 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (correct): The power supply for the initiation logic is S1-HPCS and the power supply for HPCS-P-1 is SM therefore HPCS-P-1 will start. MC-4A powers all HPCS system valves and HPCS-P-3.

B (incorrect): HPCS-V-4 and HPCS-V-12 loses power and will not open.

C (incorrect): HPCS-P-1 does start.

D (incorrect): HPCS-P-1 does start, valves do not open.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-06 A loss of power to which of the following would prevent SLC-V-4A (Squib Injection valve) from opening?

A. MC-7A B. MC-8A C. MC-7B D. MC-8B ANSWER: C KA # & KA VALUE: Knowledge of electrical power supplies to the following: Explosive valves 211000 K2.02 (RO 3.1) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000172 Page 24 SOURCE: NEW LO: None LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect): See C B (incorrect): See C C (correct): SLC-V-4A is powered from MC-7B. MC-8B is power supply to SLC-V-4B.

D (incorrect): See C

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-07 Which of the following explains how a trip of the Reactor Protection System affects the following components?

Scram Air Pilot Valves Backup Scram Valves A. Energize De-Energize B. Energize Energize C. De-energize De-energize D. De-energize Energize ANSWER: D KA # & KA VALUE: Knowledge of the physical connections and/or cause effect relationships between REACTOR PROTECTION SYSTEM and the following: Control rod drive hydraulic system. 212000 K1.06 (RO 3.5) (CFR 41.6) Tier 2 / Group 1

REFERENCE:

SD000161 Page 22 SOURCE: NEW LO: 7682 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect): See D B (incorrect): See D C (incorrect): See D D (correct): RPS de-energizes the scram air pilot valves and energizes the backup scram valves

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-08 Which of the following statements is true concerning the Intermediate Range Monitor (IRM) detector drive unit controls?

A. The IRM "DRIVE IN" pushbutton has a seal in feature and will drive in a detector until it is de-selected or it is fully inserted.

B. Pushing the "POWER ON" pushbutton turns on the IN or OUT lights only for the IRM that have been selected.

C. The IRM "SELECT" pushbuttons are latched in the "DE-SELECT" position and are backlight when de-selected.

D. The IRM "DRIVE OUT" pushbutton has a seal in feature and will withdraw a detector until it is de-selected or it is fully withdrawn.

ANSWER: A KA # & KA VALUE: Intermediate Range Monitor System: Knowledge of the purpose and function of major system components and controls. 215003 G2.1.28 (RO 4.1) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000138 Page 12 and 13; Simulator SOURCE: Bank - Modified LO00652 LO: 5456 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: The power on P/B turns on lights for all IRMs, selected or not. The select P/B latches in the select position and the lights are lit when selected. The drive out P/B does not have a seal in feature - it needs to be held depressed. The drive in P/B does have a seal in feature and will drive a detector until full in or de-selected.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-09 When performing a normal startup, under which of the following conditions would the RO and Reactivity Manager SRO declare the reactor critical?

A. SRM count rate 4x10^3 cps and steady Reactor Period 100 seconds and increasing No rod motion B. SRM count rate 6x10^3 cps and up slow Reactor Period 100 seconds and increasing Rod 27-26 being pulled from position 12 to position 14 C. SRM count rate 5x10^3 cps and up slow Reactor Period 100 seconds and decreasing Rod 27-26 being pulled from position 12 to position 14 D. SRM count rate 3x10^3 cps and up slow Reactor Period 100 seconds and steady No rod motion ANSWER: D KA # & KA VALUE: Ability to manually operate and/or monitor in the control room: SRM count rate and period 215004 A4.01 (3.9) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

PPM 3.1.2, Attachment 7.3, Note N6 SOURCE: NEW LO: 6651 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): Reactor period is increasing to infinity and power is steady.

B (incorrect): Reactor period is increasing to infinity and rod motion is occurring.

C (incorrect): Rod motion is occurring.

D (correct): Note N6 of PPM 3.1.2: Criticality usually occurs in the source range between 1x10^3 and 1x10^4 cps. For purposes of this procedure, criticality shall be identified by increasing neutron level, a constant steady period and no simultaneous control rod motion.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-10 Which of the following is the minimum disagreement between APRM flow comparator channels that will produce a rod withdrawal block signal?

A. 5%

B. 10%

C. 15%

D. 20%

ANSWER: B KA # & KA VALUE: Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including: Maximum disagreement between flow comparator channels 215005 A3.06 (RO 3.0) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000149 page 9 SOURCE: NEW LO: 11654 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): See B B (correct): A 10% difference will produce a rod block signal.

C (incorrect): See B D (incorrect): See B

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-11 The Reactor Core Isolation Cooling (RCIC) system has been manually started for vessel pressure control and is operating CST (Condensate Storage Tank) to CST.

Select the statement below that identifies the RCIC system response if CST level decreases to 1' 6".

A. The CST suction valve will close causing a RCIC turbine trip on low pump suction pressure.

B. RCIC will be operating with a Suppression Pool to Suppression Pool flow path through the full flow test line.

C. The RCIC pump suction valves swap resulting in RCIC taking a suction from the Suppression Pool and transferring water to the CSTs.

D. RCIC will take a suction from the Suppression Pool and discharge to the Suppression Pool through the RCIC minimum flow valve.

ANSWER: D KA # & KA VALUE: Knowledge of the physical connections and/or cause - effect relationships between REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following:

Condensate storage and transfer system 217000 K1.01 (RO 3.5) (CFR 41.7) Tier 2 /

Group 1

REFERENCE:

SD000180 pages 18, 19, 20, 21 and 22 SOURCE: BANK (LO00803)

LO: 5724 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): A suction flow path is maintained throughout the valve swap as one does not start to close until the other is full open.

B (incorrect): The full flow test line discharges to the CSTs.

C (incorrect): The valves in the full flow test line close when the Suppression Pool suction valve is open.

D (correct) At a CST level of 110 the RCIC pump suction valves swap. When the Suppression Pool suction valve is full open the valves in the full flow test line to the CST close.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-12 An event occurs that results in a loss of DC bus S1-2.

Which of the following describes the effect this loss has on the ADS SRVs?

A. Only the A solenoids lose power.

B. Only the B solenoids lose power.

C. The A and B solenoids lose power.

D. The A and C solenoids lose power.

ANSWER: B KA # & KA VALUE: Knowledge of electrical power supplies to the following: ADS logic 218000 K2.01 (RO 3.1) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000128 Pages 28 and 29; SD000188 Figure 1.

SOURCE: NEW LO: 5077 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect): The A solenoid is powered from DP-S1-1A and remain powered.

B (correct): A loss of bus S1-2 causes DP-S1-2A to lose power causing a loss of power to the B solenoids.

C (incorrect): The A solenoid is powered from DP-S1-1A and remain powered.

D (incorrect): The A and the C solenoids are powered from DP-S1-1A and remain powered.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-13 With Columbia operating in MODE 1, ROA-FN-1A (Reactor Building HVAC Supply Fan), and REA-FN-1A (Reactor Building HVAC Exhaust Fan), are running and REA-RIS-609C (Reactor Building Exhaust Plenum Radiation Monitor) is INOP (switch is out of operate).

Which of the following explains the effect if REA-RIS-609D receives a Hi-Hi trip signal?

A. ROA-FN-1A and REA-FN-1A trip. Only the A Standby Gas Treatment train initiates.

B. ROA-FN-1A and REA-FN-1A continue to operate. Only the B Standby Gas Treatment train initiates.

C. ROA-FN-1A and REA-FN-1A trip. Both the A and B Standby Gas Treatment trains initiate.

D. ROA-FN-1A and REA-FN-1A continue to operate. Neither Standby Gas Treatment train initiates.

ANSWER: C KA # & KA VALUE: Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following: Plant ventilation 223002 K3.11 (RO 2.8) (CFR 41.7)

Tier 2 / Group 1

REFERENCE:

SD000173 page 17, 23 and 24. SD000147 Page 20 & 21, EWD-108E-005.

SOURCE: NEW LO: 5598 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): Both trains of SGT will start. (See C for Explanation.)

B (incorrect): ROA-FN-1A and REA-FN-1A trip. (See C for Explanation.)

C (correct): A trip of Group 3 logic for RB Exhaust Plenum radiation monitors is arranged such that two monitors must trip to actuate the inboard or outboard isolation (A&B OR C&D). INOP, Downscale, or Upscale will all cause a Z isolation signal to be sent. This condition for C & D initiates closure of the corresponding inboard valves and initiates startup of SGT Train B Lead Fan and Train A Lag Fan.

Both trains of SGT will start.

D (incorrect): Both trains of SGT will start and ROA-FN-1A and REA-FN-1A trip.

(See C for Explanation.)

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-14 During potential subsequent actuations of an SRV, which of the following correctly describes the design feature that limits piping stress and containment loading?

A. The X shaped quencher that is installed at the end of each SRV tailpipe.

B. The Suppression Pool minimum water level of 192.

C. The two vacuum breakers that are installed in each SRV tailpipe that are located below the Drywell floor.

D. The two vacuum breakers that are installed in each SRV tailpipe that are located above the Drywell floor.

ANSWER: D KA # & KA VALUE: Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or interlocks which provide for the following: Prevents siphoning of water into SRV discharge piping and limits loads on subsequent actuation of SRV's 239002 K4.03 (RO 3.1)

(CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000128 page 14 SOURCE: NEW LO: 11690; 11745 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect): See D.

B (incorrect): See D.

C (incorrect): See D.

D (correct): There are two vacuum breakers installed above the drywell floor that are designed to limit containment loading due to multiple actuations of an SRV.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-15 Following a scram, RPV water level is being maintained at +36 inches using the Startup Flow Control Valves (RFW-FCV-10A/B) in Automatic.

A sudden and immediate loss of the output signal from the Startup RPV Level Controller (RFW-LIC-620) occurs.

What will happen to RPV level as a result of this failure?

RPV Level will .

A. remain at +36 inches based on the last known RPV level setpoint received by the RFWLC PLC modulating RFW-FCV-10A/B.

B. increase because the RFWLC PLC loses its level input signal and fully opens RFW-FCV-10A/B.

C. decrease because the Automatic level signal setpoint will fail downscale causing RFW-FCV-10A/B to close fully.

D. decrease to +18 inches based on setpoint setdown and modulate RFW-FCV-10A/B to maintain +18 inches.

ANSWER: A KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM: GEMAC/Foxboro/Bailey controller operation 259002 K5.01 (3.1) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000157 page 9 SOURCE: NEW LO: 11704e LOK / LOD: F/4 HANDOUT: NONE JUSTIFICATION: A (correct): S/U valve demand continues from the PLC based on the last known setpoint input from RFW-LIC-620.

B (incorrect): PLC receives level signal directly from the level detectors.

C (incorrect): S/U valve demand continues from the PLC based on the last known setpoint input from RFW-LIC-620.

D (incorrect): Setpoint setdown is a function of the Master Controller and the RFPs speed.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-16 SGT A has been running for 20 minutes due to an automatic initiation signal being present.

If a lockout on SM-7 were to occur, which of the following explains the response of the SGT Train A system?

A. The lead fan loses power. The lag fan then receives a start signal, starts, but trips on low flow. The lag fan continues to start and then trip on low flow.

B. The lead fan trips on low flow. The lag fan then receives a start signal, starts, and maintains Secondary Containment pressure.

C. The lead fan loses power. The lag fan then receives a start signal, starts, but trips on low flow. SGT remains in this condition.

D. The lead fan continue to operate and maintains Secondary Containment pressure.

ANSWER: A KA # & KA VALUE: Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM : A.C. electrical distribution 261000 K6.01 (RO 2.9) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000144 Pages 9, 13 and 14 SOURCE: NEW LO: 5825 LOK / LOD: H/4 HANDOUT: NONE JUSTIFICATION: A (correct): MC-7B-B (powered from SM-7) powers the lead fan (SGT-FN-1A1) and the lead fans valves. When power is lost the lead fan loses power and stops.

The lead fans discharge valve also loses power and remains in the open position.

The lag fan then gets a start signal on low flow and starts. The lag fans discharge valve is interlocked from opening due to the opposite fans valve not being closed.

This causes the lag fan to trip on low flow. With an initiation signal still present, the lead fan gets a start signal due to low flow on lag fan but has no power. The lag fan then gets another start signal due to low flow on the lead fan and starts. It again trips on low flow due to the discharge valve not opening. This condition continues until the initiation logic is reset.

B (incorrect): See A.

C (incorrect): See A.

D (incorrect): See A.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-17 A charcoal filter temperature high alarm has annunciated in the control room. OPS 2 was dispatched to investigate and has reported evidence of a fire has been detected in the SGT charcoal. The CRS directs actions per ABN-SGT-TEMP/RAD.

Which of the following is correct concerning the above?

To suppress the fire direction is given to.

A. place the Emergency Deluge Spray valves control switch that are located on the local control panel in FLOOD.

B. place the Emergency Deluge Spray valves control switch that are located on the Control Room backpanel in FLOOD.

C. place and hold the Emergency Deluge Spray valves control switch that are located on the local control panel in FLOOD.

D. place and hold the Emergency Deluge Spray valves control switch that are located on the Control Room backpanel in FLOOD.

ANSWER: D KA # & KA VALUE: Standby Gas Treatment System: Ability to locate and operate components, including local controls 261000 G2.1.30 (RO 4.4) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

ABN-SGT-TEMP/RAD Page 5 SOURCE: NEW LO: 5826 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): Switches are in the Control Room.

B (incorrect): Switches have to be held in the flood position or the switch returns to close and the valves close.

C (incorrect): See A.

D (correct): The Emergency Deluge Spray valves are controlled from the control room and have to be placed and HELD in the Flood position as they spring return to close which causes the valves to close. The valves only open in the flood position and the signal does not seal in.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-18 While operating at full power, if control power is lost to CB-7/1, which of the following is correct?

Based on the current breakers configuration, CB-7/1..

A. cannot be operated locally or remotely until control power is restored.

B. can only be manually opened at the breaker, as the breakers opening springs are charged via a DC motor immediately following closure of the breaker.

C. can only be manually closed at the breaker, as the breakers closing springs are charged due to the opening action of the breaker mechanism.

D. can be manually opened and manually closed once, as the breakers opening and closing springs were charged when the breaker was closed.

ANSWER: D KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to A.C. ELECTRICAL DISTRIBUTION: Breaker control 262001 K5.02 (2.6) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000182 Page 29, 30, 31and 32 SOURCE: NEW LO: 5066 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: When CB-7/1 closes the opening springs are charged by the closing of the breaker mechanism. The closing springs are charged via a DC motor. With control power lost, only one opening and one closing operation of the breaker is available without an operator manually recharging springs.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-19 The station is preparing for startup following a refueling outage. The electric plant is in a normal lineup for startup with the exception of SM-8, which is on E-TR-B (Backup Transformer) following surveillance testing. Work Control has requested the start of CW-P-1C for PMT. CW-P-1A and CW-P-1B are running.

What will be the impact of starting CW-P-1C in the current plant configuration?

Starting CW-P-1C will cause a A. degraded frequency condition on E-TR-B (Backup Transformer).

B. degraded frequency condition on E-TR-S (Startup Transformer).

C. degraded voltage condition on E-TR-B (Backup Transformer).

D. degraded voltage condition on E-TR-S (Startup Transformer).

ANSWER: D KA # & KA VALUE: Ability to predict and/or monitor changes in parameters associated with operating the A.C. ELECTRICAL DISTRIBUTION controls including: Bus voltage 262001 A1.03 (RO 2.9) (CFR 41.5) Tier 2 / Group 1

REFERENCE:

SOP-CW-START Page 7; LER 84-079 SOURCE: Last NRC Exam Q#46 LO: 5050 LOK / LOD: F/3 HANDOUT: None JUSTIFICATION: A (incorrect): See D.

B (incorrect): See D.

C (incorrect): See D.

D (correct): Starting a third CW Pump while E-TR-S is supplying house loads will cause an under voltage or degraded voltage on E-TR-S and the electrical buses it is powering. CW-P-1C is powered from SM-3. SM-3 normally powers SM-8. When SM-8 is powered from TR-B, breaker 8-3 is open. As a result, starting the third CW Pump will not impact E-TR-B. Frequency is determined by the 230kv grid, and is relatively constant. Starting the third pump will have minimal impact.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-20 With the plant operating at 100% power, the E-IN-1 inverter voltage lowers to 90% of the normal voltage.

To what source will the static switch associated with E-IN-1 transfer, and what procedure will be prioritized?

The static switch associated with E-IN-1 will transfer to the A. bypass AC source (MC-7A). Enter ABN-POWER.

B. bypass AC source (MC-7A). Enter ABN-ELEC-INV.

C. alternate AC source (MC-7F). Enter ABN-POWER.

D. alternate AC source (MC-7F). Enter ABN-ELEC-INV.

ANSWER: D KA # & KA VALUE: Ability to (a) predict the impacts of the following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Under voltage 262002 A2.01 (RO 2.6) (CFR 41.5) Tier 2 / Group 1

REFERENCE:

SD000194 (UPS) Rev 10 Page 8; ABN-ELEC-INV Rev 12 Page 3 SOURCE: Last NRC Exam Q#47 - Modified LO: LO-5896, LO-5891 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): The bypass source requires a manual transfer, and is not bumpless.

This action would result in a loss of feedwater heating and prioritization of ABN-POWER.

B (incorrect): See A.

C (incorrect): See D.

D (correct): The undervoltage condition causes a bumpless transfer to the alternate power supply. ABN-ELEC-INV will be used to determine the fault with IN-1. If the transfer werent bumpless, such as when load is transferred to the bypass source, the feedwater heater controllers would lose power and cause a loss of feedwater heating.

This would result in the prioritization of ABN-POWER to maintain reactor power LE 3486 MWt.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-21 Columbia is operating in Mode 1 when a ground alarm is received on Battery S1-2. CRO2 reports that the S1-2 ground detection meter indicates 0K (ohms) to the POS (positive) side.

Which of the following is correct for this indication?

A. The annunciator is spurious; the meter indicates no ground on S1-2.

B. The annunciator is valid; the meter indicates a severe ground on S1-2.

C. The annunciator is spurious; the Ground Test Switch has been placed in POS (positive).

D. The annunciator is valid; the Ground Test Switch has been placed in POS (positive).

ANSWER: B KA # & KA VALUE: Ability to monitor automatic operations of the D.C. ELECTRICAL DISTRIBUTION including: Meters, dials, recorders, alarms, and indicating lights.

263000 A3.01 (RO 3.2) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000188 Page 12 and Figure 8, 8A and 8BA SOURCE: BANK - MODIFIED LO001275 LO: 5261 LOK / LOD: H/2 HANDOUT: None JUSTIFICATION: A (incorrect) A value of infinity would indicate a spurious alarm and no ground.

B (correct) A ground alarm is received at 10 0K and a reading of zero indicates a severe ground.

C (incorrect) Placing the switch in POS would give a ground alarm but the indication would be +10 0K.

D (incorrect) Placing the switch in POS would give a ground alarm but the indication would be +10 0K.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-22 The plant was operating at 98% power when a loss of all feedwater occurred. All plant equipment initiated as designed. RPV level is now -20" and up fast.

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

A. The HPCS DG Generator Differential Relay trip is bypassed.

B. The HPCS DG High Crankcase Pressure trip is bypassed.

C. RCIC-V-1 will trip at +54.5 inches and has to be reset from the Control Room before RCIC can be restarted.

D. RCIC-V-1 will trip at +54.5 inches and has to be reset at the RCIC Turbine before RCIC can be restarted.

ANSWER: B KA # & KA VALUE: Ability to predict and/or monitor changes in parameters associated with operating the EMERGENCY GENERATORS (DIESEL/JET) controls including: Crank case temperature and pressure 264000 A1.04 (2.6) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000200 Page 32; SD000180 Page 19 and 32 SOURCE: BANK LO01244 (Modified slightly)

LO: 5323 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A loss of feedwater response is for HPCS and RCIC to initiate at -50. The HPCS DG High Crankcase Pressure trip is bypassed on this type of start. The Generator differential relay trip is not bypassed. RCIC-V-1 does not trip at +54.4, RCIC-V-45 closes which closes RCIC-V-13. If RCIC-V-1 did trip it depends on the initiating event as to where the valve could be reset from, the Control Room or locally.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-23 Columbia is operating at power. The DG-1 monthly operability surveillance is in progress. SM-1 has been transferred to the Startup transformer. The Engine Speed Selector Switch is in the IDLE position with speed at approximately 400 rpm.

If a DBA LOCA were to occur, which of the following describes the response of DG-1?

DG-1 ..

A. automatically ramps to rated speed and its output breaker remains open.

B. automatically ramped to rated speed and its output breaker closes.

C. remains at idle speed and its output breaker remains open.

D. remains at idle speed and its output breaker closes.

ANSWER: C KA # & KA VALUE: Ability to manually operate and/or monitor in the control room: Transfer of emergency control between manual and automatic 264000 A4.03 (3.2) (CFR 41.7)

Tier 2 / Group 1

REFERENCE:

SD000200 Page 47 SOURCE: BANK - MODIFIED LX00807 LO: 5321 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: The question gives that a DBA LOCA occur which generate a high drywell pressure signal and a low RPV level signal. In idle, these signals do not automatically make DG-1 go to rated. In idle only an undervoltage signal cause the DG to automatically ramp to rated speed. The output breaker would not close in on the bus without an undervoltage signal.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-24 With Columbia operating at full power, CIA pressure begins to lower. ABN-CIA is entered and OPS-2 reports there is a leak on the supply side of the regulator from the normal supply.

Per ABN-CIA as CIA pressure approaches 135 psig, which of the following can be lined up to supply CIA pressure to ALL components?

A. CAS B. Service Air C. CN D. Nitrogen Bottle Station ANSWER: A KA # & KA VALUE: Knowledge of the connections and / or cause effect relationships between INSTRUMENT AIR SYSTEM and the following: Containment air 300000 K1.03 (2.8) (CFR 41.2) Tier 2 / Group 1

REFERENCE:

ABN-CIA page 4, SD000156 figure 1 SOURCE: NEW LO: NONE LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (correct): ABN-CIA directs lining up CAS per SOP-CIA-OPS as pressure approaches 135 psig.

B (incorrect): Service Air does not have a connection to CIA system.

C (incorrect): CN is the normal supply to CIA which is the leak location.

D (incorrect): Nitrogen Bottle Station cannot supply the portion of the system going to the MSIVs.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-25 The power supply to the A Control Air Compressor (CAS-C-1A) is:

A. MC-2P B. MC-6C C. MC-7A D. MC-8A ANSWER: C KA # & KA VALUE: Knowledge of electrical power supplies to the following: Instrument air compressor 300000 K2.01 (RO 2.8) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000205 Page 23 SOURCE: NEW LO: 5881 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect): MC-2P is the power supply to CAS-C-1C.

B (incorrect): MC-6C is the power supply to the Service Air compressor.

C (correct): CAS-C-1A is powered from MC-7A.

D (incorrect): MC-8A is the power supply to CAS-C-1B.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-26 With Columbia operating at power which of the following would cause the breaker of an operating Reactor Closed Cooling Water (RCC) pump to open?

1. RPV water level lowering to -55

2. RPV water level lowering to -135

3. Drywell pressure rising to 2 psig

4. A loss of power to SL-81 A. Only 2 and 3 B. Only 2, 3 and 4 C. Only 1, 2, and 3 D. Only 1, 2, and 4 ANSWER: C KA # & KA VALUE: Ability to monitor automatic operations of the CCWS including: Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS 400000 A3.01 (RO 3.0) (CFR 41.7) Tier 2 / Group 1

REFERENCE:

SD000196 Page 15 SOURCE: NEW LO: 5707 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): Choice one would also cause breaker to open.

B (incorrect): Choice 4 does not cause the breaker to open, just the pump to stop.

C (correct): An operating RCC pump breaker opens on and F or an A signal (-50 or 1.68 psig). At rated power, two RCC pumps are running and at least one of them will be powered from SL-81.

D (incorrect): See B.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-27 During maintenance in the control room, electrical disconnect E-DISC-USPP/7 (Power Supply to H13-P616 on E-PP-US) is inadvertently OPENED by an electrician, de-energizing the RDCS analyzer.

CRO1 received the following alarms on P601:

• Rod Drive Control System INOP

• RSCS INOP

• RPIS or DMM INOP How does this affect control rod movement?

A. No Rod Block annunciator will be generated so insertion or withdrawal of rods may occur.

B. No Rod Block annunciator will be generated but a lockup of RDCS would prevent insertion or withdrawal of rods.

C. A Rod Block annunciator will be generated preventing insertion or withdrawal of rods.

D. A Rod Block annunciator will be generated but will not be by RMCS enforced so insertion or withdrawal of rods may occur.

ANSWER: B KA # & KA VALUE: Knowledge of the effect that a loss or malfunction of the REACTOR MANUAL CONTROL SYSTEM will have on following: Rod block monitor 201002 K3.02 (2.9) (CFR41.7) Tier 2 / Group 2

REFERENCE:

4.603.A7 (603.A7 Annunciator Panel Alarms) pages 13-14, 23-27, and 48, SD000148 page 20 SOURCE: New LO: 5795 LOK / LOD: H/4 HANDOUT: NONE JUSTIFICATION: A (incorrect): Due to a loss of RDCS analyzer unit portion of RMCS there is no ROD block generated, however this results in a loss of all normal control rod movement.

B (correct): Due to a loss of RDCS analyzer unit portion of RMCS there is no ROD block generated, however this results in a loss of all normal control rod movement.

C (incorrect): The failure is in the RDCS portion of the RMCS system instead of the RPIS portion so there are no Rod Blocks Generated. The RPIS or DMM INOP alarm is from the DMM portion.

D (incorrect): The failure is in the RDCS portion of the RMCS system instead of the RPIS portion so there are no Rod Blocks Generated. The RPIS or DMM INOP alarm is from the DMM portion.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-28 During a startup with Reactor Power at 8%, the Reactor Operator selects rod 26-15 to withdraw it from its full in position. The control rod was not properly coupled to its CRDM and is stuck in the full in position.

How will the RWM system enforce the correct operating sequence?

A. A withdraw block will not be generated unless the CRDM is moved past its withdraw limit for that group.

B. A withdraw block will not be generated until there is greater than 2 notches difference between the CRDM and actual control rod blade position.

C. A withdraw block will be generated immediately upon selecting the uncoupled rod.

D. A withdraw block cannot be generated for the stuck rod as long as it remains fully inserted.

ANSWER: A KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD AND DRIVE MECHANISM: Rod sequence patterns 201003 K5.04 (3.1) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

SD000154 pages 8 and 9 SOURCE: NEW LO: NONE LOK / LOD: F/4 HANDOUT: NONE JUSTIFICATION: A Withdraw Block is imposed when power level is below the LPSP of 22% and a withdraw error is present, an insert block error exists and a rod not identified as an insert error is selected. A Withdraw Error occurs when a rod in the currently latched group or any lower group is withdrawn past the withdraw limit for the group, or if a rod contained in a group higher than the one that is latched is withdrawn past the insert limit for the higher group.

A (correct): With the Control Rod uncoupled the rod position indicating system only knows the position of the CRDM. The CRDM position will generate a Withdraw Block.

B (incorrect): The Rod Position Indicating System cannot tell the position of the Control Rod Blade.

C (incorrect): When the Control Rod is selected the RWM will see its position as full in and will not generate a Withdraw Block.

D (incorrect): The CRDM can be positioned in such a way that it receives a Rod Withdraw Block.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-29 While operating at 90% power a catastrophic resin intrusion occurs during a backwash and precoat of the RWCU filtered demineralizers.

Which of the following are indications of this event?

The Reactor water will have and Reactor power will have .

A. high conductivity and high pH; increased B. high conductivity and low pH; decreased C. low conductivity and high pH; increased D. low conductivity and low pH; decreased ANSWER: B KA # & KA VALUE: Knowledge of the physical connections and/or cause/effect relationships between REACTOR WATER CLEANUP SYSTEM and the following: Reactor water quality 204000 K1.10 (3.3) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

SD000190 page 17 and 18 SOURCE: New LO: 5043 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: High conductivity and low pH are typical of a catastrophic resin intrusion which will also cause Reactor power to decrease.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-30 A TIP trace in the core is in progress. The drive control unit is in the Manual Forward Mode and the detector is being driven into the core.

If the reactor were to scram and RPV level were to drop to -55 while the detector is being driven into the core, which of the following is correct?

A. The TIP detector automatically stops inserting and withdraws from the core. When the detector reaches the "in-shield" position the ball valve will automatically close.

B. The TIP detector automatically stops inserting and withdraws from the core. The ball valve must be manually closed when the detector reaches the "in-shield" position.

C. The TIP detector automatically stops inserting . The MANUAL VALVE CONTROL switch must be placed in MANUAL REVERSE position to withdraw the TIP detector.

When the detector reaches the "in-shield" position the ball valve will automatically close.

D. The TIP detector automatically stops inserting . The MANUAL VALVE CONTROL switch must be placed in MANUAL REVERSE position to withdraw the TIP detector.

The ball valve must be manually closed when the detector reaches the "in-shield" position.

ANSWER: A KA # & KA VALUE: Knowledge of TRAVERSING IN-CORE PROBE design feature(s) and/or interlocks which provide for the following: Primary containment isolation: Mark-I & II (Not-BWR1) 215001 K4.01 (RO 3.4) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

SD000155 pages 15 and 16 SOURCE: NEW LO: 6989 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (correct): At -50, TIP receives an NSSS Group 4 isolation signal. The detector stops inserting and immediately withdraws from the core. The ball valve closes when the detector is in shield.

B (incorrect): The ball valve automatically closes.

C (incorrect): The detector automatically withdraws from the core.

D (incorrect): See C.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-31 Columbia has experienced a series of events. The Emergency Operating Procedures (EOPs) have been entered and current plant conditions are as follows:

- RPV water level -155" and down slow

- RPV pressure 180 psig and down slow

- Wetwell temperature 110°F and up slow

- RHR loop "A" is injecting into the RPV with flow maximized

- RHR loop "B" is in Suppression Pool cooling with cooling maximized

- All other injection sources are unavailable Which of the following statements best describes actions that need to be taken given the current plant conditions?

A. When RPV level is LE -205", Emergency Depressurize the RPV.

B. When RPV level cannot be maintained above -161", Emergency Depressurize the RPV.

C. RHR loop "A" should be removed from injection and placed into Suppression Pool cooling.

D. RHR loop "B" should be removed from suppression pool cooling and injected into the RPV.

ANSWER: D KA # & KA VALUE: RHR/LPCI: Torus/Suppression Pool Cooling Mode: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. 219000 G2.1.7 (4.4) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

PPM 5.0.10 Page 114 SOURCE: NEW LO: 8304 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: An ED is not required until RPV level cannot be restore and maintained GT -183.

With RPV pressure LT 220 psig, RHR injection is available into the RPV. With RPV level at -155 and trending down, the determination should be made to remove RHR-B from SP cooling and inject with it even though SP temperature is 110°F.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-32 Columbia is operating at rated power when a spurious trip of CB 7/1 occurs and all systems operate as designed.

Which of the following is correct for the above event?

A. Primary Containment pressure rises. The reactor will scram due to high drywell pressure.

B. Primary Containment pressure rises but stabilizes below the high drywell pressure scram setpoint.

C. The drywell cooling fans will automatically restart when TR-B closes in on SM-7 to limit the rise in Primary Containment temperature/pressure.

D. The immediate operator actions of ABN-ELEC-SM1/SM7 are designed to prevent a reactor scram on high drywell pressure.

ANSWER: B KA # & KA VALUE: Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES: Drywell cooling.

223001 K6.01 (RO 3.6) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

CGS Simulator; SD000127 Pages 11, 12, 31 SOURCE: NEW LO: 11740; 11746 LOK / LOD: H/3 HANDOUT: None JUSTIFICATION: A (incorrect) A loss of SM-7 causes a loss of half the containment cooling fans. This causes primary containment pressures to rise. Drywell pressure rises but does not cause a reactor scram.

B. (correct) A loss of SM-7 causes a loss of half the containment cooling fans. This causes primary containment pressures to rise. Drywell pressure rises but stabilizes at approximately 0.9 psig and does not cause a reactor scram.

C (incorrect) Drywell cooling fans do not automatically restart when SM-7 is repowered from TR-B.

D (incorrect) While ABN-ELEC-SM1/SM7 is entered, there are no immediate operator actions to perform.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-33 Which of the following correctly identifies the power supplies to the RHR pumps?

A. RHR-P-2A - SM-8; RHR-P-2B - SM-7; RHR-P-2C - SM-8 B. RHR-P-2A - SM-7; RHR-P-2B - SM-8; RHR-P-2C - SM-8 C. RHR-P-2A - SM-7; RHR-P-2B - SM-8; RHR-P-2C - SM-7 D. RHR-P-2A - SM-7; RHR-P-2B - SM-7; RHR-P-2C - SM-8 ANSWER: B KA # & KA VALUE: Knowledge of electrical power supplies to the following: RHR pumps 233000 K2.02 (RO 2.8) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

SD000182 Page 120 SOURCE: NEW LO: 11805 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect): See B.

B (correct): RHR-P-2A is powered from SM-7; RHR-P-2B and RHR-P-2C are powered from SM-8.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-34 While monitoring indications on H13-P602 at 100% power, CRO-1 observes the Main Steam Line Low Power drain valves, MS-V-69, MS-V-156, and MD-V-73, all indicate closed.

This is...

A. a normal indication because they are manually closed above 50% rated steam flow.

B. an abnormal indication because they should have automatically opened above 30% rated steam flow.

C. a normal indication because they automatically close above 50% rated steam flow.

D. an abnormal indication because they should have been manually opened at 30% rated steam flow.

ANSWER: C KA # & KA VALUE: Ability to monitor automatic operations of the MAIN AND REHEAT STEAM SYSTEM including: Opening and closing of drain valves as turbine load changes 239001 A3.02 (RO 2.9) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

SD000128 Page 18 SOURCE: NEW LO: 5534 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): The valves auto close above 50% rated steam flow.

B (incorrect): At this power level the valves should be closed.

C (correct): The MSL Low power drain valves (MS-V-156, 69 & MD-V-73) automatically close when main steam flow exceeds 50% of rated steam flow.

D (incorrect): See B and C.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-35 Columbia was operating at rated power, when a failure of the FWLC system caused a RPV level to increase which resulted in a reactor scram.

ONLY the following actions have been taken:

The Mode Switch has been placed in SHUTDOWN The SRM's and IRM's have been inserted The recorders on H13-P603 have been switched over to monitor IRM power level Which of the following is correct?

Based on the above, to reset RFW-P-1A, the Turbine Emerg Trip/Reset switch.....

A. must be held in RESET to allow both the HP and the LP Stop valves to open.

B. being placed in RESET will not open the HP or the LP Stop valves.

C. is placed in RESET. Both the HP and the LP Stop valves open. When both valves are full open the switch is placed in NORM.

D. is placed in RESET. Both the HP and LP Stop valves open. When both valves show dual indication the switch is placed in NORM.

ANSWER: B KA # & KA VALUE: Knowledge of the physical connections and/or cause - effect relationships between REACTOR FEEDWATER SYSTEM and the following:

Reactor water level 259001 K1.09 (RO 3.8) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

SD000151 page 27; SOP-RFT-RESTART-QC step 2.1.2 SOURCE: BANK - LO02134 LO: 5753 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect): See B.

B (correct): A RPV level of 54.5 causes a RFT trip and a reactor scram. At least two of the three +54.5" seal in's have to reset to allow the HP and LP stops to open on a turbine reset. These P/Bs have not been reset/depressed and as such the HP nor the LP stops will open.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-36 During a discharge operation from Radwaste, a high radiation condition was sensed by installed plant equipment which caused valves to close and terminated the discharge.

Which of the following evolutions could have resulted in the above plant action?

A. Discharging Floor Drain Sample Tank water from FDR-TK-9 directly to the Circulating Water blow down system.

B. Discharging Waste Sample Tank water from EDR-TK-4A directly to the Condensate Storage Tanks.

C. Discharging RWCU Phase Separator water from RWCU-TK-104A directly to the Condensate Storage Tanks.

D. Discharging Condensate Phase Separator water from CPR-TK-92A directly to the Circulating Water blow down system.

ANSWER: A KA # & KA VALUE: Ability to predict and/or monitor changes in parameters associated with operating the RADWASTE controls including: Radiation level 268000 A1.01 (2.7) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

SD000136 Page 10 and figures 1, 2, 9, and 10 SOURCE: NEW LO: 5670 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: A (correct) FDR-TK-9 can be discharged to the CW Blowdown line and has an installed radiation monitor installed.

B (incorrect) EDR-TK-4A may be discharged to the Condensate Storage tank but there is no installed radiation monitor.

C (incorrect) RWCU-TK-104A discharge goes to EDR or FDR collection tanks and not directly to the CST.

D (incorrect) CPR-TK-92A discharge goes to the EDR-TK-2 and not directly to the Circulating Water blow down system.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-37 Alarms are received on FCP-1 and FCP-3. During investigation CRO2 notes that fire main pressure is 115 psig and trending down slow.

Which of the following fire pumps should also have auto started?

A. Only FP-P-2A.

B. FP-P-2A and FP-P-2B.

C. Only FP-P-1.

D. FP-P-2A, FP-P-2B, and FP-P-1.

ANSWER: A KA # & KA VALUE: Ability to manually operate and/or monitor in the control room: Fire main pressure 286000 A4.04 (RO 2.8) (CFR 41.7) Tier 2 / Group 2

REFERENCE:

SD000177 Page 27, 28 SOURCE: NEW LO: 5377 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (correct) FP-P-2A starts at 120 psig and should have auto started.

B (incorrect) FP-P-2B starts at 110 psig and should not have auto started.

C (incorrect) FP-P-1 starts at 110 psig and should not be running at this pressure.

D (incorrect) See B and C

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-38 REMOTE INTAKE DIV 1 RAD HIGH and HI-HI alarms have annunciated and WOA-RIS-31A and 32A are confirmed to be above the HI-HI alarm setpoints. WOA-RIS-31B and 32B are indicating higher than normal but are not in alarm.

1. WEA-FN-51, Toilet/Kitchen Exhaust Fan stops

2. WMA-FN-51A, Control Room Recirc Fan starts

3. WMA-FN-54A, Emergency Filter Unit Fan starts

4. WOA-V-51C, Normal Supply Plenum Isolation valve closes Given the above list, which of the following identifies automatic actions that occur due to the high radiation condition and what procedure is entered?

A. Only 1; ABN-RAD-CR B. Only 1 and 4; ABN-RAD-HIGH C. Only 2 and 3; ABN-RAD-HIGH D. 1, 2, 3, and 4; ABN-RAD-CR ANSWER: A KA # & KA VALUE: Ability to (a) predict the impacts of the following on the CONTROL ROOM HVAC; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Extreme environmental conditions 290003 A2.02 (RO 3.1) (CFR 41.5) Tier 2 / Group 2

REFERENCE:

ARP 4.826.P1 2-3; ABN-RAD-CR Page 3; ABN-RAD-HIGH Page 2; SD000201 Pages 12 and 13; SD000173 Page 24 SOURCE: NEW LO: 5225 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (correct) WEA-FN-51 trips as a result of a high radiation condition at the remote air intakes. The radiation levels are an entry condition into ABN-RAD-CR.

B (incorrect) WOA-V-51C closes on a FAZ signal not a remote air intake high rad condition. There is no entry condition into ABN-RAD-HIGH.

C (incorrect) WMA-FN-54A starts on a FAZ signal not a remote air intake high rad condition. WMA-FN-51A starts based on WMA-FN-54A starting. There is no entry condition into ABN-RAD-HIGH.

D (incorrect) See B and C. The radiation levels are an entry condition into ABN-RAD-CR.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-39 While operating at rated power an event occurs which results in the plant conditions indicated on the TDAS display shown on figure 1.

Assuming all automatic actions occurred as expected, and without operator action, what is the status of the RRC pumps?

A. Both RRC pumps Off.

B. Both RRC pumps operating at 15 Hz.

C. One RRC pump Off and the one RRC pump operating at 60 Hz.

D. One RRC pump operating at 60 Hz and one RRC pump operating at 51 Hz.

ANSWER: C KA # & KA VALUE: Partial or Complete Loss of Forced Core Flow Circulation - Ability to use plant computers to evaluate system or component status. 295001 G2.1.19 (3.9) (CFR 41.10) Tier1 / Group 1

REFERENCE:

Simulator SOURCE: New LO: None LOK / LOD: H/3 HANDOUT: Screenshot of TDAS Containment Status Screen 5 minutes after tripping a single RRC pump showing power at approximately 70.

JUSTIFICATION: A (incorrect) Both RRC pumps secured would result in a reactor power of ~ 50% so power indicated in screenshot is too high.

B (incorrect) Both RRC pumps operating at 15Hz would result in a reactor power of

~55% power so power indicated in screenshot is too high.

C (correct) One RRC pump Secured and the other at 60 Hz would result in reactor power at ~70%.

D (incorrect) Loss of a single channel of ASD will cause a runback to 51Hz on the effected RRC pump resulting in a power of ~95% and power is too low for this to be true.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION FIGURE 1 (Question RO-39)

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-40 Columbia is operating at 100% power when a lockout on SM-2 occurs. All systems operate as designed.

With no operator action, which of the following is the resultant configuration of the feedwater system?

A. RFW-P-1A is running and RFW-P-1B is tripped.

B. RFW-P-1B is running and RFW-P-1A is tripped.

C. Both RFW-P-1A and RFW-P-1B are tripped.

D. Both RFW-P-1A and RFW-P-1B are running.

ANSWER: A KA # & KA VALUE: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: System lineups 295003 AA2.04 (RO 3.5)

(CFR 41.10) Tier1 / Group 1

REFERENCE:

ARP 4.840.A1 8-3 and 8-7; SD000182 page 120; Columbias Simulator SOURCE: NEW LO: 11576 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (correct) A lockout on SM-2 results in a loss of condensate pump 1B and condensate booster pump 2B. This results in low suction pressures for both reactor feed pumps. RFW-P-1B has a four second time delay before it trips and RFW-P-1A has a 10 second time delay. When RFW-P-1B trips, a RRC runback occurs and reactor power is reduced. This power reduction reduces RFW demand which clears the low suction pressure condition and RFW-P-1A continues to operate and restores RPV level.

B (incorrect) See A C (incorrect) See A D (incorrect) See A

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-41 During a plant shutdown, RCIC has been initiated and is maintaining RPV level. A loss of S2-1, 250 VDC power, occurs.

Which of the following explains the effect this loss will have on the RCIC system?

A. The RCIC speed controller, RCIC-FIC-600, loses power which causes RCIC-P-1 to trip on mechanical overspeed. RCIC-V-1, the turbine trip valve, closes. RCIC-V-13, the RPV injection valve, and RCIC-V-45, the steam inlet to the RCIC turbine, remain open due to loss of power.

B. RCIC will continue to operate. RCIC flow controller, RCIC-FIC-600, loses power and no adjustments to RCIC speed or flow is available. If a manual trip is required, the trip pushbutton would still operate and trip the RCIC turbine.

C. RCIC-V-1, the turbine trip valve, closes causing RCIC-P-1 to trip. Manual operation of RCIC is possible by resetting the trip locally and manually positioning the throttle valve to control pump speed.

D. RCIC-P-1 will continue to operate. RCIC speed and injection flow via RCIC-FIC-600, is still available. If a manual trip is required, the trip pushbutton would still operate and trip the RCIC turbine.

ANSWER: D KA # & KA VALUE: Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Systems necessary to assure safe plant shutdown 295004 AA1.02 (RO 3.8) (CFR 41.7) Tier1 / Group 1

REFERENCE:

SD000188 page 23 and 24 SOURCE: NEW LO: 7657 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) See D B (incorrect) See D C (incorrect) See D D (correct) On a loss of S2-1 (250VDC), RCIC will continue to operate if already operating. Operation of the manual trip pushbutton is also maintained. Flow control is still available. On a loss of S1-1 (125 VDC), RCIC would trip if running on mechanical overspeed. Also the trip valve could be reset and local control could be taken of RCIC by operating the throttle valve.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-42 Columbia is operating at rated power at the end of cycle. Events occur that result in a trip of the Main Turbine. CRO1 notes that both Reactor Recirculation Pumps have also tripped.

Which of the following describes the reason the Reactor Recirculation Pumps tripped when the Main Turbine tripped?

A. Prevent cycling SRVs.

B. Prevent power oscillations.

C. Reduce jet pump stall flow.

D. Ensure MCPR is not exceeded.

ANSWER: D KA # & KA VALUE: Knowledge of the reasons for the following responses as they apply to MAIN TURBINE GENERATOR TRIP: Recirculation pump downshift/trip 295005 AK3.02 (RO 3.4) (CFR41.5) Tier1 / Group 1

REFERENCE:

SD000129 Page 39; CGS Simulator; SD000178 page 28 and 29 SOURCE: NEW LO: 11647 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) One or two SRVs do cycle open on a MT trip at rated power.

B (incorrect) Power oscillations occur if operating in certain regions of power to flow map and is not the reason the RRC pumps trip on a MT trip.

C (incorrect) Stall flow occurs when one RRC pump is operating. While tripping both pumps would result in reduced stall flow this is not the reason the RRC pumps trip on a MT trip.

D (correct) EOC RPT mitigates the MCPR vulnerability.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-43 A condition exists that required a manual scram insertion. The mode switch was taken to the SHUTDOWN position but control rods did not insert. In response, CRO1 depresses the four RPS Reactor Scram pushbuttons.

Which of the following is the minimum combination of RPS Reactor Scram pushbuttons that have to be depressed in order to initiate a full reactor scram?

A. Depress both the A1 and the A2 pushbuttons and then depress both the B1 and the B2 pushbuttons.

B. Depress both the A1 and the B1 pushbuttons and then depress both the A2 and the B2 pushbuttons.

C. Depress the A1 pushbutton and then depress the A2 pushbutton.

D. Depress the A2 pushbutton and then depress the B2 pushbutton.

ANSWER: D KA # & KA VALUE: Knowledge of the interrelations between SCRAM and the following: RPS 295006 AK2.01 (RO 4.3) (CFR 41.7) Tier1 / Group 1

REFERENCE:

SD000161 page 12 SOURCE: NEW LO: 5953 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect) While depressing these pushbuttons will cause a full reactor scram the question asks for the minimum combination - this is not the minimum combination.

B (incorrect) See A C (incorrect) This combination will not cause a full scram.

D (correct) A full scram is inserted by depressing any combination of A1 or A2 and B1 or B2. The switches are grouped as follows: left side of mode switch is A1 and B1 and right side of the mode switch is A2 and B2.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-44 The Control Room was abandoned due to a fire. The Remote and Alternate Remote Shutdown panels have been activated. Actions are underway to place Columbia into Shutdown Cooling. The Control Room Operator is reviewing the Cooldown Temperature/Pressure Data Log which includes the following data:

Time Temperature (°F) 1000 500 1030 470 1100 430 1130 365 1200 355 1230 295 1300 250 1330 225 1400 175 1430 140 1500 135 Which of the following is the correct concerning this cooldown?

The Technical Specification limit for cooldown rate was A. exceeded during only one, one hour period.

B. exceeded during only two, one hour periods.

C. exceeded during only three, one hour periods.

D. exceeded during at least four, one hour periods.

ANSWER: B KA # & KA VALUE: Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Cooldown rate 295016 AA2.06 (RO 3.3) (CFR 41.10)

Tier1 / Group 1

REFERENCE:

ABN-CR-EVAC Page 55 SOURCE: NEW LO: 5005 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: A (incorrect) See B B (correct) Tech Spec cooldown limit is 100°F/Hr and two of the hourly readings exceed that limit (1030 to 1130 and 1200 to 1300). Columbia has an administrative limit of 80°F and that limit was exceed three times (1030 to 1130 and 1200 to 1300 and 1330 to 1430).

C (incorrect) See B D (incorrect) See B

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-45 Columbia is operating at 70% power when a complete loss of Reactor Closed Cooling Water required a reactor scram to be inserted.

Which of the following describes the reason for the required reactor scram?

The scram is directed because of the loss of cooling to the ..

A. Control Rod Drive (CRD) Pump seals.

B. Residual Heat Removal (RHR) Pump seal coolers.

C. Reactor Recirculation (RRC) Pump motors and seals.

D. Reactor Water Clean-Up (RWCU) Non-Regenerative Heat Exchangers.

ANSWER: C KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER :

Effects on component/system operations 295018 AK1.01 (RO 3.5) (CFR 41.8) Tier1

/ Group 1

REFERENCE:

ABN-RCC Page 3; ABN-RRC-LOSS Page 3 SOURCE: BANK - April 2001 NRC Exam LO: 10361 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect) Is a credible answer as RCC cools the CRD pump seals but not the reason the scram is inserted.

B (incorrect) Is a credible answer as RCC cools the RHR pump seal coolers but is not the reason the scram is inserted.

C (correct) A scram is inserted because the RRC pump motors and seals have lost cooling and they are stopped right after a scram is inserted. (Stopping the RRC Ps would require entry into ABN-RRC-LOSS which would also then require a scram).

D. (incorrect) Is a credible answer as RCC cools the RWCU NRHXs but is not the reason the scram is inserted.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-46 With Columbia operating at full power a total loss of Control Air pressure occurs.

If no operator actions are taken, which of the following is designed to be utilized, in the resultant plant configuration, to maintain RPV level?

A. RCIC-P-1 B. HPCS-P-1 C. CRD-P-1A/1B D. RFW-P-1A/1B ANSWER: A KA # & KA VALUE: Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Reactor core isolation cooling 295019 AK2.16 (RO 2.8) (CFR 41.7) Tier1 / Group 1

REFERENCE:

SD000180 Page 3; SD000174 Page 3 SOURCE: NEW LO: 5713 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: A (correct) RCIC is designed to be utilized to maintain RPV level in hot shutdown and when the vessel is isolated from the main condenser (MSIVs closed).

B (incorrect) HPCS is high pressure injection systems and all could be used but is designed to maintain RPV level during LOCA conditions.

C (incorrect) CRD is a high pressure injection system but is not designed to maintain RPV level. In this configuration, CRD could not maintain RPV level.

D (incorrect) A complete loss of control air pressure causes the MSIVs to close rendering the RFW pumps not available.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-47 Columbia is in a refueling outage with the fuel shuffle underway. A total loss of Shutdown Cooling occurs.

The CRS enters ABN-RHR-SDC-LOSS and directs RWCU flow to be maximized and RCC-V-8 (RWCU HX Outlet) to be throttled open.

Why would these actions be taken?

A. To maximize cleanup flow in the event of core damage reducing radiation levels.

B. To help promote circulation with the spent fuel pool enabling decay heat to be removed by the SFP heat exchangers.

C. To enable natural circulation flow through the reactor providing additional time to recover shutdown cooling.

D. To improve heat transfer rates providing additional time to recover shutdown cooling.

ANSWER: D KA # & KA VALUE: Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING: Maximizing reactor water cleanup flow 295021 AK3.04 (3.3) (CFR 41.5) Tier1 / Group 1

REFERENCE:

ABN-RHR-SDC-LOSS Page 12 SOURCE: New LO: NONE LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) Core damage is prevented by core submergence.

B (incorrect) Without RHR no flow path exists with the spent fuel pool.

C (incorrect) Natural circulation is promoted by the RPV water level and RWCU flow could possibly disrupt this.

D (correct) Per ABN-RHR-SDC-LOSS bases this improves heat transfer rates providing additional time to recover shutdown cooling.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-48 A large steam rupture occurs inside containment resulting in drywell pressure scram of the reactor.

Given:

All automatic actions occur as expected.

No operator actions take place.

What will happen to wetwell pressure?

A. MSIVs will close when pressure initially decreases then RPV pressure will rise, opening SRVs causing wetwell pressure to increase slightly higher than drywell pressure.

B. When the drywell pressure increases to .2 psid higher than the wetwell pressure, the wetwell to drywell vacuum breakers will open maintaining drywell pressure .2 to .5 psig higher than wetwell pressure.

C. The steam leak will pressurize the drywell until pressure can overcome the water pressure in the downcomers venting drywell to the wetwell maintaining drywell pressure approximately 5 psig higher than wetwell pressure.

D. The steam leak will continue to pressurize the drywell and with no operator action will cause a failure of the drywell floor from the downward force applied by the rupture.

ANSWER: C KA # & KA VALUE: Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Suppression chamber pressure: Plant-Specific 295024 EA2.04 (3.9) (CFR 41.10) Tier1 / Group 1

REFERENCE:

SD000127 pages 8, 22-28, 33, and simulator SOURCE: NEW LO: NONE LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) MSIVs will close eventually, but because of the large rupture, the SRVs will not open.

B (incorrect) The vacuum breakers operate in the opposite direction as stated, but at the pressures listed.

C (correct) The downcomers provide a flow path for uncondensed steam from the drywell to the wetwell during accident conditions. The downcomers are covered by approximately 12 feet of water so the pressure difference will have to overcome the pressure due to the height of that water plus any pressure on the top of the water resulting in approximately 5 psig difference.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION D (correct) The downcomers will provide protection from damage to the drywell floor from this casualty.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-49 While operating at rated power a reactor scram occurs. While verifying how high RPV pressure went after the scram, CRO1 reports that both RRC-P-1A and RRC-P-1B have tripped.

Which of the following is the minimum value that RPV pressure reached to cause the above plant response?

A. 831 psig B. 1060 psig C. 1091 psig D. 1120 psig ANSWER: D KA # & KA VALUE: High Reactor Pressure - Ability to identify and interpret diverse indications to validate the response of another indication. 295025 G2.1.45 (RO 4.3) (CFR 41.7)

Tier1 / Group 1

REFERENCE:

SD000178 page 13; SD000184 page 24 SOURCE: NEW LO: 7639 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (incorrect) At 831 psig with the mode switch in run the MSIVs close causing a reactor scram and RRC pumps to 15 Hz due to low RPV level.

B (incorrect) At 1060 psig, a reactor scram occurs but the RRC pump speed is 15 Hz due to RPV low level.

C (incorrect) 1091 psig is the first setpoint that SRVs open. The reactor would scram and RRC pumps at 15 Hz.

D (correct) At 1120 psig the reactor would scram and RRC pumps trip due to ATWS-ARI signal.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-50 Under which of the following conditions would you use CMS-TR-5 (Wetwell-Drywell Temp Recorder) point AO2 instead of using SPTM-TI-5 (Wetwell Avg. Temp.)?

A. Suppression Pool level -9 inches.

B. Suppression Pool level +2 inches.

C. Wetwell temperature 110°F.

D. Drywell temperature 330°F.

ANSWER: A KA # & KA VALUE: Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Temperature monitoring 295026 EA1.03 (3.9) (CFR 41.41.7) Tier1 / Group 1

REFERENCE:

4.601.A12 (601.A12 Annunciator Panel Alarms) page 6 and 7, Photo of Control Room panel P601 showing operator aid new SPTM-TI-5 and CMS-TR-5 SOURCE: New LO: None LOK / LOD: F/3 HANDOUT: None JUSTIFICATION: Per note under alarm drop 1-3 NOTE: Upper temperature elements will be uncovered if Suppression Pool level is at approximately 303 on wide range or -9 inches on narrow range.

A (correct) CMS-TR-5 and 6 are the upper temperature elements and will be uncovered so by operator aid on meter in Control Room and alarm drop ARP should not be used.

B (incorrect) An EOP Entry condition but not reason.

C (incorrect) Temperature when scram and boron injection is required prior to exceeding.

D (Incorrect) Temperature Emergency Depressurization is required.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-51 A refueling accident occurs resulting in the radioactivity levels at the Exclusion Area Boundary exceeding the Alert Level. PPM 5.4.1 Radioactivity Release Control is entered and the Radwaste Building HVAC is restarted.

Which of the following is the bases for restarting the Radwaste Building HVAC in PPM 5.4.1?

A. Allows for radioactivity to be released at ground level limiting the dispersion of radioactivity.

B. Results in positive pressure in the building to limit intrusion of radioactivity from the reactor building.

C. Prevents Control Room from becoming uninhabitable due to high radiation.

D. Preserves accessibility, and assures that radioactivity is discharged through an elevated, monitored release point.

ANSWER: D KA # & KA VALUE: Refueling Accidents - Knowledge of the specific bases for EOPs. 295023 G2.4.18 (3.3) (CFR 41.10) Tier1 / Group 1

REFERENCE:

PPM 5.0.10 page 315 SOURCE: NEW LO: 8477 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: From PPM 5.0.10 page 316: Operation of ventilation in this structure preserves accessibility, and assures that radioactivity is discharged through an elevated, monitored release point.

A (incorrect) Radwaste HVAC is an elevated release point, but a ground release would limit dispersion of radioactivity.

B (incorrect) Pressure in the Rad Waste building is slightly negative, but if it were positive it would limit radioactivity from the Reactor Building..

C (incorrect) Control Room is located in the Radwaste building, but it had its own ventilation system separate from the Radwaste building.

D (correct) See note above

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-52 When entering PPM 5.2.1 Secondary Containment Control, Emergency Depressurization is required on High Drywell temperature if it cannot be restored and maintained below the limit.

This is done to prevent exceeding the design temperature of which of the following components:

A. Drywell Cooling Fans B. Main Steam Isolation Valves C. ADS SRVs D. Drywell to Wetwell Vacuum Breakers ANSWER: C KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Equipment environmental qualification 295028 EK1.02 (2.9) (CFR 41.8) Tier1 / Group 1

REFERENCE:

PPM 5.0.10 page 273, FSAR Section 5.2.2.4 SOURCE: NEW LO: 8318 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) See C B (incorrect) See C C (correct) Per PPM 5.0.10 When drywell temperature cannot be restored and maintained below the ADS design temperature, emergency RPV depressurization is performed.

D (incorrect) See C

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-53 An Operating Basis Earthquake caused a rupture in the Suppression Pool. Suppression Pool water level is lowering. EOP 5.2.1, Primary Containment Control, has been entered on low Suppression Pool water level.

HPCS-P-1 has been started per PPM 5.5.23, Emergency Suppression Pool Makeup. A failure of a relay then causes HPCS-V-1, CST suction, to close and HPCS-V-15, Suppression Pool suction, to open.

Which of the following is correct?

A. Allow HPCS-P-1 to run regardless of Suppression Pool water level.

B. Secure HPCS-P-1 if Suppression Pool water level cannot be maintained above 14.5.

C. Secure HPCS-P-1 if Suppression Pool water level cannot be maintained above 14.0.

D. Secure HPCS-P-1 if Suppression Pool water level cannot be maintained above 5.5.

ANSWER: D KA # & KA VALUE: Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCS 295030 EA1.03 (RO 3.4) (CFR 41.7) Tier1 / Group 1

REFERENCE:

PPM 5.0.10 page 90; SD000174 page 11; EOP 5.2.1; PPM 5.5.23 page 4 SOURCE: NEW LO: 8389 LOK / LOD: F/3 HANDOUT: None JUSTIFICATION: A (incorrect) This is a credible distractor as HPCS-V-15 is the suction valve from the suppression pool. If HPCS-V-15 were to be closed, HPCS would be allowed to run regardless of Suppression Pool water level. With HPCS-V-15 being open, vortex limits apply.

B (incorrect) Suppression Pool level at 14.5 require LPCS and RHR-C to be secured - not HPCS.

C (incorrect) Suppression Pool level at 14 requires RHR-A and RHR-B to be secured - not HPCS.

D (correct) A suppression pool level of 5.5 requires HPCS-P-1 to be secured.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-54 During performance of PPM 5.1.1 RPV Control, why is it desired to maintain RPV level greater than -161 inches?

It ensures adequate core cooling through A. core submergence.

B. steam cooling with injection.

C. spray cooling with HPCS or LPCS injecting.

D. steam cooling without injection.

ANSWER: A KA # & KA VALUE: Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL: Core coverage 295031 EK3.02 (4.4) (CFR 41.5) Tier1 /

Group 1

REFERENCE:

PPM 5.0.10 page 17 SOURCE: New LO: 8018 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A (correct) Core submergence is at -161 inches.

B (incorrect) Steam cooling with injection is at -183 inches.

C (incorrect) Spray cooling with HPCS or LPCS is at -210 inches.

D (incorrect) Steam cooling without injection is at -201 inches.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-55 Which of the following combinations would NOT require the Control Room Supervisor to transition from PPM 5.1.1, RPV Control, to PPM 5.1.2, RPV Control - ATWS?

1. 1 rod at RPIS position 48

2. 1 rod at RPIS position 08

3. 1 rod at RPIS position 04

4. 2 rods at RPIS position 02

5. all other rods at RPIS position 04

6. all other rods at RPIS position 02

7. all other rods at RPIS position 00 A. Conditions 1 and 2 and 7 B. Conditions 1 and 4 and 5 C. Conditions 2 and 3 and 6 D. Conditions 2 and 4 and 7 ANSWER: D KA # & KA VALUE: Knowledge of the interrelationship between SCRAM CONDITION PRESENT AND REACTOR ABOVE APRM DOWNSCALE OR UNKNOWN and the following: RPIS: Plant Specific 295037 EK2.14 (RO 3.6) (CFR 41.7) Tier1 / Group 1

REFERENCE:

OI-15 Page 11 SOURCE: BANK - NRC Exam March 2009 (LO01786)

LO: 7784 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: This is required Reactor Operator knowledge at Columbia.

A (incorrect) The combination of 1 and 2 makes entry into 5.1.2 required.

B (incorrect) Item 5 makes entry into PPM 5.1.2 required.

C (incorrect) The combination of 2 and 3 makes entry into PPM 5.1.2 required.

D (correct) Per OI 15 the reactor is shutdown with one rod at any position and all others at least inserted to position 02.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-56 Events have occurred at Columbia and a radioactive release is ongoing. The CRS has entered the EOPs and the Incident Advisor has just completed a QEDPS, and, based upon the results, the Shift Manager is in the process of classifying the event.

Which of the following action levels is the first level at which protection of the general public is performed by either sheltering them in place or evacuating them to a different location; and what actions must be taken if this level is reached?

A. Site Area Emergency; Isolate all primary systems discharging into areas outside PC and SC except those required to assure adequate core cooling.

B. Site Area Emergency; Enter PPM 5.1.1, RPV Control, and insert a manual reactor scram.

C. General Emergency; Isolate all primary systems discharging into areas outside PC and SC except those required to assure adequate core cooling.

D. General Emergency; Enter PPM 5.1.1, RPV Control, and insert a manual reactor scram.

ANSWER: A KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE : Protection of the general public 295038 EK1.02 (4.2) (CFR 41.41.8) Tier1 / Group 1

REFERENCE:

Classification Notification Form; EOP 5.4.1 SOURCE: NEW LO: 11152, 8894 LOK / LOD: H/3 HANDOUT: EOP 5.4.1 flowchart JUSTIFICATION: The public is first sheltered/evacuated at a Site Area Emergency. Once a SAE is reached you would isolate all primary systems discharging into areas outside PC and SC except those required to assure adequate core cooling. A scram is inserted when GE levels are reached.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-57 The plant is operating at power when a spurious fire alarm is received in the Control Room on FCP-1.

Which of the following concerning the Fire Protection system is correct?

A. If another detector from the zone with the spurious alarm generates a fire alarm, the fire alarm will reflash.

B. A sealed in spurious alarm will prevent any valid fire alarm from that zone from annunciating.

C. When a fire alarm is received, all trouble and additional fire alarms in the building associated with the spurious alarm are masked.

D. When a fire alarm is received, all trouble and additional fire alarms in all zones throughout the plant are masked.

ANSWER: B KA # & KA VALUE: Knowledge of the interrelations between PLANT FIRE ON SITE and the following:

Sensors / detectors and valves 600000 AK2.01 (2.6) Tier1 / Group 1

REFERENCE:

SD000177 Page 31 SOURCE: BANK - MODIFIED - LO00586 LO: 7610 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A fire alarm masks all other fire alarms in that zone only. Alarms in the building and plant will annunciate as long as they are not in the same zone. A fire alarm does not reflash once alarmed.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-58 Columbia has been contacted by the BPA dispatcher and informed that the grid analysis program has projected a degraded reliability of the 230kV power source (E-TR-S) such that post trip voltage for E-TR-S will be less than the allowable limits. The CRS enters ABN-ELEC-GRID.

Which of the following is a procedural action and also correctly explains the reason for the action?

A. Announcement is made to suspend all surveillances which have a potential to trip the Reactor or Main Turbine which limits the vulnerability of the plant tripping and causing a total loss of power.

B. Announcement is made to suspend all surveillances which have a potential to trip the Reactor or Main Turbine which ensures that when the results of the STATE ESTIMATOR is determined, the readings accurately reflect actual plant performance.

C. Monitor the Generator Exciter Volts DC meter for oscillation because the power system stabilizer is not designed to automatically maintain the main generator within the Generator Capability Curve during grid disturbances.

D. Directs a plant shutdown per PPM 3.2.1, Normal Plant Shutdown, in order to limit the possibility of a Main Turbine/Generator trip and subsequent reactor scram as a result of the potential loss of offsite power.

ANSWER: A KA # & KA VALUE: Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRICAL GRID DISTURBANCES: Actions contained in abnormal operating procedure for voltage and grid disturbances 700000 AK3.02 (RO 3.6) (CFR 41.4) Tier1 / Group 1

REFERENCE:

ABN-ELEC-GRID Pages 5 and 12 SOURCE: NEW LO: 15748 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: Per ABN-ELEC-GRID, surveillances are stopped to limit the vulnerability of the plant tripping and causing a total loss of power. The CRS would approve all troubleshooting activities by BPA but that is not the reason to stop the surveillances.

The Generator Volts DC is monitored but it is designed to automatically maintain the Generator Capability Curve. A plant shutdown is not directed by the ABN but is a plausible distractor.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-59 Columbia is starting up with Reactor power currently 24%. A malfunction in the Off Gas system results in lowering main condenser vacuum (increasing main condenser backpressure). The power increase is stopped to investigate.

If main condenser backpressure reaches 6.0 in Hg, which of the following explains the resultant plant response?

The Main Turbine trips, the MSIVs ..

A. close, the Reactor scrams.

B. remain open, the Reactor scrams.

C. remain open, the Reactor remains at power, and Reactor power will increase.

D. remain open, the Reactor remains at power, and Reactor power will decrease.

ANSWER: C KA # & KA VALUE: Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM: Reactor power 295002 AA2.02 (RO 3.2) (CFR 41.10)

Tier1 / Group 2

REFERENCE:

SD000129 page 31; SD000173 page 7 and 8; SD000161 page 14 SOURCE: NEW LO: 11645 and 11646 LOK / LOD: H/4 HANDOUT: NONE JUSTIFICATION: The Main Turbine trips under all loads at 5.5 in Hg backpressure.

A (incorrect) The MSIVs remain open (dont close until 8.3 in Hg) and at 24%

power the reactor does not scram due to a MT Trip. (Reactor power GT 30%

initiates a scram).

B (incorrect) At 24% power the reactor does not scram due to a MT Trip. (Reactor power GT 30% initiates a scram).

C (correct) The MT trip causes a loss of extraction steam to the feedwater heaters which will result in a feedwater temperature decrease which will cause Reactor power to increase.

D (incorrect) See C

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-60 A plant startup is in progress. RPV level is currently +45 and going up slowly. RPV pressure is steady at 75 psig.

Which of the following could be utilized to reduce RPV level?

A. Let down utilizing the in-service Shutdown Cooling loop.

B. Reduce Control Rod Drive System drive flow.

C. Reduce injection with Reactor Core Isolation Cooling.

D. Let down utilizing the Reactor Water Cleanup System.

ANSWER: D KA # & KA VALUE: Knowledge of the interrelationships between HIGH REACTOR WATER LEVEL and the following: Reactor Water Cleanup System (ability to drain) 295008 AK2.09 (3.1) (CFR 41.5) Tier1 / Group 2

REFERENCE:

PPM 3.1.2 Flowchart blocks L4, S9, and S14 SOURCE: Bank LO01120 LO: 5033 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) Shutdown Cooling is isolated prior to 48 psig per PPM 3.1.2, Plant Startup.

B (incorrect) While CRD is in service, reducing CRD drive water flow would cause control rod speed to change but does not affect the amount of water going to the RPV.

C (incorrect) RCIC is not placed into service before 125 psig.

D (correct) RWCU is used for letdown at this point in the plant startup.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-61 PPM 5.1.2 has a caution concerning rapid injection. Which of the following is a possible consequence of rapid injection of water into the vessel during ATWS conditions?

A. Thermal shock to the feedwater nozzles.

B. Fuel damage.

C. Rapid pressure decrease exceeding 100o/hr cooldown limit.

D. Feedpump / main turbine trip on high reactor level.

ANSWER: B KA # & KA VALUE: Inadvertent Reactivity Addition - Knowledge of the operational implications of EOP warnings, cautions, and notes. 295014 G2.4.20 (3.8) (CFR 41.10) Tier1 / Group 2

REFERENCE:

PPM 5.0.10 page 67 SOURCE: Exam Bank LR00837 LO: 8499 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: From PPM 5.0.10 the reason for the caution is to warn the operator of the potential plant response if injection of cold, unborated water into the core is too rapid. This may result in a large increase in positive reactivity with an attendant reactor power excursion sufficient to substantially damage the core.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-62 If a scram is required and the mode switch has been rotated to the SHUTDOWN position, which of the following requires depressing the manual scram pushbuttons?

1. All control rods at position 04

2. Two control rods full out

3. APRM Downscale lights not illuminated

4. APRM Upscale lights illuminated A. 1, 2, 3 and 4 B. Only 2, 3 and 4 C. Only 3 and 4 D. Only 4 ANSWER: C KA # & KA VALUE: Ability to operate and/or monitor the following as they apply to INCOMPLETE SCRAM: Neutron monitoring system. 295015 AA1.07 (RO 3.6) (CFR 41.7) Tier1 /

Group 2

REFERENCE:

PPM 3.3.1 Page 7; SD000149 page 21 SOURCE: NEW LO: 6686 LOK / LOD: H/2 HANDOUT: None JUSTIFICATION: Per PPM 3.3.1, if APRMs are not downscale (LT 5% power), additional scram actions are performed and the manual scram pushbuttons are depressed.

A (incorrect) Item 1 and item 2 would not keep the reactor GT 5% power but both are criteria for entering the ATWS EOP.

B (incorrect) Item 2 would not keep the reactor GT 5% power.

C (correct) Per PPM 3.3.1, if APRMs are not downscale, additional scram actions are performed. With Mode switch not in run the APRM upscale lights are lit at 12%

power which indicates APRMs are not downscale.

D (incorrect) Item 3 would also be a reason to depress the manual scram pushbuttons but is not included in this choice because of the word only in the answer.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-63 Columbia is operating at rated power when an event occurs causing RCC-V-104 (Drywell Supply Outboard Isolation) to close.

What will be the plant response to RCC-V-104 closing and the reason why?

A. Drywell pressure will increase due to a failure of RRC pump seals fail due to a loss of cooling water.

B. Drywell pressure will increase due to a loss of cooling water to Drywell Coolers.

C. RWCU pumps will trip on High Motor Cavity temperature due to a loss of cooling water to the pump motor cooler.

D. RWCU pumps will trip on low system flow when damaged resin clogs the resin traps from a loss of cooling water flow to the NRHX.

ANSWER: B KA # & KA VALUE: Knowledge of the reasons for the following responses as they apply to INADVERTENT CONTAINMENT ISOLATION: Suppression chamber pressure response 295020 AK3.08 (3.3) (CFR 41.4) Tier1 / Group 2

REFERENCE:

ABN-RCC page 7 SOURCE: NEW LO: NONE LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) CRD purge flow will provide adequate cooling to the RRC pump seals.

B (correct) A loss of RCC cooling to the Drywell air coolers can lead to a high Drywell pressure condition.

C (incorrect) RCC-V-104 does not isolate cooling to the RWCU Pump motor cooler.

D (incorrect) RCC-V-104 does not isolate cooling to the RWCU NRHX.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-64 The plant was operating at 90% power when a transient occurred. The following conditions exist:

LD-TE-4A/B - RCIC Pump Room is 239°F ARM-RIS RCIC Pump Room is off scale high Which of the following is correct concerning these conditions?

A. There is a fire in the RCIC Pump Room and the continued operability of RCIC is in question.

B. There is a fire in the RCIC Pump Room and RCIC operation could continue under these conditions.

C. Steam is discharging into the area and the continued operability of RCIC is in question.

D. Steam is discharging into the area and RCIC operation could continue under these conditions.

ANSWER: C KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Impact of operating environment on components 295032 EK1.04 (RO 3.1) (CFR 41.8) Tier1 /

Group 2

REFERENCE:

PPM 5.0.10 Page 251, 253 and 254 SOURCE: BANK - 2001 NRC Exam (LO01271)

LO: 8456 LOK / LOD: H/3 HANDOUT: None JUSTIFICATION: A (incorrect) The indications and the basis for the indication is for a primary system discharging into the area.

B (incorrect) The continued operation of RCIC is in question with the conditions listed.

C (correct) The indications and the basis for the indication is for a primary system discharging into the area.

D (incorrect) The continued operation of RCIC is in question with the conditions listed.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-65 With the plant operating in Mode 1 an electrical fault causes the isolation of FDR-V-219 (Sump R-3 and R-4 discharge to Rad Waste).

With FDR-V-219 closed which of the following could cause Sump R-3 or R-4 to overflow?

A. RRC Pump 1A lower seal failure.

B. Service water leak from the room cooler in RHR B pump room.

C. A large packing leak on RCIC water leg pump.

D. A large packing leak on CRD pump 1A.

ANSWER: D KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL:

Electrical ground/ circuit malfunction. 295036 EK1.02 (2.6) (CFR 41.8) Tier1 /

Group 2

REFERENCE:

SD000130 figure 8 and figure 13 SOURCE: NEW LO: NONE LOK / LOD: H/4 HANDOUT: NONE JUSTIFICATION: A (incorrect) RRC pump seal drains into the EDR sump then into the R5 sump.

B (incorrect) RHR B pump room drains into the R2 sump.

C (incorrect) RCIC pump room drains into the R1 sump.

D (correct) CRD pump room drains into R3 sump.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-66 Per OI-09, Operations Standards and Expectations, which of the following conditions should be announced to the plant staff over the PA system?

A. A radioactive spill on the RW 437 level has just been reported by the laborer supervisor.

B. The starting of the Auxiliary Oil pump for the B RFP as part of a plant shutdown.

C. The starting the Turbine Seal Oil Backup pump as part of a plant startup.

D. The CRO is commencing the weekly control rod exercising surveillance.

ANSWER: A KA # & KA VALUE: Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc. 2.1.14 (3.1) (CFR 41.10) Tier 3

REFERENCE:

OI-09 Pages 11 and 36 SOURCE: NEW LO: 6086 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: Per OI-09, only the radioactive spill should be announced. None of the other choices are required to be announced.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-67 You are sent to complete a valve lineup on the High Pressure Core Spray system per SOP-HPCS-LU. During the performance of the valve lineup you note that the required position for some of the valves is C+.

Which of the following describes the meaning of the C+?

The valve is closed A. and torqued to a specific value in the comment section.

B. and logged closed in the control room logs.

C. and a cap is required to be installed.

D. for a valve located in containment.

ANSWER: C KA # & KA VALUE: Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. 2.1.29 (RO 4.1) (CFR 41.10) Tier 3

REFERENCE:

Glossary page 44 and 47 SOURCE: BANK - LO01839 LO: 9851 LOK / LOD: F/2 HANDOUT: NONE JUSTIFICATION: A ( incorrect) See C B (incorrect) See C C (correct) C+ is a closed valve that has a cap on it.

D (incorrect) +C indicates the valve is in containment.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-68 During a refueling outage with fuel shuffle in progress, the Control Room Operator is monitoring Source Range Monitors.

If SRM count rate were to increase to exceed 300 counts per second (cps) during a bundle insertion, which of the following is correct?

A. Immediately stop bundle insertion. Once counts drop to LT 300 cps, insertion can then be continued at a slower rate so as not to exceed 300 cps.

B. Immediately stop bundle insertion. Withdraw the bundle from the core and contact the Station Nuclear Engineer for evaluation.

C. Immediately stop bundle insertion. Do not withdraw the bundle from the core until directed by the Station Nuclear Engineer.

D. Insertion of the bundle may continue to full insertion. When bundle is inserted, notify the Station Nuclear Engineer for evaluation.

ANSWER: B KA # & KA VALUE: Knowledge of RO duties in the control room during fuel handling such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation. 2.1.44 (RO 3.9) (CFR 41.10) Tier 3

REFERENCE:

PPM 6.3.2 Page 18 SOURCE: NEW LO: 8829 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) If a doubling of count rate occurs the insertion would be stopped and continued at a slower rate.

B (correct) During refueling a count rate of 300 cps requires the insertion to be immediately stopped and the bundle withdrawn from the core and the SNE notified.

C (incorrect) See B D (incorrect) See B

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-69 While venting Control Rods per PPM 3.1.1 Master Startup Checklist, a Control Rod is encountered that will not withdraw using normal drive water pressure.

In order to flush air from the insert line how many minutes should the control rod be continuously inserted for?

If this fails what procedure should be used to increase drive water pressure?

A. 1 minute; ABN-ROD B. 2 minutes; ABN-ROD C. 1 minute; SOP-CRD-HCU D. 2 minutes; SOP-CRD-HCU ANSWER: B KA # & KA VALUE: Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity. 2.2.1 (4.5)

(CFR 41.5) Tier 3

REFERENCE:

PPM 3.1.1 pages 28-30 SOURCE: NEW LO: 13301 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) Per note on Page 28 of PPM 3.1.1 2 minute insert flush.

B (correct) Per note on Page 28 of PPM 3.1.1 2 minute insert flush. Per page 30 step 8.2.14.m increase drive water pressure per ABN-ROD for those rods which could not be withdrawn at normal drive water pressure.

C (incorrect) Per note on Page 28 of PPM 3.1.1 2 minute insert flush. SOP-CRD-HCU has procedure for filling and venting CRD HCUs but is not the correct procedure called for.

D (incorrect) Per note on Page 28 of PPM 3.1.1 2 minute insert flush. SOP-CRD-HCU has procedure for filling and venting CRD HCUs but is not the correct procedure called for.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-70 When a safety related motor operated valve has been manually backseated for maintenance, the valve must be declared inoperable until motor operation can be demonstrated.

Select the response below which indicates how the condition of the valve is documented until post maintenance operability requirements are completed.

A. A Caution Tag clearance order shall be issued to document the abnormal condition of the motor operated valve.

B. A Danger Tag clearance order shall be issued to prevent energizing the operator with the valve backseated.

C. A Temporary Modification Request Tag (blue tag) is attached to indicate the condition of the valve.

D. A Maintenance Work Request and its associated Problem Tag is used to document the status of the MOV.

ANSWER: A KA # & KA VALUE: Knowledge of pre- and post-maintenance operability requirements. 2.2.21 (4.0)

(CFR 41.10) Tier 3

REFERENCE:

OI-12 pg. 41, PPM 1.3.1 pg. 57 SOURCE: Bank LR01004 LO: 6259 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (correct) Per OI-12 Any MOV that is manually backseated should be Caution tagged to document the abnormal condition of this valve.

B (incorrect) This tag is a type of tag used at CGS but not for this purpose.

C (incorrect) This tag is a type of tag used at CGS but not for this purpose.

D (incorrect) This tag is a type of tag used at CGS but not for this purpose.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-71 Some instruments in the Control Room have color banding on the indicating range of the meter.

What does operation in the BLUE color banding indicate?

A. Normal Operating Range (NOR)

B. Alert Range, Operation in this range should be avoided.

C. Action Range, Operation in this range should not be allowed.

D. Technical Specification/LCS/FSAR/ODCM/EOP Requirement, Operation in this range may be a violation of licensing bases.

ANSWER: D KA # & KA VALUE: Ability to recognize system parameters that are entry-level conditions for Technical Specifications. 2.2.42 (3.9) (CFR 41.7) Tier 3

REFERENCE:

OI-45 pg. 4 SOURCE: NEW LO: NONE LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) This would be a Green Band.

B (incorrect) This would be a Yellow Band.

C (incorrect) This would be a Red Band.

D (correct) Blue banding indicates Technical Specification/LCS/FSAR/ODCM/EOP Requirement, Operation in this range may be a violation of licensing bases.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-72 After entering PPM 5.2.1 Primary Containment Control, it is determined that emergency ventilation of the primary containment is required per PPM 5.5.14 Emergency Wetwell Venting.

PPM 5.5.14 directs all personnel in the Reactor Building to be evacuated, what is the bases for this?

A. Venting the Wetwell through the SGT system at high Wetwell pressures may rupture the SGT system ducting and release high energy steam directly into the Reactor Building.

B. Venting the Wetwell through the RB HVAC system at high Wetwell pressures may rupture the RB HVAC system ducting and release high energy steam directly into the Reactor Building.

C. Venting the Wetwell through the SGT system at high Wetwell pressures may rupture the SGT system ducting and release the radioactivity directly into the Reactor Building.

D. Venting the Wetwell through the RB HVAC system at high Wetwell pressures may rupture the RB HVAC system ducting and release the radioactivity directly into the Reactor Building.

ANSWER: C KA # & KA VALUE: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. 2.3.14 (3.4) (CFR 41.12) Tier 3

REFERENCE:

PPM 5.5.14 Page 4 and 5 SOURCE: NEW LO: NONE LOK / LOD: H / 3 HANDOUT: NONE JUSTIFICATION: Caution in PPM 5.5.14: Venting the Wetwell through the SGT system at high Wetwell pressures may rupture the SGT system ducting and release the radioactivity directly into the secondary containment.

A (incorrect) Steam entering the Wetwell will have most of its energy removed condensing the steam.

B (incorrect) PPM 5.5.14 directs venting through the SGT system.

C (correct)

D (incorrect) PPM 5.5.14 directs venting through the SGT system.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-73 Which of the following is the detector type used by the Main Steam Line (MS-RIS-610A/B/C/D) and the SJAE Condenser Outlet (OG Pretreatment) (OG-RIS-612) radiation monitors?

A. Beta Scintillation detector B. Geiger Mueller detector C. Gamma Scintillation detector D. Ion Chamber detector ANSWER: D KA # & KA VALUE: Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. 2.3.15 (2.9) (CFR 41.12) Tier 3

REFERENCE:

SD000147 page 36 SOURCE: NEW LO: 5646 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) See D B (incorrect) See D C (incorrect) See D D (correct) The Main Steam line rad monitor and the OG pre-treatment monitors are Ion Chamber monitors.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-74 With the plant at power, a reactor scram occurred on high neutron flux. The following conditions now exist:

RPV level 10", rising slowly Reactor power IRM range 3 Drywell pressure 1.5 psig, rising slowly Wetwell temperature 89°F steady REA-RIS-609A-D (RB Exhaust) 15 mrem/hr PPM 5.1.1, RPV Control, has been entered on low RPV level.

Which of the following Emergency Operating Procedures (EOPs) should also be entered at this time given the above conditions?

A. PPM 5.1.2, RPV Control - ATWS and inhibit ADS and take manual control of HPCS.

B. PPM 5.2.1, Primary Containment Control and maintain Wetwell temperature with available Wetwell cooling.

C. PPM 5.3.1, Secondary Containment Control and ensure RB HVAC isolation and SGT initiation.

D. PPM 5.4.1, Radioactivity Release Control and isolate all primary systems discharging into areas outside PC and SC.

ANSWER: C KA # & KA VALUE: Knowledge of EOP entry conditions and immediate action steps. 2.4.1 (4.6) (CFR 41.10) Tier 3

REFERENCE:

EOP 5.0.10 pages 297 and 298, EOP PPM 5.3.1.

SOURCE: BANK - LR00110 LO: 8017 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: A (incorrect) PPM 5.1.2 is entered on ATWS conditions which plant conditions do not indicate.

B (incorrect) PPM 5.2.1 is entered at 1.68 psig.

C (correct) Entry condition only exists for PPM 5.3.1 due to RB Exhaust being GT 13 mrem/hr.

D. (incorrect) PPM 5.4.1 is entered on high radioactivity release rate which is not given in the stem but is plausible based on the RB exhaust monitor being the monitor given in the stem.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION QUESTION: RO-75 The MSIVs close and an ATWS occurs. Actions per PPM 5.1.2, RPV Control - ATWS, are being performed. Level / power conditions exist. Injection systems have been STOPPED AND PREVENTED (except SLC, RCIC and CRD Pumps) to lower reactor water level.

Which of the following describes why it is necessary to reduce reactor power by further lowering reactor water level?

A. Limit the energy addition to primary containment.

B. Minimize the amount of fuel damage.

C. Simplify the RPV pressure control with the SRVs.

D. To prevent the possibility of power oscillations occurring.

ANSWER: A KA # & KA VALUE: Knowledge of EOP mitigation strategies. 2.4.6 (RO 3.7)(CFR 41.10) Tier 3

REFERENCE:

PPM 5.0.10 Page 137, 145 - 148 SOURCE: BANK - LR01022 LO: 8149 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (correct) RPV level is lowered in an ATWS to reduce reactor power which limits the energy addition to primary containment.

B (incorrect) Plausible as it is something that occur as a result of lower reactor power but the basis for lowering level is to limit energy addition to PC.

C (incorrect) See B D (incorrect) See B

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 01 (76)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION Columbia was in MODE 3 with reactor pressure at 200 psig when the feeder breaker from MC-4A to battery charger HPCS-C1-1 tripped open. During the investigation, the disconnect for HPCS-B1-DG3, the battery supply to 125V DC Bus S1-HPCS, was inadvertently opened.

The CRS will...

A. declare the associated supported required features inoperable immediately and initiate actions to restore required DC electrical power subsystem to operable status immediately.

B. declare HPCS system inoperable immediately, verify RCIC operable by administrative means immediately, and restore HPCS system to operable status in 7 days.

C. restore the distribution system to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in MODE 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

D. restore terminal voltage to GE minimum float voltage within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the required battery charger to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and restore the battery to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

ANSWER: D KA # & KA VALUE: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: System lineups. 295004 AA2.04 (SRO 3.3)

(CFR 43.5) Tier 1 / Group 1

REFERENCE:

TS 3.8.4, 3.8.7; 3.5.1; 3.8.5, 3.5.2, LCO 3.0.6 SOURCE: NEW LO: 7657 LOK / LOD: H/3 HANDOUT: YES - TS 3.8.4; TS 3.8.5; TS 3.5.1; TS 3.5.2 JUSTIFICATION: The stem indicates Mode 3 therefore TS 3.8.4 is applicable. Having breaker from MC-4A opened places 125 VDC bus S1-HPCS on battery power resulting in entry into TS 3.8.4 condition B. When battery disconnect is opened, S1-HPCS becomes de-energized TS 3.8.4 condition E is entered. Due to TS LCO 3.0.6 the cascading into TS 3.8.7 is not required.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 A (incorrect) Incorrect but plausible because this choice uses the DC Sources -

shutdown TS 3.8.5 which is applicable in Mode 4 or 5 (not mode 3).

B (incorrect) If TS 3.8.7 was entered which is not required by TS LCO 3.0.6 HPCS would be declared inoperable.

C (incorrect) Incorrect but plausible because this choice uses the completion time for Div 1 and 2 DC systems from TS 3.8.7.

D (Correct) See above.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 02 (77)

A fire occurs in the Control Room and ABN-CR-EVAC is ordered by the Shift Manager. After the remote shutdown panel is activated the following indications are noted in the RSD Room:

• RPV level -48 inches and down slow.

• RPV Pressure 850 psig and up slow.

• Containment Pressure 10 psig and up slow.

• Containment Temperature 285 F and up slow.

Based on these conditions what directions should the CRS give to the RSD Room Reactor Operator?

A. Sequentially open SRVs to lower RPV pressure to 500 to 600 psig.

B. Start RCIC and maintain level -50 to +54 inches.

C. Spray the Wetwell using RHR B.

D. Spray the Drywell using RHR B.

ANSWER: B KA # & KA VALUE: Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Reactor Water Level. 295016 AA2.02 (SRO 4.3) (CFR 43.5) Tier 1 / Group 1

REFERENCE:

ABN-CR-EVAC Page 5 SOURCE: NEW LO: 6105 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: A (incorrect) Per ABN-CR-EVAC attachment 7.12 RPV pressure is maintained 800 to 1000 psig using SRVs until plant conditions are stabilized.

B (correct) Per ABN-CR-EVAC Level leg RCIC should be started per attachment 7.11 and restored between -50 and +54 inches.

C (incorrect) In PPM 5.2.1 Wetwell sprays would be required, however ABN-CR EVAC has no procedures for spraying the Wetwell.

D (incorrect) In PPM 5.2.1 Drywell sprays would be required, however ABN-CR EVAC has no procedures for spraying the Drywell

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 03 (78)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION With the plant operating in Mode 1, a forklift runs into the Div. 2 ADS Accumulator Backup Compressed Gas Bottle Rack. Current conditions are:

• Four (4) of the Div. 2 bottles are ruptured and depressurized.

• No other bottles or equipment was damaged.

• The other fifteen (15) Div. 2 bottles are at 2400 psig.

• The Backup Nitrogen Bottle in the DG Corridor is 2800 psig.

1) What is the status of the ADS SRVs?

2) What action(s) are required?

A. 1) ADS SRVs remain operable.

2) No Tech Spec entries are required because only 14 bottles are required to be in service.

Replace the empty bottles.

B. 1) ADS SRVs inoperable.

2) Enter Tech Spec 3.5.1 Condition G and be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Place the remote backup nitrogen bottle in service per SOP-CIA-OPS and then exit Tech Spec 3.5.1.

C. 1) ADS SRVs inoperable.

2) Enter Tech Spec 3.5.1 Condition G and be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Place the remote backup nitrogen bottle in service per SOP-CIA-OPS, replace one bottle, and then exit Tech Spec 3.5.1.

D. 1) ADS SRVs inoperable.

2) Enter Tech Spec 3.5.1 Condition G and be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and All 4 bottles must be replaced to exit Tech Spec 3.5.1.

ANSWER: C KA # & KA VALUE: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safety-related instrument air system loads 295019 AA2.02 (SRO 3.7) (CFR 43.5) Tier 1 / Group 1

REFERENCE:

TS 3.5.1, and TS Bases 3.5.1 SR 3.5.1.3 SOURCE: NEW LO: 5153 LOK / LOD: H/4

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 HANDOUT: TS 3.5.1 JUSTIFICATION: This question required knowledge that 17 of 19 bottles for Division 2 be in operable with an average pressure of >2200 psig of the 17 bottles. This knowledge is TS Bases knowledge not systems knowledge.

A (incorrect) If this were Division 1 then only 14 are required to be in service per TS Bases for SR 3.5.1.3. Since this is Division 2 17 are required.

B (incorrect) Because 15 are damaged the minimum number of 17 bottles per TS Bases is not met resulting in All Division 2 ADS SRVs being inoperable. By substituting the DG Corridor bottle for one of the damaged bottles the minimum number of bottles is still not being met even though the average pressure of the 16 good bottles and one depressurized bottle would average greater than 2200 psig therefore substituting with the DG corridor bottle does not restore required number of bottles available to service so TS may not be exited.

C (correct) Because 15 are damaged the minimum number of 17 bottles per TS Bases is not met resulting in All Division 2 ADS SRVs being inoperable. By replacing one bottle and substituting one with the Bottle in the DG Corridor the minimum required of bottles is met with proper pressure so TS may be exited.

D (incorrect) Because 15 are damaged the minimum number of 17 bottles per TS Bases is not met resulting in All Division 2 ADS SRVs being inoperable. Not all 4 bottles MUST be replaced prior to exiting TS. If 2 bottles are replaced then the TS may be exited.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 04 (79)

Columbia is in a refueling outage with a spiral offload of fuel ready to commence.

• The first three fuel bundles to be removed are in the SRM A quadrant.

• SRM A and SRM B are the only two (2) operable SRMs.

• Per the NCTL fuel movement will stop after the last fuel bundle is removed in the SRM B quadrant.

While removing the first fuel bundle, SRM A is damaged.

What actions, if any, are required to be taken?

A. Restore SRM A to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

B. Fully insert all insertable control rods and place reactor mode switch in shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Only one SRM channel is required to be operable therefore no actions per Technical Specifications must be taken.

D. Suspend Core Alterations except for control rod insertion and fully insert all insertable control rods in core cells containing one or more fuel assemblies.

ANSWER: D KA # & KA VALUE: Refueling Accidents - Ability to apply Technical Specifications for a system. 295023 2.2.40 (4.7) (CFR 43.5) Tier 1 / Group 1

REFERENCE:

TS 3.3.1.2; TS Table 3.3.1.2-1 SOURCE: NEW LO: NONE LOK / LOD: H/3 HANDOUT: TS 3.3.1.2; TS Table 3.3.1.2-1 JUSTIFICATION: Requires using Table 3.3.1.2-1 to determine how many SRMs are required to be in operation for this situation. Refueling is a Mode 5 evolution. With a spiral offload in progress, only one SRM channel is required to be operable per Table 3.3.1.2-1 note

b. Per the basis for TS 3.3.1.1, the operable SRM has to be in the quadrant being fueled.

A (incorrect) Would be correct answer if in Mode 2.

B (incorrect) Would be correct answer if in Mode 4.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 C (incorrect) TS 3.3.1.2. Table 3.3.1.2-1 requires 2 SRM channels to be in operation for this situation. Note (b) on Table 3.3.1.2-1 does not apply because SRM A is in the quadrant being refueled.

D (correct) SRM-A is in the region being fueled and as such is a required SRM. Per TS 3.3.1.2, this is the required action.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 05 (80)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION A plant startup is in progress with the following conditions:

• The Mode Switch is in Startup/Hot Standby

• Suppression Pool average temperature is 93°F and down slow due to the performance of OSP-RCIC/IST-C701, RCIC Post Maintenance Testing

• RCIC testing was completed 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ago

• RCIC is operable and in a standby lineup Which of the following describes a limitation in the facility license for these conditions?

The Mode Switch A. CANNOT be placed in RUN due to Suppression Pool average temperature unless a risk assessment is performed and risk management actions are established.

B. CANNOT be placed in RUN with Suppression Pool average temperature greater than 90°F even if a risk assessment is performed.

C. CAN be placed in RUN as long as Suppression Pool average temperature remains less than or equal to 105°F.

D. CAN be placed in RUN, but Suppression Pool average temperature must be less than or equal to 90°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ANSWER: A KA # & KA VALUE: Suppression Pool High Water Temperature - Knowledge of conditions and limitations in the facility license. 295026 2.2.38 (SRO 4.5) (CFR 43.1) Tier 1 /

Group 1

REFERENCE:

TS 3.6.2.1, TS Bases 3.6.2.1, LCO 3.0.4 SOURCE: BANK - 2011 NRC Exam LO: 10308 LOK / LOD: H/2 HANDOUT: TS 3.6.2.1 JUSTIFICATION: A (correct) Normally, the mode switch would be allowed to be placed in RUN

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 without performing a risk assessment. In the stem, it states that SP temperature exceeds the limit per LCO 3.6.2.1 (90°F). LCO 3.0.4b permits a Mode change if an LCO is not met after a risk assessment is performed and risk management actions are established. (correct)

B (incorrect) The Mode Switch CAN be placed in RUN provided the conditions of LCO 3.0.4b are met. This Spec does not include the LCO 3.0.4b does not apply statement. (incorrect)

C (incorrect) The RCIC surveillance is no long in-progress, so the 105°F limit is no longer applicable. LCO 3.6.2.1 is not satisfied at 93°F. (incorrect)

D (incorrect) The Mode Switch can be placed in RUN if LCO 3.0.4b is met, but temperature must be restored in 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, not 24. (incorrect)

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 06 (81)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION An offsite release is in progress when a field team reports that offsite dose surveys indicate 1.2 miles TEDE indicates 125 mrem.

The CRS should declare a A. Site Area Emergency and recommend all schools in the EPZ be evacuated.

B. General Emergency and recommend all schools in the EPZ be evacuated.

C. Site Area Emergency and recommend all areas in EPZ up to 10 miles be evacuated.

D. General Emergency and recommend all areas in EPZ up to 10 miles be evacuated.

ANSWER: A KA # & KA VALUE: High Off-Site Release Rate - Knowledge of emergency plan protective action recommendations. 295038 2.4.44 (SRO 4.4) (CFR 43.5) Tier 1 / Group 1

REFERENCE:

PPM 13.2.2 Attachment 7.1, PPM 13.1.1 Attachment 7.1 Table 4 SOURCE: NEW LO: 10189 LOK / LOD: H/3 HANDOUT: PPM 13.2.2 Attachment 7.1, PPM 13.1.1 Attachment 7.1 Table 4 JUSTIFICATION: A (correct) Dose rate for TEDE indicate SAE should be declared, and automatic protective action recommendations are to evacuate all schools in EPZ.

B (incorrect) Dose rate for TEDE indicate SAE not a GE.

C (incorrect) Dose rate for TEDE indicate SAE should be declared, however all areas in EPZ are not evacuated until a GE is declared.

D (incorrect) Dose rate for TEDE indicate SAE not a GE.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 07 (82)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION Columba is operating in Mode 1 with SM-2 being powered from TR-S due to performance of OSP-ELEC-M703 (HPCS DG Monthly Operability Test). DG-3 has been started in preparation for being paralleled to SM-4.

The BPA dispatcher has just notified the CRS that, due to continuing grid disturbances, TR-S voltage is 229kv and can NOT be restored to within Technical Specification limits. The CRS updates the crew and enters ABN-ELEC-GRID.

OPS2 then reports steady state voltage for DG3 is reading 3900 volts.

What actions are required?

A. Only enter TS 3.8.1.A and restore offsite circuit to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B. Only enter TS 3.8.1.B and restore DG-3 to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

C. Enter both TS 3.8.1 A and TS 3.8.1 B and restore offsite circuit or DG-3 to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. Enter TS 3.8.1.D and restore offsite circuit or DG-3 to operable status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ANSWER: D KA # & KA VALUE: Generator Voltage and Electric Grid Disturbances - Ability to determine operability and/or availability of safety related equipment. 700000 G2.2.37 (4.5) (CFR 43.5)

Tier 1 / Group 1

REFERENCE:

TS 3.8.1 SOURCE: NEW LO: 3995 LOK / LOD: H/3 HANDOUT: TS 3.8.1 JUSTIFICATION: Entry into ABN-ELEC-GRID is required based on BPA dispatch report. SR 3.8.1.1 requires generator voltage to be GE 3910 volts for the DG to be operable. Both TR-S and DG-3 are not operable based on the stem.

A (incorrect) TS 3.8.1 condition A would be correct if only TR-S is inoperable.

B (incorrect) TS 3.8.1 condition B would be correct if only DG-3 was inoperable.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 C. (incorrect) TS 3.8.1 condition A and B applies but Condition D is more restrictive.

D. (correct) TS 3.8.1 condition D is entered based on TR-S and DG-3 being inop.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 08 (83)

A reactor Scram has been initiated. The only actions taken were to turn the mode switch to shutdown, and insert the SRMs and IRMs. The Reactor Operators report the following conditions:

• MSIVs are closed.

• RPV level is -25 inches and trending down slow.

• Two SRVs are open.

• RPV pressure is 1030 psig and trending up slow.

• Full Core display is de-energized.

• RWM screen is blank.

• APRM downscale lights are illuminated.

• IRMs are reading 25 on range 10.

Which of the following statements is correct and what action should the CRS direct?

A. IRM readings are correct. Direct depressing the manual scram pushbuttons, initiating ARI, and injecting SLC per PPM 5.5.25.

B. IRM readings are correct. Direct depressing the manual scram pushbuttons, initiating ARI, and injecting SLC per SOP-SLC-INJECTION-QC.

C. APRM downscale lights are correct. Direct lowering reactor pressure to 500-600 psig to allow feeding with condensate booster pumps per SOP-RFW-FCV-QC.

D. APRM downscale lights are correct. Direct starting RCIC and restore level -50 to +54 inches per SOP-RCIC-INJECTION-QC.

ANSWER: B KA # & KA VALUE: Incomplete Scram - Ability to identify and interpret diverse indications to validate the response of another indication. 295015 2.1.45 (SRO 4.3) (CFR 43.5) Tier 1 /

Group 2

REFERENCE:

PPM 5.0.10 pg. 185, PPM 5.1.2 RPV Control-ATWS SOURCE: NEW LO: NONE LOK / LOD: H/3 HANDOUT: NONE

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 JUSTIFICATION: Two SRVs being open and RPV pressure going up slowly is an indication that an ATWS condition exists and the CRS should direct actions based on this. APRM downscale lights come in at 5% power and lower.

IRMs reading 25 on range 10 indicates approximately 10% power.

Each SRV has a capacity of approximately 5 to 6 percent reactor power. With two SRVs open and pressure slowly increasing Reactor power must be above 10% and therefore an ATWS condition exists. Per PPM 5.0.10 page 185 When the boron tank is emptied, it may be refilled with demineralized water per PPM 5.5.25 and injection continued.

A (incorrect) PPM 5.5.25 is not directed until the SLC tank is emptied in PPM 5.1.2 but would be directed in PPM 5.1.1 for level control.

B (correct) Correct ATWS conditions exist and SLC should be injected.

C (incorrect) Reducing pressure could result in a power spike and core damage.

D (incorrect) Level control strategy for ATWS condition would be to lower PRV level to below -65 inches to reduce power.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 09 (84)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION Columbia is in Mode 5 with the following conditions:

• OPDRV in progress

• RBHVAC shutdown for maintenance

• RB 471 west end general area temperature is 106° F for the last 90 minutes Which of the following actions is correct for this condition?

A. Initiate action to restore the area to within the limits of Condition B in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B. Initiate action to restore the area to within 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 />.

C. The required actions 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 resulting from a maintenance issue.

D. Restore the area to within the limits of Condition C within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or suspend operations with a potential for draining the reactor vessel in immediately.

ANSWER: D KA # & KA VALUE: Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Equipment operability.

295032 EA2.02 (SRO 3.5) (CFR 43.5) Tier 1 / Group 2

REFERENCE:

LCS 1.7.1 and TS 3.6.4.3 SOURCE: BANK - MODIFIED - 2009 NRC Exam LO: 9540 LOK / LOD: H/4 HANDOUT: YES - LCS 1.7.1 pages -1 through 15 and TS 3.6.4.3 JUSTIFICATION: LCS 1.7.1 RB Support Area/Rooms - 471 Open Areas (not elsewhere listed) Cond C Temp Limit is LE 104°F (106° given in stem). Note says to go to table 1.7.1-2.

Page 5 of 6 indicates that for RB 471 Open Area the equipment affected is SGT Div 1 and 2 and refers to LCO/RFO 3.6.4.3. LCO 3.6.4.3 requires suspending OPDRVs.

A (incorrect) Per LCS 1.7.1 you immediately initiate actions to restore temp.

B (incorrect) The time limit to restore is one hour.

C (incorrect) Per note at top of table 1.7.1-1, the four hours is allowed for surveillances. This temp is due to loss of ventilation.

D (correct) The one hour is allowed per LCS Action C.1 and LCO 3.6.4.3 requires suspending OPDRVs.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 10 (85)

During testing of the Standby Service Water System, the Reactor Building Equipment Sump R4 Level Hi-Hi alarm annunciates and clears five minutes later. OPS 2 investigates and reports RHR C pump room floor is dry with minimal water in the sump but that he was unable to open either door to the LPCS pump room and that water was seeping from each doors seal. PPM 5.5.27, RB 422 Max Safe Operating Level Measurement, has not been directed.

The CRS enters PPM 5.3.1, Secondary Containment Control, and ABN-FLOODING.

Based on the above, what actions will the CRS direct`?

A. Start RHR-P-2A and LPCS-P-1. Both RHR-P-2A and LPCS-P-1 remain operable.

B. Pull control power fuses for LPCS-P-1 and RHR-P-2A. Declare RHR-P-2A and LPCS-P-1 inoperable and enter Tech Spec 3.5.1 condition C.

C. Start RHR-P-2A in SPC and pull the control power fuses for LPCS-P-1. Declare both RHR-P-2A and LPCS-P-1 inoperable and enter Tech Spec 3.5.1 condition C.

D. Start RHR-P-2A in SPC and pull the control power fuses for LPCS-P-1. Only declare LPCS-P-1 inoperable and enter Tech Spec 3.5.1 condition A.

ANSWER: C KA # & KA VALUE: Ability 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. 295036 EA2.01 (SRO 3.2) (CFR 43.5) Tier 1 /

Group 2

REFERENCE:

ABN-FLOODING page 11, SOP-RHR-SPC page 9 SOURCE: NEW LO: 10295, 9540 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) RHR and LPCS pumps are declared inoperable once they are started.

B (incorrect) RHR-P-2A should be started prior to receiving the low discharge pressure alarm.

C (correct) Per ABN-FLOODING, RHR-P-2A is started and fuses for LPCS-P-1 pulled in anticipation of losing LPCS-P-2. RHR-P-2A is started in SP Cooling per SOP-RHR-SPC and that procedure states to enter RHR-SYS-A inoperable.

D (incorrect) RHR-P-2A is declared inoperable.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 11 (86)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION While operating in Mode 1 with LPCS-P-1 running for performance of OSP-LPCS/1ST-Q702 (LPCS System Operability Test), it is noticed that the LPCS pump room cooler RRA-CC-5 (LPCS Pump Room Fan Cooler Assembly) outlet isolation valve, SW-V-37 (RRA-CC-5 Return Isolation) is closed.

With SW-V-37 closed and LPCS pump room temperatures reaching 152 F, what is the impact on the LPCS system and what actions, if any, are required?

The LPCS Pump A. has exceeded its environmental qualification temperature limit and has not been evaluated under these conditions to function as designed. TS 3.5.1 will be entered immediately and LPCS pump will be declared inoperable.

B. has exceeded its environmental qualification temperature limit and has not been evaluated under these conditions to function as designed. LCS 1.7.1 will be entered and if temperatures are not restored to allowable limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS 3.5.2 will be entered declaring LPCS inoperable.

C. will continue to meet the environmental qualifications to perform its safety function during a design basis event conditions and therefore the loss of the pump room cooler will have no impact on the LPCS pump. No actions are required.

D. has exceeded its environmental qualification temperature limit and has not been evaluated under these conditions to function as designed. Since surveillance testing is being performed entry into LCS 1.7.1 can be delayed for four hours.

ANSWER: A KA # & KA VALUE: Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of room cooling. 209001 A2.07 (SRO 2.8) (CFR 43.2) Tier 2 / Group 1

REFERENCE:

LCS 1.7.1 and LCS Bases 1.7.1, OI-41 pg. 51 SOURCE: NEW LO: NONE LOK / LOD: H/4

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 HANDOUT: LCS 1.7.1, OI-41 pg. 51 JUSTIFICATION: The environmental qualification limits of the FSAR are the Table 1.7.1-1 temperatures per the LCS 1.7.1 Bases.

A (correct) The LPCS pump has exceeded its environmental qualification temperature limit of 150 F, and OI-41 page 51 requires TS 3.5.1 entered immediately and LPCS pump declared inoperable.

B (incorrect) Plausible if OI-41 is not referenced. The LPCS pump has exceeded its environmental qualification temperature limit of 150 F. LCS 1.7.1 condition A, B, and C will be entered, and after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of exceeding the temperature limits of condition C, TS 3.5.1 will be entered and LPCS declared inoperable.

C (incorrect) The LPCS pump has exceeded its environmental qualification temperature limit of 150 F.

D (incorrect) The LPCS pump has exceeded its environmental qualification temperature limit of 150 F. The Note on the top of table 1.7.1-1 does not apply to condition C temperatures if exceeded so the four hour delay is not granted for the LCS.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 12 (87)

While operating in Mode 1, the HPCS Fill Verification surveillance, OSP-HPCS-M101, is in progress.

During the surveillance, the HPCS WATER LEG PUMP DISCH PRESS LOW annunciator alarms.

What action is required, if any?

A. This is an expected alarm during the performance of the surveillance. No compensatory measures are required, continue with the surveillance.

B. Immediately verify RCIC operable by administrative means and restore HPCS to operable status within 7 days. Inhibit HPCS-P-1 from starting by pulling the control power fuses.

C. HPCS was declared inoperable to perform this surveillance. Tech Spec 3.5.1 has already been entered. No additional actions are required per tech. specs.

D. If HPCS is required to be operable, a dedicated operator is stationed at HPCS-V-88 (HPCS vent valve) and is in communication with the main control room.

ANSWER: C KA # & KA VALUE: HPCS - Knowledge of surveillance procedures. 209002 2.2.12 (SRO 4.1) (CFR Tier 2 / Group 1

REFERENCE:

OSP-HPCS-M101, TS 3.5.1 SOURCE: NEW LO: 5432 LOK / LOD: F/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) While the surveillance does state this alarm may annunciate during the performance of the surveillance, it is not an expected alarm and actions do need to be carried out if it does alarm.

B (incorrect) Tech Specs require 14 days for HPCS to be returned to service by.

C (correct) TS 3.5.1 was entered to perform the surveillance.

D (incorrect) An operator cannot maintain operability during the performance of the surveillance.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 13 (88)

Columbia is operating at 100% power, when an unisolable break develops in the steam supply line to RCIC.

The LEAK DET RCIC PIPE ROUTING AREA TEMP HI-HI alarm is received and in response the CRS enters PPM 5.3.1 Secondary Containment Control.

1) If temperature in that area continues to increase, what actions are required to be taken?

2) What is the status of RCIC-V-1 (RCIC Turbine Trip MO Valve)?

A. 1) Shutdown the Reactor per PPM 3.2.1 Normal Plant Shutdown.

2) RCIC-V-1 is closed.

B. 1) Enter PPM 5.1.1 RPV Control and scram the Reactor.

2) RCIC-V-1 is closed.

C. 1) Shutdown the Reactor per PPM 3.2.1 Normal Plant Shutdown.

2) RCIC-V-1 is open.

D. 1) Enter PPM 5.1.1 RPV Control and scram the Reactor.

2) RCIC-V-1 is open.

ANSWER: B KA # & KA VALUE: Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Steam line break. 217000 A2.15 (SRO 3.8) (CFR 43.5)

Tier 2 / Group 1

REFERENCE:

PPM 5.3.1 Secondary Containment Control; ARP - 4.601.A2 Tile 1-1 page 5 SD000180 pg 20, and 26.

SOURCE: NEW LO: 11675; 6906; 5722 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: The LEAK DET RCIC PIPE ROUTING AREA TEMP HI-HI alarm is activated by LD-MON-1B at a setpoint of 160 °F and causes automatic isolation of RCIC-V-1.

The LEAK DET RWCU/RCIC PIPE AREA TEMP HIGH alarm is activated by LD-MON-1B at a setpoint of 140 °F but causes no automatic actions.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 Student must realize which alarm causes the isolation signal to activate.

A (incorrect) One parameter exceeds its Max Safe Value but only in a single area.

Valve isolates on HI-HI temp alarm setpoint.

B (correct) One parameter exceeds its Max Safe with a primary system discharging into the secondary containment. Valve isolates on HI-HI temp alarm setpoint.

C (incorrect) One parameter exceeds its Max Safe Value but only in a single area.

Valve isolates on HI-HI temp alarm setpoint.

D (incorrect) One parameter exceeds its Max Safe with a primary system discharging into the secondary containment. Valve isolates on HI-HI temp alarm setpoint.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 14 (89)

While operating at 85% power for economic dispatch, a maintenance worker asks for permission to perform work on an air leak in the air start motor for DG3.

What assessment tools will the CRS use to determine the overall aggregate risk for this evolution prior to authorizing maintenance?

A. OI-14, Columbia Generating Station Operational Challenges and Risk Program and the Operator Aggregate Impact Index Performance Indicator.

B. PPM 1.16.6B, Voluntary Entry Into T.S. Action Statements, LCS Requirements for Operability, and the Plant Logging System.

C. 1.3.42 Troubleshooting Plant Systems and Equipment, and the Technical Issues Resolution Process.

D. PPM 1.5.14 Risk Assessment and Management for Maintenance / Surveillance Activities, and the Paragon Program.

ANSWER: D KA # & KA VALUE: EDGs - Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. 264000 2.2.17 (SRO 3.8) (CFR 43.5) Tier 2 /

Group 1

REFERENCE:

PPM 1.5.14 page 4 SOURCE: New LO: 13394 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: A (incorrect) OI-14 is the outline for the program for identifying, tracking, and correcting deficiencies which would impact Operations Department in the operations of the plant.

B (incorrect) PPM 1.16.6B, Directs how to do a voluntary Entry for TS and LAC which would be done for this evolution, while authorizing the maintenance.

C (incorrect) PPM 1.3.42 establishes guidance for control of troubleshooting activities and the Technical Issues Resolutions Process is started by Operations Department to get Engineering assessments of issues.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 D (correct) PPM 1.5.14 institutes the stations commitment and major responsibilities for implementing M-Rule section (a)(4) requirement to perform a risk assessment before maintenance activities are to be performed. Normally performed by work control, however the CRS is required to ensure Paragon is updated to reflect all maintenance.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 15 (90)

The Alternate Power Supply to IN-1 tagged out for maintenance.

An event causing the Normal Power Supply voltage to IN-1 to be 106% of the normal voltage.

What is the immediate impact to IN-1?

What will the CRS direct in this situation?

A. IN-1 static switch will transfer to the Alternate AC source resulting in a loss of power to US-PP.

Enter ABN-POWER and reduce RRC flow. Transfer the inverter to Bypass source using the Kirk Key Interlock using SOP-ELEC-IN1-OPS.

B. IN-1 static switch will transfer to the Maintenance Bypass source maintaining power to US-PP.

When the Overvoltage condition clears recover the inverter per SOP-ELEC-IN1-OPS.

C. IN-1 static switch will transfer to the Alternate AC source resulting in a loss of power to US-PP.

Enter ABN-POWER and reduce RRC flow. Enter ABN-ELEC-INV, but do not restore power to US-PP until the overvoltage condition is clear.

D. IN-1 static switch will transfer to the Maintenance Bypass source maintaining power to US-PP.

Transfer the inverter to Bypass Source using the Kirk Key Interlock using SOP-ELEC-IN1-OPS.

ANSWER: A KA # & KA VALUE: Ability to (a) predict the impacts of the following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Over voltage. 262002 A2.02 (SRO 2.7) (CFR 41.5 / 45.6) Tier 2 / Group 1

REFERENCE:

SD000194 Page 8, SOP-ELEC-IN1-OPS pg. 11-13; ABN-ELEC-INV pg. 7 SOURCE: NEW LO: NONE LOK / LOD: H/4 HANDOUT: NONE

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 JUSTIFICATION: RO knowledge is to know the impact on the UPS unit to overvoltage conditions.

Static Switch for IN-1 will transfer to the alternate source if it senses an overvoltage condition set at 105% of the normal voltage. IN-2 and 3 do not have this transfer feature.

SRO only knowledge is required to correct the condition using procedures.

SOP-ELEC-IN1-OPS have the procedural steps to transfer to the DC power supply and to transfer using the Kirk Key to the Bypass source. The SRO knowledge is to know that the 105% voltage transfer is to protect the Inverter not US-PP.

A (correct) See above B (incorrect) The Maintenance Bypass source is the second option beyond the Static Switch but requires a manual transfer.

C (incorrect) The CRS should repower US-PP if a source of power is available.

D (incorrect) The Maintenance Bypass source is the second option beyond the Static Switch but requires a manual transfer.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 16 (91)

The CRS directs power to be raised with flow from 86% to 100% following economic dispatch.

Initial conditions are:

Master Controller (RRC-M/A-R675) indicates 46 Hz Loop A and B Drive Flow (RRC-FI-R676A/B) indicates 33,000 GPM Individual Loop A and B Controllers (RRC-M/A-R676A and RRC-M/A-676B) are in AUTO indicating:

• 46 Hz Actual

• 46 Hz Demand

• 0 Hz Deviation / Bias When raising flow, the Individual Loop B Controller raise pushbutton is inadvertently depressed and the pushbutton sticks in the depressed position. The Reactor Operator does not identify the problem until after RRC-P-1Bs speed stops increasing.

What effect does the stuck pushbutton have on the RRC system, and what action will the CRS direct to mitigate the event?

Loop A Drive Flow remains the same and Loop B Drive Flow will increase A. until the Deviation / Bias meter indicates 6 Hz and then stop resulting in a Loop B Drive Flow of approximately 37,000 GPM.

The CRS will direct stopping B RRC pump and enter ABN-RRC-LOSS.

B. until the Deviation / Bias meter indicates 6 Hz and then stop resulting in a Loop B Drive Flow of approximately 37,000 GPM.

The CRS will direct lowering B RRC pump speed to 46 Hz.

C. for 15 seconds resulting in a Loop B Drive Flow of approximately 44,000 GPM.

The CRS will direct stopping B RRC pump and enter ABN-RRC-LOSS.

D. for 15 seconds resulting in a Loop B Drive Flow of approximately 44,000 GPM.

The CRS will direct lowering B RRC pump speed to 46 Hz.

ANSWER: B KA # & KA VALUE: Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadvertent recirculation flow increase. 202001 A2.05 (SRO 4.0) (CFR 43.5) Tier 2 / Group 2

REFERENCE:

Simulator, SD000184 pages 10-13, ABN-POWER

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 SOURCE: NEW LO: 11788 LOK / LOD: H/4 HANDOUT: NONE JUSTIFICATION: A (incorrect) With the flow controllers in AUTO the Deviation / Bias will be limited to 6 Hz and stop the increase in flow. Since the rise in pump speed is stopped and is now under control it is not required to be secured.

B (correct) With the flow controllers in AUTO the Deviation / Bias will be limited to 6 Hz and stop the increase in flow. ABN-POWER directs reducing the RRC flow to pre-transient value.

C (incorrect) If Individual Loop Controller was in Manual, then the flow increase would continue for 15 seconds (Stuck Pushbutton Timer), resulting in a frequency of approximately 60.5 Hz and 44,000 GPM flow. Since the rise in pump speed is stopped and is now under control it is not required to be secured.

D (incorrect) If Individual Loop Controller was in Manual, then the flow increase would continue for 15 seconds (Stuck Pushbutton Timer), resulting in a frequency of approximately 60.5 Hz and 44,000 GPM flow.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 17 (92)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION Columbia is operating in Mode 1 when the following occurs:

1st of month at 0900 - RHR-P-2A failed to start during the RHR Loop A Operability Test (OSP-RHR/1ST-Q702).

5th of month at 1000 - RHR-V-17B (RHR outboard Drywell spray) is declared INOP and is deactivated in the closed position due to having a wrong sized motor installed.

5th of month at 1600 - The cause of RHR-P-2As failure to start is corrected, and after testing, is declared operable.

What is the longest time that RHR-V-17B can remain INOP before Columbia must be in MODE 3?

A. 8th of the month at 2100 B. 9th of the month at 2100 C. 12th of the month at 2200 D. 13th of the month at 2200 ANSWER: B KA # & KA VALUE: RHR/LPCI: Containment Spray System Mode - Ability to apply Technical Specifications for a system. 226001 2.2.40 (SRO 4.7) (CFR 43.2) Tier 2 / Group 2

REFERENCE:

Tech Spec 3.6.1.5, Tech Spec 1.3 Completion Times SOURCE: NEW LO: 5783 LOK / LOD: H/3 HANDOUT: YES - TS 3.6.1.5 JUSTIFICATION: A (incorrect) This would be correct if 0900 were used as start point and the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension was not applied as allowed by TS 1.3 Completion Times.

B (correct) TS 3.6.1.5.A is entered for RHR A at time 0900 on the 1st. TS 3.6.1.5.A for RHR B is entered at 1000 on the 5th and TS 3.6.1.5.B is entered at 1000 on the 5th for loss of both spray systems. When RHR A is restored TS 3.6.1.5.B is exited,

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 but TS 3.6.1.5A remains in effect from original time entered and per TS 1.3 Completion Times an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> may be added to the completion time making the completion time 8 days, and add in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to be in mode 3 the answer is 2100 on the 9th of the month.

C (incorrect) If the time RHR-V-17B becomes inoperable is used then this would be correct.

D (incorrect) If the time RHR-V-17B becomes inoperable is used and adding in the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension then this would be correct.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 18 (93)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION With refueling operations in progress, an operator using the Monorail Hoist wants to bypass the Normal Up Interlock to change a tool. When reviewing if this is acceptable, it is noticed that the last time OSP-NSSE-W401 (Refuel Equipment Interlocks CFT) was performed was 7 days and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

1) Can the Normal Up Interlock be bypassed without violating Technical Specifications?

2) Can refueling operations continue?

A. 1) Yes it can be bypassed because it is not an interlock required by TS 3.9.1.

2) No, the Surveillance is overdue because it was required to be performed once every 7 days. TS 3.9.1.A is required to be entered and fuel movement must be suspended immediately.

B. 1) No, it is required to be operable and to change the tool a voluntary entry into TS 3.9.1 must be made and all fuel movement must be suspended.

2) No, the Surveillance is overdue because it was required to be performed once every 7 days. TS 3.9.1.A is required to be entered and fuel movement must be suspended immediately.

C. 1) Yes it can be bypassed because it is not an interlock required by TS 3.9.1.

2) Yes, refueling operations may continue, however the surveillance is overdue and must be completed within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

D. 1) No, it is required to be operable and to change the tool a voluntary entry into TS 3.9.1 must be made and all fuel movement must be suspended.

2) Yes, refueling operations may continue, however the surveillance is overdue and must be completed within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ANSWER: C KA # & KA VALUE: Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUIPMENT: Fuel handling equipment interlocks. 234000 K5.02 (SRO 3.7) (CFR 43.2) Tier 2 / Group 2

REFERENCE:

TS 3.9.1, TS SR 3.0 Surveillance Requirement Applicability SOURCE: NEW LO: 5362 LOK / LOD: H/3

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 HANDOUT: TS 3.9.1, OSP-NSSE-W401 page 3 JUSTIFICATION: A (incorrect) 1) is true the Normal Up Interlock is not a TS required interlock, 2) is false per SR 3.0.2 there is a 1.25 times the interval specified in the frequency as measured from the last performance so it must be done within 8.75 days.

B (incorrect) 1) is false the Normal Up Interlock is not a TS required interlock, 2) is false per SR 3.0.2 there is a 1.25 times the interval specified in the frequency as measured from the last performance so it must be done within 8.75 days.

C (correct) 1) is true the Normal Up Interlock is not a TS required interlock, 2) is true per SR 3.0.2 there is a 1.25 times the interval specified in the frequency as measured from the last performance so it must be done within 8.75 days.

D (incorrect) 1) is false the Normal Up Interlock is not a TS required interlock, 2) is true per SR 3.0.2 there is a 1.25 times the interval specified in the frequency as measured from the last performance so it must be done within 8.75 days.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 19 (94)

Maintenance on a system has just completed that affected a valve on the Locked Valve Checklist, .1 of PPM 1.3.29. The valve was not verified in its locked position following maintenance, however the valve can be verified closed by plant parameters. The valve is located in a Radiation Area where the exposure to one individual checking the position of the valve will be approximately 12 mrem.

Which of the following is required to verify the position of the valve?

A. Perform a partial valve lineup using the Locked Valve Checklist and the independent verification may be waived by the Shift Manager by signing the comments section of Attachment 7.1.

B. Perform a partial valve lineup using the Locked Valve Checklist and the independent verification must be performed.

C. Perform a partial valve lineup using the Locked Valve Checklist and complete Attachment 7.2 (Deviation from Locked Valve Checklist) waiving the independent verification requirement for the valve above approved by the Shift Manager.

D. Performance of a valve lineup can be waived by the Shift Manager until after the next outage by completing Attachment 7.2 (Deviation From Locked Valve Checklist).

ANSWER: A KA # & KA VALUE: Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. 2.1.29 (SRO 4.0) (CFR 45.1 / 45.12) Tier 3

REFERENCE:

PPM 1.3.29 SOURCE: NEW LO: 3038 LOK / LOD: F/4 HANDOUT: NONE JUSTIFICATION: Per PPM 1.3.29 Page 4 When significant radiation exposures are likely to occur as a result of performing an independent verification, the independent verification can be waived by the Shift Manager. As a guideline, an estimated exposure of 5 mrem for a single (one valve) verification task is considered a significant radiation exposure. However, in these situations an alternate means for independent verification that does not involve radiation exposure (e.g., observing process parameters) should be considered.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 A (correct) The waiver of the independent verification is a approved by the Shift manager by putting a note in the IV block and then writing the waiver for independent verification in the comments section.

B (incorrect) The independent verification may be waived.

C (incorrect) The Independent verification is not a deviation from the intended valve position. The Attachment 7.2 is not used for this case.

D (incorrect) If the valve were in a High Radiation Area and the valve not checked at all then this would be true.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 20 (95)

When in Mode 5, which of the following evolutions must have a Core Alteration Supervisor present?

A. Raising an irradiated fuel bundle in the fuel prep machine.

B. Movement of an irradiated fuel bundle in the spent fuel pool.

C. Movement of individual fuel pins for in-pool inspection of irradiated fuel assemblies.

D. Movement of a new fuel bundle in the spent fuel pool.

ANSWER: B KA # & KA VALUE: Knowledge of fuel handling responsibilities of SROs. 2.1.35 (3.9) (CFR 43.7) Tier 3

REFERENCE:

SWP-RXE-01 pg. 21, PPM 6.3.2 pg. 13, 28 SOURCE: NEW LO: 3038 LOK / LOD: F/4 HANDOUT: NONE JUSTIFICATION: Per PPM 6.3.2 The Core Alt Supervisor shall hold a Senior Operators License and shall have no other concurrent responsibilities during refueling.

Per SWP-RXE-01 The Core Alt Supervisor (RFAE) shall directly supervise all movement of irradiated fuel during Mode 5. Fuel movements in other Modes shall be directly supervised by either a Core Alt Supervisor (RFAE) or a Spent Fuel Pool Supervisor (RFAF) (except not in Mode 5). The only exceptions to this requirement are 1) the movement of individual fuel pins or subassemblies associated with in-pool inspection of irradiated fuel assemblies, and 2) the raising or lowering of irradiated fuel bundles in the fuel prep machines. Supervision of these activities by either a Core Alt Supervisor (RFAE) or a Spent Fuel Pool Supervisor (RFAF) is not required.

A (incorrect) See SWP-RXE-01 Statement above.

B (correct) See SWP-RXE-01 Statement above.

C (incorrect) See SWP-RXE-01 Statement above D (incorrect) See SWP-RXE-01 Statement above.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 21 (96)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION An AR EVAL is received that requires a Temporary Configuration Change (TCC). The following information is known about the TCC:

• A piece of scaffolding was dropped and struck the coupling shroud for FPC-P-3 (Suppression Pool Cleanup Pump) in RHR A pump Room. The coupling shroud was deformed and is resting against the pump coupling.

• A temporary coupling cover made of sheet metal has been made to replace the existing cover until a new cover can be manufactured using drawings which will take approximately two months to fabricate.

Given the information about the TCC and Attachment 9.1 of SWP-CM-02 what type of Temporary Configuration Change (TCC) will this be performed as?

This will be performed as a ..

A. Temporary Modification (TM) per Section 5.3.

B. Temporary Alteration in Support of Maintenance (TASM) per Section 5.5.

C. Compensatory Measure (non-intrusive) (CM) per Section 5.4.

D. Procedurally Controlled Temporary Change (PCTC) per Section 5.6.

ANSWER: C KA # & KA VALUE: Knowledge of the process for making design or operating changes to the facility.

2.2.5 (SRO 3.2) (CFR 43.3) Tier 3

REFERENCE:

SWP-CM-02 page 11 and 22 SOURCE: NEW LO: 13336 LOK / LOD: H/4 HANDOUT: Attachment 9.1 of SWP-CM-02 JUSTIFICATION: (correct) Compensatory Measure is defined as a TCC which:

is needed for to address a degraded or non-conforming condition AND the change is non-intrusive to the design function of any SCC

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 AND does not adversely impact nearby or associated SSCs AND does not affect the FSAR, TOP-TIER drawings, or procedures (incorrect) TASM, TM, and PCTC are all other types of Temporary Configuration Changes (TCCs) outlined in SWP-CM-02 meeting other criteria for selection.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 22 (97)

USE REFERENCES PROVIDED TO ANSWER THIS QUESTION While transferring SM-1 to the startup transformer for the monthly diesel surveillance test, S-1 fails to close.

The S-1 breaker is declared inoperable.

Which of the following Technical Specifications, if any, should be entered?

A. No Tech Spec LCO is applicable.

B. Enter Tech Spec LCO 3.8.1A.

C. Enter Tech Spec LCO 3.8.1D.

D. Enter Tech Spec LCO 3.8.7A.

ANSWER: B KA # & KA VALUE: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. 2.2.25 (SRO 4.2) (CFR 43.2) Tier 3

REFERENCE:

TS Bases page B3.8.1-4 SOURCE: Bank LO: 12420 LOK / LOD: H/3 HANDOUT: TS 3.8.1 and TS 3.8.7 JUSTIFICATION: A (incorrect) There are no specific tech specs for CB-S-1 but there are for the situation of CB-S1 not closing in current plant condition.

B (correct) The bases for TS 3.8.1A states that TR-S must be capable of powering SM-4 and SM-7 and SM-8.

C (incorrect) If candidates decides the failure of CB-S1 causes off site to be inop and a DG to be inop this tech spec would apply.

D (incorrect) If the Div. 1 subsystem is declared inoperable due to CB-S1 breaker then this tech spec would apply.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 23 (98)

A batch of Reactor Closed Cooling (RCC) water has to be discharged following maintenance on the system.

It has been sampled and is acceptable for release.

What procedure is used to authorize the release and who must approve the release?

A. PPM 16.10.1A, Radioactive Liquid Waste Discharge to the River; Control Room Supervisor / SM B. PPM 16.10.1A, Radioactive Liquid Waste Discharge to the River; Chemistry Supervisor C. PPM 12.2.14 Release of Liquid from Nonradioactive Systems; Control Room Supervisor / SM D. PPM 12.2.14 Release of Liquid from Nonradioactive Systems; Chemistry Supervisor ANSWER: C KA # & KA VALUE: Ability to approve release permits. 2.3.6 (SRO 3.8) (CFR 43.4) Tier 3

REFERENCE:

PPM 12.2.14 Pages 6 and 8 SOURCE: NEW LO: 11260 LOK / LOD: H/2 HANDOUT: NONE JUSTIFICATION: PPM 12.2.14 is used for systems entirely contained in the RCA, but believed to be radiologically clean.

A (incorrect) Possible if student believes the system to be contaminated.

B (incorrect) Possible if student believes the system to be contaminated, but the Chemistry Supervisor signs stating that no identifiable radioactivity is found when counted.

C (correct) Per PPM 12.2.14, the CRS or the SM approves the release.

D (incorrect) PPM 12.4.14 is used but the Chemistry Supervisor signs stating that no identifiable radioactivity is found when counted.

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 24 (99)

With one rod withdrawn in Mode 5 SR 3.9.2.1 requires the mode switch to be verified in the locked position every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in order to meet the requirements of TS 3.9.2 Refuel Position One-Rod-Out Interlock.

1) How does the Control Room ensure this requirement is met?

2) What does this interlock protect against?

A. 1) By verifying the mode switch is locked in the Refuel position and documenting it in the Control Room Operators Log.

2) A prompt reactivity excursion during refueling which could potentially result in fuel failure with subsequent release of radioactive material to the environment.

B. 1) By ensuring OSP-NSSE-W402 (Refuel Position One-Rod-Out Interlock CFT) is completed and documenting it in the Maintenance / Surveillance Log.

2) A prompt reactivity excursion during refueling which could potentially result in fuel failure with subsequent release of radioactive material to the environment.

C. 1) By verifying the mode switch is locked in the Refuel position and documenting it in the Control Room Operators Log.

2) A prompt criticality event which could result in high radiation exposure to operators on the refueling floor.

D. 1) By ensuring OSP-NSSE-W402 (Refuel Position One-Rod-Out Interlock CFT) is completed and documenting it in the Maintenance / Surveillance Log.

2) A prompt criticality event which could result in high radiation exposure to operators on the refueling floor.

ANSWER: A KA # & KA VALUE: Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. 2.3.12 (SRO 3.7) (CFR 43.6) Tier 3

REFERENCE:

PPM 6.3.2 page 10 TS 3.9.2 and TS bases 3.9.2 SOURCE: NEW LO: 13287 LOK / LOD: H/4 HANDOUT: NONE JUSTIFICATION: Per PPM 6.3.2 The reactor mode switch shall be verified to be locked in the refuel

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 position prior to a control rod being withdrawn (TS 3.9.2) by documenting this condition in the Control Room Operators Log.

Per TS bases 3.9.2 A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment.

A (correct) meets both conditions called out above.

B (incorrect) OSP-NSSE-W402 is the surveillance which covers the Channel Functional Test for TS 3.9.2. and does include steps which verify the mode switch is in the Refuel position but is not performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C (incorrect) a prompt criticality event could result in radiation exposure to personnel on the refuel floor, however the TS bases calls out the FSAR analysis consequences are release to the environment.

D (incorrect) due to both reasons called out in B and C

COLUMBIA GENERATING STATION NRC WRITTEN EXAMINATION APRIL 2015 QUESTION: SRO - 25 (100)

Concerning the systems listed in 50.72(b)(3)(iv)(B), which of the following lists events that are BOTH reportable to the NRC per PPM 1.10.1, Notifications and Reportable Events?

An actuation that occurs as the result of A. an intentional manual initiation based on plant conditions; an intentional manual initiation as part of a pre-planned sequence during testing.

B. an intentional manual initiation as part of a pre-planned sequence during testing; actual plant conditions with no testing in progress.

C. actual plant conditions that were not pre-planned during testing; a test signal generated during a calibration check.

D. actual plant conditions with no testing in progress; actual plant conditions that were not pre-planned during testing.

ANSWER: D KA # & KA VALUE: Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. 2.4.30 (SRO 4.1) (CFR 43.5) Tier 3

REFERENCE:

PPM 1.10.1 Pages 7 and 11; NUREG-1022 Rev.3 Pages 31 - 38 SOURCE: BANK - April 2011 NRC Exam LO: 6011 LOK / LOD: H/3 HANDOUT: NONE JUSTIFICATION: A (incorrect) An intentional manual initiation as part of a pre-planned sequence during testing is not reportable per 50.72.

B (incorrect) An intentional manual initiation as part of a pre-planned sequence during testing is not reportable per 50.72.

C (incorrect) A test signal generated during a calibration check is an invalid signal and is not reportable per 50.72.

D (correct) Actual plant conditions not pre-planned are reportable per 50.72.