ML13057A530

2013-02-FINAL Written Exam

ML13057A530
Person / Time
Site:	Columbia
Issue date:	02/08/2013
From:	Vincent Gaddy Operations Branch IV
To:	Energy Northwest
laura hurley
References
50-397/13-002
Download: ML13057A530 (170)

Text

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-01 If the plant is operating at 100% power when RRC-P-1A trips, what is the reason for reducing Jet Pump Loop B Flow to LE 55.9 Mlb/hr?

A. To reduce the risk of a core instability event.

B. To prevent cavitation in the jet pump nozzles.

C. To prevent excessive crack propagation in the jet pump risers.

D. To prevent reverse flow through Reactor Recirculation Loop A.

ANSWER: C KA # & KA VALUE: 295001 AK3.05 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Reduced loop operating requirements. (3.2/3.6)

REFERENCE:

ABN-RRC-LOSS Rev.010 pg.5-6; AR 240219240219SOURCE: New LO: LO-5023 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The stem states that the plant is operating on the 100% power when the trip occurs. At 100% power, a single RRC pump trip will not result in entry into the AIA.

B (incorrect): Loop flow is reduced LT 41725 gpm to prevent jet pump nozzle cavitation.

C (correct): Per ABN-RRC-LOSS, if RRC-P-1A trips, RRC Loop B Jet Pump Flow must be reduced to LE 55.9 Mlbm/hr due to a crack in the Jet Pump-17/18 riser.

D (incorrect): This is the reason for closing RRC-V-67A following the trip.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-02 Columbia is operating at 100% power when an electrical fault causes SM-7 to lockout.

How will this partial loss of A.C. power affect D.C. loads?

The load on A. 125 VDC bus S1-1 will increase because E-IN-3A/B will automatically transfer to the DC source.

B. 125 VDC bus S1-2 will increase because E-IN-2A/B will automatically transfer to the DC source.

C. 125 VDC bus S1-7 will increase because E-IN-5 will automatically transfer to the DC source.

D. 250 VDC bus S2-1 will increase because E-IN-1 will automatically transfer to the DC source.

ANSWER: D KA # & KA VALUE: 295003 AK2.06 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following: D.C. electrical loads. (3.4/3.5)

REFERENCE:

SD000182 (AC); SD000188 (DC); SD000194 (UPS)

SOURCE: New LO: LO-5896 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The loss of SM-7 would cause a loss of MC-7A. This is the A.C. alternate power supply to E-IN-3A/B. S1-1 normally supplies E-IN-3A/B, so no transfer will occur on the loss of A.C. power.

B (incorrect): MC-8A in the A.C. alternate power supply to E-IN-2A/B, and would not be disrupted by the loss of SM-7.

C (incorrect): E-IN-5 is normally powered from A.C. source MC-8A, and will

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 automatically transfer to S1-7 on a loss of A.C., but the loss of SM-7 does not disrupt power to MC-8A.

D (correct): E-IN-1 is normally powered from A.C. source MC-7A. The loss of SM-7 causes a loss of MC-7A. E-IN-1, and all loads supplied by E-IN-1, will automatically transfer to S2-1. This results in an increase in load on S2-1.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-03 With the plant operating at power, a complete loss of 125 VDC Bus S1-7 has occurred.

As a result, ________ circuit breakers can NOT be operated from the Main Control Room.

A. SM-1 B. SM-4 C. SM-7 D. SM-8 ANSWER: A KA # & KA VALUE: 295004 AA1.03 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: A.C. electrical distribution. (3.4/3.6)

REFERENCE:

ABN-ELEC-125VDC Rev.009; SD000188 (DC) Rev.009 pg.31-34 SOURCE: New LO: LO-5262 RATING: L2 ATTACHMENT: None JUSTIFICATION: A (correct): S1-7 provides control power for SM-1 circuit breakers. A loss of control power prevents the breakers from being operated from the Main Control Room.

B (incorrect): SM-4 control power is provided by S1-HPCS.

C (incorrect): SM-7 control power is provided by S1-1.

D (incorrect): SM-8 control power is provided by S1-2.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-04 With the plant operating at 90% power in Final Feedwater Temperature Reduction, the Main Turbine trips and causes a reactor scram. Plant conditions following the trip are:

• APRMs are downscale

• RPV water level is -20 and up slow

• Both RRC pumps are operating at 15 Hz

• Drywell pressure is 0.6 psig and steady Which core thermal limit can be challenged by this transient?

A. APLHGR B. LHGR C. MAPRAT D. MCPR ANSWER: D KA # & KA VALUE: 295005 AK1.02 Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP: Core thermal limit considerations. (3.2/3.6)

REFERENCE:

SD000129 (MT) Rev.011 pg.39, Tech Spec Bases 3.3.1.1 SOURCE: New LO: LO-11647 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D. RPV water level and drywell pressure provided in the stem do not indicate a LOCA.

B (incorrect): See D.

C (incorrect): See D.

D (correct): The most vulnerable time to exceed the MCPR limit is on a

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 Main Turbine trip at the End of Cycle when all control rods are withdrawn. In order to mitigate the MCPR vulnerability, the RRC pumps are tripped. The stem states Final Feedwater Temperature reduction is in progress, which occurs at the End of Cycle. The RRC pumps are operating at 15 Hz indicating a failure of EOC-RPT. This combination of conditions results in a challenge to the MCPR thermal limit.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-05 The plant was operating at 100% power with a 1/2 scram on RPS System B due to scheduled surveillance testing, when the RPS System A motor generator output breaker tripped.

Five (5) seconds after the full scram, what was controlling reactor pressure?

A. Main Turbine Bypass Valves B. Main Turbine Governor Valves C. Main Steam Relief Valves in Relief Mode D. Main Steam Relief Valves in Safety Mode ANSWER: B KA # & KA VALUE: 295006 AA1.03 Ability to operate and/or monitor the following as they apply to SCRAM: Reactor/turbine pressure regulating system. (3.7/3.7)

REFERENCE:

LO000146 (DEH); LO000129 (MT)

SOURCE: Bank LO: LO-11650 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The conditions given in the stem do not indicate a Main Turbine trip has occurred.

B (correct): The scram described in the stem does not result in a direct Main Turbine trip. DEH will remain in Turbine Follow Reactor Mode, and the Governor Valves will ramp closed to control pressure.

C (incorrect): The loss of RPS-A results in a 1/2 MSIV isolation, but RPS-B remains powered even though there is a 1/2 scram on that system. As a result, the MSIVs remain open and the SRVs are not used for pressure control.

D (incorrect): See C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-06 Columbia was operating at 100% power when Main Control evacuation was required due to a fire.

All of the immediate actions of ABN-CR-EVAC were completed. No subsequent actions have been performed.

How will the Graphics Display System (GDS) display valve status for these conditions?

A. Groups 1-7 and Group 8 (RCIC) will have green backgrounds.

B. Groups 1-7 will have green backgrounds, and Group 8 (RCIC) will have a red and green background.

C. Groups 2-7 will have green backgrounds, and Groups 1 and 8 (RCIC) will have red and green backgrounds.

D. Groups 1, 4, and 8 (RCIC) will have red and green backgrounds, Groups 2 and 7 will have red backgrounds, and Groups 3 and 6 will have green backgrounds.

ANSWER: B KA # & KA VALUE: 295016 2.1.19 Control Room Abandonment. Ability to use plant computers to evaluate system or component status. (3.9/3.8)

REFERENCE:

ABN-CR-EVAC Rev.025; SD000173 (NS4) Rev.014 pg.11, 15 and Figure 21 SOURCE: New LO: LO-11935, LO-5601 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B.

B (correct): The immediate actions of ABN-CR-EVAC require the operator to Arm and Depress MSIV Logic A, B, C, and D pushbuttons. This will cause all valves in Groups 1-7 to close. GDS will reflect all valves closed by displaying a green background for those groups. Group 8 (RCIC) is not affected by the Isolation Logic pushbuttons. RCIC-V-8 and RCIC-V-63 will remain open (red background) and RCIC-V-76 will be closed (green background).

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): See B. Group 1 contains the MSIVs and equalizing valves.

Group 1 is normally red and green, and would remain in this configuration if only the B and D isolation pushbuttons were armed and depressed.

D (incorrect): See B. The status listed in this distractor is the status following insertion of a manual scram (required by immediate actions), but before the Isolation Logic pushbuttons are armed and depressed.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-07 While operating at power, average drywell temperature was observed to slowly rise. During investigation, RCC-V-72B (RCC Inlet to CRA-CC-2B) was found closed.

What caused this partial loss of RCC?

A. CRA-FN-2B1 and CRA-FN-2B2 (Primary Containment Cooling Fans) are both deenergized.

B. An electrical fault in Relay Cabinet 2 (RC-2) caused a spurious F signal for NSSSS Group 4.

C. The output of the controller for CRA-TCV-72B (Outlet TCV for CRA-CC-2B) failed low.

D. A piping break downstream of RCC-V-72B caused an automatic high flow isolation.

ANSWER: A KA # & KA VALUE: 295018 AA2.03 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Cause for partial or complete loss. (3.2/3.5)

REFERENCE:

SD000196 (RCC) Rev.011 pg.7 and 8; EWD-11E-026 SOURCE: Bank LO: LO-5710 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (correct): With the M coils for CRA-FN-2B1 and 2B2 deenergized, RCC-V-72B receives an auto close signal.

B (incorrect): NSSSS Group 4 contains RCC containment isolation valves and the RCC Pumps, but not RCC-V-72B.

C (incorrect): CRA-TCV-72B does not provide an isolation signal to RCC-V-72B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (incorrect): RCC-V-72B does not automatically isolate on high flow.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-08 A plant startup was in progress with reactor power at 65%, when a complete loss of Control and Service Air (CAS) occurred. No operator actions were taken in response to lowering CAS pressure before the system depressurized.

In this condition, Condensate and Feedwater can NOT be aligned for injection to the RPV through RFW-FCV-10A/B (Startup Flow Control Valves), because A. COND-FCV-15A/B/C (Booster Pumps Min Flow Valves) were closed and failed open. Condensate and Feedwater are not available for level control.

B. RFW-FCV-2A/B (RFW Pumps Min Flow Valves) were closed and failed open.

Condensate is not available for level control.

C. RFW-FCV-10A/B were open and have failed closed. RFW-V-109 (Htrs 6A/B Bypass Valve) can be used for injection.

D. RFW-FCV-10A/B were closed and failed as-is. RFW-V-109 (Htrs 6A/B Bypass Valve) can be aligned for injection.

ANSWER: D KA # & KA VALUE: 295019 AK2.03 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Reactor feedwater. (3.2/3.3)

REFERENCE:

ABN-CAS Rev.007; SOP-RFW-FCV-QC Rev.010; SD000151 (RFW)

Rev.012 pg.36 SOURCE: New LO: LO-11576 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): COND-FCV-15A/B/C were closed and did fail open. While this redirects partial flow, it would not normally prevent injection through RFW-FCV-10A/B. Condensate and Feedwater are available through RFW-V-109.

B (incorrect): RFW-FCV-2A/B will fail open on a loss of air, but Condensate and Feedwater are available through RFW-V-109.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): See D.

D (correct): The stem states that reactor power is 65% during the startup.

Level control has already been transferred from RFW-FCV-10A/B to the RFTs. As a result, RFW-FCV-10A/B were already closed by procedure when the loss of air occurred. These valves fail as-is, and therefore injection to the RPV can NOT occur through them. SOP-RFW-FCV-QC allows RFW-V-109 (motor operated valve) to be used as an alternate path.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-09 With a refueling outage in progress, a loss of Shutdown Cooling occurred. An Operator has been dispatched to deenergize RHR-V-9 by opening a disconnect per ABN-RHR-SDC-LOSS, Loss of Shutdown Cooling. When the Operator reports the disconnect has been opened to remove power, the following indications are observed:

• The green light for RHR-V-9 at H13-P-601 is on

• The red light for RHR-V-9 at H13-P-601 is off

• The RHR-V-9 POWER ENABLED annunciator on H13-P602 has cleared Which of the following disconnects did the Operator open, and where is the disconnect located?

A. RHR-DISC-V/9 (Remote Disconnect for RHR-V-9) in the RPS-B Room on the Radwaste Building 467.

B. RHR-DISC-V/9 (Remote Disconnect for RHR-V-9) in the Remote Shutdown Room on the Radwaste Building 467.

C. RHR-42-8BA2A (RHR-V-9 Motor Starter Disconnect) at MC-8BA on the Reactor Building 548.

D. RHR-42-8BA2A (RHR-V-9 Motor Starter Disconnect) at MC-8BA on the Reactor Building 471.

ANSWER: A KA # & KA VALUE: 295021 2.1.30 Loss of Shutdown Cooling. Ability to locate and operate components, including local controls. (4.4/4.0)

REFERENCE:

ABN-RHR-SDC-LOSS; SOP-RHR-SDC; EWD-9E-011; SD000198 (RHR)

Rev.013 pg.9, 10, and 14 SOURCE: New LO: LO-11801 RATING: H4 ATTACHMENT: None JUSTIFICATION: A (correct): Opening RHR-DISC-V/9 will remove power to the motor operator for RHR-V-9, but not power for remote position indication. The stem states that position indication remains after the disconnect is open. Opening

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 this disconnect will also cause the RHR-V-9 POWER ENABLED alarm to clear. This disconnect is located on the north wall of the RPS-B Room on the Radwaste Building 467.

B (incorrect): See A. RHR-V-9 can be controlled from the Remote Shutdown Room, but the disconnect is not located there.

C (incorrect): Opening RHR-42-8BA2A will remove power to the motor operator for RHR-V-9, and is the disconnect the Operator should have opened by procedure. Opening this disconnect will also remove power for remote position indication, but will not affect the RHR-V-9 POWER ENABLED annunciator.

D (correct): See C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-10 According to ABN-FUEL-HAND, what is the basis for evacuating the Refuel Floor during a refueling accident?

A. To prevent an uptake of Co-60 by personnel.

B. To prevent an uptake of I-127 by personnel.

C. To minimize the spread of contamination.

D. To minimize the dose received by personnel.

ANSWER: D KA # & KA VALUE: 295023 AK3.01 Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS: Refueling floor evacuation.

(3.6/4.3)

REFERENCE:

ABN-FUEL-HAND; FSAR Table 11.1 SOURCE: New LO: LO-6897 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): Co-60 will not be released or become airborne during a refueling accident.

B (incorrect): While an uptake of radioactive gases is possible during a refueling accident, and is listed as a concern in ABN-FUEL-HAND bases, I-127 is not the isotope of concern.

C (incorrect): Personnel contamination is possible during a refueling accident, but evacuating personnel could lead to the spread of contamination, not minimize the spread.

D (correct): According to ABN-FUEL-HAND, the refuel floor is evacuated as soon as possible to minimize the dose to personnel on the floor.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-11 The plant was operating at 70% power when a steam leak occurred inside containment. The CRS has entered the EOPs due to high drywell pressure, and all immediate actions have been performed.

Which of the following is correct for these conditions?

A. HPCS-P-1 started, but has not injected. E-MC-7C and E-MC-7E have been load shed.

B. HPCS-P-1 has started and injected. All low pressure ECCS pumps are running on minimum flow.

C. Both RRC pumps have tripped. Reactor Building HVAC has isolated and the previously running fans have tripped.

D. Both RRC pumps are operating at 15 Hz. The previously running TSW pump has tripped.

ANSWER: B KA # & KA VALUE: 295024 2.4.2 High Drywell Pressure. Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

(4.5/4.6)

REFERENCE:

ABN-FAZ-QC; SD000174 (HPCS) Rev.012 pg.13 SOURCE: New LO: LO-5422, LO-5425, LO-8017 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): HPCS-P-1 will start and automatically inject at the high drywell pressure EOP entry (1.68 psig). E-MC-7C and E-MC-7E automatically load shed at 1.68 psig.

B (correct): HPCS-P-1 will start and automatically inject at the high drywell pressure EOP entry (1.68 psig). All low pressure ECCS pumps automatically start at 1.68 psig. The stem states the plant was at 70% power and the drywell pressure increase was due to a steam leak. Based on these

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 indications, reactor pressure is greater than the shutoff head of the low pressure ECCS pumps. As a result, the pumps are running on minimum flow.

C (incorrect): The RRC pumps have lost all cooling water flow, but do not automatically trip. The CRS will direct the RRC pumps to be tripped per the EOPs (not an immediate action). RB HVAC has automatically isolated and the previously running fans have tripped.

D (incorrect): A reactor scram occurred at 1.68 psig, which caused (or will cause) RPV water level to lower to approximately -20. The RRC pumps will automatically runback to 15 Hz at +13. The TSW pumps are powered from SM-75/85, which load shed at 1.68 psig in the drywell coincident with a LOOP. Additionally, if both TSW pumps were running, TSW-P-1B would trip at 1.68 psig. The normal system lineup is to have one TSW pump running.

The previously running TSW pump will still be running.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-12 With Columbia operating in MODE 1, a reactor scram occurs and all control rods fully insert. After conditions stabilize, CRO1 places the SDV High Level Bypass Switch to the BYPASS position and resets the scram by depressing both scram reset pushbuttons on H13-P603. Three (3) minutes later, the following indications exist:

• All white RPS Logic lights on H13-601 are on

• All amber Backup scram lights on H13-P601 are off

• All blue scram lights on the full core display are on

• All scram discharge volume vent and drain valves are closed Based on the information given, what caused the reactor scram?

A. A drywell pressure rise to 2 psig B. A reactor power rise to 110%

C. A reactor water level drop to -35 D. A reactor pressure rise to 1130 psig ANSWER: D KA # & KA VALUE: 295025 EK2.04 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: ARI/RPT/ATWS. (3.9/4.1)

REFERENCE:

SD000161 (RPS) Rev.015 pg.5, 12, and 13; SD000142 Rev.014 pg.21 and 22 SOURCE: Bank LO: LO-5188, LO-5189, LO-7676 RATING: H4 ATTACHMENT: None JUSTIFICATION: A (incorrect): This condition will cause a scram (1.68 psig), but when the scram is reset, the scram valves will close (all blue lights off).

B (incorrect): This condition may result in a flow-biased scram, but when the scram is reset, the scram valves will close (all blue lights off).

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): This condition will cause a scram (+13), but when the scram is reset, the scram valves will close (all blue lights off).

D (correct): At 1060 psig reactor pressure, the reactor will scram. At 1120 psig reactor pressure, the ATWS/ARI valves open to vent the scram air header. When the scram is reset, the ATWS/ARI valves remain open and the scram air header remains depressurized. As a result, the scram valves remain open (all blue lights on) and the SDV vent and drain valves remain closed.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-13 The reactor was operating at 100% power when a transient occurred that caused wetwell temperature to rapidly rise. An emergency depressurization was completed prior to HCTL being exceeded.

The basis for this emergency depressurization is to A. prevent failure of the Safety Relief Valve tail pipes.

B. prevent failure of primary containment or safe shutdown equipment.

C. ensure the pressure suppression function is not bypassed due to drywell floor failure.

D. ensure adequate NPSH is maintained for pumps required to operate after the RPV is depressurized.

ANSWER: B KA # & KA VALUE: 295026 EK3.01

REFERENCE:

PPM 5.0.10 Rev.017 pg.230 SOURCE: Modified Bank (LO00261)

LO: LO-8303 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): SRV tail pipe stresses are a concern at high wetwell levels, and are addressed by SRVTPLL.

B (correct): Per PPM 5.0.10, Depressurizing the RPV when Wetwell temperature and RPV pressure cannot be maintaine below the HCTL precludes failure of the containment or equipment necessary for the safe shutdown of the plant.

C (incorrect): Drywell floor failure due to excessive upward force could occur due to high wetwell level and drywell sprays. This is addressed in the wetwell level leg, and in ot the basis for this ED.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (incorrect): NPSH is a concern when wetwell level is low, and is addressed by vortex limits. NPSH is not the basis for ED.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-14 Given the following:

• Drywell temperature is rising

• RPV water level is below TAF and stable If actual level remains constant, Wide Range RPV level indication may begin to rise because the density of the A. variable leg rises, which results in a lower sensed dP.

B. variable leg lowers, which results in a higher sensed dP.

C. reference leg rises, which results in a higher sensed dP.

D. reference leg lowers, which results in a lower sensed dP.

ANSWER: D KA # & KA VALUE: 295028 EK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Reactor water level measurement. (3.5/3.7)

REFERENCE:

PPM 5.0.10 Rev.017 pg.58; SD000126 (NBI)

SOURCE: New LO: LO-11774 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): The density of any fluid remaining in the variable leg will lower as drywell temperature rises.

B (incorrect): The density of any fluid remaining in the variable leg will lower as drywell temperature rises, and this would result in a higher sensed d/p. A higher d/p will cause indication to lower, not rise as stated in the stem.

C (incorrect): The density of the fluid in the reference leg will lower as drywell temperature rises.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (correct): As drywell temperature rises, the density of the head of water in the reference leg lowers. This results in a lowering sensed dP, which corresponds to a rising indicated RPV water level even though actual level is below the variable leg tap.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-15 Given the following:

• Wetwell level is 22 feet

• Wetwell temperature is 200°F Of the following, which is the lowest reactor pressure that results in exceeding HCTL?

A. 500 psig B. 600 psig C. 800 psig D. 900 psig ANSWER: B KA # & KA VALUE: 295030 EA2.03 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Reactor pressure.

(3.7/3.9)

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013

REFERENCE:

PPM 5.0.10 Rev.017 pg.69 SOURCE: New LO: LO-8302 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B. If the candidate assumes the unsafe area is below and to the left of the line, 500 psig would be selected.

B (correct): The stem states wetwell is 22 feet. As this value is between the 19.2 and 23 lines, the conservative line (19.2) should be used. At 200°F in the wetwell, the HCTL is approximately 560 psig. Of the values listed, 600 psig is the lowest pressure listed that exceeds 560 psig.

C (incorrect): See B. If the candidate attempts to interpolate the line for 22, 800 psig would be selected.

D (incorrect): See B. If the candidate uses the 23 line, 900 psig would be selected.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-16 With the plant operating at 90% power, a rupture in the Condensate system caused a loss of Feedwater. RPV water level lowered to -58 before being automatically restored.

Two (2) minutes after RPV water level reached -58, what is the status of the Emergency Diesel Generators?

A. DG-1 and DG-2 are running. DG-3 is in standby.

B. DG-3 is running. DG-1 and DG-2 are in standby.

C. DG-1, DG-2, and DG-3 are running.

D. DG-1, DG-2, and DG-3 are in standby.

ANSWER: B KA # & KA VALUE: 295031 EK2.14 Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following:

REFERENCE:

ABN-LEVEL Rev.006 pg.3 and 4; CGS Simulator SOURCE: New LO: LO-10291 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B.

B (correct): At -50, DG-3 receives an automatic initiation signal, starts, and runs unloaded. Containment cooling via RCC isolates and the RCC pumps trip at -50, but drywell pressure will not rise to 1.68 psig (auto initiation for DG-1 and DG-2) in the 2 minute boundary established by the question. RPV level is stated to have reached a low value of -58. DG-1 and DG-2 start at -

129. As a result, DG-1 and DG-2 remain in standby.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-17 An emergency depressurization has been performed per PPM 5.1.2, RPV Control - ATWS. The following conditions exist:

• RPV water level is offscale low

• RPV pressure is 180 psig and down fast

• Seven (7) SRVs are open What method is used to restore RPV water level for these conditions, and what is the reason for employing this method?

A. Commence injection into the RPV, but slowly increase injection rate to prevent large power excursions and the resultant damage to the cladding.

B. Commence injection into the RPV, but slowly increase injection rate to prevent exceeding Lowered Level and the subsequent addition of energy to the wetwell.

C. Commence injection and rapidly raise injection rate into the RPV to establish water level above -183 and restore adequate core cooling.

D. Commence injection and rapidly raise injection rate into the RPV to establish water level above -161 and restore adequate core cooling.

ANSWER: A KA # & KA VALUE: 295037 EK1.02 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor water level effects on reactor power. (4.1/4.3)

REFERENCE:

PPM 5.0.10 Rev.017 pg 147-151; PPM 5.1.2 SOURCE: New LO: LO-8499 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (correct): In an ATWS core, rapid injection of cold, unborated water could cause large power excursions and core damage. Injection into the RPV must be increased slowly to preclude the possibility of large power excursions.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 B (incorrect): See A. Although RPV level should be kept below the established Lowered Level, this is not the reason for slowly raising injection rate.

C (incorrect): See A. When RPV pressure drops below MSCP, adequate core cooling no longer exists. Injection must be re-established to maintain adequate core cooling and ultimately submerge the core. The need to restore adequate core cooling does not, however, override the need to ensure injection does not result in large power excursions and core damage.

-183 is the MSCRWL.

D (incorrect): See A and C. -161 is TAF.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-18 An offsite release is in progress when the following alarms are received:

• REMOTE INTAKE DIV 1 RAD HI-HI

• REMOTE INTAKE DIV 2 RAD HI-HI Considering only these alarms, how will the Control Room HVAC system respond?

A. WOA-V-51C and WOA-V-52C (Normal Outside Air Intake Isolation Valves) close WMA-AD-51A1 and WMA-AD-51B1 (Outside Air Supply Dampers) close WMA-FN-54A and WMA-FN-54B (Emergency Fltr Supply Fans) start WEA-FN-51 (Kitchen Exhaust Fan) stops B. WOA-V-51A and WOA-V-51B (Remote Air Intakes) open WOA-V-51C and WOA-V-52C (Normal Outside Air Intake Isolation Valves) close WMA-FN-54A and WMA-FN-54B (Emergency Fltr Supply Fans) start C. WMA-AD-51A1 and WMA-AD-51B1 (Outside Air Supply Dampers) close WEA-FN-51 (Kitchen Exhaust Fan) stops WEA-AD-51 (Outlet Damper) closes D. WMA-AD-51A1 and WMA-AD-51B1 (Outside Air Supply Dampers) close WMA-FN-54A and WMA-FN-54B (Emergency Fltr Supply Fans) start WEA-FN-51 (Kitchen Exhaust Fan) stops WEA-AD-51 (Outlet Damper) closes ANSWER: C KA # & KA VALUE: 295038 EA1.07 Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: Control room ventilation. (3.6/3.8)

REFERENCE:

PPM 4.826.P1 2-3; PPM 4.826.P2 2-2; SD000201 (CRHVAC); ABN-RAD-CR SOURCE: New LO: LO-7648 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. This configuration describes pressurization mode of

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 operation, and is the response of the Control Room ventilation system to an FAZ signal.

B (incorrect): See C. WOA-V-51A or WOA-V-51B will be manually closed as directed by ABN-RAD-CR.

C (correct): If an offsite release results in Remote Air Intake Rad HI-HI in both divisions, the Control Room ventilation systems automatically aligns for recirculation mode of operation, as described in the answer.

D (incorrect): See C. WMA-FN-54A/B automatically start on an FAZ signal, not Rad HI-HI.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-19 A fire in the Main Control Room resulted in an evacuation per ABN-CR-EVAC.

Which of the following tasks must a Reactor Operator perform as part of the critical path per ABN-CR-EVAC, and what effect will this have?

A. Place E-RMS-7/71/CT and E-RMS-7/73/CT (CT Shorting Switches) in EMERG.

This removes the control room ammeters for SL-71 and SL-73 from the breakers current protection circuits.

B. Place E-RMS-8/81/CT and E-RMS-8/83/CT (CT Shorting Switches) in EMERG.

This removes the control room ammeters for SL-81 and SL-83 from the breakers current protection circuits.

C. Place E-RMS-7/71/CT and E-RMS-7/73/CT (CT Shorting Switches) in EMERG.

This shorts around the overcurrent relays for SL-71 and SL-73 to prevent an overcurrent trip.

D. Place E-RMS-8/81/CT and E-RMS-8/83/CT (CT Shorting Switches) in EMERG.

This shorts around the overcurrent relays for SL-81 and SL-83 to prevent an overcurrent trip.

ANSWER: B KA # & KA VALUE: 600000 2.4.34 Plant Fire On Site. Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

REFERENCE:

ABN-CR-EVAC Rev.025 pg.14 and 24 SOURCE: New LO: LO-7737, RO-1057 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B. SM-8, not SM-7 is critical path.

B (correct): Because the MCR was evacuated due to fire, the critical path is to start DG-2 and isolate SM-8. A Reactor Operator (CRO2) will be directed to perform Attachment 7.3 of ABN-CR-EVAC. One task performed per this

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 attachment is to place the CT shorting switches for SL-81 and SL-83 ammeters to EMERG to remove them from the protective circuits and prevent a spurious actuation.

C (incorrect): See B. SM-8, not SM-7 is critical path.

D (incorrect): See B. The ammeter, not the overcurrent relay, is shorted out of the circuit. A spurious actuation is prevented, but actuation due to an actual overcurrent condition is maintained.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-20 Refer to the capability curve on the next page to answer this question.

Main Generator conditions are as follows:

• 1200 MWe output

• 20 MVARs IN

• 75 psig hydrogen pressure

• Generator Voltage Regulator ON

• Power System Stabilizer ON A grid disturbance results in an OSCILLOGRAPH STARTED alarm and Main Generator MVARs indicating 340 MVARs IN. All other parameters remain the same.

Main Generator MVARs A. are outside of the capability curve. Place E-RMS-90V (Main Generator Exciter Voltage Adjuster) in LOWER to return MVARs to within the capability curve.

B. are outside of the capability curve. Place E-RMS-90V (Main Generator Exciter Voltage Adjuster) in RAISE to return MVARs to within the capability curve.

C. remain within the capability curve. Enter ABN-ELEC-GRID, Degraded Off Site Power Grid, and ABN-GENERATOR, Main Generator Trouble.

D. remain within the capability curve. Dispatch OPS2 to investigate the OSCILLOGRAPH STARTED alarm, and initiate a Condition Report.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 ANSWER: B KA # & KA VALUE: 700000 AA2.04 Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

VARs outside capability curve. (3.6/3.6)

REFERENCE:

SOP-MT-START Rev.022 pg.61 and Attachment 6.5 SOURCE: New LO: LO-5521, LO-7647 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B. If the candidate uses the overexcited portion of the curve for MVARs IN, they will incorrectly place the Voltage Adjuster to LOWER.

B (correct): At 1200 MWe and 75 psig hydrogen pressure, 340 MVARs IN is below the capability curve. MVARs must be reduced (from 340 to approximately 280) to return to within the capability curve. To accomplish this excitation must be increased by taking the Voltage Adjuster to RAISE.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-21 A plant shutdown was in progress per PPM 3.2.1, Normal Plant Shutdown. A plant announcement for tripping the Main Turbine had just been made, when Main Condenser backpressure began to rapidly degrade resulting in an automatic Main Turbine trip. No operator actions have been taken, and current plant conditions are as follows:

• All control rods are fully inserted

• One SRV is open for pressure control Main Condenser backpressure was ____________ when the reactor scram occurred.

A. 6.5 Hg abs B. 8.5 Hg abs C. 20.6 Hg abs D. 21.6 Hg abs ANSWER: D KA # & KA VALUE: 295002 AA2.01 Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM: Condenser vacuum/absolute pressure. (2.9/3.1)

REFERENCE:

ABN-BACKPRESSURE; PPM 3.2.1; SOP-MT-SHUTDOWN SOURCE: New LO: LO-6788 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D. If the candidate assumes the reactor scram occurred as a result of the Main Turbine trip and does not identify the MSIV closure, this value represents the Main Turbine trip setpoint for load GT 560 MWe, but LT 835 MWe.

B (incorrect): See A and D. This value exceeds the trip setpoint for Turbine loads GE 835 MWe.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): See D. This value is greater than the Main Turbine trip setpoint, but below the MSIV closure setpoint.

D (correct): The stem states that the announcement for a Main Turbine trip has been made. PPM 3.2.1 directs a power reduction to approximately 15%

prior to tripping the Main Turbine. This means the Main Turbine trip did not cause the automatic reactor scram. At 21.6 Hga, automatic MSIV closure occurs, which caused the scram.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-22 The reason RCIC-V-45 (Steam to RCIC Turbine) automatically closes on high reactor water level is to ___________.

A. prevent vessel overfill and overflow into the Main Steam Lines B. prevent RCIC Turbine blade damage due to water introduction C. limit RPV pressure reduction and prevent exceeding RPV cooldown limits D. limit steam dome pressure reduction and prevent an automatic trip of the RRC pumps ANSWER: A KA # & KA VALUE: 295008 AK3.08 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR WATER LEVEL: RCIC steam supply valve closure. (3.4/3.5)

REFERENCE:

Tech Spec Bases 3.3.5.2-3; SD000180 (RCIC) Rev.014 pg.18 SOURCE: New LO: LO-11675 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (correct): Per the references, RCIC-V-45 automatic closure is prevent vessel overfill and overflow into the Main Steam Lines.

B (incorrect): See A. This is the TS reason for Main Turbine trip and Throttle Valve closure.

C (incorrect): See A. Closure of RCIC-V-45 will stop RCIC injection and reduce cooldown, but this is not the reason for automatic closure due to high RPV water level.

D (incorrect): See A. RCIC injection reduces steam dome pressure and temperature. The RRC pumps have a delta-T cavitation trip at LT 10.7°F for 10 mins. This is not the reason for automatic closure due to high RPV water level.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-23 With the plant operating in MODE 1, which of the following parameters would require entering a Technical Specification LCO?

A. Suppression pool level is 31 feet.

B. Drywell temperature is 140 degrees.

C. Suppression pool temperature is 85 degrees.

D. Primary containment oxygen concentration is 2.5 volume percent.

ANSWER: B KA # & KA VALUE: 295012 2.2.42 High Drywell Temperature. Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

REFERENCE:

TS 3.6.1.4; 3.6.2.1; 3.6.2.2; 3.6.3.3 SOURCE: New LO: LO-8311 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): LCO 3.6.2.2 requires GE 30 9.75 and LE 31 1.75.

B (correct): LCO 3.6.1.4 requires LE 135 degrees.

C (incorrect): LCO 3.6.2.1 requires LE 90 degrees when above 1% RTP and no testing that adds heat to the Suppression Pool in progress.

D (incorrect): LCO 3.6.3.3 requires LT 3.5 volume percent.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-24 The plant is operating at 100% power, when the DRYWELL/SUPP POOL TEMP HIGH alarm is received. No other alarms are received. CMS-TR-5 (Wetwell/Drywell Temp) indicates as follows:

• Pt 220 (AVG SUPP POOL TEMP UPPER LEVEL): 81.5°F

• Pt A02 (AVG SUPP POOL TEMP LOWER LEVEL): 74.3°F Further investigation reveals that Pt 220 has been gradually increasing.

These indications are the result of A. a leaking SRV.

B. Suppression Pool thermal stratification.

C. upper Suppression Pool thermocouples becoming uncovered.

D. a short circuited thermocouple in the upper thermocouple averaging circuit.

ANSWER: B KA # & KA VALUE: 295013 AA2.02 Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Localized heating/stratification. (3.2/3.5)

REFERENCE:

PPM 4.601.A11 1-3 Rev.022 pg.7; CMS-TR-5 IMDS; EWD-25I-036 SOURCE: New LO: None RATING: H4 ATTACHMENT: None JUSTIFICATION: A (incorrect): A leaking SRV would lead to elevated wetwell temperatures, but would not cause the difference between upper and lower temperatures.

B (correct): The difference in temperature between the upper and lower average temperatures combined with the gradual rise in temperature is indicative of thermal stratification.

C (incorrect): The stem states no other alarms were received. In order for

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 the thermocouples to become uncovered, SP level must lower to approximately -9, which would have caused a Suppression Pool low level alarm.

D (incorrect): A shorted thermocouple will result in lower temperature indication. If a thermocouple in the upper group had failed this way, average temperature would have lowered, not increased.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-25 Events have occurred that have resulted in Columbia being in an ATWS. RPV water level is -100 and is being maintained with RFW. Both Standby Liquid Control Pumps are running, and the Cold Shutdown Boron Weight has been injected.

The reactor is shutdown A. at the current RPV water level, but may return to power (clear the APRM downscales) if level is raised above -65.

B. but cooldown can NOT commence until it has been determined that the existing control rod pattern can assure the reactor is shutdown.

C. but may return to power (clear the APRM downscales) if an RPV cooldown is commenced with an 80°F/HR cooldown rate.

D. and will remain shutdown if an RPV cooldown is commenced with an 80°F/HR cooldown rate.

ANSWER: D KA # & KA VALUE: 295015 AK1.02 Knowledge of the operational implications of the following concepts as they apply to INCOMPLETE SCRAM: Cooldown effects on reactor power. (3.9/4.1)

REFERENCE:

PPM 5.0.10 SOURCE: Bank (slightly modified LO01814)

LO: LO-8180 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D. Peer PPM 5.1.2, RPV level is lowered to reduce core inlet subcooling and minimize power oscillations. Raising RPV level will result in lower RPV inlet temperature and positive reactivity, but CSBW assumes an RPV temperature of 68°F to maintain the reactor shutdown.

B (incorrect): See D. If the existing rod pattern alone can always assure the reactor is shutdown, boron injection is stopped and PPM 5.1.1 is entered, which will allow an RPV cooldown to commence. Injecting CSBW will also

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 allow cooldown to commence, which the stem states has occurred.

Cooldown may commence.

C (incorrect): See D. Without CSBW injected, power could rise due to the positive reactivity addition from cooldown.

D (correct): PPM 5.1.2 step P-7 allows cooldown to commence once the CBSW has been injected. PPM 5.0.10 defines the CSBW as the least weight of soluble boron which, if injected into the RPV and mixed uniformly, will maintain the reactor shutodown under all conditions. The CSBW is utilized to assure the reactor will remain shutdown irrespective of control rod position or RPV water temperature. PPM 5.1.2 allows a cooldown rate up 100°F with CSBW injected.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-26 The plant was operating at 100% power, when the following sequence of events occurred:

T=0 LEAK DET RWCU CH A DIFF FLOW HI and LEAK DET RWCU CH B DIFF FLOW HI alarms were received on H13-P601 A2 and A3 T = 7 secs LEAK DET RWCU ROOMS TEMP HIGH alarm was received on H13-P601 A12 T = 19 secs LEAK DET RWCU ROOMS TEMP HI-HI alarms were received on H13-P601 A2 and A3 If current time is T = 25 secs, which of the following is correct concerning RWCU-V-1 (Suction Inboard Isolation) and RWCU-V-4 (Suction Outboard Isolation)?

A. RWCU-V-1 and RWCU-V-4 are both closed.

B. RWCU-V-1 and RWCU-V-4 will both close at T = 45 seconds.

C. RWCU-V-4 is closed and RWCU-V-1 will close at T = 45 seconds.

D. RWCU-V-1 is closed and RWCU-V-4 will close at T = 45 seconds.

ANSWER: A KA # & KA VALUE: 295032 EA1.05 Ability to operate and/or monitor the following as they apply to HIGHSECONDARY CONTAINMENT AREA TEMPERATURE: Affected systems so as to isolate damaged portions. (3.7/3.9)

REFERENCE:

PPM 4.601.A12 5-3; PPM 4.601.A2 2-2 and 4-2; PPM 4.601.A3 2-5 and 3-4; SD000190 (RWCU)

SOURCE: New LO: LO-5035 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): The LEAK DET RWCU ROOMS TEMP HI-HI alarm indicates

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 an isolation setpoint has been reached. RWCU-V-1 and 4 are divisional.

Receipt of both alarms indicates both valves have isolated.

B (incorrect): LEAK DET RWCU CH A(B) DIFF FLOW HI are divisional alarms that cause RWCU-V-1 and 4 to isolate after a 45 sec time delay. This distractor would be correct if RWCU-V-1 and 4 had not already closed at T =

19 secs.

C (incorrect): See A. Although the TEMP HI-HI alarms are divisional the single HIGH temperature alarm is not (i.e. a high temp in either division will cause an alarm). If the candidate assumes only the Division 2 alarm was received, they may assume that only RWCU-V-4 is closed, and RWCU-V-1 will close after the 45 sec time delay for high differential flow.

D (incorrect): See A. Although the TEMP HI-HI alarms are divisional the single HIGH temperature alarm is not (i.e. a high temp in either division will cause an alarm). If the candidate assumes only the Division 1 alarm was received, they may assume that only RWCU-V-1 is closed, and RWCU-V-4 will close after the 45 sec time delay for high differential flow.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-27 Flooding in the _____________ will cause FDR-V-607 (RCIC Floor Drain to FDR Sump R-1 Inlet Valve) to automatically close.

A. RHR-B pump room B. LPCS pump room C. HPCS pump room D. RHR-A pump room ANSWER: D KA # & KA VALUE: 295036 Knowledge of the interrelations between SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL and the following:

Secondary containment equipment and floor drain system.

REFERENCE:

PPM 4.602.A13 2-1 SOURCE: Bank LO: LO-5332 RATING: L4 ATTACHMENT: None JUSTIFICATION: A (incorrect): RHR-B drains to Sump R2.

B (incorrect): LPCS drains to Sump R4.

C (incorrect): HPCS drains to Sump R3.

D (correct): The floor drains in RHR-A pump room drain to FDR Sump R-1.

A high sump level in R-1 causes FDR-V-607 to automatically close to prevent flooding in the RCIC pump room.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-28 Following a reactor scram on high drywell pressure, RHR-V-16B (Drywell Spray Outboard Isolation) will not open using the control switch if:

A. RHR-V-17B (Drywell Spray Inboard Isolation) is open AND RHR-V-42B (LPCI Injection Valve) is closed B. RHR-V-17B (Drywell Spray Inboard Isolation) is closed AND RHR-V-42B (LPCI Injection Valve) is closed C. RHR-V-17B (Drywell Spray Inboard Isolation) is open AND RHR-V-42B (LPCI Injection Valve) is open D. RHR-V-17B (Drywell Spray Inboard Isolation) is closed AND RHR-V-42B (LPCI Injection Valve) is open ANSWER: C KA # & KA VALUE: 203000 K4.15 Knowledge RHR/LPCI: INJECTION MODE design feature(s) and/or interlocks which provide for the following: Pump runout protection.

(2.5/2.5)

REFERENCE:

SD000198 (RHR) Rev.013 pg.15 and 40.

SOURCE: New LO: LO-11801i., LO-5781 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. RHR-V-16B can be opened when RHR-V-17B is open provided the following conditions also exist:

- RHR-V-42B is closed

- LPCI B initiation signal is present

- Div 2 high drywell pressure signal is present The stem states that an automatic reactor scram occurred due to high drywell pressure. The scram setpoint is the same as the LPCI initiation setpoint and high drywell pressure signal. The RHR-V-16B interlocks are satisfied.

B (incorrect): See A and C. Procedurally RHR-V-17B is opened first and

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 then RHR-V-16B, but there is no physical interlock.

C (correct): Per SD000198 (RHR), Pump operation at runout can occur when discharging into multiple flowpath. RHR-V-16B and RHR-V-17B are interlocked to prevent opening both valves and initiating drywell sprays if an injection flow path is already established (RHR-V-42B is open).

D (incorrect): See A and C. Procedurally RHR-V-17B is opened first and then RHR-V-16B, but there is no physical interlock.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-29 Following an emergency depressurization, RHR-P-2A is restoring RPV water level. Currently:

• RPV water level is -155 and up fast

• RHR-V-42A (LPCI Injection Valve) is open

• RHR-RMS-S105 (RHR-V-42A Override Switch) is in OVERRIDE CRO3 places the control switch for RHR-V-42A in CLOSE.

Given these conditions A. the amber MANUAL OVERRIDE light for RHR-V-42A was already illuminated.

RHR-V-42A remains open and cannot be closed until RHR-RMS-S105 is placed in NORMAL.

B. the amber MANUAL OVERRIDE light for RHR-V-42A was already illuminated.

RHR-V-42A remains open and cannot be closed until RPV water level is restored above -129.

C. RHR-V-42A moves in the closed direction. The amber MANUAL OVERRIDE light for RHR-V-42A illuminates when the control switch is placed in CLOSE, and remains lighted when the control switch is returned to AUTO.

D. RHR-V-42A moves in the closed direction. The amber MANUAL OVERRIDE light for RHR-V-42A illuminates when the control switch is placed in CLOSE, but extinguishes when the control switch is returned to AUTO.

ANSWER: C KA # & KA VALUE: 203000 A4.11 RHR/LPCI: Injection Mode. Ability to manually operate and/or monitor in the control room: Indicating lights and alarms. (3.7/3.5)

REFERENCE:

EWD-9E-043; EWD-9E-097; SD000198 (RHR) pg.17 and 30 SOURCE: New LO: LO-7729, LO-7730 RATING: H4 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. The candidate may incorrectly think placing RHR-

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 RMS-S105 in OVERRIDE causes the amber MANUAL OVERRIDE light to illuminate.

B (incorrect): See C and A.

C (correct): RHR-RMS-S105 in OVERRIDE prevents automatic operation of RHR-V-42A and disables the open and close seal-in contacts allowing the valve to be throttled. When the control switch is placed in CLOSE, the valve will move in the closed direction. The MANUAL OVERRIDE light illuminated with the control switch in AUTO indicates RHR-V-42A has been placed in CLOSE with an initiation signal and reactor pressure less than 470 psig. The stem states that RPV level is -155 (initiation signal is -129) and RHR-P-2A is restoring level, which indicates RPV pressure is LT approximately 220 psig.

D (incorrect): See C. The MANUAL OVERRIDE light illuminates when the control switch for RHR-V-42A is placed in CLOSE regardless of any other plant conditions. It will remain lighted under the conditions described in C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-30

_______ is powered from E-MC-S2/1A.

A. RHR-V-6A (RHR-P-2A SDC Suction)

B. RHR-V-8 (SDC Outboard Isolation)

C. RHR-V-9 (SDC Inboard Isolation)

D. RHR-V-53A (RHR-P-2A SDC Return)

ANSWER: B KA # & KA VALUE: 205000 K2.02 Shutdown Cooling System (RHR Shutdown Cooling Mode)

Knowledge of electrical power supplies to the following: Motor operated valves. (2.5/2.7)

REFERENCE:

SD000198 (RHR) pg.48 SOURCE: New LO: LO-11805 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): Powered from E-MC-7BA, and will close.

B (correct): E-MC-S2/1 is the power supply to RHR-V-8. The valve would normally close on an isolation signal, but will remain open without power.

C (incorrect): Powered from E-MC-8BA, and will close.

D (incorrect): Powered from E-MC-7BA, and will close.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-31 The Automatic Depressurization System logic associated with LPCS is satisfied by:

A. LPCS-P-1 discharge pressure being GE 145 psig ONLY.

B. The circuit breaker for LPCS-P-1 being closed ONLY.

C. The circuit breaker for LPCS-P-1 being closed OR LPCS-P-1 discharge pressure being GE 145 psig.

D. Both the circuit breaker for LPCS-P-1 being closed AND LPCS-P-1 discharge pressure being GE 145 psig.

ANSWER: A KA # & KA VALUE: 209001 K1.05 Knowledge of the physical connection and/or cause-effect relationship between LOW PRESSURE CORE SPRAY SYSTEM and the following: Automatic depressurization system. (3.7/3.7)

REFERENCE:

SD000000192 (LPCS) Rev.012 pg.16; SD000186 (ADS) Rev.011 pg.4, 6, and 9; M520 SOURCE: New LO: LO-5485d RATING: L3 ATTACHMENT: None JUSTIFICATION: A (correct): LPCS-PS-1 and LPCS-PS-9 satisfy the ADS logic when LPCS discharge pressure is GE 145 psig.

B (incorrect): See A. LPCS-P-1 breaker position is used in the logic for LPCS-FCV-11 (Min Flow), not ADS.

C (incorrect): See A and B.

D (incorrect): See A and B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-32 Columbia is operating at 100% power.

If E-MC-7B is lost A. LPCS-P-2 (Keep Fill Pump) will be deenergized. LPCS-P-1 manual injection availability should be maintained.

B. LPCS-P-2 (Keep Fill Pump) will be deenergized. LPCS-P-1 should be started and placed in Suppression Pool Mixing per SOP-LPCS-SP.

C. LPCS-P-1 will NOT automatically inject if an initiation signal is received. Per PPM 1.3.1, the control power fuses for LPCS-P-1 should be removed.

D. LPCS-P-1 will NOT automatically inject if an initiation signal is received. Per PPM 1.3.1, the circuit breaker for LPCS-P-1 should be racked out.

ANSWER: C KA # & KA VALUE: 209001 A2.03 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 the abnormal conditions or operations: A.C. failures. (3.4/3.6)

REFERENCE:

SD000192 (LPCS) Rev.012 pg.17; PPM 1.3.1 Rev.108 pg.68 SOURCE: New LO: LO-5486 RATING: H4 ATTACHMENT: None JUSTIFICATION: A (incorrect): LPCS-P-2 is powered from E-MC-7B, and will be deenergized.

The plant is not operating in the EOPs, so LPCS should not be maintained available. Action should be taken to prevent automatic start of the pump to prevent system damage.

B (incorrect): See A and C. Placing LPCS in Suppression Pool Mixing is an option on a failure of LPCS-P-2. LPCS-V-12 (SP Return) is powered from E-MC-7BA. As a result, this flow path is not available in this condition.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (correct): E-MC-7B supplies E-MC-7BA, which powers LPCS-V-5 (Injection Valve). LPCS-P-1 cannot automatically inject. E-MC-7BA also provides power to LPCS-FCV-11 (Min Flow). This means the pump will start on an initiation signal, but operate without min flow protection. PPM 1.3.1 states, The preferred method of deactivating an ECCS pump to prevent an unplanned or unwanted start is by pulling the control power fuses only.

D (incorrect): See C. PPM 1.3.1 states, The preferred method of deactivating an ECCS pump to prevent an unplanned or unwanted start is by pulling the control power fuses only. Because of the subsequent pump start requirement following a pump circuit breaker rackout, it is recommended that breaker rackouts be limited to those cases where pulling the control power fuses alone is not adequate.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-33 If HPCS-RMS-S26 (HPCS RPV High Water Level Override Keylock Switch) is placed in the "OVERRIDE" position, then:

1) HPCS-V-4 will NOT automatically close on high reactor water level.

2) HPCS-V-4 will automatically close on high reactor water level, but placing the control switch for HPCS-V-4 in the OPEN position will bypass the high reactor water level signal.

3) HPCS-V-4 can be throttled in either the OPEN or CLOSE direction.

4) HPCS-V-4 can NOT be throttled in either direction.

5) HPCS-V-4 will not automatically open on a high pressure core spray (HPCS) initiation signal.

6) HPCS-V-4 will automatically open on a HPCS initiation signal.

A. 1, 3, and 5 B. 1, 3, and 6 C. 1, 4, and 6 D. 2, 4, and 6 ANSWER: C KA # & KA VALUE: 209002 K4.07 Knowledge of HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) design feature(s) and/or interlocks which provide for the following:

Override of reactor water level interlock. (3.5/3.7)

REFERENCE:

SD000174 (HPCS) Rev.012 pg.9 SOURCE: Bank LO: LO-7662 RATING: L2 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. (5) HPCS-V-4 will open on an initiation signal.

B (incorrect): See C. (3) HPCS-V-4 cannot be throttled. This occurs when HPCS-RMS-S25 is placed in OVERRIDE.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (correct): HPCS-RMS-S26 in OVERRIDE automatically bypasses the HPCS-V-4 high RPV water level interlock. No operator actions are required to maintain HPCS-V-4 open when the high level is reached.

D (incorrect): See C. (2) No operator action is required.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-34 The plant is operating at 100% power with OSP-SLC/IST-Q701, SLC Pumps Operability Test, in progress. A plant transient occurs, and the Operators performing the surveillance are directed to terminate the testing. The SLC system is aligned as follows:

• SLC-V-1A and SLC-V-1B (Storage Tank Outlet Valves) are closed

• SLC-V-4A and SLC-V-4B (Squib Valves) are closed and have not been fired

• SLC-V-31 (SLC Test Tank Outlet Valve) is fully open

• SLC-P-1A and SLC-P-1B are off How will the SLC system respond if the SLC SYSTEM A Keylock Switch at H13-P603 is placed in the OPER position?

A. SLC-V-1A will open and SLC-V-1B will remain closed B. SLC-V-1A will remain closed and SLC-V-1B will open C. SLC-V-1A and SLC-V-1B will open D. SLC-V-1A and SLC-V-1B will remain closed ANSWER: D KA # & KA VALUE: 211000 A1.09 Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including: SBLC system lineup. (4.0/4.1)

REFERENCE:

SD000172 Rev.012 pg.11; EWD-10E-003 SOURCE: New LO: LO-5923, LO-5925 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D. SLC-V-1A remains closed.

B (incorrect): See D. SLC-V-1B remains closed.

C (incorrect): See D. SLC-V-1A and SLC-V-1B remain closed.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (correct): SLC-V-1A/1B are interlocked with SLC-V-31 such that they will not open unless SLC-V-31 is fully closed.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-35 Which of the following process radiation monitors is powered by RPS System A?

A. MS-RIS-610A and MS-RIS-610B (Main Steam Line Rad Monitors)

B. MS-RIS-610A and MS-RIS-610C (Main Steam Line Rad Monitors)

C. REA-RIS-609A and REA-RIS-609B (Reactor Building Exhaust Rad Monitors)

D. REA-RIS-609A and REA-RIS-609C (Reactor Building Exhaust Rad Monitors)

ANSWER: B KA # & KA VALUE: 212000 K1.05 Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following: Process radiation monitoring system. (3.3/3.6)

REFERENCE:

SD000147 (PRM) Rev.012 pg.36 SOURCE: New LO: LO-5951, LO-5649 RATING: L2 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B.

B (correct): MS-RIS-610A/C are powered from RPS-A. MS-RIS-610B/D are powered from RPS-B.

C (incorrect): Unlike the MS line rad monitors, the A and B channels of REA rad monitors are Div 1 powered, and the C and D channels are Div 2 powered. (PP-7AA and PP-8AA)

D (incorrect): See C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-36 Installing all four (4) of the RPS Shorting Links causes A. the SRMs to generate scrams using one out of two taken twice logic.

B. any single SRM channel trip to generate a reactor scram (non-coincident).

C. the IRMs and APRMs to generate scrams using one out of two taken twice logic.

D. any single IRM or APRM channel trip to generate a reactor scram (non-coincident).

ANSWER: C KA # & KA VALUE: 212000 2.1.28 RPS. Knowledge of the purpose and function of major system components and controls. (4.1/4.1)

REFERENCE:

SD000161 Rev.015 pg.6 SOURCE: New LO: LO-7678 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): Installing the Shorting Links will prevent the SRMs from generating a reactor scram, but will cause the other nuclear instrumentation to use one out of two taken twice logic.

B (incorrect): Removing the RPS Shorting Links has this effect.

C (correct): The RPS Shorting Links function to bypass the non-coincident nuclear instrument trips. With the Shorting Links installed, the IRMs and APRMs can only generate a scram by satisfying the one out of two taken twice RPS scram logic.

D (correct): Removing the RPS Shorting Links has this effect.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-37 A plant startup is in progress with all IRMs indicating 20 on range 7, when a loss of 24 VDC power to IRM-G occurs.

The effect(s) of this failure is(are)

A. a 1/2 scram AND RMCS will only allow control rods to be inserted.

B. RMCS will block control rod withdrawal ONLY.

C. RMCS will not allow any control rod movement ONLY.

D. a 1/2 scram AND RMCS will not allow any control rod movement.

ANSWER: A KA # & KA VALUE: 215003 K3.02 Knowledge of the effect that a loss or malfunction of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM will have on following:

Reactor manual control system. (3.6/3.6)

REFERENCE:

SD000138 (IRM); SD000148 (RMCS)

SOURCE: New LO: LO-7637 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): The stem states reactor power is 20 on Range 7, which is below the POAH (25 on Range 8). This condition requires the MODE Switch to be in Startup/Hot Standby, which means IRM rod blocks and scram signals are enforced. A loss of 24VDC power to IRM-G results in an INOP rod withdrawal block from RMCS and a 1/2 scram from RPS-A.

B (incorrect): See A. A 1/2 scram will also be generated.

C (incorrect): See A. RMCS will allow control rod insertion and a 1/2 scram will be generated.

D (incorrect): See A. RMCS will allow rod insertion.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-38 A plant startup is in progress with IRMs B, C, and E on Range 2. All other IRMs are on Range 3.

CRO1 is withdrawing SRM-D when the RETRACT PERMIT light extinguishes.

Based on these indications, SRM-D A. counts are LE 100 cps. The detector continues to withdraw, and a rod withdrawal block is generated.

B. counts are LE 100 cps. Detector withdrawal automatically stops. A rod withdrawal block is generated.

C. period is LE +50 secs. The detector continues to withdraw. No rod withdrawal block is generated.

D. period is LE +50 secs. Detector withdrawal automatically stops. No rod withdrawal block is generated.

ANSWER: A KA # & KA VALUE: 215004 K5.03 Knowledge of the operational implications of the following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM:

Changing detector position. (2.8/2.8)

REFERENCE:

SD000132 Rev.012 pg.22 and 23.

SOURCE: New LO: LO-5942, LO-12002 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): The RETRACT PERMIT light extinguishes when counts on the associated detector are LE 100 cps. If all IRMs in the same division as that SRM are on Range 3 or higher, no rod block is generated. The stem states that IRM-B is on Range 2, which means a rod block will be generated. This does not stop detector motion, so the detector will continue to withdraw.

B (incorrect): See A. Detector motion will not stop.

C (incorrect): See A. A period of +50 secs or shorter will result in an SRM

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 Fast Period alarm and indication, but will not cause a rod block or stop detector withdrawal.

D (incorrect): See A. A period of +50 secs or shorter will result in an SRM Fast Period alarm and indication, but will not cause a rod block or stop detector withdrawal.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-39 The plant is operating at 90% power for economic dispatch. With an edge rod selected, electrical interference causes LPRM 16-49 indication to spike to 110 watts/cm2 for five (5) seconds, and then return to actual power level indication.

During this transient, the UPSC indication on the full core display for LPRM 16-49 A. will NOT illuminate, and the LPRM DET BYPASS white lights on H13-P-603 will be illuminated.

B. will NOT illuminate, but the UPSC light on H13-P608 for 16-49 will illuminate for five (5) and then clear.

C. will illuminate and remain lit until the associated Trip Reset pushbutton on H13-P608 is depressed.

D. will illuminate for five (5) seconds and then automatically clear.

ANSWER: D KA # & KA VALUE: 215005 A3.02 Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including: Full core display. (3.5/3.5)

REFERENCE:

SD000143 (LPRM) Rev.011 pg.9-10 SOURCE: New LO: LO-5501 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D. The LPRM DET BYPASS white lights will be illuminated due to the selection of an edge rod, but this does not bypass the UPSC indication on the full core display.

B (incorrect): See D. The UPSC indication on H13-P608 will seal-in, and must be manually reset.

C (incorrect): See D. The UPSC light on H13-P608, not H13-P603 will seal-in and must be reset with the Trip Reset pushbutton.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (correct): The UPSC indication for LPRMs on the full core display illuminates when the high flux setpoint (100 watts/cm2) is exceeded, and automatically reset (extinguish) when the condition clears.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-40 Columbia is operating at 100% power, when a leak develops in the steam supply line to RCIC.

OPS2 reports the sound of steam in the Reactor Building, and the LEAK DET RWCU/RCIC PIPE AREA TEMP HI alarm is received.

What is the impact to RCIC-V-63 and RCIC-V-8 (RCIC Steam Supply Isolation Valves)?

A. RCIC-V-63 has closed, and RCIC-V-8 will close when the LEAK DET RCIC PIPE ROUTING AREA TEMP HI-HI alarm is received at 160°F.

B. RCIC-V-8 has closed, and RCIC-V-63 will close when the LEAK DET RCIC PIPE ROUTING AREA TEMP HI-HI alarm is received at 160°F.

C. RCIC-V-63 and RCIC-V-8 will close when the LEAK DET RCIC PIPE ROUTING AREA TEMP HI-HI alarm is received at 160°F.

D. RCIC-V-63 and RCIC-V-8 closed when the LEAK DET RWCU/RCIC PIPE AREA TEMP HI was received at 140°F.

ANSWER: C KA # & KA VALUE: 217000 A2.15 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. (3.8/3.8)

REFERENCE:

PPM 4.601.A2 1-1; PPM 4.601.A12 6-2 SOURCE: New LO: LO-5722 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): RCIC-V-8 and RCIC-V-63 are divisional, and are not closed by the same temperature switches, but do have the same isolation setpoint.

B (incorrect): RCIC-V-8 and RCIC-V-63 are divisional, and are not closed by the same temperature switches, but do have the same isolation setpoint.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (correct): The LEAK DET RWCU/RCIC PIPE AREA TEMP HI alarm does not cause a RCIC isolation. Although RCIC-V-63 and RCIC-V-8 are divisional and are closed by different switches, their isolation setpoint is the same (160°F). Both valves will close when the LEAK DET RCIC PIPE ROUTING AREA TEMP HI-HI alarm is received.

D (incorrect): See C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-41 With the plant operating at 100% power, MS-LIS-38A (ADS Trip System A Level 3 Input) was being calibrated per ISP-MS-Q923. The equalizing valve for MS-LIS-38A was open when the following occurred:

• A LOCA inside containment

• SM-8 locked out RPV level has just reached Level 1.

Assuming no operator actions have been taken A. the ADS SRVs opened when RPV level reached Level 1.

B. the ADS SRVs will open as soon as the 105 second time delay times out.

C. ADS Division 1 and ADS Division 2 will NOT automatically initiate. The ADS SRVs will remain closed.

D. ADS Division 2 will NOT automatically initiate. ADS Division 1 will initiate when the ECCS pump permissive is met.

ANSWER: C KA # & KA VALUE: 218000 K6.03 Knowledge of the effect that a loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM:

Nuclear boiler instrument system (level indication). (3.8/3.9)

REFERENCE:

SD000186 (ADS) Rev.011 pg.5 and 6 SOURCE: Modified Bank (LX00220)

LO: LO-11878 RATING: H3 ATTACHMENT: None.

JUSTIFICATION: A (incorrect): See C. Level 1 is a permissive for ADS. The stem states a LOCA occurred, which implies the low pressure ECCS pumps are already running.

B (incorrect): See C. The 105 sec TD is a permissive for ADS actuation.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (correct): With the equalizing valve open, MS-LIS-38A will indicate upscale and will not provide the +13 confirmatory signal for ADS Logic A.

Logic A and Logic C are required for ADS Div 1 to initiate. The stem states a lockout exists on SM-8. No Div 2 ECCS will have power, which will prevent ADS Div 2 from initiating. All ADS SRVs remain closed.

D (incorrect): See C. The first part is true. But ADS Div 1 will not initiate.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-42 A hydraulic ATWS has occurred, and PPM 5.5.6, Bypass of MSIV Isolation Interlocks, has been completed.

The MSIVs will automatically close if A. Main Condenser backpressure is 28Hg B. Main Steam Tunnel temperature is 170°F C. RPV water level is -135 D. The T between the Main Steam Tunnel and the RB 522 is 85°F ANSWER: A KA # & KA VALUE: 223002 K4.08 Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following: Manual defeating of selected isolations during specified emergency conditions. (3.3/3.7)

REFERENCE:

PPM 5.5.6; SD000173 (NS4); SD000128 (MS); EWD-19E-038, 009, and 013 SOURCE: New LO: LO-7719 RATING: L2 ATTACHMENT: None JUSTIFICATION: A (correct): MS-RMS-S84 and MS-RMS-S85 are placed in BYPASS when PPM 5.5.6 is performed. This bypasses automatic MSIV closure on high temperature, high delta-T, and low RPV level. The high backpressure isolation (low vacuum) can be bypassed using keylock switches, but not the switches used per PPM 5.5.6, and not during an ATWS. The setpoint is 21.6 Hg.

B (incorrect): See A. The setpoint has been exceeded (164°F).

C (incorrect): See A. The setpoint has been exceeded (-129).

D (incorrect): See A. The setpoint has been exceeded (80°F).

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-43 When the control switch for MS-RV-3D at H13-P603 is placed in OPEN A. the C solenoid is energized from DP-S1-1A.

B. the C solenoid is energized from DP-S1-2A.

C. the A solenoid is energized from DP-S1-1A.

D. the B solenoid is energized from DP-S1-2A.

ANSWER: A KA # & KA VALUE: 239002 K2.01 Knowledge of the electrical power supplies to the following:

SRV solenoids. (2.8/3.2)

REFERENCE:

SD000128 (MS) Rev.010 pg.28.

SOURCE: New LO: LO-5537, LO-7652 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (correct): The C solenoid is energized from DP-S1-1A from H13-P603.

B (incorrect): DP-S1-2A is the power supply to the B solenoid, which is operated from H13-P631.

C (incorrect): The A solenoid is powered from DP-S1-1A, but energized when the switch at H13-P628 is operated.

D (incorrect): DP-S1-2A is the power supply to the B solenoid, which is operated from H13-P631.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-44 With the plant operating at 100% power, a LOCA resulted in a reactor scram. Both Feed Pumps tripped on Level 8, and the CRS has directed level control using Feed and Condensate.

After resetting the trip of RFW-P-1A per SOP-RFT-RESTART-QC, how should Feed Pump speed be raised, and what is the expected Feed Pump response?

A. Operating in MDVP mode, slowly raise governor valve position. Feed Pump speed will increase when the governor valve comes off the closed seat.

B. Operating in MDVP mode, slowly raise governor valve position. Feed Pump speed will begin to increase when the governor valve is approximately 60% open.

C. Operating in MDEM mode, slowly raise governor valve position. Feed Pump speed will increase when the governor valve comes off the closed seat.

D. Operating in MDEM mode, slowly raise governor valve position. Feed Pump speed will begin to increase when the governor valve is approximately 60% open.

ANSWER: B KA # & KA VALUE: 259002 A4.01 Ability to manually operate and/or monitor in the control room:

All individual component controllers in the manual mode. (3.8/3.6)

REFERENCE:

SOP-RFT-RESTART-QC; SD000157 (FWLC) Rev.015 pg.13; SD000151 (RFW) Rev.012 pg SOURCE: New LO: LO-11701 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B. This is the expected response if the Main Turbine were not tripped.

B (correct): The Feed Pumps are started and re-started in the MDVP mode of operation. Operation is not transferred to MDEM until approximately 800 rpm. The stem states the Feed Pumps tripped due to Level 8. This is also a Main Turbine trip. With the Main Turbine tripped, steam is not admitted to the Feed Pump until the governor is approximately 60% open.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): See B. Operation is not transferred to MDEM until approximately 800 rpm.

D (incorrect): See B. Operation is not transferred to MDEM until approximately 800 rpm.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-45 SGT-FN-1A1 (SGT-A Lead Fan) has been operating for ten (10) minutes following an automatic system initiation, when the fan seizes and clears its fuses.

Which valves reposition in preparation for the operation of SGT-FN-1A2 (SGT-A Lag Fan)?

A. SGT-V-3A2 (Lead Fan Inlet Valve) closes SGT-V-5A2 (Lag Fan Exhaust Valve) opens B. SGT-V-3A1 (Lag Fan Inlet) opens SGT-V-5A2 (Lag Fan Exhaust Valve) opens C. SGT-V-5A1 (Lead Fan Exhaust Valve) closes SGT-V-5A2 (Lag Fan Exhaust Valve) opens D. SGT-V-3A2 (Lead Fan Inlet Valve) closes SGT-V-5A1 (Lead Fan Exhaust Valve) closes ANSWER: A KA # & KA VALUE: 261000 A3.03 Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: Valve operation. (3.0/2.9)

REFERENCE:

SD000144 Rev.014 pg 7, 8, 9, and 14.

SOURCE: New LO: LO-5828 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (correct): When a low flow conditions exists with the Lead Fan running, the following will occur: SGT-V-3A2 closes, SGT-V-5A2 opens, and the Lag Fan starts.

B (incorrect): SGT-V-3A1 (Lag Fan Inlet) is normally open, and therefore did not reposition to open on the low flow condition.

C (incorrect): The Lead Fan Inlet valve, not the outlet (SGT-V-5A1) automatically closes on low flow.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (incorrect): The Lead Fan Inlet valve, not the outlet (SGT-V-5A1) automatically closes on low flow.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-46 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 voltage condition on E-TR-S (Startup Transformer).

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

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

D. degraded frequency condition on E-TR-B (Backup Transformer)

ANSWER: A KA # & KA VALUE: 262001 A1.03 Ability to predict and/or monitor changes in parameters associated with operating the A.C. ELECTRICAL DISTRIBUTION controls including: Bus voltage. (2.9/3.1)

REFERENCE:

SOP-CW-START Rev.007 pg.7; LER 84-079 SOURCE: Bank LO: LO-5050 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (correct): Starting a third CW Pump while E-TR-S is supplying house loads will cause an undervoltage or degraded voltage on E-TR-S and the electrical buses it is powering.

B (incorrect): Frequency is determined by the 230kv grid, and is relatively constant. Starting the third pump will have minimal impact.

C (incorrect): 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,

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 starting the third CW Pump will not impact E-TR-B.

D (incorrect): See B and C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-47 With the plant operating at 100% power, the E-IN-1 inverter voltage rises to 110% 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. alternate AC source (MC-7F). Enter ABN-POWER.

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

C. bypass AC source (MC-7A). Enter ABN-POWER.

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

ANSWER: B KA # & KA VALUE: 262002 A2.02 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.

(2.5/2.7)

REFERENCE:

SD000194 (UPS) Rev.010 pg.8; ABN-ELEC-INV Rev.006 pg.3 SOURCE: New LO: LO-5896, LO-5891 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): 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.

B (correct): The overvoltage condition causes a bumpless transfer to the alternate power supply. ABN-ELEC-INV will be used to determine the fault with IN-1.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (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.

D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-48 Which of the following correctly describes the effect(s) a complete loss of 125VDC bus S1-2 has on ATWS/ARI?

The Division 2 ATWS/ARI valves A. maintain position indication but can NOT be operated from the Control Room.

B. lose position indication but can still be operated from the Control Room.

C. lose position indication and can NOT be operated from the Control Room.

D. lose position indication and fail open.

ANSWER: C KA # & KA VALUE: 263000 Knowledge of the effect that a loss or malfunction of the D.C.

ELECTRICAL DISTRIBUTION will have on following: Systems with D.C.

components (i.e. valves, motors, solenoids, etc.)

REFERENCE:

SD000188 (DC) Rev.009 pg.26 SOURCE: Bank LO: LO-7653 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): S1-2 is power for position indication and operation.

B (incorrect): S1-2 is power for position indication and operation.

C (correct): S1-2 supplies power to the indication and control for Div 2 ATWS/ARI valves.

D (incorrect): The ATWS/ARI valves (CRD-SPV-24B through 27B) are normally deenergized, and must be energized to open and vent the scram air header. The valves used to vent the scram air header associated with RPS (CRD-SPV-117 and 118) deenergize to open, making this a plausible distractor.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-49 The Main Control Room has been evacuated per ABN-CR-EVAC. The 125 VDC Division 1 Battery voltage has been reported to be 106 VDC.

What action is required per ABN-CR-EVAC based on this indication, and why is this action being taken?

A. Emergency depressurize the RPV, because the ADS valves may fail to open less than 105 VDC.

B. Emergency depressurize the RPV, because the control power to operate RHR from the ARSD may not be available.

C. Reduce battery loads by opening 125 VDC Division 1 disconnects to extend the availability of SRV operation.

D. Reduce battery loads by opening 125 VDC Division 1 disconnects to ensure ARSD panel control power remains available.

ANSWER: A KA # & KA VALUE: 263000 2.1.23 DC Electrical Distribution. Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(3.9/4.2)

REFERENCE:

ABN-ELEC-125VDC Rev.009 pg.14; ABN-CR-EVAC Rev.025 pg.4 and 16; PPM 5.6.1 Rev.022 pg.12 and 23 SOURCE: New LO: LO-6889 RATING: H4 ATTACHMENT: None JUSTIFICATION: A (correct): Per ABN-CR-EVAC, emergency depressurization is required if the Division 1 125 VDC battery voltage is LE 108 VDC, because the SRVs may not open LT 105 VDC.

B (incorrect): ABN-ELEC-125VDC states remote operation (due to loss of control power) may not be available LT 105 VDC, which makes this distractor plausible. But this is not the reason ABN-CR-EVAC requires

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 emergency depressurization.

C (incorrect): See A. PPM 5.6.1 requires battery loads to be reduced to keep the SRVs available for a longer period of time, not ABN-CR-EVAC.

D (incorrect): See A and C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-50 The plant is operating in Mode 1 with OSP-ELEC-M701, DG-1 Monthly Operability Test, in progress. DG-1 has been operating in parallel with E-TR-S at a load of 4100 KW for the past hour.

CRO2 observes the MVAR meter for DG-1 deflect upscale, and hears OPS2 report the local MVAR meter is upscale over the radio.

CRO2 will A. reduce DG-1 load to less than 200 KW and open E-CB-DG1/7 (DG-1 Output Breaker).

B. place the Voltage Regulator Control Switch in the LOWER position until MVAR indication returns to the normal operating range.

C. trip DG-1 using the DG Trip Pushbutton, and perform OSP-ELEC-W101, Offsite Station Power Alignment Check.

D. place the Governor Control Switch in the LOWER position until MVAR indication returns to the normal operating range.

ANSWER: C KA # & KA VALUE: 264000 2.2.12 Emergency Generators (Diesel/Jet). Knowledge of surveillance procedures. (3.7/4.1)

REFERENCE:

OSP-ELEC-M701; TS 3.8.1 SOURCE: New LO: None RATING: L4 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. This is the action required to be taken if the MVAR meter deflects and stays left of zero following closure of the output breaker.

B (incorrect): See C. This is the action required to be taken if the MVAR meter deflected upscale following closure of the output breaker. This is indication of an overexcited diesel, and an attempt should be made to correct the condition.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (correct): The conditions given in the stem arean indication of voltage regulator failure. Per OSP-ELEC-M701 4.3, if the MVAR meter deflects full upscale when the DG is operating in parallel to an off-site source, Operator action is required to trip the diesel. Tripping the diesel causes it to be inoperable, and an offsite power alignment is required per TS 3.8.1.

D (incorrect): The Governor Control Switch is used to adjust real load when operating in parallel with an off-site source, not MVARS.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-51 The plant is operating at 100% power with no equipment out of service. The Main Control Room has just received the CONTROL AIR HDR PRESS LOW alarm.

If CAS pressure continues to lower, which system response is expected next?

A. The standby CAS compressors automatically start.

B. SA-PCV-2 will close to isolate the Service Air system.

C. The Service Air compressor automatically starts.

D. CAS-PCV-1 will open to bypass the CAS dryers.

ANSWER: B KA # & KA VALUE: 300000 K1.02 Knowledge of the connection and / or cause effect relationships between INSTRUMENT AIR SYSTEM and the following:

Service Air. (2.7/2.8)

REFERENCE:

SD000205 Rev.010 pg.20 SOURCE: Bank LO: LO-5878 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The standby CAS compressors start at 100 psig, and have already started.

B (correct): The CAS HDR PRESS LOW alarm is received at 95 psig. The next action is the closure of SA-PCV-2 at 80 psig.

C (incorrect): The stem states no equipment is out of service. The Service Air compressor is normally running and does not need to be started.

D (incorrect): CAS-PCV-1 opens at 75 psig (after SA-PCV-2 closes).

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-52 Columbia is operating at rated power with RCC in the following lineup:

• RCC-P-1A running

• RCC-P-1B standby (auto)

• RCC-P-1C running

• RCC-V-6 (Outside Containment Supply) open A 10 amp fuse in the close circuit for RCC-CB-P1A (RCC-P-1A Circuit Breaker) clears resulting in a loss of closing control power to RCC-P-1A.

What response, if any, does the RCC system have to this fault?

A. The RCC system continues to operate normally B. RCC-P-1A stops rotating C. RCC-V-6 closes after 10 seconds D. RCC-P-1B automatically starts ANSWER: C KA # & KA VALUE: 400000 K6.07 Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: Breakers, relays, and disconnects.

(2.7/2.8)

REFERENCE:

SD000196 Rev.011 pg.10; LO000196 slides 74-76 SOURCE: New LO: LO-5706d RATING: L4 ATTACHMENT: None JUSTIFICATION: A (incorrect): No change in pump configuration will occur, but RCC-V-6 will close.

B (incorrect): Because RCC-CB-P-1A does not have an undervoltage trip, the condition described in this distractor is consistent with a loss of motive force for RCC-P-1A, not a loss of control power.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (correct): The close circuit for RCC-CB-P1A also powers the relays associated with the 2 pumps running logic circuit for RCC-V-6. When these relays deenergize, RCC-V-6 automatically closes after a 10 second time delay.

D (incorrect): If RCC-P-1A trips, RCC-P-1B will automatically start. RCC-P-1A will not trip on the loss of control power, so RCC-P-1B will remain in standby.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-53 A plant shutdown is in progress when the RCC RAD HIGH alarm is received. RCC-RIS-607 (RCC Radiation Monitor) indicates 150 cps.

What occurred as a result of this condition?

A. ONLY RCC-V-6 (Outside Containment Supply) automatically closed.

B. RCC-V-104 (Outboard Containment Isolation) automatically closed.

C. The previously running RCC pumps have tripped.

D. RCC-V-101 (Surge Tank Vent Valve) automatically closed.

ANSWER: D KA # & KA VALUE: 400000 A3.01 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. (3.0/3.0)

REFERENCE:

SD000196 Rev.011 pg.8, 10, and 16; PPM 4.602.A5 4-6 SOURCE: New LO: LO-5709 RATING: L2 ATTACHMENT: None JUSTIFICATION: A (incorrect): Automatically closes if LT 2 RCC pumps are running.

B (incorrect): Automatically closes on FA signal.

C (incorrect): Automatically trip on FA signal.

D (correct): RCC-V-101 closes at GE 74 cps on RRC-RIS-607

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-54 Columbia has experienced a hydraulic ATWS. CRO2 has just completed overriding the RPS trip signals in preparation for scram/reset/scram and manually driving control rods. CRD-FI-606 (CRD System Flow) is observed to be upscale.

Based on this indication, which of the following is correct concerning CRD drive header pressure?

A. CRD-FCV-2A (CRD Flow Control Valve) has closed to the minimum position, which will establish normal drive header pressure for control rod insertion.

B. Reset the scram to allow CRD drive header pressure to be established for control rod insertion using the normal drive method.

C. CRD-V-34 (Charging Header Isolation) must be closed to establish normal CRD drive header pressure for driving control rods.

D. CRD-FCV-2A (CRD Flow Control Valve) is fully open, and must be throttled using CRD-FC-600 to establish normal drive header pressure.

ANSWER: B KA # & KA VALUE: 201001 K5.02 Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM:

Flow indication. (2.6/2.6)

REFERENCE:

SD000142 (CRDH) Rev.014 pg.29 and 36; PPM 5.5.11 Rev.005 pg.6 SOURCE: New LO: LO-5185 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): CRD-FI-606 upscale causes CRD-FCV-2A to close to the minimum position, but this lowers drive header pressure.

B (correct): CRD-FI-606 upscale is indication that CRD is flowing to the reactor and the scram accumulators through the scram valves. Resetting the scram will allow CRD-FC-600 and CRD-V-3 to be adjusted to establish drive header pressure and normal rod insertion to commence.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): Closing CRD-V-34 would allow CRD drive header pressure to be established, but this action is not taken unless pressure cannot be established by resetting the scram. CRD-V-34 is not required to be closed.

D (incorrect): Manual control of CRD-V-2A using CRD-FC-600 will be required to establish drive header pressure, but CRD-FCV-2A is closed to minimum position, not fully open.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-55 A plant startup is in progress when CRO1 notices an i next to control rod 30-31 on the RWM display.

What condition would cause this indication?

A. Rod 30-31 is in the currently latched group and is inserted past the insert limit for this group.

B. Rod 30-31 is in the currently latched group and is inserted past the withdraw limit for this group.

C. RWM has lost rod position information for control rod 30-31, and 30-31 is in the currently latched group.

D. RWM has lost rod position information for control rod 30-31, and 30-31 is in a lower group than the currently latched group.

ANSWER: A KA # & KA VALUE: 201006 A4.05 Ability to manually operate and/or monitor in the control room:

Rod insert error indication. (3.2/3.2)

REFERENCE:

SD000154 (RWM) Rev.014 pg.8 and 15 SOURCE: Modified Bank (LO00124)

LO: LO-5915b, LO-5908e RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): An i on RWM next to a control rod indicates an insert error. An insert error will occur if a control rod in the currently latched group is inserted past the insert limit for that group.

B (incorrect): An insert error will occur if a rod in a group lower than the currently latched group is inserted past the withdraw limit for that group.

C (incorrect): The i does not stand for information error. A loss of rod position information will be indicated by ??

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (incorrect): The i does not stand for information error. A loss of rod position information will be indicated by ??

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-56 With the plant operating in MODE 2, LCO 3.4.2, Jet Pumps, requires ______ jet pumps to be operable.

A. 17 B. 18 C. 19 D. 20 ANSWER: D KA # & KA VALUE: 202001 2.2.22 Recirculation System. Knowledge of limiting conditions for operations and safety limits. (4.0/4.7)

REFERENCE:

SD000178 (RRC) Rev.015 pg.26; LCO 3.4.2 SOURCE: Bank LO: LO-5023a RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D.

B (correct): See D.

C (incorrect): See D.

D (incorrect): In MODES 1 and 2, LCO 3.4.2 requires all jet pumps to be operable. There are 20 total jet pumps.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-57 The plant was operating at 100% power when a failure in the ASD system caused RRC-P-1B speed to begin lowering. RRC pump speeds are as follows:

• RRC-P-1A: 58.64 Hz

• RRC-P-1B: 28.74 Hz What is the impact of operating at this RRC pump speed mismatch, and what action will be taken to mitigate the condition?

Reverse flow through the jet pumps in the B loop could occur and result in A. overheating of RRC-P-1B due to inadequate flow. Stop RRC-P-1B.

B. overheating of RRC-P-1B due to inadequate flow. Reduce RRC-P-1A flow to within the allowable limits.

C. instabilities and excessive vibration in the jet pumps. Stop RRC-P-1B.

D. instabilities and excessive vibration in the jet pumps. Reduce RRC-P-1A flow to within the allowable limits.

ANSWER: C KA # & KA VALUE: 202002 A2.04 Ability to (a) predict the impact of the following on the RECIRCULATION FLOW CONTROL SYSTEM; and (b) based on these predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Recirculation pump speed mismatch between loops. (3.0/3.2)

REFERENCE:

ABN-POWER Rev.013 pg.4, 5, and 10 SOURCE: New LO: LO-6747 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. A flow path for the affected pump still exists through the nozzles.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 B (incorrect): See C and D. A flow path for the affected pump still exists through the nozzles.

C (correct): Operating RRC Pumps outside of the allowable mismatch causes instabilities and jet pump vibrations. RRC-P-1B speed has decreased such that the ratio of pumps speeds has exceeded a 2:1 ratio.

The immediate actions of ABN-POWER require the affected pump to be stopped.

D (incorrect): See C. The subsequent actions of ABN-POWER require these actions if the pump speed ratio doesnt exceed 2:1.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-58 The plant is operating at 90% power with an interior rod selected, when APRM-C fails downscale.

What effect will this have on the Rod Block Monitor (RBM) system?

A. RBM channel A will be automatically bypassed.

B. Reference power level for RBM channel A will be provided by LPRMs.

C. The RBM channel A averaging circuit output will be lowered to equal APRM-C.

D. RBM channel A will generate an INOP rod block.

ANSWER: A KA # & KA VALUE: 215002 K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the ROD BLOCK MONITOR SYSTEM: APRM reference channel. (2.8/3.0)

REFERENCE:

SD000166 (RBM) Rev.012 pg.3, 11, 13, and 19 SOURCE: Bank LO: LO-7667 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): APRM-C is the reference APRM for RBM-A. When the reference APRM is LT 30% power, the RBM channel automatically bypasses.

B (incorrect): See A. LPRMs are used to generate the RBM channel signal, and do not provide any input to the reference power signal.

C (incorrect): The RBM averaging circuit output will not lower to the APRM value. It will only be raised by the gain circuit.

D (incorrect): A rod block will be generated by the downscale APRM with the mode switch in RUN through RMCS. RBM does not generate the control rod block.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-59 The plant is operating at 100% power with the following indications:

• RFW-LI-606A (Narrow Range) indicates +36

• MS-LI-604 (Wide Range) indicates +18 Based on these indications A. no action is required. The difference in indication is caused by the pressure draw-down effect of jet pump flow past the Wide Range variable leg.

B. no action is required. The Wide Range is calibrated for 0 psig reactor pressure and the Narrow Range is calibrated for 1000 psig reactor pressure, which causes the difference in indication.

C. the Narrow Range instrument has failed. CRO1 should take manual control of FWLC and restore RPV water level based on Wide Range indication.

D. the Narrow Range instrument has failed. CRO1 should select a different Narrow Range channel for FWLC using the Reactor Vessel Level Control Channel Selector Switch at H13-P603.

ANSWER: A KA # & KA VALUE: 216000 K5.09 Knowledge of the operational implications of the following concepts as they apply to NUCLEAR BOILER INSTRUMENTATION:

Recirculation flow effects on level indications. (2.9/2.9)

REFERENCE:

SD000126 (NBI) Rev.012 pg.15 SOURCE: Modified Bank (LO02154) (October 2009 Exam)

LO: LO-5588 RATING: L2 ATTACHMENT: None JUSTIFICATION: A (correct): Per the reference, Wide Range indicators will indicate 15 to 20 lower than Narrow Range indicators at 100% power due to the draw-down effect of jet pump flow past the wide range instrument variable line connection.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 B (incorrect): See A. The Wide Range and Narrow Range instruments are both calibrated for 1000 psig pressure. Shutdown and Fuel Zone instruments are calibrated for 0 psig reactor pressure.

C (incorrect): See A. This action would be taken if controller failure was suspected.

D (incorrect): See A. This action would be taken for a failure of the selected NR instrument, or to follow the automatic response of the FWLC system to a failure.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-60 The plant is operating at rated power. RHR-P-2A is running and spraying the Wetwell for surveillance testing, when a DBA LOCA occurs.

Which of the following is correct concerning the operation of RHR-V-27A (Wetwell Spray Valve) and RHR-V-42A (Injection Valve)?

A. RHR-V-27A will remain open, and RHR-V-42A will automatically open.

B. RHR-V-27A will remain open. RHR-V-42A can NOT be opened until RHR-V-27A is closed.

C. RHR-V-27A will automatically close, and can NOT be opened until RHR-V-42A is closed.

D. RHR-V-27A will automatically close, but can be reopened regardless of RHR-V-42A position.

ANSWER: C KA # & KA VALUE: 230000 K4.03 Knowledge of RHR/LPCI: TORUS/SUPPRESSION POOL SPRAY MODE design feature(s) and/or interlocks which provide for the following: Unintentional reduction in vessel injection flow during accident conditions. (3.5/3.6)

REFERENCE:

SD000198 (RHR) Rev.013 pg.16-18.

SOURCE: New LO: LO-5781 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. If RHR-V-27A is opened with RHR-V-42 A closed after an initiation signal, RHR-V-27A will remain open when RHR-V-42A is subsequently opened.

B (incorrect): See C. RHR-V-42A is only interlocked closed by reactor pressure. A DBA LOCA depressurizes the RPV allowing RHR-V-42A to open.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (correct): RHR-V-27A closes on a LPCI initiation signal, which will occur as a result of a DBA LOCA. RHR-V-27A is interlocked closed when RHR-V-42A is open.

D (incorrect): See C and A. The reopening of RHR-V-27A depends on RHR-V-42A being closed.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-61 A plant startup is in progress with reactor power at 25%. CRO2 observes the Main Steam Line Low Power Drain Valves (MS-V-69, MD-V-73, and MS-V-156) are open with their control switches in AUTO.

The Main Steam Line Low Power Drain Valves A. will automatically close when reactor power is greater than 25%.

B. will automatically close when Main Steam flow is greater than 50% of rated flow.

C. have no automatic features, and should have been manually closed at 5% reactor power.

D. have no automatic features, and should have been manually closed at 300 psig reactor pressure.

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

REFERENCE:

SD000128 (MS) Rev.010 pg.17; SD000157 (FWLC) Rev.015 pg.17 SOURCE: Modified Bank (LX00188)

LO: LO-5534 RATING: L4 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B. Reactor power of 25% is the setpoint at which the Main Turbine Bypass Valves (MS-V-160A through D) automatically open.

B (correct): The MSL Low Power Drain Valves auto close at 50% rated steam flow as measured by the MSL flow restrictors.

C (incorrect): See B. This describes the operation of Main Steam drain valves MS-V-67A through D.

D (incorrect): See B. This describes the operation of the Main Steam Line

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 Startup Drain Valves (MS-V-21, MS-V-68, and MD-V-78) in AUTO.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-62 The Main Turbine is being latched per SOP-MT-START following a trip for testing during a plant startup. Main Turbine speed is 450 rpm when the Reactor Operator presses LATCH and OK on the DEH screen.

Which of the following describes the automatic response to this action?

Main Turbine speed A. rises to 1750 rpm. The Governor Valves ramp full open as Valve Limit Position demand ramps to 100%.

B. rises to 1750 rpm. SPEED TARGET is automatically set to 1750 rpm and SPEED RATE is set to 80 rpm/min.

C. continues to decrease, because the Throttle Valves are held closed by the Valve Position Limit demand signal at 0%.

D. continues to decrease. DEH enters SPEED CONTROL mode and sets SPEED TARGET and SPEED RATE to 0 rpm.

ANSWER: D KA # & KA VALUE: 241000 A1.13 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including: Main turbine speed. (2.7/2.7)

REFERENCE:

SOP-MT-START; SD000146 (DEH) Rev.010 pg.7 SOURCE: New LO: LO-11669 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): This describes system response when the Main Turbine is latched GT 500 rpm.

B (incorrect): This describes system response when the Main Turbine is latched GT 500 rpm.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): Main Turbine speed will continue to decrease, but the Governor Valves, not the Throttle Valves, are held closed by the VPL.

D (correct): Per references, this is the response less than 500 rpm.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-63 At 100% power, the normal suction source for the CRD Pumps is from the ______.

A. Condensate Storage Tanks B. discharge of the Reactor Feed Pumps C. discharge of the Condensate Booster Pumps D. outlet of the Condensate Filter Demineralizers ANSWER: D KA # & KA VALUE: 256000 K1.05 Knowledge of the physical connections and/or cause-effect relationships between REACTOR CONDENSATE SYSTEM and the following: CRD hydraulics system. (3.1/3.1)

REFERENCE:

SD000142 (CRDH) Rev.014 pg.5 SOURCE: Bank LO: LO-5192 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D.

B (incorrect): See D.

C (incorrect): See D.

D (correct): The Condensate system is aligned to the suction of the CRD Pumps from the discharge of the CFDs through COND-PCV-5. The CST Tanks are also normally aligned to the suction of the CRD Pumps. Because Condensate pressure is higher, it is the normal suction source.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-64 RFW-P-1B is being placed in service as the second Reactor Feed Pump per SOP-RFT-START.

Feedwater Level Control is aligned as follows:

• RFW-P-1A is in AUTO

• RFW-P-1B is in MDEM with RFW-V-102B (Pump Discharge Valve) open

• RPV level is being maintained by RFW-LIC-600 (RPV Master Level Controller) in AUTO If CRO1 depresses the UP arrow for RFW-P-1B on RFT-COMP-1, RFW-P-1B speed will A. rise and RFW-P-1A speed will lower.

B. rise and RFW-P-1A speed will remain the same.

C. rise and RFW-P-1A speed will rise.

D. remain the same and RFW-P-1A speed will remain the same.

ANSWER: A KA # & KA VALUE: 259001 A1.04 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including: RFP turbine speed. (2.8/2.7)

REFERENCE:

SOP-RFT-START Rev.019 pg.58; SD000157 (FWLC) Rev.015 pg.8,9, and 13 SOURCE: New LO: LO-5394, LO-11701, LO-11702 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): With the conditions provided in the stem, depressing the UP arrow for RFW-P-1B will cause the B feed turbine speed to rise. This will result in an increase in feed flow, which the FWLC system will detect and respond to by lowering the speed of RFW-P-1A.

B (incorrect): See A. If RFW-V-102B were not open, this distractor would be correct. RFW-V-102B is opened when RFW-P-1B discharge pressure is within 20 to 30 psi of RFW-P-1A discharge pressure. As a result, RFW-P-1B

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 is feeding the reactor in this lineup, and raising the speed of one pump will cause the speed of the other pump to lower.

C (incorrect): See A. The UP arrows on RFT-COMP-1 control feed turbine speeds individually. The response in this distractor would be expected if the INC button on RFW-LIC-600 was depressed with both pumps in AUTO.

D (incorrect): See A. The UP arrow will not raise the speed of a feed turbine if the turbine has not been reset following a trip. The speed of the other feed turbine would also remain unchanged in that situation.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-65 Columbia was operating at 85% power with ROA-FN-1A and REA-FN-1A in service. A fault in the control circuit for ROA-FN-1B caused the fan to automatically start, and prevents any manual or automatic tripping of ROA-FN-1B.

Which of the following is correct concerning this transient?

A. ROA-FN-1A will trip at 4 and allow REA-FN-1A to restore normal Secondary Containment pressure.

B. ROA-AD-5 (Outside Air Suction Damper) will begin to open when pressure reaches 3.5 and limit pressure to 4.7.

C. Secondary Containment pressure will rise until over-pressurization causes a failure of the Secondary Containment structure.

D. REA-FN-1B will automatically start due to the low dp across REA-FN-1A. The two supply, two exhaust fan lineup will restore normal Secondary Containment pressure.

ANSWER: B KA # & KA VALUE: 290001 K4.02 Knowledge of SECONDARY CONTAINMENT design feature(s) and/or interlocks which provide for the following: Protection against over pressurization. (3.4/3.5)

REFERENCE:

LER 88-007; SD000183 Rev.010 pg.8, 16, 17, and 24.

SOURCE: New LO: LO-5677 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): ROA-FN-1A will trip at 4, but so will REA-FN-1A. This will leave ROA-FN-1B as the only running fan and cause Secondary Containment pressure to continue to rise until limited by ROA-AD-5.

B (correct): This failure occurred at CGS in 1988. As a result, ROA-AD-5 was installed to limit the over-pressure transient to 4.7.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): This occurred in the CGS event. ROA-AD-5 will now prevent this from happening.

D (incorrect): The REA fans do have an auto start feature on low running fan dp. This transient would not cause a low dp across the running REA fan, though. A two supply, two exhaust fan lineup is abnormal for RB HVAC, but is a familiar configuration for Turbine Building HVAC.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-66 The plant is operating at 100%. The only reactivity manipulation this shift was lowering RRC flow

.04 Hz to maintain 100% power. Bistable flow is causing reactor power to fluctuate as high as 100.2%.

According to PPM 1.3.1, Operating Policies, Programs and Practices:

A. Action must be taken immediately to reduce power to 3486 MWt or lower.

B. The power fluctuations may be allowed to continue to occur, but action must be taken to reduce power if the 15 minute average exceeds 100%.

C. The power fluctuations may be allowed to continue to occur, but action must be taken to reduce power if the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 100%.

D. The power fluctuations may be allowed to continue to occur, but action must be taken to reduce power if the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> average exceeds 100%.

ANSWER: C KA # & KA VALUE: 2.1.1 Knowledge of conduct of operations requirements. (3.8/4.2)

REFERENCE:

PPM 1.3.1 Rev.108 pg.24 SOURCE: New LO: LO-6094 RATING: L2 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C.

B (incorrect): See C.

C (correct): Because the power fluctuations are occurring as a result of bistable flow, they may be allowed to continue. Action must be taken to reduce power if the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> average exceeds 100% regardless of the reason.

D (incorrect): See C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-67 Given the following list:

1. Reactor Operators must review Night Orders prior to assuming the shift.

2. Reactor Operators should review Night Orders after assuming the shift.

3. Night Orders are used to implement long term corrective actions.

4. Operator actions required to satisfy a Night Order are tracked through periodic Night Order audits.

5. New Night Orders should be discussed at the Shift Brief.

Which of the statements is correct concerning Night Orders?

A. Only 1 and 4 B. 1, 3, and 4 C. 2, 3, and 5 D. Only 2 and 5 ANSWER: D KA # & KA VALUE: 2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, operations memos, etc. (2.7/3.4)

REFERENCE:

PPM 1.3.1 Re.108 pg.75 and 77; OI-09 Rev.056 pg.30 SOURCE: New LO: LO-10761 RATING: L3 ATTACHMENT: None JUSTIFICATION: 1. (False) The SM is required to review Night Orders prior to taking the shift, not the CROs.

2. (True) Per PPM 1.3.1

3. (False) Per OI-09, Night Orders are used for interim corrective actions.

4. (False) Per OI-09, a trigger such as a procedure change, caution tag,

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 or a temp log should be used to track required operator actions.

5. (True) Per PPM 1.3.1, Night Orders should be discussed at shift turnover.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-68 While performing a scheduled surveillance in the Main Control Room, a switch manipulation by a Reactor Operator causes an annunciator to alarm. The annunciator tile is not flagged.

The Reactor Operator should A. announce the alarm as expected to the CRS, state the reason for the alarm, and reference the ARP.

B. announce the alarm as expected to the CRS, silence and acknowledge the alarm, and place a flag on the annunciator tile.

C. silence and acknowledge the alarm, and reference the ARP. No announcement is required.

D. announce the alarm identifying the parameter and trend if possible to the CRS, and reference the ARP.

ANSWER: D KA # & KA VALUE: 2.1.38 Knowledge of the stations requirements for verbal communications when implementing procedures.

REFERENCE:

OI-09 Rev.056 pg.20 and 21 SOURCE: New LO: None RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D. Alarms due to the performance of normal rounds and routine activities may be announced as expected. The alarm described in the stem does not meet this criteria.

B (incorrect): See D. Alarms due to the performance of normal rounds and routine activities may be announced as expected. The alarm described in the stem does not meet this criteria.

C (incorrect): See D. Alarms generated as a part of Surveillance testing are not required to be announced, only silenced and acknowledged, provided

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 the annunciator tile is flagged.

D (correct): Per OI-09, annunciators that are not flagged, even if due to an in progress surveillance, must be treated as unexpected annunciators.

Unexpected annunciators must be announced, with the parameter and trend identified if possible. The ARP for that annunciator must also be referenced.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-69 A plant startup is in progress following a refueling outage. CRO2 is starting RCC-P-1B, the first RCC pump to be started, per SOP-RCC-START. OPS2 has closed RCC-V-2B (Pump Discharge) per the procedure, and CRO2 has started RCC-P-1B. The next step of the procedure states, SLOWLY OPEN RCC-V-2A.

CRO2 should A. correct the EPN as an editorial change and proceed with the system startup.

B. correct the EPN as a verbal PCN and proceed with the system startup.

C. report the procedure issue to the CRS. The CRS will approve an editorial change to allow system startup to continue.

D. report the procedure issue to the CRS. The CRS will approve a verbal PCN to allow system startup to continue.

ANSWER: C KA # & KA VALUE: 2.2.6 Knowledge of the process for making changes to procedures. (3.0/3.6)

REFERENCE:

SWP-PRO-02 SOURCE: Bank (October 2009 NRC Exam)

LO: LO-12200 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. CRO2 cannot approve an editorial change.

B (incorrect): See C. CRO2 cannot approve a verbal PCN.

C (correct): The procedure has the operator close the discharge valve prior to starting the pump. Once started, the discharge valve should be opened to slowly fill the system. The example illustrates an obvious EPN error. Editorial changes are used for obvious EPN errors. Per SWP-PRO-02, the CRS would approve an editorial change.

D (incorrect): Although the change meets the definition of an editorial

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 change, a verbal PCN could be used to make the change. However, the CRS cannot approve a verbal PCN. Approval requires the CRS/SM and one other member of the Management/Supervisory Staff.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-70 Mechanical maintenance is required on a minimum flow valve downstream of a pumps discharge valve. The discharge valve must be danger tagged shut as an isolation boundary.

What are the tagging requirements for this situation?

A. A drain valve between the pump and the discharge valve should be tagged open to provide a flow path if the pump starts.

B. The source of power to the pump motor should be removed and danger tagged out to prevent damaging the pump.

C. If the pump has NO auto start features, the control switch for the pump should be danger tagged in the OFF position to prevent operation.

D. If the pump has an auto start feature, an Equipment Tag should be hung on the control switch to direct securing the pump following an auto start.

ANSWER: B KA # & KA VALUE: 2.2.13 Knowledge of tagging and clearance procedures. (4.1/4.3)

REFERENCE:

OI-12 Rev.033 pg.8 SOURCE: New LO: LO-6232 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B.

B (correct): OI-12 states, If work on specific equipment requires tagging in such a manner that associated equipment could be damaged if operated, then the associated equipment is to be isolated as part of the clearance.

C (incorrect): See B. To prevent pump operation, power to the pump will be isolated. The control switch will not be used as a boundary isolation.

Therefore, the control switch will have an Equipment Tag, not a Danger Tag, as part of the clearance order.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-71 The plant was operating at 100% power when a reactor scram occurred. RPV level is being maintained in automatic, and reactor pressure is being reduced using the Bypass Valves to satisfy the Shutdown Cooling interlocks.

In what Technical Specification Mode of Operation is the plant?

A. 2 B. 3 C. 4 D. 5 ANSWER: B KA # & KA VALUE: 2.2.35 Ability to determine Technical Specification Mode of Operation.

(3.6/4.5)

REFERENCE:

TS 1.1-8 SOURCE: New LO: LO-10297 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B.

B (correct): The stem states reactor pressure is being lowered to satisfy SDC interlocks. This implies reactor pressure is greater than 125 psig (135 depending on where the parameter is read). Temperature is therefore greater than 200°F. To establish automatic RPV level control, the level leg of PPM 5.1.1 must be entered. This requires the Mode Switch to be in shutdown.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-72 CRO2 is preparing a brief to lift a clearance order. The RO that signed for the clearance order hang added the following note:

OPS2 received 30 mrem in approximately 15 mins while hanging tags in the RHR Heat Exchanger Room.

Assuming plant conditions have not changed, what is the minimum level of Radiation Work Permit (RWP) that will be required?

A. Radiation Area RWP B. High Radiation RWP C. Locked Radiation RWP D. Very High Radiation Area RWP ANSWER: B KA # & KA VALUE: 2.3.7 Ability to comply with radiation work permit requirements during normal and abnormal conditions. (3.5/3.6)

REFERENCE:

PPM 11.2.7.1 Rev.038; GEN-RPP-02 Rev.030 SOURCE: New LO: LO-11261 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B.

B (correct): 30 mrem in 15 mins = 120 mrem/hr. If dose can exceed 100 mrem in one hour, the area is a High Radiation Area, and a High Rad RWP will be required at a minimum.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-73 When emergency venting the primary containment, the preferred source is:

A. The Wetwell to minimize cycling and the potential failure of the Wetwell-to-Drywell vacuum breakers.

B. The Wetwell to minimize the amount of radioactivity released by utilizing the scrubbing action of the Suppression Pool.

C. The Drywell to minimize the moisture introduction into and possible breakdown of the SGT charcoal adsorbers.

D. The Drywell to achieve a larger primary containment pressure reduction for a fixed vent duration period.

ANSWER: B KA # & KA VALUE: 2.3.11 Ability to control radiation releases. (3.8/4.3)

REFERENCE:

TM-2118 Rev.005; LO000078 slides 148-149 SOURCE: Bank LO: LO-8363 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): Venting through the Wetwell will reduce the cycling of vacuum breakers, but this is not the stated advantage of venting via the Wetwell.

B (correct): The Wetwell is the preferred path to reduce offsite radioactivity release rates.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-74 PPM 5.5.5, Overriding RCIC Low RPV Pressure Isolation Interlock, has been directed by the CRS.

Which of the following icons is next to PPM 5.5.5 on the EOP flowcharts?

A.

B.

C.

D.

ANSWER: B KA # & KA VALUE: 2.4.19 Knowledge of EOP layout, symbols, and icons. (3.4/4.1)

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013

REFERENCE:

PPM 5.0.10 Rev.017 pg.53 SOURCE: New LO: LO-8035 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): This icon means Venting may proceed exceeding offsite radioactivity release rate limits if necessary. This is not associated with PPM 5.5.5.

B (correct): PPM 5.5.5 requires defeating interlocks, as is expressed in the title of the procedure. This icon means Override/bypass of interlocks may be necessary as is permitted. The icon is beside PPM 5.5.5 in the following EOP flowcharts: PPM 5.1.1, PPM 5.1.2, PPM 5.1.4, and PPM 5.1.6.

C (incorrect): This icon means Oscillations of RPV water level and reactor power may occur and are permitted. This is not associated with PPM 5.5.5.

D (incorrect): This icon directs the operator to refer to the designated figure (in this case Figure C), and is not associated with PPM 5.5.5.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: RO-75 A Site Area Emergency has been declared, and CRO2 has been designated as the Emergency Communicator by the Shift Manager.

The Emergency Communicator will A. notify Benton and Franklin Counties, Washington State, and the DOE-RL of emergency classifications using the SCC ring down phone next to the CRS.

B. notify Benton and Franklin Counties, Washington State, and the DOE-RL of emergency classifications using the red ENS phone on CRO3s desk.

C. provide the NRC with event information via the NRC Emergency Notification System using the red ENS phone in the Shift Managers office.

D. provide the NRC with event information via the red CRASH system phone next to the Shift Technical Advisors desk in the Control Room.

ANSWER: C KA # & KA VALUE: 2.4.43 Knowledge of emergency communications systems and techniques.

(3.2/3.8)

REFERENCE:

PPM 13.4.1 Rev.041 pg.8 and 9; Form 26503 SOURCE: New LO: LO-8907 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The SCC ring down network is used to activate the Emergency Response Organization.

B (incorrect): The red ENS phone on CRO3s desk can be used to communicate with the NRC, but is not used to notify the agencies listed in this distractor.

C (correct): Per Form 26503, the red ENS phone in the Shift Managers office is the preferred method of providing the NRC information in an emergency.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 D (incorrect): The red CRASH phone next to the STAs desk is used for offsite agency notification of emergency declarations.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-76 With the plant operating at 100% power, emergent work results in a change to the number of high risk evolutions to be performed during the shift.

What tool will be used to evaluate the temporary and aggregate risk increases that result from the additional maintenance activities?

A. ORAM B. Paragon C. Sentinel D. INOP/LCO Log ANSWER: B KA # & KA VALUE: 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. (SRO 3.8) (43.5)

REFERENCE:

PPM 1.5.14 Rev.025 pg.4 SOURCE: New LO: SRO-0685, TREQ 12-0181 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): ORAM is a tool for Outage Risk Assessment.

B (correct): Per PPM 1.5.14, Paragon is the tool used at CGS to assess risk when a Change of Configuration occurs.

C (incorrect): Sentinel is a risk assessment tool, but is no longer used per CGS procedure.

D (incorrect): The INOP/LCO Log will be used to track inoperable equipment, but is not used to assess risk.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-77 Columbia was operating at 100% power with RWCU-P-1A in service. A fault in the motor operator for RWCU-V-4 (RWCU Suction Outboard Isolation Valve) caused the valve to stroke closed. After RWCU-V-4 was fully closed, the fuses in the motor operator control power circuit cleared.

RWCU-P-1A A. continued to run when RWCU-V-4 went closed, but will eventually trip on high motor cavity temperature. RWCU-V-4 is OPERABLE because it is closed and deactivated.

B. continued to run when RWCU-V-4 went closed, but will eventually trip on high motor cavity temperature. Declare RWCU-V-4 INOPERABLE per LCO 3.6.1.3, Primary Containment Isolation Valves.

C. tripped when RWCU-V-4 went to the intermediate position. RWCU-V-4 is OPERABLE because it is closed and deactivated.

D. tripped when RWCU-V-4 went to the intermediate position. Declare RWCU-V-4 INOPERABLE per LCO 3.6.1.3, Primary Containment Isolation Valves.

ANSWER: C KA # & KA VALUE: 204000 A2.10 Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures. (SRO 2.8)

(43.2)

REFERENCE:

SD000190 (RWCU) Rev.013 pg.9; LCO 3.6.1.3 pg.3.6.1.3-1; TS Bases pg.B 3.6.1.3-3; EWD-4E-015 SOURCE: New LO: LO-5037, LO-6925 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): While high motor cavity temperature is a RWCU pump trip, the running RWCU pump will trip when RWCU-V-4 is not fully open. The TS Bases for LCO 3.6.1.3 states, The normally closed PCIVs are considered

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 OPERABLE whenautomatic valves are de-activated and secured in their closed position RWCU-V-4 is not normally closed, and is therefore INOPERABLE for the conditions described in the stem.

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

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-78 The plant is in MODE 2 with control rod withdrawal in progress. Control Rod 18-43 is being withdrawn from position 22 to position 24. The reed switch contact associated with position 24 fails to close, and remains open after the control rod settles.

What will occur as a result of this fault, and what procedure will the CRS direct to correct or mitigate the effects of the failure?

A. The four rod display for Control Rod 18-43 will be two dashes (- -) after the rod has settled. Direct the performance of ABN-RPIS.

B. The four rod display for Control Rod 18-43 will be blank after the rod has settled.

Direct the performance of ABN-RWM.

C. An RPIS DATA FAULT and rod block for Control Rod 18-43 will exist. Direct the performance of ABN-RPIS.

D. RWM will block the movement of all control rods. Direct the performance of ABN-RWM.

ANSWER: C KA # & KA VALUE: 214000 A2.01 Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal indications or operations: Failed reed switches. (SRO 3.3) (43.5)

REFERENCE:

SD000148 Rev.013 pg.11 and 27; ABN-RPIS Rev.003 pg.3 and 6.

SOURCE: New LO: LO-5795, LO-5799, LO-7753, LO-7754, LO-6708 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): Two dashes indicate an odd reed switch is closed, not the failure of an even reed switch to close.

B (incorrect): The first half of this distractor is correct. The four rod display for Control Rod 18-43 will be blank due to the failure of the position 24 reed switch contact to close. Although RWM will generate a rod block because of

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 this condition, ABN-RWM does not contain the actions necessary to correct or mitigate the condition.

C (correct): An RPIS DATA FAULT will occur and a rod block will be generated for Control Rod 18-43. ABN-RPIS contains steps to insert a substitute position, clear the rob block, and move the control rod to a position with operable indication, and should be directed by the CRS.

D (incorrect): The first half of this distractor is correct. Although RWM will generate a rod block because of this condition, ABN-RWM does not contain the actions necessary to correct or mitigate the condition.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-79 Use your provided references to answer this question.

The plant is at 90% power and the crew is increasing power to 100%. The CRS then notices APRM B is indicating 98% power.

The CRS should:

A. Direct APRM B to be bypassed using a CSCO per PPM 1.3.1, and continue raising power.

B. Enter LCO 3.3.1.1 Condition A, and place the channel or the trip system in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. Enter LCO 3.3.1.1 Condition B, and place the channel or the trip system in trip within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. Enter LCO 3.3.1.1 Condition C, and restore RPS trip capability within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ANSWER: A KA # & KA VALUE: 212000 A2.06 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High reactor power. (SRO 4.2) (43.1)

REFERENCE:

PPM 1.3.1 Rev.108 pg.53; LCO 3.3.1.1-1 through 3 SOURCE: New LO: LO-5097 RATING: H3 ATTACHMENT: LCO 3.3.1.1-1 through 3 JUSTIFICATION: A (correct): APRM B is one of 3 channels in the RPS B trip system, with only 2 required channels. SR 3.3.1.1.2 requires APRM reading to be LE 2%

of calculated power. While APRM B is inoperable for failing to meet the SR, it is not a required channel. APRM B should be bypassed to prevent an inadvertent 1/2 scram, and the bypass switch should be controlled using a CSCO per PPM 1.3.1. The power increase may continue.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 B (incorrect): See A. The candidate must recall from the bases that only 2 channels are required per trip system, and therefore entry into this spec is not required.

C (incorrect): See A.

D (incorrect): See A.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-80 Columbia was operating at the End-of-Cycle with Final Feedwater Temperature Reduction in progress per PPM 3.1.11. Conditions were as follows:

• Reactor power: 91.0%

• RPV Inlet Temperature (RFW-TI-5): 357°F

• RFW-V-109 (FWH 6A/6B Bypass) open

• COND-V-144 (FWH 5A/5B Bypass) open An event occurred that resulted in the following indications:

• Reactor power: 92.5%

• RPV Inlet Temperature (RFW-TI-5): 344°F What caused the positive reactivity addition, and what procedure should the CRS direct to mitigate the event?

A. An electrical fault caused BS-V-12A (Bleed Steam Supply to FWH 4A) to close.

Direct ABN-CORE to prevent power oscillations.

B. An electrical fault caused BS-V-12A (Bleed Steam Supply to FWH 4A) to close.

Direct ABN-POWER to restore RPV Inlet Temperature to 355°F or greater.

C. A spurious high level trip of FWH 6A. Direct ABN-CORE to prevent power oscillations.

D. A spurious high level trip of FWH 6A. Direct ABN-FWH-HI/LEVEL to restore RPV Inlet Temperature to 355°F or greater.

ANSWER: B KA # & KA VALUE: 295014 AA2.03 Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION: Cause of reactivity addition. (SRO 4.3) (43.5, 43.6)

REFERENCE:

SD000163 (Bleed Steam) Rev.009 pg.15; PPM 3.1.11 Rev.008 pg.3 and 5; ABN-POWER Rev.013 pg.7, CGS Simulator; ABN-CORE Rev.014 pg.12; SER 19-92 SOURCE: New LO: LO-7615, LO-11575, LO-7607, SRO-0643, LO-6790

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 RATING: H4 ATTACHMENT: None JUSTIFICATION: A (incorrect): The first half of the distractor is correct. The closure of BS-V-12A will cause the reactivity addition and the conditions indicated in the stem of the question. Although power oscillations are more likely to occur with the last two stages of FWH bypassed, ABN-CORE does not contain the actions necessary to mitigate the effects of a FWH trip or a loss of bleed steam for these conditions. ABN-CORE would address the oscillations if they were due to a failure of FWLC, RRC Flow Control or DEH.

B (correct): ABN-POWER contains the actions to directly mitigate this transient, including control rod insertion and FWH recovery to restore RPV Inlet Temperature to 355°F or greater to comply with LCS 1.1.6.

C (incorrect): Normally, a trip of FWH 6A would cause a positive reactivity addition of at least the magnitude described in the stem. With FFTR in progress, FWH 6A is bypassed. If the heater tripped, the temperature reduction and power rise described in the stem would not occur.

D (incorrect): See C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-81 The dose limit (TEDE) the Emergency Director is authorized to approve to protect valuable property is ________.

A. 5 rem B. 10 rem C. 25 rem D. 5(N-18) rem, where N is the age of the individual ANSWER: B KA # & KA VALUE: 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (SRO 3.7) (43.4)

REFERENCE:

PPM 13.2.1 Rev.020 pg.019 SOURCE: New LO: LO-6020, LO-6129 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B.

B (correct): Per PPM 13.2.1 and EPA 400 PAGs for Emergency Workers.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-82 Use your provided references to answer this question.

Control rod withdrawal for plant startup was in progress. All SRMs were indicating 1 x 103 cps when an SRM MONITORS DOWNSCALE alarm was received, and SRM-A was reported to be indicating downscale. During investigation into the problem, the Mode Switch for SRM-C was inadvertently placed in Standby.

What action(s), if any, are required by Tech Specs, and what will the CRS direct?

A. No Tech Spec actions are required. Direct control rod withdrawal to proceed while the investigation continues.

B. No Tech Spec actions are required. Direct the Mode Switch for SRM-C be returned to Operate and SRM-A be bypassed to allow rod withdrawal to continue.

C. Restore SRM-A or SRM-C to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Direct the Mode Switch for SRM-C be returned to Operate to exit the required action statement.

D. Restore SRM-A and SRM-C to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Direct the Mode Switch for SRM-C be returned to Operate and SRM-A be bypassed to exit the required action statement.

ANSWER: C KA # & KA VALUE: 215004 2.2.44 Source Range Monitor - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(SRO 4.4) (43.2)

REFERENCE:

SD000132 (SRM) Rev.012 pg.19-25; LCO 3.3.1.2-1 and -6. SD000138 (IRM) Rev.010 pg.16 SOURCE: New LO: LO-5943, LO-12002, LO-5944 RATING: H3 ATTACHMENT: LCO 3.3.1.2-1 and -6 JUSTIFICATION: A (incorrect): See C. Tech Spec actions are required due to one (1) inoperable SRM. If the candidate does not correctly correlate SRM readings

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 to IRM power, they may assume IRMs are on range 3 and SRMs are not required per LCO 3.3.1.2.

B (incorrect): The second half of this distractor is correct, but Tech Spec action is required.

C (correct): With SRM-A downscale and SRM-C in Standby, two SRMs are inoperable. The conditions given in the stem (rod withdrawal in progress and SRMs at 1 x 103 cps) indicate the plant is in MODE 2 and IRMs are on range 1. According to Table 3.3.1.2-1, three (3) channels of SRMs are required to be operable in this condition. This means 1 required SRM is inoperable. Per LCO 3.3.1.2 Condition A, restore required SRMs to operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Restoring SRM-A or SRM-C will satisfy this required action. Placing the Mode Switch for SRM-C to Operate returns the SRM to operable, and allows the required action statement to be exited.

D (incorrect): See C. SRM-A or SRM-C must be restored to operable, not both SRMs. This information must be obtained from Table 3.3.1.2-1. These actions would be directed by the CRS, but bypassing SRM-A is not required to exit the action statement.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-83 The plant is operating at 50% power when the CRS receives a report that a required Channel Functional Test associated with the Primary Containment Pressure - High function was not performed. The test was last performed 120 days ago, and is required to be performed by the Technical Specifications every 92 days. A risk evaluation has been performed and a plan to manage the risk has been established.

Based on these conditions and per the Technical Specifications, the CRS is required to declare the associated instrumentation inoperable ______________.

A. immediately B. within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the time of discovering the missed surveillance C. when the grace period of 1.25 times the surveillance periodicity expires D. within 92 days of the time of discovering the missed surveillance ANSWER: D KA # & KA VALUE: 295024 2.4.21 High Drywell Pressure. Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (SRO 4.6) (43.5)

REFERENCE:

SR 3.3.1.1.8, SR 3.0.3, and B 3.0-16 SOURCE: New LO: LO-10303 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See D. This distractor will be selected if SR 3.0.3 is not applied.

B (incorrect): See D. SR 3.0.3 requires 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the specified frequency, whichever is greater. For the Drywell Pressure - High function, 92 days is allowed.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): See D. SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. For a 92 day frequency, the 25% extension would expire at 115 days.

D (correct): SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable when a Surveillance has not been completed within the specified frequency. A delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limits of the specified frequency, whichever is greater, applies from the point in time it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified frequency was not met.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-84 Use your provided references to answer this question.

CGS is operating at 100% power with Electrical Maintenance conducting non-intrusive breaker inspections. The Work Week Manager has just informed you that the main feeder breaker for E-PP-3CA was inadvertently opened, and cannot be closed.

Which of the following actions are required as a direct result of the loss of E-PP-3CA?

A. Monitor Offgas System Hydrogen concentration within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> until the operability of a Hydrogen Monitor is restored. If a Hydrogen Monitor is not restored within 30 days, place the unit in a condition that does not require Offgas System operation.

B. Direct Chemistry to conduct grab samples and analysis of Offgas System Hydrogen concentration within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until the operability of a Hydrogen Monitor is restored. If a Hydrogen Monitor is not restored within 30 days, initiate a Condition Report.

C. Monitor Offgas System Hydrogen concentration within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until the operability of a Hydrogen Monitor is restored. Direct the performance of ABN-OFFGAS to restore monitoring capability.

D. Direct Chemistry to conduct grab samples and analysis of Offgas System Hydrogen concentration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until the operability of a Hydrogen Monitor is restored. Direct the performance of ABN-ELEC-SM3/SM8 to restore the SL-31 distribution system.

ANSWER: B KA # & KA VALUE: 271000 2.2.36 Offgas. Ability to analyze the effects of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (SRO 4.2) (43.2)

REFERENCE:

SOP-ELEC-AC-LU Rev.032 pg.180; SD000187 (Offgas) Rev.011 pg.12; RFO 1.3.7.3-1; RFO Bases B1.3.7.3-2 and -3; ABN-OG Rev.003; ABN-ELEC-SM3/SM8 Rev.014.

SOURCE: New LO: LO-5622, LO-5612e RATING: H2

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 ATTACHMENT: SOP-ELEC-AC-LU pg.180; LCS 1.3.7.3-1 JUSTIFICATION: A (incorrect): E-PP-3CA provides power to both Offgas Hydrogen Analyzers. With both Analyzers out of service (1 required monitor inoperable), RFO 1.3.7.3 Condition A must be entered. RFO 1.0.3 does not apply, so exiting the condition of applicability is not required if operability is not restored.

B (correct): E-PP-3CA provides power to both Offgas Hydrogen Analyzers.

With both Analyzers out of service (1 required monitor inoperable), RFO 1.3.7.3 Condition A must be entered. This Condition requires Main Condenser Offgas Treatment System Hydrogen concentration to be monitored within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. The bases for RFO 1.3.7.3 states this will be accomplished by chemistry grab sample and analysis. Condition A also requires a monitor to be restored to operable within 30 days. If not restored, enter Condition B, which requires initiation of a Condition Report.

C (incorrect): The first half of this distractor is correct. See B. ABN-OFFGAS will not mitigate the loss of hydrogen monitoring, and should not be directed for this condition.

D (incorrect): The first half of this distractor is correct. See B. ABN-ELEC-SM3/SM8 contains the actions to recover from a loss of the SL-31 distribution system, but does not specifically address a loss of E-PP-3CA.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-85 The plant is operating at rated power with the Shift Manager out of the Control Room conducting a field observation. A notification was just received from the site security force that an armed attack has occurred in the Protected Area, and the Shift Manager has been taken hostage.

In this situation, the _________ will assume the role of Emergency Director until the EOF is activated.

A. STA / Incident Advisor B. Operations Manager C. Plant Manager D. Control Room Supervisor ANSWER: D KA # & KA VALUE: 2.4.38 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency director if required. (SRO 4.4)

(43.5)

REFERENCE:

PPM 13.10.1 Rev.034 pg.6 and 7 SOURCE: New LO: LO-6129 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The Incident Advisor will assist the CRS with Emergency classifications, but will not assume Emergency Director responsibilities.

B (incorrect): The Operations Manager may be the Emergency Director once the EOF is activated, but will not initially assume the role.

C (incorrect): The Plant Manager will not assume Emergency Director responsibilities.

D (correct): Per the Emergency Plan, the Control Room Supervisor will assume Shift Manager responsibilities, including emergency direction and

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 control authority, in the absence of the Shift Manager.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-86 Columbia is operating at 100% power with a calibration of SGT-SUM-1B2 (SGT Fan 1B2 Summer Circuit) in progress. Secondary Containment differential pressure is being maintained by Reactor Building HVAC with ROA-FN-1B and REA-FN-1B in service. Multiple unexpected alarms are received in the Control Room, and the following conditions are reported:

• ROA-FN-1B and REA-FN-1B are OFF

• ROA-V-1 and ROA-V-2 (RB Supply Valves) are OPEN

• REA-V-1 and REA-V-2 (RB Exhaust Valves) are OPEN

• The SEC PRESS P HIGH annunciator on P602 is in alarm

• Multiple Board R alarms have been received The CRS should direct the performance of ______________ to recover from the trip.

A. SOP-SGT-START-DIV/1-QC B. SOP-HVAC/RB-RESTART-QC C. SOP-SGT-START Section 5.2 D. ABN-HVAC, HVAC TROUBLE ANSWER: B NOTE: The answer key was changed to the correct answer being A. See ML13056A520.

KA # & KA VALUE: 295035 EA2.01 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

Secondary containment pressure. (SRO 3.9) (43.5)

REFERENCE:

Condition Report (AR-CR) 267373; Event Notification 48131; PPM 4.812.R2 Rev.019 pg.31 and 38; ABN-HVAC Rev.010 pg.5 and 15; PPM 4.602.A5 Rev.035 pg.24; SOP-SGT-START-DIV/1-QC Rev.000 pg.1-3; SOP-HVAC/RB-RESTART-QC Rev.000 pg.1-3 SOURCE: New LO: LO-6876 RATING: H3 ATTACHMENT: None

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 JUSTIFICATION: A (incorrect): SOP-SGT-START-DIV/1-QC would be used to start SGT-A and restore Secondary Containment pressure if RB HVAC was not available.

With the supply and exhaust valves open, RB HVAC should be restarted to recover from the trip.

B (correct): With the RB HVAC supply and exhaust valves open, the CRS will direct a restart of RB HVAC using SOP-HVAC/RB-RESTART-QC to recover Secondary Containment P.

C (incorrect): See A and B. SOP-SGT-START Section 5.2 is used to manually initiate SGT for normal evolutions.

D (incorrect): See A and B. Entry conditions for ABN-HVAC exists, making this a plausible procedure choice.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-87 Which of the following instruments is used to estimate core damage when the Severe Accident Guidelines have been entered?

A. Containment Radiation Monitors (CMS-RE-27A,B,E,F)

B. Containment Particulate Monitors (CMS-RE-12/1A,B)

C. Containment Noble Gas Detectors (CMS-RE-12/3A,B)

D. Reactor Building Stack Radiation Monitors (PRM-RE-1A,B,C)

ANSWER: A KA # & KA VALUE: 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (SRO 2.9) (43.4)

REFERENCE:

TM-2177 Rev.006 pg.001 and Attachment 4.2.1 SOURCE: New LO: LO-9644 RATING: L4 ATTACHMENT: None JUSTIFICATION: A (correct): Per the reference, 3 methods of estimating core damage are available, including use of the containment radiation monitors.

B (incorrect): Used to detect leakage in containment, not estimate core damage.

C (incorrect): Used to detect leakage in containment, not estimate core damage.

D (incorrect): Used to determine offsite release rates, not estimate core damage.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-88 The reactor is being started up with the following as current conditions:

• Reactor Power is 12%

• Reactor Pressure is 960 psig and stable

• The Main Turbine Bypass Valves are intermediate

• There is confirmed leakage from one safety relief valve.

The vendor representative recommends cycling the leaking SRV in an attempt to re-seat it. This evolution would be performed per __________.

A. ABN-SRV, Safety Relief Valve Opening B. PPM 4.601.A2 6-7, Relief Valve Tailpipe Temperature High C. TSP-MSRV/IST-R701, SRV Operability Surveillance D. OI-09, Operations Standards and Expectations, as a Simple Quick Act ANSWER: C KA # & KA VALUE: 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (SRO 4.7) (43.5)

REFERENCE:

TSP-MSRV/IST-R701 Rev.006; PPM 4.601.A2 6-7; ABN-SRV Rev.004; OI-09 Rev.056 pg.61, 65, and 70.

SOURCE: New LO: SRO-0149 RATING: H4 ATTACHMENT: None JUSTIFICATION: A (incorrect): See C. Although an Abnormal Procedure may contain the desired actions, the procedure cannot be used unless the existing conditions apply. This ABN is designed to mitigate a failed open SRV, not to cycle a leaking SRV, and will not be used for this activity.

B (incorrect): ARPs are used to mitigate the effects of a condition and/or reference procedures to recover from the condition, and do not establish

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 required plant conditions to conduct evolutions. This ARP will direct Suppression Pool Cooling and initiation of a Condition Report, but will not be used to cycle the SRV.

C (correct): Per OI-09 Section 32.2.2.b, Operator Fundamentals - Control, equipment must be operated using detailed operating guidance when changing reactivity, operational modes, and system alignments. The CRS must provide direction for implementation of normal and operating procedures. Operation of equipment per an approved Tech Spec Surveillance ensures the plant conditions will support the operation, and configuration control will be maintained.

D (incorrect): Simple Quick Acts are defined as actions that may be performed by trained, qualified individuals without a procedure provided that no procedure exists for the action and the task is simple, short, and routine.

An operability surveillance (procedure) does exist for this activity, and it should not be considered routine.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-89 Columbia was operating at 100% power when the output breaker for E-C1-1A, the in service 125VDC Division 1 battery charger, tripped open. The CRS declared E-C1-1A inoperable and entered LCO 3.8.4, DC Sources - Operating, Condition A.

Which of the following is correct for this event?

A. LCO 3.8.4 Condition A should NOT have been entered, because E-C1-1B remained operable and satisfied the Tech Spec requirement for operability.

B. Place E-C1-1B in service, verify E-B1-1 terminal voltage is GE 126VDC within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and verify float current is LE 2 amps within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. When these actions are complete, E-C1-1A is NOT required to be operable.

C. Place E-C1-1B in service, verify E-B1-1 terminal voltage is GE 126VDC within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and verify float current is LE 2 amps within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. E-C1-1A must be restored to operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to exit LCO 3.8.4 Condition A.

D. A balance of plant non-Class 1E battery charger may be used to restore E-B1-1 terminal voltage to GE 126 VDC within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce float current to LE 2 amps with 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. E-C1-1A must be restored to operable and placed in service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

ANSWER: B

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 KA # & KA VALUE: 263000 2.2.25 D.C. Electrical Distribution. Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (SRO 4.2) (43.2)

REFERENCE:

LCO 3.8.4-1; Tech Spec Bases B3.8.4-1 through -7 (LCO on page B3.8.4-4)

SOURCE: New LO: LO-5265 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B. Because E-C1-1A was in service when the output breaker opened and E-C1-1B is in standby and electrically isolated, LCO 3.8.4 Condition A should have been entered.

B (correct): Per LCO 3.8.4, The DC electrical power subsystems, each subsystem consisting of one battery, one battery charger, and the corresponding control equipmentare required to be OPERABLE.

Condition A has three requirements: A.1 requires the battery terminal voltage to be restored to greater than or equal to the minimum established float voltage within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The bases for LCO 3.8.4 defines this as 126VDC for the 125VDC batteries. A.2 requires battery float current to be less than or equal to 2 amps within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. A.3 requires a battery charger to be restored to operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The bases for LCO 3.8.4 states that one charger is normally in service supplying system loads. Placing E-C1-1B in service will meet this description. The bases also states that one charger is required for operability, not both chargers. E-C1-1A is therefore not required to be operable when E-C1-1B is in service and operable.

C (incorrect): See A and B.

D (incorrect): See A and B. The bases for LCO 3.8.4 states that connecting a balance of plant non-Class 1E charger to the battery is permissible making this a plausible distractor. E-C1-1A is NOT required to be restored to operable and placed in service. Placing E-C1-1B in service while E-C1-1A remains inoperable will satisfy LCO 3.8.4.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-90 A plant startup is in progress with the following conditions:

• BPA BKR 4885 is open

• BPA BKR 4888 is open

• The Main Turbine is operating at 1800 RPM A fault at the Ashe Substation causes a lockout on the 230KV system (86TS-1 lockout).

One (1) minute after the fault, what will be the status of SM-7 and SM-8, and what procedures should the CRS direct?

A. E-TR-B is powering SM-7 and SM-8. Enter PPM 5.1.1, RPV Control, and direct ABN-ELEC-GRID.

B. E-TR-B is powering SM-7 and SM-8. Enter ABN-TURBINE, and direct PPM 3.3.1, Reactor Scram.

C. DG-1 is powering SM-7 and DG-2 is powering SM-8. Enter PPM 5.1.1, RPV Control, and direct PPM 3.3.1, Reactor Scram.

D. DG-1 is powering SM-7 and DG-2 is powering SM-8. Enter ABN-TURBINE, and direct ABN-ELEC-GRID.

ANSWER: A KA # & KA VALUE: 262001 A2.03 Ability to (a) predict the impacts of the following on the A.C.

ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations: Loss of off-site power. (SRO 4.3) (43.5)

REFERENCE:

SD000182 Rev.017 pg.43, 44, 76, and Figure 9; ABN-ELEC-GRID Rev.004; PPM 3.3.1 Rev.057; PPM 5.1.1 Rev.019 SOURCE: New LO: LO-11824, LO-11823, SRO-0666 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): With both BPA Breakers 4885 and 4888 open, plant loads are

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 supplied by the Startup Transformer via the Ashe Substation. A loss of the Ashe off-site power source results in an automatic transfer of safety related busses SM-7 and SM-8 to the Backup Transformer. DG-1 and DG-2 automatically start, but do not provide power. The interruption in power to SM-7 and SM-8 causes a loss of power to RPS and a full scram. The loss of Startup power also results in a loss of feed and condensate and low RPV water level. PPM 5.1.1 will be directed to mitigate the loss of feed and condensate. ABN-ELEC-GRID will be prioritized to restore off-site power.

B (incorrect): See A. Although a reactor scram will occur and an entry condition for PPM 3.3.1 exists, PPM 5.1.1 will be directed to mitigate the consequences of the loss of off-site power and subsequent scram. The Main Turbine will not immediately trip as a result of the loss of off-site power.

C (incorrect): DG-1 and DG-2 will start as a result of the undervoltage condition on SM-7 and SM-8 caused by the loss of off-site power. The Backup Transformer will supply power to SM-7 and SM-8.

D (incorrect): See C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-91 When moving irradiated fuel in the RPV, a minimum water level of 22 feet above the reactor vessel flange is required. The TS bases for this minimum water level is to:

A. Minimize refueling personnel dose rates.

B. Retain iodine fission products in the water in the event of a refueling accident.

C. Maintain sufficient water inventory for core cooling while the reactor head is off.

D. Ensure the fuel assemblies remain at least 10 feet below the pool surface during movement.

ANSWER: B KA # & KA VALUE: 295023 AA2.04 Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS: Occurrence of fuel handling accident.

(SRO 4.1) (43.2, 43.5)

REFERENCE:

TS Bases 3.9.6-1 SOURCE: New LO: LO-6925 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): See B.

B (correct): LCO 3.9.6 Bases states, Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident.

C (incorrect): See B.

D (incorrect): See B.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-92 The plant was operating at 90% power when a failure caused all inboard MSIVs to close. Four (4) minutes later, the operating crew has performed all required immediate actions and conditions are as follows:

• RPV water level is -37 and down fast

• RPV pressure is slowly rising

• MS-RV-1B and MS-RV-1C opened automatically and remain open Considering these conditions, which of the following is correct concerning RPV pressure, and what will the CRS direct to mitigate the transient?

RPV pressure has A. NOT exceeded the Tech Spec Safety Limit. Direct PPM 5.1.1, RPV Control.

B. exceeded the Tech Spec LCO pressure, but has NOT exceeded the EOP entry pressure. Direct PPM 5.1.2, RPV Control - ATWS.

C. exceeded the Tech Spec Safety Limit. Direct PPM 5.1.1, RPV Control, and insert all insertable control rods within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. exceeded the EOP entry pressure. Direct PPM 5.1.2, RPV Control - ATWS.

ANSWER: D KA # & KA VALUE: 295037 EA2.06 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor pressure. (SRO 4.2) (43.5)

REFERENCE:

SD000128 (Main Steam) Rev.010 pg.8; SD000161 (RPS) Rev.015 pg.13; PPM 5.1.1 Rev.019; PPM 5.1.2 Rev.021; Tech Spec Bases B2.1.2-1; LCO 3.4.12-1 SOURCE: New LO: LO-5527 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The first part of this distractor is correct. Per the Tech Spec

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 Bases for the RCS pressure SL, the SRV setpoints are established to ensure the SL will not be exceeded. None of the conditions in the stem indicate a failure of the SRVs to operate as designed. Although PPM 5.1.1 will be entered based on low RPV water level and high RPV pressure, it will not be used to mitigate these conditions as the question asked. This is because of the presence of an ATWS.

B (incorrect): See D. The Tech Spec LCO pressure is 1035 psig, and this value has been exceeded. The candidate must recognize that the EOP entry pressure (1060 psig) has also been exceeded. The second part of this distractor is correct.

C (incorrect): See A and D. The second part of the distractor contains the Tech Spec required actions for the violation of a SL.

D (correct): The stem states MS-RV-1B and 1C opened automatically. The relief pressure setpoint for these valves is 1091 psig. The EOP entry pressure is 1060 psig. RPS generated a scram when the MSIVs closed and when RPV pressure exceeded 1060 psig. The stem also states that pressure is slowly rising with two SRVs open four minutes after the MSIVs closed. This indicates that the reactor did not shutdown after a scram signal was initiated. The stem states that the immediate actions have been completed, which means manual attempts to insert a scram were not successful. Based on this information, an ATWS exists. PPM 5.1.1 will be entered, but will be exited to PPM 5.1.2 per override RC-2. PPM 5.1.2 contains the actions required to mitigate the ATWS and lower RPV water level.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-93 Columbia is operating at 90% power. A malfunction in the DEH system causes the Main Turbine Governor Valves to slowly close.

Which of the following correctly identifies how reactor power will respond to this failure, and what procedure will be directed to mitigate the transient?

Reactor power will A. rise. Direct ABN-POWER, Unplanned Reactor Power Change.

B. rise. Direct ABN-PRESSURE, Unplanned Reactor Pressure Change.

C. lower. Direct ABN-POWER, Unplanned Reactor Power Change.

D. lower. Direct ABN-PRESSURE, Unplanned Reactor Pressure Change.

ANSWER: B KA # & KA VALUE: 295025 EA2.02 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor power. (SRO 4.2) (43.5)

REFERENCE:

SD000146 (DEH) Rev.010 pg.8; ABN-POWER Rev.013; ABN-PRESSURE Rev.009 SOURCE: Bank LO: LO-11666 RATING: H2 ATTACHMENT: None JUSTIFICATION: A (incorrect): The first part of this distractor is correct. See B. An entry condition for ABN-POWER exists, but ABN-PRESSURE will be used to mitigate this event.

B (correct): When the Main Turbine Governor Valves begin to close, reactor pressure will rise. The Main Turbine Bypass Valves are biased closed until reactor pressure rises enough to overcome the bias. As reactor pressure rises, reactor power will also rise. At 100% power, ABN-POWER would be directed simultaneously with ABN-PRESSURE to maintain reactor power below the license level. Because this transient occurs at 90% reactor power,

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 ABN-PRESSURE will be directed to mitigate the transient through manual operation of the Main Turbine Bypass Valves.

C (incorrect): As reactor pressure rises moderator temperature will rise, which adds negative reactivity. This would cause power to lower by itself.

But at 90% power the void coefficient is much larger, and the positive reactivity added from the collapse of voids will cause power to rise. See B.

D (incorrect): See B and C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-94 A long term Temporary Modification (TM) is being installed this shift using an approved procedure and will be identified using a Caution Clearance Order.

This TM installation A. must be recorded in the TM Log Index in the Control Room per PPM 1.3.9, Temporary Modifications.

B. must be recorded in the electronic Control Room Log per PPM 1.3.9, Temporary Modifications.

C. is NOT required to be recorded because an approved procedure is being used for configuration control per PPM 1.3.1, Conduct of Operations.

D. is NOT required to be recorded because the Clearance Order process is being used for configuration control per PPM 1.3.64, Plant Clearance Order.

ANSWER: A KA # & KA VALUE: 2.2.11 Knowledge of the process for controlling temporary design changes.

(SRO 3.3) (43.3)

REFERENCE:

PPM 1.3.9 Rev.049 pg.23 and 24 SOURCE: New LO: LO-6287, SRO-0468 RATING: L3 ATTACHMENT: None JUSTIFICATION: A (correct): Per PPM 1.3.9, the CRS/SM must record the TM information in the TM Log Index when it is issued/authorized. PPM 1.3.9 also requires a TM Log Index to be maintained in the Control Room.

B (incorrect): Although it would not be incorrect to record the installation of the TM in the electronic Control Room Log per PPM 3.1.10, this distractor involves only the requirements of the Temporary Modification process. PPM 1.3.9 does not require an electronic Control Room Log entry for the installation of a TM.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 C (incorrect): Configuration control is typically maintained using approved plant procedures or the Clearance Order process, making these plausible distractors. PPM 1.3.9 requires tracking using the TM Log Index even if approved procedures and clearance orders are being used.

D (incorrect): See C.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-95 Columbia was operating in MODE 1 when events occurred that required the control room crew to evacuate the Main Control Room. All immediate actions were completed prior to leaving. The following plant conditions now exist:

• Reactor power is 8 percent and down slow

• RPV level is -100 inches and down slow Which of the following procedures should be directed to mitigate the event?

A. PPM 5.1.1, RPV Control B. PPM 5.1.2 , RPV Control ATWS C. ABN-LEVEL, Unplanned RPV Water Level Change D. ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown ANSWER: D KA # & KA VALUE: 295031 2.4.16 Reactor Low Water Level. Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. (SRO 4.4) (43.5)

REFERENCE:

ABN-CR-EVAC Rev.025 pg.7; ABN-LEVEL Rev.006; PPM 5.1.1; PPM 5.1.2; PPM 1.3.1 Rev.107 pg.33 SOURCE: Bank LO: SRO-0251 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): An entry condition for PPM 5.1.1 exists based on RPV level less than +13. See D.

B (incorrect): Reactor power indicates an ATWS condition exists. PPM 5.1.2 contains actions to mitigate an ATWS. See D.

C (incorrect): An entry condition exists for ABN-LEVEL based on an

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 unplanned RPV water level change. See D.

D (correct): With a low RPV water level, entry conditions exist for Volume 4 (Abnormal Operating Procedures) and Volume 5 (Emergency Operating Procedures) procedures. PPM 1.3.1, Conduct of Operations, specifies that EOPs will direct the performance of Volume 4 procedures. The exception to this is ABN-CR-EVAC. Per the note in the subsequent actions, ABN-CR-EVAC supersedes EOP procedures and PPM 3.3.1

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-96 The reactor is at rated power, and wetwell level has been increasing due to a large unidentified leak. The CRS has entered PPM 5.2.1, Primary Containment Control, and has determined wetwell level cannot be maintained below the SRV Tailpipe Level Limit. All other primary containment parameters are normal.

Based on this determination, the CRS should now:

A. Enter PPM 5.1.1, RPV Control.

B. Enter PPM 5.1.3, Emergency RPV Depressurization.

C. Shutdown the reactor per PPM 3.2.1, Normal Plant Shutdown.

D. Reduce reactor power and pressure per PPM 3.2.4, Fast Power Reduction.

ANSWER: A KA # & KA VALUE: 295029 2.4.6 High Suppression Pool Water Level. Knowledge of EOP mitigation strategies. (SRO 4.7) (43.5)

REFERENCE:

PPM 5.2.1 SOURCE: New LO: LO-8383 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): Per PPM 5.2.1, when wetwell level cannot be maintained below the SRVTPLL, entry into PPM 5.1.1 is required to place the MODE Switch into SHUTDOWN and scram the reactor.

B (incorrect): See A. If wetwell level and RPV pressure cannot be maintained below SRVTPLL, ED using PPM 5.1.3 is required. This will not occur until after the reactor is scrammed and attempts to redure RPV pressure are made.

C (incorrect): A scram, not a normal shutdown, is required.

D (incorrect): A scram is required at this point.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-97 Columbia was operating at 100% power when seismic activity resulted in the following conditions:

• The 86XU and 86XUOA relays have actuated

• The Startup Transformer is locked out

• The Backup Transformer is locked out

• DG-2 is locked out

• SM-7 is locked out These conditions have existed for 16 minutes. The Emergency Director must declare a(n)

A. Unusual Event per 6.1.U.1.

B. Alert per 6.1.A.1.

C. Alert per 6.1.A.2.

D. Site Area Emergency per 6.1.S.1.

ANSWER: D KA # & KA VALUE: 295003 AA2.05 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Whether a

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 partial or complete loss of A.C. power has occurred. (SRO 4.2) (43.5)

REFERENCE:

SD000182 (AC); PPM 13.1.1; PPM 13.1.1A SOURCE: New LO: LO-6131 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The conditions for 6.1.U.1 exist, but the highest emergency declaration is required to be made.

B (incorrect): See D. A complete loss of all AC power to SM-7 and SM-8 has existed for GT 15 minutes. The stem also states the plant was at 100%

power when the event occurred, which means the plant is now in Mode 3.

6.1.A.1 is for Mode 4/5.

C (incorrect): If the candidate must recognize power has been lost to SM-8 to recognize 6.1.A.2 is not applicable.

D (correct): The actuation of 86XU and 86XUOA indicate the Main Generator is tripped and TR-N1 is not available. The stem states a lockout exists on SM-7. A lockout on TR-S, TR-B and DG-2 means SM-8 has no power. In this condition with the plant in Mode 3, 6.1.S.1 must be declared.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-98 Given the following:

• All offsite power has been lost

• SW-P-1A tripped and cannot be restarted

• DG-1 was tripped from the MCR

• RHR-P-2B has a broken shaft

• RPV water level is -40 and being maintained with HPCS-P-1

• Drywell temperature is 295°F and stable

• Drywell pressure is 7.0 psig and stable Which of the following is correct for these conditions?

Initiating drywell sprays A. is safe. Direct Service Water aligned to RHR-B, and drywell sprays to be initiated per PPM 5.5.2, RHR/SW Crosstie Lineup.

B. is safe. Direct SM-7 to be restored per ABN-ELEC-SM1/SM7, and drywell sprays to be initiated per SOP-RHR-CONT-SPRAY.

C. could cause the primary containment to fail. Direct an emergency depressurization per PPM 5.1.3, Emergency RPV Depressurization.

D. could cause the primary containment to become de-inerted. Direct an emergency depressurization per PPM 5.1.3, Emergency RPV Depressurization.

ANSWER: A

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 KA # & KA VALUE: 295028 2.1.25 High Drywell Temperature. Ability to interpret reference materials, such as graphs, curves, tables, etc. (SRO 4.2) (43.5)

REFERENCE:

PPM 5.2.1, PPM 5.0.10, PPM 5.1.3, ABN-ELEC-SM1/SM7 SOURCE: New LO: LO-8433, LO-8314 RATING: H3 ATTACHMENT: None JUSTIFICATION: A (correct): By plotting a drywell pressure of 7.0 psig and a drywell temperature of 295°F on the DSIL curve, the candidate should determine that initiation is safe (not in the prohibited region). With a lock out on SM-7, RHR-P-2A is not available for containment sprays. RHR-P-2B is stated to have a broken shaft. PPM 5.2.1 allows the use of ESP PPM 5.5.2, RHR-SW Crosstie to spray containment with Service Water through RHR-B.

B (incorrect): See A. SM-7 power could be restored per ABN-ELEC-SM1/SM7 using DG-1. The stem states that SW-P-1A, the cooling water supply for DG-1, cannot be started. Running DG-1 without cooling should not be directed while adequate core cooling is assured and other systems are available foe containment control.

C (incorrect): See A. This could occur if sprays were initiated in the prohibited region of the DSIL curve. Emergency Depressurization would be required if drywell temperature could not be restored and maintained below 330°F.

D (incorrect): See A. This could occur if sprays were initiated in the prohibited region of the DSIL curve. Emergency Depressurization would be required if drywell temperature could not be restored and maintained below 330°F.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-99 Following the new fuel receipt inspection, new fuel bundles are normally stored in ____________.

A. the Fuel Prep Machine B. the Spent Fuel Pool C. a Hi-Storm D. New Fuel Inner Containers ANSWER: B KA # & KA VALUE: 2.1.42 Knowledge of new and spent fuel movement procedures. (SRO 3.4)

(43.7)

REFERENCE:

PPM 6.2.3 Rev.036 pg.10; SD000207 (Fuel Handling) Rev.012 pg.32 SOURCE: New LO: LO-5361a RATING: L3 ATTACHMENT: None JUSTIFICATION: A (incorrect): The Fuel Prep machine is used to transfer new fuel from the inspection stand to the Spent Fuel Pool, but is not a storage location.

B (correct): The Spent Fuel Pool is the normal storage location for new fuel following the new fuel receipt inspection.

C (incorrect): Hi-Storms are used for spent fuel storage.

D (incorrect): New fuel is stored in the New Fuel Inner Containers prior to inspection.

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 QUESTION: SRO-100 A hydraulic ATWS occurred. The CRS directed the SLC System to be initiated for boron injection.

Current conditions are as follows:

• SLC System A and SLC System B control switches are in OPER

• SLC-V-4A and SLC-V-4B (Squib Valves) CIRCUIT READY lights are illuminated

• SLC DIV 1 OUT OF SERVICE and SLC DIV 2 OUT OF SERVICE annunciators are NOT in alarm Based only on these indications, the SLC System is A. injecting boron into the RPV at 82 gpm. Continue injecting boron until the existing rod pattern alone can ensure the reactor is shutdown.

B. injecting boron into the RPV at 82 gpm. Continue injecting boron until the CSBW has been injected, or SLC-TK-1 level reaches 200 gallons.

C. NOT injecting boron into the RPV. Direct PPM 5.5.8, Alternate Boron Injection Using RCIC.

D. NOT injecting boron into the RPV. Direct PPM 5.5.25, Alternate Injection Using the SLC System.

ANSWER: C KA # & KA VALUE: 211000 A2.02 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failure of explosive valve to fire. (SRO 3.9) (43.5)

REFERENCE:

SD000072 (SLC) Rev.012; PPM 5.1.2 Rev.20; PPM 5.5.8 Rev.013; PPM 5.5.25 Rev.006; PPM 4.603.A7 Rev.045 pg.65; SOP-SLC-INJECTION-QC Rev.003 pg.4; EWD-10E-006 and 009; M522 SOURCE: New LO: LO-7636, LO-5929, LO-5925, LO-5922, SRO-0552 RATING: H3 ATTACHMENT: None

COLUMBIA GENERATING STATION WRITTEN EXAMINATION EXAM KEY FEBRUARY 2013 JUSTIFICATION: A (incorrect): With the SLC System control switches in OPER, the explosive valves should have fired. The CIRCUIT READY lights illuminated indicates the valves have continuity and have not fired. Additionally, when continuity is lost, the SLC DIV 1/2 OUT OF SERVICE annunciators will alarm. With neither explosive valve fired, there is no flow path for boron to be injected to the RPV through the SLC System. If SLC were injecting, injection would be terminated when the existing rod pattern alone ensured the reactor was shutdown.

B (incorrect): See A.

C (correct): See A. With no flow path through the SLC System available, boron must be injected by an alternate means. Per PPM 5.1.2, the CRS will direct PPM 5.5.8 to inject boron using the RCIC system.

D (incorrect): The first part of this distractor is correct. See A. PPM 5.5.25 would be selected if the indications were that the explosive valves had fired, but the SLC pumps were not operating.