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Generation Station - 11-2006 - Final As-Given Written Exam, Exam Key
	ML063540193
Person / Time
Site:	Columbia
Issue date:	11/30/2006
From:	Nease R; Operations Branch IV
To:	Parrish J; Energy Northwest
References
	50-397/06-301 50-397/06-301
	Download: ML063540193 (107)
	v • d • e

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 1

EXAM KEY NOVEMBER 2006 Page 1 of 81 Columbia is at 70% power when a jet pump fails due to a loss of the nozzle (rams head) on Reactor Recirculation Loop A.

Based on this failure, the reactor operator would expect:

A. Reactor recirculation total flow input to APRM channels A, C, and E to decrease.

B. Indicated core flow to decrease.

C. Indicated flow for Recirculation Loop A to increase.

D. The failed jet pump's differential pr essure indication to be more noisy.

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295001 AA2.06 Ab ility to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Nuclear Boiler Instrumentation (3.2)

REFERENCE:

CGS Systems Description, "Reactor Recirculation System", Rev. 13, pg. 21-22; SD000178 SOURCE: New

LO: 5023 a. Predict the impacts of the RRC system of each of the following: Jet Pump Failure RATING: Knowledge: Analysi s Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - Loop flow will increase. B. Incorrect - Indicated flow will increase.

C. Correct - Flow will increase due to less resistance from the nozzle.

D. Incorrect - Jet pump d/p indication will be less noisy on the failed jet pump. COMMENTS: Ref. 10CFR55.41(5) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 2

EXAM KEY NOVEMBER 2006 Page 2 of 81 Columbia is in MODE 5 and the Benton substation 115 kV feeder is unavailable due to maintenance.

If bus SM-1 trips on undervoltage, what is the ex pected sequence of events if no operator action is taken? A. Diesel Generator 1 automatica lly starts and reenergizes SM-7.

B. Diesel Generator 1 aut omatically starts and SM-7 remains deenergized..

C. Feeder breaker 7-1 opens and breaker B-7 closes powering SM-7 from TR-B.

D. Feeder breaker 7-1 opens and the bus remains deenergized.

ANSWER: A QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295003 AA1.03 Ab ility to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER: Systems necessary to assure safe plant shutdown (4.4)

REFERENCE:

CGS System Description, AC Distribution, Rev. 12, pg. 30; SD000182

SOURCE: New

LO: 5051d. Explain or identify the system interlock or response: Identify SM7 response to primary and secondary undervoltage.

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A Correct - On a primary under voltage condition, the EDG will automatically start and the output breaker will close when the generator is up to rated speed and voltage.

B Incorrect - DG-2 cannot energize SM-7.

C Incorrect - Transformer TR-B is deenergized as given in the stem.

D Incorrect - The lockout relay does not trip.

COMMENTS: 10CFR55.41(7)

Changed distractor B 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 3

EXAM KEY NOVEMBER 2006 Page 3 of 81 Given the following:

An Extended Station Blackout is in progress RCIC is currently injecting to the RPV

If all DC power will soon be lost, powering which of the following DC buses from DG-4, the Alternate AC Station Battery Charger, will pr event a RCIC trip and allow CONTINUED RCIC injection?

A. S1-1 B. S1-2 C. S2-1 D. S1-7 ANSWER: A QUESTION TYPE: Closed Reference KA # & KA VALUE: 295004 AA2.04 Ab ility to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: System lineups (3.2)

REFERENCE:

CGS System Description, DC Distr ibution, rev. 7, pg. 23 - 30; SD000188 CGS System Description, AC Distri bution, rev. 13, pg. 29; SD000182

SOURCE: New Question

LO:

RATING: Knowledge: Fundamental Difficulty: 4

ATTACHMENT: None

JUSTIFICATION: A. Correct - Bus S1-1 by it self will allow RCIC to continue to function, without this power, RCIC will trip. B. Incorrect - There is minimal effect on RCIC from loss of this bus.

C. Incorrect - RCIC continues to run without indication on most valves.

D. Incorrect - This bus does not supply RCIC loads. COMMENTS: Ref: 10 CFR 55.41 (7) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 4

EXAM KEY NOVEMBER 2006 Page 4 of 81 The purpose of the Feedwater Level Cont rol System Setpoint Setdown is to:

A. Maintain reactor water level below level 8 following a SCRAM.

B. Lower the reactor water level setpoint when there is only one feedpump running.

C. Ensure the feedpump turbines do not overspeed following a SCRAM.

D. Lower the reactor water level setpoi nt when in single element control.

ANSWER: A QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295006 AK3.04 Knowledge of the reasons for the following responses as they apply to SCRAM: Reactor water level set point Setdown: Plant specific (3.1)

REFERENCE:

CGS System Description, Feedw ater Level Control System, Rev. 13,Section V. A. ; SD000157

SOURCE: New Question

LO: 5397 State the purpose of Setpoint Se tdown, when it initiates, how it is reset, and how it affects the FWLC system.

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Correct - The signal happens upon a scram, and it is to prevent reaching level 8. B. Incorrect - The signal only occurs on a scram.

C. Incorrect - The feedpumps will slow down automatically following a scram due to less feedwater demand. D. Incorrect - Setpoint set down has no function associated with being in single element control. COMMENTS: 10 CFR 55.41 (7) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 5

EXAM KEY NOVEMBER 2006 Page 5 of 81 ABN-RCC is being performed due to a complete loss of RCC.

The bases for this procedure requiring the operati ng crew to place all the RCC pump switches in the pull to lock position is to:

A. minimize the potential for damaging the pumps and/or motors.

B. minimize break flow in the event of a piping failure.

C. allow the system to be returned to service in a controlled manner.

D. maintain RCC inventory unt il the system is restored.

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295018 / 2.1.28 Knowledge of the purpose and function of major components and controls.

2.1.20 Ability to execute proc edure steps (4.3) [Deleted]

REFERENCE:

ABN-RCC, Rev. 3, bases for step 4.1.4 SOURCE: New

LO:

RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - see C.

B. Incorrect - see C.

C. Correct- The bases for placing the pumps in PTL is that in the event of a

loss of power to busses SL-71 and SL

-81 the pump breakers remain closed due to there not being an undervoltage breaker trip. Placing the switches in PTL ensures an orderly return to service.

D. Incorrect - see C.

COMMENTS: Ref: 10 CFR 50.41 (4) & (10) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 6

EXAM KEY NOVEMBER 2006 Page 6 of 81 Which of the following will fail in the closed direction following a complete loss of Control Air System pressure?

A. CRD flow control valves (CRD-FCV-2A/2B).

B. Intertie between the Containment Nit rogen Inerting System and the Containment Instrument Air System (CN-V-65).

C. Inboard Main Steam Isolation Valves (MS-V-22A/22B/22C/22D).

D. Reactor Closed Cooling Water heat exchanger discharge valves (RCC-V-2A/2B/2C).

ANSWER: A QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295019 AA2.02 (AK2.03 RFW) Ab ility to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safety related instru ment air system loads (see AK2.1- AK2-19)

(3.6)

REFERENCE:

CGS System Description, "Control Rod Hydraulic System", Rev. 12,Section IV. K; SD000142 SOURCE: New Question

LO: 7605. Describe the effect of a CAS failure on system loads.

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Correct - The CRD flow control valve fails closed on a loss of CAS.

They have a stop that prevents them from fully closing. B. Incorrect - The intertie valve is gagged open, so that it will stay open with a loss of CAS. C. Incorrect - The feedpump governor valves are electro-hydraulically operated. D. Incorrect - The RCCW heat exchanger discharge valves are motor operated. COMMENTS: Ref. 10 CFR 55.41 (7) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 7

EXAM KEY NOVEMBER 2006 Page 7 of 81 Columbia is in MODE 4. An inadvertent drain-down of the reactor vessel has reduced level to -55 inches.

Which of the following methods of decay heat removal from the RPV is available?

A. RHR A or B - Shutdown Cooling.

B. Injection with HPCS AND opening some SRVs.

C. Reject heat through the RW CU non-regenerative heat exchanger.

D. Run at least one Reactor Recirc Pump.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295021 AK1.03 Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING: Adequate core cooling (3.9)

REFERENCE:

CGS System Description, RHR, Rev. 11, pg. 20; SD000198

SOURCE: New Question

LO: 5781 c. List the interlocks and trips associated with the following RHR system components: RHR-V-6A/B

RATING: Knowledge: Analysis Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - V-8 closes at level 3. B. Correct - HPCS is available, as well as SRVs in this mode.

C. Incorrect - RWCU isolates at level 2.

D. Incorrect - This provides forced core cooling, but not decay heat removal. COMMENTS: Ref : 10 CFR 50.41 (7) & (8) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 8

EXAM KEY NOVEMBER 2006 Page 8 of 81 Given the following conditions :

Columbia was operating at 100% power when a LOCA occurred. Drywell pressure is 15 psig. Reactor pressure is 400 psig.

For the above conditions, which of the following operator actions will clear the interlocks for RHR-V-24A, Suppression Pool cooling, so that it can be opened?

A. Place switch RHR-RMS-S105, RHR-V-42A Valve Logic Override, in OVERRIDE.

B. Place switch RHR-RMS-S101A, RHR-V-42A Permissive Override, in TEST.

C. ONLY CLOSE RHR-V-42A, LPCI injection.

D. CONTINUALLY hold switch for RHR-V-24A in the OPEN position.

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295024 EK2.12 Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: Suppression pool cooling (3.5)

REFERENCE:

CGS System Descripti on, RHR, Rev. 11, pg. 15-16 SOURCE: New Question

LO: 5781 List the interlocks and trips a ssociated with the following RHR system components: f. RHR-V-24A/B and RHR-V-27A/B

RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None JUSTIFICATION: A. Incorrect - This will allow th rottling of 42a, but is not necessary below 470. B. Incorrect - This will allow 42A to be tested during normal ops.

C. Correct - This will clear the interlock, allow the valve to be opened and stay open as long as pressure stays below 470. D. Incorrect - This only overrides the logic if a LPCI initiation has not already occurred. COMMENTS: Ref : 10 CFR 55.41 (7) & (9) & (10) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 9

EXAM KEY NOVEMBER 2006 Page 9 of 81 The EOPs require an Emergency Depressurizati on be performed prior to exceeding the high wetwell temperature limit.

The bases for this requirement is to protect the:

A. Fuel Cladding.

B. Reactor Pressure Vessel.

C. Primary Containment.

D. Reactor Building.

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295026 EK3.01 Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Emergency/normal depressurization (3.8)

REFERENCE:

5.0.10, EOP Flowchar t Training Manual, Rev. 7, page 200

SOURCE: New Question

LO: 5629 State the three (3) purpos es of the suppression chamber.

RATING: Knowledge: Analysis Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - ECCS suction would be adversely affected by an ED, but even if primary containment was lost, CST is still available. B. Incorrect - ED is to mitigate the possibility of a LOCA, not to prevent one. C. Correct - The basis is to maintain the pressure suppression capability of the containment. D. Incorrect - Secondary containment would not be damaged by this problem. COMMENTS: Ref: 10 CFR 55.41 (9) & (10) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 10

EXAM KEY NOVEMBER 2006 Page 10 of 81 PPM 5.2.1, Primary Containment Control, has been entered and it has been determined that WW level CANNOT be maintained above 19 ft 2 in. According to the procedure, an Emergency Depressurization is required.

If the Emergency Depressurization is not started unt il level has lowered to 18 ft 6 in, what are the consequences?

A. Vortexes at the suction ECCS pumps can begin and result in air binding of the pumps. B. Suppression pool temperature indication becomes invalid.

C. Condensation of steam from the SRV downcomers cannot be assured.

D. The SRV Tail Pipe Level Limit (SRVTPLL) will be exceeded.

ANSWER: C Post Exam Comment - The licensee recommended this question be deleted because the term "SRV downcomers" is confusing. They contend the correct

term is "quencher". This comment was rejected because the steam reaches the

quencher via the downcomers and the SR V downcomers are the only ones located in the suppression pool. Based on th is, distractor C is correct as stated. QUESTION TYPE: Closed Reference

KA # & KA VALUE: [New KA] 295030 EK1.01 Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Steam Condensation (3.8) [New KA]

[KA Deleted] 295030 EK1.03 Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Heat capacity (3.8) [KA Deleted]

REFERENCE:

PPM 5.2.1, Primar y Containment Control, Re

v. 15 and 5.0.10, Flowchart Training Manual, Rev. 7, pg. 262

SOURCE: 2003 CGS Initial Licensing Exam - Modified

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - Vortexing does not begin until 17.5 ft for the RCIC system. B. Incorrect - Temperature indicati on is still valid at this level. C. Correct - The downcomer vents will be exposed, and steam will not be properly condensed. D. Incorrect - The SRVTPLL will be exceeded by raising level.

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 11

EXAM KEY NOVEMBER 2006 Page 11 of 81 COMMENTS: Ref : 10 CFR 50.41 (5) & (8) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 12

EXAM KEY NOVEMBER 2006 Page 12 of 81 Procedure 5.1.2, RPV Control - ATWS, has been ent ered. Boron Injection is required and the SLC system keylock switches have been taken to operate.

Which of the following conditions would prevent bor on injection if no other operator action is taken?

A. The SLC Test Tank Outlet Valve SLC-V-31 is OPEN.

B. The RWCU Outboard Isolati on Valve RWCU-V-4 is OPEN.

C. Storage Tank Outlet Valve SLC-V-1A OR SLC-V-1B is CLOSED.

D. Loss of continuity to squib valves SL C-V-4A & 4B AFTER keylock switches are taken to OPER.

ANSWER: A QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295037 EA1.04 Ab ility to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: SBLC (4.5)

REFERENCE:

CGS System Description, SLC, Rev. 11, pg. 9; SD000172

SOURCE: Clinton 1 7/23/2001 - Modified

LO: 5925 Describe the expected response to placing the SLC system A or B keylock switch in the operate position.

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Correct - The pump suction valves cannot open if the test valve is open. The pump suction valves not fully open prevents boron

injection if no other operator action is taken. B. Incorrect - The only input to SLC pump start circuitry is the suction valve position. C. Incorrect - Flow is still possible through one train.

D. Incorrect - This is the indicati on expected for the squib valves after firing (opening). COMMENTS: 10 CFR 50.41 (6) (7) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 13

EXAM KEY NOVEMBER 2006 Page 13 of 81 Which of the following describes the Halon system

's discharge rate in the Control Room floor modules?

Does the Halon discharge, by itsel f, make the Control Room uninhabitable?

A. Discharges fully immediately.

Control Room is still habitable.

B. Discharges fully immediately.

Control Room must be evacuated.

C. Discharges over an extended period of ti me. Control Room is still habitable.

D. Discharges over an extended period of ti me. Control Room must be evacuated.

ANSWER: C Post Exam Comment - During the exam, one applicant asked if the question involved one bottle of Halon or the Halon System. The applicant was told to

consider the question from the system perspective. QUESTION TYPE: Closed Reference

KA # & KA VALUE: 600000 AK1.02 Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire fighting (2.9)

REFERENCE:

CGS System Description, Fire Protection, Rev. 11, Pg. 11-13; SD000177 SOURCE: New Question

LO: 5376 Briefly explain the operation of the following types of fire suppression systems: e. Control Room Halon 1301 Floor Modules

RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - the dischar ge is in three stages over 12 minutes. B. Incorrect - the discharge is in three stages over 12 minutes.

C. Correct - The discharge is extended, and the Halon is designed so that the CR will still be habitable. D. Incorrect - The Halon is designed to leave the CR habitable. COMMENTS: Ref : 10 CFR 55.41 (4) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 14

EXAM KEY NOVEMBER 2006 Page 14 of 81 Columbia was operating at full power with RCIC tagged out. A trip of both Reactor Feedwater Pumps caused RPV level to drop and a Reactor scram to occur on low RPV water level. HPCS recovered RPV level and HPCS-V-4 was closed when RPV water level reached +35". Conditions

are currently:

-Reactor Pressure 700 psig (rising at 10 psig per minute) -Time after scram 5 minutes

-CRD pumps both tripped

-Drywell pressure 0.3 psig

If no additional operator actions are taken, what is the expected RPV water level response over the next 10 minutes and why?

RPV water level will-A. rise above the high RPV water level trip setpoint due to heatup.

B. rise above the high RPV water level trip setpoint due to Startup Valve leakage exceeding decay heat requirements.

C. lower below the low level alarm point due to cooldown.

D. lower below the low RPV level trip se tpoint due to steam loads reducing RPV water inventory.

ANSWER: A QUESTION TYPE: Closed Reference KA # & KA VALUE: 295008 AA2.05 Ab ility to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL: Swell (2.9)

REFERENCE:

SOURCE: Cooper Exam 8/2/2002 LO:

RATING: Knowledge: Analysi s Difficulty: 3 ATTACHMENT: None JUSTIFICATION: A. Correct - The specific volume change from the CST water to saturated liquid at 700 psig results in ~40% increase in gallons/inches of RPV level. 80" of cold water added = 112". B. Incorrect - Rx pressure is above Condensate Booster pressure.

C. Incorrect - RPV water level will rise.

D. Incorrect - RPV level will rise. BPVs are closed and other steam loads will not lower level under these conditions. COMMENTS: Ref: 10 CFR 55.41 (5) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 15

EXAM KEY NOVEMBER 2006 Page 15 of 81 Procedure ABN-CRD-MAXFLOW has been entered due to a low RPV level condition. This procedure contains a caution that states "Do not lower drive water pressure to LT 260 psid".

The bases for maintaining drive water pressure greater than 260 psid is to:

A. Maintain seal water to the reactor recirc pump seals.

B. Support control rod in sertion per the EOP's.

C. Prevent CRD pump runout at low RPV pressures.

D. Maintain cooling water to the Control Rod Drive Mechanisms.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295009 Low Reactor Water Level 2.1.32 Ab ility to explain and apply system limits and precautions (3.4)

REFERENCE:

ABN-CRD-MAXFLOW, Revision 1, Bases section CGS System Description, Control R od Drive Hydraulic System, Rev. 12

SOURCE: New Question

LO: 5186. a. Describe the impact of CRD pressure controller operation on the following CRD System parameters: Dr ive water differential pressure RATING: Knowledge: Fundamental / Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - Drive water DP and suction pressure of the pump are not directly related. B. Correct - Dp is necessary to drive rods.

C. Incorrect - This does not prevent runout, because the pump could still runout through the scram valves if it was going to. D. Incorrect - The FCV maintains cooling flow. COMMENTS: Ref. 10 CFR 55.41 (6)

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 16

EXAM KEY NOVEMBER 2006 Page 16 of 81 After a 370 day run at full power, a Group 1 isolati on resulted in a reactor scram. EOP 5.1.1, RPV Control, and EOP 5.2.1, Primary Containment Cont rol have been entered.

Given the following conditions: RPV Pressure is being maintained 800 to 1000 psig using SRVs. RPV Level is +30 inches and stable. Suppression Pool cooling is unavailable. Wetwell temperature is 106 F and rising.

Which of the following injection lineups will resu lt in the LEAST amount of heat added to the Wetwell assuming this lineup will be maintained for several hours?

A. HPCS flow from the CST.

B. RCIC flow from the CST.

C. HPCS flow from the Wetwell.

D. RCIC flow from the Wetwell.

ANSWER: B Post Exam Comment - T he NRC does not agree with the licensee's recommendation to delete Question #15.

The HPCS system is a larger pump that will add more heat to the suppression pool via the minimum flow line than

RCIC. Secondly, RCIC is drawing 800-1000 psig steam off the reactor and exhausting it to the suppression pool at a much lower enthalpy. Otherwise, the steam would be exhausted directly from the reactor to the suppression pool via the relief valves. Both these factor s would reduce the amount of heat being added to the suppression pool making RCIC the only correct answer. QUESTION TYPE: Closed Reference KA # & KA VALUE: 295013 AA1.02 Ab ility to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE

Systems that add heat to the suppression pool (3.6)

REFERENCE:

CGS System Description, RCIC, Rev. 12, Procedure 5.2.1, Primary Containment Control, Rev. 15, and Proc edure 5.1.1, RPV Control, Rev. 16 SOURCE: New Question LO: RATING: Knowledge: Analysis, Difficulty: 4 ATTACHMENT: None JUSTIFICATION: A. Incorrect - The injection flow into the RPV would be the same as RCIC, but none of the steam sent to the WW would go through RCIC, so it would have more overall enthalpy. B. Correct - Some of the steam would exhaust through RCIC vice the SRVs lowering the overall heat being added to the wetwell. C. Incorrect - using the warmer water as a suction requires less decay 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 17

EXAM KEY NOVEMBER 2006 Page 17 of 81 heat from the reactor to produce the sa me amount of steam. It is also not using the steam for work. D. Incorrect - using the warmer water as a suction requires less decay heat from the reactor to produc e the same amount of steam. COMMENTS: 10 CFR 55.41 (5), (8), (14) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 18

EXAM KEY NOVEMBER 2006 Page 18 of 81 Due to special test requirements, Columbia Gener ating Station is in MODE 2 being shut down by control rod insertion instead of a manual scram. Reac tor pressure is steady at 900 psig and reactor power is at 12 on IRM range 7.

Due to a plant problem contro l rod insertion has been stopped.

With no operator action taken, which of the following will result?

A. Reactor Scram due to hi neutron flux.

B. Reactor Scram due to MSIV closure.

C. Rod block due to SRM upscale.

D. Rod block due to IRM downscale.

ANSWER: A Post Exam Comment - T he NRC does not agree with the licensee's recommendation to delete Question #16. With the power in intermediate range 7, the reactor is critical and the turbine is off-line. In addition to the steaming, the makeup water to the reactor it will be rela tively cold and moderator temperature will be driven down. This adds positive reac tivity thereby increasing power until the scram setpoint is reached ma king A the only correct answer. QUESTION TYPE: RO

KA # & KA VALUE: 295014 AK2.06 Knowledge of t he interrelations between Inadvertent Reactivity Addition and the following:

Moderator temperature (3.4 / 3.5)

REFERENCE:

SER 24-91; SD000161

SOURCE: Bank - Modified; Analysis Difficulty: 3

LO: 5192

RATING: H3

ATTACHMENT: None

JUSTIFICATION: A. Correct - Reactor power will increase due to the effects of the cooldown causing a hi flux scram. B. Incorrect - RPV Pressure is on DEH in automatic. MSIV close at 831 in RUN. C. Incorrect - SRM rod block is bypassed.

D. Incorrect - Power increases not decreases. COMMENTS: POAH is 25 on IRM Range 8 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 19

EXAM KEY NOVEMBER 2006 Page 19 of 81 Columbia is operating at 100% power. Radiati on readings for the Reactor Building Exhaust Air Plenum begin rising on REA-RIS-609A, B, C, and D.

If the radiation source continues to give off I NCREASING amounts of radioactive gas, the Reactor Building Exhaust Plenum Radiation Monitoring System recorder outputs will rise to the 'Z' signal setpoint of _________

and then the Reactor Building Exhaust Air Plenum Radiation Monitoring System recorder output readings will _________ . A. 13 mr/hr, continue to RISE.

B. 13mr/hr, STABILIZE OR DROP.

C. 15mr/hr, continue to RISE.

D. 15mr/hr, STABILIZE OR DROP.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 295034 EA1.02 Ab ility to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Process Radiation monitoring system (3.9)

REFERENCE:

CGS System Description, PRM, rev. 11, pg. 19-20; SD000147; SD000173

SOURCE: New Question

LO: 5647 State the automatic actions associated with each of the following gaseous and liquid stream Process Radiation Monitors upon sensing high

radiation levels: g. Reactor Building Exhaust Plenum RMS RATING: Knowledge: Analysis Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - Z signal is olates flow through plenum at 13mr/hr, therefore, no more radioactive gas es would be drawn in, and what gas was already in the plenum would decay. B. Correct - Z signal isolates flow through plenum at 13mr/hr.

C. Incorrect - The Z signal setpoint is 13mr/hr.

D. Incorrect - The Z signal setpoint is 13mr/hr. COMMENTS: Ref : 10 CFR 55.41 (4) & (7) & (11) & (13) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 20

EXAM KEY NOVEMBER 2006 Page 20 of 81 Columbia is in MODE 4 twelve hours after a S CRAM with Shutdown Cooling in service on RHR A.

Reactor water level is 55 inches and pressure is 75 psig. The CRO mistakenly throttles down on RHR-V-3A, Heat Exchanger Shell Side Outlet, causing RHR flow to DECREASE to 700 gpm.

Over the next 30 minutes, if no other operator actions are taken, RPV :

A. temperature will remain STABLE.

B. level will INCREASE.

C. temperature will DECREASE.

D. level will DECREASE.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 205000 A1.05 Ab ility to predict and/or moni tor changes in parameters associated with operating the S HUTDOWN COOLING SYSTEM/MODE controls including: Reactor water level (3.4)

REFERENCE:

CGS System Descripti on, RHR, rev. 11, pg. 22; SD000198

SOURCE: Fermi 2, 12/11/1995 - Modified

LO: 7728 Describe the physical connection and / or the cause-and-effect relationships between the RHR syst em and the following: g. RPV

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - The reactor will heat up the RPV because of lower SDC flow. B. Correct - Flow through the RHR HX drops causing less cooling resulting in a temp rise causing RPV level to rise. C. Incorrect -The reactor will heat up the RPV because of lower SDC flow. D. Incorrect - The reactor will heat up the RPV because of lower SDC flow. COMMENTS: 10 CFR 55.41 (7), (14) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 21

EXAM KEY NOVEMBER 2006 Page 21 of 81 Which of the following combinations of R PV Wide Range Level (MS-LIS-37A/B) and Drywell Pressure Switches (MS-PS-48A/B/C/D) will cause the Low Pressure Core Spray System to automatically initiate?

A. MS-LIS-37B and MS-PS-48A B. MS-LIS-37A and MS-PS-48B C. MS-LIS-37A and MS-PS-48C D. MS-LIS-37B and MS-PS-48D ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 209001 K1.09 Knowledge of the physical connections and/or cause- effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: Nuclear boiler instrumentation (3.2)

REFERENCE:

CGS System Description, LPCS, Rev. 10, pg. 7; SD000192

SOURCE: New Question

LO: LO-5481 List the signals and setpoint s, which cause a LPCS initiation.

RATING: Knowledge: Fundamental Difficulty: 4

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - this is not a combination that will auto start. B. Incorrect - this is not a combination that will auto start.

C. Correct - see reference.

D. Incorrect - this will start LPCI B & C COMMENTS: 10 CFR 55.41 (7) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 22

EXAM KEY NOVEMBER 2006 Page 22 of 81 A Large Break LOCA has occurred and the RPV is depr essurized. HPCS is injecting from the CST, and LPCS is injecting. There ar e NO other injection lineups available.

To prevent Wetwell level from reaching the SRV Tail Pipe Level Limit of _______, the _______

pump should be stopped.

A. 51 ft, LPCS B. 41 ft, LPCS C. 51 ft, HPCS D. 41 ft, HPCS

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 209002 A2.12 Ab ility to (a) predict the impacts of the following on the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High suppression pool level (3.3)

REFERENCE:

Procedure 5.2.1, Primar y Containment Control, rev. 15 SOURCE: New Question

LO: 8384 Given a list, identify the statem ent that describes the purpose of terminating injection into the primary containment if wetwell level and RPV pressure cannot be maintained below the SRVTPLL.

RATING: Knowledge: Analysis Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - this is not an injection source external to the PC. B. Incorrect - this is not an inje ction source external to the PC. C. Correct - per the EOP, stop inject ion source external to the PC, the SRVTPLL is 51 ft. D. Incorrect - this is the incorrect SRVTPLL. COMMENTS: 10 CFR 55.41 (8)

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 23

EXAM KEY NOVEMBER 2006 Page 23 of 81 Columbia is operating at 90% power in three-el ement control when an inad vertent actuation of HPCS occurs.

If all systems respond as expected, with no operator action, what will indicated RPV water level do?

Indicated RPV water level will-.

A. rise, then stabilize lower than the original level.

B. rise, then stabilize higher than the original level.

C. lower, then stabilize lowe r than the original level.

D. lower, then stabilize higher than the original level.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 209002 A1.03 Ab ility to predict and/or moni tor changes in parameters associated with operating the H IGH PRESSURE CORE SPRAY SYSTEM (HPCS) controls including: Reactor water level (3.7)

REFERENCE:

CGS FSAR, Section 15.5.1, Inadver tent High-Pressure Core Spray Startup, Amendment 57

CGS System Description, Feedwater Level Control System, Rev. 13

SOURCE: New Question

LO:

RATING: Knowledge: Analysis Difficulty: 4

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - Level will initia lly swell, but the higher steam flow than feed flow will produce a level error due to higher steam flow than feed

flow, this will speed up the FP until the level error goes away in the

FWLC system. B. Correct - Level will initially sw ell. The FWLC system will have an error due to steam flow > feed flow. It will bias stabilized level higher. C. Incorrect - Level will initially swell due to pressure decrease from HPCS spray in shroud. D. Incorrect - Level will initially swell. COMMENTS: 10 CFR 55.41 (5)

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 24

EXAM KEY NOVEMBER 2006 Page 24 of 81 When the SLC System A keylock switch is taken to the 'OPER' position, what is the overall expected RWCU System response?

A. RWCU-V-4 closes AND RWCU-FCV-33 closes if open B. RWCU-V-1 closes AND RWCU-FCV-33 closes if open C. ONLY RWCU-V-4 closes D. ONLY RWCU-V-1 closes ANSWER: A QUESTION TYPE: Closed Reference

KA # & KA VALUE: 211000 A4.06 Ab ility to manually operate and/or m onitor in the control room: RWCU system isolation (3.9)

REFERENCE:

CGS System Descrip tion, RWCU, Rev. 10; SD000190 SOURCE: New Question

LO: 5931 Given one or more systems that interrelate to SLC, state the importance or function of that relationship

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Correct, SLC operation causes V-4 to close, V-4 closed causes FCV-33 to close. It is important that FCV-33 close to remain an option for

boron injection. B. Incorrect, V-4 closes.

C. Incorrect, FCV-33 closes on a V-4 closure signal.

D. Incorrect, V-4 closes. COMMENTS: Reference : 10 CFR 55.41 (6), (9) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 25

EXAM KEY NOVEMBER 2006 Page 25 of 81 Both SLC system keylock switches have been taken to OPERATE. ONE of the squib valves failed to open.

What should the operator expect the APPROXIMATE boron injection flowrate to the RPV to be if all other components operate as expected?

A. 30 gpm B. 45 gpm C. 60 gpm D. 90 gpm

ANSWER: D Post Exam Comment - The licensee recommended deleting this question because the design pumping rate of each SLC pump is 43 gpm. This makes the

total system flow 86 gpm and not 90 gpm as distractor D contains. This comment was rejected because the stem of the ques tion asks what is the "APPROXIMATE" flowrate. This was used because the actual flowrate will vary from 85 to 90 GPM due to variances in pump construction and 86 gpm would logically be approximately 90 gpm compared to any of the other distractors. QUESTION TYPE: Closed Reference

KA # & KA VALUE: 211000 A1.04 Ab ility to predict and/or moni tor changes in parameters associated with operating the ST ANDBY LIQUID CONTROL SYSTEM controls including: Valve operations (3.6)

REFERENCE:

CGS System Descrip tion, SLC, Rev. 11; SD000172

SOURCE: New Question

LO: 5922 Describe the following SLC system flowpaths: a. Normal Injection

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - dual pump flow is approximately 90. B. Incorrect - single pump output is 45 gpm, but the pumps are cross-tied. C. Incorrect - dual pump flow is approximately 90.

D. Correct - the pumps are cro ss-tied, and they are positive displacement pumps. Pump output will be 90 gpm. COMMENTS: Reference : 10 CFR 55.41 (6) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 26

EXAM KEY NOVEMBER 2006 Page 26 of 81 Changed B to 45 from 40 and D to 90 from 80 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 27

EXAM KEY NOVEMBER 2006 Page 27 of 81 Columbia is operating at 99% power with no equipm ent out of service. I&C technicians are preparing to start a test on RPS trip system A t hat will produce a half-scram. You notice one of the four RPS Scram Group lights is NO T lit on RPS trip system B.

What are your immediate actions per proc edure ABN-RPS, and what would the consequences be if the RPS A half scram is initiated?

A. Replace fuse, half-scram with no rod movement.

B. Immediately stop the work on RPS A, half-scram with no rod movement.

C. Replace fuse, one quarter of the rods scram.

D. Immediately stop the work on R PS A, one quarter of the rods scram.

ANSWER: D QUESTION TYPE: Closed Reference

KA # & KA VALUE: 212000 A2.19 Ab ility 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: Part ial system activation (Half-SCRAM)

(3.8)

REFERENCE:

Procedure ABN-RPS, Rev. 2 CGS System Description, RPS, rev. 12; SD000161

SOURCE: New Question

LO: 7683 Predict the effect(s) that a fa ilure of the RPS system will have on: a.

Scram and Backup Scram valves RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None JUSTIFICATION: A. Incorrect - One quarter of the B scram solenoids are de-energized.

When the A scram solenoids are de-energized, those rods will scram. B. Incorrect- One quarter of the rods will move.

C. Incorrect - Procedure ABN-RPS direct s you to stop work on the other trip system. D. Correct - One quarter of the rods would scram, procedure directs you to stop work on the A trip system. COMMENTS: Reference 10 CFR 55.41 (7) & (10) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 28

EXAM KEY NOVEMBER 2006 Page 28 of 81 IRM channel E is on Range 3 and is reading 10/40 scale. Over the next few minutes, reactor power doubles about 3.5 times.

Based on this change in reactor power, the CRO should expect IRM Channel E to now indicate approximately--

A. 25/40 on Range 3 B. 78/125 on Range 4 C. 12/40 on Range 5 D. 95/125 on Range 6

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 215003 A4.03 Ab ility to manually operate and/or m onitor in the control room: IRM range switches (3.6)

REFERENCE:

CGS System Description, Inte rmediate Range Monitor, Rev. 8; SD000138 SOURCE: Bank - Slightly Modified

LO: 5461 Describe the correlation between Reactor Period and IRM indication.

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: 10 to 20 on Range 3 is one doubling; 20 to 40 on Range 3 is another doubling; 40 to 80 on Range 4 is another doubling (3 total); .57 doublings

more is about 120 Range 4 or 12 Range 5. COMMENTS: Ref : 10 CFR 55.41 (7) & (1)

Revised stem 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 29

EXAM KEY NOVEMBER 2006 Page 29 of 81 Which of the following circumstances for the Source Range Monitors will generate a rod block?

A. 5 x 10 4 counts per second and all IRMs on range 3.

B. 0.9 counts per second.

C. SRM channel A mode switch in standby.

D. One detector retracted with 165 counts per second.

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 215004 K1.03 Knowledge of the physical connections and/or cause- effect relationships between SOURCE RANG E MONITOR (SRM) SYSTEM and the following: Rod control and information system: Plant specific (3.0)

REFERENCE:

CGS System Description, Sour ce Range Monitor, Rev. 10; SD000132

SOURCE: New Question

LO: 5943 List the scrams and the rod blocks generated by the SRM system.

Include the setpoints for each and when they are bypassed.

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - 10 5 counts per second is the rod block. B. Incorrect - The setpoint is 0.7 cps. C. Correct - One channel being out of operate provides a rod block. D. Incorrect - This rod block is set at 100cps. COMMENTS: Ref : 10 CFR 55.41 (7) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 30

EXAM KEY NOVEMBER 2006 Page 30 of 81 With Columbia operating at full power, annuncia tor CIA DIV 1 OUT OF SERVICE annunciates.

You then receive a call reporting that CIA-RV-5A, which is located downstream of Nitrogen backup bank bottle pressure control valve, CIA-PCV-2A, and between CIA-V-30A and CIA-V-39A, is stuck open. There is a large amount of ni trogen escaping depressurizing the line.

What supply(ies) is(are) still available to operat e ADS valves in addition to the accumulators?

A. ONLY the Main Header for ALL SRVs.

B. ONLY ADS Accumulator Header B for Division II ADS SRVs.

C. The Main Header for ALL SRVs, and ADS Accumulator Header B for ALL SRVs.

D. The Main Header for ALL SRVs, and ADS Accumulator Header B for Division II ADS SRVs.

ANSWER: D QUESTION TYPE: Closed Reference

KA # & KA VALUE: 218000 K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM: Air supply to ADS valves: Plant specific (3.6)

REFERENCE:

CGS System Description, Containm ent Instrument Air, Rev. 8; SD000156

SOURCE: New Question

LO: 7748 Determine the effect a CIA malfunction has on: b. SRVs

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - ADS Accu mulator Header B is still available. B. Incorrect - the Main header is still available.

C. Incorrect - the ADS Accumulator Header B cannot backfeed into A.

D. Correct - A header isolates, main header and B header are still available. COMMENTS: Reference : 10 CFR 55.41 (7), (8), & (10) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 31

EXAM KEY NOVEMBER 2006 Page 31 of 81 Given the following conditions :

Drywell Pressure Instrument MS-PS-48D in TRIP. Drywell Pressure Instrument MS-PS-48B failed HIGH. Procedure ABN-FAZ has been entered.

Which of the following describes the automatic actuations of the NSSSS system AND describes the operator action necessary to mitigate that impact?

A. ALL RCC to the Drywell is ISOLATED, VENT the Drywell.

B. ALL Circulating Water Pumps TRIP, Place RHR in Suppression Pool Cooling.

C. ALL TSW Pumps TRIP, Ensure a TSW pump STARTS after DG starts.

D. RWCU to the Drywell is ISOLATED, OPEN RWCU-FCV-33, Blowdown Control Valve, to prevent RWCU relief valves from lifting.

ANSWER: A QUESTION TYPE: Closed Reference KA # & KA VALUE: 223002 A2.06 Ab ility to (a) predict the impacts of the following on the PCIS/NSSSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Containment instrumentation failures (3.0)

REFERENCE:

ABN-RCC, Rev. 3, pg. 3 CGS System Description, NS4, Rev. 10, pg. 5; SD000173 SOURCE: New Question LO:

RATING: Knowledge: Analysi s Difficulty: 3 ATTACHMENT: None JUSTIFICATION: A. Correct - RCC isolated can cause drywell pressure to exceed F signal, venting may be necessary. B. Incorrect - Circ water pump A does not trip. C. Incorrect - NO pumps will trip without a LOOP signal.

D. Incorrect - RWCU does NOT isolate on drywell pressure or plant trip or ECCS initiation. COMMENTS: Ref : 10 CFR 55.41 (5) & (7) & (10) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 32

EXAM KEY NOVEMBER 2006 Page 32 of 81 The Control Room has been evacuated due to a noxi ous chemical. The following conditions exist:

RPV Level is BELOW Level 1.

RHR-P-2A AND RHR-P-2B are running.

DP-S1-2A is de-energized.

All Normal/Emergency switches at the Remote S hutdown Panel (RSP) and the Alternate Remote.

Shutdown Panel (ASP) have been taken to Emergency.

Based on the given conditions, which of the following describes operation of the SRVs?

A. ADS CAN automatically initiate AND AD S SRVs can be manually operated from the RSP. B. ADS CAN automatically initiate BUT ADS SRVs CAN NOT be manually opened from the RSP. C. ADS CAN NOT automatically initiate BU T ADS SRVs can be manually opened from the RSP. D. ADS CAN NOT automatically initiate A ND ADS SRVs CAN NOT be manually opened from the RSP.

ANSWER: C QUESTION TYPE: Closed Reference KA # & KA VALUE: 239002 K4.05 Knowledge of RELIEF/SAFETY VAL VES design feature(s) and/or interlocks which provide for the fo llowing: Allows for SRV operation from more than one location: Plant specific (3.6)

REFERENCE:

CGS System Description, ADS, Re

v. 10, pg. 8; SD000186 and RSP, Rev. 6, pg. 8 ; SD000210 SOURCE: New Question LO: 5077 List the power supplies to the ADS solenoids 5886 State the effects to associated component controls and alarms when their Power Transfer Switches are placed to the EMERGENCY position. RATING: Knowledge: Analysis Difficulty: 4

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - ADS B logic has lost power. ADS A has logic, but will be

'blocked' by the ARS emergency switches. Therefore - not ALL ADS valves will open, but some will. B. Incorrect - ADS B logic has lost power. ADS A has logic, but will be

'blocked' by the ARS emergency switches. Therefore - not ALL ADS valves

will open, but some will. C. Correct - ADS valves will still operate from the RSP because they are powered from DP-S1-2D after the transfer. D. Incorrect - ADS valves will still operate from the RSP because they are powered from DP-S1-2D after the transfer.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 33

EXAM KEY NOVEMBER 2006 Page 33 of 81 Columbia is operating at 95% power with the Feedw ater Level Control System in three-element control. RPV level is 36".

If the controlling Narrow Range Level instrument fails high, what will be the i mmediate trend of the ACTUAL RPV level, AND, what is the MAXIMU M / MINIMUM level WHEN RPV level stabilizes?

A. Up, BELOW Level 7 B. Up, ABOVE Level 7 C. Down, BELOW Level 4 D. Down, ABOVE Level 4

ANSWER: D QUESTION TYPE: Closed Reference KA # & KA VALUE: [New KA] 259002 K5.03 Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM: Water level measurement (3.1) [New KA] [KA Deleted] 259002 K5.09 Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM: Adequate core cooling: FWCI (3.8) [KA Deleted]

REFERENCE:

CGS System Description, FWLC, Rev. 13, pg. 6,16; SD000157 SOURCE: New Question LO: 5400 Predict the expected response of the feedwater level control system in both Single and Three Element Control, to a failure or malfunction of the following: Loss of the selected RPV Level Channel 9711 Describes the FWLC system malfunc tions, which will initiate a RFW CONTR SYSTEM TROUBLE alarm. RATING: Knowledge: Analysis Difficulty: 2 ATTACHMENT: None JUSTIFICATION: A. Incorrect - If controlling le vel fails high, RFPT will slow down, causing level to go down. B. Incorrect - If controlling level fails high, RFPT will slow down, causing level to go down. C. Incorrect - Level should deviate a max of 3 inches before recovering.

D. Correct - Level should deviate a max of 3 inches before recovering. COMMENTS: Ref : 10 CFR 55.41 (3) & (7) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 34

EXAM KEY NOVEMBER 2006 Page 34 of 81 An accident occurred allowing radioactive gas to be released into the Reactor Building.

If a SGT pressure controller fails, and Reactor Build ing pressure increases above zero, this would cause the offsite release rate of halogens to

___________. The operator can REDUCE the offsite release of halogens by manually___________ Standby Ga s Treatment flow to the elevated release. A. Increase, Increasing B. Increase, Decreasing C. Decrease, Increasing D. Decrease, Decreasing

ANSWER: A QUESTION TYPE: Closed Reference

KA # & KA VALUE: 261000 K3.02 Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: Off-site release rate (3.6)

REFERENCE:

CGS System Descrip tion, SGT, Rev. 12 SD0900144

SOURCE: New Question

LO: 5822 State the Reactor Building pre ssure the SGT system is designed to maintain, as well as the pressure its DPIC is set to maintain and why it is at that setting.

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Correct, unfiltered rel ease increases. More flow means less pressure. B. Incorrect - less flow will relatively increase pressure.

C. Incorrect - positive pressure creates outflow, which will increase release rate. D. Incorrect - less flow will relatively increase pressure. COMMENTS: Ref : 10 CFR 55.41 (13) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 35

EXAM KEY NOVEMBER 2006 Page 35 of 81 Columbia was operating at full power with Divis ion 1 Standby Gas Treatment train tagged out for charcoal replacement when a Large Break Loss of Coolant Accident occurred.

Flow to the elevated release point from the Standby Gas Treatment system will be ________ with only one train operating instead of two, and the offsite release will be _____________ 10 CFR Part 100, Reactor Site Criteria, limits during the accident.

A. the same, above B. the same, below C. lower, above D. lower, below

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 261000 K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM: A. C. Electrical distribution (2.9)

REFERENCE:

CGS System Description, SGT, Rev. 12, pg. 3; SD000144

SOURCE: New Question

LO: 5821 State the purpose of t he Standby Gas Treatment system.

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - The design bas es of the SGT is to maintain releases within part 100 during DBAs. B. Correct - Flow is controlled to pressure, and the design of SGT is to maintain within part 100 limits. C. Incorrect - The flow will be the same because the controller controls to a certain pressure which can be maintained with a single train. D. Incorrect - The flow will be the same because the controller controls to a certain pressure which can be maintained with a single train. COMMENTS: Ref : 10 CFR 55.41 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 36

EXAM KEY NOVEMBER 2006 Page 36 of 81 On 4160 volt AC breakers NOT in the HPCS system which of the following is correct?

The charging motor charges the __________

spring. If DC control power is lost, __________

breaker trip(s) is / are still active.

A. opening; no B. opening; the overcurrent C. closing; no D. closing; the overcurrent

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 262001 K5.02 Knowledge of the operational implications of the following concepts as they apply to A. C. ELECTRICAL DISTRIBUTION: Breaker control (2.6)

REFERENCE:

CGS System Description, AC Distr ibution, Rev. 13, Pg. 20-23; SD000182

SOURCE: New Question

LO: 5065, 5051

RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - The clos ing springs are charged by motor. B. Incorrect - The closing springs are charged by motor.

C. Correct.

D. Incorrect - The overcurrent tr ip is powered from DC and is NOT active when DC is lost. COMMENTS: Ref : 10 CFR 55.41 (7)

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 37

EXAM KEY NOVEMBER 2006 Page 37 of 81 During a Station Blackout, as the battery discharges over time loads such as a motor will draw

___________ current while running. This is because battery voltage ___________ over time.

A. less, decreases.

B. less, increases.

C. more, decreases.

D. more, increases.

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 263000 A1.01 Ab ility to predict and / or monitor changes in parameters associated with operating the D.C. EL ECTRICAL DISTRIBUTION controls including: Battery charging/discharging rate (2.5)

REFERENCE:

CGS Procedure 5.

6.1, Rev. 12, pg. 5

SOURCE: New Question

LO:

RATING: Knowledge: Analysis/ Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - As voltage drops, the DC motor will demand more current. B. Incorrect - As voltage drops, the DC motor will demand more current.

C. Correct - The battery voltage will decrease and more current is drawn from the battery as it discharges D. Incorrect - Motor current goes up, and battery voltage decreases over time. COMMENTS: Ref : 10 CFR 55.41 (8) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 38

EXAM KEY NOVEMBER 2006 Page 38 of 81 The plant has experienced a small LOCA. Drywell pr essure is 6 psig. All systems functioned as designed. Ten minutes after the acciden t, all off-site power is lost.

Which of the following automatic responses would you expect following the loss of off-site power?

A. SW-P-1A/1B will start after its 20 second time delay.

B. DG-3 will trip on high jacket water temperature.

C. A Failure to Auto Start alarm will annunciate for all DGs.

D. HPCS-P-2 will start regardless of its discharge valve position.

ANSWER: D Post Exam Comment - Answer A wa s also accepted as correct based on a post exam comment. The reason distractor A wa s thought to be incorrect is the pump would not start until the discharge valve cl osed following the restoration of power to the bus. This would take longer than the 20 second time delay. However, the

wording of the distractor states the pump will start "after its 20 second time delay."

This is a true statement and is therefore accepted as correct. QUESTION TYPE: Closed Reference

KA # & KA VALUE: 264000 A3.06 Ab ility to monitor automatic oper ations of the EMERGENCY GENERATORS (DIESEL/JET) including: Cooling water system operation (3.1)

REFERENCE:

CGS System Description, St andby Service Water, Rev. 14; SD000204

SOURCE: New Question

LO: 7744 Describe the physical connection and/or cause-and-effect relationship between Service Water and: b. Diesel Generators

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - the dischar ge valve must stroke closed after the sequencer sequences SW on, and then start to reopen. This adds to

the normal start time. B. Incorrect - this trip will still be bypassed after the LOOP.

C. Incorrect - this would trip the DG.

D. Correct - this pump auto starts regardless of its discharge valve position.

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 39

EXAM KEY NOVEMBER 2006 Page 39 of 81 COMMENTS: Ref : 10 CFR 55.41 (8) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 40

EXAM KEY NOVEMBER 2006 Page 40 of 81 Columbia was operating at full power with TR-S t agged out of service to facilitate BPA work. A leak in containment caused drywell pressure to rise to 4 psig. All systems operated as designed except that breaker CB-B-8 failed to auto close.

If, coincident with the start of DG-2 , a Generator Overcurrent condition were to occur, which of the following is correct?

A. DG-2 would start and trip due to t he overcurrent condition. RHR-P-2B and RHR-P-2C would lose power.

B. DG-2 would tie onto and re-energize SM-8. RHR-P-2C starts and 5 seconds later, RHR-P-2B starts.

C. DG-2 would start but not tie onto S M-8 due to the overcurrent condition. RHR-P-2B and RHR-P-2C would lose power.

D. DG-2 would tie onto and re-energize SM-8. RHR-P-2B starts and 10 seconds later, RHR-P-2C would start.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 264000 K3.01 Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: Emergency core cooling systems (4.2)

REFERENCE:

CGS System Descripti on, DG Pg 18, 19, 52; SD000200

SOURCE: New Question

LO: 5313 (DG) 7772 (AC)

RATING: Knowledge: Analysis Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A & C are incorrect because this trip is bypassed with a LOCA signal present. D is incorrect because the loading sequence is not correct. COMMENTS: Ref : 10 CFR 55.41 (7) & (8) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 41

EXAM KEY NOVEMBER 2006 Page 41 of 81 Which of the following are available to cool t he Control Air Compressors during a Loss of Offsite Power? A. ONLY CJW-P-1A B. ONLY CJW-P-1B C. Fire Water AND CJW-P-1A D. Fire Water AND CJW-P-1B ANSWER: D QUESTION TYPE: Closed Reference

KA # & KA VALUE: 300000 K1.04 Knowledge of the connections and/or cause effect relationships between INSTRUMENT AIR SYSTEM and the following: Cooling water to compressor (2.8)

REFERENCE:

CGS System Description, Cont rol and Service Air System, Rev. 9; SD000205 SOURCE: New Question

LO: LO 7606 Determine the affect on the CAS from the following events: b. Loss of Offsite Power RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect, this pump is shed during a LOOP. B. Incorrect, Fire Water is available due to the connections and the DG fire pump. C. Incorrect, this CJW pump is shed during a LOOP.

D. Correct , the CJW pump is powered by the diesel on a vital load center, and fire water is available through the DG fire pump. COMMENTS: Ref : 10 CFR 55.41 (4) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 42

EXAM KEY NOVEMBER 2006 Page 42 of 81 Which of the following conditions will prevent the manual insertion of any control rod using the Reactor Manual Control System?

A. RMCS Activity Control disagree.

B. IRM downscale at 14% power.

C. RPIS malfunction at 24% power.

D. RDCS Rod bypassed.

ANSWER: A QUESTION TYPE: Closed Reference

KA # & KA VALUE: 201002 K4.04 Knowledge of REACTOR MANUAL CONTROL SYSTEM design feature(s) and/or interlocks wh ich provide for the following: Single notch rod withdrawal and insertion (3.3)

REFERENCE:

CGS System Description, RMCS, Rev. 11, pg. 15; SD000148

SOURCE: New Question

LO: 5799 State the function of the following rod motion indicators: b. Activity control disagree.

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Correct - See reference. B. Incorrect - The IRM downscale will only prevent rod withdrawal.

C. Incorrect - RPIS will only cause a rod block through RWM or RSCS below 20% power. D. Incorrect - This only prevents a single rod from inserting using RMCS. COMMENTS: Ref : 10 CFR 55.41 (6) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 43

EXAM KEY NOVEMBER 2006 Page 43 of 81 If power is lost to bus SH-5, which of the following components will lose power?

A. RRC-P-1A B. COND-P-5 C. TSW-P-1A D. CRD-P-1B ANSWER: A QUESTION TYPE: Closed Reference

KA # & KA VALUE: 202001 K2.01 Knowledge of electrical power supplies to the following: Recirculation Pumps Plant specific (3.2)

REFERENCE:

CGS System Description, AC Distribution, Rev. 13; SD000182 SOURCE: New Question

LO: 5058 Identify the loads on the following buses: e. SH5, SH6

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Correct - RRC A is powered from SH-5. B. Incorrect - It is powered from SI-63 via SH-6.

C. Incorrect - Not powered from SH-5.

D. Incorrect - Not powered from SH-5. COMMENTS: Ref : 10 CFR 55.41 (6) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 44

EXAM KEY NOVEMBER 2006 Page 44 of 81 The reactor is at 75% power when a voltage tr ansient results in the loss of one Reactor Recirculation Pump. The transient has result ed in the reactor now operating in the AIA.

Based on this event, the operator should first:

A. prevent an ASD over-frequency pump trip by placing the operating loop controller in manual.

B. place the operating loop controller in manual to minimize the potential for vibration induced jet pump damage.

C. manually adjust the operating loop flow controller to exit the AIA and preclude uncontrolled power oscillations.

D. manually adjust the master flow controlle r to exit the AIA and avoid exceeding the power to flow scram setpoint.

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 202002 A2.01 Ab ility to (a) predict the impacts of the following on the RECIRCULATION FLOW CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Recirculation Pump Trip (3.4)

REFERENCE:

SD000184, RRFC, Rev 14, page 12 of 43

SOURCE: NEW

LO: 9687 - State the conditions that will cause an individual ASD controller to automatically shift from AUTO to MANUAL.

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - The loop flow c ontroller automatically shifts to manual on the trip of one of the running pumps.

B. Incorrect - The same reason as A.

C. Correct - Given the reactor is operati ng in the AIA, the first action should be to exit the AIA using flow control.

D. Incorrect - The master flow controller cannot be used with only one

reactor recirc pump running (by interlock). COMMENTS: 10CFR55.41 (5) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 45

EXAM KEY NOVEMBER 2006 Page 45 of 81 With the reactor stable at 920 psig and 36 inches, if containment temperature were to increase from 90 to 350 degrees, the indicated level on the Upset Level Range would:

A. increase as temperature increases.

B. decrease as temperature increases.

C. increase ONLY after the calibration te mperature of 135 degrees is exceeded.

D. decrease ONLY after the calibration temperature of 135 degrees is exceeded.

ANSWER: A 6QUESTION TYPE: Closed Reference KA # & KA VALUE: 216000 K5.07 Knowledge of the operational implications of the following concepts as they apply to NUCLEAR BOILER INSTRUMENTATION: Elevated temperature effects on level indication (3.6)

REFERENCE:

CGS System Description, Nuclear Boiler Instrumentation, Rev. 9; SD000126 SOURCE: New Question LO:

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: With the reactor stable, an increase in containment temperature would cause an increase in the temperature of the reference leg. This would lower the density making the reference leg "lighter". Because level is derived by a dp cell measuring the difference in weight, the decrease in

the weight of the reference leg would cause indicated level to increase

making A the correct answer. COMMENTS: Ref : 10 CFR 55.41 (5)

Added 'as temperature increases' to A and B.

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 46

EXAM KEY NOVEMBER 2006 Page 46 of 81 The plant is at 99% power when condenser vacuum begins to lower (trending to no vacuum). If this trend continues the MSIVs will close at . Per ABN-VACUUM, the operator is required to . A. 7" Hg; ONLY TRIP the Main Turbine B. 7" Hg; SCRAM the Reactor and then trip the Main Turbine C. 8.3" Hg; ONLY TRIP the Main Turbine D. 8.3" Hg; SCRAM the Reactor and then trip the Main Turbine ANSWER: D QUESTION TYPE: Closed Reference

KA # & KA VALUE: 239001 A2.08 Ab ility to (a) predict the impacts of the following on the MAIN AND REHEAT STEAM SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low condenser vacuum (3.6)

REFERENCE:

ABN-VACUUM

SOURCE: New Question

LO:

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect

- MSIVs close at 8.3" B. Incorrect - MSIVs close at 8.3" C. Incorrect - If you trip the turbine first, then the Reactor might scram on high pressure. D. Correct- MSIVs close at 8.3, tr ipping reactor first helps limit SRV usage and automatic trips. COMMENTS: Ref : 10 CFR 55.41 (7) & (10)

Reworded stem 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 47

EXAM KEY NOVEMBER 2006 Page 47 of 81 Columbia is in the process of a start up following a refueling outage. The following are some of the normal steps in placing the turbine generator on line:

1. Throttle valve / governor valve transfer

2. Bypass valves control pressure at 920 psig

3. Bypass valves close

Which of the following gives the correct sequence for these activities?

A. 1, 2, then 3 B. 2, 1, then 3 C. 3, 1, then 2 D. 1, 3, then 2

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 241000 K1.28 Knowledge of the physical connections and/or cause- effect relationships between REACTOR/TURBINE PRESSURE REGULATING SYSTEM and the following: Reactor startup (3.2)

REFERENCE:

CGS System Description, Main Turbine, Rev. 9, pg. 39-41; SD000129

SOURCE: New Question

LO:

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect B. Correct C. Incorrect D. Incorrect COMMENTS: Ref : 10 CFR 55.41 (4) & (10)

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 48

EXAM KEY NOVEMBER 2006 Page 48 of 81 Columbia was at 90% power when a malfunction of the A Moisture Separator Reheater 2 nd stage temperature controller caused the Temper ature Control Valves to go closed.

Due to the above, final steady state reactor thermal power will be ________ the original thermal power, and the main turbine governor valv es will travel in the _________ direction.

A. the same as, open B. the same as, closed C. lower than, open D. lower than, closed ANSWER: A QUESTION TYPE: Closed Reference KA # & KA VALUE: 245000 K3.03 Knowledge of the effect that a loss or malfunction of the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS will have on following: reactor power (3.9)

REFERENCE:

CGS System Description, Columbia Simulator SOURCE: New Question LO: 7747 Determine the affects of MS R Second Stage Reheater operations on the LP Main Turbine RATING: Knowledge: Analysis Difficulty: 4

ATTACHMENT: None JUSTIFICATION: A. Correct - Rx power will be the same because there has been no change in the net reactivity for the reactor. Pressure goes up due to

more steam flow through governor valves, which drives governor

valves slightly open. B. Incorrect - See A. C. Incorrect - Power stays the same

- colder feedwater offsets higher pressure. D. Incorrect - See C. COMMENTS: Ref : 10 CFR 55.41 (1) & (4) & (5) & (14) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 49

EXAM KEY NOVEMBER 2006 Page 49 of 81 Which of the following responses is expected for the Reactor Feedwater System following a complete loss of Plant Service Water (TSW)?

A. The feedpumps will eventually aut o trip on high vibration.

B. The bearing temperatures will rise on the feedpumps.

C. The feedpumps will eventually auto tr ip on high lube oil temperature.

D. The feedpump auxiliary oil pump will auto start.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 259001 K6.06 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR FEEDWATER SYSTEM: Plant service water (2.7)

REFERENCE:

CGS System Descripti on, Feedwater, Rev. 9; SD000151

SOURCE: New Question

LO: 5768 Describe how the following system s interrelate with the Feedwater system. B. TSW

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - There is no automatic trip on high vibration. B. Correct - The oil is no longer being cooled, so the bearing temperature will rise. C. Incorrect - There is no automatic trip on high lube oil temperature D. Incorrect - This pump has a start signal on a loss of pressure, not high temperature. COMMENTS: Ref. 10 CFR 55.41 (4) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 50

EXAM KEY NOVEMBER 2006 Page 50 of 81 Columbia is in the process of a plant startup after a refueling outage. The Offgas and Air Removal systems are in the process of being placed in service.

Which of the following statements correctly descri bes the operation of the Steam Jet Air Ejector 1 st Stage Pressure Control Valve, MS-PCV-16A, when it s control switch is placed in the 'AUTO' position?

MS-PCV-16A opens when-.

A. downstream pressure is LT 50 psig.

B. 1 st stage steam flow is LT 9200 lbm/hr.

C. MS-V-12A, 2 nd stage startup steam supply closes.

D. upstream steam supply pressure is GT 120 psig.

ANSWER: D Post Exam Comment - Answer A was also determined to be correct. The distractor was believed to be incorrect because the LT 50 psig causes the valve to open if the valve controller is in standby and upstream pressure is GT 120 psig.

However, MS-PCV-16A, regulates downstr eam pressure at 120 psig so if downstream header pressure is LT 50 psig and upstream pressure is GT 120 psig, the valve should be open. Ther efore, both A and D are correct. QUESTION TYPE: Closed Reference

KA # & KA VALUE: 271000 A1.15 Ab ility to predict and/or moni tor changes in parameters associated with operating the OFFGAS SYSTEM controls including: Steam supply pressures (2.7)

REFERENCE:

CGS System Description, Ai r Removal System, Rev. 10; SD000181

SOURCE: New Question

LO: 5621 Describe the physical connec tion and/or the cause-and-effect relationship between the Offgas Proce ssing system and the following: c.

Control and Service Air system AND d. Main Steam System RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - In STDBY it opens if LT 50 psig and upstream press GT 120#. B. Incorrect - Steam flow closes AR-V-2A/B/C 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 51

EXAM KEY NOVEMBER 2006 Page 51 of 81 C. Incorrect - 2 nd stage operates independently from 1 st stage valves. D. Correct - In AUTO valve opens when upstream pressure GT 120#. COMMENTS: Ref : 10 CFR 55.41 (13) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 52

EXAM KEY NOVEMBER 2006 Page 52 of 81 Columbia was operating at 99% power with #2 Re mote Intake valves, WOA-V-51B/52B, CLOSED.

If an event occurs that causes the Reactor Buildi ng Exhaust Plenum Radiation indication to raise to 16 mr/hr and stabilize, which of the following is the resultant Control Room HVAC lineup?

A. Normal Supply Plenum valves, WOA-V-51C

/52C, OPEN; #1 Remote Air Intake valves, WOA-V-51A/52A, OPEN B. Normal Supply Plenum valves, WOA-V-51C

/52C, OPEN; #1 Remote Air Intake valves, WOA-V-51A/52A, CLOSED C. Normal Supply Plenum valves, WOA-V-51C

/52C, CLOSED; #1 Remote Air Intake valves, WOA-V-51A/52A, OPEN D. Normal Supply Plenum valves, WOA-V-51C

/52C, CLOSED; #1 Remote Air Intake valves, WOA-V-51A/52A, CLOSED

ANSWER: C QUESTION TYPE: Closed Reference

KA # & KA VALUE: 290003 K1.03 Knowledge of the physical connections and/or cause- effect relationships between CONTROL ROOM HVAC and the following: Remote air intakes: Plant Specific (2.8)

REFERENCE:

CGS System Description, Control Room HVAC, Rev. 10; SD000201

SOURCE: New Question LO: 7649 Describe the CR HVAC response system response to a FAZ signal RATING: Knowledge: Analysis Difficulty: 3 ATTACHMENT: None JUSTIFICATION: A. Incorrect - The normal suppl y plenum closes on a Z signal (13 mr/hr on RB exhaust). B. Incorrect - The normal supply plenum closes on a Z signal. C. Correct - The remote air intake (manual valves) must be open for the Control Room to pressurize. D. Incorrect - With all intakes closed, the CR will not pressurize. COMMENTS: Ref. 10 CFR 55.41 (7)

Reworded stem 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 53

EXAM KEY NOVEMBER 2006 Page 53 of 81 With the mode switch in shutdown and reactor pressu re at 135 psig, the reactor would be in Mode:

A. 2 B. 3 C. 4 D. 5 ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 2.1.22 Ability to determine the Mode of Operation (2.8)

REFERENCE:

CGS Technical Specifications Table 1.1-1

SOURCE: New Question LO:

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: Steam Table

JUSTIFICATION: A. Incorrect because mode switch is not in startup. B. Correct because mode switch in shutdown, and pressure indicates temperature above 200F. C. Incorrect because pressure indicates temperature above 200F.

D. Incorrect because pressure indicates that vessel head is off. COMMENTS: Ref : 10 CFR 55.41 (5) & (10) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 54

EXAM KEY NOVEMBER 2006 Page 54 of 81 With Columbia operating at 100% power, a lower s eal cavity (Seal No. 1) pressure of ____ psig and an upper seal cavity (Seal No. 2) pressure of ____ psig would be indicative of a degraded upper seal on a Reactor Recirculation Pump?

A. 310; 910 B. 510; 710 C. 710; 510 D. 910; 310 ANSWER: D QUESTION TYPE: Closed Reference

KA # & KA VALUE: 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretations (3.7)

REFERENCE:

CGS Procedure ABN-RRC-SEAL, Rev 4, Step 1.2.

SOURCE: New Question

LO:

RATING: Analysis/Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - The indications would be the failure of the lower seal. B. Incorrect - The indications would be the failure of the lower seal.

C. Incorrect - The indications would be the failure of the lower seal D. Correct - Upper seal pressure would drop below 510 psig for a failure of upper seal. COMMENTS: Ref : 10 CFR 55.41 (3) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 55

EXAM KEY NOVEMBER 2006 Page 55 of 81 Procedure 6.3.2, Fuel Shuffling and/or Offloading and Reloading states, "Total core flow is restricted to LE 10,000 GPM drive flow via RHR and/or RRC." This applies with fuel bundles removed from the core.

The reason for this precaution is to prevent damaging the:

A. Control Rods.

B. LPRMs.

C. Jet Pumps.

D. Refueling Equipment.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 2.2.26 Knowledge of refueling administrative requirements (2.5)

REFERENCE:

CGS Procedure 6.3.2, Fuel Shu ffling and/or Offloading and Reloading, Rev.

16, pg. 12 SOURCE: New Question

LO:

RATING: Knowledge: Analysis Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect - The Control Rods are unaffected due to support from fuel. B. Correct - See reference.

C. Incorrect - Jet pumps are adequat ely supported during refueling. D. Incorrect - Cross flow of concern should only occur in between the fuel assemblies. COMMENTS: Ref : 10 CFR 55.41 (2) & (10) 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 56

EXAM KEY NOVEMBER 2006 Page 56 of 81 Which of the following is the criteria for stopping t he fuel shuffle process in procedure 6.3.2, Fuel Shuffling and/or Offloading and Reloading?

A. Doubling in the SRM period.

B. Doubling in the average SRM count rate.

C. Doubling of any single SRM period.

D. Doubling of any single SRM count rate.

ANSWER: B QUESTION TYPE: Closed Reference

KA # & KA VALUE: 2.2.28 Knowledge of new and spent fuel movement procedures (2.6)

REFERENCE:

CGS Procedure 6.3.2, Fuel Shu ffling and/or Offloading and Reloading, Rev.

16, Section 2.2 SOURCE: New Question

LO:

RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A. Incorrect -

The reference says count rate. B. Correct - See reference.

C. Incorrect - Reference says TWO doublings of any single SRM count rate. D. Incorrect - Reference says TWO doublings of any single SRM count rate. COMMENTS: Ref: 10 CFR 55.41 (10)

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 57

EXAM KEY NOVEMBER 2006 Page 57 of 81 The reason for the automatic reactor scram a ssociated with a main turbine trip is to:

A. limit cycling of the SRVs.

B. mitigate the reactor power increase.

C. prevent a main steam line rupture.

D. minimize the wetwell heatup.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 295005AK1.01 4.0/4.1 10CFR 55.41 Knowledge of the operational implications of the following concepts as they apply to Main Turbine Generator Trip: Pressure effects on reactor power. (4.0)

REFERENCE:

SD000161; TS Bases 3.3.1.1 SOURCE: Modified

LO: 5949

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: As stated in the system descripti on, the basis for the automatic scram is to limit the pressure and corresponding power increase following the closure of the throttle valves. Additionally, the ba sis for Turbine Trip LCO states the pressure and power effects on the reactor following a trip of the Main

Turbine must be limited. B is correct.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 58

EXAM KEY NOVEMBER 2006 Page 58 of 81 The reactor was operating at 99% power when the reactor scrammed. The following conditions exist: HPCS and RCIC auto started

RCIC is maintaining level in the normal band

Drywell pressure is .92 psig

Both Reactor Recirculation Pumps have tripped

CB-RPT-3A/4A and CB-RPT-3B/4B are open

Both Reactor Feed Pumps have tripped

Which of the following caused the scram?

A. Main turbine trip B. MSIV isolation C. Reactor level + 13 inches D. Reactor pressure 1060 psig

ANSWER: A QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295005 AK2.03 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: Recirculation system (3.2)

REFERENCE:

SD000178 RRC Systems text pages 10, 23 and 24

SOURCE: Bank

LO: 5023

RATING: Knowledge: Analysi s Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: Only a Main Turbine trip opens both CB-RPT-3A/4A and CB-RPT 3B/4B. A is correct. The other three choices are scram signals but would only open

CB-RPT-3A and 3B.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 59

EXAM KEY NOVEMBER 2006 Page 59 of 81 The Control Room has been abandoned and all immediate actions have been completed.

According to ABN-CR-EVAC, the RPV must be Emergency Depressurized from the Remote Shutdown Panel when indicated RPV level reaches:

A. -147".

B. -150".

C. -161".

D. -183".

ANSWER: A QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295016 AA1.06 Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT: Reactor water level (4.0)

REFERENCE:

ABN-CR-EVAC; SD000126

SOURCE: NEW

LO: 11401

RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: ABN-CR-EVAC states ED is r equired at -147". ED is performed when RPV level drops below lowest usable indicated RPV level which is -147" on the

wide range. A is correct. Emergency Depr essurization, without control room evacuation would occur at an RPV le vel of -161" (non-ATWS) and -183" (ATWS) thus C and D are incorrect. -150" is lowest meter indication for a

wide range instrument, is not usabl e, and therefore D is incorrect. COMMENTS: Reworded stem 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 60

EXAM KEY NOVEMBER 2006 Page 60 of 81 Columbia is in a refueling outage. Fuel shuffle evolutions are on-going on the Reactor Building 606' elevation. Due to an error associated with the plac ement of spent fuel bundles in the Spent Fuel Pool the 606' Fuel Pool area criticality Moni tor, ARM-RIS-2, alarms in the Control Room.

What indications of the alarming radiation m onitor are available on the Reactor Building 606' Refueling Floor?

A. A rotating amber light only.

B. A pulsing red light only.

C. A rotating amber light and an klaxon alarm.

D. A pulsing red light and an klaxon alarm.

ANSWER: C QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295023 AA1.04 Ability to operate and/

or monitor the following as they apply to REFUELING ACCIDENTS: Radiati on Monitoring equipment. (3.4 3.7)

REFERENCE:

SD000141

SOURCE: NEW

LO: 5114

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: Per the systems text all Area Radiation Monitors have a rotating beacon.

Additionally, ARM-RIS-2 has a klax on horn associated with its alarm condition. C is correct.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 61

EXAM KEY NOVEMBER 2006 Page 61 of 81 A scram occurs with Columbia at 90% power and near the end of the operating cycle. RPV level has been recovered from -65 inches and the scram has been reset.

After resetting the scram, the reactor operator not es all RPS Group white lights are illuminated but the scram discharge volume vents and drains did not open. The reactor operator also notes the scram accumulators are not recharging.

These indications would be expected if:

A. APRM power peaked at 120 percent.

B. RPV pressure peaked at 1138 psig.

C. Drywell pressure peaked at 1.9 psig.

D. Scram Discharge Volume level peaked at the 530' elevation.

ANSWER: B QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295025 EK2.04 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: ARI/RPT/ATWS: Plant specific (3.9)

REFERENCE:

SD000142

SOURCE: Bank Slightly Modified

LO: 5189

RATING: Knowledge: Analysi s Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: ATWS/ARI Logic is initiated at a RPV pressure of 1120 and a RPV level of

-50". B is correct as it is GT 1120 ps ig. A, B, and C would cause a scram but not the initiation of ATWS ARI.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 62

EXAM KEY NOVEMBER 2006 Page 62 of 81 Which of the following would preclude the use of a wide range level instrument to report RPV level while operating in the EOPs?

A. RPV Pressure of 25 psig; Drywell Tem perature of 300°F; no erratic indications observed B. RPV Pressure of 50 psig; Drywell Te mperature of 285°F; erratic indication observed C. RPV Pressure of 75 psig; Drywell Te mperature of 330°F; erratic indication observed D. RPV Pressure of 100 psig; Drywell Te mperature of 325°F; no erratic indication observed ANSWER: C QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295028 EK2.03 Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Reac tor Water level indication (3.6)

REFERENCE:

PPM 5.0.10 RPV Saturation Temperature Curve

SOURCE: New

LO: 8488

RATING: Knowledge: Analysi s Difficulty: 2

ATTACHMENT: SATURATION TEMPER ATURE CURVE from PPM 5.0.10 JUSTIFICATION: Per PPM 5.0.10 and the RPV Satu ration Temperature Curve, the answer is C as the parameters are within the unsafe region of figure A and erratic indications are observed COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 63

EXAM KEY NOVEMBER 2006 Page 63 of 81 The plant was operating at rated power when a tr ip of both Reactor Feed Pumps occurred. RPV level dropped to -75 inches before returning to the normal operating band. A ll systems operated as designed except that coincident with the reactor scr am, the feeder breaker to MC-4A tripped open.

Given these conditions, the operator should trip DG-3-.

A. by locally closing the engine fuel oil supply valve.

B. by placing the Unit Mode Selector Switch in the "MAINT" position.

C. from the control room within 6 minutes.

D. at the local control panel immediately.

ANSWER: D QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295031 Reactor Low Water Level. 2.4.24 Knowledge of loss of cooling water procedures. (3.3 / 3.7)

REFERENCE:

ABN-SW; SD000204; SD000200

SOURCE: NEW

LO: 6760, 5835, 5837

RATING: Knowledge: Analysi s Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A RPV level of -50 inches caus es DG-3 to start. The HPCS service water pump, HPCS-P-2, is powered from MC

-4A. With the feeder opening to MC-4A, this prevents HPCS-P-2 from starting. This means DG-3 is running

without service water. Per ABN-SW, DG-3 is immediately tripped. D is

correct. C is incorrect because DG-3 cannot be tripped from the control

room. Additionally the time is incorrect for DG-3 but is correct for DG-1 and DG-2 if they were running without service water. A is incorrect because it is not per procedure. B is incorrect as it would not stop local starts of the DG. COMMENTS: Reworded stem 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 64

EXAM KEY NOVEMBER 2006 Page 64 of 81 The plant was operating at 97% power when a transi ent occurred resulting in a gaseous release.

QEDPS indicates a TEDE (Whole Body) dose that requires a General Emergency classification.

The CDE (Thyroid/Iodine) dose is only 20% of t he required General Emergency dose threshold.

Based on these conditions, the operating crew shoul d conclude that the release is from the:

A. Reactor Building with SGT in service.

B. Reactor Building with SGT not in service.

C. Turbine Building with Turbine Building HVAC in service.

D. Turbine Building with Turbine Building HVAC not in service.

ANSWER: A Post Exam Comment - The licensee recommended this question be deleted because it is beyond the scope of knowledge for an RO. This recommendation is

based on the interpretation of QEDPS not being an RO task. This

recommendation was rejected because the RO was not asked to interpret the

QEDPS data as this was done in the st em of the question. This data was provided so the RO could determine t he release was being filtered thereby eliminating distractors C and D. The only gaseous filtration system is the SGT system and the RO should be able to observe from the information provided it must be in service. Because the filt ration function is one of the SGT system's primary purposes, this is testable RO knowledge. QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295038 High Off-site Release Rate EA2.04 Ability to determine and/or interpret the following as they appl y to HIGH OFF-SITE RELEASE RATE:

Source of off-site release (4.1 4.5)

REFERENCE:

SD000144 SOURCE: Bank

LO: 5821

RATING: Knowledge: Analysi s Difficulty: 3 ATTACHMENT: None

JUSTIFICATION: A filtered release, i.e. with SGT in operation, results in a relatively low CDE (Thyroid from Iodine) dose. A pr ojected dose at the site boundary high enough for a General Emergency, but with a relatively low Thyroid dose can only be the result of a release through SGT. A is correct.

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 65

EXAM KEY NOVEMBER 2006 Page 65 of 81 COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 66

EXAM KEY NOVEMBER 2006 Page 66 of 81 The plant was operating at 99% power when a Main Turbine Trip occurred but the reactor did not scram. Direction in the EOPs is given that if SRVs are cycling, manually open SRVs until pressure drops to 945 psig.

Which of the following describes the basis for this direction?

A. Maintains reactor water inventory in the Containment.

B. Maximizes the amount of st eam condensed in the wetwell.

C. Maximizes the amount of energy directed to the main condenser.

D. Maintains pressure below the scram se tpoint and allows resetting of the scram.

ANSWER: C QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295007 AK3.04 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE: Safe ty/relief valve operation: Plant specific (4.0 4.1)

REFERENCE:

5.0.10 SOURCE: Bank (slightly modified stem and modified answer to be consistent with distractors)

LO: 8053

RATING: Knowledge: Analysi s Difficulty: 3 ATTACHMENT: None

JUSTIFICATION: SRVs are opened to stop SRVs from cycling and pressure is reduced to 945 psig which is the pressure at which steam flow through the BPVs is at 100%.

C is correct.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 67

EXAM KEY NOVEMBER 2006 Page 67 of 81 The plant was operating at 89% power when a Re circulation Suction Line break caused a High Drywell Pressure reactor scram. The High Drywe ll signal has cleared and ONLY the scram has just been reset.

Which of the following is correct concerning these conditions?

A. EDR-R-5 (Sump in the CRD Pump roo m) is filling from the scram discharge header, and pumps down based on the operati on of the Fill/Pump out Timer.

B. EDR-R-5 (Sump in the CRD Pump roo m) is filling from the scram discharge header, but does not pump down due to the isol ation of the outlet discharge valve EDR-V-395.

C. FDR-R-3 (Sump in the HPCS Pump roo m) is filling from the broken RRC Suction line and pumps down based on the operati on of the Fill/Pumpout Timer.

D. FDR-R-3 (Sump in the HPCS Pump r oom) is filling from the broken RRC Suction line, but does not pump down due to the is olation of the outlet discharge valve FDR-V-220.

ANSWER: B QUESTION TYPE: RO/SRO

KA # & KA VALUE: 295036 EA2.03 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL:

Cause of the high water level (3.4 3.8)

REFERENCE:

SD000142; SD000167; SD000173

SOURCE: Bank LO: 5475 RATING: Knowledge: Analysi s Difficulty: 3 ATTACHMENT: None

JUSTIFICATION:

C and D are incorrect because the drywell outlet valves for the floor drains isolate on the high drywell pressure and does not reopen based on the scram being reset.

A is incorrect because the sump outlet isolates on the high drywell pressure and

does not reopen based on the scram being reset. B is correct because the water in

the sump comes from the SDV and it does not pump down until the "F" signal is

reset. COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 68

EXAM KEY NOVEMBER 2006 Page 68 of 81 The following plant conditions exist follo wing an extended run at rated conditions:

Reactor level was -137 inches for the last 3 minutes and is now trending up slow

SM-7 is out of service due to ongoing maintenance

HPCS-P-1 is injecting into the core

RHR-P-2B and RHR-P-2C are not running

ADS is NOT inhibited

Which of the following describes the re sponse to a manual start of RHR-P-2C?

A. When RHR-P-2C discharge pressure is GE 125 psig, all ADS SRVs will open immediately.

B. When RHR-P-2C discharge pressure is GE 125 psig for 105 seconds, all ADS SRVs will open.

C. When the breaker for RHR-P-2C clos es, all ADS SRVs will open immediately.

D. When the breaker for RHR-P-2C cl oses, all ADS SRVs will open 105 seconds later.

ANSWER: A QUESTION TYPE: RO/SRO KA # & KA VALUE: 203000 K3.03 Knowledge of the effe ct that a loss or malfunction of the RHR/LPCI INJECTION MODE will have on following: Automatic depressurization logic (4.2 4.3)

REFERENCE:

SD000186

SOURCE: Bank - Modified stem and distractors LO: 5070

RATING: Knowledge: Analysi s Difficulty: 3 ATTACHMENT: None JUSTIFICATION:

ADS will initiate when both of the following conditions are met: 105 seconds after -

129" and RHR pressure GE 125 psig. A is correct. B is not correct because it

includes the 105 seconds that have already timed out. C and D are incorrect

because they are based on breaker closure not system pressure.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 69

EXAM KEY NOVEMBER 2006 Page 69 of 81 Average Power Range Monitor (APRM) channel "C" is powered from:

A. Critical Instrument Power Inverter IN-1.

B. 125 VDC Distribution Panel DP-S1-1A.

C. 24 VDC Distribution Panel DP-SO-A.

D. Reactor Protection System Bus "A".

ANSWER: D QUESTION TYPE: RO/SRO

KA # & KA VALUE: 215005 K2.02 Knowledge of electrical power supplies to the following: APRM channels (2.6 2.8)

REFERENCE:

SD000149

SOURCE: NEW

LO: 5096

RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: As stated in SD000149 - APRM 'C' is powered from RPS A thus D is correct.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 70

EXAM KEY NOVEMBER 2006 Page 70 of 81 RCIC has been manually started for RPV pressure c ontrol and is operating in the CST to CST mode.

If CST level decreases to 1' 6", then:

A. a RCIC turbine trip will occur on low RCIC pump suction pressure causing RCIC-V-1 to close.

B. RCIC will take a suction from the S uppression Pool and discharge back to the Suppression Pool through the full flow test line.

C. RCIC will take a suction from the S uppression Pool and discharge back to the Suppression Pool through RCIC-V-19.

D. RCIC will take a suction from the Suppr ession Pool and transfer water to the CSTs through RCIC-V-59.

ANSWER: C QUESTION TYPE: RO/SRO

KA # & KA VALUE: 217000 K4.07 Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feat ure(s) and/or interlocks which provide for the following: Alternate supplies of water (3.6 3.6)

REFERENCE:

PPM 4.601.A4-3.4 SOURCE: Bank - Slightly modified

LO: 5724

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: At a CST level of 1'10" RCI C-V-31 (SP suction) opens and RCIC-V-10 (CST suction) closes. When RCIC-V-31 is fu ll open RCIC-V-22 and V-59 (full flow test line to CST) close thus C is correct. A is not correct because one suction

valve does not close until the other is full open. B is incorrect because RCIC-V-22 and V-59 close. D is incorrect because RCIC-V-19 discharges to the SP. COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 71

EXAM KEY NOVEMBER 2006 Page 71 of 81 The Control Room Operator (CRO) manually aligned the valves for the Reactor Core Isolation Cooling (RCIC) system from t he control room and started the RCIC pump during the performance of a pump operability surveillance. RCIC is now operating in the CST to CST mode.

An Equipment Operator just repor ted the presence of a small steam leak on the RCIC turbine. The CRS has directed the CRO to secure the RCIC turbine.

The CRO depresses the RCIC "MANUAL ISOLATION" pushbutton.

In response to this action, the RCIC turbine will:

A. trip and both the inboard and outboard RCI C steam supply line isolation valves (RCIC-V-63 and RCIC-V-8) will close.

B. continue to operate normally.

C. trip and ONLY RCIC-V-63 (steam supply line inboard isolation valve) will close.

D. trip and ONLY RCIC-V-8 (steam supply line outboard isolation valve) will close.

ANSWER: B QUESTION TYPE: RO/SRO

KA # & KA VALUE: 217000 A3.01 Ab ility to monitor automatic operat ions of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) in cluding: Valve operation (3.5 3.5)

REFERENCE:

SD000180

SOURCE: Bank - modified slightly

LO: 5723

RATING: Knowledge: Analysi s Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: If an initiation signal were pres ent, the isolation P/B would close RCIC-V-8 and trip the turbine. In the stem, it is clear an initiation signal is not present thus B is the correct answer and other answers are incorrect. COMMENTS: Stem revised 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 72

EXAM KEY NOVEMBER 2006 Page 72 of 81 Which of the following describes the effect of a loss of normal 480 VAC power to UPS inverter E-IN-1? A. The Static Switch provides a make bef ore break forward transfer of loads from the normal AC source, MC-7A, to the 250 VDC battery.

B. The 250 VDC battery, which supplies the in verter in parallel with the output of the rectifier fed from MC-7A, assumes the load.

C. A break before make transfer to the Kirk Key Bypass Source, MC-7F, results in a momentary (4 millisecond) loss of power to inverter E-IN-1 loads.

D. The Static Switch provides a bumpless tr ansfer of the critical UPS loads from the inverter output, to the Bypass AC source, fed from MC-7F.

ANSWER: B QUESTION TYPE: RO/SRO

KA # & KA VALUE: 262002 K4.01 Knowledge of UNINTE RRUPTIBLE POWER SUPPLY (A. C./D.

C.) design feature(s) and/or interlocks wh ich provide for the following: Transfer from preferred power to alter nate power supplies (3.1 3.4)

REFERENCE:

ABN-INV SOURCE: Bank - Slightly modified

LO: 5891

RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: The battery is the first to assu me load on a loss of normal AC as stated in the answer B. A is incorrect because ther e is no static switch involved. C is incorrect because the Kirk Key swaps power between normal and bypass

source MC-7A. D is incorrect because static switch no involved and bypass

source is MC-7F.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 73

EXAM KEY NOVEMBER 2006 Page 73 of 81 Using the attached Electrical Wire Diagram of the Plant Service Water System, which of the following statements is correct?

A. Contact S21 / S21T is closed when the breaker for TSW-P-1A is open and would be opened when the breaker for TSW-P-1A is closed.

B. When emergency power is restored, TS W-P-1A auto starts after a 10 second time delay if Standby Pump Selector Switch is in TSW-P-1A position.

C. An undervoltage on SM-85 causes TSW-P-1A to start regardless of lube water flow if start occurs within 60 seconds of undervoltage on SM-85.

D. TSW-P-1A trips when TSW-V-53A is 15%

open in the closed direction regardless of TSW-P-1A control switch position.

ANSWER: B QUESTION TYPE: RO/SRO

KA # & KA VALUE: 400000 2.1.24 Ability to obtain and interpret station electrical and mechanical drawings (2.8 3.1)

REFERENCE:

EWD - 57E - 002

SOURCE: NEW

LO: 4047

RATING: Knowledge: Analysi s Difficulty: 2

ATTACHMENT: EWD-57E-002

JUSTIFICATION: A is incorrect because Contact is an 'a' contact and follows breaker position.

C is incorrect because lube water flow has to be normal on any pump start

regardless of any time considerations. D is incorrect because the contacts above valve at 15% contacts requires the switch to be in any position other

than Auto after start. B is correct as the pump selector switch needs to be in

the TSW-P-1A position and on the pump start a 10 second time delay is enforced by TSW-RLY-62/TSW1A.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 74

EXAM KEY NOVEMBER 2006 Page 74 of 81 A few minutes after resetting a valid reactor scram signal, the CRO notes that the red and green lights for the Scram Discharge Volume Drain Valves, are both illuminated.

Based on these indications, the CRO should conclude:

A. One drain valve is intermediate and the ot her drain valve is either intermediate or full open.

B. One drain valve is intermediate and the ot her drain valve is either intermediate or full closed.

C. The outboard drain valve is full clos ed and the inboard drain valve is full open.

D. The inboard drain valve is full clos ed and the outboard drain valve is full open.

ANSWER: A QUESTION TYPE: RO/SRO

KA # & KA VALUE: 201001 A3.10 Ab ility to monitor automatic oper ations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including: Lights and Alarms (3.0 2.9)

REFERENCE:

SD000142

SOURCE: Bank - modified stem and distractor wording

LO: 5198

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: As per SD000142; both SDV drain va lve lights illuminated indicate both drain valves are at least in the intermediate position. A is correct.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 75

EXAM KEY NOVEMBER 2006 Page 75 of 81 The control room switch for Fuel Pool Circulati ng Pump A (FPC-P-1A) is in the IR-71 position and the control switch for FPC-P-1B is in the IR-69 position.

With this switch alignment:

A. taking the local control switch for the standby FPC pump to the START position will start the pump without enfor cing the start permissives.

B. and both local control switches in t he NEUTRAL position, the standby pump only auto start if the operating FPC pum p has a low discharge pressure.

C. and the local control switch for both FP C pumps in the START position, neither pump will start if there is an 'F' or 'A' signal.

D. and the local control switch for both FP C pumps in the NEUTRAL position, if the operating FPC pump trips, the st andby FPC pump will not auto start.

ANSWER: D QUESTION TYPE: RO/SRO

KA # & KA VALUE: 233000 2.1.30 Ab ility to locate and operate components / including local controls (3.9 3.4)

REFERENCE:

SD000202

SOURCE: NEW

LO: 15308

RATING: Knowledge: Analysi s Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: A is incorrect because star t permissives are still enforced with operation from the local control panel. B is incorre ct because there is no auto feature associated with the standby pumps with C/S in NEUTRAL. C is incorrect because the FPC start logic does not look at an F or A signal. D is correct

per systems text.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 76

EXAM KEY NOVEMBER 2006 Page 76 of 81 Columbia was operating at rated power when a manual scram was initiated but control rods did not insert. PPM 5.1.1 was entered and exited to PPM 5.

1.2 due to the ATWS condition. RPV level is currently -20" and reactor power is approximately 20%. All three Circ Water pumps are in operation and the Main Turbine is on line.

In accordance with the EOPs, PPM 5.5.6, By passing MSIV Low RPV Level and High Steam Tunnel Temperature Isolation Interlocks, has been performed and RCIC-V-1 has been manually closed. The EOPs now direct that RPV level be lowered in an effort to reduce reactor power. RPV level is now -60 inches and trending down slowly.

Which of the following choices indicates the correct lineup for the above conditions?

A. CW-P-1B, CW-P-1C, and the Main Turb ine tripped at -50 inches. CW-P-1A will continued to operate.

B. CW-P-1B and CW-P-1C tripped at -50 in ches. CW-P-1A continued to operate.

The Main Turbine stayed on line.

C. All three Circ Water pumps tripped at -

50 inches. The Main Turbine will trip on loss of Main Condenser Vacuum.

D. All three Circ Water pumps continued to operate. The Main Turbine stayed on line. ANSWER: B QUESTION TYPE: RO/SRO KA # & KA VALUE: 2.1.31 Ability to locate control r oom switches, controls and indications and to determine that they are correctly reflec ting the desired plant lineup (4.2 3.9)

REFERENCE:

SD000180; SD000193

SOURCE: NEW LO: 11241 RATING: Knowledge: Analysi s Difficulty: 3 ATTACHMENT: None JUSTIFICATION:

Closing RCIC-V-1 prevents RCIC start at Level 2 and tripping off the Main Turbine.

PPM 5.5.6 does nothing to the logic for the CW Pumps. CW-P-1B and CW-P-1C

will trip at Level 2. CW-P-1A does not trip on a Level 2 signal. A is incorrect as it

indicates the MT will trip. C is incorrect because it states all 3 CW pumps will trip at

-50". D is incorrect because B and C CW pumps do trip at -50". B is correct.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 77

EXAM KEY NOVEMBER 2006 Page 77 of 81 If stroke time testing a valve clos ed, the stopwatch is started when:

A. both red and green lights are illumi nated and is stopped when the red light extinguishes.

B. the control switch is rotated to the closed position and is stopped 10 seconds after the red light extinguishes.

C. the control switch begins to be rotat ed towards the closed position and is stopped when the red light extinguishes.

D. both red and green lights are illumi nated and is stopped when the red light extinguishes.

ANSWER: C QUESTION TYPE: RO/SRO

KA # & KA VALUE: 2.2.12 Knowledge of su rveillance procedures (3.0 3.4)

REFERENCE:

OSP-RHR/IST-Q702 pr ecaution and limitation 4.7

SOURCE: NEW LO: 10776

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: Per the surveillance procedures, the stopwatch is started when the control switch is turned and stopped when the valve indicates full open or closed. C

is correct. COMMENTS: Removed wording in stem

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 78

EXAM KEY NOVEMBER 2006 Page 78 of 81 You have been directed to perform a task in an area of the plant where the radiation level is 80 mr/hr. You expect the task to last 30 minutes.

This quarter you have received 987 mrem TEDE.

Based on the dose you will receive performing this task--.

A. an Administrative Dose Hold Point will become effective when you receive an additional 13 mrem TEDE.

B. an Administrative Dose Extension shall be approval by the Plant General Manager PRIOR to beginning the task.

C. an Administrative Dose Extension sha ll be approval by the Radiation Protection Manager PRIOR to beginning the task.

D. there are no Administrative Dose Hold Points associated with the completion of this task.

ANSWER: D QUESTION TYPE: RO

KA # & KA VALUE: 2.3.4 Knowledge of the radiati on exposure limits and contamination control /

including permissible levels in excess of those authorized. (2.5 3.1)

REFERENCE:

GEN-RPP-06 SOURCE: NEW

LO: 11257 RATING: Knowledge: Fundamental Difficulty: 3

ATTACHMENT: None JUSTIFICATION: As stated in GEN-RPP-06 attach ment 8.1, an administrative dose hold point occurs for a TEDE of 2 Rem. The question will have the individual dose

exceeding 1 Rem therefore no dose hold point is applicable. D is correct. If 2

rem were exceeded an Administrative Do se Hold Point would occur. RPM approval is required. If 4 rem TEDE were to be exceeded, then the Plant General Manager's approval would be required. COMMENTS: Reword stem 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 79

EXAM KEY NOVEMBER 2006 Page 79 of 81 Your electronic dosimeter reads 0 mrem when you entered a posted radiological area. Now, after spending 10 minutes in the area, it reads 20 mrem.

Based on this information, this area should be posted as a:

A. Radiation Area.

B. High Radiation Area.

C. High High Radiation Area.

D. Locked High Radiation Area.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. (2.6 3.0)

REFERENCE:

PPM 11.2.7.1; SWP-RPP-01

SOURCE: NEW

LO: 11257

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: 20 mrem times 6 (6 - 10 minute periods in an hour) is 120 mrem which is a High Radiation Area. A high Radiation Area is posted with a sign and the

words CONTACT HP PRIOR TO ENTRY. B is correct.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 80

EXAM KEY NOVEMBER 2006 Page 80 of 81 With Columbia operating at power, heavy sm oke is quickly filling the Control Room.

According to ABN-CR-EVAC, which of the following actions must be completed prior to evacuating the Control Room?

A. If Control Rods failed to insert, initiate ARI.

B. Arm and Depress MSIV Isolation Logic Pushbuttons.

C. Have the Safe Shutdown Oper ator perform Attachment 7.1.

D. Place the Mode Switch in the 'Refuel' position.

ANSWER: B QUESTION TYPE: RO

KA # & KA VALUE: 2.4.11 Knowledge of Abno rmal Condition procedures. (3.4 3.6)

REFERENCE:

ABN-CR-EVAC

SOURCE: NEW LO: 6889

RATING: Knowledge: Analysi s Difficulty: 3

ATTACHMENT: None

JUSTIFICATION: Per ABN-CR-EVAC none are immediat e operator actions except closing the MSIVs. B is correct.

COMMENTS:

2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 81

EXAM KEY NOVEMBER 2006 Page 81 of 81 Events occur which require the Shift Manager to declare an Unusual Event.

Which of the following is correct concerning emer gency notifications for this event classification?

A. The designated individual will initiate t he NRC Event Notification System within thirty minutes after the emergency event declaration.

B. The designated Equipment Operator will initiate the NRC Event Notification System within one hour after State and Local authorities are notified.

C. The designated Reactor Operator will init iate the NRC Event Notification System within one hour after the emergency event declaration.

D. The NRC Event Notification System is not required to be activated at this event classification level.

ANSWER: C QUESTION TYPE: RO

KA # & KA VALUE: 2.4.29 Knowledge of the Emergency Plan (2.6 4.0)

REFERENCE:

PPM 13.4.1

SOURCE: NEW LO: 6176

RATING: Knowledge: Fundamental Difficulty: 2

ATTACHMENT: None

JUSTIFICATION: Per PPM 13.4.1, the NRC is notif ied within one hour of event notification. C is correct.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 1

EXAM KEY NOVEMBER 2006 Page 1 of 26 Columbia is operating with the following conditions given:

The reactor is operating at 100% power Rod line is 100% OPRMs are inoperable Due to an ASD fault, both RRC-P-1A and RRC-P-1B run back to 15 Hz.

Based on the given conditions, which is correct?

The plant would be in:

A. region A of the power to flow map. ABN-POWER would be entered and the reactor would be manually scrammed.

B. both the OPRM enabled region and the Area of Increased Awareness. ABN-POWER would be entered and control rods would be inserted per the fast shutdown sequence.

C. the OPRM enabled region but no other on the power to flow map. ABN-CORE would be entered and exit from the region would be accomplished by inserting control rods per the fast shutdown sequence.

D. region A of the power to flow map. ABN-CORE would be entered and the reactor would be manually scrammed.

ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE: 295001AA2.01 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

Power/flow map (3.5 3.8) 10CFR55.43.5

REFERENCE:

SOP-RRC-START & ABN-CORE SOURCE: NEW LO: 5022 RATING: L3 ATTACHMENT: YES - SOP-RRC-START Attachment 6.1 - Two Loop Power/Flow Map JUSTIFICATION: A and B are incorrect because ABN-POWER does not give any direction for RRC pump runback. B is also incorrect because you are in region A. C is incorrect because ABN-CORE does not direct exiting the region by inserting rods. Also you are not just in the OPRM

region. D is correct. The conditions given would leave the plant in region A. ABN-CORE would be entered and a manual scram would be inserted because the OPRMs were

inoperable.

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 2

EXAM KEY NOVEMBER 2006 Page 2 of 26 Columbia Generating Station is in Hot Shutdown. All systems are operational. The feeder breaker to the HPCS Battery Charger, HPCS-C1-1, then trips open.

Based on these conditions, the CRS should declare-..

A. affected required features inoperable immedi ately and initiate actions to restore required DC electrical power subsystem to operable status immediately.

B. HPCS system inoperable immediately, verify RCIC operable by administrative means immediately, and restore HPCS system to operable status in 14 days.

C. HPCS inoperable within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or cooldown to LE 200 °F within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

D. HPCS system inoperable immediately and restore HPCS system to operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ANSWER: B QUESTION TYPE:

SRO KA # & KA VALUE: 295004AA2.04 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: System lineups (3.2 3.3)

10CFR55.43.5

REFERENCE:

TS 3.8.4B SOURCE: NEW LO: 7657 RATING: H2 ATTACHMENT:

YES - TS 3.8.4 ; TS 3.8.5; TS 3.5.1; TS 3.5.2 JUSTIFICATION:

The plant is in Mode 3. B is correct as it uses TS 3.8.4B for Mode 1, 2 or 3 as basis for answer. B is incorrect because it uses the DC shutdown TS 3.8.5. C is incorrect because it uses the completion time for Div 1 and 2 DC systems from TS 3.8.4. D is

incorrect because it uses ECCS Shutdown and HPCS is not a required operable system. COMMENTS: Revised stem COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 3

EXAM KEY NOVEMBER 2006 Page 3 of 26 The plant experienced a transient that has resulted in the following plant conditions:

Drywell temperature is 310 degrees All control rods are fully inserted Reactor pressure is 50 psig Narrow Range indicates +7 inches and has been erratic for the last 30 minutes Shutdown Flooding Range indicates +7 inches Upset Range indicates +7 inches All other level indications have been judged unreliable Based on the above, determine the correct level indication and procedure action for these conditions.

A. RPV Level should be considered unknown and PPM 5.1.4, RPV Flooding, should be entered. B. The Upset Range indication should be used and level should be recovered using PPM 5.1.1, RPV Control.

C. The Shutdown Flooding Range indication s hould be used and RPV Level should be recovered using PPM 5.1.1, RPV Control.

D. Narrow Range indication should be used and RPV level should be recovered using PPM 5.1.1, RPV Control.

ANSWER: B Post Exam Comment - Upon grading the exams it was determined A is the correct answer instead of B. This was an administrative error made during the construction of the final exam. QUESTION TYPE:

SRO KA # & KA VALUE: 295028 EA2.03 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: Reactor water level (3.9) 10CFR55.43.5

REFERENCE:

Reference:

PPM 5.0.10 EOP Figure A SOURCE: Bank LO: 8456 RATING: H3 ATTACHMENT:

EOP Caution 1 and Figure A JUSTIFICATION: RPV Pressure vs Drywell temperature is outside the saturation limit shown in Figure A making any erratic indicator unusable (Answer D) Answer B a nd C are incorrect because the indications do not meet the minimum usable levels per Caution 1. Therefore RPV Level cannot be determined and RPV COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 4

EXAM KEY NOVEMBER 2006 Page 4 of 26 Flooding is required,. Answer A.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 5

EXAM KEY NOVEMBER 2006 Page 5 of 26 The plant is in MODE 3 with the scram reset a nd RHR-P-2B in Shutdown Cooling with the following conditions:

RHR-P-2A inoperable Reactor water level +65 inches and stable RRC-P-1A in operation at 15 Hz SW-P-1B then trips and will not restart.

If this condition exists for an extended period of time, which of the following statements is correct?

A. Due to lowering RPV level, PPM 5.1.1 RPV Control would be entered to re-establish adequate core cooling.

B. Due to rising drywell temperature, PPM 5.2.1 Primary Containment Control would be entered to lower drywell temperature.

C. Due to rising reactor pressure, ABN-RHR-SDC-LOSS would be entered to re-establish shutdown cooling.

D. Due to trip of RRC-P-1A on high motor temperature, ABN-RRC-LOSS would be entered to re-establish forced core flow.

ANSWER: C QUESTION TYPE:

SRO KA # & KA VALUE: 295021AA2.06 - Ability to determine and /or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: React or pressure (3.2 3.3) 10CFR55.43.5

REFERENCE:

ABN-RHR-SDC-LOSS SOURCE: NEW LO: 5780 b RATING: H3 ATTACHMENT:

None JUSTIFICATION:

A is incorrect because the level will go up due of heat up. B is incorrect because the loss of SW has no effect on PC temperature. C is th e correct answer because the loss of cooling will cause reactor temperature and pressure to increase until Shutdown Cooling would isolate at 125 psig. D is incorrect because the RRC pumps do not trip on high temperature.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 6

EXAM KEY NOVEMBER 2006 Page 6 of 26 The plant was operating at 98% power when a transient occurred that resulted in a Drywell Floor Downcomer sheared off 6 inches below the Drywell Floor. The following conditions exist:

Drywell pressure 32 psig and stable RPV Level -155 inches and down slow Wetwell Level 29 feet and down slow 2 Control Rods Not fully inserted ARM-RIS-13 HPCS Pump Room Pegged high at 10E4 An Emergency Depressurization shall be directed per-A. PPM 5.2.1, Primary Containment Control B. PPM 5.1.1, RPV Control C. PPM 5.1.2, RPV Control - ATWS D. PPM 5.3.1, Secondary Containment Control ANSWER: A QUESTION TYPE:

SRO KA # & KA VALUE: 295024EA2.04 - Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Suppression Chamber Pressure. (3.9 3.9)

10CFR55.43.5

REFERENCE:

PPM 5.2.1 SOURCE: NEW LO: 8341 8340 RATING: H4 ATTACHMENT:

YES - PSP Curve, PPM 5.3.1 table 24 and section S JUSTIFICATION: Due to the downcomer failure, Suppression Chamber Pressure and Drywell Pressure are equal and an ED is required because the PSP curve has been exceeded.

This makes A correct. B and C are both incorrect because neither of these

procedures requires an ED above TAF. D is incorrect because 5.3.1 requires that

there be 2 areas above MSOV prior to ED.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 7

EXAM KEY NOVEMBER 2006 Page 7 of 26 Columbia was operating at 96% power when an unplanned Reactor Feedwater transient occurred. The following conditions now exist:

Reactor level +21 inches and up slow Reactor pressure 1048 psig and up fast Reactor power 27% and stable All white scram group lights NOT illuminated MSIVs Closed Drywell pressure 1.58 psig and up Suppression Pool temperature 85°F and up Reactor Building pressure -.11 inches of water Which of the following procedures should be entered first?

A. ABN-LEVEL B. ABN-PRESSURE C. PPM 5.1.1. RPV Control D. PPM 5.1.2, RPV Control ATWS ANSWER: C QUESTION TYPE:

SRO KA # & KA VALUE:

295037 2.4.6 SCRAM Condition Present and Power above APRM Downscale or Unknown: Knowledge of symptom based EOP mitigation strategies. (3.1 / 4.0)

10CFR55.43.6

REFERENCE:

PPM 5.1.1 RPV Control, PPM 5.0.10 page 100 SOURCE: NEW LO: 8017 RATING: L3 ATTACHMENT:

None JUSTIFICATION:

C is the correct answer because with the white scram group lights out, there is a scram signal present. With power at 27%, not all control rods inserted. This

requires an entry into PPM 5.1.1 RPV Control prior to the entry into PPM 5.1.2 RPV Control ATWS. The SRO must make a choice under these conditions as to

which procedure to enter. Since EOPs take precedence over ABNs, the correct

choice would be to enter the correct EOP even though both ABN-LEVEL and ABN-

PRESSURE have entry conditions..

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 8

EXAM KEY NOVEMBER 2006 Page 8 of 26 Columbia is operating at 99% power. A failure occurs in the Radwaste Building resulting in a spill of a large amount of resin from a RWCU Demineralizer 45 minutes ago.

Reactor Power 99% WEA-RIS-14 Rad Waste Bldg. Exhaust, Low 1.8E6 cpm Based on the above conditions, the CRS should enter-A. PPM 13.1.1 and PPM 5.4.1, perform a Site Evacuation, and evacuate the Columbia River, Horn Rapids ORV Park, Ringold Fishing Area, Wahluke Hunting Area, and Schools in EPZ. B. PPM 13.1.1 and PPM 5.4.1, perform a Site Evacuation, and evacuate all sections 0-2 miles and 10 miles downwind, and shelter remaining sections.

C. PPM 5.4.1 concurrently with PPM 5.1.1 and manually scram the reactor.

D. PPM 5.4.1 and Emergency Depressurize the reactor.

ANSWER: A QUESTION TYPE:

SRO KA # & KA VALUE: 295038 2.4.44 - High Offsite release rate: Knowledge of the Emergency Plan Protective Action Recommendations. (2.1 / 4.0) 10CFR55.43.5

REFERENCE:

PPM 13.2.2 rev. 15, PPM 13.1.1, rev. 34 SOURCE: NEW LO: 8893 RATING: H4 ATTACHMENT:

YES - PPM 5.4.1, rev. 12 with entry conditions, and PPM 13.1.1, rev. 34. table 3 JUSTIFICATION: A is correct because the conditions given meet the requirements for a SAE and the actions are the automatic PARS for that EAL. B is incorrect because these actions

are for a GE, which has not been reached. C and D are both incorrect because there is no primary system discharging outside of the plant.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 9

EXAM KEY NOVEMBER 2006 Page 9 of 26 Following an Emergency Depressurization due to a coolant leak, the following conditions exist:

Drywell Temperature 275 °F Reactor Pressure 25 psig Drywell Pressure 27 psig RPV Level -162 inches and stable Wetwell Pressure 22 psig RHR-P-2A and LPCS-P-1 Injecting Wetwell Level 35 ft SM-8 Locked out Based on the above plant parameters, the CRS should enter--

A. PPM 5.2.1, Primary Containment Control, and spray the Drywell regardless of adequate core cooling.

B. PPM 5.1.1, RPV Control, determine PC Flooding is required and exit to Severe Action Guidelines (SAGs).

C. PPM 5.1.1, RPV Control, and monitor Reactor Level instruments for erroneous/erratic indications.

D. PPM 5.2.1, Primary Containment Control, and lower Suppression Pool Level to LT +2 inches utilizing SOP-RHR-SPC.

ANSWER: C QUESTION TYPE:

SRO KA # & KA VALUE: 295012 2.1.25 High Drywell Temperature:

Ability to obtain and interpret station reference materials such as graphs, monographs, and tables with contain performance data (2.8 / 3.1) 10CFR55.43.5

REFERENCE:

PPM 5.0.10 rev. 9, PPM 5.2.1 rev. 16 SOURCE: NEW LO: 4104 RATING: H3 ATTACHMENT: YES - PPM 5.2.1 - Primary Containm ent Control EOP Flowchart, RPV Saturation Temperature Curve A, PCPL Curve B, P-8, P-9, P-11, P-13 and P-14 of the PC Pressure Leg, L1 on WW level leg.

JUSTIFICATION: A incorrect - conditions do not exist which require DW sprays regardless of adequate core cooling. B incorrect - conditions do not exist which require venting the Primary Containment. C correct - the combination of DW pressure and low

reactor pressure have resulted in an entry into the Sat Curve. D incorrect because the valve lineup for lowering suppression pool isolated at 1.68 psig DW pressure.

COMMENTS: Revised stem COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 10

EXAM KEY NOVEMBER 2006 Page 10 of 26 During a reactor startup with power at 28% a rod drop accident causes a power spike and has resulted in the following plant parameters:

Reactor pressure 990 psig Reactor power 31% Reactor level 36 inches MAPRAT 0.68 MCPR 1.01 LHGR 0.27 Based on given conditions, which of the following is correct?

A. Insert all operable control rods within two hours.

B. Adjust the APRM gain within six hours.

C. Verify control rod separation criteria are met and disarm the associated Control Rod drive within two hours.

D. Restore MCPR to within the limits in two hours and reduce thermal power to LT 25%

RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ANSWER: A QUESTION TYPE:

SRO KA # & KA VALUE: 295014 AA2.05 ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION: Violation of a Safety Limit IMP 4.6 10CFR55.43.2

REFERENCE:

Tech Spec 2.1, 3.1.3, 3.2.2, 3.2.4 SOURCE: NEW LO: 10304 RATING: H2 ATTACHMENT:

YES - TS 3.1.3, 3.2.2, 3.2.4 JUSTIFICATION:

A is correct because the MCPR safety limit has been violated. B is incorrect because LHGR (MFLPD ) is LT the FRTP. C is incorrect because the action is for a stuck rod. D is incorrect because one or the other conditions would be performed, not

both. COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 11

EXAM KEY NOVEMBER 2006 Page 11 of 26 Columbia was operating at full power when a transient occurred. All required EOP actions have been completed up to this point in the event. The following plant conditions now exist:

RPV Water Level is -140 inches and steady RPV Pressure is 200 psig and down slow Drywell Pressure is 95 psig and up slow Wetwell Pressure is 91 psig and up slow Wetwell Level is 46 feet and steady Wetwell Temperature is 230°F and down slow Which of the following describes the next action the CRS should take based on the above conditions?

A. Direct performance of PPM 5.5.14 which would preclude failure of the containment and subsequent loss of systems required to maintain adequate core cooling.

B. Direct performance of PPM 5.5.15 which prevents exceeding 1 rem TEDE at the site boundary during the release.

C. Direct Emergency Depressurization per PPM 5.1.3 which would preclude the failure of the SRV Tailpipe and subsequent loss of Pressure Suppression function of the wetwell.

D. Direct Emergency Depressurization pe r PPM 5.1.3 which would preclude failure of containment by assuring that RPV blowdown does not cause PCPL to be exceeded.

ANSWER: A QUESTION TYPE:

SRO KA # & KA VALUE: 295029 EA2.01 Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Suppression pool water level (3.9/3.9) 10CFR55.43.5

REFERENCE:

PPM 5.0.10 SOURCE: NEW LO: 8040 RATING: H3 ATTACHMENT:

PPM 5.2.1 PC Pressure leg steps P12, P13, P14 & PCPL Curve; L-13 & PSP Curve, WT-5 & HCTL Curve JUSTIFICATION: PPM 5.5.14 would be performed due to WW Level. 5.0.10 defines the PCPL as the limit used to preclude containment failure and subsequent loss of the ability to maintain adequate core cooling. A is correct. B is incorrect because 5.5.15 is performed if WW/L is GT 51'. C is incorrect because emergency depressurization would have already been performed. D is incorrect because ED due to HCTL is not required. COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 12

EXAM KEY NOVEMBER 2006 Page 12 of 26 The reactor was operating at 92% power with HPCS-P-1 in operation in Full Flow Test mode, Suppression Pool to Suppression Pool. A transient has occurred which resulted in a scram and the following conditions:

Reactor Building Exhaust Plenum 12 mr/hr and stable Wetwell Level -3 inches and down slow Reactor Level 22 inches and down slow Reactor Pressure 1048 psig and up slow Control Rod 30-31 Position 24 Control Rod 15-47 Position 08 Which of the following is correct concerning these conditions?

A. HPCS-V-15 remains open, PPM 5.3.1 Secondary Containment Control and PPM 5.1.2 RPV Control ATWS are entered.

B. HPCS-V-15 closes, PPM 5.2.1 Primary Containment Control is entered, and SOP-HPCS-CST/SP is utilized for Suppression Pool level control.

C. HPCS-V-15 closes, PPM 5.3.1 Secondary Containment Control is entered and PPM 5.1.2 RPV Control ATWS are entered.

D. HPCS-V-15 remains open, PPM 5.2.1 Primary Containment Control is entered, and PPM 5.5.23 Emergency Suppression Pool Makeup is utilized for Suppression Pool level control.

ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE: 209002A2.11 Ability to (a) predict the impacts of the following on the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low suppression pool level (3.3 3.5) 10CFR55.43.5

REFERENCE:

SD000174 rev. 10 page 10 and PPM 5.2.1 SOURCE: NEW LO: 8017, 5429 RATING: H3 ATTACHMENT:

NONE JUSTIFICATION: A is incorrect because there are no entry conditi ons for PPM 5.3.1. B is incorrect because HPCS-V-15 remains open and SOP-HPCS-CST/SP is incorrect. C is incorrect because HPCS-V-15 remains open and there are no entry conditions for PPM5.3.1. D is correct

because there is no low level interlock to close HPCS-V-15 and the entry for PPM 5.2.1 on SP level is given. PPM 5.5.23 is used to refill the SP per PPM 5.2.1.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 13

EXAM KEY NOVEMBER 2006 Page 13 of 26 Columbia is operating at 65% power. The last performance of the weekly RPS Manual Scram Channel Functional Test Surveillance was completed at 1200 on October 21 st. It was discovered at 1000 on October 30 th that the next performance of this surveillance had not yet been completed.

Select the statement below which correctly describes the actions which must be taken based on the above condition.

A. The missed surveillance must be completed by 0600 on October 31 st or be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B. The completion of the surveillance, if started immediately, is within Technical Specification time requirements.

C. Manage the risk impact and complete the missed surveillance by 1000 on November 6 th. D. The missed surveillance has resulted in Columbia having to be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from the time of discovery.

ANSWER: C QUESTION TYPE:

SRO KA # & KA VALUE: 212000 A2.03 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: Surveillance Testing (3.3 3.5) 10CFR55.43.3

REFERENCE:

TS 3.3.1.1 SOURCE: NEW LO: 10301 RATING: H3 ATTACHMENT:

TS 3.3.1.1 including table ``3.3.1.1-1 and SR 3.0.3 JUSTIFICATION: B is incorrect - SR 3.0.2 allows 1.25 times 7 days from last performance which would be 0600 on October 30 (8 days and 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months ) - surv eillance is late. A is based on the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from recognition of a missed surveillance and is incorrect because if the risk is managed the surveillance can go longer than Oct. 31 st . C is correct per SR 3.0.3 which has been changed to allow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or surveillance frequency if an risk impact is performed. D is incorrect as it does not take into account TS 3.0.3.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 14

EXAM KEY NOVEMBER 2006 Page 14 of 26 The plant was operating at 90% power when a seismic event caused the following:

Reactor level -25 inches and stable Scram Group Lights not illuminated Blue Scram lights 27 are illuminated SLC-P-1A and SLC-P-1B loss of power indicated Main Generator undergoing oscillations from 450 Mwe to 1100 Mwe Based on these conditions, the CRS enters-A. PPM 5.1.2 and directs boron injection with RCIC.

B. PPM 5.1.2 and directs the closure of RCIC-V-1 to prevent a Main Turbine trip.

C. ABN-POWER and directs the start of both RRC pumps at 15 Hz to stop the Main Generator oscillations.

D. ABN-POWER and directs that control rods be inserted in reverse order of the fast shutdown sequence.

ANSWER: A QUESTION TYPE:

SRO KA # & KA VALUE: 217000 2.1.20 - RCIC - Ability to execute procedural steps (4.3 4.2) 10CFR55.43.5

REFERENCE:

PPM 5.1.2, rev. 17, step Q-11.

SOURCE: NEW LO: 11145 RATING: H2 ATTACHMENT:

YES - Q10 through Q14 of PPM 5.1.2 JUSTIFICATION:

A is correct as required by PPM 5.1.2 step Q-14. MG Oscillations are in excess of 25% thermal power. B is incorrect because PPM 5.1.2 directs the use of RCIC for boron injection. C and D are both incorrect because PPM 5.1.2 takes precedent over any direction in ABN-POWER under these conditions.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 15

EXAM KEY NOVEMBER 2006 Page 15 of 26 The plant was operating at 95% power when MS-RV-5B failed in the open position and could not be closed. Suppression Pool temperature has reached 108°F and is trending up slowly.

Based on the above conditions, the CRS should enter in order to . A. ABN-SRV and immediately reduce RRC flow to 60 mlbm/hr; limit the reactor pressure/power transient associated with the SRV closure.

B. ABN-SPC and place two loops of RHR Suppression Pool Cooling in service; limit the rate of Suppression Pool heatup.

C. PPM 5.1.1 RPV Control and place the Mode Switch in SHUTDOWN; comply with Technical Specifications.

D. PPM 5.2.1 Primary Containment Control and initiate an Emergency Depressurization; prevent exceeding the Heat Capacity Temperature Limit.

ANSWER: C QUESTION TYPE:

SRO KA # & KA VALUE: 239002A2.03 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: stuck open relief valve (4.1 4.2) 10CFR55.43.5

REFERENCE:

PPM 5.2.1 rev. 16 WW temp leg; ABN-SRV; PPM 5.0.10 SOURCE: NEW LO: 8300 RATING: L2 ATTACHMENT:

NONE JUSTIFICATION: C is correct because PPM 5.2.1 block WT-4 requires entry into PPM 5.1.1 before WW temp reaches 110°F. A is incorrect because ABN-SRV directs that action as subsequent actions, not immediate actions a nd power is reduced to 90% not flow to 60 Mlbm/hr. B is incorrect because there is no procedure ABN-SPC. D is incorrect because there is no ED required until HCTL is exceeded for temperature.

COMMENTS: Revised stem and distractors COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 16

EXAM KEY NOVEMBER 2006 Page 16 of 26 The plant is operating at 96% power when a broken coupling is discovered on SW-P-1B.

The CRS is required to declare SW-P-1B inoperable and-A. DG-2 inoperable immediately.

B. prevent DG-2 start immediately.

C. its associated ECCS pumps inoperable immediately.

D. run SW-P-1A immediately to determine its operability.

ANSWER: A QUESTION TYPE:

SRO KA # & KA VALUE:

262001 2.1.11 Knowledge of less than one hour technical specification actions statements for systems: AC Electri cal Distribution (3.0 3.8) 10CFR55.43.2

REFERENCE:

TS 3.7.1 and TS 3.8.1 SOURCE: NEW LO: 9414 RATING: H2 ATTACHMENT:

NONE JUSTIFICATION:

A is correct because TS 3.7.1 directs the cascade and TS 3.8.1 applicability requires the DG inoperability. B is incorrect because DG-3 is not associated with SW-P-1B.

C is incorrect because TS 3.7.1 make no direction for considering the ECCS pumps.

D is incorrect because, while a "common cause" determination is required there is no immediate requirement for a SW-P-1A run.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 17

EXAM KEY NOVEMBER 2006 Page 17 of 26 The plant was operating at 99% power when a failure of TR-N2 caused the loss of both SH-5 and SH-6.

Which of the following actions is correct for this condition?

Enter- A. ABN-POWER, verifies operation in Region A prior to scramming the reactor.

B. ABN-RRC-LOSS, verifies operation in Region A prior to scramming the reactor.

C. ABN-POWER and immediately scram the reactor.

D. ABN-RRC-LOSS and immediately scram the reactor.

ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE: 202001A2.04 A Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM (HPCS); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Multiple Recirculation Pump trips. (3.7 3.8)

10CFR55.43.5

REFERENCE:

ABN-RRC-LOSS rev. 1, immediate actions SOURCE: NEW LO: 6733 RATING: H2 ATTACHMENT:

NONE JUSTIFICATION:

The immediate actions for ABN-RRC-LOSS state that the plant must be scrammed if both RRC pumps trip in Modes 1 or 2. D is the only correct answer.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 18

EXAM KEY NOVEMBER 2006 Page 18 of 26 The plant is at 32% power with APRM F out of serv ice and a peripheral control rod selected on the rod select matrix. APRM B then fails upscale.

Which of the following is correct?

RBM-B is inoperable,-

A. and must be restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. and must be placed in the trip condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

C. but is not required by Tech Specs until reactor power exceeds 35%.

D. but is not required by Tech Specs because a peripheral control rod is selected.

ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE: 215002 2.1.12 - Ability to apply Tech Specs for a system. (2.9 4.0) 10CFR55.43.2

REFERENCE:

Tech Spec 3.3.2.1 Am 169 SOURCE: NEW LO: 5701 RATING: H2 ATTACHMENT:

YES - Tech Spec 3.3.2.1 and table 3.3.2.1-1 JUSTIFICATION: RBM operability is required by TS anytime rector power is GE 30% unless a peripheral control rod is selected. As stated in the stem, a peripheral control rod is selected which does not require RBM operability. D is correct.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 19

EXAM KEY NOVEMBER 2006 Page 19 of 26 With Columbia operating at 100% power, a leak in the Main Condenser has caused a reactor water chlorides to reach 250 ppb (.25 ppm).

Select the statement that correctly describes the actions to be taken for the above condition.

A. Restore conductivity to within limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

B. Be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

C. Perform an orderly unit shutdown and be in cold shutdown as rapidly as operating conditions permit.

D. If chlorides not below 200 ppb within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , reduce core flow to 60 Mlbm/hr and SCRAM the reactor per PPM 3.3.1.

ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE: 256000 2.1.34 Ability to maintain primary and secondary plant chemistry within allowable limits (2.3 2.9) 10CFR55.43.5

REFERENCE:

SWP-CHE-02 Rev.11 Page 6 and Page 10 SOURCE: NEW LO: 5013 RATING: H3 ATTACHMENT: SWP-CHE-02 Rev.11 Page 1, 6, 7, 10; LCS 1.4.1 Rev. 28 pages 1 thru 4 JUSTIFICATION :

If only TS was referenced, A would be correct. A is incorrect but a viable action per LCS 1.4.1 Table 1.4.1-1. B is incorrect but an action per LCS 1.4.1 if the required completion time for condition A is not met. C is incorrect as this action would be

required if Action Level 2 was exceeded. Conductivity exceeds Action Level 3 value which require a flow reduction and scram if not below 200 ppb within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D is correct.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 20

EXAM KEY NOVEMBER 2006 Page 20 of 26 Columbia is operating at 99% power when several crew members become sick and go home four hours prior to the end of their shift. The remaining shift complement consists of 1 Senior Reactor Operator, 2 Reactor Operator's, and 1 Equipment Operator.

Which of the following describes the Technical Specification requirements concerning this situation?

A. The required Senior Reactor Operator, Reactor Operator, and Equipment Operator positions may be vacant for a period not to exceed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provided action is taken to

replace these positions within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. With less than the required shift complement, action must be taken within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to replace the required position or be in Mode 2 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Mode 3 within the

following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C. The required Senior Reactor Operator, Reactor Operator, and Equipment Operator positions may be vacant for a period not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided immediate action is

taken to replace these positions.

D. With less than the required shift complement, action must be taken within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to replace the required position or immediately take actions to place the reactor in Mode 3.

ANSWER: C QUESTION TYPE:

SRO KA # & KA VALUE: 2.1.4 Knowledge of shift staffing requirements (2.3 3.4) 10CFR50.43.2

REFERENCE:

Tech Spec 5.2.2b SOURCE: NEW LO: 6071, 6933 RATING: H2 ATTACHMENT:

NONE JUSTIFICATION :

Per TS 5.2.2b, C is correct.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 21

EXAM KEY NOVEMBER 2006 Page 21 of 26 The plant is in MODE 5 with Refueling activities in progress on the Refuel Floor.

Which of the following is considered a core alteration which would require an SRO on the Refuel Floor?

A. Withdrawal of one SRM with the control switch from the control room.

B. Withdrawal of a control rod from a cell with no fuel.

C. Movement of an irradiated fuel bundle in the Fuel Pool.

D. Reseating of a fuel bundle in the core with the refuel mast.

ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE: 2.2.29 Knowledge of SRO fuel handling responsibilities. (1.6 3.8) 10CFR55.43.7

REFERENCE:

PPM 6.3.5 rev. 10, page 3 SOURCE: Bank, 2002 NRC Exam - slightly changed.

LO: 7699 - For a given refueling operation, determine if the evolution is a Core Alteration.

RATING: L3 ATTACHMENT:

NONE JUSTIFICATION: A , B, and C are all incorrect because they do not meet the Tech Spec/Columbia Procedural definition of a core alteration. D is correct because PPM 6.3.5 specifically states the reseating of a fuel bundle during core verification is a core

alteration.

COMMENTS: Revised stem COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 22

EXAM KEY NOVEMBER 2006 Page 22 of 26 A Temporary Modification has just been installed in the plant.

Who signs and dates the "Installation Complete" block on the TMR?

A. Operations Manager B. Minor Modifications Group Supervisor C. Design Engineer D. CRS/Shift Manager ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE: 2.2.11 Knowledge of the process for controlling temporary changes. (2.5 3.4) 10CFR55.43.3

REFERENCE:

PPM 1.3.9 Rev. 39 Step 3.2.4 SOURCE: NEW LO: 8628 SRO only RATING: L3 ATTACHMENT:

NONE JUSTIFICATION: PPM 1.3.9 Temporary Modificati ons states the CRS/Shift Manager signs the "Installation Complete" block. D is the correct answer.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 23

EXAM KEY NOVEMBER 2006 Page 23 of 26 Who is responsible to perform the final review and approval of a Planned Special Exposure.

A. Radiation Protection Manager B. Plant General Manager C. Operations Manager D. Shift Manager ANSWER: B QUESTION TYPE:

SRO KA # & KA VALUE: 2.3.2 Knowledge of facility ALARA program. (2.5 2.9) 10CFR55.43.4

REFERENCE:

GEN-RPP-08 Rev. 1 page 3 SOURCE: BANK LO00257 - 2000 NRC exam slightly modified LO: 11258 RATING: H2 ATTACHMENT:

None JUSTIFICATION:

Per GEN-RPP-08 the Plant General Manager has final review/approval. B is correct.

COMMENTS: Revised stem

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 24

EXAM KEY NOVEMBER 2006 Page 24 of 26 The plant is operating at 50% power following a forced outage. A batch of nonradioactive RCC water has to be discharged following maintenance on the system. Sample results confirmed no identifiable activity other than naturally occurring isotopes.

Who authorizes the release of this RCC water?

A. Radiation Protection Manager B. Operations Manager C. Chemistry Manager D. CRS/Shift Manager ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE: 2.3.6 Knowledge of the requirements for reviewing and approving release permits. (2.1 3.1) 10CFR50.43.4

REFERENCE:

PPM 12.2.14 R4 Page 4 SOURCE: Bank - 2001 NRC Exam - slightly modified LO: 11260 RATING: L4 ATTACHMENT:

NONE JUSTIFICATION:

Per PPM 12.2.14, the CRS/Shift Manager approves the release.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 25

EXAM KEY NOVEMBER 2006 Page 25 of 26 Columbia is operating at 80% power. A surveillance concurrent with an instrument failure causes the HPCS system to inject to the RPV. Injection is secured by overriding HPCS-V-4, the HPCS injection valve, closed

and stopping HPCS-P-1.

Which of the following is true in regards to NRC reportability?

This would be a/an-A. 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report for ECCS injection into the RPV.

B. 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report for Tech Spec required shutdown.

C. 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report for valid actuation of a system.

D. 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report for single train inoperable.

ANSWER: D QUESTION TYPE:

SRO KA # & KA VALUE:

2.4.30 Knowledge of which events related to system operation/status should be reported to outside agencies. (2.2 3.6) 10CFR55.43.5

REFERENCE:

PPM 1.10.1 Rev. 27 Pages 9 - 12, NUREG 1022 3.2.6 SOURCE: NEW LO: 6011 RATING: H3 ATTACHMENT:

PPM 1.10.1 rev. 27, page 9 - 12 ; NUREG-1022 Page 45 for 3.2.6 JUSTIFICATION:

A and C are incorrect because this condition is not a valid initiation signal. B is incorrect because this situation does not require a TS shutdown. D is correct because HPCS is a single train which is now unable to perform its safety function.

COMMENTS:

COLUMBIA GENERATING STATION SENIOR REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 26

EXAM KEY NOVEMBER 2006 Page 26 of 26 A LOCA has occurred that resulted in the following conditions:

Reactor level -138 inches and stable on the Compensated Fuel Zone Reactor level off scale low on the Wide Range Reactor Pressure 105 psig and stable Wetwell temperature 199°F and up slow Wetwell level GT 51 feet Wetwell pressure 91 psig and up fast Offsite dose rate 9 mrem/hr TEDE and 5 mrem/hr CEDE Which of the following is correct concerning these conditions?

A. Enter PPM 5.4.1, Radioactivity Release Control. The Reactor should be emergency depressurized because the Offsite Release has exceeded the Alert Classification.

B. Enter PPM 5.1.1, RPV Control. The Reactor should be emergency depressurized because the HCTL has been exceeded.

C. Enter PPM 5.2.1, Primary Containment Control. Containment should be vented through the drywell, regardless of offsite release rate, to prevent the loss of systems required for adequate core cooling.

D. Enter PPM 5.2.1, Primary Containment Control. Containment should be vented through the wetwell, regardless of offsite release rate, to prevent the loss of systems required for

adequate core cooling.

ANSWER: C QUESTION TYPE:

SRO KA # & KA VALUE:

2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency conditions. (3.0 4.0) 10CFR55.43.5

REFERENCE:

PPM 5.0.10 rev. 9, pages 89, 268, & 269 SOURCE: NEW LO: 11229 RATING: H3 ATTACHMENT:

Yes - PCPL Curve; PPM 5.2.1 P-13 & P-14; HCTL Curve; PPM 5.4.1 with entry conditions JUSTIFICATION: A and B are both incorrect because neither has exceeded the limits. C is incorrect because you are directed to vent the drywell with wetwell level GT 51 feet.

@@ Line 2: / Line 2: @@
 | number = ML063540193
 | issue date = 11/30/2006
-| title = Columbia Generation Station - 11-2006 - Final As-Given Written Exam, Exam Key
+| title = Generation Station - 11-2006 - Final As-Given Written Exam, Exam Key
-| author name = Nease R L
+| author name = Nease R
 | author affiliation = NRC/RGN-IV/DRS/OB
-| addressee name = Parrish J V
+| addressee name = Parrish J
 | addressee affiliation = Energy Northwest
 | docket = 05000397
@@ Line 160: / Line 160: @@
 ANSWER: A       QUESTION TYPE: Closed Reference
 KA # & KA VALUE: 295019 AA2.02 (AK2.03 RFW) Ab ility to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safety related instru ment air system loads (see AK2.1- AK2-19)
 (3.6)
@@ Line 260: / Line 260: @@
 quencher via the downcomers and the SR V downcomers are the only ones located in the suppression pool. Based on th is, distractor C is correct as stated.      QUESTION TYPE: Closed Reference
 KA # & KA VALUE: [New KA] 295030 EK1.01 Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Steam Condensation (3.8) [New KA]
 [KA Deleted] 295030 EK1.03 Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Heat capacity (3.8) [KA Deleted]
@@ Line 616: / Line 616: @@
 KA # & KA VALUE: 212000 A2.19 Ab ility 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: Part ial system activation (Half-SCRAM)
 (3.8)
@@ Line 1,038: / Line 1,038: @@
 Reworded stem 2006 COLUMBIA GENERATING STATION REACTOR OPERATOR WRITTEN EXAMINATION QUESTION # 47
 EXAM KEY                                          NOVEMBER 2006 Page 47 of  81 Columbia is in the process of a start up following a refueling outage. The following are some of the normal steps in placing the turbine generator on line:
 : 1. Throttle valve / governor valve transfer
 : 2. Bypass valves control pressure at 920 psig
 : 3. Bypass valves close

ML063540193: Difference between revisions

Revision as of 08:36, 13 July 2019

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