ML063070432

Oyster Creek October 2006 Exam - Final RO & SRO Written Examination with Answer Key (401-5 Format) (Folder 3)

ML063070432
Person / Time
Site:	Oyster Creek
Issue date:	09/15/2006
From:	AmerGen Energy Co
To:	D'Antonio J M Operations Branch I
Sykes, Marvin D.
References
Download: ML063070432 (183)

Text

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 1 of 481.The plant was at rated power wit h all systems normally aligned and no equipment out of service. An event occurred that resulted in a reactor scram. The following annunciato rs are noted as in alarm: 4160 STATION POWER BUS 1B -

MN BRKR 1B 86 LKOUT TRIP 4160 STATION POWER BUS 1B - BUS 1B UV 4160 STATION POWER BUS 1D - MN BRKR 1D TRIP 4160 STATION POWER BUS 1D - MN BRKR 1D 86 LKOUT TRIP 4160 STATION POWER BUS 1D - BUS 1D VOLTS LO Which of the following states the correct action for these conditions?a.Confirm DW Recirc. Fans 1-1, 1-2, and 1-3 are operating, IAW ABN-48, Loss of USS 1B2b.Confirm EDG2 is supplying Bus D, IAW ABN-36, Loss of Offsite Powerc.Restart RPS MG #2, IAW ABN-51, Loss of VMCC 1B2 d.Verify air compressor 1-2 in se rvice IAW ABN-47, Loss of USS 1B1 Answer: a Justification: The indications provided show a fault and loss of power to 4160 AC busses 1B and 1D. When 1D de-energizes, EDG2 would normally start and load

onto the bus. But, since there is a fault on Bus 1D, EDG2 does not pickup Bus

1D and Bus 1D stays de-energized. With Bus 1D de-energized, Buses 1B1, 1B2 (and VMCC 1B2) and 1B3 become de-energized.

IAW ABN-48, DW Recirc. Fans 1-1, 1-2, and 1-3 should be confirmed running (powered from Bus 1A23). These fans ar e not effected by the loss of power to Bus 1D (DW recirc. fans 1-4 and 1-5 will loose power). Answer a is correct.

ABN-36 does say to start EDG2 if it is not running or supplying power to Bus 1D.

But with a fault on Bus 1D, the EDG2 cannot close onto Bus 1D. Answer b is

incorrect.

ABN-51 says that if VMCC 1B2 is de-energized (RPS MG 2 is de-energized), then to re-power RPS from transformer 1 (which is normally powered from VMCC 1A2, and is not effected by the power lo sses in the question). Since there is no power to VMCC 1B2, RPS MG 2 cannot be started. Answer c is incorrect.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 2 of 48 USS 1B1 powers air compressors 1-2 and 1-3, and is not powered given thepower losses in the question. With no pow er to USS 1B1, air compressor 1-2 will not be running. Answer d is incorrect.

295003 AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER : Cause of partial or complete loss of A.C.

power (CFR: 43.5)

OC Learning Objective: 2621.828.0.0012 (262-10444: Describe the interlock signals and setpoints for the affected system components and expected system response including power loss or failed components.)

Cognitive Level: Comprehension or Analysis

Question Type: New

Changed answers from the original due to missing CFR as noted by NRC review.

9/8/06 Rossi reviewed. Changed 'lists' to 'sta tes' in the stem. Added 'for these conditions' in the question.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 3 of 482.Given the following plant conditions: The plant is at 80% power and is recovering to 100% after removing Recirculation Pump C from service The Feedwater and Condensate Syst ems are aligned for rated power An event occurred resulting in several annunciators, and the followingindication is noted: Bus 1A ammeter indicates 0 amps Which of the following states the condi tion of the reactor following required manual operator actions?a.The reactor was manually scr ammed due to loss of feedwater pumps, IAW ABN-17, Feedwater System Abnormal Operationb.Reactor recirculation flow was lowered to 8.5 E4 gpm, IAW ABN-2, Recirculation System Failuresc.The reactor was manually scrammed due to the loss of reactor recirculation pumps, IAW ABN-2, Recirculation System Failuresd.A manual rapid power reduction has occurred due to the loss of a feedwater pump, IAW ABN-17, Feedwater System Abnormal Operation Answer: c Handouts: None

Justification: The loss of Bus 1A resu lts in the loss of feedwater pump 1A, condensate pump 1A, and reactor recirculation pumps A and E (recirculation

pump C is also on this bus, but it was se cured in the question stem). The loss of a single feedwater or condensate pump w ould require the crew to perform a rapid power reduction IAW ABN-17. Mu ltiple feedwater or condensate pumps would require a manual scram. The loss of a single recirculation pump (in 4-loop or 5-loop operation) requires that recirculat ion flow either lower flow or to insert cam rods IAW ABN-2. For multiple pump trips (in 4/5 loop), a manual scram is

required. Answer c is the correct answer.

All other answers are incorrect since they give an inappropriate action for the given conditions. (Also see procedures

317 and 301.2) 295006 AA2.06 Ability to determine and/or interpret the following as they apply to SCRAM :

Cause of reactor SCRAM (CFR: 43.5)

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 4 of 48 OC Learning Objective: 2621.828.0.

0017 (10450: Describe and interpret procedure sections and steps for plant emergency or off-normal conditions thatinvolve this system including personnel allocations and equip ment operation in accordance with applicable ABN, EOP & EOP Support Procedures.)

Cognitive Level: Comprehension or Analysis Question Type: Modified Bank

9/8/06 Rossi reviewed. Placed parts of the initial paragraph into bullets, with minor wording changes to make more clear.

Moved part of first paragraph into a new paragraph.9/20/06 NRC comments: Added 'required' in the question.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 5 of 483.The plant was shutting down for an outage, with the follo wing conditions: RPV coolant temperature is 320° F and is cooling down Shutdown Cooling pumps A and B are in service with a total SDC system flow of 4000 GPM RBCCW flow through each of the in-service SDC heat exchangers is 1000 GPM Shutdown Cooling pump C is tagged out for repair All Reactor Recirculation pumps are running RPV water level is being maintained at 155" TAF The following annunciator came into alarm: SHUT DN CLG - PUMP B TRIP The new plant conditions are as follows: Shutdown Cooling flow has been verified at 2000 GPM Investigation shows that the SDC pump B tripped on over-current Which of the following lists the requi red action to increase RPV cooling?a.Maximize RBCCW cooling water flow through the SDC loop A heat exchanger to no more than 2000 GPM IAW procedure 309.2, Reactor Building Closed Cooling Water Systemb.Maximize SDC loop A flow to no more than 3400 GPM IAW procedure 305, Shutdown Cool ing System Operationc.Increase RPV water level to above 170" TAF IAW procedure 305, Shutdown Cooling S ystem Operationd.Initiate alternate RPV cooldown IAW ABN-3, Loss of Shutdown Cooling Answer: b Handouts: None Justification: Initial conditions indicate a partial loss of shutdown cooling flow. The RAP for the tripped SDC pump directs anot her pump be started if possible (which is not possible).

In both procedures 309.2 and 305, it sti pulates that RBCCW cannot exceed 1500 GPM through a SDC heat exchanger. Answer is a incorrect.

Procedure 305 explains how to place SDC in service. The SDC pump discharge valves are throttled to establish the desired cooldown rate. The same procedure sets a limit on SDC flow of 3400 GPM through a heat exchanger. Since current NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 6 of 48 SDC flow is only 2000 GPM, then SDC flow can be increased up to 3400 GPM.

Answer b is correct.

IAW procedure 305, with reactor recirc pumps running, RPV water level shouldbe maintained within the normal band. Raising water level is only required during

a partial SDC flow loss when no reactor re circ pumps are running. Answer c is incorrect.

Initiating alternate cooling through t he cleanup system is only performed when SDC flow is lost or cannot be established, IAW ABN-3. Answer d is incorrect.

295021 AA2.02 Ability to determine and/or interpret the following as they apply to LOSS OFSHUTDOWN COOLING: RHR/shutdown cooling system flow (CFR: 43.5)

OC Learning Objective: 2621.828.0.

0045 (02602: Identify and interpret procedures for plant emergency or off-normal situations which involve the SDC System, including personnel and equipment allocations.)

Cognitive Level: Comprehension or Analysis

Question Type: New NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 7 of 484.The plant was at 90% power in preparation for recovering a control rod that was manually scrammed for test ing purposes. An electrical grid disturbance occurred resulting in a turbine trip.

The following plant conditions exist: RPV water level is 150" TAF and st eady (up from a low of 112" TAF) RPV pressure band is 900 - 1000 psig, being controlled with Isolation Condenser B Drywell pressure is 3.2 psig and rising slowly Drywell temperature is 180° F and rising slowly Torus water temperature is 91° F and steady All control rods indicate full-in The following annunciator came into alarm: ISOL COND - SHELL B LVL HI/LO The following conditions are noted: Isolation Condenser B shell water le vel is 8' and rising slowly, and that makeup is secured Chemistry reports that their samp ling of the shell water indicates greater than the expected radionuclide concentrations for the given plant conditions Which of the following procedural actions is required?

a.Initiate containment spray in the torus cooling modeb.Close the Isolation Condenser DC valves c.Emergency Depressurize d.Isolate Isolation Condenser B Answer: d HANDOUT: EOPs

Justification: Even t hough the Primary Containm ent Control EOP has been entered, torus water temperature is below the entry condition and is stable. The first action step in the EOP says to maintain torus water temperature less than 95° F and to initiate torus cooling. The EOP Users Guide says to maintain temperature and to initiate torus coo ling as required. In this case, with temperature all ready less than 95° F and not rising, torus cooling is not required.

Answer a is incorrect.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 8 of 48 The RPV Control - No ATWS EOP directs closing of the isol ation condenser DC valves when RPV water level reaches 180". Since water level was given as 150" and steady, this action is not required. Answer b is incorrect.

An emergency depressurization IAW Radioactivity Release Control EOP is

required when a General Emergency (from off-site dose) is declared. There is noindication that this has been reached. Therefore, answer c is incorrect.One entry condition into the Radioactivity Release Control EOP (EMG-3200.12) is an isolation condenser tube leak. The indi cations of this have been provided in the question stem. The first action step in this EOP is to isolate primary systems discharging outside the primary and se condary containment (such as theisolation condenser vent lines), exc ept for systems required by EOPs. Eventhough isolation condenser B is being used IAW EOPs, there are several other RPV pressure control methods available to take the place of isolation condenser B when it gets isolated (IC A, EMRVs).

Therefore, IC B should be isolated IAWthe Radioactivity Release Control EO P, and another pressure control methodutilized. Answer d is correct.

295038 2.4.6 High Off-site Release Rate Knowledge symptom based EOP miti gation strategies. (CFR: 43.5)

OC Learning Objective: 2621.845.0.

0012 (02483: Using procedure EMG-3200.12, evaluate the technical basis fo r each step and apply this evaluation to determine the correct course of action under emergency conditions.)

Cognitive Level: Comprehension or Analysis Question Type: New

9/8/06 Rossi reviewed. Added 'procedural' into the question. Placed 'The following plant conditions exist:' from the first paragr aph into a new paragraph. Placed last stem paragraph into bullets.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 9 of 485.The plant is at rated power.

Which of the following (1) would requ ire a notification, and (2) to whom this notification must be made?a.(1) An unexpected primar y containment isolation (2) Duty Station Manager of t he isolation IAW OP-OC-106-101, Significant Event Notification and Reportingb.(1) The failure of a non-APRM LPRM, with all other LPRMs operable (2) Duty Station Manager of t he LPRM failure IAW OP-OC-106-101, Significant Event No tification and Reportingc.(1) A spurious trip of the r unning TBCCW and auto start of the standby TBCCW pump

(2) PA announcement to plant pers onnel of the started pump IAW OP-AA-104-101, Communicationsd.(1) The plant scrammed with one control rod stuck full-out (2) State and local authorities IAW EP-OC-114-100, State/Local Notifications Answer: a Handouts: None Justification: OP-OC-106-101 requires the SM notify t he Duty Station Manager of any event that proceeds in a way signi ficantly different than expected. The primary containment isolation is signif icant and was not expected. The isolation also affected DW sump and equipment leakage monitoring ability. The same

requirement is also found in OP-AA-106-101 (other examples are unexpected 1/2 scram). Also, the unexpected isolation woul d require a prompt investigation (OP-AA-106-101-1001). A prompt investigation requires DSM notification also.

Answer a is correct.

OP-AA-106-101, and OP-OC-106-101 requires the SM notify the Duty Station Manager of any forced entry into a 72-hour (or less) TS shutdown LCO. IAW TS 3.2.B.2, the loss of a si ngle LPRM does not cause any APRM to be inoperable, and thus no TS entry. Answer b is incorrect.

OP-AA-104-101 requires a PA announcement prior to staring equipment where safety is a concern or when energizing/de-energizing major electrical switchgear and busses. In answer c, the component has already started and there would be no requirement no announce the start for safe ty concerns. Answer c is incorrect.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 10 of 48 If two control rods had failed to insert to at least position 04, then an emergency plan entry is required. But with one control at any position failing to insert, the reactor is assured to be shutdown. Theref ore, state/local aut horities would not be notified. Answer d is incorrect.

295020 Inadvertent Containment Isolation2.1.14 Knowledge of system stat us criteria which require the notification of plant personnel. (CFR: 43.5)

OC Learning Objective: 2621.830.0.0005 (01638: Given a description of an event, describe the following: 1) what ca tegory the event belongs in; 2) who must be notified; 3) time limit

4) any follow-up reports.)

Cognitive Level: Memory Fundamental

Question Type: New The handout were deleted. NRC comment.

9/20/06 NRC comment: Added 'rod' to answer selection d.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 11 of 486.The plant is in SHUTDOWN due to a loss of condenser vacuum (which has NOT been restored).

The following conditions currently exist: Shutdown Cooling loop A is in serv ice, with all reactor recirculation pumps in service RPV water level is in the normal band Unit Substation 1B2 is de-energi zed for emergent maintenance (it is expected the bus will be returned to service in 30 minutes) RPV coolant temperature is 220° F and is trending down slowly The following component failure has just occurred: TE-31J (Reactor Recirculation Pu mp E suction temperature element) has failed upscale Which of the following states (1) the effect on the plant and (2) the required actions?a.(1) Shutdown Cooling Pump A has tripped due to high SDC Pump A suction temperature (2) Start SDC Pump B or C to restore SDC System flow IAW

procedure 305, Shutdown Cool ing System Operationb.(1) Shutdown Cooling Pump A has tripped due to isolation valve closure (2) Bypass the failed temperat ure element and restore SDC Pump A IAW ABN-3, Loss of Shutdown Coolingc.(1) Shutdown Cooling Pump A has tripped due to isolation valve closure (2) Initiate Alternate Shutdown Cooling Using EMRVs and Core Spray IAW ABN-3, Loss of Shutdown Coolingd.(1) Shutdown Cooling Pump A has tripped due to high SDC Pump A suction temperature

(2) Initiate Alternate RPV cool down (cleanup system letdown) perprocedure 303, Reactor Cleanup Demineralizer System Answer: b Handouts: None NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 12 of 48 Justification: A high tem perature sensed on any reactor recirculation loop (350°)

will isolate SDC. When SDC IV V-17-19 closes, all SDC pumps trip. Because it has been determined that the recirculation temperature sensor has failed, ABN-3 allows bypassing the sensor and restor ing SDC flow. Answer b is correct.

It is true that 350° F SDC suction tem perature will isolate S DC and trip the SDC pump, this is not what was given in the question. Also, SDC Pumps B and C are powered by USS 1B2, which is de-ener gized. Answer a is incorrect.

As stated, the SDC IVs close on high recirc. temperature. ABN-3 directs that the failed temperature sensor be bypassed and SDC restored in step 3.2.2. Later, in step 3.2.8, it directs alternate cooli ng with core spray and EMRVs. Since step 3.2.2 can be performed, t here would be no reason to perform step 3.2.8. Answer c is incorrect.

As stated, SDC pump trips from IV posit ion, not SDC loop temperature. Also, cleanup letdown as an alternate path is not available since the condenser is not available (given in the question stem). Answer d is incorrect.

205000 A2.05 Ability to (a) predict the impacts of the following on the SHUTDOWN COOLINGSYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

System isolation (CFR: 41.5 / 43.5)

OC Learning Objective: 2621.828.0.

0045 (02602: Identify and interpret procedures for plant emergency or off-normal situations which involve the SDC System, including personnel and equipment allocations.)

Cognitive Level: Comprehension or Analysis

Question Type: New

A bullet in the question was deleted (which provided confirming information thatSDC had isolated). This was a NRC comment.

9/8/06 Rossi reviewed. Changed 'which is still being investigated' to 'which has NOT been restored' to clarify that the condenser is not available in the stem. Placed

'The following conditions currently exist:'

from the first paragraph into a new paragraph. Modified first paragraph to read better.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 13 of 487.Tech Specs 3.4.B.3 states that if the EMRV operability for the ADS function cannot be met, then reactor pressure shall be reduced to 110 psig.Which of the following provides the basis for the 110 psig?a.A small break LOCA will NOT re sult in a primary containment temperature of 281° F, which would require emergency

depressurizationb.Core spray flow into the RPV during a small break LOCA is sufficient to ensure that peak f uel centerline temperature does not exceed 2200° Fc.There is no credible event in which an RPV over-pressure condition would challenge the reactor coolant system pressure safety limit, requiring the use of the ADS valvesd.Core spray flow into the RPV during a small break LOCA is sufficient to ensure that cladding oxidation will not exceed 0.17 times the total cladding thickness before oxidation Answer: d HANDOUT: None

Justification: The relief valves of t he ADS System enable the core spray system to provide protection against the sma ll break LOCA in the event feedwater system is not available. Under the conditi ons of a small break LOCA at high RPVpressures and no feedwater available, the ADS valves will open to depressurizethe RPV to allow core spray to inject for core cooling. At an RPV pressure of 110 psig, core spray can provide the design flow necessary to maintain adequate

core cooling. Thus if the small break LO CA occurred at or less than this pressure (110 psig), the ADS valves are not requi red to depressurize the RPV to allow core spray injection. The Emergency Core Cooling System design must meet the criteria of 10CFR50.46 (Acceptance Cr iteria for Emergency Core Cooling Systems for Light Water Nuclear Power Pl ants). Two of these criteria is that ECCS will maintain peak cladding temperat ure (not fuel tem perature) less than 2200° F, and maximum cladding oxidati on of 0.17 times the total claddingthickness before oxidation. Answer d is correct, and answer b is incorrect.

All other answers are credible but are not correct ISW TS 3.4.B.3 basis.

218000 2.2.25 (ADS)

Knowledge of bases in technical spec ifications for limiting conditions for NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 14 of 48 operations and safety limits. (CFR: 43.2)

OC Learning Objective: 2621.850.0.0090 (01658: State requirements associatedwith given areas of Technical Specifications (safety limit s, LSSS, etc.).)

Cognitive Level: Comprehension or Analysis

Question Type: New

The original question provided the TS 3.4 as a handout and the question simply asked for the basis of 110 psig in t hat TS. The handout was deleted and question stem re-written to be asked without the handout (NRC comment).

9/8/06Changed 'not' to 'NOT' in answer a.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 15 of 488.The plant was at rated power, with SERVICE AIR COMPRESSOR 2 tagged out for repair.

The following annunciators came into alarm: SERVICE AIR - RCVR 1 PRESS LO SERVICE AIR - RCVR 2/INST AIR PRESS LO SERVICE AIR - RCVR 3 PRESS LO SERVICE AIR - COMPR 1 BREAKER TRIPThe INSTR AIR SUPPLY PRESS indicator shows 82 psig and lowering.

ABN-35, Loss of Instrument Air, has been entered. The SRO directed a manual reactor scram IAW the ABN. All control rods indicate full-in Which of the following actions is correct for these conditions?a.Establish and maintain RPV wate r level 138" to 175" TAF with the LRFV in MANUAL IAW ABN-1, Reactor Scramb.Establish and maintain RPV pressure 800 - 1000 psig with the Main Turbine Bypass Valves IAW EMG-3200.01A, RPV Control -

No ATWSc.Establish and maintain RPV wate r level 138" to 175" TAF using letdown from the RPV with the Cl eanup System if required, IAW ABN-1, Reactor ScramdEstablish and maintain RPV pressure 800 - 1000 psig with Isolation Condensers IAW EM G-3200.01A, RPV Control - No ATWS Answer: d Handouts: None

Justification: The first alarm (M3a) comes in at 95 psig in service air receiver 1-1.

The second alarm (M3b) comes in at either 80 psig in service air receiver 1-2 or

85 psig in the air header. The third alarm (M3c) comes in at 85 psig in the service

air receiver 1-3. This indicates a syst em-wide air loss. As provided in the question stem, ABN-35, Loss of Instrument Air, has been entered. It directs a manual reactor scram when instrument air pr essure drops to 55 psig (or if control rods begin to drift into the core). It is apparent that the instrument air pressure continues to lower.

The feedwater flow regulating valves are air-operated and lock-up on loss of air (even though they might drift). With thes e valves locked-up, the LFRV controller NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 16 of 48 in MANUAL will not function to operate t he LFRV valves (also air opewrated).

Answer a is incorrect.

With instrument air pressure loweri ng, the air-operated outside MSIVs will auto close. With these valves closed, the T BV are no longer available to control RPV pressure. Answer b is incorrect.

ABN-1 does direct to letdown through t he Cleanup System if necessary, but with the air loss, the Cleanup System letdown path is not available (cleanup will isolate). Answer c is incorrect.

Isolation condensers are available for use (even though shell makeup must be performed locally). RPV pressure control below 1045 psig is directed by EMG-3200.01A. Answer d is correct.

300000 2.4.6 (Instrument Air)

Knowledge symptom based EOP miti gation strategies. (CFR: 43.5)

OC Learning Objective: 2621.828.0.

0043 (10450: Describe and interpret procedure sections and steps for plant emergency or off-normal conditions thatinvolve this system including personnel allocation and equipment operation in accordance with applicable ABN, SDRP , EOP and EOP support procedures and EPIPs.)Cognitive Level: Comprehension or Analysis Question Type: New

9/8/06 Rossi reviewed. Deleted: with all system s normally aligned in the stem, and re-worded the stem using panel nomenclature. Placed 'The following annunciators

came into alarm:' from the first par agraph into a new second paragraph. Modified the question for focus from 'Which of the following is correct?' to 'Which of the following actions is correct for these conditions?'

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 17 of 489.The plant was starting up after a ref uel outage. The Reactor Operator was withdrawing control rods, when the control rod position indication went dark for the contro l rod being withdrawn.

Control rod position indication was NOT regained when they inserted the control rod one notch. The Operator t hen attempted to fully insert the control rod in preparation for isolati ng the control rod. Neutron monitoring showed no change in counts as the control rod was inserted.

With the control rod valved out of service, IAW procedure 302.1 Control Rod Drive System, and control rod pos ition not known, which of thefollowing Technical Specificat ions actions is required?a.The SHUTDOWN MARGIN must be verified within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,including the effects of the unknown-positioned control rodb.The reactor must be placed in the SHUTDOWN CONDITION within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sc.Immediately initiate action to fully insert all insertable control rodsd.Verify there are no more than 8 inoperable control rods valved out of service, prior to conti nuing with control rod withdrawals Answer: a HANDOUT: None Justification: Shutdown margin is dete rmined with the strongest reactivity control rod assumed fully withdrawn and all other control rods fully inserted. But since the control rod in the question has no pos ition indication, and they are unable to verify that the control rod is fully inserted, it's pos ition is unknown. Because of this, shutdown margin must be verified with this control not fully inserted. This is required IAW TS 3.2.A.2. Answer a is correct.

Only if the SDM cannot be verified withi n the time allowed (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />), the plant must be placed in the shutdown condi tion within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> IAW TS 3.2.A.3. Answer b is incorrect.If SDM cannot be met while in REFUEL mode, then TS 3.2.A.5 requires that all control rods be fully inserted. Answer c is incorrect.

TS 3.2.B.4 allows only 6 inoperable, valv ed out of service control rods. In any event, the startup cannot continue even if this verification was made. Answer d is incorrect.

214000 2.2.22 Knowledge of limiting cond itions for operations and safety limits. (CFR: 43.2)

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 18 of 48 OC Learning Objective: 2621.828.0.0011 (10451:

Given Tech Specs, identify and explain associated actions for each secti on of Tech Specs relating to this system including personnel allocati ons and equipment operation.)

Cognitive Level: Comprehension or Analysis

Question Type: New

The TS 3.2 handout was deleted. NRC comment.

9/8/06 Rossi reviewed. Re-wrote the stem for simplicity and brevity.

9/20/06 NRC comment: Added 'IAW procedur e 302.1 Control Rod Drive System' in the question.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 19 of 4810.The plant is high on Range 9 of the Intermediate Range Monitors during a startup.Which of the following lists the on-dut y shift requirements IAW Technical Specifications:a.1 Shift Manager 3 licensed Nuclear Plant Operators

2 licensed or non-licensed Nuclear Plant Operators

1 Shift Technical Advisorb.1 Unit Supervisor 3 licensed Nuclear Plant Operators

2 licensed or non-licensed Nuclear Plant Operators

1 Shift Technical Advisorc.1 Shift Manager 2 licensed Nuclear Plant Operators

3 licensed or non-licensed Nuclear Plant Operatorsd.1 Shift Manager 2 licensed Nuclear Plant Operators

3 licensed or non-licensed Nuclear Plant Operators

1 Shift Technical Advisor Answer: d HANDOUT: None - TS 6.0 NOT provided.

Justification: TS 6.2.2.2.a requires t he following: 1 SM, 2 licensed Nuclear PlantOperators, 3 licensed or non-licensed Nuclear Plant Operators, 1 Shift Technical Advisor (STA is not required in shutdo wn or refuel with the reactor < 212° F, according to TS 6.2.2.2.h.). Answer d is correct. All other answers are incorrect.

Conduct of Operations

2.1.4 Knowledge

of Shi ft Staffing requirements.

OC Learning Objective: 2621.850.0.0090 (01658: State requirements associatedwith given areas of Technical Specifications (Safety Limits, LSSS, etc.)

Cognitive Level: Comprehension or Analysis

Question Type: Bank NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 20 of 4811.Which of the following proposed plant changes would require a 10CFR50.59 Evaluation and NRC approval prior to procedural implementation? (neglect any effect s of the proposed changes on the Updated Safety Analysis Report)a.Changing the Reactor Building Vent Radiation Monitors upscale trip setting to 20 mR/hrb.Changing the RPV heat up/cooldown rates to 16° F/10 minutesc.Changing the requirement to use the Rod Worth Minimizer in the Low Power Mode until 15% power on a startupd.Changing the Scram Discharge Volume Hi-Hi setpoint to 25 gallons Answer: a HANDOUT: None

Justification: IAW 10CRF50.59(c)(1)I: A licensee may make changes in thefacility, without obtaining a license amendm ent, as long as a tech spec change is not required. A 50.59 review would need to be performed for a change that would require a change in Tech Specs. (See LS-AA-104). A change in Tech Specs

must first be approved by the NRC.

TS Table 3.1.1.j requires the hi setpoint of RB vent radiation monitors be set at <

17 mR/hr (currently set at 9 mR/hr (see RAP-10F1f)). Changing this setpoint above the TS value (to 20 mR/hr) would first need a Tech Spec change, preceded by a 50.59 review and evaluation. Answer a is correct.

TS 3.3.c.1 allows a 100° F/hr limit on heatup/cooldown rate. Startup andshutdown procedures limit this to 95° F/hr. Setting this limit to 96° F/hr (which equals 16° F/10 minites) is still less t han the TS requirement and thus does not violate TS. Answer b is incorrect.

TS 3.2.B.2 requires the RWM be operabl e on a startup up t0 10% power (and this is reflected in procedure 409, O peration of the Rod Worth Minimizer).

Requiring the RWM to be operable up to 15%

power on a startup does not effect Tech Specs. No NRC notification woul d be required. Answer c is incorrect.

TS Table 3.1.1.a requires a scram signal from < 29 gallons in the SDV, and is currently set at 26 gallons (se RAP-H1b). Setting this limit at 25 gallons is more conservative and does not violate TS. Answer d is incorrect.

Equipment Control

2.2.9 Knowledge

of the process for det ermining if the proposed change / test or experiment increases the probability of occurrence or consequences of an accident during the change

/ test or experiment.(CFR: 43.3)

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 21 of 48 OC Learning Objective: 2621.830.0.0005 (02618:

State what actions an operator and supervisor make to initiate a procedure change.)

Cognitive Level: Comprehension or Analysis

Question Type: Bank The handouts were deleted. NRC comment. Answer d was changed so that the new setpoint given was less that the ex pected known actual scram setpoint (not TS scram setpoint) for scram discharge volume.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 22 of 4812.The plant is at rated power. An NL O was required to manipulate a manual valve (located at floor level, and r equires no tools to manipulate) in a Locked High Radiation Area (LHRA). This area has a peak dose rate of

1050 mr/hr, and is routinely surv eyed by Radiation Protection.

IAW RP-AA-460, Controls for High and Very High Radiation Areas, which of the following steps are required by the Operator (besides signing onto the appropriate RWP)?1.Review the currently available survey data for the area2.Receive a briefing from the RP Tech 3.Ensure that the RP Tech accompanies you into the LHRA 4.Verify the maximum dose rate with your electronic dosimetry 5.When leaving the area, notify RP to second check you that the access is closed and lockeda.1, 2, and 3b.2, 4, and 5 c.1, 3, 4, and 5 d.1, 2, and 5 Answer: d Justification: RP-AA-460, Controls for High and Very High Radiation Areas (section 4.8), the following ar e required: 1) review survey data (it does allow RP Tech to accompany the worker into the ar ea, if there is no current survey data);

2) review/sign RWP; 3) receive an RP brief; 4) When exitin g, wait there and notify RP so that they can come and veri fy the gate/door is closed and locked.

Answer d is correct and all other answers are incorrect.

Verifying maximum dose rates is not a responsibility of the worker.

Therefore, items 1, 2 and 5 are required. Answer d is correct.

Radiation Control 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure. (CFR: 43.4)

OC Learning Objective:

Cognitive Level: Comprehension or Analysis

Question Type: New

9/8/06 NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 23 of 48 Rossi reviewed. Deleted: 'with all syst ems normally aligned' in the stem, and added 'rated'. Simplified the re maining stem for brevity.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 24 of 4813.The reactor was at rated power w hen an event occurred. Present plant conditions are as follows: Annunciator SCRAM CONTACTOR OPEN is in alarm All red scram lights are ON Annunciator ARI INITIATED is in alarm RPV water level indicates 120" TAF and rising slowly Drywell pressure is 2.2 psig and rising very slowly Drywell temperature is 170° F and rising very slowly Torus water temperature is 100° F and rising All reactor Recirculation Pump s DRIVE MOTOR switches are green-flagged (switch semaphore indicates green) Annunciators EMRV OPEN and SV/EMRV NOT CLOSED are in alarm Annunciator APRM DNSCL is NOT in alarm Annunciator ROPS BYPASSED is in alarm Which of the following states the next required operator action?a.Initiate drywell sprays IAW EMG-3200-02, Primar y Containment Controlb.Pull the open EMRV control fu ses IAW ABN-40, Stuck Open EMRVc.Perform scram reset and scram IAW EMG-3200-01B, RPV Control

- With ATWSd.Vent the scram air header IAW EMG-3200-01B, RPV Control -

With ATWS Answer: c HANDOUT: EOPs

Justification: The indications provided show that an electromatic relief valve (EMRV) is open (EMRV open and not closed alarms) and that a reactor

scrammed occurred (scram contactor open alarm and scram lights on). It also shows that the reactor is not shutdown and that power is greater than 4% (APRM downscale alarm not in), and alternate rod insertion (ARI initiated alarm) has been initiated.

Answer a is incorrect because even though drywell sprays coul d be initiated now in the drywell temperature leg, temperature is far aw ay from 281° F, and other actions are of higher priority. Answer a is incorrect.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 25 of 48 Actions to close the EMRV should be pe rformed in conjunction with EOP actions.

But pulling control fuses to close the EMRV is not an action in ABN-40. Answer b is incorrect.The initial conditions show that reacto r overfill protection (ROPS) is bypassed and that all reactor recirculation pumps have been manually tripped (green-flagged switches). The next action in RPV Control - With ATWS is to insertcontrol rods given a hydraulic ATWS ex ists (since all red scram lights are on, then all scram valves have opened and the ATWS is not electric). A possible method to insert control rods is to reset the scram, allow the scram discharge volume time to drain, and to scr am again. Answer c is correct.

Venting the scram air header (performed for an electric ATWS) will not help in inserting control rods. Answer d is incorrect.

Emergency Procedures/Plan 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.(CFR: 43.5)

OC Learning Objective: 2621-845.0.0005 (03060: During a walkthrough on the BPT or BWR simulator, demonstrate the ability to shutdown the reactor during a failure to scram situation in a timely manner IAW EMG-3200.01B, Support Procedure 21.)

Cognitive Level: Comprehension or Analysis Question Type: New

9/20/06 NRC comment: Changed 'lists' to 'states' in the question.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 26 of 4814.The reactor is at 23% power and the turbine generator has just been placed on-line. The Reactor Operator is raising reactor power by

withdrawing control rods.

Which one of the following is co rrect regarding turbine generator operation?a.A turbine vibration of 15 m ils (and trending up) requires an immediate reactor scram and turb ine trip IAW TURBINE MECH -

VIBRATION HI annunciator response procedureb.A turbine vibration of 15 m ils (and trending up) requires an immediate turbine trip ONLY IAW TURBINE MECH - VIBRATION

HI annunciator response procedurec.A loss of both stator cooling water pumps requires an immediate reactor scram and turbine trip IAW ABN-11, Loss of Generator Stator Coolingd.A loss of both stator cooling water pumps requires an immediate turbine trip ONLY IAW ABN-11, Lo ss of Generator Stator Cooling Answer: b Handouts: None

Justification: IAW RAP-Q3b (vibration high), an immediate turbine trip is required if any turbine bearing reaches 12 mils or above (and continues to increase).

Since reactor power is less than 30%, a r eactor scram is not required. Answer a is incorrect and answer b is correct.

IAW ABN-11, if a turbine runback occurs (as a result of the loss of stator cooling) or stator temperatures are rising AND reactor power is less than 30%, then generator MVARs should be m anually reduced to zero, or as low as allowed by grid conditions. At 24% power, a runba ck is not expected anyway because power is so low. No scram nor turbine trip are required. Answers c and d are incorrect.

295005 G2.4.49 Main Turbine Generator Trip / Emergency Procedures /Plan:

Ability to perform without reference to procedures those actions that require immediate operation of system co mponents and controls. (CFR: 43.5)

OC Learning Objective: 2621.828.0.0050, Objective S:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation in accordance wi th applicable ABN, SDRP, EOP & EOP support procedures and EPIPs.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 27 of 48 Cognitive Level: Comprehension or Analysis Question Type: Modified Bank

References:

RAP-Q3b, ABN-10

9/8/06 Rossi reviewed. Added 'regarding turbine generator operation' into the question for focus.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 28 of 4815.The Control Room has been evacuated due to Control Room fire. ABN-30, Control Room Evacuation, is being executed. Following the evacuation, the following conditions exist: The REACTOR MODE SELECTOR switch is in SHUTDOWN and all control rods verified full-in RPV water level is steady at 150" and adequate core cooling is assured RPV pressure is 900 psig and lowering All Core Spray Pumps and all EMRV's have been disabled IAW ABN-30 Based on the conditions given, which of the following actions must be metto comply with Technical Specifications?

1. Reduce RPV pressure to < 110 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. Place the reactor in CO LD SHUTDOWN within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.a.1 ONLYb.2 ONLY c.1 and 2d.Neither 1 or 2 Answer: c Handouts: Tech Spec 3.4

Justification: A, B and D are in correct - both 1 and 2 must be met.

C is correct - both Tech Spec action statements must be met-with the EMRV's disabled (IAW Attachment ABN-30-8) the ADS function is also disabled. Tech Spec 3.4.B (ADS) requires reactor pressu re to be reduced to less than 110 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if ADS operabi lity requirements are not me

t. Table 3.4.1 (CoreSpray) allows reduced Core Spray capability, provided several things are met:

one is that the RPV be maintained < 212° F (currently at 900 psig). Since the

requirements of the Table cannot be met, then 3.4.A.2 applies: place in Cold Shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Answer c is correct.

The NRC first thought that this question was a K/A mismatch. After discussion, they realized that it was a match and the question remained as-is.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 29 of 48 9/8/06 Rossi reviewed. He had a different answer than key. Spoke with Busk. He suggested changing the question since t he Isolation Condenser Tech Specs, under the question conditions, are not clear. Question has been modified to that shown. Busk re-reviewed.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 30 of 4816.An ATWS has occurred and all control rods have been inserted IAW Support Procedure 21.

The following plant conditions currently exist: Reactor pressure is 900 psig Reactor water level is 10 inches Torus water level is 173 inches Torus water temperature is 160 F What action is required for these conditions?a.Reduce RPV pressure to prevent exceeding TLLb.Reduce RPV pressure to prevent exceeding HCTL c.Emergency Depressurize due to exceeding HCTL d.Initiate Standby Liquid Control due to exceeding BIIT Answer: c Handouts: EMG-3200.01A, EMG-3200.02 (P rovide large figures for easier reading)Justification: A is incorrect - this would be a viable answer if HCTL was not already exceeded.B is incorrect - HCTL has already been exceeded-it's too late to lower pressure to "prevent exceeding HCTL."C is correct - HCTL has been exceeded-Emergency Depressurization is required by EMG-3200.02, Prim ary Containment Control.

D is incorrect - all rods have been insert ed-there is no requirement (or need) to initiate SLC relative to BIIT.

295026 EA2.01 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool water temperature (CFR: 43.5)

OC Learning Objective: 2621.828.0.0032, Objective J:

Identify and interpret normal, abnormal and Emergency Operating Procedures for the Primary Containment System.

Cognitive Level: Comprehension or Analysis

Question Type: Bank NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 31 of 48 9/8/06 Rossi reviewed. Changed the initial 2 sentences into one by adding 'and' between them. Placed the following into a new paragraph: 'The following plant

conditions currently exist:'

References:

EMG-3200.01A, EM G-3200.02, EOP Users Guide NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 32 of 4817.A turbine trip occurred while operating at rated power due to a loss of turbine operating oil. Current plant conditions are as follows: Reactor pressure is 1020 psig and rising slowly Reactor water level is 150 inches and rising slowly Aux flash tank pressure on Panel 7F indicates zero Which of the following should be used to control reactor pressure?a.EMRV's ONLYb.Isolation Condensers ONLY c.EMRV's and/or Isolation Condensersd.Main Turbine Bypass Valves Answer: c Handouts: None

Justification: A is incorrect. EMRVs c an be used under the given conditions, but it is not the only system available. The main turbine bypass valves are not

available (due to loss of condenser vac uum as indicated by aux flash tank pressure). With RPV level below 160 in ches, the Isolation Condensers are alsoavailable, for "stabilizing RPV pressure below 1045 psig" as directed by ABN-1.

The other options given in ABN-1 (R WCU and IC tube side vents) are not practical for the given conditions.

B is incorrect. As stated above, the EMRV s are also available. There is nothing in the question stem that makes Isolat ion Condensers unavailable. An RPV water level above 160" would make the ICs unavailable.

C is correct since both the EMRVs and the ICs are available for pressure control and can be used.

D is incorrect - the main turbine bypa ss valves are not available without main condenser vacuum above 10" Hg (Vacuum Trip 2 trips the bypass valves at 10"

Hg). Aux Flash Tank pressure at zero indicates main condenser vacuum is at

zero.295007 AA2.03 Ability to determine and/or inte rpret the following as they apply to HIGH REACTOR PRESSURE: reac tor water level. (CFR: 43.5)

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 33 of 48 OC Learning Objective: 2621.828.0.0037, Objective N:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation IAW appl icable ABN, EOP & EO P support procedures and EPIPs.Cognitive Level: Comprehension or Analysis

Question Type: New

References:

ABN-1, 307 NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 34 of 4818.Refueling is in progress when an acci dent occurs on the refuel floor. Thefollowing conditions exist five minutes later: Refueling has been suspended and the refuel floor has been evacuated ALL refuel floor radiation monitors indicate between 70 and 90 mR/hr on Panel 2R Reactor Building ventilation exhaust radiation monitors indicate 3 mR/hr on Panel 2R Reactor Building differential pressure is negative 0.25 inches WG Which of the following describes how the Reactor Building Ventilation System (RBVS) and Standby Gas Treat ment System (SGTS) should be operated during this event.a.RBVS is in service and should remain in serviceb.SGTS is in service and should remain in service c.RBVS is in service; SGTS should be placed in service d.SGTS is in service; RBVS should be placed in service Answer: d Handouts: EMG-3200.11 Justification: Under the given conditions, the high radiation le vels on the refuel floor have auto started SGTS and isolat ed normal RB ventilation (see RAP-10F4m).A is incorrect - RBVS is tripped and isolated; SGTS is in service.

B is incorrect - SGTS is in service but the conditions for placing RBVS back in service are met, as directed by t he Secondary Containment Control EOP.C is incorrect - RBVS is tripped and isolated; SGTS is in service.

D is correct - RBVS isolated on Hi Refuel Floor radiation (> 50 mR/hr w/ 2 minute time delay in either the spent fuel pool area or on the operating floor).

This also caused SGTS to initiate, main taining RB negative differential pressure.

The Secondary Containment Control EOP directs placing the RBVS in service IF it has isolated or is shutdown, AND t he drywell is not being vented through the RB supply fans (the drywell is open du ring refueling), AND RB ventilationexhaust radiation level is below 9 mR/hr, OR RB pressure is above 0 inches WG and a ground level release is imminent or in progress. Since all of theseconditions are met, RBVS should be placed back in service. This will require

overriding interlocks (Hi Refuel Floor r adiation initiation signals), resetting and NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 35 of 48 restarting RB Ventilation as direct ed by Support Procedure 50, which also causes SGTS to shutdown.

295034 EA2.01 Ability to determine and/or interpret the following as they apply to SECONDARYCONTAINMENT VENTILATION HIGH RADIATION: Ventilation radiation levels (CFR: 43.5)

OC Learning Objective: 2621.828.0.0042, Objective F:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2621.828.0.0042, Objective M:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operations IAW applicable ABN, SDRP, EOP and EOP support procedures and EPIPs.

Cognitive Level: Comprehension or Analysis

Question Type: New

References:

RAP-10F1f, RAP-10F3 m, EMG-3200.11, EOP Users Guide NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 36 of 4819.Given the following: A plant startup is in progress The reactor mode switch is in STARTUP SRM 22 is bypassed due to a failed detector IRM's 11-16 and 18 are indicating 10% on Range 8 IRM 17 is indicating 75% on Range 7 SRM 23 experiences an INOP condit ion due to a power supply failure Which statement below describes how this impacts the reactor startup?1. A withdraw rod block ____________.2. The reactor startup ____________.a.(1) is generated (2) can continue because IRM 17 can be switched to Range 8b.(1) is generated (2) CANNOT continue because the rodblock cannot be bypassedc.(1) is NOT generated (2) can continue because only two SRM's are required to be operable during a reactor startupd.(1) is NOT generated (2) CANNOT continue because more than two SRM's are required to be operable during a reactor startup Answer: a Handouts: None

Justification: A is correct - a rod block is generated due to the SRM 23 INOP condition and not ALL of the correlating IRMs (15, 16, 17 and 18) are on or

above Range 8. Since IRM 17 is at the top of the 25-75% band, it can be switched to Range 8, as directed by Pr ocedures 201 and 402.3. This will bypass the SRM 23 rod block, allowing c ontrol rod withdrawal to continue.

B is incorrect - the first statement is co rrect. However, while it is true that one SRM is already bypassed, switching IR M 17 to range 8 automatically bypassesall SRM rod block functions, which will al low control rod withdrawal to continue.

This is allowed (direct ed) by Procedures 201 and 402.3.

C is incorrect - the first statement is incorrect because a withdraw rod block IS generated. The second statement is correc t, although insignificant for the given conditions.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 37 of 48 D is incorrect - the first statement is incorrect because a withdraw rod block IS generated. The second statement is also incorrect in that Procedure 201 only requires two operable SRM's during a reac tor startup (until all IRM's are on Range 8 or above).

215004 A2.02 Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the c onsequences of those abnormal conditions or operations: SRM inop condition (CFR: 43.5)

OC Learning Objective: 2621.828.0.0029, Objective F:

Describe the interlock signals and setpoi nts for the affect ed system components and expected system response includi ng power loss of failed components.

2621.828.0.0029, Objective G:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2621.828.0.0029, Objective I:

Given normal operating procedures and documents for the system, describe or interpret the procedural steps. [Describe and interpret procedure sections or

steps and documents, under normal operat ing conditions, that involve this system.] [200s, 300s, 400s, 800s]

Cognitive Level: Comprehension or Analysis Question Type: New

References:

201, 401.4, 402.

3, RAP-G4d, GE 237E912 NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 38 of 4820.The reactor is operating at ra ted power when the SV/EMRV OPENannunciator goes into alarm. Related indications are as follows: Generator MW is slightly lower than at shift turnover Drywell pressure is 1.2 psig and steady, which is the same as at shift turnover Drywell temperature is 132 F and steady, which is 2 F higher than at shift turnover Torus water temperature is 78 F and steady, which is the same as at shift turnover The Acoustic Monitor for Safety Valve NR-28J is in the red "Valve Open Region" Tailpipe temperature for NR-28J is 245 F; all others are reading approximately 130 F This indicates ____(1)____. The corre ct action to take for this is to

____(2)____.a.(1) a Safety Valve is open (2) enter ABN-1, Reactor Scramb.(1) a Safety Valve is open (2) enter ABN-40, Stuck Open EMRVc.(1) a Safety Valve is leaking (2) commence an immediate plant shutdownd.(1) a Safety Valve is leaking (2) write an IR for Engineering to evaluate Answer: d Handouts: None

Justification: A and B are incorrect - safety valve NR-28J is leaking. If it were open, tailpipe temperature would be higher than 275 F and drywell pressure would be rising. Entering ABN-1, Reacto r Scram, would be the correct action to take if this were the case; ABN-40, Stuck Open EMRV, does not contain any guidance for an open safety valve.C is incorrect - safety valve NR-28J is leaking, but not enough to warrant an immediate plant shutdown.

D is correct - based on the given conditions, this is the appropriate action to take

for the leaking safety valve.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 39 of 48 239002 A2.02 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: Leaky SRV (CFR: 43.5)

OC Learning Objective: 2621.

828.0.0005, Objective K.2:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation IAW applicable ABN, SDRP, EOP & EOP support procedures and EPIPs.

Cognitive Level: Comprehension or Analysis

Question Type: Modified Bank

References:

RAP-B4g, ABN-40 NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 40 of 4821.The reactor was operating at rated pow er when a loss of all off-site power occurred. The following conditions currently exist: All control rods are inserted to or beyond position '04' RPV water level is 105 inches and slowly lowering Reactor pressure is being controlled at 900-1000 psig Power was restored to Bus C two minutes ago Emergency Diesel Generat or #2 failed to start It is necessary to maximize CRD flow to restore and maintain RPV water level How can this be accomplished given the current plant conditions?

a.Control CRD flow using the in-service Flow Control Valve NC-30, IAW Support Procedure 3b.Close charging header supply V-15-52, lineup and throttle CRD bypass flow at less than or equal to 150 gpm IAW Support

Procedure 3c.Cross-tie USS 1A2 to 1B2 and start a second CRD pump IAW ABN-36, Loss of Off-Site Power. Then control CRD flow using the in-service Flow Control Valve NC-30 IAW Support Procedure 3d.Cross-tie USS 1A2 to 1B2 and start a second CRD pump IAW ABN-36, Loss of Off-Site Power. Then close charging header supply V-15-52, lineup and throttle CRD bypass flow as necessary

to restore RPV water level IAW Support Procedure 3 Answer: b Handouts: None Justification: A is incorrect - the scram is not reset (since RPV level is 105 inches). Support Procedure 3 directs cont rolling CRD flow using the in-service Flow Control Valve NC-30, only if the scram is reset.

B is correct - in cases where the sc ram cannot be reset, Support Procedure 3 directs closing V-15-52, then opening V-15-237 (CRD bypass isolation) and throttling V-15-20 (CRD bypass) so as not to exceed 150 gpm (for one or two pump operation). A CAUTION in SP 3 states "Operating one CRD pump at greater than 150 gpm may re sult in a pump trip."

C and D are incorrect - USS 1A2 and 1B2 cannot be cross-tied when reactor temperature is above 212 F, EXCEPT during station blackout conditions, as directed by ABN-37, Station Blackout.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 41 of 48 201001 A2.03 Ability to (a) predict the impacts of the following on the CONTROL ROD DRIVE HYDRAULIC SYSTEM; and (b) based on thos e predictions, use procedures to correct, control, or mitigate the cons equences of those abnormal conditions or operations: Power supply failures (CFR: 43.5)

OC Learning Objective: 2621.828.0.0011, Objective 16:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation IAW appl icable ABN, EOP & EO P support procedures and EPIPs.Cognitive Level: Comprehension or Analysis

Question Type: New

References:

ABN-36, Support Procedure 3 NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 42 of 4822.Given the following: An ATWS is in progress with the MSIVs closed The conditions required by RPV Control - With ATWS for re-opening the MSIVs have been met Differential pressure across the MSIVs is 280 psid Which one of the following describes the operating restrictions for opening the MSIVs under these conditions?

Shift Manager approval __(1)__ requir ed. Damage to downstream piping

__(2)__ occur.a.(1) IS (2) MAYb.(1) IS NOT (2) MAYc.(1) IS(2) WILLd.(1) IS NOT(2) WILL Answer: b Handouts: None

Justification: A is incorrect

- SM approval is NOT required.

B is correct - Procedure 301.1 specifies t he following restrictions for opening the MSIVs: d/p 100 psid - no damage to piping; no approvals required 100 < d/p 160 psid - approaching limit where damage may occur; SM approval required d/p 160 psid - repeated opening may damage piping; open IAW EOP's. d/p 360 psig - single opening will damage piping; open IAW EOP's. C is incorrect - SM approval is not required; damage to downstream piping MAY occur.D is incorrect - damage to downstream piping MAY occur.

239001 G2.1.32 Main and Reheat Steam Syst em / Conduct of O perations: Ability to explain and apply system lim its and precautions. (CFR: 43.5)

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 43 of 48 OC Learning Objective: 2621.828.0.0026, Objective S:

Identify and interpret normal, abnormal, and emergency operating procedures forthe various steam systems and explain the reasons for the applicable

precautions and limitations.

Cognitive Level: Memory of Fundamental

Question Type: New

References:

301.1, EMG-3200.01B NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 44 of 4823.A small-break LOCA occurred while at rated power. Current plant conditions are as follows: Reactor water level is 92 inches Reactor pressure is 500 psig Drywell pressure is 13 psig Drywell bulk temperature is 450 °F TR-IA55 on Panel 8R readings are as follows:

o Point 40 = 452 °F o Point 41 = 453 °F o Point 42 = 451 °F o Point 43 = 448 °F o Point 44 = 446 °F Which of the following RPV water level instruments can be used to determine RPV water level?a.WR GEMAC ONLYb.YARWAY A & B ONLY c.NR GEMAC A & B ONLY d.NR GEMAC A & B and YARWAY A & B ONLY Answer: b Handouts: EMG-3200.02, Support Procedure 28 Justification: A is incorrect - at 451 °F and 92 inches, the WR GEMAC instrument is in the UNSAFE region of the graph in Support Procedure 28.

B is correct - at 448 °F (446 °F) and 92 inches, both Yarway instruments are in the SAFE region of the graph in Supp ort Procedure 28. None of the other instruments are in the SAFE regi on of their respective graph.

C is incorrect - at 452 °F (453 °F) and 92 inches, both NR instruments are in the UNSAFE region of the gr aph in Support Procedure 28.

D is incorrect - both NR instruments ar e in the UNSAFE region. Only the Yarway instruments are in the SAFE region.

G2.1.25 Conduct of Operations: Ability to obtain and interpret station reference materials such as graphs / monographs

/ and tables which contain performance data. (CFR: 43.5)

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 45 of 48 OC Learning Objective: 2621.828.0.0032, Objective T:

Given a set of plant conditions, interpre t Control Room and/or local Primary Containment System indications and eval uate them in terms of limits and trends using available data.Cognitive Level: Comprehensive or Analysis

Question Type: New

References:

EMG-3200.02, Support procedure 28 NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 46 of 4824.Which of the following is NOT consider ed a refueling error, as specified in Procedure 205.0, Reactor Refueling?

When loading fuel into the RPV, a ____(1)____ fuel assembly, discovered upon unlatching and lifting the grappl e, and ____(2)____ north/south movement of the bridge.a.(1) mislocated (2) afterb.(1) mislocated (2) prior toc.(1) misoriented (2) afterd.(1) misoriented (2) prior to Answer: d Handouts: None

Justification: A and B are incorrect - any mislocated fuel assembly is considered a refueling error.

C is incorrect - a misoriented fuel assembly that is not discovered until after north/south movement of the bridge is considered a refueling error.

D is correct - as stated in 205.0, "if a misoriented fuel assembly is discovered upon unlatching and lifting the grapple, and pr ior to north/south movement of the bridge, then reorient the fuel assembly per the Fuel Move Sheet immediately without declaring a refueling error."

G2.2.31 Equipment Control: Knowledge of procedures and limitat ions involved in initial core loading. (CFR: 43.7)

OC Learning Objective: 2621.812.0.0003, Objective K:

Describe, in general, refueling and f uel handling procedures to include precautions and limitations per Procedure 205 series.

Cognitive Level: Memory of Fundamental

Question Type: New

References:

205.0 NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 47 of 4825.Given the following: The plant is operating at 25% power A fire is reported and confirmed in the Reactor Water Cleanup Cage area Several minutes later, t he following parameters are noted: RPV pressure has lowered MWe has lowered Torus temperature is rising Which of the following is the correct action?a.Initiate torus cooling while maintaining reactor power constantb.Initiate torus cooling while r educing recirc. flow to minimumc.Scram the reactor and execute ABN-1, Reactor Scram d.Locally trip the recirc. pumps whose controls will be affected by the fire Answer: c Handouts: ABN-29, Attachm ent 29-1 in its entirety.

Justification:

The indications provided show that an EMRV has opened (see ABN-40) during the fire event on RB 51' (the cleanup cage is located on RB 51').

IAW ABN-29, if there is a fire on RB 51' and spurious operation of an EMRV, then the action is to scram the reactor and to close the open EMRV. Answer c is correct.Answer a and b are incorrect - it does not direct a scram, even though torus cooling may appropriate soon.

Answer d is incorrect - this would be an appropriate action if the fire was on elevation 23. Since the fire is on elevat ion 51, this would be an incorrect answer G2.4.27 Emergency Procedures/Plan: Knowl edge of fire in the plant procedure.(CFR: 43.5)

OC Learning Objective: 2621.828.0.0019, Objective E:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation in accordance wi th applicable ABN, SDRP, EOP and EOP support procedures, and EPIPs.

NRC Exam 2006-1 Senior Reactor Operator KeyNRC SRO Exam 2006-1 Key Rev 3Page 48 of 48Cognitive Level: Comprehensive or Analysis Question Type: Bank

References:

ABN-29, ABN-40 The NRC initially wanted the handout delet ed. We felt the question was not appropriate without the handout. The entir e attachment to the ABN will be provided.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 1 of 1351.The plant was at 65% power. A malfunction occurred in the master recirculation controller which caused recirculation flow and reactor power to lower. The Reactor Operator has taken all recirculation speed

controllers to MANUAL and the flow/power reduction has ceased. The

following conditions exist: Reactor power is 42% and steady Reactor recirculation flow is 6 x 10 4 GPM Which of the following actions are required?a.Manually scram the reactorb.Raise reactor recirculation flow or insert control rods c.Lower recirculation flow or insert control rods d.Raise recirculation flow or withdraw control rods Answer: bHANDOUT: OC POWER OPERATION CURVE (from 202.1)

Justification: The master controller malfunction has placed the plant in theExclusion Zone of the Power Operation Cu rve. IAW 202.1, Power Operation, theExclusion Zone is a region where reacto r operation is not allowed due to stability concerns. If the zone is entered inadvertent ly, then exit the zone by using rods or flow (the same wording is also used in 301.2, Reactor Recirculation System).

ABN-2, Recirculation System Failures, prov ides more detailed instruction if theexclusion zone is entered and how to exit the zone: exit the exclusion zone byraising pump speed and/or inserting the CRAM array (control rods). Therefore, answer b is correct.

Answer a is incorrect since scram is not the appropriate action. Answers c and d are also incorrect responses.

295001 AK1.04 Knowledge of the operational implications of the followi ng concepts as they applyto PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :

fLimiting cycle oscillation (CFR: 41.8 to 41.10)

OC Learning Objective: 2621.828.0.

0040 (00226: Identify and interpret procedures for plant emergency/off-normal situations which involve the Recirc.

System, including personnel and equipment allocations.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 2 of 135 Cognitive Level: Comprehension or Analysis Question Type: Bank

8/15/06: NRC Comments Verified that the plant conditions did not impose a control rod block. We lowered the resultant power from from 45% to 42% since the power/flow conditionsoriginally given was close to a rod block (but was not imposed).

9/8/06 Rossi reviewed. Deleted: following a shor t forced outage, and restoration of rated power is underway.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 3 of 1352.The plant was at rated power, when the following annunciator came into alarm: VITAL POWER DC PWR LOST - BUS A/B UV With the affected DC Bus at 0 volts, and in accordance with the applicableRAP, the Unit Supervisor has declared the following valves

INOPERABLE: V-16-2, Inlet Isolation Va lve to Cleanup Auxiliary Pump V-16-14, Cleanup System Inlet Isolation Valve V-14-31, Steam Inlet Valve to "A" Emergency Condenser V-14-34, Emergency Condenser NE01A Condensate Return ValveWhich of the following automatic actions should have occurred as a result of this event?a.125 VDC DC-D transfers to 125 VDC DC Ab.125 VDC DC-1 transfers to 125 VDC DC C c.125 VDC DC-E transfers to 125 VDC DC B d.125 VDC DC-2 transfers to 125 VDC DC A answer: a Justification: The given annunciator, al ong with the inoperable valves are enough to determine that 125 VDC Bus B is the effected DC bus. When voltage is lost to the bus, 125 VDC DC-D will automatically transfer from 125 VDC Bus B to 125VDC Bus A. Answer a is correct.

125 VDC DC-1, also normally supplied by Bus B, also transfers to Bus A, not Bus C. Answer b is incorrect.

125 VDC DC-E is normally powered from 125 VDC Bus A and is unaffected by the event. DC-E would transfer to Bus B on a loss of volts to Bus A. Answer c is incorrect.

125 VDC DC-2, powered from 125 VDC Bus C is unaffected by the event.

Answer d is incorrect.

References:

RAP-9XF1d, revision 2; BR 3000, Electrical Power System Key One Line Diagram, revision 8; ABN-54, DC Bus B and Panel/MCC Failures, revision

1.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 4 of 135 295004 AK1.02 Knowledge of the operational implications of the followi ng concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : Redundant D.C. powersupplies: Plant-Specific (CFR: 41.8 to 41.10)

OC Learning Objective: 2621.828.0.0012 (01121: State potential consequences on plant operation, plant equipment and environment due to failure of DC Electrical System.)Cognitive Level: Comprehensive or Analysis

Question Type: Bank

9/8/06 Rossi reviewed. Deleted: with all system s normally aligned in question stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 5 of 1353.The plant was at rated power when the following annunciators came into alarm over a short period of time: TURBINE VAC/SEALS - COND VAC LO 25 INCHES MAIN STEAM - COND VAC LO/TURB TRIP I and II TURBINE VAC/SEALS - COND VAC TRIP 1 22 INCHES TURBINE VAC/SEALS - COND VAC TRIP 2 10 INCHES Condenser vacuum continues to degrade. The following conditions currently exist: RPV water level lowered to 130" and has recovered to 170", and is stable All control rods indicate full-in Which of the following systems will be used for RPV pressure control?a.EMRVsb.Isolation Condensersc.Turbine Bypass Valves d.Isolation Condenser Vent answer: a Justification: The turbine bypass valv es are not available due to the loss of condenser vacuum (RAP-Q1c). Answer c is incorrect.

Isolation Condenser vents are unavail able since condenser vacuum is lost (Support Procedure 15 of EMG-3200.01A r equires the condenser to be intact).

Answer d is incorrect.

EMRVs are allowed IAW Support Proc edure 12 of EMG-3200-01A. Answer a is correct. (There is no indication that toru s water is too low that would preclude their use.)

The use of Isolation Condensers is prohi bited due to RPV water level of 170".

Both EMG-3200.01A and ABN-1 require R PV water level less than 160". Answer b is incorrect.

295005 AK2.07 Knowledge of the interrelations bet ween MAIN TURBINE GENERATOR TRIP and the following: Reactor Pressure Control (CFR: 41.7)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 6 of 135 OC Learning Objective: 2621.845.0.0004 (03012: Utilize appropriate EOP Support Procedures to determine various parameters required to support operation under the SBEOPs.)

Cognitive Level: Comprehension or Analysis

Question Type: New

8/15/06: NRC Comments Added that condenser vacuum continued to degrade.

9/8/06 Rossi reviewed. Re-ordered answer sele ctions from short-long. Changed the answer and justification to match new selections.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 7 of 1354.The reactor was at rated power when the Shift Manager declared the control room NOT habitable due to a to xic substance release, and that a control room evacuation is required.

Prior to leaving the control room , the following actions were taken: The reactor was scrammed and isolated The turbine was tripped Isolation Condenser B was placed into service.

While at the Remote Shutdown Panel, you have recorded the following RPV pressures:Time (hhmm)RPV Pressure (psig) 11001000 1110895 Which of the following is correct regarding the RPV cooldown rate (assume the cooldown rate does not change)?

The RPV cooldown rate is...........

Allowed by procedure 203, Plant Shutdown Allowed by Tech Specsa.Less than allowedLess than allowedb.Greater than allowedLess than allowedc.Greater than allowedEqual to allowedd.Greater than allowedG reater than allowed answer: a HANDOUT: ATTACHMENT ABN-30-4 Justification: Procedure 203, Plant Shutdown (step 6.66) requires that normal cooldown rate be limited to < 15°/10 mi nute interval = 90° F/hour. TS 3.3.C.1 limits the cooldown rate to 100° F/hour.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 8 of 135 From Attachment ABN 30-4, 1000 psig = 546.22° F, and 895 psig = 533.29° F (must interpolate). This gives a tem perature change of 12.

93° in a 10 minute period (which equals 77.58° F/hour). This cooldown rate is less than procedure 203, and less that TS 3.3.C.1. Answer a is correct. All other answers are incorrect.

295016 AA2.06 Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT : Cooldown Rate (CFR: 41.10)

OC Learning Objective: 2621.828.0.0064 (10445: Given a set of systemindications or data, evaluate and interp ret them to determine limits, trends and system status.)Cognitive Level: Comprehensive or Analysis

Question Type: New

8/15/06: NRC Comments Added to assume that the cooldown rate does not change.

9/8/06Rossi reviewed. Added "release" in stem, and changed answer selections to 3-

column format.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 9 of 1355.The reactor is at rated power.

Which of the following would require an entry into Technical Specifications if instrument air were lost to the listed system/component?a.Feedwater Control Systemb.Scram Discharge Volume c.Reactor Recirculation System d.Shutdown Cooling System Answer: b HANDOUT: None Justification: IAW ABN-35, a loss of air to the feedwater flow control system, theflow control valves will lockup, and ma y be manually controlled locally. There is no TS for these valves. Answer a is incorrect.

A loss of air to the scram discharge volume results in the vent/drain valves failing closed. TS 4.2.G requires that the SDV vent/drain valves verified open at least once per 31 days. Answer b is correct.

A loss of air to the reactor recirculation system results in the lockup of the fluid couplers, and can be manually controlled lo cally. A loss of air to the shutdown system results in the minimum flow valves failing open. There are no TS associated with recirculation pumps in manual control nor with an shutdowncooling loop. Answers c and d are incorrect.

295019 2.1.33Ability to recognize indications fo r system operating par ameters which are entry-level conditions for technical specifications. (Partial or complete loss of instrument air) (CFR: 43.2

/ 43.3) (CFR 41.7 - this is my tie to RO CFR)

OC Learning Objective: 2621.850.0.0090 (01661: Using the Tech Specs, determine if the LCO requirements ar e/are not being met and determine the appropriate plant/operator response and state the basis for response.)Cognitive Level: Comprehensive or Analysis

Question Type: New

8/15/06: NRC Comments Deleted handout TS 4.2.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 10 of 1356.While at the controls during a fuel shuffle, you are notified that an irradiated fuel bundle was dropped while being moved over the core.

Which of the following would be an expected radiation monitoring response from this event, if the des ign basis release were to occur?a.119 elevation radiation monitor C10 will indicate elevated radiationlevels, and when tripped high, will initiate the Standby Gas

Treatment System (after a time delay)b.119 elevation radiation monitor C5 will indicate elevated radiationlevels, and when tripped high, will isolate the DW vent/purge valves (after a time delay)c.119 elevation radiation monitor C9 will indicate elevated radiationlevels, and when tripped high, will initiate the Standby Gas

Treatment System (after a time delay)d.119 elevation radiation monitor B9 will indicate elevated radiationlevels, and when tripped high, will isolate the DW vent/purge valves (after a time delay)

Answer: c Justification: All the list ed radiation monitors measur e radiation levels on the refuel floor (119'). Only rad monitors C9 and B9, when tripped high, will initiate SGT after a short time delay. Rad monitors C5 and C10 initiate no protective

actions. The containment high range radi ation monitors (CHRRM), when tripped high, will isolate DW/Torus vent and purge valves. Answer c is correct. All other answers are incorrect. (see RAP-10F 1m, -10F2m, -10F3m, -10F4m, and -

10F4k.)295023 AA1.04 Ability to operate and/or monitor the following as they apply to REFUELINGACCIDENTS : Radiation Moni toring Equipment (CFR: 41.7)

OC Learning Objective: 2621.828.0.033A (00819: State any automatic actions initiated by the ARM System. State which monitors provide these actions and the setpoints.)Cognitive Level: Comprehensive or Analysis

Question Type: Modified

8/15/06: NRC Comments Added if the DB release were to occur.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 11 of 1357.The reactor was at rated power when an RPV over-pressure event occurred. One electromatic relie f valve (EMRV) opened momentarily as designed.While the EMRV was open, which of the following is correct? (select one from each part) Control Room panel EMRV position indicating lights are a(n) 1 (direct/indirect) indication of EMRV position; EMRV tailpipe temperature indication is a(n) 2 (direct/indirect)indication of EMRV position.a.(1) direct (2) directb.(1) indirect (2) directc.(1) direct (2) indirectd.(1) indirect (2) indirect Answer: b Justification: The control room panel EM RV position indicating lights show the position of the EMRV pilot valve positi on - not the EMRV. This is an indirectindication of the actual EMRV positi on. The EMRV tailpiece temperatureindicators indicate temperature in t he EMRV tailpiece. Only when the EMRV is open, will there be elevated te mperatures in the tailpiece. This provides directindication of the EMRV positi on. (See RAP-B3g, -B4g, ABN-40) 295025 EK1.03 Knowledge of the operational implications of the followi ng concepts as they apply to HIGH REACTOR PRESSURE :

Safety/relief valve tailpipe temperature/pressure relati onships (CFR: 41.8 to 41.10)

OC Learning Objective: 2621.828.0.0026 (00538: Describe Control Room and/or local steam system indications.)Cognitive Level: Comprehensive or Analysis

Question Type: New NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 12 of 135 8/15/06: NRC Comments Some question was raised over this question regarding the meaning of direct/indirect. Left as-is.

9/8/06 Rossi reviewed. Changed tailpiece to tailpipe.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 13 of 1358.In accordance with the Primary C ontainment Contro l EOP, EMG-3200.02, before bulk drywell temper ature reaches 281° F, drywell sprays are lined-up and initiated.

Which of the following states why c ontainment sprays are initiated before the bulk drywell temperature reaches 281° F?

Spraying the drywell wil l ensure that the--a.environmental qualificat ion temperature of t he EMRV solenoids is not exceededb.environmental qualificat ion temperature of t he drywell/torus vent and purge valve solenoids is not exceededc.drywell design temperature of 281° F at a design internal drywell pressure of 48 psig is not exceededd.drywell design temperature of 281° F at a design internal drywell pressure of 35 psig is not exceeded Answer: d Justification: A drywell te mperature of 281° F is the drywell design temperature at 35 psig (see USAR 6.2.1.3.5 and 3.8.

2.3.b.2 and EOP Users Guide, 2000-BAS-3200.02). As stated on page 2-24, the EQ temperatur e of safety relatedequipment is only slightly above this tem perature. The drywell design is 292° F at 44 psig and 281° F at 35 psig. Answer d is correct. All other answers are

incorrect.

295028 EK3.03 Knowledge of the reasons for the following responses as they apply to HIGHDRYWELL TEMPERATURE : Drywell Spray Operation (CFR: 41.5 / 45.6)

OC Learning Objective: 2621.845.0.

0007 (03000: Using procedure EMG-3200.02, evaluate the technical bases foe each step in the procedure and apply this evaluation to determine correct courses of action under emergency conditions.)

Cognitive Level: Memory or Fundamental

Question Type: Bank

8/15/06: NRC Comments NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 14 of 135 They commented if answers c and d were RO knowledge components. Lesson plan on Primary Containment (2621.828.0.0032) did have an objective to state the PC temperature and pressure limits.

9/8/06 Rossi reviewed. Changed 'lists' to 'states' in the question.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 15 of 1359.The reactor was at rated power, when the following annunciators came into alarm: REACTOR LEVEL - RX LVL LO I REACTOR LEVEL - RX LVL LO II Which of the following states (1) w here the Feedwater Control System will control RPV water level in AUTO (pri or to any Operator actions), and (2) the procedurally required manual operator actions to control RPV water level?Feedwater Control System Actiona.Will control RPV water level at the pre-scram level setpoint Trip two feedwater pumps when RPV water level begins to riseb.Will control RPV water level at the post-scram level setdown level

setpoint Trip two feedwater pumps when RPV water level begins to risec.Will control RPV water level at the post-scram level setdown level

setpoint Trip two feedwater pumps whenRPV water level reaches 140"d.Will control RPV water level at the pre-scram level setpoint Trip two feedwater pumps whenRPV water level reaches 140" Answer: b Justification: RAP-H5e and -H6e (RX LVL LO) require if a scram occurs, to verify actuation of the post scram level set down and to perform followup actions of ABN-1. (SP-2 of RPV Control - No ATWS also says the same correct answer.)

Following a scram and lowering RPV water level, feedwater level control will attempt to control RPV water level at t he reactor level setdown setpoint (142")(when feedwater level control is left in AUTO). ABN-1, Reactor Scram, requires that when RPV water level begins to ri se, to trip two feedwater pumps. When RPV water level reaches 140", to plac e the main feed regulating valves inmanual and close. Answer b has both correct components and is correct. All other answers provide the incorrect set point after the scram or the incorrect operator actions. (See also MDD-OC-625-B.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 16 of 135 295031 (Reactor Low Water Level)2.4.31 Knowledge of annunciators alarms and indications / and use of the

response instructions.

OC Learning Objective: 2621.828.0.0018 (10446: Identify and explain system operating controls/indications un der all plant operating conditions.)Cognitive Level: Comprehensive or Analysis

Question Type: Modified

9/8/06 Rossi reviewed. Deleted:

with all systems normally ali gned from the stem. Placed answer selections in 3-column format.

Added RPV Control - N0 ATWS Support Procedure 2 to justification. Changed feedw ater control to Feedwater Control System in the question.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 17 of 13510.The plant was at rated power when an event occurred resulting in anairborne radiological release outside of the plant structures. The current conditions exist: All control rods indicate full-in A radiological release is in-progress Which of the following states how and why the control room HVAC system should be aligned?a.System A must be run in the PURGE Mode, to remove contaminated air from the Cont rol Room, utilizing the fan onlyb.System B must be run in the PURGE Mode, to remove contaminated air from the Cont rol Room, utilizing the fan onlyc.System A must be run in the PART RECIRC Mode to maintain a positive pressure in the Control Roomd.System B must be run in the FULL RECIRC Mode to minimize the use of outside air into the Control Room Answer: c Justification: There are no automatic actions of t he control room ventilation system from any high radiation signal.

Procedure 331.1, Control Room and Ol d Cable Spreading Room Heating, Ventilation and Air Conditioning System, descr ibes the partial recirculation mode:

this mode of operation is provided to mini mize contamination infiltration into the control room by maintaining a positive pressure in the control room using partial outside air.

Section 6.1.1 of 331.1, provides guidance for a radiological release with offsite power available. With offsite power ava ilable, System B or System A should be run in PART RECIRC mode. Only when there is a loss of offsite power, shall the System be run with the fan only (to limit EDG loading). Answer c is correct.

Running System A in the PURGE mode is incorrect. Purge mode is used to remove smoke, fumes, or other undesir able odors from the control room. Also, running the systems with fans only is r equired only when combined with a loss of off-site power to reduce EDG loading. Answer a is incorrect.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 18 of 135 Running System B in the PURGE mode is incorrect. Purge mode is used to remove smoke, fumes, or other undesir able odors from the control room. Answer b is incorrect.

Running System A in the FULL RECIRC Mode is incorrect. Full Recirc mode is used to minimize the intrusion of toxic gases into the control room. Answer d is incorrect.

295038 EK3.03 Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE RELEASE RATE: Control Room Ventilation Isolation (CFR: 41.5)

OC Learning Objective: 2621.828.0.0054 (02324: Explain the basis, with use of the procedures, for the four different modes of control r oom ventilation damper alignment and the effects of the damper alignment modes on control roomhabitability.)Cognitive Level: Comprehensive or Analysis

Question Type: Modified

9/8/06 Rossi reviewed. Deleted: The outside ai r temperature is 50° F and the control room air temperature is 74° F in the stem. The first dra ft of this question includeda loss of off-site power. With the loss of power, air temperatures became a concern to make the question correct. In the final version of the question with power available, air temperatures ar e no longer a concern. Changed 'lists' to'states' in the question. Deleted duplicate word in stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 19 of 13511.The reactor was at rated power w hen the following annunciator came into alarm: TURBINE VAC/SEALS - COND VAC LO 25 INCHES The reactor operator lowered recirculation flow as directed by the associated RAP. Condenser vacuum has recovered to 25.8" and is steady. The Unit Supervisor directs you to restore RPV pressure to the

pre-event value by adjusting the elec tronic pressure regulator (EPR), in accordance with 202.1, Power Operations.

Which of the following lists t he required action and its effect?Take the EPR RELAY POSITION control switch to 1 positionwhich will cause turbine control valves to 2 .a.(1) LOWER (%)(2) close furtherb.(1) LOWER (%)(2) open furtherc.(1) RAISE (%)(2) open furtherd.(1) RAISE (%)(2) close further Answer: d Justification: As power is reduced, the EPR relay position also goes down (proportional to turbine load). To raise R PV pressure back up, the turbine controlvalves must close down some. Lowering the EPR relay position even further will

do this (Raise (%)). As the TCV close down some, RPV pressure will rise.

Therefore, the EPR relay position must be taken to the RAISE position, which will cause turbine control valves to close further, causing RPV pressure to rise.

Answer d is correct. All other answers either manipulate the switch in the incorrect direction or the plant effect is incorrect. (See also procedure 315.4.)

RAP-Q3c directs a power reduction to maintain vacuum > 25".

295002 AA1.06 Ability to operate and/or m onitor the following as they apply to LOSS OF MAIN CONDENSER VACUUM : Reactor/turbi ne pressure regulating system (CFR:

41.7)OC Learning Objective: 2621.828.0.0051 (10446: Identify and explain system operating controls/indications un der all plant operating conditions.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 20 of 135Cognitive Level: Comprehensive or Analysis Question Type: Modified

9/8/06 Rossi reviewed. Changed 'switch' to 'cont rol switch' and added 'p osition' in stem.

Corrected typo (valve to value). Underlined 'into alarm'.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 21 of 13512.The plant was at rated power, when the following annunciators came into alarm: TORUS/DRYWELL - DW SU MP HI LEAK/PWR FAIL TORUS/DRYWELL - DW PRESS HI/LO Drywell pressure peaked at, and currently indicates 1.4 psig, and the PWR FAIL has been ruled out as a cause.

Which of the following would NOT be used to determine the drywell unidentified leak rate?a.Directly, by reading the Unidentified Drywell Leakage recorder (ULRM-1) on Panel 3Fb.Calculate, given the drywell su mp flow integrator readings and times of the readingsc.Calculate, given the time betw een the drywell sump low and highlevel alarmsd.Estimate, given the drywell equipment drain tank leak rate and condenser hotwell makeup rate answer: d Justification: The given question stem identifies an increased DW pressure from an increase in DW uni dentified leakage.

Procedure 312.9, Primary Cont ainment Control, provides directions on how to calculate the DW unidentified leak ra te: take the difference between the DW sump integrator readings and divide by t he elapsed time of the readings (this isthe method used for the daily surveillanc e). There also exists an unidentified DW leakage recorder on Panel 3F. If these are not functional, procedure 351.1, The Chemical Waste/Floor Drain Operat ing Procedure, provides a method to calculate DW unidentified leakage: measure the time between the low and high sump level alarms, and divide into the sump volume between these two alarms.

Answers a, b, and c are all acceptabl e methods. Answers a, b, and c are incorrect.

Answer d is a mix of identified leak rates and possible sources of unidentified leak rates. Answer d is correct. (R efer to drawings 147474, RAP-C3h, and procedures 351.1, 351.2, and 312.9).

295010 AA2.01 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE : Leak Rates (CFR: 41.10)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 22 of 135 OC Learning Objective: 2621.828.0.

0032 (00418: Given control panelindications, interpret the cause of Pr imary Containment System alarms, aloneand in combination, as applicable.Cognitive Level: Comprehensive or Analysis

Question Type: New

8/15/06: NRC Comments Added that DW pressure peaked at 1.4 psig (to show that 2.9 psig was never reached).9/8/06 Rossi reviewed. Deleted: with all systems normally aligned in stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 23 of 13513.The plant was at rated power w hen the Secondary Containment Control EOP, EMG-3200.11, was entered due to high area temperat ures (not dueto a fire).

Which of the following area leak det ection system annunciators will result in automatic isolation of the affected system?a.Cleanup System area leak det ection: CLEANUP SYSTEM - RWCUHELB annunciatorsb.Shutdown Cooling System area l eak detection: SD HX CLG - SDHX PUMP RM TEMP HI annunciatorsc.Isolation Condenser System ar ea leak detection: ISOL COND -COND AREA TEMP HI annunciatorsd.Trunion Room area leak detection: MAIN STEAM - TRUNION RM TEMP HI annunciators Answer: a Justification: Cleanup system leaks will be annunciated by D1d and D2d (RWCU HELB at 160° F) and by D8d (CU ROOM TEMP HI). The HELB annunciators, when alarmed simultaneously, will isolate the cleanup system at 160° F area temperature. The other cleanup alarm does no auto action. Answer a is correct.

Shutdown cooling system leaks will be annunciated by C8d (SD HX PUMP RM TEMP HI) but provide no automatic actions. Answer b is incorrect.

Isolation condenser leaks will be annunc iated by C8b (COND AREA TEMP HI) but provide no automatic actions. Answer c is incorrect.

Trunion room leaks will be annunciated by J8a (TRUNION RM TEMP HI) but provide no automatic actions. Answer d is incorrect. Main steam leaks into the

steam tunnel (trunion room) are annuncia ted by J3a and J4a (FLOW HI/MN STM LINE AREA TEMP HI-HI 1 and II) but answer d is asking specifically for trunion room leak detection.

295032 EK1.03 Knowledge of the operational implications of the followi ng concepts as they apply to HIGH SECONDARY CONTAINMEN T AREA TEMPERATURE: Secondary containment leakage detection (CFR: 41.8 to 41.10)

OC Learning Objective: 2621.828.0.0039 (10449: State the function of systemalarms, alone and in combination, as applicable in accordance with the system

RAPs.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 24 of 135 Cognitive Level: Memory of Fundamental Question Type: New NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 25 of 13514.The plant is at rated power. While the NLO was investigating the Reactor Building Sump 1-7 high level alarm, he states that due to an apparent fault in the RB sump 1-7 control circ uitry, neither sump pump will start.

Which of the following actions is required?

a.Manually isolate the inputs from RB Sump 1-6 into RB Sump 1-7b.Check that RB Sump 1-6 inputs into RB Sump 1-7 have automatically isolatedc.Manually isolate Drywell Floor Drai n Sump inputs into RB Sump 1-7d.Check that the Drywell Floor Drain Sump inputs into RB Sump 1-7have automatically isolated Answer: b Justification: With the RB Sump 1-7 high alarm in (RAP-RB1C(1-7), the Secondary Containment EOP is enter ed (EMG-3200.11). All systems discharging into the sump should be isolated (except for systems required for EOPs or fire suppression - of which, neither current ly apply). RB Sump 1-6 discharges into RB Sump 1-7, and sump 1-7 automatically isolates on a high water level alarm(which was given). According to the applicable RAP, the NLO should check for

automatic valve closure of inputs from Sump 1-6 into Sump 1-7. Answer b is correct.Since sump 1-6 output automatically is olated, no manual valve manipulations to isolate are required. Answer a is incorrect.

The drywell floor drain sump discharges to the chemical waste/floor drain collection tanks (the same place where RB sump 1-7 pumps discharge to (with adischarge check valve). Theref ore, there should not be any input into RB sump 1-7 from the drywell floor drains. Answ ers c and d are incorrect. (See drawing 147434, sheet 3, rev. 58) 295036 EK 1.02 Knowledge of the operational implications of the followi ng concepts as they applyto SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL :

Electrical ground/ circuit malfunction (CFR: 41.8 to 41.10)

OC Learning Objective: 2621.828.0.0015 (1414: Describe the operation of the pumps and level instrumentation associ ated with Reactor Building Sumps 1-6and 107 including automatic isolations.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 26 of 135Cognitive Level: Comprehensive or Analysis Question Type: New

9/8/06 Rossi reviewed. Deleted: wit h all systems normally aligne d in stem. Simplified thestem information.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 27 of 13515.The plant was at rated power wit h all systems normally aligned. The following annunciators came into alarm: ISOL COND - COND AREA TEMP HI RADIATION MONITORS AREA - AREA MON HI ISOL COND - COND A FLOW HI POSSIBLE RUPTURE The Operator verifies the isolati on condenser area rad monitor is above the high setpoint (Panel 2R) and area temperature has risen (Panel 10R).

Which of the following states the ex pected Isolation C ondenser A lineup in this condition? (assume no operator actions)a.The steam supply valves, condensate return valves and vent valves indicate OPENb.The steam supply valves, condensate return valves and vent valves indicate CLOSEDc.The steam supply valves and the condensate return valves indicate OPEN, and the vent valves indicate CLOSEDd.The steam supply valves and the condensate return valves indicate CLOSED and the vent valves indicate OPEN answer: d Justification: All three annunciators point to a rupture in Isolat ion Condenser A, in the vicinity of the IC (rad levels and te mperatures: Rap-C8b, RAP-10F1k). These first annunciators have no automatic actions associated with them. The third annunciator (RAP-C3a), will isolate the steam and condensate return valves for

the associated IC A from high Dp (high flow) (See also drawing 3029, sheet 2).

The IC vent valves are unaffected by the isolation signal to the steam and condensate return valves. The vent valves, which are normally open while at

power (auto close on system initiation), remain open following the isolation signal. The only answer which lists st eam and condensate return valves closed and vent valves open, is answer d. Answer d is correct. All other answers are incorrect, since they provide the incorrect valve lineup. (The associated RAP

does require the operator to close the vent valves, but no operator action is assumed in the question.)

207000 K4.01 Knowledge of ISOLATION (EMERGENCY) CONDENSER design feature(s) and/or interlocks which provide for the fo llowing: Isolation of the system in the event of a line break (CFR: 41.7)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 28 of 135 OC Learning Objective: 2621.828.0.0023 (02030: Describe the Isolation Condenser design features and/or inte rlocks (including signals and setpoints) which provide for the following: 1) aut omatic system initia tion; 2) automatic system isolation.)Cognitive Level: Comprehensive or Analysis

Question Type: New

9/8/06 Rossi reviewed. Changed 'lists' to 'states' in the question.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 29 of 13516.Core Spray Main Pump NZ01A was being lined-up after completing pump maintenance and suction valve V-20-3 packing replacement. The surveillance test requires both flow measurements and valve timing to

prove operability.

If the System was not properly vented, which of the following states theimpact during the surveillance?a.The suction valve would take less time to closeb.The motors would run at a constant higher amperage value c.System discharge pressure w ould indicate a higher valved.System flow would indicate a lower value Answer: d Justification: Running t he pump with the system not fu lly vented would not impact the closure time of the discharge valve. Answer a is incorrect.

The pump would run with a reduced capacit

y. Answer d is correct. With a reduced capacity, motor amps would be less. Answer b is incorrect.

System discharge pressure would be less

- not greater. Answer c is incorrect.

209001 K5.05 Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM : System Venting (CFR: 41.5)

OC Learning Objective: 2621.828.0.0010 (209-10445: Given a set of systemindications or data, evaluate and interp ret them to determine limits, trends and system status.)

Cognitive Level: Memory or Fundamental

Question Type: Modified

8/15/06: NRC Comments The original question did not reflec t the K/A. The question above is new.

9/8/06 Rossi reviewed. Changed stem to prov ide the necessary information while maintaining economy of wording.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 30 of 13517.The reactor was at rated power.

An event occurred that caused the Reactor Operator to manually scram the plant. Reactor power was 22%following the scram. The following conditions exist: Reactor pressure is 1004 psig Standby Liquid Control System 1 was initiated.

Which of the following shows the expected indications for SLC System 1?PUMP ON LightSQUIBS LightPump Discharge Pressure (psig)a.ONON1085 b.ONOFF1085c.OFFON985d.ONOFF985 Answer: a Justification: There is no automatic initiation of t he SLC System. In a normal standby configuration, both the PUM PS ON and SQUIBS lights are OFF (pumps are off and squib valves are energized). When a system is manually initiated, both lights go ON (see procedure 304, and EMG-3200.01B, Support Procedure 22). SLC discharge pressure should be so me valve greater than RPV pressure.

Answer a is correct.

No other answer has this correct co mbination and are therefore incorrect.

211000 A3.02 Ability to monitor automatic operati ons of the STANDBY LIQUID CONTROLSYSTEM including: Explosive valves indicating lights: (CFR: 41.7)

OC Learning Objective: 2621.828.0.0046 (10446: Identify and explain system operating controls/indications un der all plant operating conditions.)

Cognitive Level: Memory or Fundamental

Question Type: Modified

9/8/06 Rossi reviewed. Deleted: wit h all systems normally aligned.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 31 of 13518.A plant start-up is in-progress.

The following plant conditions exist: Reactor power is on Range 8 of the Intermediate Range Monitors (IRM) IRM 11 is in BYPASS due to erratic detector output Control rods are being withdrawn to raise reactor power The following annunciator came into alarm, followed by the listed automatic system initiation: VITAL POWER DC PWR LOST

- 24 VDC PP-A PWR LOST Standby Gas Treatment System automatically initiated Which of the following lists the effect s on the IRM System from this event?a.IRMs 11-14 meters indicate downscale on Panel 3R, and a rodblock and 1/2 scram existsb.IRMs 11-14 meters indicate downscale on Panel 4F, and a rodblock only existsc.IRMs 11-14 meters indicate up scale on Panel 3R, and a rodblock and 1/2 scram existsd.Only IRMs 12-14 meters indica te upscale on Panel 4F and a rodblock and 1/2 scram exists Answer: a Justification: The indications in the st em are those of a loss of 24 VDC Panel A (RAP-9XF7d). Power is lost to 11, 12 SRMs and to 11-14 IRMs , and to the SGT trip relays (which causes SGT to auto st art). The IRM trip auxiliary relays (and the IRM drawers on Panel 5R) (see dra wings 706E812, sheet 9, 3 and 237E566, sheet 1) are powered from 24 VDC. T he trip auxiliary relays are normally energized. When power is lost, all trips are instituted (upsca le rodblock, upscale scram, inop. and rodblocks). The effected IRM drawers also show downscale on the meter from loss of power. Therefor e, answer a is corre ct: the IRM meter shows downscale, and a rodblock and 1/2 scr am exist. All other answers either provide incorrect meter indication or wrong trips.

295003 K6.05 Knowledge of the effect t hat a loss or malfunction of the following will have on the INTERMEDIATE RANGE MO NITOR (IRM) SYSTEM : Trip Units (CFR: 41.7)

OC Learning Objective: 2621.828.0.0029 (10444:

Describe the interlock signalsand setpoints for the affected system components and expected response including power loss or failed components.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 32 of 135Cognitive Level: Comprehensive or Analysis Question Type: New

8/15/06 NRC Comments They pointed out that RAP-9XF7d (24 Volt PP-A Pwr Lost) says that IRMs 15-18fail downscale, but also says on the very same page that the plant will receive a

1/2 scram from IRMs 11-14. They w anted another correct reference. The companion rap for loss of 24 Volt PP-B ment ions just IRMs 15-18 twice. See also drawing 3C-736-11-001. This shows 24 VDC Power Panel A feeding control room panel 3R (IRMs 11-14). See also drawing 706E812, sheet 9 for IRM 11.

9/8/06Rossi reviewed. Changed 'underway' to 'i n-progress' in the stem. Changed 'lists' to 'states' in stem. Corrected justification:

power is lost to SRM 21, 22 (instead of incorrect designation SRM 11, 12).

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 33 of 13519.A plant startup is in-progress. The following plant conditions exist: SRM 22 has failed and is in BYPASS The 8 th control rod has just been fully withdrawn An event occurs which results in t he loss of instrument power to SRM drawer 24.

Which of the following lists the neutr on monitoring indications from this event?a.SRM recorder (Panel 4F) has lost powerb.Channel 24 period meter (Panel 4F) indicates infinityc.SRM 24 meter (Panel 5R) indicates upscale d.Channel 24 period meter (Panel 5R) indicates downscale answer: d Justification: 24 VDC powers the SRM draw er, including the trip relays. A loss of instrument power results in the downsca le indication of the SRM meters and period meters, both on Panel 5R and 4F. Therefore, answer d is correct. (see

drawings 706E812, sheet s 4, 47, procedure 401.1)

The SRM recorder power comes from 120 VAC CIP Div 1 (see drawings706E812, sheets 3 and 4). Ther efore, the SRM recorder is still powered. Answer a is incorrect.

Answers b and c are incorrect since both fail downscale.

215004 K6.05 Knowledge of the effect t hat a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRM) SYSTEM : Trip units (CFR: 41.7)

OC Learning Objective: 2621.828.0.0029 (10444:

Describe the interlock signalsand setpoints for the affected system components and expected response including power loss or failed components.)

Cognitive Level: Memory or Fundamental

Question Type: New

9/8/06Rossi reviewed. Changed 'underway' to 'i n-progress' in the stem. Changed 'lists' to 'states' in stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 34 of 13520.The reactor is at rated power. Be low are the currently bypassed Local Power Range Monitors (LPRMs) into Average Power Range Monitors (APRMs): APRM 1 APRM 5 APRM 628-33A44-33D04-33B28-49C36-41B20-49D 36-41A Which of the following additi onal LPRM inputs to APRMs:

(1) CANNOT be bypassed (and maintain APRMs OPERABLE), as allowed by procedure 403, LPRM-APRM System Operations, and (2) the effect if bypassed?a.(1)44-33A (2) Too many LPRM inputs bypassed resulting in an automatic 1/2

scramb.(1)36-41D (2) Too many LPRM inputs bypassed in one radial location

resulting in an automatic 1/2 scramc.(1)28-49D (2) Too many LPRM inputs bypassed resulting in an automatic 1/2

scramd.(1)20-49B (2) Too many LPRM inputs bypassed in one radial location

resulting in an automatic rodblock Answer: a HANDOUT: Attachment 202.1-1, Daily APRM Status Check

Justification: Procedure 403 has 2 precaut ions: 5.2.2.2 (and 5.3.2.3) says "Each APRM requires at least 5 LPRM signals. Inadvertently bypassing a 4 th LPRMsignal will initiate an INOP trip." 5.3.

2.4 says "Failure of, or bypassing, two chambers from one radial location in any one APRM shall make that APRM channel inoperable."

Answer a bypasses the 4 th LPRM from APRM 1 and this APRM will become inoperable, and an automatic 1/2 scram occu rs. (APRM 1 and 5 are located in the same core quadrant. APRM 1 has a and c LPRM inputs. Even though the given NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 35 of 135 table does not show any LPRMs into APR M 1 from 44-33, it can be seen that LPRM 44-33D inputs into APRM 5. Ther efore,LPRM 44-33A must input into APRM 1.) Answer a is correct.

Answer b bypasses a second LPRM in the same radial location in APRM 15, which makes APRM 5 inoperable. There is no automatic function from this bypass. Answer b is incorrect.

Answer c bypasses a 3 rd LPRM in APRM 5 and is allowed. No automatic function occurs from this bypass. Answer c is incorrect.

Answer d bypasses a second LPRM in t he same radial location for APRM 6, which makes APRM 6 inoperable. There is no automatic action from this bypass.

Answer d is incorrect.

215005 2.1.32 Ability to explain and apply system lim its and precautions: APRM/LPRM (CFR:

41.10)OC Learning Objective: 2621.828.0.0029 (10444:

Describe the interlock signalsand setpoints for the affected system components and expected response including power loss or failed components.)Cognitive Level: Comprehensive or Analysis Question Type: Modified

8/15/06: NRC Comments Added "additional" LPRMS-NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 36 of 13521.The reactor is at rated power. Whic h of the following would prevent theability to determine reactor coolant system leak rate?a.Containment High Range Radiat ion Monitor indicates 45R/Hrb.Containment Airborne Particulate and Gaseous Radiation Monitoring System indicates 1x10 5 CPMc.Drywell pressure at or below 2.0 psig

d. Reactor water level at or below 86" TAF Answer: d Justification: Reactor coolant system leak rate into the containment is measured by how much water is pumped out of the pr imary containment over time. With the drywell equipment and floor sump isolation valves closed, the reactor coolantsystem leak rate cannot be determined.

Anything that causes an isolation of these valves would prevent the ability to determine reactor coolant system leak rate.RAP-C1g, CAPGRAMS Radi ation High, has no automatic actions. Answer a is incorrect.RAP-10F4k, Hi Range Rad. Monitor Abnorm., will close the torus/DW vent and purge valves at the high setpoint (not t he DW floor and equipment drain valves).

Answer b is incorrect.

Support procedure 1 of RPV Control

- No ATWS (EMG-3200-01A) and RAP-C4h, shows that the drywell equipment and floor drain IVs isolate on a primary containment isolation signal (RPV water le vel at or below 86" TAF, or drywell pressure at or above 3.0 psig). Answer c is incorrect.

Answer d is correct since the drain valves close at or below 86" TAF (see Support procedure 1).

223002 K1.14 Knowledge of the physical connections and/or cause-effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the followi ng: Containment drainage system (CFR: 41.2 to 41.9)

OC Learning Objective: 2621.828.0.0037 (02456: Describe RPS isolation logic trip signals and functions, including the following: 1) purpose/design basis; 2) setpoints; 3) conditions that allow bypassing isolation signals; 4) how bypassing isolation signals is accomplished.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 37 of 135 Cognitive Level: Memory or Fundamental Question Type: New

9/8/06 Rossi reviewed. Corrected the units in selections a and b. Changed selection c from 'at or above 2 psig' to 'at or below 2 psig'.

9/22/06 Validator comments: Changed question from 'With the reactor at power,..' to 'The reactor is at rated power. Which of-'.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 38 of 13522.Which of the following would result in reactor water level being controlled in single-element control?a.The loss of a steam flow signal to the digital control computers while at rated powerb.The loss of a feed flow signal to the digital control computers while at rated powerc.The loss of both digital contro l computers while at rated powerd.The loss of RPV water level inputs from LT-ID13A and LT-ID13B while at power Answer: c Justification: FW level control uses a steam flow signal from each of the two steam lines. When a steam flow input is lost, the system will double the good remaining steam flow input and will conti nue to use 3-element control. Answer a is incorrect.

The FW level control system uses feedwater flow from each of two feedwaterlines. When one FW flow input is lost, the system will calculate the feed flow

based upon valve position, num ber of running pumps and reactor pressure. The system will continue to use 3-element control. Answer b is incorrect.

When a single digital control computer is lost, the system will continue in 3-element control with the oper able digital control com puter. When both computers are lost, control is transferred to the Moore controllers, which will control in single-element control. Answer c is correct. (see RAP-J1c)

If RPV water level inputs ID13A and ID13B are bad, then the controller uses LT-0013C. The control system does not transfer to single element control. Answer d is incorrect. (See MDD-OC-625-B, FW and Recirc Control Systems Upgrade Modification) 259002 K4.09Knowledge of REACTOR WATER LEVEL CONTROL SYSTEM design feature(s) and/or interlocks which provide for the fo llowing: Single element control (reactor water level provides the only input) (CFR: 41.7)

OC Learning Objective: 2621.828.0.0018 (10444:

Describe the interlock signalsand setpoints for the affected system components and expected response including power loss or failed components.)

Cognitive Level: Memory or Fundamental NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 39 of 135 Question Type: New 8/15/06: NRC Comments The original answer d (Following a scram from rated power while controlling with the low flow regulating valve in MANU AL) was not a plausible distracter.

Changed with what is shown.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 40 of 13523.The plant is at rated power. The following switch position is noted: STANDBY GAS SELECT is in position SYS 2 An event occurs which automatically initiates the Standby Gas Treatment System.Five minutes after the initiation (with no operator action), which of the following is the correct fan/valv e configuration if the lead systemdeveloped/maintained a low flow signal?

System 1 Fan System 2 Fan System 2 Orifice Valve V-28-28a.ONONOPEN b.ONOFFCLOSEDc.OFFONCLOSEDd.ONOFFOPEN Answer: aJustification: On an automatic system init iation, both SGT fans start. If the leadfan develops adequate flow within the first 2-3 minutes, the lag fan will shutdown and the associated inlet/outlet valves close. If the lead fan does not develop adequate flow, the lag fan continues and the lead fan continues to run, but with the lead system inlet/outlet valves closed.

The system orifice valves are normally closed (with the systems is standby) and stays closed when the lead system starts with proper flow. If the lead r unning system sees low flow, then besides what's already been said, the lead system orifice valve also opens (and inlet/outlet valves close and the redundant system assumes the SGT function).

Therefore, 5 minutes after an auto initiati on, system 2 fan (which was selected as lead) will be running with the loop inlet/outlet valves closed and loop orifice valve open. System 1 fan is also running per forming the SGT function. Answer a is correct. (See also procedure 330)

All other answers are incorrect due to in correct status of fan/valve. (See RAP-L5b).261000 A3.01 Ability to monitor automatic operations of the STANDBY GAS TREATMENTSYSTEM including: System Flow (CFR: 41.7 )

OC Learning Objective: 2621.828.0.0042 (10445: Given a set of systemindications or data, evaluate and interpre t them to determine limits, trends, and system status.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 41 of 135Cognitive Level: Comprehensive or Analysis Question Type: Modified

8/15/06: NRC Comments Added with no operator action in the stem.

9/8/06 Rossi reviewed. Deleted: with all syst ems normally aligned from the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 42 of 13524.The plant was at 50% power when a small LOCA into the primary containment occurred. Following the scr am, all offsite power was lost and both emergency diesels could not be started. The fo llowing conditions exist: RPV water level is 113" TAF and lowering very slowly Drywell pressure is 5.1 psig and rising slowly RPV pressure is 820 psig and lowering slowly All control rods are fully inserted ABN-37, Station Blackout, has been entered o 4160 VAC Bus D has been powe red from the combustion turbine, and critical loads have been restored in accordance

with ABN-37-7 o the SBO Transformer load is currently 2.4 MWe Which of the following is the required action to restore RPV water level?

a.Lower RPV pressure as requir ed and lineup and inject condensatetransfer to core sprayb.Lower RPV pressure as required and lineup and inject fire water to core sprayc.Lineup and inject with Condensat e Pump A / Feedwater Pump Ad.Lineup and inject with the maxi mum flow using both CRD pumps Answer: d HANDOUTS: ABN 7 and EMG-3200.01AJustification: With the given conditions, only feedwater and CRD can inject into the RPV at 820 psig. There is ample room on the SBO transformer (8 MWe maximum allowable and currently at 2.

4) to start one condensate pump and one feedwater pump (0.811 MWe for the CP, and 3.141 MWe for the FWP = 3.952 additional MWe for 1 condensate/feed pump; so 8 - 2.4 = 5.6 MWe room on the SBO transformer). But, condensate pump A and feedwater pump A are powered from 4160 bus 1A, which cannot get power from bus 1B (1B is powered from the

combustion turbine). Answer c is incorrect.

To inject with condensate transfer or fire water, core spray system 1 or corespray system 2 must be unavailable. With the given conditions of power to BusD, components on core spray system 1 and core spray system 2 are available.

Therefore, conditions have not been met to inject with either condensate transfer of fire water. Answers a and b are incorrect.

Part of the lineup in ABN-37 is cross-ti eing power to USS 1A2, which suppliesCRD Pump 1A. CRD Pump 1B is also pow ered. Injection from these systems is directed. Answer d is corre ct. (Se ABN-37, EMG-3200-01A)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 43 of 135 262001 (A.C. Electrical Distribution) 2.4.6 Knowledge symptom based EO P mitigation strategies.(CFR: 41.10)Cognitive Level: Comprehensive or Analysis

Question Type: New

9/8/06 Rossi reviewed. The correct answer was changed to answer d. ABN-37 cross-ties power to USS 1A2 (for CRD 1A) and CRD 1B already is aligned to receive

power. Both can inject as required. Answ er c (the original correct answer) was changed to say condensate pump A and f eedwater pump A. Feedwater is required in the ABN, but bus 1A, wh ich supplies these pumps, cannot be powered from the combustion turbine.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 44 of 13525.The plant was at rated power. An event occurred which resulted in a loss of 125 VDC Bus C.

Which of the following com ponents has lost DC power?1.Station Blackout Transformer Remote Monitoring Panel2.4160V Switchgear 1A and 1B 3.Turbine Generator Excitation Switchgear 4.Remote Shutdown Panel Inverter 5.Emergency Diesel G enerator 1 Switchgeara.1 and 2b.2, 3, and 4 c.3 and 5 d.1 and 5 Answer: d Justification: The following components receive DC power from that listed:1.Station Blackout Transformer Remote Monitoring Panel - DC C (correct)2.4160V Switchgear 1A and 1B - 1A from DC C and 1B from DC B (incorrect)3.Turbine Generator Excitation Switchgear - DC A (incorrect)4.Remote Shutdown Panel Inverter - DC B (incorrect)5.Emergency Diesel Generator 1 Switchgear - DC C (correct)Therefore, only answer d has choices of 1 and 5. Answer d is correct and all other answers are incorrect. (See ABN

-54, ABN-55, drawing BR 3028 and D-3033).263000 K2.01 Knowledge of electrical power supp lies to the following: Major DC Loads (CFR: 41.7)

OC Learning Objective: 2621.828.0.0012 (01106:

Draw a one-line diagram of the 125 VDC Distribution System includi ng: major busses (A, B and C battery systems), battery charging power supplies, major breakers, automatic bus transfer switches, manual bus transfer switches, and major loads for each DC panel.)Cognitive Level: Memory or Fundamental

Question Type: Bank NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 45 of 135 8/15/06: NRC Comments Corrected the justification to show that answer d is correct (had said that answer a was correct, which was in error).

9/8/06 Rossi reviewed. Deleted: with all syst ems normally aligned in the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 46 of 13526.The plant was shutdown with plant pow er supplied from the normal offsite source, when a total loss of offsite power occurred. The plant responded

as designed.

Several hours later, all offsite power was restored and 4160 Bus 1A has been re-energized.

Which of the following states the co rrect method used to transfer Bus 1C from its EDG back to normal power?Synchronize EDG1 with Bus 1A, reduce load on EDG1 and then-a.place the local MODE SELECTOR switch in STOPb.place the Control Room NORMAL START switch in STOP c.open the EDG1 output breaker , then place the local MODESELECTOR switch in STOPd.open the EDG1 output breaker, then place the Control RoomNORMAL START switch in STOP Answer: b Justification: When offsite power is lo st, both EDG1 and 2 automatically start and load onto their respective emergency bus (EDG1 - Bus 1C; EDG2 - Bus 1D).

There is a precaution in procedure 341, Emergency Diesel G enerator Operation, which states that if the EDG1 (2) fast-s tarted (ie, loss of Bus 1C voltage), then it can only be shutdown from the control room. Therefore, answers a and c are incorrect since the listed switches are local at the EDG.

341 directs the Bus 1C be paralleled, reduce load (KVAR and KW) on the EDG, then place the Normal Start switch in St op. When placed in Stop, the EDG outputbreaker will open, the EDG slows to 400 RPM for 15 minutes, then shuts-down.

There is no reason to first open the EDG output breaker, then shutdown the EDG since the switch in Stop performs this function automatically. Answer d (and c) are incorrect. Answer b is correct.

264000 A4.03 Ability to manually operate and/or monito r in the control room: Transfer of emergency control between manual and automatic (CFR: 41.7)

OC Learning Objective: 2621.828.0.0013 (264-10446: Identify and explain system operating controls/indications un der all plant operating conditions.)

Cognitive Level: Memory or Fundamental

Question Type: New NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 47 of 135 8/15/06 NRC Comment They felt that the references show ed that 2 answers were plausible. Thisquestion has replaced the original question.

9/8/06 Rossi reviewed. Changed stem to prov ide the necessary information while maintaining economy of words.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 48 of 13527.The plant is at rated power.If a total loss of offsite power, were to occur, which of the followingReactor Building Closed Cooling Water valves would still have electrical

power to operate? V-5-147CCW Inlet Isolation Valve V-5-166RBCCW Outlet Isolation Valve V-5-167RBCCW Outlet Isolation Valvea.V-5-147 and V-5-166 onlyb.V-5-147 and V-5-167 only c.V-5-166 and V-5-167 only d.V-5-147, V-5-166 and V-5-167 Answer: d Justification: The valves are powered from the following busses: V-5-147MCC 1B21A V-5-166MCC 1B21B V-5-167MCC 1A21 During a loss of offsite power, busses C and D (which will be re-powered from EDG1 and EDG2 respectively) load shed. Busses 1A2 and 1B2 always get

power. These busses power 1B21A, B and 1A21. Therefore, all three RBCCW

isolation valves have power. Answer d is correct. The other answers are incorrect since they do not list all valves that hav e electrical AC power. (see 309.2) (The valve noun names provided above ar e named as in the procedure.)

400000 K2.02 Knowledge of electrical power supplies to the following: CCW Valves (CFR: 41.7)

OC Learning Objective: 2621.828.0.0016 (262-10444: Describe the interlock signals and setpoints for the effected system components and expected system response including power loss or fa iled components) 2621.

828.0.0035 (07261:

State the location and explain how to mani pulate all controls normally operated for the RBCCW System.)

Cognitive Level: Memory or Fundamental

Question Type: Bank

9/8/06 Rossi reviewed. Deleted: with all syst ems normally aligned in the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 49 of 13528.Which of the following could be indicative of a Reactor Manual Control System control rod move ment timer malfunction?a.The green INSERT light ON for 3.5 seconds during a control rod single notch ROD INb.The red WITHDRAW light ON for 3 seconds during a control rod ROD OUT NOTCHc.The amber SETTLE light ON for 5 seconds following a control rod single notch ROD IN evolutiond.The green INSERT light ON for 1 second during a control rod ROD OUT NOTCH Answer: b Justification: For a single notch-in, t he green insert light should be on for 3.5 seconds (See ABN-6, Control Rod Driv e System). Answer a is incorrect.

For a rod notch-out, the r ed withdraw light should be on for 1.5 seconds. Thegiven answer is twice as long. Answer b is correct.

The amber settle light should be on for 5 seconds following the insert or withdraw evolution. Answer c is incorrect.

The green light is on for 1 second during a control rod out notch. Answer d is incorrect.

201002 A3.04 Ability to monitor automatic operati ons of the REACTOR MANUAL CONTROLSYSTEM including: Rod movement sequence timer malfunction alarm (CFR: 41.7)

OC Learning Objective: 2621.828.0.036 (00726: Given a mode and direction for control rod movement, describe respons e of the timer, response of the CRD System, system indications and operation of controls.

Cognitive Level: Memory or Fundamental

Question Type: New NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 50 of 13529.Which of the following would result in a Technical Specification violation?a.While at power, a recirculation pump was started in an idle loop whose loop temperature was 40° F less than reactor coolant temperatureb.While at power (5-loop), over a one week period, two recirculationpumps tripped and the loops were placed in an IDLE conditionc.During a startup, while on R ange 10 of the Intermediate Range Monitors, a malfunction in the Master Recirc Speed Controller lowered recirculation flow to 38x10 6 lb/hrd.While at power, an event occurr ed which required that a single recirculation scoop tube be placed into local manual control Answer: cHANDOUT: TECH SPECS 3.3 Justification: IAW TS 3.3.C2, an idle recirculation pump shall not be started unless the loop temperature is within 50° F of the reactor coolant temperature.Since this answer gave 40° F, the operation is allowed by TS and no TS violation

is present. Answer a is incorrect.

IAW TS 3.3.F, 2 idle recirculation loops are allowed (except for starting) with no further actions. Since there is no TS ent ry violation, answer b is incorrect.

IAW TS 3.3.H.2, the minimum required recirculation flow rate while on IRM Range 10 is 39.65x10 6 lb/hr. Since flow was below this minimum, a TS action would be required. Answer c is correct (t his is also stated in procedure 301.2, Reactor Recirculation System).

Placing a recirculation MG set scoop tube in local manual control is not mentioned in TS, and therefore does not violate TS. Answer d is incorrect.

202002 2.1.33 (Recirculation Flow Control)Ability to recognize indications for syst em operating parameter s which are entry-level conditions for technical specifications. (CFR 41.10 / 43.2 / 43.3)

OC Learning Objective: 2621.850.0.0090 (016581:

Identify whether or not a Tech Spec or License Limit has been exceeded.)Cognitive Level: Comprehensive or Analysis

Question Type: Modified 8/15/06: NRC Comments NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 51 of 135 They thought that answers b and c were both correct. The question has been re-worded to ask which one would result in a TS violation.

9/8/06 Rossi reviewed. Corrected justificati on for 2 idle recirculation loops.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 52 of 13530.Which of the following power losses would cause the Rod Worth Minimizer to be INOPERABLE due to the loss of control rod position information?a.Protection System Panel Ab.Instrument Panel 4A c.Continuous Instrument Panel CIP-3 d.VACP-1 Answer: c Justification: Of the listed power supplies, only conti nuous instrument panel 3 provides electrical power to the control rod position indications. All other listed power supplies do not provide this power. Answer c is correct. (See ABN-58, GU 3C-733-11-005) 214000 K1.01 Knowledge of the physical connections and/

or cause/effect relationships between ROD POSITION INFORMATION SYSTEM and the following: RWM: Plant-Specific (CFR: 41.2 to 41.9)

OC Learning Objective: 2621.828.0.0041 (10444:

Describe the interlock signals and setpoints for the effected system components and expected system response including power loss of failed components.)

Cognitive Level: Memory or Fundamental Question Type: New NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 53 of 13531.The plant is at 2% power during a startup, with drywell inerting in-progress. Drywell oxygen concentrat ion is currently 9% and lowering slowly.If a spurious high drywell pressure signal initiated Standby Gas Treatment system, which of the following is co rrect (assume no operator action)?a.Drywell pressure will rise due to the nitrogen addition but venting at a slower rate through Standby Gas Treatmentb.Drywell pressure will lower due to nitrogen addition isolation and venting through Standby Gas Treatment systemc.The drywell oxygen indicator show s a stable valid indication due to the isolation of nitrogen and DW vent and purge valvesd.The drywell oxygen indicator no longer shows a valid indication due to the isolation of the drywell oxygen sampling system Answer: d With a high drywell pressure isolation si gnal, nitrogen into the drywell is isolated (V-23-14, -14), and the drywell atmosphere to RB HVAC isolate (V-27-1, -2). On the same isolation signal, the drywe ll oxygen sample primary containmentisolation valves also close. Therefor e, nitrogen inlet/outlets of the primary containment are isolated, and the oxyg en sampling system is also isolated.

Answer d is correct.

Answer a is incorrect since nitrogen addition is isolated. Answer b is incorrect since drywell venting is isolated. Answer c is incorrect since the oxygen reading is not valid since the oxygen samp ling system is isolated. (see BR2011, 13432.19-1, M0012, 3E-666-21-1000, USAR Table 6.2-12; 312.9) 223001 A1.06 Ability to predict and/or monitor changes in parameter s associated with operating the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES controls including:Oxygen Concentration CFR: 41.5)

OC Learning Objective: 2621.828.0.0032 (00394: Given auto isolation signals, list or identify causes, system responses, and affected primary containment system components.)Cognitive Level: Comprehensive or Analysis

Question Type: New

8/15/06: NRC Comments Added no operator actions.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 54 of 13532.A plant startup is in-progress.

The startup is continuing with no noted problems, in accordance with procedur e 201, Plant Startup. The current conditions exist: RPV pressure is 150 psig Turbine warming is in-progr ess, at the pre-warming mark The very next step is to open the turbine control valves and turbine stop valve #2 The Mechanical Pressure Regulator is set at 250 psig Turbine Bypass Valves are CLOSED Which of the following lists the met hod to open turbine control valves and turbine stop valve #2 for turbine warming?

Open the turbine control valves by using (1) , and open turbine stop valve #2 by using (2) .a.(1) the BYPASS VALVE OPENING JACK switch (2) the LOAD LIMIT CONTROL switchb.(1) the MPR Control Switch (2) the SPEED LOAD CHANGER switchc.(1) the BYPASS VALVE OPENING JACK switch (2) the MAIN STOP VALVE N O. 2 INTERNAL BYPASS switchd.(1) the LOAD LIMIT CONTROL switch (2) the MAIN STOP VALVE N O. 2 INTERNAL BYPASS switch Answer: c Justification: IAW 315.1, Turbine Generat or Startup, given that the turbine bypass valves are closed, and the , placing the BYPASS VALVE OPENINGJACK in RAISE will open the turbine control valves. Then the stop valve is

opened to admit steam by placing the MAIN STOP VALVE NO. 2 INTERNAL BYPASS in RAISE position. If the turbine bypass valves were open, then the

placing the load limit control switch to raise would open the turbine control valves; and stop valve number 2 bypass woul d open the stop valves. Answer c is correct.245000 K5.02 Knowledge of the operational implications of the followi ng concepts as they apply to MAIN TURBINE GENERATOR AN D AUXILIARY SYSTEMS : Turbine operation and limitations NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 55 of 135 (CFR: 41.5)

OC Learning Objective: 2621.828.0.0050 (10446: Identify and explain system operating controls/indications un der plant operating conditions.)

Cognitive Level: Memory or Fundamental

Question Type: Modified

9/8/06 Rossi reviewed. Changed 'The plant is st arting-up following a refuel outage' to 'a plant startup is in progress'.

9/22/06 Validator comment: Added 'at t he pre-warming mark' in the second bullet.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 56 of 13533.The plant is at rated power. You have just received a report that firedetector R5D9 (Reactor Building 51' North, Zone 1) failed its surveillance

test to detect and alarm a fire. The SRO has declared this fire detector inoperable. (There are no alarms lo cked-in from this inoperable fire detector).Your initial investigation shows that there are 8 fire detectors on RB 51' North, Zone 1, and 9 fire detectors on RB 51" North, Zone 2.

Which of the following states how this inoperable fire detector effects theability of the fire protection system to detect fires in RB 51' North and to actuate Deluge System #5?a.The fire protection system can still detect fires and actuate the fireprotection system for mitigation; no compensatory measures are requiredb.The fire protection system can still detect fires but CANNOTactuate the fire protection system for mitigation; no compensatory measures are requiredc.The fire protection system CANNOT detect fires nor actuate the fireprotection system for mitigation; an hourly fire watch patrol must be establishedd.The fire protection system CANNOT detect fires nor actuate the fireprotection system for mitigation; a continuous fire watch must be established Answer: a HANDOUT: PROCEDURE 333 (Plant Fire Protection System, Attachment 333-15) AND PROCEDURE 101.2 (OYSTER CREEK SITE FIRE PROTECTION PROGRAM, Attachment 101.2-3)

Justification: IAW procedure 645.6.031 (Atta chment 645.6.031-2), there are 8 fire detectors in 51' RB North Zone 1 (which includes the given inoperable detector).

As stated in the stem, t here are no other inoperable det ectors. Therefore, all other fire detectors in Zone 1, and all t hose in Zone 2 of 51' RB North will alarm at the fire panels when activated. Each fire detector is independent of the others to activate the alarms on the panels.

IAW procedure 333 (Attachment 333-15), act uation of Deluge System 5 (for 51' RB North) requires only 1 detector from 51' RB North Zone 1 to actuate, and 1 detector from 51' RB North Zone 2 to ac tuate. Therefore, one inoperable detector will still allow the fire pr otection system to detect and mitigate a fire in 51' RB North.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 57 of 135 Procedure 101.2 tells us that 6 detectors are required for RB 51' North Zone 1, and 7 detectors are required for RB 51 North Zone 7. With the number of operable fire detectors less than required, a fire watch patrol must be established. Therefore, fire detection in RB 51' North is still operable and functional, and no compensatory actions ar e required. Answer a is correct.

Since the system is still able to mitigate a fire, answer b is incorrect. Since the system is still able to detect and mitigate a fire in the area, no compensatory measures are required. Answers c and d are incorrect.

286000 K3.01 Knowledge of the effect that a loss or malfunction of the FIRE PROTECTIONSYSTEM will have on following: The ability to detect fires (CFR: 41.7)

OC Learning Objective: 2621.828.0.0019 (286-10445: Given a set of systemindications or data, evaluate and interpre t them to determine limits, trends, and system status.)Cognitive Level: Comprehensive or Analysis

Question Type: New

8/15/06: NRC Comments They want to see procedure 645.6.031.

No comment regarding the question as written.9/8/06 Rossi reviewed. Deleted: with all system s normally aligned and no equipment out of service from the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 58 of 13534.The plant is at rated power.

Which of the following Emergency O perating Procedures has an entry condition that is available through the plant process computer and is NOT read on the control room panels?a.EMG-3200.01A, RPV Control - NO ATWSb.EMG-3200-01B, RPV Control - With ATWS c.EMG-3200.02, Primary Containment Controld.EMG-3200.11, Secondary Containment Control Answer: c Justification: Indications RPV water le vel, RPV pressure, drywell pressure, and reactor power (entry conditions into R PV Control - No ATWS, and RPV Control -

With ATWS) can be found on the contro l room panels. Answers a and b are incorrect.

Indications for torus water temperature, drywell pressure, torus water level, and primary containment hydrogen concentrati on (entries for Primary Containment Control) can be found on control room panel

s. Bulk drywell temperature (also an entry into Primary Containment Control) is available only on the plant process computer (SPDS - Containment Cond itions screen). (When the PPC is not operable, DWT can be calculated from pl ant indications, but the question stem states that no equipment is OOS.)

Indications of the Alert emergency classi fication for radioactivity release are found from control room panel indications, and from other data sources, including procedures. Indications of an isolati on condenser tube leak are found from control room panel indications and alarms. Answer d is incorrect.

2.1.19 Ability to use plant computer to obtain and evaluate parametric information on system or component status. (CFR: 41.10)

OC Learning Objective: 2621.863.0.0007 (02233: Discuss the relevance of information shown on the PPC SPDS disp lays to the implementation of the SBEOPs. )Cognitive Level: Memory or Fundamental

Question Type: New

8/15/06: NRC Comments: Bolded NOT.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 59 of 135 9/8/06 Rossi reviewed. Deleted: with all system s normally aligned and no equipment out of service from the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 60 of 13535.A plant startup is in-progress.

The following conditions exist: The MODE Switch is in STARTUP, with control rod withdrawals in-progress IRMs 11, 12, 15, 16, 18 read 72-74 on Range 1 IRMs 13, 14, and 17 read 9 - 10 on Range 2 A malfunction in the IRM drive circui try caused IRM 13 to withdraw to the full-out position.

Which of the following states the e ffect on the plant and the required Operator actions to conti nue withdrawing control rods?a.There are panel annunciators ONLY

withdrawing control rods may continue without any other control panel manipulationsb.There are panel annunciators and a rodblock from IRM downscale ONLY; bypassing the IRM is required to continue withdrawing

control rodsc.There are panel annunciators and a rodblock from IRM downscale AND IRM detector position; bypassing the IRM is required to

continue withdrawing control rodsd.There are panel annunciators, a rodblock and a 1/2 scram; bypassing the IRM and resetti ng the 1/2 scram is required to continue withdrawing control rods Answer: c Justification: The following IRM param eters provide roblocks only (no scraminput): IRM downscale (in REFUAL and STARTUP; bypassed in Range 1 or in

RUN), detector not fully inserted (bypa ssed in RUN), and IRM high (bypassed in RUN). When the IRM comes off the full-in position, a rodblock is instituted (pluspanel annunciators). It is expected that the IRM will also go downscale as it drives to the fully withdrawn position (dow nscale also gives a rodblock except in Range 1). There are no 1/2 scrams from t hese conditions. Therefore, to continue to move control rods, IRM 13 (which is instituting a rodblock both from downscale and IRM position) must be bypassed. Answer c is correct.

Answer a is incorrect since it does not lis t rodblocks. Answer b is incorrect since it does not list all rodblocks. Answer d is incorrect since no 1/2 scram occurs. (See RAP-H7a, 237E912 and 796E212).

2.2.2 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 61 of 135 Ability to manipulate the console contro ls as required to operate the facility between shutdown and designated power levels. (CFR: 45.2)

OC Learning Objective: 2621.828.0.

0029 (10449: State the function and interpretation of system alarms, alone and in combination, in accordance with system RAPs.)Cognitive Level: Comprehensive or Analysis

Question Type: New

9/8/06 Rossi reviewed. Changed 'The plant is st arting-up after a 5-day forced outage' to

'a plant startup is in-progress' in the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 62 of 13536.The plant is shutdown for a refuel outage. A fuel shuffle is in-progress.

Which of the following, as stated by procedure 205.0, Reactor Refueling, states when a communication must be made between the Control Room Licensed Operator and the Refue ling Senior Reactor Operator?a.When a blade guide is vertically aligned and is being lowered into the coreb.When a new fuel bundle is vertica lly aligned and is being lowered into the corec.When an irradiated fuel bundle is vertically aligned and is being lowered into the spent fuel racksd.When a control rod is vertically aligned and is being lowered into the core Answer: b Justification: IAW procedure 205.0, (s ection 7.3.1) Reactor Refueling, a communication between the Refuel SRO and the CRO is required at the

commencement and completion of each move, and whenever a bundle enters or

exits the core. None of the answers provided are the commencement or completion of a step. Answer b does meet the procedural requirement in that a bundle is entering the core. Answer b is correct.

Procedure 205.0 does not require a comm unication regarding the blade guide into the core. Answer a is incorrect.

Procedure 205.0 does not require a communi cation regarding the fuel placement into the fuel pool racks. Answer c is incorrect.

Procedure 205.29, Control Rod Blade Removal and Replacement, does not require a communication as the blade enters the core. Answer d is incorrect.

2.2.28 Knowledge of new and spent fuel mo vement procedures. (CFR: 43.7 )

OC Learning Objective: 2621.812.0.0003 (00323:

State the responsibilities of the following personnel during re fueling operations IAW procedure 205.0: 1) Reactor Engineer; 2) Shift Manager; 3) Control Room Licensed Operator; 4) Bridge Operator; 5) Fuel Move Checke r; 6) Fuel Handling Director.)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 63 of 135 Cognitive Level: Memory of Fundamental Question Type: New

9/8/06 Rossi reviewed. Changed stem to prov ide the necessary information while maintaining economy of words. Changes 'lists' to 'states' in the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 64 of 13537.The plant is at rated power with the following condition: The Reactor Water Cleanup System has been removed from service to support system repair (electrical)

The outside containment motor operated isol ation valve is to be used as aclearance boundary. IAW OP-MA-109-101, Clearance and Tagging, this is allowed as long as the following conditions are met:1. The valve control station is tagged2. The electrical energy source is removed

3. The valve's handwheel is tagged Given that the valve is located in a high radiation area, which of thefollowing states who can waive the 3 rd requirement above, due to ALARA considerations?a.The Unit Supervisorb.The Radiation Protection Technician c.The Cleanup System Engineer d.The Clearance Writer (Reactor Operator)

Answer: a Justification: OP-MA-109-101, Clearance and Tagging, states that the above requirement shall only be waived by the clearance approver (who is a first line

supervisor or above) or Shift M anagement when radiol ogical or hazardous conditions exist. The same procedure defined clearance approver as an individual trained and qualified to appr ove clearances; a clearance approver should be a currently or previously licensed SRO. OP-OC-100, Oyster Creek Conduct of Operations provides the following for Shift Management: normallyconsists of an SRO licensed Shift Manager, an SRO licensed Unit Supervisor, and an SRO licensed Field Supervisor. Theref ore, answer a is correct. All other answers are incorrect.The RP Tech is not a licensed SRO. Answer b is incorrect.

The Cleanup System Engineer is not a licensed SRO. Answer c is incorrect.

The clearance writer is usually a RO licensed individual. Answer d is incorrect.

2.3.2Knowledge of facility AL ARA program. (CFR: 41.12)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 65 of 135 OC Learning Objective: RWT (Objecti ve 22: Describe the Station ALARA Program).Cognitive Level: Memory or Fundamental

Question Type: New 9/8/06 Rossi reviewed. Changed stem to prov ide the necessary information while maintaining economy of words.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 66 of 13538.The plant was at rated power, when a loss of 125 VDC DC-E occurred, and cannot be restored. The applicable ABN has been entered Which of the following states why av ailable Non-Licens ed Operators are directed to perform plant tours?a.All fire protection mitigation systems have been disabledb.Power has been lost to all control room annunciators c.Automatic trip ability for f eedwater pumps, condensate pumps and main turbine has been lostd.Power is lost to all area radiation monitors and most process radiation monitors Answer: b Justification: The loss of DC-E results in the loss of all control room annunciators.

ABN-53 requires that plant operators tour plant areas where the annunciators are lost (plant wide). Answer b is correct.The ability of the fire protection system to mitigate fires is not effected, but fireannunciation is lost in the control room. Answer a is incorrect.

DC power for feedwater and condensat e pumps (on 4160 Bus A and 4160 Bus B) are powered from DC Bus B and Bus C. Answer c is incorrect.

Radiation monitors are power by various AC power supplies (mostly vital AC).

ARMs are powered by Continuous Inst rument Panel-3 (CIP-3) (reference procedure 407.1) Answer d is incorrect.

2.4.32 Knowledge of operator response to loss of all annunciators. (CFR: 41.10)

OC Learning Objective: 2621.828.0.

0012 (10450: Describe and interpret procedure sections and steps for plant emergency or off-normal conditions thatinvolve this system including personne l allocation and equipment operation IAW applicable ABN, SDRP, EOP & EO P Support Procedures and EPIPs.)Cognitive Level: Comprehensive or Analysis

Question Type: New

9/8/06 Rossi reviewed. Deleted: with all syst ems normally aligned in the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 67 of 13539.The reactor was operating at rated pow er when a loss of all offsite power occurred. Plant conditions are as follows: Reactor water level dropped to 82 inches and is currently 110 inches and rising slowly Reactor pressure dropped to 890 psig and is currently 950 psig and rising slowly Drywell pressure has risen to 2.2 psig and is stable Both EDG output breakers have been closed for two minutes Restoration of power to plant bus es IAW ABN-36-3, Plant Electrical Distribution Restoration, has NOT yet commenced.

Which one of the following statement s is true for these conditions?a.Drywell temperature is rising because there are no drywell recirc fans runningb.Instrument air pressure is lowering because there are no air compressors runningc.Reactor Building P is zero because no reac tor building ventilation fans are runningd.Service water temperatures are rising because there are no service water pumps running Answer: b Handouts: None Justification: A is incorrect - since ther e is no LOCA signal (Hi DW pressure AND Lo-Lo RPV level), drywell recirc fans 1, 3 and 5 auto-started 2.5 seconds after power was restored to buses 1C and 1D (and USS 1A2 and 1B2). Although RPV

level dropped below the Lo-Lo level setpoi nt (86 inches), a concurrent high drywell pressure must be received to prev ent drywell recirc fans from starting when 1C/1D bus power is restored.B is correct - on loss of all offsite power, all (4) 4160V buses de-energize. The UV logic for the 1C and 1D buses trips the load breakers on buses 1C and 1D as

well as the feeder breakers to USS 1A 1 and 1B1, which power Turbine Building loads. The feeder breakers for USS's 1A2/1B2 and 1A3/1B3 remain closedsince UV trip devices on individual loads provide load shedding for these USS's.

When power is restored to buses 1C and 1D by their respective EDG's, USS's 1A1 and 1B1 remain de-energized and USS's 1A2/1B2 and 1A3/1B3 automatically re-energize. Certain loads powered from these USS's are automatically sequenced on at various time intervals to prevent overloading the EDG's, including CRD pumps, RBCCW pumps and Service Water pumps. The

service and instrument air compresso rs are powered from USS 1A1/1B1 and NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 68 of 135 therefore require manual re storation. This is di rected by ABN-36 following restoration of power to USS 1A 1/1B1 IAW Attachment ABN-36-3.

C is incorrect - reactor building ventila tion fans tripped on the loss of offsite power, but the Standby Gas Treatment System started due to the Lo-Lo RPV level signal (SGTS does not require a concurrent high drywell pressure), restoring the Reactor Building negative P.D is incorrect - service water pumps aut omatically restart two minutes after power is restored to buses 1C and 1D with no LOCA signal present. A LOCA signal (preventing restart) requires Lo-Lo RPV level AND Hi DW pressure.

295003 AK1.02 Knowledge of the operational implications of the followi ng concepts as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Load Shedding (CFR 41.5)OC Learning Objective:

2621.828.0.0016, Objective E:

Describe the interlocks signals and setpoints for the affected system components and expected system response includi ng power loss or failed components.

2621.828.0.0016, Objective M:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2621.828.0.0016, Objective O:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation IAW appl icable ABN, EOP & EO P support procedures and EPIPs.Cognitive Level: Comprehension or Analysis

Question Type: New

References:

338, ABN-36, GE 223R0173, sh. 1A

8/15/06: NRC Comments Added that DW rose to 2.2 psig and stable.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 69 of 13540.Given the following: Reactor power is 28% with a power ascension in progress The main generator is loaded to 200 MW and 50 MVAR Stator cooling water pump 1A is tagged out for maintenance DC Bus A tripped due to a ground fault The electrical fault on DC-A causes a trip of breaker 1B1M Which statement below describes t he effect of this event on main generator voltage regulation?Generator terminal voltage ____________

a.must be maintained manuallyb.will be maintained automatically c.will rise since there is a lo ss of automatic and manual voltage controld.will lower since there is a lo ss of automatic and manual voltage control Answer: c Handouts: None

Justification: A and B are incorrect -

a loss of DC-A, which powers the main generator excitation equipment, results in a loss of both automatic and manual voltage control. The CAUTION for Step 3.

2 of ABN-53 states, in part, "Loss of Main Generator voltage control will result from a loss of power to DC DistributionCenter A." In addition, a NOTE for Step 3.3 of ABN-12 states, in part, "Loss of DC control power to the excitation swit chgear will result in a loss of generator voltage control."

C is correct - a trip of 1B1M causes a loss of USS 1B1, which results in a loss of the only available stator cooling water pump. When stator flow drops below 230 gpm, a generator runback occurs. This causes an automatic load reduction on the generator. Since reactor power is below 30%, the crew will not scram the reactor (ABN-11 directs a reactor sc ram if a generator runback occurs when reactor power is above 30%). If bel ow 30% power, ABN-11 directs reducing MVARs to zero. Since there is a loss of both automatic and manual voltagecontrol, the operator will be unable to reduce generator voltage/MVARs and as generator (real) load is reduced during t he runback, generator terminal voltage will increase.

D is incorrect - a loss of stator cooli ng results in a generator runback. As the generator unloads with no method of controlling voltage, generator terminalvoltage will increase.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 70 of 135 295004 G2.1.28 Partial or Complete Loss of D.C. Po wer / Conduct of Operations: Knowledge of the purpose and function of major system components and controls. (CFR 41.5, 41.7)OC Learning Objective:

2621.828.0.0025, Objective A:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation IAW appl icable ABN, EOP & EO P support procedures and EPIPs 2621.828.0.0025, Objective G:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2621.828.0.0025, Objective H:

Identify and explain system operating controls/indicat ions under all plant operating conditions.

Cognitive Level: Comprehension or Analysis Question Type: New

References:

ABN-11, ABN-12, ABN-53

9/8/06 Rossi reviewed. Changed init ial bullet from 'A ground fault on DC Bus A causes it to de-energize' to 'DC BUS A has tripped due to a ground fault'.

9/22/06 Validator comments: ABN-11 says that a runback will occur at approximately 25% rated power (182 MW e). Changed the second bullet from 175 MW to 200 MW.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 71 of 13541.Initial plant conditions are as follows: A plant startup is in progr ess with reactor power at 12% The mode switch is in STARTUP Recirculation flow is 11 E4 gpm Reactor pressure is 1000 psigA turbine bypass valve malfunction causes: A spike in reactor pressure to 1043 psig A spike in reactor power to 40%

What is the status of the r eactor (assume no operator action)?a.At powerb.Scrammed due to high reactor pressure c.Scrammed due to high IRM neutron flux d.Scrammed due to high APRM neutron flux Answer: c Handouts: Nome

Justification: A is incorre ct - the reactor scrammed due to high IRM neutron flux.

B is incorrect - from RAP-H1f, the high r eactor pressure scram setpoint is 1045 psig.

C is correct - based on the conditions given (STARTUP, at 12% power), the

reactor is operating on IRM Range 10.

The scram setpoint for IRM Range 10 is38.4% (LSSS), which was exceeded.

D is incorrect - the APRM Hi-Hi scram se tpoint with recirc flow at 11 E4 gpm would be greater than 60%.

295006 AK2.06 Knowledge of the interrelations bet ween SCRAM and the following: Reactor power (CFR 41.5, 41.6)

OC Learning Objective:

2621.828.0.0037, Objective C:

Describe all RPS scram logic trip signals, including the following:1. Purpose / Design Basis

2. Setpoints

3. Conditions that allow bypassing scram signals

4. How bypassing scram signals is accomplished NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 72 of 135 Cognitive Level: Comprehension or Analysis Question Type: Modified Bank

References:

201, RAP-H1f, Tech Spec 2.3

8/15/06: NRC Comments Added no operator action.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 73 of 13542.Given the following: The reactor is operating at rated power on a hot summer day RBCCW & TBCCW heat exchangers are being cooled by Service Water RBCCW pump 1-1 and heat exchanger 1-1 are in service RBCCW heat exchanger 1-2 is tagged out due to a tube leak RBCCW temperatures have been trending upward The crew has entered ABN-19, RBCCW Failure ResponseWhich one of the following actions can be utilized to reduce RBCCWsystem temperatures?a.Place RBCCW pump 1-2 in service along with pump 1-1b.Increase Reactor Water Cleanup regenerative heat exchanger flowc.Lineup the A & B Fuel Pool C ooling heat exchangers to be cooled by TBCCWd.Lineup the TBCCW heat exchangers to be cooled by the Circulating Water System Answer: d Handouts: None

Justification: A is incorrect - in order to prevent flow-induced vibration (due toexcessive flow) in the RBCCW heat exchangers, Procedure 309.2 requires two RBCCW heat exchangers to be in service prior to placing a second RBCCW

pump in service.

B is incorrect - this can't be done without increasing RWCU system flow rate, which would increase the heat load on the RBCCW system.

C is incorrect - the A & B FPC heat exchangers cannot be aligned to be cooled by TBCCW; only the augmented (C) FPC heat exchanger can.

D is correct - ABN-19 directs this action Operating with 2 RBCCW pumps in servic e requires 2 heat exchangers to be in service.295018 AA1.01 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Backup systems (CFR 41.4, 41.10)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 74 of 135 OC Learning Objective: 2621.828.0.0035, Objective P:

Describe and interpret procedure sections and steps for plant emergency and off-normal conditions that involve this system including personnel allocation and equipment operation in accordance with plant procedures.

Cognitive Level: Comprehension or Analysis

Question Type: New

References:

ABN-19, 309.2, 311.1, BR 2006 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 75 of 13543.Given the following: The reactor is SHUTDOWN and a cooldown is in progress Reactor pressure is 15 psig; reactor water level is 160 inches Reactor recirculation system status is as follows:

o Loops 'A' and 'C' are ISOLATED o Loops 'B' and 'D' are IDLE o Pump 'E' is in service Shutdown cooling pumps 'A

' and 'B' are in service The auxiliary reactor water cleanup pump is in service Bus 1A de-energizes due to an electrical fault The crew places shutdown cooling pump 'C' in service What action should be taken regarding the reactor recirculation system?a.OPEN the 'B' pump discharge valve then CLOSE the 'E' pumpdischarge valveb.CLOSE the 'E' pump discharge va lve, then OPEN the 'B' pumpdischarge valvec.OPEN the 'D' pump discharge valve then CLOSE the 'E' pumpdischarge valved.CLOSE the 'E' pump discharge va lve, then OPEN the 'D' pumpdischarge valve Answer: c Handouts: None

Justification: A is incorrect - P&L 4.2.12 in Procedure 305, Shutdown Cooling System Operation, states "If the Cleanup System is in service, the B Recirc loop should not be the selected loop in those instances where one loop is required to be fully open." NOTE: the auxiliary reac tor water cleanup pump is powered from MCC 1B21 (Bus 1B) and therefore remains in service on loss of Bus 1A.

B is incorrect - for the reason stated abov e for choice A. In addition, and more importantly, closing the E pump discharge va lve would violate Tech Spec 3.3.F.4, which states "With reactor cool ant temperature greater than 212 F and irradiated fuel in the reactor vessel, at least one recirculation loop discharge valve and its associated suction valve s hall be in the open position."

C is correct - a loss of Bus 1A causes a trip of recirc pump E. P&L 4.2.11 inProcedure 305 states: " To prevent SDC Syst em flow from short-cycling the core, the E Recirc Loop Discharge Valve must be CLOSED or the E Recirc Pump running." This statement requires the operator to close the E recirc pumpdischarge valve due to the pump trip. Ho wever, Tech Spec 3.3.F.4 requires atleast one recirc loop suction and associated discharge valve to be open. To NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 76 of 135 meet both of these requirements, the co rrect action to take would be to open the D pump discharge valve, then cl ose the E pump discharge valve.

D is incorrect - this violates Tech S pec 3.3.F.4-see explanation for choice B above.295021 AA1.05 Ability to operate and/or monitor t he following as they apply to LOSS OFSHUTDOWN COOLING: Reactor recirculation (CFR 41.10)

OC Learning Objective:

2621.828.0.0038, Objective J:

Given normal operating procedure and doc uments for the system, describe or interpret the procedural steps.

2621.828.0.0038, Objective M:Given Technical Specifications, identify and explain associated actions for each section of the Technical Sp ecifications relating to th is system including personnelallocation and equipment operation.

Cognitive Level: Comprehension or Analysis

Question Type: New

References:

ABN-2, ABN-3, 301.2, Tech Spec 3.3.F.4 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 77 of 13544.The reactor was manually scrammed due to a steam leak into the Primary Containment. Current plant conditions are as follows: Reactor water level is 140 inches and steady Reactor pressure is 600 psig and lowering slowly Drywell pressure is 14 psig and rising slowly Drywell temperature is 210 F and rising slowly Torus pressure is 13 psig and rising slowly Torus water level is 153 inches and steady Torus water temperature is 110 F and rising slowly Which of the following alarms would be shown on the PPC SPDS screens for the Primary Containment?a.A RED Priority 1 alarm due to torus water temperature above 106° Fb.A RED Priority 1 alarm due to drywell temperature above 200 Fc.A YELLOW Priority 2 alarm due to torus water level above 152"d.A YELLOW Priority 2 alarm due to drywell or torus pressure above 12 psig Answer: d Justification (See OC-PPC-SRS-0001, S ystem Requirements Specification forthe Oyster Creek Safety Parameter Dis play System) IAW the reference, the following alarm priority 1 or 2 are activated when: Torus/DW pressure > that which ex ceeds PSP in EOPs: RED Priority 1 Torus/DW pressure > 12 psig: YELLOW Priority 2 Torus water level < 110": RED Priority 1 Torus water level exceeds torus load limit in EOPS: RED Priority 1 Torus water level <143" or > 154" (TS limit): YELLOW Priority 2 Torus water temperature exceeds HCTL in EOPS: RED Priority 1 Torus water temperature > 96° F with power > 2%: RED Priority 1 Torus water temperature > 95° F: YELLOW Priority 2 DW temperature > 281° F: RED Priority 1 DW temperature > 200° F: YELLOW Priority 2 With torus water temperature at 106° F, the alarm should be yellow priority 1 (also, since HCTL is not exceeded, and all control rods are inserted). Answer a is incorrect.

A DW temperature of 210 F would only be a yellow priority 2 alarm. Answer b is incorrect.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 78 of 135 A torus water level of 153" would present no alarms. Answer c is incorrect.

A torus/DW pressure > 12 psig (and less that PSP) would only be a yellow priority 2 alarm. Answer d is correct.

295024 EK2.16 Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: SPDE/ERIS/CRIDS (CFR 41.7, 41.10)

OC Learning Objective:

2621.863.0.0007, Objective O:

Discuss the relevance of information s hown on the PPC SPDS displays to the implementation of SBEOPs.

Cognitive Level: Memory of Fundamental Question Type: New

References:

EMG-3200.02, OC-PPC-SRS-0001

8/15/06: NRC CommentsThe original question did hit the K/A. The new question is as above.

9/8/06 Rossi reviewed. Replaced '(all rods in) from rated power due to a steam loss of coolant accident' with 'due to a steam l eak into the Primary Containment' in the stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 79 of 13545.The reactor was operating at rated pow er when a loss of all offsite power occurred. The transient resulted in EMRV actuation and one EMRV stuck open. Current plant conditions are as follows: Reactor water level is 80 inches and lowering slowly Reactor pressure is 350 psig and lowering slowly Torus water temperature is 145 F EDG-1 is loaded to 1400 KW EDG-2 is loaded to 1800 KW Which Containment Spray pumps, and associated ESW pumps, should be placed in the torus cooling mode?a.Two Containment Spray pumps and two ESW pumps in System 1b.Two Containment Spray pumps and two ESW pumps in System 2 c.One Containment Spray pump and one ESW pump in either System 1 OR 2d.Two Containment Spray pumps and two ESW pumps in both System 1 AND 2 Answer: a Handouts: Attachments 341-5 and 341-6 Justification: A is correct - based on the given conditions, the Primary Containment Control EOP directs plac ing both Containment Spray Systems in the Torus Cooling mode. EDG-1 has suffic ient capacity to carry two containment spray pumps and two ESW pumps, while EDG-2 does not. In addition, since 2

Core Spray pumps are running (due to Lo-Lo level) and getting ready to inject

when RPV pressure drops below ~310 psig, securing Core Spray to run 4 Containment Spray pumps is not an opt ion. Therefore, since only one Containment Spray System can be placed in service, System 1 is the correct choice.B is incorrect - a CAUTION in Suppor t Procedure 25 states

"Diesel Generator overload will result if a Containment Spray Pump and ESW pump are started with a Diesel Generator load of greater than 2150 KW."

Since EDG-2 is already loaded to 1800 KW, and a containment spra y/ESW pump combination will add ~

580 KW (as shown in Attachment 341-6), EDG-2 does not have sufficient

capacity to carry System 2 (C and D) c ontainment spray pumps and System 2 (C and D) ESW pumps-it can only carry 1 containment spray pump and 1 ESW

pump.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 80 of 135 C is incorrect - according to the Primary Containment Control EOP, if torus water temperature cannot be maintained below 95 F, two containment spray systems (4 pumps) should be operated in torus cooling, if available. Current plant conditions prevent operating all four containment spray pumps, however two pumps can and should be placed in torus cooling.

D is incorrect - in addition to the EDG load restrictions mentioned above, there is a CAUTION in Support Procedure 25 that states: "NPSH problems will develop on all operating pumps if more than 4 Containment Spray/Core Spray Main pumps are operated at the same time." With RPV level at 80 inches, Core Spray Systems 1 and 2 (2 main pumps and 2 booster pumps) would be operating, but not injecting (core spray will begin to inject when RPV pressure is ~310 psig).

Since 2 core spray pumps are running, and are needed to restore RPV level

when RPV pressure drops below 310 psig , only 2 containment spray pumps canbe placed in service.

295026 EA1.01 Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppr ession pool cooling (CFR 41.10)

OC Learning Objective:

2621.828.0.0009, Objective L:

Given normal operating procedures and documents for the system, describe or interpret the procedural steps.

Cognitive Level: Comprehension or Analysis Question Type: New

References:

EMG-3200.02, Support Procedure 25, 341 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 81 of 13546.A large un-isolable leak developed in the torus while the reactor was operating at rated power. The timeline for torus water level is as follows (all (t) times are in minutes): At t = 0, Primary Containment C ontrol entry was required due to low torus water level At t = 10, makeup to the torus was commenced using Core Spray System 1 At t = 30, torus water level was 130 inches At t = 50, torus water level was 120 inches Which statement below is true for these conditions (assume no further operator actions)?a.At t = 70 minutes, PSP will be exceededb.At t = 70 minutes, the EMRV tailpipes begin to uncover c.At t = 80 minutes, Emergency Depressurization will be requiredd.At t = 110 minutes, the toru s vent header downcomers begin to uncover Answer: a Handouts: EMG-3200.02

Justification: A is correct -

the present rate of drop in torus level is 0.5 inches per minute (130 - 120 = 10 inches; 10 inches

/20 minutes = 0.5 inches/minute).

Therefore, in 20 more minutes (t = 70 minutes), level will have dropped to 110inches, which is the point at whic h the torus vent header downcomers are uncovered, and the point at which the pressure suppression function of the primary containment can no l onger be assured. It is fo r this reason that segment A-B of the PSP curve is vertical at 110 inches.

B is incorrect - the EMRV tailpipes ar e not uncovered until torus level reaches 90inches, which will occur at t = 110 minutes given the current rate of drop in torus level.C is incorrect - Emergency Depressuri zation is required BEFORE reaching 110 inches-at t = 80 minutes, level will have dropped to 105 inches.

D is incorrect - the torus vent header downcomers are uncovered at 110 inches-at t = 110 minutes, level will have dropped to 90 inches, which is where the EMRV downcomers start to uncover.

NOTE: the above calculations ignore the fa ct that torus water level will actually drop at a quicker rate due to the round shape of the torus.

295030 EA2.01 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 82 of 135 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: S uppression pool level (CFR 41.9, CFR 41.10)OC Learning Objective:

2621.828.0.0032, Objective J:

Identify and interpret normal, abnorma l, and Emergency O perating Procedures for Primary Containment.

2621.828.0.0032, Objective T:

Interpret Primary Containment indica tions in terms of limits and trends.

Cognitive Level: Comprehension or Analysis

Question Type: New

References:

EMG-3200.02, EOP Users Guide

8/15/06: NRC Comments Added no operator actions.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 83 of 13547.Plant conditions are as follows: An ATWS is in progress, with reactor power currently at 8% SLC System 1 is injecting RPV water level is being maintained between 0 and -20 inches Fuel Zone level indicators C and D have been turned on at Panel 4F Which Fuel Zone level instrument channels will be used to control RPV water level?a.A and Bb.B and D c.C ONLY d.D ONLY Answer: b Handouts: None

Justification: A is incorrect - FZLI channel A and C are not accurate when SLC is injecting.B is correct - B and D are both available and are providing accurate levelindication. FZLI channels A and C utilize the SLC injection line as the variable leg and are therefore not accurate when SLC is injecting. Note that when all recirc pumps are tripped, FZLI channel s A and B automatically turn on.

C is incorrect - FZLI channel A and C are not accurate when SLC is injecting.

D is incorrect - FZLI channel B is also available.

295037 EA2.02 Ability to determine and/or interpret the following as they apply to SCRAMCONDITION PRESENT AND REACTO R POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor water level (CFR 41.5, 41.7)

OC Learning Objective:

2621.828.0.0055, Objective D:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2621.828.0.0055, Objective I:

Explain or describe how this system is interrelated with other plant systems.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 84 of 135 Cognitive Level: Comprehension or Analysis Question Type: New

References:

EOP Users Guide

8/15/06: NRC Comments Added reactor power at 8% in the question stem.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 85 of 13548.The control room has been evacuated due to a fire. The fire has beenextinguished. ABN-29, Plant Fires, requires the following ventilationsystems shutdown prior to purging the control room. A and B 480V Switchgear Room Ventilation System A/B Battery Room, MG Set Room Ventilation System Chemistry Laboratory Ventilation System Reactor Building Ventilation System According to ABN-29, the reason this action is taken is to prevent smoke and fumes purged from the control r oom from being brought into these areas, which could _______________a.prevent personnel accessb.cause damage to equipment c.set off automatic fi re suppression systemsd.cause a reaction with other hazardous materials Answer: c Handouts: None Justification: A is incorrect - this is not the reason stated in ABN-29.

B is incorrect - this is not the reason stated in ABN-29.

C is correct - this is t he reason stated in ABN-29.

D is incorrect - this is not the reason stated in ABN-29.

600000 AK3.04 Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE: Actions contained in the abnormal procedure for plant fire on

site. (CFR 41.10)

OC Learning Objective:

2621.828.0.0019, Objective E:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation in accordance wi th applicable ABN, SDRP, EOP and EOP support procedures, and EPIPs.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 86 of 135 Cognitive Level: Memory or Fundamental Knowledge Question Type: New

References:

ABN-29 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 87 of 13549.Given the following: A plant startup is in progr ess with reactor power at 6% The steam chest and high pressure turbine are being warmed Reactor feed pump (RFP) A is in service feeding through LFRV A A FWLC failure causes RPV level to rise to 183 inches Which of the following occurs as a result of this failure?a.RFP A trips ONLYb.The main turbine trips ONLY c.RFP A and the main turbine tripd.Neither RFP A nor the main turbine trip Answer: b Handouts: None

Justification: A is incorrect - RFP A will not trip since ROPS is bypassed.

B is correct - the main turbine must be reset if the steam chest and HP turbine are being warmed and it will trip when RPV level reaches 175 inches. For RFP A to trip, the ROPS logic must see RPV level at 181 inches and feedwater flow greater than 2.23 E6 lbm/hr. Feedwater flow will not get this high when feeding through the LFRV, which is rated fo r a maximum of 1500 gpm (1500 gpm is equal to 0.72 E6 lbm/hr).

C is incorrect - the main turb ine will trip; RFP A will not trip.

D is incorrect - the main turbine will trip.

295008 G2.1.27 High Reactor Water Level / Conduct of Operations: Knowledge of system

purpose and or function (CFR 41.5, 41.7)

OC Learning Objective:

2621.828.0.0018, Objective A:

Given plant operating conditions, descri be or explain the purpose(s)/function(s) of the system and its components.

2621.828.0.0018, Objective D:

Describe the interlock signals and setpoi nts for the affect ed system components and expected system response includi ng power loss or failed components.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 88 of 135 Cognitive Level: Comprehension or Analysis Question Type: New

References:

201, 317, RAP-H5d, RAP-H7d, ABN-10 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 89 of 13550.The reactor was initially operating at rated power. A feedwater line break inside primary containment resulted in a high drywell pressure scram.

Current plant conditions are as follows: RPV level is 88 inches and rising slowly with both CRD pumps injecting RPV pressure is being maintain ed at 800-900 psig with Isolation Condensers Drywell pressure is 13 psig and lowering slowly with drywell sprays initiated Torus water temperature is 145 F and lowering slowly Torus water level is 168 inches and rising slowly Which of the following is the most immediate reason for lowering torus water level?

To prevent exceed ing the ____________________.a.Torus Load Limitb.Heat Capacity Temperature Limit c.Primary Containment Pressure Limit d.Maximum Pressure Suppression Primary Containment Water Level Answer: a Handouts: EMG-3200.02, or, providing the LARGE figures of the graphs would be preferred Justification: A is correct - based on a to rus water level of 168 inches and rising, and a reactor pressure of 800-900 psig, the Torus Load Limit is the most

immediate concern since it will be exceed ed before any of the other limits. From Figure E (TLL) of EMG-3200.02, the Torus Load Limit that corresponds to an

RPV pressure of 800 to 900 psig is ~174 to 178 inches.

B is incorrect - since torus water temperature is lowering, the margin to the HCTL is improving.

C is incorrect - for the given torus water level and torus pressure (drywell pressure), the PCPL is of no concern-to rus pressure would have to rise above 50 psig at the given torus level for this to be a concern.D is incorrect - the MPSPCWL is 188 inches, which makes it a secondary concern relative to the Torus Load Limit.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 90 of 135 295029 EK3.02 Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Low ering suppression pool water level (CFR 41.9, 41.10)

OC Learning Objective:

2621.828.0.0032, Objective J:

Identify and interpret normal, abnormal and Emergency Operating Procedures for the Primary Containment System Cognitive Level: Comprehension or Analysis

Question Type: New

References:

EMG-3200.02, EOP Users Guide NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 91 of 13551.Consider an event in which an acci dent causes a high-energy radioactivesystem to discharge into the Reactor Building.Assuming the radioactivity release into the Reactor Building is the same in each case, which of the following results in the highest off-site release rate?Reactor Building P Standby Gas Treatment flowa.-0.10 inches WG2600 scfmb.0.0 inches WG0 scfm c.0.10 inches WG0 scfm d.0.10 inches WG2600 scfm Answer: d Handouts: None

Justification: A is incorrect - in this case there is a minimal release through SGTS, which is ~99% efficient. Since SGTS is able to maintain a negative RBP, there is no ground level release.

B is incorrect - with RB P at zero and no SGTS flow there is no release.C is incorrect - in this case only a ground level release is occurring since there is a positive pressure in the Reactor Building and no SGTS flow.

D is correct - since there is a positive pressure in the Reactor Building a ground level release is occurring, which is eq uivalent to the ground level release in choice C (based on the same RB P). In addition, since SGTS is not 100%

efficient (see UFSAR Table 6.5-1), there is some relatively small release through this path. Therefore, this case resu lts in the highest off-site release rate.

295035 EK2.03 Knowledge of the interrelations between SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE and the following:

Off-site release rate (CFR 41.8, 41.9)OC Learning Objective:

2621.828.0.0042, Objective F:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2621.828.0.0042, Objective L:

Explain or describe how this system is interrelated with other plant systems.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 92 of 135 Cognitive Level: Comprehension or Analysis Question Type: New

References:

UFSAR Table 6.5-1

8/15/06: NRC Comments Corrected answer choice labels from a, b, d, c to a, b, c, d. Answer d is correct.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 93 of 13552.Given the following: The reactor is shutdown due to a forced outage Reactor water level is 165 inches on NR GEMAC Three (3) Shutdown Cooling pumps are in service Total Shutdown Cooling System flow is 7500 gpm According to Procedure 305, Shutdown Cooling System O peration, raising shutdown cooling system flow rate may result in-a.flow-induced vibration of the shutdown cooling heat exchangersb.damage to the nuclear instrumentation due to flow-induced vibrationc.Spurious trips of the shutdown cooling pumps due to low suction pressured.Exceeding the maximum design t ube-side flow rate of the SDC heat exchangers Answer: c Handouts: None Justification: A is incorrect - Procedur e 309.2 has a P&L to limit RBCCW flow to less than 3700 gpm to prevent damage to the RBCCW heat exchanger from flow induced vibration. This is not relat ed to SDC (tube side) flow but could be a misconception.

B is incorrect - this is related to a precaution associated with Reactor Recirculation System operation (P&L 5.2.8 of Procedure 301.2), not Shutdown Cooling.C is correct - as stated in 305, "Sim ultaneous operation of all three (3) SDC Pumps at high flow rates (System Flow

>7500 gpm) may result in pump suction pressures near the trip setpoint (4 psig).

To avoid spurious pump trips, operation at system flow > 7500 gpm should be minimized.

D is incorrect - according to 305, t he maximum design tube side flow rate is 3400 gpm.205000 A1.02 Ability to predict and/or monitor changes in parameter s associated with operatingthe SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including: SDC/RHR pump flow (CFR 41.10)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 94 of 135 OC Learning Objective:

2621.828.0.0045, Objective P:

Identify and explain the normal operating procedures for the Shutdown Cooling System.Cognitive Level: Memory of Fundamental

Question Type: New

References:

301.2, 305, 309.2 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 95 of 13553.Which of the following methods of makeup to the Isolation Condensers require operation of the A or B Isolation Condenser Makeup Valves, V-11-36 or V-11-34, on Panel 5F/6F?1. Adding makeup with Demineralized Water IAW 307, Isolation Condenser System2. Adding makeup with Fire Protec tion IAW 307, Isolation Condenser System (NOT from local hose stations)3. Adding makeup with Fire Protection via local hose stations IAW 307, Isolation Condenser System4. Core Spray makeup to the Isolation Condensers IAW 308, Emergency Core Cooling System Operationa.1 and 2b.3 and 4 c.1 and 3 d.2 and 4 Answer: d Handouts: None

Justification: A, B and C are incorre ct - makeup with Demineralized Water (choice 1) is the normal method of she ll makeup when the IC's are in standby.This method fills the IC shells via grab sample lines-this flow path does not utilize V-11-36 or V-11-34. Makeup from Fire Protection via local hose stations (choice 3) utilizes the IC shell drain lines as the makeup flow path-does not

utilize V-11-36 or V-11-34.

D is correct - adding makeup with Fire Protection IAW 307 (choice 2), and Core Spray makeup IAW 308 (choice 4), are the only methods (of those given) that utilize makeup valves V-11-36 or V-11-34, which are the normal makeup supply valves from the Condensate Transfer System.

207000 A4.06 Ability to manually operate and/or monitor in the control room: Shell side makeup valves (CFR 41.8, 41.10)

OC Learning Objective:

2621.828.0.0023, Objective C:

Describe or trace (given a simplified drawi ng or P&ID) the basic flow path for the following modes of Isolation Condenser operation:

1. Standby

2. Emergency Operation

3. Sources of Shell Side Makeup NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 96 of 135 Cognitive Level: Memory of Fundamental Question Type: New

References:

307, 308 GE 148F262

8/15/06: NRC Comments They wanted me to verify that answe rs b and d did require the use of theisolation condenser makeup valves V-11-34 or V-11-36. See pages 26, 29 of 307, and page E12-2 of 308. These valves do need to be operated to fill the ICshells from fire water/core spray. Later, it was realized that the question is asking for the operation of both IC makeup valves. As written, both valves may not besimultaneously opened. The question was changed to OR instead of and.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 97 of 13554.Given the following: An automatic scram occurred while operating at rated power All control rods did NOT fully insert; reactor power is 13% Reactor pressure band is being maintained at 800 to 1000 psig Reactor water level band is being maintained at -20 to +30 inches Standby Liquid Control (SLC) System #1 is injecting into the RPV Which one of the following conditions ensures adequate SHUTDOWN MARGIN?Length of Time SLC has been Injecting SLC Tank Concentrationa.30 minutes12 weight percentb.60 minutes12 weight percent c.25 minutes15 weight percent d.35 minutes15 weight percent Answer: b Handouts: Tech Spec 3.2

Justification: A is incorrect - 30 gpm for 30 minutes yields 900 gallons of boron solution at 12 weight percent. This is outside the shaded area of Tech Spec Figure 3.2-1, which means insufficient boron solution would have been injected to ensure adequate SHUTDOWN MARGIN.

B is correct - SLC pump capacity is 30 gpm. 1800 gallons of boron at 12 weight percent would have been injected after 60 minutes. This is within the shaded area of Tech Spec Figure 3.2-1, which r epresents the acceptable values of liquid control tank volume and solution concentration which assure that, with one 30

gpm liquid control pump, the reactor can be brought to the cold shutdown condition from a full power steady state operating condition at any time in core life independent of the cont rol rod system capabilities. (The cross-hatched area of Figure 3.2-1 represents the acceptable values of liquid control tank volume and solution concentration which assure t hat the equivalency requirements of 10 CFR 50.62-ATWS Rule-are satisfied. Note the Tech Spec definition of SHUTDOWN MARGIN is: "-the amount of r eactivity by which the reactor would be subcritical when the control rod with the highest reactivity worth is fully withdrawn, all other operable c ontrol rods are fully insert ed, all inoperable control NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 98 of 135 rods are at their current position, r eactor water temperature is 68ºF, and the reactor fuel is xenon free. Determinat ion of the control rod with the highest reactivity worth includes consideration of any inoperable control rods which are not fully inserted."

C is incorrect - this is outside the s haded area of Tech Spec Figure 3.2-1 (750 gallons at 15%).

D is incorrect - 1050 gallons of 15 wei ght percent of boron is outside the shaded area of Tech Spec Figure 3.2-1.

211000 K5.03 Knowledge of the operational implications of the followi ng concepts as they apply to STANDBY LIQUID CONTROL SYSTE M: Shutdown margin (CFR 41.6)

OC Learning Objective:

2621.828.0.0046, Objective A:

Given plant operating conditions, descri be or explain the purpose(s)/function(s) of the system and its components.

2621.828.0.0046, Objective N:Given Technical Specifications, identify and explain associated actions for each section of the Technical Sp ecifications relating to th is system including personnelallocation and equipment operation.

Cognitive Level: Comprehensive or Analysis Question Type: New

References:

UFSAR 9.3.5, EOP Users Gu ide, Tech Spec 1.45/3.2.C, 612.4.001 8/15/06: NRC Comments They thought that answers c and were direct lookup. They suggested using

injection times for these answers, and t he question was modified to that shown.

9/8/06 Rossi reviewed. Placed answers in 3-column format.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 99 of 13555.Given the following: A reactor startup is in progress Reactor pressure is 500 psig Reactor water level is 160 inches Reactor power is 20% on IRM Range 8 Two steam jet air ejectors (SJAE) are in service The steam chest and high-pressure turbine are being warmed A spurious reactor isolation occurs Which of the following describes t he plant response and/or the correct action for this event?a.Commence a normal plant shutdown IAW 203, Plant Shutdownb.An automatic scram occurs, enter ABN-1, Reactor Scram ONLYc.An automatic scram occurs, ent er ABN-1, Reactor Scram and RPV Control - No ATWSd.Place the startup on hold until the failure is corrected, then re-open the MSIVs IAW 301.1, Main Steam Supply System Answer: b Handouts: None

Justification: A is incorrect

- an automatic scram will occur.

B is correct - the reactor isolation resu lts in closure of all MSIVs and main steam line drains (in addition to some other valv es). At 500 psig (<600 psig), the MSIV closure scram is bypassed. Although ther e is relatively little steam flow, the reactor isolation will cause reactor pressure to rise. As pre ssure rises, power will also rise due to collapsing steam voids. The pressure rise will cause reactor

power to increase to the IRM Hi-Hi scram setpoint, which is 38% on IRM Range

8. This will occur before pressure rises to the high-pressure scram setpoint of 1045 psig. The automatic scram will termi nate the pressure rise. NOTE: asstated in Tech Spec 2.3 (LSSS) Bases, "Below 600 psig, when the MSIV closure scram is bypassed, scram protection is provided by the IRMs." For the given conditions (at this point in the star tup), there would be one feedwater pump in service and with relatively low steam fl ow and feed flow, there would not be asignificant change in RPV level due to the MSIV closure. Since RPV level remains above 138 inches, and RPV pressure remains below 1060 psig, there

are no EOP entry conditions.

C is incorrect - the transient will not result in any EOP entry conditions.

D is incorrect - an automatic scram will occur.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 100 of 135 212000 A2.11 Ability to (a) predict the impacts of the following on the REACTOR PROTECTIONSYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Mainsteamline isolation valve closure (CFR 41.5, 41.6)

OC Learning Objective:

2621.828.0.0037, Objective D:

Describe all RPS scram logic trip signals, including the following:1. Purpose / Design Basis

2. Setpoints

3. Conditions that allow bypassing scram signals

4. How bypassing scram signals is accomplished

2621.828.0.0037, Objective F:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2621.828.0.0037, Objective N:

Describe and interpret procedure sections and steps for plant emergency or off-normal conditions that involve this system including personnel allocation and equipment operation IAW applicable ABN, SDRP, EOP & EOP support procedures, and EPIPs.Cognitive Level: Comprehensive or Analysis

Question Type: New

References:

ABN-1, EMG-3200.01A, Tech Spec 2.3 Bases, 237E566 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 101 of 13556.Given the following: A reactor startup is in progress IRM Range 6/7 correlation is requir ed by Procedure 201, Plant Startup Which of the following support personnel, if any, are required to be notified prior to performing this task IAW Pr ocedure 402.2, IRM Operation During Startup?a.An I&C Technician ONLYb.A Reactor Engineer ONLY c.An I&C Technician and a Reactor Engineerd.No support personnel are needed, this task is performed by Operations ONLY Answer: a Handouts: None

Justification: A is correct - Prerequisi te 3.2 of Procedure 402.2 requires I&C to be notified to perform the IRM Range 6/7 correlation.

B is incorrect - a Reactor Engineer is not needed for IRM Range 6/7 correlation, but is needed to perform IRM calibration IAW Procedure 1001.9.

C is incorrect - only an I&C Technician is needed for IRM Range 6/7 correlation.

D is incorrect - an I&C Technician is required to support per formance of this task. Specifically, they perfo rm any required IRM adjustments.

215004 G2.1.14 Intermediate Range Monitor (IRM) System

/ Conduct of Operations: Knowledge of system status criteria which require the notification of plant personnel. (CFR 41.10)OC Learning Objective:

2621.828.0.0029, Objective K:Given Technical Specifications, identify and explain associated actions for each section of the Technical Specifications relating to this system including personnelallocation and equipment operation.

Cognitive Level: Memory or Fundamental

Question Type: New

References:

201, 402.2 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 102 of 13557.Given the following: Reactor power is 70% Five recirc loops are in service and total recirc flow on panel 4F is 15.0 E4 gpm The "C" recirc loop flow transmitter that feeds the Total Recirc Flowindicator on panel 4F fails to 0 (zero)Recirc flow as displayed on panel 4F will read __(1)__ E4 gpm, and will result in a ___(2)___.a.(1) 12.0 (2) rod blockb.(1) 12.0(2) scramc.(1) 13.5 (2) rod blockd.(1) 13.5(2) scram Answer: a Handouts: Attachment 202.1-2

Justification: A is correct - the total recirc flow indicator on Panel 4F receives a signal from the flow monitor on Panel 5R, which inputs to APRM Channels 5 through 8. Prior to the failure, each fl ow transmitter was sensing approximately 3.0 E4 gpm, which is summed to produc e 15.0 E4 total recirc flow. Onetransmitter failing to zero results in a tota l indicated recirc flow of 12.0 E4 gpm.

This produces a 10% mismatch between the RPS Division 1 and RPS Division 2 recirc flow monitors, causing a flow comparator rod block.

B is incorrect - reactor power at 70% is well below the scram setpoint for recirc flow at 12.0 E4 gpm-the scram set point from Attachment 202.1-2 is approximately 104% power.

C and D are incorrect - one could arrive at 13.5 gpm if they thought there were 10 recirc flow inputs to the total recirc flow indicator on 4F, vice only 5.

215005 A1.04 Ability to predict and/or monitor changes in parameter s associated with operating the AVERAGE POWER RANG E MONITOR/LOCAL PO WER RANGE MONITORSYSTEM controls including: SCRAM and rod block trip setpoints (CFR 41.7)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 103 of 135 OC Learning Objective:

2621.828.0.0029, Objective F:

Describe the interlock signals and setpoi nts for the affect ed system components and expected system response includi ng power loss or failed components.

2621.828.0.0029, Objective G:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

Cognitive Level: Comprehension or Analysis

Question Type: Bank

References:

202.1, 420, UF SAR 7.5.1.8.7, RAP-H7a NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 104 of 13558.The plant was at rated power w hen an event occurred which required a manual scram. All scram-related systems function as designed.

Moments later, the Shift Manager declared the Control Room un-inhabitable and must be evacuated.

Prior to leaving the Control Room, the following switch positions are noted: The front-panel control switch for EMRV NR108A is in OFF The back-panel NORMAL-DISABLED switch for EMRV NR108B is inDISABLED Which of the following is correct r egarding the ability of the EMRVs to function in the Pressure Relief m ode and to perform the ADS function in the event of a small-break LOCA?ADS Function Pressure Relief Modea.ALL EMRVsONLY EMRVs B, C, D, and Eb.ALL EMRVsONLY EMRVs A, C, D, and Ec.ONLY EMRVs A, C, D, and EONLY EMRVs C, D, and Ed.ONLY EMRVs B, C, D, and EONLY EMRVs B, C, D, and E Answer: c Justification: (see drawing 729E182) With the EMRV control panel switch if OFF, the associated pressure switch will not function to open the EMRV on high reactor pressure (relief m ode), but the ADS function re mains unaffected. With theswitch in DISABLED, all functions of the associated EMRV are defeated.

Therefore, EMRV A (in OFF) can functi on in the ADS mode but not in the reliefmode. EMRV B (in DISABLED) will not work in any mode. Answer c is correct, and the other selections are incorrect since they list the incorrect valves with their available modes.

218000 K3.01 Knowledge of the effect that a loss or malfunc tion of the AUTOMATICDEPRESSURIZATION SYSTEM will have on following: Restoration of reactor water level after a break that does not depressurize the reactor when required (CFR 41.7)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 105 of 135 OC Learning Objective:

2621.828.0.0005, Objective I:

Describe the operation of the ADS contro ls including: Removal of ADS control logic fuses to close EMRVs.Cognitive Level: Comprehensive or Analysis

Question Type: New

References:

GE 729E182

8/15/06: NRC Comments As originally written, they thought it was too easy: fuse removal gives it away.

This question has been rewritten.

9/8/06 Rossi reviewed. Broke in itial paragraph into 2 paragr aphs. Placed answers in 3-column format.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 106 of 13559.Given the following: The reactor is operating at rated power Pressure switch PS-1A83A fails low Using the attached drawing, GE 148F 712 (see coordinates G-8), what is the effect of this failure?a.One of the five ADS/EMRVs will NOT actuate in the ADS modeb.One of the five ADS/

EMRVs will NOT actuate in the Pressure Relief modec.Two of the five ADS/EMRVs wi ll NOT actuate in the ADS moded.Two of the five ADS/EMRVs will NO T actuate in the Pressure Relief mode Answer: b Handouts: GE 148F712 (ensure large enough to read)

Justification: A is incorrect - since t he failed pressure switch senses reactor pressure, and ADS functions on RPV leve l and Drywell pressure only, the ADS mode is not affected by this failure.

B is correct - PS-1A83A provides a high reactor pressure signal to EMRVNR108A. If this pressure switch fails low, EMRV A will not open on high reactor

pressure (1065 psig).

C is incorrect - since the failed pressure switch senses reactor pressure, and ADS functions on RPV leve l and Drywell pressure only, the ADS mode is not affected by this failure.

D is incorrect - although 2 of 5 EMRV's open at 1085 psig, and 3 of 5 EMRV's open at 1105 psig, each EMRV has a dedicated pressure switch that provides a high reactor pressure signal to the re spective EMRV actuation logic. PS-1A83A provides a reactor pressure signal to EMRV NR108A only.

239002 K6.01 Knowledge of the effect t hat a loss or malfunction of the following will have on the RELIEF/SAFETY VALVES: Nuclear bo iler instrument system (pressureindication) (CFR 41.7)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 107 of 135 OC Learning Objective:

2621.828.0.0005, Objective E:

Describe the EMRV initiation logic for both over-pressure operation and operation in the ADS mode. Include the following:

1. Initiation signals and setpoints

2. Timers and setpoints

3. Control switches

4. Panel indications

2621.828.0.0005, Objective J:

State how the following systems interrelate with ADS:

1. Vessel and Primary Cont ainment Instrumentation2. Core Spray

3. NSSS

4. Vital AC Power

5. 125 VDC Power Cognitive Level: Comprehension or Analysis

Question Type: New

References:

GE 148F712, GE 729E182, sh. 1, UFSAR 5.2.2.4

8/15/06 NRC Comment They questioned wether a single failure could effect 2 EMRVs. Since 2 EMRVsopen at the same set pressure, one pressure switch failure could possibly be thought to effect 2 EMRVs. The question remained as-is.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 108 of 13560.Given the following: The reactor is operating at rated power when a small leak develops inside the drywell Drywell temperature is 175 F and drywell pressure is 2.6 psig; both are rising slowly The Unit Supervisor directs venti ng the primary containment using the Standby Gas Treatment System (SGTS) IAW Support Procedure 31 Which statement below describes how venting the primary containment IAW Support Procedure 31 affects t he suppression chamber-to-drywell vacuum breakers and the reacto r building-to-suppression chamber vacuum breakers? Assume the v enting evolution causes containment pressure to lower.

Venting from the ___(1)___ coul d cause the _____(2)_____ vacuum breakers to open.a.(1) torus (2) suppression chamber-to-drywellb.(1) torus (2) reactor building-to-suppression chamberc.(1) drywell (2) suppression chamber-to-drywelld.(1) drywell (2) reactor building-to-suppression chamber Answer: c Handouts: None

Justification: A is incorrect - for the suppression chamber-to-drywell vacuum breakers to open, suppression chamber pr essure must exceed drywell pressure by at least 0.5 psid. For the given conditions, venting from the torus will cause suppression chamber pressure to remain below drywell pressure.

B is incorrect - for the reactor buildi ng-to-suppression chamber vacuum breakers to open, suppression chamber pressure must be at least 0.5 psid less thanReactor Building pressure, which is appr oximately at atmospheric pressure (-

0.25" WG). For the given conditions, venting the torus (to atmosphere) will not cause suppression chamber pressure to go below Reactor Building pressure by 0.5 psid.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 109 of 135 C is correct - venting from the drywell will cause drywell pressure to lower relative to suppression chamber pressure and if drywell pressure is 0.5 psid less

than suppression chamber pressure, the suppression chamber-to-drywellvacuum breakers will open.

D is incorrect - for the given conditions , venting from the drywell will not cause suppression chamber pressure to go below Reactor Building pressure.

261000 A1.06 Ability to predict and/or monitor changes in parameter s associated with operating the STANDBY GAS TREATMENT SYSTEM controls including: Drywell and suppression chamber differential pressure: Mark-I (CFR 41.7, 41.9)

OC Learning Objective:

2621.828.0.0042, Objective M:

Describe and interpret procedure sections and steps for plant emergency or offnormal conditions that involve this system including personnel allocation and equipment operations IAW applicable ABN, SDRP, EOP and EOP support procedures and EPIPs.

Cognitive Level: Comprehension or Analysis

Question Type: New

References:

EMG-3200.02, EOP Users Guide, Fundamentals NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 110 of 13561.The reactor is operating at rated power when Bus 1C undervoltage relay 27-13C fails low. Annunciator BUS 1C VOLTS LO goes into alarm. Bus 1C indications on Panel 8F/9F are normal.

How does EDG # 1 respond to this event?

EDG #1 ________

a.remains in standbyb.starts and idles at 400 RPM c.fast starts, output breaker closes d.fast starts, output breaker does NOT close Answer: a Handouts: None

Justification: A is correct - the Bus 1C (and 1D) undervoltage relays are arranged in a two-out-of-three logic scheme, which requires any two relays to trip to disconnect the bus from its normal source (Bus 1A), actuate bus load

shedding, and start the EDG.

If a single relay drops out on undervoltage, theannunciator will go into alarm, but the aut omatic actions described above will not occur until a second relay drops out on undervoltage (or fails low).B is incorrect - EDG #1 will idle or fast start as needed.

C and D are incorrect - EDG #1 will not fast start until/unless at least one of the other two bus undervoltage relays (27-11C,27-12C) drop out on low voltage.

262001 K3.02 Knowledge of the effect that a loss or malfunction of the A.C. ELECTRICALDISTRIBUTION will have on following: Emergency generators (CFR 41.7)

OC Learning Objective:

2612.828.0.0013, Objective C:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2612.828.0.0013, Objective I:

Describe the interlock signals and setpoi nts for the affect ed system components and expected system response includi ng power loss or failed components.

2612.828.0.0013, Objective N:

State the function and interpretation of system alarms, alone and in combination,as applicable in accordance with the system RAPS.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 111 of 135 Cognitive Level: Comprehension or Analysis Question Type: New

References:

RAP-T3a, UFSAR 8.3.1.1.1

8/15/06 NRC Comments They thought that selection b (is prevented from starting) was not plausible. This selection has been changed to that shown.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 112 of 13562.Which one of the following annunciators would be accompanied by a loss of power to Main Steam Line Radi ation Monitors RN06A & RN06B on Panel 2R?a.IP-4 PWR LOSTb.CIP 3 PWR LOST c.24VDC PP-A PWR LOST d.PROT SYS PNL 1 PWR LOST Answer: d Handouts: None

Justification: A is incorrect - none of the MSL radiation m onitoring equipment is powered by IP-4 (see ABN-58).

B is incorrect - CIP-3 provides power to the MSL radiation monitor recorders on Panel 10F (see ABN-58).

C is incorrect - none of the MSL radiat ion monitoring equipment is powered by 24 VDC (see 340.2).

D is correct - MSL Radiation Monito rs RN06A & B on Panel 2R are powered from Protection System Panel

1. 1, breaker #10 (see 406.1).

262002 K1.14 Knowledge of the physical connections and/or cause-effect relationshipsbetween UNINTERRUPTABLE PO WER SUPPLY (A.C./D.C

.) and the following:

Main steam line radiation monitors (CFR 41.7)

OC Learning Objective:

2621.828.0.033A, Objective G:

Explain or describe how this system is interrelated with other plant systems.

2621.828.0.033A, Objective L:

State the function and interpretation of system alarms, alone and in combination,as applicable in accordance with the system RAPS.

Cognitive Level: Memory or Fundamental

Question Type: New

References:

ABN-58, 406.1, 340.2, ABN-50 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 113 of 13563.Which one of the following shows t he correct correlation between the Emergency Diesel Generator gover nor mode of operation (droop, isochronous) and the positions of the EDG

Mode Selector (PTD) Switch?

P = Peaking T = Transfer D = Deadline Droop Isochronousa.PT and Db.P and TDc.DP and Td.T and DP Answer: b Handouts: None

Justification: B is correct - the thr ee-position PTD switch has the following functions: (1) PEAKING - sets up EDG to assume 2750 KW on a normal start.

Since this operation is in parallel with the grid, the governor would be in the DROOP mode. (2) TRANSFER - sets up governor control circuitry for load transfer. Load transfer occurs between the EDG and the grid, which again

means parallel operation and the governor is in the DROOP mode. (3)

DEADLINE - sets up EDG control circuits for isochronous operation.

A, C and D are incorrect - these are a ll incorrect combinations of the Mode Selector (PTD) Switch positions and the EDG governor modes of operation.

264000 K4.03 Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: Speed droop control (CFR 41.7)

OC Learning Objective:

2621.828.0.0013, Objective H:

Identify and explain system operating controls/indic ations under all plant operating conditions.

Cognitive Level: Memory or Fundamental

Question Type: New

References:

341 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 114 of 13564.The plant was at rated power, when the following annunciator came into alarm: ROD CNTRL - CONTROL AIR PRESS LO The following conditions are noted: INSTR AIR SUPPLY PRESS indicates 77 psig and lowering very slowly An NLO in the field reports the air receivers indicate 105 psig and steady Another NLO reports that the pre-filter Dp indicator is off-scale high There are no indications of any air leaks and there are no valve mis-positions.

Which of the following lists the effect on the air systems under the given conditions, and the next expected operat or action, as outlined in ABN-35, Loss of Instrument Air? (assume no oper ator actions other than that listed)

Effect Actiona.Service Air pressure will begin to decrease When Service Air drops to 90 psig, then confirm the lag compressor has startedb.Service Air pressure will begin to decrease When Service Air pressure drops to 90 psig, then confirm Service Air valve V-6S-2 is closed and is NOT bypassedc.Instrument air pressure will continue to degrade When Instrument Air pressure drops to 75 psig, then confirm that Service Air

valve V-6S-2 is open and is NOT

bypassedd.Instrument air pressure will continue to degrade When Instrument air pressure drops to 55 psig, then scram the reactor Answer: d Handouts: None

Justification: The indications provid ed show a normal pressure at the air receivers, but a degraded air pressure dow nstream of the pos t-filters going to instrument air. Control Room instrument air pressure indication is measured justdownstream of the pre-filters, air dryers and post-filters. The annunciator NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 115 of 135 provided in the stem is sensed downstream of where the Control Room indication is sensed. There is no air leakage an d no mis-positioned valves. The only possible cause for the degraded instrument ai r, is a plug of the pre-filters, or post-filters, or an air dryer failure. T hese components effect instrument air only.

Outside air if filtered, compressed and sent to the air receivers. Air from the air receiver outlet can either go to inst rument air (after passing through the pre-filters, the air dryers, and t he post-filters) or to servic e air (which is not filteredand dried). From the given indications, t here is no abnormality with the service airside of the system. Therefore, there is no low air pressure sensed in service air,and none is expected. This would make answers a and b incorrect.

Answer b is also not a correct action from ABN-35. Answer b is incorrect.

Answer c is incorrect in that it says to confirm the service air valve open, when it should be confirmed closed.

Answer d is correct in that it refers to actions to take when instrument air is degrading to 55 psig.

300000 A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIRSYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Air dryer and filter malfunctions (CFR 41.7, 41.10)

OC Learning Objective:

2621.828.0.0043, Objective G:

Describe the interlock signals and setpoi nts for the affect ed system components and expected system response includi ng power loss or failed components.

2621.828.0.0043, Objective H:

Given a set of system indications or data, evaluate and interpret them todetermine limits, trends and system status.

2621.828.0.0043, Objective I:

Identify and explain the syst em operating controls/indications under all plant operating conditions.

2621.828.0.0043, Objective K:

State the function and interpretation of system alarms, alone and in combination,as applicable in accordance with the system RAPS.

Cognitive Level: Comprehension or Analysis

Question Type: New NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 116 of 135

References:

RAP-H1a, ABN-35, BR 2013, sh. 1 8/15/06 NRC Comments They said that the correct answer did not match the ABN-35 steps, in that the intended correct response would only be taken when service air was low - not

when instrument air was low. The question has been re-written.

9/8/06 Rossi reviewed. Deleted: with all system s normally aligned in the stem. Placed second stem paragraph and placed into bullets, and added indicator label for air pressure. Placed answers in 3-column format.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 117 of 13565.Given the following: The reactor is operating at ra ted power with the 'A' CRD pump in service The Unit Supervisor directs swapping to the 'B' CRD pump somaintenance can be performed on the 'A' CRD pump Which of the following actions, if any , must be performed prior to starting CRD pump 'B'?a.Close 'B' CRD pump discharge valveb.Open 'B' CRD pump suction valve ONLY c.Open 'B' CRD pump suction and discharge valvesd.Take manual control of the in-service flow control valve Answer: d Handouts: None

Justification: A is incorrect - this is only required during initial pump startup when placing the CRD system in service.

B and C are incorrect - as stated in Procedure 302.1, NOTE 4.3.1, "during normal operation, both CRD pumps are us ually valved to the system so that pump changeover for purposes other t han maintenance may be made from the control room."

D is correct - for routine pump c hangeover, Procedure 302.1 requires taking manual control of the in-service flow control valve (NC03A or NC03B) prior to starting the standby (alter nate) CRD pump in service.

201001 A4.01 Ability to manually operate and/or monito r in the control room: CRD pumps (CFR 41.10)OC Learning Objective:

2621.828.0.0011, Objective 13:

Given normal operating procedures and documents for the system, describe or interpret the procedural steps.

Cognitive Level: Memory or Fundamental

Question Type: New

References:

302.1 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 118 of 13566.Given the following: The reactor is operating at rated power Annunciator CCW FLOW LO goes into alarm for the 'A' reactor recirc pump The crew has entered ABN-19, RBCCW Failure Response Which one of the following describes (1) when a reactor scram is required by ABN-19, and (2) the limiting r eactor recirculation system component that this action is based on?a.(1) when one CCW FLOW LO annunciator has been in alarm for >

1 minute (2) recirc pump sealsb.(1) when more than one CCW FLOW LO annunciator has been in alarm for > 1 minute (2) recirc pump sealsc.(1) when one CCW FLOW LO annunciator has been in alarm for >

1 minute (2) recirc pump motord.(1) when more than one CCW FLOW LO annunciator has been in alarm for > 1 minute (2) recirc pump motor Answer: b Handouts: None

Justification: A is incorrect - ABN-19 di rects a reactor scram when there are two or more CCW FLOW LO alarms for greater than one minute.

B is correct - when RPV temperature is > 212 F, the mode switch is in STARTUP or RUN, and there are two or more CCW FLOW LO alarms for greater than one minute, ABN-19 directs a reactor scram and trip of all operating recirc pumps. Recirc pumps seals are the limiting component since (1) the seal temperature limits specifi ed in ABN-19 are lower for the seals than for the pump motors and (2) high seal temperatures can cause seal failure, which is of higher

consequence than high motor beari ng and/or winding temperatures.

C is incorrect - ABN-19 directs a reac tor scram when there are two or more CCW FLOW LO alarms for greater than one minute. Recirc pump seals are the limiting component.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 119 of 135 D is incorrect - recirc pump seals are the limiting component.

202001 A2.17 Ability to (a) predict the impacts of the following on the RECIRCULATIONSYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of seal cooling water (CFR 41.3, 41.10)

OC Learning Objective:

2621.828.0.0035, Objective M:

Using the procedure, identify and expl ain normal and emergency operations of the RBCCW system.

Cognitive Level: Memory or Fundamental

Question Type: New

References:

RAP-E7d, RAP-E7b, ABN-19 8/15/06 NRC Comments They asked if the CCW Low Flow alarm was to the 'A' recirc. pump only. Yes.

Each pump has its own individual alarm. The question remains as-is.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 120 of 13567.The reactor was operating at ra ted power with the 'B' RWCU pump in service when a turbine trip occurred.

The following conditions currently exist: An ATWS is in progress Reactor pressure is 950 psig RPV water level was lowered to 110"TAF and is stable The Reactor Operator initiate s Standby Liquid Control System 1 The following indications are observed:

o System 1 PUMP ON light on Panel 4F is lit o System 1 SQUIBS light on Panel 4F is NOT lit o Pump discharge pressure on Panel 4F is 1080 psig o FLOW ON annunciator on Panel 3F is NOT in alarm o SQUIB VALVE OPEN annunciator on Panel 3F is NOT in alarm What is the status of the R eactor Water Cleanup (RWCU) System?

RWCU is ____(1)____ and the 'B' RWCU pump is ____(2)____.a.(1) isolated (2) trippedb.(1) isolated (2) NOT trippedc.(1) NOT isolated (2) trippedd.(1) NOT isolated (2) NOT tripped Answer: d Handouts: None

Justification: A is incorrect - RWCU is not isolated and the 'B' RWCU pump is not tripped.

B is incorrect - RWCU is not isolated.

C is incorrect - the 'B' RWCU pump is not tripped.

D is correct - the given conditions indi cate SLC System 1 pump started (PUMP ON light is lit; 1080 psig discharge pressure) but the squib valve did not fire (SQUIBS light not lit; SQUIB VALVE OPEN annunciator not in alarm) and thereis no liquid poison flow into the reac tor (FLOW ON annunciator not in alarm).

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 121 of 135 RWCU isolation on SLC initiation occurs based on system flow > 15 gpm as sensed by FS-IL06. This flow switch al so inputs to the FLOW ON annunciator.

Since the squib valve did not fire, system flow to the reactor did not reach the 15 gpm setpoint required to isolate RWCU.

Since RWCU did not isolate, the 'B'RWCU pump is still running.

204000 K6.07 Knowledge of the effect t hat a loss or malfunction of the following will have on theREACTOR WATER CLEANUP SYSTEM: SBLC logic (CFR 41.6, 41.7)

OC Learning Objective:

2621.828.0.0039, Objective D:

Describe the interlock signals and setpoi nts for the affect ed system components and expected system response includi ng power loss of failed components.

2621.828.0.0039, Objective F:

Explain or describe how this system is interrelated with other plant systems.Cognitive Level: Comprehensive or Analysis

Question Type: New

References:

304, RAP-G1b, RAP-G2b

9/22/06: Validator comments: Added a bullet in the stem saying that RPV water level that was above the level setpoi nt for cleanup system isolation (>86").

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 122 of 13568.Given the following: The reactor is operating at rated power LPRM detector calibrations are in progress One TIP detector is IN-CORE A leak occurs inside the drywell, causing drywell pressure to exceed 3.5 psig As the TIP detector retracts from the reactor, when will its ball valveautomatically close?

As soon as the detector ________

a.is moved outside of the indexerb.is outside the primary containment c.has moved past the ball valve d.has moved into the shield chamber Answer: d Handouts: None Justification: A, B and C are incorrect

- the ball valve will not close until the detector is in the shield chamber, as i ndicated by the "in-shield" limit switch.D is correct - when the TIP system receiv es a containment isolation signal due to high drywell pressure, all detectors not "i n-shield" will retract to the in-shieldposition, the associated ball valves will close, and the purge valve will close. The ball valve receives a close signal from the detector "in-shield" limit switch (see 405.2, P&L 4.17, among ot her places in 405.2).

215001 A3.03 Ability to monitor automatic operati ons of the TRAVERSING IN-CORE PROBEincluding: Valve operation (CFR 41.7)

OC Learning Objective:

2621.828.0.0029, Objective F:

Describe the interlock signals and setpoi nts for the affect ed system components and expected system response includi ng power loss or failed components.

Cognitive Level: Memory or Fundamental

Question Type: New

References:

405.2 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 123 of 135 8/15/06 NRC Comments They did not think the original answer a (is outside the reactor vessel) was

credible. It was changed to that shown.

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 124 of 13569.Which of the following is a Fuel Po ol Cooling System design feature that prevents inadvertent draining of the spent fuel storage pool?a.Spent fuel storage pool liner telltale drain systemb.No penetrations in the spent fuel storage pool wall c.Fuel pool cooling pumps trip on Lo-Lo skimmer surge tank leveld.Anti-siphon check valves in the spent fuel storage pool diffuser lines Answer: d Handouts: None

Justification: A is incorre ct - the telltale drain system provides indication of aspent fuel storage pool liner leak-it does not prevent inadvertent draining of the pool.B is incorrect - there are penetrations in the spent fuel storage pool wall, however none are below the level of 1 f oot above the top of the stored fuel.

C is incorrect - the FPC pumps do trip on Lo-Lo skimmer surge tank level, but the purpose of this trip is to prevent loss of NPSH to the FPC pumps. Based on the design of the system (weir overflow of SFSP to SST, among others), the FPC pumps cannot drain the spent fuel storage pool.

D is correct - the diffuser lines (SFSP retu rn) are the only lines that go below the top of the stored f uel. The anti-siphon check valves prevent inadvertent draining of the SFSP during reactor cavity letdow n (siphoning from the spent fuel storage pool to the reactor cavity).

233000 K4.06 Knowledge of FUEL POOL COOLING A ND CLEAN-UP design feature(s) and/orinterlocks which provide for the following: Maintenance of adequate pool level (CFR 41.7, 41.9)

OC Learning Objective:

2621.828.8.0020, Objective A:

Given plant operating conditions, descri be or explain the purpose(s)/function(s) of the system and its components.

2621.828.8.0020, Objective II:

Describe the operation of the anti-siphon check valves and analyze how the system may respond if they fail to perform their design function.

Cognitive Level: Memory or Fundamental

Question Type: New NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 125 of 135

References:

UFSAR 9.1.2.

2, 9.1.3, 205, 311, 237E756 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 126 of 13570.Given the following: The plant is operating at rated power on four (4) recirc loops Recirc pump 'B' is out of service due to an oil leak in the motor The 4160V BUS 1B UV annunciator goes into alarm Which of the following is the correct action to take for this event?a.Scram the reactor and enter ABN-1, Reactor Scramb.Confirm operating recirc pump speeds reduced to 20 to 30 Hz c.Perform a rapid power reduction as directed by the Unit Supervisord.Enter 301.2, Reactor Recirculation System, for a scoop tube lockup on recirc pump 'D' Answer: a Handouts: None

Justification: A is correct - an undervolt age condition on bus 1B causes a trip of recirc pump D, a trip of condensate pumps B & C, and a trip of feedwater pumpsB & C. For multiple condensate and/

or feedwater pump trips, ABN-17 (Feedwater System Abnormal Conditions) directs scramming the reactor and entering ABN-1. The correct action from ABN-2 (Recirculation System Failures) is to close the D pump discharge valve.

B is incorrect - this action is directed by ABN-2 for a single recirc pump trip with only 3 loops initially in service (2 pumps remaining in service). This action would not be taken when there are 3 or 4 pumps remaining in service. ABN-2 does require recirc pump speed to be less than 33 Hz when only 3 recirc pumps are in

service, but it does not direct action to lower speed to the 20 - 30 Hz range.

C is incorrect - as stated in ABN-17, this is the correct action to take for a single condensate pump or single feedwater pump trip.

D is incorrect - an undervoltage condition on bus 1B causes a scoop tube lockup on the D recirc pump and this action would be necessary to control the speed of

the D recirc pump if it were still running.

Since it tripped on the UV condition, this action would not be taken.

259001 K2.01 Knowledge of electrical power supplie s to the following: Reactor feedwater pump(s): Motor-Driven-Only (CFR 41.7)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 127 of 135 OC Learning Objective:

2621.828.0.0017, Objective C:

Given the system logic/electrical dra wings describe the system trip signals, setpoints and expected system respons e including power loss or failed components.Cognitive Level: Comprehensive or Analysis

Question Type: Modified Bank

References:

RAP-T4c, ABN-2, ABN-17 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 128 of 13571.The following communication takes place over the radio between the Reactor Operator and the Reactor Bu ilding Operator while adding makeup to the "A" Isolation Condenser fr om the Fire Protection System:RO:"Open V eleven forty nine, makeup va lve from Fire Protection to the Isolation Condensers"NLO:"I understand, open V eleven forty ni ne, Fire Protection to Isolation Condenser Makeup"RO:"That is correct."

This communication is __________.

a.INCORRECT because the NLO paraphrased the direction given by the ROb.INCORRECT because the RO and NLO did not use the phonetic alphabet for "V"c.INCORRECT because the RO and NLO did not address each other by name or titled.CORRECT because it meets t he requirements of HU-AA-101, Human Performance Tools and Verification Practices Answer: c Handouts: None Justification: A is incorrect - paraphrasing is allowed by HU-AA-101.

B is incorrect - HU-AA-101 only require s use of the phonetic alphabet "as required, to ensure proper component identification."

In this case, C is correct - HU-AA-101 requires the sender to address the

receiver by name or title. HU-AA-101 al so states: "For non-face-to-face verbal communication, the sender and receiver s hall IDENTIFY themselves by stating their name or title."

D is incorrect - this communication does not meet the requirements of HU-AA-101.G2.18 Conduct of Operations: Ability to coordinate personnel activities outside the control room. (CFR 41.10)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 129 of 135 OC Learning Objective:

2621-PBIG.0002, Objective 3:

Demonstrate proper communications, in accordance with Conduct of Operations Manual and Human Performance proc edural requirements, including:a. Clear and concise communications

b. Three-way communications

c. Use of phonetic alphabet

d. Specifying correct component/unit/train

e. Creating understanding

f. verifying understanding (clarification/confirmation)

Cognitive Level: Memory or Fundamental

Question Type: New

References:

HU-AA-101 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 130 of 13572.Given the following: The reactor is in COLD SHUTDOWN and pre-startup evolutions are in progress The 'B' recirc pump is being placed in service and is aligned as follows:

o The MG set drive motor breaker is shut o The scoop tube is positioned at 100%

o The WARM light has just illuminated Which one of the following describes what happens when the STRT/NORM pushbutton is depressed?a.The field breaker will close i mmediately and the scoop tube will remain at 100%.b.The field breaker will close i mmediately and the scoop tube will start running backc.The scoop tube will start runni ng back and the field breaker will close when the scoop tube reaches the low speed positiond.The scoop tube will start runni ng back and the field breaker will close when the scoop tube passes through the 40% to 30% range Answer: d Handouts: None Justification: A, B and C are incorrect

- the scoop tube will run back and the field breaker will not close until the scoop tube reaches the 40-30% range.

D is correct - as soon as the STRT/NORM pushbutton is depressed the scoop tube begins to run back. When it reaches the 40-30% position, the field breaker will close and the recirc pump will start.

The scoop tube will continue to run back to the low speed position.

G2.2.1 Equipment Control: Ability to perform pre-startup procedures for the facility / including operating those controls associated with plant equipment that could affect reactivity. (CFR 41.7)

OC Learning Objective:

2621.828.0.0040, Objective G:

Explain the starting logic for the reactor recirc MG sets Cognitive Level: Memory or Fundamental

Question Type: Bank

References:

301.2 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 131 of 13573.Per RP-AA-203, Exposure Control and Authorization, occupationalworkers at Oyster Creek have an Administrative Dose Control Level (ADCL) of ___(1)___ mrem TEDE per year.

This limit can be raised to ___(2)___

mrem TEDE with written approval by the Radiation Protection Manager and the work group supervisor.a.(1) 1000 (2) 2000b.(1) 2000 (2) 3000c.(1) 1000 (2) 4000d.(1) 2000 (2) 4000 Answer: b Handouts: None

Justification: A is incorrect - the initial ADCL limit is 2000 mrem TEDE.

B is correct - RP-AA-203 states: "Admin istrative dose control levels have been established for Total Effective Dose Equivalent Limits of 2000 mrem routine cumulative TEDE/yr."

And, "To raise the ADCL to 3000 mrem TEDE in a calendar year, written approval is r equired by the Radiation Protection Manager and the work group supervisor."

C and D are incorrect - the initial ADCL limit is 2000 mrem TEDE; this can be raised to 4000 mrem TEDE with "written appr oval from the Radiation Protection Manger, a work group supervisor, and t he Station/Plant Manager." The RPMand work group supervisor can only authorize an extension to 3000 mrem TEDE.

G2.3.4 Radiation Control: Knowled ge of radiation exposure limits and contamination control / including permissible levels in excess of those authorized.(CFR 41.12)

OC Learning Objective:

Exelon GET RWT Module (Rev. 31), Objective 17:

State the Exelon Administrative Dose Control Level for TEDE and the administrative does guidelines for LDE, SDE, and TODE.

Exelon GET RWT Module (Rev. 31), Objective 18:

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 132 of 135 State the actions to be taken if an Exelon Administrative Dose Control Level or administrative guidelines are being approached.

Cognitive Level: Memory or Fundamental

Question Type: Modified Bank

References:

RP-AA-203 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 133 of 13574.The primary containment is being purged with air in preparation for a plant shutdown.Why does procedure 312.9, Primary Cont ainment Control, prohibit thesimultaneous opening of Drywell and Torus vent and purge valves?

To prevent-a.over-pressurizing the Reactor Building ventilation ductsb.exceeding the Reactor Building v entilation exhaust fan capacityc.loss of the positive differential pressure between the Drywell and Torusd.creating a pathway for steam to bypass the suppression pool during a LOCA Answer: d Handouts: None

Justification: A is incorrect - there is no reason to believe this would occur and this is not the reason for the limitation given in 312.9.

B is incorrect - there is no reason to believe this would occur and this is not the reason for the limitation given in 312.9.

C is incorrect - this is not the reason fo r the limitation given in 312.9, but could be a misconception based on the requirem ent in 312.9 to monitor Drywell and Torus pressure during purging to maintain a positive d/p.

D is correct - as stated in P&L 7.2.6 of 312.9, "When primary containment is required, simultaneous opening of dryw ell and torus vent and purge valves is prohibited. Operating with both the drywell and torus valves open creates a pathway to bypass the torus-to-drywell vacuum breakers." Procedure 312.9 also references LER 97-014, which states:

"On October 21, 1997, while reviewing industry events, it was discovered t hat a suppression pool bypass leak path existed during purging and v enting of the primary c ontainment. Two operating procedures allowed venting and purging of the torus and drywell simultaneously.If a LOCA were to occur during this ti me, a pathway for steam to bypass the suppression pool was created. Calculati ons indicated that the allowable bypassarea was exceeded but, in the unlikely event of a LOCA, the peak torus and dry-well pressure would not increase."

G2.3.9 Radiation Control: Knowled ge of the process for performing a containment purge. (CFR 41.10)

NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 134 of 135 OC Learning Objective:

2621.828.0.0065, Objective F:

Given normal operating procedures and documents for the system, describe or interpret the procedural steps.

Cognitive Level: Memory or Fundamental

Question Type: Modified Bank

References:

312.9, LER 97-014 NRC Exam 2006-1 Reactor Operator Exam KeyNRC RO Exam 2006-1 Key Rev 2Page 135 of 13575.The reactor is operating at rated power.

Which of the following annunciators correspond to an EOP entry condition? (Assume the alarms are received individually and are valid.)a.RX PRESS HIb.RX LVL HI/LO c.DW PRESS HI/LO d.ISOL COND AREA TEMP HI Answer: d Handouts: None

Justification: A is incorrect - this annunciator alarms when reactor pressure reaches 1030 psig; the EOP entry condition for high reactor pressure is 1045

psig.B is incorrect - this annunciator alarms at 146" TAF and the EOP entry on RPV water level is below 138". answer b is incorrect.

C is incorrect - this annunciator alarms at a low DW pressure of 1.0 psig or a high DW pressure of 1.4 psig. The EO P entry condition for DW pressure is 3 psig.D is correct - this annunciator alarms at 160 F and requires entry into EMG-3200.11, Secondary C ontainment Control.

G2.4.2 Emergency Procedures/Plan: Knowl edge of system set points / interlocks and automatic actions associated with EOP entry conditions. (CFR 41.10)

OC Learning Objective:

2621.828.0.0023, Objective K:

State the function and interpretation of system alarms, alone and in combination,as applicable in accordance with system RAPS.

Cognitive Level: Memory or Fundamental

Question Type: New

References:

RAP-H3f, RAP-G7c, RAP-C5e, RAP-C8b

8/15/06 NRC Comments They thought that the original answer b (DW TEMP HI) could be an EOP entry condition. It has been changed with that shown.