ML14310A408

2014-10 - Draft Written Exam

ML14310A408
Person / Time
Site:	Grand Gulf
Issue date:	10/24/2014
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
laura hurley
References
Download: ML14310A408 (199)

Text

Examination Outline Cross

-Reference Level RO 295001 Partial or Complete Loss of Forced Core Flow Circulation Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Limiting cycle oscillation: Plant Specific Tier # 1 Group # 1 K/A # 295001 AK1.04 Rating 2.5 Question 1 The plant is operating on the 100 percent rod line at rated thermal power

. Operation is in accordance with Backup Stability Protections due to the inoperability of OPRMs. Then: Both reactor recirc pumps unexpectedly downshift to slow speed Reactor power stabilizes at 45 percent of rated thermal power With these conditions, in order to limit thermal hydraulic induced power oscillations ONEP 05-01-02-III-3, Reduction in Recirculation Flow Rate, directs the crew to immediately

A. begin inserting control rods.

B. place the reactor mode switch in SHUTDOWN.

C. trip the recirculation pump with the lowest flow.

D. maximize reactor recirc pump flow s . Answer: B Explanation:

Since the core is operating in the scram region of the power

-flow curve with OPRMS inoperable, step 4.1 requires an immediate scram.

A would be correct if in the controlled entry region of the power

-flow curve.

C would be correct for trip of a single Recirc pump.

D would be correct if the pumps are manually downshifted Technical

References:

05-01-02-III-3, , Reduction in Recirculation Flow Rate References to be provided to applicants during exam: Figure 1 from 05-01-02-III-3, Power-flow map.

Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295003 - Partial or complete loss of AC Knowledge of abnormal condition procedures.

Tier # 1 Group # 1 K/A # 295003 - 2.4.11 Rating 4.0 Question 2

According to 05-1-02-I-4, Loss of AC Power

, which of the following 2 buses would constitute a station blackout if power was lost?

A. 15AA and 16AB

B. 15AA and 17AC C. 16AB and 13AD D. 16AB and 14AE Answer: A Explanation:

A is correct because of the note 05 02-I-4, Loss of AC Power, Rev 45.

B is wrong because both buses are ESF buses but 17AC is not the Div 3 bus.

C is wrong because 13AD is not an ESF but does carry a recirc pump MG set which could be plausible if an applicant doesn't understand what safety related means.

D is wrong because 14AE is not an ESF but does carry a recirc pump MG set which could be plausible if an applicant doesn't understand what safety related means.

Technical

References:

05-1-02-I-4, Loss of AC Power, Revision 45, Section 5.0 Automatic Actions References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross

-Reference Level RO 295004 Partial or Total Loss of DC Power Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Tier # 1 Group # 1 K/A # 295004 2.4.4 Rating 4.5 Question 3 The plant is operating at rated thermal power when an unplanned automatic actuation of the following ESF systems occurs:

Control Room Standby Fresh Air 'A' initiation Combustible Gas Control System 'A' initiation Containment/Drywell Division I isolation Auxiliary Building Division I isolation Standby Service Water System 'A' initiation Standby Gas Treatment System 'A' initiation Auxiliary Building Division I Ventilation isolation At this time, the crew should:

A. restore AC power and enter 05 02-V-11 Loss of Plant Service Water.

B. restore DC power and enter 05 02-V-11 Loss of Plant Service Water.

C. restore AC power and enter 05 02-III-5, Automatic Isolations.

D. restore DC power and enter 05 02-III-5, Automatic Isolations.

Answer: D Explanation:

These actuations could not result from a loss of AC, only DC. The affected systems can only be restored by the actions of ONEP 05 02-III-5, which resets the affected signals. Plant service water is affected, but 05 02-V-11 contains no instructions that will restore service water in this condition.

This event happened at GGNS as documented in LER 95

-005-00 Technical

References:

Grand Gulf Nuclear Station LER 95

-005-00. 05-1-02-III-5, Automatic Isolations References to be provided to applicants during exam: None Learning Objective:

GLP-OPS-L1100 Discuss the interrelationships of the following systems with the Plant DC System. (10) ESF Power Distribution System

- R21 (10.2)

GLP-OPS-R2100 Describe the interrelationship between the following systems and the ESF Distribution System: (25) 125 Volt DC System

- L11 (25.1)

Question Source:

New Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295005 Main Turbine Generator Trip Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: RPS Tier # 1 Group # 1 K/A # 295005 - AK2.02 Rating 3.8 Question 04

Which of the following RPS signals also generates a turbine trip signal?

A. Reactor high water level B. Reactor low water level C. High reactor pressure

D. Main steam isolation valve closure Answer: A Explanation:

A is correct because this signal generates a turbine trip to protect the turbine from carryover.

B is wrong because this signal does not directly generate a turbine trip.

C is wrong because this signal does not directly generate a turbine trip.

D is wrong because this signal does not directly generate a turbine trip.

Technical

References:

05-1-02-I-2, Turbine and Generator Trips, Rev 34 References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 295016 Control Room Abandonment Knowledge of surveillance procedures Tier # 1 Group # 1 K/A # 295016 2.2.12 Rating 3.7 Question 5 All automatic starts of 1P41

-C001A, SSW A Pump will at least be momentarily disabled by performance of surveillance procedure:

A. 06-OP-1P41-Q-004, SSW Loop A Valve and Pump Operability Test.

B. 06-OP-1P41-M-004, SSW Loop A Operability Check.

C. 06-OP-1C61-M-001, Remote Shutdown Panel and Accident Monitoring Instrument Channel Check.

D. 06-OP-1C61-R-0002, Remote Shutdown Panel Control Check.

Answer: D - We will have to check at GG but I am thinking there may be parts of "A" that will render the SSW pump inoperable. Leave it as is and we will check on it later.

Explanation:

Of these, only the Remote Shutdown Panel Control Check disables auto starts.

A, B, and C are all surveillance procedures which operate or monitor SSW components, but they do not disable any pump features.

Technical

References:

Surveillance Procedure 06

-OP-1C61-R-002, Remote Shutdown Panel Control Check, Table 1, p. 11.

References to be provided to applicants during exam: None GLP-OPS-C6100 Learning Objective: GLP-OPS-C6100 Describe the reason for disabling the Control Room controls by activating the Lockout Transfer Relays on H22

-P152 and LOCAL/REMOTE switches on Alternate Shutdown Panels during a fire emergency in the Main Control Room. (11) Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295018 Partial or Total Loss of CCW Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Securing individual components (prevent equipment damage)

Tier # 1 Group # 1 K/A # 295018 - AK3.03 Rating 3.1 Question 06 The plant has been operating full power for 5 months following a refueling outage. One of the 2 running CCW pumps has tripped and the standby pump will not start. What does ONEP

05-1-02-V-1, "Loss of Component Cooling Water", direct you to do? What is the reason this action is taken?

A. Scram the reactor; one CCW pump cannot maintain RRC motor temperatures within specification so manual action is required B. Isolate CCW to FPCCU heat exchangers; maximize cooling to RRC pumps C. Isolate CCW to RWCU heat exchangers; maximize cooling to RRC pumps D. Reduce core flow to 70 mlbm/hr; reduce the heat load on RRC pumps Answer: B Explanation:

A is wrong because this would be the appropriate response if this was a complete loss of CCW. Plausible because if the applicant believes this constitutes a complete instead of a partial loss of CCW this is the correct answer.

B is correct because this is specified in step 3.2.2 of the procedure and the note before step 3.2.2 gives the reason for the step. Also, step 3.2.1 was already accomplished because an attempt to start the standby pump was already attempted.

C is wrong because this is required by step 3.2.3. This is plausible if the applicant believes there has not been enough time since the refueling outage and they are not allowed to isolate CCW to FPCCU so the next step is to isolate CCW to RWCU D is wrong because this is specified in step 3.2.4 as a conditional step if RRC temperatures cannot be maintained within specification Technical

References:

ONEP 05-1-02-V-1, Loss of Component Cooling Water, Rev. 24, p. 3

-4 References to be provided to applicants during exam:

None. Learning Objective: N/A Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross

-Reference Level RO 295019 Partial or Total Loss of Inst. Air Ability to obtain and interpret station electrical and mechanical drawings.

Tier # 1 Group # 1 K/A # 295019 2.2.41 Rating 3.5 Question 7 When TBCW is not available to plant air compressors, aligning SSW to supply cooling water to plant air compressors requires opening valve:

A. P41-F155A B. P41-F159A C. P41-F160B D. P41-F042 Answer: A Explanation:

B is wrong because this valve supplies the A drywell purge compressor C is wrong because this is the return from the B drywell purge compressor D is wrong because this is the supply to DRWL CLRS/CCW HXS The valve descriptions are omitted because the purpose of this question is to require use of the P & ID. Giving the valve descriptions makes use of the reference unnecessary.

Technical

References:

Step 3.3 of ONEP 05 02-V-9, Loss of Instrument Air Standby Service Water System P & I Diagrams M

-1061D (G-2) References to be provided to applicants during exam:

Standby Service Water System P & I Diagrams M

-1061A, B, C, and D Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295021 Loss of Shutdown Cooling Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor water heatup/cooldown rate Tier # 1 Group # 1 K/A # 295021 - AA2.01 Rating 3.5 Question 08 The plant is shut down for 5 days for RF19 with RHR B in service and DIV 1 ESF bus out o f

service for maintenance. Initial conditions are:

RPV temperature is 110

°F Fuel shuffling has not commenced Reactor vessel is flooded up to below the flange

What is the current maximum heatup rate?

A. 50° B. 80° C. 100° D. 120° Answer: B Explanation:

Since the temperature is 110

°F the applicant should use the 120

°F curve of Attachment 1, Figure 3, of 05 02-III-1, Inadequate Decay Heat Removal, Rev. 39. Using this curve the difference between 200

°F and 120°F is 80°F and the time 5 days after shutdown is 1 hr. All other rates are plausible if the applicant chooses the wrong graph.

Technical

References:

05-1-02-III-1, Inadequate Decay Heat Removal, Rev. 39 References to be provided to applicants during exam:

All attachments from 05 02-III-1, Inadequate Decay Heat Removal, Rev. 39 Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(5)

Examination Outline Cross

-Reference Level RO 295023 Refueling Accident Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS : Area radiation levels.

Tier # 1 Group # 1 K/A # 295023 AA2.01 Rating 3.6 Question 9 Refueling activities are in progress. A CONTAINMENT FUEL HDLG AREA RADIATION HIGH alarm is received. No other alarms are received at this time. This condition could be caused by:

A. a fuel assembly being inadvertently raised too high at the refueling platform.

B. drop of a spent fuel assembly from the refueling platform.

C. a fuel assembly being inadvertently raised too high at the fuel handling platform.

D. drop of a spent fuel assembly from the fuel handling platform.

Answer: A Explanation:

Raising an irradiated fuel assembly too high could result in reduced shielding with a subsequent rise in area radiation levels.

B and D are wrong because if release of gaseous activity from a damaged fuel assembly is the cause of the area radiation alarm, this would be accompanied by gaseous process radiation alarms. C and D are wrong because the fuel handling platform is not in the containment building.

Technical

References:

ARI 04-1-02-1H13-P844-1A-A3, CONTAINMENT FUEL HDLG AREA RADIATION HIGH.

Lesson plan GLP

-RF-F1101, Fuel handling/Refueling Platfor m References to be provided to applicants during exam:

None Learning Objective: GLP-OPS-D1721 Given applicable Alarm Response Instructions (ARIs) and plant conditions, state/identify the probable causes for any alarm listed in the System Alarm Index of SOI 04-1-01-D17-1 and SOI 04 01-D21-1 (NLOs responsible for LOCAL Panels only). (19)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295024 High Drywell Pressure Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE: RHR/LPCI Tier # 1 Group # 1 K/A # 295024 - EA1.04 Rating 4.1 Question 10 The plant has scrammed after receiving a high drywell pressure signal. The amber light above the RHR A pump is lit indicating the A RHR pump breaker has-A. tripped on a load shed.

B. tripped on no minimum flow path.

C. received a close signal and is tripped.

D. received a close signal and is running.

Answer: C Explanation:

A is wrong because there would not be a reason for the pump to trip on a load shed. The applicant should assume all equipment was in a normal position, so the pump wouldn't trip.

The pump should load after LSS is complete.

B is wrong because there is no trip on no minimum flow path, but plausible since this should be an immediate action for operators. Applicants may confuse the immediate action with preventing an automatic action from occurring.

C is correct because this is the meaning of the light per Table 1 of GFIG

-OPS-E1200. D is wrong because of C.

Technical

References:

GFIG-OPS-E1200, Residual Heat Removal (RHR) System

- Figures, Rev 1 References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent) Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 295025 High Reactor Pressure Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor pressure.

Tier # 1 Group # 1 K/A # 295025 EA2.01 Rating 4.3 Question 11 A reactor vessel steam dome high pressure scram signal should be the signal that first generates a reactor scram on a:

A. single MSIV closure with reactor power at 75%.

B. closure of all MSIVs at full power.

C. turbine trip at full power.

D. pressure controller failing closed with reactor power at 75%.

Answer: D Explanation:

Interpreting reactor pressure involves determining if it is responding as expected.

A is wrong because this should not result in a scram.

B is wrong because the reactor will scram on the main steam isolation valve closure.

C is wrong because the reactor will scram on Turbine stop valve closure or low trip oil pressure.

At less than RTP, the high pressure signal should precede the high power signal.

Technical

References:

Mitigating of Core Damage (EPTS

-2) Pressure Increase Events, GLP

-OPS-MCD12 References to be provided to applicants during exam:

None. Learning Objective: GLP-OPS-MCD12 Given a pressure increase event and associated plant parameter curves, explain the behavior of the plant parameters.(2

) Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.4 3 Examination Outline Cross

-Reference Level RO 295026 Suppression Pool High Water Temp Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor SCRAM Tier # 1 Grou p # 1 K/A # 295026 - EK3.05 Rating 3.9 Question 12 The reason for transitioning out of EP

-3, Containment Control, before suppression pool temperature reaches 110 oF is to ensure:

A. There is an adequate NPSH for all pumps taking suction off of the suppression pool.

B. All pumps taking suction from the suppression pool will not become gas bound.

C. Containment pressure remains below 10 psig during a design basis accident.

D. EP-2, RPV Control, is initiated before reaching maximum boron injection initiatio n temperature.

Answer: D Explanation:

A is wrong because NPSH is only referenced for level and not temperature. This is plausible since NPSH is a combination of both.

B is wrong because gas binding of pumps is not mentioned for high suppression pool temperatures.

C is wrong because the design containment pressure is 15 psig.

D is correct because in the PSTG this is referenced as the reason for transitioning.

Technical

References:

05-S-01-PSTG, Plant Specific Technical Guidelines, Rev 05 References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(7);(8)

Examination Outline Cross

-Reference Level RO 295027 High Containment Temperature Knowledge of the reasons for the following responses as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY) : Containment spray: Plant specific

. Tier # 1 Group # 1 K/A # 295027 EK3.02 Rating 3.2 Question 13 The reason containment sprays are used to mitigate high containment temperature ONLY when containment pressure is relatively high is this:

A. ensures the steam formation will not over

-pressurize containment.

B. prevents an excessive rate of depressurization.

C. ensures RHR pump discharge pressure is adequate to prevent runout.

D. prevents excessive flow through the RHR heat exchanger.

Answer: B Explanation:

The CSIPL does not allow spray to be initiated when temperature is high and pressure is low.

Spray with high temperature and low pressure results in a rapid cooldown/depressurization which may lead to forming a negative pressure in containment with subsequent loss of containment integrity.

A is credible because steam may be produced by spray flow, but the concern is overcooling, not overpressurization.

C is credible because high pressure would limit centrifugal pump flow but this is not the basis for the CSIPL.

D is credible because low pressure would result in more HX flow but this is not the basis for the CSIPL.

Technical

References:

BWROG EP Guidelines section 17.9.

References to be provided to applicants during exam:

Learning Objective: GLP-OPS-EP3 7. State/identify the basis associated with each individual step in EP

-3. Question Source:

Bank # (note changes; attach parent) Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295028 High Drywell Temperature Knowledge of EOP entry conditions and immediate action steps.

Tier # 1 Group # 1 K/A # 295028 - 2.4.1 Rating 4.3 Question 14 Which of the following would require an entry into 05

-S-01-EP-3, Containment Control?

If you were required to enter 05

-S-01-EP-2, RPV Control, based on not being able to control a parameter in EP

-3, what would your immediate action be upon entry into EP

-2? A. Suppression Pool temperature at 85 o F Maximize Suppression Pool cooling B. Drywell pressure at 1.15 psig Prevent ECCS injection C. Drywell temperature at 145 o F Verify the Reactor Mode Switch in Shutdown D. Suppression Pool level at 18.45 ft Inhibit ADS Answer: C Explanation:

A wrong because this temperature would not require entry into the EOP. The immediate action is not an action required when entering EP

-2, but is reasonable for a high temperature.

B is wrong because the pressure would not require entry into the EOP. The immediate action is a step in EP

-2 for high drywell pressure but pressure has to be above 1.39 psig.

C is correct because this temperature is above the entry criteria of 135 oF. The immediate action is correct upon entry into EP

-2. D is wrong because the level would not require entry into an EOP. The immediate action is a step in the EP

-2 procedure that should be second nature to an applicant so they could believe it is an immediate action.

Technical

References:

05-S-01-EP-3, Containment Control 05-S-01-EP-2, RPV Control References to be provided to applicants during exam:

None. Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross

-Reference Level RO 295030 Low Suppression Pool Wtr Lvl Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: Reactor SCRAM Tier # 1 Group # 1 K/A # 295030 EK3.06 Rating 3.6 Question 15 The reason a reactor scram is required on a low suppression pool level is:

A. the drywell to containment suppression pool vents will become uncovered. B. the SRV sparger will become uncovered C. the ECCS pumps may fail due to vortexing.

D. a LOCA may result in a loss of containment..

Answer: D Explanation:

The EP technical basis states that the reason is to preserve containment integrity becaus e the SP may not adequately condense steam on a primary system break..

Technical

References:

EP-2, RPV Control EP-3, Containment control EP Technical Basis, Step References to be provided to applicants during exam: None Learning Objective:

GL P-OPS-EP3 7. State/identify the basis associated with each individual step in EP

-3. Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295031 Reactor Low Water Level Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL: Automatic depressurization system actuation Tier # 1 Group # 1 K/A # 295031 - EK3.01 Rating 3.9 Question 16 The reason for the 105 second time delay before ADS initiation is to allow enough time _______: A. To inhibit ADS if not desired.

B. To ensure systems are aligned for an emergency depressurization.

C. For RCIC to recover RPV level.

D. For HPCS to recover RPV level.

Answer: D Explanation:

A is wrong because it is not the reason for the timer exists but plausible if an applicant believes the system initiation is based on the emergency procedure.

B is wrong because emergency depressurization does use the ADS valves and proper alignment is needed but an emergency depressurization occurs much later in the emergency

procedure and it is not the basis for the timer.

C is wrong because it is the incorrect high pressure system but plausible if an applicant confuses RCIC for an ECCS system.

D is correct because the reason is specified in the training manaual.

Technical

References:

GLP-OPS-E2202, Automatic Depressurization System (ADS)

References to be provided to applicants during exam:

None. Learning Objective: Describe the signals and setpoints for the initiation of the ADS System for both automatic and manual initiation.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(8),(10)

Examination Outline Cross

-Reference Level RO 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Knowledge of EOP mitigation strategies Tier # 1 Group # 1 K/A # 295037 2.4.6 Rating 3.7 Question 17 The plant was at rated thermal power when a turbine trip occurred. Multiple control rods remained at the fully withdrawn position. Currently; Reactor power is 10% RPV level is 18" Reactor pressure is 935 psig being controlled by the turbine bypass valves

. According to EP

-2A, ATWS RPV Control, the crew should lower RPV level:

A. to -70" and maintain it as close to

-70" as possible.

B. until power is below 5% or level drops to

-167". C. to between

-191" and -167" D. to between

-191" and -70". Answer: D Explanation:

A is wrong because, while level is initially lowered to

-70" in step 7.1, step 9 subsequently directs a level band of

-191" and -70" B and C are correct only if SRVs are open.

Technical

References:

EP-2A, Revision Date 5/15/12 References to be provided to applicants during exam:

None Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(10) 55.43 Examination Outline Cross

-Reference Level RO 295038 High Off

-site Release Rate Knowledge of the interrelations between HIGH OF F-SITE RELEASE RATE and the following: Offgas System Tier # 1 Group # 1 K/A # 295038 - EK2.02 Rating 3.6 Question 18 To prevent an off

-site release greater than allowable limits, the Offgas system isolates when (a) __________ Upscale HI

-HI-HI trip channel is(are) received. The Upscale HI

-HI-HI alarm setpoint is ____________ throughout an operating cycle by Chemistry.

A. single; fixed B. single; varied C. both; fixed D. both; varied Answer: D Explanation:

A is wrong because it takes both Upscal e HI-HI-HI alarms to isolate Offgas but it takes a single alarm to take the system out of bypass which makes this answer plausible. This value can be varied throughout an operating cycle at management discretion according to procedure

08-S-03-22, INSTALLED RADIATION MONITORING SYSTEM ALARM SETPOINT DETERMINATION AND CONTROL, Rev 114.

B is wrong because the first answer is false but the second is true C is wrong because it takes both channels to generate a trip but the setpoint can be varied at any time throughout the cycle.

D is correct because it takes both channels to isolate and the setpoint can be varied.

Technical

References:

Procedure 08-S-03-22, INSTALLED RADIATION MONITORING SYSTEM ALARM SETPOINT DETERMINATION AND CONTROL, Revision 114 References to be provided to applicants during exam:

None. Learning Objective: Discuss the types of alarms received on the Radiation Monitoring Systems. (8)

Describe the automatic actions initiated by the Process Radiation Monitoring. (15)

Question Source: Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(11)

Examination Outline Cross-Reference Level RO 600000 Plant Fire On Site Ability to operate and / or monitor the following as they apply to PLANT FIRE ON SITE: Fire Alarm Tier # 1 Group # 1 K/A # 600000 AA1.06 Rating 3.0 Question 19 A rate-of-rise heat detector in the "B" RPS MG room triggers a local fire alarm. The expected fire system response is a:

A. CO 2 discharge occurs immediately.

B. CO 2 discharge occurs after a 30 second delay.

C. halon discharge occurs immediately.

D. halon discharge occurs after a 30 second delay. Answer: B Explanation:

C and D are wrong because this room is not protected by Halon.

A is wrong because the discharge is delayed to allow for evacuation.

Technical

References:

GFIG-OPS-P6400 Figure 11, Typical CO 2 Control Station.

References to be provided to applicants during exam:

Learning Objective: GLP-OPS-P6400 Describe the automatic actions/features associated with each of the following Fire Protection System components, including setpoints and bypasses (if applicable):

(6.0) Fire Protection C02 System, including automatic and manual start (6.6)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 700000 Generator Voltage and Electric Grid Disturbances Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Voltage regulator controls Tier # 1 Group # 1 K/A # 700000 - AA1.03 Rating 3.8 Question 20 What is the maximum allowable reactive load allowed during normal and emergency conditions? Which pushbuttons are used to change reactive load?

A. +253 MVARS; TVR raise/lower B. +253 MVARS; Load Demand raise/lower C. +297.5 MVARS; TVR raise/lower D. +297.5 MVARS; Load Demand raise/lower Answer: C Explanation:

A is wrong because the maximum reactive load is listed in 04 01-N40-1 as +297.5 in P&L 3.8. A -253 corresponds to the minimum reactive load allowed by procedure. This is plausible if the applicant reverses the signs on maximum load.

B is wrong because this does not meet the maximum specified in procedure. Also, changing the load with the load demand pushbuttons changes actual load and not reactive load. This is plausible if an applicant doesn't understand the difference in grid/turbine load.

C is correct because this is the correct value and the correct button D is wrong because the incorrect pushbutton is listed.

Technical

References:

04-1-01-N40-1, Main Generator and Auxiliaries, Rev 37, p.4, P&L 3.8 Describe how the output voltage of the Main Generator is varied, including any limitations. (8) References to be provided to applicants during exam:

None. Learning Objective: Identify precautions, limitations, cautions, warnings and notes which apply to a given situation and are related to the Main Generator. (14.1). Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross

-Reference Level RO 295002 Loss of Main Condenser Vac Knowledge of the interrelations between LOSS OF MAIN CONDENSER VACUUM and the following: Offgas system Tier # 1 Group # 2 K/A # 295002 AK2.07 Rating 3.1 Question 21 With the plant at rated power, for which of the following alarms would a slow, gradual loss of condenser vacuum be expected?

A. OG POST-TREAT RAD HI

-HI-HI/INOP.

B. OG RADWST VENT RAD HI.

C. OG CHAR VAULT RAD HI.

D. MSL B/MSL C RAD HI

-HI/INOP. Answer: A Explanation:

OG POST-TREAT RAD HI

-HI-HI/INOP isolates condenser off

-gas, which, according to 05 02-V-8 "Loss of Condenser Vacuum" should lead to a slow, gradual reduction in condenser vacuum. OG RADWST VENT RAD HI has no automatic actions. OG CHAR VAULT RAD HI has no automatic actions.

MSL B/MSL C RAD HI

-HI/INOP only results in a half

-trip. Technical

References:

05-1-02-V-8, Loss of Condenser Vacuum 05-1-02-II-2, Off-Gas Activity High References to be provided to applicants during exam:

Learning Objective: GLP-OPS-N6200 13. Discuss the interrelationships of the following systems with the Condenser Air Removal System. h. Process Radiation Monitoring System, D17 i. Offgas System, N64 Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO Low Reactor Water Level Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following: Reactor water cleanup Tier # 1 Group # 2 K/A # 295009 - AK2.04 Rating 2.6 Question 22 Which of the following would result in the isolation of Reactor Water Cleanup?

A. Reactor vessel water level at

-50". B. Main steam line tunnel temperature at 165 o F. C. RWCU heat exchanger room at 110 o F. D. RWCU differential flow at 70 gpm for 30 seconds.

Answer: A Explanation: A is correct because it is less than the setpoint of

-41.6" B is wrong because the setpoint is at 185 o F C is wrong because the setpoint is 120 o F D is wrong because the setpoint is 79 gpm for 45 seconds Technical

References:

GLP-OPS-G3336, Rev 16, Reactor Water Cleanup System References to be provided to applicants during exam:

None. Learning Objective: 8.6 Describe automatic actions associated with system isolation valves.

Question Source:

Bank # (note changes; attach parent)

Modified Bank # New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)- 55.43

Examination Outline Cross

-Reference Level RO 295010 High Drywell Pressure Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE: Drywell ventilation/cooling Tier # 1 Group # 2 K/A # 295010 AA1.01 Rating 3.4 Question 23 An energized white light above the associated drywell chiller hand switch in the control room indicates the chiller:

A. is in standby.

B. is in local control.

C. has tripped.

D. has auto started.

Answer: A Explanation:

B , C, and D are wrong because A is correct.

Technical

References:

04-1-02-1H13-P601-16 A-B4, ARI for DRWL PRESS HI GLP-OPS-M5100, Drywell Cooling and Drywell Chilled Water Systems References to be provided to applicants during exam: None Learning Objective: GLP-OPS-M5100. 04 State/identify the locations from which the following components of the Drywell Cooling and Drywell Chilled Water Systems can be found. Drywell Coolers Drywell Chillers Obj 6, Describe the automatic actions . . . Drywell Chillers Question Source:

Bank # (note changes; attach parent)

Modified Bank #

N ew X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295011 High Containment Temp Knowledge of the interrelations between HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY) and the following: Containment ventilation/cooling: Mark

-III. Tier # 1 Group # 2 K/A # 295011 - AK2.01 Rating 3.7 Question 24 What is/are the MINIMUM number of fan coolers required to be in service per SOI 04 01-M41, Containment Cooling System? Will containment fan coolers receive an auto start signal if a high containment temperature alarm is received?

A. 1/Yes B. 2/No C. 1/No D. 2/Yes Answer: B Explanation:

A is wrong because the minimum number required is 2 and there is no auto start on high containment temperature.

B is correct because the minimum required is 2 and the only auto start for the fan coolers is on low flow across a fan.

C is wrong because see A D is wrong because the minimum number is correct there is no auto start on high containment temperature.

Technical

References:

GLP-OPS-M4100, Containment Cooling System M41, Rev 5 Note from 04 01-M41, Containment Cooling System:

"The following operations are performed at 1H 13-P842, Unless otherwise noted. Two OR three Containment coolers Should be in service during normal operations.

" References to be provided to applicants during exam:

None. Learning Objective: 10.1 State/identify the interlocks associated with starting and stopping of fans.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross

-Reference Level RO 295012 High Drywell Temperature Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE : Pressure/temperature relationship Tier # 1 Group # 2 K/A # 295010 AK1.01 Rating 3.3 Question 25 The purpose of the RPV Saturation Temperature (RPVST) curve is to evaluate the effect of drywell temperature on:

A. reactor vessel integrity.

B. containment integrity.

C. RPV level instrumentation accuracy.

D. RPV pressure instrumentation accuracy.

Answer: C Explanation:

The RPV Saturation Temperature is a plot of the saturation temperature of water as a function of pressure. If the temperature of the water in an RPV water level instrument run exceeds this temperature, the water may start to boil, resulting in unreliable level indication.

Technical

References:

02-S-01-40, p. 42 of 45.

References to be provided to applicants during exam:

none Learning Objective: GLP-OPS- Describe the system interrelationship between the Drywell Cooling and Drywell Chilled Water Systems and the following: (10) Containment and Drywell Instrumentation and Control System (10.6)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 295013 High Suppression Pool Temperature Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Suppression pool cooling Tier # 1 Group # 2 K/A # 295013 - AA1.01 Rating 3.9 Question 26 RHR B is in service in suppression pool cooling mode due to high temperature in the suppression pool from RCIC operability testing. According to procedure 04 01-E12-1, Residual Heat Removal System, the RHR system flow must be maintained greater than _________. At what minimum temperature should suppression pool cooling be secured?

A. 7300 gpm; 65 o F B. 7300 gpm; 70 o F C. 7450 gpm; 65 o F D. 7450 gpm; 70 o F Answer: A Explanation:

A is correct because as referenced in the procedure this is the minimum flow and a caution in the procedure has you secure suppression pool cooling before you go below 65 o F. B is wrong because it is the correct flow but temperature is incorrect. This is plausible because the note references 70 oF but it doesn't explicitly have you shutdown suppression pool cooling until below 65 o F. C is wrong because the flow requirements are incorrect but plausible since this is the tech spec value for flow. The temperature is correct D is wrong because both flow and temperature are incorrect but plausible.

Technical

References:

04-1-01-E12-1, Residual Heat Removal System, Rev. 143, p.26 References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(8)

Examination Outline Cross

-Reference Level RO 295015 Incomplete SCRAM Knowledge of the operational implications of the following concepts as they apply to INCOMPLETE SCRAM : Shutdown margin Tier # 1 Group # 2 K/A # 295015 AK1.01 Rating 3.6 Question 27 Select the statement(s) that correctly complete(s) this sentence.

Following a reactor scram with two control rods at the fully withdrawn position, the crew may make a transition from EP

-2A, ATWS, to EP

-2, RPV control relying on the negative reactivity from: 1. inserted control rods.

2. injected boron.

3. current xenon concentration.

A. 1, 2 and 3 B. 1 and 2 only C. 1 and 3 only D. 1 only Answer: D Explanation:

BWROG EPGs/SAGs, Appendix B Revision 2 says "Note that the instruction requires a positive determination, not only that the reactor is shutdown, but that it will remain shutdown, without reliance upon boron, under worst

-case cold shutdown conditions. The phrase "without boron" does not imply that the condition cannot be met if boron has been injected, but that credit cannot be taken for the negative reactivity contributed by the boron. Control rod insertion alone must provide the necessary shutdown margin.

" Technical

References:

BWROG EPGs/SAGs, Appendix B Revision 2 References to be provided to applicants during exam:

Learning Objective:

GLP-OPS-EP02A 7. State/identify the basis associated with each individual step in EP

-2A. Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Leve l RO 203000 RHR/LPCI: Injection Mode Knowledge of RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following: Operation from remote shutdown panel Tier # 2 Group # 1 K/A # 203000 - K4.14 Rating 3.6 Question 28

If the RHR A INJ DISABLE handswitch is placed in the ENABLE position at the Remote Shutdown Panel, F042A, RHR A Injection Shutoff Valve, will not automatically ______:

A. Open on a LOCA signal concurrent with low reactor pressure AND no containment spray initiation signal.

B. Close on containment spray initiation signal.

C. Open on LOCA signal concurrent with low reactor pressure OR no containment spray initiation signal.

D. Open on LOCA signal.

Answer: A Explanation:

A is correct becau se it is listed as a caution in 05 02-II-1 B is wrong because it is not defeated. The valve can still be opened manually with the signal present but it will auto close C is wrong because this does not meet the caution in 05 02-II-1 D is wrong because the valve does not get an open signal solely on initiation. Plausible if an applicant believes it does Technical

References:

05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Attachment XIV References to be provided to applicants during exam:

None. Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(8)

Examination Outline Cross

-Reference Level RO 205000 Shutdown Cooling Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) :Valve operation Tier # 2 Group # 1 K/A # 295000 K5.02 Rating 2.8 Question 29 Shutdown cooling is being established using RHR loop A per SOI 1 E12-2, Shutdown Cooling and Alternate Decay heat Removal Operations.

After starting RHR pump A, the operator is procedurally directed to open a valve and immediately establish a minimum flow in order to prevent:

A. cavitation in RHR pump A.

B. discharging reactor vessel coolant to the suppression pool.

C. reactor pressure vessel thermal stratification. D. lifting the RHR pump A discharge relief valve.

Answer: B Explanation:

According to the note prior to the pump start, "Failure to establish greater than 1154 gpm within 8 seconds of pump start automatically opens F064A, establishing flow from Reactor to Suppression Pool."

A is wrong because, while inadequate pump flow could eventually lead to pump cavitation, F064A operation will prevent this.

C is wrong because, while stratification is a concern when flow is inadequate, F064A operation will prevent this.

D is wrong, because the RHR pump is centrifugal, and its discharge pressure at shutoff head is well below the lift setpoint of system safety valves.

Technical

References:

Shutdown Cooling and Alternate Decay Heat Removal Operation, SOI 04 01-E12-2, Revision 116. Caution prior to step 4.1.2c(17)

References to be provided to applicants during exam:

None Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 205000 Shutdown Cooling Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following: High pressure isolation Tier # 2 Group # 1 K/A # 205000 - K4.0 2 Rating 2.6 Question 30 Which NSSSS group isolation provides protection against high reactor pressure for the RHR system in the Shutdown Cooling Mode of Operation?

A. 2 B. 3 C. 4 D. 5 Answer: B Explanation:

A is wrong because Group 2 isolations protect for RHR to Radwaste B is correct because Group 3 isolations protect the RHR in shutdown cooling C is wrong because Group 4 isolations were part of steam condensing mode of RHR D is wrong because Group 5 isolations protect RHR test lines Technical

References:

Table 2 of GFIG

-OPS-E1200; GLP

-OPS-M701 p.36 References to be provided to applicants during exam:

None. Learning Objective: Describe the signals, setpoints and valves actuated by the RHR Shutdown Cooling Isolation Logic. (7.3)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 209001 LPCS Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM : A.C. power Tier # 2 Group # 1 K/A # 209001 K6.01 Rating 3.4 Question 31 Following a loss of MCC 15B11, an ECCS initiation signal would result in the LPCS pump:

A. failing to start.

B. starting and running at shutoff head (zero flow) regardless of RCS pressure.

C. starting, and running on mini

-flow only regardless of RCS pressure.

D. starting and running without mini

-flow protection, but will inject into the core when RCS pressure drops low enough.

Answer: C Explanation:

MCC 15B11 powers all system MOVs but not the pump. The pump will start, but no MOV can be repositioned.

A is wrong because the pump has not lost power.

B is wrong because the mini

-flow valve has no power but it is normally open.

D is wrong because the injection flowpath is normally isolated with an MOV that has no power. Technical

References:

GLP-OPS-E2100 References to be provided to applicants during exam: None Learning Objective:

GLP-OPS-E2100 Rev. 10, Obj. 7.1 State/identify the power supplies for the following LPCS system components: Motor Operated Valves Obj 16 Upon recognition of the following abnormal plant indications, predict system response and potential consequences

- AC power disturbances Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 209001 LPCS Ability to determine operability and/or availability of safety related equipment Tier # 2 Group # 1 K/A # 20900 1 - 2.2.37 Rating 3.6 Question 32 The plant is in Mode 4 with Electrical Bus 16AB OOS for planned maintenance. You've been asked to perform 06

-OP-1E21-M-01, LPCS Monthly Functional Test, and the LPCS valves are in the following position:

LPCS Injection valve

- Closed LPCS Suction Valve

- Open LPCS Minimum flow to suppression Pool

- Open LPCS Testable Check Valve

- Closed Is the LPCS system operable? Are the minimum required systems for TS 3.5.2 met?

A. Yes/Yes B. No/Yes C. Yes/No D. No/No Answer: A Explanation:

A is correct because this lineup meets the requirements of LPCS Monthly Functional Test, and since LPCS, RHR A and HPCS are all operable and only 2 are required this meets the LCO for 3.5.2.

B is wrong because the system is not operable but plausible since it is counterintuitive to have the min flow valve should be in the open position for the system to be operable.

C is wrong because the system is operable and the minimum systems are met.

D is wrong because the system is operable and the minimum systems are met.

Technical

References:

06-OP-1E21-M-01, LPCS Monthly Functional Test, Revision 106 TS SR 3.5.2 References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7);(8)

Examination Outline Cross

-Reference Level RO 209002 HPCS Ability to predict and/or monitor changes in parameters associated with operating the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) controls including:

Condensate storage tank level: BWR-5,6 Tier # 2 Group # 1 K/A # 209002 A1.09 Rating 2.6 Question 33 The reactor is shutdown with RHR A in shutdown cooling. A full flow test (CST to CST) of HPCS is in progress when E22

-FIS-N656, HPCS Pump Discharge Flow fails downscale.

Following the instrument failure, CST level will:

A. drop and RPV level will rise.

B. drop and suppression pool level will rise.

C. rise and suppression pool level will drop.

D. remain steady.

Answer: B Explanation:

The full flow test is recirculating the CST. Transmitter FT

-N056 feeds FIS

-N656. With measured flow < 1206 gpm, F012, HPCS recirc. opens, moving water from the CST to the suppression pool.

A would be correct if the recirc valve discharged to the RCS.

C would be correct if the full flow test was recirculating the suppression pool and the recirc discharged to the CST.

D would be correct if equalizing a flow transmitter caused indicated flow to fail high.

Technical

References:

GFIG-OPS-E2201, HPCS figures, Figure 2, HPCS System Overview References to be provided to applicants during exam: None Learning Objective:

The HPCS lesson plan was not provided Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 211000 SLC Knowledge of the operational implications of the following concepts as they apply to STANDBY LIQUID CONTROL SYSTEM: Explosive valve operation Tier # 2 Group # 1 K/A # 211000 - K5.05 Rating 2.5 Question 34 Following an initiation of SLC 'A', which of the following sets of indications would be observed if the 'A' squib valve fails to open?

A. SLC storage tank level decreasing, SLC pump discharge pressure higher than reactor pressure, and reactor power decreasing.

B. SLC storage tank level decreasing, SLC pump discharge pressure lower than reacto r pressure, and reactor power constant.

C. SLC storage tank level constant, SLC pump discharge pressure higher than reactor pressure, and reactor power decreasing.

D. SLC storage tank level constant, SLC pump discharge pressure higher than reactor pressure, and reactor power constant.

Answer: D Comments: An applicant could reasonably argue B and D are correct.

Explanation:

A is wrong because it describes a good start of the SLC system B is wrong because if SLC pump discharge pressure is less than reactor pressure SLC would not be able to inject, therefore it is not indicative of a failure of the squib to fire but of the system not developing pressure C is wrong because SLC storage tank level would not remain constant if SLC was initiated.

The indication describe a failed instrument D is correct because tank level would remain constant and discharge pressure would be higher than reactor pressure because the pump would be recircing due to the relief valve lifting.

Technical

References:

GLP-OPS-C4100 Rev 10, page 18

-20 References to be provided to applicants during exam:

None. Learning Objective: (12) Describe the indications of SLC injection into the reactor.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

N ew X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(6) 55.43

Examination Outline Cross

-Reference Level RO 212000 RPS Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM will have on following:

Process radiation monitoring Tier # 2 Group # 1 K/A # 212000 K3.01 Rating 3.0 Question 35 The reactor is at 5% power during a startup when EPA

-S003A trips on undervoltage. With respect to the process radiation monitors, this event would result in; A. closure of Inboard Reactor Sample Valve, B33

-F019. B. closure of Outboard Reactor Sample Valve, B33

-F020.

C. generation of a mechanical vacuum pump trip signal.

D. no automatic trip signals or isolations.

D. Explanation:

On the loss of RPS A, the power would be automatically switched to inverter 1Y87.

Therefore, none of the RPS components would be affected making "D" correct. No process monitors will lose power in this event.

Technical

References:

GLP-OPS-D1721 GG UFSAR, 15.8.1 GG UFSAR, 15.2.2.5 References to be provided to applicants during exam: None. Learning Objective Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental

Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross-Reference Level RO 215003 IRM Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM: Detectors Tier # 2 Group # 1 K/A # 215003 - K6.04 Rating 3.0 Question 36 The plant is starting up following an outage. SRM/IRM overlap has been verified and all IRMs are indicating normally on range 5 when IRM F and IRM H fail high.

What is the expected system/plant response?

A. A full RPS scram.

B. A half scram on RPS Division A AN D a rod block.

C. A half scram on RPS Division B AND a rod block.

D. A rod withdrawal block ONLY. Answer: C Explanation:

A is wrong because both detectors are on the same division so full scram logic would not be made up. Plausible if an applicant doesn't know which IRMs are on which division B is wrong because this is the wrong division of RPS C is correct this is the correct division of RPS and a rod block is generated.

D is wrong because a rod block would be generated an RPS signal would also be generated. Technical

References:

GLP-OPS-C5102 - p. 20-31 References to be provided to applicants during exam:

None. Learning Objective: Concerning the major automatic features of the IRM System, state/identify the signals and setpoints associated with an IRM scram signal, including when the signals are bypassed.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 215003 IRM Ability to monitor automatic operations of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM including: Meters and recorders Tier # 2 Group # 1 K/A # 215003 A3.01 Rating 3.3 Question 37 When an Intermediate Range Monitor's "UP" button illuminates, the operator should press the associated channel's ________ button.

A. UP B. DRIVE OUT C. DOWN D. DRIVE IN Answer: A No need for a 2 part question.

Explanation:

The UP button is backlit when the light signifies that the IRM is at 75% of scale. Pressing the button raises the channel to the next highest scale. This condition is expected during a plant startup.

There is nothing intuitive about the meaning of this light. It could mean that the detector needs to be driven in (up), or is up and should be down (driven out). If could also logically mean that the current scaling is too high, which would occur during a plant shutdown

. Technical

References:

GLP-OPS-C5102 References to be provided to applicants during exam:

None Learning Objective:

GLP-OPS-C5102 For each of the following IRM System components, state/identify methods of operation, including locations of associated controls and indications: (7) Drive mechanism (7.1) IRM range select pushbuttons (7.2)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 215004 Source Range Monitor 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 consequences of those abnormal conditions or operations: Power supply degraded Tier # 2 Group # 1 K/A # 215004 - A2.01 Rating 2.7 Question 38 The reactor is starting up following a refueling outage when output voltage on Inverter 1Y96 indicates 0.0 volts. What is an expected automatic action of this condition and what would be the immediate operator action of Alarm Response 04 02-1H13-P680, to mitigate this failure so startup can continue?

A. Control rod block; bypass SRM C B. Control rod block; bypass SRM D C. Full RPS scram; bypass SRM C D. Full RPS scram; bypass SRM D Answer: A Explanation:

A is corre ct because SRM normally only generate rod blocks unless in special situations. This inverter feeds the power supply for SRM C.

B is wrong because the first half is correct but Inverter 1Y95 feeds SRM D.

C is wrong because there are no scram signals generated from SRMs unless the shorting links are removed. This is normally done during fuel loading and would not be done while starting up. It is plausible if an applicant doesn't understand this.

D is wrong because see answer C Technical

References:

GLP-OPS-C5101 , SOURCE RANGE MONITORING (SRM) SYSTEM

- C5101, Rev 3 References to be provided to applicants during exam:

None. Learning Objective: 10.1 Describe the interrelationships of the following systems with the SRM System: 120V AC Uninterruptible Power Supply (UPS) System. 8.2 Concerning the major automatic features of the SRM System, state/identify the signals and setpoints associated with an SRM rod withdrawal block signal, including when the signals are bypassed.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 215005 APRM / LPRM Knowledge of the physical connections and/or cause-effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the following: Traversing incore probe system Tier # 2 Group # 1 K/A # 215005 K1.13 Rating 2.6 Question 39 The traversing incore probe system is used to ensure that the _______ instrumentation remains operable: A. reactor protection system

B. control rod block C. post accident monitoring D. emergency core cooling system Answer: A Explanation:

TS 3.3.1 function 2b cannot remain operable because SR 3.3.1.1.7 (LPRM calibration) cannot be met.

The other answers are incorrect because TIPS is not used in maintaining operability for these systems.

This is RO level because the RO is required to recognize conditions that require TS entry.

Technical

References:

GLP-OPS-C5105, TIP system TS 3.3.1, function 2b, SR 3.3.1.1.7.

References to be provided to applicants during exam: None Learning Objective: GLP-OPS-C5105, TIP system Describe the interrelationships of the following systems with the TIP System: (10)

Process Computer (10.1)

Local Power Range Monitors (10.4)

Given plant conditions, identify entry into the Limiting Condition for Operation (LCO) for any Technical Specification related to the TIP System. (12)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 217000 RCIC Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation Tier # 2 Group # 1 K/A # 217000 - 2.1.7 Rating 4.4 Question 40 The plant is operating at rated power when you receive SUPP POOL LVL HI alarm on P601. Control room operators should immediately verify: A. RCIC suction is from the suppression pool B. RCIC suction is from the condensate storage tank C. HPCS operability D. FCV MOV-F022, RCIC Inboard Test return, is open Answer: A Explanation:

A is correct because this is the immediate action in ARP 0 4-1-02-1H13-P601. B is wrong because this is not listed in the ARP but it could be correct if someone believes that high suppression pool level would result in RCIC inoperability/unavailability.

C is wrong because it is not listed in the ARP but could be correct if someone believes the condition makes RCIC inoperable.

D is wrong because it is listed in the immediate actions to verify the valve is closed and not open. Technical

References:

ARP 04-1-02-1H13-P601 References to be provided to applicants during exam:

None. Learning Objective: (19) GLP-OPS-E5100. Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)- 55.43 Examination Outline Cross

-Reference Level RO 218000 ADS Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM

ADS logic operation Tier # 2 Group # 1 K/A # 218000 K5.01 Rating 3.8 Question 41 The reactor has experienced a transient the following conditions currently exist:

Reactor level has been less than Level 3 for 8 minutes Reactor level has been less than Level 1 for 5 minutes All ECCS equipment has responded as designed Drywell pressure is 0.5 psig and is rising at 0.1 psig per minute.

Based on these conditions, the Automatic Depressurization System relief valves:

A. have opened.

B. will open when drywell pressure reaches 1.39 psig

C. will open 105 seconds after drywell pressure reaches 1.39 psig.

D. will open in 4.2 minutes.

Answer: D I like this question better as it is more operationally focused.

Explanation:

I nput from the high drywell pressure channel is automatically bypassed if Reactor level stays

< -150.3 for 9.2 minutes. This condition assumes a LOCA outside the Drywell with degraded high pressure injection systems, and thus takes override action to initiate ADS to recover RPV level.

\ Technical

References:

GLP-OPS-E2202 References to be provided to applicants during exam: None Learning Objective: GLP-OPS-E2202 Describe the signals and setpoints for the initiation of the ADS System for both automatic and manual initiation. (15.0)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 223002 PCIS/Nuclear Steam Supply Shutoff Ability to manually operate and/or monitor in the control room: valve closures Tier # 2 Group # 1 K/A # 223002 - A4.01 Rating 3.6 Question 42 What combination of NSSSS manual initiation pushbuttons would result in an isolation of MSIVs ONLY? A. Division 1 and Division 2 B. Division 1 and Division 4 C. Division 2 and Division 3 D. Division 2 and Division 4 Answer: A Explanation:

A is correct because either 1 and 2 or 3 and 4 will give you a MSIV isolation.

B is wrong because this would give you a MSIV closure and a Div 1 NSSSS, Containment, Drywell, and Auxiliary building isolation.

C is wrong because this would give you a MSIV closure and a Div 2 NSSSS, Containment, Drywell, and Auxiliary building isolation.

D is wrong because should create a half isolation signal.

Technical

References:

GLP-OPS-M701, Containment and Drywell Instrumentation and Control System

- M71, Rev 11 p.56 References to be provided to applicants during exam:

None. Learning Objective: State/identify the location(s) from which the Containment &

Drywell Instrumentation & Control System isolation signals can be initiated. (8.2)

Describe the logic arrangements which actuate the isolation logics listed in Objective 7. (9)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(6)

Examination Outline Cross

-Reference Level RO 239002 SRVs Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or interlocks which provide for the following: Insures that only one or two safety/relief valves reopen following the initial portion of a reactor isolation event (LLS logic): Plant

-Specific Tier # 2 Group # 1 K/A # 239002 K4.01 Rating 3.9 Question 43 The plant is at rated power when the MSIVs close causing a reactor scrama reactor isolation occurs. Reactor pressure reaches 1 120 psig resulting in m Multiple SRVs cycling initially lift, then all SRVs close. Following the initi al transient, the SRVs would be expected to automatically maintain reacto r pressure between

When pressure subsequently rises, only ____(1)_____ should lift when pressure reaches (2) . psig. (2) A. 926 and 1033 psig.F051D 1103 B. 936 and 1073 psig.F051D 1033 C. 946 and 1113 psig.F051B 1103 D. 1013 and 1113 psig.F051B 1033 Answer: B Explanation:

Low-Low Set initiates automatically when RPV pressure reaches SRV F051D's normal lift setpoint of 1103 psig.

Once initiated, six SRVs are capable of operating in the Low

-Low Set mode with the following adjusted opening and closing setpoints: One SRV, F051D, lifts at 1033 psig and blows down to 926 psig. A second SRV, F051B, lifts at 1073 psig and blows down to 936 psig. Technical

References:

GLP-OPS-E2202 References to be provided to applicants during exam:

Learning Objective: GLP-OPS-E2202 For each mode of operation of the SRVs, state/identify the number of valves which open and the associated opening setpoints. (11.2)

Formatted:

HighlightFormatted:

Indent: Left: 0.63", No bullets ornumbering Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 259002 Reactor Water Level Control Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM: Reactor water level input Tier # 2 Group # 1 K/A # 259002 - K6.08 Rating 3.5 Question 44

If the plant is in single element mode with automatic level channel selection mode, and the controlling level channel fails high the DFCS will de

-select the failed channel and select-

A. the upset range channel irrespective of the magnitude.

B. an intermediate value from the remaining narrow range and upset range channels.

C. the highest value from the remaining narrow range and upset range channels.

D. the lowest value from the remaining narrow range and upset range channels.

Answer: B Explanation:

A is wrong because this will happen only if the upset channel is the intermediate value.

B is correct because this is design of the system.

C is wrong because the system is designed to select the intermediate value.

D is wrong because the system is designed to select the intermediate value. Technical

References:

GLP-OPS-C3400, DIGITAL FEEDWATER CONTROL SYSTEM (DFCS)

- C34, Rev 12 References to be provided to applicants during exam:

None. Learning Objective: State/identify the Level Control Subsystem's response to failure of one or more Narrow Range level control channels when in the Automatic Level Channel Selection Mode. (3.6). Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 261000 SGTS Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: Off-site release rate Tier # 2 Group # 1 K/A # 261000 K3.02 Rating 3.6 Question 45 Procedure 04 01-T48-1, Standby Gas Treatment, contains a caution to prevent an automatic start of SBGT if painting in the Auxiliary Building. This is because paint fumes in the SBGT System can:

A. result in a fire due to the electric heater in the system.

B. plug the HEPA filters and reduce system flowrates.

C. reduce the efficiency of the charcoal adsorber leading to increased off

-site release rates.

D. render the demister inoperable and subsequent fouling of the charcoal.

Answer: C Explanation:

The presence of a volatile organic substance (e.g. painting) in the SBGT system reduces the efficiency of the adsorber material leading to an increased release rate of radioactive particles to the environment.

Technical

References:

GLP-OPS-T4801 References to be provided to applicants during exam:

None Learning Objective:

GLP-OPS-T4801 State/identify the design bases of the Standby Gas Treatment System. (2.0)

Question Source:

Bank # (note changes; attach parent)

Modified Bank # New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 262001 AC Electrical Distribution Ability to predict and/or monitor changes in parameters associated with operating the A.C. ELECTRICAL DISTRIBUTION controls including: Bus voltage Tier # 2 Group # 1 K/A # 262001 - A1.03 Rating 2.9 Question 46 A degraded voltage on the DIV 1 and DIV 3 ESF buses results in bus voltages to be at 3600 volts for 7 seconds. Following the transient, the status of the DIV 1 and DIV 3 EDGs would be______. A. Both EDGs running B. Neither EDG running C. DIV 1 EDG running and DIV 3 EDG NOT running D. DIV 1 EDG NOT running and DIV 3 EDG running Answer: B Explanation:

A is wrong because neither of the diesel would start.

For the DIV 1 diesel there is a 9 second time delay between 2912 volts and 3744 volts before the DIV 1 diesel starts.

For the DIV 3 diesel there is a 4 second time delay between 3045 and 3660 volts for the diesel to start. This start sequence has to be coincident with a LOCA signal. If there is no LOCA signal present coincident with the 4 second time delay then there is a 5 minute time delay until DIV 3 diesel starts.

B is correct because of the explanation in A C is wrong because see above D is wrong because see above Technical

References:

GLP-OPS-R2100.12 , LOAD SHEDDING & SEQUENCING SYSTEM AND ESF AC POWER DISTRIBUTION SYSTEM

- R21, Rev 12, p.19 GLP-OPS-P8100 , HIGH PRESSURE CORE SPRAY (HPCS) DIESEL GENERATOR SYSTEM - P81, Rev 14, p. 40 References to be provided to applicants during exam:

None.

Learning Objective: Discuss the interrelationships of the following systems with the HPCS Diesel Generator System: (18). State/identify the signals and setpoints that will produce a Bus Undervoltage signal to LSS Shed System and the actions associated with each signal. (12)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 262001 AC Electrical Distribution Knowledge of the effect that a loss or malfunction of the A.C. ELECTRICAL DISTRIBUTION will have on following: Emergency generators Tier # 2 Group # 1 K/A # 262001 K3.02 Rating 3.8 Question 47 The plant is at rated power when a grid disturbance causes the voltage on buses 15AA and 16AB to drop in a step change to a sustained value of 3700 volts at time T=0. The associated incoming offsite bus feeder breakers open at time T= ____(1)____ and T the emergency diesel generator will receive s an automatic start signal

_____a tseconds later

. time T= ____(2)____.

(1) (2) A. 0.5 0.5 B. 0.5 1.0 C. 9.0 9.0 D. 9.0 9.5 Answer: D Explanation:

With bus voltage < 3744 volts, after a 9 second delay, the offsite bus feeder breakers open. This activates the < 2912 volt timer, which, after 0.5 seconds sends a start signal to the associated diesel generator.

The 0.5 second opening of incoming breakers is credible because this is the time delay of the 70% timer. The start of the EDG simultaneous with breaker opening is credible, since this is what happens with the 70% timer.

Technical

References:

GLP-OPS-R2100 GLP-OPS-P7500 References to be provided to applicants during exam:

None Learning Objective: GLP-OPS-P7500 Describe the automatic initiation logic for a Standby Diesel Generator for degraded power conditions and a Loss of Coolant Accident (LOCA). (8.0) GLP-OPS-R2100 State/identify the signals and setpoints that will produce a bus undervoltage signal to LSS shed system and the actions associated with each signal. (12)

Question Source:

Ban k # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 262002 UPS (AC/DC)

Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.)

D.C. electrical power Tier # 2 Group # 1 K/A # 262002 - K6.02 Rating 2.8 Question 48 If DC power is lost to an ESF inverter the static switch on the inverter will (a)? When the condition is corrected the static switch for the ESF inverter will automatically/manually be transferred back to the DC source?

A. only transfer if the IN SYNC light is illuminated automatically B. transfer regardless of IN SYNC light automatically C. only transfer if the IN SYNC light is illuminated manually D. transfer regardless of IN SYNC light manually Answer: D Explanation:

Excerpts from the training material:

If the inverter output and the AC source are out of sync the static switch will not transfer. An out of sync transfer would cause major damage to the inverter and potential damage to the loads.

If the DC source has low or no voltage IN SYNC is not required, most of the inverters will auto transfer.

Upon a return of the normal source the BOP Inverters (except 1Y99) will automatically (-2 seconds of normal power) return to the normal power supply.

The ESF Inverters and 1Y99 have had the auto return feature disabled internally on the inverter. (Each inverter has a switch to enable or disable this feature.) If DC power is lost to an ESF Inverter, the DC input circuit breaker trips on undervoltage.

A is wrong because a loss of DC would result in no voltage so the IN SYNC light would not come into play. Also, only the BOP buses will automatically transfer when normal DC power is restored.

B is wrong because ESF inverters have the automatic return to normal supply disable

d. C is wrong because the inverter will transfer regardless of the IN SYNC light since the inverter has zero voltage.

D is correct because the transfer will happen regardless of the IN SYNC light and there is no automatic return to the normal source for ESF inverters.

Technical

References:

GLP-OPS-L6200, Static Inverter System

- L62, Rev 10 References to be provided to applicants during exam:

None. Learning Objective: Describe the automatic feature of the Static Inverter System that transfers power sources and which inverters have the following features: ( 4)

Inverter to alternate power source ( 4.1)

Alternate power source to inverter upon return of inverter power (4.2)

Describe the operation of the Static Inverter using: (5)

Static Switch

- Auto (5.1) Static Switch

- Manual (5.2)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 263000 DC Electrical Distribution Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Grounds Tier # 2 Group # 1 K/A # 263000 A2.01 Rating 2.8 Question 49 It is important to promptly eliminate a ground on a DC bus because:

A. one ground can cause unpredictable spurious operation of equipment.

B. subsequent grounds will be masked, and multiple grounds can produce unanalyzed results. C. a shock hazard exists at the grounded equipment.

D. a fire hazard exists due to excessive current flow at the ground location.

Answer: B Explanation:

Failure to promptly eliminate a single ground could mask subsequent grounds. Multiple grounds could lead to unpredictable spurious operation of equipment, inoperable equipment, unanalyzed loads on batteries, or unanalyzed equipment failure modes that could be expected to occur during harsh environments associated with accident conditions. The first ground does not produce adverse effects because there is no path for current flow with one ground in an ungrounded system.

Technical

References:

GLP-OPS-L1100 References to be provided to applicants during exam:

Learning Objective: GLP

-OPS-L1100 State/identify the purpose/function of Gr ound Detection Systems. (5.3)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO DC Electrical Distribution Knowledge of electrical power supplies to the following: Major D.C. loads Tier # 2 Group # 1 K/A # 263000 - K2.01 Rating 3.1 Question 50 The Diesel Generator 11 __________ is(are) powered from the ESF DC distribution system.

A. generator field flash B. fuel oil booster pump C. lube oil pump D. air compressors Answer: A Comments: The relay panel also receives AC power. This question is ok with changes made.

Explanation:

B is wrong because it is powered from a BOP bus C is wrong because it is AC powered D is wrong because it is AC powered Technical

References:

04-1-01-L11-1_123 and GLP

-OPS-P7500_024-1 References to be provided to applicants during exam: None. Learning Objective: GLP-OPS-L1100. (8) State/identify the major loads of the Plant DC System Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 264000 EDGs Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: A.C. electrical distribution Tier # 2 Group # 1 K/A # 264000 K3.02 Rating 3.9 Question 51 The voltage regulator on a standby diesel generator has a malfunction that limits its maximum voltage to 3900 VAC.

In this condition, the output breaker can be closed:

1. automatically.

2. remotely (P864) when paralleling with an offsite source.

3. remotely (P864) when energizing a dead bus.

A. 1, 2 and 3 B. 2 and 3 only C. 3 only D. 1 only Answer: C Explanation:

DG at rated voltage is an interlock for closing the breaker automatically or remotely when paralleling. It is not an interlock when all offsite feeder breakers are open Technical

References:

Table 4 of GLP

-OPS-P7500 References to be provided to applicants during exam: None Learning Objective:

GLP-OPS-P7500 State/identify the start sequence and interlocks involved in starting the Standby Diesel Generator System during the following conditions: (15.0) LOCA signal present (15.1) LOP/BUV signal present (15.2) Local Emergency Start signal present (15.3) Normal operations (15.4)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 300000 Instrument Air Knowledge of the connections and/or cause effect relationships between INSTRUMENT AIR SYSTEM and the following: Cooling water to compressor Tier # 2 Group # 1 K/A # 300000 - K1.04 Rating 2.8 Question 52 Normal cooling water for the instrument air compressors is provided by _______:

A. Plant Service Water.

B. Standby Service Water.

C. Component Cooling Water.

D. Turbine Building Cooling Water.

Answer: D Explanation:

A is wrong because it does not provide cooling but is plausible if the applicant believes that since the system is not safety related it would come from a non

-safety related source.

B is wrong because this is the emergency source of water.

C is wrong because it does not provide cooling but is plausible if the applicant doesn't realize that component cooling water only provides cooling to potentially radioactive sources.

D is correct because it is the normally aligned source of cooling water.

Technical

References:

Document where the correct answer is found References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 400000 Component Cooling Water Knowledge of electrical power supplies to the following:

CCW valve s Tier # 2 Group # 1 K/A # 40000 K2.02 Rating 2.9 Question 53 The power supply to P42

-F116, CCW RTN FRM DW is:

A. Bus 11DA B. LCC 16BB3 C. MCC17B01 D. MCC 15B11 Answer: D Explanation:

A is wrong because the "DA" shows that this is a 125V DC bus, and the valve is powered from 480 VAC.

B is wrong because the valve motor is < 50 hp. It is powered from an MCC (not an LCC). C is the power supply to the B CCW pump.

C is wrong because MCC 17B11 is a division III bus which does not power any CCW components.

Technical

References:

04-1-01-P42-1, CCW SOI GLP OPS-R2700, Normal AC Power Distribution System References to be provided to applicants during exam:

Learning Objective: GLP OPS-R2700 Identify components using the GGNS numbering system. (16)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental

Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross-Reference Level RO 201001 CRD Hydraulic Knowledge of the physical connections and/or cause

-effect relationships between CONTROL ROD DRIVE HYDRAULIC SYSTEM and the following: Condensate System Tier # 2 Group # 2 K/A # 201001 - K1.0 1 Rating 3.1 Question 54

During normal operation, the suction pressure for the control rod drive pumps is from the ________. On a loss of the condensate system the suction of the control rod drive pumps is from the _________.

A. condensate pumps; CST B. condensate booster pumps; CST C. condensate pumps; demineralized water system D. condensate booster pumps; demineralized water system Answer: A Explanation:

A is correct B is wrong because the suctions from the condensate pumps C is wrong because the service water system does not feed the CRD system but is plausible because the water system is very clean and could be used in the CRD system D is wrong because of B and C Technical

References:

GLP-OPS-C111A References to be provided to applicants during exa m: None. Learning Objective: 11.1 Describe the interrelationships between the CRWST and the CRD Hydraulic System.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(6)

Examination Outline Cross

-Reference Level RO 201003 Control Rod and Drive Mechanism Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD AND DRIVE MECHANISM

Hydraulics Tier # 2 Group # 2 K/A # 201003 K5.01 Rating 2.6 Question 55 In order to continuously withdraw a CRDM, the collet fingers are continuously hel d: A. outward by hydraulic pressure B. outward by mechanical interaction with the index tube C. inward by hydraulic pressure D. inward by mechanical interaction with the index tube Answer: A Explanation:

The collet fingers are initially opened by mechanical interaction with the index tube by a slight insertion, but they are subsequently held open by hydraulic pressure applied to a piston. The fingers move outward to unlatch the rod, they move inward to latch the rod and prevent outward movement.

Technical

References:

GLP-OPS-C111A_008-1 References to be provided to applicants during exam:

Learning Objective:GLP-OPS-C111A Describe in detail the following control rod operations: (10.0) o Control rod insertion (10.1) o Control rod withdrawal (10.2)

GLP-OPS-C111AB Describe the following CRD Mechanism evolutions and operations: (5)

Operation of the collet locking mechanism. (5.3)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 201005 RCIS Ability to predict and/or monitor changes in parameters associated with operating the ROD CONTROL AND INFORMATION SYSTEM (RCIS) controls including: First stage shell pressure/turbine load: BWR

-6 Tier # 2 Group # 2 K/A # 201005 - A1.01 Rating 3.2 Question 56 If the plant is at 65% power with 1 bypass valve full open the Rod Control and Information System will allow control rods to be withdrawn a maximum of (a) notch/notches at a time.

A. 1 B. 2 C. 4 D. 12 Answer: C Explanation:

"Above the High Power Setpoint (62% RTP), rod withdrawal is limited to two notches by the RWL. For instance, if reactor power is 63% with the bypass valves open, RC&lS will interpret the lower first stage turbine pressure as a lower reactor power. I f RC&lS sees a first stage turbine pressure equivalent to 60%, it will allow rods to be withdrawn four notches (below the HPSP) instead of the limit of two notches as required above the HPSP." Technical

References:

GLP-OPS-C1102, Rod Control and Information Systems (RC&IS), Revision 6, p. 20 References to be provided to applicants during exam:

None. Learning Objective: Describe the reason for restricting Main Steam Bypass Valve operation when at power. (7.2)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41(b)(6)

Examination Outline Cross

-Reference Level RO 204000 RWCU Ability to manually operate and/or monitor in the control room: Valve controllers Tier # 2 Group # 2 K/A # 204000 A4.02 Rating 2.9 Question 57 Reactor pressure is 150 psig during a plant heatup with RWCU blowdown to the main condenser maintaining RPV level.

Opening 1G33

-F003, RWCU SYS BLWDN FLO CONT valve too far can result in the need to subsequently:

A. replace RWCU pump seals.

B. re-open 1G33-F004.

C. backwash and precoat the F/Ds.

D. re-open the primary containment isolation group 8 valves.

Answer: B Explanation:

Opening F033 will cause non

-regenerative HX outlet temp to rise because of a loss of cooling flow in the regenerative HX. If it reaches 140°F, 1GG33

-F004 closes and the RWCU pumps trip.

A is wrong because at this reactor pressure the RWCU pumps are in the POST

-Pump mode. RWCU will isolate at 140° which is well within design of RWCU seals.

C is wrong because the demineralizer hold pump is designed to auto start as flow drops to prevent loss of precoat.

D is wrong because this activity should not result in any condition that causes the group 8 isolation.

Technical

References:

04-1-02-1H13-P680-11A-C6 ARI, RWCU FLTR DMIN INL TEMP HI 140°F.

04-1-01-G33-1, Reactor Water Cleanup, caution on top of page 55.

References to be provided to applicants during exam:

None Learning Objective: GLP

-OPS-G3336 Describe any interlocks, trips, or automatic actions associated with the following Reactor Water Cleanup System components: (8.0) System Isolation Valves (8.6)

Concerning SOI 04 01-G33-1: (10.0) Identify the reasons for the precautions, limitations, cautions, warnings and notes given in the SOI. (10.2)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 219000 RHR/LPCI: Torus/Pool Cooling Mode Knowledge of the purpose and function of major system components and controls Tier # 2 Group # 2 K/A # 219000 - 2.1.28 Rating 4.1 Question 58 The suppression pool cooling mode of RHR is designed to limit suppression pool temperature to a maximum of (a) during normal conditions to ensure suppression pool temperature does not exceed 185 o F under accident conditions. This is accomplished in the suppression pool cooling mode by regulating flow through the-A. 95 oF; F024A/B, RHR A/B test return to suppression pool B. 95 oF; F048A/B, heat exchanger A/B bypass valve C. 110 oF; F024A/B, RHR A/B test return to suppression pool D. 110 oF; F048A/B, heat exchanger A/B bypass valve Answer: B Explanation:

A is wrong because the valve listed is not throttled in the suppression pool cooling mode B is correct because this is the temperature listed in the training manual and the only valve that should be throttled in SOI 04 01-E12, Residual Heat removal System.

C is wrong because this temperature is not listed in the training manual but plausible since it is listed in the EOP (EP

-2). Also this is the incorrect valve for regulating temperature D is wrong because see answer C.

Technical

References:

GLP-OPS-E1200, Residual Heat Removal (RHR) System

- E12 SOI 04-1-01-E12, Residual Heat removal System References to be provided to applicants during exam:

None. Learning Objective: State/identify the purpose/function of the Suppression Pool Cooling mode of RHR (1.4)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 219000 RHR/LPCI: Torus/Pool Cooling Mode Knowledge of the purpose and function of major system components and controls Tier # 2 Group # 2 K/A # 219000 - 2.1.28 Rating 4.1 Question 58 The suppression pool cooling mode of RHR is designed to limit suppression pool temperature to a maximum of (a) during normal conditions to ensure suppression pool temperature does not exceed 185 o F under accident conditions. This is accomplished in the suppression pool cooling mode by regulating flow through the- A. 95 oF; F024A/B, RHR A/B test return to suppression pool B. 95 oF; F048A/B, heat exchanger A/B bypass valve C. 110 oF; F024A/B, RHR A/B test return to suppression pool D. 110 oF; F048A/B, heat exchanger A/B bypass valve Answer: B Explanation:

A is wrong because the valve listed is not throttled in the suppression pool cooling mode B is correct because this is the temperature listed in the training manual and the only valve that should be throttled in SOI 04 01-E12, Residual Heat removal System.

C is wrong because this temperature is not listed in the training manual but plausible since it is listed in the EOP (EP

-2). Also this is the incorrect valve for regulating temperature D is wrong because see answer C.

Technical

References:

GLP-OPS-E1200, Residual Heat Removal (RHR) System

- E12 SOI 04-1-01-E12, Residual Heat removal System References to be provided to applicants during exam:

None. Learning Objective: State/identify the purpose/function of the Suppression Pool Cooling mode of RHR (1.

4) Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 219000 RHR/LPCI: Torus/Pool Cooling Mode Knowledge of the purpose and function of major system components and controls Tier # 2 Group # 2 K/A # 219000 - 2.1.28 Rating 4.1 Question 58 The suppression pool cooling mode of RHR is designed to limit suppression pool temperature to a maximum of (a) during normal conditions to ensure suppression pool temperature does not exceed 185 o F under accident conditions. This is accomplished in the suppression pool cooling mode by regulating flow through the-A. 95 oF; F024A/B, RHR A/B test return to suppression pool B. 95 oF; F048A/B, heat exchanger A/B bypass valve C. 110 oF; F024A/B, RHR A/B test return to suppression pool D. 110 oF; F048A/B, heat exchanger A/B bypass valve Answer: B Explanation:

A is wrong because the valve listed is not throttled in the suppression pool cooling mode B is correct because this is the temperature listed in the training manual and the only valve that should be throttled in SOI 04 01-E12, Residual Heat removal System.

C is wrong because this temperature is not listed in the training manual but plausible since it is listed in the EOP (EP

-2). Also this is the incorrect valve for regulating temperature D is wrong because see answer C.

Technical

References:

GLP-OPS-E1200, Residual Heat Removal (RHR) System

- E12 SOI 04-1-01-E12, Residual Heat removal System References to be provided to applicants during exam:

None. Learning Objective: State/identify the purpose/function of the Suppression Pool Cooling mode of RHR (1.4)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross

-Reference Level RO 233000 Fuel Pool Cooling/Cleanup Knowledge of FUEL POOL COOLING AND CLEAN-UP design feature(s) and/or interlocks which provide for the following: Pool cooling during loss of cooling accident: BWR-6 Tier # 2 Group # 2 K/A # 233000 - K4.08 Rating 2.6 Question 60 Both fuel pool cooling water pumps are running when a LOCA signal is received. Both pumps will trip and- A. both will be automatically restarted by the sequencer.

B. neither pump can be restarted until the LOCA signal is reset.

C. either pump can be manually restarted immediately.

D. either pump can be manually restarted following load shedding and sequencing.

Answer: D Explanation:

A is wrong because the pumps are not listed as loads that automatically sequence on.

B is wrong because the LOCA signal does not lock out the pumps.

C is wrong because the LSS signal can't be in or the pump will trip and the signal will be in while loads are sequencing.

D is correct because see A, but they are listed as loads that are not locked out and can be manually started.

Technical

References:

04-1-01-R21-1, Load Shedding and Sequencing System GLP-OPS-R2100, Load Shedding and Sequencing System and ESF AC Power Distribution System References to be provided to applicants during exam:

None. Learning Objective: (17) Describe the sequencing of Loads on a LOCA signal.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7), 55.41(b)(8)

Examination Outline Cross

-Reference Level RO 245000 Main Turbine Gen. / Aux Knowledge of the effect that a loss or malfunction of the following will have on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS : Main steam Tier # 2 Group # 2 K/A # 245000 K6.08 Rati ng 3.0 Question 61 One SRV failing open at rated power will result in generator megawatts __(1)____ and feedwater injection temperature __(2)__. (1) (2) A. lowering lowering B. lowering remaining constant

C. remaining constant lowering D. remaining constant remaining constant Answer: A Explanation:

The SRV operation lowers steam pressure, EHC responds by lowering turbine steam flow. The lowered turbine steam flow leads to reduced extraction steam flow, which lowers feedwater temperature.

Technical

References:

ARI 04-01-02-1H13-P601-19A-A5, SRV/ADS VLV OPEN/DISCH LINE PRESS HI References to be provided to applicants during exam:

None Learning Objective:

GLP-OPS-E2202 Identify the system interrelationships between each of the following systems and the ADS/SRV System: (19.0)

Nuclear Boiler B13/21 (19.2)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO Reactor Feedwater Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR FEEDWATER SYSTEM: Reactor water level control system Tier # 2 Group # 2 K/A # 259001 - K6.07 Rating 3.8 Question 62

The plant is operating at steady state with reactor power at 90% when steam flow transmitter N34-N030A fails upscale. Following this failure, the following occurs:

STEAM FLOW FAIL light begins flashing DFCS TROUBLE alarm annunciates on H13

-P680 Based on this malfunction;

A. Feedflow will increase and the reactor will scram on high water level.

B. Feedflow will decrease and the reactor will scram on low water level.

C. DFCS will automatically shift to single element control and feedflow remains fairly constant. D. DFCS automatically deselects N34

-N030A, remains in 3

-element control, and feedflow remains fairly constant.

Answer: C Explanation:

A is wrong; feedflow will remain fairly constant because the DFCS shifts to single element control and maintains reactor water level at setpoint.

B is wrong; feedflow will remain fairly constant because the DFCS shifts to single element control and maintains reactor water level at setpoint.

C is correct; feedflow will remain fairly constant because the DFCS shifts to single element control and maintains reactor water level at setpoint.

D is wrong because DFCS automatically shifts to single element control.

Technical

References:

GLP-OPS-C3400.12, page 31.

References to be provided to applicants during exam:

None. Learning Objective: C3400.12 - 5.3 - State/identify the consequences of a steam flow channel failure.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b).4 55.43.5 Examination Outline Cross

-Reference Level RO 268000 Radwaste Knowledge of the effect that a loss or malfunction of the RADWASTE will have on following: Drain sumps Tier # 2 Group # 2 K/A # 268000 K3.04 Rating 2.7 Question 63 Isolating the floor drain collection tank and both waste surge tanks will result in not being able to pump down the ______.

A. Turbine Building West Equipment Drain Sump B. HPCS Room Sump C. Auxiliary Building North Equipment Drain Sump D. Turbine Building North Chemical Waste Sump Answer: B Explanation:

Of these sumps, only B pumps to these tanks (dirty RW). A and C discharge to the Equipment collection drain tanks (clean RW). D discharges to the Miscellaneous Chemical Waste Receiver Tank

Technical

References:

GLP-OPS-G1718.08 References to be provided to applicants during exam:

None Learning Objective: State/identify the purpose/function and location of the Equipment Drain Collector Tanks, Floor Drain Collector Tank and Waste Surge Tanks. (3.0) Discuss the water quality of the various inputs to the tanks from operating the plant and the importance of segregation of the waste water. (3.3)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO 27100 Offgas Knowledge of the physical connections and/or cause

-effect relationships between OFFGAS SYSTEM and the following: Plant air systems Tier # 2 Group # 2 K/A # 271000 - K1.07 Rating 2.7 Question 64 The air used in the Offgas System is supplied by the _______ Air System and on a loss of air F061A/B, Offgas Condenser Drain Valves, fail _______.

A. Instrument; open B. Instrument; close C. Service; open D. Service; close Answer: A Explanation:

A is correct. B is wrong because the valves fail open.

C is wrong because Instrument Air is used.

D is wrong because Insrument Air is used and the valves fail open.

Technical

References:

GLP-OPS-N6465, References to be provided to applicants during exam:

None. Learning Objective: (8.4) State/identify the location from where the Offgas Condenser Drain Valves, F016A/B, can be operated (14.2) Discuss the interrelationships of the Instrument Air System with the Offgas/Vault Refrigeration System Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(13)

Examination Outline Cross

-Reference Level RO 290003 Control Room H VAC Ability to monitor automatic operations of the CONTROL ROOM HVAC including: Initiation/failure of fire protection system Tier # 2 Group # 2 K/A # 290003 A3.02 Rating 3.0 Question 65 The operating control room air conditioning unit will automatically shut down if:

A. smoke is detected in the return ductwork from the control room.

B. high-high radiation is sensed in the outside air intake duct.

C. high-high temperature is sensed on an operating fresh air unit filter train.

D. a Freon leak is detected in the control room air conditioner cooling coils.

Answer: A Explanation:

Upon sensing the presence of smoke in the return ductwork from the Control Room or in the fresh air intake duct, the smoke detector causes the following actions to take place: The operating air conditioning unit shuts down.

The operating air conditioning unit inlet and outlet dampers close.

The fire detection system sounds the fire alarm.

B is wrong because this isolates the control room but does not trip the air conditioning unit. C is wrong because this shuts down the running control room standby fresh air units.

D is wrong because the Freon leak alarm has no automatic actions associated with it..

Technical

References:

ARI 04-S-02-SH13-P855-1A-B4 , CONT RM HVAC FREON HI GLP-OPS-Z5100 References to be provided to applicants during exam:

None Learning Objective: GLP-OPS-Z5100 Discuss the interrelationships of the Plant Fire Protection System with the Control Room HVAC System. (13.6)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO Conduct of Operations Knowledge of shift turnover practices.

Tier # 3 Group # K/A # 2.1.3 Rating 3.0 Question 66 What is the MAXIMUM time a shift can be one less than the minimum required by regulations? Can the exception to minimum staffing requirements be invoked during shift change for a crewperson being late or absent?

A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; Yes B. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; No C. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; Yes D. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; No Answer: D Explanation:

A is wrong because the time is less than the maximum allowed but is a logical time to get an onsite replacement.

B is wrong because of answer A. The conduct of operations procedure does not allow for this exception for tardiness or absence. The second part of the question is correct.

C is wrong because this is the correct time, but the conduct of operations procedure does not all for this exception D is correct because this is the maximum time allowed and the correct interpretation of the conduct of operations procedure.

Technical

References:

EN-OP-115, Conduct of Operations , Rev. 14, Attachment 9.3 References to be provided to applicants during exam:

None. Learning Objective: N/A Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross

-Reference Level RO Conduct of Operations Knowledge of the refueling process.

Tier # 3 Group # K/A # 2.1.41 Rating 2.8 Question 67 When removing the reactor vessel head, reactor vessel level should be maintained using the ________ range.

A. Fuel Zone B. Wide C. Upset D. Shutdown Answer: D Explanation: The target is to maintain vessel level approximately one foot below the vessel flange (196" to 200" on shutdown range). Level too high would result in water spilling over the flange. Level too low increases dose in the refueling area. Only the Shutdown Range measures level this high. Technical

References:

03-1-01-5 Refueling, Step 5.15.15 GFIG-OPS-B2101, Figure 2 References to be provided to applicants during exam:

Learning Objective: GLP-OPS-B2101 State/identify the indicating ranges of the following RPV Water Level Instruments: (5.0) o Reactor Water Level Narrow Range (5.1) o Reactor Water Level Wide Range (5.2) o Reactor Water Level Shutdown Range (5.3) o Reactor Water Level Upset Range (5.4) o Reactor Water Level Fuel Zone Range (5.

5) Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO Conduct of Operations Knowledge of conduct of operations requirements.

Tier # 3 Group # K/A # 2.1.1 Rating 3.8 Question 68 Per ENP-OP-115, Conduct of Operations, which of the following tasks is allowed to be performed without a procedure in hand?

A. Adjusting circulating water blowdown flow.

B. Initiating SLC during an ATWS.

C. Resetting a full scram.

D. Rotating CRD pumps.

Answer: A Explanation:

A is correct because the activity is specified in Attachment 9.3 of ENP

-OP-115, Conduct of Operations B is wrong because it is not listed in the procedure.

C is wrong because it is not listed in the procedure but resetting a half scram is.

D is wrong because it is not listed in the procedure.

Technical

References:

Attachment 9.3 of ENP

-OP-115, Conduct of Operations, Rev 14 References to be provided to applicants during exam:

None. Learning Objective: N/A. Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(10) 55.43 Examination Outline Cross

-Referen ce Level RO Conduct of Operations Knowledge of the process used to track inoperable alarms.

Tier # 3 Group # K/A # 2.1.43 Rating 3.0 Question 69 A single length of red tape placed diagonally across an alarm window means the annunciator:

A. card has been removed.

B. is not functioning properly due to faulty alarm circuitry.

C. inputs have been verified to cause a nuisance alarm.

D. deficiency affects technical specification operability.

Answer: B Explanation:

A is wrong because a removed card is identified by two lengths of red tape placed to form an 'X'. C is wrong because an annunciator with input bypassed due to nuisance alarm is designated with two vertical lengths of red tape.

D is wrong because this condition leads to further documentation but not a window marking Technical

References:

02-S-01-25 Deficient Equipment Identification, sections 6.2 and 6.3.

References to be provided to applicants during exam:

Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

55.43 Examination Outline Cross-Reference Level RO Equipment Control Ability to determine operability and/or availability of safety related equipment Tier # 3 Group # K/A # 2.2.37 Rating 3.6 Question 70 Can systems be considered operable if a system fails to meet acceptance criteria between required performances of the surveillance?

Can systems be considered operable when auxiliary equipment required to be operable is inoperable, but the system can still meet its surveillance requirement acceptance criteria?

A. Yes/Ye s B. Yes/No C. No/Yes D. No/No Answer: D Explanation:

A is wrong because SR 3.0.1 clearly defines that failure can to meet a SR can happen anytime and if so the LCO is no longer met. Also, the second question tests the definition of OPERABLE/OPERABILITY and the answer does not meet this definition.

B is wrong because of the first half of A.

C is wrong because of the second half A.

D is correct because it meets SR 3.0.1 and the definition of operability.

Technical

References:

Technical Specifications SR 3.0.1 and Technical Specification definition of operability References to be provided to applicants during exam:

None. Learning Objective: N/A Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross

-Reference Level RO Equipment Control Ability to track Technical Specification limiting conditions for operations.

Tier # 3 Group # K/A # 2.2.23 Rating 3.1 Question 71 A valve is re

-positioned to satisfy the requirements of a technical specification required action and is documented on an eSOMS LCOTR.

This valve would be identified with a ______ tag.

A. No B. Danger C. Test & Maintenance D. Caution Answer: D Explanation:

A no tag identifies components that cannot be tagged but require positioning when hanging and removing a tagout.

There is no LCOTR tag.

A test & maintenance tag makes the tagout holder accountable for the component's status.

Technical

References:

02-S-01-17, Control of Limiting Conditions for Operation References to be provided to applicants during exam:

Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO Radiation Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. Tier # 3 Group # K/A # 2.3.5 Rating 2.9 Question 72 Taking the mode switch from Operate to Off on an Area Radiation Monitor causes the FAIL SAFE green light to extinguish and the-A. LOCAL red light to illuminate.

B. HIGH ALARM/RESET pushbutton to illuminate.

C. AREA RAD PNL P845 TROUBLE annunciator on 1H13

-P601 to annunciate.

D. AREA RADIATION MONITORING SYSTEM FAILURE annunciator on 1H13

-P845 to annunciate.

Answer: A Explanation:

A is correct because this will come in as specified in training material B is wrong because this will not come in but plausible if an applicant believes turning the monitor off doesn't fully remove power and you need to reset the ARM also.

C is wrong because this is not the correct annunciator or panel but plausible since both are close D is wrong because this is the incorrect annunciator panel Technical

References:

GLP-OPS-D1721, Radiation Monitoring System, Rev 14, p.21 References to be provided to applicants during exam:

None. Learning Objective:

Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(11)

Examination Outline Cross

-Reference Level RO Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Tier # 3 Group # K/A # 2.3.15 Rating 2.9 Question 73 The Area Radiation monitors utilize _________ detectors.

A. Geiger-Mueller tube B. ion chamber C. scintillation D. solid state alpha and beta scintillation Answer: A Explanation:

The main steamline radiation monitoring subsystem uses ion chambers.

The process liquid monitoring subsystem uses scintillation detectors.

Ventilation release rad monitors use solid state alpha detectors and beta scintillation detector s. Technical

References:

GLP-OPS-D1721, p 19 of 68 References to be provided to applicants during exam:

None Learning Objective:

GLP-OPS-D1721 State/identify the types of detectors used for the following radiation monitors: (5) Area Radiation Monitors (5.1)

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level RO Emergency Procedures/Plan Knowledge of EOP entry conditions and immediate action steps.

Tier # 3 Group # K/A # 2.4.1 Rating 4.6 Question 74 Which of the following is an entry condition to EP

-2, RPV Control, and what is the purpose of the immediate action that follows?

A. Suppression pool temperature above 95 oF, maintain a sufficient heat sink B. Drywell pressure above 1.23 psig, ensure reactor is shutdown C. Drywell temperature above 135 oF, protect equipment in the containment D. Containment temperature above 95 oF, maintain containment integrity Answer: B Explanation:

A is wrong because it is not an entry condition for EP

-2 B is correct because it is an entry condition and the immediate action is to ensure the reactor has scrammed C is wrong because it is not an entry condition for EP

-2 D is correct because it is not an entry condition for EP

-2 Technical

References:

EP-2, RPV Control References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental

Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(10) 55.43 Examination Outline Cross

-Reference Level RO Emergency Procedures/Plan Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

Tier # 3 Group # K/A # 2.4.16 Rating 3.5 Question 75 A reactor scram has been automatically initiated due to high drywell pressure.

According to Procedure 02

-S-01-40, EP Technical Bases, Reactor Scram ONEP, 05 02-I-1 should be executed:

A. concurrently with EP

-2. B. prior to entering EP

-2.

C. only after drywell pressure is less than 1.23 psig.

D. only after exiting the emergency procedures.

Answer: A. Explanation:

According to the referenced procedure, ONEP 05-1-02-1-1should be entered and executed concurrently with EP

-2. Technical

References:

02-S-01-40, EP Technical Base s, Revision 004 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level: Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(10) 55.43

Examination Outline Cross

-Reference Level SRO 295021 Loss of Shutdown Cooling Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor water temperature Tier # 1 Group # 1 K/A # 295021 - AA2.04 Rating 3.6 Question 76 The plant is in Mode 4 and flooding up when a spurious Group 3 isolation occurs due to low reactor water level. Thirty minutes has elapsed and maintenance is unable to reset the isolation. Reactor water temperature has risen to 205°F. Primary and secondary containments are intact.

What is the current emergency classification for this event? What is the maximum time to make the initial notification to the NRC?

A. Unusual Event; 15 minutes B. Unusual Event; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Alert; 15 minutes D. Alert; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Answer: B Explanation:

A is wrong because the classification is correct since this is an unplanned event and RCS temperature is greater than 200°F but the time is incorrect.

B is correct because of A. C is wrong because an alert would only be necessary if 200°F had been exceeded for greater than 60 minutes.

D is wrong because of C.

Technical

References:

1 0-S-01-1, Activation of the Emergency Plan, Rev 122, Attachment I, page 7 10-S-01-6, Notification of Offsite Agencies and Plant On

-Call Emergency Personnel, Rev 50, Note under Step 6.1.1.d References to be provided to applicants during exam:

10-S-01-1, Activation of the Emergency Plan, Rev 122 Attachment I (all pages)

Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)- 55.43(b)(5)

Examination Outline Cross

-Reference Level S RO 295026 Suppression Pool High Water Temp.

Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

Temperature Monitoring Tier # 1 Group # 1 K/A # 295026 EA1.03 Rating 3.9 Question 77 The plant is in MODE 2.

RCIC testing is in progress in accordance with SR 3.5.3.3.

SR 3.5.3.3 was successfully performed 93 days ago.

Suppression pool average temperature is 97°F.

No other testing is in progress.

A risk assessment has not been performed LCOs 3.5.3.1 and 3.6.2.1 are provided.

Technical Specifications will allow the plant to enter MODE 1:

A. only after the RCIC testing is completed satisfactorily and suppression pool temperature is lowered to

< 95°F. B. only after suppression pool average temperature is lowered to

< 95°F but the RCIC testing can still be in progress.

C. with current suppression pool temperature only if RCIC testing is still in progress during the MODE change.

D. with current suppression pool temperature only if the RCIC testing is completed satisfactorily.

Answer: C Explanation: LCO 3.6.2.1 Actions A and B require the plant to reduce THERMAL POWER to < 1%, which will require a MODE change from MODE 1 to MODE 2. Thus, 3.0.4.a will not allow MODE 1 entry with the LCO not met.

C conditions meet LCO 3.6.2. LCO 3.5.3 is met because SR 3.5.3.3 is still within the frequency limits of SR 3.0.2.

A is wrong because the SR 3.5.3.3 is still within the frequency limits of SR 3.0.2.

B is wrong. With RCIC testing is in progress, the LCO is met at the current suppression pool temperatur

e. D is wrong because the LCO is only met at the current suppression pool temperature if the testing is still in progress.

Technical

References:

References to be provided to applicants during exam:

LCOs 3.5.3.1 and 3.6.2.1 Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.4 3(b)(2)

Examination Outline Cross

-Reference Level SRO 295027 High Containment Temperature Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY): Containment radiation levels: Mark

-III Tier # 1 Group # 1 K/A # 295027 - EA2.04 Rating 3.7 Question 78 A General Emergency has been declared.

Current conditions are

Reactor water level is

-195" and stable Approximately 15% fuel damage Containment temperature is 180°F All containment radiation monitors are off scale high Large fission product inventory in Containment Radiation levels at the Site Boundary are at 25 mRem/hr TEDE Projected radiation levels at 8 miles are 2 mRem/hr TEDE What are the Protective Action Recommendations?

A. No evacuation of surrounding areas is required.

B. Evacuate a 2

-mile radius and shelter 5 miles downwind.

C. Evacuate a 2

-mile radius and 5 miles downwind; shelter the rest of the 10

-mile EPZ; consider the use of Potassium Iodide.

D. Evacuate a 2-mile radius and 10 miles downwind; shelter the rest of the 10

-mile EPZ; consider the use of Potassium Iodide.

Answer: D Explanation:

A is wrong because all GE requires PARs to be recommended.

B is wrong because sheltering is required for the entire EPZ and not just downwind C is wrong because the dose projection at 8 miles requires an extended PAR D is correct because the conditions presented require an extended PAR Technical

References:

10-S-01-1, Activation of the Emergency Plan, Rev 122, Step 6.1.6k References to be provided to applicants during exam:

10-S-01-1, Activation of the Emergency Plan, Rev 122, Step 6.1.6k only Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)- 55.43(b)(5)

Examination Outline Cross-Reference Level S RO 295028 High Drywell Temperature Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE :

Reactor water level Tier # 1 Group # 1 K/A # 295028 EA2.03 Rating 3.9 Question 79 The crew has entered EP

-2, RPV Control and EP

-3, Containment Control. The following conditions exist:

Drywell temperature 300°F.

RPV pressure 10 psig.

Suppression pool temperature is 190°F.

Containment pressure 2 psig.

Suppression pool 23 ft.

Containment temperature 230°F All RPV level indications are erratic At this point, the SRO should:

A. Enter Emergency Depressurization.

B. Enter EP-5, RPV Flooding.

C. Initiate CTMT sprays.

D. Enter Steam Cooling Answer: B Explanation:

Conditions are in the "possible boiling" area of the RPVST. This + unstable level indication makes RPV level unknown, which requires transition to EP

-5. A is wrong because emergency depressurizations conditions are not met.

C is wrong because the given condition are in the "unsafe to initiate" region of the CSIPL D is wrong because this transition is based on level, which is unavailable.

Technical

References:

EP-2, EP-1, EP-5 References to be provided to applicants during exam Learning Objective:

Question Source: Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO 295030 Low Suppression Pool Water Level Knowledge of EOP mitigation strategies Tier # 1 Group # 1 K/A # 295030 - 2.4.6 Rating 4.7 Question 80 The plant is operating at rated power when an earthquake occurs resulting in a n unisolable suppression pool leak.

Suppression pool level is 17.5 feet and lowering with a ll available methods of making up to the suppression pool in service in accordance with EP-3 , Containment Control

. EP-4 , Auxiliary Building Control is being implemented and sump pumps are maintaining Auxiliary Building Area water levels below their max safe values.

The CRS should enter EP-2 , RPV Control

-

A. and order an Emergency Depressurization using 8 ADS SRVs. B. and order operators to reduce reactor pressure to the reduced band using the bypass valves. C. and order an Emergency Depressurization using available EP

-2, Table 3, Alternate Depressurization Systems

. D. when suppression pool level drops below 14.5 feet and order an Emergency Depressurization using available EP

-2, Table 3, Alternate Depressurization Systems

. Answer: A Explanation:

A is correct because since all methods of suppression pool makeup are in service and level is still dropping you would have to transition to EP-2 and emergency depressurize.

B is wrong because you are unable to maintain suppression pool level SPL

-9 directs you to ED. C is wrong because step ED

-4 in E-2 requires 8 ADS/SRVs to be open to ED. The only allowance to use Table 3 is contained in ED

-5. D is wrong because the procedure step has you transition before reaching 14.5 ft. Understanding that level is dropping and continuing to drop after exhausting all makeup would meet the definition of before in this case. Also use of Table 3 systems is not allowed unless you meet the conditions in ED

-5 Technical

References:

EP-2, RPV Control EP-3, Containment Control 02-S-01-40, EP Technical Bases, Rev 5 References to be provided to applicants during exam:

None.

Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)- 55.43(b)(5)

Examination Outline Cross

-Reference Level S RO 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Containment conditions/isolations Tier # 1 Group # 1 K/A # 295037 EA2.07 Rating 4.2 Question 81 The plant was at rated power when an inadvertent closure of all MSIVs occurred. The following conditions exist:

Multiple rods failed to insert.

Reactor power is 3% and steady.

Suppression pool temperature is 109°F and rising.

RPV level is 40".

The crew is implementing EP

-2A, ATWS RPV control.

At this time, the SRO should

A. exit level and pressure of EP

-2A and enter emergency depressurization.

B. exit EP-2A and enter EP

-3, Containment Control

C. exit EP-2A, flood the CTMT, and enter SAPs.

D. remain in EP

-2A and inject boron into the RPV with both SLC pumps.

Answer: D Explanation: This action is specified in the POWER leg of EP

-2A, ATWS RPV Control.

A would only be done if RPV level were low and cannot be restored.

B In this case, EP

-2A must be continued until the reactor will remain subcritical in all conditions without boron. C would only be done if RPV level were low, cannot be restored, and emergency depressurization has been performed.

D is continuing with the current procedure.

Technical

References:

EP-2A References to be provided to applicants during exam:

None Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO 700000 Generator Voltage and Electric Grid Disturbances Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: VARs outside capability curve Tier # 1 Group # 1 K/A # 700000 - AA2.04 Rating 3.6 Question 82 The plant is operating at 900MWe with a power factor of 0.9 when alarm GEN HOT SPOT MONITOR comes in.

RTD 1N41-TE-N156 reads 130°C RTD 1N41-TE-N158 reads 135° C

What procedure should be entered by the CRS to combat this transient?

A. 03-1-01-2, Power Operations B. 05-1-02-I-4, Loss of AC Power C. 05-1-02-I-2, Turbine and Generator Trips D. 04-1-01-N40-1, Main Generator and Auxiliarie s Answer: C Explanation:

A is wrong because this procedure is not entered during transients.

B is wrong because temperature is not indicative of grid instability by itself so there is no need to enter the procedure. If there were also voltage and frequency oscillations then this answer would be correct.

C is correct because according to the alarm response procedure whenever temperature is above 120°C than a generator trip is warranted.

D is wrong because this procedure should be entered if you're going to reduce MVARs to within the precautions listed in the 04 01-N40-1. This shouldn't be the conclusion since temperature is elevated also.

Technical

References:

04-1-02-1H13-P680, Alarm Response Instruction References to be provided to applicants during exam:

Figure 15 Reactive Capability Curve from GFIG

-OPS-N4151, Main Generator and Excitation System Figures, Rev. 5.

Learning Objective: Document learning objective if possible.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)- 55.43(b)(5)

Examination Outline Cross

-Refer ence Level S RO 295002 Loss of Main Condenser Vac Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required

. Tier # 1 Group # 2 K/A # 295002 2.4.38 Rating 4.4 Questio n 83 The plant was increasing power at 38% RTP when the following alarms occurred:

HP CNDSR SHELL PRESS HI IP CNDSR SHELL PRESS HI LP CNDSR SHELL PRESS HI EOC/RPT A TRIP EOC/RPT B TRIP Reactor power is stable at 25% with turbine bypass valves controlling reactor pressure

. Based on these conditions, the CRS should now:

A. enter EP-2, RPV Control, and if power is reduced to < 4%, declare an Alert.

B. enter EP-2A, ATWS RPV Control, and if power is reduced to < 4%, declare a Site Area Emergency.

C. enter 05-1-02-I-2, Turbine and Generator Trips. If power is reduced to <4%, no E

-Plan declaration is necessary.

D. enter 05-1-02-V-8, Loss of Condenser Vacuum. No E

-Plan declaration is necessary.

Answer: A Explanation:

the reactor should have scrammed because power was > 35.4% RTP when the turbine tripped. This is described by SA3 in the Classification flowcharts. EP

-2 should be implemented because a reactor scram is required and all control rods are not at position 00.

Technical

References:

EN-OP-115, Conduct of Operations.

References to be provided to applicants during exam:

Emergency Classification Flowcharts.

Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO 295011 High Containment Temp Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY): Containment humidity: Mark

-III Tier # 1 Group # 2 K/A # 295011 - AA2.03 Rating 3.2 Question 84 The plant is operating at full power when containment exhaust filter moisture high alarm is received.

Containment temperature is 90°F.

Suppression Pool temperature is 90°F Radiation levels are normal for the power level Containment sump runtimes are normal The CRS should enter:

A. EP-3, Containment Control

B. EP-2, RPV Control C. 04-S-01-M41-1, Containment Cooling System D. 04-S-01-P64-1, Fire Protection Water System Answer: C Explanation:

A is wrong because containment temperature and radiation levels would be elevated if a steam leak were occurring.

B is wrong because radiation levels would be higher and sump runtimes would be increasing to deal with the higher leakage C is correct because the temperature are slightly elevated but not above any EOP entry conditions.

D is wrong because fire water is relatively cool water it would probably wouldn't add to the amount of vapor inside containment.

Technical

References:

04-1-02-1H13-P842, Alarm Response Instruction, Rev 22 References to be provided to applicants during exam:

N one. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)- 55.43(b)(5)

Examination Outline Cross

-Reference Level S RO 295014 Inadvertent Reactivity Addition Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Tier # 1 Group # 2 K/A # 295014 2.2.42 Rating 4.6 Question 85 The plant is performing a plant startup following a refueling outage. At 1230, with the plant at 20% power, the reactor operator reports that a review of plant indications show that one hour previously, Jet pumps 5 and 6 indicated a step drop in flow.

Total core flow indicated a step drop.

Recirculation loop drive flow indicated a step increase.

Reactor power dropped slightly.

At this time, the SRO should: (LCO 3.4.3 is provided)

A. declare LCO 3.4.3 action condition A entry at 1230.

B. declare LCO 3.4.3 action condition A entry at 1130.

C. consider LCO 3.4.3 met because SR 3.4.3.1 does not have to be performed at this power level.

D. consider LCO 3.4.3 met because the established patterns of jet pump flow have not been established following the refueling.

Answer: A Explanation:

These are symptoms of a rams head ejection. The definition of OPERABLE is not met. Two jet pumps are not capable of their specified safety related function.

B is wrong because action conditions apply from the time of discovery, not the time of occurrence

. C is wrong because the SR has to be met, even though it does not have to be performed.

D is wrong because the TS basis says "During the initial weeks of operation under such conditions, while baselining new "established patterns," engineering judgment of the daily surveillance results is used to detect significant abnormalities which could indicate a jet pump failure." Technical

References:

Technical Specifications 3.4.3, 3.4.3 bases, SR 3.0.1.

References to be provided to applicants during exam: LCO 3.4.3.

Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO 203000 RHR/LPCI: Injection Mode Knowledge of EOP entry conditions and immediate action steps Tier # 2 Group # 1 K/A # 203000 - 2.4.1 Rating 4.8 Question 86 The plant has scrammed from full power due to high drywell pressure.

1 control rod is not indicating full in Drywell Pressure is 4 psig Drywell Temperature is 150 F RPV pressure is 700 psig The CRS should-A. enter EP-5, RPV Flooding.

B. remain in EP

-2, RPV Control.

C. enter EP-3, Containment Control.

D. enter EP-2A, ATWS RPV Control.

Answer: A Explanation: A is correct because all RPV level instruments, with exception of Fuel Zone, are only calibrated to 135oF. Based on this RPV level would not be able to be determined and the second override in EP

-2 would need to be enforced. B and C are plausible if the applicant misses the override condition.

D is wrong because you wouldn't need to enter EP

-2A since the reactor would be shutdown under all conditions without boron with one rod either stuck out or not indicating full in.

Technical

References:

GLP-OPS-B2101, RPV Level Instrumentation, Rev 7 References to be provided to applicants during exam:

None. Learning Objective: State/identify the calibration conditions for each of the following level instruments: (9.0) o Reactor Water Level Narrow Range (9.1) o Reactor Water Level Wide Range (9.2) o Reactor Water Level Shutdown Range (9.3) o Reactor Water Level Upset Range (9.4) o Reactor Water Level Fuel Zone Range (9.5)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank

1. New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)- 55.43(b)(5)

Examination Outline Cross

-Reference Level S RO 212000 RPS Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: : RPS bus power supply failure Tier # 2 Group # 1 K/A # 212000 A2.02 Rating 3.9 Question 87 The plant is at rated power when P680 annunciator HCU TROUBLE is received. When the operator depresses the HCU FAULT pushbutton on the Operator's Control Module, the following control rods have blinking red LEDs on the RC&IS display; 04-21 04-45 28-05 28-61 The CRS also notes the Division I MSIV lights on panels 1H13

-P622 and P623 are dark.

The SRO should direct the crew to:

A. Re-charge the affected HCU's in accordance with SOI 04 01-C11-1 Control Rod Drive Hydraulic System

. B. Transfer the 'A' RPS bus to the alternate power source per ONEP 05 02-III-2, Loss of One or Both RPS buses.

C. Transfer the 'B' RPS bus to the alternate power source per ONEP 05 02-I II-2, Loss of One or Both RPS buses.

D. Dispatch an operator to the affected HCU's inside containment to determine the fault in accordance with ARI 04-1-02-1 H 13-P680-4A2-D4, HCU TROUBLE.

Answer: B Explanation:

These are symptoms of loss of RPS bus 'A".

Technical

References:

ONEP 05-1-02-III-2, Loss of One or Both RPS buses References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO 215003 IRM Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic operation of detectors/system Tier # 2 Group # 1 K/A # 215003 - A2.05 Rating 3.5 Question 88 A plant startup is in progress with the reactor Mode Switch in STARTUP.

IRMs and APRMs have been verified to have proper overlap, all APRMs are off the downscale light, and all Mode 1 prerequisites have been met.

Inverter 1Y87 and its alternate power supply fail.

(1) Regarding TS 3.3.1.1, what is the required action?

(2) When is the Mode Switch allowed to be placed in RUN?

A. (1) Place the affected IRM channels in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(2) When directed by the Shift Manager in accordance with 03 01-1, Cold Shutdown to Generator Carrying Minimum Load.

B. (1) Be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

. (2) Only after either Inverter 1Y87 or its alternate power supply are returned to OPERABLE status.

C. (1) Place the affected IRM channels in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(2) Only after either Inverter 1Y87 or its alternate power supply are returned to OPERABLE status.

D. (1) Be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(2) When directed by the Shift Manager in accordance with 03 01-1, Cold Shutdown to Generator Carrying Minimum Load.

Answer: A Explanation:

A is correct because this power supply failure would only require entry into Condition A of TS 3.1.1. The action would already be taken so there would be 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to fix the inverter. The second part of the question tests application of TS 3.0.4. Since IRMs are not required in Mode 1 this would meet TS 3.0.4.a B is wrong because TS 3.3.1.1.H requires this only after you fail to meet the time requirements TS 3.3.1.1.A. Even though the action has been taken for A the time allowance still remains at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> before you need to enter condition H.

C is wrong because you would not need to wait until the inverter is restored.

D is wrong because of the TS discussion in B.

Technical

References:

TS 3.3.1.1 References to be provided to applicants during exam:

TS 3.3.1.1.

Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)- 55.43(b)(2)

Examination Outline Cross

-Reference Level S RO 259002 Reactor Water Level Control Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: RFP runout condition: Plant

-Specific Tier # 2 Group # 1 K/A # 259002 A2.04 Rating 3.1 Question 89 The plant is at rated thermal power when the following alarms are received:

RFPT A ACT THR BRG WR HI RFPT A TRIP Shortly thereafter, RX LVL 40"/32" HI/LO is received.

The plant stabilizes with the following conditions:

Reactor power

- 80% RFPT B speed

- 5900 rpm The procedure that contains the actions to mitigate these conditions is::

A. ARI 04-1-02-1H13-P680-3A-A3, RX LVL 40"/32" HI/LO, to mitigate the low reactor level.

B. ONEP 05-1-02-III-3, Reduction in Recirculation System Flow Rate, to mitigate the failure of the recirculation runback.

C. ONEP 05-1-02-V-7, Feedwater System Malfunctions, to mitigate RFPT B runout.

D. ARI 04-1-02-1H13-P680-2A-A2, RFPT A TRIP, to mitigate the low reactor level.

Answer: C Explanation:

RFPT B is at runout conditions and RX level is low. The mitigating actions are to lower reactor power to below 75% and ensure that RFPT speed is reduced to

< 5850 rpm. Of these, only Feedwater System Malfunctions contains these instructions.

Technical

References:

ARI 04-1-02-1H13-P680-3A-A3, RX LVL 40"/32" HI/LO ONEP 05-1-02-III-3, Reduction in Recirculation System Flow Rate ONEP 05-1-02-V-7, Feedwater System Malfunctions ARI 04-1-02-1H13-P680-2A-A2, RFPT A TRIP

References to be provided to applicants during exam:

None Learning Objective: GLP-OPS-N2100 Describe the interrelationship(s) between the Feedwater System and the following: (28.0) Nuclear Boiler System, B21 (28.10)

Concerning SOI 04 01-N21-1: (29.0) Identify precautions, limitations, cautions, warnings and notes that apply to a given situation. (29.1)

List/identify the immediate operator actions for the following Off Normal Event Procedures: (31.0) Feedwater System Malfunctions ONEP, 05 02-V-7 (31.2) Given a copy of the above listed ONEPs and plant conditions, discuss the subsequent actions to be taken with regard to the Feedwater System. (32.0

) Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO 261000 SGTS Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations: A.C. electrical failure Tier # 2 Group # 1 K/A # 261000 - A2.07 Rating 2.7 Question 90 A seismic event has resulted in the following:

Small break LOCA in the drywell Auto initiation of RCIC and HPCS Loss of offsite power with all diesels supplying power to their respective buses EP-2 has been entered

Drywell pressure is 2.0 psig and stable Drywell temperature is 120°F and stable Reactor water level is stable in the normal band Fuel handling exhaust radiation at 4.2 mr/hr and stable Suppression pool temperature at 90°F and stable RCIC equipment area temperature at 150°F and stable At this point operators are required to transition to -

A. EP-3, Containment Control, and operate all available suppression pool cooling with RHR pumps not required for adequate core cooling.

B. EP-3, Containment Control, and defeat drywell cooling isolation interlocks using Attachment 10 of EP

-1. C. EP-4, Auxiliary Building Control, and verify standby gas treatment system is in service and all automatic isolations are complete.

D. EP-4, Auxiliary Building Control, and operate RCIC room coolers to lower temperature in the RCIC room.

Answer: C Explanation:

A is wrong because suppression pool temperature is higher than normal but does not exceed the 95°F in EP

-3. B is wrong because drywell temperature has not exceeded the temperature in EP

-3. C is correct because the setpoint has been exceeded in EP

-4 D is wrong because the RCIC room temperature has not exceeded the temperature in EP

-5 Technical

References:

EP-4, Auxiliary Building Control References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)- 55.43(b)(5)

Examination Outline Cross

-Reference Level S RO 239001 Main and Reheat Steam Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Tier # 2 Group # 2 K/A # 239001 2.4.

9 Rating 4.2 Question 91 After being at rated power for 11 continuous months, the crew is performing a cooldown per 03 01-3B, Plant Shutdown to comply with LCO 3.5.1, ECCS

- Operating.

Required action condition D was entered 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ago due to failure of the B RHR motor.

Reactor water cleanup is isolated and not available.

Shutdown cooling has been placed in service using A RHR and is supplementing the cooldown from operating the turbine bypass valves

. Reactor pressure is 110 psig.

Then, the A RHR pump trips and cannot be restarted.

If no repairs are made, it will eventually be necessary to declare a(n) ___________ due to

__________.

(1) (2) A. NOUE B. NOUE. C. Alert. D. Alert Inability to reach operating mode within TS limits Unplanned loss of decay heat removal capability with irradiated fuel in the RPV.

Inability to reach operating mode within TS limits Unplanned loss of decay heat removal capability with irradiated fuel in the RPV.

Answer: A Explanation:

ONEP 05-1-02-III-1, Inadequate Decay Heat Removal says (section 3.4) to place redundant loop in shutdown cooling (not available) or place ADHRS in service (cannot be done above 200°F). It will not be possible to place the plant in MODE 4 in this condition. Decay heat removal capability has not been lost because the turbine bypass system is still available.

Technical

References:

ONEP 05-1-02-III-1 Emergency Classification Flowcharts

References to be provided to applicants during exam:

Emergency Classification Flowcharts and TS 3.5.1. Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO 241000 Reactor/Turbine Pressure Regulator Ability to interpret reference materials, such as graphs, curves, tables, etc.

Tier # 2 Group # 2 K/A # 241000 - 2.1.25 Rating 4.2 Question 92 The plant is operating at full power when alarm TURB IPC CAB FAIL is received. Turbine control and bypass valves are operating normally for this power level. I&C personnel are sent to investigate and determine the internal power supply for IPC cabinet 1H13

-P825-JC03 needs to be replaced.

Per EN-WM-100, "Work Request (WR) Generation, Screening and Classification," what priority should be assigned to the work request A. Priority 1

B. Priority 2 C. Priority 3 D. Priority 4 Answer: B Explanation:

Since the power supply has failed it falls into corrective maintenance space. Since the component presents a risk to generation but the system is still operable, because turbine and bypass valves controls are less redundant, this work request should be screened as a priority 2. All other answers are plausible depending on how the applicant judges the risk of additional failures.

Technical

References:

Per EN-WM-100, "Work Request (WR) Generation, Screening and Classification,"

Rev 9 References to be provided to applicants during exam:

Per EN-WM-100, "Work Request (WR) Generation, Screening and Classification,"

p. 1-13 and Attachment 9.1 Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41(b)- 55.43(b)(5)

Examination Outline Cross

-Reference Level S RO 259001 Reactor Feedwater Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of extraction steam Tier # 2 Group # 2 K/A # 259001 A2.04 Rating 3.4 Question 93 The plant is at rated power when P680 alarm FW HTR 6B LVL HI is received. Motor

-operated valve 1N36

-F011B, FW HTR 6B EXT STM ISOL, is stroking closed.

The FIRST thing the SRO should direct the crew to do is:

A. re-open 1N36-F011B in accordance with ARI 04 02-1H13-P680-2A-A10, FW HTR 6B LVL HI. B. initiate a reactor scram and enter ONEP 05-1-02-1-1, Reactor Scram.

C. reduce core flow at the maximum rate in accordance with ONEP 05 02-V-5, Loss of Feedwater Heating

. D. reduce core flow using the slow detent of reactor recirculation flow control in accordance with ONEP 05 02-V-5, Loss of Feedwater Heating.

Answer: C Explanation:

FW HTR 6B is a high pressure heater, so recirculation flow should be reduced rapidly per the note prior to step 2 of ONOP Loss of Feedwater Heating.

The note also says that the action should be taken when the isolation valve begins to close.

Control rods would only be inserted if the reduction in recirc flow resulted in being above the MEOD boundary line.

It may be possible to avoid a scram on this occurrence, so the procedures direct a power reduction.

Technical

References:

ONEP 05-1-02-V-5, Loss of Feedwater Heating References to be provided to applicants during exam:

Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO Conduct of Operations Knowledge of refueling administrative requirements Tier # 3 Group # K/A # 2.1.40 Rating 3.9 Question 94 According to procedure 03 01-5, Refueling, "Loads in excess of _______ pounds shall be prohibited from travel over fuel assemblies in the Spent Fuel OR Upper Containment Fuel Storage Pool Racks.

" A. 975 B. 1060 C. 1140 D. 1220 Answer: C Explanation:

Precaution and limitation 2.8 of 03 01-5, Refueling, Rev 128 defines the maximum load allowed over fuel assemblies in the spent fuel pool.

"2.8 Loads in excess of 1140 pounds Shall be prohibited from travel over Fuel assemblies in the Spent Fuel OR Upper Containment Fuel Storage Pool Racks.

" Technical

References:

03-1-01-5, Refueling, Rev 128 References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)- 55.43(b)(7)

Examination Outline Cross

-Reference Level S RO Conduct of Operations Knowledge of RO duties in the control room during fuel handling such as responding to

alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Tier # 3 Group # K/A # 2.1.44 Rating 3.8 Question 95 The plant is in MODE 5, applying the provisions of LCO 3.10.5, Single CRD Removal

- Refueling, to remove one CRD mechanism in a cell containing three fuel assemblies. CRD mechanism removal is in progress.

Then, alarm CONT ROD WITHDRAWAL BLOCK is received. The RO determines that the cause of the alarm is refueling platform near or over core and grapple loaded.

Technical specifications require the operators to immediately:

1. direct the maintenance crew to suspend the removal of the CRD mechanism.

2. insert a control rod withdrawal block.

3. direct the refueling platform operator to stop core alterations.

A. 1 only B. 1 and 2 only C. 2 and 3 only D. 1, 2, and 3.

Answer: D Explanation:

In this mode, the alarm signifies that the LCO was not met when the CRD removal began. The control rod withdrawal block was not inserted. 1, 2, and 3 are all immediate actions that are required per TS 3.10.5. While a rod block is now present, action 3 will make the rod block clear, so it will be necessary to insert one.

Technical

References:

ARI 04-1-02-1H13-P680-4A2-C5, CONT ROD WITHDRAWAL BLOCK.

LCO 3.10.5 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO Equipment Contr ol Knowledge of the process for making changes to procedures Tier # 3 Group # K/A # 2.2.5 Rating 3.6 Question 96 Which of the following situations would be appropriate for issuing a Temporary Change Notice: 1. Equipment is out of service and a change is required to complete the work and return the equipment to service.

2. A change is required to implement a surveillance that is scheduled to be performed in 4 weeks. 3. A directive change is required to support an operability review.

A. 1 and 2 ONLY B. 1 and 3 ONLY C. 2 and 3 ONLY D. 1, 2, and 3 Answer: B Explanation:

From 01-S-02-3, Author's Guide, Rev 121, Note 6.4.1, "directive change is required to implement a surveillance and the work is scheduled in the near future (i.e., two weeks)".

Since the time frame is 4 weeks in (2) then it should be issued as a direct revision instead of a TCN. This makes any answer containing (2) incorrect.

Technical

References:

01-S-02-3, Author's Guide, Rev 121, Note 6.4.1 References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)- 55.43 Examination Outline Cross

-Reference Level S RO Equipment Control Knowledge of the process for controlling temporary design changes

. Tier # 3 Gr oup # K/A # 2.2.11 Rating 3.3 Question 97 When an approved procedure installs electrical jumpers for testing purposes, Temporary Modification Tags should be applied whenever the jumpers:

A. must be independently verified.

B. are out of line of sigh

t. C. are NOT documented in the Asset Suite (AS).

D. will be in place for greater than one hour.

Answer: B Explanation:

P&L 7 of EN

-DC-136, Temporary Modifications, contains this requirement.

A is wrong because this process (testing) is excluded from T-MOD requirements when restoration is documented and verified through independent verification and the jumper remains in line of sight.

C is wrong because testing jumpers are not documented in the Asset Suite.

D is wrong because there are no time requirements. The jumpers do not need a T

-mod tag unless they are left unattended.

Technical

References:

EN-DC-136, Temporary Modifications References to be provided to applicants during exam:

Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO Radiation Control Ability to control radiation releases.

Tier # 3 Group # K/A # 2.3.11 Rating 4.3 Question 98 (1) What is the governing procedure to process liquid radioactive waste discharges?

(2) What is the governing procedure to process solid radioactive waste discharges?

A. (1) 01-S-08-11, Radioactive Discharge Controls (2) EN-RP-121, Radioactive Material Control

B. (1) EN-RP-121, Radioactive Material Control (2) 01-S-08-11, Radioactive Discharge Contro ls C. (1) 01-S-08-11, Radioactive Discharge Controls (2) 01-S-08-11, Radioactive Discharge Controls D. (1) EN-RP-121, Radioactive Material Control (2) EN-RP-121, Radioactive Material Control Answer: A Explanation:

A is correct because 01-S-08-11, control liquid and gaseous releases. Step 6.2.1 of 01

-S-08-11 refers to EN

-RP-121 for processing solid waste.

B is wrong of A. C is wrong of A.

D is wrong of A.

Technical

References:

01-S-08-11, Radioactive Discharge Controls EN-RP-121, Radioactive Material Control References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(13) 55.43(b)(5)

Examination Outline Cross

-Reference Level S RO Emergency Procedures / Plan Ability to recognize abnormal indications for system operating parameters that are entry

-level conditions for emergency and abnormal operating procedures.

Tier # 3 Group # K/A # 2.4.4 Rating 4.7 Question 99 The plant was at rated power when all four channels of MSL PIPE TNL TEMP HI alarms are received: Concurrently, A main steam line isolation occurs.

A reactor scram occurs with all control rods fully inserted.

Shortly thereafter, alarm RX LVL 3 (+10")

LO is rec eived. SP temperature is 89°F and rising. SP Level is 18.5 ft. and rising.

Then, the CRS should enter:

1. EP-2, RPV Control

2. EP-3, Containment Control

3. EP-4, Auxiliary Building Control A. 1 only B. 1 and 2 only

C. 1 and 3 only D. 1, 2, and 3 Answer: C Explanation:

At least one entry conditions is met for EP

-2 (RV level < 11.4 In).

At least one entry condition is met for EP

-4 (MSL Pipe Tunnel Temp > 185°F)

No entry conditions for EP

-3 are met.

Technical

References:

EP-2, 3, and 4

References to be provided to applicants during exam:

Learning Objective:

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content:

55.41(b)(7) 55.43 Examination Outline Cross

-Reference Level SRO Emergency Procedures/Plan Knowledge of fire protection procedures Tier # 3 Group # K/A # 2.4.46 Rating 4.2 Question 100 A fire is reported to the control room at the Independent Spent Fuel Storage Facility (ISFSI) at 12:00 pm. How many fire brigade members are required to be sent?

The fire brigade leader reports back at 12:25 pm that the fire is not extinguished. Which Emergency Action Level has been exceeded?

A. 3 Unusual Event B. 3 Alert C. 5 Unusual Event D. 5 Alert Answer: C Explanation:

A is wrong because the 5 man team is responsible for responding to the ISFSI. A UE is required by the emergency plan B is wrong because a 5 man team is required and an Alert would not be C is correct because a 5 man team is required by procedure and an unusual event would be required because the ISFSI is not safety related and there has not been a report of an damage D is wrong because the classification is incorrect Technical

References:

10-S-03-2, Response to Fires, Rev 25. Step 6.2.3.f 10-S-01-1, Activation of the Emergency Plan, Rev. 122 References to be provided to applicants during exam:

None. Learning Objective: Document learning objective if possible.

Question Source:

Bank # (note changes; attach parent)

Modified Bank #

New X Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)- 55.43(b)(5)