ML14350A131

Initial Exam 2014-301 Draft SRO Written Exam

ML14350A131
Person / Time
Site:	Brunswick
Issue date:	12/16/2014
From:	NRC/RGN-II
To:	Progress Energy Carolinas
Shared Package
ML14350A180	List: ML14350A124 ML14350A125 ML14350A131 ML14350A134 ML14350A137 ML14350A155 ML14350A158 ML14350A161 ML14350A168 ML14350A173
References
50-324/14-301, 50-325/14-301 50-324/OL-14, 50-325/OL-14
Download: ML14350A131 (78)

Text

76. S201006 1 Which one of the following completes the statements below concerning the Rod Worth Minimizer (RWM)?

The RWM channel functional test was NOT required to be performed (1) any control rod was withdrawn at <8.75% RTP in MODE 2 IAW Surveillance Requirement 3.3.2.1.2.

Currently total steam flow is (2) %.

A. (1) until (2) 18 B. (1) until (2) 20 C. (1) until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after (2) 18 D. (1) until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after (2) 20 Answer: D K/A:

201006 ROD WORTH MINIMIZER SYSTEM (RWM)

G2.04.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 / 43.5 / 45.12)

RO/SRO Rating: 4.2/4.2 Pedigree: New

Objective:

LOI-CLS-LP-07.1, Obj. 3 - Describe the operation of the RWM above and below the Low Power Setpoint (LPSP) and the Low Power Alarm Point (LPAP), including the setpoints and where the input signal originates.

LOI-CLS-LP-07.1, Ob. 10- Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM and COLR, determine the required action(s) to be taken in accordance with Technical Specifications, the TRM, or ODCM associated with the RWM System. (SRO/STA only) (LOCT)

Reference:

None Cog Level: Fundamental Knowledge Explanation:

The RWM uses total steam flow to determine reactor power. 19.1% is the LPSP and 27.8% is the LPAP.

The area between these two setpoint is the transistion zone where the RWM will alarm but not enforce the blocks. A note in the surveillance requirements allows 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn in MODE 2 to perform the SR.

Distractor Analysis:

Choice A: Plausible because when a control is withdrawn would be correct if it was not for the note in the SR. Since the Transition zone is indicated by the picture steam flow is greater than 19.1%.

Choice B: Plausible because when a control is withdrawn would be correct if it was not for the note in the SR. Part two is correct.

Choice C: Plausible because part one is correct and since the Transition zone is indicated by the picture steam flow is greater than 19.1%

Choice D: Correct Answer, see explanation SRO Basis:

Facility operating limitations in the technical specifications and their bases. [10 CFR 55.43(b)(2)]

77. S203000 1 Following a large line break DBA LOCA, plant conditions are:

Reactor water level

-50 inches (N036/N037)

Reactor pressure 5 psig Core Spray One loop available, injecting at 4800 gpm RHR One loop available, injecting at 17,000 gpm Suppression pool temp.

148°F The CRS has reached step RC/L-44 in RVCP:

Which one of the following identifies the basis for reduction in RPV injection when reactor water level is below Minimum Steam Cooling Reactor Water Level?

A. Prevent exceeding the Heat Capacity Temperature Limit.

B. Maintain long term operation of the Core Spray and RHR Pumps.

C. Minimize off-site releases per Alternative Source Term calculations.

D. Prevent exceeding design temperature limits for Primary Containment.

Answer: B K/A:

203000 RHR/LPCI: INJECTION MODE A2 Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) 16 Loss of coolant accident RO/SRO Rating: 4.4/4.5 Pedigree: 2004 Audit Exam Objective:

LOI-CLS-LP-017, Obj. 28 - Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event: d. Loss of coolant accident (LOCT)

Reference:

None Cog Level: Fundamental Knowledge Explanation:

The calculation for the NPSH to the Core Spray and RHR pumps specify establishing cooling at ten minutes into the design basis LOCA. The calculation also assumes that the temperature of the suppression pool will be at approximately 169°F at ten minutes. If containment cooling is not established, then it is possible that the Core Spray or RHR pumps will be lost due to inadequate NPSH.

Distractor Analysis:

Choice A: Plausible because the heat capacity temperature limit would be of concern if the reactor was pressurized to make sure the suppression pool could handle the heat loading.

Choice B: Correct Answer, see explanation Choice C: Plausible because RVCP has actions within it for alternate source term analysis dealing with the suppression pool.

Choice D: Plausible because while these values are high but do not exceed the design limit of 220°F in the suppression pool.

SRO Basis:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)]

This requires assessing plant conditions and based on a decision point the usage of a table and the knowledge of the EOP basis for establishing long term cooling for containment.

78. S215003 1 Unit One is performing a reactor startup, prior to the point of adding heat.

IRM C is bypassed due to erratic operation.

IRM A fails downscale.

Which one of the following completes the statements below?

Addressing ONLY Technical Specification 3.3.1.1, Reactor Protection System (RPS)

Instrumentation, requires placing the channel in trip in (1) hours.

IAW AD-OP-ALL-0101, Event Response and Notifications, the plant manager will be directly notified of this event by the (2).

(Reference provided)

A. (1) 6 (2) Shift Manager B. (1) 6 (2) Site Duty Manager C. (1) 12 (2) Shift Manager D. (1) 12 (2) Site Duty Manager Answer: D K/A:

215003 INTERMEDIATE RANGE MONITOR (IRM) SYSTEM G2.04.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11)

RO/SRO Rating: 2.7/4.1 Pedigree: new Objective:

LOI-CLS-LP-009-A, Obj. 13 - Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM and COLR, determine the required action(s) to be taken in accordance with Technical Specifications, TRM and COLR associated with the Source Range and Intermediate Range Monitoring Systems.

(SRO/STA only)

Reference:

TS 3.3.1.1 Cog Level: High

Explanation:

TS table 3.3.1.1-1requires 3 operable IRM channels per trip system. With IRM A & C inop, 3 channels are not operable for one trip system. IRM A and C are in the same trip system. Condition A must be entered.

The question does NOT address TRM.

IAW with the procedure the SM notifies the Site Duty Manager who in turn makes all other notifications.

Distractor Analysis:

Choice A: Plausible because TS Condition B would be applicable if one or more required trip systems were inoperable and would require placing in trip condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the SM does notify the Site Duty Manager but not the PM.

Choice B: Plausible because TS Condition B would be applicable if one or more required trip systems were inoperable and would require placing in trip condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the Site Duty Manager does notify the PM.

Choice C: Plausible because TS condition A is required to be entered, placing the channel in trip in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the SM does notify the Site Duty Manager but not the PM.

Choice D: Correct Answer, see explanation SRO Basis:

Facility operating limitations in the technical specifications and their bases. [10 CFR 55.43(b)(2)]

79. S215005 1 The following Unit One information is obtained in preparation for filling out the Emergency Notification Form for a low power ATWS:

1204 ONLY one rod out 1205 All rods in SLC was NOT injected ERFIS Met Data:

Upper Wind Direction 15.00 Deg Lower Wind Direction 13.00 Deg The Control Room Site Emergency Coordinator is evaluating the following lines:

Which one of the following completes the statements below IAW 0PEP-02.6.21, Emergency Communicator?

On line 9, (1) degrees should be entered for "Wind Direction from".

On Line 12, the earliest time that can be entered for Unit One "Shutdown at Time" is (2) hours.

A. (1) 13 (2) 1204 B. (1) 13 (2) 1205 C. (1) 15 (2) 1204 D. (1) 15 (2) 1205 Answer: A K/A:

215003 AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM G2.04.30 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required. (CFR: 41.10 / 43.5 / 45.11)

RO/SRO Rating: 2.4/4.4 Pedigree: New Objective:

Task - SRO Only - Complete And Approve Emergency Notification Forms For Initial And Follow-up Notifications To State/County Agencies In Accordance With 0PEP-02.6.21.

Reference:

None Cog Level: High Explanation:

Recent OE has had operators not using the one rod out difinition that is allowed by the procedure for determining that the reactor is shutdown. One rod out meets the definition for Shutdown Margin and All Rods In also meets this definition. The ENF form uses the lower wind speed from the ERFIS Met Data screen in the control room.

Distractor Analysis:

Choice A: Correct Answer, see explanation.

Choice B: Plausible because the lower wind directions is used but all rods in is not the earliest time that the reactor can remain shutdown under all conditions without boron.

Choice C: Plausible because 15 degrees is the upper wind direction not the lower wind direction and 1204 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.58122e-4 months <br /> is correct.

Choice D: Plausible because 15 degrees is the upper wind direction not the lower wind direction and all rods in is not the earliest time that the reactor can remain shutdown under all conditions without boron.

SRO Basis:

This task is an SRO Only task.

344020B502 Complete And Approve Emergency Notification Forms For Initial And Follow-up Notifications To State/County Agencies In Accordance With 0PEP-02.6.21.

EOP User's Guide Definitions:

80. S261000 1 Unit One is operating at rated power.

On 10/13/14 at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> it is discovered that this SR was last performed on 09/01/14 at 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />.

Which one of the following is correct IAW Technical Specifications?

A. The time between surveillances is acceptable IAW SR 3.0.2.

B. The time between surveillances is acceptable IAW SR 3.6.4.3.1.

C. The time between surveillances is unacceptable. Entry into the applicable Conditions and Required Actions for the missed surveillance is immediately required.

D. The time between surveillances is unacceptable. Entry into the applicable Conditions and Required Actions may be delayed for up to 31 days to permit performance of the surveillance.

Answer: D K/A:

261000 STANDBY GAS TREATMENT SYSTEM G2.02.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)

RO/SRO Rating: 3.6/4.5 Pedigree: New Objective:

LOI-CLS-LP-200-B, Ob 9 - Given plant conditions, apply the rules of Section 3.0 to determine appropriate actions in accordance with Technical Specifications. (SRO/STA Only)(LOCT)

Reference:

None Cog Level: hi Explanation:

SR has not been performed for 42 days.

38.75 max per 25% criteria of SR 3.0.2.

3.0.3 allows delay of up to 31 days from discovery of missed surv.

Distractor Analysis:

Choice A: Plausible because SR 3.0.2 allows an additional 25%.

Choice B: Plausible because if the student thinks that the surv. is monthly.

Choice C: Plausible because the surv. time is unacceptable and if SR 3.0.3 is not applied this is correct.

Choice D: Correct Answer, see explanation SRO Basis:

Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).

81. S290001 1 Unit Two was operating at rated power when the following alarms/conditions are reported:

UA-03 (2-7) Area Rad Rx Bldg High UA-03 (4-5) Process Rx Bldg Vent Rad Hi-Hi UA-05 (6-10) Rx Bldg Isolated UA-05 (4-6) SBGT Sys A Failure A-02 (5-7) Stm Leak Det Ambient Temp High UA-05 (6-7) Rx Bldg Static Press Dif-Low A-01 (3-5) HPCI Isol Trip sig A Initiated A-01 (4-5) HPCI Isol Trip sig B Initiated Steam is visible exiting the Reactor Building blowout panels The following indications are observed:

Which one of the following completes the statements below?

The release through the Reactor Building blowout panels is considered (1) release.

The highest Emergency Action Level classification for the given conditions is (2).

(Reference provided)

A. (1) a ground (2) an Alert B. (1) a ground (2) a Site Area Emergency C. (1) an elevated (2) an Alert D. (1) an elevated (2) a Site Area Emergency Answer: B

K/A:

290001 SECONDARY CONTAINMENT G2.04.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 /

43.5 / 45.3 / 45.12)

RO/SRO Rating: 4.1/4.3 Pedigree: New Objective:

LOI-CLS-LP-301-B, Obj. 9 - Given a hypothetical abnormal event and plant operating mode, use 0PEP-02.1 to properly classify or re-classify the event.

Reference:

0PEP-02.1 Cog Level: High Explanation:

Even though the release is out the 117 foot elevation this is considered a ground release. The ambient Temp High alarm indicates that a Group 4 isolation should have occurred With the inability to isolate the leak the EAL call is a loss of RCS and a loss of Containment which makes a SAE.

Distractor Analysis:

Choice A: Plausible because a ground release is correct and if only one barrier is assessed then this is correct.

Choice B: Correct Answer, see explanation Choice C: Plausible because even though the release is out the 117 foot elevation this is considered a ground release and if only one barrier is assessed then this is correct.

Choice D: Plausible because even though the release is out the 117 foot elevation this is considered a ground release and a SAE is correct.

SRO Basis:

Fuel handling facilities and procedures. [10 CFR 55.43(b)(7)]

82. S290003 1 Which one of the following identifies:

(1) the conditions which will cause the control room ventilation system to automatically align in the Fire Protection mode and (2) the required Technical Specifications (TS) / Technical Requirements Manual (TRM) actions IAW 0OP-37, Control Building Ventilation System Operating Procedure, if the system was initiated for 20 minutes with minimum local smoke and with charcoal exposure doubtful?

A. (1) Smoke detected in the Unit One Electronic Equipment room or the manual pull station tripped in the Unit Two Electronic Equipment Room.

(2) Initiate an LCO on the affected train IAW TS 3.7.3, CREV System, and TRM 3.18, CREV System-Smoke Protection Mode, ONLY.

B. (1) Smoke detected in the Unit One Electronic Equipment room or the manual pull station tripped in the Unit Two Electronic Equipment Room.

(2) Initiate an LCO on the affected train IAW TS 3.7.3, CREV System, and TRM 3.18, CREV System-Smoke Protection Mode, and take the action specified in TS 5.5.7, Ventilation Filter Testing Program (VFTP).

C. (1) Smoke detected in the Unit One Electronic Equipment room and the manual pull station tripped in the Unit Two Electronic Equipment Room.

(2) Initiate an LCO on the affected train IAW TS 3.7.3, CREV System, and TRM 3.18, CREV System-Smoke Protection Mode, ONLY.

D. (1) Smoke detected in the Unit One Electronic Equipment room and the manual pull station tripped in the Unit Two Electronic Equipment Room.

(2) Initiate an LCO on the affected train IAW TS 3.7.3, CREV System, and TRM 3.18, CREV System-Smoke Protection Mode, and take the action specified in TS 5.5.7, Ventilation Filter Testing Program (VFTP).

Answer: C K/A:

290003 CONTROL ROOM HVAC A2 Ability to (a) predict the impacts of the following on the CONTROL ROOM HVAC ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) 04 Initiation/failure of fire protection system RO/SRO Rating: 3.1/3.3 Pedigree: New Objective:

LOI-CLS-LP-037, Obj. 10 - Given plant conditions associated with the CB HVAC and EAF System determine the required action(s) to be taken in accordance with Technical Specifications, TRM, and COLR (LOCT) (SRO/STA Only)

Reference:

None Cog Level: Hi

Explanation:

The initiation is from an initiation in the U1 and U2 rooms, it can be from the detector or the manual actuation lever. A table within the operating procedure gives guidance for the LCO that needs to be entered. since the fire was in the AEER and not the washroom then the actions of TS 5.5.7 do not need to be performed.

Distractor Analysis:

Choice A: Plausible because this is halve of the logic not the full initiation logic and the TS actions are correct.

Choice B: Plausible because this is halve of the logic not the full initiation logic and the TS actions would be correct for a fire in the washroom.

Choice C: Correct Answer, see explanation Choice D: Plausible because the initiation logic is correct and the TS actions would be correct for a fire in the washroom.

SRO Basis:

Facility operating limitations in the technical specifications and their bases. [10 CFR 55.43(b)(2)]

From the operating procedure, OP-37.

83. S295001 1 Unit Two was operating at rated power when a trip of 2A RFP occurred followed immediately by a trip of the 2B Reactor Recirc pump.

The following plant conditions exist:

Reactor power 52%

Total core flow (WTCF) 36.96 Mlbm/hr OPRM system Inoperable Which one of the following completes the statements below?

The current operating point on the appropriate Power to Flow Map is (1).

Verifying the current operating point on the Power to Flow Map is directed by (2)

Supplementary actions.

(Reference provided)

A. (1) 5% Buffer Region (2) 2AOP-04.0, Low Core Flow, ONLY B. (1) 5% Buffer Region (2) 2AOP-04.0, Low Core Flow AND 0AOP-23.0, Condensate/Feedwater System

Failure, C. (1) Region B - Immediate Exit (2) 2AOP-04.0, Low Core Flow, ONLY D. (1) Region B - Immediate Exit (2) 2AOP-04.0, Low Core Flow AND 0AOP-23.0, Condensate/Feedwater System

Failure, Answer: B K/A:

295001 PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION G2.01.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

RO/SRO Rating: 4.6/4.6 Pedigree: Modified from NRC Exam 10-2 (Instead of asking which power to flow map this question asks the region on the power to flow map.)

Objective:

CLS-LP-302-C, Obj. - 4. Given plant conditions and AOP-04.0, determine the required supplementary actions.

Reference:

B2C21 Core Operating Limits Report Figures 2 and 4.

Cog Level: High

Explanation:

AOP-04 & AOP-23 both provide guidance determine the current operating point on the Power to Flow Map.

Without I&C doing adjustments to APRMs the current power to flow map is still the two loop map.

Distractor Analysis:

Choice A: Plausible because this is the correct region and AOP-4.0 does have an action but AOP-23 also has an action to determine the current operating point on the power to flow map Choice B: Correct Answer, see explanation Choice C: Plausible because this is the region on the the single loop power to flow map, not the two loop and AOP-4.0 does have an action but AOP-23 also has an action to determine the current operating point on the power to flow map.

Choice D: Plausible because this is the region on the the single loop power to flow map, not the two loop.

SRO Basis:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)]

From the two loop power to flow map:

From the single loop power to flow map:

84. S295005 1 Which one of the following completes the statements below IAW Technical Specification 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation?

Three channels of feedwater and main turbine high water level trip instrumentation are required to be operable (1).

The bases for the high water level trip instrumentation indirectly initiating a reactor scram from the main turbine trip on high reactor level is to (2).

A. (1) in MODE 1 (2) mitigate the reduction in MCPR B. (1) in MODE 1 (2) protect the main turbine from damage C. (1) when thermal power is >23% RTP (2) mitigate the reduction in MCPR D. (1) when thermal power is >23% RTP (2) protect the main turbine from damage Answer: C K/A:

295005 MAIN TURBINE GENERATOR TRIP G2.02.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2)

RO/SRO Rating: 4.0/4.7 Pedigree: New Objective:

LOI-CLS-LP-026, Obj. 29 - Given plant conditions, determine whether given plant conditions meet minimum Technical Specifications requirements associated with the Main Turbine, Gland Seal, and Moisture Separator Reheater system.

Reference:

None Cog Level: Memory Explanation:

This TS is applicable when thermal power is greater than 23% RTP. IAW the bases the reason for the scram is to mitigate MCPR.

Distractor Analysis:

Choice A: Plausible because MODE 1 is power operation.

Choice B: Plausible because MODE 1 is power operation and this is the reason for the turbine trip not the scram from the turbine trip.

Choice C: Correct Answer, see explanation Choice D: Plausible because this is the reason for the turbine trip not the scram from the turbine trip.

SRO Basis:

Facility operating limitations in the technical specifications and their bases. [10 CFR 55.43(b)(2)]

85. S295015 1 The following plant conditions exist on Unit Two:

An ATWS with a spurious Group I Isolation has occurred HPCI is injecting to the RPV to maintain RPV level A-01 (1-5) Suppression Chamber Lvl Hi-Hi is in alarm Which one of the following identifies:

(1) the reason that HPCI is re-aligned from its current suction source and (2) the procedure that contains the steps to perform the actions to transfer the HPCI suction valves?

A. (1) To prevent pump bearing damage (2) 2OP-19, High Pressure Coolant Injection System Operating Procedure B. (1) To prevent pump bearing damage (2) SEP-10, Circuit Alteration Procedure C. (1) To prevent HPCI exhaust check valve damage (2) 2OP-19, High Pressure Coolant Injection System Operating Procedure D. (1) To prevent HPCI exhaust check valve damage (2) SEP-10, Circuit Alteration Procedure Answer: B K/A:

295015 INCOMPLETE SCRAM G2.04.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

(CFR: 41.10 / 43.5 / 45.13)

RO/SRO Rating: 3.8/4.3 Pedigree: Last used on the 2010-1 NRC Exam Objective:

CLS-LP-019-A, Obj. 26g - Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event: High Suppression Pool water level

Reference:

None Cog Level: High Explanation:

Very high lube oil temperatures can result in loss of lubricating qualities in the oil and thus cause damage to the bearings. Suction for HPCI is normally aligned to the Condensate Storage Tank (CST) if it is available.

In accordance with the caution on LPC, the HPCI automatic suction transfer logic can be defeated to allow this lineup if suppression pool temperature is approaching 140°F. Step RC/L-23.

Defeat HPCI Hi Suppression Pool Level Suction Transfer is performed per SEP-10.

Distractor Analysis:

Choice A: Plausible because 2OP-19 contains direction for transferring the HPCI suction from the torus to the CST. (Section 8.9)

Choice B: Correct Answer, see explanation Choice C: Plausible because HPCI operation <2100 rpms can cause exhaust check valve damage.

2OP-19 contains direction for transferring the HPCI suction from the torus to the CST. (Section 8.9)

Choice D: Plausible because HPCI operation <2100 rpms can cause exhaust check valve damage.

SRO Basis:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)]

86. S295022 1 Unit One was at full power when all offsite power was lost.

The following is the Emergency Diesel Generator status:

DG1 Locked out on fault DG2 Running and loaded DG3 Running and loaded DG4 Locked out on fault Which one of the following completes the statements below?

The (1) CRD pump must be started to re-establish the CRD system.

(2) contains the step for placing the CRD Flow Control, C11-FC-R600, in manual with manual potentiometer at minimum setting?

A. (1) 1A (2) 1OP-08, Control Rod Hydraulic System Operating Procedure.

B. (1) 1A (2) 0AOP-02, Control Rod Malfunction/Misposition.

C. (1) 1B (2) 1OP-08, Control Rod Hydraulic System Operating Procedure.

D. (1) 1B (2) 0AOP-02, Control Rod Malfunction/Misposition.

Answer: C K/A:

295022 LOSS OF CRD PUMPS AA2 Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS:

(CFR: 41.10 / 43.5 / 45.13) 02 CRD system status RO/SRO Rating: 3.3/3.4 Pedigree: 08 NRC Exam Objective:

CLS-LP-302G, Obj. 4c. Given plant conditions and any of the following AOP's, determine the required supplementary actions: AOP-36.1.

Reference:

None Cog Level: High Explanation:

With a loss of all offsite power the E-Buses will strip the loads (CRD Pumps), there are no auto starts for these pumps, so both CRD pumps will be off. DG1 is lost which means E1 is lost and A CRD pump will not be able to be started. The steps for restart are located in the OP. The DG4 loss is a loss of the 2B CRD pump.

Distractor Analysis:

Choice A: Plausible because A CRD pump is tripped but the E-bus has no power for restarting the pump.

Unit 2 has power for the 2A CRD pump.

Choice B: Plausible because A CRD pump is tripped but the E-bus has no power for restarting the pump and this condition is an entry condition for AOP-02 but it does not provide direction to perform this step. Unit 2 has power for the 2A CRD pump.

Choice C: Correct Answer, see explanation Choice D: Plausible because this condition is an entry condition for AOP-02 but it does not provide direction to perform this step.

SRO Basis:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)]

The first part can be answered by knowing the power supply to the pump (systems knowledge.

The second part of the question (SRO Knowledge) is not systems knowledge, is not an immediate operator action, is not an entry condition for AOP/EOP, and is not purpose or mitigative strategy of the procedure. It assesses plant abnormal conditions and then selects a procedure to recover or with which to proceed.

87. S295025 1 During an ATWS on Unit One the following annunciators/indications are observed:

A-05 (3-6) Reactor Vess Hi Press Trip A-03 (1-10) Safety / Relief Valve Open SRV A, C, F, and G are cycling open Which one of the following completes the statements below?

The highest that reactor pressure reached was at least (1) psig.

The bases for Step RC/P-09 of LPC is to (2).

A. (1) 1060 (2) conserve SRV accumulator pressure B. (1) 1060 (2) minimize heat discharged to the suppression pool C. (1) 1130 (2) conserve SRV accumulator pressure D. (1) 1130 (2) minimize heat discharged to the suppression pool Answer: D K/A:

295025 HIGH REACTOR PRESSURE G2.04.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 / 45.12)

RO/SRO Rating: 4.1/4.3 Pedigree: Modified from 04 NRC Exam (added part one to the question to meet the k/a)

Objective:

LOI-CLS-LP-300-E, Obj. 11 - Given plant conditions and the Level/Power Control Procedure, determine the operator actions required to stabilize or reduce reactor pressure. (LOCT)

Reference:

None Cog Level: High

Explanation:

The SRVs listed all open at 1130 psig. The yellow lights are memory lights indicating that these had been open or are currently open. 1060 is the reactor trip signal. The bases for Step RC/P-30 is to conserve accumulator pressure (sustained opening of SRVs with no continuous pnuematic supply).

Distractor Analysis:

Choice A: Plausible because the reactor trip signal is 1060 and this is the bases for maintaining the SRVs open with a loss of pnuematic supply.

Choice B: Plausible because the reactor trip signal is 1060 and this is the bases for the step.

Choice C: Plausible because 1130 psig is correct and this is the bases for maintaining the SRVs open with a loss of pnuematic supply.

Choice D: Correct Answer, see explanation SRO Basis:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)]

This question measures the SRO's assesment of high RPV pressure conditions and the knowledge of EOP symptom based steps used to prevent SRV cycling under high RPV pressure conditions.

88. S295026 1 Unit One is at rated power. 0AOP-30.0, Safety/Relief Valve Failures, has been entered for a stuck open SRV F and the supplementary actions are being performed.

The following torus temperatures are observed:

93° F Suppression Pool Temp at location 45° on ERFIS 91° F Suppression Pool Temp at location 90° on ERFIS 90° F Suppression Pool Temp at location 135° on ERFIS 91° F Suppression Pool Temp at location 180° on ERFIS 93° F Suppression Pool Temp at location 225° on ERFIS 97° F Suppression Pool Temp at location 270° on ERFIS 112° F Suppression Pool Temp at location 315° on ERFIS 96.1° F Blk Wtr Avg Supp Pool on CAC-TR-4426-1A Which one of the following identifies:

(1) the required action IAW Technical Specification 3.6.2.1, Suppression Pool Average Temperature, and (2) the consequences of pulling SRV F fuses in the incorrect order?

(Reference provided)

A. (1) Enter Condition A.

(2) Loss of power to the SRV tailpipe temperature sensors.

B. (1) Enter Condition A.

(2) The power sensing relay would be required to shift.

C. (1) Enter Condition D.

(2) Loss of power to the SRV tailpipe temperature sensors.

D. (1) Enter Condition D.

(2) The power sensing relay would be required to shift.

Answer: B K/A:

295026 SUPPRESSION POOL HIGH WATER TEMPERATURE EA2 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: (CFR: 41.10 / 43.5 / 45.13) 01 Suppression pool water temperature RO/SRO Rating: 4.1/4.2 Pedigree: New Objective:

LOI-CLS-LP-004-A, Obj 19 - Given plant conditions determine the required action(s) to be taken in accordance with Technical Specifications, associated with the Primary Containment System. (SRO/STA only) (LOCT)

Reference:

TS 3.6.2.1 with LCO and applicability removed.

Cog Level: Hi

Explanation:

The operator will have to determine that the bulk water average temperatures are the tech spec required values not the individual azimuths readings. SRV F discharges at azimuth 310° in the torus.

Tech Spec 3.6.2.1 Condition A will be entered based on no testing and temp between 95 and 110°F. The note in the procedure states that removing the fuses out of order would cause the power sensing relay to shift. The fuses remove power to the SRV and acoustic monitor which provides green, red and amber lights for the SRV.

The torus temperatures bulk water average is calculated as follows:

Location Weight Temp 4545A.

0.1734 93 93B.

16.1262 9090C.

0.1156 91 91D.

10.5196 135 0.1156 90 10.4040 180 0.1156 91 10.5196 225 0.1156 93 10.7508 270 0.1156 97 11.2132 315 0.1734 112 19.4208 AVG 0.0752 95.29 7.1655 Bulk Water Temp Average 96.1197 Distractor Analysis:

Choice A: Plausible because Condition A is correct and a loss of power would occur to the acoustic monitor but not the temperature sensors.

Choice B: Correct Answer, see explanation Choice C: Plausible because this is a TS required action for 110°F in the suppression pool, but it is based on average water temperature not a local temperature and a loss of power would occur to the acoustic monitor but not the temperature sensors.

Choice D: Plausible because this is a TS required action for 110°F in the suppression pool, but it is based on average water temperature not a local temperature and the second part is correct.

SRO Basis:

Facility operating limitations in the technical specifications and their bases. [10 CFR 55.43(b)(2)]

89. S295030 1 Unit Two is executing RVCP with the following conditions present:

Reactor water level 55 inches (N036/N037) and rising Supp. chamber pressure 1.5 psig Core Spray One loop injecting at 3000 gpm RHR One loop injecting at 8800 gpm Supp. pool level

-5 feet 7 inches Supp. pool temp 178° F Which one of the following actions are required IAW RVCP to ensure there is no Core Spray or RHR pump damage?

(Reference provided)

A. Raise Core Spray flow to 5000 gpm and shutdown the RHR pump(s).

B. Raise Core Spray flow to 5000 gpm and lower RHR flow to 8,000 gpm.

C. Raise Core Spray flow to 3600 gpm and shutdown the RHR pump(s).

D. Raise Core Spray flow to 3600 gpm and lower RHR flow to 8,000 gpm.

Answer: C K/A:

295030 LOW SUPPRESSION POOL WATER LEVEL EA2 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: (CFR: 41.10 / 43.5 / 45.13) 02 Suppression pool temperature RO/SRO Rating: 3.9/3.9 Pedigree: Bank Objective:

LOI-CLS-LP-300-B, Obj. 17 - Given plant conditions, determine if the NPSH and/or Vortex Limit has been exceeded IAW the NPSH and Vortex Limit Graphs.

Pedigree: 07 NRC Exam

Reference:

0EOP-01-UG, Attachment 5, Figure 5, 6, 9, 10, 11, and 12.

Cog Level: Hi Explanation:

The NPSH curve must be lowered to at or below to the 0 psig curve because suppression pool level is below -31 inches. For each foot below -31 inches the pressure must be lowered 0.5 psig. 5 feet 7 inches

(-67 inches) is 3 feet below -31 inches therefore the NPSH curve is 0 psig. This results in 3600 gpm for core spray. However a suppression pool level of -67 inches is below the RHR vortex limit of 5.25 feet on Unit Two (not for Unit One). Since water level is above TAF and RPV level is rising the SCO should direct securing the RHR pump(s).

Distractor Analysis:

Choice A: Plausible because core spray injection flow should be raised to maintain or restore RPV level, this value is representative of using 1.5 psig in the supp. chamber without the adjustment for level and RHR should be secured based on the vortex figure.

Choice B: Plausible because core spray flow should be raised to maintain or restore RPV level, this value is representative of using 1.5 psig in the supp. chamber without the adjustment for level.RHR flow is above the NPSH curve for the pumps, so this would be correct if the vortex curve was not exceeded.

Choice C: Correct Answer, see explanation Choice D: Plausible because the NPSH curve must be lowered to the 0 psig curve because suppression pool level is below -31 inches. The RHR flow value is representative of using 1.5 psig in the supp. chamber without the adjustment for level.

SRO Basis:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)]

The question (SRO Knowledge) is not systems knowledge, is not an immediate operator action, is not an entry condition for AOP/EOP, and is not purpose or mitigative strategy of the procedure.

It is knowledge of when and how to implement attachments in the EOP network.

90. S295031 1 Unit One is shutting down for a forced outage IAW GP-05, Unit Shutdown, due to a failing reactor recirculation pump seal #1.

RCIC is in day 4 of a 7 day LCO.

The reactor mode switch is placed in SHUTDOWN as directed by the procedure.

The startup level control valve fails and HPCI is manually placed in service to maintain RPV water level.

Reactor water level dropped to 150 inches before being restored and maintained in the normal band.

Which one of the following completes the statement below?

IAW 0OI-01.07, Notifications, this event meets the conditions for reportability within:

(Reference provided)

A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ONLY.

B. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ONLY.

C. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

D. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Answer: B K/A:

295031 REACTOR LOW WATER LEVEL G2.04.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11)

RO/SRO Rating: 2.7/4.1 Pedigree: New Objective:

LOI-CLS-LP-201-D, Obj. 12 - Given plant conditions and an event, determine any applicable reporting requirements per OI-01.07, Notifications. (LOCT)

Reference:

0OI-01.07, Notifications, Attachment 1, Reportability Evaluation Checklist Cog Level: Hi Explanation:

Reactor water level has dropped to less that the RPS actuation signal (166 inches) thereby making this a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report. The NOTE states that manual initiation is reportable for HPCI system thereby making the HPCI injection even though it was manually accomplished an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report.

Distractor Analysis:

Choice A: Plausible because even though the plant is in a tech spec action statement this is not the reason for the shutdown. If tech specs would be the reason for the shutdown then this would be correct.

Choice B: Correct Answer, see explanation Choice C: Plausible because even though the plant is in a tech spec action statement this is not the reason for the shutdown. If tech specs would be the reason for the shutdown then this would be correct. An RPS actuation signal is recieved but not while the unit was critical which makes the four hour incorrect.

Choice D: Plausible because an RPS actuation signal is recieved but not while the unit was critical which makes the four hour incorrect and an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report for the HPCI injection is correct.

SRO Basis:

The question is linked to a task that is labeled as an SRO-only task, and the task is NOT listed in the RO task list. These are not RO tasks.

The STA task is Determine Non Emergency Reportability Requirements Independent Of The SRO Determination per OI-1.07 and NUREG-1022.

The SRO task is Determine Reportability Requirements per 0OI-1.07 Independent Of The STA Determination.

91. S295033 1 A fuel bundle has been dropped in the Unit Two Spent Fuel Pool with area radiation values as indicated on Attachment 1, Area Radiation Monitoring.

Which one of the following completes the statements below?

An area (1) exceeded the Max Norm Operating Radiation Limit.

The Alternate Source Term Implementation analysis dictates that the maximum allowable time to manually start the Control Room Emergency Ventilation system is (2) minutes.

(Reference provided)

A. (1) has (2) 15 B. (1) has (2) 20 C. (1) has NOT (2) 15 D. (1) has NOT (2) 20 Answer: B K/A:

295033 HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS G2.02.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45.12)

RO/SRO Rating: 4.2/4.4 Pedigree: New Objective:

LOI-CLS-LP-300-M, Obj. 06e - Given plant conditions and the Secondary Containment Control Procedure, determine if any of the following have been exceeded: Maximum normal operating radiation levels (LOCT)

Reference:

0EOP-01-UG Att. 10 Fig. 24, & Attachment 1, Area Radiation Monitoring Cog Level: High Explanation:

AOP-05, Radioactive Spills, High Radiation, and Airborne Activity, would be entered for the high rad conditions and for the fuel handling accident.

For the fuel handling accident the CREV system must be manually started within 20 minutes to ensure habitability of the main control room (Alternative Source Term analysis). The 15 minute requirement is contained in AOP-05, for isolating fire protection on indications of a HELB.

IAW SCCP the table indicates that the values of the indications are above the conditions for Max Norm. and some of these values are greater than the values for an Alert classification

Distractor Analysis:

Choice A: Plausible because the Max Norm rad levels have been violated and 15 minutes is the bases for closing the Fire protection valve that must be isolated for a HELB (this step is in the same procedure as starting the CREV system).

Choice B: Correct Answer, see explanation Choice C: Plausible because the student may misread the log scales for the meters and determine it has not been exceeded and 15 minutes is the bases for closing the Fire protection valve that must be isolated for a HELB (this step is in the same procedure as starting the CREV system).

Choice D: Plausible because the student may misread the log scales for the meters and determine it has not been exceeded and 20 minutes is correct.

SRO Basis:

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]

This question also requires the examinee to know the bases for performing actions stated in the emergency/abnormal procedures.

92. S295038 1 During accident conditions, the source term from the Unit One Turbine Building Ventilation must be estimated IAW 0PEP-03.6.1, Release Estimates Based Upon Stack/Vent Readings. Available data:

1-D12-RR-4548-3-2-1 Reading 7.425 E-1 Ci/cc 1-VA-FT-3358 Failed low (Turbine Building Vent Flow)

Turb Bldg HVAC 2 exhaust fans running Sample results 1.284 E+4 Ci/sec (taken 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago)

Which one of the following identifies the highest emergency action level classification that is required for these conditions?

(Reference provided)

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: B K/A:

295038 HIGH OFF-SITE RELEASE RATE EA2 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE:

(CFR: 41.10 / 43.5 / 45.13) 03 Radiation levels RO/SRO Rating: 3.5/4.3 Pedigree: New Objective:

LOI-CLS-LP-301-A, Obj. 6 - Determine data required for offsite dose projection in accordance with PEP-03.4.7, Automation of Offsite Dose Projection, and PEP-03.6.1, Release Estimates Based Upon Stack/Vent Readings. (LOCT)

Reference:

0PEP-03.6.1 Attachment 3 and 0PEP-02.1 Cog Level: High Explanation:

Per Attachment 3 the estimated release is calculated as follows:

Monitor reading (Ci/cc) X Flow (15,500) X Conversion factor (472) 7.425 E-1 X 15,500 X 472 = 5.4321 E+6 Ci/sec The threshold for an Alert is a value greater than 3.49 E+5 Ci/sec.

Distractor Analysis:

Choice A: Plausible because this is the EAL classification for the sample results.

Choice B: Correct Answer, see explanation Choice C: Plausible because if 15,500 per exhaust fan is used the calculation would be 1.0864 E+7 which would make a site area emergency correct.

Choice D: Plausible because If 7.425 E+1 is used the calculation would be 5.4321 E+8 which would make a general emergency correct.

SRO Basis:

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]

93. S400000 1 IAW Technical Specification 3.7.2, Service Water (SW) System and Ultimate Heat Sink (UHS), which one of the following inoperable pumps, by itself, would require entry into an action statement?

A. CSW Pump 1B B. CSW Pump 2B C. NSW Pump 1B D. NSW Pump 2B Answer: A K/A:

400000 COMPONENT COOLING WATER SYSTEM (CCWS)

A2 Ability to (a) predict the impacts of the following on the CCWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) 01 Loss of CCW pump RO/SRO Rating: 3.3/3.4 Pedigree: New Objective:

LOI-CLS-LP-043, Obj. 18 - Given plant conditions associated with the Service Water system determine the required action(s): b. to be taken IAW Technical Specifications. (LOCT)

(SRO/STA Only)

Reference:

none Cog Level: Fundamental Explanation:

Each unit has two NSW pumps and 3 are required IAW TS 3.7.2. So having one of these pumps inoperable does not require entry into the TS action statements. Each unit has 3 CSW pumps. IAW the bases for Unit 1 it requires specifically the 1A and 1B CSW pumps while Unit 2 bases specifically requires the 2A and 2C CSSW pumps.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because this pump inoperable on the other unit would make this correct.

Choice C: Plausible because one of the two pumps on the unit is inoperable, but theere are still 3 pumps between the units combined.

Choice D: Plausible because one of the two pumps on the unit is inoperable, but theere are still 3 pumps between the units combined.

SRO Basis:

Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

This requires knowledge of the bases document in what pumps are required for the system to be considered operable.

94. SG2.01.39 1 Which one of the following events would require you to direct a Reactor Scram in order to maintain safe operation of the facility?

1. With Unit Two in MODE 1 an electrical fire has resulted in erratic or questionable indications on numerous main control room nuclear instruments

2. An earthquake has occurred in which the National Earthquake Center reports horizontal ground accelerations of 0.08 g were registered. Operating Basis Earthquake (OBE) exceedance light is energized. All other plant indications indicate the plant is currently stable.

3. Power is at 50% and power ascension is in progress after a refuel outage. An accident on the Refuel Floor involving spent fuel has caused the Shift Manager to declare a Site Area Emergency.

4. Department of Homeland Security has increased the National Threat Advisory System (NTAS) Level to elevated threat for Brunswick County.

A. Event 1 B. Event 2 C. Event 3 D. Event 4 Answer: A K/A:

G2.01.39 Knowledge of conservative decision making practices. (CFR: 41.10 / 43.5 / 45.12)

RO/SRO Rating: 3.6/4.3 Pedigree: Bank, modified from Browns Ferry bank Objective:

LOI-CLS-LP-201-C, Obj. 14 -

Reference:

None Cog Level: High Explanation:

With an electrical fire and errratic operation of key equipment/indications then a manual scram would be required.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because a severe event has occurred.

Choice C: Plausible because a declaration of a SAE does not always necessitate a reactor scram.

Choice D: Plausible because on an immenent threat the reactor will be scrammed.

SRO Basis:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

From Pre Fire Plans:

From ASSD-01, Operator Actions:

95. SG2.02.03 1 Which one of the following completes the statements below concerning the Unit Two Turbine Bypass System?

The minimum number of inoperable turbine bypass valves that would require entry into an action statement of LCO 3.7.6, Main Turbine Bypass System, is (1).

The capacity of the Unit Two Turbine Bypass System is (2) %.

A. (1) two (2) 20.6 B. (1) two (2) 69.6 C. (1) three (2) 20.6 D. (1) three (2) 69.6 Answer: D K/A:

G2.02.03 (multi-unit license) Knowledge of the design, procedural, and operational differences between units. (CFR: 41.5 / 41.6 / 41.7 / 41.10 / 45.12)

RO/SRO Rating: 3.8/3.9 Pedigree: New Objective:

LOI-CLS-LP-25 Obj. 9, Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM and COLR, determine whether given plant conditions meet minimum Technical Specifications, the TRM or ODCM requirements for the Main Steam system

Reference:

None Cog Level: memory Explanation:

From the bases document Turbine Bypass System is inoperable if 2 Bypass valves are inoperable on U1 and 3 bypass valves are inoperable on U2. The U2 system contains 10 bypass valves while U1 only has 4 bypass valves.

Distractor Analysis:

Choice A: Plausible because U1 would be inoperable with this number of inoperable bypass valves and this is the capacity of Unit 1.

Choice B: Plausible because U1 would be inoperable with this number of inoperable bypass valves and this is the capacity of Unit 2.

Choice C: Plausible because U2 is inoperable and this is the capacity of Unit 1.

Choice D: Correct Answer, see explanation

SRO Basis:

Facility operating limitations in the TS and their bases [10 CFR 55.43(b)(2)]

From Unit One Bases:

From Unit Two Bases:

From SD-25:

96. SG2.02.37 1 The following information was obtained during the last scram timing for control rod 18-19 IAW 0PT-14.2.1, Single Rod Scram Insertion Times Test.

Control Position Time Rod Insertion Notch (Secs) 18-19 5%

46 0.438 20%

36 1.188 50%

26 2.026 90%

6 3.349 Unit One is operating at rated power when control rod 18-19 scram accumulator has depressurized and cannot be repaired for two days.

All other control rods and control rod scram accumulators are operable.

Concerning control rod 18-19, which one of the following completes the statements below?

The scram times (1) within Technical Specification 3.1.4, Control Rod Scram Times.

IAW Technical Specification 3.1.5, Control Rod Scram Accumulators the control rod (2) be declared SLOW.

(Reference provided)

A. (1) are (2) can B. (1) are (2) cannot C. (1) are NOT (2) can D. (1) are NOT (2) cannot Answer: D K/A:

G2.02.37 Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7 / 43.5 / 45.12)

RO/SRO Rating: 3.6/4.6 Pedigree: Modified from 08 NRC Exam, made the control rod slow so that it must be declared inoperable Objective:

CLS-LP-08, Obj. 18. Given plant conditions and TS, including the bases, TRM, ODCM, and COLR, determine the required actions to be taken in accordance with TS associated with CRD system. (SRO/STA Only)

Reference:

TS 3.1.5 and TS 3.1.4 Cog Level: high Explanation:

IAW TS 3.1.4 Since the last scram timing was outside of the limits for notches 26 and 36 but within the total allowable time of 7 seconds this rod would be declared SLOW based on this TS. If the examinee adds the times together it will be outside of the 7 seconds and could think that the rod is inoperable.

Control rod scram accumulators shall be operable in Modes 1and 2.

One control rod scram accumulator inoperable with reactor steam dome pressure >950 psig the required action is to declare the associated control rod scram time slow (only applicable if it was within the limits of Table 3.1.4-1 during the last scram time surv.) or declare the associated control rod inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Since it was slow this means it must be declared inop.

Distractor Analysis:

Choice A: Plausible because if the scram times were good then this answer could be correct.

Choice B: Plausible because if the scram times were good then this answer could be correct.

Choice C: Plausible because if the examinee looks at only the 3.1.4 TS then the rod is slow.

Choice D: Correct Answer, see explanation.

SRO Basis:

Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

97. SG2.03.04 1 During accident conditions, an auxiliary operator is needed to enter the reactor building for local emergency actions to prevent fuel damage. Due to elevated reactor building radiation levels, it is estimated the operator will receive 7.5 rem.

Which one of the following completes the statements below?

The estimated dose of 7.5 rem (1) exceed EPA-400 limits.

The Site Emergency Coordinator (2) authorize exceeding 10CFR20 limits IAW 0PEP-3.7.6, Emergency Exposure Controls.

A. (1) will not (2) can B. (1) will not (2) cannot C. (1) will (2) can D. (1) will (2) cannot Answer: A K/A:

G2.03.04 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 /

43.4 / 45.10)

RO/SRO Rating: 3.2/3.7 Pedigree: 07 NRC Exam Objective:

CLS-LP-102-A, Obj. 11 - State the emergency worker exposure limits listed in EPA 400 for each of the following conditions:

b. Protection of valuable property

Reference:

None Cog Level: Memory Explanation:

Per PEP-03.7.6, emergency limits follow EPA-400 guidelines of 10 rem for protection of valuable property and 25 rem for life saving action. Exceeding 10 CFR 20 limits (5 rem) requires authorization of the SEC for on site activities.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: incorrect because it may be authorized by the SEC.

Choice C: incorrect because the 7.5 rem does not exceed the EPA-400 limit.

Choice D: incorrect because the 7.5 rem does not exceed the EPA-400 limit and it may be authorized by the SEC.

SRO Basis:

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]

98. SG2.03.14 1 Unit Two is shutdown to support drywell entry due to Recirculation Pump oil level concerns. Reactor coolant temperature is 200°F.

E&RC has determined that the drywell atmosphere is not suitable for unfiltered release.

Which one of the following completes the statements below IAW 2OP-24, Section 6.3.13, Primary Containment Purging (Deinerting) Through the SBGT System?

This section (1) be performed under the current plant conditions.

If drywell pressure was above 0.7 psig, deinerting could result in (2).

A. (1) can (2) contamination of the RB 50' B. (1) cannot (2) contamination of the RB 50' C. (1) can (2) exceeding ODCM Main Stack release rates D. (1) cannot (2) exceeding ODCM Main Stack release rates Answer: A K/A:

G2.03.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12 / 43.4 / 45.10)

RO/SRO Rating: 3.4/3.8 Pedigree: 10-2 NRC Exam Objective:

CLS-LP-10 Obj. 9. Determine the effect that the following will have on SBGT:

j.High Drywell Pressure

Reference:

None Cog Level: High Explanation:

ONLY allowed to purge through purge fans if drywell atmosphere is suitable for unfiltered release. Must purge through SBGT and procedure ONLY allows purge through SBGT in Mode 4 due to LOCA concerns.

Drywell pressure above.7 psig can cause the loss of the SBGT System water seal which will cause the RB 50' to be contaminated with airborne activity.

Distractor Analysis:

Choice A: Correct Answer, see explanation Choice B: Plausible because contaminating the RB 50' is correct and deinerting the drywell cannot be commenced greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before the unit being gets 15% power. Based on not suitable for unfiltered release, deinerting will be delayed until Mode 4 is reached.

Choice C: Plausible because higher drywell pressure would provide for higher Main Stack release rates but the Main Stack Rad Hi-Hi isolation is set to ensure ODCM release rates will not be exceeded and Mode 4 is correct.

Choice D: Plausible because higher drywell pressure would provide for higher Main Stack release rates but the Main Stack Rad Hi-Hi isolation is set to ensure ODCM release rates will not be exceeded and deinerting the drywell cannot be commenced greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before the unit gets below 15% power. Based on not suitable for unfiltered release, deinerting will be delayed until Mode 4 is reached.

SRO Basis:

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]

99. SG2.04.27 1 A fire in the control building fire area requires entry into 0ASSD-01, Alternative Safe Shutdown Procedure Index. The CRS has determined that alternate safe shutdown actions are required. Both Unit One and Unit Two have been manually scrammed.

Which one of the following completes the statements below IAW 0ASSD-01?

The next action that is required is to (1).

Following this action both units will (2).

A. (1) place MSIV control switches in close (2) perform 0ASSD-01, Alternative Safe Shutdown Procedure Index concurrently with 0ASSD-02, Control Building.

B. (1) trip both Reactor Recirc pumps (2) perform 0ASSD-01, Alternative Safe Shutdown Procedure Index concurrently with 0ASSD-02, Control Building.

C. (1) place MSIV control switches in close (2) exit 0ASSD-01, Alternative Safe Shutdown Procedure Index and enter 0ASSD-02, Control Building D. (1) trip both Reactor Recirc pumps (2) exit 0ASSD-01, Alternative Safe Shutdown Procedure Index and enter 0ASSD-02, Control Building Answer: C K/A:

G2.04.27 Knowledge of fire in the plant procedures. (CFR: 41.10 / 43.5 / 45.13)

RO/SRO Rating: 3.4/3.9 Pedigree: 08 NRC Exam Objective:

CLS-LP-304, Obj. 12. Given plant conditions with an ASSD fire and the ASSD procedures, determine the appropriate operator actions to be performed for the fire.

Reference:

None Cog Level: High Explanation:

The tripping of the Recirc pumps is required for AOP-32 Plant Shutdown from Outside the Control Room, therefore a plausible option.The only action directed from the applicable section of ASSD-01 is to place the MSIV control switches to close. For a fire in the control building ASSD-01 is exited, for the other areas the procedure is perform concurrently.

Distractor Analysis:

Choice A: Plausible because number one is correct and if the fire was in a different area then perfoming concurrently would be correct.

Choice B: Plausible because tripping of the recirc pumps is an action that is performed on a control room evacuation and if the fire was in a different area then perfoming concurrently would be correct.

Choice C: Correct Answer, see explanation.

Choice D: Plausible because tripping of the recirc pumps is an action that is performed on a control room evacuation and exiting the procedure is correct.

SRO Basis:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations [10 CFR 55.43(b)(5)]

From AOP-32

100. SG2.04.45 1 The following alarms and indications exist on Unit One:

A-05 (5-6) Pri Ctmt Press Hi Trip is in alarm A-03 (6-9) Reactor Low Wtr Level Initiation is in alarm A-03 (5-1) Auto Depress Timers Initiated is in alarm Reactor coolant sample yields a result of 310 µCi/gm Iodine-131 Inboard and Outboard MSIV logic lights are illuminated No area radiation or temperatures are above Max Normal Operating Levels Which one of the following completes the statement below?

These alarms and indications establish that a loss of the _____ exists.

(Reference provided)

A. Containment AND Fuel Clad Barriers ONLY B. Reactor Coolant System AND Fuel Clad Barriers ONLY C. Reactor Coolant System AND Containment Barriers ONLY D. Containment, Reactor Coolant System AND Fuel Clad Barriers Answer: B K/A:

G2.04.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5

/ 45.3 / 45.12)

RO/SRO Rating: 4.1/4.3 Pedigree: New Objective:

LOI-CLS-LP-301-B, Obj. 7 - Identify how each of the following symptoms and indicators relate to one or more fission product barrier loss or potential loss:

a. Coolant Activity

d. Containment Pressure

g. Containment Isoation Status

Reference:

0PEP-02.1, Brunswick Nuclear Plant Initial Emergency Actions Cog Level: High Explanation:

Pri Ctmt Press Hi Trip alarm indicates that Drywell pressure is > 1.7 psig which is a loss of the RCS barrier.

The Iodine levels of the sample indicate a loss of the Fuel Cladding Barrier. While there is a failure of the Group 1 to isolate (level less than LL3) there are no given indications of the system discharging outside of its normal pathway, so a loss of the containment barrier does not exist.

Distractor Analysis:

Choice A: Plausible because a Group 1 isolation has failed and if there were indications of this discharging outside of its normal pathway then the Containment barrier would be lost and the Fuel Clad barrier is lost based on iodine levels.

Choice B: Correct Answer, see explanation Choice C: Plausible because a Group 1 isolation has failed and if there were indications of this discharging outside of its normal pathway then the Containment barrier would be lost and the RCS barrier is lost based on drywell pressure Choice D: Plausible because a Group 1 isolation has failed and if there were indications of this discharging outside of its normal pathway then the Containment barrier would be lost and the RCS barrier is lost based on drywell pressure and the Fuel Clad barrier is lost based on iodine levels.

SRO Basis:

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]