ML12284A451

2012 Dresden Nuclear Power Station Initial License Examination Proposed Written Examination RO & SRO

ML12284A451
Person / Time
Site:	Dresden
Issue date:	09/26/2012
From:	Nelson J, David Reeser Operations Branch III
To:	Exelon Generation Co
Shared Package
ML11354A123	List: ML12284A346 ML12284A451 ML12284A455 ML12284A460 ML12284A463 ML12284A470 ML12284A480 ML12291A697 ML12291A698 ML12291A700 ML12291A701
References
Download: ML12284A451 (200)

Text

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 1 of 100 ILT 11-1 (NRC 2012-301) 1 Points: 1.00 Given the following conditions TR22 has tripped due to an internal fault, resulting in a reactor scram and subsequent loss of off-site power.

The Unit 2 Isolation Condenser is in service providing reactor pressure control. The 2-1301-3, RX INLET ISOL, valve is being throttled to control cooldown.

A fire has been reported in 125 VDC Bus 2A-1.

250 VDC MCC 2 has tripped due to a short circuit.

The Fire Brigade leader has called to inform you 125 VDC Bus 2A-1 is going to be de-energized by the HVO so the fire can be extinguished.

Which of the following is true with regard to the Isolation Condenser when 125 VDC Bus 2A-1 is de-energized (assume no operator action)?

A.

It will automatically isolate.

B.

It will remain in service, but cooldown rate will be fixed.

C.

It will remain in service, but cooldown rate will be maximized.

D.

It will remain in service with the ability to control cooldown rate.

Answer:

A Question 1 Details Topic:

1 - 207000.K1.11 Comments: Objective: 207LN001.06

Reference:

DOA 6900-02 K/A: 207000.K1.11 3.4 / 3.6 K/A: Knowledge of the physical connections and/or cause-effect relationships between ISOLATION (EMERGENCY) CONDENSER and the following: Primary containment isolation system: BWR-2,3 CFR: 41.8 Safety Function: 4 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments:

a) Isolation Condenser Group 5 isolation circuits are de-energize to actuate. The power supplies are 2A-1 and 2B-1 for Unit 2. Loss of EITHER power supply will cause a complete isolation.

b) Isolation condenser DC MOVs are powered from 250VDC MCC 2A which receives power from Unit 3 250 VDC MCC 3. If power were lost to these valves, cooldown rate would be fixed.

c) 2-1301-3 valve HAND/RESET circuit is powered from the 120 VAC Instrument Bus and will remain energized. If power were lost to this logic, the 2-1301-3 valve would fully open maximizing cooldown rate.

d) Would be correct if the Group 5 circuit was energize to actuate (ie HPCI Group 4).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 2 of 100 ILT 11-1 (NRC 2012-301) 2 Points: 1.00 Unit 3 was operating at near rated power, when Bus 33-1 experienced an OVERCURRENT condition.

Which of the following be a correct plant response/indication?

A.

Unit 3 RPS EPA 3A-1 "IN" light extinguished.

B.

Unit 3 RPS EPA 3B-1 "IN" light extinguished.

C.

Annunciator 903-8 C-4 U2/3 DIESEL GEN FAIL TO START in alarm.

D.

Annunciator 903-8 E-8 ESS UPS ON DC OR ALTERNATE AC in alarm.

Answer:

A Question 2 Details Topic:

2 - 212000.K1.04 Comments: Objective: DRE212LN001.03

Reference:

DOP 0500-03 figure 2, DANs 903-8 E-5 and 903-8 E-8 K/A: 212000.K1.04 3.4 / 3.6 K/A: Reactor Protection System: Knowledge of the physical connections and/or cause effect relationships between REACTOR PROTECTION SYSTEM and the following: A.C. electrical distribution.

CFR: 41.7 Safety Function: 7 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC, 2008 NRC Comments: With an overcurrent on Bus 33-1 it will become completely de-energized, which will cause Bus 38 to become de-energized. With Bus 38 de-energized, MCC 38-2 also becomes de-energized, which is the power supply to RPS MG set A. When MG Set A coasts down, the EPA 3A-1 becomes de-energized (IN power light goes out).

EPA 3B-1 is incorrect because it is powered from MG Set B (MCC 39-2).

Annunciator for U2/3 EDG fail to start is incorrect, because with an overcurrent on Bus 33-1, the U2/3 EDG will auto start (but is prevented from closing in on the bus).

Annunciator for ESS Bus Transfer is incorrect because it is powered from Bus 39, not Bus 38.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 3 of 100 ILT 11-1 (NRC 2012-301) 3 Points: 1.00 Unit 2 was operating at near rated power when the 125 VDC 2B-1 Distribution panel de-energized.

Several hours later, the Electrical Maintenance Department completed repairs and the 2B-1 Distribution panel was RE-ENERGIZED.

What is the expected power supply to the Unit 2, Div 2 ADS logic without further operator actions?

A.

U2 ESS Bus B.

U2 Instrument Bus C.

125 VDC Distribution Panel 2A-1 D.

125 VDC Distribution Panel 2B-1 Answer:

D Question 3 Details Topic:

3 - 218000.K2.01 Comments: Objective: DRE218LN001.02

Reference:

12E-2461 sh 1, 12E-2462 sh 1 K/A: 218000.K2.01 3.1 / 3.3 K/A: Knowledge of electrical power supplies to the following: ADS logic CFR: 41.7 Safety Function: 3 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: The U2 Div 2 ADS logic has two power supplies. The normal supply is the 2B-1 Dist Panel and the alternate supply is the 2A-1 Dist Panel. When the normal supply (2B-1) is lost a normal seeking automatic transfer device will transfer to the alternate supply (2A-1). Upon the normal supply being re-energized, the normal seeking automatic transfer device will transfer power back from the alternate supply (2A-1) to the normal supply (2B-1).

a) The ESS Bus is a power supply to HPCI and PCIS logic.

b) The Instrument Bus is a power supply to the Isolation Condenser logic and controls.

c) 2A-1 would be the power supply if the automatic transfer device didnt transfer back to the normal supply.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 4 of 100 ILT 11-1 (NRC 2012-301) 4 Points: 1.00 With Unit 2 at 100% power, the following events occur:

MCC 29-1 de-energized due to an overcurrent trip.

Drywell pressure began rising and eventually reached 2.5 psig.

10 minutes later HPCI area temperature rises to 300°F.

Which of the following is correct concerning the status of the HPCI system?

A.

HPCI is running and injecting into the RPV.

B.

HPCI is running but NOT injecting into the RPV.

C.

HPCI injected into the RPV but is now isolated.

D.

HPCI did NOT inject into the RPV and is now isolated.

Answer:

C Question 4 Details Topic:

4 - 206000.K2.01 Comments: Objective: DRE206LN001.12

Reference:

DAN 902(3)-3 C-7, DAN 902(3)-5 D-11, DOP 6700-19 K/A: 206000.K2.01 3.2 / 3.3 K/A: Knowledge of electrical power supplies to the following: System valves CFR: 41.7 Safety Function: 2 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments:

A loss of MCC 29-1 results in a loss of 4-valve control power. 4-valve control power supplies half of the 4 & 5 valve isolation logic. However, HPCI 125 VDC still supplies the other half of the logic. The 5-valve will still isolate due to the high HPCI area temperature isolation signal. The sequence of events will result in HPCI injection due to the high DW pressure signal and then HPCI will isolate upon receipt of high area temperature.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 5 of 100 ILT 11-1 (NRC 2012-301) 5 Points: 1.00 A transient occurred on Unit-2 some time ago. Eventually the DEOP flowcharts were exited, and SAMGs were entered.

The following conditions exist:

Reactor Pressure 75 psig steady RPV Water Level

-200" steady Drywell Pressure 10 psig steady Primary Containment Water Level 80 ft. steady Electrical Power ALL buses are ENERGIZED Instrument Air has been lost, but all other required systems are operable In the current condition, which of the below may be used to vent the RPV?

A.

MSIV's B.

Head Vent C.

HPCI steam line D.

Isolation Condenser vent valves Answer:

B Question 5 Details Topic:

5 - 300000.K3.02 Comments: Objective: DRE295LN003.03

Reference:

SAMG-1; DEOP 500-06 K/A: 300000.K3.02 3.3 / 3.4 K/A: Knowledge of the effect that a loss or malfunction of the (INSTRUMENT AIR SYSTEM) will have on the following: Systems having pneumatic valves and controls CFR: 41.7 Safety Function: 8 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: Instrument Air is required to open various valves in all the above mentioned systems except the Head Vent. Without Instrument Air, only the head vent method is available in the current condition.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 6 of 100 ILT 11-1 (NRC 2012-301) 6 Points: 1.00 Unit 3 was operating at near rated power with the Computer UPS 250 VDC battery system O.O.S.

A transient occurs causing Bus 36 to trip on overcurrent.

Which of the following power supplies would provide power to the Unit 3 Rod Worth Minimizer?

A.

MCC 38-2 B.

Bus 25 C.

Bus 28 D.

Bus 35 Answer:

B Question 6 Details Topic:

6 - 262002.K3.05 Comments: Objective: 283LN001.06

Reference:

DOP 6800-03, 12E-6600 K/A: 262002.K3.05 2.9 / 3.0 K/A: Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following:

Rod worth minimizer CFR: 41.7 Safety Function: 6 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: The RWM on Unit 3 is powered from the Computer UPS.

a) MCC 38-2 is a normal power supply for Instrument Bus and alternate power supply for ESS Bus power on Unit 3.

b) Bus 25 will power the Computer UPS system on a loss of the batteries and Bus 36.

c) Bus 28 is an alternate power supply for ESS Bus power on Unit 2.

d) Bus 35 is a common misconception due to the Computer UPS normal power coming from Unit 3.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 7 of 100 ILT 11-1 (NRC 2012-301) 7 Points: 1.00 The high pressure sensing line for the core plate D/P instrumentation has a fitting leak of 10 gpm at the transmitter.

An ATWS event occurs and SBLC is started.

Which of the following describes the response of the SBLC system?

The SBLC system will inject__________.

A.

all of the boron into the RPV above the core plate B.

all of the boron into the RPV below the core plate C.

most of the boron into the RPV above the core plate, the rest outside the RPV D.

most of the boron into the RPV below the core plate, the rest outside the RPV Answer:

D Question 7 Details Topic:

7 - 211000.K4.06 Comments: Objective: DRE211LN001.03

Reference:

M-33, M-26, LPDRE223LN004 K/A: 211000.K4.06 2.6 / 2.7 K/A: Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Core plate differential pressure indication CFR: 41.7 Safety Function: 1 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC, 2003 Pilgrim Unit 1 Comments:

a) Incorrect, boron is injected below the core plate.

b) Would be correct if the low pressure sensing line tapped into the SBLC injection line.

c) Incorrect, boron is injected below the core plate.

d) With a leak on the high pressure sensing line that senses pressure below the core plate from a tap off the SBLC injection line, some of the boron would be injected below the core plate with the rest being lost outside the RPV.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 8 of 100 ILT 11-1 (NRC 2012-301) 8 Points: 1.00 Unit 2 was operating at near rated power when a transient occurred resulting in an ATWS.

The Unit Supervisor directed the Operating team to bypass the interlocks for the MSIV low water level AND Off Gas high radiation isolations.

Which of the following would be a correct indication of the interlocks above being properly bypassed?

A.

The Off Gas SJAE suction valves automatically open.

B.

Annunciator 902-5 B-13, CHANNEL A/B RPV LVL LO, can be reset.

C.

Annunciator 902-5 D-4, GROUP 1 ISOLATION INITIATED, can be reset.

D.

The MSIVs automatically open, UNLESS a valid MSL high rad signal is present.

Answer:

C Question 8 Details Topic:

8 - 223002.K4.08 Comments: Objective: 29502LK077

Reference:

DAN 902-5 B-13, DEOP 0500-02, 12E-2501 sht 1 and 2 K/A: 223002.K4.08 3.3 / 3.7 K/A: Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following: Manual defeating of selected isolations during specified emergency conditions.

CFR: 41.7 Safety Function: 5 PRA: No Level: Memory Pedigree: Bank History: 2010 NRC, 2012 NRC Explanation: The jumpers installed per the above procedure bypass the lo-lo RPV water level Group 1 signal. This permits the annunciator to be reset. The low level annunciator cannot be reset when these jumpers are installed. The MSIVs do NOT auto open (these jumpers prevent their closure). The MSL hi-rad signal no longer is an input that would cause a Group 1 isolation. The Off Gas suction valves are not affected by these jumpers.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 9 of 100 ILT 11-1 (NRC 2012-301) 9 Points: 1.00 A Unit 2 startup is in progress and the following conditions exist:

Reactor Mode Switch is in STARTUP.

All IRM range switches are on 9 or 10.

All IRMs are reading mid scale.

A fault in the IRM detector drive system causes all IRM detectors to begin to withdraw.

Which of the following would be the first indication of the fault in the IRM detector drive system?

A.

Rod Out Block B.

RPS Full Scram C.

IRM INOP alarm D.

IRM downscale alarm Answer:

A Question 9 Details Topic:

9 - 215003.K5.03 Comments: Objective: DRE215LN003.06

Reference:

DOP 0700-02, DAN 902(3)-5 C-10 and 15 K/A: 215003.K5.03 3.0 / 3.1 K/A: Knowledge of the operational implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM :

Changing detector position CFR: 41.5 Safety Function: 7 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments:

a) With the Mode Switch in STARTUP, IRMs must be fully inserted, or a Rod Out Block is generated.

b) INOP alarm is generated by low voltage, function switch not in operate and module not plugged in properly.

c) Scram signal is generated by high neutron flux or IRM HiHi/Inop with companion APRM downscale.

d) With the IRMs reading mid scale, downscale alarms would not immediately come in as soon as the detectors began to move.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 10 of 100 ILT 11-1 (NRC 2012-301) 10 Points: 1.00 A prerequisite of DOS 6620-01, SBO 2 DIESEL GENERATOR SURVEILLANCE AND MARGIN/FULL LOAD TESTFULL LOAD REJECT TEST, is to ensure that the bus selected for this surveillance is powered from either Transformers TR 21 or TR 22 AND the associated Emergency Diesel Generator is NOT paralleled to this bus.

What is the operational concern for this prerequisite?

A.

A trip of either Diesel Generator output breaker while in this lineup, would cause the Bus to de-energize.

B.

Large circulating currents, due to voltage differences, may result in an auto trip of both Diesel Generator output breakers.

C.

A neutral ground condition, due to voltage differences, may result in an auto trip of both Diesel Generator output breakers.

D.

A neutral voltage condition, due to phase angle differences, may result in an auto trip of both Diesel Generator output breakers.

Answer:

B Question 10 Details Topic:

10 - 264000.K5.05 Comments:

Objective: DRE264LN001.10

Reference:

DAN 902-8 D-1, DOS 6620-01 K/A: 264000.K5.05 3.4 / 3.4 K/A: Knowledge of the operational implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESEL/JET):

Paralleling A.C. power sources.

CFR: 41.5 Safety Function: 6 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC Comments: A caution in the above DAN states: Do NOT operate any longer than necessary with main and reserve feed breakers closed simultaneously. Large circulating currents may result in an auto trip of the feed breakers.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 11 of 100 ILT 11-1 (NRC 2012-301) 11 Points: 1.00 Unit 2 was operating at near rated power when a complete loss of U2 125 VDC Division 1 occurred.

Which of the following buses will no longer have breaker indication in the Control Room?

A.

Bus 23 B.

Bus 24 C.

Bus 26 D.

Bus 27 Answer:

A Question 11 Details Topic:

11 - 262001.K6.01 Comments:

Objective: DRE263LN002.12

Reference:

DOA 6900-02 K/A: 262001.K6.01 3.1 / 3.4 K/A: Knowledge of the effect that a loss or malfunction of the following will have on the A.C. ELECTRICAL DISTRIBUTION: D.C. power.

CFR: 41.7 Safety Function: 6 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC Explanation: With a loss of 125 VDC Division 1, all division 1 buses will lose remote indication and protection ability. Bus 23 is the only Div 1 bus listed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 12 of 100 ILT 11-1 (NRC 2012-301) 12 Points: 1.00 Unit 2 is operating at full power with the following conditions:

2A and 2B RBCCW pumps and heat exchangers are in service.

2/3 RBCCW pump and heat exchanger are lined up to Unit 3.

A transient occurs resulting in the following conditions:

Reactor scram due to high drywell pressure Drywell pressure is 7.3 psig and climbing at a rate of 1.0 psig per minute.

Bus 24-1 was de-energized and is now powered from the U2 EDG 1 Minute after U2 EDG picked up bus 24-1, 2A RBCCW pump tripped on overcurrent.

What action(s) is(are) required regarding the Unit 2 RBCCW system?

A.

Immediately start 2B RBCCW pump.

B.

Reset the under-voltage on Bus 24-1, then start 2B RBCCW pump.

C.

Isolate RBCCW to the drywell.

D.

Remove 2A RBCCW heat exchanger from service.

Answer:

C Question 12 Details Topic:

12 - 400000.K6.05 Comments: Objective: 208LN001.12

Reference:

DOA 3700-01 K/A: 400000.K6.05 3.0 / 3.1 K/A: Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: Pumps CFR: 41.7 Safety Function: 8 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: Given the conditions in the stem, all RBCCW pumps on Unit 2 have been lost. Bus 24-1 is the power supply to RBCCW pump 2B. RBCCW Pump 2B is interlocked to trip on a combination of Core Spray System initiation (+2 psig in the drywell) and loss of normal power to Bus 24-1.

a) The 2B RBCCW can not be restarted under these conditions until leads are lifted, per DEOP 0500-02, that prevent the 2B RBCCW pump from starting.

b) On a loss of Bus 24-1, the undervoltage relay is required to be reset to start loads powered on Bus 29, this action would not allow the 2B RBCCW pump to start, common misconception.

c) DOA 3700-01, LOSS OF COOLING BY RBCCW, immediate actions state: If a LOCA has occurred and drywell pressure is greater than 2 psig, then isolate the drywell RBCCW.

d) If a high drywell pressure signal was not present, the 2B RBCCW pump would be running and the examinee may determine removing the 2A RBCCW heat exchanger from service to prevent overloading the 2B RBCCW pump would be the correct action, common misconception.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 13 of 100 ILT 11-1 (NRC 2012-301) 13 Points: 1.00 Unit 2 was operating at near rated power during the summer months, when a scram occurred.

The following conditions exist:

RPV water temperature is 195oF.

Shutdown Cooling system is in service with the A and B loops.

The 2/3 RBCCW pump and heat exchanger are lined up to Unit 3.

If the 2A RBCCW pump trips on overcurrent, which of the following annunciators would be experienced FIRST?

A.

902-3 G-2, AREA TEMP HI B.

902-4 A-23, SDC HX/FUEL POOL WTR TEMP HI C.

902-4 B-23, SDC PP TRIP D.

902-4 H-4, 2A RECIRC LOOP WTR TEMP HI Answer:

B Question 13 Details Topic:

13 - 205000.A1.08 Comments:

Objective: DRE205001.08

Reference:

DOP 1000-03, DAN 902-4 A-23 K/A: 205000.A1.08 3.1 / 2.9 K/A: Shutdown Cooling System: Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including: Heat exchanger temperatures.

CFR: 41.5 Safety Function: 4 PRA: No Level: High Pedigree: Bank History: 2009 NRC; 2012 NRC Comments: 902-4 A-23 would be received FIRST since its setpoint is >

200oF SDC HX inlet and outlet shell side temp, which would occur when 2A RBCCW pump tripped (reduced RBCCW cooling). 902-3 G-2 would NOT be received first since the area temp hi does not come in until the SDC temps reach 150oF and this condition would not cause the area temperatures to rise that fast. 902-4 B-23 would NOT come in first since the SDC pump trip is a temperature of 339oF on the pump suction and would not occur that fast. 902-4 H-4 would NOT come in first since the recirc temperature of > 330oF would not be reached that fast.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 14 of 100 ILT 11-1 (NRC 2012-301) 14 Points: 1.00 The following conditions exist during a normal reactor startup:

Reactor power

- supercritical with SRMs indicating approximately 7 x 104 cps IRMs

- all indicating on scale on range 1 or 2 SRMs

- shorting links are installed A few moments later SRM channel 22 is indicating 5 x 105 cps.

This SRM indication results in:

A.

SRM HI and HI-HI alarms ONLY B.

SRM HI and HI-HI alarms and a rod block ONLY C.

SRM HI and HI-HI alarms, a rod block and a half-scram ONLY D.

A full reactor scram Answer:

B Question 14 Details Topic:

14 - 215004.A1.05 Comments: Objective: DRE215LN004.06

Reference:

DAN 902(3)-5 A-4 and B-12, DGP 1-1, DOP 0700-01 K/A: 215004.A1.05 3.6 / 3.8 K/A: Ability to predict and/or monitor changes in parameters associated with operating the SOURCE RANGE MONITOR (SRM) SYSTEM controls including: SCRAM, rod block, and period alarm trip setpoints CFR: 41.5 Safety Function: 7 PRA: No Level: High Pedigree: Bank History: 2012 NRC; 2006 NRC Comments: The information provided states IRM range switches are on range 1 or 2. SRM rod blocks are bypassed when IRMs are on range 8 or above, therefore the rod blocks are active. The SRM HI-HI scram is only functional when the Shorting Links are removed for refueling operations.

SRM HI and HI-HI alarms will be received, but they will be accompanied by the rod block.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 15 of 100 ILT 11-1 (NRC 2012-301) 15 Points: 1.00 Unit 2 is at 10% of rated thermal power, with APRM 1 having eight (8) of its input LPRMs bypassed, when the following occurs:

An LPRM input to APRM 1 fails DOWNSCALE.

APRM 1 indicates 1%.

The ___(1)___ light(s) will be illuminated and the Operator is required to bypass the failed LPRM

___(2)___.

A.

(1) APRM 1 DOWNSCALE ONLY; (2) ONLY B.

(1) APRM 1 DOWNSCALE ONLY; (2) AND APRM 1.

C.

(1) APRM 1 DOWNSCALE AND INOP; (2) ONLY D.

(1) APRM 1 DOWNSCALE AND INOP; (2) AND APRM 1 Answer:

B Question 15 Details Topic:

15 - 215005.A2.03 (1)

Comments: Objective: DRE215LN005.03

Reference:

DANs 902-5 C-6, DOP 0700-08, DRE215LN005 K/A: 215005.A2.03 3.6 / 3.8 K/A: Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inoperative trip (all causes).

CFR: 41.5 Safety Function: 7 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Explanation: INOP trip setpoint is met when nine (9) LPRMs are bypassed.

With 8 input LPRMs into the APRM bypassed ONLY the downscale light would illuminate. BOTH the failed LPRM AND APRM 1 should be bypassed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 16 of 100 ILT 11-1 (NRC 2012-301) 16 Points: 1.00 Unit 2 was operating at full power.

A leak in the 2A Target Rock valve bellows causes alarm 902-4 D-23, 2A TARGET ROCK RELIEF VLV INOP, to annunciate.

Which of the following describes (1) the modes (if any) the Target Rock WILL be able to operate and (2) the operator actions that may be used to mitigate the affect on the plant?

A.

(1) Safety ONLY (2) Align balance of plant steam loads including bypassing of isolations B.

(1) Relief and ADS ONLY (2) Align RWCU in recirculation mode C.

(1) Safety and ADS ONLY (2) Align RWCU in recirculation mode D.

(1) The Target Rock valve will NOT operate (2) Align balance of plant steam loads including bypassing of isolations Answer:

B Question 16 Details Topic:

16 - 239002.A2.02 Comments: Objective: 239LN001.03

Reference:

DAN 902-4 D-23 K/A: 239002.A2.02 3.1 / 3.2 K/A: Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Leaky SRV CFR: 41.5 Safety Function: 3 PRA: Yes Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Incorrect - DAN 902-4 D-23, a ruptured bellows will cause the Target Rock valve to be inoperable in Safety mode, aligning BOP loads is an action when ADS valves fail to open during an emergency depressurization b) Correct - ADS and Relief modes continue to be operable, these modes are independent of the pilot valve bellows and rely on drywell pneumatics to operate, per DEOP 100 RWCU in recirc mode is one of the alternate means for controlling pressure.

c) Incorrect - Target Rock valve to be inoperable in Safety mode d) Incorrect - The Target Rock ADS and Relief modes continue to be operable.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 17 of 100 ILT 11-1 (NRC 2012-301) 17 Points: 1.00 The following plant conditions exist:

SBGT System has received an autostart signal The 2/3 A SBGT train is running in PRIMARY The 2/3 B SBGT train is in STANDBY Which SBGT system damper automatically closed on the autostart signal?

A.

2/3-7509 XTIE DAM B.

2/3-7504A OUTSIDE AIR DAM C.

2/3-7504B OUTSIDE AIR DAM D.

2/3-7506A BACK DRAFT DAMPER Answer:

B Question 17 Details Topic:

17 - 261000.A3.03 Comments:

Objective: DRE 261LN001.03

Reference:

DOS 7500-02 K/A: 261000.A3.03 3.0 / 2.9 K/A: Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: Valve operation CFR: 41.7 Safety Function: 9 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Correct - Upon autostart signal, 2/3 A SBGT outside damper will go close, aligning 2/3 A SBGT suction from the reactor Building or HPCI Gland Exhauster discharge initially. INCORRECT - 2/3 B Outside air damper will remain open drawing in approx 300 scfm via the crosstie line between the SBGT trains, hence the 2/3 crosstie damper is open to allow cooling flow for the train that is in standby. 2/3 A SBGT back draft damper is on the discharge of the 2/3 A SBGT fan which is running, the back draft damper acts as a check valve and will open as soon as the 2/3 A SBGT fan is discharging flow to the 2/3 Chimney The 2/3 B SBGT back draft damper will be closed to prevent possible spread of contamination.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 18 of 100 ILT 11-1 (NRC 2012-301) 18 Points: 1.00 Unit 3 is operating with the following conditions:

Reactor power is 35% and steady Reactor water level is being maintained at 30 inches 3B FRV and LFRV are in Automatic; the 3A FRV is closed in manual.

3B RFP is running; and 3A and 3C Condensate Pumps are running A spurious full scram signal is received - all systems function as designed How will the FWLC system react?

The FWLC system will...

A.

maintain reactor water level at 5 inches without regard to feedwater flow.

B.

maintain reactor water level at 30 inches without regard to feedwater flow.

C.

close the 3B FRV for 25 seconds. After 25 seconds, it will control level at the current setpoint.

D.

maintain the 3B FRV's position for one (1) second. After one (1) second, it will pulse down to 30% of its current demand.

Answer:

D Question 18 Details Topic:

18 - 259002.A3.06 Comments: Objective: DRE259LN002.06

Reference:

DAN 902(3)-5 G-7 K/A: 259002.A3.06 3.0 / 3.0 K/A: Ability to monitor automatic operations of the REACTOR WATER LEVEL CONTROL SYSTEM including: Reactor water level setpoint setdown following a reactor scram: Plant-Specific CFR: 41.7 Safety Function: 2 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: Upon a scram signal, the system takes the setpoint to -10 and locks the FRVs in place for 1 second.

At 1 second (T1) the programming looks at reactor water level and makes one of the following decisions: If reactor water level is greater than -30, then:

The system pulses the FRVs to 30% of current demand, any valve in manual, receives a closed signal. 25 seconds later (T26) the system releases the valves in auto to control at the current setpoint. 19 seconds later (T45) the system ramps back to a setpoint of +30, at a rate of 10/minute.

If reactor water level is less than -30, then: The system releases the FRVs to control at the current setpoint (assuming water level is low enough to NOT enter the HPCI and IC steam lines).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 19 of 100 ILT 11-1 (NRC 2012-301) 19 Points: 1.00 Which of the following describe the (1) location and (2) action necessary, if any, to determine Unit 2 125VDC battery voltage from the Main Control Room?

A.

(1) 902-8 (2) no additional action required B.

(1) 902-8 (2) verify meter selector switch is in the '125V' position C.

(1) 923-2 (2) no additional action required D.

(1) 923-2 (2) verify meter selector switch is in the '125V' position Answer:

B Question 19 Details Topic:

19 - 263000.A4.02 Comments:

Objective: DRE263LN001.11

Reference:

DAN 902(3)-8 E-9, K/A: 263000.A4.02 3.2 / 3.1 K/A: Ability to manually operate and/or monitor in the control room: Battery voltage indicator: Plant-Specific CFR: 41.7 Safety Function: 6 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: An LED type indicator with a selector switch is located on the 902(3)-8 panel. The switch can be positioned to display the voltage of the unit's 250 /

125 / 24/48 VDC batteries.The sensing point for battery voltage is upstream of the selected battery to main bus breaker so if a trip of the breaker occurs, voltage is still shown in the display.

The 923-2 panel has voltmeters and selector switches for the 345Kv system REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 20 of 100 ILT 11-1 (NRC 2012-301) 20 Points: 1.00 Unit 2 was operating at full power when a transient occurred causing the following conditions:

Unit 2 scrammed TR-22 trip due to diffierential current Drywell pressure is 11 psig and slowly rising Reactor pressure is 300 psig and steady Reactor water level is -150 inches and slowly lowering Reactor Feed Pumps and HPCI have failed The unit supervisor has directed to use LPCI injection from the CST to recover reactor water level per DEOP 0500-03.

What actions are required?

A.

Line up one LPCI pump in EACH loop to take suction from the CST Verify closed BOTH LPCI loops containment spray and cooling valves Verify open the 2-1501-21 valve in BOTH LPCI loops B.

Line up one LPCI pump in EACH loop to take suction from the CST Verify closed BOTH LPCI loops containment spray and cooling valves Verify open the 2-1501-21 valve in the LPCI loop selected for injection ONLY C.

Line up one LPCI pump to take suction from the CST ONLY Verify closed containment spray and cooling valves for the loop lined up to take suction from the CST ONLY Verify open the 2-1501-21 valve in the LPCI loop selected for injection ONLY D.

Line up one LPCI pump to take suction from the CST ONLY Verify closed BOTH LPCI loops containment spray and cooling valves Verify open the 2-1501-21 valve in the LPCI loop selected for injection ONLY Answer:

D Question 20 Details Topic:

20 - 203000.A4.05 Comments: Objective: 203LN001.02

Reference:

DEOP 500-3 K/A: 203000.A4.05 4.3 / 4.1 K/A: RHR/LPCI: Injection Mode (Plant Specific) Ability to manually operate and/or monitor in the control room: Manual initiation controls CFR: 41.7 Safety Function: 2 & 4 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a &b) Only one pump is lined up at a time to take suction form the CST due to NPSH and vortexing concerns.

c) A knowledge of system flow paths is required to know that the system crosstie would allow some water from the CST into the opposite divisions containment spray and cooling valves, which is not allowed per procedure.

d) DEOP 0500-3 directs aligning one pump to take suction from the CST, closing all LPCI containment cooling and spray valves, and injecting only through the LPCI loop that was automatically selected for injection upon receiving a high drywell pressure signal.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 21 of 100 ILT 11-1 (NRC 2012-301) 21 Points: 1.00 Given the following sequence of events:

01:00

'B' LPCI subsystem is declared inoperable 02:00

'B' Core Spray subsystem is declared inoperable 03:00

'A' LPCI subsystem is declared inoperable 04:00

'A' Core Spray subsystem is declared inoperable Which of the following is the earliest time entry into LCO 3.0.3 is required?

A.

02:00 B.

03:00 C.

04:00 D.

05:00 Answer:

A Question 21 Details Topic:

21 - 209001.G.2.2.38 Comments:

Objective: DRE209LN001.07

Reference:

T.S. 3.5.1 K/A: 209001.G.2.2.38 3.6 / 4.5 K/A: Low Pressure Core Spray System: Knowledge of conditions and limitations in the facility license.

CFR: 41.10 Safety Function: 2 & 4 PRA: No Level: High Pedigree: New History: 2012 NRC Explanation: Per Tech Spec 3.5.1, when operating in mode 1, all ECCS systems are required to be operable, which includes BOTH Core Spray systems. If any two or more low pressure ECCS subsystems are inoperable LCO 3.0.3 is required to be entered immediately.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 22 of 100 ILT 11-1 (NRC 2012-301) 22 Points: 1.00 Unit 2 was operating at near rated power, when the following set of conditions occurred:

A scram on high DW pressure.

A loss of off-site power (LOOP).

The 2/3 cribhouse inlet temperature is 93oF and steady.

Given the above conditions, a maximum of ___(1)___ CCSW pumps may be run concurrently, to prevent ___(2)___.

A.

(1) 2; (2) overloading the 2/3 EDG B.

(1) 3; (2) overloading the 2/3 EDG C.

(1) 2; (2) choking the DGCWP flow to the 2/3 EDG D.

(1) 3; (2) choking the DGCWP flow to the 2/3 EDG Answer:

C Question 22 Details Topic:

22 - 400000.G.2.1.32 Comments:

Objective: DRE277LN001.13

Reference:

DOP 6600-05 K/A: 400000.G.2.1.32 3.8 / 4.0 K/A: Ability to explain and apply system limits and precautions.

CFR: 41.10 Safety Function: 8 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC, 2007 NRC Explanation: With the given cribhouse inlet temperature a max of 2 CCSW pumps may be run with the 2/3 EDG supplying Bus 23-1. This will prevent choking the DGCWP from the 2/3 EDG. This is an industry event at Dresden. The distractor for 3 CCSW pumps is wrong unless temp is between 83oF and 90oF. The distractors for overloading the 2/3 EDG is wrong because the actions of DGA 12 ensure power is distributed so that the CCSW pumps can run without overloading the EDGs. Given that there is a LOOP and Scram, actions of DGA 12 will be in progress.

REQUIRED

REFERENCE:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 23 of 100 ILT 11-1 (NRC 2012-301) 23 Points: 1.00 The Unit 2/3 Emergency Diesel Generator is being run per DOS 6600-01, DIESEL GENERATOR SURVEILLANCE TESTS, for the monthly run, when the EO reports that the local annunciator DG 2/3 B-4, DIESEL ENGINE TEMP HIGH, is in alarm.

The EO reports that the local engine temperature is 215oF.

Per the above DOS, which of the actions below are required to be taken NEXT?

A.

Close the Emergency Fuel shutoff valve.

B.

Reverse Diesel Generator cooling water flow.

C.

Shutdown the Diesel Generator using the fuel injection rack lever.

D.

Swap the 2/3 Diesel Generator Cooling Water Pump to its alternate power supply.

Answer:

C Question 23 Details Topic:

23 - 264000.K4.07 Comments:

Objective: DRE264LN001.05

Reference:

DOS 6600-01, DAN DG2/3 B-4 K/A: 264000.K4.07 3.3 / 3.4 K/A: Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: Local operation and control CFR: 41.7 Safety Function: 6 PRA: No Level: High Pedigree: Bank History: 2012 NRC Explanation: With the Diesel Generator engine temperature > 200oF, the Diesel should have tripped. Per the limitations and actions of the DOS, if the automatic or manual D/G shutdown fails to stop the D/G, then the preferred method for D/G shutdown is to trip by using the fuel injection rack lever.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 24 of 100 ILT 11-1 (NRC 2012-301) 24 Points: 1.00 Unit 2 is operating at near rated power, with Bus 25 OOS for maintenance, when the following occurs:

Bus 28 experiences an overcurrent condition.

What affect (if any) does this have on the Unit 2 24/48 VDC system?

A.

ONLY the 'A' battery chargers lose their AC power source.

B.

ONLY the 'B' battery chargers lose their AC power source.

C.

BOTH the 'A' and 'B' battery chargers will remain available.

D.

BOTH the 'A' AND 'B' battery chargers lose their AC power source.

Answer:

D Question 24 Details Topic:

24 - 263000.K1.01 (2)

Comments:

Objective: DRE263LN003.12

Reference:

DOP 6800-02 K/A: 263000.K1.01 3.3/3.5 K/A: Knowledge of the physical connections and/or causeeffect relationships between D.C. ELECTRICAL DISTRIBUTION and the following: A.C. electrical distribution CFR: 41.7 Safety Function: 6 PRA: No Level: High Pedigree: Bank History: 2012 NRC Explanation: All 4 (2 positive and 2 negative) chargers are powered from the Instrument Bus. With Bus 25 OOS, MCC 25-2 has no power. With Bus 28 going overcurrent, MCC 28-2 loses power. Without these two MCCs, the Instrument Bus has no power, and subsequently all 4 chargers lose AC power.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 25 of 100 ILT 11-1 (NRC 2012-301) 25 Points: 1.00 A Unit 2 startup is in progress.

All four SRMs are fully inserted and are reading approximately 10,000 CPS.

Five minutes later the NSO observes:

SRM 21 is now reading approximately 5,000 cps and trending down SRMs 22, 23 and 24 are still indicating 10,000 cps All other reactor plant parameters are unchanged Which of the following operations / malfunctions would cause the observed indications?

A.

24/48 VDC Bus A voltage is trending down at 0.1 VDC per minute.

B.

SRM 21 detector has developed a gas leak.

C.

SRM 21 SELECT switch has been depressed and is lit.

D.

SRM 21 RAMP switch has been placed in the RAMP position.

Answer:

B Question 25 Details Topic:

25 - 215004.K6.04 Comments: Objective: DRE215LN004.02

Reference:

BWR Fundamentals Ch.7 Sensors/Detectors K/A: 215004.K6.04 2.9 / 2.9 K/A: Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRM) SYSTEM : Detectors CFR: 41.7 Safety Function: 7 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: BWR Fundamentals states: In Fission chambers, if the gas pressure decreases, there will be less gas atoms to undergo ionization, therefore less electrons collected. This means a smaller electrical current will exist and the indication will fail downscale.

a) Both SRM 21 and 22 would respond to a decreasing voltage.

b) A gas leak from an SRM detector will cause the SRM to read low.

c) A lit SELECT switch would allow the SRM to be withdrawn. But withdraw for the SRM is not a seal-in signal (insert is a seal-in signal) d) the RAMP switch being placed in RAMP would only be seen if the SRM MODE switch was also in PERIOD which would not allow the SRM to display reactor power.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 26 of 100 ILT 11-1 (NRC 2012-301) 26 Points: 1.00 Unit 2 is in STARTUP Drywell and Torus inerting are in progress.

A spurious Group 2 isolation occurs.

Which of the following describes the status of the Drywell and Torus inerting?

A.

Inerting of the Drywell and Torus remains in progress.

B.

Inerting of the Torus remains in progress ONLY C.

Inerting of the Drywell remains in progress ONLY.

D.

Inerting of the Drywell and Torus is no longer in progress.

Answer:

D Question 26 Details Topic:

26 - 223002.K3.28 Comments:

Objective: 223LN005.

Reference:

902(3)-5 E-5 K/A: 223002.K3.28 2.7 / 2.8 K/A: Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following: Containment nitrogen inerting system CFR: 41.7 Safety Function: 5 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: Upon receipt of a Group 2 signal, AO 2-1601-57, 58 and 59 will go closed. This will cut off the supply of nitrogen inerting the Torus and the Drywell.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 27 of 100 ILT 11-1 (NRC 2012-301) 27 Points: 1.00 Given the following:

Unit 2 is operating at full power.

CRD exercising is in progress.

Rod L7 is moved to position 48.

RPIS indication for rod L7 indicated '48', then RED '--', then went out.

What is this an indication of and what is(are) the NSO required actions?

A.

Rod overtravel; attempt to recouple the rod B.

Rod overtravel; bypass the RWM and drive the rod to 00 C.

Failed open RPIS reed switch; enter a substitute position D.

Failed open RPIS reed switch; bypass the RWM and drive the rod to 00 Answer:

A Question 27 Details Topic:

27 - 201003.K1.03 Comments:

Objective: 201LN002.08

Reference:

DOA 0300-05; LP DRE201LN002 K/A: 201003.K1.03 3.1 / 3.1 K/A: Knowledge of the physical connections and/or cause effect relationships between CONTROL ROD AND DRIVE MECHANISM and the following: RPIS CFR: 41.6 Safety Function: 1 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) Correct, DOA 0300-05 actions for uncoupled rod with reactor power greater than 10%

b) Incorrect, This is the action for an uncoupled rod if reactor power less than 10%

c & d) Incorrect, Open reed switches would cause RPS indication to be blank.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 28 of 100 ILT 11-1 (NRC 2012-301) 28 Points: 1.00 The following plant conditions exist:

Unit 2 is at 95% power 2C LPCI pump is operating for surveillance testing 2A and 2D CCSW pumps are operating for surveillance testing LPCI full flow test valves 2-1501-38B & 20B are positioned to yield 5000 gpm to the torus Then the following occurs:

MCC 28-1 trips on overcurrent.

Drywell pressure trends up to 2.5 psig due to a small steam leak.

Which one of the following describes the response of the LPCI system under these conditions?

A.

The 2C LPCI pump will trip Then the test line valves will close B.

The 2C LPCI pump will trip The test line valves will fail as is C.

The 2C LPCI pump will NOT trip The test line valves will close D.

The 2C LPCI pump will NOT trip The test line valves will fail as is Answer:

C Question 28 Details Topic:

28 - 219000.K2.01 Comments: Objective: 203LN001.06

Reference:

DAN 902(3)-5 D-11; DOP 6700-19 K/A: 219000.K2.01 2.5 / 2.9 K/A: RHR/LPCI: Torus/Suppression Pool Cooling Mode: Knowledge of electrical power supplies to the following: Valves CFR: 41.7 Safety Function: 5 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: High Drywell pressure (+2 psig) is a LPCI initiation signal - All four pumps will get a start signal, the CCSW pumps will receive a trip signal with high drywell pressure. A LPCI loop will be selected for injection to the RPV, and all test/containment cooling valves will get a close signal. 2-1501-38B & 20B are powered from MCC 29-4 and will close as designed. 2-1501-38A & 20A are powered from MCC 28-1 and would fail as is.

a) credible if examinee believes running LPCI pumps will trip initially (like the running CCSW pumps) common misconception.

b & d) credible if examinee believes MCC 28-1 is the power to the B subsystem valves.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 29 of 100 ILT 11-1 (NRC 2012-301) 29 Points: 1.00 Units 2 and 3 are operating at full power.

Due to ongoing work in the AEER, the 125 VDC Bus 2A-1 supply to the AEER halon fire panel is secured.

A condition requiring initiation of fire suppression equipment occurs.

Regarding AUTOMATIC fire suppression in the AEER, which of the following warnings (if any) would alert personnel working in the AEER?

A.

Flashing light and siren due to the automatic initiation of CO2 injection ONLY.

B.

Flashing light and siren due to the automatic initiation of Halon injection ONLY.

C.

Flashing light and siren due to the automatic initiation of Halon AND CO2 injection.

D.

NO warning, Halon and CO2 injection must be initiated manually.

Answer:

D Question 29 Details Topic:

29 - 286000.K3.02 Comments:

Objective: 286LN002.12

Reference:

LP DRE 286LN002 K/A: 286000.K3.02 3.2 / 3.4 K/A: Knowledge of the effect that a loss or malfunction of the FIRE PROTECTION SYSTEM will have on following: Personnel protection CFR: 41.7 Safety Function: 8 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

An automatic initiation of Halon or CO2 will will cause a flashing light and siren to activate in the area of the sensed fire for 60 seconds prior to injection.

All CO2 fire panels have redundant power sources, the AEER Halon panel only has a single power source.

CO2 systems only initiate automatically to the EDG rooms and must always be manually initiated in the AEER.

Securing 125 VDC power to the AEER halon panel, would inhibit automatic initiation and require Halon to be manually initiated.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 30 of 100 ILT 11-1 (NRC 2012-301) 30 Points: 1.00 The control rod reed switches are designed to illuminate a Full Core Display red DRIFT light under which of the following conditions?

No rod motion command present and...

A.

ONLY an odd reed switch closed.

B.

ONLY an even reed switch closed.

C.

an odd AND even reed switch closed.

D.

the ROD DRIFT TEST switch in TEST.

Answer:

A Question 30 Details Topic:

30 - 214000.K4.01 Comments:

Objective: 201LN002.06

Reference:

DAN 902(3)-5 A-3 K/A: 214000.K4.01 K/A: Knowledge of ROD POSITION INFORMATION SYSTEM design feature(s) and/or interlocks which provide for the following: Reed switch locations CFR: 41.7 Safety Function: 7 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC, 2009 Hope Creek Comments:

a) Correct - Per DAN 902(3)-5 A-3, ROD DRIFT, requires Control Rod moving past an ODD numbered reed switch with NO control rod motion requested.

b) Incorrect - requires odd reed switch without rod motion command signal.

c) Incorrect - requires odd reed switch without rod motion command signal.

d) Incorrect - rod motion command must be present and odd reed switch closed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 31 of 100 ILT 11-1 (NRC 2012-301) 31 Points: 1.00 Which of the following describes how the operation of the Backup Scram Solenoid Valves will result in a reactor scram?

When actuated, the Backup Scram Solenoid Valves...

A.

relieve hydraulic control unit drive pressure to the exhaust header.

B.

isolate and vent air downstream of the Scram Dump Valves.

C.

align the overpiston volume of the rod drive mechanism to the exhaust header.

D.

isolate and vent the air header to the Scram Inlet and Outlet Valves.

Answer:

D Question 31 Details Topic:

31 - 201001.K5.08 Comments:

Objective: 201LN001.06

Reference:

M-34 K/A: 201001.K5.08 2.5 / 2.6 K/A: Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM :

Solenoid operated valves CFR: 41.5 Safety Function: 1 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments:

For added protection, the Instrument Air header to all Scram Pilot Valves (2(3)-305-117 and 2(3)-305-118) has a pair of Backup Scram Valves (2(3)-302-19A and 2(3)-302-19B). When energized, the Backup Scram Valves isolate and vent the instrument air header to the Scram Inlet (2(3)-

305-126) and Scram Outlet Valves (2(3)-305-127) to rapidly insert all control rods.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 32 of 100 ILT 11-1 (NRC 2012-301) 32 Points: 1.00 Which of the following is indicative of debris from the Drywell being discharged through the downcomers during a small LOCA?

A.

fluctuating HPCI flow B.

fluctuating CCSW flow C.

fluctuating Torus spray flows D.

fluctuating ERV tail pipe pressures Answer:

C Question 32 Details Topic:

32 - 230000 K6.05 Comments:

Objective: DRE203LN001.12

Reference:

DAP 05-09 K/A: 230000 K6.05 K/A: Knowledge of the effect that a loss or malfunction of the following will have on the RHR/LPCI: TORUS/SUPPRESSION POOL SPRAY MODE :

Suppression pool CFR: 41.7 Safety Function: 5 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: Debris discharged through the downcomers would end up in the Torus and cause clogging of the ECCS suction strainers.

a) Incorrect - during a small LOCA, HPCI suction would be from the CSTs and would not be effected by debris in the Torus.

b) Incorrect - CCSW system flow would not be effected by debris in the torus. if debris from the torus were to make it to the LPCI/CCSW coolers, only LPCI flow would be effected.

c) Correct - LPCI provides torus sprays and its suction is aligned to the torus through the ECCS suction strainers.

d) Incorrect - While ERV tailpipes are in the torus and pressures do fluctuate during valve operation, it is not indicative of debris in the torus.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 33 of 100 ILT 11-1 (NRC 2012-301) 33 Points: 1.00 Unit 3 is conducting a special test with the Main Generator conditions as follows:

Real load is 800 MWe Terminal Voltage is 17.1 kV Reactive Load is 0 MVAR Voltage Regulator Mode Switch is in MANUAL If an operator placed the Voltage Regulator Mode Switch to AUTO, and the voltage regulator failed to its MINIMUM limit for the current voltage, which of the following indicates the final value of Main Generator reactive loading three minutes later (i.e. plant stable)?

(Assume terminal voltage remains constant.)

A.

0 MVAR B.

negative (-) 70 MVAR C.

negative (-) 120 MVAR D.

negative (-) 160 MVAR Answer:

B Question 33 Details Topic:

33 - 245000.A1.08 Comments:

Objective: 245LN003.11

Reference:

DOP 6400-08 K/A: 245000.A1.08 2.5 / 2.7 K/A: Ability to predict and/or monitor changes in parameters associated with operating the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS controls including: Generator output voltage/reactive load CFR: 41.5 Safety Function: 6 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments:The examinee must identify the correct Minimum Excitation Limiter (MEL) curve from the supplied references. The MEL limits the excitation dependent on the current terminal voltage of the generator. The point on the graph will move straight down the 800MWE line until it reaches the 17.1 Kv limit Distractors indicates a misunderstanding of system operation or inability to read the graph.

REQUIRED

REFERENCES:

Figures 5, 6 & 7 from DOP 6400-08

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 34 of 100 ILT 11-1 (NRC 2012-301) 34 Points: 1.00 Unit 2 is operating at 75% power.

The following events occur simultaneously:

2D ERV spuriously opens 2D ERV vacuum breaker fails open What is (1) the expected initial response of primary containment pressure and (2) the initial action taken to mitigate the event?

A.

(1) Torus pressure increase due to steam in the torus (2) Place the 2D ERV control switch in OFF B.

(1) Torus pressure increase due to steam in the torus (2) Initiate Torus sprays C.

(1) Drywell pressure increase due to steam in the drywell (2) Place the 2D ERV control switch in OFF D.

(1) Drywell pressure increase due to steam in the drywell (2) Initiate Drywell sprays Answer:

C Question 34 Details Topic:

34 - 223001.A2.09 Comments: Objective: 223LN001.02

Reference:

Dresden 2/3 PSTGs; DEOP 200-1 K/A: 223001.A2.09 3.4 / 3.6 K/A: Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Vacuum breaker malfunction CFR: 41.5 Safety Function: 5 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) Incorrect - With the vacuum breaker failed open, steam will be discharged to the drywell.

b) Incorrect - With the vacuum breaker failed open, steam will be discharged to the drywell.

c) Correct - With the vacuum breaker failed open, 2D ERV will discharge to the drywell and per DOA 0250-01, the correct immediate action is to place the 2D ERV control switch to off.

d) Incorrect - Drywell sprays is a action directed by the Unit Supervisor per DEOP 200-1 after drywell pressure has exceeded 9 psig.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 35 of 100 ILT 11-1 (NRC 2012-301) 35 Points: 1.00 Unit 2 Emergency Diesel Generator (U2 EDG) was running for a surveillance, when the following alarms are received in the Main Control Room:

902-8 C-8, U2 DIESEL ROOM TEMP HI XL3 23-23, 2 DG ROOM FIRE B1 Which of the following is an expected automatic action as a result of the alarms above?

A.

U2 EDG trips B.

Diesel fuel oil transfer pump trips C.

U2 EDG room vent fan louvers (first floor Turbine Building) trip D.

U2 EDG air intake modulator dampers (second floor Turbine Building) close Answer:

C Question 35 Details Topic:

35 - 288000.A3.01 Comments:

Objective: DRE288LN001.06

Reference:

DAN XL3 23-23 K/A: 288000.A3.01 3.8 / 3.8 K/A: Ability to monitor automatic operations of the PLANT VENTILATION SYSTEMS including: Isolation/initiation signals.

CFR: 41.7 Safety Function: 9 PRA: No Level: High Pedigree: New History: 2012 NRC Explanation:

a) Incorrect - High temp and fire are not trips of the U2 EDG b) Incorrect - The fuel oil transfer pump does not receive a trip signal from a fire condition.

c) Correct - Per DAN XL3 23-23, this indicates a fire and the room vent fan louvers trip automatically.

d) Incorrect - The air intake modulator damper does not receive a trip signal from a fire condition.

REQUIRED

REFERENCE:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 36 of 100 ILT 11-1 (NRC 2012-301) 36 Points: 1.00 Unit 2 was operating at 40% power with all APRMs in service.

Rod A-8 is selected.

A fault cause APRM channel 4 to fail down scale.

Based on these conditions, with no operator action, Rod Block Monitor (RBM) channel 8 A.

remains in service and reset B.

is bypassed C.

shifts to its alternate reference APRM D.

generates a RBM DOWNSCALE rod block Answer:

B Question 36 Details Topic:

36 - 215002.A4.03 Comments:

Objective: 215LN002.08

Reference:

LP DRE215LN002 K/A: 215002.A4.03 2.8 / 2.8 K/A: Ability to manually operate and/or monitor in the control room: Trip bypasses: BWR-3,4,5 CFR: 41.7 Safety Function: 7 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: Examinee must know that APRM 4 is the normal reference APRM for RBM channel 8 and A-8 is an edge rod.

a) would be correct if any other APRM were to fail downscale.

b) Correct, When APRM 4 fails down scale, RBM 8 sees reactor power less than 30%. Reactor power less than 30% causes automatic bypass of the RBM, An edge rod also causes an automatic bypass of the RBM.

c) RBM will shift to its alternate reference APRM only after its normal reference APRM is bypassed by an operator.

d) RBM DOWNSCALE rod block occurs when th LPRM inputs to a RBM channel average less than 7.1% with the reference APRM >30%.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 37 of 100 ILT 11-1 (NRC 2012-301) 37 Points: 1.00 Given the following:

Unit 2 Feed Water flow is 4.7 Mlbm/hr 2A AND 2B Reactor Feedpumps are in service 2A, 2B, AND 2C Condensate pumps are in service No Condensate pump is selected for standby due to standby selector switch repairs The 2C condensate pump trips Reactor Feedpump suction pressure drops to 110 psig Annunciator 902-6 G-8, RFP SEQUENTIAL TRIP SUCTION PRESS LO, comes in.

(1) Predict the response of the Feedwater pumps (2) What immediate operator actions must be taken?

A.

(1) One Reactor Feedpump trips immediately (2) Manually inhibit the Standby RFP from starting by placing STBY SELECT switch in OFF.

B.

(1) One Reactor Feedpump trips after 3 seconds (2) Start the 2D condensate pump and ensure RFP suction pressure recovers to greater than 120 psig.

C.

(1) One Reactor Feedpump trips immediately (2) Start the 2D condensate pump and ensure RFP suction pressure recovers to greater than 120 psig, then restart a second RFP.

D.

(1) One Reactor Feedpump trips after 3 seconds (2) Manually inhibit the Standby RFP from starting by placing STBY SELECT switch in OFF, then start the 2D condensate pump and ensure RFP suction pressure recovers to greater than 120 psig.

Answer:

B Question 37 Details Topic:

37 - 259001.G.2.4.31 Comments:

Objective: 259LN001.10

Reference:

DAN 902-6 G-8, 902-6 F-5, 902-6 F-7, DOP 3200-2 K/A: 259001.G.2.4.31 4.2 / 4.1 K/A: Reactor Feedwater System: Knowledge of annunciator alarms, indications, or response procedures.

CFR: 41.10 Safety Function: 2 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: These condition lead to a sequential trip of a reactor feedpump after a time delay. Per the DANs the standby condensate/condensate booster pump should be started and if insufficient RFP suction pressure is available for the standby RFP to start, Then manually inhibit the Standby RFP from starting by placing the STBY SELECT switch in OFF. A & C are incorrect because the RFP will not trip immediately unless suction pressure drops to 100 psig or less.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 38 of 100 ILT 11-1 (NRC 2012-301) 38 Points: 1.00 With Unit-2 at 95% power, the following transient is observed:

Rise in indicated core flow.

Drop in core thermal power.

Drop in Main Generator power.

Drop in Core differential pressure.

Which of the following is causing the above operating anomaly?

A.

Failure of a jet pump B.

Partial opening of an ERV C.

Failure of a condensate demineralizer post strainer D.

Loss of speed mismatch signal from an operating recirc pump Answer:

A Question 38 Details Topic:

38 - 202001.K5.02 Comments:

Objective: 20200LK016

Reference:

DOA 0201-01 K/A: 202001.K5.02 3.1 / 3.2 K/A: Knowledge of the operational implications of the following concepts as they apply to RECIRCULATION SYSTEM: Jet pump operation: BWR-3,4,5,6 CFR: 41.5 Safety Function: 1 & 4 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: DOA 0201-01 states the symptoms of a failed Jet Pump are:Rise in indicated core flow, drop in core thermal power, drop in main generator power and drop in core D/P. None of the distractors would have all of the symptoms listed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 39 of 100 ILT 11-1 (NRC 2012-301) 39 Points: 1.00 Unit 2 is operating at 75% power in DOME pressure control when 2 of the 3 EHC pressure controller processors (<R2> and <T2>) fail low.

As a result of this failure, turbine control valves will...

A.

close resulting in reactor scram on high pressure or flux.

B.

open and the reactor will scram due to MSIV closure on low reactor pressure.

C.

remain at current demand as pressure control transfers to THROTTLE active mode.

D.

open but the Max Combined Flow Limiter will prevent an excessive pressure reduction.

Answer:

A Question 39 Details Topic:

39 - 295005.K1.01 Comments:

Objective: 241LN001.08

Reference:

DOA 5650-02 K/A: 295005.AK1.01 4.0 / 4.1 K/A: Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP : Pressure effects on reactor power CFR: 41.8 Safety Function: 3 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: DOA 5650-01 states: IF 2 out of 3 DEHC Pressure Controller Processors fail low, THEN Turbine Control Valves AND Turbine Bypass Valves close, causing a reactor scram on high pressure. The triple modular redundancy (TMR) feature of DEHC requires 2 of the 3 pressure controller processors to be functioning in order for the system to work.

The failure indicated here would result in a loss this redundancy and thus the system will trip and close the control valves. Consequently, reactor pressure and flux will rise to the scram setpoint.

b) Incorrect - this would be the result if the control or bypass valves fail open.

c) Incorrect - this occurrs when dome exceeds throttle pressure by > 200 psig.

d) Incorrect - this would be the result if the pressure control regulator failed high.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 40 of 100 ILT 11-1 (NRC 2012-301) 40 Points: 1.00 Unit 2 is operating at full power.

When the following occurs:

A fire is reported in the 2/3 Cribhouse DOA 0010-10, FIRE-EXPLOSION, is entered.

Which of the following are immediate operator actions of DOA 0010-10 to initiate fire fighting actions?

Notify security to respond to the scene Notify Coal City Fire Protection District Announce fire location on plant radio system Announce fire location on plant PA system A.

1, 2, 3 B.

1, 2, 4 C.

1, 3, 4 D.

2, 3, 4 Answer:

C Question 40 Details Topic:

40 - 600000.AK1.02 Comments:

Objective: 29501LK080

Reference:

DOA 0010-10 K/A: 600000.AK1.02 2.9 / 3.1 K/A: Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire Fighting CFR: 41.10 Safety Function: 8 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a, c & d) Incorrect, Notify Coal City FPD is an immediate action action of DOA 0010-10 only if the fire is located outside the protected area. The 2/3 Cribhouse is located inside the protected area.

b) Correct, Per DOA 0010-10, Notify security to respond to scene, Announce fire location on plant radio and Announce fire location on plant PA are immediate actions if the fire is located in the protected area.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 41 of 100 ILT 11-1 (NRC 2012-301) 41 Points: 1.00 DEOP 400-5 "Failure to Scram" allows the use of the Core Spray system as an Alternate System, while HPCI and LPCI are preferred systems.

What is the reason for restricting injection with Core Spray?

A.

It would cool the SBLC solution enough to cause excessive boron precipitation.

B.

It would cause boron to concentrate in the upper core region, thereby lowering its effectiveness at reducing core power.

C.

It would flash to steam in the core region pushing boron out of the core.

D.

It would displace boron in the entire core region, replacing it with cold, unborated water.

Answer:

D Question 41 Details Topic:

41 - 295037.EK1.03 Comments:

Objective: 29502LK033

Reference:

DEOP 400-5 bases K/A: 295037.EK1.03 4.2 / 4.4 K/A: Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Boron effects on reactor power (SBLC)

CFR: 41.8 Safety Function: 1 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Per DEOP 400-5 bases, spraying core spray would remove boron from the entire core region, replacing it with cold, unborated water, resulting in power excursion problems.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 42 of 100 ILT 11-1 (NRC 2012-301) 42 Points: 1.00 Unit 2 was operating at near rated power, when a LOCA occurred.

If RPV pressure is 100 psig and lowering slowly, which of the following set of parameters would result in a USEABLE indicated wide range RPV water level?

Drywell temperature of ___(1)___ and an indicated wide range RPV water level of ___(2)___.

A.

(1) 257 oF (2) -19 inches B.

(1) 257 oF (2) -23 inches C.

(1) 357 oF (2) +12 inches D.

(1) 357 oF (2) -11 inches Answer:

A Question 42 Details Topic:

42 - 295028.K2.03 Comments:

Objective: DRE216LN001.12

Reference:

DEOP 200 figures A, B, and C, (RPV Water Level Instruments)

K/A: 295028.K2.03 3.6 / 3.8 K/A: Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Reactor water level indication.

CFR: 41.7 Safety Function: 5 PRA: No Level: High Pedigree: Bank History: 2012 NRC; 2010 NRC Explanation: The examinee must utilize both the RPV saturation temperature and minimum indicated level tables of the above DEOP to determine that the only values that would result in a useable indication is a Drywell temperature of 257oF and a RPV level of -19 inches.

REQUIRED

REFERENCES:

DEOP 200 Figures A, B, and C, (RPV Water Level Instruments).

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 43 of 100 ILT 11-1 (NRC 2012-301) 43 Points: 1.00 A transient has occurred resulting in the following conditions:

Drywell pressure 42 psig and stable.

Torus bottom pressure is 55 psig and stable.

Primary containment water level is 27 feet and slowly increasing.

RPV level -173" and slowly decreasing.

The Unit Supervisor has directed venting of the primary containment to lower pressure.

Which of the following vent paths should the NSO align first?

A.

torus via the APCV system B.

torus via the SBGT system C.

drywell via the APCV system D.

drywell via the SBGT system Answer:

B Question 43 Details Topic:

43 - 295024.EK2.18 Comments:

Objective: 29502LK063

Reference:

DEOP 0500-04 K/A: 295024.EK2.18 3.3 / 3.4 K/A: Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: Ventilation CFR: 41.7 Safety Function: 5 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments:

a) Incorrect because per the UFSAR drywell design pressure is 62 psig, 59 psig does not require venting through APCV b) Correct, per DEOP 500-4, with the conditions given, the correct vent path is the torus via SBGT.

c) & d) Incorrect, If torus level is greater than 30 feet, venting from the drywell is directed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 44 of 100 ILT 11-1 (NRC 2012-301) 44 Points: 1.00 Unit 2 was operating at near rated power, with the following conditions:

The Unit 2A 125 VDC Battery Charger is supplying the Unit 2 125 VDC system.

The Unit 2/3 250 VDC Battery Charger is supplying the Unit 2 250 VDC system.

Then a fire caused Bus 28 to de-energize.

(1) What effect will this have on the batteries AND (2) what actions can be taken to mitigate the transient?

A.

(1) The Unit 2 125 VDC batteries ONLY will begin to discharge; (2) place the Unit 2 125 VDC Battery Charger in service.

B.

(1) The Unit 2 250 VDC batteries ONLY will begin to discharge; (2) place the Unit 2 250 VDC Battery Charger in service.

C.

(1) The Unit 2 125 AND 250 VDC batteries will begin to discharge; (2) place the Unit 2 125 AND Unit 2 250 VDC Battery Chargers in service.

D.

(1) The Unit 2 125 AND 250 VDC batteries will begin to discharge; (2) place the Unit 2 125 Alternate Battery AND Unit 2 250 VDC Battery Charger in service.

Answer:

A Question 44 Details Topic:

44 - 295004.K2.01 Comments:

Objective: DRE263LN001.02

Reference:

DOA 6900-04, DOA 6900-T1 K/A: 295004.AK2.01 3.1 / 3.1 K/A: Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: Battery charger CFR: 41.7 Safety Function: 6 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: Explanation: With the 2A charger (MCC 28-2) supplying the 125 VDC system and the 2/3 charger (MCC 29-2 or 39-2) supplying the 250 VDC system, a loss of Bus 28 will cause ONLY the 125 VDC 2A charger to lose power, which will cause the 125 VDC batteries to begin to discharge. Transferring the 125 VDC system to the Unit 2 125 VDC charger is the correct action.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 45 of 100 ILT 11-1 (NRC 2012-301) 45 Points: 1.00 Given the following conditions:

Unit 2 is at full power.

2A and 2B Instrument Air Compressors (IAC) are running.

2A IAC trips due to low lube oil pressure.

Unit 2 Instrument Air pressure is 78 psig and dropping slowly.

An operator has been dispatched to start and align 3C IAC to Unit 2.

The reason for this action is to prevent the __________.

A.

FRVs from failing closed B.

outboard MSIVs from drifting closed C.

Target Rock from becoming inoperable in ERV mode D.

Unit 2 Instrument Air to Unit 2 Service Air cross tie from opening Answer:

B Question 45 Details Topic:

45 - 295019.AK3.02 Comments:

Objective: 278LN001.12

Reference:

DOA 4700-01, LP DRE239LN001 K/A: 295019.AK3.02 3.5 / 3.4 K/A: Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Standby air compressor operation CFR: 41.5 Safety Function: 8 PRA: No Level: High Pedigree: New History:

Comments:

a) Incorrect - The FRVs lock up on a loss of air pressure.

b) Correct - A reactor scram and manually closing the MSIVs is required on a loss of instrument air due to the potential of the MSIVs drifting closed.

c) Incorrect - The target rock would be inop in ERV mode with a loss of Drywell Pneumatics.

d) Incorrect - The Unit 2 IA to Unit 2 Service Air cross-tie would already be open (85 psig).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 46 of 100 ILT 11-1 (NRC 2012-301) 46 Points: 1.00 Unit 2 and Unit 3 are operating at near rated power when Transmission Systems Operations (TSO) notifies the Control Room that the predicted post Unit trip with LOCA switchyard voltages are:

Unit 2: 325 KV Unit 3: 350 KV What are the required actions from the Operating team AND the reason for these actions?

A.

Adjust TR 32 Tap Changer; to restore system operability B.

Adjust TR 32 Tap Changer; to reduce circulating currents C.

Adjust TR 86 Tap Changer; to restore system operability D.

Adjust TR 86 Tap Changer; to reduce circulating currents Answer:

C Question 46 Details Topic:

46 - 700000.K3.02 Comments:

Objective: DRE262LN003.12

Reference:

DOA 6500-12 K/A: 700000.K3.02 3.6 / 3.9 K/A: Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

Actions contained in abnormal operating procedure for voltage and grid disturbances.

CFR: 41.10 Safety Function: 7 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC Explanation: Given the voltages, only Unit 2 is below the procedural setpoint. TR-32 would be adjusted for Unit 3; TR-86 would be adjusted for Unit 2. The actions are to adjust the TR-86 Tap Changer (to position 31) to raise VOLTS (not VARs) to restore system operability.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 47 of 100 ILT 11-1 (NRC 2012-301) 47 Points: 1.00 DEOP 0400-03 STEAM COOLING attempts to prevent peak cladding temperature from exceeding ____(1)____ WITHOUT makeup. This prevents exceeding the threshold for

____(2)____.

A.

(1) 1500°F (2) fuel rod perforation B.

(1) 1500°F (2) significant metal-water reaction C.

(1) 1800°F (2) fuel rod perforation D.

(1) 1800°F (2) significant metal-water reaction Answer:

D Question 47 Details Topic:

47 - 295031.K3.04 Comments:

Objective: 29501LK012

Reference:

DEOP 0010-00 K/A: 295031.K3.04 4.0 / 4.3 K/A: Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL: Steam cooling CFR: 41.5 Safety Function: 2 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: DEOP 0010-00 states: Steam cooling without makeup capability is employed in DEOP 0400-03 during steam cooling. With no makeup to the RPV, adequate steam flow exists as long as RPV water level remains above the Minimum Zero-Injection RPV Water Level.

When RPV water level drops below this elevation, emergency depressurization must be performed. The peak clad temperature is permitted to rise to 1800°F, the threshold for significant metal-water reaction, to maximize the heat transfer to steam and to delay the depressurization as long as possible. 1500°F, the threshold for fuel rod perforation, is the limit when injection is available.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 48 of 100 ILT 11-1 (NRC 2012-301) 48 Points: 1.00 Unit 2 is operating at full power.

A transient occurs resulting in the following conditions:

Reactor pressure spikes to 1260 psig.

Reactor water level rapidly drops to -40 inches.

Which of the following describes the expected response of the Recirc pumps?

A.

Recirc pumps trip immediately due to low reactor water level.

B.

Recirc pumps trip after 9 seconds due to low reactor water level.

C.

Recirc pumps trip immediately due to high reactor pressure.

D.

Recirc pumps trip after 9 seconds due to high reactor pressure.

Answer:

C Question 48 Details Topic:

48 - 295025.EA1.07 Comments:

Objective: 212LN002.06

Reference:

DAN 902(3)-5 A-8; 12E-6582F K/A: 295025.EA1.07 4.1 / 4.1 K/A: Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: ARI/RPT/ATWS: Plant-Specific CFR: 41.7 Safety Function: 3 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a & b) Incorrect, reactor water level has not yet reached the low level trip of -50.82" and only ARI initiates immediately on low reactor level.

c) Correct, reactor pressure >1211 psig will cause the recirc pumps to trip immediately.

d) Incorrect, Common misconception that the 9 second time delay is in effect with the low reactor water level trip of the recirc pumps.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 49 of 100 ILT 11-1 (NRC 2012-301) 49 Points: 1.00 An effluent offsite release is in progress.

Indications on the 923-7 panel have failed.

Besides the 923-7 panel, where in the Main Control Room can the effluent release rate be directly monitored?

A.

902-2 Panel B.

902-10 Panel C.

PPDS Display D.

SPDS Display Answer:

C Question 49 Details Topic:

49 - 295038.EA1.01 Comments:

Objective: 272LN004.10

Reference:

DAN 923-7 B-3 K/A: 295038.EA1.01 K/A: Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE : Stack-gas monitoring system: Plant-Specific CFR: 41.7 Safety Function: 9 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Incorrect - The 902-2 panel has the service water rad monitors b) Incorrect - The 902-10 panel has off-gas rad monitors c) Correct - The PPDS display shows total effluent release rate in the control room d) Incorrect - The SPDS display shows if there is a release in progress with a status box that changes color but not a value for rate.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 50 of 100 ILT 11-1 (NRC 2012-301) 50 Points: 1.00 Unit 3 was operating at rated power with the 2/3 EDG OOS.

A severe fire has occurred in the AEER concurrent with a loss of offsite power.

The order to evacuate the control room has been given.

As the Unit 3 NSO, in-plant actions you should perform include verifying...

A.

Bus 33-1 CUB 3, MAIN FEED FROM BUS 33, breaker OPEN.

B.

Bus 33-1 CUB 5: 3-7338, FEED TO 480V SWGR 38 (RB), breaker OPEN C.

Bus 38 CUB 4A: 3-7838-1/4, 480V REACTOR BUILDING MCC 38-1 & 38-4, breaker OPEN D.

Bus 33 CUB 10, 3-6733-3, BUS 33-1 FEED, breaker CLOSED.

Answer:

A Question 50 Details Topic:

50 - 295016.A1.04 (2)

Comments:

Objective: 29501LP083

Reference:

DSSP 0100-CR K/A: 295016.A1.04 3.1 / 3.2 K/A: Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT: A.C. electrical distribution CFR: 41.7 Safety Function: 7 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments:

a) Correct, DSSP 100-CR attachment B (U3 NSO ACTIONS) requires this breaker OPEN b) Incorrect, DSSP 100-CR attachment B (U3 NSO ACTIONS) requires this breaker CLOSED not OPEN.

c) Incorrect, DSSP 100-CR attachment B (U3 NSO ACTIONS) requires this breaker CLOSED not OPEN.

d) Incorrect, this is a DSSP 100-CR attachment E (U2 US OR DESIGNEE ACTIONS) requirement.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 51 of 100 ILT 11-1 (NRC 2012-301) 51 Points: 1.00 During refueling operations a mechanical failure in the mast results in an irradiated bundle dropping onto the fuel racks and being damaged.

Which one of the following will increase resultant refuel floor contamination levels?

A.

Fuel pool temperature being at the Tech Spec LOWER limit.

B.

Fuel pool temperature being at the Tech Spec UPPER limit.

C.

Fuel pool level being at the Tech Spec LOWER limit.

D.

Fuel Pool level being at the UPPER limit.

Answer:

B Question 51 Details Topic:

51 - 295023.AA2.03 Comments:

Objective: 233LN001.11

Reference:

DAN 902(3)-4 A-23, DRE LP233LN001 K/A: 295023.AA2.03 3.3 / 3.8 K/A: Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS : Airborne contamination levels CFR: 41.10 Safety Function: 8 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments:

a) With fuel pool temperature low in the band, the concern is dose rate from the spent fuel increasing.

b) Correct, the concern with high fuel pool temperature is iodine coming out of solution and increasing airborne contamination levels c) A low fuel pool level would result in elevated dose rates on the refuel floor.

d) High fuel pool level is a concern with the fuel pool overflowing into the vent ducts of the RB ventilation system.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 52 of 100 ILT 11-1 (NRC 2012-301) 52 Points: 1.00 Unit 2 was operating at near rated power when BOTH Recirc Pumps tripped.

Prior to the transient, RPV water level was ___(1)___ than INDICATED Fuel Zone RPV water level AND the difference will get ___(2)___ following the transient.

A.

(1) lower (2) smaller B.

(1) lower (2) larger C.

(1) higher (2) smaller D.

(1) higher (2) larger Answer:

A Question 52 Details Topic:

52 - 295001.AA2.06 Comments:

Objective: DRE216LN001.06

Reference:

LP DRE216LN001, DOP 1000-03 K/A: 295001.AA2.06 3.2 / 3.3 K/A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Nuclear boiler instrumentation CFR: 41.10 Safety Function: 1 & 4 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Fuel zone instruments range from -340 to +60 inches. During normal operation at power with reactor level +30 inches, the Fuel Zone instruments are pegged high. Fuel Zone indications are affected by anything that causes flow through the monitored jet pump. Therefore recirc pump, LPCI pump or Shutdown Cooling pump flow causes the Fuel Zone instruments to be inaccurate in the non-conservative direction (read higher than actual).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 53 of 100 ILT 11-1 (NRC 2012-301) 53 Points: 1.00 The 250VDC and 125VDC systems are aligned as follows:

The Unit 3 250VDC Charger is OOS for maintenance.

The 2/3 250VDC Charger is being supplied from Unit 3 and is on the Unit 3 250VDC system floating the battery.

The Unit 3 125 VDC Charger 3A is on the Unit 3 125VDC system floating the battery.

The Unit 3 EO reports a fire in MCC 39-2 and MCC 39-2 has been deenergized.

Which of the following describes the effect (if any) this will have on the Unit 3 250VDC and 125VDC batteries?

A.

The Unit 3 250VDC battery would be unaffected and the Unit 3 125VDC battery would be unaffected.

B.

The Unit 3 250VDC battery would be unaffected and the Unit 3 125VDC battery would be the only source of power to the Unit 3 125VDC system.

C.

The Unit 3 250VDC battery would be the only source of power to the Unit 3 250VDC system and the Unit 3 125VDC battery would be unaffected.

D.

The Unit 3 250VDC battery would be the only source of power to the Unit 3 250VDC system and the Unit 3 125VDC battery would be the only source of power to the Unit 3 125VDC system.

Answer:

C Question 53 Details Topic:

53 - 295003.AA2.03 Comments:

Objective: 263LN002.02, 263LN001.02

Reference:

DOA 6900-02, DOA 6900-04 K/A: 295003.AA2.03 K/A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER : Battery status:

Plant-Specific CFR: 41.10 Safety Function: 6 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: MCC 39-2 provides power to the 125 VDC battery charger #

3 and Unit 2/3 250 VDC battery charger. MCC 38-2 provides power to the 125 VDC battery charger # 3A. MCC 38-3 provides power to the Unit 3 250 VDC battery charger (which is OOS in the stem of the question).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 54 of 100 ILT 11-1 (NRC 2012-301) 54 Points: 1.00 Unit 2 is operating at near rated power, with the following annunciators illuminated:

902-4 B-18 DIV 1 TORUS WTR BULK TEMP HI 902-4 C-18 DIV 2 TORUS WTR BULK TEMP HI 902-3 C-13 2B ELECTROMATIC RELIEF VLV OPEN Torus Bulk temperature is 95oF and rising.

Per DOA 0250-01, RELIEF VALVE FAILURE, what IMMEDIATE actions must be performed to mitigate Torus heatup?

A.

Initiate Torus cooling and verify Torus temperature stabilizes.

B.

Cycle ADS INHIBIT switch from NORMAL to INHIBIT to NORMAL and verify reactor pressure stable.

C.

Place 2B relief valve control switch in OFF and verify feedwater control system has stabilized reactor level.

D.

Cycle 2B relief valve control switch to MANUAL and then to OFF and verify turbine control system maintains reactor pressure stable.

Answer:

C Question 54 Details Topic:

54 - 295026.G.2.4.49 Comments:

Objective: DRE223LN001.08

Reference:

DOA 250-1, DANs 902-4 B-18 and C-18, DAN 902-3 C-13 K/A: 295026.G.2.4.49 4.6 / 4.4 K/A: Suppression Pool High Water Temperature: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

CFR: 41.10 Safety Function: 5 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC, 2001 NRC Comments: With Torus temp rising (95°F DEOP entry) and indication of 2B ERV open, entry into DOA 0250-1 and DEOP 0200-1 is required.

Immediate actions taken without reference are to place relief valve control switch in OFF and verify feedwater control system has stabilized reactor level.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 55 of 100 ILT 11-1 (NRC 2012-301) 55 Points: 1.00 Unit 3 was operating at near rated power when a spurious scram occurred.

The Shift Manager has determined that a hot fast restart is to be performed.

Which of the following equipment condition(s) needs to be met to allow the Reactor Mode Switch to be placed in the REFUEL position to perform Reactor Mode Switch Interlock testing?

A.

Moderator temperature < 350oF.

B.

All Control rods are fully inserted.

C.

SRM Rod Block Functional Test complete.

D.

Refuel Position One Rod Out Interlock met.

Answer:

B Question 55 Details Topic:

55 - 295006.G.2.1.23 Comments:

Objective: 20102LK004

Reference:

DGP 2-3; DOS 0700-01; DOS 0700-05; Tech Spec 3.10.1 K/A: 295006.G.2.1.23 4.3 / 4.4 K/A: SCRAM: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

CFR: 41.10 Safety Function: 1 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC, 2008 NRC Comments: Per T.S. 3.10.1, with all control rods fully inserted and no core alterations in progress, the Reactor Mode Switch may be placed in any position for testing purposes without exiting Mode 3.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 56 of 100 ILT 11-1 (NRC 2012-301) 56 Points: 1.00 Unit 2 is in a refueling outage.

The reactor vessel head bolts are tensioned.

Reactor water level is just below the flange.

Reactor vessel water temperature is 170°F and steady.

2A and 2B Recirculation pumps are secured.

2A and 2B SDC systems are in service cooling the reactor vessel.

2C SDC pump is Out of Service (O.O.S.) for maintenance.

A fault causes 2B SDC pump to trip on over current.

Which of the following describe the MINIMUM required Tech Spec actions to be taken within one hour?

A.

Monitor reactor coolant temperature and pressure.

B.

Verify reactor coolant circulating by an alternate method.

C.

Restore required SDC subsystem(s) to OPERABLE status.

D.

Verify an alternate method of decay heat removal is available for each inoperable required SDC subsystem.

Answer:

D Question 56 Details Topic:

56 - 295021.G.2.2.39 Comments: Objective: 205LN001.11

Reference:

T.S. 3.4.8 K/A: 295021.G.2.2.39 K/A: Loss of shutdown cooling: Knowledge of less than or equal to one hour Technical Specification action statements for systems.

CFR: 41.10 Safety Function: 4 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: Examinee must first determine from the initial conditions that T.S.3.4.8 SDC - Cold Shutdown applies. And a loss of Bus 23-1 causes 2A and 2C SDC pumps to lose power.

a) Would apply if NO SDC subsystems were in operation.

b) Would apply if NO SDC subsystems were in operation.

c) Would apply if the Unit was in Hot Shutdown.

d) Correct, per T.S. 3.4.8, verify an alternate method of decay heat removal is available for each inoperable required SDC subsystem.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 57 of 100 ILT 11-1 (NRC 2012-301) 57 Points: 1.00 Given the following plant conditions on Unit 2:

Drywell pressure is 2.5 psig and rising slowly Reactor water level is 90 inches and dropping slowly HPCI system automatically initiated and is running at 3,800 rpm, injecting 5,600 gpm into the vessel CST level has decreased, but the automatic suction transfer for HPCI has failed Annunciator 902-3 A-11, 'HPCI Booster Pump Suction Pressure Low' has just alarmed Which of the following describes the response of the HPCI system under these conditions?

A.

HPCI turbine will trip on low pump suction pressure B.

HPCI turbine speed will rise and then HPCI pump flow will eventually drop to zero C.

HPCI turbine speed will remain constant and HPCI pump flow will eventually drop to zero D.

HPCI turbine speed will decrease until the MGU reaches the low speed stop Answer:

B Question 57 Details Topic:

57 - 295030.EA1.05 Comments:

Objective: DRE206LN001.08

Reference:

LP DRE206LN001 K/A: 295030.EA1.05 3.5 / 3.5 K/A: Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCI CFR: 41.7 Safety Function: 5 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: The Booster Pump low suction pressure trip is bypassed with an initiation signal present. The turbine speed control system CANNOT respond to turbine speed increases because it is positioned by a FLOW controller. As flow falls off, the MGU would go to the high speed stop but with no water to pump, HPCI will continue to operate (with the MGU at the HSS - NOT 4000 rpm but full steam power) until the turbine overspeeds.

Turbine speed would remain constant if HPCI had been started manually.

Decreasing turbine speed until the MGU reaches the low speed stop is a condition HPCI is placed in during manual startup and shutdown.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 58 of 100 ILT 11-1 (NRC 2012-301) 58 Points: 1.00 With both units operating at full power, a service water pump trip results in the following conditions:

Service water header pressure 78 psig dropping slowly.

X-area temperature 147°F and rising slowly.

Alternator air temperature 52°C and steady.

Stator cooling water outlet temperature 75oC and rising slowly.

HPCI area temperature 140°F and rising slowly.

Which one of the following actions is appropriate for these conditions?

A.

Immediately reduce load to < 38.5% and trip the main turbine.

B.

Enter DEOP 300-1 and take action to mitigate the high area temperature conditions.

C.

Reduce load per DGP 3-1 only enough to compensate for changes in system temperatures.

D.

Verify the 2/3 emergency fire pump starts and then cross-connect the fire water and service water systems.

Answer:

A Question 58 Details Topic:

58 - 295018.AA2.01 Comments:

Objective: 276LN001.08

Reference:

DOA 3900-01 K/A: 295018.AA2.01 3.3 / 3.4 K/A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Component temperatures CFR: 41.10 Safety Function: 8 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Actions per DOA 3900-01. If alternator air temperature exceeds 48°C, then the DOA directs load reduced and the main generator tripped. HPCI area temperature is below the DEOP 300-1 entry condition. Cross-connecting fire water and service water would only be done to supplement the fire protection system. Reducing load only enough to compensate for high system temperatures is NOT sufficient action based on the conditions given.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 59 of 100 ILT 11-1 (NRC 2012-301) 59 Points: 1.00 Unit 2 was operating at full power when a transient occurred resulting in the following conditions:

Loss of Offsite Power (LOOP)

Reactor scram Reactor pressure is 50 psig All scram valves are open Rod H-8 indicates position 46, all other rods indicate full in Which of the following actions (if any) would be successful in inserting H-8?

A.

Vent the scram air header.

B.

Vent the CRD overpiston area.

C.

Scram H-8 using its individual scram toggle switch.

D.

H-8 can not be inserted until power has been restored to a CRD pump.

Answer:

D Question 59 Details Topic:

59 - 295015.AK1.04 Comments:

Objective: 29501LK031

Reference:

DOA 0300-01 K/A: 295015.AK1.04 K/A: Knowledge of the operational implications of the following concepts as they apply to INCOMPLETE SCRAM : Reactor pressure: Plant-Specific CFR: 41.8 Safety Function: 1 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

Examinee is required to know that a a LOOP would result in a loss of all CRD pumps. With no CRD pumps running, the charging water header would depressurize. If the charging water header pressure is lost, the CRD accumulator will depressurize as a result of leakage through the 0305-115 accumulator check valve/0305-126 scram inlet valves (which are open) resulting in the potential inability to scram the drive at low reactor pressure.

a, b & c) Incorrect - all of these are alternate means of rod insertion but would be ineffective because no motive force exists to insert the rod under these conditions.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 60 of 100 ILT 11-1 (NRC 2012-301) 60 Points: 1.00 Unit 3 was operating at near rated power, when the following occurred:

A leaking valve has caused water to be sprayed on 'A' SBGT Charcoal Bed.

A fuel handling accident occurred, causing radiation levels at the site boundary to increase.

An NSO isolated Reactor Building ventilation and started the 'A' SBGT train.

The release rates at the site boundary will be higher than anticipated for which of the following?

A.

Iodine B.

Particulates C.

Transuranics D.

Noble Gases Answer:

A Question 60 Details Topic:

60 - 295017.AK2.12 Comments:

Objective: 27800LK027

Reference:

UFSAR Sec.6.5 K/A 295017.AK2.12 3.4 / 3.7 K/A: Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Standby gas treatment/FRVS CFR: 41.7 Safety Function: 9 PRA: No Level: High Pedigree: Bank History: 2007 NRC; 2012 NRC Explanation: The charcoal bed normally removes the Iodine, but due to being wet the efficiency of removal goes down. Transuranics are released with fuel failure but do not go airborne. Particulates are removed by the HEPA filters. Noble Gases have little hold up and would not be significantly changed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 61 of 100 ILT 11-1 (NRC 2012-301) 61 Points: 1.00 Unit 3 is operating at rated conditions.

A loss of both CRD pumps has occurred.

This condition will result in ________.

A.

degradation of the CRDM seals due to loss of cooling flow B.

loss of scram capability due to loss of charging water flow C.

loss of Unit 3 reactor water level indication due to loss of RVWLIS D.

failure of Reactor Recirc pump seals due to loss of seal cooling flow Answer:

A Question 61 Details Topic:

61 - 295022.AK3.02 Comments: Objective: 201LN003.12

Reference:

DRE LP201LN003 K/A: 295022.AK3.02 2.9 / 3.1 K/A: Knowledge of the reasons for the following responses as they apply to LOSS OF CRD PUMPS: CRDM high temperature CFR: 41.5 Safety Function: 1 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments:

a) At elevated temperatures the CRDM seals become brittle and can result in breakdown and increased scram times.

b) Incorrect - CRD accumulators maintain scram capability upon a loss of charging water flow c) Incorrect - Loss of RVWLIS will not effect reactor water level indication during normal operations d) Incorrect - seal cooling to the Recirc pumps is provided by RBCCW.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 62 of 100 ILT 11-1 (NRC 2012-301) 62 Points: 1.00 Unit 2 was operating at near rated power, when a transient occurred, and the following conditions exist:

Drywell pressure is 3.0 psig and steady.

Reactor pressure is 950 psig and steady.

RWCU blowdown rate is 200 gpm and steady.

RWCU recirculation and blowdown modes are being used to control RPV pressure.

RPV water level is 12 inches and lowering at rate of 1 inch/minute with all available systems operating.

Which of the following action would mitigate the conditions listed above?

A.

Secure blowdown AND recirculation modes.

B.

Secure blowdown mode AND maintain recirculation mode.

C.

Maintain existing blowdown AND recirculation modes.

D.

Maintain existing blowdown mode AND secure recirculation mode.

Answer:

B Question 62 Details Topic:

62 - 295009.AA1.04 Comments:

Objective: 204LN001.08

Reference:

DEOP 0100-00 K/A: 295009.AA1.04 K/A: Ability to operate and/or monitor the following as they apply to LOW REACTOR WATER LEVEL : Reactor water cleanup CFR: 41.7 Safety Function: 2 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: Securing blowdown will decrease the amount of water leaving the reactor, which mitigates the lowering RPV water level. DEOP 100 allows installing jumpers to maintain RWCU Recirc mode for pressure control.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 63 of 100 ILT 11-1 (NRC 2012-301) 63 Points: 1.00 Unit 2 was operating at full power when a spurious Group I (one) occurred.

10 seconds later the following indications exist:

2-0203-3B and 2-0203-3C ERVs indicate open on the 902-3 and 902-21 panel, all other ERVs indicate closed.

The Isolation Condenser is in a normal standby lineup.

Which of the following would indicate the HIGHEST reactor pressure reached during this event based on the above indications.

A.

1065 psig B.

1095 psig C.

1115 psig D.

1135 psig Answer:

C Question 63 Details Topic:

63 - 295020.AA2.04 Comments:

Objective: 239LN001.06

Reference:

DRE LP239LN001 K/A: 295020.AA2.04 3.9 / 3.9 K/A: Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION : Reactor pressure CFR: 41.10 Safety Function: 5 & 7 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments:

a) Incorrect, Only a reactor scram would have occurred at 1060 psig b) Incorrect, Isolation Condenser would auto initiate if greater than 1070 for more than 17 seconds c) Correct, 1110 psig is the setpoint for the low set ERVs (B and C ERVs) d) Incorrect, 1133 psig is the setpoint for the high set ERVs (D and E ERVs and the Target Rock)

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 64 of 100 ILT 11-1 (NRC 2012-301) 64 Points: 1.00 Unit 2 was operating at full power when the following events occurred:

The D Main steam line breaks in the steam tunnel.

Loss of the Instrument Bus due to a short circuit.

An NSO reports the following valves open:

RX RETURN VLV, 2-1201-7 2B FW REG ISOL, 2-3206B MSIV, 2-0203-1A Which of the following describes the status of the PCIS status box in the Safety Parameter Display System (SPDS)?

A.

GREEN B.

YELLOW C.

RED D.

CYAN Answer:

C Question 64 Details Topic:

64 - 295032.G.2.1.19 Comments: Objective: 283LN002.06

Reference:

DOP 9950-17 K/A: 295032.G.2.1.19 K/A: High Secondary Containment Area Temperature: Ability to use plant computers to evaluate system or component status.

CFR: 41.10 Safety Function: 5 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) The PCIS status box is GREEN when a PCIS isolation signal is present and all required PCIS isolation valves are shut.

b) Yellow is a pre-alarm indication on PCIS used by the parameters with analog values on SPDS.

c) Correct, A steam line break would cause a Group 1 PCIS isolation due to high secondary containment temperature in the steam tunnel. The PCIS status box is RED when a PCIS isolation signal is present and all required PCIS isolation valves are not shut. All MSIVs are required to be shut when a Group 1 signal is received.

d) Would be correct if the status of the PCIS valves were undeterminable.

A loss of the Instrument Bus would not affect the ability of the SPDS to determine valve position even though some of the PCIS valves would lose remote indication.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 65 of 100 ILT 11-1 (NRC 2012-301) 65 Points: 1.00 Unit 3 was at rated conditions when a transient occurred.

An Isolation Condenser steam leak occurred and was isolated.

Isolation Condenser area temperature is 170ºF and is too high for personnel access.

Valid Reactor Building Ventilation isolations are present on each of the following parameters:

Drywell Pressure Reactor Building Exhaust Radiation Reactor Water Level Restarting the Reactor Building Ventilation would allow safer access to the Isolation Condenser area ________.

A.

but is NOT allowed due to the Drywell Pressure isolation B.

but is NOT allowed due to the Reactor Water Level isolation C.

and may be performed after bypassing the isolation signals D.

but is NOT allowed due to the Reactor Building Exhaust Radiation isolation Answer:

D Question 65 Details Topic:

65 - 295034.K1.01 Comments:

Objective: 29502LK050

Reference:

LP 295L-S3 and DEOP 300-1 K/A: 295034.K1.01 3.8 / 4.1 K/A: Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION : Personnel protection CFR: 41.9 Safety Function: 9 PRA: Yes Pedigree: Bank Level: Memory History: 2012 NRC Explanation: Only the drywell and RPV water level isolations are allowed to be bypassed since they do not indicate a release hazard. Reactor building exhaust radiation above the isolation setpoint would be indicative of a potential radioactive release problem and would not be allowed to be bypassed.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 66 of 100 ILT 11-1 (NRC 2012-301) 66 Points: 1.00 Unit 3 was shutdown 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> ago.

Reactor vessel water temperature is 100°F.

Core offload is in progress.

With no decay heat removal, what is the time to boil per OP-DR-104-1001?

A.

3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> B.

7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> C.

11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> D.

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Answer:

D Question 66 Details Topic:

66 - Generic.2.1.25 Comments:

Objective: 29501LP005

Reference:

OP-DR-104-1001 K/A: Generic.2.1.25 3.9 / 4.2 K/A: Ability to interpret reference materials, such as graphs, curves, tables, etc.

CFR: 41.10 Safety Function: 5 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: The examinee must know that with a core offload in progress, the reactor vessel is flooded up, the dryer/separator and fuel pool gates are removed making attachment K the correct graph to use.

a) Incorrect - 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> is the time to boil while flooded to flange and dryer/separator removed.

b) Incorrect - 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> is the time to boil while flooded up, gates in and dryer/separator in.

c) Incorrect - 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> is the time to boil while flooded up, gates in and dryer/separator removed d) Correct - 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> is the time to boil for conditions stated in the question.

REQUIRED

REFERENCES:

OP-DR-104-1001 Attachments A, C, E, G, I and K.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 67 of 100 ILT 11-1 (NRC 2012-301) 67 Points: 1.00 Given the following conditions:

There is a fire in the Aux Electric Equipment Room and Control Room evacuation is required The Unit 2 NSO has been directed to SCRAM the plant All procedural means to SCRAM the plant have failed Action is immediately needed to protect Public Health and Safety What is the lowest level of authority needed to approve action to SCRAM the plant that is NOT covered by a procedure or technical specification?

A.

Any licensed Reactor Operator (RO)

B.

Any licensed Senior Reactor Operator (SRO)

C.

The Shift Manager (SM)

D.

The Shift Operations Superintendent (SOS)

Answer:

B Question 67 Details Topic:

67 - Generic.2.1.20 Comments:

Objective: 29501LE016

Reference:

DRE299L-S4 and DSSP 100-CR K/A: Generic.2.1.20 4.6 / 4.6 K/A: Ability to interpret and execute procedure steps.

CFR: 41.10 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Explanation: As stated in reference materials - An SRO must approve action that is not covered by Tech Specs/Procedure. Certain steps or conditions within this procedure may contradict normal license conditions or Technical Specification requirements, as stated in 10 CFR 50.54x: "A licensee may take reasonable action that departs from a license condition or Technical Specification requirements in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with the license conditions and Technical Specifications that can provide adequate or equivalent protection is immediately apparent." This action "shall be approved, as a minimum, by a licensed Senior Reactor Operator (SRO) prior to taking the action."

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 68 of 100 ILT 11-1 (NRC 2012-301) 68 Points: 1.00 Unit 2 CCSW pump flow requirement is 3621 gpm. Unit 3 requires 3500 gpm.

This is because Unit 2 CCSW.....

A.

pumps also supply cooling water to the control room emergency HVAC system and this flow was determined to be 121 gpm.

B.

pumps analysis is in question and the plant is conservative in setting the required flow for Unit 2 at a higher flow.

C.

pumps can be used to supply 121 gpm to the LPCI corner room coolers in an emergency, Unit 3 is NOT able to perform this auxiliary function.

D.

piping is arranged differently in the plant than Unit 3. This piping arrangement results in uneven flow to the two (2) HXs. If 3621 gpm can be shown then the HX that receives the lower flow is guaranteed to be 3500 gpm.

Answer:

A Question 68 Details Topic:

68 - Generic.2.2.03 (3)

Comments:

Objective: DRE277LN001.03

Reference:

DOS 1500-02 discussion section K/A: Generic.2.2.03 3.8 / 3.9 K/A: Knowledge of the design, procedural, and operational differences between units.

Level: Memory Pedigree: Bank History: 2012 NRC Comments: Unit 2 CCSW PP flow requirement is higher because Unit 2 must be capable of supplying the backup Control Room HVAC refrigeration condensing unit (RCU) with its design flow of 102 gpm. Due to the piping configuration and orifice sizing, the actual flow to the RCU was determined to be 121 gpm. To ensure at least 3500 gpm is being supplied to the heat exchanger, a total pump flow of 3621 gpm is required.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 69 of 100 ILT 11-1 (NRC 2012-301) 69 Points: 1.00 Unit 2 is operating at 50% power and expected to pick up load early next shift.

'D' Lift pump is being prepared to start after maintenance on the motor has been completed.

The NSO in the control room attempts to start the 'D' Lift pump.

The EO at the Lift Station reports no observable change in status locally.

Based on these events, in regards to the 'D' Lift pump, the operating crew should _________.

A.

leave the pump as is to support troubleshooting activities B.

attempt to restart the pump ONE more time ONLY C.

attempt to restart the pump a MAXIMUM of TWO more times D.

place the control switch in PULL TO LOCK at the local control panel in the Lift Station Answer:

A Question 69 Details Topic:

69 - Generic.2.2.20 Comments:

Objective: 299L029

Reference:

DOP 6500-19 K/A: Generic.2.2.20 2.6 / 3.8 K/A: Knowledge of the process for managing troubleshooting activities.

CFR: 41.10 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Correct - Per DOP 6500-19, if the equipment malfunction is not readily apparent and plant conditions permit, then leave conditions as is to aid in troubleshooting.

b) Incorrect - One more restart is a restriction for Hot Restarts c) Incorrect - Two more restarts, is a restriction for cold starts.

d) Incorrect - PTL is the action to take if the cause of the breaker trip is known.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 70 of 100 ILT 11-1 (NRC 2012-301) 70 Points: 1.00 Chemistry has reported that high coolant activity exists on Unit 2 and a fuel element failure is suspected.

If site assembly is required, which of the following actions is required per DGA 16 to prevent excessive personnel exposure?

A.

Isolating HPCI steam lines B.

Isolating Recirc sample lines C.

Isolating the Isolation Condenser D.

Isolating HPCI steam drains to the condenser Answer:

D Question 70 Details Topic:

70 - Generic.2.3.14 Comments:

Reference:

ILTS026, DGA 16 K/A: Generic.2.3.14 3.4 / 3.8 K/A: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

CFR: 41.12 Safety Function: 9 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Per DGA 16 Caution. Assembly area inside the RPA is near the feed pumps, which is against the condenser shield wall. Any flow of radioactive water to the condenser would increase dose rates in this area, so a is correct. IC drains do not go to the condenser, so b is incorrect.

Isolating HPCI steam flow would block the leakage, but would render HPCI unavailable, so this action would not be appropriate. Recirc sample drains also do not go to the condenser.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 71 of 100 ILT 11-1 (NRC 2012-301) 71 Points: 1.00 An accident has occurred at the station and you have volunteered to perform an evolution to protect valuable property. The dose rate in the area you will be entering is 30 Rem/hr.

What is the MAXIMUM time you can spend in the area performing your task without violating TEDE Radiation Exposure Limits per RP-AA-203 EXPOSURE CONTROL AND AUTHORIZATION?

A.

4 minutes B.

10 minutes C.

20 minutes D.

30 minutes Answer:

C Question 71 Details Topic:

71 - Generic.2.3.4 (2)

Comments:

Objective: 29900LK208

Reference:

RP-AA-203, EP-AA-113 K/A: Generic.2.3.4 3.2 / 3.7 K/A: Knowledge of radiation exposure limits under normal or emergency conditions.

Level: High Pedigree: Bank History: 2012 NRC Explanation:

a) Incorrect - Would exceed 2REM which is the Local TEDE limit.

b) Incorrect - Would exceed 5REM which is the Federal TEDE limit.

c) Correct - The exposure limit for protecting valuable property is 10REM TEDE. Based on 30 REM/Hr in the area, the stay time would be 20 minutes.

d) Incorrect - Would exceed 15REM which is the local TODE limit.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 72 of 100 ILT 11-1 (NRC 2012-301) 72 Points: 1.00 Unit 3 was operating at near rated power when ALL annunciators are lost EXCEPT the annunciators labeled ANNUN DC PWR FAILURE" are illuminated on the following panels:

903-3 903-4 903-5 903-6 903-7 903-8 903-55 What are the immediate actions the Operating team is required to perform?

A.

Lock out the Recirc MG set scoop tubes ONLY.

B.

Insert a manual scram, and trip the Turbine ONLY.

C.

Stabilize and control RPV water level in normal band.

D.

Insert a manual scram, trip the Turbine AND lock out the Recirc MG set scoop tubes.

Answer:

C Question 72 Details Topic:

72 - Generic.2.4.32 Comments:

Objective: 26302LK001

Reference:

DOA 6900-03 K/A: Generic.2.4.32 3.6 / 4.0 K/A: Knowledge of operator response to loss of all annunciators.

Level: Memory Pedigree: Bank History: 2012 NRC, 2010 NRC Explanation: The conditions in the stem indicate a complete loss of Div 1 125VDC. The immediate action in the above DOA is to stabilize and control RPV water level in the normal control band. The distractors depict actions that would be taken if other sections of the DC system were lost.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 73 of 100 ILT 11-1 (NRC 2012-301) 73 Points: 1.00 Unit 2 was operating at near rated power when a steam line rupture in the HPCI room occurred.

A scram was attempted. An ATWS occurred, with the following conditions:

15 control rods are at position 24 or GREATER.

All scram valves are open.

CleanUp Pump Area temperature is 175oF.

HPCI Room temperature is 235oF.

HPCI Cubicle Rad Levels are 3750 mrem.

East CRD Module Area rad levels are 2500 mrem.

West CRD Module Area rad levels are 2800 mrem.

The Unit Supervisor has directed insertion of control rods using alternate methods.

Which action is the PREFERRED method for inserting the control rods that are NOT at position 00?

A.

Vent the control rods' overpiston volumes.

B.

Bypass the RWM AND drive the control rods MANUALLY.

C.

Close the scram air header supply valve AND vent the scram air header.

D.

Close the scram discharge vents and drains AND pull power supply fuses for the scram solenoids.

Answer:

B Question 73 Details Topic:

73 - Generic.2.4.23 Comments:

Objective: DRE215LN002.06

Reference:

DEOP 0300-01, DEOP 0500-05 K/A: Generic.2.4.23 3.4 / 4.4 K/A: Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

CFR: 41.10 PRA: No Level: High Pedigree: New History: 2012 NRC Explanation: Control rods should be manually inserted per DEOP 0500-

05. Pulling the power supply fuses for the scram solenoids would not be viable, because the scram valves are already opened. Venting the scram air header or the individual overpiston areas are not viable, because of the high dose (greater than Max Safe) in the areas of the CRD module areas.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 74 of 100 ILT 11-1 (NRC 2012-301) 74 Points: 1.00 Units 2 and 3 are operating at full power.

When the following events occur:

Security reports an aircraft has struck the reactor building.

DOA 0010-18, ESCALATED SECURITY EVENT / HOSTILE FORCE INTRUSION, has been entered.

Which of the following are required IMMEDIATE actions per DOA 0010-18?

Start the 2/3 Diesel Fire Pump Maximize CRD flow on BOTH Units Isolate RWCU systems on BOTH Units Verify MO 2(3)-1301-1 AND MO 2(3)-1301-4 on BOTH Units are closed A.

1, 2, 3 B.

1, 2, 4 C.

1, 3, 4 D.

2, 3, 4 Answer:

D Question 74 Details Topic:

74 - Generic.2.4.28 Comments:

Objective: 295L010

Reference:

DOA 0010-18 K/A: Generic 2.4.28 K/A: Knowledge of procedures relating to a security event (non-safeguards information).

CFR: 41.10 Safety Function: 5 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a, b & c) Incorrect, Start the 2/3 Diesel Fire Pump is an immediate action of DOA 0010-10, FIRE/EXPLOSION d) Correct, verify MO 2(3)-1301-1 AND MO 2(3)-1301-4 ON BOTH Units are closed, isolate BOTH Unit RWCU systems and maximize CRD flow are immediate actions of DOA 10-18 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 75 of 100 ILT 11-1 (NRC 2012-301) 75 Points: 1.00 A hydraulic ATWS has occurred on Unit 2.

The Unit Supervisor has directed you to perform repeated scram resets per DEOP 0500-05.

Which of the following describes the MINIMUM electrical safety precautions required to perform this task?

All metal removed, safety glasses,...

A.

long sleeve cotton shirt and rubber gloves B.

electrical safety coat and rubber gloves C.

electrical safety coat, face shield and rubber gloves D.

electrical safety coat, face shield and rubber gloves with leather gauntlets Answer:

A Question 75 Details Topic:

75 - Generic.2.1.26 Comments:

Objective: 295L105

Reference:

SA-AA-129 K/A: Generic 2.1.26 3.4 / 3.6 K/A: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

CFR: 41.10 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: SA-AA-129 states that for the voltage in the area being worked in for DEOP 500-5 the minimum PPE is all metal removed, long sleeve cotton shirt, safety glasses and rubber gloves REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 76 of 100 ILT 11-1 (NRC 2012-301) 76 Points: 1.00 Unit 2 was operating at near rated power when a transient occurred, resulting in the following set of conditions:

Annunciator 902-7 G-3, TURB BYP VLV OPEN, alarmed.

An NSO reported RPV water level is 22 inches and trending down.

An NSO reported that Main Steam line pressure is 800 psig.

An automatic scram occurred, with all rods reported as fully inserted.

An NSO reported RPV water level is 10 inches and trending down.

The Reactor scrammed on ___(1)___ and the Unit Supervisor is required to direct ___(2)___.

A.

(1) RPV water level; (2) entering DGP 2-3, REACTOR SCRAM B.

(1) 2 out of 3 DEHC Pressure Controller Processors failed low; (2) entering DGP 2-3, REACTOR SCRAM C.

(1) Main Steam Isolation Valves closure; (2) opening the MSIVs, per DOA 5650-03, TURBINE CONTROL VALVE OR BYPASS VALVE FAILED OPEN D.

(1) Main Steam Isolation Valves closure; (2) verifying Control valves throttle closed, per DOA 5650-03, TURBINE CONTROL VALVE OR BYPASS VALVE FAILED OPEN Answer:

D Question 76 Details Topic:

76 - 295006.A2.06 Comments:

Objective: DRE245LN001.06

Reference:

DOA 5650-03 K/A: 295006.A2.06 3.5 / 3.8 K/A: Ability to determine and/or interpret the following as the apply to SCRAM: cause of reactor SCRAM.

Safety Function: 1 PRA: No Level: High Pedigree: Bank History: 2009 NRC; 2012 NRC Explanation: When MSL pressure dropped below 827 psig and with the mode switch is in RUN, a scram will occur due to 9.5% Main Steam Isolation Valves closure. The required action is to verifying Control valves throttle closed, NOT open the MSIVs.

The scram setpoint for RPV water level is 6.02 inches.

2 out of 3 DEHC Pressure Controller Processors failing low will cause Bypass valves to close NOT open.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 77 of 100 ILT 11-1 (NRC 2012-301) 77 Points: 1.00 Unit 3 is operating at near rated power with ONLY the 3A RBCCW pump and 3A RBCCW Heat Exchanger in operation, when the following occurs:

The TUBE SIDE flow through 3A RBCCW Heat Exchanger is reduced to 10%, due to fouling.

If this condition goes un-corrected, RBCCW heat exchanger outlet temperatures will ___(1)___

AND the SRO will be required to direct actions per ___(2)___.

A.

(1) DECREASE, (2) DAN 903-4 E-16, 3 A/B PUMPBACK COMPRESSOR TRIP.

B.

(1) DECREASE; (2) DOP 1200-03, RWCU SYSTEM OPERATION WITH THE REACTOR AT PRESSURE.

C.

(1) INCREASE, (2) DAN 903-4 E-16, 3 A/B PUMPBACK COMPRESSOR TRIP.

D.

(1) INCREASE; (2) DOP 1200-03, RWCU SYSTEM OPERATION WITH THE REACTOR AT PRESSURE.

Answer:

D Question 77 Details Topic:

77 - 295018.AA2.02 Comments: Objective: DRE204LN001.06

Reference:

DOA 3900-01, DOP 1200-03, P&ID M-22 K/A: 295018.AA2.02 3.1 / 3.2 K/A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER :

Cooling water temperature Level: High Pedigree: Bank History: 2012 NRC Explanation: With Service Water (tube side) being reduced through the RBCCW heat exchanger, RBCCW temperatures will increase. If a misconception that RBCCW is on the tube side is assumed, then RBCCW heat exchanger outlet temperatures would decrease (same cooling, less RBCCW flow). With RBCCW temperature increasing, this would cause the RWCU non-regen heat exchanger temperatures to rise, which will cause a system isolation. The action on a system isolation is to dial the pressure controller to zero. Unit 3 Pumpback air compressors are unaffected by a Unit 3 RBCCW problem, since BOTH units Pumpback air compressors are cooled by Unit 2 RBCCW.

SRO Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 78 of 100 ILT 11-1 (NRC 2012-301) 78 Points: 1.00 Unit 2 is at full power.

Drywell temperature is 148°F and rising.

What is the Tech Spec limit for Drywell Air Temperature AND its Bases?

A.

150°F; to ensure LOCA peak Drywell temperature DOES NOT exceed 281°F.

B.

150°F; to ensure RPV water level instruments DO NOT become unreliable due to boiling in the instrument runs in the Drywell.

C.

160°F; to ensure LOCA peak Drywell temperature DOES NOT exceed 281°F.

D.

160°F; to ensure RPV water level instruments DO NOT become unreliable due to boiling in the instrument runs in the Drywell.

Answer:

A Question 78 Details Topic:

78 - 295028.EA2.01 Comments:

Objective: 21000LK004

Reference:

Tech Spec 3.6.1.5 and the Bases K/A: 295028.EA2.01 4.0 / 4.1 K/A: Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE :Drywell temperature Safety Function: 5 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Explanation: Per the Bases the temperature limit is 150°F to ensure that the safety analysis remains valid by maintaining the expected initial conditions and ensure that the peak LOCA drywell temp does not exceed the maximum allowable temp of 281°F.

SRO Criteria: 2 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 79 of 100 ILT 11-1 (NRC 2012-301) 79 Points: 1.00 Per the Tech Spec bases, what is the MINIMUM number of operable Main Turbine Bypass Valves (BPVs) required to prevent exceeding the MCPR safety limit during a Load Reject scram?

A.

6 B.

7 C.

8 D.

9 Answer:

D Question 79 Details Topic:

79 - 295005.G.2.2.25 (2)

Comments:

Objective: 241LN001.07

Reference:

T.S. 3.7.7 K/A: 295005.G.2.2.25 3.2 / 4.2 K/A: Main Turbine Generator Trip: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Safety Function: 3 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: As stated in Tech Spec 3.7.7 and associated Bases, and Tech Spec 2.0 (Safety Limit), and the COLR, Whenever the Unit is at, or above 25% power, all BPVs must be operable to prevent exceeding the MCPR Safety Limit during a turbine trip, load reject or feedwater controller failure.

SRO per Criteria: 2 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 80 of 100 ILT 11-1 (NRC 2012-301) 80 Points: 1.00 Unit 2 was operating at full power when a transient occurred resulting in an ATWS.

SBLC failed to inject and alternate SBLC injection is in progress.

Given the following sequence of events, what is the EARLIEST time that operators in the field can secure injecting SBLC?

12:00 - All APRM DOWNSCALE lights are lit.

12:05 - All IRMs are below range 7 12:10 - All control rods are at position 04, 02 or 00.

12:15 - All control rods are fully inserted.

A.

12:00 B.

12:05 C.

12:10 D.

12:15 Answer:

C Question 80 Details Topic:

80 - 295037.G.2.4.35 Comments:

Objective: 29502LP022

Reference:

DEOP 0500-01 K/A: 295037.G.2.4.35 K/A: SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown: Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Safety Function: 1 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) Incorrect - All APRM downscales are a factor in the second override of the level leg of DEOP 0400-05 b) Incorrect - IRMs below range 7 is a factor in the first override of the power leg of DEOP 0400-05 c) Correct - Per DEOP 0500-01 limitations and actions state: IF while executing this procedure any of the following conditions are met, THEN terminate Alternate Standby Liquid Control Injection. All control rods are inserted to OR beyond position 04.

d) Incorrect - Examinee must know that all rods do not need to be inserted per DEOP 0500-01 to secure from alternate SBLC SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 81 of 100 ILT 11-1 (NRC 2012-301) 81 Points: 1.00 Unit 3 was operating at near rated power when an Instrument Air system pipe ruptured, causing Instrument Air pressure to drop rapidly to 0 psig.

The FWRVs will fail ___(1)___.

If RPV water level is dropping and feedwater is needed for injection, while executing the DEOPs, the Unit Supervisor is required to direct Operators to ___(2)___.

A.

(1) immediately; (2) valve in Main Bank Nitrogen bottles, per DOA 4700-01, INSTRUMENT AIR SYSTEM FAILURE B.

(1) immediately; (2) install handwheels and manually operate the FWRVs, per DOA 0600-01, TRANSIENT LEVEL CONTROL C.

(1) after 30 minutes; (2) valve in Main Bank Nitrogen bottles, per DOA 4700-01, INSTRUMENT AIR SYSTEM FAILURE D.

(1) after 30 minutes; (2) install handwheels and manually operate the FWRVs, per DOA 0600-01, TRANSIENT LEVEL CONTROL Answer:

D Question 81 Details Topic:

81 - 295019.G.2.4.8 Comments:

Objective: DRE278LN001.12

Reference:

DOA 4700-01, DOA 0600-01 K/A: 295019.G.2.4.8 3.8 / 4.5 K/A: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Safety Function: 8 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC; 2010 NRC Explanation: Per the critical component failure table in the Loss of Inst Air DOA, the FWRVs will operate on their backup nitrogen source for 30 minutes after a loss of Inst Air. The DOA for transient level control states that if the FWRVs can not be opened due to loss of air and feedwater/alternate injection systems are required by the DEOPs, then operate the valves manually as directed by the Unit Supervisor.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 82 of 100 ILT 11-1 (NRC 2012-301) 82 Points: 1.00 Which of the following provide the bases to maintain Torus water level above -4.5 inches, in Modes 1, 2, and 3?

Ensures a sufficient amount of water.....

A.

with the Minimum CST Volume, Long-Term Cooling is available for the Design Basis Accident.

B.

would be available to adequately condense the steam from the relief valve quenchers ONLY.

C.

with the Minimum CST Volume, in the event of a LOCA to permit recirculation cooling flow to the core.

D.

would be available to adequately condense the steam from the relief valve quenchers, downcomer lines, or HPCI turbine exhaust line.

Answer:

D Question 82 Details Topic:

82 - 295030.G.2.2.22 Comments: Objective: DRE223LN001.07

Reference:

Bases 3.6.2.2 K/A: 295030.G.2.2.22 4.0 / 4.7 K/A: Low Suppression Pool Water Level: Knowledge of limiting conditions for operations and safety limits.

Safety Function: 5 PRA: No Level: Memory Pedigree: Bank History: 2007 NRC, 2012 NRC Explanation: Answer is per the above bases. CST is not part of bases. If the suppression pool water level is too low, an insufficient amount of water would be available to adequately condense the steam from the relief valve quenchers, downcomer lines, or HPCI turbine exhaust line.

SRO per Criteria: 2 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 83 of 100 ILT 11-1 (NRC 2012-301) 83 Points: 1.00 Unit 2 was operating at near rated conditions with the 902-36 back-panel recorder TIRS 2-1640-200A, TORUS TEMP MON DIV I, Out Of Service (O.O.S.) due to a failed power supply. The following conditions exist:

RPV pressure is 1000 psig.

HPCI testing per DOS 2300-03 "HIGH PRESSURE COOLANT INJECTION SYSTEM OPERABILITY AND QUARTERLY IST VERIFICATION TEST" is in progress.

TIRS 2-1640-200B currently indicates the following:

Point 1 165oF Point 2 90oF Point 3 110oF Point 4 150oF Point 5 140oF Point 6 160oF Point 7 125oF Point 8 130oF What action(s) is/are required based on the current readings?

A.

Start all available Torus cooling ONLY.

B.

Start all available Torus cooling and scram ONLY.

C.

Secure HPCI testing and start all available Torus cooling ONLY.

D.

Secure HPCI testing, start all available Torus cooling, AND scram.

Answer:

D Question 83 Details Topic:

83 - 295013.A2.01 Comments: Objective: 29502LK011

Reference:

DEOP 200-1 K/A: 295013.A2.01 3.8 / 4.0 K/A: Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Suppression pool temperature.

Safety Function: 5 PRA: No Level: High Pedigree: Bank History: 2012 NRC, 2007 NRC Comments: The average torus temperature is 133.5°F. Per DEOP 200-1 the actions that need to be taken are secure HPCI testing, start all available Torus cooling, and scram.

SRO per Criteria: 5 REQUIRED

REFERENCES:

DEOP charts, with the entry conditions blanked out.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 84 of 100 ILT 11-1 (NRC 2012-301) 84 Points: 1.00 Unit 3 is at full power.

When the following occurs:

903-4 C-19, LPCI/CS EAST SUMP LVL HI is received in the control room.

An EO reports a large unisolable leak on the Core Spray suction piping and there is 10 inches of water on the floor in the East Corner Room.

20 minutes later:

903-4 D-19, LPCI/CS WEST SUMP LVL HI is received in the control room.

An EO reports there is 8 inches of water on the floor in the West Corner Room.

Based on these conditions, the SRO is required to direct ____(1)____ and the basis for this is

____(2)____.

A.

(1) a reactor scram per DGP 02-03 (2) a direct threat exists relative to secondary containment integrity, to equipment located in the secondary containment and to continued safe operation of the plant.

B.

(1) a reactor shutdown per DGP 02-01 (2) a direct threat exists relative to secondary containment integrity, to equipment located in the secondary containment and to continued safe operation of the plant.

C.

(1) a reactor scram per DGP 02-03 (2) to reduce to decay heat levels the energy that the RPV may be discharging to the secondary containment.

D.

(1) a reactor shutdown per DGP 02-01 (2) to reduce to decay heat levels the energy that the RPV may be discharging to the secondary containment.

Answer:

B Question 84 Details Topic:

84 - 295036.G.2.4.18 Comments: Objective: 295L049

Reference:

BWROG EPG/SAGs, Appendix B (DEOP Bases)

K/A: 295036.G.2.4.18 K/A: Secondary Containment High Sump/Area Water Level: Knowledge of the specific bases for EOPs Safety Function: 5 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: Examinee must identify that a leak from core spray suction is not a leak from a primary system and a reduction in RPV pressure will not effect the leak rate.

a) Incorrect - the reactor would be scrammed if a primary system was discharging and unisolable.

b) Correct - Per BWROG EPG/SAGs, Appendix B c) Incorrect - the reactor would be scrammed if a primary system was discharging and unisolable d) Incorrect - this is the basis for scramming the reactor with an unisolable leak.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 85 of 100 ILT 11-1 (NRC 2012-301) 85 Points: 1.00 Unit 2 is at 6% power with a startup in progress.

The RBCCW supply valves to the Drywell fail closed.

RBCCW expansion tank level is lowering at 1 inch every 10 minutes.

Drywell pressure is 1.3 psig and rising 0.1 psig every 10 minutes.

The reactor is scrammed per DGP 02-03, REACTOR SCRAM.

To mitigate the effects on containment, the Unit Supervisor will direct actions per __________.

A.

DOA 0040-01, SLOW LEAK B.

DOA 3700-01, LOSS OF COOLING BY RBCCW SYSTEM C.

DOP 1600-01, NORMAL PRESSURE CONTROL OF THE DRYWELL OR TORUS D.

DOP 5750-11, DRYWELL COOLING SYSTEM OPERATION Answer:

C Question 85 Details Topic:

85 - 295010.G.2.4.11 Comments:

Objective: 29502LK001

Reference:

DEOP 200-1 K/A: 295010.G.2.4.11 4.0 / 4.2 K/A: High Drywell Pressure: Knowledge of abnormal condition procedures.

Level: High Pedigree: New History: 2012 NRC Comments:

a) Incorrect - DOA 40-1 directs action to combat a slow primary system leak.

b) Incorrect - DOA 3700-1 does not provide direction to control containment pressure after a loss of RBCCW c) Correct - DAN 902-3 A-13, DRYWELL PRESS HI, directs venting containment per DOP 1600-01 d) Incorrect - DOP 5750-11 lists RBCCW in operation as a prerequisite to start the Drywell Cooling system SRO per criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 86 of 100 ILT 11-1 (NRC 2012-301) 86 Points: 1.00 Unit 2 was operating at near rated power, when a transient occurred, resulting in the following sequence of events:

The Reactor scrammed.

MCC 29-2 experienced an overcurrent condition.

ALL RFPs and HPCI tripped and would NOT start.

An NSO reported RPV water level of -72 inches and trending down.

An NSO reported RPV pressure of 925 psig and trending down.

An NSO placed the SBLC keylock switch to the "SYS 1 & 2" position.

An NSO reported SBLC discharge pressure of 1500 psig and BOTH SQUIB A and B lights LIT Currently ___(1)___ SBLC pump is injecting into the RPV and the Unit Supervisor is required to direct ___(2)___.

A.

(1) neither; (2) injecting alternate SBLC per DEOP 500-1, ALTERNATE STANDBY LIQUID CONTROL INJECTION.

B.

(1) neither; (2) placing the SBLC keylock switch to the "SYS 2 & 1" position per DOP 1100-02, INJECTION OF STANDBY LIQUID CONTROL.

C.

(1) one; (2) verify RWCU system isolation per DOP 1100-02, INJECTION OF STANDBY LIQUID CONTROL, locally.

D.

(1) one; (2) placing the SBLC keylock switch to the "SYS 2 & 1" position per DOP 1100-02, INJECTION OF STANDBY LIQUID CONTROL.

Answer:

B Question 86 Details Topic:

86 - 211000.A2.04 Comments: Objective: DRE211LN001.08

Reference:

DOP 1100-02 K/A: 211000.A2.04 3.1 / 3.4 K/A: Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow.

Level: High Pedigree: New History: 2012 NRC Comments: b) With the pump discharge pressure 1500 psig, it is an indication that the relief valve has lifted (above relief setpoint) and is directing pump flow back to the main tank and NOT the RPV. DOP 1100-02 directs placing the SBLC switch to the opposite position selected first.

a) Injecting with alternate boron is done when SBLC is needed for an ATWS, NOT water injection.

c) Credible if examinee believes MCC 29-2 is the power supply to RWCU Isolation valves (MCC 28-1) d) Credible if examinee believes MCC 29-2 is power supply to B SBLC pump (MCC 29-1)

SRO per criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 87 of 100 ILT 11-1 (NRC 2012-301) 87 Points: 1.00 Unit 2 was shutdown, with the following set of conditions:

NO Recirc pumps are running.

RPV water temperature is 320°F.

2A SDC pump is running and aligned to the RPV.

2B SDC pump is running and aligned to the RPV.

2C SDC pump is running and aligned to the FPC system.

If Recirc Loop water temperature trends up to 360°F, what action(s) is/are the Unit Supervisor required to direct?

A.

Control RPV water temperature/pressure per DOA 1000-1, RESIDUAL HEAT REMOVAL ALTERNATIVES only.

B.

Align 2C SDC pump to the RPV to control RPV water temperature/pressure per DOP 1000-3, SHUTDOWN COOLING MODE OF OPERATION only.

C.

Align 2C SDC pump to the RPV to control RPV water temperature/pressure per DOP 1000-3, SHUTDOWN COOLING MODE OF OPERATION and monitor fuel storage pool water level once per hour per DOA 1900-1, LOSS OF FUEL POOL COOLING.

D.

Monitor fuel storage pool water level once per hour per DOA 1900-1, LOSS OF FUEL POOL COOLING and control RPV water temperature/pressure per DOA 1000-1, RESIDUAL HEAT REMOVAL ALTERNATIVES.

Answer:

A Question 87 Details Topic:

87 - 205000.A2.01 (2)

Comments: Objective: DRE205LN001.06

Reference:

DOA 1000-01, DOA 1900-01, DAN 902-4 B-23 K/A: 205000.A2.01 3.1 / 3.3 K/A: Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Recirculation loop high temperature.

Safety Function: 4 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: The SDC Pump Suction temperature (Recirc Loop Temp) exceeds the setpoint of 339°F for SDC pump trip, for the pumps lined up to the RPV, which causes 'A' and 'B' to trip. The 'C' pump will NOT trip, as it still has suction from the Fuel Pool Cooling system. Entry into DOA 1000-01 is the only required entry.

SRO per criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 88 of 100 ILT 11-1 (NRC 2012-301) 88 Points: 1.00 Given the following conditions:

Both units were at full power.

2/3A SBGT train is running for DOS 7500-02, SBGT SYSTEM SURVEILLANCE AND IST TEST.

2/3B SBGT train is in STANDBY.

2/3A SBGT Temperature indicates 0°F at the 923-5 panel.

An EO reports that 2/3A SBGT local temperature is 250°F.

Based on these conditions, the Unit Supervisor directs the NSO to:

A.

Place the 2/3A SBGT train to OFF, ONLY, per DOS 7500-02.

B.

Place the 2/3B SBGT train to OFF and the 2/3A SBGT train to OFF per DOS 7500-02.

C.

Place the 2/3B SBGT train to START and the 2/3A SBGT train to OFF per DAN 923-5 A-6, SBGT GAS TRT SYS A TROUBLE D.

Place the 2/3B SBGT train to OFF and the 2/3A SBGT train to OFF per DAN 923-5 A-6, SBGT GAS TRT SYS A TROUBLE Answer:

C Question 88 Details Topic:

88 - 261000.G.2.2.44 Comments:

Objective: 261LN001.08

Reference:

923-5 A-6 K/A: 261000.G.2.2.44 4.2 / 4.4 K/A: Standby Gas Treatment System: 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.

Safety Function: 9 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: The correct answer is C. A control room indication of 0°F is indicative of remote temperature indication failure when SBGT is running.

A charcoal adsorber temperature of 234°F or higher requires placing the train in the cooling mode. This is accomplished by placing 2/3B SBGT train (the STANDBY train) to START and placing 2/3A SBGT train (the PRIMARY train with high charcoal temperature) to OFF.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 89 of 100 ILT 11-1 (NRC 2012-301) 89 Points: 1.00 Unit 2 was operating at near rated power, with 2A LPCI pump taken O.O.S. at 0100 on July 1, 2012, and the following conditions exist:

2A Core Spray pump is running per DOS 1400-05, CORE SPRAY SYSTEM PUMP OPERABILITY AND QUARTERLY IST TEST WITH TORUS AVAILABLE.

System Discharge pressure is 235 psig.

System flow is 4400 gpm with MO 2 1402 4A, FLOW TEST VLV full open.

What is the impact to the Core Spray system and what actions are required?

A.

2A Core Spray subsystem is INOP; immediately enter LCO 3.0.3 for shutdown B.

2A Core Spray subsystem is INOP; restore Core Spray subsystem to operability by 0100 on July 8, 2012 C.

2A Core Spray subsystem is OPERABLE restore LPCI subsystem to operability by 0100 on July 8, 2012 D.

2A Core Spray subsystem is OPERABLE; restore 2A LPCI pump to operability by 0100 on July 31, 2010 Answer:

B Question 89 Details Topic:

89 - 209001.G.2.2.40 Comments:

Objective: DRE209LN001.07

Reference:

T.S. 3.5.1, DOS 1400-05 K/A: 209001.G.2.2.40 3.4 / 4.7 K/A: Low Pressure Core Spray System: Ability to apply Technical Specifications for a system.

Level: High Pedigree: Bank History: 2007 NRC, 2012 NRC Explanation: Per DOS 1400-5 the Core Spray pump must achieve a flowrate of 4600 gpm at greater than 235 psig. Since the flowrate is not sufficient the A Core Spray pump is INOP and puts you in a 7 day clock.

The 'A' LPCI pump INOP means that you have 2 low pressure subsystem INOP driving to an immediate shutdown per T.S. 3.0.3.

SRO per Criteria: 2 REQUIRED

REFERENCES:

Tech Spec 3.5.1 with less than one hour times blanked out.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 90 of 100 ILT 11-1 (NRC 2012-301) 90 Points: 1.00 Given the following conditions:

An earthquake occurred Unit 2 and 3 scrammed from full power Unit 3 steam leak due to a packing failure of 3-2301-5, STEAM ISOL VLV which fails to operate Unit 3 lost all AC electrical power Unit 3 reactor water level -60 inches trending up at 2 inches per minute Unit 2 supplied by the Unit 2/3 Emergency Diesel Generator ONLY supplying Buses 23-1, 28 and 29 Which of the following describes the order of procedures the Unit 3 Unit Supervisor should direct to mitigate the transient?

1. DEOP 0300-01, SECONDARY CONTAINMENT CONTROL

2. DEOP 0500-03, ALTERNATE WATER INJECTION SYSTEMS

3. DOP 6620-06, POWERING UNIT 3 4KV BUSES VIA THE SBO DG 3 A.

1, 2, 3 B.

2, 1, 3 C.

3, 1, 2 D.

3, 2, 1 Answer:

C Question 90 Details Topic:

90 - 262001.G.2.4.5 Comments:

Objective: 295L089

Reference:

DGA-12; DOA 6500-01; DGP 02-03; DEOP 0100 K/A: 262001.G.2.4.5 3.7 / 4.3 K/A: A.C. Electrical Distribution: Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

Safety Function: 6 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: The priority of actions should be to restore AC electrical power (no other actions to mitigate the transient would be possible without AC power), then stopping the steam leak by shutting the 3-2301-4 valve which is AC powered then addressing reactor water level control.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 91 of 100 ILT 11-1 (NRC 2012-301) 91 Points: 1.00 Unit 2 was in STARTUP, with the RWCU system in operation and RPV pressure was 300 psig, when the Instrument Air supply line to RWCU FCV 2-1219, ruptured.

The Unit Supervisor is required to direct ________.

A.

reducing system flow per DAN 902-4 C-12 RWCU RECIRC PP DISCH PRESS LO B.

an operator to check the breaker relay targets per DAN 902-4 A-10 RWCU RECIRC PP TRIP C.

verification of system isolation per DOP 1200-03, RWCU SYSTEM OPERATION WITH THE REACTOR AT PRESSURE hard card D.

securing the RWCU system per DOP 1200-03, RWCU SYSTEM OPERATION WITH THE REACTOR AT PRESSURE hard card Answer:

D Question 91 Details Topic:

91 - 204000.A2.03 Comments:

Objective: DRE204LN001.12

Reference:

M-30, DOA 4700-01, DOP 1200-03, DANs 902-4 A-10 & C-12 K/A: 204000.A2.03 2.9 / 2.9 K/A: Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Flow control valve failure.

Level: High Pedigree: Bank History: 2012 NRC, 2009 NRC Explanation: Loss of IA to the FCV causes valve to fail closed (does not lock up). The pump experiences a dead headed situation (NOT a pump trip). Per DOP 1200-03, during a system transient, secure the system via the hard card. The high pressure isolation (downstream of the PCV) would not occur, since flow was isolated by the failing closed of the FCV.

SRO Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 92 of 100 ILT 11-1 (NRC 2012-301) 92 Points: 1.00 Unit 3 was operating at near rated power, when a transient occurred with the EHC Pressure Controller Processors.

The following parameters are reported:

RPIS has failed.

IRMs are reading 50 on range 7 and decreasing.

RPV pressure is 1065 psig and trending up at 1 psig/minute.

RPV water level dropped to -12 inches and trending up at 1 inch/minute.

Per the DEOPs, the Unit Supervisor is required to direct __________.

A.

reducing RPV pressure to 945 psig maximum RPV water level band of +8 to +48 inches B.

controlling RPV pressure between 800 to 1060 psig maximum RPV water level band of +8 to +48 inches C.

controlling RPV pressure between 800 to 1060 psig maximum RPV water level band of -40 to -50 inches D.

controlling RPV pressure between 800 to 1060 psig maximum RPV water level band of -164 to +48 inches.

Answer:

D Question 92 Details Topic:

92 - 241000.G.1.07 Comments: Objective: DRE248LN001.12

Reference:

DEOP 0400-05 K/A: 241000.G.1.07 4.4 / 4.7 K/A: Reactor/Turbine Pressure Regulating System: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Level: High Pedigree: Modified History: 2012 NRC Comments:

a) Incorrect - If ADS valves were cycling (RPV pressure > 1130psig) these would be possible bands per the DEOPs b) Incorrect - If RPIS was available these would be possible bands per the DEOPs c) Incorrect - If the second override of DEOP 400-5 were used this would be the band for level and pressure.

d) Correct - Because it is unknown if the reactor will stay shutdown under all conditions (RPIS Failure, IRM power on scale), DEOP 400-5 is entered. With reactor power less than 6%, the correct level bands are -164 to 48 inches; 800 -

1060 psig.

SRO per criteria: 5 REQUIRED

REFERENCES:

DEOP charts, with the entry conditions blanked out.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 93 of 100 ILT 11-1 (NRC 2012-301) 93 Points: 1.00 Unit 2 was operating at near rated conditions, with the 2A Fuel Pool Cooling (FPC) pump O.O.S.

for motor replacement, when Bus 29 de-energized due to an overcurrent condition.

The Unit Supervisor is required to direct monitoring fuel storage pool water temperatures and level once per hour ___(1)___ per ___(2)___.

A.

(1) locally; (2) DOA 1900-01, LOSS OF FUEL POOL COOLING B.

(1) locally; (2) DOA 1000-01, RESIDUAL HEAT REMOVAL ALTERNATIVES C.

(1) from the Main Control Room; (2) DOA 1900-01, LOSS OF FUEL POOL COOLING D.

(1) from the Main Control Room; (2) DOA 1000-01, RESIDUAL HEAT REMOVAL ALTERNATIVES Answer:

A Question 93 Details Topic:

93 - 233000.A2.09 (1)

Comments:

Objective: DRE233LN001.08

Reference:

DOA 1900-01 K/A: 233000.A2.09 2.7 / 2.9 K/A: Ability to (a) predict the impacts of the following on the FUEL POOL COOLING AND CLEAN-UP ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C. electrical power failures.

Level: High Pedigree: Bank History: 2012 NRC Comments: With A FPC pump O.O.S. and a subsequent loss of B FPC pump (loss of Bus 29), there is no forced cooling of the fuel storage pool.

Per DOA 1900-01, when forced cooling of the fuel storage pool is not available, then monitor fuel storage pool water temperatures and level once per hour LOCALLY (Main Control Room indication of spent fuel pool temperature may be erroneous due to water stagnation).

SRO per criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 94 of 100 ILT 11-1 (NRC 2012-301) 94 Points: 1.00 Unit 2 has a core offload in progress.

Which of the following is REQUIRED to be performed prior to removing control rod XX-YY from the core for blade replacement?

A.

Ensure all other control rods are fully inserted.

B.

Ensure that the control cell for control rod XX-YY is defueled.

C.

Two SRMs are operable with at least one in an adjacent quadrant.

D.

Ensure that the REFUEL INTERLOCKS for control rod XX-YY have been defeated.

Answer:

B Question 94 Details Topic:

94 - Generic.2.1.42 (3)

Comments:

Objective: 29800LK084

Reference:

DFP 0800-16 K/A: Generic.1.42 2.5 / 3.4 K/A: Knowledge of new and spent fuel movement procedures.

Level: Memory Pedigree: Bank History: 2012 NRC Explanation:

a) Incorrect - Is a requirement of the Tech Spec for single control rod drive removal 3.10.4 b) Correct - Per DFP 0800-16 prerequisites, the control cell for control rod XX-YY must be defueled prior to removing the control rod.

c) Incorrect - Per T.S. 3.3.1.2, only 1 SRM s required to be in operation when core offload is in progress.

d) Incorrect - Refueling interlocks are not required to be defeated to lift the rod out of the top core, only to withdraw the rod out the bottom for testing.

SRO per criteria: 7 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 95 of 100 ILT 11-1 (NRC 2012-301) 95 Points: 1.00 During performance of DOS 1500-02 the following pump flows were recorded:

Unit 2: 2B CCSW pump 3610 gpm Unit 3: 3B CCSW pump 3590 gpm The Unit Supervisor will declare _______.

A.

BOTH 2B AND 3B CCSW Pumps NOT operable B.

BOTH 2B AND 3B CCSW Pumps operable C.

the 2B CCSW Pump operable BUT the 3B CCSW Pump NOT operable D.

the 3B CCSW Pump operable BUT the 2B CCSW Pump NOT operable Answer:

D Question 95 Details Topic:

95 - Generic.2.2.37 Comments:

Objective: DRE277LN001.07

Reference:

DOS 1500-12, T.S. Bases 3.7.1, UFSAR 9.2 K/A: Generic.2.2.37 3.6 / 4.6 K/A: Ability to determine operability and/or availability of safety related equipment.

PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Explanation: The Unit 2 CCSW Pumps must have a flowrate > 3621 gpm (121 gpm for CREVs). Unit 3 CCSW Pumps must have a flowrate of

>3500. If these flow rates cannot be achieved, the associated pump must be declared inoperable.

SRO per Criteria: 1 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 96 of 100 ILT 11-1 (NRC 2012-301) 96 Points: 1.00 A spent fuel cask is being moved on Unit 2.

The following events occur:

Reactor Building Overhead Crane Radiation Monitor High Radiation alarms locally.

Refuel Floor ARM High Radiation alarms locally.

Overhead Crane dose rates are measured at 40 mr/hr.

Refuel Floor dose rates are measured at 150 mr/hr.

Which of the following describe the effect of the above conditions?

A.

SBGT has started and Reactor Building Overhead Crane movement is inhibited in the hoist raise function ONLY.

B.

SBGT has started and Reactor Building Overhead Crane movement is inhibited in the hoist raise and lower function.

C.

Reactor Building Ventilation system is running and Reactor Building Overhead Crane movement is inhibited in the hoist raise function ONLY.

D.

Reactor Building Ventilation system is running and Reactor Building Overhead Crane movement is inhibited in the hoist raise and lower function.

Answer:

A Question 96 Details Topic:

96 - Generic.2.3.5 Comments:

Objective: 299L042

Reference:

DFP 0850-03 K/A: Generic 2.3.5 2.9 / 2.9 K/A: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) Correct - DFP 0850-03 states RB Vent isolates and SBGT starts with Refuel Floor Rad > 100 mr/hr and RB Crane hoist raise is inhibited > 30 mr/hr.

b) Incorrect - The RB Crane lower function is not inhibited > 30 mr/hr.

c) Incorrect - RB Vent isolates with Refuel Floor rad >100 mr/hr.

d) Incorrect - RB Vent isolates with Refuel Floor rad >100 mr/hr.

SRO per Criteria: 9 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 97 of 100 ILT 11-1 (NRC 2012-301) 97 Points: 1.00 Unit 2 is operating near rated power.

HVOs and EMD have been trying to locate and isolate an 80 volt ground on the 125 VDC System.

The Unit 2 NSO then notices that two previously lit annunciators are now extinguished.

A test of the annunciators shows ALL of the annunciators on the 902-3, 5, 6, and 8 panels have been lost.

20 minutes later, the cause of the loss of the annunciators was corrected.

What is the minimum EP classification level (if any) that must be declared? (Exclude discretionary EALs)

A.

NO EP Classification B.

Unusual Event C.

Alert D.

Site Area Emergency Answer:

B Question 97 Details Topic:

97 - Generic.2.4.41 Comments:

Objective: 29501LP032

Reference:

EP-AA-1004 K/A: Generic.2.4.41 K/A: Knowledge of the emergency action level thresholds and classifications.

PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: The correct EAL is MU4 because ALL of the annunciators 902-3, 5 and 8 panels have been lost for >15 minutes (EAL Threshold Value). To upgrade to an Alert requires a significant transient in progress OR the loss of one of the following: 1) all indication needed to monitor criticality 2) core heat removal 3) fission product barrier status.

SRO per Criteria: 5 REQUIRED

REFERENCES:

EP-AA-1004 EAL Charts.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 98 of 100 ILT 11-1 (NRC 2012-301) 98 Points: 1.00 A Site Area Emergency has been declared.

The automated call out system is unavailable.

Which of the following emergency response personnel MUST be notified using pagers?

A.

Station ERO personnel ONLY B.

Station and Corporate ERO personnel ONLY C.

Station ERO personnel and NDO ONLY D.

Station and Corporate ERO personnel and NDO Answer:

D Question 98 Details Topic:

98 - Generic.2.4.40 Comments:

Objective: 295L160

Reference:

EP-AA-112-100-F-06 K/A: Generic 2.4.40 K/A: Knowledge of SRO responsibilities in emergency plan implementation.

PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Incorrect - Unusual Events are not required to have offsite notifications when using the automated call out system.

b) Incorrect - Could be selected if not known that NDO is notified using the pager system.

c) Incorrect - Could be selected if not known that EOF is required to be staffed during a site area emergency.

d) Correct - Per EP-AA112-100-F-06 step 4 SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 99 of 100 ILT 11-1 (NRC 2012-301) 99 Points: 1.00 DOP 2000-110, Attachment 1: WASTE SURGE TANK RADIOACTIVE WASTE DISCHARGE TO RIVER CARD, contains the calculation for determining the ___(1)___ flowrate and radiological monitor alarm setpoints, and, excluding designees, is REQUIRED to be verified by ___(2)___.

A.

(1) dilution (2) Unit Supervisor B.

(1) dilution (2) Shift Manager C.

(1) discharge (2) Unit Supervisor D.

(1) discharge (2) Shift Manager Answer:

D Question 99 Details Topic:

99 - Generic.3.06 Comments:

Objective: DRE268LN001.14

Reference:

DOP 2000-110 K/A: Generic.3.06 2.0 / 3.8 K/A: Ability to approve release permits.

Level: Memory Pedigree: Bank History 2012 NRC Explanation: The calculations that must be completed prior to starting the release determine the maximum discharge rate (dilution is provided on the river card as a constant). The calculations must then be verified by the Shift Manager or designee, to approve the river discharge.

SRO per criteria 4.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 100 of 100 ILT 11-1 (NRC 2012-301) 100 Points: 1.00 Both units are operating at near rated power with the following conditions:

Fuel handlers are inspecting new fuel on the Refuel Floor.

The 2/3A SBGT train is in day 2 of a seven day OOS for planned maintenance.

The 2/3B SBGT train control switch is in the PRI position.

What required action is initiated IF the planned maintenance on the 2/3A SBGT train lasted past the planned 7 day duration?

A.

Place the 2/3B SBGT train in operation.

B.

Place both units in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C.

Restore the 2/3A SBGT train to OPERABLE status.

D.

Immediately suspend movement of new fuel assemblies.

Answer:

B Question 100 Details Topic:

100 - Generic.2.23 (1)

Comments:

Objective: DRE299LN001.04

Reference:

Tech Spec Section 3.6.4.3 K/A: Generic.2.23 3.1 / 4.6 K/A: Ability to track Technical Specification limiting conditions for operations.

Level: High Pedigree: Bank History: 2007 NRC, 2012 NRC Explanation: This knowledge can be gained from the Tech Specifications section 3.6.4.3 action B.1.

Action B.1 states be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for correct answer to the question.

Fuel handlers are moving new fuel which is NOT irradiated fuel as the applicability statement states, so no need to stop fuel inspections nor place the OPERABLE SBGT subsystem in operation. Already have one OPERABLE (2/3B) SBGT subsystem so no need for restoration.

SRO per Criteria: 2 REQUIRED

REFERENCES:

TS 3.6.4.3, with less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> times removed.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 1 of 100 ILT 11-1 (NRC 2012-301) 1 Points: 1.00 Given the following conditions TR22 has tripped due to an internal fault, resulting in a reactor scram and subsequent loss of off-site power.

The Unit 2 Isolation Condenser is in service providing reactor pressure control. The 2-1301-3, RX INLET ISOL, valve is being throttled to control cooldown.

A fire has been reported in 125 VDC Bus 2A-1.

250 VDC MCC 2 has tripped due to a short circuit.

The Fire Brigade leader has called to inform you 125 VDC Bus 2A-1 is going to be de-energized by the HVO so the fire can be extinguished.

Which of the following is true with regard to the Isolation Condenser when 125 VDC Bus 2A-1 is de-energized (assume no operator action)?

A.

It will automatically isolate.

B.

It will remain in service, but cooldown rate will be fixed.

C.

It will remain in service, but cooldown rate will be maximized.

D.

It will remain in service with the ability to control cooldown rate.

Answer:

A Question 1 Details Topic:

1 - 207000.K1.11 Comments: Objective: 207LN001.06

Reference:

DOA 6900-02 K/A: 207000.K1.11 3.4 / 3.6 K/A: Knowledge of the physical connections and/or cause-effect relationships between ISOLATION (EMERGENCY) CONDENSER and the following: Primary containment isolation system: BWR-2,3 CFR: 41.8 Safety Function: 4 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments:

a) Isolation Condenser Group 5 isolation circuits are de-energize to actuate. The power supplies are 2A-1 and 2B-1 for Unit 2. Loss of EITHER power supply will cause a complete isolation.

b) Isolation condenser DC MOVs are powered from 250VDC MCC 2A which receives power from Unit 3 250 VDC MCC 3. If power were lost to these valves, cooldown rate would be fixed.

c) 2-1301-3 valve HAND/RESET circuit is powered from the 120 VAC Instrument Bus and will remain energized. If power were lost to this logic, the 2-1301-3 valve would fully open maximizing cooldown rate.

d) Would be correct if the Group 5 circuit was energize to actuate (ie HPCI Group 4).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 2 of 100 ILT 11-1 (NRC 2012-301) 2 Points: 1.00 Unit 3 was operating at near rated power, when Bus 33-1 experienced an OVERCURRENT condition.

Which of the following be a correct plant response/indication?

A.

Unit 3 RPS EPA 3A-1 "IN" light extinguished.

B.

Unit 3 RPS EPA 3B-1 "IN" light extinguished.

C.

Annunciator 903-8 C-4 U2/3 DIESEL GEN FAIL TO START in alarm.

D.

Annunciator 903-8 E-8 ESS UPS ON DC OR ALTERNATE AC in alarm.

Answer:

A Question 2 Details Topic:

2 - 212000.K1.04 Comments: Objective: DRE212LN001.03

Reference:

DOP 0500-03 figure 2, DANs 903-8 E-5 and 903-8 E-8 K/A: 212000.K1.04 3.4 / 3.6 K/A: Reactor Protection System: Knowledge of the physical connections and/or cause effect relationships between REACTOR PROTECTION SYSTEM and the following: A.C. electrical distribution.

CFR: 41.7 Safety Function: 7 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC, 2008 NRC Comments: With an overcurrent on Bus 33-1 it will become completely de-energized, which will cause Bus 38 to become de-energized. With Bus 38 de-energized, MCC 38-2 also becomes de-energized, which is the power supply to RPS MG set A. When MG Set A coasts down, the EPA 3A-1 becomes de-energized (IN power light goes out).

EPA 3B-1 is incorrect because it is powered from MG Set B (MCC 39-2).

Annunciator for U2/3 EDG fail to start is incorrect, because with an overcurrent on Bus 33-1, the U2/3 EDG will auto start (but is prevented from closing in on the bus).

Annunciator for ESS Bus Transfer is incorrect because it is powered from Bus 39, not Bus 38.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 3 of 100 ILT 11-1 (NRC 2012-301) 3 Points: 1.00 Unit 2 was operating at near rated power when the 125 VDC 2B-1 Distribution panel de-energized.

Several hours later, the Electrical Maintenance Department completed repairs and the 2B-1 Distribution panel was RE-ENERGIZED.

What is the expected power supply to the Unit 2, Div 2 ADS logic without further operator actions?

A.

U2 ESS Bus B.

U2 Instrument Bus C.

125 VDC Distribution Panel 2A-1 D.

125 VDC Distribution Panel 2B-1 Answer:

D Question 3 Details Topic:

3 - 218000.K2.01 Comments: Objective: DRE218LN001.02

Reference:

12E-2461 sh 1, 12E-2462 sh 1 K/A: 218000.K2.01 3.1 / 3.3 K/A: Knowledge of electrical power supplies to the following: ADS logic CFR: 41.7 Safety Function: 3 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: The U2 Div 2 ADS logic has two power supplies. The normal supply is the 2B-1 Dist Panel and the alternate supply is the 2A-1 Dist Panel. When the normal supply (2B-1) is lost a normal seeking automatic transfer device will transfer to the alternate supply (2A-1). Upon the normal supply being re-energized, the normal seeking automatic transfer device will transfer power back from the alternate supply (2A-1) to the normal supply (2B-1).

a) The ESS Bus is a power supply to HPCI and PCIS logic.

b) The Instrument Bus is a power supply to the Isolation Condenser logic and controls.

c) 2A-1 would be the power supply if the automatic transfer device didnt transfer back to the normal supply.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 4 of 100 ILT 11-1 (NRC 2012-301) 4 Points: 1.00 With Unit 2 at 100% power, the following events occur:

MCC 29-1 de-energized due to an overcurrent trip.

Drywell pressure began rising and eventually reached 2.5 psig.

10 minutes later HPCI area temperature rises to 300°F.

Which of the following is correct concerning the status of the HPCI system?

A.

HPCI is running and injecting into the RPV.

B.

HPCI is running but NOT injecting into the RPV.

C.

HPCI injected into the RPV but is now isolated.

D.

HPCI did NOT inject into the RPV and is now isolated.

Answer:

C Question 4 Details Topic:

4 - 206000.K2.01 Comments: Objective: DRE206LN001.12

Reference:

DAN 902(3)-3 C-7, DAN 902(3)-5 D-11, DOP 6700-19 K/A: 206000.K2.01 3.2 / 3.3 K/A: Knowledge of electrical power supplies to the following: System valves CFR: 41.7 Safety Function: 2 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments:

A loss of MCC 29-1 results in a loss of 4-valve control power. 4-valve control power supplies half of the 4 & 5 valve isolation logic. However, HPCI 125 VDC still supplies the other half of the logic. The 5-valve will still isolate due to the high HPCI area temperature isolation signal. The sequence of events will result in HPCI injection due to the high DW pressure signal and then HPCI will isolate upon receipt of high area temperature.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 5 of 100 ILT 11-1 (NRC 2012-301) 5 Points: 1.00 A transient occurred on Unit-2 some time ago. Eventually the DEOP flowcharts were exited, and SAMGs were entered.

The following conditions exist:

Reactor Pressure 75 psig steady RPV Water Level

-200" steady Drywell Pressure 10 psig steady Primary Containment Water Level 80 ft. steady Electrical Power ALL buses are ENERGIZED Instrument Air has been lost, but all other required systems are operable In the current condition, which of the below may be used to vent the RPV?

A.

MSIV's B.

Head Vent C.

HPCI steam line D.

Isolation Condenser vent valves Answer:

B Question 5 Details Topic:

5 - 300000.K3.02 Comments: Objective: DRE295LN003.03

Reference:

SAMG-1; DEOP 500-06 K/A: 300000.K3.02 3.3 / 3.4 K/A: Knowledge of the effect that a loss or malfunction of the (INSTRUMENT AIR SYSTEM) will have on the following: Systems having pneumatic valves and controls CFR: 41.7 Safety Function: 8 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: Instrument Air is required to open various valves in all the above mentioned systems except the Head Vent. Without Instrument Air, only the head vent method is available in the current condition.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 6 of 100 ILT 11-1 (NRC 2012-301) 6 Points: 1.00 Unit 3 was operating at near rated power with the Computer UPS 250 VDC battery system O.O.S.

A transient occurs causing Bus 36 to trip on overcurrent.

Which of the following power supplies would provide power to the Unit 3 Rod Worth Minimizer?

A.

MCC 38-2 B.

Bus 25 C.

Bus 28 D.

Bus 35 Answer:

B Question 6 Details Topic:

6 - 262002.K3.05 Comments: Objective: 283LN001.06

Reference:

DOP 6800-03, 12E-6600 K/A: 262002.K3.05 2.9 / 3.0 K/A: Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following:

Rod worth minimizer CFR: 41.7 Safety Function: 6 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: The RWM on Unit 3 is powered from the Computer UPS.

a) MCC 38-2 is a normal power supply for Instrument Bus and alternate power supply for ESS Bus power on Unit 3.

b) Bus 25 will power the Computer UPS system on a loss of the batteries and Bus 36.

c) Bus 28 is an alternate power supply for ESS Bus power on Unit 2.

d) Bus 35 is a common misconception due to the Computer UPS normal power coming from Unit 3.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 7 of 100 ILT 11-1 (NRC 2012-301) 7 Points: 1.00 The high pressure sensing line for the core plate D/P instrumentation has a fitting leak of 10 gpm at the transmitter.

An ATWS event occurs and SBLC is started.

Which of the following describes the response of the SBLC system?

The SBLC system will inject__________.

A.

all of the boron into the RPV above the core plate B.

all of the boron into the RPV below the core plate C.

most of the boron into the RPV above the core plate, the rest outside the RPV D.

most of the boron into the RPV below the core plate, the rest outside the RPV Answer:

D Question 7 Details Topic:

7 - 211000.K4.06 Comments: Objective: DRE211LN001.03

Reference:

M-33, M-26, LPDRE223LN004 K/A: 211000.K4.06 2.6 / 2.7 K/A: Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Core plate differential pressure indication CFR: 41.7 Safety Function: 1 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC, 2003 Pilgrim Unit 1 Comments:

a) Incorrect, boron is injected below the core plate.

b) Would be correct if the low pressure sensing line tapped into the SBLC injection line.

c) Incorrect, boron is injected below the core plate.

d) With a leak on the high pressure sensing line that senses pressure below the core plate from a tap off the SBLC injection line, some of the boron would be injected below the core plate with the rest being lost outside the RPV.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 8 of 100 ILT 11-1 (NRC 2012-301) 8 Points: 1.00 Unit 2 was operating at near rated power when a transient occurred resulting in an ATWS.

The Unit Supervisor directed the Operating team to bypass the interlocks for the MSIV low water level AND Off Gas high radiation isolations.

Which of the following would be a correct indication of the interlocks above being properly bypassed?

A.

The Off Gas SJAE suction valves automatically open.

B.

Annunciator 902-5 B-13, CHANNEL A/B RPV LVL LO, can be reset.

C.

Annunciator 902-5 D-4, GROUP 1 ISOLATION INITIATED, can be reset.

D.

The MSIVs automatically open, UNLESS a valid MSL high rad signal is present.

Answer:

C Question 8 Details Topic:

8 - 223002.K4.08 Comments: Objective: 29502LK077

Reference:

DAN 902-5 B-13, DEOP 0500-02, 12E-2501 sht 1 and 2 K/A: 223002.K4.08 3.3 / 3.7 K/A: Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following: Manual defeating of selected isolations during specified emergency conditions.

CFR: 41.7 Safety Function: 5 PRA: No Level: Memory Pedigree: Bank History: 2010 NRC, 2012 NRC Explanation: The jumpers installed per the above procedure bypass the lo-lo RPV water level Group 1 signal. This permits the annunciator to be reset. The low level annunciator cannot be reset when these jumpers are installed. The MSIVs do NOT auto open (these jumpers prevent their closure). The MSL hi-rad signal no longer is an input that would cause a Group 1 isolation. The Off Gas suction valves are not affected by these jumpers.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 9 of 100 ILT 11-1 (NRC 2012-301) 9 Points: 1.00 A Unit 2 startup is in progress and the following conditions exist:

Reactor Mode Switch is in STARTUP.

All IRM range switches are on 9 or 10.

All IRMs are reading mid scale.

A fault in the IRM detector drive system causes all IRM detectors to begin to withdraw.

Which of the following would be the first indication of the fault in the IRM detector drive system?

A.

Rod Out Block B.

RPS Full Scram C.

IRM INOP alarm D.

IRM downscale alarm Answer:

A Question 9 Details Topic:

9 - 215003.K5.03 Comments: Objective: DRE215LN003.06

Reference:

DOP 0700-02, DAN 902(3)-5 C-10 and 15 K/A: 215003.K5.03 3.0 / 3.1 K/A: Knowledge of the operational implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM :

Changing detector position CFR: 41.5 Safety Function: 7 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments:

a) With the Mode Switch in STARTUP, IRMs must be fully inserted, or a Rod Out Block is generated.

b) Scram signal is generated by high neutron flux or IRM HiHi/Inop with companion APRM downscale.

c) INOP alarm is generated by low voltage, function switch not in operate and module not plugged in properly.

d) With the IRMs reading mid scale, downscale alarms would not immediately come in as soon as the detectors began to move.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 10 of 100 ILT 11-1 (NRC 2012-301) 10 Points: 1.00 A prerequisite of DOS 6620-01, SBO 2 DIESEL GENERATOR SURVEILLANCE AND MARGIN/FULL LOAD TESTFULL LOAD REJECT TEST, is to ensure that the bus selected for this surveillance is powered from either Transformers TR 21 or TR 22 AND the associated Emergency Diesel Generator is NOT paralleled to this bus.

What is the operational concern for this prerequisite?

A.

A trip of either Diesel Generator while in this lineup, would cause the Bus to de-energize.

B.

Large circulating currents, due to voltage differences, may result in an auto trip of both Diesel Generators.

C.

A neutral ground condition, due to voltage differences, may result in an auto trip of both Diesel Generators.

D.

A neutral voltage condition, due to phase angle differences, may result in an auto trip of both Diesel Generators.

Answer:

B Question 10 Details Topic:

10 - 264000.K5.05 Comments:

Objective: DRE264LN001.10

Reference:

BWR Fundamentals Ch. 5 Motors and Generators, DOS 6620-01 K/A: 264000.K5.05 3.4 / 3.4 K/A: Knowledge of the operational implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESEL/JET):

Paralleling A.C. power sources.

CFR: 41.5 Safety Function: 6 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC a) Incorrect - Upon a trip of either diesel generator, the bus would remain energized, powered from the other diesel generator.

b) Correct - A difference in voltages between generators will cause a circulating current to flow that could exceed breaker load limits.

c) Incorrect - A neutral ground condition would be caused by a mechanical or electrical fault in the generator or associated wiring.

d) Incorrect - Large circulating currents may cause a neutral voltage condition, but with the generators operating in parallel, the phase angles would be matched.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 11 of 100 ILT 11-1 (NRC 2012-301) 11 Points: 1.00 Unit 2 was operating at near rated power when a complete loss of U2 125 VDC Division 1 occurred.

Which of the following buses will no longer have breaker indication in the Control Room?

A.

Bus 23 B.

Bus 24 C.

Bus 26 D.

Bus 27 Answer:

A Question 11 Details Topic:

11 - 262001.K6.01 Comments:

Objective: DRE263LN002.12

Reference:

DOA 6900-02 K/A: 262001.K6.01 3.1 / 3.4 K/A: Knowledge of the effect that a loss or malfunction of the following will have on the A.C. ELECTRICAL DISTRIBUTION: D.C. power.

CFR: 41.7 Safety Function: 6 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC Explanation: With a loss of 125 VDC Division 1, all division 1 buses will lose remote indication and protection ability. Bus 23 is the only Div 1 bus listed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 12 of 100 ILT 11-1 (NRC 2012-301) 12 Points: 1.00 Unit 2 is operating at full power with the following conditions:

2A and 2B RBCCW pumps and heat exchangers are in service.

2/3 RBCCW pump and heat exchanger are lined up to Unit 3.

A transient occurs resulting in the following conditions:

Reactor scram due to high drywell pressure Drywell pressure is 7.3 psig and climbing at a rate of 1.0 psig per minute.

Bus 24-1 was de-energized and is now powered from the U2 EDG 1 Minute after U2 EDG picked up bus 24-1, 2A RBCCW pump tripped on overcurrent.

What action(s) is(are) required regarding the Unit 2 RBCCW system?

A.

Immediately start 2B RBCCW pump.

B.

Reset the under-voltage on Bus 24-1, then start 2B RBCCW pump.

C.

Isolate RBCCW to the drywell.

D.

Remove 2A RBCCW heat exchanger from service.

Answer:

C Question 12 Details Topic:

12 - 400000.K6.05 Comments:

Objective: 208LN001.12

Reference:

DOA 3700-01 K/A: 400000.K6.05 3.0 / 3.1 K/A: Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: Pumps CFR: 41.7 Safety Function: 8 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: Given the conditions in the stem, all RBCCW pumps on Unit 2 have been lost.

Bus 24-1 is the power supply to RBCCW pump 2B. RBCCW Pump 2B is interlocked to trip on a combination of Core Spray System initiation (+2 psig in the drywell) and loss of normal power to Bus 24-1.

a) The 2B RBCCW can not be restarted under these conditions until leads are lifted, per DEOP 0500-02, that prevent the 2B RBCCW pump from starting.

b) On a loss of Bus 24-1, the undervoltage relay is required to be reset to start loads powered on Bus 29, this action would not allow the 2B RBCCW pump to start, common misconception.

c) DOA 3700-01, LOSS OF COOLING BY RBCCW, immediate actions state: If a LOCA has occurred and drywell pressure is greater than 2 psig, then isolate the drywell RBCCW.

d) If a high drywell pressure signal was not present, the 2B RBCCW pump would be running and the examinee may determine removing the 2A RBCCW heat exchanger from service to prevent overloading the 2B RBCCW pump would be the correct action, common misconception.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 13 of 100 ILT 11-1 (NRC 2012-301) 13 Points: 1.00 Unit 2 was operating at near rated power during the summer months, when a scram occurred.

The following conditions exist:

RPV water temperature is 195oF.

Shutdown Cooling system is in service with the A and B loops.

The 2/3 RBCCW pump and heat exchanger are lined up to Unit 3.

If the 2A RBCCW pump trips on overcurrent, which of the following annunciators would be experienced FIRST?

A.

902-3 G-2, AREA TEMP HI B.

902-4 A-23, SDC HX/FUEL POOL WTR TEMP HI C.

902-4 B-23, SDC PP TRIP D.

902-4 H-4, 2A RECIRC LOOP WTR TEMP HI Answer:

B Question 13 Details Topic:

13 - 205000.A1.08 Comments:

Objective: DRE205001.08

Reference:

DOP 1000-03, DAN 902-4 A-23 K/A: 205000.A1.08 3.1 / 2.9 K/A: Shutdown Cooling System: Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including: Heat exchanger temperatures.

CFR: 41.5 Safety Function: 4 PRA: No Level: High Pedigree: Bank History: 2009 NRC; 2012 NRC Comments: 902-4 A-23 would be received FIRST since its setpoint is >

200oF SDC HX inlet and outlet shell side temp, which would occur when 2A RBCCW pump tripped (reduced RBCCW cooling). 902-3 G-2 would NOT be received first since the area temp hi does not come in until the SDC temps reach 150oF and this condition would not cause the area temperatures to rise that fast. 902-4 B-23 would NOT come in first since the SDC pump trip is a temperature of 339oF on the pump suction and would not occur that fast. 902-4 H-4 would NOT come in first since the recirc temperature of > 330oF would not be reached that fast.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 14 of 100 ILT 11-1 (NRC 2012-301) 14 Points: 1.00 The following conditions exist during a normal reactor startup:

Reactor power

- supercritical with a period of 500 seconds IRMs

- all indicating on scale on range 4 or 5 SRMs

- shorting links are installed SRM channels are indicating:

SRM Channel 21 - 5 x 102 cps SRM Channel 22 - 5 x 105 cps SRM Channel 23 - 4 x 103 cps SRM Channel 24 - 6 x 104 cps These SRM indications result in:

A.

SRM HI and HI-HI alarms ONLY B.

SRM HI and HI-HI alarms and a rod block ONLY C.

SRM HI and HI-HI alarms, a rod block and a half-scram ONLY D.

A full reactor scram Answer:

B Question 14 Details Topic:

14 - 215004.A1.05 Comments:

Objective: DRE215LN004.06

Reference:

DAN 902(3)-5 A-4 and B-12, DGP 1-1, DOP 0700-01 K/A: 215004.A1.05 3.6 / 3.8 K/A: Ability to predict and/or monitor changes in parameters associated with operating the SOURCE RANGE MONITOR (SRM) SYSTEM controls including: SCRAM, rod block, and period alarm trip setpoints CFR: 41.5 Safety Function: 7 PRA: No Level: High Pedigree: Bank History: 2012 NRC; 2006 NRC Comments: The information provided states IRM range switches are on range 4 or 5. SRM rod blocks are bypassed when IRMs are on range 8 or above, therefore the rod blocks are active. The SRM HI-HI scram is only functional when the Shorting Links are removed for refueling operations.

SRM HI and HI-HI alarms will be received, but they will be accompanied by the rod block.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 15 of 100 ILT 11-1 (NRC 2012-301) 15 Points: 1.00 Unit 2 is at 10% of rated thermal power, with APRM 1 having eight (8) of its input LPRMs bypassed, when the following occurs:

An LPRM input to APRM 1 fails DOWNSCALE.

APRM 1 indicates 1%.

The ___(1)___ light(s) will be illuminated and the Operator is required to bypass the failed LPRM

___(2)___.

A.

(1) APRM 1 DOWNSCALE ONLY; (2) ONLY B.

(1) APRM 1 DOWNSCALE ONLY; (2) AND APRM 1.

C.

(1) APRM 1 DOWNSCALE AND INOP; (2) ONLY D.

(1) APRM 1 DOWNSCALE AND INOP; (2) AND APRM 1 Answer:

B Question 15 Details Topic:

15 - 215005.A2.03 (1)

Comments:

Objective: DRE215LN005.03

Reference:

DANs 902-5 C-3 and C-6, DOP 0700-08, DRE215LN005 K/A: 215005.A2.03 3.6 / 3.8 K/A: Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inoperative trip (all causes).

CFR: 41.5 Safety Function: 7 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Explanation: INOP trip setpoint is met when nine (9) LPRMs are bypassed. With 8 input LPRMs into the APRM bypassed ONLY the downscale light would illuminate. BOTH the failed LPRM AND APRM 1 should be bypassed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 16 of 100 ILT 11-1 (NRC 2012-301) 16 Points: 1.00 Unit 2 was operating at full power.

A leak in the 2A Target Rock valve bellows causes alarm 902-4 D-23, 2A TARGET ROCK RELIEF VLV INOP, to annunciate.

Which of the following describes the mode(s), if any, in which the Target Rock WILL be able to operate?

A.

Safety ONLY B.

Relief and ADS ONLY C.

Safety and ADS ONLY D.

The Target Rock valve will NOT operate Answer:

B Question 16 Details Topic:

16 - 239002.A2.02 Comments:

Objective: 239LN001.03

Reference:

DAN 902-4 D-23 K/A: 239002.A2.02 3.1 / 3.2 K/A: Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Leaky SRV CFR: 41.5 Safety Function: 3 PRA: Yes Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Incorrect - DAN 902-4 D-23, a ruptured bellows will cause the Target Rock valve to be inoperable in Safety mode b) Correct - ADS and Relief modes continue to be operable, these modes are independent of the pilot valve bellows and rely on drywell pneumatics to operate, per DEOP 100.

c) Incorrect - Target Rock valve will be inoperable in Safety mode d) Incorrect - Target Rock ADS and Relief modes continue to be operable.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 17 of 100 ILT 11-1 (NRC 2012-301) 17 Points: 1.00 The following plant conditions exist:

SBGT System has received an autostart signal The 2/3 A SBGT train is running in PRIMARY The 2/3 B SBGT train is in STANDBY Which SBGT system damper automatically closed on the autostart signal?

A.

2/3-7509 XTIE DAM B.

2/3-7504A OUTSIDE AIR DAM C.

2/3-7504B OUTSIDE AIR DAM D.

2/3-7506A BACK DRAFT DAMPER Answer:

B Question 17 Details Topic:

17 - 261000.A3.03 Comments:

Objective: DRE 261LN001.03

Reference:

DOS 7500-02 K/A: 261000.A3.03 3.0 / 2.9 K/A: Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: Valve operation CFR: 41.7 Safety Function: 9 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Correct - Upon autostart signal, 2/3 A SBGT outside damper will go close, aligning 2/3 A SBGT suction from the reactor Building or HPCI Gland Exhauster discharge initially. INCORRECT - 2/3 B Outside air damper will remain open drawing in approx 300 scfm via the crosstie line between the SBGT trains, hence the 2/3 crosstie damper is open to allow cooling flow for the train that is in standby. 2/3 A SBGT back draft damper is on the discharge of the 2/3 A SBGT fan which is running, the back draft damper acts as a check valve and will open as soon as the 2/3 A SBGT fan is discharging flow to the 2/3 Chimney The 2/3 B SBGT back draft damper will be closed to prevent possible spread of contamination.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 18 of 100 ILT 11-1 (NRC 2012-301) 18 Points: 1.00 Unit 3 is operating with the following conditions:

Reactor power is 35% and steady Reactor water level is being maintained at 30 inches 3B FRV and LFRV are in Automatic; the 3A FRV is closed in manual.

3B RFP is running; and 3A and 3C Condensate Pumps are running A spurious full scram signal is received - all systems function as designed How will the FWLC system react?

The FWLC system will...

A.

maintain reactor water level at 5 inches without regard to feedwater flow.

B.

maintain reactor water level at 30 inches without regard to feedwater flow.

C.

close the 3B FRV for 25 seconds. After 25 seconds, it will control level at the current setpoint.

D.

maintain the 3B FRV's position for one (1) second. After one (1) second, it will pulse down to 30% of its current demand.

Answer:

D Question 18 Details Topic:

18 - 259002.A3.06 Comments:

Objective: DRE259LN002.06

Reference:

DAN 902(3)-5 G-7 K/A: 259002.A3.06 3.0 / 3.0 K/A: Ability to monitor automatic operations of the REACTOR WATER LEVEL CONTROL SYSTEM including: Reactor water level setpoint setdown following a reactor scram: Plant-Specific CFR: 41.7 Safety Function: 2 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: Upon a scram signal, the system takes the setpoint to -10 and locks the FRVs in place for 1 second.

At 1 second (T1) the programming looks at reactor water level and makes one of the following decisions: If reactor water level is greater than -30, then:

The system pulses the FRVs to 30% of current demand, any valve in manual, receives a closed signal. 25 seconds later (T26) the system releases the valves in auto to control at the current setpoint. 19 seconds later (T45) the system ramps back to a setpoint of +30, at a rate of 10/minute.

If reactor water level is less than -30, then: The system releases the FRVs to control at the current setpoint (assuming water level is low enough to NOT enter the HPCI and IC steam lines).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 19 of 100 ILT 11-1 (NRC 2012-301) 19 Points: 1.00 Which of the following describe the (1) location and (2) action necessary, if any, to determine Unit 2 125VDC battery voltage from the Main Control Room?

A.

(1) 902-8 (2) no additional action required B.

(1) 902-8 (2) verify meter selector switch is in the '125V' position C.

(1) 923-2 (2) no additional action required D.

(1) 923-2 (2) verify meter selector switch is in the '125V' position Answer:

B Question 19 Details Topic:

19 - 263000.A4.02 Comments:

Objective: DRE263LN001.11

Reference:

DAN 902(3)-8 E-9, K/A: 263000.A4.02 3.2 / 3.1 K/A: Ability to manually operate and/or monitor in the control room:

Battery voltage indicator: Plant-Specific CFR: 41.7 Safety Function: 6 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: An LED type indicator with a selector switch is located on the 902(3)-8 panel. The switch can be positioned to display the voltage of the unit's 250 / 125 / 24/48 VDC batteries.The sensing point for battery voltage is upstream of the selected battery to main bus breaker so if a trip of the breaker occurs, voltage is still shown in the display.

The 923-2 panel has voltmeters and selector switches for the 345Kv system REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 20 of 100 ILT 11-1 (NRC 2012-301) 20 Points: 1.00 Unit 2 was operating at full power when a transient occurred causing the following conditions:

Unit 2 scrammed TR-22 trip due to diffierential current Drywell pressure is 11 psig and slowly rising Reactor pressure is 300 psig and steady Reactor water level is -150 inches and slowly lowering Reactor Feed Pumps and HPCI have failed The unit supervisor has directed to use LPCI injection from the CST to recover reactor water level per DEOP 0500-03.

What actions are required?

A.

Line up one LPCI pump in EACH loop to take suction from the CST Verify closed BOTH LPCI loops containment spray and cooling valves Verify open the 2-1501-21 valve in BOTH LPCI loops B.

Line up one LPCI pump in EACH loop to take suction from the CST Verify closed BOTH LPCI loops containment spray and cooling valves Verify open the 2-1501-21 valve in the LPCI loop selected for injection ONLY C.

Line up one LPCI pump to take suction from the CST ONLY Verify closed containment spray and cooling valves for the loop lined up to take suction from the CST ONLY Verify open the 2-1501-21 valve in the LPCI loop selected for injection ONLY D.

Line up one LPCI pump to take suction from the CST ONLY Verify closed BOTH LPCI loops containment spray and cooling valves Verify open the 2-1501-21 valve in the LPCI loop selected for injection ONLY Answer:

D Question 20 Details Topic:

20 - 203000.A4.05 Comments:

Objective: 203LN001.02

Reference:

DEOP 500-3 K/A: 203000.A4.05 4.3 / 4.1 K/A: RHR/LPCI: Injection Mode (Plant Specific) Ability to manually operate and/or monitor in the control room: Manual initiation controls CFR: 41.7 Safety Function: 2 & 4 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a &b) Only one pump is lined up at a time to take suction form the CST due to NPSH and vortexing concerns.

c) A knowledge of system flow paths is required to know that the system crosstie would allow some water from the CST into the opposite divisions containment spray and cooling valves, which is not allowed per procedure.

d) DEOP 0500-3 directs aligning one pump to take suction from the CST, closing all LPCI containment cooling and spray valves, and injecting only through the LPCI loop that was automatically selected for injection upon receiving a high drywell pressure signal.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 21 of 100 ILT 11-1 (NRC 2012-301) 21 Points: 1.00 Given the following sequence of events:

01:00

'B' LPCI subsystem is declared inoperable 02:00

'B' Core Spray subsystem is declared inoperable 03:00

'A' LPCI subsystem is declared inoperable 04:00

'A' Core Spray subsystem is declared inoperable Which of the following is the earliest time entry into LCO 3.0.3 is required?

A.

02:00 B.

03:00 C.

04:00 D.

05:00 Answer:

A Question 21 Details Topic:

21 - 209001.G.2.2.38 Comments:

Objective: DRE209LN001.07

Reference:

T.S. 3.5.1 K/A: 209001.G.2.2.38 3.6 / 4.5 K/A: Low Pressure Core Spray System: Knowledge of conditions and limitations in the facility license.

CFR: 41.10 Safety Function: 2 & 4 PRA: No Level: High Pedigree: New History: 2012 NRC Explanation: Per Tech Spec 3.5.1, when operating in mode 1, all ECCS systems are required to be operable, which includes BOTH Core Spray systems. If any two or more low pressure ECCS subsystems are inoperable LCO 3.0.3 is required to be entered immediately.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 22 of 100 ILT 11-1 (NRC 2012-301) 22 Points: 1.00 Unit 2 was operating at near rated power, when the following set of conditions occurred:

A scram on high DW pressure.

A loss of off-site power (LOOP).

The 2/3 cribhouse inlet temperature is 93oF and steady.

Given the above conditions, a maximum of ___(1)___ CCSW pumps may be run concurrently, to prevent ___(2)___.

A.

(1) 2; (2) overloading the 2/3 EDG B.

(1) 3; (2) overloading the 2/3 EDG C.

(1) 2; (2) choking the DGCWP flow to the 2/3 EDG D.

(1) 3; (2) choking the DGCWP flow to the 2/3 EDG Answer:

C Question 22 Details Topic:

22 - 400000.G.2.1.32 Comments:

Objective: DRE277LN001.13

Reference:

DOP 1500-02, DOP 6600-05, UFSAR 6.3, DGA 12 K/A: 400000.G.2.1.32 3.8 / 4.0 K/A: Ability to explain and apply system limits and precautions.

CFR: 41.10 Safety Function: 8 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC, 2007 NRC Explanation: With the given cribhouse inlet temperature a max of 2 CCSW pumps may be run with the 2/3 EDG supplying Bus 23-1. This will prevent choking the DGCWP from the 2/3 EDG. This is an industry event at Dresden. The distractor for 3 CCSW pumps is wrong unless temp is between 83oF and 90oF. The distractors for overloading the 2/3 EDG is wrong because the actions of DGA 12 ensure power is distributed so that the CCSW pumps can run without overloading the EDGs. Given that there is a LOOP and Scram, actions of DGA 12 will be in progress.

REQUIRED

REFERENCE:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 23 of 100 ILT 11-1 (NRC 2012-301) 23 Points: 1.00 The Unit 2/3 Emergency Diesel Generator is being run per DOS 6600-01, DIESEL GENERATOR SURVEILLANCE TESTS, for the monthly run, when the EO reports that the local annunciator DG 2/3 B-4, DIESEL ENGINE TEMP HIGH, is in alarm.

The EO reports that the local engine temperature is 215oF.

Per the above DOS, which of the actions below are required to be taken NEXT?

A.

Close the Emergency Fuel shutoff valve.

B.

Reverse Diesel Generator cooling water flow.

C.

Shutdown the Diesel Generator using the fuel injection rack lever.

D.

Swap the 2/3 Diesel Generator Cooling Water Pump to its alternate power supply.

Answer:

C Question 23 Details Topic:

23 - 264000.K4.07 Comments:

Objective: DRE264LN001.05

Reference:

DOS 6600-01, DAN DG2/3 B-4 K/A: 264000.K4.07 3.3 / 3.4 K/A: Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: Local operation and control CFR: 41.7 Safety Function: 6 PRA: No Level: High Pedigree: Bank History: 2012 NRC Explanation: With the Diesel Generator engine temperature > 200oF, the Diesel should have tripped. Per the limitations and actions of the DOS, if the automatic or manual D/G shutdown fails to stop the D/G, then the preferred method for D/G shutdown is to trip by using the fuel injection rack lever.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 24 of 100 ILT 11-1 (NRC 2012-301) 24 Points: 1.00 Unit 2 is operating at near rated power, with Bus 25 OOS for maintenance, when the following occurs:

Bus 28 experiences an overcurrent condition.

What affect (if any) does this have on the Unit 2 24/48 VDC system?

A.

ONLY the 'A' battery chargers lose their AC power source.

B.

ONLY the 'B' battery chargers lose their AC power source.

C.

BOTH the 'A' and 'B' battery chargers will remain available.

D.

BOTH the 'A' AND 'B' battery chargers lose their AC power source.

Answer:

D Question 24 Details Topic:

24 - 263000.K1.01 (2)

Comments:

Objective: DRE263LN003.12

Reference:

DOP 6800-02, DOA 6900-01 K/A: 263000.K1.01 3.3/3.5 K/A: Knowledge of the physical connections and/or causeeffect relationships between D.C. ELECTRICAL DISTRIBUTION and the following: A.C. electrical distribution CFR: 41.7 Safety Function: 6 PRA: No Level: High Pedigree: Bank History: 2012 NRC, 2008 Cert Explanation: All 4 (2 positive and 2 negative) chargers are powered from the Instrument Bus. With Bus 25 OOS, MCC 25-2 has no power. With Bus 28 going overcurrent, MCC 28-2 loses power. Without these two MCCs, the Instrument Bus has no power, and subsequently all 4 chargers lose AC power.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 25 of 100 ILT 11-1 (NRC 2012-301) 25 Points: 1.00 A Unit 2 startup is in progress.

All four SRMs are fully inserted and are reading approximately 10,000 CPS.

Five minutes later the NSO observes:

SRM 21 is now reading approximately 5,000 cps and trending down SRMs 22, 23 and 24 are still indicating 10,000 cps All other reactor plant parameters are unchanged Which of the following operations / malfunctions would cause the observed indications?

A.

24/48 VDC Bus A voltage is trending down at 0.1 VDC per minute.

B.

SRM 21 detector has developed a gas leak.

C.

SRM 21 SELECT switch has been depressed and is lit.

D.

SRM 21 RAMP switch has been placed in the RAMP position.

Answer:

B Question 25 Details Topic:

25 - 215004.K6.04 Comments:

Objective: DRE215LN004.02

Reference:

BWR Fundamentals Ch.7 Sensors/Detectors K/A: 215004.K6.04 2.9 / 2.9 K/A: Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRM) SYSTEM : Detectors CFR: 41.7 Safety Function: 7 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: BWR Fundamentals states: In Fission chambers, if the gas pressure decreases, there will be less gas atoms to undergo ionization, therefore less electrons collected. This means a smaller electrical current will exist and the indication will fail downscale.

a) Both SRM 21 and 22 would respond to a decreasing voltage.

b) A gas leak from an SRM detector will cause the SRM to read low.

c) A lit SELECT switch would allow the SRM to be withdrawn. But withdraw for the SRM is not a seal-in signal (insert is a seal-in signal) d) the RAMP switch being placed in RAMP would only be seen if the SRM MODE switch was also in PERIOD which would not allow the SRM to display reactor power.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 26 of 100 ILT 11-1 (NRC 2012-301) 26 Points: 1.00 Unit 2 is in STARTUP Drywell and Torus inerting are in progress.

A spurious Group 2 isolation occurs.

Which of the following describes the impact (if any) on Drywell and Torus inerting?

A.

No impact, inerting of the Drywell and Torus remains in progress.

B.

Drywell and Torus vent valves close ONLY C.

Nitrogen supply valves close ONLY.

D.

Drywell and Torus vent valves AND nitrogen supply valves close.

Answer:

D Question 26 Details Topic:

26 - 223002.K3.28 Comments:

Objective: 223LN005.

Reference:

902(3)-5 E-5 K/A: 223002.K3.28 2.7 / 2.8 K/A: Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following: Containment nitrogen inerting system CFR: 41.7 Safety Function: 5 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: Upon receipt of a Group 2 signal, the Drywell and Torus vent valves (AO 2-1601-23 and 60) and the nitrogen supply valves (AO 1601-57, 58 and 59) will go closed. This will cut off the supply of nitrogen inerting the Torus and the Drywell and isolate the vent path from the Torus and the Drywell.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 27 of 100 ILT 11-1 (NRC 2012-301) 27 Points: 1.00 Given the following:

Unit 2 is operating at full power.

CRD exercising is in progress.

Rod L7 is moved to position 48.

RPIS indication for rod L7 indicated '48', then RED '--', then went out.

What is this an indication of and what is(are) the NSO required actions?

A.

Rod overtravel; attempt to recouple the rod B.

Rod overtravel; bypass the RWM and drive the rod to 00 C.

Failed open RPIS reed switch; enter a substitute position D.

Failed open RPIS reed switch; bypass the RWM and drive the rod to 00 Answer:

A Question 27 Details Topic:

27 - 201003.K1.03 Comments:

Objective: 201LN002.08

Reference:

DOA 0300-05; LP DRE201LN002 K/A: 201003.K1.03 3.1 / 3.1 K/A: Knowledge of the physical connections and/or cause effect relationships between CONTROL ROD AND DRIVE MECHANISM and the following: RPIS CFR: 41.6 Safety Function: 1 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) Correct, DOA 0300-05 actions for uncoupled rod with reactor power greater than 10%

b) Incorrect, This is the action for an uncoupled rod if reactor power less than 10%

c & d) Incorrect, Open reed switches would cause RPS indication to be blank.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 28 of 100 ILT 11-1 (NRC 2012-301) 28 Points: 1.00 The following plant conditions exist:

Unit 2 is at 95% power 2C LPCI pump is operating for surveillance testing 2A and 2D CCSW pumps are operating for surveillance testing LPCI full flow test valves 2-1501-38B & 20B are positioned to yield 5000 gpm to the torus Then the following occurs:

MCC 28-1 trips on overcurrent.

Drywell pressure trends up to 2.5 psig due to a small steam leak.

Which one of the following describes the response of the LPCI system under these conditions?

A.

The 2C LPCI pump will trip The test line valves will close B.

The 2C LPCI pump will trip The test line valves will fail as is C.

The 2C LPCI pump will NOT trip The test line valves will close D.

The 2C LPCI pump will NOT trip The test line valves will fail as is Answer:

C Question 28 Details Topic:

28 - 219000.K2.01 Comments:

Objective: 203LN001.06

Reference:

DAN 902(3)-5 D-11; DOP 6700-19 K/A: 219000.K2.01 2.5 / 2.9 K/A: RHR/LPCI: Torus/Suppression Pool Cooling Mode: Knowledge of electrical power supplies to the following: Valves CFR: 41.7 Safety Function: 5 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: High Drywell pressure (+2 psig) is a LPCI initiation signal - All four pumps will get a start signal, the CCSW pumps will receive a trip signal with high drywell pressure. A LPCI loop will be selected for injection to the RPV, and all test/containment cooling valves will get a close signal. 2-1501-38B & 20B are powered from MCC 29-4 and will close as designed. 2-1501-38A & 20A are powered from MCC 28-1 and would fail as is.

a) credible if examinee believes running LPCI pumps will trip initially (like the running CCSW pumps) common misconception.

b & d) credible if examinee believes MCC 28-1 is the power to the B subsystem valves.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 29 of 100 ILT 11-1 (NRC 2012-301) 29 Points: 1.00 Units 2 and 3 are operating at full power.

Due to ongoing work in the AEER, the 125 VDC Bus 2A-1 supply to the AEER halon fire panel is secured.

A condition requiring initiation of fire suppression equipment occurs.

Regarding AUTOMATIC fire suppression in the AEER, which of the following warnings (if any) would alert personnel working in the AEER?

A.

Flashing light and siren due to the automatic initiation of CO2 injection ONLY.

B.

Flashing light and siren due to the automatic initiation of Halon injection ONLY.

C.

Flashing light and siren due to the automatic initiation of Halon AND CO2 injection.

D.

NO warning, Halon and CO2 injection must be initiated manually.

Answer:

D Question 29 Details Topic:

29 - 286000.K3.02 Comments:

Objective: 286LN002.12

Reference:

LP DRE 286LN002 K/A: 286000.K3.02 3.2 / 3.4 K/A: Knowledge of the effect that a loss or malfunction of the FIRE PROTECTION SYSTEM will have on following: Personnel protection CFR: 41.7 Safety Function: 8 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

An automatic initiation of Halon or CO2 will will cause a flashing light and siren to activate in the area of the sensed fire for 60 seconds prior to injection.

All CO2 fire panels have redundant power sources, the AEER Halon panel only has a single power source.

CO2 systems only initiate automatically to the EDG rooms and must always be manually initiated in the AEER.

Securing 125 VDC power to the AEER halon panel, would inhibit automatic initiation and require Halon to be manually initiated.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 30 of 100 ILT 11-1 (NRC 2012-301) 30 Points: 1.00 Which of the following conditions would cause the 902-5 A-3, ROD DRIFT, annunciator to alarm?

Reed switch at position 35 ____(1)____, ____(2)____ a rod motion signal present.

A.

(1) closed (2) with B.

(1) closed (2) without C.

(1) open (2) with D.

(1) open (2) without Answer:

B Question 30 Details Topic:

30 - 214000.K4.01 Comments:

Objective: 201LN002.06

Reference:

DAN 902(3)-5 A-3 K/A: 214000.K4.01 K/A: Knowledge of ROD POSITION INFORMATION SYSTEM design feature(s) and/or interlocks which provide for the following: Reed switch locations CFR: 41.7 Safety Function: 7 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Incorrect - requires odd reed switch without rod motion command signal.

b) Correct - Per DAN 902(3)-5 A-3, ROD DRIFT, requires Control Rod moving past an ODD numbered reed switch (closed) with NO control rod motion requested.

c) Incorrect - requires odd reed switch without rod motion command signal.

d) Incorrect - requires odd reed switch without rod motion command signal.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 31 of 100 ILT 11-1 (NRC 2012-301) 31 Points: 1.00 Which of the following describes how the operation of the Backup Scram Solenoid Valves will result in a reactor scram?

When actuated, the Backup Scram Solenoid Valves...

A.

relieve hydraulic control unit drive pressure to the exhaust header.

B.

isolate and vent air downstream of the Scram Dump Valves.

C.

align the overpiston volume of the rod drive mechanism to the exhaust header.

D.

isolate and vent the air header to the Scram Inlet and Outlet Valves.

Answer:

D Question 31 Details Topic:

31 - 201001.K5.08 Comments:

Objective: 201LN001.06

Reference:

M-34 K/A: 201001.K5.08 2.5 / 2.6 K/A: Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM :

Solenoid operated valves CFR: 41.5 Safety Function: 1 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments:

For added protection, the Instrument Air header to all Scram Pilot Valves (2(3)-305-117 and 2(3)-305-118) has a pair of Backup Scram Valves (2(3)-302-19A and 2(3)-302-19B). When energized, the Backup Scram Valves isolate and vent the instrument air header to the Scram Inlet (2(3)-

305-126) and Scram Outlet Valves (2(3)-305-127) to rapidly insert all control rods.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 32 of 100 ILT 11-1 (NRC 2012-301) 32 Points: 1.00 Which of the following is indicative of debris from the Drywell being discharged through the downcomers during a small LOCA?

A.

fluctuating HPCI flow B.

fluctuating CCSW flow C.

fluctuating Torus spray flows D.

fluctuating ERV tail pipe pressures Answer:

C Question 32 Details Topic:

32 - 230000 K6.05 Comments:

Objective: DRE203LN001.12

Reference:

DAP 05-09 K/A: 230000 K6.05 K/A: Knowledge of the effect that a loss or malfunction of the following will have on the RHR/LPCI: TORUS/SUPPRESSION POOL SPRAY MODE :

Suppression pool CFR: 41.7 Safety Function: 5 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: Debris discharged through the downcomers would end up in the Torus and cause clogging of the ECCS suction strainers.

a) Incorrect - during a small LOCA, HPCI suction would be from the CSTs and would not be effected by debris in the Torus.

b) Incorrect - CCSW system flow would not be effected by debris in the torus. if debris from the torus were to make it to the LPCI/CCSW coolers, only LPCI flow would be effected.

c) Correct - LPCI provides torus sprays and its suction is aligned to the torus through the ECCS suction strainers.

d) Incorrect - While ERV tailpipes are in the torus and pressures do fluctuate during valve operation, it is not indicative of debris in the torus.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 33 of 100 ILT 11-1 (NRC 2012-301) 33 Points: 1.00 Unit 3 is conducting a special test with the Main Generator conditions as follows:

Real load is 800 MWe Terminal Voltage is 17.1 kV Reactive Load is 0 MVAR Voltage Regulator Mode Switch is in MANUAL Generator hydrogen pressure is 30 psig If an operator placed the Voltage Regulator Mode Switch to AUTO, and the voltage regulator failed to its MINIMUM limit for the current voltage, which of the following indicates the final value of Main Generator reactive loading three minutes later (i.e. plant stable)?

(Assume terminal voltage remains constant.)

A.

positive (+) 220 MVAR B.

negative (-) 70 MVAR C.

negative (-) 120 MVAR D.

negative (-) 220 MVAR Answer:

B Question 33 Details Topic:

33 - 245000.A1.08 Comments:

Objective: 245LN003.11

Reference:

DOP 6400-08 K/A: 245000.A1.08 2.5 / 2.7 K/A: Ability to predict and/or monitor changes in parameters associated with operating the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS controls including: Generator output voltage/reactive load CFR: 41.5 Safety Function: 6 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments:The examinee must identify the correct Minimum Excitation Limiter (MEL) curve from the supplied references. The MEL limits the excitation dependent on the current terminal voltage of the generator. The point on the graph will move straight down the 800MWE line until it reaches the 17.1 Kv limit Distractors indicates a misunderstanding of system operation or inability to read the graph.

REQUIRED

REFERENCES:

Figures 5, 6 & 7 from DOP 6400-08

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 34 of 100 ILT 11-1 (NRC 2012-301) 34 Points: 1.00 Unit 2 is operating at 75% power.

The following events occur simultaneously:

2D ERV spuriously opens 2D ERV vacuum breaker fails open What is (1) the expected initial response of primary containment pressure and (2) the IMMEDIATE action taken to mitigate the event?

A.

(1) Torus pressure increase due to steam in the torus (2) Place the 2D ERV control switch in OFF B.

(1) Torus pressure increase due to steam in the torus (2) Initiate Torus sprays C.

(1) Drywell pressure increase due to steam in the drywell (2) Place the 2D ERV control switch in OFF D.

(1) Drywell pressure increase due to steam in the drywell (2) Initiate Drywell sprays Answer:

C Question 34 Details Topic:

34 - 223001.A2.09 Comments:

Objective: 223LN001.02

Reference:

Dresden 2/3 PSTGs; DEOP 200-1 K/A: 223001.A2.09 3.4 / 3.6 K/A: Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Vacuum breaker malfunction CFR: 41.5 Safety Function: 5 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) Incorrect - With the vacuum breaker failed open, steam will be discharged to the drywell.

b) Incorrect - With the vacuum breaker failed open, steam will be discharged to the drywell.

c) Correct - With the vacuum breaker failed open, 2D ERV will discharge to the drywell and per DOA 0250-01, the correct immediate action is to place the 2D ERV control switch to off.

d) Incorrect - Drywell sprays is a action directed by the Unit Supervisor per DEOP 200-1 after drywell pressure has exceeded 9 psig.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 35 of 100 ILT 11-1 (NRC 2012-301) 35 Points: 1.00 Unit 2 Emergency Diesel Generator (U2 EDG) was running for a surveillance, when the following alarms are received in the Main Control Room:

902-8 C-8, U2 DIESEL ROOM TEMP HI XL3 23-23, 2 DG ROOM FIRE B1 Which of the following is an expected automatic action as a result of the alarms above?

A.

U2 EDG trips B.

Diesel fuel oil transfer pump trips C.

U2 EDG room vent fan louvers (first floor Turbine Building) trip D.

U2 EDG air intake modulator dampers (second floor Turbine Building) close Answer:

C Question 35 Details Topic:

35 - 288000.A3.01 Comments:

Objective: DRE288LN001.06

Reference:

DAN XL3 23-23 K/A: 288000.A3.01 3.8 / 3.8 K/A: Ability to monitor automatic operations of the PLANT VENTILATION SYSTEMS including: Isolation/initiation signals.

CFR: 41.7 Safety Function: 9 PRA: No Level: Memory Pedigree: New History: 2012 NRC Explanation:

a) Incorrect - High temp and fire are not trips of the U2 EDG b) Incorrect - The fuel oil transfer pump does not receive a trip signal from a fire condition.

c) Correct - Per DAN XL3 23-23, this indicates a fire and the room vent fan louvers trip automatically.

d) Incorrect - The air intake modulator damper does not receive a trip signal from a fire condition.

REQUIRED

REFERENCE:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 36 of 100 ILT 11-1 (NRC 2012-301) 36 Points: 1.00 Unit 2 was operating at 40% power with all APRMs in service.

Rod A-8 is selected.

A fault cause APRM channel 4 to fail down scale.

Based on these conditions, with no operator action, Rod Block Monitor (RBM) channel 8 A.

remains in service and reset B.

is bypassed C.

shifts to its alternate reference APRM D.

generates a RBM DOWNSCALE rod block Answer:

B Question 36 Details Topic:

36 - 215002.A4.03 Comments:

Objective: 215LN002.08

Reference:

LP DRE215LN002 K/A: 215002.A4.03 2.8 / 2.8 K/A: Ability to manually operate and/or monitor in the control room: Trip bypasses: BWR-3,4,5 CFR: 41.7 Safety Function: 7 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: Examinee must know that APRM 4 is the normal reference APRM for RBM channel 8 and A-8 is an edge rod.

a) would be correct if any other APRM were to fail downscale.

b) Correct, When APRM 4 fails down scale, RBM 8 sees reactor power less than 30%. Reactor power less than 30% causes automatic bypass of the RBM, An edge rod also causes an automatic bypass of the RBM.

c) RBM will shift to its alternate reference APRM only after its normal reference APRM is bypassed by an operator.

d) RBM DOWNSCALE rod block occurs when th LPRM inputs to a RBM channel average less than 7.1% with the reference APRM >30%.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 37 of 100 ILT 11-1 (NRC 2012-301) 37 Points: 1.00 Given the following:

Unit 2 Feed Water flow is 4.7 Mlbm/hr 2A AND 2B Reactor Feedpumps are in service 2A, 2B, AND 2C Condensate pumps are in service No Condensate pump is selected for standby due to standby selector switch repairs The 2C condensate pump trips Reactor Feedpump suction pressure drops to 110 psig Annunciator 902-6 G-8, RFP SEQUENTIAL TRIP SUCTION PRESS LO, comes in.

(1) Predict the response of the Feedwater pumps (2) What immediate operator actions must be taken?

A.

(1) One Reactor Feedpump trips immediately (2) Run Recirc pumps to minimum.

B.

(1) One Reactor Feedpump trips immediately (2) Start the 2D condensate pump.

C.

(1) One Reactor Feedpump trips after 3 seconds (2) Run Recirc pumps to minimum.

D.

(1) One Reactor Feedpump trips after 3 seconds (2) Start the 2D condensate pump.

Answer:

D Question 37 Details Topic:

37 - 259001.G.2.4.31 Comments:

Objective: 259LN001.10

Reference:

DAN 902-6 G-8, 902-6 F-5, 902-6 F-7, DOP 3200-2 K/A: 259001.G.2.4.31 4.2 / 4.1 K/A: Reactor Feedwater System: Knowledge of annunciator alarms, indications, or response procedures.

CFR: 41.10 Safety Function: 2 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: These condition lead to a sequential trip of a reactor feedpump after a time delay. Per the DANs the standby condensate/condensate booster pump should be started. A & C are incorrect because the RFP will not trip immediately unless suction pressure drops to 100 psig or less. The student may incorrectly relate the automatic Recirc runback on the trip of a fourth condensate pump to the condition of low suction pressure, if so reducing Recirc flow to within the capacity of the running condensate pumps would be an appropriate action.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 38 of 100 ILT 11-1 (NRC 2012-301) 38 Points: 1.00 With Unit-2 at 95% power, the following transient is observed:

Rise in indicated core flow.

Drop in core thermal power.

Drop in Main Generator power.

Drop in Core differential pressure.

Which of the following is causing the above operating anomaly?

A.

Failure of a jet pump B.

Partial opening of an ERV C.

Failure of a condensate demineralizer post strainer D.

Loss of speed mismatch signal from an operating recirc pump Answer:

A Question 38 Details Topic:

38 - 202001.K5.02 Comments:

Objective: 20200LK016

Reference:

DOA 0201-01 K/A: 202001.K5.02 3.1 / 3.2 K/A: Knowledge of the operational implications of the following concepts as they apply to RECIRCULATION SYSTEM: Jet pump operation: BWR-3,4,5,6 CFR: 41.5 Safety Function: 1 & 4 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: DOA 0201-01 states the symptoms of a failed Jet Pump are:Rise in indicated core flow, drop in core thermal power, drop in main generator power and drop in core D/P. None of the distractors would have all of the symptoms listed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 39 of 100 ILT 11-1 (NRC 2012-301) 39 Points: 1.00 Unit 2 is operating at 75% power in DOME pressure control when 2 of the 3 EHC pressure controller processors (<R2> and <T2>) fail low.

As a result of this failure, turbine control valves will...

A.

close resulting in reactor scram on high pressure or flux.

B.

open and the reactor will scram due to MSIV closure on low reactor pressure.

C.

remain at current demand as pressure control transfers to THROTTLE active mode.

D.

open but the Max Combined Flow Limiter will prevent an excessive pressure reduction.

Answer:

A Question 39 Details Topic:

39 - 295005.K1.01 Comments:

Objective: 241LN001.08

Reference:

LP DRE241NL001A; DOA 5650-02 K/A: 295005.AK1.01 4.0 / 4.1 K/A: Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP : Pressure effects on reactor power CFR: 41.8 Safety Function: 3 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: DOA 5650-01 states: IF 2 out of 3 DEHC Pressure Controller Processors fail low, THEN Turbine Control Valves AND Turbine Bypass Valves close, causing a reactor scram on high pressure. The triple modular redundancy (TMR) feature of DEHC requires 2 of the 3 pressure controller processors to be functioning in order for the system to work.

The failure indicated here would result in a loss this redundancy and thus the system will trip and close the control valves. Consequently, reactor pressure and flux will rise to the scram setpoint.

b) Incorrect - this would be the result if the control or bypass valves fail open.

c) Incorrect - this occurrs when dome exceeds throttle pressure by > 200 psig.

d) Incorrect - this would be the result if the pressure control regulator failed high.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 40 of 100 ILT 11-1 (NRC 2012-301) 40 Points: 1.00 Unit 2 is operating at full power.

When the following occurs:

A fire is reported in the 2/3 Cribhouse DOA 0010-10, FIRE-EXPLOSION, is entered.

Which of the following are immediate operator actions of DOA 0010-10 to initiate fire fighting actions?

1. Notify security to respond to the scene

2. Notify Coal City Fire Protection District

3. Announce fire location on plant radio system

4. Announce fire location on plant PA system A.

1, 2, 3 B.

1, 2, 4 C.

1, 3, 4 D.

2, 3, 4 Answer:

C Question 40 Details Topic:

40 - 600000.AK1.02 Comments:

Objective: 29501LK080

Reference:

DOA 0010-10 K/A: 600000.AK1.02 2.9 / 3.1 K/A: Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire Fighting CFR: 41.10 Safety Function: 8 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a, b & d) Incorrect, Notify Coal City FPD is an immediate action action of DOA 0010-10 only if the fire is located outside the protected area. The 2/3 Cribhouse is located inside the protected area.

c) Correct, Per DOA 0010-10, Notify security to respond to scene, Announce fire location on plant radio and Announce fire location on plant PA are immediate actions if the fire is located in the protected area.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 41 of 100 ILT 11-1 (NRC 2012-301) 41 Points: 1.00 DEOP 400-5 "Failure to Scram" allows the use of the Core Spray system as an Alternate System, while HPCI and LPCI are preferred systems.

What is the reason for restricting injection with Core Spray?

A.

It would cool the SBLC solution enough to cause excessive boron precipitation.

B.

It would cause boron to concentrate in the upper core region, thereby lowering its effectiveness at reducing core power.

C.

It would flash to steam in the core region pushing boron out of the core.

D.

It would displace boron in the entire core region, replacing it with cold, unborated water.

Answer:

D Question 41 Details Topic:

41 - 295037.EK1.03 Comments:

Objective: 29502LK033

Reference:

DEOP 400-5 bases K/A: 295037.EK1.03 4.2 / 4.4 K/A: Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Boron effects on reactor power (SBLC)

CFR: 41.8 Safety Function: 1 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Per DEOP 400-5 bases, spraying core spray would remove boron from the entire core region, replacing it with cold, unborated water, resulting in power excursion problems.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 42 of 100 ILT 11-1 (NRC 2012-301) 42 Points: 1.00 Unit 2 was operating at near rated power, when a LOCA occurred.

If RPV pressure is 100 psig and lowering slowly, which of the following set of parameters would result in a USEABLE indicated wide range RPV water level?

Drywell temperature of ___(1)___ and an indicated wide range RPV water level of ___(2)___.

A.

(1) 257 oF (2) -19 inches B.

(1) 257 oF (2) -23 inches C.

(1) 357 oF (2) +12 inches D.

(1) 357 oF (2) -11 inches Answer:

A Question 42 Details Topic:

42 - 295028.K2.03 Comments:

Objective: DRE216LN001.12

Reference:

DEOP 200 figures A, B, and C, (RPV Water Level Instruments)

K/A: 295028.K2.03 3.6 / 3.8 K/A: Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Reactor water level indication.

CFR: 41.7 Safety Function: 5 PRA: No Level: High Pedigree: Bank History: 2012 NRC; 2010 NRC Explanation: The examinee must utilize both the RPV saturation temperature and minimum indicated level tables of the above DEOP to determine that the only values that would result in a useable indication is a Drywell temperature of 257oF and a RPV level of -19 inches.

REQUIRED

REFERENCES:

DEOP 200 Figures A, B, and C, (RPV Water Level Instruments).

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 43 of 100 ILT 11-1 (NRC 2012-301) 43 Points: 1.00 A transient has occurred resulting in the following conditions:

Drywell pressure 42 psig and stable.

Torus bottom pressure is 55 psig and stable.

Primary containment water level is 27 feet and slowly increasing.

RPV level -173" and slowly decreasing.

The Unit Supervisor has directed venting of the primary containment to lower pressure.

Which of the following vent paths should the NSO align first?

A.

torus via the APCV system B.

torus via the SBGT system C.

drywell via the APCV system D.

drywell via the SBGT system Answer:

B Question 43 Details Topic:

43 - 295024.EK2.18 Comments:

Objective: 29502LK063

Reference:

DEOP 0500-04 K/A: 295024.EK2.18 3.3 / 3.4 K/A: Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following: Ventilation CFR: 41.7 Safety Function: 5 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments:

a) Incorrect because per the UFSAR drywell design pressure is 62 psig, 59 psig does not require venting through APCV b) Correct, per DEOP 500-4, with the conditions given, the correct vent path is the torus via SBGT.

c) & d) Incorrect, If torus level is greater than 30 feet, venting from the drywell is directed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 44 of 100 ILT 11-1 (NRC 2012-301) 44 Points: 1.00 Unit 2 was operating at near rated power, with the following conditions:

The Unit 2A 125 VDC Battery Charger is supplying the Unit 2 125 VDC system.

The Unit 2/3 250 VDC Battery Charger is supplying the Unit 2 250 VDC system.

Then a fire caused Bus 28 to de-energize.

(1) What effect will this have on the batteries AND (2) what actions can be taken to mitigate the transient?

A.

(1) The Unit 2 125 VDC batteries ONLY will begin to discharge; (2) place the Unit 2 125 VDC Battery Charger in service.

B.

(1) The Unit 2 250 VDC batteries ONLY will begin to discharge; (2) place the Unit 2 250 VDC Battery Charger in service.

C.

(1) The Unit 2 125 AND 250 VDC batteries will begin to discharge; (2) place the Unit 2 125 AND Unit 2 250 VDC Battery Chargers in service.

D.

(1) The Unit 2 125 AND 250 VDC batteries will begin to discharge; (2) place the Unit 2 125 Alternate Battery AND Unit 2 250 VDC Battery Charger in service.

Answer:

A Question 44 Details Topic:

44 - 295004.AK2.01 Comments:

Objective: DRE263LN001.02

Reference:

DOA 6900-04, DOA 6900-T1 K/A: 295004.AK2.01 3.1 / 3.1 K/A: Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: Battery charger CFR: 41.7 Safety Function: 6 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: Explanation: With the 2A charger (MCC 28-2) supplying the 125 VDC system and the 2/3 charger (MCC 29-2 or 39-2) supplying the 250 VDC system, a loss of Bus 28 will cause ONLY the 125 VDC 2A charger to lose power, which will cause the 125 VDC batteries to begin to discharge. Transferring the 125 VDC system to the Unit 2 125 VDC charger is the correct action.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 45 of 100 ILT 11-1 (NRC 2012-301) 45 Points: 1.00 Given the following conditions:

Unit 2 is at full power.

2A and 2B Instrument Air Compressors (IAC) are running.

2A IAC trips due to low lube oil pressure.

Unit 2 Instrument Air pressure is 78 psig and dropping slowly.

An operator has been dispatched to start and align 3C IAC to Unit 2.

The reason for this action is to prevent the __________.

A.

FRVs from failing closed B.

outboard MSIVs from drifting closed C.

Target Rock from becoming inoperable in ERV mode D.

Unit 2 Instrument Air to Unit 2 Service Air cross tie from opening Answer:

B Question 45 Details Topic:

45 - 295019.AK3.02 Comments:

Objective: 278LN001.12

Reference:

DOA 4700-01, LP DRE239LN001 K/A: 295019.AK3.02 3.5 / 3.4 K/A: Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Standby air compressor operation CFR: 41.5 Safety Function: 8 PRA: No Level: High Pedigree: New History:

Comments:

a) Incorrect - The FRVs lock up on a loss of air pressure.

b) Correct - A reactor scram and manually closing the MSIVs is required on a loss of instrument air due to the potential of the MSIVs drifting closed.

c) Incorrect - The target rock would be inop in ERV mode with a loss of Drywell Pneumatics.

d) Incorrect - The Unit 2 IA to Unit 2 Service Air cross-tie would already be open (85 psig).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 46 of 100 ILT 11-1 (NRC 2012-301) 46 Points: 1.00 Unit 2 and Unit 3 are operating at near rated power when Transmission Systems Operations (TSO) notifies the Control Room that the predicted post Unit trip with LOCA switchyard voltages are:

Unit 2: 325 KV Unit 3: 350 KV What are the required actions from the Operating team AND the reason for these actions?

A.

Adjust TR 32 Tap Changer; to restore system operability B.

Adjust TR 32 Tap Changer; to reduce circulating currents C.

Adjust TR 86 Tap Changer; to restore system operability D.

Adjust TR 86 Tap Changer; to reduce circulating currents Answer:

C Question 46 Details Topic:

46 - 700000.K3.02 Comments:

Objective: DRE262LN003.12

Reference:

DOA 6500-12 K/A: 700000.K3.02 3.6 / 3.9 K/A: Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

Actions contained in abnormal operating procedure for voltage and grid disturbances.

CFR: 41.10 Safety Function: 7 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Explanation: Given the voltages, only Unit 2 is below the procedural setpoint. TR-32 would be adjusted for Unit 3; TR-86 would be adjusted for Unit 2. The actions are to adjust the TR-86 Tap Changer (to position 31) to raise VOLTS (not VARs) to restore system operability.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 47 of 100 ILT 11-1 (NRC 2012-301) 47 Points: 1.00 DEOP 0400-03 STEAM COOLING attempts to prevent peak cladding temperature from exceeding ____(1)____ WITHOUT makeup. This prevents exceeding the threshold for

____(2)____.

A.

(1) 1500°F (2) fuel rod perforation B.

(1) 1500°F (2) significant metal-water reaction C.

(1) 1800°F (2) fuel rod perforation D.

(1) 1800°F (2) significant metal-water reaction Answer:

D Question 47 Details Topic:

47 - 295031.K3.04 Comments:

Objective: 29501LK012

Reference:

DEOP 0010-00 K/A: 295031.K3.04 4.0 / 4.3 K/A: Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL: Steam cooling CFR: 41.5 Safety Function: 2 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: DEOP 0010-00 states: Steam cooling without makeup capability is employed in DEOP 0400-03 during steam cooling. With no makeup to the RPV, adequate steam flow exists as long as RPV water level remains above the Minimum Zero-Injection RPV Water Level. When RPV water level drops below this elevation, emergency depressurization must be performed. The peak clad temperature is permitted to rise to 1800°F, the threshold for significant metal-water reaction, to maximize the heat transfer to steam and to delay the depressurization as long as possible. 1500°F, the threshold for fuel rod perforation, is the limit when injection is available.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 48 of 100 ILT 11-1 (NRC 2012-301) 48 Points: 1.00 Unit 2 is operating at full power.

A transient occurs resulting in the following conditions:

Reactor pressure spikes to 1260 psig.

Reactor water level rapidly drops to -40 inches.

Which of the following describes the expected response of the Recirc pumps?

A.

Recirc pumps trip immediately due to low reactor water level.

B.

Recirc pumps trip after 9 seconds due to low reactor water level.

C.

Recirc pumps trip immediately due to high reactor pressure.

D.

Recirc pumps trip after 9 seconds due to high reactor pressure.

Answer:

C Question 48 Details Topic:

48 - 295025.EA1.07 Comments:

Objective: 212LN002.06

Reference:

DAN 902(3)-5 A-8; 12E-6582F K/A: 295025.EA1.07 4.1 / 4.1 K/A: Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: ARI/RPT/ATWS: Plant-Specific CFR: 41.7 Safety Function: 3 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a & b) Incorrect, reactor water level has not yet reached the low level trip of -50.82" and only ARI initiates immediately on low reactor level.

c) Correct, reactor pressure >1211 psig will cause the recirc pumps to trip immediately.

d) Incorrect, Common misconception that the 9 second time delay is in effect with the low reactor water level trip of the recirc pumps.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 49 of 100 ILT 11-1 (NRC 2012-301) 49 Points: 1.00 An effluent offsite release is in progress.

Indications on the 923-7 panel have failed.

Besides the 923-7 panel, where in the Main Control Room can the effluent release rate be directly monitored?

A.

902-2 Panel B.

902-10 Panel C.

PPDS Display D.

SPDS Display Answer:

C Question 49 Details Topic:

49 - 295038.EA1.01 Comments:

Objective: 272LN004.10

Reference:

DAN 923-7 B-3 K/A: 295038.EA1.01 K/A: Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE : Stack-gas monitoring system: Plant-Specific CFR: 41.7 Safety Function: 9 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Incorrect - The 902-2 panel has the service water rad monitors b) Incorrect - The 902-10 panel has off-gas rad monitors c) Correct - The PPDS display shows total effluent release rate in the control room d) Incorrect - The SPDS display shows if there is a release in progress with a status box that changes color but not a value for rate.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 50 of 100 ILT 11-1 (NRC 2012-301) 50 Points: 1.00 Unit 3 was operating at rated power with the 2/3 EDG OOS.

A severe fire has occurred in the AEER concurrent with a loss of offsite power.

The order to evacuate the control room has been given.

As the Unit 3 NSO, in-plant actions you must perform include verifying...

A.

Bus 33-1 CUB 3, MAIN FEED FROM BUS 33, breaker OPEN.

B.

Bus 33-1 CUB 5: 3-7338, FEED TO 480V SWGR 38 (RB), breaker OPEN C.

Bus 38 CUB 4A: 3-7838-1/4, 480V REACTOR BUILDING MCC 38-1 & 38-4, breaker OPEN D.

Bus 33 CUB 10, 3-6733-3, BUS 33-1 FEED, breaker CLOSED.

Answer:

A Question 50 Details Topic:

50 - 295016.A1.04 (2)

Comments:

Objective: 29501LP083

Reference:

DSSP 0100-CR K/A: 295016.A1.04 3.1 / 3.2 K/A: Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT: A.C. electrical distribution CFR: 41.7 Safety Function: 7 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments:

The student must be able to determine (with an understanding of the Aux power system) the emergency bus line-up requirements given the power sources available in the stem.

a) Correct, DSSP 100-CR attachment B (U3 NSO ACTIONS) requires this breaker OPEN b) Incorrect, DSSP 100-CR attachment B (U3 NSO ACTIONS) requires this breaker CLOSED not OPEN.

c) Incorrect, DSSP 100-CR attachment B (U3 NSO ACTIONS) requires this breaker CLOSED not OPEN.

d) Incorrect, this is a DSSP 100-CR attachment E (U2 US OR DESIGNEE ACTIONS) requirement.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 51 of 100 ILT 11-1 (NRC 2012-301) 51 Points: 1.00 During refueling operations a mechanical failure in the mast results in an irradiated bundle dropping onto the fuel racks and being damaged.

Which one of the following will increase resultant refuel floor contamination levels?

A.

Fuel pool temperature being at the Tech Spec LOWER limit.

B.

Fuel pool temperature being at the Tech Spec UPPER limit.

C.

Fuel pool level being at the Tech Spec LOWER limit.

D.

Fuel Pool level being at the UPPER limit.

Answer:

B Question 51 Details Topic:

51 - 295023.AA2.03 Comments:

Objective: 233LN001.11

Reference:

DAN 902(3)-4 A-23, DRE LP233LN001 K/A: 295023.AA2.03 3.3 / 3.8 K/A: Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS : Airborne contamination levels CFR: 41.10 Safety Function: 8 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments:

a) With fuel pool temperature low in the band, the concern is dose rate from the spent fuel increasing.

b) Correct, the concern with high fuel pool temperature is iodine coming out of solution and increasing airborne contamination levels c) A low fuel pool level would result in elevated dose rates on the refuel floor.

d) High fuel pool level is a concern with the fuel pool overflowing into the vent ducts of the RB ventilation system.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 52 of 100 ILT 11-1 (NRC 2012-301) 52 Points: 1.00 Unit 2 was operating at near rated power when BOTH Recirc Pumps tripped.

Prior to the transient, ACTUAL reactor water level was ___(1)___ than INDICATED Fuel Zone reactor water level AND the difference will get ___(2)___ following the transient.

A.

(1) lower (2) smaller B.

(1) lower (2) larger C.

(1) higher (2) smaller D.

(1) higher (2) larger Answer:

A Question 52 Details Topic:

52 - 295001.AA2.06 Comments:

Objective: DRE216LN001.06

Reference:

LP DRE216LN001, DOP 1000-03 K/A: 295001.AA2.06 3.2 / 3.3 K/A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Nuclear boiler instrumentation CFR: 41.10 Safety Function: 1 & 4 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Fuel zone instruments range from -340 to +60 inches. During normal operation at power with reactor level +30 inches, the Fuel Zone instruments are pegged high. Fuel Zone indications are affected by anything that causes flow through the monitored jet pump. Therefore recirc pump, LPCI pump or Shutdown Cooling pump flow causes the Fuel Zone instruments to be inaccurate in the non-conservative direction (read higher than actual).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 53 of 100 ILT 11-1 (NRC 2012-301) 53 Points: 1.00 The 250VDC and 125VDC systems are aligned as follows:

The Unit 3 250VDC Charger is OOS for maintenance.

The 2/3 250VDC Charger is being supplied from Unit 3 and is on the Unit 3 250VDC system floating the battery.

The Unit 3 125 VDC Charger 3A is on the Unit 3 125VDC system floating the battery.

The Unit 3 EO reports a fire in MCC 39-2 and MCC 39-2 has been deenergized.

Which of the following describes the effect (if any) this will have on the Unit 3 250VDC and 125VDC batteries?

A.

The Unit 3 250VDC battery would be unaffected and the Unit 3 125VDC battery would be unaffected.

B.

The Unit 3 250VDC battery would be unaffected and the Unit 3 125VDC battery would be the only source of power to the Unit 3 125VDC system.

C.

The Unit 3 250VDC battery would be the only source of power to the Unit 3 250VDC system and the Unit 3 125VDC battery would be unaffected.

D.

The Unit 3 250VDC battery would be the only source of power to the Unit 3 250VDC system and the Unit 3 125VDC battery would be the only source of power to the Unit 3 125VDC system.

Answer:

C Question 53 Details Topic:

53 - 295003.AA2.03 Comments:

Objective: 263LN002.02, 263LN001.02

Reference:

DOA 6900-02, DOA 6900-04 K/A: 295003.AA2.03 3.2 / 3.5 K/A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER : Battery status:

Plant-Specific CFR: 41.10 Safety Function: 6 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: MCC 39-2 provides power to the 125 VDC battery charger #

3 and Unit 2/3 250 VDC battery charger. MCC 38-2 provides power to the 125 VDC battery charger # 3A. MCC 38-3 provides power to the Unit 3 250 VDC battery charger (which is OOS in the stem of the question).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 54 of 100 ILT 11-1 (NRC 2012-301) 54 Points: 1.00 Unit 2 is operating at near rated power, with the following annunciators illuminated:

902-4 B-18 DIV 1 TORUS WTR BULK TEMP HI 902-4 C-18 DIV 2 TORUS WTR BULK TEMP HI 902-3 C-13 2B ELECTROMATIC RELIEF VLV OPEN Torus Bulk temperature is 95oF and rising.

Per DOA 0250-01, RELIEF VALVE FAILURE, what IMMEDIATE actions must be performed to mitigate Torus heatup?

A.

Initiate Torus cooling and verify Torus temperature stabilizes.

B.

Cycle ADS INHIBIT switch from NORMAL to INHIBIT to NORMAL and verify reactor pressure stable.

C.

Place 2B relief valve control switch in OFF and verify feedwater control system has stabilized reactor level.

D.

Cycle 2B relief valve control switch to MANUAL and then to OFF and verify turbine control system maintains reactor pressure stable.

Answer:

C Question 54 Details Topic:

54 - 295026.G.2.4.49 Comments:

Objective: DRE223LN001.08

Reference:

DOA 250-1, DANs 902-4 B-18 and C-18, DAN 902-3 C-13 K/A: 295026.G.2.4.49 4.6 / 4.4 K/A: Suppression Pool High Water Temperature: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

CFR: 41.10 Safety Function: 5 PRA: Yes Level: Memory Pedigree: Bank History: 2012 NRC, 2001 NRC Comments: With Torus temp rising (95°F DEOP entry) and indication of 2B ERV open, entry into DOA 0250-1 and DEOP 0200-1 is required.

Immediate actions taken without reference are to place relief valve control switch in OFF and verify feedwater control system has stabilized reactor level.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 55 of 100 ILT 11-1 (NRC 2012-301) 55 Points: 1.00 Unit 3 was operating at near rated power when a spurious scram occurred.

The Shift Manager has determined that a hot fast restart is to be performed.

Which of the following equipment condition(s) needs to be met to allow the Reactor Mode Switch to be placed in the REFUEL position to perform Reactor Mode Switch Interlock testing?

A.

Moderator temperature < 350oF.

B.

All Control rods are fully inserted.

C.

SRM Rod Block Functional Test complete.

D.

Refuel Position One Rod Out Interlock met.

Answer:

B Question 55 Details Topic:

55 - 295006.G.2.1.23 Comments:

Objective: 20102LK004

Reference:

DGP 2-3; DOS 0700-01; DOS 0700-05; Tech Spec 3.10.1 K/A: 295006.G.2.1.23 4.3 / 4.4 K/A: SCRAM: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

CFR: 41.10 Safety Function: 1 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC, 2008 NRC Comments: Per T.S. 3.10.1, with all control rods fully inserted and no core alterations in progress, the Reactor Mode Switch may be placed in any position for testing purposes without exiting Mode 3.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 56 of 100 ILT 11-1 (NRC 2012-301) 56 Points: 1.00 Unit 2 is in a refueling outage.

The reactor vessel head bolts are tensioned.

Reactor water level is just below the flange.

Reactor vessel water temperature is 170°F and steady.

2A and 2B Recirculation pumps are secured.

2A and 2B SDC systems are in service cooling the reactor vessel.

2C SDC pump is Out of Service (O.O.S.) for maintenance.

A fault causes 2B SDC pump to trip on over current.

Which of the following describe the MINIMUM required Tech Spec actions to be taken within one hour?

A.

Monitor reactor coolant temperature and pressure.

B.

Verify reactor coolant circulating by an alternate method.

C.

Restore required SDC subsystem(s) to OPERABLE status.

D.

Verify an alternate method of decay heat removal is available for each inoperable required SDC subsystem.

Answer:

D Question 56 Details Topic:

56 - 295021.G.2.2.39 Comments:

Objective: 205LN001.11

Reference:

T.S. 3.4.8 K/A: 295021.G.2.2.39 3.9 / 4.5 K/A: Loss of shutdown cooling: Knowledge of less than or equal to one hour Technical Specification action statements for systems.

CFR: 41.10 Safety Function: 4 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: Examinee must first determine from the initial conditions that T.S.3.4.8 SDC - Cold Shutdown applies, then apply the LCO requirements for the minimum required SDC subsystems.

a) Would apply if NO SDC subsystems were in operation.

b) Would apply if NO SDC subsystems were in operation.

c) Would apply if the Unit was in Hot Shutdown.

d) Correct, per T.S. 3.4.8, verify an alternate method of decay heat removal is available for each inoperable required SDC subsystem.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 57 of 100 ILT 11-1 (NRC 2012-301) 57 Points: 1.00 Given the following plant conditions on Unit 2:

Drywell pressure is 2.5 psig and rising slowly Reactor water level is 90 inches and dropping slowly HPCI system automatically initiated and is running at 3,800 rpm, injecting 5,600 gpm into the vessel CST level has decreased, but the automatic suction transfer for HPCI has failed Annunciator 902-3 A-11, 'HPCI Booster Pump Suction Pressure Low' has just alarmed Which of the following describes the response of the HPCI system under these conditions?

A.

HPCI turbine will trip on low pump suction pressure B.

HPCI turbine speed will rise and then HPCI pump flow will eventually drop to zero C.

HPCI turbine speed will remain constant and HPCI pump flow will eventually drop to zero D.

HPCI turbine speed will decrease until the MGU reaches the low speed stop Answer:

B Question 57 Details Topic:

57 - 295030.EA1.05 Comments:

Objective: DRE206LN001.08

Reference:

LP DRE206LN001 K/A: 295030.EA1.05 3.5 / 3.5 K/A: Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCI CFR: 41.7 Safety Function: 5 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Comments: The Booster Pump low suction pressure trip is bypassed with an initiation signal present. The turbine speed control system CANNOT respond to turbine speed increases because it is positioned by a FLOW controller. As flow falls off, the MGU would go to the high speed stop but with no water to pump, HPCI will continue to operate (with the MGU at the HSS - NOT 4000 rpm but full steam power) until the turbine overspeeds.

Turbine speed would remain constant if HPCI had been started manually.

Decreasing turbine speed until the MGU reaches the low speed stop is a condition HPCI is placed in during manual startup and shutdown.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 58 of 100 ILT 11-1 (NRC 2012-301) 58 Points: 1.00 With both units operating at full power, a service water pump trip results in the following conditions:

Service water header pressure 78 psig dropping slowly.

X-area temperature 147°F and rising slowly.

Alternator air temperature 52°C and steady.

Stator cooling water outlet temperature 75°C and rising slowly.

HPCI area temperature 140°F and rising slowly.

Which one of the following actions is appropriate for these conditions?

A.

Immediately reduce load to 25% and trip the main turbine.

B.

Enter DEOP 300-1 and take action to mitigate the high area temperature conditions.

C.

Reduce load per DGP 3-1 only enough to compensate for changes in system temperatures.

D.

Verify the 2/3 emergency fire pump starts and then cross-connect the fire water and service water systems.

Answer:

A Question 58 Details Topic:

58 - 295018.AA2.01 Comments:

Objective: 276LN001.08

Reference:

DOA 3900-01 K/A: 295018.AA2.01 3.3 / 3.4 K/A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Component temperatures CFR: 41.10 Safety Function: 8 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Actions per DOA 3900-01. If alternator air temperature exceeds 48°C, then the DOA directs load reduced and the main generator tripped. HPCI area temperature is below the DEOP 300-1 entry condition. Cross-connecting fire water and service water would only be done to supplement the fire protection system. Reducing load only enough to compensate for high system temperatures is NOT sufficient action based on the conditions given.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 59 of 100 ILT 11-1 (NRC 2012-301) 59 Points: 1.00 Unit 2 was operating at full power when a transient occurred resulting in the following conditions:

Loss of Offsite Power (LOOP)

Reactor scram Reactor pressure is 50 psig All scram valves are open Rod H-8 indicates position 46, all other rods indicate full in Which of the following actions (if any) would be successful in inserting H-8?

A.

Vent the scram air header.

B.

Vent the H-8 CRD overpiston area.

C.

Scram H-8 using its individual scram toggle switch.

D.

H-8 can not be inserted until power has been restored to a CRD pump.

Answer:

D Question 59 Details Topic:

59 - 295015.AK1.04 Comments:

Objective: 29501LK031

Reference:

DOA 0300-01 K/A: 295015.AK1.04 3.8 / 3.8 K/A: Knowledge of the operational implications of the following concepts as they apply to INCOMPLETE SCRAM : Reactor pressure: Plant-Specific CFR: 41.8 Safety Function: 1 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

Examinee is required to know that a a LOOP would result in a loss of all CRD pumps. With no CRD pumps running, the charging water header would depressurize. If the charging water header pressure is lost, the CRD accumulator will depressurize as a result of leakage through the 0305-115 accumulator check valve/0305-126 scram inlet valves (which are open) resulting in the potential inability to scram the drive at low reactor pressure.

a, b & c) Incorrect - all of these are alternate means of rod insertion but would be ineffective because no motive force exists to insert the rod under these conditions.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 60 of 100 ILT 11-1 (NRC 2012-301) 60 Points: 1.00 Unit 3 was operating at near rated power, when the following occurred:

A leaking valve has caused water to be sprayed on 'A' SBGT Charcoal Bed.

A fuel handling accident occurred, causing radiation levels at the site boundary to increase.

An NSO isolated Reactor Building ventilation and started the 'A' SBGT train.

The release rates at the site boundary will be higher than anticipated for which of the following?

A.

Iodine B.

Particulates C.

Transuranics D.

Noble Gases Answer:

A Question 60 Details Topic:

60 - 295017.AK2.12 Comments:

Objective: 27800LK027

Reference:

UFSAR Sec.6.5 K/A 295017.AK2.12 3.4 / 3.7 K/A: Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Standby gas treatment/FRVS CFR: 41.7 Safety Function: 9 PRA: No Level: High Pedigree: Bank History: 2007 NRC; 2012 NRC Explanation: The charcoal bed normally removes the Iodine, but due to being wet the efficiency of removal goes down. Transuranics are released with fuel failure but do not go airborne. Particulates are removed by the HEPA filters. Noble Gases have little hold up and would not be significantly changed.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 61 of 100 ILT 11-1 (NRC 2012-301) 61 Points: 1.00 Unit 3 is operating at rated conditions.

A loss of both CRD pumps has occurred.

This condition will result in ________.

A.

degradation of the CRDM seals due to loss of cooling flow B.

loss of scram capability due to loss of charging water flow C.

loss of Unit 3 reactor water level indication due to loss of RVWLIS D.

failure of Reactor Recirc pump seals due to loss of seal cooling flow Answer:

A Question 61 Details Topic:

61 - 295022.AK3.02 Comments:

Objective: 201LN003.12

Reference:

DRE LP201LN003 K/A: 295022.AK3.02 2.9 / 3.1 K/A: Knowledge of the reasons for the following responses as they apply to LOSS OF CRD PUMPS: CRDM high temperature CFR: 41.5 Safety Function: 1 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments:

a) At elevated temperatures the CRDM seals become brittle and can result in breakdown and increased scram times.

b) Incorrect - CRD accumulators maintain scram capability upon a loss of charging water flow c) Incorrect - Loss of RVWLIS will not effect reactor water level indication during normal operations d) Incorrect - seal cooling to the Recirc pumps is provided by RBCCW.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 62 of 100 ILT 11-1 (NRC 2012-301) 62 Points: 1.00 Unit 2 was operating at near rated power, when a transient occurred, and the following conditions exist:

Drywell pressure is 3.0 psig and steady.

Reactor pressure is 950 psig and steady.

RWCU blowdown rate is 200 gpm and steady.

RWCU recirculation and blowdown modes are being used to control RPV pressure.

RPV water level is 12 inches and lowering at rate of 1 inch/minute with all available systems operating.

Which of the following action would mitigate the conditions listed above?

A.

Secure blowdown AND recirculation modes.

B.

Secure blowdown mode AND maintain recirculation mode.

C.

Maintain existing blowdown AND recirculation modes.

D.

Maintain existing blowdown mode AND secure recirculation mode.

Answer:

B Question 62 Details Topic:

62 - 295009.AA1.04 Comments:

Objective: 204LN001.08

Reference:

DEOP 0100-00 K/A: 295009.AA1.04 K/A: Ability to operate and/or monitor the following as they apply to LOW REACTOR WATER LEVEL : Reactor water cleanup CFR: 41.7 Safety Function: 2 PRA: No Level: High Pedigree: Bank History: 2012 Comments: Securing blowdown will decrease the amount of water leaving the reactor, which mitigates the lowering RPV water level. DEOP 100 allows installing jumpers to maintain RWCU Recirc mode for pressure control.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 63 of 100 ILT 11-1 (NRC 2012-301) 63 Points: 1.00 Unit 2 was operating at full power when a spurious Group I (one) occurred.

10 seconds later the following indications exist:

2-0203-3B and 2-0203-3C ERVs indicate open on the 902-3 and 902-21 panel, all other ERVs indicate closed.

The Isolation Condenser is in a normal standby lineup.

Which of the following would indicate the HIGHEST reactor pressure reached during this event based on the above indications.

A.

1065 psig B.

1095 psig C.

1115 psig D.

1135 psig Answer:

C Question 63 Details Topic:

63 - 295020.AA2.04 Comments:

Objective: 239LN001.06

Reference:

DRE LP239LN001 K/A: 295020.AA2.04 3.9 / 3.9 K/A: Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION : Reactor pressure CFR: 41.10 Safety Function: 5 & 7 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments:

a) Incorrect, Only a reactor scram would have occurred at 1060 psig b) Incorrect, Isolation Condenser would auto initiate if greater than 1070 for more than 17 seconds c) Correct, 1110 psig is the setpoint for the low set ERVs (B and C ERVs) d) Incorrect, 1133 psig is the setpoint for the high set ERVs (D and E ERVs and the Target Rock)

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 64 of 100 ILT 11-1 (NRC 2012-301) 64 Points: 1.00 Unit 2 was operating at full power when the following events occurred:

The D Main steam line breaks in the steam tunnel.

Loss of the Instrument Bus due to a short circuit.

An NSO reports the following valves open:

RX RETURN VLV, 2-1201-7 2B FW REG ISOL, 2-3206B MSIV, 2-0203-1A Which of the following describes the status of the PCIS status box in the Safety Parameter Display System (SPDS)?

A.

GREEN B.

YELLOW C.

RED D.

CYAN Answer:

C Question 64 Details Topic:

64 - 295032.G.2.1.19 Comments: Objective: 283LN002.06

Reference:

DOP 9950-17 K/A: 295032.2.1.19 K/A: High Secondary Containment Area Temperature: Ability to use plant computers to evaluate system or component status.

CFR: 41.10 Safety Function: 5 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) The PCIS status box is GREEN when a PCIS isolation signal is present and all required PCIS isolation valves are shut.

b) Yellow is a pre-alarm indication on PCIS used by the parameters with analog values on SPDS.

c) Correct, A steam line break would cause a Group 1 PCIS isolation due to high secondary containment temperature in the steam tunnel. The PCIS status box is RED when a PCIS isolation signal is present and all required PCIS isolation valves are not shut. All MSIVs are required to be shut when a Group 1 signal is received.

d) Would be correct if the status of the PCIS valves were undeterminable.

A loss of the Instrument Bus would not affect the ability of the SPDS to determine valve position even though some of the PCIS valves would lose remote indication.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 65 of 100 ILT 11-1 (NRC 2012-301) 65 Points: 1.00 Unit 3 was at rated conditions when a transient occurred.

An Isolation Condenser steam leak occurred and was isolated.

Isolation Condenser area temperature is 170ºF and is too high for personnel access.

Valid Reactor Building Ventilation isolations are present on each of the following parameters:

Drywell Pressure Reactor Building Exhaust Radiation Reactor Water Level Restarting the Reactor Building Ventilation would allow safer access to the Isolation Condenser area ________.

A.

but is NOT allowed due to the Drywell Pressure isolation B.

but is NOT allowed due to the Reactor Water Level isolation C.

and may be performed after bypassing the isolation signals D.

but is NOT allowed due to the Reactor Building Exhaust Radiation isolation Answer:

D Question 65 Details Topic:

65 - 295034.K1.01 Comments:

Objective: 29502LK050

Reference:

LP 295L-S3 and DEOP 300-1 K/A: 295034.K1.01 3.8 / 4.1 K/A: Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION : Personnel protection CFR: 41.9 Safety Function: 9 PRA: Yes Pedigree: Bank Level: Memory History: 2012 NRC Explanation: Only the drywell and RPV water level isolations are allowed to be bypassed since they do not indicate a release hazard. Reactor building exhaust radiation above the isolation setpoint would be indicative of a potential radioactive release problem and would not be allowed to be bypassed.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 66 of 100 ILT 11-1 (NRC 2012-301) 66 Points: 1.00 Unit 3 was shutdown 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> ago.

Reactor vessel water temperature is 100°F.

Core offload is in progress.

Using the OP-DR-104-1001 curves provided, determine the time to boil with no decay heat removal.

A.

3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> B.

7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> C.

11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> D.

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Answer:

D Question 66 Details Topic:

66 - Generic.2.1.25 Comments:

Objective: 29501LP005

Reference:

OP-DR-104-1001 K/A: Generic.2.1.25 3.9 / 4.2 K/A: Ability to interpret reference materials, such as graphs, curves, tables, etc.

CFR: 41.10 Safety Function: 5 PRA: No Level: High Pedigree: New History: 2012 NRC Comments: The examinee must know that with a core offload in progress, the reactor vessel is flooded up, the dryer/separator and fuel pool gates are removed making attachment K the correct graph to use.

a) Incorrect - 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> is the time to boil while flooded to flange and dryer/separator removed.

b) Incorrect - 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> is the time to boil while flooded up, gates in and dryer/separator in.

c) Incorrect - 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> is the time to boil while flooded up, gates in and dryer/separator removed d) Correct - 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> is the time to boil for conditions stated in the question.

REQUIRED

REFERENCES:

OP-DR-104-1001 Attachments A, C, E, G, I and K.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 67 of 100 ILT 11-1 (NRC 2012-301) 67 Points: 1.00 Given the following conditions:

There is a fire in the Aux Electric Equipment Room and Control Room evacuation is required The Unit 2 NSO has been directed to SCRAM the plant All procedural means to SCRAM the plant have failed Action is immediately needed to protect Public Health and Safety What is the lowest level of authority needed to approve action to SCRAM the plant that is NOT covered by a procedure or technical specification?

A.

Any licensed Reactor Operator (RO)

B.

Any licensed Senior Reactor Operator (SRO)

C.

The Shift Manager (SM)

D.

The Shift Operations Superintendent (SOS)

Answer:

B Question 67 Details Topic:

67 - Generic.2.1.20 Comments:

Objective: 29501LE016

Reference:

DRE299L-S4 and DSSP 100-CR K/A: Generic.2.1.20 4.6 / 4.6 K/A: Ability to interpret and execute procedure steps.

CFR: 41.10 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Explanation: As stated in reference materials - An SRO must approve action that is not covered by Tech Specs/Procedure. Certain steps or conditions within this procedure may contradict normal license conditions or Technical Specification requirements, as stated in 10 CFR 50.54x: "A licensee may take reasonable action that departs from a license condition or Technical Specification requirements in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with the license conditions and Technical Specifications that can provide adequate or equivalent protection is immediately apparent." This action "shall be approved, as a minimum, by a licensed Senior Reactor Operator (SRO) prior to taking the action."

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 68 of 100 ILT 11-1 (NRC 2012-301) 68 Points: 1.00 The Unit 2 ____(1)____ battery contains more cells than its respective Unit 3 battery due to

____(2)____.

A.

(1) alternate 125 VDC (2) additional loads B.

(1) alternate 125 VDC (2) increased line loss C.

(1) BOP 250 VDC (2) additional loads D.

(1) BOP 250 VDC (2) increased line loss Answer:

B Question 68 Details Topic:

68 - Generic.2.2.03 (5)

Comments:

Objective: DRE263LN002.03

Reference:

LP DRE263LN002 K/A: Generic.2.2.03 3.8 / 3.9 K/A: Knowledge of the design, procedural, and operational differences between units.

Level: Memory Pedigree: New History: 2012 NRC Comments: Unit 2 alternate 125 VDC battery has five extra cells to make up for line loss due to the long cable run from battery to the 125VDC bus.

The Unit 2 alternate 125 VDC battery is located in the SBO building. The Unit 3 alternate 125 VDC battery is located in the Unit 3 Turbine Building.

The BOP 250 VDC batteries on both Units have the same number of cells and only power the non-safety related turbine building loads (1 motor).

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 69 of 100 ILT 11-1 (NRC 2012-301) 69 Points: 1.00 Unit 2 is operating at 50% power and expected to pick up load early next shift.

'D' Lift pump is being prepared to start after maintenance on the motor has been completed.

The NSO in the control room attempts to start the 'D' Lift pump.

The EO at the Lift Station reports no observable change in status locally.

Based on these events, in regards to the 'D' Lift pump, the operating crew should _________.

A.

leave the pump as is to support troubleshooting activities B.

attempt to restart the pump ONE more time ONLY C.

attempt to restart the pump a MAXIMUM of TWO more times D.

place the control switch in PULL TO LOCK at the local control panel in the Lift Station Answer:

A Question 69 Details Topic:

69 - Generic.2.2.20 Comments:

Objective: 299L029

Reference:

DOP 6500-19 K/A: Generic.2.2.20 2.6 / 3.8 K/A: Knowledge of the process for managing troubleshooting activities.

CFR: 41.10 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Correct - Per DOP 6500-19, if the equipment malfunction is not readily apparent and plant conditions permit, then leave conditions as is to aid in troubleshooting.

b) Incorrect - One more restart is a restriction for Hot Restarts c) Incorrect - Two more restarts, is a restriction for cold starts.

d) Incorrect - PTL is the action to take if the cause of the breaker trip is known.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 70 of 100 ILT 11-1 (NRC 2012-301) 70 Points: 1.00 Chemistry has reported that high coolant activity exists on Unit 2 and a fuel element failure is suspected.

If site assembly is required, which of the following actions is required per DGA 16 to prevent excessive personnel exposure?

A.

Isolating HPCI steam lines B.

Isolating Recirc sample lines C.

Isolating the Isolation Condenser D.

Isolating HPCI steam drains to the condenser Answer:

D Question 70 Details Topic:

70 - Generic.2.3.14 Comments:

Reference:

ILTS026, DGA 16 K/A: Generic.2.3.14 3.4 / 3.8 K/A: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

CFR: 41.12 Safety Function: 9 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Comments: Per DGA 16 Caution. Assembly area inside the RPA is near the feed pumps, which is against the condenser shield wall. Any flow of radioactive water to the condenser would increase dose rates in this area, so D is correct. IC drains do not go to the condenser, so C is incorrect.

Isolating HPCI steam flow would block the leakage, but would render HPCI unavailable, so this action would not be appropriate. Recirc sample drains also do not go to the condenser.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 71 of 100 ILT 11-1 (NRC 2012-301) 71 Points: 1.00 An accident has occurred at the station and you have volunteered to perform an evolution to protect valuable property. The dose rate in the area you will be entering is 30 Rem/hr.

What is the MAXIMUM time you can spend in the area performing your task without violating TEDE Radiation Exposure Limits per RP-AA-203 EXPOSURE CONTROL AND AUTHORIZATION?

A.

4 minutes B.

10 minutes C.

20 minutes D.

30 minutes Answer:

C Question 71 Details Topic:

71 - Generic.2.3.4 (2)

Comments:

Objective: 29900LK208

Reference:

RP-AA-203, EP-AA-113 K/A: Generic.2.3.4 3.2 / 3.7 K/A: Knowledge of radiation exposure limits under normal or emergency conditions.

Level: High Pedigree: Bank History: 2012 NRC Explanation:

a) Incorrect - Would exceed 2REM which is the Local TEDE limit.

b) Incorrect - Would exceed 5REM which is the Federal TEDE limit.

c) Correct - The exposure limit for protecting valuable property is 10REM TEDE. Based on 30 REM/Hr in the area, the stay time would be 20 minutes.

d) Incorrect - Would exceed 15REM which is the local TODE limit.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 72 of 100 ILT 11-1 (NRC 2012-301) 72 Points: 1.00 Unit 3 was operating at near rated power when ALL annunciators are lost EXCEPT the annunciators labeled ANNUN DC PWR FAILURE" are illuminated on the following panels:

903-3 903-4 903-5 903-6 903-7 903-8 903-55 What are the immediate actions the Operating team is required to perform?

A.

Lock out the Recirc MG set scoop tubes ONLY.

B.

Insert a manual scram, and trip the Turbine ONLY.

C.

Stabilize and control RPV water level in normal band.

D.

Insert a manual scram, trip the Turbine AND lock out the Recirc MG set scoop tubes.

Answer:

C Question 72 Details Topic:

72 - Generic.2.4.32 Comments:

Objective: 26302LK001

Reference:

DOA 6900-03 K/A: Generic.2.4.32 3.6 / 4.0 K/A: Knowledge of operator response to loss of all annunciators.

Level: Memory Pedigree: Bank History: 2012 NRC, 2010 NRC Explanation: The conditions in the stem indicate a complete loss of Div 1 125VDC. The immediate action in the above DOA is to stabilize and control RPV water level in the normal control band. The distractors depict actions that would be taken if other sections of the DC system were lost.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 73 of 100 ILT 11-1 (NRC 2012-301) 73 Points: 1.00 Unit 2 was operating at near rated power when a steam line rupture in the HPCI room occurred.

A scram was attempted. An ATWS occurred, with the following conditions:

15 control rods are at position 24 or GREATER.

All scram valves are open.

CleanUp Pump Area temperature is 175oF.

HPCI Room temperature is 235oF.

HPCI Cubicle Rad Levels are 3750 mrem.

East CRD Module Area rad levels are 2500 mrem.

West CRD Module Area rad levels are 2800 mrem.

The Unit Supervisor has directed insertion of control rods using alternate methods.

Which action is the PREFERRED method for inserting the control rods that are NOT at position 00?

A.

Vent the control rods' overpiston volumes.

B.

Bypass the RWM AND drive the control rods MANUALLY.

C.

Close the scram air header supply valve AND vent the scram air header.

D.

Close the scram discharge vents and drains AND pull power supply fuses for the scram solenoids.

Answer:

B Question 73 Details Topic:

73 - Generic.2.4.23 Comments:

Objective: DRE215LN002.06

Reference:

DEOP 0300-01, DEOP 0500-05 K/A: Generic.2.4.23 3.4 / 4.4 K/A: Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

CFR: 41.10 PRA: No Level: High Pedigree: New History: 2012 NRC Explanation: Control rods should be manually inserted per DEOP 0500-

05. Pulling the power supply fuses for the scram solenoids would not be viable, because the scram valves are already opened. Venting the scram air header or the individual overpiston areas are not viable, because of the high dose (greater than Max Safe) in the areas of the CRD module areas.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 74 of 100 ILT 11-1 (NRC 2012-301) 74 Points: 1.00 Units 2 and 3 are operating at full power.

When the following events occur:

Security reports an aircraft has struck the reactor building.

DOA 0010-18, ESCALATED SECURITY EVENT / HOSTILE FORCE INTRUSION, has been entered.

Which of the following is NOT a required IMMEDIATE action per DOA 0010-18?

A.

Start the 2/3 Diesel Fire Pump B.

Maximize CRD flow on BOTH Units C.

Isolate RWCU systems on BOTH Units D.

Verify MO 2(3)-1301-1 AND MO 2(3)-1301-4 on BOTH Units are closed Answer:

A Question 74 Details Topic:

74 - Generic.2.4.28 Comments:

Objective: 295L010

Reference:

DOA 0010-18 K/A: Generic 2.4.28 K/A: Knowledge of procedures relating to a security event (non-safeguards information).

CFR: 41.10 Safety Function: 5 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Correct, Start the 2/3 Diesel Fire Pump is an immediate action of DOA 0010-10, FIRE/EXPLOSION not DOA 10-18 b, c & d) Incorrect, verify MO 2(3)-1301-1 AND MO 2(3)-1301-4 ON BOTH Units are closed, isolate BOTH Unit RWCU systems and maximize CRD flow are immediate actions of DOA 10-18 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 75 of 100 ILT 11-1 (NRC 2012-301) 75 Points: 1.00 A hydraulic ATWS has occurred on Unit 2.

The Unit Supervisor has directed you to perform repeated scram resets per DEOP 0500-05.

Which of the following describes the MINIMUM electrical safety precautions required to perform this task?

All metal removed and...

A.

safety glasses B.

safety glasses, long sleeve cotton shirt and rubber gloves C.

safety glasses, electrical safety coat and rubber gloves D.

safety glasses, electrical safety coat, face shield and rubber gloves Answer:

B Question 75 Details Topic:

75 - Generic.2.1.26 Comments:

Objective: 295L105

Reference:

SA-AA-129 K/A: Generic 2.1.26 3.4 / 3.6 K/A: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

CFR: 41.10 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: SA-AA-129 states that for the voltage in the area being worked in for DEOP 500-5 the minimum PPE is all metal removed, long sleeve cotton shirt, safety glasses and rubber gloves REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 76 of 100 ILT 11-1 (NRC 2012-301) 76 Points: 1.00 Unit 2 was operating at near rated power when a transient occurred, resulting in the following sequence of events:

Annunciator 902-7 G-3, TURB BYP VLV OPEN, alarmed.

An NSO reported RPV water level is 22 inches and trending down.

An NSO reported that Main Steam line pressure is 800 psig.

An automatic scram occurred, with all rods reported as fully inserted.

An NSO reported RPV water level is 10 inches and trending down.

The Reactor scrammed on ___(1)___ and the Unit Supervisor is required to direct ___(2)___.

A.

(1) RPV water level; (2) entering DGP 2-3, REACTOR SCRAM B.

(1) 2 out of 3 DEHC Pressure Controller Processors failed low; (2) verifying Control valves throttle closed, per DOA 5650-03, TURBINE CONTROL VALVE OR BYPASS VALVE FAILED OPEN C.

(1) Main Steam Isolation Valves closure; (2) opening the MSIVs, per DOA 5650-03, TURBINE CONTROL VALVE OR BYPASS VALVE FAILED OPEN D.

(1) Main Steam Isolation Valves closure; (2) entering DGP 2-3, REACTOR SCRAM Answer:

D Question 76 Details Topic:

76 - 295006.A2.06 Comments:

Objective: DRE245LN001.06

Reference:

DOA 5650-03 K/A: 295006.A2.06 3.5 / 3.8 K/A: Ability to determine and/or interpret the following as the apply to SCRAM: cause of reactor SCRAM.

Safety Function: 1 PRA: No Level: High Pedigree: Bank History: 2009 NRC; 2012 NRC Explanation: When MSL pressure dropped below 827 psig and with the mode switch is in RUN, a scram will occur due to 9.5% Main Steam Isolation Valves closure. The required action is to verifying Control valves throttle closed, NOT open the MSIVs.

The scram setpoint for RPV water level is 6.02 inches.

2 out of 3 DEHC Pressure Controller Processors failing low will cause Bypass valves to close NOT open.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 77 of 100 ILT 11-1 (NRC 2012-301) 77 Points: 1.00 Unit 3 is operating at near rated power with ONLY the 3A RBCCW pump and 3A RBCCW Heat Exchanger in operation, when the following occurs:

The TUBE SIDE flow through 3A RBCCW Heat Exchanger is reduced to 10%, due to tube fouling.

If this condition goes un-corrected, RBCCW heat exchanger outlet temperatures will ___(1)___

AND the SRO will be required to direct actions per ___(2)___.

A.

(1) DECREASE, (2) DOP 3900-001, SERVICE WATER OPERATION.

B.

(1) DECREASE; (2) DOP 1200-03, RWCU SYSTEM OPERATION WITH THE REACTOR AT PRESSURE.

C.

(1) INCREASE, (2) DOP 3900-001, SERVICE WATER OPERATION.

D.

(1) INCREASE; (2) DOP 1200-03, RWCU SYSTEM OPERATION WITH THE REACTOR AT PRESSURE.

Answer:

D Question 77 Details Topic:

77 - 295018.AA2.02 Comments: Objective: DRE204LN001.06

Reference:

DOA 3900-01, DOP 1200-03, P&ID M-22 K/A: 295018.AA2.02 3.1 / 3.2 K/A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER :

Cooling water temperature Level: High Pedigree: Bank History: 2012 NRC Explanation: With Service Water (tube side) being reduced through the RBCCW heat exchanger, RBCCW temperatures will increase. If a misconception that RBCCW is on the tube side is assumed, then RBCCW heat exchanger outlet temperatures would decrease (same cooling, less RBCCW flow). With RBCCW temperature increasing, this would cause the RWCU non-regen heat exchanger temperatures to rise, which will cause a system isolation. The action on a system isolation is to dial the pressure controller to zero. Since Service flow through the heat exchanger is being reduced, the student may incorrectly choose the Service Water procedure which does not provide guidance for the condition indicated in the question.

SRO Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 78 of 100 ILT 11-1 (NRC 2012-301) 78 Points: 1.00 Unit 2 is at full power.

Drywell temperature is 148°F and rising.

What is the Tech Spec limit for Drywell Air Temperature AND its Bases?

A.

150°F; to ensure LOCA peak Drywell temperature DOES NOT exceed 281°F.

B.

150°F; to ensure RPV water level instruments DO NOT become unreliable due to boiling in the instrument runs in the Drywell.

C.

160°F; to ensure LOCA peak Drywell temperature DOES NOT exceed 281°F.

D.

160°F; to ensure RPV water level instruments DO NOT become unreliable due to boiling in the instrument runs in the Drywell.

Answer:

A Question 78 Details Topic:

78 - 295028.EA2.01 Comments:

Objective: 21000LK004

Reference:

Tech Spec 3.6.1.5 and the Bases K/A: 295028.EA2.01 4.0 / 4.1 K/A: Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE :Drywell temperature Safety Function: 5 PRA: No Level: Memory Pedigree: Bank History: 2012 NRC Explanation: Per the Bases the temperature limit is 150°F to ensure that the safety analysis remains valid by maintaining the expected initial conditions and ensure that the peak LOCA drywell temp does not exceed the maximum allowable temp of 281°F.

SRO Criteria: 2 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 79 of 100 ILT 11-1 (NRC 2012-301) 79 Points: 1.00 Per the Tech Spec bases, what is the MINIMUM number of operable Main Turbine Bypass Valves (BPVs) required to prevent exceeding the MCPR safety limit (WITHOUT thermal limit penalties applied) during a Load Reject scram?

A.

6 B.

7 C.

8 D.

9 Answer:

D Question 79 Details Topic:

79 - 295005.G.2.2.25 (2)

Comments: Objective: 241LN001.07

Reference:

T.S. 3.7.7 K/A: 295005.G.2.2.25 3.2 / 4.2 K/A: Main Turbine Generator Trip: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Safety Function: 3 PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments: As stated in Tech Spec 3.7.7 and associated Bases, and Tech Spec 2.0 (Safety Limit), and the COLR, Whenever the Unit is at, or above 25% power, all BPVs must be operable to prevent exceeding the MCPR Safety Limit (with no thermal limit penalties applied) during a turbine trip, load reject or feedwater controller failure.

SRO per Criteria: 2 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 80 of 100 ILT 11-1 (NRC 2012-301) 80 Points: 1.00 Unit 2 was operating at full power when a transient occurred resulting in an ATWS.

SBLC failed to inject and alternate SBLC injection is in progress.

Given the following sequence of events, what is the EARLIEST time that operators in the field can secure injecting SBLC?

12:00 - All APRM DOWNSCALE lights are lit.

12:05 - All IRMs are below range 7 12:10 - All control rods are at position 04, 02 or 00.

12:15 - All control rods are fully inserted.

A.

12:00 B.

12:05 C.

12:10 D.

12:15 Answer:

C Question 80 Details Topic:

80 - 295037.G.2.4.35 Comments: Objective: 29502LP022

Reference:

DEOP 0500-01 K/A: 295037.G.2.4.35 K/A: SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown: Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

Safety Function: 1 PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) Incorrect - All APRM downscales are a factor in the second override of the level leg of DEOP 0400-05 b) Incorrect - IRMs below range 7 is a factor in the first override of the power leg of DEOP 0400-05 c) Correct - Per DEOP 0500-01 limitations and actions state: IF while executing this procedure any of the following conditions are met, THEN terminate Alternate Standby Liquid Control Injection. All control rods are inserted to OR beyond position 04.

d) Incorrect - Examinee must know that all rods do not need to be inserted per DEOP 0500-01 to secure from alternate SBLC SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 81 of 100 ILT 11-1 (NRC 2012-301) 81 Points: 1.00 Unit 3 was operating at near rated power when an Instrument Air system pipe ruptured, causing Instrument Air pressure to drop rapidly to 0 psig.

The FWRVs will fail ___(1)___.

If RPV water level is dropping and feedwater is needed for injection, while executing the DEOPs, the Unit Supervisor is required to direct Operators to ___(2)___.

A.

(1) immediately; (2) valve in Main Bank Nitrogen bottles, per DOA 4700-01, INSTRUMENT AIR SYSTEM FAILURE B.

(1) immediately; (2) install handwheels and manually operate the FWRVs, per DOA 0600-01, TRANSIENT LEVEL CONTROL C.

(1) after 30 minutes; (2) valve in Main Bank Nitrogen bottles, per DOA 4700-01, INSTRUMENT AIR SYSTEM FAILURE D.

(1) after 30 minutes; (2) install handwheels and manually operate the FWRVs, per DOA 0600-01, TRANSIENT LEVEL CONTROL Answer:

D Question 81 Details Topic:

81 - 295019.G.2.4.8 Comments: Objective: DRE278LN001.12

Reference:

DOA 4700-01, DOA 0600-01 K/A: 295019.G.2.4.8 3.8 / 4.5 K/A: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Safety Function: 8 PRA: Yes Level: High Pedigree: Bank History: 2012 NRC; 2010 NRC Explanation: Per the critical component failure table in the Loss of Inst Air DOA, the FWRVs will operate on their backup nitrogen source for 30 minutes after a loss of Inst Air. The DOA for transient level control states that if the FWRVs can not be opened due to loss of air and feedwater/alternate injection systems are required by the DEOPs, then operate the valves manually as directed by the Unit Supervisor.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 82 of 100 ILT 11-1 (NRC 2012-301) 82 Points: 1.00 Which of the following provide the bases to maintain Torus water level above -4.5 inches, in Modes 1, 2, and 3?

Ensures a sufficient amount of water.....

A.

with the Minimum CST Volume, Long-Term Cooling is available for the Design Basis Accident.

B.

would be available to adequately condense the steam from the relief valve quenchers ONLY.

C.

with the Minimum CST Volume, in the event of a LOCA to permit recirculation cooling flow to the core.

D.

would be available to adequately condense the steam from the relief valve quenchers, downcomer lines, or HPCI turbine exhaust line.

Answer:

D Question 82 Details Topic:

82 - 295030.G.2.2.22 Comments: Objective: DRE223LN001.07

Reference:

Bases 3.6.2.2 K/A: 295030.G.2.2.22 4.0 / 4.7 K/A: Low Suppression Pool Water Level: Knowledge of limiting conditions for operations and safety limits.

Safety Function: 5 PRA: No Level: Memory Pedigree: Bank History: 2007 NRC, 2012 NRC Explanation: Answer is per the above bases. CST is not part of bases. If the suppression pool water level is too low, an insufficient amount of water would be available to adequately condense the steam from the relief valve quenchers, downcomer lines, or HPCI turbine exhaust line.

SRO per Criteria: 2 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 83 of 100 ILT 11-1 (NRC 2012-301) 83 Points: 1.00 Unit 2 was operating at near rated conditions with the 902-36 back-panel recorder TIRS 2-1640-200A, TORUS TEMP MON DIV I, Out Of Service (O.O.S.) due to a failed power supply. The following conditions exist:

RPV pressure is 1000 psig.

HPCI testing per DOS 2300-03 "HIGH PRESSURE COOLANT INJECTION SYSTEM OPERABILITY AND QUARTERLY IST VERIFICATION TEST" is in progress.

TIRS 2-1640-200B currently indicates the following:

Point 1 130oF Point 2 85oF Point 3 100oF Point 4 115oF Point 5 107oF Point 6 116oF Point 7 101oF Point 8 107oF What action(s) is/are required based on the current readings?

A.

Start all available Torus cooling ONLY.

B.

Start all available Torus cooling and scram ONLY.

C.

Secure HPCI testing and start all available Torus cooling ONLY.

D.

Secure HPCI testing, start all available Torus cooling, AND scram.

Answer:

C Question 83 Details Topic:

83 - 295013.A2.01 (1)

Comments: Objective: 29502LK011

Reference:

DEOP 200-1 K/A: 295013.A2.01 3.8 / 4.0 K/A: Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Suppression pool temperature.

Safety Function: 5 PRA: No Level: High Pedigree: Modified History: 2012 NRC Comments: The average torus temperature is 107.6°F. Per DEOP 200-1 the actions that need to be taken are secure HPCI testing and start all available Torus cooling.

SRO per Criteria: 5 REQUIRED

REFERENCES:

DEOP charts, with the entry conditions blanked out.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 84 of 100 ILT 11-1 (NRC 2012-301) 84 Points: 1.00 Unit 3 is at full power.

When the following occurs:

903-4 C-19, LPCI/CS EAST SUMP LVL HI is received in the control room.

An EO reports a large unisolable leak on the Core Spray suction piping and there is 10 inches of water on the floor in the East Corner Room.

20 minutes later:

903-4 D-19, LPCI/CS WEST SUMP LVL HI is received in the control room.

An EO reports there is 8 inches of water on the floor in the West Corner Room.

Based on these conditions, at a minimum, the SRO is required to direct ____(1)____ and the basis for this is ____(2)____.

A.

(1) a reactor scram per DGP 02-03 (2) a direct threat exists relative to secondary containment integrity, to equipment located in the secondary containment and to continued safe operation of the plant.

B.

(1) a reactor shutdown per DGP 02-01 (2) a direct threat exists relative to secondary containment integrity, to equipment located in the secondary containment and to continued safe operation of the plant.

C.

(1) a reactor scram per DGP 02-03 (2) to reduce to decay heat levels the energy that the RPV may be discharging to the secondary containment.

D.

(1) a reactor shutdown per DGP 02-01 (2) to reduce to decay heat levels the energy that the RPV may be discharging to the secondary containment.

Answer:

B Question 84 Details Topic:

84 - 295036.G.2.4.18 Comments: Objective: 295L049

Reference:

BWROG EPG/SAGs, Appendix B (DEOP Bases)

K/A: 295036.G.2.4.18 K/A: Secondary Containment High Sump/Area Water Level: Knowledge of the specific bases for EOPs Safety Function: 5 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: Examinee must identify that a leak from core spray suction is not a leak from a primary system and a reduction in RPV pressure will not effect the leak rate.

a) Incorrect - the reactor would be scrammed if a primary system was discharging and unisolable.

b) Correct - Per BWROG EPG/SAGs, Appendix B c) Incorrect - the reactor would be scrammed if a primary system was discharging and unisolable d) Incorrect - this is the basis for scramming the reactor with an unisolable leak.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 85 of 100 ILT 11-1 (NRC 2012-301) 85 Points: 1.00 Unit 2 is at 6% power with a startup in progress.

The RBCCW supply valves to the Drywell fail closed.

RBCCW expansion tank level is lowering at 1 inch every 10 minutes.

Drywell pressure is 1.3 psig and rising 0.1 psig every 10 minutes.

Drywell temperature is 110°F and rising 1°F every 5 minutes.

The reactor is scrammed per DGP 02-03, REACTOR SCRAM.

To mitigate the effects on containment, the Unit Supervisor will direct actions per __________.

A.

DOA 0040-01, SLOW LEAK B.

DEOP 0200-01, PRIMARY CONTAINMENT CONTROL C.

DOP 1600-01, NORMAL PRESSURE CONTROL OF THE DRYWELL OR TORUS D.

DOP 5750-11, DRYWELL COOLING SYSTEM OPERATION Answer:

C Question 85 Details Topic:

85 - 295010.G.2.4.11 Comments:

Objective: 29502LK001

Reference:

DAN 902-3 A-13 K/A: 295010.G.2.4.11 4.0 / 4.2 K/A: High Drywell Pressure: Knowledge of abnormal condition procedures.

Level: High Pedigree: New History: 2012 NRC Comments:

a) Incorrect - DOA 40-1 directs action to combat a slow primary system leak.

b) Incorrect - DEOP 0200-01 does not have an entry condition based on the given conditions c) Correct - DAN 902-3 A-13, DRYWELL PRESS HI, directs venting containment per DOP 1600-01 d) Incorrect - DOP 5750-11 lists RBCCW in operation as a prerequisite to start the Drywell Cooling system SRO per criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 86 of 100 ILT 11-1 (NRC 2012-301) 86 Points: 1.00 Unit 2 was operating at near rated power, when a transient occurred, resulting in the following sequence of events:

The Reactor scrammed.

MCC 29-2 experienced an overcurrent condition.

ALL RFPs and HPCI tripped and would NOT start.

An NSO reported RPV water level of -72 inches and trending down.

An NSO reported RPV pressure of 925 psig and trending down.

An NSO placed the SBLC keylock switch to the "SYS 1 & 2" position.

An NSO reported SBLC discharge pressure of 1500 psig and BOTH SQUIB A and B lights LIT Currently ___(1)___ SBLC pump is injecting into the RPV and the Unit Supervisor is required to direct ___(2)___.

A.

(1) neither; (2) injecting alternate SBLC per DEOP 500-1, ALTERNATE STANDBY LIQUID CONTROL INJECTION.

B.

(1) neither; (2) placing the SBLC keylock switch to the "SYS 2 & 1" position per DOP 1100-02, INJECTION OF STANDBY LIQUID CONTROL.

C.

(1) one; (2) verify RWCU system isolation per DOP 1100-02, INJECTION OF STANDBY LIQUID CONTROL, locally.

D.

(1) one; (2) placing the SBLC keylock switch to the "SYS 2 & 1" position per DOP 1100-02, INJECTION OF STANDBY LIQUID CONTROL.

Answer:

B Question 86 Details Topic:

86 - 211000.A2.04 Comments: Objective: DRE211LN001.08

Reference:

DOP 1100-02 K/A: 211000.A2.04 3.1 / 3.4 K/A: Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow.

Level: High Pedigree: New History: 2012 NRC Comments: b) With the pump discharge pressure 1500 psig, it is an indication that the relief valve has lifted (above relief setpoint) and is directing pump flow back to the main tank and NOT the RPV. DOP 1100-02 directs placing the SBLC switch to the opposite position selected first.

a) Injecting with alternate boron is done when SBLC is needed for an ATWS, NOT water injection.

c) Credible if examinee believes MCC 29-2 is the power supply to RWCU Isolation valves (MCC 28-1) d) Credible if examinee believes MCC 29-2 is power supply to B SBLC pump (MCC 29-1)

SRO per criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 87 of 100 ILT 11-1 (NRC 2012-301) 87 Points: 1.00 Unit 2 was shutdown, with the following set of conditions:

NO Recirc pumps are running.

RPV water temperature is 320°F.

2A SDC pump is running and aligned to the RPV.

2B SDC pump is running and aligned to the RPV.

2C SDC pump is running and aligned to the FPC system.

If Recirc Loop water temperature trends up to 360°F, what action(s) is/are the Unit Supervisor required to direct?

A.

Control RPV water temperature/pressure per DOA 1000-1, RESIDUAL HEAT REMOVAL ALTERNATIVES only.

B.

Align 2C SDC pump to the RPV to control RPV water temperature/pressure per DOP 1000-3, SHUTDOWN COOLING MODE OF OPERATION only.

C.

Align 2C SDC pump to the RPV to control RPV water temperature/pressure per DOP 1000-3, SHUTDOWN COOLING MODE OF OPERATION and monitor fuel storage pool water level once per hour per DOA 1900-1, LOSS OF FUEL POOL COOLING.

D.

Monitor fuel storage pool water level once per hour per DOA 1900-1, LOSS OF FUEL POOL COOLING and control RPV water temperature/pressure per DOA 1000-1, RESIDUAL HEAT REMOVAL ALTERNATIVES.

Answer:

A Question 87 Details Topic:

87 - 205000.A2.01 (2)

Comments: Objective: DRE205LN001.06

Reference:

DOA 1000-01, DOA 1900-01, DAN 092-4 B-23 K/A: 205000.A2.01 3.1 / 3.3 K/A: Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Recirculation loop high temperature.

Safety Function: 4 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: The SDC Pump Suction temperature (Recirc Loop Temp) exceeds the setpoint of 339°F for SDC pump trip, for the pumps lined up to the RPV, which causes 'A' and 'B' to trip. The 'C' pump will NOT trip, as it still has suction from the Fuel Pool Cooling system. Entry into DOA 1000-01 is the only required entry.

SRO per criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 88 of 100 ILT 11-1 (NRC 2012-301) 88 Points: 1.00 Given the following conditions:

Both units were at full power.

2/3A SBGT train is running for DOS 7500-02, SBGT SYSTEM SURVEILLANCE AND IST TEST.

2/3B SBGT train is in STANDBY.

2/3A SBGT Temperature indicates 0°F at the 923-5 panel.

An EO reports that 2/3A SBGT local temperature is 250°F.

Based on these conditions, the Unit Supervisor directs the NSO to:

A.

Place the 2/3A SBGT train to OFF, ONLY, per DOS 7500-02.

B.

Place the 2/3B SBGT train to OFF and the 2/3A SBGT train to OFF per DOS 7500-02.

C.

Place the 2/3B SBGT train to START and the 2/3A SBGT train to OFF per DAN 923-5 A-6, SBGT GAS TRT SYS A TROUBLE D.

Place the 2/3B SBGT train to OFF and the 2/3A SBGT train to OFF per DAN 923-5 A-6, SBGT GAS TRT SYS A TROUBLE Answer:

C Question 88 Details Topic:

88 - 261000.G.2.2.44 Comments: Objective: 261LN001.08

Reference:

923-5 A-6 K/A: 261000.G.2.2.44 4.2 / 4.4 K/A: Standby Gas Treatment System: 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.

Safety Function: 9 PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: The correct answer is C. A control room indication of 0°F is indicative of remote temperature indication failure when SBGT is running.

A charcoal adsorber temperature of 234°F or higher requires placing the train in the cooling mode. This is accomplished by placing 2/3B SBGT train (the STANDBY train) to START and placing 2/3A SBGT train (the PRIMARY train with high charcoal temperature) to OFF.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 89 of 100 ILT 11-1 (NRC 2012-301) 89 Points: 1.00 Unit 2 was operating at near rated power, with 2A LPCI pump taken O.O.S. at 0100 on July 1, 2012, and the following conditions exist:

2A Core Spray pump is running per DOS 1400-05, CORE SPRAY SYSTEM PUMP OPERABILITY AND QUARTERLY IST TEST WITH TORUS AVAILABLE.

System Discharge pressure is 235 psig.

System flow is 4400 gpm with MO 2 1402 4A, FLOW TEST VLV full open.

What is the impact to the Core Spray system and what actions are required?

A.

2A Core Spray subsystem is INOP; immediately enter LCO 3.0.3 for shutdown B.

2A Core Spray subsystem is INOP; restore Core Spray subsystem to operability by 0100 on July 8, 2012 C.

2A Core Spray subsystem is OPERABLE restore LPCI subsystem to operability by 0100 on July 8, 2012 D.

2A Core Spray subsystem is OPERABLE; restore 2A LPCI pump to operability by 0100 on July 31, 2010 Answer:

A Question 89 Details Topic:

89 - 209001.G.2.2.40 Comments: Objective: DRE209LN001.07

Reference:

T.S. 3.5.1, DOS 1400-05 K/A: 209001.G.2.2.40 3.4 / 4.7 K/A: Low Pressure Core Spray System: Ability to apply Technical Specifications for a system.

Level: High Pedigree: Bank History: 2007 NRC, 2009 Cert, 2012 NRC Explanation: Per DOS 1400-5 the Core Spray pump must achieve a flowrate of 4600 gpm at greater than 235 psig. Since the flowrate is not sufficient the A Core Spray pump is INOP and puts you in a 7 day clock.

The 'A' LPCI pump INOP means that you have 2 low pressure subsystem INOP driving to an immediate shutdown per T.S. 3.0.3.

SRO per Criteria: 2 REQUIRED

REFERENCES:

Tech Spec 3.5.1 with less than one hour times blanked out.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 90 of 100 ILT 11-1 (NRC 2012-301) 90 Points: 1.00 Given the following conditions:

An earthquake occurred Unit 2 and 3 scrammed from full power Unit 3 steam leak due to a packing failure of 3-2301-5, STEAM ISOL VLV which fails to operate Unit 3 lost all AC electrical power Unit 3 reactor water level -60 inches trending up at 2 inches per minute Unit 2 supplied by the Unit 2/3 Emergency Diesel Generator ONLY suppling Buses 23-1, 28 and 29 Which of the following describes the order of procedures the Unit 3 Unit Supervisor should direct to mitigate the transient?

1. DEOP 0300-01, SECONDARY CONTAINMENT CONTROL

2. DEOP 0500-03, ALTERNATE WATER INJECTION SYSTEMS

3. DOP 6620-06, POWERING UNIT 3 4KV BUSES VIA THE SBO DG 3 A.

1, 2, 3 B.

2, 1, 3 C.

3, 1, 2 D.

3, 2, 1 Answer:

C Question 90 Details Topic:

90 - 262001.G.2.4.5 Comments: Objective: 295L089

Reference:

DGA-12; DOA 6500-01; DGP 02-03; DEOP 0100 K/A: 262001.G.2.4.5 3.7 / 4.3 K/A: A.C. Electrical Distribution: Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

Safety Function: 6 PRA: Yes Level: High Pedigree: New History: 2012 NRC Comments: The priority of actions should be to restore electrical power, then stopping the steam leak then addressing reactor water level control.

SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 91 of 100 ILT 11-1 (NRC 2012-301) 91 Points: 1.00 Unit 2 was in STARTUP, with the RWCU system in operation and RPV pressure was 300 psig, when the Instrument Air supply line to RWCU FCV 2-1219, ruptured.

The Unit Supervisor is required to direct ________.

A.

reducing system flow per DAN 902-4 C-12 RWCU RECIRC PP DISCH PRESS LO B.

an operator to check the breaker relay targets per DAN 902-4 A-10 RWCU RECIRC PP TRIP C.

verification of system isolation per DOP 1200-03, RWCU SYSTEM OPERATION WITH THE REACTOR AT PRESSURE hard card D.

securing the RWCU system per DOP 1200-03, RWCU SYSTEM OPERATION WITH THE REACTOR AT PRESSURE hard card Answer:

D Question 91 Details Topic:

91 - 204000.A2.03 Comments: Objective: DRE204LN001.12

Reference:

M-30, DOA 4700-01, DOP 1200-03, DANs 902-4 A-10 & C-12 K/A: 204000.A2.03 2.9 / 2.9 K/A: Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Flow control valve failure.

Level: High Pedigree: Bank History: 2011 NRC, 2009 NRC Explanation: Loss of IA to the FCV causes valve to fail closed (does not lock up). The pump experiences a dead headed situation (NOT a pump trip). Per DOP 1200-03, during a system transient, secure the system via the hard card. The high pressure isolation (downstream of the PCV) would not occur, since flow was isolated by the failing closed of the FCV.

SRO Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 92 of 100 ILT 11-1 (NRC 2012-301) 92 Points: 1.00 Unit 3 was operating at near rated power, when a transient occurred with the EHC Pressure Controller Processors.

The following parameters are reported:

RPIS has failed.

IRMs are reading 50 on range 7 and decreasing.

RPV pressure is 1065 psig and trending up at 1 psig/minute.

RPV water level dropped to -12 inches and trending up at 1 inch/minute.

Per the DEOPs, the Unit Supervisor is required to direct __________.

A.

reducing RPV pressure to 945 psig maximum RPV water level band of +8 to +48 inches B.

stabilizing RPV pressure below 1060 psig maximum RPV water level band of -185 to +48 inches C.

reducing RPV pressure to 945 psig maximum RPV water level band of -40 to -50 inches D.

stabilizing RPV pressure below 1060 psig maximum RPV water level band of -164 to +48 inches.

Answer:

D Question 92 Details Topic:

92 - 241000.G.1.07 Comments: Objective: DRE248LN001.12

Reference:

DEOP 100 K/A: 241000.G.1.07 4.4 / 4.7 K/A: Reactor/Turbine Pressure Regulating System: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Level: High Pedigree: Modified History: 2012 NRC Comments:

a) Incorrect - If ADS valves were cycling (RPV pressure > 1130psig) these would be possible bands per the DEOPs b) Incorrect - If -185 is the low limit for level during steam cooling.

c) Incorrect - If the second override of DEOP 400-5 were used and ADS valves were cycling, this would be the band for level and pressure.

d) Correct - Because it is unknown if the reactor will stay shutdown under all conditions (RPIS Failure, IRM power on scale), DEOP 400-5 is entered.

With reactor power less than 6%, the correct level bands are -164 to 48 inches; RPV <1060 psig.

SRO per criteria: 5 REQUIRED

REFERENCES:

DEOP charts, with the entry conditions blanked out.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 93 of 100 ILT 11-1 (NRC 2012-301) 93 Points: 1.00 Unit 2 was operating at near rated conditions, with the 2A Fuel Pool Cooling (FPC) pump O.O.S.

for motor replacement, when Bus 29 de-energized due to an overcurrent condition.

The Unit Supervisor is required to direct monitoring fuel storage pool water temperatures and level once per hour ___(1)___ per ___(2)___.

A.

(1) locally; (2) DOA 1900-01, LOSS OF FUEL POOL COOLING B.

(1) locally; (2) DOA 1000-01, RESIDUAL HEAT REMOVAL ALTERNATIVES C.

(1) from the Main Control Room; (2) DOA 1900-01, LOSS OF FUEL POOL COOLING D.

(1) from the Main Control Room; (2) DOA 1000-01, RESIDUAL HEAT REMOVAL ALTERNATIVES Answer:

A Question 93 Details Topic:

93 - 233000.A2.09 (1)

Comments: Objective: DRE233LN001.08

Reference:

DOA 1900-01 K/A: 233000.A2.09 2.7 / 2.9 K/A: Ability to (a) predict the impacts of the following on the FUEL POOL COOLING AND CLEAN-UP ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C. electrical power failures.

Level: High Pedigree: Bank History: 2012 NRC Comments: With A FPC pump O.O.S. and a subsequent loss of B FPC pump (loss of Bus 29), there is no forced cooling of the fuel storage pool.

Per DOA 1900-01, when forced cooling of the fuel storage pool is not available, then monitor fuel storage pool water temperatures and level once per hour LOCALLY (Main Control Room indication of spent fuel pool temperature may be erroneous due to water stagnation).

SRO per criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 94 of 100 ILT 11-1 (NRC 2012-301) 94 Points: 1.00 Unit 2 has a spiral core offload in progress.

Which of the following is REQUIRED to be performed prior to removing control rod XX-YY from the core for blade replacement?

A.

Ensure all other control rods are fully inserted.

B.

Ensure that the control cell for control rod XX-YY is defueled.

C.

Three SRMs are operable with at least one in an adjacent quadrant.

D.

Ensure that the REFUEL INTERLOCKS for control rod XX-YY have been defeated.

Answer:

B Question 94 Details Topic:

94 - Generic.2.1.42 (3)

Comments: Objective: 29800LK084

Reference:

DFP 0800-16 K/A: Generic.1.42 2.5 / 3.4 K/A: Knowledge of new and spent fuel movement procedures.

Level: Memory Pedigree: Bank History: 2012 NRC Explanation:

a) Incorrect - Is a requirement of the Tech Spec for single control rod drive removal 3.10.4 b) Correct - Per DFP 0800-16 prerequisites, the control cell for control rod XX-YY must be defueled prior to removing the control rod.

c) Incorrect - Per T.S. 3.3.1.2, only 1 SRM s required to be in operation when spiral core offload is in progress.

d) Incorrect - Refueling interlocks are not required to be defeated to lift the rod out of the top core, only to withdraw the rod out the bottom for testing.

SRO per criteria: 7 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 95 of 100 ILT 11-1 (NRC 2012-301) 95 Points: 1.00 During performance of DOS 1500-02 the following pump flows were recorded:

Unit 2: 2B CCSW pump 3610 gpm Unit 3: 3B CCSW pump 3590 gpm The Unit Supervisor will declare _______.

A.

BOTH 2B AND 3B CCSW Pumps NOT operable B.

BOTH 2B AND 3B CCSW Pumps operable C.

the 2B CCSW Pump operable BUT the 3B CCSW Pump NOT operable D.

the 3B CCSW Pump operable BUT the 2B CCSW Pump NOT operable Answer:

D Question 95 Details Topic:

95 - Generic.2.2.37 Comments: Objective: DRE277LN001.07

Reference:

DOS 1500-02, DOS 1500-12, T.S. Bases 3.7.1, UFSAR 9.2 K/A: Generic.2.2.37 3.6 / 4.6 K/A: Ability to determine operability and/or availability of safety related equipment.

PRA: Yes Level: High Pedigree: Bank History: 2012 NRC Explanation: The Unit 2 CCSW Pumps must have a flowrate > 3621 gpm (121 gpm for CREVs). Unit 3 CCSW Pumps must have a flowrate of

>3500. If these flow rates cannot be achieved, the associated pump must be declared inoperable.

SRO per Criteria: 1 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 96 of 100 ILT 11-1 (NRC 2012-301) 96 Points: 1.00 A spent fuel cask is being moved on Unit 2.

The following events occur:

Reactor Building Overhead Crane Radiation Monitor High Radiation alarms locally.

Refuel Floor ARM High Radiation alarms locally.

Overhead Crane dose rates are measured at 40 mr/hr.

Refuel Floor dose rates are measured at 150 mr/hr.

Which of the following describe the effect of the above conditions?

A.

SBGT has started and Reactor Building Overhead Crane movement is inhibited in the hoist raise function ONLY.

B.

SBGT has started and Reactor Building Overhead Crane movement is inhibited in the hoist raise and lower function.

C.

Reactor Building Ventilation system is running and Reactor Building Overhead Crane movement is inhibited in the hoist raise function ONLY.

D.

Reactor Building Ventilation system is running and Reactor Building Overhead Crane movement is inhibited in the hoist raise and lower function.

Answer:

A Question 96 Details Topic:

96 - Generic.2.3.5 Comments: Objective: 299L042

Reference:

DFP 0850-03 K/A: Generic 2.3.5 2.9 / 2.9 K/A: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

PRA: No Level: High Pedigree: New History: 2012 NRC Comments:

a) Correct - DFP 0850-03 states RB Vent isolates and SBGT starts with Refuel Floor Rad > 100 mr/hr and RB Crane hoist raise is inhibited > 30 mr/hr.

b) Incorrect - The RB Crane lower function is not inhibited > 30 mr/hr.

c) Incorrect - RB Vent isolates with Refuel Floor rad >100 mr/hr.

d) Incorrect - RB Vent isolates with Refuel Floor rad >100 mr/hr.

SRO per Criteria: 9 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 97 of 100 ILT 11-1 (NRC 2012-301) 97 Points: 1.00 Unit 2 is operating near rated power.

HVOs and EMD have been trying to locate and isolate an 80 volt ground on the 125 VDC System.

The Unit 2 NSO then notices that two previously lit annunciators are now extinguished.

A test of the annunciators shows ALL of the annunciators on the 902-3, 5, 6, and 8 panels have been lost.

20 minutes later, the cause of the loss of the annunciators was corrected.

What is the minimum EP classification level (if any) that must be declared? (Exclude discretionary EALs)

A.

NO EP Classification B.

Unusual Event C.

Alert D.

Site Area Emergency Answer:

B Question 97 Details Topic:

97 - Generic.2.4.41 Comments: Objective: 29501LP032

Reference:

EP-AA-1004 K/A: Generic.2.4.41 K/A: Knowledge of the emergency action level thresholds and classifications.

PRA: No Level: High Pedigree: Bank History: 2012 NRC Comments: The correct EAL is MU4 because ALL of the annunciators 902-3, 5 and 8 panels have been lost for >15 minutes (EAL Threshold Value). To upgrade to an Alert requires a significant transient in progress OR the loss of one of the following: 1) all indication needed to monitor criticality 2) core heat removal 3) fission product barrier status.

SRO per Criteria: 5 REQUIRED

REFERENCES:

EP-AA-1004 EAL Charts.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 98 of 100 ILT 11-1 (NRC 2012-301) 98 Points: 1.00 A Site Area Emergency has been declared.

The automated call out system is unavailable.

Which of the following emergency response personnel MUST be notified using pagers?

A.

Station ERO personnel ONLY B.

Station and Corporate ERO personnel ONLY C.

Station ERO personnel and NDO ONLY D.

Station and Corporate ERO personnel and NDO Answer:

D Question 98 Details Topic:

98 - Generic.2.4.40 Comments: Objective: 295L160

Reference:

EP-AA-112-100-F-06 K/A: Generic 2.4.40 K/A: Knowledge of SRO responsibilities in emergency plan implementation.

PRA: No Level: Memory Pedigree: New History: 2012 NRC Comments:

a) Incorrect - Unusual Events are not required to have offsite notifications when using the automated call out system.

b) Incorrect - Could be selected if not known that NDO is notified using the pager system.

c) Incorrect - Could be selected if not known that EOF is required to be staffed during a site area emergency.

d) Correct - Per EP-AA112-100-F-06 step 4 SRO per Criteria: 5 REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 99 of 100 ILT 11-1 (NRC 2012-301) 99 Points: 1.00 DOP 2000-110, Attachment 1: WASTE SURGE TANK RADIOACTIVE WASTE DISCHARGE TO RIVER CARD, contains the calculation for determining the ___(1)___ flowrate and radiological monitor alarm setpoints, and, excluding designees, is REQUIRED to be verified by ___(2)___.

A.

(1) dilution (2) Unit Supervisor B.

(1) dilution (2) Shift Manager C.

(1) discharge (2) Unit Supervisor D.

(1) discharge (2) Shift Manager Answer:

D Question 99 Details Topic:

99 - Generic.3.06 Comments: Objective: DRE268LN001.14

Reference:

DOP 2000-110 K/A: Generic.3.06 2.0 / 3.8 K/A: Ability to approve release permits.

Level: Memory Pedigree: Bank History 2012 NRC Explanation: The calculations that must be completed prior to starting the release determine the maximum discharge rate (dilution is provided on the river card as a constant). The calculations must then be verified by the Shift Manager or designee, to approve the river discharge.

SRO per criteria 4.

REQUIRED

REFERENCES:

None.

EXAMINATION ANSWER KEY 11-1 (2012-301) NRC Exam DRESDEN GENERATING STATION Page: 100 of 100 ILT 11-1 (NRC 2012-301) 100 Points: 1.00 Both units are operating at near rated power with the following conditions:

Fuel handlers are inspecting new fuel on the Refuel Floor.

The 2/3A SBGT train is in day 2 of a seven day OOS for planned maintenance.

The 2/3B SBGT train control switch is in the PRI position.

What required action is initiated IF the planned maintenance on the 2/3A SBGT train lasted past the planned 7 day duration?

A.

Place the 2/3B SBGT train in operation.

B.

Place both units in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C.

Restore the 2/3A SBGT train to OPERABLE status.

D.

Immediately suspend movement of new fuel assemblies.

Answer:

B Question 100 Details Topic:

100 - Generic.2.23 (1)

Comments: Objective: DRE299LN001.04

Reference:

Tech Spec Section 3.6.4.3 K/A: Generic.2.23 3.1 / 4.6 K/A: Ability to track Technical Specification limiting conditions for operations.

Level: High Pedigree: Bank History: 2007 NRC, 2011 NRC Explanation: This knowledge can be gained from the Tech Specifications section 3.6.4.3 action B.1.

Action B.1 states be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for correct answer to the question.

Fuel handlers are moving new fuel which is NOT irradiated fuel as the applicability statement states, so no need to stop fuel inspections nor place the OPERABLE SBGT subsystem in operation. Already have one OPERABLE (2/3B) SBGT subsystem so no need for restoration.

SRO per Criteria: 2 REQUIRED

REFERENCES:

TS 3.6.4.3, with less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> times removed.