Text

Draft - RO & SRO Written Exam (Folder 2)
	ML070380292
Person / Time
Site:	Peach Bottom
Issue date:	11/22/2006
From:	Sykes M D; Operations Branch I
To:	Crane C M; Exelon Generation Co, Exelon Nuclear
	Sykes, Marvin D.
Shared Package
ML060800089	List: IR 05000277/2007301; ML070310224; ML070380284; ML070380292; ML070380304; ML070380334; ML070380337; ML070610478; ML070610491; ML070610496; ML070610499; ML070610509; ML070610511; ML070790539; ML071010332; ML071020269;
References
	Download: ML070380292 (196)
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0 N LOO? SNV NOIlVNIVVVX3 Question 1 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1540-3-011 Unit 2 is operating at 95% power when a recirculation flow reduction event results I268 Active No No 2.50 2 1 .oo 295001.AA2.02 N-ILT-1540-3-011 0.00 0.00 Importance:

RO 3.1 / SRO 3.2 Cognitive-Level:

Memory

References:

OT-I 12 Justification:

A. B. Incorrect - RBM not referenced as a nuclear monitoring instrument for THI. Correct - Core Thermal Hydraulic Instability (THI) may be occurring if any of the following conditions exist: *Steadily increasing confirmation counts on OPRM display with few to no resets.

Any APRM flux noise signal grows by 2 or more times its initial level,

APRM flux oscillations rise greater than or equal to 10% (peak to peak). Incorrect - No reference to period indication as a nuclear monitoring instrument for THI. Incorrect - Steadily increasing confirmation counts on OPRM display causing repetitive "OPRM Pre-trip Condition" alarms is indication of THI. C. D. \4 NRC EXAM Page: 2 of 144 12/22/06

Question 2 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5057-6A-003 A Station Blackout has occurred.

In accordance with SE-11, "Loss of Off-Site Powe 1269 Active No NO 3.00 2 1 .oo 295003AA1.04 N-l LT-5057-6A-003 0.00 0.00 Importance:

RO 3.6 I SRO 3.7

References:

SE-11, DBD P-S-OIA, P-T-13"Station Blackout" Justification:

A. *3. C. D. Incorrect - Only selected MCR annunciators are left with DC power. The remainder will be shed per SE-11, Attachment T. Incorrect - Even if a DC load shed is initiated immediately after the loss of power, then battery life may be extended beyond 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , but not as long as 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The inaccurate 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months time frame is used as a plausible distracter for any examinee who does not recall 2 or 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months battery limits under station blackout conditions.

Correct - Per SE-11 Bases the DC Load Shed on non-essential loads is to assume enough power for equipment such as ECCS Logic, RClC Logic and Control, Diesel Logic & Control, SRVs .... which are required for adequate core cooling and circuit breaker control to help restore AC power. Incorrect -The Main Control Room Emergency ventilation control power is AC, not DC. L-.' NRC EXAM Page: 4 of 144 12/22/06

-1 Question 3 Details 8 Question Type: Multiple Choice ToDic: N-ILT-5057-3C-002 Effect of a charger malfunction on System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: DC battery 1176 Active No No 3.00 2 1 .oo 295004 AK2.02 A 0.00 0.00 Importance:

RO 3.0 Cognitive-Level:

High N-l LT-5057-3C-002 SRO 3.

Reference:

PLOT 5057, Objective 3c; E-26 Justification:

A. 6. C. D. Incorrect - the battery will fully support all loads for approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with no battery charger; the bus will remain energized.

Correct - when the output breaker for charger 2BD003-1 trips, the charger no longer supplies power to the Division II 250 VDC bus. The bus loads would then be supplied by the 26 and 2D batteries.

The batteries are designed to supply loads during a DBA for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Incorrect - when the output breaker for charger 2BD003-1 trips, the charger no longer supplies power to the Division II 250 VDC bus. The bus loads would then be supplied by BOTH the 26 and 2D batteries.

Incorrect - charger 2BD003-2 must be manually placed in service ... only one charger can be in service at a time. The question stem states "assuming no operator actions." ,L/' NRC EXAM Page: 6of 144 12/22/06 ON ANS 07 NRC RO Rev 0 --.-.,, 4 ID: N-lLTb060F-11-001:

Points: 1.00 0 0 Unit 2 is operating at 20% power. A ground fault on the Grid results in the Main Generator output circuit breakers CB 215 and CB 225 automatically opening. Which one of the following describes the reaction of the plant to this trip? ASSUME NO OPERATOR ACTIONS.

A. The reactor will remain at power with the Main Turbine remaining in operation. Main Turbine Bypass Valves will automatically open maintaining reactor pressure.

B. The reactor will scram following the closure of the Main Turbine Stop and Control Valves. Main Turbine Bypass Valves will automatically open maintaining reactor pressure.

C. The reactor will remain at power with the Main Turbine Stop and Control Valves closing. Main Turbine Bypass Valves will automatically open maintaining reactor pressure.

D. The reactor will scram due to a Main Generator Lockout. Main Turbine Bypass Valves will automatically open maintaining reactor pressure.

Answer: C u' 12/22/06 NRC EXAM Page: 7 of 144

'U' EXAMINATION-ANSWER KEY vo Question 5 betails Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-5006-1 A-003 RPV level response on a scram 1259 Active No No 2.00 2 1 .oo 259006 AK3.01 N-l LT-5006-1 A-003 0.00 0.00 Importance: RO 3.8

/ SRO 3.9 Cognitive-Level:

Memory

Reference:

PLOT 5006, OBJ. la; Fundamentals Justification:

A. Incorrect - RPV level will not swell on a reactor scram ... the void collapse causes RPV level to shrink. B. Correct - the scram causes power

/ heat rate to decrease which causes voids to collapse which causes level to lower. The Digital Feedwater Control System (DFCS) sees the lower level and increases the speed of the RFPs.

C. Incorrect - DFCS will increase RFP speed to compensate for the lowering RPV water level. Also, voids will collapse causing RPV level to shrink, not swell. D. Incorrect - DFCS will increase RFP speed to compensate for the lowering RPV water level. -.-.-. NRC EXAM Page: 10 of 144 12/22/06 EXAMINATION ANSWER KEY 4 6 ID: N-ILT-1555-1414 Points: 1.00 Which one of the following is the reason why they reactor is SCRAMMED prior to evacuating the Main Control Room in accordance with SE-1, "Plant Shutdown from the Remote Shutdown Panel"? A. Ensures that inventory makeup requirements will be within HPCl capability.

B. Ensures that inventory makeup requirements will be within RClC capability.

C. Precludes rapid reactor vessel depressurization in the event that the main turbine bypass valves fail open. D. Scramming from outside the Control Room would require RPS bus power to be tripped causing concurrent isolations of all PClS valve groups. Answer: B Question 6 Details L" Question Type: Topic: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1

User Number 2: Comment: Multiple Choice N-ILT-1555-1-014 Which one of the following is the reason why they reactor is SCRAMMED prior to evac 1271 Active No No 3.00 2 1 .oo 295016.AK3.01 N-ILT-1555-1-014 0.00 0.00 Importance:

RO 4.1 / SRO 4.2 Cognitive-Level: Memory

References:

SE-1 Justification:

A. Incorrect - HPCl is used only in SE-10 at the Alternate Shutdown Panel and not applicable for this condition.

Correct - In accordance with SE-1. Incorrect - Per SE-1 bases this is the reason for closing the MSIVs, not scramming the reactor. Incorrect - MSlVs are manually closed prior to evacuation and all, Group Isolations are expected during SE-I. B. C. D. NRC EXAM Page: 11 of 144 12/22/06 P, 00' C :s)U!Od COO-31E-SE09-111-N

ai L a3aN L A3Y U3MSNV NOllVNIVVVX3

\-- INATION ANSWER KEY . 2007 NRC 0 Question 7 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number I: User Number 2: Comment: Multiple Choice N-ILT-5035-3E-001 Unit 3 is at 100% power. 3A RBCCW pump was in service and then tripped on overcur 1272 Active No No 3.00 2 1 .oo 29501 8AK3.03 N-ILT-5035-3E-001 0.00 0.00 Importance:

RO 3.11 SRO 3.3 Cog ni t ive-Level

Memory

References:

ON-I 13, ARC 31 3 (B-3)

Justification:

A. Correct - ON-1 13 and ARC 31 3 (B-3) require that a Reactor Recirc pump be secured from service if either seal cavity temperature exceeds 2OOOF or motor bearing temperature exceeds 194OF. Incorrect - Annunciators 313 (B-3) RECIRC PUMP MOTOR HI TEMP comes in at 15OOF seal cavity temperature and 185°F pump motor bearing temperature. These are below the temperatures required by ON-I13 or ARC 313 (B-3) to secure the pump. Incorrect - Speed is not required to be reduced per ON-I 13 until seal cavity temperature reaches 18OOF. The alarm comes in at 15OOF for seal cavity temperature.

Incorrect - ON-113 and ARC 313 (6-3) will require securing a recirc pump if either motor bearing or seal cavity temperature reach a predetermined value. B. C. D. --- NRC EXAM Page: 13 of 144 12/22/06 8 ID: N-ILT-1529-1G-001 Points: 1.00 v Unit 3 is operating at 100% power when the following occurs:

The 3A TBCCW pump trips on thermal overload due to excessive current.

The 3B TBCCW pump is successfully started and all TBCCW system parameters return to normal. Per NOM-C-5.2, "Resetting Protective Devices / Restoring Power", what is the LOWEST LEVEL of authority that must authorize a restart of the 3A TBCCW pump? A. Shift Management ONLY B. Shift Management and the Shift Operations Superintendent.

C. Shift Management and Engineering Management D. Shift Operations Superintendent and Electrical Maintenance Management Answer: A

., --.---- NRC EXAM Page: 14 of 144 ' 12/22/06 Question 8 Details Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1529-1G-001 Unit 3 is operating at 100% power when the following occurs: *The 3A TBCCW pump 1273 Active No No 3.00 2 1 .oo 295018.2.1.14 N-l LT- 1 529-1 G-00 1 0.00 0.00 Importance:

RO 2.5 I SRO 3.3 Cognitive-Level:

Memory

References:

NOM-C-5.2 Justification:

A. Correct - NOM-C-5.2, Section 2, requires Shift Management be notified and approve reclosure of a tripped circuit breaker. Incorrect - The SOS may be consulted but is not required per NOM-C-5.2.

Incorrect - Engineering Management may be consulted but they are not required per NOM- Incorrect - The SOS and Electrical Maintenance may be consulted but they are not required per NOM-C-5.2.

B. C. C-5.2. D. i_/ NRC EXAM Page: 15 of 144 12/22/06

.- 9 ID: N-ILT-50164402 Points: 1.00 Unit 2 is operating at 100% power with all Instrument Air and Instrument Nitrogen systems aligned normally when it experiences the following:

Annunciator NITROGEN COMPRESSOR A OR B TROUBLE (228 E-2) alarms

After investigation, the Equipment Operator reports:

The 'A' and 'B' Instrument Nitrogen Compressors are tripped.

The 'A' and '6' Instrument Nitrogen Receiver pressures are at 75 psig With no operator action, under these conditions pressure to the Inboard Main Steam Isolation Valves will be maintained by the: A. Nitrogen Bottles aligned by the auto opening of SV-8130 NB, "NB Supply." B. Containment Atmosphere Dilution System aligned by the auto opening of PCV- 7651 NB, "SGIG Pressure Control Valve." C. Nitrogen bottles aligned by the auto opening of PCV-7700 "Instrument Nitrogen Backup Pressure Control Valve." D. Instrument Air System aligned by the auto opening of A04230 NB, "NB Instrument Air Backup to Instrument Nitrogen." Answer: D ,Ll' NRC EXAM Page: 16 of 144 12/22/06 ANSWER KE Question 9 Details Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5016-4-002 Relationship between lnst N2 and lnst Air 1261 Active No No 2.00 3 1 .oo 295019 G2.1.28 N-ILT-5016-4-002 0.00 0.00 Importance:

RO 3.2 I SRO 3.3 Cognitive-Level:

High

Reference:

PLOT 5016, OBJ. 4; ARC 228 E-2 Justification:

A. 6. C. D. Incorrect - SV-8130 A/B valves are normally in the closed position.

They do not auto-open on low instrument N2 receiver pressure.

When open, pressure would only be aligned to the ADS valves. Incorrect - Alignment of the CAD system through SGlG system to supply the Instrument Nitrogen system requires manual valve alignments.

Incorrect - PCV-7700 is used in the manual alignment of the CAD Tank to supply nitrogen to the Drywell Instrument Nitrogen headers.

Correct - Instrument air will automatically backup the Instrument Nitrogen System when Instrument Nitrogen Receiver pressure drops below 85 psig. u NRC EXAM Page: 17 of 144 12/22/06 PP t 40 8 1 :36ed WW3 3NN 9o/zz/z 1 .-? a :JaMSUQ TI 20 Question 10 Details ' Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Com men t: Multiple Choice N-ILT-5010-3F-001

Unit 2 is in MODE 4 twenty-four hours after shutdown, following an extended full 1322 Active No No 3.00 2 1 .oo 295021AK1.01 N-l LT-50 1 0-3F-001 0.00 0.00 Importance:

RO 3.6 / SRO 3.8 Cognitive-Level:

High

References:

ON-125, GP-12 Justification:

A. Incorrect - With the RHR pump tripped there is no longer shutdown cooling flow from the reactor vessel to the RHR heat exchanger.

Correct - Decay heat will cause RPV coolant temperature to rise and eventually reach boiling. Reactor pressure will increase above atmospheric pressure (NOTE: Even if examinee assumes RPV head vents are open pressure will still increase since the head vents are on a 1" line and are designed for removal of non-condensibles at power or air removal for refueling or hydro test conditions. There is industry OE that confirms that bulk boiling of coolant due to lack of shutdown cooling will result in going greater than 212 F and pressurizing the RPV with the vents open).

Incorrect - Reactor pressure will increase above atmospheric.

Incorrect -With the RHR pump tripped there is no longer shutdown cooling flow from the reactor vessel to the RHR heat exchanger.

B. C. D. 'c' NRC EXAM Page: 19 of 144 12/22/06

EXAMINATION ANSWER KEY 07 NRC RO Rev 0 Question I4 Details Question Type: Topic: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1550-27A-001 Unit 2 is shutdown for a refueling outage with the fuel pool gates installed.

1274 Active No No 3.00 3 1 .oo 295023.AA.1.02 N-ILT-1550-27A-001 0.00 0.00 Importance:

RO 2.9 / SRO 3.1 Cognitive-Level:

High

References:

ARC 20C076 (D-2), ARC 20C075 (6-I), ON-I 24, FH-74 Justification:

A. Incorrect - Fuel pool level has decreased and has resulted in the alarm. The level in the skimmer surge tanks are the same.

The second pump will not start because the level in the skimmer surge tank has fallen below the low low level in the simmer surge tank. Correct - ARC 20C076 (D-2) will refer the Equipment Operator to SO 19.3.A-2, Fuel Pool Filling Skimmer Surge Tank via Normal Make-up Line which utilizes the Condensate Transfer System Incorrect - The control rod drive pump will inject into the reactor vessel and will not effect the level in the fuel pool due to fuel pool gates being installed.

Incorrect - The fuel pool gates are installed and because of this, addition of water to the cavity will have no effect on the fuel pool level.

B. C. D. 'W NRC EXAM Page: 21 of 144 12/22/06 EXAMINATION ANSWER KEY 20 RO Rev 0 L/ 12 ID: N-ILT-1540-3-00q 5 Points: f.00 Unit 2 is operating at 100% power Drywell Pressure unexpectedly rises to 1.2 psig and is trending up. OT-101, "High Drywell Pressure" has been entered.

The operating crew should IMMEDIATELY:

A. B. C. D. perform GP-3, "Normal Plant Shutdown".

perform GP-4, "Manual Reactor Scram".

scram and enter T-101, "RPV Control".

perform GP-9, "Fast Reactor Power Reduction".

Answer: B Question 12 Details 'w Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1540-3-001 Unit 2 is operating at 100% power when Drywell pressure unexpectedly rises to 1.2 22 I N-ILT-1540-3-001 Active NO No 2.00 2 1 .oo 295024 G2.4.49 A 6286.00 0.00 Importance:

RO 4.0 / SRO 4.0 Cognitive-Level:

Memory

References:

PLOT1 540.03, OT-I 01 Justification:

A. B. C. Incorrect - Not required unless both seals on a Recirc Pump fail. Correct - A GP-4 Manual Scram is required at 1.2 psig in Drywell. Incorrect - T-101 and T-102 are not required to be entered until drywell pressure reaches 2.0 psig. Incorrect - Not required by OT-101. D. v NRC EXAM Page: 22 of 144 12/22/06 N ANSW 7 NRC RO Rev 0 .\/' 13 L ID: N-ILT-5001A-5G-001 0 0 Unit 3 had been operating for 340 days when a reactor scram occurred 15 minutes after the scram, plant conditions are as follows: * * * * *

A Group 1 isolation has occurred and has not been reset. 9 Control Rods are at position

02. RClC tripped on overspeed and cannot be restarted.

HPCl is out of service. Reactor water level is +I5 inches and has remained steady Reactor pressure is 1140 psig. Why is RPV pressure high AND, assuming no operator action, what is the status of the Safety Relieve Valves (SRVs)?

A. Multiple rods still out, AND SRVs are closed

6. Decay heat generation, AND SRVs are closed.

C. Decay heat generation, AND SRVs are controlling pressure in self-actuation mode. D. Multiple rods still out, AND SRVs are controlling pressure in self-actuation mode.

Answer: C >W NRC EXAM Page: 23 of 144 12/22/06 I. .. c-c -' -../-- EXAMINATION ANS Question 13 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2:

Comment: Multiple Choice N-ILT-5001A-5G-001 Unit 3 had been operating on a 340 day run when a reactor scram occurred.

15 min 1275 Active No No 2.50 3 1 .oo 295025. K1.04 N-l LT-500 1 A-5G-00 1 0.00 0.00 Importance:

RO 3.6 / SRO 3.9 Cognitive-Level: High

References:

Tech. Spec. 3.4.3 Justification:

A. Incorrect - Control rods at position 02 will not significantly contribute to thermal power. SRVs are open at setpoint of 11 35 psig rt 1 % with no operator action and no normal heat sinks. Incorrect - SRVs are open at setpoint of 1135 psig 5 1% with no operator action and not normal heat sink. Correct - Decay heat contributes approximately 2% of thermal power 10 minutes following a scram. With no operator action and no normal heat sink available (Group 1 Isolation)

RPV pressure will rise until SRVs self-actuate at 1135psig+I%.

Incorrect - Control rods at position 02 will not significantly contribute to core thermal power. B. C. D. \--- NRC EXAM Page: 24 of 144 12/22/06 14 ID: N-lLT5007-1 B-002 Points: 1.00 - While operating Unit 3 at 100% reactor power, which the relationship between torus water temperature, torus level, and torus pressure?

of the following statements describes A RISE in torus water temperature will result in a RISE in torus water level and a DROP in torus pressure.

A. B. A DROP in torus water temperature will result in a RISE in torus water level and a RISE in torus pressure.

C. A RISE in torus water temperature will result in a DROP in torus water level and a RISE in torus pressure.

D. A DROP in torus water temperature will result in a DROP in torus water level and a DROP in torus pressure.

Answer: D Question 14 Details L\ '--' Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value:

Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5007-1 B-002 While operating Unit 3 at 100% reactor power, which ONE of the following statement 436 Active No No 2.00 2 1 .oo 295026 K2.06 N-ILT-5007-1 B-002 0.00 0.00 Importance:

RO 3.5 / SRO 3.7 Cognitive-Level:

Memory

Reference:

PLOT 5007, Obj. 1 b; T-102 Bases A. Incorrect - rise in level will NOT cause pressure to drop. B. Incorrect - Drop in temp will NOT cause level to rise. C. Incorrect - Reverse reason as A. D. Correct. ----' NRC EXAM Page: 25 of 144 12/22/06 3 :JaMsut,

EXAMINATION ANS 2007 NRk RO Rev 0 *--/ 16 ID: N-ILT-5059K-1A-002 Points: I.~o

Unit 2 is at 30% power with GP-2, "Normal Plant Start-up", in progress. Annunciator 224 (E-5) TORUS ROOM FLOOD is lit in the main control room.

A significant water leak is identified in the Torus.

The crew enters T-102, "Primary Containment Control", at a Torus level of 14.5' and lowering.

*

If Torus level cannot be maintained above 12.5 I, per T-102 "Primary Containment Control", you are required to (1) based on (2) : A. (1) Manually scram the reactor per GP-4 (2) Torus level indicators in the Main Control Room OR from SPDS. B. (1 ) Perform an Emergency Blowdown per T-I12 (2) Torus level indicators in the Main Control Room ONLY. C. (I) Manually scram the reactor per GP-4 (2) Torus level indicators in the Main Control Room ONLY. D. (1) Perform an Emergency Blowdown per T-I12 (2) Torus level indicators in the Main Control Room OR from SPDS. Answer: C

~L' 12/22/06 NRC EXAM Page: 28 of 144 EXAMINATION ANS NRC RO Rev 0 Question 16 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5059K-1A-002 Unit 2 is at 30% power. Power ascension to 100% power is in progress per GP-2 1354 N-ILT-5059K-1 A-002 Active No No 2.50 2 1 .oo 295030K2.09 0.00 0.00 Importance:

RO 2.5 / SRO 2.8 Cognitive-Level:

Memory

References:

PLOT-5059K, Management Expectation Justification:

A. Incorrect - Station management expectation is that operations will not take actions based solely on information from PMS or SPDS systems. Incorrect - An Emergency Blowdown per T-I 12 is not required per T-102 until Torus level reaches 10.5 feet. Correct - While SPDS provides a continuous indication of plant safety system status during normal, abnormal and emergency conditions, station management expectation is that operations will not take actions based solely on information from PMS or SPDS systems. Incorrect - An Emergency Blowdown per T-I 12 is not required per T-102 until Torus level reaches 10.5 feet. Also, station management expectation is that operations will not take actions based solely on information from PMS or SPDS systems.

B. C. D. .. --' NRC EXAM Page: 29 of 144 12/22/06

.. , --- 17 ID: N-ILT-2101-1-010 I Points: I .OO The following conditions exist on Unit 2 following a small LOCA: 0 0 0 0 0 0 MSlVs are closed.

0 All control rods are fully inserted.

RPV Level is -100 inches and dropping slowly. RPV Pressure is 960 psig and steady. Drywell Pressure is 4 psig. Torus Pressure is 3 psig. HPCl and RClC are both unavailable for injection Which of the following actions should be taken?

A. B. C. Lineup and start HPSW pumps to inject per T-245. Rapidly depressurize the RPV with BPVs per step T-101 RCIP-12.

Lower RPV pressure to inject with Core Spray without exceeding the Technical Specification Cooldown limits. Lower RPV pressure to inject with Condensate without exceeding the Technical Specification Cooldown limits. D. \-- Answer: D i__--' NRC EXAM Page: 30 of 144 12/22/06 Question 17 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-2101-1-010 The following conditions exist on Unit 2 following a small LOCA:

RPV level is 1279 Active No No 4.00 4 1 .oo 295031 A2.03 N-l LT-2 1 0 1 0 1 0 0.00 0.00 Importance:

RO 4.2 / SRO 4.2 Cognitive-Level:

High

Reference:

T-101 and Bases, T-I 12 and Bases, T-102 and Bases.

Justification:

A. Incorrect - Per T-245, "HPSW Injection into the RPV' placing the HPSW pumps inservice and the majority of valve manipulations are not completed until RPV pressure is below 400 psig. Incorrect - For the conditions given, the plant is not approaching a limit that requires an Emergency Blowdown (T-112) in T-102, T-103, T-104. RC/P-12, rapidly depressurize with BPVs, is not used.

In addition, the MSlVs are closed which eliminates use of BPVs. Incorrect - The Core Spray system will not inject into the RPV until RPV pressure is lower than 330 psig. Lowering pressure to below 330 psig will be a violation of the Tech Spec 1 OO°F/hr cooldown rate. Correct - T-101 steps RC/P-16 directs beginning an RPV depressurization maintaining cool down rate below lOOoF/hr. RCIP-16 along with RCIL-3 allows for using Condensate system to restore RPV level. B. C. D. <.-' NRC EXAM Page: 31 of 144 12/22/06 EXAMINATION ANSWER KEY Unit 2 conditions are as follows: * * *

The Unit has scrammed.

Seven (7) control rods located randomly throughout the core are stuck between positions 06 and 34. None of the seven control rods moved after ARI initiation. Reactor pressure is 920 psig. Reactor water level is +20 inches (stable on the narrow range).

Drywell pressure is 1 .O psig. Drywell temperature is 130OF. Torus temperature is 85OF. T-I 01, "RPV Control", Leg RC/Q Rods was entered from T-1 00, "Scram", due to ATWS condition. In accordance with T-101, "RPV Control", which one of the following describes the condition allowing exit from T-1 01 , Leg RC/Q? A. Cold shutdown boron weight has been injected into the reactor core.

B. ALL control rods, except one, are fully inserted into the reactor core.

C. Reactor power will remain below 4% under ALL conditions without boron.

D. The reactor will remain shutdown with RHR in the Shutdown Cooling mode.

Answer: B L/, NRC EXAM Page: 32 of 144 12/22/06 N ANSWER KEY ev 0 Question 18 Details Question Type: Topic: System ID: User ID: Status: Always select on test:

Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-2101-6-002 Unit 2 has scrammed, it is determined that seven (7) control rods located randomly 1323 Active No No 2.00 3 1 .oo 295037K1.07 N-ILT-2101-6-002 0.00 0.00 Importance:

RO 3.4 / SRO 3.8 Cog n i tive-Level:

High

References:

T-I 01 Bases Justification:

A. B. Incorrect - Boron would not have been injected for these reactor/containment conditions.

Correct - The only condition allowing exit from Leg RC/Q Rods is when an ATWS is no longer in progress. Note #24 states that termination of an ATWS requires determination that: (1) All rods are inserted to or beyond the maximum subcritical banked withdrawal position (MSBWP) of "02". OR (2) With any single rod fully withdrawn past 00, all other rods are fully inserted.

OR (3) The reactor will remain shutdown under all conditions on rod insertion alone regardless of boron concentration (RE calculation).

Incorrect - 4% reactor power is the reference for entry into T-101, "RPV Control". Leg RC/Q rods makes no reference to 5 4% power as an exit requirement.

Incorrect - Termination of ATWS Note

24 makes no reference to RHR Shutdown Cooling mode of operation.

C. D. l.,,' NRC EXAM Page: 33 of 144 12/22/06 EXAMINATION ANSWER KE 2007 NRC RO Rev 0 . , ID: N-lLT-G54-004 Points: 3.00 .4 I9 Field teams have been dispatched due to a Radioactivity Release. The field teams are located as follows: Field Team 1 is at the Training Center. Field Team 2 is 50 yards NORTH of the intersection of Routes 851 (Broad Street) and Lay Road (Site Access Road). Field Team 3 is at the ISFSI pad. Field Team 4 is at the Muddy Run Pumped Storage Facility.

Which of the following describes which Field Team(s) is(are) OFF-SITE for the purposes of Emergency Classification?

A. 2 and 3 ONLY B. 2 and 4 ONLY. C. 3 and 4 ONLY D. 2 and 3 and 4 ONLY Answer: B .I .--: NRC EXAM Page: 34 of 144 12/22/06 Question l9-Details

Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-G5-4-004 Field teams have been dispatched due to a Radioactivity Release. The field teams are 1280 Active No No 3.00 3 1 .oo 295038EA2.01 N-ILT-G5-4-004 0.00 0.00 Importance: RO 3.3 I SRO 4.3 Cognitive-Level: Memory

References:

FSAR Figure 2.2.5, EP-AA-1 000 Justification:

A. B. Incorrect - The ISFSI Pad is within the Figure

2.2.5 owner

controlled site boundary.

Correct - Per EP-AA-1000, "Off-Site

is the area outside the Station's "Site Boundary".

PB FSAR Figure 2.2.5 shows the site boundary.

Muddy Run Station and Route 851 are clearly not within the owner controlled area. Incorrect - The ISFSI Pad is within the Figure

2.2.5 owner

controlled site boundary.

Incorrect - The Training Center and the ISFSI Pad are both within the owner controlled area and are not considered off-site.

C. D. >- *- NRC EXAM Page: 35 of 144 12/22/06 PP 1 40 9E 36ed WVX3 3tlN 9o/zz/z c 7-, 0 *ad N 3)1 tl3MSNV NOIlVNIVVVX3 AMI .,/ 21 ID: N-ILT-1540-4-009

' , Points: 1.00 Unit 2 is at 85% power when annunciator 210 H-2, REACTOR HI-LO WATER LEVEL alarms. The following conditions exist: * * * *

RPV level is +31 inches and rising. Total feed flow is greater than total steam flow. "A" RFP speed is 4700 rpm and rising. "B' RFP speed is 4300 rpm and lowering. "C" RFP speed is 4500 rpm and steady. Based on the above indications, the -(I)- RFP is operating correctly and the

-(2)- RFP should be taken to manual control. A. (1 ) "A" (2) "C" B. (1 ) "B' (2) "A' C. (1) "B" (2) "C" D. (1 ) "A' (2) "B' Answer: B L* NRC EXAM Page: 37 of 144 12/22/06 Qy 3tlN LOO A34 Y3MSNV NOIlVNIWVX3 Points: 1.06' \.,' 22 1D: N-ILT-1540-5-004 Unit 2 was at 100% power when an unidentified leak into the primary containment caused an automatic reactor scram. The following conditions are present on Unit 2: * * *

All rods are inserted. RPV level is -5 inches and rising slowly. RPV pressure is 940 psig and dropping.

House Loads have been transferred.

Based on the above conditions, reactor recirculation pump speed is presently

-( 1 )- due to -w-- A. (1) 30% (2) a scram signal being present with RPV level less than +I 7 inches. B. (I) 30% (2) individual reactor feedpump flows less than 20% with RPV level less than +I7 inches. C. (1) 45% (2) a scram signal being present with RPV level less than +I7 inches. D. (1)45% (2) individual reactor feedpump flows less than 20% with RPV level less than +I7 inches. Answer: A .-,' NRC EXAM Page: 39 of 144 12/22/06 EXAMINATION ANSWER KE -.---. Question 22 betails ' Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-1540-5-004 Unit 2 was at 100% power when an unidentified leak into the primary containment 1284 Active No No 2.50 2 1 .oo 295009AK2.03 N-ILT- 1 540-5-004 0.00 0.00 Importance:

RO 3.1 I SRO 3.2 Cognitive-Level: High

References:

OT-I 00 Justification:

A. Correct B. C. D. Incorrect - Condition is a 45% runback. Incorrect - Condition is a 30% runback. Incorrect - Condition is a 45% runback. NRC EXAM Page: 40 of 144 12/22/06 S 0 u 23 ID: N-ILT-5007-8-006 Points: 1.00 The following Unit 3 conditions exist: * *

The reactor is at full power. Torus Cooling is in operation using the 3A and 3C RHR pumps. HPCI testing is in progress per ST-0-023-301-3, "HPCI Pump, Valve, Flow and Unit Cooler Functional and In-Service Test".

Entry into T-102 "Primary Containment Control" is required when Torus temperature exceeds - /I) , and Technical Specifications require immediately placing the mode switch in shutdown if Torus temperature exceeds (2) . A. (1 ) 95OF (2) 100aF B. (1) 95OF (2) 1 10°F C. (1) 105OF (2) llO°F D. (1) 110OF (2) 12OOF Answer: B '*/ NRC EXAM Page: 41 of 144 12/22/06 EXAMINATION ANSWER K Rev 0 Question 23 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number I: User Number 2: Comment: Multiple Choice N-ILT-5007-8-006 The following Unit 3 conditions exist:

The reactor is at full power.

Torus Coo 1285 Active No No 2.50 2 I .oo 29501 3AKL2.01 N-l LT-5007-8-006 0.00 0.00 Importance:

RO 3.8 / SRO 4.0 Cognitive-Level:

Memory

References:

ST-0-023-301

-3 Justification:

A. Incorrect - While 95OF is an entry for T-102, Tech Spec requires reactor mode switch to shutdown position if suppression pool temperature

> 11 O°F and 5 120OF. B. Correct - 95OF is the entry for T-102. T.S.

3.6.2.1 requires immediate suspension of all testing that adds heat to the suppression pool at pool temperature of > 105OF AND to immediately place the reactor mode switch in the shutdown position if suppression pool temperature

> 11 O°F and I 120OF. C. Incorrect - 95OF is the entry for T-102. Tech Spec requires reactor mode switch to shutdown position if suppression pool temperature

> 1 1 O°F and 5 120OF. D. Incorrect - 95OF is the entry for T-102. Tech Spec requires immediate suspension of all testing that adds heat to the suppression pool at pool temperature of > 7 05°FAND to immediately place the reactor mode switch in the shutdown position if suppression pool temperature

> IlOOF and 5 120OF. NRC EXAM Page: 42 of 144 12/22/06 TI 20 L- 24 ID: N-ILT-I 550-1 2A-001 Points: 1.00

Unit 3 is at 100% power at time 1125 am the '3A Control Rod Drive (CRD) pump trips on overcu rren

t. The "3B' CRD pump was previously blocked for maintenance.

At time 1133 am, multiple accumulator trouble lights illuminate on the Full Core Display for withdrawn control rods. At time 1137 am, CRD Charging Header pressure drops to below 940 psig. * *

Based on the above conditions, you are required to perform a -(I)- at time

-(2)- in accordance with ON-I 07, "Loss of CRD Regulating Function".

A. (1) Manual Scram, enter T-100 (2) 1153 am B. (1 ) Manual Scram, enter T-I 00 (2) 1157 am C. (1) GP-9 Fast Power Reduction (2) 1153 am D. (I) GP-9 Fast Power Reduction (2) 1157 am Answer: B ---- . NRC EXAM Page: 43 of 144 12/22/06 S 0 Question 24 Details Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number I: User Number 2: Comment: Multiple Choice N-ILT-1550-12A-001 Unit 3 is at 100% power at time 1125 am the '3A' Control Rod Drive (CRD) pump 1286 Active No No 2.50 3 1.00 295022 G2.1.23 N-ILT-1550-12A-001 0.00 0.00 Importance:

RO 3.9 / SRO 4.0 Cog n itive-Level: High

References:

ON-107, Tech. Spec. 3.1.5 Justification:

A. Incorrect - This is 20 minutes from accumulator alarm only, scram not yet required per ON-I07 or Tech Specs 3.1.5.

Correct - 20 minutes to restore charging header pressure once the condition of both multiple accumulator trouble alarms and low (< 940 psig) CRD charging header pressure. This agrees with Tech Spec 3.1 5. Incorrect - This is 20 minutes from accumulator alarm only fast power reduction not required per ON-I07 or Tech Spec 3.1.5.

Incorrect - ON-107 requires a scram due to both conditions (accumulator trouble alarm and low charging header pressure).

6. C. D. -. \-- NRC EXAM Page: 44 of 144 12/22/06

-. ,. --. do G61 6!sd g Pa4 GI 'V

'u' ON AN 07 NRC RO Rev Question 25 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1560-11-007 Which of the following sets of conditions allows operation of the "A' loop of RHR 1287 Active No No 2.00 3 1 .oo 295029EA1.03 N-ILT-1560-11-007 0.00 0.00 Importance: RO 2.9 / SRO 3.0 Cognitive-Level:

High

References:

T-102, Sheet 3 Justification:

A. Incorrect - This does not meet the criteria of the RHR NPSH curves for two-pump operation at all flow rates, as shown on T-102, Sheet

3. Operation is in the unsafe region of the curve when flow is above -23,000 gpm. Incorrect - This does not meet the criteria of the RHR NPSH curves for two-pump operation at all flow rates, as shown on T-102, Sheet

3. Operation is in the unsafe region of the curve when flow is above -23,000 gpm.

Incorrect - This does not meet the criteria of the RHR NPSH curves for two-pump operation at all flow rates, as shown on T-102, Sheet

3. Operation is in the unsafe region of the curve when flow is above -23,000 gpm. Correct - This meet the criteria of the RHR NPSH curves for two-pump operation at all flow &, as shown on T-102, Sheet

3. B. C. D. L-' NRC EXAM Page: 46 of 144 12/22/06 A L-' 26 2 ID: N-ILT-1560-3403 ,y Points: I -00" Unit 3 plant conditions are as follows: * *

Reactor is shutdown RPV level is -30 inches RPV pressure is 950 psig HIGH AREA TEMP alarm is up (window J-3 on panel 310 / 30C205L) RClC room temperature 13OOF due to a steam leak Reactor Building and Refuel Floor radiation levels are 2 mRlhr * *

For the above conditions which of the following statements are CORRECT? 1. The RClC Room should be evacuated per GP-15, "Local Evacuation".

2. T-I 12, "Emergency Blowdown" procedure must be performed if Torus Room temperature exceeds 125OF. 3. T-I 12, "Emergency Blowdown" procedure must be performed if RClC Room radiation levels exceeds 8 Rlhr. 4. Reactor Building Ventilation may be restored using T-222-2, "Secondary Containment Ventilation Bypass". A. 1 &4 6. 1&3 C. 3&4 D. 2&4 Answer: A \-/ NRC EXAM Page: 47 of 144 12/22/06 LO' 1)13Z&O96Z 00' 1 a :JaMSUQ P'8Z '3 1'8 1 'V i? 3tlN LOO NO1 X EX 'Question 27 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1560-3-004 While Unit 2 is operating at 100% power, the following conditions exist:

HPCl 1289 Active No No 3.00 3 1 .oo 295036EK3.03 N-ILT-1560-3-004 0.00 0.00 Importance: RO 3.8 I SRO 3.9 Cognitive-Level: High

References:

T-103 Justification:

A. B. C. D. Incorrect - T-I03 requires a T-I 12 Blowdown (SCC-IO) only if a primary system is discharging into the reactor building (SCC-7) and the water level reaches an action level in more than one area. HPCl and RClC are considered the same area per T-I 03. A leak from the Condensate Storage tank (CST) is NOT a primary system leak. Incorrect - T-103 requires a T-I12 Blowdown (SCC-IO) only if a primary system is discharging into the reactor building (SCC-7) and the water level reaches an action level in more than one area. HPCl and RClC are considered the same area per T-103.

Incorrect - T-I 03 requires a GP-4 scram and a depressurization (SCC-8) only if the leak is a primary system discharging into the Reactor Building (SCC-7).

A leak from the Condensate Storage tank (CST) is NOT a primary system leak. Correct - Per T-103, "Secondary Containment Control", water level above an alarm level (6" for HPCI/RCIC/Sump Rooms), GP-15, "Local Evacuation" (SC/L-5) and reference to SE-9, "Radioactive Spill" (SCIL-6) are required to be performed.

.-.---' NRC EXAM Page: 50 of 144 12/22/06 EXAMINATION ANSWER KEY 200 RO Rev 0 -1 28 ID: M-ILT-5010-6B-003

' Points:'d .OO 0 0 Following 400 days at rated power, Unit 2 is Shutdown with a cooldown in progress. Reactor Pressure is currently 420 psig. An electrical transient occurs resulting in the following:

Loss of 125 VDC power to the A logic of RHR. Loss of Drywell Cooling and a small steam leak cause a rise in Drywell Pressure to 2.2 psig. Which of the following describes the response of the RHR pumps? A. All RHR pumps are running, and they are injecting into the vessel.

B. All RHR pumps are running, and they are NOT injecting into the vessel. C. Only the B & D pumps are running, and they are injecting into the vessel D. Only the B & D pumps are running, and they are NOT injecting into the vessel. Answer: B %--. NRC EXAM Page: 51 of 144 12/22/06 Question 28 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5010-6B-003 Following a 400 day run at power, Unit 2 is Shutdown with a Cooldown in progress.

1151 Active No No 3.00 3 1.00 20300K2.03 N-ILT-5010-68-003 0.00 0.00 Importance:

RO 2.7 I SRO 2.9 Cognitive-Level:

High

References:

SO 10.7. B-2 Justification:

A. B. Incorrect - Reactor pressure is too high or LPCl injection.

Correct - RHR logics are cross-divisionalized such that a loss of one 125 VDC supply does not impact LPCl pump starts (unlike Core Spray). Per TRIPS, RHR pump shutoff head is 305 psig so they are not injecting.

Incorrect - Even with loss of a logic 125 VDC, all LPCl pumps are running. Incorrect - Even with loss of A logic 125 VDC all LPCl pumps are running.

C. D. 'U' NRC EXAM Page: 52 of 144 12/22/06 ON ANS 007 NRC RO Rev 0 L' 29 ID: N-lLT5010-40-003 Points: I'.OO 0 0 0 0 Unit 2 is in a forced outage The 'A Loop of SDC is in service using the 'C' RHR pump.

RPV level inadvertently lowers to -3 inches. Reactor pressure is 25 psig and stable. How will the RHR system respond to this transient?

A. The MO-25A ONLY (A Loop RHR injection valve) will close.

B. The MO-17 and MO-18 ONLY (RHR suction to recirculation loop isolation valves) will close.

C. The MO-17 and MO-18 (RHR suction to recirculation loop isolation valves) will remain open and the

'C' RHR pump will continue run. D. The MO-17 and MO-18 (RHR suction to recirculation loop isolation valves) and the MO-25A (A loop RHR injection valve) will close.

The 'C' RHR pump will trip.

Answer: D -u- NRC EXAM Page: 53 of 144 12/22/06 Question 29 Details L Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5010-40-003 Unit 2 is in a forced outage with the 'A Loop of SDC in service using the 'C' RHR 1158 Active No No 3.00 2 1 .oo 205000 K4.03 N-ILTdOl O-40-003 0.00 0.00 Importance: RO 3.8 /SRO 3.8 Cognitive-Level:

High

Reference:

PLOT5010.040; GP-8.B; GP-8.B COL Justification:

A. Incorrect - MO-25A (RHR injector valve) will also close on PClS Group I1 signal of 5 +I" RPV level with MO-17 open and MO-18 open and RPV pressure 5 70 psig. Incorrect - MO-I 7 and MO-I 8 will close on PClS Group II isolation signal of I +I" RPV level. Incorrect - MO-17 and MO-18 will close on PClS Group I1 isolation signal of 1. +I" RPV level. The C RHR pump will trip when either MO-17 or MO-18 indicate not full open.

Correct - MO-17 & 18 will close on the PClS Group I1 isolation signal of 5 +I" RPV level with MO-17 open and MO-18 open and RPV pressure 5 70 psig. B. C. D. .L-- NRC EXAM Page: 54 of 144 12/22/06 u' 30 _I ID: N-ILT-1530-3-004 Points: 1.00 Unit 3 is in MODE 3 with RPV coolant temperature at 280OF. Per procedure GP-12, Core Cooling, the operator must either: * *

Operate one RHR pump in shutdown cooling OR Operate one recirc pump OR Maintain reactor level above +50 inches What is the reason for maintaining water level above +50 inches? A. Provides for adequate level to prevent uncovering the core if a reactor coolant leak develops.

B. Provides for natural circulation between the core and the annulus region since no forced cooling flow exists. C. Provides for a sufficient volume of water to ensure core cooling via conductive heat transfer.

D. Provides for sufficient RPV water level to satisfy Technical Specification 3.9.6 "RPV Water Level" requirements.

Answer: B ---- NRC EXAM Page: 55 of 144 12/22/06

J 31 ' ID: N-ILT-5023-4G-003 Points: f.00' * *

The Unit 3 HPCl Turbine isolated from a false steam supply low pressure signal. I&C Technicians corrected the problem and the isolation signal is clear. Moments prior to resetting the HPCl isolation a small steam leak develops in the Primary Containment which brings Drywell pressure up to 3 psig. RPV pressure is 920 psig.

Which of the following statements describes the correct response of the HPCl system for the above conditions?

A. The HPCl isolation will automatically reset, then the steam supply valves (MO-15 and MO-16) must be manually opened.

B. The HPCl isolation will automatically reset, then the steam supply valves will automatically go open. C. HPCl will start only when pushbuttons 23A-S20 (AUTOIMANUAL RESET) and 23A-S26 (AUTO RESET) are depressed.

D. HPCl will start only when pushbutton 23A-S20 (AUTO/MANUAL RESET) and 23A-S25 (AUTO RESET) are depressed AND the steam supply valves MO-15 and MO-16 are manually re-opened.

Answer: B '-J NRC EXAM Page: 57 of 144 12/22/06

'u' Question 32 Details . Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5014-6C-001 Unit 3 was operating at 100% power when a plant transient resulted in the followin I290 Active No No 3.00 3 1 .oo 209001A1.05 N-l LT-50 14-6C-00 1 0.00 0.00 Importance:

RO 3.5 / SRO 3.6 Cognitive-Level:

High

References:

T-I 01, T-I 02, T-I 1 1 Justification:

A. Incorrect -Although the NPSH limit has been exceeded, the "A" Core Spray pump can be placed in service as directed by Step LR-7 of Incorrect - The Vortex limit (10.5 feet in the torus) has not been exceeded.

Correct - The given conditions result in entry into T-I 1 1, "Level Restoration". Step LR-7 of T-I 11 allows operation of Core Spray pump "A" even if the NPSH and/or Vortex limit(s) have been exceeded.

Incorrect -The NPSH limit has been exceeded.

T-Ill. B. C. D. L-. NRC EXAM Page: 60 of 144 12/22/06

/--- 3 :JaMS Ut/

N ANSWER KE NRC RO Rev 0 Question 33 Details Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete: Point Value:

Cross

Reference:

User Text:

User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-5011-41-002 Select the purpose of the Standby Liquid Control (SLC) Pump Discharge 1160 Active No No 2.00 2 1 .oo 21 1000 K5.05 N-l LT-50 1 1-41-002 0.00 0.00 Importance:

RO 2.5 I SRO 2.5 Cognitive-Level:

Memory

Reference:

PLOT501 1; PB FSAR Section 3.8.3 Justification:

A. Incorrect - The SLC tank heater is what ensures the sodium pentaborate remains in solution, although no longer a concern with enriched solution.

Incorrect - The accumulators are small volume accumulators with only several hundred psig of pressure.

This is not enough to ensure RPV injection.

Correct - Accumulators on positive displacement pumps provide for pulsation dampening, not for the other reasons listed.

Incorrect - The SLC system relief valves are set to relief at 1400 psig which is significantly higher than system normal operating pressure.

B. C. D. NRC EXAM Page: 62 of 144 12/22/06 ION ANS 2007 NRC RO Rev 0 u 34 ID: N-lLT-5060F-2B-bO2 Points: l.bo Unit 2 conditions:

* *

Reactor Power is 100%. Both RPS Busses were aligned to their normal RPS MG Set power supplies.

A loss of one off-site startup feed occurred causing a 4kV Emergency Bus Fast Transfer The transfer occurred as designed and restored power to the impacted Emergency Busses from the other startup feed. Assuming normal system response, which one of the following identifies the amount of time that power will be lost to the affected RPS MG Set and the RPS logic impact. Loss of Power to MG RPS Status A. 0.25 B. 3.25 C. 8.0 D. 13.0 Answer: B Effected RPS logic trips No RPS logic trip Effected RPS logic trips No RPS logic trip L-1 NRC EXAM Page: 63 of 144 12/22/06

.I --- 1.

EXAMINATION ANSWER KEY 2007 NRC RO Rev 0 Question 34 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1

User Number 2:

Comment: Multiple Choice N-ILT-5060F-2B-002 Unit 2 conditions:

Reactor power is 100%.

Both RPS Busses were aligned to 1162 Active No No 3.50 2 1 .oo 212000 K2.01 A 0.00 0.00 Importance:

RO 3.2 I SRO 3.3 Cognitive-Level: Memory N-l LT-5060F-2B-002

Reference:

PLOT 5060F; SO 54.7.A A. incorrect - 4 kV bus will transfer in 0.25 seconds, but there is another 3 seconds until the Emergency Bus MCC is reenergized.

B. Correct - The 4KV bus will fast transfer in 0.25 seconds, then 3 seconds later the Emergency Bus MCC will reclose providing 480 VAC power back to the RPS MG Set. No RPS half scram will occur.

C. Incorrect - 8 seconds corresponds to the time delay before the RPS MG Set Supply Breakers trip on a loss of power. This would result in a half scram.

D. Incorrect - 13 seconds corresponds to 10 seconds for the EDG to start and 3 additional seconds for emergency Bus MCC to reenergize.

The 8 second time delay trip is designed to cause a loss of RPS (and scram) rather than allowing the system to continue operation after a complete loss of off-site power. This would result in a half scram.

.. , *--- NRC EXAM Page: 64 of 144 12/22/06 S 0 .I f ID: N-ILT-5060C4A-006 Points: 1.00 v 35 * *

A Unit 2 reactor startup is in progress The required critical data documentation has just been completed Two identical failures cause the "B' and'"E' Wide Range Neutron Monitoring (WRNM) channels to fail inop simultaneously.

Which one of the following is the expected system response?

A. Alarm, rod block, AND full scram B. Alarm, rod block, AND half scram. C. Alarm ONLY. No rod blocks or scram signals. D. Alarm and rod block ONLY. NO scram signals.

Answer: A .~ --- Page: 65 of 144 12/22/06 NRC EXAM WER KE Question 35 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-5060C-4A-006 A Unit 2 Reactor startup is in progress

The required critical data documentatio 1164 Active No No 2.50 N-ILT-5060C-4A-006 2 1 .oo 215003 K1.O1 A 0.00 0.00 Importance:

RO 3.9 I SRO 3.9 Cognitive-Level: High

Reference:

PLOT 5060C; ARCS 21 1 B-I and 21 1 C-I Justification:

A. Correct - INOP failure is a "trip" signal. One in each trip system will generate a full scram, the "HighllNOP' annunciator, and a control rod block. Incorrect - A full scram will result from a trip signal in each channel. Incorrect - INOP failure is a "trip" signal. One in each trip system will generate a full scram, the "HighllNOP' annunciator, and a control rod block. Incorrect - INOP failure is a "trip" signal. One in each trip system will generate a full scram, the "HighllNOP' annunciator, and a control rod block. B. C. D. L/ 12/22/06 NRC EXAM Page: 66 of 144 EXAMINATION ANS O Rev 0 W 36 ID: N-ILT-5060C4A-005 Points: 1.00 *

A Unit 3 reactor startup and approach to critical is in progress. During a rod withdrawal from position '20' to '22', a high notch worth causes alarm 31 0 F- 3, WRNM SHORT PERIODITROUBLE, as detected by WRNM Channel G.

The URO confirms a period of 25 seconds.

Which one of the following is the correct system response, and the required operator action in accordance with procedure GP-2 "Normal Plant Startup"?

A. Acknowledge the alarm, INSERT the control rod to the full-in position.

B. Acknowledge the alarm, NO additional operator action required for this condition.

C. Acknowledge the alarm and rod withdraw block, INSERT the control rod to lengthen period to > 50 seconds. D. Acknowledge the alarm and rod withdraw block, INSERT the control rod to lengthen period to infinity.

Answer: C %, ...--, NRC EXAM Page: 67 of 144 12/22/06 EXAMINATION ANS 2007 NRC RO Rev 0 Question 36 Details Question Type: Topic: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5060C-4A-005 A Unit 3 reactor startup and approach to critical is in progress. During a rod 1154 Active No No 3.50 N-l LT-5060C4A-005 2 1 .oo 21 5003 2.4.50 0.00 0.00 importance:

RO 3.3 I SRO 3.3 Cog nitive-Level:

High

References:

ARC 2(3)10-F-3, GP-2 Justification:

A. Incorrect - A reactor period of 5 28 seconds will initiate a control rod withdraw block. Procedure GP-2, startup, provides guidance to manipulate control rods to ensure that the reactor period does not go less than 50 seconds. Incorrect - A reactor period of I 28 seconds will initiate a control rod withdraw block. Procedure GP-2, startup, provides guidance to manipulate control rods to ensure that the reactor period does not go less than 50 seconds. Correct - A reactor period of 5 28 seconds will initiate a control rod withdraw block. Procedure GP-2, startup, provides guidance to manipulate control rods to ensure that the reactor period does not go less than 50 seconds. Incorrect - A reactor period of 5 28 seconds will initiate a control rod withdraw block. Procedure GP-2, startup, provides guidance to manipulate control rods to ensure that the reactor period does not go less than 50 seconds. B. C. D. u NRC EXAM Page: 68 of 144 12/22/06 37 ID: N-lLT-5060-3A-005 Points: 1.00 L 0 0 Unit 2 is operating at 25% power.

2 APRM fails downscale (not INOP). Which of the following describes the expected response?

Receive Downscale:

A. Alarm ONLY. B. C. D. Alarm, Rod Block, AND Half scram. Alarm, Rod Block, AND Full scram. Alarm AND Rod Block; NO scram signals Answer: D --= Le' NRC EXAM Page: 69 of 144 12/22/06 b- Question 37 Details Question Type: Topic: System ID: User ID: Status: Always select OR test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5060-3A-005 With Unit 2 operating at 25% power, the #2 APRM fails donwscale (not INOP). Which 1166 N-ILT-5060-3A-005 Active NO No 3.00 2 1 .oo 21 5005 K3.03 A 0.00 0.00 Importance:

RO 3.3 / SRO 4.0 Cog ni tive-Level:

High

Reference:

PLOT 5060, ARC 21 'I C-2 Justification:

A. Incorrect - APRM downscale (5 3.2 %) in MODE 1 will generate a control rod withdraw block and downscale alarm 21 1 C-2 only. Incorrect - A scram vote signal is only generated for : B. APRM hop Trip High Neutron Flux Simulated Thermal Power High C. Incorrect -A scram vote signal is only generated for : APRM hop Trip High Neutron Flux Simulated Thermal Power High Correct - AP RM downscale (5 3.2 %) in MODE 1 will generate a control rod withdraw block and downscale alarm 21 1 C-2 only. 0. NRC EXAM Page: 70 of 144 12/22/06

EXAMINATION ANSWER KEY ..---a' Question 38 Details Question Type: Topic: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5002-1 W-001 The reactor recirculation flow comparators have a 10% mismatch alarm, APRM FLOW 1291 Active No No 3.00 2 1 .oo 215005A3.06 N-ILT-5002-1 W-001 0.00 0.00 Importance:

RO 3.0 / SRO 3.1 Cognitive-Level:

Memory

References:

ARC 21 1 A-4 Justification:

A. Incorrect - A Recirc loop and B Recirc loop flows are NOT compared to each other.

The flow corporation 10% alarm is based on the difference between any of the four APRM total drive flow values. Incorrect - There is no average recirc loop flow signal. Correct - Recirc loop flow comparator alarm setpoint is based on > 10% difference between any of the four APRM total drive flow values.

Incorrect - There is no comparator circuit between total recirc flow and the recirc MG Set speed demand. B. C. D. NRC EXAM Page: 72 of 144 12/22/06 9 :JaMSUv ION ANS 007 NRC RO Rev 0 Question 39 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete: Point Value:

Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-5013-IC-005 Both RClC and HPCl initiated in resposne to a valid Unit 3 low-low Reactor water 1293 Active No No 3.50 2 1 .oo 21 7000K1.03 A 0.00 0.00 Importance: RO 3.6 / SRO 2.6 Cognitive-Level:

High N-ILT-5013-1 C-005

References:

ARC 221 C-4 Justification:

A. Incorrect - RClC suction pressure will not be affected by MO-24 closure.

No suction valves wBJ reposition.

B. Correct - On high Torus level L 15' 6' HPCl suction from CST closes and Torus suction valves open. This swap also causes MO-24 return to CST to auto close thereby removing the RClC system flow path lock to CST. RClC flow controller will attempt to maintain flow at 600 gpm and increase turbine speed (trips at 125% rated speed).

C. Incorrect - RClC will not remain in CST-to-CST mode. System will trip on mechanical overspeed as flow controller will increase speed to maintain system flow as MO-24 closes. D. Incorrect - RClC Torus suction valves do not have an auto open function.

Realigning RClC suction to Torus must be done manually.

., ---- NRC EXAM Page: 74 of 144 12122lQ6

&--,' 40 ID: N-ILTSO13-58-001 Points:'? .OO The following conditions exist on Unit 2: * * *

A spurious PClS Group I isolation has occurred. RPV level lowered to -40 inches and was restored with RCIC.

The PRO placed RClC in the CST-to-CST mode per RRC 13.1-2, "RCIC System Operation During a Plant Event".

RPV level slowly lowers and subsequent vessel injection is required. RPV level is presently steady at +30 inches.

Which of the following actions are required by the PRO in order to re-inject RClC to maintain RPV level per RRC 13.1-2, "RCIC System Operation During a Plant Event"?

A. Throttle close MO-2-23-24, Condensate Tank Return, until RClC system discharge pressure is at least 100 psig greater than reactor pressure AND AO 13-22, Discharge Check, indicates open. B. Increase RClC turbine speed by adjusting the RClC flow controller to maintain RClC system discharge pressure at least I00 psig greater than reactor pressure AND AO-2-13-22, Discharge Check, indicates open.

C. Throttle close MO-2-13-30, Full Flow Test, until RClC system discharge pressure is at least 100 psig greater than reactor pressure AND AO-2-13-22, Discharge Check, indicated open.

D. Throttle open MO-2-13-30, Full Flow Test, until RClC system discharge pressure is at least 50 psig greater than reactor pressure AND AO-2-13-22, Discharge Check, indicates split indication.

Answer: C .I L-' NRC EXAM Page: 75 of 144 12/22/06

..---. 0 MSNV NOIlVNIVVV,X3 I

TI 20 Question 42 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross Reference

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5007G-3Q-001 Given the following: *Unit 2 is operating normally at 100% power. *Reactor Buil 1295 Active No No 3.00 2 1 .oo 223002K3.18 N-ILT-5007G-3Q-001 0.00 0.00 Importance:

RO 3.0 I SRO 3.1 Cognitive-Level:

High

References:

GP-8.D, M-I -S-23, E-277, E-278 Justification:

A. B. Incorrect - This choice indicates RBV remains in service. Incorrect - This choice indicates RBV trips, SGTS starts and maintains the same Reactor Building dlp. Correct - A loss of RPS Bus "B" results in a loss of power to PClS logic channel "B' and (among other things) a PClS Group Ill outboard half isolation. This causes Reactor Building Ventilation (RBV) to trip and isolate, and SGTS to auto-start. Slnce SGTS is required by design (and Tech Spec 3.6.4.1) to maintain Reactor Building d/p 2 0.25 inches of vacuum WG, Reactor Building dlp will become more negative when SGTS starts.

Incorrect - This choice indicates RBV trips and SGTS either does not start or maintains a less negative Reactor Building dlp. C. D. bL'- NRC EXAM Page: 80 of 144 12/22/06 Q :JaMSUQ

' ' EXAMINATION Question 43 Details " Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-5007G-6-001 Which one of the following describes the purpose and function of the Primary Cont 1296 Active No No 2.50 2 1 .oo 22300262.1.27 N-ILT-5007G-6-001 0.00 0.00 Importance:

RO 2.8 I SRO 2.9 Cognitive-Level:

High

References:

UFSAR 7.3 Justification:

A. Correct - PClS limits release of radioactive materials to the environment; it does not prevent it. PClS is normally energized

... it de- energizes to function.

A single failure would not prevent PClS from performing its intended function but may cause system actuation (Le., inadvertent isolation).

Incorrect - PClS does not prevent release of radioactive materials to the environment.

PCIS is normally energized.

Incorrect - PClS is normally energized.

A single failure may cause system actuation.

Incorrect - PClS does not prevent release of radioactive materials to the environment.

A single failure may cause system actuation.

B. C. D. NRC EXAM Page: 82 of 144 12/22/06

Question 44 Details Question Type: Topic: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5001A-4D-003 During a high reactor pressure transient on Unit 2, the Plant Reactor Operator 1155 Active No No 3.00 2 1 .oo 239002K5.0 1 N-ILT-5001A-4D-003 0.00 0.00 Importance:

RO 3.4 / SRO 3.5 Cognitive-Level:

High

References:

Tech.

Spec. 3.4.3 Justification:

A. incorrect - if 1 135 psig was the peak pressure only 4 SRV's would have the white memory lights lit.

Correct - SRV setpoints range form 1 135 psig to 1155 psig.

If all 11 white memory light are lit, then pressure reached 11 55 psig. With only the "C" & "D' SRVs still open, pressure is at lowest range value of 1135 psig.

incorrect - 1260 psig is the setpoint for safety valve (not SRV) actuation.

Incorrect - 1325 psig is the reactor coolant system pressure safety limit. B. C. D. ., .,.-?A' NRC EXAM Page: 84 of 144 12/22/06

EXAMINATION ANSWER KEY 4 Question 45 Details Question Type: Topic: System ID: User ID: Status: Always select on test:

Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number I: User Number 2: Comment: Multiple Choice N-ILT-5006-3J-001 Given the following: *Unit 3 is operating at 100% power.

A "Feedwater Controll 1297 Active No No 3.00 2 1 .oo 259002K3.04 N-ILT-5006-3J-001 0.00 0.00 Importance:

RO 2.9 I SRO 3.0 Cognitive-Level:

High

References:

UFSAR 14.5.2.2, UFSAR Figure 14.5.5 Justification:

A. Correct - This transient, as analyzed in Section 14.5.2.2 of the UFSAR results in a peak reactor pressure of -1250 psig (at the bottom of the vessel). An ATWS-RPT will occur at 1 106 psig ... approximately 13 seconds into the event. Incorrect - Recirc pumps trip on high reactor pressure.

Incorrect - Recirc pumps trip on high reactor pressure.

B. C. D. NRC EXAM Page: 86 of 144 12/22/06 KE */' . ID: N-ILT-5009A-3A-005 Points: 1.00 46 Unit 3 is operating at 100% power when a valid Group Ill PClS signal is generated. The following conditions exist:

*

Both SBGT filter trains are aligned. SBGT system flow is less than expected. Secondary Containment to atmosphere differential pressure is LESS negative than expected. Which of the following is the cause of this condition?

A. A Refuel Floor blowout panel is open. B. A large steam leak has occurred in Secondary Containment.

C. SBGT Fan Bypass Damper (PO-00522) fails to reposition as designed D. SBGT "Bq Fan Vortex Damper (PO-00528) fails to reposition as designed. Answer: C NRC EXAM Page: 87 of 144 12/22/06 K 'Question 46 Detaifs Question Topic: Type: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5009A-3A-005 Unit 3 is operating at 100% power when a valid Group 111 PClS signal is 1172 Active No No 3.00 2 1 .oo 261000A3.03 A 0.00 0.00 Importance:

RO 3.2 / SRO 3.3 Cognitive-Level:

High N - I LT-5 0 09A-3A-0 0 5

Reference:

PLOT 5009A Justification:

A. B. C. Incorrect - results in high SBGT flow with a less negative DP. Incorrect - results in normal SBGT flow with a less negative or possible positive DP. Correct - Bypass damper PO-00522 provides for minimum flow recirculation path of approx - 20% for capacity back to suction plenum.

It needs to reposition to ensure RB + Refuel Floor can be maintained at a negative pressure.

Incorrect - results in high SBGT flow with more negative DP.

D. \--' NRC EXAM Page: 88 of 144 12/22/06 3M U3MSNV NOIlVNIVVVX3:

Question 47 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5051-6-001 Unit 2 and Unit 3 are at 100% power. Using a single line diagram of the North and S I344 N-l LT-5051-6-001 Active No No 3.50 2 1 .oo 262001 K4.04 0.00 0.00 Importance:

RO 2.8 I SRO 3.1 Cognitive-Level:

High

References:

Print E-I, Station Single Line EXAMINEE MUST HAVE A COPY OF PRINT E-I IN ORDER TO ANSWER THIS QUESTION.

Justification:

A. Incorrect - 65 Breaker is a Unit output breaker and does not have a reclosure feature.

Also, for a fault on the 5010 Line the 55 Breaker would trip open as well.

B. Correct - Breakers 55 and 65 will open on the fault and only the 55 Breaker will attempt reclosure.

65 Breaker is a Unit output breaker and does not have a reclosure feature. C. Incorrect - The Unit 3 Main Generator will not lockout. Output Breaker 15 will remain closed.

D. Incorrect - Breakers 55 and 65 will open on the fault. The 55 and 65 Breaker motor operated disconnects are manually operated ONLY. -L-~ Page: 90 of 144 12/22/06 NRC EXAM ION 007 NRC 48 ID: N-ILT-5058-5C-003 Points: 1.00 *

The 20Y050 supply from the Static Inverter is in a normal lineup. A fault occurs on the 20Y050 Panel resulting in an excessive current condition

(> 300 amp setpoint).

Which one of the following statements is the expected response of the Static Inverter and the 20Y050 Panel?

The Static Inverter:

A. de-energizes when the Input Breaker (CBI) trips on overcurrent and the 20Y050 Panel de-energizes.

B. receives a shutdown signal that opens both breakers (CBI and CB2) and the 20Y050 Panel de-energizes.

C. Static Switch swaps to the Alternate Source and maintains 20Y050 energized while the fault clears.

D. Static Switch is prevented from transferring to the Alternate Source and maintains 20Y050 energized while the fault clears.

Answer: C c-. NRC EXAM Page: 91 of 144 12/22/06 Question 48 Details Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value:

Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5058-5C-003

The 20Y050 supply from the Static Inverter is in a normal lineup.

A fault occ 1156 Active No No 3.50 3 1 .oo 262002K4.01 N-ILT-5058-5C-003 0.00 0.00 Importance:

RO 3.11 SRO 3.4 Cognitive-Level:

High

References:

ARC-220 F-5 Justification:

A. Incorrect - The Static Switch will transfer to alternate source in order to maintain 20Y050 panel energized.

Incorrect - The Static Switch will transfer to alternate source in order to maintain 20Y050 panel energized.

Correct - The Static Inverter is current limited.

If a fault develops it will automatically transfer to the Alternate Source which can supply the larger current necessary to clear the fault and then transfer back to normal DC supply when fault clears. Incorrect - The Static Switch will transfer to the alternate source in order to maintain 20Y050 panel energized.

B. C. D. -. NRC EXAM Page: 92 of 144 12/22/06 ATION ANSWER KEY 2007 NRC - 49 ID: N-ILT5057-6A-002 Points: I.~O Given the following conditions on Unit 2:

2A 129250 Volt Battery Charger has been lined up and is performing an "equalize" charge on its battery. During the charge, AC power to the charger becomes unavailable and subsequently made available when the bus is reenergized by the diesel generator.

Which of the following describes the expected response of this battery charger? The 2A Battery Charger will: A. return in the "float" charge mode. B. return in the "equalize" charge mode. C. remain de-energized and cannot be restored with the diesel generator powering the bus. D. remain de-energized until manually restored as permitted by diesel generator loading. Answer: B b' 12/22/06 NRC EXAM Page: 93 of 144

.. =---- Question 49 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice: Diff icu Ity : Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5057-6A-002 Given the following conditions on Unit 2: *2A 1251250 Volt Battery Charger has 918 Active No No 3.00 N-ILT-5057-6A-002 2 1 .oo 263000 K6.01 A 0.00 0.00 Importance:

RO 3.2 / SRO 3.5 Cognitive-Level:

High

Reference:

SO 57B. 1-2 Justification:

A. B. Incorrect - the charger will return to the equalize charge mode. Correct -from Note 2 in SO 57B.1-2 "Upon a loss of AC input power, the battery charger will return to the same mode it was in once power is restored.

IF the battery charger was in the Equalize mode, THEN the timer will pick up where it was interrupted AND time out." Incorrect - the 2A battery charger is a safety- related component and is automatically restored approximately 16 seconds after the diesel generator restores power to the emergency bus. Incorrect - the 2A battery charger is a safety- related component and is automatically restored approximately 16 seconds after the diesel generator restores power to the emergency bus.

C. D. 'G ~ NRC EXAM Page: 94 of 144 12/22/06

' ID: N-ILTS057-10-001 Points: 1.00 'ci 50 t< Unit 2 was operating at full power when the following occurs on the Balance of Plant (BOP) Station Batteries:

2AD004/2BD004 BATTERY GROUND (220 H-5) alarms. The EO sent to investigate reports that the Ground Lamp indications on the 20D005 Panel are as follows: * * * "Ground Lamp A is BRIGHTLY lit". "Ground Lamp B is OUT'. "The Ground Detection Ammeter is reading mid-scale".

The crew begins to search for the ground by isolating loads in accordance with A0 57A. 1- 2, 125/250 VDC Balance of Plant Station Battery Ground Investigation.

When the grounded load is ISOLATED, all of the Ground Lamp Indications will be -(I)- lit and the ground detection ammeter will approach -(2)-. A. (1) brightly (2) zero B. (1) brightly (2) full scale C. (1) dimly (2) zero D. (1) dimly (2) full scale Answer: C ---= NRC EXAM Page: 95 of 144 12/22/06 Question 50 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5057-1 D-001 Unit 2 was operating at full power when the following occurs:

2AD00412BD004 BATTE 1177 Active No No 4.00 2 1 .oo 263000A3.01 N-ILT-5057-1 D-001 0.00 0.00 Importance:

RO 3.2 lSRO 3.,3 Cog n i tive-Level:

High

References:

A0 57A. 1-2 Justification:

A. Incorrect - Although the ammeter should approach zero, Ground Lamp indications that are brightly lit indicate a positive ground.

Incorrect - Ground Lamp indications that are brightly lit indicate a positive ground and full scale ammeter indicates a significant ground is present. Correct - With no ground present (isolated), both lights will be dim. Ammeter will read near zero due to low ground current.

Incorrect - Although the Ground Light indications should be dimly lit, a full scale ammeter reading indicates a significant ground is present. B. C. D. ---/' NRC EXAM Page: 96 of 144 12/22/06

i/ ' EXAMINA Question 51 Details Question Type: Multiple Choice Topic: N-ILT-5052-6G-004 An electrical fault and blown fuse has resulted in the loss of the Unit 3 Div II System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: 1179 Active No No 4.00 3 1 .oo 264000 K6.09 A 0.00 0.00 Importance:

Rc 3.3 / SRO Cognitive-Level:

High N-l LT-5052-6G-004 5

Reference:

PLOT5052.06G, print E-27 sht 1 Justification:

A. B. C. Incorrect - E-2 will start and be available for loading. Incorrect - E-2 will start and be available for loading. Correct - Div II 125 V DC Panel 3 PD supplies the E-4 EDG VDC logic and solenoid power. E-4 will not start.

Incorrect - Div II 125 V DC Panel 3 PD supplies the E-4 EDG VDC logic and solenoid power.

E-4 will not start. D. ~v NRC EXAM Page: 98 of 144 12/22/06 ANSWER KEY RO Rev 0 .-' 52 ID: Ea-ILTS03666-001 Points: 1.00 The Instrument Air System is in a normal lineup when the following occur: * * * *

INSTRUMENT AIR DRYER TROUBLE (216 C-4) goes into alarm. B INSTRUMENT AIR HEADER LO PRESS (216 D-4) goes into alarm. "B" Instrument Air Header Pressure (PI-24258) on Panel 20C012 is lowerinq. "B" Instrument Air Receiver Pressure (PI-24296) on Panel 20C012 is steady. The TBEO reports there is a valve malfunction on the "B' Instrument Air Dryer and that neither the "C" or "D" drying tower is in service. Which one of the following describes (1) the on-going effect on "B" instrument air header pressure, assumina no operator action is taken, and (2) what action(s) will mitigate this event?

A. (1) Pressure will continue to lower. (2) Cross-tie "A' and "6" instrument air headers.

B. (1) Pressure will continue to lower. (2) Cross-tie Unit 2 and Unit 3 "B' instrument air headers.

C. (I) Pressure will recover when Service Air Isolation PCV-2428 is fully closed. (2) Isolate the "B' Instrument Air Dryer. D. (1) Pressure will recover when Service Air Isolation PCV-2428 is fully closed. (2) Bypass the "6" Instrument Air Dryer. Answer: B

.< ...----' NRC EXAM Page: 99 of 144 12/22/06 EXAMINATION ANSWER KE Question 52 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value:

Cross

Reference:

User Text: User Number I: User Number 2: Comment: Multiple Choice N-ILT-5036-66-001 The Instrument Air System is in a normal lineup when the following occur: *INSTRU 1298 Active No No 3.00 2 1 .oo 300000A2.0 1 N-l LT-5036-6B-001 0.00 0.00 Importance:

RO 2.9 I SRO 2.8 Cognitive-Level: High

References:

ON-I 19, M-320 Justification:

A. Incorrect - Cross-tying the "A' and "B" instrument air headers will not be effective in restoring "B" instrument air header pressure since the "A" supply must pass through the "6' Air Dryer in order to supply the "B" header. Correct - The given conditions indicate both towers for the "B' Air Dryer are isolated, which means there is no flow to the "B" instrument air header from the "B" air compressorh-eceiver.. . "B" instrument air header pressure will continue to lower. The correct action to take for this, as directed in ON-I 19, is to cross-tie the Unit 2 and Unit 3 "B! instrument air headers. Incorrect - "B" instrument air header pressure will not recover when PCV-2428 closes since the supply from the "C" compressor/receiver must pass through the "6" Air Dryer in order to supply the "B' header. Incorrect - "B" instrument air header pressure will not recover when PCV-2428 closes since the supply from the "C" compressorheceiver must pass through the "B! Air Dryer in order to supply the "6" header. B. C. D. NRC EXAM Page: 100 of 144 12/22/06

'4 53 ID: N-lLTS034-6D-001 Points: f.00 Given the following:

0 0 0 Peach Bottom Unit 2 is operating at 100% power The "A' TBCCW trips on an electrical fault.

The "B" TBCCW pump is blocked. Which of the following describes the impact of this event and the associated required action?

A. Due to the imminent loss of Condensate pumps, scram the reactor IAW GP-4, "Manual Scram".

B. Due to the imminent loss of Stator Water Cooling, if Generator load is greater than 7,760 amps, perform GP-4, "Manual Scram". C. Due to a loss of lsophase Bus Cooling, reduce Main Generator load to less than 18,000 amps IAW GP-9-2, "Fast Reactor Power Reduction".

D. Due to loss of cooling to the Instrument Air compressors, immediately cross-tie the Instrument Air header with Unit 3 IAW ON-I 19, "Loss Of Instrument Air". Answer: C L.' NRC EXAM Page: 101 of 144 12/22/06 Question 53 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5034-6d-001 Loss of Cooling to TBCCW loads Impact on Ops 1263 Active No No 3.00 2 1 .oo 400000 A2.01 N-I LT-5034-6D-001 0.00 0.00 Importance:

RO 3.3 I SRO 3.4 Cognitive-Level: High

Reference:

PLOT 5034, OBJ. 6d; ON-I 13 & Bases; ON-I 18 & Bases Justification:

A. Incorrect - ON-I 18 directs monitoring Condensate pump bearing and motor oil temperatures and if any temperature is at or above 190 degrees, or if any pump vibration alarm is received, then ON-I 18 directs reducing reactor power IAW GP-9-2 and removing the affected pump(s) from service. ON-I 18 does not direct a manual scram due to imminent loss of Condensate pumps. Incorrect - Stator Cooling is cooled by Service Water, not TBCCW.

Correct - if TBCCW cannot be restored, ON-118 directs reducing Generator load to less than 18,000 amps IAW GP-9-2. As stated in ON-I 18, "isolated bus coolers are not considered vital TBCCW loads; hence a loss of TBCCW and the subsequent isolation of non- vital TBCCW loads, during swap to RBCCW, results in a loss of cooling water to these coolers. " Incorrect - with both TBCCW pumps tripped (both breaker contactors open), vital TBCCW loads (CRD pumps and Instrument Air Compressors) will automatically swap to RBCCW after a 40-second time delay. RBCCW will provide sufficient cooling to Instrument Air Compressors, preventing the need to cross-tie the Unit 2 Instrument Air header to Unit 3. B. C. D. NRC EXAM Page: 102 of 144 12/22/06 TlON ANS 2007 NRC RO Rev 0 54 ID: N-lLT-5003-1A-003 Points: 1.00 Unit 2 IS at 70% power to support a control rod pattern adjustment During a one notch withdrawal attempt the RO is unable to withdraw a control rod and notices the following.

0 Drive flow: 4 gpm 0 Drive pressure:

200 psid above reactor pressure 0 0 The control rod is selected Drive-In and Drive-Out and Settle lights and timing are normal "Rod Withdrawal Permissive" light is lit No rod withdrawal block exist Using the attached Table 2 of SO 62.1.A-2, "Withdrawingllnserting a Control Rod", select the one condition that is the cause of the stuck control rod. A. Air in the control rod drive mechanism.

B. C. Improper hydraulic control unit valve line-up.

Inlet to drive water filters HV-2-3-170 is closed. D. Worn or bad Control Rod Drive Mechanism seals. Answer A .. .-.-.-' Page: 103 of 144 12/22/06 NRC EXAM ATION ANS ' 2007NRCRORevO

' EX - Question 54 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number I: User Number 2: Comment: Multiple Choice N-ILT-5003-1A-003 Unit 2 is at 70% power to support a control rod pattern adjustment. During a one 1358 Active No No 3.00 2 1 .oo 201001 G2.4.31 N-ILT-5003-1A-003 0.00 0.00 Importance:

RO 3.3 / SRO 3.4 Cog n itive-Level: High

References:

SO 62.1 .A-2 Withdrawing/lnserting a Control Rod THE EXAMINEE WILL NEED TABLE 2 "CAUSE AND CORRECTIVE ACTION TROUBLE SHOOTING" OF QUESTION.

SO 62.1 .A-2 IN ORDER TO ANSWER THIS Justification:

A. Correct - Drive flow is high. Drive pressure is low. B Incorrect - Drive flow would be low, drive pressure would be normal. C. Incorrect - Drive flow would be low. D. Incorrect - Drive pressure would be normal.

NRC EXAM Page: 104 of 144 12/22/06 90122lZ 1 PP 1 40 SO 1 :a6ed lryW3 3tlN .---. A~NO 6 pue 7: '3 AlNO 2 '8 AlNO 1 'V EXAMINATION Question 55 Details Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5012-6F-001 A startup is in progress on Unit 3 with reactor power at 5%. Panel 30Y34 is inadv 1299 Active No No 3.00 2 1 .oo 204000KL6.08 N-ILT-5012-6F-001 0.00 0.00 Importance: RO 3.5 / SRO 3.5 Cognitive-Level:

Memory

References:

GP-8.D, GP-8.C, M-I -S-23; A0 58A.3-2 Justification:

A. B. C. Incorrect - MO-3-12-15 does not close on loss of 30Y34. Incorrect - MO-3-12-68 will also close on loss of 30Y34. Correct - Panel 30Y34 provides power to PClS outboard isolation valve logic. Loss of 30Y34 will result in isolation of the associated outboard containment isolation valves, including RWCU valves MO-3-12-18 and MO- 3-12-68. Note that a loss of Panel 30Y33 causes a loss of power to PClS inboard isolation valve logic. This in turn would result in closure of associated inboard containment isolation valves and, in the case of RWCU, a closure of the outboard containment isolation valves as well. This is due to loss of power to the NRHX high outlet temperature relay, which feeds both the inboard and outboard RWCU isolation valve logic. Note #2 in GP-8.C and GP-8.D describe the RWCU response to a loss of 20(30)Y33 and 20(30)Y34, respectively.

Incorrect - MO-3-12-15 does not close on loss of 30Y34. D. u Page: 106 of 144 12/22/06 NRC EXAM

Question 56 betails Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5062-3A-007 Unit 2 is at 100% power.

A Transient requiring control rod insertion per GP-9-2 1181 Active No No 3.00 2 1 .oo 214000K1.04 N-l LT-5062-3A-007 0.00 0.00 Importance: RO 3.2 / SRO 3.2 Cognitive-Level:

Memory

References:

ARC-21 1 D-5 Justification:

A. B. Correct - A select block stops all rod movement except scram.

Incorrect - Select block from RPlS failure stops all rod movement. The operator cannot even select any control rod for insertion.

C. D. L- 12/22/06 Page: 108 of 144 NRC EXAM ID: N-lLTS007F-lE-001 Points: 1.00 57 '*.' A Traversing In-Core Probe trace is being performed using automatic operation Which of the following states the response of the TIP system when a Group II isolation is actuated with one detector in the core ? A. The inserted detector withdraws to the bottom of core position and the associated ball valve will close. B. The inserted detector withdraws to the "inshield" position and the associated ball valve will close.

C. The trace continues unaffected by the isolation however, the isolation must be reset before any additional detectors can be inserted into the core.

D. The shear valve associated with the inserted detector fires isolating that detector. Other TIP guide tubes are isolated by the normally closed ball valve. Answer: B ---.- NRC EXAM Page: 109 of 144 12/22/06

.-e' EXAMINATION'ANSWER KEY 2007 NRC RO Question 57 Details Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-5007F-1 E-001 A Traversing In-Core Probe trace is being performed using automatic operation.

1183 Active No No 3.00 2 1 .oo 215001A3.03 A 0.00 0.00 Importance:

RO 2.5 I SRO 2.6 Cognitive-Level: Memory N-ILT-5007F-1 E-001

Reference:

PLOT 5007F; SO 7F.7.A-2 Justification

A. Incorrect - Each shear valve must be actuated by a keylock switch located on the TIP console valve control monitor panel.

Correct - On a Group II D isolation signal the TIP detector is first retracted to the inshield position and the ball valve closes. Incorrect - On a Group II D isolation signal the TIP detector is first retracted to the inshield position and the ball valve closes. Incorrect - Each shear valve must be actuated by a keylock switch located on the TIP console valve control monitor panel. B. C. D. --.,- NRC EXAM Page: 110of144 12/22/06

* * *

A low level transient occurred on Unit 2 causing a reactor scram from 100% power. HPCl and RClC initiated on low RPV level. Reactor level is + 20" and rising quickly. LT-2-02-3-072C, Wide Range Reactor Vessel Water Level, fails downscale.

All other RPV level instruments remain operable.

Assuming no operator action is taken, what is the expected response of the HPCl system as RPV level rises?

A. HPCl will trip at RPV level of +29". B. HPCl will trip at RPV level of +46". C. HPCl will isolate at RPV level of +29". D. HPCl will not trip on high RPV level and level will continue to rise. Answer: D NRC EXAM Page: 111 of 144 12/22/06 EXAMINATION ANSWER KEY Question 58 Details P Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5023-61-001 *Unit 2 had been operating at 100% power. *A low level transient occurred causing 1300 Active No No 4.00 2 1 .oo 216000K3.14 N-l LT-5023-61-00 1 0.00 0.00 Importance:

RO 3.8 / SRO 4.2 Cog n i tive-Level:

High

References:

ARC 221 B-1 HPCl TURB TRIP Justification:

A. Incorrect - HPCI high RPV level trip setpoint is +46", not 29". Trip needs input from both LT-72C and LT-72D (2 out of 2 logic). Incorrect - HPCl high RPV level trip needs input from both LT-72C and LT-72D (2 out of 2 logic). Incorrect - HPCI does not isolate at RPV level of +29". RPV level of +29" is the level at which HPCl will restart if tripped at

+46' with no operator action. Correct - HPCl RPV high level trip requires input from both LT-72C and LT-72D (2 out of 2 logic). With one transmitter downscale, the HPCl system will not trip on high level.

B. C. D. ~-2 NRC EXAM Page: 112 of 144 12/22/06 a :JaMsuV

-4' ATION A ' 2007 NRC RO R EX Question 59 Details Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-2102-5A-017 Unit 2 was operating at 100% power when a feedwater line break occurred inside con 1301 Active No No 2.50 2 1 .oo 230000A4.12 N-ILT-2102-5A-017 0.00 0.00 Importance: RO 3.8

/ SRO 3.8 Cog n i tive-Level:

High

References:

T-102.

T-I 02 Bases Justification:

A. Incorrect - The torus spray headers are assumed to be covered if torus level is above 21 feet. Correct - T-I 02 directs spraying the torus IF torus level is below 21 feet and BEFORE torus pressure reaches 9 psig. As stated in the Bases for T-102, "21 feet is the upper limit of torus level indication. Therefore, above 21 feet it is assumed that the torus spray spargers are submerged and that no spray action will occur". Incorrect - 18 feet is the torus level above which the torus-to-drywell vacuum breakers begin to submerge, however this is why drywell sprays are not initiated unless torus level is below 18 feet ... this does not prevent initiating torus sprays. Incorrect - The torus-to-drywell vacuum breakers begin to submerge at a torus level of 18 feet and rising. B. C. D. .I .u' NRC EXAM Page: 114 of 144 12/22/06 A3?i 83MSN'd .NOllt/NIlAlVX3

<Question 60 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2:

Comment: Multiple Choice N-ILT-5010-2A-001 Unit 2 is in MODE 4. The "B" Loop of RHR is lined up to cool the fuel pool per A0 1185 Active No No 2.50 2 1 .oo 233000K2.02 N-l LT-5010-2A-001 0.00 0.00 Importance: RO 2.8 / SRO 2.9 Cog n i tive-Level:

High

References:

A0 10.3-2; ARC 005 B-I Justification:

A. B. C. D. Incorrect - E4 Diesel output breaker is locked out from closing due to the E42 bus fault. Incorrect - E4 Diesel is locked out, and the 2D RHR Pump will trip of loss of E-42 bus power.

Incorrect - 2A RHR Pump is powered from the E12 Bus. Shutdown cooling will not be lost. Correct - E4 Diesel will auto start on low E-42 bus voltage, but does not load onto the E-42 bus due to the bus fault condition. With E-42 bus de-energized the 2D RHR pump has no power and therefore RHR system assist with fuel pool cooling is lost. '-' NRC EXAM Page: 116 of 144 12/22/06

'4 61 ID: N-ILT-5001 DL-5A-Ob1 Points: 1.00 0 0 Which of the following describes the INITIAL response of reactor pressure and Turbine Control Valve position to this transient?

Unit 3 is operating at rated power.

A fully withdrawn control rod scrams. Reactor Pressure TCV Position A. Decreases Open slightly B. Decreases Close slightly C. Remains constant Open slightly D. Remains constant Close slightly Answer: B U NRC EXAM Page: 117 of 144 12/22/06

EXAMINATION ANSWER KEY 2007 NRC RO Rev 0 .I -.--- 62 ID N -I LT-500 1 BAA-009 Points: i .'OO Which of the following identifies the expected positions of the Turbine Control Valves (TCVs), Combined Intermediate Valves (CIVs), and Feedwater Heater Extraction Steam Isolation Valves (ESIVs) following a turbine trip? TCVs clvs ESlVs A. Closed Closed Closed 6. Closed Closed Open C. Closed Open Closed D. Open Closed Closed Answer: B Question 62 betails \.,- Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5001 B-4A-009 Which of the following identifies the expected positin of the Turbine Control 1187 Active No No 2.50 N-l LT-500 1 B-4A-009 2 1 .oo 245000A1.03 0.00 0.00 Importance:

RO 2.7 I SRO 2.9 Cognitve-Level:

Memory

References:

SO 1 B.2.A-2, SO 1 B.2.A-3 Justification:

A. 6. Incorrect - ESlVs will only get a close signal on feedwater heater high level.

Correct - TSVs and ClVs will get a close signal on a turbine trip for turbine protection. ESlVs will only get a close signal on feedwater heater high level. Incorrect - ClVs will get a close signal on a turbine trip for turbine protection.

Incorrect - TCVs will get a close signal on a turbine trip for turbine protection.

C. D. NRC EXAM Page: 119 of 144 12/22/06 EXAMINATION ANSWE 2 CR 0 LJ , ' Points: 1 .OO 63 ID: N-ILT-5006-61-002 Unit 2 is operating normally at 100% power when:

FEEDWATER FIELD INSTRUMENT TROUBLE (201 H-I) goes into alarm. "B' main steam line flow indicator Fl-2-06-0888 on Panel 20C08A instantaneously fails downscale.

What is the impact of this malfunction on the Feedwater System (1) and what actions are required to mitigate this event (2)?

A. (1) Total feed flow will lower. (2) Verify reactor water level is being maintained by DFCS in single element control. B. (I) Total feed flow will rise.

(2) Verify reactor water level is being maintained by DFCS in three element control. C. (1) Total feed flow will remain as is. (2) Verify reactor water level is being maintained by DFCS in single element control. D. (1) Total feed flow will remain as is. (2) Verify reactor water level is being maintained by DFCS in three element control . Answer: C NRC EXAM Page: 120 of 144 12/22/06

-d Question 63 Details Question Topic: Type: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5006-61-002 Unit 2 is operating normally at 100% power when: *FEEDWATER FIELD INSTRUMENT TROU 1302 Active No No 3.50 2 1 .oo 259001A2.07

N-l LT-5006-61-002 0.00 0.00 Importance:

RO 3.7 / SRO 3.8 Cognitive-Level: High

References:

ARC 201 H-I, OT-100 Justification:

A. Incorrect - Since there was no plant transient (no change in actual feed flow, steam flow ro RPV level), DFCS will maintain RPV level as is. Incorrect - Since there was no plant transient (no change in actual feed flow, steam flow or RPV level), DFCS will maintain RPV level as is. Also, DFCS will automatically transfer to single element control.

Correct - as stated in Step 3.2 of OT-I 00, "If any feedwater flow indication is upscale or any steam line flow indication is downscale, then verify the Feedwater Level Control System is operating in single element control".

Since there was no plant transient (no change in actual feed flow, steam flow or RPV level), DFCS will maintain RPV level as is in single element control. Incorrect - DFCS will automatically transfer to single element control. B. C. D. ~, *-' NRC EXAM Page: 121 of 144 12/22/06

EXAMINATION ANSWER RC RO'Rev Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5037-4F-001 Unit 3 is operating at rated power. The following conditions exist:

Fire Water 1189 Active No No 3.00 2 1 .oo 286000 K4.06 A 0.00 0.00 Importance: RO 3.3 / SRO 3.5 Cognitive-Level:

Memory N-l LT-5037-4F-00 I

References:

ARC 201 A-3, ARC 201 C-I Justification:

A. 8. Incorrect -The MDFP will auto start at 140 psig fire system pressure.

Incorrect - When fire system pressure lowers to 140 psig the Motor Driven Fire Pump (MDFP) will auto start and at 130 psig the Diesel Driven Fire Pump (DDFP) will auto start.

Correct - When fire system pressure lowers to 140 psig the Motor Driven Fire Pump (MDFP) will auto start and at 130 psig the Diesel Driven Fire Pump (DDFP) will auto start.D. Incorrect - When fire system pressure lowers to 140 psig the Motor Driven Fire Pump (MDFP) will auto start and at 130 psig the Diesel Driven Fire Pump (DDFP) will auto start.

C. D. \-.. NRC EXAM Page: 123 of 144 12/22/06

--& 65 ID : N-l LT-5002-1 QdOl Points: 1.00 * * * *

A Unit 3 scram condition occurred due to a loss of feedwater transient.

RPV level reached -55 inches and was recovered by both HPCl and RCIC. All control rods inserted. RPV pressure is 825 psig.

A cooldown was commenced using ST-0-080-500-3, "Recording and Monitoring Reactor Vessel Temperatures and Pressure". For these conditions, what is the impact on RPV bottom head drain temperature?

A. Bottom head drain temperature is not accurate due to lack of forced circulation ONLY. B. Bottom head drain temperature is not accurate due to lack of forced circulation and RWCU out of service. C. No impact. Bottom head drain temperature is accurate due to recirculation pumps being at minimum speed. D. No impact. The bottom head drain temperature is accurate due to RWCU system remaining in service. Answer: B '-/ NRC EXAM Page: 124 of 144 12/22/06 v Question 65 Details Question Type: Topic: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5002-1 Q-001 *A Unit 3 scram condition occurred due to a loss of feedwater transint. *RPV leve 1303 Active No No 3.00 N-ILT-5002-1 Q-001 2 1 .oo 202001 K3.07 0.00 0.00 Importance: RO 2.9 / SRO 2.9 Cognitive-Level:

High

References:

T-I 00 Bases, ST-0-080-500-2 Justification:

A. Incorrect - Bottom head drain temp is not accurate mostly due RWCU being out of service (isolated at -48" RPV level).

Correct - Since RPV level went below

-48", both Recirc pump tripped and RWCU system isolated.

With no core forced circulation or RWCU system flow through the bottom head drain line, bottom head drain line temperature is not accurate.

Incorrect - Recirc pumps tripped at -48" RPV level. They are not in service. isolated at -48" RPV level. B. C. D. Incorrect - RWCU is not in service. The system ...---- NRC EXAM Page: 125 of 144 12/22/06 ON ANSW 07 NRC RO Rev 0 EX v' ID: N-ILT-1570-12-003 Points: 1.00 66 An Equipment Operator (EO) accrued the following working hours while working a forced outage.

He does NOT have an authorized "Overtime Guideline Deviation Authorization" form. Saturday NO HOURS Sunday NO HOURS Monday 06100 - 16:OO Tuesday 06100 - 23:OO Thursday 07100 - 20:OO Friday 06100 - 22100 Wednesday 0630 - 22:OO Identify by number which guidelines the EO violated.

1. 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period

2. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period 3, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period A. 1 and 3 only B. 1 and 2 only C. 1, 2, and 3 'v' D. 2 and 3only Answer: B "J NRC EXAM Page: 126 of 144 12/22/06

Question 66 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-1570-12-003 An Equipment Operator (EO) accrued the following working hours while working a 1191 N-ILT-1570-12-003 Active No No 3.00 3 1 .oo K2.1.1 B 0.00 0.00 Importance: RO 3.7 I SRO 3.8 Cognitive-Level:

High

Reference:

LS-AA-119 Justification:

A. Incorrect - There is no 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 7 day period violation.

He worked a total of 71 112 hours0.0013 days 0.0311 hours 1.851852e-4 weeks 4.2616e-5 months . Correct - 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period was violated on Tuesday (1 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ), 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period was violated due to total of Monday's and Tuesday's hours (27 total hours). Incorrect - There is no 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 7 day period violation. He worked a total of 71 1/2 hours. Incorrect - There is no 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 7 day period violation, He worked a total of 71 1/2 hours. B. C. D. u NRC EXAM Page: 127 of 144 12/22/06 K 2007 NRC RO Re X 67 ID: N-ILT-1504-1-001 Points: 1.00 According to HU-AA-104-101, "Procedure Use and Adherence", when a conflict arises between a standard procedure and a site-specific procedure, which procedure prevails?

A. The standard procedure always prevails.

8. The site-specific procedure always prevails.

C. The standard procedure prevails except when the site-specific procedure directs actions that ensure compliance with regulatory requirements.

D. The site-specific procedure prevails except when the standard procedure directs actions that ensure compliance with regulatory requirements.

Answer: C Question 67 Details b' Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1504-1-001 According to HU-AA-104-101, "Procedure Use and Adherence", when a conflict arises 1324 Active No No 2.50 2 1 .oo G2.1.20 N-l LT-1504-1-001 0.00 0.00 Importance:

RO 4.3 I SRO Cognitive-Level:

Memory

References:

HU-AA-104-101 Justification:

C. Correct - as stated in HU-AA-104-101, "Whenever a conflict arises between a standard procedure and a site-specific procedure, then the standard procedure shall prevail except when the site-specific procedure directs actions that ensure compliance with regulatory requirements".

12/22/06 NRC EXAM Page: 128 of 144 u 68 ID: N-ILT-1528-2-001 Poinfs: 1.00 Given the following:

0 0 0 0 0 The Unit 2 HPCI System was declared INOPERABLE and has been blocked out of service for maintenance.

A maintenance activity was performed on MO-2-23-057, "HPCI Torus Suction Outboard".

The maintenance activity included a valve stroke test using a partial ST-0-023-301-2, "HPCI Pump, Valve, Flow and Unit Coolers Functional and lnservice Test". The initial and second re-test OPEN stroke time for the valve was in the ALERT Range. The CLOSE stroke time for the valve was acceptable. Based on the guidance in both ST-0-023-301-2,"HPCI Pump, Valve, Flow and Unit Coolers Functional and lnservice Test", and NOM-P-11.1 "Operability", the MO-2-23-057:

A. operability status is indeterminate.

6. can be considered OPERABLE since the CLOSE stroke time was ACCEPTABLE.

C. remains INOPERABLE and must be examined to determine the root cause. ., 'J D. can be considered OPERABLE if a third and fourth OPEN stroke time is in the ACCEPTABLE Range.

Answer: C '---' NRC EXAM Page: 129 of 144 12/22/06

Question 68 Details Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-1528-2-001 Given the following: A maintenance activity was performed on MO-2-23-057, "HPCI Tor 1345 N-l LT-1528-2-00 1 Active No No 3.00 4 1 .oo 2.2.24 0.00 0.00 Importance:

RO 2.6 I SRO 3.8 Cognitive-Level:

Memory

References:

ST-0-023-301-2; NOM-P-11 .I Justification:

A. Incorrect - Both ST-0-023-301-2 and NOM-P-11.1 give clear guidance that after an allowed second stroke if the times are still unacceptable then the valve must be declared inoperable. Also, per NOB-P-I 1.1 there is no "indeterminate'ktatus.

The component is either operable or inoperable.

B. Incorrect - the valve's safety function is in the open direction.

The open stroke time must meet acceptable times or be declared inoperable.

C. Correct - Per ST-0-023-301-2 Limitations 4.3.2 and 4.3.3: Any valve that exceeds its limiting stroke time criteria shall be immediately declared INOPERABLE.

Any valve with a stroke time in the ALERT Range shall be immediately re-tested OR declared inoperable. Per NOM-P-11.1, Operability, test failures should be examined to determine the root cause and correct the problem before resumption of testing. Repetitive testing to achieve acceptable test results without identifying the root cause or correction of any problem in a previous test is not acceptable as a means to establish or verify operability. Examples include cycling a valve until acceptable stroke times are achieved.

D. Incorrect - ST-0-023-301-2 gives clear guidance that after an allowed second stroke if the times are still unacceptable then the valve must be declared inoperable.

NRC EXAM Page: 130 of 144 12/22/06 ION AN 2007 NRC RO Rev 0 Prior to inserting the TN-68 Spent Fuel Storage cask into the Fuel Pool Cask Pit per procedure SF-220 "Spent Fuel Cask Loading and Transport Operations", Fuel Pool Cooling is A. maximized and Fuel Pool level is lowered to between 232' 4" and 232' 5". B. secured and Fuel Pool level is lowered to between 232' 4" and 232' 5". C. secured and Fuel Pool level is raised to approximately 232' 6.5" (for Unit 2) or 232' 4.5" (for Unit 3). D. maximized and Fuel Pool level is raised to approximately 232' 6.5" (for Unit 2) or 232' 4.5" (for Unit 3). Answer B NRC EXAM Page: 131 of 144 12/22/06 ANS RO Rev 0 Question 69 Details ' Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number I: User Number 2: Comment: Multiple Choice N-ILT-5071-2-002 Prior to inserting the TN-68 Spent Fuel Storage cask into the Fuel Pool Cask Pit 1193 Active No No 4.00 2 1 .oo 2.2.28 A 0.00 0.00 Importance: RO 2.6 / SRO 3.5 Cognitive-Level: Memory N-ILT-5071-2-002

Reference:

PLOT 5071; SF-220 Justification:

A. B. Incorrect - SF-220 requires that Fuel Pool Cooling System is secured.

Correct - Procedure SF-220 requires that Fuel Pool Cooling system is secured and to establish Fuel Pool level between 232' 4" and 232' 5". Incorrect - SF-220 requires Fuel Pool level to be between 232' 4" and 232' 5". This is applicable to both units. Incorrect - SF-220 requires that Fuel Pool Cooling system is secured. C. D. LJ' NRC EXAM Page: 132 of 144 12/22/06 R 'd 70 ID: N-ILT-1535-4-004 Points: 1.00 A Reactivity Maneuver (ReMA) Form is required for which of the following activities?

A. Inserting control rods to clear APRM Hi alarms. B. Adjusting reactor recirculation flow to maintain full reactor power C. Unplanned insertion of a control rod for operability concerns.

D. Withdrawing control rods during continuation of a reactor startup above 25% power. Answer: D I. -, .--- NRC EXAM Page: 133 of 144 12/22/06 NSWER KEY I Question 70 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-1535-4-001 A Reactivity Maneuver (ReMA) Form is required for which of the following activities 1194 Active No No 2.50 2 1 .oo 2.2.34 N-ILT-1535-4-001 0.00 0.00 Importance:

RO 2.8 / SRO 3.2 Cognitive-Level:

Memory

References:

OP-AB-300-1003, BWR Reactivity Maneuver Guidance, GP-5 Justification:

A. Incorrect - Inserting control rods to clear APRM Hi alarm is considered a single reactivity maneuver per OP-AB-300-1003 and a ReMA is not required.

Incorrect - Routine load changes with reactor recirculation flow is considered a single reactivity maneuver per OP-AB-300-1003 and a ReMA is not required.

Incorrect - Unplanned insertion of a control rod for operability concerns is a simple reactivity maneuver per OP-AB-300-1003 and a ReMA is not required.

Correct - Per OP-AB-300-1003, "BWR Reactivity Maneuver Guidance" continuation of a reactor startup above 25% power is a complex maneuver and requires a ReMA.

B. C. D. .. ---- NRC EXAM Page: 134 of 144 12/22/06

E \rJ Question 71 Details Question Topic: Type: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1770-3-002 The Equipment Operators (EOs) need to perform a surveillance test in an area with 1196 Active No No 3.00 3 1 .oo 2.3.02 A 0.00 0.00 Importance:

RO 2.6 / SRO 3.0 Cog nitive-Level:

High N-ILT-1770-3-002

Reference:

PLOT 1770; RP-AA-400 Justification:

A. B. C. Incorrect - 1 individual for 60 minutes in an 80mR/hr field

= 80 mR total exposure.

Incorrect - 2 individuals for 35 minutes in an 80mR/hr field

= 93.3 mR total exposure.

Incorrect - 2 individuals for 15 minutes in an 80mRlhr field = 40 mR exposure plus 2 individuals for 35 minutes in an 8mR/hr field

= 9.3.mR = 49.3 mR total job exposure.

Correct - 1 individual for 30 minutes in an 80mR/hr field = 40 mR exposure plus 1 individual for 60 minutes in an 8mR/hr field

= 8mR = 48 total job exposure.

D. 'd/ NRC EXAM Page: 136 of 144 12/22/06 S K 0 %-' ID: N-ILT-17304-001 Points: 1.00 72 . An Equipment Operator has been assigned to enter the Moisture Separator Area to investigate a steam leak. The following information has been provided.

The Equipment Operator has 3280 TEDE annual Exposure. Expected dose for investigation of the steam leak is 300 mR. In accordance with RP-AA-203,"Exposure Control and authorization", which one of the following describes the action required to complete the steam leak investigation based on the above conditions?

A. Planned Special Exposure must be obtained.

B. Dose Control Level Extension must be obtained C. D. Emergency Exposure Extension must be obtained.

No action required if total exposure is less than 4000 mR Answer: B 'L/ NRC EXAM Page: 137 of 144 12/22/06 Question 72 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1730-4-001 An Equipment Operator has been assigned to enter the Moisture Seperator Area to 1197 Active No No 3.50 N-ILT-I 730-4-001 2 1 .oo 2.3.4 0.00 0.00 Importance:

RO 2.5 / SRO 3.1 Cognitive-Level:

Memory

References:

RP-AA-203, Exposure Control and Authorization Justification:

A. Incorrect - RP-AA-203 required dose extension above 2000 mR TEDE. Dose extensions are granted in 500 mR increments. The current extension is good to 3500 mR. Another extension is required to get to 3580 mR expected exposure. This evolution does not qualify as a Planned Special Exposure or Emergency Exposure Extension.

Incorrect - RP-AA-203 required dose extension above 2000 mR TEDE. Dose extensions are granted in 500 mR increments. The current extension is good to 3500 mR. Another extension is required to get to 3580 mR expected exposure. This evolution does not qualify as a Planned Special Exposure or Emergency Exposure Extension.

Incorrect - RP-AA-203 required dose extension above 2000 mR TEDE. Dose extensions are granted in 500 mR increments. The current extension is good to 3500 mR. Another extension is required to get to 3580 mR expected exposure.

This evolution does not qualify as a Planned Special Exposure or Emergency Exposure Extension.

B. Correct - Per RP-AA-203 C. D. .-;' NRC EXAM Page: 138 of 144 12/22/06

.\/' 73 ID: N-ILT-1560-3401 Points: 1.00 Fuel failure has resulted in an off-site release to the Main Stack Unit 2 conditions are as follows. * * *

The reactor was scrammed with all rods inserting The Main Stack rad release is approaching the ALERT level. Main Steam Line rad is 7,500 mrlhr and slowly rising. RPV pressure is 940 psig and controlled by EHC. Based on the above conditions, which one of the following actions is REQUIRED to control the radioactive release?

A. Start the Mechanical Vacuum Pump and depressurize to the condenser at < 100 Flhr. B. Close the Main Steam Isolation Valves and depressurize to the Suppression Pool at < 100 Flhr. C. Start the Mechanical Vacuum Pump and perform a rapid depressurization to the condenser regardless of cooldown rates. D. Close the Main Steam Isolation Valves and perform an Emergency Blowdown to the Suppression Pool.

Answer: B '---' NRC EXAM Page: 139 of 144 12/22/06 Question 73 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-1560-3-001 Fuel failure has resulted in an off-site release to the Main Stack. Unit 2 condit 1198 Active No No 3.00 3 1 .oo 2.3.1 1 N-ILT-1560-3-001 0.00 0.00 Importance:

RO 2.7 / SRO 3.2 Cognitive-Level:

High

References:

T-104, Radioactivity Release, T-I 01, RPV Control Justification:

A. Incorrect - MVP will not be started with gross fuel failure and condenser will not be used to depressurize.

Correct - T-104 requires MSlVs to be isolated to stop the rad release. Depressurization will be performed in accordance with T-101 < 100 Flhr. Incorrect - T-101 RCP/12 to rapidly depressurize is not required since rad release is not approaching the GE level (a T-I 04 blowdown limits) and a primary system breach is not in progress.

Incorrect - Emergency Blowdown is not required (or permitted) by T-I 04 since rad release is not approaching the GE level and a primary system breach is not in progress.

B. C. D. ~V' NRC EXAM Page: 140 of 144 12/22/06 EXAMINATION ANSWER KEY '4 ?4 ' ID: N-ILT-1560-2-004 Points: 1.00 Unit 3 is operating at 100% power when the following sequence of events occurs:

* * *

Spurious Group 1 Isolation All control rods insert EXCEPT 22-31, which is stuck full-out HPCl and RClC initiate and inject into the RPV RHR and Core Spray remain in standby RClC PUMP ROOM FLOOD (222-A4) alarms Assuming no operator actions have occurred up to this point, which TRIPS should the crew enter?

1. T-I 00, "Scram" 2. T-101, "RPV Control" 3. 4. T-I 02, "Primary Containment Control" T- 1 03, "Secondary Containment Control" A. 1 and 2 B. 2 and 3 C. 2 and 4 D. 3 and 4 Answer: C NRC EXAM Page: 141 of 144 12/22/06

, Question 74 Details -' Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-1560-2-004 Unit 3 is operating at 100% power when the following sequence of events occurs:

1235 Active No No 3.00 2 1 .oo 2.4.2 N-ILT-1560-2-004 0.00 0.00 Importance: RO 3.9

/ SRO 4.1 Cognitive-Level:

High

References:

T-I 00, T-I 00 Bases, T-I 01, T-I 02, T-I 03 Justification:

A. Incorrect - a Group 1 isolation from 100% power would result in a reactor scram and an RPV Lo Level condition, requiring entry into T- 101. As stated in T-100 Bases, "T-100 is entered each time the reactor scrams, provided that an entry condition for T-I 01 does not exist." Incorrect - none of the given conditions indicate an entry condition for T-102 ... since HPCl and RClC have initiated and RHR and Core Spray have not, a High Drywell Pressure condition does not exist. Correct - T-101 would be entered due to an RPV Lo Level condition and T-I 03 would be entered due to the RClC PUMP ROOM FLOOD alarm. Incorrect - there is no entry condition for T-I 02. B. C. D. -+-- NRC EXAM Page: 142 of 144 12/22/06 EXAMINATION ANSWER 07 N 75 ID: N-ILT-1540-3-009 Points: 1.00 L- Unit 2 is operating at 80% power when an electrical transient causes several Control Room annunciators to alarm, including the following:

*

A CONDENSATE PUMP BRK TRIP (203-E2) REACTOR HI-LO WATER LEVEL (210-H2) GENERATOR PROTECTION CIRCUIT ENERGIZED (206-L1) Assuming the alarms are valid, which of the following describes the appropriate operator action?

A. Perform GP-4, "Manual Scram" B. C. D. Perform GP-9-2, "Fast Reactor Power Reduction" Insert ALL GP-9-2 Appendix 1, control rods ONLY. Verify A and B Recirc Pumps runback to 45%. Answer: A .I ---- NRC EXAM Page: 143 of 144 12/22/06 ANSWER KEY 0 Question 75 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-1540-3-009 Unit 2 is operating at 90% power when an electrical transient causes several Contro 1233 Active No No 3.00 2 1 .oo 2.4.45 N-ILT-1540-3-009 0.00 0.00 Importance: RO 3.3 / SRO 3.6 Cog ni tive-Level:

High

References:

GP-9-2, Fast Reactor Power Reduction; GP-4, Manual Scram; OT-100, Reactor Low Level; OT- 11 2, UnexpectedlUnexplained Change In Core Flow; OT-I 13, Loss of Stator Cooling Justification:

A. Correct - a valid GENERATOR PROTECTION CIRCUIT ENERGIZED annunciator indicates a loss of Stator Cooling. OT-I 13 directs a manual scram IAW GP-4 if a valid loss of Stator Cooling exists and generator load is greater than 7760 amps (-23% reactor power).

Incorrect - this action is directed by OT-I 00 for a low reactor water level condition, based on availability of makeup capability. This action would be appropriate if it weren't for the loss of Stator Cooling condition.

Incorrect -the given conditions indicate a trip of the A Condensate Pump, which results in a Recirc runback to 45%, and requiring entry into OT-I 12. Inserting ALL GP-9-2 rods is required by OT-I 12 only if a recirc pump trip has occurred ... none of the given conditions suggest a recirc pump trip has occurred.

Incorrect - although this action would be correct in the case of a loss of the A Condensate Pump (which results in a Recirc runback to 45%), it is not the correct initial action due to the loss of Stator Cooling, which requires a reactor scram.

B. C. C. -4 NRC EXAM Page: 144 of 144 12/22/06

=-J 1 ID: N-lLT-!j002B-6A-001 Points: 1.00 Given the following:

? ? A station blackout occurred

? 0 0 0 0 Using SE-11, Attachment C, determine which statement below is TRUE? Unit 2 was initially operating at 100% power The following RPV level indications exist: Narrow range LI-94A (20C005A) indicates 0 inches Wide range LI-85B (20C005A) indicates

-40 inches Wide range LR-11 OA (20C004C) indicates

+20 inches Fuel Zone range LR-11 OB (2OCOO3-02) indicates +25 inches A. B. Actual RPV level is approximately

+20 inches; maintain RPV level per T-101. Actual RPV level CANNOT be determined; exit T-101 and enter T-116.

C. Actual RPV level is approximately 0 inches; restore and maintain level between +5 and +35 inches per T-I 01. D. Actual RPV level is below 0 inches, but above TAF; restore and maintain level between +5 and +35 inches per T-I 01. Answer: D Lx. NRC EXAM Page: 1 of 52 12/22/06

'd' EXAMINATION ANSWER K Question 1 Details \ Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-50026-6A-001 Given the following:

Unit 2 was initially operating at 100% power 1360 Active No No 3.00 2 7 .oo 295003AA2.02 N-ILT-5002B-6A-001 0.00 0.00 Importance: RO N/A / SRO 4.3 Cognitive-Level:

High

References:

SE-11 Attachment C; T-I01 THE EXAMINEE WILL NEED SE-11 ATTACHMENT C IN ORDER TO ANSWER THIS QUESTION.

This question salifies the requirement of I OCFR55.43(b)(5)

Justification:

A is incorrect - actual level is approximately

-40 inches. B is incorrect - actual level is approximately

-40 inches. C is incorrect - level is approximately

-40 inches. D is correct - according to SE-11, Attachment C, level indicators LI-94A (narrow range) and LI-85B (wide range) are DC powered and are therefore available during a station blackout. LR-110A and B are not available. However, the only accurate level indication in this case is LI-85B since LI-94A's lowest indication available is 0 inches. Therefore actual level is between 0 inches and TAF (-172"O at -40 inches. The correct action to take is to restore RPV level to between +5 and +35 inches per T-I 01. 'i, NRC EXAM Page: 2 of 52 12/22/06

- 2 ID: N-ILT-1555-1-015 Given the following: Unit 2 was initially operating at 100% power.

A loss of all off-site power occurred. Diesel Generator E-I failed to start. All control rods are fully inserted.

RPV level is -1 0 inches and steady. Reactor pressure is 950 psig. 2A DC POWER PANEL LO VOLTAGE (209 C-3) is in alarm. 2A DC Bus voltage at Panel 20C021 (CSR) is 90 VDC. What actions are required for these conditions?

A. Enter SE-13, "Loss of a 125 or 250 VDC Safety Related Bus" B. Restart the 2A CRD Pump in accordance with SO 3.1.-2,"CRD Hydraulic System Startup with the System Filled and Vented".

C. Place the alternate 2A battery charger in service in accordance with SO 57B.1-2, "1 251250 Volt Station Battery Charger Operations".

D. Transfer the 2A battery charger power source from E-124-T-B to E-I 34-T-B in accordance with A0 57B.6-2,"Transfer of 125V Battery Charger 2AD003 to Alternate Power and Return to Normal". Answer: A NRC EXAM Page: 3 of 52 12/22/06 ION ANS 2007 NRC SRO Rev 0 w Question 2 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1555-1-015 Given the following: *Unit 2 was initially operating at 100% power.

A loss of 1305 Active No No 3.00 3 1 .oo 295004AA2.03 N-ILT-1555-1-015 0.00 0.00 Importance: RO 2.8 I SRO 2.9 Cognitive-Level:

High

References:

ARC 209 C-3, SE-13, A0 57B.6-2 This question satisfies the requirements of 1 OCFR55.43(b)(5), Justification:

A. Correct - This is an SE-13 entry condition

... the referenced alarm and voltage on a safety- related 125 VDC distribution panel less than 107.45 VDC requires entry into SE-13.

Incorrect - Cannot start the 2A CRD Pump due to no power available to the E-I2 bus. Incorrect - Both the normal and alternate supply to the battery charger come from the same source.

Incorrect - this evolution can only be done when in MODE 4 or 5, as specified in A0 57B.6-2, Prerequisite 2.1. B. C. D. L' NRC EXAM Page: 4 of 52 12/22/06

7 NRC SRO Rev Question f Details +-. Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-5034-4B-007 *Unit 2 is operating at 100% rated power. *The 2A TBCCW pump blocked for maintena 1306 Active No No 3.00 3 1 .oo 29501 8AA2.04 N-ILT-5034-4B-007 0.00 0.00 Importance:

RO 2.9 / SRO 2.9 Cognitive-Level: High

References:

ON-I 13 This question satisfies the requirements of IOCFR55.43(b)(5).

Justification:

A. Correct - The RBCCW backup does not occur on low pressure. This swap requires that both 480 VAC MCCs are tripped. The CRD pumps are normally supplied by TBCCW. Incorrect - The RBCCW backup does not occur on low pressure. Even if the swap did occur, RBCCW would not supply lsophase Bus Coolers. Incorrect - The RBCCW backup does not occur on low pressure.

Incorrect - TBCCW no longer supplies cooling water to the Condensate Filter Demineralizer Hold pumps. B. C. D. .-.- NRC EXAM Page: 6 of 52 12/22/06 EXAMINATION ANS 2007 NRC SRO Rev 0 .4 4 ID: N-ILT-1550-22C:-OOl

* * *

Unit 3 is operating at 100% power. An explosion ruptures several Instrument Air lines in the turbine building.

All available air compressors are running.

Instrument Air pressure lowers toward 0 psig. Control rods begin to drift in. For the above conditions, per ON-I 19 "Loss of Instrument Air", the crew must enter (1) and use (2) to control RPV pressure and (3) to control RPV level. A. (1) T-100, "Scram". (2) SRVs/HPCI.

(3) HPCVRCIC.

B. (1) T-101, "RPV Control".

(2) SRVs/HPCI.

(3) Feedwater.

C. (1) T-100, "Scram" (2) Bypass Valves.

(3) HPCI/RCIC.

D. (1) T-101, "RPV Control".

(2) Bypass Valves.

(3) Condensate Pumps. Answer: A L- NRC EXAM Page: 7 of 52 12/22/06

- Question 4 Details , Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Va I ue: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1550-22C-001 *Unit 3 is operating at 100% power. *An explosion ruptures several Instrument 1308 Active No No 3.00 3 1 .oo 295019AA2.02 N-ILT-1550-22C-001 0.00 0.00 Importance:

RO n/a / SRO 3.7 Cog n itive-Level

High This question satisfies the requirements of 1 OCFR55.43(b)(5).

References:

ON-I 19 Justification:

A. Correct - Outboard MSlVs will go closed on a loss of air, therefore no steam for feed pumps or use of the main condenser for decay heat.

Condensate is available for injection however it is not preferred due to AO-9091, C RFP bypass failed open on loss of air and increasing RPV level in an uncontrolled manner. HPCVRCIC are totally unaffected by loss of air. Incorrect - CRD flow control valves fail closed on a loss of air. Incorrect - Condenser is not available for pressure control due to MSlVs going closed on loss of air. Incorrect - Condenser is not available due to MSlVs going closed on loss of air. B. C. D.

Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete: Point Value:

Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Question 5 Details Multiple Choice N-ILT-1800-2-004 During Refueling Operations RPV water level is required to be 1307 Active No No 3.00 3 1 .oo 295023G2.2.25 N-ILT-1800-2-004 0.00 0.00 Importance:

RO / SRO 3.7 Cognitive-Level: Memory This question satisfies the requirements of 1 OCFR55.43( b)( 2).

References:

Tech Spec 3.9.6 Justification:

A. 6. Incorrect - Correct basis, incorrect applicability.

Also applicable during handling of control rods.

Correct - (1) This is the basis as stated in Tech Spec 3.9.6 Bases. (2) This is consistent with the applicability statement of Tech Spec LCO 3.9.6. Incorrect - Incorrect basis, incorrect applicability.

Incorrect - Incorrect basis, correct applicability.

C. D. .~ -.- NRC EXAM Page: 10 of 52 12/22/06 EXAMIN 6 ID: N-ILT-G6-8-001 Points: 1.00 ;J Which of the following events result in Emergency Classification per EP-AA-1007, "Radiological Emergency Plan Annex for Peach Bottom Atomic Power Station", and subsequently requires notification to outside agencies?

Loss of Drywell cooling and Drywell temperature above 28OoF. 125 VDC Battery 28 at 104 volts for 30 minutes while in MODE 4. A. B. C. A leak at the Fuel Pool Cooling pump suction causing a Skimmer Surge Tank low-level alarm. D. Inability to maintain RPV pressure and Torus temperature below the HCTL curve while in MODE 3. Answer: D 'L-: NRC EXAM Page: 11 of 52 12/22/06 Question 6 Details - Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value:

Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-G6-8-001 Which of the following events require notification to State and Local authorities 1309 Active No No 2.50 3 1 .oo 295026G2.4.30 N-l LT-G6-8-001 0.00 0.00 Importance:

RO n/a / SRO 3.6 Cognitive-Level:

High This question satisfies the requirements of 1 OCFR55.43(b)(5)

References:

T-102, EP-AA-1007, EP-AA-I 14 Justification:

A. B. Incorrect - This does not meet any EAL criteria.

Incorrect - This does not meet any EAL criteria ... MA3 and MS3 specify a loss of ALL required Tech Spec safety-related 125 VDC power sources.

Incorrect - This does not meet any EAL criteria ... MU12 specifies a Skimmer Surge Tank (SST) low-level alarm AND visual observation of an uncontrolled drop in water level below the SST inlet.

Based on the fuel pool design, a leak at the fuel pool cooling pump suction would not cause fuel pool level to drop below the SST inlet (weirs).

Correct - Requires Emergency Blowdown per T-102, Step T/T-10. This requires declaration of an SAE per EAL MS5. C. D. c-, NRC EXAM Page: 12 of 52 12/22/06 ID: N-ILT-2101-1-011 Points: 1.00 u 7 * * * * * *

Unit 2 is operating at 80% reactor power. An electrical problem has resulted in the loss of all Rod Position Indication (RPIS). A few minutes later, a reactor scram occurs due to a low RPV water level transient.

Reactor power is 3.0 E-2%. RPV level lowered to -50 inches and is now +20 inches. RPV pressure is 930 psig being controlled by EHC. Scram header pressure is 0 psig. Based on the above, which one of the following describes the condition of the plant and the procedure required to address the condition?

An ATWS: A. IS in progress. Enter T-100, "Scram", and then Enter T-101, "RPV Control", at RC-1. 8. C. IS in progress. Enter T-101, "RPV Control", and concurrently execute all legs. IS NOT in progress. Enter T-101, "RPV Control", at RC-1 and concurrently execute all legs. D. IS NOT in progress. Enter T-100, "Scram" and concurrently enter GP-3, "Plant Shutdown".

Answer: B ii NRC EXAM Page: 13 of 52 12/22/06 0 US 3tlN LOO , A3)I tll3MSNV NOIlVNIVVW3 3 :JaMSUt/ 0 Aatl OtlS 3tlN LOOZ SNV NOIL N ANSWER KE RC SRO Rev 0 Question 8 Details \J Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value:

Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1540-4-010 Unit 3 was operating at 100%

power when a feedwater level control malfunction cause 1359 N-ILT-1540-4-010 Active No No 3.00 3 1 .oo 295008AA2.0 1 0.00 0.00 Importance:

RO N/A / SRO 3.9 Cognitive-Level:

High This question satisfies the requirements of 1 OCFR55.43(b)(5)

References:

OT-I 02; OT-I 10; T-I01 ----- Justification: A is incorrect - OT-I 10 directs closing the MSlVs if RPV level cannot be maintain below the bottom of the MSlVs (+I08 inches). In addition, while OT-I02 does direct maintaining reactor pressure below 1053 psig, since the reactor is scrammed, OT-I 02 is no longer applicable.

OT-I10 is executed concurrently with T- 101 "RPV Control".

B is incorrect - OT-I IO directs closing the MSlVs if RPV level cannot be maintain below the bottom of the MSlVs (+I08 inches). In addition, while OT-I02 does direct maintaining reactor pressure below 1053 psig, since the reactor is scrammed, OT-102 is no longer applicable.

OT-I10 is executed concurrently with T- 101 "RPV Control".

C is correct - when RPV pressure reaches 1050 psig, OT-I 10, which is executed concurrently with T- I01 ,"RPV Control", directs manual SRV operation using a single SRV (if possible) and prolonged SRV opening . D is incorrect - OT-I 10 directs prolonged SRV opening using a single SRV (or as few as possible) in order to minimize SRV tailpipe loading and the number of SRVs NRC EXAM Page: 16 of 52 12/22/06 INATIO 7 NRC SRO Rev 0 that are effected by higher than normal loads.

9 ID: N-ILT-5007-8407 Points: 1.00 Given the following conditions:

Unit 2 is at 100% power.

The HPCl System is in service per ST-0-023-301-2, "HPCI Pump, Valve, Flow and Unit Cooler Functional and In-Service Test". Torus Cooling is in service per SO IO. 1 ,D-2. Torus bulk averaqe temperature on SPOTMOS TIS-2-2-71A reaches 96OF during testing and the Control Room Supervisor entered T-I 02, "Primary Containment Control".

The Reactor Operator who is recording Torus temperature per ST-0-023-301 -2 observes that exhausting is at 106OF. * *

Torus water temperature at the Torus Bay location where the HPCl turbine is What action(s), if any, idare required?

A. Verify Torus water average temperature 5 1 10°F once per hour ONLY. B. C. Immediately suspend all HPCl testing since it is adding heat to the Torus. Verify Torus water average temperature 5 11 O°F once per hour Restore Torus average temperature to 5 95OF within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

AND D. No additional action required. Torus water temperature will continue to be monitored every 5 minutes while HPCl remains in service.

Answer: D w NRC EXAM Page: 17 of 52 12/22/06

. EXAMINATION ANSWER KEY 2007 NRC SRO Rev Question 9 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5007-8-007 Given the following conditons: *Unit 2 is at 100% power. *ST-0-023-301-2, HPCl 1312 Active No No 3.00 3 1 .oo 2950 1 3AA2.02 N-l LT-5007-8-007 0.00 0.00 Importance: RO 3.2 I SRO 3.5 Cognitive-Level: High This question satisfies the requirements of 10CFR55.43 (b)(2).

References:

ST-0-023-301, TS 3.6.2.1 Justification:

A. Incorrect - This addresses a required action of 3.6.2.1 .A. 1 (> 95OF average Torus temperature) which is NOT applicable due to testing in progress that is adding heat to the Torus. B. Incorrect - This action is driven by TS 3.6.2.1.C for Torus temperature

> 105OF (averaqe tem pe ratu re). Average Torus tern pera tu re presently is 96OF. C. Incorrect - This addresses required actions of 3.6.2.1 .A. 1 and A.2 (> 95OF average Torus temperature) which is NOT applicable due to testing in progress that is adding heat to the Torus. D. Correct - Tech Spec 3.6.2.1 for Torus temperature is concerned with only averaqe water temperature, not local readings.

No action is required by TS 3.6.2.1 since SPOTMOS RIS-2-2-71 A is reading an average of 96OF with testing going on that adds heat to the Torus. The next action level would be at a Torus temperature of 105OF at which all testing would be suspended per T.S. 3.6.2.1 .C and ST-0-023-301

-2. NRC EXAM Page: 18 of 52 12/22/06 AT 2 E '.-' 10 ID: N-IIT-15603-005 Points: 1.00 * *

Unit 3 is at 70% power. Annunciator 323 (E-5) A RHR PUMP ROOM FLOOD is received.

Two minutes later annunciator 326 (A-4) TORUS WATER LEVEL OUT OF NORMAL RANGE is received.

Torus level is 14.2 feet and lowering slowly.

The Equipment Operator sent to investigate the alarms reports back that he cannot get into the 'A RHR room from the 116' El. due to the door latch not releasing. Which of the following best describes the cause of the above conditions and what are the required procedure actions?

*

A. RHR Pump A suction line break ONLY Monitor Torus level. Declare an Alert. *

B. RHR Pump A and/or RHR Pump C suction line break. Immediately perform a manual scram using procedure GP-4 AND enter T-I 01, "RPV Control". Declare an Alert. * * *

C. RHR Pump A suction line break ONLY. Enter T-I 02,"Primary Containment Control". Restore Torus water level to normal, or if not possible, commence a GP- 3 shutdown.

D. RHR Pump A and/or RHR Pump C suction line break. Restore Torus water level to normal using RClC minimum flow line, or if not possible, commence a GP-3 shutdown. Enter Tech Spec 3.6.2.2 for Suppression Pool Water Level.

Answer: C

~. ....-' NRC EXAM Page: 19 of 52 12/22/06 02 I1 E RHR pump suction piping, not the C RHR pump suction.

The A and C RHR Rooms are separated at lower elevations by a water tight door, preventing cross flooding from one room to the other. The RClC System would not be a system used to restore Torus level.

T-l02step TIL-5 directs using either HPCI, Condensate transfer, or HPSW systems. Le- -*; NRC EXAM Page: 21 of 52 12/22/06 EXAMINATION ANSWER Rev 0 11 ID: N-ILT-5023-6D-001 Points: 1.00 Unit 3 conditions are as follows: 0 0 A scram occurred due to a primary system leak into the drywell.

HPCl is being used to maintain RPV level at approximately

+IO inches. A large leak occurs on the Unit 3 CST and present level is 4' 6". The PRO notifies the Control Room Supervisor that the HPCl System valve lineup is unchanqed. Based on the above conditions the Control Room Supervisor needs to direct what actions, if any? A. The PRO should continue monitoring CST level and take no other action at this time. CST level is adequate to support HPCl System operation.

B. CST level is to be recovered using SO 27.1 .A, "Condensate Transfer and Storage System Startup and Normal Operation".

C. HPCl System suction is to be transferred manually using SO 23.7.B-3, "Transfer of HPCl Suction From CST to Torus".

D. The HPCl System is to be isolated and RCIC is to be placed into service for RPV level control using RRC 13.1-3, "RCIC System Operation During a Plant Event." Answer: C \---, NRC EXAM Page: 22 of 52 12/22/06 Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5023-6D-001 Unit 3 conditions are as follows: A scram occurred due to a primary system leak in 1346 Active No No 3.00 4 I .oo 206000A2.09 N-ILT-5023-6D-001 0.00 0.00 Importance: RO 3.5 / SRO 3.7 Cog n i tive-Leve I: High This question satisfies the requirements of 1 OCFR55.43(b)(5).

References:

ARC 321 C-3; SO 23.7.8-3 Justification:

A. Incorrect - Action to swap HPCI suction to the Torus is given in ARC 321 C-3. B. Incorrect - Condensate transfer system takes suction from the CST and does not have the capability to normally makeup to the CST. C. Correct - ARC 321 C-3 provides the guidance that on Low-Low CST level he operator is to verify that HPCl suction valves automatically swap from the CST to the Torus. If no automatic function occurs, the operator is to perform SO 23.7.8-3 to manually swap suction to the Torus. D. Incorrect - RClC also takes normal suction from the CST and would be affected by the Low-Low CST level condition. With HPCI already in service and capable of maintaining RPV level, there is no need to place RCIC in service at this time.

NRC EXAM Page: 23 of 52 12/22/06

-d ID: N-ILT-1540-5405 0 0 Unit 2 is operating at 100% power when a fuel failure occurs. Main steam line radiation levels on Panel 20C010 are reading 6.0 R/hr and rising quickly. Based on the above condition what is the effect on the plant and what actions need to be taken? A. An automatic scram should have already occurred Perform GP-4 "Manual Reactor Scram". Declare an Unusual Event.

B. A Group I isolation will occur at approximately 15R/hr Scram and enter T-100, "Scram". Declare an Unusual Event.

C. An automatic scram will occur at approximately 10 Rlhr Enter OT-103, Main Steam Line High Radiation".

Perform GP-4, "Manual Reactor Scram", D. An automatic scram should have already occurred. Perform GP-4, "Manual Reactor Scram". Close the MSlVs in accordance with OT-103,"Main Steam Line High Radiation". Answer: C NRC EXAM Page: 24 of 52 12/22/06 INATIO 2007 N Question 12 Details Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Poi n t Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1540-5-005 Unit 2 was operating.at 100% power when a gross fuel failure occurred. Main steam 1314 Active No No 3.00 3 1 .oo 2 12000A2.17 N-ILT-1540-5-005 0.00 0.00 Importance:

RO n/a / SRO 4.2 Cognitive-Level: High This question satisfies the requirements of 1 OCFR55.43(b)(5).

References:

OT-I 03, EP-AA-1007 Justification:

A. Incorrect - MSL radiation levels have not yet risen to the scram setpoint of lOxNFPB, which is -1 0,000 mR/hr. At this time it would be prudent to perform GP-4 and manually scram. However, 10xNFP6, which is -10,000 mR/hr is also the threshold value for declaring an Unusual Event IAW EAL RU3. Incorrect - MSL radiation levels for the Group I isolation setpoint is 1 OxNFPB, which is -10,000 mRlhr. This is also the threshold value for declaring an Unusual Event IAW EAL RU3. Correct -When MSL radiation levels reach (or are expected to each) 8000mR/hr, OT-103 directs a manual scram IAW GP-4. Incorrect - OT-103 does not require closing the MSlVs and in fact directs actions to prevent the Group I isolation and loss of the main condenser as a heat sink

... per OT-103 bases: a GP-4 scram at 8000 mR/hr is directed to minimize the pressure transient caused by the MSlV closure that will occur if radiation levels continue to rise; in addition, performing the GP- 4 scram may cause MSL radiation levels to remain below the Group I isolation setpoint, thereby maintaining the main condenser as a heat sink. B. C. D. NRC EXAM Page: 25 of 52 12/22/06 13 ID: N-ILT-5001G$8-003 Points: 1.00 *

Unit 2 is operating at 100% power. On Day 1 annunciator 227 B-5 BLOWDOWN RELIEF VALVES BELLOWS LEAKING alarms for a failed bellows on the 'G' Safety Relief Valve. On Day 5 a failed bellows is detected on the 'D' Safety Relief Valve.

Which Technical Specification Required Actions listed below applies to the above operational conditions, if any? A. No Technical Specification Required Actions apply.

B. Restore the 'G' SRV to operable status within 14 days. C. Place the Reactor in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and be in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . D. Place the Reactor in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce Reactor Steam Dome pressure to 5 100 psig within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Answer: A --' NRC EXAM Page: 26 of 52 12/22/06

ID: N-ILT-5006-8406 Points: I.Ud -4 14 Unit 2 is in MODE 1 at 20% power. It has been identified that the Main Turbine and the Reactor Feed Pump Turbine (RFPT) high RPV water level trips from the Digital Feedwater Control System (DFCS) computer DCC-X are set at +51 inches. What actions, if any, are required for the above condition and why?

A. Place the channel in trip within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ONLY, based on one RPV high water level trip channel being inoperable.

B. Restore DFCS RPV high water level trip capability within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND place the channel in trip within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , based on RPV high water level trip capability NOT being maintained.

C. No actions required.

The wide range signals which trip HPCl and RClC provide adequate trip capability for the Main Turbine and the RFPTs.

D. No actions required. The DFCS high RPV water level trips for the Main Turbine and the RFPTs are not required to be operable at this time. However, reactor thermal power must remain

<25%. Answer: D .. -' - NRC EXAM Page: 28 of 52 12/22/06 NS Rev 0 XA Ll Question 14 Details Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1

User Number 2: Comment: Multiple Choice N-ILT-5006-8-006 Unit 2 is in MODE 1 at 20% power. It has been identified that the Main Turbine and 1361 Active No No 3.00 3 1 .oo 259002 G2.2.22 N-ILT-5006-8-006 0.00 0.00 Importance:

RO 3.4 / SRO 4.1 Cog n i tive-Level:

High

References:

Tech Spec 3.3.2.2 and Bases This question satisfies the requirements of 1 OCFR55,43(b)(2).

Justification:

A. Incorrect.

No actions are required due to RTP being

~25%. This would be the required action per Spec. 3.3.2.2.A if power were 225% and one or more DFCS high water level trip channels were inoperable.

B. Incorrect. For the conditions given, this would be the correct Tech. Spec.

action to enter per 3.3.2.2.B if power were 2 25%. No actions are required due to RTP being <25%. C. Incorrect. While the RPV level inputs from RCIC would be sufficient to still trip the Main Turbine, and the inputs from HPCl would be sufficient to still trip the RFPTs, Tech. Spec.

3.3.2.2 requires the 2 channels per trip system from the DFCS to be operable. No actions are required due to RTP being <25%. D. Correct. While the trips out of the DCC-X computer are inoperable

(> +49"), there are no actions required due to RTP being <25%. Spec. 3.3.2.2 applicability requires 2 channels per trip system to be operable 2 25% power to ensure that the fuel clad integrity Safety Limit and the cladding 1% plastic strain limit is not violated during the feedwater controller failure, max.

demand event. NRC EW\M Page: 29 of 52 12/22/06 15 , ' ID: N-ILT-5057-8-002

,. Points: 1.00 * *

Unit 2 is at 100% power. Battery charger 2AD003 was placed on EQUALIZE 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ago.

An Equipment Operator performing rounds identifies that Battery Room exhaust fan OAV36 tripped and OBV36 failed to auto start The Equipment Operator placed OBV36 in service satisfactorily. There were no Main Control Room or local panel annunciators received.

What actions, if any, need to be taken and why?

A. No actions are required as long as Turbine Building ventilation remains in service. Create an issue for the deficiency and monitor the Battery Rooms for high temperature conditions.

Due to the potential for a buildup of moisture on the batteries prepare a plan for return of the air flow detector to operable within 14 days assign a responsible person to ensure the plan is completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Due to the potential buildup of combustible gases verify the operability of the Battery Room Ventilation Exhaust System every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B. AND C. AND restore the air flow detector to operable within 14 days. D. Due to the potential buildup of combustible gases verify the operability of the Battery Room Ventilation Exhaust System every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND restore the air flow detector to operable within 14 days Answer: D

,, L-' NRC EXAM Page: 30 of 52 12/22/06 Question Type: Topic: System ID:

User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Question 15 Details Multiple Choice N-ILT-5057-8-002 *Unit 2 is at 100% power. *Battery charger 2AD003 has been placed in the equalize 1317 Active No No 3.00 3 1 .oo 263000A2.02 N-l LT-5057-8-002 0.00 0.00 Importance:

RO n/a / SRO 2.9 Cognitive-Level:

High This question satisfies the requirements of 1 OCFR55.43(b)(2).

References:

TRM 3.14.10 Justification:

A. Incorrect - Actions for Battery Room Ventilation exhaust air flow detector not being functional are required by TRM 3.14.10. Incorrect - This answer does not address all required actions of TRM 3.14.10.A and the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time frame is not correct. Should be 14 days vs. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Moisture buildup is not an issue.

Incorrect - 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months completion time is not correct. Should be 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Correct - These actions satisfy the requirements of TRM 3.14.1 O.A. B. C. D. .L-.-' Page: 31 of 52 12/22/06 NRC EXAM

,' , 16 ID: N-ILT4002-6G-003 Points: 1.00 Given the following:

Unit 2 is operating 100% power. The Total Feed Flow signal produced by Feedwater Level Control fails to "zero". The OPRM TRIP ENABLED (21 1 B-3) annunciator does NOT alarm following the transient.

Based on the above, Recirculation Pumps will (1) and the crew must initially (2) : A. (1) Runback to 30%. (2) Perform SO 2.7.A-2, "Resetting Recirculation System Upper and Lower Flow Limits". B. (1) Runback to 45%. (2) Determine current operating point on Exhibit GP-5-2, "Power Flow Operation Map"; monitor for THI. C. (1) Runback to 30%. (2) Perform A0 60A.1-2, "Alternate Method to Detect and Suppress Thermal Hydraulic Instability (THI)".

., 'v' D. (1) Runback to 45%. (2) Perform A0 60A.1-2, "Alternate Method to Detect and Suppress Thermal Hydraulic Instability (THI)".

Answer: C -u NRC EXAM Page: 32 of 52 12/22/06 TlON A .u Question 16 Details Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice: Diff icu Ity: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5002-6G-003 Given the following: *Unit 2 is operating at 100% power. *The Total Feed Flow 1318 Active No No 3.00 4 1 .oo 202002A2.07 N-l LT-5002-6G-003 0.00 0.00 Importance:

RO nla / SRO 3.3 Cognitive-Level:

High This question satisfies the requirements of 1 OCFR55.43(b)(5).

References:

ARC 21 1 B-3, OT-I 12 Justification:

A. Incorrect - The given conditions indicate the OPRM System is inoperable, requiring performance of A0 60A.1-2. Performing a recirc. system runback signal reset is not an initial priority action.

Incorrect - < 20% total feedwater flow causes a Recirc runback to 30%. Recirc runback to 30%. This would result in the OPRM trip output to RPS to be enabled when recirc flow goes below -60 percent while APRM STP is above - 30 percent. The OPRM TRIP ENABLED annunciator would be expected to alarm during/following the runback. If this annunciator does not alarm, the OPRM System would be assumed to be inoperable, in which case OT-I 12 directs performing A0 Incorrect - < 20% total feedwater flow causes a Recirc runback to 30%. B. C. Correct - 20% total feedwater flow causes a 60A. 1 . -2. D. L-' NRC EXAM Page: 33 of 52 12/22/06 i_/ 17 ID: N-lLT6001B-6A-001 Points: Id0 A Unit 2 startup is in progress with the following plant conditions:

*

Reactor power is 25%. Generator output is 200 MWe. Annunciator TURBINE STOP V. CLOSURE

& CONTROL VLV FAST CLOSURE SCRAM BYPASS (210 A-2) is lit. A relay failure causes the Power-to-Load Unbalance lockout to actuate.

The POWER LOAD UNBALANCE TRIP (206 B-I) annunciator goes into alarm.

Which of the following describes (1) the plant response and (2) the correct procedural direction for this event?

A. (1) Reactor scram ONLY. (2) Enter T-100, "Scram". B. (1) Generator lockout and turbine trip ONLY. (2) Halt GP-2 "Startup" C. (1) Generator lockout, turbine trip and reactor scram. (2) Enter T-I 00, "Scram". (1) The turbine remains online; the reactor does NOT scram. (2) Perform applicable sections of SO 1 B.2.A-2, "Main Turbine Generator Shutdown".

D. Answer: B \-' NRC EXAM Page: 34 of 52 12/22/06

.W' Question 17 Details Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete: Point Value:

Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5001 B-6A-001 A Unit 2 startup is in progress with the following plant conditions:

Reactor 1319 Active No No 3.00 3 1 .oo 245000A2.05 N-ILT-5001 B-6A-001 0.00 0.00 Importance:

RO nla / SRO 3.8 Cognitive-Level:

High This question satisfies the requirements of IOCFR55,43(b)(5).

References:

GP-2, ARC 206 B-I , TS Bases 3.3.1 .I Justification:

A. B. Incorrect - The reactor does not automatically scram. Correct - If the PLU circuit energizes, a generator lockout and turbine trip will occur.

Since reactor power is < 29.5% RTP (turbine 1 st stage pressure is

< 138.4 psig), a reactor scram will not occur as a result of the TSV/TCV closure. The turbine bypass valves will rapidly open, preventing a scram from high reactor pressure/neutron flux. The end result will be the reactor at 25% power with the turbine-generator off-line, which would necessitate halting progress on the startup per GP-2.

Incorrect - The reactor does not automatically scram. Incorrect - The PLU circuit will produce a generator lockoutlturbine trip.

C. D. ., -+-- NRC EXAM Page: 35 of 52 12/22/06 1 EXAMINA 18 ID: N-ILTS009-8-001 Porn&: 1.00' 'U' The Technical vacuum water

-(2)-. Specification leakage limit for Secondary Containment is

-(I)- cfm at -.25' of gauge and is based on maintaining Secondary Containment operability during a A. (1) 9,000 (2) fuel handling accident B. (1) 10,5000 (2) loss of coolant accident C. (1) 9,000 (2) control rod drop accident D. (1) 10,500 (2) steam line break accident Answer: B ., .-., NRC EXAM Page: 36 of 52 12/22/06 4 Question 18 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5009-8-001 The Technical Specification in- leakage limit for Secondary Containment is -(I)- 1320 Active No No 3.00 2 1 .oo 290001G2.1.32 N-l LT-5009-8-00 1 0.00 0.00 Importance: RO n/a / SRO 3.8 Cognitive-Level: Memory This question satisfies the requirements of 1 OCFR55,43(b)(2).

References:

GP-16, TS Bases 3.6.4.1 Justification:

A. B. Incorrect - 9,000 cfm is the administrative limit (GP-16); the Tech Spec limit is 10,500 cfm. Correct - 10,500 cfm is the correct Tech Spec limit. There are two accidents that take credit for Secondary Containment operability:

loss of coolant accident and fuel handling accident.

Incorrect - 9,000 cfm is the administrative limit (GP-16); the Tech Spec limit is 10,500 cfm. No credit is taken for Secondary Containment during a control rod drop accident.

Incorrect - No credit is taken for Secondary Containment during a steam line break accident.

C. D. C' NRC EXAM Page: 37 of 52 12/22/06

Question 19 DeGils Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1855-4-002 Which one of the following identifies Work Execution Centerwork Control Center Sup 1355 N-ILT-1855-4-002 Active No No 3.00 4 1 .oo G2.1.4 0.00 0.00 Importance: R 2.3 / SRO 3.4 Cognitive-Level:

Memory

References:

OP-AA-101-111, OP-PB-101-111 This question satisfies IOCFR5543(b)(2).

Justification:

A. Incorrect - WCS cannot be both STA and IA per B. Incorrect - WCS cannot be both the IA and the NRC Communicator per OP-PB-101-111.

C. Correct - If WCS is the STA then IA function is NOT required per OP-AA-101-111.

D. Incorrect - If WCS is the IA, they support the STA function, the STA is still required. OP-PB-1 01

-1 1 1. k-' NRC EXAM Page: 39 of 52 12/22/06

'-/' 20 ID: N-ILT-1526-3601 Points: 1.00 Given the following conditions:

The Control Room Supervisor (CRS) has delegated reactivity oversight of Unit 3 to a fully qualified Senior Reactor Operator (SRO) during a GP-2 startup. This has been logged in the Unified Control Room Log. During the Unit 3 startup a problem requires entry into T-103, "Secondary Containment Control". Unit 2 is operating at 75% power during this time. * *

Which of the following delineates the responsibility for command and control authority on each of the two Units for these conditions?

A. In accordance with OP-AA-101-1 I 1 ,"Roles and Responsibilities of On-Shift Personnel", the CRS shall retain command and control over both Units at all times. B. In accordance with OP-AA-103-102, "Watchstanding Practices", the Unit 3 reactivity SRO retains command control over Unit 3 until an emergency no longer exists. The CRS retains command and control over Unit 2. C. In accordance with OP-AA-101-111 ,"Roles and Responsibilities of On-Shift Personnel", the Shift Manager shall assume command and control over both Units upon his arrival in the Main Control Room.

D. In accordance with OP-AA-103-102, "Watchstanding Practices", the Unit 3 reactivity SRO immediately transfers Unit 3 command and control to the Shift Manager and provides support and backup to the CRS on both Units.

Answer: A ., --- NRC EXAM Page: 40 of 52 12/22/06 ION ANS 2007 NRC SRO Rev 0 Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1526-3-001 Given the following conditions: *The Control Room Supervisor (CRS) has delegated 1325 Active No No 2.50 3 1 .oo G2.1.6 N-ILT-1526-3-001 0.00 0.00 Importance:

RO 2.1 / SRO 4.3 Cog n itive-Level: Memory This question satisfies the requirements of 1 OCFR.55.43(b)(5).

References:

OP-AA-101-111 Justification:

A. Correct - OP-AA-101-111, "Roles and Responsibilities of On-Shift Personnel", section 4.2 states "During a transient, the Unit Supervisor (CRS) will immediately position himself as the control authority for the unit, acknowledging immediate operator actions being verbalized and taken by the Reactor Operators. After assessing the situation, the Unit Supervisor (CRS) will direct subsequent operator actions in accordance with applicable procedures until conditions are stable and a transient condition longer exists (specifically step 4.2. IO).

Incorrect - The qualified SRO never officially turned over to the Unit Supervisor (CRS). Incorrect - It is the Unit Supervisor (CRS) who maintains command and control per OP-AA- 101-1 11, not the Shift Manager. Incorrect - Neither the qualified SRO or the Shift Manager obtained command and control.

The Unit Supervisor (CRS) always had command and control.

B. C. D. -\--- NRC EXAM Page: 41 of 52 12/22/06

'4 ID: N-ILT-5010~0-064 0 0 0 0 0 Unit 2 is shutdown with all control rods fully inserted. RPV level is +25 inches. The 2A RHR Pump is running in Shutdown Cooling at 9,000 gpm flow per SO 10.1 .B-2 "RHR System Shutdown Cooling Mode Manual Start". Maintenance requests to locally set the open and closed limit switches on MO 2-10-16A "RHR Pump Min. Flow Valve". This action will require the MO 2-10-16A to be taken off of its closed seat. Authorization to perform the work (1) be granted based on (2)

A. (I) CAN (2) no plant impact if the 2A RHR Pump ONLY is first secured per SO 10.l.B-2 B. (I) CAN (2) no plant impact if the 2A RHR Pump is secured and the MO-2-10-25A "Inboard Discharge Valve" is closed first per SO 10.l.B-2.

C. (1) CANNOT (2) a PClS Group II Shutdown Cooling isolation would occur requiring entry into ON-125 "Loss or Unavailability of Shutdown Cooling".

D. (1) CANNOT (2) the 2A RHR Pump would trip on overcurrent due to excessive pump flow (pump runout) and ON-125 "Loss or Unavailability of Shutdown Cooling" would be entered. Answer: C 'L/ NRC EXAM Page: 42 of 52 12/22/06 Question 21 Details Question Topic: Type: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-ILT-5010-40-004 Unit 2 is shutdown with the 2A RHR Pump running in Shutdown Cooling at 9,000 gpm f 1362 N-ILT-5010-40-004 Active No No 2.50 2 1 .oo 205000G2.2.18 0.00 0.00 Importance:

RO 2.3 1 SRO 3.6 Cognitive-Level: Memory This question satisfies the requirements of 1 OCFR55.43(b)(5).

References:

SO 10.1 .B-2, ON-125 Justification: A. Incorrect - The work cannot occur due to reactor water being diverted to the Torus if the min. flow valve MO-16A is opened. B. Incorrect - The work cannot occur due to reactor water being diverted to the Torus if the min. flow valve MO-16A is opened. Also, to avoid diverting flow to the Torus either the SDC suction MO-17, MO-18, or MO- 15A would need to be closed. C. Correct - The work cannot occur due to reactor water would lower due to being diverted to the Torus if the min. flow valve MO-16A is opened.

To prevent this from occurring, the min. flow valve for the RHR pump is shutdown cooling is procedurally controlled closed with its feed removed during shutdown cooling operation.

ON-I 25 would have to be entered once the PClS Group II isolation occurred since shutdown cooling would become unavailable.

D. Incorrect - Even with the min. flow valve fully open during pump operation the RHR pump total flow would not exceed pump runout flow of >12, 500 gpm. NRC EXAM Page: 43 of 52 12/22/06 TlON ANSWER 22 ID: N-NLSRO-0763-2-001 Points: 1.00 Unit 2 is in a Refueling Outage. In accordance with FH-GC, "Core Component Movement-Core Transfers", which one of the following Refuel Floor activities CAN ONLY be DIRECTLY supervised by Senior Reactor Operator (SRO) or a Limited SRO? A. Cleaning recirc jet pumps in the Vessel.

B. Loading a new fuel bundle into the Vessel. C. Moving old LPRM strings to the Spent Fuel Pool. D. Shuffling of irradiated fuel in the Spent Fuel Pool. Answer: B LJ' NRC EXAM Page: 44 of 52 12/22/06 EXAMINATION A 2007 NRC SRO Question 22 Details Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text: User Number 1: User Number 2: Comment: Multiple Choice N-NLSRO-0763-2-00-1 Unit 2 is in a Refueling Outage. Which one of the following Refuel Floor activit 1327 Active No No 3.00 3 1 .oo G2.2.29 N-NLSRO-0763-2-001 0.00 0.00 Importance:

RO 1.6 / SRO 3.8 Cognitive-Level: Memory This question satisfies the requirements of I OCFR55.43( b)(6).

References:

FH-6C Section 7.0 Justification:

A. Incorrect - Cleaning jet pump may be supervised by the Designated Alternate (DA). This activity is not considered a Core Alteration.

Correct - New fuel into the core is a Core Alt and requires direct supervision by an SRO or LSRO. Does not require direct supervision of an SRO, LSRO, or DA. This activity is not considered a Core Alteration. Does not require direct supervision of SRO, LSRO, or DA. This activity is not considered a Core Alteration.

B. C. D. ~, .-*' 12/22/06 NRC EXAM Page: 45 of 52 XAMINATION ANS 2007 NRC SRO Rev 0 'I .-' 23 ID: N-ILT-1770-3-004 Poinis: 1.08 \ Equipment Operators need to enter a locked high radiation area to manually operate Primary Containment Isolation Valves in order to satisfy a Tech.Spec. required action. The highest dose rate in the area is 16,000 mRlhr. Per RP-PB-460-1001 WHICH ONE of the following describes the type of Locked High Radiation Area and the highest level of authorization required for issuing the key? Tvpe of LHRA A. Level 1 Hiqhest Authorization Required Radiation Protection Manager B. Level 1 Plant Manager C. Level 2 Radiation Protection Manager D. Level 2 Plant Manager Answer: C -L-- NRC EXAM Page: 46 of 52 12/22/06 E U' Question 23 Details Question Type: Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty: Time to Complete:

Point Value: Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1770-3-004 The Main Control Room has been abandoned due to a fire. Equipment Operators are goi 1356 Active No No 4.00 N-l LT-1770-3-004 3 1 .oo G2.3.1 0.00 0.00 Importance:

RO 2.6 / SRO 3.0 Cognitive-Level: Memory

References:

RP-AA-460:

RP-PB-460-1001 This question satisfies the requirement of 1 OCFR55.42( b)(4). Justification:

A. Incorrect - The level in incorrect. The area is a Level 2 (21 5Rlhr) which requires authorization from the RP Manager for issuing the key. B. Incorrect - The level is incorrect, and the Plant Manager's authorization is NOT required.

C. Correct - Per RP-AA-460-1001Level2 LHRA is an area with dose rates 2 15R/hr. The RP Manager must provide authorization for this entry.

D. Incorrect -While the level is correct, the RP Manager must provide authorization for this entry.

NRC EXAM Page: 47 of 52 12/22/06 NSW Rev 0 u' 24 ID: N-ILT-2117-5A-006 The following conditions exist following a GP-4 manual scram:

Reactor Power

-=1.00 EO%

RPV Pressure

RPV Level

Drywell Pressure

Scram Air Header Pressure

16 Control Rods are at position 'IO'.

12 Control Rods are at position '06'.

All other control rods are fully inserted.

930 psig and dropping +I 0 inches and rising slowly 2.2 psig and rising slowly 0 psig For the above conditions automatic initiation of the Automatic Depressurization System (ADS) is inhibited per (1) to prevent (2) . A. (1) T-101, "RPV Control" (2) exceeding 11 O°F Torus temperature before boron is injected.

B. (1) T-101, "RPV Control" (2) potential loss of, or inaccuracies in, RPV level instrumentation.

C. (1) T-I 17, "Level/Power Control" (2) core damage due to large irregular neutron flux oscillations.

D. (1 ) T-I 17, "LeveVPower Control" (2) substantial fuel damage due to a large reactor power excursion.

Answer: D NRC EXAM Page: 48 of 52 12/22/06

.*Ll Question 24 Details Question Type:

Topic: System ID: User ID: Status: Always select on test: Authorized for practice:

Difficulty:

Time to Complete: Point Value:

Cross

Reference:

User Text: User Number 1 : User Number 2: Comment: Multiple Choice N-ILT-2117-5A-006 During an ATWS, automatic initiation of the Automatic Depressurization System 1329 N-ILT-2117-5A-006 Active No No 3.00 2 1 .oo G2.4.22 0.00 0.00 Importance:

RO n/a / SRO 4.0 Cognitive-Level:

Memory This question satisfies the requirements of 1 OCF55.43(b)(5).

References:

T-I 17 Bases Justification

A. Incorrect - During an ATWS Torus temperature may exceed 11 O°F before boron injection anyway due to SRV operation

... this is not the reason for inhibiting ADS. Incorrect - Depressurization due to ADS initiation must also be accompanied by elevated Drywell temperature for this to occur ... this is the reason for inhibiting ADS. Incorrect - ADS initiation would not cause large irregular neutron flux oscillations

... it would cause a rapid reduction in reactor power due to voids. Correct - For the given conditions an ATWS exists due to control rods remaining out beyond position '02'.

From T-I 17 Bases: ADS initiation would complicate efforts to maintain RPV level within required level ranges, FURTHER, rapid and uncontrolled injection of large volumes of relatively cold, un-borated water from low pressure injection systems may occur. Wlth the reactor either critical or shutdown on boron along, the positive reactivity addition due to boron dilution and temperature reduction may result in a reactor power excursion large enough to cause substantial fuel damage.

ADS is inhibited to prevent this from hap pen i ng . B. C. D. NRC EXAM Page: 49 of 52 12/22/06 25 ID: N-ILT-1529-1 H-001 .< Points: 1.00 0 The 2D Core Spray Pump Room flood detection LS-2920D was damaged during maintenance activities in the room.

It has been verified that the LS-2920D does not function and will not bring in annunciator 226 (D-5) D CORE SPRAY PUMP ROOM FLOOD. Which of the following describes the actions to be taken for the above condition?

A. Place an equipment deficiency taglsticker on Annunciator 226 (D-5) and designate the annunciator for "Non-Preferred Use" per OP-AA-108-105, "Equipment Deficiency Identification and Documentation".

B. Place an equipment status taglsticker on Annunciator 226 (D-5) and utilize an Abnormal Component Position Sheet (ACPS) to control the abnormal condition per OP-AA-108-101 ,"Control of Equipment and System Status". C. Place an equipment deficiency tagkticker on the associated Alarm Response Card, evaluate the impact on implementing Emergency Operating Procedures, and identify any compensatory actions or additional monitoring per OP-AA-108-105, "Equipment Deficiency Identification and Documentation".

D. Place an equipment status tagkticker on the associated Alarm Response Card, document the deficiency in the Equipment Status Tag Log, and place the annunciator in MANUAL per OP-AA-108-101 ,"Control of Equipment and System Status". Answer: C -. -U- NRC EXAM Page: 50 of 52 I2122106 Question Type: Topic: System ID: User ID: Status: Always select on test:

Authorized for practice:

Difficulty:

Time to Complete: Point Value:

Cross

Reference:

User Text:

User Number 1: User Number 2: Comment: Multiple Choice N-ILT-1529-1H-001 The 2D Core Spray Pump Room flood detection LS-2920D was damaged during maintainen 1348 Active No No 3.00 3 1 .oo G2.4.33 N-ILT-1529-1 H-001 0.00 0.00 Importance:

RO 2.4 I SRO 2.8 Cog n i tive-Level

Memory References

OP-AA-108-105 Equipment Deficiency Identification and Documentation This question satisfies the requirements of 1 OCF R55.4 3( b) (5). Justification:

A. Incorrect - Peach Bottom present practice/management expectation is to place a deficiency sticker on the Alarm Response Card (ARC) associated with the annunciator. Also, use of an Equipment Status Tag (EST) is for identifying TEMPORARY abnormal equipment positioning.

For this example, nothing was placed in an abnormal position in response to the inoperable annunciator. Also, "Non-Preferred Use" is for degraded equipment issues. The Core Spray Room flood alarm system is completely inoperable, not degraded.

B. Use of an Equipment Status Tag (EST) and the Abnormal Component Position Sheet (ACPS) is for identifying TEMPORARY abnormal equipment positioning. For this example, nothing was placed in an abnormal position in response to the inoperable annunciator condition.

C. Correct - OP-AA-108-105, "Equipment Deficiency Identification and Documentation" directs several actions for this condition including evaluating the impact on implementing Emergency Operating Procedures, identifying any compensatory actions or NRC EXAM Page: 51 of 52 12/22/06 T'ION A 2007 NRC SRO Re additional monitoring, and placement of an equipment deficiency taglsticker for the deficient item. Peach Bottom present practicelmanagement expectation is to place a deficiency sticker on the Alarm Response Card (ARC) associated with the annunciator.

D. Peach Bottom present practicelmanagement expectation is to place a deficiency sticker on the Alarm Response Card (ARC) associated with the annunciator. Also, use of an Equipment Status Tag (EST) and the Abnormal Component Position Sheet (ACPS) is for identifying TEMPORARY abnormal equipment positioning.

For this example, nothing was placed in an abnormal position in response to the inoperable annunciator condition. There is no need to place the annunciator in MANUAL and it is not directed by OP- AA-108-101.

12/22/06 NRC EXAM Page: 52 of 52

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2.2.5 owner

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