ML101520530

Initial Exam 2010-301 Draft RO Written Exam

ML101520530
Person / Time
Site:	Oconee
Issue date:	05/04/2010
From:	NRC/RGN-II
To:	Duke Energy Carolinas
References
50-269/10-301, 50-270/10-301, 50-287/10-301
Download: ML101520530 (150)

Text

2010A NRC REACTOR OPERATOR EXAM I POINT Question I Unit 1 initial conditions:

Reactor in MODE 3 RCS temperature = 500°F stable RCS pressure = 885 psig stable Current conditions:

RCS pressure decreasing Pressurizer level decreasing PZR relief valve tailpipe temperature = 300°F Quench tank level increasing Quench tank pressure = 10 psig increasing Based on the above conditions, which ONE of the following describes the reason for the current conditions?

A.

1 B2 RCP Upper, Middle, and Lower seals have failed B.

1 RC-1 59 and I RC-1 60 (RXV Head vents) are leaking C. Low range RCS pressure has failed HIGH D.

1 RC-66 (PORV) is leaking to the Quench Tank 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 1 Unit 1 initial conditions:

Reactor in MODE 3 RCS temperature = 500°F stable RCS pressure = 885 psig stable Current conditions:

RCS pressure decreasing Pressurizer level decreasing PZR relief valve tailpipe temperature = 300°F Quench tank level increasing Quench tank pressure = 10 psig increasing Based on the above conditions, which ONE of the following describes the reason for the current conditions?

A. 1 B2 RCP Upper, Middle, and Lower seals have failed B. 1 RC-159 and 1 RC-160 (RXV Head vents) are leaking C. Low range RCS pressure has failed HIGH D. 1 RC-66 (PORV) is leaking to the Quench Tank

2010A NRC REACTOR OPERATOR EXAM Question I TIIGI -cpw OO8AK1.01, Pressurizer Vapor Space Accident Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident:

Thermodynamics and flow characteristics of open or leaking valves.

(3.2/3.7)

K/A MATCH ANALYSIS Requires knowledge of the throttling process and the operational implications of indications regarding PORV tailpipe temperatures ANSWER CHOICE ANALYSIS Answer: D A. Incorrect, failure of all a RCPs seals would cause seal leakage to increase and thus QT level. A LOCA would also result which would cause RCS pressure to decrease.

PZR level would decrease, however this would not cause the QT to pressurize or relief line temperature to increase.

B. Incorrect, 1 RC-1 59 and 1 RC-1 60 (RXV Head vents) discharge to the RBCUs discharge. Plausible because the manual vents on the hot legs go to the QT.

C. Incorrect, It is isolated above 600 psig. If Low range cooldown pressure were in service and LOW selected on the PORV this failure would cause the PORV to open.

D. Correct, IRC-66 (PORV) failing open would cause these indications.

Technical Reference(s): PNS-PZR Page 34

- 35 Proposed references to be provided to applicants during examination: None Learning Objective: PNS-PZR R19 Question Source: BANK Question History: Last NRC Exam: 2007 Q #31 Question Cognitive Level:

Comprehension or Analysis 2010A NRC REACTOR OPERATOR EXAM Question 1 T1/G1 - cpw 008AK1.01, Pressurizer Vapor Space Accident Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident:

Thermodynamics and flow characteristics of open or leaking valves.

(3.2/3.7)

KIA MATCH ANALYSIS Requires knowledge of the throttling process and the operational implications of indications regarding PORV tailpipe temperatures ANSWER CHOICE ANALYSIS Answer: 0 A. Incorrect, failure of all a RCPs seals would cause seal leakage to increase and thus QT level. A LOCA would also result which would cause RCS pressure to decrease.

PZR level would decrease, however this would not cause the QT to pressurize or relief line temperature to increase.

B. Incorrect, 1RC-159 and 1RC-160 (RXV Head vents) discharge to the RBCUs discharge. Plausible because the manual vents on the hot legs go to the QT.

C. Incorrect, It is isolated above 600 psig. If Low range cooldown pressure were in service and LOW selected on the PORV this failure would cause the PORV to open.

D. Correct, 1 RC-66 (PORV) failing open would cause these indications.

Technical Reference(s): PNS-PZR Page 34 - 35 Proposed references to be provided to applicants during examination: None Learning Objective: PNS-PZR R19 Question Source: BANK Question History: Last NRC Exam: 2007 Q #31 Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 2 Unit 1 plant conditions:

Reactor Trip SBLOCA has occurred Rule 2 (Loss of SCM) in progress 1A & lB Steam Generator levels = 78 XSUR increasing Based on the above conditions, which ONE of the following describes why SG levels are being increased AND the conditions required for RULE 2 to allow throttling EFDW flow?

A. Ensure secondary inventory available for heat transfer if Steam Generator feed is lost / the primary side voids to the point of allowing boiler-condenser heat transfer B. Ensure secondary inventory available for heat transfer if Steam Generator feed is lost / cooldown rates approaching Tech Spec limits C. Ensure boiler-condenser heat transfer is established / the primary side voids to the point of allowing boiler-condenser heat transfer D. Ensure boiler-condenser heat transfer is established / cooldown rates approaching Tech Spec limits 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 2 Unit 1 plant conditions:

Reactor Trip SBLOCA has occurred Rule 2 (Loss of SCM) in progress 1A & 1 B Steam Generator levels = 78" XSUR increasing Based on the above conditions, which ONE of the following describes why SG levels are being increased AND the conditions required for RULE 2 to allow throttling EFDW flow?

A. Ensure secondary inventory available for heat transfer if Steam Generator feed is lost / the primary side voids to the point of allowing boiler-condenser heat transfer B. Ensure secondary inventory available for heat transfer if Steam Generator feed is lost / cooldown rates approaching Tech Spec limits C. Ensure boiler-condenser heat transfer is established / the primary side voids to the point of allowing boiler-condenser heat transfer D. Ensure boiler-condenser heat transfer is established / cooldown rates approaching Tech Spec limits

2010A NRC REACTOR OPERATOR EXAM Question 2 TIIGI -cpw OO9EK1.01 Small Break LOCA Knowledge of the operational implications of the following concepts as they apply to the small break LOCA:

Natural circulation and cooling, including reflux boiling.

(4.2/4.7)

K/A MATCH ANALYSIS Requires knowledge of operational requirement to establish reflux (boiler condenser at ONS) boiling during SBLOCA ANSWER CHOICE ANALYSIS Answer:

D A. Incorrect: First part is plausible since that is a reason SG levels are increased as part of mitigation strategy when in the Turbine Building Flood tab of EOP. Second part is plausible since there is a concern over reaching the LOSCM setpoint prior to primary side voiding down enough to heat transfer. The second part would be correct if the question were about concerns with NOT throttling EFDW flow during approach to LOSCM stpt.

B. Incorrect: First part is plausible since that is a reason SG levels are increased as part of mitigation strategy when in the Turbine Building Flood tab of EOP. Second part is correct.

C. Incorrect: First part is correct. Second part is plausible since there is a concern over reaching the LOSCM setpoint prior to primary side voiding down enough to heat transfer. The second part would be correct if the question were about concerns with NOT throttling EFDW flow during approach to LOSCM stpt.

0. CORRECT: Per EAP-LOSM page 22 & 23

- SG heat removal must be established by feeding SGs up to levels that will promote N/C and BCM heat removal. Flow may be throttled to control cooldown rate within Tech Spec limits, but SG levels must continue to increase until LOSCM setpoint is reached if SCM 0 °F.

Technical Reference(s): EAP-LOSCM Attachment I Rule 2 Proposed references to be provided to applicants during examination:

NONE Learning Objective: EAP-LOSCM R6 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 2 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 009EK1.01 Small Break LOCA Knowledge of the operational implications of the following concepts as they apply to the small break LOCA:

Natural circulation and cooling, including reflux boiling.

(4.2/4.7)

KIA MATCH ANALYSIS Requires knowledge of operational requirement to establish reflux (boiler condenser at ONS) boiling during SBLOCA ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is plausible since that is a reason SG levels are increased as part of mitigation strategy when in the Turbine Building Flood tab of EOP. Second part is plausible since there is a concern over reaching the LOSCM setpoint prior to primary side voiding down enough to heat transfer. The second part would be correct if the question were about concerns with NOT throttling EFDW flow during approach to LOSCM stpt.

B. Incorrect: First part is plausible since that is a reason SG levels are increased as part of mitigation strategy when in the Turbine Building Flood tab of EOP. Second part is correct.

C. Incorrect: First part is correct. Second part is plausible since there is a concern over reaching the LOSCM setpoint prior to primary side voiding down enough to heat transfer. The second part would be correct if the question were about concerns with NOT throttling EFDW flow during approach to LOSCM stpt.

D. CORRECT: Per EAP-LOSM page 22 & 23 - SG heat removal must be established by feeding SGs up to levels that will promote NIC and SCM heat removal. Flow may be throttled to control cooldown rate within Tech Spec limits, but SG levels must continue to increase until LOSCM setpoint is reached if SCM S 0 of.

Technical Reference(s): EAP-LOSCM Attachment 1 Rule 2 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-LOSCM R6 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 3 Unit 1 plant conditions:

Reactor trip from 100% power RCS pressure = 88 psig decreasing Reactor Building pressure 27 psig increasing 1 B LPI Pump failed to start Based on the above conditions, which ONE of the following describes the guidance provided in EOP Enclosure 5.1 (ES Actuation) regarding the LPI pumps and system operation?

EOP Enclosure 5.1 (ES Actuation)....

A. directs continued operation with only the 1A LPI pump and no re-alignment of LPI header flows B. directs continued operation with only the IA LPI pump and manually re-aligns LPI flow down both the IA and lB LPI headers C. utilizes the 1A and the 1C LPI Pump and aligns flow down both headers with ILP-9 and 10 closed D. utilizes only the IC LPI Pump and aligns flow down both headers with 1LP-9 & 10 open 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 3 Unit 1 plant conditions:

Reactor trip from 100% power RCS pressure = 88 psig decreasing Reactor Building pressure = 27 psig increasing 1 B LPI Pump failed to start Based on the above conditions, which ONE of the following describes the guidance provided in EOP Enclosure 5.1 (ES Actuation) regarding the LPI pumps and system operation?

EOP Enclosure 5.1 (ES Actuation)....

A. directs continued operation with only the 1A LPI pump and no re-alignment of LPI header flows B. directs continued operation with only the 1A LPI pump and manually re-aligns LPI flow down both the 1A and 1 B LPI headers C. utilizes the 1A and the 1 C LPI Pump and aligns flow down both headers with 1 LP-9 and 10 closed D. utilizes only the 1 C LPI Pump and aligns flow down both headers with 1 LP-9 & 10 open

2010A NRC REACTOR OPERATOR EXAM Question 3 TI/GI -cpw 011 EK2.02 Large Break LOCA Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA:

Pumps (2.6/2.7)

KIA MATCH ANALYSIS Requires knowledge of operational requirements provided in the EOP as they relate to operation of LPI pumps during a LBLOCA ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: If ES channels 3&4 actuate and RCS pressure is < LPI shutoff head, LPIPs are checked running. If the A or the B LPI pump has failed then direction is given to close that trains injection valve (either LP-17 or 18). The operating pump is left running and aligned to its train only (although LPI will still inject through both nozzles via the crossover piping). With the addition of the LPI crossover mod, it is no longer necessary to align flow down both headers as the crossover piping in the RB ensures flow injected via both LPI nozzles.

B. Incorrect: Plausible since this would be true without credit for the LPI crossover mod.

Without an understanding of the crossover piping, the need to align flow down both headers is a plausible conclusion. Additionally, prior to the addition of the LPI crossover mod actions were required to put flow down both headers.

C. Incorrect: Plausible since aligning the C pump to the B header would replace the flow lost down that header. Additionally plausible since direction to use the C pump is contained in End. 5.1 for other failures.

D. Incorrect: Plausible direction to use the C pump down both LPI headers is contained in End. 5.1 and would be the correct actions if both the A and B LPIPs have failed.

Technical Reference(s): EPIIIAII800IOOI (EOP) End. 5.1 (ES Actuation), EAP-ESA Proposed references to be provided to applicants during examination:

NONE Learning Objective: EAP-ESA R17 Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 3 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 011 EK2.02 Large Break LOCA Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA:

Pumps (2.6/2.7)

KIA MATCH ANALYSIS Requires knowledge of operational requirements provided in the EOP as they relate to operation of LPI pumps during a LBLOCA ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: If ES channels 3&4 actuate and RCS pressure is < LPI shutoff head, LPIP's are checked running. If the A or the B LPI pump has failed then direction is given to close that trains injection valve (either LP-17 or 18). The operating pump is left running and aligned to its train only (although LPI will still inject through both nozzles via the crossover piping). With the addition of the LPI crossover mod, it is no longer necessary to align flow down both headers as the crossover piping in the RB ensures flow injected via both LPI nozzles.

B. Incorrect: Plausible since this would be true without credit for the LPI crossover mod.

Without an understanding of the crossover piping, the need to align flow down both headers is a plausible conclusion. Additionally, prior to the addition of the LPI crossover mod actions were required to put flow down both headers.

C. Incorrect: Plausible since aligning the C pump to the B header would replace the flow lost down that header. Additionally plausible since direction to use the C pump is contained in Encl. 5.1 for other failures.

D. Incorrect: Plausible direction to use the C pump down both LPI headers is contained in Encl. 5.1 and would be the correct actions if both the A and B LPIP's have failed.

Technical Reference(s): EP/1/A11800/001 (EOP) Encl. 5.1 (ES Actuation), EAP-ESA Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-ESA R17 Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 4 Unit 1 plant conditions:

Reactor power = 100%

1TA and 1TB lockout occurs Based on the above conditions, which ONE of the foNowing indicates the initial Thot and Tcold values expected once stable natural circulation flow has been established?

Thot would be approximately and Tcold would be approximately A. 562 I 532 B. 582 /532 C. 585/555 D. 605/555 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 4 Unit 1 plant conditions:

Reactor power = 100%

1 T A and 1 TB lockout occurs Based on the above conditions, which ONE of the following indicates the initial Thot and Tcold values expected once stable natural circulation flow has been established?

Thot would be approximately ___ oF and Tcold would be approximately ___ oF.

A. 562/532 B. 582/532 C. 585/555 D. 605/555

2010A NRC REACTOR OPERATOR EXAM Question 4 TIIGI

- cpw O15AA1.21 RCP Malfunctions Ability to operate and I or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

Development of natural circulation flow (4.4/4.5)

KIA MATCH ANALYSIS Requires knowledge of Thot and Tcold response expected during development of natural circ flow from a low power loss of RC flow. To monitor the development of NC flow requires the ability to predict expected temperature indications that would be indicative of NC flow.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since these values would be correct under the misconception that Tc matched up to Psat for 885 psig SG pressure. Since 885 is normal SG pressure during power ops it would be plausible to assume Tc moves to match that Tsat and then Th increased to develop the 30 40 degree delta T.

B. Incorrect: Plausible since these values would be correct for the same Tc assumptions as in A and then developed a 50 degree delta T. The 50 degree delta T is plausible since that is the normal delta T for 100% power operations.

C. CORRECT: During the transition from forced to natural circulation the cold leg temperatures should remain near the saturation temperature for the existing SG pressure and the hot leg temperatures and CETCs will increase until a stable AT between the hot and cold legs is established, generally at 30-40°F.

Since normal post trip temperature is approximately 555°F and Th would increase to 585°F.

D. Incorrect: These values would be correct if you assume the correct value for Tc however used 50 degrees as the delta T. This is plausible since 50 degree delta T is the normal value for 100% power.

Technical Reference(s):

TA-AMI Proposed references to be provided to applicants during examination:

NONE Learning Objective: TA-AMI R3 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 4 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 015AA 1.21 RCP Malfunctions Ability to operate and I or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

Development of natural circulation flow (4.4/4.5)

KIA MATCH ANALYSIS Requires knowledge of Thot and Tcold response expected during development of natural circ flow from a low power loss of RC flow. To monitor the development of NC flow requires the ability to predict expected temperature indications that would be indicative of NC flow.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since these values would be correct under the misconception that Tc matched up to Psat for 885 psig SG pressure. Since 885 is normal SG pressure during power ops it would be plausible to assume Tc moves to match that Tsat and then Th increased to develop the 30 - 40 degree delta T.

B. Incorrect: Plausible since these values would be correct for the same Tc assumptions as in A and then developed a 50 degree delta T. The 50 degree delta T is plausible since that is the normal delta T for 100% power operations.

C. CORRECT: During the transition from forced to natural circulation the cold leg temperatures should remain near the saturation temperature for the existing SG pressure and the hot leg temperatures and CETCs will increase until a stable llT between the hot and cold legs is established, generally at 30-40°F.

Since normal post trip temperature is approximately 555°F and Th would increase to 585°F.

D. Incorrect: These values would be correct if you assume the correct value for Tc however used 50 degrees as the delta T. This is plausible since 50 degree delta T is the normal value for 100% power.

Technical Reference(s): TA-AM1 Proposed references to be provided to applicants during examination: NONE Learning Objective: TA-AM1 R3 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 5 Unit 1 initial conditions:

Reactor in MODE 5 RCS pressure = 0 psig Normal LPI decay heat removal in service IC LPI pump operating Unit Blackout occurs Current conditions:

MFBs re-energized from CT-5 Based on the above conditions, which ONE of the following describes actions required by AP/26 (Loss of Decay heat Removal)?

A. Start previously running LPI pump AND initiate AP/1 1 (Recovery from Loss of Power)

B. Start previously running LPI pump AND initiate the Blackout tab of the EOP C. Feed and steam SGs to maintain CETC <246°F AND initiate AP/1 1 (Recovery from Loss of Power)

D. Feed and steam SCs to maintain CETC <246°F AND perform AP/26 End. 5.6 (Venting LPI Pumps and Suction Lines) 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 5 Unit 1 initial conditions:

Reactor in MODE 5 RCS pressure = 0 psig Normal LPI decay heat removal in service 1 C LPI pump operating Unit Blackout occurs Current conditions:

MFB's re-energized from CT-5 Based on the above conditions, which ONE of the following describes actions required by AP/26 (Loss of Decay heat Removal)?

A. Start previously running LPI pump AND initiate AP/11 (Recovery from Loss of Power)

B. Start previously running LPI pump AND initiate the Blackout tab of the EOP C. Feed and steam SG's to maintain CETC < 246°F AND initiate AP/11 (Recovery from Loss of Power)

D. Feed and steam SG's to maintain CETC < 246°F AND perform AP/26 Encl. 5.6 (Venting LPI Pumps and Suction Lines)

2010A NRC REACTOR OPERATOR EXAM Question 5 TIIGI -cpw 025AG2.4.11, Loss of Residual Heat Removal System Knowledge of Abnormal Condition Procedures (4.0/4.2)

KIA MATCH ANALYSIS Requires knowledge of mitigation strategy for AP126 (Loss of Decay Heat Removal)

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: AP126 directs ensuring previously running LPI pump is operating.

After pump is running AP directs initiating APIII which will recover from the loss of power to MFBs B. Incorrect: The first part is correct. The second part is plausible since entering the EOP and performing the blackout tab would be the correct actions if the unit were above LPI DHR.

C. Incorrect: First part is plausible since this would be a correct choice if RCS loops are full and would be utilized if LPIPs cannot be restarted. Since RCS pressure = 0 psig, the RCS loops cannot be full. Second part is correct.

D. Incorrect: First part is plausible since this would be a correct choice if RCS loops are full and would be utilized if LPIPs cannot be restarted. Since RCS pressure = 0 psig, the RCS loops cannot be full. Second part is plausible since it is an enclosure performed in the AP and it is reasonable to think that after establishing SG cooling you would be making preps to restart LPI pumps and running this end to ensure the pumps are water solid is plausible.

Technical Reference(s):

APII IAII 7001026, EAP-APG Enclosure AP26 Proposed references to be provided to applicants during examination:

NONE Learning Objective:

EAP-APG R9 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 5 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 025AG2.4.11, Loss of Residual Heat Removal System Knowledge of Abnormal Condition Procedures (4.0/4.2)

KIA MATCH ANALYSIS Requires knowledge of mitigation strategy for AP/26 (Loss of Decay Heat Removal)

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: AP/26 directs ensuring previously running LPI pump is operating.

After pump is running AP directs initiating AP/11 which will recover from the loss of power to MFB's B. Incorrect: The first part is correct. The second part is plausible since entering the EOP and performing the blackout tab would be the correct actions if the unit were above LPI DHR.

C. Incorrect: First part is plausible since this would be a correct choice if RCS loops are full and would be utilized if LPIP's cannot be restarted. Since RCS pressure = 0 psig, the RCS loops cannot be full. Second part is correct.

D. Incorrect: First part is plausible since this would be a correct choice if RCS loops are full and would be utilized if LPIP's cannot be restarted. Since RCS pressure = 0 psig, the RCS loops cannot be full. Second part is plausible since it is an enclosure performed in the AP and it is reasonable to think that after establishing SG COOling you would be making preps to restart LPI pumps and running this encl to ensure the pumps are water solid is plausible.

Technical Reference(s): AP/1/A11700/026, EAP-APG Enclosure AP26 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-APG R9 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 6 Unit 1 plant conditions:

Reactor trip occurred Total RCP seal injection flow = 0 gpm Component Cooling is unavailable Based on the above conditions, which ONE of the following describes the immediate action(s) reguired by the EOP and the reason for the action(s)?

Initiate...

A. AP/14 (Loss of Normal HPI Makeup and/or RCP Seal Injection) to restore RCP seal injection B. AP/16 (Abnormal Reactor Coolant Pump Operation) to secure all RCPs C. AP/20 (Loss of CC) to restore Component Cooling D. AP/25 (SSF EOP) to align an alternate source of seal injection 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 6 Unit 1 plant conditions:

Reactor trip occurred Total RCP seal injection flow = 0 gpm Component Cooling is unavailable Based on the above conditions, which ONE of the following describes the immediate action(s) required by the EOP and the reason for the action(s)?

Initiate...

A. AP/14 (Loss of Normal HPI Makeup and/or RCP Seal Injection) to restore RCP seal injection B. AP/16 (Abnormal Reactor Coolant Pump Operation) to secure all RCP's C. AP/20 (Loss of CC) to restore Component Cooling D. AP/25 (SSF EOP) to align an alternate source of seal injection

2010A NRC REACTOR OPERATOR EXAM Question 6 TIIGI -cpw 026AK3.03 Loss of Component Cooling Water Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water:

Guidance actions contained in EOP for Loss of CCW.

(4.0/4.2)

K/A MATCH ANALYSIS Requires knowledge of reason EOP IMAs direct initiating AP125 when RCP seal injection and CC have been lost ANSWER CHOICE ANALYSIS Answer:

D A. Incorrect: Plausible since the entry conditions for AP/14 are met and the EOP does direct entry into APs in other conditions (Ex. AP/1 1, AP/25). The EOP does not direct actions to restore normal seal injection. Seal injection flow is re-established via the RCMUP since both CC and SI have been lost.

B. Incorrect: Plausible since direction are given to trip RCPs however the directions are part if EOP Immediate Manual Actions. The EOP does not direct entry into AP/16 however the EOP does direct entry into APs in other conditions (Ex. AP/1 1, AP/25).

C. Incorrect: Plausible since the entry conditions for AP/20 are met and the EOP does direct entry into APs in other conditions (Ex. AP/1 1, AP/25). Restoring CC is not directed by the EOP. IMAs will direct initiating AP/25 to restore seal injection.

D. CORRECT: If BOTH CC and HPI Seal injection are not available then RCP seal injection must be established from the SSF RCMUP via AP125. These directions are part of EOP Immediate Manual Actions performed by the RO.

Technical Reference(s): EAP-IMA Proposed references to be provided to applicants during examination:

NONE Learning Objective:

EAP-IMA R6 Question Source:

New Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 6 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 026AK3.03 Loss of Component Cooling Water Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water:

Guidance actions contained in EOP for Loss of CCW.

(4.0/4.2)

KIA MATCH ANALYSIS Requires knowledge of reason EOP IMA's direct initiating AP/25 when RCP seal injection and CC have been lost ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since the entry conditions for AP/14 are met and the EOP does direct entry into AP's in other conditions (Ex. AP/11, AP/25). The EOP does not direct actions to restore normal seal injection. Seal injection flow is re-established via the RCMUP since both CC and SI have been lost.

B. Incorrect: Plausible since direction are given to trip RCP's however the directions are part if EOP Immediate Manual Actions. The EOP does not direct entry into AP/16 however the EOP does direct entry into AP's in other conditions (Ex. AP/11, AP/25).

C. Incorrect: Plausible since the entry conditions for AP/20 are met and the EOP does direct entry into AP's in other conditions (Ex. AP/11, AP/25). Restoring CC is not directed by the EOP. IMA's will direct initiating AP/25 to restore seal injection.

D. CORRECT: If BOTH CC and HPI Seal injection are not available then RCP seal injection must be established from the SSF RCMUP via AP/25. These directions are part of EOP Immediate Manual Actions performed by the RO.

Technical Reference(s): EAP-IMA Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-IMA R6 Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 7 Unit 1 initial conditions:

Reactor power = 90%

1 B Main Feedwater pump trips Current conditions:

Reactor power = 70% decreasing RCS pressure = 2165 psig slowly decreasing Pressurizer level = 228 inches slowly decreasing Pressurizer temperature = 640T slowly decreasing All Pressurizer heater banks controlled from Unit I control room are in AUTO and are OFF Based on the above conditions, which ONE of the following describes the status of the pressurizer and the pressurizer saturation circuit?

The pressurizer is AND the pressurizer saturation circuit A. subcooled / is responding as expected B. subcooled / has failed C. saturated I is responding as expected D. saturated / has failed 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 7 Unit 1 initial conditions:

Reactor power = 90%

1 B Main Feedwater pump trips Current conditions:

Reactor power = 70% decreasing RCS pressure = 2165 psig slowly decreasing Pressurizer level = 228 inches slowly decreasing Pressurizer temperature = 640°F slowly decreasing All Pressurizer heater banks controlled from Unit 1 control room are in AUTO and are OFF Based on the above conditions, which ONE of the following describes the status of the pressurizer and the pressurizer saturation circuit?

The pressurizer is ____ AND the pressurizer saturation circuit ___ _

A. subcooled / is responding as expected B. subcooled / has failed C. saturated / is responding as expected D. saturated / has failed

2010A NRC REACTOR OPERATOR EXAM Question 7 TIIGI -cpw 027AK2.03 Pressurizer Pressure Control System (PZR PCS) Malfunction Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following:

Controllers and positioners (2.6/2.8)

KIA MATCH ANALYSIS Requires knowledge of how controllers for Pzr saturation circuit function and the ability to diagnose a malfunction of it.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since parameters given are reasonable for the post runback condition. Normal pressurizer spray valve RC-1 would open at 2205 psig and not closed until pressure reaches 2155 psig. The decreasing RCS pressure could be explained by the decreasing Pzr level as it returns to setpoint after FDWP trip.

B. CORRECT: Saturation temp for 2165 psig is approximately 648 degrees. With the Pzr at 640 degrees it is clearly subcooled. Regarding the pressurizer level saturation circuitry, Psat must be 20 psig below actual RCS pressure before Bank 2 will energize and will not de-energize until Psat and RCS pressure (NR Med-selected RCS Pressure) are within 15 psig (5 psig dead band). With RCS pressure at 2165, pressurizer temp should be about 648°F (saturation for 2165). Saturation for actual pzr temp of 640°F is about 2045 psig therefore Bank 2 should be energized. 2205 psig.

C. Incorrect: First part is plausible since conditions in Pzr are consistent with the loss of FDWP runback. Decreasing RCS pressure is occurring concurrently with decreasing Pzr level which is a normal response if the Pzr is saturated. Second part is plausible since parameters given are reasonable for the post runback condition. Normal pressurizer spray valve RC-1 would open at 2205 psig and not closed until pressure reaches 2155 psig. The decreasing RCS pressure could be explained by the decreasing Pzr level as it returns to setpoint after FDWP trip.

D. Incorrect: First part is plausible since conditions in Pzr are consistent with the loss of FDWP runback. Decreasing RCS pressure is occurring concurrently with decreasing Pzr level which is a normal response if the Pzr is saturated. Second part is correct.

Technical Reference(s):

PNS-PZR Proposed references to be provided to applicants during examination:

NONE Learning Objective:

PNS-PZR R5, R7 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 7 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 027 AK2.03 Pressurizer Pressure Control System (PZR PCS) Malfunction Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following:

Controllers and positioners (2.6/2.8)

KIA MATCH ANALYSIS Requires knowledge of how controllers for Pzr saturation circuit function and the ability to diagnose a malfunction of it.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since parameters given are reasonable for the post runback condition. Normal pressurizer spray valve RC-1 would open at 2205 psig and not closed until pressure reaches 2155 psig. The decreasing RCS pressure could be explained by the decreasing Pzr level as it returns to setpoint after FDWP trip.

B. CORRECT: Saturation temp for 2165 psig is approximately 648 degrees. With the Pzr at 640 degrees it is clearly subcooled. Regarding the pressurizer level saturation circuitry, Psat must be 20 psig below actual RCS pressure before Bank 2 will energize and will not de-energize until Psat and RCS pressure (NR Med-selected RCS Pressure) are within 15 psig (5 psig dead band). With RCS pressure at 2165, pressurizer temp should be about 648°F (saturation for 2165). Saturation for actual pzr temp of 640°F is about 2045 psig therefore Bank 2 should be energized. 2205 psig.

C. Incorrect: First part is plausible since conditions in Pzr are consistent with the loss of FDWP runback. Decreasing RCS pressure is occurring concurrently with decreasing Pzr level which is a normal response if the Pzr is saturated. Second part is plausible since parameters given are reasonable for the post runback condition. Normal pressurizer spray valve RC-1 would open at 2205 psig and not closed until pressure reaches 2155 psig. The decreasing RCS pressure could be explained by the decreasing Pzr level as it returns to setpoint after FDWP trip.

D. Incorrect: First part is plausible since conditions in Pzr are consistent with the loss of FDWP runback. Decreasing RCS pressure is occurring concurrently with decreasing Pzr level which is a normal response if the Pzr is saturated. Second part is correct.

Technical Reference(s): PNS-PZR Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R5, R7 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 8 Unit 2 initial conditions:

Time = 1200:00 Reactor power = 100%

Both MFDW Pumps tripped Current conditions:

Time=1201:45 Reactor power = 30% decreasing Rule 1 (ATWS/Unanticipated Nuclear Power Production) in progress Loop A SCM = 0°F stable 2SA9/D2 (RC Pump Vibration High ) Actuated Rule 2 (Loss of SCM) initiated Based on the above conditions, which ONE of the following describes the required actions regarding Reactor Coolant Pumps (RCPs) provided in Rule 2 and the reason for those actions?

A. Leave RCPs operating to minimize core damage from an increase in DNBR that would occur if secured B. Leave RCPs operating to provide flow through the core for heat removal C. Secure RCPs to reduce heat input to the RCS D. Secure RCPs to prevent RCP damage 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 8 Unit 2 initial conditions:

Time = 1200:00 Reactor power = 100%

Both MFDW Pumps tripped Current conditions:

Time = 1201 :45 Reactor power = 30% decreasing Rule 1 (ATWS/Unanticipated Nuclear Power Production) in progress Loop A SCM = OaF stable 2SA9/D2 (RC Pump Vibration High) Actuated Rule 2 (Loss of SCM) initiated Based on the above conditions, which ONE of the following describes the required actions regarding Reactor Coolant Pumps (RCP's) provided in Rule 2 and the reason for those actions?

A. Leave RCP's operating to minimize core damage from an increase in DNBR that would occur if secured B. Leave RCP's operating to provide flow through the core for heat removal C. Secure RCP's to reduce heat input to the RCS D. Secure RCP's to prevent RCP damage

2010A NRC REACTOR OPERATOR EXAM Question 8 TIIGI -cpw 029EK3.12 Abnormal Transient Without Scram (ATWS)

Knowledge of the reasons for the following responses as they apply to the ATWS:

Actions contained in EOP for ATWS.

(4.4/4.7)

K/A MATCH ANALYSIS Requires knowledge of EOP directed actions for an ATWS as well as the reasons for those actions.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Leaving RCPs running is correct however the reason is due to the added heat transfer gained from forced circulation. Plausible since the second part could be correct if talking about a decrease in DNBR (instead of increase) since the loss flow that would result would cause DNBR to move towards unity.

B. CORRECT: In accordance with RULE 2, Loss of SCM, if SCMs are lost during the UNPP event, RCPs should not be tripped; they should remain in operation until power is 1% to provide flow through the core for heat removal.

Maintaining forced RCS flow is the preferred method to remove core heat (due to the increased heat transfer available).

C. Incorrect: Both parts are incorrect. Securing RCPs is plausible since this would be the correct actions if power <1 %. The reason is also plausible since securing RCPs would in fact decrease the heat input to RCS.

D. Incorrect: Both parts are incorrect. Securing RCPs is plausible since this would be the correct actions if power <1 %. The reason is also plausible since pumping a 2 phase mixture would result in pump damage due to impeller cavitation and high vibration.

Technical Reference(s):

EAP-UNPP Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-UNPP Ru Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 8 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 029EK3.12 Abnormal Transient Without Scram (ATWS)

Knowledge of the reasons for the following responses as they apply to the ATWS:

Actions contained in EOP for ATWS.

(4.4/4.7)

KIA MATCH ANALYSIS Requires knowledge of EOP directed actions for an ATWS as well as the reasons for those actions.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Leaving RCP's running is correct however the reason is due to the added heat transfer gained from forced circulation. Plausible since the second part could be correct if talking about a decrease in DNBR (instead of increase) since the loss flow that would result would cause DNBR to move towards unity.

B. CORRECT: In accordance with RULE 2, Loss of SCM, if SCMs are lost during the UNPP event, RCPs should not be tripped; they should remain in operation until power is ~ 1 % to provide flow through the core for heat removal.

Maintaining forced RCS flow is the preferred method to remove core heat (due to the increased heat transfer available).

C. Incorrect: Both parts are incorrect. Securing RCP's is plausible since this would be the correct actions if power <1 %. The reason is also plausible since securing RCP's would in fact decrease the heat input to RCS.

D. Incorrect: Both parts are incorrect. Securing RCP's is plausible since this would be the correct actions if power <1 %. The reason is also plausible since pumping a 2 phase mixture would result in pump damage due to impeller cavitation and high vibration.

Technical Reference(s): EAP-UNPP Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-UNPP R11 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 9 Unit 1 plant conditions:

SGTR tab in progress 1 B SG isolated 1A loop Tcold = 440°F decreasing 1 B S/G TUBE/SHELL DT = (-)72°F Based on the above conditions, which ONE of the following describes the reason the SGTR tab directs minimizing core SCM during cooldown AND the initial method used to reduce the SCM?

To reduce the AND reducing SCM would initially be attempted by A. primary to secondary leak rate / de-energizing Pzr heaters and cycling Pzr spray B. primary to secondary leak rate / cycling the PORV C. compressive stresses in the 1 B SG / de-energizing Pzr heaters and cycling Pzr spray D. compressive stresses in the lB SG / cycling the PORV 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 9 Unit 1 plant conditions:

SGTR tab in progress 1 B SG isolated 1A loop Tcold = 440°F decreasing 1 B S/G TUBE/SHELL DT = (-)72°F Based on the above conditions, which ONE of the following describes the reason the SGTR tab directs minimizing core SCM during cooldown AND the initial method used to reduce the SCM?

To reduce the ____ AND reducing SCM would initially be attempted by ___ _

A. primary to secondary leak rate / de-energizing Pzr heaters and cycling Pzr spray B. primary to secondary leak rate / cycling the PORV C. compressive stresses in the 1 B SG / de-energizing Pzr heaters and cycling Pzr spray D. compressive stresses in the 1 B SG / cycling the PORV

2010A NRC REACTOR OPERATOR EXAM Question 9 TI/GI -cpw 038EK3.01 Steam Generator Tube Rupture (SGTR)

Knowledge of the reasons for the following responses as they apply to the SGTR:

Equalizing pressure on primary and secondary sides of ruptured S/G.

(4.1 /4.3)

K/A MATCH ANALYSIS Requires knowing the reason for equalizing pressure on primary and secondary sides of ruptured SIG and how that is done.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: The purpose of reducing SCM during a SGTR is to reduce RCS pressure as much as possible while still maintaining SCM and RCP NPSH.

This minimizes the differential pressure between the RCS and the affected SG(s), thus minimizing the tube leak flow rate. The SGTR tab directs the operator to initially use pressurizer heaters and normal Pzr spray. If initial methods do not achieve desired results the PORV is cycled to reduce the SCM.

B. Incorrect: First part is correct. Second part is plausible since using the PORV is a strategy used in the SGTR tab to reduce SCM however it is not used unless initial methods attempted are inadequate.

C. Incorrect: First part is plausible since controlling compressive, stresses across SG tubes is a prime concern during SGTR. 1 B Tube/Shell delta T is violating the Compressive stress limit of -70°F. However, reducing SCM is not a strategy directed at correcting this issue. Feeding the isolated SG would be used to reduce the Compressive stresses. Second part is correct.

D. Incorrect: First part is plausible since controlling compressive stresses across SG tubes is a prime concern during SGTR. I B Tube/Shell delta T is violating the Compressive stress limit of -70°F. However, reducing SCM is not a strategy directed at correcting this issue. Feeding the isolated SG would be used to reduce the Compressive stresses. Second part is plausible since using the PORV is a strategy used in the SGTR tab to reduce SCM however it is not used unless initial methods attempted are inadequate.

Technical Reference(s):

EAP-SGTR, EOP reference document Proposed references to be provided to applicants during examination:

NONE Learning Objective: EAP-SGTR R9, R6 Question Source:

New Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis Question 9 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 038EK3.01 Steam Generator Tube Rupture (SGTR)

Knowledge of the reasons for the following responses as they apply to the SGTR:

Equalizing pressure on primary and secondary sides of ruptured S/G.

(4.1/4.3)

KIA MATCH ANALYSIS Requires knowing the reason for equalizing pressure on primary and secondary sides of ruptured S/G and how that is done.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: The purpose of reducing SCM during a SGTR is to reduce RCS pressure as much as possible while still maintaining SCM and RCP NPSH.

This minimizes the differential pressure between the RCS and the affected SG(s), thus minimizing the tube leak flow rate. The SGTR tab directs the operator to initially use pressurizer heaters and normal Pzr spray. If initial methods do not achieve desired results the PORV is cycled to reduce the SCM.

B. Incorrect: First part is correct. Second part is plausible since using the PORV is a strategy used in the SGTR tab to reduce SCM however it is not used unless initial methods attempted are inadequate.

C. Incorrect: First part is plausible since controlling compressive stresses across SG tubes is a prime concern during SGTR. 1 B Tube/Shell delta T is violating the Compressive stress limit of -70°F. However, reducing SCM is not a strategy directed at correcting this issue. Feeding the isolated SG would be used to reduce the Compressive stresses. Second part is correct.

D. Incorrect: First part is plausible since controlling compressive stresses across SG tubes is a prime concern during SGTR. 1 B Tube/Shell delta T is violating the Compressive stress limit of -70°F. However, reducing SCM is not a strategy directed at correcting this issue. Feeding the isolated SG would be used to reduce the Compressive stresses. Second part is plausible since using the PORV is a strategy used in the SGTR tab to reduce SCM however it is not used unless initial methods attempted are inadequate.

Technical Reference(s): EAP-SGTR, EOP reference document Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-SGTR R9, R6 Question Source: New Question History: Last NRC Exam NIA Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 10 Unit 1 plant conditions:

Both MFBs de-energized TDEFWP operating Based on the above conditions, which ONE of the following describes the status of bearing oil cooling water supply to the TDEFWP?

TDEFWP bearing oil cooling is currently provided by and it provide adequate cooling water flow until AC power has been re-established.

A. CCW / will B. HPSW / will C. CCW / will NOT D. HPSW / will NOT 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 10 Unit 1 plant conditions:

Both MFB's de-energized TDEFWP operating Based on the above conditions, which ONE of the following describes the status of bearing oil cooling water supply to the TDEFWP?

TDEFWP bearing oil cooling is currently provided by and it ___

provide adequate cooling water flow until AC power has been re-established.

A. CCW / will B. HPSW / will C. CCW / will NOT D. HPSW / will NOT

2010A NRC REACTOR OPERATOR EXAM Question 10 TIIGI -cpw 054AA1.03 Loss of Main Feedwater (MFW)

Ability to operate and I or monitor the following as they apply to the Loss of Main Feedwater (MFW):

AFW auxiliaries, including oil cooling water supply.

(3.5/3.7)

K/A MATCH ANALYSIS Loss of both MFBs will result in loss of both MFWPs. Question requires knowledge of how bearing cooling water supply to TDEFWP responds to loss of AC power.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is plausible since CCW is the normal cooling water supply to the TDEFWP bearing oil. It uses an AC pump which is not available due to loss of power. Second part is correct.

B. CORRECT: CCW is the normal cooling water supply to the TDEFWP bearing oil. It uses an AC pump which is not available due to loss of power. HPSW is the alternate supply and can provide sufficient pressure and flow via the Elevated Water Storage Tank.

C. Incorrect: First part is plausible since CCW is the normal cooling water supply to the TDEFWP bearing oil however it uses an AC pump which is not available due to loss of power. Second part is plausible since the normal supply is via an AC pump whose suction is CCW and that pump is no longer available due to the loss of power.

D. Incorrect: First part is correct. Second part is plausible since there is no AC available therefore there are no HPSW pumps operating. Additionally plausible since the normal supply is from an AC driven pump (although its suction is CCW water) which is now unavailable.

Technical Reference(s):

CF-EF Proposed references to be provided to applicants during examination: NONE Learning Objective: CF-EF R26 Question Source:

NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals Question 10 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 054AA 1.03 Loss of Main Feedwater (MFW)

Ability to operate and I or monitor the following as they apply to the Loss of Main Feedwater (MFW):

AFW auxiliaries, including oil cooling water supply.

(3.5/3.7)

KIA MATCH ANALYSIS Loss of both MFB's will result in loss of both MFWP's. Question requires knowledge of how bearing cooling water supply to TDEFWP responds to loss of AC power.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is plausible since CCW is the normal cooling water supply to the TDEFWP bearing oil. It uses an AC pump which is not available due to loss of power. Second part is correct.

B. CORRECT: CCW is the normal cooling water supply to the TDEFWP bearing oil. It uses an AC pump which is not available due to loss of power. HPSW is the alternate supply and can provide sufficient pressure and flow via the Elevated Water Storage Tank.

C. Incorrect: First part is plausible since CCW is the normal cooling water supply to the TDEFWP bearing oil however it uses an AC pump which is not available due to loss of power. Second part is plausible since the normal supply is via an AC pump whose suction is CCW and that pump is no longer available due to the loss of power.

D. Incorrect: First part is correct. Second part is plausible since there is no AC available therefore there are no HPSW pumps operating. Additionally plausible since the normal supply is from an AC driven pump (although its suction is CCW water) which is now unavailable.

Technical Reference(s): CF-EF Proposed references to be provided to applicants during examination: NONE Learning Objective: CF-EF R26 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question II Unit 1 initial conditions:

Station Blackout occurred Neither KHU automatically started Manual Emergency Start of BOTH KHUs is required Time = 1201 Keowee Emergency Start Channel A control room switch placed in START Time 1202 Keowee Emergency Start Channel B control room switch placed in START Based on the above conditions, which ONE of the following describes the time at which BOTH KHUs have received an Emergency Start signal AND the Generator Output Voltage (KV) of the KHUs that would indicate proper operation?

A. 1201 I 13.8 B. 1202 I 13.8 C. 1201

/ 4.16 D. 1202 I 4.16 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 11 Unit 1 initial conditions:

Station Blackout occurred Neither KHU automatically started Manual Emergency Start of BOTH KHU's is required Time = 1201 Keowee Emergency Start Channel A control room switch placed in START Time = 1202 Keowee Emergency Start Channel B control room switch placed in START Based on the above conditions, which ONE of the following describes the time at which BOTH KHU's have received an Emergency Start signal AND the Generator Output Voltage (KV) of the KHU's that would indicate proper operation?

A. 1201 / 13.8 B. 1202 / 13.8 C. 1201 / 4.16 D. 1202 / 4.16

2010A NRC REACTOR OPERATOR EXAM Question 11 TIIGI -cow 055EA1.02 Loss of Offsite and Onsite Power (Station Blackout)

Ability to operate and monitor the following as they apply to a Station Blackout:

Manual EDIG start (Manual start of Hydro unit acceptable).

r(4.3/4.4)

KIA MATCH ANALYSIS Requires the ability to determine how to perform a manual emergency start of both KHUs during a Station Blackout and expected indication when monitoring for proper operation.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: There are 2 switches in the Unit I and 2 control room for emergency starting the KHUs. Both switches are associated with Unit ls emergency start circuitry even though one of the switches is on the Unit 2 side of the control room. The 2 switches are a part of 2 redundant channels and either channel will emergency start both KHUs. Since the first switch was operated at 1201, that is the time both KHUs would have received an Emergency Start signal. Once operating, the normal output voltage for a KHU is 13.8 Ky.

B. Incorrect: First part is plausible since there are 2 switches and 2 KHUs therefore it would be reasonable to assume that there is a switch for each KHU (since the KHUs are redundant in themselves) and therefore chose 1202 as the time.

Additional plausibility comes from the fact that when the EOP directs manually starting both KHUs it directs using both switches. Second part is correct.

C. Incorrect: First part is correct. Second part is plausible since 4.16KV is the end voltage used by ONS.

D. Incorrect: First part is plausible since there are 2 switches and 2 KHUs therefore it would be reasonable to assume that there is a switch for each KHU (since the KHUs are redundant in themselves) and therefore chose 1202 as the time.

Additional plausibility comes from the fact that when the EOP directs manually starting both KHUs it directs using both switches. Second part is plausible since 4.16KV is the end voltage used by ONS.

Technical Reference(s): EOP End. 5.38 (Restoration of Power), EAP-BO Attach #1, EL-KHG Proposed references to be provided to applicants during examination:

NONE Learning Objective: EAP-BO R6, R7, R13 Question Source:

NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals Question 11 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 055EA 1.02 Loss of Offsite and Onsite Power (Station Blackout)

Ability to operate and monitor the following as they apply to a Station Blackout:

Manual ED/G start (Manual start of Hydro unit acceptable).

(4.3/4.4)

KIA MATCH ANALYSIS Requires the ability to determine how to perform a manual emergency start of both KHU's during a Station Blackout and expected indication when monitoring for proper operation.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: There are 2 switches in the Unit 1 and 2 control room for emergency starting the KHU's. Both switches are associated with Unit 1 's emergency start circuitry even though one of the switches is on the Unit 2 side of the control room. The 2 switches are a part of 2 redundant channels and either channel will emergency start both KHU's. Since the first switch was operated at 1201, that is the time both KHU's would have received an Emergency Start signal. Once operating, the normal output voltage for a KHU is 13.8 KV.

B. Incorrect: First part is plausible since there are 2 switches and 2 KHU's therefore it would be reasonable to assume that there is a switch for each KHU (since the KHU's are redundant in themselves) and therefore chose 1202 as the time.

Additional plausibility comes from the fact that when the EOP directs manually starting both KHU's it directs using both switches. Second part is correct.

C. Incorrect: First part is correct. Second part is plausible since 4.16KV is the end voltage used by ONS.

D. Incorrect: First part is plausible since there are 2 switches and 2 KHU's therefore it would be reasonable to assume that there is a switch for each KHU (since the KHU's are redundant in themselves) and therefore chose 1202 as the time.

Additional plausibility comes from the fact that when the EOP directs manually starting both KHU's it directs using both switches. Second part is plausible since 4.16KV is the end voltage used by ONS.

Technical Reference(s): EOP Encl. 5.38 (Restoration of Power), EAP-BO Attach #1, EL-KHG Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-BO R6, R7, R13 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 12 Which ONE of the following would indicate that the 2DIC inverter has experienced a loss of AC output voltage AND how 2KVIC panelboard would then receive power?

A. LOAD CONNECTED TO EMERGENCY light on the inverter will be illuminated AND panelboard 2KVIC will automatically be energized from panelboard 2KRA (regulated power).

B. LOAD CONNECTED TO EMERGENCY light on the inverter will be illuminated AND panelboard 2KVIC will automatically be energized from Unit 3.

C. INVERTER OUTPUT LOW light on the Inverter be illuminated AND the 2KVIC panelboard will be de-energized until manually aligned to panelboard 2KRA (regulated power).

D. INVERTER OUTPUT LOW light on the lnverter be illuminated AND the 2KVIC panelboard will be de-energized until manually aligned to receive power from its alternate unit.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 12 Which ONE of the following would indicate that the 2DIC inverter has experienced a loss of AC output voltage AND how 2KVIC panelboard would then receive power?

A. LOAD CONNECTED TO EMERGENCY light on the inverter will be illuminated AND panelboard 2KVIC will automatically be energized from panelboard 2KRA (regulated power).

B. LOAD CONNECTED TO EMERGENCY light on the inverter will be illuminated AND panelboard 2KVIC will automatically be energized from Unit 3.

C. INVERTER OUTPUT LOW light on the Inverter be illuminated AND the 2KVIC panelboard will be de-energized until manually aligned to panel board 2KRA (regulated power).

D. INVERTER OUTPUT LOW light on the Inverter be illuminated AND the 2KVIC panelboard will be de-energized until manually aligned to receive power from its alternate unit.

2010A NRC REACTOR OPERATOR EXAM Question 12 TIIGI -cpw 057AA2.15 Loss of Vital AC Electrical Instrument Bus Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

That a loss of ac has occurred.

(3.8/4.1)

K/A MATCH ANALYSIS Requires the ability to interpret indications to determine that a loss of the normal AC supply to KVID has occurred. Per NRC, global loss of AC not required to match KA.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since panelboards powered by essential inverters do have automatic backup. This would be correct if output of one of the essential inverters had been lost since their associated panelboards do have auto backup from ASCO switches and Static Transfer switches.

B. Incorrect: Plausible since this indication is available for the Essential inverters.

Additionally plausible since alternate units do back up the DC panelboards but not the AC panelboards.

C. CORRECT: If voltage on the inverter falls below 115 volts the associated output voltage low light will illuminate. If the output is lost procedures directs aligning KRA to KVIC in accordance with 0P111071010 (Operation of the Batteries and Battery Chargers) since there is no automatic backup.

D. Incorrect: First part is correct and the second part is plausible since re-alignment is a manual function and alternate units do back up the DC panelboards but not the AC panelboards.

Technical Reference(s):

ISA-I 31B7, EL-VPC Proposed references to be provided to applicants during examination:

NONE Learning Objective: EL-VPC R21R5 Question Source:

NEW Question History: Last NRC Exam NIA Question Cognitive Level: Comprehension and Analysis Question 12 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 057 AA2.15 Loss of Vital AC Electrical Instrument Bus Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

That a loss of ac has occurred.

(3.8/4.1)

KIA MATCH ANALYSIS Requires the ability to interpret indications to determine that a loss of the normal AC supply to KVID has occurred.

N global loss not ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since panelboards powered by essential inverters do have automatic backup. This would be correct if output of one of the essential inverters had been lost since their associated panel boards do have auto backup from ASCO switches and Static Transfer switches.

B. Incorrect: Plausible since this indication is available for the Essential inverters.

Additionally plausible since alternate units do back up the DC panel boards but not the AC panelboards.

C. CORRECT: If voltage on the inverter falls below 115 volts the associated output voltage low light will illuminate. If the output is lost procedures directs aligning KRA to KVIC in accordance with OP/1107/010 (Operation of the Batteries and Battery Chargers) since there is no automatic backup.

D. Incorrect: First part is correct and the second part is plausible since re-alignment is a manual function and alternate units do back up the DC panel boards but not the AC panelboards.

Technical Reference(s): 1SA-13/B7, EL-VPC Proposed references to be provided to applicants during examination: NONE Learning Objective: EL-VPC R2/R5 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 13 Plant conditions:

1 CA Battery Charger fails - output voltage = 0 VDC 1CA Battery voltage = 126 VDC 1 DCB Bus voltage

= 123 VDC Unit 2 DCAIDCB Bus voltage = 124 VDC Unit 3 DCAIDCB Bus voltage = 127 VDC Based on the above conditions, which ONE of the following will automatically supply power to 1 DIA panelboard?

A. 1DCB Bus B. bA Battery C. Unit 2 DC Bus D. Unit 3 DC Bus 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 13 Plant conditions:

1 CA Battery Charger fails - output voltage = 0 VDC 1 CA Battery voltage = 126 VDC 1DCB Bus voltage

= 123 VDC Unit 2 DCAlDCB Bus voltage = 124 VDC Unit 3 DCAlDCB Bus voltage = 127 VDC Based on the above conditions, which ONE of the following will automatically supply power to 1 DIA panelboard?

A. 1DCB Bus B. 1 CA Battery C. Unit 2 DC Bus D. Unit 3 DC Bus

2010A NRC REACTOR OPERATOR EXAM Question 13 TI/GI cpw 058AK1.01, Loss of DC Power Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation (2.8/3.1r:

K/A MATCH ANALYSIS Requires knowledge of the operational implications of failed battery charger and the operational impact of the loss of a Vital DC Battery Charger and the response by the Vital DC system ANSWER CHOICE ANALYSIS Answer: B A. Incorrect. For the Vital DC system, the 1 DCB bus is not aligned to the I DCA bus.

Plausible because 1DCB Bus is aligned to backup the essential inverters.

B. Correct. The voltage from ICA battery is higher than the backup source (Unit 2 DC Bus). Unit ICA battery will supply power.

C. Incorrect, plausible because this would be correct if the Unit 2 DC bus voltage was higher than the 1 CA battery voltage.

D. Incorrect. Unit 3s DC Bus is not connected to Unit 1. Plausible because unit 3 does backup Unit 1 in the SSF power scheme.

Technical Reference(s): Lesson Plan EL-DCD Proposed references to be provided to applicants during examination: None Learning Objective: EL-DCD R4 Question Source: Bank Question History: Last NRC Exam 2009 (modified) #14 Question Cognitive Level:

Comprehension and Analysis 2010A NRC REACTOR OPERATOR EXAM Question 13 T1/G1 cpw 058AK1.01, Loss of DC Power Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation (2.8/3.1 )

KIA MATCH ANALYSIS Requires knowledge of the operational implications of failed battery charger and the operational impact of the loss of a Vital DC Battery Charger and the response by the Vital DC system ANSWER CHOICE ANALYSIS Answer: B A. Incorrect. For the Vital DC system, the 1 DCB bus is not aligned to the 1 DCA bus.

Plausible because 1 DCB Bus is aligned to backup the essential inverters.

B. Correct. The voltage from 1 CA battery is higher than the backup source (Unit 2 DC Bus). Unit 1 CA battery will supply power.

C. Incorrect, plausible because this would be correct if the Unit 2 DC bus voltage was higher than the 1 CA battery voltage.

D. Incorrect. Unit 3's DC Bus is not connected to Unit 1. Plausible because unit 3 does backup Unit 1 in the SSF power scheme.

Technical Reference(s): Lesson Plan EL-DCD Proposed references to be provided to applicants during examination: None Learning Objective: EL-DCD R4 Question Source: Bank Question History: Last NRC Exam 2009 (modified) #14 Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 14 Unit 1 and 2 initial conditions:

A and B LPSW pump operating Current conditions:

ISA9/A9 (LPSW HEADER A PRESS LOW)

A LPSW pump amps = 15 - 35 fluctuating B LPSW pump amps = 55 stable LPSW HDR PRESS = rapidly fluctuating between 60 & 75 psig Based on current conditions, which ONE of the following describes the status of the LPSW pumps and what actions are directed by AP/24 (Loss of LPSW)?

A. The A LPSW pump is cavitating / Place the Unit 1/2 STANDBY LPSW PUMP AUTO START CIRCUIT in DISABLE then stop LPSW Pump A and start LPSW Pump C B. The A LPSW pump has a sheared shaft / Place the Unit 1/2 STANDBY LPSW PUMP AUTO START CIRCUIT in DISABLE then stop LPSW Pump A and start LPSW Pump C C. The A LPSW pump is cavitating / Start LPSW Pump C then stop LPSW Pump A D. The A LPSW pump has a sheared shaft / Start LPSW Pump C then stop LPSW Pump A 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 14 Unit 1 and 2 initial conditions:

A and B LPSW pump operating Current conditions:

1 SA9/A9 (LPSW HEADER A PRESS LOW)

A LPSW pump amps = 15 - 35 fluctuating B LPSW pump amps = 55 stable LPSW HDR PRESS = rapidly fluctuating between 60 & 75 psig Based on current conditions, which ONE of the following describes the status of the LPSW pumps and what actions are directed by AP/24 (Loss of LPSW)?

A. The A LPSW pump is cavitating 1 Place the Unit 1/2 STANDBY LPSW PUMP AUTO START CIRCUIT in DISABLE then stop LPSW Pump A and start LPSW PumpC B. The A LPSW pump has a sheared shaft 1 Place the Unit 1/2 STANDBY LPSW PUMP AUTO START CIRCUIT in DISABLE then stop LPSW Pump A and start LPSW Pump C C. The A LPSW pump is cavitating 1 Start LPSW Pump C then stop LPSW Pump A D. The A LPSW pump has a sheared shaft 1 Start LPSW Pump C then stop LPSW Pump A

2010A NRC REACTOR OPERATOR EXAM Question 14 TIIGI -cpw 0062AG2.1.20 Loss of Nuclear Service Water Ability to interpret and execute procedure steps.

(4.6/4.6)

KIA MATCH ANALYSIS Requires ability to compare LPSW system indications to AP/24 requirements and execute appropriate steps ANSWER CHOICE ANALYSIS Answer: A A. Correct: Indication given is consistent with pump cavitation on LPSW Pump A.

LPSW Pump B amps are at the normal value for existing conditions. AP124 procedural direction for cavitation is to disable the auto start feature then stop the affected pump.

B. Incorrect: Sheared shaft indication would be low amps vice fluctuating amps.

Plausible as it is the correct pump. The actions given are procedurally correct if candidate decides the pump is not cavitating.

C. Incorrect: Wrong pump is referenced as cavitating. Plausible if candidate misinterprets the data given. Procedure direction is consistent if wrong pump is selected.

D. Incorrect: The wrong pump is selected. Sheared shaft indication would be low amps vice fluctuating amps. The actions given are procedurally correct if candidate decides the pump is not cavitating and are consistent with misinterpreting the pump affected.

Technical Reference(s): AP124 (Loss of LPSW)

Proposed references to be provided to applicants during examination: None Learning Objective: SSS-LPW Obj R15, EAP-APG (R9)

Question Source: BANK Question History: Last NRC Exam: 2007 Retest #53 Question Cognitive Level:

Comprehension or Analysis 2010A NRC REACTOR OPERATOR EXAM Question 14 T1/G1 - cpw 0062AG2.1.20 Loss of Nuclear Service Water Ability to interpret and execute procedure steps.

(4.6/4.6)

KIA MATCH ANALYSIS Requires ability to compare LPSW system indications to AP/24 requirements and execute appropriate steps ANSWER CHOICE ANALYSIS Answer: A A. Correct: Indication given is consistent with pump cavitation on LPSW Pump A.

LPSW Pump B amps are at the normal value for existing conditions. AP/24 procedural direction for cavitation is to disable the auto start feature then stop the affected pump.

B. Incorrect: Sheared shaft indication would be low amps vice fluctuating amps.

Plausible as it is the correct pump. The actions given are procedurally correct if candidate decides the pump is not cavitating.

C. Incorrect: Wrong pump is referenced as cavitating. Plausible if candidate misinterprets the data given. Procedure direction is consistent if wrong pump is selected.

D. Incorrect: The wrong pump is selected. Sheared shaft indication would be low amps vice fluctuating amps. The actions given are procedurally correct if candidate decides the pump is not cavitating and are consistent with misinterpreting the pump affected.

Technical Reference(s): AP/24 (Loss of LPSW)

Proposed references to be provided to applicants during examination: None Learning Objective: SSS-LPW Obj R15, EAP-APG (R9)

Question Source: BANK Question History: Last NRC Exam: 2007 Retest #53 Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 15 Unit 3 initial conditions:

Time=1200 Reactor power = 45% stable Operating Main Feedwater Pump trips IA and AlA are lost to 3FDW-315 Current conditions:

Time=1300 RCS Tave = 550T stable Based on the above conditions, which ONE of the following describes the expected SG level and the status of 3FDW-315?

ASSUME NO OPERATOR ACTIONS HAVE OCCURRED A. 25 s/U level / failed open B. 25 5/U level / controlling SG level at setpoint C. 30 XSUR / failed open D. 30 XSUR / controlling SG level at setpoint 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 15 Unit 3 initial conditions:

Time = 1200 Reactor power = 45% stable Operating Main Feedwater Pump trips IA and AlA are lost to 3FDW-315 Current conditions:

Time = 1300 RCS Tave = 550°F stable Based on the above conditions, which ONE of the following describes the expected SG level and the status of 3FDW-315?

ASSUME NO OPERATOR ACTIONS HAVE OCCURRED A. 25" S/U level/failed open B. 25" S/U level/controlling SG level at setpoint C. 30" XSUR / failed open D. 30" XSUR / controlling SG level at setpoint

2010A NRC REACTOR OPERATOR EXAM Question 15 TIIGI -cpw 065AA2.07 Loss of Instrument Air Ability to determine and interpret the following as they apply to the Loss of Instrument Air:

Whether backup nitrogen supply is controlling valve position.

(2.8/3.2)

K/A MATCH ANALYSIS Requires interpreting plant conditions to determine how valve should be responding when being supplied by backup nitrogen supply to control valve position ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Correct level is 30. 25 is Plausible since it would be the correct level if Main FDW were supplying the SGs. 3FDW-315 would still have auto control capability. Second part is plausible since there is a loss of air supply. Being unaware of N2 backup or not knowing N2 backup is good for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would lead to this choice. Additional plausibility comes from the fact that Rule 3 does provide guidance to feed through the startup valve if FDW-315 is failed open and when feeding through the startup valve, 25 is the normal level setpoint.

B. Incorrect: Correct level is 30. 25 is Plausible since it would be the correct level if Main FDW were supplying the SGs. 3FDW-315 would still have auto control capability. Second part is correct.

C. Incorrect: First part is correct. 3FDW-315 would still have auto control capability.

Second part is plausible since there is a loss of air supply. Being unaware of N2 backup or not knowing N2 backup is good for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would lead to this choice. Additional plausibility comes from the fact that Rule 3 does provide guidance to feed through the startup valve if FDW-315 is failed open and when feeding through the startup valve AND when on EFDW, 30 XSUR is the normal level setpoint.

D. CORRECT: 3FDW-315 has backup N2 supply that insures adequate level control for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Since elapsed time is only 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the backup N2 supply would be regulating valve position as required. Since level being controlled by EFDW the correct level is 30 XSUR.

Technical Reference(s):

CF-EF Proposed references to be provided to applicants during examination: NONE Learning Objective:

CF-EF R39 Question Source:

New Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 15 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 065AA2.07 Loss of Instrument Air Ability to determine and interpret the following as they apply to the Loss of Instrument Air:

Whether backup nitrogen supply is controlling valve position.

(2.8/3.2)

KIA MATCH ANALYSIS Requires interpreting plant conditions to determine how valve should be responding when being supplied by backup nitrogen supply to control valve position ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Correct level is 30". 25" is Plausible since it would be the correct level if Main FDW were supplying the SG's. 3FDW-315 would still have auto control capability. Second part is plausible since there is a loss of air supply. Being unaware of N2 backup or not knowing N2 backup is good for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would lead to this choice. Additional plausibility comes from the fact that Rule 3 does provide guidance to feed through the startup valve if FDW-315 is failed open and when feeding through the startup valve, 25" is the normal level setpoint.

B. Incorrect: Correct level is 30".25" is Plausible since it would be the correct level if Main FDW were supplying the SG's. 3FDW-315 would still have auto control capability. Second part is correct.

C. Incorrect: First part is correct. 3FDW-315 would still have auto control capability.

Second part is plausible since there is a loss of air supply. Being unaware of N2 backup or not knowing N2 backup is good for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would lead to this choice. Additional plausibility comes from the fact that Rule 3 does provide guidance to feed through the startup valve if FDW-315 is failed open and when feeding through the startup valve AND when on EFDW, 30" XSUR is the normal level setpoint.

D. CORRECT: 3FDW-315 has backup N2 supply that insures adequate level control for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Since elapsed time is only 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the backup N2 supply would be regulating valve position as required. Since level being controlled by EFDW the correct level is 30" XSUR.

Technical Reference(s): CF-EF Proposed references to be provided to applicants during examination: NONE Learning Objective: CF-EF R39 Question Source: New Question History: Last NRC Exam NIA Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 16 Unit 1 plant conditions:

Reactor power = 50% stable Units 2 and 3 in MODE 5 Grid disturbance is in progress BOTH KHUs generating to the grid AN but ONE of the Offsite Sources required by Tech Spec 3.8.1 (AC Sources-Operating) are lost Based on the above conditions which ONE of the following describes

1) actions required by Tech Spec 3.8.1 ?

AND

2) a condition that would require manually separating BOTH KHUs from the electrical grid?

A. Immediately enter Tech Spec LCO 3.0.3 / KHU High Generator Output Voltage B. Immediately enter Tech Spec LCO 3.0.3 / KHU Low Generator Output Voltage C. Energize BOTH Standby Busses within one hour / KHU High Generator Output Voltage D. Energize BOTH Standby Busses within one hour / KHU Low Generator Output Voltage 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 16 Unit 1 plant conditions:

Reactor power = 50% stable Units 2 and 3 in MODE 5 Grid disturbance is in progress BOTH KHU's generating to the grid All but ONE of the Offsite Sources required by Tech Spec 3.8.1 (AC Sources-Operating) are lost Based on the above conditions which ONE of the following describes...

1) actions required by Tech Spec 3.8.1 ?

AND

2) a condition that would require manually separating BOTH KHU's from the electrical grid?

A. Immediately enter Tech Spec LCO 3.0.3 / KHU High Generator Output Voltage B. Immediately enter Tech Spec LCO 3.0.3 / KHU Low Generator Output Voltage C. Energize BOTH Standby Busses within one hour / KHU High Generator Output Voltage D. Energize BOTH Standby Busses within one hour / KHU Low Generator Output Voltage

2010A NRC REACTOR OPERATOR EXAM Question 16 TI/GI -cpw ro77AG2.2.22 Generator Voltage and Electrical Grid Disturbances Knowledge of limiting conditions for operations and safety limits.

(4.0/4.7)

KIA MATCH ANALYSIS Requires knowledge of LCO entry conditions bases on degraded grid ANSWER CHOICE ANALYSIS Answer: 0 A. Incorrect: First part is plausible since entering LCO 3.0.3 is generally done when a loss of safety function has occurred and it would be a reasonable misconception that the KHUs would be inoperable due to generating to the grid during a grid disturbance which would constitute a loss of both emergency power supplies.

Second part is plausible since a low switchyard voltage could result in high KHU output voltages so those conditions may exist given the ICs.

B. Incorrect: First part is plausible since entering LCO 3.0.3 is generally done when a loss of safety function has occurred and it would be a reasonable misconception that the KHUs would be inoperable due to generating to the grid during a grid disturbance which would constitute a loss of both emergency power supplies.

Second part is correct.

C. Incorrect: First part is correct Second part is plausible since a low switchyard voltage could result in high KHU output voltages so those conditions may exist given the ICs.

0. CORRECT: With one or both required offsite sources inoperable TS 3.8.1 Condition J requires energizing both SBBs within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Per AP134 if a KHU is tied to the grid and its output voltage reaches 13.2KV the unit must be separated from the grid.

Technical Reference(s):

TS 3.8.1 AP134 (Degraded Grid)

Proposed references to be provided to applicants during examination:

NONE Learning Objective:

ADM-TSS R4 EAP-APG R8 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 16 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 077 AG2.2.22 Generator Voltage and Electrical Grid Disturbances Knowledge of limiting conditions for operations and safety limits.

(4.0/4.7)

KIA MATCH ANALYSIS Requires knowledge of LCO entry conditions bases on degraded grid ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is plausible since entering LCO 3.0.3 is generally done when a loss of safety function has occurred and it would be a reasonable misconception that the KHU's would be inoperable due to generating to the grid during a grid disturbance which would constitute a loss of both emergency power supplies.

Second part is plausible since a low switchyard voltage could result in high KHU output voltages so those conditions may exist given the IC's.

B. Incorrect: First part is plausible since entering LCO 3.0.3 is generally done when a loss of safety function has occurred and it would be a reasonable misconception that the KHU's would be inoperable due to generating to the grid during a grid disturbance which would constitute a loss of both emergency power supplies.

Second part is correct.

C. Incorrect: First part is correct Second part is plausible since a low switchyard voltage could result in high KHU output voltages so those conditions may exist given the IC's.

D. CORRECT: With one or both required offsite sources inoperable TS 3.8.1 Condition J requires energizing both SBB's within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Per AP/34 if a KHU is tied to the grid and its output voltage reaches 13.2KV the unit must be separated from the grid.

Technical Reference(s): TS 3.8.1 AP/34 (Degraded Grid)

Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-TSS R4 EAP-APG R8 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 17 Unit 1 initial conditions:

Reactor power = 100%

lCHPlpumpOOS Current conditions:

1A and lB Main FDW pumps tripped Condensate Booster Pumps unavailable All EFDW pumps unavailable 1A and 1 B SG Outlet pressure = 860 psig slowly decreasing RCS pressure = 2317 psig increasing Based on the above conditions, which ONE of the following describes the required operator action(s) in accordance with the EOP?

A.

Establish SSF ASW flow to the SG and establish SG levels at 240 inches.

B.

Establish SSF ASW flow to the SG and do NOT establish a level in the SG5.

C.

Establish HPI forced cooling and open 1HP-410.

D.

Establish HPI forced cooling and open I HP-409.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 17 Unit 1 initial conditions:

Reactor power = 100%

1 C HPI pump OOS Current conditions:

1 A and 1 B Main FDW pumps tripped Condensate Booster Pumps unavailable All EFDW pumps unavailable 1 A and 1 B SG Outlet pressure = 860 psig slowly decreasing RCS pressure = 2317 psig increasing Based on the above conditions, which ONE of the following describes the required operator action(s) in accordance with the EOP?

A.

Establish SSF ASW flow to the SG and establish SG levels at 240 inches.

B.

Establish SSF ASW flow to the SG and do NOT establish a level in the SGs.

C. Establish HPI forced cooling and open 1 HP-41 O.

D. Establish HPI forced cooling and open 1 HP-409.

2010A NRC REACTOR OPERATOR EXAM Question 17 TIIGI -cpw BEO4EA2.2 Inadequate Heat Transfer Ability to determine and interpret the following as they apply to the (Inadequate Heat Transfer):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

(3.6/4.4)

K/A MATCH ANALYSIS Requires knowledge of appropriate mitigation strategy contained in plant procedures for inadequate heat transfer conditions. Demonstrating compliance with those procedures represents operation within the limitations in the facilitys license and amendments.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect. Will not be required with adequate HPI flow. Plausible because it would be correct if HPI is considered degraded.

If there is only 1 HPIP operating then actions are taken to align SSF ASW to feed the SGs. 240 is level directed by Rule 7 when feeding from SSF-ASW B. Incorrect. Will not be required with adequate HPI flow. Plausible because aligning SSF-ASW would be correct if HPI is considered degraded (only 1 HPIP available).

Not establishing a level is plausible since it is consistent with EOP guidance on feeding a dry SG with feedwater.

C. Incorrect. HP-410 will not establish flow in the B header. Plausible because HP-410 is the cross over valve for the A HPI header and valve sequence is reversed.

0. Correct. With the C HPIP inoperable, flow in the B HPI header will be inadequate which will require the operator to open HP-409.

Technical Reference(s): EAP-LOHT Rule 4 Proposed references to be provided to applicants during examination:

NONE Learning Objective: EAP-LOHT R24, R28 Question Source:

BANK EAPO7OIO2 Question History: Last NRC Exam Oconee RO 2006 Question Cognitive Level: Comprehension and Analysis Question 17 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM BE04EA2.2 Inadequate Heat Transfer Ability to determine and interpret the following as they apply to the (Inadequate Heat Transfer):

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

(3.6/4.4)

KIA MATCH ANALYSIS Requires knowledge of appropriate mitigation strategy contained in plant procedures for inadequate heat transfer conditions. Demonstrating compliance with those procedures represents operation within the limitations in the facility's license and amendments.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect. Will not be required with adequate HPI flow. Plausible because it would be correct if HPI is considered degraded. If there is only 1 HPIP operating then actions are taken to align SSF ASW to feed the SG's. 240" is level directed by Rule 7 when feeding from SSF-ASW B. Incorrect. Will not be required with adequate HPI flow. Plausible because aligning SSF-ASW would be correct if HPI is considered degraded (only 1 HPIP available).

Not establishing a level is plausible since it is consistent with EOP guidance on feeding a dry SG with feedwater.

C. Incorrect. HP-410 will not establish flow in the B header. Plausible because HP-410 is the cross over valve for the A HPI header and valve sequence is reversed.

D. Correct. With the C HPIP inoperable, flow in the B HPI header will be inadequate which will require the operator to open HP-409.

Technical Reference(s): EAP-LOHT Rule 4 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-LOHT R24, R28 Question Source: BANK EAP070102 Question History: Last NRC Exam Oconee RO 2006 Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 18 Unit 1 initial conditions:

Reactor power = 100%

1AMSLB occurs Current conditions:

Reactor has tripped RCS Tave = 544T slowly increasing 1A SG Pressure = 0 psig 1 B SG Pressure = 990 psig slowly increasing Turbine bypass valves in Auto Reactor Building pressure = 0.2 psig stable Based on the above conditions, which ONE of the following describes the status of the TDEFWP and how subsequent operation of the TDEFWP would be performed?

TDEFWP is...

A. operating and can be secured with TDEFWP control switch before AFIS is reset B. operating and can be secured with TDEFWP control switch ONLY after AFIS is reset C. NOT operating and can be started with TDEFWP control switch before AFIS is reset D. NOT operating and can be started with TDEFWP control switch ONLY after AFIS is reset 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 18 Unit 1 initial conditions:

Reactor power = 100%

1 A MSLB occurs Current conditions:

Reactor has tripped RCS Tave = 544°F slowly increasing 1 A SG Pressure = 0 psig 1 B SG Pressure = 990 psig slowly increasing Turbine bypass valves in Auto Reactor Building pressure = 0.2 psig stable Based on the above conditions, which ONE of the following describes the status of the TDEFWP and how subsequent operation of the TDEFWP would be performed?

TDEFWP is...

A. operating and can be secured with TDEFWP control switch before AFIS is reset B. operating and can be secured with TDEFWP control switch ONLY after AFIS is reset C. NOT operating and can be started with TDEFWP control switch before AFIS is reset D. NOT operating and can be started with TDEFWP control switch ONLY after AFIS is reset

2010A NRC REACTOR OPERATOR EXAM Question 18 TI/GI

- cpw BEO5EK2. I Excessive Heat Transfer Knowledge of the interrelations between the (Excessive Heat Transfer) and the following:

Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(3.8/4.0)

KIA MATCH ANALYSIS Requires knowledge of the relationship between EHT and the manual and automatic operation of the TDEFWP.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since 1FDW-315 is closed in first step of Rule 5. This makes it plausible that AFIS would not secure the TDEFWP so that it would be available to feed the B SG if needed.

B. Incorrect: Plausible since IFDW-315 is closed in first step of Rule 5. This makes it plausible that AFIS would not secure the TDEFWP so that it would be available to feed the B SG if needed. Second part is plausible since many components require manual action other than just turning switch to re-position following a safety system actuation (ox: ES components).

C. CORRECT: The TDEFWP control switch will override the AFIS interlock to close TO-I 45. TO-I 45 blocks the hydraulic oil supply to MS-95 therefore stopping steam supply to the TDEFWP. The TDEFWP switch overrides the AFIS signal and allows the operator to restart the TDEFWP as necessary to feed Steam Generators without resetting the AFIS signal.

D. Incorrect: TDEFWP would be off. Second part is plausible since many components require manual action other than just turning switch to re-position following a safety system actuation (ex: ES components).

Technical Reference(s):

CF-EF Proposed references to be provided to applicants during examination:

NONE Learning Objective: CF-EF R58 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis Question 18 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM BE05EK2.1 Excessive Heat Transfer Knowledge of the interrelations between the (Excessive Heat Transfer) and the following:

Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(3.8/4.0)

KIA MATCH ANALYSIS Requires knowledge of the relationship between EHT and the manual and automatic operation of the TDEFWP.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since 1 FDW-315 is closed in first step of Rule 5. This makes it plausible that AFIS would not secure the TDEFWP so that it would be available to feed the B SG if needed.

B. Incorrect: Plausible since 1 FDW-315 is closed in first step of Rule 5. This makes it plausible that AFIS would not secure the TDEFWP so that it would be available to feed the B SG if needed. Second part is plausible since many components require manual action other than just turning switch to re-position following a safety system actuation (ex: ES components).

C. CORRECT: The TDEFWP control switch will override the AFIS interlock to close TO-145. TO-145 blocks the hydraulic oil supply to MS-95 therefore stopping steam supply to the TDEFWP. The TDEFWP switch overrides the AFIS signal and allows the operator to restart the TDEFWP as necessary to feed Steam Generators without resetting the AFIS signal.

D. Incorrect: TDEFWP would be off. Second part is plausible since many components require manual action other than just turning switch to re-position following a safety system actuation (ex: ES components).

Technical Reference(s): CF-EF Proposed references to be provided to applicants during examination: NONE Learning Objective: CF-EF R58 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 19 Unit 1 plant conditions:

Reactor power = 70% stable Pressurizer level = 210 slowly decreasing 1HP-120 (RC VOLUME CONTROL) failed closed AP/14 (Loss of Normal HPI Makeup and/or RCP Seal Injection) initiated Based on the above conditions, which ONE of the following describes the initial actions required to control Pressurizer level AND the minimum allowed Pressurizer level (inches) in accordance with AP/14?

Throttle...

A. 1HP-26 I 200 B. IHP-26 / 80 C. 1HP-122 (RC VOLUME CONTROL BYPASS) / 200 D. 1HP-122 (RC VOLUME CONTROL BYPASS! 80 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 19 Unit 1 plant conditions:

Reactor power = 70% stable Pressurizer level = 210" slowly decreasing 1 HP-120 (RC VOLUME CONTROL) failed closed AP/14 (Loss of Normal HPI Makeup and/or RCP Seal Injection) initiated Based on the above conditions, which ONE of the following describes the initial actions required to control Pressurizer level AND the minimum allowed Pressurizer level (inches) in accordance with AP/14?

Throttle...

A. 1 HP-26 / 200 B. 1 HP-26 / 80 C. 1 HP-122 (RC VOLUME CONTROL BYPASS) / 200 D. 1 HP-122 (RC VOLUME CONTROL BYPASS / 80

2010A NRC REACTOR OPERATOR EXAM Question 19 T1/G2

- cpw 028AA1.07 Pressurizer (PZR) Level Control Malfunction Ability to operate and I or monitor the following as they apply to the Pressurizer Level Control Malfunctions:

Charging pumps maintenance of PZR level (including manual backup).

(3.3/3.3)

K/A MATCH ANALYSIS Requires knowledge of how HPIPs and valves are utilized to maintain pressurizer level following a level control valve (HP-120) failure. Since pump operation is not directly impacted by a failure of the level control valve (HP-120), manual throttling of HP-26 is how Pzr level is maintained.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: APII4 directs throttling makeup through HP-26 to maintain PZR

>200. If HP-26 fails, NEO locally open HP-122 (HP-120 bypass). EAP-APG Enclosure AP/14 page 5.

B. Incorrect: First part is correct. Second part is plausible since 80 is the pressurizer level required to maintain pressurizer heater operability. Rule 6 allows throttling provided pressurizer level is increasing and with the 80 heater cutoff it could be a reasonable misconception that 80 is the low level limit.

C. Incorrect: First part is incorrect. First part is plausible since 1HP-122 would be correct if 1 HP-26 would not open. Second part is correct.

D. Incorrect: Both parts are incorrect. First part is plausible since 1HP-122 would be correct if 1 HP-26 would not open. Second part is plausible since 80 is the pressurizer level required to maintain pressurizer heater operability. Rule 6 allows throttling provided pressurizer level is increasing and with the 80 heater cutoff it could be a reasonable misconception that 80 is the low level limit.

Technical Reference(s):

EAP-APG Enclosure APII4 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-APG R9 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 19 T1/G2 - cpw 2010A NRC REACTOR OPERATOR EXAM 028AA 1.07 Pressurizer (PZR) Level Control Malfunction Ability to operate and I or monitor the following as they apply to the Pressurizer Level Control Malfunctions:

Charging pumps maintenance of PZR level (including manual backup).

(3.3/3.3)

KIA MATCH ANALYSIS Requires knowledge of how HPIP's and valves are utilized to maintain pressurizer level following a level control valve (HP-120) failure. Since pump operation is not directly impacted by a failure of the level control valve (HP-120), manual throttling of HP-26 is how Pzr level is maintained.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: AP/14 directs throttling makeup through HP-26 to maintain PZR

>200". If HP-26 fails, NEO locally open HP-122 (HP-120 bypass). EAP-APG Enclosure AP/14 page 5.

B. Incorrect: First part is correct. Second part is plausible since 80" is the pressurizer level required to maintain pressurizer heater operability. Rule 6 allows throttling provided pressurizer level is increasing and with the 80" heater cutoff it could be a reasonable misconception that 80" is the low level limit.

C. Incorrect: First part is incorrect. First part is plausible since 1 HP-122 would be correct if 1 HP-26 would not open. Second part is correct.

D. Incorrect: Both parts are incorrect. First part is plausible since 1 HP-122 would be correct if 1 HP-26 would not open. Second part is plausible since 80" is the pressurizer level required to maintain pressurizer heater operability. Rule 6 allows throttling provided pressurizer level is increasing and with the 80" heater cutoff it could be a reasonable misconception that 80" is the low level limit.

Technical Reference(s): EAP-APG Enclosure AP/14 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-APG R9 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 20 Unit 1 initial conditions:

Reactor in MODE 6 Current conditions:

FTC Level approximately 6 inches below the 21.34 foot mark and slowly decreasing East fuel carriage is in the RB and empty West fuel carriage is in the SEP and empty Reactor Building Main Fuel Bridge in transit to the upender with a spent fuel assembly in the mast Section 4D (Fuel Transfer Canal Flooded) of AP/26 (Loss of Decay Heat Removal) initiated Based on the conditions above, which ONE of the following describes the first actions required to be taken in accordance with Section 4D (Fuel Transfer Canal Flooded)?

A. Close 1SF-I and I SF-2 (East/West Transfer Tube solations)

B. Verify SF system aligned for refueling cooling mode and stop 2B SF cooling pump C. Place the fuel assembly into the East Upender and position the West Fuel Carriage to the RB D. Place the fuel assembly into the East Upender and position the East Fuel Carriage to the SEP 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 20 Unit 1 initial conditions:

Reactor in MODE 6 Current conditions:

FTC Level approximately 6 inches below the 21.34 foot mark and slowly decreasing East fuel carriage is in the RB and empty West fuel carriage is in the SFP and empty Reactor Building Main Fuel Bridge in transit to the upender with a spent fuel assembly in the mast Section 40 (Fuel Transfer Canal Flooded) of AP/26 (Loss of Decay Heat Removal) initiated Based on the conditions above, which ONE of the following describes the first actions required to be taken in accordance with Section 40 (Fuel Transfer Canal Flooded)?

A. Close 1 SF-1 and 1 SF-2 (East/West Transfer Tube Isolations)

B. Verify SF system aligned for refueling cooling mode and stop 2B SF cooling pump C. Place the fuel assembly into the East Upender and position the West Fuel Carriage to the RB D. Place the fuel assembly into the East Upender and position the East Fuel Carriage to the SFP

2010A NRC REACTOR OPERATOR EXAM Question 20 TIIG2 036AK2.01 Fuel Handling Incidents Knowledge of the interrelations between the Fuel Handling Incidents and the following:

Fuel handling equipment.

(2.9/3.5)

KIA MATCH ANALYSIS Requires knowledge of the relationship between a fuel handling incident resulting in a decreasing fuel transfer canal water level and pieces of fuel handling equipment (Upenders and Fuel Carriage).

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible because it will be performed later however both carriages must be placed in the SFP prior to closing SF-1/2.

B. Incorrect: Plausible because it will be performed later in this section of AP/26.

C. Incorrect: East upender is the correct location. Procedure directs placing the carriages in the SEP to allow FTT Isolation valves to be closed. Misconception about which way the carriage must go to close SE-i & 2 D. Correct: Procedure directs placing the fuel assembly in transit into a safe location and specifies the upender or originallintended location and positioning the carriages in the SFP in preparation for closing the FTT Isolation valves.

Technical Reference(s): AP126 Rev 20, TS 3.9.6 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-APG Obj R9, FH-FHS Obj R7 Question Source: BANK Question History: Last NRC Exam ONS RO 2009 #62 (Re-ordered distracters)

Question Cognitive Level:

Comprehension or Analysis 2010A NRC REACTOR OPERATOR EXAM Question 20 T1/G2 -

036AK2.01 Fuel Handling Incidents Knowledge of the interrelations between the Fuel Handling Incidents and the following:

Fuel handling equipment.

(2.9/3.5)

KIA MATCH ANALYSIS Requires knowledge of the relationship between a fuel handling incident resulting in a decreasing fuel transfer canal water level and pieces of fuel handling equipment (Upenders and Fuel Carriage).

ANSWER CHOICE ANALYSIS Answer: 0 A. Incorrect: Plausible because it will be performed later however both carriages must be placed in the SFP prior to closing SF-1/2.

B. Incorrect: Plausible because it will be performed later in this section of AP/26.

C. Incorrect: East upender is the correct location. Procedure directs placing the carriages in the SFP to allow FTT Isolation valves to be closed. Misconception about which way the carriage must go to close SF-1 & 2 D. Correct: Procedure directs placing the fuel assembly in transit into a safe location and specifies the upender or original/intended location and positioning the carriages in the SFP in preparation for closing the FTT Isolation valves.

Technical Reference(s): AP/26 Rev 20, TS 3.9.6 Proposed references to be provided to applicants during examination: None Learning Objective: EAP-APG Obj R9, FH-FHS Obj R7 Question Source: BANK Question History: Last NRC Exam ONS RO 2009 #62 (Re-ordered distracters)

Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 21 Which ONE of the following describes conditions that indicate RIA-54 is unable to perform its function AND if batch releases are allowed while the RIA is inoperable?

A. Counts do not increase when Source Check is performed I Batch releases are allowed.

B. Counts do not increase when Source Check is performed I Batch releases are NOT allowed.

C. Sample pump found OFF I Batch releases are allowed.

D. Sample pump found OFF I Batch releases are NOT allowed.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 21 Which ONE of the following describes conditions that indicate RIA-54 is unable to perform its function AND if batch releases are allowed while the RIA is inoperable?

A. Counts do not increase when Source Check is performed / Batch releases are allowed.

B. Counts do not increase when Source Check is performed / Batch releases are NOT allowed.

C. Sample pump found OFF / Batch releases are allowed.

D. Sample pump found OFF / Batch releases are NOT allowed.

2010A NRC REACTOR OPERATOR EXAM Question 21 TIIG2

- cpw 059AK3.03 Accidental Liquid Radwaste Release Knowledge of the reasons for the following responses as they apply to the Accidental Liquid Radwaste Release:

Declaration that a radioactive-liquid monitor is inoperable.

(3.0/3.7)

K/A MATCH ANALYSIS i

NR( Oh to ask about functional vs operable Also can ask what is prevented by declaring RIA inoperable (meaning an unmonitored release).

Requires recognizing conditions that make RIA-54 unable to perform its function and actions required to prevent an unmonitored release when the RIA is non functional.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect. First part is plausible since it would be reasonable to assume that when RIA-54 is exposed to a source during its source check that indicated counts would increase however Limits and Precautions of PT1230/01 (RIA PT) clearly state that counts will NOT increase during a source check. Second part is correct.

B. Incorrect. First part is plausible as discussed in A above Second part is plausible since it is reasonable to assume that if the RIA is unable to provide protection then any releases through that pathway must be suspended. However batch releases (via dip sample with bucket) are still allowed since each discrete release batch is sampled for activity prior to release.

C. CORRECT. If the sample pump is OFF then the RIA is not able to monitor water in the TBS. With RIA-54 inoperable, SLC 16.11.3 allows continuing releases if sampled prior to each discrete release.

D. Incorrect. First part is correct. Second part is plausible since it is reasonable to assume that if the RIA is unable to provide protection then any releases through that pathway must be suspended. However batch releases (via dip sample with bucket) are still allowed since each discrete release batch is sampled for activity prior to release.

Technical Reference(s): PT1230101 (RIA PT) RAD-RIA SLC 16.11.3 Proposed references to be provided to applicants during examination:

NONE Learning Objective:

RAD-RIA R5 Question Source:

NEW Question History: N/A Question Cognitive Level: Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 21 T1/G2 - cpw 059AK3.03 Accidental Liquid Radwaste Release Knowledge of the reasons for the following responses as they apply to the Accidental Liquid Radwaste Release:

Declaration that a radioactive-liquid monitor is inoperable.

(3.0/3.7)

KIA MATCH ANALYSIS vs operable. Also can what i

(meaning an unmonitored release).

Requires recognizing conditions that make RIA-54 unable to perform its function and actions required to prevent an unmonitored release when the RIA is non-functional.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect. First part is plausible since it would be reasonable to assume that when RIA-54 is exposed to a source during its source check that indicated counts would increase however Limits and Precautions of PT/230/01 (RIA PT) clearly state that counts will NOT increase during a source check. Second part is correct.

B. Incorrect. First part is plausible as discussed in A above Second part is plausible since it is reasonable to assume that if the RIA is unable to provide protection then any releases through that pathway must be suspended. However batch releases (via dip sample with bucket) are still allowed since each discrete release batch is sampled for activity prior to release.

C. CORRECT. If the sample pump is OFF then the RIA is not able to monitor water in the TBS. With RIA-54 inoperable, SLC 16.11.3 allows continuing releases if sampled prior to each discrete release.

D. Incorrect. First part is correct. Second part is plausible since it is reasonable to assume that if the RIA is unable to provide protection then any releases through that pathway must be suspended. However batch releases (via dip sample with bucket) are still allowed since each discrete release batch is sampled for activity prior to release.

Technical Reference(s): PT/230/01 (RIA PT) RAD-RIA SLC 16.11.3 Proposed references to be provided to applicants during examination: NONE Learning Objective: RAD-RIA R5 Question Source: NEW Question History: N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 22 Unit 2 plant conditions:

3 of the 5 fire detectors in the West penetration room will be simultaneously removed for repair and/or replacement Based on the above conditions, which ONE of the following describes the compensatory actions required by SLC 16.9.6 (Fire Detection Instrumentation)?

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of removing the detectors....

A. Perform Channel Functional Test on remaining fire detectors AND backup fire suppression equipment is NOT required.

B. Establish an hourly fire watch AND backup fire suppression equipment is NOT required.

C. Perform Channel Functional Test on remaining fire detectors AND backup fire suppression equipment must be staged in the area.

D. Establish an hourly fire watch AND backup fire suppression equipment must be staged in the area.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 22 Unit 2 plant conditions:

3 of the 5 fire detectors in the West penetration room will be simultaneously removed for repair and/or replacement Based on the above conditions, which ONE of the following describes the compensatory actions required by SLC 16.9.6 (Fire Detection Instrumentation)?

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of removing the detectors....

A. Perform Channel Functional Test on remaining fire detectors AND backup fire suppression equipment is NOT required.

B. Establish an hourly fire watch AND backup fire suppression equipment is NOT required.

C. Perform Channel Functional Test on remaining fire detectors AND backup fire suppression equipment must be staged in the area.

D. Establish an hourly fire watch AND backup fire suppression equipment must be staged in the area.

2010A NRC REACTOR OPERATOR EXAM Question 22 T1/G2 067AA1.03 Plant fire on site Ability to operate and I or monitor the following as they apply to the Plant Fire on Site:

Bypassing of a fire zone detector.

(2.5/2.8)

KIA MATCH ANALYSIS Requires ability to take correct actions when bypassing (via removing) fire zone detectors ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is plausible since it would be correct if a Fire Barrier were determined to be inoperable per SLC 16.9.5. This would require verifying operability of detectors which would be done by performing the functional test. Additional plausibility since this test is normally a 31 day surveillance performed on the detectors. Second part is correct B. CORRECT: Per SLC 16.9.6 an hourly fire watch established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required if more than 50% of detectors in one location are inoperable. There are no requirements for backup fire suppression in this case.

C. Incorrect: First part is plausible since it would be correct if a Fire Barrier were determined to be inoperable per SLC 16.9.5. This would require verifying operability of detectors which would be done by performing the functional test. Additional plausibility since this test is normally a 31 day surveillance performed on the detectors. Second part is plausible since this could be correct if there were fire suppression equipment in the area inoperable.

D. Incorrect: Plausible since this could be correct if there were fire detection and suppression equipment in the area inoperable.

Technical Reference(s): SLC 16.9.5, 16.9.6 Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-ITS R7 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 22 T1/G2 -

2010A NRC REACTOR OPERATOR EXAM 067 AA 1.03 Plant fire on site Ability to operate and I or monitor the following as they apply to the Plant Fire on Site:

Bypassing of a fire zone detector.

(2.5/2.8)

KIA MATCH ANALYSIS Requires ability to take correct actions when bypassing (via removing) fire zone detectors ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is plausible since it would be correct if a Fire Barrier were determined to be inoperable per SLC 16.9.5. This would require verifying operability of detectors which would be done by performing the functional test. Additional plausibility since this test is normally a 31 day surveillance performed on the detectors. Second part is correct B. CORRECT: Per SLC 16.9.6 an hourly fire watch established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required if more than 50% of detectors in one location are inoperable. There are no requirements for backup fire suppression in this case.

C. Incorrect: First part is plausible since it would be correct if a Fire Barrier were determined to be inoperable per SLC 16.9.5. This would require verifying operability of detectors which would be done by performing the functional test. Additional plausibility since this test is normally a 31 day surveillance performed on the detectors. Second part is plausible since this could be correct if there were fire suppression equipment in the area inoperable.

D. Incorrect: Plausible since this could be correct if there were fire detection and suppression equipment in the area inoperable.

Technical Reference(s): SLC 16.9.5, 16.9.6 Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-ITS R7 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 23 Unit 3 plant conditions:

Loop A and Loop B SCMs = OF stable Core SCM = (-)5°F flashing with a red background Based on the above conditions, which ONE of the following describes the status of the reactor core?

A. saturated and covered B. saturated and partially uncovered C. superheated and covered D. superheated and partially uncovered 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 23 Unit 3 plant conditions:

Loop A and Loop B SCM's = O°F stable Core SCM = (_)5°F flashing with a red background Based on the above conditions, which ONE of the following describes the status of the reactor core?

A. saturated and covered B. saturated and partially uncovered C. superheated and covered D. superheated and partially uncovered

2010A NRC REACTOR OPERATOR EXAM Question 23 TI/G2

- cpw 074EG2.1.7 Inadequate Core Cooling Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(4.4/4.7)

KIA MATCH ANALYSIS Requires instrument interpretation to make an operational judgment of core performance ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since specific knowledge of the operation of the Subcooled Margin Monitors is required to determine that the core indicates superheated.

Saturated is plausible since both loops indicate saturated. The core being fully covered is plausible since both loops still indicate saturated.

B. Incorrect: Plausible since specific knowledge of the operation of the Subcooled Margin Monitors is required to determine that the core indicates superheated.

Saturated is plausible since both loops indicate saturated. Second part is correct.

C. Incorrect: First part is correct. Second part is plausible since both loops indicate saturated therefore in would be reasonable to deduce the core is still covered.

D. CORRECT: Core SCM indicating flashing negative numbers with red background is an indication of superheated conditions in the core. If the core is superheated then it is at least partially uncovered.

Technical Reference(s):

EAP-ICC, IC-RCI Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-ICC RI, IC-RCI R42 Question Source:

NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals Question 23 T1/G2 - cpw 2010A NRC REACTOR OPERATOR EXAM 074EG2.1.7Inadequate Core Cooling Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(4.4/4.7)

KIA MATCH ANALYSIS Requires instrument interpretation to make an operational judgment of core performance ANSWER CHOICE ANALYSIS Answer: 0 A. Incorrect: Plausible since specific knowledge of the operation of the Subcooled Margin Monitors is required to determine that the core indicates superheated.

Saturated is plausible since both loops indicate saturated. The core being fully covered is plausible since both loops still indicate saturated.

B. Incorrect: Plausible since specific knowledge of the operation of the Subcooled Margin Monitors is required to determine that the core indicates superheated.

Saturated is plausible since both loops indicate saturated. Second part is correct.

C. Incorrect: First part is correct. Second part is plausible since both loops indicate saturated therefore in would be reasonable to deduce the core is still covered.

D. CORRECT: Core SCM indicating flashing negative numbers with red background is an indication of superheated conditions in the core. If the core is superheated then it is at least partially uncovered.

Technical Reference(s): EAP-ICC,IC-RCI Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-ICC R1, IC-RCI R42 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 24 Unit 2 initial conditions:

Reactor power = 100%

Current conditions:

2SA2/B4 (RC AVERAGE TEMP HIGH/LOW) actuated Loop A Controlling Thot fails high (620°F)

Based on the above conditions, which ONE of the following describes the initial ICS response AND required operator actions to mitigate the failure?

ICS will Control Rods AND operator actions will include manually feedwater.

A. insert / increasing B. insert / decreasing C. withdraw / increasing D. withdraw / decreasing 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 24 Unit 2 initial conditions:

Reactor power = 100%

Current conditions:

2SA2/B4 (RC AVERAGE TEMP HIGH/LOW) actuated Loop 'A' Controlling Thot fails high (620°F)

Based on the above conditions, which ONE of the following describes the initial ICS response AND required operator actions to mitigate the failure?

ICS will ____ Control Rods AND operator actions will include manually feedwater.

A. insert / increasing B. insert / decreasing C. withdraw / increasing D. withdraw / decreasing

2010A NRC REACTOR OPERATOR EXAM Question 24 T1/G2 - cpw BAO2AK1.3 Loss of NNI-X Knowledge of the operational implications of the following concepts as they apply to the (Loss of NNI-X):

Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of NNI-X).

(3.8/3.8)

K/A MATCH ANALYSIS Requires knowledge of operational implications of plant indications of failed NNI for RCS Thot (to determine control rod response) and the remedial actions required by operators to stabilize the plant.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since for the Thot failure, indicated Tave would increase. If the operator were to respond to indicated Tave instead of actual RCS temp then increasing FDW would be the direction needed to restore Tave to near setpoint.

B. CORRECT: With Th failing high, indicated Tave increases and ICS causes control rods to drive (based on Tave error) in an attempt to restore (indicated)

Tave to setpoint. Since actual Tave is decreasing, Feedwater should be decreased to stop the temperature (and pressure) decrease.

C. Incorrect: First part is plausible since actual Tave will be decreasing therefore if ICS responded to actual Tave (or the other loop Tave) it would withdraw rods. Second part is plausible since for the Thot failure, indicated Tave would increase. If the operator were to respond to indicated Tave instead of actual RCS temp then increasing FDW would be the direction needed to restore Tave to near setpoint.

D. Incorrect: : First part is plausible since actual Tave will be decreasing therefore if ICS responded to actual Tave (or the other loop Tave) it would withdraw rods. Second part is correct.

Technical Reference(s): SAE-L074 Proposed references to be provided to applicants during examination: NONE Learning Objective: SAE-L074 R6 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 24 T1/G2 - cpw 2010A NRC REACTOR OPERATOR EXAM BA02AK1.3 Loss of NNI-X Knowledge of the operational implications of the following concepts as they apply to the (Loss of NNI-X):

Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of NNI-X).

(3.8/3.8)

KIA MATCH ANALYSIS Requires knowledge of operational implications of plant indications of failed NNI for RCS Thot (to determine control rod response) and the remedial actions required by operators to stabilize the plant.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since for the Thot failure, indicated Tave would increase. If the operator were to respond to indicated Tave instead of actual RCS temp then increasing FDW would be the direction needed to restore Tave to near setpoint.

B. CORRECT: With Th failing high, indicated Tave increases and ICS causes control rods to drive (based on Tave error) in an attempt to restore (indicated)

Tave to setpoint. Since actual Tave is decreasing, Feedwater should be decreased to stop the temperature (and pressure) decrease.

C. Incorrect: First part is plausible since actual Tave will be decreasing therefore if ICS responded to actual Tave (or the other loop Tave) it would withdraw rods. Second part is plausible since for the Thot failure, indicated Tave would increase. If the operator were to respond to indicated Tave instead of actual RCS temp then increasing FDW would be the direction needed to restore Tave to near setpoint.

D. Incorrect:: First part is plausible since actual Tave will be decreasing therefore if ICS responded to actual Tave (or the other loop Tave) it would withdraw rods. Second part is correct.

Technical Reference(s): SAE-L074 Proposed references to be provided to applicants during examination: NONE Learning Objective: SAE-L074 R6 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 25 Unit 1 initial conditions:

Reactor power = 25% slowly increasing Turbine trip Current conditions:

Reactor power = 22% decreasing Based on the above conditions, which ONE of the following describes the procedure(s) that will be utNized to direct plant activities AND the expected Steam Generator pressure (psig)?

A. EOPUNPPtab I 885 B. EOP UNPP tab I 1015 C. Plant Operating Procedures / 885 D. Plant Operating Procedures I 1015 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 25 Unit 1 initial conditions:

Reactor power = 25% slowly increasing Turbine trip Current conditions:

Reactor power = 22% decreasing Based on the above conditions, which ONE of the following describes the procedure(s) that will be utilized to direct plant activities AND the expected Steam Generator pressure (psig)?

A. EOP UNPP tab / 885 B. EOP UNPP tab / 1015 C. Plant Operating Procedures / 885 D. Plant Operating Procedures / 1015

2010A NRC REACTOR OPERATOR EXAM Question 25 TIIGI

- cpw BAO4AA2.1 Turbine Trip Ability to determine and interpret the following as they apply to the (Turbine Trip):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(3.3/3.7)

K/A MATCH ANALYSIS Requires selecting the appropriate procedures following a turbine trip ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: The turbine trip should not have resulted in a reactor trip since power was

<30%. First part is plausible since this would be the correct answer if initial reactor power had been >30% and the reactor did not trip. Second part is correct and would still be plausible for UNPP since the Gen. breakers tripping open would make TLSF false and therefore remove the 50 psi bias thus TBVs would be controlling at setpoint.

B. Incorrect: The turbine trip should not have resulted in a reactor trip since power was

<30%. First part is plausible since this would be the correct answer if initial reactor power had been >30% and the reactor did not trip. Second part is plausible with UNPP since there are conditions where AMSAC would actuate and still send the 125 psi bias to the TBVs which would result in them controlling at 1010 psig.

C. CORRECT: With Rx power < 30% a turbine trip does not result in a Rx trip. The plant would run back to 20% CTP via an ICS runback due to both Gen bkrs open. Once at 20% with turbine off line, either the Shutdown procedure or the Startup procedure would be implemented to direct the plant. With no Rx trip, the TBVs would control at setpoint (885) since the 50 psi bias to the setpoint would be removed by ICS when both Generator breakers open.

D. Incorrect: First part is correct. Second part is plausible since there are conditions where AMSAC would actuate and still send the 125 psi bias to the TBVs which would result in them controlling at 1010 psig.

Technical Reference(s):

IC-RPS, EAP-SA, APII Proposed references to be provided to applicants during examination:

NONE Learning Objective:

IC-RPS R3, EAP-SA RI Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 25 T1/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM BA04AA2.1 Turbine Trip Ability to determine and interpret the following as they apply to the (Turbine Trip):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(3.3/3.7)

KIA MATCH ANALYSIS Requires selecting the appropriate procedures following a turbine trip ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: The turbine trip should not have resulted.in a reactor trip since power was

< 30%. First part is plausible since this would be the correct answer if initial reactor power had been >30% and the reactor did not trip. Second part is correct and would' still be plausible for UNPP since the Gen. breakers tripping open would make TLSF false and therefore remove the 50 psi bias thus TBV's would be controlling at setpoint.

B. Incorrect: The turbine trip should not have resulted in a reactor trip since power was

< 30%. First part is plausible since this would be the correct answer if initial reactor power had been >30% and the reactor did not trip. Second part is plausible with UNPP since there are conditions where AMSAC would actuate and still send the 125 psi bias to the TBV's which would result in them controlling at 1010 psig.

C. CORRECT: With Rx power < 30% a turbine trip does not result in a Rx trip. The plant would run back to 20% CTP via an ICS runback due to both Gen bkrs open. Once at 20% with turbine off line, either the Shutdown procedure or the Startup procedure would be implemented to direct the plant. With no Rx trip, the TBV's,would control at setpoint (885) since the 50 psi bias to the setpoint would be removed by ICS when both Generator breakers open.

D. Incorrect: First part is correct. Second part is plausible since there are conditions where AMSAC would actuate and still send the 125 psi bias to the TBV's which would result in them controlling at 1010 psig.

Technical Reference(s):

IC-RPS, EAP-SA, AP/1 Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-RPS R3, EAP-SA R1 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 26 Unit I plant conditions:

Reactor power = 100%

2SA-181A1 1 (TURBINE BSMT WATER EMERGENCY HIGH LEVEL) actuated NEO reports water level in Turbine Building basement increasing Based on the above conditions, which ONE of the following describes the required actions directed by AP/1 0 (Turbine Building Flood)?

Manually trip the reactor and...

A. align Station ASW pump for use B. start all Main Vacuum pumps C. secure all operating CCW pumps D. place all HPSW pump switches to OFF 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 26 Unit 1 plant conditions:

Reactor power = 100%

2SA-18/A11 (TURBINE BSMT WATER EMERGENCY HIGH LEVEL) actuated NEO reports water level in Turbine Building basement increasing Based on the above conditions, which ONE of the following describes the required actions directed by AP/1 0 (Turbine Building Flood)?

Manually trip the reactor and...

A. align Station ASW pump for use B. start all Main Vacuum pumps C. secure all operating CCW pumps D. place all HPSW pump switches to "OFF"

2010A NRC REACTOR OPERATOR EXAM Question 26 TIIG2 - cpw BAO7AA2.2 Flooding Ability to determine and interpret the following as they apply to the (Flooding):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

(3.3/3.7)

K/A MATCH ANALYSIS Knowledge of mitigation strategy is required to ensure adherence to APIIO.

Adhering to APIIO ensures operation within license limitations set by TS 5.4 requiring implementing procedures recommended by Reg Guide 1.33.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since this would be a correct answer if Unit I was on LPI DHR.

In that case since Unit I is not available to provide alternate source of feedwater the Station ASW Pump is started to make Station ASW available to Units 2&3.

B. Incorrect: Plausible since there is NOTE in AP/lO (for step 4.2) which explains that actions taken will result in a loss of condenser vacuum.

C. CORRECT: APIIO directs securing all CCWPs and closing discharge valves to isolate the intake canal from the leak.

D. Incorrect: Plausible since HPSW pumps are located in TBB and are susceptible to flooding. AP/lO does provide guidance on what to do if HPSW is lost. Additionally plausible since EWST is available to supply required HPSW in absence of operating HPSW pumps. This makes the HPSW supply unique and adds plausibility to securing the HPSW pumps.

Technical Reference(s): APIIIAII700IOIO Proposed references to be provided to applicants during examination:

NONE Learning Objective: EAP-APG R8 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 26 T1/G2 - cpw BA07 AA2.2 Flooding 2010A NRC REACTOR OPERATOR EXAM Ability to determine and interpret the following as they apply to the (Flooding):

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

(3.3/3.7)

KIA MATCH ANALYSIS Knowledge of mitigation strategy is required to ensure adherence to AP/10.

Adhering to AP/10 ensures operation within license limitations set by TS 5.4 requiring implementing procedures recommended by Reg Guide 1.33.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since this would be a correct answer if Unit 1 was on LPI DHR.

In that case since Unit 1 is not available to provide alternate source of feedwater the Station ASW Pump is started to make Station ASW available to Units 2&3.

B. Incorrect: Plausible since there is NOTE in AP/10 (for step 4.2) which explains that actions taken will result in a loss of condenser vacuum.

C. CORRECT: AP/10 directs securing all CCWP's and closing discharge valves to isolate the intake canal from the leak.

D. Incorrect: Plausible since HPSW pumps are located in TBB and are susceptible to flooding. AP/10 does provide guidance on what to do if HPSW is lost. Additionally plausible since EWST is available to supply required HPSW in absence of operating HPSW pumps. This makes the HPSW supply unique and adds plausibility to securing the HPSW pumps.

Technical Reference(s): AP/1/A11700/010 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-APG R8 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 27 Unit 1 plant conditions:

Core SCM = 0°F LOCA Cooldown tab in progress CETCs = 395T slowly decreasing I LP-1 03 (POST LOCA BORON DILUTE) will NOT open Based on the above conditions, which ONE of the following valves is required to be opened in accordance with the LOCA CD tab to establish post LOCA boron dilution flow?

A. 1LP-3 B. 1LP-19 C. 1LP-104 D. 1LP-105 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 27 Unit 1 plant conditions:

Core SCM = OaF LOCA Cooldown tab in progress CETC's = 395°F slowly decreasing 1 LP-1 03 (POST LOCA BORON DILUTE) will NOT open Based on the above conditions, which ONE of the following valves is required to be opened in accordance with the LOCA CD tab to establish post LOCA boron dilution flow?

A. 1 LP-3 B. 1LP-19 C. 1LP-104 D. 1LP-105

2010A NRC REACTOR OPERATOR EXAM Question 27 TIIG2 - cpw BEO3EG2.2.3 Inadequate Subcooling Margin Knowledge of the design, procedural, and operational differences between units.

(3.8/3.9)

K/A MATCH ANALYSIS Requires knowledge of design, procedural, and operational differences between units following a SBLOCA when aligning Post Loca Boron Dilution flowpath. Due to differences in the routing of LPI piping, Unit I has a different alternate Post Loca Boron dilution flowpath than does units 2 and 3. Unit I has LP-105 in its flowpath where Units 2 & 3 would use LP-19.

ANSWER CHOICE ANALYSIS Answer:

D A. Incorrect: Plausible since 1 LP-3 is in the normal DHR drop line however the tap for the alternate PLBD flowpath is between 1 LP-2 & 1 LP-3 therefore 1 LP-3 is not required to be opened. Additional plausibility is added due to the fact that LP-3 is a valve required to be opened for Units 2 & 3.

B. Incorrect: Plausible since LP-19 would be correct for Units 2 & 3 however Unit 1 has a unique alternate path that does not use 1 LP-1 9.

C. Incorrect: Plausible since 1LP-104 isa valve in the normal PLBD flowpath however it is in series with 1 LP-1 03 therefore the failure of 1 LP-1 03 to open renders 1 LP-1 04 useless.

D. CORRECT: For Unit I, another drain line and motor operated valve, ILP-105, is installed below LP-I and LP-2 to provide a second flow path to the Reactor Building Emergency Sump. The addition of ILP-I05 was required in this flow path due to the arrangement of the Decay Heat Drop Line on Unit I. On Unit I the Drop Line does not drop straight to the suction of the LPI pumps as it does on units 2 and 3, but instead, curves back upwards before reaching the pumps, in effect, forming a loop seal. This would prevent a gravity drain path from the hot leg to the LPI pumps suction header as it exists on Units 2 and 3 and requires using ILP-105 for the alternate PLBD flowpath.

Technical Reference(s): PNS-LPI, EOP-LOCACD Proposed references to be provided to applicants during examination:

NONE Learning Objective:

PNS-LPI R27, R28 Question Source:

NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals Question 27 T1/G2 - cpw 2010A NRC REACTOR OPERATOR EXAM BE03EG2.2.3 Inadequate Subcooling Margin Knowledge of the design, procedural, and operational differences between units.

(3.8/3.9)

KIA MATCH ANALYSIS Requires knowledge of design, procedural, and operational differences between units following a SBLOCA when aligning Post Loca Boron Dilution flowpath. Due to differences in the routing of LPI piping, Unit 1 has a different alternate Post Loca Boron dilution flowpath than does units 2 and 3. Unit 1 has LP-105 in its flowpath where ~nits 2 & 3 would use LP-19.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since 1 LP-3 is in the normal DHR drop line however the tap for the alternate PLBD flowpath is between 1 LP-2 & 1 LP-3 therefore 1 LP-3 is not required to be opened. Additional plausibility is added due to the fact that LP-3 is a valve required to be opened for Units 2 & 3.

B. Incorrect: Plausible since LP-19 would be correctfor Units 2 & 3 however Unit 1 has a unique alternate path that does not use 1 LP-19.

C. I ncorrect: Plausible since 1 LP-1 04 is a valve in the normal PLBD flowpath however it is in series with 1 LP-1 03 therefore the failure of 1 LP-1 03 to open renders 1 LP-1 04 useless.

D. CORRECT: For Unit 1, another drain line and motor operated valve, 1 LP-1 OS, is installed below LP-1 and LP-2 to provide a second flow path to the Reactor Building Emergency Sump. The addition of 1LP-105 was required in this flow path due to the arrangement of the Decay Heat Drop Line on Unit 1. On Unit 1 the Drop Line does not drop straight to the suction of the LPI pumps as it does on units 2 and 3, but instead, curves back upwards before reaching the pumps, in effect, forming a loop seal. This would prevent a gravity drain path from the hot leg to the LPI pumps suction header as it exists on Units 2 and 3 and requires using 1LP-105 for the alternate PLBD flowpath.

Technical Reference(s): PNS-LPI, EOP-LOCACD Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-LPI R27, R28 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 28 Unit 3 initial conditions:

Time = 0300 Reactor power = 100%

Reactor trip CT-3 Lockout occurs Current conditions:

MFB re-energized 6900V power still unavailable HPI system leak downstream of 3HP-31 occurs 3A1 RCP SI flow = 3.9 gpm slowly decreasing 3A2 RCP SI flow = 3.7 gpm slowly decreasing 3B1 RCP SI flow = 3.5 gpm slowly decreasing 3B2 RCP SI flow = 3.4 gpm slowly decreasing Seal Inlet Header Flow 40 gpm stable Based on the above conditions, which ONE of the following describes the status of the following RCP support systems valve(s) two minutes later?

A. 3HP-21 has closed (ONLY)

B. 3HP-31 has opened (ONLY)

C. ALL individual seal return valves have closed and 3HP-21 has closed D. ALL individual seal return valves have closed and 3HP-31 has opened 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 28 Unit 3 initial conditions:

Time = 0300 Reactor power = 100%

Reactor trip CT -3 Lockout occurs Current conditions:

MFB re-energized 6900V power still unavailable HPI system leak downstream of 3HP-31 occurs 3A 1 RCP SI flow = 3.9 gpm slowly decreasing 3A2 RCP SI flow = 3.7 gpm slowly decreasing 3B1 RCP SI flow = 3.5 gpm slowly decreasing 3B2 RCP SI flow = 3.4 gpm slowly decreasing Seal Inlet Header Flow = 40 gpm stable Based on the above conditions, which ONE of the following describes the status of the following RCP support systems valve(s) two minutes later?

A. 3HP-21 has closed (ONLY)

B. 3HP-31 has opened (ONLY)

C. ALL individual seal return valves have closed and 3HP-21 has closed D. ALL individual seal return valves have closed and 3HP-31 has opened

2010A NRC REACTOR OPERATOR EXAM Question 28 T2IGI -cpw 003K4.1 1 Reactor Coolant Pump Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following:

Isolation valve interlocks (3.0/3.0)

K/A MATCH ANALYSIS Requires knowledge of RCP Seal Injection flow isolation valve interlocks and the differences in these interlocks between units.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since 3HP-21 would be closed however the individual seal return valves would also be closed. The individual SR valves not being closed is plausible since total SI flow is still normal and on Unit I the individual SR valves close based on total SI flow and not individual SI flows.

B. Incorrect. Plausible since seal injection flow to each RCP is low and 3HP-31 is the seal injection supply valve. 3HP-31 controls off of total seal injection flow so as long as total flow is at setpoint (40 gpm) 3HP-31 would not have attempted to increase SI flow. Additional plausibility from the fact that 2,3HP-3lwiII automatically close if SI flow is <4 gpm on each pump for> one minute.

C. CORRECT: There is a RCP Seal Return interlock that will automatically close Seal Return isolation valves (SRIVs) when SI flow is <4 gpm so they will be closed. On Units 2&3 if seal injection flow to ALL RCPs <4 gpm/RCP for> I mm then, (2)(3)HP-2I automatically closes (UI HP-3I must be manually closed).

D. Incorrect: The first part is correct. Second part is plausible since seal injection flow to each RCP is low and 3HP-31 is the seal injection supply valve. 3HP-31 controls off of total seal injection flow so as long as total flow is at setpoint (40 gpm) 3HP-31 would not have attempted to increase SI flow. Additional plausibility from the fact that 2,3HP-3lwill automatically close if SI flow is <4 gpm on each pump for> one minute.

Technical Reference(s): 0P12,31A/I 104/002 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-HPI R22 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals Question 28 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 003K4.11 Reactor Coolant Pump Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following:

Isolation valve interlocks (3.0/3.0)

KIA MATCH ANALYSIS Requires knowledge of RCP Seal Injection flow isolation valve interlocks and the differences in these interlocks between units.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since 3HP-21 would be closed however the individual seal return valves would also be closed. The individual SR valves not being closed is plausible since total SI flow is still normal and on Unit 1 the individual SR valves close based on total SI flow and not individual SI flows.

B. Incorrect. Plausible since seal injection flow to each RCP is low and 3HP-31 is the seal injection supply valve. 3HP-31 controls off of total seal injection flow so as long as total flow is at setpoint (40 gpm) 3HP-31 would not have attempted to increase SI flow. Additional plausibility from the fact that 2,3HP-31will automatically close if SI flow is <4 gpm on each pump for> one minute.

C. CORRECT: There is a RCP Seal Return interlock that will automatically close Seal Return isolation valves (SRIVs) when SI flow is < 4 gpm so they will be closed. On Units 2&3 if seal injection flow to ALL RCPs < 4 gpm/RCP for> 1 min then, (2)(3)HP-21 automatically closes (U1 HP-31 must be manually closed).

D. Incorrect: The first part is correct. Second part is plausible since seal injection flow to each RCP is low and 3HP-31 is the seal injection supply valve. 3HP-31 controls off of total seal injection flow so as long as total flow is at setpoint (40 gpm) 3HP-31 would not have attempted to increase SI flow. Additional plausibility from the fact that 2,3HP-31will automatically close if SI flow is <4 gpm on each pump for> one minute.

Technical Reference(s): OP/2,3/Al1104/002 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-HPI R22 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 29 Unit 1 plant conditions:

Reactor power = 100%

1CC-8 (CC RETURN PENT (54) OUTSIDE BLOCK) fails closed 1LPSW-6 (UNIT 1 RCP COOLERS SUPPLY) fails closed Based on the above conditions, which ONE of the failed valves will require ALL RCPs to be secured in accordance with AP/16 (Abnormal Reactor Coolant Pump Operation) and why?

A.

1 CC-8 / due to high RCP motor stator temperatures B.

1 CC-8 / due to high RCP radial bearing temperatures C.

1 LPSW-6 / due to high RCP motor stator temperatures D. ILPSW-6 / due to high RCP radial bearing temperatures 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 29 Unit 1 plant conditions:

Reactor power = 100%

1 CC-8 (CC RETURN PENT (54) OUTSIDE BLOCK) fails closed 1 LPSW-6 (UNIT 1 RCP COOLERS SUPPLY) fails closed Based on the above conditions, which ONE of the failed valves will require ALL RCPs to be secured in accordance with AP/16 (Abnormal Reactor Coolant Pump Operation) and why?

A. 1 CC-8 / due to high RCP motor stator temperatures B. 1 CC-8 / due to high RCP radial bearing temperatures C. 1 LPSW-6 / due to high RCP motor stator temperatures D. 1 LPSW-6 / due to high RCP radial bearing temperatures

2010A NRC REACTOR OPERATOR EXAM Question 29 T2IGI

- okm 003K6.04 Reactor Coolant Pump Knowledge of the effect of a loss or malfunction on the following will have on the RCPs:

Containment isolation valves affecting RCP operation.

(2.8/3.1)

K/A MATCH ANALYSIS Both CC-8 and LPSW-15 are Containment isolation valves and with both failed closed the candidate must assess the effect of the failure on RCP operation ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since CC is one of the cooling mediums that provides cooling to RCP5. Additionally plausible since there is a required immediate trip of RCPs if stator temps reach 295° F in AP/1 6 however CC does not cool the motor stators.

B. Incorrect: Plausible since CC is one of the cooling mediums that provides cooling to RCPs. Additionally plausible if there is a misconception regarding cooling of the RCP bearings. Seal Injection water 2.0 gpm is still available to cool the RCP bearings even though CC has been lost. AP/1 6 requires tripping RCP if radial brg temp reaches 225°F.

C. CORRECT: LPSW via LPSW-6 supplies cooling water to the oil coolers and stator air coolers. If RCPs continued to run without oil and motor cooling they would all be damaged. AP/16 requires tripping RCP when motor stator temps reach 295°F.

D. Incorrect: First part is correct. Plausible since AP/1 6 requires tripping RCP if radial brg temp reaches 225°F. Additional plausibility if there is a misconception regarding cooling of the RCP bearings. Seal Injection water 2.0 gpm is still available to cool the RCP bearings even though CC has been lost.

Technical Reference(s): PNS-CPM Pg 6 APII6 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-CPM RI, 19 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension or Analysis 2010A NRC REACTOR OPERATOR EXAM Question 29 T2/G1 - okm 003K6.04 Reactor Coolant Pump Knowledge of the effect of a loss or malfunction on the following will have on the RCPs:

Containment isolation valves affecting RCP operation.

(2.8/3.1 )

KIA MATCH ANALYSIS Both CC-B and LPSW-15 are Containment isolation valves and with both failed closed the candidate must assess the effect of the failure on RCP operation ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since CC is one of the cooling mediums that provides cooling to RCPs. Additionally plausible since there is a required immediate trip of RCP's if stator temps reach 295°F in AP/16 however CC does not cool the motor stators.

B. Incorrect: Plausible since CC is one of the cooling mediums that provides cooling to RCPs. Additionally plausible if there is a misconception regarding cooling of the RCP bearings. Seal Injection water -2.0 gpm is still available to cool the RCP bearings even though CC has been lost. AP/16 requires tripping RCP if radial brg temp reaches 225°F.

C. CORRECT: LPSW via LPSW-6 supplies cooling water to the oil coolers and stator air coolers. If RCPs continued to run without oil and motor cooling they would all be damaged. AP/16 requires tripping RCP when motor stator temps reach 295°F.

D. Incorrect: First part is correct. Plausible since AP/16 requires tripping RCP if radial brg temp reaches 225°F. Additional plausibility if there is a misconception regarding cooling of the RCP bearings. Seal Injection water -2.0 gpm is still available to cool the RCP bearings even though CC has been lost.

Technical Reference(s): PNS-CPM Pg 6 AP/16 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-CPM R1, 19 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 30 Unit 1 initial conditions:

Reactor power = 100%

IA CC pump operating Current conditions:

1 CC-7 fails closed Based on the above conditions, which ONE of the following describes the expected plant response?

A. The reactor will automatically trip and NEITHER CC Pump will be operating B. The reactor will automatically trip and BOTH CC Pumps will be operating C. Letdown will be automatically isolated and NEITHER CC Pump will be operating D. Letdown will be automatically isolated and BOTH CC Pumps will be operating 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 30 Unit 1 initial conditions:

Reactor power = 100%

1 A CC pump operating Current conditions:

1 CC-7 fails closed Based on the above conditions, which ONE of the following describes the expected plant response?

A. The reactor will automatically trip and NEITHER CC Pump will be operating B. The reactor will automatically trip and BOTH CC Pumps will be operating C. Letdown will be automatically isolated and NEITHER CC Pump will be operating D. Letdown will be automatically isolated and BOTH CC Pumps will be operating

2010A NRC REACTOR OPERATOR EXAM Question 30 T2IGI

- cpw 004K1.36 Chemical and Volume Control System Knowledge of the physical connections and/or cause-effect relationships between the CVCS and the following systems:

CCWS (2.6/2.8)

KIA MATCH ANALYSIS Requires knowledge of cause-effect relationship of loss of CC flow on the HPI system ANSWER CHOICE ANALYSIS Answer: C A. INCORRECT: Plausible since a loss of CC would ultimately result in a required reactor trip once CRD temps reach 180 degrees however it is a required manual trip and not an automatic trip. Second part is correct since both CC pumps trip when either CC-7 or CC-8 go closed.

8. Incorrect: Plausible since a loss of CC would ultimately result in a required reactor trip once CRD temps reach 180 degrees however it is a required manual trip and not an automatic trip. Second part is plausible since there is an auto start of the standby CC pump on lo CC flow however if either CC-7 or 8 close, both CC pumps automatically trip.

C. CORRECT: CC is the cooling medium for the letdown coolers. If CC is lost, letdown temperature would rise very quickly. If the letdown temperature reaches 130°F a high temperature stat-alarm will sound and at 135°F the letdown isolation valve, HP-5, will close. This happens in 1 minute with a total loss of CC flow. CC-7 closing would result in a total loss of CC flow to the letdown coolers.

D. Incorrect: First part is correct. Second part is plausible because if CC flow had decreased due to reasons other than CC-7 or CC-8 failing closed then both CC pumps would be operating since the S/B pump would start on low flow. If either CC-7 or 8 close, both CC pumps automatically trip.

Technical Reference(s): PNS-HPI, PNS-CC Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-HPI R22, PNS-CC R18,19 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals Question 30 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 004K1.36 Chemical and Volume Control System Knowledge of the physical connections and/or cause-effect relationships between the CVCS and the following systems:

CCWS (2.6/2.8)

K/A MATCH ANALYSIS Requires knowledge of cause-effect relationship of loss of CC flow on the HPI system ANSWER CHOICE ANALYSIS Answer: C A. INCORRECT: Plausible since a loss of CC would ultimately result in a required reactor trip once CRD temps reach 180 degrees however it is a required manual trip and not an automatic trip. Second part is correct since both CC pumps trip when either CC-7 or CC-8 go closed.

B. Incorrect: Plausible since a loss of CC would ultimately result in a required reactor trip once CRD temps reach 180 degrees however it is a required manual trip and not an automatic trip. Second part is plausible since there is an auto start of the standby CC pump on 10 CC flow however if either CC-7 or 8 close, both CC pumps automatically trip.

C. CORRECT: CC is the cooling medium for the letdown coolers. If CC is lost, letdown temperature would rise very quickly. If the letdown temperature reaches 130°F a high temperature stat-alarm will sound and at 135°F the letdown isolation valve, HP-5, will close. This happens in - 1 minute with a total loss of CC flow. CC-7 closing would result in a total loss of CC flow to the letdown coolers.

D. Incorrect: First part is correct. Second part is plausible because if CC flow had d.ecreased due to reasons other than CC-7 or CC-8 failing closed then both CC pumps would be operating since the SIB pump would start on low flow. If either CC-7 or 8 close, both CC pumps automatically trip.

Technical Reference(s): PNS-HPI, PNS-CC Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-HPI R22, PNS-CC R18,19 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 31 Unit 3 initial conditions:

ReactorinMODE4 LPI DHR alignment for cooldown in progress Current conditions:

3LP-1 (LPI RETURN BLOCK FROM RCS) will not open 3LP-12 (3A LPI COOLER OUTLET) failed closed Based on the above conditions, which ONE of the following describes the effect of the failures on ECCS-LPI train availability?

The 3LP-1 failure impact ECCS-LPI train availability and the failure of 3LP-12 impact ECCS-LPI train availability.

A. Does / Does B. Does / Does NOT C. Does NOT / Does D. Does NOT / Does NOT 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 31 Unit 3 initial conditions:

Reactor in MODE 4 LPI DHR alignment for cooldown in progress Current conditions:

3LP-1 (LPI RETURN BLOCK FROM RCS) will not open 3LP-12 (3A LPI COOLER OUTLET) failed closed Based on the above conditions, which ONE of the following describes the effect of the failures on ECCS-LPI train availability?

The 3LP-1 failure impact ECCS-LPI train availability and the failure of 3LP-12

__ impact ECCS-LPI train availability.

A. Does / Does B. Does / Does NOT C. Does NOT / Does D. Does NOT / Does NOT

2010A NRC REACTOR OPERATOR EXAM Question 31 T2/G1

- cpw 005K3.05 Residual Heat Removal System (RHRS)

Knowledge of the effect that a loss or malfunction of the RHRS will have on the following: ECCS (3.7/3.8)

K/A MATCH ANALYSIS Requires knowledge of the effect that malfunctions in the decay heat cooler/train will have on ECCS availability.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since 3LP-1 is a suction to the LPI pumps however it is the Decay Heat Removal suction. Since ECCS suction is from either the BWST or RBES, a failure of LP-1 has no impact on ECCS availability. Second part is correct.

B. Incorrect: First part is plausible since 3LP-1 is a suction to the LPI pumps however it is the Decay Heat Removal suction. Since ECCS suction is from either the BWST or RBES, a failure of LP-1 has no impact on ECCS availability. Second part is plausible because unit 3 has LPI bypasses around the LPI cooler so it would be plausible to determine the ECCS train is still available since I would still be able to get flow down the header using the cooler bypass valve.

C. CORRECT: 3LP-1 is a suction valve to the LPI pumps however it is the Decay Heat Removal suction. Since ECCS suction is from either the BWST or RBES, a failure of LP-1 has no impact on ECCS availability. 3LP-12 is a cooler outlet valve. Since flow through the cooler is an integral part of the ECCS train (allows LPSW to cool RBES water), failure of the valve closed renders the LPI train unavailable (and inoperable).

D. Incorrect: First part is correct. Second part is plausible because unit 3 has LPI bypasses around the LPI cooler so it would be plausible to determine the ECCS train is still available since I would still be able to get flow down the header using the cooler bypass valve.

Technical Reference(s):

LPI system drawing, PNS-LPI Proposed references to be provided to applicants during examination: NONE Learning Objective:

PNS-LPI R5, R29, R30, R14 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension or Analysis 2010A NRC REACTOR OPERATOR EXAM Question 31 T2/G1 - cpw 005K3.05 Residual Heat Removal System (RHRS)

Knowledge of the effect that a loss or malfunction of the RHRS will have on the following: ECCS (3.7/3.8)

K/A MATCH ANALYSIS Requires knowledge of the effect that malfunctions in the decay heat cooler/train will have on ECCS availability.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since 3LP-1 is a suction to the LPI pumps however it is the Decay Heat Removal suction. Since ECCS suction is from either the BWST or RBES, a failure of LP-1 has no impact on ECCS availability. Second part is correct.

B. Incorrect: First part is plausible since 3LP-1 is a suction to the LPI pumps however it is the Decay Heat Removal suction. Since ECCS suction is from either the BWST or RBES, a failure of LP-1 has no impact on ECCS availability. Second part is plausible because unit 3 has LPI bypasses around the LPI cooler so it would be plausible to determine the ECCS train is still available since I would still be able to get flow down the header using the cooler bypass valve.

C. CORRECT: 3LP-1 is a suction valve to the LPI pumps however it is the Decay Heat Removal suction. Since ECCS suction is from either the BWST or RBES, a failure of LP-1 has no impact on ECCS availability. 3LP-12 is a cooler outlet valve. Since flow through the cooler is an integral part of the ECCS train (allows LPSW to cool RBES water), failure of the valve closed renders the LPI train unavailable (and inoperable).

D. Incorrect: First part is correct. Second part is plausible because unit 3 has LPI bypasses around the LPI cooler so it would be plausible to determine the ECCS train is still available since I would still be able to get flow down the header using the cooler bypass valve.

Technical Reference(s): LPI system drawing, PNS-LPI Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-LPI R5, R29, R30, R14 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 32 Which ONE of the following describes the hicihest RCS pressure (psig) at which the I LP-I (LPI RETURN BLOCK FROM RCS) pressure interlock will allow I LP-1 to be opened and the reason I LP-I has a pressure interlock?

A. 365 I prevent overpressurizing LPI suction piping B. 365 I ensure delta p across I LP-I will allow it to open C. 420 I prevent overpressurizing LPI suction piping D. 420 I ensure delta p across 1 LP-l will allow it to open 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 32 Which ONE of the following describes the highest RCS pressure (psig) at which the 1 LP-1 (LPI RETURN BLOCK FROM RCS) pressure interlock will allow 1 LP-1 to be opened and the reason 1 LP-1 has a pressure interlock?

A. 365 / prevent overpressurizing LPI suction piping B. 365 / ensure delta p across 1 LP-1 will allow it to open C. 420 / prevent overpressurizing LPI suction piping D. 420 / ensure delta p across 1 LP-1 will allow it to open

2010A NRC REACTOR OPERATOR EXAM Question 32 T2/G1 cpw 005K4.O1 Residual Heat Removal System (RHRS)

Knowledge of RHRS design feature(s) and/or interlock(s) which provide for the following:

Overpressure mitigation system (3.0/3.2)

KIA MATCH ANALYSIS Requires knowledge of how LPI overpressure protection is accomplished. This is done by an interlock that prevents placing LPI DHR piping in service prior to being below 400 psi, ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: The I LP-I interlock prevents system overpressurization by preventing ILP-I from being opened when WR RCS pressure (via the Amphenol connector) is >400 psig.

B. Incorrect: First part is correct. Second part is plausible because waiting on a lower RCS pressure to open 1 LP-1 would in fact lower the dp across LP-1 when it is opened and there are many different valves throughout the plant where we take specific actions to ensure dp is low enough across a valve before we try to open it (Ex. MSCVs, FDW valves, etc.).

C. Incorrect: Plausible since the I LP-1 interlock prevents 1 LP-1 from being opened when WR RCS pressure (via the Amphenol connector) is >400 psig. Second part is correct.

D. Incorrect: First part is plausible since the I LP-1 interlock prevents 1 LP-1 from being opened when WR RCS pressure (via the Amphenol connector) is >400 psig. Second part is plausible because waiting on a lower RCS pressure to open I LP-1 would in fact lower the dp across LP-1 when it is opened and there are many different valves throughout the plant where we take specific actions to ensure dp is low enough across a valve before we try to open it (Ex. MSCVs, FDW valves, etc.).

Technical Reference(s): PNS-LPI pgs 52 & 53 Proposed references to be provided to applicants during examination:

NONE Learning Objective: PNS-LPI R16 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Memory or Fundamental Knowledge 2010A NRC REACTOR OPERATOR EXAM Question 32 T2/G1 - cpw 005K4.01 Residual Heat Removal System (RHRS)

Knowledge of RHRS design feature(s) and/or interlock(s) which provide for the following:

Overpressure mitigation system (3.0/3.2)

K/A MATCH ANALYSIS Requires knowledge of how LPI overpressure protection is accomplished. This is done by an interlock that prevents placing LPI DHR piping in service prior to being below 400 psi, ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: The 1 LP-1 interlock prevents system overpressurization by preventing 1 LP-1 from being opened when WR RCS pressure (via the Amphenol connector) is >400 psig.

B. Incorrect: First part is correct. Second part is plausible because waiting on a lower RCS pressure to open 1 LP-1 would in fact lower the dp across LP-1 when it is opened and there are many different valves throughout the plant where we take specific actions to ensure dp is low enough across a valve before we try to open it (Ex. MSCV's, FDW valves, etc.).

C. Incorrect: Plausible since the 1 LP-1 interlock prevents 1 LP-1 from being opened when WR RCS pressure (via the Amphenol connector) is >400 psig. Second part is correct.

D. Incorrect: First part is plausible since the 1 LP-1 interlock prevents 1 LP-1 from being opened when WR RCS pressure (via the Amphenol connector) is >400 psig. Second part is plausible because waiting on a lower RCS pressure to open 1 LP-1 would in fact lower the dp across LP-1 when it is opened and there are many different valves throughout the plant where we take specific actions to ensure dp is low enough across a valve before we try to open it (Ex. MSCV's, FDW valves, etc.).

Technical Reference(s}: PNS-LPI pgs 52 & 53 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-LPI R16 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge

2010A NRC REACTOR OPERATOR EXAM I POINT Question 33 Unit 1 initial conditions:

Reactor power = 100%

Current conditions:

RCS pressure = 1350 psig decreasing Reactor Building pressure = 4.8 psig increasing ES Channel 2 did NOT actuate Based on the above conditions, which ONE of the following describes ALL safety injection pumps that have AUTOMATICALLY started?

A.

1A HPI, lB HPI, 1A LPI, lB LPI B. 1A HPI, 1C HPI, 1A LPI, lB LPI C. 1A HPI, lB HPI, IA LPI ONLY D. IA HPI, 1A LPI ONLY 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 33 Unit 1 initial conditions:

Reactor power = 100%

Current conditions:

RCS pressure = 1350 psig decreasing Reactor Building pressure = 4.8 psig increasing ES Channel 2 did NOT actuate Based on the above conditions, which ONE of the following describes ALL safety injection pumps that have AUTOMATICALLY started?

A. 1A HPI, 1 B HPI, 1A LPI, 1 B LPI B. 1A HPI, 1 C HPI, 1A LPI, 1 B LPI C. 1A HPI, 1 B HPI, 1A LPI ONLY D. 1A HPI, 1A LPI ONLY

2010A NRC REACTOR OPERATOR EXAM Question 33 T2/GI

- cpw 006A3.05 Emergency Core Cooling Ability to monitor automatic operation of the ECCS, including:

Safety Injection Pumps.

(4.2/4.3)

K/A MATCH ANALYSIS Requires the ability to monitor automatic start of ECCS pumps based on ES channels that have actuated ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: With RB pressure > 3 psig ES 1-6 should have actuated. Since LPI pumps are on ES 3&4, both A & B LPIPs would be operating and since both the A & B HPIPs are on are on ES-I, they would be operating.

B. Incorrect: Plausible since the A HPI pump is on ES channel I but the C HPI pump is on channel 2. Additionally, all ES LPI pumps would be operating however 1C LPI pump is not an ES pump. Additional plausibility from the fact that other ES systems that have a C pump have that pump as one of the ES pumps (ex. HPI & LPSW)

C. Incorrect: Plausible since the A and B HPIPs are actuated off of the ES channel 1 and this would be correct if ES pumps were also actuated off of ES I &2 and the B LPIP was on the even numbered channel (like their actual arrangement on ES 5&6).

D. Incorrect. Plausible if you assume both HPI and LPI are on ES I &2 and the A pumps are on the odd channels and B pumps on the even channels.

Technical Reference(s): IC-ES Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-ES R14, 18 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Comprehension and Analysis 2010A NRC REACTOR OPERATOR EXAM Question 33 T2/G1 - cpw 006A3.05 Emergency Core Cooling Ability to monitor automatic operation of the ECCS, including:

Safety Injection Pumps.

(4.2/4.3)

KIA MATCH ANALYSIS Requires the ability to monitor automatic start of ECCS pumps based on ES channels that have actuated ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: With RB pressure> 3 psig ES 1-6 should have actuated. Since LPI pumps are on ES 3&4, both A & B LPIP's would be operating and since both the A & B HPIP's are on are on ES-1, they would be operating.

B. Incorrect: Plausible since the A HPI pump is on ES channel 1 but the C HPI pump is on channel 2. Additionally, all ES LPI pumps would be operating however 1 C LPI pump is not an ES pump. Additional plausibility from the fact that other ES systems that have a "C" pump have that pump as one of the ES pumps (ex. HPI & LPSW)

C. Incorrect: Plausible since the A and B HPIP's are actuated off of the ES channel 1 and this would be correct if ES pumps were also actuated off of ES 1 &2 and the B LPIP was on the even numbered channel (like their actual arrangement on ES 5&6).

D. Incorrect. Plausible if you assume both HPI and LPI are on ES 1 &2 and the A pumps are on the odd channels and B pumps on the even channels.

Technical Reference(s): IC-ES Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-ES R14, 18 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 34 Unit 1 plant conditions:

0P111A111031002, (Filling and Venting RCS) Enclosure 4.14 (Establishing Pzr Steam Bubble And RCS Final Vent) in progress RCS pressure = 40 psig with a pure steam bubble IGWD-13 (QUENCH TANK VENT OUTSIDE RB) is closed IGWD-17 (PRESSURIZER VENT) is open Based on the above conditions, which ONE of the following describes the response of QT level and pressure?

There would be a(n) in QT pressure AND a(n) in QT level.

A. increase / increase B. increase

/ minimal change C. minimal change / increase D. minimal change / minimal change 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 34 Unit 1 plant conditions:

OP/1/A/1103/002, (Filling and Venting RCS) Enclosure 4.14 (Establishing Pzr Steam Bubble And RCS Final Vent) in progress RCS pressure = 40 psig with a pure steam bubble 1GWD-13 (QUENCH TANK VENT OUTSIDE RB) is closed 1 GWD-17 (PRESSURIZER VENT) is open Based on the above conditions, which ONE of the following describes the response of QT level and pressure?

There would be a(n) ~:--_ in QT pressure AND a(n) __ in QT level.

A. increase 1 increase B. increase 1 minimal change C. minimal change 1 increase D. minimal change 1 minimal change

2010A NRC REACTOR OPERATOR EXAM Question 34 T2/G1 okm/cpw 007K5.02 Pressurizer Relief Tank/Quench Tank System (PRTS)

Knowledge of the operational implications of the following concepts as they apply to PRTS:

Method of forming a steam bubble in the PZR (3.1/3.4)

K/A MATCH ANALYSIS Requires knowledge of the QT operational parameters (pressure and level changes) that indicate Pzr steam bubble formation is complete ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible if you have the misconception that the Pzr vent line is above the water level in the QT and therefore both QT pressure and level increase as some but not all of the steam would condense as it was vented to the QT.

B. Incorrect: Plausible if you have the misconception that the Pzr vent line is above the water level in the QT and therefore both QT pressure and level increase as some but not all of the steam would condense as it was vented to the QT.

C. CORRECT: Per OP/1103/002, Pzr steam bubble formation is complete (ie, all the N2 gas is vented out of the Pzr) when a change (rise) in QT pressure of less than 0.2 psig occurs and QT level increases by 2 inches. Since the Pzr vent is underwater in the QT, when N2 is being vented it will rise to the surface and cause a corresponding increase in QT pressure therefore minimal pressure response is a sign that all of the N2 has been vented. Additionally, as water is vented it is condensed under the water level of the QT therefore minimal QT pressure change in conjunction with increasing QT level is indicative of all N2 being out of Pzr.

D. Incorrect: Plausible if you do not understand conceptually how the N2 bubble is formed or if you do not understand that the pressurizer is being vented to the QT.

Additionally plausible if you have the misconception that the QT was vented to the vent header.

Technical Reference(s): OPIIIAIIIO3IOO2, End. 4.11 pg 1; End. 4.14, pg 4 & 8 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R17 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge Question 34 T2/G1 - okm/cpw 2010A NRC REACTOR OPERATOR EXAM 007K5.02 Pressurizer Relief Tank/Quench Tank System (PRTS)

Knowledge of the operational implications of the following concepts as they apply to PRTS:

Method of forming a steam bubble in the PZR (3.1/3.4 )

KIA MATCH ANALYSIS Requires knowledge of the QT operational parameters (pressure and level changes) that indicate Pzr steam bubble formation is complete ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible if you have the misconception that the Pzr vent line is above the water level in the QT and therefore both QT pressure and level increase as some but not all of the steam would condense as it was vented to the QT.

B. Incorrect: Plausible if you have the misconception that the Pzr vent line is above the water level in the QT and therefore both QT pressure and level increase as some but not all of the steam would condense as it was vented to the QT.

C. CORRECT: Per OP/11 03/002, Pzr steam bubble formation is complete (ie, all the N2 gas is vented out of the pzr) when a change (rise) in QT pressure of less than 0.2 psig occurs and QT level increases by 2 inches. Since the Pzr vent is underwater in the QT, when N2 is being vented it will rise to the surface and cause a corresponding increase in QT pressure therefore minimal pressure response is a sign that all of the N2 has been vented. Additionally, as water is vented it is condensed under the water level of the QT therefore minimal QT pressure change in conjunction with increasing QT level is indicative of all N2 being out of Pzr.

D. Incorrect: Plausible if you do not understand conceptually how the N2 bubble is formed or if you do not understand that the pressurizer is being vented to the QT.

Additionally plausible if you have the misconception that the QT was vented to the vent header.

Technical Reference(s): OP/1/A11103/002, Enc!. 4.11 pg 1; Enc!. 4.14, pg 4 & 8 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R17 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge

2010A NRC REACTOR OPERATOR EXAM I POINT Question 35 Which ONE of the following describes the normal power supply to the 1A CC pump AND the emergency backup source of power that will be supplying the Main Feeder Buses following a Loss of Offsite Power due to a Switchyard Isolation?

A. 1XL I KHU via overhead path B.

1XL I KHU via underground path C. 1XS1 I KHU via overhead path D. 1XS1 I KHU via underground path 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 35 Which ONE of the following describes the normal power supply to the 1A CC pump AND the emergency backup source of power that will be supplying the Main Feeder Buses following a Loss of Offsite Power due to a Switchyard Isolation?

A. 1XL / KHU via overhead path B. 1 XL / KHU via underground path C. 1XS1 / KHU via overhead path D. 1XS1 / KHU via underground path

2010A NRC REACTOR OPERATOR EXAM Question 35 T2!GI

- cpw roo8K2.o2 Component Cooling Water Knowledge of bus power supplies to the following:

CCW pump, including emergency backup.

(3.0/3.2)

K/A MATCH ANALYSIS Requires knowledge of the 1A CC pump normal and emergency backup power supplies.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: IXL is the normal power supply for the IA CC pump. With no ES actuation and a switchyard isolation the KHU aligned to the overhead will energize the MFB via CT-I.

B. Incorrect: First part is correct. Second part is plausible since it would be correct if and ES actuation had occurred with the loss of offsite power.

C. Incorrect: First part is plausible since 1XS1 is a load center that does supply major components including component cooling valve 1CC-7 however it does not supply power to the CC pumps. Second part is correct.

D. Incorrect: First part is plausible since IXS1 is a load center that does supply major components including component cooling valve ICC-7 however it does not supply power to the CC pumps. Second part is plausible since it would be correct if and ES actuation had occurred with the loss of offsite power.

Technical Reference(s): PNS-CC EL-EPD Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-CC R17, EL-EPD R27,28 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension or Analysis 2010A NRC REACTOR OPERATOR EXAM Question 35 T2/G1 - cpw 008K2.02 Component Cooling Water Knowledge of bus power supplies to the following:

CCW pump, including emergency backup.

(3.0/3.2)

KIA MATCH ANALYSIS Requires knowledge of the 1 A CC pump normal and emergency backup power supplies.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: 1XL is the normal power supply for the 1A CC pump. With no ES actuation and a switch yard isolation the KHU aligned to the overhead will energize the MFB via CT -1.

B. Incorrect: First part is correct. Second part is plausible since it would be correct if and ES actuation had occurred with the loss of offsite power.

C. Incorrect: First part is plausible since 1XS1 is a load center that does supply major components including component cooling valve 1 CC-7 however it does not supply power to the CC pumps. Second part is correct.

D. Incorrect: First part is plausible since 1XS1 is a load center that does supply major components including component cooling valve 1CC-7 however it does not supply power to the CC pumps. Second part is plausible since it would be correct if and ES actuation had occurred with the loss of offsite power.

Technical Reference(s): PNS-CC EL-EPD Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-CC R17, EL-EPD R27,28 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 36 Unit 3 plant conditions:

Reactor power = 100%

Pressurizer pressure control malfunction has occurred RCS pressure = 2000 psig decreasing Based on the above conditions, which ONE of the following describes the RCS pressure at which a LOW RCS PRESSURE reactor trip will occur AND the RCS pressure setroint where Engineered Safeguards digital channels 1 and 2 will actuate?

The reactor will trip at psig and ES digital channels 1 and 2 will actuate at psig.

A. 1810 / 1600 B. 1810 / 900 C. 1720 / 1600 D. 1720 / 900 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 36 Unit 3 plant conditions:

Reactor power = 100%

Pressurizer pressure control malfunction has occurred RCS pressure = 2000 psig decreasing Based on the above conditions, which ONE of the following describes the RCS pressure at which a LOW RCS PRESSURE reactor trip will occur AND the RCS pressure setpoint where Engineered Safeguards digital channels 1 and 2 will actuate?

The reactor will trip at ____ psig and ES digital channels 1 and 2 will actuate at

____ psig.

A. 181 0 / 1600 B. 1810 / 900 C. 1720 / 1600 D. 1720 / 900

2010A NRC REACTOR OPERATOR EXAM Question 36 T2IGI

- cpw 010K3.03 Pressurizer Pressure Control Knowledge of the effect that a loss or malfunction of the PZR PCS will have on the following:

ESFAS (4.0/4.2)

K/A MATCH ANALYSIS Requires knowledge of if the effect that a malfunction of the PZR PCS wilt have on ESFAS actuation ANSWER CHOICE ANALYSIS Answer: A Plausibility of all distracters enhanced by keeping the RPS setpoint options at a higher pressure than all of the ES setpoint options A. CORRECT: The setpoint for the RPS low pressure trip is 1810 psig and ES 1&2 actuate at 1600 psig.

B. Incorrect: First part is correct. Second part is plausible since 900 psig is the setpoint for the LPI Inhibit bistable which allows bypassing LPI ES when satisfied.

C. Incorrect: First part is plausible since 1720 psig is the RPS high pressure trip when the RPS channel is placed in shutdown bypass. Second part is correct.

D. Incorrect: First part is plausible since 1720 psig is the RPS high pressure trip when the RPS channel is placed in shutdown bypass. Second part is plausible since 900 psig is the setpoint for the LPI Inhibit bistable which allows bypassing LPI ES when satisfied.

Technical Reference(s): IC-ES, lC-RPS Proposed references to be provided to applicants during examination:

NONE Learning Objective: IC-ES R14, lC-RPS R3 Question Source:

NEW Question History: Last NRC Exam nla Question Cognitive Level:

Knowledge and Fundamentals Question 36 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 010K3.03 Pressurizer Pressure Control Knowledge of the effect that a loss or malfunction of the PZR PCS will have on the following:

ESFAS (4.0/4.2)

KIA MATCH ANALYSIS Requires knowledge of if the effect that a malfunction of the PZR PCS will have on ESFAS actuation ANSWER CHOICE ANALYSIS Answer: A Plausibility of all distracters enhanced by keeping the RPS setpoint options at a higher pressure than all of the ES setpoint options A. CORRECT: The setpoint for the RPS low pressure trip is 1810 psig and ES 1 &2 actuate at 1600 psig.

B. Incorrect: First part is correct. Second part is plausible since 900 psig is the setpoint for the LPI Inhibit bistable which allows bypassing LPI ES when satisfied.

C. Incorrect: First part is plausible since 1720 psig is the RPS high pressure trip when the RPS channel is placed in shutdown bypass. Second part is correct.

D. Incorrect: First part is plausible since 1720 psig is the RPS high pressure trip when the RPS channel is placed in shutdown bypass. Second part is plausible since 900 psig is the setpoint for the LPI Inhibit bistable which allows bypassing LPI ES when satisfied.

Technical Reference(s): IC-ES,IC-RPS Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-ES R14, IC-RPS R3 Question Source:

NEW Question History: Last NRC Exam _.....;n...;;,;.l=a __ _

Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 37 Unit 2 plant conditions:

Reactor power = 100%

2B RPS Channel Low RCS Pressure Bistable failed in tripped state 2B RPS Channel in Manual Bypass Current conditions:

2C RPS Channel inadvertently placed in Shutdown Bypass Based on the above conditions, which ONE of the following describes the impact (if any) on reactor power and control room alarms?

With NO additional operator actions, reactor power will be and the associated RPS Channel C statalarm for bistable trip will be actuated.

A. 0% I Low pressure B. 0% I High pressure C. 100% / Low pressure D. 100% / High pressure 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 37 Unit 2 plant conditions:

Reactor power = 100%

2B RPS Channel Low RCS Pressure Bistable failed in "tripped" state 2B RPS Channel in "Manual Bypass" Current conditions:

2C RPS Channel inadvertently placed in "Shutdown Bypass" Based on the above conditions, which ONE of the following describes the impact (if any) on reactor power and control room alarms?

With NO additional operator actions, reactor power will be and the associated.

RPS Channel C statalarm for bistable trip will be actuated.

A. 0% / Low pressure B. 0% / High pressure C. 100% / Low pressure D. 100% / High pressure

2010A NRC REACTOR OPERATOR EXAM Question 37 T2IGI okm/cpw 01 2A4.03 Reactor Protection System Ability to manually operate and/or monitor in the control room:

Channel blocks and bypasses.

(3.6/3.6)

KIA MATCH ANALYSIS Requires the ability to monitor plant response and control room indications that occur when placing RPS Channels in Manual Bypass and Shutdown Bypass ANSWER CHOICE ANALYSIS Answer:

D A. Incorrect: Both parts are incorrect. First part is plausible since there would be a bistable tripped in both the B and C channels however with the B channel in Manual Bypass the failed bistable does not result in RPS logic seeing that channel as actuated. Since it takes 2 channels to actuate, the reactor will still be at power.

Second part is plausible in that it would be essentially true if the question were asking which would be bypassed instead of actuated. It would be plausible to believe that the bistables in question were tripped instead of bypassed. If the bistable is bypassed then the statalarm is essentially bypassed.

B. Incorrect: First part is incorrect but plausible as described in A above. Second part is correct C. Incorrect: First part is correct. Second part is plausible in that it would be essentially true if the question were asking which would be bypassed instead of actuated. It would be plausible to believe that the bistables in question were tripped instead of bypassed.

D. CORRECT: With the B channel in Manual Bypass the failed bistable does not result in RPS logic seeing that channel as actuated therefore there is only one RPS channel tripped. Since it takes 2 tripped RPS channels to generate a Reactor trip the Rx still be at power. When an RPS channel is placed in shutdown bypass, RPS automatically inserts a high RCS pressure trip set point of 1720 psig therefore the high RCS pressure bistable will have actuated.

Technical Reference(s): IC-RPS pgs 8,18,19 Proposed references to be provided to applicants during examination:

NONE Learning Objective: IC-RPS R5, R6 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension or Analysis Question 37 T2/G1 - okm/cpw 2010A NRC REACTOR OPERATOR EXAM 012A4.03 Reactor Protection System Ability to manually operate and/or monitor in the control room:

Channel blocks and bypasses.

(3.6/3.6)

KIA MATCH ANALYSIS Requires the ability to monitor plant response and control room indications that occur when placing RPS Channels in Manual Bypass and Shutdown Bypass ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Both parts are incorrect. First part is plausible since there would be a bistable tripped in both the Band C channels however with the B channel in Manual Bypass the failed bistable does not result in RPS logic seeing that channel as actuated. Since it takes 2 channels to actuate, the reactor will still be at power.

Second part is plausible in that it would be essentially true if the question were asking which would be bypassed instead of actuated. It would be plausible to believe that the bistables in question were tripped instead of bypassed. If the bistable is bypassed then the statalarm is essentially bypassed.

B. Incorrect: First part is incorrect but plausible as described in A above. Second part is correct C. Incorrect: First part is correct. Second part is plausible in that it would be essentially true if the question were asking which would be bypassed instead of actuated. It would be plausible to believe that the bistables in question were tripped instead of bypassed.

D. CORRECT: With the B channel in Manual Bypass the failed bistable does not result in RPS logic seeing that channel as actuated therefore there is only one RPS channel tripped. Since it takes 2 tripped RPS channels to generate a Reactor trip the Rx still be at power. When an RPS channel is placed in shutdown bypass, RPS automatically inserts a high RCS pressure trip set point of ~1720 psig therefore the high RCS pressure bistable will have actuated.

Technical Reference(s): IC-RPS pgs 8,18,19 Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-RPS R5, R6 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 38 Which ONE of the following would result in a trip of the 1 D RPS Channel AND the 1 D CRD Breaker?

A. Reactor Building pressure bistables in the 1A and lB RPS channels fail in the tripped state B. Reactor Building pressure bistable in the 1 D RPS channel fails in the tripped state C. Loss of 1KVID D. LossoflDCB 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 38 Which ONE ofthe following would result in a trip of the 1 D RPS Channel AND the 1 D CRD Breaker?

A. Reactor Building pressure bistables in the 1 A and 1 B RPS channels fail in the "tripped" state B. Reactor Building pressure bistable in the 1 D RPS channel fails in the "tripped" state C. Loss of 1 KVID D. Loss of 1 DCB

2010A NRC REACTOR OPERATOR EXAM Question 38 T2IGI

- cpw 01 2K2.01 Reactor Protection System Knowledge of bus power supplies to the following:

RPS channels, components, and interconnections (3.3/3.7)

K/A MATCH ANALYSIS Requires knowledge of the power supply to RPS Channels and related components (CRD breakers)

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible since failure of the 2 RB pressure bistables will cause all 4 CRD breakers to open however only the A and B RPS channels would be tripped.

B. Incorrect: Plausible since this failure would result in the RPS channel tripping however the CRD breaker still requires 2 tripped RPS channels to open therefore it will remain closed C. CORRECT: Loss of the vital power source to a particular RPS channel will result in that entire channel de-energizing, with all indicating lights off, and the channel tripped. Loss of the vital power source will also result in a trip of the individual CRD breaker associated with that RPS channel since the I2OVAC to the breakers UV coil and shunt trip relay will be lost.

D. Incorrect: Plausible since DCB is the normal supply to DID. If DID is lost KVID would be de-energized resulting in D RPS channel and D CRD breaker trip. DCB is normal supply to DID however there is an auto backup from alternate unit via isolating diodes.

Technical Reference(s):

IC-RPS Proposed references to be provided to applicants during examination: NONE Learning Objective: lC-RPS R18, 20 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals Question 38 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 012K2.01 Reactor Protection System Knowledge of bus power supplies to the following:

RPS channels, components, and interconnections (3.3/3.7)

KIA MATCH ANALYSIS Requires knowledge of the power supply to RPS Channels and related components (CRD breakers)

ANSWER CHOICE ANALYSIS Answer: 8 A. Incorrect: Plausible since failure of the 2 RB pressure bistables will cause all 4 CRD breakers to open however only the A and B RPS channels would be tripped.

B. Incorrect: Plausible since this failure would result in the RPS channel tripping however the CRD breaker still requires 2 tripped RPS channels to open therefore it will remain closed C. CORRECT: Loss of the vital power source to a particular RPS channel will result in that entire channel de-energizing, with all indicating lights off, and the channel tripped. Loss of the vital power source will also result in a trip of the individual CRD breaker associated with that RPS channel since the 120VAC to the breakers UV coil and shunt trip relay will be lost.

D. Incorrect: Plausible since DCB is the normal supply to DID. If DID is lost KVID would be de-energized resulting in D RPS channel and D CRD breaker trip. DCB is normal supply to DID however there is an auto backup from alternate unit via isolating diodes.

Technical Reference(s): IC-RPS Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-RPS R18, 20 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 39 Unit 2 initial conditions:

Reactor power = 100%

Current conditions:

MSLB occurs RCS pressure = 1580 psig slowly increasing RB peak pressure = 2.8 psig Based on the above conditions, which ONE of the following describes valves have received a signal to CLOSE?

A. 2CC-7 B. 2LWD-1 C. 2LPSW-6 D. 2LPSW-1062 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 39 Unit 2 initial conditions:

Reactor power = 100%

Current conditions:

MSLB occurs RCS pressure = 1580 psig slowly increasing RB peak pressure = 2.8 psig Based on the above conditions, which ONE of the following describes valves have received a signal to CLOSE?

A. 2CC-7 B. 2LWD-1 C. 2LPSW-6 D. 2LPSW-1062

2010A NRC REACTOR OPERATOR EXAM Question 39 T2IGI cpw 01 3A3.02 Engineered Safety Features Actuation System (ESFAS)

Ability to monitor automatic operation of the ESFAS including:

Operation of actuated equipment.

(4.1/4.2)

K/A MATCH ANALYSIS Requires knowledge of ES actuation setpoints, what components are operated from which ES digital channels.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible since it would be correct if RB pressure had reached the ES 1-6 setpoint of 3 psig.

B. CORRECT: 2LWD-1 is on ES channel 1. With RCS pressure below the ES channel I actuation setpoint for RCS pressure (1600 psig) ES I will have actuated and sent a close signal to 2LWD-1 for non essential containment isolation.

C. Incorrect: Plausible since 2LPSW-6 does receive a closed signal from ES actuation however it is not from either Channel 1 or 2. This answer would be correct if ES channel 5 had actuated which would occur at 3 psig RB pressure.

D. Incorrect: Plausible since 2LPSW-1 062 does receive a closed signal from ES actuation however it is not from either Channel 1 or 2. This answer would be correct if ES channel 6 had actuated which would occur at 3 psig RB pressure..

Technical Reference(s):

IC-ES EOP End. 5.1 Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-ES R14 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Memory or Fundamental Knowledge Question 39 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 013A3.02 Engineered Safety Features Actuation System (ESFAS)

Ability to monitor automatic operation of the ESFAS including:

Operation of actuated equipment.

(4.1/4.2)

KIA MATCH ANALYSIS Requires knowledge of ES actuation setpoints, what components are operated from which ES digital channels.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible since it would be correct if RB pressure had reached the ES 1-6 setpoint of 3 psig.

B. CORRECT: 2LWD-1 is on ES channel 1. With RCS pressure below the ES channel 1 actuation setpoint for RCS pressure (1600 psig) ES 1 will have actuated and sent a close signal to 2LWD-1 for non essential containment isolation.

C. Incorrect: Plausible since 2LPSW-6 does receive a closed signal from ES actuation however it is not from either Channel 1 or 2. This answer would be correct if ES channel 5 had actuated which would occur at 3 psig RB pressure.

D. Incorrect: Plausible since 2LPSW-1062 does receive a closed signal from ES actuation however it is not from either Channel 1 or 2. This answer would be correct if ES channel 6 had actuated which would occur at 3 psig RB pressure..

Technical Reference(s):

IC-ES EOP Enc!. 5.1 Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-ES R14 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Memory or Fundamental Knowledge

2010A NRC REACTOR OPERATOR EXAM I POINT Question 40 Unit 3 plant conditions:

Reactor power = 100%

3KVIA AC Vital Power Panelboard supply breaker trips OPEN ES Analog Channel C WR RCS pressure signal fails LOW Based on the above conditions, which ONE of the following describes which (if any) ES digital channels have actuated?

have actuated.

A.

NO channels B.

Channels I thru 4 C.

ONLY channels 2 AND 4 D.

ONLY channels I AND 3 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 40 Unit 3 plant conditions:

Reactor power = 100%

3KVIA AC Vital Power Panel board supply breaker trips OPEN ES Analog Channel "c" WR RCS pressure signal fails LOW Based on the above conditions, which ONE of the following describes which (if any) ES digital channels have actuated?

have actuated.

A.

NO channels B.

Channels 1 thru 4 C.

ONLY channels 2 AND 4 D.

ONLY channels 1 AND 3

2010A NRC REACTOR OPERATOR EXAM Question 40 T2IGI okm/cpw 013K6.01 Engineered Safety Features Actuation System (ESFAS)

Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS:

Sensors and detectors.

(2.7/3.1)

KIA MATCH ANALYSIS Requires knowledge of the effect of both a loss of power to a channels sensorsldetectors as well as a malfunction of a sensorldetector will have on ESFAS actuation ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since there is a loss of power to an analog channel. The digital channels fail in the untripped state when they lose power but the analog channels fail tripped when they lose power. Since KVIA is a supply to both the A analog and the Odd digitals, it would be plausible to determine the A analog channel does not trip therefore no digital channels would actuate.

B. Incorrect: Plausible since there are 2 analog channels tripped on RCS pressure and therefore this would be correct if there were no loss of power to the Odd digital channels.

C. CORRECT: The digital channels fail in the untripped state when they lose power but the analog channels fail tripped when they lose power. Since KVIA is a supply to both the A analog and the Odd digitals, there would be 2 Analog channels tripped on the RCS pressure parameter therefore a trip signal is sent to Digital channels 1-4. With the Odd Digital channels without power, only channels 2 and 4 would actuate.

D. Incorrect: Plausible since it would be correct if KVIA supplied the Even digitial channels instead of the Odd channels.

Technical Reference(s):

IC-ES lesson pg II, TS 3.3.5 Proposed references to be provided to applicants during examination:

NONE Learning Objective: IC-ES R2, R5, TI, ADM-ITS R7 Question Source:

NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Comprehension or Analysis 2010A NRC REACTOR OPERATOR EXAM Question 40 T2/G1 - okm/cpw 013K6.01 Engineered Safety Features Actuation System (ESFAS)

Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS:

Sensors and detectors.

(2.7/3.1 )

K/A MATCH ANALYSIS Requires knowledge of the effect of both a loss of power to a channels sensors/detectors as well as a malfunction of a sensor/detector will have on ESFAS actuation ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since there is a loss of power to an analog channel. The digital channels fail in the untripped state when they lose power but the analog channels fail tripped when they lose power. Since KVIA is a supply to both the A analog and the Odd digitals, it would be plausible to determine the A analog channel does not trip therefore no digital channels would actuate.

B. Incorrect: Plausible since there are 2 analog channels tripped on RCS pressure and therefore this would be correct if there were no loss of power to the Odd digital channels.

C. CORRECT: The digital channels fail in the untripped state when they lose power but the analog channels fail tripped when they lose power. Since KVIA is a supply to both the A analog and the Odd digitals, there would be 2 Analog channels tripped on the RCS pressure parameter therefore a trip signal is sent to Digital channels 1-4. With the Odd Digital channels without power, only channels 2 and 4 would actuate.

D. Incorrect: Plausible since it would be correct if KVIA supplied the Even digitial channels instead of the Odd channels.

Technical Reference(s): IC-ES lesson pg 11, TS 3.3.5 Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-ES R2, R5, T1, ADM-ITS R7 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 41 Unit iplant conditions:

Time = 03:00 Reactor power = 100%

lB and IC RBCUs operating in HIGH speed 1A RBCU is operable and OFF ES channels 1-6 actuate Based on the above conditions, which ONE of the following describes RBCU status one minute later?

A. lB and IC RBCU5 operating in HIGH speed and IA RBCU OFF B.

I B and C RBCUs operating in LOW speed and IA RBCU OFF C. ALL RBCUs operating in LOW speed D. ALL RBCUs will be OFF 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 41 Unit 1 plant conditions:

Time = 03:00 Reactor power = 100%

1 Band 1 C RBCUs operating in HIGH speed 1 A RBCU is operable and OFF ES channels 1-6 actuate Based on the above conditions, which ONE of the following describes RBCU status one minute later?

A. 1B and 1C RBCUs operating in HIGH speed and 1A RBCU OFF B. 1 Band C RBCUs operating in LOW speed and 1A RBCU OFF C. ALL RBCUs operating in LOW speed D. ALL RBCUs will be OFF

2010A NRC REACTOR OPERATOR EXAM Question 41 T2IGI -okm 022A3.O1, Containment Cooling System (CCS)

Ability to monitor automatic operation of the CCS, including:

Initiation of safeguards mode of operation (4.1/4.3)

KIA MATCH ANALYSIS Requires the ability to monitor RBCU operation during initiation of safeguards (ES) mode of operation ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible if the 3-minute time delay is mis-applied. Since this is the pre-Es position of the RBCUs this would be the correct answer if you understand that the 3 minute time delay is when the RBCUs got their signal to re-position to ES position but did not understand that they were all initially stopped at the point of ES actuation.

B. Incorrect: Plausible if you were not aware of the 3 minute time delay and believed that the RBCU in OFF would not actuate on ES.

C. Incorrect: When ES actuates a 3-minute time delay is in effect and once the time delay is finished then all 3 RBCU5 will start at LOW speed. This choice is plausible if you are not aware of the 3 minute time delay or believe it is less than 1 minute.

D. CORRECT: When ES actuates all operating RBCUs will stop and a 3-minute time delay is in effect. Once the time delay is finished then all 3 RBCU5 will start at LOW speed. Since the 3 minute time delay has not yet timed out all RBCUs would be off.

Technical Reference(s):

PNS-RBC pg 5,6,16,17 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-RBC R1,R5 Question Source: Modified Bank PNSI 50501-enclosed Question History: Last NRC Exam NIA Question Cognitive Level:

Comprehension or Analysis Question 41 T2/G1 - okm 2010A NRC REACTOR OPERATOR EXAM 022A3.01, Containment Cooling System (CCS)

Ability to monitor automatic operation of the CCS, including:

Initiation of safeguards mode of operation (4.1/4.3)

KIA MATCH ANALYSIS Requires the ability to monitor RBCU operation during initiation of safeguards (ES) mode of operation ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible if the 3-minute time delay is mis-applied. Since this is the pre-Es position of the RBCU's this would be the correct answer if you understand that the 3 minute time delay is when the RBCU's got their signal to re-position to ES position but did not understand that they were all initially stopped at the point of ES actuation.

B. Incorrect: Plausible if you were not aware of the 3 minute time delay and believed that the RBCU in OFF would not actuate on ES.

C. Incorrect: When ES actuates a 3-minute time delay is in effect and once the time delay is finished then all 3 RBCUs will start at LOW speed. This choice is plausible if you are not aware of the 3 minute time delay or believe it is less than 1 minute.

D. CORRECT: When ES actuates all operating RBCU's will stop and a 3-minute time delay is in effect. Once the time delay is finished then all 3 RBCUs will start at LOW speed. Since the 3 minute time delay has not yet timed out all RBCUs would be off.

Technical Reference(s): PNS-RBC pg 5,6,16,17 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-RBC R1,R5 Question Source: Modified Bank - PNS150501-enclosed Question History: Last NRC Exam NIA Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 42 Which ONE of the following describes the range of BWST levels where RBS pump suction would be aligned to both the RBES and the BWST simultaneously AND what action(s) would be required if I LP-22 failed to close when isolating the BWST?

When performing Enclosure 5. 12 (ECCS Suction Swap to RBES) both suction sources are aligned when BWST level is between (feet) AND A. 159 I stop the lB LPI pump AND lB RBS pump B. 159 I Maximize total LPI flow < 3100 gpm C. 9-6/stopthelBLPlpumpANDlBRBSpump D. 9 - 6 I Maximize LPI flow < 3100 gpm 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 42 Which ONE of the following describes the range of BWST levels where RBS pump suction would be aligned to both the RBES and the BWST simultaneously AND what action(s) would be required if 1 LP-22 failed to close when isolating the BWST?

When performing Enclosure 5. 12 (ECCS Suction Swap to RBES) both suction sources are aligned when BWST level is between (feet) AND ____ _

A. 15 - 9 / stop the 1 B LPI pump AND 1 B RBS pump B. 15 - 9 / Maximize total LPI flow < 3100 gpm C. 9 - 6 / stop the 1 B LPI pump AND 1 B RBS pump D. 9 - 6 / Maximize total LPI flow < 3100 gpm

2010A NRC REACTOR OPERATOR EXAM Question 42 T2IGI

- cpw 026A1.03, Containment Spray System (CSS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including:

Containment sump level (3.5/3.5)

K/A MATCH ANALYSIS Requires ability to monitor changes in BWST level to ensure compliance with design limits on amount of BWST water moved to RBES to provide adequate volume of water in RBES. At ONS actions are taken based on BWST level instead of Containment sump level however analysis assume certain sump levels based on what BWST level is therefore monitoring BWST level and operating controls of RBS based on that level is synonymous with using Containment sump level.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since there are actions being taken to swap the suction from BWST to RBES however the actions allowed in this range do not align suction to both sources. Second part is correct.

B. Incorrect: First part is plausible since there are actions being taken to swap the suction from BWST to RBES however the actions allowed in this range do not aUgn suction to both sources. Second part is plausible since these actions are correct if only I LPI pump is operating when isolating BWST.

C. CORRECT: At 9 in BWST, LP-19 & 20 are both opened and suction for RBS &

LPI pumps is aligned to both BWST and RBES simultaneously. When BWST level reaches 6 the BWST is isolated by closing LP-21 and LP-22. If I LP-22 fails to close, the lB LPI pump AND lB RBS pump are secured until ILP-28 is manually closed.

D. Incorrect: First part is correct. Second part is plausible since stopping the RBS pump would slow the rate of decrease of the BWST and is fact the correct answer if 1 LP 20 fails to open at 9. Second part is plausible since these actions are correct if only I LPI pump is operating when isolating BWST.

Technical Reference(s): EOP Enclosure 5.12 (ECCS Suction Swap to RBES) and EAP-LOSCM Attachment 3 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-LOSCM R34, 36 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 42 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 026A 1.03, Containment Spray System (CSS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including:

Containment sump level (3.5/3.5)

K/A MATCH ANALYSIS Requires ability to monitor changes in BWST level to ensure compliance with design limits on amount of BWST water moved to RBES to provide adequate volume of water in RBES. At ONS actions are taken based on BWST level instead of Containment sump level however analysis assume certain sump levels based on what BWST level is therefore monitoring BWST level and operating controls of RBS based on that level is synonymous with using Containment sump level.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since there are actions being taken to swap the suction from BWST to RBES however the actions allowed in this range do not align suction to both sources. Second part is correct.

B. Incorrect: First part is plausible since there are actions being taken to swap the suction from BWST to RBES however the actions allowed in this range do not align suction to both sources. Second part is plausible since these actions are correct if only 1 LPI pump is operating when isolating BWST.

C. CORRECT: At 9' in BWST, LP-19 & 20 are both opened and suction for RBS &

LPI pumps is aligned to both BWST and RBES simultaneously. When BWST level reaches 6' the BWST is isolated by closing LP-21 and LP-22. If 1 LP-22 fails to close, the 1 B LPI pump AND 1 B RBS pump are secured until 1 LP-28 is manually closed.

D. Incorrect: First part is correct. Second part is plausible since stopping the RBS pump would slow the rate of decrease of the BWST and is fact the correct answer if 1 LP-20 fails to open at 9'. Second part is plausible since these actions are correct if only 1 LPI pump is operating when isolating BWST.

Technical Reference(s): EOP Enclosure 5.12 (ECCS Suction Swap to RBES) and EAP-LOSCM Attachment 3 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-LOSCM R34, 36 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 43 Unit 3 initial conditions:

Reactor power = 35% slowly increasing Current conditions:

Reactor power = 30% decreasing PCB 58 and PCB-59 (Unit 3 Generator Output Bkrs) OPEN Turbine master in HAND OAC point O3X2060 (ICS TURBINE LOADING STATUS) = FALSE Based on the above conditions, which ONE of the following describes the operation of the Turbine Bypass Valves (TBVs)?

is being compared to Turbine Header Pressure setpoint to develop the controlling error signal AND TBVs are controlling at psig?

A. Turbine Header Pressure / 885 B. Turbine Header Pressure / 935 C. Steam Generator Outlet Pressure / 885 D. Steam Generator Outlet Pressure / 935 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 43 Unit 3 initial conditions:

Reactor power = 35% slowly increasing Current conditions:

Reactor power = 30% decreasing PCB 58 and PCB-59 (Unit 3 Generator Output Bkrs) OPEN Turbine master in HAND OAC point 03X2060 (ICS TURBINE LOADING STATUS) = FALSE Based on the above conditions, which ONE of the following describes the operation of the Turbine Bypass Valves (TBV's)?

____ is being compared to Turbine Header Pressure setpoint to develop the controlling error signal AND TBV's are controlling at psig?

A. Turbine Header Pressure I 885 B. Turbine Header Pressure I 935 C. Steam Generator Outlet Pressure I 885 D. Steam Generator Outlet Pressure I 935

2010A NRC REACTOR OPERATOR EXAM Question 43 T2/G1

- cpw 039G2.1.19, Main and Reheat Steam Ability to use plant computers to evaluate system or component status.

(3.9/3.8)

K/A MATCH ANALYSIS Requires utilizing OAC indication for Turbine Load Status Flag (TLSF) to determine the setpoint at which Main and Reheat Steam pressure is being controlled ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since it would be correct if the Turbine Master were in Automatic. Second part is correct.

B. Incorrect: First part is plausible since it would be correct if the Turbine Master were in Automatic. Second part is plausible since it would be correct of the Turbine Load Status Flag were True however both Gen output breakers being open forces the status of the flag to False.

C. CORRECT: With the Turbine master in HAND, TBVs compare Steam Generator Outlet Pressure to THP setpoint to develop the controlling error.

With the TLSV being False and no trip confirmed signal from the Rx there is no bias applied to the TBV control therefore they would control at setpoint (which is 885).

D. Incorrect: First part is correct. Second part is plausible since it would be correct of the Turbine Load Status Flag were True however both Gen output breakers being open forces the status of the flag to False.

Technical Reference(s): STG-ICS Chapter 3 Proposed references to be provided to applicants during examination:

NONE Learning Objective: STG-ICS RIO Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals Question 43 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 039G2.1.19, Main and Reheat Steam Ability to use plant computers to evaluate system or component status.

(3.9/3.8)

KIA MATCH ANALYSIS Requires utilizing OAC indication for Turbine Load Status Flag (TLSF) to determine the setpoint at which Main and Reheat Steam pressure is being controlled ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since it would be correct if the Turbine Master were in Automatic. Second part is correct.

B. Incorrect: First part is plausible since it would be correct if the Turbine Master were in Automatic. Second part is plausible since it would be correct of the Turbine Load Status Flag were True however both Gen output breakers being open forces the status of the flag to False.

C. CORRECT: With the Turbine master in HAND, TBV's compare Steam Generator Outlet Pressure to THP setpoint to develop the controlling error.

With the TLSV being False and no trip confirmed signal from the Rx there is no bias applied to the TBV control therefore they would control at setpoint (which is 885).

D. Incorrect: First part is correct. Second part is plausible since it would be correct of the Turbine Load Status Flag were True however both Gen output breakers being open forces the status of the flag to False.

Technical Reference(s): STG-ICS Chapter 3 Proposed references to be provided to applicants during examination: NONE Learning Objective: STG-ICS R10 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 44 Unit 1 initial conditions:

Reactor power = 70% stable Current conditions:

1HPE-6 (Heater 1AI Bleed Inlet) closed Based on the above conditions, which ONE of the following predicts the impact of the malfunction on Feedwater flow assuming no operator action AND the procedure which will be used to reopen 1HPE-6?

Feedwater flow will stabilize at a value than the pre-transient level AND will be used to reopen IHPE-6.

A. higher / OP/11A11106/23 (High and Low Pressure Extraction)

B. higher / OPI1/A111061002 (Condensate and FDW system)

C. lower / OPI1IAI1IO6/23 (High and Low Pressure Extraction)

D. lower / OP/I/A/I 106/002 (Condensate and FDW system) 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 44 Unit 1 initial conditions:

Reactor power = 70% stable Current conditions:

1 HPE-6 (Heater 1A 1 Bleed Inlet) closed Based on the above conditions, which ONE of the following predicts the impact of the malfunction on Feedwater flow assuming no operator action AND the procedure which will be used to reopen 1 HPE-6?

Feedwater flow will stabilize at a _____ value than the pre-transient level AND

___ will be used to reopen 1 HPE-6.

A. higher 1 OP/1/A/1106/23 (High and Low Pressure Extraction)

B. higher 1 OP/1/A/1106/002 (Condensate and FDW system)

C. lower 1 OP/1/A/1106/23 (High and Low Pressure Extraction)

D. lower 1 OP/1/A/1106/002 (Condensate and FDW system)

2010A NRC REACTOR OPERATOR EXAM Question 44 T2IGI -cpw 059A2.06, Main Feedwater (MFW) System Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of steam flow to MEW system.

(2.7/2.9)

K/A MATCH ANALYSIS Requires ability to predict the impact on FDW system when steam flow is shut off to a high pressure feedwater heater and then requires knowledge of procedures use to mitigate the consequences of the operation ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible due to the initial lOS response. Initially, CTP will begin to increase which would generally indicate an increase in EDW flow is required however ICS will attempt to maintain CTP at setpoint and will therefore decrease the demand to the remaining portions of lOS which will then actually decrease EDW flow. Since the big picture of lOS is to maintain the primary and secondary heat balance it is plausible to deduce that ICS would increase feedwater to match the initial increase in CTP. Second part is correct.

B. Incorrect: First part is plausible as described in A. Second part is plausible since OP/1/A/1106/002 (Condensate and EDW system) is the procedure used to control most EDW heater operations. Additionally, it is the procedure to which you are directed if you are not able to reopen the extraction valve.

C. CORRECT: Initially, CTP will begin to increase however ICS will attempt to maintain CTP at setpoint and will therefore decrease the demand to the remaining portions of ICS. Additionally the FDW temperature correction ckt in the FDW subsection will modify FDW demand down since FDW temperature will be lower due to the loss of extraction steam. 0P111A11106123, Enclosure 4.1 (Re-opening Extraction Valves) contains guidance for re-opening extraction valves at power.

D. Incorrect: First part is correct. Second part is plausible since OP/1/A11106/002 (Condensate and EDW system) is the procedure used to control most FDW heater operations. Additionally, it is the procedure to which you are directed if you are not able to reopen the extraction valve.

Technical Reference(s): OPIIIA/1106123, Enclosure 4.1., Re-opening Extraction Valves)

AP128 (ICS Instrument Failure)

STG-ICS STG-FHS Proposed references to be provided to applicants during examination:

NONE Learning Objective:

STG-FHS R9, 23 STG-ICS R30, 14 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 44 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 059A2.06, Main Feedwater (MFW) System Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of steam flow to MFW system.

(2.7/2.9)

KIA MATCH ANALYSIS Requires ability to predict the impact on FDW system when steam flow is shut off to a high pressure feedwater heater and then requires knowledge of procedures use to mitigate the consequences of the operation ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible due to the initial ICS response. Initially, CTP will begin to increase which would generally indicate an increase in FDW flow is required however ICS will attempt to maintain CTP at setpoint and will therefore decrease the demand to the remaining portions of ICS which will then actually decrease FDW flow. Since the big picture of ICS is to maintain the primary and secondary heat balance it is plausible to deduce that ICS would increase feedwater to match the initial increase in CTP. Second part is correct.

B. Incorrect: First part is plausible as described in A. Second part is plausible since OP/1/A/11061002 (Condensate and FDW system) is the procedure used to control most FDW heater operations. Additionally, it is the procedure to which you are directed if you are not able to reopen the extraction valve.

C. CORRECT: Initially, CTP will begin to increase however ICS will attempt to maintain CTP at setpoint and will therefore decrease the demand to the remaining portions of ICS. Additionally the FDW temperature correction ckt in the FDW subsection will modify FDW demand down since FDW temperature will be lower due to the loss of extraction steam. OP/1/A11106/23, Enclosure 4.1 (Re-opening Extraction Valves) contains guidance for re-opening extraction valves at power.

D. Incorrect: First part is correct. Second part is plausible since OP/1/A/11 06/002 (Condensate and FDW system) is the procedure used to control most FDW heater operations. Additionally, it is the procedure to which you are directed if you are not able to reopen the extraction valve.

Technical Reference(s): OP/1/A11106/23, Enclosure 4.1., Re-opening Extraction Valves) AP/28 (ICS Instrument Failure) STG-ICS STG-FHS Proposed references to be provided to applicants during examination: NONE Learning Objective: STG-FHS R9, 23 STG-ICS R30, 14 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 45 Unit 1 initial conditions:

Reactor power = 100%

Current conditions:

Reactor trip IFDW-33 (IA SU FDW Block) FAILS closed Based on the above conditions, which ONE of the following describes the expected Steam Generator levels 20 minutes after the trip?

ASSUME NO OPERATOR ACTIONS 1A SG level =

AND lB SG level =

A 25 S/U / 25 S/U B

12 S/U / 25 S/U C

30 XSUR / 25 S/U D. 30XSUR / 30XSUR 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 45 Unit 1 initial conditions:

Reactor power = 100%

Current conditions:

Reactor trip 1 FDW-33 (1A SU FDW Block) FAILS closed Based on the above conditions, which ONE of the following describes the expected Steam Generator levels 20 minutes after the trip?

ASSUME NO OPERATOR ACTIONS 1A SG level = ___ AND 1B SG level = __ _

A. 25" S/U / 25" S/U B. 12" S/U / 25" S/U C. 30" XSUR / 25" S/U D. 30" XSUR / 30" XSUR

2010A NRC REACTOR OPERATOR EXAM Question 45 T2IGI

- bank 059K3.02, Main Feedwater (MEW) System Knowledge of the effect that a loss or malfunction of the MFW will have on the following:

AFW system.

(3.6/3.7)

K/A MATCH ANALYSIS Requires knowledge of the effect that a malfunction of a FDW block valve will have on EFDW system ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since 25 is the normal post trip SG level in both SCs when on Main FDW. Since the S/U control valve has not failed it is plausible to deduce the S/U valve could still control at 25. It is also plausible to believe that the main EDW control valve would control at 25 if the SU valve did not by believing the valve composite demand would be sent to the main control valve in lieu of the startup valve controlling. IF the A SC did not decrease and initiate dryout then the second part would be correct.

B. Incorrect: Plausible since this would be correct if it took 21 in BOTH SCs to actuate EFDW on dryout protection.

C. Incorrect: Plausible since the failure is on the IA SC only. Failure to realize that BOTH MDEFWPs will start if EITHER SC reaches 21 for 30 seconds could lead to this choice.

D. CORRECT: With the SU block valve failed closed the SU control valve cannot supply FDW to the IA SG. SG level will decrease until <21 for 30 seconds which will start BOTH MDEFWPs. With both MDEFWPs operating, FDW-315 and 316 will control at 30 XSUR Technical Reference(s): CF-EF, CF-FDW Proposed references to be provided to applicants during examination:

NONE Learning Objective:

CF-EF R20, R25, R37 Question Source:

BANK (CF023704)

Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 45 T2/G1 - bank 2010A NRC REACTOR OPERATOR EXAM 059K3.02, Main Feedwater (MFW) System Knowledge of the effect that a loss or malfunction of the MFW will have on the following:

AFW system.

(3.6/3.7)

KIA MATCH ANALYSIS Requires knowledge of the effect that a malfunction of a FDW block valve will have on EFDW system ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since 25" is the normal post trip SG level in both SG's when on Main FDW. Since the S/U control valve has not failed it is plausible to deduce the S/U valve could still control at 25". It is also plausible to believe that the main FDW control valve would control at 25" if the SU valve did not by believing the valve composite demand would be sent to the main control valve in lieu of the startup valve controlling. IF the A SG did not decrease and initiate dryout then the second part would be correct.

B. Incorrect: Plausible since this would be correct if it took 21" in BOTH SG's to actuate EFDW on dryout protection.

C. Incorrect: Plausible since the failure is on the 1A SG only. Failure to realize that BOTH MDEFWP's will start if EITHER SG reaches 21" for 30 seconds could lead to this choice.

D. CORRECT: With the SU block valve failed closed the SU control valve cannot supply FDW to the 1A SG. SG level will decrease until <21" for 30 seconds which will start BOTH MDEFWP's. With both MDEFWP's operating, FDW-315 and 316 will control at 30" XSUR Technical Reference(s): CF-EF, CF-FDW Proposed references to be provided to applicants during examination: NONE Learning Objective: CF-EF R20, R25, R37 Question Source: BANK (CF023704)

Question History: Last NRC Exam NIA Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 46 Unit 1 initial conditions:

Reactor power = 100%

Unit I TDEFWP unavailable Current conditions:

Both Main FDW pumps trip lBMDEFWPfailstostart Based on the above conditions, which ONE of the following describes actions directed by the EOP to remove core decay heat?

Initiate...

A. Rule 3 (Loss of Main or Emergency Feedwater) and cross connect with an alternate unit to supply the 1 B Steam Generator B. Rule 3 (Loss of Main or Emergency Feedwater) to decrease SG pressure and feed with Condensate Booster pumps C. Rule 4 (Initiation of HPI Forced Cooling) if RCS pressure reaches 2300 psig D. EOP End. 5.9 (Extended EFDW Operation) and feed both SGs with 1A MDEFWP 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 46 Unit 1 initial conditions:

Reactor power = 100%

Unit 1 TDEFWP unavailable Current conditions:

Both Main FDW pumps trip 1 B MDEFWP fails to start Based on the above conditions, which ONE of the following describes actions directed by the EOP to remove core decay heat?

Initiate...

A. Rule 3 (Loss of Main or Emergency Feedwater) and cross connect with an alternate unit to supply the 1 B Steam Generator B. Rule 3 (Loss of Main or Emergency Feedwater) to decrease SG pressure and feed with Condensate Booster pumps C. Rule 4 (Initiation of HPI Forced Cooling) if RCS pressure reaches 2300 psig D. EOP Encl. 5.9 (Extended EFDW Operation) and feed both SG's with 1A MDEFWP

2010A NRC REACTOR OPERATOR EXAM Question 46 T2IGI 061 K6.02, Auxiliary / Emergency Feedwater (AFW) System Knowledge of the effect of a loss or malfunction of the following will have on the AFW components:

Pumps.

(2.6/2.7)

K/A MATCH ANALYSIS Requires knowledge of how AFW components are utilized based on loss of MFDWPs, TDEFWP, and one MDEFWP ANSWER CHOICE ANALYSIS Answer:

D A. Incorrect: Plausible since cross connecting with alternate unit is a mitigation strategy utilized by Rule 3 however it is applied if no EFDWPs are available on the subject unit.

B. Incorrect: Plausible since CBP feed is a strategy utilized by Rule 3 and it would be correct if the IA MDEFWP had also been lost.

C. Incorrect: Plausible since HPI FC is utilized as a strategy in RULE 4 and would be correct if the 1A MDEFWP had also been lost since it is only applied if neither SG can be fed and RCS pressure reached 2300 psi.

0. CORRECT: If only one MDEFWP is available Rule 3 will send you to End. 5.9 which will direct opening FDW-313 & 314 and feeding both SGs with one MDEFWP.

Technical Reference(s): EOP Rule 3, EAP-LOHT Attachment 3 Proposed references to be provided to applicants during examination:

NONE Learning Objective:

EAP-LOHT R26 Question Source:

NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals Question 46 T2/G1 -

2010A NRC REACTOR OPERATOR EXAM 061 K6.02, Auxiliary / Emergency Feedwater (AFW) System Knowledge of the effect of a loss or malfunction of the following will have on the AFW components:

Pumps.

(2.6/2.7)

KIA MATCH ANALYSIS Requires knowledge of how AFW components are utilized based on loss of MFDWP's, TDEFWP, and one MDEFWP ANSWER CHOICE ANALYSIS Answer: 0 A. Incorrect: Plausible since cross connecting with alternate unit is a mitigation strategy utilized by Rule 3 however it is applied if no EFDWPs are available on the subject unit.

B. Incorrect: Plausible since CBP feed is a strategy utilized by Rule 3 and it would be correct if the 1 A MDEFWP had also been lost.

C. Incorrect: Plausible since HPI FC is utilized as a strategy in RULE 4 and would be correct if the 1 A MDEFWP had also been lost since it is only applied if neither SG can be fed and RCS pressure reached 2300 psi.

D. CORRECT: If only one MDEFWP is available Rule 3 will send you to Encl. 5.9 which will direct opening FDW-313 & 314 and feeding both SG's with one MDEFWP.

Technical Reference(s): EOP Rule 3, EAP-LOHT Attachment 3 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-LOHT R26 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 47 Which ONE of the following describes actions required in that will extend the life of the Control Batteries following a loss of all AC power in accordance with EOP Enclosure 5.38 (Restoration of Power)?

Load Shed the inverter.

A.Kl B. DIA C. KSF-1 D. KOAC 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 47 Which ONE of the following describes actions required in that will extend the life of the Control Batteries following a loss of all AC power in accordance with EOP Enclosure 5.38 (Restoration of Power)?

Load Shed the inverter.

A. KI B.DIA C. KSF-1 D. KOAC

2010A NRC REACTOR OPERATOR EXAM Question 47 T2IGI

- cpw 062K3.03, AC Electrical Distribution System Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following:

DC system.

(3.7/3.9)

KIA MATCH ANALYSIS Requires knowledge of the effect of a loss of AC power will have on DC systems and actions required subsequent to the loss of AC that will impact DC system availability ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: If a blackout exists on all 3 units, End. 5.38 directs performing End. 5.32 which will open the DC input breaker to the KI inverter.

B. Incorrect: Plausible since DIA is an inverter powered from the control batteries load shedding the inverter would extend control battery life however DIA, DIB, DIC, and DID inverters remain energized from the batteries during a blackout.

C. Incorrect: Plausible since the XSF-1 inverter is an inverter powered by DC however the inverter is at the SSF and remains energized during a blackout D. Incorrect: Plausible since the KOAC inverter is an inverter that is load shed when performing Enclosure 5.32 however it is powered from the Power Batteries and not the Control Batteries.

Technical Reference(s): EL-DCD Proposed references to be provided to applicants during examination: NONE Learning Objective: EL-DCD RI Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals Question 47 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 062K3.03, AC Electrical Distribution System Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following:

DC system.

(3.7/3.9)

KIA MATCH ANALYSIS Requires knowledge of the effect of a loss of AC power will have on DC systems and actions required subsequent to the loss of AC that will impact DC system availability ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: If a blackout exists on all 3 units, Encl. 5.38 directs performing Enc!. 5.32 which will open the DC input breaker to the KI inverter.

B. Incorrect: Plausible since DIA is an inverter powered from the control batteries load shedding the inverter would extend control battery life however DIA, DIB, DIC, and DID inverters remain energized from the batteries during a blackout.

C. Incorrect: Plausible since the XSF-1 inverter is an inverter powered by DC however the inverter is at the SSF and remains energized during a blackout D. Incorrect: Plausible since the KOAC inverter is an inverter that is load shed when performing Enclosure 5.32 however it is powered from the Power Batteries and not the Control Batteries.

Technical Reference(s): EL-DCD Proposed references to be provided to applicants during examination: NONE Learning Objective: EL-DCD R1 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 48 Which ONE of the following describes the normal alignment of the Power Battery busses AND a condition in which SLC 16.8.3 (Power Battery Parameters) would require changing that alignment?

The Oconee units are normally and this would be changed if A. cross-tied / a single power battery becomes inoperable B. cross-tied / two or more power batteries become inoperable C. separated / a single power battery becomes inoperable D. separated / two or more power batteries become inoperable 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 48 Which ONE of the following describes the normal alignment of the Power Battery busses AND a condition in which SLC 16.8.3 (Power Battery Parameters) would require changing that alignment?

The Oconee units are normally ____ and this would be changed if ___ _

A. cross-tied / a single power battery becomes inoperable B. cross-tied / two or more power batteries become inoperable C. separated / a single power battery becomes inoperable D. separated / two or more power batteries become inoperable

2010A NRC REACTOR OPERATOR EXAM Question 48 T2/G1

- cpw 063K4.02, D.C. Electrical Distribution Knowledge of DC electrical system design feature(s) and! or interlock(s) which provide for the following:

Breaker interlocks, permissives, bypasses and cross-ties.

(2.9/3.2)

KIA MATCH ANALYSIS Requires knowledge of the normal design status of DC power battery busses as well as the design feature which allows cross-tie between units.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is incorrect but plausible since the normal alignment of the Control Batteries is with the units cross-connected via the isolating diodes. Second part is correct.

B. Incorrect: First part is incorrect but plausible since the normal alignment of the Control Batteries is with the units cross-connected via the isolating diodes. Second part is incorrect but plausible since SLC 16.8.3 Condition C is for having 2 or more batteries inoperable and does have immediate corrective actions however the corrective actions in that case do not include cross-tieing the busses.

C. CORRECT: Normal alignment of the power battery buses is with the units separated (EL-DCD page 35). SLC 16.8.3 Condition B is for a single power battery inoperable and requires initiating actions to cross-tie busses immediately.

D. Incorrect: First part is correct. Second part is plausible since SLC 16.8.3 Condition C is for having 2 or more batteries inoperable and does have immediate corrective actions however the corrective actions in that case do not include cross-tieing the busses.

Technical Reference(s): EL-DCD SLC 16.8.3 Proposed references to be provided to applicants during examination:

NONE Learning Objective: EL DCD R7, ADM-TSS R4 Question Source:

NEW Question History: Last NRC Exam N!A Question Cognitive Level: Knowledge and Fundamentals Question 48 T2/G1 - cpw 2010A NRC REACTOR OPERATOR EXAM 063K4.02, D.C. Electrical Distribution Knowledge of DC electrical system design feature(s) andl or interlock(s) which provide for the following:

Breaker interlocks, permissives, bypasses and cross-ties.

(2.9/3.2)

KIA MATCH ANALYSIS Requires knowledge of the normal design status of DC power battery busses as well as the design feature which allows cross-tie between units.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is incorrect but plausible since the normal alignment of the Control Batteries is with the units cross-connected via the isolating diodes. Second part is correct.

B. Incorrect: First part is incorrect but plausible since the normal alignment of the Control Batteries is with the units cross-connected via the isolating diodes. Second part is incorrect but plausible since SLC 16.8.3 Condition C is for having 2 or more batteries inoperable and does have immediate corrective actions however the corrective actions in that case do not include cross-tieing the busses.

C. CORRECT: Normal alignment of the power battery buses is with the units separated (EL-DCD page 35). SLC 16.8.3 Condition 8 is for a single power battery inoperable and requires initiating actions to cross-tie busses immediately.

D. Incorrect: First part is correct. Second part is plausible since SLC 16.8.3 Condition C is for having 2 or more batteries inoperable and does have immediate corrective actions however the corrective actions in that case do not include cross-tieing the busses.

Technical Reference(s): EL-DCD SLC 16.8.3 Proposed references to be provided to applicants during examination: NONE Learning Objective: EL DCD R7, ADM-TSS R4 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 49 Operators are preparing to synchronize KHU-2 to the grid in accordance with OP/0/A/1 106/019, (Keowee Hydro At Oconee)

The operator notes the following indications:

Grid Frequency = 59.9 cycles Keowee Frequency = 60.3 cycles Keowee 2 Line Volts = 13.7 kV Keowee 2 Output Volts = 15.2 kV Based on the above conditions, which ONE of the following describes the control that will be used to adjust the synchroscope indication and the response when ACB-2 is closed?

The will be used to adjust the synchroscope indication and A. UNIT 2 AUTO VOLTAGE ADJUSTER / ACB-2 will immediately receive a trip signal as a direct result of the line voltage differential B. UNIT 2 SPEED CHANGER MOTOR / ACB-2 will NOT receive a trip signal as a direct result of the line voltage differential C. UNIT 2 AUTO VOLTAGE ADJUSTER I ACB-2 will NOT receive a trip signal as a direct result of the line voltage differential.

D. UNIT 2 SPEED CHANGER MOTOR / ACB-2 will immediately receive a trip signal as a direct result of the line voltage differential 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 49 Operators are preparing to synchronize KHU-2 to the grid in accordance with OP/O/A/1106/019, (Keowee Hydro At Oconee)

The operator notes the following indications:

Grid Frequency = 59.9 cycles Keowee Frequency = 60.3 cycles Keowee 2 Line Volts = 13.7 kV Keowee 2 Output Volts = 15.2 kV Based on the above conditions, which ONE of the following describes the control that will be used to adjust the synchroscope indication and the response when ACB-2 is closed?

The will be used to adjust the synchroscope indication and ___ _

A. UNIT 2 AUTO VOLTAGE ADJUSTER I ACB-2 will immediately receive a trip signal as a direct result of the line voltage differential B. UNIT 2 SPEED CHANGER MOTOR I ACB-2 will NOT receive a trip signal as a direct result of the line voltage differential C. UNIT 2 AUTO VOLTAGE ADJUSTER I ACB-2 will NOT receive a trip signal as a direct result of the line voltage differential.

D. UNIT 2 SPEED CHANGER MOTOR I ACB-2 will immediately receive a trip Signal as a direct result of the line voltage differential

2010A NRC REACTOR OPERATOR EXAM Question 49 T2IGI 064A1.03, Emergency Diesel Generators (ED/G)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including:

Operating voltages, currents, and temperatures (use Hydro units if possible).

(3.1/3.4)

K/A MATCH ANALYSIS Per NRC OK not to address temperatures. Requires monitoring parameters and predicting response when operating EDIG system controls. Additionally requires ability to manipulate controls of KHU to prevent exceeding design limits as unit is brought on-line.

ANSWER CHOICE ANALYSIS Answer: B Out of tolerance circuit protection is only active for Emergency Starts. Speed Changer Motor (SCM), Auto Voltage Adjuster (AVA)

A. Incorrect: First part is wrong but plausible if it is not known that the SCM not the AVA is used to adjust the synch scope.

Second part is wrong but plausible if the OOT protection circuit is misapplied (only for Emerg Starts) in which case the breaker still would not trip as the voltage high but still within tolerance.

B. CORRECT: Keowee frequency is higher than grid so synchroscope will be spinning clockwise (CW) which will require use of the MSC. Out of tolerance circuit protection will not trip ACB 2.

C. Incorrect: Plausible in that the second part is correct. First part is wrong but plausible if it is not known that the SCM not the AVA is used to adjust the synch scope.

D. Incorrect. Plausible in that the first part is correct. Second part is wrong but plausible if the QOT protection circuit is misapplied (only for Emerg Starts) in which case the breaker still would not trip as the voltage high but still within tolerance.

Plausibility is based on memorizing the ACB auto trip feature and correctly calculating less than 10% normal voltage. Also, plausibility is hinged on the applicant knowing how the synch scope will respond to the frequency differential.

Technical Reference(s): EL-KHG, 0P101A111061019 Rev 83 Proposed references to be provided to applicants during examination: None Learning Objective: EL-KHG Ru, R4, R20, R19, R7 Question Source: BANK ELO4IIIO Question History: Last NRC Exam 2009 ONS RO NRC Exam Q#49 (Slightly changed and re-ordered given conditions and rearranged answers)

Question Cognitive Level:

Comprehension or Analysis 2010A NRC REACTOR OPERATOR EXAM Question 49 T2/G1 -

064A 1.03, Emergency Diesel Generators (ED/G)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including:

Operating voltages, currents, and temperatures (use Hydro units if possible).

(3.1/3.4)

KIA MATCH ANALYSIS to temperatures. Requires monitoring parameters and predicting response when operating ED/G system controls. Additionally requires ability to manipulate controls of KHU to prevent exceeding design limits as unit is brought on-line.

ANSWER CHOICE ANALYSIS Answer: B Out of tolerance circuit protection is only active for Emergency Starts. Speed Changer Motor (SCM), Auto Voltage Adjuster (AVA)

A. Incorrect: First part is wrong but plausible if it is not known that the SCM not the AVA is used to adjust the synch scope. Second part is wrong but plausible if the OOT protection circuit is misapplied (only for Emerg Starts) in which case the breaker still would not trip as the voltage high but still within tolerance.

B. CORRECT: Keowee frequency is higher than grid so synchroscope will be spinning clockwise (CW) which will require use of the MSC. Out of tolerance circuit protection will not trip ACB 2.

C. Incorrect: Plausible in that the second part is correct. First part is wrong but plausible if it is not known that the SCM not the AVA is used to adjust the synch scope.

D. Incorrect. Plausible in that the first part is correct. Second part is wrong but plausible if the OOT protection circuit is misapplied (only for Emerg Starts) in which case the breaker still would not trip as the voltage high but still within tolerance.

Plausibility is based on memorizing the ACB auto trip feature and correctly calculating less than 10% normal voltage. Also, plausibility is hinged on the applicant knowing how the synch scope will respond to the frequency differential.

Technical Reference(s): EL-KHG, OP/0/Al1106/019 Rev 83 Proposed references to be provided to applicants during examination: None Learning Objective: EL-KHG R11, R4, R20, R19, R7 Question Source: BANK EL041110 Question History: Last NRC Exam 2009 ONS RO NRC Exam Q#49 (Slightly changed and re-ordered given conditions and rearranged answers)

Question Cognitive Level:

Comprehension or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 50 Unit 1 conditions:

Time= 1159:40 Reactor power = 100% stable KHU-1 OOS ACB-4 closed KHU-2 gets Emergency Start signal from another unit Time 1200:00 KHU-2 speed reaches 190 RPM Time = 1200:30 KHU-2 speed = 190 RPM Based on the above conditions, which ONE of the following describes the status of KHU-2 and the procedural actions required by Unit 1 (if any) as a result of that status?

A. Emergency locked out / Enter LCO 3.0.3 immediately B. Emergency locked out I Energize both Standby Buses within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Energizing CT-4 I No additional actions required D. Energizing CT-4 I Initiate APII I (Recovery From Loss Of Power) 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 50 Unit 1 conditions:

Time = 1159:40 Reactor power = 100% stable KHU-100S ACB-4 closed KHU-2 gets Emergency Start signal from another unit Time = 1200:00 KH U-2 speed reaches 190 RPM Time = 1200:30 KHU-2 speed = 190 RPM Based on the above conditions, which ONE of the following describes the status of KHU-2 and the procedural actions required by Unit 1 (if any) as a result of that status?

A. Emergency locked out / Enter LCO 3.0.3 immediately B. Emergency locked out / Energize both Standby Buses within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Energizing CT -4 / No additional actions required D. Energizing CT-4 / Initiate AP/11 (Recovery From Loss Of Power)

2010A NRC REACTOR OPERATOR EXAM Question 50 T2IGI -cpw 064A2.02, Emergency Diesel Generators (ED/C)

Ability to (a) predict the impacts of the following malfunctions or operations on the EDIG system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Load, VARS, pressure on air compressor, speed droop, frequency, voltage, fuel oil level, temperatures.

(2.7/2.9)

K/A MATCH ANALYSIS i NRC OK to ask about lake level as malfunction and actions based on moperabilify as iesponse (However could not make RO level question from that)

Predict the impact of overspeed on KHU and use TS required actions to mitigate the consequences of the inoperability.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since it is common for TS to require immediate 3.0.3 entry when both trains (overhead and underground) of a safety function (emergency power to Oconee) have been lost.

B. CORRECT: For KHUs, If an Emergency Start is present and the unit overspeeds

(>180 RPM), then a 23 second timer starts. If the unit has not decreased to < 180 RPM within this 23 seconds, then an Emergency Lockout is generated. Since KHU-1 is already OOS then both the overhead and underground KHUs are inoperable and TS 3.8.1 Condition J requires energizing both SBBs within I hour C. Incorrect: First part is plausible since this would be correct if you did not have an emergency lockout. Second part is plausible since it would be correct if KHU-2 had operated properly.

D. Incorrect: First part is plausible since this would be correct if you did not have an emergency lockout. Second part is plausible since AP/1 I (Recovery from Loss of Power) is entered when 4160V buses lose power and it is subsequently regained. It, would be correct if your unit were counting on the KHUs to energize your MFB since that would indicate that 4160V had lost power however in this case although the KHUs are inoperable, Unit 1 has not had a loss of power therefore 4160V buses were never deenergized.

Technical Reference(s): TS 3.8.1, EL-KHG Proposed references to be provided to applicants during examination:

NONE Learning Objective:

ADM-ITS R7, EL-KHG R21 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis 2010A NRC REACTOR OPERATOR EXAM Question 50 T2/G1 - cpw 064A2.02, Emergency Diesel Generators (ED/G)

Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Load, VARS, pressure on air compressor, speed droop, frequency, voltage, fuel oil level, temperatures.

(2.7/2.9)

KIA MATCH ANALYSIS lake as malfunction and actions on as could not make RO level question from that.)

Predict the impact of overspeed on KHU and use TS required actions to mitigate the consequences of the inoperability.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since it is common for TS to require immediate 3.0.3 entry when both trains (overhead and underground) of a safety function (emergency power to Oconee) have been lost.

B. CORRECT: For KHU's, If an Emergency Start is present and the unit overspeeds (2: 180 RPM), then a 23 second timer starts. If the unit has not decreased to < 180 RPM within this 23 seconds, then an Emergency Lockout is generated. Since KHU-1 is already 005 then both the overhead and underground KHU's are inoperable and TS 3.8.1 Condition J requires energizing both SBB's within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Incorrect: First part is plausible since this would be correct if you did not have an emergency lockout. Second part is plausible since it would be correct if KHU-2 had operated properly.

D. Incorrect: First part is plausible since this would be correct if you did not have an emergency lockout. Second part is plausible since AP/11 (Recovery from Loss of Power) is entered when 4160V buses lose power and it is subsequently regained. It would be correct if your unit were counting on the KHU's to energize your MFB since that would indicate that 4160V had lost power however in this case although the KHU's are inoperable, Unit 1 has not had a loss of power therefore 4160V buses were never deenergized.

Technical Reference(s): TS 3.8.1, EL-KHG Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-ITS R7, EL-KHG R21 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 51 Unit 1 initial conditions:

Time=1200 Reactor power = 35%

1A steam generator tube leak = 2.1 gpd stable RCS activity = 0.25 pCi/mI DEl increasing Current conditions:

Time=1400 NO change in IA SG tube leak rate RCS activity = 0.65 pCi/mI DEl and increasing Based on the above conditions, which ONE of the following describes the response of the radiation monitors between 1200 and 1400?

A. 1RIA-16 (Main Steam Line Monitor) and 1RIA-40 (CSAE Off-gas) increased.

B. 1RIA-16 (Main Steam Line Monitor) increased while 1RIA-40 (CSAE Off-gas) remained constant.

C. IRIA-59 (N-16 monitor) and IRIA-40 (CSAE Off-gas) increased.

D. 1RIA-59 (N-16 monitor) increased whilelRlA-40 (CSAE Off-gas) remained constant.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 51 Unit 1 initial conditions:

Time = 1200 Reactor power = 35%

1 A steam generator tube leak = 2.1 gpd stable RCS activity = 0.25 ~Cilml DEI increasing Current conditions:

Time = 1400 NO change in 1 A SG tube leak rate RCS activity = 0.65 ~Cilml DEI and increasing Based on the above conditions, which ONE of the following describes the response of the radiation monitors between 1200 and 1400?

A. 1 RIA-16 (Main Steam Line Monitor) and 1 RIA-40 (CSAE Off-gas) increased.

B. 1 RIA-16 (Main Steam Line Monitor) increased while 1 RIA-40 (CSAE Off-gas) remained constant.

C. 1 RIA-59 (N-16 monitor) and 1 RIA-40 (CSAE Off-gas) increased.

D. 1 RIA-59 (N-16 monitor) increased while1RIA-40 (CSAE Off-gas) remained constant.

2010A NRC REACTOR OPERATOR EXAM Question 51 T2/G1 -CPW 073K5.01, Process Radiation Monitoring (PRM) System Knowledge of the operational implications as they apply to concepts as they apply to the PRM system:

Radiation theory, including sources, types, units, and effects.

(2.5/3.0)

K/A MATCH ANALYSIS Knowledge of the operational implications of process RIA responses are required to determine expected RIA response to SGTR and failed fuel. Additionally, an understanding of N-16 production and decay is needed to understand RIA-59 & 60 responses (or lack of response) to failed fuel. RIA-40 is a process monitor.

ANSWER CHOICE ANALYSIS Answer: A A. Correct: RIA-16 and 40 will respond to ALL activity, therefore an increase in RCS activity, which the stem provides with a degrading fuel failure, would cause all 3 to increase.

B. Incorrect. RIA-40 will be affected by the fuel failure as described in A. Plausible since RIA-40 is reading Air Ejector off gas flow and not directly monitoring the RCS.

C. Incorrect. RIA-40 will be affected by the fuel failure, whereas RIA 59 (N-16 detectors) will not. Plausible since RIA-59 & 60 are Main Steam Line monitors and activity that leaks to the secondary side will pass by the RIAs on the way to the Main Turbine.

D. Incorrect. RIA-40 will be affected by the fuel failure, whereas RIA 59 (N-16 detectors) will not. Plausible since RIA-40 is reading Air Ejector off gas flow and not directly monitoring the RCS.

Technical Reference(s): RAD RIA Proposed references to be provided to applicants during examination: NONE Learning Objective: RAD-RIA R2 Question Source: Bank RADOIO2O7 Question History: Last NRC Exam ONS 2006 Question Cognitive Level: Comprehension and Analysis Question 51 T2/G1 - CPW 2010A NRC REACTOR OPERATOR EXAM 073K5.01, Process Radiation Monitoring (PRM) System Knowledge of the operational implications as they apply to concepts as they apply to the PRM system:

Radiation theory, including sources, types, units, and effects.

(2.5/3.0)

KIA MATCH ANALYSIS Knowledge of the operational implications of process RIA responses are required to determine expected RIA response to SGTR and failed fuel. Additionally, an understanding of N-16 production and decay is needed to understand RIA-59 & 60 responses (or lack of response) to failed fuel. RIA-40 is a process monitor.

ANSWER CHOICE ANALYSIS Answer: A A. Correct: RIA-16 and 40 will respond to ALL activity, therefore an increase in RCS activity, which the stem provides with a degrading fuel failure, would cause all 3 to increase.

B. Incorrect. RIA-40 will be affected by the fuel failure as described in A. Plausible since RIA-40 is reading Air Ejector off gas flow and not directly monitoring the RCS.

C. Incorrect. RIA-40 will be affected by the fuel failure, whereas RIA 59 (N-16 detectors) will not. Plausible since RIA-59 & 60 are Main Steam Line monitors and activity that leaks to the secondary side will pass by the RIA's on the way to the Main Turbine.

D. Incorrect. RIA-40 will be affected by the fuel failure, whereas RIA 59 (N-16 detectors) will not. Plausible since RIA-40 is reading Air Ejector off gas flow and not directly monitoring the RCS.

Technical Reference(s): RAD RIA Proposed references to be provided to applicants during examination: NONE Learning Objective: RAD-RIA R2 Question Source: Bank RAD010207 Question History: Last NRC Exam ONS 2006 Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 52 Unit 1 initial conditions:

Time=1200 Reactor power = 100%

A and B LPSW Pumps operating C LPSW pump in AUTO Blackout Occurs Current conditions:

Time=1230 Both MFBs re-energized Based on the above conditions, which ONE of the following describes the status of the C LPSW pump 5 seconds after the MFBs have re-energized AND the system that will require the use of OP/I /A/1104/010 (Low Pressure Service Water) to return itto service once LPSW pressure has been restored?

A. operating / Reactor Building Aux Coolers B. operating / RBCU Waterhammer Prevention C. NOT operating / Reactor Building Aux Coolers D. NOT operating / RBCU Waterhammer Prevention 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 52 Unit 1 initial conditions:

Time = 1200 Reactor power = 100%

A and B LPSW Pumps operating C LPSW pump in AUTO Blackout Occurs Current conditions:

Time = 1230 Both MFB's re-energized Based on the above conditions, which ONE of the following describes the status of the C LPSW pump 5 seconds after the MFB's have re-energized AND the system that will require the use of OP/1/A/11 04/01 0 (Low Pressure Service Water) to return it to service once LPSW pressure has been restored?

A. operating 1 Reactor Building Aux Coolers B. operating 1 RBCU Waterhammer Prevention C. NOT operating 1 Reactor Building Aux Coolers D. NOT operating 1 RBCU Waterhammer Prevention

2010A NRC REACTOR OPERATOR EXAM Question 52 T2/G1

- CPW 076A2.02, Service Water System (SWS)

Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Service water header pressure.

(2.7/3.1)

KIA MATCH ANALYSIS Requires predicting the impact of low LPSW header pressure due to loss of power on LPSW pump operation and ability to determine which system will require procedure use to mitigate the consequences of the malfunction.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since it is reasonable to assume that the pump in standby will auto start once power is available and LPSW pressure is low. With a 30 minute loss of power, LPSW pressure will be well below the 70 psig start setpoint and the misconception of the 10 second time delay operation would result in this choice. Second part is correct.

B. Incorrect: First part is plausible since it is reasonable to assume that the pump in standby will auto start once power is available and LPSW pressure is low. With a 30 minute loss of power, LPSW pressure will be well below the 70 psig start setpoint and the misconception of the 10 second time delay operation would result in this choice. Second part is plausible since the LPSW RBCU Waterhammer Isolation valves (which are also addressed in AP/24) do isolate at the same low pressure setpoint of 18 psig as the RB Aux Coolers however the Waterhammer Isolation Circuitry will automatically reinstate itself as LPSW pressure returns to normal.

C. CORRECT: Following a LOOP when power has been restored to the Main Feeder Busses if pressure remains 70 psig for 10 seconds the standby LPSW pump will start. If only 5 seconds have passed since MFBs re-energize the standby pump will not be operating. Low LPSW pressure will isolate the RB Auxiliary Coolers and the RBCU Waterhammer Prevention Circuitry. Once the RB Aux Cooler isolation valves close, they must be manually re-opened. AP124 directs the operator to OP/1/A11104/010 (Low Pressure Service Water) to restore the system once pressure is restored.

D. Incorrect: First part is correct. Second part is plausible since the LPSW RB waterhammer mod valves (which are also addressed in AP/24) do reopen automatically once LPSW pressure is restored.

Technical Reference(s): SSS-LPW, AP124 Proposed references to be provided to applicants during examination: NONE Learning Objective: SSS-LPW R13, R6 Question Source: NEW Question History: Last NRC Exam 4L Question Cognitive Level: Comprehension and Analysis Question 52 T2/G1 - CPW 2010A NRC REACTOR OPERATOR EXAM 076A2.02, Service Water System (SWS)

Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Service water header pressure.

(2.7/3.1 )

KIA MATCH ANALYSIS Requires predicting the impact of low LPSW header pressure due to loss of power on LPSW pump operation and ability to determine which system will require procedure use to mitigate the consequences of the malfunction.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since it is reasonable to assume that the pump in standby will auto start once power is available and LPSW pressure is low. With a 30 minute loss of power, LPSW pressure will be well below the 70 psig start setpoint and the misconception of the 10 second time delay operation would result in this choice. Second part is correct.

B. Incorrect: First part is plausible since it is reasonable to assume that the pump in standby will auto start once power is available and LPSW pressure is low. With a 30 minute loss of power, LPSW pressure will be well below the 70 psig start setpoint and the misconception of the 10 second time delay operation would result in this choice. Second part is plausible since the LPSW RBCU Waterhammer Isolation valves (which are also addressed in AP/24) do isolate at the same low pressure setpoint of 18 psig as the RB Aux Coolers however the Waterhammer Isolation Circuitry will automatically reinstate itself as LPSW pressure returns to normal.

C. CORRECT: Following a LOOP when power has been restored to the Main Feeder Busses if pressure remains S 70 psig for 10 seconds the standby LPSW pump will start. If only 5 seconds have passed since MFB's re-energize the standby pump will not be operating. Low LPSW pressure will isolate the RB Auxiliary Coolers and the RBCU Waterhammer Prevention Circuitry. Once the RB Aux Cooler isolation valves close, they must be manually re-opened. AP/24 directs the operator to OP/1/A11104/010 (Low Pressure Service Water) to restore the system once pressure is restored.

D. Incorrect: First part is correct. Second part is plausible since the LPSW RB waterhammer mod valves (which are also addressed in AP/24) do reopen automatically once LPSW pressure is restored.

Technical Reference(s): SSS-LPW, AP/24 Proposed references to be provided to applicants during examination: NONE Learning Objective: SSS-LPW R13, R6 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 53 Initial plant conditions:

Large IA leak occurs Service air header pressure = 87 psig decreasing Turbine Building air header pressure per gage below TUBB BLDG 160

140

120

100

=

80

60

40

20

PSI

0 Based on the above conditions, which ONE of the following describes the ONLY air compressors that will be operating?

A. Diesel Air Compressors AND Primary IA Compressor B. Diesel Air Compressors AND AlA Compressors C. AlA Compressors AND Backup IA Compressors D. Primary IA Compressor AND Backup IA Compressors 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 53 Initial plant conditions:

Large IA leak occurs Service air header pressure = 87 psig decreasing Turbine Building air header pressure per gage below TURB BLDG 160 =

140 =

120 =

100 =

.~=

80 60 40 20 PSI o

Based on the above conditions, which ONE of the following describes the ONLY air compressors that will be operating?

A. Diesel Air Compressors AND Primary IA Compressor B. Diesel Air Compressors AND AlA Compressors C. AlA Compressors AND Backup IA Compressors D. Primary IA Compressor AND Backup IA Compressors

2010A NRC REACTOR OPERATOR EXAM Question 53 T2IGI

- cpw 078A4.O1, Instrument Air System (lAS)

Ability to manually operate andlor monitor in the control room:

Pressure gauges.

(3.1/3.1)

K/A MATCH ANALYSIS Requires demonstrating the ability to monitor an IA pressure gage in the control room and based on the indication determine IA compressor status.

ANSWER CHOICE ANALYSIS Answer:

D A. Incorrect. Diesel Air Compressors will Auto start at 90 psig IA pressure. Gage indicates pressure is slightly> 90 psig. Plausible if you use Service Air pressure.

Second part is correct.

B. Incorrect. Diesel Air Compressors will Auto start at 90 psig IA pressure. Plausible if you use Service Air pressure. AlA compressors start at 88 psig IA pressure therefore would not be operating.

C. Incorrect. AlA compressors start at 88 psig IA pressure therefore would not be operating. Second part is correct.

0. CORRECT. Primary IA compressor would be operating loaded and the Backup IA compressors would start at 93 psig IA pressure Technical Reference(s): SSS-IA Proposed references to be provided to applicants during examination: NONE Learning Objective: SSS-IA R48 Question Source: Modified BANK SSS040802 Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis 2010A NRC REACTOR OPERATOR EXAM Question 53 T2/G1 - cpw 078A4.01, Instrument Air System (lAS)

Ability to manually operate and/or monitor in the control room:

Pressure gauges.

(3.1/3.1)

KIA MATCH ANALYSIS Requires demonstrating the ability to monitor an IA pressure gage in the control room and based on the indication determine IA compressor status.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect. Diesel Air Compressors will Auto start at 90 psig IA pressure. Gage indicates pressure is slightly> 90 psig. Plausible if you use Service Air pressure.

Second part is correct.

B. Incorrect. Diesel Air Compressors will Auto start at 90 psig IA pressure. Plausible if you use Service Air pressure. AlA compressors start at 88 psig IA pressure therefore would not be operating.

C. Incorrect. AlA compressors start at 88 psig IA pressure therefore would not be operating. Second part is correct.

D. CORRECT. Primary IA compressor would be operating loaded and the Backup IA compressors would start at 93 psig IA pressure Technical Reference(s): SSS-IA Proposed references to be provided to applicants during examination: NONE Learning Objective: SSS-IA R48 Question Source: Modified BANK SSS040802 Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 54 Unit 2 conditions:

Reactor power = 100%

Letdown temperature= 112°F stable 2HP-5 failed closed Based on the above conditions, which ONE of the following describes requirements for manually opening 2HP-5 locally in accordance with AP/32 (Loss of Letdown) AND the MINIMUM Pressurizer level (inches) at which a manual reactor trip would be required?

A. Maintain constant communication with operator dispatched to open 2HP-5 / 400 B. Maintain constant communication with operator dispatched to open 2HP-5 / 380 C. Place the LETDOWN HI TEMP INTLK BYP switch to BYPASS / 400 D. Place the LETDOWN HI TEMP INTLK BYP switch to BYPASS / 380 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 54 Unit 2 conditions:

Reactor power = 100%

Letdown temperature = 112°F stable 2HP-5 failed closed Based on the above conditions, which ONE of the following describes requirements for manually opening 2HP-5 locally in accordance with AP/32 (Loss of Letdown) AND the MINIMUM Pressurizer level (inches) at which a manual reactor trip would be required?

A. Maintain constant communication with operator dispatched to open 2HP-5 / 400 B. Maintain constant communication with operator dispatched to open 2HP-5 / 380 C. Place the LETDOWN HI TEMP INTLK BYP switch to BYPASS / 400 D. Place the LETDOWN HI TEMP INTLK BYP switch to BYPASS / 380

2010A NRC REACTOR OPERATOR EXAM Question 54 T2IGI

- CPW 103G2.1.20, Containment System Ability to interpret and execute procedure steps.

(4.6/4.6)

K/A MATCH ANALYSIS Question requires knowledge of procedure steps regarding opening a containment isolation valve (HP-5) that has failed closed and the requirements to properly execute the steps.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since 400 is the maximum measurable pressurizer level (0-400). It is plausible to believe the required Rx trip would be at the point that pressurizer level goes off scale high however 375 is used to account for instrument errors.

B. CORRECT: Since HP-5 is a containment isolation valve operated on ES actuation, constant communication is required to maintain Administrative control over the penetration as allowed by TS 3.6.3. AP132 directs that at 375 Pzr level, trip the Rx.

C. Incorrect: First part is plausible since these directions are in HP-32 for opening HP-5 only it would be correct if HP-5 had closed due to High Letdown temperature and the valve was being opened under those directions. Second part is plausible since 400 is the maximum measurable pressurizer level (0-400). It is plausible to believe the required Rx trip would be at the point that pressurizer level goes off scale high however 375 is used to account for instrument errors.

D. Incorrect: : First part is plausible since these directions are in HP-32 for opening HP-5 only it would be correct if HP-S had closed due to High Letdown temperature and the valve was being opened under those directions. Second part is correct.

Technical Reference(s): AP132 (Loss of Letdown)

Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-APG R8 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis 2010A NRC REACTOR OPERATOR EXAM Question 54 T2/G1 - CPW 1 03G2.1.20, Containment System Ability to interpret and execute procedure steps.

(4.6/4.6)

KIA MATCH ANALYSIS Question requires knowledge of procedure steps regarding opening a containment isolation valve (HP-5) that has failed closed and the requirements to properly execute the steps.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since 400" is the maximum measurable pressurizer level (0"-400"). It is plausible to believe the required Rx trip would be at the point that pressurizer level goes off scale high however 375" is used to account for instrument errors.

B. CORRECT: Since HP-5 is a containment isolation valve operated on ES actuation, constant communication is required to maintain Administrative control over the penetration as allowed by TS 3.6.3. AP/32 directs that at 375" Pzr level, trip the Rx.

C. Incorrect: First part is plausible since these directions are in HP-32 for opening HP-5 only it would be correct if HP-5 had closed due to High Letdown temperature and the valve was being opened un'der those directions. Second part is plausible since 400" is the maximum measurable pressurizer level (0"-400"). It is plausible to believe the required Rx trip would be at the point that pressurizer level goes off scale high however 375" is used to account for instrument errors.

D. Incorrect:: First part is plausible since these directions are in HP-32 for opening HP-5 only it would be correct if HP-5 had closed due to High Letdown temperature and the valve was being opened under those directions. Second part is correct.

Technical Reference(s): AP/32 (Loss of Letdown)

Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-APG R8 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 55 Unit 3 initial conditions:

3CC-8 has been manually opened due to loss of air to the valve Current conditions:

Instrument air has been restored to 3CC-8 3CC-8 remains manually open Based on the above conditions, which ONE of the following describes whether 3CC-8 can be operated from the control room and 3CC-8s response to an ES 1-6 actuation?

3CC-8 operated from the control room and 3CC-8 automatically close.

A. can / will B. can / will NOT C. can NOT / will D. can NOT / will NOT 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 55 Unit 3 initial conditions:

3CC-8 has been manually opened due to loss of air to the valve Current conditions:

Instrument air has been restored to 3CC-8 3CC-8 remains manually open Based on the above conditions, which ONE of the following describes whether 3CC-8 can be operated from the control room and 3CC-8's response to an ES 1-6 actuation?

3CC-8 __ be operated from the control room and 3CC-8 ___ automatically close.

A. can / will B. can / will NOT C. can NOT / will D. can NOT / will NOT

2010A NRC REACTOR OPERATOR EXAM Question 55 T2IGI -CPW 103G2.4.20, Containment System Knowledge of the operational implications of EOP warnings, cautions, and notes.

(3.8/4.3)

K/A MATCH ANALYSIS EOP tabs contain no warnings, cautions, or notes relevant to Containment Systems therefore used one from an AP. Question requires knowledge of operational implications of ES actuations on NOTE in API2O pg 3 regarding CC-8 being manually opened. CC-8 is a containment isolation valve.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Both parts are incorrect but plausible because a candidate could assume that with IA restored the valve would now work from the CR and from an ES signal.

B. Incorrect: First part is incorrect but plausible because a candidate could assume that with IA restored the valve would now work from the CR however there is a mechanical lever that must be disengaged to allow the valve to operate from the control room. Second part is correct.

C. Incorrect: First part is correct. Second part is incorrect but plausible because a candidate could assume that with IA restored the valve would now work from an ES signal. Additionally, this is basically correct for FDW 315 and 316. These valves are manually throttled against spring pressure therefore when these valves are manually throttled they cannot be opened from the control room but would be able to be closed from the control room.

D. CORRECT: Once CC-8 is manually opened there is a NOTE for step 4.3 of API2O that says if manually opened, CC-8 will not operate from the control room. This is true until a lever that was locally engaged to allow manual operation of the valve is disengaged. Until then CC-8 cannot respond to a signal from the CR to close.

Technical Reference(s): PNS-CC, API2O (Loss of Component Cooling), CF-EF Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-CC R13, R14 EAP-APG R6, R8 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 55 T2/G1 - CPW 103G2.4.20, Containment System Knowledge of the operational implications of EOP warnings, cautions, and notes.

(3.8/4.3)

KIA MATCH ANALYSIS EOP tabs contain no warnings, cautions, or notes relevant to Containment Systems therefore used one from an AP. Question requires knowledge of operational implications of ES actuations on NOTE in AP/20 pg 3 regarding CC-8 being manually opened. CC-8 is a containment isolation valve.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Both parts are incorrect but plausible because a candidate could assume that with IA restored the valve would now work from the CR and from an ES signal.

B. Incorrect: First part is incorrect but plausible because a candidate could assume that with IA restored the valve would now work from the CR however there is a mechanical lever that must be disengaged to allow the valve to operate from the control room. Second part is correct.

C. Incorrect: First part is correct. Second part is incorrect but plausible because a candidate could assume that with IA restored the valve would now work from an ES signal. Additionally, this is basically correct for FDW 315 and 316. These valves are manually throttled against spring pressure therefore when these valves are manually throttled they cannot be opened from the control room but would be able to be closed from the control room.

D. CORRECT: Once CC-8 is manually opened there is a NOTE for step 4.3 of AP/20 that says if manually opened, CC-8 will not operate from the control room. This is true until a lever that was locally engaged to allow manual operation of the valve is disengaged. Until then CC-8 cannot respond to a signal from the CR to close.

Technical Reference(s): PNS-CC, AP/20 (Loss of Component Cooling), CF-EF Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-CC R13, R14 EAP-APG R6, R8 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 56 Unit 1 initial conditions:

100% power 450 EFPD Based on the above conditions, which ONE of the following events is the cause of the indications on the attached P/T display?

SEE ATTACHMENT ASSUME NO OPERATOR ACTIONS A. SBLOCA with a Reactor Trip on variable low pressure B. LBLOCA with a Reactor Trip on low RCS pressure C. MSLB with a Reactor Trip on AFIS initiation D. Loss of MFDW with a Reactor Trip due to losing both MFDWPs 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 56 Unit 1 initial conditions:

100% power 450 EFPD Based on the above conditions, which ONE of the following events is the cause of the indications on the attached PIT display?

SEE ATTACHMENT ASSUME NO OPERATOR ACTIONS A. SBLOCA with a Reactor Trip on variable low pressure B. LBLOCA with a Reactor Trip on low RCS pressure C. MSLB with a Reactor Trip on AFIS initiation D. Loss of MFDW with a Reactor Trip due to losing both MFDWPs

2010A NRC REACTOR OPERATOR EXAM Question 56 T21G2

- CPW roo2A4.o8, Reactor Coolant System (RCS)

Ability to manually operate and/or monitor in the control room:

Safety parameter display systems.

(3.4/3.7)

KIA MATCH ANALYSIS Per NRC, question using OAC P/T display vs Westinghouse SPDS display info is OK.

Requires ability to monitor OAC PIT display and interprete RCS pressure and temperature responses to determine the event in progress ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: the variable low pressure setpoint was never reached (diagonal line), Th

& Tc stayed apart as they decreased (indicating a large overcooling event).

B. Incorrect: Th & Tc decreased as pressure decreased. RB pressure is only 0.2 psi.

C. Correct: A SG pressure is 370 psi with indications of severe overcooling.

D. Incorrect: The display indicates overcooling. A loss of MFWPs would be indicated as under cooling (Temps increasing).

Technical Reference(s): SF-OlD (PTID), EAP-EHT Proposed references to be provided to applicants during examination: NONE Learning Objective: SF-0l0 R9, EAP-EHT R14 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis Question 56 T2/G2 - CPW 2010A NRC REACTOR OPERATOR EXAM 002A4.08, Reactor Coolant System (RCS)

Ability to manually operate andlor monitor in the control room:

Safety parameter display systems.

(3.4/3.7)

KIA MATCH ANALYSIS NRC, question using OAC PIT display vs Westinghouse SPDS display info is OK.

Requires ability to monitor OAC PIT display and interprete RCS pressure and temperature responses to determine the event in progress ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: the variable low pressure setpoint was never reached (diagonal line), Th

& Tc stayed apart as they decreased (indicating a large overcooling event).

B. Incorrect: Th & Tc decreased as pressure decreased. RB pressure is only 0.2 psi.

C. Correct: "A" SG pressure is -370 psi with indications of severe overcooling.

D. Incorrect: The display indicates overcooling. A loss of MFWPs would be indicated as under cooling (Temps increasing).

Technical Reference(s): SF-010 (PTID), EAP-EHT Proposed references to be provided to applicants during examination: NONE Learning Objective: SF-010 R9, EAP-EHT R14 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 57 Which ONE of the following describes the power supply for the Unit 1 Group B Pressurizer heaters?

A.

1XH B.

1XI C. 1XJ D. 1XSF 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 57 Which ONE of the following describes the power supply for the Unit 1 Group B Pressurizer heaters?

A. 1XH B. 1XI C. 1XJ D. 1XSF

2010A NRC REACTOR OPERATOR EXAM Question 57 T21G2

- CPW 011 K2.02, Pressurizer Level Control System Knowledge of bus power supplies to the following:

PZR heaters.

(3.1/3.2)

K/A MATCH ANALYSIS Question requires knowledge of power supply to Group B Pressurizer heaters as well as the Operators ability to control Group B heaters ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: The normal supply for Group B is IXSF.

B. Incorrect: Plausible since IXH is the normal supply to group E pressurizer heaters.

C. Incorrect: Plausible since 1XI is the normal supply to Group F pressurizer heaters.

D. Incorrect: Plausible since 1XJ is the normal supply to Group D pressurizer heaters.

Technical Reference(s): PNS-PZR Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R7 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 57 T2/G2 - CPW 011 K2.02, Pressurizer Level Control System Knowledge of bus power supplies to the following:

PZR heaters.

(3.1/3.2)

KIA MATCH ANALYSIS Question requires knowledge of power supply to Group 8 Pressurizer heaters as well as the Operators ability to control Group 8 heaters ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: The normal supply for Group 8 is 1XSF.

B. Incorrect: Plausible since 1 XH is the normal supply to group E pressurizer heaters.

C. Incorrect: Plausible since 1XI is the normal supply to Group F pressurizer heaters.

D. Incorrect: Plausible since 1XJ is the normal supply to Group 0 pressurizer heaters.

Technical Reference(s): PNS-PZR Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R7 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 58 Which ONE of the following describes who determines that a RB Continuous Release is allowed and after it is started what are the requirements for sampling the RB atmosphere in accordance with 0P11102/014 (RB Purge System)?

A. CRSRO / Release may continue indefinitely after initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without submitting daily sample requests.

B. CRSRO / Release may continue indefinitely provided RP assigns a new GWR number and sample results are entered in the Unit Log every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. Shift RP / Release may continue indefinitely after initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without submitting daily sample requests.

D. Shift RP / Release may continue indefinitely provided RP assigns a new GWR number and sample results are entered in the Unit Log every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 58 Which ONE of the following describes who determines that a RB Continuous Release is allowed and after it is started what are the requirements for sampling the RB atmosphere in accordance with OP/1102/014 (RB Purge System)?

A. CRSRO I Release may continue indefinitely after initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without submitting daily sample requests.

B. CRSRO I Release may continue indefinitely provided RP assigns a new GWR number and sample results are entered in the Unit Log every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. Shift RP I Release may continue indefinitely after initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without submitting daily sample requests.

D. Shift RP I Release may continue indefinitely provided RP assigns a new GWR number and sample results are entered in the Unit Log every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2010A NRC REACTOR OPERATOR EXAM Question 58 T21G2

- cpw 029A2.04, Containment Purge System (CPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Health physics sampling of containment atmosphere.

(2.5*/3.2*)

K/A MATCH ANALYSIS Requires knowledge of HP sample requirements for a continuous RB purge and who can initiate a continuous release.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: first part is incorrect. Plausible because this is who normally makes decisions about the unit. Second part is correct.

B. Incorrect: first part is incorrect. Plausible because this is who normally makes decisions about the unit. The second part is incorrect. Plausible because this is the normal required sampling frequency.

C. Correct, RP determines when to put the RB on continuous release. Once on continuous release the RB is not required to be sampled.

D. Incorrect, first part is correct. The second part is incorrect. Plausible because this is the normal required sampling frequency.

Technical Reference(s): 0P111A111021014 (RB Purge) Rev, PNS-RBP Rev 10 Proposed references to be provided to applicants during examination: None Learning Objective: PNS-RBP Obj R4,5,8 Question Source: Bank Question History: Last NRC Exam: ONS 2007 Re-test #61 (Re-ordered answers)

Question Cognitive Level:

Memory or Fundamental Knowledge 2010A NRC REACTOR OPERATOR EXAM Question 58 T2/G2 - cpw 029A2.04, Containment Purge System (CPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Health physics sampling of containment atmosphere.

(2.5*/3.2*)

KIA MATCH ANALYSIS Requires knowledge of HP sample requirements for a continuous RB purge and who can initiate a continuous release.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: first part is incorrect. Plausible because this is who normally makes decisions about the unit. Second part is correct.

B. Incorrect: first part is incorrect. Plausible because this is who normally makes decisions about the unit. The second part is incorrect. Plausible because this is the normal required sampling frequency.

C. Correct, RP determines when to put the RB on continuous release. Once on continuous release the RB is not required to be sampled.

D. Incorrect, first part is correct. The second part is incorrect. Plausible because this is the normal required sampling frequency.

Technical Reference(s): OP/1/A11102/014 (RB Purge) Rev, PNS-RBP Rev 10 Proposed references to be provided to applicants during examination: None Learning Objective: PNS-RBP Obj R4,5,8 Question Source: Bank Question History: Last NRC Exam: ONS 2007 Re-test #61 (Re-ordered answers)

Question Cognitive Level:

Memory or Fundamental Knowledge

2010A NRC REACTOR OPERATOR EXAM I POINT Question 59 Plant conditions:

Spent Fuel Storage Cask has been dropped in Unit 1&2 SEP Spent Fuel damage is visible RIA-6 and RIA-41 HIGH alarm actuates Spent Fuel Pool level = -3.5 feet decreasing Based on the above conditions, which ONE of the following describes which filters will be used to reduce off site releases and the status of the SE Pumps?

Unit Reactor Building Purge filters and the Spent Fuel Cooling pumps will be A.1

/ON B.

1

/ OFF C.2 ION D. 2 I OFF 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 59 Plant conditions:

Spent Fuel Storage Cask has been dropped in Unit 1 &2 SFP Spent Fuel damage is visible RIA-6 and RIA-41 HIGH alarm actuates Spent Fuel Pool level = -3.5 feet decreasing Based on the above conditions, which ONE of the following describes which filters will be used to reduce off site releases and the status of the SF Pumps?

Unit Reactor Building Purge filters and the Spent Fuel Cooling pumps will be A. 1 / ON B. 1 / OFF C. 2 / ON D. 2 / OFF

2010A NRC REACTOR OPERATOR EXAM Question 59 T21G2 - cpw 033A3.02, Spent Fuel Pool Cooling System (SFPCS)

Ability to monitor automatic operation of the Spent Fuel Pool Cooling System including:

Spent fuel leak or rupture.

(2.9/3.1)

K/A MATCH ANALYSIS Knowledge of automatic operation of the SF Cooling pumps on a decreasing SF Pool level and SFPV as a result of a Spent Fuel Pool accident is required.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is Plausible since Unit I and Unit 2 share Spent Fuel Pools and there is only one set of filters needed for the Spent Fuel Filtered Exhaust system.

Since there are no dedicated filters, Unit 2s filters Reactor Building Purge filters are used. Second part is plausible since 4 is the level at which SF Pumps loose suction and level is still > 4 feet.

B. Incorrect: First part is Plausible since Unit 1 and Unit 2 share Spent Fuel Pools and there is only one set of filters needed for the Spent Fuel Filtered Exhaust system.

Since there are no dedicated filters, Unit 2s filters Reactor Building Purge filters are used. Second part is correct C. Incorrect: First part is correct. Second part is plausible since 4 is the level at which SF Pumps loose suction and level is still > 4 feet.

0. CORRECT: Unit I and Unit 2 share Spent Fuel Pools and there is only one set of filters needed for the Spent Fuel Filtered Exhaust system. Since there are no dedicated filters, Unit 2s filters Reactor Building Purge filters are used. The Spent Fuel Cooling pumps have a low level trip at -2.5 feet. Since level is -3.5 feet the pumps would be off.

Technical Reference(s): RAD-RIA, FH-SFC, FH-FES Proposed references to be provided to applicants during examination: NONE Learning Objective: RAD-RIA R2 FH-FES R2, and FH-SFC R3 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals Question 59 T2/G2 - cpw 2010A NRC REACTOR OPERATOR EXAM 033A3.02, Spent Fuel Pool Cooling System (SFPCS)

Ability to monitor automatic operation of the Spent Fuel Pool Cooling System including:

Spent fuel leak or rupture.

(2.9/3.1 )

KIA MATCH ANALYSIS Knowledge of automatic operation of the SF Cooling pumps on a decreasing SF Pool level and SFPV as a result of a Spent Fuel Pool accident is required.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is Plausible since Unit 1 and Unit 2 share Spent Fuel Pools and there is only one set of filters needed for the Spent Fuel Filtered Exhaust system.

Since there are no dedicated filters, Unit 2's filters Reactor Building Purge filters are used. Second part is plausible since 4' is the level at which SF Pumps loose suction and level is still > 4 feet.

B. Incorrect: First part is Plausible since Unit 1 and Unit 2 share Spent Fuel Pools and there is only one set of filters needed for the Spent Fuel Filtered Exhaust system.

Since there are no dedicated filters, Unit 2's filters Reactor Building Purge filters are used. Second part is correct C. Incorrect: First part is correct. Second part is plausible since 4' is the level at which SF Pumps loose suction and level is still> 4 feet.

D. CORRECT: Unit 1 and Unit 2 share Spent Fuel Pools and there is only one set of filters needed for the Spent Fuel Filtered Exhaust system. Since there are no dedicated filters, Unit 2's filters Reactor Building Purge filters are used. The Spent Fuel Cooling pumps have a low level trip at -2.5 feet. Since level is -3.5 feet the pumps would be off.

Technical Reference(s): RAD-RIA, FH-SFC, FH-FES Proposed references to be provided to applicants during examination: NONE Learning Objective: RAD-RIA R2 FH-FES R2, and FH-SFC R3 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 60 Unit 2 initial conditions:

Reactor power = 100%

Current conditions:

Reactor trip Controlling 2A Steam Generator Outlet Pressure fails HIGH Based on the above conditions, which ONE of the following describes the 2A AND 2B Turbine Bypass Valves (TBVs) response?

The 2A TBVs will fully open AND the 2B TBVs will fully open A. then return to throttled position I then return to throttled position B. then return to throttled position I and remain fully open C. and remain fully open I then return to throttled position D. and remain fully open I and remain fully open 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 60 Unit 2 initial conditions:

Reactor power = 100%

Current conditions:

Reactor trip Controlling 2A Steam Generator Outlet Pressure fails HIGH Based on the above conditions, which ONE of the following describes the 2A AND 2B Turbine Bypass Valves (TBV's) response?

The 2A TBV's will fully open __ AND the 2B TBV's will fully open __

A. then return to throttled position / then return to throttled position B. then return to throttled position / and remain fully open C. and remain fully open / then return to throttled position D. and remain fully open / and remain fully open

2010A NRC REACTOR OPERATOR EXAM Question 60 T21G2

- cpw 041 K6.03, Steam Dump System (SDS) and Turbine Bypass Control Knowledge of the effect of a loss or malfunction on the following will have on the SOS:

Controller and positioners, including ICS, SIG, CRDS.

(2.7/2.9)

K/A MATCH ANALYSIS Requires knowledge of signal inputs to Turbine Bypass valve controls and the effect that a failed open controller will have on Turbine Bypass Control ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since it would be correct for normal post trip response without the instrument failure. Additionally plausible since the misconception that the TBVs continue to control from Turbine Header Pressure post trip would also lead to this choice however when the Turbine Bailey station trips to hand (which it does on a Rx trip) the controlling signal for the TBVs swaps from being Turbine Header Pressure to Steam Generator Outlet Pressure. Second part is correct.

B. Incorrect: First part is plausible since it would be correct for normal post trip response without the instrument failure. Additionally plausible since the misconception that the TBVs continue to control from Turbine Header Pressure post trip would also lead to this choice however when the Turbine Bailey station trips to hand (which it does on a Rx trip) the controlling signal for the TBVs swaps from being Turbine Header Pressure to Steam Generator Outlet Pressure. Second part is plausible since it would be correct if you mistakenly applied the failed instrument to the B TBVs instead of the A TBVs OR if you are under the misconception that both sets of TBVs control from the same SG Outlet Pressure as is the case pre-trip when using Turbine header pressure as the controlling signal.

C. CORRECT: When the reactor trips, both sets of TBVs would normally go full open in an attempt to relieve enough steam from the SGs to gain control of SG Outlet pressure. Shortly after the trip (generally less than a minute) both sets of TBVs will be back to the throttled position and in control of SG Outlet Pressure. With the controlling pressure for the A TBV failed high, the A TBV would remain full open since it will always believe that actual pressure is greater than setpoint.

D. Incorrect: First part is correct. Second part is plausible since it would be correct if you mistakenly applied the failed instrument to the B TBVs instead of the A TBVs OR if you are under the misconception that both sets of TBVs control from the same SG Outlet Pressure as is the case pre-trip when using Turbine header pressure as the controlling signal.

Technical Reference(s): STG-ICS Chapter 3 Proposed references to be provided to applicants during examination: NONE Learning Objective: STG-ICS RIO Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 60 T2/G2 - cpw 2010A NRC REACTOR OPERATOR EXAM 041 K6.03, Steam Oump System (SOS) and Turbine Bypass Control Knowledge of the effect of a loss or malfunction on the following will have on the 505:

Controller and positioners, including ICS, S/G, CROS.

(2.7/2.9)

KIA MATCH ANALYSIS Requires knowledge of signal inputs to Turbine Bypass valve controls and the effect that a failed open controller will have on Turbine Bypass Control ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: First part is plausible since it would be correct for normal post trip response without the instrument failure. Additionally plausible since the misconception that the TBV's continue to control from Turbine Header Pressure post trip would also lead to this choice however when the Turbine Bailey station trips to hand (which it does on a Rx trip) the controlling signal for the TBV's swaps from being Turbine Header Pressure to Steam Generator Outlet Pressure. Second part is correct.

B. Incorrect: First part is plausible since it would be correct for normal post trip response without the instrument failure. Additionally plausible since the misconception that the TBV's continue to control from Turbine Header Pressure post trip would also lead to this choice however when the Turbine Bailey station trips to hand (which it does on a Rx trip) the controlling signal for the TBV's swaps from being Turbine Header Pressure to Steam Generator Outlet Pressure. Second part is plausible since it would be correct if you mistakenly applied the failed instrument to the B TBV's instead of the A TBV's OR if you are under the misconception that both sets of TBV's control from the same SG Outlet Pressure as is the case pre-trip when using Turbine header pressure as the controlling signal.

C. CORRECT: When the reactor trips, both sets of TBV's would normally go full open in an attempt to relieve enough steam from the sG's to gain control of sG Outlet pressure. Shortly after the trip (generally less than a minute) both sets of TBV's will be back to the throttled position and in control of sG Outlet Pressure. With the controlling pressure for the A TBV failed high, the A TBV would remain full open since it will always believe that actual pressure is greater than setpoint.

O. Incorrect: First part is correct. Second part is plausible since it would be correct if you mistakenly applied the failed instrument to the B TBV's instead of the A TBV's OR if you are under the misconception that both sets of TBV's control from the same SG Outlet Pressure as is the case pre-trip when using Turbine header pressure as the controlling signal.

Technical Reference(s): sTG-ICs Chapter 3 Proposed references to be provided to applicants during examination: NONE Learning Objective: sTG-ICs R10 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 61 Unit 3 initial conditions:

Reactor power = 100%

3MS-112 & 3MS-173 (SSRH 3A/3B Controls) are in MANUAL 3MS-77, 78, 80, 81 (MS to SSRHs) are in AUTO Current conditions:

Main Turbine trips Based on the above conditions, which ONE of the following describes the plant response?

A. 3M5-1 12 & 3MS-173 valve demand will throttle back with load B. 3MS-1 12 & 3MS-173 valve demand will remain full open but the air will be ported off, causing the valves to close C. 3MS-77, 78, 80, and 81 will close when the air is dumped off of the valves D. 3MS-77, 78, 80, and 81 will close as SSRH inlet pressure decreases 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 61 Unit 3 initial conditions:

Reactor power = 100%

3MS-112 & 3MS-173 (SSRH 3A/3B Controls) are in MANUAL 3MS-77, 78, 80, 81 (MS to SSRH's) are in AUTO Current conditions:

Main Turbine trips Based on the above conditions, which ONE of the following describes the plant response?

A. 3MS-112 & 3MS-173 valve demand will throttle back with load B. 3MS-112 & 3MS-173 valve demand will remain full open but the air will be ported off, causing the valves to close C. 3MS-77, 78, 80, and 81 will close when the air is dumped off of the valves D. 3MS-77, 78, 80, and 81 will close as SSRH inlet pressure decreases

2010A NRC REACTOR OPERATOR EXAM Question 61 T21G2

- cpw 045K3.O1, Main Turbine Generator (MT/G) System Knowledge of the effect that a loss or malfunction of the MT/G system will have on the following:

Remainder of the plant.

(2.9/3.2)

K/A MATCH ANALYSIS Requires knowledge of the effect of a Turbine Trip on other components in the plant ANSWER CHOICE ANALYSIS Answer: D A. INCORRECT: 3MS-1 12 & 173 will close in manual when the Main Turbine trips.

Plausible because they will throttle back in a runback condition.

B. lncorrect:The signal from the Moore Controllers will direct the valves to close.

Plausible because this and the air being bled off of the valves causes them to close.

C. Incorrect: 3MS-77, 78, 80 and 81 are Motor Operated Valves. Plausible because they close in this situation D. Correct: Upon receipt of a Turbine Trip signal, and a decrease in SSRH inlet pressure, MS-77, 78, 80 and 81 will close regardless of the switch positions in the Control Room and on the Heater Panel unless power is not available.

Technical Reference(s): STG-MSR Proposed references to be provided to applicants during examination: NONE Learning Objective: STG-MSR R18 Question Source: Bank Question History: Last NRC Exam 2007 NRC exam Q #38 Question Cognitive Level:

Comprehension and Analysis 2010A NRC REACTOR OPERATOR EXAM Question 61 T2/G2 - cpw 045K3.01, Main Turbine Generator (MT/G) System Knowledge of the effect that a loss or malfunction of the MT/G system will have on the following:

Remainder of the plant.

(2.9/3.2)

KIA MATCH ANALYSIS Requires knowledge of the effect of a Turbine Trip on other components in the plant ANSWER CHOICE ANALYSIS Answer: D A. INCORRECT: 3MS-112 & 173 will close in manual when the Main Turbine trips.

Plausible because they will throttle back in a "runback condition".

B. IncorrectThe signal from the Moore Controllers will direct the valves to close.

Plausible because this and the air being bled off of the valves causes them to close.

C. Incorrect 3MS-77, 78, 80 and 81 are Motor Operated Valves. Plausible because they close in this situation D. Correct: Upon receipt of a Turbine Trip signal, and a decrease in SSRH inlet pressure, MS-77, 78,80 and 81 will close regardless of the switch positions in the Control Room and on the Heater Panel unless power is not available.

Technical Reference(s): STG-MSR Proposed references to be provided to applicants during examination: NONE Learning Objective: STG-MSR R18 Question Source: Bank Question History: Last NRC Exam 2007 NRC exam Q #38 Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 62 Unit 1 plant conditions:

OP/I/A I 106/002A (Condensate And FDW System Startup And Shutdown) End.

4.2 (Condensate And FDW System Startup) in progress 1A Hotwell pump (HWP) operating CBP suction pressure = 45 psig stable The 3 square amber lights located above the HWP PUMP AMP gages are ON Procedure directs placing a standby Hotwell pump to AUTO Based on the above conditions, which ONE of the following describes what the Hotwell pump amber lights indicate AND the I B Hotwell pump response once its control switch is placed in AUTO?

1A Hotwell pump is operating with and the lB Hotwell pump automatically start when its control switch is placed in Auto.

A. low suction pressure / will B. low discharge pressure / will C. low suction pressure / will NOT D. low discharge pressure / will NOT Fiw iA4WPjH i i 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 62 Unit 1 plant conditions:

OP/1/A 1106/002A (Condensate And FDW System Startup And Shutdown) Encl.

4.2 (Condensate And FDW System Startup) in progress 1 A Hotwell pump (HWP) operating CBP suction pressure = 45 psig stable The 3 square amber lights located above the HWP PUMP AMP gages are "ON" Procedure directs placing a standby Hotwell pump to "AUTO" Based on the above conditions, which ONE of the following describes what the Hotwell pump amber lights indicate AND the 1 B Hotwell pump response once it's control switch is placed in AUTO?

1 A Hotwell pump is operating with __ and the 1 B Hotwell pump __ automatically start when its control switch is placed in Auto.

A. low suction pressure / will B. low discharge pressure / will C. low suction pressure / will NOT D. low discharge pressure / will NOT

, WI' lA fwp '"It"'!' 1~ +--

150_

_I" 200_

D

~ ~ 150~

I'~ 100

~ :.50 ::-,""

~

~"

~

n =

',0 =

200

',' 100-

2010A NRC REACTOR OPERATOR EXAM Question 62 T21G2

- cpw 056G2.2.44, Condensate System Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(4.2/4.4)

K/A MATCH ANALYSIS Question requires interpreting light indications for HWP discharge pressure as well as understanding how placing HWP switch to auto per procedure would affect plant conditions ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: The second part is correct as the 1 B Hotwell Pump will start on low CBP suction pressure of < 55 psig when placed in Automatic, however the amber light indicates low HWP discharge pressure and not low suction pressure. Plausible since many pumps have low suction pressure indications and controls due to inadequate NPSH issues therefore having indication of low suction pressure (vs low disch pressure) is a more logical choice though incorrect in this application.

B. CORRECT: The amber light does indicate low HWP discharge pressure and illuminates at <125 psig discharge pressure. The lB Hotwell pump will start when its control switch is placed in Auto due to low CBP suction pressure <

55 psig. (CF-C Pages 21 & 22)

C. Incorrect: The amber light indicates low HWP discharge pressure and not low suction pressure. Plausible since many pumps have low suction pressure indications and controls due to inadequate NPSH issues therefore having indication of low suction pressure (vs low disch pressure) is a more logical choice though incorrect in this application. The HWP will start since CBP suction pressure is < 55 psig.

Plausible since the light above the HWP starting switch is illuminated it would be reasonable to interpret it as a protective interlock to prevent the pump from starting similar to RCP starting interlock indications.

D. Incorrect: The first part is correct however the HWP will start since CBP suction pressure is < 55 psig. Plausible since the light above the HWP starting switch is illuminated it would be reasonable to interpret it as a protective interlock to prevent the pump from starting similar to RCP starting interlock indications.

Technical Reference(s): CF-C, OPIIIAIIIO6IOO2A End. 4.2 Proposed references to be provided to applicants during examination: NONE Learning Objective: CF-C R9, R28 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis Question 62 T2/G2 - cpw 2010A NRC REACTOR OPERATOR EXAM 056G2.2.44, Condensate System Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and

. system conditions.

(4.2/4.4)

KIA MATCH ANALYSIS Question requires interpreting light indications for HWP discharge pressure as well as understanding how placing HWP switch to auto per procedure would affect plant conditions ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: The second part is correct as the 1 B Hotwell Pump will start on low CBP suction pressure of < 55 psig when placed in Automatic, however the amber light indicates low HWP discharge pressure and not low suction pressure. Plausible since many pumps have low suction pressure indications and controls due to inadequate NPSH issues therefore having indication of low suction pressure (vs low disch pressure) is a more logical choice though incorrect in this application.

B. CORRECT: The amber light does indicate low HWP discharge pressure and illuminates at <125 psig discharge pressure. The 1B Hotwell pump will start when its control switch is placed in Auto due to low CBP suction pressure <

55 psig. (CF-C Pages 21 & 22)

C. Incorrect: The amber light indicates low HWP discharge pressure and not low suction pressure. Plausible since many pumps have low suction pressure indications and controls due to inadequate NPSH issues therefore having indication of low suction pressure (vs low disch pressure) is a more logical choice though incorrect in this application. The HWP will start since CBP suction pressure is < 55 psig.

Plausible since the light above the HWP starting switch is illuminated it would be reasonable to interpret it as a protective interlock to prevent the pump from starting similar to RCP starting interlock indications.

D. Incorrect: The first part is correct however the HWP will start since CBP suction pressure is < 55 psig. Plausible since the light above the HWP starting switch is illuminated it would be reasonable to interpret it as a protective interlock to prevent the pump from starting similar to RCP starting interlock indications.

Technical Reference(s): CF-C, OP/1/A11106/002A Encl. 4.2 Proposed references to be provided to applicants during examination: NONE Learning Objective: CF-C R9, R28 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 63 Unit 3 plant conditions:

Reactor power = 100%

Fuel movement in progress in SFP 3RIA-6 (Spent Fuel Pool) in HIGH alarm Based on the above conditions, which ONE of the following describes action(s) that will occur?

A. 3RIA-6 audible alarm will automatically sound.

B. the Spent Fuel Pool Ventilation system will be automatically isolated.

C. the Spent Fuel Filtered Exhaust fans will be manually started from the Control Room.

D. the Outside Air Booster Fans will be manually started from the Spent Fuel Pool entrance area.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 63 Unit 3 plant conditions:

Reactor power = 100%

Fuel movement in progress in SFP 3RIA-6 (Spent Fuel Pool) in HIGH alarm B?sed on the above conditions, which ONE of the following describes action(s) that will occur?

A. 3RIA-6 audible alarm will automatically sound.

B. the Spent Fuel Pool Ventilation system will be automatically isolated.

C. the Spent Fuel Filtered Exhaust fans will be manually started from the Control Room.

D. the Outside Air Booster Fans will be manually started from the Spent Fuel Pool entrance area.

2010A NRC REACTOR OPERATOR EXAM Question 63 T21G2

- cpw 072K1.03, Area Radiation Monitoring (ARM) System Knowledge of the physical connections and/or cause-effect relationships between the ARM system and the following systems:

Fuel building isolation.

(3.6/3.7)

K/A MATCH ANALYSIS Per NRC, actions based on fuel damage and RIA-6 OK if distracter includes auto isolation of SFP ventilation systems Requires knowledge of any cause-effect relationship between ARM alarms and SFP isolation (none).

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: 3RIA-6 causes an audible alarm.

B. Incorrect: Plausible since it could be seen as desirable for the ventilation systems in the SFP (which exhausts to the Unit vent) to auto isolate on high radiation levels.

C. Incorrect: Plausible since AP/9 does require starting one of the Spent Fuel filtered exhaust fans when there is fuel damage in the SEP however they are NOT started from the Control Room but in the entrance area to the SEP itself.

D. Incorrect: Plausible since the Outside Air Booster Fans are manually started in accordance with AP/9 (Spent Fuel damage) when there is fuel damage in the SEP however the controls for the fans are located in the Control Room.

Technical Reference(s): API9, RAD-RIA Proposed references to be provided to applicants during examination:

NONE Learning Objective: RAD-RIA R15, FH-FHS R31 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 63 T2/G2 - cpw 072K1.03, Area Radiation Monitoring (ARM) System Knowledge of the physical connections and/or cause-effect relationships between the ARM system and the following systems:

Fuel building isolation.

(3.6/3.7)

KIA MATCH ANALYSIS N

on fuel ventilation and RIA-6 OK if distracter includes auto Requires knowledge of any cause-effect relationship between ARM alarms and SFP isolation (none).

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: 3RIA-6 causes an audible alarm.

B. Incorrect: Plausible since it could be seen as desirable for the ventilation systems in the SFP (which exhausts to the Unit vent) to auto isolate on high radiation levels.

C. Incorrect: Plausible since AP/9 does require starting one of the Spent Fuel filtered exhaust fans when there is fuel damage in the SFP however they are NOT started from the Control Room but in the entrance area to the SFP itself.

D. Incorrect: Plausible since the Outside Air Booster Fans are manually started in accordance with AP/9 (Spent Fuel damage) when there is fuel damage in the SFP however the controls for the fans are located in the Control Room.

Technical Reference(s): AP/9, RAD-RIA Proposed references to be provided to applicants during examination: NONE Learning Objective: RAD-RIA R15, FH-FHS R31 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

Question 64 2010A NRC REACTOR OPERATOR EXAM I POINT uJ D

(1)

Co uJ U

IA Pressure vs. Time 1230 Based on the graph above, which ONE of the following describes the time at which SA 141 (SA to IA Controller) will automatically open?

A. 1207 B. 1210 C. 1212 100 1200 1210 1220 TIME D. 1215 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 64 IA Pressure VS. Time 100~~------------------------------------------~

W 0::

J if) if)

W 0::

0...

90........................... :........................... :......................... '"

80........................... :........................... :...........................

70~----~~-----4*-------+------~*

---

~----~

1200 1210 1220 1230 TIME Based on the graph above, which ONE of the following describes the time at which SA-141 (SA to IA Controller) will automatically open?

A. 1207 B. 1210 C. 1212 D. 1215

2010A NRC REACTOR OPERATOR EXAM Question 64 T21G2 - cpw 079K4.01, Station Air System (SAS)

Knowledge of SAS design feature(s) andlor interlock(s) which provide for the following:

Cross-connect with lAS.

(2.9/3.2)

K/A MATCH ANALYSIS Requires knowledge of automatic cross-connect between Service air and Instrument air systems.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since 93 psig is the pressure at which the Backup IA compressors will start.

B. Incorrect: Plausible since 90 psig is the pressure at which the Diesel Air Compressors will start C. Incorrect: Plausible sine 88 psig is the pressure at which the AlA compressors will start D. CORRECT: SA to IA Controller (SA-141) valve senses the IA system pressure and opens at 85 psig to allow service air into the IA system.

Technical Reference(s): SSS-IA, AP121A117001022 Loss of Instrument Air Proposed references to be provided to applicants during examination: NONE Learning Objective: SSS-IA R52, R27 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis 2010A NRC REACTOR OPERATOR EXAM Question 64 T2/G2 - cpw 079K4.01, Station Air System (SAS)

Knowledge of SAS design feature(s) and/or interlock(s) which provide for the following:

Cross-connect with lAS.

(2.9/3.2)

KIA MATCH ANALYSIS Requires knowledge of automatic cross-connect between Service air and Instrument air systems.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since 93 psig is the pressure at which the Backup IA compressors will start.

B. Incorrect: Plausible since 90 psig is the pressure at which the Diesel Air Compressors will start C. Incorrect: Plausible sine 88 psig is the pressure at which the AlA compressors will start D. CORRECT: SA to IA Controller (SA-141) valve senses the IA system pressure and opens at 85 psig to allow service air into the IA system.

Technical Reference(s): SSS-IA, AP/2/Al1700/022 Loss of Instrument Air Proposed references to be provided to applicants during examination: NONE Learning Objective: SSS-IA R52, R27 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 65 Unit 3 plant conditions:

Reactor power = 100%

Fire in progress in area of 3TE switchgear 3TE has been de-energized Based on the above conditions, which ONE of the following pieces of equipment is NOT available?

A. 3AMDEFWP B. 3CFIPIP C. 3B MDEFWP D. 3ALPSWP 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 65 Unit 3 plant conditions:

Reactor power = 100%

Fire in progress in area of 3TE switchgear 3TE has been de-energized Based on the above conditions, which ONE of the following pieces of equipment is NOT available?

A. 3A MDEFWP B. 3C HPIP C. 3B MDEFWP D. 3A LPSWP

2010A NRC REACTOR OPERATOR EXAM Question 65 T2/G2-cpw 086K5.03, Fire Protection System (FPS)

Knowledge of the operational implication of the following concepts as they apply to the Fire Protection System:

Effect of water spray on electrical components.

(3.1/3.4)

K/A MATCH ANALYSIS er NRC put fue n a specfc location which reqwies de-energizing a Load Center or 4160V bus and ask about equipment effected.

Requires knowledge of operational implications of de-energizing 3TE to allow water spray to extinguish fire.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: 3A MDEFWP is powered from 3TD B. Incorrect: 3C HPIP is powered from 3TD.

C. CORRECT: 3B MDEFWP is powered from 3TE and would therefore NOT be available D. Incorrect: 3B LPSWP is powered from 3TC Technical Reference(s): PNS-HPI, PNS-LPI, CF-EF, PNS-RBC, SSS-LPW Proposed references to be provided to applicants during examination: NONE Learning Objective: EL-EPD R27 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 65 T2/G2 - cpw 086K5.03, Fire Protection System (FPS)

Knowledge of the operational implication of the following concepts as they apply to the Fire Protection System:

Effect of water spray on electrical components.

(3.1/3.4)

KIA MATCH ANALYSIS which requires a

or equipment effected.

Requires knowledge of operational implications of de-energizing 3TE to allow water spray to extinguish fire.

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: 3A MDEFWP is powered from 3TD B. Incorrect: 3C HPIP is powered from 3TD.

C. CORRECT: 38 MDEFWP is powered from 3TE and would therefore NOT be available D. Incorrect: 3B LPSWP is powered from 3TC Technical Reference(s): PNS-HPI, PNS-LPI, CF-EF, PNS-R8C, SSS-LPW Proposed references to be provided to applicants during examination: NONE Learning Objective: EL-EPD R27 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 66 Unit 2 plant conditions:

ReactorinMODE6 RCS Boron = 2270 ppmb Based on the above conditions, which ONE of the following describes whether RCS Boron concentration meets the requirements of OP/21A11 502/007 (Operations Defueling/Refueling Responsibilities) for core alterations AND what is the minimum number of OPERABLE Source Range NIs required by the same procedure?

A. meets I 2 B. meets I 1

C. does NOT meet I 2 D. does NOT meet I 1

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 66 Unit 2 plant conditions:

Reactor in MODE 6 RCS Boron = 2270 ppmb Based on the above conditions, which ONE of the following describes whether RCS Boron concentration meets the requirements of OP/2/A/1502/007 (Operations Defueling/Refueling Responsibilities) for core alterations AND what is the minimum number of OPERABLE Source Range NI's required by the same procedure?

A. meets / 2 B. meets / 1 C. does NOT meet / 2 D. does NOT meet / 1

2010A NRC REACTOR OPERATOR EXAM Question 66 T3-cpw 2.1.36, Conduct of Operations Knowledge of procedures and limitations involved in core alterations.

(3.0/4.1)

K/A MATCH ANALYSIS Requires knowledge of procedure limits and precautions and/or Tech Spec requirements for fuel movement.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: First part is correct since 2250

- 2950 ppm boron is required per L&P of OP/2/AI1 5021007 (Operations Defuel ing/Refuel i ng Responsibilities).

Second part is correct as both the procedure and TS require 2 Source Range NIs when moving a fuel assembly.

B. Incorrect: First part is correct. Second part is plausible since conditions other than when moving fuel or adding positive reactivity, a single Source Range NI is all that is required.

C. Incorrect: First part is plausible since 2950 is the upper end of band and it could easily be confused as the minimum boron allowed since it is above the TS required 2220 ppmb. Second part is correct.

D. Incorrect: : First part is plausible since 2950 is the upper end of band and it could easily be confused as the minimum boron allowed since it is above the TS required 2220 ppmb. Second part is plausible since conditions other than when moving fuel or adding positive reactivity, a single Source Range NI is all that is required.

Technical Reference(s): FH-FHS, OP/2/A/1502/007, TECH SPECS Proposed references to be provided to applicants during examination: NONE Learning Objective: FH-FHS R5, R20 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 66 T3 - cpw 2.1.36, Conduct of Operations Knowledge of procedures and limitations involved in core alterations.

(3.0/4.1 )

KIA MATCH ANALYSIS Requires knowledge of procedure limits and precautions and/or Tech Spec requirements for fuel movement.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: First part is correct since 2250 - 2950 ppm boron is required per L&P of OP/2/Al1502/007 (Operations Defueling/Refueling Responsibilities).

Second part is correct as both the procedure and TS require 2 Source Range NI's when moving a fuel assembly.

B. Incorrect: First part is correct. Second part is plausible since conditions other than when moving fuel or adding positive reactivity, a single Source Range NI is all that is required.

C. Incorrect: First part is plausible since 2950 is the upper end of band and it could easily be confused as the minimum boron allowed since it is above the TS required 2220 ppmb. Second part is correct.

D. Incorrect:: First part is plausible since 2950 is the upper end of band and it could easily be confused as the minimum boron allowed since it is above the TS required 2220 ppmb. Second part is plausible since conditions other than when moving fuel or adding positive reactivity, a single Source Range NI is all that is required.

Technical Reference(s): FH-FHS, OP/2/Al1502/007, TECH SPECS Proposed references to be provided to applicants during examination: NONE Learning Objective: FH-FHS R5, R20 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 67 Unit 1 initial conditions:

Reactor power 100%

LDST level = 75 stable Group 7 rod position = 94% withdrawn Makeup to LDST initiated from lB BHUT Neutron error = 0 stable Current conditions:

IHP-15 Bailey controller indicates 470 gallons added to LDST I B Bleed Transfer Pump secured Based on the above conditions, which ONE of the following would describe a diverse indication that 470 gallons of 1 B BHUT had been added to the LDST?

LDST level is approximately inches and neutron error will become A. 90 / positive B. 90 I negative C. 95 / positive D. 95 I negative 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 67 Unit 1 initial conditions:

Reactor power = 100%

LOST level = 75" stable Group 7 rod position = 94% withdrawn Makeup to LOST initiated from 1 B BHUT Neutron error = 0 stable Current conditions:

1 HP-15 Bailey controller indicates 470 gallons added to LOST 1 B Bleed Transfer Pump secured Based on the above conditions, which ONE of the following would describe a diverse indication that 470 gallons of 1 B BHUT had been added to the LOST?

LOST level is approximately ____ inches and neutron error will become ___ _

A. 90 / positive B. 90 / negative C. 95 / positive O. 95 / negative

2010A NRC REACTOR OPERATOR EXAM Question 67 T3-cpw 2.1.45, Conduct of Operations Ability to identify and interpret diverse indications to validate the response of another indication.

(4.3/4.3)

K/A MATCH ANALYSIS Requires analyzing various diverse indications to validate a SG tube leak exists.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect. First part is correct. Second part is plausible since it would be correct if the addition was from A BHUT OR if the student made the common error of incorrectly determining the direction of rod motion required to offset the boron addition.

B. CORRECT: LDST is 31.3 gal/inch. If 470 gallons of water had been added then level should have increased 15 inches which would put level at 90 inches. If B bleed had been added then Boron concentration would be decreasing which means neutron error would be building negative towards a rod push to offset the boron addition.

C. Incorrect: First part is plausible because this would be correct if you use 24 gal/inch for LDST volume however this is the Pressurizer value. Second part is plausible since it would be correct if the addition was from A BHUT OR if the student made the common error of incorrectly determining the direction of rod motion required to offset the boron addition.

D. Incorrect: First part is plausible because this would be correct if you use 24 gal/inch for LDST volume however this is the Pressurizer value. Second part is correct.

Technical Reference(s): PNS-PZR, PNS-HPI, CP-016 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R1,2,3 CP-016 R5, Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis Question 67 T3 - cpw 2010A NRC REACTOR OPERATOR EXAM 2.1.45, Conduct of Operations Ability to identify and interpret diverse indications to validate the response of another indication.

(4.3/4.3)

K/A MATCH ANALYSIS Requires analyzing various diverse indications to validate a SG tube leak exists.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect. First part is correct. Second part is plausible since it would be correct if the addition was from A SHUT OR if the student made the common error of incorrectly determining the direction of rod motion required to offset the boron addition.

B. CORRECT: LOST is 31.3 gal/inch. If 470 gallons of water had been added then level should have increased 15 inches which would put level at 90 inches. If B bleed had been added then Boron concentration would be decreasing which means neutron error would be building negative towards a rod push to offset the boron addition.

C. Incorrect: First part is plausible because this would be correct if you use 24 gal/inch for LOST volume however this is the Pressurizer value. Second part is plausible since it would be correct if the addition was from A SHUT OR if the student made the common error of incorrectly determining the direction of rod motion required to offset the boron addition.

O. Incorrect: First part is plausible because this would be correct if you use 24 gal/inch for LOST volume however this is the Pressurizer value. Second part is correct.

Technical Reference(s): PNS-PZR, PNS-HPI, CP-016 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-PZR R1,2,3 CP-016 R5, Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT Question 68 Which ONE following describes the RCS Pressure Safety Limit (psig) and what is credited with ensuring the limit is NOT exceeded?

A. 2500 I RPS trip settings.

B. 2500 I Pressurizer Spray Valve.

C. 2750 I RPS trip settings.

D. 2750 I Pressurizer Spray Valve.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 68 Which ONE following describes the RCS Pressure Safety Limit (psig) and what is credited with ensuring the limit is NOT exceeded?

A. 2500 / RPS trip settings.

B. 2500 / Pressurizer Spray Valve.

C. 2750 / RPS trip settings.

D. 2750 / Pressurizer Spray Valve.

2010A NRC REACTOR OPERATOR EXAM Question 68 T3-CPW 2.2.25, Equipment Control Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(3.2/4.2 K/A MATCH ANALYSIS Requires knowledge of the LCO for RCS Pressure safety limit and its bases.

ANSWER CHOICE ANALYSIS Answer:

D A. Incorrect: First part is plausible since 2500 psig is the RCS design pressure value.

Second part is correct.

B. Incorrect: First part is plausible since 2500 psig is the RCS design pressure value.

Second part is plausible since the pressurizer spray valve is a method of limiting RCS pressure spikes however the bases of the Safety Analysis specifically says that no credit is take for operation of the spray valve.

C. CORRECT: 2750 is the RCS Pressure Safety Limit. Safety Limit bases specifies that RPS trip setpoints are credited for ensuring safety limits are not exceeded.

D. Incorrect: First part is correct. Second part is plausible since the pressurizer spray valve is a method of limiting RCS pressure spikes however the bases of the Safety Analysis specifically says that no credit is take for operation of the spray valve.

Technical Reference(s): TS 2.1.2 (RCS Pressure Safety Limit) including bases Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-TSS R8 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 68 T3-CPW 2010A NRC REACTOR OPERATOR EXAM 2.2.25, Equipment Control Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(3.2/4.2 KIA MATCH ANALYSIS Requires knowledge of the LCO for RCS Pressure safety limit and its bases.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: First part is plausible since 2500 psig is the RCS design pressure value.

Second part is correct.

B. Incorrect: First part is plausible since 2500 psig is the RCS design pressure value.

Second part is plausible since the pressurizer spray valve is a method of limiting RCS pressure spikes however the bases of the Safety Analysis specifically says that no credit is take for operation of the spray valve.

C. CORRECT: 2750 is the RCS Pressure Safety Limit. Safety Limit bases specifies that RPS trip setpoints are credited for ensuring safety limits are not exceeded.

D. Incorrect: First part is correct. Second part is plausible since the pressurizer spray valve is a method of limiting RCS pressure spikes however the bases of the Safety Analysis specifically says that no credit is take for operation of the spray valve.

Technical Reference(s): TS 2.1.2 (RCS Pressure Safety Limit) including bases Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-TSS R8 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 69 Unit 1 plant conditions:

Reactor power = 100%

IA Core Flood Tank parameters:

o Pressure = 572 psig stable o

Level

12. 91 feet o

Boron Concentration = 2010 ppmb I B Core Flood Tank parameters:

o Pressure = 590 psig stable o

Level = 12.51 feet o

Boron Concentration = 1895 ppmb Based on the above condition, which ONE of the following describes the action(s) required (if any) in accordance with Tech Spec 3.5.1 (Core Flood Tanks)?

A. NO actions required B. Enter LCO 3.0.3 immediately C. Restore IA CFT to OPERABLE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ONLY D. Restore lB CFT to OPERABLE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ONLY 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 69 Unit 1 plant conditions:

Reactor power = 100%

1A Core Flood Tank parameters:

o Pressure = 572 psig stable o

Level = 12. 91 feet o

Boron Concentration = 2010 ppmb 1 B Core Flood Tank parameters:

o Pressure = 590 psig stable o

Level = 12.51 feet o Boron Concentration = 1895 ppmb Based on the above condition, which ONE of the following describes the action(s) required (if any) in accordance with Tech Spec 3.5.1 (Core Flood Tanks)?

A. NO actions required B. Enter LCO 3.0.3 immediately C. Restore 1A CFT to OPERABLE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ONLY D. Restore 1 B CFT to OPERABLE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ONLY

2010A NRC REACTOR OPERATOR EXAM Question 69 T3-CPW 2.2.39, Equipment Control Knowledge of less than or equal to one hour Technical Specification action statements for systems.

(3.9/4.5)

KIA MATCH ANALYSIS Question requires recalling from memory the 1 hr or less requirements of TS 3.5.1 (Core Flood Tanks).

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible since even the parameters that are out of spec on the CFTs are reasonable close to the same parameter on the other CFT. This makes it difficult to identify an out of tolerance parameter just by looking at the deltas and therefore requires specific knowledge of the allowable values to be able to determine if either CFT is out of spec.

B. CORRECT: TS 3.5.1 Condition D is for two CFTs inoperable and requires immediate entry into LCO 3.0.3. CFT pressure is required to be> 575 psig.

Since the A CFT pressure is 572, this SR is not met. CFT level is given in the spec as a value of cubic feet however the specific instrument surveillance performed in PTI600IOI identifies 12.56 to 13.44 feet as the acceptable level range. B CFT level does not meet that requirement. With two CFTs not meeting the SRs, Condition 0 is warranted which requires LCO 3.0.3.

C. Incorrect: Plausible since the A CFT is inoperable due to low CFT pressure. This would be a correct answer under the misconception that the B CFT level is acceptable.

D. Incorrect: Plausible since the B CFT is inoperable due to low CFT level. This would be a correct answer under the misconception that the A CFT level is acceptable.

Technical Reference(s): Technical Specification 3.5.1 (Core Flood Tanks)

Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-TSS R4 Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals Question 69 T3 -CPW 2010A NRC REACTOR OPERATOR EXAM 2.2.39, Equipment Control Knowledge of less than or equal to one hour Technical Specification action statements for systems.

(3.9/4.5)

KIA MATCH ANALYSIS Question requires recalling from memory the 1 hr or less requirements of TS 3.5.1 (Core Flood Tanks).

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: Plausible since even the parameters that are out of spec on the CFT's are reasonable close to the same parameter on the other CFT. This makes it difficult to identify an out of tolerance parameter just by looking at the deltas and therefore requires specific knowledge of the allowable values to be able to determine if either CFT is out of spec.

B. CORRECT: TS 3.5.1 Condition D is for two CFT's inoperable and requires immediate entry into LCO 3.0.3. CFT pressure is required to be ~ 575 psig.

Since the A CFT pressure is 572, this SR is not met. CFT level is given in the spec as a value of cubic feet however the specific instrument surveillance performed in PTl600/01 identifies 12.56 to 13.44 feet as the acceptable level range. B CFT level does not meet that requirement. With two CFT's not meeting the SR's, Condition D is warranted which requires LCO 3.0.3.

C. Incorrect: Plausible since the A CFT is inoperable due to low CFT pressure. This would be a correct answer under the misconception that the B CFT level is acceptable.

D. Incorrect: Plausible since the B CFT is inoperable due to low CFT level. This would be a correct answer under the misconception that the A CFT level is acceptable.

Technical Reference(s): Technical Specification 3.5.1 (Core Flood Tanks)

Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-TSS R4 Question Source: New Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 70 Unit 1 plant conditions:

Reactor power = 100%

Based on the above condition, which ONE of the following describes a condition that would require entry into a Tech Spec ACTIONS table?

A.

UST level

= 5.6 feet B.

BWST level = 47.3 feet C.

ID RPS channel in Manual bypass D.

230KV Dacus Black and White lines isolated 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 70 Unit 1 plant conditions:

Reactor power = 100%

Based on the above condition, which ONE of the following describes a condition that would require entry into a Tech Spec ACTIONS table?

A.

UST level = 5.6 feet B.

BWST level = 47.3 feet C.

1 D RPS channel in Manual bypass D.

230KV Dacus Black and White lines isolated

2010A NRC REACTOR OPERATOR EXAM Question 70 T3-CPW 2.2.42, Equipment Control Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

(3.9/4.6)

KIA MATCH ANALYSIS Requires analyzing several conditions and parameters and determining if they result in TS entry conditions being met.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: TS 3.7.6 requires that both the UST and Hotwell be operable and that the UST contain > 30,000 gallons.

. PT/600/01 (Periodic Instrument Surveillance) verifies this volume by requiring UST level be > 6 feet.

B. Incorrect: Plausible since Tech Specs requires the BWST to be operable and contain 350,000 gallons of Borated water. PT/600/01 (Periodic Instrument Surveillance) verifies this volume by requiring >47 feet in BWST.

C. Incorrect: Plausible since Tech Specs do require that RPS be operable, however there are 4 channels for each required function and only 3 channels are required therefore having one of the RPS channels in Manual Bypass does not result in required functions being inoperable as long as no other RPS inoperabilitys exist.

D. Incorrect: Plausible since either Dacus black or white are part of what can be credited in TS 3,8,1 for one of the two offsite sources on separate towers however since there are still more than enough offsite sources available that meet the separate tower criteria, these being out of service would not require entry into the TS ACTION table for TS 3.8.1.

Technical Reference(s): TS 3.1.6, 3.5.4, 3.3.10, and 3.3.1 PTI600/01 (Periodic Instrument Surveillance)

Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-TSS RB Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level: Knowledge and Fundamentals Question 70 T3 - CPW 2010A NRC REACTOR OPERATOR EXAM 2.2.42, Equipment Control Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

(3.9/4.6)

KIA MATCH ANALYSIS Requires analyzing several conditions and parameters and determining if they result in TS entry conditions being met.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: TS 3.7.6 requires that both the UST and Hotwell be operable and that the UST contain> 30,000 gallons.. PT/600/01 (Periodic Instrument Surveillance) verifies this volume by requiring UST level be> 6 feet.

B. Incorrect: Plausible since Tech Specs requires the BWST to be operable and contain 350,000 gallons of Borated water. PT/600101 (Periodic Instrument Surveillance) verifies this volume by requiring >47 feet in BWST.

C. Incorrect: Plausible since Tech Specs do require that RPS be operable, however there are 4 channels for each required function and only 3 channels are required therefore having one of the RPS channels in Manual Bypass does not result in required functions being inoperable as long as no other RPS inoperability's exist.

D. Incorrect: Plausible since either Dacus black or white are part of what can be credited in TS 3,8,1 for one of the two offsite sources on separate towers however since there are still more than enough offsite sources available that meet the separate tower criteria, these being out of service would not require entry into the TS ACTION table for TS 3.8.1.

Technical Reference(s): TS 3.7/.6, 3.5.4, 3.3.10, and 3.3.1 PTl600/01 (Periodic Instrument Surveillance)

Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-TSS R8 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 71 Unit 1 plant conditions:

Reactor in MODE 5 Reactor Building Main Purge in operation Based on the above conditions, which ONE of the following will cause the RB Purge fan to trip?

A. Inadvertent actuation of ES Channel 5 B.

I RIA-45 reaches ALERT setpoint C. Suction piping pressure = 5 inches of water vacuum D. I PR-3 (RB PURGE CONTROL) closed 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 71 Unit 1 plant conditions:

Reactor in MODE 5 Reactor Building Main Purge in operation Based on the above conditions, which ONE of the following will cause the RB Purge fan to trip?

A. Inadvertent actuation of ES Channel 5 B. 1 RIA-45 reaches ALERT setpoint C. Suction piping pressure = 5 inches of water vacuum D. 1 PR-3 (RB PURGE CONTROL) closed

2010A NRC REACTOR OPERATOR EXAM Question 71 T3-cpw 2.3.1 1, Radiation Control Ability to control radiation releases.

(3.8/4.3)

K/A MATCH ANALYSIS Question demonstrates ability to control radiation releases by demonstrating the ability to determine conditions that would result in terminating the release by tripping the main purge fan.

ANSWER CHOICE ANALYSIS Answer: 0 A. Incorrect: Plausible since there is ES actuations that will cause the purge fan to trip however it is ES Channels I &/or 2 therefore this would be a correct choice if asking about ES-i or ES-2 actuation.

B. Incorrect: Plausible since RIA-45 will trip the main purge fan however it takes a HIGH alarm to do so therefore this would be the correct choice if it were asking about a HIGH alarm instead of an ALERT alarm.

C. Incorrect: Plausible since vacuum in the suction piping will trip the main purge fan however the setpoint is 9 inches water.

D. CORRECT: The main purge fan will trip if PR-3 is closed while the fan is

running, Technical Reference(s): PNS-RBP Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-RBP R4 Question Source: Modified Bank Question History: Last NRC Exam 2004 NRC Exam Q #56 Question Cognitive Level:

Knowledge and Fundamentals Question 71 T3 - cpw 2010A NRC REACTOR OPERATOR EXAM 2.3.11, Radiation Control Ability to control radiation releases.

(3.8/4.3)

KIA MATCH ANALYSIS Question demonstrates ability to control radiation releases by demonstrating the ability to determine conditions that would result in terminating the release by tripping the main purge fan.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since there is ES actuations that will cause the purge fan to trip however it is ES Channels 1 &/or 2 therefore this would be a correct choice if asking about ES-1 or ES-2 actuation.

B. Incorrect: Plausible since RIA-45 will trip the main purge fan however it takes a HIGH alarm to do so therefore this would be the correct choice if it were asking about a HIGH alarm instead of an ALERT alarm.

C. Incorrect: Plausible since vacuum in the suction piping will trip the main purge fan however the setpoint is 9 inches water.

D. CORRECT: The main purge fan will trip if PR-3 is closed while the fan is

running, Technical Reference(s): PNS-RBP Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-RBP R4 Question Source: Modified Bank Question History: Last NRC Exam 2004 NRC Exam Q #56 Question Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 72 Unit 1 plant conditions:

Reactor in MODE 6 LPI DHR in progress using 1A LPI pump Fuel movement in progress Nl-3 and 4 out of service RB Purge in progress Based on the above conditions, which ONE of the following describes when Reactor Building evacuation is required?

A. 1NI-2fails low B. 1RIA-4 HIGH alarm C. 1RIA-45 HIGH alarm D. Loss of power to IA LPI pump 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 72 Unit 1 plant conditions:

Reactor in MODE 6 LPI DHR in progress using 1 A LPI pump Fuel movement in progress NI-3 and 4 out of service RB Purge in progress Based on the above conditions, which ONE of the following describes when Reactor Building evacuation is required?

A. 1 N 1-2 fails low B. 1 RIA-4 HIGH alarm C. 1 RIA-45 HIGH alarm D. Loss of power to 1A LPI pump

2010A NRC REACTOR OPERATOR EXAM Question 72 T3-cpw 2.3.12, Radiation Control Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(3.2/3.7)

K/A MATCH ANALYSIS Requires knowledge of licensed operator duties required to reduce the possibility of excessive dose being received by individuals inside containment during a boron dilution event.

ANSWER CHOICE ANALYSIS Answer:

D A. Incorrect: Plausible since there is fuel movement in progress and there are required actions that must be performed if there is a failure of one of the two required operable SR NIs during core alterations. This failure would require immediate suspension of core alterations.

B. CORRECT: RIA-4 HIGH alarm actuates with hi radiation area in the RB. The RB evacuation alarm will automatically sound when RIA-4 reaches its HIGH alarm point.

C. Incorrect: Plausible since 1 RIA-45 monitors the RB and does have automatic actions as a result of high alarms. If the alarm is received, the RB Purge will trip off however a RB evacuation is not required.

D. Incorrect: Plausible since a loss of the IA LPI pump would constitute a loss of DHR and would require entry into AP/26. AP/26 does direct making a page however it is not under these conditions. Entering AP/26 due to a loss of inventory would require a plant page therefore the misconception that this path through the AP would also require the page is reasonable.

Technical Reference(s): API3 (Boron Dilution), AP126 (Loss of DHR)

Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-APG R8 Question Source: NEW Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 72 T3 - cpw 2.3.12, Radiation Control Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

(3.2/3.7)

KIA MATCH ANALYSIS Requires knowledge of licensed operator duties required to reduce the possibility of excessive dose being received by individuals inside containment during a boron dilution event.

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since there is fuel movement in progress and there are required actions that must be performed if there is a failure of one of the two required operable SR NI's during core alterations. This failure would require immediate suspension of core alterations.

B. CORRECT: RIA-4 HIGH alarm actuates with hi radiation area in the RB. The RB evacuation alarm will automatically sound when RIA-4 reaches its HIGH alarm point.

C. Incorrect: Plausible since 1 RIA-45 monitors the RB and does have automatic actions as a result of high alarms. If the alarm is received, the RB Purge will trip off however a RB evacuation is not required.

D. Incorrect: Plausible since a loss of the 1A LPI pump would constitute a loss of DHR and would require entry into AP/26. AP/26 does direct making a page however it is not under these conditions. Entering AP/26 due to a loss of inventory would require a plant page therefore the misconception that this path through the AP would also require the page is reasonable.

Technical Reference(s): AP/3 (Boron Dilution), AP/26 (Loss of DHR)

Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-APG R8 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 73 For Operations personnel, which ONE of the following describes the required response to an Electronic Dosimeter dose alarm and when it is acceptable to deviate from that requirement?

A. Exit the area immediately and contact RP I with RP permission B. Exit the area immediately and contact RP I with emergency dose limits in effect C. Move away from the area until alarm clears I with RP permission D. Move away from the area until alarm clears I with emergency dose limits in effect 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 73 For Operations personnel, which ONE of the following describes the required response to an Electronic Dosimeter dose alarm and when it is acceptable to deviate from that requirement?

A. Exit the area immediately and contact RP / with RP permission B. Exit the area immediately and contact RP / with emergency dose limits in effect C. Move away from the area until alarm clears / with RP permission D. Move away from the area until alarm clears / with emergency dose limits in effect

2010A NRC REACTOR OPERATOR EXAM Question 73 T3-cpw 2.3.7, Radiation Control Ability to comply with radiation work permit requirements during normal or abnormal conditions.

(3.5/3.6)

K/A MATCH ANALYSIS Requires knowledge of how to respond to Dose and Dose Rate alarms determined by RWPs in both normal and abnormal conditions. Additionally requires knowledge of when it is acceptable under abnormal conditions to deviate from the RWP requirements ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since this would be the correct response to a making a temporary change to the RWP requirement but RP cannot authorize personnel to continue work with a continuous dose alarm. RAD RPP page 59 B. CORRECT: Per RAD-RPP page 59, if your dose alarm activates, exit the area and contact RP. Per OMP 1-18 when EDLs are implemented NEOs and others working under EDLs may continue to work through ED alarms.

C. Incorrect: First part is plausible since this is the correct response if a dose rate alarm occurs. Second part is plausible since this would be the correct response to a making a temporary change to the RWP requirement but RP cannot authorize personnel to continue work with a continuous dose alarm.

D. Incorrect: First part is plausible since this is the correct response if a dose rate alarm occurs. Second part is correct Technical Reference(s): RAD-RPP, OMP 1-18 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-TCA R6, RAD-RPP R9 Question Source:

Modified Bank (RAD020902)

Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 73 T3 - cpw 2.3.7, Radiation Control Ability to comply with radiation work permit requirements during normal or abnormal conditions.

(3.5/3.6)

KIA MATCH ANALYSIS Requires knowledge of how to respond to Dose and Dose Rate alarms determined by RWP's in both normal and abnormal conditions. Additionally requires knowledge of when it is acceptable under abnormal conditions to deviate from the RWP requirements ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since this would be the correct response to a making a temporary change to the RWP requirement but RP cannot authorize personnel to continue work with a continuous dose alarm. RAD-RPP page 59 B. CORRECT: Per RAD-RPP page 59, if your dose alarm activates, exit the area and contact RP. Per OMP 1-18 when EDL's are implemented NEO's and others working under EDL's may continue to work through ED alarms.

C. Incorrect: First part is plausible since this is the correct response if a dose rate alarm occurs. Second part is plausible since this would be the correct response to a making a temporary change to the RWP requirement but RP cannot authorize personnel to continue work with a continuous dose alarm.

D. Incorrect: First part is plausible since this is the correct response if a dose rate alarm occurs. Second part is correct Technical Reference(s): RAD-RPP, OMP 1-18 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-TCA R6, RAD-RPP R9 Question Source: Modified Bank (RAD020902)

Question History: Last NRC Exam NIA Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 74 Unit 1 plant conditions:

Blackout tab in progress EOP step gives direction to Initiate AP/1 1 (Recovery From Loss of Power)

Based on the above conditions, which ONE of the following describes the actions required to perform the EOP step?

A. RO will perform AP/1 1 and SRO will wait until AP/1 1 has been completed to continue in the EOP.

B. SRO will direct steps in AP/1 I and then return to the EOP once AP/1 1 is complete.

C. RO will take steps to begin AP/1 1 before SRO continues in EOP but once AP/1 I is begun the SRO can re-direct the ROs activities.

D. SRO can continue in EOP once AP/1 1 is being performed. SRO CANNOT re-direct the ROs activities until AP/1 I is completed.

2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 74 Unit 1 plant conditions:

Blackout tab in progress EOP step gives direction to "Initiate AP/11 (Recovery From Loss of Power)"

Based on the above conditions, which ONE of the following describes the actions required to perform the EOP step?

A. RO will perform AP/11 and SRO will wait until AP/11 has been completed to continue in the EOP.

B. SRO will direct steps in AP/11 and then return to the EOP once AP/11 is complete.

C. RO will take steps to begin AP/11 before SRO continues in EOP but once AP/11 is begun the SRO can re-direct the RO's activities.

D. SRO can continue in EOP once AP/11 is being performed. SRO CANNOT re-direct the RO's activities until AP/11 is completed.

2010A NRC REACTOR OPERATOR EXAM Question 74 T3-cpw 2.4.16 Emergency Procedures / Plan Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

(3.5/4.4)

KIA MATCH ANALYSIS Requires Knowledge of EOP/AP implementation hierarchy and process for coordinating AP with EOP based on criteria in step directing AP performance ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since this would be correct if the step gave direction to Perform AP/1 1. There are numerous steps in the EOP that are Perform steps and OMP 18 explains that the referenced procedure and/or steps must be performed prior to continuing in the current procedure in use. Additionally there are numerous cases of the EOP director waiting on completion of other procedures (End. 5.38).

B. Incorrect: Plausible since this would be correct if the step were an If At Any Time (IAAT) step. EOP has numerous cases were the IAAT process is used. This requires going and performing the steps directed by the IAAT and then returning to the original step.

C. CORRECT: Initiate steps require actions to begin the referenced steps or procedure and then continuing with the current procedure in use.

D. INCORRECT: Plausible since this would result in both AP/1 1 and the EOP being performed however OMP 1-18 only requires the AP be initiated and then it allows the SRO to determine the best use of available manpower. Additionally plausible since this description would be correct if talking about the RO being in a Rule.

Technical Reference(s): OMP 1-18 Proposed references to be provided to applicants during examination:

NONE Learning Objective: ADM-OMP RIO, R52 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals 2010A NRC REACTOR OPERATOR EXAM Question 74 T3 - cpw 2.4.16 Emergency Procedures 1 Plan Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

(3.5/4.4)

KIA MATCH ANALYSIS Requires Knowledge of EOP/AP implementation hierarchy and process for coordinating AP with EOP based on criteria in step directing AP performance ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since this would be correct if the step gave direction to "Perform AP/11". There are numerous steps in the EOP that are "Perform" steps and OMP 18 explains that the referenced procedure and/or steps must be performed prior to continuing in the current procedure in use. Additionally there are numerous cases of the EOP director waiting on completion of other procedures (Encl. 5.38).

B. Incorrect: Plausible since this would be correct if the step were an If At Any Time (1M T) step. EOP has numerous cases were the 1M T process is used. This requires going and performing the steps directed by the 1M T and then returning to the original step.

C. CORRECT: "Initiate" steps require actions to begin the referenced steps or procedure and then continuing with the current procedure in use.

D. INCORRECT: Plausible since this would result in both AP/11 and the EOP being performed however OMP 1-18 only requires the AP be initiated and then it allows the SRO to determine the best use of available manpower. Additionally plausible since this description would be correct if talking about the RO being in a Rule.

Technical Reference(s): OMP 1-18 Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-OMP R10, R52 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level:

Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT Question 75 Unit 1 plant conditions:

Reactor power = 100%

1SA3IB6 (FIRE ALARM) actuated Fire Alarm panel indication o

point 0202071 (Unit 1 pipe trench room 348 north end) actuated Based on the above conditions, which ONE of the following describes the initial action required by the Alarm Response Guide AND a method used in RP/1000/029 (Fire Brigade Response) to dispatch the fire brigade when it is required?

A. Dispatch a Fire Brigade qualified operator to determine validity of the alarm / Plant Paging system B. Dispatch a Fire Brigade qualified operator to determine validity of the alarm / Have Security dispatch fire brigade C. Dispatch the Fire Brigade / Plant Paging system D. Dispatch the Fire Brigade / Have Security dispatch fire brigade 2010A NRC REACTOR OPERATOR EXAM 1 POINT Question 75 Unit 1 plant conditions:

Reactor power = 100%

1 SA3/B6 (FIRE ALARM) actuated Fire Alarm panel indication o

point 0202071 (Unit 1 pipe trench room 348 north end) actuated Based on the above conditions, which ONE of the following describes the initial action required by the Alarm Response Guide AND a method used in RP/10001029 (Fire Brigade Response) to dispatch the fire brigade when it is required?

A. Dispatch a Fire Brigade qualified operator to determine validity of the alarm 1 Plant Paging system B. Dispatch a Fire Brigade qualified operator to determine validity of the alarm 1 Have Security dispatch fire brigade C. Dispatch the Fire Brigade 1 Plant Paging system D. Dispatch the Fire Brigade 1 Have Security dispatch fire brigade

2010A NRC REACTOR OPERATOR EXAM Question 75 T3 -cpw 2.4.26 Emergency Procedures / Plan Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage (3.1/3.6)

K/A MATCH ANALYSIS Per NRC, ask about activating fire brigade from Control Room OK.

Requires knowledge of actions directed by a Fire Alarm in the control room as well as knowledge of how Fire Brigade is dispatched.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: Manual actions of ARG for Fire Alarm statalarm direct dispatching a fire brigade qualified operator to assess validity of the alarm. It does NOT direct dispatching the Fire Brigade until the alarm is determined to be valid. is used to dispatch the fire brigade and the initial response is to use the plant page. This is significant since all fire brigade members do not have radios and pagers therefore the plant page is used to ensure all members are notified.

B. Incorrect: First part is correct. Second part is plausible since would be correct if asking about dispatching MERT to a medical emergency per RP/1000/016 (MERT activation...)

C. Incorrect: First part is plausible since it would be reasonable to assume that dispatching the fire brigade would be directed if a Fire Alarm is received however the ARG does not direct dispatching Fire Brigade until the alarm is determined to be valid. Second part is correct.

D. Incorrect: First part is plausible since it would be reasonable to assume that dispatching the fire brigade would be directed if a Fire Alarm is received however the ARG does not direct dispatching Fire Brigade until the alarm is determined to be valid. Second part is plausible since would be correct if asking about dispatching MERT to a medical emergency per RP/1000/016 (MERT activation...)

Technical Reference(s): RPIOIBII000IO29 (Fire brigade response) ARG for ISA3/B6, IC-FDS, RP/10001016 (MERT activation...)

Proposed references to be provided to applicants during examination: NONE Learning Objective:

lC-FDS R6 Question Source:

NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis 2010A NRC REACTOR OPERATOR EXAM Question 75 T3 -cpw 2.4.26 Emergency Procedures I Plan Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage (3.1/3.6)

KIA MATCH ANALYSIS from Control Room Requires knowledge of actions directed by a Fire Alarm in the control room as well as knowledge of how Fire Brigade is dispatched.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: Manual actions of ARG for Fire Alarm statalarm direct dispatching a fire brigade qualified operator to assess validity of the alarm. It does NOT direct dispatching the Fire Brigade until the alarm is determined to be valid. is used to dispatch the fire brigade and the initial response is to use the plant page. This is significant since all fire brigade members do not have radios and pagers therefore the plant page is used to ensure all members are notified.

B. Incorrect: First part is correct. Second part is plausible since would be correct if asking about dispatching MERT to a medical emergency per RP/1 0001016 (MERT activation... )

C. Incorrect: First part is plausible since it would be reasonable to assume that dispatching the fire brigade would be directed if a Fire Alarm is received however the ARG does not direct dispatching Fire Brigade until the alarm is determined to be valid. Second part is correct.

D. Incorrect: First part is plausible since it would be reasonable to assume that dispatching the fire brigade would be directed if a Fire Alarm is received however the ARG does not direct dispatching Fire Brigade until the alarm is determined to be valid. Second part is plausible since would be correct if asking about dispatching MERT to a medical emergency per RP/10001016 (MERT activation... )

Technical Reference(s): RP/O/B/1000/029 (Fire brigade response) ARG for 1SA3/B6, IC-FOS, RP/10001016 (MERT activation... )

Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-FOS R6 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension and Analysis