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 1

Question 1 I Unit 11 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 conditions, which ONE of the following describes the reason for the Based on the above conditions, current conditions?

A. 11 B2 RCP Upper, Middle, and Lower seals have failed 1 RC-159 B. 1 RC-1 59 and 1I RC-160 RC-1 60 (RXV Head vents) are leaking C. Low range RCS pressure has failed HIGH 1 RC-66 (PORV) is leaking to the Quench Tank D. 1

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 1I Question TIIGI --cpw T1/G1 cpw OO8AK1 008AK1.01, Pressurizer Vapor

.01, Pressurizer Vapor Space Space Accident Accident Knowledge of Knowledge of the the operational operational implications implications of of the the following following concepts concepts as as they they apply to apply Pressurizer Vapor to aa Pressurizer Vapor Space Space Accident:

Accident:

Thermodynamics and flow characteristics Thermodynamics characteristics of of open open or leaking leaking valves.

(3.2/3.7)

KIA MATCH ANALYSIS K/A MATCH ANALYSIS Requires knowledge Requires knowledge of of the the throttling throttling process process andand the the operational operational implications implications of indications regarding PORV indications PORV tailpipe temperatures ANSWER CHOICE ANALYSIS Answer: 0D 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 1 RC-1 59 and 1RC-160 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, 1IRC-66 RC-66 (PORV) failing open would cause these indications.

Reference(s): PNS-PZR Page 34 - 35 Technical Reference(s): -

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: Comprehens Comprehension ion or Analysis

2010A NRC REACTOR OPERATOR EXAM 1I POINT Question 2 Unit 11 plant conditions:

• Reactor Trip

• SBLOCA has occurred

• Rule 2 (Loss of SCM) in progress

• 1A & & 1lB 78 XSUR increasing B Steam Generator levels = 78" 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 22 Question TIIGI --cpw T1/G1 cpw OO9EK1 009EK1.01 Small Break

.01 Small Break LOCA LOCA Knowledge of Knowledge the operational of the operational implications implications ofof the the following following concepts concepts as as they they apply to the small break apply to the small break LOCA: LOCA:

Natural circulation Natural circulation and and cooling, including reflux cooling, including reflux boiling.

boiling.

(4.2/4.7)

(4.2/4.7)

K/A MATCH KIA MATCH ANALYSIS Requires knowledge Requires knowledge of of operational operational requirement requirement to establish reflux (boiler condenser at ONS) condenser 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.

D. CORRECT: Per EAP-LOSM page 22 & & 23 - SG heat removal must be established by feeding SGs up to levels that will promote NIC N/C and BCM 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 00 °F. of.

Technical Reference(s): EAP-LOSCM Attachment I1 Rule 22 Proposed references to be provided to applicants during during examination: NONE Learning Objective: EAP-LOSCM EAP-LOSCM R6 Question Source: NEW Question NEW Question History: Last Question History: Last NRC NRC Exam Exam N/ANIA Question Cognitive Cognitive Level: Knowledge Knowledge andand Fundamentals

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM 1I POINT POINT Question 33 Question Unit 11 plant Unit conditions:

plant conditions:

• • Reactor trip from Reactor trip from 100% power 100% power

• • RCS pressure =

RCS pressure 88 psig

= 88 decreasing psig decreasing

• • Reactor Reactor Building pressure =27 Building pressure 27 psig psig increasing increasing

• • 11 B B LPI LPI Pump Pump failed failed to to start Based on Based above conditions, on the above conditions, which ONEONE of the the following describes describes the guidance provided in EOP Enclosure 5.1 5.1 (ES Actuation) regarding the LPI pumps pumps and system operation?

EOP Enclosure EOP Enclosure 5.1 (ES Actuation)

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 IA LPI pump and manually re-aligns LPI flow down both the 1A IA and 1lB B LPI headers C. utilizes the 1A and the 11C C LPI Pump and aligns flow down both headers with 1ILP-9 LP-9 and 10 closed D. utilizes only the 1 IC C LPI Pump and aligns flow down both headers with 1 LP-9 & 10 1LP-9 open

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 33 Question TI/GI --cpw T1/G1 cpw 011 EK2.02 Break LOCA Large Break 011 EK2.02 Large LOCA Knowledge of Knowledge of the the operational operational implications implications ofof the the following following concepts concepts asas they they apply to the Large Break apply to the Large Break LOCA:LOCA:

Pumps Pumps (2.6/2.7)

MATCH ANALYSIS KIA MATCH KIA Requires knowledge Requires knowledge of of operational operational requirements requirements provided in in the EOP 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 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 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 C pump to the B B header would replace the flow lost down that header. Additionally plausible since direction to use the C pump is contained in Encl.

End. 5.1 for other failures.

D. Incorrect: Plausible direction to use the C C pump down both LPI headers is contained in Encl.

End. 5.1 and would be the correct actions if both the A and BB LPIPs LPIP's have failed.

Technical Reference(s)

Reference(s):: EPIIIAII800IOOI EP/1/A11800/001 (EOP) End.Encl. 5.1 (ES Actuation), EAP-ESA Proposed references to be provided to applicants during during examination: NONE NONE Learning Objective:

Objective: EAP-ESA R17 Question Source:

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

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 44 Question Unit 11 plant Unit plant conditions:

conditions:

• • Reactor Reactor power power == 100%

100%

• • 11TA and 11TB TA and lockout occurs TB lockout occurs Based on Based on the the above above conditions, conditions, which which ONE ONE of the following of the foNowing indicates indicates the the initial initial Thot Thot and and Tcold values expected Tcold expected once once stable stable natural natural circulation circulation flow has been established?

has been established?

approximately _ _ _ oF and Tcold would be approximately _ _ _ oF.

Thot would be approximately 562 I 532 A. 562/532 582 /532 B. 582/532 C. 585/555 D. 605/555

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 44 Question T1/G1 - cpw TIIGI - cpw O15AA1.21 015AA 1.21 RCP RCP Malfunctions Malfunctions Ability to Ability operate and to operate and II or monitor the or monitor following as the following as they apply to they apply to the the Reactor Reactor Coolant Pump Malfunctions Coolant Pump Malfunctions (Loss of (Loss of RC RC Flow):

Flow):

Development of Development natural circulation of natural circulation flow flow (4.4/4.5)

(4.4/4.5)

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

ANSWER CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since these values would be correct under the misconception 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 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 ATllT between the hot and cold legs is established, generally at 30-40°F.

Since normal post trip temperature is approximatel approximately y 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 is plausible since 50 degree delta TT is the normal value for 100%100% power.

Technical Reference(s)

Reference(s):: TA-AMI TA-AM1 Proposed Proposed references references toto be be provided provided to to applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: TA-AMI TA-AM1 R3 R3 Question Question Source:

Source: NEW NEW Question Question History:

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

Level: Comprehens Comprehension ion and and Analysis Analysis

2010A NRC REACTOR OPERATOR EXAM 1I POINT Question 5 Unit 11 initial conditions:

• Reactor in MODE 5

• RCS pressure = = 0 psig

• Normal LPI decay heat removal in service

• 1IC C LPI pump operating

• Unit Blackout occurs Current conditions:

MFBs re-energized from CT-5

• MFB's Based on the above conditions, which ONE of the following describes actions required by AP/26 (Loss of Decay heat Removal)?

AP/1 1 (Recovery from Loss of A. Start previously running LPI pump AND initiate AP/11 Power)

B. Start previously running LPI pump AND initiate the Blackout tab of the EOP C. Feed and steam SG's SGs to maintain CETC <<246°F AP/1 1 (Recovery from 246°F AND initiate AP/11 Loss of Power)

D. Feed and steam SG's SCs to maintain CETC <<246°F 246°F AND perform AP/26 Encl.

End. 5.6 (Venting LPI Pumps and Suction Lines)

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 55 Question TIIGI T1/G1 --cpw cpw 025AG2.4.11 025AG2.4.11,, LossLoss ofof Residual Residual HeatHeat Removal Removal System System Knowledge of Knowledge of Abnormal Abnormal Condition Condition Procedures Procedures (4.0/4.2)

(4.0/4.2)

KIA MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of mitigation mitigation strategy strategy for for AP/26 AP126 (Loss (Loss of of Decay Decay Heat Heat Removal)

Removal)

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: A Answer: A A. CORRECT:

A. CORRECT: AP/26AP126 directs directs ensuring ensuring previously previously running running LPILPI pump pump is is operating.

operating.

After pump is running After pump is running AP directs initiating AP/11 directs initiating APIII which will recover recover from from the the loss of loss of power power to MFBs to MFB's B. Incorrect:

B. part isis correct. The second Incorrect: The first part part is second part is plausible plausible since since entering the EOP 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 full LPIPs cannot be restarted. Since RCS pressure = = 0 psig, the RCS loops cannot be full. Second part is correct.

D. Incorrect:

D. Incorrect: First part is plausible since this would be a correct choice if RCS loops are LPIPs cannot be restarted. Since RCS pressure =

full and would be utilized if LPIP's = 0 psig, psig, the 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 COOling you would be making preps to restart LPI pumps and running this end encl to ensure the pumps are water solid is plausible.

Technical Reference(s)

Reference(s):: APII IAII 7001026, EAP-APG Enclosure AP/1/A11700/026, Enclosure AP26 Proposed Proposed references to to be provided provided toto applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: EAP-APG EAP-APG R9 R9 Question Question Source:

Source: NEW NEW Question Question History:

History: Last Last NRC NRC ExamExam N/A N/A Question Question Cognitive Cognitive Level:

Level: Comprehens Comprehension ion and and Analysis Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 66 Question Unit 11 plant Unit conditions:

plant conditions:

• Reactor

• Reactor trip trip occurred occurred

• • Total RCP seal Total RCP seal injection flow =

injection flow =0 gpm 0 gpm

• • Component Component Cooling Cooling isis unavailable unavailable Based on Based on the above conditions, the above conditions, which which ONE ONE ofof the following describes the following describes the the immediate immediate action(s) required action(s) reguired by the EOP by the EOP and and the the reason reason for the action(s)?

for the action(s)?

Initiate...

Initiate ...

A. AP/14 A. AP/14 (Loss (Loss of of Normal Normal HPIHPI Makeup Makeup and/or and/or RCP RCP Seal Seal Injection)

Injection) to to restore restore RCP RCP seal seal injection injection B. AP/16 (Abnormal B. (Abnormal Reactor Reactor Coolant Pump Pump Operation) to secure all all RCP's 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 66 Question TIIGI --cpw T1/G1 cpw 026AK3.03 026AK3.03 Loss Loss ofof Component Component Cooling Cooling Water Water Knowledge of Knowledge of the the reasons reasons for the following for the following responses responses as they apply as they apply to to the the Loss Loss of Component Cooling of Component Cooling Water: Water:

Guidance actions Guidance actions contained contained inin EOP for Loss EOP for Loss ofof CCW.

CCW.

(4.0/4.2)

(4.0/4.2)

KIA MATCH ANALYSIS K/A MATCH ANALYSIS Requires knowledge Requires knowledge of of reason reason EOPEOP IMA's IMAs direct direct initiating initiating AP/25 AP125 when when RCP RCP seal seal injection and injection and CCCC have have been been lost lost ANSWER CHOICE ANSWER CHOICE ANALYSIS Answer: D Answer: D Incorrect: Plausible since A. Incorrect: since the entry conditions for AP/14 are met and the EOP does direct entry into into AP's in other conditions (Ex.

APs in (Ex. AP/11 AP/1 1,, AP/25). The EOP EOP does not not direct actions to restore normal seal injection. Seal injection flow is re-established re-established via the RCMUP since both CC and SI have been lost.

B. Incorrect: Plausible since direction are given to trip RCP's 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 AP's APs in other conditions (Ex. AP/11 AP/1 1,, AP/25).

C. Incorrect: Plausible since the entry conditions for AP/20 are met and the EOP does APs in other conditions (Ex. AP/11 direct entry into AP's AP/1 1,, AP/25). Restoring CC is not directed by the EOP. IMA'sIMAs 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. AP/25. These directions are part of EOP Immediate Manual Actions performed by the RO.

Technical Reference(s)

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

Objective: EAP-IMA EAP-IMA R6 Question Question Source:

Source: New New Question Question History:

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

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC REACTOR OPERATOR EXAM 1I POINT Question 7 Unit 11 initial conditions:

• • Reactor power = = 90%

• 11 B Main Feedwater pump trips Current conditions:

• Reactor power = 70% decreasing

• RCS pressure = 2165 psig slowly decreasing

• Pressurizer level = 228 inches slowly decreasing 640T slowly decreasing

• Pressurizer temperature = 640°F

• All Pressurizer heater banks controlled from Unit 1I 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

2010ANRC 2010A NRC REACTOR REACTOROPERATOR OPERATOR EXAM EXAM Question 77 Question TIIGI T1/G1 --cpw cpw 027AK2.03 Pressurizer Pressure 027AK2.03 Pressurizer Pressure Control Control System System (PZR (PZR PCS)

PCS) Malfunction Malfunction Knowledge of Knowledge ofthe the interrelations interrelations between between the the Pressurizer Pressurizer Pressure Pressure Control Control Malfunctions and the Malfunctions and the following: following:

Controllers and Controllers and positioners positioners (2.6/2.8)

(2.6/2.8)

KIA MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of how how controllers controllers forfor PzrPzr saturation saturation circuit circuit function function andand thethe ability to diagnose a malfunction of ability to diagnose a malfunction of it.

it.

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer:

Answer: B B A. Incorrect: First A. Incorrect: First part part isis correct.

correct. Second Second partpart isis plausible plausible since since parameters parameters given given areare reasonable for the post runback reasonable for the post runback condition.condition. Normal Normal pressurizer pressurizer spray spray valve valve RC-1 RC-1 would open would open at 2205 psig at 2205 psig and and not not closed closed until until pressure pressure reaches reaches 2155 2155 psig.

psig. The The decreasing RCS pressure decreasing RCS pressure could be could be explained explained by the decreasing by the decreasing Pzr Pzr level level as as itit returns to returns to setpoint setpoint after after FDWP FDWP trip.

trip.

B. CORRECT:

B. CORRECT: Saturation temp for 2165 psig psig is is approximately approximately 648 degrees. With Pzr at 640 degrees it is clearly subcooled. Regarding the Pzr Regarding the pressurizer 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 Bank 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).

2165). Saturation for actual pzr temp of 640°F is about 2045 psig therefore Bank 2 should be energized. 2205 psig.

Bank C.

C. Incorrect:

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

psig. The decreasing decreasing RCS pressure pressure could could be be explained explained byby the decreasing decreasing Pzr Pzr level level as itit returns to to setpoint after FDWP trip.

after FDWP D. Incorrect:

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

saturated. Second Second part part is is correct.

correct.

Technical Technical Reference(s)

Reference(s):  : PNS-PZR PNS-PZR Proposed Proposed references references to to be be provided provided toto applicants applicants during during examination:

examination: NONENONE Learning Learning Objective:

Objective: PNS-PZR PNS-PZR R5, R5, R7 R7 Question Question Source:

Source: NEW NEW Question Question History:

History: Last Last NRC NRC ExamExam N/A NIA Question Question Cognitive Cognitive Level:

Level: Comprehens Comprehension ion and and Analysis Analysis

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM POINT 1I POINT Question 88 Question Unit 22 initial Unit initial conditions:

conditions:

• • Time Time == 1200:00 1200:00

• • Reactor Reactor power power = 100%

100%

• Both MFDW Pumps Pumps tripped Current conditions:

Current

• • Time Time=1201:45

= 1201 :45

• Reactor power = 30% decreasing

• Rule 11 (ATWS/Unanticipated Nuclear Power Production) in progress

• Loop A SCM = OaF 0°F stable

• 2SA9/D2 (RC Pump Vibration High) 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) (RCPs) provided in Rule 2 and the reason for those actions?

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

2010A NRC REACTOR OPERATOR EXAM Question 8 TIIGI --cpw T1/G1 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 KIA Requires knowledge of EOP directed actions for an ATWS as well as the reasons for those actions.

ANSWER CHOICE ANALYSIS Answer: B RCPs running is correct however the reason is due to the added A. Incorrect: Leaving RCP's 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 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).

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

RCPs is plausible since this would be D. Incorrect: Both parts are incorrect. Securing RCP's 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 R11 Question Source: NEW Question History: Last NRC Exam N/A NIA Question Cognitive Level: Comprehension and Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 99 Question Unit 11 plant Unit plant conditions:

conditions:

• • SGTRSGTR tab tab inin progress progress

• • 11 BB SG isolated SG isolated

• 1A loop

• 1A loop Tcold Tcold == 440°F 440°F decreasing decreasing

• • 11 BB S/G S/G TUBE/SHELL TUBE/SHELL DT DT = (-)72°F

= (-)72°F Based on Based on the the above above conditions, conditions, which which ONE ONE ofof the following describes the following describes the the reason reason the the SGTR tab directs minimizing SGTR tab directs minimizing core SCM core SCM during during cooldown cooldown AND AND the the initial initial method method used used to to reduce the reduce the SCM?

SCM?

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

primary to A. primary to secondary secondary leak leak rate rate / de-energizing de-energizing Pzr Pzr heaters heaters and and cycling cycling Pzr Pzr spray spray B. primary to secondary leak rate / cycling the PORV C. compressive stresses in the 11 B SG / de-energizing Pzr heaters and cycling Pzr spray D. compressive stresses in the 1 lB B SG / cycling the PORV

2010A NRCNRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 9 TI/GI --cpw T1/G1 cpw 038EK3.01 Steam Generator Tube Rupture Rupture (SGTR)

(SGTR)

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

SGTR:

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

of ruptured S/G.

(4.1 /4.3)

(4.1/4.3)

K/A MATCH ANALYSIS KIA Requires knowing the reason for equalizing pressure on primary and secondary sides of ruptured S/GSIG 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.

compressive, stresses across SG C. Incorrect: First part is plausible since controlling compressive tubes is a prime concern during SGTR. 1 1BB 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 IBB 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 Proposed references to be provided to applicants duringduring examination: NONE Learning Objective: EAP-SGTR R9, R6 Question Source: New New Question History: Last NRC NRC Exam N/A NIA Question Question Cognitive Cognitive Level:

Level: Comprehension Comprehension and and Analysis Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 11 POINT POINT Question 10 Question 10 Unit 11 plant Unit plant conditions:

conditions:

• • BothBoth MFB's MFBs de-energized de-energized

• TDEFWP

• TDEFWP operating operating Based on Based on the the above above conditions, conditions, which which ONE ONE of of the the following following describes describes thethe status status of of bearing oil bearing oil cooling cooling water water supply supply toto the the TDEFWP?

TDEFWP?

TDEFWP bearing TDEFWP bearing oil oil cooling cooling is currently provided is currently provided by by and and itit _ _ __

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

re-established.

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

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 10 Question 10 TIIGI -cpw T1/G1 - cpw 054AA11.03 054AA Loss of

.03 Loss of Main Feedwater (MFW)

Main Feedwater (MFW)

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

Feedwater (MFW):

AFW auxiliaries, AFW auxiliaries, including including oil cooling water oil cooling water supply.

supply.

(3.5/3.7)

(3.5/3.7)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Loss of Loss of both MFBs will both MFB's will result result in in loss loss ofof both both MFWP's.

MFWPs. Question Question requires requires knowledge of how bearing knowledge of how bearing cooling watercooling water supply supply to to TDEFWP TDEFWP responds responds to to loss loss of of power.

AC power.

ANSWER CHOICE CHOICE ANALYSIS Answer: B A. Incorrect:

Incorrect: First 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 now unavailable.

Technical Reference(s)

Reference(s):: CF-EF Proposed Proposed references to be provided to to be to applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: CF-EF CF-EF R26R26 Question Question Source:

Source: NEW NEW Question Question History:

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

Level: Knowledge Knowledge and and Fundamental Fundamentals s

OPERATOR EXAM 2010A NRC REACTOR OPERATOR 1I POINT Question 11 II Unit 11 initial conditions:

• Station Blackout occurred

• Neither KHU automatically started

• Manual Emergency Start of BOTH KHU's 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 KHUs have received an Emergency Start signal AND the Generator Output BOTH KHU's Voltage (KV) of the KHU'sKHUs 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 2010A NRC REACTOR REACTOR OPERATOROPERATOR EXAM EXAM Question 11 Question 11 r

TIIGI T1/G1 --cowcpw 055EA1 055EA 1.02 .02 Loss Loss of Offsite and of Offsite and Onsite Onsite Power Power (Station (Station Blackout)

Blackout)

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

Blackout:

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

(4.3/4.4)

(4.3/4.4)

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

operation.

ANSWER CHOICE ANSWER CHOICE ANALYSIS Answer: A Answer: A A. CORRECT: There are 22 switches switches in the Unit Unit 1I and 22 control control room room for emergency starting emergency starting the KHU's.

KHUs. Both Both switches switches are are associated associated with Unit Unit 1ls

'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. KHUs. Since the first switch was operated at 1201, that is the time both KHU's KHUs would have received an Emergency Start signal. Once operating, the normal output voltage for a KHU is 13.8 KV.

Ky.

B. Incorrect: First part is plausible since there are 2 switches and 2 KHU's 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.

KHU's Additional plausibility comes from the fact that when the EOP directs manually starting both KHUsKHU'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:

Incorrect: First part part is is plausible since since there are 22 switches and and 22 KHUs KHU's therefore itit would bebe reasonable to assume that there isis aa switch switch for each each KHU (since the KHUs KHU's areare redundant redundant in in themselves) themselves) and and therefore therefore chose 1202 1202 as as the the time.

Additional plausibility Additional plausibility comes comes from the fact the fact that that when when the the EOP EOP directs directs manually manually starting starting both both KHUs KHU's itit directs directs using using both both switches.

switches. Second Second partpart is is plausible plausible since since 4.16KV 4.16KV is the end voltage used is the end voltage used by ONS. by ONS.

Technical Technical Reference(s)

Reference(s):: EOP EOP End.Encl. 5.38 5.38 (Restoration (Restoration of of Power),

Power), EAP-BO EAP-BO Attach Attach #1,

1. 1, EL-KHG EL-KHG Proposed Proposed references references to be provided to be provided toto applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: EAP-BOEAP-BO R6, R6, R7, R7, R13 R13 Question Question Source:

Source: NEW NEW Question Question History:

History: LastLast NRC NRC ExamExam NIA N/A Question Question Cognitive Cognitive Level:

Level: Knowledge Knowledge and and Fundamental Fundamentals s

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

power?

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

power).

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

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

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

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 12 Question 12 TIIGI -cpw T1/G1 - cpw 057AA2.15 057 AA2.15 Loss Loss of Vital AC of Vital Electrical Instrument AC Electrical Instrument Bus Bus Ability Ability to determine and to determine and interpret interpret the the following following as as they they apply apply toto the the Loss Loss ofof Vital Vital AC Instrument Bus:

AC Instrument Bus:

That aa loss That loss ofof ac has occurred.

ac has occurred.

(3.8/4.1)

(3.8/4.1)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requires the Requires the ability ability to interpret indications to interpret indications toto determine determine that that aa loss loss of of the the normal normal AC supply to KVID has occurred.

AC supply to KVID has occurred. Per N NRC, global loss of AC not global loss not required to match KA.

ANSWER CHOICE ANSWER CHOICE ANALYSIS 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 panelboards 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 Additionally boards but not panelboards 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 0P11107101 OP/1107/010 0 (Operation of the Batteries and Battery Chargers) since there is no automatic backup.

D. Incorrect:

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

panelboards.

Technical Technical Reference(s)

Reference(s):: ISA-I 31B7, EL-VPC 1SA-13/B7, EL-VPC Proposed Proposed references references to be provided to be provided to to applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: EL-VPC EL-VPC R21R5R2/R5 Question Question Source:

Source: NEW NEW Question Question History:

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

Level: Comprehens Comprehension ion and and Analysis Analysis

2010A NRC REACTOR OPERATOR EXAM 11 POINT Question 13 Plant conditions:

• 11 CA Battery Charger fails - output voltage =

- = 0 VDC

• 11CA CA Battery voltage = 126 VDC

• 1 DCB Bus voltage 1DCB = 123 VDC

• Unit 2 DCAlDCB DCAIDCB Bus voltage = = 124 VDC

• DCAIDCB Bus voltage = 127 VDC Unit 3 DCAlDCB Based on the above conditions, which ONE of the following will automatically supply power to 11 DIA panelboard?

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

2010ANRC 2010A NRC REACTOR REACTOROPERATOROPERATOREXAM EXAM Question 1313 r:

Question TI/GI T1/G1 cpwcpw 058AK1 .01, Loss 058AK1.01, Loss of DC Power ofDC Power Knowledge Knowledge of the operational ofthe operational implications implications of ofthe thefollowing following concepts concepts as as they apply they apply toto Loss Loss ofofDCDC Power:

Power: Battery Batterycharger chargerequipment equipment and instrumentation and instrumentation (2.8/3.1)

(2.8/3.1 K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requir es knowledge Requires knowledge of of the the operational operational implications implications of of failed failed battery battery charger charger andand the operational the operational impact impact of the loss of the loss of of aa Vital Vital DCDC Battery Battery Charger Charger and the response and the response by the by the Vital Vital DC DC system system ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer:

Answer: B B A. Incorrect.

A. Incorrect. For For the the Vital Vital DC DC system, system, the the 11 DCB DCB bus bus isis not not aligned aligned toto the the 1I DCA DCA bus.

bus.

Plausible because Plausible because 11DCB DCB BusBus isis aligned aligned to backup the to backup the essential inverters.

essential inverters.

B. Correct.

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

C. Incorrect, plausible C. Incorrect, plausible because because this would be correct if the Unit 2 DC DC busbus voltage was was higher than the 11 CA battery voltage.

higher D. Incorrect.

D. Incorrect. Unit 3's3s DC Bus is not connected to Unit 1. Plausible because because unit unit 33 does does backup backup Unit 1 1 in the SSF power scheme.

Technical Reference(s): Lesson Plan EL-DCD Technical Reference(s):

Proposed Proposed references to be provided to applicants during during examination: None None Learning Learning Objective:

Objective: EL-DCD EL-DCD R4 R4 Question Question Source:

Source: Bank Bank Question Question History:

History: Last Last NRC NRC Exam Exam 20092009 (modified)

(modified) #14#14 Question Question Cognitive Cognitive Level:

Level: Compr ehension and Comprehension and Analysis Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT 1

Question 14 1 and 2 initial conditions:

Unit 1

• A and B LPSW pump operating Current conditions:

• 1 ISA9/A9 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 Pump C 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAMEXAM Question 14 Question 14 TIIGI --cpw T1/G1 cpw 0062AG2.1 0062AG2.1.20 Loss of

.20 Loss of Nuclear Service Water Nuclear Service Water Ability to Ability interpret and to interpret and execute execute procedure procedure steps.

steps.

(4.6/4.6)

(4.6/4.6)

MATCH ANALYSIS KIA MATCH KIA system indications to AP/24 requirements and Requires ability to compare LPSW system execute appropriate steps execute 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 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):

Reference(s): AP124 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 Question Cognitive Cognitive Level: Comprehens Comprehensionion or or Analysis Analysis

2010ANRC 2010A NRCREACTOR REACTOROPERATOR OPERATOREXAM EXAM 11 POINT POINT Question 15 Question 15 Unit33 initial Unit initial conditions:

conditions:

• • Time Time= 1200

= 1200

• Reacto

• Reactor power r power == 45%

45% stable stable

• Operat ing Main

• Operating Main Feedwater Feedw ater Pump Pump trips trips

• IA IA and and AlAAlA are are lost lost to 3FDW-315 to 3FDW-315 Current conditions:

Current conditions:

• Time=

• Time = 1300 1300

• RCS Tave

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

level and the status of 3FDW-315?

ASSUME NO ASSUME NO OPERATOR OPERATOR ACTIONSACTIONS HAVEHAVE OCCURRED OCCURRED A. 25" A. s/U level/failed 25 S/U level / failed open B. 25" B. 25 S/U 5/U level/controlling level / controlling SG level at setpoint setpoint C. 30" C. 30 XSUR XSUR / failed open D. 30" D. 30 XSUR XSUR / controlling SG level at setpointsetpoint

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 1515 TIIGI --cpw T1/G1 cpw 065AA2.07 Loss of Instrument Instrument Air determine and interpret Ability to determine interpret the following as they apply to the Loss Loss of Instrument Air:

Whether backup nitrogen supply is is controlling valve position.

(2.8/3.2)

K/A MATCH ANALYSIS KIA 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".

30. 25" 25 is Plausible since it would be the correct level if Main FDW were supplying the SG's. 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"25 is the normal level setpoint.

30. 25 is Plausible since it would be the correct level if B. Incorrect: Correct level is 30".25" Main FDW were supplying the SG's. 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 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 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> the backup N2 supply would be regulating valve position as required. Since level being controlled by EFDW the correct level is 30 30" XSUR.

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

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM POINT 1I POINT Question 16 Question 16 Unit 11 plant Unit plant conditions:

conditions:

• • Reactor Reactor power power == 50%50% stable stable

• Units

• Units 22 and and 33 inin MODE MODE 55

• Grid disturbance

• Grid disturbance isis in in progress progress

• • BOTH BOTH KHU's KHUs generating generating to to the the grid grid

• AllAN but but ONE ONE of the Offsite of the Offsite Sources Sources required required by Tech Spec by Tech Spec 3.8.1 3.8.1 (AC (AC Sources-Sources-Operating) are Operating) are lost lost Based on Based on the above conditions the above conditions whichwhich ONE ONE of of the the following following describes describes ...

1) actions required

1) actions required by by Tech Tech Spec Spec 3.8.1 3.8.1 ??

AND

2) condition that

2) aa condition that would require require manually manually separating separating BOTH BOTH KHU's KHUs from from the the electrical 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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 16 Question 16 TI/GI --cpw T1/G1 cpw Generator Voltage 077AG2.2.222 Generator ro77AG2.2.2 Voltage and and Electrical Electrical Grid Grid Disturbances Disturbances Knowledge of Knowledge limiting conditions of limiting conditions for for operations operations andand safety safety limits.

limits.

(4.0/4.7)

(4.0/4.7)

KIA MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of LCO LCO entry entry conditions conditions bases bases onon degraded degraded grid grid ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer:

Answer: D 0 Incorrect: First A. Incorrect: First part is plausible part is plausible since since entering entering LCO LCO 3.0.3 3.0.3 is is generally generally done done when when aa loss of safety loss of safety function function hashas occurred occurred and and itit would would be be aa reasonable reasonable misconception misconception that that the KHU's KHUs would be inoperable 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. 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 misconceptionmisconception that KHUs would be inoperable due to generating to the grid during a grid the KHU's 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.

IC's.

0.

D. CORRECT: With one or both required offsite sources inoperable TS 3.8.1 Condition JJ requires energizing both SBBs SBB's within 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />. Per AP134AP/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)

Reference(s):: TS 3.8.1 AP134 AP/34 (Degraded Grid)

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

Objective: ADM-TSS R4 R4 EAP-APG EAP-APG R8 Question Question Source:

Source: NEW NEW Question Question History:

History: Last Last NRC NRC ExamExam N/AN/A Question Question Cognitive Cognitive Level:

Level: Knowledge Knowledge andand Fundamental Fundamentals s

2010ANRC 2010A NRCREACTOR REACTOROPERATOROPERATOREXAM EXAM 1I POINT POINT Question 17 Question 17 Unit 11 initial Unit initial conditions:

conditions:

• • Reactor Reactor powerpower == 100%

100%

• lCHPl

• 1C HPI pump OOS pumpO OS Current conditions:

Current conditions:

• • 11A and 1lB A and Main FDW B Main FDW pumps pumps tripped tripped

• Condensate Booster

• Condensate Booster Pumps Pumps unavailable unavailable

• All EFDW pumps

• All EFDW pumps unavailable unavai lable

• 1A and 1 B SG Outlet

• 1A and 1B SG Outlet pressure pressu re == 860 860 psig psig slowly slowly decreasing decreasing

• RCS RCS pressure pressure == 23172317 psig psig increasing increasing Based on Based on the the above above conditions, conditions, which which ONE ONE of of the the following following describes describes the the required required operato r action( s) in accord ance with operator action(s) in accordance with the EOP? the EOP?

A. Establish A. Establish SSF SSF ASWASW flow to thethe SG SG and and establish establish SG SG levels levels at at 240 240 inches.

inches.

B.

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

SG5.

C.

C. Establish HPI Establish HPI forced forced cooling and open 11HP-410.

HP-41 O.

D.

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

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 17 Question 17 TIIGI -cpw T1/G1 - cpw BEO4EA2.2 Inadequate BE04EA2.2 Inadequate Heat Heat Transfer Transfer Ability to Ability determine and to determine and interpret interpret the the following following as as they they apply apply toto the the (Inadequate (Inadequate Heat Transfer):

Heat Transfer):

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

license amendments.

(3.6/4.4)

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

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 11 HPIP operating then actions are taken to align SSF ASW to feed the SG's.SGs. 240" 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 11 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 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 D. B HPI header will be inadequate which will require the operator to open HP-409.

Technical Reference(s)

Reference(s):: EAP-LOHT Rule 4 Proposed references to be provided to applicants during examination: NONE Learning Objective: EAP-LOHT EAP-LOHT R24, R28 Question Source:

Source: BANK EAPO7OIO2 EAP070102 Question History:

History: Last NRC Exam Oconee RO 2006 Question Cognitive Question Cognitive Level: Comprehens Comprehension and Analysis ion and Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 18 Question 18 Unit 11 initial Unit initial conditions:

conditions:

• • Reactor Reactor power power == 100%

100%

• • 11AMSLB occurs A MSLB occurs Current conditions:

Current conditions:

• • Reactor Reactor has has tripped tripped

• • RCSRCS TaveTave == 544°F 544T slowly slowly increasing increasing

• • 11A SG Pressure A SG Pressure == 00 psig psig

• • 11 B B SG Pressure =

SG Pressure psig slowly increasing

= 990 psig increasing

• • Turbine Turbine bypass bypass valves valves inin Auto Auto

• • Reactor Reactor Building Building pressure pressure == 0.2 0.2 psig psig stable 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 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 2010A NRC REACTOR REACTOROPERATOR OPERATOR EXAM EXAM Question 18 Question 18 TI/GI T1/G1 - cpw

- cpw BEO5EK2. I Excessive BE05EK2.1 Excessive Heat Heat Transfer Transfer Knowledge of Knowledge ofthe the interrelations interrelations between between the the (Excessive (Excessive Heat Heat Transfer)

Transfer) andand the the following:

following:

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

manual features.

(3.8/4.0)

(3.8/4.0)

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

TDEFWP.

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer:

Answer: C C A. Incorrect:

A. Incorrect: Plausible Plausible since since 11FDW-315 FDW-315 is is closed closed in in first first step step ofof Rule Rule 5.

5. This This makes makes itit plausible that plausible that AFIS AFIS would would notnot secure secure the the TDEFWP TDEFWP so that that itit would be be available to feed the B B SG if needed.

B. Incorrect: Plausible since 1IFDW-315 B. FDW-315 is closed in first step of Rule 5. This makes itit 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:

actuation (ox: ES components).

components).

CORRECT: The TDEFWP control switch will override the AFIS interlock to C. CORRECT:

C.

close close TO-I TO-145.45. TO-I TO-145 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 feed Steam Generators without resetting the AFIS signal.

D.

D. Incorrect:

Incorrect: TDEFWP TDEFWP would be be off.

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

(ex: ES components)..

Technical Technical Reference(s)

Reference(s):: CF-EF CF-EF Proposed Proposed references references to to be be provided provided to to applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: CF-EF CF-EF R58 R58 Question Question Source:

Source: NEW NEW Question Question History:

History: Last Last NRCNRC Exam Exam N/A N/A Question Question Cognitive Cognitive Level:

Level: Comprehens Comprehension ion andand Analysis Analysis

NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAM EXAM 11 POINT POINT Question 19 19 Unit 11 plant conditions:

Unit

• • Reactor Reactor power power = 70% stable 210 slowly decreasing

• Pressurizer level = 210"

• 11HP-120 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...

Throttle ...

A. 11HP-26 HP-26 /I 200 B. 1IHP-26 HP-26 / 80 C. 1 1HP-122 HP-122 (RC VOLUME CONTROL BYPASS) / 200 D. 1 1HP-122 HP-122 (RC VOLUME CONTROL BYPASS BYPASS!/ 80

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 19 Question 19 T1/G2 - cpw T1/G2 - cpw 028AA1 .07 Pressurizer (PZR) 028AA 1.07 Pressurizer (PZR) Level Level Control Control Malfunction Malfunction Ability to Ability operate and to operate and II or or monitor monitor the following as the following as they they apply apply to to the the Pressurizer Pressurizer Level Control Malfunctions Level Control Malfunctions:  :

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

backup).

(3.3/3.3)

(3.3/3.3)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of how how HPIP's HPIPs and valves are and valves utilized to are utilized to maintain maintain pressurizer pressurizer level following level following aa level level control control valve valve (HP-120)

(HP-120) failure.

failure. Since Since pump pump operation operation is is not not impacted by directly impacted by a failure of the level level control valve (HP-120),

(HP-120), manual manual throttling of HP-26 HP-26 is how Pzr level level is is maintained.

maintained.

ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: AP/14 APII4 directs throttling makeup through HP-26 to maintain PZR

>200.

>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" 80 is the pressurizer level required to maintain pressurizer heater operability. Rule 6 allows throttling provided pressurizer level is increasing and with the 80" 80 heater cutoff it could be a misconception that 80" reasonable misconception 80 is the low level limit.

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

D. Incorrect: Both parts are incorrect. First part is plausible since 1HP-122 1HP-122 would be correct if 1 1 HP-26 would not open. Second part is plausible since 80 80" is the pressurizer level required to maintain pressurizer heater operability. Rule 66 allows throttling provided pressurizer level is increasing increasing and with the 8080" heater cutoff it could be aa reasonable misconceptio misconception n that 80 80" is the low low level limit.

Technical Technical Reference(s)

Reference(s):: EAP-APG EnclosureEnclosure APII4 AP/14 Proposed Proposed references to to be be provided provided to to applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: EAP-APG EAP-APG R9 R9 Question Question Source:

Source: NEW NEW Question Question History:

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

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 20 Question Unit 11 initial Unit initial conditions:

• • Reactor in MODEMODE 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 SFPSEP and empty

• Reactor Building Main Fuel Bridge in transit to the upender with a spent fuel assembly in the mast

• Section 40 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 40 4D (Fuel Transfer Canal Flooded)?

A. Close 11SF-I SF-1 and 1I SF-2 (East/West Transfer Tube Isolations) 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 SFPSEP

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 20 Question 20 TIIG2 -

T1/G2 -

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

following:

Fuel handling Fuel handling equipment.

equipment.

(2.9/3.5)

(2.9/3.5)

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

Carriage).

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: 0D Incorrect: Plausible A. Incorrect: Plausible because because itit will be performed later be performed later however however both both carriages carriages must must be placed inin 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 SEP to allow FTT Isolation valves to be closed. Misconception Misconception about which way the carriage must go to close SF-1 SE-i & 2 D. Correct: Procedure directs placing the fuel assembly in transit into a safe location and specifies the upender or originallinten original/intendedded location and positioning the carriages in the SFP in preparation for closing the FTT Isolation valves.

Technical Reference(s):

Reference(s): AP126 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:

History: Last NRC NRC Exam Exam ONSONS RO RO 2009 2009 #62 (Re-ordered (Re-ordered distracters) distracters)

Question Question Cognitive Cognitive Level:

Level: Comprehens Comprehension ion or or Analysis Analysis

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

inoperable?

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

allowed.

B. Counts B. Counts dodo not increase when not increase when Source Source Check Check isis performed performed /I Batch Batch releases releases are are NOT NOT allowed.

allowed.

C. Sample pump C. Sample pump found found OFFOFF /I Batch Batch releases releases are are allowed.

allowed.

D. Sample D. Sample pump pump found OFF OFF /I Batch Batch releases releases are are NOTNOT allowed.

allowed.

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 21 21 TIIG2 - cpw T1/G2 - cpw 059AK3.03 Accidental LiquidLiquid Radwaste Radwaste Release Release Knowledge of the reasons Knowledge reasons for the following responses responses as they apply to the Liquid Radwaste Accidental Liquid Radwaste Release:

Release:

Declaration that a radioactive-liquid monitor is inoperable.

(3.0/3.7)

K/A MATCH ANALYSIS KIA i NR( Oh to ask about functional vs operable.

operable Also can ask what is prevented by declaring RIA iinoperable (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- 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 PT1230/01 (RIA PT) clearly state that increase however Limits and Precautions of PT/230/01 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 PT/230/01 (RIA PT) RAD-RIA SLC 16.11.3 16.11.3 Proposed references to be provided to applicants during examination: NONE NONE Learning Objective: RAD-RIA R5 Question Source: NEW NEW Question History: N/AN/A Question Cognitive Question Cognitive Level:

Level: Knowledge Knowledge and Fundamentals Fundamentals

2010A NRC REACTOR OPERATOR EXAM 1I 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 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> of removing the detectors 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 067 .03 Plant fire on site AA 1.03 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 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM 1I POINT POINT Question 23 Question 23 Unit 33 plant Unit conditions:

plant conditions:

• • Loop Loop AA and and Loop Loop BB SCM's SCMs == O°F stable OF stable

• • Core Core SCMSCM == (_)5°F flashing with

(-)5°F flashing with aa red background red background Based on Based on the above conditions, the above conditions, which which ONE ONE of of the following describes the following describes the the status status of of the the reactor core?

reactor core?

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

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAMEXAM Question 23 Question 23 T1/G2 - cpw TI/G2 - cpw 074EG2.1.7 Inadequate Core 074EG2.1.7Inadequate Core Cooling Cooling Ability to Ability to evaluate evaluate plant plant performance performance andand make make operational operational judgments judgments based based on on operating characteristic operating characteristics, s, reactor reactor behavior, behavior, and and instrument instrument interpretation.

interpretation.

(4.4/4.7)

(4.4/4.7)

KIA MATCH ANALYSIS KIA MATCH ANALYSIS Requires instrument Requires instrument interpretation interpretation toto make make an an operational operational judgment of of core core performance performance ANSWER CHOICE ANALYSIS Answer: 0D Answer:

A. Incorrect: Plausible since specific knowledge of the operation of the Subcooled Margin Monitors Margin Monitors is required 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 Reference(s):: EAP-ICC,IC-RCI Proposed references to be provided to applicants during examination: NONE Learning Objective:

Objective: EAP-ICC RI, R1, IC-RCI R42 Question Source:

Source: NEW NEW Question Question History:

History: Last Last NRC NRC Exam Exam NIANIA Question Question Cognitive Cognitive Level:

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC REACTOR OPERATOR EXAM I POINT 1

Question 24 Unit 2 initial conditions:

• Reactor power = = 100%

Current conditions:

• 2SA2/B4 (RC AVERAGE TEMP HIGH/LOW) actuated

• Loop 'A' 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 BA02AK1.3 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 KIA 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.

Incorrect: : First part is plausible since actual Tave will be decreasing therefore if ICS D. Incorrect::

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 N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT 1

Question 25 1 initial conditions:

Unit 1

• 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 utNized to direct plant activities AND the expected Steam Generator pressure (psig)?

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

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 25 Question 25 TIIGI - cpw T1/G1 - cpw BAO4AA2.1 BA04AA2.1 TurbineTurbine Trip Trip Ability to Ability determine and to determine and interpret interpret the the following following asas they they apply apply to to the the (Turbine (Turbine Trip):

Trip):

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

emergency operations.

(3.3/3.7)

(3.3/3.7)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requires selecting Requires selecting thethe appropriate appropriate procedures procedures following following aa turbine turbine trip trip ANSWER CHOICE CHOICE ANALYSIS Answer: C C A. Incorrect:

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

<<30%. First part is plausible since this would be the correct answer if initial reactor 30%. First power had had been been >30%

>30% and the reactor did not not trip. Second Second partpart is is correct and would' 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 TBVs would be controlling at setpoint.

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

<30%.

< 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'sTBVs 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 TBV's ,would 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 actuate and still send the 125125 psi bias to the TBVsTBV's which would result in in them controlling at 1010 1010 psig.

Technical Reference(s)

Reference(s):: IC-RPS, IC-RPS, EAP-SA, EAP-SA, APIIAP/1 Proposed Proposed references references to to be be provided provided to to applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: IC-RPS IC-RPS R3,R3, EAP-SA EAP-SA RI R1 Question Question Source:

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

Level: Comprehens Comprehension ion and and Analysis Analysis

NRC REACTOR OPERATOR EXAM 2010A NRC EXAM 1I POINT POINT Question 26 Unit 1I plant conditions:

Unit conditions:

• Reactor power = 100%

• 2SA-18/A11 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)?

and...

Manually trip the reactor and ...

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

2010A NRCNRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 26 TIIG2 - cpw T1/G2 -

BAO7AA2.2 BA07 Flooding AA2.2 Flooding interpret the following as they apply to the (Flooding):

Ability to determine and interpret (Flooding):

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

(3.3/3.7)

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

APIIO ensures operation within license limitations set by TS 5.4 Adhering to AP/10 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 1I was on LPI DHR.

In that case since Unit 1I 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 AP/lO (for step 4.2) which explains that actions taken will result in a loss of condenser vacuum.

C. CORRECT: APIIOAP/10 directs securing all CCWP's 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 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): APIIIAII700IOIO 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 Question NRC Exam N/A N/A Question Cognitive Cognitive Level: Knowledge and Fundamentals

2010A NRC REACTOR OPERATOR EXAM I POINT 1

Question 27 Unit 11 plant conditions:

• Core SCM = 0°F

= OaF

• LOCA Cooldown tab in progress CETCs =

• CETC's 395T slowly decreasing

= 395°F

• 1I 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. 11LP-3 LP-3 B. 1LP-19 C. 1LP-104 D. 1LP-105

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 27 Question 27 TIIG2 - cpw T1/G2 - cpw BEO3EG2.2.3 Subcooling Margin Inadequate Subcooling BE03EG2.2.3 Inadequate Margin Knowledge of Knowledge of the the design, procedural, and design, procedural, and operational operational differences differences between between units.

units.

(3.8/3.9)

(3.8/3.9)

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

ANSWER CHOICE ANALYSIS Answer: D A. Incorrect: Plausible since 11 LP-3 is in the normal DHR drop line however the tap for the alternate PLBD flowpath is between 11 LP-2 & 11 LP-3 therefore 11 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.

correct for Units 2 & 3 however Unit 11 has B. Incorrect: Plausible since LP-19 would be correctfor LP-1 9.

a unique alternate path that does not use 11 LP-19.

C. IIncorrect:

ncorrect: Plausible since 1 1LP-104 LP-1 04 is isaa valve in the normal PLBD flowpath however it is in series with 11 LP-1 03 therefore the failure of 11 LP-1 03 to open renders 11 LP-1 04 useless.

D. CORRECT: For Unit 1, I, another drain line and motor operated valve, 1 LP-1 OS, is ILP-105, installed below LP-1LP-I and LP-2 to provide a second flow path to the Reactor Building Emergency Sump. The addition of 1LP-105 ILP-I05 was required in this flow path due to the arrangement of the Decay Heat Drop Line on Unit 1. I. On Unit 1 I the Drop Line does not drop straight to the suction of the LPI pumps as it does on units 22 and 3, but instead, curves back upwards before reaching the pumps, in effect, forming aa loop seal. This would prevent aa gravity drain path from the hot leg to the LPI pumps suction header as it exists on Units 22 and 33 and requires using ILP-105 1LP-105 for the alternate PLBD flowpath.

Technical Reference(s): PNS-LPI, EOP-LOCACD Proposed references to be provided to applicants during during examination: NONENONE Learning Objective: PNS-LPI R27, R28 Question Question Source: NEW NEW Question History:

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

Level: Knowledge and and Fundamentals

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 28 Question 28 Unit 33 initial Unit initial conditions:

conditions:

• • Time Time == 0300 0300

• • Reactor Reactor power power == 100%100%

• • Reactor Reactor trip trip

• • CT CT-3 Lockout occurs

-3 Lockout occurs Current conditions:

Current conditions:

• • MFBMFB re-energized re-energized

• • 6900V 6900V power power still still unavailable unavailable

• • HPIHPI system system leak leak downstream downstream ofof 3HP-31 3HP-31 occurs occurs

• • 3A 3A11 RCP RCP SISI flow == 3.9 3.9 gpm gpm slowly decreasing slowly decreasing

• 3A2 RCP SI flow = = 3.7 gpm slowly decreasing

• • 3B1 RCP RCP SI flow = 3.5 gpm slowly decreasing

• 3B2 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 2010A NRC REACTOR OPERATOR EXAM REACTOR OPERATOR EXAM Question 28 Question 28 T2IGI --cpw T2/G1 cpw 003K4.1 003K4.111 Reactor Reactor Coolant Coolant Pump Pump Knowledge of Knowledge of RCPS RCPS design design feature(s) feature(s) and/or and/or interlock(s) interlock(s) which which provide provide for for the the following:

following:

Isolation valve interlocks Isolation interlocks (3.0/3.0)

K/A MATCH ANALYSIS KIA knowledge of Requires knowledge of RCP RCP Seal Seal Injection 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 1I 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 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 < 4 gpm so they will be closed. On Units 2&3 if seal injection flow to ALL RCPs < <44 gpm/RCP for> 1 I mm then, (2)(3)HP-21 min (2)(3)HP-2I automatically closes (U1 (UI HP-31 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 2,3HP-31will automatically close ifif SI SI flow is <4

<4 gpm on each pump pump for> one minute.

Technical Reference(s): 0P12,31A/I 104/002 OP/2,3/Al1104/002 Proposed Proposed references to be be provided provided toto applicants applicants during during examination: NONE NONE Learning Learning Objective:

Objective: PNS-HPI PNS-HPI R22 Question Source:

Question Source: NEW NEW Question Question History:

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

Level: Knowledge Knowledge and and Fundamentals Fundamentals

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 29 Question 29 Unit 11 plant Unit conditions:

plant conditions:

• • Reactor Reactor power power == 100%

100%

• • 11CC-8 CC-8 (CC RETURN PENT (CC RETURN PENT (54)

(54) OUTSIDE OUTSIDE BLOCK)

BLOCK) failsfails closed closed

• • 11LPSW-6 (UNIT LPSW-6 (UNIT 11 RCP COOLERS SUPPLY)

RCP COOLERS SUPPLY) failsfails closed closed Based on Based on the the above above conditions, conditions, which which ONE ONE of of the the failed failed valves valves will will require require ALL ALL RCPs RCPs to to be secured be secured in in accordance accordance with with AP/16 AP/16 (Abnormal (Abnormal Reactor Reactor Coolant Coolant Pump Pump Operation)

Operation) and and why?

why?

A. 11 CC-8 / duedue to high high RCP RCP motor stator temperatures B. 11 CC-8 / duedue to high RCP RCP radial bearing temperatures C. 11 LPSW-6 C. LPSW-6 / due due to high high RCP RCP motor motor stator stator temperatures D. 1ILPSW-6 LPSW-6 / due to high RCP radial bearing temperatures

2010A NRC REACTOR OPERATOR EXAM Question 29 T2IGI - okm T2/G1 -

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))

(2.8/3.1 K/A MATCH ANALYSIS KIA CC-8 and LPSW-15 are Containment isolation valves and with both failed Both CC-B 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 RCP's RCPs. RCPs if 295° F in AP/16 stator temps reach 295°F 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 2.0 gpm is still available to cool the RCP bearings bearings. Seal Injection water -2.0 AP/1 6 requires tripping RCP if radial brg temp even though CC has been lost. AP/16 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.

AP/1 6 requires tripping RCP if radial D. Incorrect: First part is correct. Plausible since AP/16 brg temp reaches 225°F. Additional plausibility if there is a misconception regarding cooling of the RCP bearings. Seal Injection water 2.0-2.0 gpm is still available to cool the RCP bearings even though CC has been lost.

Technical Reference(s): PNS-CPM Pg 6 APII6 AP/16 Proposed references to be provided to applicants during examination: NONE Learning Objective: PNS-CPM R1, RI, 19 Question Source: NEW Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM POINT 1I POINT Question 30 Question 30 Unit 11 initial Unit initial conditions:

conditions:

• • Reactor power =

Reactor power 100%

= 100%

• • 1IA CC pump A CC pump operating operating Current conditions:

Current conditions:

• 1 CC-7

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

plant response?

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

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 30 Question 30 T2IGI T2/G1 - cpw- cpw 004K1 .36 Chemical 004K1.36 Chemical and and Volume Volume Control ControlSystem System Knowledge of Knowledge ofthe the physical physical connections connections and/orand/or cause-effect cause-effect relationships relationships between the CVCS and the between the CVCS and the following systems:following systems:

CCWS CCWS (2.6/2.8)

(2.6/2.8)

KIA MATCH K/A MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of cause-effect cause-effect relationship relationship of of loss loss of of CCCC flow flow on on the the HPI HPI system system ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: C Answer: C A. INCORRECT:

A. INCORRECT: Plausible Plausible since since aa loss loss of of CC CC would would ultimately ultimately result result in in aa required required reactor reactor trip once once CRDCRD temps reach 180 degrees however reach 180 degrees however itit isis aa required required manualmanual trip trip and not and not anan automatic automatic trip.trip. Second Second part is is correct since both both CCCC pumps trip when either CC-7 or CC-8 go closed.

either

8. Incorrect: Plausible since a loss of CC would ultimately result in a required reactor B.

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

automatically C. CORRECT:

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 letdown reaches reaches 130°F 130°F aa high temperature stat-alarm will sound and at 135°F 135°F the letdown letdown isolation valve, HP-5, will close. This happens in - 11 minute with a total total loss of CC flow. CC-7 CC-7 closing would result in aa total loss loss of of CC CC flow to the letdown letdown coolers.

D.

D. Incorrect:

Incorrect: FirstFirst part part is is correct.

correct. Second Second part part is is plausible plausible because because ifif CC CC flowflow had had decreased d.ecreased due due toto reasons other other than than CC-7 CC-7 oror CC-8 CC-8 failing closed closed thenthen bothboth CC CC pumps would be operating since the pumps would be operating since the SIB pump would S/B pump would start start on on low low flow.

flow. IfIf either either CC-CC-77 or or 88 close, close, both both CCCC pumps pumps automatically automatically trip.

trip.

Technical Technical Reference(s)

Reference(s):: PNS-HPI, PNS-HPI, PNS-CC PNS-CC Proposed Proposed references references to to be be provided provided toto applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: PNS-HPI PNS-HPI R22, R22, PNS-CC PNS-CC R18,19 R18,19 Question Question Source:

Source: NEW NEW Question Question History:

History: LastLast NRC NRC ExamExam N/A N/A Question Question Cognitive Cognitive Level:

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC REACTOR OPERATOR EXAM 11 POINT Question 31 Unit 3 initial conditions:

• ReactorinMODE4 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 2010A NRC REACTOR REACTOROPERATOROPERATOR EXAM EXAM Question 31 Question 31 T2/G1 T2/G1 - cpw- cpw 005K3.05 Residual 005K3.05 Residual Heat Heat Removal Removal System System (RHRS)

(RHRS)

Knowledge of Knowledge ofthe the effect effect that that aa loss loss or malfunction of or malfunction the RHRS ofthe RHRS will will have have on on the the following:

following: ECCS ECCS (3.7/3.8)

(3.7/3.8)

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

on ECCS availability.

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer:

Answer: C C A. Incorrect: First A. Incorrect: part isis plausible First part plausible since since 3LP-1 3LP-1 is is aa suction suction toto the the LPI LPI pumps pumps however however itit is the is Decay Heat the Decay Heat Removal Removal suction.

suction. Since Since ECCS ECCS suction suction is is from from either either the the BWST BWST or or RBES, RBES, a a failure of LP-1 has no impact of LP-1 has no impact on on ECCS ECCS availability.

availability. Second Second part part is is correct.

correct.

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

C. CORRECT: 3LP-1 is a suction valve to the LPI pumps however it is the Decay C.

Heat Removal suction. Since ECCS suction is from either the BWST or RBES, Heat 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.

D. Incorrect:

Incorrect: FirstFirst part part is is correct. Second Second partpart is is plausible plausible because unit unit 33 has has LPI LPI bypasses around the LPI cooler so bypasses around the LPI cooler so itit would be would be plausible plausible toto determine determine thethe ECCS ECCS traintrain is is still still available available since since II would would still still be be able able to to get get flow flow down down the the header header using using the the cooler cooler bypass bypass valve.

valve.

Technical Technical Reference(s)

Reference(s):: LPI LPI system system drawing, drawing, PNS-LPI PNS-LPI Proposed Proposed references references to be provided to be provided to to applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: PNS-LPI PNS-LPI R5, R5, R29, R29, R30, R30, R14 R14 Question Question Source:

Source: NEW NEW Question Question History:

History: Last Last NRC NRC ExamExam N/A N/A Question Question Cognitive Cognitive Level:

Level: Comprehens Comprehension ion or orAnalysis Analysis

2010A NRCNRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 32 Which ONE of the following describes the highest RCS pressure hicihest RCS pressure (psig)

(psig) at which the 1I LP-1 LP-I (LPI (LPI RETURN RETURN BLOCK BLOCK FROM FROM RCS)

RCS) pressure pressure interlock interlock will allow 1I LP-1 LP-1 to be be opened and the reason 1I LP-1 LP-I has a pressure interlock?

A. 365 /I prevent overpressurizing LPI suction piping B. 365 /I ensure delta p across 1I LP-1 LP-I will allow it to open C. 420 /I prevent overpressurizing LPI suction piping D. 420 /I ensure delta p across 11 LP-1 LP-l will allow it to open

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 32 Question 32 T2/G1 - cpw T2/G1 cpw 005K4.O1 Residual Heat 005K4.01 Residual Heat Removal Removal System System (RHRS)

(RHRS)

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

following:

Overpressure mitigation Overpressure mitigation system system (3.0/3.2)

(3.0/3.2)

K/A MATCH ANALYSIS KIA MATCH ANALYSIS Requires knowledge Requires knowledge of of how LPI overpressure how LPI overpressure protection protection is is accomplished.

accomplished. ThisThis is is done done by an interlock by an interlock that that prevents prevents placing placing LPI LPI DHR DHR piping piping in in service service prior prior to to being below being below 400 400 psi, psi, ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: A A. CORRECT: The 1I LP-1 LP-I interlock prevents prevents system overpressurization overpressurization by preventing 1ILP-I 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 11 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 MSCVs, FDW valves, etc.).

(Ex. MSCV's, C. Incorrect: Plausible since the 1 I LP-1 interlock prevents 1 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 I LP-1 interlock prevents 1 1 LP-1 from being opened when WR RCS pressure (via the Amphenol connector) is >400 psig. Second part is plausible because waiting on aa lower RCS pressure to open I1LP-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 dp is low enough across aa valve before we try to open it (Ex. MSCVs, MSCV's, FDW valves, etc.).

Technical Reference(s)

Reference(s}:: PNS-LPI PNS-LPI pgs 52 & 53 53 Proposed Proposed references references to be provided to be provided to to applicants applicants during during examination: NONE NONE Learning Learning Objective:

Objective: PNS-LPI PNS-LPI R16R16 Question Question Source:

Source: NEW NEW Question Question History:

History: Last Last NRC NRC Exam Exam NIAN/A Question Question Cognitive Cognitive Level:

Level: Memory Memory or or Fundamental Fundamental Knowledge Knowledge

2010A NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 33 Question 33 Unit 11 initial Unit initial conditions:

conditions:

• • Reactor Reactor power power == 100%

100%

Current conditions:

Current conditions:

• • RCS pressure == 1350 RCS pressure 1350 psig psig decreasing decreasing

• • Reactor Building pressure Reactor Building pressure == 4.8 4.8 psig increasing psig increasing

• • ESES Channel Channel 22 did did NOT NOT actuate Based on the above conditions, which ONE of the following describes ALL safety injection pumps pumps that have have AUTOMATICALLY started?

A. 1A HPI, 1lB B HPI, 1A LPI, 1lBB LPI B. 1A HPI, 11C C HPI, 1A LPI, 1lB B LPI C. 1A HPI, 1lB B HPI, 1A IA LPI ONLY D. 1AIA HPI, 1A LPI ONLY

2010ANRC 2010A NRCREACTOR REACTOROPERATOR OPERATOREXAM EXAM Question 33 Question 33 T2/GI T2/G1 - cpw

- cpw 006A3.05 Emergency 006A3.05 Emergency Core Core Cooling Cooling Ability Ability to monitorautomatic to monitor automaticoperation operation ofofthe the ECCS, ECCS, including:

including:

Safety Injecti on Pumps Safety Injection Pumps. .

(4.2/4.3)

(4.2/4.3)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requir es the ability to Requires the ability monitor automatic to monitor automatic start start ofof ECCS ECCS pumps pumps based based on on ES ES channe channels ls that that have have actuated actuated ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answe Answer: Ar: A A. CORRECT:

A. CORRECT: With With RB RB pressure>

pressure > 33 psig psig ES ES 1-61-6 should should have have actuated.

actuated. Since Since LPILPI pumps are pumps are on on ES ES 3&4, 3&4, both both AA& & BB LPIP's LPIPs would would be be operating operating and since both and since both the A the A& &B B HPIP's HPIPs areare onon are are on on ES-1, ES-I, they they would would be be operating.

operating.

B. Incorrect:

B. Incorrect: Plausible Plausible since since the the AA HPI HPI pump pump is is on on ES ES channel channel 1I but the C C HPI HPI pump pump isis on channe l 2. Additi on channel Additionally, onally, all ES ES LPI pumps pumps would be operating operating however however 11C C LPI LPI pump is not pump not an ES ES pump. Additional Additional plausibility plausibility from the fact that other ES ES systems systems that have a "C" that C pump have that pump as one of the ES pumps (ex. HPI & & LPSW)

LPSW)

C. Incorrect:

C. Incorrect: Plausible Plausible since the A and B HPIP's HPIPs are actuated actuated off of the ES channel channel 11 and this and this would be correct if ES pumps were also actuated actuated off of ES 1 I &2 and the the BB LPIP LPIP was on the even numbe numbered channel (like their actual arrangement red channel arrangement on ES 5&6). 5&6).

D. Incorrect.

D. Incorrect. Plausib Plausible le if you assum assume e both HPI and LPI are on ES 1 I &2 and the AA pumps pumps are are on the odd channechannels ls and BB pumps on the even channe channels.

ls.

Techni cal Refere Technical nce(s): IC-ES Reference(s): IC-ES Propos Proposeded referen referencesces to to be be provid ed to provided to applica applicants during examin nts during examination: NONE ation: NONE Learni ng Object Learning ive: IC-ES Objective: IC-ES R14, R14, 18 18 Questi on Source Question Source:: NEW NEW Questi on History Question History:: Last Last NRC NRC Exam Exam NIA N/A Questi on Cognit Question ive Level:

Cognitive Level: Compr ehension and Comprehension and Analys Analysisis

REACTOR OPERATOR EXAM 2010A NRC REACTOR EXAM 1I POINT POINT Question 34 Unit 11 plant conditions:

Unit

• 0P111A111031002, (Filling and OP/1/A/1103/002, and Venting RCS) Enclosure Enclosure 4.14 (Establishing Pzr Steam Bubble And RCS Final Vent) in in progress

• RCS pressure = = 40 psig with a pure steam bubble

• IGWD-13 (QUENCH TANK VENT OUTSIDE RB) is closed 1GWD-13

• 1IGWD-17 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 34 Question 34 T2/G1 - okm/cpw T2/G1 okm/cpw 007K5.02 Pressurizer Relief 007K5.02 Pressurizer Relief Tank/Quench Tank/Quench Tank Tank System System (PRTS)

(PRTS)

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

apply to PRTS:

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

(3.1/3.4 )

K/A MATCH KIA MATCH ANALYSIS Requires knowledge Requires knowledge ofof the QT QT operational parameters (pressure (pressure and level level changes) that indicate PzrPzr 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 OP/1103/002, 03/002, Pzr steam bubble formation is complete (ie, all the N2 gas is vented out of the pzr) 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, OP/1/A11103/002, End.

Enc!. 4.11 pg 1; 1; End.

Enc!. 4.14, pg 4 & 88 Proposed references to be provided to applicants duringduring examination: NONE NONE Learning Objective: PNS-PZR R17 Question Source: NEW NEW Question Question History:

History: Last Last NRC NRC Exam Exam N/A N/A Question Cognitive Cognitive Level: Memory Memory or or Fundamental Fundamental Knowledge Knowledge

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

Isolation?

A. 1XL A. 1XL /I KHU via overhead KHU via overhead path path B. 11XL B. KHU via XL /I KHU via underground underground path path 1XS1 /I KHU via overhead path C. 1XS1 path D. 1XS1 D. 1XS1 /I KHU KHU via underground underground path path

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAMEXAM Question 35 Question 35 T2/G1 - cpw T2!GI - cpw Component Cooling 008K2.02 Component roo8K2.o2 Water Cooling Water Knowledge of Knowledge of bus power supplies bus power supplies to to the following:

the following:

CCW pump, including emergency CCW pump, including emergency backup. backup.

(3.0/3.2)

(3.0/3.2)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of the the 11A CC pump A CC normal and pump normal and emergency emergency backup backup power power supplies.

supplies.

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

-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 11CC-7 CC-7 however it does not supply power to the CC pumps. Second part is correct.

IXS1 is a load center that does supply major D. Incorrect: First part is plausible since 1XS1 components including component cooling valve 1CC-7 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)

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 NEW Question History:

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

Level: Comprehens Comprehension ion or or Analysis Analysis

2010A NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 36 Question 36 Unit 33 plant Unit conditions:

plant conditions:

• • Reactor Reactor power power == 100%

100%

• • Pressurizer Pressurizer pressure pressure control malfunction has control malfunction has occurred

• • RCS RCS pressure pressure = 2000 2000 psig psig decreasing Based on Based on the the above conditions, which ONE above conditions, ONE of of the following describes describes the RCS RCS pressure pressure at which a LOW at LOW RCS PRESSURE reactor RCS PRESSURE reactor trip will occur AND the RCS RCS pressure setroint where Engineered Safeguards digital channels 11 and 2 will actuate?

setpoint The reactor will trip at ____ psig and ES digital channels 11 and 2 will actuate at

____ psig.

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

2010A NRC 2010A NRC REACTOR REACTOR OPERATOROPERATOR EXAM EXAM Question 36 Question 36 T2IGI - cpw T2/G1 - cpw 010K3.03 Pressurizer Pressure 010K3.03 Pressurizer Pressure Control Control Knowledge of Knowledge of the the effect effect that that aa loss loss or or malfunction malfunction of of the the PZR PZR PCS PCS will will have have onon the following:

the following:

ESFAS ESFAS (4.0/4.2)

K/A MATCH ANALYSIS KIA knowledge of Requires knowledge of if if the the effect effect that aa malfunction malfunction ofof the PZR PZR PCS PCS will 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 11&2

&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 IC-ES,IC-RPS Proposed references to be provided to applicants during examination: NONE Learning Objective: IC-ES IC-ES R14, lC-RPS IC-RPS R3 Question Source: NEW Question History:

History: Last NRC Exam _.....;n...;;,;.l=a nla _ __

Question Cognitive Cognitive Level: Knowledge and and Fundamentals Fundamentals

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 37 Question 37 Unit 22 plant Unit conditions:

plant conditions:

• • Reactor Reactor power power == 100%

100%

• • 2B2B RPS RPS Channel Channel Low Low RCS RCS Pressure Pressure Bistable Bistable failed failed in in "tripped" tripped state state

• • 2B 2B RPS Channel in RPS Channel Manual Bypass" in "Manual Bypass Current conditions:

Current conditions:

• • 2C2C RPS RPS Channel Channel inadvertently inadvertently placed placed in in "Shutdown Shutdown Bypass" Bypass Based on Based above conditions, on the above conditions, which ONE ONE of the following describes describes the impact impact (if (if any) any) on reactor power on reactor power andand control control room room alarms?

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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 37 Question 37 T2IGI - okm/cpw T2/G1 okm/cpw 01 2A4.03 Reactor Protection 012A4.03 Reactor Protection System System Ability to Ability to manually manually operate and/or monitor operate and/or in the monitor in the control control room:

room:

Channel blocks and bypasses.

Channel blocks and bypasses.

(3.6/3.6)

(3.6/3.6)

KIA MATCH KIA MATCH ANALYSIS Requires the ability Requires ability to to monitor plant response monitor plant response andand control control room room indications indications that that occur when placing RPS occur RPS Channels Channels inin Manual Manual Bypass Bypass and Shutdown Bypass and Shutdown Bypass ANSWER CHOICE ANALYSIS Answer: D D A. Incorrect: Both parts are incorrect. First part is plausible since there would be a B and C channels however with the B channel in Manual bistable tripped in both the Band 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 aa high RCS pressure trip set point of 1720

~1720 psig therefore the high high RCS pressure bistable will have actuated.

Technical Reference(s): IC-RPS pgs 8,18,19 Proposed references to be provided to applicants during during examination: NONE NONE Learning Objective: IC-RPS R5, R6 Question Question Source:

Source: NEW NEW Question History:

Question History: Last Last NRC NRC Exam Exam N/A N/A Question Question Cognitive Cognitive Level: Comprehension Comprehension or or Analysis Analysis

NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 38 ONE ofthe Which ONE of the following would result result in in aa trip of the 11 D D RPS RPS Channel Channel AND the 11 DD Breaker?

CRD Breaker?

A. Reactor Building pressure bistables in the 11A A and 1lBB RPS channels fail in the tripped "tripped" state B. Reactor Building pressure bistable in the 11 D RPS channel fails in the "tripped" tripped state C. Loss of 11KVID KVID D. Loss LossoflDCB of 1DCB

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 38 Question 38 T2IGI T2/G1 - cpw

- cpw 01 2K2.01 Reactor 012K2.01 Reactor Protection Protection System System Knowledge of Knowledge of bus bus power power supplies supplies to the following:

to the following:

RPS channels, components, and interconnecti RPS channels, components, and interconnections ons (3.3/3.7)

(3.3/3.7)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of the the power power supplysupply to to RPS RPS Channels Channels andand related related components (CRD components (CRD breakers) breakers)

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer:

Answer: 8 B A. Incorrect:

A. Incorrect: Plausible Plausible since since failure failure of of the the 22 RB RB pressure pressure bistables bistables will will cause cause all all 44 CRD CRD breakers to breakers to open open however however onlyonly thethe A A and and BB RPS RPS channels channels would would be be tripped.

tripped.

B. Incorrect:

B. Incorrect: Plausible Plausible since since this this failure failure would would result result in the the RPS RPS channel tripping however however the CRD breaker still requires requires 2 tripped RPS RPS channels channels to open therefore itit will remain closed will C. CORRECT:

C. CORRECT: Loss of the vital power source to a particular RPS channel will result in that entire channel de-energizing, result de-energizing, with all indicating lights off, and and the the channel channel tripped. Loss of the vital power source will also result in a trip of the the individual CRD breaker associated with that RPS channel since the 120VAC individual I2OVAC to to the breakers the breakers UV coil and shunt trip relay will be lost.

D. Incorrect:

D. Incorrect: Plausible since DCB is the normal supply to DID. If DID is lost KVID would be de-energized resulting in D D RPS channel and D D CRD breaker trip. DCB is normal normal supply supply to DID however there is an auto backup from alternate unit unit via isolating diodes.

Technical Technical Reference(s)

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

examination: NONENONE Learning Learning Objective:

Objective: lC-RPS IC-RPS R18,R18, 20 20 Question Question Source:

Source: NEWNEW Question Question History:

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

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM 1I POINT POINT Question 39 Question 39 Unit 22 initial Unit initial conditions:

conditions:

• • Reactor Reactor power power == 100%

100%

Current conditions:

Current

• MSLB occurs

• • RCS RCS pressure pressure == 1580 1580 psig slowly increasing psig slowly increasing

• • RB 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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 39 Question 39 T2IGI - cpw T2/G1 cpw 013A3.02 Engineered Safety 01 3A3.02 Engineered Safety Features Features Actuation Actuation System (ESFAS)

System (ESFAS)

Ability to Ability monitor automatic to monitor operation of automatic operation of the the ESFAS ESFAS including:

including:

Operation of Operation of actuated actuated equipment.

equipment.

(4.1/4.2)

(4.1/4.2)

K/A MATCH KIA MATCH ANALYSIS Requires knowledge Requires knowledge of of ES ES actuation actuation setpoints, what components components areare operated operated from which ES from ES digital digital channels.

channels.

ANSWER CHOICE ANALYSIS Answer: B 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 1I actuation setpoint for RCS pressure (1600 psig) ES 1I 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 1 or 2. This answer would be correct if ES channel 5 had actuated which would occur at 3 psig RB pressure.

2LPSW-1 062 does receive a closed signal from ES D. Incorrect: Plausible since 2LPSW-1062 actuation however it is not from either Channel 1 1 or 2. This answer would be correct if ES channel 6 had actuated which would occur at 3 psig RB pressure pressure....

Technical Reference(s): End. 5.1 IC-ES EOP Enc!.

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 Memory or Fundamental Knowledge

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 40 Question Unit 3 Unit plant conditions:

plant

• Reactor power == 100%

Reactor 100%

• Panelboard 3KVIA AC Vital Power Panel board supply breaker trips OPEN

• ES Analog Channel "c"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 1I thru 4 C. ONLY channels 2 AND 4 D. ONLY channels 1I AND 3

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM .

Question 40 Question 40 T2IGI - okm/cpw T2/G1 okm/cpw 013K6.01 Engineered Safety 013K6.01 Engineered Safety Features Features Actuation Actuation System System (ESFAS)

(ESFAS)

Knowledge of Knowledge of the the effect of aa loss effect of loss or or malfunction malfunction onon the the following following will will have have on on the the ESFAS:

ESFAS:

Sensors and detectors.

Sensors (2.7/3.1))

(2.7/3.1 KIA MATCH K/A MATCH ANALYSIS Requires knowledge Requires knowledge of of the effect ofof both both aa loss loss of power to a channels sensorsldetectors as well as a malfunction of a sensor/detector sensors/detectors 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)

Reference(s):: IC-ES lesson pg II, 11, TS 3.3.5 Proposed references to be provided to applicants during during examination: NONE NONE Learning Objective:

Objective: IC-ES R2, R5, TI, T1, ADM-ITS R7 Question Question Source: NEW NEW Question History:

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

Level: Comprehens Comprehension or Analysis ion or Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT 1

Question 41 Unit 1iplant plant conditions:

• Time = 03:00

• Reactor power = = 100%

• 1 lB and 1IC Band C RBCUs operating in HIGH speed

• 1 1A 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?

lB and 1C A. 1B IC RBCUs RBCU5 operating in HIGH speed and 1AIA RBCU OFF I Band B. 1 B and C RBCUs operating in LOW speed and 1A IA RBCU OFF C. ALL RBCUs operating in LOW speed D. ALL RBCUs will be OFF

2010A NRC REACTOR OPERATOR EXAM Question 41 T2IGI --okm T2/G1 okm 022A3.O1, Containment Cooling System (CCS) 022A3.01, 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 RBCUs this would be the correct answer if you understand that the 3 position of the RBCU's minute time delay is when the RBCU's 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 RBCU5 will start at LOW speed. This choice is plausible delay is finished then all 3 RBCUs if you are not aware of the 3 minute time delay or believe it is less than 11 minute.

D. CORRECT: When ES actuates all operating RBCU's RBCUs will stop and a 3-minute time delay is in effect. Once the time delay is finished then all 3 RBCUs 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 PNS150501-Question History: Last NRC Exam NIA Question Cognitive Level: Comprehension or Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 42 Question 42 Which ONE of the following describes the rangerange of BWST BWST levels levels where RBS RBS pump pump suction would bebe aligned to both both the RBES RBES andand the BWST BWST simultaneously AND what action(s) would be required if 1I LP-22 failed to close when isolating 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 _ _ _ __

159 A. 15 - 9 /I stop the 1lB B LPI pump AND 1lB B RBS pump 159 B. 15 - 9 /I Maximize total LPI flow << 3100 gpm C. 99-6/stopthelBLPlpumpANDlBRBSpump

- 6 / stop the 1B LPI pump AND 1B RBS pump D. 9 - 6 /I Maximize total LPI flow < 3100 gpm

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 42 T2IGI - cpw T2/G1 - cpw 026A11.03, 026A .03, Containment Spray System (CSS) (CSS) predict and/or monitor Ability to predict monitor changes changes in in parameters parameters (to (to prevent prevent exceeding design limits) limits) associated with operating the CSS controls including:

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 aUgn suction to both sources. Second part is plausible since these actions are correct if only 1 I LPI pump is operating when isolating BWST. .

C. CORRECT: At 9' 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' 6 the BWST is isolated by closing LP-21 and LP-22. If 1 I LP-22 fails to close, the 1lB B LPI pump AND 1lB B RBS pump are secured until 1 LP-28 is ILP-28 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 11LP LP-20 fails to open at 9.9'. Second part is plausible since these actions are correct if only I1 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 duringduring examination: NONE NONE Learning Objective: EAP-LOSCM R34, 36 Question Source: NEW NEW Question History: Last NRC NRC Exam N/A N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAMEXAM 1I POINT POINT Question 43 Question 43 Unit initial conditions:

Unit 33 initial conditions:

• • Reactor power == 35%

Reactor power 35% slowly increasing slowly increasing Current conditions:

Current Reactor power = 30%

• Reactor 30% decreasing decreasing

• • PCBPCB 58 and PCB-59 58 and PCB-59 (Unit Generator Output (Unit 33 Generator Output Bkrs)

Bkrs) OPEN OPEN

• Turbine mastermaster in HAND in HAND

• • OAC point 03X2060O3X2060 (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)?

(TBVs)?

____ is being compared to Turbine Header Pressure setpoint to develop the controlling error signal AND TBV'sTBVs 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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 43 Question 43 T2/G1 T2/G1 - cpw

- cpw 039G2.1.19, Main 039G2.1.19, Main and Reheat Steam and Reheat Steam Ability to Ability to use use plant plant computers computers to to evaluate evaluate system system oror component component status.status.

(3.9/3.8)

(3.9/3.8)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requires utilizing Requires utilizing OAC indication for OAC indication for Turbine Turbine Load Load Status Status Flag Flag (TLSF)

(TLSF) to to determine the setpoint determine the setpoint at which at which Main Main and and Reheat Reheat Steam Steam pressure pressure isis being being controlled controlled ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: C C A. Incorrect:

Incorrect: First First part is plausible since since it would be be correct ifif the Turbine Master Master were in Automatic. Second in Second part part is is correct.

correct.

First part is plausible since it would be correct if the Turbine Master B. Incorrect: First 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 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 Technical Reference(s)

Reference(s):: STG-ICS STG-ICS Chapter Chapter 33 Proposed Proposed references references to be provided to be provided to to applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: STG-ICS STG-ICS RIO R10 Question Question Source:

Source: NEW NEW Question Question History:

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

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 44 Question 44 Unit 11 initial Unit initial conditions:

conditions:

• • Reactor Reactor power power == 70%70% stable stable Current conditions:

Current conditions:

• 1HPE-6

• 1HPE-6 (Heater (Heater 1A 1AI1 Bleed Bleed Inlet)

Inlet) closed closed Based on Based on the above conditions, the above conditions, which which ONE ONE of of the the following following predicts predicts the the impact impact of of the the malfunction on malfunction Feedwater flow on Feedwater flow assuming assuming no no operator operator action action AND AND the the procedure which procedure which will be will be used used to to reopen reopen 11HPE-6?

HPE-6?

Feedwater flow Feedwater flow will will stabilize stabilize atat aa _ _ _ _ _ value value than than the the pre-transient pre-transient level level AND AND

_ _ _ will will be be used used to reopen 1IHPE-6.

to reopen HPE-6.

A. higher A. higher 1/ OP/1/A/1106/23 OP/11A11106/23 (High (High and and Low Low Pressure Pressure Extraction)

Extraction)

B. higher B. higher 1/ OP/1/A/1106/002 OPI1/A111061002 (Condensate and FDW system) system)

C. lower 1/ OP/1/A/1106/23 OPI1IAI1IO6/23 (High and Low Pressure Extraction)

D. lower D. lower 1/ OP/1/A/1106/002 OP/I/A/I 106/002 (Condensate and FDW system)

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 44 Question T2IGI --cpw T2/G1 cpw 059A2.06, Main Main Feedwater Feedwater (MFW)

(MFW) System System Ability to (a) predict the impacts (a) predict impacts of the following malfunctions malfunctions or operations operations on the MFW; MFW; and (b) based on those predictions, predictions, use procedures to correct, control, use procedures or mitigate the consequences of those malfunctions or operations:

Loss of steam flow to MFW MEW system.

(2.7/2.9)

K/A MATCH ANALYSIS KIA 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 lOS response. Initially, CTP will A. Incorrect: First part is plausible due to the initial ICS begin to increase which would generally indicate an increase in FDW EDW flow is required however ICS will attempt to maintain CTP at setpoint and will therefore decrease the demand to the remaining portions of ICS lOS which will then actually decrease FDW EDW lOS is to maintain the primary and secondary heat flow. Since the big picture of ICS 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 FDW OP/1/A/11061002 EDW system) is the procedure used to control most FDWEDW 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, 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/A11106/002 OP/1/A/11 06/002 (Condensate and EDW 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): OPIIIA/1106123, OP/1/A11106/23, Enclosure Enclosure 4.1., Re-opening Extraction Valves) AP128AP/28 (ICS Instrument Instrument Failure) STG-ICS STG-FHS STG-FHS Proposed references to be provided to applicants duringduring examination: NONE NONE Learning Objective: STG-FHS R9, 23 STG-ICS R30, 14 14 Question Source: NEW NEW Question History: Last NRC NRC Exam N/A Question Cognitive Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT 1

Question 45 Unit 1 1 initial conditions:

• Reactor power = 100%

Current conditions:

• Reactor trip

• 1 IFDW-33 FDW-33 (1A (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 1B lB SG level =

= _ __

A 25 S/U / 25" A. 25" 25 S/U B. 12 S/U / 25" B 12" 25 S/U C.

C 30" 30 XSUR / 25" 25 S/U 30XSUR D. 30" XSUR / 30" 30XSUR XSUR

2010A NRC REACTOR OPERATOR EXAM Question 45 T2IGI - bank T2/G1 -

059K3.02, Main Feedwater (MFW)(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 KIA Requires knowledge of the effect that a malfunction of a FDW block valve will have on EFDW system ANSWER CHOICE ANALYSIS Answer: D 25 is the normal post trip SG level in both SG's A. Incorrect: Plausible since 25" SCs when on Main FDW. Since the S/U control valve has not failed it is plausible to deduce the

25. It is also plausible to believe that S/U valve could still control at 25". that the main FDW EDW control valve would control at 25" 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 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"21 in BOTH SG's SCs to actuate EFDW on dryout protection.

C. Incorrect: Plausible since the failure is on the 1A IA SG SC only. Failure to realize that MDEFWPs will start if EITHER SG BOTH MDEFWP's 21 for 30 seconds could lead to SC reaches 21" this choice.

D. CORRECT: With the SU block valve failed closed the SU control valve cannot supply FDW to the 1A IA SG. SG level will decrease until <21" <21 for 30 seconds MDEFWPs. With both MDEFWP's which will start BOTH MDEFWP's. MDEFWPs operating, FDW-315 and 316 will control at 30" 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 N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT 1

Question 46 Unit 1 1 initial conditions:

• Reactor power = = 100%

• .. Unit 1 I TDEFWP unavailable Current conditions:

• Both Main FDW pumps trip

• 1 lBMDEFWPfailstostart 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...

Initiate ...

A. Rule 3 (Loss of Main or Emergency Feedwater) and cross connect with an alternate unit to supply the 1 1B 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. End. 5.9 (Extended EFDW Operation) and feed both SG's SGs with 1A MDEFWP

2010A NRC REACTOR OPERATOR EXAM Question 46 T2IGI -

T2/G1 -

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 KIA Requires knowledge of how AFW components are utilized based on loss of MFDWPs, TDEFWP, and one MDEFWP MFDWP's, ANSWER CHOICE ANALYSIS Answer: 0D 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 A MDEFWP had also been lost.

correct if the 1IA C. Incorrect: Plausible since HPI FC is utilized as a strategy in RULE 4 and would be correct if the 1 1A A 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 Encl.

D. End. 5.9 which will direct opening FDW-313 & 314 and feeding both SG's 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

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 47 Question 47 Which ONE Which ONE ofof the the following following describes describes actions actions required required inin that that will will extend extend the the life life of of the the Control Batteries Control Batteries following following aa loss loss of of all all AC power inin accordance AC power accordance with with EOP EOP Enclosure Enclosure 5.38 (Restoration of 5.38 (Restoration of Power)?

Power)?

Load Shed Load Shed the the inverter.

inverter.

A.Kl A. KI B. DIA B.DIA C. KSF-1 C. KSF-1 D. KOAC D. KOAC

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 47 Question 47 T2IGI T2/G1 - cpw

- cpw 062K3.03, AC 062K3.03, Electrical Distribution AC Electrical Distribution System System Knowledge of Knowledge of the effect that the effect that aa loss loss or or malfunction malfunction ofof the the ac ac distribution distribution system system will have on the following:

will have on the following:

system.

DC system.

DC (3.7/3.9)

(3.7/3.9)

KIA MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of the the effect effect of of aa loss loss of of AC AC power power will will have have onon DC DC systems systems and actions required subsequent and actions required subsequent to to the the loss loss of of AC AC that that will will impact impact DCDC system system availability availability ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: A A. CORRECT: If a blackout exists on all 3 units, Encl. End. 5.38 directs performing End. 5.32 which will open the DC input breaker to the KI inverter.

Enc!.

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)

Reference(s):: EL-DCD Proposed Proposed references to be provided to be provided toto applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: EL-DCD EL-DCD RI R1 Question Question Source:

Source: NEWNEW Question Question History:

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

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 48 Question 48 Which ONE Which ONE ofof the the following following describes describes the the normal normal alignment alignment of of the the Power Power Battery Battery busses AND busses AND aa condition condition inin which which SLC SLC 16.8.3 16.8.3 (Power (Power Battery Battery Parameters)

Parameters) would would require require changing that alignment?

changing that alignment?

The Oconee The Oconee units units are are normally normally ____ and and this this would would be be changed changed ifif _ _ __

A. cross-tied A. cross-tied // aa single single power power battery battery becomes becomes inoperable inoperable B. cross-tied B. cross-tied / two or or more more power power batteries batteries become become inoperable inoperable C. separated // aa single C. separated single power power battery battery becomes becomes inoperable inoperable D. separated / two or more power batteries become inoperable

2010A NRC 2010A NRC REACTOR REACTOROPERATOR OPERATOREXAM EXAM Question 48 Question 48 T2/G1 T2/G1 - cpw

- cpw 063K4.02, D.C.

063K4.02, D.C. Electrical Electrical Distribution Distribution Knowledge Knowledge of of DC DC electrical electrical system system design design feature(s) feature(s) andl and! oror interlock(s) interlock(s) which which provide for the following:

provide for the following:

Breaker interlocks, Breaker interlocks, permissives, permissives, bypasses bypasses andand cross-ties.

cross-ties.

(2.9/3.2)

(2.9/3.2)

KIA MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of the the normal normal design design status status ofof DC DC power power battery battery busses busses as as well as the design feature well as the design feature which allowswhich allows cross-tie cross-tie between between units.

units.

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer:

Answer: C C A. Incorrect:

A. Incorrect: First First part part is incorrect but is incorrect but plausible plausible since since the the normal normal alignment alignment of of the the Control Batteries Control Batteries is is with the units cross-connec the units cross-connected ted via via the the isolating isolating diodes.

diodes. Second Second part is part is correct.

correct.

B. Incorrect:

B. Incorrect: First First part part is is incorrect incorrect but plausible plausible since the normal normal alignment of the Control Batteries is with the units cross-connected Control cross-connected via the isolating diodes.

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 batteries corrective actions in that case do not include cross-tieing the busses.

C. CORRECT:

C. CORRECT: Normal alignment alignment of the power battery buses is with the units separated (EL-DCD page 35). SLC 16.8.3 Condition 8B is for a single power separated battery inoperable inoperable and requiresrequires initiating actions to cross-tie cross-tie busses immediately.

immediately.

D.

D. Incorrect:

Incorrect: First part is correct. Second part is plausible since SLC 16.8.3 16.8.3 Condition CC is is for having 2 2 or more batteries inoperable and does have immediate corrective actions actions however however the corrective actions in that that case case dodo not not include cross-tieing the busses.

busses.

Technical Technical Reference(s)

Reference(s):: EL-DCD EL-DCD SLC SLC 16.8.3 16.8.3 Proposed Proposed references references to to be be provided provided toto applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: EL EL DCD DCD R7, R7, ADM-TSS ADM-TSS R4 R4 Question Question Source:

Source: NEW NEW Question Question History:

History: Last Last NRC NRC ExamExam N!A N/A Question Question Cognitive Cognitive Level:

Level: Knowledge Knowledge and and Fundam Fundamentals entals

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 49 Question 49 Operators are Operators are preparing preparing to to synchronize synchronize KHU-2 KHU-2 to to the the grid grid inin accordance accordance with with OP/0/A/1 106/019, (Keowee OP/O/A/1106/019, (Keowee HydroHydro At At Oconee)

Oconee)

The operator The operator notes notes the the following following indications:

indications:

• Grid Frequency 59.9

• Grid Frequency = 59.9 cycles

= cycles

• • Keowee Keowee Frequency Frequency == 60.360.3 cycles cycles

• Keowee 22 Line

• Keowee Line Volts Volts == 13.7 13.7 kV kV

• Keowee 2 Output Volts

• Keowee 2 Output Volts = 15.2 = 15.2 kV kV Based on Based on the the above above conditions, conditions, which which ONE ONE ofof the the following following describes describes the the control control that that will be will be used used to to adjust adjust the the synchroscope synchroscope indication indication and and the the response response when ACB-2 when ACB-2 is is closed?

closed?

The The will be will be used used to adjust adjust the synchroscope synchroscope indication and _ _ __

indication and A. UNIT A. UNIT 22 AUTO VOLTAGE ADJUSTER I/ ACB-2 will immediately 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 B. UNIT as aa direct result of the line voltage differential direct C. UNIT C. UNIT 2 AUTO VOLTAGE ADJUSTER I ACB-2 will NOT receive a trip signal as as aa direct direct result of the line voltage differential.

D. UNIT D. UNIT 2 SPEED CHANGER MOTOR I/ ACB-2 will immediately receive a trip signal Signal as aa direct result of the line voltage differential

2010A NRC REACTOR OPERATOR EXAM Question 49 T2IGI -

T2/G1 -

064A11.03, 064A .03, Emergency Diesel Generators (ED/G)

Ability to predict and/or monitor changes in parameters (to (to prevent exceeding design limits) associated with operating the ED/G system controls including:

Operating voltages, currents, and temperatures (use Hydro units ifif possible).

(3.1/3.4)

K/A MATCH ANALYSIS KIA Per NRC OK not to address temperatures. Requires monitoring parameters and predicting response when operating ED/G 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 QOT protection circuit is misapplied (only for Emerg Starts) in which plausible if the OOT 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.

0P101A111061019 Rev 83 Technical Reference(s): EL-KHG, OP/0/Al1106/019 Proposed references to be provided to applicants during examination: None Learning Objective: EL-KHG Ru, R11, R4, R20, R19, R7 Question Source: BANK ELO4IIIO 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 NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 50 Question 50 Unit 11 conditions:

Unit Time=

Time = 1159:40

• Reactor power = = 100% stable KHU-1 OOS

• KHU-100S

• ACB-4 closed

• KHU-2 gets Emergency Start signal from another unit Time = 1200:00

• KH KHU-2 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 11 (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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Energizing CTCT-4-4 /I No additional actions required D. Energizing CT-4 I/ Initiate APII AP/11I (Recovery From Loss Of Power)

2010A NRC REACTOR OPERATOR EXAM Question 50 T2IGI --cpw T2/G1 cpw 064A2.02, Emergency Diesel Generators (ED/G) (ED/C)

Ability to (a) predict the impacts of the following malfunctions or operations on EDIG system; and (b) based on those predictions, use procedures to correct, the ED/G 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 KIA 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.) 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.

KHUs, If an Emergency Start is present and the unit overspeeds B. CORRECT: For KHU's,

(>180 (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 OOS then both the overhead and underground KHU's KHU-1 is already 005 KHUs are inoperable and TS 3.8.1 Condition J requires energizing both SBB's SBBs within 1 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 AP/11I (Recovery from Loss of Power) is entered when 4160V buses lose power and it is subsequently regained. It It, would be correct if your unit were counting on the KHUs KHU's to energize your MFB since that would indicate that 4160V had lost power however in this case although the KHUs are inoperable, Unit 1 KHU's 1 has not had aa 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

NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 51 Question 51 Unit 11 initial Unit initial conditions:

• TimeTime=1200

= 1200

• Reactor power = 35%

• 11A A steam generator tube leak = 2.1 gpd stable

• RCS activity = = 0.25 ~Cilml pCi/mI DEI DEl increasing Current conditions:

• TimeTime=1400

= 1400

• NO change in 1IA A SG tube leak rate

• RCS activity = = 0.65 ~Cilml pCi/mI DEI 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. 11RIA-16 RIA-16 (Main Steam Line Monitor) and 11RIA-40 RIA-40 (CSAE Off-gas) increased.

B. 11RIA-16 RIA-16 (Main Steam Line Monitor) increased while 11RIA-40 RIA-40 (CSAE Off-gas) remained constant.

C. 1IRIA-59 RIA-59 (N-16 monitor) and 1IRIA-40 RIA-40 (CSAE Off-gas) increased.

D. 11RIA-59 whilelRlA-40 (CSAE Off-gas) remained constant.

RIA-59 (N-16 monitor) increased while1RIA-40

2010A NRC REACTOR OPERATOR EXAM Question 51 T2/G1 --CPW 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 KIA 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 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 RAD010207 Question History: Last NRC Exam ONS 2006 Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM 1

I POINT Question 52 Unit 1 1 initial conditions:

• Time =1200 Time=1200

• Reactor power = 100%

• A and B LPSW Pumps operating

• C LPSW pump in AUTO

• Blackout Occurs Current conditions:

Time=1230

• Time = 1230

• Both MFB'sMFBs 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 MFBs 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 ititto OP/I /A/1104/010 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 T2/G1 -

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, predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Service water header pressure.

(2.7/3.1))

(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 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. AP/24 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 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 4L N/A Question Cognitive Level: Comprehension and Analysis

2010A NRC REACTOR OPERATOR EXAM EXAM POINT 1I POINT Question 53 Initial Initial plant conditions:

• Large IA leak occurs

• Service air header pressure == 87 psig decreasing

• • Turbine Building air header pressure per gage below TUBB TURB BLDG BLDG 160 160 =

140 140 =

120 120 =

100 100

.~=

=

=

80 -

60 -

40 20 PSI PSI o0 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 53 Question 53 T2IGI - cpw T2/G1 - cpw 078A4.O1, Instrument Air System 078A4.01, Instrument System (lAS)

(lAS)

Ability to Ability to manually manually operate operate and/or andlor monitor monitor in in the the control control room:

room:

Pressure gauges.

Pressure gauges.

(3.1/3.1)

(3.1/3.1)

K/A MATCH ANALYSIS KIA Requires demonstrating Requires demonstrating the ability ability to to monitor monitor anan IA IA pressure pressure gage gage in in the the control control room and room and based based on the indication indication determine determine IAIA 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 D.

IA compressors would start at 93 psig IA pressure Technical Reference(s):

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 NRC Exam N/A N/A Question Cognitive Level: Comprehens Comprehension ion and Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 54 Question 54 Unit 22 conditions:

Unit conditions:

• • Reactor Reactor power power == 100%

100%

• • Letdown Letdown temperature temperature= =112°F 112°F stable stable

• 2HP-5 failed

• 2HP-5 failed closedclosed Based on Based on the the above above conditions, conditions, which which ONEONE of of the the following following describes describes requirements requirements for for manually opening manually opening 2HP-5 2HP-5 locally locally in in accordance accordance with with AP/32 AP/32 (Loss (Loss ofof Letdown)

Letdown) AND AND the the MINIMUM Pressurizer MINIMUM Pressurizer level level (inches)

(inches) at which aa manual at which manual reactor reactor trip trip would would be be required?

required?

A. Maintain Maintain constant constant communication communication with operatoroperator dispatched dispatched to to open open 2HP-5 2HP-5 // 400 400 B. Maintain constant communication communication with operator dispatched to open 2HP-5 2HP-5 / 380 380 C. Place the LETDOWN HI TEMP INTLK INTLK BYP switch to BYPASS / 400 D. Place the LETDOWN HI TEMP INTLK BYP switch to BYPASS / 380

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 54 Question 54 T2IGI - CPW T2/G1 - CPW 103G2.1.20, Containment System 103G2.1.20, Containment System Ability to Ability interpret and to interpret and execute execute procedure procedure steps.

steps.

(4.6/4.6)

(4.6/4.6)

K/A MATCH KIA MATCH ANALYSIS Question requires knowledge of procedure steps regarding opening Question requires opening a containment isolation valve (HP-5) that that has failed closed and the requirements to properly execute properly steps.

execute the steps.

ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: First part is correct. Second part is plausible since 400" 400 is the maximum (0-400). It is plausible to believe the required Rx trip measurable pressurizer level (0"-400").

would be at the point that pressurizer level goes off scale high however 375" 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 AP132 directs that at 375" 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 under those directions. Second part is plausible since 400 400" is the maximum measurable pressurizer level (0"-400").

(0-400). It is plausible to believe the required Rx trip would be at the point that pressurizer level goes off scale high however 375375" is used to account for instrument errors.

D. Incorrect::

Incorrect: : First part is plausible since these directions are in HP-32 for opening HP-55 only it would be correct if HP-S 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): AP132 AP/32 (Loss of Letdown)

Proposed Proposed references to be provided to applicants during NONE during examination: NONE Learning Objective: EAP-APG R8 Question Source:

Source: NEWNEW Question History:

Question History: Last Last NRC NRC Exam Exam N/AN/A Question Cognitive Question Cognitive Level: Comprehension Comprehension and and Analysis Analysis

2010A NRC REACTOR OPERATOR EXAM I POINT 1

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 3CC-8s response to an ES 1-6 actuation?

can be operated from the control room and 3CC-8's 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 T2/G1 CPW 103G2.4.20, Containment System 103G2.4.20, Knowledge of the operational implications of EOP warnings, cautions, and notes.

(3.8/4.3)

K/A MATCH ANALYSIS KIA 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 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 AP/20 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 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 NIA N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM 1I POINT POINT Question 56 Question 56 Unit 11 initial Unit conditions:

initial conditions:

• • 100% power 100% power

• • 450 EFPD 450 EFPD Based on Based on the the above above conditions, conditions, which which ONE ONE ofof the the following following events events is is the the cause cause of of the the indications on the attached P/T indications on the attached PIT display?display?

SEE ATTACHMENT SEE ATTACHMENT ASSUME NO NO OPERATOR ACTIONS ACTIONS A. SBLOCA with a Reactor Reactor Trip on variable low pressure B. LBLOCA B. LBLOCA with a ReactorReactor Trip on low low RCS RCS pressure 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 56 Question 56 T21G2 - CPW T2/G2 - CPW Reactor Coolant 002A4.08, Reactor roo2A4.o8, Coolant System System (RCS)

(RCS)

Ability to Ability manually operate to manually operate andlor and/or monitor monitor in in the the control control room:

room:

Safety parameter display systems.

Safety parameter display systems.

(3.4/3.7)

(3.4/3.7)

KIA MATCH KIA MATCH ANALYSIS Per NRC, NRC, question question using using OAC P/T display OAC PIT display vs vs Westinghouse Westinghouse SPDS SPDS display display info info is is OK.

OK.

Requires ability to monitor Requires monitor OAC OAC PIT PIT display and interprete interprete RCS 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"A SG pressure is -370 370 psi with indications of severe overcooling.

D. Incorrect: The display indicates overcooling. A loss of MFWPs would be indicated as under cooling (Temps increasing).

SF-OlD (PTID), EAP-EHT Technical Reference(s): SF-010 Proposed references to be provided to applicants during examination: NONE SF-0l0 R9, EAP-EHT R14 Learning Objective: SF-010 Question Source: NEW Question History: Last NRC Exam N/A NIA Question Cognitive Level: Comprehension and Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 57 Question 57 Which ONE Which ONE ofof the the following following describes describes the the power power supply supply for for the the Unit Unit 11 Group Group BB Pressurizer heaters?

Pressurizer heaters?

A. 1XH A. 1XH B. 1XI B. 1XI 1XJ C. 1XJ C.

1XSF D. 1XSF

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAMEXAM Question 57 Question 57 T21G2 - CPW T2/G2 - CPW 011 K2.02, 011 K2.02, Pressurizer Level Control Pressurizer Level Control System System Knowledge of Knowledge of bus power supplies bus power supplies to to the following:

the following:

PZR heaters.

PZR heaters.

(3.1/3.2)

(3.1/3.2)

K/A MATCH KIA MATCH ANALYSIS Question requires knowledge Question requires knowledge ofof power power supply supply to Group Group 8B Pressurizer Pressurizer heaters heaters as as well as the Operators ability to control Group Group 8B heaters heaters ANSWER CHOICE ANALYSIS Answer: A A. CORRECT: The normal supply for Group 8B is 1XSF. IXSF.

B. Incorrect: Plausible since 1IXH 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 0D 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 N/A Question Cognitive Level: Knowledge and Fundamentals

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAMEXAM 1I POINT POINT Question 58 Question Which ONE of the following describes who determines that a RB RB Continuous Release Release is is allowed and after itit is started what are the requirements requirements for sampling the RB RB atmosphere inin accordance with OP/1102/014 0P11102/014 (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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAMEXAM Question 58 Question T21G2 - cpw T2/G2 - cpw 029A2.04, Containment Purge Purge System (CPS)

(a) predict Ability to (a) predict the impacts impacts of of the following malfunctions malfunctions or operations operations on the Containment PurgePurge System; and (b) (b) based on predictions, use 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*)

(2.5*/3.2*)

K/A MATCH ANALYSIS KIA 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 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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 59 Question 59 conditions:

Plant conditions:

Plant

• Spent Fuel Storage Spent Fuel Storage Cask Cask hashas been been dropped dropped in in Unit Unit 11&2 SEP

&2 SFP

• • Spent Fuel damage Spent Fuel damage is is visible visible

• • RIA-6 RIA-6 and and RIA-41 HIGH alarm RIA-41 HIGH alarm actuates actuates

• • Spent Spent Fuel Fuel Pool Pool level level == -3.5 feet decreasing

-3.5 feet decreasing Based on Based on the the above above conditions, conditions, which which ONE ONE of of the the following following describes describes which which filters filters will will be used to be used to reduce reduce off off site releases and site releases and the the status status of of the the SF SE Pumps?

Pumps?

Unit Unit Reactor Building Reactor Building Purge Purge filters and and the Spent Spent Fuel Fuel Cooling Cooling pumps pumps will bebe A.11 //ON A. ON B. 11 / OFF C.22 /ION C. ON D. 2 /I OFF

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 59 Question 59 T21G2 T2/G2 - cpw

- cpw 033A3.02, 033A3.02, Spent Fuel Pool Spent Fuel Pool Cooling Cooling System System (SFPCS)

(SFPCS)

Ability to Ability monitor automatic to monitor automatic operation operation ofof the the Spent Fuel Pool Spent Fuel Pool Cooling Cooling System System including:

including:

Spent fuel leak Spent leak or rupture.

rupture.

(2.9/3.1))

(2.9/3.1 K/A MATCH KIA MATCH ANALYSIS Knowledge of automatic Knowledge automatic operation operation of the SF Cooling Cooling pumps on a decreasing 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 1I 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 2s filters Reactor Building Purge filters are used. Second part is plausible since 4' 4 is the level at which SF Pumps loose suction and level is still >

> 4 feet.

B. Incorrect: First part is Plausible since Unit 11 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 2s filters Reactor Building Purge filters are used. Second part is correct C. Incorrect: First part is correct. Second part is plausible since 4' 4 is the level at which SF Pumps loose suction and level is still > 4 feet.

still>

0. CORRECT: Unit 1 D. 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 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)

Reference(s):: RAD-RIA, FH-SFC, FH-FES Proposed references to be provided to applicants during during examination: NONE NONE Learning Objective: RAD-RIA R2 FH-FES R2, and FH-SFC R3 Question Question Source: NEW NEW Question History:

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

Level: Knowledge and and Fundamental Fundamentals s

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM 1I POINT POINT Question 60 Question 60 Unit 22 initial Unit initial conditions:

conditions:

• • Reactor Reactor power power == 100%

100%

Current conditions:

• Reactor trip

• • Controlling Controlling 2A Steam Generator Outlet Pressure Steam Generator Pressure fails HIGH HIGH Based on the above conditions, which ONE of the following describes the 2A AND 2B Turbine Bypass Valves (TBV's)(TBVs) response?

TBVs will fully open _ _ AND the 2B TBV's The 2A TBV's 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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 60 Question 60 T21G2 T2/G2 - cpw- cpw 041 K6.03, 041 K6.03, Steam Steam Oump Dump System System (SOS)(SDS) andand Turbine Turbine Bypass Bypass Control Control Knowledge of Knowledge of the the effect effect ofof aa loss loss oror malfunction malfunction on on the the following following willwill have have on on thethe SOS:

505:

Controller and Controller positioners, including and positioners, including ICS, SIG, CROS.

ICS, S/G, CRDS.

(2.7/2.9)

(2.7/2.9)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of signal signal inputs inputs toto Turbine Turbine Bypass Bypass valve valve controls controls andand the the effect effect that a failed open controller that a failed open controller will will have have on Turbine Bypass on Turbine Bypass Control Control ANSWER CHOICE CHOICE ANALYSIS Answer: C Answer: C A. Incorrect:

Incorrect: First First part is plausible since itit would be be correct correct for normal normal post post trip response without the instrument response instrument failure. Additionally plausible plausible since since the misconception that the TBV's misconception 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 TBV's 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 TBV's misconception 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 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 BB TBVs TBV's instead of the A TBVs TBV's OR if you are under the misconceptio misconception n that both sets of TBVs TBV's control from the same SG Outlet Pressure as is the case pre-trip when using using Turbine header pressure as the controlling signal.

C. CORRECT: When the reactor trips, both sets of TBVs TBV's would normallynormally go full open in open in an an attempt attempt to to relieve enough enough steam steam from the the SGs sG's to to gain gain control of of SG sG Outlet pressure. Shortly after the trip (generally less less than aa minute) minute) both both sets of sets of TBVs TBV's willwill be be back back to to the the throttled throttled position position and and inin control control ofof SG sG Outlet Outlet Pressure. With the controlling pressure Pressure. With the controlling pressure for the the A TBV TBV failed high, high, the the AA TBV TBV would would remain remain fullfull open open since since itit will will always always believe believe that that actual actual pressure pressure is is greater greater thanthan setpoint.

setpoint.

D. Incorrect:

O. Incorrect: First First part part is is correct. Second part correct. Second part is is plausible plausible since since itit would would be be correct correct ifif you mistakenly you mistakenly applied applied thethe failed failed instrument instrument to to the the BB TBVs TBV's instead instead of of the the AA TBVs TBV's OR if you are OR if you are underunder thethe misconceptio misconception n that that both both sets sets of of TBVs TBV's control control from from the the same same SG SG Outlet Outlet Pressure Pressure as as isis the the case case pre-trip pre-trip when when using using Turbine Turbine header header pressure pressure as as the controlling signal.

the controlling signal.

Technical Technical Reference(s)

Reference(s):: STG-ICS sTG-ICs Chapter Chapter 33 Proposed Proposed references references to to be be provided provided to to applicants applicants during during examination:

examination: NONE NONE Learning Objective: STG-ICS Learning Objective: sTG-ICs R10 RIO Question Question Source:

Source: NEW NEW Question Question History:

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

Level: Comprehens Comprehension ion and and Analysis Analysis

2010A NRC REACTOR OPERATOR EXAM 11 POINT Question 61 Unit 3 initial conditions:

• Reactor power = = 100%

• 3MS-112 & 3MS-173 (SSRH 3A/3B Controls) are in MANUAL

• SSRHs) are in AUTO 3MS-77, 78, 80, 81 (MS to SSRH's)

Current conditions:

• Main Turbine trips Based on the above conditions, which ONE of the following describes the plant response?

3M5-1 12 & 3MS-173 valve demand will throttle back with load A. 3MS-112 3MS-1 12 & 3MS-173 valve demand will remain full open but the air will be ported off, B. 3MS-112 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 61 Question 61 T21G2 T2/G2 - cpw

- cpw 045K3.O1, Main 045K3.01, Main Turbine Turbine Generator Generator (MT/G)

(MT/G) System System Knowledge of Knowledge of the the effect effect that that aa loss loss or malfunction of or malfunction the MT/G of the MT/G system system will will have have on the following:

on the following:

Remainder of Remainder of the plant.

the plant.

(2.9/3.2)

(2.9/3.2)

KIA MATCH ANALYSIS K/A MATCH ANALYSIS Requires knowledge Requires knowledge of of the the effect effect ofof aa Turbine Turbine Trip Trip on on other other components components in in the the plant plant ANSWER CHOICE ANALYSIS ANALYSIS Answer: D Answer: D INCORRECT: 3MS-112 A. INCORRECT: 3MS-1 12 & 173 173 will close in manual when the Main Turbine trips.

Plausible because Plausible because they will throttle back in a "runback runback condition".

condition.

lncorrect:The signal from the Moore Controllers will direct the valves to close.

B. IncorrectThe Plausible because this and the air being bled off of the valves causes them to close.

Incorrect: 3MS-77, 78, 80 and 81 are Motor Operated Valves. Plausible because C. Incorrect 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 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.

Reference(s): STG-MSR Technical Reference(s):

Proposed references to be provided to applicants during examination: NONE Learning Objective: STG-MSR R18 Question Source: Bank Question History:

History: Last NRCNRC Exam 2007 2007 NRC NRC exam exam Q #38 Question Cognitive Cognitive Level: Comprehens Comprehension ion and and Analysis Analysis

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM POINT 1I POINT Question 62 Question 62 Unit 11 plant Unit plant conditions:

conditions:

• • OP/1/A OP/I/A 1106/002A I 106/002A (Condensate (Condensate And And FDW FDW System System Startup Startup And And Shutdown)

Shutdown) Encl.

End.

4.2 (Condensate 4.2 (Condensate And And FDW FDW System System Startup)

Startup) in progress in progress

• • 11A Hotwell pump A Hotwell pump (HWP)

(HWP) operating operating

• • CBP CBP suction suction pressurepressure == 45 45 psig psig stable stable

• The 3

• The 3 square amber lights square amber lights located located above above the the HWP HWP PUMP PUMP AMP AMP gages gages are are "ON" ON

• • ProcedureProcedure directs directs placing placing aa standby standby Hotwell Hotwell pump pump to AUTO to "AUTO" Based on Based on the the above above conditions, conditions, which which ONEONE ofof the the following following describes describes what what the the Hotwell Hotwell pump pump amber amber lights lights indicate indicate AND AND the the 1I BB Hotwell Hotwell pump pump response response once once it's its control control switch switch is placed is placed in AUTO? in AUTO?

11A Hotwell pump A Hotwell pump is is operating operating with with _ _ and and the the 1lB B Hotwell Hotwell pump pump __ automatically automatically start start when its control switch is placed in 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 WI' lA iA4WP fwp jH '"It"'!' 1~ +--

i i

_I"

- D 200_ -

200

~ ~ 150~ 150_

=I'~. 100=--

',' 100-  ;

~

~~":.50 ~::-,""

n = ',0 =

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 62 Question 62 T21G2 T2/G2 - cpw

- cpw 056G2.2.44, Condensate 056G2.2.44, Condensate System System Ability to Ability to interpret control room interpret control room indications indications to verify the to verify the status status and and operation operation of of aa system, and system, and understand understand how how operator operator actions actions andand directives directives affect affect plant and plant and

.system conditions.

(4.2/4.4)

(4.2/4.4)

MATCH ANALYSIS K/A MATCH KIA Question requires interpreting Question requires interpreting light light indications indications for HWPHWP discharge discharge pressure as well as understanding understanding how placing HWP switch to auto per procedure would plant conditions affect plant ANSWER CHOICE ANALYSIS Answer: B A. Incorrect: The second part is correct as the 11 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 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 aa protective interlock to prevent the pump from starting similar to RCP starting interlock indications.

D. Incorrect:

Incorrect: The first part is correct however the HWP will start since CBP suction pressure is << 55 psig. Plausible since the lightlight above the HWP HWP starting starting switch is illuminated itit would be reasonable to interpret it as aa protective interlock to prevent the pump from starting starting similar to RCP starting starting interlock indications.

indications.

Technical Reference(s)

Reference(s):: CF-C, CF-C, OPIIIAIIIO6IOO2A OP/1/A11106/002A End. Encl. 4.2 4.2 Proposed Proposed references to be be provided to applicants applicants during during examination: NONE NONE Learning Learning Objective:

Objective: CF-C CF-C R9, R9, R28 R28 Question Source:

Question Source: NEW NEW Question History:

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

Level: Comprehens Comprehension ion and and Analysis Analysis

2010A NRC REACTOR OPERATOR EXAM 1I 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 B?sed 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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 63 Question 63 T21G2 T2/G2 - cpw

- cpw 072K1 .03, Area 072K1.03, Area Radiation Radiation Monitoring Monitoring (ARM)

(ARM) System System Knowledge of Knowledge ofthe the physical physical connections connections and/or and/or cause-effect cause-effect relationships relationships between the ARM system and the following between the ARM system and the following systems: systems:

Fuel building Fuel building isolation.

isolation.

(3.6/3.7)

(3.6/3.7)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Per NNRC, actions based on on fuel fuel damage andand RIA-6 RIA-6 OK OK ifif distracter distracter includes includes auto auto isolation of SFP ventilation ventilation systems Requires knowledge Requires knowledge of of any any cause-effect cause-effect relationship relationship between between ARM ARM alarms alarms and and SFP isolation (none).

SFP isolation (none).

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: A Answer: A A. CORRECT:

A. CORRECT: 3RIA-63RIA-6 causes causes anan audible audible alarm.

B. Incorrect:

B. Incorrect: Plausible since it could be seen as desirable for the ventilation systems in in the SFP (which exhausts to the Unit vent) to auto isolate on high radiation levels.

the Incorrect: Plausible since AP/9 does require starting one of the Spent Fuel filtered C. Incorrect:

exhaust fans when there is fuel damage in the SFP SEP however they are NOT started from the Control Room but in the entrance area to the SFP SEP itself.

D. Incorrect:

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 SEP however the controls for the fans are located in the Control Room.

Technical Technical Reference(s)

Reference(s):: API9, AP/9, RAD-RIA Proposed Proposed references to to be provided provided toto applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: RAD-RIA RAD-RIA R15,R15, FH-FHS FH-FHS R31 R31 Question Question Source:

Source: NEWNEW Question Question History:

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

Level: Knowledge Knowledge and and Fundamental Fundamentals s

2010A NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 64 Question 64 IA Pressure IA Pressure VS. vs. Time Time 100 100~~------------------------------------------~

uJ W 90 ........................... :........................... : ......................... '"

0::

D

J (1) if)

Co if) uJ W

0::

U 0...

<< 80 ........................... :* ........................... .: .......................... .

70~----~~-----4*-------+------~*------~----~

1200 1210 1220 1230 TIME Based on the graph above, which ONE of the following describes the time at which SA- SA 141 (SA to IA Controller) will automatically open?

A. 1207 B. 1210 C. 1212 1212 D. 1215 1215

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM 64 Question 64 T21G2 - cpw T2/G2 -

079K4.01, Station Air System (SAS)

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

Cross-connect with lAS.

(2.9/3.2)

K/A MATCH ANALYSIS KIA 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 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 1I 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?

3AMDEFWP A. 3A MDEFWP 3CFIPIP B. 3C HPIP C. 3B MDEFWP 3ALPSWP D. 3A LPSWP

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 65 Question 65 T2/G2-cpw T2/G2 - cpw 086K5.03, Fire 086K5.03, Fire Protection Protection System System (FPS)

(FPS)

Knowledge of Knowledge of the operational implication the operational implication of of the following concepts the following concepts as as they they apply apply to the Fire Protection System:

to the Fire Protection System:

Effect of Effect water spray of water on electrical spray on electrical components.

components.

(3.1/3.4)

(3.1/3.4)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS er NRC put fue n a specfc location which which requires reqwies de-energizing aa Load Center oror 4160V bus and ask about equipment equipment effected.

effected.

Requires knowledge Requires knowledge of of operational operational implications implications of of de-energizing de-energizing 3TE 3TE to allow allow water spray to 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 3B MDEFWP is powered from 3TE and would therefore NOT be available D. Incorrect: 3B LPSWP is powered from 3TC Technical Reference(s):

Reference(s): PNS-HPI, PNS-LPI, CF-EF, PNS-R8C, 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 Question Cognitive Level: Knowledge and Fundamental Fundamentals s

2010A NRC REACTOR OPERATOR EXAM 1I POINT Question 66 Unit 2 plant conditions:

ReactorinMODE6

• 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 OP/21A11 502/007 (Operations Defueling/Refueling Responsibilities) for core alterations AND what is the minimum number of OPERABLE Source Range NI's NIs required by the same procedure?

A. meets /I 2 B. meets /I 11 C. does NOT meet /I 2 D. does NOT meet /I 1 1

2010A NRC REACTOR OPERATOR EXAM Question 66 T3-cpw T3 - cpw 2.1.36, Conduct of Operations Knowledge of procedures and limitations involved in core alterations.

(3.0/4.1))

(3.0/4.1 K/A MATCH ANALYSIS KIA procedure limits and precautions and/or Tech Spec Requires knowledge of procedure 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 OP/2/AI1 5021007 (Operations Defueling/Refueling Defuel i ng/Refuel i ng Responsibilities).

Second part is correct as both the procedure and TS require 2 Source Range NIs when moving a fuel assembly.

NI's 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.

Incorrect: : First part is plausible since 2950 is the upper end of band and it could D. Incorrect::

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.

OP/2/A/1502/007, TECH SPECS Technical Reference(s): FH-FHS, OP/2/Al1502/007, 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 NIA Question Cognitive Level: Knowledge and Fundamentals

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAMEXAM POINT 1I POINT Question 67 Question 67 Unit 11 initial Unit initial conditions:

conditions:

• • Reactor Reactor power power = 100%100%

• • LOST LDST level level == 75" 75 stable stable

• • Group Group 77 rod position == 94%

rod position 94% withdrawn withdrawn

• • Makeup Makeup to LDST initiated to LOST initiated from from 1lB BHUT B BHUT

• • Neutron Neutron error error == 00 stable stable Current conditions:

Current conditions:

• • 1IHP-15 Bailey controller HP-15 Bailey controller indicates indicates 470 gallons gallons added to LOST added to LDST

• • 1I B B Bleed Bleed Transfer Pump Pump secured secured Based on the above conditions, which ONE of the following would describe a diverse indication that 470 gallons of 11 B B BHUT had been added to the the LOST?

LDST?

LDST level is approximately LOST approximately ____ inches and neutron error will become _ _ __

A. 90 / positive B. 90 /I negative C. 95 / positive D. 95 /I negative O.

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 67 Question 67 T3-cpw T3 - cpw 2.1.45, Conduct 2.1.45, Conduct of Operations of Operations Ability to Ability identify and to identify and interpret interpret diverse diverse indications indications to validate the to validate the response response ofof another indication.

another indication.

(4.3/4.3)

(4.3/4.3)

K/A MATCH K/A MATCH ANALYSIS ANALYSIS Requires analyzing Requires analyzing variousvarious diverse diverse indications indications to validate aa SG to validate SG tube tube leak leak exists.

exists.

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: B B A. Incorrect. First A. Incorrect. part isis correct.

First part correct. Second part isis plausible Second part plausible since since itit would would be be correct correct ifif the the addition was from A SHUT 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: LOST 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 LDST volume however this is the Pressurizer value. Second part is plausible for LOST since it would be correct if the addition was from A SHUT BHUT OR if the student made the common error of incorrectly determining the direction of rod motion required to offset the boron addition.

D.

O. Incorrect: First part is plausible because this would be correct if you use 24 gal/inch for LDST LOST volume however this is the Pressurizer value. Second part is correct.

Technical Reference(s)

Reference(s):: PNS-PZR, PNS-HPI, CP-016 Proposed references to be provided to applicants during examination: examination: NONE Learning Objective: PNS-PZR PNS-PZR R1,2,3 CP-016 R5, Question Source: NEW NEW Question Question History:

History: Last Last NRC NRC ExamExam N/AN/A Question Question Cognitive Cognitive Level:

Level: Comprehens Comprehensionion and and Analysis Analysis

NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAMEXAM 1I POINT POINT Question 68 Which ONE following describes the RCS RCS Pressure Pressure Safety Limit Limit (psig)

(psig) and what is is limit is credited with ensuring the limit is NOT 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 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 68 Question 68 T3-CPW T3- CPW 2.2.25, Equipment 2.2.25, Equipment Control Control Knowledge of Knowledge of the the bases bases in in Technical Technical Specifications Specifications forfor limiting limiting conditions conditions for for operations and safety operations and safety limits. limits.

(3.2/4.2 (3.2/4.2 K/A MATCH KIA MATCH ANALYSIS ANALYSIS Requires knowledge Requires knowledge of of the the LCO LCO for for RCS RCS Pressure Pressure safety safety limit limit and and its its bases.

bases.

ANSWER CHOICE ANSWER CHOICE ANALYSIS ANALYSIS Answer: D Answer: D A. Incorrect:

A. Incorrect: First First part part is is plausible plausible since since 2500 2500 psig psig is is the the RCS RCS design design pressure pressure value.

value.

Second part is correct.

B. Incorrect: First part is is plausible since 2500 psig is 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)

Reference(s):: TS 2.1.2 (RCS Pressure Safety Limit) including bases Proposed references to be provided to applicants during examination: NONE NONE Learning Objective:

Objective: ADM-TSS R8 Question Question Source:

Source: NEWNEW Question Question History:

History: Last Last NRC NRC ExamExam N/A NIA Question Question Cognitive Cognitive Level:

Level: Knowledge Knowledge andand Fundamental Fundamentals s

REACTOR OPERATOR EXAM 2010A NRC REACTOR 1I POINT Question 69 Unit 11 plant conditions:

• Reactor power = 100% 100%

• 1A IA Core Flood Tank parameters:

o Pressure = = 572 psig stable o Level = 12. 91 feet o Boron Concentration = = 2010 ppmb

• 1I 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 IA CFT to OPERABLE within 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> ONLY D. Restore 1 B CFT to OPERABLE within 1 lB 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 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 11 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 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 CFT's CFTs inoperable and requires immediate entry into LCO 3.0.3. CFT pressure is required to be 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 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 CFTs not meeting the SR's, SRs, Condition D 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 BB 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 NIA Question Cognitive Level: Knowledge and Fundamentals

NRC REACTOR 2010A NRC REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 70 Question Unit 11 plant Unit plant conditions:

• • Reactor Reactor power power == 100%

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. D RPS channel in Manual bypass 1ID D. 230KV Dacus Black and White lines isolated

2010A NRC REACTOR OPERATOR EXAM Question 70 T3-CPW 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 contain > 30,000 gallons that the UST contain> gallons... PT/600/01 (Periodic Instrument Surveillance) verifies this volume by requiring UST level be> be > 6 feet.

B. Incorrect: Plausible since Tech Specs requires the BWST to be operable and PT/600/01 (Periodic Instrument contain 350,000 gallons of Borated water. PT/600101 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 inoperabilitys exist.

required functions being inoperable as long as no other RPS inoperability's 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, 3.7/.6, 3.5.4, 3.3.10, and 3.3.1 PTl600/01 (Periodic PTI600/01 Instrument Surveillance)

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

2010A NRC REACTOR OPERATOR EXAM 1I POINT Question 71 Unit 11 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. 1I RIA-45 reaches ALERT setpoint C. Suction piping pressure == 5 inches of water vacuum D. 1 I PR-3 (RB PURGE CONTROL) closed

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 71 Question 71 T3-cpw T3 - cpw 2.3.1 1, Radiation 2.3.11, Radiation Control Control Ability to Ability control radiation to control radiation releases.

releases.

(3.8/4.3)

(3.8/4.3)

MATCH ANALYSIS K/A MATCH KIA Question demonstrates Question demonstrates ability ability to control radiation to control radiation releases releases by by demonstrating demonstrating the the ability ability to determine conditions that would result that result in in terminating the release release by by tripping the tripping main purge the main purge fan.

fan.

ANSWER CHOICE ANALYSIS Answer: D 0 A. Incorrect: Plausible since there is ES actuations that will cause the purge fan to trip however it is ES Channels 1I &/or 2 therefore this would be a correct choice if asking about ES-1ES-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):

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 Fundamental Fundamentals s

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR EXAM OPERATOR EXAM POINT 1I POINT Question 72 Question 72 Unit 11 plant Unit plant conditions:

conditions:

• • Reactor Reactor in in MODE MODE 66

• • LPILPI DHR DHR in in progress progress using using 11A LPI pump A LPI pump

• • Fuel Fuel movement movement in in progress progress

• • NI-3 Nl-3 and and 4 outout of of service service

• • RB RB Purge Purge in in progress progress Based on the above conditions, which ONE of the following describes when Reactor Building evacuation is is required?

A. 11NI-2fails N 1-2 fails low B. 11RIA-4 RIA-4 HIGH alarm C. 11RIA-45 RIA-45 HIGH alarm D. Loss of power to 1A IA LPI pump

2010A NRC REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 72 Question 72 T3-cpw T3 - cpw Radiation Control 2.3.12, Radiation Knowledge of radiological Knowledge radiological safety principles principles pertaining pertaining to licensed licensed operator requirements, fuel handling duties, such as containment entry requirements, handling responsibilities, responsibilities, access to locked high-radiation areas, aligning filters, etc.

(3.2/3.7)

K/A MATCH ANALYSIS KIA 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 NIs during core alterations. This failure would require immediate operable SR NI's 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 11 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 1AIA 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): API3AP/3 (Boron Dilution), AP126 AP/26 (Loss of DHR)

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

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 73 Question 73 For Operations For Operations personnel, personnel, which which ONE ONE of of the following describes the following describes the the required required response response to an to an Electronic Electronic Dosimeter Dosimeter dose alarm and dose alarm when itit is and when is acceptable acceptable to to deviate deviate from from that that requirement?

requirement?

A. Exit Exit the area area immediately immediately and and contact contact RPRP /I with RP RP permission permission B. Exit B. Exit the area area immediately immediately and contact contact RPRP /I with emergency emergency dose dose limits limits in in effect Move away from the area until alarm clears /I with RP permission C. Move D. Move away from the area until alarm clears /I with emergency dose limits in D. in effect

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM Question 73 Question 73 T3-cpw T3 - cpw 2.3.7, Radiation 2.3.7, Radiation Control Control Ability to Ability to comply comply with with radiation radiation work work permit permit requirements requirements during during normal normal or or abnormal conditions.

abnormal conditions.

(3.5/3.6)

(3.5/3.6)

K/A MATCH KIA MATCH ANALYSIS Requires knowledge Requires knowledge of of how how to respond respond to Dose Dose and and Dose Dose Rate Rate alarms alarms determined by RWPs in by RWP's in both both normal and abnormal conditions. Additionally Additionally requires knowledge of when it is acceptable under abnormal conditions to requires 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- 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 EDLs are implemented NEO's NEOs and others EDLs may continue to work through ED alarms.

working under EDL's 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)

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 (RAD020902))

Question History: Last NRC Exam NIA Question Cognitive Cognitive Level: Knowledge and Fundamental Fundamentals s

2010A NRC 2010A NRC REACTOR REACTOR OPERATOR OPERATOR EXAM EXAM 1I POINT POINT Question 74 Question 74 Unit 11 plant Unit plant conditions:

• • Blackout Blackout tab inin progress progress Initiate AP/11

• EOP step gives direction to "Initiate AP/1 1 (Recovery From Loss of Power)"

Power)

Based on the above conditions, which ONE of the following describes the actions required to perform the EOP step?

AP/1 1 and SRO will wait until AP/11 A. RO will perform AP/11 AP/1 1 has been completed to continue in the EOP.

AP/1 I and then return to the EOP once AP/11 B. SRO will direct steps in AP/11 AP/1 1 is complete.

AP/1 1 before SRO continues in EOP but once AP/11 C. RO will take steps to begin AP/11 AP/1 I is begun the SRO can re-direct the RO's ROs activities.

D. SRO can continue in EOP once AP/11 AP/1 1 is being performed. SRO CANNOT re-direct the RO's ROs activities until AP/11 AP/1 I is completed.

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 74 Question 74 T3-cpw T3 - cpw 2.4.16 Emergency Procedures 2.4.16 Emergency Procedures 1/ Plan Plan Knowledge of Knowledge of EOP EOP implementation implementation hierarchy hierarchy and and coordination coordination with with other other support support procedures or guidelines such procedures or guidelines such as, as, operating operating procedures, procedures, abnormal abnormal operating operating procedures, and procedures, and severe severe accident accident management management guidelines.

guidelines.

(3.5/4.4)

(3.5/4.4)

KIA MATCH KIA MATCH ANALYSIS Requires Knowledge Requires Knowledge of of EOP/AP EOP/AP implementation implementation hierarchy hierarchy and process for coordinating AP with EOP based based on criteria in step step directing AP performance ANSWER CHOICE CHOICE ANALYSIS Answer: C A. Incorrect: Plausible since this would be correct if the step gave direction to "Perform Perform AP/1 1.

AP/11". There Perform steps and OMP are numerous steps in the EOP that are "Perform" 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.(End. 5.38).

B. Incorrect: Plausible since this would be correct if the step were an If At Any Time (IAAT)

(1M T) step. EOP has numerous cases were the 1M IAATT process is used. This requires going and performing the steps directed by the 1M IAATT and then returning to the original step.

Initiate steps require actions to begin the referenced steps or C. CORRECT: "Initiate" procedure and then continuing with the current procedure in use.

D. INCORRECT: Plausible since this would result in both AP/11 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 1-18 Proposed references to be provided to applicants during examination: NONE Learning Objective: ADM-OMP RIO, R10, R52 Question Source:

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

Question Knowledge and Fundamentals

2010A NRC 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM POINT 1I POINT Question 75 Question 75 Unit 11 plant Unit plant conditions:

conditions:

• • Reactor Reactor power power == 100%

100%

• • 11SA3IB6 (FIRE ALARM)

SA3/B6 (FIRE ALARM) actuated actuated

• • Fire Fire Alarm Alarm panel panel indication indication point 0202071 oo point 0202071 (Unit (Unit 11 pipe pipe trench trench room room 348 348 north north end) end) actuated actuated Based on Based on the the above above conditions, conditions, which which ONE ONE of of the following describes the following describes thethe initial initial action action required by required by the Alarm Response the Alarm Response Guide Guide AND AND aa method method used in in RP/10001029 used RP/1000/029 (Fire (Fire Brigade Response)

Brigade Response) to to dispatch the fire brigade brigade when itit is is required?

required?

Dispatch aa Fire A. Dispatch Fire Brigade Brigade qualified qualified operator operator toto determine determine validity validity of of the alarm 1/ Plant the alarm Plant Paging system Dispatch a Fire B. Dispatch Fire Brigade Brigade qualified operator to determine validity of the alarm 1/ Have 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 2010A REACTOR OPERATOR NRC REACTOR OPERATOR EXAM EXAM Question 75 Question 75

-cpw T3 -cpw T3 2.4.26 Emergency Procedures 2.4.26 Emergency Procedures I/ PlanPlan Knowledge of Knowledge of facility facility protection protection requirements, requirements, including including fire fire brigade brigade and and portable portable fire fighting equipment fire fighting equipment usage usage (3.1/3.6)

(3.1/3.6)

K/A MATCH KIA MATCH ANALYSIS ANALYSIS Per NRC, ask about activating fire brigade from from Control Control Room Room OK.

Requires knowledge Requires knowledge of of actions actions directed directed by by aa Fire Fire Alarm in in the the control control room room asas well as well knowledge of as knowledge of how how Fire Brigade is Fire Brigade is dispatched.

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.

Attachment 2 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 RP/1000/016 activation...)

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:

Incorrect: First part is is plausible since itit would be reasonable to assume that dispatching the fire brigade would be directed if aa Fire Alarm is received however the ARG does not direct dispatching Fire Brigade until until the alarm is determined to to be be valid. Second part part is plausible since would be correct ifif asking about dispatching MERT MERT to to aa medical medical emergency emergency per per RP/1000/016 RP/10001016 (MERT activation...)

activation ... )

Technical Technical Reference(s)

Reference(s):: RPIOIBII000 IO29 (Fire RP/O/B/1000/029 (Fire brigade brigade response) response) ARGARG for for ISA3/B6, 1SA3/B6, IC-FDS, IC-FOS, RP/1000101 RP/10001016 6 (MERT activation...)

activation ... )

Proposed Proposed references references to to bebe provided provided toto applicants applicants during during examination:

examination: NONE NONE Learning Learning Objective:

Objective: lC-FDS IC-FOS R6 R6 Question Question Source:

Source: NEW NEW Question Question History:

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

Level: Comprehens Comprehension ion andand Analysis Analysis