Text

Initial Exam 2012-302 Draft RO Written Exam
	ML13025A013
Person / Time
Site:	Oconee
Issue date:	01/22/2013
From:	; NRC/RGN-II
To:	; Duke Energy Carolinas
References
	50-269/12-302, 50-270/12-302, 50-287/12-302
	Download: ML13025A013 (151)
	v • d • e

{{#Wiki_filter:FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 1 EPEOO7 EA2.04 Reactor Trip Ability to determine or interpret the following as they apply to a reactor trip: (CFR 41.7 / 45.5 / 45.6) If reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP Given the following Unit 1 conditions:

Time 0400
Reactor power = 70%
Diamond and BOTH FDW Masters in MANUAL
1TA de-energized Current conditions:
Time =0401
Reactor power = 59% decreasing

1) TheOATCwilI (1).

2) BOTH FDW Masters (2) transfer to AUTO and reduce FDW.

Which ONE of the following completes the statements above? A. 1. insert control rods and initiate emergency boration

2. will B. 1. insert control rods and initiate emergency boration

2. will NOT C. 1. perform Plant Transient Response

2. will D. 1. perform Plant Transient Response

2. will NOT Friday, October 12, 2012 Page 1 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 1 General Discussion L. Answer A Discussion Incorrect. First part is correct. Second part is plausible because it is true for the Main FDW pumps and FDW control valves. Answer 8 Discussion Correct. The reactor should have tripped due to the 2 RCPs tripping. The OATC will initiate Rule 1. This will direct inserting rods and initiating emergency boration. The FDW Masters will not revert to AUTO. Answer C Discussion Incorrect. First part is plausible because reactor power is below the fluxlflow/imbalance trip for a loss of one RCP. Second part is plausible because it is true for the Main FDW pumps and FDW control valves. Answer D Discussion Incorrect. First part is plausible because reactor power is below the flux/flow/imbalance trip for a loss of one RCP. Second part is correct. Basis for meeting the KA Question requires knowledge of how to determine an ATWS has occurred and the actions to take. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Rule I EOP IMAs Chapter 4 EPEOO7 EA2.04 Reactor Trip Ability to determine or interpret the following as they apply to a reactor trip: (CFR 41.7 / 45.5 I 45.6) If reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 2 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 2 21 APEOO8 AK2.03 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: (CFR 41.7 / 45.7) Controllers and positioners Given the following Unit 1 conditions: Time = 0400

Reactor power = 100%
Transient occurs Time = 0405
RCS pressure = 2467 psig increasing Time = 0410
RCS pressure 2427 psig decreasing Time 0415
RCS pressure = 2025 psig decreasing
1 DIB breaker #24 trips OPEN

1) At0410, 1RC-66willhaveasignaltobe (1)

2) At 0415, 1 RC-66 (2) be operated manually by the RO.

Which ONE of the following completes the statements above? A. 1. open

2. can B. 1. open

2. can NOT C. 1. closed

2. can D. 1. closed

2. can NOT Friday, October 12, 2012 Page 3 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 2 General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because RC-66 does have alternate control power. However it will not operate if power to the solenoid is lost (IDIB Breaker #24). Answer B Discussion Correct. 1RC-66 will still have an open signal. Per AP/44, IRC-66 opens at 2450 psig and closes at 2400 psig. IDIB Breaker #24 supplies power to I RC-66 solenoid, It will not operate in Auto or manual without power to the solenoid. Answer C Discussion Incorrect. First part is plasuible because 2427 is below the 2450 psig setpoint which opens IRC-66 but above close setpoint of 2400. Second part is plausible because RC-66 does have alternate control power. However it will not operate if power to the solenoid is lost (1 DIB Breaker

24)

Answer 0 Discussion Incorrect. First part is plasuible because 2427 is below the 2450 psig setpoint which opens IRC-66, but above close setpoint of 2400. Secoi part is correct. Basis for meeting the KA itioequires knowledge of the controller for IRC-66 and how it relates to a vapor space accident. Basis for Hi Cog 1 Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided PNS-PZR R30 AP/l APEOO8 AK2.03 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: (CFR 41.7 / 45.7) Controllers and positioners 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 4 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 3 3] EPEOO9 EA2.33 Small Break LOCA Ability to determine or interpret the following as they apply to a small break LOCA: (CFR 43.5 /45.13) RCS water inventory balance and Tech-Spec limits Given the following Unit 1 conditions: Time 0400

Reactor power = 100%
RCS Tave = 5790 F stable
Pzr level 220 inches stable
RC Makeup Flow = 58 gpm stable
Letdown Flow = 74 gpm stable
Seal Inlet Header Flow = 32 gpm stable
Total Seal Return Flow = 8.8 gpm stable
RB Normal Sump level 18 inches increasing
LDST level 80 inches decreasing Time = 0430
RC Makeup Flow = 85 gpm slowly increasing
Pzr level 200 inches decreasing

1) At 0400, in accordance with TS 3.4.13 (RCS Operational LEAKAGE), the calculated RCS leak rate (1) the unidentified LEAKAGE limit.

2) At 0430, in accordance with AP/2 (Excessive RCS Leakage) 1 HP-5 (2) required to be closed.

Which ONE of the following completes the statements above? A. 1. exceeds

2. is B. 1. exceeds

2. is NOT C. 1. does NOTexceed

2. is D. 1. does NOT exceed

2. is NOT Friday, October 12, 2912 Page 5 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 3 L General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plasuible because Pzr level is decreasing and makeup flow is abnormally high. Answer B Discussion Correct. Using an inventory balance to determine the RCS leakage the leakage is 7.2 gpm. This is above the unidentified TS limit of 1 gpm. Leak Rate = 58 + 32 - 74 - 8.8 = 7.2 gpm MU SI LD TSR AP/2 has a JAAT step that if makeup flow is> 100 gpm and Pzr level is decreasing then close IHP-5. Makeup flow is < 100 gpm and IHP-5 will not be closed at this time. Answer C Discussion Incorrect. First part is plausible because it does not exceed the limit for identified leakage of 10 gpm. Second part is plasuible because Pzr level is decreasing and makeup flow is abnormally high. Answer D Discussion Incorrect. First part is plausible because it does not exceed the limit for identified leakage of 10 gpm. Second part is correct. Basis for meeting the KA The question requires how to perform an inventory balance to determine RCS leakage and if the TS limit is exceeded and then the actions to take when the leak increases. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided AP/02 TS 3.4.13 EPEOO9 EA2.33 Small Break LOCA Ability to determine or interpret the following as they apply to a small break LOCA: (CFR 43.5 /45.13) RCS water inventory balance and Tech-Spec limits 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 6 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 4 EPEO1 1 EK2.02 Large Break LOCA Knowledge of the interrelations between the Large Break LOCA and the following: (CFR 41.7 I 45.7) Pumps Given the following Unit 1 conditions: Initial conditions:

Reactor power 100%

Current conditions:

RCS pressure = 328 psig decreasing
RB pressure 5 psig increasing

1) (1) Reactor Building Spray pumps are operating.

2) (2) LPI pumps are operating.

Which ONE of the following completes the statements above? A. 1.two

2. ONLY two B. 1.two

2. three C. 1. zero

2. ONLY two D. 1. zero

2. three Friday, October 12, 2012 Page 7 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 4 General Discussion Answer A Discussion Incorrect. First part is plausible if the candidate has the misconception that RBS initiates at a RB pressure of 3 psig. Second part is correct. Answer B Discussion Incorrect. First part is plausible if the candidate has the misconception that RBS initiates at a RB pressure of 3 psig. Second part is plausible because there are three LPI pumps however only 2 are ES pumps. Answer C Discussion Correct. RB is below the RBS setpoint of 10 psig so no RBS pumps will be operating. ES will start two LPI pumps at 3 psig in the RB and < 550 RCS pressure. Answer D Discussion Incorrect. First part is correct. Second part is plausible because there are three LPI pumps however only 2 are ES pumps. Basis for meeting the KA Question requires knowledge of pumps that will be operating following a LBLOCA. Basis for Hi Cog Basis for SRO only FZiiievel Cognitive Level QuestionTypJ Question Source RO Memory NEW Development References Student References Provided IC-ES R14,R18 EPEO1 1 EK2.02 Large Break LOCA Knowledge of the interrelations between the Large Break LOCA and the following: (CFR 41.7 / 45.7) Pumps 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 8 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 5 5] APEO15!017 AK3.04 Reactor Coolant Pump (RCP) Malfunctions Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 41.5,41.10 / 45.6 / 45.13) Reduction of power to below the steady state power- to-flow limit Given the following Unit 1 conditions: Initial conditions:

Reactor power 80% stable Current conditions:
1A1 RCP trips

1) The unit will run back to a MAXIMUM power level of (1) %.

2) The reason for limiting reactor power in this case is to maintain DNBR (2) the limit specified in the COLR.

Which ONE of the following completes the statements above? A. 1.65

2. above B. 1.65

2. below C. 1.74

2. above D. 1.74

2. below Friday, October 12, 2012 Page 9 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE C 1LT42 ONS SRO NRC Examination QUESTION 5 5] General Discussion Answer A Discussion Incorrect. First part is plausible because it is correct for a FDW pump trip. Second part is correct. Answer B Discussion incorrect. First part is plausible because it is correct for a FDW pump trip. Second part is plausible because a candidate could have the misconception that power is limited to prevent going above the DNBR limits. Answer C Discussion Correct. ICS will run the unit back to 74%. Power is limited to maintain DNBR above the limit in the COLR. Answer D Discussion incorrect. First part is correct. Second part is plausible because a candidate could have the misconception that power is limited to prevent going above the DNBR limits. Basis for meeting the KA Question requires knowledge of the limit of reactor power on a RCP trip with excessive imbalance and the reason for this limit. Basis for Hi Cog Basis for SRO only Job Level I Cognitive Level QuestionType Question Source RO Memoty NEW Development References Student References Provided STG-ICS Chapter 2 IC-RPS R3 APEO 15/017 AK3 .04 Reactor Coolant Pump (RCP) Malfunctions Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 41.5,41.10 / 45.6 / 45.13) Reduction of power to below the steady state power- to-flow limit 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 10 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 6 APEO25 AK2.Ol Loss of Residual Heat Removal System (RHRS) Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: (CFR 41.7 / 45.7) RFIR heat exchangers Given the following Unit 1 conditions:

Unit shutdown in progress
LPI aligned in High Pressure Mode

1) If (1) fails to control properly a loss of decay removal can occur.

2) In accordance with AP/26 (Loss of Decay Heat Removal), heat removal will be restored by (2) to control the cooldown.

Which ONE of the following completes the statements above? A. 1. 1LPSW-251

2. placing 1LPSW-251s FAIL SWITCH in FAIL OPEN and throttling 1LPSW-4 B. 1. 1LPSW-251

2. aligning the LPI system in series mode and using 1LPSW-252 C. 1. ILPSW-252

2. placing 1 LPSW-252s FAIL SWITCH in FAIL OPEN and throttling 1 LPSW-5 D. 1. 1LPSW-252

2. aligning the LPI system in switchover mode and using 1LPSW-251 Friday, October 12, 2012 Page 11 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE A 1LT42 ONS SRO NRC Examination QUESTION 6 6j General Discussion Answer A Discussion Correct. In the High Pressure Mode flow is through the IA LPI cooler only. ILPSW-251 failing closed will result in having to enter AP/26 (Loss of Decay Heat Removal). AP/26 will direct placing ILPSW-251s FAIL SWITCH in FAIL OPEN and using ILPSW-4 to control heat removal. Answer B Discussion Incorrect. First part is correct. Second part is plausible because it is correct for switchover mode. Answer C Discussion Incorrect. First part is plausible because the candidate could have the misconception that in the High Pressure Mode LPI is aligned through the B cooler. Second part is plausible because it would be correct if the first part of the question were correct. Answer D Discussion Incorrect. First part is plausible because the candidate could have the misconception that in the High Pressure Mode LPI is aligned through the B cooler. Second part is plausible because it would restore heat transfer. Basis for meeting the KA Question requires knowledge of how losing LPSW flow to Decay heat removal coolers could result in a Loss of DHR and actions required to restore heat removal. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided PNSLPI R8 High Pressure Mode Switchover Mode Series Mode AP/26 APEO25 AK2.O1 Loss of Residual Heat Removal System (RHRS) Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: (CFR 41.7 /45.7) RHR heat exchangers 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 12 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 7 APEO26 2.4.50 Loss of Component Cooling Water (CCW) tz APEO26 GENERIC Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 /43.5 /45.3) Given the following Unit 1 conditions: Initial conditions:

Time = 0400
Reactor power= 100%
Component Cooling Return Flow = 563 gpm
1SA-09/C-1 (Component Cooling Return Flow Low) actuates Current conditions:
Time 0402
Component Cooling Return Flow = 103 gpm
The Standby CC pumps has NOT started
CC Surge Tank level = 18 inches stable

1) At 0400, Statalarm ISA-09/C-1 (1) valid.

2) At 0402, in accordance with 1 SA-09/C-1 ARG, the Standby CC pump (2)_ be manually started.

Which ONE of the following completes the statements above? A. 1.is

2. will B. 1.is

2. will NOT C. 1. is NOT

2. will D. 1. is NOT

2. will NOT Friday, October 12, 2012 Page 13 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 7 General Discussion Answer A Discussion Correct. 1SA-09/C-l set point is 575 gpm. The Standby CC pump did not start and should be started after verifying that CC surge tank level is > 12 inches. Answer B Discussion Incorrect. First part is correct. Second part is plausible because it would be correct if CC level were less than 12 inches. Answer C Discussion iorrect. First part is plasuible because it is above the setpoint for Statalarm I SA-09/B-l (CRD Return Flow Low) of 138 gpm. Second pa correct. Answer D Discussion Incorrect. First part is plasuible because it is above the setpoint for Statalarm 1SA-09/B-l (CRD Return Flow Low) of 138 gpm. Second part is plausible because it would be correct if CC level were less than 12 inches. Basis for meeting the KA Question requires knowledge of alarm setpoints and controls identified in the ARG. Basis for Hi Cog Basis for SRO only Job Level j Cognitive Level j QuestionType Question Source RO Comprehension NEW Development References Student References Provided PNS-CC Rl5 1SA-09/B-1 and ISA-09/C-L APEO26 2.4.50 Loss of Component Cooling Water (CCW) APEO26 GENERIC Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 / 43.5 / 45.3) 401-9 Comments: Remarks/Status Friday, October 12, 2912 Page 14 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 8 81 APEO27 2.4.49 Pressurizer Pressure Control System (PZR PCS) Malfunction APEO27 GENERIC Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10/43.2/45.6) Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%

Current conditions:

RCS pressure = 2078 psig decreasing
Pzr level 350 inches increasing
1 SA-1 8/A-i (Pressurizer Relief Valve Flow) actuated
ALL 1 RC-66 flow monitor red lights are illuminated Immediate Manual Actions of AP/44 (Abnormal Pressurizer Pressure Control) directs the operator to (1) AND the actual setpoint (psig) for the Low RCS Pressure RPS trip is (2)

Which ONE of the following completes the above statement? A. 1. Close 1RC-4

2. 1810 B. 1. Close 1RC-4

2. 1800 C. 1. Manually trip the reactor

2. 1810 D. 1. Manually trip the reactor

2. 1800 Friday, October 12, 2012 Page 15 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 8 A General Discussion Answer A Discussion Correct: AP/44 IMAs state: IAAT all of the following conditions exist: PORV open RC pressure < 2300 psig (HIGH) or 480 psig (LOW) PZR level <375 THEN close 1RC-4. The actual low RCS pressure setpoint for the RPS trip is 1810 psig. Answer B Discussion Incorrect. First part is correct. Second part is plausible since 1800 psig is the minimum allowed TS setpoint. Answer C Discussion Incorrect: First part is plausible since it would be the correct actions if pressurizer level were> 375 and therefore IRC-4 could not be closed. Second part is correct. Answer D Discussion Incorrect: First part is plausible since it would be the correct actions if pressurizer level were> 375 and therefore IRC-4 could not be closed. Second part is plausible since 1800 psig is the minimum allowed TS setpoint. Basis for meeting the KA This question requires knowledge of AP/44 IMAs for a failed open PORV. Basis for Hi Cog r Basis for SRO only 1 1 Job Level Cognitive Level QuestionType Question Source RU Memory BANK ONS 2009B Ql Development References Student References Provided EAP-APG R9 IC-RPS R3 IAP/44 ONS 2009B Ql APEO27 2.4.49 Pressurizer Pressure Control System (PZR PCS) Malfunction APEO27 GENERIC Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10/43.2/45.6) 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 16 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 9 9 EPEO29 EA1.13 Anticipated Transient Without Scram (ATWS) Ability to operate and monitor the following as they apply to a ATWS: (CFR 41.7 / 45.5 / 45.6) Manual trip of main turbine Given the following Unit 1 conditions: Initial conditions:

Time 0400
Reactor power = 100%

Current conditions:

Time0405
BOTH Main FDW pumps trip
Reactor power = 60% decreasing
ALL Main Turbine Stop valves OPEN
Rule 1 (ATVVS/UNPP) initiated

1) At the completion of Rule 1 (1) HPI pumps will be operating.

2) At 0405 and in accordance with the UNPP tab, the Main Turbine (2) required to be manually tripped.

Which ONE of the following completes the statements above? A. 1. ONLY two

2. is B. 1. ONLY two

2. is NOT C. 1. three

2. is D. 1. three

2. is NOT Friday, October 12, 2012 Page 17 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE

    *1LT42 ONS SRO NRC Examination                                                QUESTION                    9                   9j General Discussion Answer A Discussion Correct. Rule I requires two HPI pumps to be operating. Because BOTH Main FDW pumps are tripped the UNPP tab directs tripping the Main Answer B Discussion

[Incorrect. First part is correct. Second part is plausible because it would be correct if the Main FDW pumps were not tripped Answer C Discussion Incorrect. First part is plausible because it is true for Rule 2 (Loss of SCM). Second part is correct. Answer D Discussion Incorrect. First part is plausible because it is true for Rule 2 (Loss of SCM). Second part is plausible because it would be correct if the Main FDW pumps were not tripped. Basis for meeting the KA F . Question requires knowledge of when to trip the Main Turbine during an ATWS. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestioriType Question Source RO Comprehension NEW Development References Student References Provided EOP Rule I LOP U1IPP tab EPEO29 EA1.13 Anticipated Transient Without Scram (ATWS) Ability to operate and monitor the following as they apply to a ATWS: (CFR 41.7 / 45.5 / 45.6) Manual trip of main turbine 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 18 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 10 10] EPEO38 EA2.07 Steam Generator Tube Rupture (SGTR) Ability to determine or interpret the following as they apply to a SGTR: (CFR 43.5 / 45.13) Plant conditions, from survey of control room indications Given the following Unit 1 conditions: Time = 0400

Reactor power = 100%
TDEFDW pump is OOS Time 0401
Both Main FDW pumps trip Time = 0410
1A SG level = 30 inches XSUR stable
lB SG level = 36 inches XSUR increasing

1) (1) will cause the above indications.

2) At 0410 and in accordance with Rule 7 the MAXIMUM EFDW flow allowed to each SG is (2) gpm.

Which ONE of the following completes the statements above? A. 1. A loss of instrument air to 1FDW-316

2. 600 B. 1. A loss of instrument air to I FDW-316

2. 1000 C. 1. Primary to secondary leak in the 1 B SG

2. 600 D. 1. Primary to secondary leak in the 1 B SG

2. 1000 Friday, October 12, 2012 Page 19 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 10 General Discussion Answer A Discussion Incorrect. First part is plausible because a loss of all gas to the valve positioner would cause the valve to fail open. However 1FDW-3l6 is backed upped with Nitrogen. Second part is correct. Answer B Discussion Incorrect. First part is plausible because a loss of all gas to the valve positioner would cause the valve to fail open. However IFDW-316 is 4ppe1with Nitrogen. Second part is plausible because it would be correct if the TDEFDW pump were operating.. Answer C Discussion Correct. The SG levels should be controlling at 30 inches XS{JR. lB SG level is increasing this is cause by the tube leakage. The EFDW flow limit is 600 gpm per header. This the limit for a MDEFDW pump. Answer 0 Discussion Incorrect. First part is correct. Second part is plausible because it would be correct if the TDEFDW pump were operating.. Basis for meeting the KA Question requires evaluation of control indications to determine if a SG has a tube leak. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-SGTR AP/3 I I EOP Rules EPEO38 EA2.07 Steam Generator Tube Rupture (SGTR) Ability to determine or interpret the following as they apply to a SGTR: (CFR 43.5 / 45.13) Plant conditions, from survey of control room indications 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 20 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 11 L APEO54 AK1.01 Loss of Main Feedwater (MFW) Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): (CFR 41.8 I 41.10 I 45.3 MFW line break depressurizes the S/G (similar to a steam line break) Given the following Unit 1 conditions: Time = 0200

Reactor power 100%

Time = 0202

lA SG pressure = 100 psig decreasing
IB SG pressure = 810 psig decreasing
RB pressure = 3.6 psig increasing
PZR level = 9 inches decreasing
RCS pressure = 1610 psig decreasing
CETCs = 520°F and decreasing A (1) has occurred and (2)

Which ONE of the following completes the statement above? ASSUME NO OPERATOR ACTION A. 1. Small break LOCA

2. ES Channels 3 and 4 have actuated B. 1. Small break LOCA

2. ES Channels 3 and 4 have NOT actuated C. 1. Main FDW line break

2. AFIS will have actuated to secure the Main FDW pumps D. 1. Main FDW line break

2. AFIS will NOT have actuated to secure the Main FDW pumps Friday, October 12, 2012 Page 21 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE C 1LT42 ONS SRO NRC Examination QUESTION 11 General Discussion Answer A Discussion Incorrect. First part is plausible because High containment pressure would be reached over a period of time, depending on break size. Pressurizer level would decrease due to the leak. RCS pressure would also decrease based on the leak. However CETCs would not decrease based on a small break LOCA. Second part is correct. Answer B Discussion Incorrect. First part is plausible because High containment pressure would be reached over a period of time, depending on break size. Pressurizer level would decrease due to the leak. RCS pressure would also decrease based on the leak. However CETCs would not decrease based on a small break LOCA. Second part is plausible because ES channels 3 and 4 would not have actuated on low RCS pressure. Answer C Discussion Correct. Indications are present that indicate an Excessive heat transfer which is resulting in an overcooling event. A FDW line break in the RB would result in a plant iesponse similar to a Main Steam Line break. AFIS will actuate which will secure both Main FDW pumps. Answer D Discussion rect.First part correct. Second part is plausible if the candidate has the misconception that AFIS has not actuated. Basis for meeting the KA Question requires the diagnosis of a MFW line break and the resulting loss of Main FDW pumps. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ONS 2007 RO RETEST QlO Development References ______________________ Student References Provided EAP-EHT CF-FDW R43 ONS 2007 RO RETEST QlO APEO54 AK1.0l Loss of Main Feedwater (MFW) Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): (CFR 41.8 / 41.10/ 45.3) MFW line break depressurizes the S/G (similar to a steam line break) 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 22 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 12 12j EPEO55 EA1.05 Loss of Offsite and Onsite Power (Station Blackout) Ability to operate and monitor the following as they apply to a Station Blackout: (CFR 41.7 / 45.5 / 45.6) Battery, when approaching fttiiy discharged Given the following Unit 1 conditions: Initial conditions:

Time 0400
A station blackout has occurred Current conditions:
Time= 0845
1CA bus voltage = 105 VDC At 0845 and in accordance with the Blackout tab, 1CC-8 (1)_ required to be failed closed because it fails (2)

Which ONE of the following completes the statement above? A. 1. is NOT

2. closed on a loss of IA B. 1. is NOT

2. closed on a loss of DC power C. 1.is

2. open on a loss of IA D. 1.is

2. open on a loss of DC power Friday, October 12, 2012 Page 23 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 12 General Discussion Answer A Discussion Incorrect. First part is plausible because it would be correct if less than three hour had elapsed or I CA bus voltage were greater than 105 vdc. Second part is correct. Answer B Discussion Incorrect. First part is plausible because it would be correct if less than three hour had elapsed or ICA bus voltage were greater than 105 vdc. Second part is plausible because 1CC-8 fails closed on a loss oflA. Answer C Discussion Incorrect. First part is correct. Second part is plausible because it fails open on a loss of DC to the solenoid but closed on loss of IA. The candidate could be confused and get them backwards. Answer D Discussion Correct. Per the BO tab if a station blackout has lasted greater than 3 hours and ICA bus voltage is approximately 105 vdc then 1CC-8 is required to be closed. ICC-8 fails open on a loss of DC power to the solenoid.. Basis for meeting the KA Question requires knowledge of an action to be taken due to low DC voltage during a blackout. The ICC-8 fails open on a loss of DC but closed on a loss ofIA. Basis for Hi Cog 3asis for SRO only Job Level Cognitive Level LQ5ti0Typ.e Question Source RO Comprehension NEW Development References Student References Provided EAP-BO RIO EOP Blackout tab EPEO55 EA1 .05 Loss of Offsite and Onsite Power (Station Blackout) Ability to operate and monitor the following as they apply to a Station Blackout: (CFR 41.7 / 45.5 / 45.6) Battery, when approaching fully discharged 401-9 Comments: RemarksIStatus Friday, October 12, 2012 Page 24 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 13 l3 APEO58 AK3.O1 Loss of DC Power Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: (CFR4I.5,41.lO/45.6 / 45.1) Use of dc control power by D/Gs Given the following plant conditions:

No Keowee Units are operating
Keowee Battery Charger #1 output is zero

1) The normal power supply for Battery Charger #lis 1XA fed from (1) .

2) The reason DC power is used for Keowee control power is that it will be available fora MINIMUM of approximately (2) hour(s) following a loss of ALL AC power.

Which ONE of the following completes the statements above? A. 1. 1TC

2. one B. 1. 1TC

2. four C. 1. the 230 KV switchyard

2. one D. 1. the 230 KV switchyard

2. four Friday, October 12, 2012 Page 25 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 13 13j General Discussion Answer A Discussion Incorrect. First part is plasuible because ITC is the backup power supply to IXA. Second part is coorect. Answer B Discussion Incorrect. First part is plausible because ITC is the backup power supply to 1XA. Second part is plausible because 4 hours is a common TS competion time and may be confused with how long the battery will last. Answer C Discussion Correct. With Keowee Unit 1 not operating the power supply for the battery charger is the 230 KV switchyard. The Keowee batteries will last about 1 hour. Answer D Discussion Incorrect. First part is correct. Second part is plausible because 4 hours is a common TS competion time and may be confused with how long the battery will last. Basis for meeting the KA Quest requires knowledge of the reason DC power is used for Keowee Control power. At Oconee we use the Keowee Hydro units instead of DIGs. Basis for Hi Cog Basis for SRO only Job Level RU Cognitive Level Memory L QL5ti0Type NEW L Question Source Development References Student References Provided EL-DCD APEO58 AK3.0l Loss of DC Power Knowledge of the reasons for the following responses as they apply to the Loss of DC Power: (CFR 41.5,41.10 / 45.6 /45.1) Use of dc control power by D/Gs 401-9 Comments: [RemarkslStatus Friday, October 12, 2912 Page 26 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 14 APEO62 AA1 .05 Loss of Nuclear Service Water Ability to operate and / or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): (CFR 41.7 / 45.5 / 45.6) The CCWS surge tank, including level control and level alarms, and radiation alarm Which ONE of the following would be an indication that an extended loss of LPSW has occurred? A. Spent Fuel Pool level increase B. Component Cooling Surge tank level increase C. High Pressure Injection pump motor temperature increase D. Primary instrument air compressor discharge temperature increase Friday, October 12, 2012 Page 27 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 14 l4 General Discussion Answer A Discussion Incorrect. Plausible because level would increase if cooling were lost. However the SFP is cooled by RCW. Answer B Discussion Correct. LPSW cools CC which when LPSW is lost CC temperature will increasing causing the water to swell and the CC surge tank level to increase. Answer C Discussion Incorrect. Plausible because HPT pumps are cooled by LPSW but are automatically backed up by HPSW. Answer D Discussion Incorrect. Plausible because Primary IA compressor is cooled by HPSW. Basis for meeting the KA Question requires knowledge of how a loss ofLPSW would affect CC surge tank level. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided FH-SFC PNS-CC SSS-IA PNS-HPI APEO62 AA1.05 Loss of Nuclear Service Water Ability to operate and / or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): (CFR 41.7! 45.5 / 45.6) The CCWS surge tank, including level control and level alarms, and radiation alarm 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 28 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 15 l5 APEO65 2.1.23 - Loss of Instrument Air APEO65 GENERIC Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6) Given the following Unit I conditions:

Reactor power 100%
Instrument Air Pressure decreasing
AP/22 (Loss of Instrument Air) initiated Current conditions:
Instrument Air pressure 61 psig decreasing
FDW Pump delta P OAC alarms actuate
1A & 1 B Main FDW Pump speeds are both increasing Which ONE of the following describes the actions required by AP/22?

A. Commence a plant shutdown. If at any time two or more CRD temperatures are >180°F, then trip the reactor. B. Commence a plant shutdown, If at any time SG level approaches main FDW pump trip criteria, then trip the reactor. C. Manually trip the reactor. Manually trip both main FDW pumps. D. Manually trip the reactor. Take both FDW Masters to Hand and decrease demand to zero. Friday, October 12, 2012 Page 29 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 15 is General Discussion Answer A Discussion Incorrect: Plausible in that a rapid unit shutdown is required; however the method directed is a manual Rx trip. Second part is plausible as it is the immediate trip criteria for loss of CC flow to the CRDMs during the loss of TA is valid. Answer B Discussion Incorrect: Plausible in that a rapid unit shutdown is required; however the method directed is a manual Rx trip. Second part is plausible in that OMP 1-18 dictates a Manual Rx Trip and tripping of both MFWPS if any SG reaches >96% on the OR level. Answer C Discussion Correct. AP/22 requires the reactor to be tripped when FDW is not controllable. The OAC alarm actuates at about 30 psig, well below the 65 psig where FDW valves can stop responding to control signals. Applicants need to know when the OAC alarm actuates. (See referenced SAEL and Step 4.3 of the AP) Therefore, the AP requires that the reactor be tripped and the MFDW pumps to be tripped. Answer D Discussion Incorrect: AP/22 requires MFDW pumps to be tripped immediately after the reactor is manually due to loss of FDW controllability. Plausible in that the candidate could erroneously think that Feedwater control valves (and FDW demand) would still be controllable if taken to Hand on the ICS stations. Basis for meeting the KA Question tests knowledge of AP/22 (Loss Of IA) during a loss of IA. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ONS 2009 RO Q#l6 bevelopment References Student References Provided AP/22 (Loss of Instrument Air) 555-IA, SAE-L 035 ONS 2009 RO Q#16 APEO65 2.1.23 - Loss of Instrument Air APEO65 GENERIC Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 /43.5 /45.2 / 45.6) 4O19 Comments: Remarks/Status Friday, October 12, 2012 Page 30 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 16 APEO77 AA1 .03 Generator Voltage and Electric Grid Disturbances Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: (CFR: 41.4, 41.5, 41.7,41.10/45.8) Under-excitation Given the following Unit 1 conditions: Initial conditions:

Reactor power = 80%
AP/34 (Degraded Grid) has been entered Current conditions:
Generator output voltage 17.85 KV
Generator output = 700 MWs and 450 MVARs
Generator Hydrogen pressure = 57 psig stable

1) The generator _(1) operating within the limits of AP/34 Enclosure 5.2 (Generator Reduced Capability Curve).

2) If generator output voltage cannot be maintained within the capability curve, AP/34 will direct (2)

Which ONE of the following completes the statements above? REF ERENCE PROVIDED A. 1.is

2. manually tripping the reactor and going to the EOP B. 1.is

2. opening PCB 20 and 21 and going to AP/1 C. 1. is NOT

2. manually tripping the reactor and going to the EOP D. 1. is NOT

2. opening PCB 20 and 21 and going to AP/1 Friday, October 12, 2012 Page 31 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE C 1LT42 ONS SRO NRC Examination QUESTION 16 General Discussion Answer A Discussion Incorrect. First part is plausible because it would be true if the next highest curve used (or if interpolated curve used for 17.85) vs the next lowest. Second part is correct. Answer B Discussion Incorrect. First part is plausible because it would be true if the next highest curve used (or if interpolated curve used for 17.85) vs the next lowest.. Second part is plausible because it would be correct if power were less than 50%. Answer C Discussion Correct. The generator is under excited and operating outside the limits of the capability curve. With power greater than 50% AP134 directs tripping the reactor and going to the EOP. AnswerD Discussion Incorrect. First part is correct. Second part is plausible because it would be correct if power were less than 50%. Basis for meeting the KA Question requires knowledge of generator under-excitation and required actions if it cant be corrected. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided AP/34 End. 5.2 AP/34 APEO77 AA1 .03 Generator Voltage and Electric Grid Disturbances Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: (CFR: 41.4, 41.5, 41.7, 41.10 /45.8) Under-excitation 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 32 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 17 BWEO4 EK1.2 Inadeqiate Heat Transfer Knowledge of the operational implications of the following concepts as they apply to the (Inadequate Heat Transfer): (CFR: 41.8 /41.10 / 45.3) Normal, abnormal and emergency operating procedures associated with (Inadequate Heat Transfer). Given the following Unit 1 conditions:

Time = 0400
Reactor power 100%
lCHPlpumpOOS Current conditions:
Time = 0405
1A and lB Main FDW pumps tripped
Condensate Booster Pumps unavailable
All EFDW pumps unavailable
IA and lB SG Outlet pressure = 860 psig slowly decreasing
RCS pressure 2317 psig increasing Which ONE of the following describes the required operator action(s) in accordance with the EOP?

A. Establish SSF ASW flow to the SG and establish SG levels at 240 inches. B. Establish SSF ASW flow to the SG and do NOT establish a level in the SGs. C. Establish HPI forced cooling and open 1HP-410. D. Establish HPI forced cooling and open I HP-409. Friday, October 12, 2012 Page 33 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 0 1LT42 ONS SRO NRC Examination QUESTION 17 General Discussion Answer A Discussion Incorrect. Will not be required with adequate HPI flow. Plausible because it would be correct if HPI is considered degraded. If there is only I HPIP operating then actions are taken to align SSF ASW to feed the SGs. 240 inches is level directed by Rule 7 when feeding from SSF-ASW Answer B Discussion Incorrect. Will not be required with adequate HPI flow. Plausible because aligning SSF-ASW would be correct if HPI is considered degraded (OnlY I HPIP available). Not establishing a level is plausible since it is consistent with EOP guidance on feeding a dry SG with feedwater. Answer C Discussion Incorrect. HP-410 will not establish flow in the B header. Plausible because HP-410 is the cross over valve for the A HPI header and valve sequence is reversed. Answer D Discussion Correct. Conditions are met to establish HPI F/C and opening II-TP-409 will establish flow in the B header. Basis for meeting the KA res_knowledge of appropriate mitigation strategy contained in plant procedures for inadequate heat transfer conditions. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ONS 2009A RO Q17 Development References Student References Provided EAP-LOHT Rule 4 Rule 3 BWEO4 EK1 .2 Inadequate Heat Transfer Knowledge of the operational implications of the following concepts as they apply to the (Inadequate Heat Transfer): (CFR: 41.8 /41.10 /45.3) Normal, abnormal and emergency operating procedures associated with (Inadequate Heat Transfer). 401-9 Comments: Remarks/Status 1 Friday, October 12, 2012 Page 34 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 18 BWEO5 EK3.2 Excessive Heat Transfer Knowledge of the reasons for the following responses as they apply to the (Excessive Heat Transfer) (CFR: 41.5 /41.10,45.6.45.13) Normal, abnormal and emergency operating procedures associated with (Excessive Heat Transfer). Given the following Unit I conditions: Initial conditions:

Reactor power= 100%
1SA2/A9 (MS PRESS HIGH/LOW) alarms Current conditions:
1A SG pressure = 905 psig increasing
1 B SG pressure = 22 psig decreasing
RCS temperature 530 °F increasing
The overcooling has been stopped

1) In accordance with Rule 5 (MSLB), TBVs will initially be adjusted to maintain (1) constant.

2) The reason for maintaining the above temperature constant is to help prevent (2)

Which ONE of the following completes the statements above? A. 1. CETCs

2. pressurizer swell and subsequent RCS re-pressurization B. 1. CETCs

2. compressive stresses in the isolated SG C. 1. Tcolds

2. pressurizer swell and subsequent RCS re-pressurization D. 1. Tcolds

2. compressive stresses in the isolated SG Friday, October 12, 2012 Page 35 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 18 A General Discussion Answer A Discussion borrect. Rule 5 directs maintaining CETCs constant. This is to reduce Pzr level increase and subsequent RCS re-pressurization. Answer B Discussion 1 Incorrect. First part is correct. Second part is plausible because SG stresses are a concern after isolating a SG. However in this case the stresses will be tensile instead of compressive. Answer C Discussion Incorrect. First part is plausible because Tc is the temperature that is normally used for RCS temperature control. Second part is correct. Answer D Discussion Incorrect. First part is plausible because Tc is the temperature that is normally used for RCS temperature control. Second part is plausible because SG stresses are a concern after isolating a SG. However in this case the stresses will be tensile instead of compressive. Basis for meeting the KA Requires knowledge of the reason an action is taken in Rule 5 as a result of excessive heat transfer. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-EHT EOP Rule 5 BWEO5 EK3.2 Excessive Heat Transfer Knowledge of the reasons for the following responses as they apply to the (Excessive Heat Transfer) (CFR: 41.5 / 41.10, 45.6, 45.13) Normal, abnormal and emergency operating procedures associated with (Excessive Heat Transfer). 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 36 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 19 APEOO3 AK2.05 Dropped Control Rod Knowledge of the interrelations between the Dropped Control Rod and the following: (CFR 41.7 / 45.7) Control rod drive power supplies and logic circuits Given the following Unit 1 conditions: Time = 0400

Reactor power 100%

Time 0415

Control Rod Group 3 Rod 2 position = 50% withdrawn
Control Rod Group 3 out limit light is NOT illuminated
1SA-O2JB-1 0 (CRD Asymmetric Rod Alarm) actuates Time = 0500
Voltage of one phase of the input to the Single Rod Power Supply is greater than the High High overvoltage setpoint

1) At 0415, an Asymmetric Rod Runback (1) occur.

2) At 0500, a reactor trip (2) occur.

Which ONE of the following completes the statements above? A. 1.will

2. will B. 1. will

2. will NOT C. 1. will NOT

2. will D. 1. will NOT

2. will NOT Friday, October 12, 2012 Page 37 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 19 General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because a High High over voltage on two phases will result in all four AC CRD breakers opening._This would cause a reactor trip. Answer B Discussion Correct. On a safety group an asymmetric rod alarm with a loss of the out limit light will result in a run back. Greater than two phases above the higher overvoltage setpoint will open the associated AC breakers. The alternate supply will still be energizing the CRD and reactor trip will not occur. Answer C Discussion Incorrect. First part is plausible because a rod in limit is required to get a runback on controlling rod group. Second part is plausible because a High High over voltage on two phases will result in all four AC CRD breakers opening. This would cause a reactor trip.p. Answer D Discussion Incorrect. First part is plausible because a rod in limit is required to get a runback on controlling rod group. Second part is correct. Basis for meeting the KA Question requires knowledge of the Asymmetric Rod ninback logic and the CRD power supplies. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided IC-CRI APEOO3 AK2.05 Dropped Control Rod Knowledge of the interrelations between the Dropped Control Rod and the following: (CFR 41.7 / 45.7) Control rod drive power supplies and logic circuits 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 38 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 20 201 APEOO5 2.4.4 Inoperable/Stuck Control Rod APEOO5 GEITERIC Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) Given the following Unit 1 conditions:

Power escalation in progress
Reactor power = 60% increasing Which ONE of the following would require entry into AP/1 (Unit Runback)?

A. High discharge pressure results in a lB Main Feedwater Pump trip B. A stuck Group 7 control rod resulting in a 8 percent misalignment from the group average C. A Condensate Booster pump trips resulting in Main FDW pump suction pressure of 350 psig D. A faulty relay results in PCB 21 tripping open Friday, October 12, 2012 Page 39 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 20 B General Discussion Answer A Discussion Incorrect. Plauible because it would be correct if above 65% power. Answer B Discussion Correct. A control rod that is greater than 6.5% (9) misaligned from its group average will result in an asymmetric fault. This will require entry into API1. Answer C Discussion incorrect. Plausible because a FDW pump suction pressure of less than 360 psig will cause Powdex system to bypass. (lC-14/l5 will open) and less than 235 psig will cause runback. Answer D Discussion Incorrect. Plausible because BOTH generatior output breakers open would cause a runback and require entry in to the AP. Basis for meeting the KA Question requires knowledge of AP/l entry conditions based on stuck control rod. Basis for Hi Cog Basis for SRO only Job Level__L Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided 1 A P/l CF-C IC-CRI ARG 1 SAO2-Al I APEOO5 2.4.4 Inoperable/Stuck Control Rod APEOO5 GENERIC Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 40 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 21 21 APEO28 AA2.04 Pressurizer (PZR) Level Control Malfunction Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: (CFR: 43.5 / 45.13) Ammeters and running indicators for CVCS charging pumps Given the following Unit 1 conditions: Initial conditions:

Reactor power 100%
Pzr level Channel 3 is selected
SASS in MANUAL Current conditions:
Pzr temperature B fails LOW

1) The operating HPI pump current (amps) will (1)

2) If SASS had been in AUTO, SASS would have selected Pzr level Channel (2) .

Which ONE of the following completes the statements above? A. 1. increase

2. one B. 1. increase

2. two C. 1. decrease

2. one D. 1. decrease

2. two Friday, October 12, 2012 Page 41 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 21 General Discussion Answer A Discussion Correct. Pzr temperature B feeds Pzr level channel 3. With SASS in MANUAL selected Pzr level will decrease. This will cause 1HP-120 to open. This will cause HPI pump amps to increase, if SASS were in AUTO, SASS would have selected Pzr Level Channel 1. Answer B Discussion Incorrect. First part is correct. Second part is plausible because Pzr Level Channel I and 2 are modified by Pzr temperature A. Answer C Discussion Incorrect. First part is plausible if the candidate has a misconception on the affect of a loss of temperature compensation to Pzr level. Second part is correct. Answer D Discussion rrect. First part is plausible if the candidate has a misconception on the affect of a loss of temperature compensation to Pzr level. Second part is plausible because Pzr Level Channel 1 and 2 are modified by Pzr temperature A. Basis for meeting the KA Question requires knowledge of how HPI pumps amps will respond to a Pzr level malfunction. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW L_____ Development References Student References Provided PNS-PZR R31 APEO28 AA2.04 Pressurizer (PZR) Level Control Malfunction Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: (CFR: 43.5/ 45.13) Ammeters and running indicators for CVCS charging pumps Comments: RemarkslStatus Friday, October 12, 2012 Page 42 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 22 22] APEO32 AK3.02 Loss of Source Range Nuclear Instrumentation Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR 41.5,41.10 / 45.6/45.13) Guidance contained in EOP for loss of source-range nuclear instrumentation Given the following Unit 1 conditions: Initial conditions:

Startup in progress
NI-I and Nl-2 SR 385 cps
NI-i and NI-2 WR = 5E-5%
1NI-3 and INI-4 inoperable Current conditions:
1SA-05/A-7 (1NI-1 Test/Fail) actuates Which ONE of the following describes the required action in accordance with 1SA-05/A-7 and TS 3.3.9 (Source Range Neutron Flux) and why?

A. Restore affected channel to operable status prior to increasing thermal power because, at this power level, the source range is the only reliable direct indication of reactor power B. Restore affected channel to operable status prior to increasing thermal power because, at this power level, WR Nls also provide a reliable indication of reactor power C. Open control rod drive trip breakers immediately because, at this power level, the source range is the only reliable direct indication of reactor power D. Open control rod drive trip breakers within 1 hour because, at this power level, WR NIs also provide a reliable indication of reactor power Friday, October 12, 2012 Page 43 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 22 A General Discussion Answer A Discussion J Correct. Because power is <4E-4% the channel must be restore prior to increasing power. This is because at this power level the source range is the only reliable direct indication of reactor power. Answer B Discussion Incorrect. First part is correct. Second part is plausible because WR Nis are on scale. Answer C Discussion Incorrect. Plausible because it would be correct if both required source ranges were inoperable. Answer D Discussion Incorrect. First part is plausible because it would be correct if both required source ranges were inoperable and the time frame were one hour, Second part is plausible because the WR is on scale. Basis for meeting the KA Oconee does not have actions in the EOP or APs for a failure of source range instrumentation. Discussed with chief examiner. Wrote question on Alarm response Guidance and TS actions including reasons for the actions. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memoiy NEW Development References Student References Provided TS B3.3.9 TS 3.3.9 I SA-05/A-7 APEO32 AK3.02 Loss of Source Range Nuclear Instrumentation Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR 41.5,41.10 / 45.6 /45.13) Guidance contained in EOP for loss of source-range nuclear instrumentation 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 44 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 23 APEO37 AA1 .08 Steam Generator (S/G) Tube Leak Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: (CFR 41.7 / 45.5 / 45.6) Charging flow indicator Given the following Unit 1 conditions: Initial conditions:

Time= 0400
Reactor power = 100%
1RIA-40 in HIGH alarm
1A HPI pump operating
RC Makeup flow = 60 gpm
Letdown Flow 78 gpm
Seal Inlet HDR Flow =32 gpm
Total Seal Return Flow = 8.8 gpm Current conditions:
Time=0415
lBHPlpumpstarted
RC Makeup flow = 160 gpm stable
HPI Flow Train A = 240 gpm stable
Pzr level = 209 inches stable

1) At 0400, (1) will be used to mitigate this event.

2) At 0415, the reactor (2) required to be manually tripped.

Which ONE of the following completes the statements above? A. 1. AP/31 (Primary To Secondary Leakage)

2. is B. 1. AP/31 (Primary To Secondary Leakage)

2. is NOT C. 1. Steam Generator Tube Rupture tab

2. is D. 1. Steam Generator Tube Rupture tab

2. is NOT Friday, October 12, 2012 Page 45 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE B 1LT42 ONS SRO NRC Examination QUESTION 23 23 General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because it would be correct if Pzr level were decreasing with all available HPI. Answer B Discussion Correct. Leak rate MU + SI LD TSR

                    =             -   -

Leak rate 60 + 32 -78 8.8

          =                -

Leak rate 5.2 gpm

Because leak rate less than 25 gpm AP/3 I will be used. In accordance with the SGTR tab the reactor should not be tripped because Pzr level is not decreasing with all available HPI. Answer C Discussion Incorrect. First part is plausible because it would be correct if calculated leak rate were greater than 25 gpm. Second part is plausible because it would be correct if Pzr level were decreasing with all available HPI. Answer D Discussion Incorrect. First part is plausible because it would be correct if calculated leak rate were greater than 25 gpm. Second part is correct. Basis for meeting the KA Question requires using the HPI flow indication to determine the correct procedure to use during a SG tube leak. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW L Development References Student References Provided AP/31 SGTR tab APEO37 AA1.08 Steam Generator (S/G) Tube Leak Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: (CFR 41.7 / 45.5 / 45.6) Charging flow indicator 4019 Comments: RemarkslStatus Friday, October 12, 2012 Page 46 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 24 BWAO2 AK3.4 Loss of NNI-X Knowledge of the reasons for the following responses as they apply to the (Loss of NN1-X) (CFR: 41.5 / 41.10, 45.6, 45.13) RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated. Unit 2 initial conditions:

Reactor power = 90%
Loop A Controlling Thot fails LOW (571°F)
2SA21B4 (RC AVERAGE TEMP HIGH/LOW) actuated Current conditions:
The Diamond and BOTH FDW Masters taken to HAND Which ONE of the following describes the initial action taken by an RO, and the reason for the action, in accordance with OMP 1-18 (Implementation Standard During Abnormal And Emergency Events)?

A. Increase FDW to stabilize reactor power B. Increase FDW to stabilize RCS pressure C. Insert control rods to stabilize RCS pressure D. Insert control rods to stabilize reactor power Friday, October 12, 2012 Page 47 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 24 24] General Discussion Answer A Discussion Incorrect. Plausible because increasing FDW would stabilize reactor power. However OMP 1-18 does not allow increasing FOW during performance of PTR. Answer B Discussion Incorrect. Plausible because increasing FDW would stabilize RCS pressure. However OMP 1-1 8 does not allow increasing FDW during performance of PTR. Answer C Discussion Correct. With Th failing low, indicated Tave decreases and ICS causes control rods to withdraw (based on Tave error) and FDW to decrease in an attempt to restore (indicated) Tave to setpoint. Since actual Tave is increasing, Control rods should be inserted to stop the pressure increase. Answer D Discussion Incorrect: First part is plausible since actual inserting control rods is correct. However per OMP 1 .1 8 RCS pressure is the parameter used to abilize the unit with control rods. Reducing feedwater is used to stabalize reactor power. Basis for meeting the KA 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 ROs to stabilize the plant. Basis for Hi Cog Basis for SRO only Job Level__[_Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided SAE-L074 OMP 1-18 BWAO2 AK3.4 Loss of NNI-X Knowledge of the reasons for the following responses as they apply to the (Loss of NNI-X) (CFR: 41.5 / 41.10,45.6,45.13) RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated. 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 48 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 25 BWAO5 AKI .3 Emergency Diesel Actuation Knowledge of the operational implications of the following concepts as they apply to the (Emergency Diesel Actuation) (CFR; 41.8/41.10, 45.3) Annunciators and conditions indicating signals, and remedial actions associated with the (Emergency Diesel Actuation) Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%
ACB-4 closed
KHU#1 output=88MWe
KHU #2 output = 48 MWe Current conditions:
Switchyard Isolation occurs
2SA-1 7/A-2 (GEN #1 NORMAL LOCKOUT) actuates

1) KHU #1 (1) energize Oconee Unit ls Main Feeder Buses.

2) When the Keowee emergency start signal is reset KHU #1 will (2)

Which ONE of the following completes the statements above? A. twill

2. trip B. 1. will

2. parallel to the grid C. 1. will NOT

2. trip D. 1. will NOT

2. parallel to the grid Friday, October 12, 2012 Page 49 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE A 1LT42 ONS SRO NRC Examination QUESTION 25 General Discussion Answer A Discussion Correct. With a switchyard Isolation ONS Unit I will trip and lose power until it is restored by the Keowee Unit NOT tied to the underground. KHU #1 in this case. A normal lockout will NOT prevent it from operating with an emergency start signal. However with a normal lockout signal present when the emergency start signal is reset the KI-FU will trip. Answer B Discussion Incorrect. First part is correct. Second part is plausible because it would be correct if a normal lockout were not present. Answer C Discussion Incorrect. First part is plausible if the candidate has the misconception that a normal lockout will prevent KI-TU from operating. Second part is correct. Answer D Discussion Incorrect. First part is plausible if the candidate has the misconception that a normal lockout will prevent KHU from operating. Second part is plausible because it would be correct if a normal lockout were not present. Basis for meeting the KA Oconee does not have Diesels for emergency power. This question relates to the Keowee Hydro Units which provide Oconee with emergency power. Question requires knowledge of Statalarms and related action concerning the Keowee units. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EL-KHIJ R9 OP/1106/019 2SA- I 7/A-2 BWAO5 AK1 .3 Emergency Diesel Actuation Knowledge of the operational implications of the following concepts as they apply to the (Emergency Diesel Actuation) (CFR: 41.8/ 41.10, 45.3) Annunciators and conditions indicating signals, and remedial actions associated with the (Emergency Diesel Actuation) .401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 50 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 26 26! BWEO3 EA1.1 Inadequate Subcooling Margin Ability to operate and! or monitor the following as they apply to the (Inadequate Subcooling Margin) (CFR; 41.7 !45.5 !45.6) Components, and functions of control and safety systems, including instnimentation, signals, interlocks, failure modes, and automatic and manual features. Given the following Unit 3 conditions:

C0reSCMO°F
LOCA Cooldown tab in progress
CETCs = 395°F slowly decreasing
3LP-103 (POST LOCA BORON DILUTE) will NOT open Which ONE of the following sets of valves are required to be open in accordance with the LOCA CD tab to establish post LOCA boron dilution flow?

A. 3LP-3AND3LP-19 B. 3LP-3 AND 3LP-20 C. 3LP-104 AND 3LP-19 D. 3LP-104 AND 3LP-20 Friday, October 12, 2012 Page 51 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 26 26 Given the following Unit 3 conditions:

CoreSCM=O°F
LOCA Cooldown tab in progress
CETCs = 395°F slowly decreasing
3LP-103 (POST LOCA BORON DILUTE) will NOT open Which ONE of the following sets of valves are required to be open in accordance with the LOCA CD tab to establish post LOCA boron dilution flow?

A. 3LP-3AND3LP-19 B. 3LP-3AND3LP-20 C. 3LP-104 AND 3LP-19 D. 3LP-104 AND 3LP-20 Friday, October 12, 2012 Page 52 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 26 General Discussion Answer A Discussion Correct. Both 3LP-3 AND 3LP-19 are in the post-loca alternate boron dilution flowpath. Answer B Discussion Incorrect: 3LP-3 is correct. 3LP-20 is plausible since it is on the adjacent RB Emergency Sump header. Answer C Discussion rrect: 3LP-104 is plausible since it is required for flow in the normal post loca boron dilution flowpath. 3LP-19 is correct. Answer D Discussion Incorrect: 3LP-104 is plausible since it is required for flow in the normal post loca boron dilution tlowpath. 3LP-20 is plausible since it is on the adjacent_RB_Emergency Sump header. Basis for meeting the KA Requires knowledge of components and manual features following a Loss of SCM. __________ Basis for Hi Cog Basis for SRO only r Job Level Cognitive Level rJestionType Question Source RO Memory MODIFIED ONS 2009A Q27 elopment References Student References Provided PNS-LPI R27 EOP-LOCACD BDFP drawing ONS 2009A Q27 BWEO3 EA1.1 Inadequate Subcooling Margin Ability to operate and / or monitor the following as they apply to the (Inadequate Subcooling Margin) (CFR: 41.7 / 45.5 / 45.6) Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 53 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examinafion QUESTION 27 BWEO9 EK1 .1 Natural Circulation Operations Knowledge of the operational implications of the following concepts as they apply to the (Natural Circulation Cooldown) (CFR: 41.8 /41.10, 45.3) Components, capacity, and ftinction of emergency systems. Given the following Unit 1 conditions: Initial conditions:

Reactor power 100%

Current conditions:

Condenser vacuum = 19 inches Hg stable
1TA and 1TB de-energized SG levels will be automatically controlled at Which ONE of the following completes the statement above?

A. 25 inches SUR B. 30 inches XSUR C. 50% OR D. 240 inchesXSUR Friday, October 12, 2012 Page 54 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 27 General Discussion Answer A Discussion Incorrect. Plausible because it would be correct if on Main FDW with RCPs. Answer B Discussion Incorrect. Plausible because it would be correct if on EFDW with RCPs. Answer C Discussion Incorrect. Plausible because it would be correct if on Main FDW without RCPs. Answer D Discussion Correct. At 19 inches Hg Main FDW will trip. Without 1TA and ITB (no RCPs) EFDW will control SG level at 240 inches XSUR. Basis for meeting the KA Question requires knowledge of the fuction of EFDW during Natural Circulation. Basis for Hi Cog L_____________________________ Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided CF-EFR37 BWEO9 EK1.l Natural Circulation Operations Knowledge of the operational implications of the following concepts as they apply to the (Natural Circulation Cooldown) (CFR: 41.8 / 41.10, 45.3) Components, capacity, and function of emergency systems. 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 55 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 28 28 SYSOO3 A4.03 Reactor Coolant Pump System (RCPS) Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) RCP lube oil and lift pump motor controls Given the following Unit 1 conditions: Initial conditions:

1A1 RCP start in progress
AC oil lift pump is started Current conditions:
Oil lift pump low discharge pressure does NOT clear

1) The AC oil lift pump (1)

2) In accordance with OP/i/All 103/006 (RCP Operation), in artier to start the 1A1 RCP, the associated (2)

Which ONE of the following completes the statements above? A. 1. must be manually stopped

2. DC oil lift pump will be started B. 1. must be manually stopped

2. Start/Stop switch will be placed in the bypass position C. 1. will automatically stop after a time delay

2. DC oil lift pump will be started D. 1. will automatically stop after a time delay

2. Start/Stop switch will be placed in the bypass position Friday, October 12, 2012 Page 56 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 28 28] General Discussion Answer A Discussion Incorrect. First part is plasuible because some oil lift pumps do not auto stop (ie Main Turbine). Second part is correct. Answer B Discussion Incorrect. First part is plausible because some oil lift pumps do not auto stop (ie Main Turbine). Second part is plausible because the bypass position on the switch will bypass the seal injection requirement but not the oil interlock. Answer C Discussion Correct. The AC Oil Lift pump will stop after 3 minutes and the DC oil lift pump will be started in order to start the RCP. Answer D Discussion Incorrect. First part is correct. Second part is plausible because the bypass position on the switch will bypass the seal injection requirement but oil interlock. Basis for meeting the KA uestion requires knowledge of the RCP oil system. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memoiy NEW Development References Student References Provided PNSCPMR3 OP 1103/006, RCP Operation SYSOO3 A4.03 Reactor Coolant Pump System (RCPS) Ability to manually operate andlor monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) RCP lube oil and lift pump motor controls 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 57 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 29 29] SYS 004 Al .10 Chemical and Volume Control System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: (CFR: 41.5 /45.5) Reactor power Given the following Unit 1 conditions: Time = 0400

Reactor power 95% stable
Reactor Diamond in MANUAL
BOTH FDW Masters in MANUAL
1HP-7 is adjusted to increase Letdown flow from 70 gpm to 95 gpm
Letdown temperature 95°F Time = 0415
Letdown temperature 105°F

1) As a result of the change in letdown temperature reactor power will (1)

2) The setpoint to isolate letdown flow on high letdown temperature is (2) °F.

Which ONE of the following completes the statements above? A. 1. increase

2. 130 B. 1. increase

2. 135 C. 1. decrease

2. 130 D. 1. decrease

2. 135 Friday, October 12, 2012 Page 58 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 0 1LT42 ONS SRO NRC Examination QUESTION 29 General Discussion Answer A Discussion Incorrect. First part is plausible if the candidate has the misconception that an increase in letdown temperature will reduce the RCS Boron concentration. Second part is plausible because it is the setpoint for Letdown temp high statalarm. Answer B Discussion Incorrect. First part is plausible if the candidate has the misconception that an increase in letdown temperature will reduce the RCS Boron concentration. Second part is correct. Answer C Discussion Incorrect. First part is correct. Second part is plausible because it is the setpoint for Letdown temp high statalarm. Answer D Discussion Correct. Increasing letdown temperature will cause the Demin to relapse Boron into this RCS. This will cause reactor power to decrease. The high letdown temperature setpoint is 135 degrees. Basis for meeting the KA Question requires knowledege of how changes in the CVCS will affect reactor power. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided

PNS-HPI R5, R40 SYSOO4 Al.lO Chemical and Volume Control System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: (CFR: 41.5 / 45.5)

Reactor power 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 59 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 30 3O SYSOO4 K6.26 Chemical and Volume Control System Knowledge of the effect of a loss or malfunction on the following CVCS components: (CFR: 41.7 / 45.7) Methods of pressure control of solid plant (PZR relief and water inventoly) Given the following Unit 1 conditions: Initial conditions:

Time 0400
A transient occurred resulting in a reactor trip from 100% power
Pzr level = 400 inches stable
Letdown cannot be established Current conditions:
Time= 0500
Forced Cooldown tab in progress
Pzr level = 400 inches stable
Pzr temperature = 603°F
RCS temperature = 581°F
RCS pressure = 2100 psig decreasing

1) At 0500, RCS pressure will be controlled by adjusting (1)

2) If NO operator action is taken, RCS pressure will stabilize at no lower than (2) psig.

Which ONE of the following completes the statements above? A. 1. HPIflowONLY

2. 1355 B. 1. HPIflowONLY

2. 1595 C. 1. HPI flow OR Pzr heaters

2. 1355 D. 1. HPI flow OR Pzr heaters

2. 1595 Friday, October 12, 2012 Page 60 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 30 30] General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because it is the saturation pressure for 581 degrees. Answer B Discussion Correct. The RCS is solid and subcooled. At this time Pzr heaters will not increase RCS pressure. ONLY HPI flow will control pressure (letdown not available). The saturation temperature for 603 degrees is about 1595 psig. Pressure will decease toward this value. Answer C Discussion Incorrect. First part is plausible because Pzr heaters will normally increase pressure. Second part is plausible because it is the saturation pressure for 581 degrees. Answer D Discussion Incorrect. First part is plausible because Pzr heaters will normally increase pressure. Second part is correct. Basis for meeting the KA Question requires knowledge of how RCS pressure is controlled when the Pzr is solid. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Deveopment References Student References Provided PNS-PZR R22 EOP FCD Tab SYSOO4 K6.26 Chemical and Volume Control System Knowledge of the effect of a loss or malfunction on the following CVCS components: (CFR: 41.7 / 45.7) Methods of pressure control of solid plant (PZR relief and water inventory) 4019 Comments: Remarks/Status Friday, October 12, 2012 Page 61 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 31 SYSOO5 K4.06 Residual Heat Removal System (RHRS) Knowledge of RHRS design feature(s) and/or interlock(s) which provide or the following: (CFR: 41.7) Function of RHR pump miniflow recirculation Given the following Unit 1 conditions:

Reactor in MODE 5
LPI in normal decay removal

1) The MINIMUM allowable flow per LPI pump for unrestricted operation is (1) gpm in accordance with OP/i/Ni 104/004 (Low Pressure Injection System).

2) In accordance with OP/i/Ni 104/004, if operated below the minimum flow rate the associated LPI pump (2)

Which ONE of the following completes the statements above? A. 1. 800

2. must be stopped immediately B. 1. 800

2. can still be operated C. i. 170

2. must be stopped immediately D. 1. 170

2. can still be operated Friday, October 12, 2012 Page 62 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 31 3U General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because this is true after ES with the LPI pumps deadheading. Answer B Discussion Correct. The minimum flow required is 800 gpm per pump. Per the L&P of OP/i 104/004 (LPI System) after exceeding the minimum flow time limit the pump can still be operated but is technically inoperable until performance testing is completed. Answer C Discussion Incorrect. First part is plasuible becaue 170 is the minimum flow for the HPI pumps. Second part is plausible because this is true after ES with the LPI pumps deadheading. Answer D Discussion Incorrect. First part is plasuible becaue 170 is the minimum flow for the HPI pumps. Second part is correct. Basis for meeting the KR The LPI pumps have a fixed recirc line that goes back to the suction of the LPI pumps. As a result the LPI pumps have minimum flow limits. [The question requires knowledge of this limit and actions if not met. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided PNS-LPI R35 OP/llO4/004 SYSOO5 K4.06 Residual Heat Removal System (RHRS) Knowledge of RHRS design feature(s) and/or interlock(s) which provide or the following: (CFR: 41.7) Function of R}IR pump miniflow recirculation 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 63 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 32 SYSOO5 K5.O1 Residual Heat Removal System (RHRS) Knowledge of the operational implications of the following concepts as they apply the RHRS: (CFR: 41.5 / 45.7) Nil ductility transition temperature (brittle fracture) Given the following Unit I conditions: Initial conditions:

ReactorinMODE5
RCS temperature 145°F stable
LPI in normal decay heat removal Current conditions:
RCS pressure is increasing The setpoint at which I RC-66 will open is (1) psig, and this is to prevent (2)

Which ONE of the following completes the statement above? A. 375 overpressurization of the LPI system piping B. 375 brittle fracture of the reactor vessel C. 530 overpressurization of the LPI system piping D. 530 brittle fracture of the reactor vessel Friday, October 12, 2012 Page 64 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 0 1LT42 ONS SRO NRC Examination QUESTION 32 32] General Discussion This procedure provides guidance to comply with Technical Specification (TS) 3.4.12 and SLC 1 6.5.2 as well as provide protection from a low temperature overpressure (LTOP) event. A dedicated LTOP Operator is a compensatory measure allowed by TS 3.4.12 Condition F to monitor for an initiation of an LTOP event. Entry into this Condition is required any time a dedicated LTOP Operator is established. AND HPI pump operating and capable of injecting into RCS via ll-{P-l20 (RC VOLUME CONTROL), Verify RCS pressure within the following limits: HPA logic 4.7.1 WHEN RCS > 220°F and < 325°F, verify RCS < 525 psig. 4.7.2 WHEN RCS < 220°F, verify RCS < 375 psig. The PORV is not required to be OPERABLE when no HPI pumps are running and RCS pressure < 100 psig. An OPERABLE power operated relief valve (PORV) with a lift setpoint of< 535 psig; and Answer A Discussion Incorrect. First part is plausible because 375 psig is the high pressure alarm when RCS temp is below 220°F. Second part is plausible since P1-JR mode uses LPI piping, at some point, increasing pressure would over pressurize LPI piping (true but incorrect bases). Answer B Discussion Incorrect. First part is plausible because 375 psig is the high pressure alarm when RCS temp is below 220°F. Second part is correct. Answer C Discussion Incorrect. First part is correct. Second part is plausible since DHR mode uses LPI piping, at some point, increasing pressure would over pressurize LPI piping (true but incorrect bases). Answer 0 Discussion Correct. The PORV low temperature setpoint is 530 psig and it is to help prevent RV brittle fracture. Basis for meeting the KA Question requires knowledge of how RV brittle feracture is prevented whil on LIP (DHR). Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided CPOI7R8 PNS-PZR R9 OP/I/All 104/049 SYSOO5 K5.01 - Residual Heat Removal System (RHRS) Knowledge of the operational implications of the following concepts as they apply the RHRS: (CFR: 41.5 / 45.7) Nil ductility transition temperature (brittle fracture) 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 65 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 33 331 SYSOO6 K6.1O Emergency Core Cooling System (ECCS) Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: (CFR: 41.7 / 45.7) Valves Given the following Unit 1 conditions: Time = 0400

Reactor trip from 100% power due to a LOCA Time = 0430
RCS pressure = 45 psig slowly decreasing
1LP-17 failed CLOSED

1) The actual RCS pressure setpoint that will cause the LPI pumps to start in the ES mode is (1) psig.

2) At 0430, LPI flow (2) enter the core through BOTH LPI/CFT nozzles.

Which ONE of the following completes the statements above? A. 1. 500

2. will B. 1. 500

2. will NOT C. 1. 550

2. will D. 1. 550

2. will NOT Friday, October 12, 2012 Page 66 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 33 33] General Discussion Answer A Discussion Incorrect. First part is plausible because 500 psig is the TS value LPI injection. Second part is correct. Answer B Discussion Incorrect. First part is plausible because 500 psig is the TS value LPI injection. Second part is plausible because the A LPI header injection valve is closed. Answer C Discussion Correct. 550 psig is the actual setpoint for LPI ES activation. Because of the crossover mod even with ILP-l7 closed flow will still enter the core through both nozzles. Answer D Discussion rect. First part is correct. Second part is plausible because the A LPI header injection valve is closed. Basis for meeting the KA with a valve failure. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided IC-ES I PNS-LPI R49 SYSOO6 K6.1O Emergency Core Cooling System (ECCS) Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: (CFR: 41.7 / 45.7) Valves 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 67 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 34 3 SYSOO7 A4.09 Pressurizer Relief TanklQuench Tank System (PRTS) Ability to manually operate andlor monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Relationships between PZR level and changing levels of the PRT and bleed holdup tank Given the following Unit 1 conditions:

Reactor power = 100%
Leak through 1 RC-66 (0.6 gpm)

Pzr level will (1) and initially Quench Tank (2) will increase. Which ONE of the following completes the statement above? A. 1. remain constant

2. level B. 1. remain constant

2. pressure C. 1. increase

2. level D. 1. increase

2. pressure Friday, October 12, 2012 Page 68 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 34 34 General Discusson Answer A Discussion correct. IHP-120 will mainatin Pzr level constant. QT level will increase. Answer B Discussion Incorrect. First part is correct. Second part is plausible because steam is going into the QT. Answer C Discussion Incorrect. First part is plausible because of the misconception that with a Pzr steam space leak Pzr level will increase. Second part is correct. Answer D Discussion Incorrect. First part is plausible because of the misconception that with a Pzr steam space leak Pzr level will increase. Second part is plausible because steam is going into the QT. Basis for meeting the KA Question requires knowledge of the relationship between Pzr and QT levels. Basis for Hi Cog Basis for SRO only

-]

Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided PNS-PZR R32 SYSOO7 A4.09 Pressurizer Relief TanklQuench Tank System (PRTS) Ability to manually operate and/or monitor in the control room: (CFR: 41.7! 45.5 to 45.8) Relationships between PZR level and changing levels of the PRT and bleed holdup tank 14O19 Comments: RemarkslStatus Friday, October 12, 2012 Page 69 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 35 351 SYSOO8 K1.02 Component Cooling Water System (CCWS) Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.9) Loads cooled by CCWS Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%

Current conditions:

BOTH CC pumps have TRIPPED
RCP seal injection = 8 gpm/pump Which ONE of the following describes the continued operation of the RCPs?

A. RCPs must be secured immediately B. RCPs must be secured within 30 minutes C. RCPs may continue operation indefinitely D. RCPs may operate no longer than 100 hours Friday, October 12, 2012 Page 70 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 35 General Discussion Answer A Discussion rrect, plausible because this would be correct if CC and RCP seal injection were lost. Answer B Discussion Incorrect, plausible because this is the time associated with closing LPSW valves to RCP motors after shutdown. Answer C Discussion Correct, per note in AP/20 Operation of the RCPs without CC may continue indefinitely provided RCP seal injection is functioning properl to maintain pump temperatures within limits of End 5.1 (RCP Immediate Trip Criteria) of AP/l 6 (Abnormal RCP Operation). Answer D Discussion Incorrect, 100 hours is plausible because this is the time the RCPs can operate with a failed seal or seal return or leakage flow. Basis for meeting the KA ioequires knowledge of effect of Loss of CC on operation of the RCPs. j Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK ONS 2007 Retest Q36 Development References Student References Provided AP/20 (Loss of Component Cooling) AP/16 ONS 2007 Retest Q36 SYSOO8 K1.02 Component Cooling Water System (CCWS) Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.9) Loads cooled by CCWS 401-9 Comments: J RemarkslStatus Friday, October 12, 2012 Page 71 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 36 SYSO1O K2.O1 - Pressurizer Pressure Control System (PZR PCS) Knowledge of bus power supplies to the following: (CFR: 41.7) PZR heaters Given the following Unit 1 conditions: Initial conditions:

Reactor power= 100%

Current conditions:

AP/25 (SSF Emergency Operating Procedure) in progress
SSF Aux Service Water pump is supplying BOTH SGs
SSF RC Makeup pump is supplying the RCP seals

1) MCCPXSF is supplying powerto Prz Heaters Bank2Group (1)

2) The Pzr heaters low level cutoff setpoint from the uncompensated SSF Pzr level is (2) inches.

Which ONE of the following completes the statements above? A. 1.C

2. 80 B. 1.C

2. 85 C. 1.D

2. 80 D. 1.D

2. 85 Friday, October 12, 2012 Page 72 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE B 1LT42 ONS SRO NRC Examination QUESTION 36 36j General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because it is the setpoint for the normal Pzr heater low level cutoff. Answer B Discussion Correct. Pzr heater Bank 2 Group C is powered from PXSF. The Pzr heaters low level cutoff from the SSF level instrument is 85 inches. Answer C Discussion incorrect. First part is plausible because Par Heaters bank 2 Group D are heaters used during emergencies and can be controlled from the Aux Shutdown panel. Second part is plausible because it is the setpoint for the normal Pzr heater low level cutoff. Answer 0 Discussion incorrect. First part is plausible because Pzr Heaters bank 2 Group D are heaters used during emergencies and can be controlled from the Aux Shutdown panel. Second part is correct. _______________________________ Basis for meeting the KA Question requires knowedge of power supplies to Pzr heaters. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided PNS-Pzr R7 SYSO1O K2.O1 Pressurizer Pressure Control System (PZR PCS) Knowledge of bus power supplies to the following: (CFR: 41.7) PZR heaters 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 73 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 37 SYSO12 K3.03 Reactor Protection System (RPS) Knowledge of the effect that a loss or malfunction of the RPS will have on the following: (CFR: 41.7 / 45.6) SDS Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%

Current conditions:

1A and lB Main FDW pumps trip
Actuated Statal arms:

1SA-01/E-2 (CRD TRIP BKR. A TRIP)

             > 1SA-01/E-4 (CRD TRIP BKR. C TRIP)
             > 1SA-081C-09 (DSS CH1 ACTUATED)
             > 1SA-08/C-10 (DSS CH2 ACTUATED)

1) Control Rods Groups I - 7 (1) fully inserted into the core.

2) The Turbine Bypass Valves are controlling at (2)

Which ONE of the following completes the statements above? A. 1. are

2. setpoint B. 1. are

2. setpoint plus 125 psig C. 1. are NOT

2. setpoint D. 1. are NOT

2. setpoint pIus 125 psig Friday, October 12, 2012 Page 74 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 37 3 General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because the combination of CRD breakers that are open will not cause the control rods to trip into the core and generate a trip confirm signal. The turbine will have tripped due to the AMSAC portion of AMSAC/DSS on a loss of both Main FDW pumps. With the turbine tripped the TBVs would normally control at setpoint. Answer B Discussion Correct. The combination of CRD breakers that are open will not cause the control rods to trip into the core and generate a trip confirm signal. However since both DSS channels actuated, all rods will drop into the core and a trip confirm signal will be generated. This will add a 125 psig bias to the TBVs setpoint. Answer C Discussion Incorrect. First part is plausible because the combination of CRD breakers that are open will not cause the control rods to trip into the core and generate a trip confirm signal. Second part is plausible because the combination of CRD breakers that are open will not cause the control rods to trip into the core and generate a trip confirm signal. The turbine will have tripped due to the AMSAC portion of AMSAC/DSS on a los of both Main FDW pumps. With the turbine tripped the TBVs would normally control at setpoint. Answer D Discussion Incorrect. First part is plausible because the combination of CRD breakers that are open will not cause the control rods to trip into the core and generate a trip confirm signal. Second part is correct. Basis for meeting the KA Question requires knowledge of how a malfuction of the RPS signal will affect the TBVs. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided STG-ICS IC-CRI IC-RPS R17 1 SA-08/C-9/1 0 SYSO 12 K3.03 - Reactor Protection System (RPS) Knowledge of the effect that a loss or malfunction of the RPS will have on the following : (CFR: 41.7 / 45.6) SDS 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 75 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 38 SYSO12 K5.O1 Reactor Protection System (RPS) Knowledge of the operational implications of the following concepts as the apply to the RPS: (CFR: 41.5 /45.7) DNB

1) The (1) RPS trip is designed to prevent exceeding the DNBR AND the Fuel Center Line Melt safety limit.

2) The actual setpoint for the above RPS trip is (2)

Which ONE of the following completes the statements above? A. 1. Highflux

2. 105.50%RTP B. 1. Highflux

2. 104.75%RTP C. 1. High RCS Temperature

2. 618°F D. 1. High RCS Temperature

2. 617°F Friday, October 12, 2012 Page 76 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 38 38 General Discussion Answer A Discussion Incorrect. First part is coorect. Second part is plasuible because it is the TS value. Answer B Discussion Correct. The High Flux trip is designed to prevent exceeding the DNBR and the Fuel Centerline Safety Limits. The actual setpoint is 104.75% RTP. Answer C Discussion Incorrect. First part is plasuible because it does prevent exceeding the DNBR safety limit. Second part is plausible because it is the TS value for the High RCS Temperature trip. Answer D Discussion Incorrect. First part is plasuible because it does prevent exceeding the DNBR safety limit. Second part is plausible because it is the actual value for the High RCS Temperature trip. Basis for meeting the KA

Question requires knowledge of the relation ship between RPS trips and DNBR.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided IC-RPS R3 SYSOI2 K5.OI Reactor Protection System (RPS) Knowledge of the operational implications of the following concepts as the apply to the RPS: (CFR: 41.5 / 45.7) DNB 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 77 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 39 SYSO13 K1.14 Engineered Safety Features Actuation System (ESFAS) Knowledge of the physical connections and/or cause effect relationships between the ESFAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) lAS Given the following Unit I conditions: Initial conditions:

Time = 0400
Reactor power= 100%
Air line to 1HP-5 actuator is severed Current conditions:
Time=0401
IA Header pressure has decreased to 90 psig
ES 18 actuates

1) At 0401,the Unit I Aux IA Compressor (1) operating.

2) At 0402, 1HP-5 (2) be in its required ES position.

Which ONE of the following completes the statements above? A. 1.is

2. will B. 1. is

2. will NOT C. 1. is NOT

2. will D. 1. is NOT

2. will NOT Friday, October 12, 2012 Page 78 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE C 1LT42 ONS SRO NRC Examination QUESTION 39 General Discussion Answer A Discussion Incorrect. First part is plausible because the Backup IA compressors start at 93 psig. Second part is correct. Answer B Discussion Incorrect. First part is plausible because the Backup IA compressors start at 93 psig. Second part is plausible because some valves (I HP-3 1) do fail open on a loss of IA. Answer C Discussion Correct. The Aux IA Compress will not start until IA pressures reaches 88 psig. IHP-5 fails closed on a loss of IA and its ES position is closed. Answer D Discussion Incorrect. First part is correct. Second part is plausible because some valves (1HP-31) do fail open on a loss of IA. Basis for meeting the KA [Question requires knowledge of the relationship between IA and an ES valve. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided NSHPIR4O AP/22 SYSO13 K1.14 Engineered Safety Features Actuation System (ESFAS) Knowledge of the physical connections and/or cause effect relationships between the ESFAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) lAS 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 79 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 40 SYSO22 A2.03 Containment Cooling System (CCS) Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 / 45.3 / 45.13) Fan motor thermal overload/high-speed operation Given the following Unit 1 conditions: Time = 0400

Reactor power 100%
1A and 1C RBCUs operating in HIGH speed Time = 0401
ES Channel 1 -8 actuates Time 0405
1A and lB RBCUs in LOW speed
10 RBCU in HIGH speed

1) Continued operation in this alignment could damage the 10 RBCU fan (1) due to high humidity in the RB.

2) In accordance with EOP Enclosure 5.1 (ES Actuation) the 1C RBCU will (2)

Which ONE of the following completes the statements above? A. 1. motor

2. immediately be switched to LOW speed B. 1. motor

2. ONLY be switched to LOW speed after directed by the CRS C. 1. blades

2. immediately be switched to LOW speed D. 1. blades

2. ONLY be switched to LOW speed after directed by the CRS Friday, October 12, 2012 Page 80 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE B 1LT42 ONS SRO NRC Examination QUESTION 40 4O General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because normally when automatic actions do not occur the RO is expected to take the action. Answer B Discussion Correct. The high humidity in the RB can damage the RBCU motor if operated in high speed. End. 5.1 directs the RO to inform the CRS prior to taking action to place the RBCU in the proper speed. Answer C Discussion Incorrect. First part is plausible because the candidate could have the misconception that the high humidity would put undue strain on the RBCU blades. Second part is plausible because normally when automatic actions do not occur the RO is expected to take the action. Answer D Discussion Incorrect. First part is plausible because the candidate could have the misconception that the high humidity would put undue strain on the RBCU blades. Second part is correct. Basis for meeting the KA Question requires knowledge of the impact of one RBCU remaining in high speed after ES actuation and the procedural guidance on how to] correct the issue. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided PNS-RBC R5 EOP End. 5.1 SYSO22 A2.03 Containment Cooling System (CCS) Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /45.3 / 45.13) Fan motor thermal overload/high-speed operation 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 81 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 41 41j SYSO22 K3.02 Containment Cooling System (CCS) Knowledge of the effect that a loss or malfunction of the CCS will have on the following: (CFR: 41.7 / 45.6) Containment instrumentation readings Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%
RCS leak in the reactor building
ALL RBCU5 trip and will not start Current conditions:
Reactor has been tripped
EOP subsequent actions in progress
Reactor Building pressure 3.2 psig increasing In accordance with the EOP:

1) Pzr level will be controlled at a MINIMUM of (1) inches.

2) SC level will be controlled at (2) inches S/U level.

Which ONE of the following completes the statements above? A. 1. >100

2. 2535 B. 1. >100

2. 5565 C. 1.>180

2. 2535 D. 1.180

2. 5565 Friday, October 12, 2012 Page 82 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 41 4j General Discussion Answer A Discussion Incorrect. First part is plasuible because it would be correct if RB pressure were less than 3 psig. Second part is plasuible because it would be correct if RB pressure were less than 3 psig. Answer B Discussion Incorrect. First part is plasuible because it would be correct if RB pressure were less than 3 psig. Second part is correct. Answer C Discussion Incorrect. First part is correct.. Second part is plasuible because it would be correct if RB pressure were less than 3 psig. Answer 0 Discussion Correct. Per the EOP with RB pressure greater than 3 psig Pzr level will be controlled> 180 inches and SG level will be controlled from 55 65 - inches. Basis for meeting the KA Question requires knowledge of how a loss of RB cooling in conjuction with an RCS leak would affect Pzr and SG level and where levels would then be controlled. This is due to fact that the level instruments are not temperature compensated. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EOP End. 5.5 EOP SA SYSO22 K3.02 Containment Cooling System (CCS) Knowledge of the effect that a loss or malfunction of the CCS will have on the following: (CFR: 41.7 / 45.6) Containment instrumentation readings 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 83 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 42 SYSO26 K2.Ol Containment Spray System (CSS) Knowledge of bus power supplies to the following: (CFR: 41.7) Containment spray pumps Given the following Unit I conditions: Initial conditions:

Time0400
Reactor power 20%

Current conditions:

Time= 0401
RB pressure = 14 psig increasing
lTDde-energized The B RBS pump (1) operating and the B HPI pump (2) operating.

Which ONE of the following completes the statement above? A. 1.is

2. is B. 1.is

2. is NOT C. 1. is NOT

2. is D. 1. is NOT

2. is NOT Friday, October 12, 2012 Page 84 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 42 c 42 General Discussion Answer A Discussion Incorrect, first part is plausible because it is true for the HPT pumps. Second part is correct. Answer B Discussion Incorrect, first part is plausible because it is true for the HPI pumps. Second part is plausible because it is true for the LPI pumps.. Answer C Discussion Correct, ES 1-8 will have actuated. lTD is de-energized which will prevent the lB RBS pump and the 1C HPI pump from starting. So lB RBS will be off and the lB HPI pump will be operating. Answer D Discussion Incorrect, first part is correct. Second part is plausible because it is true for the LPI pumps.. Basis for meeting the KA Question requires knowledge of RBS pump power supplies. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED ONS 2007 RO RETEST Q42 Development References Student References Provided IC-ES ONS 2007 RO RETEST Q42 SYSO26 K2.01 Containment Spray System (CSS) Knowledge of bus power supplies to the following: (CFR: 41.7) Containment spray pumps 401-9 Comments: RemarkslStatus Friday, October 12, 2912 Page 85 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 43 SYSO26 K3.02 Containment Spray System (CSS) Knowledge of the effect that a loss or malfunction of the CSS will have on the following: (CFR: 41.7 / 45.6) Recirculation spray system Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%
1TC locked out
1BS-2 breaker is open and will not close Current conditions:
ES 1 8 actuate due to a LBLOCA When the RBS system is placed in the recirculation mode:

1) Its purpose isto (1)

2) The RBS system (2) be able to perform its safety function.

Which ONE of the following completes the statements above? A. 1. entrain Iodine thus reducing offsite dose

2. will B. 1. entrain Iodine thus reducing offsite dose

2. will NOT C. 1. minimize hydrogen production due to Zirc water reaction

2. will D. 1. minimize hydrogen production due to Zirc water reaction

2. will NOT Friday, October 12, 2012 Page 86 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 43 Gener& Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because a candidate could have the misconception that either I BS-2 is normally open (lilcelHP-27) or that the A pump is powered from lTD. Answer B Discussion Correct. In the recirc mode RBS entrains Iodine to limit offsite dose due to RB leakage. With 1TC locked out and IBS-2 breaker open neither RBS will function lBS 2 is normally closed and thus will not go to its ES position Thus its safety function will not be met Answer C Discussion Incorrect. First part is plausible because it is true for Zinc and aluminum reaction. Second part is plausible because a candidate could have the misconception that either 1BS-2 is normally open (like IHP-27) or that the A pump is powered from lTD. Answer 0 Discussion Incorrect. First part is plausible because it is true for Zinc and aluminum reaction. Second part is correct. Basis for meeting the KA Question requires knowledge about how failures of the RBS system will affect the system when placed in the recirc mode. Basis for Hi Cog Basis for SRO only Job Leve Cognitive Level QuestionType Question Source RO Comprehension NEW Deveopment References Student References Provided PNS-BS Rl SYSO26 K3.02 Containment Spray System (CSS) Knowledge of the effect that a loss or malfunction of the CSS will have on the following: (CFR: 41.7! 45.6) Recirculation spray system 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 87 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 44 4] SYSO39 K5.08 Main and Reheat Steam System (MRSS) Knowledge of the operational implications of the following concepts as the apply to the MRSS: (CFR: 441.5 / 45.7) Effect of steam removal on reactivity Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%
1MS-1 12 & 1MS-173 (SSRH lA/lB Controls) are open and in MANUAL
1MS-77, 78, 80, 81 (MS to SSRHs) are open and in AUTO Current conditions:
Reactor trip
Station blackout

1) 1MS-112&1MS-173wi11 (1)

2) As a result of the loss power to 1 MS-77, 78, 80, 81, positive reactivity (2) be added to the reactor.

Which ONE of the following completes the statements above? A. 1. close

2. will B. 1. close

2. will NOT C. 1. remain open

2. will D. 1. remain open

2. will NOT Friday, October 12, 2012 Page 88 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE A 1LT42 ONS SRO NRC Examination QUESTION 44 44 General Discussion Answer A Discussion Correct. IMS-l 12 and IMS-173 will close on a loss of power even though they are in manual. IMS-77 through IMS-81 will not close on a loss of power and will cause a positive reactivity addition to occur. Answer B Discussion Incorrect. First part is correct. Second part is plausible if the candidate has the misconception that all of the valves close on a turbine trip with a loss of power or that the valves are in series and IMS-112/117 will stop the steam flow. Answer C Discussion Incorrect. First part is plausible because the valves are in manual. Second part is correct. Answer D Discussion Incorrect. First part is plausible because the valves are in manual. Second part is plausible if the candidate has the misconception that all of the valves close on a turbine trip with a loss of power or that the valves are in series and IMS-l12/117 will stop the steam flow.. Basis for meeting the KA Question requires knowledge of steam supply valves to the SSRH and the resultant affect on core reactivity. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source

RO Comprehension NEW Development References Student References Provided STG-MSR R17 SYSO39 K5.08 Main and Reheat Steam System (MRS S)

Knowledge of the operational implications of the following concepts as the apply to the MRSS: (CFR: 441.5 / 45.7) Effect of steam removal on reactivity 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 89 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 45 SYSO59 A3.03 Main Feedwater (MFW) System Ability to monitor automatic operation of the MFW. including: (CFR: 41.7 / 45.5) Feedwater pump suction flow pressure Given the following Unit 3 conditions: Initial conditions:

04:00:00
Reactor power = 70% stable
3A Main Feed Pump suction pressure 236 psig decreasing Current conditions:
04:01:25
3A Main FDW Pump suction pressure = 230 psig increasing At 04:01:25, which ONE of the following describes the status of the Main Feedwater pumps and the plant?

A. Both MFPs are operating / Plant run back in progress at 20%/rn in B. Both MFPs are operating I Reactor power is 70% stable C. 3A MFP has tripped / Plant run back in progress at 20%/mm D. 3A MFP has tripped / Reactor power is 65% stable Friday, October 12, 2012 Page 90 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE A 1LT42 ONS SRO NRC Examination QUESTION 45 45! General Discussion Answer A Discussion Correct: Conditions are met for a CBPIFdw Pump Low Suction Pressure runback (<235 psig; for >90 sec MFP trips; Runback rate = 20%/mm.). Only 85 secs has elapsed therefore runback is still in progress and MFPT is not tripped. Answer B Discussion Incorrect: Plausible in that both MFPs are still operating (< 90 secs); second part is plausible as it is the current power level and is consistent with the FDWP operating status (i.e. no runback to 65% due to FDWPT Trip) Answer C Discussion Incorrect: Plausible if the time delay is incorrectly assumed to be met and the MFP has tripped. Runback status is correct. Answer D Discussion Incorrect: Plausible if the time delay is incorrectly assumed to be met and the MFP has tripped. Second part is consistent with MFP status and the runback is incorrectly assumed to have stopped at the loss of MFP runback setpoint Basis for meeting the KA Requires knowledge of setpoints for MFP low suction pressure runback / trip and the ability to determine the status of the plant and MFW system components _________________________________ Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ONS 2009 Q43 Development References Student References Provided CF-FDW STG-ICS ONS 2009 Q43 SYSO59 A3.03 Main Feedwater (MFW) System Ability to monitor automatic operation of the MFW, including: (CFR: 41.7/45.5) Feedwater pump suction flow pressure 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 91 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 46 46 SYSO61 A3.03 - Auxiliary / Emergency Feedwater (AFW) System Ability to monitor automatic operation of the AFW. including: (CFR: 41.7 / 45.5) AFW S/G level control on automatic start Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%
PCB-21 is OPEN Current conditions:
Reactor trip due to spurious turbine trip Steam Generator levels will be controlled at Which ONE of the following completes the statement above?

A. 25 inches SUR B. 30 inches XSUR C. 240 inchesXSUR D. 50% OR Friday, October 12, 2012 Page 92 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE B 1LT42 ONS SRO NRC Examination QUESTION 46 46] General Discussion Answer A Discussion Incorrect. Plausible because it would be correct for a normal trip with Main FDW operating. Answer B Discussion Correct. With PCB-21 open a slow transfer will occur when the reactor trips. This means there is a 1.7 second delay before the E breakers close. This will cause a loss of all secondary pumps and EFDW pumps will start. SG level will be controlled at 30 inches XSUR. Answer C Discussion Incorrect. This is plausible if the candidate also believed that RCPs were also lost. They do have a slow transfer be it is not long enough for a loss of the RCPs. Answer D Discussion Incorrect. Plausible because it would be correct if main FDW were oeprating and RCPs were off. Basis for meeting the KA Question requires knowledge of where SG level would be controlled after EFDW has started for a given plant condition. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension Development References Student References Provided .EL-EPD R30 CF-EF SYSO61 A3.03 Auxiliary / Emergency Feedwater (AFW) System Ability to monitor automatic operation of the AFW, including: (CFR: 41.7 / 45.5) AFW S/G level control on automatic start 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 93 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 47 SYSO62 2.2.12 AC Electrical Distribution System SYSO62 GENERIC Knowledge of surveillance procedures. (CFR: 41.10 145.13) Given the following Unit 1 conditions:

ReactorinMODE5
OPI1IAII1O5/014 (Control Room Instrumentation Operation And Information)

Enclosure 4.4 (Mode 5) in progress

1) (1) is allowed to be credited as a required Offsite Power Source.

2) (2) is allowed to be credited as a required Emergency Power Source.

Which ONE of the following completes the statements above? A. 1. CT-5 (LCT)

2. CT-4 B. 1. CT-5(LCT)

2. Backcharged Main Transformer C. 1. CT-5 (Central Swyd)

2. CT-4 D. 1. CT-5 (Central Swyd)

2. Backcharged Main Transformer Friday, October 12, 2012 Page 94 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 47 General Discussion Answer A Discussion Incorrect. Plausible because the LCTs are offsite. Second part is correct. Answer B Discussion Incorrect. Plausible because the LCTs are offsite. Second part is plausible because a backcharged transformer could be used during an emergency. Answer C Discussion Correct. CT-S (Central Swyd) is a Offsite Power Source. A CT-4 is considered an Emergency Power Source. Answer D Discussion Incorrect. First is correct. Second part is plausible because a backcharged transformer could be used during an emergency. Basis for meeting the KA Question requires knowledge of the AC power system to complete a survellance procedure. OP/I 105/0 14 is a surveillance procedure. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided OP/i/All 105/014 SD 1.3.5 SYSO62 2.2.12 AC Electrical Distribution System SYSO62 GENERIC Knowledge of surveillance procedures. (CFR: 41 10 / 45.13) 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 95 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 48 SYSO63 2.2.22 DC Electrical Distribution System SYSO63 GENERIC Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2) Given the following plant conditions:

Uniti =Mode5
Unit2=Mode4
Unit3 = Mode 6 To satisfy TS 3.8.3 (DC Sources Operating) for Unit 2, (1) batteries are required and one of the batteries (2) required to be from Unit 1.

Which ONE of the following completes the statement above? REFERENCE PROVIDED A. 1. three

2. is B. 1. three

2. is NOT C. 1. four

2. is D. 1. four

2. is NOT Friday, October 12, 2012 Page 96 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE C 1LT42 ONS SRO NRC Examination QUESTION 48 48 General Discussion Answer A Discussion Incorrect. First part is plausible because it is correct if you dont apply LCO 3.8.3C. Second part is correct. Answer B Discussion Incorrect. First part is plasuible because it is correct if you dont apply LCO 3.8.3C. Second part is plasuible if Note 3 is not applied. Answer C Discussion Correct. A total of 4 batteries are required to satisfy LCO 3.8.3 and one of them has to be from Unit 1. Answer D Discussion Incorrect. First part is correct. Second part is plasuible if Note 3 is not applied. Basis for meeting the KA Question requires knowledge of TS 3.8.3, DC Sources Operating-Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EL-DCD R15 TS 3.8.3 TS 3.8.3 SYSO63 2.2.22 DC Electrical Distribution System SYSO63 GENERIC Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2) 1 4O19 Comments: Remarks/Status Friday, October 12, 2012 Page 97 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 49 SYSO63 K4.02 DC Electrical Distribution System Knowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Breaker interlocks, permissives, bypasses and cross-ties Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%

Current conditions:

Time 0400
1 PA Battery is inoperable At 0400, SLC 16.8.3 (Power Battery Parameters) requires:

1) Declaring the Unit 1 TD EFDW pump inoperable_ (1)

2) Cross connecting (2) buses.

Which ONE of the following completes the statements above? A. 1. immediately

2. 1PAand1PB B. 1. immediately

2. 1 PA, 2PA and 3PA C. 1. within one hour

2. 1PAandIPB D. 1. within one hour

2. 1PA,2PAand3PA Friday, October 12, 2012 Page 98 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 49 General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because the candidate could have the misconception that tying the units busses together would satisfy the SLC for some period of time. Answer B Discussion Correct. SLC 16.8.3 Condition B is for a single power battery inoperable and requires declaring the TDEFDW pump inoperable immediately. It also requires cross connecting all units PA buses. Answer C Discussion Incorrect. First part is plausible because one hour is a common TS completion time. Second part is plausible because the candidate could have the misconception that tying the units busses together would satisfy the SLC for some period of time. Answer 0 Discussion Incorrect. First part is plausible because one hour is a common TS completion time. Second part is correct. Basis for meeting the KA Requires knowledge of the normal design status of DC power battery busses as well as the design feature which allows cross-tie between units. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EL-DCD SLC 16.8.3 SYSO63 K4.02 DC Electrical Distribution System Knowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Breaker interlocks, permissives, bypasses and cross-ties 401-9 Comments: Remarks!Status Friday, October 12, 2012 Page 99 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 50 5O SYS064 Al .03 Emergency Diesel Generator (EDIG) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including: (CFR: 41.5 /45.5) Operating voltages, currents, and temperatures Given the following Unit 1 conditions: Time = 0400

Reactor power = 100%
Keowee Unit 1 output = 88 MWe
Keowee Unit 2 output 80 MWe Time = 0405
Switchyard Isolation occurs
CT-i locked out Time = 0500
Keowee Unit 1 has energized the MFB via CT-4
AP/1 1 (Recovery From Loss of Power) in progress
CT-4MEGAWATTS=15
CT-4MEGAVARS=16

1) At 0405, Keowee output voltage and frequency must be within a MINIMUM of (1) % of normal before the SK breakers can energize the Standby Bus.

2) At 0500 and in accordance with AP/li, Keowee Unit 1 output (2) required to be reduced to maintain CT-4 within limits.

Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1.5%

2. is B. 1.5%

2. is NOT C. 1. 10%

2. is D. 1. 10%

2. is NOT Friday, October 12, 2012 Page 100 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 50 General Discussion Answer A Discussion Incorrect, First part is plasuible because 5% is where the Statalarm for voltage and frequency actuates. Second part is plasuible because it is above the 100% limit line. Answer B Discussion Incorrect, First part is plasuible because 5% is where the Statalarm for voltage and frequency actuates. Second part is correct. Answer C Discussion Incorrect, First part is correct. Second part is plasuible because it is above the 100% limit line. Answer D Discussion Correct. The Out of Tolerance Circuit requires that Keowee output voltage and frequency must be within 10% of normal values before the associated breaker can close. Mifi I output reduction is not required because load up to 112% is allowed.. Basis for meeting the KA This question tests knowledge of the Out of Tolerance Circuit on the Keowee units. This circuit monitors voltage and frequency of the unit and prevents supplying power to ES equipment if frequency and voltage are not at appropriate values. Also requies the ability to appl;y the CT-4 curve to maintain with limits. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED ONS 2007 Q43 Development References Student References Provided , EL-KHG AP/l I End. 5.IA AP/ll End. 5.1A ONS 2007 Q43 SYSO64 A 1.03 Emergency Diesel Generator (ED/G) System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including: (CFR: 41.5 / 45.5) Operating voltages, currents, and temperatures 401-9 Comments: iarkslStatus Friday, October 12, 2012 Page 101 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 51 SYS 064 A3 .07 Emergency Diesel Generator (ED/G) System Ability to monitor automatic operation of the ED/G system, including: (CFR: 41.7 / 45.5) Load sequencing Given the following plant conditions: Initial conditions:

Unit 1 reactor power = 100%
Unit 3 Main Feeder Buses aligned to the Central Switchyard Current conditions:
Switchyard Isolation occurs
Unit 1 RCS pressure = 200 psig decreasing

1) lX5andlX6will (1).

2) 1 X7 will (2)

Which ONE of the following completes the statements above? A. 1. NOTload shed

2. NOT load shed B. 1. NOTload shed

2. load shed and reenergize in one minute C. 1. load shed and reenergize in thirty seconds

2. NOTload shed D. 1. load shed and reenergize in thirty seconds

2. load shed and reenergize in one minute Friday, October 12, 2012 Page 102 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 51 General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible all Load Centers do not always load shed. i.e. X5 and X6. Answer B Discussion Correct. Due to the Switchyard Isolation, both Oconee Uint I will trip and lose power. Due to the concurrent LOCA a load shed will be generated and power will come from the CT-4 and the underground. Although the SL breakers are closed, since Oconee Unit I also has an ES actuation 1X5 and 1X6 will not load shed. 1X7 will load shed and re-energized in 1 minute. Answer C Discussion Incorrect. Frist part is plasuible because it would be correct if Unit 1 did not have a LOCA. Second part is plausible all Load Centers do not always load shed. i.e. X5 and X6. Answer D Discussion Incorrect. Frist part is plasuible because it would be correct if Unit I did not have a LOCA. Second part is correct. Basis for meeting the KA Oconee does not have Diesel Generator or load sequencing. However Oconee does have Keowee Hydro Units that supply emergency power. Wej also have Load Shed which is used to protect the CT-4 and CT-5 transformers. This question requires knowledge of KHU operation and load shed during a switchyard isolation and a LOCA. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EL-PSL R6 SYSO64 A3.07 Emergency Diesel Generator (ED/G) System Ability to monitor automatic operation of the ED/G system, including: (CFR: 41.7 / 45.5) Load sequencing .401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 103 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 52 52 SYSO73 A4.03 Process Radiation Monitoring (PRM) System Ability to manually operate andlor monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Check source for operability demonstration Given the following Unit 1 conditions: Initial conditions:

Enclosure 4.9 (GWD Tank Release) of OP/1-21A11 104/018 (GWD System) in progress Current conditions:
1 RIA-37 source check is to be performed

1) The source check (1) performed by actuating 1 RIA-37 Source Check on the Enable Controls screen.

2) The source check is operable if (2)

Which ONE of the following completes the statements above? A. 1.is

2. the Process Monitor Fault Alarm is NOT received B. 1.is

2. 1 RIA-37 readings increase during the source check C. 1. is NOT

2. the Process Monitor Fault Alarm is NOT received D. 1. is NOT

2. 1 RIA-37 reading increase during the source check Friday, October 12, 2012 Page 104 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE A 1LT42 ONS SRO NRC Examination QUESTION 52 52 General Discussion Answer A Discussion Correct. The source check is performed by actuating 1RIA-37 source Check on the Enable Controls screen. The source check is operable if Process Monitor Fault is NOT received. Answer B Discussion Incorrect. First part is correct. Second part is plausible because it is a common misconception that the RIA readings will increase on a source check. Answer C Discussion Incorrect. First part is plausible because it is correct for 1RIA-38. Second part is correct. Answer D Discussion incorrect. First part is plausible because it is correct for 1RIA-38. Second part is plausible because it is a common misconception that the RIA readings will increase on a source check. Basis for meeting the KA Question requires knowledge of how a source check is performed and the expected RIA response. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memoty NEW Development References Student References Provided RD-RIA R5 PT/230/002 SYSO73 A4.03 Process Radiation Monitoring (PRM) System Ability to manually operate and/or monitor in the control room: (CFR: 41.7 I 45.5 to 45.8) Check source for operability demonstration 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 105 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 53 53 SYSO76 A2.02 Service Water System (SWS) Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 / 45/3 /45/13) Service water header pressure Given the following Unit 1 conditions: Initial conditions:

Time=1200
Reactor power 100%
A and B LPSW Pumps operating
CLPSWpumpinAUTO
Blackout Occurs Current conditions:
Time=1230
Both MFBs re-energized

1) The C LPSW pump (1) be operating 5 seconds after the MFBs are re-energized.

2) The (2) will require the use of OP/1/A/1104/010 (Low Pressure Service Water) to return them to service once LPSW pressure has been restored.

Which ONE of the following completes the statements above? A. 1. will

2. RBCU5 B. 1. will

2. Reactor Building Aux Coolers C. 1. will NOT

2. RBCUs D. 1. will NOT

2. Reactor Building Aux Coolers Friday, October 12, 2012 Page 106 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 53 53] General Discussion Answer A Discussion Incorrect: First part is plausible since it is reasonable to assume that the pump in standby vill 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. Answer B Discussion 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. Answer C Discussion 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. Answer D Discussion Correct. Following a LOOP when power has been restored to the Main Feeder Busses if pressure remains 70 psig for 10 seconds the standby LPSW pump will start. If only 5 seconds have passed since MFBs re-energize the standby pump will not be operating. Low LPSW pressure will isolate the RB Auxiliary Coolers and the RBCU Waterhammer Prevention Circuitry. Once the RB Aux Cooler isolation valves close, they must be manually re-opened. AP/24 directs the operator to OP/I IAJI 1 04/010 (Low Pressure Service Water) to restore the system once pressure is restored. Basis for meeting the KA ___________ 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. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK ONS 2009A Q52 Development References Student References Provided SSS-LPW R23 AP/24 SYSO76 A2.02 Service Water System (SWS) Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 / 45/3 / 45/13) Service water header pressure 4O19 Comments: RemarkslStatus Friday, October 12, 2012 Page 107 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 54 L SYSO78 K1.04 Instrument Air System (lAS) Knowledge of the physical connections and/or cause-effect relationships between the lAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Cooling water to compressor

1) The system that normally cools the primary instrument air compressor is (1).

2) If cooling is lost, the primary instrument air compressor will trip at a MINIMUM discharge temperature of (2) °F.

Which ONE of the following completes the statements above? A. 1. RCW

2. 240 B. 1. ROW

2. 425 C. 1. HPSW

2. 240 D. 1. HPSW

2. 425 Friday, October 12, 2012 Page 108 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 54 54] General Discussion Answer A Discussion Incorrect. First part is plausible because RCW does cool the backup IA compressors. Second part is correct. Answer B Discussion Incorrect. First part is plausible because RCW does cool the backup IA compressors. Second part is plausible because this would be correct for the backup IA compressors Answer C Discussion Correct. HPSW cools the Primary IA Compressor. The High discharge temperature trip for the Primary IA compressor is 240 degrees. Answer D Discussion Incorrect. First part is correct. Second part is plausible because this would be correct for the backup IA compressors. Basis for meeting the KA Question requires knowledge of cooling for the Primary IA compressor and effect of lossing the cooling. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided SSS-HPW R6 SSS-1AR33 SYSO78 Ki .04 Instrument Air System (lAS) Knowledge of the physical connections and/or cause-effect relationships between the lAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Cooling water to compressor 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 109 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 55 55 SYS 103 K4.04 Containment System Knowledge of containment system design feature(s) andlor interlock(s) which provide for the following: (CFR: 41.7) Personnel access hatch and emergency access hatch Given the following Unit 1 conditions: Initial conditions:

ReactorinMODE4 Current conditions:
Personnel Hatch Inner Door Red Light illuminated
Personnel Hatch Outer Door Red Light illuminated

1) In accordance with TS 3.6.2 (Containment Air Locks),

Which ONE of the following completes the statement above? A. no action is required at this time B. BOTH personnel hatch doors must be closed within one hour C. ONLY one personnel hatch door must be closed within one hour D. action must be initiated to evaluate overall containment leakage rate immediately Friday, October 12, 2012 Page 110 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 55 c 55 General Discussion Answer A Discussion Incorrect. Plausible because it is correct in MODE 5. Answer B Discussion Incorrect. First part is plausible because both doors are normally kept closed. Answer C Discussion Correct. Per TS 3.6.2 if one air lock door is inoperable then an operable door must be verified closed within 1 hour. Answer D Discussion Incorrect. First part is plausible because it is a required action if TS 3.6.2 Condition C. Basis for meeting the KA Question requires knowledge of a design feature of the RB personnel hatch. In the question the interlock that prevents both doors from being open at the same time has failed. The question asks what is the required actions from TS as a result of the failure. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided TS 3.6.2 SYS 103 K4.04 Containment System Knowledge of containment system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Personnel access hatch and emergency access hatch 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 111 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 56 56! SYSO 11 2.1.7 Pressurizer Level Control System (PZR LCS) SYSOI1 GENERIC Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 /45.12 / 45.13) Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%
SASS is in MANUAL
Pzr Level Channel 3 is selected Current conditions:
ICCM Train B loses incoming cabinet power
ICCM Train B plasma display is blank

1) Pzr Level Channel 3 will fail (1)

2) A valid Pzr level signal can only be obtained by selecting (2)

Which ONE of the following completes the statements above? A. 1. asis

2. ONLY Pzr level channel 1 B. 1. asis

2. EITHER Pzr level channel 1 OR channel 2 C. 1. low

2. ONLY Pzr level channel 1 D. 1. low

2. EITHER Pzr level channel 1 OR channel 2 Friday, October 12, 2012 Page 112 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 56 General Discussion Answer A Discussion Incorrect. First part is plausible because if the channel is taken out of service this would be correct. Second part is plausible because the candidate could have the misconception that channel B feeds channel 2 and 3. Also if SASS were in AUTO it will swap back to ONLY channel 1. Answer B Discussion Incorrect. First part is plausible because if the channel is taken out of service this would be correct. Second part is correct. Answer C Discussion Incorrect. First part is correct. Second part is plausible because the candidate could have the misconception that channel B feeds channel 2 and [3. Also if SASS were in AUTO it will swap back to ONLY channel 1. Answer D Discussion Correct. For an loss of power to the ICCM cabinet the channel, level indication will fail low. ICCM channel A feeds Pzr level channels 1 and

2. So channel 1 or 2 will indicate correctly. Either channel can be selected for a valid indication.

Basis for meeting the KA Question requires knowledge of how the Pzr level control system works and action to take when a component fails. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided PNS-Pzr R35 SYSO1 1 2.1.7 Pressurizer Level Control System (PZR LCS) SYSO11 GENERIC Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13) 401-9 Comments: arks/Status Friday, October 12, 2012 Page 113 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 57 SYSO14 K4.06 Rod Position Indication System (RPIS) Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.5 / 45.7) Individual and group misalignment Given the following Unit 1 conditions: Initial conditions:

Reactor startup in progress
Control rods in sequence mode
Withdrawing Group 5 control rods Current conditions:
Group 5 control rods = 94% withdrawn
Group 6 control rods 26% withdrawn

1) The control rod position indication used to determine if a sequence fault exits is (1)

2) Asequencefault (2) have occurred.

Which ONE of the following completes the statements above? A. 1.API

2. should B. 1. API

2. should NOT C. 1. RPI

2. should D. 1. RPI

2. should NOT Friday, October 12, 2012 Page 114 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 57 57 General Discussion Answer A Discussion Incorrect. First part is plausible because API is used for sequence enable. Second part is correct. Answer B Discussion Incorrect. First part is plausible because API is used for sequence enable. Second part is plausible if the candidate does not the current overlap is 32% which exceeds the 25% plus or minus 3% limit. Answer C Discussion Correct. RPI feeds the sequence fault. Group 5 and 6 rods ovelpaped by 32% which is outside of the 25% plus or minus 3% limit and a sequence fault should have occurred. Answer D Discussion Incorrect. First part is correct. Second part is plausible if the candidate does not the current overlap is 32% which exceeds the 25% plus or minus 3% limit. Basis for meeting the KA Question requires knowledge of design features and interlocks associated with control rod misalignment. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided IC-CRI R31 SYSO14 K4.06 Rod Position Indication System (RPIS) Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.5 / 45.7) Individual and group misalignment 4O19 Comments: RemarkslStatus Friday, October 12, 2012 Page 115 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 58 SYSO17 K6.O1 In-Core Temperature Monitor (ITM) System Knowledge of the effect of a loss or malfunction of the following ITM system components: (CFR: 41.7/45.7) Sensors and detectors Given the following Unit 1 conditions: Initial conditions:

Time= 0400
Reactor power = 100%
An ICCM Train B qualified CETC develops an open circuit in the detector Current conditions:
Time= 0430
C0reSCM=0°F
RB pressure = 8 psig

1) At 0400, this detector (1) be used in ICCM Train B Core SCM calculation.

2) At 0430, (2) CETCs will be used in ICCM Train A Core SCM calculation.

Which ONE of the following completes the statements above? A. 1.will

2. 5 B. 1. will

2. 12 C. 1. will NOT

2. 5 D. 1. will NOT

2. 12 Friday, October 12, 2012 Page 116 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 58 General Discussion Answer A Discussion Incorrect. First part is plausible if the candidate had the misconception that an open circuit caused the indication to fail high. (would be true for an RTD). Second part is correct. Answer B Discussion Incorrect. First part is plausible if the candidate had the misconception that an open circuit caused the indication to fail high. (would be true for an RTD). Second part is plausible because the ICCM uses 12 CETCs per train. Answer C Discussion Correct. This detector will fail low and not be used in the calculation. The ICCM Core SCM calculation used the 5 highest of the 12 CETCs for each train. Answer D Discussion Incorrect. First part is correct. Second part is plausible because the ICCM uses 12 CETCs per train. Basis for meeting the KA Question requires knowledge of the effect of a failure of a CETC would affect the SCM calculation. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided IC-RCI R38 BNT-CPO2P SYSO 17 K6.O 1 In-Core Temperature Monitor (ITM) System Knowledge of the effect of a loss or malfunction of the following ITM system components: (CFR: 41.7 /45.7) Sensors and detectors 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 117 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 59 59] SYSO 16 A2.O 1 Non-Nuclear Instrumentation System (NNIS) Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /45.3 /45.5) Detector failure Given the following Unit 1 conditions: Initial conditions:

Reactor power = 40%

Current conditions:

Final Feedwater temperature controlling signal fails HIGH

1) Initially, actual Feedwater Flow will (1)

2) AP/28 (lOS Instrument Failures) will ensure the Diamond Panel is in Manual and (2) are/is in Hand to stabilizethe plant.

Which ONE of the following completes the statements above? A. 1. decrease

2. 1A& 1BFDW Masters B. 1. decrease

2. Steam Generator Master C. 1. increase

2. 1A& lB FDW Masters D. 1. increase

2. Steam Generator Master Friday, October 12, 2012 Page 118 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 59 C General Discussion Answer A Discussion Incorrect: Plant response is incorrect. Plausible if the impact of the temperature compensation on the flow instrument is misapplied and assumption is that indicated flow would go down resulting in indicated flow being <FDW demand. Hand/Auto Stations identified in AP/28 are the FDW Masters only. Answer B Discussion Incorrect: Plant response is incorrect. Plausible if the impact of the temperature compensation on the flow instniment is misapplied and assumption is that indicated flow would go down resulting in indicated flow being <FDW demand. SG Master Hand/Auto Station is plausible if the failed signal is incorrectly assumed to be applied upstream of the SG Master instead of upstream of the FDW Loop Masters. Answer C Discussion Correct: FW Temp failing high will increase FDW Demand signal and raise actual FDW Flow. Hand/Auto Stations identified in AP/28 are the FDW Masters only. Answer D Discussion FW Temp failing high will increase FDW Demand signal and raise actual FDW Flow. SG Master Hand/Auto Station is plausible if the failed signal is incorrectly assumed to be applied upstream of the SG Master instead of upstream of the FDW Loop Masters. Basis for meeting the KA Requires knowledge of the effect of a failed FW Temp instrument on the ICS control of the FDW System and the appropriate FDW controls to operate to stabilize the plant Basis for Hi Cog Basis for SRO only Job Level RU L Cognitive Level Comprehension QuestionType BANK L Question Source ONS 2009 Q58 Development References Student References Provided AP/28 SAE-L087 0NS2009 Q58 SYSO 16 A2.O 1 Non-Nuclear Instrumentation System (NNIS) Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5) Detector failure 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 119 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 60 6 SYSO34 A4.02 Fuel Handling Equipment System (FHES) Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Neutron levels Given the following Unit 1 conditions:

ReactorinMODE6
Fuel assembly is being inserted into the core

1) In accordance with TS 3.9.2 (Nuclear Instrumentation), (1) is the MINIMUM number of Source Range neutron flux monitor(s) that are required to be in service at this time

2) Prior to disengaging the fuel assembly the Refueling SRO Assistant must determine the (2)

Which ONE of the following completes the statements above? A. 1.one

2. digital load reading is within the expected range B. 1. one

2. Source Range count rate is stable C. 1.two

2. digital load reading is within the expected range D. 1.two

2. Source Range count rate is stable Friday, October 12, 2012 Page 120 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 60 General Discussion Answer A Discussion Incorrect. First part is plausible because it would be correct if core geometry were not being changed. Second part is plausible because the Bridge Operator must insure this prior to disengaging the assembly. Answer B Discussion Incorrect. First part is plausible because it would be correct if core geometry were not being changed. Second part is correct. Answer C Discussion Incorrect. First part is correct. Second part is plausible because the Refueling SRO must insure this prior to disengaging the assembly. Answer 0 Discussion Correct. Two Source Ranges are required when core geometry is being changed. The Refueling SRO Assistant must ensure a stable count rate prior to disengaging a fuel assemble inserted into the core. Basis for meeting the KA knowledge of monitoring neutron level during fuel movement. Basis for Hi Cog Basis for SRO only Job Level Coiiive[evei QuestionType Question Source RO Memory NEW Development References Student References Provided TS 3.9.2 MP/0/AI1 5 00/009 FH-FHS R39 SYSO34 A4.02 Fuel Handling Equipment System (FHES) Ability to manually operate and!or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Neutron levels 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 121 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 61 SYSO41 K2.Ol - Steam Dump System (SDS)/Turbine Bypass Control Knowledge of bus power supplies to the following: (CFR: 41.7) ICS, normal and alternate power supply Given the following Unit I conditions:

Reactor power 100%

1) lOS AUTO powerto the TBVs is supplied by (1)

2) ICS HAND power to the TBVs is supplied by (2)

Which ONE of the following completes the statements above? A. 1. 1KI

2. 1KU B. 1. 1KI

2. 1KX C. 1. 1KU

2. 1KI D. 1. 1KU

2. 1KX Friday, October 12, 2012 Page 122 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 61 6l General Discussion Answer A Discussion Correct. ICS AUTO power is from 1KI. ICS HAND power is from IKU. Answer B Discussion Incorrect. First part is correct. Second part is plausible because IKX does supply power to the Main Turbine supervisor instrumentation. Answer C Discussion Incorrect. First part is plausible because IKU supplies ICS HAND power. Second part is plasuible because 1KI supplies AUTO power. Answer D Discussion Incorrect. First part is plausible because IKU supplies ICS HAND power. Second part is plausible because IKX does supply power to the Main Turbine supervisor instrumentation. Basis for meeting the KA Question requires knowledge of the power supplies to TBVs (ICS). Our supplies are AUTO and HAND power instead of normal and alternate. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EL-VPC R7 SYSO4 1 K2.O 1 Steam Dump System (SDS)/Turbine Bypass Control Knowledge of bus power supplies to the following: (CFR: 41.7) ICS, normal and alternate power supply 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 123 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 62 SYSO45 K3.Ol Main Turbine Generator (MT/G) System Knowledge of the effect that a loss or malfunction of the MT/G system will have on the following: (CFR: 41.7 / 45.6) Remainder of the plant Given the following Unit 1 conditions: Initial conditions:

Reactor power = 25% stable Current conditions:
Main Turbine trips The reactor (1) trip and the TBVs will control at (2)

Which ONE of the following completes the statement above? A. twill

2. setpoint B. 1. will

2. setpoint plus 125 psig C. 1. will NOT

2. setpoint D. 1. will NOT

2. setpoint plus 50 psig Friday, October 12, 2012 Page 124 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 62 62] General Discussion Answer A Discussion Incorrect.. First part is plausible because it would be correct if reactor power were above 29.75%. Second part is correct. Answer B Discussion jncorrect.. First part is plausible because it would be correct if reactor power were above 29.75%. Second part is plausible because it would be correct if the reactor had tripped. Answer C Discussion Correct. The reactor will not trip because power is below 27.75%. The TBVs will control at setpoint because the turbine has tripped but the reactor has not. Answer 0 Discussion Jncorrect. First part is correct. Second part is plausible because a 50 psig bias is applied if the turbine is on line. Basis for meeting the KA Question requires knowledge of parts of the plant if the MT trips. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided STG-ICS IC-RPS SYS045 K3 .01 Main Turbine Generator (MT/G) System Knowledge of the effect that a loss or malfunction of the MT/G system will have on the following: (CFR: 41.7 / 45.6) Remainder of the plant 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 125 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 63 63] SYSOO2 A3.Ol Reactor Coolant System (RCS) Ability to monitor automatic operation of the RCS, including: (CFR: 41.7 / 45.5) Reactor coolant leak detection system Given the following Unit 1 conditions:

Reactor power = 100%
Pzr level 215 inches increasing
RBNS level 10 inches increasing at 1 inch/mm
1RIA-40 = 16 cpm stable
1RIA-45 = 30 cpm stable
1RIA-47 = 1500 cpm increasing
1RIA-49 600 cpm increasing

1) The leak into the RB (1) from the RCS.

2) A high alarm on (2) will send a close signal to 1LWD-2 to isolate the RB normal sump.

Which ONE of the following completes the statements above? A. 1.is

2. 1RIA-45 B. 1.is

2. 1RIA-49 C. 1. is NOT

2. 1RIA-45 D. 1. is NOT

2. 1RIA-49 Friday, October 12, 2012 Page 126 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 63 63] General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because IRIA-45 will trip and isolate the RB purge on a high alarm. Answer B Discussion Correct. 1RIA-47 and 49 are increasing. This would indicate that the leak is in the RB. 1RIA-49 will close ILWD-2 on a high alarm. Answer C Discussion Incorrect. First part is plausible because Pzr level is increasing. Second part is plausible because lRIA45 will trip and isolate the RB purge on a high alarm. Answer D Discussion Incorrect. First part is plausible because Pzr level is increasing. Second part is correct. Basis for meeting the KA Question requires knowledge of one the TS required RCS leak detection systems (1RIA-47) and actions on RB RIA high alarm. Oconee does not have any auto actions of the RCS based on the leak detection system. Discussed with the Chief. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided RAD-RIA SYSOO2 A3.O1 Reactor Coolant System (RCS) Ability to monitor automatic operation of the RCS, including: (CFR: 41.7 / 45.5) Reactor coolant leak detection system 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 127 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 64 64 SYSO75 Kl.08 Circulating Water System Knowledge of the physical connections and/or cause-effect relationships between the circulating water system and the following systems: (CFR: 41.2 to 41.9/45.7 to 45.8) Emergency/essential SWS Given the following plant conditions: Initial conditions:

TimeO400
All Oconee units reactor power = 100%

Current conditions:

Time0415
A and C LPSW pumps are tripped and will not start
LPSW header pressure = 65 psig stable

1) At 0400, the A, B and C LPSW pumps are provided suction by (1)

2) At 0415, in accordance with AP124 (Loss of LPSW) Unit 3 LPSW pumps (2) be aligned to the Unit I and 2 LPSW system.

Which ONE of the following completes the statements above? A. 1. ONLY Unit 1 and Unit 2 CCW pumps

2. will B. 1. ONLY Unit 1 and Unit 2 CCW pumps

2. will NOT C. 1. ALL three Units CCW pumps

2. will D. 1. ALL three Units CCW pumps

2. will NOT Friday, October 12, 2012 Page 128 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 64 0 General Discussion Answer A Discussion Incorrect. First part is plausible because the Unit I &2 and unit 3s LPSW systems are separate on the discharge side. Second part is plausible because Unit I & 2 have only one LPSW pump operating and LPSW pressure is low. Answer B Discussion Incorrect. First part is plausible because the Unit 1&2 and unit 3s LPSW systems are separate on the discharge side. Second part is correct. Answer C Discussion Incorrect. First part is correct. Second part is plausible because Unit 1 & 2 have only one LPSW pump operating and LPSW pressure is low. Answer D Discussion Correct. All three units supply water to the CCW cross-over piping. The cross-over piping unit cross connect valves are normally open. This is where the LPSW pumps get their suction. AP/24 directs supplying Unit 1 & 2s LPSW with Unit 3 ONLY if NO Unit 1 &2 LPSW pumps are operating. So in this case the systems will NOT be cross connected. Basis for meeting the KA Question requires knowledge how the CCW system provides suction to the LPSW pumps. Basis for Hi Cog Basis for SRO only 1 Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided SSS-LPW R4 AP/24 SYSO75 Ki .08 - Circulating Water System Knowledge of the physical connections and/or cause-effect relationships between the circulating water system and the following systems: (CFR: 41.2 to 41.9/45.7 to 45.8) Emergency/essential SWS 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 129 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 65 65J SYSO86 Al.03 Fire Protection System (FPS) Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Fire Protection System operating the controls including: (CFR: 41.5 /45.5) Fire doors Given the following Unit 1 conditions:

Reactor power = 100%
A fire door is blocked open and cannot be closed
All fire detectors are determined to be operable Which ONE of the following describes a required action and the MAXIMUM completion time in accordance with SLC 16.9.5 (Fire Barriers)?

A. Establish hourly fire watch patrol on at least one side of the boundary immediately. B. Establish hourly fire watch patrol on at least one side of the boundary within 1 hour. C. Establish a continuous fire watch on at least one side of the boundary immediately. D. Establish a continuous fire watch on at least one side boundary within 1 hour. Friday, October 12, 2012 Page 130 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE B 1LT42 ONS SRO NRC Examination QUESTION 65 65j General Discussion Answer A Discussion a Incorrect. Plausible because an hourly fire watch is required and immediately is a common TS completion time. Answer B Discussion Correct. An hourly fire watch is required is required to be established in one hour. Answer C Discussion Incorrect. Plausible because a continuous fire watch would be required if no fire detectors were operable on either side of the Fire Barrier and immediately is a common TS completion time. Answer D Discussion Incorrect. Plausible because a continuous fire watch would be required if no fire detectors were operable on either side of the Fire Barrier and one hour is correct. Basis for meeting the KA Question requires knowledge of fire doors and action to take if one is inoperable. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided SLC 16.9.5 (Fire Barriers) SYSO86 Al .03 Fire Protection System (FPS) Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Fire Protection System operating the controls including: (CFR: 41.5/45.5) Fire doors 401-9 Comments: arksIStatus Friday, October 12, 2012 Page 131 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 66 6 GEN2.1 2.1.3 GENERIC Conduct of Operations Conduct of Operations Knowledge of shift or short-term relief turnover practices. (CFR; 41.10 / 45.13) In accordance with OMP 2-16 (Shift Turnover) which ONE of the following is required to be reviewed/completed prior to shift turnover? A. Computer Temporary Alarm Summary B. Control Room Tagout Status Report C. Computer Point Processing Log D. RIA Alarm Summary Friday, October 12, 2012 Page 132 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE 0 1LT42 ONS SRO NRC Examination QUESTION 66 General Discussion Answer A Discussion Incorrect. Plausible because it is has to be completed after shift turnover. Answer B Discussion Incorrect. Plausible because it is has to be completed after shift turnover. Answer C Discussion Incorrect. Plausible because it is has to be completed after shift turnover. Answer D Discussion Correct. OMP 2-16 requires reviewing the RIA Alarm Summary prior to shift turnover. Basis for meeting the KA Question requires knowledge of shift turn over practices. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided OMP 2-16 (Shift Turnover) GEN2.l 2.1.3 GENERIC Conduct of Operations Conduct of Operations Knowledge of shift or short-term relief turnover practices. (CFR: 41.10 / 45.13) 4O19 Comments: Remarks/Status Friday, October 12, 2012 Page 133 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 67 67 GEN2.1 2.1.40 GENERIC Conduct of Operations Conduct of Operations Knowledge of refueling administrative requirements. (CFR: 41.10 / 43.5 / 45.13) Given the following Unit I condiUons:

ReactorinMODE6
RB Purge in progress
Defueling in progress
SF Cooling aligned in refueling mode In accordance with MP/O/A/1500/009 (Defueling/Refueling Procedure), which ONE of the following would require immediate suspension of fuel handling?

A. B SFP Cooling pump trips B. I RIA-49 removed from service C. Spent Fuel Pool level = (-)1 .7 feet D. Emergency air lock doors are open with a temporary cover in place Friday, October 12, 2012 Page 134 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 67 6 General Discussion Answer A Discussion Incorrect. While the B SFP will be operating in the refueling alignment, its loss does not require stopping fuel movement. Answer B Discussion Incorrect. Plausible because IRIA-49 does monitor RB atmosphere and if it were IRIA-45 it would be correct. IRIA-45 is required to termi the RB purge. Answer C Discussion Correct. Fuel handling must be stopped I SFP/Transfer canal level is less than 1 foot. Answer D Discussion Incorrect. Plausible because it would be correct if a cover were not in place. Basis for meeting the KA Question requires knowledge of administrative requirements for stopping fuel movement. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided FH-FHS TS 3.9.3 MPh 5 00/009 GEN2.1 2.1.40 GENERIC Conduct of Operations Conduct of Operations Knowledge of refueling administrative requirements. (CFR: 41.10 / 43.5 / 45.13) 401-9 Comments; RemarkslStatus Friday, October 12, 2012 Page 135 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 68 GEN2.2 2.2.1 GENERIC Equipment Control Equipment Control Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity. (CFR: 41.5 / 41.10 / 43.5 /43.6 / 45.1) Given the following Unit I conditions: Initial conditions: Reactor in MODE 5 Current conditions Startup in progress In accordance with SOMP 01-02 (Reactivity Management):

1) (1) would require stopping control rod withdrawal.

2) The dedicated SRO (2) allowed to peer-check the withdrawal of control rods.

Which ONE of the following completes the statements above? A. 1. Actuation of ANY unexpected annunciator

2. is B. 1. Actuation of ANY unexpected annunciator

2. is NOT C. 1. Actuation of ONLY reactivity related annunciators

2. is D. 1. Actuation of ONLY reactivity related annunciators

2. is NOT Friday, October 12, 2012 Page 136 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 68 6 General Discussion I_ Answer A Discussion Correct. Per SOMP 1-02 stopping withdrawal of CRs is required upon reciept of any uinexpected annunciater and the dedicated SRO is allowed to perform peer-checks related to the reactor startup. Answer B Discussion Incorrect. First part is correct. Second is plasuible because non-dedicated SROs are not allowed to provide peer checks. Answer C Discussion Incorrect. Plausible because the candidate may have the misconception that stopping CR withdrawal is only required for alarms related to reactivity. Second part is coorect. Answer D Discussion Incorrect. Plausible because the candidate may have the misconception that stopping CR withdrawal is only required for alarms related to reactivity. Second is plasuible because non-dedicated SROs are not allowed to provide peer checks. Basis for meeting the KA Question requires knowledge of procedural requirements while withdrawing control rods including when it is required to be stopped and can the ,SRO peer check the withdrawal. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided SOMP 01-02 GEN2.2 2.2.1 GENERIC Equipment Control Equipment Control Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity. (CFR: 41.5 / 41.10 / 43.5 /43.6 / 45.1) 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 137 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 69 69j GEN2.2 2.2.43 GENERIC Equipment Control Equipment Control Knowledge of the process used to track inoperable alarms. (CFR: 41.10 /43.5 / 45.13) Given the following Unit 1 conditions: Initial conditions:

Reactor power = 100%
iSA-i 5/A-2 (SU Source Volt Monitor Logic Test) actuated Current conditions:
l&E determines the alarm actuated due to a defective alarm relay
Repairs will take 34 days
The CRS directs you to remove the nuisance alarm from service Which ONE of the following list required actions in accordance with OMP 1-02 (Rules of Practice) to remove the alarm from service?

A. 1. Add the Statalarm to the Out of Normal Alarms section of the Unit Turnover Sheet.

2. Place an OOSII&E label on the alarm window.

B. 1. Add the Statalarm to the Out of Normal Alarms section of the Unit Turnover Sheet.

2. Place a Tb Sheet or CBWO label on the alarm window.

C. 1. Add the Statalarm to the Equipment Deficiencies section of the Unit Turnover Sheet.

2. Place an OOSII&E label on the alarm window.

D. 1. Add the Statalarm to the Equipment Deficiencies section of the Unit Turnover Sheet.

2. Place a TIO Sheet or CBWO label on the alarm window.

Friday, October 12, 2012 Page 138 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 69 69] General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible because this label would be put on a piece of equipment in the control room that was removed from service. Answer B Discussion Correct. Per OMP 1-02, the Satatalarm will be added to the Out of Normal section of the Unit Turnover Sheet and a Tb Sheet or CBWO label will be plased on the alarm window. Answer C Discussion Incorrect. First part is plasuible because it is correct for equipment in the control room that was removed from service. Second part is plausible because this label would be put on a piece of equipment in the control room that was removed from service. Answer D Discussion Incorrect. First part is plasuible because it is correct for equipment in the control room that was removed from service. Second part is correct. Basis for meeting the KA Question requires knowlwdge of the process for tracking inoperable Statalarms. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided OMP 1-02 GEN2.2 2.2.43 GENERIC Equipment Control Equipment Control Knowledge of the process used to track inoperable alarms. (CFR: 41.10/43.5/ 45.13) 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 139 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 70 o] GEN2.3 2.3.14 GENERIC Radiation Control Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41 .12 / 43.4 /45.10) Given the following Unit 1 conditions:

Reactor power = 86% decreasing
36 gpm primary to secondary leak in 1 A SG
The SRO has directed the NEC to perform a task in the turbine building basement

1) The location in the TB basement that will have the greatest increase in general area dose rates is the (1)

2) If the operator receives a dose alarm while performing this task he will (2)

Which ONE of the following completes the statements above? A. 1. powdex

2. complete the task while monitoring his dose B. 1. powdex

2. immediately stop and leave the area C. 1. TBSump

2. complete the task while monitoring his dose D. 1. TBSump

2. immediately stop and leave the area Friday, October 12, 2012 Page 140 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE A 1LT42 ONS SRO NRC Examination QUESTION 70 70 General Discussion Answer A Discussion Correct. The Powdex will gather and concentrate the RCS activity making it have the highest general dose rates; since emergency dose limits (EDL5) are in affect the NEO is expected complete the task while monitoring his dose. Answer B Discussion Incorrect, First part is correct. Second part is plasuible because if EDLs were not in affect this would be correct. Answer C Discussion Incorrect: First part is plasuible because the TB Sump will be an area for secondary contamination to collect but it will not have dose rates as high as the Powdex. Second part is correct. Answer D Discussion Incorrect: First part is plasuible because the TB Sump will be an area for secondary contamination to collect but it will not have dose rates as high as the Powdex. Second part is plasuible because if EDLs were not in affect this would be correct. Basis for meeting the KA Question requires knowledge of radation hazards due to a SGU and radiological safety practices Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RU Comprehension MODIFIED ONS 2009 RU Q72 Development References Student References Provided EAP-TCA R6 ONS 2009 RU Q72 GEN2.3 2.3.14 GENERIC Radiation Control Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12 / 43.4 /45.10) 401-9 Comments: Remarks/Status Friday, October 12, 2012 Page 141 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 71 7j GEN2.3 2.3.15 - GENERIC Radiation Control Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 / 43.4/ 45.9) Given the following Unit 1 conditions: Initial conditions:

Time=0400
Mode5
RB Purge in operation Current conditions:
Time=0415
RB Purge in operation
Radiation levels in the RB increasing
The switchover acceptance range setpoint is reached At 0415, 1RIA-45 will read _(1)_ and 1RIA-46 (2) trip the Main Purge fan.

Which ONE of the following completes the statement above? A. 1. off scale high

2. will NOT B. 1. offscalehigh

2. will C. 1. zero

2. will NOT D. 1. zero

2. will Friday, October 12, 2012 Page 142 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 71 0 General Discussion Answer A Discussion Incorrect, First part is plausible because counts are increasing to the top of its range. Second part is plausible because the candidate could have the misconception that 1RIA-46 trip setpoint is abovet the switchover acceptance range setpoint. Answer B Discussion Incorrect, First part is plausible because counts are increasing to the top of its range. Second part is correct. Answer C Discussion Incorrect, First part is correct. Second part is plausible because the candidate could have the misconception that 1RIA-46 trip setpointisabovet the switchover acceptance range setpoint. Answer D Discussion -- Correct, I RIA-45 will read zero and I RIA-46 will provide the same interlock functions as 1 RIA-45 (which would include tripping Purge fan1 and closing Purge valves). Basis for meeting the KA Question requires knowledge of fixed radiation monitors. Basis for Hi Cog L________________ Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK ONS 2007 Retest Q52 Development References Student References Provided RAD-RIA R2 ONS 2007 Retest Q52 GEN2.3 2.3.15 GENERIC Radiation Control Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 / 43.4 I 45.9) 401-9 Comments: 1 Remarks/Status Friday, October 12, 2012 Page 143 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 72 72j GEN2.3 2.3.7 GENERIC Radiation Control Radiation Control Ability to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12 / 45 10) Given the following conditions:

An Operator is preparing to enter the RCA to perform a valve lineup in a high radiation area
The Radiation Work Permit (RWP) specifies that the worker must not exceed their annual ALERT exposure limit The individual workers exposure must remain less than a MAXIMUM of (1) of the annual (2)

Which ONE of the following completes the statement above? A. 1. 80%

2. NRC exposure limit B. 1. 80%

2. Duke administrative exposure limit C. 1. 90%

2. NRC exposure limit D. 1. 90%

2. Duke administrative exposure limit Friday, October 12, 2012 Page 144 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE B 1LT42 ONS SRO NRC Examination QUESTION 72 General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plasuible because a canddidate could have the misconception that the alert is a percentage of the NRC limit. Answer B Discussion Correct. alert is 80% of the annual administrative limits. Answer C Discussion First part is plausible because 90% is the excluded limit. Second part is plasuible because a canddidate could have the misconception that the alert is a percentage of the NRC limit. Answer D Discussion First part is plausible because 90% is the excluded limit. Second part is correct. Basis for meeting the KA This question requires knowledge of how to comply with an RWP. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK MNS Bank 4424 Development References Student References Provided RAD-RPP R12 MNS Bank 4424 GEN2.3 2.3.7 GENERIC Radiation Control Radiation Control Ability to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12 / 45.10) 401-9 Comments; RemarkslStatus Friday, October 12, 2012 Page 145 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 73 GEN2.4 2.4.2 GENERIC Emergency Procedures / Plan Emergency Procedures I Plan Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 / 45.7 / 45.8) Given the following Unit 1 conditions:

Reactor power = 100%

Which ONE of the following would require entry into the EOP? A. Reactor power increases to 102% B. Group 2 Rod 6 Control Rod fully inserts into the core C. Reactor Coolant System leakage in the RB of 55 gpm D. Reactor Coolant System pressure increases to 2360 psig Friday, October 12, 2012 Page 146 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 73 0 General Discussion Answer A Discussion Incorrect. Plausible because it would be correct if power inceased above 105.5%. Also it is above the ICS limit of 101%. Answer B Discussion Incorrect. Plausible because it would be correct if 2 control rods dropped into the core. Answer C Discussion Incorrect. Plausible because it would be correct if RCS leakage were greeater then Normal Makeup Capability. Also at this leak rate a reactor shut down is required but not a reactor trip. Answer D Discussion Correct. A reactor trip will occur when RCS pressure reaches 2355. If the RPS does not trip the reactor the operator is required to per OMP 1-18 Attachment A (Licensed Operator Memory Items). This would require entering the FOP. Basis for meeting the KA Question requires knowledge of events which would result into entry into the FOP. Basis for Hi Cog Basis for SRO only Job Leveif Cognitive Level QuestionType - Question Source RO Memory NEW Development References Student References Provided OMP 1-18 Attachment A GEN2.4 2.4.2 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 / 45.7 / 45.8) 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 147 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 74 7 GEN2.4 2.4.4 1 - GENERIC Emergency Procedures I Plan Emergency Procedures / Plan Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10/43.5 / 45.11) Given the following Unit 1 conditions: Time = 0400

Unusual Event is declared Time 0430
An Alert is declared

1) The lowest emergency action level which requires the Emergency Operating Facility to be activated is a(n) (1)

2) The latest time that the Alert declaration must be communicated to the State and Counties is (2)

Which ONE of the following completes the statements above? A. 1. Notification of Unusual Event

2. 0445 B. 1. Notification of Unusual Event

2. 0530 C. 1. Alert

2. 0445 D. 1. Alert

2. 0530 Friday, October 12, 2012 Page 148 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 74 C General Discussion Answer A Discussion Incorrect. First part is plasuibe because the Emergency Plan is implemented during an Unusual Event. Second part is correct. Answer B Discussion Incorrect. First part is plausible because the Emergency Plan is implemented during an Unusual Event. Second part is plausible because follow ups notifications are required every hour. Answer C Discussion Correct. The lowest emergency action level that requires the EOF to be activated is an Alert and the classification upgrade must be communicated to the state and counties within 15 minutes. Answer D Discussion Incorrect. First part is correct. Second part is plausible because follow ups notifications are required every hour. Basis for meeting the KA Question requires knowledge of actions that are taken at different emergency action levels. Basis for Hi Cog Basis for SRO only Job LeveTf Cognitive Level QuestionTypeJ Question Source RO Memory NEW Development References Student References Provided EAP-SEP R3 RP/l 000/002 GEN2.4 2.4.4 1 GENERIC Emergency Procedures / Plan Emergency Procedures I Plan Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5 / 45.1 1) 1 4O19 Comments: marksIStatus Friday, October 12, 2012 Page 149 of 212

FOR REVIEW ONLY DO NOT DISTRIBUTE - 1LT42 ONS SRO NRC Examination QUESTION 75 GEN2.4 2.4.42 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of emergency response facilities. (CFR: 41.10 / 45.11)

1) The emergency facility that assumes responsibility for communications with offsite agencies including the NRC once it is activated is the (1)

2) The lowest classification level that requires this facilitys activation is an (2) .

Which ONE of the following completes the statements above? A. 1. Technical Support Center (TSC)

2. Alert B. 1. Technical Support Center (TSC)

2. Site Area Emergency C. 1. Operations Support Center (OSC)

2. Alert D. 1. Operations Support Center (OSC)

2. Site Area Emergency Friday, October 12, 2012 Page 150 of 212

FOR REVIEW ONLY - DO NOT DISTRIBUTE 1LT42 ONS SRO NRC Examination QUESTION 75 General Discussion Answer A Discussion Correct. The TSC assumes communications with the NRC after it is activated and is required to be activated at an Alert. Answer B Discussion rrect. First part is correct. Second part is plausible if the candidate had the misconception that an Alert it was optional. Answer C Discussion Incorrect. First part is plausible because the OSC does perform required actions during an event but not communications with outside agencies. Second part is correct. Answer D Discussion Incorrect. First part is plausible because the OSC does perform required actions during an event but not communications with outside agencies. Second part is plausible if the candidate had the misconception that an Alert it was optional. Basis for meeting the KA Question requires knowledge of emergency response facilities. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK CNS 2008 NRC Exam Development References Student References Provided EAP-SEP RPI1000/002 CNS 2008 NRC Exam GEN2.4 2.4.42 GENERIC Emergency Procedures I Plan Emergency Procedures / Plan Knowledge of emergency response facilities. (CFR: 41.10 / 45.11) 401-9 Comments: RemarkslStatus Friday, October 12, 2012 Page 151 of 212}}

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