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Initial Exam 2011-302 Draft RO Written Exam
	ML113390251
Person / Time
Site:	Oconee
Issue date:	12/01/2011
From:	; NRC/RGN-II
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ES-401 - Site-Specific RO Written Examination Form ES-401-7 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Applicant Information Name:

Date: 10126/2011 Facility/Unit: Oconee Region: i El II Ill El IV El Reactor Type: W El CE El BW GE El Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination, you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the examination begins.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results Examination Value 75 Points Applicants Score Points Applicants Grade Percent L-

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION EPEOO7 EA2.06 Reactor Trip Ability to determine or interpret the following as they apply to a reactor trip: (CFR 41.7 / 45.5/45.6) 1 C Occurrence of a reactor trip Given the following Unit 1 conditions:

Reactor power = 75%

. Main Feedwater transient is in progress Which ONE of the following combinations of statalarms from iSA-I could indicate an AUTOMATIC reactor trip has occurred due to LOW RCS PRESSURE?

A.

CRD CR0 CRD CRD CRD CRD TRIP BKR A TRIP BKR B TRIP BKR C TRIP BKR 0 ELECTRONIC ELECTRONIC TRIP TRIP TRIP TRIP TRIPE TRIP F B.

CR0 CR0 CRD CRD CR0 CRD TRIP BKR A TRIP BKR B TRIP BKR C TRIP BKR D ELECTRONIC ELECTRONIC TRIP TRIP TRIP TRIP TRIP E TRIP F C.

CRD CR0 CRD CR0 CRD CRD TRIP BKR A TRIP BKR B TRIP BKR C TRIP BKR D ELECTRONIC ELECTRONIC TRIP TRIP TRIP TRIP TRIP E TRIP F D.

CR0 CR0 CRD CRD CRD CR0 TRIP BKR A TRIP BKR B TRIP BKR C TRIP BKR 0 ELECTRONIC ELECTRONIC TRIP TRIP TRIP TRIP TRIPE TRIP F m

Wednesday, August 31, 2011 Page 1 of 208

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2011B ONS SRO NRC Examination QUESTION 1 General Discussion Answer A Discussion Incorrect. Plausible since only two CRD breakers opening will result in a Reactor trip however it is specific combinations of breakers and A & C are not two of the pair that result in de-energizing CRDs.

Answer B Discussion Incorrect. Plausible since only two CRD breakers opening will result in a Reactor trip however it is specific combinations of breakers and B & D are not two of the pair that result in de-energizing CRDs.

Answer C Discussion Correct. A low RCS pressure RPS trip would attempt to open all CRD breakers. It does not take all CRD breakers opening to de-energize the CRDs. There are several pairs of CRD breakers where only two breakers are required to de-energize the CRDs. A & B are one of the pairs.

Answer D Discussion Incorrect. Plausible since the E & F contactors would result in all CRDs being inserted and would not open any of the CRD breakers however the E & F contacts are actuated from AMSACIDSS due to high RCS pressure.

Basis for meeting the KA KA requires the ability to determine the occurrence of a Rx trip. Interpreting statalarms to determine which combinations would indicate a Rx trip has occurred meets the KA.

Basis for Hi Cog Requires analyzing various statalarm combinations to determine if the correct breaker combination for a Rx trip has occurred. Also requires specifically differentiating between indications of a high pressure trip resulting in DSS vs low pressure trip.

Basis for SRO only Job Level Cognitive Level QuestionTypo Question Source RO Comprehension NEW Development References Student References Provided IC-CRI R20, IC-RPS R17 IC-CRI IC-Digital RPS EPEOO7 EA2.06 Reactor Trip Ability to determine or interpret the following as they apply to a reactor trip: (CFR 41.7 / 45.5 / 45.6)

Occurrence of a reactor trip 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 2 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 2 APEOO8 AK2.02 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)

Sensors and detectors Given the following Unit I conditions:

Reactor power = 100%

I RC-66 (PORV) is leaking past its seat

Pressurizer temperature = 648° F

Quench tank pressure = 5 psig Based on the above conditions, which ONE of the following describes the expected tailpipe temperature (°F) downstream of 1 RC-66?

A. 648 B. 272 C. 228 D. 162 Wednesday, August 31, 2011 Page 3 of 208

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2011B ONS SRO NRC Examination QUESTION 2 General Discussion Re-arranged answers to make different letter the correct answer.

Answer A Discussion Incorrect: Plausible with the same misconception made at TMI which was assuming constant temperature across the valve due to throttling process Answer B Discussion Incorrect: Plausible if one thinks that the throttling process is a constant entropy process and looks for the same entropy as at 648 degrees F -

1.27 BTU/R/lb Answer C Discussion CORRECT: The enthalpy for the steam leaving the pressurizer at 648 degreesF will be the same at 5 psig (2opsia) 1124 BTU/Ib. This enthaIpy at 20 psia constitutes a wet vapor with a temperature of 228 degrees F. Throttling processes are constant enthalpy processes and energy remains approximately the same on both sides of a throttling process.

Answer D Discussion Incorrect: Plausible because this will be the answer if 5 psig is not converted to psia.

Basis for meeting the KA Requires knowledge of Pressurizer vapor space accident (leaking PORV) on tailpipe temp by applying thermodynamic flow characteristics of a leaking valve Basis for Hi Cog Requires applying thermodynamic knowledge and demonstrating the ability to use the Mollier diagram.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RU Comprehension BANK 2009A NRC exam Question #1 Development References Student References Provided PNS-PZR R34 APEOO8 AK2.02 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)

Sensors and detectors 401-9 Comments RemarkslStatus Wednesday, August 31, 2011 Page 4 of 208

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2011B ONS SRO NRC Examination QUESTION 3 APEOI5/017 AK2.08 Reactor Coolant Pump (RCP) Malfunctions Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: (CFR 41.7 /45.7)

CCws Given the following Unit 2 conditions:

Reactor power = 100%

2HP-31 (RCP Seal Flow Control) failed CLOSED

2CC-8 failed CLOSED

ALL individual RCP seal return valves CLOSED Which ONE of the following describes the response of 2HP-21 (if any) AND the required Operator actions that will be performed?

A.

2HP-21 will remain OPEN

RCPs will be secured ONLY if Immediate Trip Criteria of AP/16 (Abnormal RCP Operation) are exceeded.

B.

2HP-21 will close automatically

RCPs will be secured ONLY if Immediate Trip Criteria of AP/16 (Abnormal RCP Operation) are exceeded.

C.

2HP-21willreman OPEN

AP/14 (Loss of HPI Normal Makeup and/or Seal Injection) will direct tripping the reactor then securing ALL RCPs immediately.

D.

2HP-21 will close automatically

AP/14 (Loss of HPI Normal Makeup and/or Seal Injection) will direct tripping the reactor then securing ALL RCPs immediately..

Wednesday, August 31, 2011 Page 5 of 208

FOR REVIEW ONLY 10 NOTDIS.TRIBUTE -

2011B ONS SRO NRC Examination QUESTION 3 3 General Discussion Answer A Discussion Incorrect. This distracter is plausible since HP-21 does not automatically close on Unit 1. Additionally, leaving the RCPs running unless Immediate Trip Criteria of AP/16 is exceeded would be correct in the case where either CC or Seal Injection were lost as long as both CC and Seal Injection were not lost simultaneously.

Answer B Discussion Incorrect. Plausible since leaving the RCPs running unless Immediate Trip Criteria of AP/16 is exceeded would be correct in the case where either CC or Seal Injection were lost as long as both CC and Seal Injection were not lost simultaneously.

Answer C Discussion Incorrect. This distracter is plausible since HP-2 1 does not automatically close on Unit 1 Answer D Discussion Correct. On Unit 2, the individual seal return valve will close when that pumps seal injection is <4 gpm and CC flow is < 575 gpm. In this case, 2HP-3 1 is closed which means all pumps seal injection are <4 gpm and CC-8 going closed will trip all running CC pumps therefore all individual seal return valves have closed. When all Unit 2 seal return valves close, 2HP-21 will close. One of the IMAs of AP/14 (Loss of HPI Normal Makeup and/or Seal Injection) is to trip the Rx and stop all RCPs if BOTH RCP seal injection and CC are lost.

Basis for meeting the KA This question requires knowledge of how a loss of component cooling can result in a loss of RC Flow based on Malfunctions that impact RCP operation.

Basis for Hi Cog Requires analyzing the impact of a valve failure and applying the analysis to actions directed by APs.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-APG R9 2AP/14 1AP/14 APEO15/017 AK2.08 Reactor Coolant Pump (RCP) Malfunctions Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: (CFR 41.7 / 45.7)

CCWS 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 6 of 208

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201 lB ONS SRO NRC Examination QUESTION 4 4 APEO22 AA2.O1 Loss of Reactor Coolant Makeup Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: (CFR 43.5/ 45.13)

Whether charging line leak exists Given the following Unit I conditions:

Initial conditions:

Reactor power = 100%

Current conditions:

Pressurizer level = 195 decreasing

LDST level = 78 decreasing Which ONE of the following has occurred?

A. Line break downstream of 1 HP-7 B. Line break downstream of IHP-120 C. 1 HP-14 has failed in the bleed position D. Loss of Instrument Air and Auxiliary Instrument Air to IHP-5 Wednesday, August 31, 2011 Page 7 of 208

FOR REVIEW ONLY DO NOT DISTRIBUTE B

2011B ONS SRO NRC Examination QUESTION 4 General Discussion Answer A Discussion Incorrect. This answer is plausible since a line break downstream of HP-7 would give many of the same indication as a line break downstream of HP-120 (Sump rates. RIA alarms, etc.). This break would result in a loss of makeup to the LDST and would therefore cause a decrease in LDST level. This answer is incorrect since 1HP-120 would still control Pzr level therefore it would remain unchanged.

Answer B Discussion Correct. With a line break downstream of HP-120, makeup to the pressurizer would be lost and pressurizer level would decrease. HP-120 would see the decrease in Pzr level and try to provide additional makeup by opening further. As HP-120 opened further, LDST level would begin to decrease since the amount of water entering the LDST would be unaffected by the failure.

Answer C Discussion Incorrect. Plausible since HP-14 failing in bleed would cause an overall loss of inventory and LDST level would be decreasing however 1HP 120 would still maintain Pzr level constant.

Answer D Discussion Incorrect. Plausible since the candidate must determine that a loss of air supply to HP-5 will result in the valve failing closed and then determine that HP-5 failing closed would result in a decrease in the water being added to the LDST and would therefore result in a decrease in LDST level.

1-IP-120 would still control Pzr level therefore Pzr level would remain unchanged. Changing one bullet in stem (Pzr level) could make this a correct answer.

Basis for meeting the KA Requires diagnosing a line break downstream of the normal makeup valve.

Basis for Hi Cog HI cog since the candidate must analyze various component failures to determine the impact on both LDST and Pzr levels.

Basis for SRO only Job Level Cognitive Level QuestionTyp Question Source RU Comprehension NEW Development References Student References Provided PNS-HPI R2,426 APEO22 AA2.01 Loss of Reactor Coolant Makeup Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: (CFR 43.5/ 45.13)

Whether charging line leak exists 401-9 Comments: RemarksiStatus Wednesday, August 31, 2011 Page 8 of 208

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2011B ONS SRO NRC Examination QUESTION 5 5j APEO25 AK2.03 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)

Service water or closed cooling water pumps Given the following Unit 3 conditions:

Reactor in MODE 5

RCS heatup in progress

LPI cooler outlet temperature = 162°F increasing

3Aand3B LPSW Pump operating

30 LPI Pump operating Which ONE of the following will result in a complete Loss of Decay Heat Removal capability?

ASSUME NO OPERATOR ACTIONS A. 3LP-1 I (3A LPI Cooler Inlet) fails closed B. 3LP-13 (3B LPI Cooler Inlet) fails closed C. 3A LPSW Pump trips concurrent with a 3T0 switchgear Lockout D. 3A LPSW Pump trips concurrent with a 3TD switchgear Lockout Wednesday, August 31, 2011 Page 9 of 208

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2011B ONS SRO NRC Examination QUESTION 5 5 General Discussion Answer A Discussion Incorrect. Plausible since this would be a correct answer if asked regarding Unit 1 or Unit 2 while they are in the High Pressure Mode of LPI alignment. Additionally plausible since the High Pressure Mode is one of the LPI Modes that are available to be used during periods where RCS pressure is at the upper end of the range of pressures where LPI is available to be used for Decay Heat Removal.

Answer B Discussion Incorrect. Plausible since this would be a correct answer if asked regarding Unit 1 or Unit 2 while they are in the Switchover Mode of LPI alignment. Additionally plausible since the Switchover Mode is one of the LPI Modes that are available to be used during periods where RCS pressure is at the upper end of the range of pressures where LPI is available to be used for Decay Heat Removal.

Answer C Discussion incorrect. Plausible since the LPSW pumps are 4160V pumps. While the LPSW pumps follow the standard TC feeds the A pump and TD feeds the B pump standard, not all safety related pumps follow that standard. Two specific examples are the Motor Driven EFDW pumps where the A MDEFWP is fed from TD and the B MDEFWP is fed from TE. The Component Cooling pumps are also fed from the TD and TB power strings.

Other components that do not follow the standard are RBCUs & HPSW Pumps. Since all pumps do not follow the TC-TD-TE standard it is plausible to believe that the 3B LPSWP is fed from 3TC (like the B HPSW Pump is fed from MFBI). Additionally, since it is plausible that the LPSW pumps do not follow the standard power supply arrangement it is also plausible that neither C nor D are correct since 3TE is still available to supply power to the B LPSW Pump (similar to the MDEFWP power supply arrangement).

Answer 0 Discussion Correci The 3B LPSWpump is fed from the 3TD switchgear. If the 3A pump trips and 3TD locks out there is no LPSW to provide cooling to the LPI coolers which results in a complete loss of DHR capability.

Basis for meeting the KA Requires knowledge of the relationship between having LPSW pumps available as a heat sink for DHR and the ability to provide Decay Heat Removal with the LPI system.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType j Question Source RO 1 Memory NEW Development References Student References Provided Obj. IC-ES R20 ES Power Supplies Ui HP Mode drawing Ul Switchover Mode drawing APEO25 AK2.03 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)

Service water or closed cooling water pumps 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 10 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 6 APEO26 AA1.O1 Loss of Component Cooling Water (CCW)

Ability to operate and! or monitor the following as they apply to the Loss of Component Cooling Water: (CFR 41.7/45.5 /45.6)

CCW temperature indications Given the following Unit 2 conditions:

Reactor power = 100%

LPSW to CC coolers is lost

CP 02A0068 (CC Cooler Outlet Temp) = 145°F increasing

Letdown temperature = 115°F increasing Which ONE of the following is the EARLIEST time that AUTOMATIC isolation of letdown will occur?

Letdown Temperature vs. Time U

(1) 0)

0)

I 0) 13 0) 4-i 0)

CL E

0)

F 0

13 0)

-J 1200 1210 1220 1230 1240 1250 Ti me A. 1205 B. 1210 C. 1215 D. 1220 Wednesday, August 31, 2011 Page 11 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 6 B General Discussion 02A0068 CC COOLER OUTLET TEMP 91 degrees F normal temperature at power Answer A Discussion Incorrect. Plausible since 130 degrees is the setpoint for the high letdown temperature statalarm.

Answer B Discussion Correct. If the letdown temperature reaches 130°F a high temperature stat-alarm will sound and at 135°F the letdown isolation valve, HP-5, will be interlocked closed to protect the demineralizer resin.

Answer C Discussion Incorrect. 140 degrees is plausible since it is the high temperature alarm setpoint for the OAC point monitoring CRD temps and CRDs are cooled by CC, Answer D Discussion Incorrect. Plausible since this is the maximum LOST temperature allowed by the limits and precautions of the HPI procedure.

Basis for meeting the KA Discussed this KA with Chief Examiner. Determined it acceptable to ask about temperature indication of components/systems cooled by CC since we take no actions based on actual CC outlet temp. This question requires the ability to monitor automatic actions that occur as a result of a loss of CC and its impact on letdown temperature.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj PNS-HPI R5, R8 PNS-HPI APEO26 AA1.01 Loss of Component Cooling Water (CCW)

Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: (CFR 41.7 / 45.5 / 45.6)

CCW temperature indications 401-S Comments: ks!Status Wednesday, August 31, 2011 Page 12 of 208

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2011B ONS SRO NRC Examination QUESTION 7 APEO27 2.2.3 Pressurizer Pressure Control System (PZR PCS) Malfunction APEO27 GENERIC (multi-unit license) Knowledge of the design, procedural, and operational differences between Units. (CFR: 41.5 141.6/ 41.7 / 4110 / 4512)

Given the following Unit 3 conditions:

Reactor power = 100%

3RC-1 has failed OPEN

3RC-3 will NOT close Which ONE of the following describes the Reactor Coolant Pump(s) that will be INITIALLY secured after the Reactor has been Manually tripped in accordance with API3IAJ 1700/044 (Abnormal Pressurizer Pressure Control)?

A. 3B1 ONLY B. 3B1 AND 3B2 C. 3A1 ONLY D. 3A1 AND 3A2 Wednesday, August 31, 2011 Page 13 of 208

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2011B ONS SRO NRC Examination QUESTION 7 General Discussion Answer A Discussion Incorrect. Plausible since on Unit 3 the Pzr spray line is located on the discharge of the 3Bl RCP and therefore securing this pump alone would significantly decrease the amount of Pzr spray through the failed open valves. Since the question asks which pumps will be initially secured it is plausible to believe that the AP would direct securing the spray pump only and then only securing other pumps if this were not sufficient.

Additionally plausible due to the process used to choose what pump to leave running and why. It is common practice to always leave the spray pump running when possible. In that context, leaving both pumps on in the loop with Pzr spray is not considered (as a function of ensuring Pzr spray available) therefore it would be plausible to believe that you only need to secure the RCP in the loop with the Pzr spray tap.

Answer B Discussion Correct. AP144 directs tripping the Rx and securing both the 3B1 and the 3B2 RCPs if RCS pressure cannot be controlled using 3RC-l and 3RC-3.

Answer C Discussion Incorrect. Plausible since the Pzr spray line is located on the discharge of the 1A1 RCP on unit I therefore securing the 3A1 RCP only is plausible based on the misconception that the Pzr spray line is on the A loop on Unit 3 as well. Under that misconception, securing the 3A1 RCP would significantly reduce spray flow through the failed valves and therefore make this choice plausible. since the question asks which pumps will be initially secured it is plausible to believe that the AP would direct securing the spray pump only and then only securing other pumps if this were not sufficient. Additionally plausible due to the process used to choose what pump to leave running and why. It is common practice to always leave the spray pump running when possible. In that context, leaving both pumps on in the loop with Pzr spray is not considered (as a function of ensuring Pzr spray available) therefore it would be plausible to believe that you only need to secure the RCP in the loop with the Pzr spray tap.

Answer D Discussion Incorrect. Plausible since this would be correct if the event occurred on Unit I.

Basis for meeting the KA Knowing the difference in Unit 1 vs. Unit 2&3 with regards to location of the Pzr spray line and differences in direction provided in AP/44 in relation to failed open spray valve and associated block valve meet the KA.

Basis for Hi Cog Basis for SRO only 1

Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided PNS-RCS R8 PNS-RCS 3AP/44 1 AP/44 APEO27 2.2.3 Pressurizer Pressure Control System (PZR PCS) Malfunction APEO27 GENERIC (multi-unit license) Knowledge of the design, procedural, and operational differences between units. (CFR: 41.5 / 41.6 /41.7 / 41.10 / 45.12) 401-9 Comments: emarks/Status Wednesday, August 31, 2011 Page 14 of 208

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2011B ONS SRO NRC Examination QUESTION 8 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 I conditions:

Initial conditions:

Time=1200:00

Reactor power = 100%

BOTH Main Feedwater pumps trip Current conditions:

Time=1201:30

Reactor power = 10% slowly decreasing

SCMs=O°F Which ONE of the following describes actions required (if any) in accordance with Rule 1 (ATWS/UNPP), Rule 2 (Loss of SCM) AND the UNPP tab?

A. Trip RCPs trip the Main Turbine B. Trip RCPs yj do NOT trip the Main Turbie C. Do NOT trip RCPs do trip the Main Turbine D. Do NOT trip RCPs do NOT trip the Main Turbine Wednesday, August 31, 2011 Page 15 of 208

FOR REVIEW ONLY DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 8 8 c General Discussion Answer A Discussion Incorrect. Tripping the RCPs is plausible since SCMs are 0 and it has been less than 2 minutes. If Rx power were < 1%, Rule 2 would direct these actions. Tripping the Main Turbine is correct.

Answer B Discussion Incorrect. Tripping the RCPs is plausible since SCMs are 0 and it has been less than 2 minutes. If Rx power were < 1%, Rule 2 would direct these actions. Not Tripping the Turbine is plausible since it would be correct if Main FDW were operating.

Answer C Discussion Correct. In accordance with Rule 2, RCPs are left on since power is> 1%. The UNPP tab will direct tripping the Main Turbine if Main FDW is lost.

Answer D Discussion Incorrect. Not tripping the RCPs is correct however not Tripping the Turbine is incorrect but plausible since it would be correct if Main FDW were operating.

Basis for meeting the KA Requires the ability to determine if the Main Turbine must be manually tripped during an ATWS.

Basis for H Cog Requires analyzing plant conditions and applying mitigation strategy of procedures to the analysis.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-UNPP R5, Rl 1 EAP-UNPP Rule I & 2 UNPP tab EPEO29 EAI.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 Wednesday, August 31, 2011 Page 16 of 208

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2011B ONS SRO NRC Examination QUESTION 9 9 BWEO5 EK3.4 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)

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.

Given the following Unit I conditions:

Reactor power = 100%

1A Main Steam Line Break occurs outside the Reactor Building Which ONE of the following:

1) is the LOWER Pressurizer level that will allow HPI to be throttled in accordance with Rule 5 ( Main Steam Line Break)?

2) describes the reason HPI must be throttled as soon as conditions allow?

A. 1. 24 increasing

2. Pressurized Thermal Shock conditions may develop as a result of repressu rization.

B. 1. 110 increasing

2. Pressurized Thermal Shock conditions may develop as a result of repressu rization.

C. 1. 24 increasing

2. Terminate the cold BWST water being injected to prevent exceeding Tech Spec Cooldown rates D. 1. 110 increasing

2. Terminate the cold BWST water being injected to prevent exceeding Tech Spec Cooldown rates Wednesday, August 31, 2011 Page 17 of 208

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2011B ONS SRO NRC Examination QUESTION 9 . 91 General Discussion Answer A Discussion Correct. HPI is throttled when Pzr level is on scale and increasing. Throttling HPI is required to prevent PTS concerns from rapid pressure increases created as Pzr level rapidly increases towards being water solid.

Answer B Discussion Incorrect. First part is plausible since it is> 100 and the other option is < 100. lOUis a common post trip pressurizer level. Controlling Pzr level> 100 is the level requirements established for this event per Rule 5 and it is therefore plausible to believe that the 100 threshold must be met prior to throttling HPI. Second part is correct.

Answer C Discussion Incorrect. First part is correct. Second part is plausible for 3 reasons. 1) HPI cooling can be a significant cooling medium responsible for a major portion of RCS cooling when HPI injection is occurring during certain Small Break Loca scenarios therefore the misconceptio that n HPI cooling could result in exceeding cooldown rates and is the bases for throttling during a MSLB is plausible. 2.) Since terminating the cooldown is a big concern during a MSLB, and cold water injection from the BWST does contribute some to the RCS cooling it is plausible to believe that the reason HPI must be throttled ASAP is due to the contribution to the cooldown being made by the HPI injection. 3) Following a MSLB, Rule 8

may require a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months hold of RCS pressure and temperature. Since HPI does contribute to the cooldown, it is plausible to have the misconception that HPI must be trrottled to control cooldown rates since Rule 8 directs a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months hold of PCS pressure and temperature.

Answer D Discussion Incorrect. First part is plausible since it is> 100 and the other option is < 100. 100 is a common post trip pressurizer level. Controlling Pzr level > 100 is the level requirements established for this event per Rule 5 and it is therefore plausible to believe that the 100 threshold must be met prior to throttling HPI. Second part is plausible for 3 reasons. I) HPI cooling can be a significant cooling medium responsible for a major portion of RCS cooling when HPI injection is occurring during certain Small Break Loca scenarios therefore the misconception that HPI cooling could result in exceeding cooldown rates and is the bases for throttling during a MSLB is plausible. 2.) Since terminating the cooldown is a big concern during a MSLB, and cold water injection from the BWST does contribute some to the RCS cooling it is plausible to believe that the reason HPI must be throttled ASAP is due to the contribution to the cooldown being made by the HPI injection. 3) Following a MSLB, Rule 8

may require a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months hold of RCS pressure and temperature. Since HPI does contribute to the cooldown, it is plausible to have the misconception that HPI must be throttled to control cooldown rates since Rule 8 directs a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months hold of PCS pressure and temperature.

Basis for meeting the KA The question requires knowledge of the reason for actions performed by an RO during an EHT event that ensure adherence to procedures and limitations in the facility license.

Basis for Hi Cog HI Cog since question requires an understanding of the reason behind actions taken during an EHT event.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj. EAP-EHT R2, Rl4 EAP-EHT BWEO5 EK3.4 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)

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 I

Wednesday, August 31, 2011 Page 18 of 208

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2011B ONS SRO NRC Examination QUESTION 10 io APEO54 AK3.04 Loss of Main Feedwater (MFW)

Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater (MFW): (CFR 41.5,41.10 / 45.6 / 45.13)

Actions contained in FOPs for loss of MFW Given the following Unit I conditions:

Reactor trip due to loss of Main Feedwater

Emergency Feedwater is NOT available

Condensate Booster Pump feed has been established

RCS leak = 80 gpm slowly increasing Which ONE of the following describes actions required and the reason for the actions in accordance with the LOHT tab:

A. Reduce running RCPs to one pump per loop to reduce heat input to RCS B. Reduce running RCPs to one pump per loop to reduce inventory lost from the RCS leak C. Reduce running RCPs to one to reduce heat input to RCS D. Reduce running RCPs to one to reduce inventory lost from the RCS leak Wednesday, August 31, 2011 Page 20 of 208

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2011B ONS SRO NRC Examination QUESTION 10 io General Discussion Answer A Discussion Correct. With CBP feed established, the LOHT tab will direct reducing number of running RCPs to one/loop while waiting on a source of feedwater. The reason is to reduce heat input to the RCS.

Answer B Discussion Incorrect. The number of RCPs is correct. The reason is plausible since there is a large RCS leak present and one of the reason that RCPs are secured during a loss of SCM event is to reduce the inventory lost out of the break.

Answer C Discussion Incorrect. Number of pumps is plausible since it would be correct if CBP feed were not available and HPI forced cooling had been established.

The reason is correct even for going to one pump in HPI FC.

Answer D Discussion Incorrect. Number of pumps is plausible since it would be correct if CBP feed were not available and HPI forced cooling had been established.

The reason is plausible since there is a large RCS leak present and one of the reason that RCPs are secured during a loss of SCM event is to reduce the inventory lost out of the break.

Basis for meeting the KA Requires knowledge of the reason reducing the number of RCPs to one/loop during a loss of heat transfer.

Basis for Hi Cog Requires analysis of conditions to determine appropriate actions that need to be taken.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-LOHT R2 EAP-LOHT APEO54 AK3.04 Loss of Main Feedwater (MFW)

Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater(MFW): (CFR 41.5,41.10 / 45.6/45.13)

Actions contained in EOPs for loss of MFW 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 21 of 208

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2011B ONS SRO NRC Examination QUESTION 11 EPEO55 EA1.O1 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)

In-core thermocouple temperatures Given the following Unit I conditions:

Initial conditions:

Reactor power = 100%

Current conditions:

Station Blackout (power has NOT been restored)

RCS Temperatures 2 minutes after trip

- Tc=550°F

- Th = 556°F

- CETCs = 558°F

SG Pressures = 1010 psig stable Which ONE of the following describes the response of RCS temperature indications during the transition but prior to establishing natural circulation flow?

RCSTcoId CETCs A. Stable Stable B. Stable Increasing C. Decreasing Stable D. Decreasing Increasing Wednesday, August 31, 2011 Page 22 of 208

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201 lB ONS SRO NRC Examination QUESTION 11 iij General Discussion Re-arranged answers Answer A Discussion Incorrect: Plausible in that Tc would be stable and CETC response could be correct if at low decay heat levels however this is a Rx trip from 100% therefore Decay heat would be high. Additionally, CETCs would be stable once Natural Circ had been established and stable.

Answer B Discussion Correct: In the time period following the trip natural circulation conditions will be developing. Thot and CETC will be increasing with Tcold being held constant by SO pressures in order to build in an adequate thermal driving head to establish flow. After flow is established CETC &

Thot will stabilize and eventually decrease as either decay heat level drops off or SO pressures are reduced Answer C Discussion Incorrect: Tcold is plausible based on assuming CETCs remain constant and Tc decreases tp establish the required delta T to ensure Natural Circulation flow occurs. CETC response could be correct if at low decay heat levels however this is a Rx trip from 100% therefore Decay heat would be high. Additionally, CETCs would be stable once Natural Circ had been established and stable.

Answer D Discussion Incorrect: Prior to stable Natural Circulation being established it is plausible to believe that Tc decreases since there is no flow in the loops and CETCs increase as part of the process of establishing the Delta T required for Natural Circ flow.

Basis for meeting the KA Requires the ability to determine the status of RCS heat removal based on the relationship between RCS Loop Temperatures and CETC temperature. The ability to monitor CETC for correct response requires the ability to predict the response of the CETCs during the blackout.

Basis for Hi Cog Requires analyzing plant conditions to determine response of temperature indications while establishing natural circulation.

Basis for SRO only 1

Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2009 NRC Q12 Development References Student References Provided Obj. TA-AM 1 R3 TA-AM1 EPEO55 EA1.01 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)

In-core thermocouple temperatures 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 23 of 208

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2011B ONS SRO NRC Examination QUESTION 12 APEO56 AK 1.03 Loss of Offsite Power Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: CFR 41.8/41.10/45.3)

Definition of subcooling: use of steam tables to determine it Given the following Unit 2 conditions:

Loss of Off Site power has occurred

Main Feeder Busses have been re-energized from CT-4

2A and 2B MDEFWPs will NOT start

RCS pressure = 1285 psig stable

RCS temperature = 577°F stable Which ONE of the following EOP tabs will be used to direct plant activities?

REFERENCE PROVIDED A. Subsequent Actions B. Blackout C. LOSCM D. LOHT Wednesday, August 31, 2011 Page 24 of 208

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2011B ONS SRO NRC Examination QUESTION 12 General Discussion Answer A Discussion Incorrect. Plausible since this would be the tab used to stabilize the plant if the RCS remained subcooled.

Answer B Discussion Incorrect. Plausible since this would be correct if the MFBs were de-energized. It is plausible to believe that you could still be in the Blackout tab even with Main Feeder Buses energized since simply fixing the entry condition for a tab is not necessarily a transfer point out of the tab. As an example, if you entered the LOHT tab you would not transfer out of it once CBP feed were established even though a loss of heat transfer no longer exists.

Answer C Discussion Correct. Using Attachment 5.18 provided as a reference, the RCS is in the saturated conditions area of the curve. With a loss of SCM, the entry conditions for LOSCM tab are met and a transfer to this tab would be appropriate.

Answer D Discussion Incorrect. Plausible since there has been a loss of off site power which means both Main Feedwater pumps will have tripped. With the failure of both MDEFWPs it is plausible to believe that the LOHT tab would be appropriate.

Basis for meeting the KA Received approval for the use of EOP End. 5.18 instead of steam tables with Chief Examiner. Question requires using End. 5.18 and plotting temperature vs. pressure to determine that you are in the Saturated Conditions section of the curve and therefore the operational implications of that would be to recognize that it meets the entry conditions for the LOSCM tab.

Basis for Hi Cog Requires plotting points on a graph and analyzing the results.

Basis for SRO only 1

Job Level Cognitive Level QuestionType j Question Source RO Comprehension NEW Development References Student References Provided EAP-SAR R20, R21 EOP End. 5.18 EOP End. 5.18 EAP-SAR APEO56 AKI.03 Loss of Offsite Power Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: CFR 41.8 / 41.10 / 45.3)

Definition of subcooling: use of steam tables to determine it 401.9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 25 of 208

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2011B ONS SRO NRC Examination QUESTION 13 l3 APEO57 2.4.46 Loss of Vital AC Electrical Instrument Bus APEO57 GENERIC Ability to veri that the alarms are consistent with the plant conditions. (CFR: 41.10/43.5 / 45.3/45.12)

Which ONE of the following represents the conditions of the statalarms on ISA-I for ES Actuation Logicchannelsone and twothatwill occurif the IDIB inverterinput breaker trips open?

A.

ES I TRIP ES2 TRIP B.

ES I TRIP ES 2 TRIP C.

ES I TRIP ES 2 TRIP D.

ES I TRIP ES2 TRIP Wednesday, August 31, 2011 Page 26 of 208

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2011B ONS SRO NRC Examination QUESTION 13 General Discussion Answer A Discussion Correct: If the Digital ES channels lose power, they fail in the untripped state. IDIB feeds KVIB. Since there is no auto backup to the vital panelboards and since 1KVIB feeds the EVEN digital channels, a loss of power to the IKVIB panelboard would not result in actuation of any of the digital ES channels.

Answer B Discussion Incorrect: KVIA panelboard provides power to the odd digital ES channels and KVIB feeds power to the even ES digital channels therefore reversing which panelboard feeds which channels is an easy misconception. RPS channels trip when they lose power and ES channels fail untripped when they lose power therefore reversing which ones trip and which ones do not trip is also a plausible misconception.

Answer C Discussion Incorrect. Plausible since KVIB panelboard (fed from 1DIB) feeds the even ES channels therefore this would be correct under the misconception that the digital ES channels actuate when they lose power (like the RPS channels do).

Answer D Discussion Incorrect: That both ES I and 2 are affected is plausible since KVIA feeds one half of the ES digital channels and KVIB feeds the other half therefore it is plausible to have a misconception about which specific channels are fed from which panelboard. Specifically, since there are 4 ES functions (Lo RCS pressure, Lo Lo RCS, Hi RB Pressure and HI HI RB pressure) and there are 4 Vital panelboards feeding the ES channels it would be plausible to believe that a single panelboard feeds each ES function.Since RPS channels trip when they lose power, it is plausible to believe that the ES channels affected would trip.

Basis for meeting the KA Requires the ability to interpret plant conditions based on a loss of a vital instrument panelboard and determine what alarms are consistent with those conditions.

Basis for Hi Cog Requires the ability to analyze impact of loss of inverter to ES digital channel alarms.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj. IC-ES R5, R12 APEO57 2.4.46 Loss of Vital AC Electrical Instrument Bus APE057 GENERIC Ability to verif that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5/45.3 / 45.12) 401-9 Comments: Remarks!Status Wednesday, August 31, 2011 Page 27 of 208

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2011B ONS SRO NRC Examination QUESTION 14 APEO58 AA2.02 Loss of DC Power Ability to determine and interpret the following as they apply to the Loss of DC Power: (CFR: 43.5/45.13) 125V dc bus voltage, low/critical low, alarm Unit I initial conditions:

Reactor power = 100%

ISA6/B2 INVERTER IDID SYSTEM TROUBLE actuated Current conditions:

NEO reports:

ISAI3/A8 INVERTER 1DID INPUT VOLTAGE LOW actuated

Inverter I DID output voltage low Based on the above conditions, which ONE of the following describes:

1) the plant response to the indications above?

2) actions directed by ISAI3/A8 if the inverter output voltage remains low?

A. 1. 1 D RPS channel trips

2. Transfer power for I KVID to Regulated Power Panel Board (1 KRA)

B. 1. 1 D RPS channel trips

2. Transfer DC bus IDID power to alternate unit (Unit 2 DCB)

C. 1. ICSAutoPowerislost

2. Transfer power for I KVID to Regulated Power Panel Board (1 KRA)

D. 1. ICS Auto Power is lost

2. Transfer DC bus I DID power to alternate unit (Unit 2 DCB)

Wednesday, August 31, 2011 Page 28 of 208

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2011B ONS SRO NRC Examination QUESTION 14 General Discussion Re-arranged answers Answer A Discussion Correct: IDID feeds KVID which supplies power to the 1D RPS channel. Loss of power to an RPS channel will result in the RPS channel tripping. There is no auto backup to the vital power panelboards therefore the ARG will direct transferring power to its backup source which is Regulated Power from IKRA.

Answer B Discussion Incorrect: First past is correct. Second part is plausible since 2DCB is the backup or alternate source for the 1DID inverter however it is an automatic backup via isolating diodes and would therefore automatically occur based on voltage from the normal source.

Answer C Discussion Incorrect: First part is plausbile since it would be correct for a loss of the Essential KI inverter instead of the Vital 1DID inverter. Second part is correct.

Answer D Discussion Incorrect: First part is plausbile since it would be correct for a loss of the Essential KI inverter instead of the Vital IDID inverter. Second part is plausible since 2DCB is the backup or alternate source for the 1DID inverter however it is an automatic backup via isolating diodes and would therefore automatically occur based on voltage from the normal source.

Basis for meeting the KA Requires the ability to interprete the consequences of low bus voltage statalarms that occur as a result of problems with Vital Inverters.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED NRC 2007 RO retest Q14 Development References Student References Provided Obj. EL-VPC R2, R5 EL-VPC APEO58 AA2.02 Loss of DC Power Ability to determine and interpret the following as they apply to the Loss of DC Power: (CFR: 43.5 / 45.13) 125V dc bus voltage, low/critical low, alarm 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 29 of 208

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2011B ONS SRO NRC Examination QUESTION 15 [ is APEO62 AK3.03 Loss of Nuclear Service Water Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: (CFR 41.4, 41.8 /45.7)

Guidance actions contained in EOP for Loss of nuclear service water Given the following Unit I conditions:

Turbine Building Flood tab initiated

Main and Emergency Feedwater have been lost Which ONE of the following describes:

I) how RCS decay heat will be removed in accordance with the TBF tab?

2) the reason for the guidance provided regarding how decay heat is removed?

A. 1. Using HPI Forced Cooling

2. Raw lake water will damage the SGs B. I. Using HPI Forced Cooling

2. SSF-ASW suction source is CCW and CCW will be isolated to stop the flooding C. I. Using SSF-ASW

2. In anticipation of losing power source to HPI pumps D. I. Using SSF-ASW

2. In anticipation of losing LPSW due to water in Turbine Building Wednesday, August 31, 2011 Page 30 of 208

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2011B ONS SRO NRC Examination QUESTION 15 l5 General Discussion Answer A Discussion Incorrect. First part is plausible since it would be correct in all other conditions regarding loss of main and emergency feedwater. Second part is plausible since it is a valid reason for using HPI FC before SSF-ASW.

Answer B Discussion Incorrect. First part is plausible since it would be correct in all other conditions regarding loss of main and emergency feedwater. Second part is plausible since CCW is the suction source for SSF-ASW and CCW is the most likely source of the flooding. There are actions taken to minimuze CCW water that can get to the Turbine Building since it is the most probable source of flooding however these action do not isolate the SSF ASWP from its suction source in the CCW inlet piping.

Answer C Discussion First part is correct. Second part is plausible since there is flooding occurring the turbine building therefore it is plausible to believe that there would be a concern over electrical shorts causing a loss of some or all electrical power sources located or connected to the turbine building.

These faults could feed back to the 4160V swgr and lock it out.

Answer D Discussion Correct. For this event, feeding SGs with raw water from SSF ASW or Station ASW is preferred over HPI forced cooling. HPI F/C is not preferred since once the BWST is depleted, water in the RBES is not expected to be available due to unavailability of LPSW for cooling. A plant cooldown will not be performed with Station ASW.

Basis for meeting the KA Discussed this KA with Chief Examiner on 5/31/11. Agreed using the LPI coolers will satisfy intent of KA. Requires knowledge of reasons for guidance contained in the EOP that are a result of loss of LPSW.

Basis for Hi Cog Requires knowledge of procedural guidance and the reasons behind the guidance.

Basis for SRO only

-1 Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj EAP-TBF R5, R6 EAP-TBF APEO62 AK3.03 Loss of Nuclear Service Water Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: (CFR 41.4, 41.8 /45.7)

Guidance actions contained in EOP for Loss of nuclear service water 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 31 of 208

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2011B ONS SRO NRC Examination QUESTION 16 161 APEO65 2.4.6 Loss of Instrument Air APEO65 GENERIC Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)

Given the following Unit I Conditions:

Time=1200

Reactor power = 100%

Instrument Air pressure = 87 psig slowly decreasing

Aux IA pressure = 78 psig decreasing

AP/1/A/1700/022 (Loss of Instrument Air) initiated Which ONE of the following describes the actions required in accordance with AP122 at Time = 1200?

A. Isolate RB Aux Coolers B. Bypass HPI Demineralizers C. Dispatch an operator to manually open ICC-8 D. MANUALLY Trip the Reactor then trip BOTH Main FDW pumps Wednesday, August 31, 2011 Page 32 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 16 j 16 General Discussion The ICs of this question represent a large leak on the AlA system such that Al is feeding AlA but is not quite able to maintain pressure. This could be due to mulitple leaks or problems with compressors feeding the IA system.

Answer A Discussion Incorrect. Plausible since this would be correct if Instrument Air pressure were < 80 psig.

Answer B Discussion Incorrect. Plausible since this would be correct if Instrument Air pressure were < 80 psig.

Answer C Discussion Correct. Once Auxiliary Instrument Air goes below 80 psig, AP/22 directs dispatching an operator to manually open CC-8.

Answer D Discussion Incorrect. Plausible since this is an action directed by AP122 based on IA pressure however the pressure threshold for this actions is 65 psig.

Basis for meeting the KA Requires knowledge of mitigation strategy used in AP/22 for decreasing IA pressure.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj EAP-APG R9 AP/22 APEO65 2.4.6 Loss of Instrument Air APEO65 GENERIC Knowledge of EOP mitigation strategies. (CFR: 41.10/43.5 / 45.13) 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 33 of 208

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2011B ONS SRO NRC Examination QUESTION 17 17j APE077 AK1.02 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)

Over-excitation Given the following Unit I conditions:

Reactor power = 100%

Generator output = 900 MWe stable

Generator Hydrogen pressure = 60 psig stable

MVARS oscillating between 200 MVARS and 300 MVARS due to Grid Disturbance

AP/34 (Degraded Grid) in progress Which ONE of the following describes the:

1) MAXIMUM MVARS allowed in accordance with AP/34 End. 5.1 (Generator Capability Curve)?

2) potential consequences of a grid disturbance that resulted in increasing MVARs above the limit?

REFERENCE PROVIDED A. 1. approximately 470

2. Excessive field heating B. 1. approximately 470

2. Excessive armature core end heating C. 1. approximately 350

2. Excessive field heating D. 1. approximately 350

2. Excessive armature core end heating Wednesday, August 31, 2011 Page 34 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 17 General Discussion Answer A Discussion Correct. Using the curve provided and conditions in the stem, 470 MVARS is the limit. This point on the curve is on the AB segment of the curve. The legend at the bottom of the curve explains that the AB segment is limited by field heating.

Answer B Discussion Incorrect. First part is correct. Second part is plausible since it is one of the options for the limiting factor of the capacity curve and if the 470 MVAR number were platted on the leading PF side of the curve it would result in being outside of the CD segment of the curve which means that the Armature core end heating limit would be being exceeded.

Answer C Discussion Incorrect. First part is plausible since it would be correct for a leading power factor. Second part is plausible since if would be correct if you mistakingly plotted 350 MW vs the 60 psig curve on the lagging PF side of the curve.

Answer D Discussion Incorrect. Plausible since this is correct if plotted on the leading PF side of the curve.

Basis for meeting the KA Requires knowledge of the operational implication of overexciting the T/G to the point the capability curve limits are exceeded. Additionally plausible since exceeding the MVAR limit on the lagging side of the curve is by definition over-exciting the generator.

Basis for Hi Cog Requires using the T/G capability curve and understanding the consequences of exceeding limits.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj STGOI5 R26,27 OP 110601 End 4.5 Cap Curve STGO15 AP/34 Cap Curve APEO77 AK1.02 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)

Over-excitation 401-9 Comments: RemarkslStatus Wednesday, August31, 2011 Page 35 of 208

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2011B ONS SRO NRC Examination QUESTION 18 18 BWEO4 EK1.3 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)

Annunciators and conditions indicating signals, and remedial actions associated with the (Inadequate Heat Transfer)

Given the following Unit I conditions:

Initial conditions:

Reactor power = 100%

TDEFWP isolated for repair Current conditions:

BOTH Main Feedwater pumps trip

IA and lB MDEFWPs fail

I SA2ID3 (RC Press High/Low) actuated Assuming NO Feedwater is restored to the SGs (Main, Emergency, CBP) which ONE of the following describes:

1) the next method of decay heat removal directed by the EOP?

2) a criteria used to deterrnine when the method of Decay Heat Removal above is required?

A. 1. Initiate HPI forced cooling

2. RCS pressure = 2300 psig B. I. Initiate HPI forced cooling

2. BOTH SGs = < 15 SU level C. I. Feed SGs with SSF ASW

2. RCS pressure = 2300 psig D. I. Feed SGs with SSF ASW

2. BOTH SGs = < 15 SU level Wednesday, August 31, 2011 Page 36 of 208

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2011B ONS SRO NRC Examination QUESTION 18 General Dscussion Answer A Discussion Correct. HPI FC is used if NO main or emergency feedwater is available to feed the SOs and it is initiated at 2300 psig RCS pressure.

Answer B Discussion Incorrect. First part is correct. Second part is plausible since 15 SO level is indicative of a dry SO and is therefore a logical place for proceeding to the next available source of decay heat removal. Additionally plausible since there is special considerations given for feeding dry SOs and 15 is the threshold value used during implementation of Rule 7 to determine acceptable feed rate. Additionally plausible since the concept of initiating HPI forced cooling is based on a loss of heat transfer and once there is no level in either SO, that is actually the point at which a loss of heat transfer occurs. Additionally plausible since some plants do use the condition of having dry SOs as the place where a transfer to the Loss of Heat transfer tab (or equivalent) should occur.

Answer C Discussion Incorrect. First part is plausible since it would be correct if main and emergency feedwater were lost due to flooding of the Turbine Building.

Second part is correct.

Answer D Discussion Incorrect. First part is plausible since it would be correct if main and emergency feedwater were lost due to flooding of the Turbine Building.

Second part is plausible since 15 SG level is indicative of a dry SG and is therefore a logical place for proceeding to the next available source of decay heat removal. Additionally plausible since there is special considerations given for feeding dry SOs and ISis the threshold value used during implementation of Rule 7 to determine acceptable feed rate.. Additionally plausible since the concept of initiating HPI forced cooling is based on a loss of heat transfer and once there is no level in either 5G. that is actually the point at which a loss of heat transfer occurs.

Additionally plausible since some plants do use the condition of having dry SOs as the place where a transfer to the Loss of Heat transfer tab (or equivalent) should occur.

Basis for meeting the KA Requires knowledge of the operational implications of a conditions indicating signal (RCS pressure) in the context of a complete loss of main and emergency feedwater.

Basis for Hi Cog Requires analyzing plant conditions and determining the correct procedural guidance based on the analysis.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj. EAP-LOHT RI EAP-LOHT BWEO4 EK1 .3 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)

Annunciators and conditions indicating signals, and remedial actions associated with the (Inadequate Heat Transfer) 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 37 of 208

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2011B ONS SRO NRC Examination QUESTION 19 APEOO1 2.4.45 Continuous Rod Withdrawal APEOOI GENERIC Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

Given the following Unit I conditions:

Initial conditions:

Reactor power = 75%

Current conditions:

ISA2/AI2 (ICS Tracking) in alarm

Neutron error = -5% (full scale low)

Reactor power = 80% increasing

BY REACTOR load limit on CTPD station illuminated

Feedwater flow increasing Which ONE of the following describes the event in progress?

A. Tave failed low B. Controlling NI failed low C. Continuous Rod Withdrawal D. Turbine Header Pressure failed high Wednesday, August 31, 2011 Page 38 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 19 General Discussion Answer A Discussion Incorrect. Plausible since it is plausible that with Tave failed low, rods would be pulling to increase Tave resulting in an increase in Rx power however the Tave error would modify feedwater to decrease to correct the perceived low Tave.

Answer B Discussion Incorrect. Plausible since these indications could be correct for a controlling NI failing high however if the controlling NI fails low the Diamond trips to hand therefore Feedwater would be decreasing as a result of the unit tracking NI power.

Answer C Discussion Correct. As rods pull, Rx power increases which results in neutron error decreasing since core thermal power demand does not change. At -5%

neutron error ICS goes to Track and Feedwater flow tracks Rx power to prevent a large mismatch which means that feedwater flow will also increase as power increases.

Answer D Discussion Incorrect. Plausible since with THP failed high, Rx power will initially increase and Neutron error will be on the negative side however Rx power would turn and begin to decrase since FDW flow will begin to decrease and rods will begin to insert due to the modifications to their demands from THP error.

Basis for meeting the KA Chief examiner said OK to diagnose a Continuous Rod Withdrawal from plant conditions including alarms. Question requires analyzing alarms and plant parameters to diagnose a continuous rod withdrawal.

Basis for Hi Cog Requires analyzing plant conditions to diagnose a malfunction.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj ICS R24,25 APEOO1 2.4.45 Continuous Rod Withdrawal APEOO I GENERIC Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 I 45.3 /45.12) 4O19 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 39 of 208

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2011B ONS SRO NRC Examination QUESTION 20 2O APEO24 AA2.O1 Emergency Boration Ability to determine and interpret the following as they apply to the Emergency Boration: (CFR: 43.5 / 45.13)

Whether boron flow and/or MOVs are malfunctioning, from plant conditions Given the following Unit I conditions:

Reactor power = 40% slowly decreasing

Rule I (ATWS/UNPP) in progress

I HP-24 and I HP-25 are OPEN

IA and IC HPI pumps operating

When IHP-26 switch was rotated to the OPEN position, both of its position indicator lights went dark

HPI flow and valve indications are as indicated below 98 347 750 760 000 600 p p __400 1307 300 1150 1160 0 _ 0 Which ONE of the following actions is directed next by Rule 1?

A. Open IHP-409 B. Open IHP-410 C. Start the lB HPI pump D. Dispatch operator to open CRD Breakers Wednesday, August 31, 2011 Page 40 of 208

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2011B ONS SRO NRC Examination QUESTION 20 20 General Discussion Answer A Discussion Incorrect. Plausible since this would be correct if the candidate had backwards which trains I-IP-409/4l0 fed or if 1HP-27 was not open.

Answer B Discussion Correct. Although the position indication for 1HP-26 indicates that the breaker or thermals may have tripped, HPI flow gage indicates that 1HP-26 is closed. OMP 1-2 requires using diverse indications to verify valve position. With RC Makeup flow approximately equal to Train flow, you can deduce that the HPI Train flow is because makeup flow is above the 60 gpm cutoff for HPI train flow gage and what you see on train flow is actually makeup flow and therefore 1HP-26 is actually closed requiring opening 1HP-410.

Answer C Discussion Incorrect. Plausible since A Train flow is low and the B HPI pump feeds the A train. Starting the B pump would actually increase flow under conditions that HP-26 was open or partially open.

Answer D Discussion Incorrect. Plausible since this would be correct if you determined that 1HP-26 was performing correctly which is plausible since there is some indication of flow in the A train.

Basis for meeting the KA Requires using plant indications to determine that 1HP-26 has malfunctioned and is closed.

Basis for Hi Cog Requires analyzing plant conditions to determine a malfunction using diverse indications.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided OIj ADM-OMP R6 EAP-UNPP R8 ADM-OMP Rule 1 Attach.

APEO24 AA2.01 Emergency Boration Ability to determine and interpret the following as they apply to the Emergency Boration: (CFR: 43.5 / 45.13)

Whether boron flow and/or MOVs are malfunctioning, from plant conditions 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 41 of 208

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2011B ONS SRO NRC Examination QUESTION 21 APEO33 AA1.03 Loss of Intermediate Range Nuclear Instrumentation Ability to operate and / or monitor the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR 41.7 / 45.5 I 45.6)

Manual restoration of power Which ONE of the following would require MANUAL restoration of power if the normal power supply is lost?

A. Wide Range INI-3 B. Turbine Bypass Valves C. Turbine Supervisory Instrumentation D. Main Feedwater Pump Motor Gear Unit Wednesday, August 31, 2011 Page 42 of 208

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2011B ONS SRO NRC Examination QUESTION 21 21 General Discussion Answer A Discussion Correct. NI-3 is in the B RPS cabinet and is powered from KVIB. KVIB is powered from the DIB inverter and the panelboard has no auto backup power supply.

Answer B Discussion Incorrect. Plausible sinceTB Vs are powered from an AC panelboard that is fed directly from an inverter similar to NI-3 however it is powered from either KI or KU which are powered from essential inverters and have an auto backup from regulated power.

Answer C Discussion Incorrect. Plausible since TSI equipment is powered from one of the essential panelboards (KIIKUIKX panelboards). The essential panelboards are backed up from Regulated power however it is an auto swap and does not require manual alignment. Plausibility comes from a misconception that TSI is powered from a vital panelboard OR from the misconception that the essential panelboards have no automatic backup.

TSI is powered from the KX panelboard.

Answer 0 Discussion Incorrect. Plausible sinceMGUs are powered from an AC panelboard that is fed directly from an inverter similar to NI-3 however it is powered from either KI which is powered from an essential inverter and has an auto backup from regulated power.

Basis for meeting the KA Discussed with chief examiner who agreed a power supply question could be used to meet intent of new KA. Question requires knowledge of th power supply to the Wide Range NIs.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj Digiati RPS R18 EL-VPC R7 Digital RPS EL-VPC APE03 3 AA 1.03 Loss of Intermediate Range Nuclear Instrumentation Ability to operate and / or monitor the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR 41.7 / 45.5 /

45.6)

Manual restoration of power 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 43 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 22 22 APEO36 AK2.O1 Fuel Handling Incidents Knowledge of the interrelations between the Fuel Handling Incidents and the following: (CFR 41.7 / 45.7)

Fuel handling equipment Unit I initial conditions:

Reactor in MODE 6

Core reload in progress Current conditions:

Fuel Transfer Canal level decreasing

East fuel carriage is positioned in the RB and contains a spent fuel assembly in the upender

West fuel carriage is in the SEP and empty

Reactor Building Main Fuel Bridge in transit to the East Upender to retrieve the Fuel Assembly

Section 4D (Fuel Transfer Canal Flooded) of AP/26 (Loss of Decay Heat Removal) initiated Based on the conditions above, which ONE of the following describes the flt action(s) required to be taken in accordance with Section 4D (Fuel Transfer Canal Flooded)?

A. Retrieve assembly from the East Upender with the Main Fuel Bridge then position the East carriage in the Spent Fuel Pool B. Verify SF system aligned for refueling cooling mode and stop 2B SF cooling pump C. Immediately position the East carriage in the Spent Fuel Pool D. Close 1SF-I and ISF-2 (East/West Transfer Tube Isolations)

Wednesday, August 31, 2011 Page 44 of 208

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201 lB ONS SRO NRC Examination QUESTION 22 22 General Discussion Re-arranged answers Answer A Discussion Incorrect. Plausible since it would be a reasonable misconception to believe that one of the upenders was designed to move assemblies from the RB and to the SFP and the other moves assemblies from the SFP to the RB. Under that misconception it would be required to retrieve the assembly in the upender prior to moving the carriage to the SFP. Additionally, the candidate may determine the assembly needs to be placed in its designated core location (one of the options listed in the AP if the assembly is in transient).

Answer B Discussion Incorrect: Plausible since these actions are directed by the AP and must be done prior to closing SF-1&2 however they are not done prior to positioning carriages in the SFP..

Answer C Discussion Correct: Since there is no fuel assembly in transient the first actions directed would be to position both fuel carriages in the SFP in preparation for closing 1SF-i & 2.

Answer D Discussion Incorrect: Plausible because these actions will be taken later to isolate the SFP and RB however both carriages must be placed in the SFP prior to closing SF-112.

Basis for meeting the KA Requires knowledge of the relationship between a fuel handling incident resulting in a decreasing fuel transfer canal water level and its impact on the operation of fuel handling equipment (Upenders and Fuel Carriage).

Basis for Hi Cog Requires analyzing plant conditions and equipment status to determine a correct course of action.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED NRC 2009A Q20 Development References Student References Provided Obj. EAP=APG R9, FH-FHS R7 2009A NRC Q20 AP/26 (Not Available Electronically)

APEO36 AK2.O1 Fuel Handling Incidents Knowledge of the interrelations between the Fuel Handling Incidents and the following: (CFR 41.7 / 45.7)

Fuel handling equipment 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 45 of 208

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2011B ONS SRO NRC Examination Q1J ESTION 23 231 APEO6 1 AA2.04 Area Radiation Monitoring (ARM) System Alarms Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: (CFR: 43.5 /45.13)

Whether an alarm channel is functioning properly Which ONE of the following describes ALL automatic actions that will occur as a result of a HIGH alarm on I RIA-4 (Reactor Building Hatch Monitor)?

A. Statalarm ONLY B. Statalarm AND a Local Alarm ONLY C. Statalarm AND the RB Evacuation alarm ONLY D. Statalarm, Local Alarm, AND the RB Evacuation alarm Wednesday, August 31, 2011 Page 46 of 208

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2011B ONS SRO NRC Examination QUESTION 23 23 General Discussion Answer A Discussion Incorrect. The H IGH alarm will actuate a Statalarm but will also result in a local alarm and the RB Evacuation alarm.

Answer B Discussion Incorrect. The H IGH alarm will actuate a Statalarm and a local alarm but will also result in the RB Evacuation alarm.

Answer C Discussion Incorrect. The H IGH alarm will actuate a Statalarm and the RB Evacuation alarm but will also actuate a local alarm.

Answer D Discussion Correct, the HIGH alarm will result in a statalarm, a local audible horn will sound, and the RB evacuation alarm will be activated.

Basis for meeting the KA Requires the ability to interpret the response of RIA-4 to determine if the alarm channel is ftinctioning properly (i.e. does the monitor actuate the RB evacuation alarm and since it does, is it actuated by the ALERT alarm or the HIGH alarm).

Basis for Hi Cog Basis for SRO only z

-1 Job L.evel Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj RAD-RIA R2 RAD-RIA APEO6I AA2.04 Area Radiation Monitoring (ARM) System Alarms Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: (CFR: 43.5 / 45.13)

Whether an alarm channel is functioning properly 401-9 Comments:

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2011B ONS SRO NRC Examination QUESTION 24 24 APEO67 AK3.04 Plant Fire On Site Knowledge of the reasons for the following responses as they apply to the Plant Fire on Site: (CFR 41.5,41.10 / 45.6/45.13)

Actions contained in EOP for plant fire on site Given the following Unit 1 conditions:

Fire in the turbine building

Reactor has been manually tripped

All Main and Emergency feedwater has been rendered NOT available

SSF-ASW aligned per APIOIAJI700IO25 (SSF Operating Procedure)

Which ONE of the following:

.1) states the MAXIMUM RCS pressure (psig) maintained in accordance with AP125?

2) describes the reason for the maximum RCS pressure?

A. 1. 2250

2. Minimize RCS inventory loss via the PORV and Code relief valves B. 1. 2250

2. Maximize Delta P across RCP seals to increase RCMUP seal injection C. 1. 2450

2. Minimize RCS inventory loss via the PORV and Code relief valves D. 1. 2450

2. Maximize Delta P across RCP seals to increase RCMUP seal injection Wednesday, August 31, 2011 Page 48 of 208

FOR REVIEW ONLY - DO NO T DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 24 General Discussion Per Chief Examiner, AOP actions accepted.

Answer A Discussion Correct.

If RCS pressure is not 2250 psig within 20 minutes, RCS inventory loss from the PORV/Codes (due to high pressure and lack of heat transfer) could create enough voiding to inhibit natural circulation, once the RCS is cooled to 555°F (TC).

Answer B Discussion Incorrect. First part is correct. Second part is plausible since there is a specific concern related to RCP seals and the RCMUP as described below:

RCS pressure is decreased below 2250 psig to ensure that the pressurizer code safety valves do not lift and to ensure that RCS pressure is below the pressure where the RCMU Pump discharge relief valve could weep or leak flow. RCS pressure must be decreased < 2250 psig to ensure that RCMU flow is not diverted from the RC pump seals.

Answer C Discussion Incorrect. First part is plausible since 2450 psig is the actuation setpoint of the PORV. Second part is correct.

Answer D Discussion Incorrect. First part is plausible since 2450 psig is the actuation setpoint of the PORV. Second part is plausible since there is a specific concern related to RCP seals and the RCMUP as described below:

RCS pressure is decreased below 2250 psig to ensure that the pressurizer code safety valves do not lift and to ensure that RCS pressure is below the pressure where the RCMU Pump discharge relief valve could weep or leak flow. RCS pressure must be decreased < 2250 psig to ensure that RCMU flow is not diverted from the RC pump seals.

Basis for meeting the KA Per Chief Examiner, AOP actions accepted. Requires knowledge of the reason that RCS pressure is reduced to < 2250 when using SSF-ASW via AP/25 following a plant fire.

Basis for Hi Cog Requires knowlede of the reason for actions directed by a procedure, Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj. EAP-SSF R29 EAP-SSF AP/25 APEO67 AK3.04 Plant Fire On Site Knowledge of the reasons for the following responses as they apply to the Plant Fire on Site: (CFR 41.5,41.10 / 45.6/45.13)

Actions contained in EOP for plant fire on site 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 49 of 208

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2011B ONS SRO NRC Examination QUESTION 25 25 BWAO4 AK2.1 Turbine Trip Knowledge of the interrelations between the (Turbine Trip) and the following:

(CFR: 41.7 /45.7)

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

Given the following Unit I conditions:

Reactor power = 100%

Which ONE of the following is an AUTOMATIC trip of the Main Turbine?

A. Condenser vacuum = 22 inches Hg B. AMSAC actuation C. Turbine Oil Fire D. AFIS actuation Wednesday, August 31, 2011 Page 50 of 208

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2011B ONS SRO NRC Examination QU ESTION 25 25 General Discussion Answer A Discussion Incorrect. Plausible since there is a low vacuum trip of the main turbine and 22 in HG is the point at which AP/27 (loss of condenser vacuum) requires a manual trip of the Main Turbine.

Answer B Discussion Correct. AMSAC will trip the Main Turbine and start all operable EFWPs need both channels of AMSAC/DSS to be enabled (2/2 logic) AND:

either Both MFPs have low hydraulic oil pressure (<75 psig)

Or Both MFPs have low discharge pressure (<770 psig)

Answer C Discussion Incorrect. Plausible since a Turbine Oil Fire is listed in the lesson plan as a turbine trip. There is a special trip mechanism that is activated during a Turbine Oil Fire that will trip the turbine and shutdown the Oil Pumps. Since it is a unique design, it is plausible to believe that it would be an automatic trip however the Turbine Oil Fire trip must be manually activated.

Answer D Discussion Incorrect. Plausible since AFIS is a system that is easily confused with AMSAC. AFIS does directly impact both Main and Emergency feedwater however it actuates from SG pressure and not Main Feedwater pump discharge pressure. The similarities in the systems impacted as well as the acronym make it plausible to confuse AFIS and AMSAC responses.

Basis for meeting the KA Requires knowledge of the interrelation between a safety system (AMSAC) and a Main Turbine trip.

Basis for Hi Cog Basis for SRO only Job Level RO L Cognitive Level Comprehension QuestionType NEW Question Source Development References Student References Provided Obj STG-EHC R23,10 STG-EHC BWAO4 AK2.1 Turbine Trip Knowledge of the interrelations between the (Turbine Trip) and the following:

(CFR: 41.7/45.7)

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

401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 51 of 208

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2011B ONS SRO NRC Examination QUESTION 26 [ 26 BWAO7 AKI .2 Flooding Knowledge of the operational implications of the following concepts as Normal, abnormal and emergency operating procedures associated with (Flooding).

Given the following Unit I condition:

Reactor power = 100%

Large COW leak occurs in Turbine Building Basement Which ONE of the following states the MAXIMUM Steam Generator level to be achieved in accordance with the Turbine Building Flood tab?

A. 95%O.R.

B. 50%O.R.

C. 240 XSUR D. 285 XSUR Wednesday, August 31, 2011 Page 52 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 26 26 General Discussion Answer A Discussion Correct. TBF tab directs increasing level to 95% OR at max allowable rate and specifies that this guidance supersedes Rule 7 Table 4 guidance.

Answer B Discussion Incorrect. Plausible since 50% OR is the level directed to be maintained when establishing Natural Circulation using Main Feedwater.

Additionally plausible since the ICs do not indicate a loss of main Feedwater.

Answer C Discussion Incorrect. Plausible since24O XSUR is the level used for Natural Circulation when on EFDW. Additionally plausible since the TBF tab takes actions based on the probability of losing pumps located in the TBB due to flooding.

Answer D Discussion Incorrect. Plausible since 285 XSUR is a threshold SG level used in the EOP to determine a course of action however it is the level at which SO blowdown is aligned during a SGTR.

Basis for meeting the KA Requires knowledge of the SO level requirements directed by the EOP associated with a Turbine Building Flood event. This matches the operational implications portion of the KA since the level must be achieved by manually controlling the feedwater valves during the SO level increase to 95%.

Basis for Hi Cog Basis for SRO only_____

Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj. EAP-TBF R4 EAP-TBF BWAO7 AK1.2 Flooding-Knowledge of the operational implications of the following concepts as Normal, abnormal and emergency operating procedures associated with (Flooding).

401-9 Comments RemarkslStatus Wednesday, August 31, 2011 Page 53 of 208

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2011B ONS SRO NRC Examination QUESTION 27 27 BWEO8 EKI.1 LOCA Cooldown Knowledge of the operational implications of the following concepts as they apply to the (LOCA Cooldown)

(CFR: 41.8/41.10/45.3)

Components, capacity, and function of emergency systems.

Which ONE of the following is the most complete list of Unit 3 conditions that ALL require 3LP-3 to be OPEN?

A. Normal LPI Decay Heat Removal B. Normal LPI Decay Heat Removal Post LOCA Boron Dilute flowpath aligned per the LOCA CD tab C. Normal LPI Decay Heat Removal Alternate Post LOCA Boron Dilute flowpath aligned per the LOCA CD tab D. Normal LPI Decay Heat Removal Post LOCA Boron Dilute flowpath aligned per the LOCA CD tab Alternate Post LOCA Boron Dilute flowpath aligned per the LOCA CD tab Wednesday, August 31, 2011 Page 54 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 27 27 C General Discussion Answer A Discussion Incorrect. Plausible since this would be correct for Unit 1, Answer B Discussion Incorrect. Plausible since the first part is correct and the second part would be correct concerning the Alternate path but not the normal flowpath.

Answer C Discussion Correct, 3LP-3 is in the DHR drop line and is required to be open for DHR alignment as well as for the alternate post loca boron dilution flowpath.

Answer D Discussion Incorrect. Plausible since 2 of the 3 conditions are correct and it is plausible to believe that both the normal and alternate boron dilution flowpaths need 3LP-3 to be opened since both flowpaths (normal and alternate) originate from the DHR drop line.

Basis for meeting the KA Requires knowledge of the functions of components necessary to align the Alternate Post Loca Boron Dilution flowpath (used only in emergency) per the LOCA CD tab of the EOP.

Basis for Hi Cog Requires analyzing system alignments and comparing the alignments to identify common components.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj PNS-LPI R27 PNS-LPI Boron Dilution Flowpath drawings BWEO8 EK1.1 LOCA Cooldown Knowledge of the operational implications of the following concepts as they apply to the (LOCA Cooldown)

(CFR: 41.8/41.10/45.3)

Components, capacity, and function of emergency systems.

401-9 Comments Remarks/Status Wednesday, August 31, 2011 Page 55 of 208

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2011B ONS SRO NRC Examination QUESTION 28 28 sYsoo3 K3.03 Reactor Coolant Pump System (RCPS)

Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: (CFR: 41.7 I 45.6)

Feedwater and emergency feedwater Given the following Unit 3 conditions:

Initial conditions:

3A & 3B Main FDW pumps tripped

ALL EFDWPs have started with 200 gpm EFW flow to each SG

3FDW-315 & 316 (3A and 3B OTSG EFW Flow Control Valves) in MANUAL

3A and 3B SG levels = 38 XSUR and stable

Unit 3 auxiliaries being provided by CT-5 Current conditions:

3FDW-315 & 316 are placed in Automatic Which ONE of the following describes the response of 3FDW-315 & 316?

A. Travel open to increases SG levels to 240 XSUR.

B. Travel open to increases SG levels to 50% on Operating level.

C. Travel closed to decrease SG level to 30 on XSUR.

D. Travel closed to decrease SG level to 25 on SU level.

Wednesday, August 31, 2011 Page 56 of 208

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2011B ONS SRO NRC Examination QUESTION 28 28 General Discussion Answer A Discussion Correct:. With RCP off the and the valves in automatic setpoint is 240 XSUR. Since level is below that setpoint the valves will open to increase level to 240.

Answer B Discussion Incorrect:. Plausible since it would be correct if on Main FDW.

Answer C Discussion Incorrect: Plausible since it would be correct if a RCP were operating however with CT-5 supplying auxiliaries, TA and TB are de-energized therefore no RCPs are operating.

Answer D Discussion Incorrect: Plausible since it would be correct if a RCP were operating and main feedwater were still available however with CT-5 supplying auxiliaries, TA and TB are de-energized therefore no RCPs are operating.

Basis for meeting the KA Requires knowledge of how a loss of RCPs affects FDW-315 and 316 when Emergency FDW is supplying the SGs.

Basis for Hi Cog Requires analyzing plant conditions to determine the setpoint that will be used for SG level control.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK CF023709 Development References Student References Provided Obj CF-EF R37 CF-EF SYSOO3 K3.03 Reactor Coolant Pump System (RCPS)

Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: (CFR: 41.7/45.6)

Feedwater and emergency feedwater 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 57 of 208

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2011B ONS SRO NRC Examination QUESTION 29 29 SYSOO4 A3.O1 Chemical and Volume Control System Ability to monitor automatic operation of the CVCS, including: (CFR: 41.7 / 45.5)

Water and boron inventory Given the following Unit I conditions:

. LDST level trend is as described below LDST Loud us. Time 0)

U 0) 0)

-J H

Co ci

-J 1200 1210 1220 1230 1240 1250 Time Which ONE of the following states the EARLIEST time that 1 HP-24 and I HP-25 will automatically open?

A. 1210 B. 1215 C. 1230 D. 1255 Wednesday, August 31, 2011 Page 58 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 29 29 C General Discussion Answer A Discussion Incorrect. Plausible since this time represents 60 LDST level which is the lo level alarm point.

Answer B Discussion tncorrect. Plausible since this time represents 55 LDST level which is the b-b level alarm point.

Answer C Discussion Correct. At 40 in LDST level, HP-24 & 25 will open to ensure adequate suction to the HPI pumps.

Answer D Discussion Incorrect. Plausible since this time represents 15 in the LDST. If 18 were assumed to be the setpoint for the interlock then this would be the correct answer. 18° is plausible since it is the bo level alarm point for the CC surge tank. It is plausible to confuse the CC surge tank and the LDST level setpoints.

Basis for meeting the KA Requires the ability to monitor the automatic swap of the HPI pump suction from its normal source of the LDST to the Borated Water Storage Tank based on available water inventory.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj PNS-HPI R8 PNS-HPI SYSOO4 A3.01 Chemical and Volume Control System Ability to monitor automatic operation of the CVCS, including: (CFR: 41.7 / 45.5)

Water and boron inventory 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 59 of 208

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2011B ONS SRO NRC Examination QUESTION 30 3O SYSOO4 A4.04 Chemical and Volume Control System Ability to manually operate and/or monitor in the control room: (CFR: 41/7 /45.5 to 45.8)

Calculation of boron concentration changes Given the following Unit I conditions:

Reactor power = 100%

RCS=625ppmb

IABHUT=975ppmb

IBBHUT=<l0ppmb

LDST level = 75 stable Which ONE of the following describes the water addition required (gallons) to raise LDST level to 90 and maintain RCS boron concentration approximately 625 ppmb?

A. IABHUT=298 lB BHUT = 167 B. IABHUT=231 IBBHUT= 129 C. IABHUT=167 lB BHUT= 298 D. IABHUT=129 lB BHUT= 231 Wednesday, August 31, 2011 Page 60 of 208

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2011B ONS SRO NRC Examination QUESTION 30 General Discussion A and B represent answers calculated using the hot and cold volumes. Answers C and D use a common math error using each volume. Since the approximate volume of water required to change RCS boron concentration does not have a thumb rule or other obvious methods that allow an operator to estimate the amount of water needed, the fact that A/B and C/B are significantly different volumes does not make C/D not plausible since there is no way to determine an approximate volume other than performing the correct calculation.

Answer A Discussion Correct Clvi +C2V2=CtVf Cl = 975 ppmb vi =x C2Oppmb V2 = Vf- Vi Cf= 625 ppmb Vf= 31 gal/in X 15 inches = 465 gal.

975X = 625 (465)

VI = 298 gal V2 = 465 298 = 167 gal Answer B Discussion Incorrect. Plausible since this number uses the gal/in of the Pressurizer (24 gal/in) instead of the gal/in of the LDST (31 gal/in) to calculate the volume of water needed to increase LDST level to 90 inches.

Answer C Discussion Incorrect. Plausible since this number reverses A BHUT and B BHUT requirments Answer D Discussion Incorrect. Plausible since this number reverses A BHUT and B BHUT requirments while using the Pzr gal/in to calculate the needed volume.

Basis for meeting the KA Requires the ability to calculate a batch addition to the LDST Basis for Hi Cog Requires performing algebraic calculations Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj OP-CPO16R1 OP-CPO16 SYSOO4 A4.04 Chemical and Volume Control System Ability to manually operate and/or monitor in the control room: (CFR: 41/7 / 45.5 to 45.8)

Calculation of boron concentration changes 4Ol9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 61 of 208

FOR REVIEW ONLY-DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 31 31 SYSOO5 2.4.9 Residual Heat Removal System (RHRS)

SYSOO5 GENERIC Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5/45.13)

Given the following Unit I conditions:

Initial conditions:

Time=1200

Reactor in MODE 5

Loops are full

Pzr level = 200 stable

HPI in operation

RC Makeup flow = 22 gpm

LPI aligned in High Pressure Mode Current conditions:

Time=1205

RC Makeup flow = 80 gpm increasing

Pzr level 140 decreasing Which ONE of the following describes the:

1) Abnormal Procedure that will be entered FIRST?

2) actions required at Time = 1205 by the Abnormal Procedure that was entered first?

A. 1. AP/2 (Excessive RCS Leakage)

2. Initiate EOP Enclosure 5.5 (Pzr and LDST Level Control)

B. 1. AP/2

2. Close IHP-5 C. 1. AP/26 (Loss of Decay Heat Removal)

2. Stop ALL LPI Pumps and close either ILP-1 or 1LP-2 D. 1. AP/26

2. Start all available HPI pumps AND remove White Tags then open I HP-409 and 1HP-410 Wednesday, August 31, 2011 Page 62 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 31 A General Discussion Answer A Discussion Correct. Entry conditions for API2 are met. For these conditions, AP/26 is entered when directed by AP/2. Prior to AP/2 directing to Go To AP/26, direction is given to initiate End. 5.5.

Answer B Discussion Incorrect. First part is correct. Second part is plausible since it would be correct if the leak were> 100 gpm per IMA of AP/2.

Answer C Discussion Incorrect. First part is plausible since AP/26 is entered and provides strategy for leak identification and isolation as well has strategy for LPI system manipulations. It has specific sections for loss of RCS inventory in this conditions which adds to plausibility. Second part is plausible since it is the actions taken ifall available HPI cannot maintain Pzr level.

Answer D Discussion Incorrect. First part is plausible since AP/26 is entered and provides strategy for leak identification and isolation as well has strategy for LPI system manipulations. It has specific sections for loss of RCS inventory in this conditions which adds to plausibility. Second part is plausible since there is a NOTE in IMAs that defines all available HPI as using HP-409 and 410. The Note is actually a precursor to an IAAT for stopping LPI pumps Basis for meeting the KA Requires knowledge of actions required when shutdown and on DHR to mitigate a loss of coolant event.

Basis for Hi Cog Requires analyzing plant conditions and applying that analysis to procedural guidance for the event.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj. EAP-APG R9 AP/26 IMAs AP/2 SYSOO5 2.4.9 Residual Heat Removal System (RHRS)

SYSOO5 GENERIC Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) 401-9 Comments: Rema rkslStatus Wednesday, August 31, 2011 Page 63 of 208

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2011B ONS SRO NRC Examination QUESTION 32 32 SYSOO6 A3.06 Emergency Core Cooling System (ECCS)

Ability to monitor automatic operation of the ECCS, including: (CFR: 41.7/45.5)

Valve lineups Given the following Unit I conditions:

Initial conditions:

Time=1200

Reactor power = 20% stable

SBLOCA occurs Current conditions:

Time=1210

RCS pressure = 410 psig slowly decreasing

RB pressure = 2.7 psig slowly increasing Which ONE of the following contains ONLY valves that received an Engineered Safeguards signal to OPEN at Time = 1210?

A. IHP-24ANDIHP-5 B. IHP-24 AND 1J-7.

C. ILP-18 AND lBS-I D. ILP-18 AND ILPSW-15 Wednesday, August 31, 2011 Page 64 of 208

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2011B ONS SRO NRC Examination QUESTION 32 32 General Discussion Answer A Discussion Incorrect. While both components are on ES-i&2, HP-5 gets a signal to Close for containement isolation.

Answer B Discussion Correct. With RCS pressure below 550 psig ES 1-4 have actuated. Since RB pressure is below 3 psig, ES 5-8 have not actuated. 1HP-24 opens on ES-i and 1LP-17 opens on ES-3 therefore they have both received a signal to open.

Answer C Discussion Incorrect. 1LP-18 would be open since it is on ES-4 however lBS-i is on ES-7 and would therefore not have received a signal to open, Answer 0 Discussion Incorrect, 1LP-18 would be open since it is on ES-4. Although 1LPSW-15 is on both ES-5 and ES-5, the setpoint for ES-5&6 is 3 pisg and therefore they would not have actuated, Basis for meeting the KA Requires the ability to monitor proper vavle alignment based on auto actuation of ECCS systems.

Basis for Hi Cog Requires analyzing plant conditions to determine appropriate ES channel actuation and applying knowledge of what actuates from each channel.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj. IC-ES R18, R14 IC-ES RZ module drawing SYSOO6 A3.06 Emergency Core Cooling System (ECCS)

Ability to monitor automatic operation of the ECCS, including: (CFR: 41.7/45.5)

Valve lineups 401-9 Comments Remarks/Status Wednesday, August 31, 2011 Page 65 of 208

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2011B ONS SRO NRC Examination QIJESTION 33 33 SYSOO6 K2.04 Emergency Core Cooling System (ECCS)

Knowledge of bus power supplies to the following: (CFR: 41.7)

ESF AS-operated valves Which ONE of the following states the power supply for 3LP-18?

A. 3XS-1 B. 3XS-2 C. 3XS-4 D. 3XS-5 Wednesday, August 31, 2011 Page 66 of 208

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2011B ONS SRO NRC Examination QU ESTION 33 33 General Discussion Answer A Discussion Incorrect. Plausible since it would be correct for I LP- 17 Answer B Discussion Correct. 1LP-18 is powered from 3XS-2.

Answer C Discussion Incorrect. Plausible since it would be correct for 1LP-19 Answer D Discussion Incorrect. Plausible since it would be correct for 1LP-20 Basis for meeting the KA Requires knowledge of bus power supplys for ESFAS valves.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj. IC-ES R20 IC-ES SYSOO6 K2.04 Emergency Core Cooling System (ECCS)

Knowledge of bus power supplies to the following: (CFR: 41.7)

ESFAS-operated valves 401-9 Comments:

1 i,arks/Status Wednesday, August 31, 2011 Page 67 of 208

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201 lB ONS SRO NRC Examination QUESTION 34 34 SYSOO7 K5.02 Pressurizer Relief Tank/Quench Tank System (PRTS)

Knowledge of the operational implications of the following concepts as they apply to PRTS: (CFR: 41.5 / 45.7)

Method of forming a steam bubble in the PZR Unit I plant conditions:

OP/l/A111031002, (Filling and Venting RCS) Enclosure 4.14 (Establishing Pzr Steam Bubble And RCS Final Vent) in progress

Quench Tank level = 84 inches

Quench Tank pressure = 0.5 psig

The Pressurizer is vented to the Quench Tank for 30 minutes Based on the above conditions, which ONE of the following describes QT parameters that would indicate that Pzr Steam Bubble Formation is complete?

QT level (inches) QT pressure (psig)

A. 84.1 0.6 B. 84.1 2.5 C. 86.1 0.6 D. 86.1 2.5 Wednesday, August 31, 2011 Page 68 of 208

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2011B ONS SRO NRC Examination QUESTION 34 34 General Discussion Answer A Discussion Incorrect: Plausible if you have the misconception that the Pzr vent line is above the water level in the QT and therefore both QT pressure and level increase only slightly as some but not all of the steam would condense as it was vented to the QT Answer B Discussion Incorrect: Plausible if you have the misconception that the Pzr vent line is above the water level in the QT and therefore the steam being vented to the QT would cause pressure to increase with little impact on level.

Answer C Discussion CORRECT: Per OP/i 103/002, Pzr steam bubble formation is complete (i.e., all the N2 gas is vented out of the Pzr) when a change (rise) in QT pressure of less than 0.2 psig occurs and QT level increases by 2 inches. Since the Pzr vent is underwater in the QT, when N2 is being vented it will rise to the surface and cause a corresponding increase in QT pressure therefore minimal pressure response is a sign that all of the N2 has been vented. Additionally, as water is vented it is condensed under the water level of the QT therefore minimal QT pressure change in conjunction with increasing QT level is indicative of all N2 being out of Pzr.

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

Basis for meeting the KA Requires knowledge of the QT operational parameters (pressure and level changes) that indicate Pzr steam bubble formation is complete Basis for Hi Cog Requires analyzing plant data and applying procedural guidance to determine if steam bubble exists.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED Modified NRC 2009A Q34 Development References Student References Provided Obj PNS-PZR R17 OP/l/A/1103/002, End. 4.14 SYSOO7 K5.02 Pressurizer Relief Tank/Quench Tank System (PRTS)

Knowledge of the operational implications of the following concepts as they apply to PRTS: (CFR: 41.5 / 45.7)

Method of forming a steam bubble in the PZR 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 69 of 208

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2011B ONS SRO NRC Examination QU ESTION 35 35 SYSOO8 K2.02 Component Cooling Water System (CCWS)

Knowledge of bus power supplies to the following: (CFR: 41.7)

CCW pump, including emergency backup Given the following Unit I conditions:

Initial conditions:

Time=1200

Reactor power = 100%

Current conditions:

Time=1201

Switchyard isolation occurs

CTI lockout occurs Which ONE of the following states the:

1) time required following the Reactor trip for the CC pumps to regain power?

2) source of power provided to the CC pumps?

A. 1.21 seconds

2. lXOandlXP B. 1.21 seconds

2. IXLandIXN C. 1.31 seconds

2. IXOandIXP D. 1.31 seconds

2. IXLandIXN Wednesday, August 31, 2011 Page 70 of 208

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2011B ONS SRO NRC Examination QUESTION 35 35 General Discussion Reworded the stem and re-arranged answers Answer A Discussion Incorrect: Power from XL & XN. Time = 31 sec. Plausible because XO and XP are also 600v supplies to vital equipment. 21 sec plausible because MFBMP and Load shed add up to 21 seconds.

Answer B Discussion Incorrect: Time = 31 sec. Plausible because MFBMP and Load shed add up to 21 seconds if the transfer to standby time is neglected.

Answer C Discussion Incorrect: Power from XL & XN. Plausible because XO and XP are also 600v supplies to vital equipment.

Answer D Discussion Correct: SWYD isolate, KHU emergency start, MFBMP (20 sec) followed by a load shed (lsec) and a transfer to standby (10 sec) occurs for a total of3l seconds.

Basis for meeting the KA Requires knowledge of power supplies to CC pumps and how they are powered following a blackout.

Basis for Hi Cog Requires analyzing plant data and determining the power switching logic response to regain power to MFBs.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK NRC 2007 Retest Q35 Development References Student References Provided Obj EL-PSL Rl4 Power supplies, PNS-PSL SYSOO8 K2.02 Component Cooling Water System (CCWS)

Knowledge of bus power supplies to the following: (CFR: 41.7)

CCW pump, including emergency backup 401-9 Comments RemarkslStatus L

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2011B ONS SRO NRC Examination QUESTION 36 36 SYSO1O K5.O1 Pressurizer Pressure Control System (PZR PCS)

Knowledge of the operational implications of the following concepts as the apply to the PZR PCS: (CFR: 41.5 / 45.7)

Determination of condition of fluid in PZR, using steam tables Given the following Unit 3 conditions:

Initial conditions:

Reactor power = 100%

Switchyard Isolation occurs Current Conditions:

Natural Circulation established

RCS pressure = 2155 psig

Tcold = 550° F stable

Pressurizer level = 220 stable

Pressurizer temperature = 628°F

1) The Pressurizeris_(1)_.

2) Pressurizer Heater Bank #2 (Groups B & D) heaters are _(2)_.

Which ONE of the following completes the statements above?

A. 1. saturated

2. energized B. 1. subcooled

2. energized C. 1. saturated

2. NOT energized D. 1. subcooled

2. NOT energized Wednesday, August 31, 2011 Page 72 of 208

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2011B ONS SRO NRC Examination QUESTION 36 36 General Discussion Answer A Discussion Incorrect. First part is plausible since it would be correct for normal Pzr temperatures. With RCS pressure, Tcold, and Pzr level at their normal values it is plausible to believe that the Pzr is in its normal state of saturated. Second part is correct.

Answer B Discussion Correct. With RCS pressure at 2150 psig, saturation temperature for that pressure is approximately 648 degrees F. With the Pressurizer temp at 628 degrees, the Pzr is subcooled. Bank 2 heaters are used in the Pzr saturation recovery circuit. As long as RCS pressure is at least 20 psig from saturation pressure of the Pzr these heaters would be energized. Additionally, the heaters are fed from lX8 which do not load shed therefore even following the Switchyard isolation, the heaters would be energized since the Pzr is subcolled by about 350 psig.

Answer C Discussion Incorrect. First part is plausible since it would be correct for normal Pzr temperatures. With RCS pressure, Tcold, and Pzr level at their normal values it is plausible to believe that the Pzr is in its normal state of saturated. Second part is plausible since RCS pressure is at 2155 therefore is the Pzr were actually saturated the Bank 2 heaters would be OFF since they turn off at 2150 psig.

Answer D Discussion Incorrect. First part is correct. Second part is plausible even as it relates to the first part based on the fact that a Switchyard isolation has occurred which makes it plasuible that iven if the Pzr were subcooled the heaters would not be on.

Basis for meeting the KA This question requires determining that the Pzr is Subcooled using steam tables.

Basis for Hi Cog Requires calculations and use of steam tables.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj. PNS-PZR R22,29 PNS-Pzr SYSOIO K5.01 Pressurizer Pressure Control System (PZR PCS)

Knowledge of the operational implications of the following concepts as the apply to the PZR PCS: (CFR: 41.5 / 45.7)

Determination of condition of fluid in PZR, using steam tables 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 73 of 208

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2011B ONS SRO NRC Examination QUESTION 37 SYSOIO K6.O1 Pressurizer Pressure Control System (PZR PCS)

Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: (CFR: 41.7 / 45.7)

Pressure detection systems Which ONE of the following failure combinations will result in I RC-1 (Pzr Spray Valve) failing OPEN when in Automatic?

A. lAand lBNRRCSpressurefailed HIGH B. I B and 10 NR RCS pressure failed HIGH C. 10 and lDNRRCSpressurefailed HIGH D. I D and I E NR RCS pressure failed HIGH Wednesday, August 31, 2011 Page 74 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 37 37 A General Discussion Answer A Discussion Correct. Median selected NR RCS pressure is used to feed 1RC-l automatic operation. The circuit uses A, B, and E RPS NR pressure signals therefore if both A and B had failed high, 1RC-l would respond by opening when in automatic.

Answer B Discussion Incorrect. Plausible since there is a combination of 2 RPS NR RCS pressure signals that if failed high will result in 1RC-l failing open in automatic.

Answer C Discussion Incorrect. Plausible since there is a combination of 2 RPS NR RCS pressure signals that if failed high will result in 1RC-l failing open in automatic.

Answer D Discussion Incorrect. Plausible since there is a combination of 2 RPS MR RCS pressure signals that if failed high will result in 1RC-l failing open in automatic.

Basis for meeting the KA Requires knowledge of the effect that a malfunction of RCS NR pressure signals will have on operation of 1RC-l (pressurizer spray valve).

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided IC-RCI R61 IC-RCI SYSOIO K6.Ol Pressurizer Pressure Control System (PZR PCS)

Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: (CFR: 41.7 /45.7)

Pressure detection systems 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 75 of 208

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2011B ONS SRO NRC Examination QUESTION 38 38 SYSO12 A2.02 Reactor Protection System (RPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; 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 /45.5)

Loss of instrument power Given the following Unit I conditions:

Reactor power = 100%

IA RPS Thot RTD power supply is lost Which ONE of the following describes:

1) ALL RPS trips affected by the failure?

2) the actions preferred in accordance with OP/I/A/I 105/014 (Control Room Instrumentation Operation And Information)?

REFERENCE PROVIDED A. 1. RCS High Outlet Temperature ONLY

2. Place MANUAL TRIP Keyswitch in TRIP.

B. 1. RCS High Outlet Temperature ONLY

2. Place affected RPS Channel MANUAL BYPASS keyswitch in BYP.

C. 1. RCS High Outlet Temperature and RCS Variable Low Pressure

2. Place MANUAL TRIP Keyswitch in TRIP.

D. 1. RCS High Outlet Temperature and RCS Variable Low Pressure

2. Place affected RPS Channel MANUAL BYPASS keyswitch in BYP.

Wednesday, August 31, 2011 Page 76 of 208

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2011B ONS SRO NRC Examination QES[ON 38 38 General Discussion We have 4 RPS channels. TS 3.3.1 only requires 3 to be operable. OP/i 105/014 gives guidance on what to do if problems arise with RPS channels.

Answer A Discussion Incorrect. First part is plausible since it is the only trip function in RPS with high temperature in its name. Second part is plausible since it would be correct if this were a required RPS channel. However only 3 RPS channels are required JAW TS 3.3.1 and since there are no other conditions given, the channel with the failed NI would be considered not required.

Answer B Discussion Incorrect. Incorrect. First part is plausible since it is the only trip function in RPS with high temperature in its name.Second part is correct.

Answer C Discussion Incorrect. First part is correct. Second part is plausible since it would be correct if this were a required RPS channel. However only 3 RPS channels are required lAW TS 3.3.1 and since there are no other conditions given, the channel with the failed NI would be considered not required.

Answer D Discussion Correct. The High Outlet Temperature trip uses Thot directly to determine if the trip setpoint has been reached. The Variable Low Pressure trip uses Thot in the formula to caculate the low pressure trip:

l1.l4Thot -4706 Since this is NOT a required channel, 1105/014 directs (per a note saying it is preferred) placing the channel in Manual Bypass is correct.

Basis for meeting the KA Requires ability to precinct the impact of a loss of power supply to a power range NI and the ability to use the procedure to determine the correct actions to take based on the failure.

Basis for Hi Cog Requires analyzing plant conditions and determining the correct actions based on guidance in a procedure.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj IC-NI R28 ADM-ITS R8 ADM-PIS R3 OP/1105/014 End 4.7 OP/i 105/14 IC-NI ADM-ITS ASM-PIS SYSOI2 A2.02 Reactor Protection System (RPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; 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 / 45.5)

Loss of instrument power 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 77 of 208

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2011B ONS SRO NRC Examination QUESTION 39 39 SYSO12 K1.06 Reactor Protection System (RPS)

Knowledge of the physical connections and/or cause effect relationships between the RPS and the following systems: (CFR: 41.2 to 41.9 I 45.7 to 45.8)

T/G Given the following Unit 3 conditions:

Reactor power = 31% decreasing

Main Turbine trips Which ONE of the following describes the plant response to the Main Turbine trip?

A. Reactor will automatically runback to 20% power and stabilize B. Reactor will automatically runback to 15% power and stabilize C. Reactor will trip and TBVs will use Turbine Header Pressure error as controlling signal D. Reactor will trip and TBVs will use SG Outlet Pressure error as controlling signal Wednesday, August 31, 2011 Page 78 of 208

FOR REVIEW ONLY DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 39 39 General Discussion Answer A Discussion Incorrect. Plausible since a runback would occur if power were <27.75% and 20% is the runback setpoint for a runback for Both Generator Breakers Open, Answer B Discussion Incorrect. Plausible since a runback would occur if power were <27.75% and 15% is the runback setpoint for a Maximum Runback via the pushbutton on the LCP.

Answer C Discussion Incorrect. Plausible since a Rx trip would occur and Turbine Header Pressure error is the normal control signal for the Turbine Bypass Valves.

Answer D Discussion J

Correct. Once above 29.75% the Turbine to Rx RPS trip is activated. Since power is 31% the trip would be active therefore the Rx would trip.

When the Turbine trips the Turbine bailey station will trip to Hand which results in transferring control of the TBVs from Turbine Header Pressure error to OTSG Outlet Pressure error.

Basis for meeting the KA Requires knowledge of the cause and effect relationship between a T/G trip and an RPS trip.

Basis for Hi Cog Requires analyzing plant conditions and then applying those conditions to knowledge of RPS and ICS operations Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj STG-ICS R3, RiO IC-RPS R3 STG-ICS Chptr 2&3 IC-RPS SYSO12 Kl.06 Reactor Protection System (RPS)

Knowledge of the physical connections and/or cause effect relationships between the RPS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

T/G 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 79 of 208

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SYSO13 K2.O1 Engineered Safety Features Actuation System (ESFAS)

Knowledge of bus power supplies to the following: (CFR: 41.7)

ESFAS/safeguards equipment control Given the following Unit I conditions:

Initial conditions:

Reactor power = 100%

I KVIB Pan elboard de-energized Current conditions:

MSLB inside the Reactor Building occurs

Lowest RCS pressure = 1137 psig

Reactor Building Pressure peaked at 32 psig Which ONE of the following describes ALL ES Actuation Logic Channels that have actuated?

A. 1,3,5,7 B. 2,4,6,8 C. 1,5,7ONLY D. 2,6,8ONLY Wednesday, August 31, 2011 Page 80 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 201 lB ONS SRO NRC Examination QUESTION 40 40 General Discussion Answer A Discussion Correct. KVIB provides power to the even digital channels. With KVIB de-energized, the even channels cannot actuate. Since RB pressure h exceeded its actuation setpoint of 10 psig, all ES channels should have actuated however since the even channels cannot actuate due to loss of power, channels 1, 3, 5, and 7 will have actuated.

Answer B Discussion Incorrect. Plausible since this would be correct if KVIA were de-energized instead of KVIB.

Answer C Discussion Incorrect. Plausible since RCS pressure has not reached the LPI injection setpoint of 550 psig and the power supplys to the Actuation Logic channels is split based on odd and even channels. Channel I&2 RCS pressure setpoint has already been reached therefore under the misconception that ES channels 3 and 4 (LPI) only actuate on low RCS pressure this would be correct however exceeding 3.0 psig in RB pressure results in actuating ES 1-6.

Answer D Discussion Incorrect. Plausible since RCS pressure has not reached the LPI injection setpoint of 550 psig and the power supplys to the Actuation Logic channels is split based on odd and even channels. Channel I &2 RCS pressure setpoint has already been reached therefore under the misconception that ES channels 3 and 4 (LPI) only actuate on low RCS pressure this would be correct however exceeding 3.0 psig in RB pressure results in actuating ES 1-6. KVIB does impact only the even ES logic channels.

Basis for meeting the KA Discussed KA with Chief Examiner. Determined that question on either control power or power to ES digitals would match.

Requires knowledge of ES powers components power supplies.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj IC-ES R2, R26 IC-ES SYSOI3 K2.01 Engineered Safety Features Actuation System (ESFAS)

Knowledge of bus power supplies to the following: (CFR: 41.7)

ESFAS/safeguards equipment control 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 81 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination 41 4I SYSO13 K5.02 Engineered Safety Features Actuation System (ESFAS)

Knowledge of the operational implications of the following concepts as they apply to the ESFAS: (CFR: 41.5 / 45.7)

Safety system logic and reliability Given the following Unit I conditions:

Initial conditions:

Reactor power = 25%

1AMSLB inside the RB occurs Current conditions:

1A NR RB pressure = 2.9 psig

1 B NR RB pressure = 3.1 psig

1 C NR RB pressure = FAULTED Which ONE of the following states ALL ES Instrument Channels (if any) that have sent an actuation signal to ES Actuation Logic Channels 1-6?

A. ICI and 1C2 ONLY B. IBI and 1B2 ONLY C. 1BI, 1B2, ICI, and 1C2 D. NONE Wednesday, August 31, 2011 Page 82 of 208

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2011B ONS SRO NRC Examination QUESTION 41 4l General Discussion Answer A Discussion Incorrect. Plausible since the required TS setpoint for RB pressure is <4 psig so it is plausible to believe that the lB Voters have NOT actuated and that only the Voters with the faulted input would be actuated. Additional plausibility comes from the operation of the OLD ES Analog channels where each channel operated independently of the others inputs.

Answer B Discussion Incorrect. Plausible since this would be correct based on the old ES analog channel logic where each channel operates independently of the other channels inputs.

Answer C Discussion Incorrect. Plausible since the B inputs are > setpoint and it would be plausible to believe that a faulted signal would result in the conservative action of actuating the output of the associated Voters.

Answer D Discussion Correct.

Input pressure in Engineering Units is compared to a Comparator setpoint of 3.0 psig.

If the signal is FAULTED, it is blocked from use in any other location in the system other than the functional trip statalarm.

UNFAULTED signals are then compared to the other (2) instrument channel inputs. The 2.MAX is calculated and used for the 2.MAX selection for actuation and to the other two instrument channels.

Basis for meeting the KA Requires knowledge of the Operational implications of the ES system logic.

Basis for Hi Cog Requires analyzing RB pressure inputs in context of the ES actuation logic to determine the status of the ES Voters.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj IC-ES R2l, R23 IC-ES SYSOI3 K5.02 Engineered Safety Features Actuation System (ESFAS)

Knowledge of the operational implications of the following concepts as they apply to the ESFAS: (CFR: 41.5 / 45.7)

Safety system logic and reliability 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 83 of 208

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2011B ONS SRO NRC Examination QUESTION 42 42 SYSO22 K1.O1 Containment Cooling System (CCS)

Knowledge of the physical connections and/or cause-effect relationships between the CCS and the following systems: (CFR: 41.2 to 41 .9 / 45.7 to 45.8)

SWS/cooling system Which ONE of the following would result in the I B RBCU being NOT Operable in accordance with Limits and Precautions of OP/i/NI 104/015 (Reactor Building Cooling System).

A. Switch in OFF B. 1XS2 de-energized C. RBCU Dampers misaligned D. LPSW inlet and outlet flow = 500 gpm Wednesday, August 31, 2011 Page 84 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 42 42 General Discussion Answer A Discussion Incorrect. Plausible since it is reasonable to believe that the RBCU would not start with the switch in the OFF position. Additionally plausible since the MDEFWPs will not start if EFDW is actuated with their swithces in the OFF position.

Answer B Discussion Incorrect. Plausible since it is logical to believe that IXS2 is in the power chain for the lB RBCU and that if 1XS2 were de-energized the RBCU could not perform its function and would therefore be inoperable. The unique power supply arrangement (see below) for the RBCUs also makes it plausible that the limits and precautions of the operating procedure would provide guidance on the impact of operability.

A RBCU Fan Motor 1 Supplied directly from X8 600V load center

2. Li Isolation breaker between X8 and fan motor is physically located in XS I MCC but is not connected electrically to XS 1.

3.0 Old supply was XS 1 but due to heavy electrical loads on XS 1, several incidents occurred where a fan motor fault tripped entire XS 1 MCC.

4. LiModilication electrically bypassed XS1 but left isolation breaker physically located there.

B RBCU Fan Motor

1. Supplied from XS3 600V MCC 2.LlDue to fewer loads on XS3, problem with A RBCU electrical supply was not a problem with B.

C RBCU Fan Motor iLl Supplied directly from X9 600V load center

2. LI Old supply was XS2, but due to same problems as with A RBCU, supply was moved to X9 with the isolation breaker physically remaining in XS2 cabinets.

Answer C Discussion Incorrect. Plausible since there are specific desired damper positions when the lB RBCU is operating that is based on which RBCUs are actually in service therefore it would be plausible to believe they would affect the RBCUs ability to distribute cool air as designed and therefore impact its operability.

Answer D Discussion Correct. L&P specifies that LPSW flow to each RBCU must be> 550 gpm Inlet Flow or >250 gpm Outlet Flow to meet flow requirements oi]

SLC 16.9.12.

Basis for meeting the KA Requires knowledge of the cause/effect relationship between RBCUs and LPSW.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj PNS-RBCRI,14,15 PNS-RBC SYSO22 K1.01 Containment Cooling System (CCS)

Knowledge of the physical connections and/or cause-effect relationships between the CCS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

SWS/cooling system 401-9 Comments; Rema rkslStatus Wednesday, August 31, 2011 Page 85 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 42 Wednesday, August 31, 2011 Page 86 of 208

FOR REVIEW ONLY DO* OT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 43 SYSO26 A3.O1 Containment Spray System (CSS)

Ability to monitor automatic operation of the CSS, including: (CFR: 41.7 / 45.5) 43 c Pump starts and correct MOV positioning Given the following Unit 1 conditions:

Reactor power = 100%

SBLOCA occurs

RB Pressure response as indicated below Reactor Builclmg Pressure us. Time (1) 0 0)

(0 U) 0)

U m

1200 1202 1204 1206 1208 1210 1212 1214 1216 Time Which ONE of the following states the EARLIEST time at which the Reactor Building Spray system will AUTOMATICALLY actuate?

A. 1203 B. 1204 C. 1210 D. 1215 Wednesday, August 31, 2011 Page 87 of 208

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2011B ONS SRO NRC Examination QU .EST.LON 43 43 General Dscussion Answer A Discussion Incorrect. Plausible since 3 psig is the actual setpoint for ES 1-6 on high RB pressure.

Answer B Discussion Incorrect. Plausible since 4 psig is the Tech Spec setpoint for ES 1-6 on high RB pressure.

Answer C Discussion Correct. 10 psig is the actual ES actuation setpoint for ES channels 5&6 Answer D Discussion Incorrect. Plausible since 15 psig is the Tech Spec required setpoint for ES channels 5&6, Basis for meeting the KA Requires the ability to monitor the automatic actuation of RB spray. Implicit in the ability to monitor automatic operation of equipment would be knowing when the equipment should automatically operate.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided OBJ IC-ES Ri, R14 IC-ES SYSO26 A3.0l Containment Spray System (CSS)

Ability to monitor automatic operation of the CSS, including: (CFR: 41.7/45.5)

Pump starts and correct MOV positioning 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 88 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QU ESTION 44 44 SYSO39 A 1.05 Main and Reheat Steam System (MRSS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: (CFR: 41.5 / 45.5)

RCST-ave Given the following Unit 3 conditions:

Reactor tripped from 100% power

A feedwater transient occurred

RCS temperature = 532°F slowly in creasing Which ONE of the following would result in stabilizing RCS temperature at approximately 532°F?

A. TBVs MUST be manually adjusted to maintain Turbine Header Pressure = 885 psig B. Turbine Header Pressure setpoint knob on the Turbine Master set to 885 psig C. Turbine Header Pressure setpoint knob on the Turbine Master set to 835 psig D. Turbine Header Pressure setpoint knob on the Turbine Master set to 760 psig Wednesday, August 31, 2011 Page 89 of 208

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2011B ONS SRO NRC Examination QUESTION 44 General Discussion Answer A Discussion Incorrect. Plausible for 2 reasons:

1. The setpoint knob that controls THP setpoint is located on the Turbine Master Bailey station. Since the Rx and Turbine have tripped, the Turbine Master has tripped to HAND and it would be plausible to believe that with the Turbine Master in HAND the setpoint knob would not be able to control TB Vs. That is because once the Turbine Master trips to HAND, the TB Vs begin using SG Outlet pressure as their controlling signal (but it does still compare SO Outlet pressure to THP setpoint for control.

2. The setpoint knob only goes as low as 600 psig therefore it would be plausible to believe that since RCS temperature is significantly below its normal value, the setpoint knob would not be able to be used to control at this low temperature due to being off scale low.

Answer B Discussion Incorrect. Plausible since this setpoint is the equivalent to 532 degrees however with the Rx tripped there is a 125 psig bias added to setpoint andi therefore requiring the setpoint knob to be set 125 psig below the desired header pressure.

Answer C Discussion Incorrect. Plausible since this setpoint would equate to having a 50 psig bias applied to the THP setpoint and that is correct when the Turbine Master in hand and the TBVs closed.

Answer D Discussion Correct, with the Rx tripped a 125 psig bias has been added to THP setpoint. Controlling THP at saturation pressure for RCS temperature is required to stabilize temp therefore considering the 125 psig bias, the setpoint would have to be 760 psig to control THP at 885 psig.

Basis for meeting the KA Requires predicting RCS temperature response to operation of main steam system controls.

Basis for Hi Cog Requires using steam tables and mathematical calculations to determine correct setpoint.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj STG-ICS RiO STG-ICS SYSO39 A 1.05 Main and Reheat Steam System (MRSS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: (CFR: 41.5 / 45.5)

RCS T-ave 401-9 Comments: Rema rkslStatus Wednesday, August 31, 2011 Page 90 of 208

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2011B ONS SRO NRC Examination QUESTION 45 45 SYSO59 K3.02 Main Feedwater (MFW) System Knowledge of the effect that a loss or malfunction of the MFW will have on the following: (CFR: 41.7 / 45.6)

AFW system Given the following Unit 3 conditions:

Initial conditions:

Time=1200

Reactor power = 100%

3B MDEFWP switch in AUTO 2

3A MDEFWP switch in AUTO 1 for testing Current conditions:

Time=1201

BOTH Main Feedwater pumps trip

3MS-87 (MS to TDEFDWP Control) fails closed Which ONE of the following describes ALL Emergency Feedwater Pumps operating at Time = 1202 assuming NO operator actions?

A. TDEFWP and 3A MDEFWP B. TDEFWP and 3B MDEFWP C. 3AMDEFWP ONLY D. 3BMDEFWPONLY Wednesday, August 31, 2011 Page 91 of 208

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2011B ONS SRO NRC Examination QUESTION 45 45 General Discussion Answer A Discussion Incorrect. TDEFWP is correct however in Auto 1, only dryout protection will start the MDEFWP and that occurs due to low SO level. 1 minute after a Rx trip, SO levels will still be well into the Operating Range.

Answer B Discussion Correct. The TDEFWP still has Aux Steam available and would therefore auto start and the MDEFWP in Auto 2 will also auto start when MFWPs are lost.

Answer C Discussion Incorrect. TDEFWP not starting is plausible since the Main Steam supply to the TDEFWP is not available due to the failure of 3MS-87 however Aux Steam is still available. In Auto 1, only dryout protection will start the MDEFWP and that occurs due to low SO level. 1 minute after a Rx trip, SO levels will still be well into the Operating Range.

Answer D Discussion Incorrect. TDEFWP not starting is plausible since the Main Steam supply to the TDEFWP is not available due to the failure of 3MS-87 however Aux Steam is still available. 3B MDEFWP would start.

Basis for meeting the KA Requires knowledge of the effect that a loss of both Main Feedwater pumps will have on Emergency Feedwater pumps.

Basis for Hi Cog Requires analyzing plant conditions and applying the analysis to the starting logic for Emergency Feedwater pumps.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj. CF-EF R12, R20, R22 SYSO59 K3.02 Main Feedwater (MFW) System Knowledge of the effect that a loss or malfunction of the MFW will have on the following: (CFR: 41.7/45.6)

AFW system 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 92 of 208

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2011B ONS SRO NRC Examination QUESTION 46 SYSO61 K6.02 Auxiliary / Emergency Feedwater (AFW) System Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: (CFR: 41.7/45.7)

Pumps Given the following Unit I conditions:

Initial conditions:

Reactor power = 100%

TDEFWP tagged out for repair

BOTH Main Feedwater pumps trip Current conditions

Rule 3 initiated

RCS pressure = 2310 psig decreasing

IA MDEFWP will not start Which ONE of the following describes actions required by the EOP to mitigate this event?

A. Initiate HPI Forced Cooling B. Initiate Condensate Booster Pump feed C. Align the I B MDEFWP to feed BOTH SGs D. Align alternate units EFDW to feed the IA SG Wednesday, August 31, 2011 Page 93 of 208

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2011B ONS SRO NRC Examination QUESTION 46 46 General Discussion Answer A Discussion Incorrect, Plausible since with RCS pressure> 2300, this would be correct if the lB MDEFWP had also failed. Plausibility comes from the fact that there has been a loss of a significant portion of the EFDW system and RCS pressure has exceeded 2300 psig.

Answer B Discussion Incorrect. Plausible since this would be correct if all EFDW had been lost and RCS pressure was <2300 psig. Plausibility comes from the fact that there has been a loss of a significant portion of the EFDW system and CBP feed would be the next desired source if all EFDW had been lost and RCS pressure were still below 2300 psig.

Answer C Discussion Correct. Initial steps in End. 5.9 (Extended EFDW Operations) will align the single MDEFWP to feed both SOs.

Answer D Discussion Incorrect. Plausible since using an alternate units EFDW to feed is one of the options if all EFDW is lost.

Basis for meeting the KA Per Chief Examiner, pumps lost should be EFDWPs to match KA. Requires knowledge of the effect of a loss of the TDEFWP and one MDEFWP has on the remaining MDEFWP Basis for Hi Cog Requires analyzing plant conditions and applying the analysis to ensure correct procedure direction.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj EAP-LORT R26 EAP-LOHT Rule 3 EOP End. 5.9 SYSO6I K6.02 Auxiliary / Emergency Feedwater (AFW) System Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: (CFR: 41.7/45.7)

Pumps 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 94 of 208

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2011B ONS SRO NRC Examination QUESTION 47 47 SYSO62 A1.O1 AC Electrical Distribution System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR: 41.5/45.5)

Significance of D/G load limits Given the following plant conditions:

Loss of Offsite power

230KV Yellow Bus locked out

Breaker SKI failed OPEN

CT-4 supplying all 3 units MFBs Which ONE of the following is the HIGHER load allowed without exceeding load limits provided in AP/II (Recovery From Loss of Power) End 5.IA (CT-4 Overload Limits)?

REFERENCE PROVIDED A. IOMW,6MVAR B. 17MW,I3MVAR C. I9MW,II.5MVAR D. 20.5 MW, 8 MVAR Wednesday, August 31, 2011 Page 95 of 208

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2011B ONS SRO NRC Examination QUESTION 47 47 General Discussion Answer A Discussion Incorrect. Plausible since this data results in being closest to the Limit with NO Forced Cooling line and would be a correct choice if no forced cooling of the transformer oil were indicated.

Answer B Discussion Correct. This data results in being the closest to the SK breaker curve without exceeding it.

Answer C Discussion Incorrect. Plausible since this data would be correct if you assumed that the area outside the last curve was the acceptable operation portion.

Plausible since there is no clear label identifying it as an unacceptable region and therefore an understanding of the load limits is required to determine it unacceptable.

Answer 0 Discussion tncorrect. Plausible since this would be correct if the SK breaker had not failed.

Basis for meeting the KA Per Chief Examiner OK to ask about CT-4 load limits.

Basis for Hi Cog Requires ability to use a complex curve.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj EAP-APG R9 AP/Il End 5.IA AP/Il End 5.1A EAP-AP1 1 SYSO62 A1.O1 AC Electrical Distribution System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR: 41.5 / 45.5)

Significance of D/G load limits 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 96 of 208

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2011B ONS SRO NRC Examination QUESTION 48 SYSO62 K4JO AC Electrical Distribution System Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Uninterruptable ac power sources Unit 1 initial conditions:

Reactor power = 100%

I DCA Bus Voltage = 125 VDC

I DCB Bus Voltage = 126 VDC

2DCA Bus Voltage = 127 VDC

2DCB Bus Voltage = 127 VDC Current conditions:

I XSI incoming feeder breaker trips Based on the above conditions, which ONE of the following is correct regarding the DC power systems assuming no operator actions are taken?

A. I DCA wiN be powered from the standby charger B. I DCB loads will be powered from Battery I CB C. 1 KX lnverter will be powered from I DCB D. I DIC Inverter will be powered from I DCB Wednesday, August 31, 2011 Page 97 of 208

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2011B ONS SRO NRC Examination QU ESTION 48 48 General Discussion Re-arranged answers Answer A Discussion Incorrect: the standby charger will not automatically power 1DCA. Plausible because the standby charger can be manually aligned to supply 1DCA.

Answer B Discussion Incorrect: DCB is not powered from 1XS1. The Battery Charger will be supplying DCB and the battery. Plausible because if 1XS2 were de energized, B would be correct.

Answer C Discussion Correct: Upon a loss of IXS1, Battery Charger 1CA de-energizes. Battery 1CA automatically picks up DC bus DCA. Essential Inverters (KX, KI and KU) are powered from DCA or DCB (whichever has the higher potential). Vital DC Buses (DIA, DIB, DIC, DID) are powered from their unit or the alternate unit (whichever has the higher potential).

Answer D Discussion Incorrect: 1DIC is supplied from the alternate unit (higher potential). Plausible because 1DIC would be supplied from 1DCB if it had the higher potential.

Basis for meeting the KA Question requires knowledge of a design feature which provides uninterruptable AC power to the panelboard by way of maintaining DC input to KX inverter.

Basis for Hi Cog Requires analyzing system conditions and detailed knowledge of how the system works.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2007 NRC Q41 Development References Student References Provided Obj EL-DCD R3 EL-DCD SYS062 K4.10 AC Electrical Distribution System Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Uninterruptable ac power sources 4019 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 98 of 208

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2011B ONS SRO NRC Examination QUESTION 49 49j SYSO63 A4.03 DC Electrical Distribution System Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Battery discharge rate Given the following plant conditions:

Initial conditions:

All three units Reactor power = 100%

Current condtions:

All Units 4160v Main Feeder Busses are de-energized

Unit 1, 2, and 3 EOP Blackout tabs in progress Based on the above conditions, which ONE of the following describes the required status of Unit 2 Essential Inverters per the EOP Enclosure 5.38 (Restoration of Power) and why?

Unit 2s Essential Inverters...

A. are de-energized to prevent inverter damage.

B. are de-energized to extend available battery life.

C. remain energized to provide power to ES channels.

D. remain energized to provide control power to 41 60v.

Wednesday, August 31, 2011 Page 99 of 208

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General Discussion Re-arranged answers Answer A Discussion Incorrect: Incorrect but plausible in that inverters could be damaged due to high current as input voltages start to decrease.

Answer B Discussion Correct: Essential Inverters KI, KU, & KX DC input breakers are opened to extend battery life per direction given from the EOP SBO tab (End.

38 and tab Step 2.38)

Answer C Discussion Incorrect: Plausible if ES Channels (are vital loads from KVIA,B,C,D) are confused with essential loads (from KI, KU, KX); vital loads must be differentiated from essential loads Answer D Discussion Incorrect: Plausible if control power (ex. for breakers, switches, etc) are incorrectly assumed to be essential inverter loads. Additionally plausii since control power for 4160V breaker operation is maintained during a blackout.

Basis for meeting the KA Requires knowledge of required actions within procedures and the ability to monitor battery discharge rate as a function of loads remaining on the battery. Additionally, de-energizing the inverters is in fact operating the battery discharge rate.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level

] QuestionType Question Source RO Memory BANK NRC 2009 Q48 Development References Student References Provided EAP-BO R8 EAP-BO EOP End 5.38 and 5.32 SYSO63 A4.03 DC Electrical Distribution System Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Battery discharge rate 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 100 of 208

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2011B ONS SRO NRC Examination QUESTION 50 5O SYSO64 A2.09 Emergency Diesel Generator (ED/G) System Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/ 43.5 /45.3 / 45.13)

Synchronization of the ED/G with other electric power supplies Given the following plant conditions:

Switchyard isolation has occurred

AP/1 I (Recovery from Loss of Power) in progress

Power has been restored to Unit I &2s Main Feeder Buses from Keowee Unit I via the overhead power path

MANUAL synchronization of KHU to the grid is in progress The PCB-8 Sync Check pushbutton (I)_ prevent out of phase closure of PCB-8 and the _(2)_ will be used to adjust Line side Potential to match Bus side Potential.

Which ONE of the following completes the statement above?

A. 1.will

2. Speed Changer Motor B. 1. will

2. Auto Voltage Adjuster C. I. will NOT

2. Speed Changer Motor D. I. will NOT

2. Auto Voltage Adjuster Wednesday, August 31, 2011 Page 101 of 208

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2011B ONS SRO NRC Examination QUESTION 50 so General Discussion Answer A Discussion Incorrect. First part is plausible since the Sync Check pushbutton is held depressed when syncing across PCB-8 to enable the syncrosocope. It is therefore reasonable to assume that due to the major expense that would be incurred if out of sync closure occurred, the sync check provided protection against that event. Second part is plausible since the Speed Changer Motor is used to adjust sync scope direction and speed in preparation for syncing to grid.

Answer B Discussion Incorrect. First part is plausible since the Sync Check pushbutton is held depressed when syncing across PCB-8 to enable the syncrosocope. It is therefore reasonable to assume that due to the major expense that would be incurred if out of sync closure occurred, the sync check provided protection against that event. Second part is correct.

Answer C Discussion Incorrect. First part is correct. Second part is plausible since the Speed Changer Motor is used to adjust sync scope direction and speed in preparation for syncing to grid Answer D Discussion Correct. The PCB-8 Sync Check pushbutton is depressed to enable the syncroscope. The Sync Check provides indication only and does not afford protection against out of phase closure of PCB-8. The Auto Voltage Adjuster is used to match Line side and Bus side voltages prior to closing PCB-8, Basis for meeting the KA Requires the ability to predict operation of the KI-IIJ sync curcuits and controls while syncing the KHU to another electrical power supply.

Basis for Hi Cog Requires detailed knowledge of systems and procedure to integrate the two in performing synchronization activities.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EAP-APG R9 AP/1 1 End 5.3 SYSO64 A2.09 Emergency Diesel Generator (ED!G) System Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Synchronization of the ED/G with other electric power supplies 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 102 of 208

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2011B ONS SRO NRC Examination QUESTION 51 5l SYSO73 K4.O1 Process Radiation Monitoring (PRM) System Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Release termination when radiation exceeds setpoint

_(1 )_ will automatically terminate a GWD tank release if its setpoint is reached.

_(2)_ is used to purge I RIA-37 and/or I RIA-38 once the GWD tank release is completed.

Which ONE of the following completes the statements above?

A. 1. ONLY IRIA-37

2. Instrument Air B. 1. ONLY IRIA-37

2. Nitrogen C. 1. 1 RIA-37 and I RIA-38

2. Instrument Air D. I. I RIA-37 and I RIA-38

2. Nitrogen Wednesday, August 31, 2011 Page 103 of 208

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2011B ONS SRO NRC Examination QUESTION 51 General Discussion Answer A Discussion Incorrect. First past is plausible since both RIA 37 and 38 are monitoring GWD tank releases and other process RIAs that monitor common things only actuated the associated interlocks from one of the RIAs. For example, both RIA 48 and 49 monitor RB environment however only RIA-49 performs the automatic interlock function. Second part is correct.

Answer B Discussion Incorrect. First part is plausible since both RIA 37 and 38 are monitoring GWD tank releases and other process RIAs that monitor common things only actuated the associated interlocks from one of the RIAs. For example, both RIA 48 and 49 monitor RB environment however only RIA-49 performs the automatic interlock function. Second part is plausible since Nitrogen is an inert gas that is commonly used as part of insturment calibrations and is a gas that is commonly used at ONS.

Answer C Discussion Correct. Both RIA 37 and RIA 38 will perform the actions of isolating a GWD tank release if its setpoint is reached and once the release is completed, Instrument Air is used to purge the remaining waste gas out of the lines associated with the RIA.

Answer D Discussion incorrect. First part is correct. Second part is plausible since Nitrogen is an inert gas that is commonly used as part of insturment calibrations and isagas that is commonly used at ONS.

Basis for meeting the KA Requires knowledge of design features (RIA interlocks) that provide for automatic termination of a waste gas tank release if their setpoints are exceeded Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj. RAD-RIA R2, R3 SYSO73 K4.O1 Process Radiation Monitoring (PRM) System Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Release termination when radiation exceeds setpoint 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 104 of 208

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2011B ONS SRO NRC Examination QUESTION 52 52 SYSO76 A4.02 Service Water System (SWS)

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

SWS valves Given the following Unit I conditions:

Main Steam Line break has occurred

Reactor Building pressure = 6 psig increasing

Restoration of LPSW to RCP motors is required Which ONE of the following describes the MINIMUM steps required to open BOTH ILPSW-6 and ILPSW-15?

A. Select MANUAL for EITHER Channel 5 or Channel 6 using pushbutton on I UB2 then operate valves using switches on IVB2 under the RZ modules B. Select MANUAL for BOTH Channels 5 AND Channel 6 using pushbuttons on I UB2 then operate valves using switches on IVB2 under the RZ modules C. Select MANUAL for EITHER channel 5 or Channel 6 using pushbutton on 1 UB2 then operate valves using switches on I UB2 D. Select MANUAL for BOTH Channels 5 AND Channel 6 using pushbuttons on I UB2 then operate valves using switches on I UB2 Wednesday, August 31, 2011 Page 105 of 208

FOR REVIEW ONLY- DO NOT IMSTRIBUTE 2011B ONS SRO NRC Examination QUESTION 52 52 D General Discussion This is testing the new Digital RPS/ES which has just been installed during the most recent Unit I outage.

Answer A Discussion Incorrect. The first part is plausible since both valves are on both Channel 5 and 6. Second part is plausible since with the installation of the new Digital ES/RPS, Switches for many of the containment isolation valves have been added under the RZ modules.

Answer B Discussion Incorrect. Although both channels must be taken to manual, the switches are located on UB2. Second part is plausible since with the installation of the new Digital ESIRPS, Switches for many of the containment isolation valves have been added under the RZ modules.

Answer C Discussion Incorrect. The first part is plausible since both valves are on both Channel 5 and 6. Second part is correct, Answer D Discussion Correct. Both valves are on both ES channel 5 and 6 therefore both digital channels must be taken to manual. Switches for the valves are now located on UB2.

Basis for meeting the KA Requires the ability to manually operate SWS valves following an ES actuation on high RB pressure.

Basis for Hi Cog Requires understanding of logic requirements regarding manual control of ES components and knowledge of new location for control switches.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RU Comprehension NEW Development References Student References Provided Ubj IC-ES R16 IC-ES SYSO76 A4.02 Service Water System (SWS)

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 /45.5 to 45.8)

SWS valves 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 106 of 208

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2011B ONS SRO NRC Examination QUESTION 53 53 SYSO76 K1.20 Service Water System (SWS)

Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: (CFR: 41.2 to 41.9 I 45.7 to 45.8)

AFW Which ONE of the following is the cooling medium for the Motor Driven Emergency Feedwater pump motors?

A. ROW B. COW C. HPSW D. LPSW Wednesday, August 31, 2011 Page 107 of 208

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2011B ONS SRO NRC Examination QUESTION 53 53 General Discussion Answer A Discussion Incorrect. Plausible since RCW does provide cooling to various secondary components (ex. HDPs).

Answer B Discussion Incorrect. Plausible since CCW does provide cooling to various secondary components (ex. TDEFWP).

Answer C Discussion Incorrect. Plausible since RCW does provide cooling to various secondary components (ex. Primary IA compressor).

Answer D Discussion Correct. LPSW provides cooling to the MDEFWP motors.

Basis for meeting the KA Requires knowledge of the physical connection between EFDW and LPSW.

Basis for Hi Cog Basis for SRO only 1

Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj CF-EF R6 CF-EF SYSO76 K 1.20 Service Water System (SWS)

Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: (CFR: 41.2 to 41.9 /

45.7 to 45.8)

AFW 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 108 of 208

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2011B ONS SRO NRC Examination QUESTION 54 54 SYS078 2.1.32 Instrument Air System (lAS)

SYSO78 GENERIC Ability to explain and apply system limits and precautions. (CFR: 41.10 /43.2 / 45.12)

Which ONE of the following will result in the automatic shutdown of a running Backup Instrument Air Compressor in accordance with Limits and Precautions of OP/I/NI 106/027 (Compressed Air System)?

A. Compressor in STD-BY 1 and IA pressure reaches 93 psig increasing B. Compressor in BASE and IA pressure reaches 100 psig increasing C. Anytime compressor is running and its discharge air temperature exceeds 425°F D. Anytime compressor is running and delta T of inlet vs. outlet air temperature exceeds 100°F Wednesday, August 31, 2011 Page 109 of 208

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2011B ONS SRO NRC Examination QUESTION 54 54 General Discussion Answer A Discussion Incorrect. Plausible since this is the starting pressure for compressors in Standby 1..

Answer B Discussion Incorrect. Plausible since this is the pressure at which the compressor unloads but it does not trip.

Answer C Discussion Correct. A temperature switch in each compressor discharge line closes if discharge air temperature exceeds 425°F which shuts off the compressor regardless of control switch position.

Answer D Discussion Incorrect. Plausible since this is guidance provided in the Limits and Preacautions of the Compressed Air procedure (1106/27) however it is regarding the Sullair Service Air compressors.

Basis for meeting the KA Requires the ability to apply the limit on compressor discharge line temperature and explain the consequences of exceeding the limit.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj SSS-IA R8, R9 SSS-IA OP-I 106-027 L&Ps SYSO78 2.1.32 Instrument Air System (lAS)

SYSO78 GENERIC Ability to explain and apply system limits and precautions. (CFR: 41.10/43.2/45.12) 401-9 Comments: RemarkslStatus

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Wednesday, August 31, 2011 Page 110 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 55 55j SYS 103 A 1.01 Containment System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including: (CFR: 415 / 455)

Containment pressure, temperature, and humidity Which ONE of the following describes the MINIMUM combination of Reactor Building Spray and Reactor Building Cooling trains required to maintain Reactor Building temperature and pressure within design limits following a LOCA from 100% power?

A. One RBS train and One RBCU B. One RBS train and Two RBCUs C. Two RBS trains and Two RBCUs D. Two RBS trains and Three RBCUs Wednesday, August 31, 2011 Page 111 of 208

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2011B ONS SRO NRC Examination QUESTION 55 55J General Discussion Answer A Discussion Incorrect. Plausible since it is the norm for many safety systems to need two trains to perform it safety function.

Answer B Discussion Correct. Three cooling trains must be available to cool the RB atmosphere following an accident. (e.g. 1 RBS Train and 2 RBCU5).

Assuming single failure criteria on both the RBS and the RBC systems, one RBS train and two RBCUs would be available in an accident.

During an accident, a minimum of two RBCUs and one RBS train are required to maintain containment pressure and temperature following a LOCA. Additionally, the one RBS train is also required to remove iodine from containment atmosphere and maintain concentrations below those assumed in the safety analysis.

Answer C Discussion Incorrect. Plausible since it is common for Tech Specs to require one more train of a system than is required for safety function and since both of these systems are contained in the same Tech spec it would be Plausible to believe that 4 of the 5 available trains were required to perform the safety function.

Answer D Discussion Incorrect. Plausible since this is the miniumum requirments to meet the Tech Spec LCO requirments when in MODE 1. Plausibility is also added since there are other systems at ONS where Tech Specs only requires the minimum required to perform safety functions.

Basis for meeting the KA Requires ability to predict the response of containment temperature and pressure based on various RBS and RBCU combinations. More specifically it requires predicting how Containment temperature and pressure will respond to a LOCA based on operability of containment systems.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj PNS-BS R16 PNS-BS SYS1O3 A1.O1 Containment System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including: (CFR: 41.5 / 45.5)

Containment pressure, temperature, and humidity 401-9 Comments: RemarkslStatus 1

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2011B ONS SRO NRC Examination QUESTION 56 56j SYSOOI K6.02 Control Rod Drive System Knowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 41.7/45.7)

Purpose and operation of sensors feeding into the CRDS Given the following Unit I conditions:

Reactor power = 100%

Control Rod Relative Position Indication (RPI) NOT available Which ONE of the following is NOT available?

A. Group In Limit and Out Limit indications B. Rod Misalignment Correction C. RPS Trip Confirm signal D. Asymmetric Rod alarm Wednesday, August 31, 2011 Page 113 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 56 56 B General Discussion Answer A Discussion Incorrect. Plausible since initiating the Group In or Out Limit indication would require knowing the Control Rod positions and it is plausible to believe that this relies on RPI however it has its own separate set of switches to make the determination.

Answer B Discussion Correct. Rod Misalignment Correction causes withdraw commands to result in some selected rods in a group not moving to achieve alignment within 1%. For example if Group 7, Rod 1 is low by 2%, it will move alone for the first 1% of withdrawal until it is within 1% of the Group Average, then all Group 7 rods will move out together. If Group 7 Rod 1 is high by 2%, then rods 2-8 will move outward for 1% of travel and then rod I will join in outward movement once it is within 1% of Group Average. The logic uses RPI indication (so a stuck rod has no effect on commands).

Answer C Discussion Incorrect. Plausible since it is reasonable to believe that a trip confirmed signal would be a result of verifying all control rods have inserted and therefore it is plausible to believe that losing a portion of control rod position circuitry could result in losing the ability to determine if a Trip Confirm signal should be sent to ICS.

Answer D Discussion Incorrect. Plausible since the Asymmetric Rod alarm requires knowing the position of the associated control rods and therefore it is plausible believe that losing the RPI indication would result in the Asymmetric rod alarm being not available.

Basis for meeting the KA Requires knowledge of the effect of a loss of control rod position sensors has on the Control Rod Drive system.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj IC-CRI R29,30 IC-CRI SYSOO1 K6.02 Control Rod Drive System Knowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 41.7/45.7)

Purpose and operation of sensors feeding into the CRDS 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 114 of 208

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2011B ONS SRO NRC Examination QUESTION 57 57 SYSOO2 K4.07 Reactor Coolant System (RCS)

Knowledge of RCS design feature(s) and!or interlock(s) which provide for the following (CFR: 41.7)

Contraction and expansion during heatup and cooldown Given the following Unit I conditions:

Reactor in MODE 3

RCS heatup in progress

1HP-14 (LDST Bypass) in Bleed

1 HP-i 6 is closed

1) The HIGHER LDST level (inches) that will automatically return I HP-i4 to normal is _(i)_

2) While 1HP-i4 is in the Bleed position there (2)_ flow through the Letdown Filters.

Which ONE of the following completes the statements above?

A. 1. <40

2. is B. 1. <40

2. is NOT C. 1. <55

2. is D. 1. <55

2. is NOT Wednesday, August31, 2011 Page 115 of 208

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2011B ONS SRO NRC Examination QUESTION 57 57j General Discussion Answer A Discussion Incorrect. First part is correct. Second part is plausible since letdown flow normally goes through the Letdown Filters in route to the LDST however with 1HP-14 in BLEED, the filters are bypassed and letdown is routed to the A BHUT.

Answer B Discussion Correct. LDST at <40 inches will return HP-14 to the NORMAL position if it is in the BLEED position. Letdown flow normally goes through the Letdown Filters in route to the LDST however with 1HP-14 in BLEED, the filters are bypassed and letdown is routed to the 1A BHUT.

Answer C Discussion Incorrect. First part is plausible since 55 is the Lo Lo LDST level alarm setpoint. Second part is plausible since letdown flow normally goes through the Letdown Filters in route to the LDST however with lI-IP-14 in BLEED, the filters are bypassed and letdown is routed to the A BHUT.

Answer D Discussion Incorrect. First part is plausible since 55 is the Lo Lo LDST level alarm setpoint. Second part is correct.

Basis for meeting the KA Requires knowledge of an interlock which provides for the expansion of RCS during heatup.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj PNS-HPI R8 PNS-HPI HPI Bleed drawing SYSOO2 K4.07 Reactor Coolant System (RCS)

Knowledge of RCS design feature(s) andJor interlock(s) which provide for the following: (CFR: 41.7)

Contraction and expansion during heatup and cooldown 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 116 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 58 58 SYSOI5 A1.07 Nuclear Instrumentation System (NIS)

Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the NIS controls including: (CFR: 41.5 45.5)

Changes in boron concentration Given the following Unit I conditions:

Time = 1200

Reactor power = 50% stable

Control Rod Group 7 = 95% withdrawn

Inadvertent RCS de-boration is in progress Time= 1215

Group 7 = 85% withdrawn Time = 1230

Group 7 = 30% withdrawn Time = 1245

Group 6 = 95% withdrawn Time = 1300

Group 6 = 35% withdrawn Which ONE of the following is the EARLIEST time that the minimum Shutdown Margin is NOT available?

REFERENCE PROVIDED A. 1215 B. 1230 C. 1245 D. 1300 Wednesday, August 31, 2011 Page 117 of 208

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2011B ONS SRO NRC Examination QUESTION 58 58 General Discussion Answer A Discussion Incorrect. Plausible since this is the first point that indicates being outside the Steady State Operating Band which is at 87% on Gp 7 however this is not the minimum SDM curve.

Answer B Discussion Incorrect. Plausible since this is the correct answer if using the SDM curve for I inoperable control rod however since there is no indication of an inoperable control rod this is the wrong curve.

Answer C Discussion Incorrect. Plausble since this point indicates entering the Restricted region of the rod curve however this does not mean that the minimum SDM has been lost, this indicates that rods are not within limits assumed in the Safety Analysis for plant operations.

Answer D Discussion Correct. Using the 4 pumps no inoperable rod curve, this is the first point that has entered the Unacceptable Operation region of the curve.

Basis for meeting the KA J

Per Chief Examiner, systems/components responding to neutron error will match KA. Requires the ability to predict when NIS controls (control rods) will exceed design Shutdown Margin limits due to an inadvertent deboration.

Basis for Hi Cog The second part of this question requires more than one mental step to determine the direction of rod travel.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided SNO-LSU R33 Rod curves COLR rod curves SYSOI5 A1.07 Nuclear Instrumentation System (NIS)

Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the NIS controls including: (CFR: 41.5 45.5)

Changes in boron concentration 401-9 Comments: Remarks!Status Wednesday, August 31, 2011 Page 118 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 59 59 SYSO6 K3.06 Non-Nuclear Instrumentation System (NNIS)

Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: (CFR: 41.7 / 45.6)

AFW system Given the following Unit 3 conditions:

Reactor power= 100%

BOTH of the 3A Steam Generator Operating Range levels fail HIGH Which ONE of the following describes consequences of the level instrument failures?

A. Main Feedwater controls both SG levels at 25 SU range B. 3A Main Feedwater pump trips, 3A MDEFWP controls 3A SG at 30 XSUR, and 3B Main Feedwater pump controls 3B SG level at 25X8tJR SO C. BOTH Main Feedwater pumps trip, ONLY the MDEFWPs start, and control both SC levels at 30 XSUR D. BOTH Main Feedwater pumps trip, ALL EFDW pumps start, and control both SC levels at 30 XSUR Wednesday, August 31, 2011 Page 119 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QU.ESrF EON 59 59 General Discussion Answer A Discussion Incorrect. Plausible since the level given would be correct following a Rx trip with Main Feedwater available. There is a high SO level trip of the Main Turbine which would occur and therefore make it plausible to either not recognize that the hi SG level indications would trip BOTH main FDW pumps or believe it took hi level indication on both SOs to trip the main Feedwater pumps. Also plausible to believe that the 3A SG would trip only the 3A MFDWP and the 3B MFDWP would control SG levels. Believing that the MFDWPs are train specific following a Rx trip is plausible since the MS side of the SOs do become train specific following a Rx trip by way of the MSSVs closing.

Answer B Discussion Incorrect. Plausible since the High SG level would trip the turbine and therefore the Rx. In that case it would be reasonable to believe that the 3A MFDWP feeds the 3A SO and the 3B MFDWP feeds the 3B SO since the MS line side of the SGs are in fact separated into trains following a trip by way of the MSSVs closing. In that context, it is plausible to believe that the 3A MFDWP would trip if the 3A SO had hi level indications. That would lead to the conclusion that the 3A MDEFWP would start and the level given is the correct level used by EFDW when controlling SO levels.

Answer C Discussion Incorrect. Plausible since SO dryout protection will only actuate the MDEFWPs so if the candidate had the misconception that the hi SG level indications caused only the MDEFWPs to start following the trip of the MFDW pumps it would lead to this choice. The level given is correct for EFDW.

Answer D Discussion Correct. The Hi SO level trip is a 2/2 circuit that trips BOTH Main Feedwater pumps on hi level in either SO. When both MFDWPs trip, all available EFDWPs will start and EFDW will control SO levels at 30 XSUR.

Basis for meeting the KA Requires knowledge of the effect that a malfunction of SO level instrumentation will have on the Emergency Feedwater system.

Basis for Hi Cog Requires analyzing plant conditions and failures and applying that knowledge to determine the impact of the failures on a different system.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj CF-FPT R8 CF-EF R20, R22 CF-FPT CF-EF SYSO16 K3.06 Non-Nuclear Instrumentation System (NNIS)

Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: (CFR: 41.7 / 45.6)

AFW system 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 120 of 208

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2011B ONS SRO NRC Examination QUESTION 60 SYSO29 A4.04 Containment Purge System (CPS)

Ability to manually operate and/or monitor in the control room: (CFR: 41.7/45.5 to 45.8)

Containment evacuation signal Given the following Unit 1 conditions:

Reactor in MODE 5

Reactor Building Purge in progress Which ONE of the following is the most complete list of interlocks that will automatically occur as a result of a I RIA-45 (Normal Vent Gas) HIGH Alarm?

A. Trip the Main Purge Fan ONLY B. Trip the Main Purge Fan AND close I LWD-2 ONLY C. Trip the Main Purge Fan AND activate the RB Evacuation Alarm ONLY D. Trip the Main Purge Fan, close ILWD-2, AND activate the RB Evacuation Alarm Wednesday, August31, 2011 Page 121 of 208

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2011B ONS SRO NRC Examination QUESTION 60 General Discussion Answer A Discussion Correct. 1RIA-45 HIGH alarm will trip the Main Purge Fan if it is running.

Answer B Discussion Incorrect. Closing 1LWD-2 is plausible since the RIA-45 alarm would be indiciative of issues inside the Reactor Building and isolating the RB normal sump is a logical step in containing the activity. Additionally, it would be correct for RIA-49.

Answer C Discussion Incorrect. Plausible since activating the RB Evacuation Alarm is a logical step to take when activity increasing in the RB to the point of causing an RIA alarm. Additionally, this would be correct for RIA-49 which monitors RB environment.

Answer D Discussion Incorrect. Closing LWD-2 and initiating the RB evacuation alarm are both plausible as described in B and C answer explanations.

Basis for meeting the KA Requires the ability to determine if a Containment evacuation signal will be automatically generated during a RB Purge that causes an RIA-45 HIGH alarm. The ability to determine if the RB evacuation alarm should have alarmed demonstrates the ability to monitor Containment Evacuation signal from the control room.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj RAD-RIA R2,16, RAD-RIA SYSO29 A4.04 Containment Purge System (CPS)

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Containment evacuation signal 401-9 Comments Remarks/Status Wednesday, August 31, 2011 Page 122 of 208

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2011B ONS SRO NRC Examination QUESTION 61 6l SYSO33 2.1.20 Spent Fuel Pool Cooling System (SFPCS)

SYSO33 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

Given the following plant conditions:

Unit I BWST in purification

Unit I &2 Spent Fuel Pool level -.3 feet slowly decreasing

AP/1-2/N1700/035 (Loss of SEP Cooling and/or Level) initiated Which ONE of the following states:

1) which units BWST can be used for SEP makeup in accordance with AP/35?

2) the HIGHEST SEP level (feet) at which the SF Cooling pumps will AUTOMATICALLY trip?

A. 1. Unit 1

2. -2.5 B. 1. Unit I

2. -4.0 C. 1. Unit2

2. -2.5 D. 1. Unit2

2. -4.0 Wednesday, August 31, 2011 Page 123 of 208

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2011B ONS SRO NRC Examination QUESTION 61 General Discussion Answer A Discussion CORRECT. According to the NOTE associatd with step 4.13 of AP/35, to use either Unit 1 or Unit 2 BWST as a makeup source the associated BWST must already be aligned in its purification alignment. The SF Cooling pump auto trip on low SFP level setpoint is -2.5 feet.

Answer B Discussion Incorrect. First part is correct. Second part is plausible since this is the level of the pump suction lines below the normal pooi level.

Answer C Discussion Incorrect. First part is plausible as it would be reasonable to assume that the BWST must NOT be aligned in another flowpath to be used as a makup source since you could believe that the intent of the guidance is to save the time it would take to re-align the BWST for use as a makup source. Second part is correct.

Answer D Discussion Incorrect. First part is plausible as it would be reasonable to assume that the BWST must NOT be aligned in another flowpath to be used as a makup source since you could believe that the intent of the guidance is to save the time it would take to re-align the BWST for use as a makup source. Second part is plausible since this is the level of the pump suction lines below the normal pool level.

Basis for meeting the KA Requires ability to determine which BWST should be used as it applies to makeup to the SFP during a loss of SFP level in accordance with procedural guidance provided in AP/35 and therefore requires the ability to interpret and execute the associated AP step.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj FH-SFC R22 FH-SFC SYSO33 21.20 Spent Fuel Pool Cooling System (SFPCS)

SYSO33 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10/43.5 / 45.12) 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 124 of 208

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2011B ONS SRO NRC Examination QUESTION 62 62 SYSO56 A2.04 Condensate System Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; 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)

Loss of condensate pumps Given the following Unit I conditions:

Reactor power = 80% stable

IA and lB CBP operating Current conditions:

Time=1200:00

iACBPtrips

Feedwater Pump suction pressure = 225 psig slowly decreasing Which ONE of the following describes the:

1) runback rate (%/min) inserted by ICS?

2) procedure that will be directed by the Procedure Director at Time = 1202:00 assuming Feedwater Pump suction remains approximately 220 psig?

A. 1.15

2. AP/i IA/i 700/001 (U nit Run back)

B. 1.15 2.EOP C. 1.20

2. AP/1 IA/i 700/001 (Unit Run back)

D. 1.20

2. EOP Wednesday, August 31, 2011 Page 125 of 208

FOR REVIEW ONLY -DO NOT IMSTRIBUTE 2011B ONS SRO NRC Examination QUESTION 62 62 General Discussion Answer A Discussion Incorrect. First part is plausible since there are ICS runbacks that incorporate the 15%/mm runback rate. Second part is plausible since it would be correct for the first 90 seconds of the transient.

Answer B Discussion Incorrect. First part is plausible since there are ICS runbacks that incorporate the 15%/mm runback rate. Second part is correct, Answer C Discussion Incorrect. First part is correct. Second part is plausible since it would be correct for the first 90 seconds of the transient.

Answer D Discussion Correct. With FDWP suction pressure <235 psig, an ICS runback is initiated. The runback rate is 20%/mm to a power level of 15% or until the low suction pressure clears. After 90 seconds, if FDWP suction pressure is still <235 psig the FDWPs will trip which will trip the Rx and require entry into the EOP to mitigate the loss of main feedwater.

Basis for meeting the KA Requires knowledge of the impact of a loss of Condensate Booster Pump and knowledge of the procedure that will be used to mitigate the event.

Basis for Hi Cog Requires analyzing plant data to determine the Unit response and the procedure that will be used to mitigate the event.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj STG-ICS R3 EAP-SA R21, R24 EAP-SA STG-ICS Intro & Chptr 2 SYSO56 A2.04 Condensate System Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; 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)

Loss of condensate pumps 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 126 of 208

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201 lB ONS SRO NRC Examination QUESTION 63 63 SYSO71 K5.04 Waste Gas Disposal System (WGDS)

Knowledge of the operational implication of the following concepts as they apply to the Waste Gas Disposal System: (CFR: 41.5 / 45.7)

Relationship of hydrogen!oxygen concentrations to flammability Which ONE of the following:

1) is the LOWER concentration of Hydrogen that would require actions to an isolated GWD tank in accordance with Limits and Precautions of OPII-21A/1104/18; Gaseous Waste Disposal System

2) describes the actions required lithe lower limit on Hydrogen is exceeded?

A. 1. 3.1%

2. Dilute the tank contents with Nitrogen B. 1. 4.1%

2. Dilute the tank contents with Nitrogen C. 1. 3.1%

2. Dilute the tank contents with 002 D. 1. 4.1%

2. Dilute the tank contents with C02 Wednesday, August 31, 2011 Page 127 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 63 63j A General Discussion Answer A Discussion Correct. If H2 concentration exceeds 3% then 1104/18 will be used to dilute the tank with N2 to reduce the H2 concentration.

Answer B Discussion Incorrect. First part is plausible since 4% is the actual lower flammability limit of H2 in air and is a recognizable threshold value from SLC 16.11.4. Second part is correct.

Answer C Discussion Incorrect. First part is correct. Second part is plausible since C02 is used to dilute the H2 concentration in the Main Electrical Generator and one of the main reasons it is used is to prevent H2 and 02 concentrations from reaching the flammability limit. Since C02 is used in another system for basically the same reason it is a plausible distracter.

Answer D Discussion Incorrect: First part is plausible since 4% is the actual lower flammability limit of H2 in air and is a recognizable threshold value from SLC 16.11.4. Second part is plausible since C02 is used to dilute the H2 concentration in the Main Electrical Generator and one of the main reasons it is used is to prevent H2 and 02 concentrations from reaching the flammability limit. Since C02 is used in another system for basically the same reason it is a plausible distracter.

Basis for meeting the KA Requires knowledge of the operational implication of H2 approaching flammability limit in a GWD tank.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory : NEW Development References Student References Provided Obj WE-GWD R12, 13, 14, 16)

WE-GWD SYSO71 K5.04 Waste Gas Disposal System (WGDS)

Knowledge of the operational implication of the following concepts as they apply to the Waste Gas Disposal System: (CFR: 41.5 / 45.7)

Relationship of hydrogen/oxygen concentrations to flammability 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 128 of 208

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2011B ONS SRO NRC Examination QUESTION 64 64 SYSO86 A3.O1 Fire Protection System (FPS)

Ability to monitor automatic operation of the Fire Protection System including: (CFR: 41.7/45.5)

Starting mechanisms of fire water pumps Which ONE of the following is the EARLIEST time that a start signal will be generated for an HPSW pump whose switch is in BASE?

EWST Level vs. Time c-n 0

cc 0) 0)

-J H

w 40000 1200 1210 1220 1230 Time A. 1225 B. 1220 C. 1210 D. 1205 Wednesday, August 31, 2011 Page 129 of 208

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2011B ONS SRO NRC Examination QUESTION 64 64 General Discussion Answer A Discussion Incorrect, Plausible since 50,000 gal would be correct for a pump in Standby Answer B Discussion Correct. 60,000 gal EWST level is the setpoint for starting an HPSW pump in BASE. Time 1220 corresponds to reaching 60,000 gal.

Answer C Discussion

=

]

Incorrect. Plausible since 80,000 gal is the setpoint for automatically stopping a running HPSW pump.

Answer D Discussion Incorrect. Plausible since 90,000 is the setpoint for Normal Level indication (No amber or white light lit).

Basis for meeting the KA Requires knowledge of what will auto start an HPSW pump. The ability to determine if the HPSW pump should have started is integral to the ability to monitor automatic operation of the pump.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj. SSS-HPW R5 55 S-HP W SYSO86 A3.01 Fire Protection System (FPS)

Ability to monitor automatic operation of the Fire Protection System including: (CFR: 41.7 / 45.5)

Starting mechanisms of fire water pumps 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 130 of 208

FOR REVIEW ONLY -DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination SYSO79 K1.O1 Station Air System (SAS)

QUE STI ON 65 65j D Knowledge of the physical connections and/or cause-effect relationships between the SAS and the following systems: (CFR: 41.2 to 41.9/45.7 to 458) lAS IA PRESSURE 101 uJ cr D

C)

C) w U

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

A. 1207 B. 1210 C. 1212 D. 1215 Wednesday, August 31, 2011 Page 131 of 208

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2011B ONS SRO NRC Examination QUESTION 65 65 General Discussion Answer A Discussion Incorrect: Plausible since 93 psig is the pressure at which the Backup IA compressors will start.

Answer B Discussion Incorrect: Plausible since 90 psig is the pressure at which the Diesel Air Compressors will start Answer C Discussion Incorrect: Plausible sine 88 psig is the pressure at which the AlA compressors will start Answer 0 Discussion CORRECT: SA to IA Controller (SA-l41) valve senses the IA system pressure and opens at 85 psig to allow service air into the IA system.

Basis for meeting the KA Requires knowledge of automatic cross-connect between Service air and Instrument air systems.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK NRC 2009A Q64 Development References Student References Provided Obj 555-IA R52 SYSO79 KI.01 Station Air System (SAS)

Knowledge of the physical connections and/or cause-effect relationships between the SAS and the following systems: (CFR: 41.2 to 41.9/45.7 to 45.8) lAS 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 132 of 208

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2011B ONS SRO NRC Examination QU EST[ON 66 66 GEN2.l 2.1.14 GENERIC Conduct of Operations Conduct of Operations Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc. (CFR: 41.10 /

43.5 /45.12)

Which ONE of the following lists of items will ALWAYS require a plant page (except for unanticipated automatic starts and emergency situations) in accordance with OMP 1-02 (Rules of Practice)?

A. Opening PCB-18 Closing PCB-18 B. ALL AP entries Starting I Bi RCP C. Starting and stopping a 41 60V Motor Closing PCB-18 D. Starting and stopping 1BI RCP Starting a 41 60V Motor Wednesday, August 31, 2011 Page 133 of 208

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2011B ONS SRO NRC Examination QUESTION 66 66 General Discussion Answer A Discussion 1

Correct. Per OMP 1-2 both opening and closing of PCBs requires a plant page.

Answer B Discussion Incorrect. First part is plausible since many APs require a plant page when entered however not all do. The page is specified within the AP if required. Second part is correct.

Answer C Discussion Incorrect. First part is plausible since starting a 41 60V motor does require a plant page. Second part is correct.

Answer D Discussion Incorrect. Starting RCPs requires a plant page but securing the RCPs do not. Second part is correct.

Basis for meeting the KA Requires knowledge of specific criteria that require plant wide pages.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED ILT39 Audit & NRC 2009B Q66 Development References Student References Provided Obj ADM-OMP R57,58 ADM-OMP OMP 1-02 GEN2.l 2.1.14 GENERIC Conduct of Operations Conduct of Operations Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc. (CFR: 41.10 I 43.5 /45.12) 401-9 Comments: Rema rkslStatus Wednesday, August 31, 2011 Page 134 of 208

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2011B ONS SRO NRC Examination QUESTION 67 L GEN2.1 2.1.27 GENERIC Conduct of Operations Conduct of Operations Knowledge of system purpose andior function. (CFR: 417)

Which ONE of the following describes the design purpose of the Main Feedwater and Main Turbine RPS Anticipatory trips?

A. Reducing the possibility of lifting the PORV B. Reduce the possibility of lifting the Pressurizer Code relief valves C. Serve as a backup to the RPS High RCS Pressure trip D. Serve as a backup to the RPS High RCS Temperature trip Wednesday, August 31, 2011 Page 135 of 208

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2011B ONS SRO NRC Examination QUESTIQN 67 67 General Discussion Answer A Discussion Correct, the RPS anticipatory trips function are to reduce the possibility of lifting the PORV.

Answer B Discussion Incorrect. Plausible since the Code relief valves on the pressurizer could be challenged by rapidly increasing RCS pressure and the anticipatory trips do increase the margin available to prevent the Code relief valves from lifting. Additionally plausible since the Code relief valves (and not the PORV) are credited to help prevent exceeding the RCS pressure Safety Limit therefore it would be plausible that the RPS trips were designed to help prevent getting to the Code RVs setpoint.

On the surface it may appear that there is a subset issue with A and B however that is not the case. If the Answer C Discussion Incorrect. Plausible since it would be an RPS High Pressure trip that would be most likely to trip the Rx in the event of a loss of Main FDW or Main Turbine if the two Anticipatory trips were not available.

Answer D Discussion Incorrect. Plausible since either of the events would result in rapidly increasing RCS temperature therefore it is plausible to believe that the anticipatory trips were a backup to the High Temperature trips.

Basis for meeting the KA Requires knowledge of the design purpose of RPS Anticipatory trips.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj IC-RPS R9 IC-RPS GEN2.1 2.1.27 GENERIC Conduct of Operations Conduct of Operations Knowledge of system purpose and/or function. (CFR: 41.7) 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 136 of 208

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2011B ONS SRO NRC Examination QUESTION 68 68 GEN2.l 2.1.5 GENERIC Conduct of Operations Conduct of Operations Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 / 43.5 / 45.12)

Given the following plant conditions:

Uniti Reactorpower=100%

Unit 2 Reactor in MODE 3 In accordance with OMP 2-1 Attachment D (SSF Staffing Requirements), which ONE of the following:

1) states restrictions on the RO designated to man the SSF if required?

2) describes the minimum actions required for one of the designated SSF ROs to take a short trip to the station canteen?

A. 1. MUST remain in the Horseshoe area of the Control Room

2. The RO must be relieved by another licensed operator that is NOT part of the minimum staffing before leaving the designated area B. 1. MUST remain in the Horseshoe area of the Control Room

2. A method of communication must be established to enable notification of the requirement to activate the SSF before leaving the designated area C. 1. Can be anywhere inside the Control ROOM CAD doors

2. The RO must be relieved by other licensed operator that is NOT part of the minimum staffing before leaving the designated area D. 1. Can be anywhere inside the Control ROOM CAD doors

2. A method of communication must be established to enable notification of the requirement to activate the SSF before leaving the designated area Wednesday, August 31, 2011 Page 137 of 208

FOR REVIEW ONLY-DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 68 68 General Discussion Answer A Discussion Incorrect. First part is plausible as there are specific requirements where ROs must remain in the horseshoe area as described in Att B of OMP 2-

1. Second part is plausible since it would be correct if both Units required the SSF to be operable and the Control Room were staffed with the minimum staffing under those conditions (4 ROs). Second part is plausible since it would be correct with minimum staffing if only one of the units were above MODE 4 and therefore only one unit required an SSF RO.

Answer B Discussion Incorrect. First part is plausible as there are specific requirements where ROs must remain in the horseshoe area as described in Att B of OMP 2-

1. Second part is correct, Answer C Discussion Incorrect. First part is correct. Second part is plausible since it would be correct if both Units required the SSF to be operable and the Control Room were staffed with the minimum staffing under those conditions (4 ROs). Second part is plausible since it would be correct with minimum staffing if only one of the units were above MODE 4 and therefore only one unit required an SSF RO.

Answer D Discussion Correct. Per ATT D of AOM 2-1 the SSF RO must be between the CAD doors however when both units require the SSF RO, it is acceptable for one of the two SSF ROs to leave the control room area for short periods of time as long as a method of communication is established to enable notification of the requirement to man the SSF.

Basis for meeting the KA Requires ability to use OMP 2-01 requirements regarding requirments placed on the RO designated as the SSF RO.

Basis for Hi Cog Requires analyzing the status of Units 1 and 2 and then applying the requirements of OMP 2-01 to that analysis.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj ADM-OMP R5 ADM-OMP OMP 2-01 Attach. B &D GEN2.1 2.1.5 GENERIC Conduct of Operations Conduct of Operations Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 / 43.5 / 45.12) 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 138 of 208

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2011B ONS SRO NRC Examination QUESTION 69 GEN2.2 2.2.12 GENERIC Equipment Control Equipment Control Knowledge of surveillance procedures. (CFR: 41.10 / 45.13)

Given the following Unit I conditions:

Reactor power = 100%

RO is performing PT/11A106001001 (Periodic Instrument Surveillance)

A Tech Spec required surveillance does not meet its acceptance criteria

There is NOT an outstanding Surveillance Evaluation for the affected surveillance Which ONE of the following describes ALL actions required (if any) to disposition the affected surveillance sign off block once Enclosure 13.12 (Surveillance Evaluation) has been initiated in accordance with PT/I /N0600/001?

A. Leave the block empty B. Put your initials in the block ONLY C. Initial the block AND docu ment that a Surveillance Evaluation is in effect D. Put an asterisk (*) in the block and explain at the bottom of the page that a Surveillance Evaluation has been initiated Wednesday, August 31, 2011 Page 139 of 208

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 69 69 General Discussion Answer A Discussion Incorrect. Plausible since the sign off block represents a step in the procedure and it is the normal operating practice to not sign off a step that has not been performed.

Answer B Discussion Incorrect. Plausible since this is partially correct. Additionally, you are allowed to initial steps whose intent has been completed and it is plausible to believe that initiating the Surveillance Evaluation would represent performing the intent of the step since it will evaluate the operability of the associated equipment.

Answer C Discussion Correct. Per instructions contained in the body of the enclosure of the procedure, once the Surveillance Evaluation has been initiated, record Surveillance Evaluation in effect and place your initials in the block.

Answer D Discussion Incorrect. Plausible since this would be partially correct however the procedure specifies putting your initials in the block as well.

Basis for meeting the KA Requires knowledge of PT/600/0 I (Periodic Instrument Surveillance) and how to disposition surveillances that are 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 T&Q 2620001 PT/600/01 End MODE l&2 GEN2.2 2.2.12 GENERIC Equipment Control Equipment Control Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 140 of 208

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2011B ONS SRO NRC Examination QUESTION 70 GEN2.2 2.2.13 GENERIC Equipment Control Equipment Control Knowledge of tagging and clearance procedures. (CFR: 41.10 /45.13)

Which ONE of the foHowing tags would be used for configuration control of I HP-409 in accordance with NSD-500 (Red Tags/Configuration Control Tags)?

A.

B. White Tag C. OORT tag D. CORT tag Wednesday, August 31, 2011 Page 141 of 208

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2011B ONS SRO NRC Examination QUESTION 70 70 General Discussion

Answer A Discussion Incorrect. Plausible since Red Tags are tags addressed in NSD 500 however they are used for personal safety.

Answer B Discussion Correct. White tags are used for configuration control of components and systems Answer C Discussion Incorrect. Plausible since OORT tags are tags used in the field during equipment maintenance and are addressed by NSD 500. OORT tags are used to re-assign operations control of a component that is owned by Chemistry to Operations. Since the component in question is owned by Operations, an OORT tag is plausible since it begins with an 0 (for Operations).

Answer D Discussion Incorrect. Plausible since CORT tags are tags that are use in the field during maintenance activities and are addressed by NSD 500. A CORT tag would be used to re-assign operational control of a component owned by operations to Chemistry. It is plausible to believe that a CORT tag is for configuration control since CORT tags are used on components where Operations is the Owner Control Group and Operations is the owner control group for HP-409.

Basis for meeting the KA Requires generic knowledge of the tagging process defined by NSD 500 Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj ADM-SD R6 NSD 500 GEN2.2 2.2.13 GENERIC Equipment Control Equipment Control Knowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13) 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 142 of 208

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2011B ONS SRO NRC Examination QUESTION 71 7l GEN2.2 2.2.36 GENERIC Equipment Control Equipment Control Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR; 41.10 /43.2/45.13)

Given the following Unit I conditions:

Reactor power = 100%

Which ONE of the following would result in a Tech Spec LCO being NOT met?

A. I B Core Flood Tank pressure = 579 psig due to leakage B. IA RBCU breaker racked out for breaker repair C. I C LPI pump isolated for maintenance D. I B HPI pump switch in OFF for testing Wednesday, August 31, 2011 Page 143 of 208

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2011B ONS SRO NRC Examination QUESTION 71 General Discussion Answer A Discussion Incorrect. Plausible since the low pressure alarm is at 585 psig however TS 3.5.lrequires pressure> 575 psig Answer B Discussion Correct, when above MODE 3, TS 3.6.5 requires all 3 RBCUs be operable.

Answer C Discussion Incorrect. Plausible since LPI pumps are ES required pumps per IS 3.5.3 however only the A and B LPI pumps are required.

Answer D Discussion Incorrect. Plausible since it is reasonable to believe that with the pump switch in OFF the pump will not respond as required and while this will prevent the pump from starting on low seal injection flow, it does not prohibit the required auto start on an ES signal as required by TS 3.5.2 Basis for meeting the KA Requires the ability to analyze component status based on maintenance and determine its impact on the requirements of various Tech Spec LCOs.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level Questionlype Question Source RO Memory NEW Development References Student References Provided Obj. ADM-ITS R8 TS 3.5.1 15 3.6.5 T5 3.5.2 15 3.5.3 GEN2.2 2.2.36 GENERIC Equipment Control Equipment Control Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR:

41.10/43.2/45.13) 401-9 Comments: RemarkslStatus Wednesday, August 31, 2011 Page 144 of 208

FOR REVIEW ONLY DO NOT DISTRIBUTE - V 2011B ONS SRO NRC Examination QUESTION 72 721 GEN2.3 2.3.5 GENERIC Radiation Control Radiation Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 / 41.12 / 43.4 / 45.9)

Which ONE of the following Area Radiation Monitors will sound a LOCAL alarm (do NOT include any associated Statalarms) to indicate increased radiation levels?

A. I RIA-I (Control Room Monitor)

B. I RIA-15 (High Pressure Injection Pump Room Monitor)

C. 1 RIA-17 (B Main Steam Line Monitor)

D. 1 RIA-56 (High Range Stack Monitor)

Wednesday, August 31, 2011 Page 145 of 208

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2011B ONS SRO NRC Examination QUESTION 72 General Discussion Answer A Discussion Incorrect. Plausible since this is an area RIA monitor and most Area monitors provide a local alarm.

Answer B Discussion Correct. IRIA-15 has a local alarm.

Answer C Discussion Incorrect. Plausible since this is an area RIA monitor and most Area monitors provide a local alarm.

Answer D Discussion Incorrect. Plausible since this is an area RIA monitor and most Area monitors provide a local alarm.

Basis for meeting the KA Requires ability to use a fixed radiation monitor and its alarm. The ability to determine whether or not a local alarm would sound to alert you of increasing radiation levels would be required as part of the ability to use effectively use the radiation monitor.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References IStudent References Provided Obj RAD-RIA R2 RAD-RIA GEN2.3 23.5 GENERIC Radiation Control Radiation Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 / 41.12 / 43.4 / 45.9) 401-9 Comments: rksIStatus Wednesday, August 31, 2011 Page 146 of 208

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2011B ONS SRO NRC Examination QUESTION 73 73 GEN2.3 2.3.13 GENERIC Radiation Control Radiation Control Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 /43.4 /45.9 / 45.10)

Given the following Unit I conditions:

IA GWD tank release in progress

I RIA-37 HIGH alarm actuates

I SA-8/B9 (Process Monitor Radiation High) actuates Which ONE of the following describes the

1) automatic actions that will occur?

2) procedure that contains actions that must be performed prior to re-initiating the release?

A. I. Closes the GWD tank outlet valves and isolates the Waste Gas Exhauster but does NOT trip the running GWD compressors

2. OP/I -21A111041018 (GWD System) ONLY B. I. Closes the GWD tank outlet valves, isolates the Waste Gas Exhauster, AND trips running GWD compressors

2. OP/1-2/A/1104/018 (GWD System) ONLY C. I. Closes the GWD tank outlet valves and isolates the Waste Gas Exhauster but does NOT trip the running GWD compressors

2. AP/18 (Abnormal Release of Radioactivity) and OP/I -2/NI 104/018 (GWD System) ONLY D. I. Closes the GWD tank outlet valves, isolates the Waste Gas Exhauster, AND trips running GWD compressors

2. AP/18 (Abnormal Release of Radioactivity) and OP/l -2/NI 104/018 (GWD System) ONLY Wednesday, August 31, 2011 Page 147 of 208

FOR REVIEW ONLY- DO NOT DISTRIBUTE 2011B ONS SRO NRC Examination QUESTION 73 73 General Discussion Answer A Discussion Correct. A HIGH alarm from RIA-37 will close all of the GWD tank outlet valves and isolate the Waste Gas Exhauster. The associated ARG will direct going to OP/1-2/A/l 104/0 18 (GWD System) to provide additional guidance on what to do with the release that has now been terminated.

The entry conditions for AP/1 8 are not met.

Answer B Discussion Incorrect. First part is plausible since it is partially correct in that a HIGH alarm from RIA-37 will close all of the GWD tank outlet valves and isolate the Waste Gas Exhauster. Tripping the GWD compressors is plausible under the misconception that it is the GWD compressors that are provideing the driving force for the tank release. Second part is correct.

Answer C Discussion Incorrect. First part is correct. Second part is plausible since for both RIA-54 (Turbine Building Sump) and RIA-45 (RB Purge), there are actions in AP/l 8 that must be performed prior to going to the associated OP to take actions to resume the release.

Answer D Discussion Incorrect. First part is plausible since it is partially correct in that a HIGH alarm from RIA-37 will close all of the GWD tank outlet valves and isolate the Waste Gas Exhauster. Tripping the GWD compressors is plausible under the misconception that it is the GWD compressors that are provideing the driving force for the tank release. Second part is plausible since for both RIA-54 (Turbine Building Sump) and RIA-45 (RB Purge), there are actions in AP/18 that must be performed prior to going to the associated OP to take actions to resume the release.

Basis for meeting the KA Requires knowledge of actions pertaining to license duties that are directed by a Radiological Safety Procedure (GWD system procedure) in response to radiation monitor alarms.

Basis for Hi Cog Basis for SRO_only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj. EAP-APG R9 RAD-RIA R2 1 SA8/B9 RAD-RIA GWD drwg GEN2.3 2.3.13 GENERIC Radiation Control Radiation Control Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 43.4 / 45.9 / 45.10) 401-9 Comments: Remarks/Status Wednesday, August 31, 2011 Page 148 of 208

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2011B ONS SRO NRC Examination QUESTION 74 74 GEN2.4 2.4.16 GENERIC Emergency Procedures / Plan Emergency Procedures I Plan Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. (CFR: 41.10/43.5 / 45.13)

Given the following Unit I conditions:

Initial conditions:

Reactor power = 90% slowly decreasing

SG Primary to Secondary leak rate 6 gpm stable

APIIIAII700IO3I (Primary to Secondary Leakage) in progress

Unit shutdown in progress Current conditions

Reactor power = 60% slowly decreasing

SG Primary to Secondary leak rate = 28 gpm slowly increasing Which ONE of the following describes the actions required in accordance with plant procedures?

A. Continue unit shutdown using AP/31 B. Exit AP/31 and go directly to SGTR tab C. Exit AP/31, perform IMAs, then go to SGTR tab D. Perform AP/31 in parallel with performing the SGTR tab Wednesday, August 31, 2011 Page 149 of 208

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2011B ONS SRO NRC Examination QUESTION 74 74 General Discussion Answer A Discussion Incorrect. Plausible since AP/3 1 would still be in effect if the leak rate were <25 gpm and some APs are self contained APs that direct unit shutdowns and/or power decreases Answer B Discussion Correct. There is an IAAT in AP/31 that directs going tothe EOP if SG leak rate reaches 25 gpm. The entry conditions of the EOP direct going directly to the SGTR tab with leak rate > 25 gpm.

Answer C Discussion Incorrect. Plausible since exiting API3 1 and entering the EOP is correct and this would be the correct FOP path for most FOP entries however entering while on line with a SO tube leak is a unique exception that requires going direclty to the SGTR tab.

Answer D Discussion Incorrect. Plausible since most APs are performed in parallel with the FOP when EOP entry is required.

Basis for meeting the KA Requires knowledge of the hierarchy of the FOP vs. AP/31.

Basis for Hi Cog Requires analyzing plant conditons and determining a course of action based on the analysis.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Obj EAP-SAR21 AP3 I FOP entry conditions L

EAP-SA GEN2.4 2.4.16 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. (CFR: 41.10 / 43.5 / 45.13) 401-9 Comments:

I RemarkslStatus Wednesday, August 31, 2011 Page 150 of 208

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201 lB ONS SRO NRC Examination QUESTION 75 75 GEN2.4 2.4.49 GENERIC Emergency Procedures / Plan Emergency Procedures I Plan 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 I conditions:

Initial conditions:

Reactor power = 100%

Both Main Feedwater pumps trip Current conditions:

REACTOR TRIP pushbutton has been depressed

Reactor power = 3% slowly decreasing Which ONE of the following describes the NEXT action required in accordance with EOP Immediate Manual Actions?

A. OT-& Rule 1 B. Align HPI Emergency Boration C. Verify RCP seal injection available D. Depress the Turbine TRIP pushbutton Wednesday, August 31, 2011 Page 151 of 208

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2011B ONS SRO NRC Examination QUESTION 75 75 General Discussion Answer A Discussion Incorrect. Plausible since this would be correct if power level was >5%.. Additional plausibility since there is a 1% power threshold for actions within Rule 2 therefore it is plausible to believe that if power is still> 1%, going to Rule is required.

Answer B Discussion Incorrect. Plausible since this is one of the first actions taken by Rule 1 during an ATWS. It is plausible to believe these actions are part of IMAs since it is in IMAs that the ATWS is diagnosed and aligning emergency boration is critical to the successful mitigation of the ATWS.

Answer C Discussion Incorrect. Plausible since this is an action taken in IMAs however it is done after the main turbine is tripped.

Answer D Discussion Correct. Since Rx power is < 5% the next action is to depress the Turbine Trip pushbutton.

Basis for meeting the KA Requires the ability to perform EOP IMAs from memory.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Obj EAP-SA R24 EOP-SA IMAs of EOP GEN2.4 2.4.49 GENERIC Emergency Procedures / Plan Emergency Procedures / Plan 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 Wednesday, August 31, 2011 Page 152 of 208

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