Text

(2) Robinson 2016-301 Draft RO Written Exam
	ML16124A938
Person / Time
Site:	Robinson
Issue date:	04/27/2016
From:	; Division of Reactor Safety II
To:	; Duke Energy Progress
References
	50-261/OL-16-301
	Download: ML16124A938 (202)
	v • d • e

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination EPE007 EK2.02 - Reactor Trip QUESTION 1 1 D Knowledge of the interrelations between a reactor trip and the following: (CFR 41.7 / 45.7)

Breakers, relays and disconnects ...................................

Given the following:

The plant is at 100% power

Maintenance and Testing of A Reactor Trip Breaker is in progress

The A Reactor Trip Bypass Breaker has been placed in service

The A Reactor Trip Breaker has been removed from service

A transient condition exists requiring a reactor trip

The OATC depresses the manual Reactor trip pushbutton Which ONE of the following describes the plant response?

The A Reactor Trip Bypass Breaker opens due to actuation of its ____(1)____ and the B Reactor Trip Breaker opens due to actuation of its ____(2)____.

A. (1) UV trip (2) shunt trip ONLY B. (1) shunt trip (2) UV and shunt trip C. (1) shunt trip (2) UV trip ONLY D. (1) UV trip (2) UV and shunt trip Thursday, December 03, 2015 Page 1 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 1 1 D General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the "A" reactor trip breaker will receive both an UV and a shunt trip actuation. This is plausible because the trip breaker receives an UV trip actuation.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the "A" reactor trip bypass breaker does not receive a shunt trip unless the pushbutton is depressed locally at the panel. Plausible because this trip is actuated by a pushbutton at the panel.

Answer C Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because the "A" reactor trip bypass breaker does not receive a shunt trip unless the pushbutton is depressed locally at the panel. This is plausible because this trip is actuated by a pushbutton at the panel.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to Technical Specifications Bases 3.3.1 Reactor Protection System (RPS) Instrumentation (p B 3.3-6 & B 3.3-7; Rev 0) BACKGROUND Reactor Trip Switchgear, when the required logic matrix combination is completed, the RPS relay logic output voltage signal is removed, the undervoltage coils are de-energized, the breaker trip lever is actuated by the de-energized undervoltage coil, and the RTBs and bypass breakers are tripped open. This allows the shutdown rods and control rods to fall into the core. In addition, to the de-energization of the undervoltage coils, each RTB is also equipped with a shunt trip device that is energized to trip the breaker open upon receipt of a reactor trip signal from the RPS relay logic.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the interrelationship between a reactor trip manual pushbutton and the components used to trip the reactor trip and reactor trip bypass breakers.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must demonstrate knowledge of the components used to trip the reactor trip and reactor trip bypass breakers when the manual reactor trip pushbutton is depressed, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK RPS-001 4 Development References Student References Provided Technical Specifications Bases 3.3.1 Reactor Protection System (RPS)

Instrumentation 10 CFR 55.41.7 EPE007 EK2.02 - Reactor Trip Knowledge of the interrelations between a reactor trip and the following: (CFR 41.7 / 45.7)

Breakers, relays and disconnects ...................................

Remarks/Status 7/17/15 Developed from bank question RPS-001 4 7/20/15 Reviewed by exam supervisor 11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 2 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 2 APE008 2.1.30 - Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) 2 C APE008 GENERIC Ability to locate and operate components, including local controls. (CFR: 41.7 / 45.7)

With the plant at 100% power an event occurs. (SEE GIVEN CONDITIONS ON NEXT PAGE)

(1) Which ONE of the following describes the action(s) required to complete the immediate actions?

After verification of the immediate actions the CRS directs you to contact an Auxiliary Operator to open the supply breaker for RC-535, PORV BLOCK valve.

(2) Which ONE of the following identifies the MCC that provides power to RC-535, PORV BLOCK valve?

A. (1) Close RC-535 ONLY (2) MCC-5 B. (1) Close RC-535 ONLY (2) MCC-6 C. (1) Close RC-535 and control pressurizer heaters and normal PZR spray valves (2) MCC-6 D. (1) Close RC-535 and control pressurizer heaters and normal PZR spray valves (2) MCC-5 Thursday, December 03, 2015 Page 3 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 2 2 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because step 3 of AOP-019 is also an immediate action step to control pressurizer heaters and normal PZR spray to restore RCS pressure and MCC-5 is the other emergency train MCC power supply. This is plausible because closing a PORV block valve for a stuck open PORV is immediate action step 2.b rno action. In addition, most redundant emergency safeguards components have one component powered from MCC-5 and the other train component powered from MCC-6. For PORV block valves both are powered from MCC-6.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because step 3 of AOP-019 is also an immediate action step to control pressurizer heaters and normal PZR spray to restore RCS pressure. This is plausible because closing a PORV block valve for a stuck open PORV is immediate action step 2.b rno action.

Answer C Discussion Correct. 1st part correct, 2nd part correct. 1st part, according to AOP-019 (p3&4; Rev 20) Steps 2 & 3, with given conditions which show PCV-456 open with switch in the close position and pressurizer pressure less than the PORV setpoint of 2335 psig the next correct action IAW AOP-019 is to apply step 2.b rno action, IF any PZR PORV cannot be closed, THEN close its PORV BLOCK valve. The associated PORV block valve for PCV-456 is RC-535. The last immediate action of AOP-019 step 3 is, Control the Normal PZR Spray Valves AND PZR Heaters to Restore RCS Pressure to the Desired Control Band. 2nd part, according to EDP-003 (p29; Rev 68) PORV Block valves RC-535 and RC-536 power supplies are both MCC-6.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because MCC-5 is the other emergency train MCC power supply and not the power supply for the PORV block valves. This is plausible because most redundant emergency safeguards components have one component powered from MCC-5 and the other train component powered from MCC-6. For PORV block valves both are powered from MCC-6.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to locate and operate components, including local controls as related to a Pressurizer Vapor Space Accident. This is accomplished by giving conditions for a stuck open PZR PORV requiring knowledge of the required immediate actions related to this event at the RTGB and knowledge of the location of the power supplies for the breakers associated with the PZR PORV Block Valves. Technical Specifications LCO 3.4.11 Condition B for one PORV inoperable and not capable of being manually cycled required action B.2 requires remove power from the associated block valve within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Basis for Hi Cog The question is at the Analysis cognitive level because the operator must analyze the given indications of pressurizer pressure being less than the PZR PORV setpoint and lowering combined with the indication of PZR PORV PCV-456 indicating open with its control switch in the closed position and from this information determine that PCV-456 failed to close and was attempted to be closed during performance of the immediate actions, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Technical Specifications LCO 3.4.11 Amendment No. 203 AOP-019 (Rev 20)

EDP-003 (Rev 68) 10 CFR 55.41.7 APE008 2.1.30 - Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

APE008 GENERIC Ability to locate and operate components, including local controls. (CFR: 41.7 / 45.7)

Remarks/Status 7/20/15 New question developed 7/20/15 Reviewed by Exam Supervisor, with 2 Editorial Enhancements 1) Unless AOP-019 and AOP-025 are identical, may need to include a picture of pressure indicators to aid in diagnosis; 2) add RC-535 nomenclature to the stem Thursday, December 03, 2015 Page 4 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 2 2 C 9/10/15 Exam team response, IA of both AOP-019 and AOP-025 are identical to support the correct answer choice, therefore a pressure indicator is NOT required. RC-535 nomenclature was already in the stem following the first instance, added to the second instance in "(2)"

question statement.

9/30/15 question picture edited to show controller in manual with 60% demand to drive the answer for ALL required immediate actions, not just the PORV.

11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 5 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination EPE009 EA1.10 - Small Break LOCA QUESTION 3 3 C Ability to operate and monitor the following as they apply to a small break LOCA: (CFR 41.7 / 45.5 / 45.6)

Safety parameter display system ....................................

Given the following:

The plant is at 100% power

A CV Spray pump is out of service for maintenance Subsequently:

A SBLOCA occurred at 0930 coincident with a Loss of Offsite Power

B EDG tripped immediately after starting

At 1000, the crew is performing the actions of EOP-E-1, LOSS OF REACTOR OR SECONDARY COOLANT with the following current plant conditions:

CV pressure: 11 psig and rising slowly

CV Sump level: 178 inches and rising

CV radiation level: 0.5 R/hr

A Loop Cold Leg temperature: 338 °F

B Loop Cold Leg temperature: 425 °F

C Loop Cold Leg temperature: 427 °F Which ONE of the following correctly completes the statements below?

CSF-5 Containment critical safety function status tree color is ____(1)____.

CSF-4 Integrity critical safety function status tree color is ____(2)____.

A. (1) Yellow (2) Yellow B. (1) Yellow (2) Orange C. (1) Orange (2) Yellow D. (1) Orange (2) Orange Thursday, December 03, 2015 Page 6 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 3 3 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the conditions of CV pressure greater than 10 psig with no CV Spray pumps running results in an orange path for CSF-5 Containment. This is plausible because the given conditions for CSF-4 Integrity result in a Yellow path due to an RCS cold leg temperature dropped greater than 100 °F in last 60 minutes along with one RCS cold leg temperature between 325 °F and 355 °F. CSF-5 Containment does contain a Yellow path which would be possible if the conditions were different, example if CV radiation was 5 R/hr or greater vice 0.5 R/hr which was given.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because the conditions of CV pressure greater than 10 psig with no CV Spray pumps running results in an orange path for CSF-5 Containment. CSF-4 Integrity conditions result in a yellow path due to an RCS cold leg temperature dropped greater than 100 °F in last 60 minutes with one RCS cold leg temperature between 325 °F and 355 °F. This is plausible because if an RCS loop cold leg temperature was below 325 °F vice 355 °F, CSF-4 Integrity would be orange. CSF-5 Containment does contain a Yellow path which would be possible if the conditions were different, example if CV radiation was 5 R/hr or greater vice 0.5 R/hr which was given.

Answer C Discussion Correct. 1st part correct, 2nd part correct. According to CSFST, Critical Safety Function Status Trees (p 6-8; Rev 7) , CSF-4 Integrity yellow path directing operators to go to FRP-P.2 are met due to an RCS cold leg temperature dropped greater than 100 °F in the last 60 minutes with one RCS cold leg below 355 °F but greater than 325 °F. CSF-5 Containment orange path directing operators to go to FRP-J.1 are met due to CV pressure greater than 10 psig combined with no spray pumps running. No spray pumps are running due to A CV Spray pump out of service for maintenance and B CV Spray pump will not start due to no power available due to Loss Of Offsite Power combined with B EDG tripping.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because CSF-4 Integrity conditions result in a yellow path due to an RCS cold leg temperature dropped greater than 100 °F in last 60 minutes with one RCS cold leg temperature between 325 °F and 355 °F. This is plausible because CSF-5 Containment orange path directing operators to go to FRP-J.1 are met due to CV pressure greater than 10 psig combined with no spray pumps running. If an RCS loop cold leg temperature was below 325 °F vice 355 °F, CSF-4 Integrity would be orange.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to monitor and analyze plant parameters given in the Safety parameter display system during a SBLOCA Event. The given parameters are related to the parameters monitored by SPDS.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must analyze bits of information, plant conditions and determine correct color level of the CSFST, in order to answer the question.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided CSFST, Rev. 7 10 CFR 55.41.10 EPE009 EA1.10 - Small Break LOCA Ability to operate and monitor the following as they apply to a small break LOCA: (CFR 41.7 / 45.5 / 45.6)

Safety parameter display system ....................................

Remarks/Status 7/27/15 New Question developed 8/3/15 Reviewed SAT by Exam Supervisor 11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 7 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 3 3 C Thursday, December 03, 2015 Page 8 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination EPE011 EK3.15 - Large Break LOCA QUESTION 4 4 A Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: (CFR 41.5 / 41.10 / 45.6 / 45.13)

Criteria for shifting to recirculation mode ...........................

Given the following:

With the plant at 100% power a Large Break LOCA occurs

Containment pressure rises and peaks at 34.5 psig within 20 seconds Which ONE of the following identifies the criteria for entry into ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, AND what is the basis for this criteria?

A. When the RWST lowers to 27%, ONLY; So that core flow is maintained during the re-alignment to the Recirculation Mode.

B. When EITHER the RWST lowers to 27% OR the CV Sump level reaches 354 inches; So that core flow is maintained during the re-alignment to the Recirculation Mode.

C. When the RWST lowers to 27%, ONLY; So that Containment pressure is maintained less than design limits during the re-alignment to the Recirculation Mode.

D. When EITHER the RWST lowers to 27% OR the CV Sump level reaches 354 inches; So that Containment pressure is maintained less than design limits during the re-alignment to the Recirculation Mode.

Thursday, December 03, 2015 Page 9 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 4 4 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to the Entry Conditions of ES-1.3 (p2 of 39; Rev 0), the procedure is entered from Step 18 of E-1 on low RWST Level, Step 9 of ECA-2.1 on Low RWST Level and other procedure and Foldout Pages whenever the RWST reaches the Switchover Setpoint. According to the E-1 Foldout Page (Rev 4) the switchover setpoint is 27% in the RWST. According to the E-1 Background Document (p36 of 58; Rev 4) the reason for this criteria is to ensure that so that core flow is maintained during the re-alignment to the Recirculation Mode.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because CV Sump Level is NOT considered in the criteria to enter ES-1.3. This is plausible because once entry into ES-1.3 is made, CLR cannot be aligned unless establishing 354 inches in the CV Sump is possible; and the operator may incorrectly believe that it is this criteria that is more binding.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because maintaining Containment pressure less than design limits during the re-alignment to the Recirculation Mode is NOT the reason for the established criteria for entry into ES-1.3. This is plausible because at step 5 of ES-1.3 the operator will reduce ECCS flow and leave both one SI Pump and one CV Spray pump running. Coupled with the information given in the conditions regarding a sever/extreme challenge to the Containment Critical Safety Function, the operator may incorrectly believe that Containment Pressure is more of a concern that Core Cooling.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the reasons for the criteria for shifting to recirculation mode as it applies to the Large Break LOCA.

Basis for Hi Cog The question is at the memory cognitive level because the operator must bits of information about the major strategies, including why a major transition is made at a specific plant condition, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided E-1 Basis Document (p36 of 58; Rev 4)

E-1 Foldout Page (Rev 4)

ES-1.3 (p2 of 39; Rev 0) 10 CFR 55.41.10 EPE011 EK3.15 - Large Break LOCA Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: (CFR 41.5 / 41.10 / 45.6 / 45.13)

Criteria for shifting to recirculation mode ...........................

Remarks/Status 7/28/15 New question developed 8/3/15 Reviewed by Exam Supervisor with Editorial Enhancements, 1) In stem change "reason" to "basis"; 2) Distractor B & D, need to make "or" bold or underline 9/10/15 Exam team incorporated changes as discussed, OR is capital and bold.

11/19/15 In references, changed "E-1 Background Document" to "E-1 Basis Document" (Little) 11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 10 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 4 4 A Thursday, December 03, 2015 Page 11 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE022 AA2.03 - Loss of Reactor Coolant Makeup QUESTION 5 5 B Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: (CFR 43.5/ 45.13)

Failures of flow control valve or controller . . . . . . . . . . . . . . . . . . . . . . .

Given the following:

The plant is at 100% power, when the following annunciators alarm in the control room:

APP-001-B4, RCP SEAL INJECTION HI/LO FLOW

APP-001-B6, LP LTDN LN HI TEMP

APP-001-D6, LP LTDN HI PRESS

APP-001-E6, LP LTDN RELIEF HI TEMP Which ONE of the following identifies the cause of these alarms?

A. Seal Injection Filter clogged B. HCV-121, Charging Flow Valve failed closed C. TCV-144, NRHX Temperature Control Valve failed closed D. CVC-204A, Letdown Line Isolation Stop Valve failed closed Thursday, December 03, 2015 Page 12 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 5 5 B General Discussion Answer A Discussion Incorrect. This is incorrect because Seal Injection Filter clogging will not affect Letdown temperature or pressure negatively. It will result in an increased charging flow through the Regenerative Heat Exchanger. This is plausible because it will result in APP-001-B4 alarming on low seal injection flow.

Answer B Discussion Correct. According to OP-301-1 (p 126 & 136; Rev 64) Caution prior to Step 19 states, If charging flow is significantly less than letdown flow, letdown temperature may rise rapidly and could cause Annunciators APP-001-B6 and APP-001-A6. Caution prior to Step 6.6.4.2 states, If sufficient charging flow is not maintained the potential for lifting the letdown relief is elevated. According to OP-301 (p 9; Rev 112) Step 5.7.4 states, Flow may be diverted to the Reactor Coolant Pump seals as HIC-121 is throttled closed.

HCV-121 closed results in all charging flow going to the RCP seals resulting in APP-001-B-4 alarming. HCV-121 closing isolates cooling to the Regenerative Heat Exchanger resulting in letdown line high temperature and pressure resulting in RTGB alarms APP-001-B6, D6 and E6.

Answer C Discussion Incorrect. This is incorrect because closed will not affect RCP seal injection flow. This is plausible because it will result in a letdown high temperature alarm APP-001-A6, LTDN FLOW HI TEMP DEMIN BYPD.

Answer D Discussion This is incorrect because CVC-204A closed will not affect RCP seal injection flow to cause all the alarms listed. This is plausible because CVC-204A closing will isolate letdown flow path downstream of the Regenerative Heat Exchanger. This would result in lifting of letdown pressure relief valve causing annunciator APP-001-E6 to alarm. Additionally with letdown isolated, PZR level will begin to rise above programmed level, then causing the Charging pump that is running in auto to slow down and would reduce the flow to the RCP Seals, and could cause APP-001-B4 to alarm due to LO FLOW.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to interpret RTGB annunciators and determine the cause to be Charging Flow Valve (HCV-121) being closed which resulted in a loss of charging line flow.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must put together bits of information and analyze this information to determine the cause, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided OP-301-1 Rev. 64 OP-301 Rev. 112 ST-021 Rev. 3a APP-001 Rev. 60 10 CFR 55.41.7 APE022 AA2.03 - 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)

Failures of flow control valve or controller . . . . . . . . . . . . . . . . . . . . . . .

Remarks/Status 8/10/15 New question developed 9/8/15 Exam Supervisor review with Editorial comment, 1) Should we give them Charging Pump status in auto?

9/10/15 Exam team response is that if the candidates are not told otherwise then the plant operates as expected, and the running charging pump at 100% should be in auto and assumed as such.

11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 13 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 5 5 B Thursday, December 03, 2015 Page 14 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE025 2.4.47 - Loss of Residual Heat Removal System (RHRS)

QUESTION 6 6 A APE025 GENERIC Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR:

41.10 / 43.5 / 45.12)

Given the following:

A plant cooldown is in progress

The A RHR Train is in service

At 0700 RCS temperature was 130°F, and the cooldown rate was adjusted for a constant 50°F/hour Subsequently, the following is observed:

Time RCS Temp 0715 145 0730 161 0745 178

RHR flow is 3000 gpm

FCV-605, RHR HX BYPASS, controller demand is 60%

HCV-758, RHR HX OUTLET FLOW TO COLD LEGS, controller demand is 100%

RCS pressure has been controlled between 350-375 psig during this time Which ONE of the following correctly completes the statements below?

1) The cause of the observed condition is a loss of ____(1)____ flow through the RHR HX.

2) If the current heat up rate trend continues as it has for the last 45 minutes, by 0800 the Technical Specification RCS Heat up limit ____(2)____ be exceeded.

(REFERENCES PROVIDED)

A. (1) CCW (2) will B. (1) RHR (2) will C. (1) CCW (2) will not D. (1) RHR (2) will not Thursday, December 03, 2015 Page 15 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 6 6 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to ST-003 (p16; Rev 4a) RHR flows through the tubes and CCW flow around the tubes of the RHR HX. The indications provided show that flow exists in the RHR System and that the RHR HX Outlet valve is fully OPEN. However, RCS temperature is rising at an increasing rate. Consequently, CCW flow must have been interrupted to the RHR HX. The current trend in RCS temperature is rising at an increasing rate. The trend developing is a 15°F rise in the first 15 minutes, 16°F rise in the second 15 minutes, and 17°F rise in the third 15 minutes. If this trend does not change, by 0800 the RCS will be expected to rise by 18°F, which will yield a 66°F heat up since 0700. During the cooldown the operator is recording cooldown rates every fifteen minutes on Attachment 1 of GP-007. P&L 3.4 of GP-007 (p7 of 149; Rev 101) indicates that TS LCO 3.4.3 states, RCS pressure, RCS temperature, and RCS Heatup and Cooldown rates shall be maintained within the limits specified in Figures 3.4.3-1 and 3.4.3-2, and a note is provided prior to this stating that this Figure is located in the Curve Book. Clearly, this Figure will be available to the operator during the cooldown evolution. With this Figure available, the operator will conclude that the allowable heatup rate is 60°F in one hour, and conclude that TS limit will be exceeded.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because RTGB indications show that there is full RHR flow through the "A" RHR HX. This is plausible because the operator may incorrectly interpret the RTGB indications, the controller demand positions, as some controllers at RNP are reverse acting and would therefore require a 0% demand signal to be open (Which is what the KA is requiring the operator to do).

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the current trend shows that the RCS Heatup will be 62°F heat up since 0700. This is plausible because the operator may incorrectly interpret the RCS trend or use Figure 3.4.3-1 (Which is what the KA is requiring the operator to do).

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. This is accomplished by placing the operator in a situation where an RCS cooldown is in progress and RHR cooling is subsequently lost. The operator will be required to evaluate typical RTGB indications and diagnose which fluid flow through the RHR HX has been lost. Then, the operator will be required to identify the trend on the RCS temperature rise (X, X+1, X+2, etc), and predict what the temperature will be at a specific future time. The operator is provided with a reference (That would normally be available during an RCS cooldown) that will identify the maximum permissible Heat Up Rate.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall or use a reference to determine bits of information (flowpaths through the RHR HX, RCS TS Heatup rate) and then use this information when evaluating RCS temperature trends to predict and outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided ST-003 (p16; Rev 4a) Figure 3.4.3-1 10 CFR 55.41.10 APE025 2.4.47 - Loss of Residual Heat Removal System (RHRS)

APE025 GENERIC Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR:

41.10 / 43.5 / 45.12)

Remarks/Status 8/13/15 New question developed 9/8/15 Exam Supervisor reviewed SAT, however the 6th bullet requires an editorial change, correct the nomenclature for HCV-758.

Thursday, December 03, 2015 Page 16 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination CORRECTED.

QUESTION 6 6 A 11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 17 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE027 AK2.03 - Pressurizer Pressure Control System (PZR PCS) Malfunction QUESTION 7 7 B Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: (CFR 41.7 / 45.7)

Controllers and positioners ........................................

Given the following:

The plant is at 100% power

All Pressurizer heaters are energized

RCS pressure is 2270 psig and rising Which ONE of the following completes the statement below?

Based on plant conditions and no operator action taken, ___(1)___ failed low and

____(2)____, PZR PORV, WILL cycle open and closed.

A. (1) PT-444 (2) PCV-455C B. (1) PT-444 (2) PCV-456 C. (1) PT-445 (2) PCV-456 D. (1) PT-445 (2) PCV-455C Thursday, December 03, 2015 Page 18 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 7 7 B General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because PT-444 failing low will prevent PCV-455C from opening, PCV-455C requires an open signal from PT-444 at 2335 psig. This is plausible because PT-444 is the transmitter that provides a comparison signal for PZR pressure controller PC-444J to control PZR pressure in AUTO and PCV-455C receives a signal from PT-444.

Answer B Discussion Correct. According to AOP-025 Basis Document (p 14; Rev 24) Section C-Pressurizer Transmitter Failure, PT-444 is used for automatic control of PZR heaters and sprays and also PCV-455C. PT-445 is used for control of PCV-456. PT-444 failed low: Indicated pressure would be low causing PZR heaters to energize. PZR spray and PCV-455C could not respond to the actual pressure increase that would occur.

Answer C Discussion Incorrect. 1St part wrong, 2nd part correct. This is incorrect because PT-445 is not the transmitter that provides a comparison signal to PZR pressure controller PC-444J for controlling PZR pressure. This is plausible because PT-445 is a PZR pressure transmitter that provides a signal to operate PCV-456.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because PT-445 is not the transmitter that provides a comparison signal to PZR pressure controller PC-444J for controlling PZR pressure and PT-444 failing low will prevent PCV-455C from opening. This is plausible because PT-445 is a PZR pressure transmitter that provide a signal to operate a PZR PORV and PCV-455C is a PZR PORV.

Basis for meeting the KA The KA is matched because the operator must demonstrate an understanding of the system response to a malfunction of pressurizer pressure control. This is demonstrated by failure of PT-444 which provides a comparison signal for PZR pressure controller PC-444J and the effect on the PZR PORVs.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must put together bits of information to determine the effect on plant pressure, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 027 AK 2.03 1 Development References Student References Provided AOP-025 Basis Document, RTGB INSTRUMENT FAILURE, Rev. 24 10 CFR 55.41.7 APE027 AK2.03 - Pressurizer Pressure Control System (PZR PCS) Malfunction Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: (CFR 41.7 / 45.7)

Controllers and positioners ........................................

Remarks/Status 7/16/15 Bank question used for RO 7 7/27/15 Reveiwed by Exam Supervisor with Editorial Enhancement comment, 1) May need to add more to the stem to help clarify that the PZR PORV operation may not be tied to the transmitter failure.

9/10/15 Exam team response is that they have enough information to correctly answer the question. With the transmitter failed low, the PORV cycling will not be the one associated with the failed transmitter.

11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 19 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 7 7 B Thursday, December 03, 2015 Page 20 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination EPE029 EK2.06 - Anticipated Transient Without Scram (ATWS)

QUESTION 8 8 B Knowledge of the interrelations between the ATWS and the following: (CFR 41.7 / 45.7)

Breakers, relays, and disconnects ...................................

Given the following:

An ATWS has occurred

Immediate Actions of FRP-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, are in progress Which ONE of the following correctly completes the statements below?

The MINIMUM number of breakers in the Rod Drive MG Set Room that will need to be opened to trip the reactor locally is ____(1)____.

The MINIMUM number of breakers in the 4 KV Room that will need to be opened to trip the reactor locally is ____(2)____.

A. (1) one (2) one B. (1) one (2) two C. (1) two (2) one D. (1) two (2) two Thursday, December 03, 2015 Page 21 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 8 8 B General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because both MG Sets must be deenergized to trip the reactor locally from the 4KV Room. This is plausible because the reactor can be tripped locally by opening only one Reactor Trip Breaker and opening one MG Set Motor Input Breaker will trip one Rod Drive MG Set.

Answer B Discussion Correct. 1st part correct, 2nd part correct. According to FRP-S.1 (p 3; Rev 22) Step 1 rno, requires opening BOTH MG Set Generator Output breakers or BOTH MG Set Motor Input Breakers. According to ST-007, Rod Control System (p 4; Rev.4) Each MG Set has the capacity to meet the power requirements of the entire system. Normally, the MG set are operated in parallel so that a loss of one MG set does not cause a dropped control rod. If the reactor trip breakers open, power is lost to all four power cabinets and, therefore, to the CRDMs. Without current flow, the stationary coils deenergize and the stationary latches disengage allowing the rods to drop into the core by gravity.

Answer C Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because opening only one reactor trip breaker will trip the reactor and both MG Sets must be deenergized to trip the reactor locally from the 4KV Room. This is plausible because the reactor can be tripped locally by opening only one Reactor Trip Breaker and opening one MG Set Motor Input Breaker will trip one Rod Drive MG Set.

Answer D Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because opening only one reactor trip breaker will trip the reactor. This is plausible because the reactor can be tripped locally by opening only one Reactor Trip Breaker and opening both MG Set Motor Input Breaker will trip the reactor.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the interrelations between an ATWS and breakers associated with tripping the reactor.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recallt bits of information, specifically the electrical distribution to supply power to the control rods and the breakers to open to remove that power, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided FRP-S.1 Rev. 22 ST-007 Rev. 4 10 CFR 55.41.7 EPE029 EK2.06 - Anticipated Transient Without Scram (ATWS)

Knowledge of the interrelations between the ATWS and the following: (CFR 41.7 / 45.7)

Breakers, relays, and disconnects ...................................

Remarks/Status 7/27/15 New question developed 8/3/15 Reviewed by Exam Supervisor SAT 11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 22 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination EPE038 EK1.04 - Steam Generator Tube Rupture (SGTR)

QUESTION 9 9 D Knowledge of the operational implications of the following concepts as they apply to the SGTR: (CFR 41.8 / 41.10 / 45.3)

Reflux boiling ...................................................

Given the following:

A major accident has occurred

One S/G has experienced a tube rupture

The core is being cooled by reflux cooling Which ONE of the following correctly completes the statements below?

1) The operator must take action to ensure ____(1)____ in order to promote reflux cooling.

2) Reflux cooling will occur in the ruptured S/G as long as the ruptured S/G ____(2)____

remains less than that of the RCS.

A. (1) water levels are maintained in the intact S/Gs (2) pressure B. (1) the intact S/Gs are fully depressurized (2) pressure C. (1) the intact S/Gs are fully depressurized (2) temperature D. (1) water levels are maintained in the intact S/Gs (2) temperature Thursday, December 03, 2015 Page 23 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 9 9 D General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because Reflux cooling will occur in the ruptured S/G as long as the ruptured S/G temperature, not pressure, remains less than that of the RCS. This is plausible because the operator may mis-understand how Reflux Boiling works and/or when evaluating whether or not the Steam Generators are required as a secondary heat sink in FRP-H.1 the pressures of both the RCS and intact S/Gs are considered.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and C.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because reflux occurs independently of secondary side pressure control. This is plausible because EOP-ECA-1.1, a potential procedure to deal with a SGTR, and subsequent large break LOCA, contains actions to depressurize the intact S/Gs to atmosphere in an effort to promote any kind of core cooling.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to EOP-ECA-1.1-BD (p 63; Rev 2), EOP Step 35 KNOWLEDGE: If the RCS is not full of liquid at this time, it is especially important to keep the secondary system adequately full of water to promote reflux cooling. Reflux cooling is the mechanism by which steam that is generated in the RCS enters the S/G tubes and is condensed by the cold water on the S/Gs secondary side. This liquid then remains in the primary system and promotes cooling. According to LP LOF016R, (p89; Rev 0), refluxing is a heat and mass transfer process. Some of the steam produced by the core is condensed in the steam generator tubes, flows back down the hot legs to the core, thus transferring energy from the core to the steam generators. Refluxing will continue until the RCS temperature drops below the steam generator temperature, resulting in a loss of condensation.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of operational implications (need to feed S/Gs, Ruptured S/G can function as a heat sink if temperature below that of RCS) of reflux boiling as it applies to the SGTR.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must demonstrate an understanding of the phenomenon of Reflux Cooling/Boiling, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EOP-ECA-1.1-BD (p 63; Rev 2)

LOF016R, (p89; Rev 0) 10 CFR 55.41.5 EPE038 EK1.04 - Steam Generator Tube Rupture (SGTR)

Knowledge of the operational implications of the following concepts as they apply to the SGTR: (CFR 41.8 / 41.10 / 45.3)

Reflux boiling ...................................................

Remarks/Status 8/5/15 Bank question 038 EK1.04 1 used.

8/12/15 Reviewed by Exam Supervisor with Editorial Enhancement comment, 1) Revise last bullet in stem to make information regarding SI flow more obvious (don not use old rev?)

9/11/15 discussed with Exam Supervisor, changed SI to ECCS to make more apparent that there is no flow into the core.

11/17/15 Fleet reviewed determined question could potentially be unsat due to possibly having 2 correct answers. Revised to eliminate that problem.

11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 24 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 9 9 D Thursday, December 03, 2015 Page 25 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE040 AK1.05 - Steam Line Rupture QUESTION 10 10 C Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture: (CFR 41.8 / 41.10 / 45.3)

Reactivity effects of cooldown .....................................

Given the following:

The plant is at 100% RTP

A Main Steam Line Break occurs on the 72 header

A Main Steam Line Isolation Signal is generated Which ONE of the following correctly completes the statements below?

(1) The Main Steam Isolation Signal was generated as a result of ____(1)____.

(2) The reactivity effects due to this event are more severe when ____(2)____.

A. (1) Hi steam line P (2) RCS Boron = 200 ppm B. (1) Hi steam line P (2) RCS Boron = 1200 ppm C. (1) Hi steam line flow with a low Tave or low steam line press (2) RCS Boron = 200 ppm D. (1) Hi steam line flow with a low Tave or low steam line press (2) RCS Boron = 1200 ppm Thursday, December 03, 2015 Page 26 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 10 10 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the location of the steam line break will cause all steam line pressures to lower at the same rate, and therefore not cause a Hi steam line delta P. This is plausible because the high steam line delta P signal generates a safety injection signal.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See A & D.

Answer C Discussion Correct. 1st part correct, 2nd part correct. According to FRP-S.1-BD (p 21; Rev 22) Step 12, An uncontrolled cooldown of the RCS is indicated by either an uncontrolled RCS temperature reduction of an uncontrolled S/G pressure reduction. Such an RCS cooldown could add a significant amount of positive reactivity to the core, depending on the current value of the moderator temperature coefficient. According to Westinghouse Owners Group Emergency Response Guideline E-2 Faulted Steam Generator Isolation (p 2; Rev 2) the return to criticality for cooldown events concern is highest near the end of core life when Moderator Temperature Coefficient is most negative. According to ST-048 (p 4 & 5; Rev 4) a break downstream of the MSIVs results in a Hi steam line flow with low Tave or low steam line press safety injection safety injection signal and a main steam isolation signal. Hi Steam Flow (37.25% flow at no load to 20% load, increases linearly to 109% at full load) detected by at least one sensor on two of three steam lines, coincident with low Tavg (543°F) or low steam line pressure (614 psig), generates a Safety Injection signal and closes all MSIVs. Two flow controllers on each steam line are used to sense high steam line flow. This circuit is designed to detect steam line breaks downstream of the MSIVs.

Steam Line Pressure measurement is utilized for steam line break protection. Low steam line pressure (614 psig) in two of three main steam lines or Low Tavg (543°F) in two of three loops, coincident with high steam line flow in two-of-three main steam lines, will initiate the Steam Line Isolation and Safety Injection signals. This is to protect against: a steam line break downstream of the main steam check valves, a feed line break, and/or an inadvertent opening of a SG safety. In addition, each steam line pressure measurement is compared with a main steam header pressure measurement to determine if a high steam line differential pressure exists. A coincidence of two-of-three steam line differential pressures (100 psid) in any one steam line, that is, steam line pressure lower than main steam header pressure, will initiate a Safety Injection signal. With a break on the 72" header all the steam lines and the header will be effected and subject to the same pressures, therefore the safeguards trip that will occur is the Hi Steam Line Flow coincident with Low Tavg or Steam Line Pressure.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because End of Life has a higher MTC than BOL. This is plausible because low power has more mass in S/G to boil off resulting in larger cool down and MTC at BOL may have a positive MTC.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the operational implications of the reactivity effects of cooldown as related to a Steam Line Rupture in particular plant conditions which result in the most severe reactivity event.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must put together bits of information and analyze this information , in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED MNS #3569 Development References Student References Provided FRP-S.1-BD Rev 22 ST-048 Rev 4 Westinghouse Owners Group Emergency Response Guideline E-2 Rev 2 10 CFR 55.41.14 APE040 AK1.05 - Steam Line Rupture Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture: (CFR 41.8 / 41.10 / 45.3)

Reactivity effects of cooldown .....................................

Remarks/Status 8/5/15 Bank question MNS #3569 used Thursday, December 03, 2015 Page 27 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 10 10 C 8/12/15 Reviewed by Exam Supervisor, SAT 11/20/15 rewrote question based on input from operations to better meet KA and make the question less GFES.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 28 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE054 AA1.03 - Loss of Main Feedwater (MFW)

QUESTION 11 11 A Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW):(CFR 41.7 / 45.5 / 45.6)

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

Given the following:

The plant is at 100% power, when a loss of both MFW pumps and Service Water occurs

CST level is 12% and rapidly lowering

The SDAFW pump is out of service

The crew is implementing EOP-ES-0.1, REACTOR TRIP RESPONSE Which ONE of the following is required IAW OP-402, AUXILIARY FEEDWATER SYSTEM to supply the OIL COOLER for operation of the B MDAFW pump under these conditions?

A. Fire Water B. Potable Water C. Deepwell Water D. Component Cooling Water Thursday, December 03, 2015 Page 29 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 11 11 A General Discussion Answer A Discussion Correct. According to OP-402 (p49; Rev 96) Note before Step 8.4.2.2.d states, when Deepwell Water is being supplied to the AFW Pumps, the MDFAFW Pumps should have cooling water to the oil coolers and sealing water to the seals supplied from the Firemain VIA direction from the EOP Network. If this is NOT done, lube oil temperature should NOT exceed 140°F (IAW Tech Manual) AND no damage to the seals will occur. According to AOP-022 (p24; Rev 35) Caution before Step 19 states, Normal cooling water supply to MOTOR DRIVEN AFW PUMP B is unavailable. Pump operation prior to establishing emergency cooling could result in pump damage. Step 19 states to align Emergency Cooling Water to MOTOR DRIVEN AFW PUMP B using Attachment 2 while continuing With this procedure. Attachment 2 provides instructions for aligning Fire Water via hoses to provide cooling water to the MDAFW Pumps.

Answer B Discussion Incorrect. This is incorrect because IAW OP-402 and AOP-022 cooling water is supplied to the "B" MDAFW Pump via Fire Water System under these plant conditions. This is plausible because Potable Water can be aligned for cooling other plant pump bearings such as SI pump thrust bearings.

Answer C Discussion Incorrect. This is incorrect because IAW OP-402 and AOP-022 cooling water is supplied to the "B" MDAFW Pump via Fire Water System under these plant conditions. This is plausible because Deepwell pumps do provide a water supply to the suction of the MDAFW Pumps.

Answer D Discussion Incorrect. This is incorrect because IAW OP-402 and AOP-022 cooling water is supplied to the "B" MDAFW Pump via Fire Water System under these plant conditions. This is plausible because some plant pumps are provided cooling from the CCW System.

Basis for meeting the KA The KA is matched because the operator must demonstrate an understanding of the operation of the MDAFW pumps related to the water supply for oil cooler when Service Water is unavailable.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall from memory knowledge of the alternate cooling supply for the MDAFW pumps when Service Water is unavailable, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK AFW-015 1 Development References Student References Provided OP-402 Rev. 96 EOP-E-0 Rev. 6 AOP-022 Rev. 35 10 CFR 55.41.7 APE054 AA1.03 - Loss of Main Feedwater (MFW)

Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW):(CFR 41.7 / 45.5 / 45.6)

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

Remarks/Status 8/11/15 Bank question AFW-015 1 used.

9/8/15 Reviewed by Exam Supervisor SAT 11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 30 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination EPE055 EK3.02 - Loss of Offsite and Onsite Power (Station Blackout)

QUESTION 12 12 D Knowledge of the reasons for the following responses as the apply to the Station Blackout : (CFR 41.5 / 41.10 / 45.6 / 45.13)

Actions contained in EOP for loss of offsite and onsite power ..........

Given the following:

The crew is implementing EOP-ECA-0.0, LOSS OF ALL AC POWER, mitigating an Extended Loss of AC Power (ELAP)

A depressurization of all S/Gs was initiated at 10:00

The following timeline of events is observed:

Time S/G Pressures Pzr Level 10:15 500 psig 13%

10:45 190 psig 0%

Which ONE of the following correctly completes the statements below?

1) The EARLIEST time at which the depressurization is required to be stopped is

____(1)____.

2) The reason the depressurization is stopped is to ____(2)____.

A. (1) 1015 (2) prevent injection of accumulator nitrogen into the RCS B. (1) 1045 (2) ensure that no reactor head voiding will occur C. (1) 1015 (2) ensure that no reactor head voiding will occur D. (1) 1045 (2) prevent injection of accumulator nitrogen into the RCS Thursday, December 03, 2015 Page 31 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 12 12 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the operator is directed to depressurize the S/Gs to 290 psig regardless of the condition of Pzr Level. This is plausible because the operator may incorrectly believe that Pressurizer level must be maintained greater than 14% or on-scale.

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because at time 10:15 the pressurizer level falls below 14%, which was previously checked in the last step before the depressurization. This is plausible because Rx Head voiding is specifically identified in a series of Cautions and Notes prior to initiating the S/G depressurization. A note is provided stating that "PRZR level may be lost AND Reactor Vessel upper head voiding may occur due to depressurization of S/Gs. Depressurization should NOT be stopped to prevent these occurrences." The operator may misunderstand the notes and cautions, and believe that the loss of RCS inventory and PZR level less than 14% is reason to stop the depressurization.

Answer C Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and B.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to ECA-0.0 (p34 of 104; Rev 4), the operator is directed to DEPRESSURIZE Intact S/Gs To 290 PSIG IF RCP Seal Cooling Is Lost. Prior to this the operator is cautioned that S/G pressures should NOT be lowered to LESS THAN 190 psig to prevent injection of SI Accumulator nitrogen into the RCS. Additionally, a Note is provided to alert the operator that PRZR level may be lost AND Reactor Vessel upper head voiding may occur due to depressurization of S/Gs. Depressurization should NOT be stopped to prevent these occurrences. Consequently the earliest time that the depressurization should be stopped is 1045, and the reason for stopping the depressurization is to prevent injection of SI Accumulator nitrogen into the RCS.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the reason for stopping the S/G depressurization, a Major Action Item in ECA-0.0, before RCS pressure reaches 190 psig.

Basis for Hi Cog The question is at the Comprehensive/Analysis cognitive level because the operator must recall the reason for a Caution in ECA-0.0, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided ECA-0.0 (p34 of 104; Rev 4) 10 CFR 55.41.10 EPE055 EK3.02 - Loss of Offsite and Onsite Power (Station Blackout)

Knowledge of the reasons for the following responses as the apply to the Station Blackout : (CFR 41.5 / 41.10 / 45.6 / 45.13)

Actions contained in EOP for loss of offsite and onsite power ..........

Remarks/Status 7/17/15 Bank question EPP-001-03 030 used 7/20/15 Reviewed by Exam Supervisor, with Editorial Comment, 1) What is the benefit from asking the exact pressure? Is this required to be known from memory?

9/10/15 Exam team response, they are choosing between 2 exact pressures, which they should be able to easily determine the reason between the selection of one over the other in the procedure in question.

9/29/15 changed question to new question.

11/05/15 discussed with the NRC chief examiner as one of the early submission items. The following Recommendations were made:

Thursday, December 03, 2015 Page 32 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 12

- Remove unnecessary information from current initial conditions. We can discuss specifics.

12 D

- Give two times (30 minutes apart) with S/G pressure and pressurizer level. For the first time, have pressurizer level less than 14%. For the second time, have S/G pressure at 190 psig and pressurizer level going off scale low.

- Ask: IAW ECA-0.0, what is the earliest time that the S/G depressurization is required to be stopped AND the reason for stopping the depressurization (preventing injection of accumulator nitrogen into the RCS or ensuring that no Reactor Vessel upper head voiding will occur).

11/09/15 question was revised based on the above comments.

11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 33 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE056 AA2.11 - Loss of Offsite Power QUESTION 13 13 D Ability to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 / 45.13)

Operational status of service water booster pump .....................

Given the following:

The plant is at 100% power when a loss of off-site power occurs

Fifteen seconds later an SI actuation occurred due to a Large Break LOCA Which ONE of the following correctly completes the statements below?

The BOP will expect to see the SW Booster Pumps started on the ____(1)____.

SW Booster Pump suction pressure ____(2)____ have to be at least 30 psig for the SW Booster Pump to start.

A. (1) Blackout Sequencer (2) does B. (1) Blackout Sequencer (2) does NOT C. (1) SI Sequencer (2) does D. (1) SI Sequencer (2) does NOT Thursday, December 03, 2015 Page 34 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 13 13 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because the blackout sequencer did start, however, the SW Booster Pumps won't be started until 20 seconds after the blackout signal. Since an SI will occur due to the LBLOCA, the SI sequencer takes over. The SW Booster Pumps will start on the SI Sequencer. Plausible since the Blackout sequencer does start the SW Booster Pumps. Suction pressure does not matter for the SI sequencer. Plausible because suction pressure needs to be >30 psig for the SW Booster Pumps to start on the blackout sequencer.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the blackout sequencer did start, however, the SW Booster Pumps won't be started until 20 seconds after the blackout signal. Since an SI will occur due to the LBLOCA, the SI sequencer takes over. The SW Booster Pumps will start on the SI Sequencer. Plausible since the Blackout sequencer does start the SW Booster Pumps. Suction pressure does not need to be 30 psig to start on the SI sequencer. Plausible because the SI sequencer does not need to be > 30 psig to start the SW Booster Pump.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the SW Booster Pumps do get started on the SI Sequencer however Suction pressure does not need to be > 30 psig to start the SW Booster Pump on the SI Sequencer. This is plausible because if the SW Booster Pump suction pressure would be required to be > 30 psig to be started by the Blackout Sequencer.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to ST-006 (p30-31&39-41; Rev 4), on a LOOP both EDGs start and load onto Buses E-1 and E-2 in 10 seconds. If there was not a subsequent SI the SW Booster Pumps require a SW Pump running and Service Water Header Pressure of 30 psig to start. The "A" SW Pump starts at 20 seconds on the Blackout Sequencer and the "B" SW Pump starts at 25 seconds on the Blackout Sequencer. Since the SI actuation occurred 15 seconds after the LOOP the EDGs have already started and loaded onto emergency buses E-1 and E-2 however, the SW Pumps and therefore the SW Booster Pumps have not yet started on the Blackout Sequencer. When the SI actuation occurs after the Blackout Sequencer, the Blackout Sequencer stops wherever it is, and the SI sequence starts at the beginning. Once the SI actuation occurs with power to buses E-1 and E-2 at 20 seconds "A" and "C" SW Pumps start on their respective SI Train and either SW Pump will supply a start signal to both SW Booster Pumps without requiring the minimal pressure of 30 psig on the SW Header.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to interpret plant conditions to determine the operational status of SW Booster Pumps for a LOOP combined with a subsequent LOCA.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must put together bits of information and analyze this information to determine how the SW Booster Pumps would respond for the given plant conditions, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 056 AA1.15 1 Development References Student References Provided ST-004 Rev. 5a ST-006 Rev. 4 10 CFR 55.41.7 APE056 AA2.11 - Loss of Offsite Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 / 45.13)

Operational status of service water booster pump .....................

Remarks/Status 8/11/15 Bank question 056 AA1.15 1 used.

9/8/15 Reviewed by Exam Supervisor, SAT 11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 35 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 13 13 D Thursday, December 03, 2015 Page 36 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE057 AA2.15 - Loss of Vital AC Electrical Instrument Bus QUESTION 14 14 B Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: (CFR: 43.5 / 45.13)

That a loss of ac has occurred ......................................

Given the following:

The plant is operating at 100% power

A Startup Transformer lockout occurs Which ONE of the following describes how Instrument Buses 1 and 4 will respond to this event?

A. Both Instrument Buses 1 and 4 will de-energize B. Instrument Bus 1 will remain energized; Instrument Bus 4 will momentarily de-energize until the Emergency Diesel Generator re-powers the associated Emergency Bus C. Instrument Bus 4 will remain energized; Instrument Bus 1 will momentarily de-energize until the Emergency Diesel Generator re-powers the associated Emergency Bus D. Both Instrument Buses 1 and 4 will remain energized Thursday, December 03, 2015 Page 37 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 14 14 B General Discussion Answer A Discussion Incorrect. This is incorrect because MCC-5 supplies power directly to Instrument Bus 1, and it is still energized. This is plausible because if a Station Blackout were to occur, both Buses 1 and 4 will de-energize, and/or if this event occurred in Mode 3 when both Bus 2 and 3 were being powered by the SUT, both Buses 1 and 4 will de-energize.

Answer B Discussion Correct. According to ST-039 (p15; Rev 4b) during normal power operation 4KV Buses 1, 2, 4 and 5 are supplied via the UAT, and 4KV Bus 3 is supplied via the SUT. Consequently, when the SUT Lockout occurs, Bus 3 will be de-energized. When Bus 3 is de-energized, 480V Bus E-2 will de-energize causing the B DG to start and re-power Bus E-2. During the time that E-2 is de-energized, Instrument Bus 4, which is normally powered from MCC-6, will be de-energized as well. According to ST-016 (p52; Rev 4) if undervoltage occurs on one or both emergency buses the bus will isolate by opening and locking out normal supply and tie breakers. Load shedding will occur with the exception of MCC-5, -10, and -16 for E1 and MCC-6, -9, and -18 for E2. The EDG will start and sequence loads back on to the emergency bus. According to AOP-024-BD (p3 of 62; Rev 41), Buses 1 and 4 are supplied directly from MCC-5 and 6 (480V Emergency Buses), and Buses 2 and 3 are supplied from battery backed inverters. Consequently, if MCC-6 is de-energized, Instrument Bus 4 will also be de-energized.

Answer C Discussion Incorrect. This is incorrect because MCC-5 supplies power directly to Instrument Bus 1, and it is still energized, and MCC-6 supplies power directly to Instrument Bus 4, and it is de-energized. This is plausible because the operator may incorrectly believe that at 100% power the SUT powers Bus 2 and the UAT powers Bus 3. If so, one would conclude that this is correct.

Answer D Discussion Incorrect. This is incorrect because MCC-6 supplies power directly to Instrument Bus 4, and it is de-energized. This is plausible because the operator may incorrectly believe these two buses are powered by a DC Bus, via an Inverter, such as, or rather than Instrument Bus 2 and 3.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to determine that a loss of ac has occurred as it applies to a Loss of Vital AC Instrument Bus. In this case the operator must identify the status of two instrument Buses after an SUT lockout (Loss of AC).

Basis for Hi Cog The question is at the Comprehensive/Analysis cognitive level because the operator must recall the normal power supply to Instrument Buses 1&4, and then use this information to predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided ST-039 (p15; Rev 4b)

ST-016 (p52; Rev 4)

AOP-024-BD (p3 of 62; Rev 41) 10 CFR 55.41.7 APE057 AA2.15 - Loss of Vital AC Electrical Instrument Bus Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: (CFR: 43.5 / 45.13)

That a loss of ac has occurred ......................................

Remarks/Status 7/28/15 New question developed 8/5/15 Reveiwed by Exam Supervisor, as UNSAT for Job-Link and Minutia, with following comments, 1) These alarms are not on the inverter, does this meet the K/A?; 2) Do we require operators to know APPs from memory?

8/12/15 Reviewed by Fleet NRC Exam Writer, with following comment, 1) Question may better match the K/A if conditions are in Mode 5 with a loss of offsite power, and a 2 by 2 is created contrasting IB 3 & IB 4 and if the WILL / WILL NOT be energized.

9/11/15 discussed RNP site specific equipment and the inability to meet the KA based on the lack of alarms and indications in the control room with the NRC Chief Examiner. The NRC Chief Examiner randomly selected APE057 AA2.15 as a replacement KA.

Thursday, December 03, 2015 Page 38 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination New question developed.

QUESTION 14 14 B 11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 39 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE058 AA1.02 - Loss of DC Power QUESTION 15 15 A Ability to operate and / or monitor the following as they apply to the Loss of DC Power: (CFR 41.7 / 45.5 / 45.6)

Static inverter dc input breaker, frequency meter, ac output breaker, and ground fault detector .............................

Given the following:

The plant is at 100% power

A loss of MCC-A (DC Bus A) occurs Which ONE of the following correctly completes the statements below?

Instrument Bus 2 and 7 will ____(1)____.

Subsequently, DC Bus A is restored, and the crew is restoring Inverter A.

IAW OP-601, DC SUPPLY SYSTEM, to start up Inverter A, the operator will close the

____(2)____.

A. (1) de-energize (2) DC Input Breaker first, then the AC Output breaker B. (1) de-energize (2) AC Output breaker first, then the DC Input Breaker C. (1) remain energized (2) DC Input Breaker first, then the AC Output breaker D. (1) remain energized (2) AC Output breaker first, then the DC Input Breaker Thursday, December 03, 2015 Page 40 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 15 15 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to ST-016 (p31; Rev 4), Instrument Bus 2 is powered from the "A" Battery Bus through a 7.5KVA inverter (Inverter A). The inverter converts a 125Vdc supply to a 120Vac, single-Ø, 60 HZ output and provides a very reliable AC power source for the instrument Bus. According to OP-601 (p22-23; Rev 58) and Figures 36 and 50 of ST-016 (p32 and 44; Rev 4) the Inverter DC Input Breaker and AC Output Breaker are manually operated breakers with no automatic functions, and the DC input breaker is closed first, then the AC output breaker.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the Inverter A Breakers are closed DC breaker first, then AC breaker second. This is plausible because the operator may incorrectly believe that these breakers are closed in the reverse order.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the inverter supplies 60 Hz AC power. This is plausible because the rod control system uses 58.3 Hz as the frequency for that system.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and D.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to monitor, by identifying the expected Inverter DC input and AC output breaker position, when the DC power supply is lost to the Inverter.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall bits of information, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided ST-016 (p31-32 & 44; Rev 4)

OP-601 (p22-23; Rev 58)

EPP-26 Background Document (p3 of 11; Rev 16) 10 CFR 55.41.7 APE058 AA1.02 - Loss of DC Power Ability to operate and / or monitor the following as they apply to the Loss of DC Power: (CFR 41.7 / 45.5 / 45.6)

Static inverter dc input breaker, frequency meter, ac output breaker, and ground fault detector .............................

Remarks/Status 7/29/15 New question developed.

8/5/15 Reviewed by Exam Supervisor, SAT 8/12/15 Reviewed by Fleet NRC Exam Writer, with following editorial comments, 1) Second bullet is diagnosis, considering replacing with a breaker tripped open, as that would be more appropriate and the operate then is required to diagnose.; 2)Consider a question around the inverter indications keying on the "monitor" part of the K/A.

11/20/15 rewrote question based on input from operations on a different direction we could go for "operate/monitor" of this system.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 41 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 15 15 A Thursday, December 03, 2015 Page 42 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination WE04 2.4.46 - LOCA Outside Containment QUESTION 16 16 C WE04 GENERIC Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

Given the following:

The plant was at 100% power when a reactor trip and SI occurred

The crew has entered EOP-E-1, LOSS OF REACTOR OR SECONDARY COOLANT

The crew suspects a LOCA outside containment on RHR-754A, RHR PUMP A DISCHARGE Consider the following RTGB Annunciators:

1. APP-001-E4, RHR PIT A HI LEVEL

2. APP-036-H1, WDBRP TROUBLE

3. APP-036-D8, PROCESS MONITOR HI RAD IAW EOP-E-1, which ONE of the following identifies the additional RTGB Annunciators that would be ILLUMINATED to confirm the suspected location of this LOCA?

A. 1, ONLY B. 1 and 3, ONLY C. 2 and 3, ONLY D. 2, ONLY Thursday, December 03, 2015 Page 43 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 16 16 C General Discussion Answer A Discussion Incorrect. This is incorrect because APP-001-E4, RHR PIT A HI LEVEL will not alarm (The leak location is in pipe alley, not the RHR pit).

This is plausible because the operator may incorrectly believe that the discharge valve of the RHR pump is located in the RHR pump PIT where the pumps suction valve is located.

Answer B Discussion Incorrect. This is incorrect because APP-001-E4, RHR PIT A HI LEVEL, will NOT alarm because the RHR-754A discharge valve of the pump is in pipe alley. This is plausible because the operator may incorrectly believe that the discharge valve of the RHR pump is located in the RHR pump PIT where the pumps suction valve is located.

Answer C Discussion Correct. According to Drawing 5379-1484 Sheet 1 of 1 (Sheet 1 of 1; Rev 42) the RHR-754A, RHR PUMP "A" DISCHARGE valve is in pipe alley, with a remote operator to allow operation from the Gas Analyzer room. When the leakage occurs from this location, the floor drains will direct water to the Aux bldg. "A" Sump, According to Drawing 5379-0920 sheet 1 of 8 (Sheet 1 of 8; Rev 49). According to EOP-E-0 (p 20; Rev 6) Step 20 and EOP-E-1 (p 17; Rev 6) Step 16, the following conditions in the Aux building are checked and evaluated to determine if a LOCA outside containment is occurring;

- R-3, PASS PANEL AREA,

- R-4, CHARGING PUMP ROOM,

- R-6 SAMPLING ROOM,

- RI-14C, Plant Effluent NG-LO,

- LI-615A, RHR PIT "A" LEVEL INDICATOR,

- LI-615B, RHR PIT "B" LEVEL INDICATOR,

- Aux Bldg Sump Tank "A" and

- Aux Bldg Sump Tank "B".

Therefore, with the leak in the Aux Bldg, in pipe alley, APP-036-H1 will alarm due to APP-011-13, Sump Tank "A" HI-LO LEVEL on the WASTE DISPOSAL BORON RECYCLE PANEL (LEFT PANEL). APP-036-D8 PROCESS MONITOR HI RAD would be in alarm if any of those indicated Area Monitors were affected by the leak based on size and its location.

Answer D Discussion Incorrect. This is incorrect because APP-036-D8, PROCESS MONITOR HI RAD would alarm if any of the area monitors listed in EOP-E-1 reach their alarm setpoints. This is plausible because the operator may not believe that the LOCA would cause area radiation monitors to alarm, if they believed that the inventory loss was not of high enough activity to cause the alarms. (i.e. if the water out the break was RWST and not RCS as would be the case in a LOCA outside containment)

Basis for meeting the KA The KA is matched because the operator must demonstrate the Ability to verify that the alarms are consistent with the plant conditions during a LOCA Outside Containment. This is accomplished by providing the operator with a set of plant conditions reflecting a LOCA Outside Containment and also providing a group of potential alarms to consider. Then, the operator is required to choose which alarms will occur, confirming their ability verify that the alarms are consistent with the plant conditions during a LOCA Outside Containment.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information, and use this information to predict an outcome in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Drawing 5379-1484 Sheet 1 of 1 (Sheet 1 of 1; Rev 42)

Drawing 5379-0920 Sheet 1 of 8 (Sheet 1 of 8; Rev 49)

APP-001-D7 (p3, & 40 of 62; Rev 60)

APP-036-D7, D8 & H1 (p37-39 & 65 of 98; Rev 88)

APP-011-13 (p3 & 15; Rev 18)

EOP-E-0 (p 20; Rev 6) Step 20 EOP-E-1 (p 17; Rev 6) Step 16 10 CFR 55.41.4 Thursday, December 03, 2015 Page 44 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination WE04 2.4.46 - LOCA Outside Containment QUESTION 16 16 C WE04 GENERIC Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

Remarks/Status 8/11/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Reviewed by Operations SRO, with Editorial comment, recommend changing "intersystem LOCA through" to just "leakage past" in the stem to avoid confusion. Additionally the crew would be in EOP-E-1 and not be able to get to ECA-1.2 based on leak loacation.

9/10/15 Exam team incorporated recommended changes.

9/30/15 changed question to new question.

11/5/15 discussed with NRC chief examiner as one of the early submittal items. He stated the second part question appears to be asking procedure transition criteria, which is SRO knowledge. Question is Unsatisfactory due to license level mismatch. Also, I am not sure that a leak in this location would always cause an APP-036-D7 alarm. The FSAR states that an alarm in this location would most likely be VCT or RWST water because there are so many check valves downstream. The R-4 alarm setpoint is 50 millirem per hour. Did or can you validate on the simulator? Why ask or mention D8, since it is given in all four answer choices?

11/9/15 exam team reevaluated the alarms and plausibility for a LOCA outside containment in the location that was presented in the stem. The team determined that there were better alarm indications that could indicate the LOCA outside containment that a question could be written too.

Changed question to reflect.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 45 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination WE05 EK3.4 - Loss of Secondary Heat Sink QUESTION 17 17 B Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink)

(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:

The reactor was tripped from 100% power due to a loss of MFW

A spurious SI has actuated

RCS Pressure - 2300 psig

All S/G pressures - 1050 psig

S/G WIDE RANGE levels are as follows:

WIDE RANGE: A - 10%,

WIDE RANGE: B - 15%,

WIDE RANGE: C - 9%

AFW flow is indicating 0 gpm

A RED condition exists on CSF-3, Heat Sink

EOP-E-0, REACTOR TRIP OR SAFETY INJECTION, Immediate Actions are complete and have just been verified Which ONE (1) of the following correctly completes the statement below?

The crew should ____(1)____ because ____(2)____.

A. (1) remain in EOP-E-0 until directed to transition out (2) the B S/G is still a viable Heat Sink B. (1) remain in EOP-E-0 until directed to transition out (2) rules of usage require continuing in EOP-E-0 C. (1) immediately go to FRP-H.1, RESPONSE TO A LOSS OF SECONDARY HEAT SINK (2) RCS Bleed and Feed is required D. (1) immediately go to FRP-H.1, RESPONSE TO A LOSS OF SECONDARY HEAT SINK (2) RCS Bleed and Feed can still be avoided Thursday, December 03, 2015 Page 46 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 17 17 B General Discussion Answer A Discussion Incorrect. This is incorrect because the "B" Steam Generator is NOT a viable Heat Sink. This is plausible because it is the only Steam Generator that is above the Feed and Bleed threshold of less than13%, and the operator may incorrectly believe that it substantiates remaining in EOP-E-0.

Answer B Discussion Correct. According to OMM-22 (p13 & 16 of 56; Rev 45) Step 5.2.c, EOP-E-0 is the normal entry point for the EOPs, and according to Step 52.t, Each EOP ends with either a specific transition to another EOP, if further Operator guidance is required, or with the plant being maintained in a steady-state condition. Under the stated conditions E-0 is still in progress, and must be completed until a transition is directed. Note that a transition to FRP-H.1 will be made if steps 5-8 of EOP-E-0 are ineffective at restoring a Secondary Heat Sink. On the other hand, if Steps 5-8 are able to establish at least 300 gpm of feedwater flow to the Steam Generators, then the crew will continue in EOP-E-0 and not make the transition to FRP-H.1.

Answer C Discussion Incorrect. This is incorrect because the EOP rules of usage do not permit an immediate transition to FRP-H.1 under the stated plant conditions (i.e. EOP-E-0 complete through Step 4 with an SI actuated). This is plausible because if the operator were in FRP-H.1, an RCS Bleed and Feed would be required.

Answer D Discussion Incorrect. This is incorrect because the EOP rules of usage do not permit an immediate transition to FRP-H.1 under the stated plant conditions (i.e. EOP-E-0 complete through Step 4 with an SI actuated). This is plausible because the operator may incorrectly believe that Bleed and Feed criteria have not been met, but are close, and could be avoided if the operator were in FRP-H.1.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the reasons for the EOP transitions (An RO/SRO function) as they apply to the Loss of Secondary Heat Sink and the assurance that procedures are adhered to and the limitations in the facilities license and amendments are not violated. The License requires adherence to the Quality Assurance Program which requires that the plant be operated under all conditions by approved procedures.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must apply known rules of EOP usage, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED FR-H.1-2 006 Development References Student References Provided OMM-22 (p13 & 16 of 56; Rev 45) 10 CFR 55.41.10 WE05 EK3.4 - Loss of Secondary Heat Sink Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink)

(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.

Remarks/Status 8/10/15 Modified question FR-H.1-2 8/25/15 During review of SO#78 it was noted that there was a similar RO question. Both RO#17 and SO#78 need to be review to ensure no overlap and double jeopardy.

9/8/15 Reviewed by Exam Supervisor, SAT Thursday, December 03, 2015 Page 47 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 17 17 B 9/9/15 Exam team reviewed both questions for the concern above, and determined that the information that the questions are testing is NOT overlap, and should not be double jeopardy.

11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 48 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination WE11 EK1.1 - Loss of Emergency Coolant Recirculation QUESTION 18 18 B Knowledge of the operational implications of the following concepts as they apply to the (Loss of Emergency Coolant Recirculation)

(CFR: 41.8 / 41.10 / 45.3)

Components, capacity, and function of emergency systems.

Given the following:

The plant has tripped from 100% power due to a LOCA

The crew is operating in EOP-ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION

RWST level is at 16%

RWST Makeup has been established IAW Supplement P, EMERGENCY MAKEUP TO THE RWST

The only ECCS Pump drawing on the RWST is the A SI Pump and pump flow has been adjusted to 150 gpm Which ONE of the following correctly completes the statements below?

1) The EARLIEST that the A SI Pump must be stopped is when the RWST Level lowers to ____(1)____.

2) Assuming these conditions remain the same, it is expected that over the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the RWST level will ____(2)____.

A. (1) 9%

(2) rise B. (1) 9%

(2) lower C. (1) 13%

(2) rise D. (1) 13%

(2) lower Thursday, December 03, 2015 Page 49 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 18 18 B General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the RWST will not rise over the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This is plausible because the operator may incorrectly believe that the Supplement P flow rate is higher than it actually is (The Makeup System is limited in capacity only because of the need to keep the RWST Boron concentration higher).

Answer B Discussion Correct. 1st part correct, 2nd part correct. According to the ECA-1.1 (Foldout Page Item 2 and p1; Rev 2), which is in effect from the start of the procedure, if the RWST level lowers to 9% the SI Pump will need to be stopped. Additionally, the body of the procedure (p4 & 23; Rev 2) directs the operator to check RWST level less than 9%, and if so, directs that all pumps taking suction from the RWST be stopped. According to Supplement P of EOP-Supplements (p64 of 81; Rev 2) Step 11& 12, the RWST flowrate will be 84.25 gpm, which is 40 gpm lower than the RWST outflow. Consequently, the RWST level will lower over the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by about 9600 gallons.

Answer C Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and D.

Answer D Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the earliest that the A SI Pump must be stopped is 9%. This is plausible because According to the ECA-1.1 Foldout Page Item #3 if the CST lowers to 13% it must be re-filled (The operator may confuse the Tank Levels).

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the operational implications of components (RWST Low-Low Setpoint) and capacity (Makeup System when delivering to the RWST) of emergency systems as they apply to the Loss of Emergency Coolant Recirculation.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must compare one flowrate to another and predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided ECA-1.1 (Foldout Page Item 2, p1,4&23; Rev 2)

EOP-Supplements (p64 of 81; Rev 2) 10 CFR 55.41.7 WE11 EK1.1 - Loss of Emergency Coolant Recirculation Knowledge of the operational implications of the following concepts as they apply to the (Loss of Emergency Coolant Recirculation)

(CFR: 41.8 / 41.10 / 45.3)

Components, capacity, and function of emergency systems.

Remarks/Status 8/12/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 11/21/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 50 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE005 AK2.02 - Inoperable/Stuck Control Rod QUESTION 19 19 C Knowledge of the interrelations between the Inoperable / Stuck Control Rod and the following: (CFR 41.7 / 45.7)

Breakers, relays, disconnects, and control room switches ........................................................

Given the following:

The plant is at 50% power

During the performance of OST-011, ROD CLUSTER CONTROL EXERCISE & ROD POSITION INDICATION, ONE Control Bank B rod is 17 steps below the rest of the bank

The crew is performing the steps to realign the rod IAW AOP-001, MALFUNCTION OF REACTOR CONTROL SYSTEM Which ONE of the following correctly completes the statements below?

1) IAW AOP-001, OPEN the Control Bank B lift coil switch(es) for ____(1)____.

2) Realign the rod using the ____(2)____ position of the rod bank selector switch.

A. (1) the misaligned rod ONLY (2) CB B B. (1) the misaligned rod ONLY (2) M (MANUAL)

C. (1) all but the misaligned rod (2) CB B D. (1) all but the misaligned rod (2) M (MANUAL)

Thursday, December 03, 2015 Page 51 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 19 19 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect becauseAOP-001 directs the operator to open the lift disconnect switches for all the rods in the affected bank except for the rod that is to be recovered. This is plausible because if it were desired not to move the affected rod but move the other rods in the affected bank, this action would then be correct.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and D.

Answer C Discussion Correct. 1st part correct, 2nd part correct. According to AOP-001 (p 50 & 52; Rev 33) Steps 40 & 42.b, Place Lift Coil Disconnect Switches for all rods in the affected bank, EXCEPT the misaligned rod, in the OFF position. Select the affected bank with the ROD BANK SELECTOR Switch.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because AOP-001 directs placing the Rod Bank Selector Switch in the affected bank position. This is plausible because normally Control Bank "B" rods are moved in the M (MANUAL) position during normal system operation.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the interrelationship between lift coil disconnect switches and rod bank selector switch positions required for recovery of a misaligned control rod.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall from memory the required lift disconnect switch and bank selector switch position for recovering a Control Bank "B" misaligned rod IAW AOP-001, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 005 AK2.02 1 Development References Student References Provided AOP-001 Rev. 33 10 CFR 55.41.10 APE005 AK2.02 - Inoperable/Stuck Control Rod Knowledge of the interrelations between the Inoperable / Stuck Control Rod and the following: (CFR 41.7 / 45.7)

Breakers, relays, disconnects, and control room switches ........................................................

Remarks/Status 8/12/15 Used Bank Question 005 AK2.02 1 9/8/15 Reviewed by Exam Supervisor, SAT 11/21/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 52 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE028 AK1.01 - Pressurizer (PZR) Level Control Malfunction QUESTION 20 20 A Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions: (CFR 41.8 /

41.10 / 45.3)

PZR reference leak abnormalities ..................................

Given the following:

The plant is at 100% power

Pressurizer level transmitter LT-459 is selected for control

The reference leg for LT-459 develops a slow leak LT-459, CH I PRZR LEVEL LT-460, CH II PRZR LEVEL Which ONE of the following describes the correct PZR LEVEL instrument response?

LT-459 __LT-460__

A. Increases Decreases B. Decreases Remains the same C. Increases Remains the same D. Decreases Increases Thursday, December 03, 2015 Page 53 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 20 20 A General Discussion Answer A Discussion Correct. According to RNP-LOF0020R, Sensors and Detectors, Rev0, Pg 101, a reference leg leak will cause LT-459 to see a rising level and LT-460 will lower due to the Charging Pump slowing down due to input from LT-459.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See C & D.

Answer C Discussion Incorrect. This is incorrect because LT-460 would not remain the same. Plausible because response for LT-459 is correct and operator must understand system response for controlling level channel.

Answer D Discussion Incorrect. This is incorrect because the LT-459 indication increases for the given failure. Plausible if the operator does not understand reference leg leaks verses variable leg leaks. The LT-460 response would be correct if LT-459 did lower. Charging pump speed would rise therefore PZR level would rise.

Basis for meeting the KA The KA is matched because the operator must demonstrate that they understand the effect the PZR reference leak has on both the transmitter that the leak is on, and the plant due to the inventory loss.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must determine which transmitter is effected and determine the effects that both that channel and the controlling channel will have on the plant in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 028 AK1.01 1 Development References Student References Provided RNP-LOF0020R, Sensors and Detectors, Rev0, Pg 101 10 CFR 55.41.7 APE028 AK1.01 - Pressurizer (PZR) Level Control Malfunction Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions: (CFR 41.8 /

41.10 / 45.3)

PZR reference leak abnormalities ..................................

Remarks/Status 7/17/15 Used Bank question 028 AK1.01 1 7/23/15 Reviewed by Exam Supervisor, with Editorial Enhancement comment, 1) Add Noun name for LT-459 to Stem.

9/11/15 Exam team made change as proposed.

11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 54 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE032 AA2.02 - Loss of Source Range Nuclear Instrumentation QUESTION 21 21 B Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR: 43.5 / 45.13)

Expected change in source range count rate when rods are moved ........

Given the following:

A reactor startup is being commenced

Source Range count rate (CR0) prior to rod withdrawal is as follows:

N31: 3.0E+001 cps

N32: 3.0E+001 cps The crew has completed the MINIMUM number of doublings to allow taking the reactor critical IAW GP-003, NORMAL PLANT STARTUP FROM HOT SHUTDOWN TO CRITICAL.

The OATC is getting ready to pull to criticality when the following indications are observed:

N31: 2.4E+002 cps

N32: 1.2E+002 cps Which ONE of the following correctly completes the statement below?

1) IAW GP-003, Source Range Channel ____(1)____ is indicating AS EXPECTED.

2) When the Source Range Channel with the unexpected reading is removed from service per the appropriate OWP, the reactor startup ____(2)____ continue.

A. (1) N31 (2) may B. (1) N31 (2) may not C. (1) N32 (2) may D. (1) N32 (2) may not Thursday, December 03, 2015 Page 55 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 21 21 B General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because startup cannot continue. This is plausible because the operator may incorrectly believe that the startup may continue with one SR Channel OOS.

Answer B Discussion Correct. 1st part correct, 2nd part correct. According to GP-003 (p44; Rev 105) a Note prior to Step 8.2.28 states that the approach to criticality should take approximately four doublings of the indicated Reference Count Rate (CR0) under ideal conditions. The target count rate is intended to serve as a known stable reactivity state suitable for data taking and criticality predictions. In practice, the operator pulls rods and stabilizes at the third doubling, and then pulls to criticality. Based on this, the Source Range countrate should be 240 cps (2.4E002) when at the third doubling [1st doubling 60 cps, 2nd doubling 120 cps, 3rd doubling 240 cps]. With N32 reading half of the expected value at criticality, N32 is reading unexpectedly. According to GP-003 (p13; Rev 105) Step 5.4.8, The following requirements apply to the Source Range Nuclear Instruments when in MODE 2 below P-6: (ITS Table 3.3.1-1 item 4). IF one Source Range channel becomes inoperable, THEN immediately suspend operations involving positive reactivity additions, Also reference ITS 3.3.1, Required Action I.1.

Answer C Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and D.

Answer D Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because Channel N31 is reading as expected. This is plausible because the operator may incorrectly believe that the expected countrate change at the time of criticality takes place in three doublings, rather than 4.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to interpret indication for a failed SR NI channel during a reactor startup and determine the required action for this condition.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and use this information to diagnose a failed SR NI channel, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided GP-003 (p13 and 44; Rev 105) 10 CFR 55.41.7 APE032 AA2.02 - Loss of Source Range Nuclear Instrumentation Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR: 43.5 / 45.13)

Expected change in source range count rate when rods are moved ........

Remarks/Status 8/18/15 Modified Bank question NIS-012 3 9/8/15 Reviewed by Exam Supervisor, with editorial comment, 1) Why would it fail to 3 cps? Need to make the reading after failure more realistic.

9/11/15 Exam team discussed with exam supervisor and changed failure reading to be consistent with the bottom of the scale.

11/17/15 Fleet review determined that the question as written doesn't really hit the KA, and is a difficult KA to write a question to. The exam team wrote a new question to better meet the KA.

11/21/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 56 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 21 21 B Thursday, December 03, 2015 Page 57 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE033 AK3.01 - Loss of Intermediate Range Nuclear Instrumentation QUESTION 22 22 C Knowledge of the reasons for the following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR 41.5,41.10 / 45.6 / 45.13)

Termination of startup following loss of intermediate range instrumentation .............................................

Given the following:

A Reactor startup is in progress

Based on the following indications, the startup has been placed on hold:

-9

N-35 = 5 x 10 amps

-11

N-36 = 1 x 10 amps

N-31 and N-32 are 120 cps Which ONE of the following is the reason for placing the startup on hold?

A. N-35 is over compensated B. N-36 is over compensated C. N-35 is under compensated D. N-36 is under compensated Thursday, December 03, 2015 Page 58 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 22 22 C General Discussion Answer A Discussion Incorrect. This is incorrect because overcompensation of the IR signal causes a reduction of the IR signal output. However, N-35 is reading high as compared to SR NIs. This is plausible because N-35 is the IR NI channel that is not in alignment with the SR NI channel indications.

Answer B Discussion Incorrect. This is incorrect because N-36 is indicating correctly as compared to SR NI indications. This is plausible because N-36 is reading lower than N-35 and therefore if SR NI indication was reading higher and in alignment with N-35, then this would be the correct answer.

Answer C Discussion Correct. According to ST-010 (p 24; Rev 3) Overcompensation of the IR signal can occur if the compensating voltage is set too high causing a reduction of the IR signal out. Undercompensation of the IR signal occurs when the compensation voltage is set too low causing an increase of the IR signal out. Overcompensation causes the IR output signal to be nonconservative. According to LOF0020R (p 153; Rev.0) Figure 49 shows Source Range and Intermediate Range indications such that 120 CPS SR indication is approximately 1 x 10-11 amps IR indication.

Answer D Discussion Incorrect. This is incorrect because N-36 is indicating correctly as compared to SR NI indications. This is plausible because undercompensation of an IR NI is the cause of these conditions and N-36 can fail by this method.

Basis for meeting the KA The KA is matched because the startup has been terminated due to a malfunction (i.e. in effect a loss) of an intermediate range instrument. The applicant must have knowledge of the reasons for the "loss" of the intermediate range instrument to determine the correct response.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because it requires more than one mental step. The applicant must first recall from memory the relationship of SR indication and IR indication at various power levels. The applicant must then determine which of the IR channels in malfunctioning from the indications provided and the reason for the malfunctioning indication.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK NIS-010 10 Development References Student References Provided ST-010 Rev. 3 LOF0020R Rev. 0 GP-003 Rev.105 10 CFR 55.41.7 APE033 AK3.01 - Loss of Intermediate Range Nuclear Instrumentation Knowledge of the reasons for the following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR 41.5,41.10 / 45.6 / 45.13)

Termination of startup following loss of intermediate range instrumentation .............................................

Remarks/Status 7/21/15 Used Bank question NIS-010 10 8/5/15 Reviewed by Exam Supervisor, UNSAT with following comment, 1) 50 cps on SRNI does not align to 5 X 10 -11amps. Need to change this to be more realistic since this could result in 2 correct answers.

9/8/15 Operations SRO reviewed question, concurs that the SRNI and N-36 readings do not align, which could cause the candidates to conclude that both IRNIs are undercompensated.

9/11/15 exam team incorporated change to make the SRNI readings correspond more closely to the properly compensated IRNI reading.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 59 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 22 22 C Thursday, December 03, 2015 Page 60 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE060 AK2.01 - Accidental Gaseous-Waste Release QUESTION 23 23 D Knowledge of the interrelations between the Accidental Gaseous Radwaste Release and the following: (CFR 41.7 / 45.7)

ARM system, including the normal radiation-level indications and the operability status ................................

Given the following:

An accidental gaseous radwaste release is in progress

R-1, Control Room Area Radiation Monitor, alarms

R-1 is indicating 8.62E-2 mR/hr and stable Which ONE of the following correctly completes the statements below?

1) R-1 ____(1)____ operating properly.

2) When R-1 alarms, the CR HVAC will automatically shift to the Emergency ____(2)____

Mode.

A. (1) is (2) Recirculation B. (1) is (2) Pressurization C. (1) is NOT (2) Recirculation D. (1) is NOT (2) Pressurization Thursday, December 03, 2015 Page 61 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 23 23 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. See B & C.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because this radiation level is normal, and no alarm should have occurred. This is plausible because the operator may incorrectly interpret the exponential reading of the instrument, and incorrectly conclude that the Control Room reading is higher than it actually is.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because when R-1 alarms it automatically shifts to the Emergency Pressurization Mode. This is plausible because this is one of three modes of system operation (Normal, Emergency Pressurization, and Emergency Recirculation) and the operator could incorrectly believe that when the alarm occurs, the system shifts to the Emergency Recirculation Mode.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to ST-019 (p12; Rev 4) the R-1 detector range is from .01 mrem/hr to 107 mrem/hour.

The indicated reading is .0862 mrem/hour. According to OST-20 (p23 of 61; Rev 55) the R-1 reading is recorded by the operator every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and would be expected to be known by the operator. Since the alarm has occurred, and radiation levels are normal, the operator will conclude that R-1 is not working properly. According to ST-036 (p 10; Rev 4a) Emergency Pressurization initiates upon High Radiation (i.e.

Alarm occurs) or Safety Injection signal.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the interrelations between the Accidental Gaseous Radwaste Release and R-1 (Control Room Area Rad monitor), including the normal radiation-level indications and the operability status. This is accomplished by creating conditions in which R-1 has alarmed pre-maturely with normal radiation levels indicated, and requiring the operator to identify whether or not the instrument is operating properly.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall bits of information, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided ST-019 (p12; Rev 4)

OST-20 (p23 of 61; Rev 55)

ST-036 (p 10; Rev 4a) 10 CFR 55.41.11 APE060 AK2.01 - Accidental Gaseous-Waste Release Knowledge of the interrelations between the Accidental Gaseous Radwaste Release and the following: (CFR 41.7 / 45.7)

ARM system, including the normal radiation-level indications and the operability status ................................

Remarks/Status 8/18/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, UNSAT, not required to know R-1 setpoint from memory.

9/8/15 Operations SRO reviewed question, UNSAT same comment as above.

9/14/15 edited the first part of the question to make the contrast between the normal RM reading which should be knowledge from memory, and based on a reading of such, with the channel in alarm is it operating correctly?

11/21/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 62 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 23 23 D Thursday, December 03, 2015 Page 63 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE067 AA1.09 - Plant Fire On Site QUESTION 24 24 A Ability to operate and / or monitor the following as they apply to the Plant Fire on Site: (CFR 41.7 / 45.5 / 45.6)

Plant fire zone panel (including detector location) .....................

Given the following:

The plant is operating at 100% power

The following is observed on the Fire Alarm Console: (SEE GIVEN CONDITIONS ON NEXT PAGE)

The Fire Brigade has been alerted

An evacuation of the fire area has been performed Which ONE of the following correctly completes the statements below?

1) It is expected that the automatic water suppression system ____(1)____ actuate for this fire event.

2) Based on this fire event, the entry conditions ____(2)____ met for DSP-001, DEDICATED SHUTDOWN DIAGNOSTIC.

A. (1) will (2) are B. (1) will (2) are NOT C. (1) will NOT (2) are D. (1) will NOT (2) are NOT Thursday, December 03, 2015 Page 64 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 24 24 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to APP-044 (p60 of 398; Rev 28) The automatic Water Suppression System will actuate if the opposite train fire alarm is present. According to DSP-001 (p2 of 26; Rev 14), any fire that has the potential to damage plant components/controls and/or their power/control cables when Tavg is greater than 200°F, will require DSP-001 to be implemented. According to AOP-041 (p7 of 28; Rev 7) Step 8, if this fire alarm is in the Auxiliary Building, then DSP-001 will need to be implemented while AOP-041 is being carried out.

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because DSP-001 must be implemented. This is plausible because there are some areas of the plant for which this procedure does not need to be implemented, such as the Fuel or Radwaste buildings.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the automatic Water Suppression System will actuate. This is plausible because there are conditions (i.e. only one train actuated) under which the auto system would not be expected to actuate.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate Ability to monitor the Fire Alarm Console (FAC) as it applies to the Plant Fire on Site, including detector location. This is accomplished by requiring that the operator observe a Screen Shot of the FAC Screen showing alarms in specific locations within the plant, and then answering questions relating to what has been observed. The 2nd part of the question in particular demonstrates their ability to associate a detector location with the need to implement a specific procedure (i.e. DSP-001).

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must interpret a Screen Shot of the FAC Screen showing alarms in specific locations within the plant, and then answering questions relating to what has been observed, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided APP-044 (p60 of 398; Rev 28)

AOP-041 (p7 of 28; Rev 7)

DSP-001 (p2 of 26; Rev 14) 10 CFR 55.41.4 APE067 AA1.09 - Plant Fire On Site Ability to operate and / or monitor the following as they apply to the Plant Fire on Site: (CFR 41.7 / 45.5 / 45.6)

Plant fire zone panel (including detector location) .....................

Remarks/Status 8/11/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 9/17/15 Exam team moved the picture of the FAC to a reference page, to fit the entire question and answer choices on the same page.

11/21/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 65 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 24 24 A Thursday, December 03, 2015 Page 66 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE068 AK3.12 - Control Room Evacuation QUESTION 25 25 C Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation: (CFR 41.5,41.10 / 45.6 / 45.13)

Required sequence of actions for emergency evacuation of control room .................................................

Given the following:

With the plant at 100%, a confirmed bomb threat has been received

The caller specified that the bomb was located in the Control Room

The following timeline of events is recorded:

2110 The Control Room was evacuated after all appropriate actions were taken 2145 A Control operator in the Charging Pump Room informs the SM that:

N51 reads 3.0E5 cps and is slowly lowering

Pressurizer pressure is 2225 psig

Pressurizer level is 35%

The A Charging Pump is operating Which ONE of the following correctly completes the statements below?

IAW Attachment 1 of AOP-004, CONTROL ROOM INACCESSIBILITY, the operator will raise Pressurizer level to at LEAST greater than ____(1)____, and then stop the running Charging Pump.

The basis for this action is to ____(2)____.

A. (1) 60%

(2) prevent exceeding the Charging Pump starting duty limitations B. (1) 60%

(2) ensure that the reactor is subcritical C. (1) 86%

(2) ensure that the reactor is subcritical D. (1) 86%

(2) prevent exceeding the Charging Pump starting duty limitations Thursday, December 03, 2015 Page 67 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 25 25 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. See B & D.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the plant conditions require raising level to 86%. This is plausible because the procedure has the operator check the SG levels at 60% in Attachment 3.

Answer C Discussion Correct. 1st part correct, 2nd part correct. According to AOP-004 (p11-12 of 32; Rev 25) the operator will be directed to establish 1 Charging Pump ONLY, and then be directed to check the Source Range < 105 cps. Since Source Range cps are above this level, the operator will go to the RNO, and proceed to Step 17; which will direct the operator to raise the quantity of borated water in the RCS by raising Charging Pump speed and stop the running charging pump after PRZR level is > 86%.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the reason the procedure allows PRZR level to be adjusted to >86% to address the reactivity issue. This is plausible because on subsequent steps where the operator is controlling PZR level, they do so by starting and stopping of the charging pump. There is a note that reminds the operator of the starting duty limitations on the charging pumps prior to the continuous action step, however this is not the basis for the high end of the level band, and the procedure in fact has the operator select a different charging pump on subsequent starts to prevent challenging starting duty limitations. The operator could confuse this note as a basis for the high end of the level band.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the reasons for the required sequence of actions for pressurizer level control in Attachment 1 of AOP-004 based on plant conditions.

Basis for Hi Cog The question is at the Comprehensive/Analysis cognitive level because the operator must recall a step in a sequence of steps or the reason for the step in order to answer the question correctly Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided AOP-004 (p11-12 of 32; Rev 25) 10 CFR 55.41.10 APE068 AK3.12 - Control Room Evacuation Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation: (CFR 41.5,41.10 / 45.6 / 45.13)

Required sequence of actions for emergency evacuation of control room .................................................

Remarks/Status 8/4/15 New question developed.

8/11/15 Reviewed by Exam Supervisor, Editorial Enhancement commments,1) Need to specify location of fire in the stem. If the fire is in the control room, would they be in DSP-002 instead of AOP-004? Would this change the correct answer?

9/11/15 Exam team discussed with Exam Supervisor and changed first bullet into 2 bullets to avoid procedure use direction and possibility for a second answer choice to be correct.

9/30/15 new question developed.

11/17/15 Fleet review indicated that there were plausibility issues with the answer that were to avoid a water solid condition, as if that were the case, then why would we be filling the PZR to "AT LEAST" or a "MINIMUM" Found another piece in the procedure to contrast that with.

12/1/15 question edited to address concern AOM-shift had with possible high-miss.

Thursday, December 03, 2015 Page 68 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 25 25 C 12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 69 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination APE076 AA2.04 - High Reactor Coolant Activity QUESTION 26 26 A Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity: (CFR: 43.5 / 45.13)

Process effluent radiation chart recorder........................................

Given the following:

The plant is at 100% power

The R-9, Letdown Line Area, indication on Radiation Monitoring System Recorder RR-1 is slowly rising Which ONE of the following correctly completes the statements below?

1) The cause of the RMS alarm is ____(1)____.

2) If the trend continues, it is expected that ____(2)____.

A. (1) High RCS Activity (2) RR-1 will alarm BEFORE the R-9 ratemeter B. (1) A letdown line leak in Containment (2) RR-1 will alarm BEFORE the R-9 ratemeter C. (1) High RCS Activity (2) the R-9 ratemeter will alarm BEFORE RR-1 D. (1) A letdown line leak in Containment (2) the R-9 ratemeter will alarm BEFORE RR-1 Thursday, December 03, 2015 Page 70 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 26 26 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to OMM-14 (p9 of 21; Rev 51) Step 8.3.3, the basis for R-9 (Failed Fuel Monitor) setpoint is that the setpoint is set low enough to detect a small leak in the fuel rods in time to begin corrective action and clean-up. According to AOP-005-BD (p15 of 29; Rev 32), states that the reason that E&C is called when R-9 is in alarm is because RCS activity may have been the cause of the R-9 alarm condition, and that the step provides instruction for sampling the RCS for activity. According to OMM-14 (p6 of 21; Rev 51), P&L 5.1, the RR-1 alarm setpoints are chosen at 1.5 times the base-line value to allow RR-1 RMS channel to alarm before the respective ratemeter.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because R-9 is an area monitor that is located outside Containment. This is plausible because AOP-005 identifies R-9 as the Letdown Line Monitor, and the operator may incorrectly believe that it is associated with the letdown line in Containment.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the R-9 ratementer setpoint is 3000 mr/hr corresponding which is low enough to detect a small leak in the fuel rods in time to begin corrective action and clean-up, and the RR-1 alarm setpoint is set 1.5 times above the baseline value to ensure that it alarms prior to the ratemeter. This is plausible because the operator may not know the basis for the RR-1 Alert level setpoints.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to determine and interpret the process effluent radiation chart recorder as it applies to High Reactor Coolant Activity. This is accomplished by requiring the operator to analyze a rising trend in the recorder for R-9 (An Area Monitor, monitoring a process [letdown flow radioactivity level]) and identify the cause, and then identify which will provide the first indication of the trend (Which is the Recorder).

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must demonstrate an understanding of the significance of a rising trend on R-9, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED 076 AA2.01 001 Development References Student References Provided OMM-14 (p6 and 9 of 21; Rev 51)

AOP-005-BD (p15 of 29; Rev 32) 10 CFR 55.41.11 APE076 AA2.04 - High Reactor Coolant Activity Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity: (CFR: 43.5 / 45.13)

Process effluent radiation chart recorder........................................

Remarks/Status 8/4/15 Modified Bank Question 076 AA2.01 001 8/11/15 Reviewed by Exam Supervisor, Editorial Enhancement comment, 1) Eliminated 2 Distractors due to a "True/False" condition, prior to even reading the stem.

9/30/15 removed the term sonalert, as operators do not commonly use this term.

11/19/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 71 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 26 26 A Thursday, December 03, 2015 Page 72 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination WE09 2.1.20 - Natural Circulation Operations QUESTION 27 27 C WE09 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

Given the following:

EOP-ECA-0.1. LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED has been implemented

The crew is checking for Natural Circulation with the following conditions:

RCS subcooling based on CETs are 33 °F and stable

CETC are 555°F and stable

S/G pressures are stable at 900 psig

RCS Hot Leg temperatures are 555 °F and stable

RCS Cold Leg temperatures are 545 °F and slowly rising

S/G NR levels are 21% and stable Which ONE of the following correctly completes the statements below?

Based on the indications above, Natural Circulation flow ____(1)____ been established.

IAW EOP-ECA-0.1, the Operators will ____(2)____.

A. (1) has (2) maintain current steam dump rate B. (1) has (2) maintain current S/G levels C. (1) has NOT (2) raise the rate of dumping steam D. (1) has NOT (2) raise S/G levels Thursday, December 03, 2015 Page 73 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 27 27 C General Discussion The question is significantly modified because the given information for Steam Generator Pressure and RCS Cold Leg temperature have changed along with providing values for RCS Hot Leg temperature and S/G levels. In addition, the question has changed to a two part question, thus changing the required answers/distractors. According to NUREG-1021, ES-401 Section D.2.f, paragraph 1, bullet 4; to be considered a significantly modified question, at least one pertinent condition in the stem and at least one distractor must be changed from the original bank question. Changing the conditions in the stem such that one of the three distractors in the original question becomes the correct answer would also be considered a significant modification Answer A Discussion Incorrect. 1st part wrong, 2nd wrong. This is incorrect because RCS Cold Leg temperatures are not at saturation temperature for S/G pressure.

This is plausible because if RCS Cold Leg Temperatures were at saturation temperature for the S/G pressure then the rate of dumping steam would be maintained at the current rate.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because RCS Cold Leg temperatures are not at saturation temperature for S/G pressure. This is plausible because if RCS Cold Leg Temperatures were at saturation temperature for the S/G pressure then maintaining current S/Gs levels is an appropriate procedure action.

Answer C Discussion Correct. According to EOP-ECA-0.1 (p 18; Rev 0) Step 17, requires dumping more steam from intact S/Gs if any of the following criteria are not met:

CHECK RCS Subcooling based on Core Exit T/Cs - GREATER THAN 18 F [37 F]

CHECK S/G pressures - STABLE OR LOWERING

CHECK RCS Hot Leg temperatures - STABLE OR LOWERING

CHECK Core Exit T/Cs - STABLE OR LOWERING

CHECK RCS Cold Leg temperatures - AT SATURATION TEMPERATURE FOR S/G PRESSURE Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because RCS Cold Leg temperatures are not at saturation temperature for S/G pressure, therefore the procedure required action is to dump more steam from intact S/G(s) to establish Natural Circulation. This is plausible because S/G levels are required to be maintained during the cooldown of the RCS in accordance with EOP-ECA-0.1 Step 9.b.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to interpret plant conditions for establishing Natural Circulation and demonstrate knowledge of the required procedure actions to establish Natural Circulation when conditions are not met.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must analyze plant conditions and determine if Natural Circulation has been established and then determine the required procedure action(s) based upon this analysis, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED 055 EK 1.02 1 Development References Student References Provided EOP-ECA-0.1 Rev. 0 10 CFR 55.41.7 WE09 2.1.20 - Natural Circulation Operations WE09 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

Remarks/Status 7/22/15 Modified Bank question 055 EK1.02 1 7/28/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, Editorial comment, 1) Make more obvious than SG not at saturation.

Thursday, December 03, 2015 Page 74 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 27 27 C 9/11/15 exam team discussed with OPS SRO that we will let the question go to validation as is, and determine if changes are required based on outcome.

12/1/15 question revised to provide realistic SG pressure and show not at or trending to saturation.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 75 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS003 A3.01 - Reactor Coolant Pump System (RCPS)

QUESTION 28 28 C Ability to monitor automatic operation of the RCPS, including: (CFR: 41.7 / 45.5)

Seal injection flow ...............................................

Given the following:

The plant is operating at 100% power when an event occurs.

Without any manual operator actions, the OATC observes the following:

Which ONE of the following identifies the event that has occurred?

A. A SBLOCA in containment has occurred.

B. LT-459, CH I PRZR LEVEL, has failed low.

C. A load rejection to 80% power has occurred.

D. Charging valve HIC-121, CHARGING FLOW, has failed shut.

Thursday, December 03, 2015 Page 76 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 28 28 C General Discussion Answer A Discussion Incorrect. This is incorrect because in a SBLOCA, charging speed will increase to make up for the inventory loss in the RCS and try to keep pressurizer level on program level. This would result in a rise in seal injection flow, and depending on the size of the leak, would have seal injection flow stay higher than normal This is plausible because the transient would affect the seal injection flow.

Answer B Discussion Incorrect. This is incorrect because the LT-459 failure in the low direction will cause the pressurizer level control system to believe that pressurizer level is lower than program and attempt to restore that level, speeding up the charging pump. This will cause seal injection flow to rise, and will only stabilize when the charging pump slows down when pressurizer level is on the new programed value. This is plausible because the level transmitter failure will cause charging pump speed to change, and affect seal injection flow.

Answer C Discussion Correct. This is correct because the load rejection places a transient on the plant that initially causes an insurge into the pressurizer. In accordance with ST-021 CVCS (p102; Rev 3a) the charging pump speed will slow down in an attempt to lower inventory and keep pressurizer level on program. As the plant stabilizes the level control system will then speed up to keep the pressurizer level stable at the new program level for power following the load rejection. Seal injection flow lowers with the charging speed initially and then will rise back up to a value near its initial flow.

Answer D Discussion Incorrect. This is incorrect because the normal charging line flow valve, HIC-121 failing shut will increase the pressure in the charging line, and thereby increase the amount of charging flow that will be supplied to the seal injection line (all charging flow diverted that way). This is plausible because it will affect seal injection flow and cause it to change, however it would go up. The operator may believe that location of HIC-121 in the flow path would cause seal injection flow to lower.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to monitor and interpret the RCP seal injection flow.

Basis for Hi Cog The question is at the Analysis cognitive level because the operator must analyze the indications provided and make a determination as to the plant transient that cause the seal injection flow to change as indicated, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided ST-021 CVCS Rev 3a, pg 102 10 CFR 55.41.7 SYS003 A3.01 - Reactor Coolant Pump System (RCPS)

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

Seal injection flow ...............................................

Remarks/Status 7/29/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, UNSAT, due to distractor D could be partially correct. For smaller leaks in the charging line, the CCPs will speed up to compensate, seal injection will return to almost the original value. Can we change "D" to either HCV-121 failed shut or Seal Injection filters are clogged?

9/8/15 Operations SRO reviewed question, requires further discussion.

9/11/15 exam team discussed with OPS SRO his concerns over the difficulty. It was agreed that changing distractor D to HIC-121 failing shut would eliminate the possibility of having 2 correct answers.

11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 77 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 28 28 C Thursday, December 03, 2015 Page 78 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS003 K4.07 - Reactor Coolant Pump System (RCPS)

QUESTION 29 29 B Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following : (CFR: 41.7)

Minimizing RCS leakage (mechanical seals) ..........................

Given the following:

The plant is in Mode 3

APP-003-E2, VCT HI/LO PRESS, has alarmed

Actual VCT pressure is 15 psig Which ONE of the following correctly completes the statement below?

If VCT pressure continues to lower, RCP #1 Seal Leakoff flow will ____(1)____ and RCP

2 Seal leakoff flow will ____(2)____.

A. (1) lower (2) rise B. (1) rise (2) lower C. (1) rise (2) rise D. (1) lower (2) lower Thursday, December 03, 2015 Page 79 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 29 29 B General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. See C and D.

Answer B Discussion Correct. 1st part correct, 2nd part correct. According to the RCS Text (p16; Rev 6) the shaft seal section consists of three devices. They are the No. 1 controlled leakage, film riding face seal, and the No. 2 and 3 rubbing face seals. During normal system operation the charging pump(s) provide approximately 8 gpm injection flow to each RCP. The injection enters the pump between the thermal barrier and the pump bearing. The flow is then divided with approximately 5 gpm flowing down past the thermal barrier into the RCS and approximately 3 gpm flowing up past the pump bearing. The outlet from the No. 1 seal discharges to the Volume Control Tank (VCT). RCP No. 1 seal leak off flow is normally limited to 3.2 gpm for the purposes of meeting Appendix R (Fire Protection) and SBO (Station Black Out) requirements. The VCT maintains a back pressure of at least 15 psig to ensure a flow through the No. 2 seal. Because #1 and the #2 seal operate in close association with one another, the automatic adjustment of one will affect the other. When VCT pressure lowers, the differential pressure across the #1 Seal will rise, causing the seal leakoff flow to rise. When this occurs the flow to the #2 Seal from the #1 Seal will lower causing its leakoff flow to be lower.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the leakoff from the #2 Seal will lower. This is plausible because the operator may not understand how the Seal operates, and why it is important to maintain a minimum of 15 psig in the VCT.

Answer D Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the leakoff from the #1 Seal will rise. This is plausible because the operator may not understand how the Seal operates, and why it is important to maintain a minimum of 15 psig in the VCT.

Basis for meeting the KA The KA is matched because the operator must demonstrate the knowledge of an RCP seal package design feature (i.e. operate #1 Seal with a minimum backpressure) which if operated properly will limit RCP seal leakage.

Basis for Hi Cog The question is at the Comprehensive cognitive level because the operator must demonstrate an understanding of how the RCP Seal Package works, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 003 K1.04 003 Development References Student References Provided RCS Text (p16; Rev 6) 10 CFR 55.41.7 SYS003 K4.07 - Reactor Coolant Pump System (RCPS)

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

Minimizing RCS leakage (mechanical seals) ..........................

Remarks/Status 7/28/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, expressed concerned over the Safe Shutdown Seals system being new, how much has it been discussed with the class?

9/11/15 Exam team discussed with the OPS SRO and determined that we should let this go to validation.

9/29/15 changed question to one found in bank on RCP seals.

11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 80 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 29 29 B Thursday, December 03, 2015 Page 81 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS004 K2.06 - Chemical and Volume Control System QUESTION 30 30 C Knowledge of bus power supplies to the following: (CFR: 41.7)

Control instrumentation ...........................................

Given the following:

The plant is operating at 100% power

Due to RTGB indications the crew has entered AOP-024, LOSS OF INSTRUMENT BUS

An Auto Makeup to the VCT has initiated on the loss of instrument bus Which ONE of the following correctly completes the statement below?

The Auto makeup has occurred because VCT Level transmitter ____(1)____ has de-energized on a loss of Instrument Bus ____(2)____.

A. (1) LT-112 (2) 7 B. (1) LT-112 (2) 9 C. (1) LT-115 (2) 7 D. (1) LT-115 (2) 9 Thursday, December 03, 2015 Page 82 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 30 30 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because LT-112 does not control automakeups. This is plausible because the operator may incorrectly believe that this level transmitter is responsible for controlling the VCT level as it is only one of two level transmitters associated with that system.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and D.

Answer C Discussion Correct. According to EDP-008 (p15 of 23; Rev 33) LT-115 is powered from Instrument Bus 7A. According to ST-021 (p34 of 136; Rev. 3a)

LT-115 also provides inputs to LCV-115A (backup to LT-112, high level override), the VCT Automatic Makeup Control System, and the VCT HI/LO LVL alarm, APP-003-E3 (HI 48.6", LO 17.2"). On a failure of power to this level transmitter, the VCT level control system will see the VCT at 0 level and an auto makeup will commence.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because LT-115 is powered by Instrument Bus 7A. This is plausible because Bus 9 provides power to LT-112.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of bus power supplies to the primary VCT Level Control Instrument.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall a power supply in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EDP-008 (p15 of 23; Rev 33)

ST-021 CVCS (p34 of 136; Rev 3a) 10 CFR 55.41.7 SYS004 K2.06 - Chemical and Volume Control System Knowledge of bus power supplies to the following: (CFR: 41.7)

Control instrumentation ...........................................

Remarks/Status 8/4/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, UNSAT, is the KA asking for the power supply to the components or is it asking about the power supplies to the actual instrument busses? Are the students required to know this from memory?

9/8/15 Operations SRO reviewed question, UNSAT, as there is no way that this should be information required from memory.

9/29/15 changed question to new question.

11/11/15 edited the question to isolate the power supply to the IB 7 vs 9.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 83 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS005 A2.02 - Residual Heat Removal System (RHRS)

QUESTION 31 31 B Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, 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)

Pressure transient protection during cold shutdown ....................

Given the following:

A plant cooldown is in progress

The A RHR Train is in service

RCS temperature is 190°F

RCS Pressure is 350 psig

The RCS is in Solid Plant Operation Subsequently, a transient occurs resulting in the following:

RCS temperature rises to 195°F

RCS Pressure rises to 420 psig

PZR PORV PCV-456 indicates OPEN

PZR PORV PCV-455C indicates CLOSED The crew has entered AOP-019, MALFUNCTION OF RCS PRESSURE CONTROL.

Which ONE of the following correctly completes the statements below?

1) Based on plant conditions, both PORVs should have ____(1)____.

2) The operator will FIRST ____(2)____ to control or mitigate the consequences of this event.

A. (1) remained closed (2) close PCV-456 B. (1) opened (2) stop all Charging Pumps and RCPs C. (1) opened (2) operate PZR sprays and heaters to control pressure D. (1) remained closed (2) place the PCV-456 LTOPP Arming Switch to NORMAL Thursday, December 03, 2015 Page 84 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 31 31 B General Discussion Answer A Discussion Incorrect. This is incorrect because the setpoint for the PORVs opening in the LOW PRESSURE Mode is 400 psig, and both valves should have opened. This is plausible because the setpoint is temperature dependent and will rise as temperature rises. In the situation provided, a temperature rise has taken place so the operator may incorrectly believe that the setpoint is higher than 420 psig. If so, the operator would conclude that both valves should have remained CLOSED. If the operator believed that the valves should have remained CLOSED, it is reasonable that the operator would believe that a required action is to CLOSE the failed OPEN PORV.

Answer B Discussion Correct. According to ST-059 (p31; Rev 4), LTOPP control is required to be activated when the RCS is cooled down below 365ºF, to minimize Pressurized Thermal Shock (PTS) concerns. The LTOP controller uses the lowest of TE-410, TE-420 and TE-430 to determine RCS temperature and pressure as sensed by PT-500 and PT-501. The lift setpoint is variable based upon auctioneered low RCS temperature. At an RCS temperature of 365ºF, the pressure setpoint is 400 psig. The setpoint of the Comparators PC-502 and PC-503 are raised as RCS temperature is increased. The setpoint will not lower below 400 psig. Consequently, under the current plant conditions, BOTH PORVs should have OPENED. According to AOP-019 (p3 of 33; Rev 20), Step 1, the operator will check the LTOPP Arming Switches in NORMAL. Since they are in the LOW PRESSURE position, the operator will perform the RNO which directs them to Stop all Charging Pumps and RCPs.

Answer C Discussion Incorrect. This is incorrect because Operate Pzr sprays and heaters to control pressure is not an expected action in these plant conditions (Solid Plant Ops). This is plausible because this would be the required action if the plant were NOT in SOLID Plant Operation.

Answer D Discussion Incorrect. This is incorrect because the setpoint for the PORVs opening in the LOW PRESSURE Mode is 400 psig, and both valves should have opened. This is plausible because the setpoint is temperature dependent and will rise as temperature rises. In the situation provided, a temperature rise has taken place so the operator may incorrectly believe that the setpoint is higher than 420 psig. If so, the operator would conclude that both valves should have remained CLOSED. If the operator believed that the valves should have remained CLOSED, it is reasonable that the operator would believe that a required action is Place the PCV-456 LTOPP Arming Switch to NORMAL in an attempt to Close PCV-456.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to predict the impacts of a pressure transient protection during cold shutdown on the RHR System (Which in Solid Plant Operation Includes the LTOPs), and based on this, use AOP-019 to correct, control, or mitigate the consequences.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall how the LTOP works, predict whether or not it should be working under the current plant conditions, and identify what actions need to be taken based on the transient, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK Development References Student References Provided ST-059 (p31; Rev 4)

AOP-019 (p3 of 33; Rev 20) 10 CFR 55.41.7 SYS005 A2.02 - Residual Heat Removal System (RHRS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, 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)

Pressure transient protection during cold shutdown ....................

Remarks/Status 8/4/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, SAT Thursday, December 03, 2015 Page 85 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 31 31 B 11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 86 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS005 K6.03 - Residual Heat Removal System (RHRS)

QUESTION 32 32 B Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 / 45.7)

RHR heat exchanger ..............................................

Given the following:

A Train RHR is in service

Instrument Air pressure lowers to 50 psig Which ONE of the following correctly completes the statement below?

RCS temperature will INITIALLY ____(1)____, and RHR System Flow will INITIALLY

____(2)____.

A. (1) Rise (2) Rise B. (1) Rise (2) Lower C. (1) Lower (2) Lower D. (1) Lower (2) Rise Thursday, December 03, 2015 Page 87 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 32 32 B General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because system flow will lower. This is plausible because the operator may incorrectly believe that FCV-605 will fail OPEN.

Answer B Discussion Correct. 1st part correct, 2nd part correct. According to ST-003 (p19-20; Rev 4a) HCV-758, RHR Loop Temperature Control valve, will fail closed on a Loss of Instrument Air; and HCV-605, RHR Loop Flow Control Valve, will fail closed on a Loss of Instrument Air. Consequently, both valves will fail closed under the stated plant conditions. The purpose of HCV-758 is to control the rate of temperature changing during RCS Heat up/Cooldown. Consequently, if this valve fails CLOSED, the Heat Exchanger is essentially removed from service, and the RCS Temperature will start to rise. The purpose of HCV-605 is to maintain specified flow rate in RHR loop. However, on a loss of air, the valve will fail CLOSED and all system flow will stop.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the RCS Temperature will initially rise when flow through the RHR HX is stopped. This is plausible because the operator may incorrectly believe that HCV-758 fails OPEN on a loss of Air.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and C. This is plausible because the operator may incorrectly believe that both HCV-758 and HCV-605 fail OPEN on a loss of Air.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the effect of a loss RHR heat exchanger will have on the RHRS. This is accomplished by requiring the operator to identify that the RCS Temperature will rise initially when RHR flow through the RHR HX is stopped.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and then apply this information to stated condition, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK Development References Student References Provided ST-003 (p19-20; Rev 4a) 10 CFR 55.41.4 SYS005 K6.03 - Residual Heat Removal System (RHRS)

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

RHR heat exchanger ..............................................

Remarks/Status 8/5/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, commented if an leak with IA pressure lowering to 50 PSIG was run on the simulator to observe effects?

9/11/15 Exam team discussed with OPS SRO that the 50 PSIG is low enough to cause effects on the RHR TCV and FCV as presented.

11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 88 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 32 32 B Thursday, December 03, 2015 Page 89 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS006 A1.13 - Emergency Core Cooling System (ECCS)

QUESTION 33 33 C Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: (CFR: 41.5 / 45.5)

Accumulator pressure (level, boron concentration) ....................

Given the following:

The plant is at 100% RTP

The following timeline is observed:

TIME A SI accumulator pressure A SI accumulator level 1108 637 psig 68 percent 1111 646 psig 70 percent 1114 660 psig 75 percent Which ONE of the following correctly completes the statement below?

The EARLIEST time at which an RTGB annunciator alarm would occur is at time ____(1)____,

due to ____(2)____ reaching its alarm setpoint.

A. (1) 1111 (2) APP-002-A4, SI ACCUM A HI/LO LVL B. (1) 1114 (2) APP-002-B4, SI ACCUM A HI/LO PRESS C. (1) 1111 (2) APP-002-B4, SI ACCUM A HI/LO PRESS D. (1) 1114 (2) APP-002-A4, SI ACCUM A HI/LO LVL Thursday, December 03, 2015 Page 90 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 33 33 C General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the APP-002-A4, SI ACCUM A HI/LO LVL will not alarm until 75%.

This is plausible because the APP discusses that if level has risen greater than 70 GALLONS, not due to an addition from the RWST, that action needs to be taken IAW ITS 3.5.1.4. The operator may confuse these numbers and determine that the 70 is for percent of level and the alarm setpoint.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the earliest time that an alarm would be received is time 1111. This is plausible if the operator incorrectly believes that the APP-002-B4, SI ACCUM A HI/LO PRESS alarms at 660 psig, which is the high end of the ITS LCO 3.5.1 limit.

Answer C Discussion Correct. 1st part correct, 2nd part correct. According to APP-002-B4 (p19; Rev 70) the accumulator high pressure alarm setpoint is 646 PSIG.

At time 1111, the "A" SI accumulator reaches that setpoint. According to APP-002-A4 (p7; Rev 70) the accumulator high level alarm setpoint is 75% and the level at time 1111 is only 70%. Therefore this is the earliest time that an alarm would be received and that alarm would be due to pressure.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because the alarm setpoint for the accumulator pressure would have been reached at time 1111. This is plausible because SI accumulator level reaches its alarm setpoint of 75% at this time, and the operator may incorrect believe that the pressure has not reached an alarm setpoint yet.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to monitor changes in the accumulator pressure instrumentation by demonstrating knowledge of the cause of the alarm and the alarm setpoints.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall the alarm setpoints for the accumulator high pressure alarm and the high level alarm, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided APP-002-A4 p7 Rev 70 APP-002-B4 p19 Rev70 10 CFR 55.41.7 SYS006 A1.13 - Emergency Core Cooling System (ECCS)

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

Accumulator pressure (level, boron concentration) ....................

Remarks/Status 7/29/15 Used Bank question SI-009 1.

9/8/15 Reviewed by Exam Supervisor, SAT 12/1/15 switched the level and pressure columns.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 91 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 33 33 C Thursday, December 03, 2015 Page 92 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS006 K5.01 - Emergency Core Cooling System (ECCS)

QUESTION 34 34 B Knowledge of the operational implications of the following concepts as they apply to ECCS: (CFR: 41.5 / 45.7)

Effects of temperatures on water level indications ....................

Given the following:

The plant was at 100% power when a LOCA occurs inside Containment Which ONE of the following correctly completes the statements below?

1) While operating in the EOP Network, the LOWEST pressure in which Adverse containment values will be used is ____(1)____.

2) Based on containment temperature, level detectors with wet reference legs will indicate____(2)____ than the actual level.

A. (1) 4 psig (2) lower B. (1) 4 psig (2) higher C. (1) 10 psig (2) lower D. (1) 10 psig (2) higher Thursday, December 03, 2015 Page 93 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 34 34 B General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because reference legs will become less dense due to the temperature rise of the containment atmosphere in which they are in. The lower density will cause the D/P sensed to be smaller and the level indicated to be higher than actual, NOT lower. This is plausible because the operator may incorrectly recall the way in which the D/P cell uses the reference and variable leg densities to indicate level.

Answer B Discussion Correct. 1st part correct, 2nd part correct. According to OMM-022 (p 5; Rev 45) Definition 2, Adverse Containment Conditions: If the CV pressure is greater than or equal to 4 psig, then adverse containment conditions exist. When adverse setpoints are provided, they will be enclosed by brackets [ ]. According to LOF0020R (p 99; Rev 0) Section 4.6, Temperature variations can significantly affect the level measurement accuracy. An increase in ambient temperature will cause the density of the reference leg to decrease. This will result in a lower D/P sensed by the D/P cell and indicated level will be greater than actual level. Any pressure variations in the environment are felt on both sides of the open vessel DP level instrument and subsequently cancel each other. The dry and wet reference leg level detectors are only exposed to system pressure and therefore are not affected.

Answer C Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and D.

Answer D Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because 4 psig is the CV pressure when exceeded adverse values will be used. This is plausible because 10 psig is the CV pressure when exceed will actuate Containment Spray and Phase "B" isolation.

Basis for meeting the KA The KA is matched because the operator must demonstrate a knowledge of how the containment adverse conditions (temperature) will effect water level indications.

Basis for Hi Cog The question is at the Memory level because the operator must recall the value at which adverse containment numbers are used, and the operation of a D/P cell in a level measurement instrument.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED PATH-1-003 4 Development References Student References Provided OMM-022 Rev. 45 LOF0020R Rev. 0 10 CFR 55.41.10 SYS006 K5.01 - Emergency Core Cooling System (ECCS)

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

Effects of temperatures on water level indications ....................

Remarks/Status 8/12/15 Modified bank question PATH-1-003 4 9/8/15 Reviewed by Exam Supervisor, SAT 11/17/15 Fleet review determined that we did not hit the intent of the KA. The second part of the question was changed to make the question a better KA match.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 94 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 34 34 B Thursday, December 03, 2015 Page 95 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS007 K3.01 - Pressurizer Relief Tank/Quench Tank System (PRTS)

QUESTION 35 35 B Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: (CFR: 41.7 / 45.6)

Containment ....................................................

Given the following:

The plant is at 100% power

Containment pressure is 0.2 psig

Containment temperature is 94°F Subsequently:

A load rejection results in a reactor trip

Following the trip, a Pressurizer Safety valve opens, and will NOT reseat

The PRT rupture disks function as designed

Containment pressure is rising at 0.1 psig every 5 minutes

Containment temperature is rising at 2°F every 5 minutes Assuming these conditions remain constant, which ONE of the following identifies the Containment Technical Specifications LCOs that will be affected one hour from now?

A. Both LCO 3.6.4, Containment Pressure, and LCO 3.6.5, Containment Air Temperature, will be exceeded.

B. Only LCO 3.6.4, Containment Pressure, will be exceeded.

C. Only LCO 3.6.5, Containment Air Temperature, will be exceeded.

D. Neither LCO 3.6.4, Containment Pressure, nor LCO 3.6.5, Containment Air Temperature, will be exceeded.

Thursday, December 03, 2015 Page 96 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 35 35 B General Discussion Answer A Discussion Incorrect. This is incorrect because LCO 3.6.5 will be exceeded at 65 minutes. This is plausible because the operator may not know the Containment Air Temperature LCO, and be unable to successfully solve the problem.

Answer B Discussion Correct. According to LCO 3.6.4 (p3.6-13; Amendment 176) the Containment pressure is limited to 1.0 psig during Modes 1-4. According to LCO 3.6.5 (p3.6-14; Amendment 176) the Containment temperature is limited to 120°F during Modes 1-4. Since the Containment pressure is rising at 0.1 psig every five minutes, it will reach the LCO limit in 45 minutes. Since the Containment temperature is rising at 2°F every five minutes, it will reach the LCO limit in 65 minutes.

Answer C Discussion Incorrect. This is incorrect because the reverse is true. This is plausible because the operator may not know the either the Containment Pressure or the Containment Air Temperature LCO, and be unable to successfully solve the problem.

Answer D Discussion Incorrect. This is incorrect because LCO 3.6.4 will be exceeded at 45 minutes. This is plausible because the operator may not know the Containment Pressure LCO, and be unable to successfully solve the problem.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the effect that a loss the PRTS (Rupture Disk Releases) will have on the Containment (i.e. TS Limits are challenged).

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and then use the information to predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided LCO 3.6.4 (p3.6-13; Amendment 176)

LCO 3.6.5 (p3.6-14; Amendment 176) 10 CFR 55.41.10 SYS007 K3.01 - Pressurizer Relief Tank/Quench Tank System (PRTS)

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

Containment ....................................................

Remarks/Status 7/29/15 New question developed 8/12/15 Reviewed by Fleet NRC Exam Writer, with the following comment, 1) question may be too hard as currently written.

9/8/15 Reviewed by Exam Supervisor, SAT.

11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 97 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS008 A4.01 - Component Cooling Water System (CCWS)

QUESTION 36 36 A Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5)

CCW indications and controls .....................................

Given the following:

T= 1305:10

A loss of off-site power has occurred concurrent with a reactor trip

The crew enters EOP-E-0, REACTOR TRIP OR SAFETY INJECTION T= 1307:40

Immediate actions of EOP-E-0 are complete

The following parameters are noted:

RCS pressure is 2110 psig

SG pressures are all 910 psig and lowering slowly

CV pressure is 0.7 psig and rising slowly Which ONE (1) of the following describes the CCW Pump indication available in the control room?

A. "B" and "C" CCW Pump RED indicating lights are LIT.

"A" CCW Pump GREEN indicating light is LIT.

B. All 3 CCW Pump RED indicating lights are LIT.

C. "B" and "C" CCW Pump GREEN indicating lights are LIT; "A" CCW Pump RED indicating light is LIT.

D. "B" and "C" CCW Pump RED indicating lights are LIT; "A" CCW Pump GREEN and RED indicating lights are EXTINGUISHED.

Thursday, December 03, 2015 Page 98 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 36 36 A General Discussion Answer A Discussion Correct. According to ST-013 (p52; Rev 1) during a Loss of Offsite Power, all three CCW pumps are load shed from their respective buses.

CCW pumps "B" & "C" which are supplied from emergency bus E1 and E2 will trip on bus undervoltage via the 27 Bus Undervoltage Relays, and CCW pump "A" will trip when the undervoltage trip device of the feeder breaker senses the loss of voltage on the DS bus. On Train A, CCW Pump "B" will auto restart after 30 seconds via the safeguards sequencer blackout (LOOP) logic after Emergency Diesel Generator "A" output breaker closes. On Train B, CCW Pump "C" - auto restart after 30 seconds via the safeguards sequencer blackout (LOOP) logic after Emergency Diesel Generator "B" output breaker closes. Consequently, the "B" and "C" CCW Pump Red status lights will be LIT. According to ST-056 (p53; Rev 5) the DS Bus which powers the "A" CCW Pump will be energized 95 seconds after the under voltage condition detected on the DS bus occurs. Since the "A" CCW load shed on the undervoltage condition on the DS Bus, it will only start if an AUTO start signal occurs to this pump. According to ST-013 (p52; Rev 1), CCW Pump "A" will auto start on low CCW pressure, with power available on the bus.

However, the low pressure condition in the CCW System would have cleared at the 30 second point when the "B" and "C" CCW Pumps auto-started on the Blackout Sequencer signal. Consequently, the "A" CCW Pump Green status light will be LIT.

Answer B Discussion Incorrect. This is incorrect because the "A" CCW Pump is NOT running. This is plausible because it would be correct if an SI and/or CV Spray signal existed.

Answer C Discussion Incorrect. This is incorrect because the "B" and "C" CCW Pumps are running. This is plausible because the all-pumps OFF condition is a direct result of the LOOP, and the operator may incorrectly believe that no additional automatic actions occur with these pumps based on plant conditions.

Answer D Discussion Incorrect. This is incorrect because the "A" CCW Pump Green light will be LIT. This is plausible because the operator may correctly believe that the "A" CCW Pump is OFF, but incorrectly believe that it does not have Control Power to the breaker.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to monitor CCW Pump motor breaker status lights at the RTGB.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and then use this information to predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 008 A4.01 1 Development References Student References Provided ST-013 (p52; Rev 1)

ST-056 (p53; Rev 5) 10 CFR 55.41.7 SYS008 A4.01 - Component Cooling Water System (CCWS)

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

CCW indications and controls .....................................

Remarks/Status 8/26/15 Used Bank question 008 A4.01 1 9/8/15 Reviewed by Exam Supervisor, with editorial comments, 1)most of the information in the stem is not needed; 2) need to add a statement that all auto systems operated as designed.

9/8/15 Operations SRO reviewed question, comment was "Time?"

9/11/15 Exam team discussed with reviewers that Times are given and the information in the stem is required to ensure the operator candidate can evaluate if a SI has occurred.

Thursday, December 03, 2015 Page 99 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 36 36 A 11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 100 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS010 K6.04 - Pressurizer Pressure Control System (PZR PCS)

QUESTION 37 37 C Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: (CFR: 41.7 / 45.7)

PRT ...........................................................

The plant is at 100% power when the following occurs:

A load rejection results in a PZR PORV actuation

The PORV will not fully seat

The depressurization led to a reactor trip and safety injection

PZR pressure reduces to 1400 psig and stabilizes

Safety Valve tailpiece temperature indicates 338°F with the PRT at 100 psig Which ONE of the following correctly completes the statements below?

When the PRT rupture disc ruptures, the Safety Valve tailpiece temperature will ____(1)____.

The rate of RCS depressurization will ____(2)____ when that occurs.

A. (1) rise (2) increase B. (1) rise (2) remain the same C. (1) lower (2) increase D. (1) lower (2) remain the same Thursday, December 03, 2015 Page 101 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 37 37 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because when the rupture discs rupture, the constant enthalpy in the throttling process will be evident by a lowering temperature in the tailpiece, as the pressure that the PORV is relieving too is now lower. This is plausible because the operators may believe that with the disc rupture, increased flow of steam out of the PZR PORV due to less resistance to that flow will result in higher temperatures in the line. This misconception is what led operators at TMI to believe that they did not have a PORV relieving.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See A & D.

Answer C Discussion Correct. Ist part correct, 2nd part correct. According to GFES LOF0012R (p 90 Rev.0) the throttling process is a constant enthaply process, that was not clearly recognized by operators during the event at Three Mile Island. As the enthalpy remains constant in this process, we find that the downstream pressure is what determines the temperature in that point in the system. That constant pressure line is then followed to the saturation line and indicates the temperature that the Safety Valve tailpiece will indicate. For 100 psig, (115 psia) right before the rupture discs rupture, that temperature will be 338 degrees F. Once that disc ruptures and the pressure downstream is then released to the CV, the constant enthalpy line will then be extended to that new lower pressure, and the tailpiece will indicate a lower temperature. According to LOF0014R (p96 Rev.0) flow is calculated as proportional to the square of the D/P in the system. This concept applies for break flow as well. When the rupture discs rupture, the D/P will increase, therefore the RCS depressurization rate will increase as well.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because once the PRT discs rupture, the pressure in the PRT will be reduced to close to what CV ambient is. The flow out of the open PORV will increase as it is proportional to the square root of the D/P. This is plausible because the operator may believe that in a constant throttling process, which is what this is, the rate of depressurization is also constant.

Basis for meeting the KA The KA is matched because the operator must demonstrate their knowledge of how a blown rupture disk (malfunction) in the PRT will affect the PZR pressure control system, specifically the indicated tail piece temperatures and the rate of RCS depressurization.

Basis for Hi Cog The question is at the Comprehension cognitive level because the operator must understand the thermodynamic processes in a throttling process, analyze conditions, a then determine how those indications will change when a loss of the PRT (rupture disc ruptures) occurs, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided GFES LOF0012R 10 CFR 55.41.14 SYS010 K6.04 - 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)

PRT ...........................................................

Remarks/Status 7/28/15 Used Bank question ZG-THERMO CHAP 3-008 1 8/12/15 Reviewed by Fleet NRC Exam Writer, and provided the following comments, 1) PSIA conversion as a plausible distractor is sometimes not accepted, specifically by our chief examiner; 2) another plausible distractor would be the SAT temp for 1400 PSIG, the same mistake TMI made.

9/8/15 Reviewed by Exam Supervisor, SAT 11/17/15 Fleet review indicate that the question as written did not hit the KA. The question was rewritten to ask for the operators knowledge of Thursday, December 03, 2015 Page 102 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 37 what happens to the PZR when the PRT rupture discs rupture (effect due to PRT loss).

37 C 11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 103 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS012 A3.06 - Reactor Protection System (RPS)

QUESTION 38 38 D Ability to monitor automatic operation of the RPS, including: (CFR: 41.7 / 45.5)

Trip logic .......................................................

Which ONE of the following identifies a condition which will cause an automatic reactor trip?

A. RCP bus frequency reads 58.0 Hz when THREE Power Range channels read 5%

RTP.

B. Flow in ONE Reactor Coolant Loop reads 91% when ALL Power Range channels read 35% RTP.

C. ONE Intermediate Range NI channel reads 1 x10E-6 AMPS when TWO Power Range channels read 15% RTP.

D. ONE Source Range NI channel reads 5 x10E5 CPS when TWO Intermediate Range NI channels read 9 x10E-11 AMPS.

Thursday, December 03, 2015 Page 104 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 38 38 D General Discussion Answer A Discussion Incorrect. This is incorrect because below P-7 (10% Turbine Power) the TWO LOOP LO FLOW TRIP is blocked . This is plausible because if all RCP bus frequencies are less than 58.2Hz then all RCPs would be expected to trip and the student may believe that all RCPs being tripped would result in a Reactor Trip even below P-7.

Answer B Discussion Incorrect. This is incorrect because below P-8 (40% Reactor Power) the ONE LOOP LO FLOW TRIP is blocked. This is plausible because the flow given is below the LO FLOW TRIP setpoint but you would require two loops to have low flow for the trip to occur.

Answer C Discussion Incorrect. This is incorrect because IR TRIP is normally blocked by the operator by procedure when the reactor power is above 10%, but the current equivalent power at which the trip occurs is actually 25%. This is plausible because the IR level given is approaching the current equivalency for the trip setpoint for the IR TRIP.

Answer D Discussion Correct. According to ST-011 RPS (p20; Rev 5) Chapter 4, Instrumentation, Reactor Trips, 1. Source Range, the SR TRIP will occur above 1 x105 CPS. This trip can be blocked when 1 out of 2 IR channels are above 1 x10-10 AMPS, but for the conditions listed the IR is still below that value.

Basis for meeting the KA The KA is matched because the operator must demonstrate understanding of the various trip signals that the Reactor Protection System will initiate a reactor trip on.

Basis for Hi Cog The question is at the Analysis cognitive level because the operator must analyze each of the conditions in each answer choice, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK RPS-006 37 Development References Student References Provided ST-011 RPS Rev 5 10 CFR 55.41.7 SYS012 A3.06 - Reactor Protection System (RPS)

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

Trip logic .......................................................

Remarks/Status 7/20/15 Used Bank question RPS-006 37 9/8/15 Reviewed by Exam Supervisor, with editioral comments, 1) change "a reactor trip" to "an automatic reactor trip"; 2) re arrange distractors to have shortest on top, longest on bottom.

9/8/15 Incorporated Exam Supervisors comments.

11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 105 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS012 K1.05 - Reactor Protection System (RPS)

QUESTION 39 39 B 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)

ESFAS .........................................................

Given the following:

With the plant at power a safety injection actuation occurs

The Feedwater Isolation OVRD/RESET key switches are placed in OVRD/RESET Which ONE of the following identifies the Feedwater System components that are AVAILABLE for operation?

A. The Feed Reg Valves are available; The Feed Reg Bypass Valves are available.

B. The Feed Reg Valves are NOT available; The Feed Reg Bypass Valves are available.

C. The Feed Reg Valves are available; The Feed Reg Bypass Valves are NOT available.

D. The Feed Reg Valves are NOT available; The Feed Reg Bypass Valves are NOT available.

Thursday, December 03, 2015 Page 106 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 39 39 B General Discussion Answer A Discussion Incorrect. This is incorrect because the Feed Reg Valves cannot be opened. This is plausible because the operator may not know how the FWIS circuit functions.

Answer B Discussion Incorrect. This is incorrect because the Feed Reg Valves cannot be opened. This is plausible because the operator may not know how the FWIS circuit functions.

Answer C Discussion Correct. According to ST-027 (p37-38; Rev 6) the feedwater isolation circuitry is a portion of the safeguards system that isolates water flow from the main feedwater system to the steam generators under certain conditions (2/3 Hi Hi steam generator level and safety injection). There are three key operated switches on the RTGB that are used to reset and/or override the Feedwater Isolation signals (2/3 Hi-Hi steam generator level and safety injection). Placing the Feedwater Isolation OVRD/RESET key switches in the OVRD/RESET position would allow the main feedwater pumps to be started and the feedwater regulating bypass valves to be opened. The main feedwater regulating valves would still remain isolated, and will remain isolated until the reactor trip breakers are closed.

Answer D Discussion Incorrect. This is incorrect because the Feed Reg Valves cannot be opened. This is plausible because the operator may not know how the FWIS circuit functions.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the physical connections and/or cause effect relationships between the RPS and the Feedwater Isolation Signal (ESFAS).

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall what conditions are needed to re-open the FRVs after a FWIS based on SI actuation, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided ST-027 (p37-38; Rev 6) 10 CFR 55.41.7 SYS012 K1.05 - 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)

ESFAS .........................................................

Remarks/Status 7/29/15 new question developed.

8/12/15 reviewed by Fleet NRC Exam writer, with the following comments; 1) Typical ">2 part question". Better to ask in traditional 2 X 2 format. i.e. FRV are/are not, contrasted with something else.

9/2/15 revised question to make it a 2 X 2 contrasting the FRV and FRV bypass availability.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 107 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 39 39 B Thursday, December 03, 2015 Page 108 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS013 K5.02 - Engineered Safety Features Actuation System (ESFAS)

QUESTION 40 40 C 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 ..................................

Which ONE of the following correctly completes the statements below regarding the Containment Spray Actuation logic scheme?

The MINIMUM number of Containment Pressure channels that must sense pressure greater than the Hi-Hi pressure setpoint to generate an automatic Containment Spray Actuation is

____(1)____.

The Containment Spray Actuation bistables are ____(2)____ to actuate.

A. (1) Two (2) Energize B. (1) Two (2) De-energize C. (1) Four (2) Energize D. (1) Four (2) De-energize Thursday, December 03, 2015 Page 109 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 40 40 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and D.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the MINIMUM number of pressure channels is four (2 in each 2/3 scheme). This is plausible because the 4 psig Hi Containment Pressure signal is simply 2 of 3 logic, and requires a MINIMUM of Two. The operator may not know that the Hi-Hi Containment Pressure uses a different logic scheme.

Answer C Discussion Correct. 1st part correct, 2nd part correct. According to ST-024 (p22; Rev 4) CS Actuation will automatically occur when a Containment Hi-Hi Pressure signal is sensed at 10 psig. According to ST-024 (p19; Rev 4) There are nine (9) transmitters located in the Aux. Bldg. near the IVSW tank area. Three are used for the (2 out of 3) HI pressure SI signal at 4 psig (PC-951B, 953B, & 955B) and six supply the HI-HI pressure signal actuation at 10 psig (PC-950, 951A, 952, 953A, 954 & 955A). These six transmitters are arranged in a 2 of 3 logic scheme, taken twice.

Consequently, a MINIMUM four pressure transmitters will be needed to actuate to auto actuate Containment Spray. According to ST-024 (p23; Rev 4) Containment pressure bistables for spray actuation are energize-to-actuate. This differs from other ESF actuations. The purpose is to minimize the possibility for an inadvertent spray signal due to power interruption.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the bistables are energize-to actuate. This is plausible because all of the other ESF bistables are De-energize to function, and the operator may incorrectly believe that the Containment Spray Actuation scheme is similar.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the operational implications of the safety system logic (2/3 taken twice vs. 2/3) and reliability (energize v. de-energize to actuate) as they apply to the Containment Spray Auto Actuation (ESFAS).

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall bits of information regarding the Containment Spray Actuation logic scheme in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided ST-024 (p19, 22& 23; Rev 4) 10 CFR 55.41.7 SYS013 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 ..................................

Remarks/Status 7/20/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 110 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS022 A4.04 - Containment Cooling System (CCS)

QUESTION 41 41 B Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Valves in the CCS ................................................

Given the following:

The plant is operating at 100% power

The Service Water valve alignment for HVH-1 (CV Recirc Fan A) is as shown below Which ONE of the following identifies the operational restrictions, if any, placed on HVH-1?

A. HVH-1 may run continuously, there are no restrictions.

B. HVH-1 may operate for brief periods of up to 15 minutes.

C. HVH-1 must immediately be secured and cannot be run under these conditions.

D. HVH-1 may be started in emergencies ONLY, but it must run for at least 15 minutes before shutdown.

Thursday, December 03, 2015 Page 111 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 41 41 B General Discussion Answer A Discussion Incorrect. This is incorrect because HVH-1 may NOT run continuously when there is no SW flow to the Motor. This is plausible because the operator may incorrectly diagnose the fact that SW flow has been stopped to the Fan Motor. For instance, if V6-35A was closed, rather than V6-33A, this would be correct.

Answer B Discussion Correct. According to OP-921 (p8 of 56; Rev 55) Caution prior to Step 6.1.1.2, states that Continuous operation of HVH-1, 2, 3, or 4 fan motor without cooling water to the fan motor may result in damage to the fan motor. However, HVH-1, 2, 3, or 4 may be operated for brief periods of up to 15 minutes without cooling water to the fan motor per ESR 95-00700.

Answer C Discussion Incorrect. This is incorrect because HVH-1 may be run under situation where there is no SW flow to the Motor. This is plausible because prolonged periods of operation will result in damage to the fan motor.

Answer D Discussion Incorrect. This is incorrect because when started under these conditions, the fan cannot run longer than 15 minutes. This is plausible because restriction would be appropriate under certain starting duty requirements. For instance, if the fan has been started two times in the last hour and was run at least 15 minutes and stopped in one of these runs, then one additional start is allowed with NO waiting period. The operator may confuse the restrictions.

Basis for meeting the KA The KA is matched because the operator must demonstrate the Ability to monitor the CCS Valves in the control room.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must analyze indicated conditions and based on that apply precautions and limitations, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided OP-921 (p8 of 56; Rev 55) 10 CFR 55.41.9 SYS022 A4.04 - Containment Cooling System (CCS)

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

Valves in the CCS ................................................

Remarks/Status 7/20/15 new question developed 9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, recommended showing the picture with the fan 'ON' to eliminate confusion.

9/11/15 exam team will change picture to show HVH-1 in ON, but distractor D may need to be changed.

9/14/15 exam team changed picture, no changes required to distractors/answer choices.

11/24/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 112 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 41 41 B Thursday, December 03, 2015 Page 113 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS026 A2.08 - Containment Spray System (CSS)

QUESTION 42 42 C Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; 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)

Safe securing of containment spray when it can be done) ...............

Given the following:

At 12:00 a LOCA occurred inside containment causing automatic actuation of the CV Spray System

At 12:05 CV pressure peaked at 15 PSIG

CV pressure since peaking has been lowering at 0.2 PSIG per minute at a constant rate.

Assuming the rate of pressure decrease remains constant, which ONE of the following correctly completes the statement below?

IAW EOP-E-1, LOSS OF REACTOR OR SECONDARY COOLANT, the EARLIEST the Containment Spray Pumps can be stopped is between ____(1)____.

Prior to stopping the CV Spray pumps and placing the system in a normal alignment, the operator is directed to ____(2)____.

A. (1) 12:25 and 12:35 (2) reset containment spray AND phase B B. (1) 12:25 and 12:35 (2) reset containment spray ONLY C. (1) 12:55 and 13:05 (2) reset containment spray AND phase B D. (1) 12:55 and 13:05 (2) reset containment spray ONLY Thursday, December 03, 2015 Page 114 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 42 42 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the pressure requirement to secure the pumps is containment pressure less than 4 PSIG. This is plausible because the student may incorrectly believe that the pressure drop to less than 10 PSIG in containment, the pressure at which a Spray signal is actuated, is the value at which the Containment spray system can be secured. If the operator candidate calculates the time to less than 10 PSIG they will arrive at a time of 12:30.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and D.

Answer C Discussion Correct. 1st part correct, 2nd part correct. According to EOP-E-1 (p11; Rev 4) Step 12, the containment spray pumps can be stopped when the containment pressure is less than 4 PSIG. The operator will calculate that the time the containment pressure is less than 4 PSIG based on the current rate of decrease is 55 minutes. Therefore 13:00 is the earliest time that pressure would have been reduced to less than 4 PSIG.

According to EOP-E-1 (p11 of 26; Rev. 6) Step 12 d. has the operator reset the containment spray signal by placing the CONTAINMENT SPRAY Key Switch to OVRD/RESET and RETURN to NORMAL, AND RESET Containment Isolation Phase B, prior to securing the CV Spray pumps.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the operator has to reset the containment spray signal by placing the CONTAINMENT SPRAY Key Switch to OVRD/RESET and RETURN to NORMAL, AND RESET Containment Isolation Phase B, prior to securing the CV Spray pumps. This is plausible because the operator may only believe that the containment spray key switch is required to be reset prior to securing the pumps.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to determine the CV pressure at which the CV spray pumps can be secured, calculate the time to reach that pressure, and understand the procedural requirements to safely secure the CV spray pumps.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must calculate the time at which the CV pressure will be less than the value at which the CV spray pumps can be secured and understand what must be reset to safely secure those pumps, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided EOP-E-1, pg 11 Rev 4 10 CFR 55.41.7 SYS026 A2.08 - Containment Spray System (CSS)

Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; 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)

Safe securing of containment spray when it can be done) ...............

Remarks/Status 7/22/15 new question developed 9/8/15 Reviewed by Exam Supervisor, SAT 9/11/15 OPS SRO reviewed the question, with Editorial comments, that the 41 minute time may be a high miss part of this question. Consider making the question a 2 X 2 with the CV pressure <4 and <10 and the time 41 min and some other number.

9/30/15 consider changing rate at which pressure is lowering to 0.3 psig.

Thursday, December 03, 2015 Page 115 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination 12/1/15 question modified to remove pyschometric flaws.

QUESTION 42 42 C 12/1/15 add to last bullet prior to securing the spray pumps and closing discharge valves.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 116 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS039 A1.10 - Main and Reheat Steam System (MRSS)

QUESTION 43 43 B 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)

Air ejector PRM .................................................

Given the following:

The plant is at 100% power

The crew entered AOP-035, S/G TUBE LEAK

The crew has initiated Attachment 5, R-15 Monitoring

Current S/G Tube leakrate based on the R-15 reading is 40 gpd

The leakrate has risen steadily over the last four hours

The leakage cannot be isolated to one Steam Generator Which ONE of the following correctly completes the statement below?

The LOWEST change in the S/G leakrate that will require a plant shutdown be initiated IAW AOP-035 is an ADDITIONAL ________ ?

A. 20 gpd B. 40 gpd C. 60 gpd D. 80 gpd Thursday, December 03, 2015 Page 117 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 43 43 B General Discussion Answer A Discussion Incorrect. This is incorrect because there will be no plant shutdown required if SG Tube leakage rises to 60 gpd. This is plausible because AOP-035 actually has three gpd SG Tube Leakrates that require action (30, 75 and 100); and the operator may incorrectly believe that 60 gpd is one of them.

Answer B Discussion Correct. According to Attachment 5 of AOP-035 (p67 of 69; Rev 29), there are two Continuous Actions based on SG Tube Leakage that will require a plant shutdown. The first is step 20, and the second is Step 22. Step 20 checks the SG Tube Leakage greater than 100 gpd, and Step 22 checks the SG Tube Leakage greater than 75 gpd. Each of these steps requires a plant shutdown to be initiated. Because of this, if the SG Tube leakage rises by an additional 40 gpd the actions associated with CA Step 21 will be required.

Answer C Discussion Incorrect. This is incorrect because even though a plant shutdown will be required if SG Tube Leakage rises an additional 60 gpd, this is not the LOWEST change requiring a plant shutdown. This is plausible because AOP-035 actually has three gpd SG Tube Leakrates that require action (30, 75 and 100); and the operator may incorrectly believe that 100 gpd is lowest requiring a plant shutdown.

Answer D Discussion Incorrect. This is incorrect because even though a plant shutdown will be required if SG Tube Leakage rises an additional 80 gpd, this is not the LOWEST change requiring a plant shutdown. This is plausible because AOP-035 actually has three gpd SG Tube Leakrates that require action (30, 75 and 100); and the operator may incorrectly believe that 120 gpd is one of them.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to monitor changes in parameters to SG tube leakage limits associated with operating R-15 (Air Ejector Monitor).

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall bits of information (SG Leakrates requiring plant shutdowns within AOP-035), in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided AOP-035 (p67 of 69; Rev 29) 10 CFR 55.41.10 SYS039 A1.10 - 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)

Air ejector PRM .................................................

Remarks/Status 7/27/15 new question developed 9/8/15 Reviewed by Exam Supervisor, SAT 11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 118 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS059 K4.02 - Main Feedwater (MFW) System QUESTION 44 44 C Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Automatic turbine/reactor trip runback ..............................

Given the following:

The plant is operating at 100% power

An inadvertent actuation of the Feedwater Isolation Signal (FWIS) occurs Which ONE of the following describes how this actuation affects the Reactor and the Main Turbine?

A. BOTH the Reactor and the Main Turbine receive trip signals directly from the FWIS.

B. The Reactor receives a trip signal directly from the FWIS, and causes a Main Turbine trip.

C. The Main Turbine receives a trip signal directly from the FWIS and the Reactor will trip because the Main Turbine tripped with power above P-8.

D. NEITHER the Reactor nor the Main Turbine receive trip signals directly from the FWIS. However, the Reactor will trip on plant conditions created by the FWIS actuation and cause a Main Turbine trip.

Thursday, December 03, 2015 Page 119 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 44 44 C General Discussion Answer A Discussion Incorrect. This is incorrect because the reactor does not receive an automatic trip signal directly from FWIS. This is plausible because the operator may incorrectly believe that it does.

Answer B Discussion Incorrect. This is incorrect because the reactor does not receive an automatic trip signal directly from FWIS. This is plausible because the operator may incorrectly believe that the reactor does not receive an automatic trip signal directly from FWIS, and that the Main Turbine does NOT. If this were so, the Rx Trip would generate a main Turbine Trip.

Answer C Discussion Correct. According to ST-006 (p28; Rev 4) Step when the FWIS is received it will shut all of the feedwater regulating valves, feedwater bypass valves, feedwater block valves, and trip the Main Feedwater Pumps and the Turbine. According to ST-011 (p19; Rev 5) when the Main Turbine Trips with reactor power higher than P-8, a Rx Trip signal will be generated.

Answer D Discussion Incorrect. This is incorrect because the Main Turbine does receive an automatic trip signal directly from FWIS. This is plausible because the operator may incorrectly believe that it does NOT. If this were so, a Rx Trip signal would be generated on the plant conditions created by the FWIS (Lo-Lo S/G Level), and a Main Turbine Trip signal would be generated by the Rx Trip.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of MFW design feature(s) and/or interlock(s) (i.e. FWIS) which provide for an automatic turbine/reactor trip.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must simply recall what component manipulations are generated when an FWIS actuation occurs, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided ST-006 (p28; Rev 4)

ST-011 (p19; Rev 5) 10 CFR 55.41.7 SYS059 K4.02 - Main Feedwater (MFW) System Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Automatic turbine/reactor trip runback ..............................

Remarks/Status 7/30/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 120 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS061 K5.01 - Auxiliary / Emergency Feedwater (AFW) System QUESTION 45 45 C Knowledge of the operational implications of the following concepts as the apply to the AFW: (CFR: 41.5 / 45.7)

Relationship between AFW flow and RCS heat transfer .................

Given the following:

With the plant at 100% power, an ATWS event occurred

The crew entered FRP-S.1, RESPONSE TO NUCLEAR POWER GENERATION - ATWS

All SG Narrow Range levels are OFF-Scale LOW

Total AFW flow is 200 gpm Which ONE of the following identifies the action required, if any, necessary to ensure that the MINIMUM conditions are established for maintaining a Secondary Heat Sink IAW FRP-S.1?

A. Raise total AFW flow by at least 100 gpm.

B. Raise total AFW flow by at least 250 gpm.

C. Raise total AFW flow by at least 400 gpm.

D. No further action is required, a Secondary Heat Sink currently exists.

Thursday, December 03, 2015 Page 121 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 45 45 C General Discussion Answer A Discussion Incorrect. This is incorrect because a total feed flow of 600 gpm is required under the current plant conditions to establish an adequate Secondary Heat Sink. This is plausible because a normal adequate Secondary Heat sink is defined in part as total Feed Flow of 300 gpm.

Answer B Discussion Incorrect. This is incorrect because a total feed flow of 600 gpm is required under the current plant conditions to establish an adequate Secondary Heat Sink. This is plausible because a normal adequate Secondary Heat sink is defined in part as total Feed Flow of 300 gpm, and the operator may incorrectly believe that because of the ATWS event an additional 150 gpm is required.

Answer C Discussion Correct. According to CSF-3 of CSFST (p5 of 9; Rev 7) under normal conditions a Secondary Heat Sink is defined as at least one S/G Narrow Range Level > 9% [18%] or total FW flow > 300 gpm. However, an ATWS event presents the operator with a unique challenge. According to FR-S.1 (p8 of 30; Rev 22) Step 10, with all S/G Narrow Range Levels <9%, the operator is directed to establish 600 gpm total feed flow to the S/Gs until at least one S/G narrow Range Level is > 9%. According to the FR-S.1 Background Document (p19 of 44; Rev 22) this is done because ATWS analyses have shown that the specified AFW flow (discharge flow of two motor driven AFW pumps at S/G design pressure or 600 gpm) is acceptable to adequately remove the heat generated from power operation prior to reactor shutdown. If AFW flow is not greater than the specified value, it is important to raise AFW flow in order to maintain secondary heat sink. Consequently, AFW flow must be raised an additional 400 gpm.

Answer D Discussion Incorrect. This is incorrect because a Secondary Heat Sink does NOT exist. This is plausible because there is some AFW flow, and the operator may incorrectly believe that this AFW flow equates to an adequate Secondary Heat Sink.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the operational implications of the relationship between AFW flow and RCS heat transfer as it applies to an ATWS Event in which there is inadequate level in the S/G to ensure that a Secondary Heat Sink is established. Under these conditions the operator must know that AFW must be raised above that which is normally required to adequately remove heat.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall a piece of information (i.e. need a minimum of 600 gpm during ATWS Event); and then use that information under a given set of plant conditions to predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided CSF-3 of CSFST (p5 of 9; Rev 7)

FR-S.1 (p8 of 30; Rev 22) Step 10 FR-S.1 Background Document (p19 of 44; Rev 22) 10 CFR 55.41.10 SYS061 K5.01 - Auxiliary / Emergency Feedwater (AFW) System Knowledge of the operational implications of the following concepts as the apply to the AFW: (CFR: 41.5 / 45.7)

Relationship between AFW flow and RCS heat transfer .................

Remarks/Status 7/28/15 new question developed 9/8/15 Reviewed by Exam Supervisor, SAT, but with comment, Add "IAW FRP-S.1" to ensure there is no confusion between this procedure and the CSFST.

9/8/15 Incorporated Exam Supervisors comments.

9/8/15 Operations SRO reviewed question, the values put us right at 600 GPM, and FRP-S.1 states > 600 GPM, which may be splitting hairs but Thursday, December 03, 2015 Page 122 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination could cause confusion.

QUESTION 45 45 C 11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 123 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS062 K1.04 - AC Electrical Distribution System QUESTION 46 46 C Knowledge of the physical connections and/or cause-effect relationships between the ac distribution sys- tem and the following systems : (CFR:

41.2 to 41.9)

Off-site power sources ...........................................

Given the following:

The plant is at 100% power

A reactor trip has occurred Which ONE of the following identifies the breakers which will automatically CLOSE one minute after the reactor trip?

Breaker nomenclature:

52/10 4KV BUS 1-2 TIE BKR 52/12 START-UP TRANSFORMER TO 4KV BUS 2 52/17 START-UP TRANSFORMER TO 4KV BUS 3 52/19 4KV BUS 3-4 TIE BKR A. 52/10 and 52/12 B. 52/10 and 52/17 C. 52/12 and 52/19 D. 52/17 and 52/19 Thursday, December 03, 2015 Page 124 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 46 46 C General Discussion Answer A Discussion Incorrect. This is incorrect because 52/10 will not automatically close (Already is CLOSED). This is plausible because the operator may incorrectly believe that initially Breakers 52/17 and 52/19 were closed supplying Bus 3 and 4 from the SUT; and that upon Main Generator lockout, all that is needed is for the SUT to re-power Busses 1 & 2. If so, the operator would believe that 52/10 and 52/12 would auto close.

Answer B Discussion Incorrect. This is incorrect because neither breaker will automatically close (Both are already is CLOSED). This is plausible because the operator may incorrectly believe that initially Breakers 52/12 is closed and 52/17 is open supplying Bus 2 from the SUT (And buses 1, 3, 4 and 5 are powered from the UAT); and that upon Main Generator lockout, the SUT must re-power Busses 1, by 52/10 closing & 3 and 4 by 52/17 closing with 52/19 already closed. If so, the operator would believe that 52/10 and 52/17 would auto close.

Answer C Discussion Correct. According to ST-039 (p15&48; Rev 4b) during normal power operation 4KV Buses 1, 2, 4 and 5 are supplied via the UAT, and 4KV Bus 3 is supplied via the SUT. A reactor and/or turbine trip actuation (with or without a concurrent SI signal) will initiate a fast, dead bus transfer of loads from the UAT to the SUT after a 60 second time delay (provided at least 1 generator breaker, 52/8 or 52/9, is closed during the 60 second period). The transfer operation automatically trips breakers 52/7 and 52/20 while breakers 52/12 & 52/19 are simultaneously closed.

The transfer is initiated by tripping either lockout relay 86P or 86BU (which will occur after the 60 second delay).

Answer D Discussion Incorrect. This is incorrect because 52/17 will not automatically close (Already is CLOSED). This is plausible because the operator may incorrectly believe that initially Breakers 52/12 and 52/10 were closed supplying Bus 1 and 2 from the SUT (And buses 3, 4 and 5 are powered from the UAT); and that upon Main Generator lockout, the SUT must re-power Bus 3, by 52/17 closing & Bus 4 by 52/19 closing. If so, the operator would believe that 52/17 and 52/19 would auto close.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the physical connections to off-site power sources (4kV breakers 52/10, 12, 17, 19, & 20), as well as a cause-effect relationship between the ac distribution system and off-site power sources is required (230kV switchyard breaker protection scheme leading to trip of main generator).

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and use the information to predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 062 K1.04 001 Development References Student References Provided ST-039 (p15&48; Rev 4b) 10 CFR 55.41.4 SYS062 K1.04 - AC Electrical Distribution System Knowledge of the physical connections and/or cause-effect relationships between the ac distribution sys- tem and the following systems : (CFR:

41.2 to 41.9)

Off-site power sources ...........................................

Remarks/Status 7/19/15 used bank question from previous NRC exam 9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, commented that he doesnt like the question. Is there anything we can ask on Fast Bus XFER?

9/29/15 used a bank question on fast bus xfer to replace question.

Thursday, December 03, 2015 Page 125 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination 11/25/15 AOM-SHIFT approved question for 60 day submittal.

QUESTION 46 46 C Thursday, December 03, 2015 Page 126 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS062 K2.01 - AC Electrical Distribution System QUESTION 47 47 B Knowledge of bus power supplies to the following : (CFR: 41.7)

Major system loads ..............................................

Given the following:

The plant is at 100% power

Breaker 52/20, UNIT AUX TO 4KV BUS 4 BKR, trips on fault Which ONE of the following correctly completes the statement below?

Circulating Water pump(s) _________ lost power.

A. A" & "B" have B. "B" & "C" have C. "A" & "C" have D. "B" ONLY has Thursday, December 03, 2015 Page 127 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 47 47 B General Discussion Answer A Discussion Incorrect. This is incorrect because the "A" CW pump is powered from 4KV bus 1 which will remain energized. This is plausible because all the CW pumps are powered from 4KV busses.

Answer B Discussion Correct. According to OP-603 (p179; Rev 123) the electrical distribution system startup lineup lists the CW pump B breaker on 4160V Bus 4 and CW pump C breaker on 4160V Bus 5.

Answer C Discussion Incorrect. This is incorrect because the "A" CW pump is powered from 4KV bus 1 which will remain energized. This is plausible because the operator may recall that the loss of power to 4KV bus 4 will also cause a loss of 4KV bus 5, and all the CW pumps are powered from 4KV busses, but may confuse which pump is powered from which bus.

Answer D Discussion Incorrect. This is incorrect because breaker 52/24 supplies power to 4kV bus 5 from 4kV bus 4. CW pumps "B" power from 4kV bus 4 loses power, and because 4kV bus 5 also loses power, CW pump "C" is also lost.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the CW pumps power supplies.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must remember which busses have lost power when the 4KV 4 bus feeder breaker trips and CW pumps powered from those busses, and the operation of 52/24, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK Development References Student References Provided OP-603 p179 CWD B-190628-SH01344 10 CFR 55.41.7 SYS062 K2.01 - AC Electrical Distribution System Knowledge of bus power supplies to the following : (CFR: 41.7)

Major system loads ..............................................

Remarks/Status 7/23/15 used bank question 062 K2.01 1 9/8/15 Reviewed by Exam Supervisor, UNSAT, as the second part of the question gives a clue to the answer of the first part. Recommend removing second part of question, changing "pumps" to "pump(s)" in first part and first distractor to ""B" ONLY" 9/11/15 Exam team changed question based on exam supervisor comments above.

11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 128 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS063 A3.01 - DC Electrical Distribution System QUESTION 48 48 D Ability to monitor automatic operation of the DC electrical system, including: (CFR: 41.7 / 45.5)

Meters, annunciators, dials, recorders, and indicating lights .............

Given the following:

The plant is in Mode 6

A EDG is under clearance

A and B battery chargers are in service

A-1 and B-1 battery chargers are in standby Subsequently, a loss of the STARTUP TRANSFORMER occurs.

Which ONE of the following identifies the annunciator response on APP-036 over the FIRST MINUTE following the STARTUP TRANSFORMER loss?

A. BATT CHARGER B/B-1 TROUBLE alarms and then clears, ONLY B. BATT CHARGER A/A-1 TROUBLE alarms and remains in alarm BATT CHARGER B/B-1 TROUBLE alarms and remains in alarm C. BATT CHARGER B/B-1 TROUBLE alarms and remains in alarm, ONLY D. BATT CHARGER A/A-1 TROUBLE alarms and remains in alarm BATT CHARGER B/B-1 TROUBLE alarms and then clears Thursday, December 03, 2015 Page 129 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 48 48 D General Discussion Answer A Discussion Incorrect. This is incorrect because the BATT CHARGER A/A-1 TROUBLE alarm will alarm and remain in as power is not restored due to the A EDG being under clearance. The BATT CHARGERs are designed to automatically restart following power restoration. This is plausible because the BATT CHARGER B/B-1 alarm will alarm in this event.

Answer B Discussion Incorrect. This is incorrect because the BATT CHARGER B/B-1 TROUBLE alarm will clear once power is restored to the E-2 busses from the B EDG. The BATT CHARGERs are designed to automatically restart following power restoration. This is plausible because the candidate may believe that the BATT CHARGER needs to be re-set, causing the alarm to remain locked in. The STANDBY BATT CHARGER will be tripped on the AC INPUT FAILURE (APP-036-D1/ D2) which the candidate may believe causes the alarm to stay in until the charger is manually reset.

Answer C Discussion Incorrect. This is incorrect because the BATT CHARGER A/A-1 TROUBLE alarm will also alarm and remain in as power is not restored due to the A EDG being under clearance. The BATT CHARGERs are designed to automatically restart following power restoration. This is plausible because the BATT CHARGER B/B-1 alarm will alarm in this event.

Answer D Discussion Correct. According to APP-036-D1/D2 (p27, 30; Rev 88) BATT CHARGER A/A-1 TROUBLE or BATT CHARGER B/B-1 TROUBLE is received following a Loss of Off-Site Power (LOOP), the In Service Battery Charger will automatically restart. The annunciator may alert to momentary DC overvoltage condition during the battery charger start. The IN SERVICE BATT CHARGER should restart and the TROUBLE alarm will clear at that point.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to interpret the annunciators received when AC Input power is lost to the BATT CHARGERS and subsequently restored.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must analyze the plant conditions given and determine the annunciators that would be received based on those conditions, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided APP-036-D1/D2 10 CFR 55.41.7 SYS063 A3.01 - DC Electrical Distribution System Ability to monitor automatic operation of the DC electrical system, including: (CFR: 41.7 / 45.5)

Meters, annunciators, dials, recorders, and indicating lights .............

Remarks/Status 8/13/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT.

9/8/15 Operations SRO reviewed question, Editorial comments, 1) Can we add all the Auxiliaries on the SUT; 2) OR can we bold "Plant in MODE 6" 11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 130 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 48 48 D Thursday, December 03, 2015 Page 131 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS064 A2.18 - Emergency Diesel Generator (ED/G) System QUESTION 49 49 A 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)

Consequences of premature opening of breaker under load .............

Given the following:

The plant is at 100% power

OST-401-1, EDG A SLOW SPEED START testing is in progress

Current A EDG load is 2500 KW Subsequently, a Loss of Offsite Power (LOOP) occurs and the operator observes:

Bus E-1 is de-energized

EDG A stops Which ONE of the following correctly completes the statement below?

1) EDG A tripped on overspeed because ____(1)____.

2) When attempting to restart EDG A IAW OP-604, DIESEL GENERATORS A AND B, the Parallel Switch must be placed in the ____(2)____ position.

A. (1) The EDG A Output Breaker opened on overcurrent (2) ISOL B. (1) The EDG A Output Breaker opened on overcurrent (2) PARALLEL C. (1) The 480V E-1 Normal Supply Breaker (52/18B) tripped on Bus E-1 undervoltage (2) ISOL D. (1) The 480V E-1 Normal Supply Breaker (52/18B) tripped on Bus E-1 undervoltage (2) PARALLEL Thursday, December 03, 2015 Page 132 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 49 49 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to OP-604 (p115 of 323; Rev 109) a Caution in Section 8.4.1, states that "If a Loss Of Offsite Power (LOOP) occurs while the Diesel Generator is operating in parallel, the Diesel Generator's output breaker may trip due to overload.

Either the amptector at the breaker panel or the overcurrent relay at the generator panel could trip the breaker. The diesel engine may trip due to overspeed when the output breaker opens at full load." According to Step 8.4.1.3, the operator is directed to place the Parallel Switch in the ISOL position to restart EDG "A".

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because according to Step 8.4.1.3, the operator is directed to place the Parallel Switch in the ISOL position to restart EDG "A". This is plausible because the switch is a two-position Switch, and the operator may not know the significance of each position.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because according to ST-006 (p41; Rev 4) with the EDG Output Breaker closed, the E-1 undervoltage scheme is isolated, and furthermore, with the EDG powering the Bus, voltage on E-1 will be retained even after the LOOP. This is plausible because under normal circumstances (i.e. EDG is in standby) the undervoltage scheme on E-1 will operate and open the Normal Supply Breaker to E-1.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to predict what will happen to the EDG system if the output breaker is opened prematurely. This is accomplished by providing the operator with the indications that would be expected on an LOOP with the EDG paralleled to its associated Bus, and then requiring the operator to identify what caused the expected conditions (If the operator can identify what caused expected results, the operator can also predict the expected results if a cause is proposed).

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall the consequences of opening an output breaker with a generator still loaded, and remember the value to which the generator must be unloaded, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided OP-604 (p115 of 323; Rev 109)

ST-006 (p41; Rev 4) 10 CFR 55.41.7 SYS064 A2.18 - 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)

Consequences of premature opening of breaker under load .............

Remarks/Status 8/11/15 new question developed 9/8/15 Reviewed by Exam Supervisor, UNSAT, based on the second part of the question does not meet the K/A, part two of the question must be related to the first part.

9/8/15 Operations SRO reviewed question, DSRO won't know part 2. Agree UNSAT.

9/30/15 validator comments confirmed above, question changed to new question.

Thursday, December 03, 2015 Page 133 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination 11/25/15 AOM-SHIFT approved question for 60 day submittal.

QUESTION 49 49 A Thursday, December 03, 2015 Page 134 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS064 K2.02 - Emergency Diesel Generator (ED/G) System QUESTION 50 50 D Knowledge of bus power supplies to the following: (CFR: 41.7)

Fuel oil pumps ..................................................

Which ONE of the following supplies power to the A and B EDG Fuel Oil Transfer Pumps?

A. MCC-10 and MCC-9 respectively B. MCC-9 and MCC-10 respectively C. MCC-6 and MCC-5 respectively D. MCC-5 and MCC-6 respectively Thursday, December 03, 2015 Page 135 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 50 50 D General Discussion Answer A Discussion Incorrect. This is incorrect because MCC-5 and MCC-6 are the power supplies for the EDG Fuel Oil Transfer Pumps. This is plausible because MCC-10 is in the same area as MCC-5 and MCC-9 is in the same area of MCC-6. In addition, the power supplies for the EDG Fuel Oil Transfer Pumps are from MCCs.

Answer B Discussion Incorrect. See A.

Answer C Discussion Incorrect. This is incorrect because MCC-6 supplies the "B" EDG Fuel Oil Transfer Pump and MCC-5 supplies the "A" EDG Fuel Oil Transfer Pump. This is plausible because MCC-5 and MCC-6 both supply the EDG Fuel Oil Transfer pumps.

Answer D Discussion Correct. According to EDP-003, MCC Buses (p 26 & 32; Rev 68) EDG Fuel Oil Transfer Pump "A" is listed as a load on MCC-5 and EDG Fuel Oil Transfer Pump "B" is listed as a load on MCC-6.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the power supplies of the EDG Fuel Oil Transfer Pumps.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall from memory the power supplies for the EDG Fuel Transfer Pumps, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK EDG-003 5 Development References Student References Provided EDP-003 10 CFR 55.41.7 SYS064 K2.02 - Emergency Diesel Generator (ED/G) System Knowledge of bus power supplies to the following: (CFR: 41.7)

Fuel oil pumps ..................................................

Remarks/Status 8/4/15 used bank question EDG-003 5 9/8/15 Reviewed by Exam Supervisor, SAT.

11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 136 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS073 A4.01 - Process Radiation Monitoring (PRM) System QUESTION 51 51 D Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Effluent release .................................................

Given the following:

The reactor is at 100%

The crew is performing a containment vacuum relief IAW OP-921, CONTAINMENT AIR HANDLING

R-12, CV AIR OR PLANT STACK, NOBLE GAS, alarms V12-12, CV VAC RELIEF V12-13, CV VAC RELIEF APP-036-D7, AREA MONITOR HI RAD APP-036-D8, PROCESS MONITOR HI RAD Which ONE of the following correctly completes the statements below?

(1) The crew would expect to see ____(1)____ annunciator flashing.

(2) Due to the conditions above, ____(2)____ will automatically close.

A. (1) APP-036-D7 (2) V12-12 & V12-13 ONLY B. (1) APP-036-D8 (2) V12-12 & V12-13 ONLY C. (1) APP-036-D7 (2) V12-12, V12-13 & the CV Intake Damper D. (1) APP-036-D8 (2) V12-12, V12-13 & the CV Intake Damper Thursday, December 03, 2015 Page 137 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 51 51 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the CV Intake Damper closes as well. This is plausible because in actuality R-12 alarming auto closes V12-12 and 13 ONLY, and these valves closing will close the damper; and the operator may not know this.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because R-12 is a process monitor not an area monitor. This is plausible because the radiation monitor system consists of both area and process monitors where at least one of each include automatic functions (i.e. R-1, an area monitor has a control function), and the operator may not know this.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to ST-019 (p7; Rev 4) R-12 is a process monitor. According to ST-019 (p68; Rev 4) and ST-037 (p36; Rev 3) when R-12 enters the alarm condition, the CV Purge Supply and Exhaust Valves, and the Vacuum relief Valves (V12-12 and V12-13) will automatically close. According to ST-037 (p36; Rev 3) the Outdoor Air Makeup system has an air motor operated damper that opens when the containment purge fan HVE-1A or HVE-1B is started or when the Vacuum Relief system is energized. According to OP-921 (p38 of 56; Rev 55) Step 6.4.2.6, V12-12 and V12-13 are opened for this procedure. When these valves are opened, the Containment Damper will open as well. When R-12 alarms, a signal will be sent to close V12-12 and 13, and when these valves close the Containment Damper will close as well Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to monitor an effluent release in the control room, as evidenced by predicting what valves will close on an R-12 alarm.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information, and then use this information to predict how the Containment Purge System will respond overall, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2013 NRC exam Development References Student References Provided ST-019 (p7 & 68; Rev 4)

ST-037 (p36; Rev 3)

OP-921 (p38 of 56; Rev 55) 10 CFR 55.41.11 SYS073 A4.01 - Process Radiation Monitoring (PRM) System Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Effluent release .................................................

Remarks/Status 7/19/15 used bank question 2013 NRC exam 9/8/15 Reviewed by Exam Supervisor, SAT 11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 138 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS073 2.4.9 - Process Radiation Monitoring (PRM) System QUESTION 52 52 B SYS073 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:

RHR has just been placed in service IAW GP-007, PLANT COOLDOWN FROM MODE 3 TO MODE 5

RCS temperature: 248°F

RCS pressure: 349 psig Subsequently:

The crew has just entered AOP-005, RADIATION MONITORING SYSTEM, due to a valid alarm on R-12, CV AIR & Plant Vent Particulate

The crew has determined that R-12 is alarming due to an RCS leak Which ONE of the following is the correct method for implementing plant procedures to mitigate this event?

AOP-016, EXCESSIVE PRIMARY PLANT LEAKAGE A. Complete AOP-005 in its entirety, THEN Go to AOP-016 B. Perform BOTH AOP-005 and AOP-016 concurrently C. When directed exit AOP-005, and Go to AOP-016 D. Immediately exit AOP-005, and Go to AOP-016 Thursday, December 03, 2015 Page 139 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 52 52 B General Discussion Answer A Discussion Incorrect. This is incorrect because AOP-005 Attachment 12 Step 13 states, Go to AOP-016, Excessive Primary Plant Leakage, while continuing with this procedure. This is plausible because the last step of the main body of AOP-005 directs transition to procedure and step in effect.

Answer B Discussion Correct. According to AOP-005, Attachment 12 (p 32; Rev 32) Step 13, Go to AOP-016, Excessive Primary Plant Leakage, while continuing with this procedure.

Answer C Discussion Incorrect. This is incorrect because AOP-005 Attachment 12 Step 13 states, Go to AOP-016, Excessive Primary Plant Leakage, while continuing with this procedure. This is plausible because AOP-005 directs the operator to Go to AOP-016.

Answer D Discussion Incorrect. This is incorrect because AOP-005 Attachment 12 Step 13 states, Go to AOP-016, Excessive Primary Plant Leakage, while continuing with this procedure. This is plausible because the crew would meet the entry conditions for entry into AOP-016 based upon the given information that R-12 is alarming due to an RCS leak.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the mitigation strategy for a loss of coolant accident when shutdown as it relates to implementation of the abnormal operating procedure for radiation monitoring.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall from memory how AOP-005 and AOP-016 are implemented, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED AOP-005-08 12 Development References Student References Provided AOP-005 Rev. 32 10 CFR 55.41.10 SYS073 2.4.9 - Process Radiation Monitoring (PRM) System SYS073 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)

Remarks/Status 8/18/15 modified bank question AOP-005-08 12 9/8/15 Reviewed by Exam Supervisor, SAT but with following comment, ensure that the AOP-016 is the correct procedure based on plant conditions given in the stem.

9/8/15 exam team verified that AOP-016 is the correct procedure chioce.

9/8/15 Operations SRO reviewed question, concerned that when AOP-005 kicks out to AOP-016 it is pretty much the end of AOP-005. While technically concurrent at that point, close enough to lead to confusion.

11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 140 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 52 52 B Thursday, December 03, 2015 Page 141 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS076 K1.15 - Service Water System (SWS)

QUESTION 53 53 D 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)

FPS ............................................................

Which ONE of the following identifies components normally cooled by Service Water that can be cooled by the Fire Water System during an emergency?

A. The Circulating Water Pumps and the Control Room HVAC Cooling Units, ONLY.

B. The Motor Driven AFW Pumps and the Control Room HVAC Cooling Units, ONLY.

C. The Circulating Water Pumps, the Control Room HVAC Cooling Units AND the SI Pump Thrust Bearings.

D. The Motor Driven AFW Pumps, the Control Room HVAC Cooling Units AND the SI Pump Thrust Bearings.

Thursday, December 03, 2015 Page 142 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 53 53 D General Discussion Answer A Discussion Incorrect. This is incorrect because it includes the Circulating Water Pumps, and does NOT include the AFW and SI Pumps. This is plausible because Attachment 5 of AOP-022 is provided to supply emergency cooling water to the Circulating Water Pumps, however the cooling medium is the Potable Water System. Additionally, the operator may incorrectly believe that the AFW and SI Pumps are cooled by the Primary Water Pumps in an emergency.

Answer B Discussion Incorrect. This is incorrect because it does NOT include the SI Pumps. This is plausible because the SI Pumps can be supplied by the Primary Water Pumps, and the operator may incorrectly believe that this is the ONLY source that can supply cooling to the SI Pumps.

Answer C Discussion Incorrect. This is incorrect because it includes the Circulating Water Pumps, and does NOT include the AFW Pumps. This is plausible because of AOP-022 is provided to supply emergency cooling water to the Circulating Water Pumps, however the cooling medium is the Potable Water System. Additionally, the operator may incorrectly believe that the AFW are cooled by the Primary Water Pumps in an emergency (It is actually the SI Pumps that are provided cooling by the PW Pumps).

Answer D Discussion Correct. According to AOP-22-BD (p31; Rev 35) Attachment 1 provides instructions to locally align emergency cooling to the SI Pumps from the Fire Water System, PRIMARY WATER PUMP A, or PRIMARY WATER PUMP B; Attachment 2 provides instructions to locally align emergency cooling to the MDAFW Pump associated with the leaking SW Header from the Fire Water System.; and Attachment 3 aligns emergency cooling water to Control Room HVAC WCCU-1A and WCCU-1B from the Fire Water System.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the physical connections between the SWS and the FPS.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must simply recall which components, normally cooled by Service Water, can be cooled by the Fire Water System in an emergency, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided AOP-22-BD (p31; Rev 35) 10 CFR 55.41.4 SYS076 K1.15 - 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)

FPS ............................................................

Remarks/Status 7/29/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 143 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS078 K3.03 - Instrument Air System (IAS)

QUESTION 54 54 D Knowledge of the effect that a loss or malfunction of the IAS will have on the following: (CFR: 41.7 / 45.6)

Cross-tied units .................................................

Given the following:

The plant is at 100% power

D IA Compressor is running

Station Air Compressor is running Subsequently:

IA Header pressure is 79 PSIG and lowering

The crew is implementing AOP-017, LOSS OF INSTRUMENT AIR SA-5, STATION AIR TO INST AIR CROSS CONNECT SA-220, SA TO IA CROSS CONNECT BYPASS FILTER ISOLATION SA-221, SA TO IA CROSS CONNECT BYPASS FILTER ISOLATION Which ONE of the following correctly completes the statements below?

1) IAW AOP-017, the PREFERRED method to cross connect IA from Station Air is by opening

____(1)____ .

2) IA will no longer be used to supply Breathing Air to ____(2)____.

A. (1) SA-5 (2) minimize IA loads ONLY B. (1) SA-5 (2) minimize IA loads and prevent harm to users of Breathing Air C. (1) SA-220 and SA-221 (2) minimize IA loads ONLY D. (1) SA-220 and SA-221 (2) minimize IA loads and prevent harm to users of Breathing Air Thursday, December 03, 2015 Page 144 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 54 54 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. This is incorrect because SA-220 and SA-221 is the preferred flow path directed by AOP-017 to pass air through a filter to remove contaminants prior to passing into oil free IA Header. Station Air is oil lubricated, even though it is filtered and directed through Air Dryer "D" it is not for breathing air. Therefore, the RC Personnel must be notified to stop use of the Breathing Air System to prevent potential harm to users of Breathing Air. This is plausible because SA-5 would be opened if SA-220 and SA-221 could not be open IAW AOP-017 and stopping the use of Breathing Air does reduce the air load on the IA System.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because SA-220 and SA-221 is the preferred flow path directed by AOP-017 to pass air through a filter to remove contaminants prior to passing into oil free IA Header. This is plausible because SA-5 would be opened if SA-220 and SA-221 could not be open IAW AOP-017.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because Station Air is oil lubricated, even though it is filtered and directed through Air Dryer "D" it is not for breathing air. Therefore, the RC Personnel must be notified to stop use of the Breathing Air System to prevent potential harm to users of Breathing Air. This is plausible because stopping the use of Breathing Air does reduce the air load on the IA System.

Answer D Discussion Correct. According to AOP-017-BD (p 5&7; Rev 42) Step 8 Note and Step 12, the preferred step is addressed in the left column of this procedure (Open SA-220, SA-221). This will allow the Service Air to pass through a filter to remove contaminants prior to passing into oil free Instrument Air Header. The RNO addresses the alternate method which is to open cross connect valve SA-5; this agrees with the intent expressed in OP-905, Instrument & Station Air System which states: "SA-5 should not be used to interconnect the Station Air with the Instrument Air Systems unless the filter is unavailable and no other source of Instrument Air is available." The Station Air Compressor is oil lubricated and is therfore not to be used for breathing air. Therefore, the RC Personnel must be notified to stop use of the Breathing Air. Also the need to stop the use of breathing air is due to the reduction of air pressure.

Basis for meeting the KA NOTE: At one time Unit 1 and Unit 2 Air Systems could be cross connected, however with the decommissioning of Unit 1 this practice has been terminated. Cross-tied units refers to cross-tying IA and/or Station Air Systems. The KA is matched because the operator must demonstrate knowledge of the affect that cross connecting the Station Air System with the IA system will have on the Breathing Air System.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall why SA-220 and SA-221 are preferred flow path and the reasons why Breathing Air is not used when Station Air is cross connected with IA, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided AOP-017-BD Rev. 42 10 CFR 55.41.7 SYS078 K3.03 - Instrument Air System (IAS)

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

Cross-tied units .................................................

Remarks/Status 8/17/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 11/25/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 145 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 54 54 D Thursday, December 03, 2015 Page 146 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS103 2.1.7 - Containment System QUESTION 55 55 D SYS103 GENERIC Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)

Given the following:

At T=0 min, the plant experienced multiple faulted SGs in Containment

At T=8 min, the crew has transitioned to, and is implementing, FRP-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK, due to an ORANGE condition on RCS Integrity

CV Spray Flow is 0 gpm The following observations are made at subsequent times:

T=10 min Containment pressure is 10 psig, 2 HVH units running T=20 min Containment pressure is 22 psig, 1 HVH unit running T=30 min Containment pressure is 30 psig, 0 HVH units running T=40 min Containment pressure is 42 psig, 0 HVH units running Which ONE of the following identifies the EARLIEST time at which the operator would be required to suspend FRP-P.1 and enter FRP-J.1, RESPONSE TO HIGH CONTAINMENT PRESSURE?

A. T=10 min B. T=20 min C. T=30 min D. T=40 min Thursday, December 03, 2015 Page 147 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 55 55 D General Discussion Answer A Discussion Incorrect. This is incorrect because these conditions will only result in an Orange condition on Containment, which will not permit the transition to FRP-J.1. This is plausible because an Orange condition exists and the operator may incorrectly believe that Containment is a higher priority than Integrity. If so, this would the earliest time of transition to FRP-J.1.

Answer B Discussion Incorrect. This is incorrect because these conditions will only result in an Orange condition on Containment, which will not permit the transition to FRP-J.1. This is plausible because an Orange condition exists and the operator may incorrectly believe that Containment is a higher priority than Integrity. If not, conditions have degraded since T=10, and the operator may incorrectly believe that and Orange condition developed between T=10 and T=20, and that Containment is a higher priority than Integrity; or that a Red condition has developed on Containment. If so, this would the earliest time of transition to FRP-J.1.

Answer C Discussion Incorrect. This is incorrect because these conditions will only result in an Orange condition on Containment, which will not permit the transition to FRP-J.1. This is plausible because an Orange condition exists and the operator may incorrectly believe that Containment is a higher priority than Integrity. If not, conditions have degraded since T=20, and the operator may incorrectly believe that and Orange condition developed between T=20 and T=30, and that Containment is a higher priority than Integrity; or that a Red condition has developed on Containment. If so, this would the earliest time of transition to FRP-J.1.

Answer D Discussion Correct. According to OMM-22 (p26 of 56; Rev 45) Step 5.2.6.l, if any ORANGE terminus is encountered, the operator is expected to monitor all of the remaining trees and if no RED-condition is encountered, suspend any EOP in progress and perform the FRP required by the ORANGE terminus. If during the performance of an ORANGE-condition FRP, any RED-condition or higher priority ORANGE-condition arises, then the RED or higher priority ORANGE condition is to be addressed first, and the original ORANGE-condition FRP suspended. According to OMM-22 (p25 of 56; Rev 45) the six Status Trees are always evaluated in the following sequence (order of priority): (1) Subcriticality (S), (2) Core Cooling (C), (3) Heat Sink (H), (4) Integrity (P), (5) Containment (J), (6) Inventory (I). Since Integrity is a higher priority than Containment, in order for the operator to transition to FRP-J.1, a Red condition is required to exist on Containment. According to CSFST [CSF-5, Containment]

(p8 of 9; Rev 7) the only condition that will yield a Red condition is Containment Pressure greater than 42 psig. Consequently the earliest time of transition to FR-J.1 is T=40.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to evaluate plant performance in the Containment and make operational judgments (When is FRP-J.1 entry required) based on operating characteristics (Sequential loss of cooling equipment), reactor behavior (Containment pressure is rising), and instrument interpretation (CV spray flow, pressure instrumentation).

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information, and then use this information to predict an outcome (When is transition required) in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided OMM-22 (p25-26 of 56; Rev 45)

CSFST [CSF-5, Containment] (p8 of 9; Rev 7) 10 CFR 55.41.10 SYS103 2.1.7 - Containment System SYS103 GENERIC Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)

Remarks/Status 8/18/15 new question developed Thursday, December 03, 2015 Page 148 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination 9/8/15 Reviewed by Exam Supervisor, SAT QUESTION 55 55 D 9/8/15 Operations SRO reviewed question, Editorial comment, would we need to give the HVH units?

12/1/15 question modified to remove doubt as to whether or not the crew would stay in FRP-P.1, or believe they would immediately leave.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 149 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS001 K6.13 - Control Rod Drive System QUESTION 56 56 C Knowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 41.7/45.7)

Location and operation of RPIS ....................................

With the plant at 100%, the crew has entered AOP-001, MALFUNCTION OF REACTOR CONTROL SYSTEM (SEE GIVEN CONDITIONS ON NEXT PAGE)

Which ONE of the following correctly completes the statements below?

AOP-001 was entered due to ____(1)____. The CRS directs I&C personnel to go to the

____(2)____ to investigate the cause.

A. (1) a dropped control rod (2) 1BD Power Cabinet in the Rod Control Room B. (1) a dropped control rod (2) Logic Cabinet in the Rod Control Room C. (1) an IRPI failure (2) IRPI System Rack in the Unit 2 Cable Spread Room D. (1) an IRPI failure (2) IRPI System Rack in the Hagan Room Thursday, December 03, 2015 Page 150 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 56 56 C General Discussion Answer A Discussion Incorrect. This is incorrect because there was no indication of a prompt drop on reactor power indications. The second part is incorrect because this is the wrong Power Cabinet. This is plausible because the IRPI indication failure is the same as a dropped rod and would be correct if there was a prompt drop in reactor power. The second part is plausible because I&C would be sent to a power cabinet. (The distractor uses a wrong power cabinet to ensure distinction between the second part of this distractor and the second part of the B distractor.)

Answer B Discussion Incorrect. This is incorrect because there was no indication of a prompt drop on reactor power indications. This is plausible because the IRPI indication failure is the same as a dropped rod and would be correct if there was a prompt drop in reactor power. The second part is a correct cabinet to investigate for a rod control failure and associated location if it was a dropped rod.

Answer C Discussion Correct. According to AOP-001 (p 6; Rev 33) Step 14, Determine the status of IRPI as follows:

a. Analyze the below indications for an IRPI problem:

IRPI Indication oIndicator off-scale High OR Low with NO flux effects

Dropped Rod Indication with no flux changes According to ST-009 (p 4; Rev 5), The IRPI System rack mounted electronic equipment is located in the RPI Racks in the Unit 2 Cable Spread Room (Computer Room).

Answer D Discussion Incorrect. This is incorrect because the IRPI System rack is located in Unit 2 Cable Spread Room. This is plausible because other system racks such as RPS racks are located in the Hagan Room.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the effect of a failure of IRPI for a control rod and the location of IRPI equipment.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must analyze plant indications and determine the cause, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided AOP-001 Rev. 33 ST-009 Rev. 5 10 CFR 55.41.7 SYS001 K6.13 - Control Rod Drive System Knowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 41.7/45.7)

Location and operation of RPIS ....................................

Remarks/Status 7/28/15 New question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 151 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 56 56 C Thursday, December 03, 2015 Page 152 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS002 K4.01 - Reactor Coolant System (RCS)

QUESTION 57 57 D Knowledge of RCS design feature(s) and/or interlock(s) which provide for the following : (CFR: 41.7)

Filling and draining the RCS .......................................

Given the following:

GP-001, FILL AND VENT OF THE REACTOR COOLANT SYSTEM, is in progress

The crew is preparing to start an RCP to sweep gases in support of venting the RCS

The PZR is at 100%

S/G secondary water temperatures are 52°F greater than RCS cold leg temperatures Which ONE of the following correctly completes the statements below?

1) IAW GP-001, the FIRST method used to reduce the delta T between the S/G and the RCS cold leg temperatures is to ____(1)____ .

2) This is done ____(2)____ when a RCP is started.

A. (1) drain and fill the S/Gs (2) to ensure number 1 seal delta P remains > 210 psid B. (1) ADJUST HIC-758, RHR HX DISCH FLOW (2) to ensure number 1 seal delta P remains > 210 psid C. (1) drain and fill the S/Gs (2) to prevent a low temperature overpressure event D. (1) ADJUST HIC-758, RHR HX DISCH FLOW (2) to prevent a low temperature overpressure event Thursday, December 03, 2015 Page 153 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 57 57 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the reason for the delta T limitation is to prevent a low temperature overpressure event when the RCP is started. This is plausible because maintaining number 1 seal delta P > 210 psid is a requirement anytime a RCP is running.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because GP-001 directs use of HIC-758 to attempt to heat up the RCS first. This is plausible because draining and filling S/Gs is the second method used IAW GP-001.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to GP-001 (p 74; Rev 68) Step 6.5.25, if any S/G temperature is more than 50°F above its associated Loop Cold Leg temperature, THEN perform the following: a. Adjust HIC-758 (RHR HX DISCH FLOW), to raise RCS temperature within the allowable heatup rates. B. IF RCS temperature can not be raised sufficiently to achieve a 50°F or less differential temperature, THEN perform the following: (2) Cool the S/Gs by filling and draining per the following:

-OP-402, Auxiliary Feedwater System, for filling

-OP-406, Steam Generator Blowdown/Wet Layup System, for draining According to OP-101 (p 12; Rev 74) Step 5.2.1.3, to avoid an overpressure transient, an RCP shall NOT be started when the RCS is water solid and any steam generator temperature is 50°F greater than the temperature of the RCS.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the RCS design feature related to filling and venting the RCS in that the S/G secondary water temperature must be within 50°F of the RCS cold leg temperatures prior to starting a RCP to prevent a low temperature overpressure event.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall from memory the methods used in GP-001 to equalize S/G secondary water temperature with RCS cold leg temperature prior to starting a RCP and the reason for this requirement, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK GP-001-03 15 Development References Student References Provided GP-001 Rev. 68 OP-101 Rev. 74 10 CFR 55.41.10 SYS002 K4.01 - Reactor Coolant System (RCS)

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

Filling and draining the RCS .......................................

Remarks/Status 8/17/15 used bank question GP-001-03 15 9/8/15 Reviewed by Exam Supervisor, with Editorial comment, 1) Add "PZR is at 100%

9/11/15 exam team incorporated change as proposed by exam supervisor.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 154 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 57 57 D Thursday, December 03, 2015 Page 155 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS011 A1.02 - Pressurizer Level Control System (PZR LCS)

QUESTION 58 58 B Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including: (CFR: 41.5 / 45.5)

Charging and letdown flows ........................................

Given the following:

The plant is at 35% power and stable

The A Charging Pump is running in AUTO

The B Charging Pump is running in MANUAL and adjusted to 50% output on Charging Pump Speed Controller SC-151

Letdown flow is 105 gpm Subsequently:

Letdown flow is lowered 45 gpm

Reactor power is lowered to 30% and stabilized

Pressurizer Level lowers to, and has stabilized at 31.5%

Which ONE of the following correctly completes the statements below?

1) The Pressurizer Level Control System ____(1)____ operating as expected for these plant conditions.

While Letdown was being lowered APP-003-F3, CHG PMP LO SPEED, alarmed.

2) According to OP-301, CHEMICAL AND VOLUME CONTROL SYSTEM, the operator will be required to adjust the speed of the ____(2)____ Charging Pump to clear this alarm?

A. (1) is (2) A B. (1) is (2) B C. (1) is NOT (2) A D. (1) is NOT (2) B Thursday, December 03, 2015 Page 156 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 58 58 B General Discussion Answer A Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the "B" Charging Pump is in Manual and must be adjusted. This is plausible because with the current Charging Pump arrangement it is the "A" Charging Pump causing the low speed alarm, and the operator may incorrectly believe that the speed of this pump needs to adjusted.

Answer B Discussion Correct. 1st part correct, 2nd part correct. According to OP-301 (p20 of 140; Rev 112), the Pressurizer Program Level ramps linearly between zero and 100% power. With Rx power at 0% (Tavg 547°F), Pressurizer level is 22.2%, and with Rx power at 100% (Tavg 575.9°F), Pressurizer level is 53.3%. Consequently, when power is lowered to 30%, Pressurizer level will stabilize at 31.5%.

([53.3%-22.2% level]/100% power x 30% power) + 22.2% level = 31.5% level According to OP-301 (p29-30 of 140; Rev 112) IF the CHG PMP LO SPEED alarm (APP-003-F3) is actuated, THEN GO TO an earlier section of OP-301. According to OP-301 (p24 of 140; Rev 112) IF the CHG PMP LO SPEED alarm (APP-003-F3) is actuated from the Charging Pump on automatic control with two Charging Pumps running, THEN PERFORM the following: REDUCE the second Charging Pump speed in manual to clear the alarm.

Answer C Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and D.

Answer D Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the PLCS is operating exactly as expected. This is plausible because the programmed band at 30% is between 22.2-53.3%, and the operator could predict an incorrect value of the linearly varying programmed Pressurizer level. Additionally, if the stabilized value of Pressurizer Level were changed to 34%, this would be correct.

Basis for meeting the KA The KA is matched because the operator must demonstrate the Ability to monitor changes in parameters to prevent exceeding design limits (programmed Pzr Level) associated with operating the charging and letdown flow controls. In this case the operator must demonstrate knowledge of the allowable band for the Pressurizer Programmed Level and predict what the level should be when power is adjusted, and then identify which Charging Pump speed must be adjusted during the process of reducing normal letdown flow.

Basis for Hi Cog The question is at the Comprehensive Analysis cognitive level because the operator must recall bits of information, and use the information to predict an outcome (correct programmed level) and determine required actions (Which Charging Pump speed to adjusted) in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided OP-301 (p20, 24, 29-30 of 140; Rev 112) 10 CFR 55.41.10 SYS011 A1.02 - Pressurizer Level Control System (PZR LCS)

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

Charging and letdown flows ........................................

Remarks/Status 8/19/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, UNSAT with following comment, 1) All bands are allowable, since they are w/in the 6-20. Need to use "minimum" and "maximum" in the stem to prevent this.

9/8/15 Operations SRO reviewed question, commented that question should be rewritten as the crew will still attempt to maintain 8-13 GPM Thursday, December 03, 2015 Page 157 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 58 58 B 9/11/15 exam team edited the question to add MINIMUM and MAXIMUM in front of "allowed flows" since we are looking for the band 10/14/15 due to validation issues with the use of the expanded band, this question was changed to move away from that.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 158 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS014 K5.01 - Rod Position Indication System (RPIS)

QUESTION 59 59 C Knowledge of the operational implications of the following concepts as they apply to the RPIS: (CFR: 41.5 / 45.7)

Reasons for differences between RPIS and step counter ................

Given the following:

With the plant at 100% power, a grid disturbance caused a load rejection to 80%

On Control Bank D, a Group 1 rod failed to move with the group Which ONE of the following describes the relative comparison of the various rod position indications and the reason?

A. IRPI, ERFIS and Bank D Group 1 Step Counter are all higher than the Bank D Group 2 Step counter because the affected rod failed to move.

B. IRPI and the Bank D Group 1 Step Counter are higher than ERFIS for the affected rod and the Bank D Group 2 Step Counter because the affected rod failed to move.

C. IRPI and ERFIS for the affected rod are matched and indicate higher than both Bank D Group Step Counters because Group Step Counters are driven by bank demand.

D. IRPI indicates higher than ERFIS for the affected rod and both Bank D Group Step Counters because ERFIS is driven by individual demand and the Group Step Counters by bank demand.

Thursday, December 03, 2015 Page 159 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 59 59 C General Discussion Answer A Discussion Incorrect. This is incorrect because IRPI and ERFIS will be matched but higher than the group step counters. This is plausible because the affected rod will actually be higher than group 2 control rods.

Answer B Discussion Incorrect. This is incorrect because IRPI and ERFIS will be matched and higher than both group step counters. This is plausible because the affect rod will actually be higher than group 2 control rods Answer C Discussion Correct. According to ST-009 (p 1; Rev 5), the Rod Control System provides the operator with a "demand rod bank position" (step counters) and not actual individual rod positions. The Individual Rod Position Indication (IRPI) system provides an actual position information for each control rod.

Answer D Discussion Incorrect. This is incorrect because IRPI feeds ERFIS. This is plausible because IRPI for the affected control rod will actually be higher than the other rods in its group and the other group.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of the response of IRPI versus step counter indication for a misaligned control rod.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must analyze the effect on IRPI and step counter indications for a misaligned rod during a load rejection event, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 014 K5.01 1 Development References Student References Provided ST-009 Rev. 5 10 CFR 55.41.7 SYS014 K5.01 - Rod Position Indication System (RPIS)

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

Reasons for differences between RPIS and step counter ................

Remarks/Status 7/27/15 used bank question 014 K5.01 1 9/8/15 Reviewed by Exam Supervisor, SAT 11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 160 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS017 A4.01 - In-Core Temperature Monitor (ITM) System QUESTION 60 60 C Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Actual in-core temperatures .......................................

Given the following:

The crew is performing FRP-C.1, RESPONSE TO INADEQUATE CORE COOLING Which ONE of the following correctly completes the statements below?

1) IAW FRP-C.1 the crew will monitor ____(1)____ for inadequate core cooing conditions.

2) The MINIMUM temperature at which transition to SACRM-1, SEVERE ACCIDENT CONTROL ROOM MANAGEMENT INITIAL RESPONSE, is required with available RCPs running is ____(2)____.

A. (1) loop RTDs (2) greater than 1200°F and rising B. (1) loop RTDs (2) greater than 2300°F and rising C. (1) Core Exit TCs (2) greater than 1200°F and rising D. (1) Core Exit TCs (2) greater than 2300°F and rising Thursday, December 03, 2015 Page 161 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 60 60 C General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because FRP-C.1 directs use of Core Exit TCs for monitoring temperature. This is plausible because loop RTD temperatures are provided on the ICCM Panel and hot leg temperatures are used in later steps of FRP-C.1 to ensure RCS pressure is less than the shutoff head of the RHR pump as a check that core cooling has been restored prior to exiting the procedure.

1200°F is the correct temperature.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and D.

Answer C Discussion Correct. 1st part correct, 2nd part correct. According to FRP-C.1-BD (p 14; Rev 56) Step 6, check Core Exit T/Cs - Less than 1200°F, to check if excessive core exit temperatures, symptomatic of an inadequate core cooling condition, still exist. According to FRP-C.1-BD (p 42; Rev 56)

Step 22, if the operator enters this step and core exit T/C temperatures are greater than 1200°F and rising and RCPs are running in all available RCS cooling loops, the operator should transition to the SAMGs. This condition indicates that all attempts to restore core cooling have failed and core damage cannot be prevented and the operator should go to SAMGs.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because 1200°F and rising is the minimum temperature IAW FRP-C.1 which core damage cannot be prevented. This is plausible because 2300°F is the maximum temperature range for core exit T/Cs.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability to monitor in-core temperatures in the control room. This is accomplished by the candidate determining which parameter is monitored for adequate core cooling IAW FRP-C.1.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall from memory what parameter is monitored in FRP-C.1 to determine inadequate core cooling conditions exist or not and at what temperature above which core damage may occur if core cooling cannot be restored, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 017 A1.01 1 Development References Student References Provided FRP-C.1-BD Rev. 56 10 CFR 55.41.10 SYS017 A4.01 - In-Core Temperature Monitor (ITM) System Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Actual in-core temperatures .......................................

Remarks/Status 8/17/15 used bank question 017 A1.01 1 9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, Editorial, can we change the stem statement 2) to ask at what temp would we transition to SAMGs?

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 162 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS027 K2.01 - Containment Iodine Removal System (CIRS)

QUESTION 61 61 A Knowledge of bus power supplies to the following: (CFR: 41.7)

Fans ...........................................................

Which ONE (1) of the following is the power supply for HVE-3, CV Air Iodine Removal Fan?

A. MCC-5 B. MCC-6 C. MCC-9 D. MCC-10 Thursday, December 03, 2015 Page 163 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 61 61 A General Discussion Answer A Discussion Correct. According to EDP-003 (p23 of 68; Rev 68) HVE-3 is powered from MCC-5.

Answer B Discussion Incorrect. This is incorrect because HVE-3 is powered from MCC-5. This is plausible because HVE-3 & 4 dont have train designators, and there is no convention at RNP that odd numbers are A Train.

Answer C Discussion Incorrect. This is incorrect because HVE-3 is powered from MCC-5. This is plausible because MCC-9 is vital power for Train B, fed from MCC-6. Many safety-related components are fed from these "secondary" MCCs, such as AFW valves, Service Water valves, etc.

Answer D Discussion Incorrect. This is incorrect because HVE-3 is powered from MCC-5. This is plausible because MCC-10 is vital power from Train A, fed from MCC-5.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of bus power supplies to HVE-3.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must simply recall the power supply to HVE-3 in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2014 NRC exam Development References Student References Provided EDP-003 (p23 of 68; Rev 68) 10 CFR 55.41.8 SYS027 K2.01 - Containment Iodine Removal System (CIRS)

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

Fans ...........................................................

Remarks/Status 7/19/15 used bank question from 2014 NRC exam 9/8/15 Reviewed by Exam Supervisor, SAT, but commented that we could change it to a fan that's power supply is more readily known.

9/9/10 discussed with exam team and reviewed 2014 NRC exam scores. Question did well on 2014 exam. Decided to leave unchanged.

9/8/15 Operations SRO reviewed question, Editorial comment, is there a more common fan?

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 164 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS029 2.1.23 - Containment Purge System (CPS)

QUESTION 62 62 B SYS029 GENERIC Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6)

Consider the following plant evolutions:

1. A Containment Entry is being made at power

2. Core Alterations are taking place with the Equipment Hatch removed Which ONE of the following identifies the evolutions, if any, in which Containment Purge MUST be in operation IAW OP-921, CONTAINMENT AIR HANDLING?

A. 1 ONLY B. 2 ONLY C. BOTH 1 and 2 D. NEITHER 1 NOR 2 Thursday, December 03, 2015 Page 165 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 62 62 B General Discussion Answer A Discussion Incorrect. This is incorrect because CV purge is not required for Containment Entry at power. This is plausible because the operator may incorrectly believe that CV purge is necessary to maintain habitability requirements for emergency entry of CV at power.

Answer B Discussion Correct. According to Technical Specification LCO 3.9.7, Containment Purge Filter System (p3.9-11; Amendment 195), the Containment Purge Filter System shall be OPERABLE and operating during movement of recently irradiated fuel assemblies in containment. According to GP-010 (p78 of 88; Rev 87), Attachment 10.7, IF Core Alterations or movement of irradiated fuel in Containment AND the equipment hatch is removed, THEN ENSURE CV Purge and Filter System is in operation with ventilation isolation from R-11 and R-12 defeated.

Answer C Discussion Incorrect. This is incorrect because CV purge is not required for Containment Entry at power. This is plausible because the operator may incorrectly believe that CV purge is necessary to maintain habitability requirements for emergency entry of CV at power.

Answer D Discussion Incorrect. See A and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate the Ability to perform specific system and integrated plant procedures during all modes of plant operation. This is accomplished by requiring the operator to identify when specific procedure OP-921 must be used to place CV Purge in operation during Refueling, Plant Cooldown and At Power evolutions.

Basis for Hi Cog The question is at the Memory because the operator must recall bits of information (when Containment Purge System is required to be in operation), in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided LCO 3.9.7 (p3.9-11; Amendment 195)

GP-010 (p78 of 88; Rev 87) 10 CFR 55.41.10 SYS029 2.1.23 - Containment Purge System (CPS)

SYS029 GENERIC Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6)

Remarks/Status 8/18/15 used bank question G2.3.9 2 9/8/15 Reviewed by Exam Supervisor, UNSAT, as operators are not required to memorize exact procedure steps.

9/8/15 Operations SRO reviewed question, agree UNSAT, too much memory, would be high miss.

9/14/15 new question developed to replace.

9/30/15 based off all validators feedback, the plant with normal do this while it is not required by procedure for the accumulator venting.

Therefore the question was changed to a 2 by and evolution 3 removed.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 166 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 62 62 B Thursday, December 03, 2015 Page 167 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS033 K3.02 - Spent Fuel Pool Cooling System (SFPCS)

QUESTION 63 63 A Knowledge of the effect that a loss or malfunction of the Spent Fuel Pool Cooling System will have on the following: (CFR: 41.7 / 45.6)

Area and ventilation radiation monitoring systems .....................

Given the following:

A leak has occurred on the Spent Fuel Pit Heat Exchanger

No fuel handling activities are in progress

SFP level is 36 feet 6 inches and lowering slowly Which ONE of the following correctly completes the statements below?

1) ____(1)____ will provide indication of rising radiation levels due to lowering SFP level.

2) When R-21, Fuel Handling Building Upper Level noble gas monitor alarms

____(2)____ will automatically trip.

A. (1) R-5, Spent Fuel Pit Area (2) HVE-15 B. (1) R-5, Spent Fuel Pit Area (2) HVE-15A C. (1) R-30, Fuel Handling Building Lower Level (2) HVE-15 D. (1) R-30, Fuel Handling Building Lower Level (2) HVE-15A Thursday, December 03, 2015 Page 168 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 63 63 A General Discussion Answer A Discussion Correct. 1st part correct. 2nd part correct. According to AOP-005-BD (p 9; Rev 32) Attachment 5 Step 4 basis, low level in the Spent Fuel Pit may be the cause of the Radiation Monitor Alarm. Attachment 5 of AOP-005 is for Area Monitor R-5, Spent Fuel Pit Area. Therefore, lowering level in SFP may result in rising area radiation levels which will be detected by R-5. According to systems LP-19 RMS Rev 4. and APP-010-E7, HVE-14/15 AIR FLOW LOST/OVLD a high alarm on R-21 will result in a trip of HVE-15..

Answer B Discussion Incorrect. 1st part correct. 2nd part incorrect. 1st part, According to AOP-005-BD (p 9; Rev 32) Attachment 5 Step 4 basis, low level in the Spent Fuel Pit may be the cause of the Radiation Monitor Alarm. Attachment 5 of AOP-005 is for Area Monitor R-5, Spent Fuel Pit Area.

Therefore, lowering level in SFP may result in rising area radiation levels which will be detected by R-5. 2nd part incorrect, this is plausible because HVE-15A is the charcoal filtered fan that serves the spent fuel pool. However, it is only run during fuel handling activities or during testing of the ventilation system. A high alarm on R-21 will not result in a trip of HVE-15A if running. Reference systems LP-19 RMS Rev 4.

This is plausible if the candidate incorrectly determines that a high alarm on R-21 will result in a trip of either HVE-15 or HVE-15A.

Answer C Discussion Incorrect. 1st part wrong. 2nd part correct. This is incorrect because R-5 monitors the area radiation levels for the SFP. This is plausible because R-30 monitors ventilation exhaust from the Fuel Handling Building Lower Level in the area below the SFP by the Gas Decay Tanks.

The SFP Heat Exchanger is at the same elevation as the Gas Decay tanks in the Fuel Handling Building.

Answer D Discussion Incorrect. 1st part wrong. 2nd part wrong. See B & C.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge (which rad monitors monitor what areas/processes) of lowering level in SFP due to a SFP Cooling System malfunction.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall from memory which area radiation monitor will provide indication of a lowering level in the SFP and which process monitor will monitor a leak from the SFP Heat Exchanger.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided AOP-005-BD Rev. 32 ST-019 Rev. 4 10 CFR 55.41.4 SYS033 K3.02 - Spent Fuel Pool Cooling System (SFPCS)

Knowledge of the effect that a loss or malfunction of the Spent Fuel Pool Cooling System will have on the following: (CFR: 41.7 / 45.6)

Area and ventilation radiation monitoring systems .....................

Remarks/Status 8/12/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, may be missed due to the R-20 part of question, changed to contrast a different concept.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 169 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 63 63 A Thursday, December 03, 2015 Page 170 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS035 A2.05 - Steam Generator System (S/GS)

QUESTION 64 64 A Ability to (a) predict the impacts of the following mal- functions or operations on the GS; 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)

Unbalanced flows to the 5/Gs ......................................

Given the following:

The plant is operating at 100% power

The SDAFW Pump is OOS Subsequently:

A large Steam Break occurs inside Containment on the B Steam Generator

V2-16A, AFW Header Discharge Valve, has failed CLOSED

The B MDAFW Pump trips on auto-start Which ONE of the following correctly completes the statements below?

Without operator action a Secondary Heat Sink ____(1)____ be established.

IAW EOP-E-0 Foldout Criteria, when the operator does take action, the operator will

____(2)____.

A. (1) will (2) isolate AFW flow to the B S/G B. (1) will (2) reduce AFW flow in the B S/G to 60 gpm C. (1) will NOT (2) isolate AFW flow to the B S/G D. (1) will NOT (2) reduce AFW flow in the B S/G to 60 gpm Thursday, December 03, 2015 Page 171 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 64 64 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to ST-042 (p7, 10 & 14; Rev 5), on the trip both MDAFW Pumps will auto start, the AFW Header Discharge will auto open, and the AFW control valves are preset to control flow at 325 gpm from each pump. When the "B" MDAFW Pump trips, 325 gpm of AFW flow will be delivered to the "B" and "C" S/G ("A" S/G Iso valve failed CLOSED) via the "A" MDAFW Pump, however because the pressure in the "B" S/G is lower than that of the "C" S/G, all, or most of the AFW flow will enter the "B" S/G (unbalanced flows). According to EOP-E-0 (p19 of 38; Rev 6) Step 17, a heat sink is defined as (1) Total AFW flow to S/G(s) GREATER THAN 300 GPM or (2) S/G Narrow Range level in at least one S/G GREATER THAN 9%. Since there is a total of 325 gpm of AFW flow, a heat sink will exist. According to Item #2 of the EOP-E-0 Foldout Page (pFoldout; Rev 6), The Faulted S/G Criteria will require the operator to take action after Step 5. Accordingly, IF ANY S/G pressure is lowering in an uncontrolled manner OR has completely depressurized, AND ANY S/G is NOT Faulted (Which is the case); the operator will RESET SI, and CLOSE AFW Header Discharge Valve for Faulted S/G(s):

V216B (S/G B). The operator will then PERFORM Supplement D, Deenergizing AFW Valves For AFFECTED S/G, and MAINTAIN total feed flow GREATER THAN 300 gpm to the "C" S/G UNTIL the "C" S/G Narrow Range level is GREATER THAN 9% [18%].

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the operator will isolate AFW flow to the "B" S/G based on the EOP-E-0 Foldout Page Item #2. This is plausible because if all three S/Gs were depressurizing the operator would be directed to reduce AFW flow to 60 gpm per S/G as the only means available to control the cooldown rate; and the operator may incorrectly believe that it applies here. If so, the operator would reduce the "B" S/G flow to 60 gpm and allow the remaining AFW flow making up the Heat Sink, to flow into the "C" S/G.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because an AFW flow of greater than 300 gpm creates a Heat Sink, and this will occur without operator action (even though the AFW flow all or mostly flowing into a Faulted S/G). This is plausible because the operator may not know the 300 gpm threshold, or incorrectly believe that the Heat Sink cannot be defined by a Faulted S/G.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate the ability predict the impacts of the unbalanced flows to the S/Gs on whether or not a Secondary Heat Sink exists; and then use the EOP-E-0 Foldout Page to correct, control, or mitigate the consequences.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and analyze plant conditions to determine the required actions, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided ST-042 (p7, 10 & 14; Rev 5)

EOP-E-0 (p19 & Foldout; Rev 6) 10 CFR 55.41.10 SYS035 A2.05 - Steam Generator System (S/GS)

Ability to (a) predict the impacts of the following mal- functions or operations on the GS; 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)

Unbalanced flows to the 5/Gs ......................................

Remarks/Status 8/19/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, commented that it is technically correct, but does that make it a good question? Lot of memorization.

Thursday, December 03, 2015 Page 172 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 64 64 A 9/30/15 changed question out with new question.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 173 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination SYS068 K1.07 - Liquid Radwaste System (LRS)

QUESTION 65 65 D Knowledge of the physical connections and/or cause effect relationships between the Liquid Radwaste System and the following systems:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

Sources of liquid wastes for LRS ...................................

Which ONE of the following discharges to the Reactor Coolant Drain Tank?

A. Charging pump seals B. RCV-609, CC SURGE TANK VENT C. CVC-203 A/B, LETDOWN RELIEF VALVES D. CVC-389, EXCESS LETDOWN DIVERSION Thursday, December 03, 2015 Page 174 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 65 65 D General Discussion Answer A Discussion Incorrect. This is incorrect because the charging pumps seals drain to the AUX BLDG Sump Tank which drains to the WHUT. This is plausible because this is another tank in the Liquid Rad Waste System.

Answer B Discussion Incorrect. This is incorrect because RCV-609 drains to the WHUT. This is plausible because this is another tank in the Liquid Rad Waste System.

Answer C Discussion Incorrect. This is incorrect because CVC-203 A/B relieve to the PRT. This is plausible because this is another tank in the Liquid Rad Waste System.

Answer D Discussion Correct. According to ST-023 (p8; Rev 2) and ST-021 (p81, Rev 3a) the Excess Letdown diversion is sent to the RCDT via CVC-389.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the physical connections into the RCDT.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must simply recall where the Letdown Excess Valve diverts to in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2013 NRC exam Development References Student References Provided ST-023 (p8; Rev 2)

ST-021 (p81, Rev 3a) 10 CFR 55.41.3 SYS068 K1.07 - Liquid Radwaste System (LRS)

Knowledge of the physical connections and/or cause effect relationships between the Liquid Radwaste System and the following systems:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

Sources of liquid wastes for LRS ...................................

Remarks/Status 7/19/15 used bank question used on 2013 NRC exam 9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, Editorial comments, can be aligned? Normally and fails to VCT?

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 175 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.1 2.1.29 - GENERIC - Conduct of Operations QUESTION 66 66 C Conduct of Operations Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. (CFR: 41.10 / 45.1 / 45.12)

Which ONE of the following correctly completes the statements below?

IAW OPS-NGGC-1303, VERIFICATION PRACTICES, guidance for waiving Independent Verification (IV) requirements states that the IV may be waived if the operators were expected to ____(1)____ to complete the verification.

In these cases, ____(2)____ Verification will be used as an alternative to Independent Verification.

A. (1) be in an area where the dose rate is greater than 100 mrem/hour (2) Concurrent B. (1) receive a total exposure of greater than 10 mrem (2) Concurrent C. (1) receive a total exposure of greater than 10 mrem (2) Functional D. (1) be in an area where the dose rate is greater than 100 mrem/hour (2) Functional Thursday, December 03, 2015 Page 176 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 66 66 C General Discussion Answer A Discussion Incorrect. 1st part wrong. 2nd part wrong. See B and D.

Answer B Discussion Incorrect. 1st part correct. 2nd part wrong. This is incorrect because the alternative verification method, when waiving an Independent Verification requirement is a Functional Verification (A Concurrent Verification would leave the operators no better off with respect to exposure). This is plausible because the Functional and Concurrent Verifications are two additional types of verification, and the operator may not know the difference between the two.

Answer C Discussion Correct. 1st part correct. 2nd part correct. According to OPS-NGGC-1303 (p18 of 29; Rev 12) an Independent Verification may be waived if excessive radiation exposures would result. As a guideline, an exposure of greater than 10 mrem to conduct the INDEPENDENT/CONCURRENT VERIFICATION would be considered excessive. Individual situations should be determined on a case-by-case basis by the respective supervisor. In these situations, an alternate means such as FUNCTIONAL VERIFICATION not involving radiation exposure (such as observing process parameters) should be utilized.

Answer D Discussion Incorrect. 1st part wrong. 2nd part correct. This is incorrect because the criteria is NOT dose rate based. This is plausible because dose rate and exposure are closely related concepts, and the operator may incorrectly believe that the waiver criteria is based on doserate, rather than total exposure.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of how to conduct component lineups that require Independent Verifications; specifically under what conditions these requirements may be waived, and alternatives.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall bits of information, and then utilize this information to analyze a set of conditions and predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided OPS-NGGC-1303 (p18 of 29; Rev 12) 10 CFR 55.41.10 GEN2.1 2.1.29 - GENERIC - Conduct of Operations Conduct of Operations Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. (CFR: 41.10 / 45.1 / 45.12)

Remarks/Status 7/30/15 new question developed 9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, verify that OPS-NGGC-1303 is still a valid procedure.

9/10/15 exam team verified OPS-NGGC-1303 is still valid.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 177 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 66 66 C Thursday, December 03, 2015 Page 178 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.1 2.1.36 - GENERIC - Conduct of Operations QUESTION 67 67 D Conduct of Operations Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7)

Given the following plant conditions:

- The plant is in MODE 6 for refueling operations

- Core alterations are in progress Which ONE (1) of the following completes the statements below?

IAW LCO 3.9.2, NUCLEAR INSTRUMENTATION, ___(1)___ source range neutron flux monitor(s) shall be OPERABLE.

NI-51 and NI-52, ___(2)___ capable of providing audible indication inside CV.

A. (1) one (2) are B. (1) two (2) are C. (1) one (2) are NOT D. (1) two (2) are NOT Thursday, December 03, 2015 Page 179 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 67 67 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because NI-51 and NI-52(PAM Monitors) are NOT capable of providing audible indication inside the CV. This is plausible if the candidate incorrectly believes that the PAM monitors, NI-51 and NI-52 provide the exact same indications that the NI-31 or NI-32 can provide, however only NI-31 or NI-32 can provide the audible indication inside the CV.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because two SR NI's shall be OPERABLE per ITS 3.9.2. This is plausible if the candidate incorrectly believes that only one is required, since the control room also has the N51 and N52 (PAM Monitors) available for indication.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to Technical Specification LCO 3.9.2 (p3.9-2; Amendment 190), Two source range neutron flux monitors shall be OPERABLE during Mode 6. According to ST-010 (p2, 8, 11 and 19; Rev 3), the NIS consists of ten independent channels: Two SR Channels (N31 and N32), two IR Channels, four PR Channels, and two R.G. 1.97 Channels. The R.G 1.97 (N51 and N52) channels, or PAM instruments, include both Source and Wide Range channels. The N31 and N32 have audible count rate available in the CV, while the N51 and N52 do not.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of procedures (Tech Specs with < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ACTION statements) and limitations (minimum number of Source Range Instruments needed to conduct Core Alterations) involved in core alterations.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall bits of information in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK G2.1.40 1 Development References Student References Provided TS LCO 3.9.2 (p3.9-2; Amendment 190)

ST-010 (p2, 8, 11 and 19; Rev 3) 10 CFR 55.41.10 GEN2.1 2.1.36 - GENERIC - Conduct of Operations Conduct of Operations Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7)

Remarks/Status 7/19/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT, but with the comment/question, are RO candidates required to memorize ITS actions?

9/8/15 exam team response, yes they are less than one hour action statements.

9/8/15 Operations SRO reviewed question, similar comment, is the below the line info RO required?

9/10/15 exam team response remains that < one hour action statements are RO required knowledge.

12/1/15 AOM-Shift found an appropriate bank question to replace this.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 180 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 67 67 D Thursday, December 03, 2015 Page 181 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.1 2.1.45 - GENERIC - Conduct of Operations QUESTION 68 68 B Conduct of Operations Ability to identify and interpret diverse indications to validate the response of another indication. (CFR: 41.7 / 43.5 / 45.4)

Given the following:

A plant startup is in progress IAW GP-005, POWER OPERATION

NIS Power Indication is 50%

ERFIS Net MWe is 342 MWe Which ONE of the following indications would require the crew to stop the startup and stabilize reactor power and perform OST-010, POWER RANGE CALORIMETRIC DURING POWER OPERATION DAILY?

A. Tavg is 562.8°F B. Loop T is 33°F C. MWe Net Recorder is 336 MWe D. Calorimetric reactor power is 51.5%

Thursday, December 03, 2015 Page 182 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 68 68 B General Discussion Answer A Discussion Incorrect. This is incorrect because Tavg is not a redundant Reactor Power indication per Attachment 1 of GP-005. This is plausible because Tavg is programmed via control rod movement to vary linearly with reactor power between 0% and 100%. This value of Tavg is indicative of 54% power when on program. However, it could be a Tavg indicative of 50% power that is 1.3°F higher than program.

Answer B Discussion Correct. According to Precaution & Limitation 12 of GP-005 (p7 of 156; Rev 128), all indications of reactor power level listed on Attachment 1, Reactor Power Redundant Indications should be monitored and compared during the power level changes. Indications such as core T, and turbine first-stage pressure should be compared to NI indications at least once per 10% during the power level changes. If all indications do not agree within 5% of each other during power level changes, then Reactor Power should be stabilized, OST-010, Power Range Calorimetric During Power Operation (ERFIS) performed, and plant management contacted for further instructions. According to Attachment 1 of GP-005 (p127 of 156; Rev 128) Average Reactor Power, Loop T, 1st Stage Pressure and Net Mwe Max are redundant indications. At 50% power, Core T is 28.5°F and varies linearly to 100% power. Consequently, with an NIS Power Range indication of 50%, it would be expected that Loop T is approximately 28.5°F. Since actual Loop T is 33°F, and 15% higher than expected, the crew must stabilize plant power and perform OST-10.

Answer C Discussion Incorrect. This is incorrect because this is only 6 MWe different from the ERFIS Net MWe indication, and Attachment 1 indicates that the ERFIS and Net recorder normally indicate a 6 Mwe difference. This is plausible because the operator may not know about the Note and incorrectly believe that the difference is too large.

Answer D Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the OST is required to be performed when there is a 5% difference in redundant indications during plant manuevering in OP-105 or GP-005. This is plausible because the operator may incorrectly believe that the OST is required for a difference criteria of only 1.5%, which would be true under steady state conditions.

Basis for meeting the KA The KA is matched because the operator must demonstrate the Ability to identify and interpret diverse indications to validate the response of NIS Power Range instruments during a reactor startup. This is accomplished by requiring that the operator choose a parameter indicating at a specific value, which would require stabilizing reactor power and performing a calibration of the NIS PR instruments.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and use the information to predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided GP-005 (p7, 127 of 156; Rev 128) 10 CFR 55.41.10 GEN2.1 2.1.45 - GENERIC - Conduct of Operations Conduct of Operations Ability to identify and interpret diverse indications to validate the response of another indication. (CFR: 41.7 / 43.5 / 45.4)

Remarks/Status 7/14/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, why did we reference OP-105 vice GP-005?

11/5/15 GP-005 is referenced.

Thursday, December 03, 2015 Page 183 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination 12/2/15 AOM-SHIFT approved question for 60 day submittal.

QUESTION 68 68 B Thursday, December 03, 2015 Page 184 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.2 2.2.38 - GENERIC - Equipment Control QUESTION 69 69 A Equipment Control Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)

Given the following:

The plant is at 100% power

The FWUFM System has been OOS for five days Subsequently:

An AO reports a steam leak near the Main Steam Isolation Valves

The following indications are noted in the Control Room:

Tavg is LOWERING

Steam flow and feed flow have RISEN

A Power Limit Warning alarm on ERFIS has been received

Reactor power is slowly RISING Which ONE of the following identifies the MAXIMUM thermal power limit allowed by the operating license under these conditions?

A. 2300 MWt B. 2307 MWt C. 2339 MWt D. 2346 MWt Thursday, December 03, 2015 Page 185 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 69 69 A General Discussion Answer A Discussion Correct. According to OMM-001-2 (p16; Rev 92) Step 5.2.3, Diverse indications of Reactor power should be monitored at all times to ensure the Reactor is operated at less than or equal to 2339 MW thermal when the FWUFM System is in service and 2300 MW thermal when FWUFM is not in service as required by TRM 3.25 limitation. Also according to OMM-001-2 (p18/19; Rev 92) Step 5.2.7/11, when a Power Limit Warning alarm has been received, or plant indications show an actual transient condition, action shall be taken to maintain reactor power less than or equal to 2339 MWt using the Valve Position Limiter. However, because the FWUFM System is OOS, this limit must be adjusted to 2300.

Answer B Discussion Incorrect. This is incorrect because with an alarm present and a transient condition the steady state allowance of 100.3% maximum power level is not permitted. This is plausible because this power level would be consistent with 100.3 % of the FWUFM OOS power limit.

Answer C Discussion Incorrect. This is incorrect because the FWUFM System is OOS which limits maximum power to 2300 MWt. This is plausible because it is 100% RTP identified in the Operating License.

Answer D Discussion Incorrect. This is incorrect because FWUFM System is OOS which limits maximum power to 2300 MWt. This is plausible because it is 100%

RTP identified in the Operating License, and under steady state conditions OMM-001-2 permits power at any given time to be 100.3%, or 2346 MWt.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of both the License Power Level (2339 MWt) Limit and the Condition that the plant must operate within the confines of the Technical Specifications (In this case the TRM). In the case presented, the operator must select the maximum power permitted under the stated plant conditions (In the TRM and OMM).

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information (max license power level and conditions permitted) and then apply this information to stated plant conditions in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED 035 K5.01 001 Development References Student References Provided OMM-001-2 (p16, 18/19; Rev 92) Step 5.2.3, 7 and 11 10 CFR 55.41.10 GEN2.2 2.2.38 - GENERIC - Equipment Control Equipment Control Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)

Remarks/Status 7/14/15 modified bank question 035 K5.01 001 9/8/15 Reviewed by Exam Supervisor, with editorial comments, 1) stem has information not needed to answer question; 2) need to verify initial conditions are accurate {can we be <2300Mwt and still be @ 100% RTP? (have to be less than 2300 Mwt when FWUFM OOS for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )}

9/8/15 Operations SRO reviewed question, stated in response to Exam Supervisor comment that YES, they adjust NI to 100% at 2300 NWt per the TRM.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 186 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 69 69 A Thursday, December 03, 2015 Page 187 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.2 2.2.39 - GENERIC - Equipment Control QUESTION 70 70 A Equipment Control Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7 / 41.10 / 43.2 / 45.13)

Given the following:

The plant is at 100% power

It has been determined that LCO 3.2.3, Axial Flux Difference is outside the target band.

Which ONE of the following correctly completes the statement below?

IAW Technical Specification LCO 3.2.3 ACTION.

(1) Restore AFD to within target band within ___(1)___.

(2) If AFD cannot be restored within target band within the allowable time, Reduce THERMAL POWER to < 90% RTP or 0.9 APL, whichever is less, within an additional

___(2)___.

A. (1) 15 minutes (2) 15 minutes B. (1) 15 minutes (2) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. (1) 30 minutes (2) 15 minutes D. (1) 30 Minutes (2) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Thursday, December 03, 2015 Page 188 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 70 70 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to Tech Spec LCO 3.2.3 ACTION Condition A (p3.2-8; Amendment 176), the operator must restore AFD to within target band within 15 minutes. According to ACTION Condition B, if the Required Action and associated Completion Time of Condition A not met, the operator must If AFD cannot be restored Reduce THERMAL POWER to < 90% RTP or 0.9 APL, whichever is less, within 15 minutes.

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is too long. This is plausible because the RO candidate recognizes that they are responsible for all TS ACTION that is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or less, and there are many ACTIONs required by Tech Specs within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because 30 minutes is too long. This is plausible because the RO candidate recognizes that they are responsible for all TS ACTION that is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or less, and Condition C of LCO 3.2.3 has a 30 minute ACTION.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of less than or equal to one hour Technical Specification action statements such as LCO 3.2.3.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must simply recall bits of information in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Tech Spec LCO 3.2.3 ACTION Condition A&B (p3.2-8; Amendment 176) 10 CFR 55.41.10 GEN2.2 2.2.39 - GENERIC - Equipment Control Equipment Control Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7 / 41.10 / 43.2 / 45.13)

Remarks/Status 7/20/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, SAT 9/8/15 Operations SRO reviewed question, RO level knowledge, below the line ITS info?

9/10/15 exam team response is that < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months tech spec action items are RO required knowledge.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 189 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.3 2.3.13 - GENERIC - Radiation Control QUESTION 71 71 A 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:

The plant is in Refueling Mode and Core Alterations are in progress

The Refueling SRO reports that a fuel assembly was accidently dropped from the manipulator crane and has landed on top of the core area Subsequently:

AOP-013, FUEL HANDLING ACCIDENT, has been entered

CV Particulate Radiation Monitor R-11 reading is stable

CV Noble Gas Radiation Monitor R-12 reading is increasing Which ONE of the following correctly completes the statements below?

IAW AOP-013 An evacuation of from the Containment Vessel ____(1)____ required.

The initiation of OMM-033, IMPLEMENTATION OF CV CLOSURE, ____(2)____ required.

A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT Thursday, December 03, 2015 Page 190 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 71 71 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to AOP-013 (p3 of 13; Rev 32) step 1, personnel evacuation from containment is required. Step 4 of AOP-013 has the operator determine if implementation of CV closure is required based on rising indication or alarm on either R-11 or R-12, which is given information.

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because initiation of OMM-33 is required. This is plausible is the operator incorrectly concludes that both R-11 and R-12 are required to be rising or in alarm. However, step 4 directs to the operator to check R-11 AND R-12 and if EITHER is rising or in alarm, CV closure IAW OMM-033 is required.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because an evacuation is required. This is plausible because the operator may incorrectly conclude that fuel handling personnel are to remain in the area to store fuel in a safe location. This is based on wording in the AOP-013 basis document for step 1 which states "Consideration was given to provide instructions for any refueling activities to store fuel in some safe location. However, it is considered that too many unknowns are possible which may place personnel in danger of exposure. Systems and structures as designed provide for safe locations of fuel. Also, most likely, the fuel being handled was the one that is involved in the accident.

Therefore, the first priority is to protect personnel in the affected area".

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements (in this case an evacuation or prohibition of entry), and fuel handling responsibilities (what actions are required in the event of a fuel handling accident). This is accomplished by requiring that the operator identify whether or not establishing Containment Closure is required in response to rising indications or alarms on the involved radiation monitors.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall bits of information in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK AOP-013-03 10 Development References Student References Provided AOP-005 (p7&9 of 61; Rev 32)

AOP-005-BD (p5&7 of 29; Rev 32) 10 CFR 55.41.10 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)

Remarks/Status 7/15/15 modified bank question 076 G2.4.26 001 9/8/15 Reviewed by Exam Supervisor, UNSAT, as RO's are not required members of the fire brigade AND SRO's will not qualify for fire brigade until after license class is complete.

9/14/15 exam team drafted new question that would meet the KA.

12/1/15 AOM-Shift recommended replacement bank question.

Thursday, December 03, 2015 Page 191 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination 12/2/15 AOM-SHIFT approved question for 60 day submittal.

QUESTION 71 71 A Thursday, December 03, 2015 Page 192 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.3 2.3.4 - GENERIC - Radiation Control QUESTION 72 72 B Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10)

Given the following:

An operator has been assigned work in the radiologically controlled area

The dose rate in the area is 500 mR/hr

The operator has a current yearly dose of 0.5 Rem TEDE Which ONE of the following identifies the MAXIMUM time that the operator can perform work BEFORE reaching the Duke Energy Annual Administrative Dose Limit without receiving an extension?

A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months C. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months Thursday, December 03, 2015 Page 193 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 72 72 B General Discussion Answer A Discussion Incorrect. This is incorrect because 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months of work are permitted. This is plausible because this would be correct if the operator incorrectly believed that the Admin Limit was 1 Rem Annually.

Answer B Discussion Correct. According to PD-RP-ALL-0001 (p17 of 57; Rev 3) the Table associated with Step 5.2.2.1 states that the Duke Energy Annual Administrative Dose Limit is 2 Rem Annually. There is an allowable exposure of 1500 mrem before the Admin Limit is reached. This would permit the operator to work for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months before the limit is reached.

Answer C Discussion Incorrect. This is incorrect because 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months of work are permitted. This is plausible because this would be correct if the operator incorrectly believed that the Admin Limit was 2.5 Rem Annually.

Answer D Discussion Incorrect. This is incorrect because 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months of work are permitted. This is plausible because this would be correct if the operator incorrectly believed that the Admin Limit was 4 Rem Annually.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of radiation exposure limits under normal conditions, specifically with respect to local administrative TEDE limits.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall a bit of information (Annual Dose Limit) and use this information to predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK G2.3.4 1 Development References Student References Provided PD-RP-ALL-0001 (p17 of 57; Rev 3) 10 CFR 55.41.12 GEN2.3 2.3.4 - GENERIC - Radiation Control Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10)

Remarks/Status 7/20/15 used bank question G2.3.4 1 9/8/15 Reviewed by Exam Supervisor, SAT 11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 194 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.4 2.4.11 - GENERIC - Emergency Procedures / Plan QUESTION 73 73 A Emergency Procedures / Plan Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13)

Given the following:

A plant cooldown is in progress

RCS Temperature is 430°F

RCS pressure is 725 psig

The SI Accumulators have been isolated Subsequently, an RCS leak occurs and the operating crew enters AOP-016, EXCESSIVE RCS LEAKAGE.

Which ONE of the following correctly completes the statements below?

(1) AOP-016 will ensure that a MINIMUM of ____(1)____ Charging Pumps are running with their speed adjusted to maximum.

(2) If Pressurizer level is lowering uncontrollably with maximum charging flow, under the current plant conditions the crew would transition to ____(2)____.

A. (1) Two (2) AOP-033, SHUTDOWN LOCA B. (1) Two (2) EOP-E-0, REACTOR TRIP OR SAFETY INJECTION C. (1) Three (2) AOP-033, SHUTDOWN LOCA D. (1) Three (2) EOP-E-0, REACTOR TRIP OR SAFETY INJECTION Thursday, December 03, 2015 Page 195 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 73 73 A General Discussion Answer A Discussion Correct. 1st part correct, 2nd part correct. According to AOP-016 (p4-6; Rev 23) Steps 6-11, the operator will attempt to start two Charging Pumps, if available, and establish maximum charging flow. Also according to AOP-016 (p6-7; Rev 23) Steps 15-16 and 18-19, the operator will check Pzr level lowering with maximum charging flow, and if so, check RCS pressure and temperature to select a recover procedure. Since RCS pressure is less than 1000 psig, the status of the SI Accumulators must be checked. Since the SI Accumulators are isolated, the crew will transition to AOP-033.

Answer B Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because the crew will transition to AOP-033. This is plausible because if the SI Accumulators were unisolated, or pressure were greater than 1000 psig, the transition would be made to EOP-E-0.

Answer C Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because only two charging pumps will be started in the initial steps (prior to transition to AOP-033) of AOP-016. This is plausible because if the crew directly entered AOP-033 and bypassed entry into AOP-016, the crew would be directed to start all three Charging Pumps if available.

Answer D Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of abnormal condition procedures such as the number of Charging Pumps started in AOP-016 and the entry conditions for AOP-033.

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and use this information to select the correct mitigation procedure, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided AOP-016 (p4-7; Rev 23) 10 CFR 55.41.10 GEN2.4 2.4.11 - GENERIC - Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13)

Remarks/Status 7/20/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, with editorial comment, in stem change "shutdown" to "cooldown" 9/10/15 OPS SRO reviewed the question and recommended changing the stem to have RCS Temp at 340 degrees.

9/11/15 exam team made changes as proposed.

11/29/15 AOM-Shift recommended changes in conditions to be more realistic. Temp 430F and pressure 725 psig 11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 196 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 73 73 A Thursday, December 03, 2015 Page 197 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.4 2.4.25 - GENERIC - Emergency Procedures / Plan QUESTION 74 74 D Emergency Procedures / Plan Knowledge of fire protection procedures. (CFR: 41.10 / 43.5 / 45.13)

Given the following:

The plant is at 100% power

The Control Room receives a confirmed report of a fire in the Unit 2 Cable Spread Room (Fire Zone 19) at 0730 Which ONE of the following correctly completes the statements below?

(1) IAW AOP-041, RESPONSE TO FIRE EVENT, the operator must ____(1)____.

(2) The LATEST time that this action may be complete is ____(2)____.

A. (1) close both Pressurizer PORV Block Valves (RC-535 and RC-536), ONLY (2) 0740 B. (1) close both Pressurizer PORV Block Valves (RC-535 and RC-536), ONLY (2) 0745 C. (1) close both Pressurizer PORV Block Valves (RC-535 and RC-536) AND place the PCV-455C and 456 Isolate Switches to the ISOLATE position (2) 0745 D. (3) close both Pressurizer PORV Block Valves (RC-535 and RC-536) AND place the PCV-455C and 456 Isolate Switches to the ISOLATE position (4) 0740 Thursday, December 03, 2015 Page 198 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 74 74 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because the Pzr PORV Block valves must be closed as well. This is plausible because according to the AOP-041 Background Document the PZR PORV block valves are closed based on defense in depth; and the operator may incorrectly believe that it is only needed to isolate the PORV circuits.

Answer B Discussion Incorrect. 1st part wrong, 2nd part wrong. See A and C.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because 0740 is the LATEST time that this action may be complete and still meet the requirements of the NOTE prior to Step 3. This is plausible because the operator may incorrectly confuse Time Critical Tasks completion times with 15 minutes, rather than 10 minutes. (restoring RCP seal cooling or isolate seals before restoration of charging if 15 min elapsed)

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to AOP-041 (p4 of 28; Rev 7) Step 3&4, if there is a confirmed fire in Fire Zone 19 the operator is directed to Place the PCV-455C and 456 Isolate Switches to the ISOLATE position AND close both Pressurizer PORV Block valves (RC-535 and RC-536) within 10 minutes.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of the actions contained within the fire protection procedures (AOP-041).

Basis for Hi Cog The question is at the Comprehension/Analysis cognitive level because the operator must recall bits of information and use this information to predict an outcome, in order to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided AOP-041 (p4 of 28; Rev 7) 10 CFR 55.41.10 GEN2.4 2.4.25 - GENERIC - Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of fire protection procedures. (CFR: 41.10 / 43.5 / 45.13)

Remarks/Status 7/16/15 new question developed.

9/8/15 Reviewed by Exam Supervisor, UNSAT, based on K/A not met, question is asked on an AOP, not a Fire Protection Procedure.

9/11/15 exam team discussed with the exam supervisor that the question is asked on a procedure that mitigates a fire, and believes the KA is matched. Question unchanged at this time.

11/5/15 discussed with NRC chief examiner as one of the early submission items. He recommended a number of things that should apply throughout the exam. 1st, whenever possible, state IAW a procedure. 2nd, if something is the same in all four of the answer choices, move it to the stem and reword the question to a "required/not required" format for the other answer part.

11/9/15 added IAW a procedure as per the chief examiners comments. The answer and distractor choices need to remain as is, since the two part is an action, and a time required to complete it. If we take the first part, that is common to all answer choices out, then it cues the candidate that the action must be done, in order for there to be a time to complete it.

11/29/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 199 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 74 74 D Thursday, December 03, 2015 Page 200 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination GEN2.4 2.4.31 - GENERIC - Emergency Procedures / Plan QUESTION 75 75 D Emergency Procedures / Plan Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3)

Given the following:

The plant has tripped from 100% power

The crew has entered EOP-E-0, REACTOR TRIP OR SAFETY INJECTION

The BOP is verifying automatic safety injection actuation IAW Attachment 1, AUTO ACTION VERIFICATION, which ONE of the following identifies two RTGB Annunciators whose status is required to be checked?

A. APP-001-C1, RCP THERM BAR COOL WTR HI FLOW; and APP-002-F7, INSTR AIR HDR LO PRESS B. APP-001-C1, RCP THERM BAR COOL WTR HI FLOW; and APP-008-F7, SOUTH SW HDR LO PRESS C. APP-036-D2, BATT CHARGER B/B-1 TROUBLE; and APP-002-F7, INSTR AIR HDR LO PRESS D. APP-036-D2, BATT CHARGER B/B-1 TROUBLE; and APP-008-F7, SOUTH SW HDR LO PRESS Thursday, December 03, 2015 Page 201 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination QUESTION 75 75 D General Discussion Answer A Discussion Incorrect. 1st part wrong, 2nd part wrong. See B and C.

Answer B Discussion Incorrect. 1st part wrong, 2nd part correct. This is incorrect because APP001C1 is not checked as part of Attachment 1. This is plausible because it is checked in the body of EOP-E-0, at Step 10, as one of two means to check RCP Seal Cooling.

Answer C Discussion Incorrect. 1st part correct, 2nd part wrong. This is incorrect because APP002F7 is not checked as part of Attachment 1. This is plausible because it is checked in the body of EOP-E-0, at Step 25, when restoring Instrument Air to containment.

Answer D Discussion Correct. 1st part correct, 2nd part correct. According to Attachment 1 of EOP-E-0 (p33; Rev 6) Step 15, the operator is directed to check APP036D2, BATT CHARGER B/B1 TROUBLE extinguished as a means to check the Battery Chargers are energized. According to of EOP-E-0 (p30; Rev 6) Step 7.c, the operator is directed to check APP008F7, SOUTH SW HDR LO PRESS extinguished as a means to check the proper operation of the service water system.

Basis for meeting the KA The KA is matched because the operator must demonstrate Knowledge of annunciator alarms use in the EOPs, specifically the annunciators used in EOP-E-0.

Basis for Hi Cog The question is at the Memory cognitive level because the operator must recall bits of information to answer the question correctly.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided EOP-E-0 (p30 and 33; Rev 6) 10 CFR 55.41.10 GEN2.4 2.4.31 - GENERIC - Emergency Procedures / Plan Emergency Procedures / Plan Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3)

Remarks/Status 7/16/15 modified bank question 022 AA2.03 9/8/15 Reviewed by Exam Supervisor, UNSAT, question does not need to be this difficult to satisfy a generic K/A.

9/14/15 new question developed to better meet the KA. Exam team, Supervisor and OPS SRO concur on changes.

11/17/15 Fleet review determined that we did not write a question that met the KA, as it did not apply to the Emergency plan or procedures. The exam team wrote a new question to meet the KA.

12/2/15 AOM-SHIFT approved question for 60 day submittal.

Thursday, December 03, 2015 Page 202 of 202

FOR REVIEW ONLY - DO NOT DISTRIBUTE ILC15 RNP RO NRC Examination EPE007 EK2.02 - Reactor Trip QUESTION 1 1 D Knowledge of the interrelations between a reactor trip and the following: (CFR 41.7 / 45.7)

Breakers, relays and disconnects ...................................

Given the following:

The plant is at 100% power

Maintenance and Testing of A Reactor Trip Breaker is in progress

The A Reactor Trip Bypass Breaker has been placed in service

The A Reactor Trip Breaker has been removed from service

A transient condition exists requiring a reactor trip

The OATC depresses the manual Reactor trip pushbutton Which ONE of the following describes the plant response?