ML15119A469

Initial Exam 2015-301 Draft RO Written Exam

ML15119A469
Person / Time
Site:	Turkey Point
Issue date:	04/29/2015
From:	Division of Reactor Safety II
To:	Florida Power & Light Co
References
Download: ML15119A469 (328)

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PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 003 K4.07 Importance Rating 3.2 Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following:

Minimizing RCS leakage (mechanical seals)

Proposed Question: RO Question # 1 Which ONE of the following completes the statement below?

RCP #1 Seal leakoff is reduced by raising A. Reactor Coolant Drain Tank pressure B. Volume Control Tank pressure C. Reactor Coolant Pump Standpipe level D. RCP Seal Injection Flow Proposed Answer: B Explanation (Optional):

A. Incorrect: Plausible to remember that RCP #2 Seal Leakoff flowpath is directed to the RCDT and that raising RCDT pressure would have an effect on #1 seal leakoff. #1 seal leakoff, however, is not affected by RCDT pressure. It does affect #2 seal leakoff.

B. Correct: Seal leakoff flow is directed to the VCT at a design flowrate of 3 gpm.

C. Incorrect: Plausible to remember seal leakoff flow is directed to the standpipe at a design flowrate of 100 cc/hr, however, this is flowrate for #3 RCP seal.

D. Incorrect: Plausible to remember that if injection is high, more flow is getting to the seals. However, #1 Seal leakoff is based on the VCT pressure, not on seal injection flow.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Technical Reference(s): 561 3-M-3041, Sheet 3, Rev. 31 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: [P6902008 Obj.#3 4a 4b (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 DC Cook Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

COG = iF, DIF B PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Examination Outline Crossreference: Level RO SRO Tier# 2 Group# 1 KJA# 003 K4.07 Importance Rating 3.2 Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following:

Minimizing RCS leakage (mechanical seals)

Proposed Question: RO Question # 1 Which ONE of the following completes the following statement?

RCP #2 seal minimizes RCS leakage during normal operations by directing leakoff of A.

B.

3 gph to the Reactor Coolant Drain Tank 3 gpm to the Volume Control Tank.

z C. 100 cc/hr to the Reactor Coolant Pump Standpipe.

D. 100 cc/hr directly to the Reactor Coolant Drain Tank.

Proposed Answer: A Explanation (Optional):

A. Correct: RCP #2 Seal Leakoff flowpath is directed to the #2 Seal Standpipe, which flows through an orifice to the RCDT.

B. Incorrect: Plausible to remember seal leakoff flow is directed to the VCT at a design flowrate of 3 gpm, however, this is flowpath and flowrate for #1 RCP seal.

C. Incorrect: Plausible to remember seal leakoff flow is directed to the standpipe at a design flowrate of 100 cc/hr, however, this is flowrate for #3 RCP seal.

D. Incorrect: Plausible to remember seal leakoff flow is directed to the RCDT at a design flowrate of 100 cc/hr, however, this is flowrate for #3 RCP seal and flowpath for the #2 RCP seal leakoff via the standpipe.

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 5613-M-3041 Sheet3 Rev. 31 Technical Reference(s): , , (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO 6902008, Obj. #3, 4a, 4b Learning Objective: (As available)

Question Source: Bank # x Modified Bank # (Note changes or attach parent)

New Question History:

Question Cognitive Level:

Last NRC Exam:

Memory or Fundamental Knowledge Comprehension or Analysis 2008 x

DC Cook z

10 CFR Part 55 Content: 55.41 55.43 Comments:

COG = iF, DIF = B FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# I KIA# 004 A1.09 Importance Rating 3.6 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: RCS pressure and temperature Proposed Question: RO Question # 2 Given the following conditions:

• Unit4isinMODE5.

• The Pressurizer is solid with OMS in low pressure operation.

• 4B RHR loop is in operation.

• RCS pressure control is in manual.

Subsequently:

• Operator manually lowers demand on TC-4-144A, Letdown Temperature Controller.

Which ONE of the following identifies the plant response with no additional operator action?

A. NRHX CCW flow decreases.

B. Shutdown Margin increases.

C. RCS pressure decreases.

D. RCS temperature increases.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible to believe that lowering TC-4-144A, Letdown Temperature PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Controller demand would cause the valve to close. However, lowering demand on the controller opens the valve to increase CCW flow and lower the temperature.

B. Incorrect. Plausible to believe that a temperature change would affect Shutdown Margin. However, a decrease in RCS temperature would add positive reactivity and decrease shutdown margin.

C. Correct. With PCV-4-145 in manual control, changes to RCS pressure would not be adjusted for by the automatic setpoint of PCV-4-145. An increase in heat removal from letdown would reduce RCS temperature and since the plant is solid, RCS pressure.

D. Incorrect. Plausible to believe that lowering TC-4-144A, Letdown Temperature Controller demand would cause the valve to close, which would cause RCS temperature to increase. However, lowering demand on the controller opens the valve to increase CCW flow and lower the temperature.

Technical Reference(s): 4-OP-050 Rev 13 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902121A Obj. 9, 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

IN IT! A L S CK9 VERIF 5.2.2 (Contd)

CAUTIONS

• Letdown temperature should be monitored closely when adjusting Low 1

Pressure Letdown Controller, PCV-4-145. Temp Cntl Vlv for CCW From Non Regen HX Outlet, TCV-4-144, may not respond fast enough to changes in letdown temperature.

• Letdown pressure should monitored closely when ang Low Pressure Letdown Controller, PCV-4-145. Pressure shall not exceed 600 psig as indicated on P1-4-145.

• Letdown flow shall be limited to 120 gpm to prevent channeling of demineralizer resin.

17. Adjust Low Pressure Letdown Controller, PCV 145 as necessary to obtain adequate letdown flow to warm RHR System piping.

18. Using 4-OP-047, CVCS Charging and Letdown, start additional Charging Pumps as required to match the increased letdown flow.

19. Ensure RHR loop boron concentration is as required for a refueling outage (RFO) or for a non-refueling outage (non-RFO) as follows:

a. IF the RHR loops have been sampled in the previous 14 days AND no intervening evolutions to reduce RHR loop boron concentration have occurred, THEN obtain last RHR loop boron concentration from Chemistry.

b. the RHR loops have NOT been sampled in the previous 14 days OR there have been intervening evolutions to reduce RHR loop boron concentration, THEN have Chemistry obtain new sample AND report the results to the Unit 4 Reactor Operator.

c. Record applicable RHR loop boron concentration: ppm

d. Check the Rl-lR loop boron concentration is equal to or greater than the current RCS boron concentration.

(1) IF RHR loop boron concentration is less than the minimum required cold shutdown boron concentration, THEN continue letting down to CVCS until RHR loop boron concentration is greater than the minimum required cold shutdown RCS boron concentration before continuing this procedure.

(2) IF Unit 4 is in an action statement which prohibits positive reactivity changes, THEN continue letting down to CVCS until RHR loop boron concentration is within 20 ppm of the RCS before continuing this procedure W97:IJCIIcIsIcI slab

• ..* *\

I

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KJA# 004 K5.43 Importance Rating 3.6 Knowledge of the operational implications of the following concepts as they apply to the CVCS:

Saturation, subcooling, superheat in steam/water Proposed Question: RO Question # 3 Given the following conditions:

• Unit 3 is at 100% power.

• Letdown flow is at 105 gpm Which ONE of the following completes the statement below?

To remove a Letdown Orifice from service, PCV-3-145, Low Pressure Letdown Controller, is placed in Manual and adjusted (1) to maintain the downstream fluid in a (2) state.

A. (1) open (2) subcooled B. (1) closed (2) subcooled C. (1) open (2) saturated D. (1) closed (2) saturated Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part wrong. 2 nd part right. Plausible to believe that opening PCV-3-145 would reduce pressure, thereby reducing temperature and affecting the heat transfer rate to maintain the letdown flow subcooled. Removing one orifice from service reduces PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION letdown flow from 105 to 60 gpm. However, PCV-3-145 is adjusted closed to maintain pressure and subcooling.

B. Correct. Removing one orifice from service reduces letdown flow from 105 to 60 gpm.

At 105 gpm, two charging pumps are running. Since the charging pump is positive displacement and flow is controlled by speed of the pump, pump speed must be reduced to match charging to letdown and maintain PRZ level. This prevents erosion damage to the letdown orifices. PCV-3-145 is closed to increase system pressure and maintain subcooling.

C. Incorrect. jst part wrong. 2 part wrong. Plausible to believe that opening PCV-3-145 would reduce pressure, thereby reducing temperature and affecting the heat transfer rate and to confuse saturated with subcooled.. Removing one orifice from service reduces letdown flow from 105 to 60 gpm. The rapid reduction in downstream pressure on closure of the orifice isolation valve would cause letdown line flow flashing and this effect would lead to a belief that the flow is initially in a saturated state. However, PCV 3-145 is adjusted closed to maintain pressure and subcooling.

D. Incorrect. 1st part right. 2 part wrong. Plausible to believe that opening PCV-3-145 would reduce pressure, thereby reducing temperature and affecting the heat transfer rate and to confuse saturated with subcooled.. Removing one orifice from service reduces letdown flow from 105 to 60 gpm. The rapid reduction in downstream pressure on closure of the orifice isolation valve would cause letdown line flow flashing and this effect would lead to a belief that the flow is initially in a saturated state. However, PCV 3-145 is adjusted closed to maintain pressure and subcooling.

Technical Reference(s): 3-OP-047 Rev 8 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902113 Obj. 5,8,9 (As available)

Question Source: Bank #

Modified Bank # 91123 (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 Calvert Cliffs Question Cognitive Level: Memory or Fundamental Knowledge PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A # 004 K5.43 Importance Rating 3.6 Knowledge of the operational implications of the following concepts as they apply to the CVCS:

Saturation, subcooling, superheat in steam/water Proposed Question: RO Question # 3 -J The Low Pressure Letdown Control Valve PCV-3-145 prevents saturated conditions between the (1) and Letdown Orifice Stop Valves to prevent erosion damage. The Low Pressure Letdown Control Valve (2) to increase pressure.

A.

B.

(1) Non-regenerative Heat Exchanger (2) opens (1) Non-regenerative Heat Exchanger (2) closes z

C. (1) Regenerative heat exchanger (2) opens D. (1) Regenerative heat exchanger (2) closes Proposed Answer: B Explanation (Optional):

A. Incorrect. Valve opens to decrease pressure and closes to increase pressure.

Plausible to think that opening a valve will increase downstream pressure and closing a valve will decrease downstream pressure, but here the pressure control is for the upstream section.

B. Correct. Per SD-013 ( CVCS), page 20, the purpose of the Low Pressure Letdown Control Valve PCV-3-145 is to prevent flashing of hot liquid to steam between the Letdown Orifice Stop Valves and the Non-regenerative Heat Exchanger and to Control letdown system pressure at approximately 275 psig. This prevents erosion damage to the letdown orifices. PCV-3-145 opens to decrease system pressure and closes to increase system pressure.

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 C. Incorrect. Prevents flashing of hot liquid to steam between the Letdown Orifice Stop Valves and the Non-regenerative Heat Exchanger. Plausible to believe that it would be between the Letdown Orifice Stop Valves and the Regenerative Heat Exchanger since that is the first major component in line from the RCS. The interlocks between LCV 460 and PCV-3-200A,B,C prevent flashing damage to the Regenerative Heat Exchanger. Valve opens to decrease pressure and closes to increase pressure.

Plausible to think that opening a valve will increase downstream pressure and closing a valve will decrease downstream pressure, but here the pressure control is for the upstream section.

D. Incorrect. Prevents flashing of hot liquid to steam between the Letdown Orifice Stop Valves and the Non-regenerative Heat Exchanger. Plausible to believe that it would be between the Letdown Orifice Stop Valves and the Regenerative Heat Exchanger since

-J that is the first major component in line from the RCS. The interlocks between LCV 460 and PCV-3-200A,B,C prevent flashing damage to the Regenerative Heat Exchanger.

Technical Reference(s):

SD-013. Chemical and Volume Control System (1 1/15/2012)

Page 20.

(Attach if not previously provided) z Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank# 91123 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 Calvert Cliffs Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 FOR OFFICIAL USE ONLY- LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Comments:

-J z

(9 0

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

INIT 7.11.2.3 (Contd)

f. Verify OR place PCV-3-145, Low Pressure Letdown Controller, Manual-Auto Station in Manual.

g. While monitoring letdown pressure on P1-3-145. adjust Low Pressure Letdown Controller, PCV-3-145, to obtain a letdown pressure of 90 psig to 100 psig.

NOTES I

• Substeps 7. 11.2.3.h and 7.11.2.3.1 should be performed concurrently using two

• operators; one to perform the valve maneuvers, and the other to perform any I PCV-3-145 adjustments necessary to prevent going beyond the pressures specified.

p I

• To limit the letdown pressure spike, prompt operation of PCV-3-145 is required when an I increase in letdown flow or pressure is observed due to the letdown orifice stop valve opening.

I

h. WHEN letdown pressure is stable in Manual, THEN Open the desired closed orifice stop valve.

i. Adjust Low Pressure Letdown Controller, PCV-3-145, in Manual to limit the pressure spike to 400 psig as indicated on P1-3-145.

j. WHEN temperature on TI-3-144 is stable at 115-121°F, THEN place TC-3-144A in Auto. (N/A if swapping orifice stop valves)

k. Adjust PC-3-459G, Charging Pump Master Controller in Manual to maintain PZR Level and Letdown temperature.

N OTE With two charging pumps running and in Automatic control a prolonged pressunzer high level may cause the Pressurizer Master Controller to reduce speed on the running charging pumps below the pump oil trip speed. I

4. IF it is desired to decrease letdown flow OR to remove a letdown orifice 4rom service, THEN perform the following:

a. Verify OR place TC-3-144A in Manual.

b. IF two charging pumps are running, THEN take manual control of running charging pumps while maintaining Pressurizer Level and Letdown temperature.

c. nsu 7 PCV-3-145, Low Pressure Letdown Controller, Manual-Auto Station is in Manual.

d While monitoring letdown pressure on P1-3-145, adjust Low Pressure Letdown Controller, PCV-3-145, to obtain a letdown pressure of 300 psig.

MQ7AAi ICI,kIn.,,I,I,

INIT 7.11 .2.4 (Contd)

NOTES Substeps 7. 11.2.4.e, 7. 11.2.4.I and 7. 11.2.4.g should be performed concurrently using two operators, one to perform the valve maneuvers, and the other to perform any V.3-145 adjustments necessaiy to prevent going beyond the pressures specified.

PC

• The operator adjusting PCV-3-145hould take a bump closed as soon as a decrease in flow rate or pressure is detected to prevent fiashing.

e. WHEN letdown pressure is stable in Manual, THEN Close the desired open orifice stop valve.

f. Place the control switch for the Orifice Stop Valve that was Closed in Substep 7.11 .2.4.e to Auto.

g. Adjust Low Pressure Letdown Controller, PCV-3-145, in Manual to maintain pressure above 150 psig as indicated on PI-3-l45.

h. Set TC-3-144A to 50 percent demand.

i. WHEN temperature is stable at 115-121°F, THEN place TC-3-144A in Auto.

5. IF desired flow rate has been obtained, THEN adjust Low Pressure Letdown Controller, PCV-3-145, in Manual to normal operating band.

6. IF desired flow rate has been obtained, THEN place Low Pressure Letdown Controller, PCV-3-145, in Auto.

7. IF two charging pumps are running AND only one pump is required, THEN perform the following:

a. Verify all running charging pumps are in Manual speed control.

b. Slowly decrease the charging pump speed on the pump being removed from service.

c. WHEN minimum demand on the Demand Meter (20 to 25 percent) for the pump to be shutdown has been established, THEN turn the pump control switch to Off.

CAUUON Automatic charging pump control should NOT be restored until pressurizer level is at program level to prevent charging pump trip on low speed.

8. WHEN Pressurizer level is at program level, THEN restore automatic charging pump control.

9. Verify Letdown Demineralizer Divert Valve, TCV-3-143, is in the Auto position AND the green Demins light is lit.

10. Enter completion of this procedure subsection in the Unit Narrative Log.

(fl7.fjfj

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 005 K4.11 Importance Rating 3.5 Knowledge of RHRS design feature(s) and/or interlock(s) which provide or the following:

Lineup for low head recirculation mode (external and internal)

Proposed Question: RO Question # 4 Given the following conditions:

• Unit 3 is in MODE 4.

• RCS pressure is 370 psig.

• Crew prepares to place RHR in service.

• MOV-3-750, Loop 3C RHR Pump Suction Stop Valve, is open.

• MOV-3-862B, RHR Suction from RWST, is open.

• MOV-3-751, Loop 3C RHR Pump Suction Stop Valve, will not open.

Which ONE of the following completes the statements below?

MOV-3-751 (1) prevented from opening by RCS pressure.

MOV-3-751 (2) prevented from opening by MOV-3-862B.

A. (1)is (2) is not B. (1)isnot (2) is not C. (1) is (2) is D. (1)isnot (2) is Proposed Answer: D PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Incorrect: 1st part wrong. 2 part wrong. Plausible to believe that RCS pressure is preventing MOV-3-751 from opening since RHR cannot be placed in service with RCS pressure greater that 450 psig and RCS temperature greater than 350°F and student confused those two parameters. RCS pressure will close MOV-3-750 and MOV-3-751 when the Loop A Hot leg pressure reaches 515-535 psig, so RCS pressure is not preventing MOV-3-751 from opening. Plausible to believe that since MOV-3-750 is open, the valve interlocks have been satisfied. However, the interlock states that MOV 3-751 cannot be opened unless MOV-3-862B and MOV-3-863B are closed.

B. Incorrect: 1st part right. 2 nd part wrong. RCS pressure will close MOV-3-750 and MOV 3-751 when the pressure reaches 515-535 psi , 50 RCS pressure is not preventing 9

MOV-3-751 from opening. Plausible to believe that since MOV-3-750 is open, the valve interlocks have been satisfied. However, the interlock states that MOV-3-751 cannot be opened unless MOV-3-862B and MOV-3-863B are closed.

C. Incorrect: 1st part wrong. 2 part right. Plausible to believe that RCS pressure is preventing MOV-3-751 from opening since RHR cannot be placed in service with RCS pressure greater that 450 psig and RCS temperature greater than 350°F and student confused those two parameters. RCS pressure will close MOV-3-750 and MOV-3-751 when the Loop A Hot leg pressure reaches 515-535 psig, so RCS pressure is not preventing MOV-3-751 from opening. The interlock states that MOV-3-751 cannot be opened unless MOV-3-862B and MOV-3-863B are closed.

D. Correct: RCS pressure will close MOV-3-750 and MOV-3-751 when the pressure reaches 515-535 psig, so RCS pressure is not preventing MOV-3-751 from opening.

The interlock states that MOV-3-751 cannot be opened unless MOV-3-862B and MOV 3-863B are closed.

Technical Reference(s): 5613-M-3050, Sh. 1, Rev. 36 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO LP6902121A(0612112012)Obj.3 Learning Objective: (As available) 8 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 006 K2.02 Importance Rating 2.5 Knowledge of bus power supplies to the following: Valve operators for accumulators Proposed Question: RO Question # 5 Given the following conditions:

• A large break LOCA has occurred on Unit 3.

• The 3B 4kV Bus is faulted and locked-out.

• The crew is performing actions of 3-EOP-E-1, Loss of Reactor or Secondary Coolant.

• Conditions are met to isolate the SI Accumulators.

Which ONE of the following completes the statement below and identifies the Accumulator Isolation Valves capable of being closed?

MOV-3-865A and (1) , Accumulator Discharge Isolation valves, can be closed from (2)

A. (1) MOV-3-865B (2) VPB B. (1) MOV-3-865C (2) VPB C. (1) MOV-3-865B (2) MCC using pushbutton D. (1) MOV-3-865C (2) MCC using pushbutton PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part wrong. 2nd part right. MOV-3-865A, B and C are powered from Vital MCCs A, B, and C respectively. Since 3B 4kV bus powers the 3B Vital LC which in turn powers the 3B Vital MCC, MOV-33-865B cannot be energized due to the lockout.

Plausible to believe that the MOVs may be powered from 3A and 3C MCCs. On Unit 4, MOV-4-865A, B and C are powered from Vital MCCs C, A, and B respectively.

B. Correct. MOV-3-865A, B and C are powered from Vital MCCs A, B, and C respectively.

Since 3B 4kV bus powers the 3B Vital LC which in turn powers the 3B Vital MCC, MOV 33-865B cannot be energized due to the lockout.

C. Incorrect. 1st part wrong. 2nd part wrong. MOV-3-865A, B and C are powered from Vital MCCs A, B, and C respectively. Since 3B 4kV bus powers the 3B Vital LC which in turn powers the 3B Vital MCC, MOV-33-865B cannot be energized due to the lockout.

Plausible to believe that the MOVs may be powered from 3A and 3C MCCs. On Unit 4, MOV-4-865A, B and C are powered from Vital MCCs C, A, and B respectively. Also plausible to believe the MOVs can be operated remotely from the breaker, however, the pushbutton is an overload reset only.

D. Incorrect. 1st part right. 2nd part wrong. MOV-3-865A, B and C are powered from Vital MCCs A, B, and C respectively. Since 3B 4kV bus powers the 3B Vital LC which in turn powers the 3B Vital MCC. Plausible to believe the MOVs can be operated remotely from the breaker, however, the pushbutton is an overload reset only.

5610-T-E-1 591, Sh. 1, Rev. 74 Technical Reference(s): 5613-E-1O, Sh. 1, Rev. 68 (Attach if not previously provided) 5614-E-10, Sh. 1, Rev. 56 Proposed References to be provided to applicants during examination: NO LP6902121BObj.3 Learning Objective: (As available)

Question Source: Bank # 100855 Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam: 2011 Point Beach PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-154 DRAFT NRC EXAM SECURE NFORMATON Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 006 K2.02 Importance Rating 2.5 Knowledge of bus power supplies to the following: Valve operators for accumulators Proposed Question: RO Question # 5 Given the following:

• A large break LOCA has occurred on Unit 3.

• The 3B 4kV Bus is faulted and locked-out.

• The crew is performing actions of 3-EOP-E-1, Loss of Reactor or Secondary Coolant.

• Conditions are met to isolate the SI Accumulators.

Which ONE of the following completes the following and identifies the Accumulator Isolation valves that are capable of being closed at this time?

(1) , Accumulator Discharge Isolation valves, can be closed (2) z A. (1) MOV-3-365A and MOV-3-365B (2) remotely from VPB.

B. (1) MOV-3-365A and MOV-3-365C (2) remotely from VPB.

C. (1) MOV-3-365A and MOV-3-365B (2) remotely from MCC using pushbutton.

D. (1) MOV-3-365A and MOV-3-365C (2) remotely from MCC using pushbutton.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part wrong. 2nd part right. MOV-3-865A, B and C are powered from Vital MCCs A, B, and C respectively. Since 3B 4kV bus powers the 3B Vital LC which in turn powers the 3B Vital MCC, MOV-33-865B cannot be energized due to the lockout.

Plausible to believe that the MOVs may be powered from 3A and 3C MCCs. On Unit 4, MOV-4-865A, B and C are powered from Vital MCCs C, A, and B respectively.

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 B. Correct. MOV-3-865A, B and C are powered from Vital MCCs A, B, and C respectively.

Since 3B 4kV bus powers the 3B Vital LC which in turn powers the 3B Vital MCC, MOV 33-865B cannot be energized due to the lockout.

C. Incorrect. 1st part wrong. 2nd part wrong. MOV-3-865A, B and C are powered from Vital MCCs A, B, and C respectively. Since 3B 4kV bus powers the 3B Vital [C which in turn powers the 3B Vital MCC, MOV-33-865B cannot be energized due to the lockout.

Plausible to believe that the MOVs may be powered from 3A and 3C MCCs. On Unit 4, MOV-4-865A, B and C are powered from Vital MCCs C, A, and B respectively. Also plausible to believe the MOVs can be operated remotely from the breaker, however, the D.

pushbutton is an overload reset only.

Incorrect. 1st part right. 2nd part wrong. MOV-3-865A, B and C are powered from Vital MCCs A, B, and C respectively. Since 3B 4kV bus powers the 3B Vital LC which in turn

-J powers the 3B Vital MCC. Plausible to believe the MOVs can be operated remotely from the breaker, however, the pushbutton is an overload reset only.

Technical Reference(s):

LP 6902121B Accumulators, (07/29/2104), Slide 31 (Attach if not previously provided) z Proposed References to be provided to applicants during examination: NO LP 6902121B Accumulators, Learning Objective: (07/29/2104), Obj 3. (As available)

Question Source: Bank # 100855 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2011 Point Beach Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 007 A3.01 Importance Rating 2.7 Ability to monitor automatic operation of the PRTS, including: Components which discharge to the PRT Proposed Question: RO Question # 6 Given the following conditions:

• Unit4isinMODE4.

• PCV-4-473, PRT Nitrogen Regulator, is adjusted to maintain 6.5 psig in the PRT.

• P1-4-472, PRT PRESS, reads 6.5 psig and stable.

• Ll-4-470, PRT LVL, reads 52% and stable.

• Containment Temperature is 115°F and stable.

Subsequently, six hours later:

• Pl-4-472, PRT PRESS, reads 60 psig and is rising slowly.

• Ll-4-470, PRT LVL, reads 52% and stable.

• Pressurizer Safety Temperatures are reading in a range of 110°F to 120°F.

• Pressurizer Relief Temperature is reading 115°F.

Which ONE of the following correctly identifies the cause of the subsequent conditions?

A. PCV-4-473, PRT Nitrogen Regulator has failed open with PRT level being controlled automatically.

B. RV-4-203, Normal Letdown Relief has failed open with PRT level being controlled automatically.

C. RV-4-203, Normal Letdown Relief has failed open with PRT level being controlled manually.

D. PCV-4-473, PRT Nitrogen Regulator has failed open with PRT level being controlled manually.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible to believe the Nitrogen Regulator failed and that the PRT level is controlled automatically since the RCDT pumps are used to transfer the contents of the PRT as well as gravity drain to the Containment Sumps. The RCDT pumps do maintain RCDT level automatically, but not the PRT.

B. Incorrect. Plausible to believe the Normal Letdown Relief has failed open, since that would cause a pressure increase, however, there would also be a level increase. It is also plausible that the PRT level is controlled automatically since the RCDT pumps are used to transfer the contents of the PRT, as well as gravity drain to the Containment Sumps. The RCDT pumps do maintain RCDT level automatically, but not the PRT.

C. Incorrect. Plausible to believe the Normal Letdown Relief has failed open, since that would cause a pressure increase, however, there would also be a level increase.

D. Correct. PCV-4-473, PRT Nitrogen Regulator has failed open with PRT level being controlled manually. This would account for the pressure increase and no level change.

PRT level control is always manual.

561 0-M-3065 SH 1 Rev 30 Technical Reference(s): 5614-M-3041 SH 2 Rev 41 (Attach if not previously provided) 5614-M-3047 SH 1 Rev 21 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 008 A1.02 Importance Rating 2.9 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the CCWS controls including: CCW temperature Proposed Question: RO Question # 7 Given the following conditions:

• Unit 3 is in MODE 4.

• TCV-3-143, L/D Demineralizer Divert Valve, is aligned to AUTO.

Subsequently:

• Annunciator A315, LTDN DEMIN HI TEMP/FLOW DIVERTED, alarms.

• TI-3-140, Regen HX Letdown outlet temperature is 165°F.

• TI-3-143, Non-Regen Hx Letdown temperature has risen to 145°F.

• FI-3-620, local NRHX CCW flow is reading 90 gpm.

Which ONE of the following completes the following statements?

The RCO will manually adjust (1) to reduce Letdown Temperature to prevent exceeding the design limits (2)

A. (1) TC-3-144A, L/D Temp Controller (2) of the CVCS Demineralizer resin B. (1) TCV-3-143, L/D Demineralizer Divert Valve (2) for VCT temperature C. (1) TC-3-144A, L/D Temp Controller (2) for VCT temperature D. (1) TCV-3-143, L/D Demineralizer Divert Valve PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION (2) of the CVCS Demineralizer resin Proposed Answer: A Explanation (Optional):

A. Correct. 1st part right. 2nd part right. Crew is directed by 3-ARP-097.CR.A A315 to take manual control of TCV-3-144A to raise CCW flow to the Non-Regenerative Heat Exchanger to reduce Letdown Temperature. This is to prevent exceeding the design limits of the CVCS Demineralizer resin.

B. Incorrect. 1st part wrong. 2nd part wrong. Plausible since TCV-3-143 does divert letdown flow when the alarm comes in, but does not affect temperature. An increased temperature may provide concern about exceeding the design limits of the VCT.

C. Incorrect. 1st part right. 2nd part wrong. Plausible to believe an increased temperature may provide concern about exceeding the design limits of the VCT.

D. Incorrect. 1st part wrong. 2nd part right. Plausible since TCV-3-143 does divert letdown flow when the alarm comes in, but does not affect temperature.

Technical Reference(s): 3-ARP-097.CR.A, Rev. 11 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO LP6902113 Obj.5b Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

Nomenclature for valve names for TCV-3-144A and TCV-3-143 not provided to provide support the plausibility of the distractors.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

ENCLOSURE 1 (Page 1 of 4)

CONTROL ROOM FUNCTIONS AND INDICATIONS LOST ON LOSS OF 3P06 FUNCTIONS, Operating Lock up of Pressurizer Pressure Controllers causing spray valves to stay as is FCV-3-478, A Feedwater Control Valve On Backup Controller Lose Auto and Manual 3A Charging Pump Control causing Auto Lock-up Lose Auto Speed Control of 3B and 3C Charging Pumps Lose the Auto Makeup Control to the Volume Control Tank Lose power to Control Relay from MOV-3-l 15C which opens LCV-3-1 15B Letdown Isolation Pressurizer heaters de-energize Lose Auto and Manual control of PCV-3-l45, Letdown Pressure Controller Loss of 3B Diesel Load Sequencer, 3C23B-l deenergized Lose AM SAC A Processor Lose the Ability to Block the Source Range Trip Lose Feedwater Isolation signal (Reactor Trip with Tavg 554°F)

Loss of power to hand/auto station for CV 1607 which fails closed F

a a a a NOTES I . The following conditions exist which affect Pressurizer Pressure control:

I a Pressurizer Pressure Controller PC-444J A UTO LOCKUP

• PZR Spray Valve Controllers AUTO LOCKUP,

• PZR heaters deenergized I

• Letdown isolation

• 3A charging pump AUTO LOCKUP

• 38 AND 3C Charging pump loss of auto speed control L a _ _ a _ a W97:mr/cls/fm

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 010 K2.02 Importance Rating 2.5 Knowledge of bus power supplies to the following: Controller for PZR spray valve Proposed Question: RO Question # 8 Which ONE of the following identifies the bus which provides power to the Spray Valve Controller for PCV-3-455B, Pressurizer Spray Valve, Loop B?

A. 3D01 B. 3D23 C. 3P06 D. 3P08 Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible since PCV-3-456, PRZ PORV, is powered from 3D01. However, PCV-3-455B is powered from 3P06 and 3P09.

B. Incorrect. Plausible since PCV-3-455C, PRZ PORV, is powered from 3D23. However, PCV-3-455B is powered from 3P06 and 3P09.

C. Correct. PCV-3-455B is powered from 3P06 and 3P09.

D. Incorrect. Plausible since PCV-3-455C and PCV-3-456 also received power from 3P08.

However, PCV-3-455B is powered from 3P06 and 3P09.

Technical Reference(s): 3-ONOP-003.6, Rev. 1 (Attach if not previously provided)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed References to be provided to applicants during examination: NO Learning Objective: LP69021O8AObj.2 LP 6902260 Obj. 2, 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Corn ments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

ENCLOSURE 1 (Page I of 4)

CONTROL ROOM FUNCTIONS AND INDICATIONS LOST ON LOSS OF 3P06 FUNCTIONS, Operating Lock up of Pressurizer Pressure Controllers causing spray valves to stay as is FCV-3-478. A Feedwater Control Valve On Backup Controller Lose Auto and Manual 3A Charging Pump Control causing Auto Lock-up Lose Auto Speed Control of 3B and 3C Charging Pumps Lose the Auto Makeup Control to the Volume Control Tank Lose power to Control Relay from MOV-3-I 15C which opens LCV-3-115B Letdown Isolation Pressurizer heaters de-energize Lose Auto and Manual control of PCV-3-145, Letdown Pressure Controller Loss of 3B Diesel Load Sequencer, 3C23B-l deenergized Lose AM SAC A Processor Lose the Ability to Block the Source Range Trip Lose Feedwater Isolation signal (Reactor Trip with Tavg 554°F)

Loss of power to hand/auto station for CV-3-1607 which fails closed F

NOTES I

• The following conditions exist which affect Pressurizer Pressure control:

I a a

Pressurizer Pressure Controller PC-444J A UTO LOCKUP PZR Spray Valve Controllers AUTO LOCKUP

. PZR heaters deenergized I

• Letdown isolation

• 3A charging pump AUTO LOCKUP I a 3B AND 3C Charging pump loss of auto speed control L

W97:mr/cls/fm

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 012 A2.03 Importance Rating 3.4 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Incorrect channel bypassing Proposed Question: RO Question # 9 Given the following conditions:

• Unit 4 is MODE 3.

• Tavg is 547° F.

• Reactor Trip Breakers are open.

• Containment Pressure Channel I Hi-Hi (PS-4-2056) is removed from service to support l&C surveillance testing.

• During testing, Containment Pressure Channel I Hi (PS-4-2007) fails high.

• While removing PS-4-2007 from service, l&C inadvertently causes Containment Pressure Channel III Hi (PS-4-2009) to trip.

Which of the following completes the following sentence?

A reactor trip signal is generated due to a (1) and entry into 4-EOP-E-0 (2) required.

A. (1) Safety Injection Signal.

(2) is B. (1) Containment Isolation signal.

(2) is C. (1) Safety Injection Signal.

(2) is NOT D. (1) Containment Isolation signal.

(2) is NOT PTN L-15-1. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: A Explanation (Optional):

A. Correct. The technician de-energizes the power to Containment Pressure Channel I Hi-Hi (PS-4-2056), which would prevent energizing its logic relays, so an automatic Phase B will not occur unless both of the other two Hi-HI signals actuate. However, Containment Pressure Channel I Hi (PS-4-2007) failed high and was removed from service by pulling the fuses for the channel which energizes logic relays for that channel. The inadvertent trip of Containment Pressure Channel Ill Hi (PS-4-2009) completes the logic and generates a Phase A signal which generates a Safety Injection signal and this in turn generates a Reactor Trip Signal. Since 4-EOP-E-0 is applicable in MODES 1, 2, and 3 (greater than 1000 psig), entry into 4-EOP-E-0 is required.

B. Incorrect. 1st part wrong. 2 part right. Plausible to believe since the Containment Pressure Hi logic 2/3 was met, that the Containment Isolation signal would generate a Reactor Trip. However, the inadvertent trip of Containment Pressure Channel Ill Hi (PS-4-2009) completes the logic and generates a Phase A signal which generates a Safety Injection signal and this in turn generates a Reactor Trip Signal. Since 4-EOP-E-0 is applicable in MODES 1,2, and 3 (greater than 1000 psig), entry into 4-EOP-E-0 is required.

C. Incorrect. l part right. 2d part wrong. Plausible to believe that since the Unit is in MODE 3 and reactor trip breakers are open, entry into 4-EOP-E-0 is not required.

However, since 4-EOP-E-0 is applicable in MODES 1, 2, and 3 (greater than 1000 psig), entry into 4-EOP-E-0 is required.

D. Incorrect. 1 st part wrong. 2 part wrong. Plausible to believe since the Containment Pressure Hi logic 2/3 was met, that the Containment Isolation signal would generate a Reactor Trip. However, the inadvertent trip of Containment. Pressure Channel Ill Hi (PS-4-2009) completes the logic and generates a Phase A signal which generates a Safety Injection signal and this in turn generates a Reactor Trip Signal. Plausible to believe that since the Unit is in MODE 3 and reactor trip breakers are open, entry into 4-EOP-E-0 is not required. Since 4-EOP-E-0 is applicable in MODES 1, 2, and 3 (greater than 1000 psig), entry into 4-EOP-E-0 is required.

Technical Reference(s): 5610-T-L1 sh 11 Rev 31 .

(Attach if not previously provided) 5614-T-L1 sh 2 Rev 1 Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902163 Obj. 8 (As available)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 012 K3.04 Importance Rating 3.8 Knowledge of the effect that a loss or malfunction of the RPS will have on the following:

ESFAS Proposed Question: RO Question # 10 Given the following conditions:

• Unit 4 is at 100%.

• Pressurizer Pressure Channel I PT-4-455 fails and is removed from service in accord ance with 4-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Chann els.

Which ONE of the following, from the remaining in-service channels, identifies the minimum number of RPS and ESF actuation logic channels required to initiate a Reactor Trip and Safety Injection on Low Pressurizer Pressure?

Reactor Trip Safety Injection Channels Channels A. 2 1 B. 1 1 C. 2 2 D. 1 2 Proposed Answer: B Explanation (Optional):

A. Incorrect, wrong logic. See B. 1/2 and 2/3 are credible distractors because the applicant must know initial trip logic and what state bistables will be in after action is taken.

B. Correct. Reactor Trip and SI is 2/3 for PZR pressure. Channel 455 feeds both circuits.

When a protection channel is removed from service, bistables are tripped in all cases except for the AUTO Containment Spray actuation. Thus, a Reactor Trip and SI will occur if either of the two remaining bistables trip. 1/2 and 2/3 are credible distractors PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION because the applicant must know initial trip logic and what state bistables will be in after action is taken.

C. Incorrect, wrong logic. See B. 1/2 and 2/3 are credible distractors because the applicant must know initial trip logic and what state bistables will be in after action is taken.

D. Incorrect, wrong logic. See B. 1/2 and 2/3 are credible distractors because the applicant must know initial trip logic and what state bistables will be in after action is taken.

5610-T-L1, Sh. 11, Rev. 31 Technical Reference(s): 5614-T-L1, Sh. 2, Rev. 1 (Attach if not previously provided) 5610-T-L1, Sh. 18, Rev. 17 Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902163 Obj. 7, 8 (As available)

Question Source: Bank # 87080 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Callaway Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 013 K6.01 Importance Rating 2.7 Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS:

Sensors and detectors Proposed Question: RO Question # 11 Given the following conditions:

• Unit 4 is at 50% power.

• PT-4-494, C S/G Pressure Transmitter Channel II, fails high.

• The bi-stables for the failed channel are tripped in accordance with 4-ONOP-049. 1, Deviation or Failure of Safety Related or Reactor Protection Channels.

Subsequently:

• PT-4-495, C S/G Pressure Transmitter Channel III, fails low.

Which ONE of the following describes how the plant will respond and why?

An automatic Safety Injection will (1) Loop C Lo Stm Pressure inputs to the comparators are indicating a header pressure 100 psig (2) S/G pressure.

A. (1) occur because 2/3 (2) greater than B. (1) occur because 2/3 (2) less than C. (1) NOT occur because 1/3 (2) greater than PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION D. (1) NOT occur because 1/3 (2) less than Proposed Answer: A Explanation (Optional):

A. Correct. An SI will occur because 2 of 3 bistables are tripped; on the first failure ONOP 049.1 trips high and low pressure bistables. So on the 2nd failure low, now 2/3 pressure comparators are indicating a header pressure 100 psig greater than S/G pressure.

B. Incorrect. Plausible to believe that header pressure needs to be 100 psig less than S/G pressure. An SI will occur because 2 of 3 bistables are tripped; on the first failure ONOP-049.1 trips high and low pressure bistables. So on the 2nd failure low, now 2/3 pressure comparators are indicating a header pressure 100 psig greater than S/G pressure.

C. Incorrect. Plausible need to realize that ONOP-049.1 trips all bistables for the failed channel. An SI will occur because 2 of 3 bistables are tripped; on the first failure ONOP 049.1 trips high and low pressure bistables. So on the 2nd failure low, now 2/3 pressure comparators are indicating a header pressure 100 psig greater than S/G pressure.

D. Incorrect. Plausible need to realize that ONOP-049.1 trips all bistables for the failed channel and plausible to believe that header pressure needs to be 100 psig less than S/G pressure. An SI will occur because 2 of 3 bistables are tripped; on the first failure ONOP-049.1 trips high and low pressure bistables. So on the 2nd failure low, now 2/3 pressure comparators are indicating a header pressure 100 psig greater than S/G pressure.

Technical Reference(s): 5610-T-L1 sheet 19 rev. 26 (Attach if not previously provided) 4-ONOP-049. 1, Rev. 2A, p. 69 Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902163 Obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Corn rnents:

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

Procedure No. Procedure

Title:

Page:

69 Deviation or Failure of Safety Related Approval Date 4-ONOP-049.1 or Reactor Protection Channels 3/31/13 ATTACHMENT 4 (Page 48 of 53)

FAILED CHANNEL BISTABLE LIST P-4-494 Steam Generator C Pressure Ref Dwgs 5610-T-D-18B; 5610-T-L1, Sh 11 and 19 Max Deviation As Compared to other Channels 120 PSIG DEVIATION RACK BISTABLE BISTABLE STATUS FUNC No. No. FUNCTION ANNUNCIATOR LOGIC AFFECTED LIGHT TION HI Steam Line LOOP C 13 BS-4-494 PH-P3 HI C 9/3 MAIN STEAMLINE 2/3 comparators (header pressure AP SI HI AP s 100 psig >SIG pressure) on 1/3 SIGs PC494 C CONTROL RELATED P RX PROTECTION RELATED S SAFETY INJECTION RELATED 1 W97:/JWB/cls/emc/cls

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 022 K2.01 Importance Rating 3.0 Knowledge of power supplies to the following: Containment cooling fans Proposed Question: RO Question # 12 Given the following plant conditions:

• Unit 3 is operating at 100% power.

• The breaker to 3A Vital MCC has tripped OPEN.

Which ONE of the following components has lost power?

A. 3B Auxiliary Building Exhaust Fan B. 3A Main Steam Penetration Cooling Fan C. 3A Normal Containment Cooler Fan D. 3B CRDM Cooler Fan Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible since MCC loads do not follow strict ABC to ABC conventions for power. On Unit 4, similar loads are powered from completely different MCCs. Power to this fan is supplied from the 3B Vital MCC B. Incorrect. Plausible since MCC loads do not follow strict ABC to ABC conventions for power. On Unit 4, similar loads are powered from completely different MCCs. Power to this fan is supplied from the 3B Non-vital MCC C. Correct. Plausible since MCC loads do not follow strict ABC to ABC conventions for power. On Unit 4, similar loads are powered from completely different MCCs. 3V1A (3A NCC Fan) is powered from 3A Vital MCC and will lose power if the MCC is lost.

D. Incorrect. Plausible since MCC loads do not follow strict ABC to ABC conventions for power. On Unit 4, similar loads are powered from completely different MCCs. Power to PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION this fan is supplied from the 3C Vital MCC Technical Reference(s): 561 3-E-1 0 Sh 1, Rev 68 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902129 Obj. 3 (As available)

Question Source: Bank # 86946 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Harris Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 022 K3.01 Importance Rating 2.9 Knowledge of the effect that a loss or malfunction of the CCS will have on the following:

Containment equipment subject to damage by high or low temperature, humidity, and pressure Proposed Question: RO Question # 13 Given the following conditions:

• Unit 4 is at 100% power.

Subsequently:

• A fault occurs on the 4C SG inside containment.

• A Reactor Trip and Safety injection occur.

• All four Normal Containment Cooler Fans trip.

• 4C SG completely depressurizes.

• The Pressurizer is empty.

• Containment temperature is 197°F and lowering.

• Containment pressure is 15 psig and lowering.

• The plant conditions are stabilized.

Which ONE of the following identifies the containment instrumentation that is inoperable under the given conditions?

A. Particulate/Gas Monitors, R-11/R-12 B. High Range Radiation Monitors, Rl-4-631 lA/B C. 4C SG Water level Narrow range PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. Pressurizer Level Proposed Answer: A Explanation (Optional):

A. Correct. Containment temperature above 125°F or sample pressure of 5 psig will close isolation MOVs and isolate R-11/12 from containment. The Phase A signal from High Containment Pressure of 4 psig will also isolate R-11/12.

B. Incorrect. Plausible to believe that adverse containment conditions will render the High Range Monitor inoperable. However, this instrument is qualified as ESF instrumentation and will function in adverse containment conditions.

C. Incorrect. Plausible to believe that adverse containment conditions will render the 4C SG narrow range level instrument inoperable, since there is also no level indication..

However, this instrument is qualified as ESF instrumentation and will function in adverse containment conditions.

D. Incorrect. Plausible to believe that adverse containment conditions will render the Pressurizer level instrument inoperable, since there is also no level indication..

However, this instrument is qualified as ESF instrumentation and will function in adverse containment conditions.

Technical Reference(s): 5610-T-L1 Sh. 11 Rev. 31 .

4-OSP-203.1 Rev 14C (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902163 Obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

ATTACHMENT 7 (Page 2 of 15)

CONTAJNMENT ISOLATION PHASE A & B TEST INITIALS CKD VERIF 2.0 Test Procedure

1. Perform the Phase A, Pretest Lineup using Tables 1, 2 and 3 of this Attachment.

CAUTION Step 2 will trip R-11 and R-12, which is required during the movement of irradiated fuel or core alterations (Tech Spec 3.9.13). The action statement for Tech Spec 3.9.13 requires the containment ventilation isolation valves be maintained closed and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the Control Room Ventilation System be placed in the recirculation mode.

2. Momentarily c1 he LEFT PHASE A Containment. isolation actuate pushbuttOn,

3. IF core alterations are in progress, THEN verify the requirements of Tech Spec 3.9.13 are satisfied by performing Table 4 of this attachment.

4. Check visually the status of Containment Isolation Rack Lockout relays using Table 3 of this attachment.

5. Record the Post-Isolation Status using Tables I and 2 of this attachment.

6. Perform the following to maintain control room ventilation in recirculation mode when the containment ventilation isolation lockout relay is reset:

a. Ensure Emergency Inlet Damper D-3 control switch is in Open.

b. Ensure Emergency Inlet Damper D-2 control switch is in Open.

7. Manually Reset the following Lockout relays at 4QR50:

a. Circuit #2 Cntmt Isol Phase A Lockout CIAI I

b. Circuit #2 Cntmt Isol Phase A Lockout CIAI 12

c. Circuit #2 Cntmt Iso! Phase A Lockout CIAI 13

d. Circuit #2 Cntmt Vent Iso! Lockout CIVIl AI9fl1flI

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KJA# 026 A4.01 Importance Rating 4.5 Ability to manually operate and/or monitor in the control room: CSS controls Proposed Question: RO Question # 14 Given the following conditions:

• A Unit 3 Reactor Trip and Safety Injection occurs.

• RCS pressure is 1000 psig and stable.

• Containment pressure rises to 16 psig and continues to slowly rise.

• Annunciator H5/2, CNTMT ISOLATION ACTIVATED, is in alarm.

• The crew is performing 3-EQ P-E-0, Reactor Trip or Safety Injection.

• The RCO is checking Containment Spray requirements lAW Attachment 3.

Which ONE of the following describes whether Containment Spray (CS) is required and the next required operator action(s) in accordance with Attachment 3?

A. CS is required. Manually initiate CS using CS pushbutton. Verify CS Isolation Valves open.

B. CS is required. Manually initiate CS by starting the CS pumps. Manually open CS Isolation Valves.

C. CS is not required. Verify SI is reset. Verify SI Amber Lights on VPB are ALL BRIGHT.

D. CS is not required. Verify SI is reset and stop ALL RCPs.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible to think it is required since Annuciator H512, CNTMT ISOLATION ACTIVATED, is in alarm, but there are multiple inputs to this annunciator that do not required CS, although CS actuation will also actuate this alarm. CS would be required if containment pressure was >20 psig.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION B. Incorrect. Plausible to think it is required since Annunciator H5/2, CNTMT ISOLATION ACTIVATED, is in alarm, but there are multiple inputs to this annunciator that do not required CS, although CS actuation will also actuate this alarm. CS is not required. CS would be required if containment pressure was >20 psig.

C. Correct. CS is not required. RCO checks Containment pressure has remained less than 20 psig and then Attachment 3 has RCO verifying SI reset and Safety Injection Valve Amber lights on VPB are ALL BRIGHT.

D. Incorrect. Plausible since Containment pressure is still rising. RCPs may have been stopped due to RCP trip criteria on FOP, but would not be stopped due to Phase B (20 psig) which did not occur nor containment isolation (4 psig) 0-ADM-21 1 Rev 3 Technical Reference(s):

(Attach if not previously provided) 5610-T1-L Sh 11 Rev. 31 Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902163 Obj. 8 (As available)

Question Source: Bank # 87498 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 North Anna Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

3 EMERGENCY AND OFF-NORMAL OPERATING PROCEDURE USAGE 26 of 46 PROCEDURE NO.:

O-ADM-21 1 TURKEY POINT PLANT 4.7 Procedure Adherence for Emergency and Off Normal (Abnormal)

Procedures (continued)

2. B. (continued)

(5) Actions taken to protect personnel or equipment whenever an imminent threat may exist for events such as high energy line breaks or electrical bus faults. This includes, but is NOT limited to the following:

• Closing Main Steam Isolation Valves when a fault in the Main Steam System is identified

• De-energizing an electrical bus in response to a report of a personnel electrocution (6) If redundant stand-by equipment is available and ready, the operator is permitted to start the redundant equipment for failed or failing operating equipment.

Immediate follow up of applicable ARPs and ONOPs (AOPs) shall occur as required.

C. The operator should announce his intentions to perform a prudent operator action. This will allow the remaining crew members to independently evaluate the current plant conditions and validate or refute, as necessary, the need for the manual action.

D li1anual actuation of Containment Spray and Containment Phase B Isolation shall NOT be performed before the setpoint is reachedj

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 039 A3.02 Importance Rating 3.1 Ability to monitor automatic operation of the MRSS, including: Isolation of the MRSS Proposed Question: RO Question # 15 Given the following conditions:

• Unit4isinMODE2.

• Warm-up of the Main Steam lines is in progress.

• Main Steam Isolation Bypass Valves are open.

• Main Steam Isolation Valves were just opened.

Subsequently:

• A major steam line break occurs just upstream of the Turbine Stop Valves.

• Annunciator C81 1, MAIN STEAMLINE LO PRESS, is lit.

• RCS Tavg is 542° F.

Which ONE of the following completes the statement below?

The Main Steam Isolation Valves (1) and Main Steam Isolation Bypass Valves (2)

A. (1) automatically close (2) must be manually closed B. (1) automatically close (2) automatically close C. (1) must be manually closed (2) automatically close PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. (1) must be manually closed (2) must be manually closed.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part right. 2nd part wrong. Plausible to not remember that Safety Injection is not blocked when in MODE 2 and Main Steam Line Isolation will generate automatic closure of both the MSIVs and MSIV Bypass valves.

B. Correct. Main Steam Line Isolation from High steam Flow and Low Tavg will close MSIVs and Bypass valves.

C. Incorrect. 1st part wrong. 2nd part right. Plausible to not remember that Main Steam Line Isolation from High steam Flow and Low Tavg will close MSIVs and bypass valves, since the Unit is in MODE 2 and believe that manual action must be taken to isolate main steam.

D. Incorrect. 1st part wrong. 2nd part wrong. Plausible to not remember that Main Steam Line Isolation from High steam Flow and Low Tavg will close MSIVs and bypass valves, since the Unit is in MODE 2 and believe that manual action must be taken to isolate main steam.

5610-T-L1 Sh. 11 Rev 31 Technical Reference(s): .

(Attach if not previously provided) 5610-T-L1 Sh. 19 Rev 26 Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902163 Obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE NFORMATlON Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 059 A4.03 Importance Rating 2.9 Ability to manually operate and monitor in the control room: Feedwater control during power increase and decrease Proposed Question: RO Question # 16 Given the following conditions:

• Unit 3 Reactor power is currently on hold at 8%.

• The RCO is manually controlling SG levels at 50%.

Subsequently:

• The Unit Supervisor directs raising power to 30%.

• During Turbine load increase, the Turbine Control valves opened rapidly.

• Reactor power increases to 14%.

• All SG level deviations are in alarm.

Which ONE of the following completes the following statements?

The initial SG level deviation alarms occur because SG narrow range levels indicate 5% (1) than programmed level. The RCO must eventually throttle (2) the Feedwater Regulating Bypass valves to maintain S/G levels.

A. (1) less (2) open B. (1) greater (2) close PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. (1) less (2) close D. (1) greater (2) open Proposed Answer: 0 Explanation (Optional):

A. Incorrect: 1st part wrong. 2nd part right. Plausible to believe that increasing steam flow would reduce S/G inventory and therefore level. The rapid increase in steam demand caused swell in the SIGs, which would be indicated by a rising level.

B. Incorrect: 1st part right. 2nd part wrong. Plausible to believe that since the higher level was due to a transient, feed flow might not need adjusting. Increased steam demand would require increased feed flow to maintain level, in spite of the initial swell in the S/Gs due to the steam flow increase.

C. Incorrect: 1st part wrong. 2nd part wrong. See A and B.

D. Correct: The rapid increase in steam demand caused swell in the SIGs, which would be indicated by a rising level. Increased steam demand would require increased feed flow to maintain level, in spite of the initial swell in the SIGs due to the steam flow increase.

Technical Reference(s): 5613-T-D-17, Rev 3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 059 2.1.28 Importance Rating 4.1 Conduct of Operations: Knowledge of the purpose and function of major system components and controls.

Proposed Question: RO Question # 17 Given the following:

• Unit 4 is at 60% power.

Subsequently:

• A Condensate header rupture occurs.

• Both Main Feedwater Pumps trip.

• After 10 seconds, SG levels are 30% narrow range and lowering.

Which ONE of the following completes the following statements?

An automatic Reactor Trip setpoint (1) exceeded. AFW Pumps (2) running.

A. (1) is (2) are B. (1) is (2) are NOT C. (1) is NOT (2) are D. (1) is NOT (2) are NOT PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: C Explanation (Optional):

A. Incorrect. RPS trip setpoint was not exceeded, because the steam flow/feed flow mismatch is not active until SG level reaches 16%. Plausible to believe that only the steam flow/feed flow mismatch is necessary. AFW pumps started when both Feed Pumps were tripped.

B. Incorrect. See A. AFW pumps are running, but plausible to believe that AFW wont auto start until level is 16% in any SG and omit the loss of all feed pumps start.

C. Correct. Logic for Reactor trip is not met as noted above, but AFW auto start met based on information provided in stem.

D. Incorrect. Logic for Reactor trip is not met as noted above, AFW pumps are running, but plausible to believe that AFW wont auto start until level is 16% in any SG and omit the loss of all feed pumps start.

Technical Reference(s): 5610-T-L1 Sh 15 Rev 35 (Attach if not previously provided) 5614-T-L1 Sh 2 Rev 1 Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902163 Obj. 7 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 North Anna Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 55.43 Comments:

KA Match: Item evaluates cause-effect releationship of Main Feed Pump trip vs. AFW pump auto start circuitry.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 061 K5.01 Importance Rating 3.6 Knowledge of the operational implications of the following concepts as the apply to the AFW:

Relationship between AFW flow and RCS heat transfer Proposed Question: RO Question # 18 Given the following conditions:

• Unit 3 is increasing power following a refueling outage.

• Unit 3 experiences a Reactor trip from 25% power.

• The crew completes 3-EO P-F-U, Reactor Trip and Safety Injection, and enters 3-EOP-ES-0.1, Reactor Trip Response.

• Pressurizer level is 14% and slowly decreasing.

• All Steam Generator Narrow Range levels are between 12% and 15% and slowly rising.

• Steam Generators pressures are approximately 990 psig and slowly decreasing.

• Tavg is 544° F and slowly decreasing.

• RCS pressure is 2125 psig and slowly decreasing.

Which ONE of the following identifies the crews initial required response in accordance with 3-EOP-ES-0.1 to address the conditions above?

A. Establish Emergency Boration.

B. Reduce Auxiliary Feedwater Flow.

C. Close MSIVs and bypass valves.

D. Initiate a Safety Injection and return to 3-EOP-E-0.

Proposed Answer: B Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. Incorrect. Plausible since 3-EOP-ES-0.1 addresses boration at highest rate possible if RCS Tavg drops to less than 537°F in Continuous Action Summary, B. Correct. 3-EOP-ES-0.1 step 1 .c RNO directs the reduction of AFW to address the cooldown.

C. Incorrect. 3-EOP-ES-0.1 step 1 .c RNO directs the closing of MSIVs and bypass valves if the cooldown continues in step 1 .c RNO making this choice plausible, but this action is after the reduction of AFW flow.

D. Incorrect. 3-EOP-ES-0.1 CAUTION prior to step 1 directs the initiation of SI and return to E-0 but only if SI is actuated. Actuation of SI is not warranted for the stated conditions. Plausible for the candidate to misuse the data trends in stem and conclude that SI will be required.

3-EOP-ES-O.1 Reactor Trip Technical Reference(s):

(Attach if not previously provided)

Response, Rev 10 Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902323 Obj. 2, 3 (As available)

Question Source: Bank # 87680 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 Sequoyah Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

10 REACTORTRIP RESPONSE 6of68 PROCEDURE NO.:

3-EOP-ES-0.1 TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

1. (continued)

c. Check RCS Average Tempeture& c. Perform the following:

using DCS

--- 1) !E TAvE is decreasing, iABLE THEN perform the following:

BETWEEN 545°F AND 547°F a) Stop dumping steam:

OR TRENDING DOWN TO 547°F switch in MANUAL.

Verify Steam Dump To Atmosphere valves closed.

b) Verify SIG Blowdown Isolation valves CLOSED.

c) IF cooldown continues THEN reduce total feed flow to between 400 and 450 gpm Narrow Range Level greater, than 7% in at least one S/G.

d) IF cooldown continues, AND is due to excessive steam flow, THEN close Main Steamline Isolation and Bypass valves.

2) IF TAVE greater than 547°F AND increasing, THEN:

Dump steam to Condenser.

OR Dump steam using SIG Steam Dump To Atmosphere valves.

3) IF RCS TAVE less than 537°F, THEN initiate and continue a boration of at least 20 gpm using Attachment 12 unt Shutdown Margin can be verified using PLANT CURVE BOOK.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 061 K6.02 Importance Rating 2.6 Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Pumps Proposed Question: RO Question # 19 Given the following conditions:

• The Auxiliary Feedwater (AFW) system receives an auto-start signal.

• A AFW Pump governor malfunctions and AFW Turbine speed rises.

Which ONE of the following describes the effect of this event?

The A AFW Pump trips on over-speed at (1)

The Train 1 AFW Flow Control Valves will (2)

A. (1) 5900 rpm (2) remain throttled B. (1) 5900 rpm (2) fully open C. (1) 6500 rpm (2) remain throttled D. (1) 6500 rpm (2) fully open Proposed Answer: D PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Incorrect. 1 st part wrong. 2nd part wrong. Plausible to remember normal speed as the overspeed setpoint. Plausible because they will get an open signal and the controllers do have a preset demand for 135 gpm which would keep them throttled. However, they will go to and remain in the fully open position as a result of the lack of flow condition.

B. Incorrect. 1 St part wrong. 2nd part right. Plausible to remember normal speed as the overspeed setpoint. They will go to and remain in the fully open position as a result of the lack of flow condition.

C. Incorrect. 1st part right. 2 part wrong. Plausible because they will get an open signal and the controllers do have a preset demand for 135 gpm which would keep them throttled. However, they will go to and remain in the fully open position as a result of the lack of flow condition.

D. Correct. The overspeed trip setpoint is 6500 rpm. The flow control valves received an open signal and they will go to and remain in the fully open position as a result of the lack of flow condition.

5610-T-L1 Sheet 15 Rev 35 Technical Reference(s): .

(Attach if not previously provided) 3-OSP-075.10 Rev 2 Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902123 Obj. 6, 9 (As available)

Question Source: Bank # 88592 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

2 AFW FLOW CONTROL VALVE OPERABILITY TEST 4 of 21 PROCEDURE NO.:

3-OSP-075.1O TURKEY POINT UNIT 3 1.0 PURPOSE AND SCOPE 1.1 Purpose This procedure provides instructions for cycling the AFW Flow Control Valves for troubleshooting and/or Post Maintenance Testing.

1.2 Scope 1.2.1 Frequency None 1.2.2 Applicability None 1.2.3 MODE Restrictions None 2.0 PRECAUTIONS AND LIMITATIONS 2.1 Precautions

1. Testing two trains of AFW at a time will place Unit 3 in Technical Specification 3.0.3.

2. Immediate notification of the Unit Supervisor is required ifj2y Functional Criteria is NOT met or any malfunctions or abnormal condition occurs.

3. A train of AFW will be NOT be OPERABLE with y AFW Hand Indicating Controller NOT selected to automatic mode of operation

4. A train of AFW will be NOT be OPERABLE with jjy AFW Hand Indicating Controller NOT set at 135 gpm 2.2 Limitations I Hand Indicating Controllers (HIC-3-1401A HIC-3-1401 B, HIC-3-1457A, HIC-3-1457B, HIC-3-1458A and HIC-3-1458B) for auxiliary feedwater regulating valves shall be in Auto mode of operation with each HIC set for 135 gpm flow except as directed by this procedure.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 062 2.2.39 Importance Rating 3.9 Equipment Control: Knowledge of less than or equal to one hour technical specification action statements for systems.

Proposed Question: RO Question # 20 Given the following plant conditions:

• Unit3isinMODEl.

• Unit4 isin MODE 1.

• A lightning strike damages the Unit 3 Startup Transformer.

Which ONE of the following statements correctly describes the required response in accordance with TS 3.8.1.1?

Within one hour, the Unit 3 crew (1) required to demonstrate the OPERABILITY of the Unit 4 Startup Transformer and (2) required to demonstrate the OPERABILITY of the Unit 3 EDGs.

A. (1)is (2) is B. (1)isnot (2) is C. (1)is (2) is not D. (1)isnot PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION (2) is not Proposed Answer: C Explanation (Optional):

A. Incorrect. 1st part right. 2nd part wrong. TS 3.8.1.1 requires both Startup Transformers and their associated circuits. Action 3.8.1.1.a states that With one of two startup transformers or an associated circuit inoperable, demonstrate the OPERABILITY of the other startup transformer and its associated circuits by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

Plausible to believe that if the Units Startup Transformer is lost, then the EDGs need to be OPERABLE.

B. Incorrect. 1st part wrong. 2nd part wrong. Plausible to believe that since this is a problem with Unit 3, that the Unit 4 transformer is not affected. TS 3.8.1.1 requires both Startup Transformers and their associated circuits. Action 3.8.1.1.a states that With one of two startup transformers or an associated circuit inoperable, demonstrate the OPERABILITY of the other startup transformer and its associated circuits by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. Plausible to believe that if the Units Startup Transformer is lost, then the EDGs need to be OPERABLE.

C. Correct. TS 3.8.1.1 requires both Startup Transformers and their associated circuits.

Action 3.8.1.1.a states that With one of two startup transformers or an associated circuit inoperable, demonstrate the OPERABILITY of the other startup transformer and its associated circuits by performing Surveillance Requirement 4.8.1.1 .1 .a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

D. Incorrect. 1st part wrong. 2nd part right. Plausible to believe that since this is a problem with Unit 3, that the Unit 4 transformer is not affected. TS 3.8.1.1 requires both Startup Transformers and their associated circuits. Action 3.8.1.1.a states that With one of two startup transformers or an associated circuit inoperable, demonstrate the OPERABILITY of the other startup transformer and its associated circuits by performing Surveillance Requirement 4.8.1.1.1 .a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

Technical Reference(s): TS 3.8.1.1 Rev 293 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Learning Objective: LP 6902138 Obj. 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-]. DRAFT NRC EXAM SECURE INFORMATION

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3 8 11 As a minimum the rC electrical po shall be OPERABLE a Two startup iers and their asuits and

b. Three separate and independent diesel generators* including,

1) For Unit 3, two (3A and 3B); for Unit 4, one (3A or 3B) each with:

a) A separate skid-mounted fuel tank and a separate day fuel tank with an OPERABLE solenoid valve to permit gravity flow from the day tank to the skid mounted tank, and with the two tanks together containing a minimum of 2000 gallons of fuel oil.

b) A common Fuel Storage System containing a minimum volume of 38,000 gallons of fuel,**

c) A separate fuel transfer pump,**

d) Lubricating oil storage containing a minimum volume of 120 gallons of lubricating oil, e) Capability to transfer lubricating oil from storage to the diesel generator unit, and f) Energized MCC bus (MCC 3A vital section for EDG 3A, MCC 3K for EDG 3B).

2) For Unit 3, one (4A or 4B); for Unit 4, two (4A and 4B) each with:

a) A separate day fuel tank containing a minimum volume of 230 gallons of fuel, b) A separate Fuel Storage System containing a minimum volume of 34,700 gallons of fuel, c) A separate fuel transfer pump, and d) Energized MCC bus (MCC 4J for EDG 4A, MCC 4K for EDG 4B).

• Whenever one or more of the four EDGs is out-of-service, ensure compliance with the EDG requirements specified in Specifications 3.5.2 and 3.8.2.1.

• • A temporary Class Ill fuel storage system containing a minimum volume of 38,000 gallons of fuel oil may be used for up to 10 days during the performance of Surveillance Requirement 4.8.1.1.21.1 for the Unit 3 storage tank while Unit 3 is in Modes 5, 6, or defueled. If the diesel fuel oil storage tank is not returned to service within 10 days, Technical Specification 3.8.1.1 Action b and 3.8.1.2 Action apply to Unit 4 and Unit 3 respectively.

TURKEY POINT UNITS 3 & 4 3/4 8-1 AMENDMENT NOS. 197 AND 191

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one of two startup transformers or an associated circuit inoperable, demonstrate the OPERABILITY of the other startup transformer and its associated circuits by performing Surveillance Requirement 4 8 1 1 1 a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter If the inoperable startup transformer is the associated startup transformer and became inoperable while the unit is in MODE 1 reduce THERMAL POWER to 30% RATED THERMAL POWER within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or restore the inoperable startup transformer and associated circuits to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. If THERMAL POWER is reduced to 30% RATED THERMAL POWER within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if the inoperable startup transformer is associated with the opposite unit restore the startup transformer and its associated circuits to OPERABLE status within 30 days of the loss of OPERABILITY, or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. If the inoperable startup transformer is the associated startup transformer and became inoperable while the unit was in MODE 2, 3, or 4 restore the startup transformer and its associated circuits to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This ACTION applies to both units simultaneously.

b. With one of the required diesel generators inoperable, demonstrate the OPERABILITY of the above required startup transformers and their associated circuits by performing Surveillance Requirement 4.8.1 .1.1 .a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. If the diesel generator became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned preventative maintenance or testing, demonstrate the OPERABILITY of the remaining required diesel generators by performing Surveillance Requirement 4.8.1.1 .2.a.4 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless the absence of any potential common mode failure for the remaining diesel generators is determined. If testing of remaining required diesel generators is required, this testing must be performed regardless of when the inoperable diesel generator is restored to OPERABILITY. Restore the inoperable diesel generator to OPERABLE status within 14 days** or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With one startup transformer and one of the required diesel generators inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1 l.a on the remaining 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if inoperability is associated with Action Statement 3.8.1.1.c.

TURKEY POINT UNITS 3 & 4 3/4 8-2 AMENDMENT NOS. 215 AND 209

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 063 K1.02 Importance Rating 2.7 Knowledge of the physical connections and/or cause-effect relationships between the dc electrical system and the following systems: AC electrical system Proposed Question: RO Question # 21 Which ONE of the following describes the effect of placing the yellow NORMAL/ISOLATE switch to ISOLATE on the 4B HHSI Pump breaker cubicle?

A. Enables the pumps control switch on the 4KV Breaker cubicle door.

B. Aligns backup fuses into the 4B HHSI Pump control circuit.

C. Allows local starts at the SI Pump Room.

D. Disables all 4B HHSI Pump Breaker protective trip signals.

Proposed Answer: B Explanation (Optional):

A. Incorrect. The CS on the 4B HHSI Pump is already enabled. Plausible because if the 4B HHSI Pump was not safety related equipment, placing the yellow switch in ISOLATE would enable the CS on the cubicle door.

B. Correct. Per the references, placing the yellow Normal/Isolate switch to ISOLATE substitutes backup fuses into the 4B HHSI pump trip and close circuits.

C. Incorrect. The local PB is not disabled by placing the yellow switch to ISOLATE.

Plausible because this switch will disable the control room CS.

D. Incorrect. The 4B HHSI Pump breaker trip signal from bus stripping is not disabled when the yellow switch is placed in ISOLATE. Plausible because the breaker close signal from the sequencer is disabled by the yellow switch.

Technical Reference(s): 5614-E-25 sheet 3B rev. 6 . .

5614-E-25 sheet 3B2 rev. 3 (Attach if not previously provided)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed References to be provided to applicants during examination: N Learning Objective: 6900160 E03.b (As available)

Question Source: Bank # 90469 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 K/A# 063 K4.04 Importance Rating 2.6 Knowledge of dc electrical system design feature(s) and/or interlock(s) which provide for the following: Trips Proposed Question: RO Question # 22 Given the following conditions:

• The 3A RHR Pump is started.

• After the pump starts, DC control power is lost.

Which ONE of the following completes the following sentence?

The 3A RHR Pump Breaker 3AA15 A. can NOT be tripped remotely.

The breakers blue light is extinguished.

B. can be tripped remotely.

The breakers blue light is extinguished.

C. can NOT be tripped remotely.

The breakers blue light is on.

0. can be tripped remotely.

The breakers blue light is on.

Proposed Answer: A Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. Correct. When DC control power is lost to a running 4KV pump, the only way to trip the pump is locally by mechanical actuation of the trip coil.

B. Incorrect. 1st part wrong. 2 nd part right. Plausible to remember that there are three separate DC fuses for the RHR pump breaker and to believe that not all three can fail.

However, the given information states that all control power is lost, which means that the three circuits are de-energized. The trip coils are energized to actuate and with no power available the only way to trip the pump is locally by mechanical actuation of the trip coil.

C. lncorrect.lst part right. 2 part wrong. Plausible to believe that the blue light energizes with loss of DC power since a blue lockout light flashes for a lockout. However, DC control power provides power to the blue light when DC power is available.

D. Incorrect. 1 st part wrong. 2 part wrong. Plausible to remember that there are three separate DC fuses for the RHR pump breaker and to believe that not all three can fail.

However, the given information states that all control power is lost, which means that the three circuits are de-energized. The trip coils are energized to actuate and with no power available the only way to trip the pump is locally by mechanical actuation of the trip coil. Plausible to believe that the blue light energizes with loss of DC power since a

blue lockout light flashes for a lockout. However, DC control power provides power to the blue light when DC power is available.

Technical Reference(s): 5613-E-25 Sh. 4A Rev 4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank # 90056 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Braidwood Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

The question meets the K/A, requires examinee knowledge of DC electrical system design features that provide for trips. 4KV pump control circuits have trip coils that are energized to actuate.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 064 A1.03 Importance Rating 3.2 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including: Operating voltages, currents, and temperatures Proposed Question: RO Question # 23 Given the following conditions:

• The 4A EDG is running in parallel with 4A 4KV bus.

• The 4A EDG is running at 2800 KW and 500 KVAR out in the LAG.

Subsequently:

• The RCO positions the Voltage Regulator Switch to LOWER.

Which ONE of the following predicts the initial behavior of EDG parameters?

A. (1) VARs increase (2) current and temperature lower B. (1) VARs decrease (2) current and temperature lower C. (1) VARs increase (2) current and temperature rise D. (1) VARs decrease (2) current and temperature rise Proposed Answer: B PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Incorrect: Plausible to incorrectly remember action to adjust reactive load. Increas ing vars will cause more heating of the generator. When the RCO momentarily positioned the 4A EDG Volt Regulator to Raise and 4A EDG Amps decreased, 4-OP-0 23, Step 7.1.2.4.k.(3) directs the operator to raise voltage until amps increase, Increas ing vars will cause more heating of the generator.

B. Correct: When the RCO momentarily positioned the 4A EDG Volt Regulator to Raise and 4A EDG Amps decreased, 4-OP-023, Step 7.1 .2.4.k.(3) directs the operato r to raise voltage until amps increase. Increasing vars will cause more heating of the generator.

Plausible to believe for a constant load, raising voltage will decrease amps and with load constant, temperature remains stable.

C. Incorrect: Plausible to incorrectly remember action to adjust reactive load and to believe for a constant load, lower voltage will increase amps and with load constant, temperature remains stable. When the RCO momentarily positioned the 4A EDG Volt Regulator to Raise and 4A EDG Amps decreased, 4-OP-023, Step 7.1.2.4.k.(3) directs the operator to raise voltage until amps increase. Increasing vars will cause more heating of the generator.

D. Incorrect: Plausible to incorrectly remember action to adjust reactive load. Increas ing vars will cause more heating of the generator. When the RCO momentarily positioned the 4A EDG Volt Regulator to Raise and 4A EDG Amps decreased, 4-OP-0 23, Step 7.1.2.4.k.(3) directs the operator to raise voltage until amps increase. Increas ing vars will cause more heating of the generator.

Technical Reference(s): 4-OP-023 Rev 8 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

NO Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE NFORMA1iON Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

1NIT 7.1 .2.4 (Contd)

g. Verify 4A Diesel Generator frequency is between 59.4 and 60.6 Hz on the A Diesel Hertz indicator.

a When the diesel generator breaker closes, the following indicators will de-energize:

I

• WHITE SYNC LIGHTS

• INCOMING voltmeter SYNCHROSCOPE

• RUNNING voltmeter I Voltage indication remains available on the A Diesel Kilovolts indicator.

h. WHEN the Synchroscope pointer is at 12 oclock position, THEN Close the diesel generator breaker by placing the EDG A or 4A 4KV Bus 4AA20 switch to the Close position (spring return to normal).

(1) Verify the Diesel Generator Breaker 4AA20 has Closed (breaker green light is Off and red light is On).

i. Place the EDG A Sync to 4A 4KV Bus 4AA20 Synchroscope switch to Off.

j. Turn the A Diesel Gen Speed Changer in the Raise direction AND slowly increase diesel generator load to approximately 1000 KW on A Diesel Kilowatts indicator.

• NOTE I The following voltage adjustment will place the generator reactive load in lag.

I

k. Perform the following to adjust the reactive load:

(1) 4 V hile monitoring the A Diesel Amps indicator, momentarily position the A Diesel Gen Volt Regulator to Raise.

(2) IF A Diesel Amps increased, THEN perform the following:

(a) Slowly Lower the voltage until amps stop decreasing and start to increase (lead).

(b) Slowly Raise the voltage until amps increase (slightly in lag).

OR (3) j A Diesel Amps decreased, THEN slowly Raise the voltage until amps increase (slightly in lag).

W2OO3:FRZ/mr/cs/cs

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 064 K1.04 Importance Rating 3.6 Knowledge of the physical connections and/or cause-effect relationships between the ED/G system and the following systems: DC distribution system Proposed Question: RO Question # 24 Given the following conditions:

• A bus fault has caused 125 VDC Electrical Distribution Bus 3D01 to de-energize.

Which ONE of the following plant components will be directly affected by this loss of DC power?

A. Loss of Train 2 Feedwater Isolation capability B. Loss of all 3D Switchgear control power C. Loss C AFW Pump control and protection D. Loss of 3A EDG DC power Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible because this occurs with a loss of DC Bus 3D23.

B. Incorrect. Plausible because this occurs with a loss of DC Bus 4D23 and 4D01.

C. Incorrect. Plausible because this occurs with a loss of DC Bus 3D23.

D. Correct. Bus 3D01 provides DC power to 3A EDG.

• 3-ONOP-003.4 Rev 3 Technical Reference(s): (Attach if not previously provided) 3-ONOP-003.5 Rev 2A PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902139 Obj. 5 (As available)

Question Source: Bank # 98648 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Comanche Peak Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

2A LOSS OF DC BUSES 3D23 AND 3D23A (3B) 4 of 25 PROCEDURE NO:

3-ONOP-003.5 TURKEY POINT UNIT 3 1.0 PURPOSE Provides instructions to stabilize the plant and recover 3D23 and 3D23A, 3B DC BUS, in the event they are de-energized with the unit initially in MODE 1, Power Operation, and the Auxiliary Transformer supplying Plant loads. This procedure is performed after the unit has been stabilized per 3-EOP-ES-O.1, Reactor Trip Response.

2.0 ENTRY CONDITIONS 2.1 Indications

• Reactor and Turbine Trip due to loss of DC Power (3D23-8) to Reactor Trip Breaker B undervoltage coil.

• DC Load Center 3D23 and 3D23A, 3B DC BUS, voltmeter indicates voltage is zero.

• IF initially aligned to 3D23 and 3D23A, THEN 480V LC 4H and 4160V Swgr 4D control power transferred to 3D01 and 3DO1A.

• Any of the following inverters that are in service transfer to CVT

• 3Y05, INVERTER

• 4Y05, INVERTER

• 3Y06, SPARE INVERTER

• IF LC 3C was available and LC 3H was aligned to LC 3D, THEN [C 3H will transfer to LC 3C.

• IF LC 3C was NOT available, THEN LC 3H and MCC 3D are deenergized.

• Various valves fail as indicated on Attachment 1, Valve Failure Positions For Loss of DC Bus 3D23 and 3D23A (3B).

power to Backup Generator it Relay

• Loss of 3B Bus Load Sequencer NOTE The 3B EDG does NOT have black start capabilities.

• Lossof3BEDG

• Loss of Train 2ability of SIG Feedwater Bypass valves

REVISION NO.: PROCEDURE TITLE: PAGE:

2A LOSS OF DC BUSES 3D23 AND 3D23A (38) 5 of 25 PROCEDURE NO.:

3-ONOP003.5 TURKEY POINT UNIT 3 2.1 Indications (continued)

. Loss of AFW Pump C control and protection

• Loss of Train 2 AFW

• Loss of Train B Safeguards Equipment

• Loss of 480V Load Centers 3B and 3D Breaker control power NOTE Position indication for the Main Steam Isolation Valves on the console will be unavailable, position will need to be determined locally.

• Loss of power to Main Steam Isolation Valves, Channel B, valves remain open, only Channel A power available to close

• Loss of power to Unit 3, 94/AST, Turbine Trip ability, only 94/ASB available

• Loss of power to Unit 4, 94IASB, Turbine Trip ability, only 94/AST available NOTE Bypass Feedwater Inlet Isolation Valves failed open places Unit 3 in Action Statement for TS 3.7.1 .7

• Bypass Feedwater Inlet Isolation Valves fail open, require local manual operator override to close

• POV-3-477, S/G A FW BYPASS FCV-3-479 INLET ISOL VLV

• POV-3-487, SIG B FW BYPASS FCV-3-489 INLET ISOL VLV

• POV-3-497, S/G C FW BYPASS FCV-3-499 IN LET ISOL VLV 2.2 Alarms

• X114 Sequencer 3B Trouble

• X1/5 DC LC 3B Trouble

REVISION NO.: PROCEDURE TITLE:

PAGE:

LOSS OF DC BUS 3D01 AND 3DO1A (3A)

PROCEDURE NO.: 5 of 31 3-ONOP-003.4 TURKEY POINT UNIT 3 2.1 Indications (continued)

NOTE IF backup protection is provided from 3D23, then the automatic generator trip and automatic fast transfer from Auxiliary to Startup Transformers should still be functional.

• Loss of power to Primary Generator Lockout Relay,

• Loss of power to 4KV Bus 3A and 3B switchgear synchronism check relays, preventing manual electrical closure of 4KV Bus 3A and 3B Feeder Breakers from the Auxiliary or Startup Transformers

• Loss of 3A Bus Load Sequencer

• oss of DC Power to 3A EDG

• Loss of Train 1 feedwater isolation capability of SIG Feedwater Bypass valves.

• Loss of Train 1 AFW

• Loss of Train A Safeguards Equipment

• Loss of 480V Load Centers 3A and 3C Breaker control power

• Loss of Unit 3 Startup Transformer annunciation and protection

• Loss of CVCS Blender (FCV-3-1 1 3B fails closed) 2.2 Alarms

• X113 Sequencer 3A Trouble

• X1/1 DC LC 3A Trouble

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 073 A2.01 Importance Rating 2.5 Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, contro l, or mitigate the consequences of those malfunctions or operations: Erratic or failed power supply Proposed Question: RO Question # 25 Given the following conditions:

• Unit 3 and 4 are at 100% power

• Liquid release is in progress lAW 0-NOP-061.11C, Controlled Liquid Releas From e Monitor Tank A.

Subsequently:

• Annunciator H 1/4, PRMS HI RADIATION, alarms.

• Annunciator H116, PRMS CHANNEL FAILURE, alarms.

• Unit 3 RCO reports no indications are available for Process Radiation Monito r R-18 due to no power.

Which ONE of the following describes the required response?

A. RCV-18 is closed. Enter TS Action Statement.

B. RCV-18 is closed. Enter 3-ONOP-067, Radioactive Effluent Release.

C. RCV-18 is NOT closed. Direct Chemistry to sample effluent in accordance with ODCM.

D. RCV-18 is NOT closed. Direct Chemistry to sample effluent in accordance with 3-ARP-097.CR.H, H116.

Proposed Answer: B Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. Incorrect. Plausible to believe since Technical Specifications contain LCO for Radiation Monitoring and the ARP references TS to refer to. However R-1 8 is not in Tech Specs.

B. Correct. 3-ARP-097.CR.H, Control Room Response, Panel Hl/4 directs the operator to 3-ONOP-067 for expected automatic actions. Since this is a High Radiation alarm as well as a channel failure, RCV-018 would have closed. Panel H1/6, directs the RCO to check that the failure is valid, and if R-18 failed, to terminate the liquid release if one is in progress.

C. Incorrect. Plausible to believe since 3-ARP-097.CR.H Control Room Response, Panel H1/6 checks for Channel failure. Panel H 1/6, however contains no guidance for sampling. 3-ONOP-067 provides guidance to direct Chemistry to sample, however, the RNO would not be entered since the release will have been terminated earlier.

D. Incorrect. Plausible to believe since terminating the release is the correct action and if the valve is not closed, to sample the liquid. Panel H 1/6, directs the RCO to check that the failure is valid, and if R-18 failed, to terminate the liquid release if one is in progress.

Panel H1/6, however contains no guidance for sampling. 3-ONOP-067 provides guidance to direct Chemistry to sample, however, the RNO would not be entered since the release will have been terminated earlier.

Technical Reference(s): 3-ARP-097.CR.H Rev 6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902242 Obj 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

6 7 CONTROL ROOM RESPONSE - PANEL H PROCEDURE NO.:

WINDOW:

3-ARP-097.CR.H TURKEY POINT UNIT 3 114 (Page 1 of 1)

CAUSES: 1. High radiation in one of systems monitored by PRMS

2. PRMS system component failure H114 PRMS HI RADIATION DEVICE: SETPOINT: LOCATION:

• R-1 1 Variable with each PRMS channel N/A

• R-12

• R-14

• R-15

• R-17A

• R-17B

• R-18

• R-19

• R-20 ALARM CONFIRMATION

1. CHECK the following:

• Countrate meter on each PRMS drawer in Rack QR-66

• Alarm indicators on each drawer in Rack QR-66 OPERATOR ACTIONS

1. IF alarm is on R-11, R-12, R-14, R-17AIB, R-18, or R-20, THEN REFER TO 3-ONOP-067, Radioactive Effluent Release for expected automatic actions.

2. IF alarm is on R-15 or R-19, THEN REFER TO 3-ONOP-071.2, Steam Generator Tube Leakage for expected automatic actions.

3. IF alarm is on R-14, R-17A, R-17B, R-18, or R-19, THEN CHECK alarm valid as follows:

A. CHECK FAIL/TEST light NOT LIT.

B. PUSH FAIL/TEST light (meter reading of 288 or 289K)

C. PUSH SOURCE CHECK light (should get meter increase).

D. PUSH HIGH ALARM light to determine if meter level is above high alarm setpoint.

4. ENSURE required automatic actions.

5. IF alarm ison R-11, R-12, R-14, R-17AIB, R-18, OR R-20, THEN REFER TO 3-ONOP-067, Radioactive Effluent Release.

6. IF alarm is on R-15 OR R-19, THEN REFER TO 3-ONOP-071.2, Steam Generator Tube Leakage.

7. REFER TO TS 3.3.3, 3.4.6, and 3.9.13 for additional required actions.

REFERENCES:

Tech Spec Sections 3.3.3, 3,4.6, and 3.9.13 PC/M 07-055, R-1 5 Steam Jet Air Ejector Monitor Replacement

REVISION NO.: PROCEDURE TITLE: PAGE:

6 9 CONTROL ROOM RESPONSE - PANEL H PROCEDURE NO.:

WINDOW:

3-ARP-097.CR.H TURKEY POINT UNIT 3 1/6 (Page 1 of 1)

CAUSES: 1. Loss of detector counts for three minutes (30 seconds for R20)

2. Loss of power to PRMS drawer 116

3. RANGE switch NOT in normal position(except R15 and R20)

4. Loss of power to R20 Local Ratemeter in Pipe and Valve Room.

CHANNEL FAILURE DEVICE: SETPOINT: LOCATION:

• RANGE switch In other than normal position PRMS drawer (except R15 and R20)

OR N/A PRMS drawer

• FAIL relay K-3 (except R15 and R20)

ALARM CONFIRMATION

1. CHECK the following:,

• Fail lamp on PRMS drawer or on R20 Local Ratemeter in Pipe and Valve Room

• Loss of power to PRMS channel or to R20 Local Ratemeter in Pipe and Valve Room

• Except for Ri 5, loss of detector counts for three minutes (30 seconds for R20)

• For Ri5, green OPERATE LED is OFF OPERATOR ACTIONS

1. IF NOT under test, THEN DETERMINE which channel is alarming and RETURN switches or power alignment to normal.

2. CHECK for channel failure.

3. IF R-14 fails, THEN STOP gas decay release.

4. IF R-18 fails, THEN STOP liquid release.

5. IF R-19 fails, THEN SECURE SIG blowdown.

6. REFER TO Tech Spec 3/4.3.3, 3/4.4.6, and 3/4.9.13.

REFERENCES:

Tech Spec Sections 3/4.3.3, 3/4.4.6 and 3/4.9.13

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMAT!ON Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 076 2.4.31 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of annunciator alarms, indications, or response procedures.

Proposed Question: RO Question # 26 Given the following conditions:

Units 3 and 4 are at 100% power with a normal electrical alignment.

• 3A and 3B ICW Pumps are running.

Subsequently:

• Annunciator 13/4, TRAVELING SCREEN GENERAL TROUBLE, alarms

• Annunciator 14/2, ICWP A/B/C TRIP, alarms

• Annunciator 14/4, ICW HEADER A/B LO PRESS, alarms

• 3A1 and 3A2 Traveling Screen Wash Pumps are ON.

• 3A ICW Pump RED light is lit.

Which ONE of the following identifies (1) the next required RCO response in accordance with 3-ARP-097.CR.I and (2) the required procedure to address the given conditions?

A. (1) Start 3B ICW Pump (2) 3-ONOP-019, Intake Cooling Water Malfunction B. (1) Start 3B ICW Pump (2) 3-ONOP-Ol 1, Screen Wash System/Intake Malfunction.

C. (1) Start 3C ICW Pump (2) 3-ONOP-019, Intake Cooling Water Malfunction PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. (1) Start 30 ICW Pump (2) 3-ONOP-Ol 1 Screen Wash System/Intake Malfunction Proposed Answer: C Explanation (Optional):

A. Incorrect. 1 st part wrong. 2 part right. Plausible to believe that since 3B ICW was running and now is not, that the correct action is to restart the previously running pump.

This is the procedural guidance in case of loss of an RHR pump, to attempt to restart the previously running pump. However, 3-ARP-097.CR.l, Panel 14/2 states to start any standby pump.

B. Incorrect. 1st part wrong. 2 part wrong. Plausible to believe that since 3B lOW was running and now is not, that the correct action is to restart the previously running pump.

This is the procedural guidance in case of loss of an RHR pump, to attempt restart to the previously running pump. However, 3-ARP-097.CR.l, Panel 14/2 states to start any standby pump. Also plausible that the required procedure would be 3-ONOP-Ol 1, Screen Wash System/Intake Malfunction, since there is indication of issues with screen wash. However, the loss of an 10W pump is more of an immediate concern.

C. Correct. 3-ARP-097.CR.l, Panel 14/2 states to start any standby pump. Then the APR provide guidance to enter Enter 3-ONOP-019, Intake Cooling Water Malfunction,.

D. Incorrect. 1 st part right. 2 part wrong. Plausible that the required procedure would be 3-ONOP-Ol 1, Screen Wash System/Intake Malfunction, since there is indication of issues with screen wash. However, the loss of an ICW pump is more of an immediate concern.

Technical Reference(s): 3-ARP-097.CR.l Rev 11 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902154 Obj. 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE:

PAGE:

11 CONTROL ROOM RESPONSE 25 PROCEDURE NO.: - PANEL I WINDOW:

3-ARP-097.CR.l TURKEY POINT UNIT 3 4/2 (Page 1 of 1)

CAUSES: 1. Motorfailure

2. Pump failure 1412

3. Excessive AP across traveling screens ICWPAIBIC TRIP DEVICE: SETPOINT: LOCATION:

• 3AA19-174-RELAY (3A ICWP) 105 amps N/A TOO

• 3AB 17-174-RELAY (38 ICWP) 105 amps TDO

• 3AD05-1 74-RELAY (3C ICWP) 105 amps TDO ALARM CONFIRMATION CHECK for any 1GW pump mismatched breaker indication.

OPERATOR ACTIONS

1. START any standby ICW pump.

2. DISPATCH an operator to check for cause of lOW pump tripping.

3. DISPATCH an operator to tripped lOW pump 4KV Bus breaker to identif y any breaker targets.

4. IF an overload condition develops on pump started above OR more complicated problems are indicated with lOW System, THEN GO TO 3-ONO P-019, Intake Cooling Water Malfunction.

5. REFER TO TS 3.7.3 for additional required actions.

6. ADVISE Electrical or Mechanical Maintenance of problem.

REFERENCES:

1. Tech Spec 3/4.7.3

2. 5613-E-27, Sh 2A and 2A1, ICW Pump 3A/3AA19

3. 5613-E-27, Sh2Band2Bl, ICWPump3B/3AB17

4. 5613-E-27, Sh 2C and 2C1, ICW Pump 3C/3AD05

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 078 A4.01 Importance Rating 3.1 Ability to manually operate and/or monitor in the control room: Pressure gauges Proposed Question: RO Question # 27 Given the following conditions:

• Units 3 and 4 are at 100% power.

• The 4CM Instrument Air compressor is out for maintenance.

• 3CM is running in LEAD, 3CD is in LAG, 4CD is in STANDBY-LAG.

• The Instrument Air systems are cross-tied.

Subsequently:

• Unit 4 Annunciators 1611, INSTR AIR SYSTEM HI TEMP/LO PRESS, alarms.

Which ONE of the following identifies the minimum pressure on P1-4-1444, INST AIR PRESS, if it cannot be maintained, when a Unit 4 Reactor Trip is required per 0-ONOP-Ol 3, Loss of Instrument Air?

A. 60 psig B. 65 psig C. 80 psig D. 90 psig Proposed Answer: B Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION A. Incorrect. Plausible to remember that 0-ONOP-013 FOP discusses major impacts for a dual unit loss of Instrument Air of less than 60 psig. However, 0-ONOP-013 FOP requires a Reactor Trip if IA pressure cannot be maintained >=65 psig.

B. Correct. 0-ONOP-013 FOP requires a Reactor Trip if IA pressure cannot be maintained

>=65 psig.

C. Incorrect. Plausible to remember the closure setpoint of 80 psig for CV-3/4-1 605, Unit 3/4 Instrument Air Crosstie Isolation Control Valves. However, 0-ONOP-013 FOP requires a Reactor Trip if IA pressure cannot be maintained >65 psig.

D. Incorrect. Plausible to remember that 0-ONOP-Ol 3 FOP discusses major impacts for a single unit loss of Instrument Air of less than 90 psig. However, 0-ONOP-013 FOP requires a Reactor Trip if IA pressure cannot be maintained >65 psig.

Technical Reference(s): 0-ONOP-013, Rev 7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902145, Obj. 8 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New . X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Corn rnents:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

PROCEDURE NO.:

LOSSOFINSTRUMENTAIR I FOLDOUT 0-ONOP-013 TURKEY POINT PLANT FOLDOUT PAGE For Procedure O-ONOP-013 UNIT TRIP CRITERIA IF Instrument Air System can NOT be maintained greater than 65 psig, requires isolation of the Auxiliary Building, Containment Building, or Turbine Building Air Headers, OR results in the inability to maintain SIG levels on either unit, AND Instrument Air Compressors are unable to restore pressure, THEN:

• TRIP the affected unit AND ENTER the appropriate EOP

• CONTINUE to restore Instrument Air to the unit(s).

MAJOR COMPONENT IMPACTS

1. A single unit loss of Instrument Air of less than 90 psig may result in the loss of function to the following valves if the Nitrogen Backup Systems can NOT be maintained to the components:

a. Pressurizer PORVs

b. Steam Dumps to Atmosphere

c. Train 2 of AFW (same unit), operate FCVs in MANUAL to conserve Nitrogen.

d. Train 1 of AFW (opposite unit), operate FCVs in MANUAL to conserve Nitrogen.

e. Unit 3 EDG Fuel Oil Transfer capability

2. A single unit loss of Instrument Air of less than 65 psig results in the loss or partial loss of function depending on the spring bench setting of the following valve(s):

a. Letdown Isolation Valves

b. Feedwater Reg Valves

c. Feedwater Bypass Valves

d. Unit 3A and 3B Emergency Diesel Generator Day Tank Level Control Valve

3. A dual unit loss of Instrument Air of less than 60 psig results in the loss of additional functions:

a. Charging Pump speed control fails to high speed

b. Primary Water Makeup capability

c. Loss of automatic transfer of Charging Pump Suction to the RWST

d. Unit 3 ECC valves fail OPEN after a delay of 20 or more minutes.

e. Unit 4 ECC valves fail OPEN

f. Unit 3 inability to control RCS cooldown using HCV-3-758 and FCV-3-605 and may require stopping the RHR Pump to stop a cooldown, if in progress when air was lost.

g. Unit 4 inability to control RCS cooldown using HCV-4-758 and FCV-4-605 and may require stopping the RHR Pump to stop a cooldown, if in progress when air was lost.

4. 0-ONOP-105, Control Room Evacuation, provides useful guidance in determining components required for a plant cooldown. This procedure should be referenced if a cooldown must be performed without the availability of Instrument Air. Plant Management Staff and Engineering should be involved in any decision to attempt a cooldown without Instrument Air.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 1 KIA# 103 A2.05 Importance Rating 2.9 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations Emergency containment entry Proposed Question: RO Question # 28 Given the following conditions:

• Unit3isinMODEl.

• Containment is exited after an emergent Containment Entry to investigate RCS leakage.

• Containment Airlock Doors (Personnel Access) are able to be opened at the same time due to broken linkage.

• The inside airlock door is latched closed with some leakage, but within LLRT limits.

• The outside airlock door can NOT be latched closed.

NOTES

• TS LCO 3.6.1.2 is Containment Leakage.

• TS LCO 3.6.1.3 is Containment Air Locks.

Which ONE of the following describes (1) the inoperable component and (2) the Technical Specification LCO that is NOT met for Containment?

A. (1) The Containment Inside Airlock Door (2) LCO 3.6.1.2 B. (1) The Containment Inside Airlock Door (2) LCO 3.6.1.3 C. (1) The Containment Outside Airlock Door (2) LCO 3.6.1.2 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. (1) The Containment Outside Airlock Door (2) LCO 3.6.1.3 Proposed Answer: D Explanation (Optional):

A. Incorrect: 1st part wrong. 2 part wrong. Plausible to believe that the inside air lock door is inoperable since there is leakage past the seal and that TS 3.6.1.2 has not been met. However, leak rate on the inside door meets the measured overall integrated containment leakage rate and therefore LCO 3.6.1.2 is met.

B. Incorrect. 1 st part wrong. 2,d part right. Plausible to believe that the inside air lock door is inoperable since there is leakage past the seal. However, the inside door is closed and the leak rate on the inside door meets the measured overall integrated containment leakage rate.

C. Incorrect: 1 st part right. 2 part wrong. Plausible to believe that with one door unable to latch, the leak rate TS 3.6.1.2 is NOT met. The personnel air lock doors are only a part of the requirement for containment leakage of TS 3.6.1.2. However, leak rate on the inside door meets the measured overall integrated containment leakage rate and therefore LCO 3.6.1.2 is met.

D. Correct: TS 3.6.1.3 provides for actions if the [CO is NOT met and specifies each door as being OPERABLE or inoperable. TS 3.0.1 requires compliance with the [CO or if not met, the associated action statement shall be met. Since the outside door cannot be closed, it is considered inoperable, but with the inside door closed, TS 3.6.1.3.a.1 action is met.

Technical Reference(s): PTN Technical Specifications .

Rev 293 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902126 Obj. 9 (As available)

Question Source: Bank # 88962 Modified Bank # (Note changes or attach parent)

New PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question History: Last NRC Exam: 2009 Millstone 2 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited in accordance with the Containment Leakag Rate Testing rogram.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the measured overall integrated containment leakage rate exceeding 1.0 La within one hour, initiate action to I be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Restore the I overall integrated leakage rate to less than 0.75 La and the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 La prior to increasing the Reactor Coolant System temperature above 200°F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the required test schedule and shall be determined in conformance with the criteria specified in the Containment Leakage Rate Testing Program.

TURKEY POINT UNITS 3 & 4 3/4 6-2 AMENDMENT NOS. 192 AND 186

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, or during the performance of containment air lock surveillance and/or testing requirements, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate in accoce with the Containment Leakge Rate Testing Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the OPERABLE air lock door closed;

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days;

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TURKEY POINT UNITS 3 & 4 3/4 6-3 AMENDMENT NOS. 260 AND 255

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. Following each closing, at the frequency specified in the Containment Leakage Rate Testing Program, by verifying that the seals have not been damaged and have seated properly by vacuum testing the volume between the door seals in accordance with approved plant procedures.

b. By conducting overall air lock leakage tests in accordance with the Containment Leakage Rate Testing Program.

c. At least once per 24 months by verifying that only one door in each air lock can be opened at a time.

TURKEY POINT UNITS 3 & 4 3/4 6-4 AMENDMENT NOS. 260 AND 255

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 001 K6.08 Importance Rating 2.9 Knowledge of the effect of a loss or malfunction on the following CRDS components:

Purpose and position switch of alarm for high flux at shutdown Proposed Question: RO Question # 29 Given the following conditions:

• Unit 3 is in a refueling outage.

• The Upper Internals Assembly is being lifted out of the Reactor Vessel to set into the lower cavity.

Subsequently:

• B411, SOURCE RANGE HI FLUX AT SHUTDOWN, alarms.

Which ONE of the following correctly completes the statements below?

The Source Range high flux alarm (1) initiate a containment evacuation alarm.

If flux continues to increase, then in accordance with B411 (2)

A. (1) WILL (2) initiate Emergency Boration B. (1) WILL (2) adjust the High Flux at Shutdown alarm setpoint C. (1)WILLNOT (2) initiate Emergency Boration D. (1)WILLNOT (2) adjust the High Flux at Shutdown alarm setpoint PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMA11ON Proposed Answer: A Explanation (Optional):

A. Correct. The purpose of the Source Range Hi Flux at Shutdown Alarm is to enhance shutdown reactivity monitoring and to automatically commence a Containment evacuation. The Control Room announces the Containment evacuation in accordance with 3-ARP-097.CR.B 4/1. An increase in source range counts is an indication that the RCCA(s) are still latched. During the Upper Internals lift, it will cause positive reactivity addition and increase source range counts.

B. 1st part right 2nd part wrong. Plausible to remember adjusting the High Flux at shutdown setpoint in accordance with 3-OSP-059.6, High Flux at Shutdown, since this is done in 3-GO P-503 after entering MODE 5.

C. 1st part wrong. 2nd part right. Plausible to believe since there is an action in 3-GOP-301 to Block the High Flux at Shutdown Alarms prior to withdrawing Shutdown banks and student may believe that this is still blocked.

D. 1st part wrong. 2nd part wrong. Plausible to believe since there is an action in 3-GOP-301 to Block the High Flux at Shutdown Alarms prior to withdrawing Shutdown banks and student may believe that this is still blocked. Plausible to remember adjusting the High Flux at shutdown setpoint in accordance with 3-OSP-059.6, High Flux at Shutdown, since this is done in 3-GOP-503 after entering MODE 5.

Technical Reference(s): 3-ARP-097.CR.B4/1 RiO .

(Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO 6918104-Obj.7 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-]. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE; PAGE:

10 CONTROL ROOM RESPONSE 22

- PANEL B PROCEDURE NO.:

WINDOW:

3-ARP-097.CR.B TURKEY POINT UNIT 3 41 (Page 1 of 1)

CAUSES: 1. Neutron flux in reactor increased to alarm setpoint

2. SR NI malfunction B411 SOURCE RANGE HI FLUX AT SHUTDOWN DEVICE: SETPOINT: LOCATION:

Source Range detectors: Half decade above count rate at shutdown. Variable N/A

• N.31 and resets at each shutdown

• N-32 NOTE if the annunciator is in alarm from a spike on either N-31 or N-32, there is a two second time delay in the Source Range High Flux at Shutdown circuitry to prevent the actuation of the Containment Evacuation Alarm.

ALARM CONFIRMATION

1. CHECK count trend on NI level recorder on console.

2. CHECK both source range indicators for increase since shutdown.

OPERATOR ACTIONS

1. IF a startup is NOT in progress, THEN ENSURE actuation of Containment Evacuation alarm.

2. ANNOUNCE containment evacuation over the Page System.

3. IF a startup is in progress, THEN BLOCK the Containment Evacuation alarm.

4. IF count rate has changed due to a planned change in plant condition such as, heatup, boron concentration change, etc., THEN ADJUST the High Flux at Shutdown alarm setpoint using 3-OSP-059.6, High Flux at Shutdown, to maintain a one-half decade above indicated source range count rate.

5. IF rods are withdrawn AND count rates have changed due to changing plant conditions that were NOT planned, THEN TRIP the reactor.

6 IF flux continues to increase THEN BORATE using 3-ONOP-046 I Emergency Boration.

7. INVESTIGATE for possible dilution/cooldown of RCS.

8. IF SR NI malfunction, THEN PERFORM 3-ONOP-059.5, Source Range Nuclear Instrumentation Malfunction.

REFERENCES:

Tech Spec Sections 3.3.1 and 3.9.2

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 011 A4.05 Importance Rating 3.2 Ability to manually operate and/or monitor in the control room: Letdown flow controller Proposed Question: RO Question # 30 Given the following conditions:

• Unit4isinMODE5.

• Letdown is in service.

• The RCS is solid.

• A plant cooldown and depressurization is in progress.

Which ONE of the following describes the action to lower RCS pressure?

A. Raise the setpoint of HCV-4-142, RHR LID to CVCS.

B. Raise the setpoint of PCV-4-145, Low Pressure LTDN Controller.

C. Lower the setpoint of HCV-4-142, RHR L/D to CVCS.

D. Lower the setpoint of PCV-4-145, Low Pressure LTDN Controller.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible to believe that rasing the setpoint from RHR L/D to CVCS will increase flow from the RCS and thereby lower RCS pressure. However, HCV-4-142 is not throttled and fully open per 4-GOP-305 while PCV-4-145 will adjust to maintain set pressure.

B. Incorrect. Plausible to believe that raising the setpoint will open PCV-4-145, which would decrease RCS pressure. However, unlike the Excess LTDN Flow Controller, PCV-4-137, which opens the valve when the setpoint is raised, raising the setpoint on PCV-4-145 closes the valve which increases pressure.

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Incorrect. Plausible to believe that lowering the setpoint from RHR L/D to CVCS will decrease flow and thereby lower the backpressure on PCV4-145. However, HCV 142 not throttled and fully open per 4-GOP-305 while PCV-4-145 will adjust to maintain set pressure.

D. Correct. Lowering the setpoint of PCV-4-145 will open the valve increasing flow and lowering RCS pressure. PCV-4-145 will modulate and adjust to maintain the new set pressure.

Technical Reference(s): 4-GOP-305 Rev 13 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP69O2113Obj5B (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

INIT fl QA RECORD PAGE 5.17.2.6.a (Contd)

(3) WHEN the required boron concentration for the shutdown has been verified, THEN stop Reactor Coolant Pumps using 4-NOP-041 .0 IA, 4A Reactor Coolant Pump Operations, 4-NOP-041.01B, 4B Reactor Coolant Pump Operations, and 4-NOP-041.OIC, 4C Reactor Coolant Pump Operations, as appropriate.

(4) Lower the setpoint of PCV-4-145 to reduce RCS pressure to th range of 300 to 324 psig. [Commitment Step 2.3.3J

b. Open the Aux Spray Isolation, CV-4-3 11.

c. Close the Pressurizer spray valves:

(I) PRZ Spray Loop B, PCV-4-455B (2) PRZ Spray Loop C, PCV-4-455A

d. Close the charging line isolation valves:

(1) Loop A Charging Isolation, CV-4-3l0A (2) Loop C Charging Isolation, CV-4-3 I OB CAUTIONS V

• The Pressurizer cooldown rate limit of 190°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is 10°F below the Technical Specification limit of 200°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

• Do NOT re-align auxiliary spray prior to de-energizing Pressurizer heaters.

e. Control Pressurizer cooldown at a rate NOT to exceed 190°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period by adjusting charging pump speed.

f. Notify Primary Chemistry when PZR temperature is less than 300°F.

W97:JWB/cls/cls/ab

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 017 K5.02 Importance Rating 3.7 Knowledge of the operational implications of the following concepts as they apply to the ITM system: Saturation and subcooling of water Proposed Question: RO Question # 31 Given the following conditions:

• Unit 3 trips due to Loss of Offsite Power.

• 3A and 3B EDGs energize their respective 4KV buses.

• 3-EOP-ES-0.2, Natural Circulation Cooldown, is being implemented.

Subsequently during the cooldown:

• Average CET temperature is 550°F and rising.

• RCS pressure is 545 psig and rapidly lowering.

• Pressurizer level is 11% and lowering.

Which ONE of the following actions is required in accordance with 3-EOP-ES-0.2?

A. Open PORVs to restore decay heat removal.

B. Initiate SI and Phase A to re-establish plant control.

C. Start an extra Charging Pump to raise Pressurizer level.

D. Close Pressurizer Spray Valves to raise RCS pressure.

Proposed Answer: B Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. Incorrect. Plausible to believe since 3-EOP-ES-0.2 provides guidance to depressurize RCS using a PORV and this strategy is used in the EOP network for re-establishing decay heat removal. However, RCS subcooling is less than 19°F, which requires SI and Phase A.

B. Correct. RCS is superheated by 71°F. 3-EOP-ES-0.2 Foldout Page SI and Phase A actuation criteria is <19°F subcooled based on Core Exit TCs.

C. Incorrect. Plausible to believe since pressure is lowering, starting a Charging pump would raise Pressurizer level, compress the steam space, raise RCS pressure and improve subcooling. However, 3-EOP-ES-0.2 Foldout Page SI and Phase A actuation criteria is <19°F subcooled based on Core Exit TCs and criteria has been met.

D. Incorrect. Plausible to believe that natural circulation is providing a driving head for pressurized spray and closing the spray valves would raise RCS pressure. However, 3-EOP-ES-0.2 Foldout Page SI and Phase A actuation criteria is <19°F subcooled based on Core Exit TCs and criteria has been met.

Technical Reference(s): 3-EOP-ES-0.2 Rev 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: Yes Learning Objective: LP6902324Obj.6 8 10 (As available)

Question Source: Bank #

Modified Bank # 88323 (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 Ginna Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KJA # 017 K5.02 Importance Rating 3.7 Knowledge of the operational implications of the following concepts as they apply to the ITM system: Saturation and subcooling of water Proposed Question: RO Question # 31 -J Plant conditions occurred as follows:

• The unit tripped due to a Loss of Offsite power.

Both DIGs started and energized the required loads.

All equipmellt responded as designed.

ES-0.2. Natural Circulation Cooldown is being implemented SuhseqLlently. during the cooldown. the following conditions exist:

• Containment pressure is 5.5 psig and slowly rising.

z

• Average core exit thermocouple (CET) temperature is 420°F and stable.

• RCS pressure is 500 psig and stable.

• Pressurizer level is 20% and stable.

Given the above conditions and provided references:

I. Per Fig- 1.0 Figure Miii Subcooling, what is the status of the RCS?

a tid

2. Based on the status of the RCS. which action is required per ES-0.2, Natural Circulation Coo ldown Foldout page?

A. I. The RCS is subcooled.

2. DO NOT initiate SI and CI B. 1. The RCS is inadequately subcooled.

2. Initiate SI and CI.

C. I. The RCS is inadequately suhcooled.

2. DO NOT initiate SI and CI.

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 D. I The RCS is suhcooled.

2. Initiate SI and CI Proposed Answer: B Explanation (Optional):

A.

B.

I Ihis would he correct iladverse containment conditions did not exist.

2. Correct action if adverse containment conditions did not exist I. RCS is inadequately suhcooled by 40°F.

-J

2. Action required per LS0.2. Natural Circulation Cooldow n Foldout page z

C. I. RCS is inadequately suhcooled by 40°F.

2. SI and Cl are required D. I This would he correct if adverse containment conditions did not exist.

2. Action required per ES0.2. Natural Circulation Cooldown FoldoLit page Technical Reference(s): Fig 1.0 Figure Mm Subcooling ES-0.2, Natural Circulation (Attach if not previously provided)

Proposed References to be provided to applicants during examination: Yes Given a set of plant and equipment conditions, evaluate the conditions to determine the applicable procedure, and from the procedure Learning Objective: determine the appropriate (As available)

EXPECTED ACTIONS or RESPONSE NOT OBTAINED instructions to implement in ES-0.2, Natural Circulation Cooldown Question Source: Bank # 88323 Modified Bank # (Note changes or attach parent)

New FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Question History: LastNRCExam: 2008 Ginna Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 Comments: -J z

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

2 NATURAL CIRCULATION COOLDOWN FOLDOUT PROCEDURE NO.:

3-EOP-ES-O.2 TURKEY POINT UNIT 3 FOLDOUT PAGE For Procedure 3-EOP-ES-O.2

1. SI ACTUATION CRITERIA

a. H 3-EOP-ES-0.2, NATURAL CIRCULATION COOLDOWN, was entered with Unit 3 in Mode 1, 2, or 3 reater than 1000 psig), AND either condition listed below occurs, THEN actuate S ate Containment Isolation Phase A, and go to 3-EQ P-E-0, REACTOR TRIP OR SAFETY INJECTION, Step 1:

RCS Subcooling basédon Core Edt TCs,,LESS THAN 19°I, OR PRZ level CAN NOT BE MAINTAINED GREATER THAN 7%

b. H 3-EOP-ES-0.2, NATURAL CIRCULATION COOLDOWN, was entered with Unit 3 in Mode 3 (less than 1000 psig), or Mode 4, AND either condition listed below occurs, THEN go to 3-ONOP-041.7, SHUTDOWN LOCA [MODE 3 (LESS THAN 1000 PSIG)

OR MODE 4], Step 1:

RCS Subcooling based on Core Exit TCs LESS THAN 19°F OR PRZ level CAN NOT BE MAINTAINED GREATER THAN 7%

2. CST MAKEUP WATER CRITERIA IF CST level decreases to less than 12%, THEN add makeup to CST using 3-NOP-018.01, CONDENSATE STORAGE TANK (CST).

3. AFW SYSTEM OPERATION CRITERIA

a. IF two AFW Pumps are operating on a single train, THEN one of the pumps shall be shut down within one hour of the initial start signal.

b. IF two AFW Trains are operating AND one of the AFW Pumps has been operating at low flow of 80 gpm or less for one hour, THEN that AFW Pump shall be shut down.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 015 K2.01 Importance Rating 3.3 Knowledge of bus power supplies to the following: NIS channels, components, and interconnections Proposed Question: RD Question # 32 Given the following conditions:

• A Unit 3 Reactor Startup is in progress.

• Power level is 1 x 10-8 amps.

Subsequently:

• 120V Vital Instrument Panel 3P07 is lost.

Which ONE of the following completes the following statement?

A Reactor Trips occurs due to the loss of (1) , and as power lowers below bE-b amps (2) will energize.

A. (1) IR Nl-35 (2) SR NI-31 B. (1)IRNI-35 (2) SR NI-32 C. (1)IRNI-36 (2) SR Nl-31 D. (1)IRNI-36 (2) SR NI-32 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible to believe if examinee incorrectly remembers loads on 3 P07. Loss of 3P07 will result in a loss of Channel II, which will de-energize SR N-32 and IR N-36.

N-35 still has power. N-36 bistable causes reactor trip when it de-energizes.

B. Incorrect. Plausible to believe if examinee incorrectly remembers loads on 3P07. N-35 still has power. N-36 bistable causes reactor trip when it de-energizes.

C. Correct. Loss of 3P07 will result in a loss of Channel II, which will de-energize SR N-32 and IR N-36. N-36 bistable causes reactor trip when it de-energizes. SR N-31 will still have power and re-energize.

D. Incorrect. First part is correct but second part is incorrect because loss of 3P07 will result in a loss of Channel II, which will de-energize SR N-32. Also possible to confuse P-6 and whether or not EITHER SR instruments will automatically energize post-trip.

Technical Reference(s): 3-ONOP-003.7 Rev 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902104 Obj. 5 (As available)

Question Source: Bank # 86018 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 McGuire Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

ENCLOSURE I (Page2of3)

CONTROL ROOM FUNCTIONS AND INDICATIONS LOST ON FAILURE OF VITAL INSTRUMENT PANEL 3P07 INDICATIONS (Contd)

Tl-3-422B B Loop Ovpwr AT Tl-3-422A B Loop AT Tl-3-422C B Loop Overtemp AT Tl-3-422D B Loop Temp Avg P1-3-456 Pressurizer Pressure OH II Ll-3-460 Pzr Level ProtlOont OH II Fl-3-415 A Loop ROS Flow F 1-3-425 B Loop RCS Flow F 435 O Loop ROS Flow TR-3-41 0 Wide Range T-Oold Recorder (Red, Blue, Green)

TR-3-41 3 Wide Range T-Hot Recorder P1-3-474 A Stm Gen Pressure OH II P1-3-484 B Stm Gen Pressure OH II P1-3-494 O Stm Gen Pressure OH II LI-3-475 A Stm Gen Level OH Ii Ll-3-485 B Stm Gen Level OH II LI-3-495 O Stm Gen Level OH II LR-3-477 SG B Wide Range Level Recorder (Blue)

P1-3-468 Stm Header Pressure HCV-3-1 21 Charging flow to Regen Hx (Valve Goes Full Open)

FR-3-1 13 Blender Flow Recorder 3001 Boric Acid Totalizer 3C0 1 Primary Water Totalizer N-3-32 Source Range Counts CH II N-3-32 Source Range Start-up Rate CR II N-3-36 Inter Range Current CH II N-3-36 Inter Range Start-up Rate CH II N-3-42 Power Range N-3-42 Axial Flux Difference NR-3-45B Nuclear Instrumentation (NIS) Recorder FR-3-488 B SG Feedwater Flow Recorder FCV-3-488 3B SG Feedwater Control Valve (On Backup Controller)

FCV-3-489 3B SG Feedwater Bypass Valve (Fails Closed)

HIC-3-1401A Aux Feedwater Train I Controller H 10-3-1 457A Aux Feedwater Train I Controller H 10-3-1 458A Aux Feedwater Train I Controller 3002 OST Low Level Amber Light LI-3-6384A OST Level NIS RackChII (N-32 N-36 N-42)

RAR-3-631 1A Oont High Rad Monitor and % H2 Recorder 3005 Channel I Status Lights 3C05 Channel II Status Lights Fl-3-943 SI Cold Leg Flow A!Q7IrIq/I!fm

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RD SRO Tier# 2 Group# 2 KIA# 029 A2.01 Importance Rating 2.9 Ability to (a) predict the impacts of the following mal- functions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Maintenance or other activity taking place inside containment Proposed Question: RD Question # 33 Given the following conditions:

• Unit4isinMODE6.

• The crew has commenced rod unlatching.

Subsequently:

• Annunciator 1416, CNTMT SUMP HI LEVEL, alarms.

• Reactor Cavity water level is 56 feet and lowering.

• Bubbles are rising around the outside of the Reactor Vessel.

• R-1 1, Containment Air Particulate Monitor, is in alarm Which ONE of the following completes the following statements based on the given conditions?

Containment Purge Isolation valves (1) required to be closed in accordance with 4-ONOP-033.2, Refueling Cavity Seal Failure.

Normal Containment Coolers are (2) secured.

A. (1)are (2) automatically B. (1) are NOT (2) manually C. (1)are (2) manually PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. (1) are NOT (2) automatically Proposed Answer: C Explanation (Optional):

A. Incorrect. 1st part right. 2 part wrong. Plausible to believe that since a Containment Phase A Isolation is required to be initiated, the Normal Containment Coolers are automatically secured. However, they have no automatic trip function.

B. Incorrect. 1st part wrong. 2rd part right. Plausible to believe that the Containment Purge Isolation valves do not need to be closed since they are normally closed and de-energized. However, 4-ONOP-033.2, Attachment 2 states to check that they are closed.

C. Correct. 4-ONOP-033.2, Refueling Cavity Seal Failure, directs the crew to evacuate all personnel from containment, initiate containment isolation Phase A, as well as verify that purge is secured (pumps and valves), and to Trip all Normal Containment Coolers.

D. Incorrect. 1 st part wrong. 2nd part wrong. Plausible to believe that the Containment Purge Isolation valves do not need to be closed since they are normally closed and de-energized. However, 4-ONOP-033.2, Attachment 2 states to check that they are closed. Also plausible to believe that since a Containment Phase A Isolation is required to be initiated, the the Normal Containment Coolers are automatically secured.

However, they have no automatic trip function.

Technical Reference(s): 4-ONOP-033.2, Rev 4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Learning Objective: LP 6902607 Obj 4 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO,: PROCEDURE TITLE: PAGE:

REFUELING CAVITY SEAL FAILURE 9 of 30 PROCEDURE NO.:

4-ONOP-033.2 TURKEY POINT UNIT 4 ATTACHMENT I Control Room (Page 1 of 6)

[STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

1. EVACUATE all personnel from Containment.

A. ANNOUNCE the following over the plant PA System:

Attention all personnel in Unit 4 Containment, evacuate Unit 4 Containment.

B. ACTUATE the Containment Evacuation alarm.

C. ANNOUNCE the following over the plant PA System:

Attention all personnel in Unit 4 Containment, evacuate Unit 4 Containment.

2. DIRECT Radiation Protection to:

• LIMIT access to Containment.

• LIMIT access to Spent Fuel Building.

• MONITOR radiation levels in Containment and Spent Fuel Building.

• MONITOR Containment for airborne contamination.

• MONITOR cumulative whole body doses for personnel in Containment and Spent Fuel Building.

• MONITOR cumulative thyroid doses for personnel in Containment and Spent Fuel Building

REVISION NO.: PROCEDURE TITLE: PAGE:

REFUELING CAVITY SEAL FAILURE 10 of 30 PROCEDURE NO.:

4-ONOP-033.2 TURKEY POINT UNIT 4 ATTACHMENT I Control Room (Page 2 of 6)

I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I CAUTION Check that Containment is evacuated prior to performing Attachment 1, Step 3

3. ISOLATE Containment as follows:

A. INITIATE Containment Isolation Phase A.

B. CHECK Containment Isolation Phase A valve white lights on VPB are all BRIGHT.

C. ENSURE Unit 4 Containment purge exhaust and supply fans are OFF.

D. ENSURE Unit 3 Containment purge exhaust and supply fans are OFF.

E. CHECK of the following Containment IF y purge valve is NOT CLOSED, Purge Supply AND Exhaust Isolation THEN PULL fuses for jy OPEN purge Valves are CLOSED. valve, (fuses located behind VP B):

POV-4-2600, CONTAINMENT

• XEP, POV-4-2600, CONTAINMENT PURGE SUPPLY ISOL (O.C.) PURGE SUPPLY ISOLATION (O.C.)

POV-4-2601, CONTAINMENT

• XLAG, POV-4-2601, CONTAINMENT PURGE SUPPLY ISOL (l.C.f PURGE SUPPLY ISOLATION (l.C)

• POV-4-2602, CONTAINMENT

• XEQ for POV-4-2602, CONTAINMENT PURGE EXHAUST ISOL (O.Cj PURGE EXHAUST ISOLATION (O.C.)

• POV-4-2603, CONTAINMENT

• XLAH for POV-4-2603 PURGE EXHAUST ISOL (I.C.) CONTAINMENT PURGE EXHAUST ISOLATION (I.C.)

REVISION NO.: PROCEDURE TITLE: PAGE:

REFUELING CAVITY SEAL FAILURE 1 1 of 30 PROCEDURE NO.:

4-ONOP-033.2 . TURKEY POINT UNIT 4 ATTACHMENT I Control Room (Page 3 of 6)

I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I F. CLOSE the following:

• CV-4-2819, CNTMT INSTRUMENT AIR BLEED ISOL (IC)

• CV-4-2826, CNTMT INSTRUMENT AIR BLEED ISOL (00)

G. ENSURE aN normal Containmen?

coolers are tripped AND associated outlet dampers are CLOSED.

H. ENSURE Containment sump pumps are tripped.

I. ENSURE Control Room Ventilation in emergency recirculation mode per Attachment 4

4. CLOSE all inner and outer Containment access hatches.

5. CALCULATE time for SFP to reach 200°F based on Attachment 1, Step 6 performance using present or project level loss per jjy of the following:

• 0-ADM-051, Outage Risk Assessment And Control

• Plant Curve Book

6. STOP SFP cooling pumps.

7. SHUTDOWN all non-essential electrical equipment in the Containment lower levels.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 041 A3.05 Importance Rating 2.9 Ability to monitor automatic operation of the SDS, including: Main steam pressure Proposed Question: RO Question # 34 Given the following conditions:

• A Unit 3 startup is in progress.

• The crew prepares to synch the Main Generator to the grid.

• Rod Control is in Manual.

• Steam Dumps to Condenser Mode Selector Switch is in Manual.

• Tavg is 550°F.

• Tref is 548°F.

• Power Level is 7%.

Subsequently:

• The Turbine trips.

Which ONE of the following completes the following sentence for the initial plant response to the Turbine trip?

The Condenser Steam Dumps will (1) and (2)

A. (1) modulate open on the Turbine trip signal (2) close when the plant stabilizes at the same power level B. (1) quick open (2) slowly modulate closed based on the Tavg signal until Tavg is 547°F C. (1) quick open PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION (2) close when the plant stabilizes at the same power level D. (1) modulate open on the Steam Header pressure (2) close to maintain the Steam Header Pressure Controller setpoint Proposed Answer: D Explanation (Optional):

A. Incorrect. 1st part wrong. 2 part wrong. Plausible to believe that that the Condenser Steam dumps will open on the turbine trip signal, however, with the Mode Selector switch in Manual, the turbine trip signal provides no input into the control logic. Since the Condenser Steam dumps are capable of 27% power, it is plausible to believe that opening them would increase reactor power. However, they are in the pressure mode of control and will open and modulate to maintain set pressure (1005 psig).

B. Incorrect. 1 st part wrong. 2 part wrong. Plausible to believe since a quick open signal is part of the logic. However, for a turbine trip, the Tavg-Tref delta needs to be >7.5°F.

Also, since the Mode Selector is in Manual, the Quick open logic is defeated. Also plausible to believe the Condenser Dumps will modulate until Tavg is 547°F since the Condenser Steam dump Tavg Control program is designed to bring Tavg to a no load setpoint of 547°F with the Mode Selector in Auto. However, the program is bypassed when the Mode selector is in manual and receives its input from Steam Header Pressure.

C. Incorrect. jst part wrong. 2 rd part wrong. Plausible to believe since a quick open signal is part of the logic. However, for a turbine trip, the Tavg-Tref delta needs to be >7.5° F.

Also, since the Mode Selector is in Manual, the Quick open logic is defeated. Since the Condenser Steam dumps are capable of 27% power, it is plausible to believe that opening them would increase reactor power. However, they are in the pressure mode of control and will open and modulate to maintain set pressure (1005 psig).

D. Correct. They are in the pressure mode of control and will open and modulate to maintain set pressure (1005 psig).

Technical Reference(s): 5610-T-L1 Sh 22A Rev 14 .

(Attach if not previously provided) 3-GOP-3d Rev 30 Proposed References to be provided to applicants during examination:

PTN L-1S-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Learning Objective: LP 6902118 Obj 4, 5, 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

Procedure No Procedure

Title:

Page 58 Approval Date 3-GOP-301 Hot Standby to Power Operation 9/3/14 INIT 5.43 j the Steam Dumps to Condenser are to be used for startup, THEN verify the following:

5.43.1 Place the Mode Selector switch to Manual.

5.43.2 Verify Steam Header Pressure Controller output signal is less than 1 0 percent.

5.43.3 Place the Steam Dump to Condenser Control switch to On.

5.43.4 Verify receipt of Annunciator C 8/3, STEAM DUMP ARMED/ACTUATED.

5.43.5 Verify steam dump valves controlling steam header pressure at 1005 psig.

(.) 5.44 IF the reactor has NOT been made critical since the last refueling, THEN perform the following:

REACTOR ENGINEERING VERIFICATION POINTS Reactor Engineering shall:

(,) 1. Verify that the total rod worth is within 10 percent of design value.

(,) 2. Verify or update shutdown boron curves in the Plant Curve Book (Hot and Cold Shutdown).

(,) 3. Verify that the moderator temperature coefficient is proven less than or equal to plus 5 pcm/°F.

(.) 4. Verify that the HZP unrodded measured temperature coefficient ensures that the 100 percent power xenon free MTC is less than or equal to Zero OR Figure 13, Section 2 of the Plant Curve Book has been updated to reflect the measured data.

(,) 5. If an HZP flux map was NOT performed, then notify the Shift Manager that a flux map will be required at less than or equal to 30 percent power.

(,) 6. Verify that all Reactor Engineering surveillance tests required by 0-OSP-200. 1, Schedule of Plant Checks and Surveillances, and 0-ADM-215, Plant Surveillance Tracking Program, have been completed.

(,) 7. Verify that 3-OSP-059. 7, NIS Setpoint and Calibration Predictions for a New Cycle Startup, has been completed for applicable unit.

(,) 8. Verify new intermediate range alarm setpoints have been installed as necessary. (If setpoints did NOT change, then mark this step N/A.)

Reactor Engineering Supervisor or Designee: /

Signature Print Date W2003:TNM/cls/cls/ab

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A# 045 A1.06 Importance Rating 3.3 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MT/G system controls including: Expected response of secondary plant parameters following T/G trip Proposed Question: RO Question # 35 Given the following conditions:

• Unit 3 is at 100% power Subsequently:

• BOP manually trips the Turbine due to high vibration.

Which ONE of the following identifies how secondary system parameters respond?

PT-3-447, Turbine Inlet Pressure, (1) and Condenser Steam Dumps control RCS temperature (2)

A. (1) lowers (2) at no load Tavg B. (1) rises (2) at no load Tavg C. (1) lowers (2) within 5°F of program D. (1) rises (2) within 5°F of program PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: A Explanation (Optional):

A. Correct. 1 st part right. 2 part right. With the Mode Selector switch in Auto, Tavg Control will open and modulate the Steam Dumps to bring Tavg to no load of 547°F.

Turbine impulse pressure is the pressure to the High Pressure turbine downstream of the Turbine Control Valves, which will decrease when the Turbine Control Valves and Stop Valves close on the Turbine Trip.

B. Incorrect. 1 st part wrong. 2nd part right. Plausible to believe if incorrectly remembers that the Turbine Inlet Pressure taps off upstream of the Turbine Stop Valves. However, Turbine impulse pressure is the pressure to the High Pressure turbine downstream of the Turbine Control Valves, which will decrease when the Turbine Control Valves and Stop Valves close on the Turbine Trip.

C. Incorrect. 1 5t part right. 2 nd part wrong. Plausible to believe that since the Load Rejection Program closes the Condenser Steam Dumps at Tavg-Tref of 5°F, the same would occurs for the turbine trip. However, the logic for the turbine trip, With the Mode Selector switch in Auto, Tavg Control will open and modulate the Steam Dumps to bring Tavg to no load of 547°F.

D. Incorrect. 1 5t part wrong. 2 part wrong. Plausible to believe if incorrectly remembers that the Turbine Inlet Pressure taps off upstream of the Turbine Stop Valves. However, Turbine impulse pressure is the pressure to the High Pressure turbine downstream of the Turbine Control Valves, which will decrease when the Turbine Control Valves and Stop Valves close on the Turbine Trip. Plausible to believe that since the Load Rejection Program closes the Condenser Steam Dumps at Tavg-Tref of 5°F, the same would occurs for the turbine trip. However, the logic for the turbine trip, With the Mode Selector switch in Auto, Tavg Control will open and modulate the Steam Dumps to bring Tavg to no load of 547°F.

5613-T-L1 SH 22A Rev 14 Technical Reference(s): 5613-M-3089 SH 1 Rev 43 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Learning Objective: LP 6902118 Obj. 5, 9 (As available)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Corn ments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 055 K1.06 Importance Rating 2.6 Knowledge of the physical connections and/or cause effect relationships between the CARS and the following systems: PRM system Proposed Question: RD Question # 36 Given the following conditions:

• Unit 4 is at 100% power.

• Main Condenser vacuum is lowering.

• The crew enters 4-ONOP-014, Main Condenser Loss of Vacuum.

Which ONE of the following describes (1) the initial required action and (2) radiation monitoring capabilities?

A. (1) Place the Steam Jet Air Ejector Hogging Jet in service.

(2) SJAE SPING is not OPERABLE.

B. (1) Place the Steam Jet Air Ejector Hogging Jet in service.

(2) SJAE SPING is OPERABLE.

C. (1) Place the Standby Steam Jet Air Ejector in service.

(2) SJAE SPING is not OPERABLE.

D. (1) Place the Standby Steam Jet Air Ejector in service.

(2) SJAE SPING is OPERABLE.

Proposed Answer: A Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. Correct. In accordance with 4-ONOP-014, Immediate Operator actions are to place the Steam Jet Air Ejector Hogging Jet in service if vacuum is lowering. 4-ONP-014, Step 3.2.5 directs the RCO to notify Chemistry to taking compensatory actions for loss of SJAE R-15 SPING.

B. Incorrect. 1st part right. 2 part wrong. Plausible to believe the Hogging Jet Air Ejector discharges to the same location as the Steam Jet Air Ejectors. However, the Hogging Jet Air Ejector discharges to atmosphere, bypassing R-15.

C. Incorrect. 1st part wrong. 2nd part right. Plausible to believe that if a standby SJAE is available, the prudent action would be to place it in service. However, 4-ONOP-014 does not provide guidance to place the standby SJAE on line as part of the Immediate Operator Actions.

D. Incorrect. 1 part wrong. 2 part wrong. Plausible to believe that if a standby SJAE is available, the prudent action would be to place it in service. However, 4-ONOP-014 does not provide guidance to place the standby SJAE on line as part of the Immediate Operator Actions. Plausible to believe the Hogging Jet Air Ejector discharges to the same location as the Steam Jet Air Ejectors. However, the Hogging Jet Air Ejector discharges to atmosphere, bypassing R-15.

Technical Reference(s): 4-ONOP-014 Rev 6 .

5614-M-3014 SF1 3 Rev 24 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902282 Obj 3 6 (As available)

Question Source: Bank #

Modified Bank # 101355 (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Seabrook Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 K/A # 055 K1.06 Importance Rating 2.6 Knowledge of the physical connections and/or cause effect relationships between the CARS and the following systems: PRM system Proposed Question: RO Question # 36 -J Given the following plant conditions:

The plant is at 100% power.

• RM-6505, Condenser Air Evacuation Discharge Radiation Monitor is in ALARM.

Which of the following describes the significance of this alarm?

A. Radiation level on RM-6505 isolates SG Blowdown.

z B. Radiation level on RM-6505 provides an input to the calculation for an approximate value of Primary to Secondary Leak Rate.

C. Radiation level on RM-6505 is used to determine the need for a reactor trip and SI per 0S1227.02, Steam Generator Tube Leak.

D. Radiation level on RM-6505 is used to determine which secondary systems need to be isolated per 0S1227.02, Steam Generator Tube Leak.

Proposed Answer: B Explanation (Optional):

A. Incorrect but plausible. There are radiation monitors associated with SG Blowdown that will isolate the systems flash tank however RM-6505 does not provide that function.

B. Correct. RM-6505 is used for the calculated primary to secondary leak rate and also used in 0S1227.02 if the value must be calculated manually.

C. Incorrect but plausible. The leak rate is used to determine the rate of plant downpower however reactor trip and SI criteria are based on the threshold of FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 maintaining >7% pressurizer level utilizing two charging pumps.

D. Incorrect but plausible. RM-6505 indications are indicative of primary to secondary leakage, however it is indicative of leakage into the steam generators and not indicative of any specific secondary system.

Technical Reference(s): 0S1227.02 Steam Generator .

Tube Leak (Attach if not previously provided)

-J Proposed References to be provided to applicants during examination: NO z

Learning Objective: L1190l02 (As available)

Question Source: Bank# 101355 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Seabrook Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

6 MAIN CONDENSERLOSS OF VACUUM 5of16 PROCEDURE NO.:

4-ONOP-014 TURKEY POINT UNIT 4 I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 3.0 OPERATOR ACTIONS 3.1 Immediate Actions WARNING Hot water may be emitted from the silencer causing the potential for personnel injury.

1. CHECK Steam Jet Air Ejector is able iLACE SJAE hogging jet in service as to restore condenser vacuui follows:

A. OPEN 4-30-043, STEAM SUPPLY TO HOGGING JET VALVE.

B. Slowly OPEN 4-30-044, STEAM SUPPLY TO HOGGING JET VALVE until 4-Pl-1 597, HOGGING JET SUPPLY PRESSURE is between 250 and 260 psig.

C. OPEN 4-30-010, CONDENSER AIR REMOVAL TO HOGGING JET.

End of Section 3.1

REVISION NO.: PROCEDURE TITLE: PAGE:

6 MAIN CONDENSER LOSS OF VACUUM 7of 16 PROCEDURE NO.:

4-ONOP-014 TURKEY POINT UNIT 4 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 3.2 Subsequent Actions (continued)

3. CHECK Main Turbine in MANUAL A. SELECT OUT MW CNTRL.

control.

B. SELECT OUT TIP CNTRL.

4. CLOSE 4-30-045, HOGGING JET DRAIN.

5. NOTIFY Chemistry that the Hogging Jet has been placed in service rendering the SJAE SPING andy PRMS Channel R-15 monitors inoperable requiring compensatory actions to ensure compliance with the following:

• Technical Specification 3.3.3.3, Table 3.3-5, Action 34

• ODCM, Table 3.1-1, Action 3.1.3

6. CHECK both SJAEs are in service. PLACE standby SJAE in-service using Attachment 1, Placing Standby Steam Jet Air Ejectors in Service

7. CHECK condenser vacuum is being LOWER Turbine load to the Operating maintained in accordance with region on Attachment 2, Condenser Attachment 2, Condenser Vacuum Vacuum Limitations, per one of the Limitations, permissible region. following:

• 4-GOP-103, Power Operation To Hot Standby

• 4-GOP-i 00, Fast Load Reduction

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 K/A# 068 2.4.11 Importance Rating 4.0 Emergency Procedures / Plan: Knowledge of abnormal condition procedures.

Proposed Question: RO Question # 37 Which ONE of the following is a condition which causes X612, RADWASTE BLDG PANEL C46 TROUBLE, to alarm in the Control Room?

A. Spent Resin Storage Tank High Level B. Waste Monitor Tank A High Level C. Reactor Coolant Drain Tank Unit 3 High Level D. Waste Liquid High Radiation Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible to believe Spent Resin alarm is in the RWB since the resin is stored there. This alarm is an entry condition into 0-ARP-097.WB.A, Waste/Boron North Annunciator Response.

B. Correct. This alarm is an entry condition into 0-ARP-097.RB, Radwaste Building Process Control Panel C-46 Annunciator Response.

C. Incorrect. Plausible to believe Reactor Coolant Drain Tank alarm is in the RWB since the Waste Monitor Tanks and Waste Holdup Tank #2 reside there. This alarm is an entry condition into 0-ARP-097.WB.B, Waste/Boron South Annunciator Response.

D. Incorrect. Plausible to believe Waste Liquid Hi Radiation alarm is in the RWB since the Waste Monitor Tanks and Waste Holdup Tank #2 reside there. This alarm is an entry condition into 0-ARP-097.WB.B, Waste/Boron South Annunciator Response.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L45-1 DRAFT NRC EXAM SECURE NFORMATlON 0-ARP-097CR.X Rev 3 0-ARP-097.RB Rev 1 Technical Reference(s): (Attach if not previously provided) 0ARP097.WB.A Rev 0A 0-ARP-097.WB.B Rev 0 Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902149 Obj 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

3 35 CONTROL ROOM RESPONSE - PANEL X PROCEDURE NO.: WINDOW:

O-ARP-097.CR.X 6/2 TURKEY POINT PLANT (Page 1 of 1)

CAUSES: High Level in RWB Sump Room or High or low level in Waste Holdup Tank #2 X 12 or in any Waste Monitor Tank RADWASTE BLDG PANEL C46r TROUBLE DEVICE: SETPOINT: LOCATION:

Common alarm relay N/A Radwaste Building Control Panel NOTE NO Control Room indications.

ALARM CONFIRMATION N/A OPERATOR ACTIONS NOTIFY operator to perform actions required by O-ARP-097.RB, Radwaste Building Process Control Panel C-46 Annunciator Response.

REFERENCES:

Reference drawing and documents for each of the 10 active alarm inputs to this annunciator are listed in O-ARP-097.RB, Radwaste Building Process Control Panel C-46 Annunciator Response.

REVISION NO.: PROCEDURE TITLE: PAGE:

1 RADWASTE BUILDING PROCESS CONTROL PANEL C-46 PROCEDURE NO.:

O-ARP-097.RB TURKEY POINT PLANT ANNUNCIATOR PANEL 0-46 LJ LJ WASTE WASTE WASTE LS-3568 HOLDUP MONITOR MONITOR 1 HISUMP ABANDONED ABANDONED ABANDONED TANK#2 TANK A TANK A LEVEL ALARM ALARM ALARM HI-HI LEVEL HI LEVEL LO LEVEL LS-1078A LS-1082A LS-1082A PAGE 4 PAGE 7 PAGE 10 PAGE 13 PAGE 16 PAGE 19 PAGE 22 1/21 WASTE WASTE WASTE HOLDUP MONITOR MONITOR ABANDONED ABANDONED 2 SPARE TANK #2 ALARM ALARM TANKB TANKB ABANDONED HI LEVEL HI LEVEL ALARM LO LEVEL LS-1078A LS-10#8A LS-1088A PAGE 5 PAGE 8 PAGE 11 PAGE 14 PAGE 17 PAGE 20 PAGE 23 1/3I WASTE WASTE WASTE HOLDUP MONITOR MONITOR ABANDONED 3 SPARE TANK #2 ALARM ABANDONED ALARM TANK C TANK C FLASHER LO LEVEL HI LEVEL LO LEVEL LS-1078B LS-1089A LS-1089A PAGE 6 PAGE 9 PAGE 12 PAGE 15 PAGE 18 PAGE 21 PAGE 24 1 2 3 4 5 6 7

REVISION NO.: PROCEDURE TITLE:

PAGE:

WASTE/BORON NORTH PANEL ANNUNCIATOR RESPONSE PROCEDURE NO.:

O-ARP-097.WB.A V

TURKEY POINT PLANT ANNUNCIATOR PANEL WASTE/BORON NORTH

_J LJ LJ SPENT RESIN SPENT RESIN LAUNDRY & HOT LAUNDRY & HOT LAUNDRY & HOT 1 STORAGE TANK STORAGE TANK SPARE SHOWER TANK C SHOWER TANK A SHOWER TANK B SPARE SPARE HOLDUP TANK A HI LEVEL HI PRESSURE HI-LO LEVEL HI-LO PRESSURE HI-LO LEVEL HI-LO LEVEL PAGE 4 PAGE 8 PAGE 11 PAGE 14 PAGE 17 PAGE 20 PAGE 23 PAGE 26

_J PAGE 29 LJ WASTE HOLDUP WASTE HOLDUP HYDROGEN NITROGEN UNIT3 UNIT4 2 TANK TANK SUPPLY HEADER SUPPLY HEADER CHARGING PUMPS CHARGING PUMPS HI/HI-LO LEVEL SPARE SPARE HOLDUP TANK B HI LEVEL LO PRESSURE LO PRESSURE LUBE WATER LUBE WATER HI-LO PRESSURE LO LEVEL LO LEVEL PAGE 6 PAGE 9 PAGE 12 PAGE 15 PAGE 18 PAGE 21 PAGE 24 PAGE 27 1/31 PAGE 30 GAS DECAY GAS DECAY GAS DECAY GAS DECAY TANK GAS DECAY GAS DECAY GAS DECAY STANDBY SELECTOR 3 TANK A TANK B TANK C TANK D HI PRESSURE TANK E TANK F SWITCH SPARE HOLDUP TANK C HI PRESSURE HI PRESSURE HI PRESSURE HI PRESSURE HI-LO PRESSURE HI PRESSURE REQUIRE PAGE? PAGE 10 PAGE 13 REPOSITIONING PAGE 16 PAGE 19 PAGE 22 PAGE 25 PAGE 28 PAGE 31 1 2 3 4 5 6 7 8 9

REVISION NO.: PROCEDURE TITLE:

PAGE:

WASTE/BORON SOUTH PANEL ANNUNCIATOR RESPONSE 7

PROCEDURE NO.:

O-ARP-097.WB.B TURKEY POINT PLANT ANNUNCIATOR PANEL SOUTHWASTE/BORON i/il HOLDUP TANK A HOLDUP TANKS HOLDUP TANK C HOLDUP TANKS 1 HI-LO LEVEL MONITOR TANK A MONITOR TANKS VENT HEADER HI-LO LEVEL HI-LO LEVEL HI PRESSURE HL-LO LEVEL SPARE SPARE HI-LO LEVEL HI PRESSURE PAGE4 PAGE7 PAGE 11 PAGE 14 1/21 PAGE 18 PAGE22 PAGE25 PAGE28 2/21 5/21 PAGE31 6/2!

MOISTURE MOISTURE MOISTURE MOISTURE REACTOR000LANT REACTOR COOLANT SEPARATOR SEPARATOR SEPARATOR SEPARATOR REACTOR COOLANT REACTOR COOLANT 2 NO 1 PLANT STACK DRAIN TANK DRAIN TANK NO 1 NO 2 NO 2 DRAIN TANK DRAIN TANK LOLEVEL HI RADIATION UNIT 3 UNIT 3 HILEVEL LOLEVEL HILEVEL UNIT 4 UNIT 4 HILEVEL LOLEVEL HILEVEL LOLEVEL PAGE 5 PAGE 8 PAGE 12 PAGE 15 PAGE 20 PAGE 23 PAGE 26 PAGE 29 PAGE 32 REACTOR COOLANT REACTOR COOLANT REACTOR COOLANT REACTOR COOLANT iJ DRAIN TANK DRAIN TANK DRAIN TANK 3 DRAIN TANK WASTE LIQUID UNIT 3 UNIT 4 UNIT 3 SPARE SPARE UNIT 4 HI RADIATION SPARE SPARE HI PRESSURE HI PRESSURE HI TEMPERATURE HI TEMPERATURE PAGES PAGE 10 PAGE 13 PAGE 17 PAGE 21 PAGE 24 PAGE 27 PAGE 30 PAGE 33 1 2 -

4 5 6 7 8 9

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 2 Group# 2 KIA# 035 K3.01 Importance Rating 4.4 Knowledge of the effect that a loss or malfunction of the SIGS will have on the following: RCS Proposed Question: RO Question # 38 Given the following conditions:

• Unit 3 startup is in progress per 3-GOP-301, Hot Standby to Power Operation.

• Unit 3 is at 25% power and stable.

• Turbine is in MW Control.

Subsequently:

• CV-3-1606, A S/G Atmospheric Steam Dump fails open.

Which ONE of the following identifies the effect of this failure?

A. Main Generator electrical output lowers by 10% due to increased steam flow through the Atmospheric Steam Dump.

B. The Pressurizer experiences an insurge due to a change in Tavg.

C. The Turbine Control valves move closed to restore the Turbine Inlet pressure.

D. Reactor Coolant Cold Leg temperature lowers and results in a Reactor power rise.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible to remember that the Steam Dump to Atmosphere (SDTA) capacity is rated at 10% total steam flow for the steam dumps and if the SDTA failed it would bypass the immediate steam flow to the turbine. However, with only one failure full open, the SDTA is steam flow is about 3  %. The Turbine is in MW control and the Turbine Control valves will open in response to TCS to maintain MW output.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION B. Incorrect. Plausible to believe since Pressurizer will insurge with a Tavg increase.

Examinee may confuse increasing AT with increasing Tavg. The SDTA opening will increase AT, decrease Tavg and causing a Pressurizer outsurge.

C. Incorrect. Plausible to believe since with TCS in TIP control EHC would try to maintain Turbine Impulse pressure. However, the system would open the Control valve and also the Turbine is in MW Control.

D. Correct. When the SDTA opens, it is capable of 3.5 % steam flow equivalent. This will have an effect in decreasing Tavg, increasing feed flow and increasing Reactor Power.

Technical Reference(s): 3-GOP-301 Rev 30 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Learning Objective: LP 6902127A Obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

INIT 5.54.5 WHEN the GCB closes, THEN perform the following:

I. Observe the East Bus Breaker indicating lights to verify breaker closure (red On; green Off).

2. IF main generator load is less then 10 MWe, THEN increase load to approximately 10 MWe using the Turbine Speed/Pwr Control switch or using the Raise button on HMI TCS Load Control Screen (each touch of the Raise button equals approximately I MWe).

3. Place synchroscope in the Off position.

4. Match the flag on the East Bus Generator GCB Control Switch by taking the switch to Close.

5. Verify the Inadvertent Protection Scheme Armed amber light above the synchroscope is Off.

6. Verify Generator Amps are within 2 percent on all three phases.

I NOTES I

• Top cessve changes in S/G pressure and leve the SDTA valves should be I

checL it they are responding prior to each additional load step.

I load shall be coordinated with the operator controlling steam ge7!PJor levels.

a S a I

I 5.55 Perform the following to increase turbine load: [Commitment Step 2.3.11 - CAPR]

5.55.1 Monitor automatic control program values using the Plant Curve Book Section IV, Figure 5 AND notify the Shift Manager of any unexpected deviations.

5.55.2 Raise Turbine Load using 3-NOP-089, Main Turbine.

1. Below 40 MWe, only the following methods are available.

• Raise/Lower Buttons on Load Control Screen

• Speed/Load Turbine Control Switch

2. AbMWe, the! lowing methods are available:,

• Raise/Lower Buttons on Load Control Screen

• Speed/Load Turbine Control Switch

• Megawatt Control

• TIP Control W201 O:TNM/cls/ab/cls

Procedure No Procedure

Title:

Page 69 Approval Date:

3-GOP-301 Hot Standby to Power Operation 9/3/14 INIT NOTE The following step is performed by the operator controlling steam generator levels and pressures.

a 5.55.3 IF the Steam Dump to Atmosphere (SDTA) valves are being used, THEN operate SDTAs to control Main Steam pressure and Tave per the following steps until all SDTA valves are Closed and the Tavg Tref DeltaT is within the band provided by the US. (Reference Attachment 5 for operation of the SDTA controllers.)

I. Verify the SDTA controllers in Automatic are Closing the SDTA valves as steam is drawn off to the turbine.

2. Slowly Close the SDTA valve in Manual to balance steam flow with the SDTA valves in Automatic AND make minor adjustments to Tavg, as necessary.

NOTES i i i

• When the SD TA valves are operating properly, there should be a balance between the I SDTA valves closing and main turbine steam usage, with little perturbation in main I steam header pressure as load is increased.

I I

. The SDTA valves can be verified to be closing by observing main steam header I pressure recover as the main generator is loaded.

I I

. The SDTA controller settings may be adjusted in small increments as necessaiy to I maintain steam generator levels.

I I

. It should NOT be necessa,y to close the SDTA valve in manual as a pre-emptive I action when the main generator output breaker is Closed.

a a a I

3. Observe main steam header pressure while loading the main generator to maintain a balance between the SDTA valves closing and the steam being used to increase load.

4. WHEN steam generator levels and pressures stabilize following a load increase, THEN notify the operator controlling the main generator to increase load by 5 to 10 MWe.

5. Continue monitoring and controlling in the steps above until Substep 5.5 5.3.6 below is completed.

W201 O:TNM/cls/ab/cls

INIT 5.55.3 (Contd)

NOTES The SDTA Ws should be closed by approximately 40 MWe. I I

. Once Turbine Power has exceeded 40 MWe, MW CNTRL and TIP CNTRL become enabled on the TCS. I a a I

6. WHEN the SDTA valves in automatic are Closed. THEN ensure the SDTA valve in Manual is Closed AND Tavg/Tref are within the band provided by US.

7. Align the SDTA controllers for automatic operation as follows:

a. Verify Steam Dump to Atmosphere Valve, CV-3-1606, is Closed.

(1) Adjust the controller setpoint to 1005 psig.

(2) Ensure the controller is in Auto.

b. Verify Steam Dump to Atmosphere Valve, CV-3-l607, is Closed.

(1) Adjust the controller setpoint to 1005 psig.

(2) Ensure the controller is in Auto.

c. Verify Steam Dump to Atmosphere Valve, CV-3-1608, is Closed.

(1) Adjust the controller setpoint to 1005 psig.

(2) Ensure the controller is in Auto.

8. Perform the following to align the steam dump to condenser for Auto:

a. Place the Steam Dump to Condenser Control switch in the On position.

b. Momentarily place the Mode Selector switch to Reset.

c. Place the Mode Selector Switch to Auto.

5.55.4 IF the steam dump to condenser (SDTC) valves are being used, THEN perform the following:

1. Verify the SDTC valves are closing as steam is drawn off to the turbine.

2. WHEN load has increased sufficiently to cause the SDTC valves to fully Close, THEN place the Mode Selector Switch to Reset, then to Auto.

3. Verify Annunciator C 8/3, STEAM DUMP ARMED/ACTUATED, clears.

W201 O:TNM/cls/ab/cls

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 007 EK1.02 Importance Rating 3.4 Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Shutdown margin Proposed Question: RO Question # 39 Given the following plant conditions:

• Unit 3 trips from 100% power on a spurious SI actuation.

• Crew enters 3-EQ P-E-0, Reactor Trip or Safety Injection.

• RCS pressure is 2150 psig and recovering.

• RCS Tavg stabilizes at 535°F.

Which ONE of the following correctly completes the following statement?

To ensure adequate Shutdown Margin, a minimum boration rate of (1) is required. With no operator action taken, Shutdown Margin over the next hour (2)

A. (1)2ogpm (2) rises B. (1)45gpm (2) rises C. (1)2ogpm (2) stays the same D. (1)4sgpm (2) stays the same PTN L-15-1 DRAFT NRC EXAM SECURE NFORMATON

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: A Explanation (Optional):

A. Correct. 3-EOP-E-0 states: IF High-Head SI Pumps have NOT iniected, AND any RCS Tavg is less than 537°F, THEN initiate and continue a boration of at least 20 qpm until Shutdown Margin is verified using PLANT CURVE BOOK. The buildup of xenon will added negative reactivity over the next 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, increasing SDM.

B. Incorrect. 1st part wrong, 2nd part right. Plausible since that is the maximum boration rate and plausible to believe that would be the required flowrate.

C. Incorrect. 1st part right, 2nd part wrong. Plausible to believe that if no action is taking place and the plant is stabilizing and forgetting the buildup of xenon, that the SDM would not change.

D. Incorrect. 1st part wrong. 2nd part wrong. Plausible since that is the maximum boration rate and plausible to believe that would be the required flowrate. Plausible to believe that if no action is taking place and the plant is stabilizing and forgetting the buildup of xenon, that the SDM would not change.

3-EOP-E-0 Rev 10 Technical Reference(s): .

(Attach if not previously provided)

Proposed References to be provided to applicants during examination: N LP6902321 Obj4 10 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

10 REACTOR TRIP OR SAFETY INJECTION 13 Of 55 PROCEDURE NO.:

3-EOP-E-0 TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

• 10. Check RCS Temperatures:

a. Check RCPs ANY RUNNING a. Go to Step 1 0.e.

b. Check RCS Average temperatures b. Perform the following:

using DCS

1) IF TAVE is decreasing, STABLE BETWEEN 545°F THEN perform the following:

AND 547°F a) Stop dumping steam.

b) IF cooldown continues AND is due to excessive feed flow, TRENDING DOWN TO 547°F THEN reduce total feed flow to 400 gpm unt narrow range level greater than 7%[27%] in at least one S/G.

c) IF cooldown continues, AND is due to excessive steam flow, THEN close Main Steamline Isolation and Bypass valves.

2) IF TAVE greater than 547°F AND increasing, THEN perform y of the following:

Dump steam to Condenser.

OR Dump steam using S/G Steam Dump To Atmosphere valves.

3) IF High-Head SI Pumps have NOT injected, AND y RCS TAVE IS bless than 537°F, THEN initiate arid continue a boration of at Ieasd 20 gpm using Attachment 7 jl Shutdown Margin is verified using PLANT CURVE BOOK and go to Step 11.

C. IF all the following conditions exist:

• HHSI Pumps have NOT injected

• Fuel burnup 18,000 MWD/MTU

• RCS TAVE less than 547°F THEN initiate and continue a boration of at least 20 gpm using Attachment 7 I Shutdown Margin can be verified using PLANT CURVE BOOK

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 009 EK1.01 Importance Rating 4.2 Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Natural circulation and cooling, including reflux boiling Proposed Question: RO Question # 40 Given the following conditions:

• A loss of all AC power occurs on Unit 3.

• The crew enters 3-EOP-ECA-0.0, Loss of All AC Power.

• Pressurizer level is 5% and lowering.

• Containment Sump level is rising.

• The crew prepares to depressurize the SGs.

Which ONE of the following actions (1) ensures natural circulation or reflux boiling cooling is sufficient and (2) lists the applicable recovery procedure after power is restored?

A. (1) Verify total AFW flow between 400 and 450 gpm.

(2) 3-EOP-ECA-0. 1, Loss of All Power Recovery Without SI Required B. (1) Maintain >7% Narrow Range Level in at least one SG.

(2) 3-EOP-ECA-0.1, Loss of All Power Recovery Without SI Required C. (1) Verify total AFW flow between 400 and 450 gpm.

(2) 3-EOP-ECA-0.2, Loss of All Power Recovery With SI Required D. (1) Maintain >7% Narrow Range Level in at least one SG.

(2) 3-EOP-ECA-0.2, Loss of All Power Recovery With SI Required Proposed Answer: D PTN L-15-I. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Incorrect. 1 st part wrong. 2 part wrong. Plausible because AFW flow provides for core cooling and is Step 4 in 3-EOP-ECA-O.O. Natural circulation will normally occur on a LOOP with the SDTA valves removing excess heat as long as the SG tubes remain covered. When the RCS loses subcooling, natural circulation is interrupted and heat removal shifts to reflux boiling, which also requires that the SG levels be> 7% to ensure the tubes are covered to continue the cooling process. Plausible to believe entry into 3-EOP-ECA-O.1, since SI has not actuated. However, entry into 3-EOP-ECA-O.2 Loss of ,

All Power Recovery With SI Required, is correct due to Pressurizer level less than 7%

and SI has not actuated.

B. Incorrect. 1st part right. 2 nd part wrong. Plausible to believe entry into 3-EOP-ECA-O.1, since SI has not actuated. However, entry into 3-EOP-ECA-O.2 Loss of All Power Recovery With SI Required, is correct due to Pressurizer level less than 7% and SI has not actuated.

C. incorrect. 1st part wrong. 2nd part right. Plausible because AFW flow provides for core cooling and is Step 4 in 3-EOP-ECA-O.O. Natural circulation will normally occur on a LOOP with the SDTA valves removing excess heat as long as the SG tubes remain covered. When the RCS loses subcooling, natural circulation is interrupted and heat removal shifts to reflux boiling, which also requires that the SG levels be> 7% to ensure the tubes are covered to continue the cooling process.

D. Correct. This describes the action that ensures sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates. Entry into 3-EOP-ECA-O.2, Loss of All Power Recovery With SI Required, is correct due to Pressurizer level less than 7% and SI has not actuated.

Technical Reference(s): 3-EOP-ECA-O.O Rev 6 BDEOPECA-o.o Rev 6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902324 Obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

6 LOSS OF ALL AC POWER 20 of 88 PROCEDURE NO.:

3-EOP-ECA-O.0 TURKEY POINT UNIT 3 I STEP I ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I CAUTION

. SIG pressures shall NOT be decreased to less than 130 psig to prevent injection of Accumulator nitrogen into the RCS.

. S/G Narrow Range Level is required to be maintained greater than 7%[27%] in at least one intact S/G. If level can NOT be maintained, S/G depressurization is required to be stopped jffljl level is restored in at least one SIG, NOTE

. SIGs are required to be depressurized at a rate sufficient to maintain a cooldown rate in the RCS Cold Legs near 100°F/hr. This will minimize RCS inventory loss while cooling the RCP seals in a controlled manner.

. Although PRZ level may be lost and reactor vessel upper head voiding may occur due to depressurization of S/Gs, depressurization shall NOT be stopped to prevent this.

14. Depressurize Intact SIGs To 230 psig:

a. Check S/G Narrow Range Levels a. Perform the following:

GREATER THAN 7%[27%] IN AT LEAST ONE SIG 1) Maintain maximum feed flow untU Narrow Range Level greater than 7%[27%] in at least one S/G.

2) WHEN Narrow Range Level greater than 7%[27%] in at least one SIG, THEN do Step 14.b through Step 14.e.

Continue with Step 15.

b. Manually dump steam using S/G Steam Dump To Atmosphere valves to maintain cooldown rate in RCS Cold Legs at less than 100°F/hr

Page 45 BD-EOP-ECA-O.O Loss of All AC Power 8/15/14 WOG Procedure Step 16 CAUTION 2 PTN Procedure Step 14 CAUTION 2 S/G Narrow Range Level is required to be maintained greater than 7%127%1 in at least one intact SIG. If level can NOT be maintained, S/G depressurization is required to be stopped until level is restored in at least one S/G.

BASIS:

During thrapid&#ation performed in the next step, S/G level could drp out of the narrow range resulting in a loss of adequate heat sink, if this situation occurs, the depressurization should be stopped and AFW flow reestablished until S/G level is restored in to greater than 7%[27%J. The analysis for ECA-O.O requires that the level in at least one intact SG be above the top of the SG U-tubes to ensure that sufficient heat transfer capability exist to remove heat from the RCither via natural circulation or reflux boiling after the RCS saturates. This is accomplished in step 10 which requires maximum AFW flow to be maintained to the intact SGs until level in at least one intact SG is in the narrow range, and the second caution before Step 14 (which requires that level in at least one intact SG be maintained in the narrow range during SG depressurization). Once these conditions are met, Step 14 directs the operator to dump steam (depressurize intact SGs) at a rate sufficient to expeditiously reduce RCS temperature and pressure (which in turn will reduce the rate of RCS inventory loss through the RCP seals) while cooling the RCP seals in a controlled manner (to avoid further damage due to thermal shock). An RCS cooldown rate of 100°F/hr accomplishes this. Step 14 is structured to provide flexibility in depressurizing the intact SGs. The depressurization may be performed using one, more than one, or all intact SGs, although the preferred method to depressurize the intact SGs is to uniformly release steam from all intact SGs at a controlled rate that will not cause the level in the SGs to drop out of the narrow range, and not cause the RCS cooldown rate to exceed IOOF/hr. If cooling down at, or near 100F/hr will cause level in the SGs to drop out of the narrow range, it is acceptable to keep one intact SG isolated (with level in the narrow range) while depressurizing the other (i.e., one or more than one) intact SGs.

Depressurization in this manner may avoid the potential to lose level in all intact SGs which would require the depressurization to be stopped until level could be restored in at least one intact SG. If the depressurization is performed with one intact SG isolated, the isolated intact SG should eventually be unisolated and depressurized once the concern for losing level in all intact SGs no longer exists.

STEP DEVIATIONS FROM WOG GUIDELINES:

TYPE DESCRIPTION 8 The WOG guidelines do not provide distinct definitions for the terms should and shall. The word should was changed to is required to to denote a requirement.

PLANT SPECIFIC SETPOINTS:

7% S/G level just in the narrow range plus normal channel accuracy. (EOP Setpoint M.2) 27% S/G level just in the narrow range plus normal channel accuracy. post-accident transmitter errors and reference leg process errors. (EOP Setpoint M.3)

W201 O:/DH/In/cls/cls

Page 48 BD-EOP-ECA-0.0 Loss of All AC Power 8/15/14 WOG Procedure Step 16 PTN Procedure Step j Depressurize Intact S/Gs To 230 psig BASIS:

This step depressurizes the intact SGs, thereby reducing RCS temperature and pressure to reduce RCP seal leakage and minimize RCS inventory loss.

During SG depressurization, SG level must be maintained above the top of the SG U-tubes in at least one SG. Maintaining the U-tubes covered in at least one SG will ensure that sufficient heat transfer capability exists to remove heat from the RCS via either natural circulation or reflux boiling after the RCS saturates Step a requires that SG level be in the narrow range in at least one SG before SG depressurization is initiated in substep b. If level is not in the narrow range in at least one SG, substep a RNO instructs the operator to maintain maximum AFW flow until narrow range level is established in one SG. When narrow range level is established, SG depressurization can be started or continued via substep b.

Substep b instructs the operator to reduce SG pressures by depressurizing the intact SGs.

Depressurization should be accomplished by opening the steam dump to atmosphere valves on the intact SGs to establish a cooldown rate of less than 100°F/hr in the RCS cold legs. By maintaining RCS cooldown rate less than 100°F/hr, the RCP seal temperatures are reduced in a controlled manner to prevent thermal shock.

Once depressurization is initiated, the depressurization rate should be controlled to maintain RCS cooldown rate near 100°F/hr. The depressurization rate should be sufficiently fast to expeditiously reduce SG pressures, but not so fast that SG pressures and RCS cooldown cannot be controlled. It is important that the depressurization not reduce SG pressures in an uncontrolled manner that undershoots the pressure limit, thus permitting potential introduction of nitrogen from the accumulators into the RCS.

During SG depressurization, AFW flow may have to be increased to maintain the required SG narrow range level. Control of AFW flow will have to be performed from the Control Room or locally depending on plant specific design. Full AFW flow should be established to any SG in which level drops out of the narrow range.

(Continued on next page)

W201 O:/DH/In/cls/cls

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 011 2.1.20 Importance Rating 4.6 Conduct of Operations: Ability to interpret and execute procedure steps.

Proposed Question: RO Question # 41 Given the following conditions:

• Unit 3 Reactor trips from 100% power.

• SI and Phase A actuate.

• Crew enters 3-EOP-E-0, Reactor Trip or Safety Injection.

• The BOP is directed to perform Attachment 3 of 3-EOP-E-0.

• The BOP checks the step to verify Containment Spray requirements.

• Containment pressure is 17 psig and rising slowly.

Which ONE of the following identifies the actions the crew must take?

A. Continue with the remaining steps of the attachment. If Containment Spray is required, then a peer check must be obtained prior to actuation.

B. Continue with the remaining steps of the attachment and monitor Containment pressure to verify Containment Spray NOT required.

C. Request permission from the Unit Supervisor. Manually initiate Containment Spray and Phase B prior to allowing an automatic actuation.

D. Perform an Update to the crew of a prudent operator action. Then, manually initiate Containment Spray and Phase B prior to allowing an automatic actuation.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible to think that since if actuation is required and as such would it be an irreversable action, that a peer check would be required. However, this is a continuous action step and since the criteria has been achieved, the RCO must perform the RNO, which does not require peer checks (0-ADM-211, step 4.6.11.A.)

B. Correct. 3-EOP-E-0, Attachment 3, step 10, is a continuous action step. 0-ADM-21 1, Emergency and Off-Normal Procedure Usage, defines a Continuous Action Step as actions provided in a procedure which direct or imply continuous performance throughout the remainder of the procedure.

C. Incorrect. Plausible to believe that the Unit Supervisor would need to provide authorization to manipulate controls to prevent an automatic actuation. 0-ADM-21 1, step 4.7.2.0 specifically states that manual actuation of Containment Spray and Containment Phase B isolation shall NOT be performed before the setpoint is reached.

D. Incorrect. Plausible to believe that a Crew update would be performed since 0-ADM-211 provides such guidance for a prudent operator action. 0-ADM-21 1, step 4.7.2.D specifically states that manual actuation of Containment Spray and Containment Phase B isolation shall NOT be performed before the setpoint is reached.

Technical Reference(s): 3-EOP-E-0 Rev 10 (Attach if not previously provided) 0-ADM-21 1 Rev 3 Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902320 Obj 1 4-6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATiON 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE:

PAGE:

10 PROCEDURE NO.:

REACTOR TRIP OR SAFETY INJECTION I 39 of 55 3-EOP-E-0 TURKEY POINT UNIT 3 ATTACHMENT 3 Prompt Action Verifications (Page 6 of 12)

I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

+ 10. Verify Containment Spray NOT Required:

a. tontainment pressure HAS a. Perform the following:

REMAINED LESS THAN 20 PSIG:

1) IF Containment Spray NOT

• PR-3-6306A initiated, THEN manually initiate Containment Spray.

• PR-3-6306B 2) Verify Containment Isolation Phase B has actuated.

3) Verify Containment Isolation Phase B Valve white lights on VPB are aB bright.

4) IF Containment Isolation Phase B Valve did NOT close, THEN manually or locally isolate affected Containment Penetration.

5) Stop jj RCPs.

11. Verify SI RESET Reset SI.

12. Verify SI Valve Amber Lights On VPB Manually align valves to establish proper SI ALL BRIGHT alignment for an injection flowpath.

REVISION NO.: fPRO CEDURE TITLE: PAGE:

3 EMERGENCY AND PROCEDURE NO.: OFF-NORMAL OPERATING PROCEDURE USAGE 25 Of 46 O-ADM-21 1 TURKEY POINT PLANT 4.7 Procedure Adherence for Emergency and Off Normal (Abnormal)

Procedures

1. When performing actions within the EOPs and ONOPs (AOPs),

operators are expected to comply with written direction. In order to comply, operators are expected to systematically implement the procedure to mitigate accidents and failures; however, unanticipated circumstances may cause an event to proceed differently than was anticipated when the procedure was written. As a result, certain situations may interrupt or require a deviation from the procedure in effect. The following provides additional guidance for the infrequent cases when a deviation is appropriate.

NOTE Operator action taken in anticipation of automatic actions or action taken to correct a failed or incomplete action is NOT considered a deviation.

2. Prudent (Manual) Operator Actions A. Prudent Operator Actions are those actions that may be done concurrently with the implementation of the Emergency Procedures or ONOPs that do NOT prevent or conflict with the performance of the EOP or ONOP (AOP).

B. Prudent Operator Actions shall be limited to the following:

(1) 4f the setpoint for an amatic actuation signal is reached and the actuation fails or, if a condition will cause an automatic action to occur, the operator should manuaIly initiate the signal.

(2) Taking manual action to control plant parameters if an automatic control system is out of service or malfunctioning.

(3) Taking manual compensatory actions for failed equipment (e.g., isolating leakage, stopping a malfunctioning pump, etc.).

(4) Taking manual action to compensate for a failed automatic action. If multiple components do NOT actuate due to a failed automatic signal (such as one train of SI failing to actuate), the operators should attempt to manually actuate the failed signal (if possible) prior to manually operating individual components.

REVISION NO.: PROCEDURE TITLE: PAGE:

3 EMERGENCY AND PROCEDURE NO: OFF-NORMAL OPERATING PROCEDURE USAGE 26 of 46 O-ADM-21 1 TURKEY POINT PLANT 4.7 Procedure Adherence for Emergency and Off Normal (Abnormal)

Procedures (continued)

2. B. (continued)

(5) Actions taken to protect personnel or equipment whenever an imminent threat may exist for events such as high energy line breaks or electrical bus faults. This includes, but is NOT limited to the following:

• Closing Main Steam Isolation Valves when a fault in the Main Steam System is identified

• De-energizing an electrical bus in response to a report of a personnel electrocution (6) If redundant stand-by equipment is available and ready, the operator is permitted to start the redundant equipment for failed or failing operating equipment.

Immediate follow up of applicable ARPs and ONOPs (AOPs) shall occur as required.

C. The operator should announce his intentions to perform a prudent operator action. This will allow the remaining crew members to independently evaluate the current plant conditions and validate or refute, as necessary, the need for the manual action.

D. Manual actuation of Containment Spray and Containment Phase B Isolation shall NOT be performed before the setpoint is reached.

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 015 AA1.05 Importance Rating 3.8 Ability to operate and / or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): RCS flow Proposed Question: RO Question # 42 Given the following conditions:

• Unit 4 is at 38% power and stable.

• 4C Loop RCS Flow Meters are: 86%, 89%, and 91%.

Which ONE of the following identifies the appropriate procedural response?

A. Enter 4-ONOP-041 .1, Reactor Coolant Pump Off-Normal.

B. Enter 4-ONOP-049. 1, Deviation or Failure of Safety Related or Reactor Protection Channels.

C. Enter 4-EQ P-E-0, Reactor Trip or Safety Injection.

D. Enter 4-EOP-FR-S.1, Response To Nuclear Power Generation / ATWS.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible to believe that since the problem appears to be with RCS flow, the RCP off-normal procedure would be the appropriate guidance. However, a reactor trip is required since the plant is not permitted to operate with only two loops.

4-ARP-097.CR.B Panel B1/3 provides guidance for loss of flow indication and if 2 of 3 channels indicate less than 90%, the correct action is to trip the reactor and enter 4-EOP-E-0, Reactor Trip or Safety Injection.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION B. Incorrect. Plausible to believe that the flow instrument has failed since flow is still indicated. Ssince the problem appears to be with RCS flow, that 4-ONOP-049.1, Deviation or Failure of Safety related or Reactor Protection Channels would be the appropriate guidance.

C. Correct. 4-ARP-097.CR.B Panel B113 provides guidance for loss of flow indication and if 2 of 3 channels indicate less than 90%, the correct action is to trip the reactor and enter 4-EQ P-E-0, Reactor Trip or Safety Injection.

D. Incorrect. Plausible to believe that the reactor should have tripped on low flow in the RCS and since it did not, the examinee may believe an ATWS has occurred, requiring an entry into 4-EOP-FR-S.1, Response To Nuclear Power Generation / ATWS.

However, power is below P-8 and will not trip for loss of one loop. And if the reactor should have tripped, the correct response is to enter 4-EOP-E-0, and if manual trip fails, then transition to 4-EOP-FR-S-1 is required.

Technical Reference(s): 4-ARP-097.CR.B Rev 7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP6902916Obj.3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

7 6 CONTROL ROOM RESPONSE - PANEL B PROCEDURE NO.: WINDOW:

4-ARP-097.CR.B TURKEY POINT UNIT 4 1/3 (Page 1 of 1)

CAUSES: One or more RCS loop flow channels below 90% due to:

• RCS loop flow channel failure(s) B113

• RCP pump trouble

• Vital 4KV bus undervoltage

• Vital 4KV bus under frequency RX COOLANT LOOP C LO FLOW DEVICE: SETPOINT: LOCATION:

• FC-434 One or more less than 90% N/A

• FC-435

• FC-436 ALARM CONFIRMATION

1. CHECK RCS loop C flow indications on VPA and status lamps on VPB to pinpoint affected channels.

2. CHECK 4C RCP tripped by observing pump breaker indicating lights AND PUMP current meter on VPA.

3. CHECK 4B 4KV Bus voltage and frequency.

4. CHECK RCS loop C Th and T recorders on VPA.

OPERATOR ACTIONS

1. IF RX power is above P-8 AND two out of three RCS Loop C flow channels are below 90%, THEN:

A. ENSURE the reactor is tripped.

B. PERFORM 4-EOP-E-0, Reactor Trip or Safety Injection.

2. IF RX power is below P-8 AND two out of three RCS Loop cre belowr 90%, THEN:

A. TRIP the reactor.

B.PERFORM 4-EOP-E- or Trip or Safety Injection.

3. IF only one RCS Loop C flow channel is below 90%, THEN:

A. CHECK channel power supply in rack.

B. REFER TO 4-ONOP-049. 1, Deviation or Failure of Safety Related or Reactor Protection Channels.

REFERENCES:

1. FPL Drawing 5614-M-3041

2. FPL Logic Diagram 5610-T-L1, Sheet 20

3. Tech Spec Sections 3.3.1, 3.4.1.1, 3.4.1.2, and 3.4.1.3

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Facility: Turkey Point Vendor: WEC Exam Date:

Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# 022 AK3.01 Importance Rating 2.7 Knowedge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: Adjustment of RCP seal backpressure regulator valve to obtain normal flow Proposed Question: RO Question # 43 Given the following conditions:

• Unit 3 is at 100% power.

• RCS Pressure is 2235 psig.

• 3A Charging Pump is running in automatic.

• HCV-3-121, Charging Flow To Regen Heat Exchanger, is throttled.

Subsequently:

• Annunciator A6/5, RCP LABYRINTH SEAL LO AP, alarms.

• Annunciator A5/1, CHARGING PUMP A TRIP, alarms.

• The RCO starts 3C Charging Pump.

• The RCO balances Charging and Letdown flows.

• VCT level is stable.

• RCP Seal Injection flows are 4 to 5 gpm.

Which ONE of the following completes the following sentence?

The RCO will throttle (1) HCV-3-121 to (2) to address the given conditions?

A. (1) closed (2) increase flow to the Regenerative Heat exchanger PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION B. (1) open (2) reduce flow to the Regenerative Heat exchanger C. (1) closed (2) increase flow to the RCP seals D. (1) open (2) reduce flow to the RCP seals Proposed Answer: C Explanation (Optional):

A. Incorrect. 1st part right. 2 part wrong. Plausible to believe, since 3-ARP-097.CR.A states that the operator should adjust HCV-3-121, Charging Flow To Regen Hx to increase seal injection flow and it is plausible to incorrectly remember the minimum RCP seal injection flow rate and believe that the correct action is to increase flow to the Regenerative Heat exchanger. However, closing HCV-3-121 reduces flow to the Regenerative Heat exchanger and increases flow to the RCP Seals.

B. Incorrect. 1st part wrong. 2 nd part right. Plausible to confuse what the effect of adjusting HCV-3-121 has on the RCP Seals and believing that it will increase flow to the RCP seals by opening HCV-3-121, although opening HCV-3-121 will increase flow to the Regenerative Heat exchanger.

C. Correct. 3-ONOP-041.1 and 3-NOP-041.O1A provide guidance to increasing seal injection flow by throttling CLOSED HCV-3-121.

D. Incorrect. 1 st part wrong. 2nd part wrong. Plausible to believe since 3-ONOP-041.1 and 3-NOP-041,9A provide guidance by stating that RCO needs to adjust HCV-3-121 and RCO incorrectly determines that opening the valve will increase flow. Seal Injection flow is increased by throttling CLOSED HCV-3-121 and it is plausible to incorrectly remember the minimum RCP seal injection flow rate and believe that the correct action is to increase flow to the Regenerative Heat exchanger.

Technical Reference(s): 3-ONOP-041.1 Rev 7 (Attach if not previously provided) 3NOP-041 .O1A Rev 4 Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902205 Obj. 7, 8 (As available)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

3A REACTOR COOLANT PUMP OPERATIONS 9 Of 37 PROCEDURE NO.:

3-NOP-041 .O1A TURKEY POINT UNIT 3 4.11 Starting 3A Reactor Coolant Pump (continued)

9. IF thermal barrier differential pressure is less than jg inches of water, as indicated on PI-3-131A, A RCP THERMAL BARRIER D/P, THEN:

NOTE A minimum of 6pm required for RCPSeal Injection.

A. While RCO monitors DOS ROP Detailed Data Summary display for flow changes, SNPO locally ADJUST 3-297A, 3A RCP SEAL WATER SUPPLY, to obtain 6 to 13 gpm.

B. IF 3-297A is full OPEN AND seal injection flow is less than 6 gpm, THEN, while SNPO monitors RCP seal injection flow locally, RCO ADJUST HCV-3-121, CHARGING FLOW TO REGEN HX, CLOSED to maximize seal injection flow.

NOTE

• Following initial verification of RCP COW flows and S/G temperatures, re-verification is N/A for subsequent RCP starts if, at discretion of the Shift Manager, NO intervening activities occurred to adversely impact those flows or temperatures.

• Some flow indicators are located in different locations. Two operators are needed in constant communication when adjustments are made.

• Adjustments should be done in small increments, no more than 1/8 of a turn at a time.

10. ADJUST the following COW flows to the 3A RCP are required:

• ADJUST 3-728A, ISO VLV FOR COW RTN FROM THRML BARRIER COOLING COIL A, to maintain FI-3-630 between 21 to 28 gpm.

• ADJUST 3-723A, ISO VLV FOR COW FROM LOWER BEARING OIL COOLING COIL A, to maintain Fl-3-628 between 5 and 10 gpm.

• ADJUST 3-724A, ISO VLV FOR COW FROM UPPER BEARING OIL COOLING COIL A, to maintain FI-3-629 between 155 and 170 gpm for Normal Operations OR between 138 and 145 gpm with RHR in service.

Procedure No.. Procedure

Title:

Page:

Approval Date:

3-ONOP-041.1 Reactor Coolant Pump Off-Normal 3/29/11 STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

CAUTION Care must be exercised when throttling HCV-3-121 in the closed direction.

Throttling this valve completely closed can cause the Charging Pump discharge relief valve to lift resulting in a possible loss of charging if the relief valve falls to reseat.

j4 Check Proper Seal Injection Flow

a. Verify at least one charging pump - a. Start a charging pump.

RUNNING

b. Check Annunciator A 6/6, SEAL WATER INJ b. IF standby seal injection filter available for use, FILTER HI AP ALARM OFF THEN place it in service using 3-OP-047, CVCS CHARGING AND LETDOWN

c. Check seal injection flow between 6 and 13 c. Perform the following:

gpm to each RCF

1) Adjust Charging Flow to Regan Heat Exchanger, HCV-3-121, to establish seal lnjection flow of 6 to 13 gpm per RCP.

2) IF unable to establish 6 to 13 gpm using HCV-3-121, THEN locally throttle open affected RCP Seal injection throttle valves:

3-297A for RCP A 3-297B for RCP B 3-297C for RCP C

3) IF seal injection flow of 6 to 13 gpm to each RCP is established, THEN observe NOTE prior to Step 15 AND go to STEP 15.

4) IF unable to establish 6 to 13 gpm flow, THEN adjust Charging Flow to Regan Heat Exchanger, HCV-3-121, to provide proper charging flow.

5) Record time of loss seal injection:

6) WHEN time since loss of seal injection equals 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, THEN perform 3-GOP-103, POWER OPERATION TO HOT STANDBY, to shut down Unit 3 AND stop the affected RCP within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of time recorded above. Return to Step 3.

\A19fl1 fl//h1rIc/rI

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 025 AK2.03 Importance Rating 2.7 Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: Service water or closed cooling water pumps Proposed Question: RO Question # 44 Given the following conditions:

• Unit 3 is in MODE 4 in 3-GOP-305, Hot Standby to Cold Shutdown.

• Unit 3 is cooling down with the 3B RHR loop in accordance with 3-OP-050, Residual Heat Removal System.

• 3A CCW Heat Exchanger is out of service for maintenance with required pumps in pull-to-lock.

Subsequently:

• The 3B CCW pump breaker trips on motor overload, causing an electrical transient resulting in a momentary loss of the 3B 4KV Bus.

• 3B 4KV Bus power is restored on the 3B EDG.

Which ONE of the following completes the statement below?

Shutdown cooling is restored when 3B RHR pump (1) and 3C CCW Pump (2)

A. (1) auto starts from the sequencer (2) auto starts from the sequencer B. (1) is manually started (2) auto starts from the sequencer C. (1) auto starts from the sequencer PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION (2) is manually started D. (1) is manually started (2) is manually started Proposed Answer: D Explanation (Optional):

A. Incorrect. 1st part wrong. 2id part wrong. Plausible to believe both RHR and COW pumps will start on the sequencer since the sequencer will have actuated on loss of power to the 3B 4KV bus. However, there is no SI signal, so the RHR pumps will not have started and with one COW Heat Exchanger out for maintenance, both 3A and 3C COW pumps are in pull to lock. Both 3B RHR and 30 COW pumps will be started manually.

B. Incorrect. part right. 2 d

part wrong. Plausible to believe that either RHR or COW pumps will start on the sequencer since the sequencer will have actuated on loss of power to the 3B 4KV bus. However, there is no SI signal, so the RHR pumps will not have started and with one COW Heat Exchanger out for maintenance, both 3A and 30 COW pumps are in pull to lock. Both 3B RHR and 3C CCW pumps will be started manually.

C. Incorrect. 1 part wrong. 2 part right. Plausible to believe that either RHR or COW pumps will start on the sequencer since the sequencer will have actuated on loss of power to the 3B 4KV bus. However, there is no SI signal, so the RHR pumps will not have started and with one COW Heat Exchanger out for maintenance, both 3A and 30 COW pumps are in pull to lock. Both 3B RHR and 3C COW pumps will be started manually.

D. Correct. There is no SI signal, so the RHR pumps will not have started and with one COW Heat Exchanger out for maintenance, both 3A and 30 COW pumps are in pull to lock. Both 3B RHR and 30 COW pumps will be started manually.

Technical Reference(s): 3-OP-050 Rev 13 (Attach if not previously provided) 3-ONOP-050 Rev 4A Proposed References to be provided to applicants during examination: N Learning Objective: LP69O2121AObj.9 (As available)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

3.0 PREREOUISITES NOTE Most prerequisites for operation of the RHR System depend upon the mode in which the

• system is operated and will be established by the applicable General Operation Procedure.

a a a a a a 1 3.1 The RHR System valve/breaker alignment has been verified at cold shutdown, by completion of the following attachments.

3.1.1 Attachment2 3.1.2 Attachment 3 3.1.3 Attachment4 3.1 .4 Attachment 5 3.2 All instruments and control devices are in service for the Residual Heat Removal System operation with no surveillance required, outstanding PWOs, clearances, or Temporary System Alterations that affect system operability as per the following:

3.2.1 O-OSP-200.1. Schedule of Plant Checks and Surveillances. (No surveillances have exceeded the date required on the missed surveillance sheet.)

3.2.2 Temporary System Alteration (TSA) Log 3.2.3 Clearance Log 3.2.4 Out-of-Service Log 3.3 The alternate shutdown panel alignment has been verified by satisfactory completion of 3-NOP-300, Alternate Shutdown Panel, for the 3B RHR Pump.

3.4 One of the following conditions exist prior to valving in CCW to both RI-IR Heat Exchangers: [Commitment Step 2.3.8]

3.4.1 Aii CCW Heat Exchangers are operable AND in service, OR 3.4.2 TWO CCW Heat Exchangers are operable AND in service AND two CCW Pumps are maintained in PULL-TO-LOCK.

W97:/JC/ab/ablcls

4.8 When RHR is in operation for cooldown, the following applies for CCW System operation:

4.8.1 The 3A or 3C CCW pump powered from the A 4KV bus is required to be operable to support operability of the RHR Loop A when required to be operable.

4.8.2 The 3B or 3C CCW pump powered from the B 4KV bus is required to bej operable to support operability of the RHR Loop B when required to be operable.

4.8.3 If only two CCW Heat Exchangers are in service and both RHR Loops are required, operation of the CCW System is restricted to only one operating CCW pump and the remaining CCW pumps are to be placed in PULL-TO-LOCK. The operating CCW pump shall be capable of being powered from an operable Emergency Diesel Generator. The second CCW Pump is considered functional for the second RHR Loop and auto-start of the second pump by the sequencer is not required. [Commitment Step 2.3.8]

4.8.4 At least one CCW Heat Exchanger is required for each RHR loop required to be operable. [Commitment Step 2.3.3]

4.8.5 If CCW is required to be isolated to a RHR Hx while RHR is in service and RCS temperature is greater than 180°F, the Rl-IR side of the Heat Exchanger shall be isolated first.

4.8.6 If necessary for temperature control, either MOV-3-749A OR MOV-3-749B may be closed without considering the associated RHR loop or Heat Exchanger inoperable so long as the following conditions are met:

I. RCS temperature is less than 1 80°F.

2. CCW flow for each operating CCW pump remains above the minimum flow value of 3,235 gpm.

3. An appropriate administrative control is placed on the control switch for the MOV that is closed (i.e., caution tag, ECO.).

4.8.7 If throttling of the CCW side of the RHR Hx is required to limit a cooldown due to problems with FCV-3-605 or HCV-3-758, the following apply:

I. Only MOV-3-749A or MOV-3-749B shall be used to prevent invalidating the CCW System flow balance.

2. A minimum of 1,000 gpm CCW flow as indicated on ultrasonic flow instrument shall be maintained for the Heat Exchanger with reduced flow.

3. The CCW Surge Tank shall be monitored to prevent overflowing the surge tank due to the expansion of water in the RHR Hx.

4. The System Engineer is aware of the need to change CCW valve positions.

4.8.8 In order to minimize CCW Heat Exchanger Tube degradation, the manufacturers maximum recommended flow limit for long term operation through each CCW Heat Exchanger is 6,840 gpm. When RHR is placed in service, the total flow on the CCW Headers should be maintained below 13,680 gpm to ensure this limit is not exceeded when a CCW Heat Exchanger is removed from service.

[Commitment Step 2.3.8]

NQ7L IflIh/hfrk

Procedure No.: Procedure

Title:

Page:

13 Approval Date:

3-OP-050 Residual Heat Removal System 3/15/14 5.0 STARTUP/NORMAL OPERATION 5.1 Preparation for Placing RFIR in Operation for Cooldown INIT Date/Time Started: /

CAUTION To prevent exceeding the flow restrictions on the CCW Heat Exchangers when two RHR Heat Exchangers are valved in, THREE CCW Heat Exchangers are required to be in seivice. If this condition can NOT be met, two CCW Pumps are required to be placed in PULL-TO-LOCK and the associated LCO in accordance with Technical Specifications 3/4.7.2 entered. (Commitment Step 2.3.8)

NOTE I I I Preparation steps may be performed in any order at Shift Managers discretion without I changing intent or order of system lineups.

I I 5.1.1 Initial Conditions

1. All applicable prerequisites listed in Section 3.0 are satisfied.

2. The RCS temperature is less than 375°F.

3. Component Cooling water flow to RHR Pump Seal Water Heat Exchangers is at least l0gpm.

4. One of the following conditions exist: [Commitment Step 2.3.8]

a. IF three CCW Heat Exchangers are operable AND in service, THEN verify two CCW Pumps are in service OR

b. only two CCW Heat Exchangers are operable and in service, THEN verify one CCW Pump in service AND two CCW Pumps shall be placed in PULL-TO-LOCK.

W97:/JC/ab/ab/cls

I STEP1 IACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I CAUTION RCS Cooldown Rate shall be maintained LESS than 90 degrees per hour.

6 Establish Conditions For Restarting An RHR Pump

a. RHR pumps BOTH STOPPED a. GotoStep7.

b. Close RHR Heat Exchanger Outlet Flow valve, HCV-3-758

c. Close RHR Heat exchanger Bypass Flow valve, FCV-3-605

d. Verify MOV-3-750 and MOV-3-751 d. GotoStepli.

OPEN

e. Start the previously running RHR pump e. Start the Standby RHR pump.

1) IF neither RHR pump can be restarted, THEN perform the following:

a) Direct appropriate personnel to restore at least one RHR pump to operable status.

b) GotoStepli.

f. Return RHR Heat Exchanger Bypass Flow valve, FCV-3-605, to AUTOMATIC operation increasing flow in increments of 500 gpm until desired flow is established

g. Open RHR Heat Exchanger Outlet Flow valve, HCV-3-758, as necessary to maintain desired RCS temperature A1a71rj-u,Io!ni,,i,,,,

STEP ACTION!EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

CAUTION CCW System load requirements of 3-NOP-030, COMPONENT COOLING WATER SYSTEM, shall NOT be exceeded.

20 Maintain Proper CCW System Alignment for RCP Operation

a. CCW Heat Exchangers THREE IN

a. Perform the following:

SERVICE

1) Start or stop CCW pumps as necessary to establish ONLY ONE RUNNING CCW PUMP.

2) IF MOV-3-749A and MOV-3-749B are open, THEN stop and place in PULL-TO-LOCK all except one gunning CCW pump.

3) Go to Step 20d)

b. CCW pumps ONLY TWO RUNNING

b. Start or stop CCW pumps as necessary to establish ONLY TWO RUNNING CCW PUMPS.

c. Check CCW from RHR Heat Exchangers - c. Perform the following:

AT LEAST ONE CLOSED

1) Isolate one Emergency Containment

• MOV-3-749A Cooler by placing one ECC Control

• MOV-3-749B Switch in STOP AND go to Step 20d.

2) IF unable to isolate one ECC, THEN stop all RCP5 AND verify natural circulation using ATTACHMENT 1.

3) Go to Step 24.

d. Verify B CCW header flow NORMAL

d. Perform the following:

1) Verify natural circulation using ATTACHMENT 1. IF natural circulation can NOT be verified, THEN increase dumping steam.

2) Go to Step 24.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 027 AA1.01 Importance Rating 4.0 Ability to operate and I or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions: PZR heaters, sprays, and PORVs Proposed Question: RO Question # 45 Given the following conditions:

• Unit 4 is at 80% power.

• 4B Steam Generator Feed Pumps breaker trips.

Subsequently

• PC-4-444J, Pressurizer Pressure Controller, does not respond in automatic.

Which ONE of the following identifies (1) the operators required response and (2) the expected setpoint for PORV response?

PC-4-444Js output is initially (1) to mitigate the pressure transient. At (2) demand on PC-4-444J, PORV PCV-4-455C is expected to open.

A. (1) raised (2) 42.5%

B. (1) raised (2) 92%

C. (1) lowered (2) 42.5%

D. (1) lowered (2) 92%

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part right. 2nd part wrong. Plausible to remember the normal demand output on PC-4-444J at 2235 psig as the open demand signal for PCV-4-455C.

B. Correct. Spray Valves and Heaters operate to the demand signal from PC-4-444J.

Spray Valves start to open at 50% demand and are full open at 75%. B/U Heaters are ON at 30% and OFF at 34%. Control Group Heaters are ON at 35% and OFF at 50%.

PC-4-444J sends a demand signal to PCV-4-455C, which is closed at 82.5% and full open at 92.5%. PORV-456 is not affected, since its signal is generated by pressure from PT-4-445. Setting the Spray Valve control station to MANUAL bypasses the signals from PC-4-444J. Setting the Pressurizer Heater switches to ON or OFF bypasses the signals from PC-4-444J. When the switch for PCV-4-455C is changed from AUTO to CLOSE or OPEN, the PC-4-444J demand signal is bypassed.

C. Incorrect. 1st part wrong. 2nd part wrong. Plausible to incorrectly remember how the demand signal from PC-4-444J affects pressurizer pressure control. Other controllers nearby on the console have demand signals that will either open or close valves on an increasing demand signal. Plausible to remember the normal demand output on PC 444J at 2235 psig as the open demand signal for PCV-4-455C.

D. Incorrect. 1St part wrong. 2nd part right. Plausible to incorrectly remember how the demand signal from PC-4-444J affects pressurizer pressure control. Other controllers nearby on the console have demand signals that will either open or close valves on an increasing demand signal.

Technical Reference(s): 5610-T-D-16B Rev 18 (Attach if not previously provided) 4-ONOP-041 .5 Rev OA Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902 109A Obj. 3, 9 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

Procedure No. Procedure

Title:

Page:

Approval Date:

4-ONOP-041.5 Pressurizer Pressure Control Malfunction 12/17/07 STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

0 Foldout page is required to be monitored throughout this procedure. I S S S a I CAUTION The Master Controller should be operated carefully (Normal controller output for 2235 psig is 425 percent demand; 92 percent demand will open PCV-4-455C). If the following conditions are met, an excessive increase in controller output could cause Power Operated Relief Valve PCV-4-455C to open:j

1. PCV-4-455C hand switch in AUTO.

? Pressurizer pressure is greater than or equal to 2000 psig, or OMS switch in LO Press Ops.

I Check PZR Pressure Control Instrument Loop Not Failed

a. Check PT-4-444 NOT FAILED by

a. Perform the following:

comparison with adjacent pressure channels and known plant parameters 1) Verify PCV-4-455C Q MOV-4-536 CLOSED.

2) Take manual control of PC-4-444J, PZR PRESS CONTROL5

3) IF manual control of PC-4-444J is NOT effective, THEN perform the following:

Take manual control of PZR spray valves.

Take manual control of PZR heaters.

b. Check PT-4-445 NOT FAILED by

b. Perform the following:

ters 1) Verify PCV-4-456 OR MOV-4-535 W97:TNM/bc/nwfcls

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 029 EA1.03 Importance Rating 3.5 Ability to operate and monitor the following as they apply to a ATWS: Charging pump suction valves from VCT operating switch Proposed Question: RO Question # 46 Given the following conditions:

• Unit 3 experiences an Anticipated Transient without Scram (ATWS).

• 3-EOP-FR-S.1, Response To Nuclear Power GenerationlATWS is in progress.

• The RCO is initiating Emergency Boration.

• Neither 3A nor 3B Boric Acid Pumps starts.

Which ONE of the following describes the required response to initiate Emergency Boration?

A. Open MOV-3-350, Emergency Boration Valve.

B. Close LCV-3-1 1 5C, VCT Outlet to Charging Pump Suction.

C. Close FCV-3-113B, Blender to Charging Pump.

D. Open 3-356, Manual Emergency Boration Valve.

Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible since this is the first valve that would be opened lAW 3-EOP-FR-S.1, Emergency Boration, EXPECTED RESPONSE. However, with the Boric Acid pumps not starting, closure of Close LCV-3-1 1 5C is the correct response.

B. Correct. To initiate Emergency Boration under the given conditions, LCV-3-1 1 5C, VCT Outlet to Charging Pump Suction must be closed and then de-energized. This will cause LCV-3-1 15B, RWST Outlet to Charging Pump Suction, to open.

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Incorrect. Plausible since this is the valve that would be opened lAW 3-EOP-FR-S.1, Emergency Boration, RNO step 4.e when MOV-3-350 failed to open. However this step would be skipped when the Boric Acid pumps failed to start.

D. Incorrect. Plausible since this valve that would be opened lAW 3-EOP-FR-S. 1, Emergency Boration, RNO step 4.e.3. However this step would be skipped when the Boric Acid pumps failed to start.

Technical Reference(s): 3-EOP-FR-S.1 Rev 4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902113 Obj. 5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

RESPONSE TO NUCLEAR POWER GENERATION/ATWS 7 of 22 PROCEDURE NO.:

3-EOP-FR-S.1 TURKEY POINT UNIT 3 STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

4. Initiate Emergency Boration Of RCS:

a. Verify SI RESET

b. Verify Charging Pumps AT LEAST ONE RUNNING IN MANUAL

c. Stop Makeup System

d. Manually start Boric Acid Pump 3A or d. Align Charging Pump suction to the 3B RWST as follows:

1) Hold closed LCV-3-1 I 5C Control switch.

2) Direct an operator to open Breaker 30669 for LCV-3-1 150.

3) WHEN 30669 is open, THEN release LCV-3-115C Control switch.

4) GotoStep4.f.

e. Open MOV-3-350, Emergency e. Perform the following:

Boration Valve

1) Open FCV-3-1 13A, Boric Acid To Blender.

2) Open FCV-3-1 13B, Blender Flow To Charging Pump.

3) Locally open 3-356, Manual Emergency Boration Valve.

4) WHEN 3-356, Manual Emergency Boration Valve is open, THEN close FCV-3-113B, Blender To Charging Pump.

5) Continue with Step 4.f.

f. Open HCV-3-121, Charging Flow To Regen Heat Exchanger

PTN L-15-1 DRAFT NRC EXAM SECURE )NFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 038 EK3.08 Importance Rating 4.1 Knowledge of the reasons for the following responses as the apply to the SGTR: Criteria for securing RCP Proposed Question: RO Question # 47 Given the following conditions:

• A SGTR occurs on 3B SIG.

• 3-EOP-E-3, Steam Generator Tube Rupture is in progress.

• A cooldown is commenced to target temperature.

Subsequently:

• RCS Subcooling lowers to 14°F.

Which ONE of the following actions must be taken and why?

A. Trip the Reactor Coolant Pumps to minimize the potential for RCP damage when an RCS depressurization is initiated.

B. Trip the Reactor Coolant Pump on the affected loop to minimize RCS inventory loss.

C. Keep the Reactor Coolant Pumps running to prevent the automatic opening of the SDTA.

D. Keep the Reactor Coolant Pumps running because a controlled RCS cooldown is in progress.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible since RCP trip criteria is met and normally it would be reasonable to trip the RCPs. However, 3-EOP-E-3, FOP, RCP Trip Criteria requires that a cooldown is NOT initiated.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION B. Incorrect. Plausible since RCP trip criteria is met and normally it would be reasonable to to trip the RCPs. Tripping RCPs for inventory loss is done for other accidents such as SBLOCA. However, 3-EOP-E-3, FOP, RCP Trip Criteria requires that a cooldown is NOT initiated.

C. Incorrect. Plausible since RCP trip criteria is met and normally it would be reasonable to to trip the RCPs. The SDTA is mentioned many times in the first six steps of the procedure such that it is plausible to remember that as a reason for tripping the RCPs once the criteria is met. However, 3-EOP-E-3, FOP, RCP Trip Criteria requires that a cooldown is NOT initiated.

D. Correct. 3-EOP-E-3, FOP, RCP Trip Criteria requires that a cooldown is NOT initiated.

Technical Reference(s): 3-EOP-E-3 Rev 7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902339 Obj. 5f (As available)

Question Source: Bank# 101448 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Point Beach Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

STEAM GENERATOR TUBE RUPTURE FOLDOUT PROCEDURE NO.:

3-EOP-E-3 TURKEY POINT UNIT 3 FOLDOUT PAGE For Procedure 3-EOP-E-3

1. ADVERSE CONTAINMENT CONDITIONS

a. IF either condition listed below occurs, THEN use [Adverse Containment Setpoints]:

Containment atmosphere temperature 180°F OR Containment radiation levels 1.3x10 5 R/hr

b. WHEN Containment atmosphere temperature returns to less than 180°F, THEN Normal Setpoints can again be used.

c. WHEN Containment radiation levels return to less than 1.3x105 R/hr, THEN Normal Setpoints can again be used [f the TSC determines that Containment Integrated Dose has NOT exceeded Rads.

2. RCP TRIP CRITERIA

a. f all conditions listed below occut THEN trip fl RCPs:

• High-Head SI pumps AT LEAST ONE RUNNING AND SI FLOWPATH VERIFIED

• RCS Subcooling LESS THAN 19°F[41°F]

• Controlled RCS cooldown NOT initiated

b. IF Phase B actuated, THEN trip a RCPs.

3. SI RE-INITIATION CRITERIA IF either condition listed below occurs after Section 3.0, Step 21, THEN manually start SI Pumps as necessary to restore RCS subcooling and PRZ level and go to 3-EOP-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT SUBCOOLED RECOVERY DESIRED, Step 1:

RCS Subcooling based on Core Exit TCs LESS THAN 1 9°F[73°Fj OR PRZ level CAN NOT BE MAINTAINED GREATER THAN 7%[48%]

4. SECONDARY INTEGRITY CRITERIA IF any S/G pressure is decreasing in an uncontrolled manner OR has completely depressurized, AND that S/G has NOT been isolated, AND is NOT needed for RCS cooldown, THEN go to 3-EOP-E-2, FAULTED STEAM GENERATOR ISOLATION, Step 1.

5. COLD LEG RECIRCULATION SWITCHOVER CRITERIA IF RWST level decreases to less than 155,000 gallons, THEN go to 3-EOP-ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1.

6. CST MAKEUP WATER CRITERIA IF CST level decreases to less than 12%,

THEN add makeup to CST using 3-NOP-Ol 8.01, CONDENSATE STORAGE TANK (CST).

7. MULTIPLE TUBE RUPTURE CRITERIA IF, after identification of a ruptured S/G, y intact S/G level increases in an uncontrolled manner OR any intact S/G has abnormal radiation, THEN stabilize the plant and return to 3 -EOP-E-3, STEAM GENERATOR TUBE RUPTURE, Step 1.

8. LOSS OF OFFSITE POWER OR SI ON OTHER UNIT IF SI has been reset AND subsequently either offsite power is lost OR SI actuates on the other unit, THEN restore safeguards equipment and at least one Computer Room Chiller to required configuration.

Refer to Attachment 3 for essential loads.

9. RHR SYSTEM OPERATION CRITERIA IF RHR flow is less than 1100 gpm, THEN the RHR Pumps shall be shut down within 44 minutes of the initial start signal.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 055 2.2.38 Importance Rating 3.6 Equipment Control: Knowledge of conditions and limitations in the facility license.

Proposed Question: RO Question # 48 Given the following conditions:

• Unit 3 is at 100% power.

• 3A EDG is out of service for maintenance.

• A fault on the Unit 3 Startup Transformer generates a Transformer Lockout and a 3B 4KV Bus Lockout which do not reset.

• 3-EOP-ECA-0.0, Loss of All AC Po,er, is in progress.

Which ONE of the following statements identifies (1) the preferred source of power restoration and (2) the design basis battery duration?

A. (1) Unit 4 through SBO Tie (2) 30 minutes B. (1) Unit 4 through SBO Tie (2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. (1) Unit 4 Startup Transformer (2) 30 minutes D. (1) Unit 4 Startup Transformer (2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Proposed Answer: D PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMAT)ON Explanation (Optional):

A. Incorrect. 1 st part wrong. 2 nd part wrong. Plausible to believe since the SBO tie is a method used to re-energize the 3A 4KV bus, however, the Unit 4 Startup Transformer is first in the procedure and preferred over the SBO tie. Plausible to believe 30 minutes since power needs to be restored to at least one 4KV bus (3A OR 3B) within 10 minutes to satisfy station blackout requirements and that the 30 minutes is the design basis for the station batteries. However, in accordance with BD-EOP-ECA-0.0, Step 12 caution provides the basis for battery capacity stating that Batteries 3A, 4A, and Spare can carry their expected emergency loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a loss of all AC power event with the Batteries starting out at 80% of their design capacity.

B. Incorrect. 15t part wrong. 2 part right. Plausible to believe since the SBO tie is a method used to re-energize the 3A 4KV bus, however, the Unit 4 Startup Transformer is first in the procedure and preferred over the SBO tie.

C. Incorrect. 1 st part right. 2 part wrong. Plausible to believe 30 minutes since power needs to be restored to at least one 4KV bus (3A OR 3B) within 10 minutes to satisfy station blackout requirements and that the 30 minutes is the design basis for the station batteries. However, in accordance with BD-EOP-ECA-0.0, Step 12 caution provides the basis for battery capacity stating that Batteries 3A, 4A, and Spare can carry their expected emergency loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a loss of all AC power event with the Batteries starting out at 80% of their design capacity.

D. Correct. The Unit 4 Startup Transformer is first in the procedure and preferred over the SBO tie. In accordance with BD-EOP-ECA-0.0, Step 12 caution provides the basis for battery capacity stating that Batteries 3A, 4A, and Spare can carry their expected emergency loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a loss of all AC power event with the Batteries starting out at 80% of their design capacity.

Technical Reference(s): 3-EOP-ECA-0.0 Rev 7 .

BD-EOP-ECA-0.0 Rev 7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Learning Objective: LP 602138 Obj 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

LOSS OF ALL AC POWER 58 of 88 PROCEDURE NO.:

3-EOP-ECA-O.O TURKEY POINT UNIT 3 ATTACHMENT 6 3A 4KV Bus Restoration (Page 2 of 7)

STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

4. Energize 3A 4KV Bus From Unit 4 Startup Transformer:

a. Check Unit 4 Startup Transformer a. Observe NOTE prior to Potential white light ONr Attachment 6, Step 5 and go to Attachment 6, Step 5.

b. Locally unlock and rack in 3AA22, b. Observe NOTE prior to 3A 4KV Bus Emergency Tie To Unit 4 Attachment 6, Step 5 and Startup Transformer go to Attachment 6, Step 5.

c. Close 3AA22, 3A 4KV Bus c. Locally close breaker.

Emergency Tie To Unit 4 Startup Transformer

d. Check 3A 4KV Bus ENERGIZED d. Observe NOTE prior to Attachment 6, Step 5 and go to Attachment 6, Step 5.

e. Maintain loading on the opposite unit Startup Transformer Tie Line less than 600 amps

f. Go to Attachment 6, Step 15

REVISION NO.: PROCEDURE TITLE: PAGE:

LOSS OF ALL AC POWER 59 Of 88 PROCEDURE NO.:

3-EOP-ECA-O.O TURKEY POINT UNIT 3 ATTACHMENT 6 3A 4KV Bus Restoration (Page 3 of 7)

I STEi] I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I NOTE Power needs to be restored to at least one 4KV bus (3A OR 3B) within 10 minutes to satisfy station blackout requirements.

5. Determine If Station Blackout Tii Line Go to Attachment 6, Step 14.

Is Available:

• Check3B4KVBus DE-ENERGIZED

• Check at least one of the following ENERGIZED:

4A4KVBus 4B4KVBus

6. Check 3D 4KV Bus Lockout Relay Reset 3D 4KV Bus Lockout Relay.

RESET 3D 4KV Bus lockout relay can NOT be reset, THEN go to Attachment 6, Step 14.

7. Check 3D 4KV Bus Perform the following:

ALIGNED TO 3A 4KV BUS:

• 3ADO1, Supply From 4KV Bus 3A a. Open 3AB19, Feeder To 4KV Bus 3D.

CLOSED b. Open 3AD06, Supply From 4KV Bus 3B

• 3AA17, FeederTo 4KV Bus 3D CLOSED c. Close 3ADO1, Supply From 4KV Bus 3A.

d. Close 3AA17, Feeder To 4KV Bus 3D.

e. 3D 4KV Bus can NOT be aligned to 3A 4KV Bus, THEN go to Attachment 6, Step 14.

Page 41 BD-EOP-ECA-0.O Loss of All AC Power 8/15/14 BASIS DOCUMENT WOG Procedure Step N/A PTN Procedure Step 12 CAUTION I Attachment 3, Step 1 is required to be performed within the first 60 minutes of a loss of all AC power event if both the *A1 and *A2 Battery Chargers are inoperable.

BASIS:

Batteries *A and Spare will sTteir connected shutdown loads for two hours when they are at 80% of rated capacity, provided the non-essential DC loads listed on Attachment 3, Step I are de-energized within the first 60 minutes of the duty cycle (i.e. within the first 60 minutes of the loss of all AC power event).

STEP DEVIATIONS FROM WOG GUIDELINES:

TYPE DESCRIPTION 3 This caution was added to address a plant specific concern with 3A, 4B, and Spare Batteries sizing criteria not covered by the WOG guideline. Refer to Safety Evaluation JPN-PTN-SEEJ-91 -013, Rev. I, for additional information.

PLANT SPECIFIC SETPOINTS:

60 minutes Time required by the Safety Evaluation JPN-PTN-SEEJ-91-013 to ensure that Batteries 3A, 4A, and Spare can carry their expected emergency loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a loss of all AC power event with the Batteries starting out at 80% of their design capacity.

W201 O:/DH/cls/cls/cls

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Facility: Turkey Point Vendor: WEC Exam Date:

Exam Type: RO Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# 057 AA2.12 Importance Rating 3.5 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: PZR level controller, instrumentation, and heater indications Proposed Question: RO Question # 49 Given the following conditions:

• Unit 3 is at 3% power while performing a reactor startup.

• Vital Instrument Panel 3P06 loses power.

• 3-ONOP-003.6, Loss of 120V Vital Instrument Panel 3P06, is in progress.

• The US directs that Charging flow be reduced to the minimum required to maintain RCP Seal Injection.

Which ONE of the following completes the statements below?

Due to the loss of 3P06, the Pressurizer Control Heaters (1) 3-ONOP-003.6 directs the reduction of Charging flow to (2)

A. (1) remain on (2) reduce the PRZ fill rate to prevent lifting a PRZ PORV.

B. (1)de-energize (2) reduce the PRZ fill rate to prevent lifting a PRZ PORV.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. (1) remain on (2) reduce the potential for thermal shock of the RHX.

D. (1) de-energize (2) reduce the potential for thermal shock of the RHX.

Proposed Answer: B Explanation (Optional):

A. Incorrect. 1st part wrong. 2uid part right. Plausible since loss of 3P07 and 3P08 will result in only a possible loss of Pressurizer Heaters, while loss of 3P06 results in loss of Pressurizer Control and Backup Heaters.

B. Correct. Loss of 3P06 results in loss of Pressurizer Control and Backup Heaters. Basis document for step 3 of 3-ONOP-003.6 states that the basis for minimizing the fill rate of the Pressurizer is to extend the time frame for recovery without lifting a PRZ PORV due to compressing the bubble.

C. Incorrect. 1st part wrong. 2 part wrong. Plausible since loss of 3P07 and 3P08 will result in only a possible loss of Pressurizer Heaters, while loss of 3P06 results in loss of Pressurizer Control and Backup Heaters. Plausible to believe that since letdown is isolated, the heat exchanger will cool down too quickly, undergoing thermal shock.

However there are no cooldown restrictions for the regenerative heat exchanger.

D. Incorrect. 1st part right. 2 part wrong. Plausible to believe that since letdown is isolated, the heat exchanger will cool down too quickly, undergoing thermal shock.

However there are no cooldown restrictions for the regenerative heat exchanger.

Technical Reference(s): 3-ONOP-003.6 Rev. 1 (Attach if not previously provided)

BD-ONOP-003.6 Rev 0 Proposed References to be provided to applicants during examination:

Learning Objective: LP 6902260 Obj 5, 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

1.0 PURPOSE This procedure provides instructions to be followed upon receipt of Loss of I 20V Vital Instrument Panel 3P06.

2.0 SYMPTOMS OR ENTRY CONDITIONS 2.1 Indications 2.1.1 Power Range N-41 Failure (NIS Racks Channel I Lights Out) 2.1 .2 Loss of Channel I Vital Instrumentation/Indications 2.1.3 Transfer of Feedwater Control to Backup Controller for Steam Generator A 2.1 .4 Loss of Power to Pressurizer pressure control Auto/Manual Station (auto lockup) 2.1 .5 Loss of Power to the Pressurizer Spray Valve Auto/Manual Station (auto lockup) 2.1.6 Loss of Pressurizer Heaters (Control and Backup 2.1.7 Isolation of CVCS Letdown Flow 2.1 .8 Loss of Power to Pressurizer Level Auto/Manual Station (auto lockup) 2.1.9 Loss of Power to 3A Charging Pump Auto/Manual Station (auto lockup) 2.1.10 PCV-3-456 Auto Open Disabled (if in OMS LOW PRESSURE OPS) 2.2 Alarms 2.2.1 F 1/2, VITAL AC BUS INVERTER TROUBLE 2.2.2 B 6/5, POWER RANGE LOSS OF DETECTOR VOLTAGE 2.2.3 B 7/1, NIS/RPI ROD DROP ROD STOP 2.2.4 C 6/1, SG A LEVEL DEVIATION CNTRL TROUBLE 2.2.5 A 1/5 RCP SEAL LEAKOFF HI FLOW 2.2.6 A 6/4, RCP SEAL WATER LO DP 2.2.7 A 7/6. RCP C SEAL WATER BYPASS LO FLOW (if CV-3-307 Open)

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Page 6 BD-ONOP-003.6 Loss of 120V Vital Instrument Panel *P06 10/7/02 PROCEDURE STEPS CAUTION This caution alerts the operator to a potential automatic reactor trip.

I NOTES I

I I

• Immediate actions are those actions which the operator should be able to perform I before opening and reading the emergency procedures. Although the operator should memorize immediate actions, they need not be memorized verbatim. The I

operator should know them well enough to complete the intent of each step.

• Provides reminders that *P06 is the RED channel and that Enclosure I is available to determine what functions, indications, and controls are lost.

This step checks if a reactor trip has occurred. If a reactor trip has not occurred the operator is directed to check if a reactor trip is required. If a reactor trip is required, the operator is directed to manually trip the reactor and perform EOP-E-O concurrently. If a trip is not required, the operator is directed to the applicable procedure step.

2. This step is written to direct the operator to check for loss of RHR if the unit is not operating in MODES 1-3. Automatic RHR flow control is lost. MOV*75O fails closed when PC*4O3 loses power and PCV*142 fails closed which may lead to an RCS overpressure condition.

3. Ti loss of *P06 directly affects the normal control of pressurizer pressure and level.

Operator attention to the pressurizer is necessary to maintain pressure and level in normal ranges.

Minimizing the fill rate of the pressurizer will extend the time frame for recovery without lifting a PZR PORV due to compressing the bubble.

b. Power operated relief valve, PCV-455C, receives its control signal from pressure comparator. PC-444A. Upon loss of *P06, PC-444A output signal locks up as is, with the possibility of maintaining PCV-455C in the open position. For this reason, pressurizer PORVs should be verified shut to prevent inadvertent depressurization of the RCS.

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PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Facility: Turkey Point Vendor: WEC Exam Date:

Exam Type: RO Examination Outline Cross-reference: Level RD SRO Tier# 1 Group# 1 KIA# 058 AK1.01 Importance Rating 2.8 Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation Proposed Question: RO Question # 50 Given the following conditions:

• Unit 3 is in MODE 3.

• Vital 480V MCC 3B is out of service.

Subsequently:

• Vital DC Bus 3D23 loses power due to a fault on the bus.

• The crew is restoring power to DC Bus 3D23 in accordance with 3-ONOP-003.

5, Loss of DC Bus 3D23 and 3D23A (3B).

• The fault has been isolated.

Which ONE of the following identifies (1) the battery charger that is still OPER ABLE and (2) the expected battery charger voltage?

A. (1)3B1 (2) 120 volts B. (1)3B1 (2) 135 volts PTN L-15-2 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. (1)3B2 (2) 135 volts D. (1)3B2 (2) 120 volts Proposed Answer: C Explanation (Optional):

A. Incorrect 1 st part wrong. 2 r part wrong. Plausible need to recall/determine correct power supplies. 3B1 BC is powered from MCC 3B; 3B2 BC is powere d from MCC 4D, therefore 3B2 BC is available. Plausible since 120 volts is nomin al line voltage for AC Instrument Bus.

B. Incorrect 1 st part wrong. 2 part right. Plausible need to recall/determine correct power supplies and order of restoration. 3B1 BC is powered from MCC 3B; 3B2 BC is powered from MCC 4D, therefore 3B2 BC is available.

C. Correct Bi BC is powered from MCC 3B; 3B2 BC is powered from MCC 4D, therefore 3B2 BC is available. TS 4.8.2.1 OPERABILITY requires battery voltag e 129 volts and 3-ONOP-003.5 states to check voltage between 131 to I4OVDC on the charager.

D. Incorrect 1 st part right. 2 part wrong. Plausible since 120 volts is nominal line voltage for AC Instrument Bus.

TS 4.8.2.1 Technical Reference(s): 3-ONOP-003.5 Rev. 2A (Attach if not previously provided) 5610-T-E-1592 Sh I Rev 45 Proposed References to be provided to applicants during examination:

N Learning Objective: 6900253 E08 (As available)

Question Source: Bank# 93366 Modified Bank # (Note changes or attach parent)

New PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question History: Last NRC Exam: 2010 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

D.C. SOURCES LIMITING CONDITION FOR OPERATION ACTION: (Continued)

b. With one of the required battery banks inoperable, or with none of the full-capacity chargers associated with a battery bank OPERABLE, restore all battery banks to OPERABLE status and at least one charger associated with each battery bank to OPERABLE status within two hours*

or be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This ACTION applies to both units simultaneously.

SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and its associated full capacity charger(s) shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1) The parameters in Table 4.8-2 meet the Category A limits, and

2) The total battery terminal voltage is greater than or equal to 129 volts on float charge and he battery charger(s) output voltage is 129 volts, and

3) If two battery chargers are connected to the battery bank, verify each battery charger is supplying a minimum of 10 amperes, or demonstrate that the battery charger supplying less than 10 amperes will accept and supply the D.C. bus load independent of its associated battery charger.

b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 105 volts (108.6 volts for spare battery D-52), or battery overcharge with battery terminal voltage above 143 volts, by verifying that:

1) The parameters in Table 4.8-2 meet the Category B limits,

2) The average electrolyte temperature of every sixth cell is above 60°F, and

3) There is no visible corrosion at either terminals or connectors, or verify battery connection resistance is:

Battery Connection Limit (Micro-Ohms) 3B, 4A inter-cell I termination < 29 inter-cell (brace locations) < 30 transition cables < 125 or total battery connections < 1958 Battery Connection Limit (Micro-Ohms) 3A, 4B, D-52 inter-cell I termination < 35 inter-cell (brace locations) < 40 transition cables < 125 or total battery connections 2463

c. At least once per 18 months by verifying that:

1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,

• Can be extended to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the oppsite unit is in MODE 5 or 6 and each of the remaining required battery chargers is capable of being powered from its associated diesel generator(s).

TURKEY POINT UNITS 3 & 4 3/4 8-14 AMENDMENT NOS. 252 AND 248

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMA11ON Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# 062 2.4.4 Importance Rating 4.5 Emergency Procedures / Plan: Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

Proposed Question: RO Question # 51 Given the following conditions:

• Unit 3 is operating at 95% power.

• ICW Header pressure is 8 psig on PI-3-1619 and P1-3-i 620.

• CWP Discharge pressure is 8 psig on all four Circ Water Pumps.

• TPCW Discharge pressure is 98 psig P1-3-1468.

• CCW Header Supply pressure is 105 psig on P1-3-640.

Which ONE of the following procedures is entered for required response?

A. 3-ONOP-019, Intake Cooling Water Malfunction B. 3-ONOP-030, Component Cooling Water Malfunction C. 3-ONOP-008, Turbine Plant Cooling Water Malfunction D. 3-ONOP-Ol 1, Screen Wash System/Intake Malfunction Proposed Answer: A Explanation (Optional):

A. Correct. ICW Header pressure of 8 psig indicates an ICW malfunction, normal is 16 psig.

B. Incorrect. Plausible since CCW pressure is lower than normal but still within the operating band.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Incorrect. Plausible since TPCW pressure is lower than normal but still within the operating band.

D. Incorrect. Plausible since CWP pressure is higher than normal but still within the operating band.

• 3-ONOP-019 Rev 2 Technical Reference(s): *

(Attach if not previously provided) 3-ARP-097.CR.l Rev 11 Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902154 Obj. 10 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

1.0 PURPOSE 1.1 This procedure provides instructions to be followed in the event of a malfunction or failure in the Intake Cooling Water (ICW) System.

2.0 SYMPTOMS OR ENTRY CONDITIONS 2.1 Visible evidence of excessive system leakage.

2.2 Catastrophic failure due to passage of heavy loads across system piping or electrical ducts.

2.3 Annunciators

2.3.1 H 8/5, CCW HX OUTLET HI TEMP 2.3.2 I 4/I, ICWP A/B/C MOTOR OVERLOAD 2.3.3 1 4/2, ICWP A/B/C TRIP 2.3.4 1 4/3, ICWP A/B/C MOTOR BRG HI TEMP 2.3.5 14/4, ICW HEADER A/B LO PRESSj 2.3.6 1 5/4, TPCW HI TEMP/LO PRESS

3.0 REFERENCES

/RECORDS REOUIRED/COMMITMENT DOCUMENTS 3.1 References 3.1.1 Technical Specifications

1. Section 3.7.3, Intake Cooling Water System 3.1.2 FSAR

1. Section 9, Auxiliary and Emergency Systems 3.1.3 Plant Drawing I. 5613-M-3019, Sht 1 and Sht2, Intake Cooling Water System 3.1.4 Procedures I. 3-GOP-103, Power Operation to Hot Standby

2. 3-ONOP-008, Turbine Plant Cooling Water Malfunction

3. 3-ONOP-Ol I, Screen Wash System/Intake Malfunction

4. 3-ONOP-030, Component Cooling Water Malfunction

5. 3-ONOP-100, Fast Load Reduction

6. 3-NOP-0 13, Instrument Air System

7. 3-NOP-019, Intake Cooling Water System W97:IJBS/In/mr/In

REVISION NO.: PROCEDURE TITLE: PAGE:

11 27 CONTROL ROOM RESPONSE - PANEL I PROCEDURE NO.: WINDOW:

3-ARP-097.CR.l TURKEY POINT UNIT 3 (Page 1 of 1)

CAUSES: 1. Leak in ICW System

2. Trip of a running ICW pump 1414 ICw HEADER AIB LO PRESS DEVICE: SETPOINT: LOCATION:

• PS-3-1619 (A HDR) 10 psig N/A

• PS-3-1620 (B HDR)

ALARM CONFIRMATION

1. CHECK lOW header pressure indicators, P1-3-1619 or 3-1620 less than or equal to 10 psig on VPA.

2. IF operating a single lOW Pump, THEN CHECK total ICW flow is less than 18,500 gpm.

OPERATOR ACTIONS

1. START standby ICW 5TJmp using 3-NOP-Ol 9, Intake Colin Water Sistem.

2. Locally CHECK lOW piping and heat exchangers for leaks.

3 REFER TO 3-ONOP-019, Intake Cooling Water MalfuncticIj

4. IF operating a single ICW Pump AND total lOW flow is greater than 18,500 gpm, THEN immediately REDUCE total lOW flow by performing the following:

A. THROTTLE TPCW Combined Outlet Valve, 3-50-401, while maintaining TPCW Hx outlet temperature less than 105°F.

B. THROTTLE 3-50-406, CCW HX lOW OUTLET SPOOL PIECE BYPASS and 3 407, CCW HX lOW OUTLET SPOOL PIECE ISOL while maintaining minimum ICW flows through CCW Hxs as determined by 3-NOP-019, Intake Cooling Water System.

5. IF unable to reduce total ICW flow through a single ICW Pump to less than 18,500 gpm, THEN REDUCE unit load using 3-GOP-103, Power Operation to Hot Standby, to limit heat input into TPCW AND THROTTLE TPCW Hx ICW flows using TPCW COMBINED OUTLET VALVE, 3-50-401, until total ICW flow is below 18,500 gpm.

6. IF a single lOW Pump has operated at flows greater than 18,500 gpm, THEN REFER TO 3-NOP-019, Intake Cooling Water System.

REFERENCES:

1. FPL Dwg 5613-M-3019, Sh 1

2. FPL EWD 5610-E-27, Sh 25, Misc. Alarms

3. PTN-BFSM-98-016, Affects of Opening 3/4-50-402 While 3/4-50-401 is Fully Open

4. PC/M 02-0 18, ICW Header Low Alarm Setpoint Change

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# 077 AK3.02 Importance Rating 3 Knowledge of the reasons for the following responses as they apply to Generator Voltage and Electric Grid Disturbances: Actions contained in abnormal operating procedure for voltage and grid disturbances Proposed Question: RO Question # 52 Given the following conditions:

• Unit 3 is at 95% power.

• System Dispatch contacts the Unit Supervisor regarding reactive load oscillations.

• The Switchyard voltage is stable.

• Voltage Regulator Selector Switch is ON.

• The TCS MVARs are changing.

Subsequently:

• The crew enters 3-ONOP-090, Abnormal Generator MW/M VAR Oscillation.

• The U3 Turbine Operator reports the Minimum Excitation Module #5 light is lit.

Which ONE of the following describes the operators response to stabilize the Main Generator?

A. Raise Main Generator AC Voltage Regulator setpoint to raise MVARs towards LAG.

B. Raise Main Generator DC Voltage Regulator setpoint to raise MVARs towards LAG.

C. Lower Main Generator AC Voltage Regulator setpoint to raise reactive load amps.

D. Lower Main Generator DC Voltage Regulator setpoint to raise reactive load amps.

Proposed Answer: A Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. Correct. This is a mechanism described in 3-ONOP-090 to adjust for MVAR swings, B. Incorrect. Plausible since this is a mechanism described in 3-ONOP-090 to adjust for MVAR swings with the Voltage Regulator placed in TEST position. However, Voltage Regulator Selector Switch is ON C. Incorrect. Plausible since 3-ONOP-090 provides guidance to lower voltage if VARS are high. However, the Minimum Excitation Module #5 light is lit, indicating that VARS are too low and voltage needs to be increased.

D. Incorrect. Plausible since 3-ONOP-090 provides guidance to lower voltage if VARS are high. However, the Minimum Excitation Module #5 light is lit, indicating that VARS are too low and voltage needs to be increased. Also, this is a mechanism described in 3-ONOP-090 to adjust for MVAR swings with the Voltage Regulator placed in TEST position. However, Voltage Regulator Selector Switch is ON Technical Reference(s): 3-ONOP-090 Rev 0 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Learning Objective: [P6902137 Obj. 5, hA (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

STEP ACTIONIEXPECTED RESPONSE 1 I RESPONSE NOT OBTAINED I

CAUTIONS

• Transferring voltage regulator to MANUAL is NOT always the safest mode of operation and should be avoided as much as possible.

• Failure to maintain voltages matched between the units may be causing the Voltage Regulator oscillations.

• The Power System Stabilizer is designed to dampen MW and MVAR oscillations without any operator intervention and should be maintained in operation as much as possible.

• Monitoring Main Generator RTDs is required if TPCW flow or temperature is changed due to the effect on the Main Generator hydrogen leakage. An increase in hydrogen leakage is expected if the gas temperature to rotor temperature gradient increases (Reference CR 2008-803).

NOTES i I . Type A - Slow MW swings of 3 or more seconds per swing. I Type B MVAR swings ONLY.

I -

I I

• Type C - Slowly building, poorly dampened MW (rotor) swings of approximately I l second per swing (when less than 30 MW they will typically involve only PTN and I Ciystal River.)

I I

. Type D System wide MW oscillations of approximately 1 second per swing, suddenly I -

appearing as opposed to slowly building up.

I Determine Type MWIMVAR Swing Being Experienced on the Unit

a. Type A NONE IN PROGRESS

a. GotoStep2.

b. Type B NONE IN PROGRES

b. GotoStep7.

c. Type C - NONE IN PROGRESS c. Observe NOTES prior to Step 16 JQ go to Step 16.

d. Type D NONE IN PROGRESS

d. Observe NOTES prior to Step 16 AND goto Step 16.

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STEP ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

4 Trigger The Digital Fault Recorder 5 Contact The Following Personnel For Assistance In Dampening The Secondary System Swings

• System Engineer/Component Specialist

• System Protection Supervisor

• Appropriate Maintenance discipline supervisor 6 Return to Step I 7 Check With System Dispatcher To Return to Step 1.

determine The Following

• No system disturbance is in progress OR

• No voltagedrop has occurred in the switchyard OR

• No voltage fluctuations are occurring in the switchyard W97:/DH/In/in/cls

STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

8 Check for the Following Indications IF Unit 4 has low VARS OR is in the lead with a MINIMUM EXCITATION ALARM,

. MVAR meter LOW VARS

- THEN raise VARS on Unit 4 to move off the minimum excitation limiter AND clear the OR MINIMUM EXCITATION related alarms. Go to Step 10.

• Ann E 8/2, GEN FIELD FORCINGNOLT REG LIMITING IN ALARM OR

• At the Exciter Switchgear, the MINIMUM EXCITATION module #5 light ON -

9 Raise Voltage Using the AC Regulator to Move Off the Minimum Excitation Limiter AND Clear Any MINIMUM EXCITATION Related Alarms 10 Check Units 1, 2, and 4 VARS NORMAL

- IF unit(s) has low VARS OR is in the lead, THEN have them raise their voltage to match voltage(s) of the other Unit(s).

11 Check MVAR Meter NORMAL

- IF VARS are high on this unit, THEN lower voltage to match voltage of other unit(s).

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PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# E04 EK2.1 Importance Rating 3.5 Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

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

Proposed Question: RO Question # 53 Given the following conditions:

• 3-EOP-ECA-1.2, LOCA Outside Containment, has been entered.

• The crew closed MOV-3-744A & B, RHR Discharge to Cold Leg Isolation Valves.

• The leak is between the 3B RHR Heat Exchanger and 3-HCV-758, RHR HX Outlet Flow Control Valve.

Which ONE of the following completes the following statement?

In accordance with 3-EOP-ECA-1 .2, isolation of the LOCA outside containment can be verified based on (1)

Local operator actions (2) for Alternate RHR to be available for plant cooldown.

A. (1) increasing RCS pressure (2) are required B. (1) increasing RCS pressure (2) are NOT required C. (1) decreasing Auxiliary Building radiation (2) are required D. (1) decreasing Auxiliary Building radiation (2) are NOT required PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: A Explanation (Optional):

A. Correct lAW below discussion B. Incorrect. 1 st part right. 2 part wrong. Plausible Alternate RHR is train separated, but need to realize that the trains cross-connect on the heat exchanger outlet and manual valves must be repositioned.

C. Incorrect. 1st part wrong. 2nd part right. Plausible Auxiliary Building radiation levels used as criteria for entering ECA-1 .2, however, validation of leak isolation criteria is increasing RCS pressure.

D. Incorrect. 1st part wrong. 2 part wrong. Plausible Auxiliary Building radiation levels used as criteria for entering ECA-1 .2, however, validation of leak isolation criteria is increasing RCS pressure. Alternate RHR is train separated, but need to realize that the trains cross-connect on the heat exchanger outlet outlet and manual valves must be repositioned.

3-EOP-ECA-1 .2 Rev 2 Technical Reference(s): BD-EOP-ECA-1 .2 Rev 3 (Attach if not previously provided) 5613-M-3050 sheet 1 rev. 36 Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902333 Obj. 5 (As available)

Question Source: Bank # 92133 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 PTN L-15-1. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Corn ments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

2 LOCA OUTSIDE CONTAINMENT 6 of 8 PROCEDURE NO.:

3-EOP-ECA-1 .2 TURKEY POINT UNIT 3 I STEP IIACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

2. Try To Identify And Isolate Break:

a. Verify SI RESET a. Reset SI.

ib. Close RHR Dcharge To Cold Leg Isolation valves:

I. 4AOV-3-744A MOV-3-744B c3 R 1 CS pressurD c Go to Step 3 BLE OR DECREASING 1

T

d. Open RHR Discharge To Cold Leg Isolation valves:

• MOV-3-744A

• MOV-3-744B

e. Close SI To Cold Leg Isolation e. Locally close valve(s).

valves:

• MOV-3-843A

• MOV-3-843B

f. RCS pressure f. Go to Step 3.

STABLE OR DECREASING

g. Open SI To Cold Leg Isolation Valves:

• MOV-3-843A

• MOV-3-843B

h. Contact RP for survey of Auxiliary Building to determine source of high radiation

REVISION NO.: PROCEDURE TITLE: PAGE:

2 LOCA OUTSIDE CONTAINMENT 7 of 8 PROCEDURE NO.:

3-EOP-ECA-1 .2 TURKEY POINT UNIT 3 I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

3. heck If Break Is Isolated:

RCS pressure INCREASING a. Go to 3-EOP-ECA-1 .1, LOSS OF EMERGENCY COOLANT RECIRCULATION, Step 1.

b. Go to 3-EOP-E1, LOSS OF REACTOR OR SECONDARt, COOLANT, Step 1 End of Section 3.0

Page 8 BD-EOP-ECA-1.2 LOCA Outside Containment 7/31/14 WOG Procedure Step 3 BASIS DOCUMENT PTN Procedure Step 3 N

Check If Break Is Isolated BASIS:

This step instructs the operator to check RCS pressure to determine if the break has been isolated by previous actions. If the break is isolated in the previous step, a significant RCS pressure increase will occur due to the SI flow filling up the RCS with break flow stopped.

The operator transfers to E-l, LOSS OF REACTOR OR SECONDARY COOLANT, if the break has been isolated, for further recovery actions. If the break has not been isolated, the operator is sent to ECA-l.1. LOSS OF EMERGENCY COOLANT RECIRCULATION, for further recovery actions since there will be no inventory in the sump.

It should be noted that for some breaks SI flow may cause an RCS pressure increase independent of break isolation. It should also be noted that for larger breaks, RCS repressurization may be delayed following break isolation. Additionally, if the RCS is saturated or a cooldown is in progress, RCS repressurization will proceed more slowly. Other means of verifying break isolation should be checked. For example, increasing RVLIS trend due to injection flow, decreasing trends in local abnormal conditions and local observation (if practical) may be useful.

STEP DEVIATIONS FROM WOG GUIDELINES:

TYPE DESCRIPTION N/A PLANT SPECIFIC SETPOINTS:

N/A WCO/fm/fm/cls

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# E05 EK2.2 Importance Rating 3.9 Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the following:

Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Proposed Question: RO Question # 54 Given the following conditions:

• The Reactor trips due to a loss of offsite power.

• A loss of all feedwater occurs.

• 3-EOP-FR-H.1, Loss of Secondary Heat Sink, is in progress.

• A source of feedwater is NOT restored.

• The crew is establishing RCS bleed and feed when only one Pressurizer PORV opens.

Which ONE of the following correctly completes the statement below?

Based on these plant conditions, the RCS bleed path is (1) and the crew should (2) , while continuing efforts to re-establish a source of feedwater to the SGs.

A. (1) adequate (2) depressurize SG5 to less than 360 psig B. (1) adequate (2) open all RCS Vents C. (1) inadequate (2) depressurize SGs to less than 360 psig PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. (1) inadequate (2) open all RCS Vents Proposed Answer: D Explanation (Optional):

A. Incorrect. 1 st part wrong. 2n,d part wrong. Plausible to believe that one PORV provides sufficient bleed path for RCS Bleed and Feed operations. However, With only one PORV open, 3-EOP-FR-H.1 directs crew to the RNO for Verify Adequate RCS Bleed Path to open the RCS and PZR vents. Also plausible to believe the correct action is to depressurize the SG to 360 psig, since that is an earlier step than Feed and Bleed.

However, the crew would have transitioned over the action to depressurize the SGs to 360 psig in order to initiate Feed and Bleed.

B. Incorrect. 1 st part wrong. 2 part right. Plausible to believe that one PORV provides sufficient bleed path for RCS Bleed and Feed operations. However, With only one PORV open, 3-EOP-FR-H.1 directs crew to the RNO for Verify Adequate RCS Bleed Path to open the RCS and PZR vents.

C. Incorrect. 1st part right. 2 part wrong. Plausible to believe the correct action is to depressurize the SG to 360 psig, since that is an earlier step than Feed and Bleed.

However, the crew would have transitioned over the action to depressurize the SGs to 360 psig in order to initiate Feed and Bleed.

D. Correct. With only one PORV open, 3-EOP-FR-H.1 directs crew to the RNO for Verify Adequate RCS Bleed Path to open the RCS and PZR vents.

Technical Reference(s): 3-EOP-FR-H.1 Rev 6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902930 Obj. 3 (As available)

Question Source: Bank# 92017 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Turkey Point PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognftive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

6 RESPONSE TO LOSS OF SECONDARY HEAT SINK PROCEDURE NO.: 23 of 60 3-EOP-FR-H.1 TURKEY POINT UNIT 3 STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

17. Verify Adequate RCS Bleed Path: Perform the following:

• PRZPORVsBOTH OPEN a. Continue attempts to open PRZ

• PRZ PORV Block Valves PORVs and PRZ PORV Block Valves.

BOTH OPEN

b. Install fuses for RCS Vent Valves:

• RV-3-1 01 for SV-3-631 BA

• RV-3-1 02 for SV-3-631 8B

• RV-3-1 03 for SV3-631 9A

• RV-3-1 04 for SV-3-631 9B

• RV-3-1 05 for SV-3-661 2

• RV-3-1 06 for SV-3-661 1

c. WHEN power is restored to RCS vent valves, THEN open all RCS Vents:

• SV-3-631 8A

• SV-3-6318B

• SV-3-6319A

• SV3-631 9B

• SV-3-661 1 SV-3-661 2

18. Reset Containment Isolation Phase A And Phase B

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group#

KIA# Eli EA2.l Importance Rating 3.4 Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation) Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Proposed Question: RO Question # 55 Given the following conditions:

• A large break LOCA occurs on Unit 3.

• While performing 3-EOP-ES-l .3, Transfer To Cold Leg Recirculation, the crew discovers emergency coolant recirculation capability is lost.

• Containment pressure peaked at 22 psig and is now 18 psig.

Which ONE of the following completes the statements below?

The Containment Spray Pumps must be operated in accordance with (1)

As RWST level lowers to 55,000 gallons, (2) Containment Spray Pump(s) is/are required to be operating..

NOTE 3-EOP-E-1, Loss of Reactor or Secondary Coolant 3-EOP-ECA-1.1, Loss of Emergency Coolant Recirculation A. (1)3-EOP-E-l (2) One B. (1)3-EOP-E-l (2) Zero PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. (1) 3-EOP-ECA-1.1 (2) One D. (1) 3-EOP-ECA-1.1 (2) Zero Proposed Answer: D Explanation (Optional):

A. Incorrect. 1 st part wrong. 2 part wrong. Plausible to believe that 3-EOP-E-1 would be re-entered since that is the procedure from which 3-EOP-ES-1 .3 was transitioned. Also plausible that at least one CS pump would be required since containment pressure is still high at 18 psig. However, due to loss of emergency coolant recirculation capability, 3-EOP-ECA-1.1 is the procedure to provide guidance and the FOP criteria for loss of pump suction (RWST empty), requires that the CS, RHR, and HHSI pumps be stopped and placed in PULL-TO-LOCK.

B. Incorrect. 1 t part wrong. 2nd part right. Plausible to believe that 3-EOP-E-1 would be re-entered since that is the procedure from which 3-EOP-ES-1 .3 was transitioned.

However, due to loss of emergency coolant recirculation capability, 3-EOP-ECA-1.1 is the procedure to provide guidance C. Incorrect. 1 st part right. 2 nd part wrong. Plausible that at least one CS pump would be required since containment pressure is still high at 18 psig. However, due to loss of emergency coolant recirculation capability, 3-EOP-ECA-1.1 is the procedure to provide guidance and the FOP criteria for loss of pump suction (RWST empty), requires that the CS, RHR, and HHSI pumps be stopped and placed in PULL-TO-LOCK.

D. Correct. Due to loss of emergency coolant recirculation capability, 3-EOP-ECA-1 .1 is the procedure to provide guidance and the FOP criteria for loss of pump suction (RWST empty), requires that the CS, RHR, and HHSI pumps be stopped and placed in PULL TO-LOCK.

Technical Reference(s): 3-EOP-ECA-1 .1 Rev 3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902338 Obj. 5 (As available)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Source: Bank # 91539 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Callaway Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

LOSS OF EMERGENCY COOLANT RECIRCULATION 4 of 69 PROCEDURE NO.:

3-EOP-ECA-1.1 TURKEY POINT UNIT 3 1.0 PURPOSE This procedure provides actions to restore Emergency Coolant Recirculation capability, to delay depletion of the RWST by adding makeup and reducing outflow, and to depressurize the RCS to minimize break flow. In addition, this procedure provides actions to reestablish core cooling in the event of Recirculation Sump blockage.

2.0 SYMPTOMS AND ENTRY CONDITIONS This procedure is entered from:

1) E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 16, when cold leg recirculation capability can NOT be verified AND Foldout Page Item 7, in the event Recirculation Sump blockage occurs.

2) ES-1 .3, TRANSFER TO COLD LEG RECIRCULATION, Steps 2, ii, 14, 15, 17, 23 and Attachment 2, Step 1, and ES-i .4, Step 16 and Attachment 2, when at least one flow path from the sump can NOT be established or maintained OR Recirculation Sump blockage occurs.

3) ECA-i .2, LOCA OUTSIDE CONTAINMENT, Step 3, when a LOCA outside Containment can NOT be isolated.

4) Foldout Page Item 3 of 3-EOP-ES-i .4, and Item 2 of 3-FOP-ES-i .3, H Emergency Coolant Recirculation is lost after it had been established.

5) Operation of Containment Spray Pumps by referring to this procedure is referenced in FR-Z.i, RESPONSE TO HIGH CONTAINMENT PRESSURE, CAUTION prior to Step 3 and Step 3 when this procedure is in effect.

REVISION NO.: PROCEDURE TITLE:

PAGE:

LOSS OF EMERGENCY COOLANT RECIRCULATION PROCEDURE NO.: FOLDOUT 3-EOP-ECA-1 .1 TURKEY POINT UNIT 3 FOLDOUT PAGE For Procedure 3-EOP-ECA-1 .1

1. ADVERSE CONTAINMENT CONDITIONS IF either condition listed below occurs, THEN use [Adverse Containment Setpoints]:

Containment atmosphere temperature 180°F OR Containment radiation levels 1.3x10 5 R/hr WHEN Containment atmosphere temperature returns to less than 180°F, THEN Normal Setpoints can again be used.

WHEN Containment radiation levels return to less than 1.3x10 5 R/hr, THEN Normal Setpoints can again be used [I the TSC determines that Containment Integrated Dose has NOT exceeded i0 5 Rads.

2. RESTORATION OF EMERGENCY COOLANT RECIRCULATION CRITERIA IF emergency coolant recirculation capability is restored during this procedure, THEN return to procedure and step in effect.

3. LOSSOFPUMPSUCTION IF RWST level lowers to 60,000 g I

.0, Step 27 to stop pump(s) taking suction from the RWST and place switches in R Containment Spray Pumps RHRPumpsl High-Head SI Pumps aligned to Unit 3 RWST IF RWST level lowers to 20,000 gallons, AND Charging Pump suction is aligned to RWST, THEN stop Charging Pumps.

4. CST MAKEUP WATER CRITERIA IF CST level decreases to less than 12%, THEN add makeup to CST using 3-NOP-018.01, CONDENSATE STORAGE TANK (CST).

5. LOSS OF OFFSITE POWER OR SI ON OTHER UNIT IF SI has been reset AND subsequently either offsite power is lost OR SI actuates on the other unit, THEN restore safeguards equipment, and at least one Computer Room Chiller, to required configuration.

Refer to Attachment 2 for essential loads.

6. U4 RWST SUCTION SOURCE STOP CRITERIA IF SI has been established from U4, AND U4 RWST level lowers to less than 60,000 gallons, THEN y running High-Head SI Pump taking suction from the U4 RWST is required to be stopped.

7. CONTAINMENT FLOODING CRITERIA IF SI has been established from U4, AND Containment Recirculation Sump Level increases to greater than 448 inches, THEN y running pumps providing RCS makeup are required to be stopped:

Containment Spray Pumps RHR Pumps High-Head SI Pumps Charging Pumps

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA# E12 EA2.2 Importance Rating 3.4 Ability to determine and interpret the following as they apply to the (Uncontrolled Depressurization of all Steam Generators) Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Proposed Question: RO Question # 56 Given the following conditions:

• 3-EOP-ECA-2. 1, Uncontrolled Depressurization of All Steam Generators, is in progress.

• RCS temperature decreases from 547°F to 422° F in the last hour.

• The crew adjusts AFW flow.

• SG NR levels are all off-scale low.

• 3A SG Safety Valve reseats.

Which ONE of the following describes the AFW flow requirement and the action for the next procedure transition?

A.

• AFW flow is at a minimum of 50 gpm per SG.

• Immediately transition to 3-EOP-E-2, Faulted SG Isolation.

B.

• AFW flow is at a minimum of 400 gpm per SG.

• Immediately transition to 3-EOP-E-2, Faulted SG Isolation.

C.

• AFW flow is at a minimum of 50 gpm per SG.

• When 3A SG pressure rises, transition to 3-EOP-E-2, Faulted SG Isolation.

D.

• AFW flow is at a minimum of 400 gpm per SG.

• When 3A SG pressure rises, transition to 3-EOP-E-2, Faulted SG Isolation.

Proposed Answer: C PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Incorrect. Plausible because 1 condition of SG isolation is met. AFW reduction is correct because RCS temperature has dropped by more than 100 degrees in the last hour. Transition not made until pressure increase is observed.

B. Incorrect. Feed flow should be reduced, but plausible because this is normal flow when NR level is below 7%.

C. Correct. Feed reduction is correct because RCS temperature has dropped by more than 100 degrees in the last hour. Transition not made until pressure increase is observed.

D. Incorrect. Incorrect flow, but correct transition. Plausible because 400 gpm is normal flowrate and would be required based on RCS cooldown rate < 100°F per hour. The applicant must determine that the requirement was exceeded.

Technical Reference(s): 3-EOP-ECA-2.1 Rev 6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Learning Objective: LP 6902335 Obj. 5 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Ginna Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

6 UNCONTROLLED DEPRESSURIZATION PROCEDURE NO OF ALL STEAM GENERATORS 5 of 64 3-EOP-ECA-2.1 TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED 3.0 OPERATOR ACTIONS CAUTION If AFW Pumps are the only available source of feed flow, steam supply to the AFW Pumps is required to be maintained from at least one S/G.

NOTE FoIdout page is required to be monitored throughout this procedure 1

1. Check Secondary Pressure Boundary:

a. Verify the following isolation valves a. Manually close valves.

CLOSED:

IF valves can NOT be closed,

• Main Steamline Isolation valves THEN dispatch operator to locally

• Main Steam Isolation Bypass isolate affected S/G one loop at a valves time using:

• One of the following in each Main Attachment 3 for SIG A Feed Line:

Feedwater Control valves Attachment 4 for SIG B Attachment 5 for S/G C Feedwater Isolation valves

• One of the following in each Main Feed Bypass line:

Feedwater Bypass valves OR Feedwater Bypass Isolation valves

• SIG Steam Dump To Atmosphere valves

• SIG Blowdown Isolation Valves

• SIG Sample valves

REVISION NO.: PROCEDURE TITLE: PAGE:

6 UNCONTROLLED DEPRESSURIZATION PROCEDURE NO.: OF ALL STEAM GENERATORS of 64 3-EOP-ECA-2.1 TURKEY POINT UNIT 3 STEP ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED I CAUTION

• A minimum feed flow of 50 gpm is required to be maintained to each S/G with a Narrow Range Level less than 7%[27%1.

• Low range flow indication is NOT available when using Main Feedwater instrumentation and an alternate source of feedwater. Changes in RCS temperature and S/G level can be used to control feedwater flow.

• Feed flow is required to be initiated slowly to avoid excessive RCS cooldown and to limit thermal stress in SIGs.

NOTE Shutdown Margin is required to be monitored during RCS cooldown.

2. Control Feed Flow To Minimize RCS Cooldown:

a. Check cooldown rate in RCS a. Decrease feed flow to 50 gpm to each CoId Legs LESS THAN 100°F/HRJ S/G.

1 Go to Step 2.c)

b. Check Narrow Range Level in all b. Control feed flow to maintain Narrow S/Gs LESS THAN 50%

Range Level less than 50% in all SIGs.

c. Check RCS Hot Leg temperatures c. Control feed flow or dump steam to STABLE DECREASING stabilize RCS Hot Leg temperatures.

IF adequate feed flow to stabilize Hot Leg temperatures OR 400 gpm is NOT available, THEN go to 3-EOP-FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, Step 1

REVISION NO.: PROCEDURE TITLE: PAGE:

6 UNCONTROLLED DEPRESSURIZATION PROCEDURE NO.: OF ALL STEAM GENERATORS FOLDOUT 3-EOP-ECA-2.1 TURKEY POINT UNIT 3 FOLDOUT PAGE For Procedure 3-EOP-ECA-2.1

1. ADVERSE CONTAINMENT CONDITIONS

a. IF either condition listed below occurs, THEN use [Adverse Containment Setpoints]:

Containment atmosphere temperature 180°F OR Containment radiation levels 1.3x10 5 R/hr

b. WHEN Containment atmosphere temperature returns to less than 180°F, THEN Normal Setpoints can again be used.

c. WHEN Containment radiation levels return to less than 1.3x10 5 R/hr, THEN Normal Setpoints can again be used if the TSC determines that Containment Integrated Dose has NOT exceeded Rads.

2. SI RE-INITIATION CRITERIA IF either condition listed below occurs, THEN manually start SI Pumps as necessary to restore RCS subcooling and PRZ level:

RCS subcooling based on Core Exit TCs LESS THAN 1 9°F[LESS THAN ADVERSE VALUE IN TABLE BELOW]

SI ADVERSE SUBCOOLING TABLE RCS PRESSURE (PSIG) ADVERSE SUBCOOLING VALUE

< 2485 AND 2000 35 °F

< 2000 AND 1500 45 °F

<1500AND1000 55°F

<1000AND500 110°F

<500 160°F OR PRZ level CAN NOT BE MAINTAINED GREATER THAN 7%[48%]

3. E-2TRANSITIbN CRITERV 4

IF any S/G pressure increapQ SI termination in Section 3 0 Step 9 through Step 17 is j

in progress, THEN go to 3-EOE ED STEAM GENERATOR ISOLATION, Step 1.

4. E-3 TRANSITION CRITERIA IF any S/G level increases in an uncontrolled manner OR any S/G has abnormal radiation, THEN manually start SI Pumps as necessary and go to 3-EOP-E-3, STEAM GENERATOR TUBE RUPTURE, Step 1.

5. COLD LEG RECIRCULATION SWITCHOVER CRITERIA IF RWST level decreases to less than 155,000 gallons, THEN go to 3-EOP-ES-1 .3, TRANSFER TO COLD LEG RECIRCULATION, Step 1.

6. CST MAKEUP WATER CRITERIA IF CST level decreases to less than 12%, THEN add makeup to CST using 3-NOP-018.01, CONDENSATE STORAGE TANK (CST).

7. LOSS OF OFFSITE POWER OR SI ON OTHER UNIT IF SI has been reset AND either offsite power is lost OR SI actuates on the other unit, THEN restore safeguards equipment, and at least one Computer Room Chiller, to required configuration.

Refer to Attachment 2 for essential loads.

8. RHR SYSTEM OPERATION CRITERIA IF RHR flow is less than 1100 gpm, THEN the RHR Pumps shall be shut down within 44 minutes of the initial start signal.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 001 AA2.05 Importance Rating 4.4 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal: Uncontrolled rod withdrawal, from available indications Proposed Question: RO Question # 57 Given the following conditions:

• Unit 4 is at 50% steady state power.

• Tavg is matched with Tref at 562° F.

• Automatic VCT makeup is in progress.

• TCS is in MW control at 410 MW.

• The Rod Motion Control Selector Switch is placed in AUTO after moving Control Bank D to 161 steps.

• The Axial Flux Difference is -3 when the Rod Motion Control Selector Switch is placed in AUTO.

Subsequently, a few minutes later:

• Unit 4 is at 51% and increasing.

• Tavg is 3°F higher than Tref.

• TCS is in MW control at 412 MW with Turbine Control Valves closing.

• The Axial Flux Difference is -0.5 and quickly trending more positive.

Which ONE of the following identifies the next required action?

A. Reduce turbine load at a slower rate in MW control.

B. Stop the auto makeup and commence Emergency Boration.

C. Reduce turbine load at a slower rate using the Speed/Power Switch.

D. Stop outward rod movement by placing the Rod Motion Control Selector Switch to MAN.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: D Explanation (Optional):

A. Incorrect. This alternative is plausible since 4-NOP-089 recommends the Main Turbine should be maintained in MANUAL control for steady state full power operation. Also, Step 4.2.4 gives directions for incremental Turbine Load changes. The candidate may determine power needs to be restored to the previous level. Tavg increasing is an indication that Control Rods have been withdrawn. The slight increase in Turbine power is a direct effect of an increase in steam pressure. As rods withdrawl, the flux in the Core shifts toward the top making Axial Flux Difference less negative.

B. Incorrect. This alternative is plausible since a VCT makeup is in progress. The assumption is positive reactivity is being added. Therefore, Tavg is increasing. rise A in Tavg creates a mismatch with Tavg. Tavg-Tref. However, in the case of a dilution, the water is equally mixed throughout the Core. Therefore, any effect on Axial Flux Difference is minimized. In this scenario, VCT makeup will have no effect since it is blended flow.

C. Incorrect. This alternative is plausible since 4-NOP-089 recommends the Main Turbine should be maintained in MANUAL control for steady state full power operation and the Speed/Power Switch can be used to lower Turbine Load.

D. Correct. The Tavg increase without a Tref (secondary) increase is an indication of positive reactivity addition. The decrease in Axial Flux Difference is another positive indication that there is a withdrawal of Control Rods. The immediate response of 4-ONOP-028, Step 4.4, Continuous Withdrawal of an RCC Control Bank, is to place the Rod Motion Control Selector switch to MAN.

4-ONOP-028 Rev 2 Technical Reference(s): 4-NOP-089 Rev 17 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902207 Obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X PTN L-15-1. DRAFT NRC EXAM SECURE IN FORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

Procedure No Procedure

Title:

Page:

7 Approval Date:

4-ONOP-028 Reactor Control System Malfunction 11/21/07 2.4 Continuous Withdrawal of an RCC Control Bank 2.4.1 RCCs stepping out as indicated on the RPIs or group demand step counters, and not manually initiated by the operator.

2.4.2 Tavg increases more than 1.5 degrees F above Tref 2.4.3 Annunciators I. B 4/4, TAVG/TAVG TREF DEVIATION

2. B 4/5, RCS HI LO TAVG

3. B 6/3, POWER RANGE OVERPOWER ROD STOP 2.5 Control Bank D Demanded Past ARO Position 2.5.1 Control Bank D group demand step counters indicate greater than the ARO position (228, 229 or 230 steps as defined in the Plant Curve Book, Section 7, COLR).

AI07-I I\AI/In/n,rkh

Procedure No Procedure

Title:

Page:

9 Approval Date:

4-ONOP-028 Reactor Control System Malfunction 11/21/07 CAUTIONS

• If the Rod Control System is inoperable due to Urgent Failure or other cause, the Shift Manager shall be notified immediately.

• If a transient occurs and the Reactor cannot be stabilized by boration/dilut ion or changes in turbine load, the Reactor shall be tripped and a transition made to 4-EOP-E-O, REACTOR TRIP OR SAFETY INJECTION.

NOTES 1

I I

I

• Boration/Dilution or changes in turbine load will affect shutdown margin and axial offset. If plant conditions permit, the Shift Manager shall be consulted for methods I used to achieve and maintain stable plant conditions. I

• Failure of RCCs to move when demandeci (e.g., Rod Control Urgent Failure),

• constitutes inoperabiity of the associated RCCs. The requirements of T.S. 3.1.3.1 I apply.

a a a a I

4.0 IMMEDIATE ACTIONS 4.1 Immovable RCC 4.1.1 IF the Rod Motion Control Selector is in Auto, THEN place in the MAN position.

4.1 .2 DO NOT withdraw any control banks until the RCCs have been aligned.

4.2 Failure of an RCC Control Bank to Insert with Reactor Control in Automatic 4.2.1 Place the Rod Motion Control Selector switch to the MAN position.

4.3 Continuous Insertion of an RCC Control Bank 4.3.1 Place the Rod Motion Control Selector switch to the MAN position.

4.3.2 ii RCC control cannot be maintained manually, THEN trip the Reactor and Turbine and go to 4-EOP-E-O.

REACTOR TRIP OR SAFETY INJECTION.

4.4 Continuous Withdrawal of an RCC Control Bank 4.4.1,, Place the Rod Motion Control Selector switch to the MAN position.

4.4.2 IF RCC control cannot be maintained manually, THEN trip the Reactor and Turbine and go to 4-EOP-E-O, REACTOR TRIP OR SAFETY INJECTION.

4.5 Control Bank D Demanded Past ARO Position 4.5.1 None W97:/JWB/lnlmr/ab

REVISION NO.: PROCEDURE TITLE: PAGE:

17 MAIN TURBINE 24 of 59 PROCEDURE NO.:

4-NOP-089 TURKEY POINT UNIT 4 4.2.3 Load Control with Turbine Inlet Pressure Control (TIP CNTRL) in NOTE

• Each TOUCH of the RAISE/LOWER buttons on the TCS LOAD CONTROL screen will adjust power approximately 1 MWe.

• The SPEED/LOAD TURBINE CONTROL switch can also be used to Raise or Lower Load in small increments (approximately 1 MWe per momentary actuation or approximately 42 MWe/Minute ramp if constantly held in Raise or Lower).,

• TCS LOAD MW CNTRL and TIP CNTRL are NOT selectable until load is initially raised above 40 MWe using the HMI RAISE button or the SPEED/LOAD TURBINE CONTROL switch.

• Placing TCS in MANUAL will reset the target ramp rate to default. If a turbine ramp is stopped and TCS is placed in MANUAL, the ramp rate must be re entered and confirmed in TCS.

• The Main Turbine should be maintained in MANUAL control for steady state full power operation.

CAUTION Do NOT CONFIRM a MW or TIP target unless GO will be executed immediately after selecting CONFIRM. If the load change is NOT executed, the MW or TIP target must be cleared from TCS. Selecting GO then selecting HOLD will clear the target from TCS. On the TCS LOAD CONTROL screen, check that LOAD TARGET equals LOAD SETPOINT. Reference AR 1868630.

1. Raise / Lower Turbine Load with ramp feature with TIP CNTRL IN as follows:

A. SELECT IN on TIP CNTRL on LOAD CONTROL screen.

B. REPEAT the following steps until desired load is reached:

(1) SELECT AND CONFIRM TIP TARGET on LOAD CONTROL screen.

(2) SELECT AND CONFIRM TIP RAMP RATE on LOAD CONTROL screen.

(3) SELECT GO.

REVISION NO.: PROCEDURE TITLE: PAGE:

17 MAIN TURBINE 26 of 59 PROCEDURE NO.:

4-NOP-089 TURKEY POINT UNIT 4 4.2.4 Change Turbine Load Incrementally NOTE

• Each TOUCH of the RAISE/LOWER buttons on the TCS LOAD CONTROL screen will adjust power approximately 1 MWe.

• The SPEED/LOAD TURBINE CONTROL switch can also be used to Raise or Lower Load in small increments (approximately 1 MWe per momentary act uation or approximately 42 MWe/Minute ramp if constantly held in Raise or Lower).

• TCS LOAD MW CNTRL and TIP CNTRL are NOT selectable until load is initially raised above 40 MWe using the HMI RAISE button or the SPEED/LOAD TURBINE CONTROL switch.

• The Main Turbine should be maintained in MANUAL control for steady state full power operation.

CAUTION Prior to Incremental Turbine Load changes, on the TCS LOAD CONTROL screen, check that LOAD TARGET equals LOAD SETPOINT. Reference AR 1868630.

1. CHECK LOAD TARGET equals LOAD SETPOINT.

2. LOWER Turbine Load on HMI as follows:

• PRESS LOWER on LOAD CONTROL screen.

3. RAISE Turbine Load on HMI as follows:

• PRESS RAISE on LOAD CONTROL screen.

4. LOWER Turbine Load using the SPEED/LOAD TURBINE CONTROL switch as follows:

• Momentarily ROTATE the SPEED/LOAD TURBINE CONTROL switch to LOWER.

5. RAISE Turbine Load using the SPEED/LOAD TURBINE CONTROL switch as follows:

• Momentarily ROTATE the SPEED/LOAD TURBINE CONTROL switch to RAISE.

End of Section 4.2.4

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA 028 AK3.03 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunctions: False indication of PZR level when PORV or spray valve is open and RCS saturated Proposed Question: RO Question # 58 Given the following conditions:

• Loss of Offsite Power occurs on Unit 3.

• 3A and 3B EDGs are powering the 3A and 36 4KV Buses.

• 3-EOP-ES-0.4, Natural Circulation Cooldown with Steam Void in Vessel (Without RVLMS) is in progress.

• RCS pressure is 1635 psig.

• Pressurizer level is 30%.

• Prior to depressurizing the RCS, Pressurizer PORV PCV-3-456 fails open.

Which ONE of the following identifies the reason for the initial rapidly increasing Pressurizer level during this event?

A. The steam space in the Pressurizer collapses allowing more makeup to be injected immediately into the RCS by the HHSI Pumps.

B. Pressurizer level reference legs flash which results in an increase in indicated level.

C. Safety Injection Accumulators inject into the RCS which increases Pressurizer level.

D. Reactor upper head region voiding occurs which results in mass transfer from the Reactor Head to the Pressurizer.

Proposed Answer: D PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Incorrect. This is plausible since the failed PORV removes energy from the Pressurizers steam space. However even with pressure at 1635 psig, the HHSI Pumps will not immediately inject into the RCS since they are not running. They are not running since SI is not required and are blocked in Step 11 of 3-EOP-ES-0.2, Natural Circulation Cooldown. Upper head region voiding is the reason why there is a rapidly increasing PRZ level.

B. Incorrect. This is plausible since Pressurizer level is an affected transmitter during adverse Containment conditions. Also, an assumption is made that a rapid depressurization affects saturation conditions in the reference leg. This would be true if the reference leg was a wet internal reference leg. However, the pressurizer reference leg is an external reference leg feed by condensing pots and at containment ambient temperature. Upper head region voiding is the reason why there is a rapidly increasing PRZ level.

C. Incorrect. Plausible to believe the SI Accumulators might be injecting. However, the crew is instructed in Step 8 of 3-EOP-ES-0.4 to stop the depressurization at 800 psig.

This pressure is not low enough to allow the Accumulators to inject. However, the upper head region voiding is why there is a rapidly increasing PRZ level.

D. Correct. The note prior to Step 8 of 3-EOP-ES-0.4 states that the upper head region may void during depressurization. This failure results in a similar depressurization and forms a bubble in the RCS Upper Head which displaces water to the Pressurizer.

Technical Reference(s): 3-EOP-ES-0.4 Rev 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902326 Obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 5 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE:

PAGE:

2 NATURAL CIRCULATION COOLDOWN PROCEDURE NO.: WITH STEAM VOID IN VESSEL (WITHOUT RVLMS) 13 of 36 3-EOP-ES-O.4 TURKEY POINT UNIT 3 I STEP II ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I CAUTION RCS depressurization below 675 psig may result in SI Accumulator injection and loss of void size monitoring capability.

NOTE The upper head region may void during depressurization. This will result in a rapidly increasing PRZ level.

8. Depressurize RCS:

a. Check RCS pressure a. Observe NOTE prior to Step 9, and GREATER THAN 800 PSIG go to Step 9.

b. Check Letdown IN SERVICE b. Use one PRZ PORV.

Go to Step 8.d.

c. Establish Auxiliary Spray using c. Use one PRZ PORV.

Attachment 2

d. Depressurize RCS unt either of the following conditions satisfied:

RCS pressure LESS THAN 800 PSIG OR PRZ level GREATER THAN 90%

e. Stop RCS depressurization

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KJA# 051 AA1.04 Importance Rating 2.5 Ability to operate and I or monitor the following as they apply to the Loss of Condenser Vacuum: Rod position Proposed Question: RO Question # 59 Given the following conditions:

• Unit 3 is at 100% power.

Subsequently:

• E513, CONDENSER LO VACUUM alarms.

• Crew enters 3-ONOP-014, Main Condenser Loss of Vacuum.

• The Crew commences a Fast Load Reduction lAW 3-GOP-100.

• Annunciator B8/1, ROD BANK LO LIMIT alarms.

Which ONE of the following describes first action required by 3-GOP-100?

A. Withdraw Control Rods until B811, ROD BANK LO LIMIT alarm clears.

B. Immediately borate the RCS at least 16 gpm.

C. Slow or stop the Turbine load reduction.

D. Trip the Reactor and enter 3-EOP-E-0.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible since it is guidance in 3-GOP-100, however is occurs after the turbine load reduction is stopped.

B. Incorrect. Plausible since this action is required by B8/2 Annunciator alarm.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Correct. This is the first step after B8/i alarms in 3-GOP-100.

D. Incorrect. Plausible since this action is on the foldout page of 3-GOP-100 for Tavg-Tref

> 6°F and was former guidance in previous versions in 3-GOP-i 00 for B812 alarm.

Technical Reference(s): 3-GOP-iOO Rev 7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902282 Obj. 4,5 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

STEP ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

12 Monitor Annunciator B 811, ROD BANK LO IF Annunciator B 8/1 alarms, THEN perform the LIMIT CLEAR following: E

a. low or stop the load reduction.

b. Place rods in manual to stop control rod insertion.

c. Use manual rod control if needed for Tavg control.

d. IF not borating, THEN perform the following to borate:

1) Set boric acid totalizerto 50 gallons.

2) Determine boric acid flow rate as directed by the Unit Supervisor.

3) Place the Reactor Makeup Selector Switch to BORATE.

4) Place the RCS Makeup Control Switch to START.

5) Adjust the setpoint on the Boric Acid Controller FC-3-1 1 3A to the desired flow rate as indicated on FR-3-1 13.

e. Withdraw Control Rods to establish TavglTref AT up to +3°F Tavg GREATER THAN Tref.

f. Repeat Step 12 RNO steps until Annunciator B 8/1 is CLEAR and to prevent B 8/2, ROD BANK AIB/C/D EXTRA LO LIMIT from alarming.

W2fl1 flflfl/mr1cIsIrs

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 060 AA1.02 Importance Rating 2.9 Ability to operate and I or monitor the following as they apply to the Accidental Gaseous Radwaste Release: Ventilation system Proposed Question: RO Question # 60 Given the following conditions:

• Waste Gas Decay Tank D contains high-activity gas.

• Waste Gas Decay Tank D Relief Valve develops a flange leak that is slowly dispersing into the Aux Bldg.

• R-14, Plant Vent Gas Monitor, alarms.

• Crew enters 3-ONOP-067, Radioactive Effluent Release.

Which ONE of the following identifies the plant response to an R-14 alarm, if any occurs?

A. Aux Bldg Exhaust Fans trip.

B. Aux Bldg Supply Fans trip.

C. No effect on Aux Bldg or Control Room Ventilation.

D. Control Room Ventilation shifts to recirculation mode.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible to believe that high activity in the Auxiliary Building will trip the Exhaust Fans to prevent radioactive release to the public. However, the Auxiliary Building Exhaust and Supply fans have no automatic trips associated with R-14.

Radiation Monitor R-14. R-14 will trip closed the Gas Release Header Isolation Valve, RCV-14, not the fans.

PTN L-15-I. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION B. Incorrect. Plausible to believe that high activity in the Auxiliary Building will trip the Supply Fans to prevent radioactive release to the public. However, the Auxiliary Building Exhaust and Supply fans have no automatic trips associated with R-14. Radiation Monitor R-14. R-14 will trip closed the Gas Release Header Isolation Valve, RCV-14, not the fans.

C. Correct. Flange leakage would enter general Auxiliary Building Spaces where it would be exhausted by the Auxiliary Building Exhaust fan and eventually pass by Plant Vent Radiation Monitor R-14. R-14 will trip closed the Gas Release Header Isolation Valve, RCV-14, and will affect neither the Control Room nor the Auxiliary Building ventilation systems.

D. Incorrect. Plausible to believe that high activity in the Auxiliary Building will shift Control Room Ventilation to recirculation mode. However, R-14 will trip closed the Gas Release Header Isolation Valve, RCV-14, and will affect neither the Control Room nor the Auxiliary Building ventilation systems.

3-ONOP-067 Rev 6 Technical Reference(s): 5610-M-3060 Sh. 1 Rev 25 (Attach if not previously provided) 5610-M-3061 Sh. 13 Rev 9 Proposed References to be provided to applicants during examination: N Learning Objective: LP69021500bj.4 6 9 LP 6902155 Obj. 3 5 (As available)

Question Source: Bank# 93195 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 Sequoyah Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

STEP ACTION/EXPECTED RESPONSE I ITi 42 Check For Release To Atmosphere

a. Verity RCV-01 4- CLOSED a. Perform the following:

1) Close RCV-014 by reducing the hand loader pressure to zero.

2) Verify the valve stem position indicator on RCV-01 4 indicates the valve is fully closed.

3) jf RCV-014 is not fully closed, THEN ensure the following GDT valves CLOSED:

• 4638A, A GDT to Plant Vent

• 4638B, B GDT to Plant Vent

• 4638C, C GDT to Plant Vent

• 4638D, D GDT to Plant Vent

• 4638E, E GDT to Plant Vent

• 4638F, F GDT to Plant Vent

b. Check if High Alarm on monitor caused by excessive release rate of gas decay tank b. Go to Step 42f.

c. Check count rate on all stack monitors

c. Go to Step 42f.

DECREASING

d. Perform following prior to recommencing gas release

1) Notify the Shift Manager and Chemistry of problem with gas release

2) Resample affected gas decay tank

3) Resubmit gas release permit to the Shift Manager for approval

e. Return to Step 1

f. Check following: f. Perform the following:

1) Isolate the affected tank.

• All gas decay tank pressures less than 100 psi 9

2) Transfer contents of affected tank to another gas decay tank using

• No gas decay tank pressure decreasing 0-OP-061.15, WASTE GAS in an uncontrolled manner SYSTEM.

W97/fm/fm/cls

PTN L-15-1 DRAFT NRC EXAM SECURE IN FORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# 068 AK2.07 Importance Rating 3.3 Knowledge of the interrelations between the Control Room Evacuation and the following:

ED/G Proposed Question: RO Question # 61 Given the following conditions:

• A fire was confirmed in the Cable Spreading Room that was affecting plant equipment.

• The crew is implementing 0-ONOP-105, Control Room Evacuation.

• The site has a loss of offsite power.

• All emergency safeguards equipment operates as required.

Which ONE of the following identifies the EDG operation in accordance with 0-ONOP-105?

At the point when control of Shutdown Systems is established, the A. 3A EDG will need to be shutdown B. 4B EDG will need to be shutdown C. 3A EDG will remain loaded D. 4B EDG will remain loaded Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible to remember that the A train EDGs are verified running instead of being secured within the attachments. 0-ONOP-105 directs securing the A train EDGs and verifying that the B train EDGs are operating if a loss of offsite power.

ATTACHMENT 24, Maintaining a Safe, Stable Configuration Following Control Room Evacuation, provides additional guidance for starting the 3A and 4A EDGs.

B. Correct. 0-ONOP-1 05 directs securing the A train EDGs and verifying that the B train EDGs are operating for a loss of offsite power.

PTN L-15-]. DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Incorrect. Plausible to remember that the 3A14B train EDGs are verified running for train separation instead of A train EDGs being secured within the attachments.

O-ONOP-105 directs securing the A train EDGs and verifying that the B train EDGs are operating for a loss of offsite power. ATTACHMENT 24, Maintaining a Safe, Stable Configuration Following Control Room Evacuation, provides additional guidance for starting the 3A and 4A EDGs.

D. Incorrect. Plausible to remember that the 3B14A train EDGs are verified running for train separation instead of A train EDGs being secured within the attachments. 3B was the original site design; 4A EDG was added during Dual Unit Outage. 0-ONOP-105 directs securing the A train EDGs and verifying that the B train EDGs are operating for a loss of offsite power. ATTACHMENT 24, Maintaining a Safe, Stable Configuration Following Control Room Evacuation, provides additional guidance for starting the 3A and 4A EDGs.

Technical Reference(s): O-ONOP-105 Rev 11 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902252 Obj. 8 (As available)

Question Source: Bank# 101437 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Point Beach Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

11 CONTROL ROOM EVACUATION 46 of 221 PROCEDURE NO.:

O-ONOP-105 TURKEY POINT PLANT ATTACHMENT 13 Unit Supervisor (Page 6 of 15)

I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

7. SHUT DOWN 3A EDG:

A. CHECK 3A EDG RUNNING. GO TO Attachment 13, Step 7.0.

B. STOP 3A EDG using Emergency Stop Switch on EDG Engine Control Panel.

C. PLACE Master Control Switch to OFF.

8. PROCEED to 3B EDG Room.

NOTE The 3B EDG synchronizing switch and the 3B EDG panel voltmeter can be used to determine 4KV Bus 3B status; however, final confirmation should come from the Unit 3 RO.

CAUTION 3B EDG may trip during transfer from the Control Room to Local. If a trip occurs, the diesel should be allowed to coast down before restarting.

9. DETERMINE if 3B EDG should be STOPPED:

A. CHECK 3AB20, OBSERVE NOTE and CAUTION prior to BUS, OPEN. Attachment 13, Step 10 and GO TOr Attachment 13, Step 10.

B. PLACE EDG Bkr 3AB20 Synchronizing Switch to ON.

REVISION NO.: PROCEDURE TITLE: PAGE:

11 CONTROL ROOM EVACUATION 126 of 221 PROCEDURE NO.:

O-ONOP-105 TURKEY POINT PLANT ATTACHMENT 17 Non-Fire Brigade Nuclear Plant Operator (Page 3 of 8)

(STEP I ( ACTION/EXPECTED RESPONSE ( I RESPONSE NOT OBTAINED I NOTE In the event of a Control Room evacuation, 4A 4KV Bus is de-energized by the Unit Supervisor. Attachment 13, Step 3

13. PROCEED to 4A EDG Control Room.

14.SHUTDOWN 4A EDG:

A. CHECK 4A EDG RUNNING. GO TO Attachment 17, Step 14.C.

B. STOP 4A EDG using EMERGENCY STOP pushbutton.

C. PLACE CS-i, MASTER CONTROL SWITCH, to LOCAL.

D. PLACE CS-19, ISOLATION SWITCH, to ISOLATE.

E. PLACE CS-18, ISOLATION SWITCH, to ISOLATE.

F. RESET 186G L/O Relay.

G. On Control Panel 4C12A, ENSURE 4AA20, DIESEL GENERATOR 4A TIE BREAKER, is OPEN.

15. PROCEED to 4B EDG Control Room.

REVISION NO.: PROCEDURE TITLE:

PAGE:

11 CONTROL ROOM EVACUATION 127 of 221 PROCEDURE NO.:

O-ONOP-105 TURKEY POINT PLANT ATTACHMENT 17 Non-Fire Brigade Nuclear Plant Operator (Page 4 of 8)

I STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 6.DETERMlNE if 4B EDG should be STOPPED:

. On Control Panel 4C12B, CHECK OBSERVE NOTE and CAUTION prior to 4AB21, DIESEL GENERATOR 4B TIE Attachment 17, Step 17 and GO TO BREAKER, OPEL Attachment 17, Step 17.

B. SELECT BUS Side on VOLTMETER.

C. CHECK 4B 4KV Bus Volts between OBSERVE NOTE and CAUTION prior to 3950 and 4350 Volts. Attachment 17, Step 17 and GO TO Attachment 17, Step 17 D. WHEN directed by the Unit 4 RO, THEN OBSERVE NOTE and CAUTION prior to STOP 4B EDG. Attachment 17, Step 17 and GO TO Attachment 17, Step 17.

E. PLACE CS-i, MASTER CONTROL SWITCH, to LOCAL.

F. PLACE CS-19, ISOLATION SWITCH, to ISOLATE.

G. PLACE CS-18, ISOLATION SWITCH, to ISOLATE.

H. STOP 4B EDG by placing the ENGINE CONTROL switch to STOP.

I. GO TO Attachment 17, Step 20.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 K/A # 076 2.4.46 Importance Rating 4.2 Emergency Procedures I Plan: Ability to verify that the alarms are consistent with the plant conditions.

Proposed Question: RO Question # 62 Given the following conditions:

• R-3-20, Letdown Radiation Monitor, causes H1/4, PRMS HI RADIATION, to alarm.

• Crew enters 3-ONOP-041 .4, Excessive Reactor Coolant System Activity.

Which ONE of the following completes the statements below?

High Dose Equivalent 1-131 levels in the RCS (1) an indication of failed fuel.

High Co-60 levels in the RCS (2) an indication of failed fuel.

A. (1)is (2) is NOT B. (1)is (2) is C. (1)IsNOT (2) is D. (1)isNOT (2) is NOT Proposed Answer: A PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Explanation (Optional):

A. Correct. 3-ARP-097.CR H1/4 directs crew to 3-ONOP-067, Radioactive Effluent Release, then to 3-ONOP-041 .4, which maximizes letdown and refers to TS 3.4.8 and TS 3.4.8 action c. requires shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if DE 1-131 is >60 microcuries/gram.

DE 1-131 is an indication of fuel failure, while CO-60 is an activation product from RCS piping.

B. Incorrect. lS part right. 2n,d part wrong. Plausible to believe 00-60 is a by-product of fission since it is produced by neutron activation, but its presence in the RCS is due activation of CO-59 found in RCS piping.

C. Incorrect. 1st part wrong. 2 part wrong. Plausible to remember that Xenon-135 is a fission product that directly impacts core reactivity, is produced from lodine-135, and not associate lodine-131 as a fission product. Plausible to believe CO-60 is a by-product of fission since it is produced by neutron activation, but its presence in the RCS is due activation of 00-59 found in RCS piping.

D. Incorrect. 1st part wrong. 2 part right. Plausible to remember that Xenon-135 is a fission product that directly impacts core reactivity, is produced from lodine-135, and not associate lodine-131 as a fission product.

3-ARP-097.CR.H Rev 6 Technical Reference(s): 3-ONOP-067 Rev 6 .

3-ONOP-041 .4 Rev 4 (Attach if not previously provided)

TS 3.4.8 Rev 293 Proposed References to be provided to applicants during examination: No Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Also meets 10CFR55.43(b) item 2 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

6 7 CONTROL ROOM RESPONSE - PANEL H PROCEDURE NO.:

WINDOW:

3-ARP-097.CR.H TURKEY POINT UNIT 3 1/4 (Page_1 of_1)

CAUSES: 1. High radiation in one of systems monitored by PRMS

2. PRMS system component failure H114 PRMS HI RADIATION DEVICE: SETPOINT: LOCATION:

• R-1 1 Variable with each PRMS channel N/A

• R-12

• R-14

• R-15

• R-17A

• R-17B

• R-18

• R-19

• R-20 ALARM CONFIRMATION

1. CHECK the following:

• Countrate meter on each PRMS drawer in Rack QR-66

• Alarm indicators on each drawer in Rack QR-66 OPERATOR ACTIONS

1. IF alarm is on R-11, R-12, R-14, R-I7AIB, R-18, or R-20, THEN REFER TO 3-ONOP-067, Radioactive Effluent Release for expected automatic actions.

2. IF alarm is on R-15 or R-19, THEN REFER TO 3-ONOP-071.2, Steam Generator Tube Leakage for expected automatic actions.

3. IF alarm is on R-14, R-17A, R-17B, R-18, or R-19, THEN CHECK alarm valid as follows:

A. CHECK FAIL/TEST light NOT LIT.

B. PUSH FAIL/TEST light (meter reading of 288 or 289K)

C. PUSH SOURCE CHECK light (should get meter increase).

D. PUSH HIGH ALARM light to determine if meter level is above high alarm setpoint.

4. ENSURE required automatic actions.

5. IF alarm is on R-11, R-12, R-14, R-17AJB, R-18, OR R-20, THEN REFER TO 3-ONOP-067, Radioactive Effluent Release.

6. IF alarm is on R-15 OR R-19, THEN REFER TO 3-ONOP-071.2, Steam Generator Tube Leakage.

7. REFER TO TS 3.3.3, 3.4.6, and 3.9.13 for additional required actions.

REFERENCES:

Tech Spec Sections 3.3.3, 3.4.6, and 3.9.13 PC/M 07-055, R-1 5 Steam Jet Air Ejector Monitor Replacement

Procedure No.: Procedure

Title:

e:

Approval Date:

3-ONOP-067 Radioactive Effluent Release 5/27/10 STEP ACTION/EXPECTED RESPONSE I I RESPONSE NOTOBTAINED I

CAUTION IF more than one high radiation event is occurring, the operator should prioritize actions to minimize OFFSITE DOSE.

NOTES i

• Prioritization should include consideration of release rate, size of leak, isolable or not, I etc.

I

• Step 3 RNO actions should be performed in the determined order of priority. I L

3 CHECK PRMS HIGH ALARM OFF - Perform the following:

• Check Ri 1 AND Ri 2 HIGH ALARMS -

• IF R-1 1 AND R-1 2 HIGH ALARM IS ON, OFF THEN go to Step 16.

• Check R-1 7A AND R-1 7B HIGH ALARMS -

• IF R-1 7A OR R-1 7B HIGH ALARM IS OFF ON, THEN go to Step 29.

• Check R-14 HIGH ALARM OFF -

IF R-14 HIGH ALARM IS ON, THEN go to Step 42.

• Check R-18 HIGH ALARM OFF -

IF R-20 HIGH ALARM IS ON, THEN

• Check R-20 HIGH ALARM OFF -

perform 3-ONOP-041 .4, EXCESSIVE REACTOR COOLANT SYSTEM ACTIVITY, while continuing with this procedure.

IF R-18 HIGH ALARM IS ON, THEN perform the following:

a. Verify RCV-O1 8- CLOSED.

b. a Liquid Release is in progress, THEN terminate the release.

c. Inform the Shift Manager of R-18 alarm.

d. Determine and correct the cause of the R-18 high alarm before commencing another liquid release.

W97/fm/fm/cls

REVISION NO.: PROCEDURE TITLE: PAGE:

EXCESSIVE REACTOR COOLANT SYSTEM ACTIVITY 4 of 10 PROCEDURE NO.:

3-ONOP-041 .4 TURKEY POINT UNIT 3 1.0 PURPOSE This procedure provides instructions to be followed in the event of unusually high Reactor Coolant System (RCS) specific activity levels that may be caused by:

• Crud bursts

• Demineralizer resin exhaustion

• Fuel element failures 2.0 ENTRY CONDITIONS

• High alarm on R-3-20, REACTOR COOLANTTDOWN

• Radiochemical analyses show rising levels of fission products

• A rise in Auxiliary Building radiation levels near the letdown, purification, and Volume Control System components

• A rise in containment high range or area radiation monitors

REVISION NO.: PROCEDURE TITLE: PAGE:

EXCESSIVE REACTOR COOLANT SYSTEM ACTIVITY 6 of 10 PROCEDURE NO.:

3-ONOP-041 .4 TURKEY POINT UNIT 3 STEP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 3.2 Subsequent Actions (continued)

CAUTION Opening both 60 gpm Orifice isolation valves may cause letdown flow to exceed 120 gpm due to higher than expected flows passing through the orifices that are worn. This may cause channeling through the demineralizers.

2. iCK R-3-20, REACTOR COOLANT EXIT this procedure.

LETDOWN Monitor high alarm is the result of high letdown piping radiation levels.

A. CHECK the following to maximum RCS letdown cleanup flow:

(1) Letdown 60 gpm orifice isolation PLACE CV-3-200B and CV-3-2000 or valves open: available orifice isolations valves in service to maximize letdown in accordance with

a. CV-3-200B. 3-NOP-047, CVCS Charging and Letdown.

b. CV-3-200C.

(2) At least two charging pumps START at least two charging pumps.

running.

NOTE If standby demineralizer is placed in operation, Chemistry should monitor RCS pH and add Lithium as necessary.

B. REQUEST Chemistry determine the decontamination factor (DF) of the demineralizer(s) in operation.

PTN L-154 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 K/A # E02 EK3.4 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to the (SI Termination) 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.

Proposed Question: RO Question # 63 Given the following conditions:

• Unit 3 experiences a Safety Injection.

• Total AFW flow is throttled to 450 gpm.

• The crew transitions from 3-EOP-E-0, Reactor Trip or Safety Injection to 3-EOP-E-1, Loss of Reactor or Secondary Coolant.

• The crew is determining SI Termination criteria with the following:

- Containment temperature is 165°F and slowly decreasing.

- Pressurizer level is 17% and rising.

- RCS subcooling is 58°F and stable.

- RCS pressure is 1550 psig and stable.

- SG Levels are 5% and rising.

Which ONE of the following describes the correct procedural response to these conditions?

A. Terminate SI since all criteria is satisfied.

B. Do NOT terminate SI since RCS pressure is too low.

C. Do NOT terminate SI since SG levels are too low.

D. Do NOT terminate SI since PZR level is too low.

Proposed Answer: B PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMA11ON Explanation (Optional):

A. Incorrect. Plausible to believe that all criteria for SI termination are met if examinee does not recognize that RCS pressure is too low.

B. Correct. RCS pressure is too low; SI termination requires 1625 psig for RCS pressure.

C. Incorrect. Plausible to believe since SG levels are below the setpoint of 7%. However, the criteria is a SG level >7% OR total AFW feedflow > 400 gpm.

D. Incorrect. Plausible to remember adverse value for PZR level and not realize that containment is not adverse.

Technical Reference(s): 3-EOP-E-1 Rev 5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902328 Obj. 4, 6 (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Seabrook Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Corn ments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 2 K/A # E02 EK3.4 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to the (SI Termination)

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.

Proposed Question: RO Question # 63

-J The following plant conditions exist:

z

- the plant has experienced a small break LOCA

- total EFW flow has been throttled to 550 GPM based on RCS temp less than 557 degrees.

- the crew has transitioned from E-0, reactor trip or safety injection to E-l, loss of reactor or secondary coolant and now to ES-i .1, SI Termination in order to reduce ECCS flow

-plant parameters are as follows:

- containment pressure is 1 .5 psig and slowly decreasing

- pressurizer level is 40% and increasing

- RCS subcooling is 43 degrees and stable

- RCS pressure is 1950 psig and stable

- the crew is terminating SI

- after placing the first CCP in standby, RCS pressure starts to slowly decrese.

Which of the following describes the correct procedural response to these conditions?

A. Restart the CCP and go to E-0, reactor trip or safety injection B. Transition to ES-i .2, post LOCA Cooldown and Depressurization C. Restore normal charging path and control charging flow to maintain pressurizer level D. Initiate safety injection and transition to E-i, loss of reactor or secondary coolant Proposed Answer: B FOR OFFICIAL USE ONLY- LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Explanation (Optional):

A. incorrect. The OAS page is ES-i .1 only directs manual restart of ECCS pumps if RCS subcooling is lost or Pressurizer level can not be maintained greater than 7%. In this case, because the RCS PRESSURE is SLOWLY decreasing a transition to ES-i .2 is more appropriate. The OAS also would direct a transition back to E-i, step 1 if RCS inventory was truly challenged, not E-O, reactor trip or safety injection B. Correct. Step 2 of ES-i .1 directs a transition to ES-i .2, post LOCA cooldown and depressurizat ion, if RCS pressure is not stable or increasing after securing the first centrifugal charging pump C. incorrect. Restoring normal charging path and controlling charging flow to maintain pressurizer level isnot directed in either ES-i .i or ES-i .2 D. incorrect. In this case a SLOW degradation of RCS pressure after stopping the first centrifugal charging pump would only require a transition to ES-i .2.

Additionally, if the RCS pressure did lead to re-initiation criteria the procedure directs manually start EGOS pumps as required, not initiate safety injection Technical Reference(s): ES1.i SI TERMINATION, step 2 and OAS page (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Li 2261 O5RO, Summarize the major actions of ES-il, L1226l O6RO, state the bases for the following notes or cautions for ES 1.1, manual actions required to Learning Objective: restart safeguards equipment (As available) following SI reset. L1203l O1RO, Recognize the symptoms and entry conditions for E-l, L12041, Recongnize the symptoms and entry conditions for ES-i .2 Question Source: Bank # x Modified Bank # (Note changes or attach parent)

New FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Question History: Last NRC Exam: 2009 Seabrook Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43

.I.I.IIIIJ Comments:

z FOR OFFICIAL USE ONLY- LOIT 1-15-1 EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

LOSS OF REACTOR OR SECONDARY COOLANT 9 of 45 PROCEDURE NO.:

3-EOP-E-1 TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I f RESPONSE NOT OBTAINED I

11. Check if SI Flow Should be Terminated:

a. RCS Subcooling based on Core Exit a. Go to Step 12.

TCs GREATER THAN 19°F

[GREATER THAN Foldout Page Item 3 Adverse Value°F1

b. Secondary heat sink: b. IF neither condition satisfied, THEN go to Step 12.

Total feed flow to intact SIGs GREATER THAN 400 GPM OR Narrow Range Level in at least one intact S/G GREATER THAN 7%[27%]

c. RCS pressure: c. GotoStepl2.

• Pressure GREATE1 THAN 1625 PSIG[1 950 PSIG]

• Pressure STABLE OR INCREASING

d. PRZ level 1 d. Try to stabilize RCS pressure with GREATER THAN 7%[48%] Normal PRZ Spray.

Go to Step 12.

e. Charging capability AVAILABLE e. Go to Step 12.

f. Go to 3-EOP-ES1 .1, SI TERMINATION, Step I

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# E08 EK1.1 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to the (Pressurized Thermal Shock) Components, capacity, and function of emergency systems.

Proposed Question: RO Question # 64 Given the following conditions:

• Unit 3 experiences a major Steam Line Break inside containment concurrent with a Loss of Off-Site Power.

• 3A and 3B 4KV Buses are powered from the Emergency Diesel Generators.

• Containment Pressure Hi signal is actuated.

• Containment Temperature is 193°F.

• The crew is performing 3-EOP-FR-P. 1, Response to Imminent Pressurized Thermal Shock.

• Both HHSI and RHR Pumps have been stopped and placed in AUTO.

• 3A Charging Pump is running with 40 gpm Charging flow.

• Letdown is unavailable.

Which ONE of the following is (1) the preferred method to depressurize the plant and (2) the earliest allowable CET subcooling temperature to terminate the depressurization?

A. (1) Use Auxiliary Spray (2) 19°F B. (1) Use Auxiliary Spray (2) 82°F C. (1) Open one Pressurizer PORV (2) 19°F PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION D. (1) Open one Pressurizer PORV (2) 82°F Proposed Answer: D Explanation (Optional):

A. Incorrect. 1st part wrong. 2 part wrong. It is plausible to assume with no RCPs running that Charging flow will be used for Auxiliary Spray. However, Auxiliary Spray is the third option listed in 3-EOP-.FR-P.1. Step 15 RNO states to use one PRZ PORV. Plausible to believe that the examinee will not realize adverse containment conditions exist and that the termination point is RCS subcooling based on CETs < 83°F.

B. Incorrect. 1st part wrong. 2nd part right. It is plausible to assume with no RCPs running that Charging flow will be used for Auxiliary Spray. However, Auxiliary Spray is the third option listed in 3-EOP-FR-P.1. Step 15 RNO states to use one PRZ PORV.

C. Incorrect. 1st part right. 21 part wrong. Plausible to believe that the examinee will not realize adverse containment conditions exist and that the termination point is RCS subcooling based on CETs < 83°F.

D. Correct: Since no RCPs are running or can be started, Step 15 RNO states to use one PRZ PORV. Step 15.b states to depressurize RCS until RCS Subcooling based on Core Exit TCs is LESS THAN 29°F [83° F].

Technical Reference(s): 3-EOP-FR-P.1, Rev 3, Step 15 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: No Learning Objective: LP 6902336 Obj. 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 5541 8 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

3 RESPONSE TO IMMINENT PRESSURIZED THERMAL PROCEDURE NO SHOCK CONDITION 16 of 42 3-EOP-FR-P.1 TURKEY POINT UNIT 3 I STEP II ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I NOTE If RCPs are NOT running, voiding may occur in RCS during depressurization.

This will result in a rapidly increasing PRZ level.

15. Depressurize RCS To Decrease RCS Subcooling:(Attachment 9 May Be Used As A Reference)

a. Use Normal PRZ Spray a. Use PRZ PORV.

IF RCS can NOT be depressurized using any PRZ PORV, THEN establish Auxiliary Spray using Attachment 5.

b. Depressurize RCSunfilyofthe following conditions satisfied:

RCS Subcooling bsed on Core Exit TCs LESS THAN 29°F[83°F]

OR PRZ Level GREATER THAN 73%[60%]

OR RCS pressure LESS THAN 150 PSIG [200 PSIG]

c. Stop RCS depressurization

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KIA# ElO EA2.1 Importance Rating 3.2 Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS) Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Proposed Question: RO Question # 65 Given the following conditions:

• Unit 4 Reactor trips due to a loss of offsite power.

• The crew performs 4-EOP-ES-O.3, Natural Circulation Cooldown with Steam Void in Vessel (With RVLMS).

Subsequently:

• Train 4A RVLMS fails.

• Pressurizer level is 6% and lowering rapidly.

Which ONE of the following identifies the required procedural response?

A. Remain in 4-EOP-ES-O.3, Natural Circulation Cooldown with Steam Void in Vessel (With RVLMS).

B. Transition to 4-EOP-E-1, Loss of Reactor or Secondary Coolant.

C. Transition to 4-EOP-ES-O.O, Rediagnosis.

D. Transition to 4-EOP-E-O, Reactor Trip or Safety Injection.

Proposed Answer: D Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-154 DRAFT NRC EXAM SECURE INFORMATION A. Incorrect. Plausible to believe that since only one train of RVLMS failed, the other is available and the procedural requirement is to continue in 4-EOP-ES-O.3. However, being unable to maintain PRZ level >7% requires a Phase A, SI, and transition to 4-EOP-E-O, step 1.

B. Incorrect. Plausible to believe that the decreasing PZR level is an indication of a LOCA and a transition to 4-EOP-E-1, Loss of Reactor or Secondary Coolant, would be the appropriate procedure transition. However, and being unable to maintain PRZ level

>7% requires a Phase A, SI, and transition to 4-EOP-E-0, step 1.

C. Incorrect. Plausible to believe that since 4-EOP-ES-0.0 provides a mechanism to allow the operator to determine or confirm the most appropriate Post Accident Recovery procedure. However, entry conditions also include an SI is actuated or required, which is not yet in place. Being unable to maintain PRZ level >7% requires a Phase A, SI, and transition to 4 EOP-E-0, step 1.

Plausible to believe that the decreasing PZR level is an indication of a LOCA and with the plant already shutdown, a transition to 4-ONOP-041 .7, Shutdown LOCA, would be the appropriate procedure transition. However, the plant had been at power, the crew is already in the EOP network, and being unable to maintain PRZ level >7% requires a Phase A, SI, and transition to 4-EOP-E-0, step 1.

D. Correct. Being unable to maintain PRZ level >7% requires a Phase A, SI, and transition to 4-EQ P-E-0, step 1.

Technical Reference(s): 4-EOP-ES-O.3 Rev 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination:

Learning Objective: LP 6902324 Obj. 6 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

2 NATURAL CIRCULATION COOLDOWN PROCEDURE NO.: WITH STEAM VOID IN VESSEL (WITH RVLMS) FOLDOUT 4-EOP-ES-O.3 TURKEY POINT UNIT 4 FOLDOUT PAGE For Procedure 4-EOP-ES-O.3

1. SI ACTUATiThLIRtERIA

a. 11 4-EOP-ES-0.2, NATURAL CIRCULATION COOLDOWN, was entered with Unit 4 in Mode 1, 2, or 3 (greater than 1000 psig), AND either condition listed below occurs, THEN actuate SI, actuate Containment Isolation Phase A, and go to 4-EQP-E-0, REACTOR TRIP OR SAFETY INJECTION, Step 1 RCS Subcooling based on Core Exit TCs LESS THAN 19°F OR 1

P RZ level CAN NOT BE MAINTAINED GREATER THAN 7%

b. 4-EOP-ES-0.2, NATURAL CIRCULATION COOLDOWN, was entered with Unit 4 in Mode 3 (less than 1000 psig), or Mode 4, AND either condition listed below occurs, THEN go to 4-ONOP-041 .7, SHUTDOWN LOCA [MODE 3 (LESS THAN 1000 PSIG)

OR MODE 4], Step 1:

RCS Subcooling based on Core Exit TCs LESS THAN 19°F OR PRZ level CAN NOT BE MAINTAINED GREATER THAN 7%

2. CST MAKEUP WATER CRITERIA IF CST level decreases to less than 12%, THEN add makeup to CST using 4-NOP-018.01, CONDENSATE STORAGE TANK (CST).

3. AFW SYSTEM OPERATION CRITERIA

a. IF two AFW Pumps are operating on a single train, THEN one of the pumps shall be shut down within one hour of the initial start signal.

b. IF two AFW Trains are operating AND one of the AFW Pumps has been operating at low flow of 80 gpm OR less for one hour, THEN that AFW pump shall be shut down.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 1 KIA# Gi 2.1.3 Importance Rating 3.7 Conduct of Operations: Knowledge of shift or short-term relief turnover practices.

Proposed Question: RO Question # 66 Which ONE of the following identifies the required reviews prior to assuming the Unit 3 RCO responsibility in accordance with O-ADM-202, Shift Relief and Turnover?

A. Check LMS for quals, review eSOMs for clearances, and check Critical Safety Function Status Trees B. Check LMS for quals, review TSA log book (TCC Index), and review Schedule of Plant Checks and Surveillances (Red Book)

C. Check Watchstander Out of Service Book, review eSOMs for clearances, and perform a Minimum Equipment List check D. Check Watchstander Out of Service Book, review TSA log book (TCC Index), and review Annunciator Status Log Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible to believe that the status of CSFs would be checked, however, unless the unit is NOT in the EOP network, these would not be required.

B. Incorrect. Plausible to believe that the RCO needs to review Schedule of Plant Checks and Surveillances (Red Book), however this is the responsiblility of the SM/US/WCCS.

C. Incorrect. Plausible to believe that the Minimum Equipment List check would be completed, howevere, the equipment status lists are completed by the SNPO/NPO/ANPOs.

D. Correct. Watchstander Out of Service Book, review TSA log book (TCC Index), and review Annunciator Status Log are some of the required reviews for the oncoming RCO.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Technical Reference(s): 0-ADM-202 Rev 6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902022 Obj. 1, 2 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN LOIT Exam Bank Item 1.1.23.22.1.6. Administered to RCO Group 19 Audit Exam.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

5.0 PROCEDURE 5.1 Shift Turnover Requirements 5.1.1 Operations Department perscnne1 ccnthe off-going shift should complete the applicable Shift Relief Checklists on a form similar to Attachments I through 7, or cmp1ete applicable section of the online turnover report.

5.1.2 Under normal conditions, shift turnovers should be made at the applicable operator work station.

5.1.3 The off-going shift should make checks and remarks on the required Shift Relief Checklists or online turnover report in such a manner as to inform the oncoming shift of the following (as a minimum):

1. Current plant / watchstation status, equipment status, alarm status, and any abnormal or infrequent lineups.

2. Transients, procedures, evolutions, and other work in progress on the watchstation.

3. Evolutions / activities performed last shift and evolutions / activities expected for the upcoming shift.

4. Review of station logs, out of specification readings, and any turnover notes.

5. Unusual events that have occurred during the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6. If the Plant is in Mode 5 or 6 or is defueled, provide shutdown risk status, current protected/operating train, and expected changes during the next shift.

5.1.4 Prior to assuming the shift. the oncoming operators should review pertinent documents such as logbooks, logsheets, and special instructions as specified in the following substeps and discuss with the shifts supervision watchstanding standards and expectations if not normally assigned to that shift. This discussion need only take place prior to assuming the first watch with a new shift if the operator will be working with that shift continuously for some period of time (e.g., if working 5 consecutive shifts to maintain license active with the same crew, discussion is only required prior to first shift). This discussion is not required for the SM:

1. The oncoming Shift Manager (SM) and Unit Supervisor (US) shall perform the following:

a. Review the Special Instruction Book to ensure that all applicable instructions have been read.

b. Review the Night Orders to ensure that pertinent information is reviewed with appropriate shift personnel as soon as is practical after shift turnover.

c. Review the Watchstander Out-of-Service Book to ensure that the individuals assigned to stand shift are qualified to assume the shift.

[Commitment Step 2.3.1]

W2003/MBS/nls/cls/els

5.1 .4.1 (Contd)

d. Review the Equipment Out-of-Service Books.

e. If any item(s) could not be completed, notify the Shift Manager (SM).

2. Oncoming Field Supervisor shall perform the following:

a. Review the Special Instruction Book to ensure that all applicable instructions have been read.

b. Review the Watchstander Out-of-Service Book to ensure that the individuals assigned to stand shift are qualified to assume the shift.

[Commitment Step 2.3.1]

c. Review the Equipment Out-of-Service Books.

d. If any item(s) could not be completed, notify the Shift Manager (SM).

3. Oncoming Reactor Control Operators (RO) shall perform the following:

a. Review the RO Rounds Log readings recorded in the Plant Computer System. As a minimum, the Abnormal / Noted Log Readings for the tour date relieving.

b. Review the Equipment Out of Service (EOOS) Logbooks.

c. During the board walkdown, the RO should familiarize with the Control Room deficiencies addressed in the CRDL and any flags or STAR stickers for items out of normal configuration, and discuss with the off-going RO any applicable contingencies or compensatory actions.

d. Review active clearances prepared in accordance with O-ADM-2 12, Tn-Plant Equipment Clearance Orders, for the assigned unit(s) back to the last shift worked (in Modes I through 4).

e. Review the Special Instruction Book to ensure that all applicable instructions have been read.

f. Review the Temporary System Alteration (TSA) Logbook.

g. Review the Watchstander Out-of-Service Book to ensure that the individuals assigned to stand shift are qualified to assume the shift.

[Commitment Step 2.3.1]

h. Review the Annunciator Status Log (ASL) for applicable compensatory actions for defeated or locked in alarms.

i If any item(s) could not be completed, notify the Shift Manager (SM).

AIflflflO IRAOOI...I- !-.I., I-I,.

5.1 .4 (Contd)

4. The oncoming RO should announce to the crew within the surveillance area that he / she has assumed the unit responsibility.

5. Oncoming Nuclear Operators/Senior Nuclear Plant Operators (NO/SNPO).

Nuclear Plant Operators (NPO). Associated Nuclear Plant Operators (ANPO) shall perform the following:

a. Review their respective watchstation Operator Rounds Log readings recorded in the Plant Computer System. As a minimum, the Abnormal/

Noted Log Readings for the tour date relieving.

b. If any item(s) could not be completed, notify the Field Supervisor.

5.1.5 Oncoming operators should review the shift relief checklists, check the relief boxes provided, indicating acceptance of the shift OR complete the online turnover report for the applicable watchstation (preferred).

5.1.6 Off-going operators should not be relieved until the equipment they are responsible for is in a stable condition (Refer to Section 4.0).

5.1.7 Off-going operators with fire team duties or 0-ONOP-105, Control Room Evacuation, duties shall ensure that the operators relieving them of these duties are aware of and accept their assignments. New assignments should be noted on the Unit 3 RO Logbook Program.

5.1.8 The off-going Field Supervisor shall assign the duties of 0-ONOP-105, Control Room Evacuation, to an oncoming operator and note the assignment on the Shift Relief Checklist and in the Unit 3 RO Logbook Program.

5.1.9 The off-going Work Control Center Supervisor (WCCS) and/or (US) assigned to the Work Control Center (WCC) are not required to complete Attachment I, but should perform a verbal turnover to oncoming operator if possible.

5.1.10 If the LAN Narrative Logbook Program is OOS, then the oncoming and off-going operator(s) should sign the Shift Relief Checklist(s) indicating review of the checklist for completion, and completion of the turnover.

CATTAuIMENT 3, (Page I of 1)

- UNIT DUTY SHIFT RELIEF CHECKLIST Relief Time: Date: Unit 4 Off-going RO..print name) Oncoming RO: (pnnt name)

PARTj-.Tab completed by Off-going RO Unit 3 Mode: ..  % Power

• Gross MWe *(* Record N/A in Modes 2-6)

Unit Mqde  % Power

• Gross MWe *(* Record N/A in Modes 2-6)

Sourçe Rngei N31 or N32r---- cps Intermediate Range: N35 or N36 amp

(**RedOrd N/ in Mode 1)1 ReacUVlChanges: Dilutin A gals, Boration gals, D-Bank steps Tavg IF / .Presure psig RCS Boron ppm ROS Leakage Determined Date Time GRP in progress # / Tan[ 4 LRP in progress 4 Tank 4 Ctmt Purge Permit in p/ogress#

WTP 000S S OSeryiceto Boric Acid Storage Tk Levels A

iB C Accumulator Reference Level/Date A . SI /. B /______ C I_____________

Flags (STAR stickers) affixed to all Unit and ftomrnin Cohotrd Room items that are out of normal configuration.

Abnormal Annunciator Status Reason: I Major EOOS/LCOs. i Operations, Procedures, Transients, or Significant Maintenance in Phigress/*:

• Mode 5. 6, Defueled

- - Provide shutdown risk status, protected/operating train, expected changes during next shift.

E8!IZ - To be ReewedIcompieted by Oncorving RD prior to shift relief (Check boii .

LI RD Logbook LI In compliance with hcerise restrictions, if any Q RO Logs Abnormal/Noted readings Q SCBA correcliwi iense readiiy avaiiabie if corrective lenses LI TSA Logbook required by license LI Waik Down Control Boards C] If on overtime this shift rcci overtime on the next shift in this LI Ciesrance Book (in Modes 1-4> for clearances position, or if a shift trade is invoived. sign overtime sheet issued since iast shift worked. (Attachment 8 of O-ADM-202)

(No check required ri Modes 5-6) LI Controi Room Dwiiciendy Log LI Watchstander Out-of-Service Book LI EOOS Logbooks (All LI check blending station controiiers are set for current RCS boron concentration LI Watch Station eaniiness ..1 LI Special Instruction Book LI if first watch with a new shift discuss LI Reviev Annunciator.Stalsis Slseea with shift supervision their watchstanding C] Revien LMS to verify no overdue items standards and expectations quaFfications that wouid prevent assumii Required prior to accepting shift responsibftity.

PART 3- To be accomplished by RD pnor to end of shift (Check box)

C] Notify Field Supervisor of any schedule or status changes and ensure the Overtime Manager is updated.

Shift Responsibility Turned Over By:

Shift ResponsibilityAccepted By:

F486 Rev, 5 tO-ADM-202) lPJOfttIMRfkkIrleIrle

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 1 K/A# Gi 2.1.42 Importance Rating 2.5 Conduct of Operations: Knowledge of new and spent fuel movement procedures.

Proposed Question: RO Question # 67 Given the following conditions:

• Operators are preparing for Unit 4 core off-load.

• Reactor is shutdown for 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br />.

• Refueling Canal boron concentration is 2395 ppm.

• SFP Level is 56ll.

Which ONE of the following identifies if core off-load may begin or not and the reason?

A. Core offload may begin since all requirements are met.

B. Minimum SEP level is NOT met to begin core offload.

C. Minimum Refueling Canal boron is NOT met to begin core offload.

D. Minimum time is NOT met to begin core offload.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible to not remember that TS 3.9.3, Decay Time, requires the reactor must be subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to irradiated fuel movement and therefore to believe that all requirements are met.

B. Incorrect. Plausible to believe that SEP level minimum is NOT met since the SFP Low Level alarm Annunciator Hi/l alarms at 57O. However, TS 3.9.11, Water level Storage Pool, requires that SEP level be greater than or equal to 5610.

C. Incorrect. Plausible to remember that minimum Boron concentration is 2400 ppm.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION However, this is the Boron concentration for the RWST when in MODES 1-4.

D. Correct. TS 3.9.3, Decay Time, requires that the reactor be subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to fuel movement.

Technical Reference(s): TS 3.9.3 Rev 293 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902144 Obj. 10 (As available)

Question Source: Bank # 99784 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritic.... hours.

APPLICABILITY: During movemeneI in the reactdi vessel.

ACTION:

With the reactor subcritical for less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, suspend all operations involving movement of irradiated fuel in the reactor vessel.

SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel.

TURKEY POINT UNITS 3 & 4 3/4 9-3 AMENDMENT NOS. 223 AND 218

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 2 KIA# G2 2.2.17 Importance Rating 2.6 Equipment Control: Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, coordination with the transmission system operator.

Proposed Question: RO Question # 68 Given the following conditions:

• Unit 4 is at 100% power.

• 4A and 4B Component Cooling Water Pumps are posted as Guarded Equipment.

• Maintenance requests permission to erect scaffolding inside the Guarded Area.

Which ONE of the following identifies the position responsible for authorizing the work inside the Guarded Area per OP-AA-102-1 003, Guarded Equipment?

A. Work Week Manager B. Maintenance Manager C. Shift Technical Advisor D. Shift Manager Proposed Answer: D Explanation (Optional):

A. Incorrect: Plausible to believe since weekly work activities and questions are coordinated and tracked by the Work Week Manager. However, the Shift Manager has the responsibility to authorize work activities performed on Protected Train Equipment, Guarded Equipment or beyond a Guarded Equipment Posting.

B. Incorrect: Plausible to believe since the Work Activity Supervisor has responsibilities within OP-AA-102-1003, that a Maintenance Supervisor would get authorization from the Maintenance Manager. However, the Shift Manager has the responsibility to PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION authorize work activities performed on Protected Train Equipment, Guarded Equipment or beyond a Guarded Equipment Posting.

C. Incorrect: Plausible to believe since the Shift Technical Advisor has responsibilities within OP-AA-102-1003 and provides support to the Shift Manager in identifying equipment to be guarded and to shift operators in in selecting the methods (tags, signs, barriers, etc.) and placements of Guarded Equipment postings. However, the Shift Manager has the responsibility to authorize work activities performed on Protected Train Equipment, Guarded Equipment or beyond a Guarded Equipment Posting.

D. Correct: Per OP-AA-102-1003, Guarded Equipment, the Shift Manager has the responsibility to authorize work activities performed on Protected Train Equipment, Guarded Equipment or beyond a Guarded Equipment Posting.

Technical Reference(s): OP-AA-102-1003 Rev 5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: LP 6902025 Obj. 2 (As available)

Question Source: Bank #

Modified Bank # 100878 (Note changes or attach parent)

New Question History: Last NRC Exam: 2012 Point Beach Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

FOR OFHCIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 2 K/A# G2 2.2.17 Importance Rating 2.6 Equipment Control: Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, coordination with the transmission system operator.

Proposed Question: RO Question # 68

-J 1P-15A Safety Injection Pump is posted as GUARDED EQUIPMENT. A contractor has called you, Unit 1 Reactor Operator, and requested permission to erect scaffolding z

inside the posted area.

Are activities allowed to be performed in the area of IP-15A and if so, with whose authorization?

A. No activities are allowed.

B. Yes, Unit 1 OS.

C. Yes, the Work Control Center Manager.

D. Yes, the Operations Manager.

Proposed Answer: D Explanation (Optional):

A. Incorrect: Generally no activities are allowed around and above guarded equipment due to the safety significant need of the equipment. Exceptions are approved by the Operations Manager.

B. Incorrect: Plausible as the unit SRO is asked or informed for virtually everything.

C. Incorrect: Plausible, as routine work activities and questions are handled by the WCC.

D. Correct: Per NP 2.1.8 the Operations Manager must give permission.

FOR OFFICIAL USE ONLY- LOIT L-15-1 EXAM SECURE INFORMATION

FOR OFFICIAL USE ONLY - LOIT L-15-1 EXAM SECURE INFORMATION Initial Rough Draft 091214 NP 2.1.8, Guarded Equipment, Technical Reference(s): (Attach if not previously provided)

Section 3.2 Proposed References to be provided to applicants during examination: NO Knowledge of the process for Learning Objective:

managing maintenance activities during power operations, such as risk assessments, work prioritization, and (As available)

J coordination with the transmission system operator. (SD 86.2 02.02.17)

Question Source: Bank #

Modified Bank #

New 100878 (Note changes or attach parent)

Z Question History: Last NRC Exam: 2012 Point Beach Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 Comments:

FOR OFFICIAL USE ONLY- LOIT L-15-1 EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

GUARDED EQUIPMENT 6 Of 103 PROCEDURE NO.:

OP-AA-1 02-1003 NUCLEAR FLEET ADMINISTRATIVE 3.0 RESPONSIBILITIES 3.1 Shift Manager

1. Identify equipment to be guarded. The Shift Manager may utilize the site-specific information attachment as a reference when determining what equipment to guard. However, the final determination of what equipment to guard is always at the discretion of the Shift Manager, in accordance with the requirements of this procedure.

2 activities performed on Protected Train Equipment, Guarded Eq!n4efor beyond a Guarded Equipment Posting Exemptions from this requirement are contained within the site-specific information attachments. The Shift Manager may delegate this responsibility to a designee.

3 Ensures rnneI are briefed on Guarded Equipment prior to authorizing entry beyond the posting. The responsibility for the briefing may be delegated to a SRO cognizant of the conditions and basis for the posting.

4. When equipment guarding is no longer required, authorizes removal of the postings.

3.2 Shift Technical Advisor (STA)

1. Supports the Shift Manager in identifying equipment to be guarded.

2. Supports shift operators, as necessary, in selecting the methods (tags, signs, barriers, etc.) and placements of Guarded Equipment postings..

3.3 Work Control Senior Reactor Operator or STA

1. Supports the Shift Manager in identifying equipment to be guarded.

2. Evaluates work requests for unrelated equipment that may jeopardize Guarded Equipment (e.g., installation / removal of scaffolding in the vicinity of Guarded Equipment).

3. When it is required to Guard Equipment on an emergent basis, perform a review of scheduled and/or in-progress activities to determine what work should be postponed or stopped to prevent adversely affecting the newly Guarded Equipment.

3.4 Unit Supervisor I Control Room Supervisor

1. Supports the Shift Manager in identifying equipment to be guarded.

2. Supports shift operators, as necessary, in selecting the methods (tags, signs, barriers) and placements of tags, signs and/or barriers for Guarded Equipment.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 2 KIA# G2 22.18 Importance Rating 2.6 Equipment Control: Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Proposed Question: RO Question # 69 Given the following plant conditions:

• Unit4 isin MODE 5.

• 4A RHR loop is in operation.

Which ONE of the following satisfies a criterion for the RCS Loop Filled requirement of O-ADM-051, Outage Risk Assessment and Control?

A. 4B RCP is operating.

B. Reactor Coolant System pressure is 50 psig.

C. Unit 4 Steam Generator Wide Range levels are 11%.

D. No intervening evolutions that could introduce air into S/G U-tubes.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Plausible to believe that an RCP operating would be sufficient to consider the RCS loops filled. However, no such criteria exist in 0-ADM-051 for the loops to be considered filled.

B. Incorrect. Plausible, since the RCS must be pressurized, however RCS pressure must be greater than 100 psig. NOTE prior to Step 5.1.1 .2 states that RCS loops cannot be considered a valid loop once RCS pressure has been decreased below 100 psig.

Plausible to remember this NOTE as RCS pressurized below 100 psig.

C. Incorrect. Plausible to believe since 0-ADM-051 states that at least two S/G levels greater than 10% is a criteria, however that requirement specifies Narrow Range not PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Wide Range and 0-ADM-051 does use Wide Range indication for some instrumentation channels, but SIG level is not one of them.

D. Correct. One of the criteria in 0-ADM-051 is that RCS filled and vented with no intervening evolutions that could introduce air into the SIGs. (Step 5.1.1.2.a)

Technical Reference(s): O-ADM-051 Rev 14A (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902032 Obj. 2 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN LOIT Exam Bank Item #1.1.23.32.3.8 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

5.0 PROCEDURE In the event of Control Room evacuation during refueling activities, 0-ONOP- 105, Control Room Evacuation, shall be followed in accordance with notations addressing unit mode (see note prior to Section 4.0 in 0-ONOP- 105, Control Room Evacuation, for example).

51 General Precautions and Instructions 5.1.1 The following items are general precautions and limitations that should be observed when operating either unit in a shutdown condition:

1. The figures and notes provided by this procedure in the Outage Risk Assessment Notebook are not to be used by Operations or any other plant personnel in lieu of controlled plant procedures, design documents, and Technical Specifications. The purpose of this information is to provide adequate information to assist Operations in recognizing and responding to abnormal configurations that may impact the Key Safe Shutdown Functions and Equipment. They may be used also as a guideline for outage planning.

CAUTION The RCS loops can not be considered a valid coolant loop once RCS pressure has been decreased below 100 psig until Substep 5.1.1.2 has been completed.

2. The RCS loops shall not be considered filled per Tech Spec 3.4.1.4.1 unless the following conditions are met:

a. The RCS has been filled and vented with no intervening evolutions that could introduce air into the steam generators (e.g., RCS level has emained above the reactor vessel nozzles)

AND b The RCS is pressurized (either water solid or with a pressurizer steam bubble) and above 100 psig.

AND 1

c At least two steam generators are filled to greater than 10 percent Narrow Range Level.

3. Entering a Reduced Inventory condition shall be prior approved by the Operations Manager and Plant Manager.

W97:DJT/cls/ab/cls

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 3 KJA# G3 2.3.12 Importance Rating 3.2 Radiation Control: Knowledge of Radiological Safety Principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Proposed Question: RO Question # 70 Given the following conditions:

• Unit 4 is in MODE 6 with a core offload is in progress.

Subsequently:

• A fuel assembly is horizontal in the Fuel Transfer Cart in the SFP Transfer Canal.

• Rl-4-1422B, Unit 4 Spent Fuel Pit South Wall, alarms.

• Rl-4-1422B indicates FAIL.

Which ONE of the following describes the crews required actions in the Spent Fuel Pool per 4-ONOP-038. 1, Loss of Refueling Equipment or Support Function?

A.

• Immediately notify the Outage Control Center.

• Place the fuel assembly in its designated location.

B.

• Suspend all fuel transfer movements.

• Leave the fuel assembly in the fuel transfer cart in the fully lowered position.

C.

• Immediately exit the Spent Fuel Pool area.

• Report to the RCA Control Point for monitoring.

D.

• Verify area dose rates using Electronic Pocket Dosimeter.

• Continue core offload activities.

Proposed Answer: B Explanation (Optional):

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. Incorrect. Plausible to believe since the Outage Control Center is the central outage coordination point. Also plausible to believe that since the fuel assembly is in transit, it needs to be place in its designated storage location as soon as possible. However, this failure reduces Minimum Equipment as per 4-NOP-040.03, Fuel Handling and Insert Shuffle in the Spent Fuel Pit, and requires immediate suspension of fuel movement.

4-ONOP-038.1, Loss of Refueling Equipment or Support Function, requires suspension of all reactor cavity, SFP, and fuel transfer movements. There is a NOTE prior to Step 3 of 4-ONOP-038. 1 which states that although the preferred location for a fuel assembly is in its intended location in the SFP storage rack, a backup location is fully lowered in the transfer canal in the SFP.

B. Correct. This failure reduces Minimum Equipment as per 4-NOP-040.03, Fuel Handling and Insert Shuffle in the Spent Fuel Pit, and requires immediate suspension of fuel movement. 4-ONOP-038.1, Loss of Refueling Equipment or Support Function, requires suspension of all reactor cavity, SFP, and fuel transfer movements. There is a NOTE prior to Step 3 of 4-ONOP-038. 1 which states that although the preferred location for a fuel assembly is in its intended location in the SEP storage rack, a backup location is fully lowered in the transfer canal in the SFP.

C. Incorrect. Plausible since this appears to be a radiological event. However, there is no evidence of any airborne radioactive contamination, so monitoring by itself would not be required. Guidance in 4-ONOP-038.1, Loss of Refueling Equipment or Support Function, does not specifically require an evacuation, and since there is no evidence of EPDs in alarm, an evacuation would not be necessary.

D. Incorrect. Plausible to believe since there is still an ARMS Channel OPERABLE (Rl-4-1408B, Unit 4 SEP Canal Area Radiation Monitor), and no evidence of EPDs in alarm, it would be permissible to move the fuel assembly. However, this failure reduces Minimum Equipment as per 4-NOP-040.03, Fuel Handling and Insert Shuffle in the Spent Fuel Pit, and requires immediate suspension of fuel movement. 4-ONOP-038.1, Loss of Refueling Equipment or Support Function, requires suspension of all reactor cavity, SFP, and fuel transfer movements. There is a NOTE prior to Step 3 of 4-ONOP-038.1 which states that although the preferred location for a fuel assembly is in its intended location in the SFP storage rack, a backup location is fully lowered in the transfer canal in the SFP.

4-ONOP-038.1 Rev 0 . .

Technical Reference(s): (Attach if not previously provided) 4-NOP-040.03 Rev 11 Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902285 Obj. 1, 4, 5 (As available)

Question Source: Bank #

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

11 FUEL HANDLING AND INSERT SHUFFLE IN THE SPENT FUEL PIT 22 of 36 PROCEDURE NO.:

4-NOP-040.03 TURKEY POINT UNIT 4 ATTACHMENT I Pre-Shuffle Minimum Equipment Checklist (Page 2 of 5)

Minimum Nights Time Allowed Equipment! Days Requirement Before Stoppin Applicable Remarks Conditions Initials Initials Tech. Spec.

for Shuffle Fuel Movement ARMS RI-4-1408B and Remote Rl-4-1422B OPERABLE IF Area Monitor is NOT operable, THEN Remote/Local ,jMMEDIATE N/A Local INSTALL a portable monitor with an alarm.

Indications and Alarms OPERABLE IF Tube Gate is OPEN, THEN CHECK daily.

Greater than Spent Fuel Pit Boron IF Tube Gate is CLOSED, THEN CHECK most orequalto N/A IMMEDIATE 3.9.14.b Concentration recent SFP boron sample, and an additional 2300 ppm sample is NOT required.

PRMS SFP Area High 1 1 Table 3.3-4 CHECK twice per shift. Either R-14 OR Plant Gaseous Radioactivity OPERABLE IMMEDIATE Vent SPING Channel 5 satisfies monitoring Item 2b Monitor 2 2 Action 28 requirement.

Communications 1 1 headsets/other reliable CONTINUOUS IMMEDIATE N/A CHECK twice per shift.

communication system Control Room to SEP 2 2 1 1 RP Coverage CONTINUOUS IMMEDIATE N/A CHECK twice per shift.

2 2

STEP ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

Suspend All Reactor Cavity, SFP, And Fuel Transfer Fuel Movements 2 Notify The Shift Manager Of The Problem NOTES The following steps to place a fuel assembly in a preferred location after stopping refueling activities may not be necessaiy. It may be possible to suspend activities with a fuel assembly in transit and temporarily leave the fuel assembly in the As Left location.

• The preferred location for a fuel assembly is in its intended location in the reactor core or the SFP storage rack. A backup location for a fuel assembly is fully lowered in the transfer canal in the SFP.

• Per Tech Specs, the suspension of CORE ALTERATIONS because of the loss of required equipment or support function shall not preclude completion of movement of a component to a safe location.

I.... _ a a a I 3 Check If Fuel Assembly Should Be Moved To A Preferred Location

a. Check if either of the following applies a. IF fuel movement will be suspended with no further movement of fuel assemblies in transit, THEN go to Step 7.

The fuel handling stations will be vacated The Refueling SRO directs that any fuel assemblies being moved shall be placed in a preferred location W97:ev/In/cls

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 3 KIA# G3 2.3.13 Importance Rating 3.4 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.

Proposed Question: RO Question # 71 Given the following conditions:

• Unit 3 is at 100% power.

• Reactor Engineering is in the Control Room performing Unit 3 flux map.

• Maintenance is preparing to perform a 0-ADM-537, Boric Acid Corrosion Program, inspection at the Unit 3 Seal Table.

Which ONE of the following correctly completes the statements below?

The (1) must give permission for a Containment entry at power. For the stated conditions above, the inspection at the Seal Table (2) allowed.

A. (1) Plant General Manager (2) is B. (1) Shift Manager (2) is C. (1) Plant General Manager (2) is NOT D. (1) Shift Manager (2) is NOT PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Proposed Answer: C Explanation (Optional):

A. Incorrect. 1 st part right. 2 part wrong. Plausible to believe since the inspection does not involve any entry into the system and is simply a visual observation for evidence of boric acid. However, 0-ADM-009 requires that the drive switches be tagged out on a clearance, or if work is being performed on the Flux Map System, authorization be obtained from the Radiation Protection Manager. In this case, no work is being performed on the Flux Map System, therefore access is denied until the clearance is hung.

B. Incorrect. 1 st part wrong. 2 part wrong. Plausible to believe since the containment entry and exit data sheets are reviewed by the Shift Manager and Completed Documentation of Approval for entry by Shift Manager and RPM (RP-AA-1 03-1002, High Radiation Area Controls) are listed in Attachment 11. However, Section 3.5 of 0-ADM-009 states that the Plant General Manager shall approve all power entries. Also plausible to believe since the inspection does not involve any entry into the system and is simply a visual observation for evidence of boric acid. However, 0-ADM-009 requires that the drive switches be tagged out on a clearance, or if work is being performed on the Flux Map System, authorization be obtained from the Radiation Protection Manager.

In this case, no work is being performed on the Flux Map System, therefore access is denied until the clearance is hung.

C. Correct. Section 3.5 of 0-ADM-009 states that the Plant General Manager shall approve all power entries and 0-ADM-009 requires that the drive switches be tagged out on a clearance, or if work is being performed on the Flux Map System, authorization be obtained from the Radiation Protection Manager. In this case, no work is being performed on the Flux Map System, therefore access is denied until the clearance is hung.

D. Incorrect. 1 st part wrong. 2r,d part right. Plausible to believe since the containment entry and exit data sheets are reviewed by the Shift Manager and Completed Documentation of Approval for entry by Shift Manager and RPM (RP-AA-103-1002, High Radiation Area Controls) are listed in Attachment 11. However, Section 3.5 of 0-ADM-009 states that the Plant General Manager shall approve all power entries.

Technical Reference(s): 0-ADM-009 Rev 17 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTh L-15-1 DRAFT NRC EXAM SECURE INFORMA11ON Question Source: Bank # 99795 Modified Bank # (Note changes or attach parent)

New Question History Last NRC Exam: 2010 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10CFRPart5SContent 55.41 12 55.43 Comments:

Pm L-15-1 DRAFT NRC EXAM -SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

17 CONTAINMENT ENTRIES WHEN CONTAINMENT INTEGRITY IS ESTABLISHED 10 of 71 PROCEDURE NO.:

O-ADM-009 TURKEY POINT PLANT 3.4 Job Supervisor (continued)

8. Perform briefing as required per Attachment 2, Containment Entry Brief, for job personnel for the following:

A. SA-AA-1 00-1 008, Heat Stress Control B. OP-AA-1 000, Conduct of Infrequently Performed Tests or Evolutions 3.5 Plant General Manager Plant General MaII pour entries.

3.6 Operations Manager

1. Designates individuals to conduct a final visual containment closeout inspection.

2. Approves items on Attachment 12, Approved Equipment Checklist for Mode 5 When Containment Integrity is Established, (F703).

3.7 Radiation Protection Shift Supervisor (RPSS)

1. Assign RP coverage.

2. Perform brief as required per Attachment 2, Containment Entry Brief for the following:

A. RWP I Radiological Requirements B. Flux Mapper status / location of detectors

3. Ensures temporary RP postings for Locked High Radiation Areas and Very High Radiation Areas are listed on a form similar to Attachment 9, Engineering Evaluated Equipment for Mode 4 and Mode 3 when Containment Integrity is Established, of this procedure and are removed from containment prior to entering MODE 2 from MODE 3.

REVISION NO.: PROCEDURE TITLE: PAGE:

17 CONTAINMENT ENTRIES WHEN CONTAINMENT INTEGRITY IS ESTABLISHED 20 of 71 PROCEDURE NO.:

0-ADM-009 TURKEY POINT PLANT 4.2 Prerequisites for Entry (continued)

4. (continued)

C. Perform one of the following: *

(1) Verify a clearance for drive switches to the Flux Map System held by the Radiation Protection Manager oç.

qualified personnel designated by the Radiation Protection Manager

[Commitment Section 6.2, Commitment 2], or (25 If work is to be performed on the Flux Map System an a clearance is undesirable, obtain authorization from the Radiation Protection Manager to proceed with containment entry without the clearance, and document that authorization in the RPSS Log.

a. ontinuous Radiation Protection coverage is required when work is performed on the Flux Map System without a clearance.

5. The Entry Team shall perform the following:

A. Review RWP and ALARA requirements.

B. Complete appropriate section of form similar to Attachment 11, Crew Preparation Checklist.

C. Note the following: NO ACCESS areas in containment when entering at power.

(1) 58 elevation - reactor cavity (2) 14 elevation - reactor sump NOTE The Radiation Protection Manager or designee may allow entries within the biological shield wall while the reactor is critical (e.g., during low power operations, low power physics testing) on a case by case basis.

(3) 14 elevation - inside the biological shield wall (4) All elevations during planned power transients (does NOT include low power physics testing)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 3 KIA# G3 2.3.14 Importance Rating 3.4 Radiation Control: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question: RO Question # 72 Which ONE of the following evolutions describes a normal situation where radiological exposure could change rapidly by the work in progress?

A. Operations performs O-NOP-061.16, Spent Resin Storage Tank, for resin transfer and determines tank level by draining water below the top of resin.

B. Maintenance uses guidance from O-ADM-009, Containment Entries When Containment Integrity Is Established, to work on the Personal Hatch Inner Door inside Containment.

C. Chemistry performs required RCS samples after a 15% power change to verify O-ADM 651, Nuclear Chemistry Manual, limits are not exceeded.

D. Radiation Protection performs RP-TP-105-3007, Operation of the GEM-5 Gamma Exit Monitor, using the required source to perform functional alarm checks.

Proposed Answer: A Explanation (Optional):

A. Correct. O-NOP-061.16 provides direction to the Control Room or Resin Transfer Supervisor to make a notification over the Public Address System of potentially changing radiological conditions due to resin transfer. Reducing the water level below the top of the resin creates a situation where radiological exposure could change rapidly by the work in progress.

B. Incorrect. Plausible to believe that work in containment might involve rapidly changing radiological conditions. However, the work on the personnel hatch door does not involve opening of any systems and is outside the bio-wall in a low dose area. Reducing the water level below the top of the resin creates a situation where radiological exposure could change rapidly by the work in progress.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Incorrect. Plausible to believe taking a sample on the RCS could involve a situation where radiological exposure could change rapidly. However, with no indication of core damage or radiation process alarms, this would not result in rapidly changing radiological conditions.

D. Incorrect. Plausible to believe that a source would cause a situation where radiological exposure could change rapidly, however, this is a sealed source that has a constant output and will not change radiological conditions.

Technical Reference(s): O-NOP-061 .16 Rev I (Attach if not previously provided)

Proposed References to be provided to applicants during examination: NO Learning Objective: NUC GET RWT GEN RQL Obj 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

1 SPENTRESINSTORAGETANK 10of44 PROCEDURE NO.:

O-NOP-061 .16 TURKEY POINT PLANT 4.2.2 Resin Sluice

1. iEQUEST Control Room or Resin Transfer Supervisor make a notification over the Public Address System of potentially changing radiological conditions due to resin transfer.

2. OPEN 1918, SRST TO PORTABLE CASK ISOLATION VALVE.

NOTE Opening 4-220, PRIMARY WATER TO SPENT RESIN HEADER, will pressurize the Spent Resin Header.

3. OPEN 4-220, PRIMARY WATER TO SPENT RESIN HDR.

4. OPEN 1802, SPENT RESIN HEADER TRANSFER LINE CROSS CONNECT.

5. EXAMINE Spent Resin Header for leaks.

6. IF any header leaks are identified, THEN:

A. CLOSE 4-220, PRIMARY WATER TO SPENT RESIN HDR.

B. CLOSE 1802, SPENT RESIN HEADER TRANSFER LINE CROSS CONNECT.

C. CONTACT Maintenance Supervisor.

7. PERFORM the following to backflush the resin header:

A. OPEN CV-5098, RESIN TRANSFER TO CONTAINER.

B. START dewatering pump to dewater the Shielded Poly High Integrity Container (HIC) AND OBSERVE pump operability.

C. WHEN a solid stream of water flows into HIC, THEN:

(1) CLOSE 1802, SPENT RESIN HEADER TRANSFER LINE CROSS CONNECT.

(2) CLOSE 4-220, PRIMARY WATER TO SPENT RESIN HDR.

(3) CLOSE CV-5098, RESIN TRANSFER TO CONTAINER.

8. WHEN water is removed from HIC, THEN STOP dewatering pump.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 4 KIA# G4 2.4.13 Importance Rating 4.0 Emergency Procedures I Plan: Knowledge of crew roles and responsibilities during EOP usage.

Proposed Question: RO Question # 73 Given the following conditions:

• Alignment is complete for Unit 3 hot leg recirculation in accordance with 3-EOP-ES-1 .4, Transfer to Hot Leg Recirculation.

• Plant conditions are stabilized for the accident in progress.

• The highest Critical Safety Function Status Tree (CSFST) is a yellow path on Inventory.

Which ONE of the following identifies the CSFST monitoring requirement in accordance with 3-EOP-F-0, Critical Safety Function Status Trees?

A. Monitor continuously.

B. Monitored every 10 to 20 minutes.

C. Suspend monitoring since conditions are stable.

D. Suspend monitoring after 3-EOP-F-0 is initially performed.

Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible if examinee believes that J2Y non-satisfied condition (yellow path or higher) must be continuously monitored; however, this is only true if an extreme (red path) or severe (orange-path) challenge exists or plant conditions are changing rapidly.

B. Correct. If no extreme (red-path) or severe (orange-path) challenge exists and plant conditions are not changing rapidly, then CSFSTs are to be monitored every 10 to 20 minutes.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Incorrect. Plausible if examinee believes that stable plant conditions warrant suspension of CSFST monitoring; however, monitoring is a continuous action step and must continue until the yellow-path condition is corrected (and a procedural transition has occurred) or the plant is stable in Mode 5 with RHR cooling.

D. Incorrect. Plausible if examinee believes that stable plant conditions and a complete pass through 3-EOP-F-0 warrant suspension of CSFST monitoring; however, monitoring is a continuous action step and must continue until the yellow-path condition is corrected (and a procedural transition has occurred) or the plant is stable in Mode 5 with RHR cooling.

3-EOP-F-0 Rev 3 Technical Reference(s): NOTE prior to Step 1; (Attach if not previously provided)

Attachment 1, Step 8 Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902353 Obj. 4 (As available)

Question Source: Bank # 98705 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Turkey Point Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

CRITICAL SAFETY FUNCTION STATUS TREES 6 of 19 PROCEDURE NO.:

3-EOP-F-O TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 2 Iitor Critical Safety Functions Using Rules Of Usage Provided In Attachment I

3. Report Results To Control Room Operators 4 Determine If Critical Safety Function WHEN additional Critical Safety Function Monitoring Can Be Stopped: monitoring is required by Attachment 1, THEN observe NOTES prior to Step 1 and The emergency condition has been return to Step 1.

corrected AND a transition to the ppropriate plant procedure has Continue with procedure and step in effect.

een performed OR The plant is stable in CoIi Shutdown with lWR cooling established OR TSC staff has determined that Critical Safety Function monitoring is NO longer required

5. Return To Procedure And Step In Effect End of Section 3.0

REVISION NO.: PROCEDURE TITLE: PAGE:

CRITICAL SAFETY FUNCTION STATUS TREES 18 of 19 PROCEDURE NO.:

3-EOP-F-O TURKEY POINT UNIT 3 ATTACHMENT I Rules of Usage for Critical Safety Function Status Trees (Page 2 of 2)

6. IF during function restoration to address a Critical Safety Function Challenge, a higher priority challenge is diagnosed, THEN the operator should terminate the ongoing response and initiate function restoration to address the higher priority Critical Safety Function Challenge.

7. jf an Extreme Challenge (RED PATH) exists a Severe Challenge (ORANGE PATH) exists OR plant conditions are changing rapidly, THEN Critical Safety Function Status Trees shall be monitored continuously.

8. IF an Extreme Challenge (RED PATH) does NOT exist AND a Severe Challenge (ORANGE PATH) does NOT exist AND plant conditions are NOT changing rapidly, THEN Critical Safety Function Status Trees shall be monitored every 10 to 20 minutes.

End of Attachment I

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 4 KIA# G4 2.4.25 Importance Rating 3.3 Emergency Procedures / Plan: Knowledge of fire protection procedures.

Proposed Question: RO Question # 74 Given the following conditions:

• Unit 3 is at 100% power.

• Work is ongoing in the Unit 3 480V Load Center Rooms.

• The foreman requests the fire door between A & B and C & D Load Centers be opened to allow better air circulation for the workers comfort.

• A Fire Protection Impairment has NOT been issued.

Which ONE of the following identifies the policy concerning the propping open of fire doors?

A. May be open without compensatory actions as long as work is ongoing.

B. May be open if an hourly roving fire watch is provided.

C. Cannot be opened solely for comfort of personnel.

D. Cannot be opened for more than 30 minutes.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible if examinee believes that worker comfort and safety are equivalent and, therefore, a fire door that is opened for personal safety reasons requires no compensatory actions; however, even under such conditions a Fire Protection Impairment must be requested/authorized.

B. Incorrect. Plausible if examinee recognizes that 0-ADM-01 6 allows fire doors to be left open on a continuous basis (without a Fire Protection Impairment), as long as a watch is provided; however, a roving watch does not meet the procedural requirements and the door must be attended continuously.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION C. Correct. 0-ADM-016 specifically states that required work activities involving fire doors do not include keeping the doors open for the comfort of personnel.

D. Incorrect. Plausible if examinee recognizes that 0-ADM-016 allows fire doors to be left open on a continuous basis, to support material transfer, for a limited period of time; however, this period of time may be extended to one full hour.

Technical Reference(s): 0-ADM-016 Rev 9 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: LP 6902038 Obj. 2 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:

PTN LOIT Exam Bank Item # 1.1.23.38.2.4 PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

Procedure No. Procedure

Title:

Page:

32 Approval Date:

O-ADM-016 Fire Protection Program 8/8/13 5.4 Control of Fire Spread with Fire Rated Assemblies 5.4.1 Uncontrolled fire growth from one train of systems necessary to achieve and maintain hot standby conditions from its redundant counterpart is prevented by fire rated assemblies which include: -

1. Fire Barriers which consist of walls, ceilings, floors, and electrical raceway protection, that are designed and constructed to provide a resistance to fire, i-neasured in hours and determined by specific tests or evaluations. Fire barriers provide a means of limiting fire travel through compartrnentalization and containment.

2. Fire Door assemblies are designed for the protection for passageways through fire barriers.

3. Fire Dampers are provided to maintain the integrity of identified fire barriers necessary to protect safe shutdown capability. Fire dampers are also used to isolate an area prior to halon system actuation.

5.5 Plant Policies Regarding Specific Fire Protection Applications 5.5.1 Fire Doors As applied to Fire Doors, Required by work activities is defined as those activities necessary to support the transferring of materials or laying objects through the open Fire Door on a continuous basis for a period of time NOT TO EXCEED 1 HOUR. These activities DO NOT include keeping the Fire Doors open for the comfort of personnel.

[Commitment Step 2.3. 1]

I Fire doors may be open when required by work activities without an impairment as long as door is being attended and door can be closed if a fire should occur.IZJAn open fire door shall never be left unattended, unless authorized by an FPI 1

2. A Fire Protection Impairment (FPI) may be requested to keep certain fire doors open for the safety of personnel during some work activities (hot weather conditions, etc.).

3. Fire door latches are never to be taped or otherwise rendered inoperable as they are required by code to be Positive Latching.

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION Examination Outline Cross-reference: Level RO SRO Tier# 3 Group# 4 KIA# G4 2.4.47 Importance Rating 4.2 Emergency Procedures I Plan: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Proposed Question: RO Question # 75 Given the following conditions:

• Unit 3 is at 100% power.

Subsequently:

• Unit 3 experiences an accident with the following conditions:

Pressure AFW Flow SIG A 950 psig and lowering 260 gpm SIG B 290 psig and lowering 340 gpm S/G C 940 psig and lowering 230 gpm NR Level WR Level SIG A 21.5% and lowering 55.8% and lowering SIG B 80.3 % and rising 55.9% and rising SIG C 18.6 % and lowering 54.5% and lowering Containment pressure: 21 .5 psig and rising Containment Sump level: 36.2 inches and rising Pressurizer pressure: 1274 psig and lowering Pressurizer level: 1 % and lowering Based on the indicated parameters, which ONE of the following identifies the initial transition from 3-EOP-E-0 (Reactor Trip or Safety Injection), after the immediate operator actions are complete?

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION A. 3-EOP-ES-O.O, Rediagnosis B. 3-FOP-F-i, Loss of Reactor or Secondary Coolant C. 3-EOP-E-2, Faulted Steam Generator D. 3-EOP-E-3, Steam Generator Tube Rupture Proposed Answer: C Explanation (Optional):

A. Incorrect. Plausible if examinee believes that the entry conditions for this procedure have been met (i.e., SI is in service or required, 3-EOP-E-O is completed/exited, and no Functional Restoration Procedure is in progress).

B. Incorrect. Plausible since Pressurizer pressure and level are lowering and Containment pressure and sump level are rising; however, with no indication of open Pressurizer PORVs and/or elevated Containment radiation levels, and RCS breach is not indicated.

C. Correct. The decreasing pressure (and increasing level) in S/G B is an indication of a faulted S/G; rising Containment pressure and sump level suggest a steam-line break inside containment.

D. Incorrect. Plausible since S/G B level is rising; however, this is due to the initial swell caused by the steam-line break (i.e., the pressure drop in the SIG). The decreasing pressure in S/G B, which is not indicative of a SGTR, suggests a secondary-side fault; additionally, a SGTR would also be accompanied by indications of rising radiation levels in the secondary plant.

Technical Reference(s): 3-EOP-E-O Rev 10 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Learning Objective: PTN 6900321; Obj. 4, 7 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CER Part 55 Content: 55.41 10 55.43 Comments:

PTN L-15-1 DRAFT NRC EXAM SECURE INFORMATION

REVISION NO.: PROCEDURE TITLE: PAGE:

10 REACTOR TRIP OR SAFETY INJECTION 16 of 55 PROCEDURE NO.:

3-EOP-E-0 TURKEY POINT UNIT 3 I STEP I I ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

11. Check PRZ PORVs, Spray Valves And Excess Letdown Isolated:

a. PORVs - CLOSED a. IF PRZ pressure less than 2335 psig, THEN manually close PORVs.

H y PRZ PORV can NOT be closed, THEN manually close its Block Valve.

IF Block Valve can NOT be closed, THEN perform the following:

1) Monitor Critical Safety Functions using 3-EOP-F-0, CRITICAL SAFETY FUNCTION STATUS TREES.

2) Go to 3-EOP-E-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.

b. Normal PRZ Spray valves b. IF PRZ pressure less than 2250 psig, CLOSED THEN manually close valves.

IF valve(s) can NOT be closed, THEN stop RCP(s) as necessary to stop spray flow.

C. CV-3-31 1, Auxiliary Spray Valve c. Manually close Auxiliary Spray Valve.

CLOSED IF Auxiliary Spray Valve can NOT be closed, THEN close HCV-3-121, Charging Flow To Regen Heat Exchanger.

d. Excess Letdown d. Perform the following:

NOT IN SERVICE 1) Close CV-3-387, Excess Letdown Isolation Valve From Cold Leg To Excess Letdown Heat Exchanger.

2) Close HCV-3-137, Excess Letdown Flow Controller.

REVISION NO.: PROCEDURE TITLE: PAGE:

10 REACTOR TRIP OR SAFETY INJECTION 17 of 55 PROCEDURE NO.:

3-EOP-E-0 TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

12. Check If RCPs Should Be Stopped:

a. RCPsANYRUNNING a. GotoStepl3.

b. High-Head SI Pump b. Go to Step 13.

AT LEAST ONE RUNNING, AND SI Flowpath VERIFIED

c. RCS subcooling c. Go to Step 13.

LESS THAN 19°F[41°F]

d. Stop a RCPs

13. Check If S!Gs Are FauIted,

a. Check pressures in all S/Gs a. Go to Step 14.

ANYS/G PRESSURE DECREASING IN AN UNCONTROLLED MANNEf OR ANY S/G COMPLETELY DEPRESSURIZED

b. Perform the following:

1) Monitor Critical Safety Functions using 3-EOP-F-O, CRITICAL SAFETY FUNCTION STATUS TREES

2) Go to 3-EOP-E-2, FAULTED STEAM GENERATOR ISOLATION, Step 1

REVISION NO.: PROCEDURE TITLE: PAGE:

10 REACTOR TRIP OR SAFETY INJECTION 18 of 55 PROCEDURE NO.:

3-EOP-E-0 TURKEY POINT UNIT 3 I STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

14. Check If SIG Tubes Are Ruptured:

a. Check levels in aN SIGs and a. Go to Step 15.

secondary radiation levels:

M SIG level INCREASING IN AN UNCONTROLLED MANNERS OR Condenser Air Ejector Radiation R-15 HIGHER THAN NORMAL OR

¶s/G Blowdown Radiation R-19

[-jHER THAN NORMAL OR DCS S/G OR secondary radiation readings HIGHER THAN NORMAL OR LocaI steamline radiation HIGHER THAN NORMAL

b. Perform the following:

1) Monitor Critical Safety Functions using 3-EOP-F-O, CRITICAL SAFETY FUNCTION STATUS TREES

2) Go to 3-EOP-E-3, STEAM GENERATOR TUBE RUPTURE, Step 1

REVISION NO.: PROCEDURE TITLE: PAGE:

10 REACTOR TRIP OR SAFETY INJECTION 19 of 55 PROCEDURE NO.:

3-EOP-E-0 TURKEY POINT UNIT 3 STEP II ACTIONIEXPECTED RESPONSE I I RESPONSE NOT OBTAINED I

15. Check If RCS Is Intact? Perform the following:

Containment radiation NORMAL 1. Monitor Critical Safety Functions using 3-EOP-F-O, CRITICAL SAFETY

• Containment pressure NORMAL: FUNCTION STATUS TREES.

PR-3-6306A PR-3-6306B

2. 0 to 3-EOP-E-1, LOSS OF a

REACTOR OR SECONDARY COOLANT, Step 1.

• Containment Sump level NORMAL:

• LI-3-6308A

• Ll-3-6308B