ML101590396

Turkey Point, 2007-301 Reactor Operator Written Exam (Section 2)

ML101590396
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Site:	Turkey Point
Issue date:	03/28/2010
From:	NRC/RGN-II
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Download: ML101590396 (202)

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NRC DRAFT 07124/07 0#1 Unit 3 is at 100% power with all systems in normal alignment except for 3A S/G level transmitter, L T-3-474 (red channel) which has failed LOW. Operators have tripped associated bistables in accordance with 3-0NOP-049.1, "Deviation or Failure of Safety Related or Reactor Protection Channels." Subsequently, 3A S/G feed flow transmitter, FT-3-477 (blue channel), fails LOW. Which ONE of the following describes the correct operator response?

A. Perform 3-0NOP-049.1 again and trip all bistables associated with FT-3-477.

B. Perform 3-0NOP-049.1 again and trip all bistables associated with FT-3-477 except for the "FW to SF Mismatch Logic" bistable, BS-3-478B-1.

C. Declare Unit 3 is in TS 3.0.3 and be in Hot Standby within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. D. Trip the reactor and transition to 3-EOP-E-0, "Reactor Trip or Safety Injection" NRC DRAFT 07/24/07 Q #1 ANSWER: D KA: 000007EA2.02 Able to determine or interpret following as they apply to reactor trip: Proper actions to be taken if the automatic safety functions have not taken place. 4.3/4.6 10CFR55: 41.b.5, 41.b.7, 41.b.10

Reference:

5610-T-L 1, Sheet 2 5610-T-L 1, Sheet 19 5610-T-D-17,Sheet1 3-0NOP-049.1, Pages 33 & 45 Cog Level: 2 Comprehension Level 2 because the operator must recognize that a red (Channel 1 ) S/G level transmitter with its bistables tripped includes a low level trip bistable.

When the blue channel (Channel 3) feed flow transmitter fails, a steam flow/feed flow mismatch logic is made up that combines with the existing low level signal to generate a reactor trip signal. New Question Response Analysis:

A. Incorrect because the reactor should have tripped. Operators should transition to EOP-E-O. Plausible, because this is the normal response to a failure of FT-3-477 if a reactor trip signal had not been generated.

B. Incorrect because the reactor should have tripped. Operators should transition to EOP-E-O. Plausible, because this is the response to a failure of FT-3-477 following a failure of L T-3-474 if a reactor trip signal had not been generated.

C. Incorrect because the reactor should have tripped. Operators should transition to EOP-E-O. Plausible, because the unit would be in TS 3.03 if the reactor had not tripped. D. Correct per the references.

The reactor should have tripped as a result of the Low S/G level with steam flow/feed flow mismatch signal generated by the FT-3-477 failure. 2 NRC DRAFT 07/24/07 0#2 Following a pressure transient in the RCS, a pressurizer safety valve lifted and subsequently remained open.

• Operators are performing 3-EOP-ES-1.2, "Post LOCA Cooldown and Depressurization.

• RCS subcooling is less than the minimum required.

• With RCS T-hot less than 350°F, operators start one RHR pump and stop one HHSI pump. Which ONE of the following describes the effect of starting the RHR pump and stopping the HHSI pump? A. Secondary heat sink will be improved by RHR flow collapsing reactor head voids after stopping the HHSI pump. B. RCS subcooling will decrease after stopping the HHSI pump until the RHR pumps injects into the RCS. C. Subsequent steam generator U-tube cooldown will be enhanced by the addition of RHR flow through the steam generators.

D. RCS subcooling will NOT be affected by stopping the HHSI pump because the RHR pump is already injecting into the RCS. 3 NRC DRAFT 07/24/07 Q#2 ANSWER: B KA: 000008AK3.05 Knowledge for the reasons for the following responses as they apply to the PRZ vapor space accident:

ECCS terminating or throttling criteria.

4.0/4.5 10CFR55: 41.b.5, 41.b.7, 41.b.8, 41.b.10

Reference:

3-EOP-ES-1.2, Step 17 BO Cog Level: 2 Comprehension Level 2 because the operator must analyze plant conditions and apply thermodynamic principles to determine the effect starting and stopping various injection flows has on subcooling and the effect that the HHSI pumps have on maintaining pressure higher and thus subcooling greater. New Question Response Analysis:

A. Incorrect because the RHR pump is no more effective at collapsing head voids than the previously running HHSI pump and this is not the reason given by the reference.

Plausible because head voids should be expected for a steam space LOCA and SI pumps do help collapse head voids. B. Correct per the reference.

C. Incorrect because the addition of RHR flow will not significantly increase flow though the S/Gs and this is not the reason given by the reference. Plausible because S/G U-tube cooldown would be enhanced by additional cooling flow. O. Incorrect because the pressure must drop until it is less than the RHR pump shutoff head. Thus subcooling must decrease when the HHSI pump is shut off. 4 NRC DRAFT 07124/07 Q#3 Operators are responding to a large break LOCA on Unit 4.

• When the BOP attempts to reset 81 on VPB, a circuit failure occurs and the 81 signal CANNOT be reset.

• Operators have transitioned to 4-EOP-E8-1.3, "Transfer to Cold Leg Recirculation" and expect to subsequently establish "piggy-back" recirc alignment.

Which ONE of the following describes the effect of being unable to reset 81? A. The running RHR pump(s) can NOT be stopped preventing operators from closing RHR pump suction valves, MOV-4-862A

& B. B. Once stopped, the RHR pump(s) can NOT be restarted until the 81 signal is reset. C. RHR discharge to cold leg isolation valves, MOV-4-744A

& B, can NOT be closed requiring local closure of BOTH RHR heat exchanger manual outlet valves, 4-759A & B. D. RHR discharge to cold leg isolation valves, MOV-4-744A

& B, can NOT be closed requiring local closure of EITHER RHR heat exchanger manual outlet valve, 4-759A or B. 5 NRC DRAFT 07/24/07 Q#3 ANSWER: C KA: 000011 EA2.02 Able to determine and interpret the following as they apply to a LBLOCA: Consequences to RHR of not resetting SI. 3.3/3.7 10CFR55: 41.b.7

Reference:

4-EOP-ES-1.3, Step 19 RNO 5613-M-3050, Sheet 1 5610-T-L 1, Sheet 11 Cognitive Level: 2 Comprehension Level 2 because the operator must understand integrated plant operations, interlocks and response not obtained actions to recognize the effect of not being able to reset SI. Once recognized that the 744 valves cannot be closed, the operator must determine the correct response to close both RHR HX outlet valves per EOP-ES-1.3, Step 19. New Question Response Analysis:

A. Incorrect because the running RHR pumps can be stopped by placing their control switches in Pull-to-Lock.

Plausible, because there is a continuous start signal to the RHR pumps as a result of the active SI signal present. B. Incorrect because the running RHR pumps can be manually started from VPB. Plausible, because the RHR pumps can be stopped by placing their control switches in Pull-to-Lock.

C. Correct per the references.

If SI cannot be reset, there is a continuous OPEN signal to MOV-4-744A

& B. If the valves cannot be closed, EOP-ES-1.3, Step 19 RNO requires closure of both RHR heat exchanger outlet valves. D. Incorrect because If the valves cannot be closed, EOP-ES-1.3, Step 19 RNO requires closure of both RHR heat exchanger outlet valves, not just one. Plausible because If SI cannot be reset, there is a continuous OPEN signal to MOV-4-744A

& B. 6 NRC DRAFT 07/24/07 0#4 Unit 3 is at 100% power when annunciator 8 2/5, RCP 8 OIL RESERVOIR HIILO LEVEL, alarms. Which ONE of the following describes the correct operator response?

If the cause is verified to be from a: A. Low reservoir level, plot 38 RCP motor bearing temperatures every 15 minutes. 8. Low reservoir level, stop the 38 RCP and isolate CCW flow to the oil coolers within 30 minutes. C. High reservoir level, plot 38 RCP motor and shaft vibration every 15 minutes. D. High reservoir level, stop the 38 RCP and increase CCW flow to the oil coolers within 30 minutes. 7 NRC DRAFT 07/24/07 Q#4 ANSWER: A KA: 000015/17AA1.02 Able to operate and/or monitor the following as they apply to the RCP Malfunctions (Loss of RC flow): RCP Oil Reservoir level and alarm indicators.

2.8/2.7 10CFR55: 41.b.7

Reference:

3-ARP-097.CR, Ann. B2/5 3-0NOP-041.1 , Steps 8 and 43 Cognitive Level: 1 Recall New Question Response Analysis:

A. Correct per 3-0NOP-041.1, Step 43. B. Incorrect because this is the action to be taken if high reservoir level is discovered, not low reservoir level as stated in the response.

Plausible because this would be the correct action if the problem was discovered to be high reservoir level. C. Incorrect because bearing temperatures should be plotted, not vibration.

Plausible because this is the correct frequency of monitoring.

D. Incorrect because this is the incorrect response for high reservoir level. CCW flow should be isolated, not increased Plausible because CCW flow to the oil coolers must be altered within 30 minutes in the event of high reservoir level.. 8 NRC DRAFT 07/24/07 0#5 Unit 3 is at 100% power with all systems in normal alignment.

3A Charging pump is operating.

3-267, 3A Chrg pump suct, isolation valve, experiences a body-to-bonnet leak resulting in a loss of charging flow. Which ONE of the following describes the correct operator response and the reason for that response?

A. Verify all charging pumps are stopped and then close valve 3-268, Chrg Pump Suct Hdr X-Conn valve, because no charging pumps can deliver flow to the RCS. 8. Stop the 3A Charging pump and start the 38 Charging pump after closing valve 3-268, Chrg Pump Suct Hdr X-Conn valve, because only 38 Charging pump can deliver flow to the RCS. C. Stop the 3A Charging pump and start 3C Charging pump after closing valve 3-268, Chrg Pump Suct Hdr X-Conn valve, because only the 3C Charging pump can deliver flow to the RCS. D. Start either of the 38 or 3C Charging pump after closing valve 3-268, Chrg Pump Suct Hdr X-Conn valve, because either 38 or 3C Charging pump can deliver flow to the RCS. REFERENCE PROVIDED 9 NRC DRAFT 07/24/07 Q#5 ANSWER: C KA: 000022AK3.07 Knowledge for the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: Isolating charging.

3.0/3.2 10CFR55: 41.b.5, 41.b.10

Reference:

3-0NOP-047.1, Step 4.1 5613-M-3047,Sheet2 NOTE: PROVIDE 5613-M-3047, Sheet 2 AS A REFERENCE Cog Level: 2 Comprehension Level 2 because the RO must evaluate the location of the leak and determine that the leak location disables both the 3A and 3B Charging pumps. The 3C Charging pump is still viable but valve 3-268 must be closed first. New Question Response Analysis:

A. Incorrect because 3C Charging pump is still available once valve 3-268 is closed. Plausible because all charging should be stopped and because even if charging could not be recovered, valve 3-268 should be closed to isolate the leak. B. Incorrect because the 3B Charging pump is disabled as well as the 3A Charging pump. Plausible because the 3A Charging pump needs to be stopped and valve 3-268 should be closed. C. Correct because the 3A Charging pump needs to be stopped and only the 3C Charging pump can deliver flow but only after valve 3-268 is closed. D. Incorrect because the 3B Charging pump is disabled as well as the 3A Charging pump. Plausible because the 3C Charging pump can deliver flow but only after valve 3-268 is closed. 10 NRC DRAFT 07124/07 Q#6 Unit 4 is in Mode 4 and on RHR cooling.

• A Loss of Off-Site Power (LOOP) occurs.

• The 4A sequencer fails to start its components.

Which ONE of the following describes the correct operator response regarding restoration of CCW pump and RHR pump operation?

The operator will: A. NOT have to start any CCW pumps or RHR pumps. B. NOT have to start any CCW pumps but will start one RHR pump. C. have to start one CCW pump and will NOT have to start any RHR pumps. D. have to start one CCW pump and start one RHR pump. 11 NRC DRAFT 07/24/07 Q#6 ANSWER: D KA: 000025AA1.04 Able to operate and/or monitor the following as they apply to the Loss of RHR System: Closed cooling water pumps. 2.8/2.6 10CFR55: 41.b.7

Reference:

4-0NOP-004, Steps 7 and 8 5614-T-L1, Sheet 12A& 12B Cog Level: 2 Comprehension Level 2 because the operator must evaluate the effect of the loss of the 4A sequencer on the final CCW and RHR pump configuration and then the operator must recall the CCW and RHR pump requirements of ONOP-004.

New Question Response Analysis:

A. Incorrect because the operator will have to start one CCW pump and start one RHR pump. Plausible because if the sequencer had operated properly, the operator would not have had to start any CCW pumps. B. Incorrect because the operator will have to start one CCW pump and start one RHR pump. Plausible because if the sequencer had operated properly, the operator would not have had to start any CCW pumps and the operator will have to start one RHR pump. C. Incorrect because the operator will have to start one RHR pump. Plausible because the operator will have to start one CCW pump. D. Correct because the operator will have to start one CCW pump and start one RHR pump. 12 NRC DRAFT 07/24/07 0#7 Operators are responding to a main steam line break inside Unit 3 containment.

• Containment pressure peaked at 26 psig and is now 15 psig.

• The 3B SG WR level is 15% and lowering.

• No charging pumps are running

• Step 10 of 3-EOP-E-1, "Loss of Reactor or Secondary Coolant," directs establishment of maximum charging flow.

• Seal Water Return temperatures are at 245°F.

• The Unit Supervisor orders closure of the local seal injection valves, 297 AlBIC, before starting the first charging pump. Which ONE of the following describes why the local seal injection valves are closed at this time? A. CCW flow to the thermal barriers has been lost. RCP seals already have elevated temperatures and initiation of cold seal injection flow will cause RCP damage. B. CCW flow to the thermal barriers has been lost. RCP seal temperatures are still in the allowed range and maximum charging flow to the RCS loops via the normal charging path is required.

C. CCW flow to the thermal barriers has NOT been interrupted.

Seal injection is NOT needed to maintain seal integrity and maximum charging flow to the RCS loops via the normal charging path is required.

D. CCW flow to the thermal barriers has NOT been interrupted.

Seal injection is NOT needed to maintain seal integrity and initiation of cold seal injection flow will cause RCP damage. 13 NRC DRAFT 07/24/07 Q#7 ANSWER: A KA: 000026AK3.03 Knowledge for the reasons for the following responses as they apply to the loss of CCW: Guidance contained in EOP for loss of CCW. 4.0/4.2 10CFR55: 41.b.5, 41.b.10

Reference:

3-EOP-E-1, Step 10 RNO SD 5610-T-L1, Sheet 11 5613-M-3030, Sheet 5 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the references.

S. Incorrect because RCP seal temperatures are already heated up per the basis document.

Plausible because CCW flow to the thermal barriers has been lost. C. Incorrect because CCW flow to the thermal barriers has been lost. Plausible because maximum charging flow to the RCS loops via the normal charging path is required.

D. Incorrect because CCW flow to the thermal barriers has been lost. Plausible because initiation of cold seal injection flow will cause RCP damage. 14 NRC DRAFT 07/24/07 Q#8 Unit 4 is initially at 70% power with all systems in automatic except the master pressurizer pressure controller, PC-4-444J, which is in manual due to an instrument failure.

• 4B Steam Generator Feed pump (SGFP) breaker subsequently trips. Which ONE of the following describes the effect of this event on the RCS and the correct operator response?

The resulting turbine runback will initially cause RCS temperature to: A. increase resulting in expansion of reactor coolant into the pressurizer causing RCS pressure to increase.

The RO will drive the PC-4-444J controller output higher (toward 100%) to stabilize pressure.

B. increase resulting in expansion of reactor coolant into the pressurizer causing RCS pressure to increase.

The RO will drive the PC-4-444J controller output lower (toward 0%) to stabilize pressure.

C. decrease resulting in contraction of reactor coolant from the pressurizer causing RCS pressure to decrease.

The RO will drive the PC-4-444J controller output higher (toward 100%) to stabilize pressure.

D. decrease resulting in contraction of reactor coolant from the pressurizer causing RCS pressure to decrease.

The RO will drive the PC-4-444J controller output lower (toward 0%) to stabilize pressure.

15 NRC DRAFT 07/24/07 Q#8 ANSWER: A KA: 000027AK1.02 Knowledge of the operational implications of the following concepts as they apply to the pressurizer pressure control m'alfunctions:

Expansion of liquids as temperature increases.

2.8/3.1 10CFR55: 41.b.5, 41.b.7, 41.b.10, 41.b.14

Reference:

4-0NOP-041.5, Step 1.a RNO SO-009, Figure 22 Cog Level: 2 Comprehension Level 2 because the operator must analyze that with turbine load reduced first, Tavg will initially increase and then be higher than Tref, and pressurizer level will increase raising pressure.

The system will need more spray output to maintain programmed pressure.

This requires an increase in controller output to open sprays. New Question Response Analysis:

A. Correct per the references.

B. Incorrect because the RO will drive the PC-4-444J controller output higher (toward 100%) to stabilize pressure.

Plausible because the resulting turbine runback will initially cause RCS temperature to increase resulting in expansion of reactor coolant into the pressurizer causing RCS pressure to increase.

C. Incorrect because the resulting turbine runback will initially cause RCS temperature to increase resulting in expansion of reactor coolant into the pressurizer causing RCS pressure to increase.

Plausible because the RO will drive the PC-4-444J controller output higher (toward 100%) to stabilize pressure O. Incorrect because the resulting turbine runback will initially cause RCS temperature to increase resulting in expansion of reactor coolant into the pressurizer causing RCS pressure to increase and the RO will drive the PC-4-444J controller output higher (toward 100%) to stabilize pressure.

Plausible the response is consistent if the operator does not realize that a turbine runback causes the RCS temperature to increase.

16 NRC DRAFT 07124/07 0#9 Unit 4 was at 100% power with all systems in normal alignment when the following events occurred:

• 4C S/G faulted outside containment.

• The reactor failed to trip automatically or manually.

• All other safeguards systems actuated as required.

Operators are performing Step 14 of 4-EOP-FR-S.1, "Response to Nuclear Power Generation/ATWS," which directs them to "Verify steam supply aligned to both trains of AFW pumps from intact S/G( s )." Which ONE of the following describes the correct operator response? (Note: AFSS-4-006 and AFSS-4-007 are AFW steam supply cross connect valves.) Direct the NSO to locally: A. open AFSS-4-006 only. B. open AFSS-4-007 only. C. close AFSS-4-007 and then open AFSS-4-006.

D. open AFSS-4-007 and then close AFSS-4-006.

17 NRC DRAFT 07124/07 Q#9 ANSWER: 0 KA: 000029G2.1.30 As it relates to the ATWS event: Able to locate and operate components, including local controls.

3.9/4.0 10CFR55: 41.b.4, 41.b.7,41.b.10

Reference:

4-EOP-FR-S.1, Step 14.d.RNO 5614-M-3075,Sheet1 Cog Level: 2 Comprehension Level 2 because the operator must recognize that the goal is to restore steam supply from two intact steam generator trains and to maintain train separation and maintain steam supply during the process. The correct way to achieve this is to locally open AFSS-4-007 and then close AFSS-4-006.

New Question Response Analysis:

A. Incorrect because the RO will direct the NSO to open AFSS-4-007 and then close AFSS-4-006 lAW EOP-FR-S.1, Step 14.d.RNO.

Plausible because the procedure step does not specify the desired position of AFSS-4-006.

It merely directs operators to manipulate AFSS-4-006.

B. Incorrect because the RO will direct the NSO to open AFSS-4-007 and then close AFSS-4-006 lAW EOP-FR-S.1, Step 14.d.RNO.

Plausible because it is correct (but not complete) to locally open AFSS-4-007.

C. Incorrect because the RO will direct the NSO to open AFSS-4-007 and then close AFSS-4-006 lAW EOP-FR-S.1, Step 14.d.RNO.

Plausible because the procedure step does not specify the desired position of AFSS-4-006 or AFSS-4-006.

It merely directs operators to manipulate AFSS-4-006 and AFSS-4-006.

D. Correct per the references 18 NRC DRAFT 07/24/07 Q#10 Operators are performing 3-EOP-E-3, "Steam Generator Tube Rupture."

• All RCPs are stopped. After dumping steam from intact S/Gs to increase subcooling, the RO has been directed to open one pressurizer PORV. Which ONE of the following describes the effect of opening the PORV on pressurizer level, RCS subcooling and S/G tube break flow rate? Pressurizer level RCS subcooling S/G break flow A. Decreases Decreases Increases B. Increases Decreases Decreases C. Increases Increases Increases D. Decreases Increases Decreases 19 NRC DRAFT 07/24/07 Q#10 ANSWER: B KA: 000038EK1.02 Knowledge of the operational implications of the following concepts as they apply to the SGTR: Leak rate vs pressure drop. 3.2/3.5 10CFR55: 41.b.5, 41.b.10, 41.b.14

Reference:

3-EOP-E-3, NOTE prior to Step 25, Step 25 3-EOP-E-3, Step 25 BO Cog Level: 1 recall New Question Response Analysis:

A. Incorrect because pressurizer level will increase, not decrease and break flow will decrease, not increase.

Plausible because subcooling will decrease.

B. Correct per the references.

C. Incorrect because subcooling will decrease, not increase and break flow will decrease, not increase.

Plausible because pressurizer level will increase.

O. Incorrect because pressurizer level will increase, not decrease and subcooling will decrease, not increase.

Plausible because break flow will decrease 20 NRC DRAFT 07/24/07 Q #11 Unit 4 is at 50% power when a large main steam line break outside containment downstream (turbine side) of the 4B MSIV and non-return check valve occurs. Which ONE of the following describes the response of the Unit 4 MSIVs and the reason for that response?

A. None of the Unit 4 MSIVs will automatically close as NO protective signal is generated for this condition.

B. 4B MSIV will automatically close due to high steam flow on the 4B S/G with a low 4B S/G pressure.

4A and 4C MSIVs will NOT automatically close. C. All of the Unit 4 MSIVs will automatically close due to high steam flow with low Tavg or low SG pressure sensed on all S/Gs. D. All of the Unit 4 MSIVs will automatically close due to high differential pressure between each S/G and the main steam header. 21 NRC DRAFT 07124/07 Q #11 ANSWER: C KA: 000040AK2.01 Knowledge of the interrelations between the steam line rupture and the following:

Valves. 2.6/2.5 10CFR55: 41.bA, 41.b.7, 41.b.8

Reference:

5610-T-L 1, Sheet 11 5610-T-L 1, Sheet 19 5614-M-3072,Sheet1 Cog Level: 2 Comprehension Level 2 because the operator must recognize that the location of the break determines the response of the MSIVs. Because the break is downstream of the MSIV/non-return check valve, it does not matter that it is located on the 4B header. All S/Gs will feed the break and all MSIVs will close. New Question Response Analysis:

A. Incorrect because all Unit 4 MSIVs will close. Plausible because the stem implies the problem is associated with only the 4B steam header. If the problem was really associated with only one header, no protective signal would be generated.

B. Incorrect because all Unit 4 MSIVs will close. Plausible because the break is located on the 4B steam header and 4B MSIV will automatically close. C. Correct per the references D. Incorrect because high steam line 1 S/G pressure differential does not close MSIVs. Plausible because all of the MSIVs will automatically close 22 NRC DRAFT 07/24/07 Q #12 With Unit 3 critical at 10-8 amps IR power, 3A main feedwater header sheared immediately outside Unit 3 containment.

Which ONE of the following describes the effect on Unit 3? 3A S/G will: A. NOT depressurize.

SI will NOT actuate. B. NOT depressurize.

Low pressurizer pressure reactor trip and SI will occur. c. depressurize.

Low pressurizer pressure reactor trip and SI will occur. D. depressurize.

Hi Steam Line SI will actuate. 23 NRC DRAFT 07/24/07 Q #12 ANSWER: D KA: 000054AK1.01 Knowledge of the operational implications of the following concepts as they apply to the Loss of Main FW: MFW line break depressurizes the S/G (similar to a steam line break) 4.1/4.3 10CFR55: 41.bA, 41.b.7, 41.b.8

Reference:

561 O-T-L 1, Sheet 11 561 O-T-L 1, Sheet 19 5613-M-3074,Sheet3 Cog Level: 2 Comprehension Level 2 because the operator must recognize that a feedwater header break at that location will depressurize the 3A S/G and the depressurization will be sensed on the 3A steam header which will result in the High steamline L'1P logic (1/3 S/Gs) being initiated which is not blocked at low power levels. New Question Response Analysis:

A. Incorrect because 3A S/G will depressurize and SI will actuate. Plausible because the SI protection is needed for a steam line break, not a feed line break. B. Incorrect because 3A S/G will depressurize and the low pressurizer pressure trip is blocked and the 3A s/g will depressurize.

Plausible because SI would be expected subsequently C. Incorrect because the low pressurizer trip is blocked. Plausible because the S/G will depressurize and SI would be expected subsequently D. Correct per the references.

24 NRC DRAFT 07/24/07 Q #13 Operators are performing 4-EOP-ECA-0.0, "Loss of All AC Power" and are in the process of depressurizing all intact S/Gs to 180 psig. What is the consequence if the SG depressurization is NOT stopped until 50 psig? A. A reactor vessel head void could become large enough to partially uncover the core. B. A Nitrogen gas bubble in the pressurizer could cause a loss of pressure control. C. The pressurizer would go water-solid resulting in a loss of pressure control. D. Nitrogen could enter the S/G U-tubes disrupting natural circulation.

25 NRC DRAFT 07124/07 Q #13 ANSWER: D KA: 000055EK1.02 Knowledge of the operational implications of the following concepts as they apply to the station blackout:

Natural Circulation Cooling. 4.1/4.4 10CFR55: 41.b.3, 41.b.4, 41.b.7

Reference:

3-EOP-ECA-O.O BD, CAUTION before Step 26 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because there is too little nitrogen to be released to cause core uncovery and a head void may occur but will not uncover the core, this is not the reason given by the reference.

Plausible because nitrogen is expected to accumulate in the reactor vessel head and the rapid depressurization will cause a head void. B. Incorrect because nitrogen accumulating in the pressurizer will not result in a loss of pressure control which is being provided by the PORVs (no spray or heaters) and this is not the reason given by the reference.

Plausible because nitrogen is expected to migrate to the pressurizer (RCS high point). C. Incorrect because the pressurizer would be expected to empty under these conditions.

Plausible because large pressurizer level changes are expected and a loss of pressurizer pressure control is expected if the pressurizer empties. D. Correct per the references.

26 NRC DRAFT 07124/07 0#14 Unit 3 is at 100% power when the following events occur:

• Unit 3 reactor automatically trips.

• Unit 3 MSIVs automatically close.

• Unit 3 annunciators go dark. Which ONE of the following identifies the procedure operators will implement after the unit is stabilized using 3-EOP-ES-0.1, "Reactor Trip Response"?

A. 3-0NOP-005.4, "4KV Bus 3A, 3B or 3D Ground" B. 3-0NOP-003.9, "Loss of 120 VAC Vital Instrument Panel 3P09" C. 3-0NOP-003.5, "Loss of DC Buses 3023 and 3D23A (3B) D. 3-0NOP-003.4, "Loss of DC Bus 3001 and 3D01A (3A) 27 NRC DRAFT 07/24/07 Q#14 ANSWER: D KA: 000058G2.4.4 As it relates to the loss of DC power event: Able to recognize abnormal indications for system operating parameters which are entry level conditions for emergency and abnormal operating procedures.

4.0/4.3 10CFR55: 41.b.7,41.b.10

Reference:

3-0NOP-003.4, Step 1.1 & 2.1, 2.2 & 3.5 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because operators are directed to enter ONOP-003.4 immediately following EOP-ES-0.1.

Plausible because a reactor trip, MSIV closure or loss of annunciators could be related to various grounds. B. Incorrect because operators are directed to enter ONOP-003.4 immediately following EOP-ES-0.1.

Plausible because the first step of ONOP-003.9 procedure checks for a reactor trip and analyzes if one is needed and loss of 3P09 previously required a reactor trip. C. Incorrect because all of the symptoms listed are indicative of a loss of 3D01, not 3D23. Plausible because the reactor will automatically trip and MSIVs will automatically close upon a loss of 3D23 D. Correct per the references 28 NRC DRAFT 07124/07 0#15 Operators are responding to a line break in the Instrument Air (IA) system.

• IA pressure as seen by PI-3-1444 (VPA) is 91 psig and lowering.

• All available IA compressors are running and crossties are open. The NSO is sent to investigate and reports the following local IA pressure indications:

PI-3-1516, Turbine Area: PI-3-1517, Containment Area: PI-3-1518, Aux Bldg, Intake Area, Control Room: 86 psig lowering 75 psig lowering 88 psig lowering Which ONE of the following describes the correct operator response?

A. Isolate IA to the Containment.

Trip the reactor and perform 3-EOP-E-O, "Reactor Trip or Safety Injection." B. Isolate IA to the Containment.

If IA pressure continues to decrease, then trip the reactor and perform 3-EOP-E-O, "Reactor Trip or Safety Injection".

C. Do NOT isolate IA to the Containment. Trip the reactor and perform 3-EOP-E-O, "Reactor Trip or Safety Injection." D. Do NOT isolate IA to the Containment.

If IA pressure continues to decrease, . then trip the reactor and perform 3-EOP-E-O, "Reactor Trip or Safety Injection".

29 NRC DRAFT 07/24/07 Q #15 ANSWER: A KA: 000065G2.1.23 As it relates to the loss of instrument air event: Able to perform specific system and integrated plant procedures during all modes of plant operation.

3.9/4.0 10CFR55: 41.bA, 41.b.10, 45.2

Reference:

0-ONOP-013:

Cog Level: 2 Comprehension CAUTION before Step 16, NOTE before Step 16, Step 17 RNO, FO Page Unit Trip Criteria Level 2 because the operator must compare the indicated containment IA pressure to IA system pressure and calculate the difference to determine if the pressure drop exceeds 10 psig and then apply procedure actions as required.

New Question Response Analysis:

A. Correct per the references.

B. Incorrect because the reactor must be tripped when IA is isolated to containment regardless of subsequent air pressure.

Plausible because isolating IA to containment should isolate the leak, mitigating the event. C. Incorrect because IA must be isolated to containment when its pressure drop exceeds 10 psi. Plausible because the reactor must be tripped under these conditions.

D. Incorrect because IA must be isolated to containment when its pressure drop exceeds 10 psi. Plausible because the reactor must be tripped under these conditions.

30 NRC DRAFT 07124/07 Q #16 Unit 3 operators are performing 3-EOP-ECA-1.2, "LOCA Outside Containment."

• After closing the RHR Discharge to Cold Leg Isolation valves, MOV-3-744A

& B, and SI to Cold Leg Isol valves, MOV-2-843A

& B, RCS pressure is still decreasing.

Which ONE of the following describes the correct operator response and the reason for the response?

A. Return to Step 1 of 3-EOP-ECA-1.2, "LOCA Outside Containment," and perform the isolation steps again as the leak is NOT isolated.

B. Transition to 3-EOP-ES-1.2, "Post LOCA Cooldown and Depressurization" as the steps in ECA-1.2 have been successful in isolating the leak so subsequent cooldown and depressurization will be required to go onto RHR cooling. C. Transition to step in effect in 3-EOP-E-1, "Loss of Reactor or Secondary Coolant" as the steps in ECA-1.2 have NOT been successful and the completion of 3-EOP-E-1 will mitigate the situation.

D. Transition to 3-EOP-ECA-1.1 , "Loss of Emergency Coolant Recirculation" as the steps in ECA-1.2 have NOT been successful so the ECA-1.1 actions to prolong injection from the RWST will be required.

31 NRC DRAFT 07/24/07 Q #16 ANSWER: D KA: W/E04EA2.1 Able to determine and interpret the following as they apply to the LOCA OC: Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

3.4/4.3 10CFR55: 41.b.5,41.b.10

Reference:

3-EOP-ECA-1.2:

Step 3.a.RNO Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because the correct transition is to go to EOP-ECA-1.1 if pressure is still stable or decreasing.

Plausible because returning to step 1 implies another effort to identify and isolate the leak which is reasonable under these circumstances.

B. Incorrect because the correct transition is to go to EOP-ECA-1.1 if pressure is still stable or decreasing.

Plausible because the stem implies operators are responding to a LOCA and with pressure still stable, EOP-ES-1.2 is a logical choice to deal with an active LOCA. C. Incorrect because the correct transition is to go to EOP-ECA-1.1 if pressure is still stable or decreasing.

Plausible because per the same procedure step, the transition would be to EOP-E-1 if pressure were increasing.

D. Correct per the references 32 NRC DRAFT 07124/07 Q #17 Upon entry into 4-EOP-ECA-1.1, "Loss of Emergency Coolant Recirculation" the RO reports Unit 4 RWST level has decreased to 59,000 gallons. Which ONE of the following describes the correct operator response?

A. Depressurize all intact S/Gs to atmospheric pressure at a rate NOT to exceed an RCS cooldown rate of 100°F/hour to ensure the Unit 4 Accumulators fully inject. B. Stop the running Unit 4 HHSI, RHR and CS pumps. Align and start one Unit 3 HHSI pump to deliver flow from Unit 3 RWST. C. Stop the running Unit 4 RHR and CS pumps. Continue to run one Unit 4 HHSI pump for as long as it will deliver flow from Unit 4 RWST. D. Establish minimum charging to deliver flow from Unit 4 RWST. When minimum charging flow has been established, stop the running HHSI RHR and CS pumps. 33 NRC DRAFT 07/24/07 Q#17 ANSWER: B KA: W/E11EK2.2 Knowledge of the interrelations between the loss of emergency coolant recirc and the following:

facility's heat removal systems incl primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

3.9/4.3 10CFR55: 41.b.7,41.b.10

Reference:

4-EOP-ECA-1.1:

Steps 1, 30, 31, 32 Cog Level: 2 Comprehension Level 2 because the operator must link the knowledge that RWST level < 60,000 gallons requires stopping the safety pumps. The 60,000 criteria is provided in Step 1 of the procedure and stopping the pumps is in Step 31. Additionally the operator must realize that stopping the safety pumps stops safety injection flow to the core and an alternate method of core cooling (opposite unit HHSI flow) is immediately required.

New Question Response Analysis:

A. Incorrect because the correct response is to stop all safety pumps pulling from that RWST and align the opposite unit's HHSI pump. Plausible because depressurizing S/Gs is a subsequent procedure step but if done it will be performed at the maximum rate. B. Correct per the references C. Incorrect because all safety pumps are stopped, including HHSI pumps. Plausible because the RHR and CS pumps are stopped and continuing to run HHSI pumps would continue to provide core cooling flow. D. Incorrect because the safety pumps need to be stopped now. Plausible because the charging pumps are probably already running (but at maximum speed) and will be allowed to run until RWST level drops to 20,000 gallons. 34 NRC DRAFT 07/24/07 Q #18 Unit 3 operators are performing 3-EOP-FR-H.1, "Response to Loss of Secondary Heat Sink." The BOP is attempting to restore AFW flow when the RO reports the following S/G levels:

• 3A S/G:

• 3B S/G:

• 3C S/G: 35% WR 25% NR 30% NR lowering lowering lowering The BOP is unable to immediately restore secondary heat sink. The STA reports containment temperature has increased to 190°F. Which ONE of the following describes the correct operator response?

A. Maintain RCPs running. Continue to perform procedure steps to restore secondary heat sink. B. Maintain RCPs running. Immediately initiate bleed and feed. C. Stop RCPs. Continue to perform procedure steps to restore secondary heat sink. D. Stop RCPs. Immediately initiate bleed and feed. 35 NRC DRAFT 07/24/07 Q #18 ANSWER: D KA: W/E05EK1.2 Knowledge of the operational implications of the following concepts as they apply to the Loss of secondary heat sink: Normal, abnormal and emergency operating procedures associated with the loss of secondary heat sink. 3.9/4.5 10CFR55: 41.bA, 41.b.8, 41.b.10

Reference:

3-EOP-FR-H.1

CAUTION prior to Step 2, Step 12 Cog Level: 2 Comprehension Level 2 because the operator must evaluate the S/G levels provided based on adverse containment conditions

(>180°F).

The operator must also recognize that 35% WR S/G level is < 32% NR level and bleed and feed criteria are met. New Question Response Analysis:

A. Incorrect because RCPs need to be tripped immediately.

Plausible because operators will continue efforts to restore secondary heat sink. B. Incorrect because RCPs need to be tripped immediately.

Plausible because operators will immediately initiate bleed and feed. C. Incorrect because bleed and feed needs to be established immediately.

Plausible because RCPs need to be stopped. D. Correct per the references 36 NRC DRAFT 07/24/07 Q #19 Unit 4 is in Mode 6 when the RO is notified that a spent fuel element has dropped onto the Region II fuel racks. Which ONE of the following identifies control room indications that can be used to determine if the fuel element cladding has been breached?

A. CVCS Letdown PRMS R-20 monitor alarms B. Spent Fuel Building Area rad monitors alarm C. Containment Building Area rad monitors alarm D. Containment Atmosphere PRMS R-11 or R-12 monitors alarm 37 NRC DRAFT 07/24/07 Q #19 ANSWER: B KA: 000036AA2.01 Able to determine and interpret the following as they apply to the fuel handling incidents:

ARM system indications.

3.2/3.9 10CFR55: 41.b.10, 41.b.11

Reference:

4-0NOP-033.3:

Section 2.2 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because the letdown monitor alarming is not identified as an indication of a dropped rod and this dropped rod is in the SFP, not the containment.

Plausible because there is an indirect path from the SFP to the letdown monitor (through the keyway, transfer canal and RHR system) B. Correct per the reference C. Incorrect because the event did not occur in containment.

Plausible because the question does not state that the event occurred in the Spent Fuel Pool area. D. Incorrect because the event did not occur in containment.

Plausible because the question does not state that the event occurred in the Spent Fuel Pool area. 38 NRC DRAFT 07/24/07 Q#20 Unit 4 is 100% power with all systems in normal operation when a tube leak occurs in the 4A S/G. The tube leak increases to 250 gpm over a period of 5 minutes. Which ONE of the following describes the effect on charging and letdown flow rates as the leak starts (before operator response) and AFTER operators have performed the prompt actions of 4-EOP-E-0, Reactor Trip or Safety Injection?

Early in S/G Tube Leak Event After Prompt Actions Complete Chging Flow Ltdwn Flow Chging Flow Ltdwn Flow A. Unchanged Unchanged Zero Unchanged B. Increasing Decreasing Maximum Zero C. Increasing Unchanged Zero Zero D. Unchanged Unchanged Zero Unchanged 39 NRC DRAFT 07124/07 Q#20 ANSWER: C KA: 000037 AK3.03 Knowledge for the reasons for the following responses as they apply to the SGTL: comparison of makeup flow and letdown flow for various modes of operation.

3.1/3.3 10CFR55: 41.b.7

Reference:

4-0NOP-071.2, FO Page Item 1.b, 561 O-T-L 1, Sheet 32A SO-009, Page 37, SO-013, Page 18 Cog Level: 2 Comprehension Level 2 because the operator must evaluate the size of the SGTL. Before SI actuation, Charging pump speed will increase in response to lowering pressurizer level and letdown flow rate will be constant.

SI actuation will result in Phase A actuation which will isolate letdown. Charging pumps will be tripped directly by the SI actuation.

New Question Response Analysis:

A. Incorrect because charging pump speed will increase before operators respond and letdown will isolate following SI /actuation as a result of Phase A isolation.

Plausible because letdown flow will be unchanged initially and charging flow will go to zero after SI actuation.

B. Incorrect because letdown flow will initially be unaffected and charging flow will go to zero after SI actuation.

Plausible because charging pump speed will increase before operators respond and letdown flow will isolate following Phase A actuation.

C. Correct because charging pump speed will increase before operators respond and charging pumps will trip when SI actuates.

Letdown flow will initially be unaffected and letdown flow will isolate following Phase A actuation.

O. Incorrect because charging pump speed will increase before operators respond and letdown will isolate following SI actuation as a result of Phase A isolation.

Plausible because Letdown flow will initially be unaffected and charging pumps will trip when SI actuates.

40 NRC DRAFT 07/24/07 Q#21 Operators are performing 3-0NOP-100, "Fast Load Reduction" in response to rapidly dropping condenser vacuum.

• Unable to recover vacuum, operators trip the unit and enter the EOPs. C 8/3, STEAM DUMP ARMED / ACTUATED, has alarmed. Two minutes after the unit trip and with the plant stable at 545°F, the BOP notes annunciator C 8/3 has cleared. Which ONE of the following describes the reason for annunciator C 8/3 clearing AND the status of the Steam Dump to Condenser (SDTC) System? The SDTC arming signal cleared when: A. vacuum dropped below 20". The SDTC System is disabled.

B. vacuum dropped below 20". The SDTC System remains fully functional.

C. Tavg dropped to the low Tavg setpoint.

The SDTC System is disabled.

D. Tavg dropped to the low Tavg setpoint.

The SDTC System remains fully functional.

41 NRC DRAFT 07/24/07 Q#21 ANSWER: A KA: 000051AK3.01 Knowledge for the reasons for the following responses as they apply to the loss of condenser vacuum: loss of steam dump capability upon loss of condenser vacuum. 2.8/3.1 10CFR55: 41.bA,41.b.5

Reference:

561 O-T-L 1, Sheet 22A Cog Level: 2 Comprehension Level 2 because the operator must recognize that 20" is a condition that may be expected for this situation and is a signal that disables the SDTC while the 545°F Tavg is a condition that could also be expected for this situation but does not disable the SDTC. Additionally the operator must recognize that the alarm may extinguish because the system has been disabled or has merely been reset and in this case the alarm is out because the system has been disabled.

New Question Response Analysis:

A. Correct per the references B. Incorrect because the SDTC system has become disabled.

Plausible because it became disabled when vacuum dropped below 20". C. Incorrect because even though Tavg did drop below the no-load Tavg under the influence of the SDTC system, no-load Tavg is not a signal that disables the system. Plausible because Tavg did drop to no-load Tavg following the trip and the system has been disabled.

D. Incorrect because the SDTC system has been disabled.

Plausible because Tavg did drop below the no-load Tavg setpoint following the trip. 42 NRC DRAFT 07/24/07 0#22 The Shift Manager directs evacuation of the control room.

• At the Alternate Shutdown Panel (ASP), the RO inserts handles into the yellow bordered switches and places them in the LOCAL position.

Which ONE of the following describes the purpose of the transfer switches associated with the AFW flow control valves (FCVs)? A. transfers control of Train 2 FCVs to the ASP B. enables local control of Train 2 FCVs C. aligns "8" train electrical control to Train 1 FCVs D. aligns backup nitrogen to the Train 1 FCVs 43 NRC DRAFT 07/24/07 Q#22 ANSWER: A KA: 000068G2.1.28 As it relates to the control room evacuation event: Knowledge of the purpose and function of major system components and controls.

3.2/3.3 10CFR55: 41.b.4, 41.b.7, 41.b.8

Reference:

O-ONOP-1 05, Attachment 3, NOTE prior to Step 6, Step 6 SD-153, Page 41, Figure 4C, 40 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference B. Incorrect because local control of the AFW valves is independent of the transfer switch. Plausible because these are the same Train 2 FCVs. C. Incorrect because the affected valves are Train 2 not Train 1. Plausible because the transfer switches transfer electrical control. D. IN correct because transfer of backup nitrogen is independent of the transfer switch. Plausible because Train 1 FCVs have backup nitrogen.

44 NRC DRAFT 07124/07 Q#23 The crew is doing a post LOCA cooldown and depressurization in accordance with 4-EOP-ES-1.2, "Post LOCA Cooldown and Depressurization."

• 4B RCP has been started for forced cooling and RCS pressure control.

• Two ECCs are running

• All NCCs are operating Containment pressure just increased greater than 20 psig. Which ONE of the following describes actions associated with operation of the Normal Containment Coolers (NCCs)

AND the bases for those actions? A. Continue Operation of the NCCs to maintain RCP operation and to assist in containment pressure reduction.

B. Continue operation of the NCCs to maximize containment cooling and atmosphere circulation to prevent stratification and eliminate pockets of hydrogen.

C. Stop all NCCs to start the 3rd Emergency Containment Coolers without violating CCW System load requirements.

D. Stop all NCCs because the 4B RCP must be secured and to prevent violating CCW System load requirements.

45 NRC DRAFT 07/24/07 Q#23 ANSWER: D KA: W/E14EK3.3 Knowledge for the reasons for the following responses as they apply to the high containment pressure:

manipulation of controls required to obtain desired operating results during abnormal and emergency situations.

3.5/3.5 10CFR55: 41.b.7, 41.b.8, 41.b.10

Reference:

4-EOP-F-O, Enclosure 5 4-EOP-FR-Z.1, Step 1 4-EOP-FR-Z.1, Step 1 BD Cog Level: 2 Comprehensive Level 2 because the operator must recognize that that the conditions have changed requiring transition to EOP-FR-Z.1 which will direct securing of the RCPs which in turn allows the NCCs to be secured. New Question Response Analysis:

A. Incorrect because the NCCs and RCPs have to be stopped per EOP-FR-Z.1.

Plausible because running NCCs would help to maximize containment pressure reduction and cool the RCP if it was left running. B. Incorrect because the NCCs have to be stopped per EOP-FR-Z.1.

Plausible because running NCCs would help to maximize containment air circulation.

C. Incorrect because procedural limitations restrict operators from running three ECC. Plausible because an additional ECC would help with the containment pressure reduction.

D. Correct, the NCCs are required to be stopped lAW FR-Z.1 guidance.

46 NRC DRAFT 07/24/07 0#24 3-EOP-FR-C.2, "Response to Degraded Core Cooling," directs operators to depressurize the RCS to inject the accumulators into the RCS. Which ONE of the following describes how the RCS will be depressurized to inject the accumulators and how operators will prevent accumulator nitrogen injection?

A. Open one pressurizer PORV until RCS pressure is less than 180 psig. Isolate accumulators when RCS pressure is less than 350 psig. B. Open both pressurizer spray valves until RCS pressure is less than 180 psig. Isolate accumulators when RCS pressure is less than 350 psig. C. Dump steam from intact S/Gs until SG Pressure is less than 180 psig. Isolate accumulators when RCS CET subcooling is less than 30°F. D. Dump steam from intact S/Gs until S/G pressure is less than 80 psig. Isolate accumulators when RCS hot leg temperatures are less than 340°F. 47 NRC DRAFT 07124/07 Q#24 ANSWER: D KA: W/E06EK1.1 Knowledge of the operational implications of the following concepts as they apply to the degraded core cooling: components, capacity and function of emergency systems. 3.6/4.0 10CFR55: 41.b.3, 41.bA, 41.b.5, 41.b.7

Reference:

3-EOP-FR-C.2, Steps 13 and 15 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because RCS pressure is reduced by dumping steam from intact S/Gs. Plausible because opening a PORV is a method often used during EOP implementation to depressurize the RCS. B. Incorrect because RCS pressure is reduced by dumping steam from intact S/Gs. Plausible because opening pressurizer spray valves is a method often used during EOP implementation to depressurize the RCS. C. Incorrect because the parameter used to judge when to stop dumping steam is S/G pressure, the stopping criteria in ECA-O.O is S/G Pressure at 180psig. Plausible because reducing subcooling implies reducing RCS pressure also D. Correct per the reference 48 NRC DRAFT 07/24/07 Q#25 In accordance with 3-EOP-ES-1.1, "SI Termination," which ONE of the following describes why seal return is restored to the VCT AND why RCS pressure must be greater than 100 psi above VCT pressure before establishing seal return flow? Seal return is restored to the VCT to: A. stop # 1 sealleakoff flow from flowing to the PRT. RCS pressure must be higher than VCT pressure to prevent Hydrogen intrusion from the VCT into the RCS. B. stop # 1 sealleakoff flow from flowing to the PRT. RCS pressure must be higher than VCT pressure to prevent reverse flow from the VCT into the RCS. C. reinitiate

1. 1 seal leakoff flow. RCS pressure must be higher than VCT pressure to prevent Hydrogen intrusion from the VCT into the RCS. D. reinitiate

1. 1 sealleakoff flow. RCS pressure must be higher than VCT pressure to prevent reverse flow from the VCT into the RCS . 49 NRC DRAFT 07124/07 Q#25 ANSWER: B KA: W/E02EK3.1 Knowledge for the reasons for the following responses as they apply to the SI termination:

Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

3.3/3.6 10CFR55: 41.b.2, 41.b.3, 41.b.10

Reference:

3-EOP-ES-1.1, Step 20 BO 5613-M-3047,Sheet3 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because the stated reason for the required 100 psi differential in the Basis document is to prevent backflow from VCT to RCS, not to prevent Hydrogen intrusion from the VCT into the RCS. Plausible because stopping seal return flow to the PRT is the reason for restoring normal seal return. B. Correct per the references C. Incorrect because stopping seal return flow to the PRT is the reason for restoring normal seal return. Note that even though seal return was isolated, #1 sealleakoff still existed to the PRT. Plausible because Hydrogen is present in the VCT which can find its way to the RCS if a flow path is established.

O. Incorrect because stopping seal return flow to the PRT is the reason for restoring normal seal return. Note that even though seal return was isolated, #1 sealleakoff still existed to the PRT. Plausible because the stated reason for the required 100 psi differential in the Basis document is to prevent backflow from VCT to RCS, 50 NRC DRAFT 07124/07 Q#26 A RCS leak has occurred that required a manual reactor trip based on the inability to maintain Pressurizer level. Operators are performing 4-EOP-ES-1.2, "Post LOCA Cooldown and Depressurization." The US directs the RO to "depressurize the RCS to refill the pressurizer." Which ONE of the following identifies the preferred order of methods to depressurize the RCS as directed by 4-EOP-ES-1.2?

A. 1) Normal Spray 2) PRZ PORV 3) Aux Spray B. 1) Normal Spray 2) Aux Spray 3) PRZ PORV C. 1) Aux Spray 2) PRZ PORV 3) Steam Dump D. 1) Aux Spray 2) Normal Spray 3) PRZ PORV 51 NRC DRAFT 07124/07 Q#26 ANSWER: A KA: W/E03EK1.3 Knowledge of the operational implications of the following concepts as they apply to LOCA Cooldown and Depressurization:

annunciators and conditions, indicating signals and remedial actions associated with the (LOCA cooldown and depressurization.

3.5/3.8 10CFR55: 41.b.3, 41.b.7, 41.b.10

Reference:

4-EOP-ES-1.2, Step 10 Cog Level: 1 Recall Bank Question Response Analysis:

A. Correct per the reference B. Incorrect because a pressurizer PORV is preferred over Aux Spray .. Plausible because normal spray is the preferred first method to reduce pressure.

C. Incorrect because steam dump is not a method directed by ES-1.2 and a pressurizer PORV is preferred over Aux Spray. Plausible because all three methods listed are effective methods of RCS pressure reduction under these conditions.

D. Incorrect because Normal spray is the first choice directed by ES-1.2. Plausible because all three methods listed are effective methods of RCS pressure reduction under these conditions and all are methods directed by ES-1.2. 52 NRC DRAFT 07/24/07 0#27 In accordance with 4-EOP-FR-P.1, "Response to Imminent Pressurized Thermal Shock Condition", which ONE of the following describes the purpose of starting an RCP while responding to a pressurized thermal shock condition?

An RCP is started to: A. minimize the time to reach Mode 5 by allowing a faster cooldown rate with forced cooling than with natural circulation.

B. restore normal pressurizer spray capability to subsequently reduce RCS pressure to the right of the 60° Ihr cooldown curve. C. mix the cold SI water with warm RCS water and decrease the likelihood of a PTS condition.

D. mix the water in the vessel and loops to ensure boron concentration is equal throughout the reactor coolant system. 53 NRC DRAFT 07124/07 Q#27 ANSWER: C KA: W IE08G2.1 .28 RCS overcooling

-PTS, As it relates to the PTS event: Knowledge of the purpose and function of major system components and controls.

3.2/3.3 10CFR55: 41.b.2, 41.b.3, 41.b.7, 41.b.10

Reference:

3-EOP-FR-P.1, Step 27 BD Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because starting an RCP does not allow a faster cooldown rate in FR-P.1. Plausible because the normal cooldown rate limit with RCPs is 1 OO°F/hr and without RCPs is 25°F/hr. B. Incorrect because RCS pressure will be subsequently maintained within a band between the minimum subcooling curve and the 200°F subcooling curve. Plausible because even though EOP-FR-P.1 does not specify which RCP should be started, operators will normally start an RCP with pressurizer spray capability.

C. Correct per the reference D. Incorrect because mixing of boron in the loops is not a basis described in the EOP-FR-P.1 basis document.

Plausible because running a RCP provides this desirable benefit. 54 NRC DRAFT 07/24/07 0#28 Operators are cooling down the plant in accordance with 3-GOP-305, "Hot Standby to Cold Shutdown".

3B RCP pump bearing temperature has been slowly increasing and has just reached the alarm setpoint value. The present plant conditions are as follows:

• RCS pressure is 600 psig.

• 3B RCP #1 seal llP is > 400 psid.

• 3B RCP seal injection flow is 10 gpm.

• 3B RCP #1 sealleakoff flow is 0.9 gpm.

• 3B RCP # 1 sealleakoff isol valve, CV-3-303B, is open. Which ONE of the following describes the correct operator response?

A. Increase seal injection flow to the 3B RCP to greater than 13 gpm. B. Open the RCP Seal Bypass Valve, CV-3-307.

C. Close the 3B RCP #1 seal leakoff isol valve, CV-3-303B.

D. Increase CCW flow through the 3B RCP thermal barrier. 55 NRC DRAFT 07124/07 Q#28 ANSWER: B KA: 003A4.07 Ability to manually operate and/or monitor in the control room RCP seal bypass. 2.6/2.6 10CFR55: 41.b.5,41.b.10

Reference:

3-GOP-305, Step 5.3.5.9.d 3-0NOP-041.1 , Step 35 Cog Level: 2 Comprehension Level 2 because the operators must evaluate plant conditions and determine that opening the bypass valve would improve the RCP conditions.

Many plant conditions must be met in order to use the bypass valve and the operator must analyze the current conditions to determine that it is appropriate and desired in this situation.

New Question Response Analysis:

A. Incorrect because the required procedural response is to open the RCP Seal Bypass Valve, CV-3-307.

Plausible because the RCP sealleakoff flow rate is low and the RCP bearing temperature is increasing.

B. Correct per the reference.

C. Incorrect because the required procedural response is to open the RCP Seal Bypass Valve, CV-3-307.

Plausible because CV-303B will be closed in a subsequent step in the same procedure (GOP-305, Step 5.19.14,9.b) and would terminate the flow of hot water to the VCT D. Incorrect because the required procedural response is to open the RCP Seal Bypass Valve, CV-3-307.

Plausible because increasing CCW through the thermal barrier would be beneficial to pump bearing cooling if seal injection flow was lost. 56 NRC DRAFT 07/24/07 0#29 Unit 3 is at 100% power when VCT level transmitter, L T-3-115, fails high. Which ONE of the following describes the effect on the plant assuming NO operator action? VCT level will: A. decrease.

Auto-makeup will NOT occur. Auto-swap to the RWST will occur. B. decrease.

Auto-makeup will NOT occur. Auto-swap to the RWST will not occur. C. increase.

Auto-makeup will start and stop automatically.

LCV-115A will auto-divert to the CVCS HUT. D. increase.

Auto-makeup will start but NOT stop automatically.

LCV-115A will auto-divert to the CVCS HUT. 57 NRC DRAFT 07/24/07 Q#29 ANSWER: B KA: 004K 1 .23 Knowledge of the physical connection and/or cause-effect relationships between the CVCS and the RWST. 3.4/3.7 10CFR55: 41.b.6,41.b.7

Reference:

3-0NOP-046.4, Step 28, CAUTION and NOTE prior to Step 29 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because auto-swap to the RWST will not occur (2 of 2 low levels required).

Plausible because VCT level will decrease and auto-makeup will not occur. B. Correct per the reference.

C. Incorrect because VCT level will decrease and auto-makeup will not occur. Plausible because LCV-115A will auto-divert to the CVCS HUT. D. Incorrect because VCT level will decrease and auto-makeup will not occur. Plausible because LCV-115A will auto-divert to the CVCS HUT. 58 NRC DRAFT 07/24/07 Q#30 Unit 3 is operating at 75% power when an Instrument Air System leak reduces the IA pressure to 60 psig on both units. Operator actions are unsuccessful in restoring IA pressure.

Which ONE of the following describes the effect of this event on Unit 3 pressurizer level and the correct operator response?

Pressurizer level will: A. decrease.

Trip the reactor and perform EOP-E-O, "Reactor Trip or Safety Injection." B decrease.

Start additional charging pumps and increase speed using 3-0P-047, "cvcs Charging and Letdown." C increase.

Trip the reactor and perform EOP-E-O, "Reactor Trip or Safety Injection." D. increase.

Open additional letdown orifice isolation valves using 3-0P-047, "cvcs Charging and Letdown." 59 NRC DRAFT 07/24/07 Q#30 ANSWER: C KA: 004A2.11 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

loss of lAS. 3.6/4.2 10CFR55: 41.bA, 41.b.5, 41.b.1 0

Reference:

0-ONOP-013, FO Page Cog Level: 2 Comprehension Level 2 because the operator must recognize that the loss of IA will result in the running charging pump(s) going to full speed and the letdown valves closing resulting in increasing pressurizer level. The operator must also relate the plant conditions and system responses to the need for a reactor must be tripped when IA pressure drops to <65 psig. New Question Response Analysis:

A. Incorrect because pressurizer level will increase.

Plausible because the correct response to IA pressure at 60 psig is to trip the reactor and enter EOP-E-O. B. Incorrect because pressurizer level will increase.

Plausible because if level were to decrease, starting additional charging pumps would be the appropriate operator C. Correct per the reference D. Incorrect because opening additional orifice valves is not possible due to the loss of IA. Plausible because pressurizer level will increase.

60 NRC DRAFT 07/24/07 Q #31 Unit 4 is in Mode 4 with RHR cooling in service. The automatic controller circuit for RHR Heat Exchanger Bypass Flow control valve, FCV-4-605 has failed resulting in the FCV going closed. Which ONE of the following describes the correct operator response?

Place FCV-4-605 controller in Manual and raise flow to: A. between 3000 and 3750 gpm. If manual control is NOT possible, direct the NSO to locally control FCV-4-605 to raise flow to between 3000 and 3750 gpm. B. > 3750 gpm. If manual control is NOT possible, direct the NSO to locally control FCV-4-605 to raise flow to > 3750 gpm. C. between 3000 and 3750 gpm. If manual control is NOT possible, open RHR Heat Exchanger Outlet Flow control valve, HCV-4-758 to raise flow to between 3000 and 3750 gpm. D. > 3750 gpm. If manual control is NOT possible, open RHR Heat Exchanger Outlet Flow control valve, HCV-4-758 to raise flow to > 3750 gpm. 61 NRC DRAFT 07124/07 Q#31 ANSWER: A KA: 005G2.1.30 As it relates to RHR, ability to locate and operate components, including local controls.

3.9/304 10CFR55: 41.bo4, 41.b.7, 41.b.8, 41.b.10

Reference:

4-0NOP-050, Step 7 Cog Level: 1 memory New Question Response Analysis:

A. Correct per the reference B. Incorrect because flow should be increased to between 3000 and 3750 gpm. Plausible because if control room control is not possible, the NSO will be directed to locally control FCV-605. C. Incorrect because operators are directed to use FCV-605, not HCV-758 to locally control flow. Plausible because flow should be increased to between 3000 and 3750 gpm. D. Incorrect because flow should be increased to between 3000 and 3750 gpm and operators are directed to use FCV-605, not HCV-758 to locally control flow. Plausible because a previous revision of ONOP-050 had operators increase flow using HCV-758. 62 NRC DRAFT 07/24/07 0#32 The RO manually tripped the reactor due to decreasing pressurizer pressure.

An automatic SI occurred on the trip depressurization.

RCS pressure is currently 1700 psig and slowly lowering.

One minute after the automatic SI signal actuated, the RO depressed both SI reset pushbuttons on VPB. Which ONE of the following describes the response of the safety injection system? A. SI immediately reset when the reset pushbuttons were depressed.

B. SI reset occurs one minute after the reset pushbuttons were depressed.

C. SI reset occurs two minutes after the reset pushbuttons were depressed.

D. SI did NOT reset but will reset if the operators push the reset buttons after the 2 minute timer has timed out. 63 NRC DRAFT 07124/07 Q#32 ANSWER: B KA: 006A4.0B ECCS -Ability to manually operate and/or monitor in the control room, 4.2/4.3 10CFR55: 41.b.7,41.b.B

Reference:

5610-T-L 1, Sheet 11 Cog Level: 2 Comprehension Level 2 because the operator must recognize that the SI reset logic is different if the initiating event was SI due to automatic actuation or manual actuation.

The SI signal automatically actuated, therefore the system "remembers" the reset attempt and SI will reset after an additional minute (2 minutes total). The operator must relate that even though RCS pressure is below the automatic SI setpoint, the system will allow manual reset in two minutes. If the SI signal was manually actuated, the system does not remember the reset attempt within the first 2 minutes and SI will not reset until another reset attempt is made after the 2 minute timer times out. New Question Response Analysis:

A. Incorrect because SI did not reset in this situation.

Plausible because SI normally resets immediately when operators push the manual reset pushbuttons as it is normally more than two minutes after the initiation signal. SI will reset after one more minute, B. Correct. There is a two minute delay so SI will reset within one minute. C. Incorrect because SI was already reset. It reset one minute earlier as the 2 minute timer is from initiation signal and not the reset signal. Plausible because the reset logic does use a 2 minute timer and this would be a correct response if the RO attempted to reset immediately.

D. Incorrect.

The 2 minutes must be over before the reset will work from a manual actuation.

Plausible because this would apply if SI was manually initiated.

64 NRC DRAFT 07/24/07 Q#33 Operators are preparing to reduce PRT pressure which has approached its upper limit of 10 psig. Which ONE of the following describes the process for reducing PRT pressure under these conditions?

Verify the PRT nitrogen regulator, PCV-3-473, is: A. aligned to the PRT. Start a waste gas compressor.

Open the PRT vent valve, CV-3-549.

When PRT pressure reaches 6 psig to 8 psig, then close CV-3-549.

B. aligned to the PRT. Start a waste gas compressor.

Open the PRT vent valve, CV-3-549.

When PRT pressure reaches vent header pressure, then close CV-3-549.

C. isolated from the PRT. Start a waste gas compressor.

Open the PRT vent valve, CV-3-549.

When PRT pressure reaches 6 psig to 8 psig, then close CV-3-549.

D. isolated from the PRT. Start a waste gas compressor.

Open the PRT vent valve, CV-3-549.

When PRT pressure reaches vent header pressure, then close CV-3-549.

65 NRC DRAFT 07/24/07 Q#33 ANSWER: A KA: 007A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRT controls including maintaining PRT pressure.

2.7/2.9 10CFR55: 41.b.3, 41.b.4, 41.b.10

Reference:

3-0P-041.3, Section 7.3.2 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference B. Incorrect because PRT pressure must reach 6 to 8 psig before closing CV-549. Plausible because the PRT nitrogen regulator, PCV-3-473, is aligned to the PRT. C. Incorrect because the PRT nitrogen regulator, PCV-3-473, is aligned to the PRT. Plausible because when PRT pressure reaches 6 psig to 8 psig, then CV-3-549 is closed. D. Incorrect because the PRT nitrogen regulator, PCV-3-473, is aligned to the PRT and because PRT pressure must reach 6 to 8 psig before closing CV-549. Plausible because vent header pressure is lower than 6 to 8 psig and would represent a greater pressure reduction than that provided by the procedure instructions.

66 NRC DRAFT 07124/07 0#34 Unit 3 operators are responding to a Loss of All AC Power (LOAAC).

• The BOP is using 3-0NOP-004.2, "Loss of 3A 4KV Bus," to restore the 3A 4KV Bus which failed to load onto its EDG because of a failure of the 3A sequencer.

• As the BOP verifies 3A 4KV bus stripping, he notes that the 3A Bus loads that were running before the LOMC still show breaker closed indication (red light).

• The BOP places each of the 3A Bus components' control switches in the STOP position.

Which ONE of the following describes the response of the 3A CCW pump breaker light indication on VPB after the BOP places its control switch in STOP? The 3A CCW pump green light will: A. remain on. Its red light will remain off. B. immediately go off. Its red light will energize and remain on. C. remain on for 10 seconds at which time it will go out and the red light will go on. D. remain on for 30 seconds at which time it will go out and the red light will go on. 67 NRC DRAFT 07124/07 Q#34 ANSWER: C KA: OOBA4.0B Ability to manually operate and/or monitor in the control room: CCW pump control switch. 3.1/3.B 10CFR55: 41.b.4,41.b.7

Reference:

5610-T-L 1, Sheet 24D Cog Level: 3 Analysis/Application Level 3 because the operator must recognize that the 3A sequencer failure includes the bus stripping portion of the sequencer as evidenced by the operator's observation that the 3A Bus loads that were running before the LOAAC still have red breaker light indication.

Bus stripping failed to work and bus stripping is the signal that blocks the timed low pressure auto-start of CCW pumps. When the operator takes the 3A CCW pump to off, the breaker opens but 10 seconds later it closes in because bus stripping did not block the low pressure auto-start.

New Question Response Analysis:

A. Incorrect because its green light will remain on for only 10 seconds at which time it will go out and the red light will come on. Plausible because this is the way the system would respond if the sequencer/bus stripping had not failed. B. Incorrect because its green light will remain on for only 10 seconds at which time it will go out and the red light will come on. Plausible because more than ten seconds has elapsed and a common misconception is that the breaker will immediately close. C. Correct per the reference D. Incorrect because its green light will remain on for only 10 seconds at which time it will go out and the red light will come on. Plausible because this is the way 3C CCW pump indication would respond under these conditions and this is the same train pump. 68 NRC DRAFT 07/24/07 Q#35 Unit 3 is at 70% power with all systems in normal operation.

The control group of pressurizer heaters malfunctions and goes to minimum output. Which ONE of the following describes the effect on the pressurizer pressure control system AND the operator response required by 3-0NOP-041.5, "Pressurizer Pressure Control Malfunction"?

Both groups of backup heaters will automatically operate: A. immediately.

Dispatch an operator to locally verify the control group heater breakers are closed. B. immediately.

Manually cycle backup heaters as needed to maintain pressure at 2235 psig. C. when pressurizer pressure drops to 2210 psig. Dispatch an operator to locally verify the control group heater breakers are closed. D. when pressurizer pressure drops to 2210 psig. Manually cycle backup heaters as needed to maintain pressure at 2235 psig. 69 NRC DRAFT 07/24/07 Q#35 ANSWER: C KA: 01 OA2.01-Ability to (a) predict the impacts of the following malfunctions or operations on the PRZ PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Heater failures.

3.3/3.6 10CFR55: 41.b.3, 41.bA, 41.b.7

Reference:

3-0NOP-041.5, Step 12 5610-T-D-16B, Sheet 1 Cog Level: 2 Comprehension .Level2 because the operator must realize that the Control Group going to minimum will result in a decrease in pressurizer pressure.

The operator must then recall that the backup groups will both energize at 2210 psig and the correct response is to check the control group heater status. New Question Response Analysis:

A. Incorrect because the backup heaters will not energize until pressure drops to 2210 psig. Plausible because an operator should be dispatched to locally verify the control group heater breakers are closed. B. Incorrect because the backup heaters will not energize until pressure drops to 2210 psig and operators are not directed to use the backup heaters to control pressure at 2235 psig. Plausible because normal pressure is 2235 psig and controlling pressure at that value is easily done using the backup heaters. C. Correct per the reference D. Incorrect because operators are not directed to use the backup heaters to control pressure at 2235 psig. Plausible because the backup heaters will not energize until pressure drops to 2210 psig. 70 NRC DRAFT 07/24/07 Q#36 Operators are performing a shutdown of Unit 4 reactor.

• Power level is currently 50%.

• First stage pressure transmitter, PT-4-447, fails high. Which ONE of the following describes the effect of this failure on the normal operation of the reactor protection system during a plant shutdown?

At the current power level, a turbine trip signal can: A. cause a reactor trip. When power is less than 10%, turbine trip will be able to cause reactor trip. B. cause a reactor trip. When power is less than 10%, turbine trip will NOT be able to cause reactor trip. c. NOT cause a reactor trip. When power is less than 10%, turbine trip will be able to cause reactor trip. D. NOT cause a reactor trip. When power is less than 10%, turbine trip will NOT be able to cause reactor trip. 71 NRC DRAFT 07124/07 Q#36 ANSWER: A KA: 012K4.06 Knowledge of the RPS design feature( s) and/or interlock( s) which provide for automatic or manual enable/disable of RPS trips. 3.2/3.5 10CFR55: 41.b.7

Reference:

561 O-T-L 1, Sheet 2, Sheet 17 Cog Level: 2 Comprehension Level 2 because the operator must recognize that PT-447 failing high has no immediate affect on the RPS system but results in the inability to clear permissive P-7. Normally when turbine power reduces below 10%, the "At-Power" trips which include the turbine tripping the reactor are cleared. The clearing logic requires both PT-447 and PT-446 to be below 10%. With PT-447 failed high it cannot clear and the "At-Power" trips remain instated at all power levels. New Question Response Analysis:

A. Correct per the reference B. Incorrect because when power is <10%, turbine trip will be able to cause a reactor trip. Plausible because a turbine trip signal can presently cause a reactor trip C. Incorrect because a turbine trip signal can presently cause a reactor trip. Plausible because when power is <10%, turbine trip will be able to cause a reactor trip. D. Incorrect because a turbine trip signal can presently cause a reactor trip and because when power is <10%, turbine trip will be able to cause a reactor trip. Plausible because the turbine trip is not normally able to cause a reactor trip below 10% power. 72 NRC DRAFT 07/24/07 0#37 Unit 3 reactor power is at 40% with all systems in normal alignment except for NIS power range channel N-41 which is OOS with bistables tripped. Subsequently power range channel N-42 fails low. Which ONE of the following describes the immediate effect of the N-42 failure on the reactor trip logic? A. The reactor trip logic is NOT affected by the N-42 failure. B. The RCS Loop Low Flow reactor trip logic is disabled by the N-42 failure C. The RCS Loop Low Flow reactor trip logic changed to 2 of 3 flow channels to trip on 1 of 3 loops. D. The reactor trip logic disables the at-power reactor trips with 2 of 4 channels < 10% power. 73 NRC DRAFT 07/24/07 Q#37 ANSWER: A KA: 012K6.02 Knowledge of the effect of a loss or malfunction of the following will have on the RPS: Redundant Channels.

2.9/3.1 10CFR55: 41.b.7

Reference:

561 O-T-L 1, Sheet 2, Sheet 17, Sheet 20 Cog Level: 2 Comprehension Level 2 because the operator must recognize that the throwing of N-41 bistables inserted a "power above P-8" signal to the logic. This effectively changed the trip logic from 2 of 4 channels on 2 of 3 loops to 1 of 3 channels on 2 of 3 loops. When N-42 failed low it could not generate a "power above P-8" signal, effectively disabling its input to the logic. However the logic still recognizes inputs from N-43 and N-44 at 40% power and the loop trip logic remains the same even though N-42 input has been disabled.

New Question Response Analysis:

A. Correct per the reference B. Incorrect because the RCS Loop Low Flow reactor trip logic was not affected by this N-42 failure. Plausible because the N-42 input to the RCS Loop Low Flow reactor trip logic is disabled.

C. Incorrect because the RCS Loop Low Flow reactor trip logic was not affected by this N-42 failure. Plausible because the trip logic was initially changed to 1 of 3 NIS channels to trip when the N-41 bistables were originally thrown. D. Incorrect because the at-power trips require 3 of 4 power range channels be below 10% to be disabled.

Plausible because the reactor trip logic does disable the at-power reactor trips when < 10% power. 74 NRC DRAFT 07/24/07 0#38 Operators are responding to a main steam line break outside containment.

• All safety systems functioned as designed.

• S I has been reset.

• The BOP goes behind VPB to reset Containment Isolation Signals. The RO continues to monitor the control boards and notes the status of annunciator H 5/2, CNTMT ISOLATION ACTIVATED.

Which ONE of the following correctly describes the status of annunciator H 5/2? Before the BOP begins to reset Containment Isolation relays, H 5/2 is: A. in alarm, because Phase A and Containment Ventilation Isolation relays are tripped. After the BOP resets relays, H 5/2 is NOT in alarm because Phase A relays are reset. B. in alarm, because Phase A and Containment Ventilation Isolation relays are tripped. After the BOP resets relays, H 5/2 is still in alarm because Containment Ventilation Isolation relays are still tripped. C. not in alarm, because Phase B relays have not tripped. After the BOP resets relays, H 5/2 is NOT in alarm because Phase B relays have NOT tripped and Phase A relays are reset. D. NOT in alarm, because Containment Ventilation Isolation relays have NOT tripped. After the BOP resets relays, H 5/2 is still in alarm because Phase A relays cannot be reset under these conditions.

75 NRC DRAFT 07/24/07 Q#38 ANSWER: B KA: 013A4.02 Ability to manually operate and/or monitor in the control room reset of ESFAS. channels.

4.3/4.4 10CFR55: 41.b.7,41.b.9

Reference:

4-EOP-E-1, Step 7 4-ARP-097.CR, H 5/2 5610-T-L1, Sht.11 Cog Level: 2 Comprehension Level 2 because the operator must recognize that a steamline break outside containment will trigger Phase A and Containment Ventilation Isolation, but not Phase B. Additionally the operator must recognize that the alarm is active when any of the 3 signals have not been reset. It takes the reset of all 3 signals to clear the alarm. New Question Response Analysis:

A. Incorrect because after the BOP resets relays, H 5/2 is still in alarm because Containment Ventilation Isolation relays are still tripped. Plausible because before the BOP begins to reset Containment Isolation relays, H 5/2 is in alarm, because Phase A and Containment Ventilation Isolation relays are tripped. B. Correct per the reference C. Incorrect because before the BOP begins to reset Containment Isolation relays, H 5/2 is in alarm, because Phase A and Containment Ventilation Isolation relays are tripped. After the BOP resets relays, H 5/2 is still in alarm because Containment Ventilation Isolation relays are still tripped. Plausible because Phase B relays have not tripped and after the BOP resets relays, Phase A relays are reset. D. Incorrect because before the BOP begins to reset Containment Isolation relays, H 5/2 is in alarm, because Phase A and Containment Ventilation Isolation relays are tripped. After the BOP resets relays, H 5/2 is still in alarm because Containment Ventilation Isolation relays are still tripped. Plausible because after the BOP resets relays, H 5/2 is still in alarm. 76 NRC DRAFT 07124/07 Q#39 Operators are responding to a simultaneous LOOP/LOCA event. 3A EDG failed to start. Which ONE of the following describes the status of the Emergency Containment Coolers (ECCs) after sequencing is complete?

3AECC 3B ECC 3C ECC A. de-energized running running B. de-energized stopped running C. running stopped de-energized D. running running de-energized 77 NRC DRAFT 07124/07 Q#39 ANSWER: C KA: 022K2.01 Knowledge of the power supplies to the containment cooling fans. 3.0/3.1 10CFR55: 41.b.8,41.b.10

Reference:

SD-029, Page 15 561 0-T-E-1591, Sheet 1 5613-T-L 1, Sheet 128 Cog Level: 2 Comprehension Level 2 because the operator must recognize that the sequencers will auto-start their respective ECCs if power to the bus is available.

3A ECC is powered from the "8" train (38 MCC, bkr 30650). 3C ECC is powered from the "A" train 3C MCC, bkr 30729). 38 ECC is powered from either train (swing bus) but does not get an auto start by the sequencer even though the bus will energize.

New Question Response Analysis:

A. Incorrect because 3C ECC is de-energized and 38 does not auto-start.

Plausible because 3A safeguards equipment is normally powered from 3A Train which is de-energized.

B. Incorrect because 3C ECC is de-energized. Plausible because 3A safeguards equipment is normally powered from 3A Train which is energized.

C. Correct per the references D. Incorrect because 38 ECC does not auto-start.

Plausible because 3A ECC is running and 3C ECC is de-energized.

78 NRC DRAFT 07124/07 0#40 Following a large break LOCA, operators have transitioned to 3-EOP-ES-1.3, "Transfer to Cold Leg Recirculation," and are determining if "piggy-back" recirculation will be required.

The RO reports the following information:

• RHR flow on FI-3-605 is 2800 gpm

• Containment pressure is 12 psig.

• Containment temperature is 178°F.

• All ECF spray valves are closed. Which ONE of the following is correct regarding subsequent Containment Spray Pump (CSP) operation?

Operators will run: A. No CSPs. B. One CSP with suction provided directly from the RWST. C. One CSP with suction provided from the discharge of the running RHR pump(s). D. Two CSPs with suction provided from the discharge of the running RHR pump(s). 79 NRC DRAFT 07/24/07 Q#40 ANSWER: C KA: 026K4.01 Knowledge of the CSS design feature( s) and/or interlock( s) which provide for: source of water for CSS, including recirculation phase after LOCA 4.2/4.3 10CFR55: 41.b.7,41.b.8

Reference:

3-EOP-ES-1.3, Steps 18, 21, 23, 25, 26 Cog Level: 2 Comprehension Level 2 because the operator must recognize that one of the four criteria to discontinue CSP operation is not satisfied.

Additionally the operator must recall that one CSP was put in pull-to-Iock in Step 2 of EOP-ES-1.3 and must remain in pull-to-Iock subsequently.

Finally the operator must recognize that subsequent procedure steps change the CSP suction source from the RWST to the discharge of the RHR pumps. New Question Response Analysis:

A. Incorrect because 1 CSP must be run under these conditions.

Plausible because 3 of the 4 listed parameters support not starting a CSP. B. Incorrect because when a CSP is subsequently started, cold leg recirculation will have been aligned. Plausible because up to this time the running CSP has been aligned to the RWST. C. Correct per the reference D. Incorrect because one CSP was put in pull-to-Iock earlier in EOP-ES-1.3 and it will remain in pull-to-Iock.

Plausible because the running CSP will have its suction provided from the discharge of the running RHR pump(s). 80 NRC DRAFT 07124/07 0#41 Operators are performing 3-EOP-E-1, "Loss of Reactor or Secondary Coolant" and the following plant conditions exist:

• RCS Pressure is 1000 psig

• All RCPs are Off

• Containment Temperature is 220°F

• Containment Pressure is 19 psig and rising

• Two ECCs are running

• All Available Charging Pumps are running

• The RCS is saturated

• RWST level is 220,000 gallons Containment Pressure has just increased to 21 psig. The operators entered 3-EOP-FR-Z.1, "Response to High Containment Pressure" and are assessing Containment Spray pumps (CSPs) status. Which one of the following describes the correct plant and operator response?

A. Neither CSP auto-started.

Manually start both CSPs. B. Neither CSP auto-started.

Manually start one CSP and place the other CSP in Pull-To-Lock.

C. Both CSPs auto-started.

Verify both CSPs auto-started.

D. Both CSPs auto-started.

Manually stop one CSP and place it in Standby. 81 NRC DRAFT 07124/07 Q#41 ANSWER: A KA: 026A 1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including containment pressure 3.9/4.2 10CFR55: 41.b.7,41.b.8

Reference:

5610-T-L 1, Sheet 11 5613-T-L 1, Sheet 12A 30EOP-E-0, Attachment 3, Step 12 3-EOP-FR-Z.1 , Step 8 Cog Level: 2 Comprehension Level 2 because the operator must analyze plant conditions and determine that both CSPs are needed but neither auto-started based on the sequencers having been reset when SI was reset in EOP-E-O. New Question Response Analysis:

A. Correct because neither CSP auto started because the sequencer was reset when the operators reset SI in E-O and both CSPs are needed lAW FR-Z.1, Step 8 guidance.

B. Incorrect because both CSPs are needed lAW 3-EOP-FR-Z.1, Step 8 guidance.

Plausible because neither CSP auto-started because the sequencer was reset when the operators reset SI in E-O. C. Incorrect because neither CSP auto-started.

Plausible because operators are directed to verify both CSPs are running lAW 3-EOP-FR-Z.1, Step 8 guidance.

D. Incorrect because neither CSP auto-started.

Plausible because operators are directed to manually stop one CSP and place it in Standby if containment pressure is < 14 psig. 82 NRC DRAFT 07/24/07 0#42 Operators have responded to a small break LOCA on Unit 3 and are performing 3-EOP-ES-1.2, Post-LOCA Cooldown and Depressurization." The RO is preparing to initiate an RCS cooldown to Cold Shutdown by dumping steam from intact steam generators. . Which ONE of the following describes why the RO must limit steam flow rate to maintain the allowable cool down rate below 100°F/hr?

Excessive steam flow rate may result in: A. termination of RCS natural circulation flow. B. challenging the integrity status tree for pressurized thermal shock limits. C. automatic closure of the MSIVs which isolates the condenser steam dumps. D. exceeding the capability of the AFW system to maintain S/G levels above 6% which will require stopping the cooldown.

83 NRC DRAFT 07124/07 Q#42 ANSWER: B KA: 039K5.05 Knowledge of the operational implications of the following concept as it applies to the MRSS: Basis for the RCS cooldown limits 2.7/3.1 10CFR55: 41.b.3, 41.bA, 41.b.5, 41.b.10

Reference:

3-EOP-ES-1.2, Step 6, BD Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because excessive steam flow will not prevent RCS NC flow. Plausible because excessive steam flow may temporarily retard NC flow. B. Correct per the reference C. Incorrect because excessive steam flow will not occur under these conditions. Only possible when you cooldown at maximum rate .. D. Incorrect only ECA-O.O requires stopping the cooldown rate if SG level is less than 6% in all SGs. 84 NRC DRAFT 07124/07 Q#43 Unit 4 is at 80% power with all systems in auto when the 4A S/G pressure transmitter associated with the controlling S/G steam flow transmitter fails LOW. Which ONE of the following describes the immediate effect on 4A: 1) S/G steam flow signal to the feed reg valve controller

2) feed reg valve 3) S/G level A. 1) steam flow signal decreases

2) feed reg valve opens. 3) S/G level increases.

B. 1) steam flow signal decreases

2) feed reg valve closes. 3) S/G level decreases.

C. 1) steam flow signal increases

2) feed reg valve opens. 3) S/G level increases.

D. 1) steam flow signal increases

2) feed reg valve closes. 3) S/G level decreases.

85 NRC DRAFT 07/24/07 Q#43 ANSWER: B KA: 059K1.03 Knowledge of the physical connection and/or cause-effect relationships between the main feedwater and the S/Gs. 3.1/3.3 10CFR55: 41.b.4, 41.b.7

Reference:

561 O-T-0-17, Sheet 1 5610-T-0-18B,Sheet1 SO-11, Page 32 Cog Level: 2 Comprehension Level 2 because the operator must realize that when the pressure density compensation input to the steam flow signal fails low, the result is the steam flow signal to the feed reg valve controller also goes low. Steam flow is compared to feed flow and the resulting imbalance causes the feed reg valve to go closed with a resulting drop in S/G level. New Question Response Analysis:

A. Incorrect because for this failure the steam flow signal decreases, the feed reg valve closes and the S/G level decreases.

Plausible because the steam flow signal will decrease.

B. Correct per the references C. Incorrect because for this failure the steam flow signal decreases, the feed reg valve closes and the S/G level decreases.

Plausible because this response provides an accurate description of the effect of the pressure input failing high. O. Incorrect because for this failure the steam flow signal decreases, the feed reg valve closes and the S/G level decreases.

Plausible because the feed reg valve will close and the S/G level will decrease.

86 NRC DRAFT 07124/07 Q#44 Unit 4 is operating at 40% power with all systems in automatic operation.

• The sensing line for first stage pressure transmitter, PT-4-446, which is selected for control on VPA, shears off at the connection to the pressure transmitter.

• Several minutes later, the BOP selects PT-4-447 as the controlling channel as directed by the procedure.

Which ONE of the following describes the response of S/G levels assuming they are left in automatic throughout the entire evolution?

S/G levels will: A. remain stable at 60%. B. lower from 60% and stabilize at 50% and then return to 60%. c. lower from 60% and remain stable at 50%. D. lower from 60% continuously until PT-4-447 is selected and then rise to 50% 87 NRC DRAFT 07/24/07 Q#44 ANSWER: B KA: 059A3.02 Ability to monitor automatic operation of the MFW, including programmed levels of the S/G. 2.9/3.1 10CFR55: 41.bA,41.b.7

Reference:

561 0-T-D-17 Sheet 1 5610-T-D-18A Sheet 1 Cog Level: 3 Analysis Level 3 because the operator must recognize that when PT-446 is selected for control, it is providing programmed level input to the S/G level control program. A sheared sensing line will cause the pressure transmitter to call for a 0% power S/G level of 50%. At 40% power, the desired programmed level has just increased to 60%. When PT -446 fails low, the programmed level input changes to 50% and not o %. Feed reg valves respond and level drops to 50% on all S/Gs. Operator must know that even though failed low, the control loop is limited at 50% level or the level would continue to decrease.

They then must realize that when PT-4-447 is selected, the SG level control system will return level back to 60%. New Question Response Analysis:

A. Incorrect because S/G levels will lower from 60% and will stabilize at 50% and then return to 60% when PT 447 is selected.

Plausible because at 40% power the level program has just increased from 50% to 60%. B. Correct per the reference.

C. Incorrect because S/G levels will lower from 60% and will stabilize at 50% and then return to 60% when PT-4-447 is selected.

Plausible because other instrument failures will cause S/G levels to rise continuously until a turbine trip occurs. D. Incorrect because S/G levels will lower from 60% and will stabilize at 50% and then return to 60% when PT-4-447 is selected.

Plausible because other instrument failures will cause S/G levels to lower continuously until a reactor trip occurs. 88 NRC DRAFT 07124/07 Q#45 With Unit 4 at 30% power during a plant startup, the RO discovers the running Steam Generator Feed Pump (SGFP) start/stop hand switch in the green-flagged position.

Which ONE of the following describes the consequences of this discovery?

A. An AFW system auto-start signal has been disabled.

B. A condensate pump auto-start signal has been disabled.

C. The running SGFP can NOT trip as a result of bus stripping.

D. The turbine can NOT run back as a result of a SGFP trip. 89 NRC DRAFT 07124/07 Q#45 ANSWER: A KA: 061 K4.02 Knowledge of the AFW design feature( s) and/or interlock( s) which provide for: AFW automatic start upon loss of MFW pump, S/G level, blackout or SI. 4.5/4.6 10CFR55: 41.bA, 41.b.7

Reference:

5610-T-L1, Sheet 15, Notes 1 and 5 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference B. Incorrect because the Condensate pump auto-start feature is unaffected by this event. Plausible because Condensate pumps do have auto-start features but they are affected by the status of other Condensate pumps, not SGFPs. C. Incorrect because the S/G feed pump trip signal from bus stripping is unaffected.

Plausible because green flag indication implies to the AFW start logic that the SGFP has been manually stopped but it does not input to the bus stripping logic in the same manner. D. Incorrect because the S/G feed pump trip signal into the turbine run back logic is unaffected.

Plausible because green flag indication implies to the AFW start logic that the SGFP has been manually stopped but it does not input to the turbine runback logic in the same manner. 90 NRC DRAFT 07124/07 Q#46 Unit 3 is at 100% power with all systems in normal alignment.

The 3C main steam line non-return check valve body ruptures resulting in a large Main Steam Line Break (MSLB) at the 3C non-return check valve. Which ONE of the following describes the automatic response of the AFW system? A. Only 3A and 3B AFW pumps start and deliver flow. 390 gpm total AFW flow will be delivered to all S/Gs. B. Only 3A and 3C AFW pumps start and deliver flow. 780 gpm total AFW flow will be delivered to all S/Gs. C. Only 3B and 3C AFW pumps start and deliver flow. 390 gpm total AFW flow will be delivered to all S/Gs. D. All AFW pumps start and deliver flow. 780 gpm total AFW flow will be delivered to all S/Gs. 91 NRC DRAFT 07/24/07 Q#46 ANSWER: D KA: 061A3.01 Ability to monitor automatic operation of the AFW, including:

AFW startup and flows. 4.2/4.2 10CFR55: 41.bA, 41.b.7, 41.b.8

Reference:

5613-M-3072, Sheet 1 5610-T-L 1 Sheet 11 5610-T-L 1 Sheet 15 Cog Level: 3 Analysis/Application Level 3 because the operator must analyze the following to come to the correct conclusion.

A steam line break at the 3C non-return valve will not cause the 3C S/G to blow dry because the 3C MSIV will close. SI actuates due to Hi steamline flow with low Tavg/ low SG pressure when all 3 S/Gs feed the break at the 3C return valve. This SI signal closes all 3 MSIVs. Both trains of AFW get a start signal due to SI. With all 3 S/Gs pressurized, all AFW pumps auto start and provide flow to all S/Gs. The AFW flow controllers are preset at 130 gpm each. 130 gpm X 6 AFW reg valves = 780 gpm. New Question Response Analysis:

A. Incorrect because all AFW pumps start and deliver flow. 780 gpm total AFW flow will be delivered to all S/Gs. Plausible if the operator assumes the faulted 3C S/G blows down and 3C S/G is providing steam supply to the 3C AFW pump. B. Incorrect because all AFW pumps start and deliver flow. Plausible if the operator recognizes 3A and 3C AFW pumps are in different trains. In that case 780 gpm total AFW flow will be delivered to all S/Gs. C. Incorrect because all AFW pumps start and deliver flow. 780 gpm total AFW flow will be delivered to all S/Gs. Plausible if the operator assumes the faulted 3C S/G blows down. 3A AFW pump would lose its steam supply and 390 gpm total AFW flow would be delivered to all S/Gs. D. Correct per the references 92 NRC DRAFT 07/24/07 Q#47 Tech Spec 3.8.1.1, AC Power Sources, allows a unit to continue power operation at less than 30% power for up to 30 days if its Startup Transformer becomes inoperable.

Which ONE of the following describes the basis for allowing continued operation at less than 30% power? At 30% power: A. the two loop low flow / two RCP breaker open reactor trip logic has been instated providing additional reactor protection in the event of a LOOP. B. S/G pressure is greater than full power S/G pressure providing additional motive force to the steam driven AFW pumps in the event of AFW actuation.

C. fewer components are loaded onto the 4KV Buses and Load Centers resulting in lower bus and load center operating temperatures.

D. decay heat has been reduced, automatic feedwater control can be maintained and the RCPs continue to run. 93 NRC DRAFT 07/24/07 Q#47 ANSWER: D KA: 062G2.2.25 As it relates to the AC Electrical Distribution:

Knowledge of the bases in Tech Specs for LCOs and safety limits. 2.5/3.7 10CFR55: 41.b.1,41.b.5,41.b.7,41.b.8

Reference:

0-ADM-536, Page 94; Section 3/4.8.1 Cog Level: 1 Recall Bank Question Response Analysis:

A. Incorrect because the basis for allowing continued operation at 30% power is decay heat has been reduced, automatic feedwater control can be maintained and natural circulation conditions are avoided by staying at power and not shutting down. Plausible because at 30% power, the two loop low flow / two RCP breaker open reactor trip logic has been instated.

B. Incorrect because the basis for allowing continued operation at 30% power is decay heat has been reduced, automatic feedwater control can be maintained and natural circulation conditions are avoided by staying at power and not shutting down. Plausible because at 30% power, S/G pressure is greater than full power S/G pressure.

C. Incorrect because the basis for allowing continued operation at 30% power is decay heat has been reduced, automatic feedwater control can be maintained and natural circulation conditions are avoided by staying at power and not shutting down. Plausible because at 30% power, fewer components are loaded onto the 4KV Buses and Load Centers. D. Correct per the reference 94 NRC DRAFT 07/24/07 0#48 Both Units are at 100% power with normal system alignments except for 40 MCC which is out of service. Subsequently the feeder breaker to 3C MCC trips open, de-energizing 3C MCC. Which ONE of the following describes the effect on the Vital DC electrical system? A. 3A DC Bus can only be expected to maintain voltage to shutdown loads above 120 volts for four hours. B. 3A DC Bus can only be expected to maintain voltage to shutdown loads above 105 volts for two hours. C. 3B DC Bus can only be expected to maintain voltage to shutdown loads above 120 volts for four hours. D. 3B DC Bus can only be expected to maintain voltage to shutdown loads above 105 volts for two hours. 95 NRC DRAFT 07124/07 Q#48 ANSWER: B KA: 063K1.03 Knowledge of the physical connection and/or cause-effect relationships between the DC Electrical Distribution and the battery charger and battery. 2.9/3.5 10CFR55: 41.b.4, 41.b.7, 41.b.8

Reference:

5610-T-E-1592 Sheet 1 SD-144, Page 9 Cog Level: 2 Comprehension Level 2 because the operator must recognize that 4D MCC is the power supply to the 3A2 battery charger and 3C MCC is the power supply to the 3A 1 battery charger. Loss of both results in no battery chargers for the 3A DC Bus. The operator must then recall that the 3A battery is sized to maintain voltage to shutdown loads above 105 volts for two hours. New Question Response Analysis:

A. Incorrect because 3A DC Bus can only be expected to maintain voltage to shutdown loads only above 105 volts for only two hours. Plausible because it is the 3A DC Bus which is affected.

B. Correct per the reference C. Incorrect because the 3A DC Bus is the affected bus. Plausible because 120 volts is the approximate normal voltage for all DC buses. D. Incorrect because the 3A DC Bus is the affected bus. Plausible because 3A DC Bus can only be expected to maintain voltage to shutdown loads above 105 volts for two hours. 96 NRC DRAFT 07124/07 0#49 Unit 3 is in Mode 3 when the input breakers to vital DC buses 3D23 and 3D23A trip open deenergizing the DC buses. Which ONE of the following describes the effect on the Unit 3 EDG(s)? A. 3A EDG has lost control and field flashing power. 3A EDG does NOT have black start capability.

B. 3B EDG has lost control and field flashing power. 3B EDG does NOT have black start capability.

C. 3A EDG has lost control and field flashing power. 3A EDG is still black start capable. D. 3B EDG has lost control and field flashing power. 3B EDG is still black start capable. 97 NRC DRAFT 07/24/07 Q#49 ANSWER: 8 KA: 064K2.03 EDGs -Knowledge of the power supplies to the control power. 2.9/3.3 10CFR55: 41.bA, 41.b.7, 41.b.8

Reference:

3-0NOP-003.5, Step 3.5 & Attachment 3, 8kr 3D23A-28 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because 38 EDG has lost control and field flashing power. Plausible because 3A EDG does not have black start capability.

8. Correct per the reference C. Incorrect because 38 EDG has lost control and field flashing power and because 3A EDG does not have black start capability.

Plausible because both Unit 4 EDGs have black start capability.

D. Incorrect because 38 EDG does not have black start capability.

Plausible because 38 EDG has lost control and field flashing power. 98 NRC DRAFT 07/24/07 0#50 Operators are performing the monthly test run of the 3A EDG using 3-0SP-023.1 , "Diesel Generator Operability Test." The BOP is attempting to verify 3A EDG is in LAG. He momentarily positions the diesel generator voltage regulator in RAISE. Assuming the 3A EDG is in the LAG, which ONE of the following describes the EDG response and subsequent operator actions required to verify the EDG is in LAG? 3A EDG amps: A. increased.

Slowly lower EDG voltage until amps stop decreasing and start to increase.

Then slowly raise EDG voltage until amps increase.

B. increased.

Slowly raise EDG voltage until amps stop increasing and start to decrease.

Then slowly lower EDG voltage until amps increase.

c. decreased.

Slowly lower EDG voltage until amps stop decreasing and start to increase.

Then slowly raise EDG voltage until amps decrease.

D. decreased.

Slowly raise EDG voltage until amps stop decreasing and start to increase.

Then slowly lower EDG voltage until amps decrease.

99 NRC DRAFT 07124/07 Q#50 ANSWER: A KA: 064G2.1.23 As it relates to the EDG, ability to perform specific system and integrated plant procedures during all modes of plant operation.

3.9/4.0 10CFR55: 41.bA,41.b.7

Reference:

3-0SP-023.1, Step 7.1.2.29.k Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference B. Incorrect because the operator should subsequently slowly lower EDG voltage until amps stop decreasing.

Plausible because 3A EDG amps increased.

C. Incorrect because 3A EDG amps increased.

Plausible because the operator should slowly lower EDG voltage until amps stops decreasing.

D. Incorrect because 3A EDG amps increased and because the operator should slowly lower EDG voltage until amps stop decreasing.

Plausible because if the EDG is in the LEAD, the subsequent operator action would be to subsequently raise the EDG voltage. 100 NRC DRAFT 07/24/07 Q #51 Unit 3 is at 100% power with all systems in normal alignment wherJ the following occur;

• Annunciator H 1/6, PRMS CHANNEL FAILURE, alarms.

• Steam Generator Liquid Sample Monitor, R3-19, fail lamp on the PRMS drawer is illuminated

• R3-19 display and recorder reading are failed LOW

• R3-19 warning and high alarm lamps are off. Which ONE of the following describes the plant response and the correct initial operator response?

A. S/G blowdown does NOT automatically isolate. Manually secure Blowdown.

B. S/G blowdown does NOT automatically isolate. Maintain blowdown in operation but re-align the discharge to the main condenser.

C. S/G blowdown has automatically isolated.

Verify S/G blowdown flow control valves, FCV-6278 A, B, & C and blowdown tank to canal level control valve, LCV-3-6265B are closed. D. S/G blowdown has automatically isolated.

Verify S/G blowdown isolation valves, MOV-6275 A, B, & C and S/G liquid sample valves MOV-1425/1426/1427 are closed. 101 NRC DRAFT 07/24/07 Q #51 ANSWER: A KA: 073A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the PRMS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

detector failure. 2.7/3.2 10CFR55: 41.b.10, 41.b.11, 41.b.12, 41.b.13

Reference:

3-0NOP-071.2, Step 3 RNO 3-ARP-097.CR, H1/6, Step 2f. Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the references B. Incorrect because blowdown needs to be manually secured. Plausible because blowdown is still in service and operator may think that with it aligned to the condenser (recent plant modification does not allow flow to condenser) continued operation is acceptable.

C. Incorrect because S/G blowdown has not automatically isolated.

Plausible because verifying the blowdown flow control valves are closed is an expected response if blowdown automatically isolates due to high radiation as stated on the meter. D. Incorrect because S/G blowdown has not automatically isolated.

Plausible because verifying the blowdown flow control valves are closed is an expected response if blowdown automatically isolates due to Phase A. 102 NRC DRAFT 07/24/07 Q#52 Unit 4 operators are responding to a loss of Intake Cooling Water (ICW). Which ONE of the following identifies plant conditions that will require operators to trip Unit 4 reactor and turbine? A. Exciter hot air temperature increases to: B. Turbine bearing temperature increases to: C. CCW temperature increases to: D. TPCW temperature increases to: 103 NRC DRAFT 07124/07 Q#52 ANSWER: C KA: 076A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Service Water controls including reactor and turbine building closed cooling water temperatures.

2.6/2.6 10CFR55: 41.bA, 41.b.5, 41.b.8

Reference:

4-0NOP-019 FO Page Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because reactor trip is not required until exciter air temperatures increase beyond 90°C. Plausible because exciter air temperatures should increase during this event. B. Incorrect because reactor trip is not required until turbine bearing temperatures increase beyond 180°F. Plausible because turbine bearing temperatures should increase during this event. C. Correct per the references.

D. Incorrect because removing load/reactor trip is not required until TPCW temperatures increase beyond 110°F. Plausible because TPCW temperatures should increase during this event. 104 NRC DRAFT 07124/07 0#53 Unit 3 operators are responding to decreasing Instrument Air (IA) Pressure.

Which ONE of the following describes the use of Service Air as backup to the IA System? Service Air valves may be opened: A. regardless of IA pressure but only after all available instrument air compressors have been started. B. if IA pressure drops below 95 psig regardless of the number of available instrument air compressors started. c. if IA pressure drops below 95 psig and after all available instrument air compressors have been started. D. as backup to IA without restrictions associated with IA pressure or starting of available IA compressors.

105 NRC DRAFT 07/24/07 Q#53 ANSWER: C KA: 078K1.02 Knowledge of the physical connection and/or cause-effect relationships between the Instrument air and the service air. 2.7/2.8 10CFR55: 41.b.4,41.b.10

Reference:

O-ONOP-013, Steps 3 and 4 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because service air valves may be opened only if IA pressure drops below 95 psig and after any available instrument air compressors have been started. Plausible because service air valves may be opened after any available instrument air compressors have been started B. Incorrect because service air valves may be opened only if IA pressure drops below 95 psig and after any available instrument air compressors have been started. Plausible because service air valves may be opened if IA pressure drops below 95 psig. C. Correct per the references.

D. Incorrect because service air valves may be opened only if IA pressure drops below 95 psig and after any available instrument air compressors have been started. Plausible because service air valves are opened as backup to instrument air. 106 NRC DRAFT 07/24/07 0#54 The plant was operating at 100% power when:

• A small break LOCA has occurred.

• SI was manually actuated due to decreasing pressurizer level.

• Phase A was NOT manually actuated following the manual SI actuation.

• Containment pressure has risen to 17 psig.

• The RCS is saturated.

• Two ECCs and 3 ECFs are operating.

While performing EOP-E-O, containment isolation signals were reset as directed by Attachment

3. All Attachment 3 actions were completed before transitioning from EOP-E-O. After the transition to EOP-E-1, the LOCA break size increased and containment pressure increased to 24 psig. Based upon the above conditions which ONE of the following describes the status of the containment isolation signals? A. Only Phase A containment isolation has occurred and has been reset. B. Only Phase B containment isolation has occurred and has been reset. C. Both Phase A and Phase B containment isolations have automatically actuated but neither Phase A or Phase B containment isolations have been reset. D. Both Phase A and Phase B containment isolations have automatically actuated and only Phase A has been reset. 107 NRC DRAFT 07/24/07 Q#54 ANSWER: D KA: 103K4.06 Knowledge of the Ctmt design feature(s) and/or interlock(s) which provide for containment isolation system 3.1/3.7 10CFR55: 41.b.7,41.b.9

Reference:

5610-T-L 1 Sheet 11 3-EOP-E-0, Attachment 3, Step 12 Cog Level: 3 Analysis Level 3 because the operator must recognize that a manual SI does not actuate Phase A. He then must analyze the plant condition to determine that an automatic SI has occurred and that this will actuate a Phase A. He then must remember that the E-O attachment 3 resets both SI and Phase A & B if actuated (Phase B was not at the time of attachment 3). If containment pressure increases above 20 psig, a Phase B isolation signal will still be generated even with the Phase A reset. New Question Response Analysis:

A. Incorrect because the pressure increasing above 20 psig will cause a Phase B even after the SI and other actuation signals were reset in E-O. Plausible because the Phase A signal was reset in attachment

3. B. Incorrect because the plant conditions would result in an automatic SI signal even after the manual SI was actuated.

This would result in a Phase A signal as well. Plausible because the attch 3 does reset Phase B if it had actuated at that time. C. Incorrect because both phase A and B have actuated.

Plausible because the operator could assume that another signal above 20 psig would actuate Phase A again therefore it would not be reset. D. Correct per the reference 108 NRC DRAFT 07/24/07 0#55 A S/G tube rupture occurred on Unit 3. All safety systems functioned as designed including Phase A Containment Isolation.

Which ONE of the following describes an effect of Phase A on the containment and actions operators will take to respond to the Phase A signal? A. Instrument Air to Containment has been isolated.

Operators will reset Phase A when directed by 3-EOP-E-O, Attachment 3, "Prompt Action Verifications".

B. CVCS letdown from containment has been isolated.

Operators will reset Phase A when directed by 3-EOP-E-O, Attachment 3, "Prompt Action Verifications".

C. CCW valves to the RCPs in containment have been closed. Operators will reset Phase A after initiating a controlled cooldown while performing 3-EOP-E-3.

D. Normal Containment Coolers have been tripped. Operators will reset Phase A after initiating a controlled cooldown while performing 3-EOP-E-3.

109 NRC DRAFT 07/24/07 Q#55 ANSWER: B KA: 103A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the Ctmt; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Phase A & B Isolation.

3.5/3.8 10CFR55: 41.b.7

Reference:

3-EOP-E-O, Attachment 3, Step 13 3-EOP-E-3, Steps 13 and 19 5613-M-3047,Sheet1 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because Instrument Air has not been isolated.

Plausible because operators will reset Phase A when directed by 3-EOP-E-O, Attachment

3. B. Correct per the references.

Letdown isolates and operators will reset Phase A when directed by 3-EOP-E-O, Attachment 3 C. Incorrect because CCW valves to the RCPs in containment have not been closed and operators will reset Phase A when directed by 3-EOP-E-O, Attachment

3. Plausible because CCW valves are closed upon Phase Band E-3 does contain a step to reset Phase A. D. Incorrect because operators will reset Phase A when directed by 3-EOP-E-O, Attachment

3. Plausible because Normal Containment Coolers have been tripped 110 NRC DRAFT 07/24/07 0#56 Unit 4 is at 100% power at the beginning of core life with:

• Control Rods are at 0-230.

• Rod control is in automatic Control Rod H-8 drops into the core:

• All automatic control systems responded as designed

• Tavg stabilizes rF below Tref. Which ONE of the following describes the effect of this event on the CVCS system after RCS parameters have stabilized?

A. Low Pressure Letdown Control valve, PCV-3-145, is throttled closed resulting in lower VCT level B. Low Pressure Letdown Control valve, PCV-3-145, is throttled open resulting in higher VCT level. C. The CVCS demineralizers have absorbed boron resulting in a positive reactivity insertion.

O. The CVCS demineralizers have released boron resulting in a negative reactivity insertion.

III NRC DRAFT 07/24/07 Q#56 ANSWER: C KA: 001 K3.01 Knowledge of the effect that a loss or malfunction of the CROS will have on the CVCS. 2.9/3.0 10CFR55: 41.b.1, 41.b.5

Reference:

3-0P-047, Step 4.10 SO 013, Page 10 Cog Level: 2 Comprehension Level 2 because the operator will have to analyze that the change in Tavg will result in lowering letdown temperatures.

Cooler water through the CVCS demineralizers will result in increased absorption of boron in the demineralizers, resulting in lower boron concentration in the VCT and a subsequent positive reactivity insertion.

New Question Response Analysis:

A. Incorrect because PCV-145 responds to letdown header pressure downstream of the letdown orifices.

While RCS pressure and letdown header pressure may temporarily respond to this RCS temperature change . the control systems will soon restore these pressures to their normal values. Plausible because if PCV-3-145 throttled closed, it would result in lower VCT level. B. Incorrect because PCV-145 responds to letdown header pressure downstream of the letdown orifices.

While RCS pressure and letdown header pressure may temporarily respond to this RCS temperature change the control systems will soon restore these pressures to their normal values. Plausible because if PCV-3-145 throttled open, it would result in higher VCT level. C. Correct per the reference O. Incorrect because the CVCS demineralizers will absorb more boron resulting in a positive reactivity insertion.

Plausible because if the CVCS demineralizers released boron it would result in resulting in a negative reactivity insertion.

112 NRC DRAFT 07/24/07 0#57 Unit 4 reactor is at 100% power when 4A RCP rotor seizes. The 4A RCP stops rotating instantly.

Which ONE of the following describes the initial effect of this event on the Reactor Coolant System? A. Pressurizer level and pressure will decrease.

DNBR will decrease and Clad Temperature Limits will NOT be exceeded.

Fuel rod failure will not occur. B. Pressurizer level and pressure will decrease.

The DNBR will decrease and Clad Temperature Limits will be exceeded.

Fuel rod failure may occur. C. Pressurizer level and pressure will increase.

DNBR will decrease and Clad Temperature Limits will be exceeded.

Fuel rod failure may occur. D. Pressurizer level and pressure will increase.

DNBR will decrease and Clad Temperature Limits will NOT be exceeded.

Fuel rod failure will NOT occur. 113 NRC DRAFT 07124/07 Q#57 ANSWER: C KA: 002K6.02 Knowledge of the effect of a loss or malfunction of the following RCS components:

RCP, 3.6/3.8 10CFR55: 41.b.2, 41.b.3, 41.b.14

Reference:

FSAR Section 14.1.9 Cog Level: 2 Comprehension Level 2 because the operator must relate the type of RCP failure to the effects on the RCS parameters.

He then must analyze how the changing parameters relate to DNBR and determine that even though RCS pressure is increasing, the reduction in core flow causes DNB to occur and then understand the affect of DNB on the fuel rods. New Question Response Analysis:

A. Incorrect because pressurizer level and pressure will increase.

Plausible because Total core cooling flow will decrease and the DNB Ratio will decrease and core damage will occur. B. Incorrect because pressurizer level and pressure will increase.

Total core cooling flow will decrease and core damage will occur. C. Correct per the references D. Incorrect because DNB limits will be exceeded and core damage will occur. 114 NRC DRAFT 07/24/07 Q#58 Operators are performing a startup on Unit 3 and power is 5% with all pressurizer controls in automatic when the controlling pressurizer level transmitter, L T 459A, fails LOW. Which ONE of the following describes the effect of this failure assuming NO operator action? Pressurizer level will continuously:

A. decrease.

Letdown isolation will occur pressurizer level reaches setpoint.

B. decrease.

The reactor will trip on low pressurizer pressure when setpoint is reached. c. increase.

The reactor will trip on high pressurizer level when setpoint is reached. D. increase.

Pressurizer PORV(s) will open when setpoint is reached. 115 NRC DRAFT 07/24/07 Q#58 ANSWER: 0 KA: 011A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRZ LCS controls including PRZ level and pressure.

3.5/3.6 10CFR55: 41.b.3, 41.b.5

Reference:

5610-T-D -15 Sheet 1 5610-T-D-16B Sheet 1 Cognitive Level: 3 Application

/ Analysis Level 3 because the operator must first recognize that the L T-459A signal feeds back to the control loop. When L T-459A indicates level is too low, the control loop will send a signal to charging pumps to speed up. Additionally the L T-459A failure causes letdown isolation.

This combination results in rising pressurizer level. If power were above P-7 (10%), the reactor would trip at 92% level. At 5% power, the 92% level trip is not enabled. Level will continue to rise, eventually covering the spray nozzles and taking the RCS water solid. Pressure will rise and PORVs will open at 2335 psig. New Question Response Analysis:

A. Incorrect because pressurizer level will continuously increase.

Plausible because pressurizer level would decrease if L T-459A failed high instead of low and letdown isolation does normally occur at 14%. B. Incorrect because pressurizer level will continuously increase.

Plausible because pressurizer level would decrease if L T-459A failed high instead of low and .the reactor would normally trip on low pressurizer pressure when the pressurizer empties. C. Incorrect because the reactor will not trip on high pressurizer level at this power level. Plausible because pressurizer level will continuously increase.

D. Correct per the references 116 NRC DRAFT 07124/07 0#59 Operators are performing a plant start up on Unit 3. Reactor power is 15%. Which ONE of the following describes the direct effect of depressing the "Source Range Less than P-6 Push to Reinstate" pushbutton?

The NIS source ranges will: A. reinstate causing a SR Hi Flux reactor trip. B. reinstate but a SR Hi Flux reactor trip will NOT occur. c. NOT reinstate and a SR Hi Flux reactor trip will NOT occur. D. NOT reinstate until power is reduced below 10% when the SRs will automatically trip the reactor. 117 NRC DRAFT 07/24/07 Q#59 ANSWER: C KA: 015A4.03 Ability to manually operate and/or monitor in the control room: trip bypasses.

3.8/3.9 10CFR55: 41.b.7

Reference:

5610-T -L 1 Sheet 16 Cog Level: 2 Comprehension Level 2 because the operator must recognize that when power is >10%, the P-10 permissive is enabled and sends a block signal to both the reinstatement of the source ranges and to the source range trip signal. Also, pushing the re-instate pushbuttons does not affect the future automatic operation of the detectors as there is no seal in for these pushbuttons.

New Question Response Analysis:

A. Incorrect because the NIS source ranges will not reinstate and reactor trip will not occur. Plausible because if power was below P-1 0, this is the action that would occur. B. Incorrect because the NIS source ranges will not reinstate.

Plausible because reactor trip will not occur. C. Correct per the references D. Incorrect because the source ranges will not automatically reinstate and trip the reactor when power drops below 10%. Plausible because the SRs will not reinstate immediately but will reinstate (without trip) when power subsequently drops below P-6. 118 NRC DRAFT 07/24/07 Q#60 Unit 3 core off-load is in progress when the refueling SRO inside containment reports a spent fuel element has dropped and a large amount of gas bubbles are originating from the damaged element. Which ONE of the following describes an expected plant response and the required operator response?

A. Only Containment Ventilation Isolation should automatically occur. Manually initiate Control Room Ventilation Isolation and verify the isolation alignments as directed by 3-0NOP-033.3, "Accidents Involving New or Spent FueL" B. Only Control Room Ventilation Isolation should automatically occur. Manually initiate Containment Ventilation Isolation and verify the isolation alignments as directed by 3-0NOP-033.3, "Accidents Involving New or Spent FueL" . C. Neither Containment nor Control Room Ventilation Isolation should automatically occur. Manually initiate both and verify the isolation alignments as directed by 3-ONOP-033.3, "Accidents Involving New or Spent FueL" D. Both Containment and Control Room Ventilation Isolation should automatically occur. Verify the isolation alignments as directed by 3-0NOP-033.3, "Accidents Involving New or Spent FueL" . 119 NRC DRAFT 07/24/07 Q#60 ANSWER: 0 KA: 034A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the fuel handling system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

dropped fuel element. 3.6/4.4 10CFR55: 41.b.9, 41.b.10, 41.b.11, 41.b.12

Reference:

5610-T-L 1 Sheet 11 3-0NOP-033.3, Steps 2.2, 4.3, 5.1.1.1 Cog Level: 2 Comprehension The SRO should recognize that bubbles from a dropped assembly are an indication the cladding has been breached.

Along with the nitrogen gas bubbles will be gaseous fission products so that when detected by the containment gas monitor, an automatic isolation signal will be generated.

He then must recognize that when moving fuel, containment integrity is set and that both the control room and the containment ventilation system isolate on high PRMS signals. New Question Response Analysis:

A. Incorrect because both containment and control room ventilation Isolation has automatically occurred.

Plausible because containment ventilation Isolation should automatically occur and control room ventilation isolation can be manually initiated.

B. Incorrect because both containment and control room ventilation Isolation has automatically occurred.

Plausible because control room ventilation Isolation should automatically occur and containment ventilation isolation can be manually initiated.

C. Incorrect because both containment and control room ventilation Isolation has automatically occurred.

Plausible because control room ventilation Isolation and containment ventilation isolation can be manually initiated.

D. Correct per the references 120 NRC DRAFT 07124/07 Q #61 Unit 4 is in Mode 3 with the MSIVs closed as operators commence a reactor start up. 4A main steam line safety valve, "RV-4-1400", fails open and the 4A S/G rapidly depressurizes.

Which ONE of the following describes the response of the safeguards system to this failure? Safety injection will: A. NOT actuate with the MSIVs closed. B. actuate when Tavg decreases to 543°F. c. actuate when the 4A S/G pressure decreases to 485 psig. D. actuate when the 4A S/G pressure decreases to 900 psig. 121 NRC DRAFT 07/24/07 Q #61 ANSWER: C KA: 035K1.14 Knowledge of the physical connection and/or cause-effect relationships between the S/G and the ESF. 3.9/4.1 10CFR55: 41.b.7,41.b.8

Reference:

5610-T-L 1 Sheet 19 561 0-T-O-18B Sheet 1 Cog Level: 3 Analysis Level 3 because the operator must recognize that when a S/G depressurizes, SI usually results due to high steam line However with the MSIVs closed the 100 psid differential pressure that normally triggers this SI signal can't be generated because the main steam lines would be depressurized.

That's why the main steam header pressure inputs have been nulled at 585 psig so that even with the MSIVs closed, when any S/G's pressure drops to 485 psig, the 100 psid SI actuation logic is made up. New Question Response Analysis:

A. Incorrect because SI will actuate when the 4A S/G pressure decreases to 485 psig. Plausible because with the MSIVs closed, the header differential pressure cannot reach the 100 psid normally needed to actuate SI. B. Incorrect because SI will actuate when the 4A S/G pressure decreases to 485 psig. Plausible because 543°F is an SI setpoint that when reached will make up part of SI logic. C. Correct per the references O. Incorrect because SI will actuate when the 4A S/G pressure decreases to 485 psig. Plausible because 900 psig would result in 100 psid at hot zero power that is an actuation signal if the MSIVs were open. 122 NRC DRAFT 07/24/07 0#62 3-EOP-E-3, "Steam Generator Tube Rupture," directs operators to use the Steam Dump to Condenser system to initiate a cooldown at maximum rate prior to RCS depressurization.

Which ONE of the following describes how the Reactor Operator will accomplish this cooldown using the Steam Dump to Condenser system? . Place the Mode Selector Switch in: A. AUTO. Bypass the low Tavg interlock.

Place the Hagan controller in MANUAL and use the UP arrow to open the steam dump valves. B. AUTO. Place the Hagan controller in AUTO and adjust the pot setting to the desired setpoint to open the steam dump valves. C. MANUAL. Bypass the low Tavg interlock.

Place the Hagan controller in MANUAL and use the UP arrow to open the steam dump valves. D. MANUAL. Place the Hagan controller in AUTO and adjust the pot setting to the desired setpoint to open the steam dump valves. 123 NRC DRAFT 07/24/07 Q#62 ANSWER: C KA: 041G2.1 .28 As it relates to the SO Sys: Knowledge of the purpose and function of major system components and controls.

3.2/3.3 10CFR55: 41.bA,41.b.7

Reference:

3-EOP-E-3, Step 19 SO-105 Page 17 Cog Level: 2 Comprehension Level 2 because the operator must understand all the operating modes of the Hagan controller and EOP-E-3, Step 19 does not provide specific guidance regarding how to use the SOTC system to achieve the desired results. As stated in the correct response, this is a several step process and requires the operator to determine how the controller responds in various modes of operation, this particular evolution requires the controller to be in both the manual mode and manual control of the hagan process output (manual-manual mode). The operator must also understand that this action would result in the SOTC valves closing if he did not also bypass the low Tavg interlock.

New Question Response Analysis:

A. Incorrect because the Mode Selector Switch will be placed in MANUAL. Plausible because the Hagan controller will be placed in MANUAL and the UP arrow will be used to open the steam dump valves. B. Incorrect because the Mode Selector Switch will be placed in MANUAL and the Hagan controller will be place in MANUAL. Plausible because placing the Hagan controller in AUTO and adjusting the pot setting would open the steam dump valves if the Mode Selector Switch were placed in MANUAL. C. Correct per the references O. Incorrect because the Hagan controller will be place in MANUAL. Plausible because the Mode Selector Switch will be placed in MANUAL 124 NRC DRAFT 07/24/07 0#63 Unit 3 is at 100% power when the steam jet air ejector common steam supply valve, 3-30-020, is inadvertently closed. Which ONE of the following describes the effect of closing valve 3-30-020?

Assume no other operator action occurs. A. Condenser vacuum will decrease.

The turbine will trip when condenser vacuum reaches 25" Hg. B. Condenser vacuum will decrease.

The turbine will trip when condenser vacuum reaches 20" Hg. C. Main generator megawatts will decrease.

The reactor will trip when the setpoint is reached. D. Main generator megawatts will decrease.

The reactor will trip when the turbine trips due to high vibration.

125 NRC DRAFT 07124/07 Q#63 ANSWER: B KA: 055K3.01 Knowledge of the effect that a loss or malfunction of the CARS will have on the main condenser.

2.5/2.7 10CFR55: 41.b.5,41.b.7

Reference:

5613-M-3014, Sheet 3 3-0NOP-014, Step 3.1, 5.5.2 Cog Level: 1 fundamental knowledge New Question Response Analysis:

A. Incorrect because the turbine will trip when condenser vacuum reaches 20" Hg. Plausible because condenser vacuum will decrease B. Correct per the references C. Incorrect because turbine efficiency will decrease if the operator assumes reactor power would have to increase to maintain the same output. D. Incorrect because there is no high vibration trip on the turbine. 126 NRC DRAFT 07/24/07 0#64 Unit 3 is at 75% power when the following occurs;

• annunciator D 5/3, SGFP A SUCTION LO PRESS, alarms

• annunciator D 7/4, LP HEATER BYPASS OPEN alarms

• CV-3-2011 indicates open

• a turbine runback is NOT in progress Which ONE of the following correctly describes the first operator response?

A. Close CV-3-2011 if SGFP suction pressure is greater than 220 psig. B. Reduce turbine load to restore SGFP suction pressure to greater than 220 psig. C. Start th'e standby Condensate Pump to restore SGFP suction pressure to greater than 220 psig. D. Stop the 3A SGFP to start the runback as it should have tripped when the SGFP A SUCTION LO PRESS annunciator actuated.

127 NRC DRAFT 07/24/07 Q#64 ANSWER: C KA: 056G2.4.50 As it relates to the condensate system: Ability to verify system alarm setpoints and operate cont.rols identified in the alarm response manual. 3.3/3.3 10CFR55: 41.b.4

Reference:

3-ARP-097.CR, D 7/4 Step 1 & 3.b. 3-ARP-097.CR, D 5/3 Step 2 & 3.a. Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because the ARP directs operators to start the standby Condensate Pump to restore SGFP suction pressure to > 220 psig. Plausible because closing CV-2011 would help to restore SGFP suction pressure.

B. Incorrect because the ARP directs operators to start the standby Condensate Pump to restore SGFP suction pressure to > 220 psig. Plausible because reducing turbine load would help to restore SGFP suction pressure.

C. Correct per the references D. Incorrect because the ARP directs operators to start the standby Condensate Pump to restore SGFP suction pressure to> 220 psig. Plausible because stopping the 3A SGFP would help to restore 3B SGFP suction pressure.

128 NRC DRAFT 07/24/07 0#65 Chemistry has reported that the concentration of oxygen in the in-service gas decay tank is 4.2% and the hydrogen concentration is 4.8%. Which ONE of the following describes the correct operator response?

Stop addition of waste gas to the gas decay tank and reduce the: A. oxygen concentration to < 4% as soon as possible.

B. hydrogen concentration to < 4% as soon as possible.

C. oxygen concentration to < 4% within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. D. hydrogen concentration to < 4% within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. 129 NRC DRAFT 07/24/07 Q#65 ANSWER: A KA: 071 K5.04 Knowledge of the operational implications of the following concept as it applies to the WGD system: relationship of H2/02 concentrations to flammability.

2.5/3.1 10CFR55: 41.b.10,13

Reference:

0-ONOP-061, NOTE prior to Step 5 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference.

B. Incorrect because ONOP-061 directs operators to reduce the oxygen concentration to < 4% as soon as possible.

Plausible because with concentrations of both gases> 4%, ONOP-061 directs operators to stop addition of waste gas to the gas decay tank. C. Incorrect because ONOP-061 directs operators to reduce the oxygen concentration to < 4% as soon as possible.

Plausible because with oxygen concentration>

2%, ONOP-061 directs operators to reduce the oxygen concentration within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. D. Incorrect because ONOP-061 directs operators to reduce the oxygen concentration to < 4% as soon as possible.

Plausible because with concentrations of both gases> 4%, ONOP-061 directs operators to stop addition of waste gas to the gas decay tank. 130 NRC DRAFT 07/24/07 0#66

• Unit 3 is at 90% power.

• Tavg = 630°F

• RCS pressure is 2235 psig Which ONE of the following describes the required operator response?

A. Place the unit in Mode 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. B. Reduce Tavg to less than 625°F within 15 minutes. C. Reduce power to less than 75% within 15 minutes D. Reduce RCS pressure to less than 2235 psig within 5 minutes. REFERENCE PROVIDED 131 NRC DRAFT 07/24/07 Q#66 ANSWER: A KA: G.2.1.11 Knowledge of less than one hour technical specification action statements for systems. 3.0/3.8 10CFR55: 41.b.5

Reference:

TS 2.1.1 TS Figure 2.1-1 Provide as reference:

Tech Spec Figure 2.1-1 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference.

B. Incorrect because the TS required response is to shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Plausible because reducing Tavg to less than 625°F would restore conditions to within the acceptable operating region and because there are other short term Tech Specs that require a 15 minute response.

C. Incorrect because the TS required response is to shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Plausible because reducing power to less than 75% would restore conditions to within the acceptable operating region and because there are other short term Tech Specs that require a 15 minute response.

D. Incorrect because the TS required response is to shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Plausible because reducing pressure within 5 minutes is a required action if the other PTN Safety Limit (press >2735 psig) is violated.

132 NRC DRAFT 07124/07 Q#67 Operators have evacuated the Control Room. The Unit 3 RO is performing Attachment 3 of 0-ONOP-1 05, "Control Room Evacuation" and has determined that RCS boration is required.

Which ONE of the following describes the correct orders the RO should give the SNPO to successfully borate the RCS? "Locally open the: A. Emergency Boration Valve, MOV-3-350.

After I start the 3C Charging pump, I will direct you to increase the 3C Charging pump speed controller setpoint to 6 psig." B. Emergency Boration Valve, MOV-3-350.

After I start the 3B Charging pump, I will direct you to increase the 3B Charging pump speed controller setpoint to 12 psig." C. Manual Boration Valve, 3-356. After I start the 3B Charging pump, I will direct you to increase the 3B Charging pump speed controller setpoint to 6 psig." D. Manual Boration Valve, 3-356. After I start the 3C Charging pump, I will direct you to increase the 3C Charging pump speed controller setpoint to 12 psig." 133 NRC DRAFT 07/24/07 Q#67 ANSWER: B KA: G.2.1.8 Conduct of Operations:

Ability to coordinate personnel activities outside the control room. 3.8/3.6 10CFR55: 41.b.7, 41.b.10

Reference:

0-ONOP-105, Attachment 3, Page 65, Step 23 Cognitive Level: 1 Recall New Question Response Analysis:

A. Incorrect because the 3C Charging pump is the wrong pump (it will have no power) and the SNPO should increase the Charging pump speed controller setpoint to 12 psig. Plausible because MOV-350 is the correct valve to open and 6 psig is the initial charging pump controller setting. B. Correct per the references C. Incorrect because valve 3-356 is the wrong valve and the SNPO should increase the Charging pump speed controller setpoint to 12 psig.. Plausible because 3-356 is a manual emergency boration valve used for boration under different circumstances and 6 psig is the initial charging pump controller setting and 3B Charging pump is the correct pump. D. Incorrect because valve 3-356 is the wrong valve and the 3C Charging pump is the wrong pump (it will have no power) Plausible because 3-356 is a manual emergency boration valve used for boration under different circumstances and because 12 psig is the correct charging pump controller setting. 134 NRC DRAFT 07/24/07 0#68 Which ONE of the following is an accurate comparison of the Unit 3 SFP radiological monitoring system to the Unit 4 SFP radiological monitoring system? A gaseous radioactive release from: A. both SFPs are monitored by their own separate Unit 3 and Unit 4 SFP SPING detectors located in each SFP vent duct. B. both SFPs are monitored by a common SFP SPING located in the SFP common vent duct. C. Unit 3 SFP is monitored by the plant vent monitor while Unit 4 SFP is monitored by a separate U4 SFP SPING. D. Unit 4 SFP is monitored by the plant vent monitor while the Unit 3 SFP is monitored by a separate U3 SFP SPING. 135 NRC DRAFT 07124/07 Q#68 ANSWER: D KA: G.2.2.4 Ability to explain the variations in control board layouts, systems, instrumentation and procedural actions between units at a facility.

2.8/3.0 10CFR55: 41.b.11

Reference:

SD-041 Page 20, SD-068 Page 33, 5614-M-3034, Sheet 1, 5610-M-3060 Sheet 1, 5613-M-3034 Sheet 1, Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because a gaseous radioactive release from both SFPs is monitored differently.

Plausible because only unit 4 is vented to the plant vent while unit 3 has a SPING detector in its vent duct.. B. Incorrect because a gaseous radioactive release from both SFPs is monitored differently.

Plausible because only unit 4 is vented to the plant vent while unit 3 has a SPING detector in its vent duct.. C. Incorrect because a gaseous radioactive release from both SFPs is monitored differently.

Plausible because only unit 4 is vented to the plant vent while unit 3 has a SPING detector in its vent duct.. D. Correct per the references 136 NRC DRAFT 07/24/07 0#69 Which ONE of the following temporary alterations would require a "prior PNSC approved Temporary System Alteration (TSA),,? A. lifting of a wire to perform a troubleshooting procedure on a circuit associated with turbine runback logic. B. addition of a jumper in the start circuit of the 3A HHSI Pump which will remain OOS on an ECO while the jumper is in place. C. addition of a temporary power supply for a Phase A Containment Isolation actuation circuit until a permanent power supply can be obtained.

D. installation of drain rigs on the pressurizer in preparation for fill and vent following refueling.

137 NRC DRAFT 07124/07 Q#69 ANSWER: C KA: G.2.2.11 Knowledge of the process for controlling temporary changes. 2.5/3.4 10CFR55: 41.b.10

Reference:

ADM-503, Step 5.1.4, Step 5.6.1.1 ADM-503, Att. 1, Page 1 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because lifting of a wire to perform troubleshooting on a circuit is specifically excluded from requiring a TSA by ADM-503 (Ref. Step 4.1.2.5). Plausible because lifting a wire is a temporary alteration.

B. Incorrect because addition of a jumper in a circuit of a component which will remain OOS on an ECO while the jumper is in place is specifically excluded from requiring a TSA by ADM-503 (Ref. Step 4.1.5.9).

Plausible because adding a jumper to a circuit is a temporary alteration.

C. Correct per the references.

D. Incorrect because installation of drain rigs connected to floor drains is specifically excluded from requiring a TSA by ADM-503 (Ref. Step 4.1.2.1 ). Plausible because installing a drain rig is a temporary alteration.

138 NRC DRAFT Q#70 Refueling core off-load is being conducted on Unit 3. The following off-load data has been recorded:

0800 0900 1000 1100 1200 1300 1400 1500 Cumulative

1. of fuel elements off-loaded o 3 9 15 22 30 37 47 07/24/07 SFP temperature 138°F 138°F 139°F 139°F 141°F 142°F 14rF 151°F Which ONE of the following identifies the first time at which the core off-load should have been stopped as required by 3-0P-040.2, "Refueling Core Shuffle"?

A. 1200 B. 1300 C. 1400 D. 1500 139 NRC DRAFT 07124/07 Q#70 ANSWER: A KA: G.2.2.28 Knowledge of new and spent fuel movement procedures.

2.6/3.5 10CFR55: 41.b.1, 41.b.11, 41.b.13

Reference:

3-0P-040.2, Steps 4.5 and 4.6 Cog Level: 1 recall New Question Response Analysis:

A. Correct because the SFP temperature increased to 141°F at 1200. This exceeds the OP-040.2 limit of 140°F. B. Incorrect because the off-load should have been stopped at 1200. Plausible because the cumulative off-load limit and the instantaneous off-load limits have been reached at 1300. C. Incorrect because the off-load should have been stopped at 1200. Plausible because the cumulative off-load limit has been exceeded at 1400. D. Incorrect because the off-load should have been stopped at 1200. Plausible because the instantaneous off-load limit has been exceeded at 1500. 140 NRC DRAFT 07/24/07 Q #71 A plant worker is assigned to work in a radiation field under the following conditions:

• The job is located 2 meters from a 1" valve that is reading 60 mr/hr at 1 meter distance.

• Additionally, the job location general area dose rate from other radiation sources is equal to the 10CFR20 minimum dose rate that defines a Radiation Area.

• The operator's cumulative documented dose for the year is 940 mrem.

• An extension of the exposure guidelines has NOT been granted. Determine the number of hours he may work without receiving a dose extension?

A. 1 hr B. 2 hrs C. 3 hrs D. 4 hrs 141 NRC DRAFT 07124/07 Q#71 ANSWER: C KA: G.2.3.1 Knowledge of 10CFR20 and related facility radiation control requirements.

2.6/3.0 10CFR55: 41.b.12

Reference:

0-ADM-600, Section 5.7.1.4 1 OCFR20 -Radiation Area Definition Cog Level: 3 Analysis Level 3 because the operator must recall the administrative limits and then calculate the remaining stay time by determining the total dose rate. The operator must recognize to use the inverse square law to calculate that the valve reading 60 mr/hr at 1 meter will cause a 15 mr/hr dose rate at the 2 meter work location.

However, the operator must recognize that the general area dose (10CFR20 dose rate = 5 mr/hr minimum for Rad Area) needs to be added as well as the worker is also exposed to this. This will allow 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of stay time before exceeding the 1000 mr guideline.

Administrative Limit is 1000 mr/yr without an extension.

Operator has 60 mr remaining exposure.

New Question Response Analysis:

A. Incorrect-plausible if you improperly calculate the 60 mr/hr at 1 meter remains constant out to 2 meters (plane source calculation).

B. Incorrect-plausible if you do not account for the inverse square law and only reduce the dose by % (30 mr/hr) instead of % (a common mistake) C. Correct -the 15 mr/hr from the valve is added to the 5 mr/hr general area dose rate to have 20 mr/hr total resulting in 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of stay time. D. Incorrect because the dose from the valve at 2 meters is 15 mr/hr but the general area dose rate must be added to it. 142 NRC DRAFT 07/24/07 Q#72 Given the following conditions at a work site:

• Radiation level is 40 mrem/hr

• Radiation level with shielding is 10 mrem/hr

• Time for one worker to place shielding is 15 minutes

• Time to conduct the task with one worker is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

• Time to conduct the task with 2 workers is 20 minutes. Assumptions:

• If shielding is used, it is to be installed by one worker only.

• The worker placing shielding will be exposed to a dose rate of 40 mr/hr. Which ONE of the following would result in the lowest total whole body dose? A. Conduct the task with two workers with shielding.

B. Conduct the task with two workers without shielding C. Conduct the task with one worker with shielding.

D. Conduct the task with one worker without shielding.

143 NRC DRAFT 07/24/07 Q#72 ANSWER: A KA: G.2.3.2 Knowledge of facility ALARA program. 2.5/2.9 10CFR55: 41.b.9, 41.b.12, 41.b.13

Reference:

0-ADM-600, Step 5.1.1.2 Cognitive Level: 3 Application

/ Analysis Level 3 because the operator must recall the ALARA rule which calls for the total collective exposure to be minimized.

Then the operator must calculate the collective exposure for each option and choose the alternative with the lowest total collective exposure.

Bank Question Response Analysis:

A. correct because the total dose would be 10 mr to the worker placing the shielding plus 3.3 mr to each worker resulting in 16.6 total mr. B. incorrect because each worker receives 13.3 mr for a total dose of approximately 26.6 mr. C. incorrect because the worker would receive 20 mrem. (10 mrem while placing the shielding and 10 mr while performing the work). D. Incorrect because this would result in one worker receiving 40 mr. 144 NRC DRAFT 07/24/07 0#73 An operator has volunteered to be a member of an Emergency Response Team that will enter a high radiation area to rescue a person from a non-life threatening situation.

• The operator has already received 2 REM Total Effective Dose Equivalent (TEDE) this year.

• The dose projection for each team member for this rescue is 4 REM TEDE. Which ONE of the following is correct regarding this situation?

The operator may: A. NOT be part of the rescue team. If the operator participated, the additional 4 REM would exceed the annual 10CFR20 limit but would NOT exceed the allowable exposure limit of Enclosure 1 of O-EPIP-20111, "Re-Entry" .. B. NOT be part of the rescue team. If the operator participated, the additional 4 REM would exceed the annual 10CFR20 limit and would exceed the allowable exposure limit of Enclosure 1 of O-EPIP-20111, "Re-Entry".

C. be part of the rescue team. The additional 4 REM will NOT exceed the annual 10CFR20 limit and will NOT exceed the allowable exposure limit of Enclosure 1 of O-EPIP-20111, "Re-Entry" . D. be part of the rescue team. The additional 4 REM will exceed the annual 10CFR20 limit but will NOT exceed the allowable exposure limit of Enclosure 1 of O-EPIP-20111, "Re-Entry" . 145 NRC DRAFT 07/24/07 Q#73 ANSWER: D KA: G.2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

2.5/3.1 10CFR55: 41.b.12

Reference:

O-EPIP-20111, Step 5.1.1.1, 5.1.1.4, 5.1.1.8, Enclosure 1 10CFR20.1201 Cog Level: 1 Recall New Question Response Analysis:

A. Incorrect because the operator may be a member of the team in spite of violating the 10CFR20 limits. Plausible because the annual 10CFR20 limit will be exceeded and the EPIP-20111 limit of 10 REM will not be exceeded.

B. Incorrect because the operator may be a member of the team and the additional 4 REM would not violate the EPIP limit of 10 REM. Plausible because the annual 10CFR20 limit will be exceeded.

C. Incorrect because the 10CFR20 limits will be violated.

Plausible because the operator may be a member of the team and the EPIP-20111 limit of 10 REM will not be violated.

D. Correct because the operator may be a member of the team in spite of violating the 1 OCFR20 limits. The EPIP-20111 limit of 10 REM will not be violated.

146 NRC DRAFT 07/24/07 Q#74 Unit 4 tripped from 100% power. The following conditions are observed by the control room operators upon completion of 4-EOP-E-O:

• Steam Generator narrow range levels:

• Steam Generator pressures:

• 3A S/G AFW max. flow rate:

• 3B S/G AFW max. flow rate:

• 3C S/G AFW max. flow rate:

• Pressurizer level:

• RCS pressure:

• RCS Cold Leg temperatures:

off-scale low 800 psig 100 gpm 125 gpm 125 gpm off-scale low 600 psig 330°F Which ONE of the following identifies the correct procedure the operators should transition to when 4-EOP-E-0 is complete?

A. 4-EOP-E-1, "Loss of Reactor or Secondary Coolant" B. 4-EOP-E-2, "Faulted Steam Generator Isolation" C. 4-EOP-FR-P.1, "Response to Imminent Pressurized Thermal Shock Condition" D. 4-EOP-FR-H.1, "Response to Loss of Secondary Heat Sink" 147 NRC DRAFT 07/24107 Q#74 ANSWER: A KA: G.2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry level conditions for emergency and abnormal operating procedures.

4.0/4.3 10CFR55: 41.b.10

Reference:

4-EOP-E-0, Step 15 4-EOP-F-0, Enclosures 3 and 4 Cog Level: 2 Comprehension Level 2 because the operator must recognize that a LOCA exists and a secondary break does not. The Operator must then apply the criteria of EOP-F-O to determine that a red or orange path does not exist for heat sink or integrity.

New Question Response Analysis:

A. Correct per the references B. Incorrect because a LOCA exists and the correct transition is to EOP-E-1, not EOP-E-2. Plausible because S/G pressures are lower than expected for the post-trip condition and containment pressure is at a value that can be caused by a MSLB. c. Incorrect because a LOCA exists and the correct transition is to EOP-E-1, not EOP-FR-P.1.

Plausible because RCS cold leg temperatures are above the threshold value of 320°F that would force transition to FR-P.1. D. Incorrect because a LOCA exists and the correct transition is to EOP-E-1, not EOP-FR-H.1.

Plausible because S/G NR levels are off-scale low and AFW flow rates are barely above the threshold value of 345 psig that would force transition to FR-H.1. 148 NRC DRAFT 07/24/07 Q#75 Operators are performing 3-EOP-FR-C.2, "Response to Degraded Core Cooling" and are currently depressurizing all intact S/Gs. The STA updates the Shift Manager that presently the CSF status trees for Containment (due to high containment pressure) is Orange and Integrity is Red. All other CSF status trees are green or yellow. Which ONE of the following describes the correct operator response?

A. Immediately transition to 3-EOP-FR-Z.1, "Response to High Containment Pressure." B. Immediately transition to 3-EOP-FR-P.1, "Response to Imminent Pressurized Thermal Shock Condition." C. Continue in 3-EOP-FR-C.2 until complete.

Then transition to 3-EOP-FR-Z.1 , "Response to High Containment Pressure." D. Continue in 3-EOP-FR-C.2 until complete.

Then transition to 3-EOP-FR-P.1 , "Response to Imminent Pressurized Thermal Shock Condition." 149 NRC DRAFT 07/24/07 Q#75 ANSWER: D KA: G.2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal and emergency evolutions.

2.9/3.6 10CFR55: 41.b.10

Reference:

0-ADM-211, Step 5.10.6 3-EOP-FR-C.2 Caution prior to Step 13 Cognitive Level: 2 Comprehension Level 2 because the operator must recall the Caution in FR-C.2 that directs that C.2 be completed prior to transitioning to the Red Path on Integrity.

The caution states that you must complete FR-C.2 prior to transitioning to FR-P.1. This is contrary to the rules of usage and require the operator to analyze and evaluate that the completion of C2 is higher importance than the P1 transition for the given plant conditions.

Additionally this exception to the rule is identified in 0-ADM-211.

New Question Response Analysis:

A. Incorrect because the operators should continue in 3-EOP-FR-C.2 until complete.

Then transition to 3-EOP-FR-P.1.

Plausible because an Red path exists on Integrity.

B. Incorrect because the operators should continue in 3-EOP-FR-C.2 until complete.

Then transition to 3-EOP-FR-P.1.

Plausible because an Red exists associated with vessel integrity.

C. Incorrect because the operators should transition to 3-EOP-FR-P.1 following completion of 3-EOP-FR-C.2.

Plausible because an orange path exists due to containment high pressure.

D. Correct per the reference 150 NRC DRAFT 07124/07 0#76 Operators are performing 3-EOP-ES-1.2, "Post LOCA Cooldown and Depressurization," and are preparing to start an RCP. Primary system parameters are:

• PRZ Level:

• RVLMS Plenum Level:

• RVLMS Head Level: 73% Full NOT Full Which ONE of the following describes the effect on pressurizer level when an RCP is started under these conditions and the correct operator response?

Pressurizer level will: A. drop when an RCP is started. Transition to 3-EOP-FR-1.3, "Response to Voids in Reactor Vessel." B. NOT be affected when an RCP is started. Transition to 3-EOP-FR-1.3, "Response to Voids in Reactor Vessel." C. drop when an RCP is started. Continue to implement 3-EOP-ES-1.2.

D. NOT be affected when an RCP is started. Continue to implement 3-EOP-ES-1.2.

151 Draft Submittal (Pink Paper)

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NRC DRAFT 07/24/07 Q#76 ANSWER: C KA: 000009EA2.04 Able to determine and interpret the following as they apply to Small Break LOCA: PRZ Level. 3.8/4.0 10CFR55: 43.b.5

Reference:

3-EOP-ES-1.2, Step 25 and BD 3-EOP-F-0, Enclosure 6 3-EOP-FR-1.3, CAUTION prior to Step 1 Cog Level: 2 Analysis Level 2 because the operator must determine that the RVLMS and pressurizer level indications indicate a void in the vessel head and predict that starting an RCP will collapse the void, lowering the pressurizer level. The operator must recall that starting an RCP is desirable in these circumstances to provide forced cooling and normal pressure control during the cooldown.

Finally these conditions are to be expected and there is no need to transition to the identified yellow path procedures.

Additionally FR-1.3 should not be implemented when SI pumps are operating.

New Question Response Analysis:

A. Incorrect because FR-1.3 should not be implemented when SI pumps are operating. Plausible because pressurizer level will drop when an RCP is started as coolant flow to the head region collapses the void, lowering the pressurizer level. B. Incorrect because pressurizer level will drop when an RCP is started as coolant flow to the head region collapses the void, lowering the pressurizer level. Plausible because a transition to FR-I-3 would address the head void. Additionally the operator must recognize FR-1.3 should not be implemented when SI pumps are operating.

C. Correct per the reference D. Incorrect because pressurizer level will drop when an RCP is started as coolant flow to the head region collapses the void, lowering the pressurizer level. Plausible because the operators should continue to implement ES-1.2. 152 NRC DRAFT 07/24/07 0#77 3-EOP-ECA-2.1, "Uncontrolled Depressurization of All Steam Generators," directs operators to control feed flow to minimize RCS cooldown.

RCS cold reg temperatures have decreased from 54rF to 340°F in the past 60 minutes. Which ONE of the following describes the correct operator response?

A. Decrease feed flow to 345 gpm. Transition to 3-EOP-FR-P.1, "Response to Imminent Pressurized Thermal Shock Condition." B. Decrease feed flow to 25 gpm to each S/G. Transition to and complete 3-EOP-FR-H.1, "Response to Loss of Secondary Heat Sink." C. Continue to implement 3-EOP-ECA-2.1.

Decrease feed flow to 25 gpm to each S/G. D. Continue to implement 3-EOP-ECA-2.1.

Decrease total feed flow to 345 gpm. 153 NRC DRAFT 07/24/07 Q#77 ANSWER: C KA: W/E12EA2.2 Able to determine and interpret the following as they apply to Uncontrolled Depressurization of all S/Gs: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

3.4/3.9 10CFR55: 43.b.1 & b.5

Reference:

3-EOP-ECA-2.1, Step 3 3-EOP-F-0, Enclosure 4 3-EOP-FR-H.1, CAUTION prior to Step 1 Cog Level: 2 Comprehension Level 2 because the operator must recognize that with a cool down> 100°F/hour, ECA-2.1 will require feed flow to be reduced to 25 gpm per S/G. This reduction in feed flow will result in a red path on heat sink, but because the reduction of feed flow was procedurally driven, performance of FR-H.1 is not desirable.

New Question Response Analysis:

A. Incorrect because RCS temperatures have not dropped far enough to trigger an orange path on Integrity and feed flow should be reduced to 345 gpm .. Plausible because RCS temperatures have dropped 22rF in the past 60 minutes. B. Incorrect because transition to FR-H.1 is not required because feed flow has been reduced due to procedural requirements.

Plausible because operators will decrease feed flow to 25 gpm to each S/G. C. Correct per the reference D. Incorrect because feed flow needs to be reduced to 25 gpm to each S/G as directed by EOP-ECA-1.2.

Plausible because operators will continue to implement 3-EOP-ECA-2.

154 NRC DRAFT 07/24/07 0#78 Unit 4 is at 100% power with all systems in normal alignment.

• The 4A CCW pump is being removed from service.

• The 4A CCW pump CS has been placed in Pull-to-Lock but its breaker is still racked in. The switchyard subsequently de-energizes.

Which ONE of the following correctly describes Unit 4 sequencer(s) response and the first procedure operators will use to verify correct CCW system alignment?

A. 4A sequencer will close the 4C CCW pump breaker. 4B sequencer will close the 4B CCW pump breaker. 4-EOP-ES-O.1, "Reactor Trip Response." B. 4A sequencer will close the 4C CCW pump breaker. 4B sequencer will close the 4B CCW pump breaker. 4-0NOP-004, "Loss of Offsite Power." C. 4A sequencer will attempt to close the 4A CCW pump breaker. 4B sequencer will close the 4B CCW pump breaker. 4-EOP-ES-0.1, "Reactor Trip Response." D. 4A sequencer will attempt to close the 4A CCW pump breaker. 4B sequencer will close the 4C CCW pump breaker. 4-0NOP-004, "Loss of Offsite Power." 155 NRC DRAFT 07124/07 Q#78 ANSWER: C KA: 000056AA2.47

-Able to determine and interpret the following as they apply to Loss of Off-site Power: Proper operation of the EOG load sequencer.

3.8/3.9 10CFR55: 43.b.5

Reference:

4-EOP-ES-0.1, Step 17 4-0NOP-004, First NOTE prior to Step 1 5614-T-L 1, Sheet 12A & 128, 5610-T-L 1, Sheet 1240 5610-T -L 1, Sheet 152A Cog Level: 2 Comprehension Level 2 because the operator must recognize that if the 4A CCW pump is in PTL it cannot start. If the 4A CCW pump breaker has not been racked out, the 4A sequencer will still try to close its breaker in the event of a LOOP. 4C CCW pump can be started by either sequencer depending on the alignment of 40 bus. However if both 4A and 48 CCW pumps are racked in, 4C CCW pump will not get a start signal from a sequencer.

Operators will transition from EOP-E-O to EOP-ES-0.1 in response to the reactor trip caused by the LOOP. ES-0.1, Step 17 is the first procedural opportunity to verify proper CCW operation.

Note that it is checked a second time in ONOP-004 after operators complete ES-0.1. New Question Response Analysis:

A. Incorrect because 4A sequencer will attempt to close the 4A CCW pump breaker. Plausible because 48 sequencer will close the 48 CCW pump breaker and operators will verify the CCW alignment first using 4-EOP-ES-0.1, "Reactor Trip Response." 8. Incorrect because 4A sequencer will attempt to close the 4A CCW pump breaker and operators will verify the CCW alignment first using 4-EOP-ES-0.1, "Reactor Trip Response." Plausible because 48 sequencer will close the 48 CCW pump breaker C. Correct per the reference O. Incorrect because 48 sequencer will close the 48 CCW pump breaker and operators will verify the CCW alignment first using 4-EOP-ES-0.1, "Reactor Trip Response." Plausible because 4A sequencer will attempt to close the 4A CCW pump breaker. 156 NRC DRAFT 07/24/07 0#79 Unit 3 is at 75% power when the following occur;

• Automatic VCT makeup STARTS with VCT level at 50% on LT-112.

• 3B main feedwater flow control valve transfers to MANUAL mode.

• OSPDS channel "A" is DEENERGIZED

• Train 1 AFW flow control valves fail CLOSED After completion of any immediate operator actions and with the plant stable, the US should; A. Restore 3A EDG to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or initiate a plant shutdown to reach Mode 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> because 3A EDG auto-start capability has been lost. B. Restore 3P07 to service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or initiate a plant shutdown to reach Mode 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> because AFW actuation logic has been degraded.

C. Verify that all Train B emergency equipment is operable within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or initiate a plant shutdown to reach Mode 3 within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> because 3A EDG auto-start capability has been lost. D. Place 3A 4kv bus and 3A/3C 480V load center bus stripping logic in the tripped condition within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> because AFW actuation logic has been degraded.

157 Q#79 ANSWER: KA: B 000057G2.1.33 NRC DRAFT 07/24/07 As it relates to Loss of Vital AC Inst. Bus: Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications. ( 3.4 1 4.0 ) 10CFR55: 43.b.2

Reference:

3-0NOP-003.7 Enclosure 1, NOTE 1 Tech Specs. 3.0.3 -Page 3/40-1 Cog Level: 2 Comprehension Level 2 because the operator must analyze indications, determine that they are caused by a loss of 3P07 and relate that the most limiting impact results in a loss of the bus stripping relays. He must then relate that the loss of bus stripping results in a loss of an auto start signal to the AFW actuation as well as the EDG output breaker auto closure failure. He then must determine that the TS actions of 3.0.3 are required based on AFW availability.

New Question Response Analysis:

A. Incorrect, indications are for a loss of 3P07 therefore operators must take action within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore the Vital AC panel or initiate a plant shutdown to reach Mode 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> because AFW actuation logic has been degraded.

Plausible because 3A EDG auto-start capability has been degraded and 3A EDG tech specs are also applicable (but less restrictive

). B. Correct per the reference C. Incorrect, indications are for a loss of 3P07 therefore operators must take action within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore the Vital AC panel or initiate a plant shutdown to reach Mode 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> because AFW actuation logic has been degraded.

Plausible because all Train B emergency equipment needs to be verified operable within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. D. Incorrect, indications are for a loss of 3P07 therefore operators must take action within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore the Vital AC panel or initiate a plant shutdown to reach Mode 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> because AFW actuation logic has been degraded.

Plausible because operators will need to place 3A 4kv bus and 3A/3C 480V load center bus stripping logic in the tripped condition.

158 NRC DRAFT 07/24/07 0#80 Unit 4 is at 100% power with all systems in normal alignment when the 4A ICW pump breaker trips open on over-current.

The 4A ICW pump is declared inoperable and removed from service.

Which ONE of the following describes the Technical Specification limits associated with the 4A ICW pump being out of service? Unit 4 is in a: A. 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.

The action statement time can be extended to seven days if 40 4KV Bus is realigned to 4A 4KV Bus. B. 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.

The action statement time can NOT be extended by realigning 40 4KV Bus. C. seven day action statement.

The action statement time can be extended to thirty days if 40 4KV Bus is realigned to 4A 4KV Bus. O. seven day action statement.

The action statement time can NOT be extended by realigning 40 4KV Bus. 159 NRC DRAFT 07/24/07 Q#80 ANSWER: A KA: 000062G2.2.22 As it relates to Loss of Nuclear Svc Water: Knowledge of limiting conditions for operations

& safety limits. 3.4/4.1 10CFR55: 43.b.2

Reference:

Tech Spec 3.7.3 Actions a and b 561 0-T-E-1591, Sheet 1 Cog Level: 2 Comprehension Level 2 because the operator must recognize that the 4A ICW pump is powered from 4A 4KV Bus. If plant components are in normal alignment, 4B and 4C ICW pumps share the same train power supply resulting in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.

Swapping 40 Bus to 4 A Bus places 4C ICW pump on the "A" Train increasing the Action Time to 7 days. New Question Response Analysis:

A. Correct per the reference B. Incorrect because the action statement time can be extended to seven days if 40 4KV Bus is realigned to 4A 4KV Bus. Plausible because Unit 4 is in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.

C. Incorrect because Unit 4 is in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement.

Plausible because 7 days is the correct action statement time for a CCW pump and the action statement time can be extended to 30 days (for a CCW pump) if 40 4KV Bus is realigned to 4A 4KV Bus. o. Incorrect because Unit 4 is in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement and the action statement time can be extended if 40 4KV Bus is realigned to 4A 4KV Bus. Plausible because 7 days is the correct action statement time for a CCW pump. 160 NRC DRAFT 07/24/07 Q #81 Unit 3 is at 100% power when a loss of all feedwater event occurs. As the crew transitions from 3-EOP-E-0, the STA informs them a red path exists on Heat Sink. The operators successfully initiate bleed and feed. Which ONE of the following describes the reportability of this event? The event must be reported to the: A. NRCOC within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. B. NRC Resident within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. C. State of Florida within 15 minutes and to the NRC immediately following State notification.

D. State of Florida within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and to the NRC immediately following State notification.

REFERENCE PROVIDED 161 NRC DRAFT 07/24/07 Q #81 ANSWER: C KA: W /E05G2.4.30 As it relates to Loss of secondary Heat Sink: Knowledge of which events related to system operations/status should be reported to outside agencies.

2.2/3.6 10CFR55: 43.b.5

Reference:

0-EPIP-201 01, Enclosure 1, Category 5, Item C.3 0-EPIP-201 01, CAUTIONS before Step 5.6.1.11 Cog Level: 1 Recall New Question Provide EPIP-20101 Enclosure 1 as a reference Response Analysis:

A. Incorrect because the event must be reported to the State of Florida within 15 minutes and to the NRC immediately following State notification in accordance with 0-EPIP-201 01, "Duties of the Emergency Coordinator." Plausible because notification to the NRCOC lAW ADM-115 is required in the event of the occurrence of significant events as defined in 1 OCFR50. 72. B. Incorrect because the event must be reported to the State of Florida within 15 minutes and to the NRC immediately following State notification in accordance with 0-EPIP-20101, "Duties of the Emergency Coordinator." Plausible because notification to the NRC Resident lAW ADM-115 is required in the event of the occurrence of significant events. C. Correct per the reference D. Incorrect because the event must be reported to the State of Florida within 15 minutes and to the NRC immediately following State notification in accordance with 0-EPIP-201 01, "Duties of the Emergency Coordinator." Plausible because the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time frame applies to NRC notification and certain other notifications described in 0-ADM-115.

162 NRC DRAFT 07/24/07 0#82 Unit 3 is at 50% power with control Bank "0" at 180 steps as I&C investigates a power cabinet non-urgent failure alarm. A momentary power interruption in the 2BO power cabinet control power circuitry causes all 3 of the Bank "0" group 2 rods to drop 140 steps into the core where they are gripped and held normally by the stationary grippers being re-energized. , The US should direct the operators to; A. Verify rod control is in AUTOMATIC and reduce turbine load using 3-0NOP-100, "Fast Load Reduction", to lower Tref to within 3°F of Tavg. B. Verify rod control is in MANUAL and reduce turbine load using 3-0NOP-100, "Fast Load Reduction", until all Bank "0" rods are aligned within 12 steps C. Manually trip the reactor, verify the reactor is tripped, perform 3-EOP-E-0 and stabilize the plant using 3-EOP-ES-0.1.

O. Manually trip the reactor, verify the reactor is tripped and Emergency Borate per 3-0NOP-046.1.

163 NRC DRAFT 07/24/07 Q#82 ANSWER: C KA: 000003G.2.4.6 As it relates to the dropped control rod event: Knowledge of symptom based EOP mitigation strategies.

3.1/4.0 10CFR55: 43.b.5

Reference:

3-ONOP-028.3, Step 1 3-EOP-E-0, Step 1, 3-EOP-ES-0.1, Step 5 Cog Level: 2 Comprehension The SRO must recognize that even though the dropped rods did not fully insert a reactor trip is required.

He then must analyze that the control rods that were dropped initially would still be trippable as they were re-gripped by the stationary grippers and should be free to fall on the trip. He then must understand that the EOP network only addresses rods that do not fully insert on a trip by emergency borating using ONOP-46.1, dropped rods are not addressed unless they fail to insert. New Question Response Analysis:

A. Incorrect because the reactor would be manually tripped. Plausible because Tavg is restored to Tref by borating and reducing turbine load in ONOP-28.3 which is used for dropped rod recovery.

ONOP-028.3 requires the Rod Control Selector Switch to be placed in MANUAL. B. Incorrect because the reactor would be manually tripped. Plausible because Tavg is restored to Tref by borating and reducing turbine load in ONOP-28.3 which is used for dropped rod recovery.

C. Correct as the event described was multiple dropped rods that were relatched by the normal operation of the stationary grippers, they should still be trippable and fully insert when the manual trip is ordered. The EOPs only address rods that are stuck out of the core. D. Incorrect because the EOPs only address stuck rods and these rods should still insert on the reactor trip. Plausible because the step in ES-0.1 directs the operator to Emergency Borate using 3-0NOP-046.1 to compensate for any stuck rods. 164 NRC DRAFT 07/24/07 Q#83 Operators are increasing Unit 3 power to 100%.

• With power at 95% power, annunciator B 9/3, SHUTDOWN ROD OFF TOP / DEVIATION, alarms.

• The RO reports that control rod H-8 (0 Bank) has stopped moving and is now 14 steps lower than other 0 Bank control rods.

• Subsequent investigation reveals H-8 isstuck and untrippable.

Which ONE of the following describes the Tech Spec required operator response and the basis for the action? A. Place Unit 3 in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to ensure adequate Shutdown Margin requirements are met. B. Place Unit 3 in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to reduce the effect of the stuck rod on the subsequent Xenon Oscillation.

c. Declare H-8 inoperable and reduce power to < 75% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, confirm all Bank "0" rod positions are within 2:. 12 steps of each other to ensure adequate Shutdown Margin requirements are met. D. Declare H-8 inoperable and reduce power to < 75% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, confirm all Bank "0" rod positions are within 2:. 18 steps of each other to reduce the affect of the stuck rod on the subsequent Xenon Oscillation.

165 NRC DRAFT 07/24/07 Q#83 ANSWER: A KA: 000005G.2.2.22 As it relates to the Inoperable/Stuck Control Rod: knowledge of limiting conditions for operations and safety limits. 3.4/4.1 10CFR55: 43.b.2

Reference:

3-0NOP-028, Step 5.1.4.2 Tech Spec 3.1.3.1 Action a. 0-ADM-536, Section %.1.3 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference.

B. Incorrect because Xenon Oscillations are not the basis for the requirement to be in MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Plausible because the reactor must be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Misaligned rods could be an initiating event for Xe oscillations which do challenge the power distribution limits. C. Incorrect because the unit must be placed in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Plausible because the rod must be declared inoperable and reducing power to <75% would be a correct response if the rod was trippable.

Reducing power using boration would increase shutdown margin. D. Incorrect because the unit must be placed in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Xenon oscillations are not the basis. Plausible because reducing power to <75% and confirming all rods in the bank are within 18 steps would reduce peaking factors and would be a correct response if the rod were not untrippable.

166 NRC DRAFT 07/24/07 0#84 Unit 4 is at 80% power during end of life (EOL) power coast down operation.

The auto function of the master charging pump controller fails causing Charging pump speed to go to minimum.

• Operators respond and take manual control of Charging pump speed.

• Tavg is being allowed to "droop" to maintain the reactor critical and is presently 3°F BELOW Tref as directed by Reactor Engineering.

Which ONE of the following identifies the pressurizer level that should be maintained and the basis for maintaining that level. Pressurizer level should be: A. 44%. Pressurizer level program is determined by reactor power to ensure that the steam volume in the pressurizer is smaller at higher power to prevent a high level reactor trip following a 100% load rejection.

B. 44%. Pressurizer level program is determined by Tavg to ensure that the water volume in the pressurizer is higher at higher temperatures to prevent the pressurizer heaters from uncovering during a 10% step increase in turbine load. C. 47%. Pressurizer level program is determined by reactor power to ensure that the steam volume in the pressurizer is high enough at any power level to prevent water relief through the safeties on a 50% load rejection.

D. 47%. Pressurizer level program is determined by Tavg to ensure that the water volume in the pressurizer at any power level is high enough to prevent requiring a safety injection following a reactor trip. 167 NRC DRAFT 07/24/07 Q#84 ANSWER: B KA: 000028AA2.08 Able to determine and interpret the following as they apply to Pressurizer Level Malfunction:

PRZ level as a function of power level. 3.1/3.5 10CFR55: 43.b.5

Reference:

4-0NOP-041.6, Step 4.2 & Enclosure 1, SO-009 pages 8&9 Cog Level: 3 Application

/ Analysis Level 3 because the SRO must recognize that the automatic pressurizer level control system would have been maintaining level lower based on the lower than normal Tavg for the present power level. He must calculate the programmed pressurizer level for the given plant conditions and relate that Tavg is being maintained at 3°F below Tref and that prz level reference signal is generated from Tavg and not from power. Tref at 80% power would normally be 568.6°F which would demand 47% pressurizer level but since Tavg is 3°F low, the required pressurizer level would be 44%. Calculation:

22% to 53% level. 80% X 31 % program level span = 25%. 25% + 22% = 47% for 80% power but Tavg is not on program so the pressurizer level must be maintained at 44% to equate to the 565.6°F Tavg. New Question Response Analysis:

A. Incorrect because prz level is a function of Tavg and a high level reactor trip will be generated from a complete load rejection.

Plausible because the level is based on Tavg and is correct for the current Tavg. B. Correct. C. Incorrect because pressurizer level is a function of Tavg and a high steam volume would be a cause of the heaters uncovering.

A high level reactor trip will be generated before water relief through the safety valves. Plausible because the level is correct for the current reactor power and water will not be released through the safety valves. O. Incorrect because pressurizer level should be 44%. Plausible because the level is correct for the current reactor power and the level is maintained high enough to not require a SI following a reactor trip. 168 NRC DRAFT 07/24/07 0#85 With reactor power at 80%, PRMS-R-15 alarms and count-rate increases rapidly and stabilizes at 5,000 cpm. Operators enter 3-0NOP-071.2, "Steam Generator Tube Leakage," and attempt to quantify the leakage using the R-15 Primary to Secondary Leak Rate Graph in the Plant Curve Book. Unit 3 Condenser air in-leakage is constant at 7 scfm. All of the S/G tube leakage is coming from the 3B S/G. Which ONE of the following describes Technical Specification implications.

The RCS primary to secondary leakage has: A. NOT exceeded the Tech. Spec. limit. B. NOT exceeded the Tech. Spec. limit. However, be in Mode 3 in:s; 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. C. exceeded the Tech. Spec. limit. Be in Mode 3 in :s; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. D. exceeded the Tech. Spec. limit. Be in Mode 3 in :s; 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. REFERENCE PROVIDED 169 Q#85 ANSWER: C NRC DRAFT 07/24/07 KA: 000037G2.1.33 As it relates to Steam Generator Tube Leak: Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

3.4/4.0 10CFR55: 43.b.2

Reference:

3-0NOP-071.2, Step 7, 9A TS 3.4.6.2.c PCB Section 5, Figure 15, Unit 3 R-15 Primary to Secondary Leak Cog Level: 3 Application 1 Analysis Level 3 because the operator must locate the correct point on the R-15 curve, read the value from the curve (which is less than the TS limit) and then relate the air leakage to the need to multiply the value by 7. Then the operator must compare the multiplied value to the TS limit of 500 gpd and determine the TS limit has been exceeded.

The operator must then recall the guidance of ONOP-071.2 to be in Mode 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> which is more restrictive than the TS limit of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. New Question Examiner note: ONOP 71.2 and TS have been frozen earlier revision.

5000 cpm = 122 gpd from graph. Multiply 122 X 7 scfm = 854 gpd which exceeds TS limit of 500 gpd from one S/G. Provide PCB Section 5, Figure 15 as a reference Response Analysis:

A. Incorrect because S/G tube leakage has exceeded the Tech. Spec. limit and the unit must be in Mode 3 in s; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Plausible because if the curve is read directly without multiplying by 7, the S/G tube leakage will not have exceeded the Tech. Spec. limit. B. Incorrect because S/G tube leakage has exceeded the Tech. Spec. limit and the unit must be in Mode 3 in s; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Plausible because if the curve is read directly without multiplying by 7, the S/G tube leakage will not have exceeded the Tech. Spec. limit and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is the time identified by Tech. Specs to go from Mode 1 to Mode 3. C. Correct per the reference D. Incorrect because the unit must be in Mode 3 in s; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Plausible because S/G tube leakage has exceeded the Tech. Spec. limit which allows 6 hrs to HSB. 170 NRC DRAFT 07124/07 Q#86 Unit 3 is at 100% power with the 3A Normal Containment Cooler (NCC) out of service on a clearance (ECO). The 3B NCC subsequently trips. Which ONE of the following describes the effect on RCP operation and the correct operator response?

A. RCP stator temperatures will start to rise. If stator temperatures exceed 248°F, trip the reactor, transition to 3-EOP-E-O, "Reactor Trip or Safety Injection," and stop all RCPs. B. RCP stator temperatures will start to rise. If stator temperatures exceed 248°F, contact the Electrical Maintenance Supervisor as directed by 3-0NOP-0-41.1, "Reactor Coolant Pump Normal," for authorization to continue RCP operations above 248°F. C. RCP pump bearing temperatures will start to rise. If bearing temperatures exceed 225°F, trip the reactor, transition to 3-EOP-E-0, "Reactor Trip or Safety Injection," and stop all RCPs. D. RCP pump bearing temperatures will start to rise. If bearing temperatures exceed 225°F, contact the Electrical Maintenance Supervisor as directed by 3-0NOP-0-41.1, "Reactor Coolant Pump Normal," for authorization to continue RCP operations above 225°F. 171 NRC DRAFT 07/24/07 Q#86 ANSWER: A KA: 003A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the RCP; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Problems associated with RCP motors, including faulty motors and current, and winding and bearing temperature problems.

2.7/3.1 10CFR55: 43.b.5

Reference:

3-0NOP-041.1, Step 44, FO page Step 1 & 5 SD-29, Page 10 and Fig. 1 Cog Level: 2 comprehension

-Level 2 because the SRO must recognize that NCCs provide cooling to the RCP motor compartments thereby cooling the stator and not the pump bearings.

He then must understand that two NCCs cannot maintain the containment ambient temperature constant at full power, and that an increase in containment temperature will cause an increase in the stator temperatures of the RCPs, he then must recall the threshold temperature to take action and the correct action to take, New Question Response Analysis:

A. Correct per the reference.

B. Incorrect because If stator temperatures exceed 248°F, operators should trip the reactor, transition to 3-EOP-E-0, "Reactor Trip or Safety Injection," and. stop all RCPs. Plausible because RCP stator temperatures will start to rise and 248°F is the temperature requiring operators to take action. Also the Electrical Supervisor may authorize operation in excess of limits for stator temperature if the authorization is given before starting the RCP. C. Incorrect because RCP pump bearing temperatures are dependent upon CCW conditions, not the running of NCCs. Plausible because if the pump bearing temperatures increase to > 225°F, the operators should trip the reactor, got to EOP-E-O and stop RCPs. D. Incorrect because RCP pump bearing temperatures are dependent upon CCW conditions, not the running of NCCs. Plausible because if the pump bearing temperatures increase, 225°F is the temperature requiring operators to take action. 172 NRC DRAFT 07/24/07 Q#87 Operators are performing 3-EOP-ECA-1.1, "Loss of Emergency Coolant Recirculation," in response to a LOCA and failure of both RHR pumps. As directed by 3-EOP-ECA-1.1, both Containment Spray pumps (CSPs) have been stopped. The STA subsequently reports an Orange path on Containment pressure.

Which ONE of the following describes the correct operator response and the basis for the action? A. Continue performing ECA-1.1. Maintain CSPs stopped in accordance with the guidance of ECA-1.1. Maximizes availability of the RWST inventory for injection flow. B. Continue performing ECA-1.1. Start CSPs in accordance with the guidance of FR-Z.1. Containment barrier protection is a higher priority than injection flow as long as RWST level is > 60,000 gallons. C. Transition to FR-Z.1, "Response to High Containment Pressure." Start CSPs in accordance with the guidance of FR-Z.1. Containment barrier protection is a higher priority than injection flow as long as RWST level is > 60,000 gallons. D. Transition to FR-Z.1, "Response to High Containment Pressure." Maintain CSPs stopped in accordance with the guidance of ECA-1.1. Maximizes availability of the RWST inventory for injection flow. 173 NRC DRAFT 07/24/07 Q#87 ANSWER: 0 KA: 006G2.4.6 As it relates to Emergency Core Cooling: Knowledge of symptom based EOP mitigation strategies.

3.1/4.0 10CFR55: 43.b.5

Reference:

3-EOP-ECA-1.1, Step 8 3-EOP-FR-Z.1, CAUTION Prior to Step 8 Cog Level: 3 Analysis Level 3 because the situation will require the SRO to analyze competing priorities for the remaining RWST injection water and determine that even though a higher priority procedure (FR-Z-1) would start the CSPs the EOP mitigating strategy to follow is to conserve the RWST inventory to protect the core. The SRO must understand that a transition to FR-Z.1 is required but that a caution in FR-Z.1 states that the CSPs should be operated iaw the requirements of ECA-1.1. The SRO must relate the plant conditions and determine that maintaining the reduced injection flow established in ECA-1.1 is more of a priority than running the CSPs under these conditions even though an Orange path on the containment exists. New Question Response Analysis:

A. Incorrect because the Orange path on containment integrity must be addressed but the EOP mitigating strategy will not run a CSPs if recirculation is not available in order to maintain injection flow for as long as possible.

B. Incorrect because the Orange path on containment integrity must be addressed but the CSPs will not be run if recirculation is not available in order to maintain injection flow for as long as possible.

C. Incorrect because the Orange path on containment integrity must be addressed but the CSPs will not be run if recirculation is not available in order to maintain injection flow for as long as possible.

Plausible because transition will be made to FR-Z.1 which will direct that the CSPs to be started unless the RWST is needed for injection flow. D. Correct per the reference 174 NRC DRAFT 07/24/07 Q#88 Operators are responding to a reduced Intake Cooling Water (ICW) flow due to Intake Cooling Water System leakage. Which ONE of the following describes the effect (if any) on the Normal Containment Coolers (NCCs) ability to remove heat from the Containment, the consequences of that effect, and the correct operator response?

The NCCs ability to remove heat is: A. degraded.

Containment temperatures will INCREASE toward the TS limit of 125°F and equipment environmental qualifications WILL be challenged.

Initiate Containment Purge in accordance with O-OP-053, "Containment Purge System." B. degraded.

Containment temperatures will INCREASE toward the TS limit of 125°F and equipment environmental qualifications WILL be challenged.

Isolate the affected portion of the ICW system and establish at least one operable ICW header. C. unaffected.

Containment temperatures will NOT CHANGE and equipment environmental qualifications will NOT be challenged.

Initiate Containment Purge in accordance with O-OP-053, "Containment Purge System." D. improved.

Containment temperatures will DECREASE and equipment environmental qualifications will NOT be challenged.

Isolate the affected portion of the ICW system and establish at least one operable ICW header. 175 NRC DRAFT 07/24/07 Q#88 ANSWER: B KA: 022A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of service water. 2.9/3.2 10CFR55: 43.b.5

Reference:

3-0NOP-019, Step 5 RNO, SD-029, Page 8, SD-040, Page 7 Tech Specs 3.6.1.5 & Basis Cog Level: 2 comprehension Level 2 because the operator must recognize that the ICW system provides heat removal for the CCW system which provides heat removal for containment via the Normal Containment coolers and the Control Rod Drive Mechanism coolers. The actions of several ONOPs will reduce heat loads on the system but ultimately the CCW temperatures will rise causing the containment temperature to rise. New Question Response Analysis:

A. Incorrect because there is no procedural direction to initiate a Containment Purge. Plausible because the NCCs ability to remove heat is degraded and Containment temperatures will increase toward the TS limit of 125°F and equipment environmental qualifications will be challenged and because initiating a containment purge would reduce containment temperatures.

B. Correct per the reference C. Incorrect because the NCCs ability to remove heat is degraded.

Containment temperatures will increase and eqUipment environmental qualifications will be challenged and because there is no procedural direction to initiate a Containment Purge. Plausible because initiating a containment purge would reduce containment temperatures D. Incorrect because the NCCs ability to remove heat is degraded.

Containment temperatures will increase and equipment environmental qualifications will be challenged.

Plausible because operators will Isolate the affected portion of the ICW system and establish at least one operable ICW header. 176 NRC DRAFT 07124/07 Q#89 Both units are at 100% power with all systems in normal alignment.

• 0-OSP-075.11, "Auxiliary Feedwater Inservice Test," is being performed to test the "A" AFW pump.

• The Shift Manager is notified that the "A" AFW pump developed 300 gpm at a maximum obtainable speed of 5800 RPM. Which ONE of the following describes the correct operator response and the Tech. Spec. implications of this event? Declare the "A" AFW pump: A. operable and return it to service. Align "A" AFW pump to Train 1 to restore full compliance with the TS LCO. B. inoperable but available and return it to service. Align "A" AFW pump to Train 2 to restore full compliance with the TS LCO. C. inoperable and remove it from service. Align "c" AFW pump to Train 1 to extend the AFW TS action time from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 30 days. D. inoperable and remove it from service. Align "B" AFW pump to Train 2 to extend the AFW TS action time from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 30 days. 177 NRC DRAFT 07/24/07 Q#89 ANSWER: C KA: 061G2.2.22 As it relates to the AFW system: knowledge of limiting conditions for operations and safety limits. 3.4/4.1 10CFR55: 43.b.2

Reference:

3-0P-075, Section 7.1 0-OSP-075.11, Step 7.1.6.1 TS 3.7.1.2 Actions 1 and 3 Cog: Level: 2 comprehension Level 2 because the operator must recognize that 5800 RPM and 300 gpm do not meet the criteria for a SAT test. and the pump must be declared inoperable.

Then the operator must recognize that even though either "8" or "c" AFW pump can be aligned to Train 1, OP-075 directs that "c" AFW pump be aligned to Train 1 and finally it is this action that will restore 2 independent AFW trains and increase the Action Time to 30 days. New Question Response Analysis:

A. Incorrect because the "A" AFW pump failed to develop enough flow or RPMs and must be declared inoperable and taken OOS. Plausible because the "A" AFW pump was OOS during the Inservice Test and would normally be returned to service and aligned to Train 1 if the test had been SAT. 8. Incorrect because the "A" AFW pump failed to develop enough flow or RPMs and must be declared inoperable and taken OOS. Plausible because the "A" AFW pump nearly delivered the required flow and might be considered (erroneously) "available" for service. C. Correct per the reference D. Incorrect because operators need to align "c" AFW pump to Train 1 to extend the AFW TS action time from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 30 days. "8" AFW pump is already aligned to Train 2 but until "C" pump is aligned to Train 1, the TS Action Time is still 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Plausible because "8" AFW pump is normally aligned to Train 2 .. 178 NRC DRAFT 07124/07 0#90 Unit 4 is in Mode 1 with all systems operable, when Containment particulate and gas monitors, PRMS-11/12, Sample Inlet Temperature switch, TS-3640, fails HIGH. Which ONE of the following describes the Tech Spec implications of this failure? The TS-3640 failure: A. did NOT affect operability of R-11 or R-12. The applicable Tech Spec LCD is satisfied.

B. disabled R-11 only. Verify R-12 is operable and the associated Tech Spec LCD will be satisfied.

C. disabled R-12 only. Verify R-11 is operable and the associated Tech Spec LCD will be satisfied.

D. disabled both R-11 and R-12. Until operability of R-11 and R-12 is restored, comply with the applicable Tech Spec Action Statement.

179 NRC DRAFT 07/24/07 Q#90 ANSWER: 0 KA: 073G2.1.33 As it relates to Process Radiation Monitoring:

Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

3.4/4.0 10CFR55: 43.b.2

Reference:

SD-068, Page 27 and Figures 19 and 20 TS 3.4.6.1 Cog Level: 2 comprehension Level 2 because the operator must recognize that TS-3640 failing high triggers the high temperature alarm at 125°F and closes the inlet and outlet valves and stops the sample pumps. Because the sample flows through both R-11 and R12 in series, both R-11 and R-12 have lost sample flow. The TS Action Statement applies if both R-11 and R-12 are inoperable.

If either R-11 or R-12 are satisfied, the LCO is satisfied.

New Question Response Analysis:

A. Incorrect because the temperature switch failure disabled both R-11 and R-12. Plausible because if the switch failure had not disabled R-11 and R-12, the applicable Tech Spec LCO would be satisfied.

B. Incorrect because the temperature switch failure disabled both R-11 and R-12. Plausible because if R-12 remained operable, the applicable Tech Spec LCO would be satisfied.

C. Incorrect because the temperature switch failure disabled both R-11 and R-12. Plausible because if R-11 remained operable, the applicable Tech Spec LCO would be satisfied.

D. Correct per the reference 180 NRC DRAFT 07/24/07 Q#91 Unit 3 is at 100% power with all systems in normal alignment and control rods fully withdrawn.

• Annunciator F 4/6, RPIS POWER TROUBLE, alarms.

• The field operator reports that the 3001 supply breaker to the RPI inverter has tripped open. Predict the impact of this failure on the rod position indication (RPI) system and the actions required.

RPls will be powered from the GVT: A. automatically.

RPls indication may have changed. Initially implement 3-0NOP-028.1 , "RGG Misalignment." Then transition to 3-0NOP-028.2, "RGG Position Indication Malfunction." B. automatically.

RPls indication will NOT change. Initially Implement 3-0NOP-028.2, "RGG Position Indication Malfunction." G. after manual transfer.

RPls indication may have changed . Initially implement 3-0NOP-028.1, "RGG Misalignment." Then transition to 3-0NOP-028.2, "RGG Position Indication Malfunction." o. after manual transfer.

RPls indication will NOT change. Initially Implement 3-0NOP-028.2, "RGG Position Indication Malfunction." 181 NRC DRAFT 07/24/07 Q#91 ANSWER: A KA: 014A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

loss of power to the RPls. 3.1/3.6 10CFR55: 43.b.5

Reference:

SO-006 Page 11 and Figure 5, 3-0NOP-028.1, Step 1.1, 5.6 3-0NOP-028.2, Step 1.1, 5.3 3-ARP-097.CR, F 4/6 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference B. Incorrect because ONOP-028.2 is not a direct entry procedure and station OE has demonstrated that the output voltage of the CVT is different than the inverter resulting in all RPls indicating differently after transfer.

Operators must first perform actions of ONOP-028.1 and then transition to ONOP-028.2 when directed.

Plausible because The RPI power supply has transferred from the inverter to the CVT. C. Incorrect because the RPI power supply has auto-transferred from the inverter to the CVT. Plausible because the RPls will indicate lower than before the power interruption and operators will Initially implement 3-0NOP-028.1, "RCC Misalignment," then transition to 3-0NOP-028.2, "RCC Position Indication Malfunction." O. Incorrect because the RPI power supply has auto-transferred from the inverter to the CVT and because ONOP-028.2 is not a direct entry procedure.

Operators will Initially implement 3-0NOP-028.1, "RCC Misalignment," then transition to 3-0NOP-028.2, "RCC Position Indication Malfunction." Plausible because the primary procedure operators will use to respond to this event is ONOP-028.2 182 NRC DRAFT 07124/07 0#92 A required continuous fire watch has been established in the Auxiliary Building at the 3C MCC.

• ARMS Channel 15, Aux Bldg North N/S Corridor, alarms and its digital display increases to 80 mr/hr.

• The RO depresses the Ch.15 HIGH alarm PB on the ARMS control panel and notes that the display reads 5 mr/hr, the same value as the high alarm value shown In Attachment 1 (ARMS Setpoint List and Locations) of 0-ONOP-066, "High Area Radiation Monitoring System Alarm." Which ONE of the following describes the validity of this reading and the correct operator response?

The alarm is: A. valid. Evacuate persons in the affected area. Do NOT include the fire watch person in the evacuation.

B. valid. Evacuate persons in the affected area. Include the fire watch person in the evacuation.

C. NOT valid. Silence the alarm. Notify Radiation Protection to survey the area. D. NOT valid. Silence the alarm. Notify Radiation Protection to install a portable radiation monitor. 183 NRC DRAFT 07/24/07 Q#92 ANSWER: B KA: 072G2.1.14 As it relates to Area Radiation Monitoring:

Knowledge of system status criteria which require the notification of plant personnel.

2.5/3.3 10CFR55: 43.bA, 43.b.5

Reference:

O-ONOP-066, Step 2 and CAUTION prior to Step 3 Cog Level: 2 comprehension Level 2 because the operator must recognize that the alarm is valid and that the continuous fire watch location at the 3C MCC is adjacent to the ARMS Channel 15 location and that response to an ARMS alarm takes priority over a continuous fire watch. New Question Response Analysis:

A. Incorrect because fire watch personnel should be included in the evacuation.

Plausible because the alarm is valid and a local area evacuation is required.

B. Correct per the references.

C. Incorrect because the alarm is valid. Plausible because the alarm may be silenced if the alarm is determined to not be valid and because Radiation Protection will subsequently be called to survey the area. D. Incorrect because the alarm is valid. Plausible because the alarm may be silenced if the alarm is determined to not be valid and because Radiation Protection will subsequently be called to install a portable radiation monitor. 184 NRC DRAFT 07/24/07 Q#93 Operators are responding to a large break LOCA when the ST A reports the following QSPDS Core Exit Thermocouple readings:

Quad 1 Quad 2 Quad 3 Quad 4 CET Temp CET Temp CET Temp CET Temp R7 660 K5 1250 H8 415 K11 666 P8 590 K3 ??? F9 835 N15 680 N6 604 J2 2300 E8 ??? H13 602 N4 625 G6 675 B10 905 H9 615 M11 ??? G1 1560 B5 660 E14 622 M9 ??? F5 2300 E12 688 L8 670 F3 1780 Which ONE of the following describes the core condition and the direction the Shift Manager should give to the Unit Supervisor?

CETs J2 and F5: A. should be included in the core condition evaluation.

A Red Path exists. Transition to 3-EOP-FR-C.1, "Response to Inadequate Core Cooling." B. should be included in the core condition evaluation.

An Orange Path exists. Transition to 3-EOP-FR-C.2, "Response to Degraded Core Cooling." C. should NOT be included in the core condition evaluation.

A Red Path exists. Transition to 3-EOP-FR-C.1, "Response to Inadequate Core Cooling." D. should NOT be included in the core condition evaluation.

An Orange Path exists. Transition to 3-EOP-FR-C.2, "Response to Degraded Core Cooling." 185 NRC DRAFT 07/24/07 Q#93 ANSWER: A KA: 017 A2.02 Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Core Damage 3.6/4.1 10CFR55: 43.b.5

Reference:

3-EOP-F-0, Enclosure 2 and Basis Page 20 Cog Level: 3 Analysis Level 3 because the operator must analyze all of the core temperature data and then recall that the evaluation is based on the five highest CETs including the failed CETS and that threshold for Red Path is 1200°F (Orange Path = 700°F). Review of the QSPDS readings reveals that including the failed CETs reveals a Red Path Condition and FR-C.1 needs to be implemented.

This must be determined even though that if the failed CETs are not included, five CETs are greater than 700°F and the operator will conclude erroneously that only an Orange Path exists. New Question Response Analysis:

A. Correct per the references B. Incorrect because a Red Path exists and transition should be made to 3-EOP-FR-C.1, "Response to Inadequate Core Cooling." Plausible because the conditions for an Orange Path also exist and EOP-FR-C.2 is the correct procedure to transition for an Orange Path. C. Incorrect because the failed CETs should be included in the evaluation.

Plausible because a Red Path exists and operators should transition to 3-EOP-FR-C.1, "Response to Inadequate Core Cooling." D. Incorrect because the failed CETs should be included in the evaluation.

Plausible because the conditions for an Orange Path also exist and FR-C.2 is the correct procedure to transition for an Orange Path. 186 NRC DRAFT 07/24/07 0#94 With Unit 3 at 80% power, operators are responding to a Xenon oscillation.

The RO reports the following values of Axial Flux Difference (AFD):

• N-41 : +13

• N-42: +12

• N-43: +16

• N-44: +16 Which ONE of the following describes the condition of the reactor's AFD and the effect (if any) of allowing continued operation with these conditions?

The reactor AFD is: A. within the operational space. Operation at this power level may continue indefinitely.

B. within the operational space. With the positive AFD, Hot channel factors may be exceeded if a short term transient occurs. c. outside the operational space. Hot channel factors may be exceeded as a result of continued operation at this power level. D. outside the operational space. Hot channel factors will not be exceeded as a result of continued operation at this power level but may be exceeded if a short term transient occurs. REFERENCE PROVIDED 187 NRC DRAFT 07124/07 Q#94 ANSWER: C KA: G2.1.7 Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.

4.4 10CFR55: 43.b.5

Reference:

PCB Section 5, Figure 1A O-ADM-536, Attachment 1, Section 3/4.2 PCB Section 7, Figure 1 Cog Level: 2 comprehension Level 2 because the operator will have to plot the given AFD values on the reference provided.

Then the operator will have to recognize that 2 values are outside the operational space. 2 of the 4 values outside implies that AFD is out of the operational space. The operator will then have to relate that being outside of the operational space means Hot channel factors may be exceeded as a result of continued operation at this power level. New Question Provide PCB Section 5, Figure 1A and Section 7, Figure 1 as references Response Analysis:

A. Incorrect because the AFD is outside the operational space. Plausible because 2 of the four AFD values are within the operational space and if AFD was within the operational space, operation at this power level may continue indefinitely.

B. Incorrect because the AFD is outside the operational space. Plausible because 2 of the four AFD values are within the operational space and if AFD was within the operational space, hot channel factors would not be exceeded as a result of continued operation at this power level. C. Correct per the references.

D. Incorrect because hot channel factors may be exceeded as a result of continued operation at this power level. Plausible because the AFD is outside the operational space. 188 NRC DRAFT 07/24/07 0#95

• Unit 3 RCS temperature is 400°F.

• Unit 4 RCS temperature is 310°F. Which ONE of the following identifies the minimum required water inventory in the Condensate Storage Tank system and describes the basis for that required inventory?

A. 210,000 gallons. Ensures sufficient water to maintain Unit 3 at Hot Standby condition for a minimum of 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. B. 210,000 gallons. Ensures sufficient water to maintain Unit 4 at Hot Standby condition for a minimum of 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. C. 420,000 gallons. This is sufficient water to maintain both Units at these conditions for a minimum of 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and then cool down both reactor coolant systems to Mode 4. D. 420,000 gallons. This is sufficient water to maintain both units at these conditions for a minimum of 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. 189 NRC DRAFT Q#95 ANSWER: A KA , G2.1.32 Ability to explain and apply all system limits and precautions.

3.8 10CFR55: 43.b.2

Reference:

3-0P-018.1, Step 4.2 Tech Specs 3.7.1.3 0-ADM-536, Section 3/4.7.1.3 Cog Level: 2 Comprehension 07/24/07 Level 2 because the operator must evaluate the RCS temperatures provided and determine that Unit 3 is in Mode 3 and Unit 4 is in Mode 4. Only after this determination is made can the operator apply the correct Tech. Specs. New Question Response Analysis:

A. Correct per the reference B. Incorrect because there is sufficient water to maintain only Unit 3 at these conditions for a minimum of 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. Unit 4 would be on RHR at this temperature and its AFW system valved out. Plausible because 210,000 gallons is the required amount of water inventory in the CSTs. C. Incorrect because only 210,000 gallons is the required amount of water inventory in the CSTs and Unit 3 is the only unit that would need CST water because Unit 4 would be on RHR at this temperature and its AFW system valved out. Plausible because the numbers shown are the numbers provided by the Tech Spec Basis document.

D. Incorrect because only 210,000 gallons is the required amount of water inventory in the CSTs and Unit 3 is the only unit that would need CST water because Unit 4 would be on RHR at this temperature and its AFW system valved out. Plausible because the numbers shown are the numbers provided by the Tech Spec Basis document.

190 NRC DRAFT 07/24/07 Q#96 Both Units are at 100% power. The 3A EDG is out of service and a 14 day Tech Spec Action Statement is in effect for Unit 3. ECO requests have been submitted by Maintenance to remove the following equipment from service:

• 3A Containment Spray Pump

• 3A Emergency Containment Filter

• 3A Residual Heat Removal Pump

• Pressurizer Heater Control Group If taken out of service, which ONE of the following components from the list above would reduce the Unit 3 effective Tech Spec Action Time from the current 14 days to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />? A. 3A Containment Spray Pump B. 3A Emergency Containment Filter C. 3A Residual Heat Removal Pump D. Pressurizer Heater Control Group 191 NRC DRAFT 07124/07 Q#96 ANSWER: B KA: G2.2.24 Ability to analyze the effect of maintenance activities on LCO status. 3.8 10CFR55: 43.b.2

Reference:

TS 3.8.1.1, Action d.1 O-OSP-023.3, Step 6.1.2 and Attachment 2 Step 8 Cog Level: 2 comprehension Level 2 because the operator must recognize that the 3A ECF is a Train B safety component and then determine the impact of that piece of equipment being out of service during the maintenance activity.

The SRO must analyze that this reduced the Action time to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if the operators placed a Train B safety component OOS while the 3A EDG is OOS. New Question Response Analysis:

A. Incorrect because 3A Containment Spray Pump is Train A equipment and being out of service will not shorten the TS Action time from the 14 days associated with the EDG to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Plausible because taking the 3A CSP OOS wi" introduce an additional TS Action time to comply with but it will be 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, not 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. B. Correct per the reference C. Incorrect because 3A Residual Heat Removal Pump is Train A equipment and being out of service wi" not shorten the TS Action time from the 14 days associated with the EDG to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Plausible because taking the 3A RHRP OOS wi" introduce an additional TS ACtion time to comply with but it wi" be 7 days, not 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. D. Incorrect because the pressurizer control group heaters are Train A equipment and being out of service wi" not shorten the TS Action time from the 14 days associated with the EDG to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Plausible because while the control group heaters do not have a Tech Spec LCO, the BU Group "A" heaters which are also Train "A" equipment do and taking the BU Group "A" heaters OOS wi" introduce an additional TS Action time to comply with but it will be 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, not 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. 192 NRC DRAFT 07/24/07 Q#97 Which ONE of the following activities requires direct supervision by a licensed Senior Reactor Operator?

A. Underwater TV camera surveillance in the refueling cavity B. Reactor vessel irradiation specimen removal C. Lower internals removal D. Control rod unlatching 193 NRC DRAFT 07/24/07 Q#97 ANSWER: 0 KA: G2.2.27 Knowledge of the refueling process. 3.5 10CFR55: 43.b.6/b.7

Reference:

4-0P-038.1, Step 4.1 & Section 4.6 Cog Level: 1 recall Bank Question Response Analysis:

A. Incorrect because underwater TV camera surveillance in the refueling cavity is not considered a Core Alteration and as such does not require direct supervision by an SRO. Plausible because this activity typically occurs directly over the core while the upper internals are removed and irradiated fuel in place. B. Incorrect because reactor vessel irradiation specimen removal is not considered a Core Alteration and as such does not require direct supervision by an SRO. Plausible because this activity occurs directly over the reactor while the reactor head is removed. C. Incorrect because lower internals removal is not considered a Core Alteration and as such does not require direct supervision by an SRO. Plausible because this activity occurs directly over the reactor while the reactor head and upper internals are removed. D. Correct per the reference 194 NRC DRAFT 07/24/07 Q#98 Core off-load is in progress on Unit 3.

• One fuel assembly is latched on the manipulator crane.

• Another fuel assembly is vertical in the upender on the Spent Fuel Pit side. A refueling cavity seal failure occurs and cavity level begins decreasing rapidly. Which ONE of the following describes the correct operators' responses?

A. Insert the assembly on the manipulator into the core and leave it latched. Lay the assembly in the upender down and leave it in the transfer cart. B. Insert the assembly on the manipulator into the core and unlatch it. Relatch the assembly in the upender and move it to a fuel rack. C. Do NOT Insert the assembly on the manipulator into the core but move it to the edge of the core and leave it suspended.

Lay the assembly in the upender down and leave it in the transfer cart. D. Do NOT Insert the assembly on the manipulator into the core but move it to transfer canal. Relatch the assembly in the upender and move it to a fuel rack. 195 NRC DRAFT 07/24/07 Q#98 ANSWER: A KA: G2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

3.3 10CFR55: 43.b.4/b.7

Reference:

3-0NOP-033.2, Step 5.1.2, 5.1.3 & NOTE Prior to Step 5.2.4 Cog Level: 1 Recall New Question Response Analysis:

A. Correct per the reference.

B. Incorrect because the operator should lay the assembly in the upender down and leave it in the transfer cart. Plausible because operators should insert the assembly on the manipulator into the core but should leave it latched. C. Incorrect because operators should insert the assembly on the manipulator into the core and leave it latched. Plausible because the operator should lay the assembly in the upender down and leave it in the transfer cart. D. Incorrect because operators should insert the assembly on the manipulator into the core and leave it latched and because the operator should lay the assembly in the upender down and leave it in the transfer cart. Plausible because unlatch the assembly held by the manipulator and relatching the assembly in the upender and move it to a fuel rack are both workable options open to the operators but neither is directed by ONOP-033.2.

196 NRC DRAFT 07124/07 0#99 Unit 3 has experienced a loss of offsite power (LOOP).

• The 38 EDG has energized the 38 4KV bus as designed.

• 3A EDG locked out and can NOT be restarted.

• SI was NOT required and did NOT occur. Which ONE of the following describes the mitigation strategies provided by 3-EOP-ES-O.2, "Natural Circulation Cooldown," for the present plant conditions?

Limit the RCS cooldown rate to: A. 25°F/hr and maintain RCS subcooling 80°F to avoid the possibility of thermal shocking the reactor vessel. 8. 25°F/hr and maintain RCS subcooling 226°F to avoid the possibility of reactor vessel head void formation.

C. 60°F/hr and maintain RCS subcooling 80°F to avoid the possibility of thermal shocking the reactor vessel.. D. 60°F/hr and maintain RCS subcooling 226°F to avoid the possibility of reactor vessel head void formation.

197 NRC DRAFT 07/24/07 Q#99 ANSWER: B KA G2.4.7 Knowledge of the event based EOP mitigation strategies.

3.8 10CFR55: 43.b.5

Reference:

3-EOP-ES-0.2, Step 10,17,18 Cog Level: 2 Comprehensive Level 2 because the SRO must relate the loss of the 3A 4KV bus to the present plant conditions and determine that only one CRDM cooling fan is available.

He then must relate this knowledge to the effects the CRDM cooling fans have on a natural circulation cooldown and determine that the lower cooldown rates and much higher subcooling numbers must be applied for the given plant conditions as heat removal from the reactor head is diminished with only one CRDM fan .. New Question Response Analysis:

A. Incorrect because operators should maintain RCS subcooling

226°F to avoid the possibility of reactor vessel head void formation.

Plausible because operators should limit the RCS cooldown rate to 25°F/hr. B. Correct per the reference C. Incorrect because operators should limit the RCS cooldown rate to 25°F/hr and maintain RCS subcooling 2 226°F to avoid the possibility of reactor vessel head void formation.

Plausible because maintaining subcooling

> 80°F is an appropriate strategy if both CRDM coolers are in operation and 60°F/hour is an attainable cooldown rate that might be viewed as appropriate considering the off-normal event. D. Incorrect because operators should limit the RCS cooldown rate to 25°F/hr Plausible because operators should maintain RCS subcooling 2 226°F to avoid the possibility of reactor vessel head void formation and 60°F/hour is an attainable cooldown rate that might be viewed as appropriate considering the off-normal event. 198 NRC DRAFT 07/24/07 Q #100 Unit 4 has experienced a large break LOCA and containment pressure has peaked at 28 psig. The Emergency Coordinator (EC) has NOT yet received any indication of potential radioactive releases from Containment.

Which ONE of the following describes the Protective Action Recommendation (PAR) the EC should make during initial contact with the State? A. No PAR should be made until confirmation of a radioactive release is made. B. Shelter all persons within a 2 mile radius of the site and shelter all persons out to 5 miles in the downwind sectors. C. Evacuate all persons within a 2 mile radius of the site and shelter all persons out to 5 miles in the downwind sectors. D. Evacuate all persons within a 2 mile radius of the site and evacuate all persons out to 5 miles in the downwind sectors. REFERENCE PROVIDED 199 NRC DRAFT 07124/07 Q#100 ANSWER: B KA: G2.4.44 Knowledge of emergency plan protective action recommendations.

4.0 10CFR55: 43.b.5

Reference:

3-EPIP-201 01, Enclosure 1, Category 1 (General Emergency) 3-EPIP-20101, Step 5.1.4 3-EPIP-20101, Attachment 3, Page 1 of 6 Cog Level: 2 comprehension Level 2 because the operator must determine that the combination of a large break LOCA with 28 psig inside containment as classifiable as a General Emergency.

Then the operator must recall that the minimum classification for a GE, even without a release is "Shelter all persons within a 2 mile radius of the site and shelter all persons out to 5 miles in the downwind sectors." Provide EPIP-20101, Enclosure 1 as a Reference New Question Response Analysis:

A. Incorrect because the EC should issue the PAR to shelter all persons within a 2 mile radius of the site and shelter all persons out to 5 miles in the downwind sectors. Plausible because there is no indication of potential radioactive releases from Containment.

B. Correct per the reference C. Incorrect because the EC should issue the PAR to shelter all persons within a 2 mile radius of the site and shelter all persons out to 5 miles in the downwind sectors. Plausible because evacuation could be considered more conservative than sheltering in the event of a General Emergency.

D. Incorrect because the EC should issue the PAR to shelter all persons within a 2 mile radius of the site and shelter all persons out to 5 miles in the downwind sectors. Plausible because evacuation could be considered more conservative than sheltering in the event of a General Emergency.

200