ML100470918

Millstone Station, Unit 2 - Draft Written Exam

ML100470918
Person / Time
Site:	Millstone
Issue date:	11/20/2009
From:	D'Antonio J M Operations Branch I
To:	Dominion Nuclear Connecticut
Hansell S
Shared Package
ML092400035	List: IR 05000336/2010301 ML100220055 ML100470870 ML100470905 ML100470918 ML100610136 ML100670496 ML100670507 ML101160391
References
50-336/10-301, TAC U01799, FOIA/PA-2011-0115 50-336/10-301
Download: ML100470918 (105)

Text

s . Question #: 1 Student Handout? Lower Order?Question ID: 9000018 RO SRO l-SRO Ques. # 1 Rev. o Selected for Exam Origin: New C Past NRC Exam? The plant automatically tripped on High Pressurizer Pressure due to an inadvertent closure of the Main Turbine Control Valves. During the performance of EOP 2525, Standard Post Trip Actions, the crew reported that Bus 24D is deenergized due to a fault and that Power Operated Relief Valve (PORV), RC-404, is stuck open. All other equipment operated as designed.

Upon entry into EOP 2532, Loss of Coolant Accident, the following conditions exist: -Containment pressure is 4.5 psia and slowly -Reactor vessel is 43% and slowly going -CET temperatures are 578°F and -RCS pressure is 1310 psia and -Pressurizer level is -Steam generator levels are both 41 % and going up Which of the following actions must the Unit Supervisor/Shift Manager perform to preserve a A Direct the Technical Support Center to develop a plan to restore RWST level. D B Direct the Balance of Plant Operator to align Condenser Air Removal to the Unit 2 Stack. C Direct the Reactor Operator to place the SItCS Pump Miniflow switches in "OPERATE". D Direct the crew to commence a controlled cooldown and depressurization.

Justification D IS CORRECT: With RCS pressure stable at 1310 psia and the PORV still open, RCS inventory is being lost faster than can restore it. The steps for the cooldown and subsequent depressurization must be pulled forward (performed out of sequence) allow RCS pressure to be reduced below HSPI shut off head to allow adequate Safety Injection A is incorrect; Although RWST level is lowering, there is NO need to develop a plan to restore RWST level at this time (perform out of Plausible because step *I of EOP 2532 directs the US or SM to have the TSC develop a plan for restoring level in the RWST if LOCA is determined to be outside of Containment.

Examinee may not remember that this step is performed ONLY if the LOCA outside of B is incorrect; With Containment pressure >4.42 psia, MSI has actuated and the MSIVs are closed resulting in a loss of vacuum; therefore.

the is no need to align Condenser Air Removal to the Unit 2 Plausible because this is a procedurally directed step. Step 15 states, "If EBFAS has initiated and the Condenser is available, align Condenser Air Removal to Unit 2 stack." If the examinee does not realize that MSI has actuated, then this step may performed out of C is incorrect; The SIICS Pump Miniflow switches are not placed in 'OPERATE" until RWST level is Plausible because the examinee may feel that a Sump Recirc Actuation Signal is imminent; therefore, it would be appropriate perform this step out of I EOP 2532, "LOCA" and OP 2260, "EOP Users Guide" Comments and Question Modification History Bob K. -D-3/C, K Corrected typo in distractor B. Bill M. -D-2/C, K Angelo -D-2/C; No comments.

NRC KIA SystemlEIA System 009 Small Break LOCA Generic:

NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.20 R04.6 SR04.6 CFRLink (CFR: 41.10143.5/45.12)

Ability to interpret and execute procedure steps. Page 1 of 77 Printed on 10/28/2009 at 13:20 s . Question #: 1 Question ID: 9000018 RO SRO Student Handout? Lower Order? l-SRO Ques. # 1 Rev. o Selected for Exam Origin: New C Past NRC Exam? The plant automatically tripped on High Pressurizer Pressure due to an inadvertent closure of the Main Turbine Control Valves. During the performance of EOP 2525, Standard Post Trip Actions, the crew reported that Bus 24D is deenergized due to a fault and that Power Operated Relief Valve (PORV), RC-404, is stuck open. All other equipment operated as designed.

Upon entry into EOP 2532, Loss of Coolant Accident, the following conditions exist: -Containment pressure is 4.5 psia and slowly rising. -Reactor vessel is 43% and slowly going down -CET temperatures are 578°F and stable -RCS pressure is 1310 psia and stable -Pressurizer level is 100%. -Steam generator levels are both 41 % and going up slowly. Which of the following actions must the Unit Supervisor/Shift Manager perform to preserve a Safety Function?

A Direct the Technical Support Center to develop a plan to restore RWST level. D B Direct the Balance of Plant Operator to align Condenser Air Removal to the Unit 2 Stack. C Direct the Reactor Operator to place the SItCS Pump Miniflow switches in "OPERATE". D Direct the crew to commence a controlled cooldown and depressurization.

Justification J D IS CORRECT: With RCS pressure stable at 1310 psia and the PORV still open, RCS inventory is being lost faster than Charging can restore it. The steps for the cooldown and subsequent depressurization must be pulled forward (performed out of sequence) to allow RCS pressure to be reduced below HSPI shut off head to allow adequate Safety Injection flow. A is incorrect; Although RWST level is lowering, there is NO need to develop a plan to restore RWST level at this time (perform step out of sequence).

Plausible because step *I of EOP 2532 directs the US or SM to have the TSC develop a plan for restoring level in the RWST if the LOCA is determined to be outside of Containment.

Examinee may not remember that this step is performed ONLY if the LOCA is outside of Containment.

B is incorrect; With Containment pressure >4.42 psia, MSI has actuated and the MSIVs are closed resulting in a loss of Condenser vacuum; therefore.

the is no need to align Condenser Air Removal to the Unit 2 stack. Plausible because this is a procedurally directed step. Step 15 states, "If EBFAS has initiated and the Condenser is available, then align Condenser Air Removal to Unit 2 stack." If the examinee does not realize that MSI has actuated, then this step may be performed out of sequence.

C is incorrect; The SIICS Pump Miniflow switches are not placed in 'OPERATE" until RWST level is Plausible because the examinee may feel that a Sump Recirc Actuation Signal is imminent; therefore, it would be appropriate to perform this step out of sequence.

Refel"lmces I EOP 2532, "LOCA" and OP 2260, "EOP Users Guide" Comments and Question Modification History Bob K. -D-3/C, K Corrected typo in distractor B. Bill M. -D-2/C, K Angelo -D-2/C; No comments.

NRC KIA SystemlEIA System 009 Small Break LOCA Generic:

NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.20 R04.6 SR04.6 CFRLink (CFR: 41.10143.5/45.12)

Ability to interpret and execute procedure steps. Page 1 of 77 Printed on 10/28/2009 at 13:20 Question #: 1 Student Handout? o Lower Order? Question 10: 9000018 RO SRO p-SRO Que *. # 1 Rev. o Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 EOP 2532 Revision 029 I Page 21 of95 Loss of' Coolant Accident ! INSTR CONTING ENCY ACfIONS NOTE Res coolduwn should be initiated within one hour after the event to conserve condensate inventory and comply with the Long Term Cooling Analysis. ReS cooldown rate greater than 40"F/hr should be maintained until the steam dump/bypass or atmospheric dump valves are full open. The starling point for the ReS couktOV\'ll should be the Tc or eET temperature', where ReS has stabilized. TC should used for monitoring ReS cooldown if in forced or nalmal circulation.

eET." should be used for all other cases. i I NOTE Technical Specification cooldowll rates should be observed during the cooldowll.

The cooldown rates are as follows: l. Res Tc grcaterthan 22W F the cooldoWJl rate is !f)()OF/1u.

2. Res Tc less than or equal to 220 c F the cooldown rate is 500F/hr. 1>('1;01'01 ControHed Cooldown "17.INITIATE a controlled cooldown using the steam dumps to establish shutdown cooling entry conditions.

17.1 INITIATE a controlled cooldown lIsing the ADVs to establish shutdown cooling entry conditions.

STOP THINK ACT REVIEW Page 2 of Printed on 10/28/2009 at 13:20 Question #: 1 Question 10: 9000018 RO SRO Student Handout? p-SRO Que *. # 1 Rev. o Selected for Exam Origin: New i I Millstone Unit 2 Loss of' Coolant Accident INSTR UC"TIONS EOP 2532 Revision 029 I Page 21 of95 ! CONTING ENCY ACfIONS NOTE I. Res coolduwn should be initiated within one hour after the event to conserve condensate inventory and comply with the Long Term Cooling Analysis.

2. ReS cooldown rate greater than 40"F/hr should be maintained until the steam dump/bypass or atmospheric dump valves are full open. 3. The starling point for the ReS couktOV\'ll should be the Tc or eET temperature', where ReS has stabilized.

4. TC should used for monitoring ReS cooldown if in forced or nalmal circulation.

eET." should be used for all other cases. NOTE Technical Specification cooldowll rates should be observed during the cooldowll.

The cooldown rates are as follows: l. Res Tc grcaterthan 22W F the cooldoWJl rate is HKloF/1u.

2. Res Tc less than or equal to 220 c F the cooldown rate is 500F/hr. 1>('1;01'01 ControHed Cooldown "17.INITIATE a controlled cooldown using the steam dumps to establish shutdown cooling entry conditions.

STOP THINK 17.1 INITIATE a controlled cooldown lIsing the ADVs to establish shutdown cooling entry conditions.

ACT REVIEW o Lower Order? Past NRC Exam? Page 2 of 77 Printed on 10/28/2009 at 13:20 Question #: 1 RO SRO Student Handout? Lower Order? Question ID; 9000018 J-SRO Ques. Rev. o Selected for Exam Origin; New D Past NRC Exam? Procedural steps listed in alphanumeric order arc sequential steps and shall be addressed in that sequence.

Exceptions to this are as follows. Asteriskcd steps, within the ORP or selected FRPs being implemented.

muy be brought forward to correct or preserve a Safety Function. StCp1' Illay be performed Ollt of order after they have been accomplished once, if they arc Continuously Applicable step, as indicated by an asterisk. Steps with recurrent actions (i.e., the step will be performed repeatedly during the procedure) should be checked off in the pI acckeeper when started. Since these steps will be performed repeatedly.

the placekceper is marked with a "COllt," in the "Done" column. This signifies that the step is continuously performed and will not be completed during the performance of the EOP. Bu!leted lists are provided within a step when anyone of several alternative actions arc equally acceptable to perform. The preferred method is listed as the first alternative. It is acceptable for the SM or US to dircctthe performance of a task out of sequence, if the actions do not interfere with maintaining an existing Safety Function (e.g., transferring power supplies for "B" charging pump, preparing for power restoration).

1.9.2 Instructions

and Contingency Actions The EOPs are formatted with [nstructiollS and Contingem..y Actions. The instructions column presents the optimal method <mu sequence for accomplishing a specific ta..k. The contingencies column contains actions to be performed if the optimum method cannot he accomplished. If the expected response is obtained (left column), the operator proceeds to the next step or sub step in tile Instructions

{."olumn (left column). OP 226f1 STOP THINK REVIEW Rev. 009-03 130f60 Page 3 of Printed on 10/28/2009 at 13:20 Question #: 1 J-SRO Ques. # Question ID; 9000018 RO SRO Student Handout? Lower Order? Rev. o Selected for Exam Origin; New D Past NRC Exam? b. Procedural steps listed in alphanumeric order arc sequential steps and shall be addressed in that sequence.

Exceptions to this are as follows.

• Asteriskcd steps, within the ORP or selected FRPs being implemented.

muy be brought forward to correct or preserve a Safety Function.

• StCp1' Illay be performed Ollt of order after they have been accomplished once, if they arc Continuously Applicable step, as indicated by an asterisk.

c. Steps with recurrent actions (i.e., the step will be performed repeatedly during the procedure) should be checked off in the pI acckeeper when started. Since these steps will be performed repeatedly.

the placekceper is marked with a "COllt," in the "Done" column. This signifies that the step is continuously performed and will not be completed during the performance of the EOP. d. Bu!leted lists are provided within a step when anyone of several alternative actions arc equally acceptable to perform. The preferred method is listed as the first alternative.

e. It is acceptable for the SM or US to dircctthe performance of a task out of sequence, if the actions do not interfere with maintaining an existing Safety Function (e.g., transferring power supplies for "B" charging pump, preparing for power restoration).

1.9.2 Instructions

and Contingency Actions a. The EOPs are formatted with [nstructiollS and Contingem

.. y Actions. The instructions column presents the optimal method <mu sequence for accomplishing a specific ta .. k. The contingencies column contains actions to be performed if the optimum method cannot he accomplished.

b. If the expected response is obtained (left column), the operator proceeds to the next step or sub step in tile Instructions

{."olumn (left column). STOP THINK Page 3 of 77 OP 226f1 REVIEW Rev. 009-03 130f60 Printed on 10/28/2009 at 13:20 SRfZ)Exam Questions* . Question #: 2 RO SRO Student Handout? D Lower Order?Question ID: 9000019 I-SRO Quell. # 2 Rev. 0 Selected for Exam Origin: New D Past NRC Exam? The plant tripped from 100% power due to a Large Break LOCA. The crew successfully completed all actions of ='OP 2525, Standard Post Trip Actions, and are presently performing EOP 2532, Loss of Coolant Accident.

The following conditions exist approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the trip:

• SRAS actuated approximately 15 minutes ago.

• Containment pressure is 5 psig and slowly lowering.

• RCS pressure is 360 psia and slowly lowering.

• CET temperatures indicate 434°F and slowly lowering.

• HPSI Pump current and flow are fluctuating.

Which of tile following describes the cause of the HPSI Pump current and flow fluctuations, and the initial action that must be directed? Hot water in the Containment Sump is flashing to steam in the HPSI Pump Start at least 2 CAR Fans in Fast The HPSI Pumps are showing signs of cavitation due to Containment Sump clogging.

Secure both Containment Spray Pumps. Boron is beginning to plate out in the core causing alternately high and low HPSI Establish Hot Leg and Cold Leg D Total Safety Injection flow is higher than necessary for the present Throttle the HPSI Injection valves as Justification I B IS CORRECT; Sump clogging will cause a lower suction pressure in all the running SI pumps. A lower suction pressure will the HPSI Pumps to cavitate.

EOP 2532 directs the CS pumps be secured (if not needed) to limit the competition for sump suction EOP 2532 also requires Containment Spray Pumps to remain in operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for Iodine scrubbing; however, cooling (maintaining adequate SI flow to the core) takes precedence over Iodine A is incorrect; The Containment Spray System and CAR Coolers are designed to lower containment sump temperature enough prevent cavitation of the HPSI Plausible because the SRAS caused the HPSI Pump suctions to swap from the RWST (cool water) to the Containment Sump water). Starting 2 CAR Fans in Fast speed would help to lower the Containment Sump temperature; however, there is no guidance to perform this C is incorrect; Boron Precipitation is analyzed to occur after a large break LOCA; however, it is NOT analyzed to occur until 8-10 event. It's NOT likely that Boron would be solidifying in the core at this plausible because accident analysis shows boron Precipitation will occur after a large break LOCA. The examinee may remember the lime frame for Boron Precipitation (8-10 hours after the LOCA). Simultaneous Hot Leg and Cold Leg Injection is appropriate acton for Boron D is incorrect; Total Safety Injection flow is likely above the SI flow curve. The curve is based on having only one train of SI in (Accident Analysis).

However, the additional flow does NOT adversely impact core Plausible because the examinees should know that Safety Injection flow is higher than required for core cooling and may conclude throttling HPSI is appropriate.

EOP 2532 has steps for throttling HPSI I EOP-2532, SI. 50, Indications of CTMT Sump Comments and Question Modification Bob K. -D**4fW (talked self out of Corrected typo in distractor Bill M. -D-2fC, Angelo -D-4fC. Fair NRC KIA System/E/A System 011 Large Break LOCA Number E.A2.10 RO 4.5 SRO 4.7 CFR Link (CFR 43.5 f 45.13) Ability to determine or interpret the following as they apply to a Large Break LOCA: Verification of adequate core cooling Page 4 of Printed on 10/2812009 at 13:20 Question #: 2 I-SRO Quell. # 2 SRfZ)Exam Questions* . Question ID: 9000019 RO SRO Rev. 0 Selected for Exam Student Handout? Origin: New D Lower Order? D Past NRC Exam? The plant tripped from 100% power due to a Large Break LOCA. The crew successfully completed all actions of ='OP 2525, Standard Post Trip Actions, and are presently performing EOP 2532, Loss of Coolant Accident.

The following conditions exist approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the trip:

• SRAS actuated approximately 15 minutes ago.

• Containment pressure is 5 psig and slowly lowering.

• RCS pressure is 360 psia and slowly lowering.

• CET temperatures indicate 434°F and slowly lowering.

• HPSI Pump current and flow are fluctuating.

Which of tile following describes the cause of the HPSI Pump current and flow fluctuations, and the initial action that must be directed?

A Hot water in the Containment Sump is flashing to steam in the HPSI Pump suctions.

Start at least 2 CAR Fans in Fast speed. B The HPSI Pumps are showing signs of cavitation due to Containment Sump clogging.

Secure both Containment Spray Pumps. C Boron is beginning to plate out in the core causing alternately high and low HPSI flow. Establish Hot Leg and Cold Leg Injection.

D Total Safety Injection flow is higher than necessary for the present conditions.

Throttle the HPSI Injection valves as needed. Justification I B IS CORRECT; Sump clogging will cause a lower suction pressure in all the running SI pumps. A lower suction pressure will cause the HPSI Pumps to cavitate.

EOP 2532 directs the CS pumps be secured (if not needed) to limit the competition for sump suction flow. EOP 2532 also requires Containment Spray Pumps to remain in operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for Iodine scrubbing; however, core cooling (maintaining adequate SI flow to the core) takes precedence over Iodine scrubbing.

A is incorrect; The Containment Spray System and CAR Coolers are designed to lower containment sump temperature enough to prevent cavitation of the HPSI Pumps. Plausible because the SRAS caused the HPSI Pump suctions to swap from the RWST (cool water) to the Containment Sump (hot water). Starting 2 CAR Fans in Fast speed would help to lower the Containment Sump temperature; however, there is no procedural guidance to perform this action. C is incorrect; Boron Precipitation is analyzed to occur after a large break LOCA; however, it is NOT analyzed to occur until 8-10 event. It's NOT likely that Boron would be solidifying in the core at this time. plausible because accident analysis shows boron Precipitation will occur after a large break LOCA. The examinee may NOT remember the lime frame for Boron Precipitation (8-10 hours after the LOCA). Simultaneous Hot Leg and Cold Leg Injection is the appropriate acton for Boron Precipitation.

D is incorrect; Total Safety Injection flow is likely above the SI flow curve. The curve is based on having only one train of SI in service (Accident Analysis).

However, the additional flow does NOT adversely impact core cooling. Plausible because the examinees should know that Safety Injection flow is higher than required for core cooling and may conclude that throttling HPSI is appropriate.

EOP 2532 has steps for throttling HPSI flow. References I EOP-2532, SI. 50, Indications of CTMT Sump Clogging Comments and Question Modification History Bob K. -D**4fW (talked self out of answer) Corrected typo in distractor D. Bill M. -D-2fC, K Angelo -D-4fC. Fair question.

NRC KIA System/E/A System 011 Large Break LOCA Number E.A2.10 RO 4.5 SRO 4.7 CFR Link (CFR 43.5 f 45.13) Ability to determine or interpret the following as they apply to a Large Break LOCA: Verification of adequate core cooling Page 4 of 77 Printed on 10/2812009 at 13:20 Question #: 2 RO rill SRO Student Handout? Lower Order? Question ID: 9000019 II-SRO Ques. # 2 Rev. o rill Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 EOP 2532 Revision 029 Pagc42 of 95 Loss of Coolant Accident CONTINGENCY ACTIONS NOTE Degradation in HPS] pump performance.

post SRAS. may be indicative of debris fouling the CTMT sump screen. Checking tIPS] pump flow greater than 30 gplll ensures minimum now requirements are met f(}1' pump protection when ReS pressure is high und prohibiting now. This presents differently than the slimp blockage issue. HPSI Pump Post SRAS I)erformance Criteria "SOJF SRAS has actuated, CHECK for adequate HPSI flow by ObSelYing ALL of the fulluwing:: Flow greater than or equal to 30 gpm for each operating pump Motor current stable Stable HPSI pump discharge prcs!'.1.uc (continue)

STOP THINK I, SO.t unable to maintain HPSl flow due to high ReS pressure, STOP ONE HPSI pump to establish the following for the operating HPSI pump:

• Flow greater than or equal to 30gpm Motor current stable Stable HPSI pump discharge pres!\ure If. HPSI pump performance degradation is due to CTMT sump clogging, (suction problem) PERFORM the following, as necessary, to attempt restoration of HPSI tlow: CfMT pressure can he maintained less than 54 psig. AND alieasl ONE complete facility of CAR fans is operating, STOP CS pumps. (<."ontinllc)

ACT REVIEW Page 5 of Printed on 1 0/28/2009 at 13:20 Question #: 2 Question ID: 9000019 RO rill SRO Student Handout? II-SRO Ques. # 2 Rev. o rill Selected for Exam Origin: New I, Millstone Unit 2 Loss of Coolant Accident INSTRucnONS EOP 2532 Revision 029 Pagc42 of 95 CONTINGENCY ACTIONS NOTE Degradation in HPS] pump performance.

post SRAS. may be indicative of debris fouling the CTMT sump screen. Checking tIPS] pump flow greater than 30 gplll ensures minimum now requirements are met f(}1' pump protection when ReS pressure is high und prohibiting now. This presents differently than the slimp blockage issue. HPSI Pump Post SRAS I)erformance Criteria "SOJF SRAS has actuated, CHECK for adequate HPSI flow by ObSelYing ALL of the fulluwing::

• Flow greater than or equal to 30 gpm for each operating pump

• Motor current stable

• Stable HPSI pump discharge prcs!'.1.uc SO.t unable to maintain HPSl flow due to high ReS pressure, STOP ONE HPSI pump to establish the following for the operating HPSI pump:

• Flow greater than or equal to 30gpm

• Motor current stable

• Stable HPSI pump discharge pres!\ure 50.2 If. HPSI pump performance degradation is due to CTMT sump clogging, (suction problem) PERFORM the following, as necessary, to attempt restoration of HPSI tlow: a. CfMT pressure can he maintained less than 54 psig. AND alieasl ONE complete facility of CAR fans is operating, STOP CS pumps. (continue)

(<."ontinllc)

STOP THINK ACT REVIEW Lower Order? Past NRC Exam? Page 5 of 77 Printed on 1 0/28/2009 at 13:20 Question #: 3 Question ID: 9000003 RO'l/i SRO Student Handout? Lower Order? I-SRO Ques. # 3 Rev. o Selected for Exam Origin: New Past NRC Exam? While operating at 100% power, the RCP A UPPER SEAL PRES HI annunciator alarms. While referring to the appropriate Annunciator Response Procedure, the RCP A BLEED-OFF FLOW HI annunciator alarms. Within a minute, the RCP A BLEED-OFF FLOW HI annunciator clears and the RCP A BLEED-OFF FLOW LO annunciator alarms and remains lit. Numerous annunciators associated with "A" RCP seals also alarm. Which of the following describes the reason for this sequence of annunciators and the direction that must be given? The "A" RCP Excess Flow Check Valve has seated. Manually trip the reactor and turbine, then stop the "An RCP. The "A" RCP Middle Seal has failed. Evaluate the condition of the other seals to confirm no other degradation or failures. The Bleedoff Pressure Controller, PIC-215, has malfunctioned.

Using the Foxboro Controller, restore "An Rep Bleedoff pressure and flow to the normal band. o The RCP Bleedoff Relief Valve has inadvertently opened. Evaluate "An RCP Seal pressures determine whether or not the "A" RCP may remain in Justification I A is CORRECT; The Rep A UPPER SEAL PRES HI annunciator is indicative of a failure of the nAn RCP Middle or Lower Seal (or combination of both). This resulted the a high Bleedoff flow through the nN RCP Seals resulting in a RCP A BLEED-OFF FLOW annunciator.

At 10 gpm. the Excess Flow Check Valve will close causing the RCP A BLEED-OFF FLOW HI annunciator to clear the RCP A BLEED-OFF FLOW LO to annunciate

<<0.75 gpm). At this point, the RCP Seal package has NO cooling flow and the must be tripped. Procedurally, the reactor and turbine are tripped prior to tripping the affected B is incorrect; The individual indications provided could be a result of a failure of the nAn RCP Middle Seal; however, the together would indicate a loss of Bleedoff flow through the nAn RCP seals requiring the RCP to be stopped. If a Middle Seal had then the action is correct If Bleedoff Pressure Controller, PIC-215, had malfunctioned, then the action would be Plausible because the Annunciator Response Procedures for the Upper Seal High Pressure and Bleedoff High Flow annunciators that these alarms may be indicative of a failed Middle C is incorrect; The above indications could be indicative of a failure of the RCP Bleedoff Pressure Controller; however, all 4 would have similar annunciators.

Additionally, the RCP Bleedoff Relief Valve would cycle to provide flow through all 4 of the Seals. Use of the Foxboro Controller will NOT be effective.

If the RCP Bleedoff Relief Valve has inadvertently opened, then the would be Plausible because a failed or failing RCP Bleedoff Pressure Controller could give the above indications; however, all 4 RCPs would D is incorrect; Opening of the RCP Bleedoff Relief Valve would likely result in a high Bleedoff flow on all 4 RCPs and NOT a Bleedoff flow annunciator.

Additionally, nAn RCP Seal pressures are dependent on seal conditions and not necessarily the flow path Plausible because the examinee may be confused on how the Bleedoff Relief Valve would impact an individual RCP Bleedoff flow pressure.

The examinee may also be confused as to whether each RCP had a bleedoff relief I ,A,RP 2590B-068 Comments and Question Modification History Bob K. -D-3/C Bill M. -D-3/C, 50150 Angelo -D-4/C; No comments.

NRC KJA SystemtEtA System 015 Reactor Coolant Pump Malfunctions NRC KJA Generic System 2.1 Conduct of Operations Number 2.1.7 RO 4.4 SRO 4.7 CFR Link (CFR: 41.5/43.5/45.12 I Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, instrument Page 6 of Printed on 10/28/2009 at 13:20 Question #: 3 Question ID: 9000003 RO'l/i SRO Student Handout? Lower Order? I-SRO Ques. # 3 Rev. o Selected for Exam Origin: New Past NRC Exam? While operating at 100% power, the RCP A UPPER SEAL PRES HI annunciator alarms. While referring to the appropriate Annunciator Response Procedure, the RCP A BLEED-OFF FLOW HI annunciator alarms. Within a minute, the RCP A BLEED-OFF FLOW HI annunciator clears and the RCP A BLEED-OFF FLOW LO annunciator alarms and remains lit. Numerous annunciators associated with "A" RCP seals also alarm. Which of the following describes the reason for this sequence of annunciators and the direction that must be given? A The "A" RCP Excess Flow Check Valve has seated. Manually trip the reactor and turbine, then stop the "An RCP. B The "A" RCP Middle Seal has failed. Evaluate the condition of the other seals to confirm no other degradation or failures.

C The Bleedoff Pressure Controller, PIC-215, has malfunctioned.

Using the Foxboro Controller, restore "An Rep Bleedoff pressure and flow to the normal band. o The RCP Bleedoff Relief Valve has inadvertently opened. Evaluate "An RCP Seal pressures to determine whether or not the "A" RCP may remain in operation.

Justification I A is CORRECT; The Rep A UPPER SEAL PRES HI annunciator is indicative of a failure of the nAn RCP Middle or Lower Seal (or a combination of both). This resulted the a high Bleedoff flow through the nN RCP Seals resulting in a RCP A BLEED-OFF FLOW HI annunciator.

At 10 gpm. the Excess Flow Check Valve will close causing the RCP A BLEED-OFF FLOW HI annunciator to clear and the RCP A BLEED-OFF FLOW LO to annunciate

<<0.75 gpm). At this point, the RCP Seal package has NO cooling flow and the RCP must be tripped. Procedurally, the reactor and turbine are tripped prior to tripping the affected RCP. B is incorrect; The individual indications provided could be a result of a failure of the nAn RCP Middle Seal; however, the indications together would indicate a loss of Bleedoff flow through the nAn RCP seals requiring the RCP to be stopped. If a Middle Seal had failed, then the action is correct If Bleedoff Pressure Controller, PIC-215, had malfunctioned, then the action would be appropriate.

Plausible because the Annunciator Response Procedures for the Upper Seal High Pressure and Bleedoff High Flow annunciators state that these alarms may be indicative of a failed Middle Seal. C is incorrect; The above indications could be indicative of a failure of the RCP Bleedoff Pressure Controller; however, all 4 RCPs would have similar annunciators.

Additionally, the RCP Bleedoff Relief Valve would cycle to provide flow through all 4 of the RCP Seals. Use of the Foxboro Controller will NOT be effective.

If the RCP Bleedoff Relief Valve has inadvertently opened, then the action would be appropriate.

Plausible because a failed or failing RCP Bleedoff Pressure Controller could give the above indications; however, all 4 RCPs would be affected.

D is incorrect; Opening of the RCP Bleedoff Relief Valve would likely result in a high Bleedoff flow on all 4 RCPs and NOT a low Bleedoff flow annunciator.

Additionally, nAn RCP Seal pressures are dependent on seal conditions and not necessarily the flow path of Bleedoff.

Plausible because the examinee may be confused on how the Bleedoff Relief Valve would impact an individual RCP Bleedoff flow and pressure.

The examinee may also be confused as to whether each RCP had a bleedoff relief valve. References I ,A,RP 2590B-068 Comments and Question Modification History Bob K. -D-3/C Bill M. -D-3/C, 50150 Angelo -D-4/C; No comments.

NRC KJA SystemtEtA System 015 Reactor Coolant Pump Malfunctions NRC KJA Generic System 2.1 Conduct of Operations Number 2.1.7 RO 4.4 SRO 4.7 CFR Link (CFR: 41.5/43.5/45.12 I 45.13) Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.

Page 6 of 77 Printed on 10/28/2009 at 13:20 Student Handout?

• QLlestion#:

3 Question ID; 9000003 RO Lower Order? l-SRO Ques. # 3 Rev. o Selected for Exam Origin; New D Past NRC Exam? 5/26109 Approval Effective Date

____---, Setpoint:

2 gpm RCP A OFF FLOW HI AUTOMATIC FUNCTIONS None NOTE This alarm may be indicative of seal stage failure. One seal failure can cause high bleed off flow alarm. Operation may continue with this alarm present providing seal bleed off temperature is within limits and seal differential pressures indicates that only one of the three lower seal stages I A vapor stage failure will require a plant trip. CQBBECTIVE NOTE If seal flow reaches I () gpm, "A" Rep controlled bleedoff excess flow check valve doses to preventl)lockagc of bleedoff 1] ow from other Reps. Refer To the following guidance and DETERlVlINE if "I\.' Rep controlled bleedoff excess How check valve ha'i closed: High blecdoff flow alarm annunciates and dears follo\ved by 10\'1' hleedoff flow alarm which remains lit Seal pressures rise on all 3 seals (provided vapor seal is inlact) IF "PI.' Rep controlled bleedoff excess flow check valve has dosed, PERFORM the following: 'rRJP reactor and turbine. STOP I\.' Rep. Go To EOP 2.':;25, "Standard Post Trip Actions'" and PERFORM required actions. Page 7 of Printed on 10/28/2009 at 13:20 Question ID; 9000003 RO SRO Student Handout? Lower Order?

• QLlestion#:

3 l-SRO Ques. # 3 Rev. o Selected for Exam Origin; New D Past NRC Exam? Approval Date Setpoint:

2 gpm 5/26109 AUTOMATIC FUNCTIONS

1. None Effective Date NOTE

____ ---, RCP A OFF FLOW HI This alarm may be indicative of seal stage failure. One seal failure can cause high bleed off flow alarm. Operation may continue with this alarm present providing seal bleed off temperature is within limits and seal differential pressures indicates that only one of the three lower seal stages I failed. A vapor stage failure will require a plant trip. CQBBECTIVE ACTIONS NOTE If seal flow reaches I () gpm, "A" Rep controlled bleedoff excess flow check valve doses to preventl)lockagc of bleedoff 1] ow from other Reps. 5. Refer To the following guidance and DETERlVlINE if "I\.' Rep controlled bleedoff excess How check valve ha'i closed: '" High blecdoff flow alarm annunciates and dears follo\ved by 10\'1' hleedoff flow alarm which remains lit

• Seal pressures rise on all 3 seals (provided vapor seal is inlact) 6. IF "PI.' Rep controlled bleedoff excess flow check valve has dosed, PERFORM the following:

Il.1 'rRJP reactor and turbine. 6.2 STOP I\.' Rep. 6.3 Go To EOP 2.':;25, "Standard Post Trip Actions'" and PERFORM required actions. Page 7 of 77 Printed on 10/28/2009 at 13:20 (No "Parents t, ("

Question #: Question ID: 9000004 RO SRO Student Handout? Lower Order? l-SRO Ques. N 4 Rev. o Selected for Exam Otigin: New D Past NRC Exam? The plant is operating at 100% power with the "B" Auxiliary Feedwater (AFW) Pump out of service for maintenance.

Then the following events occur:

• Automatic plant trip due to a Steam Generator Tube Rupture (SGTR) on #2 Steam Generator (SG).

• Loss of the RSST and VA-20 at the time of trip.

• Shortly after the trip, a Safety Injection Actuation signal (SIAS) automatically actuated.

• All other plant systems respond as designed.

Which one of the following actions must the US perform during EOP 2525, Standard Post Trip Actions, to mitigate thl3 consequences of this event and what is the reason for this action? Dispatch a PEa to the Hot Shutdown Panel, C-21, to throttle open the #2 Atmospheric Dump Valve. This will permit a cooldown of both Hot Leg temperatures to 515°F. D Direct the BOP to swap the control power supply switch for the Terry Turbine to Facility 1. This will allow the operator to maintain both S/G levels in the prescribed bands. Dispatch a PEa to manually operate the "B" Auxiliary Feedwater Regulating Valve, 2-FW-43B.

This will prevent excessive auxiliary feedwater from overfilling the affected SG. D Direct the BOP to close #2 S/G Steam Supply to the Terry Turbine, MS-202, after the disconnect is closed. This will minimize the radioactive release from the affected SG. Justification I C IS CORRECT; On a loss of normal power, Condensate is lost; therefore, main Feedwater is lost. The loss of VA-20 will cause "B" Aux Feed Regulating Valve to fail open. EOP 2525 requires at least two Auxiliary Feed Pumps to be started. In order to overfeeding

1. 2 SG, the "B" Aux Feed Regulating Valve, 2-FW-43B, must be either closed or isolated A is incorrect; A loss of VA-20 will result in a loss of power to the #2 ADV from ALL remote locations.

The #2 ADV can ONLY operated locally with the handwheel in manual. (A loss of VR-21 will result in the loss of control to the #2 ADV from C-05. The may then be controlled from Plausible because the e>:aminee may think that the FaCility 2 components controlled from Hot Shutdown Panel, C-21 are powered VR-21 or VA-40 and are NOT affected by a loss of "B" is incorrect; Control power supply to the Turbine Driven Auxiliary Feedwater Pump is from DV-20, NOT VA-20; therefore, power supplies will have NO impact on the availability of the Turbine driven Auxiliary Feedwater Plausible because the examinee may not remember that the power supply for the TDAFP is DV-20 NOT "D" is incorrect;

1. 2 S/G Steam Supply to the Terry Turbine, MS-202, will be closed to minimize the release of radioactive steam the Terry Turbine exhaust; however, this action CANNOT be performed in EOP 2525. This action is only performed in EOP 2534, lowering both hot leg temperatures to <515°F, when isolating the affected Plausible because this action will be performed at a later time and for the stated EOP 2525, AOP Comments and Question Modification Bob K. -Bill M. -D-2/C, Angelo -D-4/C; Change "close" to "operate" In correct answer. -NRC KIA SystemtEtA System 038 Steam Generator Tube Rupture (SGTR) KIA Selected 2.1 Conduct of Operations Number 2.1.30 R04.4 SRO 4.0 CFR Link (CFR: 41.7/45.7)

Ability to locate and operate components.

including local controls.

NRC KIA Generic System Page 8 of Printed on 10/28/2009 at 13:20 (No "Parents t, ("

Question #: Question ID: 9000004 RO SRO Student Handout? Lower Order? l-SRO Ques. N 4 Rev. o Selected for Exam Otigin: New D Past NRC Exam? The plant is operating at 100% power with the "B" Auxiliary Feedwater (AFW) Pump out of service for maintenance.

Then the following events occur:

• Automatic plant trip due to a Steam Generator Tube Rupture (SGTR) on #2 Steam Generator (SG).

• Loss of the RSST and VA-20 at the time of trip.

• Shortly after the trip, a Safety Injection Actuation signal (SIAS) automatically actuated.

• All other plant systems respond as designed.

Which one of the following actions must the US perform during EOP 2525, Standard Post Trip Actions, to mitigate thl3 consequences of this event and what is the reason for this action? A Dispatch a PEa to the Hot Shutdown Panel, C-21, to throttle open the #2 Atmospheric Dump Valve. This will permit a cooldown of both Hot Leg temperatures to 515°F. D B Direct the BOP to swap the control power supply switch for the Terry Turbine to Facility 1. This will allow the operator to maintain both S/G levels in the prescribed bands. C Dispatch a PEa to manually operate the "B" Auxiliary Feedwater Regulating Valve, 2-FW-43B.

This will prevent excessive auxiliary feedwater from overfilling the affected SG. D D Direct the BOP to close #2 S/G Steam Supply to the Terry Turbine, MS-202, after the disconnect is closed. This will minimize the radioactive release from the affected SG. Justification I C IS CORRECT; On a loss of normal power, Condensate is lost; therefore, main Feedwater is lost. The loss of VA-20 will cause the "B" Aux Feed Regulating Valve to fail open. EOP 2525 requires at least two Auxiliary Feed Pumps to be started. In order to prevent overfeeding

1. 2 SG, the "B" Aux Feed Regulating Valve, 2-FW-43B, must be either closed or isolated locally. A is incorrect; A loss of VA-20 will result in a loss of power to the #2 ADV from ALL remote locations.

The #2 ADV can ONLY be operated locally with the handwheel in manual. (A loss of VR-21 will result in the loss of control to the #2 ADV from C-05. The ADV may then be controlled from C-21.) Plausible because the e>:aminee may think that the FaCility 2 components controlled from Hot Shutdown Panel, C-21 are powered from VR-21 or VA-40 and are NOT affected by a loss of VA-20" "B" is incorrect; Control power supply to the Turbine Driven Auxiliary Feedwater Pump is from DV-20, NOT VA-20; therefore, swapping power supplies will have NO impact on the availability of the Turbine driven Auxiliary Feedwater Pump. Plausible because the examinee may not remember that the power supply for the TDAFP is DV-20 NOT VA-20 "D" is incorrect;

1. 2 S/G Steam Supply to the Terry Turbine, MS-202, will be closed to minimize the release of radioactive steam from the Terry Turbine exhaust; however, this action CANNOT be performed in EOP 2525. This action is only performed in EOP 2534, after lowering both hot leg temperatures to <515°F, when isolating the affected S/G. Plausible because this action will be performed at a later time and for the stated reason. References]

EOP 2525, AOP 2504D. Comments and Question Modification History Bob K. -D-4/C Bill M. -D-2/C, K Angelo -D-4/C; Change "close" to "operate" In correct answer. -RLC NRC KIA SystemtEtA System 038 Steam Generator Tube Rupture (SGTR) KIA Selected NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.30 R04.4 SRO 4.0 CFR Link (CFR: 41.7/45.7)

Ability to locate and operate components.

including local controls.

Page 8 of 77 Printed on 10/28/2009 at 13:20 Question ID: 9000004 0 RO .". SRO Rev. o Millstone Unit 2 Loss of 120 VAC Vital Instrument Panel

[...] 1.0 PURPOSE Ll Ohjt!f:livt! Selected for Exam AOP 2504D Rc\iliion 003-07 POdge 3 or22 This procedure provides instructions 10 be performed upon the In.'isnf 120 \'nlt AC Vital Instrument Panel VA-Xl. l.2 I)h('u.sslon The los.> of VA-20 cause." the 10."" of Illany almWlciators, indications, FPC inputs, interlocks and Equipment thal affecls Unit operation includes the folklwing:

FW-Sl B, #2 SO fails as is on 10&<; of power, and closes when power is restlll'ed

.. FW-4lB, #2 SO FRV hypas...., closes and control from COS is lost "e" charging pump lo....es the plunger flush PUlllJ' .. If cilalUlel"Y" is co III 1'011 ing pressurizer level, charging will go to maximuill and letdown will go to minimum .. If prcs..<;ul'izer pressure control i!>selected to channel "Y,"auto ctltllrol of prc&<;.uri1..cr sprays are Io!'t .. All pressurizel' heaten. are deenergized due to pressurizer level channel Y sending II pres..\urizeI' level low low heater cutout. '10 restore pl'essuri.lcr healer!'!., the heater select switch must be placed in the channel ,i X" position.

• #2 SO DlJMP MS control fronl C"()S.. C-21 and ('-10 is lost. .. Sleam dump 10 condenser pressure controller (PIC-4216) from C-05 is 101,1. Control from Foxboro computeds available.

.. #2 AUK feed reg valve fails (lllcn and manual t-1.1ntrol from C -05 is lost. '!.Or<U

• • • Question #: 4 Student Handout? Lower Order? I-SRO Ques. # 4 Origin: New Past NRC Exam? Page 9 of 77 Printed on 10/28/2009 at 13:20 '!.Or of VA-20 cause." the 10."" of Illany almWlciators, indications, FPC inputs, interlocks and Equipment thal affecls Unit operation includes the folklwing:

FW-Sl B, #2 SO fails as is on 10&<; of power, and closes when power is restlll'ed

.. FW-4lB, #2 SO FRV hypas. ... , closes and control from COS is lost "e" charging pump lo .... es the plunger flush PUlllJ' .. If cilalUlel"Y" is co III 1'011 ing pressurizer level, charging will go to maximuill and letdown will go to minimum .. If prcs..<;ul'izer pressure control i!>selected to channel "Y,"auto ctltllrol of prc&<;.uri1..cr sprays are Io!'t New .. All pressurizel' heaten. are deenergized due to pressurizer level channel Y sending II pres..\urizeI' level low low heater cutout. '10 restore pl'essuri.lcr healer!'!., the heater select switch must be placed in the channel ,i X" position.

• #2 SO DlJMP MS control fronl C"()S .. C-21 and ('-10 is lost. .. Sleam dump 10 condenser pressure controller (PIC-4216) from C-05 is 101,1. Control from Foxboro computeds available.

.. #2 AUK feed reg valve fails (lllcn and manual t-1.1ntrol from C -05 is lost. Lower Order? Past NRC Exam? Page 9 of 77 Printed on 10/28/2009 at 13:20 Question #: 4 Student Handout? Lower Order? Question ID: 9000004 =J RO SRO I-SRO Ques. # 4 Rev. 0 ." Selected for Exam Orlgin: New Past NRC Exam? EOP 2525 Revision 023 Millstone Unit 2 16 uf26 Standard Post Trip Actions CONTINGENCY ACTIONS 6. (IX)fHinued) CHECK that at leas.t one c.1 RESTORE level 10 40 to 70% steam generator has. BO'nl of in at least one steam generatm' the following conditions mel:

ANY of the Level is 10 to 80%.

• Main feeciw'ater Main feedwBter Of

• Motor-driven auxiliary'ru'o auxiliaty (redwater feedwater pump pumflS are operating lO TDAFW Pump. Refer 1h restore level 40 to 70%. Appendix 6, "TDAFW Pump Normal Starrup." TDAFW Pump. ReferTo Appendix 7, UTDAF\\l Pump Abnormal Startup." CHECK [hal RCS s.ubcooling dJ RESTORE steam. generator is gI'estel' than 01' eq ual h) level 40 to 70o/t,l by performing 30"E ONE of the following: FEED each unaffected steam genenHor greater thall300 gplll. FEED the least affected s.1eal'U generalOf greater [han 300 gpm. Page 10 of Printed on 10/28/2009 at 13:20 Question #: 4 I-SRO Ques. # 4 Question ID: 9000004 =J RO SRO Student Handout? Lower Order? Rev. 0 ." Selected for Exam Orlgin: New Past NRC Exam? Millstone Unit 2 Standard Post Trip Actions EOP 2525 Revision 023 INSTRUCTIONS

6. (IX)fHinued)

c. CHECK that at leas.t one steam generator has. BO'nl of the following conditions mel:

• Level is 10 to 80%.

• Main feedwBter Of 'ru'o auxiliaty (redwater pumflS are operating lO restore level 40 to 70%. d. CHECK [hal RCS s.ubcooling is gI'estel' than 01' eq ual h) 30"E Page 10 of 77 16 uf26 CONTINGENCY ACTIONS c.1 RESTORE level 10 40 to 70% in at least one steam generatm' ANY of the

• Main feeciw'ater

• Motor-driven auxiliary feedwater pump

• TDAFW Pump. Refer 1h Appendix 6, "TDAFW Pump Normal Starrup."

• TDAFW Pump. ReferTo Appendix 7, UTDAF\\l Pump Abnormal Startup." dJ RESTORE steam. generator level 40 to 70o/t,l by performing ONE of the following:

• FEED each unaffected steam genenHor greater thall300 gplll.

• FEED the least affected s.1eal'U generalOf greater [han 300 gpm. Printed on 10/28/2009 at 13:20

Question #: 4 Question ID: 9000004 RO BRO Student Handout? D Lower Order? II-SRO Ques. # 4 Rev. 0 Selected for Exam Origin: New Past NRC Exam? EOP 2525 Revision 023 Page 18 of26 Millstone Unit 2 Standard Post Trip Actions CONTINGENCY ACTIONS (continued)

I, CHECK that NONE of the b.I IF feed is available to BOTH following steam plant radiation steam generators.

monitors have an unexplained TH ROITLE feed to the alarm or indicate an steam generator with the unexplained rise in activity; highest radiation readings to maintain level 40 to 45<X: by ! Stea,lll Plant Radiation performing ANY of the Ii Monitors following:

I RM-S099, Steam Jet Air OPERATE associated Ejector main feed reg bypass RM-4262. SG Blowdown valve, FW-41A t)f FW-41B. RM-4299A and B, Main Steam Line 2) AFAS has actuated, PERFORM the following:RM-4299C.

Main Sleam Line 2

• PLACE the auxiliary feed "OVERRIDE!

MAN/STARr!

RESET" I handswitches in "PULL TO LOCK", OPERATE the associated aux feed ! reg valve, FW-43A or FW-43B. (continue)

Page 11 of Printed on 10/28/2009 at 13:20 : Question #: 4 Question ID: 9000004 RO BRO Student Handout? D Lower Order? II-SRO Ques. # 4 Rev. 0 Selected for Exam Origin: New Past NRC Exam? ! I I ! Millstone Unit 2 Standard Post Trip Actions EOP 2525 Revision 023 "7 I, INSTRUCTIONS (continued)

b. CHECK that NONE of the following steam plant radiation monitors have an unexplained alarm or indicate an unexplained rise in activity; Stea,lll Plant Radiation Monitors

• RM-S099, Steam Jet Air Ejector

• RM-4262. SG Blowdown

• RM-4299A and B, Main Steam Line I

• RM-4299C.

Main Sleam Line 2 (continue)

Page 11 of 77 Page 18 of26 CONTINGENCY ACTIONS b.I IF feed is available to BOTH steam generators.

TH ROITLE feed to the steam generator with the highest radiation readings to maintain level 40 to 45<X: by performing ANY of the Ii following:

1) OPERATE associated main feed reg bypass valve, FW -41A t)f FW-41B. 2) AFAS has actuated, PERFORM the following:

• PLACE the auxiliary feed "OVERRIDE!

MAN/STARr!

RESET" handswitches in "PULL TO LOCK",

• OPERATE the associated aux feed reg valve, FW-43A or FW-43B. (continue)

Printed on 10/28/2009 at 13:20 Question #: ., RO SRO Student Handout? Lower Order? Question ID: 9000004 J-SRO Ques. # 4 Rev. 0 Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 Emergency Operating Procedure Technical Guide EOP 2525, Standard Post Trip Actions Page...11L of 38 Step Number 7 Determine Status of Containment Isolation The intent of the Containment Isolation safety function is to ensure that containment atmospheric conditions are acceptable or that mitigative actions are initiated.

INSTRUCTIONS Containment Isolation serves to ensure that radioactivity is contained inside the containment building.

The acceptance criteria are designed to check that a normal containment environment exists or that the operator is alerted to an off-normal condition.

Containment pressure greater than the maximum expected normal containment pressure, high radiation inside or outside containment, or the steam plant are indications that more than an uncomplicated reactor trip has occurred.

CONTINGENCY ACTIONS If a steam plant radiation monitor is in alarm, steps are provided to secure feedwater to the most affect steam generator.

Since the steam generator level rises due to the leakage from the RGS, adding additional makeup could result in a potential overfill.

Contingency actions are designed to ensure that the containment is isolated when containment pressure reaches the CIAS setpoint.

Mditionally actuation for SIAS, EBFAS, and MSI are checked, along with the CIAS actuation.

Once complete facility of Control Room Air Conditioning (CRAGS) is also checked in service. The CRACS is checked to satisfy the Control Room Habitability Analysis.

JUSTIFICATION FOR DEVIATION The EPG checks containment pressure first, then radiation monitors.

MP2 checks the radiation monitors first, then containment pressure.

The Containment Temperature and Pressure Control and Containment Combustible Gas Control Safety Functions both require the Unit Supervisor to obtain containment pressure.

By placing containment pressure last in this Safety Function, it allo'NS the Unit Supervisor to proceed through the next two Safety Functions without having to query the PPO for this information.

MP2 adds a step to check radiation monitors outside containment This would alert the operator of a LOCA occurring outside containment.

This is also consistent with the guidance in the EPG for a LOCA. MP2 adds a Contingency Action to throttle/isolate feedwater to the most affected steam generator following a SGTR. This action is necessary to prevent a possible overfill of the ruptured steam generator.

Page 12 of 77 Printed on 10/28/2009 at 13:20 Question #: ., Question ID: 9000004 RO SRO J-SRO Ques. # 4 Rev. 0 Selected for Exam Millstone Unit 2 Emergency Operating Procedure Technical Guide Student Handout? Lower Order? Origin: New Past NRC Exam? EOP 2525, Standard Post Trip Actions Page...11L of 38 Step Number 7 Determine Status of Containment Isolation The intent of the Containment Isolation safety function is to ensure that containment atmospheric conditions are acceptable or that mitigative actions are initiated.

INSTRUCTIONS Containment Isolation serves to ensure that radioactivity is contained inside the containment building.

The acceptance criteria are designed to check that a normal containment environment exists or that the operator is alerted to an off-normal condition.

Containment pressure greater than the maximum expected normal containment pressure, high radiation inside or outside containment, or the steam plant are indications that more than an uncomplicated reactor trip has occurred.

CONTINGENCY ACTIONS If a steam plant radiation monitor is in alarm, steps are provided to secure feedwater to the most affect steam generator.

Since the steam generator level rises due to the leakage from the RGS, adding additional makeup could result in a potential overfill.

Contingency actions are designed to ensure that the containment is isolated when containment pressure reaches the CIAS setpoint.

Mditionally actuation for SIAS, EBFAS, and MSI are checked, along with the CIAS actuation.

Once complete facility of Control Room Air Conditioning (CRAGS) is also checked in service. The CRACS is checked to satisfy the Control Room Habitability Analysis.

JUSTIFICATION FOR DEVIATION The EPG checks containment pressure first, then radiation monitors.

MP2 checks the radiation monitors first, then containment pressure.

The Containment Temperature and Pressure Control and Containment Combustible Gas Control Safety Functions both require the Unit Supervisor to obtain containment pressure.

By placing containment pressure last in this Safety Function, it allo'NS the Unit Supervisor to proceed through the next two Safety Functions without having to query the PPO for this information.

MP2 adds a step to check radiation monitors outside containment This would alert the operator of a LOCA occurring outside containment.

This is also consistent with the guidance in the EPG for a LOCA. MP2 adds a Contingency Action to throttle/isolate feedwater to the most affected steam generator following a SGTR. This action is necessary to prevent a possible overfill of the ruptured steam generator.

Page 12 of 77 Printed on 10/28/2009 at 13:20 Question #: 5 Question ID: 9082581 RO .". SRO C Student Handout? ii'l Lower Order? J-SRO Ques. # 5 Rev. 0 ii'l Selected for Exam Origin: New Past NRC Exam? The plant has experienced a loss of VA-10 while in Mode 5 with Shutdown Cooling in operation.

Assuming RBCCW flow was NOT diverted from the SOC Heat Exchangers by any other system, which of the following actions would be performed outside the Control Room and what is the reason for performing these actions in the listed order? 1. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the DA counterclockwise direction as directed by the Control Room. 2. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2-SI-306, SOC Total Flow Control Valve, must be opened first to provide minimum flow for the operating LPSI Pump. 1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the ii'lB clockwise direction as directed by the Control Room. 2. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-657, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rate. 1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheelln the DC countorclockwise direction as directed by the Control Room. 2. Place 2-SI**306, SOC Total Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2-SI-657, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rate. D 1. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-306, SOC Total Flow Control Valve, must be opened first to provide minimum flow for the operating LPSI Pump. *Justification I B IS CORRECT; AOP 2572, Loss of SOC, requires the PEO to first place SI-657 in local manual control to restore cooling to the SI-657 is a reverse operating valve (I.e., clockwise rotation is open, counterclockwise is close). On a loss of power (VA-10), fails closed; thBrefore, SI-657 must be rotated in the clockwise direction to open it. SI-306 is also a reverse operated valve. fails open on a loss of power (VA-1O), and must be rotated in the counterclockwise direction to throttle it closed. SI-306 is last to allow more fiow through the SOC Heat Exchangers, if necessary to provide additional cooling flow to the A is incorrect; The order of local valve operations is incorrect.

Although SI-657 fails closed, SI-306 fails open; therefore, there is need to establish minimum flow protection for the running LPSI Plausible because the examinee may think that it is more important to initiate flow through the heat exchanger bypass (Total Control valve) than to initiate flow through the SOC Heat Exchanger.

This may allow for a more controlled initiation of the C is incorrect; The order of valve operations is correct; however, the direction of valve rotation is incorrect.

Both valves are Plausible because the examinee may not remember that both valves are reverse o is incorrect.

The order of local valve operations is incorrect and the direction of valve rotation is Plausible because the examinee may think that this sequence allows for more control of the cooldown.

Additionally, the examinee not remember that both of these valves are reverse References I AOP 2572 Loss of SOC, section 8.0 . Comments and Question Modification Bob K. 3/W (memory on reverse Bill M. 3/C, Angelo 3/C; No Page 13 of Printed on 10128/2009 at 13:20 Question #: 5 Question ID: 9082581 RO .". SRO C Student Handout? ii'l Lower Order? J-SRO Ques. # 5 Rev. 0 ii'l Selected for Exam Origin: New Past NRC Exam? The plant has experienced a loss of VA-10 while in Mode 5 with Shutdown Cooling in operation.

Assuming RBCCW flow was NOT diverted from the SOC Heat Exchangers by any other system, which of the following actions would be performed outside the Control Room and what is the reason for performing these actions in the listed order? DA ii'lB DC 1. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2-SI-306, SOC Total Flow Control Valve, must be opened first to provide minimum flow for the operating LPSI Pump. 1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-657, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rate. 1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheelln the countorclockwise direction as directed by the Control Room. 2. Place 2-SI**306, SOC Total Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2-SI-657, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rate. D D 1. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-306, SOC Total Flow Control Valve, must be opened first to provide minimum flow for the operating LPSI Pump. *Justification I B IS CORRECT; AOP 2572, Loss of SOC, requires the PEO to first place SI-657 in local manual control to restore cooling to the RCS. SI-657 is a reverse operating valve (I.e., clockwise rotation is open, counterclockwise is close). On a loss of power (VA-10), SI-657 fails closed; thBrefore, SI-657 must be rotated in the clockwise direction to open it. SI-306 is also a reverse operated valve. SI-306 fails open on a loss of power (VA-1O), and must be rotated in the counterclockwise direction to throttle it closed. SI-306 is operated last to allow more fiow through the SOC Heat Exchangers, if necessary to provide additional cooling flow to the RCS. A is incorrect; The order of local valve operations is incorrect.

Although SI-657 fails closed, SI-306 fails open; therefore, there is no need to establish minimum flow protection for the running LPSI Pump. Plausible because the examinee may think that it is more important to initiate flow through the heat exchanger bypass (Total flow Control valve) than to initiate flow through the SOC Heat Exchanger.

This may allow for a more controlled initiation of the cooldown.

C is incorrect; The order of valve operations is correct; however, the direction of valve rotation is incorrect.

Both valves are reverse operating.

Plausible because the examinee may not remember that both valves are reverse operated.

o is incorrect.

The order of local valve operations is incorrect and the direction of valve rotation is incorrect.

Plausible because the examinee may think that this sequence allows for more control of the cooldown.

Additionally, the examinee may not remember that both of these valves are reverse operated.

References I AOP 2572 Loss of SOC, section 8.0 . Comments and Question Modification History Bob K. 3/W (memory on reverse acting) Bill M. 3/C, 50/50 Angelo 3/C; No comments.

Page 13 of 77 Printed on 10128/2009 at 13:20 Question #: 5 Student Handout? lower Order?Question ID: 9082581 RO SRO I-SRO Ques. # 5 Rev. 0 Selected for Exam Otfgin: New Past NRC Exam? NRC KIA System/E/A System 057 Loss of Vital AC Electrical Instrument Bus Generic KIA S ....t **1"I System 2.4 Emergency Procedures

/Plan NRC KIA Generic Number 2.4.35 RO 3.8 SRO 4.0 CFRlink (CFR:41.10/43.5/45.13Knowledge of local auxiliary operator tasks during an emergency and the resultant operational Page 14 of 77 Printed on 10/28/2009 at 13:20 Question #: 5 Question ID: 9082581 RO SRO Student Handout? lower Order? I-SRO Ques. # 5 Rev. 0 Selected for Exam Otfgin: New Past NRC Exam? NRC KIA System/E/A System 057 Loss of Vital AC Electrical Instrument Bus Generic KIA S .. .. t ** 1"I NRC KIA Generic System 2.4 Emergency Procedures IPlan Number 2.4.35 RO 3.8 SRO 4.0 CFR link (CFR: 41.10 143.5/45.13)

Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. Page 14 of 77 Printed on 10/28/2009 at 13:20 Question #: 5 Student Handout? Lower Order?Question/D:

9082581 DRO SRO IJ-SRO Ques. 1/ 5 Rev. 0 Selected for Exam Origin: New D Past NRC Exam? Millstone Unit 2 AOP 1572 Re\"ision 009-04 Loss of Shutdown Cooling Puge 45 of 70 loss of Pom.'r or Air hi SI-657. SI*306 ur INSTRUCTIONS CONTINGENCY NOTE Lo&.<;of power or ail' [(1 SDC 11m" control valves ha<; the follm\>lng affect: 51-657 failscinsed SI*J06 faiL\ npen to ils limit stop (mid-Jx)sition) Loss of VA*lO fail!> both valves. SI*306 can be throltled from its maximum open position (maximum flow lim it stop posilion) when diverting additional flow through SIX.' heat exchange!'s is required. Obtaining reference positions of SDC flow control valve1\. may be helpful during SOC restoration.

If lilossof VA-lO has occurred, the reference positions are onlv available as archive data in the PPC (FPC analog pomts 2S1657 and tSI306}. OBSERVE applit"ablc cnntl'OlIcl's 01' FPC analog points 10 obtain a reference position f*ol' SDC flow Cl,)ntrol valves: Output of FlC*306 01' archive PPC data fm 251306 Output of HIC*365701' archive PPC data fm 251657 for Ihe failed valvels),ADJUST Ihe (lOnll'oJIcI' output In match actual vaN" position. FIC-30t) HIC-3657 Page 15 of Printed on 10/28/2009 at 13:20 Question #: 5 I Question/D:

9082581 DRO SRO Student Handout? IJ-SRO Ques. 1/ 5 Rev. 0 Selected for Exam Origin: New Millstone Unit 2 AOP 1572 Re\"ision 009-04 Loss of Shutdown Cooling Puge 45 of 70 8.0 loss of Pom.'r or Air hi SI-657. SI *306 ur Bufh INSTRUCTIONS CONTINGENCY ACTIONS NOTE L Lo&.<;of power or ail' [(1 SDC 11m" control valves ha<; the follm\>lng affect:

• 51-657 failscinsed

• SI*J06 faiL\ npen to ils limit stop (mid-Jx)sition)

1. Loss of VA*lO fail!> both valves. 3. SI*306 can be throltled from its maximum open position (maximum flow lim it stop posilion) when diverting additional flow through SIX.' heat exchange!'s is required.

4. Obtaining reference positions of SDC flow control valve1\. may be helpful during SOC restoration.

If lilossof VA-lO has occurred, the reference positions are onlv available as archive data in the PPC (FPC analog pomts 2S1657 and tSI306}. 8.1 OBSERVE applit"ablc cnntl'OlIcl's 01' FPC analog points 10 obtain a reference position f*ol' SDC flow Cl,)ntrol valves:

• Output of FlC*306 01' archive PPC data fm 251306

• Output of HIC*365701' archive PPC data fm 251657 for Ihe failed valvels),ADJUST Ihe (lOnll'oJIcI' output In match actual vaN" position.

• FIC-30t)

• HIC-3657 Lower Order? D Past NRC Exam? Page 15 of 77 Printed on 10/28/2009 at 13:20 SRO Exam Questions Only"(No "Parents" Or Question #: 5 Question ID: 9082581 D RO Ii'l SRO D Student Handout? Ii'l Lower Order? Il-SRO Que" If 5 Rev. 0 Ii'l Selected for Exam Origin: New D Past NRC Exam? Millstone Unit 2 Loss of Shutdown Cooling AOP 2572 Re"ision 009-04 Page 490f70 INSTRUCTIONS CONTINGENCY A(TIONS 'V' CAUTION 'V' Cale should be u..<;ed when establishing SOC heat exchanger tlowdue tn the lxHcntial1 for water in the SDC heat exchanger to be much cooler than temperature.

Initiating flow slowly alk)\\ls temperatures to equalize.

NOTE 1.51-657 it, a reverse-operating valvc; counterclockwise rotation of the handwlll:el closes the valve and clockwi.\e rotation of IllI: handwheel opens the valve. 2. When establishing RCS cooldowll

!'ate, optimum temperature respnrt.\C is adlie\ed by maintaining 51-657 between 35 and 60%, open.

If 51.-657 had a loss of Ix)wer or air, PERFORM the following:

a. csmblishillg cooldown, Refer 10 SP 2602B.

'Iempemmre, Pre&'iI.lt'e Verification," !lUG PERFORM the following:

• MONITOR RCS cooldown rale usingT351Y.

• ENSURE syslem response is within cooldmvn

h. PERFORM the fnllo\\ling to take manual control of 51-657 ('W' ESF Room): 1) UNLOCK thc manual handwhcel on valve. 2) CLOSE instrument air supply valve and VENT valve operatOl'.

J) LOOSEN stem hex nut a; rcquired to allow slcm mOltement.

4) ROTATE 51-657 handwheel as directed bv the COillrol Room. -(continue)

Page 16 of 77 Printed on 10/28/2009 at 13:20 SRO Exam Questions Only"(No "Parents" Or Question #: 5 Question ID: 9082581 D RO Ii'l SRO D Student Handout? Il-SRO Que" If 5 Rev. 0 Ii'l Selected for Exam Origin: New Millstone Unit 2 Loss of Shutdown Cooling AOP 2572 Re"ision 009-04 Page 490f70 INSTRUCTIONS CONTINGENCY A(TIONS 'V' CAUTION 'V' Cale should be u..<;ed when establishing SOC heat exchanger tlowdue tn the lxHcntial1 for water in the SDC heat exchanger to be much cooler than temperature.

Initiating flow slowly alk)\\ls temperatures to equalize.

NOTE 1.51-657 it, a reverse-operating valvc; counterclockwise rotation of the handwlll:el closes the valve and clockwi.\e rotation of IllI: handwheel opens the valve. 2. When establishing RCS cooldowll

!'ate, optimum temperature respnrt.\C is adlie\ed by maintaining 51-657 between 35 and 60%, open.

If 51.-657 had a loss of Ix)wer or air, PERFORM the following:

a. csmblishillg cooldown, Refer 10 SP 2602B.

'Iempemmre, Pre&'iI.lt'e Verification," !lUG PERFORM the following:

• MONITOR RCS cooldown rale usingT351Y.

• ENSURE syslem response is within cooldmvn

h. PERFORM the fnllo\\ling to take manual control of 51-657 ('W' ESF Room): 1) UNLOCK thc manual handwhcel on valve. 2) CLOSE instrument air supply valve and VENT valve operatOl'.

J) LOOSEN stem hex nut a; rcquired to allow slcm mOltement.

4) ROTATE 51-657 handwheel as directed bv the COillrol Room. -(continue)

Ii'l Lower Order? D Past NRC Exam? Page 16 of 77 Printed on 10/28/2009 at 13:20 i Question #: 5 .. Student Handout? Question ID: 9082581 RO [;f] Lower Order? I-SRO Ques. # 5 Rev. o [;fi Selected for Exam Origin: New Past NRC Exam? ! l\'IiUstone Unit Loss of Shutdown AOP 2572 Rc\"ision Puge 52 or CONTINGENCY At'TIQNS NOTE Sl-JlXJ i!> a reverse-operating valve; cmmlerc/ockwisf' rotation of the hal1dwhee.-I closes the va.lve and clockwise rot'1I10n ofthe handwheel opens the valve. A'io n.cceS!ioary.

ESTABLISH manual cOlHnil of SI-306 11'; ES'I:A.llLlSI I

between operators at Ihe (uP:' nSF Room) and CLo..l.)E instrument air SI-306, SIX: total now OPEN pNcnc:k on ins[J"ument pressure regulator VENT SI: *306, SDC total ,:ontrol, valve UNLOCK and REMOVE chain :from manual handy.heeL ROTATE manual handwheel ALlGN holes in outer shaft hole in inner 1 N5nRT lhe pin into shaft holes. ENSURE SOC total l:ontml. valve position on the manual actuator is throuJed open 1£desired 10 manually valve, POSITION handwheel as directed by Control _8.10 WHEN recovery from manual operations is desired, PERFORM HC'tioll!io specified by the SM/US. Page 17 of Printed on 10/28/2009 at 13:20 .. Question #: 5 Question ID: 9082581 RO SRO Student Handout? [;f] Lower Order? I-SRO Ques. # 5 Rev. o [;fi Selected for Exam Origin: i ! l\'IiUstone Unit 2 Loss of Shutdown Cooling INSTRUCTIONS AOP 2572 Rc\"ision 009-04 Puge 52 or 70 CONTINGENCY At'TIQNS G NOTE Sl-JlXJ i!> a reverse-operating valve; cmmlerc/ockwisf' rotation of the hal1dwhee. -I closes the va.lve and clockwise rot'1I10n ofthe handwheelopens the valve. &.9 A'io n.cceS!ioary.

ESTABLISH local manual cOlHnil of SI-306 11'; follows: II. ES'I:A.llLlSI I communications between operators at Ihe valve (uP:' nSF Room) and Control Room. b. CLo..l.)E instrument air supplyfO!'

SI-306, SIX: total now control. c. OPEN pNcnc:k on ins[J"ument air pressure regulator and VENT SI: *306, SDC total flow ,:ontrol, valve operator.

d. UNLOCK and REMOVE chain :from manual handy.heeL

c. ROTATE manual handwheel and ALlGN holes in outer shaft with hole in inner shaft. f. 1 N5nRT lhe pin into shaft holes. g. ENSURE SOC total Oo,,\-' l:ontml. valve position indic.alor on the manual actuator is at throuJed open position.

h. 1£ desired 10 manually operate valve, POSITION 81-306 handwheel as directed by the Control Room. _8.10 WHEN recovery from manual operations is desired, PERFORM HC'tioll!io specified by the SM/US. Page 17 of 77 New Past NRC Exam? Printed on 10/28/2009 at 13:20 Question #: Lower I-SRO Que,. If 6 Rev. 2 Selected for Exam Origin: Bank Past NRC Question ID: 3100002 RO SRO Student Handout? The plant is operating at 100% power when the Balance of Plant (BOP) operator reports that Instrument Air header pressure is at 95 psig and lowering.

Immediately following, the Turbine Building PEO reports a large unisolable leak just downstream of the "0" Instrument Air Dryer After Filters. Assuming Instrument Air header pressure continues to lower, at what pressure in the Instrument Air System must the Unit Supervisor (US) direct a manual reactor trip (by procedure) and why? Prior to reaching 85 psig. When pressure drops below 85 psig the crew is procedurally directed to crosstie Station Air with Unit 3. Operation in tllis alignment will result in all components supplied by Instrument Air being inoperable, which is an unanalyzed condition. When pressure lowers to less than 85 psig. At approximatley 85 psig the Instrument Air/Station Air Crosstie valve opens. Continued operation with Station Air supplied to valves and controllers will result in erratic operation of components due to the high moisture content of Station Air. When pressure lowers to less than 80 psig. The loss of many important controls.

such as Feedwater.

could degrade plant conditions at the time of the trip; therefore, the reactor must be tripped when control of important systems could become challel1ged. Prior to reaching 80 psig. The Auxiliary Feed Regulating Valves will lock up with less than 80 psig supply pressure.

The reactor must be tripped to allow the initial automatic opening of these valves and begin feeding Steam Generators.

Justification I C IS CORRECT; AOP 2563, Discussion section 1.2. When IA pressure lowers to less than 80 psig, the Feed Regulating Valves may lock up resulting in over feeding of Steam Generators after the trip. Additionally, the Steam Dumps may not open resulting in opening of the Main Steam Safeties as the only initial means of removing decay heat. A is incorrect; Although it is less desirable to operate with Unit 2 cross tied with Unit 3, there are NO restrictions; therefore, requirements to Plausible because the examinee may feel that continued operation with Station Air supplied by Unit 3 (and, subsequently Station crosstied to Instrument Air) is NOT B is incorrect; Although the Instrument Air/Station Air Crosstie valve automatically opens at -85 psig. continued operation with Air cross tied to Instrument Air is Plausible becaJse the examinee may remember that the Station air Cross Tie valve opens. he/she may think that continued with Station Air supplying Instrument Air is NOT D is incorrect; The Auxiliary Feed Regulating Valves have back up air that will ensure their operation for a limited duration even during complete loss of Instrument Plausible becaJse the examinee may feel that the potential loss of Auxiliary Feedwater requires an immediate Reactor I AOP 2563, Loss of Instrument Air. Section 1.2 Comments and Question Modification Sob K. -D-2/C (only need to know Changed pressure values on "A", "6" and "D" and reworded "An and "6" slightly.

-Sill M. -D-2/C, K (90 psig is too Changed values in distractors "A" and "S" to 85 pslg vs 90 psig. More plausible

.* Angelo -D-3/C; No NRC KIA System/E/A System 065 Loss of Instrument Air Number AA2.06 RO 3.6* SRO 4.2 CFR Link (CFR: 43.5/45.13) Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to trip reactor if instrument air pressure is decreasing Page 18 of Printed on 10/28/2009 at 13:20 Question #: Question ID: 3100002 RO SRO Student Handout? Lower Order? I-SRO Que,. If 6 Rev. 2 Selected for Exam Origin: Bank Past NRC Exam? The plant is operating at 100% power when the Balance of Plant (BOP) operator reports that Instrument Air header pressure is at 95 psig and lowering.

Immediately following, the Turbine Building PEO reports a large unisolable leak just downstream of the "0" Instrument Air Dryer After Filters. Assuming Instrument Air header pressure continues to lower, at what pressure in the Instrument Air System must the Unit Supervisor (US) direct a manual reactor trip (by procedure) and why? A Prior to reaching 85 psig. When pressure drops below 85 psig the crew is procedurally directed to crosstie Station Air with Unit 3. Operation in tllis alignment will result in all components supplied by Instrument Air being inoperable, which is an unanalyzed condition.

B When pressure lowers to less than 85 psig. At approximatley 85 psig the Instrument Air/Station Air Crosstie valve opens. Continued operation with Station Air supplied to valves and controllers will result in erratic operation of components due to the high moisture content of Station Air. When pressure lowers to less than 80 psig. The loss of many important controls.

such as Feedwater.

could degrade plant conditions at the time of the trip; therefore, the reactor must be tripped when control of important systems could become challel1ged.

D Prior to reaching 80 psig. The Auxiliary Feed Regulating Valves will lock up with less than 80 psig supply pressure.

The reactor must be tripped to allow the initial automatic opening of these valves and begin feeding Steam Generators.

Justification I C IS CORRECT; AOP 2563, Discussion section 1.2. When IA pressure lowers to less than 80 psig, the Feed Regulating Valves may lock up resulting in over feeding of Steam Generators after the trip. Additionally, the Steam Dumps may not open resulting in opening of the Main Steam Safeties as the only initial means of removing decay heat. A is incorrect; Although it is less desirable to operate with Unit 2 cross tied with Unit 3, there are NO restrictions; therefore, NO requirements to trip. Plausible because the examinee may feel that continued operation with Station Air supplied by Unit 3 (and, subsequently Station Air crosstied to Instrument Air) is NOT allowed. B is incorrect; Although the Instrument Air/Station Air Crosstie valve automatically opens at -85 psig. continued operation with Station Air cross tied to Instrument Air is acceptable.

Plausible becaJse the examinee may remember that the Station air Cross Tie valve opens. he/she may think that continued operation with Station Air supplying Instrument Air is NOT allowed. D is incorrect; The Auxiliary Feed Regulating Valves have back up air that will ensure their operation for a limited duration even during a complete loss of Instrument Air. Plausible becaJse the examinee may feel that the potential loss of Auxiliary Feedwater requires an immediate Reactor trip. References I AOP 2563, Loss of Instrument Air. Section 1.2 Comments and Question Modification History Sob K. -D-2/C (only need to know "80#") Changed pressure values on "A", "6" and "D" and reworded "An and "6" slightly.

-RLC Sill M. -D-2/C, K (90 psig is too high.) Changed values in distractors "A" and "S" to 85 pslg vs 90 psig. More plausible

.* RJA Angelo -D-3/C; No comments.

NRC KIA System/E/A System 065 Loss of Instrument Air Number AA2.06 RO 3.6* SRO 4.2 CFR Link (CFR: 43.5/45.13)

Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to trip reactor if instrument air pressure is decreasing Page 18 of 77 Printed on 10/28/2009 at 13:20

...** ...

Question #: 6 RO SRO Student Handout? Lower Order? Question ID: 3100002 II-SRO Ques. # 6 Rev. 2 [Yi. Selected for Exam Origin; Sank o Past NRC Exam? Millstone Unit 2 AOP 2563 Revision 009-07 Loss of Instrument Air Page 3 of 32 1.0 PURPOSE 1.1 o bjf'Ctive This procedure provides the operator with specific steps to be taken following a significant drop in instrument air pressure and actions to mitigate the effects of a reactor trip when the I nstrument Air System is flot available.

1.2 Discussion

This procedure is implemented when instrument air pressure is lowering below normal values. With a complete loss of instrument air, continued steady state plant operation is flot possible.

TIle loss of many important controls, such as in the fcedwater system, could degrade plant conditions at the time of a reactor or turbine trip. Therefore.

the reactor is tripped immediately when instrument air pressure Imve:Ts to the point where control of important systems is questionable.

This may be indicated by system response or instrument air header pressure of less than 80 psig. Should instrument air header pressure drop suddenly, as in the case of (l main header rupture. the only initial means of decay and sensible heat removal is the main steam safeties.

Subsequent1y, manual control of the atmospheric dump valves and usc of the Auxiliary Fcedwater System can mitigate the pressure transient, thereby removing sensible and decay heat from the ReS in a controJled manner. Page 19 of 77 Printed on 10/28/2009 at 13:20 Question #: 6 Question ID: 3100002 II-SRO Ques. # 6 Rev. 2 RO SRO [Yi. Selected for Exam Student Handout? Origin; Sank Lower Order? o Past NRC Exam? Millstone Unit 2 Loss of Instrument Air AOP 2563 Revision 009-07 Page 3 of 32 1.0 PURPOSE 1.1 o bjf'Ctive This procedure provides the operator with specific steps to be taken following a significant drop in instrument air pressure and actions to mitigate the effects of a reactor trip when the I nstrument Air System is flot available.

1.2 Discussion

This procedure is implemented when instrument air pressure is lowering below normal values. With a complete loss of instrument air, continued steady state plant operation is flot possible.

TIle loss of many important controls, such as in the fcedwater system, could degrade plant conditions at the time of a reactor or turbine trip. Therefore.

the reactor is tripped immediately when instrument air pressure Imve:Ts to the point where control of important systems is questionable.

This may be indicated by system response or instrument air header pressure of less than 80 psig. Should instrument air header pressure drop suddenly, as in the case of (l main header rupture. the only initial means of decay and sensible heat removal is the main steam safeties.

Subsequent1y, manual control of the atmospheric dump valves and usc of the Auxiliary Fcedwater System can mitigate the pressure transient, thereby removing sensible and decay heat from the ReS in a controJled manner. Page 19 of 77 Printed on 10/28/2009 at 13:20 Question #: 7 Question ID: 9000020 RO SRO Student Handout? Lower Order? l-SRO Ques. # 7 Rev. 0 Selected for Exam Origin: New Past NRC Exam? The reactor is at 100% power with the CEA Motion surveillance in progress.

When Group 7 CEA #1 is tested. CEAPDS indicates it inserts two steps. then slips an additional 20 steps. The appropriate actions were taken to stabilize RCS temperature and the following conditions were observed:

.. Reactor power stable at -99%.. Only Upper Electrical Limit lights are energized on the core mimic .. CEA #1 indicates 158 steps withdrawn on CEAPDS .. CEA #1 indicates 178 steps withdrawn on the PPC .. CEA Motion Inhibit (CMI) alarms on .. CEAPDS Group Deviation indication for CEA Fifty (50) minutes after CEA #1 slipped, all required actions per AOP-2556, CEDS Malfunctions, have completed including plant power Also, I&C reports the circuit malfunction that caused CEA #1 to slip has been repaired and the CEA can be Which one of the following describes actions that must be taken to recover CEA #1 and what is administrative concern of those Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA reCOVE!ry.

CEA #1 Pulse Count Indication and the CMI will be INOPERABLE while the CEA is being recovered. CEA #1 Upper Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Reed Switch Indication for CEA #1 and the CMI will be INOPERABLE while the CEA is being recovered.

D Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered . .., CEA #1 Upper Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered.

Justification I D -CORRECT; CMI is triggered based on the CEA #1 deviation from the other CEAs in Group 7, therefore it must be bypassed recover the CEA. When the CMI is bypassed, it is considered INOPERABLE.

Also, the UEL reed switch indicates it is stuck on. failure is not considered a failure of the CEA Indication System, but does require additional action be taken to recover the A -WRONG; CEA pulse counting indication is no longer accurate for CEA #1, so it is reset to the actual slipped rod position (based reed switches) before CEA recovery is attempted.

This does not make the pulse counting indication Plausible; Examinee may think the reed switch input to the Core Mimic also triggers the Upper Core Stop on the PPC, because Mimic reed switches reset the PPC pulse counts and the UCS is armed under the existing B -WRONG; These are the correct interlocks/inhibits that must be bypass to recover the CEA. However, bypassing the reed input to the UEL does not make it Plausible; Examinee may think bypaSSing UEL interlock has same impact as bypassing CMI interlock, especially where I&C "must lift lead" in the field to bypass the UEL C -WRONG; The CMI and the UEL interlock must be bypassed to withdraw the slipped Plausible; Examinee may think bypassing the CMI, which is based on reed switch input, bypasses all interlocks based on reed input (including the reed switches that drive the References AOP 2556, Pages 3, 13 and Comments and Question Modification History Bob K. D-4/C (Add to stem that all required actions of AOP-2556 have been completed up to CEA Reworded stem per comment -Bill M. -D-3/C, SO/50 (Need to reword last bullet. Meaning is not Split last bullet to specify alarms on C*04 and CEAPDS indication for CEA #1 .* Angelo -D-5/W; Difficult but Page 20 of Printed on 10/28/2009 at 13:20 Question #: 7 Question ID: 9000020 RO SRO Student Handout? Lower Order? l-SRO Ques. # 7 Rev. 0 Selected for Exam Origin: New Past NRC Exam? The reactor is at 100% power with the CEA Motion surveillance in progress.

When Group 7 CEA #1 is tested. CEAPDS indicates it inserts two steps. then slips an additional 20 steps. The appropriate actions were taken to stabilize RCS temperature and the following conditions were observed:

.. Reactor power stable at -99% . .. Only Upper Electrical Limit lights are energized on the core mimic . .. CEA #1 indicates 158 steps withdrawn on CEAPDS . .. CEA #1 indicates 178 steps withdrawn on the PPC . .. CEA Motion Inhibit (CMI) alarms on C-04 .. CEAPDS Group Deviation indication for CEA #1 Fifty (50) minutes after CEA #1 slipped, all required actions per AOP-2556, CEDS Malfunctions, have been completed including plant power changes. Also, I&C reports the circuit malfunction that caused CEA #1 to slip has been repaired and the CEA can now be recovered.

Which one of the following describes actions that must be taken to recover CEA #1 and what is the administrative concern of those actions? A Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA reCOVE!ry.

CEA #1 Pulse Count Indication and the CMI will be INOPERABLE while the CEA is being recovered.

B CEA #1 Upper Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Reed Switch Indication for CEA #1 and the CMI will be INOPERABLE while the CEA is being recovered.

D C Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered . . ., D CEA #1 Upper Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered.

Justification I D -CORRECT; CMI is triggered based on the CEA #1 deviation from the other CEAs in Group 7, therefore it must be bypassed to recover the CEA. When the CMI is bypassed, it is considered INOPERABLE.

Also, the UEL reed switch indicates it is stuck on. This failure is not considered a failure of the CEA Indication System, but does require additional action be taken to recover the CEA. A -WRONG; CEA pulse counting indication is no longer accurate for CEA #1, so it is reset to the actual slipped rod position (based on reed switches) before CEA recovery is attempted.

This does not make the pulse counting indication INOPERABLE.

Plausible; Examinee may think the reed switch input to the Core Mimic also triggers the Upper Core Stop on the PPC, because Core Mimic reed switches reset the PPC pulse counts and the UCS is armed under the existing conditions. B -WRONG; These are the correct interlocks/inhibits that must be bypass to recover the CEA. However, bypassing the reed switch input to the UEL does not make it INOPERABLE.

Plausible; Examinee may think bypaSSing UEL interlock has same impact as bypassing CMI interlock, especially where I&C "must lift a lead" in the field to bypass the UEL interlock. C -WRONG; The CMI and the UEL interlock must be bypassed to withdraw the slipped CEA. Plausible; Examinee may think bypassing the CMI, which is based on reed switch input, bypasses all interlocks based on reed switch input (including the reed switches that drive the UEL). References I AOP 2556, Pages 3, 13 and 14 Comments and Question Modification History I Bob K. D-4/C (Add to stem that all required actions of AOP-2556 have been completed up to CEA recovery.)

Reworded stem per comment -RLC Bill M. -D-3/C, SO/50 (Need to reword last bullet. Meaning is not clear.) Split last bullet to specify alarms on C*04 and CEAPDS indication for CEA #1 .* RJA Angelo -D-5/W; Difficult but fair. Page 20 of 77 Printed on 10/28/2009 at 13:20 Question #: 7 Question ID; 9000020 RO BRO Student Handout? Lower Order? I-SRO Que".*11 7 Rev. 0 -'. Selected for Exam Origin; New Past NRC Exam? NRC KIA System/E/A System 003 Dropped Control Rod Number AA2.04 RO 3.4* SRO 3.6* CFR Link (CFR: 43.5 f 45.13) Ability to determine and interpret the following as they apply to the Dropped Control Rod: Rod motion stops due to dropped rod Millstone Unit 2 AOP 2556 Revision 016-10 CEA Malfunctions Page 3 of 55 1.0 PURPOSE U This AOP contains EOP relat.ed

1.1 Objective

procedure provides instructions for the foUo\\ing malfunctions which could affect CEAs, CEDS, ACTM or CEA position indications: Multiple misaligned or untrippable CEAs Misaligned CEA misaligned greater than ]() steps Inoperable eEA Position Indication System Inoperable CMI circuit Trippablc eEA Untrippablc CEA 1.2 J)jsclIssion FoHowing a CEA drop, operator action should be directed tnward returning the plant to a stable condition.

At high power levels, if no action is taken following the CEA drop, reactor power will return to approximately the initial power level. but at a reduced core average temperature (due to positive reactivity feedback from the negative moderator tcmpt'Taturc coefficient).

The following actions will minimize the affects of the CEA drop transient. Dropped CEAs result in reactof and turbine power mismatch.

This mismatch is nulled by reducing turbine load to match reactor power (RCS temperatures not changing). Dropped CEAs also result in undesirable neutron nux patterns. correct operator R'Sponse, the time span over which these pattern.'i cxi'it is minimized. It is desirable to record as much data as pos.<;ible concerning abnormal flux patterns existing during and subsequent to rod drop. Proper use of PPC, as outlined in this procedure, produces data necessary fOf subsequent analysis r _____

_____________________________

_ I If during Cf':::A motion. a core mimic light fails to clear or light and all other : : indications Hre normal. it is acceptable to request I&C Department to temporarily I : lift leads for respective CEA upper electrical limit reed switch until it is aligned. : I This is not considered a CEA position indication problem. I -----------------------------------------------_Level of Use STOP THINK ACT REVIEW Continuous Page 21 of Printed on 10/28/2009 at 13:20 Question #: 7 Question ID; 9000020 RO BRO Student Handout? Lower Order? I-SRO Que".*11 7 Rev. 0 -'. Selected for Exam Origin; New Past NRC Exam? NRC KIA System/E/A System 003 Dropped Control Rod Number AA2.04 RO 3.4* SRO 3.6* CFR Link (CFR: 43.5 f 45.13) Ability to determine and interpret the following as they apply to the Dropped Control Rod: Rod motion stops due to dropped rod Millstone Unit 2 CEA Malfunctions AOP 2556 Revision 016-10 Page 3 of 55 1.0 PURPOSE U This AOP contains EOP relat.ed 1 .1 Objective procedure provides instructions for the foUo\\ing malfunctions which could affect CEAs, CEDS, ACTM or CEA position indications:

• Multiple misaligned or untrippable CEAs

• Misaligned CEA misaligned greater than ]() steps

• Inoperable eEA Position Indication System

• Inoperable CMI circuit

• Trippablc eEA

• Untrippablc CEA 1.2 J)jsclIssion FoHowing a CEA drop, operator action should be directed tnward returning the plant to a stable condition.

At high power levels, if no action is taken following the CEA drop, reactor power will return to approximately the initial power level. but at a reduced core average temperature (due to positive reactivity feedback from the negative moderator tcmpt'Taturc coefficient).

The following actions will minimize the affects of the CEA drop transient.

• Dropped CEAs result in reactof and turbine power mismatch.

This mismatch is nulled by reducing turbine load to match reactor power (RCS temperatures not changing).

• Dropped CEAs also result in undesirable neutron nux patterns. correct operator R'Sponse, the time span over which these pattern.'i cxi'it is minimized.

• It is desirable to record as much data as pos.<;ible concerning abnormal flux patterns existing during and subsequent to rod drop. Proper use of PPC, as outlined in this procedure, produces data necessary fOf subsequent analysis r _____

_____________________________

_ I If during Cf':::A motion. a core mimic light fails to clear or light and all other : : indications Hre normal. it is acceptable to request I&C Department to temporarily I : lift leads for respective CEA upper electrical limit reed switch until it is aligned. : I This is not considered a CEA position indication problem. I -----------------------------------------------_

.. Level of Use Continuous STOP THINK ACT Page 21 of 77 REVIEW Printed on 10/28/2009 at 13:20 Question #: 7 Student Handout? Lower Order? Question ID; 9000020 RO SRO J-SRO Ques. # 7 Rev. a Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 AOP 2556 Revision 016-10 CEA Malfunctions Page 13 of 55 CONTINGENCY ACTIONS lE. necessary, PERFO RM the following to modify dropped CEA position on PPC to match dropped eM position indicated on "CEAPDS MONITOR:" OBSERVEdroppedCEAp05ition on "CEAPDS MONITOR." Using "CEA Positions Menu" on ppc. SELECT "CEA EDITOR" PERFORM directions as on Page 22 of Printed on 10/28/2009 at 13:20 Question #: 7 Question ID; 9000020 RO SRO Student Handout? Lower Order? J-SRO Ques. # 7 Rev. a Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 CEA Malfunctions AOP 2556 Revision 016-10 INSTRUCflONS lE. necessary, PERFO RM the following to modify dropped CEA position on PPC to match dropped eM position indicated on "CEAPDS MONITOR:" I.l. OBSERVEdroppedCEAp05ition on "CEAPDS MONITOR." h Using "CEA Positions Menu" on ppc. SELECT "CEA POSITION EDITOR" and PERFORM directions as indicated on ppc. Page 22 of 77 Page 13 of 55 CONTINGENCY ACTIONS Printed on 10/28/2009 at 13:20 Question #: 7 --Student Handout? Lower Order? Question ID: 9000020 RO SRO I-SRO Ques. # 7 0 ." Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 AOP 2556 Revision 016-10 CEA Malfunctions Page 14 of 55 CONTINGENCY ACTIONS 4.19 As

-SELECT applicable CEA 15 STEPS" or "FULL OIl "CEAPDS 4.20 PRESS "MANUAL

-MI" and CHECK PRESS applicable "INHIBIT BYPASS" CHECK the Appropriate group "INHIBIT BYPASS," "CEA MOTION INIHBlT annunciator lit (BA-19, (depends on group PRESS applicable

_4.23 .LOG entry into TIS LCO, I ACTION B.l (CMl Page 23 of Printed on 10/28/2009 at 13:20 Question #: 7 Question ID: 9000020 RO SRO --Student Handout? Lower Order? I-SRO Ques. # 7 Rev. 0 ." Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 CEA Malfunctions AOP 2556 Revision 016-10 INSTRUCtIONS 4.19 As desired, -SELECT applicable CEA group" 15 STEPS" or "FULL SCALE" OIl "CEAPDS MONITOR." 4.20 PRESS "MANUAL INDIVIDUAL.

-MI" and CHECK lighllit.

_4.21 PRESS applicable group "INHIBIT BYPASS" and CHECK the following:

• Appropriate group red "INHIBIT BYPASS," lit * "CEA MOTION INIHBlT BYP" annunciator lit (BA-19, C-(4) (depends on group selected)

PRESS applicable "GROUP SELECTION." _4.23 .LOG entry into TIS LCO, 3.1.3.1, I ACTION B.l (CMl bypassed).

Page 23 of 77 Page 14 of 55 CONTINGENCY ACTIONS Printed on 10/28/2009 at 13:20 Question #: 8 Question ID: 9000005 []RO SRO Student Handout? L;{l Lower Order? I-SRO Que". "I 8 Rev. 0 Selected for Exam Otfgln: New Past NRC Exam? The reactor is manually tripped from 100% power due to a Steam Generator Tube Rupture (SGTR). On the trip, the RSST is lost due to grid instabilities.

All other systems respond normally.

EOP 2525, Standard Post Trip Actions, is entered. What is thH direction for controlling the affected SG level during the performance of EOP 2525, Standard Post Trip Actions, and what is the basis for this direction? Secure all feedwater to the affected SG. This will provide the maximum volume to accept water from the tube rupture and still allow a cooldown to 515°F to isolate the affected SG. Maintain the affected SG level 40 to 70%. This will maintain the SG tube covered to allow a cooldown to 515'F and still maintain adequate volume to accept water from the tube rupture. Maintain at least 300 gpm feedwater to the affected SG for Heat Removal. The addition of clean water will provide dilution of radioactivity which will lower the release to the environment. Feed the affected SG to maintain level 40 to 45%. This will cover the SG tube to allow for Iodine scrubbing and still allow adequate volume to accept water from the tube rupture. Justification I D IS CORRECT; If Steam Plant Radiation Monitors are in alarm, EOP 2525 requires feed flow to maintained to the SG with the radiation reading to maintain level 40 to 45%. The lower limit of 40% is above the top of the tubes. Keeping the tubes covered for Iodine scrubbing and limits the gaseous release to the environment (The loss of off-site power results in a loss of vacuum which "equires the use of the ADVs for heat removal control).

The upper limit of 45% allows for a significant volume to water from the RCS through the broken A is incorrect; Feedwater should NOT be secured to the affected SG until after the cool down to 515°F. The addition of allows for Iodine scrubbing and limits the radioactive release to the Plausible because, up until a few years ago, feedwater was secured to the affected SG immediately after a SGTR was B is incorrect.

If affected SG level is maintained higher than 45%, then subsequent leakage into the SG from the RCS may result overfilling the SG and could ultimately result in radioactive water being discharged out the affected SG ADV to the Plausible because the normal post trip SG level is 40 to C is incorrect.

While it is true that the addition of feed water will provide additional Iodine scrubbing and dilution of contaminants, too much feedwater flow will result in overfilling the SG and could ultimately result in radioactive water being out the affected SG ADV to the Plausible because EOP 2525 requires feeding the unaffected (or least affected)

SG at greater than 300 gpm .

I OP 2260, Unit 2 EOP User EOP 2525, Standard post Trip Comments and Question Modification Bob K. -D-2fC (need only "40% -45%" for No changes Bill M. D-2/C, Angelo -D-2fC; No NRC KIA System/E/A System 060 Accidental Gaseous Radwaste Release KIA NRC KIA Generic System 2.4 Emergency Procedures IPlan Number 2.4.18 RO 3.3 SRO 4.0 CFR Link (CFR: 41.10 143.1145.13) Knowledge of the speci'ic bases for EOPs. Page 24 of Printed on 10/28/2009 at 13:20 Question #: 8 Question ID: 9000005 []RO SRO Student Handout? L;{l Lower Order? I-SRO Que". "I 8 Rev. 0 Selected for Exam Otfgln: New Past NRC Exam? The reactor is manually tripped from 100% power due to a Steam Generator Tube Rupture (SGTR). On the trip, the RSST is lost due to grid instabilities.

All other systems respond normally.

EOP 2525, Standard Post Trip Actions, is entered. What is thH direction for controlling the affected SG level during the performance of EOP 2525, Standard Post Trip Actions, and what is the basis for this direction?

A Secure all feedwater to the affected SG. This will provide the maximum volume to accept water from the tube rupture and still allow a cooldown to 515°F to isolate the affected SG. B Maintain the affected SG level 40 to 70%. This will maintain the SG tube covered to allow a cooldown to 515'F and still maintain adequate volume to accept water from the tube rupture. C Maintain at least 300 gpm feedwater to the affected SG for Heat Removal. The addition of clean water will provide dilution of radioactivity which will lower the release to the environment. D Feed the affected SG to maintain level 40 to 45%. This will cover the SG tube to allow for Iodine scrubbing and still allow adequate volume to accept water from the tube rupture. Justification I D IS CORRECT; If Steam Plant Radiation Monitors are in alarm, EOP 2525 requires feed flow to maintained to the SG with the highest radiation reading to maintain level 40 to 45%. The lower limit of 40% is above the top of the tubes. Keeping the tubes covered allows for Iodine scrubbing and limits the gaseous release to the environment (The loss of off-site power results in a loss of condenser vacuum which "equires the use of the ADVs for heat removal control).

The upper limit of 45% allows for a significant volume to accept water from the RCS through the broken tube(s). A is incorrect; Feedwater should NOT be secured to the affected SG until after the cool down to 515°F. The addition of feedwater allows for Iodine scrubbing and limits the radioactive release to the environment.

Plausible because, up until a few years ago, feedwater was secured to the affected SG immediately after a SGTR was diagnosed.

B is incorrect.

If affected SG level is maintained higher than 45%, then subsequent leakage into the SG from the RCS may result in overfilling the SG and could ultimately result in radioactive water being discharged out the affected SG ADV to the environment.

Plausible because the normal post trip SG level is 40 to 75%. C is incorrect.

While it is true that the addition of feed water will provide additional Iodine scrubbing and dilution of radioactive contaminants, too much feedwater flow will result in overfilling the SG and could ultimately result in radioactive water being discharged out the affected SG ADV to the environment.

Plausible because EOP 2525 requires feeding the unaffected (or least affected)

SG at greater than 300 gpm . . References I OP 2260, Unit 2 EOP User Guide EOP 2525, Standard post Trip Actions Comments and Question Modification History Bob K. -D-2fC (need only "40% -45%" for answer) No changes Bill M. D-2/C, K Angelo -D-2fC; No comments.

NRC KIA System/E/A System 060 Accidental Gaseous Radwaste Release KIA NRC KIA Generic System 2.4 Emergency Procedures IPlan Number 2.4.18 RO 3.3 SRO 4.0 CFR Link (CFR: 41.10 143.1145.13)

Knowledge of the speci'ic bases for EOPs. Page 24 of 77 Printed on 10/28/2009 at 13:20 SRO Question #: 8 Question ID: 9000005 RO fll SRO Student Handout? Ii'i Lower Order? Rev. o Selected for Exam Otfgln; New 8 Attachment 1 EOP 2525, "Standard Post Trip Actions," Implementation Guide (Sheet'" of 11 ) Due to current MSSV blowdown setpolnts ( -gg{) psia), it is possible for MSSV(s) 10 be open post-trip when the steam dumps are not available have SG pressure within the normal control band. To assess if a MSSV is actually stuck open, it is recommended the CO adjust ADV automatic setpoinl(s) to the lower end of the control band. Additionally, temperature outside the normal band should also be used to IL'isess a stuck open MSSV. 1fT e is less than 53()"F, the operator should determine if feedwater flow is excessive to one or hoth SGs and adjust or isolate feedwater now as required.

The operator should determine if the cause of the excessive flow is due to control system malfunction (e.g., FRV not closing on the trip or FRV bypass valves not at the required position following the trip), or due to fecdwater flow to a SG blowing dO\vn from an ESDE. In the case of malfunctioning equipment, the operator should attempt to adjust feedwatcr flow manually, since a component failure should not result in the loss of a SG [or heat removal. The operator is required to isolate AFW to the affected SG within 30 minutes following the generation of MSlS during an ESDE. For scenarios where isolation is not possible from the Control Room, allowance must be made for Incal operation of FW-43A( B) or FW-44. It has been validated that it will take approximately 15 minutes to close FW-43A(B) or FW-44 locallv. Therefore.

isolation of AFW 10 an affected steam gcne;ator.

from the control room, must be attempted within IS minutes of a MSIS, UTe is less than 53WF ANll the ESDE has been terminated.

the operator is required to operate the ADV or steam dumps to stabilize Tc. Temperature should not be allowed to restore to the normal band following an ESDE. if SG level is lowering and both MDAFW pumps are not operating, the operator is required to start the TDAFP within 10 minutes following a Loss of Normal Fcedwatcr. If a SGTR has occurred, the operator is expected to throttle feed to the most affected SG as necessary to maintain level low in the band (40 to 45tH:). This will aid in maintaining SG pressure during the cooldown and aid in scrubbing radioactive iodine. TIle top of the sa tube bundle is 33fJ*;:'.

If break flow is restoring level to this band then feed flow is not neccssarv

[this mav not be assessed until verification of Containment Isolatiori'l*

• OP2260 STOP THINK ACT REVIEW Rev. 009-03 2Sof60 Page 25 of Printed on 10/28/2009 at 13:20 SRO Question #: 8 Question ID: 9000005 RO fll SRO Student Handout? Ii'i Lower Order? 8 Rev. o Selected for Exam Otfgln; New Attachment 1 EOP 2525, "Standard Post Trip Actions," Implementation Guide (Sheet'" of 11 ) h. Due to current MSSV blowdown setpolnts ( -gg{) psia), it is possible for MSSV(s) 10 be open post-trip when the steam dumps are not available have SG pressure within the normal control band. To assess if a MSSV is actually stuck open, it is recommended the CO adjust ADV automatic setpoinl(s) to the lower end of the control band. Additionally, temperature outside the normal band should also be used to IL'isess a stuck open MSSV. c. 1fT e is less than 53()"F, the operator should determine if feedwater flow is excessive to one or hoth SGs and adjust or isolate feedwater now as required.

The operator should determine if the cause of the excessive flow is due to control system malfunction (e.g., FRV not closing on the trip or FRV bypass valves not at the required position following the trip), or due to fecdwater flow to a SG blowing dO\vn from an ESDE.

• In the case of malfunctioning equipment, the operator should attempt to adjust feedwatcr flow manually, since a component failure should not result in the loss of a SG [or heat removal.

• The operator is required to isolate AFW to the affected SG within 30 minutes following the generation of MSlS during an ESDE. For scenarios where isolation is not possible from the Control Room, allowance must be made for Incal operation of FW -43A( B) or FW -44. It has been validated that it will take approximately 15 minutcs to close FW-43A(B) or FW-44 locallv. Therefore.

isolation of AFW 10 an affected steam gcne;ator.

from the control room, must be attempted within IS minutes of a MSIS. d. UTe is less than 53WF ANll the ESDE has been terminated.

the operator is required to operate the ADV or steam dumps to stabilize Temperature should not be allowed to restore to the normal band following an ESDE. e. if SG level is lowering and both MDAFW pumps are not operating, the operator is required to start the TDAFP within 10 minutes following a Loss of Normal Fcedwatcr.

f. If a SGTR has occurred, the operator is expected to throttle feed to the most affected SG as necessary to maintain level low in the band (40 to 45tH:). This will aid in maintaining SG pressure during the cooldown and aid in scrubbing radioactive iodine. TIle top of the sa tube bundle is 33(}i':.lfbreak flow is restoring level to this band then feed flow is not neccssarv

[this mav not be assessed until verification of Containment Isolatiori'l*

• STOP THINK ACT Page 25 of 77 OP2260 REVIEW Rev. 009-03 2Sof60 Printed on 10/28/2009 at 13:20 Question #: 9 RO SRO Student Handout? Lower Order?Question ID: 9000006 l-SRO Ques. # 9 Rev. o Selected for Exam Origin: New o Past NRC Exam? The plant is in MODE 5 performing a normal cooldown for refueling. "B" LPSI Pump is in service supplying both SDC Heat Exchangers.

RCS to SDC Temperature, T351X, is presently reading 18rF with RCS pressure being held at 150 pSia. Suddenly, Bus 24D is deenergized due to a fault. Fifteen minutes after the loss of Bus 24D, the following conditions are reported:

-RCS pressure is 155 psia and slowly -RCS to SDC Temperature, T351X, is reading 186°F and -CET temperatures are 205°F and slowly -RVLMS indicates vessel level at -Both S/G levels are 60% and -Containment is being NO other operator actions have been taken. Which of tile following notifications must be made? o A General Interest, Echo LJ B Unusual Event, Delta-One C Alert, Charlie-One o D Site Area Emergency, Charlie-Two Justification C IS CORRECT; MP-2*-EPI-FAP06-002, Equipment Failure, EA2, Uncontrolled temperature increase >10"F that results in temperature

>200"F. Due to the loss of SDC flow, T351X, is no longer providing an accurate RCS temperature.

The operator use CET temperatures to determine the actual change in RCS A is incorrect; The loss of Bus 24D would result in an Undervoltage actuation on Facility 2. Per RAC 14, Non-Emergency Events, an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report is required. (General Interest, Echo); however, the loss of SDC with a temperature rise of >10"F is a classification.

The highest classification must be reported.

Other details may be included in the initial Plausible if examinee thought that this was the only reportable B is incorrect.

Per MP-26-EPI-FAP06-002, Equipment Failure, EU1(2.), Uncontrolled temperature increase >10"F, a classification Unusual Event Delta-One may be reported; however, the loss of SDC with a temperature rise of >10"F is a higher classification. highest classifiGation must be Plausible if examinee diel not look at the higher Alert classification for Inability to Maintain Cold D is incorrect; Per MP-26-EPI-FAP06-002, Equipment Failure, ES2(1.), No RCS Heat Removal via Steam Generators AND Through Cooling NOT effective AND Shutdown Cooling NOT in service. This may appear to be a logical choice in that none of the core cooling methods are being utilized; however, Steam Generators are available for Heat Removal when RCS temperature is enough to cause steaming AND Once Through Cooling would likely be effective if it were Plausible because the examinee may believe that no core cooling method is presently being utilized, therefore, core cooling is serious References MP-26-EPI-FAP06-002, Millstone Unit 2 Emergency Action Levels, and RAC-14, Non-Emergency Station Comments and Question Modification Bob K. -Bill M, -D-21Y, K (Add, "No other operator actions have been taken." to the bullet Added "No other operator actions have been taken," to the bullet list, per recommendation. Angelo -D-3/C: No NRC KIA System/E/A Generic KIA Selected NRC KIA Generic System System 074 2.4 Inadequate Core Cooling Emergency Procedures IPlan Page 26 of 77 Printed on 10/2812009 at 13:20 Question #: 9 Question ID: 9000006 l-SRO Ques. # 9 Rev. o RO SRO Selected for Exam Student Handout? Origin: New Lower Order? o Past NRC Exam? The plant is in MODE 5 performing a normal cooldown for refueling. "B" LPSI Pump is in service supplying both SDC Heat Exchangers.

RCS to SDC Temperature, T351X, is presently reading 18rF with RCS pressure being held at 150 pSia. Suddenly, Bus 24D is deenergized due to a fault. Fifteen minutes after the loss of Bus 24D, the following conditions are reported:

-RCS pressure is 155 psia and slowly rising. -RCS to SDC Temperature, T351X, is reading 186°F and stable. -CET temperatures are 205°F and slowly rising. -RVLMS indicates vessel level at 100%. -Both S/G levels are 60% and stable -Containment is being evacuated.

NO other operator actions have been taken. Which of tile following notifications must be made? o A General Interest, Echo LJ B Unusual Event, Delta-One C Alert, Charlie-One o D Site Area Emergency, Charlie-Two Justification J C IS CORRECT; MP-2*-EPI-FAP06-002, Equipment Failure, EA2, Uncontrolled temperature increase >10"F that results in RCS temperature

>200"F. Due to the loss of SDC flow, T351X, is no longer providing an accurate RCS temperature.

The operator must use CET temperatures to determine the actual change in RCS temperature.

A is incorrect; The loss of Bus 24D would result in an Undervoltage actuation on Facility 2. Per RAC 14, Non-Emergency Station Events, an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report is required. (General Interest, Echo); however, the loss of SDC with a temperature rise of >10"F is a higher classification.

The highest classification must be reported.

Other details may be included in the initial report. Plausible if examinee thought that this was the only reportable event. B is incorrect.

Per MP-26-EPI-FAP06-002, Equipment Failure, EU1(2.), Uncontrolled temperature increase >10"F, a classification of Unusual Event Delta-One may be reported; however, the loss of SDC with a temperature rise of >10"F is a higher classification.

The highest classifiGation must be reported.

Plausible if examinee diel not look at the higher Alert classification for Inability to Maintain Cold SID. D is incorrect; Per MP-26-EPI-FAP06-002, Equipment Failure, ES2(1.), No RCS Heat Removal via Steam Generators AND Once Through Cooling NOT effective AND Shutdown Cooling NOT in service. This may appear to be a logical choice in that none of the 3 core cooling methods are being utilized; however, Steam Generators are available for Heat Removal when RCS temperature is high enough to cause steaming AND Once Through Cooling would likely be effective if it were initiated.

Plausible because the examinee may believe that no core cooling method is presently being utilized, therefore, core cooling is in serious jeopardy.

References MP-26-EPI-FAP06-002, Millstone Unit 2 Emergency Action Levels, and RAC-14, Non-Emergency Station Events Comments and Question Modification History Bob K. -D-3/C Bill M, -D-21Y, K (Add, "No other operator actions have been taken." to the bullet list) Added "No other operator actions have been taken," to the bullet list, per recommendation.

RJA Angelo -D-3/C: No comments.

NRC KIA System/E/A System 074 Inadequate Core Cooling Generic KIA Selected NRC KIA Generic System 2.4 Emergency Procedures IPlan Page 26 of 77 Printed on 10/2812009 at 13:20 Question #: 9 Question ID: 9000006 RO!YfJ BRO Student Handout? Lower Order? I-SRO Ques. # 9 Rev. 0 !iii Selected for Exam Origin: New Past NRC Exam? Number 2.4.30 RO 2.7 SRO 4.1 CFR Link (CFR: 41.10 /43.5/45.11)

Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Failure of Automatio Reactor Trip AI'!D Manual Trip WIIs SL.Il.X:essful I EA2 I IINABII.ITY TO MAINTAIN COLD SID I I Mode 5. 6 1. Uncontrolled ReS Temperature Increase>

10"F That Results in RCS Temperature>

2OO"F 2. Inadvertent Criticality I EA3 I boss OF ANNUNCIATORSfmANSIENj I Mode 1, 2, 3. 4 I Loss of Most (75%) MOB Arlnunoiators

> 15 Minutes ANIl EITHER of the Following:

Significant Transient in Progress Instrumentation

1. Loss of Cooling,.

15 Minutes AND Refuel Pool Water Level" 35 Fl.. 6 In. 2. Unoontrolloc RCS TemlX,rature " 10'F 3. RCS Boron Concentration

<: Minimum Required I EU2 I "'---M-o-d-e-S-, 0-"" 1. Unoxlntrollec Spent Fuel Pool Water Level Decrease Causing Loss of Suction Flow 2. Uncontrollec Refuel POOl Water Level Decrease Requiring Conlai"menl Evacuation tlli.O. All Spent Fuel Assemblies In Safe Storage Lociltions I EU3 I r=w-SS-O-F I Mode 1, 2. 3, 4 I Loss of Most (75%) MCB AnnJnciators

,. 15 Minutes AND SPDS QB ICC Instrumentation Available Mode ALL 1. Loss of ALL Onsile Electronic Communications Methods 2. Loss of ALL I':lectronic Communications Methods With Govemment AgenCIes I SHUTDOWN Leo EXCEEDED I I Mode 1, 2, 3. 4 I Unit NOT Brought To Required Mode Within Applicable LCO Action Stalemerl Time l imils Page 27 of 77 Printed on 10/28/2009 at 13 :20 Question #: 9 Question ID: 9000006 RO!YfJ BRO Student Handout? Lower Order? I-SRO Ques. # 9 Rev. 0 !iii Selected for Exam Origin: New Past NRC Exam? Number 2.4.30 RO 2.7 SRO 4.1 CFR Link (CFR: 41.10 /43.5/45.11)

Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Failure of Automatio Reactor Trip AI'!D Manual Trip WIIs SL.Il.X:essful I EA2 I IINABII.ITY TO MAINTAIN COLD SID I I Mode 5. 6 1. Uncontrolled ReS Temperature Increase>

10"F That Results in RCS Temperature>

2OO"F 2. Inadvertent Criticality I EA3 I boss OF ANNUNCIATORSfmANSIENj I Mode 1, 2, 3. 4 I Loss of Most (75%) MOB Arlnunoiators

> 15 Minutes ANIl EITHER of the Following:

Significant Transient in Progress Instrumentation

1. Loss of Cooling,.

15 Minutes AND Refuel Pool Water Level" 35 Fl.. 6 In. 2. Unoontrolloc RCS TemlX,rature " 10'F 3. RCS Boron Concentration

<: Minimum Required I EU2 I "'---M-o-d-e-S-, 0-"" 1. Unoxlntrollec Spent Fuel Pool Water Level Decrease Causing Loss of Suction Flow 2. Uncontrollec Refuel POOl Water Level Decrease Requiring Conlai"menl Evacuation tlli.O. All Spent Fuel Assemblies In Safe Storage Lociltions I EU3 I r=w-SS-O-F I Mode 1, 2. 3, 4 I Loss of Most (75%) MCB AnnJnciators

,. 15 Minutes AND SPDS QB ICC Instrumentation Available Mode ALL 1. Loss of ALL Onsile Electronic Communications Methods 2. Loss of ALL I':lectronic Communications Methods With Govemment AgenCIes I SHUTDOWN Leo EXCEEDED I I Mode 1, 2, 3. 4 I Unit NOT Brought To Required Mode Within Applicable LCO Action Stalemerl Time l imils Page 27 of 77 Printed on 10/28/2009 at 13 :20 Question #: 10 Question ID: 9000007 ORO !ill SRO Student Handout? !ill Lower Order? I-SRO Ques. 1# '10 Rev. 0 !ill Selected for Exam Origin: New Past NRC Exam? An RCS chemistry sample taken at 100% power indicates 6 micro-curies/gram DOSE EQUIVALENT 1-131. Which of the following describes the required action and the basis for that action? With the specific activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131, be in COl.D SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after detection.

Isotopic analysis of the primary coolant must be performed once per hour when activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131. The hourly sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and there is significant conservatism built into the RCS specific activity limit. With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131,B verify DOSE EQUIVALENT 1-131 .::s 60 micro-curies/gram once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Operation may for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 curies/gram The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> sampling period allows time to obtain and analyze a sample. There is a low probability of steam line break or S/G tube rupture in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and it is expected that normal coolant concentration would be restored within 48 With the specific activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131, uC in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from the time of detection and in COLD SHUTDOWN within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of detection.

The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power in an orderly manner and prevent exceeding the radiological release limit a: the site boundary from an assumed LOCA. With tne specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, uD lower the RCS specific activity to.::s 1.0 micro-curies/gram DOSE EQUIVALENT 1-131 within the next hours or be in HOT STANDBY within the following 6 It is expected that normal coolant Iodine concentration would be restored within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. If adequate time is provided to achieve HOT STANDBY to prevent exceeding Control Room dose from an assumed . Justification I B is CORRECT; TSAS 3.4.8a. and b. state: a. With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 60 micro-curies/gram once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. b. With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, but 60 micro-curies/gram, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 micro-curies/gram limit. The Basis for TS 3.4.8 states: With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> must be taken to demonstrate that the specific activity IS 60 micro-curies/gram.

Four hours is required to obtain and analyze a sample. Sampling is continued every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to provide a trend. The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The completion time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it is expected that, if there were an Iodine spike, normal coolant Iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period. A is incorrect; TSAS 3.4.8 does not require the plant to achieve COLD SHUTDOWN within 36 Plausible because several other Tech Spec Action Statements require the plant to achieve COLD SHUTDOWN within 36 (Example:

COlltainment Integrity, TSAS C is incorrect; TSAS 3.4.8 does not require the plant to achieve HOT SHUTDOWN within 6 Plausible because several other Tech Spec Action Statements require the plant to achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> COLD SHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. (Example:

Specific Activity, TSAS 3.4.8c., DOSE EQUIVALENT 1-131 >60 D is incorrect; TSAS 3.4.8 does not require the plant to achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if RCS coolant specific cannot be lowored to 1.0 micro-curies/gram DOSE EQUIVALENT 1-131 in 36 Plausible if the examinee confuses the time requirements and the actual References I Tech Spec 3.4.8, Specific Activity, and applicable Bases. Comments and Question Modification History Bob K. -D-4/W (Could not remember requirement details. Did not believe should be required from Changed distractors A and C to 0.1 micro-curlesfgram instead of the original 1.0 micro-curies/gram.

The different require the examinee to remember the actual limit and not just the action required if the limit is exceeded. Page 28 of Printed on 10/28/2009 at 13:20 Question #: 10 Question ID: 9000007 ORO !ill SRO Student Handout? !ill Lower Order? I-SRO Ques. 1# '10 Rev. 0 !ill Selected for Exam Origin: New Past NRC Exam? An RCS chemistry sample taken at 100% power indicates 6 micro-curies/gram DOSE EQUIVALENT 1-131. Which of the following describes the required action and the basis for that action? A With the specific activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131, be in COl.D SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after detection.

Isotopic analysis of the primary coolant must be performed once per hour when activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131. B uC uD The hourly sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and there is significant conservatism built into the RCS specific activity limit. With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 .::s 60 micro-curies/gram once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 curies/gram limit. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and it is expected that normal coolant Iodine concentration would be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. With the specific activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from the time of detection and in COLD SHUTDOWN within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of detection.

The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power in an orderly manner and prevent exceeding the radiological release limit a: the site boundary from an assumed LOCA. With tne specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, lower the RCS specific activity to.::s 1.0 micro-curies/gram DOSE EQUIVALENT 1-131 within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. It is expected that normal coolant Iodine concentration would be restored within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. If not, adequate time is provided to achieve HOT STANDBY to prevent exceeding Control Room dose limits from an assumed LOCA. . Justification I B is CORRECT; TSAS 3.4.8a. and b. state: a. With the specific actiVIty of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 60 micro-curies/gram once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. b. With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, but 60 micro-curies/gram, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 micro-curies/gram limit. The Basis for TS 3.4.8 states: With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> must be taken to demonstrate that the specific activity IS 60 micro-curies/gram.

Four hours is required to obtain and analyze a sample. Sampling is continued every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to provide a trend. The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The completion time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it is expected that, if there were an Iodine spike, normal coolant Iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period. A is incorrect; TSAS 3.4.8 does not require the plant to achieve COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Plausible because several other Tech Spec Action Statements require the plant to achieve COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Example:

COlltainment Integrity, TSAS 3.6.1.1).

C is incorrect; TSAS 3.4.8 does not require the plant to achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Plausible because several other Tech Spec Action Statements require the plant to achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. (Example:

Specific Activity, TSAS 3.4.8c., DOSE EQUIVALENT 1-131 >60 curies/gram.)

D is incorrect; TSAS 3.4.8 does not require the plant to achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if RCS coolant specific activity cannot be lowored to 1.0 micro-curies/gram DOSE EQUIVALENT 1-131 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Plausible if the examinee confuses the time requirements and the actual limit. References I Tech Spec 3.4.8, Specific Activity, and applicable Bases. Comments and Question Modification History I Bob K. -D-4/W (Could not remember requirement details. Did not believe should be required from memory.) Changed distractors A and C to 0.1 micro-curlesfgram Instead of the original 1.0 micro-curies/gram.

The different limits require the examinee to remember the actual limit and not just the action required if the limit is exceeded.

RJA Page 28 of 77 Printed on 10/28/2009 at 13:20 Question #: 10 Question ID: 9000007 L RO SRO Student Handout? fi'l Lower Order? I-SRO Ques. 11 '10 Rev. o .!lJ Selected for Exam Origin: New Past NRC Exam? Bill M. -D-4/W, G (Could not remember requirement details. Examinees should get this if they studied Tech Specs and Bases,) Angelo -D-5/C; Difficult but fair. Changed 1.0 in "Au to 0.1 to match number in first sentence,

• RLC NRC KIA System/E:tA System 076 High Reactor Coolant Activity Number AA2,02 RO 2.8 SRO 3.4 CFR Link (CFR: 43.5 /45.13) Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity:

Corrective actions required for high fission product activity in RCS Page 29 of 77 Printed on 10/2812009 at 13:20 Question #: 10 Question ID: 9000007 L RO SRO Student Handout? fi'l Lower Order? I-SRO Ques. 11 '10 Rev. o .!lJ Selected for Exam Origin: New Past NRC Exam? Bill M. -D-4/W, G (Could not remember requirement details. Examinees should get this if they studied Tech Specs and Bases,) Angelo -D-5/C; Difficult but fair. Changed 1.0 in "Au to 0.1 to match number in first sentence,

• RLC NRC KIA System/E:tA System 076 High Reactor Coolant Activity Number AA2,02 RO 2.8 SRO 3.4 CFR Link (CFR: 43.5/45.13)

Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity:

Corrective actions required for high fission product activity in RCS Page 29 of 77 Printed on 10/2812009 at 13:20 Question #: 10 Question ID: 9000007 RO SRO Student Handout? V Lower Order? I-SRO Ques. # 10 Rev. 0 \l.] Selected for Exam Origin: New REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to: 1.0 EQUIVALENT 1-131. and b. 1100 IICilgram DOSE EQUIVALENT XE-133. APPLICABILITY:

MODES 1,2.3,4. ACTION: With the specific activity of the primary coolant> 1.0 DOSE EQUIVALENt 1-131, verify DOSE EQUIVALENT 1-131 s 60 once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. With the specific activity of the pri1ll31Y coolant> 1.0 DOSE EQUIVAL£t..it 1-131 but s 60 operationlllay continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while eff011s are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 !lCi/gram limit. Specification 3.0.4 is 110t applicable. With the specific activity of the primaIY coolant> 1.0 DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval, or> 60 DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hOUl'S and in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. With the specific activity of the primary coolant> 1100 DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore DOSE EQUIVALENT XE-133 to within the 1100 limit. Specification 3.0.4 is not applicable. With the specific activity of the primary coolant> 1100 DOSE EQUIVALENT XE-133 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time intervaL be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN \\'ithill 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Page 30 of Printed on 10/28/2009 at 13:20 Question #: 10 Question ID: 9000007 RO SRO Student Handout? V Lower Order? I-SRO Ques. # 10 Rev. 0 \l.] Selected for Exam Origin: New REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to: a. 1.0 EQUIVALENT 1-131. and b. 1100 IICilgram DOSE EQUIVALENT XE-133. APPLICABILITY:

MODES 1,2.3,4. ACTION: a. With the specific activity of the primary coolant> 1.0 DOSE EQUIVALENt 1-131, verify DOSE EQUIVALENT 1-131 s 60 once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. b. With the specific activity of the pri1ll31Y coolant> 1.0 DOSE EQUIVAL£t..it 1-131 but s 60 operationlllay continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while eff011s are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 !lCi/gram limit. Specification 3.0.4 is 110t applicable.

c. With the specific activity of the primaIY coolant> 1.0 DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval, or> 60 DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hOUl'S and in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. d. With the specific activity of the primary coolant> 1100 DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore DOSE EQUIVALENT XE-133 to within the 1100 limit. Specification 3.0.4 is not applicable.

e. With the specific activity of the primary coolant> 1100 DOSE EQUIVALENT XE-133 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN \\'ithill 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Page 30 of 77 Printed on 10/28/2009 at 13:20

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Question #: 11 RO SRO Student Handout? D Lower Order?Question ID: 9000008 I-SRO Ques. II 11 Rev. 0 Selected for Exam Origin: New D Past NRC Exam? A plant heatup has just been started and the following conditions presently exist: -RCS Temperature is at 205°F and slowly -RCS pmssure is stable at the minimum allowed for "A" and "8" RCP

-"A" and "8" RCPs have just been

-Shutdown Cooling has just been

-"C" and "D" RCP breakers have just been racked Then, the "8" RCP trips when the breaker's overcurrent relay actuated due to being jarred while moving staging (NOT an actual overcurrent condition).

Which of the following actions are required under the present conditions? A Immediately secure the "A" RCP, raise RCS pressure, then start "C" and "D" RCPs. B Immediately place Shutdown Cooling back in operation, then secure the "A" RCP. C Immediately start "C" and "D" RCPs, then secure the "A" RCP. o Immediately start the "C" RCP and operate it with the "A" RCP. Justification I A -CORRECT The NPSH required pressure for "N &"8" RCPs is based on both pumps operating, therefore, the "A" RCP must immediately secured. With a plant heatup in operation, OP-2201 states RCS pressure should be raised as required to allow for the of the available RCPs, then they should be 8 WRONG; If NPSH requirements are not met, the RCP should be immediately Plausible; Would be chosen if avoiding the loss all RCS flow (violation of Tech. Specs.) is considered above possible RCP C -WRONG; The N PSH required pressure for "A" & "8" RCPs is less than that for "c" & "0" RCPs (see OP-2201. Attach. 2 & 3). If RCS pressure s at the !!linimum allowed for "A" &"8" RCP operation (initial conditions), it must be below the minimum pressure for "C" & "0" Rep Plausible; Would be chosen if "c" & "0" RCP NPSH required pressure is believed to be lower than "A" & "8" o WRONG; Two pump operation is specific to the applicable pumps as one pump aids in the NPSH requirements of the other. Plausible; Would be chosen if two pump operation is known, but believed to be any two RCPs operating Simultaneously.

References I OP-2301. Pg. 24; Caution and Pg. 68; Attach. 6. Conditional Actions. ;Comments and Question Modification Histo!X I 80b K. -D-4fW (Reword to make clear "relay failure" caused breaker to trip and is a quick fix). Reworded stem to clairify problem is with the breaker's overload relay and is a relatively quick fix. -RLC 8i11 M. 3/C, K Angelo -D-2fW; Changed RCS temperature to 20S*F to allow all four RCP breakers to be closed. -RLC NRC KIA SystemtEtA System 003 Reactor Coolant Pump System (RCPS) ----.---.--!J Generic KIA Selected System 2.1 Conduct of Operations Number 2.1.23 R04.3 SRO 4.4 CFR Link (CFR: 41.10/43.5/45.2

/ 45.6) Ability to perfcrm specific system and integrated plant procedures during all modes of plant operation.

NRC KIA Generic Page 31 of 77 Printed on 10/28/2009 at 13:20 Question #: 11 Question ID: 9000008 RO SRO Student Handout? I-SRO Ques. II 11 Rev. 0 Selected for Exam Origin: New A plant heatup has just been started and the following conditions presently exist: -RCS Temperature is at 205°F and slowly rising. -RCS pmssure is stable at the minimum allowed for "A" and "8" RCP operation.

-"A" and "8" RCPs have just been started. -Shutdown Cooling has just been secured. -"C" and "D" RCP breakers have just been racked up. D Lower Order? D Past NRC Exam? Then, the "8" RCP trips when the breaker's overcurrent relay actuated due to being jarred while moving staging (NOT an actual overcurrent condition).

Which of the following actions are required under the present conditions?

.................................................................................. A Immediately secure the "A" RCP, raise RCS pressure, then start "C" and "D" RCPs. B Immediately place Shutdown Cooling back in operation, then secure the "A" RCP. C Immediately start "C" and "D" RCPs, then secure the "A" RCP. o Immediately start the "C" RCP and operate it with the "A" RCP. Justification I A -CORRECT The NPSH required pressure for "N & "8" RCPs is based on both pumps operating, therefore, the "A" RCP must be immediately secured. With a plant heatup in operation, OP-2201 states RCS pressure should be raised as required to allow for the start of the available RCPs, then they should be started. 8 WRONG; If NPSH requirements are not met, the RCP should be immediately secured. Plausible; Would be chosen if avoiding the loss all RCS flow (violation of Tech. Specs.) is considered above possible RCP damage. C -WRONG; The N PSH required pressure for "A" & "8" RCPs is less than that for "c" & "0" RCPs (see OP-2201. Attach. 2 & 3). If the RCS pressure s at the !!linimum allowed for "A" & "8" RCP operation (initial conditions), it must be below the minimum required pressure for "C" & "0" Rep operation.

Plausible; Would be chosen if "c" & "0" RCP NPSH required pressure is believed to be lower than "A" & "8" RCPs. o WRONG; Two pump operation is specific to the applicable pumps as one pump aids in the NPSH requirements of the other. Plausible; Would be chosen if two pump operation is known, but believed to be any two RCPs operating Simultaneously.

References I OP-2301. Pg. 24; Caution and Pg. 68; Attach. 6. Conditional Actions. ;Comments and Question Modification Histo!X I 80b K. -D-4fW (Reword to make clear "relay failure" caused breaker to trip and is a quick fix). Reworded stem to clairify problem is with the breaker's overload relay and is a relatively quick fix. -RLC 8i11 M. 3/C, K Angelo -D-2fW; Changed RCS temperature to 20S*F to allow all four RCP breakers to be closed. -RLC NRC KIA SystemtEtA System 003 Reactor Coolant Pump System (RCPS) ----.---.--!J Generic KIA Selected --------NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.23 R04.3 SRO 4.4 CFR Link (CFR: 41.10/43.5/45.2

/ 45.6) Ability to perfcrm specific system and integrated plant procedures during all modes of plant operation.

Page 31 of 77 Printed on 10/28/2009 at 13:20 I SROE,x.arnQuestions Only (NoUParents" Or Question #: 11 D Student Handout? D Lower Order?Question ID: 9000008 []RO SRO P-SRO Ques. ,¥ 11 Rev. 0 Selected for Exam Origin: New D Past NRC Exam? CAUTION Two RCPs in the same loop are started to ensure proper NPSH for the pumps, To ensure proper NPSH, the second pump should be started immediately after first pump starting current has decayed. Upon RCP start the initiation of bypass spray /low and any potential in --sur&e, may result in a lowering oCRCS pressure.

Prompt operation of pressurizer heaters (proportional and backup), should be anticipated.

NOTE When starting RCPs. it is desirable to maintain RCS pressure between RCP MNPSH and 265 psi a, as indicated on Attachment 2, 3 or the applicable PPC NPSH display, to accommodate any pressure change on start. If a temperature rise is expected upon RCP start, this must also be considered part of the RCS heatup. After starting RCPs, perform the following: SOC to RCS temperature, T351 Y, must be lowered rapidly to compensate for the extra heat input from the RCPs. SOC to RCS temperature, T351 y, is then raised to allow heatup to progress. Monitor pressurizer pressure.

Refer"Ib OP 2301C, "Reactor Coolant Pump Operation" and PERFORM applicable actions to start selected Reps (C-03). WHEN RCPs are operating, Refer 'Tb Attachments 1 and 2 or 3, or applicable PPC displays, and VERIFY RCS pressure to between MNPSH and 265 psi a (C-03, ppe). Using "SOC SYS AX FLOW CNTL, HIC-365T' (C-Ol) and "SOC SYS TOTAL FLOW, FIC-306", AOJUST SOC System LlT to obtain value calculated in step 4.4.7 for concurrent SOC/RCP operation with {)"F heatup rate. I\s necessmy.

STABIUZE RCS temperature OR CONTINUE plant heatup. CLOSE manual disconnect switch, 89-S1652 (for SI -652, "SOC SYS suer CTMT ISOL," west wall of Control Room). CLOSE "2-S1-651, MANUAL OISCONNECTSWITCH, NS1651," (14'6" Aux Bldg, west wall, across from B51 enclosure) STOP THINK ACT REVIEW Rev. 24 of __LI__ __LI___ __LI___ ------1.1 __ __LI____ Level 0 Contini4.4.12 4.4.13 4.4.14 4.4.15 4.4.16 4.4.17 4.4.18 Page 32 of Printed on 10/28/2009 at 13:20 SROE,x.arnQuestions Only (NoUParents" Or Question #: 11 Question ID: 9000008 []RO SRO D Student Handout? D Lower Order? P-SRO Ques. ,¥ 11 Rev. 0 Selected for Exam Origin: New D Past NRC Exam? I 1. 2. l. 2. __ LI __ __ LI ___

__ LI ___ ------1.1 __ __ LI ____ Level 0 V CAUTION V Two RCPs in the same loop are started to ensure proper NPSH for the pumps, To ensure proper NPSH, the second pump should be started immediately after first pump starting current has decayed. Upon RCP start the initiation of bypass spray /low and any potential in --sur&e, may result in a lowering oCRCS pressure.

Prompt operation of pressurizer heaters (proportional and backup), should be anticipated.

NOTE When starting RCPs. it is desirable to maintain RCS pressure between RCP MNPSH and 265 psi a, as indicated on Attachment 2, 3 or the applicable PPC NPSH display, to accommodate any pressure change on start. If a temperature rise is expected upon RCP start, this must also be considered part of the RCS heatup. 4.4.12 After starting RCPs, perform the following:

• SOC to RCS temperature, T351 Y, must be lowered rapidly to compensate for the extra heat input from the RCPs.

• SOC to RCS temperature, T351 y, is then raised to allow heatup to progress.

• Monitor pressurizer pressure.

4.4.13 Refer"Ib OP 2301C, "Reactor Coolant Pump Operation" and PERFORM applicable actions to start selected Reps (C-03). 4.4.14 WHEN RCPs are operating, Refer 'Tb Attachments 1 and 2 or 3, or applicable PPC displays, and VERIFY RCS pressure to between MNPSH and 265 psi a (C-03, ppe). 4.4.15 Using "SOC SYS AX FLOW CNTL, HIC-365T' (C-Ol) and "SOC SYS TOTAL FLOW, FIC-306", AOJUST SOC System LlT to obtain value calculated in step 4.4.7 for concurrent SOC/RCP operation with {)"F heatup rate. 4.4.16 I\s necessmy.

STABIUZE RCS temperature OR CONTINUE plant heatup. 4.4.17 CLOSE manual disconnect switch, 89-S1652 (for SI -652, "SOC SYS suer CTMT ISOL," west wall of Control Room). 4.4.18 CLOSE "2-S1-651, MANUAL OISCONNECTSWITCH, NS1651," (14'6" Aux Bldg, west wall, across from B51 enclosure) fUse JOUS Contini STOP THINK ACT REVIEW OP2201 Rev. 031-12 24 of 109 Page 32 of 77 Printed on 10/28/2009 at 13:20 Question #: 11 Student Handout? Lower Order? 11 Rev. o Selected for Exam Origin: New Past NRC Exam? Question ID: 9000008 RO SRO Attachment 6 Plant Heatup Conditional Actions (Shl:c! 1 or 5) l.E in MODE 3 AND one RCP loop is not operable, LOG in to ledl Spec Statement at any time. RCP operation cannot continue, and it is necessary to restore PERFORM the PERFORM one of lhe following: STOP affected RCPs. IE Jess than two RCPs will remain running, STOP all RCPs and LOG into the following,

• MODE "rSAS 3.4.:1.2 AC'T'ION b

• MODE 4. TSAS ACfION c Refer To applicable RCP NPSH Altachment, or PPC NPSH display, and CHECK NPSH for running RCPs is met.

• IF NPSH for running RCPs is not met, STOP RCPs. IF applicable, Refer lb one of the following and COOLDOWN plant to less than 300"£ EOP 2528, "Loss of Offsite Power/Loss of Forced Circulation" OP 2207, "Plant Cooldown" VERIFY all HPSI pump control switches are in "PULL TO LOCK." at any time, during heatup, one RCP is lost (initially two RCPs operating SDCnoJ in service), PERFORM the TRIP remaining RCP. LOG into the fOllowing: MODE TSAS 3.4.1.2 ACllON b MODE 4, TSAS 3.4.1.3 ACnON c Refer Tb EOP 2.')28, "Loss of Offsile Power/Loss of Forced Circulation." ADJUST RCS pressure to establL<ih adequate NPSH for RCP operation as specified in Attachment

4. IF 2 Reps are available to be operated, Refer To OP 2301 C and START 2 RCPs. 3.5.1 LOG out of the following:

• MODE

3.4.1 ACI10N

b

• TSAS 3.4.1.3 ACTION c OP220.1 I L evel of Use STOP THINK ACT Rev. 031-12 68 of 109 Continuous Page 33 of Printed on 10/28/2009 at 13:20 Question #: 11 11 Question ID: 9000008 RO SRO Rev. o Selected for Exam Attachment 6 Plant Heatup Conditional Actions (Shl:c! 1 or 5) Student Handout? Origin: New 1. l.E in MODE 3 AND one RCP loop is not operable, LOG in to ledl Spec Action Statement 3.4,1.2. 2. at any time. RCP operation cannot continue, and it is necessary to restore SOC. PERFORM the following:

2.1 PERFORM

one of lhe following:

• STOP affected RCPs.

• IE Jess than two RCPs will remain running, STOP all RCPs and LOG into the following,

• MODE "rSAS 3.4.:1.2 AC'T'ION b

• MODE 4. TSAS ACfION c

• Refer To applicable RCP NPSH Altachment, or PPC NPSH display, and CHECK NPSH for running RCPs is met.

• IF NPSH for running RCPs is not met, STOP RCPs. 2.2 IF applicable, Refer lb one of the following and COOLDOWN plant to less than 300"£

• EOP 2528, "Loss of Offsite Power/Loss of Forced Circulation"

• OP 2207, "Plant Cooldown" 2.3 VERIFY all HPSI pump control switches are in "PULL TO LOCK." 3. at any time, during heatup, one RCP is lost (initially two RCPs operating AND SDCnoJ in service), PERFORM the following:

3.1 TRIP remaining RCP. 3.2 LOG into the fOllowing:

• MODE TSAS 3.4.1.2 ACllON b

• MODE 4, TSAS 3.4.1.3 ACnON c 3.3 Refer Tb EOP 2.')28, "Loss of Offsile Power/Loss of Forced Circulation." 3.4 ADJUST RCS pressure to establL<ih adequate NPSH for RCP operation as specified in Attachment

4. 3.5 IF 2 Reps are available to be operated, Refer To OP 2301 C and START 2 RCPs. 3.5.1 LOG out of the following:

• MODE

3.4.1 ACI10N

b

• TSAS 3.4.1.3 ACTION c I Level of Use Continuous STOP THINK ACT REVIEW OP220.1 Rev. 031-12 68 of 109 Lower Order? Past NRC Exam? Page 33 of 77 Printed on 10/28/2009 at 13:20

, Question #: 12 o Student Handout? Lower Order?Question 10; 9600016 RO SRO I-SRO Ques. # 12 Rev. 0 Selected for Exam Orlgln; Mod Past NRC Exam? A Rapid Downpower at 50%/hr is in progress due to an RCS leak in containment that exceeds administrative limit. The following plant conditions presently

• Plant power is 93% and dropping at the intended rate.

• PressLlrizer level is 65% and stable.

• ReS pressure is 2250 pSia and stable.

• One charging pump is running, Letdown is at approximately 30 gpm.

• Addin!;; boric acid to the charging pump suction to maintain the desired rate of power reduction.

• Pressurizer Sprays in progress.

• C02/3 annunciator in alarm; D-37, "PZR PRESSURE SELECTED CHANNEL DEVIATION HI/LO"

• Channel "Y" Pressurizer Level and Pressure controlling normally.

Then, during the load reduction, Pressurizer Level Channel "X" fails to zero (0) and the following occur:

• All control systems respond as designed to the failure.

• C02/3 annunciator in alarm; A-38, "PRESSURIZER CH X LEVEL HIILO".

• C02/3 annunciator in alarm; C-38, "PRESSURIZER CH X LEVEL LO-LO".

• PPC alarms on Monitor #2 indicative of the instrument failure. Which of the following actions must the Unit Supervisor direct and why? Per the ARP for C-38; shift pressurizer heater control to channel "Y" and restore pressurizer heaters, to ensure adequate margin from DNB is maintained. Per SP-2602A RCS Leakage; deselect Pressurizer Level Channel "X" from the leak rate calculation, to ensure valid trending of RCS Leak Rate by the PPC. o C Per the ARP for A-38; place the standby charging pumps in "Pull-To-Lock", to prevent the rate of the plant downpower from accelerating above 50%/hr. o Per AOP-2575, Rapid Downpower; shift pressurizer heater control to channel "Y" and restore the pressure controller setpoint, to prevent a plant trip on TM/LP. J;;stification A -CORRECT: Even though Ch. "Y" is the contrOlling channel of PZR pressure, Ch. "X" failing to zero will trip all PZR heaters. heater control switch must be selected to ignore the failed channel and the heater breakers must be manually reclosed (they will auto close even though the automatic controls are calling for more heater output). If this is action is not taken in a very timely RCS pressure will drop below the minimum required by Tech. Spec. to ensure adequate DNB margin, due to the lower pressure setpoint necessary to "Forcing PZR B -WRONG: The high rate of power change exceeds the PPC program capabilities for calculating an accurate RCS leak Plausible:

The failed PZR instrument could impact the RCS leak rate calculation and potentially be of concern if the rate of change were C WRONG: The failed instrument effects heaters only and has no effect on the standby Plausible:

The alarms received for this failure would be identical to those received had the instrument been aligned to start the standby chargirg pumps, as implied by the ARP. If this occurred, the rate of power drop would accelerate D -WRONG: All PZR heaters are effected by this failure, not just the Backup heaters (as in a loss of control power). Restoring setpoint to a normal setting will not recover RCS pressure and the plant will trip on Plausible:

Although the AOP gives guidance to force sprays, it does not allow for heater recovery on a failed channel. OP-2204, Changes contains the detailed guidance used by operators globally to commence, and secure from, forcing pressurizer However, this guidance simply states to adjust controller setpoin!, as necessary, to maintain pressure at the desired I ARP-2590B-2150 Alarm C-38, PZR Level Comments and Question Modification History Bob K. -D-3/C (change to match given reactivity Modified stem and choice ..c.. to reflect only one charging pump selected to run, others in standby Also changed down power rate to 50%/hr to match the applicable Reactivity plan. -Bill M. D-3/C, K (Able to rule out distaractor C due to down power rate of 30%/hr, which does not match the stem. Change to Distractor more plausible.

May have resulted in a Changed down power rate in distractor

..c.. to 50%/hr per recommendation. Angelo -D-3/C: No NRC KIA SystemtEtA System 004 Chemical and Volume Control System Page 34 of 77 Printed on 10/28/2009 at 13:20 , Question #: 12 Question 10; 9600016 RO SRO o Student Handout? Lower Order? I-SRO Ques. # 12 Rev. 0 Selected for Exam Orlgln; Mod Past NRC Exam? A Rapid Downpower at 50%/hr is in progress due to an RCS leak in containment that exceeds the administrative limit. The following plant conditions presently exist:

• Plant power is 93% and dropping at the intended rate.

• PressLlrizer level is 65% and stable.

• ReS pressure is 2250 pSia and stable.

• One charging pump is running, Letdown is at approximately 30 gpm.

• Addin!;; boric acid to the charging pump suction to maintain the desired rate of power reduction.

• Pressurizer Sprays in progress.

• C02/3 annunciator in alarm; D-37, "PZR PRESSURE SELECTED CHANNEL DEVIATION HI/LO"

• Channel "Y" Pressurizer Level and Pressure controlling normally.

Then, during the load reduction, Pressurizer Level Channel "X" fails to zero (0) and the following occur:

• All control systems respond as designed to the failure.

• C02/3 annunciator in alarm; A-38, "PRESSURIZER CH X LEVEL HIILO".

• C02/3 annunciator in alarm; C-38, "PRESSURIZER CH X LEVEL LO-LO".

• PPC alarms on Monitor #2 indicative of the instrument failure. Which of the following actions must the Unit Supervisor direct and why? A Per the ARP for C-38; shift pressurizer heater control to channel "Y" and restore pressurizer heaters, to ensure adequate margin from DNB is maintained.

B Per SP-2602A RCS Leakage; deselect Pressurizer Level Channel "X" from the leak rate calculation, to ensure valid trending of RCS Leak Rate by the PPC. o C Per the ARP for A-38; place the standby charging pumps in "Pull-To-Lock", to prevent the rate of the plant downpower from accelerating above 50%/hr. o Per AOP-2575, Rapid Downpower; shift pressurizer heater control to channel "Y" and restore the pressure controller setpoint, to prevent a plant trip on TM/LP. J;;stification 1 A -CORRECT: Even though Ch. "Y" is the contrOlling channel of PZR pressure, Ch. "X" failing to zero will trip all PZR heaters. The heater control switch must be selected to ignore the failed channel and the heater breakers must be manually reclosed (they will not auto close even though the automatic controls are calling for more heater output). If this is action is not taken in a very timely manner, RCS pressure will drop below the minimum required by Tech. Spec. to ensure adequate DNB margin, due to the lower pressure control setpoint necessary to "Forcing PZR Sprays". B -WRONG: The high rate of power change exceeds the PPC program capabilities for calculating an accurate RCS leak rate. Plausible:

The failed PZR instrument could impact the RCS leak rate calculation and potentially be of concern if the rate of power change were less. C WRONG: The failed instrument effects heaters only and has no effect on the standby pumps. Plausible:

The alarms received for this failure would be identical to those received had the instrument been aligned to automatically start the standby chargirg pumps, as implied by the ARP. If this occurred, the rate of power drop would accelerate dramatically. D -WRONG: All PZR heaters are effected by this failure, not just the Backup heaters (as in a loss of control power). Restoring the setpoint to a normal setting will not recover RCS pressure and the plant will trip on TM/LP. Plausible:

Although the AOP gives guidance to force sprays, it does not allow for heater recovery on a failed channel. OP-2204, Load Changes contains the detailed guidance used by operators globally to commence, and secure from, forcing pressurizer sprays. However, this guidance simply states to adjust controller setpoin!, as necessary, to maintain pressure at the desired value. References I ARP-2590B-2150 Alarm C-38, PZR Level Lo-Lo Comments and Question Modification History J Bob K. -D-3/C (change to match given reactivity plan) Modified stem and choice .. c .. to reflect only one charging pump selected to run, others in standby Also changed the downpower rate to 50%/hr to match the applicable Reactivity plan. -RLC Bill M. D-3/C, K (Able to rule out distaractor C due to down power rate of 30%/hr, which does not match the stem. Change to 50%/hr. Distractor more plausible.

May have resulted in a 50/50) Changed down power rate in distractor

.. c .. to 50%/hr per recommendation.

RJA Angelo -D-3/C: No comrnents.

NRC KIA SystemtEtA System 004 Chemical and Volume Control System Page 34 of 77 Printed on 10/28/2009 at 13:20 Question #: 12 Question ID: 9600016 :"::1 RO li'l SRO Student Handout? li'l Lower Order? I-SRO Ques.1f 12 Rev. 0 li'l Selected for Exam Origin: Mod Past NRC Exam? Number A2.15 RO 3.5 SRO 3.7 CFR Link (CFR: 41.5/43/5/45/3/45/5)

Ability to (a) predict the *mpacls 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:

High or low PZR level 03/04/04 Approval Effective Date Setpolnt:

20% C-38 PRESSURIZER CHX lEVELLO LO AUTOMATIC:

FUNCTIONS IF "SEL SW" is in "X+Y" position.

all heaters dc-energize.

COBRECTIVE ACTIONS OBSERVE aCluallevcl on pressurizer lcvc\ recorder, LR-110, pressurizer level controllers

((,-03) and ppc. IE annunciator is not valid. SHIFT pressurizer level control to channel "Y." 2.1 SHIFT pressurizer heater control "SEL SW" to channel Y. VERIFY the following:

• Available backup charging pumps are running (C-02).

• Letdown 110w is at minimum of 28 gpm on "LTDN FWW. FI-202" (C-02). IF level cannot be restored or continues to lower, Refer To AOP 2568. "Reactor Coolant System Leak." WHEN annunciator clears, VERIFY all required heaters energize. WI lEN level rises (0 4';(; belo\'v set point, VERIFY second back up charging pump stops. WHEN level rises (0 3% below setpoint, VERIFY first back up charging pump stops. VERIFY levI..'! is n..'Storeu to set point. IF alarm was caused by channd X malfunctioning.

SUBMIT Trouble Report 10 I&C Department. Refer To Technical Specifications LeOs 3.3.3.5 and 3.3.3.R to determine ACTION Statement requirements.

SUf.PORTING INFORMATION Initiating Devices

• L,C-I10XL Computer Points

• L1IOX Possibk Causes

• Controller malfunction

• RCS inventory Inss Technical Spccil'icaliott.'\

LeOs: 3.4.4, 3.3.3.5 and 3.3.3.8 Procedures

• OP 2304A. "Volume Control Portion of CVCS"

• AOP 2568, "Reactor Coolant System Leak" Control Room Drawings

• 25203-32007, 57 Annunciator Card Location:

TBlO-J12 ARP 2590B-21S Rev. 000 Page 1 of 1 Page 35 of Printed on 10/28/2009 at 13:20 I Question #: 12 Question ID: 9600016 :"::1 RO li'l SRO Student Handout? li'l Lower Order? I-SRO Ques.1f 12 Rev. 0 li'l Selected for Exam Origin: Mod Past NRC Exam? Number A2.15 RO 3.5 SRO 3.7 CFR Link (CFR: 41.5/43/5/45/3/45/5)

Ability to (a) predict the *mpacls 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:

High or low PZR level I Approval Date 02/16/04 Effective Date 03/04/04 Setpolnt:

20% AUTOMATIC:

FUNCTIONS C-38 PRESSURIZER CHX lEVELLO LO I. IF "SEL SW" is in "X + Y" position.

all heaters dc-energize.

COBRECTIVE ACTIONS 1. OBSERVE aCluallevcl on pressurizer lcvc\ recorder, LR-110, pressurizer level controllers

((,-03) and ppc. 2. IE annunciator is not valid. SHIFT pressurizer level control to channel "Y." 2.1 SHIFT pressurizer heater control "SEL SW" to channel Y. 3. VERIFY the following:

• Available backup charging pumps are running (C-02).

• Letdown 110w is at minimum of 28 gpm on "LTDN FWW. FI-202" (C-02). 4. IF level cannot be restored or continues to lower, Refer To AOP 2568. "Reactor Coolant System Leak." 5. WHEN annunciator clears, VERIFY all required heaters energize.

6. WI lEN level rises (0 4';(; belo\'v set point, VERIFY second back up charging pump stops. 7. WHEN level rises (0 3% below setpoint, VERIFY first back up charging pump stops. H. VERIFY levI..'! is n .. 'Storeu to set point. 9. IF alarm was caused by channd X malfunctioning.

SUBMIT Trouble Report 10 I&C Department.

10. Refer To Technical Specifications LeOs 3.3.3.5 and 3.3.3.R to determine ACTION Statement requirements.

SUf.PORTING INFORMATION I. Initiating Devices 2. 3. 4. 5. 6. 7.

• L,C-I10XL Computer Points

• L1IOX Possibk Causes

• Controller malfunction

• RCS inventory Inss Technical Spccil'icaliott.'\

LeOs: 3.4.4, 3.3.3.5 and 3.3.3.8 Procedures

• OP 2304A. "Volume Control Portion of CVCS"

• AOP 2568, "Reactor Coolant System Leak" Control Room Drawings

• 25203-32007, 57 Annunciator Card Location:

TBlO-J12 Page 35 of 77 ARP 2590B-21S Rev. 000 Page 1 of 1 Printed on 10/28/2009 at 13:20 Question #: 12 Question ID: 9600016 RO li'l SRO Student Handout? li'l Lower Order? I-SRO Ques. /I 12 Rev. 0 li'l Selected for Exam Origin: Mod o Past NRC Exam? I I A-aS I JEilldicated high or low level was caused by controller or transmitter malfunction I (other than Reactor Regulating System inputs), PERFORM the following: SlUFf "LTDN FLOW CNTL, HIC-ll()"'

in "MAN" (C-02). ADJUST '-LTDN CNTL. HIC-llO" to stabilize Pressurizer k'\o'c1 Letdown flow IF desired, COMMENCE forcing Pressurizer sprays. SHIFf Pressurizer level control to channel "Y" (C-03). RESTORE Letdown to automatic as follows: 8.5.1 ADJUST bias to "0". using black thumbwhceL

8.5.2 SHiff

"LTDN FLOW CNTL.IHC-llO'*

to "AUTO." 85.3 ADJUST bias to restore Pressurizer level to setpoint.

85.4 SHIFf Pres,<;urizcr heater control "SEL SW" to channel "Y." As RESET the following Pressurizer heater breakers:

.. "PROP HTR GROUP .. "PROP HTR GROUP * "IBACKUP HTRS GROUP .. "BACKUP IITRS GROUP "BACKUP llTRS GROUP 3" "BACKUP HTRS GROUP 4" [F instrument malfunction is determined not to be the cause oflow level, Refer To the foUowing.

a,.,> applicable:

2512, "'Loss of AU Charging'"

• AOP 2568, "Reactor Coolant System .. AOP 2569, "Steam Generator Thbe IF <lL'l.uallevel

\vas high or low, VERIFY level is restored to normal. IE alarm was caused by channel "X" malfunctioning, SUBMIT Trouble Report to lnstrumcntatlon

& Control Department.

ARP 2590B-213 Page 36 of Printed on 10/28/2009 at 13:20 Question #: 12 Question ID: 9600016 I-SRO Ques. /I 12 Rev. 0 RO li'l SRO li'l Selected for Exam Student Handout? Origin: Mod li'l Lower Order? o Past NRC Exam? I A-aS I K JEilldicated high or low level was caused by controller or transmitter malfunction I (other than Reactor Regulating System inputs), PERFORM the following:

RI SlUFf "LTDN FLOW CNTL, HIC-ll()"'

in "MAN" (C-02). 8.2 ADJUST '-LTDN CNTL. HIC-llO" to stabilize Pressurizer k'\o'c1 and Letdown flow (C-02). IU IF desired, COMMENCE forcing Pressurizer sprays. 8.4 SHIFf Pressurizer level control to channel "Y" (C-03). B.S RESTORE Letdown to automatic as follows: 8.5.1 ADJUST bias to "0". using black thumbwhceL

8.5.2 SHiff

"LTDN FLOW CNTL.IHC-llO'*

to "AUTO." 85.3 85.4 ADJUST bias to restore Pressurizer level to setpoint.

SHIFf Pres,<;urizcr heater control "SEL SW" to channel "Y." 9. As RESET the following Pressurizer heater breakers:

.. "PROP HTR GROUP I" .. "PROP HTR GROUP 2'-* "IBACKUP HTRS GROUP J" .. "BACKUP IITRS GROUP 2" * "BACKUP llTRS GROUP 3" * "BACKUP HTRS GROUP 4" 10. [F instrument malfunction is determined not to be the cause oflow level, Refer To the foUowing.

a,.,> applicable:

• 2512, "'Loss of AU Charging'"

• AOP 2568, "Reactor Coolant System Leak" .. AOP 2569, "Steam Generator Thbe Leak" 11. IF <lL'l.uallevel

\vas high or low, VERIFY level is restored to normal. 12. IE alarm was caused by channel "X" malfunctioning, SUBMIT Trouble Report to lnstrumcntatlon

& Control Department.

ARP 2590B-213 I Page 36 of 77 Printed on 10/28/2009 at 13:20 Question #: 13 Question ID; 53730 RO SRO Student Handout? Ii'! Lower Order? l-SRO Ques. # 13 Rev. 4 Selected for Exam Origin: Bank Past NRC Exam? A reactor startup is in progress using CEA withdrawal.

The RO has just stopped withdrawing Group # 7 CEAs and makes the following announcements:

• The reactor is critical.

• Startup rate is positive and stabilizing at -1.5 DPM. Which of tile following actions should the Reactivity SRO A Commence Emergency Boration until the reactor is o B Insert the Group #7 CEAs to lower the startup rate below 0.5 DPM. C Trip the reactor and Go to EOP 2525, "Standard Post Trip Actions".

D Insert all CEAs per OP-2206, "Reactor Shutdown" and notify RE. iJUstification*****l C CORRECT OP 2202, "Reactor Startup" Conditional Actions require 1l:im!i.!:Jg the reactor, and transitioning to EOP-2525, if a SUR of 1.0 DPM is sustained. A -WRONG: This action may be acceptable if an abnormal count rate was due to an uncontrolled cooldown during the reactor Plausible:

This action would stop the power rise and shutdown the reactor, but it is unacceptable with a high B -WRONG: This is acceptable if SUR has not yet exceeded 1.0 Plausible:

Correct action if SUR briefly spiked above 1.0 DPM or, stabilized just below 1.0 D -WRONG: To slow for a high startup rate, even if it is just barely above the threshold for Plausible:

Correct action if criticality were occurring earlier than predicted (outside of acceptable I OP-2202; Pg. 36, Attach. 5, Rx Startup Conditional Comments and Question Modification Bob K. -D-3/W (Did not remember "trip" Bill M. -D-2/w, K (Did not remember the actual trip criteria Angelo -D*3/C; No NRC KIA System/E/A System 012 Reactor Protection System -..... .... --i]

System 2.4 Emergency Procedures IPlan Number 2.4.4 RO 4.5 SRO 4.7 CFR Link (CFR:

Ability to recognize abnormal Indications for system operating parameters that are entry-level conditions for emergency and operating NRC KIA Generic Page 37 of 77 Printed on 10/28/2009 at 13:20 Question #: 13 l-SRO Ques. # 13 (NQ " Question ID; 53730 RO SRO Rev. 4 Selected for Exam Student Handout? Ii'! Lower Order? Origin: Bank Past NRC Exam? A reactor startup is in progress using CEA withdrawal.

The RO has just stopped withdrawing Group # 7 CEAs and makes the following announcements:

• The reactor is critical.

• Startup rate is positive and stabilizing at -1.5 DPM. Which of tile following actions should the Reactivity SRO direct? A Commence Emergency Boration until the reactor is subcritical.

o B Insert the Group #7 CEAs to lower the startup rate below 0.5 DPM. C Trip the reactor and Go to EOP 2525, "Standard Post Trip Actions".

D Insert all CEAs per OP-2206, "Reactor Shutdown" and notify RE. iJUstification*****l C CORRECT OP 2202, "Reactor Startup" Conditional Actions require 1l:im!i.!:Jg the reactor, and transitioning to EOP-2525, if a SUR of 1.0 DPM is sustained. A -WRONG: This action may be acceptable if an abnormal count rate was due to an uncontrolled cooldown during the reactor startup. Plausible:

This action would stop the power rise and shutdown the reactor, but it is unacceptable with a high SUR. B -WRONG: This is acceptable if SUR has not yet exceeded 1.0 DPM. Plausible:

Correct action if SUR briefly spiked above 1.0 DPM or, stabilized just below 1.0 DPM. D -WRONG: To slow for a high startup rate, even if it is just barely above the threshold for "excessive".

Plausible:

Correct action if criticality were occurring earlier than predicted (outside of acceptable limits). References I OP-2202; Pg. 36, Attach. 5, Rx Startup Conditional Actions Comments and Question Modification History Bob K. -D-3/W (Did not remember "trip" criteria)

Bill M. -D-2/w, K (Did not remember the actual trip criteria value.) Angelo -D-3/C; No comments.

NRC KIA System/E/A System 012 Reactor Protection System -..... .... --i]

NRC KIA Generic System 2.4 Emergency Procedures IPlan Number 2.4.4 RO 4.5 SRO 4.7 CFR Link (CFR: 41.10/43.2/45.6)

Ability to recognize abnormal Indications for system operating parameters that are entry-level conditions for emergency and abnormal operating prOC1'ldures.

Page 37 of 77 Printed on 10/28/2009 at 13:20 SROExam Questions Question #: 13 Student Handout? Lower Order?Question ID: 53730 RO SRO II-SRO Ques. II '13 4 Selected for Exam Origin: Bank Past NRC Exam? Attachment 5 Reactor Startup Conditional Actions (Sheet 1 of 3) 1. at any time, the following

"'onditions occur, PERFORM the specified action: IE T avg lowers to between 515 and 525 "F ANn the reactor is critical, Refer to OP 26t9A-OOl, "Control Room Daily SUlveiUancc;'

and RECORD RCS temperature once every hour. IF Tavg lowers to Jess than 515 c'F AND the reactor is critical, PERFORM the following: RAISE Tavg to than 515 OF within 15 minutes. IF Tavg is not greater than 515 OF within is minutes, PLACE plant in HOT STANDBY condition within the next 15 minutes. Refer lb T/SLCO :U.1.5 and DETERMINE applicability. an uncontrolled cooldown occurs (Tc less than 500 OF), PERFORM the following: TRIP reactor and INITIATE EOP "Stanoaro Post 11'ip Actions." STOP one of the 4 operating RCPs (C-04). Refer Tb AOP 2558, "Emergency Horation," and INITIATE emergency boration. Refer lb TIS LCO 3.4.9.1 and DETERMINE applicability. at any time a sustaincd SUR of 1.0 dpm is tlchicvcd, TRIP reactor Go To EOP 2525, "Standard Post Trip IE at nny time during reactor startup, it appears that criticality is reached, or is predicted to be reached, outside plus or minus 0.5%.6.Q (0.9% liQ for initial startup after refueling) band of ECp, PERFORM the following: INSERT all CEA regulaling groups in sequence (C-04). REQUEST Chemistry Department sample and determine RCS boron c(mccmration. INITIATE a CR for Reactivity Management tracking. Refer To OP 2208, "Reactivity Calculations" and, independent of CEA positron, VERIFY adequate SHUTDOWN MARGIN using OP 2208-013, "Shutdmm Margin Determination." NOTIFY Reactor Engineering.

OP 22HZ Level of Use STOP THINK ACT REVIE.W Rcv.02i-06 Continuous 36 of 56 Page 38 of Printed on 10/28/2009 at 13:20 SROExam Questions Question #: 13 II-SRO Ques. II '13 1. *

• Question ID: 53730 RO SRO Student Handout? Rev. 4 Selected for Exam Origin: Bank Attachment 5 Reactor Startup Conditional Actions (Sheet 1 of 3) Lower Order? Past NRC Exam? at any time, the following

"'onditions occur, PERFORM the specified action: IE T avg lowers to between 515 and 525 "F ANn the reactor is critical, Refer to OP 26t9A-OOl, "Control Room Daily SUlveiUancc;'

and RECORD RCS temperature once every hour. IF Tavg lowers to Jess than 515 c'F AND the reactor is critical, PERFORM the following:

• RAISE Tavg to than 515 OF within 15 minutes.

• IF Tavg is not greater than 515 OF within is minutes, PLACE plant in HOT STANDBY condition within the next 15 minutes.

• Refer lb T/SLCO :U.1.5 and DETERMINE applicability.

• an uncontrolled cooldown occurs (Tc less than 500 OF), PERFORM the following:

• TRIP reactor and INITIATE EOP Actions." "Stanoaro Post 11'ip

• STOP one of the 4 operating RCPs (C-04).

• Refer Tb AOP 2558, "Emergency Horation," and INITIATE emergency boration.

• Refer lb TIS LCO 3.4.9.1 and DETERMINE applicability.

2. at any time a sustaincd SUR of 1.0 dpm is tlchicvcd, TRIP reactor and Go To EOP 2525, "Standard Post Trip Actions." 3. IE at nny time during reactor startup, it appears that criticality is reached, or is predicted to be reached, outside plus or minus 0.5%.6.Q (0.9% liQ for initial startup after refueling) band of ECp, PERFORM the following:

3.1 INSERT

all CEA regulaling groups in sequence (C-04). 3.2 REQUEST Chemistry Department sample and determine RCS boron c(mccmration.

3.3 INITIATE

a CR for Reactivity Management tracking.

3.4 Refer

To OP 2208, "Reactivity Calculations" and, independent of CEA positron, VERIFY adequate SHUTDOWN MARGIN using OP 2208-013, "Shutdmm Margin Determination." 3.5 NOTIFY Reactor Engineering.

Level of Use Continuous STOP THINK ACT Page 38 of 77 REVIE.W OP 22HZ Rcv.02i-06 36 of 56 Printed on 10/28/2009 at 13:20

0 Question #: 14 Question ID: 9000021 [] RO IiI'l SRO Student Handout? IiI'l Lower Order? [*SRO Ques. # 14 Rev. o IiI'l Selected for Exam Origin: New Past NRC Exam? The plant IS in Mode 6 with the following conditions;

• Core m*load in progress and approximately half way completed.

• !lA!I LPSI pump running for Shutdown Cooling (SOC) operation.

• !lB" LPSI pump in standby, aligned for SOC use. * !lA" train of Spemt Fuel Pool (SFP) cooling in service. Then, "A" LPSI pump is lost due to a breaker fault. When the "B" LPSI pump is started, it seizes and on breaker The Unit Supervisor (US) then directs the RO to recover SOC using the !lB" Containment Spray (CS) Which of the following additional directions must the US give, while SOC flow is being supplied by a IiI'l A All fuel movement In containment must remain secured. B SOC supplementing of SFP cooling must be secured. C Containment must remain evacuated of non-essential personnel.

D D Containment Closure must be fully set with all access doors closed. Justification I A

• CORRECT: When SDC is being supplied by a CS pump, does not constitute an OPERABLE train of SDC. Therefore, all movement in CTMT must be 8* WROt-<G; A CS pump has the capacity to supply SDC flow and supplement SFP cooling, but just barely, if a train of SFP cooling in Plausible; Examinee may believe with the limited capacity of a CS pump (compared to a LPSI pump), supplementing SFP cooling not possible (it would not be at the beginning of the C -WRONG; CTMT evacuation would probably occur if a CS pump needed to be used for recovery as is required if SDC flow be restored in 15 minutes. However, once the CS pump restores flow, evacuation is no longer Plausible; Examinee may realize SDC is not considered fully operational being supplied by a CS pump and, therefore, evacuation of CTMT be maintained.

This is true if a plant heatup to over 190°F D

• WRONG; CTMT Closure must be set on initial loss of SOC flow, but once heat removal is regained, it may Plausible; Examinee moly believe that with a CS pump supplying RHR needs (SDC not operable), CTMT closure must be References AOP-2572, Pg. 3, Discussion Section and Pages 28 & Comments and Question Modification History Bob K .* D*3/W (Did not remember SOC not Operable with CS Pump Corrected a typo in Distractor B.* Bill M. -D-4/w, G (Did not realize that SOC was inoperable, which requires CORE ALTERATIONS to remain Angelo* D-5/C. Difficult but NRC KJA System/E/A System 026 Containment Spray System (CSS) Gl'!lrll'!ric KIA Selected NRC KJA Generic System 2.4 Emergency Procedures IPlan Number 2.4.9 RO 3.8 SRO 4.2 CFR Link (CFR: 41.10 143.5/45.13Knowledge of 'ow powerlshutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal)

Page 39 of 77 Printed on 10/28/2009 at 13:20 0 Question #: 14 [*SRO Ques. # 14 Question ID: 9000021 [] RO IiI'l SRO Student Handout? IiI'l Lower Order? Rev. o IiI'l Selected for Exam The plant IS in Mode 6 with the following conditions;

• Core m*load in progress and approximately half way completed.

• !lA!I LPSI pump running for Shutdown Cooling (SOC) operation.

• !lB" LPSI pump in standby, aligned for SOC use. * !lA" train of Spemt Fuel Pool (SFP) cooling in service. Origin: New Past NRC Exam? Then, "A" LPSI pump is lost due to a breaker fault. When the "B" LPSI pump is started, it seizes and trips on breaker overload.

The Unit Supervisor (US) then directs the RO to recover SOC using the !lB" Containment Spray (CS) pump. Which of the following additional directions must the US give, while SOC flow is being supplied by a CS pump? IiI'l A All fuel movement In containment must remain secured. B SOC supplementing of SFP cooling must be secured. C Containment must remain evacuated of non-essential personnel.

D D Containment Closure must be fully set with all access doors closed. Justification I A

• CORRECT: When SDC is being supplied by a CS pump, does not constitute an OPERABLE train of SDC. Therefore, all fuel movement in CTMT must be secured. 8* WROt-<G; A CS pump has the capacity to supply SDC flow and supplement SFP cooling, but just barely, if a train of SFP cooling is in operation.

Plausible; Examinee may believe with the limited capacity of a CS pump (compared to a LPSI pump), supplementing SFP cooling is not possible (it would not be at the beginning of the outage). C -WRONG; CTMT evacuation would probably occur if a CS pump needed to be used for recovery as is required if SDC flow cannot be restored in 15 minutes. However, once the CS pump restores flow, evacuation is no longer necessary.

Plausible; Examinee may realize SDC is not considered fully operational being supplied by a CS pump and, therefore, require evacuation of CTMT be maintained.

This is true if a plant heatup to over 190°F occurs. D

• WRONG; CTMT Closure must be set on initial loss of SOC flow, but once heat removal is regained, it may stop. Plausible; Examinee moly believe that with a CS pump supplying RHR needs (SDC not operable), CTMT closure must be maintained.

References

• 1 AOP-2572, Pg. 3, Discussion Section and Pages 28 & 31 Comments and Question Modification History I Bob K .* D*3/W (Did not remember SOC not Operable with CS Pump supplying).

Corrected a typo in Distractor B .* RJA Bill M. -D-4/w, G (Did not realize that SOC was inoperable, which requires CORE ALTERATIONS to remain suspended.)

Angelo* D-5/C. Difficult but fair. NRC KJA System/E/A System 026 Containment Spray System (CSS) Gl'!lrll'!ric KIA Selected NRC KJA Generic System 2.4 Emergency Procedures IPlan Number 2.4.9 RO 3.8 SRO 4.2 CFR Link (CFR: 41.10 143.5/45.13)

Knowledge of 'ow powerlshutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Page 39 of 77 Printed on 10/28/2009 at 13:20 Question #: 14 RO V SRO Student Handout? Lower Order? Question ID: 9000021 l-SRO Ques. if '14 Rev. 0 Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 AOP 2572 Revision 009-04 Loss of Shutdown Cooling Page 3 of70 1.0 PURI'OSE Ll Objective This procedure provides actions for recovering from a partial or total loss of shutdO\\n cooling. 1.2 Discussino During SDC operation.

there may not he flow past the loop RTDs. Core inlet and outlet temperatures are accurately measured during those conditions using SDC to RCS temperature, T35 t Y, and ReS to SDC temperature, T351X, respectively.

The average of these indicators provides a temperature that is equivalent to the average ReS temperature in the core. Containment Closure if; established when aU of the following conditions exist: The equipmcnt door is dosed and held in place by a minimum of four bolts. A minimum of one door in each airlock is closed . Each penetration providing direct access from the containment atmosphere to the outside atmospherc is either: Closed by a manual or automatic isolation valve, blind flange, or equivalent, or Cupable of being closed under administrative control The use of the CS pump for decay heat removal doc') not meet the definition of an Operable SDC train (LCO 3.9.8). Therefore no fuel movement is permitted when a CS pump is aligned to SDC per this procedure . .1 .3 AppHcubility This procedure is applicable in MODEs 4, 5, 6 and Defucled.

Use of the CS pumps is limited to MODE 6 <lIld Defucled.

Level of Use STOP THINK ACT REVIEW Continuous Page 40 of Printed on 10/28/2009 at 13:20 Question #: 14 I Question ID: 9000021 RO V SRO Student Handout? Lower Order? l-SRO Ques. if '14 Rev. 0 Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 Loss of Shutdown Cooling AOP 2572 Revision 009-04 Page 3 of70 1.0 PURI'OSE Ll Objective This procedure provides actions for recovering from a partial or total loss of shutdO\\n cooling. 1.2 Discussino During SDC operation.

there may not he flow past the loop RTDs. Core inlet and outlet temperatures are accurately measured during those conditions using SDC to RCS temperature, T35 t Y, and ReS to SDC temperature, T351X, respectively.

The average of these indicators provides a temperature that is equivalent to the average ReS temperature in the core. Containment Closure if; established when aU of the following conditions exist:

• The equipmcnt door is dosed and held in place by a minimum of four bolts.

• A minimum of one door in each airlock is closed .

• Each penetration providing direct access from the containment atmosphere to the outside atmospherc is either:

• Closed by a manual or automatic isolation valve, blind flange, or equivalent, or

• Cupable of being closed under administrative control The use of the CS pump for decay heat removal doc') not meet the definition of an Operable SDC train (LCO 3.9.8). Therefore no fuel movement is permitted when a CS pump is aligned to SDC per this procedure . .1 .3 AppHcubility This procedure is applicable in MODEs 4, 5, 6 and Defucled.

Use of the CS pumps is limited to MODE 6 <lIld Defucled.

Level of Use Continuous STOP THINK ACT REVIEW Page 40 of 77 Printed on 10/28/2009 at 13:20 Question #: Lower Order? l-SRO Ques. # 14 Rev. 0 Selected for Exam Origin: New Past NRC Exam? Question ID; 9000021 RO SRO Student Handout? Millstone Unit 2 Loss of Shutdown Cooling AOP 2572 Revision 009-04 Page 28 of70 INSTRUCtIONS 5.10 (continued)

2) CHECK prcssurc at 2-RW -66, "SFPClRW Purificathm Return Sample Stop," les.., than 30 psig. Table to FPC in service Fuel Assemblies Off-Loaded Flow to SFP (}-80 81-170 171-2l.7 1100 b. IF a CS pump is in service, PERFORM the following:

1) THROTTLE the following valves to obtain Table I,{) flow splits:

• 51-615, "LPSI INJ VLV5" LOOP IA

• SI -625.. 4'LPSI INJ VLVS" LOOP 1 B

• SI-635, "LPSI INJ VLVS" LOOP 2/\

• SI-645, "LPSIINJ VLVS" LOOP 2B

• 2-RW-1S, "SDC to SF PC SlOp" (continue) o 300 600 Level of Use Continuous STOP THINK ACT CONTINGENCY ACTIONS 2) IFprcs..,urc at 2-RW-66 is greater than 30 psig, PERFORM the following:

a. THROTrLE2-RW-15, "SDC to SFPC SLOp" to obtain less than 30 psig at 2-RW-66. b. CONTACT Engineering for additional guidance on decay heat removal. Flow splits with SFPC !lot in service Flow to RFP Flow to SFP 1250 450 750 950 400 1300 REVIEW Page 41 of 77 Printed on 10/28/2009 at 13:20 Question #: Question ID; 9000021 RO SRO Student Handout? Lower Order? l-SRO Ques. # 14 Rev. 0 Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 Loss of Shutdown Cooling AOP 2572 Revision 009-04 Page 28 of70 INSTRUCtIONS 5.10 (continued)

2) CHECK prcssurc at 2-RW -66, "SFPClRW Purificathm Return Sample Stop," les.., than 30 psig. Table to FPC in service Fuel Assemblies Off-Loaded Flow to SFP (}-80 81-170 171-2l.7 1100 b. IF a CS pump is in service, PERFORM the following:

1) THROTTLE the following valves to obtain Table I,{) flow splits:

• 51-615, "LPSI INJ VLV5" LOOP IA

• SI -625 .. 4'LPSI INJ VLVS" LOOP 1 B

• SI-635, "LPSI INJ VLVS" LOOP 2/\

• SI-645, "LPSIINJ VLVS" LOOP 2B

• 2-RW-1S, "SDC to SF PC SlOp" (continue) o 300 600 Level of Use Continuous STOP THINK ACT CONTINGENCY ACTIONS 2) IFprcs..,urc at 2-RW-66 is greater than 30 psig, PERFORM the following:

a. THROTrLE2-RW-15, "SDC to SFPC SLOp" to obtain less than 30 psig at 2-RW-66. b. CONTACT Engineering for additional guidance on decay heat removal. Flow splits with SFPC !lot in service Flow to RFP Flow to SFP 1250 450 750 950 400 1300 REVIEW Page 41 of 77 Printed on 10/28/2009 at 13:20 Question #: 14 1 Question ID: 9000021 RO SRO Student Handout? Lower Order? l-SRO Que,. # 14 Rev. 0 Selected for Exam Origin: New Past NRC Exam? ! MiJistone Unit 2 AOP 2572 Revision 009-04 of Shutdown Cooling Page 31 of70 INSTR UC'TIONS CONTINGENCY ACTIONS NOTE 2-SI-306 has designed leakby that diverts flow around the SOC heat exchangers and could challenge heat removal with CS pumps supplying SOc.

If a CS pump is in on SOC and sufficient cooling: cannot be obtained with 2-SI-306c1oscd, CLOSE the applicable LPSI to SDC heat exchanger isolation valve: 2-*S1 -452, LPSI Pump Discharge to SDC Heat Exchanger 2-51 -453, LPSI Pump Discharge 10 "D" SDC Ileat Exchanger

_5.18 REPEAT steps 5.12 through 5.17 as needed to c{mtrol RCS temperature.

_5.19 WHEN ready to shift SOC from a CS pump to a LPSI pump. PERFORM Attachment

8. "Realigning LPSI to Supply SDC and SFPC." _5.20 WliEti RCS pressure is stable AND RCS temperature is less than 200"F and stable. STOP Containment Closure activities.

_5.21 Go lh Section 10.0. Level of Use STOP THINK ACT REVIEW Continuous Page 42 of Printed on 10/28/2009 at 13:20 ! Question #: 14 1 Question ID: 9000021 RO SRO Student Handout? Lower Order? l-SRO Que,. # 14 Rev. 0 Selected for Exam Origin: New Past NRC Exam? MiJistone Unit 2 AOP 2572 Revision 009-04 of Shutdown Cooling Page 31 of70 INSTR UC'TIONS CONTINGENCY ACTIONS NOTE 2-SI-306 has designed leakby that diverts flow around the SOC heat exchangers and could challenge heat removal with CS pumps supplying SOc.

If a CS pump is in on SOC and sufficient cooling: cannot be obtained with 2-SI-306c1oscd, CLOSE the applicable LPSI to SDC heat exchanger isolation valve:

• 2-*S1 -452, LPSI Pump Discharge to SDC Heat Exchanger

• 2-51 -453, LPSI Pump Discharge 10 "D" SDC Ileat Exchanger

_5.18 REPEAT steps 5.12 through 5.17 as needed to c{mtrol RCS temperature.

_5.19 WHEN ready to shift SOC from a CS pump to a LPSI pump. PERFORM Attachment

8. "Realigning LPSI to Supply SDC and SFPC." _5.20 WliEti RCS pressure is stable AND RCS temperature is less than 200"F and stable. STOP Containment Closure activities.

_5.21 Go lh Section 10.0. Level of Use Continuous STOP THINK ACT Page 42 of 77 REVIEW Printed on 10/28/2009 at 13:20 (N'()"f Question #: 15 Question ID: 9000010 =J RO SRO Student Handout? Lower Order? l-SRO Que". 15 Rev. o Selected for Exam Origin: New Past NRC Exam? The requimd surveillance must be performed after repairs were made to the "A" Service Water Strainer Flush Valve, 2-SW-90A.

The completed surveillance indicates the valve stroke time is slightly above the Maximum 'Normal Limit", but is below the Maximum "Acceptable Limit". All other parameters are within the "Acceptable Limits". Which of the following describes the condition of the "A" Service Water Pump and the required action? The "A" Service Water Pump remains inoperable.

Perform the required repairs to the "A" Service Water Strainer Flush Valve, 2-SW-90A, then perform the required surveillances to restore the "A" Service Water Pump to OPERABLE. The "A" Service Water Pump remains inoperable.

Obtain a different set of test equipment and immediately the "A" Service Water Strainer Flush Valve, 2-SW-90A, again to verify that the previous data was accurate.

D The "A" Service Water Pump is considered OPERABLE.

Place the "A" Service Water Pump Strainer in an Testing Program and test weekly to ensure the "A" Service Water Strainer Flush Valve, 2-SW-90A, remains OPERABLE. The "A" Service Water Pump is considered OPERABLE.

If the "A" Service Water Strainer Flush Valve, 2-SW*90A, exceeds the Maximum "Normal Limit" on an immediate retest, then declare the "A" Service Water Pump inoperable.

Justification D IS CORRECT; Per SP 2612A, "An Service Water Pump Tests, Attachment 2, the first failure of a valve stroke test to be within "Normal" limits does not render that component inoperable; however, an immediate retest must be performed.

If a second failure "Normal" time lim't has occurred, then the component is A is incorrect; Even though the Service Water Pump did not meet the "Normal" limit criteria it may be considered however, a second set of data must be taken. Repairs are NOT Plausible because the examinee may consider the component inoperable from the first set of failed data. If the component considered inoJerable, tllen typically, a component must be repaired to restore it to B is incorrect; Even though the Service Water Pump did not meet the "Normal" limit Criteria the surveillance procedure allows it to considered OPERABLE.

Different test equipment may be obtained to verify the previous Plausible because the examinee may consider the component inoperable from the first set of failed data. Verifying previous data different test equipment is allowed by procedure and is C is incorrect; The surveillance procedure allows the Service Water Pump to be considered OPERABLE.

The Service Water may be tested on a more frequent basis; however, the stoke time must be performed immediately after the initial failure to meet "Normal" Plausible because the examinee may not be aware of the need to immediately perform a second stoke time I SP 2612A, "A" Service Water Pump Comments and Question Modification History Bob K. -D-3/W (Reword answer to say "immediate Reworded correct answer to include the term "immediate retest". Bill M. -D-31W. K (Didn'! realize an immediate retest was Angelo -D-4/C; No NRC KiA System/E/A System 076 Service Water System (SWS) NRC KiA Generic System 2.2 Equipment Control Number 2.2.12 RO 3.7 SRO 4.1 CFR Link (CFR: 41.10/45.13) Knowledge of surveillance procedures.

Page 43 of Printed on 10/28/2009 at 13:20 Question #: 15 l-SRO Que". 15 Question ID: 9000010 =J RO SRO Rev. o Selected for Exam Student Handout? Lower Order? Origin: New Past NRC Exam? The requimd surveillance must be performed after repairs were made to the "A" Service Water Strainer Flush Valve, 2-SW-90A.

The completed surveillance indicates the valve stroke time is slightly above the Maximum 'Normal Limit", but is below the Maximum "Acceptable Limit". All other parameters are within the "Acceptable Limits". Which of the following describes the condition of the "A" Service Water Pump and the required action? A The "A" Service Water Pump remains inoperable.

Perform the required repairs to the "A" Service Water Strainer Flush Valve, 2-SW-90A, then perform the required surveillances to restore the "A" Service Water Pump to OPERABLE.

B The "A" Service Water Pump remains inoperable.

Obtain a different set of test equipment and immediately the "A" Service Water Strainer Flush Valve, 2-SW-90A, again to verify that the previous data was accurate.

D C The "A" Service Water Pump is considered OPERABLE.

Place the "A" Service Water Pump Strainer in an Testing Program and test weekly to ensure the "A" Service Water Strainer Flush Valve, 2-SW-90A, remains OPERABLE.

D The "A" Service Water Pump is considered OPERABLE.

If the "A" Service Water Strainer Flush Valve, 2-SW*90A, exceeds the Maximum "Normal Limit" on an immediate retest, then declare the "A" Service Water Pump inoperable.

Justification J D IS CORRECT; Per SP 2612A, "An Service Water Pump Tests, Attachment 2, the first failure of a valve stroke test to be within the "Normal" limits does not render that component inoperable; however, an immediate retest must be performed.

If a second failure of "Normal" time lim't has occurred, then the component is inoperable.

A is incorrect; Even though the Service Water Pump did not meet the "Normal" limit criteria it may be considered OPERABLE; however, a second set of data must be taken. Repairs are NOT required.

Plausible because the examinee may consider the component inoperable from the first set of failed data. If the component is considered inoJerable, tllen typically, a component must be repaired to restore it to OPERABLE.

B is incorrect; Even though the Service Water Pump did not meet the "Normal" limit Criteria the surveillance procedure allows it to be considered OPERABLE.

Different test equipment may be obtained to verify the previous data. Plausible because the examinee may consider the component inoperable from the first set of failed data. Verifying previous data with different test equipment is allowed by procedure and is correct. C is incorrect; The surveillance procedure allows the Service Water Pump to be considered OPERABLE.

The Service Water Pump may be tested on a more frequent basis; however, the stoke time must be performed Immediately after the initial failure to meet the "Normal" limit. Plausible because the examinee may not be aware of the need to immediately perform a second stoke time test. References I SP 2612A, "A" Service Water Pump Tests Comments and Question Modification History I Bob K. -D-3/W (Reword answer to say "immediate retest").

Reworded correct answer to include the term "immediate retest". RJA Bill M. -D-31W. K (Didn'! realize an immediate retest was required.)

Angelo -D-4/C; No comments.

NRC KiA System/E/A System 076 Service Water System (SWS) NRC KiA Generic System 2.2 Equipment Control Number 2.2.12 RO 3.7 SRO 4.1 CFR Link (CFR: 41.10/45.13)

Knowledge of surveillance procedures.

Page 43 of 77 Printed on 10/28/2009 at 13:20 Question #: 1S Student Handout? Lower Order? Question ID; 9000010 RO SRO l-SRO if 15 Rev. o Selected for Exam Origin; New IF testing 2-SW-90A "W SERVICE WATER PUMP STRAINER FLUSH," PERFORM the foUo\\-ing:

NOTE l'NO Operators are required to lime strainer llush valves due to location of switches and quick operation of valves. a. Refer 'Ii> OP 232HA "Sodium Hypochlorite System," and VERIFY Sodium Hypochlorite Injection to Service Water I @ Pump is terminated.

h. LOG ENTRY inlo TSAS 3.7.4.1 and TRMAS 7.1.21.A Co ill at any time, valve does not stroke fullv. Go To Attachn;cnt I. " d. PLACE /!\' Service Water Pump Strainer control switch in "HAND" (C-58A). I@ e. PRESS '1\" Service Water Pump Strainer "START" Inltton and MEASURE open stroke time. f. RECORD 2-SW-90A open stroke time on SP 2612A-mH.

g. PRESS Service Water Pump Strainer "STOP" button and MEASURE close stroke time. h. RECORD 2-SW-t)OAclose stroke time on SP 2612A-003.

I. PLACE '/!\ Service Water Pump Strainer control switch in 'J\unY' (C-SMA). 1[::;\I 0 j. DOCUMENT 2-SW-I}OA "Normal" limits Results on SP 2612A-003.

SP 2612A STOP THINK ACT REVIEW Rev.OlO-OS H of 36 Level of

k. DOCUMENT 2-SW-9UA Operational Readiness Results on SP 2612A-003.

I. IF Operational Readiness Results "UNSAT;' Go To Attachment

1. m. IE "Normal" limits Results "UNSAT," Go To Attachment

2. n. LOG EXIT from TSAS 3.7.4.1 and TRMAS 7. L2 LA. Page 44 of Printed on 10/28/2009 at 13:20 Question #: 1S l-SRO if 15 Question ID; 9000010 RO SRO Student Handout? Rev. o Selected for Exam Origin; New 4.1.5 IF testing 2-SW-90A "W SERVICE WATER PUMP STRAINER FLUSH," PERFORM the foUo\\-ing:

NOTE Lower Order? l'NO Operators are required to lime strainer llush valves due to location of switches and quick operation of valves. Level of Use Continuous

a. Refer 'Ii> OP 232HA "Sodium Hypochlorite System," and VERIFY Sodium Hypochlorite Injection to Service Water I @ Pump is terminated.

h. LOG ENTRY inlo TSAS 3.7.4.1 and TRMAS 7.1.21.A Co ill at any time, valve does not stroke fullv. Go To Attachn;cnt I. " d. PLACE /!\' Service Water Pump Strainer control switch in "HAND" (C-58A). e. PRESS '1\" Service Water Pump Strainer "START" Inltton and MEASURE open stroke time. f. RECORD 2-SW-90A open stroke time on SP 2612A-mH.

g. PRESS Service Water Pump Strainer "STOP" button and MEASURE close stroke time. h. RECORD 2-SW-t)OAclose stroke time on SP 2612A-003.

I@ PLACE '/!\ Service Water Pump Strainer control switch in 1[::;\ 'J\unY' (C-SMA). I 0 I. j. DOCUMENT 2-SW-I}OA "Normal" limits Results on SP 2612A-003.

SP 2612A STOP THINK ACT REVIEW Rev.OlO-OS H of 36 k. DOCUMENT 2-SW-9UA Operational Readiness Results on SP 2612A-003.

I. IF Operational Readiness Results "UNSAT;' Go To Attachment

1. m. IE "Normal" limits Results "UNSAT," Go To Attachment

2. n. LOG EXIT from TSAS 3.7.4.1 and TRMAS 7. L2 LA. Page 44 of 77 Printed on 10/28/2009 at 13:20

• Question #: II-SRO Ques. II 15 15 Question ID: Rev. 9000010 RO SRO 0 Selected for Exam Student Handout? Origin; New D Lower Order? D Past NRC Exam? Attachment 2 Actions for 1ST Data Outside "Normal" Limits (Shee! t or 1) NOTE The first failure of a valve stroke time test to be within 1ST "Normal" limits does not render that component INOPERABLE, but an immediate rt'test must be performed. .IE a second failure of "Normal" stroke time limit has occurred, Go lh Attachment L VERIFY the following meet test requirements:

• Test prerequisites System condit ions Procedure performance REVIEW data and DETERMINE if test equipment is providing accurate information. lb retest component, PERFORM the foUowing: OBTAIN new applicable Hmn data sheets (new cover sheet not required). ENTER the following 011 applicable OPS Form cover sheet "Comments" section: "Retest of (!ipeew' compoflelll) required, addiliolla'data siteetJ altae/led . .. INDICATE on new OPS Form data sheets that data is from retest and AITACH to original Form. Go To applicable section of this procedure and PERFORM retest. SP 2612A I Le vel of Use STOP THINK ACT REVIEW Rev. OW-OS 350f36 Continuous Page 45 of Printed on 10/28/2009 at 13:20

• Question #: 15 Question ID: 9000010 RO SRO Student Handout? D Lower Order? II-SRO Ques. II 15 Rev. 0 Selected for Exam Origin; New D Past NRC Exam?

• Attachment 2 Actions for 1ST Data Outside "Normal" Limits (Shee! t or 1) NOTE The first failure of a valve stroke time test to be within 1ST "Normal" limits does not render that component INOPERABLE, but an immediate rt'test must be performed.

L .IE a second failure of "Normal" stroke time limit has occurred, Go lh Attachment L 2. VERIFY the following meet test requirements:

• Test prerequisites

• System condit ions

• Procedure performance

3. REVIEW data and DETERMINE if test equipment is providing accurate information.

4. lb retest component, PERFORM the foUowing:

4.1 OBTAIN

new applicable Hmn data sheets (new cover sheet not required).

4.1 ENTER

the following 011 applicable OPS Form cover sheet "Comments" section: "Retest of (!ipeew' compoflelll) required, addiliolla'data siteetJ altae/led

... 4.3 INDICATE on new OPS Form data sheets that data is from retest and AITACH to original Form. 4.4 Go To applicable section of this procedure and PERFORM retest. I Level of Use Continuous STOP THINK ACT Page 45 of 77 SP 2612A REVIEW Rev. OW-OS 350f36 Printed on 10/28/2009 at 13:20 Question #: 15 Question ID: 9000010 -RO SRO Student Handout? Lower Order? II-SRO Ques. #15 Rev. o Selected for Exam Origin; New Past NRC Exam? Attachment 1 Actions for 1ST Data Outside Limits (Sheet 1 of t) 1. CONSIDER component not OPERABLE and NOTIFY SM or US. o 2. IE ill MODE 1,2,3 or 4, LOG into TS 3.7.4.1, and TRMAS 7.1.21 A m; required.

1 G) 3. SUBMiT CR and RECORD CR number in applicable Form cover sheet. 4. NOTIFY tbe following:

• Coordinator

• System Engineer Level of SP 2612A THINK ACT REVIEW Rev. 010-08 34 of 36

• Page 46 of 77 Printed on 10/28/2009 at 13:20 I Question #: 15 II-SRO Ques. #15 Question ID: 9000010 -RO SRO Rev. o Selected for Exam Student Handout? Origin; New I Attachment 1 Actions for 1ST Data Outside Limits (Sheet 1 of t) 1. CONSIDER component not OPERABLE and NOTIFY SM or US. 2. IE ill MODE 1,2,3 or 4, LOG into TS 3.7.4.1, and TRMAS 7.1.21 A m; required.

3. SUBMiT CR and RECORD CR number in applicable Form cover sheet. 4. NOTIFY tbe following:

• Coordinator

• System Engineer Level of Use Continuous THINK ACT SP 2612A REVIEW Rev. 010-08 34 of 36 Lower Order? Past NRC Exam? 1 o G)

• Page 46 of 77 Printed on 10/28/2009 at 13:20 SROcExam Questions Only (No "Ptlrellts" Or Question #: J-SRO Ques. # 16 16 Question ID: Rev. 9000012 0 [J RO Selected for Exam SRO 0 O Student Handout? rigin: New 0 0 Lower Order? Past NRC Exam? The plant is in MODE 6 with the following cond itions: -Fuel mOl/ement is in progress.

-The Personnel Airlock Doors are open -The Equipment Hatch is open. -Containrnent Purge is in operation.

-Containrnent Atmosphere Radiation Monitor, RM-8123, is out of service for repairs. The Auxiliary Building PEO has just reported that the blower for Containment Atmosphere Radiation Monitor, RM-8262, has tripped and is very hot to the touch. Which of the following actions must be taken and why? o A Immediately suspend CORE ALTERATIONS and establish Containment Closure prior to resuming fuel movement, to ensure a potential fuel handling accident in Containment is NOT released to the environment.

Immediately suspend CORE ALTERATIONS and restore the Radiation Monitor blower prior to OB resurr ing fuel movement, to ensure a potential fuel handling accident in Containment is NOT released to the environment. Ensure a conltrol room operator is specifically assigned to close the Containment Purge Valves within 30 minutes of an event, to ensure Containment Closure is reestablished in case of a fuel handling accidemt in Containment. D Restore the Containment Purge Valves to OPERABLE status within the next 30 minutes or immediately close the Purge Valves, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment.

Justification I C IS CORRECT; TS 3.!l.4 requires that Containment Purge Valves either be closed by an automatic isolation or be capable of closed under administrative control. This means that a specific individual is designated as available to close the Purge Valves 30 minutes of a fuel handling accident in A is incorrect; CORE ALTERATIONS do NOT need to be suspended and Containment Closure is still Plausible if the examine,e believes that the Purge Valves need to be closed by an automatic isolation signal. (Only one Radiation Monitor needs to be OPERABLE to initiate and automatic closure of all 4 Purge valves.) The examinee may also believe the loss of the only remaining Radiation Monitor (and automatic isolation of the Purge Valves) results in a loss of Containment (Containment Closure must be set Q[ available during CORE B is incorrect; CORE ALTERATIONS do NOT need to be suspended; however, it would be appropriate to have the Radiation blower Plausible if the examine,e believes that the Purge Valves need to be closed by an automatic isolation D is incorrect.

In MODE. 6, the Purge Valves are still considered OPERABLE even if they are NOT able to be closed by an isolation Plausible because Tech Spec 3.6.3.1 requires each Containment Isolation Valve to be OPERABLE (in MODES 1,2,3, and 4). valves are demonstrated OPERABLE by verifying the automatic signal functions Q[ the valves are closed and secured. This Spec NOT apply to the Containment Purge Valves in MODE I Tech. Spec. 3.9.4 LCO; Containment Comments and Question Modification History Bob K. -D-3/C (Change "Designate" to "Ensure" for control room Minor rewording of choices "C" and "0" per above comments -Bill M. -NOT VALIDATED.

Inadvertently selected the answer for #17 and did not see this question.

When discussed afterwards, felt that this was an LOD of 3 and that he would have known the correct Angelo -D-3/C; No NRC KIA System/EtA I

KIA selecteD NRC KIA Generic System System 029 2.1 Containment Purge System (CPS) Conduct of Operations Page 47 of 77 Printed on 10/2812009 at 13:20 SROcExam Questions Only (No "Ptlrellts" Or Question #: 16 Question ID: 9000012 [J RO SRO 0 Student Handout? 0 Lower Order? J-SRO Ques. # 16 Rev. 0 Selected for Exam Origin: New 0 Past NRC Exam? The plant is in MODE 6 with the following conditions:

-Fuel mOl/ement is in progress.

-The Personnel Airlock Doors are open -The Equipment Hatch is open. -Containrnent Purge is in operation.

-Containrnent Atmosphere Radiation Monitor, RM-8123, is out of service for repairs. The Auxiliary Building PEO has just reported that the blower for Containment Atmosphere Radiation Monitor, RM-8262, has tripped and is very hot to the touch. Which of the following actions must be taken and why? o A Immediately suspend CORE ALTERATIONS and establish Containment Closure prior to resuming fuel movement, to ensure a potential fuel handling accident in Containment is NOT released to the environment.

OB Immediately suspend CORE ALTERATIONS and restore the Radiation Monitor blower prior to resurr ing fuel movement, to ensure a potential fuel handling accident in Containment is NOT released to the environment. C Ensure a conltrol room operator is specifically assigned to close the Containment Purge Valves within 30 minutes of an event, to ensure Containment Closure is reestablished in case of a fuel handling accidemt in Containment.

o D Restore the Containment Purge Valves to OPERABLE status within the next 30 minutes or immediately close the Purge Valves, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment.

Justification I C IS CORRECT; TS 3.!l.4 requires that Containment Purge Valves either be closed by an automatic isolation or be capable of being closed under administrative control. This means that a specific individual is designated as available to close the Purge Valves within 30 minutes of a fuel handling accident in Containment.

A is incorrect; CORE ALTERATIONS do NOT need to be suspended and Containment Closure is still available.

Plausible if the examine,e believes that the Purge Valves need to be closed by an automatic isolation signal. (Only one Containment Radiation Monitor needs to be OPERABLE to initiate and automatic closure of all 4 Purge valves.) The examinee may also believe that the loss of the only remaining Radiation Monitor (and automatic isolation of the Purge Valves) results in a loss of Containment Closure. (Containment Closure must be set Q[ available during CORE ALTERATIONS.)

B is incorrect; CORE ALTERATIONS do NOT need to be suspended; however, it would be appropriate to have the Radiation Monitor blower repairelj.

Plausible if the examine,e believes that the Purge Valves need to be closed by an automatic isolation signal. D is incorrect.

In MODE. 6, the Purge Valves are still considered OPERABLE even if they are NOT able to be closed by an automatic isolation signa!. Plausible because Tech Spec 3.6.3.1 requires each Containment Isolation Valve to be OPERABLE (in MODES 1,2,3, and 4). These valves are demonstrated OPERABLE by verifying the automatic signal functions Q[ the valves are closed and secured. This Spec does NOT apply to the Containment Purge Valves in MODE 6. References I Tech. Spec. 3.9.4 LCO; Containment Penetrations Comments and Question Modification History I Bob K. -D-3/C (Change "Designate" to "Ensure" for control room operator).

Minor rewording of choices "C" and "0" per above comments -RLC Bill M. -NOT VALIDATED.

Inadvertently selected the answer for #17 and did not see this question.

When discussed afterwards, Bill felt that this was an LOD of 3 and that he would have known the correct answer. Angelo -D-3/C; No comments.

NRC KIA System/EtA System 029 Containment Purge System (CPS) I KIA selecteD NRC KIA Generic System 2.1 Conduct of Operations Page 47 of 77 Printed on 10/2812009 at 13:20 Question #: 16 Question ID: 9000012 RO SRO Student Handout? Lower Order? J-SRO Que,. 16 Rev. 0 Selected for Exam Origin; New Past NRC Exam? Number 2.1.32 RO 3.8 SRO 4.0 CFR Link (CFR: 41.10 { 43,2 ( 45.12) Ability to explain and apply system limits and precautions.

September

20. 2004 CONTATh?.fENT PENETRATIONS LIMITING CONDITION FOR OPERATION 3,9.4 The contailUllent penetrati01l8 shall be in the following status: a. The equipment door shall be either: 1, closed and held in place by a minimum offour bolts, or 2. open under administrative control'" and capable of being closed and held in place by a minimum of four bolts. b. The pel'soooel air lock shall be either: 1. closed by one personnel air lock door. or 2. capable of being closed by an OPERABLE personnel air lock door. lUlder administmtive control"'.

and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shaH be either: I. Closed by a manual or automatic isolation valve. blind flange, or equivalent.

or 2. Be capable of being closed under administrative control '" APPLICABILITY:

Dm'iug movement of irradiated fuel assemblies within contaimllent.

ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of in-adiated fuel assemblies in the contaitullent.

Administrative controls shall emmre that appropriate personnel are aware that the equipment door, persollnel air lock door and/or oUler containment penetrations are open" and that a specific illdividual(s) is designated and available to close the equipment door, personnel air lock door and/or other containment penetrations within 30 llliuutes if a fuel handling accident occurs. Any obstructions (e,g" cables and hoses) fhat could prevent closure of the equipment door. a persOimel air lock door and/or other contaimnent penetratiOlllll11st be capable of being quickly removed, -UNIT 2 3/4 9-4 Amendment No. 66,85,93, 'iM5, 284 Page 48 of 77 Printed on 10/28/2009 at 13:20 Question ID: 9000012 RO SRO Student Handout? Rev. 0 Selected for Exam Origin; New Question #: 16 J-SRO Que,. 16 Number 2.1.32 RO 3.8 SRO 4.0 CFR Link (CFR: 41.10 { 43,2 ( 45.12) Ability to explain and apply system limits and precautions.

September

20. 2004 CONTATh?.fENT PENETRATIONS LIMITING CONDITION FOR OPERATION 3,9.4 The contailUllent penetrati01l8 shall be in the following status: a. The equipment door shall be either: 1, closed and held in place by a minimum offour bolts, or 2. open under administrative control'" and capable of being closed and held in place by a minimum of four bolts. b. The personnel air lock shall be either: 1. closed by one personnel air lock door. or 2. capable of being closed by an OPERABLE personnel air lock door. lUlder administmtive control"'.

and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shaH be either: I. Closed by a manual or automatic isolation valve. blind flange, or equivalent.

or 2. Be capable of being closed under administrative control '" APPLICABILITY:

Dm'iug movement of irradiated fuel assemblies within contaimllent.

ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of in1ldiated fuel assemblies in the contaitullent.

Administrative controls shall emmre that appropriate personnel are aware that the equipment door, persollnel air lock door and/or oUler containment penetrations are open" and that a specific illdividual(s) is designated and available to close the equipment door, personnel air lock door and/or other containment penetrations within 30 llliuutes if a fuel handling accident occurs. Any obstructions (e,g" cables and hoses) fhat could prevent closure of the equipment door. a persOimel air lock door and/or other contairunent penetratiOlllll11st be capable of being quickly removed, -UNIT 2 3/4 9-4 Amendment No. 66,85,93, 'iM5, 284 Lower Order? Past NRC Exam? Page 48 of 77 Printed on 10/28/2009 at 13:20

,. . o Question #: 17 RO Iill SRO Student Handout? Lower Order?Question ID: 9079010 I-SRO Ques. # 17 Rev. o Iill Selected for Exam Origin: Mod D Past NRC Exam? The plant was at 100% power when CONVEX ordered Main Generator output be lowered from 900 MWe 600 MWe in 15 AOP 2557. "Emergency Generation Reduction", was initiated and the following conditions now " Group', CEAs are at 170 steps " Main Generator output is 610 MWe and slowly " "A" Ste3m Dump Bypass Valve is 75% open and " "BYPASS TO CND", PIC-4216 output is 83% and " "B", "C' and "0" Steam Dump Bypass Valves are open 75% and " "STEAM DUMP TAVG CNTL", TIC-4165 output is 85% and " "RC LOOP 1 COLD LEG TEMP HI" annunciator has just alarmed (C02/3; " "RC LOOP 2 COLD LEG TEMP HI" annunciator has just alarmed (C02/3; " RCS TGoid is 550 OF and slowly rising Which one of the following actions should the US direct? ** I *************************************

I ********** I

• I **** I ***********************

Transl'er control of the steam dumps to Foxboro IA control and lower Tcold to program. A ';;j Lower the setpoint on the "A" steam dump and log into the DNB Technical Specification.

B Immediately trip the reactor and go to EOP 2525, "Standard Post Trip Actions".

C D Insert CEAs until Tcold is back on program and both C02/3 alarms have cleared.

J B -CORRECT: Turbine load has been lowered ahead of the "A" Steam Dump Valve controller setpoint, as indicated by output being than PIC-4216.

Therefore, raising the output of PIC-4216 to open the "A" steam dump is the most action to restore Temperature to program. This is an expected possible action if the setpoint on PIC-4216 is not lowered initially.

However, Tcold is alreadv above the DNB Tech. Spec. limit, so the LCO must be A -WRONG; "Transferring the "A" steam dump to Foxboro IA control would immediately fail the valve closed, making things far worse. Plausible:

This action is prudent if upon initiation of the procedure, it was noted the controller on C05 was not responding properly.

The indications given show the C05 controller may not be operating correctly, when in fact, the setpolnt must be lowered to ensure the nAn steam dump stays ahead of the other 3 valves. C -WRONG; AOP 255i' requires RCS temperature to be maintained within 10°F of program or a plant trip is required.

AOP 2557 maintains reactor power constant, therefore, Tcold should be -545°F per Attachment

1. RPS indication (and C02/3 alarms) indicate Tcold is >1=549°F, whicl' is <: 10°F above program value. Plausible; If plant power level is extracted from the Main Generator output, then Tcold should be -540°F. This would mean that Tcold is> 10°F above the program value and a trip is required. D -WRONG; Driving in CEAs will lower RCS temperature by lowering reactor power but, RCS temperature is because generator load was not controlled Plausible:

This is an acceptable action if temperature is out of band due to turbine load reduction being ahead of reactor reduction.

HOViever, reactor power is not being reduced, by References I AOP 2537, Emergency Generation Reduction, Pages 8 & 9 Comments and Question Modification History Bob K. D-3/W (Did not read "2557" and thought Rx power was being reduced. Upon second thought, was able to dertermine correct Bill M. -D-3/W K (Doesn't feel that picking up load on the Turbine is appropriate for this condition.

Feels that adjusting Dumps is more Revised distractor "A" slightly to transfer control of all the steam dumps to the Foxboro IA vs only the "A" steam Revised "S" to lower the setpoint on the "A" steam dump vs pick up load on the Turbine.* Mike C .* /W, Lower controller outputs in stem to put valves at 75% open. Done -Angelo -D-4/C; No NRC KIA System/E/A System 045 Main Turbine Generator (MT/G) System KIA NRC KIA Generic System 2.4 Emergency Procedures IPlan Page 49 of 77 Printed on 10/28/2009 at 13:20 ,. . o Question #: 17 Question ID: 9079010 RO Iill SRO Student Handout? Lower Order? I-SRO Ques. # 17 Rev. o Iill Selected for Exam Origin: Mod D Past NRC Exam? The plant was at 100% power when CONVEX ordered Main Generator output be lowered from 900 MWe to 600 MWe in 15 minutes. AOP 2557. "Emergency Generation Reduction", was initiated and the following conditions now exist: " Group', CEAs are at 170 steps withdrawn. " Main Generator output is 610 MWe and slowly lowering. " "A" Ste3m Dump Bypass Valve is 75% open and stable. " "BYPASS TO CND", PIC-4216 output is 83% and stable. " "B", "C' and "0" Steam Dump Bypass Valves are open 75% and stable. " "STEAM DUMP TAVG CNTL", TIC-4165 output is 85% and stable. " "RC LOOP 1 COLD LEG TEMP HI" annunciator has just alarmed (C02/3; C-34). " "RC LOOP 2 COLD LEG TEMP HI" annunciator has just alarmed (C02/3; D-34). " RCS TGoid is 550 OF and slowly rising (RPS). Which one of the following actions should the US direct? ** I *************************************

I ********** I

• I **** I ***********************

A Transl'er control of the steam dumps to Foxboro IA control and lower Tcold to program. ';;j B Lower the setpoint on the "A" steam dump and log into the DNB Technical Specification.

C Immediately trip the reactor and go to EOP 2525, "Standard Post Trip Actions".

D Insert CEAs until Tcold is back on program and both C02/3 alarms have cleared.

J B -CORRECT: Turbine load has been lowered ahead of the "A" Steam Dump Valve controller setpoint, as indicated by TIC-4165 output being than PIC-4216.

Therefore, raising the output of PIC-4216 to open the "A" steam dump is the most immediate action to restore Temperature to program. This is an expected possible action if the setpoint on PIC-4216 is not lowered enough initially.

However, Tcold is alreadv above the DNB Tech. Spec. limit, so the LCO must be entered. A -WRONG; "Transferring the "A" steam dump to Foxboro IA control would immediately fail the valve closed, making things far worse. Plausible:

This action is prudent if upon initiation of the procedure, it was noted the controller on C05 was not responding properly.

The indications given show the C05 controller may not be operating correctly, when in fact, the setpolnt must be lowered to ensure the nAn steam dump stays ahead of the other 3 valves. C -WRONG; AOP 255i' requires RCS temperature to be maintained within 10°F of program or a plant trip is required.

AOP 2557 maintains reactor power constant, therefore, Tcold should be -545°F per Attachment

1. RPS indication (and C02/3 alarms) indicate Tcold is >1=549°F, whicl' is <: 10°F above program value. Plausible; If plant power level is extracted from the Main Generator output, then Tcold should be -540°F. This would mean that Tcold is> 10°F above the program value and a trip is required. D -WRONG; Driving in CEAs will lower RCS temperature by lowering reactor power but, RCS temperature is "out-of-program" because generator load was not controlled properly.

Plausible:

This is an acceptable action if temperature is out of band due to turbine load reduction being ahead of reactor power reduction.

HOViever, reactor power is not being reduced, by procedure.

References I AOP 2537, Emergency Generation Reduction, Pages 8 & 9 Comments and Question Modification History I Bob K. D-3/W (Did not read "2557" and thought Rx power was being reduced. Upon second thought, was able to dertermine the correct answer.) Bill M. -D-3/W K (Doesn't feel that picking up load on the Turbine is appropriate for this condition.

Feels that adjusting Condenser Dumps is more appropriate.

Revised distractor "A" slightly to transfer control of all the steam dumps to the Foxboro IA vs only the "A" steam dump. Revised "S" to lower the setpoint on the "A" steam dump vs pick up load on the Turbine .* RJA Mike C .* /W, Lower controller outputs in stem to put valves at 75% open. Done -RLC Angelo -D-4/C; No comments.

NRC KIA System/E/A System 045 Main Turbine Generator (MT/G) System KIA NRC KIA Generic System 2.4 Emergency Procedures IPlan Page 49 of 77 Printed on 10/28/2009 at 13:20 Question #: 17 Question ID: 9079010 RO SRO Student Handout? D Lower Order? J-SRO Ques. # 17 Rev. o ". Selected for Exam Origin: Mod Past NRC Exam? Number 2.4.47 RO 4.2 SRO 4.2 CFR Link (CFR: 41.10,43.5/45.12)

Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Millstone Unit 2 Emergency Generation Reduction AOP 2557 Revision 006 -06 Page 8 of 16 I INSTRUCtIONS CONTINGENCY ACfIONS CAUTION L PPC calorimetric is inaccurate due to SG level transients.

The most accurate available indication of reactor power is power. 2. Turbine exhaust hood tCflll?crature greater than or equal to 225°F requires ImmulI!lurbine tnp. I@ WHEN transferring steam load from main turbine to steam dump and bypass valves, MONITOR the following:

• Omden.'icr backpressurc (PPC point P5127)

• Turbine exhaust hood lemperature (UR 4500 "TURBINE TEMP & EXPANSlON," recorder points 8 and 9, PPC points T4319 and T4320) * ('ondensate header flow and pressure

• MVARs 10 WHEN reducing turbine load, MAINTAIN "t':" steam uump bypa&,> valw 20 to 1(J{J9i; open as foll{Jws:

• Using "STM DUMPTAVG CNTL, TIC-4165", THROTTLE open "B," hC:' and "D" steam dump bypass valves Level of Use Continuous STOP TH1N\< ACT REVIEW , Page 50 of 77 Printed on 10/28/2009 at 13:20 Question #: 17 J-SRO Ques. # 17 Question ID: 9079010 RO SRO Student Handout? D Lower Order? Rev. o ". Selected for Exam Origin: Mod Past NRC Exam? Number 2.4.47 RO 4.2 SRO 4.2 CFR Link (CFR: 41.10,43.5/45.12)

Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

I Millstone Unit 2 Emergency Generation Reduction INSTRUCtIONS AOP 2557 Revision 006 -06 Page 8 of 16 CONTINGENCY ACfIONS CAUTION L PPC calorimetric is inaccurate due to SG level transients.

The most accurate available indication of reactor power is power. 2. Turbine exhaust hood tCflll?crature greater than or equal to 225°F requires ImmulI!lurbine tnp. I@ WHEN transferring steam load from main turbine to steam dump and bypass valves, MONITOR the following:

• Omden.'icr backpressurc (PPC point P5127)

• Turbine exhaust hood lemperature (UR 4500 "TURBINE TEMP & EXPANSlON," recorder points 8 and 9, PPC points T4319 and T4320) * ('ondensate header flow and pressure

• MVARs 10 WHEN reducing turbine load, MAINTAIN "t':" steam uump bypa&,> valw 20 to 1(J{J9i; open as foll{Jws:

• Using "STM DUMPTAVG CNTL, TIC-4165", THROTTLE open "B," hC:' and "D" steam dump bypass valves Level of Use Continuous STOP TH1N\< ACT REVIEW , Page 50 of 77 Printed on 10/28/2009 at 13:20 It) Question #: 17 Student Handout? Lower Order? Question ID: 9079010 RO SRO l-SRO Ques. II 17 Rev. 0 Selected for Exam Origin: Mod Past NRC Exam? Millstone Unit 2 AOP 2557 Revision 006-06 Emergency Genera tion Page 9 of 16 Reduction INSTRUCTIONS Using one of the following, ADJUST tmbinc to desired load:

l.JMI.T POTi, "LOAD SELECTOR.

INCREASE" and "LOAD SELECTOR, DECREASE" buttons CONTINGENCY NOTE Receipt of annunciators DA-37. "Ill COND Dfr AND DB-37. "HI CONI> DIS TEMP" is expected during this evolution (C-Oti/07). annunciator DA-37 (C-06/07), "HI COND Dfr or DB-37 (C-06/07), "In COND DIS TEMP:' is received, Refer 1b ARP 2590E, "Alarm Responsc for Control Room Panels, C-Ofil7." 13 desired load is achieved, STABILIZE turbine load and ReS temperature.

_3.14 ENSURE pressurizer level 35 to 70l1'1:. THINK IF the pressurizer level control system is nol operating properly in automaticallv, RESTORE and MAINTAIN pressurizer level 35 to 7(}% by performing ANY of (he following: OPERATE the pressurizer level control system. Manually OPERATE charging and letdown. REVIiEW Page 51 of Printed on 10/28/2009 at 13:20 It) Question ID: 9079010 RO SRO Student Handout? Lower Order? Question #: 17 l-SRO Ques. II 17 Rev. 0 Selected for Exam Origin: Mod Past NRC Exam? Millstone Unit 2 Emergency Genera tion Reduction INSTRUCTIONS

_3.11 Using one of the following, ADJUST tmbinc to desired load:

• l.JMI.T POTi, * "LOAD SELECTOR.

INCREASE" and "LOAD SELECTOR, DECREASE" buttons AOP 2557 Revision 006-06 Page 9 of 16 CONTINGENCY ACTIONS NOTE Receipt of annunciators DA-37. "Ill COND Dfr AND DB-37. "HI CONI> DIS TEMP" is expected during this evolution (C-Oti/07).

3.12 annunciator DA-37 (C-06/07), "HI COND Dfr or DB-37 (C-06/07), "In COND DIS TEMP:' is received, Refer 1b ARP 2590E, "Alarm Responsc for Control Room Panels, C-Ofil7." 13 desired load is achieved, STABILIZE turbine load and ReS temperature.

_3.14 ENSURE pressurizer level 35 to 70l1'1:. THINK 3.14.1 IF the pressurizer level control system is nol operating properly in automaticallv, RESTORE and MAINTAIN pressurizer level 35 to 7(}% by performing ANY of (he following:

a. OPERATE the pressurizer level control system. b. Manually OPERATE charging and letdown. REVIiEW Page 51 of 77 Printed on 10/28/2009 at 13:20 Question #: 18 D Student Handout? D Lower Order?Question ID: 9000011 RO SRO I-SRO Ques. # 18 Rev. 0 Selected for Exam Origin: New Past NRC Exam? The plant is in normal operation at 100% power, when a Fire System Trouble annunciator is received on C06/? and Zone 45 on Fire Panel, C-26. The Auxiliary Building PEO subsequently calls from the West DC Switchgear Room and reports the following:

• One Ion Chamber smoke detector is in alarm.

• The Halon strobe lights and horn are pulsating slowly.

• All other smoke detectors are operating normally (not in alarm).

• There is no smoke or fire in the area. The detector appears to have failed. Which of the following describes the impact of the above conditions, and the direction the US will give? The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.7.9.4, "Halon Fire Suppression System", provide backup fire suppression and establish a fire watch, when the room has cleared. The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.3.3.7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable and a fire watch must be established. The Fire Suppression system is warning that a discharge will occur after a timer countdown.

Per TRM 3.7.9.4, "Halon Fire Suppression System", provide backup fire suppression and establish a fire watch, when the room has cleared. o The Fire Suppression system is warning that a discharge will occur after a timer countdown.

Per TRM 3.3.3.'7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable, establish a fire watch, when the room has cleared. Just:ification J B -CORRECT; The East and West DC switchgear rooms require two zones (one photoelectric smoke detector and one ion smoke detector) to initiate a halon release. Activation of one smoke detector zone, ion or photoelectric, will cause the strobe and horn to pulse slowly. However, the TRM requires all detectors to be functioning or the system is inoperable A WRONG; The Halon system is not made inoperable because the detection system has a Plausible; Examinee may think that due to the "false" activation of a sensor, the system should be prevented from any activation and the Halon system can no longer C -WRONG; Activation of a second smoke detector of the opposite type, but in the same room, will cause the strobe and horns for the affected room pulse QUICKLY. The flashing lights will operate, and a 60 second pre-discharge time delay will begin. Upon expiration of the time delay the Halon System will discharge and the strobe and horn will sound steadily.

Plausible; Examinee may think that the SLOWLY pulsating horn and strobe light warn of a timer countdown to discharge halon, in which case, the Halon system would then be inoperable and this action would be correct. D -WRONG; Only one detector failing in the activate mode would cause the given Plausible; Examinee may think that the pulsating horn and strobe lights indicate that the failed detector has caused a full malfunction and a discharge is imminent.

If the system were actually triggered due to multiple detector failures, this would be correct choice . References I 1. OP 2341A, "Fire Protection System", Pg 4, Discussion section. 2. ARP 25901, "Alarm for Fire Panel, C-26" (Zone 45), Pg 67-69 Comment.s and Question Modification History Bob K. -D-3/W (Change choices to say "per" applicable procedure and make correct answer reference applicable Modified question to utilize procedures where applicable action is directed.* Bill M. -D-3/C, Angelo -D-3/C; No NRC KIA System/E/A System 086 Fire Protection System (FPS) Number A2.03 RO 2.7 SRO 2.9 CFR Link (CFR: 41.5 I 43.5 145.3145.13) Ability to (a) predict the impacts of the following mal-functions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadvertent actuation of the FPS due to circuit failure or welding Page 52 of Printed on 10/28/2009 at 13:20 Question #: 18 Question ID: 9000011 RO SRO D Student Handout? D Lower Order? I-SRO Ques. # 18 Rev. 0 Selected for Exam Origin: New Past NRC Exam? The plant is in normal operation at 100% power, when a Fire System Trouble annunciator is received on C06/? and Zone 45 on Fire Panel, C-26. The Auxiliary Building PEO subsequently calls from the West DC Switchgear Room and reports the following:

• One Ion Chamber smoke detector is in alarm.

• The Halon strobe lights and horn are pulsating slowly.

• All other smoke detectors are operating normally (not in alarm).

• There is no smoke or fire in the area. The detector appears to have failed. Which of the following describes the impact of the above conditions, and the direction the US will give? A The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.7.9.4, "Halon Fire Suppression System", provide backup fire suppression and establish a fire watch, when the room has cleared. B The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.3.3.7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable and a fire watch must be established.

C The Fire Suppression system is warning that a discharge will occur after a timer countdown.

Per TRM 3.7.9.4, "Halon Fire Suppression System", provide backup fire suppression and establish a fire watch, when the room has cleared. o The Fire Suppression system is warning that a discharge will occur after a timer countdown.

Per TRM 3.3.3.'7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable, establish a fire watch, when the room has cleared. Just:ification J B -CORRECT; The East and West DC switchgear rooms require two zones (one photoelectric smoke detector and one ion smoke detector) to initiate a halon release. Activation of one smoke detector zone, ion or photoelectric, will cause the strobe and horn to pulse slowly. However, the TRM requires all detectors to be functioning or the system is inoperable A WRONG; The Halon system is not made inoperable because the detection system has a failure. Plausible; Examinee may think that due to the "false" activation of a sensor, the system should be prevented from any subsequent activation and the Halon system can no longer trigger. C -WRONG; Activation of a second smoke detector of the opposite type, but in the same room, will cause the strobe and horns for the affected room pulse QUICKLY. The flashing lights will operate, and a 60 second pre-discharge time delay will begin. Upon expiration of the time delay the Halon System will discharge and the strobe and horn will sound steadily.

Plausible; Examinee may think that the SLOWLY pulsating horn and strobe light warn of a timer countdown to discharge halon, in which case, the Halon system would then be inoperable and this action would be correct. D -WRONG; Only one detector failing in the activate mode would cause the given alarms. Plausible; Examinee may think that the pulsating horn and strobe lights indicate that the failed detector has caused a full system malfunction and a discharge is imminent.

If the system were actually triggered due to multiple detector failures, this would be the correct choice . . References I 1. OP 2341A, "Fire Protection System", Pg 4, Discussion section. 2. ARP 25901, "Alarm for Fire Panel, C-26" (Zone 45), Pg 67-69 Comment.s and Question Modification History I Bob K. -D-3/W (Change choices to say "per" applicable procedure and make correct answer reference applicable ARP). Modified question to utilize procedures where applicable action is directed .* RLC. Bill M. -D-3/C, K Angelo -D-3/C; No comments.

NRC KIA System/E/A System 086 Fire Protection System (FPS) Number A2.03 RO 2.7 SRO 2.9 CFR Link (CFR: 41.5 I 43.5 145.3145.13)

Ability to (a) predict the impacts of the following mal-functions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadvertent actuation of the FPS due to circuit failure or welding Page 52 of 77 Printed on 10/28/2009 at 13:20 SRO Exarri Only (No Question #: 18 o Student Handout? o Lower Order?Question ID: 9000011 ORO SRO I-SRO Ques. # 18 0 Selected for Exam Origin: New o Past NRC Exam? Setpolnt: I ZONE 45 Trouble -Detector string failure Fire WEST DC SWGR Detector string alarm ROOM FlP-6

• Halon release from pressure ________

______________

AUIQMAnCJ:UNCIIONS I\ctivatioll of nne smoke detector (photoelectric or inn) causes the following to occur: Alarmed detectors red light illuminates. On FLP-o, alarmed zone annunciates. Lm:ally strobe and homs pulses slowly. On FLP-6, detectors location is shown on graphic annunciator. Alarm signal is sent to C-26 Zone 45. Alarm lamp for graphic annunciator for east 120 volt switchgear Activation of another smoke detector of opposite type (ion or photoelectric), the following tn Ahu'med detectors red light illuminates. On FLP-6, alarmed zone annunciates. Locally stmbe and horns pulses quickly. The following closes: l-HV-138VB. "Supply to 'B' Battery Room From DC Rooms" 2-HV-60lB. "Cable Vault to "B' DC Room SWGR. Room Fire Damper"' 60 second pre -discharge t tme delay begins. Upon expiration of time dclay Halon System discharges and strobe and horn sounds steadily. Flashing lights operates. Upon expiration or time dclay halon system discharges into west 120 volt room. ARP25901 Level of Use I STOP THINK ACT REVIEW Rev. 002-08 Reference 670f104 Page 53 of Printed on 10/28/2009 at 13:20 Question #: 18 I-SRO Ques. # 18 SRO Exarri Only (No Question ID: 9000011 ORO SRO Rev. 0 Selected for Exam o Student Handout? Origin: New o Lower Order? o Past NRC Exam? Setpolnt:

I ZONE 45 I

• Trouble -Detector string failure

• Fire -* Detector string alarm

• Halon release from pressure WEST DC SWGR ROOM FlP-6 ________

______________

AUIQMAnCJ:UNCIIONS

1. I\ctivatioll of nne smoke detector (photoelectric or inn) causes the following to occur:

• Alarmed detectors red light illuminates.

• On FLP-o, alarmed zone annunciates.

• Lm:ally strobe and homs pulses slowly.

• On FLP-6, detectors location is shown on graphic annunciator.

• Alarm signal is sent to C-26 Zone 45.

• Alarm lamp for graphic annunciator for east 120 volt switchgear room illuminates.

2. Activation of another smoke detector of opposite type (ion or photoelectric), causes the following tn occur:

• Ahu'med detectors red light illuminates.

• On FLP-6, alarmed zone annunciates.

• Locally stmbe and horns pulses quickly.

• The following closes:

• l-HV-138VB. "Supply to 'B' Battery Room From DC Rooms"

• 2-HV-60lB. "Cable Vault to "B' DC Room SWGR. Room Fire Damper"'

• 60 second pre -discharge t tme delay begins.

• Upon expiration of time dclay Halon System discharges and strobe and horn sounds steadily.

• Flashing lights operates.

• Upon expiration or time dclay halon system discharges into west 120 volt room. Level of Use I Reference STOP THINK ACT Page 53 of 77 REVIEW ARP25901 Rev. 002-08 670f104 Printed on 10/28/2009 at 13:20 Question #: 18 Student Handout? Lower Order? Question ID: 9000011 D RO" SRO I-SRO Que5. # 18 Rev. o Selected for Exam Origin: New Past NRC Exam? ZONE 45 WARNING When (lalon Systems are actuated, the affected area is neither oxygen dcficienl nor toxic; however, extended exposure to Halon may have harmful effects. NOTE Activation of the manual pull station CUlIseS ..1 halon release aftent five second time delay. An abort switch can he lIsed to prevent a halon until the atlected panel can be reset. If the abort switch is turned back before the panel i.;; reset, the Haton discharges after a ten second time delay. The reset is im;jde the respective FLP and a valve lock key is required to get into the panel. The manual pull station activation overrides the abort. Refer To Attachment 6, "FLP-6 and 6A Zone 45" and DETERMINE cause of alarm. 11::' fire alam1 is valid, PERFORM the following:

I@ DETERMINE location of fire. IF alarm was cause by actuation of Halon Fire Suppression System AND fire is verified, PERfORM the following: EVACUATE affected area. Refer 'Ii.) AOP 2559, "Fire" and PERfORM applicable actions. If alarm is due to Halon discharge AND no fire is present. PERFORM the following: V CAUTION When ventilating care must be taken not to discharge products of combustion into non -affected rooms. PERfORM actions to ventilate affected area. Refer To OP 2341A, "Fire Protection System;' and REMOVE appropriate Halon System from service. ARP 25901 level of usE] STOP THINK ACT REVIEW Rev. 002-08 Reference L....-_ 68 of 104 Page 54 of Printed on 10/28/2009 at 13:20 Question #: 18 I-SRO Que5. # 18 Question ID: 9000011 D RO" SRO Rev. o Selected for Exam Student Handout? Lower Order? Origin: New Past NRC Exam? ZONE 45 8 WARNING 8 When (lalon Systems are actuated, the affected area is neither oxygen dcficienl nor toxic; however, extended exposure to Halon may have harmful effects. NOTE Activation of the manual pull station CUlIseS .. 1 halon release aftent five second time delay. An abort switch can he lIsed to prevent a halon until the atlected panel can be reset. If the abort switch is turned back before the panel i.;; reset, the Haton discharges after a ten second time delay. The reset is im;jde the respective FLP and a valve lock key is required to get into the panel. The manual pull station activation overrides the abort. 1. Refer To Attachment 6, "FLP-6 and 6A Zone 45" and DETERMINE cause of alarm. 2. 11::' fire alam1 is valid, PERFORM the following:

2.1 DETERMINE

location of fire. IF alarm was cause by actuation of Halon Fire Suppression System AND fire is verified, PERfORM the following:

• EVACUATE affected area.

• Refer 'Ii.) AOP 2559, "Fire" and PERfORM applicable actions. 2.3 If alarm is due to Halon discharge AND no fire is present. PERFORM the following:

V CAUTION V When ventilating care must be taken not to discharge products of combustion into non -affected rooms. 2.3.1 PERfORM actions to ventilate affected area. 23.2 Refer To OP 2341A, "Fire Protection System;' and REMOVE appropriate Halon System from service. level of usE] Reference L....-_ STOP THINK ACT REVIEW ARP 25901 Rev. 002-08 68 of 104 I@ Page 54 of 77 Printed on 10/28/2009 at 13:20 Question #: 18 Student Handout? Lower Order? Question ID; 9000011 RO SRO J-SRO Ques. # 18 Rev. o Selected for Exam Origin: New Past NRC Exam? [IONE45 I 2.3.3 POST fire watch as necessary.

2.3.4 NOTIFY

Fire Marshall.

2.3.5 SUBMIT

lroublc Report to Maintenance Department. IE alarm is due to electrical malfunction, SUBMIT Trouble Report t.o Maintenance En continued operation, CONSIDER supplemental room cooling. As applicable, Refer 11) Technical Requirements Manual, and DETERMINE INFORMATION Initiating Devices FPL-fi

• Detector string, FSD-49 3 ion detectors (smoke) 3 photoelectric detectors (smoke)

• PS-7696

• 1IS-76% A & B (Manual Electric Release) Computer Points FLP-fi TE8436 technical Requirements Manual.Section II, subsection 1.0, table A.3.1..4 E..3.1 Procedures OP 234 lA, "Fire Protection System" AOP 2559, "Fire" AOP 2579F, "Fire Procedure for Hot Standby Appendix "R" Fire Area R -10" AOP 2579FF, "Fire Procedure for Cooldowtl and Cold Shutdown Appendix "Fe' Fire Area R -10 and R -8" Level of ARP25901 STOP THINK ACT REVIEW Rcv.OO2-08 69 of 104 Page 55 of Printed on 10/28/2009 at 13:20 Question #: 18 J-SRO Ques. # 18 Question ID; 9000011 Rev. o RO SRO Selected for Exam Student Handout? Lower Order? Origin: New Past NRC Exam? [IONE45 I 2.3.3 POST fire watch as necessary.

2.3.4 NOTIFY

Fire Marshall.

2.3.5 SUBMIT

lroublc Report to Maintenance Department.

3. IE alarm is due to electrical malfunction, SUBMIT Trouble Report t.o Electrical Maintenance Department.

4. En continued operation, CONSIDER supplemental room cooling. 5. As applicable, Refer 11) Technical Requirements Manual, and DETERMINE system operability.

INFORMATION

1. Initiating Devices

• FPL-fi

• Detector string, FSD-49

• 3 ion detectors (smoke)

• 3 photoelectric detectors (smoke)

• PS-7696

• 1IS-76% A & B (Manual Electric Release) 2. Computer Points

• FLP-fi

• TE8436 3. technical Requirements Manual.Section II, subsection 1.0,

• table A.3.1..4

• E .. 3.1 4. Procedures

• OP 234 lA, "Fire Protection System"

• AOP 2559, "Fire"

• AOP 2579F, "Fire Procedure for Hot Standby Appendix "R" Fire Area R -10"

• AOP 2579FF, "Fire Procedure for Cooldowtl and Cold Shutdown Appendix "Fe' Fire Area R -10 and R -8" Level of Use Reference STOP THINK ACT Page 55 of 77 REVIEW ARP25901 Rcv.OO2-08 69 of 104 Printed on 10/28/2009 at 13:20 Question #: 18 Lower Order?Question ID: 9000011 RO ..t. SRO Student Handout? Past NRC Exam?I-SRO Que, ... 18 Rev. 0 'iiJ Selected for Exam Origin: New TECHNICAL REQUIREMENTS 3/4,3,3 MQNIIQRING INSIRUMIiNIATIQN 3'4,3,3,7 FIRE PETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3 3."l As ;:1 minimum. the fire detection instrumentation for each fire detection zone in TRM Table 3.3-10 shall be OPERABLE APPLiCABILITY; Whenever equipment In that fire detection zone is required to be OPERABLE.

ACTIQtj; With the number of OPERABLE fire detection instrument(s}

less than the of OPERABLE requirements of TRM Table Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, establish a fire watch patrol to inspect the zone(s} with the inoperable instrument(s) at least once per hour unless the instrument(s) is located inside the containme'nt.

Roving fire watches must monitor the area of the device in question, as TECHNICAL TAaLE FIRE PElEs:IIQtf Minimum Minimum Total No. Chanrwtls Total No. Channels oiC..I:uwn.t.ls at ChaooeJ$ 4.16 & 6.9 kV Swftchgear Room (56'6') (40) 4 3 4.16 & 6.S' kV Swftchgear Room (31'6') (18) 4 480 V West Switchgear Room (36'6') (18) 2 480 V East Switchgear Room (31')"6-)

(28) 2 East DC Equipment Room (43 Alarm) (FlP 5) I) West DC Equipment Room (45 Alarm) (FLP 6) 6 Page 56 of Printed on 10/28/2009 at 13:20 Question #: 18 Question ID: 9000011 RO ..t. SRO I-SRO Que, ... 18 Rev. 0 'iiJ Selected for Exam TECHNICAL REQUIREMENTS 3/4,3,3 MQNIIQRING INSIRUMIiNIATIQN 3'4,3,3,7 FIRE PETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION Student Handout? Lower Order? Origin: New Past NRC Exam? 3.3 3."l As ;:1 minimum. the fire detection instrumentation for each fire detection zone in TRM Table 3.3-10 shall be OPERABLE APPLiCABILITY; Whenever equipment In that fire detection zone is required to be OPERABLE.

ACTIQtj; With the number of OPERABLE fire detection instrument(s}

less than the minimum of OPERABLE requirements of TRM Table 3.3-10: a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, establish a fire watch patrol to inspect the zone(s} with the inoperable instrument(s) at least once per hour unless the instrument(s) is located inside the containme'nt.

Roving fire watches must monitor the area of the device in question, as 4. 4.16 & 6.9 kV Swftchgear Room (56'6') (40) 4.16 & 6.S' kV Swftchgear Room (31'6') (18) 480 V West Switchgear Room (36'6') (18) 480 V East Switchgear Room (31')"6-)

(28) East DC Equipment Room (43 Alarm) (FlP 5) West DC Equipment Room (45 Alarm) (FLP 6) TECHNICAL REQUIREMENT TAaLE 3.a-3D FIRE PElEs:IIQtf INSTRUMgNTS Minimum Total No. Chanrwtls Total No. oiC..I:uwn.t.ls at ChaooeJ$ 4 4 2 2 I) 6 Minimum Channels 3 3 1 1 e e Page 56 of 77 Printed on 10/28/2009 at 13:20 Question #: 19 Question ID: 9000013 RO SRO Student Handout? Lower Order? l-SRO QUf:S. # 19 Rev. 0 Selected for Exam Origin: New Past NRC Exam? While operating at Beginning of Life (BOL), 100% power, the "A" Main Feed Pump high vibration annunciator alarmed. After subsequent investigation and troubleshooting, the System Engineer and Maintenance agreie that the pump must be removed from service within the next 60 minutes to prevent severe dal1age. The crew has just entered AOP 2575, Rapid Downpower.

Which of the following statements describes the method that must be utilized to perform this evolution? Use Reactivity Plan RE-G-08 to reduce power to 70%, secure the Feed Pump, then transition to OP 2208, Attachment 5, Reactivity Thumbrules, for maintaining power with riSing Xenon. Use R:eactivity Plan RE-G-05 until 90% power, then transition to AOP 2575, Attachment 7, Boration/Power Reduction Rates to continue the power reduction required to secure the Feed Pump. Use Heactivity Plan RE-G-11 to reduce power to the appropriate level, secure the Feed Pump, then raise power to the appropriate level using OP 2204, Load Changes, and a new Reactivity Plan. DOUse Heactivity Plan RE-G-10 to reduce power to the appropriate level, secure the Feed Pump, then transilion to OP 2393, Core Power Distribution and Monitoring, to maintain AS!. J ustificatioJ C IS CORRECT: React,!vity Plan RE-G-11 provides the Boration rate and CEA insertion required to down power to 55% within the one hour time constraint.

OP 2322, Main Feedwater, requires power to be at or less than 55% to remove the first Main Feed Pump from service. This 'eactivity plan also provides the dilution rate to maintain power at 55% to compensate for Xenon. OP 2322, Main Feedwater, provides guidance to raise power as high as 75% once the affected Main Feed Pump is secured. As a result, a new reactivity plan must be developed to allow raiSing power. A is incorrect; Reactivity Plan RE-G-08 only provides Boration rate and CEA Insertion required to downpower to 70%. Although OP 2208, Attachment 5, provides guidance for maintaining power with changing reactivity conditions (e.g., Xenon), AOP 2575 provides direction for use of a reactivity plan appropriate for the evolution being performed.

Additionally, power must be at or below 55% to remove a Main Feed pump from service. Plausible beca'Jse this guidance is adequate to reduce power to 70%. OP 2322 provides guidance on continued operation with one Main Feed Pump up to 7'5% power. Examinee may think that a feed pump can be removed from service at less than or equal to 75%. B is incorrect; Reactivity Plan RE-G-05 only provides Boration rate and CEA Insertion required to downpower to 90%. Although AOP 2575, Attachment 7, provides guidance for reducing power if a reactivity plan is not available, guidance exists for use of a reactivity plan appropriate for the evolution being performed.

Plausible because this guidance is adequate to remove the feed pump from service, but does not provide the complete guidance for reactivity control after the pump is secured. o is incorrect; Reactivity Plan RE-G-10 provides guidance on reducing power to 55%; however, the reduction rate will NOT allow the Feed Pump to be removed from service in the required 60 minute time limit. OP 2393 provides guidance for maintaining ASI control with CEAs, but this guidance does NOT take into consideration the amount of Boric Acid to used to counteract Xenon. Plausible because this reactivity plan will allow performance of the downpower to remove the Main Feed Pump from service, but it does NOT allow reducing power fast enough to meet the 60 minute time limit. The examinee may mistakenly assume that the reactivity plan does NOT inclL'de the effects of Xenon; therefore, the power reduction rate would actually be higher than indicated on RE-G-10. References IRE-G-03, Rapid Down Power Reactivity Plans (Question Reference

1. 1 of AOP 2575, Rapid Oownpower (NOT provided to

'Comments and Questi()n Modification HistOry I Bob K. 3/W ("Bad question, "0" could be correct. Reword stem to say "within the next 60 minutes" and reword "C" & "0" to include 2204 and new reactivity plan). Swapped Distaractor "A" and "8". Added OP 2204 to the correct answer (C.). Changed reactivity Plan to RE-G-10 instead of RE-G-14. (RE-G-14 may be appropriate as long as the crew stops at the required power level. Deleted OP 2204 and Inserted OP2393. RJA Bill M. 2/C, 0( (Doesn't feel that B is plausible.

No recommendation.)

Discussed.

No modification needed. Distractors will allow the required downpower; therefore, all are plausible.

-RJA Angelo 4/C; No comments.

NRC KIA System/E/A System 2.1 Conduct of Operations NRC KIA Generic System 2.1 Conduct of Operations Page 57 of Printed on 10/28/2009 at 13:20 Question #: 19 Question ID: 9000013 RO SRO Student Handout? Lower Order? l-SRO QUf:S. # 19 Rev. 0 Selected for Exam Origin: New Past NRC Exam? While operating at Beginning of Life (BOL), 100% power, the "A" Main Feed Pump high vibration annunciator alarmed. After subsequent investigation and troubleshooting, the System Engineer and Maintenance agreie that the pump must be removed from service within the next 60 minutes to prevent severe dal1age. The crew has just entered AOP 2575, Rapid Downpower.

Which of the following statements describes the method that must be utilized to perform this evolution?

A Use Reactivity Plan RE-G-08 to reduce power to 70%, secure the Feed Pump, then transition to OP 2208, Attachment 5, Reactivity Thumbrules, for maintaining power with riSing Xenon. B Use R:eactivity Plan RE-G-05 until 90% power, then transition to AOP 2575, Attachment 7, Boration/Power Reduction Rates to continue the power reduction required to secure the Feed Pump. C Use Heactivity Plan RE-G-11 to reduce power to the appropriate level, secure the Feed Pump, then raise power to the appropriate level using OP 2204, Load Changes, and a new Reactivity Plan. DOUse Heactivity Plan RE-G-10 to reduce power to the appropriate level, secure the Feed Pump, then transilion to OP 2393, Core Power Distribution and Monitoring, to maintain AS!. J ustificatio J C IS CORRECT: React,!vity Plan RE-G-11 provides the Boration rate and CEA insertion required to down power to 55% within the one hour time constraint.

OP 2322, Main Feedwater, requires power to be at or less than 55% to remove the first Main Feed Pump from service. This 'eactivity plan also provides the dilution rate to maintain power at 55% to compensate for Xenon. OP 2322, Main Feedwater, provides guidance to raise power as high as 75% once the affected Main Feed Pump is secured. As a result, a new reactivity plan must be developed to allow raiSing power. A is incorrect; Reactivity Plan RE-G-08 only provides Boration rate and CEA Insertion required to downpower to 70%. Although OP 2208, Attachment 5, provides guidance for maintaining power with changing reactivity conditions (e.g., Xenon), AOP 2575 provides direction for use of a reactivity plan appropriate for the evolution being performed.

Additionally, power must be at or below 55% to remove a Main Feed pump from service. Plausible beca'Jse this guidance is adequate to reduce power to 70%. OP 2322 provides guidance on continued operation with one Main Feed Pump up to 7'5% power. Examinee may think that a feed pump can be removed from service at less than or equal to 75%. B is incorrect; Reactivity Plan RE-G-05 only provides Boration rate and CEA Insertion required to downpower to 90%. Although AOP 2575, Attachment 7, provides guidance for reducing power if a reactivity plan is not available, guidance exists for use of a reactivity plan appropriate for the evolution being performed.

Plausible because this guidance is adequate to remove the feed pump from service, but does not provide the complete guidance for reactivity control after the pump is secured. o is incorrect; Reactivity Plan RE-G-10 provides guidance on reducing power to 55%; however, the reduction rate will NOT allow the Feed Pump to be removed from service in the required 60 minute time limit. OP 2393 provides guidance for maintaining ASI control with CEAs, but this guidance does NOT take into consideration the amount of Boric Acid to used to counteract Xenon. Plausible because this reactivity plan will allow performance of the downpower to remove the Main Feed Pump from service, but it does NOT allow reducing power fast enough to meet the 60 minute time limit. The examinee may mistakenly assume that the reactivity plan does NOT inclL'de the effects of Xenon; therefore, the power reduction rate would actually be higher than indicated on RE-G-10. References I

RE-G-03, Rapid Down Power Reactivity Plans (Question Reference

1. 1 of 3) AOP 2575, Rapid Oownpower (NOT provided to examinees)

'Comments and Questi()n Modification HistOry I Bob K. 3/W ("Bad question, "0" could be correct. Reword stem to say "within the next 60 minutes" and reword "C" & "0" to include 2204 and new reactivity plan). Swapped Distaractor "A" and "8". Added OP 2204 to the correct answer (C.). Changed reactivity Plan to RE-G*10 instead of RE-G-14. (RE-G-14 may be appropriate as long as the crew stops at the required power level. Deleted OP 2204 and inserted OP2393. RJA Bill M. 2/C, 0( (Doesn't feel that B is plausible.

No recommendation.)

Discussed.

No modification needed. Distractors will allow the required downpower; therefore, all are plausible.

-RJA Angelo 4/C; No comments.

NRC KIA System/E/A System 2.1 Conduct of Operations NRC KIA Generic System 2.1 Conduct of Operations Page 57 of 77 Printed on 10/28/2009 at 13:20 I . Question #: 19 Question ID: 9000013 0 RO SRO Student Handout? o Lower Order? l-SRO Que;;. # 19 Rev. 0 Selected for Exam Otfgin: New Past NRC Exam? Number 2:1.43 RO 4.1 SRO 4.3 CFR Link (CFR: 41.10/43.6/45.6)

Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plalt, fuel depletion, etc. RWST -2pump BAST -1%/min RWST -1 pump RE-G-05 RE-G-06 RE-G-04 BA vol -880 gals BA vol -880 gals BA vol -204 gals 90% BA flow -20 gpm CEApas -180steps BA flow -44 gpm BAflow-88gpm CEA pas -180 steps CEA pas -180 steps f---.-------jf----------+

............-o Time -10 min. Time-20 min. Time -10 Illill. ----I Rate -30%/hr Rate -60%/hr Rate -60%/hr f RE-G-09 RE-G-07 RE-G-08 BA vol-2458 gals BA vol -2640 gals BA vol -606 gals 70% BA flow-44 gpm BA flow-88 gpm SA flow -20 gpm CEA pas -180 steps CEA pas -180 steps CEA pas -180 steps (No CEAs)

• Time-56 min. Time -30 min . ! Time -30 min. Rate -32%/hr Rate -60%/hr I Rate -60%/hr RE-G-10 RE-G-11 RE-G-12 BA vol -720 gals BA flow-44 gpm BA vol-3080 gals SA vol -3315 gals BA flow -88 gpm BA flow -18 gpm 55% CEA pas -154 steps CEA pas -154 steps CEA pas -154 steps Time-70 min. i Time-38 min.

• Time -40 min. _._--+---------11 Rate -39%/hr I Rate -71%/hr I Rate -68%/hr I RE-G-13 RE-G-14 RE-G-15 BA vol -4400 gals SA vol -5120 gals BA vol-1100 gals BA flow-44 gpm BA flow -88 gpm BA flow -16 gpm CEA pas -140 steps 15% CEA pas -140 steps CEA pas -140 steps I Time-100 min. Time-58 min. Time -80 min. I Rate -51 %/hr ! Rate -88°J6/hr iRate -64 %/hr Page 58 of 77 Printed on 10/28/2009 at 13:20 . Question #: 19 l-SRO Que;;. # 19 Number 2:1.43 Question ID: 9000013 0 RO SRO Student Handout? Rev. 0 Selected for Exam Otfgin: New RO 4.1 SRO 4.3 CFR Link (CFR: 41.10/43.6/45.6) o Lower Order? Past NRC Exam? Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plalt, fuel depletion, etc. 90% RWST -1 pump RE-G-04 BA vol -880 gals BA flow -44 gpm CEApas -180steps RWST -2pump RE-G-05 BA vol -880 gals BAflow-88gpm CEA pas -180 steps BAST -1%/min RE-G-06 BA vol -204 gals BA flow -20 gpm CEA pas -180 steps f---.-------jf----------+

............ --Time-20 min. o Rate -30%/hr Time -10 Illill. Rate -60%/hr f 70% (No CEAs)

• 55% RE-G-07 BA vol-2458 gals BA flow-44 gpm CEA pas -180 steps Time-56 min. Rate -32%/hr RE-G-10 BA vol-3080 gals BA flow-44 gpm CEA pas -154 steps RE-G-08 BA vol -2640 gals BA flow-88 gpm CEA pas -180 steps Time -30 min . Rate -60%/hr RE-G-11 SA vol -3315 gals BA flow -88 gpm CEA pas -154 steps Time-70 min. i Time-38 min. _._--+-----Rate -39%/hr I Rate -71%/hr I

+-15% RE-G-13 RE-G-14 BA vol -4400 gals BA flow-44 gpm CEA pas -140 steps SA vol -5120 gals BA flow -88 gpm CEA pas -140 steps Time-100 min. Time-58 min. Rate -51 %/hr ! Rate -88°J6/hr Page 58 of 77 Time -10 min. ----I Rate -60%/hr RE-G-09 BA vol -606 gals SA flow -20 gpm CEA pas -180 steps ! Time -30 min. I Rate -60%/hr RE-G-12 BA vol -720 gals BA flow -18 gpm CEA pas -154 steps

• Time -40 min. ----11 I Rate -68%/hr I -RE-G-15 BA vol-1100 gals BA flow -16 gpm I CEA pas -140 steps I Time -80 min. iRate -64 %/hr Printed on 10/28/2009 at 13:20 I Question #: 19 RO SRO Student Handout? Lower Order? Ques60n ID: 9000013 J-SRO Ques. # 19 Rev. 0 Selected for Exam Origin: New Past NRC Exam? ]\Ilillstone Unit 2 AOP 2575 Revision 004-01 Rapid Downpower Page 40f35 Reactivity plans are provided and should be lIsed if the initial conditions specified in the plans (lOOOf initial power. ARO. desired final power level) are approximately as specified.

These plans arc an approximation of the required boration and CEA movement required to reach the desired power while controlling the ASI oscillation.

TIle reactivity plan CEA positioning will maintain ASI within COLR limits. OP-2393, "Core Power Monitoring Distribution and Control." provides direction for maintaining A.I)I control within specified bands during conditions, transient conditions, or at the direction of Reactor Engineering.

The reactivity plans provide the above Reactor Engineering direction for ASI control. ASl control during the down power in accordance with the reactivity plan is prefered, however, it shou Id not interfere with event mitigation.

Once reactor power is stabilized, ASl shDuld be maintained in accordance with OP 2393 or the reactivity plan. The first page of the reactivity plan provides the boration rate to initiate the down power and the desired CEA position for ASI control. The second page of the plan contains more detailed information for stabilizing the plant at the desired power level. TIllS page should not be interpreted as procedural direction and deviation from this guidance is allowable to achieve the desired power level within the desired time. The third page provides a prediction of the relative ASl trend during the down power. The ASI trend should not be used as an indication of a true absolute ASl value. If at the completion of the down power, it was noted that significant deviation from the plan was required to achieve the desired power level, reactor engineering should be promptly informed.

Core Reactivity affecls from Res temperature changes vary significantly over core life, based upon ReS boron concentration and resultant Moderator Temperature Coefficient (MTC) value. At 100(Yt:* power, beginning of life (BOL). after xenon equilibrium, the value of MTC is much negative than at the end of life (EOL). This means that at BOL, a change of 1°F ReS temperature

\...iII cause approximately a t!z % change in power, whereas a 1°F ReS temperature change at EOL will cause approximately a 2% change in power. 1. I Applicability This procedure is applicable in Mode 1 at power levels grealer than 20<:'1, when an emergency power reductioll is required.

Level of Use STOP THINK ACT AfMEW Continuous Page 59 of 77 Printed on 10/28/2009 at 13:20 Question #: J-SRO Ques. # 19 Ques60n ID: 9000013 RO SRO Student Handout? 19 Rev. 0 Selected for Exam Origin: New ]\Ilillstone Unit 2 Rapid Downpower AOP 2575 Revision 004-01 Page 40f35 Reactivity plans are provided and should be lIsed if the initial conditions specified in the plans (lOOOf initial power. ARO. desired final power level) are approximately as specified.

These plans arc an approximation of the required boration and CEA movement required to reach the desired power while controlling the ASI oscillation.

TIle reactivity plan CEA positioning will maintain ASI within COLR limits. OP-2393, "Core Power Monitoring Distribution and Control." provides direction for maintaining A.I)I control within specified bands during conditions, transient conditions, or at the direction of Reactor Engineering.

The reactivity plans provide the above Reactor Engineering direction for ASI control. A.l)l control during the down power in accordance with the reactivity plan is prefered, however, it shou Id not interfere with event mitigation.

Once reactor power is stabilized, ASl shDuld be maintained in accordance with OP 2393 or the reactivity plan. The first page of the reactivity plan provides the boration rate to initiate the down power and the desired CEA position for ASI control. The second page of the plan contains more detailed information for stabilizing the plant at the desired power level. TIllS page should not be interpreted as procedural direction and deviation from this guidance is allowable to achieve the desired power level within the desired time. The third page provides a prediction of the relative ASl trend during the down power. The ASI trend should not be used as an indication of a true absolute ASl value. If at the completion of the down power, it was noted that significant deviation from the plan was required to achieve the desired power level, reactor engineering should be promptly informed.

Core Reactivity affecls from Res temperature changes vary significantly over core life, based upon ReS boron concentration and resultant Moderator Temperature Coefficient (MTC) value. At IOO(Yt:* power, beginning of life (BOL). after xenon equilibrium, the value of MTC is much negative than at the end of life (EOL). This means that at BOL, a change of 1°F ReS temperature

\ ... iII cause approximately a I!Z % change in power, whereas a 1°F ReS temperature change at EOL will cause approximately a 2% change in power. 1. I Applicability This procedure is applicable in Mode 1 at power levels grealer than 20<:'1, when an emergency power reductioll is required.

Level of Use Continuous STOP THINK ACT AfMEW Lower Order? Page 59 of 77 Printed on 10/28/2009 at 13:20 Question #: 19 Student Handout? Lower Order? I-SRO Ques. # 19 Rev. 0'111 Selected for Exam Ot1gin: New Past NRC Exam? Question ID: 9000013 RO SRO Unit 2 AOP 2575 Revision 004-01 Rapid Downpower Page 70f35 INSTRUCI-IONS CONTINGENCY ACTIONS CAUTION In the case of a dropped eEA, rod motion is not used to initiate dO\\"11 power. JEuol down powering due to a dropped rod, lNSERT Group 7 CEAs [) +/-2 steps to initiate downpower.

Refer lb PPC or Reactor 3.5.1 IE reactor is flot at the reactivity plan Engineering Curve and Data Book initial conditions.

Refer To and OBTAIN reactivity plan for the Attachment

7. and DETERMINE inital reactor power condition and desired rate of load reduction for time desired load reduction.

in core life.

to borate from the RWST (prefem:d method) PE RFORM the following; ENSURE at least one charging pump operating. ENSURE CH-l96, vcr makeup bypass, dosed. ENSURE CH-504. RWST to charging suction, open. OPEN CH-ln. d.l IE CH-192, RWST isolation, RWST isolation.

can not be opened, Go 1b step 3.8. (:L()SE C:H..501, vcr outlet isolation.

1'. CHECK charging flow f.1 START additional charging pumps at desired rate. as needed and balance charging and letdown. e,O 1h step 3.9. Bal\cd on required rate of down power, START additional charging pumps as necessary and balance charging and letdown .. I I i I Level of Use STOP THINK AOT REVIEW Continuous Page 60 of Printed on 10/28/2009 at 13:20 Question #: 19 I-SRO Ques. # 19 Question ID: 9000013 RO SRO Student Handout? Rev. 0'111 Selected for Exam Ot1gin: New Unit 2 Rapid Downpower INSTRUCI-IONS AOP 2575 Revision 004-01 Page 70f35 CONTINGENCY ACTIONS CAUTION In the case of a dropped eEA, rod motion is not used to initiate dO\\"11 power. JEuol down powering due to a dropped rod, lNSERT Group 7 CEAs [) +/- 2 steps to initiate downpower.

Refer lb PPC or Reactor 3.5.1 Engineering Curve and Data Book and OBTAIN reactivity plan for the inital reactor power condition and desired load reduction.

to borate from the RWST (prefem:d method) PE RFORM the following;

a. ENSURE at least one charging pump operating.

b. ENSURE CH-l96, vcr makeup bypass, dosed. c. ENSURE CH-504. RWST to charging suction, open. d. OPEN CH-ln. RWST isolation.

e. (:L()SE C:H .. 501, vcr outlet isolation.

1'. CHECK charging flow at desired rate. g. e,O 1h step 3.9. Bal\cd on required rate of down power, START additional charging pumps as necessary and balance charging and letdown .. IE reactor is flot at the reactivity plan initial conditions.

Refer To Attachment

7. and DETERMINE desired rate of load reduction for time in core life. d.l IE CH-192, RWST isolation, can not be opened, Go 1b step 3.8. f.1 START additional charging pumps as needed and balance charging and letdown. I I i I Level of Use Continuous STOP THINK AOT REVIEW Lower Order? Past NRC Exam? Page 60 of 77 Printed on 10/28/2009 at 13:20 Question #: 20 Question ID: 9000014 [J RO SRO D Student Handout? Lower Order? I-SRO Ques. # 20 Rev. 0 Selected for Exam Origin: New D Past NRC Exam? Whicll of tile following actions require authorization by the Refueling SRO? D When the grapple will NOT disengage the top of a fuel assembly, snap or twang the hoist cable to release the grapple. When an overload occurs, use the hand crank on the refuel machine hoist to free the fuel assembly DB from the guide pins. In an 13mergency, insert a fuel assembly into the core and ungrapple it provided NO other fuel DC assemblies are adjacent. If an underload occurs prematurely, raise the fuel assembly, pull the mast detent pin, rotate slightly, and re,insert the assembly.

Justification I D IS CORRECT; OP 2209A, Attachment 3, provides a listing of SRO responsibilities.

Included is statement that requires from the Refuel SRO to perform various action contained in Attachment

4. Under "Difficulty Inserting", the Refueling SRO responsible for authorizing the Mast Detent Pin to be pulled. If necessary, he/she may also authorize a slight rotation of the masl allow inserting a fuel A is incorrect.

A Caution in Attachment 4 states, "Snapping or twanging the hoist cable is prohibited." However, the cable may manipulated or pulled and gently released to eliminate a grapple Plausible because the examinee may be confused about what actually constitutes a hoist cable B is incorrect; While Attachment 4 allows the Refuel SRO to authorize several actions to free a fuel assembly (seen as an manually mani Julating the cable hoist is NOT one of them. The refuel machine may be moved in either horizontal plane to free Plausible because Attachment 4 allows the Refuel SRO to authorize movement of the refuel machine manually (hand crank) in horizontal direGtion, just NOT in the vertical C is incorrect; A fuel assembly cannot be left unsupported, even in an Plausible because, in an emergency, the SRO may authorize a fuel bundle to be inserted into any open (unsupported) location in core; however, the grapple must remain attached .

I OP 2209A, Re"ueling Comments and Question Modification History Bob K. -D-5/C (Operator involvement in refuel operations is very limited. Question is still Bill M. -D-4/W, G (Didn't know due to limited involvement in fuel movement.

Question is Angelo 5MI; Not an operator NRC KIA System/l=/A System 2.1 Conduct of Operations I

KIA NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.35 RO 2.2 SRO 3.9 CFR Link (CFR: 41.10 /43.7) Knowledge of the fuel-handling responsibilities of SROs. Page 61 of Printed on 10/28/2009 at 13:20 Question #: 20 I-SRO Ques. # 20 Question ID: 9000014 [J RO SRO Rev. 0 Selected for Exam D Student Handout? Origin: New Whicll of tile following actions require authorization by the Refueling SRO? Lower Order? D Past NRC Exam? D A When the grapple will NOT disengage the top of a fuel assembly, snap or twang the hoist cable to release the grapple. DB When an overload occurs, use the hand crank on the refuel machine hoist to free the fuel assembly from the guide pins. DC In an 13mergency, insert a fuel assembly into the core and ungrapple it provided NO other fuel assemblies are adjacent. D If an underload occurs prematurely, raise the fuel assembly, pull the mast detent pin, rotate slightly, and re,insert the assembly.

Justification I D IS CORRECT; OP 2209A, Attachment 3, provides a listing of SRO responsibilities.

Included is statement that requires authorization from the Refuel SRO to perform various action contained in Attachment

4. Under "Difficulty Inserting", the Refueling SRO is responsible for authorizing the Mast Detent Pin to be pulled. If necessary, he/she may also authorize a slight rotation of the masl to allow inserting a fuel assembly.

A is incorrect.

A Caution in Attachment 4 states, "Snapping or twanging the hoist cable is prohibited." However, the cable may be manipulated or pulled and gently released to eliminate a grapple hang-up. Plausible because the examinee may be confused about what actually constitutes a hoist cable manipulation.

B is incorrect; While Attachment 4 allows the Refuel SRO to authorize several actions to free a fuel assembly (seen as an overload), manually mani Julating the cable hoist is NOT one of them. The refuel machine may be moved in either horizontal plane to free an assembly.

Plausible because Attachment 4 allows the Refuel SRO to authorize movement of the refuel machine manually (hand crank) in either horizontal direGtion, just NOT in the vertical direction.

C is incorrect; A fuel assembly cannot be left unsupported, even in an emergency. Plausible because, in an emergency, the SRO may authorize a fuel bundle to be inserted into any open (unsupported) location in the core; however, the grapple must remain attached . . References I OP 2209A, Re"ueling Operations Comments and Question Modification History I Bob K. -D-5/C (Operator involvement in refuel operations is very limited. Question is still acceptable).

Bill M. -D-4/W, G (Didn't know due to limited involvement in fuel movement.

Question is acceptable.)

Angelo 5MI; Not an operator task. NRC KIA System/l=/A System 2.1 Conduct of Operations I

KIA NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.35 RO 2.2 SRO 3.9 CFR Link (CFR: 41.10 /43.7) Knowledge of the fuel-handling responsibilities of SROs. Page 61 of 77 Printed on 10/28/2009 at 13:20 Question #: 20 Question ID: 9000014 RO SRO o Student Handout? Lower Order? P-SRO Ques. # 20 Rev. 0 Selected for Exam Origin: New Past NRC Exam? Attachment 3

Responsibilities During Refueling Operations (SheetS of 6) 5. Rrfucl SRO

• Authority Stop CORE ALTERATIONS when deemed necessary. Stop or defer any activity around the rcfuel nour which would jcollardizc the safety of personnel or ellllipment.

• Be present on the Refuel Floor amI responsible for maintaining OPS Procedures us required during the following CORE ALTERATIONS:

• Fuel << Moving/replacing

• CEA shuffle in the reactor vessel << Removing the upper guide structure from the reactor vessel 10 {Jm;oupling of CEA extension shafts

• Removal of CEA extension shafts from the UGS << Recoupling of CEA extension shafts Any other CORE ALTERATION as determined by Reactor Engineering As necessmy, the Refuel SRO following consultation with Reador Engineering, authorizes that specified guidelines on Attachment 4 he performed by refueling personnel. General monitoring responsibilities include but are not limited to the following: Ensure the Exclusion Area around the refuel pool is maintained per MA-AA-I02, "Foreign Material Excl\lSion" and DNAP-2000. "Dominion Work Management Process." Ensure proper radiological practices arc maintained around the refuel pool. Ensure safe load path,,, per MP 27 .l2B2, "Overhcad Cranc Operating Information," alld MP 2712131, "Control Heavy Load"' arc maintained. Ensure genera! safety of personnel and C{luipment. Ensure communications between refuel floor and the Omtrol Room arc maintained during CORE ALTERATIONS. Halt CORE ALfERXnONS if communications are losl. Ensure refueling equipmcnt is operated in accordance with OPS-FH 2.15, "Refueling Machine Operation." .----L-e-ve-I-o-t OP2209A STOP THINK ACT REVIEW Rev. 026-(}() 46 of 64 Page 62 Printed on 10/28/2009 at 13:20 Question #: 20 P-SRO Ques. # 20 Question ID: 9000014 RO SRO o Student Handout? Rev. 0 Selected for Exam Origin: New Attachment 3

Responsibilities During Refueling Operations (SheetS of 6) 5. Rrfucl SRO

• Authority

• Stop CORE ALTERATIONS when deemed necessary.

• Stop or defer any activity around the rcfuel nour which would jcollardizc the safety of personnel or ellllipment.

• Be present on the Refuel Floor amI responsible for maintaining OPS Procedures us required during the following CORE ALTERATIONS:

• Fuel Shuffle << Moving/replacing sources

• CEA shuffle in the reactor vessel << Removing the upper guide structure from the reactor vessel * {Jm;oupling of CEA extension shafts

• Removal of CEA extension shafts from the UGS << Recoupling of CEA extension shafts Lower Order? Past NRC Exam? 10

• Any other CORE ALTERATION as determined by Reactor Engineering

• As necessmy, the Refuel SRO following consultation with Reador Engineering, authorizes that specified guidelines on Attachment 4 he performed by refueling personnel.

• General monitoring responsibilities include but are not limited to the following:

• Ensure the Exclusion Area around the refuel pool is maintained per MA-AA-I02, "Foreign Material Excl\lSion" and DNAP-2000. "Dominion Work Management Process."

• Ensure proper radiological practices arc maintained around the refuel pool.

• Ensure safe load path,,, per MP 27 .l2B2, "Overhcad Cranc Operating Information," alld MP 2712131, "Control Heavy Load"' arc maintained.

• Ensure genera! safety of personnel and C{luipment.

• Ensure communications between refuel floor and the Omtrol Room arc maintained during CORE ALTERATIONS.

• .. Halt CORE ALfERXnONS if communications are losl. Ensure refueling equipmcnt is operated in accordance with OPS-FH 2.15, "Refueling Machine Operation." .----L-e-ve-I-o-t OP2209A STOP THINK ACT Page 62 of?? REVIEW Rev. 026-(}() 46 of 64 Printed on 10/28/2009 at 13:20 "Parents" Or "UtIIUl Question #: 20 I Question ID: 9000014 RO SRO o Student Handout? Lower Order? II-SRO Ques. # 20 Selected for Exam Orlgin: New Past NRC Exam?Rev. 0 Attachment 4 Guidelines For Fuel Movement Operations (Sheet::!

of 5) CAUTION Snapping or twanging of the hoist cable is prohibited. Hoist Cable Manipulatioll

-Manipulation or pulling on the hoist cable is typically recommended only to assist in the following: Allow engagement of the bottom nozzle with the <-'ore support plate guide pins. Facilitate fuel movement when "hang-ups" or problems with grapple engagement or disengagement are encountered. Facilitate fuel movement when potential grid interferences are encountered. Difficull,j' Inserting inserting a fuel assembly into the core, SFP storage rack, fuel elevator.

or upender can, an underload occurs, PERFORM the following:

L RAISE the assembly until the underload is cleared. 2. CHECK alignment of fuel assembly tUld fixture. 3. IF relX)sitioning a fuel assembly nl<lllllBlIy over the core, ENSURE spreader is raised. 4. As necessary, REPOSITION fuel a"tsembiy and TRY reinserting.

5. IF an underload is experienced again, PERFORM any of the following: PULL the RFM Illasl detent pin out tmd TRY reinserting. ROTATE the RFM mast slightly in the clockwise or counterclockwise direction and TRY reinserting. ENSURE fuel assembly is raised 4" from the core support plate to dear the guide pins and HAND CRANK the RFM up to 0.3" in any direction and TRY reinserting. IF the above docs not clear underload, manipUlate the hoist cable to free the fuel assembly from potential grid interferences.

Level OP2209A STOP THINK Rev. tJ26-06 49 of 64 Page 63 of Printed on 10/28/2009 at 13 :20 "Parents" Or "UtIIUl Question #: 20 I Question ID: 9000014 RO SRO o Student Handout? Lower Order? II-SRO Ques. # 20 Rev. 0 Selected for Exam Orlgin: New Attachment 4 Guidelines For Fuel Movement Operations (Sheet::!

of 5) CAUTION Snapping or twanging of the hoist cable is prohibited.

• Hoist Cable Manipulatioll

-Manipulation or pulling on the hoist cable is typically recommended only to assist in the following:

• Allow engagement of the bottom nozzle with the <-'ore support plate guide pins.

• Facilitate fuel movement when "hang-ups" or problems with grapple engagement or disengagement are encountered.

• Facilitate fuel movement when potential grid interferences are encountered.

• Difficull,j' Inserting inserting a fuel assembly into the core, SFP storage rack, fuel elevator.

or upender can, an underload occurs, PERFORM the following:

L RAISE the assembly until the underload is cleared. 2. CHECK alignment of fuel assembly tUld fixture. 3. IF relX)sitioning a fuel assembly nl<lllllBlIy over the core, ENSURE spreader is raised. 4. As necessary, REPOSITION fuel a"tsembiy and TRY reinserting.

5. IF an underload is experienced again, PERFORM any of the following:

5.1 PULL the RFM Illasl detent pin out tmd TRY reinserting.

5.2 ROTATE

the RFM mast slightly in the clockwise or counterclockwise direction and TRY reinserting.

5.3 ENSURE

fuel assembly is raised 4" from the core support plate to dear the guide pins and HAND CRANK the RFM up to 0.3" in any direction and TRY reinserting.

5,4 IF the above docs not clear underload, manipUlate the hoist cable to free the fuel assembly from potential grid interferences.

Level STOP THINK OP2209A Rev. tJ26-06 49 of 64 Past NRC Exam? Page 63 of 77 Printed on 10/28/2009 at 13 :20 Question #: 21 Question ID: 9000024 ORO !.,I SRO o Student Handout? Lower Order? I*SRO Ques. # 21 Rev. 0 Selected for Exam Origin: New Past NRC Exam? The plant is operating at 100% power when ISO New England and CONVEX operators notify Millstone Station that a "Degraded Voltage" condition exists. Voltage on the 4.16 kV buses is presently 3,900 volts. Based on this information, which one of the following describes actions that the Unit Supervisor must direct. per the applicable procedures?

A Rack out the 6.9 and 4.16 kV breakers to the RSST and slow-start both Emergency Diesel Generators.

'r'£ B Terminate surveillance testing of any safety related pumps and motors and secure them, if possible.

C Commence a plant down power and secure all unnecessary equipment as the lower power permits. D Ensure the "E" and "F" Instrument Air compressors are operating in the "Lead" and "Standby" modes. Justification I B

• CORRECT; To limit the risk of damage to safety related motor windings due to the higher current flows that would be expected, unnecessary nmning of these components must be A WRONG; The RSST breakers are not disabled until voltage drops below 88.9% of rated voltage. The EOG are verified as only if the RSST is in Plausible; Examinee may believe that in order to prevent transferring to the RSST, which is getting power from a degraded voltage, the RSST breakers must be disabled so they will NOT close on a possible plant trip. Also, prestaging the EDGs with a start" would put minimur1 stress on the machines which are destined to carry all plant AC loads. However, this action is premature over conservat;ve for the: given C WRONG; The applicable AOP Does not direct a plant down power be commenced as the loss of power to the grid is a far impact than any gains by securing Plausible; Examinee may believe that because the applicable AOP directs that unnecessary loads be secured to help with degraded voltage, and a trip from a lower power level is preferred, that lowering power to allow securing of components is o . WRONG; The restart of the vital Instrument Air Compressors is handled by EOP-2525 post-trip.

There is no benefit to prestaging their alignment as they must be locally "reset" on a loss of power regardless.

Plausible; Examinee may realize that on a probable trip from loss of the grid, Instrument Air recovery will require operator action, as the compressor that is normally running is not vitally powered. Therefore, in order to ensure Instrument Air remains available (and a Vital Auxiliary Safety Function is preserved), the prestaging of the IA compressor alignment is a logical action. References I AOP-2580, Pg. 6, Step 3.1 Comments and Question Modification History Bill M. 0-2/C, K Angelo D-4/C; No comments.

NRC KIA SystemtEtA System 2.2 Equipment Control ....'n..rll'K/A NRC KIA Generic System 2.2 Equipment Control Number 2.:Z.17 RO 2.6 SRO 3.8 CFR Link (CFR: 41.10 143.5/45.13Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, prioritization, Clnd coordination with the transmission system Page 64 of 77 Printed on 10/28/2009 at 13:20 Question #: 21 Question ID: 9000024 ORO !.,I SRO o Student Handout? Lower Order? I*SRO Ques. # 21 Rev. 0 Selected for Exam Origin: New Past NRC Exam? The plant is operating at 100% power when ISO New England and CONVEX operators notify Millstone Station that a "Degraded Voltage" condition exists. Voltage on the 4.16 kV buses is presently 3,900 volts. Based on this information, which one of the following describes actions that the Unit Supervisor must direct. per the applicable procedures?

A Rack out the 6.9 and 4.16 kV breakers to the RSST and slow-start both Emergency Diesel Generators.

'r'£ B Terminate surveillance testing of any safety related pumps and motors and secure them, if possible.

C Commence a plant down power and secure all unnecessary equipment as the lower power permits. D Ensure the "E" and "F" Instrument Air compressors are operating in the "Lead" and "Standby" modes. Justification I B

• CORRECT; To limit the risk of damage to safety related motor windings due to the higher current flows that would be expected, all unnecessary nmning of these components must be terminated.

A WRONG; The RSST breakers are not disabled until voltage drops below 88.9% of rated voltage. The EOG are verified as running only if the RSST is in service. Plausible; Examinee may believe that in order to prevent transferring to the RSST, which is getting power from a degraded grid voltage, the RSST breakers must be disabled so they will NOT close on a possible plant trip. Also, prestaging the EDGs with a "slow start" would put minimur1 stress on the machines which are destined to carry all plant AC loads. However, this action is premature and over conservat;ve for the: given conditions.

C WRONG; The applicable AOP Does not direct a plant down power be commenced as the loss of power to the grid is a far worse impact than any gains by securing equipment.

Plausible; Examinee may believe that because the applicable AOP directs that unnecessary loads be secured to help with the degraded voltage, and a trip from a lower power level is preferred, that lowering power to allow securing of components is logical. o . WRONG; The restart of the vital Instrument Air Compressors is handled by EOP-2525 post-trip.

There is no benefit to prestaging their alignment as they must be locally "reset" on a loss of power regardless.

Plausible; Examinee may realize that on a probable trip from loss of the grid, Instrument Air recovery will require operator action, as the compressor that is normally running is not vitally powered. Therefore, in order to ensure Instrument Air remains available (and a Vital Auxiliary Safety Function is preserved), the prestaging of the IA compressor alignment is a logical action. References I AOP-2580, Pg. 6, Step 3.1 Comments and Question Modification History Bill M. 0-2/C, K Angelo D-4/C; No comments.

NRC KIA SystemtEtA System 2.2 .... 'n .. rll'K/A NRC KIA Generic System 2.2 Equipment Control Equipment Control Number 2.:Z.17 RO 2.6 SRO 3.8 CFR Link (CFR: 41.10 143.5/45.13)

Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, Clnd coordination with the transmission system operator.

Page 64 of 77 Printed on 10/28/2009 at 13:20 Question #: 21 Student Handout? Lower Order? Question ID: 9000024

[J RO SRO [-SRO Ques. # 21 Rev. 0 Selected for Exam Origin: New 3.1 IF surveillances of safety related -pumps and motors are in TERMINATE surveillances dcgratlcd voltage REQUEST the SM refer COP 200.S. "Response to NE/CONEX Emergencies

'" CHECK actual degraded condition exists by observation Al\¥ of the following 41fiOvoil bus 24C OR voltage less than 3,900 48{) voil bus 22E 22F less than 440 Millstone Unit 2 Degraded Voltage 3.0 Degraded Voltage INSTRUCTIONS AOP 2580 Revision 003-04 Page 60f12 CONTINGENCY ACTIONS I of Use STOP THINK AST Page 65 of Printed on 10/28/2009 at 13:20 I Question #: 21 [-SRO Ques. # 21 Question ID: 9000024

[J RO SRO Student Handout? Rev. 0 Selected for Exam Origin: New Millstone Unit 2 Degraded Voltage 3.0 Degraded Voltage INSTRUCTIONS 3.1 IF surveillances of safety related -pumps and motors are in progress, TERMINATE surveillances during dcgratlcd voltage conditions.

3.2 REQUEST

the SM refer to COP 200.S. "Response to ISO NE/CONEX Emergencies and Alerts." '" 3.3 CHECK actual degraded voltage condition exists by observation of Al\¥ of the following conditions:

• 41fiOvoil bus 24C OR 240 voltage less than 3,900 volts

• 48{) voil bus 22E 22F voltage less than 440 volts I I of Use ContinuOI.Js STOP THINK AOP 2580 Revision 003-04 Page 60f12 CONTINGENCY ACTIONS AST REVIEW Lower Order? Page 65 of 77 Printed on 10/28/2009 at 13:20 SRO Exam Questions Only (No "Originals'*)

Question #: 22 Question ID: 9000015 D RO SRO D Student Handout? Lower Order? I-SRO Ques. If 22 Rev. 0 Selected for Exam Origin: New D Past NRC Exam? The Auxiliary Building PEa has just noted an active boric acid leak on the bottom of a flange on CH-198, "RCP Bleedoff Pre,ssure Control Valve to VCT". The leak is very small (2-3 drops per minute), but boric acid deposits from the leak are corroding a pipe support bracket located below the flange. Whictl of ttle following administrative control documents require action be taken to control this leak? A Final Safety Analysis Report, Chapter 15, License Renewal, Aging Management Programs D B Technical Spedfications, Reactor Coolant System Leakage, LCO 3.4.6.2 D C Technical Requirements Manual, Containment Isolation Valves, LCO 3.6.3.1 D D Operational Configuration Control, OP-AA-1500, Alternate Plant Configurations, Attachment 5 Justification 1 A -CORRECT; FSAR, Chapter 15, License Renewal, gives the requirements for a Boric Acid Corrosion control program.

what a boric acid leak is corroding is not a factor in the need to control, stop and repair any problems caused by boric acid B -WRONG; The given system degradation does NOT impact the Tech. Spec. for RCS Plausible; RCP Bleedoff, which is part of the Letdown System, is connected directly to the RCS and has RCS water flowing through however, it is t\OT considered RCS leakage because it can be isolated.

Additionally, it cannot be considered pressure leakage due to it's C -WRONG; -he TRM does not cover boric acid corrosion control and this small a leak would not effect the valve Plausible; Exarninee may incorrectly believe that because this valve is part of the CTMT Isolation Specification in the TRM, leakage from the flange would be covered D -WRONG; Operation Configuration Control deals with control of plant system configuration based on component position not configuration changes due to failures or degradation (corrosion) of system Plausible; Examinee malY think that if a component of a system cannot perform its function due to corrosion, then the configuration rr'ust be I FSAR, Chapter 15, License Renewal, Pg. 15-2 & 3; Boric Acid Corrosion Control Comments and Questilln Modification History Bob K. -D-2/C (Change leak to "letdown to clean RW system" and add admin procedures to all Changed locat.ion of leak to a flange on CH-198. Added specific sections or chapters to each distractor. Bill M. -D-2fW, K (Assumed the leak impacted the RCS Leakage Tech Spec. due to pressure boundary leakage. Discussed realized flange leakage is NOT pressure boundary leakage. Additionally, this leak path is isolable; therefore, not RCS Angelo -D-4/C; No NRC KIA System/E/A System 2.2 Equipment Control KIA Selecte(D NRC KIA Generic System 2.2 Equipment Control Number RO 3.6 SRO 4.5 CFR Link (CFR: 41.7/41.10/43.1

/45.13) Knowledge of conditions and limitations in the facility license. Page 66 of 77 Printed on 10/28/2009 at 13:20 SRO Exam Questions Only (No "Originals'*)

Question #: 22 I-SRO Ques. If 22 Question ID: 9000015 D RO SRO Rev. 0 Selected for Exam D Student Handout? Lower Order? Origin: New D Past NRC Exam? The Auxiliary Building PEa has just noted an active boric acid leak on the bottom of a flange on CH-198, "RCP Bleedoff Pre,ssure Control Valve to VCT". The leak is very small (2-3 drops per minute), but boric acid deposits from the leak are corroding a pipe support bracket located below the flange. Whictl of ttle following administrative control documents require action be taken to control this leak? A Final Safety Analysis Report, Chapter 15, License Renewal, Aging Management Programs D B Technical Spedfications, Reactor Coolant System Leakage, LCO 3.4.6.2 D C Technical Requirements Manual, Containment Isolation Valves, LCO 3.6.3.1 D D Operational Configuration Control, OP-AA-1500, Alternate Plant Configurations, Attachment 5 Justification 1 A -CORRECT; FSAR, Chapter 15, License Renewal, gives the requirements for a Boric Acid Corrosion control program. Specific:ally what a boric acid leak is corroding is not a factor in the need to control, stop and repair any problems caused by boric acid leakage. B -WRONG; The given system degradation does NOT impact the Tech. Spec. for RCS leakage. Plausible; RCP Bleedoff, which is part of the Letdown System, is connected directly to the RCS and has RCS water flowing through it; however, it is t\OT considered RCS leakage because it can be isolated.

Additionally, it cannot be considered pressure boundary leakage due to it's location. C -WRONG; -he TRM does not cover boric acid corrosion control and this small a leak would not effect the valve operability.

Plausible; Exarninee may incorrectly believe that because this valve is part of the CTMT Isolation Specification in the TRM, that leakage from the flange would be covered here. D -WRONG; Operation Configuration Control deals with control of plant system configuration based on component position changes, not configuration changes due to failures or degradation (corrosion) of system components.

Plausible; Examinee malY think that if a component of a system cannot perform its function due to corrosion, then the system configuration rr'ust be affected.

References I FSAR, Chapter 15, License Renewal, Pg. 15-2 & 3; Boric Acid Corrosion Control Comments and Questilln Modification History I Bob K. -D-2/C (Change leak to "letdown to clean RW system" and add admin procedures to all choices).

Changed locat.ion of leak to a flange on CH-198. Added specific sections or chapters to each distractor.

RLC Bill M. -D-2fW, K (Assumed the leak impacted the RCS Leakage Tech Spec. due to pressure boundary leakage. Discussed and realized flange leakage is NOT pressure boundary leakage. Additionally, this leak path is isolable; therefore, not RCS leakaage.)

Angelo -D-4/C; No comments.

NRC KIA System/E/A System 2.2 Equipment Control KIA Selecte(D NRC KIA Generic System 2.2 Equipment Control Number RO 3.6 SRO 4.5 CFR Link (CFR: 41.7/41.10/43.1

/45.13) Knowledge of conditions and limitations in the facility license. Page 66 of 77 Printed on 10/28/2009 at 13:20 i Question #: 22 Question ID; 9OO001S [J RO iV-! SRO Student Handout? Lower Order? [*SRO Que,. # 22 Rev. 0 Selected for Exam Origin; New C Past NRC Exam? 115.2.1.3 Boric Acid Corrosion

---

Iprogram Description Boric Acid Corrosion corresponds to NUREG-1801, Section XLMI 0 "Boric Acid Corrosion." The program manages the aging effect of loss of material and ensures that systems, structures, and components susceptible to boric acid corrosion are properly monitored.

The program uses rvisual inspec::tions to detect the boric acid leakage source, path, and any targets of the leakage. It ensures that boric acid corrosion is consistently identified, documented, evaluated, trended, and effectively repaired.

The Boric Acid Corrosion program provides both detection and analysis of leakage of borated water inside containment.

The General Condition Monitoring program is the rimary method for detecting borated water leakage outside containment The analysis of the leakage is performt:d through the Boric Acid Corrosion program. Any necessary corrective actions are implemented through the Corrective Action Program. oric Acid Corrosion program implements the requirements of: -NRC Bulletin 2001-01 (Reference 15.2-15) -NRC Bulletin 2002-01 (Reference 15.2-16) -NRC Bulletin 2002-02 (Reference 15.2-17) -NRC Bulletin 2003-02 (Reference 15.2-18) -NRC Order EA-03-009 (Reference 15.2-19) -NRC Bulletin 2004-01 (Reference 15.2-20) e acceptance criterion is the absence of any boric acid leakage or precipitation.

If boric acid leakage or precipitation is found by any personnel, it is required to be reported using the Corrective Action Program. Corrective actions for conditions that are adverse to quality are erformed in accordance with the Corrective Action Program as part of the Quality Assurance rogram. The corrective action process provides reasonable assurance that deficiencies adverse quality are either promptly corrected or are evaluated to be acceptable.

Page 67 of 77 Printed on 10/28/2009 at 13:20 Question #: 22 Question ID; 9OO001S [J RO iV-! SRO Student Handout? Lower Order? [*SRO Que,. # 22 Rev. 0 Selected for Exam Origin; New C Past NRC Exam? 115.2.1.3 Boric Acid Corrosion

---

i Iprogram Description Boric Acid Corrosion corresponds to NUREG-1801, Section XLMI 0 "Boric Acid Corrosion." The program manages the aging effect of loss of material and ensures that systems, structures, and components susceptible to boric acid corrosion are properly monitored.

The program uses rvisual inspec::tions to detect the boric acid leakage source, path, and any targets of the leakage. It ensures that boric acid corrosion is consistently identified, documented, evaluated, trended, and effectively repaired.

The Boric Acid Corrosion program provides both detection and analysis of leakage of borated water inside containment.

The General Condition Monitoring program is the rimary method for detecting borated water leakage outside containment The analysis of the leakage is performt:d through the Boric Acid Corrosion program. Any necessary corrective actions are implemented through the Corrective Action Program. oric Acid Corrosion program implements the requirements of: -NRC Bulletin 2001-01 (Reference 15.2-15) -NRC Bulletin 2002-01 (Reference 15.2-16) -NRC Bulletin 2002-02 (Reference 15.2-17) -NRC Bulletin 2003-02 (Reference 15.2-18) -NRC Order EA-03-009 (Reference 15.2-19) -NRC Bulletin 2004-01 (Reference 15.2-20) e acceptance criterion is the absence of any boric acid leakage or precipitation.

If boric acid leakage or precipitation is found by any personnel, it is required to be reported using the Corrective Action Program. Corrective actions for conditions that are adverse to quality are erformed in accordance with the Corrective Action Program as part of the Quality Assurance rogram. The corrective action process provides reasonable assurance that deficiencies adverse quality are either promptly corrected or are evaluated to be acceptable.

Page 67 of 77 Printed on 10/28/2009 at 13:20 Question #: 22 RO SRO Student Handout? Lower Order? Question ID: 9000015 I-SRO Que!'. II 22 Rev. 0 Selected for Exam Origin: New Past NRC Exam? UNITED NUCLEAR REGULATORY WASHINGTON, D.C.

DOMINION NUCLEAR CONNECTICUT, INC. DOCKET NO. 50-336 (MILLSTONE POWER STATION. UNIT NO.2) RENEWED FACILITY OPERATING LICENSE Renewed License No. DPR-65 The U.S. Nuclear Regulatory Commission (the Commission) having previously made findings set forth in License No. DPR-55 issued on September 26, 1975 has now The application to renew License DPR-65 filed by Dominion Nuclear Connecticut, Inc. (DNC), complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Part 54 Chapter 1, and all required notifications to other agencies or bodies have been duly made; Construction of the Millstone Power Station, Unit 2, (facility) has been substantially completed in conformity with Construction Permit No. CPPR-76 and the application, as amended, the provisions of the Act and the rules and regulations of the Commission;

......, .....--

!C. Actions have been identified and have been or will be taken with respect to : (1) managing the effects of aging during the period of extended operation on the: functionality of structures and components that have been identified to require review under 10 CFR 54.21 (a)(1 }, and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21 (c), such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in .........

................................The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance: (i) that the activities authorized by this renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; Page 68 of Printed on 10/28/2009 at 13:20 Question #: 22 I-SRO Que!'. II 22 Question ID: 9000015 RO SRO Student Handout? Lower Order? Rev. 0 Selected for Exam Origin: New UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-001 DOMINION NUCLEAR CONNECTICUT, INC. DOCKET NO. 50-336 (MILLSTONE POWER STATION. UNIT NO.2) RENEWED FACILITY OPERATING LICENSE Renewed License No. DPR-65 Past NRC Exam? 1. The U.S. Nuclear Regulatory Commission (the Commission) having previously made the findings set forth in License No. DPR-55 issued on September 26, 1975 has now found that A. The application to renew License DPR-65 filed by Dominion Nuclear Connecticut, Inc. (DNC), complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Part 54 Chapter 1, and all required notifications to other agencies or bodies have been duly made; B. Construction of the Millstone Power Station, Unit 2, (facility) has been substantially completed in conformity with Construction Permit No. CPPR-76 and the application, as amended, the provisions of the Act and the rules and regulations of the Commission;

...... , ..... --..........................*****.....................******.***.................*.*.*****.****** ! C. Actions have been identified and have been or will be taken with respect to : (1) managing the effects of aging during the period of extended operation on the: functionality of structures and components that have been identified to require review under 10 CFR 54.21 (a)(1 }, and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21 (c), such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in .........

................................

.. D. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; E. There is reasonable assurance: (i) that the activities authorized by this renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; Page 68 of 77 Printed on 10/28/2009 at 13:20 SRO Question #: 23 Student Handout? Lower Order?Question ID: 8000044 RO SRO I-SRO Que:;. # 23 Rev. 2 Selected for Exam Odgin: Bank Past NRC Exam? The Rad. Waste PEO has just brought an Aerated Radioactive Waste (ARW) Monitor Tank discharge permit to ttle Shift Manager for review and approval.

Upon reviEwing th,e permit and ARW system status, the SM has noticed that the ARW monitor tank was sampled by chemistry for the generation of the discharge permit with a level of 85%. However, the tank now has an actual level of 95%. Which of the following actions are required in order for the Shift Manager to approve discharging the ARW Monitor Tank? *** " ****** It **** I It

• It

• It ** It It " It ** " *** It It It

• It It " I ** " It ***** It It It " It

• It It ** It *** It It It It *** It

• It It It It It It *** Re-caiculate the amount of the discharge based on the new tank volume, and note this on the existing discharge permit when complete. Re-mi)( the tank for the required period of time, then resample the tank and generate a new discharge permit based on the new sample. Re-sa'llple the tank and generate a second discharge permit and discharge the tank based on the most conservative of the two permits. Re-mix the tank contents to ensure thorough mixing with the previously sampled contents discharge the tank on the existing Justification B CORRECT; The SM must verify, when approving the permit for release, that the ARW discharge is being done in "batch" This means the tank contents must be a discrete quantity with a known level of radioactivity.

Once the tank showed signs of input, the conte!nts were no longer known. Therefore, the tank must be re-sampled before it could be legally A -WRONG; The same discharge permit cannot be used, even if the tank volume is now Plausible; This is actually what is done for every discharge to ensure the actual amount of the discharge is correctly documented. must be done because the total amount listed on the permit assumes every gallon of tank volume will be discharged, but the pump can NOT pump the tank down to zero and often trips off line with several percent left in the C -WRONG; Administrative requirements state that if the Rad. Monitor is operable it will be used per existing Plausible; This is what is done if the ARW discharge Rad. Monitor is NOT operable and the tank must still be discharged based plant D -WRONG; -his must be done before the tank is resampled for generation of a .!:l§Yl! Plausible; This is acceptable if the contents had been known (by sampling) but had stratified and was not indicating properly on Discharge Rad I REMODCM, Rad Waste Sampling Requirements (Batch ,Comments and Question Modification History Bob K. D-2/C (OK -possibly make more challenging by adding "recirc times" to No changes Bill M. D-2/C, Angelo -D-2/C; No NRC KIA System/E/A Generic KlA Selected ..........System 2.3 Radiation Control NRC KIA Generic System 2.3 Radiation Control Number 2.3.6 R02.0 SRO 3.8 CFR Link (CFR: 41.13/43.4/45.10) Ability to approve release permits. Page 69 of Printed on 10/28/2009 at 13:20 Question #: 23 I-SRO Que:;. # 23 SRO Question ID: 8000044 RO SRO Rev. 2 Selected for Exam Student Handout? Lower Order? Odgin: Bank Past NRC Exam? The Rad. Waste PEO has just brought an Aerated Radioactive Waste (ARW) Monitor Tank discharge permit to ttle Shift Manager for review and approval.

Upon reviEwing th,e permit and ARW system status, the SM has noticed that the ARW monitor tank was sampled by chemistry for the generation of the discharge permit with a level of 85%. However, the tank now has an actual level of 95%. Which of the following actions are required in order for the Shift Manager to approve discharging the ARW Monitor Tank? *** " ****** It **** I It

• It

• It ** It It " It ** " *** It It It

• It It " I ** " It ***** It It It " It

• It It ** It *** It It It It *** It

• It It It It It It *** A Re-caiculate the amount of the discharge based on the new tank volume, and note this on the existing discharge permit when complete. B Re-mi)( the tank for the required period of time, then resample the tank and generate a new discharge permit based on the new sample. C Re-sa'llple the tank and generate a second discharge permit and discharge the tank based on the most conservative of the two permits. D Re-mix the tank contents to ensure thorough mixing with the previously sampled contents and discharge the tank on the existing permit. Justification J B CORRECT; The SM must verify, when approving the permit for release, that the ARW discharge is being done in "batch" mode. This means the tank contents must be a discrete quantity with a known level of radioactivity.

Once the tank showed signs of additional input, the conte!nts were no longer known. Therefore, the tank must be re-sampled before it could be legally discharged. A -WRONG; The same discharge permit cannot be used, even if the tank volume is now known. Plausible; This is actually what is done for every discharge to ensure the actual amount of the discharge is correctly documented.

This must be done because the total amount listed on the permit assumes every gallon of tank volume will be discharged, but the discharge pump can NOT pump the tank down to zero and often trips off line with several percent left in the tank. C -WRONG; Administrative requirements state that if the Rad. Monitor is operable it will be used per existing guidelines.

Plausible; This is what is done if the ARW discharge Rad. Monitor is NOT operable and the tank must still be discharged based on plant needs. D -WRONG; -his must be done before the tank is resampled for generation of a .!:l§Yl! permit. Plausible; This is acceptable if the contents had been known (by sampling) but had stratified and was not indicating properly on the Discharge Rad Monitor. References I REMODCM, Rad Waste Sampling Requirements (Batch Discharge). ,Comments and Question Modification History I Bob K. D-2/C (OK -possibly make more challenging by adding "recirc times" to choices) No changes made. Bill M. D-2/C, K Angelo -D-2/C; No comments.

NRC KIA System/E/A System 2.3 Radiation Control Generic KlA Selected ..........

NRC KIA Generic System 2.3 Radiation Control Number 2.3.6 R02.0 SRO 3.8 CFR Link (CFR: 41.13/43.4/45.10)

Ability to approve release permits. Page 69 of 77 Printed on 10/28/2009 at 13:20 Question #: 23 Question ID; 8000044 D RO .'Il. SRO Student Handout? Lower Order? 23 Rev. 2 Selected for Exam Origin: Bank D Past NRC Exam? Table I.C. -2 Millstone Unit 2 Radioactive Liquid Waste Sampling and Analysis Program Liquid Release Sample 'tyPe and Minimum Analysis 'fYpe of Activity Lower Limit Source Freqncncy Frequency Analysis of Detection (LLD}A --_......

(",Cilml)

A.Batch ReleaseR I I LClcan Wm:tc Moni* Grab sample prior 10 Prior to each batch i Principalpanuna

5" w-7 I I tor '!lInk. Aerated each batch release release I Emittcrs c I WaSlc M4.1llilOf I J-lJl 1 x HI-Ii I Tank and SIC,11ll I I Gcncratolllulk lJ. I Cc-144 5 x 10-(1 I Dis.solvcd

& I x 10-5 Entrained 2.Col1dcnsa lc Monthly H-3 1 x 10-5 Polishing Fucility Cornpositc E.G. -Waste Ouarterlv Gross alpha 1 x 10-7 Neutralization SumpE. ComposltcF.*G.

Sr-81), Sr-9(l 5 x Fe-55 I x W-6 B.Continuous Rele;ase LStcam C.rencrator Daily Grab S,tmplel.&

Weekly Principal Gamma 5 x 10-1 Blowdowc 1t prior to aligning to Composih.;E.(t EmiHerse.

2.Scrvicc W:lter long Island Sound for 1-131 I x IO-() Effluent]'

RBCCWsump Cc-144 5 x W-e. 3.1urbin{)

SumpsL, MonLhly Grab

• Monlhly Dissolved

& I x III 5 Sample Enlrained GascsK. 4.RBCCW SurnpM. Week.ly Grab .or Com-Monthly H-3 N. I x to :'\ positc Composih:F.*o.

Week.ly Cmnpo:"ilc Quarterlv Gross alpha I x (()-1 CompositcF.*G.

Sr-H9, Sr-90 5 x IU-x t-"c-55 1 x 10-6 TABLE I.C.-2 ! TABLE NOTATIONS i Page 70 of 77 Printed on 10/28/2009 at 13:20 Question #: 23 Question ID; 8000044 D RO .'Il. SRO Student Handout? Lower Order? 23 Rev. 2 Selected for Exam Origin: Bank D Past NRC Exam? Table I.C. -2 Millstone Unit 2 Radioactive Liquid Waste Sampling and Analysis Program Liquid Release Sample 'tyPe and Minimum Analysis 'fYpe of Activity Lower Limit Source Freqncncy Frequency Analysis of Detection (LLD}A --_ ...... --------------

---

(",Cilml)

A.Batch ReleaseR I I LClcan Wm:tc Moni* Grab sample prior 10 Prior to each batch i Principalpanuna

5" w-7 I I tor '!lInk. Aerated each batch release release I Emittcrs c I WaSlc M4.1llilOf I J-lJl 1 x HI-Ii I Tank and SIC,11ll I I Gcncratolllulk lJ. I Cc-144 5 x 10-(1 1:----------

---

---

I Dis.solvcd

& I x 10-5 Entrained Gases"-2.Col1dcnsa lc Monthly H-3 1 x 10-5 Polishing Fucility Cornpositc E.G. -Waste Ouarterlv Gross alpha 1 x 10-7 Neutralization SumpE. ComposltcF.*G.

Sr-81), Sr-9(l 5 x Fe-55 I x W-6 B.Continuous Rele;ase LStcam C.rencrator Daily Grab S,tmplel.&

Weekly Principal Gamma 5 x 10-1 Blowdowc 1t prior to aligning to Composih.;E.(t EmiHerse.

2.Scrvicc W:lter long Island Sound for 1-131 I x IO-() Effluent]'

RBCCWsump Cc-144 5 x W-e. 3.1urbin{)

SumpsL, MonLhly Grab

• Monlhly Dissolved

& I x III 5 Sample Enlrained GascsK. 4.RBCCW SurnpM. Week.ly Grab .or Com-Monthly H-3 N. I x to :'\ positc Composih:F.*o.

Week.ly Cmnpo:"ilc Quarterlv Gross alpha I x (()-1 CompositcF.*G.

Sr-H9, Sr-90 5 x IU-x t-"c-55 1 x 10-6 TABLE I.C.-2 ! TABLE NOTATIONS i Page 70 of 77 Printed on 10/28/2009 at 13:20 I Question #: 23 Question ID: 8000044 RO SRO ::::J Student Handout? Lower Order? l-SRO Que" # 23 Rev. 2 Selected for Exam Origin: Bank Past NRC Exam? ********1 this tible. Prior to the sampling.

each batch shall be isolated and at least two tankJsump 1 volumes shall be recirculated or equivalent mixing provided.

If the steam generator bulk can 1 not be recirculated prior to batch discharge.

samples will be obtained by representative I 1____________________________________________

_1 compositing during I The LLD will be 5 x 10-7 The principal gamma emitters for which this LLD applies exch;sively the following radionuclides:

Mn-54, Fe-59, Co-58, Co-SO. 2n-65, Cs-'134. Cs--137. and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 10-JtCifmL This list does not mean that only these nuclides are to be detected and Other peaks which are measurable and Identifiable, together with the above nuclides, shall be identified and reported, Nuclides which are below the LLD for the analyses should not reported as being present at the lLD level. When unusual circumstances result in a priori hightlf than required, the reasons shall be documented in the Radioactive Effluent EQ!J/Je Steam Generator IE tho applicable batch gamma activity is not greater than 5 x 10-7 \ICi/ml, ntE.ti the and analysis schedule for gross alpha., Sr-89. Sr-90. Fe-55 are not MP-22-REC-BAPOl STOP TH1NK ACT REVIEW Rev. 026-00 12 of 165 Page 71 of Printed on 10/28/2009 at 13:20 Question #: 23 l-SRO Que" # 23 Question ID: 8000044 RO SRO Rev. 2 Selected for Exam ::::J Student Handout? Lower Order? Origin: Bank Past NRC Exam? ********1 I this tible. Prior to the sampling.

each batch shall be isolated and at least two tankJsump 1 volumes shall be recirculated or equivalent mixing provided.

If the steam generator bulk can 1 not be recirculated prior to batch discharge.

samples will be obtained by representative I 1 compositing during discharge.

I 1 ____________________________________________

_ C. The LLD will be 5 x 10-7 The principal gamma emitters for which this LLD applies are exch;sively the following radionuclides:

Mn-54, Fe-59, Co-58, Co-SO. 2n-65, Mo-99, Cs-'134. Cs--137. and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 10-6 JtCifmL This list does not mean that only these nuclides are to be detected and reported.

Other peaks which are measurable and Identifiable, together with the above nuclides, shall also be identified and reported, Nuclides which are below the LLD for the analyses should not be reported as being present at the lLD level. When unusual circumstances result in a priori LLDs hightlf than required, the reasons shall be documented in the Radioactive Effluent Release Repmt. D. EQ!J/Je Steam Generator Bulk: IE tho applicable batch gamma activity is not greater than 5 x 10-7\ICi/ml, ntE.ti the sampling and analysis schedule for gross alpha., Sr-89. Sr-90. Fe-55 are not required.

STOP TH1NK ACT Page 71 of 77 MP-22-REC-BAPOl REVIEW Rev. 026-00 12 of 165 Printed on 10/28/2009 at 13:20 SROExalIl Q.u.estlons Qnly (No "Parents tt Or ffOf:iIfiD'als U ) Question #: ,24 Student Handout?Question ID: 9000016 RO o Lower Order? I-SRO Ques. it :24 Rev. o Selected for Exam Origin: New Past NRC Exam? The plant was operating normally at 100% power when the crew manually tripped the plant due to a tube rupture on #2 Steam Generator.

The crew successfully performed EOP 2525, Standard post Trip Actions, and entered EOP 2534, Steam Generator Tube Rupture. The following conditions exist: -SIAS, C!AS, and EBFAS have been

-"A" and "B" RCF)s are running with adequate -Main StElam Liml Radiation Monitor, RM 4299C, are presently reading 1.5 Rlhr and -Condenser Air is aligned to the Unit 2 -The crew is in trle process of lowering both hot leg temperatures to less than or equal to -MSI has been overridden to maintain steam flow to the

-The Unit 2 Stacl<. Gaseous Radiation Monitor, RM 8132B, is in alarm reading 800 cpm and Which of the following statements describes the procedurally directed method used to limit the release of radiation to the environment?

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• I SecurEl all Main Exhaust Fans and direct the Chemist to ensure the 95,000 microcurie/sec release limit will NOT be exceeded. Ensure all flow from the Main Condenser to the Steam Jet Air Ejector Radiation Monitor, RM-5099, has been secured. Override and start the remaining Main Exhaust Fans and ensure all Radwaste Ventilation supply fans are providing adequate flow. D Continue the cooldown and isolate #2 Steam Generator when both hot leg temperatures are less than or equal to 515°F. *Justification I D IS CORRECT; A continuing reading or 1.5 Rlhr on the Main Steam Line Radiation Monitors, 4299C, is indication of fuel failure. highly contaminated steam flowing to the Condenser, the Steam Jet Air Ejectors in service, Condenser Air Removal aligned to Unit Stack, and a Main Exhaust Fan running, the release to the atmosphere will continue.

This alignment must be maintained to cooldown of thE' affected S/G to the condenser.

If the Atmospheric Dumps were used instead, the release to the environment would considerably A is incorrect; Securing all Main Exhaust Fans would no longer allow Condenser Air Removal to remain in operation, resulting in a of vacuum and the inability to maintain steaming to the Main Condenser.

Heat Removal would need to be established through Atmospheric Dumps, which would result in a considerably higher release to the Plausible because the Annunciator Response for an alarm on the Unit 2 Stack Gaseous Radiation Monitor, RM 8132B, in the event of LOCA is to stop all Main Exhaust Fans. Exceeding the release rate limit of 95,000 microcurie/sec is reportable.

The unit is already at least an Alert, Charlie-One due to the B is incorrect; The Steam Jet Air Ejector Radiation Monitor, RM-5099, monitors the air flow from the Main Condenser through Condenser Air System. Stopping this air flow would result in the loss of vacuum and the inability to maintain steaming to Main Condenser.

Heat Removal would need to be established through the Atmospheric Dumps, which would result in a higher release to the Plausible becaLse Condenser Air Removal Flow may be stopped which would limit the release of radiation, but only for a short of time. (l.oss of vacuum leads to the use of the Atmospheric C is incorrect; Starting all Main Exhaust Fans and the Radwaste Ventilation Supply Fans will provide more dilution and lower reading on the Unit 2 Stack Gaseous Radiation Monitor, RM 8132B; however, the actual release of radioactivity will still be the In fact it may bEl slightly higher in that short lived activity will be discharged sooner due to the increase in air Plausible becaLse an examinee may recognize the lower reading on the Unit 2 Stack Gaseous Radiation Monitor, RM 8132B, as decrease in the radioactive release I EOP 2534, Steam Generator Tube Rupture and associated Technical Comments and Question Modification Bob K. -D-3/C Bill M. D-2/C, Angelo -D-3fC; Changed Main Steam Line rad monitor to the Facility 2 side to match affected SG. -NRC KIA System/E/A System 2.3 Radiation Control Page 72 of Printed on 10/28/2009 at 13:20 SROExalIl Q.u.estlons Qnly (No "Parents tt Or ffOf:iIfiD'als U) Question #: ,24 Question ID: 9000016 RO SRO Student Handout? o Lower Order? I-SRO Ques. it :24 Rev. o Selected for Exam Origin: New Past NRC Exam? The plant was operating normally at 100% power when the crew manually tripped the plant due to a tube rupture on #2 Steam Generator.

The crew successfully performed EOP 2525, Standard post Trip Actions, and entered EOP 2534, Steam Generator Tube Rupture. The following conditions exist: -SIAS, C!AS, and EBFAS have been verified.

-"A" and "B" RCF)s are running with adequate NPSH. -Main StElam Liml Radiation Monitor, RM 4299C, are presently reading 1.5 Rlhr and stable. -Condenser Air is aligned to the Unit 2 Stack. -The crew is in trle process of lowering both hot leg temperatures to less than or equal to 515°F. -MSI has been overridden to maintain steam flow to the Condenser.

-The Unit 2 Stacl<. Gaseous Radiation Monitor, RM 8132B, is in alarm reading 800 cpm and rising. Which of the following statements describes the procedurally directed method used to limit the release of radiation to the environment?

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• I

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• • • • • • • • I ****** I I I *** I

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I *** I

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A SecurEl all Main Exhaust Fans and direct the Chemist to ensure the 95,000 microcurie/sec release limit will NOT be exceeded.

B Ensure all flow from the Main Condenser to the Steam Jet Air Ejector Radiation Monitor, RM-5099, has been secured. C Override and start the remaining Main Exhaust Fans and ensure all Radwaste Ventilation supply fans are providing adequate flow. D Continue the cooldown and isolate #2 Steam Generator when both hot leg temperatures are less than or equal to 515°F. *Justification I D IS CORRECT; A continuing reading or 1.5 Rlhr on the Main Steam Line Radiation Monitors, 4299C, is indication of fuel failure. With highly contaminated steam flowing to the Condenser, the Steam Jet Air Ejectors in service, Condenser Air Removal aligned to Unit 2 Stack, and a Main Exhaust Fan running, the release to the atmosphere will continue.

This alignment must be maintained to allow cooldown of thE' affected S/G to the condenser.

If the Atmospheric Dumps were used instead, the release to the environment would be considerably hiuher. A is incorrect; Securing all Main Exhaust Fans would no longer allow Condenser Air Removal to remain in operation, resulting in a loss of vacuum and the inability to maintain steaming to the Main Condenser.

Heat Removal would need to be established through the Atmospheric Dumps, which would result in a considerably higher release to the environment.

Plausible because the Annunciator Response for an alarm on the Unit 2 Stack Gaseous Radiation Monitor, RM 8132B, in the event of a LOCA is to stop all Main Exhaust Fans. Exceeding the release rate limit of 95,000 microcurie/sec is reportable.

The unit is already in at least an Alert, Charlie-One due to the SGTR. B is incorrect; The Steam Jet Air Ejector Radiation Monitor, RM-5099, monitors the air flow from the Main Condenser through the Condenser Air System. Stopping this air flow would result in the loss of vacuum and the inability to maintain steaming to the Main Condenser.

Heat Removal would need to be established through the Atmospheric Dumps, which would result in a considerably higher release to the environment.

Plausible becaLse Condenser Air Removal Flow may be stopped which would limit the release of radiation, but only for a short period of time. (l.oss of vacuum leads to the use of the Atmospheric Dumps.) C is incorrect; Starting all Main Exhaust Fans and the Radwaste Ventilation Supply Fans will provide more dilution and lower the reading on the Unit 2 Stack Gaseous Radiation Monitor, RM 8132B; however, the actual release of radioactivity will still be the same. In fact it may bEl slightly higher in that short lived activity will be discharged sooner due to the increase in air flow. Plausible becaLse an examinee may recognize the lower reading on the Unit 2 Stack Gaseous Radiation Monitor, RM 8132B, as a decrease in the radioactive release rate. References I EOP 2534, Steam Generator Tube Rupture and associated Technical Guide Comments and Question Modification History Bob K. -D-3/C 10K) Bill M. D-2/C, I{ Angelo -D-3fC; Changed Main Steam Line rad monitor to the Facility 2 side to match affected SG. -RLC NRC KIA System/E/A System 2.3 Radiation Control Page 72 of 77 Printed on 10/28/2009 at 13:20 Question #: 24 Question ID: 9000016 RO SRO Student Handout? Lower Order? \-SRO Ques. # 24 Selected for Exam Origin: New Past NRC Exam?Rev. NRC KIA System 2.3 Radiation Control Number 2.:1.14 RO 3.4 SRO 3.8 CFR Link (CFR: 41.12 143.4 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or Millstone Unit 2 Steam Thbe Rupture EOP 2534 Revision 025-01 Page 11 of67 I INSTRUCnONS Align Condenser Air Removal to Unit 2 Stae.k CONTINGENCY ACTIONS "7. EBE>\S has initiated AN.llthe conde nser is available, the condenser air removal "ystem to Unit 2 slack: H. ENSURE condenser air removal fan, MF-SSA or MF-55B. is mIming. b. IF. condenser air removal fan MF-55A is operating, ENSURE makeup damper, Ell-171, is open. c. OPEN EB-57, ","(mdenser air removal 10 Unit 2 stack d. ENSURE AC-ll, Purge exhaust filter outlet damper is cklsed. e. OPEN AC-59, Outside air makeup damper. f. START ONE main exhaust fan. g. ENSURE HV-llS, Radwaste exhaust damper is closed. h. STARr F-20, Fuel handling area supply fan. L ENSURE HV-I.73, Exhaust mod die,charge damper is in "MOD" position.

j. PLACE AC-S9, Outside air makeup damper to "MID" position.

STOP THINK REVIEW Page 73 of 77 Printed on 10/28/2009 at 13:20 Question #: 24 \-SRO Ques. # 24 NRC KIA Generic Number 2.:1.14 Question ID: 9000016 RO SRO Student Handout? Lower Order? Rev. o Selected for Exam Origin: New Past NRC Exam? System 2.3 Radiation Control RO 3.4 SRO 3.8 CFR Link (CFR: 41.12 143.4 145.10) Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Millstone Unit 2 Steam Thbe Rupture I INSTRUCnONS Align Condenser Air Removal to Unit 2 Stae.k "7. EBE>\S has initiated AN.llthe conde nser is available, the condenser air removal "ystem to Unit 2 slack: H. ENSURE condenser air removal fan, MF-SSA or MF-55B. is mIming. b. IF. condenser air removal fan MF -55A is operating, ENSURE makeup damper, Ell -171, is open. c. OPEN EB-57, ","(mdenser air removal 10 Unit 2 stack d. ENSURE AC-ll, Purge exhaust filter outlet damper is cklsed. e. OPEN AC-59, Outside air makeup damper. f. START ONE main exhaust fan. g. ENSURE HV-llS, Radwaste exhaust damper is closed. h. STARr F-20, Fuel handling area supply fan. L ENSURE HV-I.73, Exhaust mod die,charge damper is in "MOD" position.

j. PLACE AC-S9, Outside air makeup damper to "MID" position.

STOP THINK EOP 2534 Revision 025-01 Page 11 of67 CONTINGENCY ACTIONS REVIEW Page 73 of 77 Printed on 10/28/2009 at 13:20 SROExam Questions Only (No!'Parents" Or Question #: 24 ORO SRO o Student Handout? o Lower Order?Question ID: 9000016 II-SRO Que,. I/-24 Rev. 0 Selected for Exam Origin: New o Past NRC Exam? Millstone Unit 2 Emergency Operating Procedure TI9chnicai Guide EOP 2534, Generator Tube Page 25 of 126 Step Number 7 Align Condenser Air Removal to Unit 2 Stack The intent of this, step is maintain condenser vacuum in order to use the steam dumps for an RCS cOCildown..and to minimize the possibility of airborne activity backdrafting from the Millstone stack tl) Unit 2 after EBFAS has actuated.

INSTRUCTIONS EBFAS, the Unit 2 Enclosure Building Ventilation is aligned to the Millstone stack, and isolates the Condenser Air Removal System from the Millstone stack. The lineup that exists does not provide for enough flow to maintain condenser vacuum. The isolation damper has a history of leakage which results in the Millstone stack gaseous releases backdrafting intol Unit 2 if the condenser air removal fans are not operating.

This step is designed to minimize the possibility of this occurring.

The operator is directed to ensure a condenser air removal fan is operating and to align a flow path to the Unit 2 Stack for the condenser air removal system following an EBFAS actuation.

The cJperator must also ensure EB-171 is open if condenser air removal fan "An is operating.

This is due to the difference in size between the A and B fans. Additional ventilation fans are started and dampers are opened to balance the flow. CONTINGENCY ACTIONS None JUSTIFICATION FOR DEVIATION MP2 adds this plant specific step to ensure that a vacuum is maintained in the main condenser, if the! main condenser is available.

The use of the steam dumps is the preferred method to perform a plant cooldown.

The steam dumps help maintain secondary inventory and during a SGTR will minimize offsite exposure.

REFERENCES Safety Function Requirements Manual, Section 2.8.2.6.3 "Radiological Consequences of a Steam Generator Tube Rupture" (2/2000) AR 99014319 Resolve Issue With Starting Main Exhaust Fans in EOP 2534 Page 74 of 77 Printed on 10/28/2009 at 13:20 SROExam Questions Only (No!'Parents" Or "Originals")

Question #: 24 Question ID: 9000016 II-SRO Que,. I/-24 Rev. 0 ORO SRO Selected for Exam o Student Handout? Origin: New o Lower Order? o Past NRC Exam? Millstone Unit 2 Emergency Operating Procedure TI9chnicai Guide EOP 2534, Steam Generator Tube Rupture Page 25 of 126 Step Number 7 Align Condenser Air Removal to Unit 2 Stack The intent of this, step is maintain condenser vacuum in order to use the steam dumps for an RCS cOCildown

.. and to minimize the possibility of airborne activity backdrafting from the Millstone stack tl) Unit 2 after EBFAS has actuated.

INSTRUCTIONS EBFAS, the Unit 2 Enclosure Building Ventilation is aligned to the Millstone stack, and isolates the Condenser Air Removal System from the Millstone stack. The lineup that exists does not provide for enough flow to maintain condenser vacuum. The isolation damper has a history of leakage which results in the Millstone stack gaseous releases backdrafting intol Unit 2 if the condenser air removal fans are not operating.

This step is designed to minimize the possibility of this occurring.

The operator is directed to ensure a condenser air removal fan is operating and to align a flow path to the Unit 2 Stack for the condenser air removal system following an EBFAS actuation.

The cJperator must also ensure EB-171 is open if condenser air removal fan "An is operating.

This is due to the difference in size between the A and B fans. Additional ventilation fans are started and dampers are opened to balance the flow. CONTINGENCY ACTIONS None JUSTIFICATION FOR DEVIATION MP2 adds this plant specific step to ensure that a vacuum is maintained in the main condenser, if the! main condenser is available.

The use of the steam dumps is the preferred method to perform a plant cooldown.

The steam dumps help maintain secondary inventory and during a SGTR will minimize offsite exposure.

REFERENCES Safety Function Requirements Manual, Section 2.8.2.6.3 "Radiological Consequences of a Steam Generator Tube Rupture" (2/2000) AR 99014319 Resolve Issue With Starting Main Exhaust Fans in EOP 2534 Page 74 of 77 Printed on 10/28/2009 at 13:20 Question #: 25 Student Handout? Lower Order?Question ID: 9800061 RO SRO l-SRO Que!,. 1/ 25 Rev. o Selected for Exam Origin: Mod Past NRC Exam? The plant is stable at 100% power with the Turbine Driven Auxiliary Feedwater Pump out of service for planned maintenance.

Then, the plant trips due to a loss of off site power (state wide blackout).

resulting in the following conditions:

-The "A" Main Steam header ruptures in containment on the -Busses 24B and 240 are de-energized due to a bus fault on -Facility One SIAS. CIAS. EBFAS, MSI and CSAS have all fully

-All feedwater has been secured to the #1 Steam Generator -All other plant systems and components that have power are functioning as The crew is evaluating numerous alarms and indications caused by the power loss and subsequent ESD. Which of the following alarm indications will require actions to be taken in EOP 2525? C-05 alarms indicating a Excess Steam Demand on the #1 SG. and CoOS alarm indicating VR-21 is energized. C-02/3 alarms indicating RCS temperatures are abnormally low and dropping, and both Boric Acid Pumps are de-energized. C-05 alarms indicating both SG levels abnormally low. and only one Aux. Feedwater pump is feeding just thl:l #2 SG. D C-01 alarms indicating CTMT Spray has actuated, and C-01 indicating only two CAR fans and one CS pump are operating.

iJustification A -CORRECT; All alarms and indications mentioned in the four choices are expected for the given event, a loss of the RSST and with a subsequent ESD on the "A" Main Steam header. VR-21 is deenergized based on the loss of 240. This will prevent the Atmospheric Dump Valve (ADV) from being operated from the control room (after about 10 minutes) and the ESD in the building prevents local operation.

Therefore, immediate action is required to get an operator to C21 (Remote Shutdown Panel) control RCS temperature when the affected SG boils dry (thus preventing B -WRONG; This gives indication of an excessive cooldown of the RCS with a potential problem with boric acid injection.

the other facility of power is available to allow automatic alignment of a boric acid source to the remaining charging pump, which sufficient (although not optimum) to meet "reactivity Plausible; Procedure steps will give guidance to align additional boron injection, but this is above the required C WRONG; Feedwater will feed enough to recover level in the unaffected steam generator, by Plausible; The SG levels are abnormal compared to an uncomplicated trip due to volume shrinkage from the ESD. Procedures guidance to start additional AFW pumps, as necessary (none available in this case) to return SG levels to the operating band However, the only available option with the given conditions is "Once-Through-Cooling", which can not be done in o WRONG; One facility of CTMT Cooling and Pressure Control is certainly NOT optimum during and ESD, but it is designed to be sufficient to maintain CTIIIIT Integrity, provided all feed is secured to the affected SG in the required time frame.

gives guidance to repower and start all available ESAS components.

However, this is not reguired to prevent ",. a design limit. References I EOP 2525, Step 17 and AOP 2501, Pg. 10, Diagnostic Chart Comments and Question Modification History Bob K. D-4/C (OK -good question).

Bill M. D-3/C, K Mike C. -Modify stem to eliminate possibility of improving AFW condition in 2525. -Done -RLC Angelo -D-3/C; Added til) stem that all feedwater is secured to #1 SG.* RLC NRC KIA System/E/A System 2.4 Emergency Procedure IPlan NRC KIA Generic System 2.4 Emergency Procedures IPlan Number RO 4.1 SRO 4.3 CFR Link (CFR: 41.10 143.5145.3/45.12) Ability to priori!jze and interpret the significance of each annunciator or alarm. Page 75 of 77 Printed on 10/28/2009 at 13:20 Question #: 25 Question ID: 9800061 RO SRO Student Handout? Lower Order? l-SRO Que!,. 1/ 25 Rev. o Selected for Exam Origin: Mod Past NRC Exam? The plant is stable at 100% power with the Turbine Driven Auxiliary Feedwater Pump out of service for planned maintenance.

Then, the plant trips due to a loss of off site power (state wide blackout).

resulting in the following conditions:

-The "A" Main Steam header ruptures in containment on the trip. -Busses 24B and 240 are de-energized due to a bus fault on 240. -Facility One SIAS. CIAS. EBFAS, MSI and CSAS have all fully actuated.

-All feedwater has been secured to the #1 Steam Generator (SG). -All other plant systems and components that have power are functioning as designed.

The crew is evaluating numerous alarms and indications caused by the power loss and subsequent ESD. Which of the following alarm indications will require actions to be taken in EOP 2525? A C-05 alarms indicating a Excess Steam Demand on the #1 SG. and CoOS alarm indicating VR-21 is energized.

B C-02/3 alarms indicating RCS temperatures are abnormally low and dropping, and both Boric Acid Pumps are de-energized.

C C-05 alarms indicating both SG levels abnormally low. and only one Aux. Feedwater pump is feeding just thl:l #2 SG. D C-01 alarms indicating CTMT Spray has actuated, and C-01 indicating only two CAR fans and one CS pump are operating.

iJustification J A -CORRECT; All alarms and indications mentioned in the four choices are expected for the given event, a loss of the RSST and 240, with a subsequent ESD on the "A" Main Steam header. VR-21 is deenergized based on the loss of 240. This will prevent the "B" Atmospheric Dump Valve (ADV) from being operated from the control room (after about 10 minutes) and the ESD in the Enclosure building prevents local operation.

Therefore, immediate action is required to get an operator to C21 (Remote Shutdown Panel) to control RCS temperature when the affected SG boils dry (thus preventing PTS). B -WRONG; This gives indication of an excessive cooldown of the RCS with a potential problem with boric acid injection.

However, the other facility of power is available to allow automatic alignment of a boric acid source to the remaining charging pump, which is sufficient (although not optimum) to meet "reactivity control".

Plausible; Procedure steps will give guidance to align additional boron injection, but this is above the required amount. C WRONG; Feedwater will feed enough to recover level in the unaffected steam generator, by deSign. Plausible; The SG levels are abnormal compared to an uncomplicated trip due to volume shrinkage from the ESD. Procedures give guidance to start additional AFW pumps, as necessary (none available in this case) to return SG levels to the operating band (40-70%).

However, the only available option with the given conditions is "Once-Through-Cooling", which can not be done in EOP-2525.

o WRONG; One facility of CTMT Cooling and Pressure Control is certainly NOT optimum during and ESD, but it is designed to be sufficient to maintain CTIIIIT Integrity, provided all feed is secured to the affected SG in the required time frame.

gives guidance to repower and start all available ESAS components.

However, this is not reguired to prevent ",. a design limit. References I EOP 2525, Step 17 and AOP 2501, Pg. 10, Diagnostic Chart Comments and Question Modification History Bob K. D-4/C (OK -good question).

Bill M. D-3/C, K Mike C. -Modify stem to eliminate possibility of improving AFW condition in 2525. -Done -RLC Angelo -D-3/C; Added til) stem that all feedwater is secured to #1 SG .* RLC NRC KIA System/E/A System 2.4 Emergency Procedure IPlan NRC KIA Generic System 2.4 Emergency Procedures IPlan Number RO 4.1 SRO 4.3 CFR Link (CFR: 41.10 143.5145.3/45.12)

Ability to priori!jze and interpret the significance of each annunciator or alarm. Page 75 of 77 Printed on 10/28/2009 at 13:20 II Question #: 25 Student Handout? Lower Order? Question ID: 9800061 RO SRO II-SRO Ques. /I 25 Rev. 0 Selected for Exam Origin: Mod Past NRC Exam? r---Millstone Unit 2 Standard Post Trip Actions EOP 2525 Revision 023 Page 24 of26 INSTRUCTIONS Subsequent Actiuns (cuntinued)

_17. II: steam generator pressure is less than 572 psia AND the most affected steam generator has boiled dry, as indicated by CET temperature rising, OPERATE the ADV for the least afft":Cted steam generator to stabilize CET temperature.

CONTINGENCY ACTIONS Page 76 of 77 Printed on 10/28/2009 at 13:20 Question #: 25 Question ID: 9800061 RO SRO II-SRO Ques. /I 25 Rev. 0 Selected for Exam r---Millstone Unit 2 II Standard Post Trip Actions EOP 2525 Student Handout? Lower Order? Origin: Mod Past NRC Exam? Revision 023 Page 24 of26 INSTRUCTIONS CONTINGENCY ACTIONS Subsequent Actiuns (cuntinued)

_17. II: steam generator pressure is less than 572 psia AND the most affected steam generator has boiled dry, as indicated by CET temperature rising, OPERATE the ADV for the least afft":Cted steam generator to stabilize CET temperature.

Page 76 of 77 Printed on 10/28/2009 at 13:20 Question #: 25 Lower Order? Question ID: 9800061 RO SRO Student Handout? 25 Rev. o Selected for Exam Origin: Mod Past NRC Exam? Millstone Unit 2 AOP 2501 Revision 001-02 Diagnostic for Loss of Page 10 of 10 Elcctrical Powcr Attachment 1 Lost Control Power (Sheet I of I ) Loss ofVR-ll Loss ofVR-21 Loss of VA-tO Loss ofVA-20 JAdintioEl (dlf'lItK;Ut'd)

Len S1(II.' C-Ol & CEAPDS Right side C-Ol & Core i\Iimlc nA" safety (nannel'>

lSi: "A"RPS "B" safety channels lSi: "S"RPS PA S,"t.m lost RRS Ba ttny Bit: la", tor -. 10 min. "fttr YR-lllo... C -02 LddoWII bolato, (High Teolp **ns"r failure) RRS battery BlV ,,1!il!Ible PZR at program ..!:!glll:ry lkIlleted

.. Chg Pp suctIOn to RWST PZR Ld control in Ch PZR **!polnt fait. to zero fZB I rl t!lllll.:!ll ill !::b "x,. -Maximum Cbazgillg lieN" -Minimum LetdowIl n"w Minimum Charging flow -Mllximum Letdown flow -Maximum Charging flow -Minimum Letdown itow C-03 Lose AIll'ZR Hute.. (Select in "X" Of "X+Y") Los. Alll'ZR H**ters (Select 1n "Y" or ,*x-Y") PZR Backup Hoator, All fOllr hank. ""a,'ailable

.. .. Lo.., All PZR, Heaters {Select ill "X" or '*X*'[") Lo.eAllPZR Hulen (Select In "Y" Of C-03 Btl" (.;"8"gl"'g l'ump' Both 3llII! ."towalically (Dead c""ll'ol circuit) Btl! Charginlll'ump'

'!pllt () PZR Spray AUf 0 Conn... l available, [Manual sptay nol Wi mC-IOOF only] mC-IOOF dead (MJ" .pray) {Auto cooIrol wI mC-lOOE if Pte... control in Ch. "Xff] C-03 Loop Itfmp 10 RRS 10.1 (RRS alii" bypane.) Loop 1 t...mp to RRS 10.1 (RRS Buto hyp...",) L.."p I temp inplll !<l K:C, LTO?, RRS all lost (RRS aul0 Loop 2 **mp input to lCC, LTOP, RRS an 10s1 (RRS auro b)'P.'oro)

C-05 TIC 4165 Dead CTa,'!! cnll'l) PIC 4223 Dead (ill ADV) PIC-4216 OfI:ad ("'A . CDV) PIC-4124 Dead (#2 ADV) C--05 Cond"n,,,,.

51.am Dumps All fOlll fail"d. dosed (Op= Wi Quick Open only) C..ud...n.u SI."m Dump. PIC-4216

{Colllll w! TIC-416; @ COj Of P1C-4216@FoxborolA]

I V b Qul<k Op.n "" program ** lIIIlll:tylkll leled ** No QO ro any dump valv" C-05 #1 ADY (.I"IC-4223)

""."il. Manual anI:)I: (Input p"'''

ftOZffi whm pow.. lost) lmw:y BiU 112 AD" IPlC-4224) avoil (Input press",,!

frozen .t u.!. wben pow'" ""a, lo.t), *"

deilleted

• • #2 AD\, faib clo;ed [Control from ClllCHl] #1 ADY ",mote control 1001 (C-05 & C-21 "OIItrl. dead) (Lac&!-Manual only] III AD\, remote coulroll"", (C-ll5, -21, -10 colltrl dead) {Local-Manual only J C-05 MA" SGFl' Inseli dark. troIs still work (Inwcatioll Oil PPC) UB" SGFP Insert dark, con troh still work (Indkation on PPC) ill FR" :\I:oi.. -Fail ** i, [Local-Man

...! only] (Clos** on power ",.tore) 112 FR'*

Fad". i. [L"".. l-MlUlUlIlouIy] (Clo..,. on power C-05 SGFP :Mini ... u... 1"10"-Cutd dead, valve fail clo.ed "B" SGFP M:lm...nm 1:(0'" Cnttl dead, yah" fail dosed #1 FR\, BYP"ssllIllS do,.d (All centrol lo.t) liZ FRV Bypa** fails c1o.ed (All contml i. 1o") C-05 (B*n Rad Monitor deod) Blowdown isolates (SJAE Rad Monitor dead) In Au>: FRv !ill!S open [Local-Manual only) 112 Au>: FR'-fail. open [Local-M>mualoul}']

STOP THINK ACT REVIEW Page 77 of 77 Printed on 1012812009 at 13:20 Question #: 25 Question ID: 9800061 RO SRO Student Handout? Lower Order? 25 Rev. o Selected for Exam Origin: Mod Past NRC Exam? Millstone Unit 2 Diagnostic for Loss of Elcctrical Powcr AOP 2501 Revision 001-02 Page 10 of 10 1--------Loss ofVR-ll JAdintioEl Len S1(II.' C-Ol & (dlf'lItK;Ut'd)

CEAPDS PA S,"t.m lost C -02 LddoWII bolato, (High Teolp ** ns"r failure) C-03 Lose AIll'ZR Hute .. (Select in "X" Of "X+Y") C-03 Btl" (.;"8"gl"'g l'ump' Both 3llII! ."towalically (Dead c""ll'ol circuit) C-03 Loop Itfmp 10 RRS 10.1 (RRS alii" bypane.) C-05 C--05 C-05 #1 ADY (.I"IC-4223)

""."il. Manual anI:)I: (Input p"'''

ftOZffi whm pow .. lost) C-05 MA" SGFl' Inseli dark. C<>n-troIs still work (Inwcatioll Oil PPC) C-05 SGFP :Mini ... u ... 1"10"-Cutd dead, valve fail clo.ed C-05 (B*n Rad Monitor deod) Attachment 1 Lost Control Power (Sheet I of I ) Loss ofVR-21 Loss of VA-tO Right side C-Ol & nA" safety (nannel'>

lSi: Core i\Iimlc "A"RPS RRS Ba ttny Bit: la", tor -. 10 min. "fttr YR-lllo ... RRS battery BlV ,,1!il!Ible Chg Pp suctIOn to PZR at program RWST .. !:!glll:ry lkIlleted

.. PZR ** !polnt fait. to zero -fZB I rl t!lllll.:!ll ill !::b "x,. Minimum Charging flow -Maximum Charging flow -Mllximum Letdown flow -Minimum Letdown itow Los. Alll'ZR H ** ters Lo.., All PZR, Heaters (Select 1n "Y" or ,*x -Y") {Select ill "X" or '*X* '[") PZR Backup Hoator, All fOllr hank. ""a,'ailable

.. .. IHIC-Btl! Charginlll'ump' PZR Spray AUf 0 Conn ... l un-

'!pllt () available, [Manual sptay COll-nol Wi mC-IOOF only] Loop 1 t ... mp to RRS 10.1 L .. "p I temp inplll !<l K:C, (RRS Buto hyp ... ",) LTO?, RRS all lost (RRS aul0 TIC 4165 Dead CTa,'!! cnll'l) PIC 4223 Dead (ill ADV) Cond"n,,,,.

51.am Dumps All fOlll fail"d. dosed (Op= Wi Quick Open only) I V b Qul<k Op.n "" program ** lIIIlll:tylkll leled ** No QO ro any dump valv" lmw:y BiU

1. 1 ADY ",mote control 1001 112 AD" IPlC-4224) avoil (C-05 & C-21 "OIItrl. dead) (Input press",,!

frozen .t val-(Lac&!-Manual only] u.!. wben pow'" ""a, lo.t), *"

deilleted

• • #2 AD\, faib clo;ed [Control from ClllCHl] UB" SGFP Insert dark, con ill FR" :\I:oi .. -Fail ** i, troh still work [Local-Man

... ! only] (Indkation on PPC) (Clos ** on power ",.tore) "B" SGFP M:lm ... nm 1:(0'" #1 FR\, BYP"ssllIllS do,.d Cnttl dead, yah" fail dosed (All centrol lo.t) Blowdown isolates In Au>: FRv !ill!S open (SJAE Rad Monitor dead) [Local-Manual only) STOP THINK ACT REVIEW Page 77 of 77 Loss ofVA-20 "B" safety channels lSi: "S"RPS PZR Ld control in Ch

-Maximum Cbazgillg lieN" -Minimum LetdowIl n"w Lo.eAllPZR Hulen (Select In "Y" Of mC-IOOF dead (MJ" .pray) {Auto cooIrol wI mC-lOOE if Pte. .. control in Ch. "Xff] Loop 2 ** mp input to lCC, LTOP, RRS an 10s1 (RRS auro b)'P.'oro)

PIC-4216 OfI:ad ("'A . CDV) PIC-4124 Dead (#2 ADV) C .. ud ... n.u SI."m Dump. PIC-4216

{Colllll w! TIC-416; @ COj Of P1C-4216@FoxborolA]

III AD\, remote coulroll"", (C-ll5, -21, -10 colltrl dead) {Local-Manual only J 112 FR'*

Fad". i. [L"" .. l-MlUlUlIlouIy] (Clo..,. on power liZ FRV Bypa ** fails c1o.ed (All contml i. 1o") 112 Au>: FR'-fail. open [Local-M>mualoul}']

Printed on 1012812009 at 13:20 SRO EXAM Answer Key Order Based Sequenc:e

1. Stem ID -1# 1 9000018 -0 D Direct the crew to commence a controlled cooldown and depressurization.

1. 2 9000019 -0 The HPSI Pumps are showing signs of cavitation due to Containment Sump clogging.

Secure both Containment Spray Pumps. # 3 I 9000003 -0 A The "A" RCP Excess Flow Check Valve has seated. Manually trip the reactor and turbine, then stop the "A" RCP. # 4 9000004-0 Dispatch a PEO to manually operate the "B" Auxiliary Feedwater Regulating Valve,2-FW-43B.

This will prevent excessive auxiliary feedwater from overfilling the affected SG. # 5 9082581-0 1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-657, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rate. # 6 3100002 -2 When pressure lowers to less than 80 psig. The loss of many important controls, such as Feedwater, could degrade plant conditions at the time of the trip; therefore, the reactor must be tripped when control of important systems could become challenged.

1. 7 90000:ZO -0 CEA #1 Upper Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered.

r# 9000005-0 Feed the affected SG to maintain level 40 to 45%. This will cover the SG D tube to allow for Iodine scrubbing and still allow adequate volume to accept water from the tube rupture. # 9 J 9000006 -0 C Alert, Charlie-One Page 1 of Printed on 10/29/2009 at 14:55 SRO EXAM Answer Key Order Is Based 011 Sequenc:e

1. Stem ID -Rev 1# 1 9000018 -0 D Direct the crew to commence a controlled cooldown and depressurization.

1. 2 9000019 -0 B The HPSI Pumps are showing signs of cavitation due to Containment Sump clogging.

Secure both Containment Spray Pumps. # 3 I 9000003 -0 A The "A" RCP Excess Flow Check Valve has seated. Manually trip the reactor and turbine, then stop the "A" RCP. # 4 9000004-0 C Dispatch a PEO to manually operate the "B" Auxiliary Feedwater Regulating Valve,2-FW-43B.

This will prevent excessive auxiliary feedwater from overfilling the affected SG. # 5 9082581-0 B 1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-657, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rate. # 6 3100002 -2 C When pressure lowers to less than 80 psig. The loss of many important controls, such as Feedwater, could degrade plant conditions at the time of the trip; therefore, the reactor must be tripped when control of important systems could become challenged.

1. 7 90000:ZO -0 D CEA #1 Upper Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered.

r# 8 9000005-0 D Feed the affected SG to maintain level 40 to 45%. This will cover the SG tube to allow for Iodine scrubbing and still allow adequate volume to accept water from the tube rupture. # 9 J 9000006 -0 C Alert, Charlie-One Page 1 of 3 Printed on 10/29/2009 at 14:55 SRO EXAM Answer Key Order 15 Based on Sequence # StemID -Rev # 10 9000007 -0 B With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 60 curies/gram once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 micro-curies/gram limit. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and it is expected that normal coolant Iodine concentration would be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. # 11 9000008 -0 A Immediately secure the "A" RCP, raise RCS pressure, then start "C" and "D" RCPs. # 12 96000'16 -0 Per the ARP for C-38; shift pressurizer heater control to channel "Y" and restore pressurizer heaters, to ensure adequate margin from DNB is maintained.

1. 13 537:30 -4 C Trip the reactor and Go to EOP 2525, "Standard Post Trip Actions".

1. 14 9000021 -0 A All fuel movement in containment must remain secured. # 15 90000'10 -0 The "A" Service Water Pump is considered OPERABLE.

If the "A" Service Water Strainer Flush Valve, 2-SW-90A, exceeds the Maximum "Normal Limit" on an immediate retest, then declare the "A" Service Water Pump inoperable.

1. 16 90000'12 -0 Ensure a control room operator is specifically assigned to close the Containment Purge Valves within 30 minutes of an event, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment.

1. 17 9079010 -0 B Lower the setpoint on the "A" steam dump and log into the DNB Technical Specification.

1. 18 90000'11 -0 The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.3.3.7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable and a fire watch must be established.

1. 19 90000'13 -0 Use Reactivity Plan RE-G-11 to reduce power to the appropriate level, secure the Feed Pump, then raise power to the appropriate level using OP 2204, Load Changes, and a new Reactivity Plan. # 20 90000'14 -0 D If an underload occurs prematurely, raise the fuel assembly, pull the mast detent pin, rotate slightly, and reinsert the assembly.

Page 2 of Printed on 10/29/2009 at 14:55 SRO EXAM Answer Key Order 15 Based on Sequence # StemID -Rev # 10 9000007 -0 B With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 60 micro-curies/gram once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 micro-curies/gram limit. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and it is expected that normal coolant Iodine concentration would be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. # 11 9000008 -0 A Immediately secure the "A" RCP, raise RCS pressure, then start "C" and "D" RCPs. # 12 96000'16 -0 A Per the ARP for C-38; shift pressurizer heater control to channel "Y" and restore pressurizer heaters, to ensure adequate margin from DNB is maintained.

1. 13 537:30 -4 C Trip the reactor and Go to EOP 2525, "Standard Post Trip Actions".

1. 14 9000021 -0 A All fuel movement in containment must remain secured. # 15 90000'10 -0 D The "A" Service Water Pump is considered OPERABLE.

If the "A" Service Water Strainer Flush Valve, 2-SW-90A, exceeds the Maximum "Normal Limit" on an immediate retest, then declare the "A" Service Water Pump inoperable.

1. 16 90000'12 -0 C Ensure a control room operator is specifically assigned to close the Containment Purge Valves within 30 minutes of an event, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment.

1. 17 9079010 -0 B Lower the setpoint on the "A" steam dump and log into the DNB Technical Specification.

1. 18 90000'11 -0 B The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.3.3.7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable and a fire watch must be established.

1. 19 90000'13 -0 C Use Reactivity Plan RE-G-11 to reduce power to the appropriate level, secure the Feed Pump, then raise power to the appropriate level using OP 2204, Load Changes, and a new Reactivity Plan. # 20 90000'14 -0 D If an underload occurs prematurely, raise the fuel assembly, pull the mast detent pin, rotate slightly, and reinsert the assembly.

Page 2 of 3 Printed on 10/29/2009 at 14:55 SRO EXAM Answer Key Order Is Based on Sequence # Stem ID Rev # 21 9000024 -0 Terminate surveillance testing of any safety related pumps and motors and secure them, if possible.

1. 22 90000'15 -0 Final Safety Analysis Report, Chapter 15, License Renewal, Aging Management Programs # 23 8000044-2 Re-mix the tank for the required period of time, then resample the tank and generate a new discharge permit based on the new sample. # 24 I 90000'16 -0 Continue the cooldown and isolate #2 Steam Generator when both hot leg temperatures are less than or equal to 515°F. # 25 9800061 -0 C-05 alarms indicating a Excess Steam Demand on the #1 SG, and C-08 alarm indicating VR-21 is de-energized.

Page 3 of Printed on 10/29/2009 at 14:55 Order Is Based on SRO EXAM Answer Key Sequence # Stem ID Rev # 21 # 22 # 23 # 24 I # 25 9000024 -0 B Terminate surveillance testing of any safety related pumps and motors and secure them, if possible.

90000'15 -0 A Final Safety Analysis Report, Chapter 15, License Renewal, Aging Management Programs 8000044-2 B Re-mix the tank for the required period of time, then resample the tank and generate a new discharge permit based on the new sample. 90000'16 -0 D Continue the cooldown and isolate #2 Steam Generator when both hot leg temperatures are less than or equal to 515°F. 9800061 -0 A C-05 alarms indicating a Excess Steam Demand on the #1 SG, and C-08 alarm indicating VR-21 is de-energized.

Page 3 of 3 Printed on 10/29/2009 at 14:55 Senior Reactor Operator NRC License Upgrade Question # ....!.J The plant automatically tripped on High Pressurizer Pressure due to an inadvertent closure of the Main Turbine Control Valves. During the performance of EOP 2525, Standard Post Trip Actions, the crew reported that Bus 24D is deenergized due to a fault and that Power Operated Relief Valve (PORV), RC-404, is stuck open. All other equipment operated as designed.

Upon entry into EOP 2532, Loss of Coolant Accident, the following conditions exist: Containment pressure is 4.5 psia and slowly rising. -Reactor vessel is 43% and slowly going down CET temperatures are 578°F and stable RCS pressure is 1310 psia and stable -Pressurizer level is -Steam gEmerator levels are both 41 % and going up Which of the following actions must the Unit Supervisor/Shift Manager perform to preserve a Safety A Direct the Technical Support Center to develop a plan to restore RWST level. B Direct the Balance of Plant Operator to align Condenser Air Removal to the Unit 2 Stack. C Direct the Reactor Operator to place the SI/CS Pump Miniflow switches in "OPERATE".

D Direct the crew to commence a controlled cooldown and depressurization.

Page 1 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # ....!.J The plant automatically tripped on High Pressurizer Pressure due to an inadvertent closure of the Main Turbine Control Valves. During the performance of EOP 2525, Standard Post Trip Actions, the crew reported that Bus 24D is deenergized due to a fault and that Power Operated Relief Valve (PORV), RC-404, is stuck open. All other equipment operated as designed.

Upon entry into EOP 2532, Loss of Coolant Accident, the following conditions exist: Containment pressure is 4.5 psia and slowly rising. -Reactor vessel is 43% and slowly going down CET temperatures are 578°F and stable RCS pressure is 1310 psia and stable -Pressurizer level is 100%. -Steam gEmerator levels are both 41 % and going up slowly. Which of the following actions must the Unit Supervisor/Shift Manager perform to preserve a Safety Function?

A Direct the Technical Support Center to develop a plan to restore RWST level. B Direct the Balance of Plant Operator to align Condenser Air Removal to the Unit 2 Stack. C Direct the Reactor Operator to place the SI/CS Pump Miniflow switches in "OPERATE".

D Direct the crew to commence a controlled cooldown and depressurization.

Page 1 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Question # 21 The plant t!"ipped from 100% power due to a Large Break LOCA. The crew successfully completed all actions of EOP 2525, Standard Post Trip Actions, and are presently performing EOP 2532, Loss of Coolant Accident.

The following conditions exist approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the trip:

• SRAS actuated approximately 15 minutes ago.

• Contail1ment pressure is 5 psig and slowly lowering.

• RCS p"essure is 360 psia and slowly lowering.

• CET temperatures indicate 434°F and slowly lowering.

• HPSI Pump current and flow are fluctuating.

Which of the following describes the cause of the HPSI Pump current and flow fluctuations, and the initial action that must be directed? Hot water in the Containment Sump is flashing to steam in the HPSI Pump Start at least 2 CAR Fans in Fast The HPSI Pumps are showing signs of cavitation due to Containment Sump Secure both Containment Spray Boron is beginning to plate out in the core causing alternately high and low HPSI Establish Hot Leg and Cold Leg Total Safety Injection flow is higher than necessary for the present Throttle the HPSI Injection valves as Page 2 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Exam Question # 21 The plant t!"ipped from 100% power due to a Large Break LOCA. The crew successfully completed all actions of EOP 2525, Standard Post Trip Actions, and are presently performing EOP 2532, Loss of Coolant Accident.

The following conditions exist approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the trip:

• SRAS actuated approximately 15 minutes ago.

• Contail1ment pressure is 5 psig and slowly lowering.

• RCS p"essure is 360 psia and slowly lowering.

• CET temperatures indicate 434°F and slowly lowering.

• HPSI Pump current and flow are fluctuating.

Which of the following describes the cause of the HPSI Pump current and flow fluctuations, and the initial action that must be directed?

A Hot water in the Containment Sump is flashing to steam in the HPSI Pump suctions.

Start at least 2 CAR Fans in Fast speed. B The HPSI Pumps are showing signs of cavitation due to Containment Sump clogging.

Secure both Containment Spray Pumps. C Boron is beginning to plate out in the core causing alternately high and low HPSI flow. Establish Hot Leg and Cold Leg Injection.

D Total Safety Injection flow is higher than necessary for the present conditions.

Throttle the HPSI Injection valves as needed. Page 2 of25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Question # -=-J While operating at 100% power, the RCP A UPPER SEAL PRES HI annunciator alarms. While referring to the appropriate Annunciator Response Procedure, the RCP A BLEED-OFF FLOW HI annunciator alarms. Within a minute, the RCP A BLEED-OFF FLOW HI annunciator clears and the RCP A BLEED-OFF FLOW LO annunciator alarms and remains lit. Numerous annunciators associated with "A" RCP seals also alarm. Which of the following describes the reason for this sequence of annunciators and the direction that must be given? A The "A" RCP Excess Flow Check Valve has seated. Manually trip the reactor and turbine, then stop the "A" RCP. B The "A" RCP Middle Seal has failed. degradation or failures.

Evaluate the condition of the other seals to confirm no other C The Bleedoff Pressure Controller, PIC-215, has malfunctioned.

Using the Foxboro Controller, restore "A" RCP Bleedoff pressure and flow to the normal band. D The RCP Bleedoff Relief Valve has inadvertently opened. Evaluate "Au RCP Seal pressures to determine whether or not the "A" RCP may remain in operation.

Page 3 of25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # -=-J While operating at 100% power, the RCP A UPPER SEAL PRES HI annunciator alarms. While referring to the appropriate Annunciator Response Procedure, the RCP A BLEED-OFF FLOW HI annunciator alarms. Within a minute, the RCP A BLEED-OFF FLOW HI annunciator clears and the RCP A BLEED-OFF FLOW LO annunciator alarms and remains lit. Numerous annunciators associated with "A" RCP seals also alarm. Which of the following describes the reason for this sequence of annunciators and the direction that must be given? A The "A" RCP Excess Flow Check Valve has seated. Manually trip the reactor and turbine, then stop the "A" RCP. B The "A" RCP Middle Seal has failed. Evaluate the condition of the other seals to confirm no other degradation or failures.

C The Bleedoff Pressure Controller, PIC-215, has malfunctioned.

Using the Foxboro Controller, restore "A" RCP Bleedoff pressure and flow to the normal band. D The RCP Bleedoff Relief Valve has inadvertently opened. Evaluate "Au RCP Seal pressures to determine whether or not the "A" RCP may remain in operation.

Page 3 of25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Up2rade Exam Question # 4 The plant is operating at 100% power with the "B" Auxiliary Feedwater (AFW) Pump out of service for maintenanGe.

Then the following events occur:

• Automatic plant trip due to a Steam Generator Tube Rupture (SGTR) on #2 Steam Generator '" Loss of the RSST and VA-20 at the time of '" Shortly after the trip, a Safety Injection Actuation signal (SIAS) automatically " All other plant systems respond as Which one of the following actions must the US perform during EOP 2525, Standard Post Trip Actions, to mitigate the consequences of this event and what is the reason for this action? Dispatch a PEO to the Hot Shutdown Panel, C-21, to throttle open the #2 Atmospheric Dump Valve. will permit a cooldown of both Hot Leg temperatures to Direct the BOP to swap the control power supply switch for the Terry Turbine to Facility 1. This will allow the operator to maintain both S/G levels in the prescribed bands. Dispatch a PEO to manually operate the "B" Auxiliary Feedwater Regulating Valve, 2-FW-43B.

This prevent excessive auxiliary feedwater from overfilling the affected Direct the BOP to close #2 S/G Steam Supply to the Terry Turbine, MS-202, after the disconnect is This will minimize the radioactive release from the affected Page 4 of Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Up2rade Exam Question # 4 The plant is operating at 100% power with the "B" Auxiliary Feedwater (AFW) Pump out of service for maintenanGe.

Then the following events occur:

• Automatic plant trip due to a Steam Generator Tube Rupture (SGTR) on #2 Steam Generator (SG). '" Loss of the RSST and VA-20 at the time of trip. '" Shortly after the trip, a Safety Injection Actuation signal (SIAS) automatically actuated. " All other plant systems respond as designed.

Which one of the following actions must the US perform during EOP 2525, Standard Post Trip Actions, to mitigate the consequences of this event and what is the reason for this action? A Dispatch a PEO to the Hot Shutdown Panel, C-21, to throttle open the #2 Atmospheric Dump Valve. This will permit a cooldown of both Hot Leg temperatures to 515°F. B Direct the BOP to swap the control power supply switch for the Terry Turbine to Facility 1. This will allow the operator to maintain both S/G levels in the prescribed bands. C Dispatch a PEO to manually operate the "B" Auxiliary Feedwater Regulating Valve, 2-FW-43B.

This will prevent excessive auxiliary feedwater from overfilling the affected SG. D Direct the BOP to close #2 S/G Steam Supply to the Terry Turbine, MS-202, after the disconnect is closed. This will minimize the radioactive release from the affected SG. Page 4 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Question# The plant tlas expE3rienced a loss of VA-10 while in Mode 5 with Shutdown Cooling in operation.

Assuming RBCCVII flow was NOT diverted from the SOC Heat Exchangers by any other system, which of the following aGtions would be performed outside the Control Room and what is the reason for performing these actions in the listed order? 1. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in counterclockwise direction as directed by the Control 2. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2-SI-:306, SOC Total Flow Control Valve, must be opened first to provide minimum flow for the operating LPSI Pump. 1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in clockwise din3ction as directed by the Control 2. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in counterclockwise direction as directed by the Control 2-SI-(357, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rate. '1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2-SI-1357, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rat,s. 1. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the direction as directed by the Control 2. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-:306, SOC Total Flow Control Valve, must be opened first to provide minimum flow for the operating LPSI Pump. Page 5 of Pri nted on 1 0/28/2009 at 13: 18 Senior Reactor Operator NRC License Exanl Question# The plant tlas expE3rienced a loss of VA-10 while in Mode 5 with Shutdown Cooling in operation.

Assuming RBCCVII flow was NOT diverted from the SOC Heat Exchangers by any other system, which of the following aGtions would be performed outside the Control Room and what is the reason for performing these actions in the listed order? A 1. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2-SI-:306, SOC Total Flow Control Valve, must be opened first to provide minimum flow for the operating LPSI Pump. B 1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the clockwise din3ction as directed by the Control Room. 2. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-(357, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rate. C '1. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2-SI-1357, SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired RCS cooldown rat,s. D 1. Place 2-SI-306, SOC Total Flow Control Valve, in manual and turn the handwheel in the clockwise direction as directed by the Control Room. 2. Place 2-SI-657, SOC Heat Exchanger Flow Control Valve, in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. 2-SI-:306, SOC Total Flow Control Valve, must be opened first to provide minimum flow for the operating LPSI Pump. Page 5 of 25 Pri nted on 1 0/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Question # ....!.J The plant is operating at 100% power when the Balance of Plant (BOP) operator reports that Instrument Air header pressure is at 95 psig and lowering.

Immediately following, the Turbine Building PEa reports a large unisolable leak just downstream of the "D" Instrument Air Dryer After Filters. Assuming I nstrument Air header pressure continues to lower, at what pressure in the Instrument Air System must the Unit Supervisor (US) direct a manual reactor trip (by procedure) and why? A Prior to reaching 85 psig. When pressure drops below 85 psig the crew is procedurally directed to crosstie Station Air with Unit 3. Operation in this alignment will result in all components supplied by Instrument Air being inoperable, which is an unanalyzed condition.

B When pressure lowers to less than 85 psig. At ap:Jroximatley 85 psig the Instrument Air/Station Air Crosstie valve opens. Continued operation with Station Air supplied to valves and controllers will result in erratic operation of components due to the high moisture content of Station Air. C When pressure lowers to less than 80 psig. The loss of many important controls, such as Feedwater, could degrade plant conditions at the time of the trip; tllereforEI, the reactor must be tripped when control of important systems could become challenged.

D Prior to reaching 80 psig. The Auxiliary Feed Regulating Valves will lock up with less than 80 psig supply pressure.

The reactor must be tripped to allow the initial automatic opening of these valves and begin feeding Steam Generators.

Page 6 of 25 Printed on 10/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Exam Question # ....!.J The plant is operating at 100% power when the Balance of Plant (BOP) operator reports that Instrument Air header pressure is at 95 psig and lowering.

Immediately following, the Turbine Building PEa reports a large unisolable leak just downstream of the "D" Instrument Air Dryer After Filters. Assuming I nstrument Air header pressure continues to lower, at what pressure in the Instrument Air System must the Unit Supervisor (US) direct a manual reactor trip (by procedure) and why? A Prior to reaching 85 psig. When pressure drops below 85 psig the crew is procedurally directed to crosstie Station Air with Unit 3. Operation in this alignment will result in all components supplied by Instrument Air being inoperable, which is an unanalyzed condition.

B When pressure lowers to less than 85 psig. At ap:Jroximatley 85 psig the Instrument Air/Station Air Crosstie valve opens. Continued operation with Station Air supplied to valves and controllers will result in erratic operation of components due to the high moisture content of Station Air. C When pressure lowers to less than 80 psig. The loss of many important controls, such as Feedwater, could degrade plant conditions at the time of the trip; tllereforEI, the reactor must be tripped when control of important systems could become challenged.

D Prior to reaching 80 psig. The Auxiliary Feed Regulating Valves will lock up with less than 80 psig supply pressure.

The reactor must be tripped to allow the initial automatic opening of these valves and begin feeding Steam Generators.

Page 6 of 25 Printed on 10/28/2009 at 13: 18 Senior Reactor Operator NRC License Up2rade Question # 7 The reactor is at 100% power with the CEA Motion surveillance in progress.

When Group 7 CEA #1 is tested, CEAPDS indicates it inserts two steps, then slips an additional 20 steps. The appropriate actions were taken to stabilize ReS temperature and the following conditions were observed:

• Reactor power stable at -99%.

• Only Upper Electrical Limit lights are energized on the core mimic.

• CEA #1 indicates 158 steps withdrawn on CEAPDS.

• CEA #1 indicates 178 steps withdrawn on the PPC.

• CEA Motion Inhibit (CMI) alarms on C-04

• CEAPDS Group Deviation indication for CEA #1 Fifty (50) minutes after CEA #1 slipped, all required actions per AOP-2556, CEDS Malfunctions, have completed, including plant power Also, I&C reports the circuit malfunction that caused CEA #1 to slip has been repaired and the CEA can now Which one of the following describes actions that must be taken to recover CEA #1 and what is the concern of those Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery.

CEA #1 Count Indication and the CMI will be INOPERABLE while the CEA is being recovered. CEA #1 Upper Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Reed Switch Indication for CEA #1 and the CMI will be INOPERABLE while the CEA is being recovered. Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered. CEA #1 Uppm Electrical Limit must be overridden and the CMI must be bypassed for CEA Only the CMI will be INOPERABLE while the CEA is being Page 7 of Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Up2rade Exam Question # 7 The reactor is at 100% power with the CEA Motion surveillance in progress.

When Group 7 CEA #1 is tested, CEAPDS indicates it inserts two steps, then slips an additional 20 steps. The appropriate actions were taken to stabilize ReS temperature and the following conditions were observed:

• Reactor power stable at -99%.

• Only Upper Electrical Limit lights are energized on the core mimic.

• CEA #1 indicates 158 steps withdrawn on CEAPDS.

• CEA #1 indicates 178 steps withdrawn on the PPC.

• CEA Motion Inhibit (CMI) alarms on C-04

• CEAPDS Group Deviation indication for CEA #1 Fifty (50) minutes after CEA #1 slipped, all required actions per AOP-2556, CEDS Malfunctions, have been completed, including plant power changes. Also, I&C reports the circuit malfunction that caused CEA #1 to slip has been repaired and the CEA can now be recovered.

Which one of the following describes actions that must be taken to recover CEA #1 and what is the administrative concern of those actions? A Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery.

CEA #1 Count Indication and the CMI will be INOPERABLE while the CEA is being recovered.

B CEA #1 Upper Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Reed Switch Indication for CEA #1 and the CMI will be INOPERABLE while the CEA is being recovered.

C Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered.

o CEA #1 Uppm Electrical Limit must be overridden and the CMI must be bypassed for CEA recovery.

Only the CMI will be INOPERABLE while the CEA is being recovered.

Page 7 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Question # --.!.J The reacto" is manually tripped from 100% power due to a Steam Generator Tube Rupture (SGTR). On the trip, the RSST is lost due to grid instabilities.

All other systems respond normally.

EOP 2525, Standard Post Trip Actions, is entered. What is thEl direction for controlling the affected SG level during the performance of EOP 2525, Standard Post Trip Actions, and what is the basis for this direction?

A Secure all feedwater to the affected SG. This will provide the maximum volume to accept water from the tube rupture and still allow a cooldown to 515°F to isolate the affected SG. B Maintain the affected SG level 40 to 70%. This will maintain the SG tube covered to allow a cooldown to 515°F and still maintain adequate volume to accept water from the tube rupture. C Maintain at least 300 gpm feedwater to the affected SG for Heat Removal. The addition of clean water will provide dilution of radioactivity which will lower the release to the environment.

D Feed the affected SG to maintain level 40 to 45%. This will cover the SG tube to allow for Iodine scrubbing and still allow adequate volume to accept water from the tube rupture. Page 8 of 25 Printed on 10/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Exam Question # --.!.J The reacto" is manually tripped from 100% power due to a Steam Generator Tube Rupture (SGTR). On the trip, the RSST is lost due to grid instabilities.

All other systems respond normally.

EOP 2525, Standard Post Trip Actions, is entered. What is thEl direction for controlling the affected SG level during the performance of EOP 2525, Standard Post Trip Actions, and what is the basis for this direction?

A Secure all feedwater to the affected SG. This will provide the maximum volume to accept water from the tube rupture and still allow a cooldown to 515°F to isolate the affected SG. B Maintain the affected SG level 40 to 70%. This will maintain the SG tube covered to allow a cooldown to 515°F and still maintain adequate volume to accept water from the tube rupture. C Maintain at least 300 gpm feedwater to the affected SG for Heat Removal. The addition of clean water will provide dilution of radioactivity which will lower the release to the environment.

D Feed the affected SG to maintain level 40 to 45%. This will cover the SG tube to allow for Iodine scrubbing and still allow adequate volume to accept water from the tube rupture. Page 8 of 25 Printed on 10/28/2009 at 13: 18 Senior Reactor Operator NRC License Up2rade Exam Question # .....!..J The plant is in MODE 5 performing a normal cooldown for refueling. "B" LPSI Pump is in service supplying SDC Heat RCS to SDC Temperature, T351 X, is presently reading 187"F with RCS pressure held at 150 Suddenly, 13us 24D is deenergized due to a fault. Fifteen minutes after the loss of Bus 24D, the conditions are

-RCS pressure is 155 psia and slowly -RCS to SDC Temperature, T351X, is reading 186°F and -CET temperatures are 205°F and slowly -RVLMS indicates vessel level at -Both S/G levels are 60% and -Containment is being NO other operator actions have been Which of the following notifications must be A General Interest, Echo B Unusual Event, Delta-One C Alert, Charlie-One D Site Area Emergency, Charlie-Two Page 9 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Up2rade Exam Question # .....!..J The plant is in MODE 5 performing a normal cooldown for refueling. "B" LPSI Pump is in service supplying both SDC Heat RCS to SDC Temperature, T351 X, is presently reading 187"F with RCS pressure being held at 150 psia. Suddenly, 13us 24D is deenergized due to a fault. Fifteen minutes after the loss of Bus 24D, the following conditions are reported:

-RCS pressure is 155 psia and slowly rising. -RCS to SDC Temperature, T351X, is reading 186°F and stable. -CET temperatures are 205°F and slowly rising. -RVLMS indicates vessel level at 100%. -Both S/G levels are 60% and stable -Containment is being evacuated.

NO other operator actions have been taken. Which of the following notifications must be made? A General Interest, Echo B Unusual Event, Delta-One C Alert, Charlie-One D Site Area Emergency, Charlie-Two Page 9 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Up2rade Exam Question # An RCS chemistry sample taken at 100% power indicates 6 micro-curies/gram DOSE EQUIVALENT 1-131. Which of the following describes the required action and the basis for that action? A With the specific activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131, be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after detection.

Isotopic analysis of the primary coolant must be performed once per hour when activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131. The hourly sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and there is significant conservatism built into the RCS specific activity limit. B With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 .:::. 60 micro-curies/gram once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 micro-curies/gram limit. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and it is expected that normal coolant Iodine concEmtration would be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. C With the specific activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from the time of detection and in COLD SHUTDOWN within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of detection.

The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power in an orderly manner and prevent exceeding the radiological release limit at the site boundary from an assumed LOCA. With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, lower the RCS specific activity to.:::. 1.0 micro-curies/gram DOSE EQUIVALENT 1-131 within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. It is expected that normal coolant Iodine concentration would be restored within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. If not. adequate time is provided to achieve HOT STANDBY to prevent exceeding Control Room dose limits from an assumed LOCA. Page 10 of Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Up2rade Exam Question # An RCS chemistry sample taken at 100% power indicates 6 micro-curies/gram DOSE EQUIVALENT 1-131. Which of the following describes the required action and the basis for that action? A With the specific activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131, be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after detection.

Isotopic analysis of the primary coolant must be performed once per hour when activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131. The hourly sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and there is significant conservatism built into the RCS specific activity limit. B With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 .:::. 60 micro-curies/gram once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 micro-curies/gram limit. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line break or S/G tube rupture in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and it is expected that normal coolant Iodine concEmtration would be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. C With the specific activity of the primary coolant> 0.1 micro-curies/gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from the time of detection and in COLD SHUTDOWN within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of detection.

The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power in an orderly manner and prevent exceeding the radiological release limit at the site boundary from an assumed LOCA. o With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, lower the RCS specific activity to.:::. 1.0 micro-curies/gram DOSE EQUIVALENT 1-131 within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. It is expected that normal coolant Iodine concentration would be restored within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. If not. adequate time is provided to achieve HOT STANDBY to prevent exceeding Control Room dose limits from an assumed LOCA. Page 10 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Question # 11 I A plant heatup has just been started and the following conditions presently -RCS Temperature is at 205°F and slowly -RCS pressure is stable at the minimum allowed for "A" and "S" RCP

-"A" and "S" RCPs have just been

-Shutdown Cooling has just been

-"C" and "0" RCP breakers have just been rack.ed Then, the "S" RCP trips when the breaker's overcurrent relay actuated due to being jarred while moving (NOT an actual overcurrent Which of the following actions are required under the present A ImmE!diately secure the "A" RCP, raise RCS pressure, then start "C" and "0" RCPs. B Imm*!diately place Shutdown Cooling back. in operation, then secure the "A" RCP. C Immediately start "C" and "0" RCPs, then secure the "A" RCP. D Immediately start the "C" RCP and operate it with the "A" RCP. Page 11 of25 Pri nted on 1 0/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Exam Question # 11 I A plant heatup has just been started and the following conditions presently exist: -RCS Temperature is at 205°F and slowly rising. -RCS pressure is stable at the minimum allowed for "A" and "S" RCP operation.

-"A" and "S" RCPs have just been started. -Shutdown Cooling has just been secured. -"C" and "0" RCP breakers have just been rack.ed up. Then, the "S" RCP trips when the breaker's overcurrent relay actuated due to being jarred while moving staging (NOT an actual overcurrent condition).

Which of the following actions are required under the present conditions?

A ImmE!diately secure the "A" RCP, raise RCS pressure, then start "C" and "0" RCPs. B Imm*!diately place Shutdown Cooling back. in operation, then secure the "A" RCP. C Immediately start "C" and "0" RCPs, then secure the "A" RCP. D Immediately start the "C" RCP and operate it with the "A" RCP. Page 11 of25 Pri nted on 1 0/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Question # 12 I A Rapid Downpower at 50%/hr is in progress due to an RCS leak in containment that exceeds the administrative limit. The following plant conditions presently exist: " Plant power is 93% and dropping at the intended rate. " Pressurizer level is 65% and stable. " RCS pressure is 2250 psia and stable. " One charging pump is running, Letdown is at approximately 30 gpm . .. Adding boric acid to the charging pump suction to maintain the desired rate of power reduction.

.. Forcing Pressurizer Sprays in progress . .. C02/3 annunciator in alarm; D-37, "PZR PRESSURE SELECTED CHANNEL DEVIATION HI/LO" '" Chann*ll"Y" Pressurizer Level and Pressure controlling normally.

Then, during the load reduction, Pressurizer Level Channel "X" fails to zero (0) and the following occur: " All control respond as designed to the failure.

• C02/3 annunciator in alarm; A-38, "PRESSURIZER CH X LEVEL HI/LO". " C02/3 annunciator in alarm; C-38, "PRESSURIZER CH X LEVEL LO-LO". " PPC alarms on Monitor #2 indicative of the instrument failure. Which of the following actions must the Unit Supervisor direct and why? Per the ARP for C-38; shift pressurizer heater control to channel"Y" and restore pressurizer heaters. to ensure adequate margin from DNB is maintained. Per SP-2602A RCS Leakage; deselect Pressurizer Level Channel "X" from the leak rate calculation, to ensure valid trending of RCS Leak Rate by the PPC. Per tile ARP for A-38; place the standby charging pumps in "Pull-To-Lock", to prevent the rate of the plant down power from accelerating above 50%/hr. Per AOP-25i'5, Rapid Downpower; shift pressurizer heater control to channel"Y" and restore the pressure controller setpoint, to prevent a plant trip on TM/LP. Page 12 of Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # 12 I A Rapid Downpower at 50%/hr is in progress due to an RCS leak in containment that exceeds the administrative limit. The following plant conditions presently exist: " Plant power is 93% and dropping at the intended rate. " Pressurizer level is 65% and stable. " RCS pressure is 2250 psia and stable. " One charging pump is running, Letdown is at approximately 30 gpm . .. Adding boric acid to the charging pump suction to maintain the desired rate of power reduction.

.. Forcing Pressurizer Sprays in progress . .. C02/3 annunciator in alarm; D-37, "PZR PRESSURE SELECTED CHANNEL DEVIATION HI/LO" '" Chann*ll"Y" Pressurizer Level and Pressure controlling normally.

Then, during the load reduction, Pressurizer Level Channel "X" fails to zero (0) and the following occur: " All control respond as designed to the failure.

• C02/3 annunciator in alarm; A-38, "PRESSURIZER CH X LEVEL HI/LO". " C02/3 annunciator in alarm; C-38, "PRESSURIZER CH X LEVEL LO-LO". " PPC alarms on Monitor #2 indicative of the instrument failure. Which of the following actions must the Unit Supervisor direct and why? A Per the ARP for C-38; shift pressurizer heater control to channel"Y" and restore pressurizer heaters. to ensure adequate margin from DNB is maintained.

B Per SP-2602A RCS Leakage; deselect Pressurizer Level Channel "X" from the leak rate calculation, to ensure valid trending of RCS Leak Rate by the PPC. C Per tile ARP for A-38; place the standby charging pumps in "Pull-To-Lock", to prevent the rate of the plant down power from accelerating above 50%/hr. D Per AOP-25i'5, Rapid Downpower; shift pressurizer heater control to channel"Y" and restore the pressure controller setpoint, to prevent a plant trip on TM/LP. Page 12 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # 13 1 A reactor startup is in progress using CEA withdrawal.

The RO has just stopped withdrawing Group # 7 CEAs and makes the following announcements:

• The reactor is critical.

• Startup rate is positive and stabilizing at -1.5 DPM. Which of the following actions should the Reactivity SRO direct? A Commence Emergency Boration until the reactor is B Insert the Group #7 CEAs to lower the startup rate below 0.5 C Trip the reactor and Go to EOP 2525, "Standard Post Trip D Inser: all CEAs per OP-2206, "Reactor Shutdown" and notify Page 13 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # 13 1 A reactor startup is in progress using CEA withdrawal.

The RO has just stopped withdrawing Group # 7 CEAs and makes the following announcements:

• The reactor is critical.

• Startup rate is positive and stabilizing at -1.5 DPM. Which of the following actions should the Reactivity SRO direct? A Commence Emergency Boration until the reactor is subcritical.

B Insert the Group #7 CEAs to lower the startup rate below 0.5 DPM. C Trip the reactor and Go to EOP 2525, "Standard Post Trip Actions".

D Inser: all CEAs per OP-2206, "Reactor Shutdown" and notify RE. Page 13 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Exam Question# The plant is in Mode 6 with the following conditions:

• Core re-Ioad in progress and approximately half way completed.

• "A" LPSI pump running for Shutdown Cooling (SOC) operation.

• "8" LPSI pump in standby, aligned for SOC use. * "A" train of Spent Fuel Pool (SFP) cooling in service. Then, "A" LPSI pump is lost due to a breaker fault. When the "8" LPSI pump is started, it seizes and trips breaker The Unit Supervisor (US) then directs the RO to recover SOC using the "8" Containment Spray (CS) Which of tile following additional directions must the US give, while SOC flow is being supplied by a CS A All fuel movement in containment must remain secured. B SOC supplementing of SFP cooling must be secured. C Containment must remain evacuated of non-essential personnel.

D Containment Closure must be fully set with all access doors closed. Page 14 of 25 Pri nted on 1 0/28/2009 at 13: 18 Senior Reactor Operator NRC License Exam Question# The plant is in Mode 6 with the following conditions:

• Core re-Ioad in progress and approximately half way completed.

• "A" LPSI pump running for Shutdown Cooling (SOC) operation.

• "8" LPSI pump in standby, aligned for SOC use. * "A" train of Spent Fuel Pool (SFP) cooling in service. Then, "A" LPSI pump is lost due to a breaker fault. When the "8" LPSI pump is started, it seizes and trips on breaker overload.

The Unit Supervisor (US) then directs the RO to recover SOC using the "8" Containment Spray (CS) pump. Which of tile following additional directions must the US give, while SOC flow is being supplied by a CS pump? A All fuel movement in containment must remain secured. B SOC supplementing of SFP cooling must be secured. C Containment must remain evacuated of non-essential personnel.

D Containment Closure must be fully set with all access doors closed. Page 14 of 25 Pri nted on 1 0/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Question# The required surveillance must be performed after repairs were made to the "A" Service Water Strainer Flush Valve, 2-SW-90A.

The completed surveillance indicates the valve stroke time is slightly above the Maximum 'Normal Limit", but is below the Maximum "Acceptable Limit". All other parameters are within the "Acceptable Limits". Which of the following describes the condition of the "A" Service Water Pump and the required action? A The "A" Service Water Pump remains inoperable.

Perform the required repairs to the "A" Service Water Strainer Flush Valve, 2-SW-90A, then perform the required surveillances to restore the "A" Service Water Pump to OPERABLE.

B The "A" Service Water Pump remains inoperable.

Obtain a different set of test equipment and immediately retest the "A" Service Water Strainer Flush Valve, 2-SW-90A, again to verify that the previous data was accurate.

C The "A" Service Water Pump is considered OPERABLE.

Place the "A" Service Water Pump Strainer in an Augmented Testing Program and test weekly to ensure the "A" Service Water Strainer Flush Valve, SW-BOA, remains OPERABLE. The 'A" Service Water Pump is considered OPERABLE.

If the "A" Service Water Strainer Flush Valve, SW-BOA, exceeds the Maximum "Normal Limit" on an immediate retest, then declare the "A" Service Water Pump inoperable.

Page 15 of Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question# The required surveillance must be performed after repairs were made to the "A" Service Water Strainer Flush Valve, 2-SW-90A.

The completed surveillance indicates the valve stroke time is slightly above the Maximum 'Normal Limit", but is below the Maximum "Acceptable Limit". All other parameters are within the "Acceptable Limits". Which of the following describes the condition of the "A" Service Water Pump and the required action? A The "A" Service Water Pump remains inoperable.

Perform the required repairs to the "A" Service Water Strainer Flush Valve, 2-SW-90A, then perform the required surveillances to restore the "A" Service Water Pump to OPERABLE.

B The "A" Service Water Pump remains inoperable.

Obtain a different set of test equipment and immediately retest the "A" Service Water Strainer Flush Valve, 2-SW-90A, again to verify that the previous data was accurate.

C The "A" Service Water Pump is considered OPERABLE.

Place the "A" Service Water Pump Strainer in an Augmented Testing Program and test weekly to ensure the "A" Service Water Strainer Flush Valve, 2-SW-BOA, remains OPERABLE.

o The 'A" Service Water Pump is considered OPERABLE.

If the "A" Service Water Strainer Flush Valve, 2-SW-BOA, exceeds the Maximum "Normal Limit" on an immediate retest, then declare the "A" Service Water Pump inoperable.

Page 15 of 25 Printed on 10/28/2009 at 13:18 c Senior Reactor Operator NRC License Upgrade Question # 16 I The plant is in MODE 6 with the following

-Fuel movement is in

-The Personnel Airlock Doors are -The Equipment Hatch is -Containment Purge is in

-Containment Atmosphere Radiation Monitor, RM-8123, is out of service for The Auxiliary Building PEO has just reported that the blower for Containment Atmosphere Radiation RM-8262, has tripped and is very hot to the Which of tile following actions must be taken and suspend CORE ALTERATIONS and establish Containment Closure prior to resuming fuel mOVE!ment, to ensure a potential fuel handling accident in Containment is NOT released to the environment.

ImmEldiately suspend CORE ALTERATIONS and restore the Radiation Monitor blower prior to resuming fuel B mOVE!ment, to ensure a potential fuel handling accident in Containment is NOT released to the environment.

Ensure a control room operator is specifically assigned to close the Containment Purge Valves within 30 minutes of an event, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment.

Restore the Containment Purge Valves to OPERABLE status within the next 30 minutes or immediately D close the Purge Valves, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment.

Page 16 of Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # 16 I The plant is in MODE 6 with the following conditions:

-Fuel movement is in progress.

-The Personnel Airlock Doors are open -The Equipment Hatch is open. -Containment Purge is in operation.

-Containment Atmosphere Radiation Monitor, RM-8123, is out of service for repairs. The Auxiliary Building PEO has just reported that the blower for Containment Atmosphere Radiation Monitor, RM-8262, has tripped and is very hot to the touch. Which of tile following actions must be taken and why? A suspend CORE ALTERATIONS and establish Containment Closure prior to resuming fuel mOVE!ment, to ensure a potential fuel handling accident in Containment is NOT released to the environment.

B c D ImmEldiately suspend CORE ALTERATIONS and restore the Radiation Monitor blower prior to resuming fuel mOVE!ment, to ensure a potential fuel handling accident in Containment is NOT released to the environment.

Ensure a control room operator is specifically assigned to close the Containment Purge Valves within 30 minutes of an event, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment.

Restore the Containment Purge Valves to OPERABLE status within the next 30 minutes or immediately close the Purge Valves, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment.

Page 16 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Up2rade Question # ...!:....I The plant was at 100% power when CONVEX ordered Main Generator output be lowered from 900 MWe to MWe in 15 AOP 2557, "Emergency Generation Reduction", was initiated and the following conditions now

• Group? CEAs are at 170 steps withdrawn.

• Main Generator output is 610 MWe and slowly lowering.

• "A" Steam Dump Bypass Valve is 75% open and stable. * "BYPASS TO CND", PIC-4216 output is 83% and stable. * "8", "C" and "0" Steam Dump Bypass Valves are open 75% and stable. * "STEAM DUMP TAVG CNTL", TIC-4165 output is 85% and stable .. "RC LOOP 1 COLD LEG TEMP HI" annunciator has just alarmed (C02/3; * !lRC LOOP 2 COLD LEG TEMP HI" annunciator has just alarmed (C02/3; 0-34) .. RCS Teold is 550 OF and slowly riSing Which one of the following actions should the US direct? A Transfer control of the steam dumps to Foxboro IA control and lower Tcold to program. B Lower the setpoint on the "A" steam dump and log into the DNB Technical Specification.

C ImmE!diately trip the reactor and go to EOP 2525, "Standard Post Trip Actions".

D Insert CEAs until Tcold is back on program and both C02/3 alarms have cleared. Page 17 of 25 Printed on 10128/2009 at 13:18 Senior Reactor Operator NRC License Up2rade Exam Question # ...!:....I The plant was at 100% power when CONVEX ordered Main Generator output be lowered from 900 MWe to 600 MWe in 15 minutes. AOP 2557, "Emergency Generation Reduction", was initiated and the following conditions now exist:

• Group? CEAs are at 170 steps withdrawn.

• Main Generator output is 610 MWe and slowly lowering.

• "A" Steam Dump Bypass Valve is 75% open and stable. * "BYPASS TO CND", PIC-4216 output is 83% and stable. * "8", "C" and "0" Steam Dump Bypass Valves are open 75% and stable. * "STEAM DUMP TAVG CNTL", TIC-4165 output is 85% and stable . .. "RC LOOP 1 COLD LEG TEMP HI" annunciator has just alarmed (C02/3; C-34). * !lRC LOOP 2 COLD LEG TEMP HI" annunciator has just alarmed (C02/3; 0-34) . .. RCS Teold is 550 OF and slowly riSing (RPS). Which one of the following actions should the US direct? A Transfer control of the steam dumps to Foxboro IA control and lower Tcold to program. B Lower the setpoint on the "A" steam dump and log into the DNB Technical Specification.

C ImmE!diately trip the reactor and go to EOP 2525, "Standard Post Trip Actions".

D Insert CEAs until Tcold is back on program and both C02/3 alarms have cleared. Page 17 of 25 Printed on 10128/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Question # ....!!..J The plant is in normal operation at 100% power, when a Fire System Trouble annunciator is received on C06/7 and Zone 45 on Fire Panel, C-26. The Auxiliary Building PEO subsequently calls from the West DC Switchgear Room and reports the following:

• One Ion Chamber smoke detector is in alarm.

• The Halon strobe lights and horn are pulsating slowly.

• All other smoke detectors are operating normally (not in alarm).

• There is no smoke or fire in the area. The detector appears to have failed. Which of the following describes the impact of the above conditions, and the direction the US will give? The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM "Halon Fire Suppression System", provide backup fire suppression and establish a fire watch, when room has The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.3.3.7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable and a fire watch must be established. The Fire Suppression system is warning that a discharge will occur after a timer countdown.

Per TRM 3-1.9.4, "Halon Fire Suppression System", provide backup fire suppression and establish a fire watch, when the room has cleared. The Fire Suppression system is warning that a discharge will occur after a timer countdown.

Per 3.3.3.7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable, establish a watch, when the room has Page 18 of Printed on 10/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Exam Question # ....!!..J The plant is in normal operation at 100% power, when a Fire System Trouble annunciator is received on C06/7 and Zone 45 on Fire Panel, C-26. The Auxiliary Building PEO subsequently calls from the West DC Switchgear Room and reports the following:

• One Ion Chamber smoke detector is in alarm.

• The Halon strobe lights and horn are pulsating slowly.

• All other smoke detectors are operating normally (not in alarm).

• There is no smoke or fire in the area. The detector appears to have failed. Which of the following describes the impact of the above conditions, and the direction the US will give? A The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.7.9.4, "Halon Fire Suppression System", provide backup fire suppression and establish a fire watch, when the room has cleared. B The Fire Suppression system is alarming as a warning of a potential for a discharge.

Per TRM 3.3.3.7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable and a fire watch must be established.

C The Fire Suppression system is warning that a discharge will occur after a timer countdown.

Per TRM 3-1.9.4, "Halon Fire Suppression System", provide backup fire suppression and establish a fire watch, when the room has cleared. D The Fire Suppression system is warning that a discharge will occur after a timer countdown.

Per TRM 3.3.3.7, "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable, establish a fire watch, when the room has cleared. Page 18 of 25 Printed on 10/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Question # 19 I While operating at Beginning of Life (BOL), 100% power, the "A" Main Feed Pump high vibration annunciator alarmed. After subsequent investigation and troubleshooting, the System Engineer and Maintenance agree that the pump must be removed from service within the next 60 minutes to prevent severe damage. The crew has just entered AOP 2575>, Rapid Downpower.

Which of the following statements describes the method that must be utilized to perform this evolution?

A Use Reactivity Plan RE-G-OB to reduce power to 70%, secure the Feed Pump, then transition to OP 220B, Attachment Reactivity Thumbrules, for maintaining power with rising Xenon. B Use Reactivity Plan RE-G-05 until 90% power, then transition to AOP 2575, Attachment 7, Boration/Power Reduction Rates to continue the power reduction required to secure the Feed Pump. C Use Reactivity Plan RE-G-11 to reduce power to the appropriate level, secure the Feed Pump, then raise power to the appropriate level using OP 2204, Load Changes, and a new Reactivity Plan. D Use Reactivity Plan RE-G-10 to reduce power to the appropriate level, secure the Feed Pump, then transition to OP 2393, Core Power Distribution and Monitoring, to maintain AS!. Page 19 of 25 Printed on 10/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Exam Question # 19 I While operating at Beginning of Life (BOL), 100% power, the "A" Main Feed Pump high vibration annunciator alarmed. After subsequent investigation and troubleshooting, the System Engineer and Maintenance agree that the pump must be removed from service within the next 60 minutes to prevent severe damage. The crew has just entered AOP 2575>, Rapid Downpower.

Which of the following statements describes the method that must be utilized to perform this evolution?

A Use Reactivity Plan RE-G-OB to reduce power to 70%, secure the Feed Pump, then transition to OP 220B, Attachment Reactivity Thumbrules, for maintaining power with rising Xenon. B Use Reactivity Plan RE-G-05 until 90% power, then transition to AOP 2575, Attachment 7, Boration/Power Reduction Rates to continue the power reduction required to secure the Feed Pump. C Use Reactivity Plan RE-G-11 to reduce power to the appropriate level, secure the Feed Pump, then raise power to the appropriate level using OP 2204, Load Changes, and a new Reactivity Plan. D Use Reactivity Plan RE-G-10 to reduce power to the appropriate level, secure the Feed Pump, then transition to OP 2393, Core Power Distribution and Monitoring, to maintain AS!. Page 19 of 25 Printed on 10/28/2009 at 13: 18 Senior Reactor Operator NRC License Upgrade Question # 20 I Which of tile following actions require authorization by the Refueling SRO? When the grapple will NOT disengage the top of a fuel assembly, snap or twang the hoist cable to release the When an overload occurs, use the hand crank on the refuel machine hoist to free the fuel assembly from the guide pins. In an emergemcy, insert a fuel assembly into the core and ungrapple it provided NO other fuel assemblies are adjacent. If an underload occurs prematurely, raise the fuel assembly, pull the mast detent pin, rotate slightly, reinsert the Page 20 of Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # 20 I Which of tile following actions require authorization by the Refueling SRO? A When the grapple will NOT disengage the top of a fuel assembly, snap or twang the hoist cable to release the B When an overload occurs, use the hand crank on the refuel machine hoist to free the fuel assembly from the guide pins. C In an emergemcy, insert a fuel assembly into the core and ungrapple it provided NO other fuel assemblies are adjacent.

o If an underload occurs prematurely, raise the fuel assembly, pull the mast detent pin, rotate slightly, and reinsert the elssembly.

Page 20 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Question # 21 ul The plant is operating at 100% power when ISO New England and CONVEX operators notify Millstone Station that a "Degraded Voltage" condition exists. Voltage on the 4.16 kV buses is presently 3,900 volts.

Based on this information, which one of the following describes actions that the Unit Supervisor must direct, per the applicable procedures?

A Rack out the 6.9 and 4.16 kV breakers to the RSST and slow-start both Emergency Diesel Generators.

B Terminate surveillance testing of any safety related pumps and motors and secure them, if possible.

C Commence a plant down power and secure all unnecessary equipment as the lower power permits. D Ensure the "E:" and "F" Instrument Air compressors are operating in the "Lead" and "Standby" modes. Page 21 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Exam Question # 21 ul The plant is operating at 100% power when ISO New England and CONVEX operators notify Millstone Station that a "Degraded Voltage" condition exists. Voltage on the 4.16 kV buses is presently 3,900 volts.

Based on this information, which one of the following describes actions that the Unit Supervisor must direct, per the applicable procedures?

A Rack out the 6.9 and 4.16 kV breakers to the RSST and slow-start both Emergency Diesel Generators.

B Terminate surveillance testing of any safety related pumps and motors and secure them, if possible.

C Commence a plant down power and secure all unnecessary equipment as the lower power permits. D Ensure the "E:" and "F" Instrument Air compressors are operating in the "Lead" and "Standby" modes. Page 21 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Question # 22 I The Auxiliary Building PEO has just noted an active boric acid leak on the bottom of a flange on CH-198, "RCP Bleedoff Pressure Control Valve to VCT". The leak is very small (2-3 drops per minute), but boric acid deposits from the leak are corroding a pipe support bracket located below the flange. Which of the following administrative control documents require action be taken to control this leak? A Final Safety Analysis Report, Chapter 15, License Renewal, Aging Management Programs B Technical Specifications, Reactor Coolant System Leakage, LCO 3.4.6.2 C Technical Requirements Manual, Containment Isolation Valves, LCO 3.6.3.1 D Operational Configuration Control, OP-M-1500, Alternate Plant Configurations, Attachment 5 Page 22 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # 22 I The Auxiliary Building PEO has just noted an active boric acid leak on the bottom of a flange on CH-198, "RCP Bleedoff Pressure Control Valve to VCT". The leak is very small (2-3 drops per minute), but boric acid deposits from the leak are corroding a pipe support bracket located below the flange. Which of the following administrative control documents require action be taken to control this leak? A Final Safety Analysis Report, Chapter 15, License Renewal, Aging Management Programs B Technical Specifications, Reactor Coolant System Leakage, LCO 3.4.6.2 C Technical Requirements Manual, Containment Isolation Valves, LCO 3.6.3.1 D Operational Configuration Control, OP-M-1500, Alternate Plant Configurations, Attachment 5 Page 22 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question# The Rad. Waste PEO has just brought an Aerated Radioactive Waste (ARW) Monitor Tank discharge permit to the Shift Manager for review and approval.

Upon reviewing the permit and ARW system status, the SM has noticed that the ARW monitor tank was sampled by chemistry for the generation of the discharge permit with a level of 85%. However, the tank now has an actual level of 95%. Which of the following actions are required in order for the Shift Manager to approve discharging the ARW Monitor Tank? A Re-calculate the amount of the discharge based on the new tank volume, and note this on the existing discharge permit when complete.

B Re-mix the tank for the required period of time, then resample the tank and generate a new discharge perm t based on the new sample. C Re-sample tank and generate a second discharge permit and discharge the tank based on the most conservative of the two permits. D Re-mix the tank contents to ensure thorough mixing with the previously sampled contents and discharge the tank on the existing permit. Page 23 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question# The Rad. Waste PEO has just brought an Aerated Radioactive Waste (ARW) Monitor Tank discharge permit to the Shift Manager for review and approval.

Upon reviewing the permit and ARW system status, the SM has noticed that the ARW monitor tank was sampled by chemistry for the generation of the discharge permit with a level of 85%. However, the tank now has an actual level of 95%. Which of the following actions are required in order for the Shift Manager to approve discharging the ARW Monitor Tank? A Re-calculate the amount of the discharge based on the new tank volume, and note this on the existing discharge permit when complete.

B Re-mix the tank for the required period of time, then resample the tank and generate a new discharge perm t based on the new sample. C Re-sample tank and generate a second discharge permit and discharge the tank based on the most conservative of the two permits. D Re-mix the tank contents to ensure thorough mixing with the previously sampled contents and discharge the tank on the existing permit. Page 23 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # 24 I The plant was operating normally at 100% power when the crew manually tripped the plant due to a tube on #2 Steam Generator.

The crew successfully performed EOP 2525, Standard post Trip Actions, and EOP 2534, Steam Generator Tube The following conditions -SIAS, CIAS, and E8FAS have been

-"A" and "8" RCPs are running with adequate -Main Steam Line Radiation Monitor, RM 4299C, are presently reading 1.5 R/hr and -Condenser Air Removal is aligned to the Unit 2 -The crew is in the process of lowering both hot leg temperatures to less than or equal to -MSI has. been overridden to maintain steam flow to the

-The Unit 2 Stack Gaseous Radiation Monitor, RM 81328, is in alarm reading 800 cpm and Which of the following statements describes the procedurally directed method used to limit the release radiation to the Secure all Main Exhaust Fans and direct the Chemist to ensure the 95,000 microcurie/sec release limit will NOT be exceeded. Ensure all flow from the Main Condenser to the Steam Jet Air Ejector Radiation Monitor, RM-5099. been Override and start the remaining Main Exhaust Fans and ensure all Radwaste Ventilation supply fans are providing adequate flow. Continue the cooldown and isolate #2 Steam Generator when both hot leg temperatures are less than equal to Page 24 of Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # 24 I The plant was operating normally at 100% power when the crew manually tripped the plant due to a tube rupture on #2 Steam Generator.

The crew successfully performed EOP 2525, Standard post Trip Actions, and entered EOP 2534, Steam Generator Tube Rupture. The following conditions exist: -SIAS, CIAS, and E8FAS have been verified.

-"A" and "8" RCPs are running with adequate NPSH. -Main Steam Line Radiation Monitor, RM 4299C, are presently reading 1.5 R/hr and stable. -Condenser Air Removal is aligned to the Unit 2 Stack. -The crew is in the process of lowering both hot leg temperatures to less than or equal to 515°F. -MSI has. been overridden to maintain steam flow to the Condenser.

-The Unit 2 Stack Gaseous Radiation Monitor, RM 81328, is in alarm reading 800 cpm and rising. Which of the following statements describes the procedurally directed method used to limit the release of radiation to the environment?

A Secure all Main Exhaust Fans and direct the Chemist to ensure the 95,000 microcurie/sec release limit will NOT be exceeded.

B Ensure all flow from the Main Condenser to the Steam Jet Air Ejector Radiation Monitor, RM-5099. has been secured. C Override and start the remaining Main Exhaust Fans and ensure all Radwaste Ventilation supply fans are providing adequate flow. D Continue the cooldown and isolate #2 Steam Generator when both hot leg temperatures are less than or equal to 515°F. Page 24 of 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # ... 251 The plant is stable at 100% power with the Turbine Driven Auxiliary Feedwater Pump out of service for planned maintenance.

Then, the plant trips due to a loss of off site power (state wide blackout), resulting in the following conditions:

-The "A" Main Steam header ruptures in containment on the trip. Busses 24B and 240 are de-energized due to a bus fault on 240. -Facility One SIAS, CIAS, EBFAS, MSI and CSAS have all fully actuated.

-All feedwater has been secured to the #1 Steam Generator (SG). -All other plant systems and components that have power are functioning as designed.

The crew is evaluating numerous alarms and indications caused by the power loss and subsequent ESD. Which of the following alarm indications will require actions to be taken in EOP 2525? C-05 alarms indicating a Excess Steam Demand on the #1 SG, and C-OB alarm indicating VR-21 is C-02/3 alarms indicating RCS temperatures are abnormally low and dropping, and both Boric Acid Pumps are de-energized. C-05 alarms indicating both SG levels abnormally low, and only one Aux. Feedwater pump is feeding the #2 C-01 alarms indicating CTMT Spray has actuated, and C-01 indicating only two CAR fans and one CS pump are Page 25 Printed on 10/28/2009 at 13:18 Senior Reactor Operator NRC License Upgrade Exam Question # ... 251 The plant is stable at 100% power with the Turbine Driven Auxiliary Feedwater Pump out of service for planned maintenance.

Then, the plant trips due to a loss of off site power (state wide blackout), resulting in the following conditions:

-The "A" Main Steam header ruptures in containment on the trip. Busses 24B and 240 are de-energized due to a bus fault on 240. -Facility One SIAS, CIAS, EBFAS, MSI and CSAS have all fully actuated.

-All feedwater has been secured to the #1 Steam Generator (SG). -All other plant systems and components that have power are functioning as designed.

The crew is evaluating numerous alarms and indications caused by the power loss and subsequent ESD. Which of the following alarm indications will require actions to be taken in EOP 2525? A C-05 alarms indicating a Excess Steam Demand on the #1 SG, and C-OB alarm indicating VR-21 is energized.

B C-02/3 alarms indicating RCS temperatures are abnormally low and dropping, and both Boric Acid Pumps are de-energized.

C C-05 alarms indicating both SG levels abnormally low, and only one Aux. Feedwater pump is feeding just the #2 SG. D C-01 alarms indicating CTMT Spray has actuated, and C-01 indicating only two CAR fans and one CS pump are Page 25 of25 Printed on 10/28/2009 at 13:18